ML20059E599
| ML20059E599 | |
| Person / Time | |
|---|---|
| Site: | 05200001 |
| Issue date: | 10/27/1993 |
| From: | Fox J GENERAL ELECTRIC CO. |
| To: | Poslusny C Office of Nuclear Reactor Regulation |
| References | |
| NUDOCS 9311030273 | |
| Download: ML20059E599 (3) | |
Text
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- GENuclear Energy ;
GerreralElectre Company 175 Curtner Avenue, San Jose. CA 95125 October 27,1993 Docket No.52-001 -
Chet Poslusny, Senior Project Manager Standardization Project Directorate Associate Directorate for Advanced Reactors and License Renewal Office of the Nuclear Reactor Regulation
Subject:
Submittal Supporting Accelerated ABWR Schedule - 8 Ilour RCIC Capability
Dear Chet:
Enclosed are SSAR markups addressing the recent GE/NRC discussions pertaining to 8-hour RCIC capability.
Please send a copy of this transmittal to George Thomas.
Sincerely, bY k Fox Advanced Reactor Programs cc:
Alan Beard (GE)
Norman Fletcher (DOE)
Ed Nazareno (GE) o snu oso 020129 in' 22! mgo,. p i:I y
_PDR
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23A6100 Riv. 2 ABi&R standardsafetyAnalysis Report (4) Complete plant shutdown with loss of normal feedwater before the reactor is depressurized to a level where the shutdown cooling system can be placed in operation.w 4h ct cy a.d q ko O
o u <- 3'.
(5) L s of AC power.
The RCIC 'ystem is designed to perform its function without AC power for at least 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> upporting systems such as DC power and the water supply will support the i
RCIC System during this time period. Without AC power, RCIC room cooling will not be available. However, room temperature will not reach the equipment maximum environmental temperature withinfhours.g a
4 Analyses of the G5T water supply volume, DC powerload capacity and RCIC room heatup to support the 2-hour RCIC operation during loss of AC power are provided in Table 5.4-1b, Table 5.4-1c and Figure 5.4-15 espectively. Table 5.4-lb shows a margin of 151% for the CST water. The 2-hour an is based on IEEE-485, for the Dhision I battery (Table 5.4-Ic) shows an additional argin and 118% of the required capacity including the 15% design margin and 25 againg factor with no load shedding. Figure I
5.4-14 shows the RCIC room is well below e maximum equipment environmentallimit,
of 66'C at 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> following loss of A ower with an initial room temnernmre of 32 Ci These analyses will be repeatedMpTh'e COL applican(for the as built facility to demonstrate thephour capability. See Section 5.4 for the COL license informatiqn.
6
%M P e d
- wh s c. s During loss of AC power, the RCIC System, when started at water Level 2, is capable of preventing water level from dropping below the level which ADS mitigates (Level 1).
This accounts for decay heat boiloff and primary system leakages.
Following a reactor scram, steam generation will continue at a reduced rate due to the core fission product decay heat. At this time, the turbine bypass system will divert the steam to the main condenser, and the feedwater system will supply the makeup water required to maintain reactor vessel inventory.
In the event that the reactor vessel is isolated and the feedwater supply unavailable, relief valves are provided to automatically (or remote manually) maintain vessel pressure within desirable limits. The water level in the reactor vessel will drop due to continued steam generation by decay heat. Upon reaching a predetermined low level, the RCIC System will be initiated automatically. The turbine-driven pump will supply demineralized makeup water from (1) the condensate storage tank (CST) to the reactor vessel and (2) the suppression pool. Seismically installed level instrumentation is provided for automatic transfer of the water source with manual override from CST to suppression pool on receipt of either a low CST water level or high suppression pool level signals (CST water is primary source). The turbine will be driven with a portion of the decay heat steam from the reactor vessel and will exhaust to the suppression pool.
S.4 18 Component and Subsystem Design - Amendment 32
23A6100 thv. 2 ABWVR senderdsafny Analysis Report a
5.4.14.3 Safety Evaluation The flexibility and seismic / dynamic analyses to be performed for the design of adequate component support systems include all temporary and transient loading I
conditions expected by each component. Provisions are to be made to provide spring-type supports for the initial dead weight loading due to hydrostatic testing of steam systems to prevent damage to this type support.
5.4.14.4 Inspection and Testing After completion of the installation of a support system, all hangers and snubbers are to be visually examined to assure that they are in correct adjustment to their cold setting position. Upon hot startup operations (Subsection 3.9.2.1.2), thermal growth will be l
observed to confirm that spring-type hangers will function properly between their hot l
and cold setting positions. Final adjustment capability is provided on all hangers and l
snubbers. Weld inspections and standards are to be in accordance with ASME Code l
Section III. Welder qualifications and welding procedures are in accordance with ASME Code Section IX and NF-4300 of ASME Code Section III.
5.4.15 COL License informatioE We~ TsdoMow W\\w g.4.i g, i -Te., q T r e COL applicants will test the[t$a'riiisolation valves in actual operating conditions 2
(70 kg/cm g,286'C).
5.4.16 References 5.4-1 Design and Performance of Genem! Electric Boiling Water ReactorMain Steamline Isolation Valves, General Electric Co., Atomic Power Equipment Depanment.
March 1969 (APED-5750).
g. 4. t s. 2.
An lyes of 8_Lo u F C.I C ga k g\\g M usi)
. g v o v s cl e COL c 4-b w e \\
-Q,, k cA T M k h k" \\
py
+.
ein- - s h 4c
- e. cr c s b ou twp
, L %
(s~ s a s_o_4 b -
- s. 4. c.t )
5 4-55 Component and Subsystem Design - Amendment 32 - Orstt
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