ML18053A308

Proposed Tech Specs,Updating Low Temp Overpressure Protection Requirements

ML18053A308
Person / Time
Site:	Palisades
Issue date:	04/12/1988
From:	CONSUMERS ENERGY CO. (FORMERLY CONSUMERS POWER CO.)
To:
Shared Package
ML18053A306	List: ML18053A304 ML18053A308 ML18053A332
References
NUDOCS 8804190049
Download: ML18053A308 (21)

Text

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Consumers Power Company Palisades Plant Docket 50-255 TECHNICAL SPECIFICATION PAGE CHANGES Low Temperature Overpressure Protection April 12, 1988 20 Pages TSP1187-0218-NL04

3.1 PRIMARY COOLANT SYSTEM (Continued) 3 .1.1 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 s.ufficient 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 3.1.1.e(4) and 3.1.1.e(5), above, the secondary pressure shall not exceed the primary pressure by more than 350 psi.

f. Nominal primary system operation pressure shall not exceed 2100 psia.

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

power operation:

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

Where: T. 1 t in e p

= reactor inlet temperature in F 0

= 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 initial 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

' . ~* 't. -~ .'._ * ' . . .

Amendment No. Jt, JJ,

~TSP1t81(0218-NL04

COO~ SYSTEM

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

~econdary 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 6

hot zero power (532°F) of 126.9 x 10 lb/h minus 6f jo account for flow measurement uncertainty and core flow bypass. 3 A steady state DNB analysis was also performed (assuming 115% overpower, 50 psi for pressure uncertainty, 3% for flow measurement uncertainty, and 3% 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 115% overpower is equal to 1.30. The result of this steady state DNB analysis was the following equation for limiting reactor inlet temperature:

2 T. l 1n et < 541.0 + 0.03938 (P-2060) + 0.00004843 (P-2060) +

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 110.0-x 10 < Ve~sel Flow< 130 x 10 6 Lb/h 6

The requirement that the steam generator temperature be < the PCS I temperature when forced circulation is initiated in the PCS /

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 /

. 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.

(5) "Palisades Plant Overpressurization Analysis," June, 1977, and "Palisades Plant Primary Coolant System OverpressurizationSubsystem Description," October, 1977.

3-3 Amendment No. it, jt, TSP1187-0218-NL04

3.1 PRIMARY COOLANT SYSTEM (Cont'd) 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. I

a. Allowable combinations of pressure and temperature for any heatup rate shall be below and to the right of the applicable limit line as I shown on Figure 3-1. The average heatup rate in any one hour time I period shall not exceed the heatup rate limit when one or more PCS I cold leg is less than the corresponding "Cold Leg Temperature" I below. I I

• Cold Leg Temperature Heatup Rate Limit I I

1 < 190°F 20°F/Hr I

2. > 190°F and < 310°F 40°F/Hr I

3. > 310°F and < 450°F 60°F/Hr I

4. > 450°F 100°F/Hr I I

b. Allowable combinations of pressure and temperature for any cooldown -

rate shall be below a~d 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 I the corresponding "Cold Leg Temperature" below: I I

• Cold Leg Temperature Cooldown Rate Limit I I

1.

2.

4S0°F 100°F/Hr I 300°F and < 450°F 60°F/Hr I

3. -< 300°F and >180°F 40°F/Hr I

4. < 180°F 20°F/Hr I I

c. Allowable combinations of pressure and temperature for inservice testing during heatup are as shown in Figure 3-3. The maximum I heatup and cooldown rates required by Sections a and b, above, I are applicable. Interpolation between limit lines for other than I the noted temperature change rates is permitted in 3.1.2a, b or c. 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 4S0°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

_Amendment No. " , 'J7, TSP1187-0218-NL04

. l 3.1.2 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 a.ccordance 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 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(S)~ .

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

During unit start-up and shutdown, the rates of temperature and pressure changes are limited. A plant heattip and cooldown limit of 100°F per hour is consistent with the design number of cycles .and satisfies 2

stress limits for cyclic operation. ( )

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

- . - (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 Amendment No. ~~' ~7, TSP1187-0218-NL04

. IL . '

3.1.2 Heatup and Cooldown Rates (Cont'd)

Basis (Cont'd) evaluated. During cooldown, the 1/4 thickness location is always mor~ 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 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 ?0°F per hour.

The revised pressure-temperature limits are applicable to reactor vessel inner wall fluences of up to 1.8 x 1019nvt. 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

) 0 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 ~ 450°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 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 more 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, 14, 15, 16 & 17). //

At 450°F or above, the PCS safety valves provide overpressure I protection for heatup or cooldown rates < 100°F/hr. I 3-7 Amendment. No. 89, 97, 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 Surveillance 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 50, 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) EA-PAL-85-101 "Calculation of PCS Pressure Increase from Adding II 133 gpm (3 charging pumps) Before the PORVs Open," November 4, II 1987. II (14) EA-PAL-LTOP-880119 - "Calculation of Required PORV Capacity to II Maintain the PCS Below Appendix G," January 19, 1988. II (15) EA-PAL-LTOP-880120 Rev. A - PORV Flow Capacity at Expected II LTOP Conditions" February 15, 1988. II (16) EA-PAL-LTOP-880121 - "Calculation of Time for Operator to Act II for HPSI and Bubble" - January 20, 1988. II 3-8 Amendment No. ~9. 97, TSP1187-0218-NL04

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3.1.8 e

OVERPRESSURE PROTECTION SYSTEMS I I

LIMITING CONDITIONS FOR OPERATION I I

I REQUIREMENTS. I I

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

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

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

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

APPLICABILITY: When the temperature of one or more of the I primary coolant system cold legs is less than 430°F. I I

ACTION: I I

a. With one PORV inoperable, either restore the inoperable PORV to I operable status within 7 days or depressurize and within the I 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 I vent or open both PORV pilot valves and both PORV block valves. I I

I

b. With both PORVs inoperable, depressurize and within the next I 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 I open both PORV pilot valves and both PORV block valves. I I

c. The provisions of Specifications 3.0.3 and 3.0.4 are not I applicable. I Basis I I

There are three pressure transients which could cause the PCS I pressure to exceed the pressure limits required by 10CFR50 I Appendix G. They are: (1) a charging/letdown imbalance, (2) the I start of a high pressure safety injection (HPSI) pump, and I (3) initiation of forced circulation in the PCS when the steam l generator temperature is higher than the PCS temperature. I II 3-25a Amendment No. Ji, TSP1187-0218-NL04

. J. ..

3 .1.8 OVERPRESSURE PROTECTION SYSTEMS I I

LIMITING CONDITIONS FOR OPERATION I I

I 3 .1.8. Basis (continued)

Analysis (Reference 1, 4 & 5) shows that when three charging pumps //

are operating and letdown is isolated and a spurious HPSI occurs, II the PORV setpoints ensure that 10CFR50 Appendix G pressure limits //

will not be exceeded. Above 430°F, the pressurizer safety valves //

prevent 10CFR Appendix G limits from being exceeded by a charging/ //

letdown imbalance (Reference 2). II II The requirement that steam generator temperature be < the PCS //

temperature when forced circulation is initiated in the PCS I ensures that an energy addition caused by heat transferred from the I secondary system to the PCS will not occur. Tqis requirement I applies only to the initiation of forced circulation (the start of I the first primary coolant pump) with one or more of the PCS cold I leg temperatures < 450°F.

• I I

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

The requirement for the PCS to be depressurized and vented by an I opening> 1.3 square inches (Reference 3), 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 I limits will not be exceeded when one of the PORVs is assumed to I fail per the "single failure" criteria 10CFR50 Appendix A, I Criterion 34. I Section 3.3.2.g(6) requires a dedicated operator when the PORVs //

are inoperable as allowed in Section 3.1.8.1.c. Analysis justifying //

this condition is referenced in the Basis for Section 3.3. //

References I I

1. EA-PAL-85-101 "Calculation of PCS Pressure II Increase From Adding 133 gpm (3 charging pumps) Before the //

PORVs Open," November 4, 1987. //

2. Technical Specification 3.1.2. I

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

4. EA-PAL-LTOP-880119 - "Calculation of Required PORV Capacity //

to Maintain the PCS Below Appendix G Curves," January 19, 1988. //

5. EA-PAL-LTOP-880120 - "Palisades LTOP PORV Flowrate Capacity //

When PCS Temperature is 300°F or Greater." *January 20, 1988. II II 3-25b Amendment No. 72, TSP1187-0218-NL04

. l .*

3.3 EMERGENCY CORE COOLING SYSTEM (Cont'd) 3.3.2 g. HPSI Pump operability shall be as follows: II II

1) Both HPSI Pumps shall be renciered inoperable wh-enever II PCS temperature is < 300°F unless the reactor vessel II head is removed. II II

2) One, and only one, HPSI Pump shall be operable whenever II PCS temperature is > 350°F but < 430°F. II II

3) At least one HPSI Pump shall be operable whenever PCS II temperature is~ 430°F but <460°F. II II

4) Both HPSI Pumps shall be operable whenever the PCS II temperature is~ 460°F. II II

5) One HPSI pump may be made inoperable when the reactor II is subcritical and the PCS temperature is > 460°F, II provided the requirements of Section 3.3.2:c tre met. II

6) Whenever PCS temperature is between 385°F to 430°F and II LTOP system is not armed, then a dedicated licensed II operator shall be stationed in the control room to II terminate an inadvertent HPSI Pump start and stop Charging II Pumps as necessary to limit PCS pressure. II II

7) Safety Injection Actuation System (SIAS) testing shall not II be performed while the PCS is between 300°F and 430°F. II HPSI pump testing may be performed below 430_°F provided II the HPSI pump manual discharge valve is closed. II 3.3.3 Prior to returning to the Powei 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 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 c~ndition. (l)

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 />.

3-30 Amendment No. $1, 101, TSP1187-0218-NL04.

. l .*

3.3 EMERGENCY CORE COOLING SYSTEM (Cont'd)

Basis The normal procedure fot 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 severity to the design basis accident is not possible and the engineered safeguards' systems are not required.

The SIRW tank contains a minimum of 250,000 gallons of water containing 1720 ppm boron. This is sufficient boron concentration to provide a 5% shutdown margin with all control rods withdrawn and a new core at a temperature of 60°F.

Heating steam is provided to maintain the tank above 40°F to prevent freezing. The 1% boron (1720 ppm) solution will not precipitate out above 32°F. The source of steam during normal plant operation is extraction steam line in the turbine cycle.

The limits for the safety injection tank pressure and volume assure the required amount of water injection during an accident and are based on values used for the accident analyses. The minimum 186-inch level corresponds to a volume of 1103 ft 3 a~d the maximum 198-inch level corresponds to a volume of 1166 ft 3 .

Prior to the time the re~ctor is brought critical, the valving of the safety injection system must be checked for correct alignment and appropriate valves locked. Since the system is used for shutdown cooling, the valving will be changed and must be properly aligned prior to start-up of the reactor.

The operable status of the various systems and components is to be demonstrated by periodic tests. A large fraction of these tests will be performed while the reactor is operating in the power range. If a component is found to be inoperable, it will be possible in most cases to effect repairs and restore the system to full operability within a relatively short time. For a single component to be inoperable does not negate the ability of the system to perform its function, but it reduces the redundancy provided in the reactor design and thereby limits the 1

Motor-operated valves shall be placed in the closed position and power supplies deenergized.

3-31 rt,Vt/J.fttM~;,t It Amendment No. ~~.

TSP1187-0218-NL04

. I .* .

3.3 EMERGENCY CORE COOLING SYSTEM Basis (continued) ability to tolerate additional equipment failures. To provide maximum assurance that the redundant component(s) will operate if required to do so, the redundant component(s) is to be tested prior to initiating repair of the inoperable component. If it develops that (a) the inoperable component is not repaired within the specified allowable time period; or (b)' a second component in the same or related system is found to be inoperable, the reactor will initially be put in the hot shutdown condition to provide for reduction of the decay heat from the fuel and consequent reduction of cooling requirements after a postulated loss-of-coolant accident. This will also permit improved access for repairs in some cases. After a limited time in hot shutdown, if the malfunction(s) is not corrected, the reactor will be placed in the cold shutdown condition utilizing normal shutdown and cooldown procedures. In the cold shutdown condition, release of fission products or damage of the fuel elements is not considered possible.

The plant operating procedures will require immediate action to effect repairs of an inoperable component and, therefore, in most cases, repairs will be completed in less than the specified allowable repair times. The limiting times to repair are intended to: (1) Assure that operability of the component will be restored promptly and yet, (2) allow sufficient time to effect repairs using safe and proper procedures.

The requirement for core cooling in case of a postulated loss- .

of-coolant accident while in the hot shutdown condition is significantly reduced below the requirements for a postulated loss-of-coolant accident during power operation. Putting the reactor iri the hot shutdown condition reduces the consequences of a loss-of-coolant accident and also allows more free access to some of the engineered safeguards components in order to effect repairs.

Failure to complete repairs within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> of going to the hot shutdown condition is considered indicative of a requirement for major maintenance and, therefore, in such a case, the reactor is to be put into the cold shutdown condition.

With respect to the core cooling function, there t~)functional redundancy over most of the range of break sizes.

Adequate core cooling for the break spectrum up to and including the 42~inch double-ended break is assured with the minimum safety injection which is defined as follows: For the system of four passive safety injection tanks, the entire contents of one tank are assumed to be unavailable for emergency core cooling. In addition, of the two high-pressure safety injection pumps and the two low-pressure safety injection pumps, only one of each type is assumed to operate; and also 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 3-32 Amendment No. 9~,

TSP1187-0218-NL04.

- t " ~

3.3 EMERGENCY CORE COOLING SYSTEM Basis (continued) demonstrate that the maximum fuel clad temperatures that could occur over the break size spectrum are well below the melting temperature of zirconium (3300°F).

Malfunction of the Low Pressure Safety Injection Flow control valve could defeat the Low Pressure Injection feature of the ECCS; therefore, it is disabled in the 'open' mode (by isolating the air supply) during plant operation. This action assures that it will not block flow during Safety Injection.

The inadvertent 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 one HPSI pump is inoperable eliminates unanalyzed PCS II mass additions due to inadvertent two pump starts. Both HPSI *II pumps starting in conjunction with a charging/letdown imbalance //

may cause 10CFR50 Appendix G limits to be exceeded when the PCS //

temperature is < 430°F. When the PCS temperature is > 430°F, //

the pressurizer safety valves ensure that the PCS pressure will //

not exceed 10CFR50 Appendix G limits when one or both HPSI //

pumps are started. II The requirement to have one HPSI train operable above 350°F II 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 350°F, //

the pressure would drop to the level where low pressure safety //

injection can prevent core damage. //

Analysis (Reference 3) 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%, and LTOP is not armed, //

operator action, within 2.9 minutes of the time letdown is //

isolated concurrent with HPSt, can prevent the PCS pressure from //

exceeding 10CFR50 Appendix G pressure limits. A dedicated //

operator is required under these conditions to ensure that //

mitigatory action is initiated within 2.9 minutes. //

3-33 Amendment No. lt, $1, 101, TSP1187-0218-NL04

3.3 ,EMERGENCY CORE COOLING SYSTEM Basis (continued)

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, (3) EA-PAL-LTOP-880121 "Calculation of Time for Operation to Act I for HPSI and Bubble", January 20, 1988. /

3-33a Amendment No. lt, Jt, t0t, TSP1187-0218-NL04

.,, -

• lf:t!- ..

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 both open PORV pilot valves are used as an alternative I to venting the PCS, then verify both PORV pilot valves and I both PORV block valves are open at least once per 7 days. I 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.

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 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 estimated average

. unsafe .failure rate of 1.14 x 10 -5 failure/hour P.er channel. This estimation is based on limited operating experience at conventional and nuclear plants. An "unsafe failure" is defined as one which negates channel operability and which, due to its nature, is revealed only when the channel is tested or attempts to respond to a bonafide signal.

4-2 Amendment No. 1$, $1, TSP1187-0218-NL04

4.6 SAFETY INJECTION AND CONTAINMENT SPRAY SYSTEMS TESTS 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 4.6.1 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 system. 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 I that all components have received the safety injection I signal in the proper sequence and timing (ie, the appropriate I pump breakers shall have opened and closed, and all valves I shall have completed their travel). I

b. Both high pressure safety injection pumps, P-66A and P-66B II shall be demonstrated 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 /> II whenever the temperature of one or more of the PCS cold legs II is < 300°F unless the reactor head is removed. II 4.6.2 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. 7i, ~~.

TSP1187-0218-NL04

4.6 SAFETY INJECTION AND CONTAINMENT SPRAY SYSTEMS TESTS (Contd) 4.6.3

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.

b. Acceptable levels of performance shall be that the pumps start, reach their rated shutoff heads at minimum recirculation flow, and operate for at least fifteen minutes.

4.6.4 Valves Deleted 4.6.S Containment Air Cooling System

a. Emergency mode automatic valve and fan operation will be checked for operability during each refueling shutdown.

b. Each fan and valve required to function during accident conditions will be exercised at intervals not to exceed three months.

Basis The safety injection system and the containment spray system are principal plant safety features that are normally inoperative during reactor operation.

Complete systems tests cannot be performed when the reactor is operating because a safety injection signal causes containment isolation and a containment spray system test requires the system to be temporarily disabled. The method of assuring operability of these systems is therefore to combine systems tests to be performed during annual plant shutdowns, with more frequent component tests, which can be performed during reactor operation.

The annual systems tests demonstrate proper automatic operation of the safety injection and containment spray systems. A test signal is applied to initiate automatic action and verification made that the components rece:i,ve the safety injection in the proper sequence.

The test demonstrates the operation of the valves, pump circuit breakers, and automatic circuitry. (1, 2) .

4-40 Amendment No 7Z, 77, TSP1187-0218-NL04

p ** )~ ;.,, . 1..

4.6

• SAFETY INJECTION AND CONTAINMENT SPRAY SYSTEMS TESTS Basis (continued)

(Contd)

I 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.

When the PCS cold leg temperature is less than 300°F, the start II of one HPSI pump could cause the Appendix G lir:iits to be II exceeded; therefore, both pumps are rendered inoperable. II References (1) FSAR, Section 6.1.3.

(2) FSAR, Section 6.2.3.

4-41 Amendment No.

TSP1187-0218-NL04