ML20212D134
| ML20212D134 | |
| Person / Time | |
|---|---|
| Site: | Rancho Seco |
| Issue date: | 12/05/1986 |
| From: | SACRAMENTO MUNICIPAL UTILITY DISTRICT |
| To: | |
| Shared Package | |
| ML20212D124 | List: |
| References | |
| TAC-64359, NUDOCS 8612310298 | |
| Download: ML20212D134 (24) | |
Text
RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Definitions 140 F.
Pressure is defined by Specification 3.1.2.
A refueling shutdown refers to a shutdown to replace or rearrange all or a portion of the fuel
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assemblies and/or control rods.
1.2.7 Refueling Operation An operation involving a change in core geometry by manipulation of fuel, or control rods when the reactor vessel head is removed.
1.2.8 Refueling Interval
- Time between normal refuelings of the reactor, not to exceed 24 nonths for the l
first refueling and 18 months thereafter without prior approval. of the NRC.
1.2.9 Startup The reactor shall be considered in the startup mode when the shutdown margin is reduced with the intent of going critical.
1.2.10 Remain Critical A technical specification that requires that the reactor shall not remain critical shall mean that an uninterrupted nomal hot shutdown procedure will 152x be completed within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> unless otherwise specified.
1.2.11 Tavg is defined as the arithmetic average of the
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At operating conditions TAVG coolant temperatures in the hot and cold legs of the loop with the greater number of reactor coolant pumps operating, if sdch a distinction of loops can be made.
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1.2.12 Heatup - Cooldown Mode The heatup-cooldown mode is the range of reactor coolant temperature greater than 200 F and less than 525 F.
l 1.3 OPERABLE A component or system is operable when it is capable of performing its intended function within the required range. The component or system shall be considered to have this capability when:
(1) it satisfies the limiting conditions for operation defined in Specification 3, (2) it has been tested periodically in accordance with Specification 4, and has met its perfomance requirements, (3) the system has available its normal and energency sources of power, and (4) its required auxiliaries are capable of performing their intended function. When a system or component is detemined to be inoperable solely because its normal power source isinoperable or its emergency power source is inoperable, it may be considered OPERABLE for the purpose of satisfying the requirements of its applicable Limiting Condition for Operation provided its redundant system or component is OPERABLE with an OPERABLE normal and emergency power source.
- See page 1-2b Proposed Amendment No. 152 1-2 8612310298 861205 PDR ADOCK 05000312 P
RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Definitions 1.15 0FFSITE DOSE CALCULATION MANUAL (ODCM)
An DFFSITE DOSE CALCULATION MANUAL (0DCM) shall be a manual containing the methodology and parameters to be used in the calculation of offsite dose due to radiioactive gaseous and liquid effluents and in the calculation of gaseous and liquid effluent monitoring instrumentation alarm / trip setpoints and specific details of the environmental radiological monitoring program.
1.16 RESTRICTED AREA That portion of the site property, the access to which is controlled by security fencing, equipment and personnel.
1.17 SITE B0UNDARY The boundary of the SMUD owned property.
1.18 DOSE EQUIVALENT I-131 The DOSE EQUIVALENT I-131 shall be that concentration of I-131 (microcurie / gram) which alone would produce the same thyroid dose as the quantity and isotopic mixture of I-131, I-132, I-133, I-134 and 1-135 actually present. The thyroid dose conversion factors used for this calculation shall be those listed in Table III of TID-14844, "Calculatio.n of Distance Factors for Power and Test Reactor Sites".
1.19 MEMBER (S) 0F THE PUBLIC MEMBER (S) 0F THE PUBLIC shall include all individuals who by virtue of their occupational status have no formal association with the plant. This category shall include non-employees of the licensee who are permitted to use portions of the site for recreational, occupational, or other purposes not associated with plant functions.
This category shall not include non-employees such as vending machine servicemen or postmen who, as part of their formal job function, occasionally enter an area that is controlled by the licensee for purposes of protection of individuals from exposure to radiation and radioactive materials.
152> 1.20 VECTOR LOGIC A set of circuitry in each channel of the EFIC system which once AFW has been initiated determines whether AFW to a steam generator should be allowed or terminated and the signal output for each EFIC channel to the AFW valves associated with that channel.
Proposed Amendment No.152 1-7
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RANCHO SECO UNIT 1 c
TECHNICAL SPECIFICATIONSLimiting Conditions for Operation 3.
LIMITING CONDITIONS FOR OPERATION 3.1 REACTOR COOLANT SYSTEM
_ Applicability Applies to the operating status of the reactor coolant system.
Objective To specify those limiting conditions for operation of the reactor coolant system which must be met to ensure safe reactor operations.
3.1.1 OPERATIONAL COMPONENTS Specification 3.1.1.1 Reactor Coolant Pumps A.
Pump combinations permissible for given power levels shall be as shown in specification table 2.3-1.
.B.
The boron concentration in the reactor coolant system shall not be reduced unless at least one reactor coolant pump or one decay heat removal pump is circulating reactor coolant.
C.
Operation at power with two pumps shall be limited to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> in any 30 day period.
3.1.1.2 Steam Generators 152>
A.
Two steam generators shall be operable whenever the reactor coolant average temperature is above 280 F, except as described in 3.1.1.2.B.
B.
With one or more steam generator (s) inoperable due to excessive leakage per 3.1.6.9, bring the reactor to cold shutdown conditions within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />.
C.
With one or more steam generator (s) inoperable due to steam generator defective tube (s), restore the inoperable generator (s) to operable status prior to increasing reactor coolant average temperature above 200*F.
j 3.1.1.3 Pressurizer Safety Valves A.
The reactor shall not remain critical unless both Pressurizer Coolant System code safety valves are operable.
B.
When the reactor is subcritical, at lqast one Pressurizer code safety valve shall be operable if all reactor coolant system openings are closed, except for hydrostatic tests in accordance with ASME Boiler and Pressure Vessel Code, Section l
l III.
Proposed Amendment No. 152 l
3-1
RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS j
t Limiting Conditions for Operation 3.1.1.4 Pressurizer Electromatic Relief Yalve A.
The nominal setpoint of the pressurizer electromatic relief valve shall be 2450 psig
- 10 psig except when required for cold overpressure protection.
3.1.1.5 Decay Haat Removal A.
At least two of the coolant loops listed below shall be operable when the coolant average temperature is below 280*F.
except during fuel loading and refueling.
1.
Reactor Coolant Loop (A) and its associated steam generatar and at least one associated reactor coolant pump, 2.
Reactor Coolant Loop (B) and its associated steam gene,rator and at least one associated reactor coolant pump, 3.
Decay Heat Removal Loop ( A) 4.
Decay Heat Removal Loop (B)
With less than the above required coolant loops OPERABLE, immediately initiate corrective action to return the required coolant loops,to OPERABLE status as soon as possible; be in COLD SHUTDOWN within 20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br />.
3.1.1.6 Reactor Coolant System High Point Vents A.
The vent path on Loop A and vent path on Loop B shall be operable and closed during power operation.
B.
The vent path on the pressurizer shall be operable and closed during power operation.
?,.
With one of the above reactor coolant system vent paths inoperable, STARTUP and/or POWER OPERATION may continue provided the inoperable vent path is maintained closed with power removed from the valve actuator of all the valves in the inoperable vent path; restore the inoperable vent path to OPERABLE status within 30 days.
If the status is not restored to operable in 30 days, be in HOT STANDBY within 12 nours 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 />.
D.
With two or more of the above reactor coolant system vent paths inoperable; maintain the inoperable vent paths closed with power removed from the valve actuators of all the valves in the inoperable vent paths, and restore at least (two) of the vent paths to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.
If the status is not restored to operable in 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, be in HOT STANDBY within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTOOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
Proposed Amendment No.152 3-2
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i RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Limiting Conditions for Operation Bases A reactor coolant pump or decay heat removal pump is required to be in operation before the boron concentration is reduced by dilution with makeup water. Either pump will provide mixing which will prevent sudden positive reactivity changes caused by dilute coolant reaching the reactor. One decay heat removal pump will circulate the equivalent of the reactor coolant system volume in one half hour or less.
(1)
The decay heat removal system suction piping is designed for 300 F and 300 psig; thus, the system can remove decay heat when the reactor coolant system is below this temperature.
(2) (3)
One Pressurizer code safety valve is capable of preventing overpressurization when the reactor is not critical since its relieving capacity is greater than that required by the sum of the available heat sources which are pump energy, pressurizer heaters, and reactor decay heat.
(4) Both Pressurizer code safety valves are required to be in service prior to criticality to conform to the system design relief capabilities. The code safety valves prevent overpressure for rod withdrawal accidents.
(5) The Pressurizer code safety valve lift set point shall be set at 2500 psig
- 1 percent allowance for error and each valve shall be capable of relieving 345,000 lb/hr of saturated steam at a pressure not greater than 3 percent above the set pressure.
The electromatic relief valve setpoint was established to prevent operation of the Safety Valves during transients.
Two pump operation is limited until further ECCS analysis is performed.
152> When the reactor is not critical but TAV is above 280* F, one steam generator provides sufficient heat removal capability for removing decay heat. However, single failure considerations require that both steam generators be operable.
When TAV is below 280*F, a single reactor coolant loop or DHR loop provides sufficient heat removal capability for removing decay heat; but single failure considerations require at least two loops be OPERABLE. Thus, if the reactor coolant loops are not OPERABLE, this specification requires two DHR loops to be OPERABLE.
The purpose of the high point vents is to vent noncondensible gases from the RCS which may inhibit core cooling during natural circulation.
In compliance with 10CFR50 Appendix R the power to all the valve actuators in the vent path has been removed.
REFERENCES (1)
FSAR Tables 9.5-2, 4.2-1, 4.2-2, 4.2-4, 4.2-5, 4.2-6 (2)
FSAR paragraph 9.5.2.2 and 10.2.2 (3)
FSAR paragraph 4.2.5 (4)
FSAR paragraph 4.3.8.4 and 4.2.4 (5)
FSAR paragraph 4.3.6 and 14.1.2.2.3 Proposed Amendment No. 152 3-2a L
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RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS
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Limiting Conditions for Operation 3.4 STEAM AND POWER CONVERSION SYSTEM Applicability Applies to the operability of the turbine cycle during normal operation and for the removal of decay heat.
Objective To specify minimum conditions of the turbine cycle equipment necessary to assure the required steam relief capacity during normal operation and the capability to remove decay heat from the reactor core.
Specification 152>
3.4.1 The reactor shall not be brought or remain above 280F with irradiated fuel in the pressure vessel unless the following conditions are met:
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A.
Capability to remove decay heat by use of two steam generators as specified in 3.1.1.2.
B.
One turbine bypass valve or one atmospheric dump valve per steam generator shall be operable.
C.
A minimum of 250,000 gallons of water shall be available in the condensate storage tank.
D.
Two main steam system safety valves are operable per steam generator.
E.
Both auxiliary feedwater trains (i.e., pumps and their flow paths) are operable.
F.
Both trains of main feedwater isolation on each main feedwater line are operable.
With less than the above required components operable, be on decay 4
heat cooling within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.
Proposed Amendment No.152 3-23
RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Limiting Conditions for Operation 152>
3.4.2 The reactor shall not be brought or remain critical unless the following conditions are met:
A.
Capability to remove decay heat by use of two steam generators as specified in 3.1.1.2.
B.
One turbine bypass valve or one atmospheric dump valve per steam generator shall be operable except that: (1) with one less than the minimum number of valves, restore the inoperable valve within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in hot shutdown within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and on decay heat cooling within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />; (2) with two less than the minimum number of valves, restore at least one inoperable valve within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> or be in 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 on decay heat cooling within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> C.
A minimum of 250,000 gallons of water shall be available in the condensate storage tank except that with less than the minimum volume, restore the minimum volume within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> or be in 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 on decay heat cooling within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
D.
Seventeen of the eighteen main steam safety valves are operable except that with less than the miniraum number of valves, restore the inoperable valve (s) within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> or be in 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 on decay heat cooling within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
E.
Four turbine throttle stop valves are operable except that with less than the minimum number of valves, restore the inoperable valve (s) within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> or be in 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 on decay heat cooling within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
F.
Both auxiliary feedwater trains (i.e., pump and their flow path) are operable except that:
(1) With one auxiliary feedwater train inoperable, restore the train 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 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 on decay heat cooling within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
(2) With both auxiliary feedwater trains inoperable, the reactor shall be made subcritical within four hours and the reactor shall be on decay heat cooling within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
Proposed Amendment No. 152 3-23a
t RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Limiting Conditions for Operation 152>
3.4.2 G.
Both trains of main feedwater isolation on each main feedwater line are operable except that:
(1) With one main feedwater isolation train inoperable, restore the train 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 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 on decay heat cooling within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
(2) With both main feedwater isolation trains inoperable, the reactor shall be made subcritical within four hours and the reactor shall be on decay heat cooling within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
Bases The feedwater system and the turbine bypass system are normally used for i
152>
decay heat removal and cooldown above 280 F.
Main feedwater is supplied by operation of a condensate pump and main feedwater pump.
If neither main feed pump is available, feedwater can be supplied to the steam generators by an auxiliary feedwater pump. Steam relief would be through the turbine bypass system to the condenser, if available, or through the system's atmospheric relief valves.
l The auxiliary feedwater system is designed to provide sufficient flow on loss of main feedwater to match decay heat plus Reactor Coolant Pump heat input to the Reactor Coolant System before solid pressurizer operation could occur.(4)
The 250,000 gallcus of water in the condensate storage tank is sufficient to remove decay heat (plus Reactor Coolant pump heat for two pumps) for approximately 13 hours1.50463e-4 days <br />0.00361 hours <br />2.149471e-5 weeks <br />4.9465e-6 months <br />. This volume provides sufficient water to remove the decay heat for approximately 5.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> and to subseguently cool the plant to the DHR system pressure at a cool down rate of 50 F/hr (1).
The minimum relief qapacity of seventeen steam system safety valves is 13,329,163 lb/hr.L2; This is sufficient capacity to protect th system under the design overpower condition of 112 percent.(3) e steam 152>
Both trains of main feedwater isolation on each main feedwater line are required to be operable. Train A of main feedwater isolation is comprised of main feedwater control valves, main feedwater block valves and startup control valves. Train B of main feedwater isolaticn is comprised of the main feedwater isolation valves.
Proposed Amendment No. 152 3-24
RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Limiting Conditions for Operation REFERENCES 152>
(1)
B and W Document 32-1141727-00, " Heat Removal Capability of SMUD CST," March 1984.
(2)
FSAR paragraph 10.3.4 i
(3)
FSAR Appendix 3A, Answer to Question 3A.5 I
152>
(4)
B and W Calculation 86-1123794-99, "SMUD AFW System Following Loss of Feedwater," March 1981 I
i Proposed Amendment No.152 3-24a l
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RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Limitiing Conditions for Operation 3.5 INSTRUMENTATION SYSTEMS 3.5.1 OPERATIONAL SAFETY INSTRUMENTATION Applicability-Applies to unit instrumentation and control systems.
Objective To delineate the conditions of the unit instrumentation and safety circuits necessary to assure reactor safety.
Specifications 3.5.1.1 Startup and operation are not pemitted unless the requirements of Tchle 3.5.1-1, Columns A and B are met.
3.5.1.2 In the event the number of protection channels operable falls below the limit given under Table 3.5.1-1, Columns A and B, operation shall be limited as specified in Column C.
152>
In the event the number of operable Process Instrumentation or EFIC system channels is less than the Total Number of Channel (s), restore the inoperable channels to operable status within 7 days, or be in at least hot shutdown within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
If the number of 152>
operable channels is one less than the minimum channels operable,-
either restore the inoperable channels to operable within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> or be in at least hot shutdown within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
If the number of operable channels is two less than the minimum channels operable, the reactor shall be made subcritical within four hours and on decay heat cooling within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
3.5.1.3 For on-line testing or in the event of a protection instrument or channel failure, a key operated channel bypass switch associated with each reactor protection channel will be used to lock the channel trip relay in the untripped state as indicated by a light.
Only one channel shall be locked in this untripped state at any one
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tine.
3.5.1.4 The key operated shutdown bypass switch associated with each reactor protection channel shall not be used during reactor power operation.
3.5.1.5 During startup when the intemediate range instrument comes on scale, the overlap between the intermediate range and the source range instrumentation shall not be less than one decade.
If the overlap is less than one decade, the flux level shall be maintained in the source range until the one decade overlap is achieved.
Proposed Amendment No.152 3-25
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RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Limiting Conditions for Operation 3.5.1.6 In the event that one of the trip devices in either of the sources
-supplying power to the control rod drive mechanisms fails in the untripped state, the power supplied to the rod drive mechanisms through the failed trip device shall be manually removed within 30 minutes. The condition will be corrected and the remaining trip devices shall be tested within eight hours.
If the condition is not corrected and the remaining trip devices are not tested within the eight-hour period, the reactor shall be placed in the hot shutdown condition within an additional four hours.
152>
3.5.1.7 For calibration or maintenance of an Emergency Feedwater Initiation and Control (EFIC) channel, a key operated " maintenance bypass" switch associated with each channel will be used which will prevent the initiate signal from being' transmitted to the Channel A and B trip logic. Only one channel shall be locked into " maintenance bypass" at.any one time, j
3.5.1.8 If a channel of the RPS is in bypass, it is permissible to bypass only the corresponding channel of EFIC.
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Bases Every reasonable effort will be made to maintain all safety instrumentation in operation. A startup is not permitted unless three power range neutron instrument channels and two channels each of the following are operable:
four reactor coolant temperature instrument channels, four reactor coolant flow instrument channels, four reactor coolant pressure instrument channels, four pressure-temperature instrument channels, four flux-imbalance flow instrument channels, four power-number of pumps instrument channels, and four high reactor building pressure instrument channels. The safety features actuation system must have two analog channels functioning correctly prior to startup. EFIC system instrumentation as required by Table 3.5.1-1 must be operable.
I Operation at rated power is permitted as long as the systems have at least the redundancy requirements of Column B (Table 3.5.1-1).
This is in agreement with redundancy ard single failure criteria of IEEE 279 as described in FSAR section 7.
152>
The four reactor protection channels were provided with key operated maintenance bypass switches interlocked to allow on-line testing or maintenance on only one channel at a time during power operation. Each channel is provided alarm and lights to indicate when that channel is bypassed.
Proposed Amendment No. 152 3-25a
RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Limiting Conditions for Operation Bases (Continued)
Each reactor protection channel key operated shutdown bypass switch is provided with alarm and lights to indicate when the shutdown bypass switch is being used. There are four shutdown bypass keys in the control room under the administrative control of the shif t supervisor.
The keys will not be used during reactor power operation.
There are four reactor protection channels.
Normal trip logic is two out of four. Required trip logic for the power range instrunentation channels is 152>
two out of three. The EFIC trip logic is two times one-out-of-two taken twice. Minimum trip logic on other instrumentation channels is one out of-two.
The EFIC system is designed to automatically initiate AFW when:
1.
all four RC pumps are tripped, 2.
RPS has tripped the reactor on anticipatory trip indicating loss of main feedwater, 3.
the level of either steam generator.is low, 4.
either steam generator pressure is low, or 5.
SFAS ECCS actuation (high RB pressure or low RCS pressure).
The EFIC system will isolate main feedwater to any steam generator when the pressure goes below 600 psig.
The EFIC system is also designed to isolate or feed AFW according to the following logic:
If both SGs are above 600 psig, supply AFW to both SGs If one SG is below 600 psig, supply AFW to the other SG If both SGs are below 600 psig but the pressure difference between the two SGs exceeds 100 psig, supply AFW only to the SG with the higher pressure If both SGs are below 600 psig and the pressure difference is less than 100 psig, supply AFW to both SGs At cold shutdown conditions all EFIC initiate and isolate functions are manually or automatically bypassed. When pressure in both steam generators is greater than 750 psig, the following bypassed initiation signals will have been automatically reset:
- 1) Loss of 4 RC pumps, 2) low steam generator pressure, 3) low steam generator level.
Proposed Amendment No. 152 3-26
t RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Limiting Conditions for Operation 152 Since the EFIC receives signals from the RPS it is important that only corresponding channels be placed in " maintenance bypass." If a channel of RPS is in maintenance bypass, only the corresponding channel of EFIC can be bypassed. An interlock feature also prevents bypassing more than one EFIC channel at a time. These interlocking features allow the EFIC system to take a single failure in addition to having one channel in maintenance bypass.
Various RPS test features can inhibit initiate signals to the EFIC system and degrade the EFIC system below acceptable limits if the RPS channel is not in bypass. Therefore, no testing should be performed on a RPS instrument string which supplies an output to EFIC without placing that RPS channel in bypass.
The EFIC system is designed to allow testing during power operation. The EFIC system can be tested from its input terminals to the actuated device controllers without placing the channel in key locked " maintenance bypass."
A test of the EFIC trip logic will actuate one of two relays in the controllers. The two relays are tested individually to prevent automatic actuation of the component.
Each EFIC channel key operated maintenance bypass switch is provided with alarm and lights to indicate when the maintenance bypass switch is being used.
4 The source range and intermediate range nuclear flux instrumentatiop0*## "
overlap by one decade. This decade overlap will be achieved at 10-amps on the intermediate range scale.
Power is normally supplied to the control rod drive mechanisms from two separate parallel 480 volt sources. Redundant trip devices are employed in each of these sources.
If any one of these trip devices fails in the untripped state on-line repairs to the failed device, when practical, will be made, and the remaining trip devices will be tested. Eight hours is aaple time to test the remaining trip devices and in many cases make on-line repai rs.
Proposed Anendment No. 152 3-26a
RANCHO SECL UNIT 1 TECHNICAL SPECIFICATIONS Lf;iting Conditions f:r Oper: tion Table 3.5.1-1 (Continued)
INSTRUMENTS OPERATING CONDITIONS (C)
(A)
(8)
Operator Action if Functional Unit Total Number of Minimum Channels Conditions of Columns A Channels Operable and B Cannot be Met 9.
Reactor Building Purge Isolation 2
1 Operation may continue provioed the purge on high radiation inlet and outlet valves of the inoperable channel (s) are closed and their respective breakers de-energized or comply with 3.5.1.2.
At cold shutdown or refueling, each of the purge inlet and outlet valves will be closed.
152>
Emergency Feedwater Initiation and Control (EFIC) System 1.
AFW Initiation a.
Manual 2
2 See 3.5.1.2.
b.
Low Level, SGA or B 4/SG 3/SG(Note 1)
See 3.5.1.2.
May be bypassed below 750 psig OTSG pressure, c.
Low Pressure, SGA or B 4/SG 3/SG(Note 1)
See 3.5.1.2.
May be bypassed below 750 psig OTSG pressure.
d.
Loss of MFW Anticipa-tory Reactor Trip 4
3(Note 1)
See 3.5.1.2.
Loss of MFW Anticipatory Reactor Trip is effectively typassed in RPS below 20 percent power.
e.
Loss of 4 RC Pumps 4
3(Note 1)
See 3.5.1.2.
May be bypassed below 750 psig OTSG pressure, f.
Automatic Trip Logic 2
2 See 3.5.1.2.
2.
SG-A Main Feedwater Isolation a.
Manual 2
2 See 3.5.1.2.
b.
Low SGA Pressure 4
3(Note 1)
See 3.5.1.2.
May be bypassed below 750 psig OTSG pressure.
c.
Automatic Trip Logic 2
- 2 See 3.5.1.2.
4 Note 1 The number of minimum channels operable may be reduced to 2 provided one-nf the inoperable channels is in a tripped state.
Proposed Amendment No.152 3-30a
+,_
m.,
p V
e v
r-
. [ #s, x.
RANCHO SECO UNIT 1 1520 TECHNICAL SPECIFICATIONS T le3.5.1-1(Continued)-
INSTRUMENTS OPERATIM CONDITIONS I
~
(0)
(A)
(B) rator Action if Functional Unit Total Number of Minimum Channels Codditions of Columns A Channels Operable and 8 Cannot be Met 3.
SG.B Main Feedwater Isolation 4.
Hanual 2
2 See 3.5.1.2.
b.
Low SGB Pressure 4
3(Note 1)
See 3.5.1.2. ' May be bypassed below 750 psig OTSG pressure.
c.
Automatic Trip Logic 2
2 See 3.5.1.2.
4.
FJW Valve Commands (Vector) a.
Vector Enable 2
2..
See 3.5.1.2.
b.
Vector Module 4
3 See 3.5.1.2 c.
Control Enable 2
2 See 3.5.1.2 d.
Control Module 2
2 See 3.5.1.2 Note 1 The number of minimum channels operable may be reduced to 2 provided one of the inoperable channels is placed in a 4
tripped state.
Proposed Amen <twnt No.152 3-300
'l i
RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Limiting Conditions for Operation 3.5.3 SAFETY FEATURES ACTUATION SYSTEM SETPOINTS Applicability This specification applies to the safety features actuation system actuation setpoints.
Objective To provide for automatic initiation of the safety features actuation system in the event of a breach of reactor coolant system integrity.
Specification The safety features actuation setpoints and permissible bypasses shall be as follows:
Functional Unit Action Setpoint High Reactor Building Reactor Building spray valves ***
130 psig pressure
- Reactor Building spray pumps ***
130 psig kigh pressure injection, and start of Reactor Building cooling and Reactor Building isolation.
14 psig 152><
Low pressure injection,EFIC AFW initiate 14 psig Low reactor coolant system High pressure injection, and start pressure **
of Reactor Building cooling and Reactor Building Isolation
?_1600 psig
~.
Low pressure injection,EFIC AFW initiate 2.1600 psig 152><
Automatic Actuation Logic All above Not Applicable Manual All above Not Applicable 152><
- May be bypassed during Reactor Building leak rate test.
- May be bypassed below 1850 psig and is automatically reinstated above 1850 psig.
- Five-minute time delay.
Proposed Amendment No. 152 3-34
RANCHO SECO UNIT 1 TECHNICAL SPECIFICATICOS Surveillance Standards Table 4.1-1 (Continued)
INSTRUENT SURVEILLANCE REQUIREENTS Channel Description Check Test Calibrate Remarks i 42.
Reactor Bu11 ding drain i
accumulation tank level NA NA R
43.
Incore neutron detectors M(1)
NA NA (1) Check functioning, including functioning of cosaputer readout and/or recorder readout, t
44.
a.
Process and area radi-ation monitoring system W
M Q
i b.
Containment Area Monitors W
NA R
l
- 45. Emergency plant radiation 4
l Instruments M(1)
NA R
(1) Battery check 46.
Environmental air nonitors M(1)
NA R
(1) Check functioning 47.
Strong motion accelerometer Q(1)
NA R
(1) Battery check 1
152>< 48.
Deleted
- 49. Pressurizer Water I.evel H
NA R
l
- 53. Auxili ary Feedwater Flow l
Rate M
NA R
- 51. Reactor Coolant Sy stem Sub-
]
cc<oling Margin Monitor M
NA R
1
- 52. EMOV Power Position
)
Indicator
]
(Primary Detector)
M NA R
- 53. EM3V Position Indicator (Backup Detector)
M NA R
T/C or Acoustic 54.
EHOV Glock Valve Position Indicator M
NA R
+
- 55. Safety Valve Position In-dicator (Primary Detector)
M NA R
T/C t
Proposed Amendment No.152 4-7b
RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Surveillance Standards TABLE 4.1-1 (Continued)
INSTRUMENT SURVEILLANCE REQUIREMENTS i
Channel Description Check Test Calibrate Rena s 152> 68. AFW Initiation a.
Manual H/A H
N/A b.
Low Level SGA or B S
M R
c.
Los Pressure SGA or S S
tl R
d.
Loss of MFW Anticipa-tory Reactor Trip S
M R
e.
Loss of 4 RC Pumps 5
M H/A f.
Automatic Trip Logic N/A M
N/A 69.
SGA Hain Feedwater Line Isolation a.
Manual H/A H
ff/A h.
Autoisatic Trip Logic
!!/A 11 fi/A 70.
SGB Hain Feedwater Line Isolation a.
Manual H/A fi ff/A b.
Autoiaatic Trip Logic N/A H
N/A
< Proposed Amendment No.152 4-7d
r RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Surve111anca Standards TABLE 4.1-1 (Continued)
INSTRUMENT SURVEILLANCE REQUIREMENTS Channel Description Check Test Calibrate Remarts 152> 71. AFW Valve Commands (Yector) a.
AFW Initiation Autornatic Trip Logic Tripped N/A M
N/A b.
SGA Pressure Low 5
M R
c.
SGB Pressure Low S
M R
d.
SG Pressure Difference SGA Pressure >
S M
R
.SGB P m s su r,e _,,,,
SPA Pressure >
SGA Pressure 5
M R
72.
AFW Control Yalve Control j
a.
Ibnual/ Auto in Manual N/A M
N/A b.
AFW Initiation Auto-matic Trip Logic Tripped N/A H
N/A l'
- 73. SG Level Control Setpoint 5 election a.
Manual / Auto in Manual N/A M
N/A b.
AFW Initiation Auto-matic Trip Logic Tripped N/A M
N/A c.
Loss of 4 RC Pumps S
M N/A 74 ADV Control Valve Control a.
Manual / Auto in Manual N/A M
N/A S = Each shift M =11onthly P = Prior to each startup if not done previous week D = Daily Q, Quarterly R = Once during the refueling interval I
W = Weekly SY = Semiannual 1
Proposed Amendment No.152 4-7e 1
on.
RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Surveillance Standards TABLE 4.1-2 MINIMUM EQUIPMENT TEST FREQUENCY ltem Test Frequency 1.
Control rods Rod drop times of Each refueling shutdown all full length rods 2.
Control rod movement Movement of each rod Every two weeks 3.
Pressurizer Setpoint Note 3 code safety valves 4.
Main Steam safety Setpoint Note 3 valves 5.
Refueling system Functional Each refueling interval interlocks prior to handling fuel 152>< 6.
Turbine throttle Movement of each valve Monthly stop valves 7.
Reactor Coolant Leakage Calculated inventory weekly System Leakage check daily 8.
Charcoal and high Charcoal and HEPA filter Each refueling interval and efficiency filters for iodine and particul-at any time work on filters ate removal efficiencies, could alter their integrity D0P test on HEPA filters.
Freon test on charcoal filter units.
9.
Fire pumps and power Functional Monthly supplies
- 10. Reactor Building Functional Each refueling interval isolation trip
- 11. Spent fuel cooling Functional Each refueling interval l
system prior to fuel handling
- 12. Turbine Overspeed Calibration Each refueling interval Trips
- 13. Internals Vent Manual Actuation, III Each refueling interval Remot9}isualinspec-Valves tion,12 and verify that valve not stuck open.
Proposed Amendment No.152 4-8
RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Surveillance Standards TABLE 4.1 2 (Continued)
MINIMUM EQUIPMENT TEST FREQUENCY Item Test Frequency
- 14. Reactor Coolant Functional tqst of Each refueling interval System High Point each valveI43 Vents
- 15. Low Temperature Functional (5)
Prior to RCS temperature Overpressure decreasing below 350'F Protection (Et10V) 152> 16. Main Feedwater Isolation Valves a.
Main Feedwater Functional Each refueling interval.
Isolation Valves b.
Main Feedwater Functional Each refueling interval.
Block Valves c.
Startup Feedwater Functional Each refueling interval.
Control Valves d.
Main Feedwater Functional Each refueling interval.
Control Valves
- 17. Turbine Throttle Cycle Each refueling interval.
Stop Valves 4
1.
Verifying through manual actuation that the valve is fully open with a force of < 400 lbs. (applied vertically upward).
2.
Check visually accessible surfaces to evaluate observed surf ace irregularities.
3.
Tested in accordance with Section XI of the ASME Boiler and Pressure Yessel Code and applicable Addenda as required by 10 CFR 50, Section 50.55a(g), except where specific written relief has been granted by the NRC pursuant to 10 CFR 50, Section 50.55a(g)(6)(1).
4.
Cycle each valve in the vent path through at least one complete cycle of full travel from the control room and verify the flow of gas through the system vent path.
Verify all manual isolation valves in each vent path are locked in the open position.
5.
EMOV block valve closed during test.
Proposed Amendment No. 152 4-8a
RANCHO SECO UNPT 1 TECHNICAL SPECIFICATIONS Surveillance Standards 4.8 AUXILIARY FEE 0 WATER PUMP PERIODIC TESTING Applicability Applies to the periodic testing of the turbine and motor driven auxiliary feedwater pumps.
Objective To verify that the auxiliary feedwater pump and associated valves are operable.
Specification 4.8.1 Monthly on a staggered test basis at a time when the average reactor coolant system temperature is >305*F, the turbine / motor driven and motor driven auxiliary feedwater pumps shall be operated on recirculation to the condenser to verify proper operation.
Separate tests will be performed in order to verify the turbine driven capability and the motor driven capability of auxiliary feedwater pump P-318.
The monthly test frequency requirement shall be brought current within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> after the average reactor coolant system temperature is >305'F for the motor driven puups. The turbine driven capability shat 1 be brought current within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> of obtaining 5 percent reactor power.
152>
Acceptable performance will be indicated if the pump starts and operates for fifteen minutes at a flow rate of greater than or equal to 760 gpm at a discharge pressure sufficient to drive that flow through the most restrictive flow path to a single steam generator which is at a pressure of 1050 psig.
4.8.2 At least once per 18 months:
1.
Verify that each automatic valve in the flow path actuates to its correct position upon receipt of each auxiliary feedwater actuation test signal.
2.
Verify that each auxiliary feedwater pump starts as designed automatically upon receipt of each auxiliary feedwater actuation test signal.
4.8.3 All valves, including those that are locked, sealed, or otherwise secured in position, are to be inspected monthly to verify they are in the proper position.
i Proposed Amendment No. 152 4-39
RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Surveillance Standards 4.8.4 Prior to startup following a refueling shutdown or any cold shutdown of longer than 30 days duration, conduct a test to demonstrate that the motor-driven AFW pumps can pump water from the CST to the steam generator.
152><
Bases The monthly test frequency will be sufficient to verify that the turbine / motor driven and motor driven auxiliary feedwater pumps are operable. Verification of correct operation will be made both from the control room instrumentation and direct visual observation of the pumps.
The OPERABILITY of the auxiliary feedwater system ensures that the Reactor Coolant System can be cooled down to less than 305*F from normal operating conditions in the event of a total loss of off-site power.
The electric driven auxiliary feedwater pu:aps are capable of delivering a 152> total feedwater flow of 760 gpm at a pressure of 1050 psig to the entrance of the steam generators. The steam driven auxiliary feedwater pump is capable of delivering a total feedwater flow of 760 gpm to the entrance of the steam generators over the steam generator operating range of 800 psig to 1050 psig.
This capacity is utilized as analytical input to the Loss of Hain Feedwater Analysis which is the design basis event for AFW flow requirements.
4 Proposed Amendment No.152 4-39a
I ENCLOSURE 3 Design Basis Report - ECN AS415 I
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