ML20086C523
| ML20086C523 | |
| Person / Time | |
|---|---|
| Site: | Arkansas Nuclear  |
| Issue date: | 06/22/1995 |
| From: | ENTERGY OPERATIONS, INC. |
| To: | |
| Shared Package | |
| ML20086C510 | List: |
| References | |
| NUDOCS 9507070129 | |
| Download: ML20086C523 (7) | |
Text
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i Tha OPERABILITY of th3 Saismic Monitoring Instrumentation ensuras thet
' sufficient capability is available to promptly determine the magnitude of a 1
. seismic event and evaluate the response of those features important to safety.
This capability is required to permit comparison of the measured response to that used in the design basis for the facility to determine if plant shutdown is required pursuant to Appendix "A" of 10CFR Part 100.
The instrumentation is i
consistent with the recommendations of Safety Guide 12, " Instrumentation for Earthquake," published March 19, 1971, and NUREG-0800 Section 3.7.4,
" Seismic i
Instrumentation."
To support loss of main feedwater analyses, steam line/feedwater line break analyses, SBLOCA analyses, and NUREG-0737 requirements, the EFIC system is designed to automatically initiate EFW when:
1.
all four RC pumps are tripped 2.
both main feedwater pumps are tripped 3.
the level of either steam generator is low 4.
either steam generator pressure is low 5.
ESAS ECCS actuation (high R3 pressure or low RCS pressure)
The EFIC system is also designed to isolate the affected steam generator on a steam line/feedwater line break and supply EFW to the intact generator according to the following logic:
If both SG's are above 600 psig, supply EFW to both SG's.
If one SG is below 600 psig, supply EFW to the other SG.
If both SG's are below 600 psig, but the pressure difference between the two SG's exceeds 100 psig, supply EFW only to the SG with the higher pressure.
If both SG's are below 600 psig and the pressure difference is less than 100 psig, supply EFW to both SG's.
At cold shutdown conditions all EFIC initiate and isolate functions are bypassed except low steam generator level initiate.
The bypassed functions will be automatically reset at the values or plant conditions identified in Specification 3.5.1.15.
" Loss of 4 RC pumps" initiate and " low steam generator pressure" initiate are the only shutdown bypasses to be manually initiated during cooldown.
If reset is not done manually, they will automatically reset. Main feedwater pump trip bypass is automatically removed above 10% power.
The principal function of the control Room Isolation-Hich Radiation is to provide an enclosed environment from which the unit can be operated followine an uncontrolled release of radioactivity. The hiah radiation isolation function provides assurance that under the recuired conditions, an isolation sianal will be civen.
Due to the unicue situation of the sha ed control room habitability, one ANO-1 control room isolation channel receives a hich radiation sianal from the ANO-1 control room area radiation monitor (RE-8001) and the redundant channel receives a hich radiation sianal from the ANO-2 control room ventilation process monitor (2RITS-8750-1).
With no channel of the control room radiation monitoring system operable, the CREVS must be placed in a condition that does not reouire the isolation to occur.
To ensure that the ventilation system has been placed in a state eouivalent to that which occurs after the hiah radiation isolation has occurred, one OPERABLE train of the CREVS is placed in the emercency recirculation mode of operation.
Reactor operation can continue indefinitely in this state.
The 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> Completion Time is a sufficient amount of time in which to take the Reouired Action.
REFERENCE i
Amendment No. 461 43c j
9507070129 950622 PDR ADOCK 05000313 p
.~
4 PSAR, S ction 7.1
- FSAR,,Section 2.7.6 1
1 i
I l
h i
1 3
v 3.8.15 Storago in the rpent fusi pocl shall bs rastrictcd to fusi acocmbliss having initial enrichment less than or equal to 4.1 w/o U-235.
The
-provisions of Specification 3.0.3 are not applicable.
1 3.8.16 Storage in Region 2 (as shown on Figure 3.8.1) of the spent fuel pool shall be further restricted by burnup and enrichment limits specified in Figure 3.8.2.
In the event a checkerboard storage configuration is deemed necessary for a portion of Region 2, vacant spaces adjacent to the faces of any fuel assembly which doe = not meet the Region 2 burnup criteria (non-restricted) shall be physically blocked before any such fuel assembly may be placed in Region 2.
This will prevent inadvertent fuel assembly insertion into two adjacent storage locations. The provisions of Specification 3.0.3 are not applicable.
3.8.17 The boron concentration in the spent fuel pool shall be maintained (at all times) at greater than 1600 parts per million.
3.8.18 Durino the handlina of irradiated fuel in the reactor buildino. the control room emeraency air conditionine system and the control room emeroency ventilation system shall be operable as required by Specification 3.9.
Bases Detailed written procedures will be available for use by refueling personnel.
These procedures, the above specifications, and the design of the fuel handling equipment as described in Section 9.6 of the FSAR incorporating built-in interlocks and safety features, provide assurance that no incident could occur during the retueling operations that would result in a hazard to public health and safety.
If no change is being made in core geometry, one flux monitor is sufficient.
This permits maintenance on the instrumentation.
Continuous monitoring of radiation levels and neutron flux provides immediate indication of an unsafe condition.
The requirement that at least one decay heat removal loop be in operation ensures that (1) sufficient cooling capacity is available to remove decay heat and maintain the water in the reactor pressure vessel at the refueling temperature (no rmally 14 0
- F), and (2) sufficient coolant circulation is maintained through the reactor core to minimize the effects of a boron dilution incident and prevent boron stratification. (1)
The requirement to have two decay heat removal loops operable when there is less than 23 feet of water above the core, ensures that a single failure of the operating decay heat removal loop will not result in a complete loss of decay heat removal capability. With the reactor vessel head removed and 23 feet of water above the core, a large heat sink is available for core cooling, thus in the event of a failure of the operating decay heat removal loop, adequate time is provided to initiate emergency procedures to cool the core.
The shutdown margin indicated in Specification 3.8.4 will keep the core subcritical, even with all control rods withdrawn from the core. (8)
Although the refueling boron concentration is sufficient to naintain the core keff s 0.99 if all the control rods were removed from the core, only a few control rods will be removed at any one time during fuel shuffling and Amendment No. 44, M, M,%,M7, M3, 59a 441,444
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PROPOSED ANO-I TECHNICAL SPECIFICATION CHANGES l
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4 Th2 OPERABILITY of tha Salemic Monitoring Instrumentation ensuras that sufficient capability is available to promptly determine the magnitude of a
.se sm c event and evaluate the response of those features important to safety.
i i This capability is required to permit comparison of the measured response to that used in the design basis for the facility to determine if plant shutdown is required pursuant to Appendix "A" of 10CFR Part 100.
The instrumentation is consistent with the recommendations of Safety Guide 12, " Instrumentation for Earthquake," published March 19, 1971, and NUREG-0800 Section 3.7.4, " Seismic Ins t r2menta tion. "
To support loss of main feedwater analyses, steam line/feedwater line break analysas, SBLOCA analyses, and NUREG-0737 requirements, the EFIC system is designe.d to automatically initiate EFW when:
i 1.
all four RC pumps are tripped 2.
both main feedwater pumps are tripped 3.
the level of either steam generator is low 4.
either steam generator pressure is low 5.
ESAS ECCS actuation (high RB pressure or low RCS pressure)
The EFIC system is also designed to isolate the affected steam generator on a steam line/feedwater line break and supply EFW to the intact generator according to the following logic If both SG's are above 600 psig, supply EFW to both SG's.
If one SG is below 600 psig, supply EFW to the other SG.
If both SG's are below 600 psig, but the pressure difference between the two SG's exceeds 100 psfg, supply EFA only to the SG with the higher pressure.
If both SG's are below 600 psig and the pressure difference is less than 100 psig, supply EFW to both SG's.
At cold shutdown conditions all EFIC initiate and isolate functions are bypassed except low steam generator level initiate. The bypassed functions will be automatically reset at the values or plant conditions identified in Specification 3.5.1.15.
" Loss of 4 RC pumps" initiate and " low steam generator pressure" initiate are the only shutdown bypasses to be manually initiated during cooldown.
If reset is not done manually, they will automatically reset. Main feedwater pump trip bypass is automatically removed above 10% power.
The principal function of the Control Room Isolation-High Radiation is to provide an enclosed environment from which the unit can be operated following an uncontrolled release of radioactivity. The high radiation isolation function provides assurance that under the required conditions, an isolation signal will be given.
Due to the unique situation of the shared control room habitability, one ANO-1 control room isolation channel receives a high radiation signal from the ANO-1 control room area radiation monitor (RE-8001) and the redundant channel receives a high radiation signal from the ANO-2 control room ventilation process monitor (2RITS-8750-1). With no channel of the control room radiation monitoring system operable, the CREVS must be placed in a condition that does not require the isolation to occur. To ensure that the ventilation system has been placed in a state equivalent to that which occurs after the high radiation isolation has occurred, one OPERABLE train of the CREVS is placed in the emergency recirculation mode of operation.
Reactor operation can continue indefinitely in this state. The 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> Completion Time is a sufficient amount of time in which to take the Required Action.
l Amendment No. 161 43c
l REFERENCE
.FSAR[ Section 7.1 FSAR, Section 2.7.6 l
I i
j i
l i
i 1
Amendment No.
43d I
I I
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e 3."8.15, storage in the spent fuel pool shall be restricted to. fuel assemblies having initial enrichment less than or equal to 4.1 w/o U-235.
The provisions of Specification 3.0.3 are not applicable.
3.8.16 Storage in Region 2 (as shown on Figure 3.8.1) of the spent fuel pool shall be further restricted by burnup and enrichment limits specified in Figure 3.8.2.
In the event a checkerboard storage configuration is deemed necessary for a portion of Region 2, vacant spaces adjacent to the faces of any fuel assembly which does not meet the Region 2 burnup criteria (non-restricted) shall be physically blocked before any such fuel assembly may be placed in Region 2.
This will prevent inadvertent fuel assembly insertion into two adjacent storage locations. The provisions of Specification 3.0.3 are not applicable.
3.8.17 The boron concentration in the spent fuel pool shall be maintained (at all times) at greater than 1600 parts per million.
3.8.18 During the handling of irradiated fuel in the reactor building, the control room emergency air conditioning system and the control room emergency ventilation system shall be operable as required by Specification 3.9.
i Bases Detailed written procedures will be available for use by refueling personnel.
These procedures, the above specifications, and the design of the fuel handling equipment as described in Section 9.6 of the FSAR incorporating built-in interlocks and safety features, provide assurance that no incident could occur during the refueling operations that would result in a hazard to public health and safety.
If no change is being made in core geometry, one flux monitor is sufficient. This permits maintenance on the instrumentation.
Continuous monitoring of radiation levels and neutron flux provides immediate indication of an unsafe condition.
The requirement that at least one decay heat removal loop be in operation ensures that (1) sufficient cooling capacity is available to remove decay heat and maintain the water in the reactor pressure vessel at the refueling temperature (normally 140* F), and (2) sufficient coolant circulation is maintained through the reactor core to minimize the effects of a boron dilution incident and prevent boron stratification. (8)
The requirement to have two decay heat removal loops operable when there is less than 23 feet of water above the core, ensures that a single failure of the operating decay heat removal loop will not result in a complete loss of decay heat removal capability. With the reactor vessel head removed and 23 feet of water above the core, a large heat sink is available for core cooling, thus in the event of a failure of the operating decay heat removal loop, adequate time is provided to initiate emergency procedures to cool the core.
The shutdown margin indicated in Specification 3.8.4 will keep the core suberitical, even with all control rods withdrawn from the core. (* ) Although the refueling boron concentration is sufficient to naintain the core keff 5 0.99 if all the control rods were removed from the core, only a few control rods will be removed at any one time during fuel shuffling and c
Amendment No. 47,M,W,M,4M,4M, 59a 44,4W j