ML19270H801
| ML19270H801 | |
| Person / Time | |
|---|---|
| Site: | Zion File:ZionSolutions icon.png |
| Issue date: | 12/31/1979 |
| From: | COMMONWEALTH EDISON CO. |
| To: | |
| Shared Package | |
| ML19270H798 | List: |
| References | |
| NUDOCS 8001030634 | |
| Download: ML19270H801 (14) | |
Text
r ATTACHMENT 1 Technical Specification Changes to Steam Generator Emergency Heat Removal (T.S. Section 3.7 and 4.7)
Pages Modified; iii, viii, ix, x, 156, 158, 159, 161, 162, 163 Pages Added:
159a, 161a, 162a i
1677 352 s001030 -k39'
Table of Contents (continued)
SURVEILLANCE LIMITING CONDITION FOR OPERATION REQUIREMENT PAGE 3.7 Steam Generator Emergency Heat Removal 4.7 156 3.7.1 Steam Line Safety and Relief Valves 4.7.1 156 3.7.2 Auxiliary Feedwater Pump System, Per Unit 4.7.2 158 4.7.3 159a 3.7.3 Auxiliary Feedwater Supply System Bases 3.8 Emergency Core Cooling and Core Cooling Support 4.8 164 4.8.1 164 3.8.1 Centrifugal Charging Pump System 4.8.2 168 3.8.2 Safety Injection Pump System 3.8.3 Residual Heat Removal Pump System 4.8.3 170 3.8.4 System Testing of Centrifugal Charging, Safety Injection, and Residual Heat Removal Pump Systems 4.8.4 173 4.8.5 174 3.8.5 Accumulator System
.4.8.6 175 3.8.6 Component Cooling System 4.8.7 178 3.8.7 Service Water System 4.8.8 180 3.8.8 Hydrogen Control Systems 3.8.9 Equipment for Evaluating Post LOCA 4.8.9 184 Bases 197 3.9 Containment Isolation Systems 4.9 197 3.9.1 Isolation Valve Seal Water System 4.9.1 3.9.2 Penetration Pressurization Systems 4.9.2 198 3.9.3 Containment Isolation Valves 4.9.3 199~
3.9.4 Main Steam Isolation Valves and Bypasses 4.9.4 200 3.9.5 Containment Integrity 4.9.5 201 Bases 3.10 Containment Structural Integrity 4.10 212 4.10.1 212 3.10.1 Containment Leakage Rate Testing 3.10.2 Containment Tendon Testing 4.10.2 215 3.10.3 End Anchorages and Adjacent Concrete Surfaces 4.10.3 217 Inspection CN Containment Liner Inspection 4.10.4 218 3.10.4 4.10.5 219 3.10.5
[j Containment Pressure 4.10.6 219 3.10.6 Containment Temperature
(:
Bases 4.11 222 3.11 Ln Radioactive Liquids LN Bases 4.12 230 3.12 Radioactive Gases Bases iii
LIST OF TABLES Page Table 30 3.1-1 Reactor Protection System-Limiting Operating Conditions and Setpoints 33 3.1-2 Reactor Protection System Instrument Numbers 88 3.3.2-1 RT Testing Results NOT 106 3.3.4-1 In Service Inspection Program 122 l-3.3.5-1 Reactor Coolant Systems and Chemistry Specifications 129 3.4-1 Engineered Safeguards Actuation System-Limiting Conditions of Operation and Setpoints 132 3.4-2 Engineered Safeguards System Instrument Numbers Neutron Flux High Trip Points with Steam Generator Safety Valves Inoperable 160a 3.7-1 Four Loop Operation 160b 3.7-2 Neutron Flux High Trip Points with Steam Generator Safety Valves Inoperable-Three Loop Operation 3.15-1 Equipment Requirement with Inoperative 4KV E.S.S.
Bus 268 3.15-2 Equipment Inoperable with Inoperative 4KV E.S.S. Bus 269 OK
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35 4.1-1 Reactor Protection System Testing And Calibration Requirements t,
741 l4.3.B-1 'D Minimum Number:of/:SteamrGeneratorsitBbe= Inspected During Inservi'ce Inspections l 4. 3.B-2 Steam Generator Tube Inspection 74j 134 4.4-1 Engineered Safeguards System Testing and Calibration Requirements 136 4.4-2 Engineered Safety Equipment Actuation Test 148 4.5-1 Containment Fan Cooler Components viii
LIST OF TABLES (Continued)
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Table Page 4.6-1 Containment Spray System Components 153 4.7-1 Steam Generator Safety Valves, Set Pressures, Orifice Sizes and 160 Steam Flows 4.7-2 Auxiliary Feedwater Pump System 161 4.7-3 Auxiliary Feedwater Supply System 161a 4.8-1 Centrifugal Charging 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 18b 4.8-5 Component Cooling Pump System 189 4.8-6 Service Water Pump System 190 4.8-7 Hydrogen Control System 192 4.9-1 Isolation Seal Water System 203 4.9-2 CK Penetration Pressurization System 204
~4 4.9-3 Containment Isolation Valves 205 v.
4.9-4 ty, Main Steam Isolation Valves 208 LD 4.11-1 Radioactive Liquid Waste Sampling and Analysis 226 4.12-1 Pathways of Release 236 4.12-2 Radioactive Gaseous Waste Sampling and Analysis 237 4.12-3 Effluent Gaseous Waste Monitors 239 4.14-1 Process and Internal Monitoring 252 ix
LIST OF TABLES (Continued)
Page Table 4.15-1 4160-Volt Engineered Safeguard Bus Main, Resdrve, and Standby Feeds 270 4.16-1 Environmental Radiological Monitoring Program (ERMP) 276~
4.16-2 Practical Lower Limits of Detection for ERM" 280a 284 4.17-1 Charcoal Filters 285 4.17-2 HEPA Filters 4.19-1 Failed Fuel Monitoring Instrumentation 295 4.21-1 Fire Protection Instruments 295p 295r 4.21-2 Fire Suppression Water System 295s 4.21-3 Sprinkler Systems 295t 4.21-4 CO Systems 2
295u 4.21-5 Fire Hose Stations 332 6.3.1 Boundary Doors For Flood Conditions 328a 6.6.2 Special Reports
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O LIMITINd CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT 3.7 STEAM GENERATOR EMERGENCY HEAT REMOVAL
~4.7 STEAM GENERATOR EMERGENCY HEAT REMOVAL Applicability:
Applicability:
Applies to auxiliary feedwater pump system, Applies' to surveillance of auxiliary auxiliary feedwater supply system, and steam feedwater pump system, auxiliary feedwater generator safety valves.
supply system and steam generator safety.
valves.
Objective:
Objective:
To insure adequate plant cooldown capabili-To insure availability of the above system ties upon loss of normal feedwater flow and and valves, loss of main c;ndenser vacuum.
Specification:
Specification:
1.
Steam Line Safety and Relief Valves, l.
Steam Line Safety and Relief Valves l
per unit
- l per unit.
A.
Twenty ASME code safety valves A.
Ten steam generator safety valves (5 per steam generator) shall be per unit shall be tested for set operable whenever the reactor.is pressure at each refueling ou,tage.
heated above 350 F except as Testing shall be done by a calibrated 0
~specified in 3.7.1.C, 3.7.1.D, au::iliary lif ting device or by bench and 3.7.1.E testing on compressed gas.
At Jaast two of the valves tested shall be from each orifice size
("Q" or "R").
All valves on a unit shall have been
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tested at the end of each second re-fueling outage.
The valves and the
-y corresponding set pressures and orifice sa sizes are identified in Table 4.7-1.
L' B.
Deleted B.
Deleted LT1
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156
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i LIMITING CONDITION FOR OPERATION f
SURVEILLANCE REQUIREMENT 3.7.1 E.
When the reactor is operating on 3 loops, 4.7.1 E.
Not applicable at least two code safety valves associated i
with the remaining steam generator must be operable.
i F.
If these conditions cannot be met the F.
Not applicable the reactor shall be brought to the hot i
shutdown conditions within four hours.
After a maximum of 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> in the hot shutdown conditions, if the system is not operable the reactor shall be broughti i
i to the cold shutdown condition within, i
24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
2.
Auxiliary feedwater pump system, per unit, 2
.i Auxiliary Feedwater numn system, per unit (Table 4.7-2)
A.
Three auxiliary feedwater pump systems A.
Surveillance and testing of the shall be operable whenever the reactor auxiliary feedwater pump system's is above 350aF except as specified shall be performed as follows; in 3.7. 2.B, 3.7. 2. C, and 3.7. 2.D.'
the auxiliary feedwater pumps shall be started manually from the control room each month.
Performance will be j
I acceptable if:
s 1) the pump starts upon actuation, operates for at least 17 minutes on recirculation flow, and the discharge pressure and flow are ch within +1095 of a point on the N
pump holid curve, and
2) the pump provides at least 105 gallons per minute flow to each q,
steam generator, en oo 158
SURVEILLANCE REQUIREMENT LIMITING CONDITION FOR OPERATION is above 350*F, the 4.7.2.B The surveillance and testing of the 3.7.2.B.
Whenever the reactor turbine-driven auxiliary feedwater pump turbine driven auxiliary feedwater pump system shall be operable except that veri-as required by 4.7.2.A need not be fication by testing need not be completed performed prior to exceeding 350*F until required by Section 4.7.2.B.
but must be performed within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> af ter achieving hot standby.
C.
From and after the date that one or the C. No additional surveillances are required.
three auxiliary feedwater pump systems is made or found to be inoperable for any reason, reactor operation is permissible only during the succeeding 7 days provided that during these 7 days the two remaining auxiliary feedwater pump systems are operable.
D.
If the conditions in 3.7. 2.C cannot be D.
Not applicable met the reactor shall be brought to hot shutdwon and below 350* F within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
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]PF SURVEILLANCE REQUIREENT LIE! TING CONDITION FOR OPERATION 3.7.3 Auxiliary feedwater supply system 4.7.3 Auxiliary feedwater supply system (Table 4. 7-3)
A.
The condensate storage tank (s) shall be operable with a minimum contained A.
Surveillance.and testing of the volume of 170,000 gallons of water auxiliary feedwater supply system shall lined up to each unit with its be as follows:
reactor above 350 OF, except as specified in 3.7.3.B and 3.7.3.C.
1)
The contained water volume shall be verified to be within its limits at ~
least every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
2)
The manual valves for the lined-up tank (s) for each unit shall be verified locked open each month.
B.
From and after the time that the B.
When the service water system is the conditions in 3.7.3.A cannot be required supply system, the system met, continued reactor operation is per~
shall be demonstrated operable at least, missib,le provided that at least ons of daily by stroking the power-operated the two following criteria are met:
service water. supply valves to at'least two operable auxiliary feedwater pumps 1)
Within the next four hours the from the control room.
Performance will conditions of 3.7.3.A are restored, be acceptable if valve motion is or indicated upon actuation.
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2)
Within the next four hours the operability of the service water system as a backup supply to the CT' auxiliary feedwater pumps is
((j demonstrated, and that the conden-sate storage tank's (s) are. restored to operable status within the next ta on 7 days.
CD C.
If the conditions in 3.7.3.A.and C.
Not applicable
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3.7.3.B cannot be met the reactor (s) shall be brought to hot shutdown and below 350 0F within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
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Component Name Component Number Auxiliary Feedwater Pump-1A (2A)
FWOO4 (turbine driven)
Auxiliary Feedwater Pump-1B (2B)
FW005 (motor driven)
Auxiliary Feedwater Pump-1C (2C)
FWOO6 (motor driven)
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O Auxiliary Feedwater Pump Sy:: tem TABLE 4.7-2 161
Component Name Component Number Condensate Storage Tank SC001 Auxiliary Feedwater Pump -1A (2A)
MOV 4WO102 service water supply valve Auxiliary Feedwater Pump -1B (2B)
MOV -SW0101 service water supply valve Auxiliary Feedwater Pump - 1B (2B)
MOV 4WO104 service water supply valve Auxiliary Feedwater Pump - 1C (2C)
MOV-SWQ103 service water supply valve Auxiliary Feedwater Pump - 1C (2C)
MOV-SWO105 service water supply valve Ch N
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Auxiliary Feedwater Supply System TABLE 4.7-3 161a
Bases:
3.7 The twenty code steam safety valves During 3 loop operation, at least two safety per unit have a total combined capa-valves are retained in service for any unlikely city to relieve the total steam pressurization on the non-operating steam flow of one unit.
These valves generator, assure code overpressure protection is provided (1).
In the event Although not required for safe operation of the that one or more of the safety valves unit, the four atmospheric steam relief valves are inoperable, the loop steam flows per unit, provide additional decay heat removal are restricted to the maximum re-capability.
These valves, which are air or lieving capacity of the most-restric-electric motor operated, are also manually tive operating loop.
This is controllable from the control room, and are accomplished by reduction of the installed to prevent unnecessary operation of Power.Ragge Neutron Flux High Set-the steam generator safety valves.
(1) point Trip such that reactor power is limited to be less than the thermal The auxiliary feedwater pump systems provide a power required to produce steam flow very reliable source of flow to the steam gener-in excess of the relieving capacity ators for decay heat removal.
Either the steam of the most restrictive loop.
The driven auxiliary feedwater pump or one of the reactor trip setpoints are derived two motor driven auxiliary feedwater pumps can on the following basls:
supply the required flow of a unit.
(2)
For 4 loop operation:
Suction to the auxiliary feedwater pumps is TSVC-ISVC SP =
x 109%
profided by the condensate sterage tank or, TSVC as a backup, the service water system.
A minimum of 170,000 gallons are required to For 3 loop operation:
provide for 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> at hot standby followed TSVC-ISVC by 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> cooldown at 50'F per hour with SP =
x 75%
steam discharge to the atmosphere concurrent TSVC with total loss of offcite power.
(3)
This is sufficient to reduce the reactor coolant where:
system temperature to below 350*F when the Residual Heat Removal System may be placed SP = Reactor Trip Setpoint in operation, q
TSVC = Total safety valve relieving y
capacity per steam generator C3 ISVC= Inoperable safety valve
(.1)
FSAR, Section 10.3 CD relieving capacity per steam generator.
(2)
PSAR, Section 14.1.9 LN (3)
FSAR, Section 6.7.2 162
Manually operated valves are available which will allow the condensate storage tanks to be cross connected.
Therefore, water for the auxiliary feedwater pumps may be supplied from either tank.
However, when operated in a cross-connected mode, the supply lines will be lined up such that the minimum of 170,000 gallons of water will be drawn upon immediately without additional actions.
When a condensate storage tank is shared 0
by both units operating above 350 F,
a minimum volume of 340,000 gallons shall be available.
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Bases:
>i, i.i 4.7; The testing of at least two safety:
! valves of each orifice size assures that.
a representative sample of valves is ii.i tested at each refueling.
The testing-interval assures the availability of the safety valves and of the auxiliary feedwater lpumpsystem.
The four hour delay in the surveillance and testing of the turbine driven auxiliary feed-i l water pump until the reactor has reached the i hot standby condition is to prevent unnecessary cooldown of the reactor coolant system during periods when the reactor is not available as a heat source.
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