ML20244A954
| ML20244A954 | |
| Person / Time | |
|---|---|
| Site: | Beaver Valley |
| Issue date: | 04/05/1989 |
| From: | DUQUESNE LIGHT CO. |
| To: | |
| Shared Package | |
| ML20244A948 | List: |
| References | |
| NUDOCS 8904180285 | |
| Download: ML20244A954 (3) | |
Text
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PLANT SYSTEMS
. BASES
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3/4.7.7 CONTROL ROOM EMERGENCk HABITABILITY SYSTEM The OPERABILITY of the control room emergency habitability system ensures that the control room will remain habitable for operations personnel during and following all credible accident conditions.
The ambient air temperature is controlled to prevent exceeding the i
allowable equipment qualification temperature for the equipment and l
instrumentation in the control room.
The OPERABILITY of this system J
in conjunction with control room design provisions is based on limiting the radiation exposure to personnel occupying the control room to 5 rem or less whole body, or its equivalent.
This limitation is consistent with the requirements of General Design Criteria 19 of Appendix "A", 10 CFR 50.
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3/4.7.8 SUPPLEMENTAL LEAK COLLECTION AND RELEASE SYSTEM (SLCRS)
The OPERABILITY of the SLCRS provides for the filtering of postulated radioactive effluents resulting from a Fuel Handling Accident (FHA) and-form leakage of LOSS OF COOLANT ACCIDENT (LOCA) activity from systems outside of the Reactor Containment
- building, such as Engineered Safeguards Features (ESF) equipment, prior to their release to-the environment.
This system also collects potential leakage of LOCA activity from the Reactor Containment building penetrations into the contiguous areas ventilated by the SLCRS except for the Main Steam Valve Room and Emergency Air Lock.
The operation of this system was assumed in calculating the postulated offsite doses in the analysis for a FHA.
System operation was also assumed in' that portion of the Design Basis Accident (DBA) LOCA analysis which addressed ESF leakage following the LOCA, however, no credit for SLCRS operation was taken in the DBA LOCA analysis for collection and filtration of Reactor Containment building leakage even though an unquantifiable amount of contiguous area penetration leakage: would in fact be collected and filtered.
Based on the results of the i
- analyses, the SLCRS must be OPERABLE to ensure that ESF leakage following the postulated DBA LOCA and leakage resulting from a FHA will not exceed 10 CFR 100 limits.
3/4.7.9 SEALED SOURCE CONTAMINATION The limitations on sealed source contamination ensure that the total body or individual organ irradiation does not exceed allowable limits in the event of ingestion or inhalation of the source material.
The limitations on removable contamination for sources requiring leak
- testing, including alpha emitters, is based on 10 CFR 70.39(c) limits for plutonium.
Leakage of sources excluded from the requirements of this specification represent less than one maximum permissible body burden for total body irradiation if the source material is inhaled or ingested.
g g3 Sealed sources are classified into three groups according to their mmz
- use, with surveillance requirements commensurate with the probability
$8' of damage to a
source in that group.
Those sources which are O@
frequently handled are required to be tested more often than those l
co which are not.
Sealed sources which are continuously enclosed within ng a
shielded mechanism (i.e.,
sealed sources within radiation
@M monitoring or boron measuring devices) are considered to be stored o5 and need not be tested unless they are removed from the shielded m<
mechanism.
l 3/4.7.10 and 3/4.7.11 RESIDUAL HEAT REMOVAL SYSTEM (RHR)
SM Deleted BEAVER VALLEY - UNIT 1 B 3/4 7-5
V y, :
3/4.7 PLANT SYSTEMS 7
7 BASES
'li 3/4.7.9. SEALED SOURCE CONTAMINATION The limitations on sealed source removable contamination ensure that the total body or ir.dividual organ irradiation does'not exceed allowable limits in the' event of ingestion or inhalation of the source material..The limitations o
on removable contamination for. sources requiring leak testing, including alpha emitters, is based on 10 CFR 70.39(c) limits for plutonium.
Leakage of sources excluded from the requirements of this specification represent less than one-maximum permissible body burden for total body irradiation'if the source material-is inhaled or ingested.
? ' ADb 1-MbtT /
3/4.7.10'and 3/4.7.11 RESIDUAL HEAT REMOVAL SYSTEM (RHR)
Deleted 3/4.7.12' SNUBBERS Allisnubbers are required OPERABLE to ensure that the structural integrity of the. reactor coolant system and all other safety-related systems is main.
tained during and following a seismic or other similar event initiating dynamic loads.
Snubbers excluded from this inspection program are those installed on
'(c
.nonsafety-related systems and.then.only if their failure or. failure of the system.on which they are installed, would have no adverse effect on any safety-related system.
.The' visual inspection frequency. is based upon maintaining a constant level of finubber protection to. systems.
Therefore, the required inspection interval varies inversely with the observed snubber failures and is determined by the number of inoperable snubbers found ~during an inspection.
Inspections performed before that interval has elapsed may be used as a new reference point to determine the next inspection.
When the cause of the rejection of a snubber is clearly established and remedied for that snubber and for any other snubbers that may be generically
(
-susceptible, and verified OPERABLE by inservice functional testing, that snubber may be exempted from being counted as inoperable. Generically, susceptible snubbers are those which are of a specific make or model and have the same design features directly related to rejection of the snubber by visual inspection, or are similarly located or exposed to the same environmental conditions such as temperature, radiation and vibration.
When a snubber is.found inoperable, an engineering evaluation is performed, in addition to the' determination of the snubber mode of failure, in order to' determine if any safety-related component or system has been adversely affected by the inoperability of the snubber. The engineering evaluation shall det. ermine whether or not the snubber mode of failure has imparted a significant effect or 1
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degradation on the supported compt nent or syste::1
-B BVER VALLEY - UNIT 2 8 3/4 7-5 J
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7.
. INSERT 1 Scaled sources are classified into three groups according to their
- use, with surveillance requirements commensurate with the probability of. damage to a
source in that group.
Those sources which are j'
frequently handled are required to be tested more often than those which 'are not.
Sealed sources which are continuously enclosed within a
shielded mechanism (i.e.,
sealed sources within radiation monitoring or boron measuring devices) are considered to be stored and need not be tested unless they are removed from the shielded mechanism.
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