ML19341C458
| ML19341C458 | |
| Person / Time | |
|---|---|
| Site: | Zion File:ZionSolutions icon.png |
| Issue date: | 02/17/1981 |
| From: | COMMONWEALTH EDISON CO. |
| To: | |
| Shared Package | |
| ML19341C455 | List: |
| References | |
| NUDOCS 8103030545 | |
| Download: ML19341C458 (20) | |
Text
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APPENDIX A TECHNICAL SPECIFICATION CHANGE NUMBER 81-1 PAGES MODIFIED DESCRIPTION vii Figure 6.1.1 title change and two missing figure titles added.
vili Table 3.8.9-1 and two titles added to listing. An incorrect title corrected.
ix Table 4.8.9-1 added to listing x
Page x retyped to accomodate increased size of
" List of Tables".
No changes to text.
75 The requirements for an operable pressurizer have been upgraded to require 150KW of heaters and a specified water level.
76 Page 76 retyped to accomodate insertions. No changes to text.
77 The requirements for power operated relief valves (PORV) and their associated block valves have been included in the technical specifications.
78 Bases for the pressurizer, PORV and block valves have been added.
131 Section V, " Auxiliary Feedwater" has been added to Table 3.4-1, " Engineered Safeguards Actuation System - Limiting Conditions on Operation and Setpoints" The descriptions for the Permissives, P-ll and P-12, have been expanded.
131a Page 131a has been added to accomodate increased size of Table 3.4-1.
No changes to text on this page.
133 Page 133 has been retyped to accomodate change to Table 3.4-2.
No change to the text on this page.
133a Page 133a has been added with the inclusion of Section V, " Auxiliary Feedwater" to Table 3.4-2,
" Engineered Safeguards System Instrument Numbers" 135 Section V, " Auxiliary Feedwater" has been added to Table 4.4-1, " Engineered Safequards System Testing and Calibration Requirements" 184 Specification 3.8.9 and 4.8.9 have been rewritten describing requirements for accident monitoring instrumentation 810 30 3 D 56
PAGES MODIF IED DESCRIF ION 192a Table 3.8.9-1 entitled " Accident Monitoring Instrumentation" added to Technical Speicification 192b Table 4.8.9-1 entitled " Accident Monitoring Instrumentation Surveillance Requirements" 195 D7e basis for Specification 3.8.9 has been rewritten.
The basis for the hydrogen recombiner has been rewritten. The reference to the first refueling of unit 1 has been deleted.
300 Section 6.1.0 now indicates requirements for Shift Technical Advisor. Page 300 is not copy of existina Technical Specifications, but of previous submittal. (Zion Tech Spec change No. 79-7, January 24, 1980) 331 Fiqure 6.1.1 now indicates requirement for Shift Technical Advisor.
LIST OF FIGURES Figure Page 3.3.2-1 Reactor Coolant System Heatup Limitations 84 3.3.2-2 Reactor Coolant System Cooldown Limitations 85 3.3.2-3 Fffect of Fluence and Copper Content on Shift of 86 RTNDT for Reactor Vessel Steels Exposed to 550 degrees F Temperature 3.3.2-4 Fluence at 1/4T and 3/4T as a Function of Full Power 87 Service Years 3.4-1 High Steam Line Flow Setpoint 131a 4.16-1 Location of Fixed Environmental Radiological Monitoring Station 278 6.1.1 Minimum Shift Crew Composition 331 J.
I vii
LIST OF TABLES Table Page 3.1-1 Reactor Protection System-Limiting Operations 30 Conditions and Setpoints 3.1-2 Reactor Prouection System Instrume.;t Numbers 33 3.3.2-1 RTNDT Testing Results 88 l
3.3.5-1 Reactor Coolant Systems and Chemistry Specifications 122 3.4-1 Engineered Safeguards Actuation System-Limiting 129 Conditions on Operation and Setpoints 3.4-2 Engineered Safeguards System Instrument Numbers-132 3.7-1 Maximum Allowable Power Range Neutron Flux High Trip 160a Setpoint With Inoperaole Steam Line Safety Valves During Four Loop Operation 3.7-2 Maximum Allowable Power Range Neutron Flux High Trip 1600 Setpoint with Inoperable Steam Line Safety valves During Three Loop Operation l
3.8.9-1 Accident Monitoring Instrumentation 192a 3.15-1 Equipment Requirement witn Inoperative 4KV E.S.S. Bus 268 3.15-2 Equipment Inoperdole with Inoperative 4KV E.S.S. Bus 269 4.1-1 Reactor Protection System Testing and Calibration Requirements 35 t
4.3.B-1 Minimum Numoer of Steam Generators to be Inspected During loservice Inspection 741 4.3.B-2 Steam Generator Tube Inspection 74j 4.4-1 Engineered Safeguards System Testing and Calibration 134 Requirements viii
LIST OF TABLES (cont)
Table Page 4.4-2 Engineered Safety Equipment Actuation Test 136 4.5-1 Containment Fan Cooler Lomponents 148 4.6-1 Containment Spray System Components 153 4.7-1 Steam Generator Safety Valves, Set Pressures, Orifice Sizes 160 and Steam Flows 4.7-2 Auxiliary Feedwater Pumps 161 4.8-1 Centrifugal Cnarging Pump System 185 4.8-2 Safety Injection Pump System 186 4.8-3 Residual Heat Removal Pump System 187 4.8-4 Accumulator Tanks 188 4.8-5 Component Cooling PJmp System 189 4.8-6 Service Water Pump System 190 4.8-7 Hydrogen Control System 192 4.8.9-1 Accident Monitoring Instrumentation Surveillance Requirements 1920 4.9-1 Isolation Seal Water System 203 4.9-2 Penetration Pressurization System 204 4.9-3 Containment Isolation Valves 205 4.9-4 Main Steam Isolation Valves 208 ix
LIST OF TABLES (cont)
Tecle Page 4.11-1 Radioactive Liquid Waste Sampling and Analysis 226 j
4.12-1 Pathways of Release 236 4.12-2 Radicactive Gaseous Waste Sampling and Analysis 237 4.12-3 Effluent Gaseous Waste Monitor 239 4.14-1 Process and Internal Monitoring 252 4.15-1 4160-Volt Engineered Safeguard Bus Main, Reserve and StancDy Feeds 270 4.16-1 Zion Standard Radiological Monitoring Program 276 4.17-1 HEPA/ Charcoal Filter Systems Surveillance Requirements 284 4.17-2 Particulate Filter System Surveillance Requirements 286 4.19-1 Failed Fuel Monitoring Instruments 295 6.6.2 Special Reports 328a 6.3.1 Boundary Doors for Flood Concitions 332 x
4 LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT.
3.3.1 C.
Pressurizer Safety Valves 4.3.1 C.
Pressurizer Safety Valves 1.
At least one code safety valve shall 1.
Safety valves RC8010A, RCodiOG and be operable anenever the vessel is RC8010C shall be tested for set closed, except during hydrostatic pressure at each refueling outage. Testing shall be done oy tests.
a calibrated auxiliary lifting 2.
All code safety valves shall be device or oy bench testing using operable wherever the reactor compressed gas. Tne valve coolant temperature is above 200'F.
setpoints snall be 2485 + 1% psig.
D.
Pressurizer 2.
Not Applicable.
The pressurizer shall be OPERABLE with at D.
Pressurizer least 150kw of pressurizer heaters and a water level not to exceed 92%.
1.
The pressurizer volume snall De determined to be witnin its limit APPLICABILITY: Modes 1, 2, 3 at least once per snift.
ACTION:
a.
With the standby AC on-site power 2.
The'standoy AC on-site power
~
supply (diesel generators) to the supply (diesel generators) for the pressurizer heaters inoperable; pressurizer heaters shall be either restore the inoperaole demonstrated OPERABLE at least standby AC on-site power supply once per 18 months oy transferring (diesel generators) within seven power from the normal to the days, or be in at least HOT SHUTDOWN stanooy AC on-site power supply within the following four hours.
(diesel generators) ano energizing the heaters.
b.
With the pressurizer otherwise inoperable, be in at least HOT SHUTDOWN with the reactor trip breakers open within six hours and
.in the HOT SHUTDOWN condition with Tavg less than 350'F within the following six nours.
75
L MITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT 3.3.1 E.
Loop Stop Valves 4.3.1 E.
Loop Stop Valves 1.
No more than one loop shall be isolated 1.
'Not-' Applicable.
unless the Reacter Coolant System is connected to the Residual Heat Removal
.,'~'~~~
o System and the Residual Heat Removal System is operable.
2.
Whenever a reactor coolant loop is 2.
The coron concentration in tne isolated, the boron concentration in the isolated loop shall be oetermineo isolated loop shall be maintained at a at least every 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
value greater than or equal to the boron concentration in the active loops.
3.
Isolated Reactor Coolant Loop Startup 3.
Isolated Reactor Coolant Loop Startup n'nenever startup of an isolated reactor coolant loop is initiated the following conditions shall be met:
a.
All the channels, including a.
Tne interlocks associated witn redundant channels, of the Loop Stop the loop stop valves shall be Valve Interlock System of tne verified every refueling isolated loop are operable, except outage.
as specified in 3.3.1.E.3.f.
In the event this condition is not satisfied, the loop must remain isolated.
b.
Tne reactor shall De in a shutdown D.
Not Applicable, conditions prior to opening either stop valve and througnout tne timing interval required oy 3.3.1.E.3.a prior to opening the cold leg stop valve.
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Actuation Channel
.No. of Minimum Minimum Operator Action Description (Per Unit)
No. o f Channels Operable Degree of if column 3 or 4 Channels to trip Channels 64+
Redundancy ooo cannot ce met o Setpointsoo V.
Manual 1/ pump 1/ pump 1/ pump 0
Maintain Hot Soutdown***
N.A.
2.
Auton.atic 2
1 2
1 Maintrin Hot Shutdown ***
N.A.
3.
Steam Generator (S/G)
Water Level low-low i Start Motor 2 per C/G Driven Pumps 3 per S/G any le~ S/G 2 per S/G 1 per S/G Maintain Hot Snutoown*** 10%
Narrow Ranga 11 Start Turbine 2 per S/G Driven Pumps 3 per S/G any 2/4 S/G 2 per S/G 1 per S/G Maintain Hot Snutoown*** 10%
Narrow Rangs 4.
Undervoltage-RCP busses 75%
Start Turbine Driven Pump 4-1/ bus 2
3 1
Maintain Hot Snutdown*** RCP dus Voltage 5.
S.I. Start Motor ano Turbine Driven Pumps 2
1 2
1 Maintain Hot Snutoown***
N.A.
6.
Station Blackout 3-1/ bus 2
2 1
Maintain Hot Shutdown *** Time Start Motor and Turbine Driven Pump Depenaent on Voltage PERMISSIVES Setpoint +
P-11 Pressurizer pressure (2/3) celow 1915 psig allows manual block of safety injection actuation during a plant cooldown.
l P-12 Tavg (2/4) Delow 540'F allows manual block of High Steam Flow safety injection actuation if borated to greater than cold shutdown conditions.
SEE FOOTNOTES ON FOLLOWING PAGE.
ENGINEERED SAFEGUARDS ACTUATION SYSTEM - LIMITING CONDITIONS FOR OPERATION AND SETPOINTS TABLE 3.4-1 (CONTINUED) 131
+If minimum conditions are not met within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, the unit shall be in the COLD SHUTDOWN condition within an additional 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
++Setpoints are + established tolerances for instrument channel and setpoint errors as specif!Pd in "Chann21 Accuracies, Overall Channel Accuracies and Set-point Tolerances for y[ NES Process I and C Reactor Protection ana Lontrol Systems" August 30,1971 - CEW-2652.
Setting.
T..c instruments shall no* be set to exceed a Limiting Safety System
+++Tnis channel may be placed in the bypass moae during periods of active testing during safeguards eqeipment testing as specified in Section 4.4.2.
- Requires simultaneous actuation of two switenes.
- ' Maintain Hot Shutdown' means maintain or be in HOT SHUTDOWN witnin four hours if tne unacceptable condition arises during power operation.
ENGINEERED SAFEGUARDS ACTU4 TION SYSTEM - LIMITING PP;JITIONS FOR OPERATION AND SETPOINTS (Footnotes to Tacle)
TABLE 3.4-1 131a
Channel Description Device Designation III.
CONTAINHENT ISOLATION A)
Phase A 1.
Manual Actu_tica N.A.
2.
Safety Injection Section I of this table.
B)
Phase B 1.
Manual Actui. tion N.A.
2.
Automatic Actuation N.A.
3.
High-High Containment Pressure PT-CS19, PT-CS20, PT-CS21, PT-CS22 IV.
STEAM LINE ISOLATION 1.
Manual Actuation N.A.
2.
Automatic Actuation N.A.
3.
High-High Containment Pressure PT-CS19, PT-CS20, PT-CS21, PT-CS22 4.
High Steam Line Flow in
- a. Flow:
FT-512, FT-513, FT-522, FT-523 Coincidence with Low-Low Tavg FT-532, FT-533, FT-54?., FT-543 or Low Steam Line Pressure
- b. Temperature:
TE-411A, TE-411B, TE-421A, TE-421B TE-431A, TE-431B, TE-441A, TE-441B Pressure:
PT-516, PT-326, PT-536 PT-546 ENGINEERED SAFEGUAADS SYSTEM INSTRUMENT hv oERS TABLE 3.4-2 Continued 133
Channel Description Device Description V.
Manual NA 2.
Automatic NA 3.
Steam Generator LC-5178, LC-5278, LC-5378, LC-5478 Water Level Low-low LC-5188, LC-5288, LC-5388, LC-5488 LC-5198, LC-5298, LC-5398, LC-5498 4.
Undervoltage - RCP Busses Start Turbine Driven Pump 27-1, 27-2, 27-3, 27-4 5.
SI Start Motor and See Section I of this Table.
Turbine Driven Pumps 6.
Station Blackout Start Unit I 427x2-142, 427x2-143, 427x2-144 Motor and Turbine Unit II 427x2-242, 427x2-243, 427x2-244 Driven Pumps PERMISSIVES P-ll Pressurizer oressure - PT-455, PT-456, PT-457 P-12 Temperature - TE-AllA, TE-4118, TE-421A, TE-421B, TE-431A, TE-4318, TE-441A, TE-4418 rNGINEERED SAFEGUARDS SYSTEM INSTRUMENT NUMBERS TABLE 3.4-2 (Continued) 133a
ACTUATION CHANNEL CHAT @EL
.CH rsEL CHANNEL DESCRIPTION CHECK CALIBRATION FitNCT. TON TEST j
IV.
STEAMLINE ISOLATION 1
i 1.
Manual Actuation N.A.
N.A.
R 2.
Automatic Actuation N.A.
N.A.
M 3.
High-High Containment Pressure See Item II Aoove
.I 4.
High Steam Line Flow in Coincidence with Low-Low Tavg See Item I Above I
or l
Low Steam Pressure V.
Manual N.A.
N.A.
R 2.
Automatic N.A.
N.A.
M 3.
R M
Water Level Low-Low 4.
Undervoltage - RCP Busses N.A.
R R
5.
Safety Injection See Item I on Page 134 6.
Station Blackout N.A.
R R
PERMISSIVES 1.
P-ll N.A.
N.A.
M 2.
P-12 N.A.
N.A.
M NOTE: Specified intervals may be adjusted 225% to accommooate test senedules S - Once per shift M - Once per month N.A. - Not applicaole R - Once per refueling shutdown - calibration of tnese instruments may be done as much as six months prior to tne start of refueling outage and still satisfy this requirement.
ENGINEERED SAFEGUARDS SYSTEM TESTING AND CALIBRATION REQUIREMENTS TABLE 4.4 (Continued) i
LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT 3.8.9 The accident monitoring instrumentation 3.8.9 Each accident monitoring instrumentation channels shown.in Table 3.8.9-1 shall De enannel snall be demonstrated OPERABLE by 0PERABLE.
performance of the IN$TRUMENT CHANNEL i
CALIBRATION at the frequencies snown in i
APPLICABILITY:
Modes 1, 2 and 3 l
Table 4.8.9-1.
ACTION:
a.
With the number of OPERABLE accident monitoring instrumentation channels less than the Required Number of Channels shown in Table 3.8.9-1, s
(Col. 2) either restore the inoperable channel (s) to OPERABLE status within seven days, or the unit shall be placed in at least the HOT SHUTDOWN mode, with T less ayg than 350'F (Mode 4), within the next
'i 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.*
1 D.
With the num0er of OPERABLE accident monitoring instrumentation channels I
less tnan the Minimum Number of 1
Cnannels shewn in Taole 3.8.9-1, (Col. 3) either restore the inoperable channel (s) to OPERABLE status within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> or the unit shall be placed in at least the HOT SHUTOOWN mode, witn Tava less than 350'F, (Mode 4) within be next 12 nours.
c.
Tne provisions of Specifications 3.0.4 are not applicable.
- This action does not apply to the PORV Position Indication or. the PORV Block Valve Position Indication if the Block Valve on the associated line is known to be closed either by verification within seven days Or by system status knowledge prior to indication failure.
184
TABLE 3.8.9-1 ACCIDENT MONITORING INSTRUMENTATION 1.
2.
3.
INSTRUMENT (PARAMETER)
TOTAL NO.
REQUIRED NO.
MINIMUM NO.
0F CHANNELS OF CHANNELS OF CHANNELS 1.
Containment Pressure (Narrow Range) 4 2
2.
1 Reactor Coolant Outlet Temperature - THot (Wide Range) 4-1/ Loop 2
1 3.
Reactor Cool. ant Inlet Temperature - TCold (Wide Range) 4-1/ Loop 2
4.
Reactor Coolant Pressure (Wide Range) 1 2
2 5.
Steam Line Pressure 1
12-3/SG 2/SG 6.
Pressurizer Water Level 1/SG 3
2 1
7.
Steam Generator Water Level (Narrow Range) 12-3/SG 2/SG 1/SG 8.
Steam Generator Water Level (Wide Range) 4-1/SG 2
9.
Refueling Water Storage Tank Level 1
2 2
10.
Auxiliary Feedwater Flow 1
4-1/SG 2
1 11.
Reactor Coolant System Subcooling Margin
- 2*
2*
12.
PORV Position Indication **
1*
2/ valve 2/ valse 13.
PORV Block Valve Position Indication 1/ valve 1/ valve 1/ valve 0
14.
Safety Valve ?osition Indication 3-1/ valve 2
1 procedure performed by the operator.The Reactor Coolant Subcooling Margin is Getermined by two analysis or, 2) a acoustical monittring system (backup indication).The PORV Position Indication r:onsists y inoication) and 2) an 192a
TABLE 4.8.9-1 ACCIDENT MONITORING INSTRUMENTATION SLRVEILLANCE REQUIREMENTS CHANNEL CHAR 4NEL INSTRUMENT (PARAMETER)
CHECK CALIBRATION 1.
Containment Pressure (Narrow Range)
M R
2.
Reactor Coolant Outlet Temperature - T ot (Wide Range)
M R
H 3.
Reactor Coolant Inlet Temperature - TCold (Wide Range)
M R
4.
Reactor Coolant Pressure (Wide Range)
M R
5.
Steam Line Pressure M
R 6.
Pressurizer Water Level M
R 7.
Steam Generator Water Level (Narrow range)
M R
8.
Steam Generator Water Level (Wide range)
M R
9.
Refueling Water Storage Tank Level M
R 10.
Auxillary Feedwater Flow M
R
- 11. Reactor Coolant System Subcooling Margin M
R 12.
PORV Position Indication M
R 13.
PORV Block Valve Position Indication M
R
- 1. 4. Safety Valve Position Indication M
R M: Montnly R: Rerueling 1920
_ _ _ _ - _ _ _ - _ - ~
A hydrogen recombiner system is installed to remove the If during this seven day period a third service water hydrogen and oxygen gases that accumulate in the pump becomes inoperable, the unit will be brought to containment atmosphere following a loss-of-coolant HOT SHUTDOWN within four hours unless one of the accident. (9) The system operability requirement inoperable pumps are returned to operable status, becomes effective following successful preoperational The requirement for independent standby AC and DC-testing.
pov.er supplies for an operating service water pump l Until completion of c3'culatory water pump maintenance iront the unit precludes the possibility of loeing two service water pump dedicated to the same unit because program, scheduled for completion about Decembar 31, of the failure of a common power supply.
1981, one service water pump at a time may be taken out of sarvice to perform maintenance on the associated The OPERABILITY of the accident monitoring circulating water pump.
In order to work on a instrumentation ensures that sufficient information circulating water pump, an intake plenum must be is available on selected plant parameters to monitor drained.
The intake plenum is a common reservoir for a and assess these variables during and fcllowing an service water cump and a circulating water pump.
This accident.
This cacdbility is consistent with the leaves five service water pumps in service, which will recommendations of Regulatory Guide 1.97, provide sufficient coverage for any postulated loss of
" Instrumentation for Light-Water-Cooled Nuclear Power coolant accident coincident with a loss of off-site Plants to Assess Plant Conditions During and power and any other single failure of an active Following an Accident," Decenter 1975 and NUREG-0578, comconent.
With this tenporary specification, should a "TMI-2 Lessons Learned Task Force Status Report and second service water cump 5ecome inoperable, it must be Short-Term Recommendations".
returned to operable status within seven days or the unit will be brought to HOT SHUTDOWN.
(1)
FSAR Chapter 9 (2)
FSAR Section 6.2 f
(3)
FSAR Section 6.2.3 (4)
FSAR Section 14.3 i
(5)
FSAR Section 9.3 (6)
FSAR Section 9.6 & FSAR answer to question 9.1 (7)
FSAR Section 14.3.6 (8)
FSAR Answer to Question 9.9 (9)
FSAR Section 6.8 195
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Unit in Mooes 1, 2, 3, 4 Unit 1 Unit 1 No or and Position Unit Unit 2 Unit 2 Shift Engineer or 1
1 2
Shift Forenan l Shift Control Room Engineer None 1
1 Required Nuclear Station Operator 1
2 3
Equipment Operator or 2
3 4
Eouipment Attendant Radiation Protection Person 1
1 1
TOTAL 6
8 11 MINIMUM
- 6 7
10
- The minimum numoer refers only for the case of snift snortage, caused by a sudden sickness or home emergency.
tbtes:
1.
Senior reactor Operator (SRO) shall be present on-site at all times unen there is fuel in the reactor.
2.
A licensed person shall De in the control room at all times whenever fuel is in either reactor.
3.
Two licensed people shall be in the control room during reactor startups, shutdowns, operation, and other periods sucn as planned control rod manipulations.
MINIMUM SHIFT CREW COMPOSITION Figure 6.1.1 331
)
ATTACHMENT 3 Zion Station Units 1 and 2 NRC Docket Nos. 50-295 and 50-304 Proposed License Conditions Systems Integrity The Licensee shall implement a program to reduce leakage from systems outside containment that would or could contain highly radioactive fluids during a serious transient or accident to as low as practical levels.
This program shall include the following:
1.
Provisions establishing preventive maintenance and periodic visual inspection requirements, and 2.
Integrated leak test requirements for each system at a frequency not to exceed refueling cycle intervals.
Iodine Monitoring The licensee shall implement a program which will ensure the capability to accurately determine the airborne lodine concentration in vital areas under accident conditions.
This program shall include the following:
1.
Training of personnel, 2.
Procedres for monitoring, and 3.
Provisions for maintenance of sampling and analysis equipment.
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