Text

Proposed Tech Specs Permiting Use of Variable PORV Relief Valve Setpoint for Low Temp Overpressure Protection & Increasing Operating Margin During Plant Heatup & Cooldown
	ML18054B009
Person / Time
Site:	Palisades
Issue date:	09/22/1989
From:	; CONSUMERS ENERGY CO. (FORMERLY CONSUMERS POWER CO.)
To:	;
Shared Package
ML18054B007	List: L-89-098, Reactor Pressure Vessel Temp Limits Determination; ML18054B006; ML18054B009; ML18054B010; ML18054B012;
References
	NUDOCS 8909280092
	Download: ML18054B009 (51)
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ATTACHMENT I Consumers Power Company Palisades Plant Docket 50-255 PROPOSED TECHNICAL SPECIFICATION PAGE CHANGES September 22, 1989 19 Pages TSP0889-0101-MD01-NL04

3.1 PRIMARY COOLANT SYSTEM (Cont'd) 3.1.1 Operable Components (Cont'd)

h.

Initiation of forced circulation shall not occur at PCS cold II leg temperatures < 430°F unless the Shutdown Cooling System II is isolated from the PCS and the steam generator temperature II does not exceed the PCS cold leg temperature by more than the II

~T limit below.

II Cold Leg Temperature

1.

> 120°F and < 170°F

2. > 170°F and < 210°F

3. > 210°F

~T Limit 100°F 20°F 100°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 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 .

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 li 1

t.

( l)

Th coo ng or one primary coo ant pump is in opera ion.

e 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 550 terminate the boron dilution under asymmetric flow conditions.

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.

Administrative procedures will provide for use of pressurizer sprays I

II II II II II II II to maintain a nominal spread between the boron concentration in (2) the pressurizer and the primary system during the addition of boron.

3-ld Amendment No

~7, $$, 117, 11$,

TSP0889-0101-MD01-NL04

3.1 PRIMARY COOLANT SYSTEM (Contd)

Basis (Contd)

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 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 cooling system to assure a redundant heat removal system for the reactor.

Calculations have been performed to demonstrate that a pressure differential of 1380 psiC3) can be withstood by a tube uniformily 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 pressure 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 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 of+ 40°F.

The transient analyses were performed assuming a vessel flow at hot zero power (532°F) of 124.3 x 106 lb/hr minus 6% to account for flow measurement uncertainty and core flow bypass.

A DNB analysis was performed in a parametric fashion to determine the core inlet temperature as a function of pressure and flow for which the minimum DNBR is equal to 1.17.

This analysis includes the following uncertainties and allowances: 2% of rated power for power 3-2 Amendment No l0, $t, ttS, TSP0889-0101-MD01-NL04 I

3.1 PRIMARY COOLANT SYSTEM (Cont'd)

Basis (Cont'd) measurement; +/-0.06 for ASI measurement; +/-50 psi for pressurizer pressure; +/-7°F for inlet temperature; and 3% measurement and 3%

bypass for core flow.

In addition, transient biases were included in the derivatio? ~f the following equation for limiting reactor inlet temperature:

4 Tinlet ~ 543.3 +.057S(P-2060) + O.OOOOS(P-2060)~72 + l.173(W-120) -

.Ol02(W-120)**2 The limits of validity of this equation are:

1800 < Pressure < 2200 Psia 100.0-x 106 < Ve;sel Flow < 130 x 106 Lb/h ASI as sho~ in Figure 3.0 With measured primary coolant system flow rates > 130 M lbm/hr, limiting the maximum allowed inlet temperature to the Trnlet LCO at 130 M lbm/hr increases the margin to DNB for higher PCS flow rates.

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-T power.

The measured ASI calculated from the excore detector signals and adjusted for shape annealing (Yr) and the core power constitute an ordered pair (Q,Yr).

An 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 forced circulation is initiated in the PCS ensures that an energy addition caused by heat transferred from the secondary system to the PCS will not occur.

This requirement applies only to the initiation of forced circulation (the start of the first primary coolant pump) when the PCS cold 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 ~SAR, Section 14.3.2.

(2)

Updated FSAR, Section 4.3.7.

(3)

Palisades 1983/1984 Steam Generator Evaluation and Repair Program Report, Section 4, April 19, 1984 (4)

ANF-87-lSO(NP), Volume 2, Section 15.0.7..1 (5)

ANF-88-108 I

(6)

Consumers Power Company Engineering Analysis EA-A-NL-89-14-1

//

3-3 Amendment No

~t, it, tt1, ttS, TSP0889-0101-MD01-NL04

3.1 PRIMARY COOLANT SYSTEM (Continued) 3.1.2 Heatup and Cooldown Rates The primary coolant pressure and the system heatup and 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 I

limit line as shown on Figures 3-1 and 3-2.

The average heatup I

or cooldown rate in any one hour time period shall not exceed I

the heatup or cooldown rate limit when one or more PCS cold leg I

is less than the corresponding "Cold Leg Temperature" below.

Cold Leg Temperature Heatup/Cooldown Rate Limit 1

< 170°F 20°F/Hr

2.

> 170°F and < 250°F 40°F/Hr

3.

> 250°F and < 350°F 60°F/Hr

4.

> 350° F 100° F/Hr I

II II I

I Whenever the shutdown cooling isolation valves (MOV3015 and I

MOV3016) ar~ open, the primary coolant system shall not be heated I

at a rate of more than 40°F/Hr. when the "Cold Leg Temperature"

/

is >170°F.

//

b.

Allowable combinations of pressure and temperature for inservice I

testing during heatup are as shown in Figure 3-3.

The maximum heatup and cooldown rates required by Section a. above, are I

applicable.

Interpolation between limit lines for other than the noted temperature change rates is permitted in 3.l.2a.

I

c.

The average heatup or cooldown rates for the pressurizer shall I

not exceed 200°F/hr in any one hour time period.

Whenever the I

Shutdown Cooling isolation valves (MOV3015 and MOV3016) are OPEN, I

the pressurizer shall not be heated at a rate of more than I

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. 11, ~l, ii, ~1, ll1, TSP0889-0101-MD01-NL04

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 shall be up4ated in accordance with the following criteria and procedure:

1.

US Nuclear Regulatory Commission Regulatory Guide 1.99 Revision 2 has been used to predict the increase in trans1t1on 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 fast neutron exposure (E ~ 1 MeV).

Such a conversion shall be made consistent with the dosimetry evaluation of capsule w-290(12).

3.

The limit lines in Figures 3-1, 3~2 and 3-3 are based on the requirements of Reference 9, Paragraphs IV.A.2 and IV.A. 3.

These lines reflect a preservice hydrostatic test pressure of 2400 psig and a vessel flange material reference temperature of 60°F(8).

Basis All components in the primary coolant system are designed to withstand the effects of cyclic loads due to primary system temperature and pressure changes.Cl)

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 maximum plant heatup and cooldown limit of 100°F per hour is consistent with the design number of cycles and satisfies stress limits for cyclic operation.(2)

The reactor vessel plate 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 at the fluence to which the Figures 3-1, 3-2 and 3-3 apply.(lO)

The unirradiated RTNDT has been determined to be -S6°F.Cll)

An RTNDT of -S6°F is used as an unirradiated value to which irradiation effects are added.

In addition, 3-5 Amendment No. i1,,l, ii, Sf, f1, ll1, 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

( )

and specific component function and has a maximum NDTT of +40°F. 5 As a result of fast neutron irradiation in this region of the core, there will be an increase in the RT with operation.

The integrated 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.

f Since the neutron spectra and the flux measured at the samples and reactor vessel inside radius should be nearly identical, 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 ln 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 the US NRC Regulatory Guide.(10)

To compensate for any increase in the RT caused by irradiation, limits on the pressure-temperature relationship are periodically changed to stay within the stress limits during heatup and cooldown.

Reference 7 provides a procedure for obtaining the allowable loadings for ferritic pressure-retaining materials in Class 1 components.

This procedure is based on the principles of linear elastic fracture mechanics and involves a stress intensity factor prediction which is a lower bound of static, dynamic and cracf arrest critical values.

The stress intensity factor computed J) is a function of RTNoT, 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 7, reactor vessel wall locations at 1/4 and 3/4 thickness are limiting. It is at these locations that the crack propagation associated with the hypothetical flaw must be arrested.

At these locations, fluence attenuation and thermal gradients have been TSP0889-0101-MD01-NL04 3-6 Amendment No. i1,,t, ~~' S~, ~1,

tt1, I

I I

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 ~ressure-temperature limits are applicable to reactor vessel inner wall fluences of up to 1.8 x l019nvt.

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 criticality condition which defines a temperature below which "the core cannot be made critical (strictly based upon fracture mechanics' considerations) is 371°F.

The most limiting wall location is at 1/4 thickness.

The minimum criticality temperature, 371°F is the minimum permissible temperature for the inservice system hydrostatic 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 I

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 critical.

3-7 TSP0889-0101-MD01-NL04 Amendment No. 11, ~t, s~, SS, s~,

~1, tt1,

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 I

Valves (PORVs) provide overpressure protection; at 430°F or above, I

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 heatup curves only with respect to margin for primary membrane stress.l 7 J 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)

(5)

(6)

(7)

. (8)

(9)

{10)

{11)

{12)

{13)

{14)

{lS)

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, 1979, submitted to the NRC by Consumers Power Company letter dated July 2, 1979.

FSAR, Section 4.2.4.

(Deleted)

I ASHE Boiler and Pressure Vessel Code,Section III, Appendix G, "Protection Against Non-Ductile Failure," 1974 Edition

US Atomic Energy Counnission Standard Review Plan, Directorate of Licensing, Section S.3.2, "Pressure-Temperature Limits."

10 CFR Part SO, Appendix G.,

'~Fracture Toughness Requirements,"

May 31, 1983 as amended November 6, 1986.

I US Nuclear Regulatory Commission, Regulatory Guide 1.99, I

Revision 2, May, 1988.

I Combustion Engineering Report CEN-189, December, 1981.

"Analysis of Capsules T-330 and W-290 from the Consumers Power Company Palisades Reactor Vessel Radiation Surveillance Program," WCAP-10637, September, 1984.

"An*alysis of Fast Neutron Exposure of the Palisades Reactor I

Pressure Vessel" by Westinghouse Electric Corporation, March 1989.

I Consumers Power Company Engineering Analysis EA-FC-809-13, Rev 1

II "Pressure Response Effect of VLTOP with Replacement" PORVs. 11 I

Consumers Power Company Engineering Analysis EA-A-PAL-89-98 I

"Palisades Pressure and Temperature Limits."

I 3-8 Amendment No. 11, ~l, jj, 8~,

~1, lt1, I

I TSP0889-0l01-MD01-NL04

3.1.8 OVERPRESSURE PROTECTION SYSTEMS LIMITING CONDITIONS FOR OPERATION 3.1.8.1 REQUIREMENTS Two power operated relief valves (PORVs) with a lift setting below and/or to the right of the curve in 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

inch yent 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.

I

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 to exceed the pressure limits required by 10CFRSO 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 260°F and 430°F, the PORV setpoints ensure that 10CFR50

//

Appendix G pressure limits will not be exceeded.

Below 260°F,

//

overpressure protection is still provided by the PORVs but HPSI I

operability is precluded by the limitations of Technical I

Specification 3.3.2 g.

Above 430°F, the pressurizer safety I

valves prevent 10CFR50 Appendix G limits from being exceeded.

I 3-25a Amendment No. jl, 11, 111, TSP0889-0101-MD01-NL04

3.1.8 OVERPRESSURE PROTECTION SYSTEMS LIMITING CONDITIONS FOR OPERATION 3.1.8.

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

programmed 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 is provided by two independent channels of measurement, I

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 I

for potential 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 10CFRSO Appendix G pressure limits will not be exceeded when one of the PORVs is assumed to fail per the "single failure" criteria 10CFRSO 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.

/-

References

1.

2.

3.

4.

Technical Specification 3.3.2 Technical Specification 3.1.2.

Consumers Power Company Engineering Analysis EA-FC-809-13, Rev 1 "Palisades Plant Overpressurization Analysis" June 1987 and "Palisades Plant Primary Coolant System Overpressurization Subsystem Description" October 1977

3-25b Amendment No. 111, I

I II I

I I

I TSP0889-0l01-MD01-NL04

-- L TOP LIMIT CURVE 2,800..........., _. _

F..i.g_~.re.. -~-~-4....... _

0 I

0 0

I o

I 2,400 2,000 1,600 ii I

I 0

0 I

I I

O I

~ 1,200 I

I I

0 I

I o

........... \\........... *-........... '*.......... -**.......... -*-.......... *'..........,,.......

t 800 I

I I

0 I

0 t *** - **** - *** * ***********, *** - *** - *** -.* ********** *,*...

- ****** '\\ **** - **

400 z

0.

50 100 150 200 250 300.

350 400 PCS Degrees F

3.3 EMERGENCY CORE COOLING SYSTEM (Continued) 3.3.3

g.

HPSI pump operability shall be as follows:

1) If the reactor head is installed, both HPSI pumps shall be rendered inoperable when:

I I

a.

The PCS temperature is < 260°F, or II

b.

Shutdown cooling isolation valves M0-3015 and M0-3016 I

are open.

I

2)

Two HPSI pumps shall be operable when the PCS temperature I

is > 325°F.

I

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 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 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 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. H, 101, 111, TSP0889-0101-MD01-~L04

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 is6lating 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 valves to close.

Insuring both HPSI pumps are inoperable when the PCS temperature I

is < 260°F or the shutdown cooling isolation valves are open

//

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 10CFRSO Appendix G limits to be exceeded I

when the PCS temperature is < 260°F.

When the PCS temperature

//

is ~ 430°F, the pressurizer safety valves ensure that the PCS I

pressure will not exceed 10CFRSO Appendix G.

I The requirement to have both HPSI trains operable above 325°F I

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.

Therefore, when the PCS I

temperature is ~260°F and ~325°F operation of the HPSI system

//

would not cause the 10CFRSO Appendix G limits to be exceeded I

nor is HPSI system operation necessary for core cooling.

I HPSI pump testing with the HPSI 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 PCS.

References (1) FSAR, Section 9.10.3; (2) FSAR, Section 6.1, TSP0889-0101-MD01-NL04 3-33 Amendment No. tt, U, t0t, *t t1, 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 overpressure protection except when the vent pathway is provided with a valve which is locked, sealed, or otherwise secured in the open position, then verify these valves open at least once per 31 days.

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 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 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.

Tiie power range safety channels and ~T 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 can be expected to remain within acceptable tolerances if recalibration is performed at each refueling shutdown interval.

Substantial calibration shifts within a channel (essentially a channel failure) will be revealed during routine checking and testing procedures.

Thus, minimum calibration frequencies of one-per-day for the power range safety channels, and once each refueling shutdown for the process system channels, are considered adequate.

The minimum testing frequency for those instrument channels connected to the reactor protective system is based on an estimate¢ average unsafe failure rate of 1.14 x 10-5 failure/hour per channel.

This estimation is based on limited operating experience at conventional and nuclear plants.

An*"unsafe failure" is defined as one which negates channel operability and wh1ch, due to its nature, is revealed only when the channel is tested or attempts to respond to a bonafide signal.

4-2 Amendment No. tS, St, tt1, ttS, TSP0889-0101-MD01-NL04 I

I

4.6 SAFETY INJECTION AND CONTAINMENT SPRAY SYSTEMS TESTS 4.6.1 4.6.2 Applicability Applies to the safety injection system, the containment spray system, chemical injection system and the containment cooling system tests.

Objective To verify that the subject systems 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 be applied b

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 in the proper sequence and timing (ie, the appropriate pump breakers shall have opened and closed, and all valves shall have completed their travel).

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 < 260°F or if shutdown cooling valves M0-3015 and M0-3016 are open unless the reactor head is removed.

Containment Spray System

a.

System 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 tripping 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 visual observations indicate all components have operated satisfactorily.

4-39 Amendment No.

~t, 11, j6, 111, TSP0889-0101-MD01-NL04 II I

4.6 SAFETY INJECTION AND CONTAINMENT SPRAY SYSTEMS TESTS (Continued)

Basis (continued)

During reactor operation, the instrumentation 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 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, 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 performance.

With the reactor vessel head installed when the PCS cold leg temperature is less than 260°F, or if the shutdown cooling system isolation valves 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

4-41 Amendment No. tt1, TSP0889-0101-MD01-NL04 I

II I

I I

I

ATTACHMENT II Consumers Power Company Palisades Plant Docket 50-255 EXISTING TECHNICAL SPECIFICATION PAGES MARKED UP WITH PROPOSED CHANGES September 22, 1989 31 Pages TSP0889-0101-MD01-NL04

3.1 PRIMARY COOLANT SYSTEM 3.1.l Applicability 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 primary 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 being made in th* 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 exception:

Before removing a pump from service, thermal power shall be reduced as specified in Table 2.3.l and appropriate corrective action implemente~. With one pump out of service, return the pump to service "¥-lthin 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months (1 eturn to four-p~**.1p operation) or be in hot sh**i ii.own (or below) *. tth the reactui,*.ripped (from the*C-06 panel, opening th& 42-01 and 42-02 circuit breakers) within". the nm-~ 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . Start-up (abo-,rs hot* shutdown) with les* than four pumps is not permitted aud p~~er operation with less than three pumps is not permittedo

c.

The measured* four pr~ry coolant pumps opti~ating reactor vessel flow shall be 124.3 x 106 lb/hr or greater, when corrected to 532°1.

d. Both steam generators shall be capabla of performing their heat transfer function whenever th* average temperature of the primary coolant is above 325~1.

e. Maximum primary system pressure differentials shall not exceed the following:

(1)

Deleted 3-lb Amendment No Jl, 8J, tt8, 119 December 12, 1988 TSP1088-0181-NL04 I

3. l

3. 1.1 PRIMARY COOLANT SYSTEM (Continued)

Operable Components (Continued)

(2)

Hydrostatic tests shall be conducted in accordance with

~

applicable paragraphs of Section 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 50 psi where Po is nominal operating pressure.

(3)

Primary side leak tests shall be conducted at normal operating pressure. The temperature shall be consistent with applicable fracture toughness criteria for ferritic materials and shall be selected such that the differential pressure across the steam generator tubes is not greater than 1380 psi.

(4)

Maximum secondary hydrostatic test pressur* shall not exceed 1250 psia. A minimum temperature of loo*r is required. Only ten cycles are permitted.

(5)

Maximum secondary leak test pressure shall not exceed 1000 psia. A minimum temperature of 100°P is required.

(6)

In performing the tests identified in 3.l.l.e(4) and 3.l.l.e(5), above, the secondary pressur* shall not exceed the prilliary pressure by more than 350 psi.

f. Nominal primary system operation pressur* shall not exceed 2100 psia.

g. The:.r".:eiictor *inlet td".. ;>erature (indicated) 1:1i1all not e,..eeed the'.valu* given by the following equation at steady state power operation:

I Ti 1 t s 543.3 +.0575(P-2060) + 0.00005(P-2060)**2 + l.173(W-120) -

I n 8

.0102(W-120)**2 -

I Wh*re: Tinl*t *

reactor inl*t temp*rature 1D. *po P

nominal operating pressur* in psia W

total recirculating mass flaw in 106 lb/h corrected to th* operating temperatur*

conditions.

When th* ASI exceeds the limits spacified in Figure 3.0, within 15 minutes, initiat* correctiv* actions to restor* th* ASI to th* acceptabl* region. Restore th* ASI to acceptable values within on* hour or ba at less than 70% of rated pow*r within the following two hours.

I I

I If the measured primary coolant system flow rate is greater than 130 M lbm/hr, the maximum inlet temperatur* shall b*

less.than or equal to the Tinlet LCO at 130 M lbm/hr.

3-lc Amendment No JZ, JZ, 8J, 1%1, 118 Novembar 15, 1988

't)

TSP1088-0181-NL04.

3. l

3. l. l Operable Component3 (Cont'd)

h.

During init l pri y coolant ?Ump starts (i.e., initiation of forced cir ion), secondary system temperature in the

. f"'f"'

steam genera hall be < the PCS cold leg temp*rature

_J,,-/tfS~'

unless the S col eg temperature is~ 450°F.

i.

The PCS shall not be heated or maintained above J25°F unless a minimum of 375 kW of pressurizer heater capacity is available from both buses lD and lE.

Should heater capacity from either bus lD and lE fall below 375 kW, eith*r restore the inoperable heaters to provid* 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 th* next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

Basis Wh&n primary coolant boron concentration is b*ing changed, the process must be uniform throughout th* primary coolant system

. volume to prevent stratification of primary coolant at lower boron concentration which could r*sult in a r*activity insertion.

Sufficient mixing of the primary coolant is assur*d if one shutdown cooling or o~* primary coolant pump is in operation.Cl)

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

By imposing a minimum shutdown cooling pump flov rate of 2810 gp111, suffici*nt time is provided for th* operatof 6jo terminate th* boron dilution under asymmetric flow conditions.

The presauriz*r volume is r~lativ*ly inactive, therefore wil.1. t*nd to have a boron conc*ntratio*~* high*r th,~:. r...Jt of th* PF~* ) coolant sys~*to..S.'1ring a dilutioi. **j)eration.

Ad*inistrative procadures will provide for u** of prassuriz*r sprays to maintain a nominal spread b*tw**n th* boron concentration in the pres1urizer and th* primary 1y1t.. ~uring th* additiO!l of boron. <2>

I I

Th* FSAll scf aty analyai* waa perf ormad aasuming four primary coolant I

pump* were operat1n1 for accidents that occur durin1 reactor I

operation. Therefore, reactor startup above hot shutdown i* not I

permitted unlHa ~11 four primary coolanc pump* are operating.

I Oparation with three p~illary coolant pump* i* permitted for I

a limited time to allow th* restart of a stopped P\\Dlllt or for I

reactor internal* vibration monitoring and teating.

I Requiring th* plant to ba in hot shutdown with th* reactor tripped I

from th* C-06 panal, opening the 42-01 and 42-02 circuit breakers, I

assur*s an inadvertent rod bank withdrawal will not b* initiated I

by th* contra~ room operator. Both steam generator* ara required I

to be operable whenever th* tnll'eratura of the primary coolanc is I

gr*atar than th* d**ian temperature of th* shutdown cool1n1 *Y*t..

I to a*aure a r*dundant heat removal sy1t.. for the reactor.

I 3-ld Amendment No 17, lZ1, 118 Noveaber 1,, 1988 TS11088-0181-NL04 1

3.1 PRIMARY COOLANT SYSTEM (Cont'd) 3.1.1 Operable Components (Cont'd)

h.

Initiation of forced circulation shall not occur at PCS cold II leg temperatures< 430°F unless the Shutdown Cooling System II is isolated from the PCS and the steam generator temperature II does not exceed the PCS cold leg temperature by more than the II

~T limit below.

II Cold Leg Temperature

1.

> 120°F and < 170°F

2.

> 170°F and < 210°F

3.

> 210°F

~T Limit 100°F 20°F 100°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 from both buses lD and IE.

Should heater capacity from either bus ID and lE fall below 375 kW, either restore the inoperable heaters to provide at least 375 kW of heater capacity from both buses ID 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 .

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 in operation. (l)

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 operato{ 5

~o terminate the boron dilution under

asymmetric flow conditions.

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.

Administrative procedures will provide for use of pressurizer sprays I

II II II II II II II to maintain a nominal spread between the boron concentration in

( 2) the pressurizer and the primary system during the addition of boron

3-ld Amendment No ~7. SJ, tt7, tt~.

TSP0889-0I01-MDOI-NL04

3.1 PRIMARY COOLANT SYSTEM (Contd)

Basis (Contd)

Administrative procedures will provide for use of pressurizer 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 witq 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 cooling system to assure a redundant heat removal system for the reactor.

Calculations have been performed to demonstrate that a pressure differential of 1380 psiC3) can be withstood by a tube uniformily 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 pressure 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 ASME BPV Section XI (1971).

The differential maintains stresses in the steam generator tube walls within code allowable stresses.

The minimum temperature of lOOaF for pressur1z1ng the steam generator secondary *side is set by the NDTT of theo11a;ws; cover of+ 40°F.

The *transient analyses were performed assuming a vessel flow at hot zero power (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 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 allo~ances: 2% of rated power for power 3-2.

Amendment No U, U, ug, TSP0889-0101-NL04

I *

J. l PR!~.ARY COOLA..~T SYSTE~ (Conc'd)

~

(Cont'd)

The Axial Shape Index alarm channel is being used to monitor the ASI co ensure chat th* 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 fro~ the excore detector signals and adjusted for shape annealing (YI) and the core power constitute an ordered pair (Q,YI).

An alarm signal is activated before the ordered pair exc.iQd th... boundariu specified in F:Lgui:'& 3.0.

The requirement that the steam generator temperature be < the PCS temperature when forced circulation is initiated in the PCS ensures that au energy addition caused by heat tran*f erred from the secondary systea to the PCS will not occur.

This requirement applies only to the initiation of forced circulation (the start of th* first primary coolant pump) when th* PCS cold 1*1 tempera cure is <

F.

J 1 u 1llt'" '15o*r che prs uiuy Yal*****

References (1)

(2)

(3)

(4)

(~

Updated Updated FSAll, S*ct Pal1~ades 1983/198 Proisra!. Report, S*

ANF-87-150(NP), Vo

(!clacud, ANF-88-108 (b}CPto tff,d1*'--;,"

ft 14.3.2.

on 4.3.7.

Steam Generator Evaluation tior. ~. April 19, 1984

2, Section 15.0.7.1 and Rep... Jr

"""/1t31r 4A-19-NJ.-i9-.1J1-I I

I I

11~,, 14-At*":r I~-.,')

s Aow.r o/I,

T 1!1"'1 rk h~,:r, d em ti, 7;,;,.r wh. °""- ~J"1c1o~ Cn/,i6:S~.sT...,,,,,

.,:s-,sol*H cl I l-1!Wt -H.c. Rs;j,tne.,d c,,;1!1.1/t:;/i~ l'l'f'I*~,, ;,.,; 7i*1id 1.vJ~.

-lit s7

~.,,,, c ~.,.-ro., -r...,, ~*Tv"'<...

1:S- )1ek, ~llw} ~

P~J C~lcJ

/,&..,...,,,,PJl*7~~

3-3 Allendmaac No JI, JZ, 117, 118 Nov-.ber 15, 1981

3.1 PR!~RY COOLANT SYSTEM (

3. l. 2 Heacup and Cooldown RaCe9 Th* primary coolant pre99ur* and ch* syscem heacup and cooldown races shall b* limited in accordance with Figure 3-t, Figure 3-Z and as follows.

.,J S*Z.

0,_ L**'~*..,,.

I

a.

Allowabl* combinations o pressure and tempE*a e for any heatup ~

C..tJe>/c/0,,.,,,,,

race shall be below and to th* right of the plicable limit line I

as shown on FigureSJ-t

!h* average heatup rate in any one hour

/

time period shall not exceed the heacup rate limit when one or I

more PCS cold lag is leu than the corresponding "Cold Leg

/

Temperature" below.

I H;:cup Rae= Limit 20.F/'Rr 4o*r/Hr 60.F/Hr loo*r/Hr Allowabl\\ combinations of pressure and temp*rature f o oldovn rate shall b* below and to the right of applicable lines a* shown on Figure 3*2. The av

cooldoWft rat*

t *zceed the Cooldovn Rate Lim en' one or more PCS cold than the corrHpond old Le1 Temperature" below:

Cooldova Rate Limit loo*r1ar 6o*r/Hr 4o*r/Hr 2o*r/Hr

~;t. Allowable cmb1Daeioaa of prHaura and tmperatura for !Uervic*

tHU.n1 durtn1 heacu, ara aa 1havn in Figura 3-3. The mazimull heaeup and c:ooldOVD rat** raquirad by Sacticnia *--~d b, above, are applicable. Interpolation between limit line* for other than th* noted t111P*ratura chan1* rat** i* permitted in 3.l.2a, b or c.

/.,~co.IJ~

t.~* l. The &Taraae heatup'tate8 for the preamurizar shall DOC ezcead

,J.oo.Wr/hr 111 any on* hour tiu pariod_,.._._....-i~.......,..,.._

t111P**11111** ,,., ch** *1a*r, 3*4 1hutdovn coolinl 1y1t.. i*

Amendment No. Z7, II. JJ. 97. 117 November 14. 1981 111J-.~ -tAt s,~_,.J,,,,.. ~IHJ""°/.,,,;..1.,;,;..,

I I

I TSP1l87-0218*NL04

..,.,,_,1H> (lflO !DI~ 1M" N~.I#') 44L ~".1 Tk.

pu....,1-/z.,. rA.1/,,. *.,,1,~ J,. "d.r~.-..;..,/

,,.,,.~ oHllWI,o_. F ///>..

3.1.2 Heatup and Cooldown Rates (Continued) d1.

Before the radiation exposure of the reactor vessel exceeds the exposure for which the figures apply. Figures 3-1. 3-2 and 3-3 shall be updated in accordance with th* folloW'ing criteria and procedure:

~b-1"ML US Nuclear Regulatory Commission Regulatory Guide l.99~s been used to predict th* 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.

1.

2.

Before the end of th* int*srated power period for which Figures 3-1, 3-2 and 3-3 apply, th* limit lines on the figures shall be updated for a new integrated pov*r period.

The total integrated reactor thermal paver from start-up to the end of the nev pov*r period shall be couverted to an.

equivalent integrated fast neutron exposure (! ~ l M1V).

Such a conversion shall be made consistent with th*

3.

Basis dosimetrjo evaluation of cap*ul1 w-290CU>.

Th* limit lines in Figures 3*1, 3-2 and 3-3 ar* based on the requirements of Reference 9, Paragraphs.IV.A.2 and IV *. A.3.

Th111 lines r&f l1ct a preservic* hydrostatic test pressure of 2400 psig and a vessel flange material r1f 1r*nc1 tl!lll'*ratur1 of 60 1 F(S).

Ali compon1nta in tn* primary coolAht sy.cm are d'sip1d to withstand th* ef f 1cta of cyclic load* due to primary sy*t.. tl!lll'1ratur1 and pressure chana***(l) These cyclic load* are introduced by normal unit load transient*, reactor trips and *tart-uit and shutdovn operation.

During unit start-up and shutdown, th* rat** cf t1mp1ratur* and prusur* chan1** ar* limited. A m.aximwl plant heatup and cooldovn limit of ioo*r per hour is consistent with th* design number of cycl** and satisfie*.*tr*** limits for cyclic operation. <2>

Th* reactor v1ss*l plate and material opposite th* care has been purch.. *d to a specified Charpy v~Natch t1ac re*ult of 30 ft-lb or greater at an ND'l'T of + 1o*r or les1. Th* vessel weld has th*

highest RTNDT of plat*, w*ld and HAZ materials at the flu*nc* to (10) which th* Figures 3-1, 3-2 and 3-3 apply.

Th* unirradiated RTNDT has been determined to b* -56°P.Cll)

AD RTNDT*of -56 1 P ia used as an I

unirradiatad valu* to which irradiation effects are added.

Ia. addition, TSP1187-0218-NL04 3-5 Amendment No. %1, It, JJ, 89, 97, 117 November 14, 1988 I J

J.1.2

-r.4c 1117iM11iJJ,,r,,,,,,~,,.. /1>1 /tl~v)r}/v,... -/-f.lc.,.,.,1,.,,

"~.-/ °"'~/levlfP'f',J "3,;,6 ~._.,.,. 1.3,,-r,J,i,Jt; Dcr.:ra:.

an ~~~~~re1 t/:/:_if/!/f-DC zt1-r.(

c1-~:3.;s..,,c7~,,,.r.

the plate has been volumetrically inspected by ultrasonic te.st using both longitudina nd shear wave methods. The remainin1 material in the r*actor ve el, and other primary coolant system components, meets the appropr

design code requirements and specific component function and

& maximum NDTr of +40.F.(5)

As a result of fast neutron irradiation in t region of the core, there will be an increase in the RT with operation. ihs teela:iqscs up# SS pr 'ti3S SIU ;*te1r 0 d h'! UJIUICS (i I 1 !Iii) fltaii Of

h* g"nrcr uesul are ~au*i're~ ip huion j 1: 3:' Iii *ks PB/1? -~

pl n ia,..uasu U; &adi:IR II 1 u tie nu Sines the neut~on sp~et~a and thQ flu: m=~;urad at th= :Ampl== ~nd reactor vessel inside radius should be nearly identical, the measured transition shift frOll a sample can be applied to the adjacent section of the reactor vessel for later sta1e1 in plant life equivalent to the difference in calculated flwc ma1nitude.

The maai~ exposure of the reactor vessel will be obtained from the measured sample exposure by application of the calculated azimuthal neutron flua variation. The predicted iTHDT shift for the base metal has been predicted ba1ed upon 1urveillanca data and the US HRC Regulatory Cuide.ClO)

To compen1ata for any increa1e.

in the IT caused by irradiation, limits on the pra11ura-temperature relationship are periodically chan1sd to stay within the 1tre11 Limits during heatup and cooldo,Jn.

Reference 7 provide1.a procedure for obtaining the allowable loadin11 for ferritic pre11ure-retainin1 11aterial1 in Cla11 1 components. Thi1 procedure is baaed on the principle* of linear elaat;,: fracture mech~"ics and involve" a 1tru1 intensity factor pradi.~;ip::.. which is c; 'O"':.. bound of I!:. .if.*. dyna11ic ant:~ ~rec:"

arrsat c-..i.tical valua1.

it* 1tra11 intenl;;.1;y factor computed( 7) is a function of iTtfDT, operatin1 temperature, and ve11el wall temperature 1radieat1.

Pressure-temperature liait calculational procedures for the reactor coolant pre11ure boundary are defined in ief erence 8 ba1ed upon ieferuce 7. The liaic linH of Piaurs* 3-1 through 3-l, **

con1ider a 54 p1i pTa11ura allowance to account for the fact that pres1ur* i1.. aaured in tb* pressurizer rather than at the ve11el beltliae. Ia addition, for calculation.al purpo1e1, 5*r.... 10 p1i CUAO...,..., taken a1 *a~urment error allowance~ for. temperature,...._

pr

1*,111t 1lJ*

By Reference 7, rea or vessel wall location* at 1/4 and 3/4 thickne11 are lim" ia1. It is at th***

location* that the crack propagation as1 1ated with the hypothetical flaw must be arrested. A the1a locations, fluenca attenuation and.thermal 1radient1 ha TSPll87=0218*NL04 Amendment No. 21, 4!, SS, It, t7, 117 November 14, 1988

3.1.2 Heatup and Cooldown Rates (Continued)

~

(Cont'd) evaluatad. During cooldown, the 1/4 thickness location is always more limiting in that the RTNDT is higher than that at the

~ ~. /

~~'b

'G 4 ~

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.

~ ""

~

a t! l

\\I l Figures 3-1 through 3-3 define stress limitations only from a fracture mechanic_,' point of view.

~ t"~

"~ ~

"'~.t Other considerations may be more restrictive with respect to

\\;:t 3. l 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 f f ~

capacity ter.ds to restrict both normal heatup and cooldown rates

~ ~ l to less than 60°F per hour.

~, ~

The revised prassura=tQmpQrature limits are applicabla to reactor ! ~ f vessel inner wall fluences of up to 1.8 x l0 19nvt. The application ~i~

of appropriate fluence attenuation _factors (Ref erenca 10) at the I ~ _,

1/4 and 3/4 thickness locations results in RTNDT shifts of 241°F

"""" ~~

/?1° ~P',

respectively, for the limiting weld material. The y ~

criticality condition which defines a temperature below which I 'I !

the core cannot be made critical (strictly based upon fracture i ~*!*

mechanics' considerations) is 371°F.

The most limiting wall I Ji-a location is at 1/4 thickness. The minimum criticality

~

~.J temperature, 37l°F is the minimum permissible temperature for

~.J~

the inservice **ystem hydrostat:!.c pressure test.. That temperature

e. t...1 r is calculated..ised upon 2310 ~sig inservice h~-*.rostatic tes1;

~ '? i pressure.

s.JO i l~

Th* restriction of average hea.. ~d cooldown rates to loo*r/h. i...._l /

when all PCS cold legs are !~~,-~nd the maintenance of a

~ f\\./

pressure-temperature relationship under the heatup, cooldown and inaervice test curve* of Figuras 3-1, 3-2 and 3-3, respectively, I ensur** that the requirements of References 6, 7, 8 and 9 are met.:.:.J The cote operational limit applies only when the reactor is critical.

The heatup and cooldown rate restrictions app11a*ble *+1* the tema*r*n1r* gt on* o- **** ai 1h1 PCS cold 111' 1 e l1aa #i. a 453 2 are consistent with the analyses performed for lov temperature

~!t;j~ssure protection (LTOP) (References 13, 14, 15, 16 & 17).

o1"'"11J/30 For above, the PCS safety valves provide overpres*ure protection for heatup or cooldown rates ~ 100°!'/hr.

Amendment No. 27,,t, If, JJ, If, f7. 117 November 14, 1988 8&/ow'l3o*~ ~

Po~o~-,:,JI&/,~

I I

TSP1187-0218-NL04 f*l11co (P~RV,-)p~,"J,, t:rd-lpuo.ou..., p~-,;,,r,;...;

J.l.2 Heatup and Cooldown Rates (Cor

~

(Continued)

The criticality temperature is deter1llined per Reference a and the core operational curves adhere to the requirements of Reference 9.

The inservice test curves incorporate allowances for the thel"ll!al gradients associated with the heatup curve used to attain inservice test pressure.

These curves differ from heatup curves 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)

(5)

(6)

(7)

(8)

(9)

(10)

(11)

(12)

(14)

(U)

FSAR, Section 4.2.2.

ASME Boiler and Pressure Vessel Coda,Section III, A-2000.

Battelle Columbus Laboratories Report, "Palisades Pressure Vessel Irradiation Capsule Program:

Unirradiated Mechanical PropertiH," August 25, 1977.

Bat:tel,l* Columbus Laboratories Report, "Palisades Nuclear Plant Reactor Vessel Surveillance Program:

Capsule A-240," March 13, 1979, submitted to the NRC by Consumers Power Col!lpany letter dated July 2, 1979.

FSA!l, Section 4.2.4.

)

Ui 7&n 1 "T h17*leror7 Conmhdsm, hplaeuc Gdide i.'9-, ( O.i.l\\6J

"!f'uu 1i lliliduil !11MidCY ca fhdltccd RadiaU1a ;**8* =..

i1a111s V1*11 1 Ma11*t1l1;" l*lJ; 1071:

ASM! Boiler and Pressure Vessel Code,Section III, Appendix G, "Protection Against Non-Ductile Failure," 1974 Edition.

US Atomic Energy Commission Standard Review Plan, Directorate of Licensing, Section 5.3.2, "Pressure-Tqperature Limits."

10 CFR Part 50, Appendix G, "Fracture Toughnes* Requirements,"

May 31, 19830.0 °"""._'6-J N..,...,Aw,,, 1916

US Nuclear Regulatory C01111ission, Regulatory Guida l.99, &wall Revision 2, Jfsd': 'PH. lll114t1lfll' Combu*tion !nginaarin1 Report CEH-189, Deceaber, 1981.

"Analyd* of Cap*ulH T-330 and W-290 fl'Oll the Con*umers Power COl!lpany Pali**d** Reactor Vessel Radiation Surveillance Progra," WCAP-10637, September,. 1984.

ii P\\t II 101 "C1hn 1 Uh* d PCI PHH11H iBHIAH h** 'U'*I UJ IJ& (3 11l1s1.. I,_,a) Babu *he HM'u 8pca;ll W1*wMH '1,

~-

et Nii ls'llP 118119

"&alcalattca cf laqa1nd P81:9 eapacUy ID Ha#n,,1P th* !Cl ld*r,,,, **,._ & :" cfaaaat) U, 1'88.

IYs PAis ~!8P 888119 Rssc h P81"' flbi CIPIE!Cj it l!s:puu11j.

L;QJ C1a*'9,1a1" l*'r*a_,. U a 19811 11.t PtJ.S 1i;o1 '90111

. "Clll'Pl1t1 n of r1m* hr Opuutct co Ace

- - 1WII *** lu'r'rl*"

7 *mt**~ ag, 10111.

l!lt !!!II HH1 8/9 81 "Peltaulca Plaut fd:maq euotlilt Sfitft RTUl 11.. *Tnp*r*tnr= T1*1u i'H i\\,,...... Q...... LiSHI lctlat and P-***11=* V**** 1 Cz'*" P1rt*'** o.

3-8 Amendment Mo. %7, It, JJ, IJ, f7' 117 November 14, 1988 TSP1187-0218-NL04 I

I I

I

~_, Ul 1-' Vl

- I*

w l

IO p

R E

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50 PALISADES PLANT PRESSURE AND *~"£MPERATURE LIMITS FOR HEATUP For nuence to 1.1x10* nvt Figure J-1 100 f /Hi' 80f/Hr 60f/Hr 40f/Hr 15 IGO 135 150 ll:i aoo 225 250 a~

JOO Ja5

>>O l~

400 42S 4)0

*ERATURE DEGREES F

FIGURE 3-1 PALISADES PRESSURE AND TEMPERATURE LIMITS FOR HEATUP PRESS PSIG F = 1.8 x 1019 n/cm2 (No measurement uncertainty included) 100 w

I

\\C 3000 r=-*-1----*---*-1**---------1 * **-** *--r****--*--** -t* *** *****- *1-- * --*--1.. ---* *-**1-* * *** ***-1-- ***--- *t** -* * ***

t * -- **

r* * **--* -t* - * * - - -- t i

r i

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ao:

1 2750 '

_J_... *--**.L--~- ***+*** -*** ****~*-... *-***;-**-*- ****-L - -* --* - -*....

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l

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1000 ~-

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,-,, -l,rTTI, TrT lrrn!rrJ-*TTTjrn,~,,-,-,- -~

0 50 75 100 125 150 175 200 225 250 275 300 325 350 375 400.

425 450 TEMPERATURE DEGREES F

~

0 <..

ID p

a* !

D',....

ll ci" I-'

.JO""

~

0.

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\\()"&!'I).

CD""'ol~

(D....

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-I*

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w w

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R E s 1 s u R 1 E

PALISADES PLANT PRESSURE AND TEMPERATURE LIMITS FOR COOLOOWN for Fluenc* to 1.a X 1 o** nvt

~,*:*3 *--~------..--~-------.---.....,---,---.---- **---

&-----1----+~

-=-1-~--~

50 100 125

~

--~

+- --t--1

~--GiE-=--t---f---t---t---t-*--t--t 1-t-i

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-+--- --- ---*-l---

I I

175

'Jo.>

225 250 275 300 Jl25 J~

J/)

4o.l

":J!>

"~0 TEMPERATURE*OEGREES f

FIGURE 3-2

PALISADES PRESSURE & TEMPERATURE LIMITS FOR COOLOOWN PRESS PSIG l 'J 2

F' = l.H X 10 n/cm (No Measurement Uncertainty Included) aooo 2750 2500 2250 I

I

~

I 'JI

~

I/IA IJ (j /).

~~ ~r 2000 1750 1500 w

I I-' 1250 0

~

1000 CD ::s

~

CD 750

s ct

~

0 500 "S.N.

~~

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{l 60 40 20

~

80

~

100

~

80

-~ ~

~

v

-- ~

j(

20 1--1.. 1--l

~..

0

'l::R.

0 I

I at 50 75 100 125 150 175 200 225 250 275.

300 325 350 375 400

.. 25

.. 50 l;;S.

TEMPERATURE DEGREES F

~ li i

i:s G'

Cl g

t-'

~

0.

\\0 "G. N.

CX> ""'l ""'l CX>..

..... "10;

..... )"-l

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Hi

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~

w I *

~

PALISADES PLANT PRESSURE AND TEMPERATURE LIMITS fOR HYDROSTATIC TESTING "IS

100 for fluence TO I.IX 1011 nvl figure J-J TEMPERATURE DEGREES f

FIGURE 3-3 PALISADES PRESSURE AND TEMPERATURE LIMITS FOR HYDRO PRESS PSI6 19 2

f = 1.8 X 10 n/cm 3000 2750 2500 2250 2000 1750 1500 w

I 1250 I-'

I-'

~

1000 en f:j

~

750 en f:j c+

!2:

500 0.

'So N.

250

~~

Ill /j;v 0

I/II I/

20

/; %//;'

a '!/

40 I !//#'

I~

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~ 7 v

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i

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.......- '/

~

~....

0

......... l'-1.

~..

I l::R.

l::R.

50 75 100 125 150 175 200 225 250 275 300 325 350 375 400

..C25 450

~

'So TEMPERATURE DEGREES F

~

'4

~

- ' 1 i

d

~

~ "

1 J. 1. 8 OVERPRESS~RE PROTECTION SYSTE~S LI~ITI~G CONDITIONS FOR OPERATION J.1.8.l REQUIREMENTS

.a rnaeg th* *1..,U"iY1' 2 cf CP 2

gr TO'U il siu PWlll:asy ua 1 ?Pt SJ' 8 t 2

m sc 1 d 1181 ts 1 398°F; cc mlteueoez the sltatdcsnt a&a 7 ing.

.i:solae!:H :aloes (H8V 3613 and H8V 391&) *are apH: 1 *ma pcn*r nparar94 **lief vilv@Y (PORO§) with a life Bl't1ai gf < J!Q psia sb.al 1 h1 np*nala; !!' a l!'IHtn* ceelaru s,.u111 : en*.;-

....2 C 1 i~'J?I'? iHhU !hall h 1p11t; n heh pnpu p1' at ualuu and both iORU Hee le "llluu 1hall a a 1p aa.

au rilaa *h* Oapasutan o:i cue H &&l'a si tha pt!maz; s;stem cotd

' taeep., Two power operated relief valves (PORVs) c *

a.

b.

c.

a lift sattin~.......-.._.,.......,.,....

..,.~"",....,..,...... ~

When the temperature of one or more of the coolant system cold legs is less than 43o*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, depressurize and within th* 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 POIV *'l** valves and bo~ii' POlV block valv***

The provieion* of Specification* 3.0.3 and 3.0.4 are not applicable.*

Bui a There are three pr**aure transients which could cau** the PCS pressure to exceed th* pres*ure limit* required by lOC~O Appendix G.

They are:

(1) a cbarging/letdova imbalance, (2) the start of a high presaura safety injection (BPSI) pump, and (3) initiatiou of f orcad circulation in the PCS when th* st*..

ganerat~r temperature ia higher than the PCS temperature.

3-25a Amendaent No. J!, 7Z. 117

Novaber 14. 1988-TSP1187-0218-NL04

/.

I I

I l

I I

I I j.

I I

l I

I I

I

3. l. 8 L!MIT!NG

.3.l.8.

r 13.,/ow A,c*,:, Ol'&tpu.o,,1.4.. p1.c'T;,r,~ _,.;-.3"'T,//

~/J,,d -'11 -r/4. PIJRY.r., J,~r HPJ/ ~*

J,,'J,~)L _.:r p>t1e/vd,,,t:/

St:RE l?ROTECT"rON SYSTE::

.1.

~ '-.

1

...,.. -r!

_ ~

~ _ J

'/,

.,~, 1&11., '""" 11 G I,,., J"' "'P

~.,,..,,,I tt11 I

COND!T!ON OR OPERATION

. 3 u.1 l~17'iW..,1,.J,z.

~

Basis (coneinu'li.dl 1J.,7w.-., I.. I.Do,: -..J "/JO*~

I I

Analysis (Referenc~; 4,,.. shows that when three charging pUlllps are operating and 1 down is isolated and a spurious HPSt occurs1 the PORV setpoints e ur* chat LOCFRSO Appendix G pressure limits will noc be exceeded.

Above 430°F, th* pressurizer safety valves prevent 10CFR Appendix G limits from being exceed*d.*1* 1 a**i*!n2/

latdm,m tmpalapc* (i1i1***** i);

tbs **~*itemaae Eftit sc1am g1a11acot cempatal*r* -* ' the ?Ci tegperacu** waea fa**** eir~~laeioa ts 1Hit1itid ta ia1 pcs O.J2'"TH CR** aa '""'" addtetua caused by n11e cransfc:,.. fn*

ch* uu***., s)*U* eu the res *111 nut uccat: !hta n*... '-**H*

app1111 1al1 eo eh* '*'*'**!i:ea cf !ottiel etzcalattua hh* ua1 pf £h* f1r1t p*Mle.,, 1111.. I JU.,) Vila CUC Ct liUiW Of Iha P99 s;gld I te=p*rlt 11r** < 45g*r I

I I

IMS.Jtf-r

1.3 square inch** (Reference 3), or by openin& boch I POIV pilot valve* and both l'OIV block valve* when oae or both I PORV* are inoperable ensure* that the 10~0 Appendiz G pres*ure I limits will not be exceeded when one of the l'OIV* 1* ***umed to I fail per th* -"single f:t,:ure" crit~ iocruo _Mpendiz A,, I Crict~ion 34. S1;,1.- ~ ;tlllll 1*. #d,:,,,.rrr~..,,,,,.,,..,. "'ro ~ ~-C. ,,,,_,114,, PMwvl4#11t1tl,,.J,,,1J,h.,.,.,c..r, -t-AP~ltll.r,.,,..., 1:M. "" 0~ S1a1,1a a,;,a,1(6) **.. '*** a 4**,ecead eperceaz chi& Cfti POI'* I ar* '******~'*.. alleva* 6a le11i1a iulul l **

a1l1*** Jll*t'i>tna I

tb'* 11a*l*ie* '9 *1fataat14 ta cha Basta *** i***'oe 1.3, I b'l l/ufll.,,l# +J.... C-#rtt,_.J 'S11/if~ I~ ~ 11r s PI tJIN'1Ti;.,... *. Relerencea ~ "nc.J.,,.,,,,1 S,,.,,:J.c.;r,:.r., 3.a. 'Z.

1. il6 P*li 89 181 "Ci1EU1attoa of P&I P1aaau:c !ac:1a1a rs..

-Add1DI 13\\.ISB (a l~HIMI,,... ~ lef~H tba '8RTa Ufaa, Yoseph** +a 19171

2. Technical Spec:ificatioD 3.1.2.

I I I I I I +!'- "1ali1adea Plant 0Verpr111urizatioa Analy1i1," June 1977 and Pali*ad** Plat Primary Coolant Sy1t** OVerprlHUriZ&tioD Sub*yatm DHcripctoa. ~* October, 1977. I I I

s.

,..u1

1 TOI 1181 H "eatewl:au= a' !*qutwd P9M' Cip&ttcy to v1 1a1a6a *~* PGI lctuw *PP*ad~ I &*.. **>>" JCUUZi) l's 1911.

&1 l\\lu L;G' 11e1aa "fld11ad1a ttUI 1010 110111e1 eapceta,. ma.a Pea **** ****11** ** aea*r oz a11at1t.",....,. an 1011 .S CPC t;.,,.,,,;_,~llC.,4... J,,.,,,-., IA _,e -~O"l-13 3-25b TS!1187-0218-.Ht.04 Amendment Mo *. 117 Movmber 14,. 1988 I I I /. 3.1. 8 OVERPRESSURE PROTECTION SYSTEMS l l 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 programmed and calibrated 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 limit 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 provided by two independent channels of measurement, control, actuation, and valves, 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 instrument inaccuracies, measurement error, and instrument drift. This will ensure that at no time between calibration intervals will the combination of PCS temperature and pressure exceed the limits of Figure 3-4 without PORV actuation. When the shutdown coo1ing system is not isolated (M0-3015 and* M0-3016 open) from the PCS, assurance that the shutdown cooling system 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.l 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 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 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 10CFRSO Appendix G pressure limits will not be exceeded when one of the PORVs is assumed to fail per the "single failure" criteria 10CFRSO 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-lJ,/f.v/

4.

"Palisades Plant Overpressurization Analysis" June 1987 and "Palisades Plant Primary Coolant System Overpressurization Subsystem Description" October 1977. 3-25b Amendment No. ll1, I ;, I I I I 2,800 2,400 2,000 ~ 1,600 ii z a. ~ 1,200 t 800 50 . L TOP LIMIT. CURVE . __........,.............. _f_i_g_~.r~... a.~.4........................................... 0 0 0 0 I 0 o 0 I o 0 0 I I I I 0 I 0 o o o I 0 o o \\.

- - - - I*******

0 I 0 0 I 0 0

*. * * * * * * *

r ***********,* ***********, ************, ******** -

100 150 200 250 300 350 400 PCS Degrees F i / 3.3 EMERGENCY CORE COOLING SYSTEM Applicability Appli** co th* operating stacua of the *m*rgency cor* cooling syscam. ObjtlCCiV* To ~ssura operability of equipment required to ramav* decay heat from cha core in either em*rgency or normal shutdown situations. Specif icat1on* Safety Injection and Shutdown Cooling Sy1cema 3.3.l !he reactor shall not be made critical, excapc for lov-cemparatura phyaica taata, Wlleaa all of th* followiDa. conditiona are mac:

a. The SI&W tank cont&ina DOC laaa than* 250,000 iallon8 of water with a boron coacancration of ac laaac 1720 ppa but Dot mora than 2000 ppm ac a tamparaeu~* noc laaa chaD 40 8 1e

b. All four Safety Injection canka are operable and pr***urized ta at laaat ZOO p*il with a tank liquid. level of at leaac 186 inch**

'~ (55.5%) ~ a maximma level of 198 inch** (59%) with a boron concentration of at leaac 1720 ppa but not more than 2000 ppa.

c. Ona lov-preaaura Safety IAjection pump i* operable on each bua *

d. Ona high-preaaura Safety Injection pump i* operable on each bua.

- Boch ~hutdown heat ex~haagera and both compoaenc cooling heac exc~ll ~ger* are oparat: ~;.e.

£. Piping and valve* ahall be operable to provide cwo flov path* from the SIRW tank co the prim&~ coaliDg ayacaa.

g. All valv*** pip~ and iACarlocka a*aoc:l.&ted. with th* above component* aad required ta f wactiou durills accident coaditiona are operable.

h. The Lov-Pru*ur* Safecy Injection !'lov Concrol Valve CV-3006 shall be opened aad diaabled (by isolaciDa cha air 1upply) to prevent apurioua cloaura.

1. tha Safecy Injection boccle mocor-operaced isol&cion valve* shall be opened with cha electric power 1upply co th* valve motor diacoDDected.

J. Iha Safecy Injection miniflow valve* CV-3027 and 3056 ahall be opened with HS-3027 aad 3056 poaiciona to maiAtaiD them open

3-29 TSP1Z8S-0354~NL04 Amandmenc No. 7*, lOl February 10, l987

J.3 E?-4.ERGENCY CORE COOLING SYSTEM (Contd) 3.3.2 Duriag power operacioa, th* requirements of 3.3.l may be modified to allov oae of the followial'coudicions to be true ac any one time. If the system is not restored co meet the requirements of 3.3.l withiD Che time period specified below, the reactor shall be placed in a hot shutdowu condition within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . If the*requirameucs of 3.3.1 are aot met 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 hour*.

a.

b.

c.

One safety injection tank may be inoperable for a period of no more than ona*hour. On* low-pr*s*ure safety inj1ctioa pump may be inoperable provided th9 pump i* re*tored to operable atatua withiJI 24 houra. Th* other lov-preHure safety iDj1ctioa pump shall be t**t*d t~ damolltltrato operability prior to iDitiatiDI r1pair of the inoperable pump.* Oae high-pre**ure safety injection pump may be inoperable provided the pump i* re*tor1d to operable atatu. within 24 hour*. The other high-pre*.ure safety iajectioa pump shal1 be tHted to daonatrat* operability prior to initiating repair of the inoperable pump.

d.

Oaa *hutdown heat ezchanger aad one compoaent cooliD1 water heat achanger may be inoperable for a period of no more than 24 hour*.

i." e.

Any valvec. i"*'*rlocka or pip:aJig directly aHO:,ciatad with one of th* above ca11ponaaca and required to fuuctioa during accident conditiona 1hall be dined to be part of that compoaeat and shall meac th*.... r1quir... at* aa liated for that componeat.

f.

Any valve, interlock or pipe a**ociated with tha *afecy injeccia11 aacl *hutdOVD cooliDI *Y*Cea and which i* nae covered under 3.3.2* abO"I* but, which ia required to funcciam durin1 accidnc c:oaditiou, may be inoperable for a period of no mr* than 24 hour** Prior ca initiattq repair*, all valve* a!Ul iAterlocka 1A the *Y*t* that provide the duplicate fUDCtioa aball b* te*ted to daoucrat* operability. 3-29* Amndmlllc N'o 2t 51 September 10, 1979 3.3 EMERGENCY CORE COOLING SYST'E... ,~~-*----*

g.

. *. ~ HPSI Pump operability shall be as follows: Both HPSI Pumps shall be rendered CS temperature is < 300°F unless ad is removed.

2) mum of l HPSI pump may be operable cure is ~ 300°F but < 350°F.

3) ly one, HPSI Pump shall be ra is ~ 350°F but < 43

4)

At least one whenever PCS

6)

One HPSI pump may be ade noperable when the reactor is subcritical and he PCS mperature is ~ 460°F, provided the ction 3.3.2.c are met.

7)

Whenever PCS emperature is becwe 38S°F to 430°1 and LTOP syste is not armed. then a de cated licensed operator all be stationed in the c trol room to termina an inadvertent HPSI Pump sta and stop Charging Pumps s necessary to limit PCS pressure.

8) ty Injection Actuation System (SIAS) te b performed while the PCS is between 300°1 a d 430°F *.

PSI 'pump -~_esting ma} be pf!rformed b* _..ow 430° rovicl d th* HPSI ~ manual discnarge valve is closed. 3.3.3 Prior to returning to the Power Operation Condition after every time th* plane haa been placed in the Refueling Shutdown Condition, or the Cold Shutdown Condition for more than 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 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 maintenance, repair or replacement, the following conditions shall be mat:

a. All.pressure isolation valves listed in Tabl* 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.l cannot be demonstrated, at least two valves in each high pressure line having a non-functional valve must b* !n and remain in, th* mode correspondinR ta th*

isolated condition.Cl) Motor-operated valves shall ba placed in th* clo**d position and power supplies deenargized. 3-30 TSP0189-0002-NLP4 Amendment No. JI, ti!, 117 Novemb*r 14, 1988. 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 3.3 EMERGENCY CORE COOLING SYSTEM (Continued) 3.3.3

g.

HPSI pump operability shall be as follows:

1) If the reactor head is installed, both HPSI pumps shall be rendered inoperable when:

~b,D La

a.

The PCS temperature is < 400 F, or I I I

b.

Shutdown cooling isolation valves M0-3015 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 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 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 previqus 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 isolatidn valves listed'.'"in.Table 4.3.l :;\\all 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 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.(!) lHotor-operated valves shall be placed in the closed position and power supplies deenergized. 3-30 Amendment No. Sl, l~l, 111, TSP0889-010 l-NL04. J

~

3.3 E!fERGENCY CORE COOLING SYSTE~

~

(concinued) demonstrate chat cha maximum fuel clad cem~eracures that could occur over the break size spectrum are will below chi melcing t1mp1rature of zirconium (3300°F).

Malfunction of the Low Pressure Safety Injection Flow control valve could defeat the Low Pressure Injection f1atura of th*

ECCS; therafore, it is dtsabl1d in the 'open' modi (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 valve* in conjunction with a LOCA ha* not been analyzed.

To provide a1surance that this will not occur.

th*** 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 th* 480 volt breaker* are opened.

Thu*, a failure of a breaker and a switch are required for any of th* valve* to clo***

'19.. ~

-f--

~ll':iru"=h9 aa1 &H,_, h iaepn18h e11mt*atH aawwwl1nd PSS

~~

mq* additiop1 due tg 1nad1"HIBI Mle P*P Uatts. !lath !P9I

~

~p:a:::ria~:;:s ~~:*:::::

1

=1;: ;:.*,:~::*:::~::::*::=!=::!*;21 temp*ncure h 1 438 1P.

Wrilil the Pt§ tnparature I. ! 4]0'P,..

th* pr11*nr1... HfatJ oaJ:vli insure E&ac Ebe 1'c!!i praHdfi Ulll not exceed 1 ocnso !JPfldd::ta: s l:tm:tts Wltid bbl df boEii &PSI I

I I

I P'PRP' ere.. Hted Tha raquirment to have... BPSI traidoparabla above..,..,,_

I provides added ***uranca that the ef fact* of a LOCA occuring I

under LTOP condition* would be mitigated. If a LOCA occurs when I

the prw!y sy*t.. tmpar~tura b lH* than or equal to ¥'* ur I the prH.\\:*re would drop t:: *the i.val when' low praHure safety I

injection can t,"'avent core dam.... ~**

I I

I he BPSI pump mamaal dt.c:hars* valve I

ce the cloHd valve eU.minatH the I

- Cini bain1 th* cauae of a ma.* addition I

I Raf arencaa (1) FSAJt, Sect 9.10.l; (2) FSA&, Sec 1011. 6.1, (3) !A-PAL-P.S80121 "Calculation. of Time for Operation ta Act I

for HP and lubbl*"* January 20, 1981.

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 Pressuce Injection feature of ~he 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 valves to close.

Insuring both HPSI pumps are inoperable when the PCS temperature

< ~°F or the shutdown cooling isolation valves are open eliminates PCS mass additions due to inadvertent HPSI pump starts.

Both HPSI pumps starting in conjunction with a charging/letdown imbalance may cause 10CFR50 Appendix G limits to be exceeded when the PCS temperature is < 2"1°F.

When the PCS temperature is ~ 430°F, the pressurizer safety valves ensure that the PCS pressure wi 11 not exceed 10CFR50 Appendiz c, 1 5 d t, I 11 Gilt oz hath ttpsr P'!Tf? ?ZS tnt9d 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~ pressure safety injection can prevent core damage.

5 I

J HPSI pump testing with the HPSI 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 PCS.

References (1) FSAR, Section 9.10.3; (2) FSAR, Section 6.1, TSP0889-0l01-NL04 I

I I

I

4. O SL"RVE::'.LLA...'lCE RECU!R~E~7S

4. O. l 4.0.2 Surveillance requiremencs shall be applicable durin~ the reaccor operacing condicions associaced wich individual Limicing Condicions for Operacion unless oChQrwise staced in an individual surveillance requiremenc.

L"nless 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 interval, and

b. A total maximum combined interval tim* for any thru consecutiv* surveillanc* intervals not to exceed 3.25 times th* specified survaillanca interval.

4.1 INSTRUMENTATION AND CONTROL 4.1. l Applicability Applies to th* reactor protectiv* system and oth*r critical instrumentation and controls.

Objectiv*

To specify cha minimwa frequency and type of surveillance to be appli9d to critical plant instrum*ntation and controls.

Specification*

Calibration. t~*ting, and checking of instrum*nt ehannels, rea~or prote~tiv* r.;stem and engi~~~r*d safegua~*~*~~yst.. logic channels and miacellaneou* in*trument sy*tema and control* shall be p*rformed.. specified in 4.1."l and in Table* 4.1.1 to 4.1.3.

Overpre*aur* Protection Sy*t*ll9 a.. Each PORV 1hall b* demonstrated operable by:

1. Perfor-..nc* of a channel functional teat on th* POIV actuacioa chaael, but u:cluding valve operation, within

31. daya prior to anter1D1 a condition in vbich the PORV 1* required 01terable and at leaac once per ll day*

thereafter when the POIV i* requir*d operable.

2. Performance of a channel calibration on tba PORV actuacioa channel ac lea*C once per 18 maath*.

3. VertfyiDI the PORV iaolation valve ia opea &C lea*C once per 72 hour* vb1n tba roav i* bein1 uaecl for pvarpre*aure protection.

4. Te*t1D1 in accordance vitb th* in*ervice in*pection raqutr.. enca for ASMI Section II, Sect101l IWV Cat*1ory C valve*.

4-1

. Aaaadmenc tfo JI, 51 September 10, :979 TS"P t uut-n" na_.;;:. ~ *

b.

The PCS vent(s) shall be verified to be open ae ~east 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 overpressure prot:eccion except when the '-'enc pathway is provided with a valve which is locked, seaied, or otherwise secured in the open position, then verify these valves open ac lease once per 31 days.

c.

When boch open PORV.pilst valves are used as an alcernative co vencing ch* PCS, Chen verify boch PORV ~valves and both PORV block valve1 are open at lease once per 7 days.

Bads Failures such as blown instrument fuses, defective indicacors, and faulted amplifhrs which result in "up1cale" or "downscale" indication can be easily recognized by simple observation of th* functioning of an instrument or sysc... 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 systelll8 when the plant is in operation, a checkin1 frequency of onc*-p*r-shif t is deemed adequate for reactor and steam 1y1t.. in1trumentation. Calibration* are performed to insure the p~***ntat1on and acqui1ition of accurate information.

The pov*r range safety channel* and ~T power channel* ara are calibrated daily again1t a heat balance standard to account for enor* induced by chan1~1 rod pattern* and core phy1ics parmneters.

Ocher c:hannel* ar* subject only to th*'"drift" *nor* i:ndu.:ed within th* in1truaentation it1elf and, con1equ*ntly, can tolerat* lon1*r intarval1 between calibration. Proc*** 1y1t..

in*trumentation enor* induced by drift can be expected to remain within accaptable tolerance* if recalibration i*

performed at **ell rafualilll 1hutdavn int*rval.

Sub1tantial calibration 1bift1 vi.thin a chann*l (****ntially a channel f ailura) vill be rav*al*d durin1 routine checkin1 and tHtiDS procadure1. Thu, minimua calibration frequencial Of one-per-day for th* paver ran1* *af*ty channel** and once **ell rafaelizll 1hutdavn for tb* proc*** 11*t.. channel1, are con1idered adequate.

Th* aiDtmu. te1t1D1 frequancy for tho** i111trum*nt channel*

connected to th* reactor protective 1y1c.. 11 ba1e~5 on an

- Ciliated avera1* un*af* failura rate of 1.14 z 10 failure/hour per ch&DDel; Thi* **t:blation 1* b&1ed on 11ait*d operatiDI experienca at conv*ntional and nuclear plut1.

AD "unaate failure" i* defined ** on* vbicb n*1*te* channel operability and vbicb, due to it* natura, i* ravealed oaly vban th* channel i* t**ted or attempt* to re.,ond to a bonatid* *isnaJ..

4-2 Allnchunt lfo. * * n1. 118 Navlmbar 15, 1981 I

a

4. 6 SAFETY rNJECTtON AHO CONT A! NM~T SPRA '{

SY~,.;_,,,.

4.6.l 4.6.2 Applicability Appli** to the safety injection system, the containment spray sy*tem, chemical injection system and the containment cooling sy*t*m te1c1.

Objective To verify that the subject system* will respond promptly and perform their intended function1, if required.

Spee i fi cation 1 Safety Injection Sycte.m

a.

System tests shall b* performed at eacb reactor refuelin1 interval. A te1t safety injection 1i1nal will be applied.

to initiate operation of the 1y1t*** 'nle safety injection and shutdown coolin1 sy1t** pump motor1 may be de-ener1ized for this te1t. The system vill be con1idered satisfactory if control board indication and visual ob1ervation1 indicate that all coaponents have received the safety injection

si111al in th* proper sequence and tiain1 (ie, th* appropriate pump breaker* shall have opened and clo1ed, and all valve*

shall have completed their travel).

b.

Both bi1b pre11ure safety injection pump*, P-66A !!!!& P-661 shall be d..an1trated inoperable at lea1t once per 12 houri whenever the t*perature of one or more of tb9 PCS cold le11 i1 < 1tt*r unle11 the reactor bead ia r..aved.

."8li

.I.. '0.,:

Containment S ra S 1U*

a.

ball be performed at eacb reactor refuelin1 teat sball be performed vitb th* iaalation valve* in

1pray supply line* at the containment block8d peratioa of the 1y1t.. i* initiated by trippin1 actuatioa inatruaentation.

b.

every fiv* J9ar1 the 1pray noulH 1ball be ta be opea. *

e.

teat vill be coa1idered 1ati1f actary if vi1ual

nation* indicau all companenta have operatM dtfac:corily.

c>- 1J.rJ,,,7Jo~ e_o.,,/11ie,, 110Jv44

!11/D-301, ~

t¥l~

4-39 Allendaenc Yo. si, 1J, *** 117 November 14, 1981 TSP1187-0218-tn.04 I

I I

4.6 4.6.J 4.6.4 4.6.5 SAFETY rNJECTrON AND CONTArNHENT S?RA'l

a.

The safety injection pump1, shutdown cooling pump1, and containment spray pump1 shall be started at intervals not to eKceed three months.

Alternate manual starting between control room console and the local breaker shall be practiced in the test program.

b.

Acceptable level* of performance stt.11 be that the pumps start, reach their rated shutoff head1 at miniaaam recirculation flow, and operate for at leaat fifteen minutes.

Valves Containment Air Cooling Sr1tem

a.

Emer1enc1 mode automatic valve and fan operation will be checked for operability durin1 eacb refuelin1 shutdown.

b.

Each fan and valve required to function durin1 accident condition* vill be ezerci1ed at interval* not to ezceed three months.

i)

The safety injection s11t.. and the containment spray 1y1te* are princip1:~ plant safety featur11 thAt are Mrmally inoperae.~ve.

duria1 reactor operation.

C0919leta 1711:.. tHtl cannot be perf ormd wha the ructor ii operatia1 becaa.. a 1af1ty injection 1ip1al cau111 contaia.ent i1olatioa and a coatai11111at spray 1y1t.. te1t require* the sy1t** to be temporarily di1abled.

Tile.. chod of a11urin1 operability of th*** 171t..

1 i1 therefore to combine 171t..

tHU to b9 perfoned duria1 annual plant 1butdow1, vitb. more f requeat co..-*t CHU, which cm be performed duria1 reactor operaciaa.

Tba aaaual 171c.. t11t1 dm1Da1trate proper automatic operation of the 1afac7 injection and coatai11111nc 1pra1 17at.... A t11t 1i1aal i1 applied to initiate automatic action and verification mad* that the compoaaat1 receive the 1afat7 injection in the proper sequi.ace. The test d1.an1trat11 the operation of the valve*, puap circuit brukar1, and auto.. tic circuitry. (1, 2) 4-40 Amendment 10. St, 7J, 11, 117 Mov.. ber 14, 1911 TSP111T-0211-llL04 I

I

4.6 SAFETY

SPRAY SYSTEMS TESTS (Conti.nued)

~

During reactor operation, the instrumentation ~hich is depended on to initiate safety injection and containment spray is g*n*rally checked daily and the initiating circuits are tested monthly.

In addition, the active components (pumps and valves) ar* 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. Th* test interval of three months is based on the judgment that morefrequent 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.

Verifica~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 plugaing mecbanis11 available, the retut every five years is considered to be mor* than adequate.

Other systems that are also important to the emergency coolin&

fUiletion are th* SI taDKi, the campon*nt coolin1 1yitWiil, the service water syste11 and the containment. air coolers.

'11le SI tanks are a pa11ive safety feature.

In accordance with the specifications, the water volume and pressure in the SI tanks

&H checked periodically.

'11le ocher sy1t*s *ntioned operate when the reactor is in operation and by th*** means are continuously monitored for satisfactory performance.

esld le& tE&&MtitUti is la** tha* lOo*r, -~-

- - rt a' on* HPII '!Ml' 11~1* aauaa eha Appulldii G tlSiEI ea Y*

&a b* ****adedt ch111fot1, bt5tb puapl ifi PIB&iti& taapatablcu=

References (1)

FSAI., Section 6.1.3.

(2)

FSAI., Section 6.2.3.

4-41 Amnd*ntMo. 117 Mov*ber 14, 1981 TSP1187-0218-NL04 I

I I

4.6 SAFETY INJECTION AND CONTAINMENT SPRAY SYSTEMS TESTS (Continued)

Basis (continued)

During reactor operation, the instrumentation 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 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, 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 performance.

With the reactor vessel ~nsta~led when the PCS cold leg temperature is less than ~°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 Amendment No. 117, I

I I

I

_

ML18054B009

Text

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