ML20093J818
| ML20093J818 | |
| Person / Time | |
|---|---|
| Site: | LaSalle  |
| Issue date: | 09/21/1984 |
| From: | Eisenhut D Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20093J821 | List: |
| References | |
| NUDOCS 8410170032 | |
| Download: ML20093J818 (12) | |
Text
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- .- B.
Paragraph 2.C.(7) is added as follows:
(7) ' Low Pressure in Pum) Discharge of the Control Rod Drive (Section 4.6.2, SSE1 #2, and Section 7.2.3.2, SSER #7)
Prior to completion of the.startup test program, the licenset shall install instrumentation that would automatically shut down the reactor in the event of low control rod drive pump discharge pressure. This automatic scram shall be activated during startup and refueling modes only.
3.
This amendment is effective as of date of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION Darrell G. Eisenhut, Director Division of Licensing
Enclosure:
Changes to the Technical Specifications
) ate cf Issuance: SEP 21884 Cth OELDhly..
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s COMMONWEALTH EDISON COMPANY DOCKET N0. 50-374-
-LA SALLE COUNTY STATION, UNIT 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 4 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 Commonwealth Edison-Company, dated July 31,1984, as supplemented by letters dated August 1 & 2,'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 ir.10 CFR Chapter I;
.B.
The facility will operate in conformit.! 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, and (ii) that such activities will be conducted in compliance with the Connission's regulations set forth in 10 CFR Chapter I; D.
The issuance of this amendment will not be inimical to the common 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 i
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:
l The Technical Specifications contained in Appendix A, as revised through Amendment No. 4, and the Environmental e
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.
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UNITED STATES E
NUCLEAR REGULATORY COMMISSION o
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a wasnwaTow, o.c.2 cess y.....)
COMONWE'ALTH EDISON COMPANY DOCKET NO. 50-374 LA SALLE COUNTY STATION, UNIT 2 AMENDMENT TO FACILITY GPERATING LICENSE
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Amendment No. 4 i
License No. NPF-18 1.
The Nuclear Regulatory Commission (the Comission or the NRC) having found that:
A.
The application for amendment filed by the Comonwealth Edison Company, dated July 31,1984, as supplemented by letters dated August 1 & 2, 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; 8.
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:
(1) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (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 common 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:
l 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 l
revised through Amendment No. 4, and the Environmental Protection Plan contained in Appendix B, are hereby l
incorporated in the license. The licensee shall operate the i
facility in accordance with the Technical Specifications and I
the Environmental Protection Plan.
i r
o
~n 2-B.
Paragraph 2.C.(7) is added as follows:
(7) Low Pressure in Pum) Discharge of the Control Rod Drive (Section 4.6.2 SSE' #2, and Section 7.2.3.2, SSER #7)
Prior to completien of the startup test program, the licensee shall install instrumentation that would automatically shut down
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the reactor in the event of low control rod drive pump discharge pressure. This automatic scram shall be activated during startup and refueling modes only.
3.
This amendment is effective as of date of issuance.
FOR THE NUCLEAR REGULATORY COMISSION O
arrell Eisenhut, Director M
Division of Licensing
Enclosure:
Changes to the Technical Specifications Date of Issuance: ggp g g
)
i
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ENCLOSURE TO LICENSE AMENDMENT NO._4 FACILITY OPERATING LICENSE NO. NPF-18 DOCKET NO. 50-374 4
Replace.he following pages of the Appendix "A" Technical Specifications with the enclosed pages. The revised pages are identified by Amendment number and contain vertical lines indicating the area of change.
. REMOVE INSERT 2-4 2-4 8 2-13 3/4 1-10 3/41-1d 3/4 3-3 3/4 3-3 3/4 3-5 3/4 3-5 3/4 3-6 3/. 3-6 3/4 3-8 3/4 3-8 8 3/4 1-3 8 3/4 1-3 C
TABLE 2.2.1-1 REACTOR PROTECTION SYSTEM INSTRUMENTATION SETPOINTS FUNCTIONAL UNIT TRIP SETPOINT ALLOWABLE VALUES
${
1.
Intermediate Range Monitor, Neutron Flux-High 1 120 divisions of
-< 122 divisions full scale of full scale 2.
Average Power Range Monitor:
g Z
a.
Neutron Flux-High, Setdown
$ 15% of RATED 1 20% of RATED m
THERMAL POWER THERMAL POWER b.
Flow Biased Simulated Thermal Power - Upscale
- 1) Two Recirculation Loop Operation a) Flow Biased 1 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 5 0.66W + 45.7% with 5 0.66W + 48.7% with a maximum of a maximum of b) High Flow Clamped i 113.5% of RATED 1 115.5% of RATED
,y THERMAL POWER THERMAL POWER c.
Fixed Neutron Flux-High 5 118% of RATED 1 120% of RATED' THERMAL POWER THERMAL POWER 3.
Reactor Vessel Steam Dome Pressure - High 5 1043 psig 5 1063 psig 4.
Reactor Vessel Water Level - Low, level 3 1 12.5 inches above 1 11 inches above instrument zero*
instrument zero*
5.
Main Steam Line Isolation Valve - Closure i 8% closea
~< 12% closed 6.
Main Steam Line Radiation - liigh 1 3 x full
-< 3.6 x full power background power background 7.
Primary Containment Pressure - High 1 1.69 psig 5 1.89 psig
!g 8.
Scram Discharge Volume Water Level - High 5 767' Sh" 1 767' Sk"
- R 9.
Turbine Stop Valve - Closure 1 5% closed i 7% closed
.a i@
10.
Turbine Control Valve fast Closure, Trip Oil Pressure - Low 3 500 psig 3 414 psig
.2 11.
Reactor Mode Switch 5.utdown Position N.A.
N.A.
12.
Manual Scram N.A.
N.A.
- See Bases Figure B 3/4 3-1.
1 )
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~ REACTIVITY CONTROL SYSTEM SURVEILLANCE REQUIREMENTS
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4.1.3.5 Each codtrol rod scram-accumulator shall be determined OPERA 8LE:
a.
At least oqce 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.
2.
Measuring and recording the time that each individual accumulator check valve maintains the associated accumulator pressure above the alarm setpoint with no control rod drive pump operating.
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LA SALLE - UNIT 2 3/4 1-10 Amendment No. 4
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TABLE 3.3.1-1 (Continued)
E REACTOR PROTECTION SYSTEM INSTRUMENTATION N
r-APPLICABLE MINIMUM OPERABLE 0PERATIONAL CHANNELS PER
/
FUNCTIONAL UNIT CONDITIONS TRIP SYSTEM (a)
' ACTION 7'.
Primary Containment Pressure - High 1, 2 2
ff)
I9) 1,.
8.
Scram Discharge Volume Water
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Level - High 15(h$'
2 3
III I3) 9.
Turbine Stop Valve - Closure I
4 6
10.
Turbine Control Valve Fast Closure, I ')
Id)
I Valve Trip System Oil Pressure - Low 2
6 w
11.
Reactor Mode Switch Shutdown 3
Position 1, 2 1
1 lw 3, 4 1
7
- O 5
1 3
12.
Manual 5 cram 1, 2 1
1 3, 4 1
8 5
1 9
I 1
! B-9 e
I
TA8LE 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 <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> for required surveillance without placing the ti'p 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 LPRM inputs to an APRM channel.
4 (d) This function is not required to be OPERABLE when the reactor pressure vessel head is unbolted or removed per Specification 3.10.1.
(e) 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.
l (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 pressure is < 140 psig, equivalent to THERMAL POWER less than 30% of RATED THERMAL POWER.
(j) Also actuates the EOC-RPT system.
- Not required for control rods removed per Specification.3.9.10.1 or 3.9.10.2.
C LA SALLE - 06 NIT 2 h-
M*nfment _No1_4 _ _
TABLE 3.3.1-2 c-2 REACTOR PROTECTION SYSTEM RESPONSE TIMES vsN E
FUNCTICNAL UNIT RESPONSE TIME c:3 (Seconds)
-4 1.
N a.
Neutron Flux - High*
NA b.
Inoperative NA
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2.
Average Power Range Monitor
- a.
Neutron Flux - High, Setdown NA b.
Flow Biased Simulated Thermal Power-Upscale 1 0.09 c.
Fixed Neutron Flux - High 5 0.09 d.
Inoperative NA i
3.
Reactor Vessel Steam Dome Pressure - High 1 0.55 w
4.
Reactor Vessel Water Level - Low, Level 3
< 1.05 2;
5.
Main Steam Line Isolation Valve - Closure I 0.06 y
6.
Main Steam Line Radiation - High SA 4
as 7.
Primary Containment Pressure - High NA 8.
Scram Discharge Volume Water Level - High NA 9.
Turbine.Stop Valve - Closure 1 0.06 10.
Turbine Control Valve Fast Closure, Trip Oil Pressure - Low
$ 0.08, 11.
Reactor Mode Switch Shutdown Position NA
'12.
Manual Scram NA EI
- Neutron detectors are exempt from response time testing.
Response time shall be measured e
{p from the detector output or from the input of the first electronic component in the e
channel.
5
- Not including simulated thermal power time constant.
- Measured from start of turbine control valve fast closure.
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- s TABLE 4.3.1.1-1 (Continued)
E REACTOR PROTECTION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS hk CHANNEL OPERATIONAL CHANNEL FUNCTIONAL CHANNEL CONDITIONS FOR WHICH FUNCTIONAL UNIT CHECK TEST CALIBRATION SURVEILLANCE REQUIRED c:
J5
-4 8.
Scram Discharge Volume Water N
Level - High NA M
R
~T, 2, 5 9.
Turbine Stop Valve - Closure NA M
R 1
10.
Turbine Control Val.e Fast Closure Valve Trip System Oil Pressure - Low NA M
R 1
11.
Reactor Mode Switch Shutdown Position NA R
NA 1,2,3,4,5 i
12.
Manual Scram NA M
NA 1,2,3,4,5 P
ca (a) Neutrom detectors may be excluded from CHANNEL CALIBRATION, 1
(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.
(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 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 El channel gain adjustment made in compliance with Specificativ.i 3.2.2 shall not be included in g
determining the absolute difference.
5 (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 w
system.
(g) Measure and compare core flow to rated core flow.
(h) This calibration shall consist of verifying the 6 i 1 second simulated thermal power time constant.
1
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14 REACTIVITY CONTROL SYSTEMS 8ASES CONTROL ~R005 (Continued)
Control rod coupling integrity is required to ensure compliance with the analysis of the rod drop accident in the FSAR. -The overtravel position feature
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provides the only positive means of determining that a rod is properly coupled and therefore this check must be performed prior to achieving criticality after 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 contributet 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 excer.sive wear on the system componentsr.
3/4.1.4 CONTROL ROD 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 RUM 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 erroneous rod withdrawal from locations of high power density during high power operation.
Two channels are provided.
Tripping one of the channels will block c
erroneous rod withdrawal soon enough to prevent fuel damage.
This system backs up the written sequence used by the operator for withdraival of control rods.
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-. -. -... - -.