ML20196B744
| ML20196B744 | |
| Person / Time | |
|---|---|
| Site: | Quad Cities  |
| Issue date: | 01/29/1988 |
| From: | COMMONWEALTH EDISON CO. |
| To: | |
| Shared Package | |
| ML20196B742 | List: |
| References | |
| 4152K, NUDOCS 8802120107 | |
| Download: ML20196B744 (10) | |
Text
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,b ATTAQ9ENT 1 PROPOSED CHANGES TO APPENDIX A TECHNICAL SPECIFICATIONS FOR OUAD CITIES STATION UNITS 1 AND 2 FACILITY OPERATING LICENSES DPR-29 AND DPR-30 Revised Pages: 1.1/2.1-2a (DPR 29) 3.2/4.2-5a (DPR 29) 3.2/4.?-12 (DPR 29) 1.1/2.1-2a (DPR 30) 3.2/4.2-57 'DPR 30) '
3.2/4.2-12 ')PR 30)
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OPR-29 The definitions used above for the ,
APRM scram trip apply. In the event
, of operation with a maximum fraction I' limiting power density (MFLPD) greater than the fraction of rated power (FRP), the setting shall be .
modified as follows: )- >
FeP 7
$1 (.58WD + $0) MFLPD f
?
The definitions used above for the APRM scram trip apply.
The ratio of FRP to MFLPD shall be .
set equal to 1.0 unless the actual operating value is less than 1.0.' in which case the actual operating value will be used.
This may also be performed by increasing the APRM gain by the inverse ratic. NFLPD/FRP. which accompitsbes the same degree of protection as reducing the trip setting by FRP/MFLPD.
C. Reactor low water level scram setting shall be 144 inches above the top of the active fuel" at normal cperating conditions.
D. Reactor low water level (CC5 initiation shall be 1 84 inches above l the top cf the active fuel' at normal operating conditions.
E. Turbine stop valve scram shall be i 10% valve closure from full open.
F. Turbine control valve fast closure scram shall initiate upon actuation of the fast closure solenoid valves which trip the turbine control valves.
G. Main steamline isolation valve closure scram shall be i 10% valve closure f rom f ull open.
H. Main steamline low-pressure ini-t14 tion of main steamline isolation valve closure shall oe 1 825 psig.
' Top of active fuel is cefired to be 360 inches atove vessel zero ($ee Bases 3.2).
i 1.1/2,1-2a Acencrent No.
07298
.)
1 ~ _+.,_
i Qua0-CITIES OPR-29
. TABLE 3.1-2
!NSTRUMENTATION THAT INITIATES OR CONTROLS THE C0et AND CONTAINMENT C0OLING SYSTEMS Minimum number of Operable or Tripped Instrume channals Trin Function Trio Laval tattina Remarki 4 Reactor low low 184 inches above 1. In conjunction with low-water level top of active fuel' reactor pressure initiates core spray and LPCI.
- 2. In conjunction with high-drywell pressure 120-second time delay and low-pressure core cooling interlock initi-4tes auto blowdown.
- 4. Initiates startirg of diesel generators.
4E4} 12.5 psig 1. Initiates core spray. LPCI.
High-drng}1*III pressurek HPCI. and 58GTS.
- 2. In conjunction with low low water level. 120-second time delay, and low-pressure core cooling interlock initiates auto blowdown.
- 3. Initiates starting of diesel generators.
- 4. Initiates isolation of contral room ventilation.
2 Reactor low 300 psigip1350 plig 1. Permissive for opening core pressure spray and LPCI admission valves.
- 2. In conjunction with low low reactor water level initiates core spray and LPCI.
Containeent spray Prevents inadvertent operation of interlock containment spray during accident conditions.
2 I 2/3 core height 12/3 core height 4.3 containment 0.5 psigipil.5 psig high pressure 2 Timer auto 1120 seconds In conjunction with low low blowdown reactor water level, high-drywell pressure, and low-pressure core cooling interlock initiates auto blowdown.
4 Low-pressure 100 psigip1150 psig Defers APR actuation pending con-core cooling pump firnution of low-pressure core discharge pressure cooling system operation.
2/8U5($] Undervoltage on 3045 2 5% volts 1. Initiates starting of diesel emergency buses generators.
- 2. Permisst e for starting ECCS pumps.
- 3. Removes nomessential loads from buses.
- 4. Bypasses degraded voltage timer.
- Top of active fuel is defined at 360" above vessel sero for all water levels used in the LOC A analysis i
3.2/4.2-12 AmendPent No.
07238 l.
WBWG
. . OPR-29 3.2 LIMITING CON 0! TION FDA OPERATION BASES In addition to reactor protection instrumentation which initiates a reactor scram, protective instrumentation has been provided which initiates action to mitigate the consequences of accidents which are beyond the operator's ability to control. or tenninates operator errors before they result in serious consequences. This set of specifications provides the limiting conditions of operation for the primary system isolation function, initiation of the emergency core cooling system, control rod block and standby gas treatment systems. The objectives of the specifications are (1) to assure the effectiveness of the protective instrumentation when required by preserving its capability to tolerate a single f ailure of any component of such systems even during periods when portions of such systems are out of service for maintenance and (2) to prescribe the trip settings required to assure adequate perfornunce. When necessary, one channel may be made inoperable for brief intervals to conduct required functional tests and calibrations. Some of the settings on the instrumentation that initiates or controls core and containnent cooling have tolerances explicitly stated where the high or low values are both critical and may have a substantial effect on safety. It should be noted that the setpoints of other instrumentation. where only the high or low end of the setting has a direct bearing on safety. are chosen at a level away f rom the normal operating range to prevent inadvertent actuation of the safety system involved and exposure to abnormal situations.
Isolation valves are installed in those lines that penetrate the primary containment and must be isolated during a loss-of-coolant accident so that the radiation dose limits are not exceeded during an accident condition. Actuation of these valves is initiated by the protective instrumentation which serves the condition for which isolation is required (this instrumentation is shown in Table 3.2.1). Such instrumentation must be available whenever primary containment it'.egrity is required. The objective is to isolate the primary containnent so that the guidelines of 10 CFR 100 are not exceeded during an accident.
The instrumentation which initiates primary system isolation is connected in a dual bus arrangement. Thus the discussion given in the basis for Specification 3.1 is applicable here.
The low reactor level instrunentation is set to trip at > 8 inches on the level instrument (top of active fuel is defined to the 360 inches above vessel zero) and after allowing for the full power pressure drop across the steam dryer the low-level trip is at 504 inches above vessel zero. or 144 inches above the top of active fuel. Retrofit 8x8 fuel has an active fuel length 1.24 inches longer than earlier fuel designs.
However, present trip setpoints were used in the LOCA analyses *. This trip initiates closure of Group 2 and 3 primary containnent isolation valves but does not trip the recirculation pumps (reference SAR Sectice 7.7.2). For a trip setting of 504 ihches above vessel zero and a 60-second valve closure time, the valves will be closed before perforation of the cladding occurs even for the maximum break; the setting is therefore adequate.
The low low reactor level instrumentstion is set to trip when reactor water level is 1 444 inches above vessel zero (with tcp of active fuel defined as 360 inches above l vessel zero. -59 inches is 84 inches above the top of active fuel). This trip initiates closure of Group 1 primary containment isolation valves (reference SAR Section 7.7.2.2) and also activates the ECC subsystems, starts the emergency diesel generator, and trips the rectreulation pumps. This trip setting level was chosen to be low enough to prevent spurious operation but high enough to inittate ECC5 operation and primary system isolation so that no melting of the fuel cladding will occur and so that postaccident cooling can be accomplished and the guidelines of 10 CFR 100 will not be enceeded. For the complete circumferential break of a 28-inch rectreulation line and with the trip setting given above. ECCS initiation and primary isolation are initiated and in time to meet the above criteria. The instrumentation also covers the full spectrum of breaks and meets the above criteria.
The high-drywell pressure instrwnentation is a backup to the water level instrumentation and, in addition to initiating ECCS. It causes isolation of Group 2 tsolation valves.
For the breaks discussed above, tnis instrwnentation will initiate ECCS operation at about the same tire as the low low water level instrunentation; thus the results given above are applicable here also. Group 2 1 solation valves include the drywell vent, purge and sump tso14 tion valves. Hign-drywell pressure activates only these valves because high drywell pressure could occur as the result of non-saf ety-related causes such as not purging the drywell air during start-up. Total system isolation is not desirable for these conditions, and only the valves in Group 2 are reoutred to close.
The low low water level instrumentation initiates protection for the full spectrum of loss-of-coolant acetdents and causes a trip of Group 1 primary system isolation valves.
l
- toss of coolant accident analysts for Dresden Units 2 & 3 and Quad Cities Units 1 &
- 2. NE00 24146A. April, 1979 3.2/4.2-Sa Amendment No.
( 07238/0336Z
- ^
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OPR-10 l i- s The definitions used above for the APRM scram ta'e apply. In the event of operatiot .~th a maximum fraction limiting powc density (MFtPD) greater than the fraction of rated power (FRP) the setting shall be modified as follows:
$i (.58Wh + 50) _ FRPMFLPD The definitions used above for the APRM scram trip apply.
The ratio of FRP to MFLPD shall be set equal to 1.0 unless the actual operating value is less than 1.0, in which case the actual operating value will be used.
This adjustment may also be performed by increasing the APRM gain by the inverse ratio. MFLPC/FRP which accomplishes the same degree of protection as reducing the trip setting by FRP/MFLPO.
C. Reactor low water level scram setting shall be 144 inches above the top of the active fuel
- at normal operating condit ions.
D. Reactor low water level ECCS initiation shall be 1 84 inches above l the top of the active fuel' at normal operating conditions.
E. Turbine stop valve scram shall be i 10% valve closure from full open.
F. Turbine control valve fast closure scram shall initiate upon actuation of the fast closure solenoid valves which trip the turbine control valves.
G. Main steamline isolation valve closure scram shall be i 10% valve closure from full open.
H. Main steamline low-pressure ini-tiation of main steamline isolation valve closure shall be 1 825 psig.
"Top of active fuel is defined to te 360 inches above vessel zero (See Bases 3.2).
1.1/2.1-2a Amensment No.
07294
s 00AO-CITIES OPR-30 3.2 1.!ti! TING CON 0! TION FOR OPERATION 8ASES t'
In addition to reactor protection instrumentation which initiates a reactor
. scram, protective instrumentation has been provided which initiates action to mitigate the consequences of accidents which are beyond the operator's obtlity to control, or terminates operator errors before they result in serious consequences. This set of specifications provides the limiting conditions of operation for the primary system isolation function. initiation of the s emergency core cooling system, control rod block and standby gas treatment systems. The objectives of the specifications are (1) to assure the effectiveness of the protective instrumentation when required by preserving its capability to tolerate a single failure of any component of such systems even 1
during periods when portions of such systems are out of service for maintenance and (2) to prescribe the trip settings required to assure adequate performance. When necessary, one channel may be made inoperable for brief intervals to conduct required functional tests and calibrations. Some of the settings on the instrumentation that initiates or controls core and containment cooling have tolerances explicitly stated where the high or low values are both critical and may have a substantial effect on safety. It should be noted that the setpoints of other instrumentation, where only the high or low end of the setting has a direct bearing on safety are chosen at a level away from the normal operating range to prevent inadvertent actuation of the safety system involved and exposure to abnormal situations.
Isolation valves are installed in those 11nes that penetrate the primary containment and must be isolated during a loss-of-coolant accident so that the radiation dose limits are not exceeded during an accident condition. Actuation of these valves is initiated by the protective instrumentation which serves the condition for which isolation is required (this instrumentation is shown in Table 3.2.1). Such instrumentation must be available whenever primary containment integrity is required. The objective is to isolate the primary containment so that the guidelines of 10 CfR 100 are not exceeded during an accident.
The instrumentation which initiates primary system isolation is connected in a dual bus arrangement. Thus the discussion given in the basis for $pecification 3.1 is appitcable here.
The low reactor level instrumentation is set to trip at > 8 inches on the level instrument (top of active fuel is defined to be 360 inches above vessel aero) and af ter allowing for the full power pressure drop across the steam dryer the low-level trip is at 504 inches above vessel zero, or 144 inches above the top of active fuel. Retrofit 8x8 fuel has an active fuel length 1.24 inches longer than earlier fuel destpns. However, present trip setpoints were used in the LOCA analyses (NE00-24146A, April 1979). This trip initiates closure of Group 2 and 3 primary containment isolation valves but does not trip the recirculation pumps (reference SAR section 7.7.2). For a trip setting of $04 inches above vessel zero (144 inches above top of active fuel) and a 60-second valve closure time, the valves will be closed before perforation of the cladding occurs even for the maximum break: the setting is therefore adequate.
The low low reactor level instrumentation is set to trip when reactor water level is 1 444 inches above vessel zero (with top of active fuel defined as 360 inches above vessel zero. -59 inches is 84 inches above the top of active l fuel). This trip initiates closure of Group 1 primary containment isolation valves (reference SAR Section 7.7.2.2) and also activates the ECC subsystems.
starts the emergency diesel generator, and trips the recirculation pumps. This trip setting level was chosen to be low enough to prevent spurious operation but high enough to initiate ECC$ operation and primary system isolation so that no melting of the fuel cladding will occur and so that postaccident cooling can be accomplished and the guideltnes of 10 CFR 100 will not be exceeded. For the ,
complete circumferential break of a 28-inch recirculation line and with the trip setting given above. ECCS initiation and primary isolation are initiated and in time toofmeet full spectrum theand breaks above criteria.
meets The instrumentation the above criteria. also covers the The high-drywell pressure instrumentation is a backup to the water level instrumentation and. in addition to initiating ECCS it causes tsolation of Group 2 1 solation valves. For the breaks discussed above. this instrumentation will initiate ECCS operation at about the same time as the low low water level instrumentation; thus the results given above are applicable here also. Group 2 isolation valves include the drywell vent. purge and sump isolation valves.
High-drywell pressure activates only these valves because high drpell pressure could occur as the result of non-saf ety-related causes such as not purging the drywell air during start-up. Total system isolation is not desirable for these conditions, and only the valves in Group 2 are reoutred to close, The low low water level instrumentation initiates protection for the full spectrum of ,
loss-of-coolant accidents and causes a trip of Group 1 primary system isolation valves.
07248/03362 3.2/4.2-5a Amendment No.
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DPR-30 TABLE 3.2 2 INSTRUMENTATION THAT INIT! Aft $ OR CONTROLS THE Cott AND CONTAINMENT COOLING SYSTEMS Minimum Number of Operable or Tripped Instrumept, thannalt w M Function Trio Laval tattina Ramarki 4 Reactor low low 184 inches above 1. In conjunction with low-water level top of active fuela reactor pressure initiates l core spray and LPCI.
- 2. In conjunction with high-dryseell pressure 120-second time delay and low-pressure core cooling interlock initi-ates auto blowdown.
- 4. Initiates starting of diesel generators.
4(4) High-dryy 1 12.5 psig 1. Initiates Core Spray. LPCI.
pressuret
- I3I HPCI. and $BGTS.
- 2. In conjunction with low low water level.120-second time delay, and low-pressure core cooling interlock initiates auto blowdown.
- 3. Initiates starting of diesel generators.
- 4. Initiates isolation of control room ventilation.
2 Reactor low 300 pstgipi350 psig 1. Permissive for opening core pressure spray and LPCI aenission valves.
- 2. In conjunstion with low low-reactor mater level initiates core spray and LPCI.
Containment spray Prevents inadvertent operation of interlock containment spray during accident conditions.
2 3 2/3 core height 12/3 core height l3l 4.< containment 0.5 p51g1911.5 psig high pressure 2 Timer auto 1120 seconds In conjunction with low low blowdown reactor water level, high-drpeell pressure. and low-pressure core cooling interlock initiates auto blow-down.
4 Low-pressure 100 psigipi l50 psig Defers APR actuation pending con-core cooling pump firmation of low pressure core discharge pressure cooling system operation.
Undervoltage on 3045 i $1 volts 1. Initiates starting of diesel 2/BU$($)
emergency buses generators.
- 2. Permissive for starting (CCS pump 5.
- 3. esmoves nonessential loads from buses.
- 4. Bypasses degraced voltage timer.
- Top of active fuel ss defined at 360" above vessel aero for all water levels used in the LOCA analysis.
07240 3.2/4.2-12 Amenenent No.
ATTAQ9ENT 2 SUte1ARY OF CHANGES Eight (8) changes to the Quad Cities Station Units 1 and 2 Technical specifications have been identified and are listed below as follows:
- 1) Page 1.1/2.1-2a, Item (D). DPR-29 and 30 (a) Delete "84 inches (+4 inches /-0 inch)" and replace with 7t84 inches"
- 2) Page 3.a/4.2-5a. DPR-29 and 30 Fifth Paragraph (a) Insert "2" before the number 444 inches so the sentence now reads
...when reactor. water level islt444 inches..."
(b) Delete "high enought to prevent spurious operation but low enough" and replace with "low enough to prevent spurious cjeration". This change replaces the word "high" with "low" and "Icw" with "high".
6
- 3) Page 3.2/4.2-12. Table 3.2-2. DPR-29 and 30 (a) Under column titled "Trip Level Setting" for the Reactor Low Low Water Level Trip Function, delete "(+4 inches /-0 inch)" so it instead reads as follows, "184 inches above top of active fuel *".
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ATTAQWENT 3 d I
I EVALUATION CF SIGNIFICANT HAZ/JtDS CONSIDERATION AND DESCRIPTION OF PROPOSED AMENDMENT. REQUEST l l
l There are two types of changes associated with this proposed amendment request;.the first change deletes the +4/-0 inch tolerance for the i Reactor Low Low Level Trip Setpoint while the second change corrects a wording error in the Limiting condition for Operation basis of Section-3.2 of the i I
Technical Specifications.
1 Review of Station records has indicated that the-upper limit for core. :
.and containment cooling initiation setpoint is not always met. . Current procedures have resulted in cases where the core and containment cooling systems would have initiated above the water levels addressed in the Technical specifications if an actual low level condition occurred. This would ,
result in unnecessary challenges and activations of the core and containment cooling systems. A review of the Technical Specification Bases has shown that i the upper limit has no safety significance. Deleting the upper tolerance will also help facilitate calibration of the Reactor Low Low Water Level instrumentation. j i
The-second change interchanges the words "high" and "low" in a ,
sentence that explains the setting of the Reactor Low Low Water Level Trip. [*
This change is administrative in nature and is being sought to insure that the Bases are consistent with the information pertaining to the setpoint in the ;
Technical specifications.
BASIS FOR NO SIGNIFICANT HAEARDS DETERMINATION i t
&==annwealth ddison has evaluated this proposed amendment and ,
determined that it involves no significant hazards consideration. In j accordance with the criteria of 10 CFR 50.92(c), a proposed amendment to an l-
. operating license involves no significant hazards considerations if operation of the facility, in accordance with the proposed amendment, would not: [
- 1) Involve a significant increase in the probability or consequences of an j accident previously evaluated because: i l
(a) The actual setpoint for the Reactor Low Low Level is not being f a
changed and therefore the probability or consequences of an accident !
previously evaluated is unchanged. .
(b) The second change merely corrects a wording error in the basis of the !
Technical Specifications. This change is considered to be l
- administrative in nature. !
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- 2) create the possibility of a new or different kind of accident from any ,
accident previously evaluated because:
(a) The upper bound of the tolerance range associated with the Reactor Low Low Water Level Trip Setpoint serves no safety function and the conservatism associated with the actual setpoint is not being reduced. The actual setpoint which would have been used in accident evaluation is unchanged. Hence, the proposed amendment does not create the possibility of a new or different kind of accident than which was previously evaluated.
(b) The second change merely corrects a wording error in the Basis of the Technical Specification. This change is considered to be administra-tive in nature.
I
- 3) Involve a significant reduction in the margin of safety because:
(a) The proposed amendment does not change the actual trip setpoint but merely reflects a change to the tolerance. As a result, the margin of safety has not been decreased as a result of this proposed amendment.
(b) The second change merely corrects a wording error in the Basis of the Technical Specification. This change is considered to be administra- !
tive in nature, ,
Therefore, since the proposed license amendment satisfies the criteria specified in 10 CPR 50.92, commonwealth Edison has determined that a no significant hazards consideration exist for these items. We further request their approval in accordance with the provisions of 10 CFR 50.91(a)(4).
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