ML19351D736
| ML19351D736 | |
| Person / Time | |
|---|---|
| Site: | Yankee Rowe |
| Issue date: | 01/15/1962 |
| From: | Bryon R US ATOMIC ENERGY COMMISSION (AEC) |
| To: | Coe R YANKEE ATOMIC ELECTRIC CO. |
| References | |
| NUDOCS 8011170737 | |
| Download: ML19351D736 (4) | |
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.7 PL4" C72 T<;cket 1:c. 50 2p JAN 1 51562 Yankee Atomic Electric Con:-4ny l&1 Stuart Street Yioston 16, F.assachusetts Attention:
1*r. Lot;er J. Coe Vice Frosident Gentlement Receipt is acknowledged of your "Propened Chr. age No.
13" dated January 5,1962, " Proposed chance ro. ih=
dated Jcnurry 10, 1962, and y:rar letter conceming the decirn and fabrication of Ocrs III for the Tr.nhec reacter dated Jortatry 3,1962 Sinrtrely m rc, Distritration P. A. Ecrric-2 Doc. Room E
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?orna1 Sapol.
Robert H. Drynn, Chief Ia_ihRreadings Tcccerch & Power Eccctor Lafety Str.nch CTEduards Divisien cf Licensinr. and Eent1ction I
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Neutron Irradiation of Reactor Vessels
.(Resume of data from survey of 8/23/61 with additional general information)
Data Sheet No.11-1 1.
Na=e of plant Yankee Nuclear Power Station, Docket 50-29
- 1
'E 2.
Name of owner Yankee Atomic Electric Company 3
Startup date 8/19/60 r
4.
Nominal thermal power
- - 485,000 Kw 13 2
5 Ther=al neutron flux (average) 2 x 10 n/cm,,,,
71 3 Kv/ liter 6.
Power density (average) 7 core dimensions:
diameter 75.6 in.
I.
"=
height
- 90 in.
_ aA 8.
Coolant
- pressurized water 4
Reactor Vessel j:g:
r=
9 Builder
- Babcock and wilcox co=pany 10.
Design pressure 2500 psig Desi n temperature 650*F 11.
6 12.
Internal diameter
- 9 f t.1 in.
=.
13 Wall thickness 7 7/8 in.
- 14. Base metal A-302-Gr. B 15 Ultimate strenSth of base cetal 55,200-95,000 psi
- 16. Yield strength of base metal 59,100 psi s
17 Charpy test not indicated
- 18. Assumed initial NDT not indicated 19 Inspection performed at the mill
- as per ASTM Standards 20.
Method of. welding not indicated 21.
Welding inspection Radiography 7::
- L..
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Data Sheet No. 11-2 22.
Type and method of cladding 0.109 in. thick sheet of S.S. applied to the base material by the Babcock and Wilcox spot weldin6 process 23 Inspection of cladding
- not indicated 24.
Heat treatment of vessel
- The vessel material was quenched and tempered to obtain the desired impact y29 properties.
During the. fabrication period the material was stress relieved at certain intervals. No detrimental influence on impact strength is caused by stress relieving since it was performed ata temperature lower than the tempering temperature.
Desi n code used
- ASME Section VIII 25 6
qg
- 26. Allowable stress at design temp.
20,000 psi.
_ i:=
according to code egc 27 Stress at belt line due to internal : 14,650 psi pressure
- 28. Stress due to gacca heating 1,130 psi 29 Transient stresses due to heatup 4,200 psi /4,600 psi
- and shutdown
- 30. Thermal stress resulting from a 4,000 psi loss of flow
-NVT-at Reactor Vessel Belt Line
- 31. Flux spectrum established
- by ca:.culation
- 32. Plant lead factor
- 0.8 2
33 Yearly integrated flux -nyt-2 3 x 10 n/cm (neutrons between 0.625 MeV and 0.8 MeV) 18 2
mr
- 34. Yearly integrated fast flux -nyt-1.02 x 10 n/cm (neutrons above 0.8 MeV) 1E 2
5 10 x 10 n/cm 35 Integrated fast flux accumulated
=
to date
- 36. Estimated vessel life 20 calendar years 19 2
37 Estimated integrated fast flux 2.04 x 10 n/cm i
over vessel life (belt line)
'l
-l 1
~
i:S+.7 Data Sheet No. 11-3
~"~
Program for monitoring the effects of neutron irradiation on the mechanical properties of the reactor vessel material
=-
- 38. Origin of test specimens original and reference material 39 Planned number of tensile test
- up to 192
=?1 specimens
~~'
- 40. Planned number of Charpy test up to 168 specimens
==.;; ;
- 41. Total number of specimens (tensile 12 or 24
+ Charpy) in one test lot--
- 42. Imposed licitations on the operating:
pressure reduced to 500 psi at temp.
conditions of the reactor vessel below 300*F 9E 43 Data when limitations will be in immediately effect
=:
- 44. Proposed steps which may extend the :
none considered service life of the reactor vessel e
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- ^. :.
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