ML20002E007
| ML20002E007 | |
| Person / Time | |
|---|---|
| Site: | Big Rock Point File:Consumers Energy icon.png |
| Issue date: | 01/12/1971 |
| From: | Walke G CONSUMERS ENERGY CO. (FORMERLY CONSUMERS POWER CO.) |
| To: | Morris P US ATOMIC ENERGY COMMISSION (AEC) |
| Shared Package | |
| ML20002E008 | List: |
| References | |
| NUDOCS 8101260067 | |
| Download: ML20002E007 (3) | |
Text
-
- wrf,
/
potKETED h
'K f%
USAEC y
Consumers
'x\\
5~ ggs&
-==
.s L
POWBr c g/ ;
~
\\C% )
U 4
'4 si Company s
'/
/
\\
/
' t.
%Il jy y h t -
ugen Avenue, Jackson, Michigan 49201. A'ee Code 517788-05S0 l A, General Omcas.
.,,-G.,,-
January 12, 1971
'y /,..'p.
Regulatorj Filo Cy.
Dr. Peter A. Morris, Director Re: Docket 50-155 Division of Reactor Licensing License DPR-6 US Atomic Energy Commission Washington, DC 20545
Dear Dr. Morris :
This letter is written to complete Consumers Power Company's response of September 11, 1970 to your letter of August 6,1970 con-cerning sensitized stainless steel components within the primary coolant pressure boundary.
Consumers Power's letter, dated September 11, 1970, deferred providing maximum stress levels for sensitized stainless steel compon-ents (a portion of Question No 2) until a later date. These stress levels are summarized below:
Stress Intensity (ksi)
Component Membrane Peak Alternating Reactor Vessel Instrument Nozzle Extension (795-4) 91 8.7 6.2 Steam Outlet Nozzle Extension (T95-13) 10.0 9.6 63 Letdown Nozzle Extension (795-17) 90 8.7 6.4 Recirculation Nozzle Extension (796-3) 10 5 10.1 6.6 Poison Nozzle Extension (796-8) 91 8.8 49 vent ;;;.,zzle Flange (807-3) 7.2 58 57 Core Support Bracket (802-16) 4.0 11.8
<11.8 Core Support Plate Bracket (802-18) 1.4 15 7
<l5 7 Diffuser Bracket (802-32) 05 39
<39 Steam Drum Downcomer Nozzle Extension (103-2) 11.2 10.1 13 1 Rizer Nozzle Extension (103-8) 99 8.9 53 Vent Nozzle Extension (1G4-7) 51 57 10.2 l
, *: J n
l r
i 2
Dr. Peter.A. Morris January 12, 1971.
These stress values were calculated by Combustion Engineering, Inc at the. request of. Consumers Power Company. A copy of this report (Report 6k60-A), less the detailed calculation sheets, is attached.
A su= mary of the design practices for the steam drum internal brackets is included in the attached report. Stress levels.for the reactor vessel stub tubes were previously submitted v4.th Amendment 8 to the Final Hazards Summary Report dated November 11+, 1961.
Due to a coordination error on the part of Consumers Power
' Company, stress levels for the reactor vessel upper support bracket (798-8,12) were inadvertently not calculated. Therefore, these stress levels are not included.
With regard to the plans developed for surveillance and nondestruction tests of the sensitized stainless steel components (Question 9), the September 11, 1970 letter stated that Consumers.
Power was attempting.to develop equipment to volumetrically examine the reactor vessel steam outlet nozzles from the inside of the vessel-during 1971. This equipment vill not be available prior to the 1971-refueling outage; therefore, plans to inspect the reactor vessel out-let nozzles have been deferred until 1972.
Yours very truly, a-RBS/dmb G. J. Walke Nuclear Fuel Management Administrator m
, *, -