05000259/LER-1982-020-03, During Normal Operation, Unidentified Drywell Leakage Increased & Exceeded Limits. Caused by vibration-induced Fatigue Crack in Reactor Water Cleanup Sys Test Connection.Line Rerouted: Difference between revisions

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{{Adams
#REDIRECT [[05000259/LER-1982-020-03, Unidentified Drywell Leakage Increased in Excess of Tech Spec Limits.Caused by Vibration Induced Fatigue Crack in 3/4-inch Reactor Water Cleanup Sys Test Connection.Line Rerouted]]
| number = ML20027B901
| issue date = 09/27/1982
| title = During Normal Operation, Unidentified Drywell Leakage Increased & Exceeded Limits. Caused by vibration-induced Fatigue Crack in Reactor Water Cleanup Sys Test Connection.Line Rerouted
| author name = Roberts W
| author affiliation = TENNESSEE VALLEY AUTHORITY
| addressee name =
| addressee affiliation = NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II)
| docket = 05000259
| license number =
| contact person =
| document report number = LER-82-020-03X, LER-82-20-3X, NUDOCS 8209300291
| package number = ML20027B899
| document type = LICENSEE EVENT REPORT (SEE ALSO AO RO), TEXT-SAFETY REPORT
| page count = 3
}}
{{LER
| Title = During Normal Operation, Unidentified Drywell Leakage Increased & Exceeded Limits. Caused by vibration-induced Fatigue Crack in Reactor Water Cleanup Sys Test Connection.Line Rerouted
| Plant =
| Reporting criterion =
| Power level =
| Mode =
| Docket = 05000259
| LER year = 1982
| LER number = 20
| LER revision = 3
| Event date =
| Report date =
| ENS =
| abstract =
}}
 
=text=
{{#Wiki_filter:NRC FORM 36S U. S. NUCLE AP REGULATORY COMMISSION 87 M UPDATE REPORT, LICENSEE EVENT REPORT PREVIgDS RLFORT DATE 7/22/82
. CONT'ROL BLOCK: l l
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lh (PLEASE PRINT OR TYPE ALL BrCJIRED INFORMATION) 1 6
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7 8 9 LICENSEE CODE 14 15 LICENSE NUMUER 25 26 LICENSE TYPE JO 67 CAT $8 CON'T l0l1l
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60 61 DOCK ET NUMB ER 68 63 EVENT DATE 74 75 REPORT DATE 80 EVENT DESCRIPTION AND PROBABLE CONSEQUENCES h [O 7 71-1 Durine normal operation. unidentified drvve11 leakane as determined bv flow l
lO#3l ]
recorder and flow' integrator suddenly increased and exceeded limits specified 101. l l in T.S.
3.6.C.1.
In accordance with T.S.
3.6.C.3 an orderly shutdown was I
10151I initiated and the reactor was in cold shutdown condition within ~ 24 hours.
There I
10 ls l l was no dancer to the health or enfety of the nublic.
The leakane never l
l0l7ll approached normal makeup canacity.
All enercency systems were available and l
l O 181 I operable.
I 7
8 9 80 SYSTEM
 
==CAUSE==
CAUSE COMP.
VA LVE COCE COC-E SUBCODE COMPONENT CODE SUSCODE SUBCODE 10191 l c lc l@ Ln_]@ l ^ 1@ I P l T l P l E l X l X lh [A_lh l Zl h 7
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to it 12 13 18 19 10 SEOUENTIAL OCCURRENCE REPORT REVISION L
EVENT YEAR REPORT NO.
CODE TYPE NO.
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CAUSE OESCRIPTION AND CORRECTIVE ACTIONS h.
I 1 I O I l Leaknee resulted fron a vibration-induced f aticue crack in a 3/4-inch roactor water I
IiiilI cleanun system test connection.
Inndecuate brncini of the test connection line 1
l i 121 l allowed high-frecuency vibration.
The line was rerouted to decrease suscentibility l
IiI3l l to vibration.
Insp$ tion was completed for similar conditions on unit 3.
Unit 2 1
4 I i ( 4 I l will be inspected durine the evcle 4 refueline outace.
I J
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80 STA S
% POWER OTHER STATUS DSO RY DISCOVERY DESCRIPTION l l 5 l [_gj@ l 0 l 9 l 5 l@l NA l
lAl@l Operator observation l
COSTENT i2 n
44 4s 40 80 ACnviTY RELEAcEO OF RELE ASE AMOUNT OF ACTIVITY LOCATION OF RELE ASE l l6l[_Z)h,[Zj@l NA l
l NA l
PERSONNEL EXPOSU ES DESCRIPTION @
NUMPER TYPE l i i il 10 l 0 10 l@l2 l@l NA l
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PDR ADOCK 05000259 iv I i l 9 l [Z_J@l 6
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NRC USE ONLY l
TVA nress release - 1/,n/99 l
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7 8 9 13 68 69 80 3 William A. Roberts (205) 729-0788 NAME OF PREPARER PHONE:-
 
* ' Tennessee Valley Authority Form BF 17 r-3-
Browns Ferry Nuclear Plant BF 15.2 j
2/17/82 L$R SUPPLEMENTAL INFORMATION
{
BFRO _ 259 /_3LO_20 R2 Technical Specification Involved 3.6.c.1 & 3.6.c.3 i
Reported Under Technical Specification _6.7 2.b.(2)* Date Due NRC 5
Event Narrative:
f Unit 1 - 1,045 We (steady state)
Unit 2 - 1,085 We (preconditioning up to full load)
Unit 3 - 0 W e (refuel outage)
During nonnal operation, unidentified drywell leakage as determined by the floor drain sump pump flow from FR-77-6 and flow integrator FQ-77-6 suddenly increased and exceeded limits specified in Technical Specification 3.6.c.1.
In accordance with Technical Specificaticn 3 6.C.3, an orderly shutdown was initiated and the reactor was in cold shutdown within 24 hours. There was no danger to the health and safety of the public. All emergency systems were available and operable. An inspection identified the leakage as being from a crack in a 3/4-inch test connection which ties into the 6-inch pipe adjacent to reactor water cleanup I
i system valve FCV-69-1. 2 1s test line and its associated manual isolation valves j
(69-583 and 69-584) were inadequately supported which led to high-frequency fatigue and subsequent cracking of the line.
Repair was made by removing a 90-degree elbow and shortening the line thus orientating the test valves in vertical plane perpendicular to the 6-inch line instead of being orientated in a horizontal plane parallel to the 6-inch pipe. W e crack in the 3/4-inch pipe occurred above a socket weld and extended around approximately 60 percent of the circumference. The broken test line was repaired.
l.
(see attached page)
I
* Previous Similar Events:
NONE Retention:
?criod - Lifetime; Responsibility - Document Control Supervisor
((
* Revision :
~
 
1 2-LER SUPPLEMENTAL INFORMATION BFRO-50-259/82020 R2 The average ficw rate'of water from the cracked line was approximately 12 gallons per minute. The total amount or water lost during the event was approximately 17,200 gallons. These values were derived frcm the actual ficw rate as recorded on strip chart FR-77-6 for a 24-hour period on March 20, 1982.
During the inspection to locate the leak, it was noted that this vent connecticn had previously been provided with a vibration support. Support requirements for this vent connection were not snown on TVA drawings. The support had apparently been removed at some time for maintenance or modificaticn work. Reinstallatien had not been performed because vibration support requirements were not shown on 3
design drawings. The fatigue crack occurred because the support had not been reinstalled on the line.
An evaluation of any suspect units 1 and 2 safety-related piping system will be conducted to verify that similar problems do not. exist on other test, vent, or drain connections. If any discrepancies are found, appropriate corrective acticn will be taken. This inspection will be performed during the unit 1, cycle 5 and unit 2-cycle 4 refueling outages presently scheduled for completion July 17, 1983, ana January 30, 1983, respectively.
t t
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4
~ -,. -.
}}
 
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