ML13330A061
| ML13330A061 | |
| Person / Time | |
|---|---|
| Site: | San Onofre  |
| Issue date: | 06/05/1980 |
| From: | Baskin K Southern California Edison Co |
| To: | Eisenhut D Office of Nuclear Reactor Regulation |
| References | |
| NUDOCS 8006190497 | |
| Download: ML13330A061 (4) | |
Text
Southern California Edison Company P. o. BOX 800 2244 WALNUT GROVE AVENUE ROSEMEAD. CALIFORNIA 91770 K. P. BASKIN TELEPHONE MANAGER, NUCLEAR ENGINEERING J
(213) 572-1401 AND LICENSING June 5,
1980 Director of Nuclear Reactor Regulation Attention: D. G. Eisenhut, Director Division of Project Management U.S. Nuclear Regulatory Commission Washington, D.C. 20555 Gentlemen:
Subject:
Docket No. 50-206 Turbine Disc Inspection San Onofre Nuclear Generating Station Unit 1 My letter of March 18, 1980 provided information concerning the design and operation of the San Onofre Nuclear Generating Station Unit 1 low pressure turbines. That letter also committed Southern California Edison (SCE) to perform an ultrasonic inspection of the low pressure turbine discs during the Spring, 1980 refueling outage. That inspection has been completed and the results are discussed below.
A general physical inspection indicated the turbines to be clean, with no corrodent buildup or pitting. There was some blade erosion. The ultrasonic inspection of the low pressure discs was completed by Westinghouse Electric Corporation. The discs were free of indications with one exception.
The No. 5 disc, governor end, of the No. 1 low pressure turbine displayed a tangential aim indication in the disc bore area at balance hole No. 13.
This indication could not be confirmed with the use of the radial aim technique.
Therefore, no depth could be estimated. Based upon Westinghouse's experience to date in the verification process, the indication is below normal detectable depth and is of no concern at this time.
Based upon the results of this inspection, the turbines will be returned to normal operation at the end of the refueling/maintenance outage.
If you have any questions, please let me know.
Very truly yours,
- f. so4116 '0 -1 7.
Southern California Edison Company a
5 P.O. BOX 800 C
2244 WALNUT GROVE AVENUE
- ROSEMEAD, CALIFORNIA 91770 C7 June 6, 1980 U. S. Nuclear Regulatory Commission Office of Inspection and Enforcement Region V 1990 North California Boulevard Suite 202, Walnut Creek Plaza Walnut Creek, California 94596 Attention: Mr. R. H. Engelken, Director DOCKET NO. 50-206 SAN ONOFRE - UNIT 1
Dear Sir:
Reference:
- 1) Letter dated May 5, 1980 from SCE (J. M. Curran) to NRC (R. H. Engelken)
The referenced letter provided notification of linear indications found on the surface of the bore in Reactor Coolant Pump A and B Flywheels.
The indications were detected in the intermittent welds of the flywheel bores by a dye penetrant examination.
No indications were found in the base metal or heat affected zone of either flywheel.
The purpose of the welds is to prevent relative motion between the two flywheel plates during fabrication.
The indications are service induced.
After investigation by the flywheel manufacturer, Westinghouse, it was determined that the welds in the bore of the flywheels should be removed.
This is consistent with current fabrication practices employed by Westinghouse, which were implemented subsequent to manufacture of San Onofre flywheels, to prevent in service overstressing welds not designed for load bearing capability.
Load carrying capability is assumed by welds on the outer surface and bolting joining the two halves. The welds were removed by machining out an annular area of each bore, removing both the welds and their heat-affected zones. The annular area removed at the seam between the two plates is a U-shaped groove with no sharp corners and chamfered at the keyways.
This configuration introduces no additional stress concentrations.
The groove in the A flywheel is approximately 1 3/32" deep and 3 5/16" wide The groove in the B flywheel is approximately 1 4/32" deep and 3 1/16" wide.
U. S. Nuclear Regulatory Commission Page 2 The calculated nominal tensile stress in the flywheel bore in the as manufactured condition (at 1200 RPM) is approximately 10,120 psi. The calculated nominal tensile stress in the modified condition is approximately 10, 150 psi.
The calculation of the modified condition assumes the ID of the flywheel bore is increased by 2 1/2". The increase of approximately 30 psi in the calculated nominal tensile stress in the flywheel bore is negligible.
The reduction in moment of inertia of the flywheel due to the removal of the annular area is less than 0.02%. The effect of this reduction in moment of inertia on coastdown characteristics of the Reactor Coolant Pump has been determined to have no significant affect.
Subsequent to removal of the annular area of the bore including the welds and heat-affected zones, a dye penetrant examination of the groove was performed and no indications were found.
It has been concluded with Westinghouse's concurrence that, as a result of the above modifications, the mechanical and structural integrity of the modified flywheel has not been degraded and the original criteria for design is unaltered.
It has been determined that no unreviewed safety questions exist as defined in 10CFR50.59. Therefore, we consider the modifications made appropriate for returning the flywheels to a fully serviceable condition.
If you should require additional information concerning this occurence, please contact me.
Sincerely, H. L. Ottoson Manager of Nuclear Operations
Attachment:
Licensee Event Report 80-025 cc: Director, Nuclear Reactor Regulation (30)
Director, Office of Management Information & Program Control (3)
Director, Nuclear Safety Analysis Center (1)
NRC FORM 366 U. S. NUCLEAR REGULATORY COMMISSION
('7.77)
A LICENSEE EVENT REPORT CONTROL BLOCK:
(PLEASE PRINT OR TYPE ALL REQUIRED INFORMATION) 6 1 Ic IA IS 10 IS ' 101 01 01 -
0001 1 0 1 -I 0l0 31 41 1 1 l 1 01 l01 I 7
8 9
LICENSEE CODE 14 15 LICENSE NUMBER 25 26 LICENSE TYPE 30 57 CAT 58 CON'T II
]
REP O
51o oo 21 o
51 015l8l 0
l10 16o 6l810 SOURCE 1____________________________
7 8
60 61 DOCKET NUMBER 68 69 EVENT DATE
.74 75 REPORT DATE 80 EVENT DESCRIPTION AND PROBABLE CONSEQUENCES3 During a dye penetrant inspection of the Reactor Coolant Pump Flywheel bores in a refueling outage, linear indications were found in the welds. These welds 0
4 are located in the flywheel lamination seam. No indications were found in 'the base material. There were no adverse affects on public health or safety.
F07 1 r81 I 7
8 9 80 SYSTEM CAUSE CAUSE COMP.
VALVE CODE CODE SUBCODE COMPONENTCODE SUBCODE SUBCODE
[O]
IC lB (@ LsJ@ LBJ0 IM 10 IT 10 lR Ix I9 LzJ@ [zi 7
8 9
10 11 12 13 18 19 20 SEQUENTIAL OCCURRENCE REPORT REVISION LER/RO [EVENT YEAR REPORT NO.
CODE TYPE NO.
RUE R
18101 j
1012151
-1.
121L1 Lx1 1-1 L0J 21 22 23 24 26 27 28 29 30 31 32 ACTION FUTURE EFFECT SHUTDOWN ATTACHMENT NPRD-4 PRIME COMP.
COMPONENT TAKEN ACTION ON PLANT METHOD HOURS 22 SUBMITTED FORM SUB.
SUPPLIER MANUFACTURER LA.JLzJ@ LI@ [..J 10 1 010 10l
[12 NJ@
AJ IW 1 12 10 &
33 34 35 36 37 40 41 42 43 44 47 CAUSE DESCRIPTION AND CORRECTIVE ACTIONS The welds in the bore of the flywheels were installed to prevent relative I
1motion between the two lamination plates during fabrication. They were not l
designed for the service conditions of the flywheel and have been removed from the bore.
7 8
9 80 FACILITY METHOD OF STATUS
%POWER OTHER STATUS DISCOVERY DISCOVERY DESCRIPTION 1
LHJ @
10 l gI N.A l.iI Dye penetrant inspection 7
8 9
10 12 13 44 45 46 80 ACTIVITY CONTENT RELEASED OF fELEASE AMOUNT OF ACTIVITY LOCATION OF RELEASE 1
I@ IJ Z.1J31 N.A.
I IN.A.I 7
8 9
10 11 44 45 80 PERSONNEL EXPOSURES NUMBER TYPE DESCRIPTION 0O 0 10 IQ Z ]1 N.A..
7 8
9 11 12 13 80 PERSONNEL INJURIES NUMBER DESCRIPTIONS 0
d101 01I N.A.
7 8
9 11 12 80 LOSS OF OR DAMAGE TO FACILITY TYPE DESCRIPTION W L.JI N.A.
7 8
9 10 s0 PUBLICITY ISSUED DESCRIPTION @
U Z 18 N.A.
7 8
9 10 68 69 80 0
NAMEOFPREPARER J. M.Curran PRONE(714) 492-7700 IL
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