ML20101K784
| ML20101K784 | |
| Person / Time | |
|---|---|
| Site: | LaSalle  |
| Issue date: | 12/17/1984 |
| From: | Schwencer A Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20101K787 | List: |
| References | |
| NUDOCS 8501020003 | |
| Download: ML20101K784 (10) | |
Text
[T pn nea ft uq[o, UNITED STATES i
NUCLEAR REGULATORY COMMISSION
'D E
WASHING TON, D. C. 20555
(...../
DEC'17 1984' COMMONWEALTH EDIS0N COMPANY DOCKET NO. 50-374 LA SALLE COUNTY STATION, UNIT 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 6 i
License No. NPF-18 1.
The Nuclear Regulatory Comission (the Comission or the NRC) having found that:
A.
The application for amendment filed by the Comonwealth Edison Company, dated September 25, 1984, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Comission's regulations set forth in 10 CFR Chapter I; B.
The facility will operate in conformity with the application, the-provisions of the Act, and the regulations of the Comission; C.
There is a reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, c,d (ii) that such activities will be conducted in compliance with the Comission's regulations set forth in 10 CFR Chapter I*
D.
The issuance of this amendment will not be inimical to the comon defense and security or to the health and safety of the public; and E.
The issuance of this amendment is in accordance with 10 CFR Part 51 of the Comission's regulations and all applicable requirements have been satisfied.
2.
Accordingly, the license is amended as follows:
A.
Page changes to the Technical Specifications as indicated in the enclosure to this license amendment and paragraph 2.C.(2) of the Facility Operating License No. NPF-18 is hereby amended to read as follows:
The Technical Specifications contained in Appendix 'A, as revised through Amendment No. 6, and the Environmental Protection Plan contained in Appendix B, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.
B.
Paragraph 2.C.(7) is satisfied.
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This amendment'is effective as of date of issuance.
7 FOR.THE NUCLEAR REGULATORY COMMISSION A'. Schwencer, Chief.
. Licensing Branch No. 2 Division of Licensing
Enclosure:
. Changes to'the Technical Specifications 2
.Date of Issuance: December 17, 1984
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ENCLOSURE TO LICENSE AMENDMENT NO. 6 FACILITY OPERATING LICENSE NO. NPF-18 DOCKET NO. 50-374 Replace the following pages of the Appendix "A" Technical-Specifications with the enclosed pages. The revised pages are identified by Amendment number and contain a vertical line indicating the area of change.
REMOVE
- INSERT 2-4 2-4 B 2-13 3/4 1-10 3/4 1-10 3/4 3-3 3/4 3-3 3/4 3-6 3/4 3-6 3/4 3-8 3/4 3-8 8 3/4 1-3 8 3/4 1-3 i
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REACTIVITY CONTROL SYSTEMS BASES CONTROL RODS (Continued)
In addition, the automatic CRD charging water header low pressure scram i
(see Table 2.2.1-1) initiates well before any accumulator loses its full capa-bility to insert the control rod.
With the 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.
Control rod coupling integrity is required to ensure compliance with the 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 and therefore this check must be performed prior to achieving criticality af ter completing CORE ALTERATIONS that could have affected the control rod drive coupling integrity.
The subsequent check is performed 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 rod 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 rods are OPERABLE and not so frequent as to cause excessive wear on the system components.
3/4.1.4 CONTROL R0D PROGRAM CONTROLS 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 /gm 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 /gm.
Thus requiring the RSCS and RWM to be OPERABLE when THERMAL POWER is less than or equal to 20% of RATED THERMAL POWER provides adequate control.
The RSCS and RWM provide automatic supervision to assure that out-of-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 prevent fuel damage in the event of l
erroneous rod withdrawal from locations of high power density during high power operation. Two channels are provided.
Tripping one of the channels will block i
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 Amendment No. 6
TABLE 4.3.1.1-1 (Centinutd)
REACTOR PROTECTION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS 7
TABLE 4.3.1.1-1 (Continued) s REACTOR PROTECTION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS e
CHANNEL OPERATIONAL-ll CHANNEL FUNCTIONAL CHANNEL CONDITIONS FOR WHICH mo FUNCTIONAL UNIT CHECK TEST CALIBRATION SURVEILLANCE REQUIRED e
8.
Scram Discharge Volume Water Level - High NA M
R 1, 2, 5 9.
Turbine Stop Valve - Closure NA M
R 1
10.
Turbine Control Valve Fast Closure Valve Trip System Oil Pressure - Low NA M
R 1
so 11.
Reactor Mode Switch 23 Shutdown Position NA R
NA 1,2,3,4,5 ws 12.
Manual Scram NA M
NA 1,2,3,4,5 13.
Control Rod Drive a.
Charging Water Header Pressure - Low NA M
R 2, 5 b.
Delay Timer NA M
R 2, 5
~
(a) Neutron detectors may be excluded from CHANNEL CALIBRATION.
(b) The IRM, and SRM channels shall be determfaed 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.
(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 APRM channel to conform to the power values 2,
calculated by a heat balance during OPERATIONAL CONDITION 1 when THERMAL POWER > 25% of RATED i
THERMAL POWER.
Adjust the APRM channel if the absolute difference is greater than 2%.
Any APRM channel gain. adjustment made in compliance with Specification 3.2.2 shall not be. included in EL determining the absolute difference.
2 (e) This calibration shall consist of the adjustment of the APRM flow biased channel to conform to a calibrated flow signal.
-(f) The LPRMs shall be calibrated at least once per 1000 effective full power hours (EFPH) using the TIP system.
(g) Measure and compare core flow to rated core flow.
(h) This calibration shall consist of verifying the 611 second simulated thermal power time constant.
TABLE 3.3.1-2 c.
REACTOR PROTECTION SYSTEM RESPONSE TIMES p:
TABLE 3.3.1-2 m
REACTOR PROTECTION SYSTEM RESPONSE TIMES E
Z RESPONSE TIME m
FUNCTIONAL UNIT (Seconds) 1.
a.
Neutron Flux - High*
NA b.
Inoperative NA 2.
Average Power Range Monitor
- a.
Neutron Flux - High, Setdown NA b.
Flow Biased Simulated Thermal Power-Upscale 1 0.09 w
c.
Fixed Neutron Flux - High 5 0.09 30 d.
Inoperative NA 3.
Reactor Vercel Steam Dome Pressure - High 1 0.55 4.
Reactor Vessel Water Level - Low, Level 3 1 1.05 5.
Main Steam Line Isolation Valve - Closure 5 0.06 6.
Main Steam Line Radiation - High NA 7.
Primary Containment Pressure - High NA 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 5 0.08, 11.
Reactor Mode Switch Shutdown Position NA 12.
Manual Scram NA 13.
Control Rod Drive F
a.
Charging Water Header Pressure - Low NA i
3 b.
Delay Timer NA 5
" Neutron detectors are exempt from response time testing.
Response time stall be measured y
from the detector output or from the input of the first electronic component in the channel.
- Not including simulated thermal power time constant.
m
- Measured from start of turbine control valve fast closure.
9 TABLE 3.3.1-1 (Continued)
REACTOR PROTECTION SYSTEM INSTRUMENTATION TABLE 3.3.1-1 (Continued) a REACTOR PROTECTION SYSTEM INSTRUMENTATION APPLICABLE MINIMUM OPERABLE OPERATIONAL CHANNELS PER FUNCTIONAL UNIT CONDITIONS TRIP SYSTEM (a)
ACTION 7.
Primary Containment Pressure - High 1, 2(#)
2(9) 1
~
8.
Scram Discharge Volume Water Level - High 1(hj, 2
1 5
2 3
9.
Turbine Stop Valve - Closure 1(I) 4(d) 6 10.
Turbine Control Valve Fast Closure, Valve Trip System 011 Pressure - Low 1(I) 2(3) 6
' 11.
Reactor Mode Switch Shutdown Position 1, 2 1
1-3, 4 1
L 5
1 3
12.
'1, 2
1 1
~'
3, 4 1
8 5
1 9
- 13.. Control Rod Drive a.
Charging, Water Header 2(h).
2 1
Pressure - tow 5
2 3
b.
Delay-Timer' 2
2 1
Ch) 2 3
S I
REACTIVITY CONTROL SYSTEM SURVEILLANCE REQUIREMENTS 4.1.3.5 Each control rod scram accumulator shall be determined OPERABLE:
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 disarmed 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 alarm setpoint 940 + 30, -0 psig on decreasing pressure.
o l'
l LIMITING SAFETY SYSTEM SETTING i
BASES REACTOR PROTECT 0N SYSTEM INSTRUMENTATION SETPOINTS (Continued)
- 13. Control Rod Drive (CRD) Charging Water Header Pressure - Low
-The Hydraulic Control Unit (HCU) scram accumulator is precharged with high pressure nitrogen (N ).
When the Control Rod Drive (CRD) pump is activated, 2
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, 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 irt a reduction of the accumulator energy and thereby degrades normal scram performance of the CRD's in the absence of sufficient reactor pressure.
The CR0 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, i
normal scram performance, is assured at all reactor pressures.
An adjustable time-delay relay is provided for each pressure tracsmitter/ trip channel to
. protect against inadvertant 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 logic.
The trip function is active in STARTUP and REFUEL modes because
- reactor pressure may be insufficient to assist the CRD scram action.
TABLE 2.2.1-1 REACTOR PROTECTION SYSTEM INSTRUMENTATION SETP0lNTS FUNCTIONAL UNIT TRTP SETPOINT ALLOWABLE VALUES 1.
Intermediate Range Monitor, Neutron Flux-High
-< 120 divisions of
< 122 divisions full scale
- of full scale 2.
Average Power Range Monitor:
a.
Neutron Flux-High, Setdown
-< 15% of RATED
< 20% of RATED THERMAL POWER THERMAL POWER b.
Flow Biased Simulated Thermal Power - Upscale
- 1) Two Recirculation Loop Operation a) Flow Biased
< 0.66W + 51% with a
< 0.66W + 54% with a
- maximum of
- maximum of b) High Flow Clamped 5 113.5% of RATED 1 115.5% of RATED THERMAL POWER THERMAL POWER
- 2) Single Recirculation Loop Operation a) Flow Biased
< 0.66W + 45.7% with
< 0.66W + 48.7% with
- a maximum of
- a maximum of b) High Flow Clamped 5 113.5% of RATED
-< 115.5% of RATED THERMAL POWER THERMAL POWER c.
Fixed Neutron Flux-High
-< 118% of RATED
< liO% of RATED THERMAL POWER
- THERMAL POWER
(
3.
Reactor Vessel Steam Dome Pressure - High 5 1043 psig i 1063 psig 4.
Reactor Vessel Water Level - Low, Level 3
> 12.5 inches above
> 11 inches above instrument zero*
- instrument zero*
5.
Main Steam Line Isolation Valve - Closure 1 8% closed
.< 12% closed 6.
Main Steam Line Radiation - High
-< 3 x full
< 3.6 x full power background power background 7.
Primary Containment Pressure - High 5 1.69 psig 5 1.89 psig 8.
Scram Discharge Volume Water Level - High 5 767' Sh" i 767' Sh" 9.
Turbine Stop Valve - Closure 1 5% closed 1 7% closed 10.
Turbine Control Valve Fast Closure, Trip Oil Pressur,e - Low 1 500 psig 1 414 psig
{
11.
Reactor Mode Switch Shutdown Position N.A.
N.A.
12.
Manual Scram N.A.
N.A.
13.
Control Rod Drive h
a.
Charging Water Header Pressure-Low 1 1157 psig 1 1134 psig b.
Delay Timer 5 10 seconds 5 10 seconds See Bases Figure B 3/4 3-1.
-