ML20078R708
| ML20078R708 | |
| Person / Time | |
|---|---|
| Site: | LaSalle  |
| Issue date: | 12/14/1994 |
| From: | COMMONWEALTH EDISON CO. |
| To: | |
| Shared Package | |
| ML19311B586 | List: |
| References | |
| NUDOCS 9412270176 | |
| Download: ML20078R708 (256) | |
Text
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ATTACHMENT B PROPOSED CHANGES TO THE TECHNICAL ,
SPECIFICATIONS FOR OPERATING LICENSES NPF-11 AND NPF-18 '
i NPF-11 NPF-18 B 2-9 B 2-9 3/4 3-1 3/4 3-1 Inserts A and B Inserts A and B '
3/4 3-5 3/4 3-5 3/4 3-7 3/4 3-7 3/4 3-8 3/4 3-8 Insert C Insert C 3/4 3-9 3/4 3-9 l Insert D Insert D 3/4 3-14 3/4 3-14 1 3/4 3-20 3/4 3-20 i 3/4 3-21 3/4 3-21 3/4 3-22 3/4 3-22 3/4 3-23 ,
3/4 3-25 3/4 3-26 3/4 3-26 Insert E Insert E 3/4 3-27 3/4 3-27 3/4 3-27(a) 3/4 3-27(a) 3/4 3-32 3/4 3-32 3/4 3-33 3/4 3-33 3/4 3-34 3/4 3-34 3/4 3-35 3/4 3-35 3/4 3-36 3/4 3-36 3/4 3-38 3/4 3-38 3/4 3-39 3/4 3-39 3/4 3-41 3/4 3-41 3/4 3-46 3/4 3-46 3/4 3-47 3/4 3-47 3/4 3-49 3/4 3-49 3/4 3-50 3/4 3-50 Insert F Insert F 3/4 3-52 3/4 3-52 ki\nla\1asa110\actatil.wpf56 9412270176 941214 PDR ADCCK 05000373 P PDR
ATTACHMENT B PROPOSED CHANGES TO THE TECHNICAL SPECIFICATIONS FOR OPERATING LICENSES NPF-11 AND NPF-18 NPF-11 NPF-18 3/4 3-54 3/4 3-55
.3/4 3-55 3/4 3-56 Insert G Insert G 3/4 3-57 3/4 3-58 Insert H Insert H 3/4 3-59 3/4 3-59 3/4 3-86 3/4 3-86 Inserts I and J Inserts I and J 3/4 3-87 3/4 3-87 Insert K Insert K .
3/4 3-89 3/4 3-89 B 3/4 3-1 B 3/4 3-1 Inserts L and M Inserts L and M B 3/4 3-2 B 3/4 3-2 Inserts N and O Inserts N and O B 3/4 3-3 B 3/4 3-3 B 3/4 3-3a B 3/4 3-3a Insert P Insert P B 3/4 3-4 B 3/4 3-4 Inserts Q, R, and S Inserts Q, R, and S B 3/4 3-6 B 3/4 3-6 Insert T Insert T i
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, . 2.2 LIMITING SAFETY SYSTEM SETTINGS J l
l BASES I
l 2.2.1 REACTOR PROTECTION SYSTEM INSTRUMENTATION SETPOINTS 1
The Reactor Protection System instrumentation setpoints specified in '
j Table 2.2.1-1 are the values-at which the reactor trips are set for each i
- parameter. The Trip Setpoints have been selected to ensure that the reactor i 4 core and reactor coolant system are prevented from exceeding their Safety !
Limits during normal operation and design basis anticipated operational occur- l rences and to assist in mitigating the consequences of accidents. Operation with a trip set less conservative than its Trip Setpoint but within its speci-fied Allowable Value is acceptable on the basis that the difference between !
i each Trip Setpoint and the Allowable Value is equal to or less than the drift allowance assumed for each trip in the safety analyses. l l
- 1. Intermediate Range Monitor. Neutron Flux - High The IRM system consists of 8 chambers, 4 in each of the reactor trip systems. The IRM is a 5 decade 10 range instrument. The trip setpoint of 120 divisions of scale is active in each of the 10 ranges. .Thus s', the IRM is ranged up to accommodate the increase in power level, the trip setpoint is also ranged up. The IRM instruments provide for overlap wich both the APRM ,
4 and SRM systems.
i The most significant source of reactivity changes during the power increase !
is due to control rod withdrawal. In order to ensure that the IRM provides the required protection, a range of rod withdrawal v::idents have been analyzed. l The results of these analyses are in Section 15.4.1.2 of the FSAR. The most 1
severe case involves an initial condition in which THERMAL POWER is at approxi- !
mately 1% of RATED THERMAL POWER. Additional conservatism was taken in this l analysis by assuming the IRM channel closest to the control rod being withdrawn i is bypassed. The results of this analysis show that the reactor is shutdown !
and peak power is limited to 1% of RATED THERMAL POWER with the peak fuel enthalpy well below the fuel failure threshold of 170 cal /gm. Based on this ,
analysis, the IRM provides protection against local control rod errors and :
continuous withdrawal of control rods in' sequence and provides backup pro- l tection for the APRM.
- 2. Average Power Range Monitor For operation at low pressure w a low flow during STARTUP, the APRM scram ;
setting of 15% of RATED THEIB4AL POWER provides adequate thermal 6.argin between the setpoint and the Safety Limits. The margin accommodates the anticipated ;
maneuvers associated with power plant startup. Effects of increasing pressure l' at zero or hw void content.are minor and cold water from sources available Tem :
during startup is not much colder than that already in the system. coefficients areI M RWM. Of all the possible sources of reactivity input, uniform control rod Because l
withdrawal is the most probable cause of significant power increase. l r
i LA SALLE - UNIT 1 B 2-9 t
rNFO OM1- No CHAN GO l LIMITING SAFETY SYSTEM SETTINGS ,
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8ASES .. I REACTOR PROTECTION SYSTEM INSTRUMENTATION SETPOINTS (Continued)
Average Power Range Monitor (Continued) the flux distribution associated with uniform rod withdrawals does not involve high local peaks and because several rods must be moved to change power by a i significant amount, the rate of power rise is very slow. Generally the heat flux is in near equilibrium with the fission rate. In an assumed uniform rod withdrawal approach to the trip level, the rate of power rise is not more than 5% of RATED THERMAL POWER per minute and the APRM system would be more than I adequate to assure shutdown before the power could exceed the Safety Limit. l The 15% neutron flux trip remains active until the mode switch is placed in i the Run position.
The APRM trip system is calibrated using heat balance data taken during steady state conditions. Fission chambers provide the basic input to the system and therefore the monitors respond dir2ctly and quickly to changes due to transient operation for the case of the Fixed Neutron Flux-High 1185 setpoint; i.e, for a power increase, the THERMAL POWER of the fuel will be less than that indicated by the neutron flux due to the time constants of the heat transfor associated with the fuel. For the Flow Biased Simulated Thermal Power-Upscale setpoint, a time constant of 611 seconds is introduced into the flow biased APRM in order to simulate the fuel thermal transient characteristics.
A more conservative maximum value is used for the flow biased setpoint as shown in Table 2.2.1-1.
The APRM setpoints were selected to provide adequate margin for the Safety Limits and yet allow operating margin that reduces the possibility of unnecessary shutdown. The flow referenced trip setpoint must be adjusted by the specified formula in Specification 3.2.2 in order to maintain these margins when MFLPD is greater than or equal to FRTP.
- 3. Reactor Vessel Steam Dome Pressure-High High pressure in the nuclear system could cause a rupture to the nuclear syv cm process barrier resulting in the release of fission products. A pressure increase while operating will also tend to increase the power of the reactor by compressing voids thus adding reactivity. The trip will quickly reduce the neutron flux, counteracting the pressure increase. The trip setting is slightly higher than the operating pressure to normit normal operation without spurious trips. The setting provides for a wide margin to the maximum allowable design pressure and takes into account the location of the preMure measurement compared to the highest pressure that occurs in the syssa during a transient.
This trip setpoint is effective at low power / flow conditions when the turbine stop valve closure trip is bypassed. For a turbine trip under these conditions, the transient analysis indicated an adequate margin to the thermal hydraulic limit.
l LA SALLE - UNIT 1 B 2-10 e _ _ . _ - _ _ _ ._- ._ - _ _ - - - _ - - - - -_ - - _ - - -
0 3/4.3 IhSTRUMENTATION 3/4.3.1 REACTOR PROTECTION SYSTEM INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.1 As a minimum, the reactor protection system instrumentation channels -
shown in Table 3.3.1-1 shall be OPERABLE with the REACTOR PROTECTION SYSTEM RESPONSE TIME as shown in Table 3.3.1-2.
APPLICABILITY: As shown in Table 3.3.1-1. MS{
80 TION:
[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 channels and/or trip system in the tripped condition within I hour. l l
- b. With the number of OPERABLE channels less than required by the Minimum I OPERABLE Channels per Trip System requirement for both trip systems, place at least one trip system in the tripped condition within I hour and take the ACTION required by Table 3.3.1-1.
SURVEILLANCE RE0VIREMENTS 4.3.1.1 Each reactor protection system instrumentation channel shall be demonstrated OPERABLE by the performance of the CHANNEL CHF.CK, CHANNEL FUNCTIONAL TEST and CHANNEL CALIBRATION operations for the OPERATIONAL CONDITIONS and at the frequencies shown in Table 4.3.1.1-1. l 4.3.1.2 LOGIC SYSTEM FUNCTIONAL TESTS and simulated automatic operation of all channels shall be performed at least once per 18 months. l 4.3.1.3 The REACTOR PROTECTION SYSTEM RESPONSE TIME of each reactor trip functional unit shown in Table 3.3.1-2 shall be demonstrated to be within its ;
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 reactor trip system.
%W.5EK7'
- nnel per trip system, an inoperable.
/*Withadesknprovidingonlyonec 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 L
be restored to OPERABLE status within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> or the ACTION required by Table 3.3.1-1 for that Trip Function shall be taken.
- If more channels are inoperable in one trip system than in the other, select -
(thattripsystemtoplaceinthetrippedcondition,exceptwhenthiswoul cause the Trip Functi_on to occur. , _
LA SALLE - UNIT 1 3/4 3-1 Amendment No. 94
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i ATTACHMENT B PROPOSED CHANGES TO THE TECHNICAL !
l SPECIFICATIONS FOR OPERATING LICENSES NPF-11 AND NPF-18 1 1
INSERT A
- a. With one channel required by Table 3.3.1-1 inoperable in one or more Functional Units, place the inoperable channel and/or that trip system in the tripped condition
- within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. ,
- b. With two or more channels required by Table 3.3.1-1 inoperable in one or more Functional Units:
- 1. Within one hour, verify sufficient channels remain OPERABLE or tripped
- to maintain trip capability in the Functional Unit, and
- 2. Within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, place the inoperable channel (s) in one trip
- aystem and/or that trip system ** in the tripped condition *,
and
- 3. Within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, restore the inoperable channels in the other trip system to an OPERABLE status or tripped *.
- c. Otherwise, take the ACTION required by Table 3.3.1-1 for the Functional Unit. 1 INSERT B An inoperable channel or trip system need not be placed in the tripped condition where this would cause the Trip Function to occur. In these cases, if the inoperable channel is not restored i to OPERABLE status within the required time, the ACTION required by Table 3.3.1-1 for the Functional Unit shall be taken. ,
This ACTION applies to that trip system with the most inoperable l channels; if both trip systems have the same number of inoperable channels, the ACTION can be applied to either trip system.
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TABLE 3.3.1-1 Y -
u,
_ REACTOR PROTECTION SYSTEM INSTRUMENTATION N
p, APPLICABLE MINIMUM OPERABLE -
e FUtlCllGilAL UNIT OPLRATIONAL CONDITIONS CilANNELS PER TRIP SYSTEM (a) ACTION W
5*
- 1. !.termediate Range Monitors: k
.. Neutron Flux - liigh 2 3 1 y
O 3, 4 2 2 5(b) 3 3 3 ,
li. Inoperative Q
2 3 1 3, 4 2 2 5 3 3 g
- 2. Average Power Range Monitor: IC} I m a. Neutron Flux - Illgh, Setdown 2 2 g 3 2 1
2 ,
y 5(b) 2 3 r* h. Flow Riased Simulated Thermal Q Power-Upscale -
1 . 2 4
- c. Fixed Neutron Flux-High 1 2 -
4
- d. Inoperative 1, 2 3 2 2 1 2
1 5 2 3
- 3. Reactor Vessel Steam Dome l'ressure - High 1,2(d) b 2 1 4
- 4. Re.ictor Vessel Water Level - Low, level 3 1, 2
- 2 1
- 5. Itain Steam Line Isolation Valve - .'
Closure 1g') 4 4
- 6. H.iin Steam Line Radiation - ~
liigh 1,2(d) 2 5 -
__, -,___.e _ _ - - - - - - - - - - - - - ^ ~ ^ '
TA8LE 3.3.1-1 (Continued) 5 v.
REACTOR PROTECTION SYSTEM INSTRUNENTATION O .
l; APPLICA8tE MINIMUM OPERA 8LE
. OPERATIDMAL CHANNELS PER e FUNCTIONAL UNIT COM0!TIONS TRIP SYSTEM (a) ACTION
- 7. Primary Containment Pressure - High 1, 2 III 2 III 1
- 8. Scram Olscharge Volume Water Level - High 1 2 5(hj, 2 3 1
y
- 9. Turbine Stop Valve - Closure I III -
4 III 6 O
- 10. Turbine Control Valve Fast Closure, Valve Trip System Oil Pressure - Low I III 2 III 6
$ 11. Reactor Mode Switch Shutdown - N w Position 1, 2 1 1 4 3, 4 1 7 l
5 1 3
- 12. Manual Scram 1, 2 1 1 Q 3, 4 1 8 5 1 9 Q
- 13. Cdntrol Rod Drive
- a. Charging Water Header 2 2 1 Pressure - Low 5(h) 2
- b. Delay Timer 2 3
1 Q
2(h) 5 2 3 Q
^
f
, . 3 t -
a
.if i
ZN Pc) ONLY No CHANGES -
, REACTOR PhuTEGH ON SYSTEM INSTRUMENTATION .
ACTION ACTION 1 -
Se in at least HDT SHUTDOW within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
ACTION 2 -
Verify all insertable control rods to be inserted in the core and lock the reactor mode switch in the Shutdown position within one hour.
ACTION 3 -
Suspend all operations involving CORE ALTERATIONS
- and insert all insertable control rods within one hour. **
ACTION 4 -
Se in at least STARTUP within 5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />.
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ACTION 5 -
Se in STARTUP with the mair, steam line isolatien' valves closed within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> or in at least NOT SWTDOWN within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
ACTION 6 -
Initiate a reduction in THERMAL POWER within 15 minutes and reduce turbine first stage pressure to $ 140 psig, equivalent l to THERMAL POWER 1ess thsn 3GK of RATED THERMAL POWER, within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />. '
ACTION 7 -
Verify all insertable control rods to be insertad within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.
l ACTION 8 -
Lock the reactor mode switch in the Shutdown position within I hour.
g- !
ACTION 9 -
Suspend all operations involving CORE ALTERATIONS,* and insert ,
all insertable control rods and lock the reactor mode switch in the SHUTDOWN position within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />. !
g ;
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- Except movement of IRM, SAN or special movable detectors, or replacement of '
LPRM strings provided 53M instrumentation is OPERABLE per Specification 3.9.2. i
. I LA SALLE - UNIT 1 3/4 3-4 Amendment No. 18 t
,w. -
TABLE 3.3.1-1 (Continued) ,
REACTOR PROTECTION SYSTEM IN$TRUMENTATION j TABLE NOTATIONS (a) A channel may be placed in an inoperable status for up hours for required surveillance without placing the channel in the tripped ]
condition provided at least one OPERA 8LE channel in the same trip systae is monitoring that parameter.
(b) The " shorting links" shall be removed from the RPS circuitry prior to and l during the time sty control red 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 LPM inputs per i level er less than 14 LPRM inputs to an APRM channel.
(d) This function is not required to be OPERABLE when the reacter pressure vessel head is unbelted or removed per Specification 3.10.L (e) This function shall be autaastically bypassed when the reacter sede switch is not in the Run position.
(f) This function is not required to be OPERA 8LE when PRIMARY CONTAIfeeff l INTEGRITY is not required. !
(g) Also actuates the staney gas treatment systan.
(h) With ary centrol red withdrawn. Net applicable to control rods removed per Specification 3.9.10.1 er 3.9.10.2.
(1) This function shall be autamatically bypassed when turbine first stage pressure is < 140 psig, equivalent to THERMAL POWER 1ess than SOE of RATEDTHERMAIPOWER.
(j) Also actuates the IDC-RPT system.
"Not requires for control rods removed per ifications 3.9.10.1 er 3.9.10.2.
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LA SALLE - UNIT 1 3/4 3-5 Amerumment No. 30 i
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, TA8tE 3.3.1-2 y REACTOR PROTECTION SYSTEM RESPONSE TIMES i
l-m ,
e RESPONSE TIME g FIF.TIONAL UNIT (Seconds)
Q l
- a. Neutron Flux - High* NA k
y
- b. Inoperative NA
- 2. Average Power Range Monitor *
- a.
b.
Neutron Flux - High, Setdown Flow 8tased Simulated Thermal Power-Upscale MA .. 9
< 0.09
- c. Fixed Neutren Flux - High < 0.09 Q
- d. Inoperative liA -
w 3. Reactor Vessel Steam Dome Pressure - High < 0.55 1 4. Reactor Vessel Water Level - Low, Level 3 7 1.05 '
w 5. Main Steam Line Isolation Valve - Closure 7 0.06 N J. 6. Main Steam Line Radiation - High liA $
- 7. Primary Contalnment Pressure - High NA
- 8. Scram Discharge Volume Water tevel - High NA Q
- 9. Turbine Stop Valve - Closure -< 0.06
Trip 011 Pressure - Low < 0.08, @
- 11. Reactor Mode Swltch Shutdown PosILIon lin
- 12. Manual Scram NA
~
A !
- 13. Control Rod Drive
- a. Charging Water Header Pressure - Low g
NA
- b. Delay Timer NA D
~
{
s
" 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.
- l s **Not including simulated thermal power ting constant.
- Measured from start of turbine control valve fast closure.
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TABLE 4.3.1.1-1 REACTOR PROTECTION SYSTEN INSTRUNENTATION SURVEllLANCE REQUIREMENTS E CHANNEL OPERATIONAL i~ FUNCTIONAL CHANNEL CONDITIONS FOR WNICH
'" CHANNEL TEST CAllBRATION'*' SURVEILLANCE REOUIRED
. CHECK FUNCTIONAL UNIT U intemediate Range Monitors g
l.
- a. Neutron Flux - High S/U,5 S/U,W R 2 5 W R 3,4,5 NA 2, 3 [, 4, 5 Inoperative NA W b.
- 2. Average Power Range Monitor:"'
- a. Neutron Flux - High, S /U ,5 S/U, W SA b
Setdown 3 5 5 W SA 48
- b. Flow Blased Simulated Thermal S/U'*',
n W'* , SA , R
Power-Upscale S. D"' FQ) '
l m
' c. Fixed Neutron Flux - U, W'* , SA 1 High S
(' d. Inoperative NA # NA 1, 2, 3, 5
- 3. Reactor Vessel Steam Dome 1, 2 AQ Pressure - High NA Q
- 4. Reactor Vessel Water Level - 1, 2 Low, Level 3 5 #Q R
- 5. Main Steam Line Isolation R 1 l Valve - Closure NA Q
- 6. Main Steam Line Radiation - 1, 2 3E g High 5 ArQ l R 1
- 7. Primary Containment Pressure - 1, 2 M
High NA p@ q
[
to G F1 L . _ . . - _ . _ . _ _ - - _ _ _ _ - . - . . . - - - . _ _ _ - _ _ - - - - - - _ - - - . - - _ _ _ - _
--r - ,-
1ABIE 4.3.1.1-1 (Continued)
, ,- REACTOR.PR0l[Cil0N SYSTEM INSTRUMENTAll0N SURVEILLANCE REQUIREMENTS
'l.' tilANNEL OPERAll0NAL
- , CilANN[l IUNCT10NAL CllANNEL CONDITIONS FOR WillCH IUNC110NAL UNil . l.Ill t.K TEST CAllBRATION SURVElllANCE REQUIRED-SE 8. Scram Discharge Volume Water G Level - liigh ~NA 6 R 1, 2, 5 I
_. 9. Turbine Stop Valve - Closure NA R 1
- 10. Turbine Control Valve fast Closure Valve Trip System Oil I I Pressure - Low NA Q R
- 11. Reactor Mode Switch Shutdown Position NA R NA 1,2,3,4,5
- 12. Manual Scram NA W NA 1,2,3,4,5 l
- 13. Control Rod Drive
- a. Charging Water Header Pressure - Low NA M R 2, 5 w b. Delay limer NA M R 2, 5 A
4'
=
(a) Neutron detectors may be excluded from CHANNEL Call 8 RATION.
(b) 1he IRMg*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) 1his calibration shall consist of the adjustment of the APRM channel to conform to the power levels calculated by a heat balance during OPERATIONAL CONDITION 1 when THERMAL POWER 2 25% of RATED THERMAL .
The APRM Gain Adjustment factor (GAF) for any channel shall be equal to the power value deter-POWER.
mined by the heat balance divided by the APRM reading for that channel.
2 Within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, adjust any APRM channel with a GAF > 1.02. In addition, adjust any APRM channel within
$ 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, (1) if power is greater than or equal to 90% of RAl[D THERMAL POWER and the APRM channel GAF is Q < 0.98, or (2) if power is less than 90% of RATED THERMAL POWER and the APRM reading exceeds the power ,
M value determined by the heat balance by more than 10% of RATED THERMAL POWER. Until any required APRM 5 adjustment has been accomplished, notification shall be posted on the reactor control panel.
g (e) This calibration shall consist of the adjustment of the APRM flow biased channel to conform to a __
n calibrated flow signal.
(f) The LPRMs shall he calibrated at least once per 1000 effective full power hours (EIPil). .
NSERT 'C" (g) Measure and compare core Ilow Io rated core iInw.
(h) 1his calibration shall consist of ven if ying the 6
- I second simulated thermal power time constant.
I
1 ATTACHMENT B PROPOSED CHANGES TO THE TECHNICAL SPECIFICATIONS FOR OPERATING LICENSES NPF-11 AND NPF-18 INSERT C
- The provisions of Specification 4.0.4 are not applicable for a period of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after entering OPERATIONAL CONDITION 2 or 3 when shutting down from OPERATIONAL CONDITION 1. '
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INSTRUMENTATION 3/4.3.2 ISOLATION ACTUATION INSTRUMENTATION LIMITING CONDITION FOR OPERATION 9
3.3.2 The isolation actuation instrumentation channels shown in Table 3.3.2 shall be OPERABLE with their trip setpoints set consistent with the values shown in the Trip Setpoint column of Table 3.3.2-2 and with ISOLATION SYSTEM RESPONSE TIME as shown in Table 3.3.2-3.
APPLICABILITY: As shown in Table 3.3.2-1.
ACTION:
- a. 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.
- b. 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 trip system in the tripped con-dition within one hour. l
- c. With the number of OPERABLE channels less than required by the Minimum l OPERABLE Channels per Trip Syst,er eequirement for both tr,jp systems, place at least one trip system ... the tripped condition within ne hour and take the ACTION required by Table 3.3.2-1.
GRT "b" i
- 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 ours or the ACTION required by Table 3.3.2-1 for that Trip Function shalT be taken.
"If more channels are inoperable in one trip system than in the other, select that trip system to place in the tripped condition except when this would cause the Trip Function to occur.
"*An inoperable channel need not be placed in the t ir pped condition where this would cause the Trip Function to occur. In these cases, the inoperable channel shall be restored to OPERABLE status within I hour or the ACTION required by .
Table 3.3.2-1 for that Trip Function shall be taken.
LA SALLE - UNIT 1 3/4 3-9 Amendment No. 94 i
ATTACHMENT B PROPOSED CHANGES TO THE TECHNICAL SPECIFICATIONS FOR OPERATING LICENSES NPF-11 AND NPF-18 INSERT D
- b. With the number of OPERABLE channels less than required by the Minimum OPERABLE Channels per Trip System Requirement for one trip system, either
- 1. Plc.ce the inoperable channel (s) and/or trip system in the tripped condition
- Within a) 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> for trip functions without an OPERABLE channel, 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 c) 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
- 2. Take the ACTION required by Table 3.3.2-1.
t
- c. With the number of OPERABLE channels less than required by the Minimum OPERABLE Channels per Trip System requirement for both i trip systems, *
- 1. Place at least one trip system ** in the tripped condition ***
within one hour, and
- 2. a) Place the inoperable channel (s) in the remaining trip system in the tripped condition *** within
- 1) 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> for trip functions without an OPERABLE channel,
- 2) 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
- 3) 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 b) Take the ACTION required by Table 3.3.2-1.
1 4
kt\nlaslasalle\eotstii wpf60 4
. . - -a-.. . . - c- =
T*BLE 3.3.2-1 (Continued)
.u.
wrTMn ISOLATION ACTUATION INSTRUMENTATION ACTION ACTION 20 -
. Se ACTION 21 -
nthe atnext least 24NOT hours.SHUTDOWN within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD Se in at least STARTUP with the associated isolation valve closed within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> or be in at least NOT SHUTDOWN w 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTDOWN within the next ACTION 22 -
24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
Close the affected system isolation valves within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> and ACTION 23 - declare the affected. system inoperable.
ACTION 24 - Se in at least STARTUP within 5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />. "
ACTION 25 -
treatment system operating within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />. Establish S Lock the affected system isolation valves closed within I hour ACTION 26 - and declare the affected system inoperable. !
Provided each other that the manual initiation function is OPERABLE each line, group valve, inboard or outboard, as applicable, in ,
restore the manual initiation function to OPERA 8LE '
function to 0PERABLE status within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />; othe a.
Se in$NLTTDOWN COLD at least HOT withinSHUTDOWN the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, within or the n' ext 12 h
- b. i Close the affected system isolation valves within the next hour and declare the affected system in operable.
g ..
May be bypassed stop valves closed. with reactor steam pressure < 1043 psig and all turbin ~
When handling irradiated fuel in the secondary containment and duri i
- ALTERATIONS and operations with a potential for draining the reacto During reactor vessel. CORE . ALTERATIONS and operations with a potential.for dra (a)
. (b) See Specification 3.5.3. Table 3.6.3-1 for valves in each lve group. !
- A channel may be placed in an inoperable status for up to urs for required surveillance without placing the channel in the trippea -
system is monitoring that parameter. condition provi l i In addition for those trip systems '
with a design providing only one channel per trip system, the channel m be placed in an inoperable status for up to 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> for required surveillan testing without placing the channel in the tripped condition provided that the redundant isolation valve, inboard or outboard, as applicable in each ,
(c) Alsodant actuates valve the is OPERABLE, standby gas treatment or systas.
place the trip system in .
(d) I (e)
Also actuates secondary containment ventilation iso
- Table 3.8.5.2-1.
(f) Closes only RWCU system inlet outboard valve. I LA SALLE - UNIT 1 3/4 3-14 Amendment No. 26
,-----,---n . , . - . - --- , - - - - - -. .--_m . ~ - - - ,,,-.-----,-------_--.n.._.,n.. ,_, ---,-,--. a--.-- -----,--n
TABLE 4 _1_ ,
ISOLATION ACTUATION IltSTRtspTATION SURVEILLAIEE REllWIRDUTS '
g CMf1NEL* OPERATISML '.
l CMMMEL FifMCTIO M L '
CHAleIEL COMITIONS FOR 1RilCN h
TRIP FUNCTION _ CMcK TEST CALIBRATICII SURVEILLAIEE MeulMB j E A. AUTomTIC INITIATION I
- 1. PRIMARY CONTAra9 H T ISOLAT10ft
- s. Reacter Vessel Water Level i
" 1) Low, Level 3
- 2) Low Low, Level 2 4Q R 1, 2, 3
-W 9 R 1, 2, 3
! 3) Low Low Low, Level 1 , 4f' Q R 1, 2, 3 l l
i b.
c.
Drywell Pressure - Nigh Nefn Steam Line M4 fMQ T
q 1,2,3
- 1) Radfatten - High 5 4rQ '
R 1, 2, 3
- 2) Pressure - Low
- 3) Flow - MIgh M4 j 4F 4tQQ i 4 1 d.
10 4 j R 1,2,3 Nefn Steam Line Tunnel 1 Temperature - Nigh 104 R w
2
- e. Condenser Vacuum - Low IIA h> MQ d q 1, 2 3 1e 2E , 3*
- f. Main Steam Line Tunnel y ,, a Temperature - Nigh ,
10 4 --W G R 1, 2, 3
- l U 2. SEC0084RY CONTAllelENT ISOLATI0li
- a. Reacter ButIdfng Vent Exhaust b.
Plenum Radiatten - Nigh Drywell Pressure - Nigh 5 XQ R 1, 2, 3 and **
- c. Reacter Vessel Water M4 ' fat-Q q -
1,2,3 XQ i
Level - Low Low, Level 2 M4 R 1, 2, 3, and I '
- d. Fuel Peel Vent Exhaust Radiatten - High ' ~ ,JP"ki 5 R 1, 2, 3 and **
- i
- 3. REACTOR M4TER CLEAIIUP SYSTEM ISOLATICII '
A Flow - Nigh
- a. 5 t XQ 1 -
R 1, 2, 3 k" -
- b. Heat Exchanger Area c.
Temperature - Nigh Heat Exchanger Area 10 4 l
Tk 4 1,2,3 P. Ventilatten AT - Nfgh IIA I
.4t~ k 4 1,2,3 -
g d. SLCS laItIatIen 11 4 NA 1, 2, 3
. e. Reacter Vessel Water
= Level - Low Low, Level 2 NA .WQ R 1, 2, 3 l
1
I e-Q .-
t i
G ISOLATION ACTUATION IN5flRMENTAileN $1RtVEILLANCE REQtilREfENTS i CHANNEL CHANNEL OPERATIONAL TRIP (UNCTION - FUNCil0NAL CHANNEL
_ CHECK CONDITIONS FOR WHICH e YEST CALIORATION SURVEILLANCE REQUIRE 8 4.
REACTOR CORE 150 tail 0N C00LilIG SYSTEM 150LAT10N y a. RCIC Steam line Flow - Nigh 1
' b. RCIC Steam Supply Pressure -
IIA Q tt 1, 2, 3 .
Low i
- c. Ilcic TurtIne Enheest Dfaphrage IIA - XQ I 4 1, 2, 3 ,
Pressure - High
, XQ (}
NA q
- d. RCIC Egefpment Room 1, 2, 3 t
Temperature - High /
4
- e. RCIC Steam Line Tunnel IIA T #k 4 1, 2, 3 Temperature - High
- f. RCIC Steam Line Tunnel 18 4
- k f 4 1,2,3 i
& Temperature - High
- MA #9 q w g. Drywell Pressure - High 1, 2, 3
- Q IIA
) h. RCIC Equipment Room I
q 1, 2, 3 y & Temperature - High U
IIA f JVQ j q 1,2,3 i
- 5. ~
g 1551 SYSTEM STEAM CONDENSING MDOE 150LAT10N
! a. IIHR Equipment Area a b.
Temperature - High ItHR Area Cooler Temperature -
IIA t
- Q q* 1, 2, 3
,i i
c.
High RHR Neat Exchanger Steam IIA 4
- Q q 1, 2, 3 Sepply Flow - High 44 #O q 1, 1, 3 5
i f .
l [
a F . . -
p* . '
~
1 l ' 1
&. ' l TABLE 4.3.2.1-1 (Continued)
G . 1
. ISOLATION ACTUATI0ll Ill5TRt#ENTATI0lt SURVEILLAllCE REQUIREIEllTS p
p, l CHAlelEL OPERATIONAL I
, CHANNEL FullCTIONAL CHAlWIEL CONDITIONS FOR tellCM c TRIP FUNCTION CHECK TEST Call 8 RAT 10ll x SURVEILLABICE REQUIRED ,
- 6. RHR SYSTEM SmlT00tel C00LilIG IRIDE 150LAT10ll
- a. Reacter Vessel Water Level - 1 Low, Level 3 R 1, 2, 3
- b. Reacter Vessel ( l
)
(RHR Cut-in Permissive) c.
Pressure - High RHit pump Section Flow - High M }MI q 1, 2, 3 i
l
- d. RHR Area Teeperature - NI 11 4 NA
'% q.
q 1, 2, 3 i
- e. RHIt Equipe at Area AT - H gh i JttQ 1, 2, 3 l' M4 i q 1,2,3 y B. MANUAL INITIATI0ll -
,, j af 1. Inboard Valves MA R 10 4 1,2,3 y 2. Outboard Valves 10 4 R II4 1,2,3
- 3. Inboard Valves HA R 11 4 1, 2, 3 and **,#
- 4. Outboard Valves MA 1 IIA 1, 2, 3 and **,# l S. Inboard Valves HA h 10 4 1, 2, 3 !
- 6. Outboard Valves N4 R II4 1,2,3
- 7. Outboard Valve Mit G 11 4 1,2,3
When reacter steam pressure > 1943 psig and/or any tuttine step valve is open.
- When handling Irradiated fuel in the secondary containment and during CSE ALTERATICII5 and operattens with a potential for draining the reacter vessel.
l [ FDuring CORE ALTERATIONS and aperattens with a potential for draining the reacter vessel.
- o .
en 5 -
INSTRUNENTATION r 1
(- ' 3/4.3.3 EMERGENCY CORE COOLING SYSTEN ACTUATION INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.3 The emergency core cooling system (ECCS) actuation instrumentation '
l 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 and with EMERGENCY CORE COOLING SYSTEM RESPONSE TIME as shown in Table 3.3.3-3.
APPLICABILITY: As shown in Table 3.3.3-1. .
ACTION:
- a. With an ECCS actuation insteunentation channel trip setpoint less conservative than the value shown in the Allowable Values column of i Table 3.3.3-2, declare the channel inoperable until the channel is .
restored to OPERA 8LE status with its trip setpoint adjusted consistant '
with the Trip Setpoint value.
- b. With one or more ECCS actuation instrumentation channels inoperable, take the ACTION required by Table 3.3.3-1. !
- c. With either ADS trip system "A" or "3" inoperable, restore the inoperable trip system to OPERABLE status within:
j
- 1. 7 days, provided that the NPCS and RCIC systems are OPERA 8LE. . !
- 2. 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 NOT 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 re3ctor steam done pressure to less than or equal to 122 psig within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
i SURVEILLANCE REQUIRENENTS i 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 perfe wed at least once per 18 monthsh l
4.3.3.3 The ECCS RESPONSE TIME of each ECCS trip function shown in Table 3.3.3-3 ;
shall be demonstrated to be within the limit at least once per 18 renths. t Each test shall include at 'teast 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.
l Q urve111ance is performed during Refuel lufor LPCI A, 8"The
,_and c- specified 38-month interi LA5ALLE UNIT-1 _ 3/4 3-23 Amendment No. 24
- . x -
- . g naununnan a nununns n, sss ' '
lII
- [ 5' aa4 5 aaa4 5 eaaaeaea
-E NE l a a r; a a a r; .a saa aaaa \
a : ..
.i a
I 2
s s
. 2 2
c-.!2 \\u, %%- 5a5 I
1c.ss =I"i
=
3 55
~
- !.- - s A
m : -
4 s
E
.5 a
s
- 5g
-s as
- 5 ~
! 5 5 j, 3gE 3 ]g' .
3t} g a .
- - m . 3, ,
- 3,, t - 3c g
)
1 8 2 8: #. o a Ism E E y g 5: se J .3 3 =' L E tses:Gl'sE ! f B 3 , 3s :j . E -E 1 s ! f! !$5f+5 aII!s. I 5 3..?iE! il ! j - 3E II-.1 3 E 3 3 ! .
- : m 31 s3 I.!Ee 3 =-
8 5}3=.lk... }E EElE E j! }=! j i }}} ls!s k A di4 4iaa liA444 asi 3 . A LA SALLE - UNIT 1 3/4 3-25 Amendment 100 23 4 -re - my , y -w+-- -,-,---ev-wy-w --"
TABLE 3.3.3-1(Continued)
- v. EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUENTATION
, ? . A MINIfRM OPERABLE APPLICABLE CHANNELS PER TRIP OPERATIONAL j E TRIP FL'NCTION FUNCTION I *) COMITIONS ACTION N C. DIVISION 3 TRIP SYSTEM
- 1. HPCS SYSTEM
- a. Reactor Vessel Water level - Low, Low, level 2 4 1, 2, 3, 4*, 5* 35
, b. Drywell Pressure - Hi ph 4 1,2,3 35
- c. Reactor Vessel Water .evel-High, level 8 2 1,2,3,4*,5* 32 -
- d. Deleted
- e. Deleted
- f. Pisap Discharp Pressure-Hi (Bypass) 1 1, 2, 3, 4*, 5* 31 w g. HPCe Systen Flow Rate-Low PereTssive) 1 1, 2, 3, 4* , 5* 31 :'
i
) , h. Manual Initiation 1/ division 1, 2, 3, 4*, 5* 34 Y D. LOSS OF POWER MINIftM APPLICABLE M TOTAL NO. INSTitUNENTS OPERABLE OPERATIONAL OF INSTIMENTS TO TRIP INSTIMENTd COMITIONS ACTION
- 1. B 2/ bus 2/ bus 2/ bus 1, 2, 3, 4**, 5** 37 i 4.16 Losskw Eme,Ea N us Undervoltage of Vo1
- 2. .16 kv Emergency Bus Undervoltage 2/ bus 2/ bus 2/ bus 1, 2, 3, 4**, 5** 37 (Degraded Vo tage)
A channel instrument may be laced in an ino duri periods of required J (a) s;rveillance without placi the trip systes/perable status for up tochannel/ instrument in the ition provided at tripped least one other OPERABLE channel nstrument in the same trip system is monitoring that parameter. (b Also actuates the associated division distsel generator. ,
= d Prsvides si 1 to close HPCS pump dische valve only on 2-out-of-21 Ic. g i
n the stem is required to
~
# %c Applicable OPERABLE per Specification . 5. 2 or 3. 5. 3. -
F
** Required when ESF t is required to be OPERABLE.
(
~
# Not required to be LE when reactor steam does pressure is < 122 psig. ,
l
m t ATTACHMENT B PROPOSED CHANGES TO THE TECHNICAL SPECIFICATIONS FOR OPERATING LICENSES NPF-11 AND NPF-18 INSERT E (d) A channel / instrument may be placed in an inoperable status for up to 2 hours during periods of required surveillance without i placing the trip system / channel / instrument in the tripped : condition provided at least one other OPERABLE ! channel / instrument in the same trip system is monitoring that parameter. i b i 1 [ t l I l k:\nla\lasalle\aotatti.wpf61 l
L TABLE 3.3.3-1 (Continued) EMERGENCY CORE C0OLING SYSTEM ACTUATION INSTRUMENTATION j AG.I.105 ACTION 30 - With the number of OPERABLE channels less than required by the Minimum OPERABLE Channels per Trip Function requirement: !
- a. With one channel inoperable, place the inoperable channel in-the tripped condition within ou r declare the l associated system inoperable, yg ,
- b. With more than one channel inoperable, declare the -
associated system inoperable. n g ACTION 31 - With the number of OPERABLE channels less han' required by the Minimum OPERABLE channels per Trip Functio place the r7 inoperable channel in the tripped condition within dKihou'h; restore the inoperable channel to OPERABLE status within 7 days or declare the associated system inoperable. gy M ACTION 32 - With the number of OPERABLE channels less than required by the Minimum OPERABLE Channels per Trip Function requirement declare i the associated ADS trip system or ECCS inoperabl y / % ; t y k J ACTION 33 - With the number of OPERABLE channels less than the Minimum OPERABLE Channels per Trip Function requirement, place the_ 4/ , inoperable channel in the tripped condition within ene nogr. y ACTION 34 - With the number of OPERABLE channels less than required by the Minimum 0PERABLE Channels per Trip Function requirament, restore the inoperable channel to OPERABLE status withinIBZhTurn or declare the associated AUS trip system or ECCS inoperab' e. ACTION 35 - With the number of OPERABLE channels less than required by the ; Minimum OPERABLE Channels per Trip Function requirement
- a. For one trip system, >1 ace that trip system in the tripped condition within dos tou6 or de 1 re the HPCS system l inoperable. e 44.f Q ' '
- b. For both trip systems, declare the HPCS system inoperable.
1 ACTION 36 - Deleted ACTION 37 - With the number of OPERABLE instruments less than the Minimum Operable Instruments, place the inoperable instrument (s) in the tripped condition within I hour or declare the associated l: ' emergency diesel generator inoperable and take the ACTION required by Specification 3.8.1.1 or 3.8.1.2 as appropriate. I 3/4 3-27 Amendment No. 94 LA SALLE - UNIT 1
TABLE 3.3.3-1 (Continued) EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION ACTION ACTION 38 With the number of OPERABLE channels less than required by ' the Minimum OPERA 8LE Channels per trip function requirements:
- a. With one channel inoperable, remove the inoperable channel within ene houP; restore the inoperable channel to OPERABLE 24 M status'within 7 days or declare the associated ECCS systems .
inoperable. ,
- b. With both channels inoperable, restore at least one channel to OPERA 8LE status within one hour or declare the associated ECCS systems inoperable.
t i e i LA SALLE - UNIT 1 3/4 3-27(a) Amendment 10
-\ .
TABLE 4.3.3.1-1 i E - EIERGUICY CORE COOLING SYSTEM ACTETION INSTIMENTATIOlt SURVEILLAllCE REQUIRDENTS l I;; CHAIRIEL OPERATIOML , CMNNEL FUNCTIOML CHANilEL COMITI0lls FOR llNICH l g TRIP FUNCTION CHECK TEST CALIBRATIOll SURVEILLAllCE REQUIRES
]A.DIVISIONITRIPSYSTEM
- 1. IIHR-A (LPCI HDDE) AW LPCS SYSTEM
- a. Reactor Vessel Water Level -
b. Low Low Low, level 1 Drywell Pressure - Nish
#Q R 1, , 3, (*, 5*
- c. LPCS Pump Olscharge Flow-Low M
#9-M W.
0 1, 1 3
, 3, 4*, 5*
Q
- d. LPCSandLPCIAInjectionValve l l InjectionLinePressureLow >
)
- Interlock M JP Q R 1,2,3,4*,5*
i w e. LPCSandLPCIAInjectionValve
- A Reactor Pressure Low Interlock M .XQ / R 1,2,3
- 5*
w fi LPCI Pump A Start Time Delay Relay M WQ l (' 12 *, 5*
- g g. LPCI Pump A Flow-Low M Q ( .
1,, 2,,33 *, 5*
- h. Manual Initiation M IA 1, 2, 3, 4*, 5*
l l 2. AUTOMTIC DEPRESSURIZATI0ld SYSTEM TRIP SYSTEM "A"#
- s. Reactor Vessel Water Level -
! Low Low Low, level 1 R
- b. Drywell Pressure-Nigh W GI I O 1 1,232l~ 3 Initiation Timer X Qt
- c. M Q 1, 2, 3
, d. Reactor Vessel Water Level - t iXQ
^
Low Level 3 .JWS R 1, 2, 3 i [ e. LPCSfumpDischarge fE f i i
- f. LPCIPumpADgcharge Pressure-Ni M XQ 1 Q 1,2,3 ,
- Pressure-Nigh MA f X Q (j q 1, ,3 y
- a. Manual Initiation MA R i IIA 1, ,3
- 5. Drywell Pressure Bypass Timer M XQ g 1 ,3 U l. Manual Inhibit M R llA 1,, ,3 S
. ~
j Y* g TABLE 4.3.3.1-1(Continued) g g, EERGENCY CORE COOLING SYSTEM ACTMTION INSTRIBUTATION SURVE s CHAMEL CHANNEL FUNCTIOEL OPEMTIOEL CHANNEL S TRIP FUNCTION CHECK TEST COMITIONS FOR 1RitCH -
" CALIBRATION SURVEILLANCE REQUIRED
! B. DIVISION 2 TRIP SYSTEM
- 1. RHR B A M C (LPCI MDDE) l
- a. Reactor Vessel Water Level - '
Low Low Low, level 1 9'
- b. D 11 Pressure - High R 1, 2, 3
- c. IBandCInjectionValve ,
-M- Q ' Q 1, 2, 3, 4*, 5*
Injection Line Pressure Low j Interlock l w d. LPCI Pump B Start Time Delay Relay M M WQ R 1, 2, 3, 4*, 5* i s. LPCI Pump Discharge Flow-Low- M i MQ t 1, 2, 3, 4*, 5* w f. Manual Initiation M ( 1, 2, 3, 4*, 5* ' LPCIBandCInjectionValve IA
$ g. 1,2,3,4*,5* l Reactor Pressure Low Interlock M XQ R 1, 2, 3, 4*, 5*
- 2. AUTOMATIC DEPRESSURIZATION SYSTEM TRIP SYSTEM "B"#
t
- a. Reactor Vessel Water Level -
Low Low Low, Level 1 fr 9 R 1,2,3 i b. c. Drywell Pressure-Nigh Initiation Timer
,MQ q 1, 2, 3 M XQ 1,2,3 I
[ d. Reactor Vessel Water Level - Q ow Level 3 M5 f &Q[ .
- e. MumpBandCDischarge R 1,2,3 g Pressure-Hi -
M t ! f. ManualInttIap1on M 0 1 ,3 - ! F h. Drywell Pressure Bypass Timer llA 1, ,3 M -
- 1. Manual Inhibit M R - g 1, 1,
,3
,3 ,
i CL3
~
- - - , . . - - . , - - - . - . . . - , , - , - , - - . . - ~ - - - - + - --.-e.4---m. . - ,--,++,m_.e.- - , - , . ____--.-,---#-. ~,_i.+ w , .
. . - . - - - - - - - - ~ , . , _ . . . - -
3 : - TABLE 4.3.3.1-1(Continued) EfERGUICY CORE COOLING SYSTEM ACTUATION IllSTIMENTATI0lt SURVEILLANCE E s , CH4l#lEl OPERATIOML CHANNEL FullCTIONAL CHAlWlEL COMITIONS FOR )RilCH hTRIPFUNCTION CHECK TEST CALItitATI0lt SURVEILLANCE REQUIRED C. DIVISI0ll 3 TRIP SYSTEM
- 1. HPCS SYSTEM
- a. Reactor Vessel Water Level -
i b. Low Low, Level 2 Drywell Pressure-High ,WQ
#k '
R 1,2,3 1, 2, 3, 4*, 5*
- c. Reactor Vessel Water Level-High Q j /
Level 8 .&-Q
- d. Deleted I i
R 1, 2, 3, 4*, 5* , e. Deleted 3:' f. Pump Discharp Pressure-High M 1(Q 1, 2
- g. HPC5 System Flow Rate-Low m
( , , 5* Y h. Manual Initiation M ( 1 , , , 5* St fA 1 , , , 5* D. LOSS OF POWER
- 1. 4.16 kV Emergency Bus Under-voltage (LossofVoltage M M R , gen, ne
- 2. 4.16 kV Emergency Bus Umer)- 1
- voltage (DegradedVoltage)
M M R l' ,
, , 4**, te y t required to be OPERABLE when reactor steam done pressure is less than er equal to 122 psig.
- the system is required to be OPERABLE after being manually realigned, as appilcable, per -
IficatRon 3.5.2. ~ f E
** 1 red when ESF equipment is required to be OPEllABLE.
! 3F .! 1 e
~
] 1 ,
~
I
INSTRUMENTATION 3/4.3.4 RECIRCULATION PUMP TRIP ACTUATION INSTRUMENTATION ATVS RECIRCULATION PUMP TRIP SYSTEM INSTRUMENTATION LIMITING CONDITION FOR OPERATION l 3.3.4.1 The anticipated transient without scram recirculation pump trip : j (ATWS-RPT) system instrumentation channels shown in Table 3.3.4.1-1 shall be OPERABLE with their trip setpoints set consistent with the values shown in the Trip Setpoint column of Table 3.3.4.1-2. , i APPLICABILITY: OPERATIONAL CONDITION 1. , ACTION-
- a. With an ATVS recirculation pump trip system instrumentation channel <
trip setpoint less conservative than the value shown in the Allowable Values column of Table 3.3.4.1-2, declare the channel inoperable until the channel is restored to OPERABLE status with its trip l setpoint adjusted consistent with the Trip Setpoint value. , i
- b. With the number of OPERABLE channels one less than required by the ,
i Minimum OPERABLE Channels per Trip System requirement for one or both trip systems place,the inopera la channel (s) in the tripped condition within 6 hou f </ fourg
- c. With the' number of OPERABLE channels two or more less than required by the Minimum OPERABLE Channels per Trip System requirement f ov/T trip system and:
- 1. If the inoperable channels consist of one reactor vessel water >
level channel and one reactor vessel pressure channel, place both inoperable channels in the tripped condition within , i or, if this action will initiate a pump trip, declare the trip system inoperable.
- 2. If the inoperable channels include two reactor vessel water j level channels or two reactor vessel pressure channels, declare the trip system inoperable.
- d. With one trip system inoperable, restore the inoperable trip system to OPERABLE status within 72 hours or be in at least STARTUP within the next 6 hours.
- e. With both trip systems inoperable, restore at least one trip system i
to OPERABLE status within 1 hour or be in at least STARTUP within the next 6 hours. SURVEILLANCE REQUIREMENTS .6 4.3.4.1.1 Each ATVS recirculation pump trip system instrumentation channel shall be demonstrated OPERABLE by the performance of the CHANNEL CHECK, CHANNEL FUNCTIONAL TEST and CHANNEL CALIBRATION operations at the frequencies shown in Table 4.3.4.1-1.
- 4.3.4.1.2 LOGIC SYSTEM FUNCTIONAL TESTS and simulated automatic operation of l all channels shall be performed at least once per 18 months.
3/4 3-35 Amendment No. 79 LA SALLE - UNIT 1
.o TABLE 3.3.4.1-1 -
ATWS RECIRCULATION Pulf TRIP SYSTEM INSTRIDENTATION E TRIP FUNCTION NIMINUM OPERABLE C g LS PER TRIP SYS1EM
-E q 1. Reactor Vessel Water Level - 2 l g Low Low, Level 2 .
- 2. Reactor Vessel Pressure-High 2 l d
(*I
- One channel in one trip system may be placed in an inoperable status for up for requiral surveillance provided that all other channels are OPERABLE.
u E, n E l . . l l__ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ -- _ . - - - - - - _ _ _ _ _ _ _ _ _ _ . _ _ _ . _ , _ - _ _ _ _ _ _ . _ _ _
TABLE 3.3.4.1-2 9 ATW5 RECIRCULATION PUMP TRIP SYSTEN INSTR M NTATION SETPOINTS E TRIP ALLOWA8tE e TRIP FUNCTION SETPOINT VALUE E Reactor Vessel, Water Level - 1.
>N i
>- 50 inches *
~ >- 57 inches *
[ Lew Low, Level 2 ~ 2. Q Reactor Vessel Pressure-High 5 1135 psig i 1150 psig 3
" See Bases Figure 83/4 3-1.
p o3 u 3 @ d R. -
%D e
e o eos m -- , .ww
m TABLE 4.3.4.1 9 p ATWS RECIRCULATION PUNP TRIP ACTUATION INSTRISENTATION SURVEILLANCE REQUIRDENTS CHANNEL CHAsseEL FUNCTIONAL CHAfGIEL e TRIP FUNCTION CHECK TEST CALIBRATION e
$ 1. Reactor Vessel Water Level - S gQ g Low Low, Level 2
- 2. Reactor Vessel Pressure - High 5 AQ q l Y '
M 1 f 5, a
?
l l . 1 I
\
O INSTRUMENTATION D,b END-OF-CYCLE RECIRCULATION PUMP TRIP SYSTEM INSTRUMENTATION LIMIU NG CONDITION FOR OPERATION 3.3.4.2 The end-of-cycle recirculation pump trip (EOC-RPT) system instrumenta- ' tion channels shown in Table 3.3.4.2-1 shall be OPERABLE with their trip ! setpoints set consistent with the values shown in the Trip Setpoint column of . Table 3.3.4.2-2 and with the E W-0F-CYCLE RECIRCULATION PUMP TRIP SYSTEM RESPONSE : TIME as shown in Table 3.3.4.2-3. APPLICABILITY: OPERATIONAL CONDITION 1, when THERMAL POWER is greater than or . equal to 3(m of RATED THERMAL POWER. , ACTION:
- a. With an end-of-cycle recirculation pump trip system instrumentation channel trip setpoint less conservative than the value shown in the ,
Allowable Values column of Table 3.3.4.2-2, declare the channel inoperable until the channel is restored to OPERABLE status with the channel setpoint adjusted consistent with the Trip Setpoint value.
- b. With the number of OPERABLE channels one less than required by the Minimum DPERABLE Channels per Trip System requirement for one or both trip systems, >1 ace the inoperable _ channel (s) in the tripped
. condition within a mud g2 WE) ~
j
- c. With the number of OPERABLE channels two or more less than required .
by the Minimum OPERABLE Channels per Trip System requirement (s) for ' one trip system and: l
- 1. If the inoperable channels consist of one turbine control valve channel and one turbine stop valve channel uplace both inoperable channels in the tripped condition within g g g Q Q
- 2. If the inoperable channels include two turbine control valve channels or two turbine stop valve channels, declare the trip system inoperable.
d
- d. With one trip system inoperable, restore the inoperable trip systen to OPERABLE status within 72 hours. Otherwise, either:
- 1. Increase the MINIMUM CRITICAL POWER RATIO (MCPR) Limiting Condition for Operation (LCO) to the E0C-RPT inoperable value per Specifica-tion 3.2.3 within the next I hour or,
- 2. Reduce THERMAL POWER to less than 30% of RATED THERMAL POWER with- l in the next 6 hours.
- e. With both trip systems inoperable, restore at least one trip system to OPERABLE status within I hour. Otherwise, either:
LA,SALLE - UNIT 1 3/4 3-39 Amendment No. 58
1 rNFO ONL Y - NO C/W US , 1 INSTRUMENTATION END-OF-CYCLE RECIRCULATION PUMP TRIP SYSTEM INSTRUMENTATION LIMITING CONDITION FOR OPERATION
- 1. Increase the MINIMUM CRITICAL POWER RATIO (MCPR) Limiting Condition for Operation (LCO) to the EOC-RPT inoperable value i per Specification 3.2.3 within the next I hour or,
- 2. reduce THERMAL POWER to less than 30% of RATED THERMAL POWER within the next 6 hours.
SURVEILLANCE REOUIREMENTS 4.3.4.2.1 Each end-of-cycle recirculation pump trip system instrumentation channel shall be demonstrated OPERABLE by the performance of the CHANNEL FUNCTIONAL TEST and CHANNEL CALIBRATION operations at the frequencies shown in ' Table 4.3.4.2.1-1. 4.3.4.2.2 LOGIC SYSTEM FUNCTIONAL TESTS and simulated automatic operation of all channels shall be performed at least once per 18 months. 4.3.4.2.3 The END-OF-CYCLE RECIRCULATION PUMP TRIP SYSTEM RESPONSE TIME of each trip function shown in Table 3.3.4.2-3 shall be demonstrated to be within its limit at least once per 18 months. Each test shall include at least the logic of one type of channel input, turbine control valve fast closure or turbine stop valve closure, such that both types of channel inputs are tested at least once per 36 months. The time allotted for breaker arc suppression shall be verified by test at least once per 60 months. l 1 l LA SALLE - UNIT 1 3/4 3-40 Amendment No. 94
i 4
- E TA8LE 3.3.4.2-1 M
; g
. r- END-OF-CYCLE RECIRCULATION Ptw TRIP SYSTEM INSTRtNNTATION i 5 s MINIfRM .
I , g _ TRIP FUNCTION OPERA 8LECHANNEgg) PER TRIP SYSTEM
- 1. Turbine Step Valve - Closure 2(b)
- 2. Turbine control Valve - Fast Closure 2(b) ,.
t Y b { w I'}A trip system may be placad in an inoperable status for up to provided that the other trip system is OPERA 8tE. hours for required survelliance g (b)This function shall be automatically bypassed when turbine first stage pressure is less then or equ 140 psig, equivalent to THERMAL POWER less than 30% of RATED THEIM4L POWER.
! l .
(
~
g g . y t
T ABli 4.3.4.2.1-1 INU-OF-CYCLE RICIRfU1.All0N PUMP 1 RIP SYSilM SURVElllANCE REQUIREMENTS E
- v. CHANNEL M FUNCTIONAL CHANNEL Ui 1 RIP FUNCTION TEST CAllBRATION
- 1. Turbine stop Valve-Closure I h Q R
] 2. Turbine Control Valve-Fast Closure Q R l
M w m Q Q R . n g ~ 4
'r' '
R k o
. 1, ,y 5
l fMFo ONLY - NO ChMN'GES
~ .
. INSTRUMENTATION 3/4.3.5 REACTOR CORE ISOLATION COOLING SYSTEM ACTUATION INSTRUMENTATION
.s LIMITING CONDITION FOR OPERAT.ON l
3.3.5 The reactor core isolation cooling (RCIC) system actuation instru- j mentation enannels shown in Table 3.3.5-1 shall be OPERABLE with their trip setpoints set consistent with the values shown in the Trip Setpoint column of Table 3.3.5-2. APPLICABILITY: l OPERATIONAL CONDITIONS 1, 2 and 3 with reactor steam done pressure greater than 150 psig. ACTION:
- a. With a RCIC :iystem actuation instrumentation channel trip setpoint less conservative than the value shown in the Allowable Values column of Table 3.3.5-2, de:lare the channel inoperable until the i channel is restored to OPERABLE status with its trip setpoint adjusted consistent with the Trip Setpoint value. ,
i b. With one or more RCIC system actuation instrumentation channels inoperable, I take the ACTION required by Table 3.3.5-1. SURVEILLANCE REQUIREMENTS - t 4.3.5.1 Each RCIC system actuation instrumentation channel shall be demonstrated OPERABLE by the performance of the CHANNEL CHECK, CHANNEL ' FUNCTIONAL TEST and CHANNEL CALIBRATION operations at the frequencies shoen in Table 4.3.5.1-1. 4.3.5.2 LOGIC SYSTEM FUNCTIONAL TESTS and simulated automatic operation of all channels shall be performed at least once per 18 months. ! l I I
. \
LA SALLE - UNIT 1 3/4 3-45
5 TABLE 3.3.5-1 u. N HEACTOR CORE ISOLATION COOLING SYSTEM ACTUATION INSTRUMENTATION ' G
. MINIMUM OPERABLE e TUNCllut.lAL UNITS CHANNELSPE{,)
TRIP SYSTEM ACTION
] a. Iteactor Vessel Water Level - Low Low, level 2 2 50 1,.
!!cactor Vessel Water Level - liigh, Level 8 2(b) 51
- c. . Ibnual Initiation I ICI ' '
, 52 U.
1 ~ oi ~ [ (a) A Ehannel may be placed.in an inoperable status for up to hours for required surveillance without pl. icing the trip system in the tripped cone'ition provided at least one other OPERA 8tE channel In the s.ime trip system is monitoring that paraeeter. - (b) One trip system with two nut of-two logic.
~
(c) Simil e chainiel. ~ e 1 e _ _ -_ _. __ _ _ _ _ _ _ -- e __ ___ __ _ _ _ . _ _ _ . . - - _____ ______ _ _ _ _ _ _ _ _ . _ _ _ _ . _ _ _ _ _ _ _ _ . _ _ _ _ . _ _ _ _ . . _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ . _ _ - . -
TABLE 3.3.5-1 (Continued) REACTOR CORE ISOLATION COOLING SYSTEM ACTUATION INSTRUMENTATION , f ACTION 50 - With the number of OPERAQLE channels less than required by the
- Minimum OPERABLE Channels per Trip System requirement:
l
- a. For one trip system, place the inoperable channel in the tripped condition within o declare the RCIC system inoperable.
g gg
- b. For both trip systems, declam the RCIC system inoperable.
ACTION S1 - With the number of OPERABLE channel less than required by the minimum OPERA 8LE Channels per Trip System requirement, declare theRCICsysteminoperably yyjpg ACTION 52 - With the number of OPERABLE channels less than required by the Minimum OPERABLE Channels per Trio System requi nt, restore the inoperable channel to OPERA 8LE status withi ' hours or ' declare the RCIC system inopersole. ' I t 4 LA 3A..i *,.N T1 3/4 3-47
- . - - - - - - , - - - - - - - ~ - - -
' 5 TABLE 4.3.5.1-1
~
p REACTOR CORE'150tATION C00tlMG SYSTEM ACTMTICII INSTRINE8TsATION SURVEILLANCE R k CHANNEL e CMNNEL FUNCTIONAL CHAISIEL g FUNCTIONAL UNITS CHECK TEST CALIBRATICII
" a. Reactor Vessel Water level -
tow tow, tevel 2 M /Q R l -
- b. Reactor Vessel Water JWIS /k R Level - High, tevel 8 !
- c. Manual Initiation M R M R .
3 W 4 I E e a
. _ - . _ . _ - - - _ - _ _ . _ _ _ - - _ . . - _ _ _ - . _ _ . _ _ _ _ _ _ _ . _ - . _ _ _ _ _ - - - - _ _ _ _ - _ _ -______.__-___& . -_ wv-- -
INSTRUMENTATION 3/4.3.6 CONTROLRODWhTHDRAWALBLOCKINSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.6 The control rod withdrawal block ' instrumentation channels shown in , Table 3.3.6-1 shall be OPERABLE with their trip setpoints set consistent with the values shown in the Trip Setpoint column of Table 3.3.6-2. APPLICABILITY: As shown in Table 3.3.6-1. ACTION: ,
- a. With a control rod withdrawal block instrumentation channel trip setpoint less conservative than the value shown in the Allowable Values column of Table 3.3.6-2, 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 the number of OPERABLE channels less than required by the Minimum OPERABLE Channels per Trip System requirement, take the ACTION required by Taoir 3.3.6-1.
SURVEILLANCE RE0U'REMENTS 4.3.6 Each of the above required control rod withdrawal block trip systems and instrumentation channels shall be demonstrated OPERABLE by the performance of the CHANNEL CHECX, CHANNEL FUNCTIONAL TESThand CHANNEL CALIBRATION operations for the OPERATIONAL CONDITIONS and'at the frequencies shown in Yf Table 4.3.6-1. QusEET n_- "F" . LA SALLE - UNIT 1 3/4 3-50
ATTACHNENT B PROPOSED CHANGES TO THE TECHNICAL SPECIFICATIONS FOR OPERATING LICENSES NPF-11 AND NPF-18 INSERT F A channel may be placed in an inoperable status for up to 6 hours for required surveillance (or 12 hours for repair) without placing the trip system in the tripped condition provided at.least one other OPERABLE channel in the same trip system is monitoring that ' parameter. I t ki\nla\lasalle\actstii.wpf62
O p ' TABLE 3.3.5-1 (Continued)
, CONTROL R00 WITHDRAWAL BLOCK INSTRUMENTATION ACTION ACTION 60 - Declare the RBM inoperable and take the ACTION required by Specification 3.1.4.3.
ACTION 61 - With the number of OPERABLE channels: i
- a. One less than required by the Minimum OPERABLE Channels per Trip '
Function requirement restore the inoperable channel to OPERABLE status within 7 days,or place the inoperable channel in the tripped condition within the next hour. *
- b. Two or more less than required by the Miniaue OPERA 8LE Channels per Trip Function requirement, place at least one inoperable channel in the
- tripped condition within one hour.
ACTION 62 - With the number of OPERABLE Channels less than required by the Minimum OPERABLE Channels per Trip Function requirement, place the inoperable channelinthetrippedconditionwithing sn Q2 A.Q ~ ' With THERMAL POWER 1 30% of RATED THERMAL POWER. . Witn core than one control rod withdrawn. Not a removed per Specification 3.9.10.1 or 3.9.10.2. pplicable to control rods < r
- a. The RBM shall be automatically bypassed when a peripheral control rod is selected. ,
- b. This function shall be automatically bypassed if detector count rate is 1 100 cps or the IRM channels are on range 3 or higher.
- c. This function shall be aut:matica11 channels are on range 8 or higner. y bypassed when the associated IRM
- d. This function shall be automatically bypassed when the IRM channels are on l
range 3 or higher.
- e. This function shall be automatically bypassed when the IRM channels are on range 1.
i LA-5ALLE - UNIT 1 3/4 3-52
$ n
.48
[ TABLE 4.3.5-1
\
5 ' CONillOL ROO WITISRAMAL SLOCK INSTRIBENTATION SURVEILUUICE . wlREBE , CHANNEL CHANNEL OPERATIONAL FUNCTIONAL CfWWRIEL _ TRIP FUNCTICII CO W ITIONS FOR l#IICH __CNECK TEST CALIBRATICIII *I y 1. SURVEILLANCE REQUIRES R00 SLOCK IWIIITOR - H a. Upscale M4 f54 q
- b. Inoperative 10 4 l SM Q 1*
- c. M.A. 1* l Downecale
- 2. AMBE 11 4
'I*'.
- 4 1*'
- n. Flow Blased $1mulated Thermal Power-Upscale 54
- b. Inoperatfwe 11 4
.R Q SA 1 M4 SM N.A.
- c. Downecale IIA 1. 2. 5
, d. Neutron Flux-Migh N4 j
5 4((b).R b).K 1L Q SA SA 1 5 1 3. SOURCE RAflGE NOIIITOR$ *"" Y a. Detector not full fn I M SMI I.W M.A.
- b. Upscale 5 l
4
~
- c. Inoperettve M4 NA SM SM
.W q N.A.
d5 35
- d. Downscale f14 54(b).W l l'
- 4. INTElWIEDIATE RAIIGE NDfffTOR$
.W q 75 i *
- a. Detector not full fn 35 M4 SM(b) .W N.A.
- b. Upscale M4 54 q l
- c. Inoperative .W 75
- . d. Downscale 91 4 M4 SW9).W SM .W N.A.
q Q5r. 5 g
- 5. SCflAN DISCH41tGE VOLINE '
- a. Water Level-Hfgh 10 4 q '
N 1. 2. 5**
. b. Scram Dfscharge Volume f
Swftch In Bypese
=
10 4 (Jr Q N.A. 5** l
- 6. REACT 0lt C00UUIT SYSTEN RECIltCULATI0lt FLOW i i
g a. Upscale 11 4 54 k q
- b. Inoperative 1 g c. Cooperator M4 N4 l SM(b) Q N.A. 1 l
Q q 1 S _ _ _ _ . . - _ - _ . ___-_-_..___.___.__--- - . - _ - - - - - - --.__-- _ - _ _ - -- -,- ,.,-..e.-
,',7 TABLE 4.3.6-1 (Continued) CONTROL R0D WITHDRAWAL B' LOCK INSTRUMENTATION SURVEILLANCE REQUIREME NOTES: '
- a. Neutron detectors may be excluded from CHANNEL CALIBRATION. "
- b. Within 24 hours prior to startup, if not performed within the previous 7 cays.
- c. Includes reactor manual control multiplexing
- system input.
With THERMAL POWER > 30% of RATED THERMAL POWER. With more than one control rod withdrawn. Not applicable to control " rods removed per Specification 3.9.10.1 or 3.9.10.2. . QSGQCT*6" - e 4 l l l LA SALLE - UNIT 1 3/4 3-55 '
o ATTACHNENT B PROPOSED CHANGES TO THE TECHNICAL ; SPECIFICATIONS FOR OPERATING LICENSES NPF-11 AND NPF-18 INSERT G ,
*** The provisions of Specification 4.0.4 are not applicable for a period of 24 hours after entering OPERATIONAL CONDITION 2 or 3 when shutting down from OPERATIONAL' CONDITION 1.
l l l I ks\nla\laea11e\aotstil.wpf63
\
TN FC ML1- Afo CHA W GE5 i 1 INSTRUMENTATION j 3/4.3.7 MONITORING INSTRUMENTATION RADIATION MONITORING INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.7.1 The radiation wnitoring instrumentation channels shown in Table i 3.3.7.1-1 shall be OPEVA E* with their alarm / trip setpoints within the j specified limits. ]l APPLICABILITY: As shown in Table 3.3.7.1-1. I ACTION: )
- a. With a radiation monitoring instrumentation channel alare/ trip setpoint exceeding the value shown in Table 3.3.7.1-1, adjust the setpoint to within the limit within 4 hours or declare the channel inoperable.
1
- b. With one or more radiation monitoring channels inoperable, take the ACTION required by Table 3.3.7.1-1.
- c. The provisions of Specification 3.0.3 are not applicable.
SURVEILLANCE REOUIREMENTS 4.3.7.1 Each of the above required radiation monitoring instrumentation channels shall be demonstrated OPERABLE by the performance of the CHANNEL CHECK, CHANNEL FUNCTIONAL TEST and CHANNEL CALIBRATION operations for the conditions and at the frequencies shown in Table 4.3.7.1-1. l
. l l
l l l 1 l l l l "The normal or emeigency power source may be inoperable in OPERATIONAL CONDITION 4 or 5 or when defueled. LA SALLE - UNIT 1 3/4 3-56 Amendment No. 94
- - _ ______ _________________a
9 g TABLE 3.3.7.1-1 , G m RADIATION MONITORING INSTRUMENTATION
-* MINIMUM CilANNELS APPLICABLE ALARM / TRIP INSThititLHTAT10N OPERABLE MEASUREMENT CONDITIONS _SETPOINT RANGE ACTION
- a. Hsin Control Room 2/ intake 1,2,3,5 and
- 3.5 mR/hr Atmospheric Control 0.1 to 10,000 mR/hr 70 Sr. e s Radiation H.iiteering Subsystem -
7 . e { NOTES
*When irradiated fuel is being handled in the secondary containment. '
mvseer we - i l s
~
i . . . e e _ _ _ _ _ . . _ _ . _ _ _ _ _ _ _ ___ -- .r. . , . - - - - - - . , , ~ - - -,
. .- - _ . . - ,,-e e ..%, . , ,
k 1 ATTACHMENT B PROPOSED CHANGES TO THE TECF.NICAL SPECIFICATIONS FOR OPERATING LICENSES NPF-11 AND NPF-18 i INSERT H
- A channel may be placed in an inoperable status for up to 6 hours for required surveillance testing without placing the Trip System in the tripped condition, provided at least one other operable channel in the same Trip System is monitoring that Trip Function.
i I f i l E i kt\nla\lasalle\actetil.wpf64 l ll
~
.rN Fo wuf - No CWAWGES. i
~ .
TABLE 3.3.7.1-1 (Continued) - RADIATION MONITORING INSTRLMENTATION sits : ACTION 70 -
- a. With one of the required monitors inoperable, place the inoperable channel in the downscale tripped condition within 1 hour; restore the inoperable channel to l OPERAti.E status within 7 days, or, within the next 6 hours, initiate and maintain operation of the control ,.
room emergency filtration system in the pressurization - mode of operation. i
- b. With both of the required monitors ';noperable; initista '
and maintain aparation of tr.a contml room emergency filtration system in the pmssurization mode of operation within 1 hour.
. l i
3 LA SALLE - UNIT 1 3/4 3-58 Amendment No. 18 l -o . . . . . . _ . . . . . .
- i. :
E - i W TABLE 4.3.7.1-1 ' l k
- RADIATION NONITORING INSTRUNENTATION SURVEILLANCE REQUIREMENTS c . -
i3 OPERATIONAL 4 CHANNEL CilANNEL CONDITIONS FOR FUNCTIONAL CHANNEL INSTRUMLNTAT10N _ CIECK WHICH SURVEILLANCE TEST CAllBRATION _ REQUIRED
- a. f t.iln Control Room Atwispheric Control System Radiation .
t lionitoring Sissystem j 5 XQ i g R 1,2,3,5 and,*
~u u .
4 NOTES 4 so When irradiated fuel is being handled in the secondary containment. O h O O e
. I t
- a-
- _, ,,e s _.-.-n.--,- - - - - - -, w- -- ---e n - m - ---- , _ _ ___ - - - - - - - - _
( 1NSTMBENTATION ,
~
3/4.3.8 FEEDWATER/ MAIN TUR8INE TRIP SYSTEM ACTdATION INSTRUMENTATION i LIMITIls COICITION FOR OPERATION 3.3.8 The feedwater/ main turbine trip system actuation instroentation channels shown in Table 3.3.8-1 shall be OPERABLE with their trip setpoints set consistent with the values shown in the Trip Setpoint colan of Table 3.3.8-2. APPLICABILITY: OPERATIONAL CONDITION 1. .n/g7-ACTION: a. channel trip setpoint less conserva tan actuation instruman Withafeedwater/mainturbinetripbwethanthevalueshowninthe Allowable Values column of Table 3.3.8-2 declare the channel
- inoperable and either place the inoperable channel in the tripud (
condition until the channel is restored to OPERA 8tf status wit 1 its I
/. trip setpoint adjusted consistent with the Trip Setpoint value, or declare the associated system inoperable. /
b. With the number of OPERABLE channels one less than required by the ' Minimum OPERABLE Channels per Trip System requirement, restore the inoperable channel to OPERABLE status within 7 days or be in at least STARTUP within the next 6 hours. i c. With the number of OPERABLE channels two less than required by the - Minimum DPERABLE Channels per Trip System requirement, restore at least one of the inoperable channels to OPERABLE status within 72hoursorbein___atleastSTARTUPpinthenext6 hours. y l SURVEILLANCE REQUIREMENTS I 4.3.8.1 Each feedwater/ main turbine trip system actuation instrumentation channel shall be demonstrated OPERABLE by the performance of the CNANNEL CHECK, CHANNEL FUNCTIONAL TEST and CHANNEL CALIBRATION operations at the frequencies shown in Table 4.3.8.1-1. 4.3.8.2 LOGIC SYSTEM FUNCTIONAL TESTS and simulated automatic operation of all channels shall be performed at least once per 18 months.* i O j r (=1he specified 18 month interval may be waived for Cycle I proyided the QsurveillanceisperformedduringRefuel1. LA SALLE - UNIT 1 3/4 3-86 ' Amendment No. 85
ATTACHNENT B i PROPOSED CHANGES TO THE TECHNICAL ! l SPECIFICATIONS FOR OPERATING LICENSES NPF-11 AND NPF-18 l l INSERT I
- a. With a feedwater/ main turbine trip system actuation ;
instrumentation channel trip setpoint less conservative than- i the value shown in the Allowable Values column of Table 3.3.8- ' 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 the number of OPERABLE channels one less than that l required by the Minimum OPERABLE Channels per Trip System i requirement:
- 1. Within 7 days, either place the inoperable channel in the :
tripped
- condition or restore the inoperable channel to '
OPERABLE status.
- 2. Otherwise, be in at least STARTUP within 6 hours. :
- c. With the number of OPERABLE channels two less than required by the Minimum OPERABLE Channels per Trip System requirement:
- 1. Within two hours place or verify at least one inoperable channel in the tripped
- condition, and restore either t inoperable channel to OPERABLE status within 72 hours, or , !
- 2. Be in at least STARTUP within the next 6 hours.
INSERT J An inoperable channel need not be placed in the tripped condition where this would cause the Trip Function to occur.
)
e $ O M bh8 N OC .h eNh
1
. . l l
l l E c 55 E $ I- Q '
- EwS .
E
- E e" 5 Em e
s- gr ' k . 1 e W 4 5 d y $, a
~
9 m 2 e W
- a E ,
3 . R .% .
=
- =.
k 5
= m t C b E 3 .
- m. S e e , ,Q ,
E t s l u 3
) L4 I i
i t W n 5 2 2) e-h - - k E. LA SALLE - UNIT 1 3/4 3.g7 Amendment No. 85
ATTACHNENT B ! t PROPOSED CHANGES TO THE TECHNICAL SPECIFICATIONS FOR OPERATING LICENSES NPF-11 AND NPF-18 i INSERT K A channel may be placed in an inoperable status for up to 6 hours for required surveillance testing without placing the Trip System in the tripped condition. ; 6 f I i I I ( kr\nla\lanalle\aotstii.wpf66
g TELE 3.3.8-2 g E FEEWATER/NAIN TUR81NE TRIP SYSTEM ACTUATION INSTRt#E TRIP FUNCTION ALLOWBLE l E TRIP SETPOINT _ VALUE i y a. Reactor Vessel Water Level-High. Level 8 w < 55.5 inches *
< 56.8 inches
- l see Bases H gure B 3/4 3-1.
i w O Y B %r 4 4
-r k
f n
=
m
.I "
a
. I. . .
e eP en' l a a gpg- ee m a
~
.. \
4 5 8
- I
,s oN i Y ~Y <
ci E
~
e 5 = 3 E m w 5" 5 , 0; - i Di i .
!E l 5 j i W .5 E
E s - l i e -. 5 5% E E 12 k
=
25 LA SALLE - WIT I 3/4 3-89 Amendment No. 85
, -.-,.w
l 3/4.3 1NSTRUMENTATION ! BASES ! l 3 /4.3.1 REACTOR PROTECTION SYSTEM INSTRUMENTATION l The reactor protection system automatically initiates a reactor scram to:
- a. Preserve the integrity of the fuel cladding.
- b. Preserve the integrity of the reactor coolant system.
- c. Minimize the energy which must be adsorbed following a loss-of- l 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 l both trip systems will produce a reactor scram. The system meets the intent : of IEEE-279, 1971, for nuclear oower niant protection systems. Specified. surve_illance intervals for usIV-Closure, say-uosure. m-aa-a: mad the. !
.A< Manual 5craa7have been detemined in accordance w' th NEDC-30851P-A, "Technica Specification Improvement Analyses for BWR Reactor Protection System", March 1981k The bases for the trip settings of the RPS are discussed in the bases ; ~ for Specification 2.2.1. 4 Qyggg7- eg ,
The measurement of response time at the.specified frequencies provides assurance that the protective functions associated with each channel are com-plated within the time limit assumed in the accident analysis. No credit was i taken for those channels with response times indicated as not applicable. Response time may be demonstrated by any series of sequential, overlapping 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 (1) inplace, onsite or offsite test measurements, or l (2) utilizing replacement sensors with certified response times. ; Surve/ Maece. 8.*o( *f *k bMa H e e, ouikJt. b*[e.t
.- ~ ~ ,
T~ " LA SALLE - UNIT 1 B 3/4 3-1 AMENDMENT NO. 95 l 1 l
ATTACHNENT B PROPOSED CHANGES TO THE TECHNICAL SPECIFICATIONS FOR OPERATING LICENSES NPF-11 AND NPF-18 INSERT L and MDE-83-0485 Revision 3, " Technical Specification Improvement Analysis for the Reactor Protection System for LaSalle County Station, Units 1 and 2", April 1991. INSERT M When a channel is placed in an inoperable status solely for performance of required surveillances, entry into LCO and required ACTIONS may be delayed, provided the associated function maintains RPS trip capability. 1 kt\nla\lasalle\aotatil.wpf67
O INSTRUMENTATION BASES I m 3/4.3.2 ISOLATION ACTUATION INSTRUMENTATION This specification ensures the effectiveness of the instrumentation used- - to mitigate the consequences of accidents by prescribing the OPERABILITY trip setpoints and response times for isolation of the reactor systems. When necessary, one channel may be inoperable for brief intervals to conduct required surveillance. Both channels of each trip system for the main steam tunnel ambient tesiperature and ventilation system differential temperature may be placed in an inoperable status for up to 4 hours for required reactor building ventilation system maintenance and testing and 12 hours for the required secondary containment Leak Rate test without placing the trip system in the tripped condition. This will allow for maintainina tie 111mbi11tv of the ventilation system and secondary containment.75ces of tte 3 rip settings
~
may have tolerances explicitly stated where both the high and low values are critical and may have a substantial effect on safety. "he setpoints of other instrumentation, where only the high or low end of the setting h:ve 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 A.C. operated valves, it is assumed that the A.C. I 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. The safety analysis considers an allowable inventory loss l which in turn determines the valve speed in conjunction with the 13 second delay. 3/4.3.3 EMERGENCY CORE COOLING SYSTEM ACTUATION 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, trip setpoints and response times 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. (IN0T"Oh i LA SALLE - UNIT I B 3/4 3-2 AMENDMENT NO. 98
. . . . . .a
ATTACHNENT B PROPOSED CHANGES TO THE TECHNICAL SPECIFICATIONS FOR OPERATING LICENSES NPF-11 AND NPF-18 INSERT N Specified surveillance intervals and surveillance and maintenance outage times have been determined in accordance with NEDC-30851P-A, Supplement 2, " Technical Specification Improvement Analyses for BWR Isolation Instrumentation Common to RPS and ECCS Instrumentation", March 1989, and with NEDC-31677P-A, " Technical Specification Improvement Analysis for BWR Isolation Actuation Instrumentation", July 1990. When a channel is placed in an inoperable status solely for performance of required surveillances, entry into LCO and required ACTIONS may be delayed, provided the associated function maintains primary containment isolation capability. INSERT O Specified surveillance intervals and surveillance and maintenance outage times have been determined in accordance with NEDC-30936P-A,
" Technical Specification Improvement Methodology (With Demonstration for BWR ECCS Actuation Instrumentation)", Parts 1 and 2, December 1988, and RE-025 Revision 1, " Technical Specification Improvement Analysis for the Emergency Core Cooling System Actuation Instrumentation for LaSalle County Station, Units 1 and 2", April 1991. When a channel is placed in an inoperable status solely for performance of required surveillances, entry into LCO and required ACTIONS may be delayed, provided the associated function maintains ECCS initiation capability.
ki\nla\ lana 11e\aotstLi.wpf68
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=
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INSTRUMENTATION , BASES . 3/4.3.4 RECIRCULATION PUMP TRIP ACTUATION INSTRUMENTATION l The anticipated transient without scram system provides a means of limiting the conseq(ATWS) uences of therecirculation unlikely pump trip i occurrence of a failure to scram during an anticipated transient. The
; response of the plant to this postulated event falls within the envelope of I study events in General Electric Company Topical Report NEDD-1034g, dated March 1971 and NED0-24222, dated December, 1979, and Appendix G of the FSAR.
The end-of-cycle recirculation pump trip (E0C-RPT) system is a part of the Reactor Protection System and is an essential safety supplement to the reactor trip. The purpose of the EOC-RPT is to recover the loss of themal margin which occurs at the end-of-cycle. The physical phenomenon invol.ved is ' that the void reactivity feedback due to a pressurization transient can add positive reactivity to the reactor system at a faster rate than the control rods add negative scram reactivity. Each E0C-RPT system trips both recircula-tion pumps, reducing coolant flow in order to reduce the void collapse in the core during two of the most limiting pressurization events. The two events for which the EOC-RPT protective feature will function are closure of the i turbine stop valves and fast closure of the turbine control valves. ' A generic analysis, which provides for continued operation with one or i both trip systems of the E0C-RPT system inoperable, has been perfomed. The analysis determined bounding cycle independent MINIMUM CRITICAL POWER RATIO (MCPR) Limiting Condition for Operation (LCD) values which must be used if the
- EDC-RPT system is inoperable. These values ensure that adequate reactivity margin to the MCPR safety limit exists in the event of the analyzed transient with the RPT function inoperable. The analysis results are further discussed i
, in the' bases for Specification 3.2.3. ) A fast closure sensor from each of two turbine control valves provides input to the EOC-RPT system; a fast closure sensor from each of the other two turbine control valves proviaes input to the second E0C-RPT system. Similarly, a position switch for each of two turbine stop valves provides input to one EOC-RPT system; a position switch from each of the other two stop valves provides input to the other E0C-RPT system. For each EOC-RPT system, i the sensor relay contacts are arranged to fom a 2-out-of-2 logic for the fast j closure of turbine control valves and a 2-out-of-2 logic for the turbine stop ~ l valves. The operation of either logic will actuate the E0C-RPT system and trip both recirculation pumps. Each EOC-RPT system may be manually bypassed by use of a keyswitch which is administrative 1y controlled. The manual bypasses and the automatic Operating Bypass at less than 30% of T THERMAL POWER are annunciated i ~ the control room. y swvel//ancain(e wAf=hce {afaye Mer Specified surveillance intervalbhave been detemined in accordance with the following:
- 1. %DC-30851P-A, " Technical Specification Improvement Analyses for BWR Reactor Protection System", March 1988.
I LA SALLE - UNIT 1 B 3/4 3-3 AMENDMENT NO. 95
,- - -.-. - c- , - ,---- , - - - , - - - - , - - - - - - - - - - - - - - , - - - - . - - - - - - - - - - - - - - - -
INSTRUMENTATION i BASES 3/4.3.4 RECIRCULATION PUMP TRIP ACTUATION INSTRUMENTATION l
- 2. GENE-770-06-1-A,
- Bases for Changes to Surveillance Test Intervals and 1 Allowed Dut-of-service Times for Selected Instrumentation Technical .
i specifications", December 1992. The EOC-RPT system response time is the time assumed in the analysis ! between initiation of valve motion and complete suppression of the electric j arc, i.e., 190 as, less the time allotted for sensor response, i.e., 10 ms, i and less the time allotted for breaker are suppression detemined by test, as correlated to manufacturer's test results, i.e., 83 as, and plant pre-operational test results.
~
L ENS 5RTb I l l I 4 I I LA SALLE - LNIT 1 B 3/4 3-3 a AMENDMENT NO. 95 l
e ATTACHNENT B f PROPOSED CHANGES TO THE TECHNICAL
)
SPECIFICATIONS FOR OPERATING LICENSES NPF-11 AND NPF-18 : l INSERT P : i When a channel is placed in an inoperable status solely for ' performance of required surveillances, entry into LCO and required ACTIONS may be delayed, provided the associated function maintains the applicable RPT initiation capability. l l l l l i l l l i I i ki\nla\lasalle\aotstli.wpf69
i INSTRUMENTATION SASES ! 3/4.3.5 REACTOR CORE ISOLATION COOLING SYSTEM ACTUATION INSTRUMENTATION The reactor core isolation cooling system actuation instrumentation is i j provided to initiate actions to assure adequate core cooling in the event of ' reactor isolation from its primary heat sink and the loss of feedwater flow to i the reactor vessel without providi actuation of any of the emergency core I cooling equipment. __ pgj 7 ,,g ) 3/4.3.6 CONTROL ROD WITHDRAWAL SLOCK INSTRUMENTATION " The control rod block functions are provided consistent with the require-ments of the specifications in Section 3/4.1.4, Control Rod Program Controls and l Section 3/4.2 Power Distribution Limits. The trip logic is arranged so that a ; trip in any one of the inputs will result in a control _ rod block. l 3/4.3.7 MONITORING INSTRUMENTATION 3/4.3.7.1 RADIATION MONITORING INSTRUMENTATION The OPERASILITY of the radiation monitoring instrumentation ensures that; (1) the radiation levels are continually measured in the areas served by. the individual channels, and (2) the alarm or automatic action is initiated when the radiation level trip setp61nt is exceeded. g 7 s3,, pswr , 3.4.3.7.2 SEISMIC MONITORING INSTRUMENTATION The OPERASILITY of the seismic monitoring instrumentation ensures that suffic-ient capability is available to promptly detenmine the magnitude of a seismic event and evaluate the response of those features important to safety. This capability is required to permit comparison of the measured response to that used in the design basis for the unit. This instrumentation is consistent with the recommen-dations of Regulatory Guide 1.12 " Instrumentation for Earthquakes", April 1974. 3/4.3.7.3 METEOROLOGICAL MONITORING INSTRUMENTATION The OPERABILITY of the meteorological monitoring instrumentation ensures that sufficient meteorological data is available for estimating potential radiation doses to the public as a result of routine or accidental release of j radioactive materials to the atmosphere. This capability is required to evaluate i i the need for initiating protective measures to protect the health and safety of l the public. This instrumentation is consistent with the recommendations of !' Regulatory Guide 1.23 "Onsite Meteorological Programs," February,1972. 3/4.3.7.4 REMOTE SHUTDOWN MONITORING INSTRUMENTATION The OPERABILITY of the remote shutdown monitoring instrumentation ensures that sufficient capability is available to permit shutdown and maintenance of - HOT SHUTDOWN of the unit' from locations outside of the control rce. This ; capability is required in the event control room habitability is lost and is consistent with General Design Criteria 19 of 10 CFR 50. j LA SALLE - UNIT 1 B 3/4 3-4 Amendment No. 58 I
--- - - - - - + - - , -
t t 4 i ATTACHMENT'B PROPOSED CHANGES TO THE TECHNICAL SPECIFICATIONS FOR OPERATING LICENSES NPF-11 AND NPF ! INSERT Q ; Specified surveillance intervals and surveillance and maintenance i t outage times have beer.t determined in accordance with GENE-770-06-2-A,
" Addendum To Bases for Changes to Surveillance Test Intervals and >
Allowed Out-of-Service Times for Selected Instrumentation Technical l Specifications (BWR RCIC Instrumentation)", December 1992. When a channel is placed in an inoperable status solely for performance of required surveillances, entry into LCO and required ACTIONS may be delayed, provided the associated function maintains RCIC initiation - capability, i INSERT R , i Specified surveillance intervals and surveillance and maintenance t outage times have been determined in accordance with NEDC-30851P-A, Supplement 1, " Technical Specification Improvement Analysis for BNR Control Rod Block Instrumentation", October 1988, and GENE-770-06-1-A,
" Bases for Changes to Surveillance Test Intervals and Allowed Out-Of-Service Times for Selected Instrumentation Technical Specifications",
December 1992. When a channel is placed in an inoperable status solely for performance of required surveillances, entry into LCO and required ACTIONS may be delayed, provided the associated function maintains Control Rod Block capability. INSERT S Specified surveillance intervals and surveillance and maintenance I outage times have been determined in accordance with GENE-770-06-1-A, j
" Bases for Changes to Surveillance Test Intervals and Allowed Out-Of- i Service Times for Selected Instrumentation Technical Specifications",
December 1992. When a channel is placed in an inoperable status solely for performance of required surveillances, entry into LCO and , required ACTIONS may be delayed, provided the associated function maintains initiation capability. j l
'I keinla\lasalle\aotatii.wpf70 l
l 1 1
l BASES 3/4.3.7.10 DELETED i 3/4.3.7.D EXPLDSIVE GAS DONITORIIE INSTRLBENTATION . This instrissentation provides for monitoring (and controlli )theconcentrations of potentially explosive pas mixtures in the waste pas hal system. , 3/4.3.7. H LODSE-PART DETECTION SYSTEN The DPERABILITY of the loose-part detection system ensures that sufficient capability is available to detect loose metallic parts in the primary system and avoid or mitigate damal of-service times and surve'p to primary system cogonents. The allowable out-11ance requirements are consistent with the recom- - - mandations Primary System of Regulatory Guide 1.133, Reactors."" Loose-Part Detection of Light-Water-Cooled P 3/4.3.E FEEDWATER/NAIN TURBINE TRIP SYSTEN ACTUATION INSTRISENTAT The feedwater/ main turbine trip system actuation instrumentation is - provided to initiate the feedwater systas/ main turtpine trip system in the event i of reactor vessel water level equal to or p ater than the level 8 setpoint associated with a feedwater control?er fai ure to prevent overfilling the reactor vessel which any result in hi safety / relief valve discharge lines. gh pressur,e liquid discharge through the
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't S e er e esp LA SALLE - UNIT 1 8 3/4 3-6
. Amendment No. 85
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i ATTACIOtENT B PROPOSED CHANGES TO THE TECHNICAL SPECIFICATIONS FOR OPERATING LICENSES NPF-11 AND NPF-18 INSERT T Specified surveillance intervals and surveillance and maintenance outage times have been determined in accordance with GENE-770-06-1-A,
" Bases for Changes to Surveillance Test Intervals and Allowed Out-Of-Service Times for Selected Instrumentation. Technical Specifications",
December 1992. When a channel is placed in an inoperable status solely for performance of required surveillances, entry into LCO and required ACTIONS may be delayed, provided'the associated function maintains Feedwater System / Main Turbine Trip System actuation capability. I ki\nla\lasalle\actstii.wpf71
l o 2.2 LIMITING SAFETY SYSTEM SETTINGS BASES 2.2.1 REACTOR PROTECTION SYSTEM INSTRUMENTATION SETPOINTS
. 1 The Reactor Protection System instrumentation setpoints specified in l Table 2.2.1-1 are the values at which the reactor trips are set for each !
parameter. The Trip Setpoints have been selected to ensure that the reactor l core and reactor coolant system are prevented from exceeding their Safety I Limits during normal operation and design basis anticipated operational occur-rences and to assist in mitigating the consequences of accidents. Operation with a trip set less conservative than its Trip Setpoint but within its speci-fied 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.
- 1. Intermediate Range Monitor, Neutron Flux - High The IRM system consists of 8 chambers, 4 in each of the reactor trip systems. The IRM is a 5 decade 10 range instrument. The trip setpoint of 120 divisions of scale is active in each of the 10 ranges. Thus as the IRM is ranged up to accommodate the increase in power level, the trip setpoint is also ranged up. The IRM instruments provide for overlap with both the APRM and SRM systems.
The most significant source of reactivity changes during the power increase is due to control rod withdrawal. In order to ensure that the IRM provides the required protection, a range of rod withdrawal accidents have been analyzed. The results of these analyses are in Section 15.4.1.2 of the FSAR. The most severe case involves an initial condition in which THERMAL POWER is at approxi-mately 1% of RATED THERMAL POWER. Additional conservatism was taken in this analysis by assuming the IRM channel closest to the control rod being withdrawn is bypassed. The results of this analysis show that the reactor is shutdown and peak power is lirited to 1% of RATED THERMAL POWER with the peak fuel enthalpy well below ne fuel failure threshold of 170 cal /gm. Based on this analysis, the IRM provides protection against local control rod errors and continuous withdrawal of control rods in sequence and provides backup pro-tection for the APRM.
- 2. Average Power Range Monitor For operation at low pressure and low.flo'w during STARTUP, the APRM scram setting of 15% of RATED THERMAL POWER provides adequate thermal margin between the setpoint and the Safety Limits. The margin accommodates the anticipated maneuvers associated with power plant startup. Effects of increasing pressure at zero or low void content are minor and cold water from sources available during startup is not much colder than that already in the system. Tempe ature coefficients are small and control rod patterns are constrained by the C J'11Ed)RWM. Of all the possible sources of reactivity input, uniform control rod withdrawal is the most probable cause of significant power increase. Because LA SALLE UNIT 2 B 2-9
ZNFO ONLY - MO CHANGES - l LIMITING SAFETY SYSTEM SETTINGS BASES REACTOR PROTECTION SYSTEM INSTRUMENTATION SETPOINTS (Continued) Average Power Range Monitor (Continued) the flux distribution associated with uniform rod withdrawals does not involve high local peaks and because several rods must be moved to change power by a significant amount, the rate of power rise is very slow. Generally the heat flux is in near equilibrium with the fission rate. In an assumed uniform rod withdrawal approach to the trip level, the rate of power rise is not more than 5% of RATED THERMAL POWER per minute and the APRM system would be more than adequate to assure shutdown before the power could exceed the Sa The 15% neutron flux trip r.emains active until the mode switch sifety Limit. placed in the Run position. The APRM trip system is calibrated using heat balance data taken during steady state conditions. Fission chambers provide the basic input to the system and therefore the monitors respond directly and quickly to changes due to transient operation for the case of the Fixed Neutron Flux-High 118% setpoint; i.e, for a power increase, the THERMAL POWER of the fuel will be less than that indicated by the neutron flux due to the time constants of the heat transfer associated with the fuel. For the Flow Biased Simulated Thermal Power-Upscale setpoint, a time constant of 6 i 1 seconds is introduced into the flow biased APRM in order to simulate the fuel thermal transient characteristics. A more conservative maximum value is used for the flow biased setpoint as shown in Table 2.2.1-1. The APRM setpoints were selected to provide adequate margin for the Safety Limits and yet allow operating margin that reduces the possibility of unnecessary shutdown. The flow referenced trip setpoint must be adjusted by the specified formula in Specification 3.2.2 in order to maintain these margins when MFLPD is greater than or equal to FRTP.
- 3. Reactor Vessel Steam Dome Pressure-High High pressure in the nuclear system could cause a rupture to the nuclear system process barrier resulting in the release of fission products. A pressure increase while operating will also tend to increase the power of the reactor by compressing voids thus adding reactivity. The trip will quickly reduce the neutron flux, counteracting the pressure increase. The trip setting is slightly higher than the operating pressure to permit normal operation without spurious trips. The setting provides for a wide margin to the maximum allowable design pressure and takes into account the location of the pressure measurement compared to the highest pressure that occurs in the system during a transient.
This trip setpoint is effective at low power / flow conditions when the turbine stop valve closure trip is bypassed. For a turbine trip under these conditions, the transient analysis indicated an adequate margin to the thermal hydraulic limit. LA SALLE - UNIT 2 B 2-10
- .. . -- .. . - ~ .. . - - . - -
l l , 3 /4.3 INSTRUMENTATION 3/4.3.1 REACTOR PROTECTION SYSTEM INSTRUMENTATION i 1 LIMITING CONDITION FOR OPERATION 3.3.1. As a minimum, the reactor protection system instrumentation channels - shown in Table 3.3.1-1 shall be OPERABLE with the REACTOR PROTECTION SYSTEM RESPONSE TIME as shown in Table 3.3.1-2. -- APPLICABILITY: As shown in Table 3.3.1-1. #M ACTION: _ ( With the number of OPERABLE channels less than required by the Minimum j OPERABLE Channels per Trip System requirement for one trip system, place the inoperable channels and/or trip system in the tripped condition
- fl within I hour. l
- 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 I hour nd take the ACTION required by Table 3.3.1-1. _ SURVEILLANCE REOUIREMENTS 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.I'-1.
4.3.1.2 LOGIC SYSTEM FUNCTIONAL TESTS and simulated automatic operation of 1 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 shown in Table 3.3.1-2 shall be demonstrated to be within its 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 reactor trip system. A.SE l 6With a design providing only one channel per trip system, an inoperable channel need not be placed in the tripped condition where this would cause J the Trip Function to occur. In these cases, the inoperable channel shall be l l restored to OPERABLE status within 2 hours or the ACTION required by ' Table 3.3.1-1 for that Trip Function shall be taken. t L **If more channels are inoperable in one trip system than in the other, select
- that trip system to place in the tripped condition, except when this would 1 t cause le Trip Eunction to-y LA SALLE - UNIT 2 3/4 3-1 Amendment No. 78
l
\
1 ATTACHMENT B PROPOSED CHANGES TO THE TECHNICAL SPECIFICATIONS FOR OPERATING LICENSES NPF-11 AND NPF-18 INSERT A
- a. With one channel required by Table 3.3.1-1 inoperable in one or more Functional Units, place the inoperable channel and/or that trip system in the tripped condition
- within 12 hours,
- b. With two or more channels required by Table 3.3.1-1 inoperable in one or more Functional Units:
- 1. Within one hour, verify sufficient channels remain OPERABLE or tripped
- to maintain trip capability in the Functional Unit, and
- 2. Within 6 hours, place the inoperable channel (s) in one trip system and/or that trip system ** in the tripped condition *,
and
- 3. Within 12 hours, restore the inoperable channels in the other trip system to an OPERABLE status or tripped *.
- c. Otherwise, take the ACTION required by Table 3.3.1-1 for the Functional Unit.
INSERT B
- An inoperable channel or trip system need not be placed in the tripped condition where this would cause the Trip Function to occur. In these cases, if the inoperable channel is not restored to OPERABLE status within the required time, the ACTION required by Table 3.3.1-1 for the Functional Unit shall be taken.
** This ACTION applies to that trip system with the most inoperable channels; if both trip systems have the same number of inoperable channels, the ACTION can be applied to either trip system.
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E TABLE 3.3. -1 (Continued) , g REACTOR TEM INSTAt30fiATION E .
. TABLE 3.3.1-1 (Continued)
M REACTOR i. itssION SYSTEM IIISTalSENTATION g
- APPLICABLE MINIIERI OPERABLE l OPERATIONAL CHAleELS PER i FuleCTICIIAL INIIT ,, - ', COMITICIe5 TRIP SYSTEM (a) ACTICII
- 7. Primary Centalspee(Pressure - Nigh 1, IIII 2I8) 1
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- 3 ,
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'tw Velve Trip System til Pressure - Law 1gg) 2g) 5 t
. . O l 11. Reacter Itsde Switch Shutdeun -
Positten 1, 2 1 1 3, 4 1 7 5 1 3 . ! D
- 12. finnual Scram 1, 2 1 1 th i 3, 4 1 8
5 1 3 h j 13. Centrol Red Drive ( a. Charging .eter Header Pressure - Low 2(h) 5 2 2 1 3 g b. Delay Timer 2(h) 2 1 - 5 2 3 i . I
a 1HFO MLV - NO CMANGES I TABLE 3.3.1-1 (Continued) REACTOR PROTICTION SYSTB INSTRUM UTATION ACTION STATWENTS . ACTION 1 - Se in at least NOT SHUTDOW within 12 hours. i 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 hour. ACTION 3 - Sus;end all operations involving CORE ALTRATIONS* and insert all - insertahle centrol reds'within one hour. ACTION 4 - Se in at feest STARTUP within 5 hours. ACTISI 5 - Se.in STARTUP'with the main s b line isolation valves c1'osed within 6 hours or in at least HUT SitlTDOWN within 22 hours.
~
ACTION E - Initista a reduct1en in THENUL POWER within 15 minutes and reduce tuttine first stage pressure te < 140 psig, equivalent to TMNHL POWER less then 305 of RATEDh POWER, within 2,hourt. ACTION 7 ' - Verify aiT insertable control rods to be inserted within 1 hour. ACTION 8 - Lack the reactor mode switch in the Shutdown position within
, 1 hour. -
ACTION 9 - Suspend all operations involving CORE ALTRATIONS,* and insert i all insertable control rods and lock the reactor mode switch in ! the SWTDOW position within 1 hour. I
'..i:Ic
*Except movement erIWt, SWI, or special movable detectors, or replacement of ~ ~
LPWI strings prerlded.5EM instrumentation is OPERABLE per Specification 3.9.2. ij~$'1%$.&M - ' 0.ve q%.l . ;t. . G G e LA 5ALLE - UNIT 2 3/4 3-4
TABLE 3.3.1-1 (Continued) REACTOR Pfl0TECT10N SYSTEM INSTRUNENTATION TABLE NUTATIONS (a) A channel may be placed in an insperehle stitus for up te hours for l required surveillance without placing the channel in the tripped condition l ' provided at least one OPERABLE channel in the some trip Jystem is monitoring that parameter. . (b) The " shorting links" shall be removed from the RP5 circuitry prior to and
. during the time any control red is withdrawn
- and during shutdown margin essenstrations perfereed per Specification 3.10.3.
(c) An APM channel is insperehle if there are less then 2 LPM imputs per level er less than 34 LPM impets to an APM channel. - (d) his function is not required to be OPERABLE tenen the reacter pressure i vessel head is enholted or removed per specification 3.30.1. l (e) This function shall be autametically hypassed when the reactor made switch is not in the hun positism. . (f) This function is met required to be OPERABLE.when PRDERY CONTAlmelf INTEGRITY is not required. (g) Alme actmetas the staney gas treatment system.- (h) With any control red withdrawn. Itet applicable ta control rods removed per Specification 3.9.30.1 or 3.9.30.2. (1) This function shall he autamatically bypassed when turbine first stage pressure is < 140 psig, equivalent to TNEMAL POWER 1ess than 30E of RATD T M NEE POWER. (j) Also actuates the WC-RPT systan. . anet required for sentrol rods removed per specification 3.9.30.1 or 3.9.10.2. P . 3/4 3-5 Amenchment No.17 LA 5ALLE - UNIT 2
5 TABLE 3.3.1-2 ' REACTOR P90TECTION SYSTER RESP = TIES W l-m TASLE 3.3.1-2 - REACTOR P90TFCTION SYSTEN RESPONSE TIMES E ' q . y, c u /- ; - RESPONSE TIE FWCT10mL WIT h ' ,;- . (Seconds) h
- 1. Intermediats Aqppe peelterst
- a. Neutron Files 'Nigh" ' '
Q M
- h. Insperative M
- 2. Average Power Range Moniter*
- l'
- a. Neutron Fleet - Nigh, Setdeun '
M. ..
- h. Fleu Blased Slaulated Thermal Feuer-$ scale < e.es i w c. Fixed Neutron Fleet - Nigh I 8.09 4 i i d. Insperative b '
- 3. Reacter Vessel Steen Dome Pressure - Nigh i 6.55 -
- 4. Reacter Vessel Water Level - Law, Level 3 '
< 1.95 g
- 5. Main Steam Line Isolatten Valve - Closure < 8.06
- 6. Main Steam Line Radiatten - Nigh b b-7.
8. PrimaryCentainmentPressure-Nip Scram 31scharge Volume Water Level - Nigh E , kh M . 9. Tertine Step Valve - Closure i
- 10. Turbine Centrol Valve Fast Closure. -
1 0.05 ( Trip 011 Pressure - Lou < S.088
- 11. Reacter Nede Sultch Shutdeus Posities b
- 12. Manual Scram M
- 13. Centrol Red Drive g a. Charging Water Needer Pressure - tw m
- b. Delay Timer M
" " Neutron detectors are exempt free roepense time testing. Response ties shall be esasured from the detector output er from the tegnet of the first electronic component in the
] channel. , l * **Not Including simulated thermal power time constant.
#Nessured from start of turbine control valve fast closure. .
TABLE 4.3.1.1-1 e-REACTOR PROTECTION SYSTEN INSTRUMENTATION SURVEILLMEE REWIRDIENTS P CHANNEL OFERATIONAL
. CHANNEL FUNCTIONAt. CHANNEL CONDITIONS FOR WHICH FUNCTIONAL UNIT CHECK TEST CAllBRATION(a) SURVEILLANCE REOUIRED 0 1. Intermediate Range Monitors m a. Neutron Flux - High S/U*', S S/U"', W R 2 S W R 3 4, 5
- b. Inoperative NA W NA 2 5
- 2. Average Power Range Monitor:"'
- a. Neutron Flux - High, l Setdown S/U,5 S/U"3, W SA 2 O
, S W M 5 i b. Flow Blased Simulated Thermal Power-Upscale S, D"' S/U"3,#@ W'* SA , R'"' I w c. Fixed Neutron Flux - . 1 High S S ,- Q W'*, SA 1 w d. Inoperative NA q NA 1, 2, 3, 5
- 3. Reactor Vessel Steam Dome Pressure - High NA M Q 1, 2
- 4. Reactor Vessel Water Level -
Low, Level 3 JW$ XQ R 1, 2
- 5. Main Steam Line Isolation Valve - Closure M Q R I 3E 6. Main Steam Line Radiation -
g x High S M R 1, 2 G 7. Primary Containment Pressure - I High NA I q 1, 2
=
1ABLE 4.3.1.1-1 (Continued) g REACTOR PR0iEC110N SYSTEN IN51RUMENTAll0N SURVEILLANCE REOUIREMENTS y CHANNEL DPERATIONAL p CHANNEL FUNCIIONAL CHAletEl CONDITIONS FOR WillCil
',' FUNCil0NAL UNIT - ClllCK TEST Call 8 RATION SURVEILLANCE REQUIRED ji 8. Scram Discharge Volume Water M Level - High NA G R 1, 2, 5 m 9. Turbine Stop Valve - Closure NA Q R I I
- 10. Turbine Control Valve Fast
- Closure Valve Trip System 011 Pressure - Low NA Q R I l
- 11. Reactor Mode Switch Shutdown Position NA R NA 1,2,3,4,5
- 12. Manual Scram NA W NA 1,2,3,4,5 1
- 13. Control Rod Drive
- a. Charging Water Header Pressure - Low NA M R 2, 5
- b. Delay Timer NA M R 2, 5 i ;
Y m (a) Neutron detectors may be excluded from CHANNEL CALIBRATION. (b) TheIRMdndSRMchannelsshallbedeterminedtooverlapforatleast1/2decadesduringeachstartupand 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 hours 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 levels calculated by a heat balance during OPERATIONAL CONDITION I when THERMAL POWER ;t 25% of RATED THERMAL , POWER. The APRM Gain Adjustment Factor (GAF) for any channel shall be equal to t power value deter-
- i mined by the heat balance divided by the APRM reading for that channel.
>. Within 2 hours, adjust any APRM channel with a GAF > I.02. In addition, adjust any APRM channel within d 12 hours, (1) If power is greater than or equal to 90% of RATED THEIIMAL POWER and the APRM channel GAF is 5 < 0.98, or (2) if power is less than 90/. of RATED THEllMAL POWER and the APRM reading exceeds the power 3
- value determined by the heat balance by more than 10% of RATED THEllMAL POWER. Untti any required APRM adjustment has been accomplished, notification shall be posted on the reactor control panel.
z (e) This calibration shall consist of the adjustment of the APRM flow biased channel to conform to a P calibrated flow signal.
~
* (f) The LPRMs shall be calibrated at least once per 1000 effective full power hours (EFPM).
(g) Measure and compare core flow to rated core flow. JNJERT (h) This calibration shall consist of verifying the 6 i I second simulated thermal power time constant. "C 4 e
.l ,
l
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l 1 ATTACHMENT B ' PROPOSED CHANGES TO THE TECHNICAL SPECIFICATIOBIS FOR OPERATING LICENSES NPF-11 AND LPF-18 INSERT C
- The provisions of Specification 4.0.4 are not applicable for a i period of 24 hours after entering OPERATIONAL CONDITION 2 or 3 when shutting down from OPERATIONAL CONDITION ~1.
I I L kt\nla\lasalle\aotstil.wpf$9 l
o INSTRUMENTATION 3 /4. 3. 2 ISOLATION ACTUATION INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.2 The isolation actuation instrumentation channels shown in Table 3.3.2 shall be OPERABLE with their trip setpoints set consistent with the values shown in the Trip Setpoint column of Table 3.3.2-2 and with ISOLATION SYSTEM RESPONSE TIME as shown in Table 3.3.2-3. APPLICABILITY: As shown in Table 3.3.2-1. ACTION:
- a. 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. A th the number of OPERABLE channels less than required by the l Minimum OPERABLE Channels per Trip System requirement for one trip system, place the inoperable channel (s) and/or trip system in the tripped condition
- within one hour. i l
- 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. _
ZAISERT bl l l i 1 I 1
*An inoperable channel need not be placed in the tripped condition where tSis' would cause the Trip Function to occur. In these cases the inoper-able channel shall be restored to OPERABLE status withinGIto,urs or the ACTION required by Table 3.3.2-1 for that Trip Function shill be taken.
**If more channels are inoperable in one trip system than in the other, select that trip system to place in the tripped condition except when this would cause the Trip Function to occur.
***An inoperable channel need not be placed in '.he tripped condition where ,
this would cause the Trip Function to occur. In these cases, the inoper- l able channel shall be restored to OPERABLE sr.atus within 1 hour or the ACTION required by Table 3.3.2-1 for that Tr p Function shall be taken. LA SALLE - UNIT 2 3/4 3-9 Amendment No. 78 l
I ATTACHMENT B PROPOSED CHANGES TO THE TECHNICAL SPECIFICATIONS FOR OPERATING LICENSES NPF-11 AND NPF-18 INSERT D
- b. With the number of OPERABLE channels less than required by the Minimum OPERABLE Channels per Trip System Requirement for one trip system, either
- 1. Place the inoperable channel (s) and/or trip system in the tripped condition
- within 1 a) 1 hour for trip functions without an OPERABLE channel, b) 12 hours for trip functions common to RPS Instrumentation, and c) 24 hours for trip functions not common to RPS Instrumentation.
or
- 2. Take the ACTION required by Table 3.3.2-1.
- c. With the number of OPERABLE channels less than required by the Minimum OPERABLE Channels per Trip System requirement for both i trip systems,
- 1. Place at least one trip system ** in the tripped condition ***
within one hour, and
- 2. a) Place the inoperable channel (s) in the remaining trip system in the tripped condition *** within j i
- 1) 1 hour for trip functions without an OPERABLE channel,
- 2) 12 hours for trip functions common to RPS Instrumentation, and
- 3) 24 hours for trip functions not common to RPS Instrumentation or b) Take the ACTION required by Table 3.3.2-1.
k o n t.u...u.s.oe st u .wo r s o
]
TABLE 3.3.2-1 (Continued) ISOLATION ACTUATION INSTRUMENTATION ! l ACTION STATEMENTS ACTION 20 - Be inthe at next least24 NOT hours.SHUTDOWN within 12 hours and in COLD SHUTDO ACTION 21 - Be in at least STARTUP with the associated isolation valves { closed within 6 hours or be in at least HOT SHUTDOWN within 12 hours and in COLD SHUTDOWN within the next 24 hours. l ACTION 22 - Close the affected system isolation valves within 1 hour and ! ACTION 23 - declare the affected system inoperable. l ACTION 24 - Be in at least STARTUP within 6 hours. l Establish SECONDARY CONTAINMENT INTEGRITY with the standby gas ACTION 25 - treatment systes operating within 1 hour. t Lock the affected system isolation valves closed within 1 hour ! ACTION 26 - and declare the affected system inoperable. : Provided each other that the manual initiation function is OPERABLE for each line, group valve, inboard or outboard, as applicable, in : restore the manual initiation function to OPERABLE ' status within 24 hours; otherwise, restore the manual initiation function to OPERABLE status within 8 hours; otherwise: a. Be in at least HOT SHUTDOWN within the next 12 hours and in
- b. COLD SHUTDOWN within the following 24 hours, or Close the affected systes isolation valves within the next hour and declare the affected system in operable. l TABLE NOTATIONS May stop bevalves bypassed closed.with reactor steam pressure < 1043 psig and all turbine When handling irradiated fuel in the secondar/ containment and during CORE
# ALTERATIONS and operations with a potential for draining the reactor vessel.
During reactor CORE vessel. ALTERATIONS and operations with a potential for draining the (a) 9 ; See Specification 3.6.3, Table 3.6.3-1 for valves in eachhalve group. (b) A channel may be placed in an inoperable status for up to w nours for required surveillan,7 without placing the channel in the tripped condi- [ i tion provided at is monitoring thatleast one other OPERABLE channel in the same trip system parameter. In addition for those trip systems with a ' design providing only one channel per trip system, the channel may be placed in an inoperable status for up to 8 hours for required surveillance testing without placing the channel in the tripped condition provided that the redundant isolation valve, inboard or outboard, as applicable, in each line is operable and all required actuation instrumentation for that re-dundant (c) Also actuates valve is OPERABLE, W31 ace the trip system in the tripped condition. the standby gas treatment system. (d) (e) A channel is OPERABLE if 2 of 4 instruments in that channel are OPERABLE Also Tableactuates 3.6.5.2-1. secondary containment ventilation isolation dampers per . (f) Closes only RWCU system inlet outboard valve. LA SALLE - UNIT 2 3/4 3-14 : Amendment No. 61 i ' 1
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- e. -
g TABLE 4.3.2.1-1 (Continued) % ISOLATION ACTUATION INSTRUMENTATION SURVEILLANCE REQUIREMENTS E
' CHANNEL OPERATIONAL CHANNEL FUNCTIONAL CHANNEL CONDITIONS FOR WHICH TRIP FUNCTION CHECK TEST CALIBRATION SURVEILLANCE REQUIRED m 4. REACTOR CORE ISOLATION COOLING SYSTEM ISOLATION
- a. RCIC Steam Line Flow - High NA XQ J Q 1,2,3
- b. RCIC Steam Supply Pressure - %
c. Low RCIC Turbine Exhaust Diaphragm NA 5WQ Q 1, 2, 3 d. Pressure - High RCIC Equipment Room NA f [Q { i Q 1,2,3 e. Temperature - High RCIC Steam Line Tunnel NA IYQ Q 1, 2, 3 f w ) f. Temperature - High . RCIC Steam Line Tunnel NA WQ ! Q 1,2,3 r w A- M Temperature - High NA O 1,2,3 (Q
! Q h g. 'Drywell Pressure - High NA i ( Q 1,2,3
- h. RCIC Equipment Room a Temperature - High NA Wk Q 1, 2, 3 g
- 5. RHR SYSTEM STEAM CONOENSING MODE ISOLATION 4
- a. RHR Equipment Area M b.
Temperature - High RHR Area Cooler Temperature - NA , (Q l Q 1, 2, 3 YW n 4 l j High NA l Q 1,2,3
- c. RHR Heat Exchanger Steam Supply Flow - High NA Xh Q 1,2,3 F
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Amendment No.10. l l l
r% . . 5 TABLE 3.3.3-1 (Continued)
!C r- EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRLSENTATION Fi e
MININUM OPERABLE APPLICABLE g CHAlWELS PER TRIP OPERATIONAL Q TRIP FUNCTION FUNCTION I ") COMITIONS I ACTION re C. DIVISION 3 TRIP SYSTEM
- 1. HPCS SYSTEM ,
- a. Reactor Vessel Water Level - Low, Low, Level 2 4 ) 1, 2, 3, 4*, 5* 35
- b. Drywell Pressure - High 1, 2, 3 35
! c. Reactor Vessel Water Level-High, level 8 4(*) 2 ) 1, 2, 3, 4*, 5* 32 i d. Deleted { e. Deleted w f. Pump Discher p Pressure-High (Bypass) 1 1, 2, 3, 4 , 5* 31 l 3 g. HPCS System Flow Rate-Low (Permissive) 1 1, 2, 3, 4 , 5* 31 w h. Manual Initiation 1/divlston 1, 2, 3, 4 , 5* 34 P@
- D. LOSS OF POWER MINIllM TOTAL NO. INSTRU- OPERABLE APPLICABLE
- 0F INSTRU- IENTS TO INSTRU p OPERATIONAL BENTS TRIP IENTSCW) COMITIONS ACTION
- 1. 4.16 kV Emergency Bus Undervoltage 2/ bus 2/ bus 2/ bus 1, 2, 3, 4**, 5** 37
- (Loss of Voltage)
- 2. 2/ bus 2/ bus 2/ bus 1, 2, 3, 4**, 5** 37 I
4.16 kV Emergency) (Degraded Vo'tage Bus Undervoltage
. TABLE NDTATION f i
' (a) A channel / instrument may be placed in an inoperable status for up hours during periods of required
'* surveillance without placin F one other OPERA 8LE channel /g the instrument trip same in the systes/ tripchannel system is / instrument monitoring that in the tripped condition provided at l parameter. .
(b) Also actuates the associated division diesel generator.
*- fc) Provides signal to close HPCS pump discharge valve only on 2-out-of-2 1 ic. g l Appilcable when the system is required to be OPERABLE per Specification .5.2 or 3.5.3.
m Required when ESF equipment is required to be OPERABLE. l *
# Not required to be OPERA 8LE when reactor steam dome pressure is 5 122 psig. . .
Ginr" Erb l
i i l ATTACHMENT B PROPOSED CHANGES.TO THE TECHNICAL SPECIFICATIONS FOR OPERATING LICENSES NPF-11 AND NPF-18 i INSERT E A channel / instrument may be placed in an inoperable status for i (d) up to 2 hours during periods of required surveillance without placing the trip system / channel / instrument in the tripped condition provided at least one other OPERABLE channel / instrument in the same trip system'is monitoring that parameter. i o ki\nla\lasa11e\actstii.wpf61 +
TABLE 3.3.3-1 (Continued) EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION ACTION ACTION 30 - With the number of OPERABLE channels less than required by the Minimum OPERABLE Channels per Trip Function requirement:
- a. With one channel inoperable, place inoperable channel .
in the tripped condition within ne hou r declare __the associated system inoperable. + {qp g
- b. With more than one channel inoperable, declare the associated system inoperable. uir. ieg ACTION 31 - With the number of OPERABLE channels less th required by the place th Minimum OPERABLE channels per Trip Functio inoperable channel in the tripped condition within Me l restore the inoperable channel to OPERABLE status wit n 7 days "
or declare the associated system inoperable. _ ars, ACTION 32 - With the number of OPERABLE channels less than required by the Minimum OPERABLE Channels per Trip Function requirement, declare the associated ADS trip system or ECCS inoperabl[/l/,,24 J, @ ACTION 33 - With the number of OPERABLE channels less than the Minimum OPERABLE Channels per Trip Function requirement, place the , y inoperable channel in the tripped condition within Ena houp. ACTION 34 - With the number of OPERABLE channels less than required by the Minimum OPERABLE Channels per Trip Function require nt, restore the inoperable channel to OPERABLE status within ourn or declare the associated ADS trip system or ECCS inoperab'e. ACTION 35 - With the number of OPERABLE channels less than required by the Minimum OPERABLE Channels per Trip Function requirement
- a. For one trip system, glace that trip system in the tripped condition within dne 1ouf)or declare the HPCS system inoperable. &- g./ 4Q
- b. For both trip systems, declare the HPCS system inoperable.
ACTION 36 - Deleted ACTION 37 - With the number of OPERABLE instruments less than the Minimum Operable Instruments, place the inoperable instrument (s) in the l tripped condition within 1 hour or declare the associated 1 emergency diesel generator inoperable and take the ACTION ' required by Specification 3.8.1.1 or 3.8.1.2 as appropriate. I l l l l I LA SALLE - UNIT 2 3/4 3-27 Amendment No. 78 s j
~
s i TABLE 3.3.3-1 (Continuedl BWIBENCY C0ftf COOLIM SYSTEN ACTUATION INSTRUMENTATION , 1 M ) ACTION 38 iffth the susber of OPGIABLE channels less than required by ) the Rfnimus OPGtASLE Channels per trip function requirements:
- s. ittth one channeldneperable, remove the inoperable channel
" ** i a ' - hedrj restore the inoperable channel to g 4N# OPERABLE stdtus within 7 days or declare the associated ECCS systems insperable.
- b. ittth both channels inoperable, restore at least one .
chesnel ta OPERASLE status within ene hour er declare the .
,, eseociated ECCS system inoperable.
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^
3 i * - E TA8LE 4.3.3.1-1 (Continued) *: g i. r-ENERGENCY CORE COOLING SYSTEN ACTUATION INSTRINENTATION SURVEILLANCE REQUIRFENTS E e CHANNEL OPERATI9NAL g CHANNEL FUNCTIONAL CNANNEL COMITIONS FOR WICN q TRIP FUNCTION CHECK TEST CALIBRATION SURVEILLANCE REQUIRED ' C. DIVISION 3 TRIP SYSTEN
- 1. HPCS SYSTEN ,
- s. Reactor Vessel Water Level - -
Low Low, Level 2 W4l R 1, 2, 3, 4*, $* *
- b. Drywell Pressure-High .WQ l q 1,2,3
- c. Reactor Vessel Water Level-High j i i d.
Level 8 Deleted
- h. 1 I
-WQl
'/
R 1, 2, 3, 4*, 5* ' i ' w '
- e. Deleted q ,
! i f. Pump Discharge Pressure-High M Mil Q 1, 2, 3, 4*, $*
- w g. HPCS System Flow Rate-Low M q 1, 2, 3, 4*, 5* '
I j h. Manuel Initiation M R NA 1, 2, 3, 4*, 5* ' l D. LOSS OF POWER - ! 1. 4.16 kV Emergency Bus Under- NA NA R 1, 2, 3, 4** 5** . ! voltage (Loss of Voltage) , i l 2. 4.16 kV Emergency Bus Under- M M R 1, 2, 3, 4**, 5** i voltage (Oegraded Voltage) ( , TABLE WTATIONS l f
?+
#Not required to be OPERABLE when reactor steam done pressure is less than er equal to 122 psig.
*When the system is required to be OPERABLE after being manually realigned, as appilcable, per y Specification 3.5.2. ,
l ** Required when ESF equipment is required to be OPERABLE. i y 1 S ' l l -
. ~e
INSTRUMENTATION 3/4.3.4 RECIRCULATION PUMP TRIP ACTUATION INSTRUMENTATION ATVS RECIRCULATION PLMP' TRIP SYSTEM INSTRUMENTATI3N LIMITING CONDITION FOR OPERATION 3.3.4.1 The anticipated transient without scras recirculation pump trip (ATWS-RPT) systes instrumentation channels shown in Table 3.3.4.1-1 shall be i OPERABLE with their trip setpoints set consistent with the values shown in the Trip Setpoint column of Table 3.3.4.1-2. APPLICABILITY: OPERATIONAL CONDITION 1. ACTION:
- a. With an ATWS recirculation pump trip systes instrumentation channel trip setpoint less conservative than the value shown in the Allowable Values column of Table 3.3.4.1-2, declare the channel inoperable untti the channel is restored to OPERABLE status with its trip setpoint adjusted consistent with the Trip Setpoint value. '
- b. With the number of OPERABLE channels one less than required by the Minimim 0PERABLE Channels per Trip System requirement for one er both trip systemsalace the inonerable channel (s) in the tripped condition withing noun _. ^- Jc/ h@
- c. With the number of OPERABLE channels two or more less than required
, by the Minimum 0PERABLE Channels per Trip System requirement for_one trip systes and: {y/,nn
- l
- 1. If the inoperable channels consist of one reactor vessel water i level channel and one reactor vessel pressure channel . place ,
both inoperable channels in the tripped condition within q,hgger-or, if this action will initiate a pump trip, declare the trip system inoperable.
- 2. If the inoperable chant)els include two reactor vessel water !
level channels or two reactor vessel pressure channels, declare the trip system inoperable.
- d. With one trip system inoperable, restore the inoperable trip system to 0PERABLE status within 72 hours or be in at least STARTUP within the next 6 hours.
- e. With both trip systems inope able, restore at least one trip systes '
to 0PERA8LE status within I hour or be in at least STARTUP within the next 6 hours. - SURVEILLANCE REQUIREMENTS 4.3.4.1.1 Each ATWS recirculation pump trip system instrumentation channel shall be demonstrated OPERA 8LE by the performance of the CHANNEL CHECK, ; CHANNEL FUNCTIONAL TEST and CHANNEL CALIBRATION operations at the frequencies i shown in Table 4.3.4.1-1. .:. _ 4.3.4.1.2 LOGIC SYSTEM FUNCTIONAL TESTS and simulated automatii: operation of b all channels shall be performed at least once per 18 months. l LA SALLE - UNIT 2 3/4 3-35 Amendment No. 63 l
b 5 *
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- LA SALLE - UNIT 2 3/4 3-36
- Amendment No. 63
- . . . - , , - -- -- --- ,m--- -,.
TA8LE 3.3.4.1-2 9
, ATWS RECIRCULATION PUMP TRIP SYSTEM INSTRUMENTATION SETPOINTS '
l [ TRIP ALLOWABLE
, TRIP FUNCTION SETPOINT VALUE
- 1. Reactor Vessel, Water Level - >- 50 inches * >- 57 inches
- Low Low, level 2 c.
- 2. Reactor Vessel Pressure-High 1 1135 psig i 1150 psig R
T1 O O 5
- i. . - ,
h g '
.O k
O b a O 3
" See Bases Figure 83/4 3-1.
g - 5 TABLE 4.3.4.1-1 I W ATWS RECIRCULATION PUNP TRIP ACTUATION INSTRLROTATION SURVEILLANCE REQUIRDUTS { , 1-CHANNEL CHANNEL FUNCTIONAL CHANNEL g TRIP FUNCTION CHECK TEST CALIBRATION ,,
- 1. Reactor Vessel Water Level - S R Low Low, Level 2 I i
- 2. Reactor Vessel Pressure - High 5 YQ Q i.
W 1 f .. . n F O
INSTRUMENTATION [ . END-OF-CYCLE RECIRCULATION PUMP TRIP SYSTEM INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.4.2 The end-of-cycle recirculation pump trip (EOC-RPT) system instrumenta- ' tion channels shown in Table 3.3.4.2-1 shall be OPERABLE with their trip setpoints set consistent with the values shown in the Trip Setpoint column of Table 3.3.4.2-2 and with the END-OF-CYCLE RECIRCULATION PUMP TRIP SYSTEM - RESPONSE TIME as shown in Table 3.3.4.2-3. APPLICABILITY: OPERATIONAL CONDITION 1, when THERMAL POWER is greater than or equal to 30% of RATED THERMAL POWER. ACTION:
- a. With an end-of-cycle recirculation pump trip system instrumentation channel trip setpoint less conservative than the value shown in the Allowable Values column of Table 3.3.4.2-2, declare the channel inoperable until the channel is restored to OPERABLE status with the '
channel setpoint adjusted consistent with the Trip Setpoint value.
- b. With the number of OPERABLE channels one less than required by the Minimum OPERABLE Channels per Trip System requirement for one or both trip systems, niace the ino condition within@perable pq channel (s) in the tripped
- c. With the number of OPERABLE channels two or more less than required by the Minimum OPERABLE Channels per Trip System requirement (s).for one trip system and:
- 1. If the inoperable channels consist of one turbine control valve channel and one turbine stop valve channel, place both _ ,
inoperable channels in the tripped condition within
- 2. If the inoperable channels include two turbine control valve channels or two turbine stop valve channels, declare the trip system inoperable.
- d. With one trip system inoperable, restore the inoperable trip system to OPERABLE status within 72 hours, otherwise, either:
- 1. Increase the MINIMUM CRITICAL POWER (MCPR) Limiting Condition for Operation (LCO) to the E0C-RPT inoperable value per Speci-fication 3.2.3 within the next I hour, or
-- 2. Reduce THERMAL POWER to less than 30% of RATED THERMAL POWER !
within the next 6 hours. l
- e. With both trip systems inoperable, restore at least one trip system to OPERABLE status within 1 hour,
- 1. Increase the MINIMUM CRITICAL POWER (MCPR) Limiting Condition for Operation (LCO) to the E0C-RPT inoperable value per Speci-fication 3.2.3 within the next I hour, or
- 2. Reduce THERMAL POWER to less than 30% RATED THERMAL POWER within the next 6 hours.
.: eti i r mHT ? 3/4 3-39 Amendment No. 78 A: -- - m . -- - w- - S w .s m-
.TNFo ONLY- N0 CHANGEJ .
INSTRUMENTATTON
$URVEILLANCE REQUIREMENTS 4.3.4.2.1 Each and-of-cycle recirculation pg trip syptam instrumentation
. channel shall be demonstrated OPERABLE by the perfomance of the CHANNEL ;
FUNCTIONAL TEST and CHAMIEL CALIBRATION operations at the frequencies shown in t Table 4.3.4.2.1-1. : 4.3.4:2.2 LOGIC SYSTEM PUNCTIONAL TESTS and simulated automatic operation of all channels shall be performed at least once per 18 months. 4.3.4.L 3 The END-OF-CYCLE RECIRCULATION PUNP TRIP SYSTEM RESPONSE TIME of i each trip function shown in Table 3.3.4.2-3 shall be demonstrated to be within ! its limit at least once per 18 months. Each test.shall include at least the . logic of one type of channel input, turtrine control valve fast closure or
. turbine stap valve closure, such that both types of channel inputs are tasted at least once per 35 months: The time allotted for breaker arc suppression shall be verified by test at least once per'60 months.
e e e i 1
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I l u .- e LA SALLE - UNIT 2 3/4 3-40
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A & O LA SALLE - LMIT 2 3/4 3-41 . l
g TABLE 4.3.4.2.1-1
% END-OF-CYCLE RECIRCULATION PUMP TRIP SYSTEM SURVEILLANCE REQUIREMENTS I
R
, CHANNEL FUNCT10NAL CHANNEL E TRIP FUNCTION TEST CAllBRATION a
~ 1. Turbine stop Valve Closure 0
- 2. Turbine Control Valve-Fast Closure Q N
t T1 O O 5 ~ k. y E
$ I k
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ZN FO ONLY- NO CHAhlGS3 t- . INSTR M NTATIN - 3/4.3.5 REACTOR' CORE ISOLATION COOLING SYST S ACTUATI M NSTituMENTATION ,
. .L - Ani CO W TION M OP R ATI M l
3.3.5 The reacter core isolation coolins RCIC) system actuation instru-mentation channels shown .in Tele 3.3.k1 s(hall be OPRABLE with their trip setpoints set consistant with the values shown in the Trip 5etpoint column of Table 3.3.F L APPLICABILITY: OPRATIONAL COMITIONS 1, 2 and 3 with reactor steam l dame pressure greater than 150 peig. 3: . -
., s. With a RCIC systas actuation instrumentation channel trip setpoint
*less conservative then the valut shown in the. Allowable Values colina of T4ble 3.3.k2, declare the channel inoperable until tho' -
channel is restored to OP RABLE status with its trip setpoint - - - ad,justad consistent with the Trip 5etpoint value. .
- h. With one er mers RCIC actuation instmoentation channels inoperable, take the required by Table 3.3.5-2. ,
SURVEILLANCE REDUIltEMENTS 4.3.5.1 Each RCIC system actuation instrumentation channel shell be dear...strated OPDABLE by the performance of the CHANNEL CHECK, CHANNEL
, FUNCTIONAL TEST and CH4 fem. CALIBRATION. operations at the frequencies shown in Table 4.3.5.1-1. ,
4.3.5.2 LDGIC SYSTM FUNCTIONAL TESTS and simulated automatic operation of all channels shall be performed at least once per 18 months. e i LA SALLE - UNIT 2 3/4 3-45 _ - - _ - - - _ _ . _ _ _ - . _ - _ _ _ _ _ _ _ - _ _ _ - _ - _ _ _ _ _ - _ _ - - - _ _ - _ _ _ _ ~ _ _ _ _ . - - _ _ . _ _ _ _ _ _ _ . . _ _ _ _ _ _ . _ _ . - _ - _ _ _ . - _ . _ _ _ _ _ - _ _ _ . _ _ - - . . _ _ _ .
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LA SALLE - tmIT 2 3/4 3-46 l
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l TABLE 3,3,5-1 (Continued) 1 ACTIDIL 50 - Wth the meter of OPRABM channel's less than required by the i Maimue OPRABE Channels per Trip System requirement: -
- a. For one trip system place the inoperable channel in the . i
. tripped condition Ethin dine new0er declaru the RCIC l system inoperable. . - %
4_ (2N heinr7$
- h. For both trip systans, declare the RCIC system inoperable.
ACTIOli 51 - Wth the naber of OPRARE channel less then required by the - ' misimum OPERABLE Chesnels per Trip requirement, declare
. the M p inoperabl y , g fj,_ 2 y 4 g ,
ACTIO11 51 - Wtk the ember of OPDA33 channels less then regtired by the Maimme CPRASE Channels per Trip Systes requi , restore the insperable channel tai OPDABE status within shours or declare the KIC system inoperable. 20 . l l l LA SALM - LIIIT 2 3/4 3-47 , l
I
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I a s,~ . ,s i , - E s il3. 1$ j=' z 5 - l g a j 2 ; E 3 s I a I~E .. l C '4 i a LA SALLE - UNIT 2 3/4 3-49 Amendment No.10
. . J *, . . q. , . . . , * . ... q :.. ~- _
. ~ , . ,
p-Dif m elTATION .
~ '
3/4.3.5 ColmWL ROD WITWitedAL"BLDCK IIISTitl#EffATI0li . Louumi C0peITI0ll P0lt OPetATION I q 1.3.5 The control red witiutraal black instrumentation channels shown in Tele 3.3.k1 shall be OPGIABLE with their tMy setpoints set.consistant with the values shown in the Trip 5etpoint column of Table 3.3.5-L , 4 APPLICABILITY: As shown in Table 1.3.5-1. - M:
- a. With a control red withdraal black instrumentation channel tMp -
setpoint less asneervative than the value shown in the A11eunble Valume colume of Tele 1.3.ht, declare the.chennel inoperele .until
- the channel.is restored to OP RABLE status with its tM p setpoint adjustad consistent wie the Trip 5stpoint value.
- k. Wfth the number of OPRABLE channels lors then required by the Nintaan OPDABLE Channels per Trip System requirement, take the ACTIst required by Table 1.3.PL ,
SUlivEILLABICE xxuuiais ;
- 4. 3. 5 Each of the above required control r'ed withermal block trip systems and instrumentation channels shall be demonstrated OP RABLE hy the performance of the CHAleIEL OECK, CHAleIEL PUIICTIOllAL TESTM CHAlelEL UC**^.TIOli operstiene for the OPSATIollAL COISITICII5 and at the frequenc'ee shown in ;
Table 4.3.Pl. ,
, t
* ==
rHsent "F"
- l i
, i 1 LA SALLE - UNIT 2 3/4 3-50' !
i ATTACHNENT B I PROPOSED CHANGES TO THE TECHNICAL i SPECIFICATIONS FOR OPERATING LICENSES NPF-11 AND NPF-18 1 , INSERT F A channel may be placed in an inoperable status for up to 6 hours for required surveillance (or 12 hours for repair) without placing - the trip system in the tripped condition provided at least one other OPERABLE channel in the same trip system is monitoring that parameter. i P l l l i I O kt\nla\lasa11esmotetii.wpt62 1
, , ,; , . . s .. .,.. ,
7 TABLE 3,3.5-1 (Continued) - CONT 11DL 1100 WITHDittWAL BLOCK IltSTIUMENTATION .
.- M ACTION 50 Declare the IWI inoperable and take the ACTION required try Specification 3.L4.3. _
ACTIDit 81 - WIth the nuber of OPDABE channels: .
- a. One less then regoired try the flinimus OPDABLE Channels per Trip Punction regefrement, resters the Inoperable channel ts OPDASLE states within 7 days or place the inoperable channel in'the tripped ,
condition withis the nest hour. b. i , True er more less then required by the Minimus'OPDABLE Channels per Trip Punction requirement, place at least one inoperable channel in the tripped cenettien within ene hour. ACTICII E2. - Wfth the nusher of OPDASW Channels fees then required by. the Minimus WERABLE' Channels .per-Trip Function requirement place the inoperable .
,, channel fa tem tripped condition within M U hours With TMBEL POWS 1,305 of RATED THERIAL POWEL With men then one centrol red withdram. Not appi.icable to control rods removed per Specification 3.9.10.1 er 3.9.10.L
- a. The RBI ' shelf be mutamatically trypassed when a peripheral control red is >
selected.
- k. This function shall be autametically 1,Mesed if detector count rata is
- 3,100 cps or-the INI channels'are en range 3 er higher. ,_
- c. Thir functies shall be automatically trypassei when the usectated Ilpt l
- channels are en range 8 er higher.
- l l . .
- d. This function shall be automatically lypassed when the Ist channels are en l l range 3 or higher. ,
- s. This function shall be automatically trypassed when the 1st channels are on l
. range 1.
~ -
.m .o ,
l
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l %. LA-SALLE - LMIT 2 3/4 3-52 l 4 l
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4 g TASLE 4.3.5-1 . i g: CeNTROL Ree WITlenAMat stats InsipigENfATIsli setVEILIANCE REeBleE9EWil b,. . CIIMeEL erERATIGIIAL CIIMeEL FISICTIONAL ~ CNMBEL CeleITIONS Fat 48I101 . TRIP FleICTI411 CIECE TEST ' CALleaAited 'I $leVElt!AIICE IEtulRES , , 1. Det SLSCE ISIIITOR .
- a. upscale N.A.
(b f c g g. i b. Insperative . .. N.A. . I bc 9 S* A* I* '
- c. Beunseale .
. N. A. 4~ 18
] ,. , , , , l 3. MEl . j a. Flow Blaced Slantated . Thermal peuer- W ale N.A. 5A 1
- b. Insperative N. A. N.A. 1, 3, 5 ,;
i
. c. Daunseale N.A. 5A l d. Itsutree Fliot-Nish N.A. 54 . E
+ 3. 50 MACE SMIGE 181111085
' ' i Y a. Detector met fell in N.A. ,W N.A. '
,5
- E. h. Upscale N.A. ,W g c.
d. Insperative Downscale . N.A. N.A.
,U
,W : q n.A. j,'el r 5 5
l . 4. INTEWESIATE RAlIGE 1811I18R5 i , l a. Betector not fell la II. A. ,W . N.A. 7 5 j b. Upscale N.A. ,W d 5 c. d. Insperative Downscale N.A. N.A.
,W
,W q
.A. y ,,g 55 .
- 5. Scaml sISCManet yeticIE
- a. Isoter Level-Nigh N.A. -
q 8 1, 2, 5**
- b. Scram Stocharge Volume ,
Sultch in typass - N.A. G. II. A. 5**
- 5. REACTOR C00LANT SYSTEN RECIRCMLATIeII FLOW ,
- s. W ale ,
N.A.
~
gQ q ' 1
- b. Insperative N.A. I gQ , 1 N.A. I
- c. Camparater N.A. ,sVOL . 4 .
1
- f. .
,. TABLE 4,3.5-1 (Castinued) - -
C0irTML MB ti!TIEHted4L BLOCK IltsTapeffATION SURVEILLANCE REQUIREMENT 5 L TABLEISTATIOll5 t (a) neutron detectors may be emeluded from CHAISEL CALIBRATION. (b)ttithin 34 hours prior to startup, if not performed within the - previoue 7 days. (c) Includes reacter manual sentrol moltiplexing system input. 9tth 71EB14L PelER 3,315 of AATB 7)Emi4L POWEL .
- " tilth more thes one control ud wtthereun. list applicable to cantal
. rede removed por $peciffection 3.9.10.1 or 3.9.30.L - .
. 4 O 6 IN'S ER.T "Ei" ,
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LA SALLE - UNIT 1 3/4 3-56 ,
ATTACHNENT B PROPOSED CHANGES TO THE TECHNICAL SPECIFICATIONS FOR OPERATING LICENSES NPF-11 AND NPF-18 INSERT G
*** The provisions of Specification 4.0.4 are not applicable for a period of 24 hours after entering OPERATIONAL CONDITION 2 or 3 when shutting down from OPERATIONAL CONDITION 1.
I i I l kt\nlaglanelleseotst11.wpf63
INSTRUMENTATION Y' CMMNI . 3/4.3.7 MONITORING INSTRUMENTATION RADIATION MONITORING INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.7.1 The radiation monitoring instrumentation channels shown in Table 3.3.7.1-1 shall be OPERABLE
- with their alarm / trip setpoints within the .
specified limits. APPLICABILITY: As shown in Table 3.3.7.1-1. ACTION: .
- a. With a radiation monitoring instrumentation channel alarm / trip setpoint exceeding the value shown in Table 3.3.7.1-1, adjust the setpoint to within the limit within 4 hours or declare the channel inoperable.
- b. With one or more radiation monitoring channels inoperable, take the ACTION required by Table 3.3.7.1-1.
- c. The provisions of Specification 3.0.3 are not applicable.
SURVEILLANCE REOUIREMENTS 4.3.7.1 Each of the above required radiation monitoring instrumentation , channels shall be demonstrated OPERABLE by the performance of the CHANNEL CHECK, CHANNEL FUNCTIONAL TEST and CHANNEL CALIBRATION operations for the conditions and at the frequencies shown in Table 4.3.7.1-1. V t
*The normal or emergency power source may be inoperable in OPERATIONAL CONDITION 4 or 5 or when defueled.
3/4 3-57 Amendment No. 78 LA SALLE - UNIT 2
~
> TA8LE 3.3.7.1-1 l !g RADIATION MONITORING INSTRUMENTATION i
F en MINIMUM CHANNELS APPLICA8LE ALARM / TRIP MEASUREMENT l g INSTRUMENTATION OPERABLE CONDITIONS SETPOINT RANGE
- ACTION
- a. Main Control Room 2/ intake 1,2,3,5 and
- 3. " ' '/hr 0.1 to 10,000 mR/hr 70 i Atmospheric Control
! System Radiation i Monitoring Subsystem TABLE NOTATIONS
$ *When irradiated fue' is being handled in the secondary containment.
"l,
=
f _ERYh ACTION STATEMENT ACTION 70 -
- a. With one of the required monitors inoperable, place the inoperable channel in the downscale tripped condition within 1 hour; restore the inoperable channel to OPERABLE status within 7 days, or, within the next 6 hours, initiate and maintain operation of the control room emergency filtration system in the pressurization mode of operation.
- b. With both of the required monitors inoperable, initiate and maintain operation of the control room emergency filtration system in the pressurization mode of operation within 1 hour.
o
. , , _- ,,n v -n. w- , ^ - t ,.*------e-- -,e e -7 - -==---is. w- -*-- m a -. _------m- =__- m_---
ATTACHMENT B i PROPOSED CHANGES TO THE TECHNICAL SPECIFICATIONS FOR OPERATING LICENSES NPF-11 AND NPF-18 INSERT H A channel may be placed in an inoperable status for up to 6 hours for required surveillance testing without placing the Trip System in the tripped condition, provided at least one other operable channel in the same Trip System is monitoring that Trip Function. t 9 kannla\lasalle\aotatil.wpf64
TABLE 4.3.7.1-1 e- [ RADIATION MONITORING INSTRLMENTATION SURVEILLANCE REQUIREMENTS r
- OPERATIONAL
' CHANNEL CONDITIONS FOR CHANNEL FUNCTIONAL CHANNEL WHICH SURVEILLANCE INSTRUNENTATION CHECK TEST CALIBRATION REQUIRED na a. Main Control Room Atmospheric Control System Radiation Nonitoring Subsystem 5 )[(Q R 1,2,3,5 and
- e g NOTES y, When irradiated fuel is being handled in the secondary containment.
m e
INSTRUMENTATION 3/4.3.8 FEEDWATER/ MAIN TURBJK TAIP SYSTEM ACTUATION INSTRUMENTATION LIMITINGCONDITIONFOROPERATjDN 3.3.8 The feedwater/ main turbine tri) system actuation instrumentation channels ' shown in Table 3.3.8-1 shall be OPERAILE with their trip setpoints set consistent with the values shown in the Trip Setpoint column of Table 3.3.8-2.
' APPLICABILITY: OPERATIONAL CONDITION 1.
ACTION: ERT'$ With a feedwater/ main turbine trip system actuation instrumentation channel trip setpoint less conservative than the value shown in the Allowable Values column of Table 3.3.8-2 declare the channel inoperableandeitherplacetheinoperablechannelinthetrip>ed j; condition until the channel is restored to OPERABLE status wit) its 1 trip setpoint adjusted consistent with the Trip Setpoint value, or declare the associated system inoperable. f b. With the number of OPERABLE channels one less than required by the Minimum OPERABLE Channels restore the inoperable channel to OPERABLE status per withinTrip 7 daysSystem or i*quirement, be in at least STARTUP within the next 6 hours. 1
- c. With the number of OPERABLE channels two less than required by the Minimum OPERABLE Channels per Trip System requirement, restore at ;
least one of the inoperable channels to OPERABLE status within 72 hours or be in at least STARTUP within the next 6 hours. SURVEILLANCE REQUIREMENTS 4.3.8.1 Each feedwater/ main turbine tri system actuation instrumentation channel shall be demonstrated OPERABLE the performance of the CHANNEL CHECK, CHANNEL FUNCTIONAL TEST and CHAN L CALIBRATION operations at the frequencies shown in Table 4.3.8.1-1. 4.3.8.2 LOGIC SYSTEM FUNCTIONAL TESTS and simulated automatic operation of all channels shall be perfomed at least once per 18 months. . (rMse*T"D LASALLE - UNIT 2 3/4 3-86 Amendment No. 69
.. __ ~ _ . . _ . . _ . - - . - - - - - - - --- -- -- ---- - --
'i
ATTACHMENT B ! PROPOSED CHANGES TO THE TECHNICAL SPECIFICATIONS FOR OPERATING LICENSES NPF-11 AND NPF-18 ! INSERT I
- a. With a feedwater/ main turbine trip system actuation instrumentation channel trip setpoint less conservative than the value shown in the Allowable Values column of Table 3.3.8-2, declare the channel inoperable intil the channel is restored to OPERABLE status with its trip setpoint adjusted consistent with the Trip Setpoint value.
- b. With the number of OPERABLE channels one less than that required by the Minimum OPERABLE Channels per Trip System requirement: ,
- 1. Within 7 days, either place the inoperable channel in the tripped
- condition or restore the inoperable channel to L OPERABLE status.
- 2. Otherwise, be in at least STARTUP within 6 hours.
- c. With the number of OPERABLE channels two less than required by i the Minimum OPERABLE Channels per Trip System requirement:
- 1. Within two hours place or verify at least one inoperable ;
channel in the tripped
- condition, and restore either inoperable channel to OPERABLE status within 72 hours, or,
- 2. Be in at least STARTUP within the next 6 hours.
INSERT J An inoperable channel need not be placed in the tripped i condition where this would cause the Trip Function to occur. i l 1
)
1 l l k \nla\lasalle\aotstil.wpf65 I 1
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- 5 .
E I= i s E k
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=
LA SALLE - UNIT 2 3/4 3-87 Amendment No. 69
i 4 ATTACEMENT E ! PROPOSED CHANGES TO THE TECHNICAL SPECIFICATIONS FOR OPERATING LICENSES NPF-11 AND NPF-18 . i INSERT K A channel may be placed in an irioperable status for up to 6 hours for required surveillance testing without placing the Trip System in the tripped condition. , b t I i I i l I e I i ka\ nim \lasalle\aotatil.wpf66 I
l' i 5 W l-m e TABLE 3.3.8-2 E Q FEEDWATER/ MAIN TUR8INE TRIP SYSTEM ACTUATION INSTRLSENTATION SETPOINTS i ALLOWASLE TRIP FUNCTION TRIP SETPOINT VALUE
- a. Reactor Vessel Water Level-High, Level 8 < 55.5 inches * < 56,0 inches
- g R
D Y r-3 i o o b
-r
%n l
i E F
*See Bases Figure 8 3/4 3-1.
m
- _ - _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ m - g + w . -1 ,
m n- = g y y
5 E F 7 TABLE 4.3.8.1-1 FEEDWATER/ MAIN TURSIME TRIP SYSTEM ACTUATION INSTitLSEENTATION SURVEILL N CHANNEL ) CHANNEL FUNCTIONAL TRIP FUNCTION CHANNEL CHECK TEST . CALIBRATION
- a. Reactor Vessel Water Level-High, $
Level 8 f(G R Y. tse I I
.F e
,_v , , .m ,-,-, e i , - - - _ _ - - - - - - - - - - - - - - - - - - -
3 /4,3 INSTRUMENTATION BASES 3/4.3.1 REACTOR PROJ The reactor protection system automatically initiates a reactor scram to:
- a. Preserve the integrity of the fuel cladding.
- b. Preserve the integrity of the reactor 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 perfom 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-279,1971, for nuclear power nian". orotection nystems. Specified i surveillance intervals nor n.v-oor<ure. "SV-Closure. Tv-r mmes: and t d 9 anual scray have been determ E ned in accordance with NEDC-30851P-A, " Technical l Specification Improvement Analyses for BWR Reactor Protection System", March 988d The bases for the trip settings of the RPS are discussed in the bases for Specification 2.2.1. g3 e The measurement of response time at the specified frequencies provides assurance tnat the protective functions associated with each channel are com-pleted within the time limit assumed in the accident analysis. No credit was taken for those channels with response times indicated as not applicable. ! Response time may be demonstrated by any series of sequential, overlapping or total channel test measurement, provided such tests demonstrate the total i channel response time as defined. Sensor response time verification may be demonstra,ted by either (1) inplace, onsite or offsite test measurements, or (2) utilizing replacement sensors with certified response times. ; e v - wJ surn;/bce. anJ m inbace. %+.3Wh ! LA SALLE - UNIT 2 B 3/4 3-1 AMENDMENT NO. 79
-,e -,,-,.--e------ c---ve,- ,,,-----,-,--r,--- ----,,,v-- ,w .- - - , - - , - - - - , , - - , - - - , - - -
ATTACIDEENT B PROPOSED CHANGES TO THE TECHNICAL , SPECIFICATIONS FOR OPERATING LICENSES NPF-11 AND NPF-18 INSERT L and MDE-83-0485 Revision 3, " Technical Specification Improvement Analysis for the Reactor Protection System for LaSalle County Station, Units 1 and 2", April 1991. i I J INSERT M When a channel is placed in an inoperable status solely for performance of required surveillances, entry i'nto LCO and required ACTIONS may be delayed, provided the associated function maintains RPS trip capability. - i l 1 l l J i i i i 1 kr\nla\lanalle\aotstli.wpf67 l
I l
. 1 INSTRUMENTATION RASES 3/4.3.2 ISOLATION ACTUATION INSTRUMENTATION k W '
This specification ensures the effectiveness of the instrumentation used - to mitigate the consequences of accidents by prescribing the OPERABILITY trip setpoints and response times for isolation of the reactor systems. When ; necessary, one channel may be inoperable for brief intervals to conduct required surveillance. Both channels of each trip system for the main steam l tunnel ambient temperature and ventilation system differential temperature may ' be placed in an inoperable status for up to 4 hours for required reactor building ventilation system maintenance and testing and 12 hours for the required secondary containment Leak Rate test without placing the trip system in the tripped condition. This will allow for maintainina 3he 4 imhnity of ; the ventilation system and secondary containment.75ces of the 3r p 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 and of the setting have a direct bearing on safety, are established at a level away from the normal operating 1 range to prevent inadvertent actuation of the systems involved. 1 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 A.C. operated valves, it is assumed that the A.C. l power supply is lost and is restored by startup of the emergency diesel l generators. In this event, a time of 13 seconds is assumed before the valve , 1 starts to move. The safety analysis considers an allowable inventory loss l which in turn determines the valve speed in conjunction with the 13 second delay. l l l l 3/4.3.3 EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRLMENTATION 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, trip setpoints and response times 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 staa time. LA SALLE - UNIT 2 B 3/4 3-2 AMENDMENT NO. 82
I ATTACHMENT B ! i PROPOSED CHANGES TO THE TECHNICAL- I SPECIFICATIONS FOR OPERATING LICENSES NPF-11 AND NPF-18 INSERT N Specified surveillance intervals and surveillance and maintenance outage times have been determined in accordance with NEDC-30851P-A, Supplement 2, " Technical Specification Improvement Analyses for BWR Isolation Instrumentation Common to RPS and ECCS Instrumentation", March 1989, and with NEDC-31677P-A, " Technical Specification Improvement Analysis for BWR Isolation Actuation Instrumentation", July 1990. When a channel is placed in an inoperable status solely for performance of required surveillances, entry into LCO and required ACTIONS may be delayed, provided the associated function maintains primary containment isolation capability. ' INSERT O Specified surveillance intervals and surveillance and maintenance ; outage times have been determined in accordance with NEDC-30936P-A,
" Technical Specification Improvement Methodology (With Demonstration for BWR ECCS Actuation Instrumentation)", Parts 1 and 2, December 1988, and RE-025 Revision 1, " Technical Specification Improvement >
Analysis for the Emergency Core Cooling System Actuation Instrumentation for LaSalle County Station, Units 1 and 2", April 1991. When a channel is placed in an inoperable status solely for performance of required surveillances, entry into LCO and required ACTIONS may be delayed, provided the associated function maintains ECCS initiation capability. o i kitnla dasalle\aotst11.wpf68
, INSTRUMENTATION BASES 3/4.3.4 RECIRCULATION PUMP TRIP ACTUATION INSTRUMENTATION The anticipated transient without scram (ATWS system provides a means of limiting the consequence)s of the unlikelyrecirculation pum occurrence of a failure to scram during an anticipated transient. The -
response of the plant to this postulated event falls within the envelope of i study events in General Electric Company Topical Report NED0-10349, dated ! March 1971 and NEDO-24222, dated December, 1979, and Appendix G of the FSAR. i The end-of-cycle recirculation pump trip (E0C-RPT) system is a part of the Reactor Protection System and is an essential safety supplement to the reactor trip. The purpose of the E0C-RPT is to recover the loss of thermal margin which occurs at the end-of-cycle. The physical phenomenon involved is ! that the void reactivity feedback due to a pressurization transient can add positive reactivity to the reactor system at a faster rate than the control rods add negative scram reactivity. Each E0C-RPT system trips both recircula-1 tion pumps, reducing coolant flow in order to reduce the void collapse in the core during two of the most limiting pressurization events. The two eventss for which the EOC-RPT protective feature will function are closure of the turbine stop valves and fast closure of the turbine control valves. A generic analysis, which provides for continued operation with one or both trip systems of the E0C-RPT system inoperable, has been performed. The analysis determined bounding i:ycle independent MINIMUM CRITICAL POWER RATIO (MCPR) Limiting Condition for Operation (LCO) values which must be used if the EOC-RPT system is inoperable. These values ensure that adequate reactivity margin to the MCPR safety limit exists in the event of the analyzed transient with the RPT function inoperable. The analysis results are further discussed in the bases for Specification 3.2.3. , A fast closure sensor from each of two turbine control valves provides , input to the EOC-RPT system; a fast closure sensor from each of the other two turbine control valves provides input to the second EOC-RPT system. Similarly, a position switch for each of two turbine stop valves provides input to one EOC-RPT system; a positi'on switch from each of the other two stop valves provides input to the other EOC-RPT system. For each E0C-RPT system, the sensor relay contacts are arranged to form a 2-out-of-2 logic for the fast ! closure of turbine control valves and a 2-out-of-2 logic for the turbine stop l valves. The operation of either logic will actuate the E0C-RPT system and l trip both recirculation pumps. l Each EOC-RPT system may be manually bypassed by use of a keyswitch which is administrative 1y controlled. The manual bypasses and the automatic Operating Bypass at less_than 30%_ of RATED _ THERMAL POWER are annunciated in the control room. g 34,y.f N,n,,, omd wnm.nc, afg fm,3g Specified surveillance intervalsthave been determined in accordance with the following:
- 1. NEDC-30851P-A, " Technical Specification Improvement Analyses for BWR Reactor Protection System", March 1988.
LA SALLE - UNIT 2 B 3/4 3-3 AMENDMENT NO. 79 l
INSTRUMENTATION BASES 3/4.3.4 RECIRCULATION PUMP TRIP ACTUATION INSTRUMENTATION
- 2. GENE-770-06-1-A, " Bases for Changes to Surveillance Test Intervals and Allowed Out-of-service Times for Selected Instrumentation Technical Specifications", December 1992.- "
The E0C-RPT system response time is the time assumed in the analysis between initiation of valve motion and complete suppression of the electric arc, i.e.,190 ms, less the time allotted for sensor response, i.e.,10 ms, and less the time allotted for breaker arc suppression determined by test, as correlated to manufacturer's test results, i.e., 83 as, and plant pre-operational test results. i YWf 1 LA SALLE - UNIT 2 B 3/4 3-3a AMENDMENT NO. 79
ATTACHNENT B PROPOSED CHANGES TO THE TECHNICAL SPECIFICATIONS FOR OPERATING LICENSES NPF-11 AND NPF-18 INSERT P When a channel is placed in an inoperable status solely for performance of required surveillances, entry into LCO and required ACTIONS may be delayed provided the associated function maintains the applicable RPT initiation capability. 1 i i i I kt\nIS\laSalle\ectStil,Wpf69 1
- '. )
1 l l l ~ INSTRUNENTATION BASES 1 I 3/4.3.5 REACTOR CORE ISOLATION COOLING SYSTEM ACTUATION INSTRUMENTATION The reactor core isolation cooling system actuation instrumentation is l provided to initiate actions to assure adequate core cooling in the event of reactor isolation from its primary heat sink arid the loss of feedwater flow to the reactor vessel witho idinc actuation of any of the emergency core cooling souionent.
' ERT3Q !
374.3.6 CONTROL R00 WITHDRAWAL BLOCK IN$TRUMENTATION The control red block functions are provided consistent with the ! requirements of the specifications in Section 3/4.1.4, Control Rod Program Controls and Section 3/4.2 Power Distribution Limits. The trip logic is '
- 6'9ck -arranged so that a trip in any one of p inputs wil_] result in a control rod >
.. 3 74 . 3. 7 MDNITORING INSTRUMENTATION qasemDQ -
3/4.3.7.1 RADIATION MONITORING INST *LM NTATION
- t The'0PERABILITY of the radiation monitoring instrumentation ensures that; (1) the radiation levels are continually measured in the areas served by the s individual channels, and (2) the alam or automatic action is initiated when the radiation level trip setpoint is exceeded. -
3.4.3.7.2 SEISMIC MDNITORING INSTRUMENTATION
- The OPERABILITY of the seismic monitoring instrumentation ensures that sufficient capability is available to promptly detemine the magnitude of a seismic event and evaluate the response of those features important to safety.
This capability is required to pemit comparison of the measured response to that used in the design basis for the unit. This instrumentetion is consistent i with the recommendations of Regulatory Evide 1.12 "Instroentation for Earthquakes". April 1974. ) i 3/4.3.7.3 METEOROLOGICAL MDNITORING INSTRUMENTATION , ! The OPERABILITY of the meteorological monitoring instroentation ensures that sufficient metsomlogical data is available for estimating potential radiation doses to the public as a result of routine er accidental release of radioactive materials to the atmosphere. This capability is required to evaluate the need for initiating protective measures to protect the health and safety of the public. This instrumentation is consistent with the recommenda- " tions of Regulatory Guide 1.23 "0nsite Meteorological Programs," February, 1972. 3/4.3.7.4 REpWTE SHLTTDOWN MONITORING INSTRUMENTATION The OPERABILITY of tne remote shutdown monitoring instrumentation ensures that sufficient capability is available to pemit shutdown and maintenance of HOT SHUTDOWN of the unit from locations outside of the control room. This capability is required in the event control room habitability is lost an,,d is consistent with General Design Criteria 19 of 10 CFR 50. LA SALLE - UNIT 2 8 3/4 3-4 Amendment No. 41 1
ATTACHMENT B PROPOSED CHANGES TO THE TECHNICAL SPECIFICATIONS FOR OPERATING LICENSES NPF-11 AND NPF-18 INSERT Q Specified surveillance intervals and surveillance and maintenance outage times have been determined in accordance with GENE-770-06-2-A,
" Addendum To Bases for Changes to Surveillance Test Intervals and Allowed Out-of-Service Times for Selected Instrumentation Technical Specifications (BWR RCIC Instrumentation)", December 1992. When a channel is placed in an inoperable status solely for performance of required surveillances, entry into LCO and required ACTIONS may be delayed, provided the associated function maintains RCIC initiation capability.
INSERT R Specified surveillance intervals and surveillance and maintenance outage times have been determined in accordance with NEDC-30851P-A, Supplement 1, " Technical Specification Improvement Analysis for BWR Control Rod Block Instrumentation", October 1988, and GENE-770-06-1-A,
" Bases for Changes to Surveillance Test Intervals and Allowed.Out-Of-Service Times for Selected Instrumentation Technical Specifications",
December 1992. When a channel is placed in an inoperable status solely for performance of required surveillances, entry into LCO and required ACTIONS may be delayed, provided the associated function maintains Control Rod Block capability. INSERT S Specified surveillance intervals and surveillance and maintenance outage times have been determined in accordance with GENE-770-06-1-A, l
" Bases for Changes to Surveillance Test Intervals and Allowed Out-Of- l Service Times for Selected Instrumentation Technical Specifications", '
December 1992. When a channel is placed in an inoperable status solely for performance of required surveillances, entry into LCO and required ACTIONS may be delayed, provided the associated function maintains initiation capability. 1 o kiinlasiana11etootst11.wpf70
INSTRUMENTATION 8ASES 3/4.3.7.11 EXPLOSIVE GAS MONITORING INSTRUMENTATION This instrumentation provides for monitoring (and controlling) the con- l centrations of potentially explosive gas mixtures in the waste gas holdup l system. 3/4.3.7.12 LOOSE-PART DETECTION SYSTEM The OPERABILITY of the loose part detection system ensures that sufficient capability is available to detect loose metallic parts in the primary system cnd avoid or mitigate damage to primary system components. The allowable out-of-service times and surveillance requirements are consistent with the recom-mandations of Regulatory Guide 1.133, " Loose-Part Detection Program for the Primary System of Light-Water-Cooled Reactors." 3/4.3.8 FEEDWATER/ MAIN TUR8INE TRIP SYSTEM ACTUATION INSTRUMENTATION, The feedwater/ main turbine trip system actuation instrumentation is provided to initiate the feedwater system / main turbine trip system in the cvent of reactor vessel water level equal to or greater than the level 8 sstpoint associated with a feedwater controller failure to prevent over- i filling the reactor vessel which may result in high pressure liquid dis-chargethroughthesafety/reliefvalvedischargelines.4l _ i l l LA SALLE - UNIT 2 8 3/4 3-6 Amendment No. 69 l 1 I
ATTACHNENT B PROPOSED CHANGES TO THE TECHNICAL SPECIFICATIONS FOR OPERATING LICENSES NPF-11 AND NPF-18 i INSERT T ! Specified surveillance intervals and surveillance and maintenance oiltage times have been determined in accordance with GENE-770-06-1-A,
" Bases for Changes to Surveillance Test Intervals and Allowed Out-Of-Service Times for Selected Instrumentation Technical Specifications",
December 1992. When a channel is placed in an inoperable status solely for performance of required surveillances, entry into LCO and required ACTIONS may be delayed, provided the associated function maintains Feedwater System / Main Turbine Trip System actuation capability. t 6 l t e l l ks\nla\ lana 11e\actstil.wpf71 1 1
I ATTACMGMT C l SIGNIFICANT HAEARDS CONSIDERATION -l Comed has evaluated the proposed Technical Specification Amendment for i LaSalle County Station Units 1 and 2, which: a) Extends the STI and . AOT for certain actuation instrumentation in the reactor protection, i isolation, emergency core cooling, recirculation pump trip, reactor -i core isolation cooling, control rod withdrawal. block, monitoring, and ; feedwater/ main turbine trip systems; b) Proposes a change to the l Feedwater/ Main Turbine Trip LCO action statements to achieve ' consistency with presently existing instrumentation LCOs; c) Deletes i the surveillance of the APRM Neutron Flux - High,'Setdown functional : unit in Operational Condition 1; d) Revising the applicability of the- ; Provisions of Specification 4.0.4 to several Reactor Protection System ! and Control Rod Withdrawal Block Instrumentation Surveillance l Requirements; e) Add the requirement for performing a shiftly channel r check for the applicable RPS, PCIS, ECCS, and RCIC instrumentation ' channels equipped with master trip units; and, f) Administrative changes to correct typographical errors and to delete cycle specific ' footnotes which are no longer applicable. It has been determined that - the changes do not constitute a Significant Hazards Consideration. , Based on the criteria for defining a significant hazards consideration established in 10CFR50.9., operation of LaSalle County Station Units 1 and 2 in accordance wit: the proposed amendment will not:
- 1) Involve a significant increase in the probability or consequences of an accident previously evaluated because:
l
- a. The proposed changes increase the STI and AOT for actuation- !
instrumentation supporting RPS, ECCS, Isolation, CRBF, RCIC, ATWS-RPT, EOC-RPT, Monitoring, and Feedwater/ Main Turbine Trip i System Actuation functions. i There are no changes in instrumentation configuration and [ function, and no instrumentation setpoints are changed. Because of this there is no change in the probability of occurrence of an accident or the consequences of an accident or the , consequences of malfunction of equipment. With respect to the probability of equipment malfunction, topical reports prepared by GE demonstrate that there is a reduction in scram frequency for the RPS, but in the case of the ECCS there is a small increase in the unavailability of the water injection functior.. This increase in unavailability was judged acceptable by GE. ' The NRC concurred with this conclusion in its review and approval of the topical reports. The proposed changes are . consistent with the Safety Evaluation Reports issued for the l topical reports. ; i i i ks\nla\lasalle\antstii.wpf72
)
l ATTACHMEBPC C SIGNIFICANT HAEARDS CONSIDERATION
- b. The changes proposed for the Feedwater/ Main Turbine Trip LCO i action statements provide actions which are consistent with presently existing instrumentation LCOs. The. design and i function of the feedwater/ main turbine trip instrumentation to 6 trip the feedwater pumps-and the main turbine upon detection of i a Level 8 event is not altered. The probability and/or !
consequences of'this moderate frequency transient are not I increased. r
- c. The APRM Neutron Flux - High, Setdown scram setting provides ,
adequate thermal margin between the setpoint.and the safety . limits for' operation at low pressure and low flow during a plant _ j startup. This function remains in effect until the mode switch is placed in the Run (Operational Condition 1) position, at ! which time it is bypassed. Deleting the requirement.for the surveillance of the APRM Neutron Flux - High, Setdown functional ! unit in Operational Condition 1 is appropriate since its ! function is not applicable in this mode. This deletion serves ! to achieve consistency between Technical Specification Tables ! and the Bases section. l 4
- d. The changes associated with Specification 4.0.4 are . _
administrative in nature and are intended to provide the plant ; operators with better guidance for its application. In cases -! where complete surveillances cannot be achieved, such as during ! a plant shutdown, then the required surveillances will be l performed within 24 hours of entering the Mode or condition in ! which the surveillance is required. The stabilization of the l plant will be of primary consideration. This change does not l affect the evaluation for any accident presented in Chapter _15 i of the UFSAR. The APRM Fi::ed Neutron Flux - High quarterly l functional tests most of the APRM channel equipment associated ! with the APRM Neutron Flux - High, Setdown scram. 1 Additionally, the expected result of the functional tests ; associated with the SRMs, IRMs, and APRMs is to demonstrate the l operability of the instrumentation. Therefore, 24 hours is a ' reasonable time to permit the surveillances to be performed upon ! entering the mode or condition in which the surveillance is required.
- e. The proposal to include the performance of channel checks as !
requirements of technical specifications is administrative in i nature. Presently, channel. checks performed for the applicable analog instrumentation in reactor vessel water level applications is controlled solely by procedure. Adding this requirement to the technical specifications provides for the appropriate controls of the surveillances, above and beyond that presently controlled by procedure. , kt\nla\lasalle\aotatis.wpf'3 I I
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-.-- _-, , , . , , . - ,-, , , . , - . . ..m., ., , , - , , - - ,
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. ATTACHMENT C .
SIGNI7ICANT MAEARDS CONSIDERATION
- f. The proposed administrative changes are offered to correct
. typographical errors and delete cycle specific footnotes which :
are no longer applicable. The nature of the changes precludes them from impacting previously analyzed accidents. ; The proposed changes therefore do not involve a significant increase l in the probability or consequences.of an accident previously. 'l evaluated. l
- 2) Create the possibility of a new or different kind of accident from i any accident previously evaluated because:
- a. The proposed changes increase the STI and AOT for'certain- -
actuation instrumentation in the RPS, ECCS, Isolation, CRBF, ! RCIC, ATWS-RPT, EOC-RPT, Monitoring, and Feedwater/ Main Turbine ! Trip systems. There are no changes in instrumentation ! configuration and function, and no instrumentation setpoints are - changed. I
- b. The changes to the Feedwater/ Main Turbine Trip LCO action statements allow the plant operators a maximum degree of operational flexibility, while maintaining the instrumentation and protection needed for terminating the feedzater controller failure transient. The single failure proof criterion of the level sensors is maintained, and the logic of the protective instrumentation is not compromised. The changes to the LCO ,
action statements do not constitute a change to the facility or- l its operation as described in the Safety Analysis Report. I
- c. Deleting the requirement for surveilling the APRM Neutron Flux - l High, Setdown functional unit in Operating Condition 1 does not degrade thermal margins. The margin accommodates the anticipated maneuvers associated with plant power ascension. i During a plant shutdown, rod insertion maneuvers, recirculation !
flow reduction, and xenon build-in all contribute to negative , reactivity insertion which precludes the degradation and l violation of thermal margins. The functions of the APRMs required to be OPERABLE in Operational Condition 1 which are in . effect remain to ensure that reactor core thermal margins arc l not compromised. l
- d. The conduct of neutron instrument functional tests in the plant ,
i mode or condition in which the trips are applicable eliminates unnecessary testing during normal plant operations. The ! expected result of the functional testing is to demonstrate the , operability of the instruments. The failure of any single instrument channel will neither cause nor prevent either a reactor scram or a control rod block. kt\nla\lasalle\aotatii.wpt74 l I r
,i ATTACNMENT C SIGNIFICANT NAEARDS CONSIDERATION I
- e. Including the performance of channel checks for'the applicable l~
analog instrumentation as part of the technical specifications transfers control of the required surveillances from procedure . 'to the technical specifications, as appropriate. -The. ; administrative natu o of this change does not alter the functions, setpoints, or configuration of the associated .i instrumentation. - T
- f. The administrative nature of the. changes prevents them from ;
affecting the functions, setpoints, or configuration of the associated instrumentation from being affected by the changes. The proposed changes do not create the possibility for an accident or malfunction of a different type than any previously evaluated in the > UFSAR.
- 3) Involve a significant reduction in the margin of safety because:
- a. Setpoints are based upon the drift occurring during an 18 month-calibration interval. The bases in the Technical Specifications !
either do not discuss STI, or state "...one channel may be ! inoperable for brief intervals to conduct required ! surveillance." The proposed changes are bounded by the analyses of the topical reports. These analyses, which.were prepared by GE and approved by the NRC, examined the effects of extending ; STI and AOT and found that the proposed changes would not
- involve a significant' reduction in the margin of safety. ,
- b. The proposed changes to the~ turbine trip LCO action statements do not change any of the settings of the Level 8 setpoints. . The single failure criteria of the multiple level sensors which sense and detect the Level 8 setpoint remains intact. The LCO '
maintains the requirement that no single instrument failure will , prevent the feedwater pump turbines and main turbine trip on a valid Level 8 signal. Scram trip signals from the turbine i retain the design feature that a single failure will neither - initiate nor impede the initiation of a reactor scram (trip). !
- c. The setting, function, and conditional requirements of the APRM Neutron Flux - High, Setdown function are not altered. This :
change serves to achieve consistency between two Technical i Specifications Tables. This eliminates the need for surveilling i a function in a mode which is not applicable. The functions of ! the APRMs required to be OPERABLE in Operational Condition 1 remain to ensure that reactor core thermal margins are not compromised.
- d. The reference to 4.0.4 applicability will assist to ensure -
consistent interpretation of the technical specifications by the ; ki\nla\lasalle\aotatil.g f75 f
ATTACHMENT C SIGNIFICANT HAZARDS CONSIDERATION plant operators. This assists in ensuring that the plant is operated within technical specification limitations. This change does not affect trip instrumentation setpoints, and the scram function of the RPS is assured by the weekly functional testing of the Manual Scram,
- e. Including the instrumentation channel checks as part of technical specification requirements provides an appropriately regimented method of controlling the conduct of the surveillances. None of the functions, setpoints, or configuration of the associated analog instrumentation is affected by this administrative change,
- f. The administrative nature of the changes serves to provide more concise guidance to tho plant operating staff, and as such do not impact the safety margin.
The proposed changes do not significantly reduce the margin of safety as defined in the basis for any Technical Specification. Guidance has been provided in " Final Procedures and Standards on No Significant Hazards Considerations," Final Rule, 51 FR 7744, for the application of standards to license change requests for determination of the existence of significant hazards considerations. This document provides examples of amendments which are and are not considered likely to involve significant hazards considerations. This proposed amendments most closely fit the example of a change which may either result in some increase to the probability or consequencer, of a previously analyzed accident or may reduce in some way a safety margin, but where the results of the change are clearly within all acceptable criteria with respect to the system or componen* specified in the applicable Standard Review Plan. This proposed amendment does not involve a significant relaxation of the criteria used to establish safety limits, a significant relaxation of the bases for the limiting safety system settings or a significant i relaxation of the bases for the limiting conditions for operations. l Therefore, based on the guidance provided in the Federal Register and the criteria established in 10 CFR 50.92(c), the proposed change does not constitute a significant hazards consideration. l ki\nla\lasalle\actst11.wpf76
ATTACHMENT D ENVIRONMENTAL ASSESSMENT STATEMENT APPLICABILITY REVIEW j Commonwealth Edison has evaluated the proposed amendment against the criteria for identification of licensing and regulatory actions , requiring environmental assessment in accordance with 10 CFR 51.21. ! It has been determined that the proposed changes meet the criteria for a categorical exclusion as provided under 10 CFR 51.22 (c) (9) . This i conclusion has been determined because the changes requested do not pose significant hazards consideration or do not involve a significant increase in the amounts, and no significant changes in the types, of any effluents that may be released offsite. Additionally, this request does not involve a significant increase in individual or cumulative occupational radiation exposure. I i t h h k \nla\lasalle\aotatii,qf77 l l
ATTACHMENT I NON-PROPRIETARY VERSION OF THE GENERAL ELECTRIC TOPICAL REPORT TECHNICAL SPECIFICATION IMPROVEMENT ANALYSIS , FOR THE REACTOR PROTECTION SYSTEM FOR LASALLE COUNTY STATIONS, UNITS 1 AND 2 l MDE-83-0485 REV. 3. DRF C71-00072-1, APRIL 1991 l l
)
i 1 I 1 kt\nla\lasalle\aotetii.wpf82
MDE-83-0485-NP Rev 3 DRF C71-00072-2 t APRIL 1991 : GENERAL ELECTRIC COMPANY TECHNICAL SPECIFICATION IMPROVEMENT ANALYSIS FOR THE REACTOR PROTECTION - SYSTEM FOR LASALLE COUNTY STATION ' UNITS 1 AND 2 (THIS REPORT HAS BEEN PREPARED FOR COMMONWEALTH EDISON COMPANY THROUGH THE TECHNICAL SPECIFICATION IMPROVEMENT COMMITTEE i OF THE BWR OWNERS' GROUP) H. X. Hoang R. B. Ninomiya APPROVED BY: N A. E. Rogers, Manager Reliability Engineering Services GE NUCLEAR ENERGY SAN JOSE, CALIFORNIA
MDE-83-0485-NP REV 3 , GENERAL ELECTRIC COMPANY ; I The information contained in this document is furnished for the purpose of providing the members of the BWR Owners' Group with plant specific analysis related to changes to the Reactor Protection System Technical Specification testing intervals and allowable out-of-service times. The - information is to be used solely for this purpose. Any other use of th'e information is unauthorized, and as to such unauthorized use, neither the General Electric Company nor any of the contributors to this document makes any warranty or representation (express or implied) with raspect to the accuracy, completeness, or usefulness of the information , contained in this document, or that the use of such information may not infringe privately owned rights of others. General Electric Company assumes no responsibility for liability or damage of any-kind which may result from use of the information contained in this document. Except as specified in the original General Electric Proposal No. 176-1008-EBO, Paragraphs 1.16.1 and 1.16.2, the information shall not be reproduced or furnished to third parties or made public without the prior express , written consent of the General Electric Company. r f 6
.j.
MDE-83-0485-NP REV 3 GENERAL ELECTRIC COMPANY TABLE OF CONTENTS l h i
- o. gggg t
- 1. INTRODUCTION 1
~
- 2. EVALUATION METHOD 2
- 3. RESULTS OF RPS EVALUATION 5 !
- 4.
SUMMARY
AND CONCLUSIONS 9
- 5. REFERENCES 10 APPENDIX A: RPS EVALUATION FOR LA SALLE COUNTY STATION UNITS 1 AND 2 l
l
-ii-
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MDE-83-0485-NP REV 3 GENERAL ELECTRIC COMPANY
- 1. INTRODUCTION This report extends the generic study of modifying the technical specification requirements of the Reactor Protection System (RPS) on a plant specific basis for LaSalle County Station (LSCS) Units 1 and 2.
The generic study (Reference 1) provides a technical basis to modify the surveillance test frequencies and allowable out-of-service time of the RPS from the generic technical specifications. The generic study also -
~
provides additional analyses of various known different RPS configurations to support the application of the generic basis on a plant specific basis. The generic basis and the supporting analyses were utilized in this plant specific evaluation. The results of the plant specific evaluation for LSCS are presented herein. This report represents the latest plant configuration of LSCS as of December 1990. l
)
1 l
p
~
! .. MDE.83-0485-NP REV 3' GENERAL ELECTRIC COMPANY , i i 5 2. EVALUATION METHOD. The plant specific evaluation of the modification of the surveillance j test frequencies and allowable out-of-service time of the RPS was' i performed in the following steps:.
- a. Gather plant specific information on the RPS from the Comonwealth Edison Company (Ceco). The information includes -
the following: * (1) Elementary Diagram of the RPS and related systems. (2) RPS description such as plant F.inal Safety' Analysis Report (FSAR). (3) Technical specifications on the RPS. (4) RPS surveillance test procedures. I l The latest revision of Items 1,2, and 3 above were supplied by j CECO. Item 4 above was provided by CECO in the form of a questionnaire identifying the differences between the procedure used in the generic evaluation and the procedure used at LSCS. Section I of the checklist in Appendix A was used to identify the data source of the plant specific information.
- b. Construct the plant specific RPS configuration from the plant specific information. Questions "A" through "H" in Section II of the checklist were used for this process,
- c. Compare the plant specific RPS configuration with the generic RPS configuration using the generic RPS elementary diagram, RPS description, technical specification requirements, and other generic inputs. Section III of the checklist was used for this process.
l 2-
MDE-83-0485-NP REV 3 GENERAL ELECTRIC COMPANY . I
- d. Classify the differences into three categories:
i (1) Obvious items which have no effect on the reliability of the RPS. Examples of these "no effect" items are component name differences, symbol differences, and other minor r.onfunctional differences, Disposition of the obvious "no effect" items does not require additional i analysis. - (2) Potential differences which require considerable engineering judgment for disposition because of the :i functional differences. Examples of these potential differences are separate channels for manual . scram as opposed to non-separate channel in the generic plant and dual redundant contacts per sensor relay in the - applicable trip channels as opposed to a single set of contacts in the generic plant. The disposition of such items would require engineering assessment as shown in Appendix K of Reference 1. (3) Potential differences which require additional analyses
- to evaluate the effect on the RPS reliability. Examples of such potential differences are using HFA relays as opposed to using both Potter and Brumfield relays and Agastat relays in the generic evaluation. Disposition of these items would require additional analyses to compare with the generic results. These analyses are documented in Reference 1.
- e. Compile a list of plant specific differences of Category (2) and (3).
( - 3-y-T e--' - + - -
]
- r. I l
MDE-83-0485-NP REV 3 GENERAL ELECTRIC COMPANY
- f. Assess the reliability effect of the differences identified in Step (e) on the generic results. The results of the assessment are documented in Section III of the checklist.
- g. Document the results of the plant specific evaluation.
The above seven step process is documented in Appendix A of this report. , 4
MDE-83-0485-NP REV 3 GENERAL ELECTRIC COMPANY
- 3. RESULTS OF RPS EVALUATION The results of the plant specific' evaluation of the RPS for LSCS are documented in Appendix A of this report. The results show that the RPS configuration of LSCS has the following differences which are classified Category (2) or (3):
The term " generic model" means the RPS configuration used in the generic analysis (Reference 1). MDE-83-0485-NP REV 3 , GENERAL ELECTRIC COMPANY i 0 6-
i 1 l i i MDE-83-0485-NP REV 3 i GENERAL ELECTRIC COMPANY i I I 1 em O Y l 4 l r i l l i I
1 l i MDE-83-0485-NP REV 3 GENERAL ELECTRIC COMPANY k I i s e O f I 1
}
I 4 I I 8-
- MDE-83-0485-NP REV 3 GENERAL ELECTRIC COMPANY i !
4
SUMMARY
AND CONCLUSIONS ; A plant specific evaluation of modifying the surveillance test frequencies and allowable out-of-service time of'the RPS from the technical specifications of LSCS has been performed. The evaluation utilized the generic basis and the additional analyses documented in Reference 1. The results indicated that the RPS configuration for LSCS , has several differences compared to the RPS configuration in the gener.ic evaluation. These differences and the assessment of their effects on the RPS failure frequency are shown in Appendix A. The analysis , reported in Reference 1 shows that these differences would not significantly affect the improvement in plant safety due to the changes ! in the technical specifications based on the generic analysis. ; Therefore, the generic analysis in Reference 1 is applicable to LSCS. t I i l l l l l o
MDE-83-0485-NP 'REV 3 GENERAL ELECTRIC COMPANY .5. REFERENCES (1) " Technical Specification Improvement Analyses for BWR Reactor Protection System," General Electric Company, NEDC-30851P-A, March 1988. 7 p b f 1
)
MDE-83-0485-NP REV 3 - GENERAL ELECTRIC COMPANY 4 APPENDIX A
+
~
RPS EVALUATION CHECKLIST FOR LA SALLE COUNTY STATION UNITS 1 AND 2 l l i t i i S i-t A-1
L. U L . NiDE-R3-0485-NP REV 3 GENERAL ELECTRIC COMPANY Section ! - RPS Plant Specific Data Source Utility: Commonwealth Edison Company Plant: LaSalle County Station Units 1 and 2
' Source Number
- 1. RPS Elementary a) Unit 1 - IE-1-4215AA through AM b) Unit 2 - 1E-2-4215AA through AM
- 2. RPS IED 732E152A, Rev. 6, 4 sh.
- 3. RPS MG Set Contr.:.1 System Elementary a) Unit 1 - 1E-1-4216AA, AB, AC b) Unit 2 - IE-2-4216AA, AB, AC
)
- 4. RPS Interconnection Scheme Elementary 807E167TD, Rev. 5, 3 sh. I
^
- 5. RPS Design Specification 22A3083AN, Rev. 10
- 6. FSAR Section 7.1.2 and 8.1.2, Rev. 6, April 1990 ,
'l
- 7. Technical Specifications Section 3/4.3 Amendment No. 75 (Unit 1) l 3/4.3 Amendment No. 59 (Unit 2)
- 8. Surveillance Test Procedure Checklist EDT BOA-8509
- 9. Others: EDT BOA-8537, Jim Marshall to Lynne Rash 03/07/85 A-2
MDE-83-0485-NP REV 3 GENERAL ELECTRIC COMPANY I Section I -' RPS Plant Specific Data Source , Revision No. of RPS Elementary Drawings Unit 1 Unit 2 RPS Elementary 1E-1-4215AA Rev. F 1E-2-4215AA Rev. D - 1E-1-4215AB Rev. F 1E-2-4215AB Rev. D 1E-1-4215AC Rev. AG IE-2-4215AC Rev. AA ; 1E-1-4215AD Rev. AJ 1E-2-4215AD Rev. AA 1E-1-4215AE Rev. AM 1E-2-4215AE Rev. AE 1E-1-4215AF Rev. AJ 1E-2-4215AF Rev. Y 1E-1-4215AG Rev. F 1E-2-4215AG Rev. C 1E-1-4215AH Rev. J 1E-2-4215AH Rev. K 1E-1-4215AJ Rev. D 1E-2-4215AJ Rev. B IE-1-4215AK Rev. S 1E-2-4215AK Rev. L 1E-1-4215AL Rev. N 1E-2-4215AL Rev. M 1E-1-4215AM Rev. G 1E-2-4215AM Rev. E 1E-1-4207ZB Rev. C IE-2-4207ZB Rev. E' 1E-1-4232AW Rev. B 1E-2-4232AW Rev. A. 1E-1-4232AX Rev. B 1E-2-4232AX Rev. A RPS MG Set Control 1E-1-4216AA Rev. H 1E-2-4216AA Rev. K System Elementary 1E-1-4216AB Rev. C 1E-2-4216AB Rev. C 1E-1-4216AC Rev. A 1E-2-4216AC Rev. A A-3
. l l MDE-83-0485-NP REV 3 GENERAL ELECTRIC COMPANY l 1 i Section II - RPS Configuration Data l
- Data A. RPS System QA1A Source i
- 1. Number of trip systems 2 (2,6) ;
~
- 2. Number of logic channels per trip system For Automatic Scram 2, (1,2) j For Manual Scram 2 (1) ;
l
- 3. Power supply source for each channel MG Set (2,6)
- 4. Operation mode
- De-energize to trip Yes (1,6) !
- 5. Logic arrangement
- one-out-of-two twice Yes (6)
- 6. Electrical Protection Assemblies (EPAs) Yes (6)
- 7. Design requirement IEEE-279 (6) ,
The numbers shown in the Data Source column refer to the documents listed in Section I. A-4
l MDE-83-0485-NP ! REV 3 ' GENERAL ELECTRIC COMPANY Section II - RPS Configuration Data
- 8. RPS Sensors
- 1. Identify the type, total number, and number per RPS channel for the following RPS sensors. ;
Total Number / Data l Iygg gym 19r RPS Channel Source j
- APRM Analog 6 2 (1,5) l
- Turbine Stop Valve Switch 8 2 (1,5) l
-l
- Turbine Control Valve Switch 4 1 (1,5) l
- MSIV Position Switch 8 4 (1,5) ;
- MSL Radiation Gamma 4 1 (1,5) l Detector !
- Level 8 (High WaterLevel) N/A N/A N/A N/A
- Level 3 (Low I Water Level) Analog 4 1 (1,5)
- SDV Level !
Type 1 (Analog) Analog 4 1 (1,5) 1 Type 2 (Switch) Switch 4 1 (1,5)
- High Reactor i Pressure Switch 4 1 (1,5) !
- High Drywell l Pressure Switch 4 1 (1,5) l
- Manual Trip Switch 4 1 (1,5)
- Mode Switch Trip Switch 1 1 (1,5)
- Low Condenser Vacuum N/A N/A N/A (1,5) l
- Low Scram Air Header Pressure N/A N/A N/A (1,5)
- Low CRD Charging Analog 4 1 (1)
Water Header Pressure ) l A-5
MDE-83-0485-NP REV 3 GENERAL ELECTRIC COMPANY Section II - RPS Configuration Data Data B. RPS Sensors (Cont'd) Source
- 2. Turbine Stop Valve closure logic arrangement Closure of 3 out of 4 valves initiates scram (1)
- 3. Turbine Stop Valve closure monitoring ,
Position switches (6)
- 4. Turbine Control Valve fast closure monitoring 011 Pressure Switches (6)
- 5. MSIV closure logic arrangement Closure of 3 out of 4 steamlines initiates scram (1)
- 6. Diversity in SDV level sensors Yes (1,5) !
- 7. Number of MSL 4 (6) l
- 8. List of available bypasses (5)
IRM Trip Bypass No
- Noncoincident Neutron Monitoring System l
]
Trip Bypass No j
- RPV High Water Level RPS Trip Bypass N/A
- Turbine Stop Valve RPS Trip Bypass Yes
- Turbine Control Valve RPS Trip Bypass Yes
- MSIV Closure RPS Trip Bypass Yes
- SDV High Water Level Trip Bypass Yes Reactor Mode Switch " Shutdown" mode Trip Bypass No )
i A-6 1 l
y MDE-83-0485-NP REV'3 GENERAL ELECTRIC COMPANY Section II - RPS Configuration Data Data C. Sensor Relays Dalg Jgggg
- 1. Types of relays GE Type HFA,HMA,CR2820 (1)
Agastat Type GP,EGP,E
- 2. Number of pairs of contacts per relay in the trip channel 2 (1)
- 3. List type of relay for each RPS sensor (1)
Potter & Brumfield Anastat lifA fal
- APRM x Turbine Stop Valve x Turbine Control Valve x fiSIV Position x MSL Radiation x x Level 3 x SDV Level x Type 1 (Analog) x Type 2 (Switch) x - High Reactor Pressure x High Drywell Pressure x x Manual Trip x Mode Switch Trip x x x Low Condenser Vacuum N/A Level 8 N/A Low Scram Air Header Pressure N/A Low CRD Charging Water Header Pressure x A-7
J MDE-83-0485-NP REV 3 GENERAL ELECTRIC COMPANY 1 1 Section II - RPS Configuration Data Data i D. Scram Contactors Data Source i
- 1. Type of scram contactors GE Type CR105, (1)
~
CR205, and CR305
- 2. Total number of scram contactors 8 (1) l
- 3. Number of scram contactors per channel 2 (1) ;
i
- t Manual scram channel shares the same scram contactors as its corresponding. j auto scram channel. _
r S i I h i I r A-B
t MDE-03-0485-NP REV 3 : GENERAL-ELECTRIC COMPANY ; t
}
t Section II - RPS Configuration Data r i 5 Data E. Air Pilot Solenoid Valves ErtA Source _. i i ! 1. Number of solenoid valves per control rod drive 7 (2) i
- 2. Number of solenoid operators per *: dive 1 (2) j l
i e i I l' A-9 i
MDE-83-0485-NP REV 3 i GENERAL ELECTRIC COMPANY i l Section II - RPS Configuration Data i Data , F. Backuo Scram Q111 Source !
- 1. Type of scram contactors I for Backup Scram Valves GE Type CR105, 11) ,
CR205, and CR305" l
- 2. Number of scram contactors -
per Backup Scram Valve 4 (1) ;
- 3. Same RPS scram contactors are ,
used to actuate Backup Scram Valves Yes (1) i
- 4. Operator mode
- energized to trip Yes (1)
- 5. Test requirement for i Backup Scram Valves Not specified in Tech. Spec. (1) e A-10 i
}
MDE-83-0485-NP REV 3 GENERAL ELECTRIC COMPANY t Section II - RPS Configuration Data i Data , G. RPS Tech. Soec. Reauirements Source
- l. Calibration Frequency for LPRM :
At least once per 1000 Effective Full Power Hours 47)
- 2. Calibration frequency for trip units.
Every refueling outage except for Reactor High Pressure and Drywell High Pressure trip units calibration done every 3 months (7)
- 3. Frequency of Logic System Functional Tests Every 18 months (7) :
- 4. Allowable time to place an inoperable channel or ,
trip system in the tripped conditions when the j number of operable channels is less than the required minimum operable channels per trip system. I hour (7)
- 5. Exception to item 4.
Restore to operable status within 2 hrs (7) '
- 6. Allowable time to place a trip system in the tripped conditions when the number of operable channels is less than the required minimum operable channels for both trip systems.
I hour (7) '
- 7. Exception to item 6 due to surveillance test.
May be in inoperable status up to 2 hrs two hours for surveillance (7) ;
- 8. Complete the Table on the following page.
A-ll l
MDE-83-0485-NP REV 3 GENERAL ELECTRIC COMPANY REACTOR PROTECTION SYSTEM INSTRUMENTATION REQUIREMENTS Channel Minimum Channel Functional Channel Operable Channels l Check Test Calibration Per Trip System i Generic Plant Generic Plant Generic Plant Generic Plant Functional Unit Model Specific Model SDecific Model Specific Model Specific
- 1. Average Power Range Monitor:
- a. Flow Blased Simulated S,D S,D S/U,W S/U,W W,SA,R W,SA,R 3 2 Thermal Power - High
- b. Neutron Flux - High S S S/U,W S/U,W W SA W,SA 3 2
- c. Inoperative N/A N/A W W N/A N/A 3 2 l2.ReactorVesselSteamDome S N/A M M R Q 2 2 j Pressure - High
- 3. Reactor Vessel Water Level - S S M M R R 2 2 Low, tevel 3
- 4. Reactor Vessel Water Level - S N/A M N/A R N/A 2 N/A High, Level 8
' 5. Main Steam Line Isolation N/A N/A M M R R 4 4 Valve - Closure
- 6. Main Steam Line Radiation - S S M M R R 2 2 High
- 7. Drywell Pressure - High 5 N/A N M R Q 2 2
- 8. Main Condenser Vacuum - Low N/A N/A N/A N/A N/A N/A N/A N/A i
A-12
MDE-83-0485-NP REV 3 GENERAL ELECTRIC COMPANY REACTOR PROTECTION SYSTEM INSTRUMENTATION REQUIREMENTS Channel Minimum Channel functional Channel Operable Channels Check Test Calibration Per Trip System Generic Plant Generic Plant Generic Plant Generic Plant Functional Unit Model Specific Model SDecific Model Specific Model Specific
- 9. Scram Discharge Volume Water Level - High Type 1 - Analog S N/A M M R R 2 2 Type 2 - Switch N/A N/A M M R R 2 2
- 10. Turbine Stop Valve - Closure N/A N/A M M R R 4 4
- 11. Turbine Control Valve Fast N/A N/A M M R R 2 2 Closure Valve Trip System 011 Pressure - Low
- 12. Reactor Mode Switch N/A N/A R R N/A N/A 2 1 Shutdown Position
- 13. Manual Scram N/A N/A M M N/A N/A 2 I
- 14. Low Air Header Scram N/A N/A N/A N/A N/A N/A N/A Pressure - Low N/A i
l A-13
l MDE-83-048ENP REV 3 GENERAL ELECTRIC COMPANY '1 REACTOR PROTECTION SYSTEM INSTRUMENTATION REQUIREMENTS Channel Minimum Channel Functional Channel Operable Channels
, Check Test Calibration Per Trip System Generic Plant Generic Plant Generic Plant Generic Plant Functional Unit Model Specific Model Specific Model Specific Model Specific
- 15. Control Rod Drive:
- a. Charging Water Header N/A N/A N/A M N/A R N/A 2 Pressure - Low i
- b. Delay Timer N/A N/A N/A M N/A R N/A 2 S - Shift M - Monthly S/U - Startup D - Daily Q - Quarterly N/A - Not Applicable i W = Weekly R - Refueling SA - Semi Annual 1
l A-14
MDE 83-0485-NP REV 3 GENERAL ELECTRIC COMPANY l Section II - RPS Configuration Data Data H. RPS Surveillance Tests Procedure Source !
- 1. The following components are all tested as part (8) ,
of an individual channel functional test:
- a. Individual channel sensor (s), e.g., Transmitters and Trip Units, switches, flux or radiation -
l sensors. -
- b. Associated logic relay (s)
- c. Associated scram contactors List any plant specific differences from the above. l
RESPONSE
We do not verify bypassed during functional, only during logic test. Also we don't functionally test rad detectors or flux detectors themselves. (Just electronic trip units).
- 2. When an individual sensor channel is in test or repair, (8) is associated logic channel tripped or is the sensor channel jumpered? State which of the two conditions applies to your plant. If any other condition exists in your plant, describe.
RESPONSE
The instrument, when in test, is not functional (i.e. valved out, out of operation, etc.). However the channel may be tripped (1/2 scram) or untripped throughout the course of the , surveillance. In general I p.:ess jumpered would apply although we do not jumper :.ensors during surveillance. During , surveillance / repair, the Tech. Spec. time clock would apply ' and require tripping the channel within 2 hours (surveillance) or 1 hour (repair). Up to this time the channel may or may not be tripped. A-15
MDE-33-0485-NP REV 3 GENERAL ELECTRIC COMPANY I Section II - RPS Configuration Data (Cont'd) Data H. RPS Surveillance Tests Procedure (Cont'd) Source
- 3. For those plants which do not place individual channels (8) in a tripped condition during test or repair, it is assumed in the GE analysis that only the individual sensor and associated logic relay is placed in an inoperable condition during test or repair of the ~
individual channel. If this assumption is not true for your plant, list the components (from sensor to - scram contactors) which are placed in inoperable condition during test or repair.
RESPONSE
This is true, only the channel is affected.
- 4. The following number of individual scram (8) contactor actuations are assumed in the GE analyses for each channel functional test:
- a. APRM channel functional tests -
2 actuations per scram contactor pair in each trip logic channel.
- b. MSIV closure channel function tests -
4 actuations per scram contactor pair in each trip logic channel,
- c. Other channel functional tests - '
1 actuation per scram contactor pair in each trip logic channel. List any differences from the above for your specific plant.
RESPONSE
4.a. - LSCS - 4/contactor pair 4.b. - True - 4 4.c. - Normally true A-16 l
MDE.83-0485-NP REV 3 GENERAL ELECTRIC COMPANY Section II - RPS Configuration Data (Cont'd) Data H. RPS Surveillance Tests Procedure (Cont'd) Source
- 5. Do plant procedures allow simultaneous (8) inoperable conditions (failed condition) of diverse sensors in a given logic channel?
RESPONSE _
~
Yes within the limit of the Tech. Spec. time clocks described above. A-17
l l MDE-83-0485-NP REV 3 GENERAL ELECTRIC COMPANY l l 1 1 l Section III - Assessed Reliability Effect of RPS Configuration Differences Plant Specific Assessed Reliability BWR Generic Model Difference Effect A. RPS System
- 1. Generic model has No plant specific _
two trip systems, difference .
- 2. Generic model has No plant specific two logic channels difference per trip system for automatic scram.
- 3. During operation, No plant specific the trip systems are difference energized and trip when de-energized.
- 4. The RPS logic is one- No plant specific out-of-two twice, i.e., difference one out of two logic channels will trip an individual system and trip of both systems is required for scram.
- 5. Generic model has No plant specific Electrical Protection difference Assemblies (EPAs).
- 6. Each RPS channel can No plant specific be manually tripped difference from the Control Room using the manual scram circuits.
- 7. Generic model has MG No plant specific set power supply, difference A-18
MDE-83-0485NP REV 3 GENERAL ELECTRIC COMPANY Section III - Assessed Reliability Effect of RPS Configuration Differences Plant Specific Assessed Reliability BWR Generic Model Difference Effect B. Sensors
- 1. Generic model has Switches used for Analog Trip Unit / pressure sensors. -
Transmitter for - pressure sensors. lA. Generic model has No plant specific Analog Trip Unit / difference Transmitter for level sensors.
- 2. Minimum number of No plant specific sensors is one per difference RPS channel fir each scram variab' .
- 3. Generic mod". has Six APRM monitors eight APRF ..onitors w/ two APRM shared with two :r RPS by two channels.
channel.
- 4. Stop Valve Closure No plant specific trip logic is a difference reduced three-of-four required for trip.
- 5. Stop Valve Closure is No plant specific monitored by limit difference switches.
A-19
MDE 83-0485-NP REV 3 GENERAL ELECTRIC COMPANY Section III - Assessed Reliability Effect of RPS Configuration Differences Plant Specific Assessed Reliability BWR Generic Model Difference Effect B. Sensors (Cont'd)
- 6. Turbine Control Valve No plant specific _
fast closure is Difference ~ monitored by control oil pressure.
- 7. MSIV closure trip No plant specific logic requires difference isolation of three out of four steam lines to scram.
- 8. Generic model has a No Level 8 trip No significant level 8 (High Reactor effect as demon-Water Level) Trip. strated by analysis in Reference 1.
- 9. Generic model has No plant specific diverse Scram Dis- difference charge Volume (SDV) level sensors.
- 10. Generic model has No plant specific 4 main steamlines. difference
- 11. Generic model does No plant specific not have a direct difference scram on low condenser vacuum.
- 12. Generic model does No plant specific not have a direct difference i scram on low air header pressure. l 1
- 13. Generic model does not CRD Charging Water address CRD Charging Header' Pressure -
Water Header Pressure Low scram signal
- Low scram signal.
A-20 l
l i MDE-83-0485-NP REV 3 ! GENERAL ELECTRIC COMPANY l l l l Section III - Assessed Rei! ability Effect of RPS Configuration Differences I Plant Specific Assessed Reliability BWR Generic Model Difference Effect C. Sensor Relays -
- 1. For all transients GE type HFA,HMA,&
there are at least CR2820, and Agastat two scram variables type GP,EGP, & E with different type relays used for all - logic relays (either scram variables - Agastat or Potter
& Brumfield).
- 2. Each sensor relay Two sets of contact has a single pair pair per sensor of contacts in relay the applicable trip channel.
D. Scram Contactors
- 1. All Scram Contactors Scram Contactors are one type (GE Type are GE Type CR105, CR105). CR205, and CR305
- 2. Eight scram contactors No plant specific (two per RPS channel) difference perform the trip (two per RPS channel) function.
E. Air Pilot Solenoid Valves
- 1. Generic model has Two HCU valves with dual solenoid operators single solenoid for each individual HCU operators. Tripping air pilot valve. De- of both valves is l energizing both sole- required for indi- l noids results in a vidual control rod I scram of the individual scram. !
control rod. ' A-21
MDE.83-0485-NP REV 3 GENERAL ELECTRIC COMPANY Section III - Assessed Reliability Effect of RPS Configuration Differences Plant Specific Assessed Reliability BWR Generic Model Difference Effect F. Backup Scram
- 1. Actuation of backup No plant specific ~
scram valves are con- difference ' trolled by same output scram contactors as RPS
- 2. Trip logic for backup No plant specific scram valves is an difference energized to trip versus de-energized to trip for individual HCU air pilot valves.
1
- 3. Backup scram valves No requirement are tested during data available 4 shutdown at least once per 18 months. l r
l i A-22 i
MDE.83-0485.NP REV 3 GENERAL ELECTRIC COMPANY Section Ill - Assessed Reliability Effect of RPS Configuration Differences . Plant Specific Assessed Reliability BWR Generic Model Difference Effect G. Technical Specifications 4 and Surveillance Test ! Procedure:
- 1. Generic model uses See Section II.G -
BWR6 Standard Tech- (f this Appendix nical Specifications for plant specific , which requires: differences. Allowable out-of-service time: I hr Test time: 2 hrs . Test frequency: IW for APRM IM for others Calibration frequency: IM for trip units R for transmitters
- 2. Generic model assumes Four actuations 2 actuations per scram per scram contactor contactor pair in each pair for both APRM trip logic for the and MSIV Closure MSIV closure channel channel functional functional test, and test and one-one actuation for the actuation for the other scram variables. other scram vari- l This leads to 272 total ables. This leads to actuations of each 352 total actuations scram contactor per of each scram ,
year. contactor per year. A-23
)
I ATTACHMENT J ; NON-PROPRIETARY VERSION OF THE GENERAL ELECTRIC TOPICAL REPORT I TECHNICAL SPECIFICATION INPROVEMENT ; ANALYSIS FOR THE EMERGENCY CORE COOLING j SYSTEM ACTUATION INSTRUMENTATION FOR ! LASALLE COUNTY STATION, UNITS 1 AND 2
- RE-025, REV 1, DRF C71-00072-1, APRIL 1991 4
l l 4 I [ t t ki\nle\ lama 13e\actstii.wpf83 ,
C RE-025-NP Rev 1
-DRF C71-00072-1 ;
APRIL 1991 GENERAL ELECTRIC COMPANY l - i TECHNICAL SPECIFICATION IMPROVEMENT ANALYSIS FOR THE EMERGENCY CORE COOLING _ SYSTEM ACTUATION INSTRUMENTATION FOR. . . LASALLE COUNTY STATION, j UNITS 1 AND'2 i (THIS REPORT HAS BEEN PREPARED FOR COMMONWEALTH EDISON COMPANY THROUGH-THE TECHNICAL SPECIFICATION IMPROVEMENT COMMITTEE ' OF THE BWR OWNERS' GROUP) t C. Ha R. B. Ninomiya : f f/ V APPROVED BY: ,
. E. Rogers, Manager Reliability Engineering Services j GE NUCLEAR ENERGY ;
SAN JOSE, CALIFORNIA 4
i i GENERAL ELECTRIC COMPANY REV 1 l R E-025-NP i The information contained in this document is furnished for the purpose of providing the members of the BWR Owners' Group with plant specific analysis related to changes to the Emergency Core Cooling System actuation instrumentation Technical Specification testing intervals _ and allowable out-of-service times. No other use, direct or indirect.. of the document or the information it contains is authorized. The information shall not be reproduced or furnished to third parties or made public without the prior express written consent of the General Electric Company. IMPORTANT NOTICE REGARDING CONTENTS OF THIS REPORT Please Read Carefully The only undertakings of General Electric Company respecting information in this document are contained in the contract between the purchasing customer and the General Electric Company as referenced in General Electric Proposals Number 355-1525, Revisions 1 and 2, and nothing contained in this document shall be construed as changing the contract. The use of this information by any who has not contracted for its use for any purpose other than that for which it is intended, is not authorized; and with respect to any unauthorized use, General Electric Company makes no representation or warranty, and assumes no liability as to the completeness, accuracy, or usefulness of the information contained in this document.
-i-
+
GENERAL ELECTRIC COMPANY REV 1 } RE-025-NP i TABLE OF CONTENTS Et91
- 1. INTRODUCTION 1
- 2. EVALUATION NETH00 2 .
- 3. RESULTS OF ECCS EVALUATION 4
- 4.
SUMMARY
AND CONCLUSIONS 7
- 5. REFERENCES 8 APPENDIX A: ECCS ACTUATION INSTRUMENTATION A-1 EVALUATION FOR THE LASALLE COUNTY STATION, UNITS 1 AND 2 1
1 GENERAL ELECTRIC COMPANY REV 1 R E.025-NP 4
'l.. INTRODUCTION This report extends the generic study of modifying the technical i specification requirements of the emergency core cooling system (ECCS) on a plant specific basis for'LaSalle County Station, Units 1 and 2, BWR Ss.
The generic study (References 1 and 2) provides a technical basis to modify the surveillance test intervals and allowable out-of-service times of the . ECCS actuation instrumentation from those of the generic technical I
~
specifications. The generic study also provides additional analyses of , ( various known different ECCS configurations to support the application of the generic basis on a plant specific basis. The generic basis'and the supporting analyses were utilized in this plant specific evaluation. The results of the plant specific evaluation for LaSalle are presented herein. s This report represents the latest plant configuration of LSCS as of December 1990. i l l l l 1
GENERAL ELECTRIC COMPANY REV 1 1 RE-025-NP l
'2. EVALUATION METHOD I The plant specific evaluation of the modification of the surveillance test frequencies and allowable out-of-service times of the ECCS actuation. ,
- instrumentation was performed in the following steps
- :
- a. Gather plant specific information on the ECCS from Commonwealth Edison Company (CECO). The information includes the following: ,
~
(1) Piping and. Instrumentation Diagrams (P&lDs) of ECCS, reactor core isolation cooling (RCIC) system, emergency service water systems, and air systems to ADS valves. : (2) Elementary Diagrams of the ECCS, RCIC, and related systems. (3) ECCS, RCIC and electric power distribution system descriptions such as those in the plant Final Safety Analysis Report (FSAR). (4) Technical specifications on the ECCS, RCIC, the suppression chamber, and the electrical systems. (5) Information on ECCS surveillance test procedures. (6) Dependency matrices showing dependence of ECCS and RCIC systems on support systems and on actuation instrumentation. (7) Available data on actuation instrumentation failures. The latest revisions of the above items were supplied by CECO. Section I of the checklist in Appendix'A was used to identify the data source of the plant specific information.
- b. Construct the plant specific ECCS configuration from the plant specific information. Sections "A" through "E" in Section II of the Appendix A checklist was used for this process.
- c. Compare the plant specific ECCS configuration with the generic ECCS configuration using the generic ECCS fault trees, ECCS description, technical specification requirements, and other 2-
GENERAL ELECTRIC COMPANY. REV 1 RE-025-NP 1
. . l generic inputs.: Section III of the checklist was used for this !
process. l r
- d. -Classify the differences in ECCS system design, in support systems, and in' instrumentation, into three categories:
(1) Differences which obviously have no negative effect on the reliability of the ECCS. Examples of these "no effect" _ items are component name differences, symbol differences, , and other minor non-functional differences. Other effects , not requiring analysis are those in which the specific plant' has greater redundancy than the generic model. . Disposition: of the items with obviously no negative effect is done with "no analysis required". (2) Differences which require engineering judgement for ; disposition because of the functional differences. Examples 'I of these differences are the use of shared room cooling , systems in a specific plant compared with individual room l cooling systems in the generic plant. The disposition of ! such items would require engineering assessment in a " simple { study" as shown in Appendix F of Reference 2. : (3) Differences which require additional analyses'to evaluate , the effect on the ECCS reliability. Examples of such. I differences are the use of two diesel generators and two electrical systems in a specific plant compared with a ; larger number of diesel generators and electrical systems in the generic evaluation. Disposition of these items would require additional analyses (" Modify fault trees and perform analysis.") to compare with the generic results. These 4 analysis are documented in Reference 2.
- e. Compile a list of plant specific differences of Categories (2) and (3).
l 3 j i
.r.- , .__ ...,_.--.,__m.,_m __m . , - _..~.._._,m_._,,_._.._.. . . . . , _ _ _ , . _ , . . . , _ _ _ , . _ _ , _ . , _ . ,,
GENERAL ELECTRIC COMPANY REV 1 R E-025-NP 1
- f. Assess the reliability effect of the differences identified in Step (e) on the generic results. The results of the assessment are documented in Section III of the checklist, Appendix A.
- g. Document the results of the plant specific evaluation.
The above seven step process is dor.umented in Appendix A of this report.
- 3. RESULTS OF ECCS EVALUATION The results of the plant specific evaluation of the ECCS for LSCS are documented in Appendix A of this report. The results show that the ECCS and support systems configuration of LSCS has three differences from the BWR 5/6 generic model* which are classified Category (3), and one which is in Category (2).
3.1 Detailed Analyses The LSCS differences in Category (3), requiring detailed analysis, are as follows:
- a. The generic model has the LPCI and LPCS injection valve permissive signal from RPV pressure, using one-out-of-two-twice logic; LSCS uses 1/2 RPV pressure signals plus 1 valve pressure l signal for this permissive.
l
- b. The generic model has no ADS inhibit switch, LSCS has an ADS inhibit switch,
- c. Injection valves in the generic model are stroke tested quarterly, at LSCS the valve stroke test is performed at refueling, which could be as long as 18 months. l
)
The term " generic model" means the ECCS configuration used in the generic analysis. l
.n %<
GENERAL ELECTRIC COMPANY REV 1 R E-025-N P l l
)
l l i I
GENERAL ELECTRIC COMPANY REY l R E-025-NP 1 l 3.2 Simole Studies , The LSCS difference in Category (2), requiring a simple study, is as follows: l i
- a. The generic model has one diesel generator for each electrical i division, LSCS had two dedicated DGs at each unit, with a fifth 4 DG shared between the two units.
1 l
"The diesel-generator sets have ample capacity to supply all power required for the safe shutdown of both units in the event of a total loss of offsite power. Ample capacity is provided for the condition i in which one unit may be involved in a loss-of-coolant accident while j the remaining unit is being shut down without loss of coolant, as well as for the condition in which both units are concurrently being shut j down without loss-of-coolant accidents." !
I i
GENERAL ELECTRIC COMPANY REV 1 RE-025-NP
- 4.
SUMMARY
AND CONCLUSIONS A plant specific evaluation of modifying the surveillance test intervals and allowable out-of-service times of the ECCS from the technical specifications of LSCS has been performed. The evaluation utilized the plant specific information supplied by CECO and the teneric basis and the additional analyses documented in References 1 and 2. The results indicate that the ECCS configuration for LSCS is similar to the ECCS configuration in the generic evaluation, with four significant differences. The differences between LSCS and the generic model have been modeled by envelope cases SA and SC of Reference 2, plus one simple study, which show that the proposed changes to ECCS actuation instrumentation Technical Specifications would meet the acceptance criteria in Reference 2. Therefore, the generic basis in References I and 2 is applicable to LSCS. 7-
GENERAL ELECTRIC COMPANY , REV 1 RE-025-NP t
- 5. REFERENCES (1) D.,B. Atcheson, et al., "BWR Owner's Group Technical Specification Improvement Methodology (with Demonstration for BWR ECCS Actuation Instrumentation) Part 1", General Electric Company, NEDC-30936P-A, December 1988.
(2) D. B. Atcheson, et al., "BWR Owner's Group Technical _ Specification Improvement Methodology (with Demonstration for - BWR ECCS Actuation Instrumentation) Part 2", General Electric Company, NEDC-30936P-A, December 1988. t b t 8-
I b
..t GENERAL ELECTRIC COMPANY REV 1 i RE-025-NP t i
t
)
r l 1 APPENDIX A .. ECCS ACTUATION INSTRUMENTATION EVALUATION FOR LASALLE COUNTY STATION, 'l UNITS'l AND 2 i l i 1 i A-1
GENERAL ELECTRIC COMPANY REV 1 RE-025-NP ^ Section I - ECCS Plant Specific Data Source Utility: Commonwealth Edisce Co. Plant: LaSalle County Station, Units 1 & 2 Source Number
- 1. ECCS and RCIC P& ids
- 2. Emergency Service Water P& ids .
- 3. Electrical Drawings
- 4. Instrumentation Logic Diagrams
- 5. ECCS Fault Trees
- 6. Final Safety Analysis Report
- 7. Technical Specifications
- 8. Other Drawings
- 9. Dependency Matrices
- 10. Failure Data I 1
- 11. Test Procedure Questionnaire '
- 12. Telephone Call Records l
- 13. NEDC-30936P-A, Part 1 I
A-2 l
-- l
GENERAL ELECTRIC COMPANY REV 1 ' R E-025-NP l Section II - ECCS Configuration Data f A. ECCS System Generic Difference Data * , LES BWR 5/6 lyMil Source
- 1. Number of:
LPCS Pumps / Loops 1/1 1/1 N 1 l LPCI Pumps 3 3 N 1 ADS Valves 7 8 Y l _ HPCS Pumps 1 1 N 1
- 2. Needed for Success, Number of:
LPCS Pumps / Loops 1/1 1/1 N 6 , LPCI Pumps 1 1 N 13 i ADS Valves 3 3 N 13
- 3. Number of:
Diesel Generators 3** 3 Y 3 Electrical Divisions 3 3 N 3 The numbers shown in the Data Source column refer to the documents listed in Section 1.
** 2 Dedicated diesel generators per unit, 1 DG is shared between Units 1 & 2 L A-3
- r. . ;
i L, GENERAL ELECTRIC COMPANY REV 1 RE 025-NP Section II - ECCS Configuration Data B. SUPPORT SYSTEM DEPENDENCIES The dependencies each front line ECCS system has on the listed support ' subsystems for both the generic and specific plant are shown. FRONT LINE SYSTEMS SUPPORT \ ---LPCI--- ADS ADS DIESELS SUBSYSTEMS \A B C LPCS A B RCIC HPCS A B C OFFSITE AC POWER X X X X X X ONSITE AC POWER DIVISION 1 X X X X r DIVISION 2 X X X X DIVISION 3 X ONSITE DC POWER , DIVISION 1 X X X X X j DIVISION 2 X X X X X > DIVISION 3 X X ! SERVICE WATER . EMERGENCY LOOP A X X X X EMERGENCY LOOP B X X X EMERGENCY LOOP C X X WATER SUPPLY CONDENSATE TANK X X SUPPRESSION POOL X X X X X X AIR CONTROL DIV 1 X ! CONTROL DIV 2 X CONTAIN. INSTR. DIV 1 X ! CONTAIN. INSTR. DIV 2 X : ROOM COOLING t LPCI X X X LPCS X RCIC X HPCS X DIESELS X X X X - IN BOTH GENERIC AND SPECIFIC BWR 5/6s l G = ONLY IN GENERIC BWR 5/6 S - ONLY IN SPECIFIC BWR 5/6 :
* - DG IS SHARED BETWEEEN UNITS 1 & 2 l
A-4
GENERAL ELECTRIC COMPANY REV 1 R E-025-NP l 1 Section II - ECCS Configuration Data C. INSTRUMENTATION DEPENDENCIES The dependencies each front line ECCS system has on the listed actua- i tion instrumentation for the generic and specific plants are shown. i FRONT LINE SYSTEMS ACTUATION \ LPCI LPCI LPCI ADS ADS' INSTRUMENTATION \ A B C LPCS A B RCIC HPCS
~
RPV WATER LEVEL 1 (LOW LOW LOW) 821-N707 A/C X X X ; B21-N707 B/D X X X RPV WATER LEVEL 2 (LOW LOW) i B21-N710 A/B/C/D X B21-N706 A/B/C/D X RPV WATER LEVEL 3 (LOW) 821-N708A X B21-N708B X ; RPV WATER LEVEL 8 (HIGH) 821-N709 A/B X B21-N705 A/B X RPV PRESSURE LOW , 821-N413 A/C X X B21-N413 B/D X X N698 A/E G G N698 B/F G G E21-N413 S E12 N413A S E12-N413B S E12-N413C S X - IN BOTH GENERIC AND SPECIFIC BWR 5/6s G - ONLY IN GENERIC BWR 5/6 S - ONLY IN SPECIFIC BWR 5/6 A-5
GENERAL ELECTRIC COMPANY REV 1 R E-025-NP , Section II - ECCS Configuration Data C. INSTRUMENTATION DEPENDENCIES (Continued) The dependencies each front line ECCS system has on the listed actua-tion instrumentation for the generic and specific plants are shown. FRONT LINE SYSTEMS ACTUATION \ LPCI LPCI LPCI ADS ADS INSTRUMENTATION \ A B C LPCS A B RCIC HPCS
~
DRYWELL PRESSURE HIGH B21-N048 A/C X X X B21-N048 B/0 X X X LPCI PUMP DISCHARGE PRESSURE HIGH E12-N016 A/N019 A X E12-N016 B/C X L E12-N019 B/C X LPCS PUMP DISCHARGE PRESSURE HIGH > E12-N001/N009 X ADS .'IMER K3SA X K35L' X DRYWELL DRESSURE BYPASS TIMER , K36A/K37A X l K36B/K378 X ADS INHIBIT SWITCH S26A S S26B S l MANUAL INITIATION SWITCH (1/ LOOP) X X X X X X X X l X - IN BOTH GENERIC AND SPECIFIC BWR 5/6s G - ONLY IN GENERIC BWR 5/6 S - ONLY IN SPECIFIC BWR 5/6 A-6
GENERAL ELECTRIC COMPANY REV 1 R E-025-NP ; l l i Section II - ECCS Configuration Data j C. INSTRUMENTATION DEPENDENCIES (Continued) The dependencies each front line ECCS system has on the listed actua-tion instrumentation for the generic and specific plants. FRONT LINE SYSTEMS RELATED NON-ACTUATION \ LPCI LPCI LPCI ADS ADS - INSTRUMENTATION \ A B C LPCS A B RCIC _HPCS LPCI/LPCS PUMP DISCHARGE FLOW LOW E12-N010 AA X BA X CA X E21-N004 X CST LEVEL LOW E51-NO35 A/B X E22-N001 A/B X SUPPRESSION POOL - WATER LEVEL HIGH N655 C/G G E22-N002 A/il X X - IN BOTH GENERIC AND SPECIFIC BWR 5/6s G = ONLY IN GENERIC BWR 5/6 S - ONLY IN SPECIFIC BWR 5/6 l i I l A-7
GENERAL ELECTRIC COMPANY REV 1 R E-025-NP Section II - ECCS Configuration Data , D. Minimum Number of Sensors, Channels, or Components for Failure, LS 1 & 2 l A: - MINIMUM NUMBER SENSOR FAILURES REQUIRED TO FAIL TRIP FUNCTION
- B: - MINIMUM NUMBER SENSOR FAILURES REQUIRED TO FAIL FUNCTION - TOTAL C: - MINIMUM NUMBER OF SENSOR TYPES REQUIRED TO FAIL FUNCTION DIFFERENT FROM GENERIC TRIP (Y/N)
FUNCTION A B C B C -- LPCS PUMP 1 RPV WATER LEVEL 1 (LOW LOW 2 2 N N INITIATION LOW) AND 1 DRYWELL PRESSURE LPCS INJ VALVE 1 RPV LOW PRESSURE 1 1 Y N LPCI PUMP 1 RPV WATER LEVEL 1 AND 2 2 N N INITIATION 1 DRYWELL PRESSURE LPCI INJ VALVES 3 RPV LOW PRESSURE 3 1 N N ADS INITIATION 2 RPV WATER LEVEL 1 OR 2 1 N N 2 RPV WATER LEVEL 3 (LOW), OR 2 PUMP DISCH PRESS, OR 2 DRYWELL PRESS ADS TIME DELAY 2 TIMERS 2 1 N N HPCS 2 RPV LEVEL 2 (LOW LOW) 4 2 N N INITIATION AND 2 ORYWELL PRESSURE HPCS LEVEL 8 2 RPV LEVEL 8 (HIGH) 2 1 N N HPCS INJ VALVE 2 RPV LEVEL 2 AND 4 2 N N 2 DRYWELL PRESSURE HPCS WATER 2 CST LEVEL AND 2 4 2 Y N SOURCE SUPPRESSION POOL LEVEL RCIC INITIATION 2 RPV LEVEL 2 2 1 N N RCIC LEVEL 8 2 RPV LEVEL 8 2 1 N N RCIC WATER 2 CST LEVEL 2 1 N N SOURCE RCIC INJ VALVE 2 RPV LEVEL 2 2 1 N N Based on data sources 4 & 6. For Level 8, trip function is false isolation of system. i A-8
i i GENERAL ELECTRIC COMPANY REY I R E-025-NP I l 1 Section II - ECCS Configuration Data E. ECCS Instrumentation and related subsystems Surveillance Requirements
- SURVEILLANCE REQUIREMENTS ** DIFFERENCE GENERIC 5/6 LES M C.QE iPlal SYSTEM REACTOR WATER LEVEL I (LOW LOW LOW) M M N DRYWELL PRESSURE HIGH H M N REACTOR PRESSURE LOW M M N MANUAL INITIATION R R N -
~
LK1 REACTOR WATER LEVEL 1 M M N DRYWELL PRESSURE HIGH M M N REACTOR PRESSURE LOW H H N PUMP START TIME DELAY RELAY M M N INJECTION VALVE DIFFERENTIAL PRESSURE LOW M N/A Y MANUAL INITIATION R R N lip _C1 REACTOR WATER LEVEL 2 (LOW LOW) M M N ORYWELL PRESSURE HIGH M M N CST LEVEL LOW M M N SUPPRESSION POOL LEVEL HIGH M M N REACTOR WATER LEVEL 8 M M N 1 MANUAL INITIATION R R N l ElC REACTOR WATER LEVEL 2 (LOW LOW) M M N REACTOR WATER LEVEL 8 M M N MANUAL INITIATION R R N ADS REACTOR WATER LEVEL 1 M M N DRYWELL PRESSURE HIGH M M N ADS TIMER M M N CORE SPRAY PUMP DISCHARGE PRESSURE M M N 1 LPCI PUMP DISCHARGE PRESSURE M M N REACTOR WATER LEVEL 3 (LOW) M N I MANUAL INITIATION R 11 ADS DRYWELL PRESSURE BYPASS TIMER M 'M N ADS INHIBIT SWITCH N/A R Y Based on Technical Specifications, data source No. 7.
- M - MONTHLY, W = WEEKLY, - REFUELING, R Q - QUARTERLY CSD - COLD SHUT DOWN A-9
_______j
GENERAL ELECTRIC COMPANY REV 1 RD025-NP Section II - ECCS Configuration Data E. ECCS Instruraentation and related subsystems Surveillance Requirements * (Continued) SURVEILLANCE REQUIREMENTS ** DIFFERENCE t GENERIC 5/6 L1G1 .(,Y/J1 INJECTION YALYI STROKE 1111 Q CSD/R Y DIESEL GENERATOR H M N -. ELECTRIC POWER ESSENTIAL AC W W N ESSENTIAL DC W W N ESSENTIAL AC BUSSES W W N Based on Technical Specifications, data source No. 7.
- M - MONTHLY, W = WEEKLY, - REFUELING, Q - QUARTERLY R
CSD - COLD SHUT DOWN I l A-10 I
GENERAL ELECTRIC COMPANY REV 1 R E-025-NP Section III - ECCS Configuration Differences Classification (LaSalle County Station) Plant Specific Classification (Justifi-BWR 5/6 Generic Model Difference cation if Insignificant) A. ECCS System Differences
- 1. RC7C and HPCS have LSCS has independent -
common portion of supply lines from CST . supply header from for RCIC and HPCS. CST.
- 2. Generic has 8 ADS LSCS has 7 ADS valves, valves.
B. Support System Differences -
- 1. Onsite AC power Each unit has 2 inde-has 3 independent pendent divisions, Divs divisions. 2 & 3. Div 1 for each unit is from the shared diesel generator.
C. Instrumentation and Procedures Differences
- 1. Opening of LPCS & Opening of low pres-LPCI injection sure system injection valves has RPV valves requires one pressure permis- valve pressure signal sive signal from plus 1/2 RPV pressure 1/2-twice logic. signals.
- 2. Containment spray No containment spray signal could pre- signal interlock.
vent LPCI operation.
- 3. Injection valves Injection valve stroke are stroke tested tests are performed at quarterly. cold shutdown.
- 4. No /SS inhibit ADS has inhibit switch.
switch. A-11 i J
GENERAL ELECTRIC COMPANY REV 1 R E-025-NP Section III - ECCS Configuration Differences Classification (LaSalle County Station) Plant Specific Classification (Justifi-BWR 5/6 Generic Model Cifference cation if Insignificant) C. Instrumentation and Procedures Differences (Cont'd) i
- 5. Analog trip units LSCS uses process -
are used, switches and analog - trip units in sensor channels. A-12
l l ATTACHMENT K l GENERAL ELECTRIC LETTER TO R.H. MIROCHUA (COMMONWEALTH EDISON COMPANY), TECHNICAL SPECIFICATION IMPROVEMENT FOR BWR INSTRUMENTATION TRANSMITTAL OF DELIVERABLES LASALLE COUNTY STATION EBO-90-246, MAY 1, 1991 I l l l l I l l l I l 1 j 1 i I l l l l i i ki\nla\lasalle\actstii.wpt84 l 1 1
., .~ - . . . . . .-
O - GE Nuclear Energy E80-90-246 2Yle's?zN5$?."s re 20t coa swa aoszt 7CES?]2929 May 1, 1991 Mr. R. H. Mirochna Commonwealth Edison Company Executive Towers West III 1400 Opus Place, Suite 400 Downers Grove, IL 60515
SUBJECT:
TECHNICAL SPECIFICATION IMPROVEMENT FOR BWR INSTRUMENTATION TRANSMITTAL OF DELIVERA8LES LA SALLE COUNTY STATION
References:
- 1. Ceco Purchase Order 328658, Release No. NU-77, dated November 13, 1990.
- 2. GE Letter No. EBO-90-414, Same Subject, Budgetary Estimate No. 295-1BFH4-HA0-90, W. Arndt to R. Mirochna, dated October 5, 1990.
Dear Mr. Mirochna:
Enclosed please find the following items which complete GE's contracted scope of work of References 1 and 2:
- 1. Report MDE-83-0485, Revision 3 " Technical Specification Improvement Analysis for the Reactor Protection System for LaSalle County ,
Station Ur.its 1 and 2", dated April 1991.
- 2. General Electric Company Affidavit for Report MDE-83-0485. I
- 3. Report RE-025, Revision 1, " Technical Specification Improvement !
Analysis for the Emergency Core Cooling System Actuation ! Instrumentation for LaSalle County Station, Units 1 and 2", dated l April 1991.
- 4. General Electric Company Affidavit for Report RE-025.
- 5. A draft submittal letter to the NRC.
- 6. Copy of GE Letter No. 0G9-1219 32D, Clarification of Limerick 1 & 2 Proposed Technical Specification Changes Common to Reactor Protection System or ECCS Actuation Instrumentation, W. Sullivan to BWROG Technical Specification Committee, dated December 22, 1989.
- 7. Copy of GE Letter No. 0G90-319-32D, Clarification of Technical Specification Changes Given in ECCS Actuation Instrumentation Analysis, W. Sullivan and J. Klapproth to M. Wohl, dated March 22, 1990.
Report MDE-83-0485, Revision 3, ccepletes Reference 2 scope of work Item 1. ! Report RE-025, Revision 1, completes Reference 2 scope of work Item 2. l i
o s R. H. Mirochna May 1, 1991 To complete the Reference 2 scope. of work Items 3 through 5, GE has verified that tae conclusions of the Licensing Topical Reports of scope of work Items' 3 through 5 are still applicable to LaSalle County Station and that plant specific reports are not necessary. The draft NRC submittal letter contains the results of this verification and is provided in a format which will facilitate Ceco's amendment application. The draft NRC submittal letter is consistent with amendment applications submitted by other plants. In addition to referencing scope of work Items I through 5, the draft NRC submittal letter also addresses the following specific items:
- 1. End-of-Cycle Recirculation Pump Trip (EDC-RPT) System.
- 2. RCIC System.
- 3. Clarification of on the intent of technical specification mark-up for ECCS Actuation Instrumentation. -
- 4. Clarification of RPS Limited Condition of Operation (LCO).
This document transmittal to Ceco fulfills all GE's comettaents and provides all deliverables for the Reference 1 Ceco Purchase Order. GE thanks you for the opportunity to perform these services for Ceco. If you have any questions, please call me, or Ron Kinomiya at '(408) 925-2077, I at your convenience. Sincerely,
//
Will am D. Arndt Senior Customer Service Engineer (708) 573-3964 Attachments cc: LEC.n EE J. W. Gieseker w/o att. J. C. Elliott w/o att. D ~.ET'Lockwood'i G. L. Hayes w/o att. E. Li Seckinger J. E. Kusky J. D. Williams w/o att. R. B. Ninomiya w/o att. Chron System File: 4.Z12.0 e-- . -
i
. 1 4
1 DRAFI l l U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555
SUBJECT:
LaSalle County Station, Units 1 and 2 Technical Specifications Change Request
Dear Sir:
Comonwealth Edison Company hereby submits Technical Specificatkns Change Request No. in accordance with 10 CFR 50.90, requesting an amendment to the Technical Specifications (TS) (Appendix A) of Operating License Nos. and . Information supporting this Change Request is contained in Attachment I to this letter. Attachment 2 provides a list of references used to justify this Change Request, and Attachment 3 provides the proposad TS change pages. This submittal requests changes to the TS to. extend surveillance test intervals and allowable out-of-service items for instrumentation supporting the Reactor Protection System (RPS), Emergency Core Cooling System (ECCS), and Isolation Actuation including comon instrumentation. This letter also submits the following documents which provide additional information supporting this Change Request, as enclosures, a) Enclosure 1 - General Electric (GE) Document No. OG9-1219-32D, letter W. P. Sullivan, GE, to Boiling Water Reactor Owners' Group (BWROG) Technical Specification Committee, dated December 22, 1989:
" Clarification of Limerick 1 and 2 Proposed Technical Specification Changes Common to Reactor Protection System or ECCS Actuation Instrumentation".
b) Enclosure 2 - GE Document No. MFN 024-90, OG90-319-320, letter from W. P. Sullivan, GE, to Millard L. Wohl, NRC, dated March 22, 1990:
" Clarification of Technical Specification Changes Given in ECCS Actuation Instrumentation Analysis)".
c) Enclosure 3 - GE Document No. MOE 83-0485-3, " Technical Specification Improvement Analysis for the Reactor Protection System for LaSalle County Station Units 1 and 2", dated April, 1991. d) Enclosure 4 - GE document No. RE-025-1, " Technical Specification Improvement Analysis for Emergency Core Cooling System Actuation Instrumentation for LaSalle County Station Units 1 and 2", dated April, 1991. DRAFI
DRAIT i Please note that applications and accompanying affidavits in accordance with 10 CFR 2.790(b)(1), to withhold from public disclosure GE Document Nos. . . MDE-83-0485-3, " Technical Specification Improvement Analysis for the Reactor Protection System for LaSalle County Station Units I and 2", dated April, > 1991, and RE-025-1, " Technical Specification Improvement Analysis for , EmergencyCoreCoolingSystemActuationInstrumentationforLaSalleCounty Station Units 1 and 2 , dated April, 1991, are included with Enclosures 3 and 4, respectively. Accordingly, we request that the documents identified above - (Enclosures 3 and 4) be withheld from public disclosure in accordance with - 10 CFR 2.790(a)(4). If you have any questions regarding this matter, please contact us. . Sincerely yours, Attachments - Enclosures t DRAFI ,y,
.____ _m --_~ , ,., - er--,r- -
ATTACHNENT 1 LASALLE COUNTY STATION UNITS I AND 2 DOCKET NOS. LICENSE NOS. l TECHNICAL SPECIFICATIONS CHANGE REQUEST
" REDUCED TESTING 0F REACTOR PROTECTION SYSTEN, ENERGENCY CORE COOLING SYSTEN, ISOLATION ACTUATION AND COPMON INSTRUNENTATION" .
Supporting Information for Changes - 6 pages
O Commonwealth Edison Company (CECO), Licensee under Facility Operating Licenses and for.LaSalle County Station (LSCS), Unit I and l Unit 2, respectively, hereby requests that the Technical Specifications (TS)* - l contained in Appendix A of the Operating Licenses be amended as proposed i herein to extend surveillance test intervals (STIs) and allowable I out-of-service times (A0Ts) for the actuation instrumentation supporting i Reactor Protections System (RPS), Emergency Core Cooling System (ECCS), and i Isolation Actuation, including instrumentation common to the Control Rod Block Function (CR8F), the Reactor Core Isolation Cooling (RCIC) system, i End-of-Cycle Recirculation Pump Trip (E0C-RPT) system, and the isolation , instrumentation common to RPS and/or ECCS. The proposed changes will minimize : unnecessary testing and remove excessive restrictive A0Ts that could : potentially degrade overall plant safety and availability. We request the changes proposed herein to be effective fifteen (15) days after ; issuance of the Amendments. This Change Request provides a discussion and description of the propo::ed TS i changes and a safety assessment of.the proposed TS changes. Discussion and Descrintion of Pronosed Chances Licensing Topical Report (LTR;, "8WR Owners' Group Response to NRC Generic i Letter 83-28, Item 4.5.3", (Reference 1) provided justification for the ; acceptability of current RPS STIs. In addition, Reference 1 established a i basis for extending STIs and A0Ts for RPS based on reliability analyses which l estimate RPS failure frequency. The analyses were further developed in other l LTRs (References 2 through 6) to provide justification for extending TS STIs ' and A0Ts for the RPS, ECCS, and Isolation Actuation including common instrumentation. References 2 through 6 also included proposed TS changes to facilitate implementation of the analyses results. References 2 through 6 were submitted to the NRC by the Boiling Water Reactor Owners' Group (BWROG)
)
and subsequently approved as detailed in NRC Safety Evaluation Reports (SERs). , These SERs describe the acceptability of both the analyses and the proposed TS changes provided to the NRC. In addition, the NRC SERs provided criteria for plant specific hpl.ementation of the generically approved TS changes. Our compliance with these criteria is discussed in the Safety Assessment of this Change Request. The Change Request proposed TS changes to the actuation instrumentation supporting the RPS, ECCS, and Isolation Actuation including instrumentation common to the CR8F and the isolation instrumentation common to the RPS and/or ECCS. These changes are specifically designated in the TS mark-ups of References 2 through 6 and therefore,.are not further discussed here. We are also proposing TS changes to instrumentation common to RPS and/or ECCS, but which are not specifically designated in References 2 through 6. These proposed changes are addressed in the analyses of References 2 through 6, but ; were not specifically designated in the TS mark-ups submitted as part of 1 References 2 through 6. These changes will provide a complete consideration ' l L I
. .- .. - . .. . _ = - - - . _.
of all systems / components initiated by RPS, ECCS, or Isolation Actuation instrumentation which are tested on a monthly schedule and are NRC approved for testing on a quarterly schedule as detailed in References 2 through 6. All changes are shown in Attachment 3. However, only those changes not . specifically designated in References 2 through 6 are described below.
- 1. The E0C-RPT system uses trip functions common to RPS. Therefore, we propose to change the EOC-RPT system STIs and A0Ts on TS pages 3/4 3-39, 3/4 3-41, and 3/4 3-44 to conform to the TS changes made for RPS instrumentation. Enclosure 1 details the fact that the analysis of Reference 2 bounds the proposed TS changes to E0C-RPT.
- 2. The RCIC system uses trip functions comon to ECCS and therefore, we propose changes to TS pages 3/4 3-46, 3/4 3/47, 3/4 3-4g to be consistent with other TS changes for ECCS instrumentation. Changes to these TS were not specifically included in the TS mark-ups provided to the NRC in Reference 5, although they are addressed in the Reference 5 analysis. .
Enclosure 2, GE document No. MFN 024-g0, letter from W. P. Sullivan, GE, to Millard L. Wohl, NRC, dated March 22,1990, " Clarification of Technical Specification Changes Given in ECCS Actuation Instrumentation Analysis", provided mark-ups for these changes to the RCIC system TS to the BWROG TSC.
- 3. The proposed TS changes to TS Section 3/4 3.3.3 on page 3/4 3-27 and 27(a) provide a 24 hour A0T for ECCS instrumentation which is consistent with the analysis in Reference 5. The proposed wording differs from the TS ,
l ' mark-up of Reference 5 which implies an allowance of 24 hours before < taking the action of TS Table 3.3.3-1. Enclosure 2 provides a clarification on the intent of the Reference 5 TS mark-up and also provides revised wording. We have proposed a TS change consistent with Enclosure 2.
- Safety Assessment The effect on safety of the proposed extensions to the STIs and A0Ts of the actuation instrumentation supporting the RPS, ECCS, and Isolation Actuation including the instrumentation common to the CRBF, and the isolation 2instrumentation comon to the RPS and/or ECCS has been addressed in References through 6. Further, the NRC has detailed their acceptance of the analyses '.
and conclusions of References 2 through 6 in SERs (included in References 2 through 6). The SERs conclude that implementation of the TS changes proposed in References 2 through 6 would provide an overall enhancement to plant safety and that the proposed changes to TS are acceptable subject to the Licensee documenting 1) plant-specific applicability, 2) that instrument ~ drift is bounded by the generic analysis assumptions, and 3) confirmation that differences between plant specific and generic RPSs were included in the plant-specific analysis. These acceptance conditions are addressed below.
- 1. A plant-specific review of the LTR's (References 2 through 6) applicability to LSCS has been conducted. For the RPS, the review
i O compared the LSCS RPS configuration and surveillance test procedure with the generic RPS evaluated in the LTR. The differences between the two were identified and the reliability effect of the differences was ; assessed. The differences and their effect are documented in a separate- ' GE report, Enclosure 3, document 83-0485-3, " Technical Specification Improvement Analysis for the Reactor Protection System for LaSalle County l Station Units I and 2", dated April, 1997. The report identifies l differences which were dispositioned by either engineering assessment or i additional analyses. The report concluded that these differences would ' not significantly affect the improvement in plant safety which would be obtained through the TS changes evaluated in the generic analysis and that the generic analysis is applicable to LSCS Units 1 and 2. ! For ECCS, a similar review was conducted. The results are documented in a i separate GE report, Enclosure 4, document No. RE-025-1, " Technical . Specification Improvement Analysis for Emergency Core Cooling System i Actuation Instrumentation for LaSalle County Station Units I and 2," dated . April, 1991. The report concludes that the ECCS configuration for LSCS is ' similar to the generic analysis with only seven differences. The differences were dispositioned by engineering assessment or additional . analyses. The results indicate that the proposed changes to ECCS instrumentation would meet the acceptance criterion of Reference 5, Part 2. The changes to the instrumentation common to the CRBF and the isolation instrumentation common to the RPS and/or ECCS are addressed in References 3 and 4, respectively. These changes are bounded by the l generic analyses in References 2 and 5 and LSCS plant-specific analyses given in Enclosures 1 and 3. i For Isolation Actuation instrumentation, analyses were provided in Reference 6 that bounded the plant-specific differences. Appendix C of l Reference 6 provides the LSCS Unit I and 2 STIs and calibration intervals i that were included in the study. I
- 2. In 1988, the NRC issued additional guidance regarding instrument drift (Reference 7). This letter states that " licensees need only confirm that the setpoint drift which could be expected under the extended STIs has been studied and either (1) has been shown to remain within the existing i allowance in the RPS (for BWRs) instrument setpoint calculation or (2) '
that the allowance and setpoint have been adjusted to account for the additional expected drift". Present setpoint calculations for LSCS are based on an eighteen (18) month calibration interval. Therefore, drift occurring during a three-month STI falls within the existing drift allowance.
- 3. We have reviewed the GE plant-specific report for LSCS and have verified that the differences between the LSCS and generic RPS were included in the plant-specific analysis. Therefore, the generic analysis in Reference 2 is applicable to LSCS.
l As discussed above, we have conformed to the guidance provided in References 2 l through 6 in the three areas to be addressed by Licensees to ensure the
acceptability of proposed TS changes. As noted previously, several changes are also proposed which are not specifically referenced in the NRC SERs given in References 2 through 6. The following discussion addresses the l acceptability of these proposed changes. - '
- 1. The E0C-RPT is initiated by signals common to the RPS. These signals (turbine stop valve closure, and turbine control valve low hydraulic ;
pressure) were not identified as common trip functions in the RPS TS improvement analysis (Reference 2). Although STI changes to the cosmon EOC-RPT trip functions were not explicitly identified in the Reference 2 analysis, the changes can be considered bounded by this analysis. The basis for this conclusion is siellar to the basis established in Reference 3.
- 2. Analysis of the effects of extending A0Ts and STIs for the RCIC system instrumentation was completed and found acceptable as detailed in Reference 5. However, proposed changes to thee TS were not specifically given in Reference 5. This does not affect the acceptability of these proposed changes, since the methods and results of Reference 5 were found ;
acceptable as documented in Reference 5. Recognizing that mark-ups to the RCIC system instrumentation TS had not been previously included in Reference 5, GE provided TS mark-ups for all GE BWR product lines, incorporating the extended STIs and A0Ts for RCIC system instrumentation. The mark-ups were provided in Enclosure 2.
- 3. Also discussed in Enclosure 2 is a clarification of the applicability of I the 24-hour TS A0T for ECCS Actuation Instrumentation. The change provides a 24-hour ACT in those TS Action Statements which are applicable to specific instrumentation. The intent of the change is to preclude the allowance of 24 hours before taking the action specified in TS. Table 3.3.3.1. Action "b" of TS paragraph 3/4 3.3, as written in Reference 5 ,
implies a 24 hour A0T before taking any action in TS Table 3.3.3-1. The ) change we have proposed accurately reflects the intent of the Reference 5 ; analysis. This change, therefore, is necessary to obtain the overall ' enhancement to safety that is possible by extending STIs and A0Ts. References 2 through 6 provided TS changes based on review of the LTRs. 1 We have proposed TS changes consistent with those previously approved and specifically designated in References 2 through 6. In addition, several changes are proposed which are not explicitly referenced in the NRC SERs, but are covered by the analyses detailed in References 2 through 6 and Enclosures 3 and 4, and are acceptable as discussed above. In sumary, the NRC criteria for demonstrating the applicability and acceptability of all proposed changes has been shown to be met, as detailed above. We, therefore, conclude that the changes proposed will minimize unnecessary testing and relax excessively restrictive A0Ts, and will provide an overall enhancement to plant safety. 4
Information Sunnartina a Findina of No Sianificant Hazards Consideration We have concluded that the proposed changes to the LSCS TS, which extend STIs and A0Ts for the RPS, ECCS, and Isolation Actuation instrumentation including instrumentation common to RPS and/or ECCS, do not constitute a Significant Hazards Consideration. In support of this determination, an evaluation of , each of the three standards set forth in 10 CFR 50.g2 is provided below. ; I) The proposed changes do not involve a significant increase in the probability or consequences of an accident previously evaluated. The proposed TS changes increase the STIs and A0Ts for actuation instrumentation supporting RPS, ECCS, and Isolation Actuation including instrumentation common to the CR8F, RCIC system, E0C-RPT, and isolation i functions. There are no changes in any of the affected systems . I themselves. Since there are no such changes, there can be no change in the probability of occurrence of an accident or the consequences of an accident or the consequences of malfunction of equipment. Regarding the - probability of malfunction of equipment, LTRs prepared by GE showed that for the RPS there is a reduction in scram frequency, but that in the ECCS case, there is a small increase in the unavailability of the water injection function. This increase in unavailability was judged acceptable by GE. The NRC in its review of the LTRs (References 2 through 6), concurred with this conclusion. The changes proposed are consistent with these SERs given in References 2 through 6 with several additions. These additional changes are bounded by the analyses of References 2 through 6 as detailed in this Change Request and in Enclosures I and 2. Therefore, the proposed changes do not involve a significant increase in the probability of consequences of an accident previously evaluated.
- 2) The proposed changes do not create the possibility of a new or different i kind of accident from any accident previously evaluated. j l
The proposed TS changes do not create the possibility for an accident or malfunction of a different type than any evaluated previously in the Final Safety Analysis Report (FSAR). The proposed changes increase the STIs and A0Ts for the RPS, ECCS, and Isolation Actuation instrumentation including common instrumentation. There are no changes in the RPS, ECCS, Isolation Actuation or common systems themselves. Since there are no such changes, there is no possibility for an accident or malfunction of a different type than any evaluated previously.
- 3) The proposed changes do not involve a significant reduction in a margin of safety. '
The proposed TS changes do not reduce the margin of safety as defined in the basis for any TS. The proposed TS changes do not change any setpoints in the RPS, ECCS, Isolation Actuation instrumentation, or common systems, or their levels of redundancy. Setpoints are based upon the drift occurring during the 18-month calibration interval. The proposed changes extend STI3 and A0Ts. The bases in the TS either do not discuss ST!s, or ! state "...one channel may be inoperable for brief intervals to conduct
, . - , , - - , , , , , , , , - - , , - - ---,,,-,.,-c--,- - - - - , , - , , , ,
required surveillance." The proposed TS changes discussed in References 2 ' through 6 as well as the additional changes discussed in this Change Request and Enclosures I and 2, are bounded hy the analyses in References 2 through 6. These analyses (References 2 through 6) prepared by GE and - reviewed and approved by the NRC examined the effects of extending STIs and A0Ts and found that the proposed changes would not involve a significant reduction in a margin of safety. Information Sunnartina an Environment Assessment An environmental assessment is not required for the changes proposed by the Change Request because the requested changes conform to the criteria for
" actions eli 51.22(c)(g).gible Thefor categorical exclusion",
will haveasnospecified impact oninthe 10CFR l requested changes environment. 1 The proposed changes do not involve a significant hazards consideration as discussed in the preceding section. The proposed changes do not involve a significant change in the types or significant increase in the amounts of any effluents that may be released offsite. In addition, the proposed changes do not involve a significant increase in individual or cumulative occupational radiation exposure. Conclusion The (To Be Provided by CECO) have reviewed these proposed changes to the TS-and determined that they do not involve an Unreviewed Safety Question and will not endanger the health and safety of the public. i l I 6 6-
4 l ATTACIOlENT 2 i i REFERENCES
- 1. S. Viswesvaran, et al., "BWR Owners' Group Response to NRC Generic. Letter 83-28, Itu 4.5.3, " General Electric Company, NEDC-30844A, March 1988
- 2. W. P. Sullivan, et al., " Technical Specification Improvement Analyses for BWR Reactor Protection System", General Electric Company, NEDC-30851P-A, March, 1988
- 3. S. Visweswaran, et al., " Technical Specification Improvement Analysis for BWR Control Rod Block Instrumentation", General Electric Company, NEDC-30851P-A, Supplement 1, October,1988
- 4. L. G. Frederick, et al., " Technical Specification Improvement Analysis for BWR Isolation Instrumentation Common to RPS and ECCS Instrumentation", >
General Electric Company, NEDC-30851P-A, Supplement 2 March,1989
- 5. D. B. Atcheson, et al., "BWR Owners Group Technical Specification Improvement Methodology (With Demonstration for BWR ECCS Actuation Instrumentation", Parts 1 and 2, General Electric Company, NEDC-30936P-A, December, 1988 '
- 6. W. P. Sullivan, et al., " Technical Specification Improvement Analysis for BWR Isolation Actuation Instrumentation", General Electric Company, '
NEDC-316770P-A, July, 1990
- 7. C. E. Rossi, NRC, to R. F. Janacek, BWROG, " Staff Guidance for Licensee Determination that the Drift Characteristics for Instrumentation Used in RPS Channels are Bounded by NEDC-30851P Assumptions when the Functional Test Interval is Extended from Monthly to Quarterly", April 27, 1988 i
.i i
l l ATTACIMENT 3 PROPOSED TECHNICAL SPECIFICATION CHANGES FOR LASALLE COUNTY STATION - List of Attached Change Pages for LaSalle County Station Unit 1 Technical Specifications 3/4 3-1 3/4 3-5 3/4 3-7 3/4 3-8 3/4 3-9 3/4 3-14 3/4 3-20 3/4 3-21 3/4 3-22 3/4 3-26 3/4 3-27 3/4 3 ?7(a) 3/4 3-4 3/4 3-33 3/4 3-34 3/4 3-39 3/4 3-41 3/4 3-44 3/4 3-46 3/4 3-47 3/4 3-49 3/4 3-50 3/4 3-52 3/4 3-54 B 3/4 3-1 B 3/4 3-2 B 3/4 3-3 8 3/4 3-4 The markups are based on Unit 1 Technical Specifications, Amendment 75 plus Modifications #M-1-1-87-096 and #M-1-1-87-097. i I l
3/4.3 INSTRtMENTAT{g 3/4.3.1 REACTOR PROTECTION SYSTEN INSTRUMENTATION LIMITING CON 0! TION FOR OPERATION 3.3.1 As a sinfeue, the reactor protection system instrumentation channels ' " shown in Table 3.3.1-1 shall be OPERA 8LE with the REACTOR PROTECTION SYSTEN RESPONSE TIME as shown in Table 3.3.1-2. APPLICA8!LITY: As shown in Table 3.3.1-1. ACTION: One feng,virtel CN^^* \ l^*PtfSC N
, ,, m ,,_ pu ,, y ,g un;yg, gat
- a. Withlti: ': O f ^^'" "" 'l :t:xx?: ?::- L. . -; t nd b OT~Ff2 the inoperable 2 channels "7;;nt: and/or
;;r T tripd
'; dy-^ r - Gystem in the tripped condition; ry * , place within * ' - The I $2. hour.s, ;reefaia==
o +* < r M J t. af "---ifie=*ia= L o a am net -1 t**1a. l
+tian_ 4Ae. ACTsoed re g ver eg 6ee Te Ma_ $.3
- t. "t th: ^2; ":7 ;f ""5"'"L*. t:; ;t: ?:r n; 'nd by th: "":' r p g.{. ^?;?f22 Ch 7 !; e.7 Trip Oy.^r n; 's- -_;t 7:7 i;^'.
tr'; y:'- :, hg 94ase-et ? r t r: tri; : rd t9: th ^*'!^*' n; 'y:' nd
"" ': 'M tr';;;d x d"': c'th'n 1 i:::
h Td? ? ?l-L SURVEILLANCE REQUIREMENTS 4.3.1.1 Each reactor protection system instrumor.tation channel shall be demonstrated OPERA 8LE by the performance of the CHANNEL CHECK, CHANNEL FUNCTIONAL TEST and CHANNEL CALIBRATION eperations for the OPERATIONAL CONDITIONS and at the frequencies shown in Table 4.3.1.1 1. 4.3.1.2 LOGIC SYSTEM FUNCTIONAL TESTS and slaulated automatic operation of all channels shall be performed at least once per la months.$
- 4. 3.1. 3 The REACTOR PROTECTION SYSTEM RESPONSE TIE of each reacter trip functional unit shown in Table 3.3.1-2 shall be demonstrated to be within its limit at least once per la months. Each test shall include at least one channel per trip systes such that all channels are tested at least once every N times la months where N is the total number of redundant channels in a specific reactor trip system.
[nsert xx . =^ ; dx'F. h pt:x enr'dia; =?y :n::.it: d ': "'; tri;;;d f;;n? ;;-2:7tri;"':y r'r,t!dr '0:r:;:-0 tM '?: c': 2:?
). 7;nt';; t: :xur. :: ;n x : , ; ' :;:N'.1; :P;;x? 2;it : rn' ^ nd Ob ^
w ^7;C.: et.t.e ;t%; . O ?; n ;7 ; !" n; ind h T9 - ? ' 1-1 fer 2:^ T '; 7:20'= f;? ' h ' M-If = n 9 x x!: ;n 'n;:rd!: 'r x: tr'; :y;^ r t';; in 'J; :^';r, ni;;; tMt t '; :y:tes-4e-p1x: ' - tri;;:d :x 4+eere-:4^. ~ 727.' t'.i; Z Id r r e the Trt; P:::ti-- 9 rr " . 1 tche l Mhe spectfled 18-month interval may be waived for Cycle 1 provided the surveillance is performed during Refuel 1. which is to commence no later than October 27, 1945. LA SALLE - UNIT 1 3/4 3-1 Amendment No. 30 l
---- - - l
Innere "A" to IECS Technical Snecifiention 3.3.1.b
- b. With the number of OPERABLE. channels less than allowed by Action a.;
- 1. ensure each required Functional Unit maintains trip capability in j each Trip System within 1 hour, l 1
- 2. ensure for each required Functional Unit, the minimum OPERABLE l channels per Trip system in one trip system and/or the trip systets !
are OPERABLE or in the tripped condition within 6 hours,
- 3. place all inoperable channels and/or associated trip system (s) in the tripped condition
- within 12 hours, otherwise take the ACTION required by Table 3.3.1-1.
Insere "B" to LSCS Technical Snecification 3.3.1.b !
- An inoperable channt.1 and/or the trip system need not be placed in the <
tripped condition if this would cause tho' Trip Function to occur. When a trip system can be placed in the tripped condition without causing the Trip Function to occur, place the most degraded trip system in the tripped condition; if both systems have the same nusber of inoperable channels, place either trip system in the tripped condition.
l 4 TABLE 3.3.1-1 (Continued) REACTOR PROTECTION $YSTEN INSTRUNENTATION TASLE NOTATIONS (a) A channel may be placed in an inoperable status for up to hours for required surveillance without placing the channel in the tripped l condition provided at least one OPERA 8LE 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 ary control red is withdrawn" and during shutdown margin demonstrations perfomed per Specification 3.10.3.
(c) An APRM channel is inoperable if there are less than 2 LPAM inputs per level er less than 14 LPAM inputs to an APRM channel. ' (4) This functica is not required to be OPERABLE when the reactor pressure vessel head is unbelted or removed per Specification 3.10.1. (e) This function shall be automatically bypassed when the reactor mode switch is not in the Run positten. ; (f) This function is not required to be OPERA 8LE when PRIMARY CONTA1 M ff INTEGRITY is not required. - t (9) Also actuates the staney ges treatment system. ' (h) With any centrol red withdrawn. Not applicable to centrol rods removed per Specification 3.9.10.1 or 3.9.10.2. (1) This function shall be automatically bypassed when tuttine first stage pressure is < 140 psig, equivalent to THEIMAL POWER Inss than 3 5 of RATED THERMAI POWER. (j) Also actuates the EOC-RPT systat.
"Not requires for control rods removed per ifications 3.9.10.1 or 3.9.10.2.
l l l l l j LA SALLE - UNIT 1 3/4 3-5 Amenchment No. 30
I . TABLE 4.3.1.1-1 g REACTOR PROTECTION SYSTEM INSTRtMENTATION SURVEILLANCE REQUIREMENTS b CHANNEL OPERATIONAL e CHANNEL FUNCTIONAL CHANNEL CONDITIONS FOR WHICH FUNCTIONAL UNIT CHECK TEST Call 8 RATION I *) SURVEILLANCE REQUIRED
- 1. Intermediate Range Monitors
" S/U(b) 3 3fy(c).W l a. Neutron Flux - High , R 2 S W R 3,4,5
- b. Inoperative M W M 2,3,4,5
- 2. Average Power Range Monitor: III
- a. Neutron Flux - High, Setdown S/U(b) 5 S/U(C),W
, SA 1, 2 S W M 3, 5 w b. Flow Blased Simulated Thermal 1 Power-Upscale S. DI9) S/UIC) +Q, W Id)I'),SA,R(h) j w c. Fixed Neutron Flux -
J, . High S S/UIC) , -4> Q W(d), 3g j
- d. Inoperative M -W6 Q .
M 1, 2, 3, 5
- 3. Reactor Vessel Steam Dome ,
Pressure - High M MQ Q 1, 2
- 4. Reactor Vessel Water Level - 5 Low, Level 3 JW -47 4 R 1, 2 l
- 5. Mair. Steam Line Isolation Valve - Closure M MbQ R 1
- 6. Main Steam Line Radiation -
[ High S WQ R 1, 2
- 7. Primary Containment Pressure -
a High M -#* Q. Q 1, 2
.= .
n _._________c -__ __ - _ . . - - --- , - - -.-% .-r-.-s-_, v - . . - - = __ - _ __ _ _ _ .
l . TA8LE 4.3.1.1-1 (Continued) .
- REACTOR PROTECil0N SYSTEM INSTRUIENIAll0N SURVE!LLANCE REQUIREMENTS ,
t ' 5 CHANNEL OPERAi10NAL
- v. CHANNEL FUNCI10NAL CHANNEL CONDI1 IONS FOR WHICH FUNCTIONAL UNIT CHECK IEST Call 8AAll0N SURVEILLANCE REQUIRtD
! f l ' 8. Scram Olscharge Volume Water ) Level - High NA -M-* Q R 1, 2, 5 ' E " q 9. Turbine Stop Valve - Closure NA -M+ Q R 1 w 10. Turbine Control Valve Fast
- Closure Valve Trip Systes Oil i Pressure - Lou NA -M-74 R* 1
! 11. Reactor Mode Switch l Shutdown Position MA R MA 1,2,3,4,5
- 12. Manual Scram 10 4 -M+ W NA 1,2,3,4,5
- 13. Control Rod Drive
- a. Charging Water Header w Pressure - Lou NA M R 2, 5 ,
A b. Delay Timer IIA M R 2, 5 ! (a) Neutron detectors may be excluded free CHANNEL CALISR L,4. ,. i (b) The ligt, and Slut channels shall be determined to overlap for at least 1/2 decades during e'ach startup
- and the IM and APM channels shall be determined to overlap for at least 1/2 decades during each
, . controlled shutdown, if not performed within the prevlees 7 days. j (c) Within 24 hours prior to startup, if not performed within the previous 7 days. (d) This calibration shall consist of the adjustment of the APM channel to conform to the power levels calculated by a heat balance during OPERATIONAL ColelTION 1 when THElWIAL POWER > 25% of RATED THERMAL POWER. The APM Gain Adjustment Factor (GAF) for any channel shall be equal to the power value deter-mined by the heat balance divided by the APlWI reading for that channel. Within 2 hours, adjust any APM channel with a GAF > 1.02. In addition, adjust any APIIM channel within 2- 12 hours (1) if power is greater than oc equal to 90% of RATED THEIIMAL POWER and the APlWt channel GAF is 5 < 0.98, or (2) if power is less than 905 of RATED THElWIAL POWER and the APIIM reading exceeds the power
$ value determined by the heat balance by more than 10% of RATED THEllMAL POWER. Until any required APIIM 5 adjustment has been accomplished, notification shall be posted on the reactor control panel.
- f. (e) This calibration shall consist of the adjustment of the APilM flow biased channel to conform to a ,
z calibrated flow signal. P (f) The LPRMs shall be calibrated at least once per 1000 ef fective full power hours (EIPH). u, (g) Measure and compare core flow to rated core flow.
" (h) This calibration shall consist of verifying the 6 t I second simulated thermal power time constant.
- "The specified 18-month interval may be waived for Cycle I provided the surveillance is performed durisul Refuel 1, which is to commence no later than October 27, 1985.
INSTRUMENTATI'0N 3/4.3.2 ISCLATION ACTUATION INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.2 The isolation actuation instrumentation channels shown in Table 3.3.2-1 shall be OPERA 8LE with their trip setpoints set consistent with the values shown in the Trip Setpoint column of Table 3.3.2-2 and with ISOLATION SYSTEM RESPONSE TIME as shown in Table 3.3.2-3. APPLICABILITY: As shown in Table 3.3.2-1. ACTION: a.
~
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 OPERA 8LE status with its trip setpoint ad, justed consistent with the Trip Setpoint value. w us . . _ _ _- .e....e _u___,. ,__. .u_ _.,__2 u.. .u u,_,_
..e....e . ... ___
fn .$c.e + _ . . _
,m___
..-.....s.. ,. . . . , . , - . _ . .
.u ... _.-. . , . . . , _._, _ _ ,
--). _i.______ . . . _ , _ . _ . . m _ _-..., _ i r_ ._ s- _. _ m e. _ . .., ._,, ._ ,._ .__. . m_.-
- 4. .u ___2
, ,__. . > u,_ -_ - .,,-- --
.s. . . . . .
. . . . . . . . . . . . . _ - . . , u.. ,___....,___
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. . . . . _ . . . . 4 a__.n.as. , - _
. <,. , a a .-- -
... ___s,_ mi.
-rr - -
g
- c. With the number of OPERABLE channels less n required by the Minimum OPERABLE Channels per Trip Syst requirement for both trip a_ - >
systems, place at least one trip syst in the tripped conditiofG!.*!D l within one hour and take the ACTION required by Table 3.3.2-1. ll Jn.wr p .._2.__,_._.-.. __.s_ ___2
.-.-,.....m... t_ _ , _ _ _ _ . _ .t_
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.. ,, __ ......1___
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,@4
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_ m s. . N, & _ce..e ti Tr'p T .; tier, t. .e; .r. Ir. ths ___.____2
.. .. e . . . . e . . . . . . .
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__ .u
.nev. ,n.u . _ , . ___..,__2
- hrna:1
- m. . ,
m,.,_&,. _ .._,,__.m.
- e. _
..- u i. , .. - u ..u_
^^ii . ; .e h ere i n;;- 01: '- +h a+ he <=1 art
.=_ _ _ _ _ _ _ _ _ _. ,_ .n_
- t ' ..._. ; - e *,__
__ ,> - t "_ - i a
.. . ..,,.,. - - r.... ... .. ..,-_,._.2 , . . . . .
.-..r..x.....
_- __. __ .,_.,2
- th: Trip T;;-t h . t: ::: 7.
"**-- .w..m in;:rcih ch: xt,:::d
___ _ , __ __ .<. .m_ --t M p'_ red
'- t% t ';;:d ::rd't'- - t:-- th'?
._ .m - __ __ __ .m _ >_ __ .,_ _m____,
... . .. . I m .ts t . WV U .W h... S., .. Ms....g b. .. '- * ' ' " ' ' ' ' * '
.L ,, u .f g L_ ew.- ,-'"_"J__2 L.
. um .. ._ r- e . . . . . . ..w. .J..&.,._.
___.___2 it _ a..c.
'vm. . g m f
_.. . ww . - . . .w .- . . ... . .y .. . . g.... .g
.&.. AAhmeam, ak.11 h_.- 4.ba.
.N _&.._ .....
p $. .am. --- . .. 7 - - - . - - - - - LA SALLE UNIT-1 3/4 3-9 Amendment No. 26
O Inmart "A" to LSCS Techniemi Snacification 3.3.2.b
- b. With the number of OPERABLE channels less than required by the Mini-l mum OPERABLE Channels per Trip System requirement for one trip systein: *
- 1. If placing the inoperable channel (s) in the tripped condition I would cause an isolation, the inoperable channel (s) shall be restored to OPERABLE status within 6 hours or take the ACTION required by Table 3.3.2-1 for the affected trip function shall be taken.
- 2. If placing the inoperable channel (s) in the tripped condition would not cause an isolation, the inoperable channel (s) and/or that trip system shall be placed in the tripped condition within:
a) 12 hours for trip functions common to RPS Instrumentation; and b) 24 hours for trip functions not common to RPS Instrumentation. The provisions of Specification 3.0.4 are not applicable. Inmart "B" to LSCS Technical Snacification 3.3.1.c 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.
TABLE 3.3.2-1 (Continued) ISOLATION ACTUATION INSTRUMENTATION ACTION ACTION 20 - Be in at least HDT SHUTDOWN within 12 hours and in COLD SHUTDOWN with the next 24 hours. ACTION 21 - Be in at least STARTUP with the associated isolation valves closed within 6 hours or be in at least NOT SHUTOOWN within 12 hours and in COLD SHUTDOWN within the next 24 hours. ACTION 22 - Close the affected system isolation valves within 1 hour and declare the affected system inoperable. . ACTION 23 Be in at least STARTUP within 8 hours. ACTION 24 - Establish SECONDARY CONTA!WiENT INTEGRITY with the standby gas treatment system operating within 1 hour. ACT!0N 25 - Lock the affected system isolation valves closed within.1 hour and declare the affected system inoperable. ACTION 26 - Provided that the manual initiation function is OPERABLE for L each other group valve, inboard or outboard, as applicable, in ' each line, restore the manual initiation function to OPERABLE status within 24 hours; othe mise, restore the manual initiation function to OPERABLE status within 8 hours; othe mise:
- a. Se in at least NOT SHUTDOWN within the next 22 hours and in COLD SHUTDOWN within the following 24 hours, or
- b. Close the affected system isolation valves within the next >
hour and declare the affected system in operable. El - May be bypassed with reactor steam pressure i 1043 psig and all turbine stop valves closed. ; When handling irradiated fuel in the secondary containment and during CORE ! ALTERATIONS and operations with a potential for draining the reactor vessel.
# During CORE ALTERATIONS and operations with a potential for draining the reactor vessel. -
(a) See specification 3.6.3, Table 3.6.3-1 for valves in each a group.
, (b) A channel m y be placed in an inoperable status for up hours for required surveillance without placing the channel in the tripped l condition provided at least one other OPERABLE channel in the same trip system is monitoring that parameter. In addition for those trip systems ' '
with a design providing only one channel per trip system, the channel may i be placed in an inoperable status for up to B hours for required surveillance tanting without placing the channel in the tripoed condition provided that ' the redundant isolation valve, inboard or outboard, as applicable, in each line is operable and all requimd actuation instrumentation for. that redun-dant valve is OPERABLE, or place the trip system in the tripped condition. , (c) Also actuates the standby gas treatment system. (d) A channel is OPERABLE if 2 of 4 instruments in that channel are OPERABLE. (e) Also actuates secondary containment ventilation isolation dampers per i Table 3.6.5.2-1. (f) Closes only RWCU system inlet outboard valve. 1 LA SALLE - UNIT 1 3/4 3-14 Amendment No. 26
-, , , ., . , . - . - . ~ . , - --
I TABLE 4.3.2.1-1 _ ISOLATION ACTUATION INSTRUNENTATION SURVEILLANCE REQUIRENENTS r-
- CHANNEL-OPERATIONAL CHANN[L FUNCTIONAL CHANNEL
% TRIP FUNCTION CHtCK CONDITIONS FOR milch TEST CAllBRA110N SURVEILIANCE REQUIRED
!. E A. AUTOMATIC INITIATION
- 1. PRINARY CONTAlfe9ENT ISOLATION 6
- a. Reactor Vessel Water Level
" 1) Low, Level 3 NA -M* Q R 1, 2, 3
- 2) Low Low, level 2 NA -M* Q R 1, 2, 3
- 3) Low Low Low, Level 1 NA -MF Q. R 1, 2, 3
- b. Drywell Pressure - High l NA -M* Q Q 1, 2, 3
- c. Main Steam Line
- 1) Radiation - High 5 -M* Q R 1, 2, 3
- 2) Pressure - Low MA -Ma> Q Q l
- 3) Flow - High NA M* Q R 1, 2, 3
- d. Main Steam Line Turnel Temperature - High MA -MA Q R 1, 2, 3 w e. Condenser Vacuum - Low NA -M> Q Q 1, 2*, 3*
1 f. Main Steam Line Tunnel w a Temperature - High
, NA -M* Q R 1, 2, 3 E! 2. SECONDARY CONTAlfetENT ISOLATION
- a. Reactor Building Vent Exhaust i Plenum Radiation - High 5 -M99 R 1, 2, 3 and **
- b. Drywell Pressure - High NA -M
- Q Q 1, 2, 3
- c. Reactor Vessel Water Level - Low Low, Level 2 NA -M
- Q R 1, 2, 3, and
- d. Fuel Pool Vent Exhaust Radiation - High 5 -M* Q R 1, 2, 3 and **
- 3. REACTOR WATER CLEANUP SYSTEM ISOLATION y a. A Flow - High 5 -M* Q R 1, 2, 3 5 b. Heat Exchanger Area .
R Temperature - High NA -M7 Q Q 1, 2, 3 2 c. Heat Exchanger Area 3 Ventilation AT - High NA -MA Q Q 1,2,3 z d. SLCS Initiation MA R NA 1, 2, 3 P e. Reactor vessel Water g Level - Low Low, Level 2. NA -M* Q R 1. 2, '3 __ . _ _ . _ _ _ _ . _ _ . _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ -_m ___ ___ . _ _ _
s.
. +
ISOLATION ACTUATION INSTRtmENTATION SURVEILLANCE REQUIREMENTS g . CHANNEL e- OPERATIONAL CHANNEL FUNCTIO ML G TRIP FUNCTION CHECK TEST CHANNEL CONDITIONS FOR letICH CALIBRATION SURVEILLANCE REQUIRE 0 h 4. REACTOR CORE ISOLATION COOLING SYSTEM ISOLATION
" a. KIC Steam line Flow - High M -M6 Q
- b. KIC Steam Supply Pressure - Q 1, 2, 3 Low M -#*Q Q 1, 2, 3
- c. RCIC Tertine Exhaust Ofaphrage Pressure - Nfgh M
- d. RCIC Equipment Room -4PQ Q 1, 2, 3 Temperature - High M
- e. RCIC Steam Line Tunnel
-56 Q Q 1, 2, 3 Temperature - High M -1PQ Q 1, 2, 3
- f. RCIC Steam Line Tunnel
& Temperature - Nfgh M -#* Q w g. Dryvell Pressure - High Q 1,2,3 M MQ ) h. RCIC Equipment Room Q 1, 2, 3 w A Temperature - High M w PQ Q 1, 2, 3 " l
- 5. RHR SYSTEM STEAM CONDENSING IWOE ISOLATION ' '
- a. MR Equipment Area A Temperature - Hfgh M q'
- b. RHR Area Cooler Temperature - -4PQ 1, 2, 3 High M
- c. RHR Neat Exchanger Steam 4t**Q Q 1, 2, 3 S g ly Flow - High M
-M* Q Q 1, 2, 3 ~
I g a .I ~., e _ _ _ _ _ _ _ . _ _ __ __ . . . . .___._ __ ..-_.,-. .._~ .._..._ _._._.-. ,,_. _ _ _ _ _ _ _ . _ _ . m ___._ _,.4,_ . - . _ ,
I ' - TABLE 4.3.2.1-1 (Continued) 5
!C ISOLATION ACTUATION INSTRt3ENTATION SURVEILLANCE REQUIREENTS h CHANNEL OPERATIONAL
, , CHANNEL FUNCTIONAL CHANNEL CONDITIONS FOR WHICH TRIP FUNCTION CHECK TEST CALIBRATION E SURVEILLANCE REQUIRED
-4 6. RHR SY$ TEM SHUT 90lRI COOLING NDOE ISOLATION
- a. Reactor Vessel Water Level -
Low, Level 3 M -ft* Q R 1, 2, 3 , b. Reactor Vessel l (RHR Cut-in Permissive) i Pressure - High M MQ Q 1, 2, 3
- c. RHR Pump Suction Flow - Nigh NA 4% Q q 1, 2, 3
- d. RHR Area Temperature - High NA #q q 1, 2, 3
- e. RHR Equipment Area AT - High M -MpQ Q 1,2,3 y B. MANUAL INITIATION -
y 1. Inboard Valves NA R M 1, 2, 3 ' y 2. Outboard Valves NA R MA 1, 2, 3
- 3. Inboard Valves MA R NA 1, 2, 3 and **,#
- 4. Outboard Valves NA R NA 1, 2, 3 and **,#
i
- 5. Inboard Valves NA R NA 1, 2, 3
- 6. Outboard Valves MA R NA 1, 2, 3
- 7. Outboard Valve NR R NA 1,2,3 "When reactor steam pressure > 1043 psig and/or any turbine stop valve is open.
**When handling irradiated fuel in the secc~fary containment and during CORE ALTERATIONS and operations with a potential for draining the reactor .assel.
[ #0uring CORE ALTERATIONS and operations with a potential for draining the reactor vessel. n
.I .
U .
~ e . - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
~
i TABLE 3.3.3-1 (Continued) E g EDERGENCY CORE COOLING SYSTEN ACTUATION INETRtBENTATION r-
- E NININUM OPERABLE APPLICA8tE
~ CNAf81ELS PER TRIP OPERATIONAL g TRIP FUNCTION FUNCTION (a) C00eITIONS ACTION ti i
y C. DIVISION 3 TRIP SYSTEM i 1. NPCSSYSiEM j a. Reactor Vessel Water Level - Low, Low, Level 2
- b. Drywell Pressure - High 4$) 1, 2, 3, 4*, 5* 35
- c. Reactor Vessel Water Level-High Level 8 4fI 2
1, 1, 2, 2, 3 3, 4*, 5* 35 32
- d. Condensate Storage Tank Level-Low 1, 2, 3, 4*, 5*
- e. 2(d) 36 Suppression Pool Water Level-Nigh 2 1, 2, 3, 4*, 5* 36
- f. Pump Discharge Pressure-High (Bypass) 1 1,2,3,4*,5* 31
! R g. NPCS System Flow Rate-Low (Permissive) 1
- 1, 2, 3, 4*, 5* 31
! h. Manual Initiation 1/ division 1, 2, 3, 4*, Sa 34 R w, , g O. LOSS OF POWER NINIfRSO AP."LICABLE TOTAL NO. INSTRt#ENTS OPERABLE OPERATIONAL OF INSTRINENTS TO TRIP I ") C00SITIONS INSTRt#ENTS ACTION
- 1. 4.16 kw Emergency Bus Undervoltage 2/ bus 2/ bus 2/ bus 4**, 5**
1, 2, 3, 37 (Loss of Voltage)
- 2. 4.16 kw Emergency Bus Undervoltage 2/ bus 2/ bus 2/ bus 1, 2, 3, 4 **, 5** 37 (Degraded Voltage) g (a) A channel instrument any be placed in an inoperable status for up to hours during periods of required survel11ance without placing the trip system / channel /f astrument in the tripped condition provided at least one i other OPERABLE channel /instrement in the same trip system is monitoring that parameter.
(b) Also actuates the associated division diesel generator. f (c) Provides signal to close NPCS pump discharge valve only on 2-out-of-2 logic. -
, (d) Provides signal to NPCS pump suction valves only.
g Applicable when the system is required to be OPERABLE per Specification.3.5.2 or 3.5.3. g ** Required when ESF equipment is required to be OPERABLE. ' Not required to be OPERABLE when reactor steam dome pressure is 5 122 psig. -
,= ,
=
1 3 e
. . . -- - . -= - - -. -
TA8LE 3.3.3-1 (Continued) EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION ACTION ACTION 30 - W(th the number of OPERABLE channels less than required by the Minimum OPERABLE Channels per Trip Function requirement: *
- a. With one channel inoperable, place the inoperable channel in the tripped condition within6ne hour 9er declare the associated system inoperable.
b. (- 2f Aoors" With more than one channel inoperable, declare the associated system inoperable. - ACTION 31 - With the number of OPERA 8LE channels less t required by the i Minimum OPERA 8LE channels per Trip Function, lace the inoperable channel in the tripped condition withindFne houD restore the inoperable channel to OPERA 8LE status within 7 days or declare
, the associated system inoperable. '
ACTION 32 - With the number of OPERABLE channels less than required by the Minimum declare the associated AOS trip system or ECCS sac inoperabi ACTION 33 - 25 Aours. With the number of OPERA 8LE channels less than the Minimum OPERABLE Channels per Trip Function requirement, place the inoperable channel in the tripped condition within x h:x. 25 Aove.5, ACTION 34 - With the number of OPERA 8LE channels less than required by the Minimum OPERABLE Channels per Trip Function requirement, restore the inoperable channel to OPERA 8LE status within urs or declare the associated ADS trip system or ECCS rable. i ACTION 35 - Y With the number of OPERABLE channels less than required by the Minimum 0PERA8LE Channels per Trip Function requirement
- a. For one trip system, place that trip system in the tripped condition withindne hourfpor declare the NPCS system inoperable. (.gp Aoves* ,
{
- b. For both trip systems, declare the HPCS system inoperable.
ACTION 36 - With the number of OPERABLE channels less than required by the Minimum 0PERA8LE Channels per Trip Function requirement, place at least one inoperable channel in the tripped condition within gne hour">of declare the NPCS system inoperable. Ef- Aours ACTION 37 - With the number of OPERABLE instruments less than the Minism Operable Instruments, place the inoperable instrument (s) in the tripped condition within 1 hour" or declare the associated ' emergency diesel generator inoperable and take the ACTION required by Specification 3.8.1.1 or 3.8.1.2 as appropriate.
"The provisions of Specification 3.0.4 are not applicable.
LA SALLE - UNIT 1 3/4 3-27 Amendment No. 41 l _ . _ _ _ _ _ . ~ _ - _ _ . _ _ _ ,__
^
l TABLE 3.3.3-1 (continued) EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION 1 ACTION 1
. . J ACTION 38 With the number of OPERA 8LE channels less than required by the Minimum OPERABLE Channels per trip function requirements: ,
- a. With one channel inoperable, remove the inoperable channel within ene nogB; restore the inoperable channel to OPERA 8LE status wit in 7 days or declare the associated ECCS systems inoperable.
2.Y kour5
- b. With both channels inoperable, restore at least one channel to OPERA ECC5 8LEinoperable.
systems status within one hour or declare the' associated I f \ i LA SALLE - UNIT 1 3/4 3-27(a) Amendment 10
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i TABLE 4.3.3.1-1 (Continued) 5 EERGENCY CORE COOLING SYSTEM ACTUATION INSTRISENTATION SURVEILLANCE REQUIREENTS M E
- CHANNEL .
OPERATIONAL i CH4801EL FUNCTIONAL CHANNEL CONDITIONS FOR WICH I E TRIP Ft#ICTION CNECK TEST CALIBRATION SURVEILLANCE REQUIRED 1 :1 ! -g C. DIVISION 3 TRIP SYSTEM
- 1. HPCS SYSTEN I
- a. Reactor Vessel Water Level -
Low Low, Level 2 ,N4-5 -40* Q R 1, 2, 3, 4*, 5* , b. Drywell Pressure-High M4 -M* Q Q 1,2,3
- c. Reactor Vessel Water Level-High Level 8 MS -4P Q R 1, 2, 3, 4*, 5*
- d. Condensate Storage Tank Level -
Loir N4 WQ Q 1, 2, 3, 4*, 5* R e. Suppression Pool Water <
- Level - High Pump Discharge Pressure-High II4 #Q Q 1, 2, 3, 4* , 5*
g f.
- g. HPCS System Flow Rate-Low NA -IPQ Q 1, 2, 3, 4* , 5" M4 -4 4 Q Q 1, 2, 3, 4*, 5*
3
- h. Manuel Initiation N4 R M4 1, 2, 3, 4*, 5*
D. LOSS OF POWER '
- 1. 4.16 kV Emergency Bus Under-
, voltage (Less of Voltage) N4 10 4 R 1, 2, 3, 4**, 5** I
- 2. 4.16 kV Emergency Bus Under- M4 NA R 1, 2, 3, 4**, 5**
voltage (Oegraded Voltage) l l
#Not required to be OPERABLE when reactor steam done pressure is less than or equal to 122 psig.
*When the system is required to be OPERASLE after being manually realigned, as e vitcable, per i
specification 3.5.2. . [ ** Required when E5F equipment is required to be OPERABLE. . g ***The specified 18-month interval may be waived for Cycle 1 provided the surveillance is performed during { Refuel 1, which is to commence no later than October 27, 1985. E ~ 1 y -
INSTRUMENTATION END-OF-CYCLE RECIRCULATION PUMP TRIP SYSTEM INSTRUMENTATION LIMITING CONDITION FOR OPERATION . 3.3.4.2 The end-of-cycle recirculation pump trip (E0C-RPT) system instrumenta-tion channels shown in Table 3.3.4.2-1 shall be OPERA 8LE with their trip setpoints set consistent with the values shown in the Trip Setpoint column of Table 3.3.4.2-2 and with the END-OF-CYCLE RECIRCULATION PUMP TRIP SYSTEM RESPONSE TIME as shown in Table 3.3.4.2-3.- APPLICABILITY: OPERATIONAL CONDITION 1, when THERMAL POWER is greater than or equal to 30s of RATED THERMAL P0hER. ACTION:
- a. With an end-of-cycle recirculation pump trip system instrumentation channel trip setpoint less conservative than the value shown in the Allowable Values column of Table 3.3.4.2-2, declare the channel inoperable until the channel is restored to OPERABLE status with the channel setpoint adjusted consistent with the Trip Setpoint value.
- b. With the number of OPERABLE channels one less than required by the Minimum OPERA 8LE Channels per Trip Systes requirement for one or both trip systems. alace the inoperable channel (s) in the tripped condition withinQ_ wur.312. hours,
- c. With the number of OPERABLE channels two or more less than required by the Minimum OPERABLE Channels per Trip System requirement (s) for one trip system and:
- 1. If the inoperable channels consist of one turbine control valve channel and one turbine stop valve channel. place both inoperable l channels in the tripped condition withind hovo i2. hours. l
- 2. If the inoperable channels include two turbine control valve channels or two turbine stop valve channels, declare the trip system inoperable.
- d. With one trip system inoperable, restore the inoperable trip system l to OPERABLE status within 72 hours. Otherwise, either l
- 1. Increase the MINIMUM CRITICAL POWER RATIO (MCPR) Limiti Condition i for Operation (LC0) to the E0C-RPT inoperable value per cifica-tion 3.2.3 within the next I hour or,
- 2. Reduce THERMAL POWER to less than 305 of RATED THERMAL POWER with-in the next 6 hours.
- e. With both trip systems inoperable, restore at least one trip system to OPERA 8LE status within I hour. Otherwise, either:
LA SALLE - UNIT 1 3/4 3-39 Amendment No. 58
1 1 g TABLE 3.3.4.2-1 g Ele-OF-CYCLE RECIRCULATION Ptw TRIP SYSTEM INSTRISENTATION r-2 5 MINIM 24 I ' g TRIP FINICTION OPERA 8tECHANNE(j) PER TRIP SYSTEM y 1. Turbine Stop Valve - Closure 2(b) ,
- 2. Turbine Control Valve - Fast Closure 2(b) ,
l
$ 8 I*I
{ A trfp system may be placed in an inoperable states for up to s for required survellism:e w provided that the other trip system is OPERABLE. (b)This function shall be automatically bypassed when turbine first stage pressure is less than or equal to 140 psig, equivalent to THElW14L POWER 1ess than 3GK of RATED THENt4L POWER. l f . i 2 F f
1 TABLE 4.3.4.2.1-1 1 E y, END-OF-CYCLE RECIRCULATION PUNP TRIP SYSTEM SURVEILLANCE REQUIREMENTS
?
E CHANNEL FUNCTIONAL CHANNEL TRIP FUNCTION g TEST CALIBRATION
- 1. Turbine Stop Valve-Closure
-M+ Q R ,
- 2. Turbine Control Valve-Fast Closure -M+ Q R Y
t e e
. , _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ '--^ ' " ^ '
TA8LE 3.3.5-1 9 vi REACTOR CORE ISOLATION COOLING SYSTEM ACTUATION INSTRUMENTATION E MINIMUM OPERABLE e e FUNCTIONAL UNITS CHANNELSPE{*) TRIP SYSTEM ACTION 5
- a. Reactor Vessel Water Level - Low Low, Level 2
] 2 50
- b. Reactor Vessel Water Level - High. Level 8 2(b) $j
- c. Manual Initiation I ICI 52 i
R. b (a) A channel may be placed in an inoperable status for up to hours for required survelliance without placing the trip system in the tripped condition provided at least one other OPERABLE channel in the same trip system is monitoring that parameter. (b) One trip system with two-out-of-two logic. (c) Single channel. i
TA8LE 3.3.5-1 (Continued) . . REACTOR CORE ISOLATION COOLING SYSTEM ACTUATION INSTRUMENTATION ACTION 50 - With the number of OPERABLE channels less than required by the Minimum OPERABLE Channels per Trip System roccirement:
- a. For one trip system, place the inoperabia channel in the tripped condition within drpe houtor declare the RCIC system inoperable. (- 2f hours
- b. For both trip systems, declare the RCIC system inoperable.
ACTION 51 - With the number of OPERA 8LE channel less than required by the minimum OPERABLE Channels per gigen, 2 f System Aoves. requirement, declare theRCICsysteminoperabpt4 ACTION 52 - With the number of OPERABLE channels less than required by the Minimum OPERA 8LE Channels per Trip System requi t, restore h inoperable channel to OPERABLE status within urs or
.clare the RCIC system inoperable.
2Y l l D LA SALLE - UNIT 1 3/4 3-47 i W
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INSTRUMENTATION 3/4.3.6 CONTROL R00 WITHDRAWAL BLOCK INSTRUMENTATION LIMITING CON 0! TION FOR OPERATION 3.3.G The control rod withdrawal block instrumentation channels shown in ' Table 3.3.6-1 shall be OPERA 8LE with their trip setpoints set consistent with the values shown in the Trip Setpoint column of Table 3.3.6-2. , APPLICA81LITY: As shown in Table 3.3.6-1. ACTION:
- a. With a control rod withdrawal block instrumentation channel trip setpoint less conservative than the value shown in the Allowable Values column of Table 3.3.6-2, declare the channel inoperable until the channel is restored to OPERA 8LE status with its trip setpoint adjusted consistent with-the Trip Setpoint value.
- b. With the number of OPERA 8LE channels less than required by the Minimum OPERABLE Channels per Trip System requirement, take the ACTION required by Table 3.3.6-1.
t SURVE!LLANCE REQUIREMENTS s , l 4.3.6 Each of the above required control rod withdrawal block trip systems and instrumentation channels shall be demonstrated OPERA 8LE 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 4.3.6-1. InSeet h LA SALLE - UNIT 1 3/4 3-50
Inmart "A" to LSCS Technical Snacification 4.3.6 A channel may be placed in an inoperable status for up to 6 hours for i required surveillance (or 12 hours for repair) without placing the trip - - l system in the tripped condition provided at least one other OPERABLE in the I same trip system is monitoring that parameter. 1 l l e 6 i l
TABLE 3.3.6-1 (Continued) l J i CONTROL R00 WITH0RAWAL BLOCK INSTRUMENTATION ACTION - ACTION 60 - Declare the R8M inoperable and take the ACTION required by Specification 3.1.4.3. , ACTION 61 - With the number of OPERABLE channels: ;
- a. One less than required by the Minimum 0PERABLE Channels per Trip !
Function requirement, restore the inoperable channel to OPERABLE status within 7 days or place the inoperable channel in the tripped condition within the next hour.
- b. Two or more less than required by the Minimum OPERA 8LE Channels per Trip Function requirement, place at least one inoperable channel in the tripped condition within one hour.
ACTION 62 - With the number of OPERA 8LE Channels less than required by the Minimum OPERA 8LE Channels per Trip Function requirement, place the inoperable channel in the tripped condition withindEne houb gg i 2. hours. J With THERMAL POWER 3 30% of RATED THERMAL POWER. With more than one control rod withdrawn. Not applicable to control rods removed per Specification 3.9.10.1 or 3.9.10.2.
- a. The R8M shall be automatically bypassed when a peripheral control rod is selected,
- b. This function shall be automatically bypassed if detector count rate is 1 100 cps or the IRM channels are on range 3 or higher.
- c. This function shall be automatically bypassed when the associated IRM channels are on range 8 or higher.
- d. This function shall be automatically bypassed when the IRM channels are on range 3 or higher.
- e. This function shall be automatica11y bypassed when the IRM channels are on range 1.
4 LA-SALLE - UNIT 1 3/4 3-52
TABLE 4.3.5-1
' 5 .
Y CONTROL 200 WIT)WRAMAL BLOCK INSTRINGENTATION SURVEILLANCE REQUIREMENTS h CHANNEL OPERATIONAL
' CHANNEL FUNCTIONAL CHAISIEL CONDITIONS FOR t#4ICH TRIP FUNCTION CHECK TEST CALIBRATION I "I SURVEILLANCE REQUIRED
!. E a-*
- 1. ROD OLOCK IWIIITOR
- a. Upecale NA 5/U l b. Insperettwe b c c kN 1*
NA 5/U Q N.A. 1* ' l 5/U b)(c) t
- c. Downscale NA c)Q Q 1*
, 2. APNI
- a. Flow Blased 5faulated 3
Thermal Power-Upscale NA 5/U N 5A I b. Inoperative IIA
- c. 5/U(b ,
Q . . 1, 2, 5 Downscale IIA 5/U b) Al+ q SA 1
, d. Neutron Flux-Migh NA 5/U b) b q . SA 2, 5 1 3. SOURCE RAf0GE IWHITORS I Y a. Detector not full in IIA 5/U(b) W N.A. 2, 5 i I b.
l
$ scale IIA 5/U ,W Q 2, 5
- c. Insperative 11 4 5/U N.A. 2, 5
- d. Downscale IIA 5/U(b),W ,W 2, 5 l
Q
- 4. INTEIDEDIATE RAf1GE IWIIITOR$
j
. a. Detector not full in IIA 5/U ,W N.A. 2, 5 l
- b. Upscale IIA W Q 2, 5
) c. Inoperative IIA 5/U(b),W 5/U N.A. 2, 5
. d. Downscale IIA 5/U(b),W l ,
l , Q 2, 5 l l
- 5. SCRAN DISCHARGE Voll#E g a. ,Weter Level-Hfgh NA Q R 1, 2, 5**
- b. Scram Discharge Volume Swttch in Bypass IIA -ft* Q N.A. 5** l
= 6. REACTOR COOLANT-SYSTEN RECIRCULATION FLOW g a. Upscale IIA 5/U ;ft* Q Q
~
1
- b. Inoperative IIA N.A. 1 l g c. Ceaperator NA 5/U(b) 5/U p# Q q Q 1 o
3/4.3 INSTRUMENTATION BASES 3/4.3.1 REACTOR PROTECTION 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 reactor 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 main-tenance. When necessary, one channel.nay be made inoperable for brief intervals to conduct required surveillance. The reactor protection system is made 19 of two independent trip systems. I 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, 1971, for nuclear power plant protection syst The bases for the trip settings of the RPS are discussed in the base Lf pecification 2.2.1.
"Inse.ctCA_)
The measurement of response time at the specified frequencies provides , assurance that the protective functions associated with each channel are com- ' plated within the time limit assumed in the accident analysis. No credit was taken for those channels with response times indicated as not applicable. ; Response time may be demonstrated by any series of sequential, overlapping or i total channel test measurement, provided such tests demonstrate the total channel response time as defined. Sensor response time verification may be demonstrated by either (1) 'nplace, onsite or offsite test measurements, or (2) utilizing replacement sensors with certified response times. l l LA SALLE - UNIT 1 8 3/4 3-1
heart *A* to ISCS Techn{ c,y gneef fication 3 /4.3,1 Specified surveillance intervals and surveillance and maintenance outage times have been determined in accordance with NEDC-30851P-A, " Technical -
- Specification Improvement Analysis for SUR Reactor Protection System," March !
1988. l I i i
t INSTRUMENTATION 8ASES 3/4.3.2 ISOLATION ACTUATION INSTRUMENTATION This specification ensures the effectiveness of the instrumentation used to mitigate the consequences of accidents by prescribing the OPERABILITY trip M g setpoints and response times for isolation of the reactor systems. When necessary, one channel may be inoperable for brief intervals to conduct required surveillance. h Tome of the trip settings may have tolerances explicitly stated where both the high geg and low values are critical and may have a substantial effect on safety. The. set-f*' oints of other instrumentation, where only the high or low end of the setting have a id ect bearing on safety, are established at a level away from the normal operating range to prevent inadvertent actuation of the systems involved, i 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 is concurrent with the 13 second diesel i startup. The safety analysis considers an allowable inventory loss in each case i which in turn detemines the valve speed in conjunction with the 13 second delay. l It follows that checking the valve speeds and the 13 second time for emergency w! power establishment will mstablish the response time for,the isolation functions. l 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. i 3/4.3.3 EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION The emergency core cooling system actuation ins +~ eentation is prov!ded to initiate actions to mitigate the consequences of i.:idents that are beyond l the ability of the operator to control. This specification provides the , OPERA 8ILITY requirements, trip setpoints and response times that will ensure i ef fectiveness of the systems to provide the .1esign 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. fNRd h : LA SALLE - UNIT 1 8 3/4 3-2
Inmart "A" to M CS Technical Snacification 3/4.3.2 Specified surveillance intervals and maintenance outage times have been determined in accordance with NEDC-31677P-A, " Technical Specification Improvement Analysis for BWR Isolation Actuation Instrumentation", June 1990. - inmart "B" to UCS Technical Seacification 3/4.3.3 Specified surveillance intervals and maintenance outage times have been determined in accordance with NEDC-30936P-A, Parts 1 and 2, "BWR Owners' Croup Technical Specification Improvement Methodology (with Demonstration for BWR ECCS Actuation Instrumentation)" December 1988, plus letter OG90-319 32D to - M. L. Wohl from W. P. Sullivan dated March 22, 1990, " Clarification of Technical Specification Changes Civen in ECCS Actuation Instrumentation Analysis". l 1 1 I
. .~ _ - - - - _ . -- . _ _ - - - - . _ . - -- . _ - _ _ - .
INSTRUMENTATION BASES 3/4.3.4 RECIRCULATION PUMP TRIP ACTUATION INSTdUMENTATION - The anticipated transient without scram (ATWS) recirculation pump trip ! system provides a means of limiting the consequences of the unlikely occurrence
- of a failure to scram during an anticipated transient. The response of the plant to this postulated event falls within the envelope of study events in General Electric Company Topical Report NE00-10349, dated March 1971 and l NEDD-24222, dated December, 1979, and Appendix G of the FSAR.
The end-of-cycle recirculation pump trip (E0C-RPT) system is a part of : the Reactor Protection System and is an essential safety supplement to the reactor trip. The purpose of the E0C-RPT is to recover the loss of thermal margin which occurs at the end-of-cycle. The physical phenomenon inw1ved is that the void reactivity feedback due to a pressurization transient can add positive reactivity to the reactor system at a faster rate than the control ' rods add ne Each E0C-RPT system trips both recircula-tion pumps,gative scram reducing reactivity. coolant flow in order to reduce the void collapse in the core during two of the most limiting pressurization events. The two events for which the E0C-RPT protective feature will function are closure of the turbine stop valves and fast closure of the turbine control valves. A generic analysis, which provides for continued operation with one or both trip systems of the E0C-RPT system inoperable, has been performed. The analysis determined bounding cycle independent MINIMUM CRITICAL POWER RATIO (MCPR) Limiting Condition for Operation (LCD) values whict) must be used if the E0C-RPT system is inoperable. These values ensure that adequate reactivity margin to the MCPR safety limit exists in the event of the analyzed transient with the RPT function ' inoperable. The analysis results are further discussed in the bases for Speci-fication 3.2.3. A fast closure sensor from each of two turbine control valves provides ! input to the E0C-RPT system; a fast closure sensor from each of the other two turbine control valves provides input to the second EOC-RPT system. Similarly, a position switch for each of two turbine stop valves provides input to one EOC-RPT system; a position switch from each of the other two stop valves provides input to the other E0C-RPT system. For each E0C-RPT system, the sensor relay contacts are arranged to form a 2-out-of-2 logic for the fast closure of turbine control valves and a 2-out-of-2 logic for the turbine stop valves. The operation of either logic will actuate the E0C-RPT system and i yMO~gripbothrecirculationpumps. fa
@ is administrative 1y controlled.Each The manual EOC-RPT sses and thesystem automaticmay be manuall Operating Bypass at less than 305 of RATED THERMAL are annunciated in the control room.
i The E0C-RPT system response time is the time assumed in the analysis between initiation of valve motion and complete suppression of the electric arc, i.e. ,190 as, less the time allotted for sensor response, i.e. ,10 as, and less the time allotted for breaker are suppression determined by: test, as correlated to manufacturer's test results, i.e. , 83 as, and plant pre-operational test results. LA SALLE - UNIT 1 8 3/4 3-3 Amendment No. 58 {
Inmart "A" to IJCS Technfem1 Snecification 3 /4. 3.4 Specified surveillance intervals and surveillance and maintenance outage times l have been determined in accordance with NEDC-30851P-A, " Technical - - Specification Improvement Analysis for BWR Reactor Protection System," March 1988. t e B P 1
t . l
. INSTRUMENTATION BASES 3/4.3.5 REACTOR CORE ISOLATION COOLING SYSTEM ACTUATION INSTRUMENTATION -
t The reactor core isolation cooling system actuation instrumentation is provided to initiate actions to assure adequate core cooling in the event of reactor isolation from its primary heat sink and the loss of feedwater flow to the reactor vessel without providing actuation of any of the emergency core ' A cooling equipment. - 3/4.3.6 CONTROL R00 WITHORAWAL BLOCK INSTRUMENTATION The control rod block functions are provided consistent with the require-ments of the specifications in Section 3/4.1.4, Cor. trol Rod Program Controls and Section 3/4.2 Power Distribution Limits. The trip logic is arranged so that a in any one of the inputs will result in a control rod block. g@ 3/4.3.7 MONITORING INSTRUMENTATION 3/4.3.7.1 RADIATION MONITORING INSTRUMENTATION The OPERABILITY of the radiatson monitoring instrumentation ensures that; i (1) the radiation levels are continually measured in the areas served by the individual channels, and (2) the alarm or automatic action is initiated when the radiation level trip setpoint is exceeded. 3.4.3.7.2 SEISMIC MONITORING INSTRUMENTATION The OPERA 8ILITY of the seismic monitoring instrumentation ensures that suffic-isnt capability is available to promptly detaraine the magnitude of a seismic event and evaluate the response of those features important to safety. This capability is required to permit comparison of the measured response to that used in the . design basis for the unit. This instrumentation is consistent with the recommen-dations of Regulatory Guide 1.12 " Instrumentation for Earthquakes" April 1974. 3/4.3.7.3 METEOROLOGICAL MONITORING INSTRUMENTATION The OPERABILITY of the meteorological monitoring instrumentation ensures 1 that sufficient meteorological data is available for estimating potential radiatior6 doses to the public as a m sult of routine or accidental release of radiesctive materials to the atmosphere. This capability is required to evaluate the need for initiating protective measures to protect the health and safety of the public. This instrumentation is consistent with the recommendations of
; Regulatory Guide 1.23 "0nsite Meteorological Programs," February, 1972.
1 3/4.3.7.4 REMOTE SHUTDOWN MONITORING INSTRUMENTATION l The OPERABILITY of the remote shutdown monitoring instrumentation ensures I that sufficient capability is available to pensit shutdown and maintenance of HDT SHUTDOWN of the unit from locations outside of the control room. This capability is required in the event control room habitability is lost and is 1 consistent with General Design Criteria 19 of 10 CFR 50. l 1 LA SALLE - UNIT 1 B 3/4 3-4 Amendment No. 58 I i
4 i Inmart "A" to M CS Technical Snacification 3/4:3.5 specified surveillance intervals and maintenance outage times have been determined in secordance with NEDC 30936P-A, Parts 1 and 2. " BUR Owners' Croup
- Technical Specification Improvement Methodology (with Demonstration for BUR ECCS Actuation Instrumentation)", December 1983, plus letter OG90-319-32D to M. L. Wohl from W. P. Sullivan dated March 22, 1990, " Clarification of Technical Specification Changes Given in ECCS Actuation Instrumentation Analysis".
l Inmart "B" to MCS Technical Snacifiention 3/4.3.6 Specified surveillance intervals and maintenance outage times have been determined in accordance with NEDC 30851P-A, Supplement 1, " Technical Specification Improvement Analysis for SUR Control Rod Block Instrumentation". October 1988.
ATTACHMENT L BWR OWNERS GROUP LETTER TO B.K. GRIMES (NUCLEAR REACTOR REGULATION) BWR OWNERS GROUP (BWROG) TOPICAL REPORTS ON TECHNICAL SPECIFICATION IMPROVEMENT ANALYSIS FOR BWR REACTOR PROTECTION SYSTEMS-USE FOR RELAY AND SOLID STATE PLANTS (NEDC-30884 AND NEDC-30851 P) BWROG 92102, NOVEMBER 4, 1992 - k antas1.a.11.s.otstti.wpfas
SWR
~
oiuneas eaoug < i" , = = ,; clo Southom Nudoor Operoting Company P.O. Box 1995. 81n 8052 a Birm6ngham, N. 35201 BUROG 92102 ~ November 4, 1992 Brian K. Crimes, Director Division of Operating Reactor Support Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, D.C. 20555 Attention: Christopher 1. Crimes, Chief Technical Specifications Branch
Subject:
BVR OWNERS' CROUP (BVROG) TOPICAL REPORTS ON TECHNICAI, SPECIFICATION IMPROVEMENT ANALYSIS FOR BUR REACTOR PROTECTION SYSTEMS - USE FOR REIAT AND SOI.ID STATE FIANTS (NEDC-30884 AND NEDC-30851F)
Reference:
Letter, C.E. Rossi (NRC) to G.J. Beck (BWROC), same subject, dated July 26, 1991
Dear Mr. Crimes:
In the reference letter the NRC expressed concern that model TS ACTION "a", proposed in NEDC 30851P-A, would allow continued plant operation for up to 12 hours with a combination of failures that could prevent a reactor scram function from completing its logic when called upon (i.e. , loss-of-function). This could occur for a relay type plant if, for example, both channels of the high reactor pressure function (which has a one-out-of-two-twice logic) were inoperable in one trip system. The reference letter also noted that the BWROG was preparing clarifying language (i.e., revised model TS ACTIONS) to address this concern, and to be used as an industry standard in future amendments implementing the RPS topical report. In response to the above, the BWROC has worked with Carl Schulten of your Staff to develop the model TS ACTIONS provided in Enclosure 1. The indicated changes to ACTIONS 3.3.la and 3.3.lb and their footnotes ensure that appro-priate actions are taken to avoid an extended loss-of function period in any RPS Functional Unit. A discussion of the application and justification for the revised model TS ACTIONS is provided in Enclosure 2. 1 l
J
' B.K. Grimes, NRC BWROG 92102 November 4,1992 Page 2 The enclosed infomation has been endorsed by a substantial number of the members of the BWROG; however, it should not be interpreted as a commitment of any individual member to a specific course of action. Each member must formally endorse the BUROG position in order for that position to become the member's position.
- Very truly yours, C. L. Tully, Chairperson BWR Owners' Group EXEC 6T/CLT/JDF/rt Enclosures 2 cc: CS Schulten (NRC) ,
IA England, BUROC Vice Chairman BWROC Primary Representatives of Participating Utilities BWROG Technical Specifications Comunittee-D LS Gifford, GE SJ Stark, GE l r
+
l l BWROG-92102 NEDC-30851P-A ENCLOSURE 1 l . November 4, 1992 l Table 5 9 CHANCES TO RE1AY RPS TECHNICAL SPECIFICATION l 3/4.3 INSTRUMENTATION 3/4.3.1 REACTOR PROTECTION SYSTEM INSTRUMENTATIQH LIMITING CONDITION FOR OPERATION 3.3.1 As a minimum, the reactor protection sy .en instrumentation channels l shown in Table 3.3.1-1 shall be OPERABLE wit'.- ;.he REACTOR PROTECTION SYSTEM RESPONSE TIME as shown in Table 3.3.1 2. l APPLICABILITY: As shown in Table 3.3.1-1. ACTION:yINSERT 1 [a. U the numbe of OPERA. d. t .sannels ic than requir by the Mini 7 ERABLE Cha els per T ). t/ stem req ement for o trip system' place the inoper e channe s) and/or th trip system n the trippe conditio within hours, provisions Specificat 3.0.4 are no applicable
- b. Vi the numbe of OPERABLE annels less e an required the Min um 0 LE Cha is per Trip ystem requir nt for both rip syst ,
lace at le e one trip stem ** in the ripped cond on withi one hour and take the ACTION req ired by Table .3.1-1. SURVEILIX4CE REOUIREMENTS 4.3.1.1 Each reactor procaction system instrumentation channel shall be demonstrated OPERABLE by the performance of the CHANNEL CHECY., 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 14CIC SYSTEM FUNCTIONAL TESTS and simulated automatic operation of all channels shall be performed at least once per 18 months, y ZNSERT 2. pan operable annel ne not be pl ed in the ipped co ition wher this wo d cause e Trip Fu tion to o r. In th a cases, e inopera annel sh be rest ed to OP LE status thin f_ urs or th CTION equired table 3. .1 1 for t t Trip F ion shal e taken.
*If more hannels a inoperab in one er system an in the er, p1 e the t p system th more i perable c els in e tripped nditio (exc c when th would cau the Trip unction occur.
5 33
fl BWROG-92102 ENULOSURE1 November 4, 1992 (continued) ;
~
INSERT 1: '
. i
- s. With one channel required by Table 3.3.1-1 inoperable in one or more i Functional Units, place the inoperable channel and/or that trip system in '
tre tripped condition
- within 12 hours. [The provisions of Specification 3.0.4 are not applicable.]
- b. Wi'.h two or more channels required by Table 3.3.1-1 inoperable in one or more Functional Units:
- 1. Within one hour, verify sufficient channels remain OPERABLE or l
tripped
- to maintain trip capability in the Functional Unit, and
- 2. Within 6 hours, place the inoperable channel (s) in one trip systes !
and/or that trip system ** in the tripped condition *, and
- 3. Within 12 hours, restore the inoperable channels in the other trip system to an OPERABLE status or tripped *. ;
otherwise, take the ACTION required by Table 3.3.1-1 for the Functional i Unit. i l l l INSERT 2+ ! i {
*An inoperable channel or trip system need not be placed in the tripped j condition where this would cause the Trip Function to occur. In these J cases, if the inoperable channel is not restored to OPERABLE status within the required time, the ACTION required by Table 3.3.1-1 for the Functional Unit shall be taken.
**This ACTION applies to that trip system with the most inoperable channels; if both trip systsas have the same number of inoperable channels, the ACTION can be applied to either trip system.
t
BWROG-92102 ENCLOSURE 2
'- November 4, 1992 l
i I Anolication and Justification for channes to Nmj-30851P-A. Table 5-9 For ACTION 3.3.la, with one channel required by Table 3.3.1-1 inoperable in . one or more Functional Unit (s) (i.e., any number of Functional Units having only one inoperable channel in each Functional Unit), the entire RPS scram capability remains intact, assuming no additional single failure. Therefore, a loss-of-function is not possible for the rewritten ACTION 3.3.1.a. The ( action that allows continued operation for 12 hours was evaluated and the reliability of the system shown to be acceptable in NEDC-30851P-A. 1 Within 12 hours the inoperable channels and/or trip system must be placed in the tripped condition. This action restores the'RPS capability to accommodate a single failure and allows operation to continue with no further restrictions. If the inoperable channel (s) and/or trip system is not placed in the tripped condition within the required time (12 hours for ACTION a), then the ACTIONS required by Table 3.3.1-1 must be taken, which require the operators to take actions to compensate for the inoperable RPS channels' function. For ACTION 3.3.lb, with two or more channels inoperable in any Functional Unit, the Reactor Protection System any not be capable of performing its intended function (i.e., a " loss of scram function" may 4xist), depending on which two (or more) channels are inoperable. In this cc..dition, during the l period allowed to place the inoperable channels and/or trip system in the i I tripped condition, if a valid trip signal was received, a failure to produce a scram signal for that Functional Unit could result. Therefore, ACTION b.1 requires that steps be taken within one hour to ensure the Functional Unit maintains trip capability. This one hour period allows the operator time to evaluate the situation and to repair or trip the channels. One hour is reasonable considering the diversity of sensors available to provide trip signals, and the low probability of an event requiring the initiation of a scram. One hour is also consistent with the current Technical Specification requirement for placing inoperable channels in the tripped condition. In addition, if it has been verified that a loss-of-function situation does not l exist, an allowance of 6 hours is provided by ACTION b.2 before the operator l l 1s required to place the inoperable channel (s) in one Trip System (or one l entire Trip System), in the tripped condition. This 6 hour requirement limits the time the RPS scram logic for any Functional Unit may be degraded in both Trip Systems. Six hours is considered acceptable based on the remaining capability to trip, the diversity available to provide trip signals, the low probability of extensive numbers of inoperabilities affecting all diverse functions, and the low probability of an event requiring the initiation of a l
7 BWROG-92102 ENC'LOSURE 2 November 4, 1992 (continued) i scram. By the end of the six hour period, the ACTION b.2 requirement that one Trip System will have its inoperable channels placed into the tripped
~
condition provides a similar level of RPS availability as found in ACTION a above, and evaluated in NEDC-30851P-A to be acceptable for a 12 hour allowable - outage time. I Within 12 hours, per ACTION b.3, all the inoperable channels in the other trip system will have been restored to OPERABLE status, or else the inoperable channels will be placed in trip. For all of the proposed ACTIONS, if the ; inoperable channels are not placed in trip within the applicable required time (1, 6, or 12 hours), then the ACTIONS required by Table 3.3.1-1 must be taken, which requires the operators to take actions to compensate for the inoperable RPS channels' function. i l l l l t f- ---/}}