ML20005E785
| ML20005E785 | |
| Person / Time | |
|---|---|
| Site: | Braidwood  |
| Issue date: | 12/29/1989 |
| From: | James H, Querio R COMMONWEALTH EDISON CO. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| BW-89-3230, LER-89-016-05, NUDOCS 9001110063 | |
| Download: ML20005E785 (9) | |
Text
-
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l Commonwealth ECson f'
' Braidwood Nucbir Pr nr St ti:n s
- Roote C1, Box 84 .
. Braceville, Illinois 80407.. -
.v*'
- Telephone 815/458-2801 -
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December 29,1989 )
BW/89-3230 n
U. S. Nuclear Regulatory. Commission :
Document Control Desk 1 Washington, D.C. 20555 '
Dear Sir:
-e . The enclosed Licensee Event Report from Braidwood Generating -
Station is being transmitted to you in accordance with the requirements of 10CFR50.73(a)(2)(1) which requires a 30-day written repart. ,
y This report is number 89-016-00; Docket No. 50-456.
Very truly yours, R. E. Querto Station Manager Braidwood Nuclear Station REQ /JDW/jfe (7126z)
Enclosure:
Licensee Event Report No. 89-016-00
/_
l l cc: NRC RegionIII Administrator NRC Resident Inspector INPO Record Center CECO Distribution List hl L
1 3 691229 h00AAh000xosooo436 s
PDC (p& t
/ \ \
LICENSEE EVENT REPORT (LER)
[ Form Rev 2.0 F cility Name (1);
- D;cket NumbTr (2) Paoe (3) sraldwood 1* ! 01 51 01 01 01 41'51 6 1!ofl0l8 Title (4) Residual Heat Removal Pump Suction Relief Valve Premature Actuation and Failure to Reseat Due to i
Deficient Work Practices and Personnel Error,
_,1 vent Date ($1 LER Number (6) Report Date (7) Other Facilities Involved (8)
Month Day Year Year /// Sequential /j//j Revision Month Day Year Facility Names Docket Number (s) j/p/j
/ Number /j// Number l
None 01510101Ol l l
~
01 51 01 01 Ol l l
_lj 2 01 1 81 9 81 9 Ol116 01 0 1l2 21 9 81 9 I I
THl$ REPORT IS SUBMITTED PURSUANT TO THE REQUIREMENTS OF 10CFR OPGM " * " " * "
MODE (9) 73.7)(b) 5 20.402(b) _ 20.405(c) _ 50.73(a)(2)(iv) _
POWER _ 20.405(a)(1)(1) _ 50.36(c)(1)' _ 50.73(a)(2)(v) _ 73.71(c)
LEVEL _ 20.405(a)(1)(ii) _ 50.36(c)(2) _ 50.73(a)(2)(vil) _
Other (Specify (10) 0!0!O _ 20.405(a)(1)(lit) X_ 50.73(a)(2)(1) _ 50.73(a)(2)(viii)(A) in Abstract
/ /,/ / / /,/ / / /,// / / / /,/ / /,// / / /,/ /, _ 20.405(a)(1)(iv) _ 50.73(a)(2)(ii) _ 50.73(a)(2)(viii)(B) below and in
/}////' / ///}/////j//}/////'
/ / / /' /
j _ 20.405(a)(1)(v) _ 50.73(a)(2)(lii) _ 50.73(a)(2)(x) Text)
LICENSEE CONTACT FOR THIS LER (12)
Name TELEPHONE NUteER Howard James, Technical Staff Engineer 2482 5 415181 l218101 COMPLETE ONE LINE FOR EACH COM ON FAILURE DESCRIBED IN THIS REPORT (13)
CAUSE SYSTEM COMPONENT MANUFAC- REPORTABLE CAUSE SYSTEM COMPONENT MANUFAC-- REPORTABLE TURER TO NPRDS TURER TO NPPDS X Bl P Rl VI l Cl71110 YES l l l l l l l A Bl P Rl Vi l Cl71110 YES l l l l l l l SUPPLEMENTAL REPORT EXPECTED (14) Expected ' Month I Day l Year Submission
- (
lyes (If ves. complete EXPECTED SUBMISSION DATE) X l NO l ll l ABSTRACT (Limit to 1400 spaces, i.e, approximately fif teen single-space typewritten lines) (16)
At 0142 on December 1,1989 while drawing a bubble in the Pressurizer (PZR), Reactor Coolant System (RCS) prsssure slowly increased from 375 psig to 404 psig. At this time the IB Residual Heat Removal (RH) pump suction relief valve, which had a setpoint of 450 psig, actuated and remained open. Charging flow was increased but PZR level indicated 0% by 0151. Reactor Operators concluded that an RH pump suction relief valve had lif ted because Hold Up Tank levels were increasing rapidly. The operating train of RH, 1A, was isolated at 0155. At 0200 RCS pr:ssure reached 272 psig and stabilized. RCS level was at the lower portion of the PZR surge line and flow into the RCS was equal to the flow exiting the RCS. At 0215 the Licensed supervisors decided to return the second Charging pump to service per 10CFR50.54(x). At 0235 the second charging pump was started. At 0245 PZR level was above 0%. At 0319 field reports identified that the IB RH pump suction relief had actuated. At 0350 the B RH Train was isolated which terminated the event. Approximately 64,000 gallons had relieved through the valve. The i
cause of the early lift was dirt between the valve spindle and guide sleeve which affected valve lift setpoint
! cdjustment. The cause for the valve remaining open was an incorrect nozzle ring setting due to a personnel j orror. Maintenance procedures will be reviewed. Training will be conducted. No previous occurrences.
2966m(010290)/2 1
ry ., e
~LIc mtrr EVENT REPORT (LER) TEXT CONTINllATION' - Fem Rev' 2.0 h.v
'E LER NLDSER (6): Pane (3)
(FACILITY NANE-(1) . DOCKET NUPSER (2)-
l N
W Year = /// = Sequential // Revision
+
fff
/// Number
/j/j f
// Number i
- Braiod
- f'1- 0l5101010141516 8l9 - Ol116 - 0 l' O 01 2 0F Dl'8 -
l TEXT- Energy Industry Identification System (EIIS) codes are identified in the text as (XX) >
l
!A.L PLANT CONDITIONS PRIOR TO EVENT: ,
Unit: Braidwood 1; ~ Event Date: December 1, 1989; Event Time:. 0142;-
INode:,5-ColdShutdown;. Rx Power: 0%;
'RCS (AB) Temperature / Pressure: 170. degrees F/350 psig
.B. DESCRIPTION OF EVENT: ;
P.;
- There were no systems or components inoperable at the beginning of the event which contributed to the severity of '
lthe event. ;
Y :The Pressurizer.(PZR) was water solid. The B and D Reacter Coolant Pumps were in operation and both trains of ~
.g.
-Residual Heat Removal (RHR)~[BP) were aligned to the RCS for Shutdown Cooling with the A loop in operation.- ';
iLetdown to the Chemical Volume and Control System (CV) (CB) was being supplied from the lA RHR system.
At 0055 on December 1,1989 in preparation for drawing a bubble in the PZR,_ the Nuclear Station ' Operator (NS0)-
(Licensed Reactor Operator) increased letdown flow from 97 to 110 gpm and energized all 3 PRZ Backup Heaters. l This was in accordance with Operating Procedure Bw0P RY-5, Drawing a Bubble in the Pressurizer. P2R pressure and-- .
. temperature were approximately 350 psig and 404 degrees F at the time.
- At 0126 the NSO observed that RCS pressure had risen to 395 psig. To stabilize RCS pressure the NSO performed the *
~following: _
- l. Increased Letdown flow. ,
' 2. Decreased CV charging flow to minimum.
- 3. Deenergized 2 of the PZR Backup Heaters.
l Dy 0139 the NSO had increased letdown flow to the maximum value, however, RCS pressure was still slowly increasing.
l At 0142 with the RCS pressure at 404 psig, the IB RHR Pump suction relief valve actuated. This corresponded to approximately 410 psig at the elevation of the IB RHR Pump suction relief valve. This was 46 psig below its ,l setpoint of 450 psig +/- 1%. '!
At 0144 the NSO observed that the PZR level had returned to scale from an off scale high condition. The level was !
decreasing at a rapid rate. The NSO began reducing letdown flow in an attempt to stabilize PZR level.
At 0149 the NSO increased charging in an attempt to maintain pressurizer level. The following actions were also taken at approximately the same time:
The Shif t Engineer (SE) (Senior Reactor Operator) (SRO) and the Shif t Foreman (SF)($RO) were notified and ll Immediately reported to the Control Room. The SF assumed the SRO direct supervisory responsibilities upon l- being briefed of the situation by the Station Control Room Engineer (SCRE) (SRO) and the NSO.
The NS0s monitored control room indicators that would be indicative of a leak inside the containment. There was no indication of leakage inside the containment.
2966m(010290)/3 a _ _. _ _ _ _ _ _ _ _ __ __ ___ _ _ . _ _ _ _ ___ ___ ___ .___ _ _
F LICENSEE EVENT REPORT (LERI TEXT CONTINUATION ' Fom Rev 2.0 FACILITY NANE (1)
.. DOCKET NUNBER (2) LER NUtBER (6) Paoe (3) -
Year /// Sequential /jj/j
/ Revision
- //j/j f
Number /// Number j
.Itaidwaod _1 015 10 l Q l 014LsLn B 19 - 01iis - 01 0 01 3 Or_ _o t e
. TEXT Energy Industry Identification System (ElIS) codes =are identified in the text as (XX)
The NSO directed the available Equipment Attendants ('EA) (non-licensed operators) to check the areas of the Auxillary Building for signs of Unit 1 CV or RHR system leakage. ,
The NSO contacted the Radwaste Control Room Operator (RWCRO) (non-licensed operator) and requested a.
check of Radweste indications for signs of an increase in the input to Radwaste. The RWCR0 reported that both Recycle Hold Up Tanks (HUT) (GA) were increasing at a rapid rate.
Based on this information, the Control Room Personnel concluded that one or both of the RHR Pump suction-relief valves had lifted and not reset.
At 0151 the NSO observed that PZR level indicated 0%.
At 0153 the NSO increased CV charging to maximum for the 1A CV Pump and isolated letdown flow. The NS0s and the SF decided to isolate the 1A RHR Train in an attempt to isolate the leak. This was based on the assumption that a system perturbation would most probably have occurred in the operating loop.
.At 0155 the NSO shutdown and isolated the 1A RHR train and placed the IB RHR train in operation. . The NSO ;
opened the CV pump suction isolation valve from the Refueling Water Storage Tank (RWST) (BR/BQ) and closed t the outlet valves from the Volume Control Tank (VCT).
At 0200 the NSO that had been monitoring the RCS presture reduction since the start of the event, observed that the No. I seal differential pressure for the IB Reactor Coolant Pump (RCP) was approaching the minimum permissible value of 200 psid. The NSO shutdown the IB RCP at this time. Shortly after the pump was shut down RCS Pressure reached 272 psig. The NS0s and SF believed that the increase in the RCS pressure reduction ,
rate that occurred about this time was the result of the RCP shutdown..' The NS0s and SF observed that the No 1 Seal differentia 1' pressure for the ID RCP was above 200 psid. They decided to keep the ID RCP in operation as long as the No. 1 Seal differential pressure was 200 psid or greater. They believed that this would prevent another increase in RCS pressure reduction rate. It would also provide PZR spray flow capability when PZR level was recovered. 3 The RCS pressure stabilized at 272 psig. This was the lowest pressure achieved in the RCS during the event.
Flow into the RCS from the 1A CV pump was equal to the amount of flow exiting the RCS. The recovery portion of the event started at this point.
At approximately 0202, an NSO directed an EA who had been checking for signs of leakage in the Auxillary .
Building to-isolate the IRH8734A, Nanual Isolation Valve from the Discharge of the 1A RH Pump to the CV Letdown Header. This valve was located in the Containment Penetration area of the Auxiliary Building on the 364' elevation.
At approximately 0205 the EA isolated the IRH8734A. The EA identified a puddle on the floor and water ,
dripping down from a cable pan. The EA determined that this water came from a relief valve located above the cable pan that was dripping a small amount of water at this time. The EA was not aware of the valve's actual identity or functlon. The EA conveyed the information to the Unit 2 SF who was on the 364' elevation outside the penetration area. The Unit 2 SF had been assisting with the in plant leakage search for Unit 1. The 1 Unit 2 SF conveyed this information to the Control Room. During the information transfer from the EA to the l Unit 2 SF, and the Unit 2 SF to the NS0s and SF in the control room, this relief valve was perceived to be the IRH8708A, the 1A RHR Pump suction relief valve. This perception created the mind set that the IRH8708A l was the valve that had actuated.
2966m(010290)/4
1 f
LICENSEE EVENT REPORT (LER) TEXT CONTINUATION Form Rev 2.0
.. . . DOCKET NUPSER (2) LER NUPBER (6) Pace (3)
FACILITYNAME(1):
Year ///
fff Sequential /jf//
f Revision
. /// Number /// Number Braidwood 1 0 l 5 1 0 1 0 1 0 l 41 51 6 8l9 . Ol116 - 01 0 01 4 0F 01 8 TEXT Energy Industry Identification System (E!IS) codes are identified in the text as [XX)
At 0215 the SRO Licensed Shift Supervisors, the SF and SE,' decided to return the IB CV pump to service. This decision was mads pursuant to the requirements of 10CFR50.54(x). The Technical Specifications specify that a maximum of 1 Centrifugal Charging Pump shall be operable in Mode 5. The SRO Supervisors viewed this as a necessary action to protect the health and safety of the public. The 1B CV pump was readily available by racking in its breaker. The PRZ level had been indicating 0% for 24 minutes. There were no other pumps readily evallable to perform the required function.
At 0227 an Alert was was classified as per Emergency Ac'. ton Level 2.n. of Braidwood Emergency / Implementing Procedures (Bw2P), " Primary System Leakage is beyond the makeup capabilities of available charging pumps".
At 0235 the NSO Isolated the 1 RH8701B, the second of the in-series RHR suction isolation valves for. the IA RHR pump. This was performed due to suspected leakage of the downstream isolation valve, IRH8701A, This suspicion was generated by the mindset created by the misidentification of the leaking relief valve as the IRH8708A combined with reports from the RWCR0 that the rate of increase in the HUTS was decreasing. The NSO started 1B CV Pump at this time.
At 0237 the Nuclear Accident Reporting System (NARS) notification was made to Illinois State agencies to declare the Alert.
At 0245.the NSO observed that the PZR Cold Calibrated (Cold Cal) Level Channel had increased above 0%. The NSO shutdown the IB-CV Pump. This was due to a concern about thermal shock to the PZR. RCS pressure was 310 psig. The RWCR0 reported that HUT levels were still increasing.
At 0254 the NSO restarted the IB CV pump. This was due to the PZR Cold Cal Level decreasing below 0% and the RCS pressure decreasing to 301 psig.
i l At 0302 the NSO observed that PZR Cold Cal I,evel indication was above 0%. The NSO reduced CV pump flow by throttilng the ICV 121, CV Charging Header Flow Control Valve.
- j. At 0312 an NSO instructed an EA who was on the 364' elevation, to proceed to the IRH87088, IB RHR Pump
- Suction Relief Valve, and determine if it had actuated.
At 0319 the EA and the Unit 2 SF identified that the IRH8708A did not appear to be open based on the absence of noise at the valve. They also identified that the IRH87088 was open. This was based on the loud screaming noise that could be heard.
At approximately 0322 the NSO directed an EA,who had been dispatched to the 364' area, to reopen and then close the IRH8734A. This was performed to provide a flowpath to depressurize the isolated 1A RH train.
Control Room personnel believed that this would ensure that the IRH8708A had reseated.
At 0326 the appropriate ENS notification was made pursuant to 10CFR50.72(a)(1) and 50.72(b)(1)(1)(B).
At 0335 the Unit 2 SF identified the relief valve that was spraying water as the OAB8634, a relief valve connected to the common header where the RHR suction relief valves discharged.
At 0345 an EA was stationed by the IRH8708A to observe for signs of valve actuation when the 1A RHR train was realigned to the RCS.
2966m(010290)/5
F 4 s
^
LICENSEE EVENT REPORT (LER) TEXT-CONTINUATION Form Rev 2.0 LER NUNBER (6) Paoe (31
- FACILITV NAE (1). '
. . DOCKET NUPSER (2)
Year Sequential // Revision j ,
///
f77
/// . Number
/j/
f7
// Number
.p
- i. Brali. :f I ' 0 l 51010 l 0141516 8l9 - 0l116 - 0 lO 01 5 0F 01 8 4 ' TEXT; -Energy Industry Identification System (EIIS) codes are identified in the text as [XX)
At 0346_the NSO. opened the 1A RHR train suction isolation' valves, 1RH8701A and B.
At 0348 the EA verified that the IRH8708A had not actuated.
At 0349 the NSO placed the 1A RHR train in operation. The EA by the suction relief valve verified that it was-not lifting.
At 0350 the NSO shut down and isolated the IB RHR. Train. The NSO observed that PRZ Level had started to increase significantly.
At 0353 the NSO shutdown the IB CV Pump and stable plant conditions were established.
At 0427 SE transferred Command and Control to the Technical Support Center.
r At 0435 the event was terminated.
At 0445 the appropriate NARS notification was made to terminate the event.
'At 0512 the appropriate NRC notification via the ENS phone system was made pursuant to 10CFR50.72(c)(1)(iii) to terminate the event.
A review of the data collected during the event has identified the following:
- 1. The total water volume passed through the IB RHR Pump suction relief valve was approximately 64,000 gallons.
- 3. The relief valve remained open approximately 111 minutes.
- 4. The' lowest level the system attained is believed to be the lower portion of the Pressurizer Surge g
Line, Reactor Vessel Level indicated 100% throughout the entire event.
- 5. The PZR Liquid Space and Surge Line Temperature Indicators experienced an indicated cool down in excess of 200 degrees F in a one hour period. This occurred when PZR level was re-established during the event. The PZR Surge Line temperature indicator decreased from 374 degrees F at 0254 to 164 degrees F at 0354 The PZR Liquid Space temperature indicator decresced from 444 degrees F at 0143 to 174 degrees F at 0243. These were normally expected occurrences for an event of this nature. PZR temperature was approximately 440 degrees F and RCS temperature was approximately 170 degrees F at the start of the event.
E On December 02, 1989 the Onsite Review of the engineering evaluation to determine the effects of the cooldown on the PZR was completed. The evaluation concluded that the structural integrity of the PZR was acceptable for continued operation.
- This event was evaluated by an NRC Augmented Inspection Team and an event investigation team from INP0.
l l
l 2966m(010290)/6
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v g 4 E
LICENSEE EVENT REPORT (LER) TEXT' CONTINUATI0h' Form Rev 2.0_
k LER NUpeER (6) Pace (3) ,
U ,FACIlll;YNAME(1)7 -
. . DOCKET NUPSER (2)'.
' Year /// Sequential /// . Revision f77 fff
. /// Number /// Number _
'O l 5101010 l 41516 Bl9 0l116 01 0 01 6 0F '01 8 k" ~ Braidwood_J 3
TEXT.. Energy Industry. Identification System (Ells) codes are identified in the text as (XX)
-This event is being reported pursuant to 10CFR50.73(a)(2)(1): -i
- Any operation or condition prohibited by the plants Technical Specifications.
- Any deviation from the plants Technical Specifications ' authorized pursuant to 10CFR50.54(x).
C.' Cause of Event:
The root cause of this event was the premature actuation of the relief valve combined with an incorrectly set; nozzle ring.
.The 1RHB7088 valve was systematically removed, tested, and disassembled for root cause analysis to. determine failure mode. A representative of the Manufacturer, Crosby Valve and Gauge Company, participated in the disassembly and analysis at the request of Braidwood Station. A summary of the investigation results is detailed below.
The Valve was bench tested upon removal. The valve lif ted at 411 psig, 407 psig and 405 psig on three successive lift tests.
The nozzle ring setting was checked. The nozzle ring was found to be out of adjustment high by 233 notches.- >
With the Nozzle ring at this setting the valve would not reseat until pressure was significantly _below the lift setpoint. This mis-setting occurred when the Nozzle ring was set during maintenance prior to 3 installation in April of 1988. The cause of this mis-setting was a failure to follow procedure. The Mechanical Maintenance personnel (HM) (non-licensed maintenance personnel) who adjusted the nozzle ring set . '
the ring 110 notches from the " upper locked" position instead of 343 notches as required by procedurs BwMP -
3305-072, Disassembly-Reconditioning-Reassembly of Crosby Style JB-TD-WR.("L" Orifice) Relief Valve. The procedure requires the nozzle ring to be taken from the "as found" position to the upper locked position during disassembly. The number of notches are counted and document in step F.2.c.1). This step was
- performed and the number of notches was documented as 343. During reassembly the MM was directed to take the ring to the upper locked position and lower it the number of notches recorded in step F.2.c.1). For this event the Ptf took the final position setting from a Station Traveler which provided the Manufacturers setting of 110 notches below "zero position". Zero position was not the upper locked position but the position where the nozzle ring was level with the valve seat. This caused the "as left" nozzle ring setting to be 110 notches below the upper locked position instead of 110 notches below the zero position which was 233 notches ,
lower.
The nozzle ring setting error was corrected and the valve was bench tested to determine if*the improper setting had affected the lift setpoint. The valve lifted at 400 psi, essentially the same lift value as the incorrect setting.
The valve lift setpoint was systematically adjusted with bench tests performed after each adjustment. The valve responded in the correct manner after each adjustment for both increasing and decreasing setpoint adjustments.
The valve was disassembled, decontaminated and inspected. Grooves were found on both the spindle and the spindle guide. It is suspected that a piece of sand grit or metal grinding became lodged between the spindle and the spindle guide during maintenance performed on the valve in April of 1988. As the adjusting bolt was turned to set the lift setpoint the spindle rotated. With the foreign material in the gap between the spindle and spindle guide the adjusting process made the groove circumferential. During a valve lif t, presumably the lif t setpoint verification prior to installation, the foreign material became dislodged. The vendor representative stated that the increased fric+. ion from a piece of foreign material between the spindle and guide would account for the 45 psi disparity between the as left setpoint value from the maintenance test performed in April of 1980 and the actual valve lift setpoint experienced on December 1,1989.
2966m(010290)/7
1 LICENSEE EVENT REPORT (LER) TEXT CONTINUATION Form _Rev 2.0
- FACILITY NAME (1)
DOCKET NUMBER (2); LER NUteER (6) Pace'(3)
Sequential Revision
' Year // /
j/j/j
// Number j///j/j
/ Number
,JrejdEDod1 0l51010l0141516 819 - 0l116 - 01 0 Ol 7 0F_ 01 8 TEXT Energy Industry identificatf u 'ystem (EIIS) codes are identified in the' text as (XX)
D. Safety Analysis:
.This event had no effect on the Safety of the Plant or the public. After the initial level decrease, the water level appears to have remained in the lower portion of the surge line until it was recovered. The reactor had been in an extended refuel outage. A significant potential for a temperature increase from decay heat did not exist.
' Under the worst case condition of a failed open RH relief valve occurring during initial system alignment i after extended Unit Operation there would still be no effect as this is enveloped in Section 5.4 of the Updated final Safety Analysis Repor'.. The design of the RH system provides- for complete isolation of an RH loop using redundant isolation vsives. Adequate leakage dection capability is provided by Radwaste input instrumentation. And ph % cal train separation provides the capability to operate indefinitely with one train isolated.
E. Corrective Actions:
Immediate corrective actions:
- 1. Letdown was isolated and charging was increased to maximum.
- 2. The RH trains were isolated in a systematic manner to stop the leakage.
- 3. Stable plant conaltions were established.
- 4. The faulty valve was removed and replaced with one from stores.
- 5. The area that was sprayed with water from the the AB relief valve was cleaned up arid decontaminated.
- 6. An Engineering evaluation was conducted on the effects of the indicated cooldown rates on the PZR.
Based on the results of this evaluation it has been concluded that the effects were insignificant and continued operation is acceptable,
- 7. The documentation for valve lif t and reset settings for the the 1A , 2A, and 2B RHR Pump suction relief valves has been reviewed. Base on this review, it has been concluded that these three relief valves are set correctly.
- 8. The valve that was removed from the 1B RH pump suction line was rebuilt, tested and installed on the 1A RH pump suction line. This was performed to facilitate bench testing of the 1A RH pump suction relief valve. This was done in response to a Recommendation from the NRC Augmented Inspection Team. The nozzle ring setting was correct. The valve actuated at 465 psig on the first test. 5 psi above the acceptance criterla of 455 psig +/- 1%. The valve setting was adjusted and the valve was returned to stores.
Additional Corrective Actions:
- 1. Maintenance procedures for Crosby relief valves will be reviewed to ensure that instructions are clear and adequate. This will be tracked to completion by action item 456-200-89-19301.
2966m(010290)/8
c:
l'
-t LICENSEE EVENT REPORT (LER) TEXT CONTINUATION ' Form Rev 2.Q_ '
- FA IL(TY NAME (1)
DOCKET NtMIER (2) , LER NtMBER (6) _ _ Paae (3) __
Year Sequential ///, Revision $
. /j//j f ff
- /// Number /// ' Number Braldeced 1 0 l 5-l 0 1 0 l 0 1 41 51 6 819 - 0l116 - 0l 0 Of 8 0F 01 8 TEXf Energy Industry Identification System (EIIS) codes are identified in the text as [XV.) .;
- 2. This event will be covered with Licensed personnel as part of the requalification program. This action will be tracked to completion by action item 456-200-89-19302.
p
' 3. . This event will be covered with appropriate Haintenance personnel as part of a Training Tallgate session. This session will stress the following:
8- . >
- a. Proper verification of reference point when adjusting blowdown rings on relief valves.
~
- b. Ensuring a clean environment is maintained during assembly / disassembly of components.
- c. Proper techniques for determination of lif t setpoints for relief valves.
U This action will be tracked to completion by action item 456-200-89-19303.
4 Braidwood Station will request that the Westinghouse Owners Group develop a Mode 4/5 LOCA Procedure ,
citing this event as an example. This action will be tracked to completion by action item 456-200-89-19304.
F. Previous Occurrences:
'There have been no similar previous occurrences.
G. Component Failure Data:
Manufacturer Nomenclature Model Number ,
Crosby Valve and Cauge Co Relief valve JB-35-TD-WR 2966m(010290)/9