ML19220B885
| ML19220B885 | |
| Person / Time | |
|---|---|
| Site: | Crane  |
| Issue date: | 08/05/1968 |
| From: | Deyoung R US ATOMIC ENERGY COMMISSION (AEC) |
| To: | Boyd R US ATOMIC ENERGY COMMISSION (AEC) |
| References | |
| NUDOCS 7904270588 | |
| Download: ML19220B885 (3) | |
Text
N DIS"'RI3CICN:
. v Euppl. N O*-1 Feelin6 C& r3 Reediis AD/RT Reading 4' 5
- 3E3 R. S. Soyd, Asaistant Director for Recctor Projects, OFI THEU: Saul Levine, Assistant Director for Reacter Techncica. DF'", _
OY3TZR CRZZX UNIT 2, DOCIrf 30 50 320 Ve have revieved the info =ation sutcitted in the PSX1 ani M.eninent 2 for the Jersey Central Power & L16ht Ceepany's Oyster Creek Unit F P,ation. Cn the ba. sis of S. S. Pavlicki's evaluatico, ve find that the inforsation suh21tted is deficient in the areas of design criteria for Class I no4* cal syste=s and the:=al shock on reacter cct;cnents.
Incle ed is a list of additional information reqaired +4 couplete ot:r ovaluation. Craft copies of the enclosure vere given +4 R. R.
Povell, of Reactor Projects Branch No. 2, on July 31, 1968, for his guidance in processing the Oys*4r Creek Unit 2 application.
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CngraalS.8 ed :s R. C. Ct f CVG R. C. D=Yeung, Chief RT:845 Con w "~ nt & Ccapenent Technology 3 ranch LRL:CLCT3:33P Division of Reactor Licens1=g Zaclosure:
As stated above cc R. L. Tedesco, DEL R. R. Povell, DEL (3) 79042705E8 bec:
S. 3. Pavlicki R. C. DeYoung
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OYSTER CRM NUCLEAR STATICN - C'iI" 2 (LCCKET NO. 50-320)
SEISMIC I:ESIGN Appendix 5A (page 5A-5) states the design criteria for Class I syste=s and cc=penents, and refers to a method of analysis described in the Flcrida Power Corporation's PSR for Crystal River Unit 3, N cket No.
50-302, /cend:ent No. 2 (Supple =ent No.1) dated Febmary 7,1968, G estion 9 11.
To clarify the proposed design criteria, eenfirm that all Class I systems and components vill be designed usit y the loading ccmbinations and stress 11=its listed in page 9 11-5 s a.ised 4-8-68) of the Crystal River Unit 3 application. Also, compare the propcsed design criteria with the criteria for e=ergency and fault conditions recently approved by the ASME Sectiot. III Cc==ittee.
Previde strain limits for the principal materials of const:uction fv2 the leading ec=bination including si=altanecus =axi=
earthquake and loss-of-ecolant loads.
Describe =ethods that vill be used for seismic analysis of class I syste=s and ec=ponents. As an exe=ple, describe the analytical rodel to be used for the reactor coolant syste=, including location af tu. aped = asses and support conditions.
Identify specific reactor inte=als which =ust maintain their functional perfor=ance capabilities to assure safe shutdown of,the reactor. Provide calculated (or estimated) sax'-"' 11=its of defor=ation er stress, at which inability to function occurs, fer each co=penent i ntified. Also, supply the calculated (cr estimated)
=ax1=um design 11:10 value, and the expected defor:ation er stress.
In all cases identify the applicable lc # ng ec=bination and state the proposed =argin of safety.
For reactor internals provide infor:ation that vill permit evaluation of the effect of irradiation en the material properties and on the preposed defor:ation li=1ts.
m v.AL SHOCK CN R7CTOR CCVPONE*T2S With regard to ther=al shock on reactor ec=ponents, educed by operation of the e ergency cere cooling syntes (ICCS), the PSAR (page 4-16) refers to the results reported in Metropc'.1*m Edison Cc=pany's PSAR for Three-Mile Island Nuclear Station (Docket No.
50-269), Anend ent so. 3 (supplement so. 2), dated sover2er 6,1967,
Questien 11. ? ovide details of the brittle fracture analysis, including equations used to correlate crack size with the calculated stress intensity f actor (~g),andupdateresultsoftheanalysis.
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