ML20136G262

Proposed Tech Spec Changes Providing for CRD Charging Water Header Low Pressure Scram W/Trip Setpoint of Greater than or Equal to 1,157 Psig.Significant Hazards Consideration Encl

ML20136G262
Person / Time
Site:	LaSalle
Issue date:	11/13/1985
From:	COMMONWEALTH EDISON CO.
To:
Shared Package
ML20136G253	List: ML20136G251 ML20136G262
References
NUDOCS 8511220295
Download: ML20136G262 (10)

Text

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ATTACHMENT B PROPOSED CHANGE TO APPENDIX A TECHNICAL SPECIFICATION TO OPERATING LICENSE NPF-ll REVISED PAGE: 2-4a B 2-13 (new page) 3/4 1-10 3/4 3-3 3/4 3-6 3/4 3-8 B 3/4 1-3 t

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. TABLE 2.2.1-1 (Continued) g REACTOR PROTECTION SYSTEM INSTRt3ENTATION SETPOINTS g . ALLOWBLE r FUNCTIONAL INIIT TRIP SETPOINT VALUES E

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9. Turbine Stop Valve - Closure i 5% closed i 7% closed f

E y 10. Turbfne Centrol Valve Fast Closure, p Trip 011 Pressure - Lou 1 500 psig 1 414 psig I *

11. Reacter flode Swttch Shutdoun Pos1t1on 10 4 -

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12. Itanuel Scram NA - IIA l

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LIMITING SAFETY SYSTIDI SETTING W O

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BASIS REACTOR PROTECTION SYSTEM INSTRUMENTATION SETPOINTS (CONTINUED)

13. Control Rod Drive (CRD) Chareine Water Header Pressure - tow The Hydraulic control Unit (HCU) scram accumulator is procharged with high pressure nitrogen (N 2). When the Control Rod Drive (CRD) pump is activated, the pressurized charging water forces the accumulator piston down to mechanical stops. The piston is maintained seated against this mechanical stop with normal charging water pressure. 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 setpoint 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 inadverjltant scram due to pressure fluctuations in the charging line. -

Four channels of pressure transmitter / trip unit combinations measure the charging water header pressure using one-out-of-two-twice logici The trip function is active in STARTUP and REFUEL modes because reactor pressure may be insufficient to assist the CRD scram action.

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REACTIVI W CONTROL SYSTEM -

, SURVEILLANCE REQUIREMENTS 4.1.3.5 Each control rod scras accumulator shall be determined OPERA 8LE:

a. At least once per 7 days by v'erifying that the indicated pressure is greatar than or equal to 940 psig unless the control rod is inserted and disarmed or scrammed.

b. At leait once per 18 months by:

4 1. Performance of a:

a) CHANNEL FUNCTIONAL TEST of the leak detectors, and

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• b) CHANNEL CALIBRATION of the pressure detectors, with the -

alarm setpoint 940 + 30 -0 psig on decreasing pressure.

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

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TABLE 3.3.1-1 (Centlausd) ,

w REACTOR P90TECT10N SYSTEN INSTR 19fMTAT1011 r"-

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" APPLICA8LE MINIlest OPERABLE OPERATIONAL CHANNELS PER FialCII(Want IRIIT CGISIT10115 TRIP SYSTEM (a) . ACTICII

7. Prienary Centalament Pressure e Migh 1, 2 III 2 I8I

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8. Scram Bischarge Volume Water level - Nigli 1 2 1 r 5(h), 2 3

9. Imtine Step Valve - Closure I III 4 III 6

10. Turbine Centrol Valve Fast Closure.

Valve Trip Systes 011 Pressure - Law I III 2 ISI 6

11. R...cter Mode Switch Shutdoun Position 1, 2 1 1 * -

y . 3, 4 1 7 w 5 1 3

12. ILmual Scram . 1, 2 -

1 1 3, 4 1 8 5 1 9 13 C. A .I R . J ' b e: < s. -

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TABLE 3.3.1-2 5

on REACTOR PROTECTION SYSTEM RESPONSE TIMES N

m i RESPONSE TIME e FUNCTIONAL UNIT '(Seconds) i'i

1. Intermediate Range Monitors:

a. Neutron Flux - High* HA

b. Inoperative NA

2. Average Power Range Monitor *

a. Neutron Flux - High, Setdown NA ,,

b. Flow Biased Simulated Thermal Power-Upscale < 0.09

c. Fixed Neutron Flux - High 30.09

d. Inoperative NA

3. Reactor Vessel Steam Dome Pressure - High < 0.55 w 4. Reactor Vessel Water Level - Low, Level 3 7 1.05 1 S. Main Steam Line Isolation Valve - Closure 7 0.06 w 6. Main Steam Line Radiation - High NA E 7. Primary Containment Pressure - High . NA

8. Scram Discharge Volume Water Level - High NA

9. Turbine Stop Valve - Closure -< 0.06

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

11. Reactor Mode Switch Shutdown Position HA

12. Manual Scram NA p 7

" Neutron detectors are exempt from response time testing. Response time shall be measured from the detector output or from the input of the first electronic component in the channel.

• • Not including simulated thermal power time constant.

1. Measured from start of turbine control valve fast closure.

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3 TABLE 4.3.1.1-1 (Continued)

E m REACTOR PROTECTION SYSTEH INSTRUMENTATION SURVEILLANCE REQUIREMENTS N

g CilANNEL OPERATIONAL

, CilANNEL FUNCTIONAL CilANNEL CONDITIONS FOR WillCil c FUNCTIONAL UNIT CllECK TEST CALIBRATION SURVEIll ANCE REQUIRED z

U 8. Scram Discharge Volume Water P . Level - High NA H R 1, 2, 5

9. Turbine Stop Valve - Closure NA H R 1

10. Turbine control Valve Fast Closure Valve Trip System Oil Pressure - Low NA H R* 1

11. Reactor Mode Switch Shutdown Position NA R NA 1,2,3,4,5 M

• 12. Manual Scram NA H NA 1,2,3,4,5

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(a) Neutron detectors may be excluded from CilANNEL CALIBRATION.

-(b) The IRH, and SRH channels shall be determined to overlap for at least 1/2 decades during each startup and the IRH and APRH channels shall be determined to overlap for at least 1/2 decades during each controlled shutdown, if not performed within the previous 7 days.

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

(d) This calibration shall consist of the adjustment of the APRH channel,to conform to the power values calculated by a heat balance duriug OPERATIONAL CONDITION 1 when TilERHAL POWER > 25% of RATED TilERHAL POWER. Adjust the APRM channel if the absolute difference is greater tiian 2%. Any APRM channel gain adjustment made in compliance with Specification 3.2.2 shall not be included in determining the absolute difference.

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

(f) The LPRHs shall be calibrated at least once per 1000 effective full power hours (EFPil) using the TIP k system.

g (g) Heasure and compare core flow to ra.ted core flow.

2 (h) This calibration shall consist of verifying the 6 & I second simulated thermal power time constant.

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• *The specified 18-month interval may b'e waived for Cycle I provided the surveillance is performed during Refuel 1.

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.' REACTIVITY CONTROL SYSTEMS BASES CONTROLR005(Continued) som+

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! W ,37 Control rod coupling integrity is required to ensure compliance with the

! P *J . analysis of the red drop accident in the FSAR. The overtravel position feature provides the only positive seans of determining that a rod is properly coupled .

and therefore this check sust be performed prior to achieving criticality after i

completing CORE ALTERATIONS that could have affected the control rod drive i

! coupling integrity. The subsequent check is performed as a backup to the initial l demonstration.

In order to ensure that the control rod patterns can be followed and there- .

i j fore that other parameters are within their limits, the control rod position '

I indication systas sust be OPERABLE.

The control rod housing support restricts the outward movement of a control i rod to less than 3 inches in the event of a housing failure. The amount of rod i i reactivity which could be added by this small amount of rod withdrawal is less l

! than a normal withdrawai increment and will not contribute to any damage to the <

j primary coolant systas. The support is not required when there is no pressure to act as a driving force to rapidly eject a drive housing. ,  ;

O The required surveillance intervals are adequata to datannine that the

'j rods are OPERA 8LE and not so frequent as to cause excessive wear on the systas components.

3/4.1.4 CONTROL R00 PROGRAM CONTROLS j

i Control rod withdrawal and insertion sequences are established to assure that the saximum insequence individual control rod or control rod segments which

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are withdrawn at any time during the fuel cycle could not be worth enough to .

1.- result in a peak fuel enthalpy greater than 280 cal /gs in the event of a control -

1 rod drop accident. The specified sequences are characterized by homogeneous,

! e scattered patterns of control rod withdrawal. When THERMAL POWER is greater F than 20% of RATED THERMAL POWER there is no possible rod worth which, if dropped l I at the design rata of the velocity limitar, could result in a peak enthalpy of  ;

280 cal /gs. Thus requiring the RSC5 and RW to be OPERA 8LE when THERMAL POWER j 1

is less than or equal to 20% of RATED THERMAL POWER provides adequata. control. i l l j The RSC5 and RWM provide automatic supervision to assure that out-of- l sequence rods will not be withdrawn or insertad. l j

l The analysis of the rod drop accident is presented in Section 15.4.9 of l l

the FSAR and the techniques of the analysis are presented in a topical report, j j Reference 1, and two supplements, References 2 and 3.

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J  ! The R8M is designed to automatically prevent fuel damage in the event of i

' erroneous rod withdrawal fres locations of high power density during high power

)! l operation. Two channels are provided. Tripping one of the channels will block

-) erroneous rod withdrawal soon enough to prevent fuel damage. This systas backs v

! up the written sequence used by the operator for withdrawal of control rods.  !

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8 3/4 I-3 LA SALLE - UNIT 1 '

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Insert to page B 3/4 1_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 adequate stored energy is available for normal scram action.

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ATTACHMENT C SIGNIFICANT HAZARDS CONSIDERATION Commonwealth Edison has evaluated the proposed Technical Specification Amendment and determined that it does not represent a significant hazards consideration. Based on the criteria for defining a significant hazards consideration established in 100rR50.92, operation of LaSalle County Station Unit 1 in accordance with the proposed amendment will not:

1) Involve a significant increase in the probability or consequences of an accident previously evaluated because this change to the Technical Specifications provides greater assurance that the scram function will mitigate the consequences of a postulated accioent. This CRD charging water header low pressure scram is discussed in the FSAR and in LaSalle County Station Safety Evaluation Report supplement 7. The setpoints are based on uncertainties allowed by the calibrated range of the pressure transmitter and trip units.

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 greater 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 with respect to the system or component; the consequences of previously evaluated accidents will not be increased and the margin of safety will not ce decreased. Therefore, based on the guidance provided in the Federal Register and the criteria established in 10CFR50.92(c), the proposed change does not constitute a significant hazards consideration.

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