ML18052B429

Proposed Tech Specs,Lowering Relief Valve Setting Over Temp Range & Providing Low Temp Overpressure Protection Up to Temp Where Pressurizer Safety Valves Provide Protection Against Brittle Fracture

ML18052B429
Person / Time
Site:	Palisades
Issue date:	12/22/1987
From:	CONSUMERS ENERGY CO. (FORMERLY CONSUMERS POWER CO.)
To:
Shared Package
ML18052B428	List: ML18052B427 ML18052B429
References
NUDOCS 8712290287
Download: ML18052B429 (15)

Text

ATTACHMENT Consumers Power Company Palisades Plant Docket 50-255 TECHNICAL SPECIFICATION PAGE CHANGES Low Temperature Overpressure Protection December 22, 1987 I'*.,..8.71~0287 871222 i *. ;.

'1,

• PDR 11'.\\DOCK 05000255 *.

p :...

• • .{.

PDR *J J.:. J*

14 Pages TSP1187-0218-NL04

3. 1 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 pressure shall not exceed 1250 psia.

A minimum temperature of 100°F is required.

Only ten cycles are permitted.

(5)

Maximum secondary leak test pressure shall not exceed 1000 psia.

A minimum temperature of 100°F is required.

(6)

In performing the tests identified in 3.l.l.e(4) and 3.l.l.e(5), above, the secondary pressure shall not exceed the primary p~essure by more than 350 psi.

f.

Nominal primary system operation pressure shall not exceed 2100 psia.

g.

The reactor inlet temperature (indicated) shall not exceed the value given by the following equation at steady state 100%

power operation:

T

< 538.0 + 0.03938 (P-2060) + 0.00004843 (P-2060) 2 + 1.0342 (W-120.2)

.. inlet Where: Ti 1

= reactor inlet temperature in F 0

n et p = nominal operating pressure in psia 6 W = total recirculating mass flow in 10 lb/h corrected to the operating temperature conditions.

Note:

This equation is shown in Figure 3-0 for a variety of mass flow rates.

h.

During init_ial primary coolant pump starts (i.e., initiation I

of forced circulation), secondary system temperature in the I

steam generators shall be < the PCS cold leg temperature I

unless the PCS cold leg temperature is > 450°F.

I 3-lc Proposed TSP1187-0218-NL04

3.1 PRIMARY C~ANT SYSTEM (Continued) allowed during normal operation, so that substantial safety margin exists between this pressure differential and the pressure differential required for tube rupture.

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 126.9 x 106 lb/h minus 6t3ro'account for flow measurement uncertainty and core flow bypass.

A steady state DNB analysis was also performed (assuming 1157. overpower, 50 psi for pressure uncertainty, 3% for flow measurement uncertainty, and 37. for core flow bypass) in a parametric fashion to determine the core inlet temperature as a function of *pressure and flowc£~r which the minimum DNBR at 1157. overpower is equal to 1.30.

The result of this steady state DNB analysis was the following equation for limiting reactor inlet temperature:

T

_< 541.0.+ 0.03938 (P-2060) + 0.00004843 (P-2060) 2 +

inlet 1.0342 (W-120.2)

A temperature measurement uncertainty of 3°F was subtracted from this limit in arriving at the LCO given in' Section 3.1.1.g.

The nominal full power inlet temperature is 2°F less than the value given in Section 3.1.1.g to allow for drift within the temperature control band.

Thus, a total uncertainty of 5°F is applied to the limiting reactor inlet temperature equation.

The limits of validity of this equation are:

1850 < Pressure < 2250 psia 6

6 110.0 x 10

< Vessel Flow < 130 x.10 Lb/h The requirement that the *steam generator temperature be < the PCS I

temperature when forced circulation is initiated in the PCS I

ensures that an energy addition caused by heat transferred from I

the secondary system to the PCS will not occur *. This requirement I

applies only to the initiation of forced circulation (the start I

. of the first primary coolant pump) when the PCS cold leg I

temperature is < 450°F.

At or above 450°F, the PCS safety valves I

• prevent the PCS pressure from exceeding 10CFRSO Appendix G limits.

I References (1)

FSAR, Sections 6.1.2.2 and 14.3.2.

(3)

XN-NF-77-18.

(2)

FSAR, Section 4.3.7 (4)

XN-NF-77-22.

(S)

"Palisades Plant Overpressurization Analysis," June, 1977, and "Palisades Plant Primary Coolant System Overpressurization Subsystem Description," October, 1977.

3-3 Proposed TSP1187-0218-NL04

3.1 PRIMARY COOLANT SYSTEM (Cont'd) 3.1.2 Heatup and Cooldown Rates The primary coolant pressure and fhe system heatup and cooldown rates shall be limited in accordance with Figure 3-1, Figure 3-2 and as follows when the PCS is not vented through a vent equal to or greater I

than 1.3 square inches.

I

a.

b.

Co Allowable combinations of pressure and temperature for any heatup rate shall be below and to the right of the applicable limit line as shown on Figure 3-1.

The average heatup rate in any one hour time period shall not exceed the heatup rate limit when one or more PCS cold leg is less than the corresponding "Cold Leg Temperature" below.

1

2.

3.

4.

• Cold Leg Temperature

< 190°F

-> 190°F

> 310°F

> 450°F and < 310°F and < 450°F Heatup Rate Limit 20°F/Hr 40°F/Hr 60°F/Hr 100°F/Hr Allowable combinations of pressure and temperature for any cooldown rate shall be below and to the right of the applicable limit lines as shown on Figure 3-2.

The average cooldown rate shall not exceed the Cooldown Rate Limit when one or more PCS cold legs is less than**

the corresponding "Cold Leg Temperature" below:

• Cold Leg Tem~erature Cooldown Rate Limit

1.

> 450°F 100°F/Hr

2.

> 300°F and < 450°F 60°F/Hr

3.

< 300°F and >180°F 40°F/Hr

4.

< 180°F 20°F/Hr Allowable combinations of pressure and temperature for inservice testing during heatup are as shown in Figure 3-3.

The maximum heatup and cooldown rates required by Sections a and b, above, are applicable.

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

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

d.

1.

The average heatup rates for the pressurizer shall not exceed I

100°F/hr in any one hour time period when the PCS cold leg I

temperature is less than 450°F.

I

2.

The average cooldown rate for the pressurizer shall not exceed I

200°F/hr for any one hour time period.

I

• Use shutdown cooling return temperature if the shutdown cooling system is in I

operation and all PCP's are off.

I 3-4 Proposed TSP1187-0218-NL04

3. L 2 Heatup and Cooldown Rates (Cont'd)

e.

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 the following criteria and procedure:

1.

US __ Nuclear Regulatory Commission Regulatory Guide 1. 99 has been used to predict the increase in transition temperature based on integrated fast neutron flux and surveillance test data.

If measurements on the irradiated specimens show increase above this curve, a new curve shall be constructed such that it is above and to the left of all applicable data points.

2.

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

The total integrated reactor thermal power from start-up to the end of the new power period shall be converted to an equivalent integrated fast neutron exposure (E ~ 1 MeV).

Such a conversion shall be made consistent with the dosimetry evaluation of capsule W-290(l 2).

3.

The limit lines in Figures 3-1, 3-2 and 3-3 are ba$ed 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(B).

Basis All components in the primary coolant system are designed to withstand the effects of cyclic loads due to primary system temperature and (1) pressure changes.

These cyclic loads are introduced by normal unit load transients, reactor trips and start-up and shutdown opera_tion.

During unit start-up and shutdown, the rates of temperature and pressure changes are limited.

A plant heatup and cooldown limit of 100°F per I

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 wela has the highest RTNDT of plate, weld and RAZ materials at the fluence to which the Figures 3-1, 3-2 and 3-3 apply. (lO)

The unirradiated RTNDT (11) has been determined to be -56°F.

An RTNDT of -56°F is used as an unirradiated value to which irradiation effects are added.

In addition, 3-5 Proposed TSP1187-0218-NL04

3.1.2 Heatup and Cooldown Rates (Cont'd)

Basis (Cont'd) evaluated.

During cooldown, the 1/4 thickness location is always more limiti~g in th~t 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 mechanic's 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.

'rhe revised pressure-temperature limits are applicable to react.or vessel inner wall fluences of up to 1.8 x l0 19nvt.

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 183°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 ins.ervice system hydrostatic pressure test.

That temperature is calculated based upon 2310 psig inservice hydrostatic test pressure.

The restriction of average heatup and cooldown rates to 100°F/h when I

all PCS cold legs are > 450°F and the maintenance of a pressure-I 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 6, 7, 8 and 9 are met.

The core operational limit applies only when the reactor is critical.

The heatup and cooldown rate restrictions applicable when the I

temperature of one or tnore of the PCS cold legs is less than 450°F I

are consistent with the analyses performed for low temperature I

overpressure protection (LTOP) (References 13 and 14).

At 450°F or I

above, the PCS safety valves provide overpressure protection for

./

heatup or cooldown rates < 100°F/hr.

I 3-7 Proposed TSP1187-0218-NL04

3.1.2 Heatup and Cooldown Rates (Cont'd)

Basis (Cont'd)

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. (7)

Due to the shifts in RTNDT' NDTT requirements associated with nonreactor vessel materials are, for all practical purposes, no longer limiting.

References (1)

FSAR, Section 4.2.2.

(2)

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

(3)

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

Unirradiated Mechanical Properties,"

August 25, 1977.

(4)

Battelle Columbus Laboratories Report, "Palisades Nuclear Plant Reactor Vessel Surveillan.ce Program:

Capsule A-240," March 13, 1979, submitted to the NRC by Consumers Power Company letter dated..

July 2, 1979.

(5)

FSAR, Section 4.2.4.

(6)

US Nuclear Regulatory Commission, Regulator Guide 1.99, "Effects of Residual Elements on Predicted Radiation Damage to Reactor Vessel Materials," July, 1975.

(7)

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

(8)

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

(9) 10 CFR Part SO, Appendix G, "Fracture Toughness Requirements,"

May 31, 1983.

(10) US Nuclear Regulatory Commission, Regulatory Guide 1.99, Draft Revision 2, April, 1984.

(11) Combustion Engineering Report CEN-189, December, 1981.

(12) "Analysis of Capsules T-330 and W-290 from the Consumers Power Company Palisades Reactor Vessel Radiation Surveillance Program,"

WCAP~10637, September, 1984.

(13) Analysis for E-PAL-87-lOlG "Calculation of PCS Pressure Increase From Adding 133 gpm (3 charging pumps) Before the PORVs Open,"

November 4, 1987.

(14) Analysis for E-PAL-85-lOlG "Palisades Heatup and Cooldown Curves," November 4, 1987.

3-8 Proposed TSP1187-0218-NL04 l

I I

I I

I

3.1.8 OVERPRESSURE PROTECTION SYSTEMS LIMITING CONDITIONS FOR OPERATION 3.1.8.1 REQUIREMENTS

a.

When the temperature of one or more of the primary coolant system cold legs is < 300°F, or whenever the shutdown cooling isolation valves (MOV-3015 and MOV-3016) are open, two power operated relief valves (PORVs) with a lift setting of < 310 psia shall be operable, or a reactor coolant system vent of-

> 1.3 square inches shall be open, or both PORV pilot valves and both PORV block valves shall be open.

b.

When the temperature of one or more of the primary system cold legs is < 430°F, two power operated relief valves (PORVs) with a lift setting of < 575 psia shall be operable except as specified in section-c. below.

c.

When a bubble is formed in the pressurizer and the actual pressurizer level is < 60 percent and the temperature of all the primary coolant system cold legs is > 385°F, PORV operability is not required.

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 and within the next 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> either vent the PCS through a > 1.3 square inch vent or open both PORV pilot valves and both PORV block valves.

b.

With both PORVs inoperable, depressurize and within the next 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> either vent the PCS through a > 1.3 square inch vent or open both PORV pilot valves and both PORV block valves.

c.

The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.

I I

I I

I I

I I

I I

I I

I I

I I

I I

I I

I I

I

/*

I I

I I

I I

I I

I I

I I

I I

I I

I I

3-25a Proposed TSP1187-0218-NL04

3.1.8 OVERPRESSURE PROTECTION SYSTEMS LIMITING CONDITIONS FOR OPERATION Basis There are three pressure transients which could cause the PCS pressure to exceed the pressure limits required by 10CFR50 Appendix -G.

They are:

(1) a charging/letdown imbalance, (2) the start of a high pressure safety injection (HPSI) pump, and (3) initiation of forced circulation in the PCS when the steam generator temperature is higher than the PCS temperature.

Analysis (Reference 1) shows that when three charging pumps are operating and letdown is isolated, the PORV setpoints of < 310 psia up to 300°F and < 575 psig between 300°F and 430°F ensure-that 10CFR50 Appendix-G pressure limits will not be exceeded.

Analysis (Reference 2) further shows that if the PCS temperature is > 385°F and there is a bubble in the pressurizer and the actual pressurizer liquid level is < 60%, operator action within 10 minutes of the time letdown is isolated can prevent the PCS pressure from exceeding 10CFR50 Appendix G pressure limits.

Above 430°F, the pressurizer safety valves prevent 10CFR Appendix G limits from being exceeded by a charging/letdown imbalance (Reference 3)

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) with one or more of the PCS cold leg temperatures < 450°F.

Requiring the PORVs to be operable when the shutdown cooling system is not isolated (M0-3015 and M0-3016 open) from the PCS ensures that the shutdown cooling system will not be pressurized above its design pressure.

The requirement for the PCS to be depressurized and vented by an opening > 1.3 square inches (Reference 5), or by opening both PORV pilot valves and both PORV block valves when one or both PORVs are inoperable ensures that the 10CFR50 Appendix G pressure limits will not be exceeded when one of the PORVs is assumed to fail per the "single failure" criteria 10CFR50 Appendix A, Criterion 34.

3-25b I

I I

I

- I I

I I

I I

I I

I I

I I

I I

I I

I I

I

/*

I I

I I

I I

I I

I I

I I

I I

I I

I I

I I

I I

Proposed TSP1187-0218-NL04

3.1.8 OVERPRESSURE PROTECTION SYSTEMS LIMITING CONDITIONS FOR OPERATION 3.1. 8 Basis (continued)

References

1.

Analysis for E-PAL-87-lOlG "Calculation of PCS Pressure Increase From Adding 133 gpm (3 charging pumps) Before the PORVs Open,"

November 4, 1987.

2.

Analysis for E-PAL-85-lOlG "Palisades Heatup and Cooldown Curves," November 4, 1987.

3.

Technical Specification 3.1.2.

4.

"Palisades Plant Overpressurization Analysis," June 1977 and "Palisades Plant Primary Coolant System Overpressurization Subsystem Description," October, 1977.

3-25c

. TSP1187-0218-NL04 I

I I

I I

I

. I I

I I

I I

I I

I I

I I

I I

I Proposed

3.3 EMERGENCY CORE COOLING SYSTEM (Cont'd)

g.

The cont.rol fuses for the high pressure safety injection pump (P-66A and P-66B) motors shall be removed when the temperature*

of one or more of the primary system cold legs is < 430°F unless the reactor head is removed.

3.3.3 Prior to returning to the Power Operation Condition after every time the plant has been placed in the Refueling Shutdown Condition, or the Cold Shutdown Condition for more than 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> and testing of Specification 4.3.h has not been accomplished in the previous 1

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. (1)

c. If Specification a. and* b. cannot be met, an orderly shutdown shall be initiated and the reactor shall be in hot shutdown condition within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, and cold shutdown within the next 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

Basis The normal procedure for starting the reactor is, first, to heat the primary coolant to near operating temperature by running the primary coolant pumps.

The reactor is then made critical by withdrawing control rods and diluting boron in the primary coolant. (l) With this mode of start-up, the energy stored in the primary coolant during the approach to criticality is substantially equal to that during power operation and, therefore, all engineered safety features and auxiliary cooling systems are required to be fully operable.

During low-temperature physics tests, there is a negligible amount of stored energy in the primary coolant; therefore, an accident comparable in Motor-operated valves shall be placed in the closed position and power supplies deenergized.

3-30 Proposed TSP1187-0218-NL04 I

I I

• /

3.3 EMERGENCY CORE COOLING SYSTEM (cont) that 25% of their combined discharge rate is lost from the primary coolant system out the break.

The transient hot spot fuel clad temperatures for the break sizes considered are shown on FSAR Figures 14.17.9 to 14.17.13.

These demonstrate that the maximum fuel clad temperatures that could occur over the break size spectrum are well below-the melting temperature of zirconium (3300°F).

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

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

The inadvertent 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*o~ the valves to close.

Removal of the control fuses for the HPSI pumps eliminates PCS I

mass additions due to inadvertent pump starts.

HPSI pump starts I

in conjunction witn a charging/letdown imbalance may cause 10CFR50 I

Appendix G limits to be exceeded when the PCS temperature is I

< 430°F.

When the PCS temperature is ~ 430°F, the pressurizer I

safety valves ensure that the PCS pressure will not exceed 10CFR50

/

Appendix G limits when one or both HPSI pumps are started.

I References (1)

FSAR, Section 9.10.3.

(2)

FSAR, Section 6.1.

3-33 Proposed TSP1187-0218-NL04

b.

The PCS vent(s) shall be verified to be open at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> 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 ooth open PORV pilot valves are used as an alternative to I

'venting the PCS, then verify both PORV pilot valves and both I

PORV block valves are open at least once per 7 days.

I Basis I{

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.

The power range safety channels 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 Tevealed 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 estimated average

- -s- -

unsafe failure rate of 1.14 x 10 failure/hour per channel.

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

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

4-2 Proposed TSP1187-0218-NL04

4.6 SAFETY INJECTION AND CONTAINMENT SPRAY SYSTEMS TESTS 4.6.1 4.6.2-4.6.3 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 I

interval. *A test safety injection signal will be applied to I

initiate operation of the sys.tern.

The safety injection and I

shutdown cooling system pump motors may be de-energized for I

this test.

The system will be considered satisfactory if I

control board indication and visual observations indicate that I

all components have received the safety injection signal in I

,the proper sequence and timing (ie, the appropriate pump

/-

breakers shall have opened and closed, and all valves shall I

have completed their travel).

I

b.

Both high pressure safety injection pumps shall be demonstrated**

I inoperable at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> whenever the temperature I

of one or more of th~ PCS cold legs is < 430°F by verifying I

that the control system fuses and their fuse holders for the

/

HPSI pump breakers (P66A, P66B) have been removed from the I

circuit unless the reactor head is removed.

I 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.

a.

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

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

4-39 Proposed TSP1187-0218-NL04

4.6 SAFETY INJECTION AND CONTAINMENT SPRAY SYSTEMS TESTS (Contd)

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.

Unless the reactor head is removed, the control fuses for both HPSI I

pump breakers are removed from the circuit when the PCS temperature I

is < 430°F to prevent a mass addition and the accompanying PCS I

pressure increase which would result if a HPSI pump were started, I

That resultant pressure increase could cause the pressure limits I

required by 10CFR50 Appendix G to be exceeded.

I References (1) FSAR, Section 6.1.3.

(2) FSAR, Section 6.2.3.

4-41 Proposed TSP1187-0218-NL04