Text

Proposed Tech Spec 4.7.3.c.2 & Bases 3/4.7.3,revising 18-month Surveillance Requirements to Account for Flow Differences Between Test Flow Path & Normal Flow Path
	ML20062K424
Person / Time
Site:	LaSalle
Issue date:	08/11/1982
From:	; COMMONWEALTH EDISON CO.
To:	;
Shared Package
ML20062K418	List: ML20062K416; ML20062K424;
References
	4730N, NUDOCS 8208170187
	Download: ML20062K424 (23)
	v • d • e

n ;

ATTACHMENT A L ASALLE COUNTY STATION UNIT 1 TECHNICAL SPECIFICATION CHANGE REQUEST NPF-4/82-ll

Subject:

Unit 1 RCIC Flow Test at Steam Pressure 150 + 15 psig.

References (a): G. I. Zwarich letter to R. H. Holyoak dated Augus t 6,1982.

(b): Sargent & Lundy Calculation No . RI-16, Rev.1, approved Augus t 6, 1982.

(c): Sargent & Lundy Calculation No. RI-17, Rev. O, approved Augus t 6, 1982.

BACKGROUND The proposed Technical Specification change has been requested as a result of the inability of the RCIC turbine to meet the literal words of Technical Specification 4.7.3.c.2. During initial Startup Test STP-14, it was determined that RCIC cannot develop enough power at 150 psig to push 600 gpm though the Test Flowpath due to the large pressure drop in the test line.

Discussion It is requested that Technical Specification 4.7.3.c.2 and Bases 3/4.7.3 be changed as indicated on the enclosed marked up pages. These changes provide that the turbine's capability be demonstrated by pumping through the test flow path at a different flow and pressure. The test flow path data will be demonstrated to be equivalent to or greater than that which is required to provide 600 gpm to the reactor vessel after correcting for line losses.

The current 1E51-F022 globe valve manufactured by Anderson Greenwood was initially assumed to have approximately 80 ft. of pressure drop in 1975. However, manufacturer's data supplied af ter 1975 indicated the valve has approximately 115 pounds of pressure drop. In addition to this dif ference in pressure drop, the RCIC test line is being repaired at this time'due to underground piping problems. The new line will be longer and has additional 90 degree bends that will increase the present test line pressure drops. Therefore, it is necessary that the Technical Specification change be made.

,1 P208170187 820811 I PDR P

ADOCK 05000373 !

PDR --

Initial Startup Test data in Table 1 and plotted on Figure 1 indicates the current test line permits on 590 gpm flow at a discharge pressure o f 430 pounds. Line losses, as calculated in References (a),

(b), and (c) indicate that a pressure of 258 psig or greater at the Test Line will ensure a flow o f 600 gpm into the reactor vessel at a reactor pressure o f 165 psig. The equivalent pressure at the "T" at 590 gpm is 416 psig. The re fo re , from Table 1 and Figure 1 it can be demonstrated that approximately 150 psig excess (416-258) is available for injection to the vessel. Actual Startup Test data will be taken to determine the vessel injection line loss and flow capacity.

Tables 2 and 3 are provided for additional information. Table 3 indicates RCIC has no problem in the test flow path at 940 psig reactor pressure.

Conclusion Commonwealth Edison Company finds no unreviewed safety questions in this Technical Specification change.

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PLANT SYSTEMS .

3 4_.7.3 REACTOR CORE ISOLATION COOLING SYSTEM U

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LIMITINGCONDITIONFOROPERATION

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3.7.3 Th s reactor core isolation cooling (RCIC) system shall be OPERABLE with an OPERABLE \ flow path capable of taking suction from the suppression pool and transferring the water to the reactor pressure vessel.

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APPLICABILITY: 'OPERATIONALCONgITIONS1,2,and3withreactorsteamdome pressure greater than 150 psig ACTION:

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a. With a RCIC discharge line " keep filled" pressure alarm instrumenta-tion channnel ' inoperable, perform Surveillance Requirement 4.7.3.a.1 at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

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b. With the RCIC system inoperable, operation may continue provided the HPCS system is OPERABLE; restore the RCIC system to OPERABLE status within 14 days or be iq at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and reduce reactor steam dome pressure to less than or equal to 150 psig within the folldwing 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

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SURVEILLANCE REQUIREMENTS

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C 4.7.3 The RCIC system shall be demonstrated OPERABLE:

a. At least once per 31 days by:

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1. Verifying by venting at the highNpoint vents that the system piping from the pump discharge valve to the system isolation valve is filled with water, \

2. PerformanceofaCHANNELFUNCTIONALYESTofthedischargeline "keepfilled"pressurealarminstrumentagion,and

3. Verifying that each valve, manual, power' operated or automatic in the flow path that is not locked, sealedsor otherwise secured in position, is in its correct position. \.\

4. Verifying that the pump flow controller is in the correct position. \

b. At least once per 92 days by verifying that the RCIC pum\p gdevelops a flow of greater than or equal to 600 gpm in the test flow p,ath with a system head corresponding to reactor vessel operatir.g pressure when steam is being supplied to the turbine et 1000 + 20, - 80 psi'g.*

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.* The provisions of Specification 4.0.4 are not applicable provided the surveillance is performed within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after reactor steam pressure is N

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n* adequate to perform the tests.

( #See Special Test Exception 3.10.7. N LA SALLE - UNIT 1 3/4 7-7

PLANT SYSTEMS SURVEILLANCE REQUIREMENTS

c. At least once per 18 months by: ,

2 1. Performing a system functional test which includes simulated

@ g automatic actuation and verifying that each automatic valve in

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o the flow path actuates to its correct position, but may exclude w o actual injection of coolant into2 the reactor _Nesse.lm o fs2. ~ ^ ~ ~

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AO pp 7e~ElfyIIIg thaVthe sPHe~m3ill develop a flow of greate70iiiN or equa(to'600 gpmKthe test flow path when 7 steam is c upplied

u. *o n Q { 3. sto-the_turbipeatia_preisutg_oful50~+15pygfi q}

L Performing a CHANNEL CALIBRATION of the discharge line " keep g filled" pressure alarm instrumentation and verifying the low '

a ,I d.j pressure setpoint to be > 62 psig.

y d. E By demonstrating MCC-121y and the 250-volt battery ** and charger **

1 o w OPERABLE:

ei 3 e bb V a t c 1.

At least once per 7 days by verifying that:

n MU a) MCC-121y is energized, and has correct breaker alignment, x9M indicated power availability from the charger and battery, 2 3 2R

{$g@g

,a

b,)

and voltage en the panel with an overall voltage of greater than or equal to 250 volts.

The electrolyte level of each pilot cell is above the plates, W

'O fg;22 c) The pilot cell specific gravity, corrected to 77 F, is v> hr greater than or equal to 1.200, and a d[ 5 Wy . d) The overall battery voltage is greater than or equal to 250 volts.

3 gdf > $ j 2. At least once per 92 days by verifying that:

$d h p O- a) The voltage of each connected battery is greater than or P c"t o equal to 250 volts under float charge and has not decreased g o 3 more than 12 volts from the value observed during the 2

> TO.d q original test, b) The specific gravity, corrected to 77 F, of each connected

{ ]3 tg3@

-t cell is greater than or equal to 1.195 and has not decreased more than 0.05 from the value observed during the previous

% 1. test, and d 4. [ f 'j c) The electrolyte level of each connected cell is above the plates.

3. At least once per 18 months by verifying that:

a) The battery shows no visual indication of physical damage or abnormal deterioration, ana b) Battery terminal connections are clean, tight, free of corrosion and coated with anti-corrosion material.

The provisions of Specification 4.0.4 are not applicably provided the surveillance is performed within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after reactor steam pressure is adequate to perform the tests.

- The Unit 2 RCIC system battery and charger raay be in service to supply Unit 1 RCIC system provided that they are demonstrated OPERABLE per Unit 1 Technical Specifications. This footnote shall be deleted upon issuance of an Operating License for Unit 2.

LA SALLE - UNIT 1 3/4 7-8

3/4.7 PLANT SYSTEMS

, U BASES 3/4.7.1 CORE STANDBY COOLING SYSTEM - EQUIPMENT COOLING WATER SYSTEMS The OPERABILITY of the core standby cooling system - equipment cooling water systems and the ultimate heat sink ensure that sufficient cooling capacity is available for continued operation of safety-related equipment during normal and accident conditions. The redundant cooling capacity of these systems, assuming a single failure, is consistent with the assumptions used in the accident conditions within acceptable limits.

3/4.7.2 CONTROL ROOM AND AUXILIARY ELECTRIC EQUIPMENT ROOM EMERGENCY FILTRATION SYSTEM The OPERABILITY of the control room and auxiliary electric equipment room emergency filtration system ensures that the rooms will remain habitable for operations personnel during and following all design basis accident conditions.

The OPERABILITY of this system in conjunction with room design provisions is based on limiting the radiation exposure to personnel occupying the rooms to 5 rem or less whole body, or its equivalent. This limitation is consistent with the requirements of General Design Criteria 19 of Appendix "A", 10 CFR Part 50. Cumulative operation of the system with the heaters OPERABLE for 10 hours1.157407e-4 days 0.00278 hours 1.653439e-5 weeks 3.805e-6 months over a 31 day period is sufficient to reduce the buildup of moisture on the adsorbers and HEPA filters.

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3/4.7.3 REACTOR CORE ISOLATION C0OLING SYSTEM The reactor core isolation cooling (RCIC) system is provided to assure adequate core cooling in the event of reactor isolation from its primary heat sink and the loss of feedwater flow to the reactor vessel without requiring l actuation of any of the Emergency Core Cooling System equipment. The RCIC l system is conservatively required to be OPERABLE whenever reactor pressure exceeds 150 psig even though the LPCI mode of the the residual heat removal (RHR) system provides adequate core cooling up to 350 psig.

The RCIC system specifications are applicable during OPERATIONAL CONDITIONS 1, 2 and 3 when reactor vessel pressure exceeds 150 psig because RCIC is the primary non-ECCS source of core cooling when the reactor is pressurized.

With the RCIC system inoperable, adequate core cooling is assured by the OPERABILITY of the HPCS system and justifies the specified 14 day out-of-service period.

The surveillance requirements provide adequate assurance that RCICS will be OPERABLE when required. Although all active components are testable and full flow can be demonstrated by recirculation during reactor operation, a complete functional test requires reactor shutdown The pump discharge piping is maintained full to prevent water hammer damage and to start cooling at the earliest possible moment.

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LA SALLE - UNIT 1 B 3/4 7-1 MW/%s, c. M 9 % 4< L e m n W [*P8% owl % +occl., &

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PLANT SYSTEMS B SES 3/4.7.4 SEALED SOURCE CONTAMINATION Thelhmitationsonremovablecontaminationforsourcesrequiringleak j testing, including alpha emmitters, is based on 10 CFR 70.39(c) limits for plutonium. This limitation will ensure that leakage from byproduct, source, and special nuclear material sources will not exceed allowable intake values.

Sealed sources are classified into three groups according to their use, with surveillance requirements commensurate with the probability of damage to a source in that group. . Those sources which are frequently handled are required to be tested more often than those which are not. Sealed sources which are continuously enclosed within a shielded mechanism, i.e., sealed sources within I radiation monitoring or bqron measuring devices, are considered to be stored and need not be tested unless they are removed from the shielded mechanism.

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3/4.7.5 FIRE SUPPRESSION SYS'TEMS The OPERABILITY of the fire suppression systems ensures that adequate fire suppression capability is available to confine and extinguish fires occurring in any portion of the fac1}ity where safety related equipment is located. The fire suppression systemxconsists of the water system, deluge and/or sprinklers, C07 systems, and fire hose stations. The collective capability of the fire suppression syste'ms is adequate to minimize potential damage to safety related equipment and is'a major element in the facility fire protection program. '

In the event that portions of the fire uppression systems are inoperable, alternate backup fire fighting equipment is required to be made available in the affected areas until the inoperable equipmentxis restored to service.

When the inoperabic fire fighting equipment is intended for use as a backup means of fire suppression, a longer period of time i's allowed to provide an alternate means of fire fighting than if the inoperable equipment is the primary means of fire suppression. '

The surveillance requirements provide assurance that the minimum OPERABILITY requirements of the fire suppression systems are met.

In the event the fire suppression water system becomes inoperable, immediate corrective measures must be taken since this system provides the major fire suppression capability of the plant. The requirement for a twenty-four hour report to the Commission provides for prompt evalu'ation of the acceptability of the corrective measures to provide adequate fire suppression capability for the continued protection of the nuclear plant.

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LA SALLE - UNIT 1 8 3/4 7-2 h

TABLE 1 L ASALLE COUNTY STATION: UNIT 1 I E 51- F02-2, full flow test line globe valve, was throttled to record the following pressures, speeds and flows. Flow controller set at 600 gpm. Reactor pressure 163 psig.

Flow Discharge Pressure Suction Pressure Speed apm psia psia rpm 590 430 21 2850 545 470 21 2900 500 500 21 3000 445 560 22 3100 395 610 22 3200

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ATTACHMENT B Status of Tech Spec Change Reauests NRC
Topic Submitted Ac tio n NPF-ll/82-7 SRM Countrate 6/14/82 Issued Am. 2 )
(Required prior to source 7/09/82 i decay below 3 cps).
NPF-11/82-8 Revise RCIC suction delta ~p 7/02/82 Issued Am. 2 alarm setpoint (Required 7/09/82 prior to pressurization).
NPF-ll/82-9 Add commitment to complete 7 /14/8 2 Issued Am . 3 I torque checks on bolting on 7/15/82 l S/R valves outside containment l (NRC Requested to be submitted 1 ASAP).
NPF-ll/82-10 Add ECCS delta p specification ----
to reflect modification (License Condition 2.C.C)
NPF-ll/82-11 Revise RCIC surveillance to 8/11/82 account for flow differences between test flow and normal flow paths (Required prior to restart af ter next shutdown) .
4730N

ML20062K424

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