ML19326B378

AO-50-313/74-01A:on 740624,metallurgical Test of Cracked Weld on Decay Heat Removal Sys, (AO 50-313/74-02A) Revealed Fatigue as Failure Mode.Caused by Design & Installation Error.Flow Orifice Installed

ML19326B378
Person / Time
Site:	Arkansas Nuclear
Issue date:	11/27/1974
From:	Cavanaugh W, Phillips J, Rueter D ARKANSAS POWER & LIGHT CO.
To:
Shared Package
ML19326B373	List: ML19326B372 ML19326B378
References
NUDOCS 8004150751
Download: ML19326B378 (4)

Text

_ _ _ _ . - .

- 3 '

.i i

lU 1. Abnormal Occurrence Report No.

50-313/74-2A

2. Report Date: N;veder 8,1974 3.

Occurrence Date: June 26, 1974 4

Facility: Arkansas Nuclear One-Unit 1 '

o Russellville, Arkansas 5

I Identification of Occurrence: See AOR 50-313/74-2 Dated 7/2/74 .

6. Conditions Prior" to Occurrence: See AOR 50-313/74-2 Steady-State Power Reactor Power FMth 4 Hot Standby Net. Output .

I MWe Cold Shutdown Percent %f Full Power  %

Refueling Shutdown Rottine Startup Orsration i

j -

Routine Shutdown Operation .

Load Changes During Routine Power Operation .

Other (Specify)

\ 7. Description of Occurrence: .

i See AOR 50-313/74-2 '

1 i

. i

!l N

/ .

, July 5, 1974 NSP-10, Rev. O I

Page1od3 .,j.;

800415075I

~' .

I , .

\v'.

I Abnormal Occurrence Report N,o. 50-313/74-2A Sheet 2

8. Designation of Apparent Cause of Occurrence:

Design _ X Procedure Manufacture Unusual Service Condition Including Installation / X Environmental Construction

• Component Failure Operator

~

d Other (Specify) .

Metallurgical examination of the cracked weld revealed fatigue as the failure mode. Consequently, an extensive fluid dynamics analysis and testing program of the decay heat system was conducted. The results

.of this program indicated that .the crack was caused by' excessive vibra-tion in the system piping, resulting from cavitation of the pump dis-charge valves. His cavitation ranged from heavy in the pump testing mode to slight in the decay heat removal mode. . Extrapolation of test (G

\

data to the low pressure injection mode indicated that -the cavitation would be non-existent in that mode.

(Continued on page 2a)

9. Analysis of Occurrence:

'I

} See AOR 50-313/74-2

!, 1 i .

i

10. Corrective Action:

A flow orifice will be installed in each decay heat loop to eliminate -

cavitation and corresponding pipe vibration in both pump testing and l decay heat removal modes. The orifice will be by-passed during the I low pressure injection mode.

Flow orifices will also be installed in the reactor building spray pump test flow path. Rese orifices will be located in the recircu-lation line back to the borated water storage tank which is not in

• the flow path used to deliver water to the reactor building spray headers.

(Continued on page 2a)

/9 '

V .

-- l July 5, 1974 NSP l'D., Ret. O Page 2 of 3 l t -

^

l s -

/,-~k Abnormal Occurrence Report No. 50-313/74-2A Sheet 2a i )

l 4

8. Designation of Apparent Cause of Occurrence: (Continued)

A similar analysis and ' testing program was performed for the reactor building spray system, and similar cavitatien and vibration was dis-

covered in the pump testing mode. Since the

ctual reactor building spray operation is accomplished with all valves in the full open position, the cavitation and corresponding vibration problems will not exist during the safety related system op trating mode. ,

10. Corrective Action: (Continued) .

, Pending arrival of materials needed to implement these changes, the i decay heat pump testing mode will be modified to pass flow through a ,, ;

3" control valve, rather than an 8" control valve, in the non-safety f

related portion of the system. For the decay heat system, this will have the effect of moving the largest pressure drop in the system to this non-safety related piping. This will significantly reduce the cavitation and vibration in the safety related portion of the system.

e I

a f -s

)

t

--- ~ ---- - -

y m. y

• 3 g...., .

Abnormal Occurrence Report No. f0-313/74-2A Sheet 3

11. Failure Data: #

See AOR 50-313/74-2 r

I _

's

12. Reviews and Approvals:

Reviewed and Approved by: Plant Safety Cotaittee Yes ()Q No ( )

Plant Superintendent Yes O() No ( )

Reference:

J(,d 4 - /, P Y Date: ///2/[79 --

Reviewed by: _f tate: // 2 ffV Licensing rvisor

\ Approved by: p A r : =h Date:

i b nager of Nuclear Servi

/t 4

,, , 7[ $

Approved by: fff/p/ ),4- Date: /!- "

/ Safety Review Committee Approved by Me)'/ge[M-A/ Date: // f - 7 [  ;

/ Director of Power Production Approved 'ay: ,

I L- /[, c' , ' -- Date: // 1 'i-

Senior Vice Pre'ident s

]

l.

i t ..

July S, 197 NSP-10,.Rev. O l6 Page 3 of 3 -

--p, --- - . - - - - , --p.