ML20037D153
| ML20037D153 | |
| Person / Time | |
|---|---|
| Site: | Crane  |
| Issue date: | 05/18/1981 |
| From: | METROPOLITAN EDISON CO. |
| To: | |
| Shared Package | |
| ML20037D145 | List: |
| References | |
| NUDOCS 8105220060 | |
| Download: ML20037D153 (32) | |
Text
l 3.1.3 MINIMUd CONDITIONS FOR CRIfICALITY Applicability Applies to reactor coolant system conditions required prior to criticality.
Objective a.
To limit the magnitude of any power excursions resulting from reactivity insertion due to moderator pressure and moderator temperature coeffi-cients.
b.
fo assure that the reactor coolant system will not go solid in the event of a rod withdrawal or startup accident.
To assure sufficient pressurizer heater capacity to maintain natural c.
circulation conditions during a loss of offsite power.
Specification 3.1.3.1 The reactor coolant temperature shall be above 5250F except for portions of low power physics testing when the requiements of Specification 3.1.9 shall apply.
3.1.3.2 Reactor coolant temperature shall be above DTT +10'F.
3.1.3.3 When the reactor coolant temperature is below the minimum tem-perature specified in 3.1.3.1 above, except for portions of low power physics testing when the requirements of Specification 3.1.9 shall apply, the reactor shall be suberitical by an amount equal to or greater than the calculated reactivity in-sertion due to depressurization.
3.1.3.4 Pressurizer 3.1.3.4.1 The reactor shall be maintained suberitical by at least one percent Ak/k until a steam bubole is formed and an indicated water level between 80 and 385 inches is established in the pressurizer.
(a)
With the pressurizer level outside the required band, be in at least HOT SHUTDOWN with the reactor trip breakers open within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
3.1.3.4.2 A minimum of 107 kw of pressurizer heaters, from each of two pressurizer heater groups shall be OPERABLE.
Each OPER\\BLE 107 kw of pressurizer heaters shall be capable of receiving power from a 480 volt ES bus via the established manaul transfer scheme.
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f (a) With the pressurizer inoperable due to one (1) inoperable '
emergency power supply to the pressurizer heaters either restore the inoperable emergency power supply within 7 days or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in HOT SHUTDOWN within the following 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
(b) With the pressurizer inoperable due to two (2) inoperable emergency power supplies to the pressurizer heaters either restore the inoperable emergency power supplies within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or De in at least HOT SEANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in HOT SHUTDORN within the following 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
3.1.3.5
'afety rod groups shall be fully witndrawn prior to any other reduction in shutdown margin by deboration or regulating rod withdrawal during the approach to criticality with the follou-ing exceptions:
a.
Inoperable rod per 3.5.2.2.
b.
Physics Testing per 3.1.9.
Bhutdown margin may not he reduced below 1%,6,k/k per 3.5.2.1.
c.
d.
Exercising rods per 4.1.2.
Following safety rod withdrawal, the regulating rods shall be positioned within their position limits as defined by Specifi-ration 3.5.2.5 prior to deboration.
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Bases At the beginning of life of the initial fuel cycle, the moderator tempera-ture coefficient is expected to be slightly positive at operating tempera-tures with the operating configuration of control rods. - (1)' Calculations show that above 5250F the positive moderator coef ficient is acceptable.-
Since the moderator temperature coefficient at icwer temperatures will be -
less negative ~or more positive than at operating temperature, (2) startup and operation of the reactor when reactor _ coolant temperature is less than 525 F is prohibited except where necessary for low power physics tests.
The potential reactivity insertion due to the moderator pressure coefficient (2) that could result from depressurizing the coolant from 2100 psia to saturation pressure of 900 psia is approximately 0.1 percent Ak/k.
During physics tests, special operating precautions will be taken.
In addi-tion, the strong negative Doppler coefficient (1) and the small integrated a k/k would limit the magnitude of a power excursion resulting from a reduc-tion of moderator density.
The requirement that the reactor is not to be made critical below DTT+ 10 F provides increased assurances that the proper relationship between primary coolant pressure and temperatures will be maintained relative to the NDTT of the primary coola.L. system.
Heatup to this temperature will be accomplished by operating the reactor coolant pumps.
If the shutdown margin required by Specification 3.5.2 is maintained, there is no possibility of an accident.t1 criticality as a result of a decrease of coolant pressure.
i The availability of at least 107 kw in pressurizer heater capability is suf fic ient to maintain primary system pressure assuming normal system heat losses. Emergency power to aeater groups 8 or 9, supplied via a manual transfer scheme, assures redundant capabilipy upon loss of offsite power.
j The requirements that the safety rod grour.s be fully withdrawn before criti-cality ensures shutdoun capability during startup. This does not p rohibit rod latch confirmation, i.e., withdrawal by group to a maximum of 3 inches withdrawn of all seven groups prior to safety rod withdrawal.
The requirements for regulating rods being within their red position limits ensurec that the shutdown margin and ejected rod criteria at hot zero power are not violated.
REFE RENCES (1)
FSAR,-Section 3.
l.
(2)
FSAR, Section 3.2.2.1.
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3.1.12 Pressurizer Power Operated Relief Valve (PORV) and Block Valve Applicability Applies to the settings, and conditions for isolation-of the PORV.
Objective j To prevent the possibility of inadvertently overpressurizing or ldepressurimingtheReactorCoolantSystem.
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I Specification 3.1.12.1 The PORV shall not be taken out of service, nor shall it be isolated from the system (except that the PORV may be iso-lated to limit leakage to within the limits of specification 3.1.6) unless one of the following is in effect:
High Pressure Injection Pump breakers are racked out or a.
MU-V16A/B/C/D and MU-V217 are closed.
b.
Head of the Reaccor Vessel is removed.
c.
T is above 320*F.
ave 3.1.12.2 The PORV settings shall be as follows, within the tolerances of f; 25 psi and f; 12*F:
Above 275'F - 2450 psig i
Below 275'F - 485 psig 3.1.12.3 If the reactor vessel head is installed and T is <2 75*F, ave High Pressure Injection Pump breakers shall not be racked in unless:
MU-V16 A/B/C/D and MU-Vll7 are closed, and a.
b.
Pressurizer level is f 220 inches.
3.1.12.4 PORV and dLock Valve lhe PORV and the associated block valve shall be OPERABLE during HOT STANDBY, STARE UP, AND POWER OPERATION:
With the PORV ' inoperable, within I hour either restore a.
the PORV to OPERABLE status or close the associated block valve and remove power from the block valve; otherwise, be in at lea ;. : HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
b.
With the PORV block valve inoperable, within I hour either restore the PORV block valve to OPERAdLE status or close the PORV (verify closed) and remove power from the PORV.
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Bases If the PORV is removed from service, sufficient measures are incorporated to p re vent overpressurization by either elimiaating tne high pressure sources
';r flovpaths or assuring that the RCS is open to atmosphere.
In order to prevent exceeding leakage rates specified in T.S. 3.1.6. the PORV may be isolated.
The PORV setpoints are specified with tolerances assumed in the bases for Technical Specification 3.1.2.
With RCS temperatures less than 275'F and the makeup pumps running, the high pressure injection valves are closed and pressurizer level is maintained less than 220 inches to prevent overpressurization in the event of any single failure.
Both the PORV and the PORV block valve should be operable during the 110T STANDBY, STARTUP, and POWER OPERATION.
If either the PORV or the PORV block valve are inoperable the PORV discharge line should be isolated to prevent potential uncontrolled RCS depressurization.
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3.4 DECAY HEAT REMOVAL - TURBINE CYCLE Applicability Applies to the operating status of equipment that functions to remove decay heat, utilizing the secondary side of the Steam Generators.
Objective To define the conditions necessary to assure immediate availability of the Emergency Feedwater (EFW) System and Main Steam Safety Valves.-
_ Specification 3.4.1 With the deactor Coolant System temperature greater than 250*F, three l
independent EFW pumps and associated flow ' paths
- shall be OPERABLE with:
1 a.
Two EFW pumps, each capable of being powered from an OPERABLE emergency bus, and one EFW pump capable of being powered from an OPERABLE steam supply system.
b.
With one pump or flow path
- inoperable, restore the inoperable pump or flow path to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in COLD SdUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. With more than one EFW pump or flow path
- inoperable, restore the inoperable pumps or flow paths
- to operable status or be suberitical within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, in at least H0f SHUTDOWN within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, and in COLD SHUTDOWN within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
c.
Four of six turbine bypass valves are OPERABLE.
d The condensate storage tanks (CSTl shall be OPERABLE sith a minimum of 150,000 gallons of condensate available in each CST. With a CST inoperable, restore the CST to operability within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in at least HOT SHUTDOWN within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, and COLD SHUTD04N within the next 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
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- For the purpose of this requirement, an OPERABLE flow path shall mean an unobstructed path from the water source to the pump and f rom the pump I'
to a steam generator.
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3.4.3 With the reactor coolant system temperature greater than 250*F, all eighteen (18) main steam safety valves shall be operable _or, if any are not operable, the maximum overpower trip setpoint (see Table 2.3-1) shall be reset as follows:
Maximum Number of Maximum Overpower Safety Valves Disabled on frip Setpoint
'Any Steam Generator
(%'of Rated Power) 1 92.4 2
79.4 3
66.3 With more than 3 main steam safety valves inoperable, restore at least fif teen (15) main steam safety valves to operable status within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> or be in at least HOT STANDBY within the next o hours and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
Bases A reactor shutdown following power operation requires removal of core decay heat.
Normal decay heat removal is iy the steam generators with the steam dump to the condenser when RCS temperature is above 250*F and by the decay heat removal system below 250*F.
Core dacay heat can be continuously dissipated up to 15 percent of full power via the steam bypass to the con-denser as fewedwater in the steam generator is coverted to steam by heat absorption.
Normally, the capability to return feedwater flow to the steam generators is provided by the main feedwater system.
The main steam safety valves will be able to relieve to atmosphere the total steam flow if necessary.
If Main Steam Safety Valves are inoperable, the power level must be reduced, as stated in Technical Specification 3.4.3, such that the remaining safety valves can accomodate the decay heat.
In the unlikely event of complete loss of of f-site electrical power to the station, decay heat removal is by either the steam-driven emergency feed-water pump, or two half-sized motor-driven pumps.
Steam discharge is to the atmosphere via the main steam safety valves and controlled atmospheric relief valves, and in the case of the turbine driven pump, from the turbine exhaust. (1)
Both motor-driven pumps are required initially to remove decay heat with one eventually sufficing.
The minimum amount of water in the condensate storage tanks, containe6 in Technical Specification 3.4.2, will allow cooldown to 250*F with steam being discharged to the atmosphere. Af ter cooling to 250*F, the decay heat removal system is used to achieve further cooling.
An unlimited emergency feedwater supply is available from the river via either of the two motor-driven reactor building emergency cooling water pumps for an indefinite period to time.
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The requirements of Technical Specification 3.4.1 assure that before the reactor is heated to above' 2500 F, adequate auxiliary feedwater capacity is available. One turbine driven pump full capacity (920 gpm) and the two half-capacity motor-driven _ pumps (460 gpm each) are specified. How-ever, only.one half-capacity motor-driven pump is necessary to supply auxiliary feedwater. flow to the steam generators in the onset of a small break loss-of-coolant accident.
REFERENCES (1).
FSAR Section 10.2.1.3.
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reactor coolan?. temperature instrument channels, four reactor coolant flow instrument channels, four reactor coolant pressure instrument channels, four pressure-temperature instrument. channels, four flux-imbalance flow instru-ment channels, four power-number of pumps instrument channels, and four high reactor building pressure instrument channels.
The reactor trip, on loss of feedwater, may be bypassed below 7% reactor power.
The reactor trip, on turbine trip, may be bypassed below 20% reactor power.
The safety features actuation system must have two analog channels functioning correctly prior to startup. The anticipatory reactor trips on loss of feedwater pumps and turbine trip have been added to reduce the number of challenges to the safety valves and power operated relief valve but have not been credited in the safety analyses.
Operation at rated power is permited as long as the systems have at least the redundancy requirements of Column "B" (Taole 3.5-1).
This is in adree-ment with redundancy and single f ailure criteria of IEEE 279 as described in FSAR Section 7.
The re a re four reactor protection channels.
Normal trip logic is two out of four. Required trip logic.#or the power range instrumentation channels is two out of three. Minimum trip logic on other instrumentation channels is one out of two.
The four reactor protection channels were provided with key operated bypass switches to allow on-line testing or maintenance on only one channel at a
time during power operation. Each channel is provided alann and lights to indicate when that channel is by passed.
There will be one reactor protec-tion system bypass switch key permitted in the control room.
Each reactor protection channel key operated shutdown bypass switch is pro-vided with alarm and lights to indicate when the shutdown bypass switch is being used.
Power is nor,aally supplied to the control rod drive mechanisms from two separate parallel 460 volt sou rc e s.
Redundant trip devices are employed in each of these sources.
If any one of these trip devices fails in the un-tripped state on-line repairs to the failed device, when practical, will be made, and the remaining trip devices will be tested. Eight hours is ample time to test the remaining trip devices and in many cases make on-line re pa irs.
RE FERE NCE FSAR. Section 7.1 3-28
TABLE 3.5-1 Continued INSTRUMENTS OPERATING CONDITIONS (A)
(B)
Minimum Minimum (C)
Operable Degree of Operator Action if Conditions Functional Units Channels Redundancy of Column I. Cannot be Het Reactor Protection System 8.
Reactor coolant pressure a.
High reactor coolant pressure instrument channels 2
1 Maintain hot shutdown i
b.
Low reactor coolant pressure instrument y
channels 2
1 Maintain hot shutdown 9.
Power / number of pumps instrument channels 2
1 Maintain hot shutdown 10.
High reactor building pressura channels 2
1 Maintain hot shutdown Other Reactor Trips 1
1 1.
Loss of Feedwater 2,
1-Maintain less than 7% indicated power 2.
Turbine Trip 22 12 Maintain less than 20% indicated power 1.
Bypass of the feedwater pump trip signal may be placed it; effect when, indicated reactor power is less than 7%.
The bypass will be removed when reactor power is raised above 7%.
2.
The main turbine trip bypass may be placed in effect when irdicated power is less than 20%.
The bypass will be removed when the reactor power is increased above 20%.
TABLE 3.5-1 Ccntinued INSTRUtENTS OPERATING CONDITIONS Functional Unit (A)
(B)
(C)
Minimum Minimum Engineered Safety Features Operable Degree of Operator Action if Conditions Channels Redundancy of Column A cannot be met (a) 3.
Reactor Building Isolation and Reactor Building Cooling System a.
Reactor Building 4 psig Instrument Channel 2
1 Hot Shutdown b.
tbnual Pushbutton 2
1
' Hot Shutdown c.
RPS Trip 2
1 Hot Shutdown d.
Reactor Building 30 psig 2
1 Hot Shutdown u
O N
e.
RCS Pressure less than 1600 psig 2
1 Hot Shutdown (a)
If minimum conditions are not ag; within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, the unit shall then be placed in a cold shutdown condition.
(b) Also initiates Low Pressure injection.
4.
Reactor Building Spray System a.
Reactor Building 30 psig Instrument Channel 2 (b) 1 Hot Shutdown b.
Spray Pump Manual Switches (c) 2 1
Hot Shutdown (a) 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 then be placed in a cold shutdown condition.
(b) Two out of three switches in each actuation channel operable.
(c) Spray valves opened by manual pushbutton listed in item 3 above.
TABLE 3.5-1 Continued INSTRUltENTS OPERATING CONDITIONS Functional Unit (A)
(B)
(C)
Minimum Minimum Engineered Safety Features Operable Degree of Operator Action if Conditions Channels Redundancy of Column A cannot be met (a) 5.
Emergency Feedwater System Loss of Feedwater or RCP 2
1 Pump (all four) - Start Hot Shutdown.
Motor and Turbine Pumps Y
u Y
(a) 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 then be placed in a cold shutdown condition.
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3.5.3 ENGINEERED SAFEGUARDS PROTECTION SYSTEM ACTUATION SETPOINTS
. Applicability This specification applies to the engineered safeguards protection system
{
actuation setpoints.
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Objective To provide for automatic initiacion of the engineered safeguards protection system in the event of a breach of Reactor Coolant System integrity.
Specification j
3.5.3.1 The engineered safeguards protection system actuation setpoints and
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permissible bypasses shall be as follows:
Initiating ~ Signal Function Setpoint 4
High Reactor Building Reactor Building S ray f 30 psig P
Pressure (1)
Reactor Building isolation 4 30 psig i
I High-Pressure Injection 4 psig Low-Pressure Injection 4 psig Start Reactor Building Cooling & Reactor Building I
Isolation
$ 4 psig Low Reactor Coolant High Pressure Injection p 1600(2) and l
System Pressure 5 500(3) psig i
Low Presure Iniection y16bO(2)and i
2 500(3) psig Reactor Building Isolation 3 1600 psig (2) l (1) May be bypassed tor reactor building leak rate test.
(2) May be bypassed below 1750 psig and is automatically reinstated i
above 1750 psig.
(3) May be bypassed below 900 psig and is automatically reinstated above 900 psig.
Bases l
High Reactor Building Pressure The basis for the 30 psig and 4 psig setpoint for the high pressure signal l
is to establish a setting which we'sid be reached in adequate time in the l
event of a LOCA, cover a spectrum of break sizes and yet be f ar enough above normal operation maximum internal pressure to prevent spurious initiation.
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..,-.,,,--.-...m,
......,,,,,~,,...___,,..,.-,.,,,___..-,,,..__,,_.,.__,,..,,_,..,m,..
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Low'Reac tor ~ Coolant ' System' Pressure The basis-for the 1600 and 500 psig low reactor coolant pressure setpoint for high and low pressure injection initiation is to establish a value'which is high enough such that protection is 'provided for the entire spectrum of break sizes and is far enough below normal operating pressure to. prevent
' spurious initiation.
Bypass of HPI below 1750 psig, and LPI below 900 psig, prevents ECCS actuation during nor:ptl system cooldown.
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l 3.5.5 ACCIDENT MONITOR'ING INSTRUMENTATION.
Applicability Applies _to1the operability requirements for theLinstrument identi--
4 fied in Table 3.5-2 during START UP or-POWER OPERArION.
Objective To assure operabilityLof key ' instrumentation useful.in diagnosing situations which could lead to inadequate core cooling.-
Specification 3.5.5.1 The minimum number of channels or alternate indications, _ identified L for the instruments in Table 3.5-2, shall be OPERABLE. 'With the-number of instrumentation channels less than~the minimum required and the alternate indication-inoperable, restore the inoperable'.
channel (s) to OPERABLE status within seven -(7) ' days aor be in at.
least HOT SHUTDOWN. within the next twelve (12) hours.
Prior to start-up following a COLD SHUTDOWd, the minimum number of-channels shown in Table 3.5-2 shall be OPERABLE.
Bases The Saturation Margin Monitor provides a quick and reliable means for determination of saturation temperature margins. The hand -
calculation of' saturation pressure and saturation temperature; margins can be easily and quickly performed _since it'only requires knowledge of RC System loop temperatures'and system pressure,'and the use of steam tables; accordingly, hand calculation provides a suitable alternate indication for the Saturation Margin Monitor.
Discharge flow from the two (2) pressurizer code safety valves and the PORV is measured by differential pressure transmitters connect-ed across elbow taps downstream of each valve. A delta pressure indication from each, pressure _ transmitter 'is available in the control room to indicate' code safety or relief valve line flow. An alarm is also provided i.n the control room to indicate that discharge -
from a pressurizer code safety or relief valve is oc_ curing. In;addi -
tion, an acoustic monitor is provided to detect flow in the PORV discharge line. An alarm is provided in the control room for the acoustic monitor.
t t-j In the event that a delta pressure monitor or the acoustic monitor becomes inoperable, access to the containment would most likely be-required; however, a reactor shutdown to-allow containment access for this repair is nor justifiable due to the existence of alter-i nate means of detecting and monitoring code safety or relief valve j
discharge flow. The alternate means include discharge line thermo-couples sad Reactor Coolant Drain Tank indications.
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The Emergency Feedwater' System is provided with two channels of flow istrumentation on each of the two discharge lines. Local' flow' indication is also available for the emergency feedwater system.
Although the prtasurizer has multiple level indications,.the sepa-rate indications are selectable via a switch !or display on a sin-
~J gle display.
Pressurizer level, however, car: also be determined via the patch panel and the computer lod.
Although the instruments identified in Table 3.5-2 are significant in diagnosing situations which could lead to inadequate core cool-ing, loss of any one of the instruments oin Table 3.5-2 would not prevent continued, safe, reactor operation provided that the alter-nate indication is operable. : Loss of an instrunent and its alter-nate indication would degrade the reactor operators diagnostic capability and, thus, should be restored within seven (7) days.
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TABLE 3.5-2 4
ACCIDENT MONITORING INSTRUMENTS
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MINIMUM ALTERNATE FUNCTION INSTRUMENTS NUMBER OF CHANNELS NUMBER OF CHANNELS INDICATION.
1 Saturation Margin Mor:itor 1
1 2
Safety / Relief Valve Differ-1 per discharge line 1 per discharge line Discharge line thermo-ential Pressure Monitor couples or Reactor Coolant Drain Tank level and pressure i
3 PORV Acoustic Monitor 1
1 Discharge line iLenmo-couples or Reactor Coolant Drain Tank level and pressure 4
Emergency Feedwater Flow 2 per flow path 1 per flow path Steam Generator water 1
level and EFW pump I
discharge pressure 5
Pressurizer Level 1
1 Pressurizer Level (patch panel) or Computer Log If the Saturation Margin Monitor is inoperable, the operability requirement for the Saturation Margin Monitor is satisfied by-manual determination saturation temperature margins.
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3.19 Separation'of TMI-l'and'TMI-2 Applicability Applies to interconnections between TMI-l and IMI-2 which have the.
potential for transferring significant quantities of contamination between units.
Objective To control the transf er of radioactivity from THI-2 to TMI4 via system interties.
L...
Specification We The isolation devices for the system interconnections listed in 3.19.1 Table. 3.19-1 shall remain inplace unless written approval has been received f rom the NRC.
If approval for use of interconnections is
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l received, use shall proceed under preestablished procedures.
No additional TMI-1/TMI-2 interconnections, with the potential of 3.19.2 transf erring significant quantities of radioactivity, shall be created without prior NRC ap?roval.
Observed defeat of an isolation device, which separates a tieline 3.19.3 between Units 1 and 2, without prior NRC approval shall be reported '
to the NRC as a special report within 30 days.
s Bases f6 Interconnections exist between TMI-l and TMI-2 that have the potential for transferring contamination to TMI-l as a result of restoration activities at These interconnections should remain isolated unless approval for TMI-2.
their uae is received from the NRC.
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Table'3.19-1 TMI-1/TMI-2 Interconnections
-( l). Unit 2 Reactor Coolant Bleed Holdup Tank - Unit 1 Reactor Coolant Waste Evaporator (2) Unit 1' Miscellaneous Waste Evaporator - Unit 2 Evaporator Condensate Test Tanks (3) Unit 2 Neutralizer Tanks, Contaminated Drain Tanks, Reactor Coolant Bleed Holdup Tank 6, Auxiliary Building Sump Tanks and Miscellaneous Waste Holdap Tanks - Unit 1. Liquid Waste Disposal' System (4) Unit 1 Evaporator Concentrate - Unit 2 Evaporator Concentrate (5) Unit 1 Spent Ion Exchange Resin - Unit 2 Spent Ion Exchange Resin-8 e
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Channel subject only to "drif t" errors induced within the instrumentation itself can tolerate longer intervals between calibrations. Process system instrumentation errors induced by drif t can be expected to remain within acceptable tolerances if recalibration is performed at the intervals of each refueling period.
Substantial calibration shif ts within a channel (essentially a channel f ail-ure) will be revealed during routine checking and testing procedures.
Thus, minimum calibration frequencies set forth are considered acceptable.
Testing On-line testing of reactor protection channels is required once every four weeks on a rotational or perfectly staggered basis.
The rotation scheme is designed to reduce the probability of an undetected failure existing within the system and to minimize the likelihood of the same systematic test errors being introduced into each redundant channel.
The rotation schedule for the reactor protection channels is as follows:
Channels A, B, C, & D Before Startup, when shutdown greater than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> Channel A One Week Af ter Startup Channel B Two Weeks After Startup Channel C Three Weeks Af ter Startup Channel D Four Weeks After Startup The reactor protection system instrumentation test cycle is continued with one channel's instrumentation tested each week.
Upon detection of a failure that prevents trip action in a channel, the instrumentation associated with the protection parameter failure will be t~ sted in the remaining channels.
e If actuation of a safety channel occurs, assurance will be required that actuation was within the limiting safety system setting.
The protection channels coincidence logic and control rod drive trip breakers are trip tested every four weeks. The trip test checks all logic combinations and is to be performed on a rotational basis.
The logic and breakers of the four protection channels shall be trip tested prior to startup when the reactor has been shutdown for greater than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
Dis-covery of a f ailure that prevents trip action requires the testing of the instrumentation associated with the protection parameter failure in the remaining enannels.
4-2
4 t
For purposes :of surveillance, reactor trip'on' loss.of, feed' water and reactor trip on turbine trip are considered reactor protection system channels.
The equipment testing and system sampli'ng frequencies specified in-Table 4.1-2 and Table 4.1-3 :are cor idered adequate to maintain th'e equipment and -
systems in a safe operational status.
REFERENCE
' (1). FSAR, Section 7.1.2.3.4 4
e i
a i
1 I
i 4
A 4
4 F
i t
e f
4-2a 1
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.. - +--
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TABLE 4.1-1 (Continu:d)
INSTRU!ENTS OPERATING CONDITIONS l
CHANNEL DESCRIPTION CHECK TEST CALIBRATE REMAPXS 19.
Reactor Building Emergency Cooling and Isciation System Channels a.
Reactor Building S(l)
M(1)
R (1) When CONTAINMENT INTEGRITY is required 4 psig Channels b.
RCS Pressure 1600 psig S(l)
M(1)
NA (1) When RCS pressure >l750 psig c.
RPS Trip S(l)
M(1)
NA (1) When CONTAINMENT INTEGRITY is required l
d.
Reactor Building 30 psig S(l)
M(1)
R (1) When CONTAINMENT INTEGRITY is required i
20.
Reactor Building Spray NA Q
NA System Logic Channel l
21.
Reactor Building Spray System Analog Channels a.
Reactor Building NA M
R j
30 psig Channels 22.
Pressurizer Temperature S
NA R
Channels l
l 23.
Control Rod Absolute Position S(l)
NA R
(1) Check with Relative Position Indicator 24.
Control Rod Absolute Position S(l)
NA R
(1) Check with Absolute Position Indicator i
25.
Core Flooding Tanks a.
Presure Channels S(l)
NA-R (1) When Reactor Coolant system pressure is greater than 700 psig b.
Level Chanceis S(l)
NA' R
l 26.
Pressurizer Level Channels S
NA R
27.
Makeup Tank Level Channels D(1)
NA R
(1) When Makeup and Purification System is in operation
TABLE 4.1-1 (Continued)
CHANNEL DESCRIPTION CIIECK TEST CALIBRATE REMARKS j
38.
Steam Generator Water Level W
NA R
{
39.
Turbine Overspeed Trip NA R
NA t
40.
Sod.ium Thiosulfate Tank Level NA NA R
Indicator
,i 41.
Sodium Hydroxide Tank level NA NA R
Indicator 42.
Diesel Generator Protective NA NA R
Relaying f
43.
4 KV ES Bus Undervoltage Relays NA M(1)
R (1) Relay operation will be (Diesel Start) checked by local test push-4 buttons.
I z.
4 44.
Reactot Coolant Pressure S(1)
M R
(1) When reactor coolant system Dil Valve Interlock Bistable
.is pressurized above 300 psig' or Taves is greater than 200*F.
I j
45.
Loss of Feedwater Reactor Trip S(l)
M(1)
R (1) When reactor power exceeds 10% power i
i 46.
Turbine Trip / Reactor Trip S(l)
M(1)
R (1) When reactor power exceeds 20% power j
47.a Pressurizer Code Safety Valve and S(l)
R (1) When T is greater than ave
. [
PORV tailpipe flow monitors 525*F 1
~
47 b PORV - Acoustic / Flow NA M
R (1) When T is greater than ave 525'F.
l 48.
PORV Setpoint NA M(1)
R (1) When T is. greater than' ave
~
525*F.
Excluding valve operation.
I
TABLE 4.1-1 (Continued)
CHANNEL DESCRIPTION CllECK TEST CALIBRATE RDfARKS 49.
Saturation Margin Monitor S(l)
M(1)
R(1)
(1) When Tave is greater than 525 F.'
T 50.
Emergency Feedwater Flow NA M(1)
R (1) When ave is greater than Instrumentation 250 F.
51.
Emergency Feedwater Initiation T
a.
Loss of RCP's or Feedwater NA Q(1)(2)
R (1) When ave is greater than 250 F.
nc u es logic test only.
S - Each Shift T/W - Twice per week D - Daily B/M - Every 2 months R - Each Refueling Period NA - Not applicable W - Weekly Q - Quarterly
/ - ven w wee s M - Monthly P - Prior to each startup if not done previous week ~
T B'
TABLE 4.1-2 MINIMUM EQUIPMENT TEST FREQUENCY 1.
Control Rods-Rod drop times of all.
Each refueling shutdown full length rods 2.
Control Rod Movement of each rod Every, two weeks, when reac-Movement tor is critical ~
3.
Pressurizer Safety Setpoint*
50%~each refueling period.
Valves 4.
. Main Steam Safety Setpoint 25% each refueling period 5.
Refueling System Functional
. Start of each refueling Interlocks
. period 6.
Main Steam (See Section 4.8)
Isolation Valves 7.
Reactor Coolant Evaluate
. Daily, when reactor coolant System Leakage-system temperature'is greater than 525'F 8.
Deleted 9.
Spent Fuel Cooling Functional 6ach refueling period prior System to fuel handling 10.
Intake Pump House (a) Silt Accumulation-Each refuelind period Floor Visual inspection of (Elevation 262 Ft.
Intake Pump House, Floor 6 in.)
(b) Silt Accumulation Quarterly Measurement of Pump House Floor 11.
Pressurizer diock Functional **
Quarterly Valve (RC-V2)
The set point of the pressurizer code safety valves shall be in accordance with ASME Boiler and Pressurizer Vessel Code,Section III, Article 9, Winter, 1968.
Function shall be demonstrated by operating the valve through one complete cycle of full travel.
4-3
d Table 4.1-2 (Continued)
I t em Test Frequency 12 Isolation devices Visual Inspection Semiannually on Unit 1/ Unit 2 interconnections (a) Unit 2 Reactor Coolant Bleed Holdup Tank - Unit 1 Reactor Coolant Waste Evaporator l
(b) Unit 1 Miscellaneous Waste Evapo ra to r - Uni t 2 Eva po r-ator Condensate Tes t Tanks (c) Unit 2 Neutralizer Tanks, Con-tainment Drain Tanks, Reactor Cooland Bleed Holdup Tanks, Auxiliary Building Sump Tanks and Miscellaneous. Waste Holdup Tanks - Unit 1 Liquid W' ste a
Disposal System i
I Of (d) Unit 1 Evaporator Concentrate -
Unit 2 Evapo ra to r Co ncent ra td ""
(e) Unit 1 Spent Ion Exchange Resin -
Unit 2 Spent Ion Exhange Resin i
J J
4.5 EMERGENCY' LDOADING SEQUENCE AND' R)RER TRANSFER ~ ' EMERGENCY ~ CORE COOLING SYSTEM AND REACTOR BUILDING COOLING SYSTEM PERIODIC TESTING 4.5.1 EMERGENCY LO4 DING SEQUENCE Applicability Applies to periodic testing requirements for safety actuation systems.
Objective To verify that the emergency loading sequence and automatic power transfer is operable.
Specifications
- 4. 5.1.1 Sequence and Power' Transfer Test During each refueling interval, a test shall be conducted to demon-a.
strate that the emergency loading sequence and power transfer is operable.
b.
The test sill be considered satisf actory if the following pumps and fans have been successfully started and the following valves have completed their travel on preferred power and transferred to the emergency power as evidenced by the control board component operat-ing lights, and either the station computer or pressure / flow indi-cation.
- M. U. Pump
- D. H. Pump and D. H.
Injection Valves and D. H. Supply Valves
- d.
B. Cooling Pump
- R. B. Ventilators
- D. H. Closed Cycle Cooling Pump
- N. S. Closed Cycle Cooling Pump;
- D. H. River Cooling Pump
- N. S. River Coolind Pump
- D.
H. and N. S. Pump Area Cooling Fan
- Screen House Area Cooling Fan
- Spray Pump.
(Initiated in coincidence with a 2 out of 3 d. 3.
30 psi Pressure Test Signal.)
- Motor Driven Emergency Feedwater Pump 4.5.1.2 Sequence rest a.
At intervals not to exceed 3 months, a test shall be conducted to demonstrate that the emergency loading sequence is operable, this test shall be perfonned on either preferred power or emerdency power.
4-39
c.
Each time data are recorded, new data shall be compared with old to detect signs of abuse or deterioration.
d.
The battery will be subjected to a load test at a frequency not to exceed refueling periods. The battery voltage as a function of time will be monitored to establish that the battery performs as expected during this load test.
Bases The tests specified.are designed to demonstrate that one diesel generator will provide power for operation of. safeguards equipment.
They also assure that the emergency generator control system and the control systems for the safeguards equipment will function automatically in the event of a loss of normal a-c station service power or upon the receipt of an engineered safe-guards Actuation Signal.
The automatic tripping of manually transferr-ad loads, on an Engineered Safeguards Actuation Signal, protects the diesel generators from a potential over-load condition. The testing frequency specified is intended to identify and permit correction of any mechanical or electrical deficiency before it can result in a system failure.
The fuel oil supply, starting circuits, and controls are continuously monitored and any faults are alarmed and indicated. An abnormal condition is these sys-tems would be signaled without having to place the diesel generators on test.
Precipitous failure of the station battery is extremely unlikely.
The sur-veillance specified is that which has been demonstrated over the years to provide an indication of a cell becoming unserviceable long before it fails.
The PORV has a remotely operated block valve to provide a positive shutoff capability should the relief valve become inoperable.
The electrical power for both the relief valve and the block valves is supplied from an ESF power source to ensure the abiity to seal this possible RCS leakage path.
The requirement that a minimum of 107 kw of pressurizer heaters and their associated controls be capable of being supplied electrical power from an emergency bus provides assurance that these heaters can be energized during a loss of of fsite power condition to maintain natural circulation.
s REFERE NCE (1)
FSAR, Section 8.2 l
l l
l 4-47
4.9 EMERGENCT FEEDRATER SYSTEM PERIODIC TESTING Applicability Applies to the periodic testing of the turbine driven and two motor-driven Emergency Feedwater pumps, associated actuation signal, and valves.
Objective To verify that the Emergency Feedwater._(EFW) System is capable of performing its design function.
Specification
-4.9.1 TEST
- 4. 9.1.1 Whenever the Reactor Coolant System temperature is greater than 250*F, the EFW pumps shall be tested in the recirculation mode in accordance with the requirements and acceptance criteria of ASME Section XI Article IWP-3210. The test frequency shall be at least every 31 days
! of plant operation at Reactor Coolant Temperature above 250*F.
4.9.1.2 During testind of the EFW System when the reactor is in STARTUP or B3WER OPERATION, if one steam generator flow path is made inoper-able, a dedicated qualified individual who is in communication with the control room shall be continuously stationed at the EFW manual valves. On instruction from the Control Room Operator, the individual shall realign the valves from the test mode to their operational alignment.
4.9.1.3 At least once per 31 days each valve listed in Table 4.9-1 shall be verified to be in the status specified in Table 4.9-1, when required to be operable.
4.9.1.4 On a quarterly basis, verify that:
(a).
each EFW pump starting logic actuates upon receipt of an EFW test signal, and (b).
that the manual control (HIC-849/850) valve station functions properly.
4.9.1.5 On a quarterly busis, EFV-30A and 30B shall be checked for proper operation by cycling each valve over its full stroke.
4.9.1.6 Irior to start-up, following a cold shutdown f.ir refueling, conduct a test to demonstrate that the motor driven E 7W pumps can pump water from the condensate tanks to the Steam Ganerators.
- For the purpose of this requirement, an OPERABLE flow path shall mean an unobstracted path from the water source to the pump and from the pump to a Steant Generator.
4-52 c
v r,
4.9.2 ACCEPIANCE CRITERIA These tests shall be considered satisfactory if control board indication and visual observation of the equipment demonstrates that all components have operated properly.
Bases The 31 day testing frequency will be sufficient to verify that the turbine driven and two motor-driven EFW pumps are operaole and that.the associated valves are in the correct alignment. A5ME SEction XI Article IWP-3210 specifies requirements and acceptance standards for the testing of nuclear safety related pv2ps.
Compliance with the normal acceptance criteria assures that tha 4FW pumps are operating as expected. The test frequency of 31 days (nominal) has been demonstrated by the B&W Emergency Feedwater Reliability Study to assure an appropriate level of reliability.
If testing indicates that the flow and/or pump head for a particular pump is not within the normal acceptance standard an evaluation of the pump performance shall be completed within 96 hours0.00111 days <br />0.0267 hours <br />1.587302e-4 weeks <br />3.6528e-5 months <br /> or the pump declared inoperable. For the case of the EFW System, the system shall be considered operable if under the worst case single pump f aiture, a minimum of 500 gpm can be delivered when steam generator pressure is 1050 psig and one steam generation is isolated.
A flow of 500 gpm, at 1050 psig head, ensures that sufficient flow can be delivered to either Steam Generator.
The surveillance requirements ensure that the overall EFW System 1 functional capability is maintained.
4-52a
. Table 4.9-1 Status of EFW Valves-Valve ~No.
Status ~
C0-V-10A Open
- CO-V-10B Open EF-V-1A
.Open EF-V-1B1
.Open EF-V-2A Open EF-V-2B Open 4
MSV-2A Open MSV-2B Open
' EF-V4 Locked Closed l
EF-V5 Locked closed' i
EF-V6
~ Locked Open EF-V10A Lucked Open EF-V10B Locked Open l
EF-V-16A Locked Open 4
j' EF-V-16B Locked Open i
EF-V-20A
. Locked Open EF-V-20B Locked open e
EF-V-22 Locked Open r
4 4
w==e k
4 4
4 e
4-52b i'
.,.. -, _,. -... _,. -.. _. _ ~ _.,, _ _ _ _, _.
EMERGENCY POWER SYSTEM PERIODIC TESTS Applicability:
Applies to periodic testing and surveillance requirement of the emergency power system.
Objective:
To verify that the emergency power system will respond promptly and perperly when required.
Specification:
The following tests and surveillance shall be performed as stated:
4.6.1 Diesel Generators a.
Manually-initiated start of the diesel generator, followed by manual synchronization with other power sources and assumption of load by the diesel generator up to the nameplate rating (3000 kw).
This test will be conducted every month on each diesel generator. Normal plant opera-tion will not be affected.
b.
Automatically start and loading the emergency diesel generator in accordance with specification 4.5.1.1 b/c including the following:
(1). Verify that the diesel ' generator starts from ambient condition upon receipt of the ES signal and is ready to load in 6. 10 seconds.
(2.).
Verify that the diesel block loads upon simulated loss of off-site power in f[.30 seconds.
4 (3).
The diesel operates with the permanently connected and auto-connected load for ji 5 minutes.
~
(4).
The diesel engine does not trip when the generator breaker is opened while carrying emergency loads.
(5).
The diesel generator block loads and operates forj! 5 minutes upon reclosure of the diesel generator breaker.
(6).
That the pressurizer heaters breaker on the emergency bus can-not be closed until the safeguards signal is bypassed and can be closed following bypass.
(7).
That following input of the Engineered Safeguard Signal, it shall be verified that the circuit breakers, supplying power to the manually transferred loads for pressurizer heater Groups 8 and 9, have been tripped.
c.
Each diesel generator shall be given an inspection at least annually in accordance with the manufactorer's recommendations for this class of standby service.
4.6.2 Station Batteries a.
The voltage, specific gravity, and liquid level of each cell will be measured and recorded monthly.
b.
The voltage and specific gravity of a pilot cell will be measured and recorded monthly.
4-46
__