ML19338B827
| ML19338B827 | |
| Person / Time | |
|---|---|
| Site: | Rancho Seco, Crane |
| Issue date: | 03/20/1978 |
| From: | Harris C, Palmer H AFFILIATION NOT ASSIGNED |
| To: | |
| References | |
| TASK-TF, TASK-TMR NUDOCS 8001180330 | |
| Download: ML19338B827 (2) | |
Text
L6@OAPI LfELWHICSMALUATI0ll 0F TifE REACT 0p VESSEL OF TifE SifUD UNIT AT RA!!CIIO
{?, '3R0 F0ffil!E UHANTIclVATDTTifMSTUif OW 3/2DTia
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A Ifactumi hchanics[andlystslof the reactor vessel has been perforned using t
the conservative approads outlined in the ASME Code,Section III, Appendi:c G.
Specific details of the analytical approach are documented in BMJ Topical Report HAW-1004GA, Rev. 1.
Analyses were performed on the two most critical areas of the reactor vessel -
the beltline region and the outlet nozyle region.
At the lowest temperature during the transient the material is still in the upper shelf region (ductilo behavior).
Due to the low level of radiat.icn, degradation to beltline region materials is not significant enough to produce a shi ft in the transition reference temperature.
Consequently, the beltline region materials are also at the upper shelf toughness region.
Fcctors-of-safety for thu beltline region and outlet nozzles have been calcu-lated as follows:
F. of S. =
K K
Im It where K is the reference stress intensity factor IR K, is the stress intensity factor due to pressure g
i K
is the stress intensity factor due to the thermal It gradient through the thickness Tau factor-of-safety for the beltline region is 1.8.
The factor-of-safety for the outlet nozzles is 1.2.
Octniled calculations are provided on sheet 3 of 3 It should be recognized that this is a very conservative analysis.
A postulated fiant size of 1/4t is assumed for the beltline region and the 3.0" nozzle corner flau is assurred for the outlet nozzle.
The material fracture toughness (refer-ence stress intensity value) used is 200 ksi/Tii.
Finally, the calculation of the stresces associated ylth the transient is conservative, i
hb 0M Charles E. Ilarris, PE March 23 1978 YS lienrik S. Palme March 23, 1978 l
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Docket'Ho. 50-346
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. Toledo Edison Cbspany y.
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ATDI: Mr. Iowell E. Roe vice President, Facilities s
- y.. Develop:: ant,' '
Edisco Plaza 300 Madison Avenue Toledo, Ohio 43652 N
Gentlemen: \\
SUBJDCT: IUm3P CIRCUIATICH TE3T - DAVIS BESSE, UNIT to.1 We have evaluated y request, as specified in your letter of February 13, 1978, that Davis Bess unit 1 be relieved /rca the requirements gf con-s ducting a natural circulb ion test.
"We vill still require that a natural rculation test be conducted in accordance with your cocnitaen and st plan described on page 14-103, Revi.= ion No. 27 of'the FSAR to our position sent to you in our letter dated Januhry 24, 1977.
January 24, 1977 letter described the reasons sup prting the staff po io
[ tonaucting the,n*"al _circul' tion test al the first opportunity after However, we find your other request for roceedina to 1001 power and
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- the current tual c(isis joAe acceptable, I t in any event, the Jte -
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~ hould be corx5uctna witnist at least 120 lays rot. the date of this letter.
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Our basis for allcw u to proceed as stated abave is based upon the -.,
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-l coceleted coonee No. I tests which have conf acceptable natural.,1 circulation, and your calculational results pre ted to the NRC staff in.
Sethesda, Maryland on February 14, 1978, which d nstrated that the.
y.
Davis Besse, UnJt 1177 FA raised loop dcsign shoul have natural circu-rienst cot parable to the carpleted res'kts of the Ocnnee lation flow Ib. 1 tes a
Sincerely, D
" M~@Md h l.
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00 f'ker S. Boyd, Director Q
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Divison of Project Management offi of Nuclear Reactor Regulation
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