Text

Proposed Tech Specs Adding Reactor Scram on Low CRD Pump Discharge Pressure
	ML20091J382
Person / Time
Site:	LaSalle
Issue date:	05/24/1984
From:	; COMMONWEALTH EDISON CO.
To:	;
Shared Package
ML20091J371	List: ML20091J366; ML20091J382;
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	NUDOCS 8406050492
	Download: ML20091J382 (13)
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ATTACHMENT 2 Proposed Change to Appendix A Technical Specifications to Operating License NPF-18 Revised pages.: 2-4 8 2-13 3/4 1-10 3/4 3-3 3/4 3-5 3/4 3-6 3/4 3-8 8 3/4 1-3 8671N 8406050492 840524 PDR ADOCK 05000374 P PDR 1

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i TADu 4.2.1-1 REACTOR PR01ECT10N SYSTEM INSTRtNIENTATION SETrolNTS 5 ft#fCTIONAL UNIT i TRIP SETP0lNT ALLOWADLE VALUES p

1. Intermediate Range Monitor, Neutron Flux-Ilfgh 4 120 divisions of i 122 divisions

} full scale of full scale

' 2. Average Power Range Monitor: g Z a. Neutron Flux-High, Setdown 5 15% of RATED $ 20% of RATED u t, TilERHAL POWER THERMAL POWER

h. "aw Blased Simulated Tiiermal Power - tipscale .

1) Two Recirculation loop Operation a) Flow filascel Ii ~< 0.66W + 51% with a ~< 0.66W + 54% with a maximum of maximum of b) liigh riow Clamped i 113.5% of RATED $ 115.5% of RATED THERMAL POWER IHERMAL POWER ,

2) Single Recirculation loop Operation .

a) Flow Diased ' < 0.66W + 45.7% with -< 0.66W + 40.7% with D

- a maxista of a maximum of M, b) liigh Flow Clamped i 113.5% of RATED $ 115.5% cf RATED T . THERMAL POWER

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s THERMAL POWER z

c. Fixed Neutron Flux-liigh ~< 118% of RATED < 120% of RATED

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I j TilERMAL POWER THERMAL POWER

3. Reactor Vessel Steam Dome Pressure - High

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5 1043 psig $ 1963 psig -1

4. Reactor Vessel Water Levei - Low, level 3 1 12.5 inches above instrument zero*

1 11 inches above Instrument zero*

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5. Main Steam Line Isolation Valve - Closure 3 8% closed 3 12% closed

6. Main Steam Line Radiation - High 5 3 x full 1 3.6 x full l power background power background

7. Primary Containment Pressure - liigh 5 1.69 psig 5 1.89 psig

8. Scram Discharge Volume Water Level - liigh 5 767' 5%" < 767' 5%"

9. Turbine Stop Valve - Closure 5 5% closed S 7% closed

10. Turbine control Valve Fast Closure, Trip 011 Pressure - Low , > 500 psig 1 414 psig

11. Reactor Mode Switch Shutdown Position N.A. N. A.

12. Manual Scran 1 N.A. N.A.

See Bases Figure R 3/4 3-1.

7:3. ce,drol 12e4 trive. 2 IlQ

.. char w wa4ec beehu pesore-low > gg7f5;

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JPEAcrot PRO 76 C 7/ CW/ 3YSTEM /N57tdMEN 7A TieAf .5E7PQ/NTS ouAHnd) ll .

LIMITING SAFETY SYSTEM SETTING B AS r '

13. Control Rod Drive (CRD) Charging Water Header Pressure - Low The Hydraulic Control Unit (HCU) scram accumulator is precharged with high pressure nitrogen (N 2). 'When the Control Rod Drive (CRD) pump is activated, the.pressuri-zed charging water forces the accumulator piston down to mechanical stops. The piston is maintained seated against this mechanical stop with normal charging water pressure, typically above 1400 psig. If the charging water header pressure decreases below the N2 pressure, such as would be the case with high leakage through the .

check valves of the CRD charging water lines, the accumulator piston would eventually rise off its stops.

This results in a reduction of the accumulator energy and thereby degrades normal scram performance of the CRD's in the absence of sufficient reactor pressure.

The CRD low charging water header pressure trip set-point initiates a scram at the charging water header pressure which assures the seating of the accumulator piston. With this trip setpoint, full accumulator capability, and therefore, normal scram performance, is assured at all reactor pressures. An adjustable time-delay relay is provided for each pressure transmitter /

trip channel to protect against inadvertant scram due to .

pressure fluctuations in the charging line.

Four. channels of pressure transmitter / pressure indi-r cating switch combinations measure the charging water

header pressure using~one-out-of-two-twice logic. The l trip function is automatically bypassed in RUN mode r

because reactor pressure is available there to assist l_ the CRD scram action. A keylock switch bypass is avail-1

_ _ able in the SHUTDOWN and REFUEL modes to allow the scram _ __ _

reset of:the RPS and to establish nominal /CRD valve l- i l line up.

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REACTIVITY CONTROL SYSTEM

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SURVEILLANCE REQUIREMENTS' 4.1.3.5 Each control rod scram accumulator shall be determined OPER' BLE: A

a. At least once per 7 days by verifying that the indicated pressure is -

greater than or equal to 940 psig unless the control rod.is inserted and disamed or scrammed.

b. At least once per 18 months by:

1. Performance of a:

a) CHANNEL FUNCTIONAL TEST of the leak detectors, and b) CHANNEL CALIBRATION of the pressure detectors, with the alam setpoint 940 + 30, -0 psig on decreasing pressure.

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2. Measuring and recording the time that each individual accumulator check valve maintains the associated accumulator pressure above the alam setpoint with no control rod drive pump operating.

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LA SALLE - UNIT 2 3/4 1-10

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TABLE 3.3.1-1 (Continued)

REACTOR PROTECTION SYSTEM INSTRUMENTATION- l.

W ;

E APPLICABLE MININim OPERA 8LE OPERATIONAL CHANNELS PER CONDITIONS- TRIP SYSTEM (a) ACTION FUNCTIONAL UNIT r 1, 2 III Z I8I 1

7. Primary Containment Pressure q High N !

8. Scram DI.scharge Volume Water'i Level - High 1 2 1 5(hj, 2 3 Turbine Stop Valve - Closure 1(I) 4 IN 6 9.

10. Turbine Control Valve Fast Closure. UI 6 Valve Trip System Oil Pressure - Low I III 2 w 11. Reactor Mode Switch Shutdown I D Position 1, 2 1 ,

7 3, 4 1 w 3 1

0 5

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12. Manual Scram 3, 4 1 8 ,

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13. dou tral R d Drive, .

4. Ch e r .* Wahr Nradec N ssor - Lois 2N 2. I i 5 04 I 3 L. %tay Thec 2 '"I 2 1

, 5 CM 2 3 l

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NO CHAMGES T9'0R R.cFERE A/CE CA/LY

- TABLE'3.3.1-1 (Continued) .

REACTOR PROTECTION SYSTEM INSTRUMENTATION ACTION STATEMENTS .

ACTION 1 -

Be in at least HOT SHUTDOWN within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

ACTION 2 -

Verify all insertable control rods to be inserted in the core and lock the reactor mode switch in the Shutdown position within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

( ACTION 3 -

Suspend all operations involving CORE ALTERATIONS

and insert all insertable control rods within one hour.

ACTION 4 -

Be in at least STARTUP within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

ACTION 5 -

Be in STARTUP with the main steam line isolation valves closed within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months or in at least HOT SHUTDOWN within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

ACTION 6- -

' Initiate a reduction in THERMAL POWER within 15 minutes and reduce turbine first stage pressure to < 140 psig, equivalent to THERMAL POWER less than 30% of RATED ~ THERMAL POWER, within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months .

ACTION 7 -

Verify all insertable control rods to be inserted within I hour.

ACTION 8 -

Lock the reactor mode switch in the Shutdown position within I hour.

l ACTION 9 -

Suspend all operations involving CORE ALTERATIONS," and insert -

all insertable control rods and lock the reactor mode switch in l the SHUTDOWN position within I hour.

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"Except movement of IRM, SRM, or special movable detectors, or replacement of LPRM strings provided SRM instrumentation is ~0PERABLE per Specification 3.9.2.

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TABLE 3.3.1-1 (Continued)

REACTOR PROTECTION SYSTEM. INSTRUMENTATION TABLE NOTATIONS (a) .A channel may be placed in an inoperable status for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months for required surveillance without placing the trip system in the tripped condition provided at least one OPERABLE channel in the'same trip system is monitoring that parameter.

(b) The " shorting links" shall be removed from the RPS circuitry prior to and during the time any control rod is withdrawn

and during shutdown margin demonstrations performed per Specification 3.10.3.

(c) An APRM channel is inoperable if there are less than 2 LPRM inputs per level or less than 14 LPRK inputs to an APRM channel.

(d) This function is not required to be OPERABLE when the reactor pressure vessel head is unbolted or removed per Specification 3.10.1.

(a) This . function shall be automatically bypassed when the reactor mode switch is not in the Run position.

(f) This function is not required to be OPERABLE when PRIMARY CONTAINMENT INTEGRITY is not required.

(g) Also actuates the standby gas treatment system.

(h) With any control rod withdrawn. Not applicable to control rods removed -

per Specification 3.9.10.1 or 3.9.10.2.

(i) This function shall be automatically bypassed when turbine first stage c pressure is < 140 psig, equivalent to THERMAL POWER less than 30% of l:.

,p

_. RATED THERMAE POWER. - -- - - - - - - - ~

(j) Also actuates the EOC-RPT systas.

(.D We*N reoclec pressore 4 95"O fst .

"Not required for control rods removed per Specification 3.9.10.1 or .

3.9.10.2.

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TA8LE 3.3.1-2 ,

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, - REACTOR PROTECTION SYSTEM RESPONSE TIES

! O t E

l RESPONSE TIME FUNCTIONAL UNIT (Seconds)

[

z U 1. Intermediate Range Monitors: .

M a. Neutron Flux - High" '

NA

b. Inoperative l. , NA I 2. Average Power Range Monitor *

a. Neutron Flux - High, Setdown NA ..

b. Flow Blased Simulated Thermal Power-Upscale 5 0.09

c. Fixed Neutron Flux - High < 0.09

d. Inoperative >

NA ,

3. Reactor Vessel Steam Dome Pressure - High 5 0.55 y 4. Reactor Vessel Water Level - Low, tevel 3 $ 1.05

, *- 5. Main Steam Line Isolation Valve - Closure < 0.06 j ! y 6. Main Steam Line Radiation - High NA 1 m 7. Primary Containment Pressure - High NA j 8. Scram Discharge Volume Water Level - High NA ,

9. Turbine Stop Valve - Closure 5 0.06 *

10. Turbine Control Valve Fast Closure, Trip Oil Pressure - Low < 0.08,

11. Reactor Mode Switch Shutdown Position NA l NA l

12. Manual Scram l

Neutron detectors are exeept from response time testl'ng. Response time sha11 be measured i from the detector output or from the input of the first electronic component in the i channel.

l3 **Not including simulated thermal power time constant.

! # Measured from start of turbine control valve fast closure.

-l3. Cou4 col I2od "Delue.

g. Char ; Walee NeedovPressore -l.ew g

b. Alny %w #A O

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TABLE 4.3.1.1-1 ,

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REACTOR PROTECTION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS 5 OPERATIONAL CHANNEL k .

CilANNEL FUNCTIONAL CllANNEL CONDITIONS FOR WHICH I *I SURVEILLANCE REQUIRED FUNCTIONAL UNIT CHECK TEST CALIBRATION

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! U 1. Intermediate Range Monitors 4 to a. Neutron Flux - High S/U'IbI,5 S/UICI,_W R 2 I :$ W R 3,4,5 t

b. Inoperative lNA W NA 2,3,4,5 l .

2. Average Power Range Monitor: III i a. Neutron Flux - High.

j Setdown S/U(b) 5 S/U(c),W.

SA 1, 2 S W SA 3, 5

b. Flow Blised Simulated Thermal I8I ,

! w Power-Upscale S. D S/U IC} ,W WI }(*} ,SA,R(h) j D c. Fixed Neutron Flux -

W Id) , SA w High jS S/U IC) ,W 1

-s d. Inoperative W NA 1. ,2, 3, 5

, l NA

3. Reactor Vessel Steam Dome {

Pressure - High NA M q 1, 2 i

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4. . Reactor Vessel Water Level -

tow, Level 3 5 M R 1, 2 g N O i l

] 5. Main Steam Line Isolation ~

M R 1 m I Valve - Closure < NA m #*O l "#

l* 6. Main Steam Line Radiation -

Qh High iS M R I, 2 i

ni m t M

7. Primary Containment Pressure -

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, High i NA M q 1, 2 g i i

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E . REACTOR PROTECTION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS l ,

CHANNEL OPERATIONAL h

CHANNEL FUNCTIONAL CHANNEL CONDITIONS FOR WHICH .

FUNCTIONAL UNIT i CHECK TCST CALIBRATION SURVEILLANCE REQUIRED

-[z U 8. ScramDischargeVolumeWaterl N Level - High NA M R 1,2,5 .

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9. Turbine Stop Valve - Closure ! NA M . R 1 1

10. Turbine control Valve Fast Closure Valve Trip System 011 Pressure - Low NA M R 1

11. Reactor Mode Switch -

Shutdown Position MA 'R NA 1, 2, 3,'4, 5 w .

A '12. Manual Scram NA M MA 1,2,3,4,5

l f (a) Neutron detectors may be excluded from CHANNEL CAllBRATION.

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(b) The IRM, and SRM channels shall be determined to overlap for at least 1/2 decades during each startup and the IRM and APRM channels shall be determined to overlap for at least 1/2 decades during each

! controlled shutdown, if not performed within the previous 7 days.

3 (c) Within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months prior to startup, if not performed within the previous 7 days.

! (d) This calibration sha11 consist of the adjustment of the APRM channel to conform to the power values

calculated by a heat balance during OPERATIONAL CONDITION I when 1HERMAL POWER.> 25% of RATED I IllERMAL POWER. Adjust the APRM channel if the absolute difference is greater LEen 2%. Any APRM i channel gain adjustment made in compilance with Specification 3.2.2 shall not be included in determining the absolute difference. .

(e) This calibration shall consist of the adjustment of the APRM flow biased channel to conform to a calibrated flow signal.

N. (f) The LPRMs shall be calibrated at least once per 1000 effective full power hours (EFPH) using the TIP system.

g (g) Measure and compare core flow to rated core flow.

N (h) This calibration shall consist of Jerifying the 611 second simulated thermal ' power time constant.

13. Coo-61 Rod 3rivo

a. CLag;ag wa4w Heoc(so bsvre. AIA M f Y '

l>. My Thw ,

NA M. R 2g

REACTIVITY CONTROL SYSTEMS BASES CONTROL R005 (Continued) sMsuk --~w Control rod coupling integrity is required to ensure compliance with the Ollow.di analysis of the rod drop accident in the FSAR. The overtravel position feature provides the only positive means of determining that a rod is properly coupled FAG and therefore this check must be performed prior to achieving criticality after conoleting CORE ALTERATIONS that could have affected the control rod drive coupling integrity. The subsequent check is perfomed as a backup to the' initial demonstration.

In order to ensure that the control rod patterns can be followed and there-fore that other parameters are within their limits, the control rod position indication system must be OPERABLE.

The control rod housing support restricts the outward movement of a control rod to less than 3 inches in the event of a housing failure. The amount of red reactivity which could be added by this small amount of rod withdrawal is less than a normal withdrawal increment and will not contribute to any damage to the primary coolant system. The support is not required when there is no pressure to act as a driving force to rapidly eject a drive housing.

The required surveillance intervals are adequate to determine that the i rods are OPERABLE and not so frequent as to caus. excessive wear on the system components.-

l l 3/4.1.4 CONTROL R00 PROGRAM CONTROLS

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Control rod withdrawal and insertion sequences are established to assure that the maximum insequence. individual control rod or control rod segments which are withdrawn at any time during the fuel cycle could not be worth enough to result in a peak fuel enthalpy greater than 280 cal /ga in the event of a control _

' rod drop accident. The specified sequences are characterized by homogeneous, scattered patterns of control rod withdrawal. When THERMAL POWER is greater than 20% of RATED THERMAL POWER, there is no possible rod worth which, if dropped at the design rate of the velocity limiter, could result in a peak enthalpy of 280 cal /gn. Thus requiring the RSCS and RWM to be OPERABLE when THERMAL POWER

-is less than or equal 9 20% of RATED THERMAL POWER provides adequate control.

The RSCS and RWM provide automatic supervision to assure that out-of-l sequence rods will not be withdrawn or inserted.

The analysis of the rod drop accident is presented in Section 15.4.9 of the FSAR and the techniques of the analysis are presented in a topical report, Reference 1, and two supplements, References 2 and 3.

The RBM is designed to automatically pre. vent fuel damage in the event of erroneous rod withdrawal from locations of high power density during high power operation. Two channels are provided. Tripping one of the channels will block erroneous rod withdrawal soon enough to prevent fuel damage. This system backs up the written sequence used by the operator for withdrawal of control rods.

LA SALLE - UNIT 2 B 3/4 1-3 l.

Insert $ pc ga 3 3/4. L - 3 .

In addition, the automatic CRD charging water header low pressure scram (see Table 2.2.1-1) initiates well before

- any accumulator loses its full capability to insert the control rod. With this added automatic scram feature, the surveillance of each individual accumulator check valve is no longer necessary to demonstrate adecuate stored energy is available for normal scram action. .

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f bd ATTACHMENT 3 Technical. Specification Change Request LaSalle County. Station Unit 2

. Commonwealth Edison has evaluatad the proposed Technical Specification Amendment and'deterined that it does not represent a significant hazards consideration. Based on the criteria for defining a significant hazards consideration established in 10 CFR 50.92, operation of LaSalle County Station Unit 2 in accordance with the proposed

amendment will not

.1)- Involve a significant increase in the probability of consequences of an accident previously evaluated because~this change to the Technical LSpecification;provides greater assurance that the scram function will mitigate the . consequences of a postulated accident. This CRD charging water _ header. low pressure scram is discussed in the FSAR and the designLwas approved in LaSalle County Station Safety Evaluation

. Report Supplement 7.

2). Create the possibility of a new or'different kind of accident from any previously evaluated because this change does not eliminate any previously required scram function but adds an additional one to better ensure _ automatic control rod insertion capability under all plant operating conditions.

3) Involve a significant reduction in the margin of safety because this change maintains or increases the likelihood that proper control rod scram capability will be available during all plant conditions.

~ Based on the preceding discussion, it is concluded that the proposed system change clearly falls within all acceptable criteria._ The consequences of previously evaluated accidents will not be increased and the margin of safety will not be decreased. Therefore, based on the

-guidance provided-in the Federal Register and the criteria established in 10 CFR 50.92(e), the proposed change does not constitute a significant hazards consideration.

8671N-

ML20091J382

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