ML19320A270
| ML19320A270 | |
| Person / Time | |
|---|---|
| Site: | Arkansas Nuclear  |
| Issue date: | 11/15/1972 |
| From: | Schwencer A US ATOMIC ENERGY COMMISSION (AEC) |
| To: | Phillips J ARKANSAS POWER & LIGHT CO. |
| References | |
| NUDOCS 8004210557 | |
| Download: ML19320A270 (3) | |
Text
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DISTRR w IION AEC PDR DKnuth Local PDR HDenton i
Docket PWR Branch Chiefs RP Reading RWKlecker FWR4 Reading OGC a auer 0 (3)
Decket No. 50-313 NOV 151972 RSBoyd RMBernero RCDeYoung EBoulbourne DSKovholt JCalVo FSchroeder LRiani RTedesco Mr. J. D. Phillipe Time President & Chief Engimane RRMaccary Arh==,=c Feuer & Light coupesy Sixth and Fine Streets Fine Bluff, Arka=*ma 71601 Deer Mr. Phillip4, On the basia of our conti==ing :eview of the Final Safety Analysis Report for the Arkansas N=elaar One - Unit No. 1, we find that we need additional information to couplete our evaluation. The specific information required is listed in the sueleeure.
In order to mai-semi = our licensing review schedule we will need a completely adequate response by December 8, 1972.
T1==== inform as withis seven (7) days after receipt of this latter of your confirmation of the schedule or the date you will be ehle to meet. If ye: e===at meet our specified date
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or if yocr reply is not fully x;:---4ve to our requests it is hiehty likely that the overall schad=1m for sempleting the licensing review for this project will have to be extended.
184=*= reassignment of the staff's efforts will require completiam of the mer assignment prior to returning to this project, the extant of extensies will most likely be great 6r them the extent of delay in your response.
Sincerely, Origin 8I STgM5r
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% A.Pellier
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A. Seksemeer, Chief Fressurised Water Res.etors Branch No. 4 Directorate of Lie===ing n=*1=aure:
Request for Additfamat Information es w/eacl:
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REQUEST FOR ADDITIONAL INFORMATION ARKANSAS NUCLEAR ONE - UNIT 1 DOCKET NO. 50-313 9.0 AUXILIARY AND EMERGENCY SYSTEMS m;r response to Request for Information 9.47 should include 9.47 v
e scussion of the esults of an evaluation of the effect on the reactor vessel supports and the reactor cavity liquid ceal that could result from the postulated dropping of the heavy components indicated.
14.0 SAFE"Y ANALYSIS 14.11 Your analysis of the steam line break accident assumes that only pne of the steam generators blows down through the break.- Your
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' analysis indicates that the cooling effect of this blowdown can cause the plant to return to power momentarily 44.5 seconds after the break occurs. In your design the main steam block valves, just outside the reactor building, are the boundary between seismic category I and seismic category II steam piping; these valves are closed by a manual switch in the control room. Thus, if the category II piping of both main steam lines fails in a major seismic event, as we would assume, then both steam generators will blow down through the breaks until the main steam block valves are closed. Manual actuation is not acceptable to assure timely closure of these valves. Therefore, you should present the results of your analysis for the simultaneous rupture of both steam lines and attendant blowdown of both steam generators, or you should explain what changes to your system you will make to prevent this occurrence.
14.12 In your analysis of recovery from a steam line break accident you indicate that at least one steam generator must be intact for use with the emergency feedwater system to remove reactor decay heat.
Our evaluation indicates that delivery of this emergency feedwater may be frustrated by a number of single failures under steam line break accident conditions, for example:
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o a. Emergency feedwater inlet valve failure. If the steam line break occurs within the isolation boundary of a steam generator, that generator may not be usable for decay heat removal. The supply of energency feedwater to the intact steam generator requires that the electric-motor-operated inlet valve open on signal. If the motor operator of this valve jams, preventing opening by eithe electric signal or manually, the supply of emergency feed.-
water is blocked from the intact steam generator.
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- b. Emergency feed pump failure.
To assure an emergency feedwater supply you have provided both a turbine-driven and an electric-motor-driven emergency feed pump. If a seismic event causes failure of Category II sections of the main steam piping, these emergency feed pumps are needed for decay heat removal. Figure 1-6 of the FSAR indicates thac these two pumps lie side-by-side in the same compartment of the auxiliary building. Thus, it appears that a single failure of one of these pumps can cause the failure of the other by flooding or mechanical damage.
- c. Control system failure.
It is not clear to us from your system description how the Integrated Control System (ICS) controls the normal and emergency feedwater valves and pumps. It appears that failure of the ICS, say in a seismic event, could leave the emergency feedwater system inoperative.
Provide the results of your ana3ysis to show that no such single failure will present the supply of adequate emergency feedwater in the event of a steam system frilure, or show that emergency feedwater is not needed to shut the plant down safely. In your analysis it is appropriate to assume that offsite power is lost at the time the accident occurs.
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