ML19256D231
| ML19256D231 | |
| Person / Time | |
|---|---|
| Site: | Crane  |
| Issue date: | 08/01/1967 |
| From: | Pawlick S US ATOMIC ENERGY COMMISSION (AEC) |
| To: | Long C US ATOMIC ENERGY COMMISSION (AEC) |
| References | |
| NUDOCS 7910170734 | |
| Download: ML19256D231 (4) | |
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P00R BREM Charles C. Long, Chief i!
.c-Reactor Projects Branch #3, DEL THRU:
- 1. C. DeYoung, Chief Cantainment & C m - nt Technology Branch, DRL Distributicn-Stefan S. Pawlicki Sucol.
Containmaant & Camponent Technology Branch, DEL DRE'Readin2 ea gEEEE:MJM2M]dND CM SHOCK Gt REACTOR YESSEL cSCTS Re
-.3 C&CTB: DEL: SSP ET-100 60 -2h s
We are engaged presently in a study directed at the resolution of tae problem of a cold water shock on the reactor vessel, induced by opera-tism of the emergency core cooling system (ECCS). The cold water shock has a potential for causing a reactor vessel f ailure, particu-larly through a brittle fracture of the vessel. It could also cause distertion of reactor vessel internals inpeding the effectiveness of core eaa14n=.
M splicants, so far, Westinghouse (Enbinson Unit No. 2) and B & W
- , have analyzed the problem. In response to the staff's ques-tisa,both have psovided a brief m ry of their analysis considering three modes of failure: dactile yielding, brittle fracture and fatigue failure. They conclude that there will be no loss of vessel integrity des to the thermal shock during operation of the er.ergency cora ecoling system. The deenf1= of the calculational procedure, acwever, verc not presented, and the staff reported to the ACRS (Oconee, Report ;;o. 2) that the problem cannet be considered resolved. We reported 2130, t.-.c we are engaged in reviewing the Westinghouse approach to ene proala and expect to meet with B & W in the future for a similar dcta na-revi a.
The thermal shock on reactor vessel received considerable actanacn at the last ACRS meeting on the Oconee plant. The cccanittec reco:: aend2a that the Aegulatory Staff review analyses of the thersal saccA affects on the integrity of the reactor vessel.
In order to be able to perform such a review, Reactor Technology will moed some details of the calculational procedures employed by tha appli-cants. Attmehad to this memorsadum is a list of information ta be inc1=Jad in the report on the thermal shock analysis. We request that this list be formally ubmitted to the afstropolitan Zaist.,n company in e=== action with their Three Mila Island application.
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