ML17291A893
| ML17291A893 | |
| Person / Time | |
|---|---|
| Site: | Columbia  |
| Issue date: | 06/26/1995 |
| From: | Clifford J NRC (Affiliation Not Assigned) |
| To: | |
| Shared Package | |
| ML17291A894 | List: |
| References | |
| NUDOCS 9507070435 | |
| Download: ML17291A893 (25) | |
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UNITED STATES NUCLEAR REGULATORY COMMISSION WASHING'TONs D.C. 20555-0001 S
INGTON PUB IC POWER SUPPLY SYST M
6 ->>7 NUC AR PROJ T NO.
2 AMENDMENT TO F CILITY OP RATING LICENSE Amendment No. i39 License No.
NPF-21 2.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by the Washington Public Power Supply System (licensee) dated July 12,
- 1994, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Commission'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 rules and regulations of the Commission; C.
There is 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 Commission's regulations; 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 Commission's regulations and all applicable requirements have been satisfied.
Accordingly, the license is amended by changes to the Technical Speci-fications as indicated in the attachment to this license amendment and paragraph 2.C.(2) of Facility Operating License No.
NPF-21 is hereby amended to read as follows:
9507070435 950b2b PDR ADQCK 05000311I7 P
(2) e ic t'on a
d vi mental Protection Pla The Technical Specifications contained in Appendix A, as revised through Amendment No.
139 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.
3.
This amendment is effective as of its date of issuance to be implemented within 30 days of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION Ja es W. Clifford, Senior Project Hanager Project Directorate IV-2 Division of Reactor Projects III/IV Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications Date of Issuance:
dune 26, 1995
N M
ACHMENT TO LIC S
AM NOMENT 139 T FAC TY 0 RAT NG GENS NO. NPF-2 CK NO. 50-397 Replace the 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 areas of change.
The corresponding overleaf pages are also provided to maintain document completeness.
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=.
XX1 3/4 3-1 3/4 3-6 3/4 3-10 3/4 3-11 3/4 3-12*
3/4 3-19 3/4 3-20 3/4 3-21 3/4 3-25 3/4 3-26*
3/4 3-33 B 3/4 3-1 8 3/4 3-2 I SERT XX1 3/4 3-1 3/4 3-6 3/4 3-10 3/4 3-11 3/4 3-19 3/4 3-20 3/4 3-21 3/4 3-25 3/4 3-33 B 3/4 3-1 B 3/4 3-2
- These pages are unchanged.
They are being reissued to retain consistency as being overleaf pages.
'f.
~i
LIST OF TA S
TAB~
IND PAG
- 1. 1 SURVEILLANCE FREQUENCY NOTATION..............'...............
1-9 1.2 OPERATIONAL CONDITIONS......................................
1-10 2.2.1-1 REACTOR PROTECTION SYSTEM INSTRUMENTATION SETPOINTS.........
2-4 B2.1.2-1 UNCERTAINTIES USED IN THE DETERMINATION OF THE FUEL CLADDING SAFETY LIMIT..................................
B 2-3 3.2.3-1 HCPR OPERATING LIHITS FOR RATED CORE FLOW...................
Deleted 3.3. 1-1 REACTOR PROTECTION SYSTEM INSTRUMENTATION...................
3/4 3-2 4.3. 1. 1-1 REACTOR PROTECTION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS..................................
3/4 3-7 3.3.2-1 ISOLATION ACTUATION INSTRUMENTATION.........................
3/4 3-12 3.3.2-2 ISOLATION ACTUATION INSTRUMENTATION SETPOINTS........ -...... 3/4 3-16 4.3.2.1-1 ISOLATION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS................................................
3/4 3-22 3.3.3-1 3 ~ 3 ~ 3 2
EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION.............................................
3/4 3-26 EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION SETPOINTS...................................
3/4 3-30 4.3.3.1-1 EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION SURVEILLANCE REQUIREMENTS...................
3/4 3-34 3.3.4. 1-1 3.3.4.1-2 4.3.4.1-1 ATWS RECIRCULATION PUMP TRIP SYSTEM INSTRUMENTATION..............................
ATWS RECIRCULATION PUMP TRIP SYSTEH INSTRUMENTATION SETPOINTS....................
ATWS RECIRCULATION PUMP TRIP ACTUATION INSTRUMENTATION SURVEILLANCE REQUIREMENTS....
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o 3/4 3 40 WASHINGTON NUCLEAR UNIT 2 XX1 Amendment No. 94, 139
ST F TAB S
3.3.4.2-1 3.3.4.2-2 3.3.4.2-3 END-OF-CYCLE RECIRCULATION PUMP TRIP SYSTEM INSTRUMENTATION
. 3/4 3-43 END-OF-CYCLE RECIRCULATION PUMP TRIP SETPOIKTS....
3/4 3-44 END-OF-CYCLE RECIRCULATION PUMP TRIP SYSTEM
RESPONSE
TINES
. 3/4 3-45 4.3.4.2.1-1 END-OF-CYCLE RECIRCULATION PUMP TRIP SYSTEM SURVEILLANCE REQUIREMENTS..............
3/4 3-46 3.3.5-1 3.3.5-2 4.3.5.1-1 3.3.6-1 3.3.6-2 4.3.6-1 3.3.7.1-1 4.3.7.1-1 3.3.7.3-1 4.3.7.3-1 3.3.7.4-1 4.3.7.4-1 METEOROLOGICAL MONITORING INSTRUMENTATION
. 3/4 3-65 METEOROLOGICAL MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS
. 3/4 3-66 REMOTE SHUTDOWN MONITORING INSTRUMENTATION...... 3/4 3-68 REMOTE SHUTDOWN MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS
. 3/4 3-69 REACTOR CORE ISOLATION COOLING SYSTEM ACTUATION INSTRUMENTATION............. -..... 3/4 3-48 REACTOR CORE ISOLATION COOLING SYSTEH ACTUATION INSTRUMENTATION SETPOINTS..............
3/4 3-50 REACTOR CORE ISOLATION COOLING SYSTEM ACTUATION INSTRUHENTATION SURVEILLANCE REQUIREMENTS......
3/4 3-51 CONTROL ROD BLOCK INSTRUMENTATION
. 3/4 3-53 CONTROL ROD BLOCK INSTRUMENTATION SETPOINTS
. 3/4 3-55 CONTROL ROD BLOCK INSTRUMENTATION SURVEILLANCE REQUIREMENTS
. 3/4 3-56 RADIATION MONITORING INSTRUMENTATION......... 3/4 3-59 RADIATION MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS..............
3/4 3-60 WASHINGTON NUCLEAR - UNIT 2 xxii Amendment No. 131
3/4.3 INSTRUMENTATION 3 4.3.
R CT P 0 CT ON SYSTEM INSTRUM NTAT ON LIMITING CONDITION OR OPERATION 3.3. 1 As a minimum, the reactor protection system instrumentation channels shown in Table 3.3. 1-1 shall be OPERABLE.
A~PP I CA I ITY:
A h
i i b1 3.3.l-l.
ACTION:
a.
With the number of OPERABLE channels less than required by the Minimum OPERABLE Channels per Trip System requirement for one trip
- system, place the inoperable channel(s) and/or that trip system in the tripped condition* within twelve hours.
The provisions of Specification 3.0.4 are not applicable.
b.
With the number of OPERABLE channels less than required by the Minimum OPERABLE Channels per Trip System requirement for both trip
- systems, place at least one trip system** in the tripped condition within 1 hour-and take the ACTION required by Table 3.3. 1-1.
SURVEILLANCE RE UIREMENTS 4.3. 1. 1 Each reactor protection system instrumentation channel shall be demonstrated OPERABLE by the performance of the CHANNEL CHECK, CHANNEL FUNCTIONAL TEST and CHANNEL CALIBRATION operations for the OPERATIONAL CONDITIONS and at the frequencies shown in Table 4.3.1.1-1.
4.3. 1.2 LOGIC SYSTEM FUNCTIONAL TESTS and simulated automatic operation of all channels shall be performed at least once per 18 months.
4.3. 1.3 The REACTOR PROTECTION SYSTEM RESPONSE TIME of each reactor trip functional unit shall be demonstrated to be within its limit at least once per 18 months.
Neutron detectors are exempt from response time testing.
Each test shall include at least one channel per trip system such that all channels are tested at least once every N times 18 months where N is the total number of redundant channels in a specific reactor trip system.
- An inoperable channel need not be placed in the tripped condition where this would cause the Trip Function to occur.
In these
- cases, the inoperable channel shall be restored to OPERABLE status within six hours after the channel was first determined to be inoperable or the ACTION required by Table 3.3. 1-1 for that Trip Function shall be taken.
- Ifmore channels are inoperable in one trip system than in the other, place the trip system with more inoperable channels in the tripped condition, except when this would cause the Trip Function to occur.
WASHINGTON NUCLEAR UNIT 2 3/4 3-1 Amendment No. 90, 139
TABLE 3.3.1-1 REACTOR PROTECTION SYSTEH INSTRUMENTATION FUNCTIONAL UNIT t
Intemediate Range Honitors:
a.
Neutron Flux - High APPLICABLE OPERATIONAL CONDITIONS 2
3, 4'(b)
HINIHUH OPERABLE CHANNELS PER TRIP SYSTEH a
b.
Inoperative 2.
Average Power Range Honitor(c):
a.
Neutron Flux - High, Setdown 2
3, 4 5
2 3
5(b) b.
C.
d.
Flow Biased Sieulated Theraa'l Power - High Fixed Neutron Flux - High Inoperative 1
1 1, 2 3
5 3.
Reactor Vessel Steam Dose Pressure - High 4.
Reactor Vessel Mater Level - Low, Level 3 5.
Hain Steatw Line Isolation Valve <<
Closure 1, 2(e) l(d)
D41 (C ti d) 0 PROT CT ON S
S NSTRUM NTA 0
TA LE NOTAT ONS (a)
(b)
(c)
(d)
(e)
(g)
(h)
A channel may be placed in an inoperable status for up to six hours 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.
The "shorting links" shall be removed from the RPS circuitry prior to and during the time any control rod is withdrawn* and shutdown margin demonstrations are being performed per Specification 3.10.3.
An APRH channel is inoperable if there are less than 2
LPRH inputs per level or less than 14 LPRH inputs to an APRH channel.
This function shall be automatically bypassed when the reactor mode switch is not in the Run position and reactor pressure ( 1060 psig.
This function is not required to be OPERABLE when the reactor pressure vessel head is removed per Specification 3. 10. 1.
This function is not required to be OPERABLE when PRINRY CONTAINMENT INTEGRITY is not required.
Also actuates the standby gas treatment system.
With any control rod withdrawn.
Not applicable to control rods removed per Specification 3.9.10.1 or 3.9.10.2.
This function shall be automatically bypassed based on turbine first stage pressure when THERNL POWER is less than 30X of RATED THERMAL POWER.
Also actuates the EOC-RPT system.
- Not required for control rods removed per Specification 3.9.10.1 or 3.9. 10.2.
WASHINGTON NUCLEAR UNIT 2 3/4 3-5 Amendment No. 99,i37
THIS PAGE INTENTIONALLY BLANK WASHINGTON NUCLEAR UNIT 2 3/4 3-6 Amendment No. 7&+H~,139
TABLE 4.3.1.1-1 (Continued)
REACTOR PROTECTION SYSTEM INSTRUMENTATION SURVEILLANCE RE UIREMENTS TABLE NOTATIONS (a)
Neutron detectors may be excluded from CHANNEL CALIBRATION.
'I (b)
The IRM and SRM channels shall be determined to overlap for at least 1/2 decade during each startup after entering OPERATIONAL CONDITION 2 and the IRM and APRM channels shall be determined to overlap for at least l/2 decade 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 > 25K of RATED THERMAL POWER.
Adjust the APRM channel if the absolute di7ference is greater than 2X of RATED THERMAL POWER.
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 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 6 4 1 second simulated thermal power time constant.
(i)
DELETED (j)
With any control rod withdrawn.
Not applicable to control rods removed per Specification 3.9.10.1 or 3.9.10.2.
WASHINGTON NUCLEAR - UNIT 2 3/4 3-9 Amendment No. ii2
NST U
T ON 3 4.3 ATI UA ON NSTRUM NTATION LIMITING CON ITION FOR OPERATION 3.3.2 The isolation actuation instrumentation channels shown in Table 3.3.2-1 shall be OPERABLE with their trip setpoints set consistent with the values shown in the Trip Setpoint column of Table 3.3.2-2.
~LA TY:
A h
i T bl
.3.2-1.
~CT~O:
a ~
b.
With an isolation actuation instrumentation channel trip setpoint less conservative than the value shown in the Allowable Values column of Table 3.3.2-2, declare the channel inoperable until the channel is restored to OPERABLE status with its trip setpoint adjusted consistent with the Trip Setpoint value.
With the number of OPERABLE channels less than required by the Minimum OPERABLE Channels per Trip System requirement for one trip system:
l.
If placing the inoperable channel(s) in the tripped condition would cause an isolation, the inoperable channel(s) shall be restored to OPERABLE status within a) 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> for trip functions common to RPS Instrumentation; and b) 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> for trip functions not common to RPS Instrumentation.
or the ACTION required by Table 3.3.2-1 for the affected trip function shall be taken.
OR 2.
If placing the inoperable channel(s) in the tripped conditions would not cause an isolation, the inoperable channel(s) and/or that trip system shall be placed in the tripped condition within a) b)
12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> for trip functions common to RPS Instrumentation; and 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> for trip functions not common to RPS Instrumentation.
The provisions of Specification 3.0.4 are not applicable.
c.
With the number of OPERABLE channels less than required by the Minimum OPERABLE Channels per Trip System requirement for both trip
- systems, place. at least one trip system* in the tripped condition within one hour and take the ACTION required by Table 3.3.2-1.
- Place one trip system (with the most inoperable channels) in the tripped condition.
The trip system need not be placed in the tripped condition when this would cause the isolation to occur.
WASHINGTON NUCLEAR UNIT 2 3/4 3-10 Amendment No. 99,139
~IN T
NN NTN TN SURV A
C R
U M
TS 4.3.2. 1 Each isolation actuation instrumentation channel shall be demonstrated OPERABLE by the performance of the CHANNEL CHECK, CHANNEL FUNCTIONAL TEST and CHANNEL CALIBRATION operations for the OPERATIONAL CONDITIONS and at the frequencies shown in Table 4.3.2.1-1.
4.3.2.2 LOGIC SYSTEM FUNCTIONAL TESTS and simulated automatic operation of all channels shall be performed at least once per 18 months.
4.3.2.3 The ISOLATION SYSTEM RESPONSE TIME of each isolation trip function shall be demonstrated to be within its limit at least once per 18 months.
Radiation detectors are exempt from response time testing.
Each test shall include at least one channel per trip system such that all channels are tested at least once every N times 18 months, where N is the total number of redundant channels in a specific isolation trip system.
WASHINGTON NUCLEAR - UNIT 2 3/4 3-11 Amendment No. 449,139
TRIP FUNCTION 1.
PRIMARY CONTAINMENT ISOLATION a.
Low, Level 3
2)
Low Low, Level 2
b.
Drywell Pressure - High c.
DELETED 2)
Pressure - Low 3)
Flow - High d.
Main Steam Line Tunnel Temperature - High e.
Main Steam Line Tunnel h Temperature - High Condense Vacuum - Low Manual Initiation f.
go 2.
SECONDARY CONTAINMENT ISOLATION 5(g) 1, 2, 4 4
5(b)(g) 2'(d) 1 1
1 2
5(b)(g) 2 2
2/gr oup 1/group 1/group TABLE 3.3.2-1 ISOLATION ACTUATION INSTRUMENTATION VALVE GROUPS MINIMUM OPERATEO BY OPERABLE CHANNELS SIGNAL PER TRIP SYSTEM a
APPLICABLE OPERATIONAL CONDITION 1, 2, 3
1 1, 2, 3 1,2,3 1,2,3 1
2t 3*
1, 2, 3
1, 2, 3 ACTION 20 20 20 23 21 21 21 21 24 24 24 aO b.
C.
d.
Reactor Building Vent Exhaust Plenum Radiation - High Drywell Pressure - High Reactor Vessel Water Level - Low Low, Level 2 Manual Initiation 3(b)(e) 3(b)(e) 3(b)(e) 3(b) 3(b) 2 1/group 1/group 1, 2, 3, and **
25 1, 2, 3
25 1, 2, 3, and 8 25 1, 2, 3
24
- It 24 o
THIS PAGE INTENTIONALLY BLANK WASHINGTON NUCLEAR UNIT 2 3/4 3-19 Amendment No. 448, ~39
THIS PAGE INTENTIONALLY BLANK k
WASHINGTON NUCLEAR - UNIT 2 3/4 3-20
,J, Amendment No. 48,139
THIS PAGE INTENTIONALLY BLANK WASHINGTONNUCLEAR UNIT 2 3/4 3-21 Amendment No.~39
TABLE 4.3.2. 1-1 ISOLATION ACTUATION INSTRUMENTATION SURVEILLANCE RE UIREMENTS TRIP FUNCTION 1.
PRIMARY CONTAINMENT ISOLATION a.
Reactor Vessel Water Level-1)
Low, Level 3
2)
Low Low, Level 2
b.
Drywell Pressure - High c.
DELETED 2)
Pressure - Low 3)
Flow - High d.
Main Steam Line Tunnel Temperature - High e.
Main Steam Line Tunnel h Temperature - High f.
Condenser Vacuum - Low g.
Manual Initiation 2.
SECONDARY CONTAINMENT ISOLATION CHANNEL CHECK S
N.A.
N.A.
N.A.
S N.A.
N.A.
N.A.
N.A.
CHANNEL FUNCTIONAL TEST SA SA Q
CHANNEL CALIBRATION R
R N.A.
OPERATIONAL CONDITIONS FOR WHICH SURVEILLANCE RE UIRED 1, 2, 3
1, 2, 3
1, 2, 3
1 1, 2, 3 1, 2, 3
1, 2 3
1'k 3*
1, 2, 3
a.
b.
C.
d.
Reactor Building Vent Exhaust Plenum Radiation - High Drywell Pressure - High Reactor Vessel Water Level - Low Low, Level 2 Manual Initiation S
N.A.
N.A.
N.A.
R N.A.
1 2
3 and *"
1,2,3, and8 1, 2, 3, and **
~TTATTCA 3 4.3 G
00 ING SYSTEM ACTUATION INSTRUMENTATION LIMITING CO ITION FOR OPERATION 3.3.3 The emergency core cooling system (ECCS) actuation instrumentation channels shown in Table 3.3.3-1 shall be OPERABLE with their trip setpoints set consistent with the values shown in the Trip Setpoint column of Table 3.3.3-2.
3 1:
A 1
1 1 31 3.3.3-1.
ACTION:
a.
With an ECCS actuation instrumentation channel trip setpoint less conservative than the value shown in the Allowable Values column of Table 3.3.3-2, declare the channel inoperable until the channel is restored to OPERABLE status with its trip setpoint adjusted consistent with the Trip Setpoint value.
b.
With one or more ECCS actuation instrumentation channels inoperable, within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> take the ACTION required by Table 3.3.3-1.
c.
With either ADS trip system "A" or "B" inoperable, restore the inoperable trip'system to OPERABLE status:
- 2. Within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.
Otherwise, be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and reduce reactor steam dome pressure to less than or equal to 128 psig within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
SURVEILLANCE RE UIREMENTS 4.3.3. 1 Each ECCS actuation instrumentation channel shall be demonstrated OPERABLE by the performance of the CHANNEL CHECK, CHANNEL FUNCTIONAL TEST, and CHANNEL CALIBRATION operations for the OPERATIONAL CONDITIONS and at the frequencies shown in Table 4.3.3. 1-1.
4.3.3.2 LOGIC SYSTEM FUNCTIONAL TESTS and simulated automatic operation of all channels shall be performed at least once per 18 months.
4.3.3.3 The ECCS
RESPONSE
TIME of each ECCS trip function shall be demonstrated to be within the limit at least once per 18 months.
Each test shall include at least one channel per trip system such that all channels are tested at least once every N times 18 months where N is the total number of redundant channels in a specific ECCS trip system.
WASHINGTON NUCLEAR UNIT 2 3/4 3-25 Amendment No. 494-~, "39
TABLE 3.3.3-1 BlERGENCY CORE COOLING SYSTEH ACTUATION INSTRUMENTATION TRIP FUNCTION A.
DIVISION I TRIP SYSTEH 1.
RHR-A LPCI MODE 5 LPCS SYSTEM a.
Reactor Vessel Water Level " Low Low Low, Level= 1 b.
Drywell Pressure - High c.
LPCS Pump Discharge Flow-Low (Minimum Flow) d.
Reactor Vessel Pressure-Low (LPCS Permissive) e.
Reactor Vessel Pressure-Low (LPCI Permissive) f.
LPCI Pump A Start Time Delay Relay g.
LPCI Pump A Discharge Flow-Low (Hinimum Flow) h.
Manual Initiation 2.
AUTOMATIC DEPRESSURIZATION SYSTEH TRIP SYSTEH "A"t 1
1 1/division 1, 2, 3, 4*, 5" 1 j 2g 3 1, 2, 3, 4*, 5+
lg 2 ~ 3$
4 5*
- 1) 2, 3, 4, 5*
MINIMUMOPERABLE APPLICABI E CHANNELS Pg)
OPERATIONAL TRIP SYSTEH CONDITIONS ACTION 30 31 32 33 32 33 32 31 34 aO b.
C.
d.
e.f.
g.
Reactor Vessel Water Level - Low Low Low, Level 1 ADS Timer Reactor Vessel Water Level - Low, Level 3 (Permissive)
LPCS Pump Discharge Pressure-High (Pump Runninq)
LPCI Pump A Discharge Pressure-High (Pump Running)
Manual Initiation Inhibit Switch 2'
1 2
2 2/division 1/division 1, 2, 3 1, 2, 3 1, 2, 3 1, 2, 3
30 32 32 32 32 35 35
THIS PAGE INTENTIONALLY BLANK WASHINGTON NUCLEAR UNIT 2 3/4 3-33 Amendment No. f39
TABLE 4.3.3. 1-1 EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION SURVEILLANCE RE UIREMENTS TRIP FUNCTION A.
DIVISION I TRIP SYSTEM AND LPCS SYSTEM a.
eac or esse a er eve Low Low Low, Level 1
b.
,Drywell Pressure - High c.
LPCS Pump Discharge Flow-Low (Minimum Flow) d.
Reactor Vessel Pressure-Low (LPCS Permissive) e.
Reactor Vessel Pressure-Low (LPCI Permissive) f.
LPCI Pump A Start Time Delay Relay g.
LPCI Pump A Flow-Low (Minimum Flow) h.
Manual Initiation TRIP SYSTEM "A"8lift L t-Low Low Low, Level 1 b.
ADS Timer c.
Reactor Vessel Water Level-Low, Level 3 (Permissive) d.
LPCS Pump Discharge Pressure-High (Pump Running) e.
LPCI Pump A O>scharge Pressure-High (Pump Running) f.
Manual Initiation g.
Inhibit Switch 2.
AUTOMATIC OEPRESSURIZATION SYSTEM CHANNEL CHECK S
N.A.
N.A.
N.A.
N.A.
N.A.
N.A.
N.A.
S N.A.
N.A.
N.A.
N.A.
N.A.
CHANNEL FUNCTIONAL CHANNEL TEST CALIBRATION R
N.A.
R N.A.
N.A.
OPERATIONAL CONDITIONS FOR WHICH SURVEILLANCE RE UIREO 1j 2j 3j 4*j 5*
1, 2, 3 1, 2, 3, 4", 5" 1, 2, 3, 4*, 5*
1, 2, 3, 4*, 5" 1, 2, 3, 4", 5" 1, 2, 3, 4", 5" 1, 2, 3
1, 2, 3
1, 2, 3 1, 2, 3 1 2, 3 1, 2, 3
1, 2, 3
3 4.3 NSTRUMENTAT ON BASES 3 4.3.
R ACTOR PROT T ON SYSTEM INSTRUMENTATION The reactor protection system automatically initiates a reactor scram to:
a.
Preserve the integrity of the fuel cladding.
b.
.Preserve the integrity of the reacto} coolant system.
c.
Minimize the energy which must be adsorbed following a loss-of-coolant accident, and d.
Prevent inadvertent criticality.
This specification provides the limiting conditions for operation necessary to preserve the ability of the system to perform its intended function even during periods when instrument channels may be out of service because of maintenance.
When necessary, one channel may be made inoperable for brief intervals to conduct required surveillance.
The reactor protection system is made up of two independent trip systems.
There are usually four channels to monitor each parameter with two channels in each trip system.
The outputs of the channels in a trip system are combined in a logic so that either channel will trip that trip system.
The tripping of both trip systems will produce a reactor scram.
The system meets the intent of IEEE-279 for nuclear power plant protection systems.
Specified surveillance intervals and surveillance and maintenance outage times have been determined in accordance with NEDC 30851 P,
"Technical Specification Improvement Analyses for BWR Reactor Protection System,"
as approved by the NRC and documented in the SER (letter to T. A. Pickens from A. Thadani dated July 15, 1987).
The bases for the trip settings of the RPS are discussed in the bases for Specification 2.2. 1.
The RPS instrumentation that provides I) the Turbine Throttle Valve-Closure and 2) Turbine Governor Valve Fast Closure, Valve Trip System Oil Pressure - Low trip signals measures first stage turbine pressure to initiate a trip signal.
The Load Rejection safety analysis (FSAR 15.2.2) bases initial conditions on rated power and specifies turbine bypass operability at greater than or equal to 30X of rated thermal power.
Because first stage pressure can vary depending on operating conditions, the qualifying notes describing when the turbine bypass feature is to be disabled specify a turbine first stage pressure corresponding to less than 30X RTP (turbine first stage pressure is dependent on the operating parameters of the reactor,
- turbine, and condenser).
Therefore, because a value for turbine first stage pressure cannot be precisely fixed and because pressure measurement initiates the trip, the Technical Specification refers to a pressure associated with a specific Rated Thermal Power value rather than a value for pressure.
The measurement of response time at the specified frequencies provides assurance that the protective functions associated with each channel are completed within the time limit assumed in the safety analyses.
No credit was taken for those channels with response times indicated as not applicable.
Response
time may be demonstrated by any series of sequential, over lapping or total channel test measurement, provided such tests demonstrate the total channel response time as defined.
Sensor response time verification may be demonstrated by either (I) inplace, onsite or offsite test measurements, or (2) utilizing replacement sensors with certified response times.
The response time limits are contained in FSAR Chapter 7.
WASHINGTON NUCLEAR UNIT 2 B 3/4 3-1 Amendment No. 90-,487, "39
INSTRUH T TIO BASES 3 4.3.2 SO T ON CT TION NSTRUMENTATION This specification ensures the effectiveness of the instrumentation used to mitigate the consequences of accidents by prescribing the OPERABILITY trip setpoints for isolation of the reactor systems.
When necessary, one channel may be inoperable for brief intervals to conduct required surveillance.
Some of the trip settings may have tolerances explicitly stated where both the high and low values are critical and may have a substantial effect on safety.
The setpoints of other instrumentation, where only the high or low end of the setting have a direct bearing on safety, are established at a level away from the normal operating range to prevent inadvertent actuation of the systems involved.
Except for the MSIVs, the safety analysis does not address individual sensor response times or the response times of the logic systems to which the sensors are connected.
For D.C.-operated
- valves, a 3-second delay is assumed before the valve starts to move.
For A.C.-operated valves, it is assumed that the A.C. power supply is lost and is restored by startup of the emergency diesel generators.
In this event, a time of 13 seconds is assumed before the valve starts to move.
In addition to the pipe break, the failure of the D.C.-
operated valve is assumed; thus the signal delay (sensor response) is con-current with the 13-second diesel startup.
The safety analysis considers an allowable inventory loss in each case which in turn determines the valve speed in conjunction with the 13-second delay.
It follows that checking the valve speeds and the 13-second time for emergency power establishment will establish the response time for the isolation functions.
However, to enhance overall system reliability and to monitor instrument channel response time trends, the isolation actuation instrumentation response time shall be measured and recorded as a part of the ISOLATION SYSTEM RESPONSE TIME.
The response time limits are contained in FSAR Chapter 7.
Operation with a trip set less conservative than its Trip Setpoint but within its specified Allowable Value is acceptable on the basis that the difference between each Trip Setpoint and the Allowable Value is equal to or less than the drift allowance assumed for each trip in the safety analyses.
3 4.3.3 HERG NCY C
COO ING SYST M ACTUAT ON INSTRUMENTATION The emergency core cooling system actuation instrumentation is provided to initiate actions to mitigate the consequences of accidents that are beyond the ability of the operator to control.
This specification provides the OPERABILITY requirements and trip setpoints that will ensure effectiveness of the systems to provide the design protection.
Although the instruments are listed by system, in some cases the same instrument may be used to send the actuation signal to more than one system at the same time.
The response time limits are contained in-FSAR Chapter 7.
Operation with a trip set less conservative than its Trip Setpoint but within its specified Allowable Value is acceptable on the basis that the difference between each Trip Setpoint and the Allowable Value is equal to or less than the drift allowance assumed for each trip in the safety analyses.
WASHINGTON NUCLEAR UNIT 2 B 3/4 3-2.
Amendment No. i39