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| ATTACHMENT 5-m -
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| Commonwealth Edson s
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| Production ~
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| Training Department Braidwood
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| V MALFUN'CTIONS AND k , I N I T' I A L CONDITIONS 4
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| l?7 :BRAIDWOOD SIMULATOR
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| , . MALFUNCTION LISTING AUXILIARY SYSTEMS Iij-.
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| l Aux-1 Loss of Instrument Air
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| . Aux-2 RWST Leak Aux Cire Water Pump Trip Aux-4 Essential Service Water Pump Trip
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| ; Aux-12. Loss of Service Air l
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| Aux-13. Loss of Non-Essential Service Watdr
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| ' Aux-17 ' Inadvertent' Control Room Ventilation Isolation-Aux-18 Non-Essential Service Water Header Leak 1
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| O 1 638M/263M/1 8/87
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| (
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| ~BRAIDWOOD SIMULATOR MALFUNCTION
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| | |
| ==Title:==
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| Loss of Instrument Air ID: AUK-1 1
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| N0: 6.3.4.1.1
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| :n-
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| == Description:==
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| Rupture in various air headers causing loss of IA pressure.
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| (3 ;
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| }
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| Variations: Date: 3/12/89 q Rev: 5 1 l
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| 1 Selectable Steps Inputs Comments
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| : 1. Select location SA, TURB, AUK, FUEL, CONT, NRCV
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| : 2. Select leak rate 0-1500 SCFS SA: Supply hdr from SA receiver TURB: U-1 TB hdr.
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| AUK: U-1 Aux Bids hdr.
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| FUEL: Fuel Handling Bidg and U-2 Aux Blds.
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| CONT: U , Containment IA hdr.
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| NRCV: IA 2" supply to Non- !
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| return Check Valves, j Brief Plant Response Section 1 - Causes low pressure in station air system later followed by low presourc in IA system since supply to IA is lost.
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| Can be isolated by LOA's.
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| 3 0 305M/83M/2 5/89
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| ---m_.___ - _ - _ - _ _ _ _ _ _ _ _ _ . - _ -
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| -. a .- ..m=-.a--- ------;-w--.---c---
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| BRAIDWOOD SIMULATOR MALFUNCTION
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| | |
| ==Title:==
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| Loss of Instrument Air ID: AUX-1 Brief Plant Response: (continued)
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| Section 2 - Causes IA low pressure, this section supplies steam dumps, FW,
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| 'HD, and ES valves and MS. Can cause a reactor trip before
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| ' operator action of LOA's can be administered (if leak rate is large).
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| l l
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| Section 3 - Causes IA hdr low pressure alarm depending on how fast the rest of the system can sense rupture. On low IA pressure, CVCS 1stdown stop= due to closure of CV-8152 and various other valves in the flowpath. If operating PDF, charging flow goes to 52 spa.
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| and you still have seal water return.
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| If LOA's isolate hdr, you will receive pressurizer level deviations.
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| Section 4 - Same as 3, except it can be isolated by LOA's and maintain plant operations on U-1.
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| Section 5 - Loss of IA in containment will cause a los af pressurizer spray and a loss of CVCS letdown .nd excess letdown. If rupture is in containment, it can be isolas.ed from the control room on the IPM11J panel. Could receive pressurizer pressure and level deviations.
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| Sectica 6 - Loss of IA to this section would cause IA bdr low pressure alarm and allow the HD and ES non-returnable check valves to shut.
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| O This will cause a loss of efficiency oir secondary cycle.
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| Can be isolated by L.'A's and plant operations continued at lower
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| -power-level.
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| 305M/83M/3 5/89
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| Titlo Leco of Instrument Air ID: AUX-1 Brief Plant Response: (continued) h LOA's - Local Operator Actions
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| : 1. Close various manual valves in IA system (LOA Aux 15) a. ISA113 (No.1 Instrument Air Supply Line)(RSAV113).
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| (LOA Aux 2) b. IIA 013 (No.1 Receiver Outlet)(RIAV013).
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| (LOA Aux 1) c. IIA 001 (No.1 Instrument Air Dryer Supply Line).(RIAV001)
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| (LOA Aux 12) d. DIA114 (U-1 TB ring hdr x-tie)(RIAV114).
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| (LOA A u 5) e. DIA055 (TB ring hdr x-tie)(RIAV055).
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| (LOA Aux 7) f. DIA097A (U-1 TB IA ring hdr supply to Aux Bldg)(RIAV097A).
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| (LOA Aux 6) 3 IIA 060 (U-1 IA Supply to HD and ES valves)(RIAV060).
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| (LOA Aux 10) h. DIA107 (Aux Blds x-tie)(RIAV107).
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| (LOA Aux 9) 1. DIA101 (Aux Blds x-tie)(RIAV 101).
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| (LOA Aux 13) j. IIA 147 (Instrument Supply Header x-tie)(RIAV 147)
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| (LOA Aux 8) k. OIA978 (U-2 TB IA ring header supply to Aux Bids)(RIAV 97B). .
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| Suggested Instructor Action: ,
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| None.
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| Events:
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| : 1) DVR 06-01-88-239: SAC moisture separator ruptures.
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| : 2) LER 20-01-88-025: Loss of Turbine Building air header.
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| ld 305M/83M/4 5/89
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| AM - f DEVIATION INVESTIGATION REPORT (OIR)
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| F m Rn 2.D Facility Name PAGE Byron Nuclear Power Station 1 0 3 Title V Moisture Senarator Rusture Resultino in Air Pressure Drom and Auto Start of the Standbv Camaressor EVENT DATE DIR NupeER rep 0RT DATE i
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| // SEQUENTIAL // REVISION PENTH DAY YEAR STA taff T . YEAR / IANGER / IA9ere peNTH DAY YEAR 1 POWER LEVEL 1 12 2 10 8 18 0 16 0 11 e la -
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| 2 13 1 9 -
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| 010 0 12 Ol2 819 1 10 10 C00ffACT FOR THIS DIR NAfE TrtrPHONE NupeER AREA CODE 1 D. Brindle. Onoratine Ennineer Ext. 2218 8l115 2l3I4l-l5I4I4l1 CDerLETE ONE LIlE FOR EACH CONPolENT FA URE DEScafRFD IN THIS REPORT CAUSE SYSTEM COMPONENT M41rJFAC- REPORTABLE CAUSE SYSTEN COMPONENT MANUFAC- REPORTABLE T11ere TO NPRDS TURER TO NPRDS X LlF IS lE lP f 12 11 Il N I I I I l I I I l I l l I I l l l l l SUPPLEMENTAL REPORT EXPECTED TENTH DAY YEAR p
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| SP) MISSION l YES fif ven. enmolate EXPECTED StmWISSION DATE1 [I ND l l TEXT Energy Industry Identification System (E!!S) codes are identified in the text as [XX)
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| A. PLANT COPEITIONS PRIOR TO EVENT:
| |
| Event Date/ Time 12-20-as_/,1140 Unit 1 MODE 1 - Power Onoration Rx Power 99.9 RCS (AB) Temperature / Pressure Normal Oneratina Unit 2 MODE 1 - Power Oneration Rx Power 40 RCS [AB) Temperature / Pressure Normal Oneratino B. DESCRIPTION OF EVENT:
| |
| There were no systema or components inoperable at the beginning of this event which contributed to its severity. At 1140 hours on December 20. 1988, the Unit 2 Station Air Compressor (SAC) (LF) auto started on low header pressure. Investigation into the low header pressure revealed a rupture in the moisture separator on the Unit 1 SAC. The Unit 1 SAC was shutdown and ths moisture separator was isolated from the system. Nuclear Work Request 863350 was written to replace the Unit 1 SAC moisture separator. The moisture separator was replaced and the Unit 1 SAC was made available on December 23, 1988. The Unit 1 SAC is Non-Safety Related and as such was not declared inoperable following the eventi however, th'eUnit 1 SAC was made unavailable due to the moisture separator failure. On .lanuary 9, 1989 the Unit 2 SAC moisture separator developed a pin hole leak. The Unit 2 SAC was shutdown and the Unit 1 SAC s s manually started after the leak was identified by an equipment attendant (Non-licensed). The Unit 2 SAC was made unavailable due to the leak. The moisture separator will be replaced under Nuclear Work Request B63631.
| |
| There were no manual or automatic safet) system actuations due to this event and plant conditions were stable at all times. All operator actions Man during this event were correct.
| |
| (0220R/0025R/012089)
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| ----]
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| | |
| DEVIATION INVESTIGATION REPORT TEXT CONTINUATION Fom pov 2.0 FACILITY NAE DIR NLDSER PAGE SEiUENTIAL REVISION g}
| |
| g $TA LINIT YEAR Ntteft NLDSER Q - .
| |
| Byron Nuclear Power Station 0 16 0 11 a la -
| |
| 213l9 -
| |
| 0 lO 2 0F 0I3 TEXT Energy Industry Identification System (EIIS) codes are identified in the text as [XX] j C. CAUSE OF EVENT:
| |
| The cause of the events is due to a combination of factors. The drain line which automatically drains the moisture separator on the Unit 1 SAC was found plugged by the Mechanics replacing the moisture separator.
| |
| The drain line for the Unit 2 SAC moisture separator was found plugged on December 30, 1988 as part of the-investigation into the Unit 1 SAC moisture separator failure. The accumulation of water caused corrosion of the interior of the moisture separators. The corrosion eventually caused weakening of the moisture separators which finally- resulted in rupture. Corrosion of the moisture separator interiors cannot be prevented, but could be kept to a minimum by keeping the moisture separators drained.
| |
| It is unknown how long elther drain was plugged since there were no work requests written prior to the events which described the condition. The operator rounds currently do not require the operator to manually drain the moisture separators.
| |
| D. SAFETY ANALYSIS:
| |
| There were no safety consequences due to this event. The Unit 2 SAC was available and Auto-Started to maintain air system pressure during the Unit I rupture. If the Unit 2 SAC had not been available, then only the Unit 0 SAC would have been in operation to maintain system pressure. During the Unit 2 moisture ,
| |
| separator leak, the Unit 1 SAC was manually started to maintain air system pressure. If the Unit 1 SAC had not been available, then only the Unit 0 SAC would have been in operation to maintain system pressure.
| |
| ) Braidwood station normally operates only one SAC and therefore it could be expected that the Unit 0 SAC at l Byron could maintain air system pressure by itself. However, the capability of the Unit 0 SAC to maintain {
| |
| system pressure is highly dependant on air usage and air system leakage which is unique to Byron. If the j Unit 0 SAC could not maintain sufficient pressure, then a plant transient and possibly a Unit trip may have resulted.
| |
| E. CORRECTIVE ACTIONS:
| |
| The moisture separator for the Unit 1 SAC was replaced and the drain line was cleared. The manual drain lines for both the Unit 0 and the Unit 2 SAC moisture separator were checked for flow. The Unit 2 SAC moisture separator manual drain was found plugged. Nuclear Work Request 863631 was initiated to clear the drain line and replace the Unit 2 SAC moisture separator which developed a leak on January 9,1989.
| |
| Procedure revisions were initiated to the operator rounds procedures (90P 199-A28 and 80P 199-A47) to require manual draining of the moisture separators. Implementation of the procedure revisions is tracked i by Action Item Record (AIR) 454-225-89-0025. A preventive maintenance request (BMP 3200-T15) was initiated to place the moisture separators on a 5 year inspection interval. The update to the preventive maintenance program is tracked by AIR 454-225-89-0024.
| |
| l An Ultrasonic (UT) inspection was performed on the Unit 0 SAC moisture separator to deterwine if its walls
| |
| , were thinning. The UT inspection of the Unit 0 moisture separator did not show thinning to the extent that was found on the Unit I and Unit 2 SAC's.
| |
| 9 l
| |
| k l %
| |
| l l
| |
| (0220R/0025R/012089)
| |
| | |
| _ _ _ _ . + . . . _.__-.u.------;-:
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| .____.._..g ,
| |
| l
| |
| , DEVIATION INVESTIGATION REPORT TEXT CONTINUATION
| |
| , Form Rev 2.0 FACILITY NAME DIR NUBBER PAGE O SEQUENTIAL REVISION
| |
| ( STA UNIT fEgt NLDSER NUPRER 8vron Nuclear Power Statlan 0 16 0 11 8 18 -
| |
| 213l9 -
| |
| 0 l0 3 0F 0l3 TEXT Energy Industry Identification System (EIIS) codes are identified in the text as [XX]
| |
| F. RECURRING EVENTS SEARCH AND ANALYSIS:
| |
| I a) EVENT SEARCH (DIR. LER) l No similar previous events found at the station.
| |
| b) IMluSTRY SEARCH (OPEX's NPRDS)
| |
| No relevant information found c) 19E The drain line for the Unit 2 SAC moisture separtor was previously found plugged on September 30, 1987. (849467) d) ANALYSIS Drain line plugging will be more quickly identified once manual draining of the moisture separator is added to the operator rounds.'
| |
| G. COMPONENT FAILURE DATA:
| |
| J MANUFACTURER NDPENCLATURE PSDEL NUPSER MFG PAR 1 NlteER U-1 ITT Fluid Handling Div. Meisture Separator Cat. No. ME71008 Manufactured 1977 U-2 ITT Fluid Handling Div. Noisture Separator Cat. No. ME72500 Manufactured 1979 H. OTHER RELATED DOCUPENTS:
| |
| NUREG 1275.Living SOER 81-09 I. EFFECTIVENESS REVIEW:
| |
| Scheduled for completion 2/1/90 J. ADDITIONAL DATA:
| |
| a) Affected Technical Specification: None b) Procedures: B0P 199-A28 and BOP 199-A47 c) Cause Code: XPRP911P99 d) Equipment Involved: ISA02A and 25A02A Holsture Separators e) Other: Corrosion 1
| |
| (0220R/0025R/012009)
| |
| | |
| . g%t-/
| |
| LICENSEE EVENT REPORT (LER)
| |
| Form Re Q Docket Number (2) Pane (3,
| |
| : Facility Name (1) ]
| |
| 0'' "
| |
| 01 5l Of 01 01 41 51 6 1lofl0l4 h
| |
| \,.
| |
| Title (4) Manual Reactor Trips due to approaching Low Low Steam Generator Levels as a result of Loss of Instrument Air Event Date (5) LER Number (6) Resort Date (7) Other Facilities Involved (8)
| |
| Year Sequential Revision Month Day Year Facility Names Docket Number (s)
| |
| Month Day Year /// //
| |
| 77j
| |
| /// W&r /jj/
| |
| f
| |
| // Number Braidwooni Unit 2 015101010141Sl7
| |
| ~ ~
| |
| 11 1 Il 5 81 8 81 8 01215 010 1l2 11 3 81 8 0151010101 l l THIS REPORT IS SUBMITTED PURSUANT TO THE REQUIREMENTS OF 10CFR p
| |
| (Check one or more of the followino) f11) 20.402(b) _ 20.405(c) JL 50.73(a)(2)(tv) _ 73.71(b) !'
| |
| POWER _ 20.405(a)(1)(1) _ 50.36(c)(1) _ 50.73(a)(2)(v) _ 73.71(c)
| |
| LEVEL 20.405(a)(1)(ll) 50.36(c)(2) 50.73(a)(2)(vii) _
| |
| Other (Specify l9 l6 _ _ _
| |
| (101 0 _ 20.405(a)(1)(lii) _ 50.73(a)(2)(1) _ 50.73(a)(2)(viii)(A) in Abstract 20.405(a)(1)(iv) 50.73(a)(2)(li) 50.73(a)(2)(v111)(B) below and in
| |
| /,/,/,/,/,/,/,/,/,/,/,/,/,/,/,/,/,/,/,/,/,/,/,/,/,/, _ _ _
| |
| 50.73(a)(2)(iii) 50.73(a)(2)(x) Text)
| |
| '/jj
| |
| /'j' '/'/j/'
| |
| ,/ // //// /,/j/'j' //////,
| |
| j/jjjjj/j'j/' /jjj'jjj' /' _ 20.405( a)( 1 )( v) _ _
| |
| LICENSEE CONTACT FOR THIS LER (12)
| |
| TELEPHONE NUMER Name AREA CODE Paul Nvkara. Technical Staff Enoineer Ext. 2477 8l 1 15 4lSl81 l218lOf CONPLETE ONE LINE FOR EACH COMn0N N FAILURE DESCRIBED IN THIS REPORT (13)
| |
| SYSTEM COMPONENT MANUFAC- REPORTA8LE CAUSE SYSTEM COMPONENT MANUFAC- REPORTABLE CAUSE .
| |
| TURER TO NPRDS TURER TO NPRDS B LI D' Pl Sl Fl
| |
| * MI41716 N l l l l l l l 1 1 I I I I i 1 I I I I I I SUPPLEMENTAL REPORT EXPECTED (14) Expected Month I Day I Year Submission
| |
| * l l lyes (If ves. canelate EXPECTED SUBMISSION DATE) X l NO I ABSTRACT (Limit to 1400 spaces, i.e. approximately fif teen single-space typewritten lines) (16)
| |
| At 0904 on November 15, 1988 low instrument air pressure was observed. The rapid decrease in the instrament air header pressure uused the feedwater regulating valves to go closed. This decreased flow to the steam generators on both units. At 0908 both Units were manually tripped due to decreasing steam generator levels. The cause of this event was inadequate installation of a coupling in the instrument air header, line CIA 058 during construction. The inadequate solder joint was stressed by contract personnel standing on the line. The instrument air header was isolated and the line repaired by replacing the joint. The line was inspected upstream and downstream of the break for other possible leaks that may have occurred as a result of the break. Two other joints were repaired f or pinhole leaks. Additional pipe supports will be added to the header. A letter was issued on November 16, 1988 to all site personnel reemphasizing the need for all personnel to exercise care in working around all plant equipment. There have been no previous occurrences of a reactor trip as a result of a loss of instrument air.
| |
| j /'"
| |
| l (w 2390m(121288)/19
| |
| ---_---_a--__-__-_m__m_ _ _ _ _ _ , _
| |
| | |
| t:
| |
| n LICENSEE EVENT REPORT fLER) TEXT CONTINUATION Form Rev 2.0 FACILITY NANE (1) DOCKET NUleER (2) LER HUpeFR (6) Pane (3)
| |
| Year /// 5equential / Revision ffj/ ff
| |
| //j//
| |
| Braidwood Unit 2 // Number / Number Ir]
| |
| \h 0 I 5 l 0 1 0 1 0 1 41 51 6 Bl8 - 01215 - 0l 0 0I 2 0F 01 4 TEXT Energy Industry Identification System (EII5) codes are identified in the text as [XX)
| |
| A. PLANT CONDITIONS PRIOR TO EVENT:
| |
| Unit: Braidwood I; Event Date: November 15, 1988; Event Time: 0908; Node: 1 - Power Operation: Rx Power: 96%;
| |
| 4 [A8) Temperature / Pressure: 585 degrees F/2235 psig Unit: Braidwood 2; Event Date: November 15, 1988; Event Time: 0908; Node: 1 - Power Operation; Rx Power: 791; RCS [A8) Temperature / Pressure: 578 degrees F/2240 psig
| |
| : 8. DESCRIPTION OF EVENT:
| |
| At 0904 on November 15, 1988 low instrument air (IA) [LO) receiver pressure was observed by control room personnel. Operators were dispatched to check for instrument air leaks, The rapid decrease in the instrument air header pressure caused the feedwater (FW) [5J) regulating valves. (2)1FW510. (2)1FW520, p (2)1FW530, and (2)1FW540, to go closed. This decreased flow to the steam generators (SG) [J8) on both units.
| |
| t At 0908, both Units were manually tripped due to decreasing steam generator levels.
| |
| 1 Operator actions decrease 6 the severity of this event since the reactors were manually tripped prior to any Engineered Safety feature (EF) [JE) actuation.
| |
| The Auxiliary Feedwater (AF) [8A] pumps automatically started to maintain steam generator levels as designed. ;
| |
| The appropriate NRC notification via the ENS phone system was made at 1000 pursuant to 10CFR50.72(a)(1)(1),
| |
| and 10CFR50.72(b)(2)(li). j l
| |
| Braidwood Station met with NRC Region III personnel on December 6,1988, to discuss this event and proposed I corrective actions.
| |
| This event is being reported pursuant to 10CFR50.73(a)(2)(iv) - any event or condition that resulted in manual or automatic actuation of any Engineered Safety Feature, including the Reactor Protection System.
| |
| C. CAUSE OF EVENT:
| |
| The root cause of this event was inadequate insta11atM sf a copper coupling in the instrument air header.
| |
| line OIA058 during construction. The inadequate solder joint was stressed by contract personnel standing on the line.
| |
| When operators arrived at the break location, there was evidence that the line had been used to stand on while painting another line above it. The painting of the line above had abruptly ended directly above the break and a wet paint roller was found on the floor below the broken line. It was verified that a painter j l [(,)/ had been standing on the line.
| |
| 2390m(121498)/20 l
| |
| i
| |
| [
| |
| | |
| ~ '
| |
| tfEENSEE EVENT REPORT (LER) TEXT CONTINUATION Form Rev 2.0 FACILITY NAME (1) DOCKET NUPSER (2) LER NUPSER (6) Paos (3)
| |
| Year /// Sequential /// Revision fjj fff Braidwood Unit 2 /// Number /// Number
| |
| -m (j) '
| |
| 0 l 51010 l 0 l dl El 6 8l8 - 01215 - 01 0 01 3 0F 01 4 TEXT Energy Industry Identification System (EIIS) codes are identified in the text as (XX]
| |
| C. CAUSE OF EVENT: (continued)
| |
| The resulting loss of instrument air caused the feedwater regulating valves to go closed. This resulted in a reduction of feedwater flow to the steam generators leading to the manual reactor trips.
| |
| D. SAFETY ANALYSIS:
| |
| This event had no offeet on the safety of the plant or the pubile. All engineered safety systems operated as designed.
| |
| Under the worst case conditions of a plant operating at 100Y. power with a loss of instrument air and no operator action, there would be no additional adverse impact on the safety of the plant er public as this is enveloped by the Final Safety Analysis Report (FSAR), Process Auxiliaries.
| |
| E. CURRECTIVE ACTIONS:
| |
| The issnediate corrective action was to recover steam generator levels and establish stable conditions.
| |
| The instrument air header was isolated and the line repaired by replacing the joint. The line was also inspected upstream and downstream of the break fur other possible leaks which may have occurred as a result g
| |
| of the break. Two other joints were repaired for pinhole leaks. Additional pipe supports will be added to the header. This will be tracked to completion by action item 456-200-88-26701.
| |
| PWR Engineering will evaluate the solder quality for portions of the IA System by sampling a few of these joints during the Unit 2 surveillance outage. Any additional actions will be based on the results of this sample. This will be tracked to completion by Action Item 456 200-88-26702.
| |
| Braidwood letter 88-1439 was issued on Movember 16, 1988 to all site personnel reemphasizing the need for all personnel to exercise care in working around all plant equipment.
| |
| F. PREVIOUS OCCURRENCES:
| |
| There have been no previous occurrences of a reactor trip as a result of a loss of instrument air.
| |
| \
| |
| 2390m(121288)/21 e
| |
| | |
| LICENSEE EVENT REPORT (LER) TEXT CONTINUATION rarm Rev 2.0 DOCKET NUPSER (2) LER NLDRER (6) Pane (3)
| |
| . FACILITY NAM (1)
| |
| Year ///
| |
| fff Sequential /jj//
| |
| f Revision
| |
| /// Neer /// Number Braidwood Unit 2 fN
| |
| .\ 3 i 5 l 0 1 0 1 0 1 41 51 6 aI8 - 0l215 - 01 0 01 4 0F 01 4 TEXT Energy Industry identification System (EIIS) codes are identified in the tent as [XX]
| |
| G. COMPONENT FAILURE DATA:
| |
| M6nufacturer Nomenclature Model Number MFG Part Number Mueller Brass Elbow, tubine: 45 deg N/A N/A 4 inch copper
| |
| - Mueller Brass Elbow tubing; 90 deg N/A N/A 4 inch copper Mueller Brass Coupling; tubing; N/A N/A 4 inch copper Mueller Brass Turbine; copper; N/A N/A 4 inch x 20 ft.
| |
| Type K Hard temper ASTM 5888 t
| |
| l 4
| |
| i a
| |
| 2390m(121288)/22 i
| |
| I l
| |
| | |
| ( BRAIDWOOD SIMULATOR MAT.VUNCTION l~
| |
| t
| |
| | |
| ==Title:==
| |
| RWST Leak ID: AUX-2 u NO: 6.3.4.1.2
| |
| ; (~ -
| |
| | |
| == Description:==
| |
| Leak in'the RWST.
| |
| '\
| |
| Variations: Date: 8/1/86 Rev: 3
| |
| . Selectable l Steps Inputs Comments l- 1. Select leak size 0-s0,000 gym
| |
| : 2. Select ramp time 0-99,999 sec.
| |
| Brief Plant Response:
| |
| RWST level lowers through low level alarm, LO-2 level and empty setpoints.
| |
| Suggested Instructor Action:
| |
| None.'
| |
| Events:
| |
| None O
| |
| ~
| |
| 1.s,.:. ses ,es , e,..
| |
| | |
| ._7_ _-
| |
| BRAIDWOOD SIMULATOR MALPURCTION
| |
| | |
| ==Title:==
| |
| Cire Water Pump Trip ID: AUX-3 .
| |
| NO: 6.3.4.1.3
| |
| ./ 9
| |
| | |
| == Description:==
| |
| Selected cire water psamp trips due to bearing failure.
| |
| Q' Variations: Date 3/12/89 Rev: 5 Selectable Steps Inputs Comuments
| |
| : 1. Select faulty pump AUX 3A 3A = 1A CW ptany AUX 35 3B = 1B CW pump
| |
| . AUX 3C 3C = IC CW pump Brief Plant Response Depending on power level, condenser vacuum lowers, turbine trip may occur.
| |
| (Ex trips if sbove P-8)
| |
| Suggested Instructor Action:
| |
| To restart the tripped pump, first clear the malfunction.
| |
| ' Events:
| |
| LJ 1. LER 06-01-85-062: Ex trip on OTAT due to loss of two cire water pumps.
| |
| : 2. DVR 6-1-85-274: Motor auto-synchronizer failure, motor would BRt switch from induction to synchronous.
| |
| : 3. DVR 06-01-87-080: IC Cire Water Pump Trips.
| |
| : 4. DVR 06-01-87-172: 1A Cire Water Pump Trips.
| |
| : 5. DVR 06-02-88-114: 2A Cire Water Pump Trips.
| |
| l 305M/83M/6 5/89
| |
| -.. - - -~ . - ------
| |
| l
| |
| | |
| ~ ~ -- - - -
| |
| LICENSEE EVENT eEPORT (Lgg)
| |
| Facility.came (1) Cocket Rumcer (2) fue '3)
| |
| Byren. Unit 1 of gl of of of 41 si a t lofl0 ! 2 avtR ?tMPEDaTURE CELTa T DEACTCR TRIP / LOST OF TWO CIRC WATER PUMPS l LIR Number (6) eeoort Date (7)
| |
| OtMer Facilities involved (9)
| |
| { }entDate(El
| |
| ...th Day Year Year sequential /j/j/ Revtsion Month Day Year Factitty = aws ! Decket No-eerts)
| |
| Number j/// Number I of El of of of f f
| |
| ~~~ ~~~
| |
| of s ?! 7 si s el E oisi2 oIo af7 21 6 al 5 of si of of of I l THIS REPORT 15 $UBMITTED PURSUANT TO THE REQUIREMENTS OF 10CFR OPERATING 1 (Check one or more of the followinal f111 20.402(b) 20.405(c) .I. 50.73(a)(2)(1v) _ 73. 71 ( b )
| |
| POWER , 20.405(a)(1)(1) _ 50.36(c)(1) 50.73(a)(2)(v) 73.71(c)
| |
| LEVEL _, 20.405(a)(1)(11) 50.36(c)(2) _ 50.73(a)(2)(vt1) Other (spectfy (101 0 !9!9 _ 20.405(a)(1)(111) _ 50.73(a)(2)(1) 50.73( a )( 2)(v111)( A) in Abstract
| |
| ,_ 20.405(a)(1)(tv) 50.73(a)(2)(11) _ 50.73(a)(2)(v111)(8) below and in
| |
| _ 20.405(a)(1)(v) _._ 50.73(a)(2)(i t t ) _,._ 50.73(a)(2)(x) Text)
| |
| LifENSEE CONTaET FOR THis lee (121 Nane TELEPHONE NUMBER AREA CODE Joe Reister system Ene1neer Ext. 2378 a1i!E fl il 41 -l El 41 41 COMPLETE ONE LINE FOR EACH COM T FAfluff DEtfefaED IN THIS REPORT fill CAUSE SYSTEM COMPONENT MANUFAC- REPORTABLE / CAUSE SYSTEM COMPONENT ' MANUFAC- REPORTA8LE TURER TO NPRDS / TURER TD NPRDS a K IE vi Al al T Fl il 7l E N / l- l l l l B K lE $lPl P Wl Il 21 0 W / l l 1 l I f .
| |
| . SUPPLEMENTAL REPORT EXPECTED f141 frpected Month l Day I Year Submission fYes fff vet. comnlete EXPECTED SUEMfsif0N DATE1 I NO I I I ABSTRACT (Limit to 1400 spaces. i.e. approntmately fif teen single-space typewritten Itnes) (16)
| |
| At 1550 CST on 6/27/85 an automatic reactor trip was initiated on overTemperature Otfrerenttal Temperature.
| |
| OTDT. The trip resulted from the loss of two circulating water pumps which tripped due to overheating of their exciter control circuits. The 0T07 setpoint was reached due to turbtne load being shed on the loss of the circulating water pumps and subsecuent RCS heatup and pressure increase. The loss of the circulating water pumps prevented the operation of the Steam cump (Turbine Bypass) valves to mitigate the RCS pressure and temperature transtant. When RCS pressure reached the PORV opening setpoint, the elevated temperature, combined with the sudden drop in pressure. caused the 0T07 setpotnt to be exceeded.
| |
| The exciter control circuits for all three circulating water pumps were located by design in a comon control cabinet. The circutts have subsequently been separated into three individual cabinets.
| |
| s a ,
| |
| (0639M) i
| |
| | |
| f tfCEW?EE EVENT DE70ef (LED) TEXT CONTINUATION f' FACILITY NA.NE (1) 00CKET NUMBER (2) LER wuweER f6) l sm -
| |
| Year g// Seccantiallg Revision
| |
| / Number l'/ // Mumcer 01E IOIO I e I al El a e15 0 16 I2 0 l 0 Of 2. cr y ;
| |
| avron. Unit 1 - -
| |
| TEXT At 1550 CST on 6/27/85 an automatic reactor trip was initiated on Overtemperature Differential Temperature.
| |
| OTDT. The trip resulted from the loss of two circulating water pumps which tripped due to overheating of their exciter control circuits. The OTDT setpoint was reached due to turbine load being shed on the loss of the circulating water pumps and subsequent RCS heatup and pressure increase. The loss of the circulating water pumps prevented the operation of the Steam Dump (Turbine Sypass) valves to mittgate the RCS pressure and tanq>erature transient. When RCS pressure reached the PORV opening setpoint. the elevated temperature. combined with the sudden drop in pressure, caused the OTOT setpotnt to be exceeded.
| |
| The CW pumps tripped due to overheated exciter control circuit components. Since all three pump exciter circuits were in a common cabinet when the IA CW Pump circuit caught on fire. the 15 CW pump circuit was S aequently affected by the heat and the 15 CW pump tripped as well. Although the exact cause of the overheating condition could not be pinpointed; the f ailure was most likely caused by the proximity of the three exciter control circuits in a single cabinet. The heat generated by these circutts could not be properly dissipated.
| |
| The consequences of the event were etntmized by quick operator action. Upon seatng ,
| |
| that that the 1A CW pump had tripped, turbine load was innediately ramped down such that condenser vacuum might be maintatned.
| |
| However. when the second circ water pump tripped, the turbine rangs rate proved to be inadequate and the reactor protection system tripped the unit.
| |
| Since no safety limits were met or exceeded. no adverse consequences resulted from this event. Cire water is a non-safety related system. No provisions are made for redundancy while at full power. The reactor
| |
| /] ,/ protection system performed adeguately per design to mitigate the consequences of this event.
| |
| The circ water pump exciter control circuits including variable transformers and surge protection diodes.
| |
| were replaced like for like; however they were separated into 3 separate cabinets. one per pump. The pumps have since been tested and are performing satisfactorily. There has been no prior occurrence of this event.
| |
| h
| |
| [
| |
| N/
| |
| (0639M)
| |
| | |
| / OEv! Af DN lt ESTICM ::N REPORT TITLE pact CW PUMP GREarER TRIP i ! Ort 3 l 1
| |
| : EVENT DATE 019 NUMBER REPORT DATE /// //////,
| |
| k f/j SECUENTIAL // REVISION /ff////W#
| |
| MONTH DAY YEAR STA UNIT YEAR // NUMBER // NUMBER MONTH DAY YEAR t, //
| |
| POWER ,
| |
| ' EVE' '
| |
| of e el I RI s of a of 1 El 5 -
| |
| 2 17 1 4 -
| |
| oIa 0 19 26 8 15 of H s ffff CONTACT FOR THis O!R NAME TELEPHONE NUMBER AREA CODE Joe Ratster. System Encineer Ext. 2378 R l 1 l E 2 11 Ial - l 5 IaiaIt COMPLETE ONE LINE FOR EACH CUMPONEN LURE DESCRIBE 0 fN THit REPORT CAUSE SYSTEM COMPONENT MANUFAC- REPORTA8LE CAUSE SYSTEM COMPONENT MANUFAC-' REPORTA8LE TURER TO NPRDS TURER TO NP905 x riE I of 11 s ! I I N I I I l l l t l I I I 1 l I I I I i i
| |
| $ SUPPLEMENTAL REPORT EXPECTED MONTH ' OAY YEAR SU8 MIS $10N I YES ftff,ves. comolete EXPEffED SURMISSION DATE) l NO TEXT WHaf HAPPENEO?
| |
| At 1627 and 1635 on 9/1/85, the 1A Cire. Water Pump trtpped approutmately 7 seconds af ter each start attempt.
| |
| WHAT WAS THE ROOT CAu$1?
| |
| The Pump Motor Auto-syncheontzer failed. This is a small hermetically sealed device making detalled inspection difficult. The exact cause for the device fatture is unknown. As a result of this malfunction. the pump mode of, operation fatled to switch from induction to synchronous, causing pump trip on high current.
| |
| HOW DID IT A t tJCT PLANT ANQfj1 PUBLIC SAFETY!
| |
| * The pump was being started in advance of an up power manauver. The pump trip had no affect on safety.
| |
| hat 17 HAPPENED REFORE?
| |
| %o previous failures.
| |
| WH AT WAS DONE TO CORRECT THE CON 0f TION AN(L HOW ARE WE MLNiTDMENT RECURRENCE?
| |
| The Auto Synchrontrer was replaced and the pump has been operating satisfactorily since. No further action is required.
| |
| l A
| |
| (0638M)
| |
| | |
| OEVIATICN IN%EST!GAf!ON REPORT j .]
| |
| TITLE pact it CIPCULATINr. WATER PUMP TRIP DUE TO a LOOSE LFAD EVENT DATE DIR NUMBER REPORT DATE SEQUENTIAL REVISION
| |
| .fH UNIT f/ NUMBER g// NUMgi&_ MONTH DAY DAY YEAR JTA YEAR // YEAR 1 POWER 015 115 Bl7 016 011 817 01 al 0 -
| |
| 010 0 17 Pl0 Al7 9 17.15 CONTACT FOR THIS DIE NAME TELEPHONE NUMBER AREA CODE T. Didier. Oneratina Eneineer Ext. 2217 al 1 15 213 14 I-lE I4 14 l 1 CDMPLETE ONE LINE FOR EACH COMPONEN FA LORE DE$rRIBED IN THIS REPORT CAUSE SYSTEM COMPONENT MANUFAC- REPORTABLE / CAUSE SYSTEM COMPONENT MANUFAC- REPORTABLE TURER TO NPRDS / TURER TO NPPDS i l ! l l l I '/ l l l I I l I I I l I l I l / I I I I I SUPPLEMENTAL REPORT EXPERTED MuNTH DAY YEAR a SUBMISSION I YES fif ves. comolete EXPECTED SUBMISSION DATE1 l NO TEXT A. PLANT CONDITIONS PRIOR TO EVENT:
| |
| Event Date/ Time 06/15/B7 / 1930 1 - Power anaration Rx Power 97.5 RCS [AB) Temperature / Pressure Normal p Unit 1 MODE
| |
| \") Unit 2 MODE .K/A_- N/A Rx Power N/A . RC5 (A8) Temperature / Pressure N/A S. DESCRIPTION OF EVENT:
| |
| On 06/15/87 at 1930 hours Byron Unit I was in Mode 1 (Power Operation) at a power level of 97.5%. The 1C Circulating Water (CW) (KE] pump tripped off causing a rapid drop in the main condenser vacuum. The unit was ramped down from 1120 MWe to 980 MWe at 50 MW/ min. U-1 Hogger vacuum pump was started and the plant was stabilized at 83% power, the rods were borated out and Byron Operating Abnormal Procedure 1 BOA - SEC 3 was entered.
| |
| C. CAUSE OF FVENT:
| |
| The cause of the IC CW Pump Trip was determined to be a loose lead on a moving contact which touched the mounting screw for the stationary contact. This energized the PR27 overcurrent relay. The short caused by the loose lead and the resulting trip was apparently initiated af ter the door to Bus 143 Cub #12 was closed ,
| |
| by an Equipment Operator (ED).
| |
| P I I (1546M/0183M) l
| |
| _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _l_-___. ____ -__
| |
| | |
| DEVIATICN INVE$TIGATION REPORT TEXT CONTINUATION TITLE DIR NUM2ER PACE SEQUENTIAL REVISION STA UNIT YEAR NUMBER NUMBER CIRCULATING WATER PUMP TRIP DUE TO A LOOSE LEAD 016 011 R17 Of El 0 -
| |
| 010 02 0F 012 TEXT D. SAFETY ANALYSIS:
| |
| The circulating water system is non-safety related, but all three circulating water pumps are required for optimum plant operation at high power levels. The plant was ramped down in a controlled manner without any other abnormal events occurring. The public safety was not compromised at any time during this event.
| |
| E. CORRECTIVE ACTIONS:
| |
| The PR 27 overcurrent relay was removed from the IC CW Pump breaker Bus 143 Cub #12 to facilitate repairs and all the leads were tightened and relanded to prevent a re-occuring short. The IC CW pump was restarted and the plant was returned to full power.
| |
| F. PREVIOUS OCCURRENCES:
| |
| 5 DVR NUMBER E NONE G. COMPONENT FAILURE DATA:
| |
| a) MANUFACTURER NOMENCLATURE MODEL NupeER MFG PART NUMBER
| |
| /G
| |
| (
| |
| v) Not Applicable b) RESULTS OF NPRDS SEARCH:
| |
| Not Applicable i
| |
| i l
| |
| )
| |
| )
| |
| )
| |
| f P
| |
| (1546M/0183M)
| |
| L__________.__
| |
| | |
| DEVIATION INVESTIGATION REPORT Alix - 3 ITLE Pact CW PUMP iRIP DUE TO SHORTED WIPE IN THE PUMP EXCITER TRANSFORMER 1
| |
| EVENT DATE DIR NUMBER #EPORT DATE
| |
| // SEQUENTIAL // REVI5!DN MONTH DAY YEAR $TA UNIT YEAR / NUMBER NUMBER MONTH DAY YEAR 1 PDWER
| |
| ' EVE' 11 2 21 1 al 7 of 6 ' of 1 al 7 --
| |
| 1 1 7f 2 --
| |
| oIo 012 011 81 8 I al e CONTACT FOR THIS DIR TELEPHONE NUMBER CAME AREA CODE Don Brindle. U-2 Oneratine Enoineer Ert. 221s ai1 I$ 2 l1 la1-l5l4 l4 l 1 COMPLETE ONE LINE FOR EACH COMPONE URE DESCRIBED IN THIS REPORT CAUSE SYSTEM CDNPONENT MANuFAC- REPORTA8tE CAUSE SYSTZM CDMPONENT MANUFAC- REPORTA8LE TURER TO NPRDS TURER TO NPRD$
| |
| X NIN lE IX fc E 11 12 10 W l l l l l 1 l I I l I I I I l l l l 1 SUPPLEMENTAL REPORT EXPECTED NnNTH DAY YEAR p
| |
| SUBM1551DN l YES fif yes. camolete EXPECTED suRMIs5 ION DATE1 X l NO TEXT A. PLANT CONDITION $ PRIDE TO EVENT:
| |
| Event Date/ Time 12/21/a7 / 0429 Unit 1 MODE 1 - Power coeration Rx Power 80. . RCS [ A8] Temperature / Pressure Normal Oneratina Unit 2 MODE 1 - Power comration Rx Power 29 RCS ( A8] Temperature / Pressure Norwal Oneratino
| |
| : 8. DESCRIPTION OF EVENT:
| |
| On 12/23/87, at 0429, with Unit 1 at 80% reactor power, the 1A Circulating Water Pump Tripped from he Field Excitation Trip Relay.
| |
| As a result of the loss of 1A Ctreulating Water Pump an increase in condenser back pressure occurred. The UI hogger vacuum pump was placed into service and the unit was ramped down approximately 10 megawatts.
| |
| Proper condenser vacuum was subsequently reestablished.
| |
| Primary plant conditions followeo the transient in a normal manner throughout the event. There were no other systems or components inoperable at the beginning of the event which contributed to the event.
| |
| C. CAUSE OF EVENT:
| |
| Investigation by station Electrical Maintenance revealed that the center tap on the automatic transformer for the motor exciter field had shorted causing a loss of excitation, thus, tripping 1A Circulating Water Pump. The damage to the center tap connection and associated wire was severe enough that the exact cause of the short could not be determined.
| |
| 1
| |
| +
| |
| I 1
| |
| (1827M/0209M)
| |
| Q_-_-_---____.
| |
| | |
| DEVIATION INVESTIGATION REPORf TEXT CONT!%UATION TITLE DIR NupeER PACE SEQUENTIAL REVISICN CW PUMP TRIP DUE TO $HORTED WIRE -
| |
| sTA UNrf YEAR NUPSER NUMBER IN THE PUMP EXCITER TRANSFORMER of 1 a 17 t I7l2 -
| |
| o 10 0 12 0l6 -
| |
| 2 0F <
| |
| l TEXT D. SAFETY ANALYSIS: l The loss of a circulating water pump is not considered a safety failure. A controlled reduction of power -
| |
| ts all that is required to prevent a loss of vacuum in the main condenser. No safety systems were affected by this event. The reactor protection system and $/G PORV's were operable had the turbine tripped on loss j of codenser vacuum. There was no effect on the health and safety of the pubite. 2 E. CORRECTIVE ACT10ks:
| |
| The automatic transformer and approximately 3 f t of wtre from the transformer to the local terminal block were replaced. The automatic transfonners for 15. IC and 2A. 2B. 2C Circulating Water Pumps were inspected and all connections were intact.
| |
| The 1A motor was meggered and a polarization test performed per SHP 4200-52. The 1A Circulating Water Pump was restarted at 1419 on 12/23/87 and performed satisfactorily.
| |
| F. PREVIOUS DCruRRENCES:
| |
| There have been no previous occurrences of t'his type of failure on the transformer for the motor exciter field on the CW Pumps. A trend review was initiated based on two previous occurrences (DVR 6-86-009 and
| |
| ( DVR 6-1-87-80). It was determined that though ea:h instance resulted in loss of excitation to CW pumps the V causes were separate and unrelated. (Trend 87-36)
| |
| DVR NUPGER J1D,1 NONE G. [0MPONENT FAftuRE DATA:
| |
| a) MANUFACTURER NOMENCLATURE MODEL NtDGER MFG PART NUMBER Electric Machinery Automatte Transformer Synchro Pac 11 SI #789044 b) REsults or NPeD$ 1[AerH' N/A c) REtutts DF NUCLEAR WORK REDuffT f MWR) SEARCH There are no previous NWRs written for this particular problem.
| |
| O I I (1827M/0209H)
| |
| | |
| AUK- 3 DEVIATION IWESTIGATION REPORT (DIR) racility Nome PAGE g Byron Nuclear Power Station i
| |
| 1 0 l 0 L '.m Title 2A Circulating Water Pune Trin EVENT DATE DIR NUPSER REPORT DATE p
| |
| SEQUENTIAL REVISION HDNTH DAY YEAR STA UNIT YEAR ((/
| |
| / NUMare
| |
| ((
| |
| // quPRER PONTH DAY YEAR 1 NR 1 Il 1 I4 8 18 0 16 0 12 8 IB -
| |
| 1 11 I 4 -
| |
| 0IO 1l2 *t 11 Ala I l l CONTACT FOR THIS DIR NAPE TEtEPHONE NupBER AREA CODC T. Tulon. Annistant Superintendent Doeratine Ext. 2213 8l1 l5 213I4I-15l4I4l1 COMPLETE ONE LINE FOR EACH COMPONENT FA URE DESERIBED IN THIS REPORT j CAUSE SYSTEM COMPONENT MANUFAC- REPORTABLE CAUSE SYSTEM COMPONENT MANUFAC- REPORTABLE l TURER TO NPRDS TURER TO NPRDS X NIN XIFINIR S 12 I4 15 YES I I l I l l !
| |
| l I I l l l l l I l I l SUPPLEMENTAL REPORT EXPECTED PONTH DAY YEAR p
| |
| SUBMISSION
| |
| ~l YES fif ves. commlete EXPECTED SUBMISSION DATE)[l ND TEXT Energy Industry Identification System (EIIS) codes are identified in the text as (XX]
| |
| p A. PLANT C0feITIONS PRIOR TO EVENT: .
| |
| \
| |
| Event Date/ Time 11-14-88 /1133 tinit 1 MODE 1 - Power Goeration Rm Power .,,_101. RCS [AB) Temperature / Pressure . Normal Doeratine Unit 2 MODE _Z__ - . Power Goeration Rx Power ,_D1_ RCS (AB) Temperature / Pressure Normal Operatino B. DESCRIPTION OF EVENT:
| |
| At 1133 on November 14, 1986 wIth Unit 2 at 53% Reactor power, the 2A Circulating Water pump tripped causing a decrease in condenser vacuum (due to an increase in hotwell temperature). It was not necessary to ramp down the Unit during this event due to the quick response of the Unit Nuclear Station Operator (N5041 censed) in starting the standby Circulating Water pump and one of the hoggers (mechanical vacuum Pump).
| |
| l Upon investigation of the 2A CW pump trip by a Equipment Operator (EO) the exciter field relay and feed breaker for the exciter were found to be tripped.
| |
| Nuclear Work Request B62317 was written to allow for further investigation of the trip by the Electrical l Maintenance Department.
| |
| All operator actions were correct and served to ritigate potential consequences of the event. No safety system actuations occurred during the event.
| |
| (0202R/0024R/122188)
| |
| | |
| BRAIDWOOD SIMULATOR MALFUNCTION 3~
| |
| ' ID:, ' AUX-12
| |
| | |
| ==Title:==
| |
| Loss of Service Air
| |
| 'NO: 6.3.4.1.12-Unicader valve. sticks open causing loss f]
| |
| | |
| == Description:==
| |
| | |
| of service air and instrument air..
| |
| V.
| |
| Variations: Date: 3/12/89 Rev: 4 Selectable Steps Inputs Comments 12A = Unit 0 compressor
| |
| : 1. _Salect compressor valve AUX 12A to fail AUX 128 125 = Unit 1 compressor AUX 12C 12C = Unit 2 compressor
| |
| : 2. Select value 0-100 percent open
| |
| : 3. Select ramp time 0-99,999 sec.
| |
| Brief, Plant Response (based: Plant being at full power)
| |
| Unioeder valve opens or opens gradually effectively acting as a-service air load without allowing makeup by the compressor. Thue, the service air and instrement air header pressure dropa. Eventually instrument. air will be lost. The first annunciator received is SAC RCVR PRESS LOW.
| |
| Suggested Instructor Action:
| |
| None.
| |
| Events:
| |
| None i
| |
| i l
| |
| L O 305M/83M/15 5/89
| |
| ------__---__._______m_ _ _ _ _ _ , _ , _ , , _ _ _ _ _ _ _ , _ _ _
| |
| | |
| BRAIDWOOD STMULATOR MALFUNCTION
| |
| | |
| ==Title:==
| |
| Essential Service Water Pump' Trip ID: AUX-4 t
| |
| NO: 6.3.4.1.4 l /~
| |
| | |
| == Description:==
| |
| Selected service water pump trips or fails to start due to a faulty breaker.
| |
| Variations: Date: 8/1/86 aev: 3 Selectable Steps Inputs Comments
| |
| : 1. Select faulty service. AUK 4A 4A - 1A SK pump water pump AUI4B 4B - 1B SK pump Brief Plant Response:
| |
| When the service water pump trips, the reduced service water flow causes an increase in temperature on those components and systems being cooled by the service water system. These include the component cooling water system, and various individual components.
| |
| Suggested Instructor Action: .
| |
| f~ To restart the tripped pump, first clear the malfunction.
| |
| 1 Events:
| |
| None s
| |
| GP
| |
| \
| |
| 1459B:4 305M/83M/7 8/86 L__.________________ _
| |
| | |
| BRAIDWOOD SIMULATOR MALFUNCTION
| |
| | |
| ==Title:==
| |
| Loss of Service Air ID: AUK-12 NO: 6.3.4.1.12 Unloader valve sticks open causes loss
| |
| .t
| |
| | |
| == Description:==
| |
| | |
| 'N of service air.
| |
| Variations: Date: 8/1/86 Rev: 3 Selectable Steps Inputs Comments
| |
| : 1. Select compressor valve AUI12A 12A = Comon compressor -
| |
| to fail AUK 128 128 = Unit 1 compressor AUIl2C 12C = Unit 2 compressor
| |
| : 2. Select value 0-100 percent open
| |
| : 3. Select ramp time 0-99,999 sec.
| |
| Brief Flant Response:
| |
| Unloader opens or' opens gradually effectively acting as a service air load without allowing makeup by compressor. Thus, the header pressure drops.
| |
| Eventually Instt. ament Air will be lost.
| |
| Suggested Instructor Action:
| |
| None.
| |
| Events:
| |
| None l
| |
| O 14598:4 305M/83M/15 8/86
| |
| | |
| BRAIDWOOD SIMULATOR MALFUNCTION
| |
| | |
| ==Title:==
| |
| Loss of Non-Essential Service Water ID: AUK-13 NO: 6.3.4.1.13
| |
| ,f~s
| |
| | |
| == Description:==
| |
| Selected non-RSW pump trip causing loss on non-ESW cooling.
| |
| (dJ Variations: Date: 8/1/86 Rev: 4 Selectable Steps Inputs Connents
| |
| : 1. Select pump AUX 13A 13A = OA WS pump AUK 13B 13B = OB WS pump AUK 13C 13C = OC WS pump Brief Plant Response:
| |
| Miscellaneous components such as oil coolers and non-essential coolers are cooled by non-essential service water. Temperatures will rise on these components and possibly require shutdown of components or power retuction.
| |
| Suggested Instructor Action:
| |
| Clear malfunction after students have identified, compensated for it, and
| |
| . ordered the malfunction to be repaired or returned to service.
| |
| Events:
| |
| Kore l
| |
| l 9
| |
| (g W 1459B:4 305M/83M/16 8/86
| |
| | |
| BRAIDWOOO SIMULATOR MALFUNCTION
| |
| | |
| ==Title:==
| |
| - Inadvertent Control Room Ventilation Isolation ID: AUK-17 NO: 6.3.4.1.17 l
| |
| y~~g
| |
| | |
| == Description:==
| |
| High smoke or high radiation occurs in
| |
| () the Control Room Ventilation Syster,.
| |
| Variations:
| |
| * Noter Selection 2 will function only
| |
| * Date: 9/19/88
| |
| * if train B of Control Room HvAC is in
| |
| * Rev: 4
| |
| * service. For train A of Control Room *
| |
| * HVAC, see Malf. RMS-1D.
| |
| * Selectable Steps Inputs Comments Select failure 1, 2 1 = High Smoke 2 = High Radiation Belef Plant Response: (IC-17,.100%, all systems in automatic)
| |
| High Smoke - Recire charcoal adsorber for both trains is placed in operation.
| |
| The only annunciator received is MCR IONIZATION HIGH.
| |
| High Radiatien - Recire charcoal adsorber placed in service, normal intake dampers shut and makeup air fan starts. No annunciator is received, however an RM-;l alarm occurs.
| |
| Suggested Instructor Action:
| |
| None.
| |
| Events:
| |
| : 1) LER 06-01-87-008: Inadvertent VC ESF Actuation (Set JPLCRI(x) = T where I = selected train).
| |
| : 2) LER 20-01-87-031: CR HVAC Shifts To Emergency Makeup Mode.
| |
| : 3) LER 06-01-87-021: Control Room Ventilation Actuation.
| |
| : 4) LER 06-01-88-001: Control Room Ventilation Actuation.
| |
| : 5) LER 20-01-88-001: Control Room Ventilation Actuation.
| |
| : 6) LER 20-01-88-010: Control Room Ventilation Actuation.
| |
| : 7) LER 20-01-88-011: Control Room Ventilatloa Actuation.
| |
| O 305M/83M/21 9/88
| |
| | |
| LICEN5EE EVENT SEPCRT (1.EP)
| |
| Factitty came (1) c, . Docket Numeer (2) _8 ne (1) ron. Of $l 01. Of 01 al Sl a 31 of 0l1 fkfbRED SAFETY F URE ACTU ION OF CONTROL #00M VENTflaTICN DUE TO PROCEDURAL INADEDUACY Event Cath (51 LER Number (6) Reeert Date f71 Other Facilities Involved f a)
| |
| Month Oay Year Year /
| |
| p 5eouential //j jj/ Revision Month Day Year raettity Names i Oceket Numeerfs)
| |
| I
| |
| / Number /// Number eyren. Unit 2 01 El el Of 01 el El E 011 214 al7 / al7 oIDIe Nclo w 014 211 al7 -
| |
| ' ~ M * *m t it tuaMf ?TMURSUANT TO THE REQUIREMENTS, OF 10CFR of El of of of al Q.8 (Check one er more of the followine) (11) -
| |
| o.
| |
| 6 20.402(b) , , , 20.405(c) L 50.73(a)(2)(tv) _ 73.71(b)
| |
| POWER ,_ 20.40$(a)(1)(t) .,,,,, 50.36(c)(1) 50. 73 ('s )( 2 ) ( v ) , 73.71(c)
| |
| LEVEL 20.405(a)(1)(11) 50.36(c)(2) 50.73(a)(2)(v11) Other (Specify (101 ol0 l 0
| |
| ,,,,, 20.405(a)(1)(itt) ,. 50.73(a)(2)(t) ,,,, 50.73(a)( 2)(viii)( A) in Abstract )
| |
| ////////////////////////// ,_ 20.405(a)(1)(tv) ,_ 50.73(a)(2)(til 50.73(a)(2)(viit)(B) below and in
| |
| ////////////////////////// ,,,,_ 20.405(a)(1)(v) SO.73(a)(2)(111) 50.73(a)(2)(x) Text)
| |
| LICENSEE CONTACT FOR THft LER *121 Name TELEPHONE NUMeER AREA CODE L. Sues. Asttstant tunerintendent Oneratine Ext. 2211 R l 1 l5 21 al 41 -l il 414l l COMPLETE ONE LINE FOR EACH COMPON N FAILURE DESCRIeE0 IN THIS REPORT (111 CAUSE SYSTEM COMPONENT MANUFAC- REPORTABLE CAUSE SYSTEM COMPONENT MANUFAC- REPORTABLE l
| |
| TURER TO NPROS TURER TO NPROS 1 1 I l l l l I l l I I l l l
| |
| , I I I W l l I I I I I I I I l SUPPLEMENTAL REPORT EXPECTED (141 Expected MontM l Day i verr' Submission lyes (If ves. comolate EXPECTED SueMIS$f0N DATE) l N0 l l l Af5 TRACT (Limit to 1400 spaces, i.e. approximately fifteen single-space typewritten lines) (16) t on Maren 24 1987, at 0252. shtft operating Personnel were in the process of switching the Main Control Room (MCR) Ventilation Systern from Train "A" to Train *B". The Main Control Room ventilation System was teing operated conservatively in its safeguard mode. In attempting to start the 'B' Train it unexpectedly automatically switched to its safeguard mode. The cause of the event was that a safeguard signal was present for the *B" Train at the time of the switchover. The cause of the signal is unknown. Corrective actions will revise the MCR Ventilation Start and Shutdown procedures to reset the safeguard actuation signal prior to starting or stopping a Train. Caution cards have been placed on the MCR handswitches until the procedures are revised. There have been previous inadvertent Main Control Room Ventilation Safeguard switchovers, however this is the first of this nature.
| |
| O V) c (131EM/0157M)
| |
| \
| |
| | |
| LICENSEE EVENT DEPCOT ftfo) TEXf CONTINUnif04 FACILITY NAME (1) DOCKET NUMBER (2) LER NUMBER f61 pue ( 21 Year 5eguential // Revision
| |
| /// Number /// Number 1 TEXT 0f510 10 l 0 l al Si a 917 01 Of 8 0 10 Of2 cr of3 B yron. Unit 1 - -
| |
| E ne. iii tr.Gustry Identification System (E!!5) codes are icentified 9n the text as (su)
| |
| A. PLANT CONDfff0NS PRf0R TO EVENT:
| |
| This event involve'd eouipment cononon to both Units I and 2. ;
| |
| Unit 1 MODE _i,,,, . Refuelina Ru Power N/A RCS (at] Temperature / Pressure 81*F /de-eressurtred Unit 2 MODE 1 - Power Onerations Rs Power & RCS (AB) Temperature / Pressure Normal oeerattne
| |
| : 8. DESCRIPTION OF EVENT:
| |
| on March 24 1987 at 0252 shift Operating Personnel (Licensed ) were in the process of switching the Main Control Room Ventilation System (V!](VC) from Train "A" to Train *B" to support surveillance testing of the
| |
| '18" Diesel Generator (EK). The Matn Control Room Ventilation System was betng conservative %y operated in its safeguard mode for unrelated safeguard circuitry concerns. Upon attempting to start the "B" fratn's supply fan, the Train unexpectedly automatically switched to its safeguard emergency mode.
| |
| Main Control Room Ventilation outside air intake radiation monitors (IL), which would generate a safeguard actuation signal, were insnediately checked and found not to be indicating any abnormal levels of radioactivity.
| |
| Tratn *B" was left operating in its safeguard mode. There were no systems or components inoperable that contributed to thh event. In addttion, there were no other safety system actuations. This event is a reportable pursuant to 10CFR 50.73 (a)(2)(tv).
| |
| (d
| |
| * C. CAUtf 0F EVENT:
| |
| The intermediate cause of this event was that a Control Room Vent dation safeguard signal was present for the "B" Train at the time the Train was started causing the automatic actuation. A safeguard actuation will only occur on a Control Room ventilation Train wnen it is operating. Tratn '8' of VC was shutdown for approximately 5 days prior to this actuation. Sometime during this time period a safeguard signal was generated. This signal will remain untti it is reset.
| |
| There are two sources for a safeguard actuation s*gnal for the Main Control Roam Ventilation system; a Safety Injection on either Unit or a high radiation signal from the respective Train's outstde air intake radiation monttor. With the "A" Train previously operating in the safeguard mode the actual cause of the signal for the 'B' Train is indeterminable.
| |
| D. $AFETY ANALYSIS:
| |
| There were no safety consequences of this event since the 'B' Train VC system actuated properly to its Safeguard condition. The 'A' Train of VC was also operable at the time of the event. The safety consequences would be the same had this event happened under any other credible set of initial conditons.
| |
| E. c0RRECTIVE ACTIONS:
| |
| Main Control Room Ventilation Starting and Shutdown procedures will be revised to reoutre the Safeguard signal to be reset prior to starting or stopping a VC Train. In the interim, while the procedures are being g revised. Caution Cards have been placed on the Main control Room handswitches requiring a reset ' prior to switch manipulation. Procedure revisions are being tracked by Action item Record 6-87-077.
| |
| (t348M/0157M)
| |
| | |
| '~
| |
| t!CENSEE EVENT nE70pt (trR) Test CcNyrNuatrey l FACILITY NAME (1) DOCKET NUMBER (2) LER quwgEp f6) pue f 1)
| |
| Year /// 5equential /// Revtsion
| |
| _ Number Number Byron. Unit 1 O l 5 1 0 l 0 l 0 I el 51 4 R17 - of 01 B - O l0 013 or O!)
| |
| [ TEXT Energy Industry Identification System (E!!S) codes are identtfted in the text as (xx]
| |
| F, PREVfDUS OCCURRENCE 1:
| |
| There have been previous inadvertent VC safeguard switchovers, however this is the fiyst of this
| |
| ~~
| |
| nature.
| |
| * G. COMPONENT FAILURE. DATA * -
| |
| a) MANUFACTURER N0MENCLATURf MODEL NUMBEg ArG PART NUMBER No components failed b) REtut71 of NPe01 SEARfM:
| |
| NONE N/A E
| |
| i f
| |
| s, l
| |
| l 6
| |
| l a
| |
| (134sM/0157M)
| |
| I .. - _ _ _ _ _ _ _ _ _ - . _ _ _ - _ _ _ - - _ - _ -_
| |
| | |
| , y
| |
| ,< - ,. i; .-1 .. - i...,;; i,
| |
| -s ni tPi. <: ** '/- v i tJ O t ? l t 'l; F .h: 1 f 1 8's :
| |
| l 'i i l; Aif. . n . , '. . t : ; /
| |
| CllE.bilt.fil' C n't I F. '.i l f .Oilli L 'I ll)H i
| |
| ( ) ."H.n it c . I'Uhn i I tiLilT 'l i s :
| |
| w/
| |
| C 01. Rili4 Y LOC (UEPT a PERSail: CElliRAL FILE UnTE SENi 07/3i/87 1 5 0 0 J '.: C i DVR 20-1-07-206; CoilIROL ROOri VEi!TILATION CHIFT TO THE EMERGENCY i ini( t::UP (10DE AS A RECULT OF SPtlRIUUS ACTUhTIuil OF A ItADIATtullil0NITOR D
| |
| 't i i l? Le i c:T I ON SEVERITY LEVEL: LER HO: 87-03i CRITERION:
| |
| WG UndSIllG [1 Ell: DW RESPUMUE DUE D6f E INSPECTOR llRS :
| |
| OR OR G / Pt. RCOH : UP /llAHER 10 U( SET .U Y SY S I Eit IL VF
| |
| '4NP LEPT /UUPV 1 GEL /ROUNTREE CURRECIIVE ACI c i d i!- 1ER90H / R/F OUIACE e s ei . 1rp nii / PRIORt1Y i.l t i t i EPSui! / 1R: #WHS C I 4 DY/DW PROCEDURE i o,.;"..i.1 1 Pia : C I:Y : D: L: O: Z: I!OD: Oflu : F09: PIC: (JFG :
| |
| F f ' d '_ R G ORIGIllAL DUE DAIE: 07/13/07 BIAIUS: CuiiPLh11:
| |
| : 1. I rlil< l li HEFUPr * ,
| |
| R D '( FOR CLOSURE 04/26 .,
| |
| 1,'IE;;'il t .E l 'OlWI i OCFR'50. 73 ( A ) ( 2 ) ( I V ) ORIG E /.1 T DAIE O / / O.V > : ,
| |
| ,, 1. Wa'll'G PEPORT
| |
| . 456/O'7-023 ORIG CLOGED 07/P :/f:.
| |
| (v ) 13f.:toTURE i E.E. FITZPATRICh DATE ColiPLETED 07/l' (rER .t:V R 'y.-- i 0 7 - 2 0 6
| |
| ! ARIPT'10H:
| |
| Ai i900 DH 06/13/87, 1RAlil OA OF COi!IROL ROOli VENT.I L (iT I Ui! Sil1 F 1 L P TU R GIRC ilODE DUE TO SPII' LNG Oil Coll!ROL ROOl1 OUIGIDE AIR 1 d f f.l; E Rett'ilATION HONITORS OPR31J AND OPR32J. F R 0ii illE Ril- i l f ilS iOltY LOG: OPR32J GAS CHAl!HEL UENT INTO ALARM AT 1908:41;
| |
| . P R ; i ..I f; . .M t.llA P H E L WEN 1 T.O ALLRT AI i 7 0 0 : 4 '.' ; 6 P R3'2 J I' AR 1 i CUL ATF M it if 411,L i'Ei! f TO ALARM Ai 1700:47 AND (iGAIN hT 1709:17. PluiPI R Vein ILATION SUI TCHOVER UAG VERIFTED. ALL. t:H ANUELG RE'll.lRilED TO HORhnL BY 1910:19. OPR32J PAR TIEULnTE FILTElt WAG REi10VED n oit D ailN I ED . NO RAD)OACT[VITY l Y.:t ullb .
| |
| OPic 14 A'l It!G ENGI NEER ' G ColitiEi!I S :
| |
| M NN1 E ODD L FIOilAL 8 R. YUNGK 06/14/07
| |
| .,t i'U!! kEPORinPLE EVEN1 p ,
| |
| v ;
| |
| i I
| |
| I l
| |
| l
| |
| : 4. .s su ,
| |
| > ;;- ,.a
| |
| , ~ 3 vih P.1 PT I Oi! ( L i.o 1T -
| |
| b OC Fi Ut-1 Gui1 Hoi 4Y :
| |
| TEXT
| |
| .x A. PL Ar!I' CUNDITIONS PR [UR ID l~ VENT x,
| |
| U N .t f DknIDiJUOD i; EVENT DAIE JUi!E 13, 1987; EVENT lihE 1908 NODE 3 - HOT STAHDBY RX POUER 0% RCS (nB)
| |
| TENiERATURE/P'iESSURL 391 DEGREES F/1300 PS1G w
| |
| 4 B. DESCRIPTION OF EVENT K
| |
| ). IJU r.t tnPUt80 rlTS OR SYSTEMS WERE Ill0PERAPLE AT 1HE BEGir4tlING OF 1 HE LVEllT UH1CH COffl RIDUTED 10 THE EVEllT.
| |
| 1
| |
| > ni 1 000 ud JUNE 13, i907, IT UAB DISCOVERED THROUGH HAIt! cot 1 TROL Rourt AlH!!!ilCI AIIOri THA1 'l R A I tl On UF ~lllE COtt1ROL ROOM VEHfILATIuit S ib i lC rl ( V :l ; ilAD SHlFTED TO I'IS ErlERGLriCY MAKEUP h0DE OF OPERATIOei.
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| > THE ACIUATION NAS DOE IO A HICH RADIAf1011 SIGNAL FROM THE OPR32J vet!TIL ATION R ADI ATIOl1 N0H l'i OR (IL) SAMPLIt!G FROM THIS 1 RAIr1 OF VErli IL Al T ON.
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| ) '
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| .t r' ni 19iO ALL Morf1 TOR CitAt!r!!!'L fiCTIVITY READINGS RETURi>ED 10
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| , ?!O P h A L AND THE L1HEUP FOR CONTItOL Ruaii VENTILA I ION SY5 ITEM
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| > (JAS St4;SEQUENTL Y RE'I UPNED 10 flu fulnL . 'lllE HOUITORS R E ri A 1. >E D OPER6DLE filit0UGHOUT Tile DURAT[ON OF THE EVEt1T.
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| x
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| ). t i ,'u l 1 COrit'I t 10:43 REnn liti o O fio.l E I lip ot ic,I lO U T lilE DURNI I f itt OF 1HE EVlii Tt . Ol'ER A I OR ACT l ut19 all' t l ilER lt! CREASED OR DECREAGEp Tite SEVERIIY OF THE EVENT. lHIS EVENT IS REPORTADL.E PURSUAi!I
| |
| ) TO iOCFP50.73(A)(2)(IV) - ANY EVElli OR CONDITION TilAT RESULTED 1 r4 tini!U AL Of: All f Onn l'IC oCfunfION OF AHY l'NG J NL"E R F L t, A FE T Y FE A lllRE , J NCLUD1l!G THE REriCTOR PROTECTION SYOfEM.
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| .g.
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| C. CAUSE OF EVENT:
| |
| > x Tile ROO T C AllSE OF T HE E d;..:a lT t)AS A RADIO KEYED LJITHIN iHE DEGIGNATED EXCLUSION AREA i1 EAR RADIAT10H HONITOR OPit32J. Il lin D
| |
| ) DEEN PREVIOUSL Y DEMut4STRAIED fHAT KEYING A RADIO (JITillH THE EXCLUSION AREA PRODUCES Tile It4STRUMENTATION RESPONSE O DW I''v E D IN IllIS E f 2N1. 'I HE ELECTRIC CURRENT PULSE GENERATED
| |
| ) BY THE RADIO IS A eOISE SIGNAL UHICH ULTIHATELY GETS TPANSLA1ED INIO AC1IV11Y BY THE hum.lTOR. A REVIELJ OF THE GECURITY ACCEG9 GbMPUTER LOG I,l A S PER F ORiiE ') TO IDENTIF/ ALL PERSONNEL LN THE
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| ? '') :
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| r 16: G 0 H '".Ulirin p Y <1,iliarT
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| +
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| s
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| . Nc. . i UI Of R. >t. . lite Ui ll i PEv%UN li' il't .kl i: (4.l l H n P no lO W(.B A GifL ilR I l Y GUARD (Cull f R Af' TOR ) . T!!E lff!!P t TY GUnRD WAS lii'i CRVIElJED AND S tal ED THAL IHE RAD 10 Wns NOT lt liY E D Wi l b fiJ liiE EXCLUSION AREA. THE GUARD DID !!OTE THnT THERE tJAS A PUN .L BIL I T Y' 1 H AT THE RADIO MIGHT HAVE DEEN 1HADVER TEri1 LY REYED AG A RESUL1 UF "DUHPING". THE AlSLE IS <!nRROW ENOUGH Thal
| |
| :: :'E"SOr! HiW I MG IllROUGH THE ROOM hUST ErlfER WITH 10 FOO1 EXCLUS10H nl2EAS OF THE HUMITURS RENDERING THEH SUSCEPT[ULE TO h D .t. S E Ii>DUL LD AC i LIA11 Oti .
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| w D. B AFETY - (u 84L (S I S :
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| litERE WAS ;'O I f iP ACT OH PLAH1 SAFEfY, BECAUUE T11ERE WAS
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| # 10 AC'iUAL R Ais l O ACT I V I. T Y PRE 912.H I . A S6HPL E OF 'IIIE P ART ICULn1 E F :li.11 R Oil '<Pi!32J WAS TAKEN AT 2040 HOURS br THE RADIA1.lut4 CHEMISIRY DEPARTHENT THIS SAriPLE (F1.LE l018613665) VERIFIED l i k' ) (HERE WAS HD RADIDACT1VITY. 1F illIS EVEti f MAD OCCUltPED IN REGiOrlSE I O 111CH RAD I on :T1V r t'l f , 'lllE SY'11 CM l ' l-:S P O ? "; E trodLD HnVE tit.::EN Tite SAME. THROUGilOU T ' file DURAIIOff UF f l it E'/ E H T , lllE ilEDUtlDANT CHANNEL OPR31J RAliIn110H MONITOR Uns OFEP%LE. ,
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| E CORRECTIVE ACTIONS:
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| s_.
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| S t il! COM1PiiL RuuM V ENT IL A r i fn1 S v SIEM L1HEttP WAS RETURilED r0 linRHR _
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| THis E'/EH1 IS COtlSIDERED TO L3E AH ISOLnTED OCCURREllCE AllD AS A RESULT NO ADDITIONAL ACTION IS COrl f EMPL A TED.
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| . (.
| |
| M F. FREVIOUS OCCURRENCES:
| |
| i Hl;RE HAVI: UCEH NO PREVIOUS OCCUltRENCES DUE TO KEYIt1G A
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| . knD10 C4USli!L nN EilGINEERED '3 Al ETY FEnfuRF ACTU AT 1ui! .
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| 't j
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| C FALLED COMPOrJEHT DATA: j j
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| i0 E(4U1PMEl#1 FAILURE 4( )
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| - - E M I)- --
| |
| "'.J'' F LEliEH i i TO DVil 20- 1 -0 7 -206 /l..ER 87-031 i
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| l 4
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| V i
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| L__ _ _ _ _ _ _ _ _ __
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| . . . 50 E, . Ji' . .
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| t.,U rif inR Y (CONT)
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| ] E I 10t!
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| .J 1,..
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| ,y A. l'L Ai!1 CONDI(10NG t R i Uli TO C.vkNi Ot TtlRR ENCk i Uis! 'l l DRAThWOOD i; EVEt! f DATF IUt1E 13, 1907; EVE (!T fliiE 193D 110DE 3 - Il0T STAR 4DBY; RX POWER: 0%; RCS (AB) TIIli PCP ATURE/ FFD> lure 39i DEGREEU F/i300 PSIG
| |
| * O L L ui'.9 E f !W ;/ :
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| Uil l i DRAIDWuGD 1; EVElli DATE OCTOGER 16, 1737. EVLHT 1 J flE 11 fiODE 2 ST ARTilP ; RX PUHEll: 3%; IPCS (AB) T EMP ER A TilRE/ PRES Gilp r- a ,
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| T'L C R t.:ES F /2?'M PCIG 4
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| * ri-U ., DESCRIPTION UF EVEllT
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| * OCCURRENCE I w
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| ni 1908 DN JUNE 13, 1 V d i', I t UW3 f>ISCOVERtID 'I HROUCil Coil f ROL P 0011
| |
| * ANNiltJ C I N IOil Til AT TRAIN GA OF CONTROL RfJOM VEHTILa TJ UN sis 1Eh (V1) f] ut!J F 7 ED TO ITS EMERGENC/ rink EUP MODE OF OPERATION DUE fu A H i f. H
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| () RADiATT0li GIGNAL FROM illE Ol'R32J VEN f ILATION RADI AT10;! hui! I i Ol' util'L.1 r!G F H nh IHIS :RAi1 OF V E r4 I I L A T i O H . ?40 DEGRADED Tt P UC h irl t J. i i.
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| FAILED C0hPONENI'S CONik f t:UTED TO THIS EVENT. PL Ar4T C th usI l l ud 'i liit6 L NED G f aul.E f HH OUC elWI INE DUR AT IOi! OF THE EVFH1 O P E.R o T h u
| |
| * eCilONG HAD NO I t1F LUE t:1.E Gil lilE SEVERTTY OF Tile E V L t>I - nLL lo ii il 1 in Lth>WNEL nLtiV1TY R t- r:1> 1 : ': ItE IIIRNED TD NOReiAL Di i9iO IIIA I EVI i!Ii' file n o tJ I T O k G W E R E N t'.VI.R PECLARED IHOFERnBLE lilROUGHOUT THE Dt)Ra! !.'is
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| . t :t-T Hl:. LVENI.
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| p.
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| OdtlRREtlCE ?:
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| 4 AT 112o OH OCTODER 16, 1987, THE ti A I ri COr4 TROL R00i1 OUT51DE Are .l i ! ) ; . i F IRAlf4 OA RADIATION MONT. TOR, OPP 3i ! CP [lMD INTO fi1Gli RADI A T JnN @ .,?..
| |
| CAU91tlG THE IRAItf 'o' V E H i tl.. n 1 T O N OF lHE CONIROL ROOH TO cil.llrf it i EtiERGENCY MAKEUP MODF OF OPERATION. 2119 EVENT UAS ALEO D1F' J . 'I o V1A CONIROL ROOH AllNUNCIATION. T ilFR E WERE NO DEC MMEli 91 tM l' + ' '
| |
| * FAILED COMPONENTS C0tifh10UlJUG TO fitE EVENT, AND AGAIN, i l:Ef< E m., i e:)
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| L:F F EC T ON PLAl4T CollDI f L ONS. Oi'ERATOR ACTION Hob ild i M V .-i C f i. :i ! !;er SEVErtITY OF THE EVENT., CHAT!NEL ACT1VLTY READINGS RE FUD t!ED 10 ! i;i i,J L
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| * DY i152 lil AT HORNING. TilIS I IliE , THE h0t4Il0R WAS (. ECL ARE L 1 din ' i.
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| PENDING Tile OUTCOME OF NilCLFhR WORK REQUEST it A 1669 4 . LCOAR bi w 3.J 1-iA HAL ENTERED. Till iiUNirOR W/.fs RETURNED TO 9ERVJN Af ! v' 9
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| '# utiierER 3. , 198i, si il{SE OCCUP.RE i!CES ARE 1.EING PEPORTED PURSUANT TO 10CFRSO.7 3 ( A ) ( .? > ( J V )
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| f.N Y EVEiEr OR CONDITION TilAT RESULTED IN HAi4UAL OR AU T OiMTIC AC'Un110N OF AH( Ei!GINEERED 04FL~ TY FEATURE, INCLUDING THE REACTOR PROTEC1 ION T/STEH.
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| ~
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| $. G,USL OF CVEi!i'
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| .+
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| O f.$ t, ', ! k II. E. h !I.. k( i x
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| Tile HOO T CAUGE OF THE EVENT WA3 A RADIO BEIMG KEYED DEPEr.iEDLY HEAR THE OPR.X2J RADIATLOH HONITOR. IT IS CllSPECTED TllAT SE r;UR I TY PI: l" a.fi h lEl . Mui b-L I CEHOh D COrli t' AC I OP , 1HADVERIEHitY KE(ED (i RADIO 1:! i l lt
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| ]l1MEDIAIE AREn OF THE RnD1ATION M uf4 T T U R , THE LLECTRIC CURRENI P UI.. G E GE NEl': n TED Tt Y THE RADIO IS A HUISE 91 C iIn L HiliCH UL11f1AIELY GEis ill Ai!Gl . A TE D :IHTO ACiIVITY ltY fIIE M0ilI i OP . ALSO, THE AJSLE IS UnRlWU 1: "Jt %il f i t -I A A P El'.O O H HOV li'G l i lP.f il P ;H t i !!! l '. O Oi l llos 10 P o S O u t' fill.ll li l8 iCEi OF IllE MONIfuRS, li AK li!G II f .. n '.11 E R I tJK A MOH)iOR TO PJCKUP A PAbIU
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| 'i f 1 V E .
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| t Qi',. '' lilld.E t!!M ? :
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| (\> ) l HF PODi CAUS OF 1 H .19 EVEl!I LS UNK rlOldH . II IS 5'IROWLLY D F L I f..V E D in L:E A linD10 DEIttG KEYED W ITHIt' file E/CLUSION AREA OF THE OPR31 J .
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| U E '.. ~ :U N E I!:E hAGNITUDE OF )dE SPlVF NA3 '30 IIJ Gli AS to IND1CalE vi b .V 10 1 R ANCri1G510H MEAR THE MUNITOR. l'REV10U9 EXPERIEilCE OF A SPECIFIU COhr OMEilT FMLURE FOUND Tils t T H E tlO I V SPIKE INDUCED n SPURIGUS ACTIVITY 91GNAL OF, ROUGHLY, TUICF Tile IIIG H R AD.i AT 10N SI"1 P u lt!i . Til "i l'lL * *PTO ekt 16 I N C .l D E rlT , THE t!OISE $PI!:F WAG ALh0Sr a DECADE OVEP i t 'I:
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| GE 'i P O 1 N i' . ALSO, INVESTIGATION UNDER NUCLEAR UORK REQtlEST ta i 6a91 E H
| |
| * il i : 11 0ROrCR OPERATION OF THE MONIVUR. FURIHER, A GECUR1fY L api' I of
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| !! i 3 lTP Y , TnMEN VERY SOON AFTER Tl!E EVEHf, 51 TOWED G E VITR AL C O N il'i ' 'I'lH t 't:.R - t.q ii lL L blE R E PRESEf!T IN T Hli iknii l 'h' HVAC P.O O ri ei file FIriE Ol' ! ! !!-
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| EVElli .
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| h 'Oh b ! [ l 'l f $ , ,S e y
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| O t.uiR R Ei !L E i
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| 'i l lif p F IdM el O I Pj P f:.' C l UN PL Ai!T 09 PU 9LII; SAFET( BECnUSE 1HERE WAL s us ACTUAL ACiJVI1Y PRESENT. A SAhiLE OF illE PARiICUL/4[E FILTER OU UPP 2J IJ/a l ni.Ep r. f 2046 I!OURG Pr INE Re iOI A f .10:1 CllE h1 S I R Y DEPARTMENI, Am 1H19 5.At1Pl.E (FILE 10106i3 MS/ VChtFIED iilAT i!!E PE WA9 Fl m
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| ,,- ,t .:
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| I :i.I . *
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| * 38' e i t:'t 1 Ort Suf enARY ' UUi!T )
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| '' n o l f i.:il l I V [ l i iF THIS L V Ei fi ilAD OCCURRED DUPI 4G con:T: R C I.< I. i i. , a < i ul'ERn r10f t , THE GAnf Ci ti't,il ilie.riC E'.i Wl'lIL D linVE ARI9 fin. l l tir uuG ilin i f it!
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| ! i JKt . ! .I Ort OF 'lllE EVES t I , i I :' O!''R H J KnD J.n i 10t1 riOnI TOk !ni , n V.T L i n Q . t i i. -
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| P.EDutJD ANT COVER AGE; TilIS siUf!Il OR ALSO REG I S'I EltED THE HUISE GPa.l'E.
| |
| OCCURRENCE 2:
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| x AL Ai r!, THERE WAS !!O [MPAC f Or4 l'L Ai4 I OP. PUBLIC SAFE'l( bi! C A U S E r. i W.Y -
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| ESTABLISHED THAT ACTIVIlY DlDN'T CAUSE THE ACTUATION. ALSO, lilE Ru tdnJD ANT Ol'R32J W AG OPE R'il l uf f AL 'lllR Olig H O U T Tile DUP. ATION UF IIII' EVEi1T .
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| .".( l j
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| E. L ut< RI- C f I VE AC f IOilO :
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| Oct UnPENCE 1 l NO iTi8kECT1VE AC T IOr4 UAS FAKEH 1: R O f1 liiIS EVENT, Sli!CE lilIS (J.M cot!SIDERED AH ISOLolEO 1:WlAf4CE OF PERSONNEL ERROR.
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| A OCCURRENCE 2 4
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| NUCLEAR tg..t R K REQUEST : A i e J. 94 UAC WRIIIEN rd IN ITI ATE 'I R OUDLE3110G l l OC BY INSTRUhENT MAINTENANCE PERSOIIrlEL. TECHNICI ANS VEKXF L ED Tlii I' i ; u.:
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| O noril10R UAS FUNC TIOr4ING PRO!'ERLY.
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| OBSERVATION F OR 1HREE DAfG N 1. Ti l ilO IRREGULAR DEHAVtOR.
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| lIAG lllEN KCIUhhED [O SFHV.lCE AT 2246 Or! OCTOBER 20, 1987.
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| THE ciONITOR WAS THEti F.itit Di Wl:..
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| Tl!E t'ii 8IIUK
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| +
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| t 'UUSI L. [L L f 1 E9 OF F U I' lilE R I P O I L t, I I t!G litESE COr1 TROL R O Ori V E N i j l ,.l iti.i t'iOili rullS Abril HS I I!.kR(.fi t li ri ti .i t s l' I .:11, 6kL CURREllTLY DE li!G I: ?!'L U h T D .
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| 3 F ,. PREVIOUS OCCURREHCE5' BOTH OCCURRENCES OF liIADVL RT ENT ACTUATION DUE TO A RADIO f:E If M i i: 'i !" r ARE LISTED IN THIS REPORT.
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| n G. f. OriPONEr1T FAILURE DATA w
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| DUhil 4
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| . . g n p -.
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| 9 s
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| k
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| | |
| LICE 2SEE EVE;iT REFC2T (LER) gy y - / ")
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| Facility Name (1) occket Number (2) l ana f31 avran unit 1 el El el of al al El 4 1 larlol3 l VteTILATIM affuaTIM out To offfRf atiffou SYSTEM VOLTAst TRAmsttsf Lassa M OFFifft tIhr fif PPED tvant Rata is) Lt3 ~ r f81 Ranart Data (7) Othae Fac111tian involved fal Month Bay Year Year // Sequential // Revision Month Day Year Factitty = I oncket m rfn1 f * ' - -
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| * 3
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| / /
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| avaam umff 2 al El el el of als is
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| ~ ~
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| al e 11 7 al 7 al 7 oi211 aie eie 21 s al 7 of El of of al i I g,, lTHIL RCPORT 25 $WMITTED PtR$UANT TO THE REQUIREMENTS OF 10CFR
| |
| ""I IN 1 I 2ee's ana er anre af the fallautaal iIll
| |
| ' '* 48U b) te.405(c) .X. se.73(a)(2Hiv) _ 73.71(b)
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| P0mR _, za. costa)(1H t) Es.36(c)(1) se.73(aH2)(v) _. 73.71(c)
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| LEvtL 20.405(aH1 Hit) se.34(cH 2) le.73(aH2Hvt1) Other (specify fini a!e!7 ,,. 20.405(a)(1)(tit) St.73(aH2)(1) .,_ le.73(a)(2Hv111)(A) in Abstract N _
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| 20.405(a)(1)(tv) 20.aelta)(1)(v) _
| |
| se.73(a)(2 Hit) se.73(aH2)(ttt)
| |
| _ 58.73(aH2)(vitt)(a)
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| St.73(aH2)(x) tfrratar comTAff FM TNf t LEE f121 below and in Text) game Trt ra*aaaa mLasara AttA CISE N.*--- 1M aev Annurmaea *~ m itar tut. 22am ai1 lE 21 11 al .I El al all I CfmWLtTE out List FM rarM cfornaraf FAf t ter attratara in TMit REPET flin CAUSE SYSTDI CWe'Outui MAmuFAC- REPORTABLE CAUSE SYSTEM CIBr e taf MANUFAC- REPORTABLE /
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| Timen To mPant Times to apens i I I I I I I I i l I I i i i lui i l l I I I I I I a l tirPLEDENTAL REPGRT EXPtetta ital Expectet h th I Day l Year
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| ( } Submission lyan fif van. emmeleta EXPffftD ""f ttfou Raft 1 1 I mR Date (15) l I I l lI l A85 TRACT (Limit to less spaces, t.e. appreminately fifteen single-space typewritten Itnes) (16)
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| On September 17.1987. at e612. with Unit 1 in power operetten (Mede 1) at 971 reacter pouer and Unit 2 in pouer operation (Mode 1) at e3% reacter power. precess radiatten mentters OPR31J (Main Control Room Outside Air Intake 'A*) and OPR32J (Main Control Reen Outside Air Intake 'A') sensed an underveltage condition and transferred to the inter 1cek mode. The interfack signal transferred the mata control roen ventilation system to its Engineered Safety Features configuration. A trip of an offsite 34eKV transmission line j caused a voltage transtant on the Comnenusalth Edtsen grid editch caused the mentters to transfer to the (
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| interlock mode. Modifications to louer the underveltage trip setpoints were installed in 1985 and have reduced the monitor's sonsttivity te voltage transients caused by large pesup starts and most grid disturbances. This is considered an isolated occurrence and ne further corrective action is planned at this time. Aa event stetlar to this has occurred in the past (LER e6-e26-es).
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| O '
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| i (165eM/etteM) j 4
| |
| | |
| t fCENSEE EVENT REPORT f tER) TEXT r0NTINuaff0N J FACILITY NAME (1) DOCKET NUMBER (2) LER NUMBER f6) pace (1) I Year / Sequential /// Revision
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| //p/
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| // N^r p/// Number Byron. Unit 1 0'l C l 0 l 0 1 0 l 41 El 4 ai7 - 012 l1 - 0l0 01 2 0F of 1 '
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| l TEXT Energy Industry Identification System (EIIS) codes are identified in the text as (xx] 1 1
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| A. PLARY_ Con #fitons raion_.in EVinT.
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| I -
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| Event Cate/ Time 9/17/a7 / 0612 Unit t MODE l - Pnmer naaration Rx Power 1ZL , RCS(A8] Temperature / Pressure Marmal anaratina !
| |
| Unit 2 MODE 1 - Power anaration Rx Power & RCS(AB] Temperature / Pressure Normal eneratina
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| : 5. description DF EVENT:
| |
| On September 17, 1987, at 0612. with Unit 1 in power operation (Mode 1) at 171 reactor power and Unit 2 in power operation (Mode 1) at 831 reactor power. process radiation monitors opt 31J (Main Control Room Outside Atr Intake 'A') and OPR32J (Main Control Room Outside Air Intake 'A') (It] sensed an undervoltage condition and transferred to the interlock mode. The interlock signal automatically transferred the Main Control Room Ventilation system (V!] to its Engineered Safety Features (ESF) (JE] configuration. Operators verified the Main Control Room ventilation System ESF lineup. Both process radiatier, monitors tamediately returned to normal operating condition once the voltage trans'ent passed. Operator actions were correct during this event and the plant was maintained in a stable condition during this event. No plant systems or components were previously inoperable that could have contributed to this event. This event is reportable per 10CFR58.73 (a)(2)(iv).
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| R C. tAUSE OF EVERT:
| |
| i' I A trip of an offstte 345CV transatssion Itne (line 0621 - Syron Station to the Cherry Valley Transmission Substation) caused a voltage transtant on the Commonwealth Edison grid. The bus voltage sensed by the j monitors momentarily dropped below the undervoltage trip setpoint of 9013 VAC imich is 75% of nominal bus voltage. This voltage drop caused the monttors to transfer to the interlock mode. There was no personnel error involved in this event.
| |
| D. 1AFETY AMALYSII:
| |
| There was no effect on plant or public safety. The transfer of the Main Control Room Ventilation System to the makeup mode of operation is an ESF actuation which establishes filtration of radioactive contamination from the air supplied to the main centrol room. At no time was the filtering capaht11ty required since no airborne activity existed during this event. The monitors immediately returned to norinal operating condition once the voltage transient passed. The safety consequences would have been the same if this event had occurred under a more severe set of initial conditions.
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| \
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| C '
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| (1650M/0196M)
| |
| | |
| LICENSEE EVENT DEPCRT (LER) TEXT CONTINUATION FACILITY NAME (1) DOCKET NUM8ER (2) LER NUMeER f61 pace (1)
| |
| Year // 5equential // Revision p/j/p/
| |
| / Number ,/,p/
| |
| // N d ar avron. unit 1 o i 1 ! o I o I o I al si a al7 - oi2l 1 - o Io of 3 aF of 1 l TEXT Energy Industry Identification System (E!!$) coces are identified in the text as (xx]
| |
| E. C.QlL9EflVLACT10k1:
| |
| Modifications were installed in 1985 that reduced the undervoltage trip setpoint on all process and area rae11stion monitors from 100 VAC to 90 VAC. Past experience indicates that the setpoint modification has reducsd the monitor's sensitivity to voltage transients caused by large pump starts and most grid disturbances. No further corrective action will be taken since this is considered an isolated occurrence.
| |
| F. PREVIOUS QCtuttikCES:
| |
| LER NiseER lilL1 86-026-00 Control Room venttistion Due te Lightning Induced Distribution System voltage Transient G. COMPONENT FAILtRE DATA:
| |
| a) MARKlFACTLEER NOMEMCLATLRE tME S klBGER MFG PART ML3GER NOt Applicable b) RESULTS OR hPROS SEARCH:
| |
| Not Applicable tu l l (1658M/0194M)
| |
| | |
| 40k-1]
| |
| LICENSEE EVENT REPORT (LER) j Facility Name (1) Docket Number (2) Page (3)
| |
| Byron. Unit 1 01 51 01 01 01 dl 51 4 1lofl0l3 Mfh NAh NR^Wf Nk SkKbONMSUN O cvent Date (5) LER Number (6) Report Date (7) Other Facilities Involved (8)
| |
| Month Day Year Year /// Sequential / Revision Month Day Year Facility Names l Docket Nunter(s) fj/j/
| |
| / Nunber
| |
| //j/
| |
| ff
| |
| // Number
| |
| ' I BYRON, UNIT 2 01 51 01 01 01 41 51 5 01 2 01 9 81 8 81 8
| |
| ~~
| |
| 010l1
| |
| ~~
| |
| 010 l l l 0151010101 l i THIS REPORT IS SUBMITTED PURSUANT TO THE REQUIREMENTS OF 10CFR
| |
| ^
| |
| (Check one or more of the followino) (11)
| |
| M DE W I 20.405(c) 50.73(a)(2)(iv) 73.71(b) 20.402(b) _ _X_ _
| |
| POWER _
| |
| 20.405(a)(1)(1) _ 50.36(c)(1) _ 50.73(a)(2)(v) 73.71(c)
| |
| LEVEL 20.403(a)(1)(ii) 50.36(c)(2) 50.73(a)(2)(vii) Other (Specify 0l9l8 _ __ _ _ _
| |
| (10) _ 20.405(a)(1)(lii) _ 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) Delow and in
| |
| /////////j/j//jj////////jj}//
| |
| / / / _ ,
| |
| : 20. 405 ( a ) ( 1 ) ( v ) _ _ 50.73(a)(2)(lii) __,50.73(a)(2)(x) Text)
| |
| LICENSEE CONTACT FOR THIS LER (12)
| |
| Name TELEPHONE NUMBER AREA CODE Fred Hornbeak. Senior Staff Engineer Ext. 2822 J8l115 21 31 41 -l 51 41 All COMPLETE ONE LINE FOR EACH COMPON FAILURE DESCRIBED IN THIS REPORT (13)
| |
| CAUSE SYSTEM COMPONENT MANUFAC- REPORTABLE CAUSE SYSTER COMPONENT MANUFAC- REPORTABLE TURER TO NPROS TURER TO NPRDS I I I I I I I I I I I I I I I IC I 1 I I I I I I I I
| |
| - SUPPLEMENTAL REPORT EXPECTED (14) Expected Month i Day I Year Submission
| |
| [ lyes (If yes, conclete EXPECTED SUBMISSION DATE) Tl NO l ll ll l ABSTRACT (Limit to 1400 spaces, i.e approximately fifteen single-space typewritten lines) (16)
| |
| On February 9, 1988, at 0103 with both Byron units in power operation (Mode 1), there was a spike on the process radiation monitor for the Main Control Room Outside Air Intake. This spike caused an automatic transfer of the main control room ventilation system to its Engineered Safety Features (ESF) configuration. Samples were taken and no actual radioactivity was found. Following troubleshooting and monitoring of the system, the cause of the spike was deemed indeterminable and the ventilation system was returned to the normal operating mode. There have been other noise spike induced ventilation system actuations at Byron Station.
| |
| O I I (1915M/0206M) 4581 8/88
| |
| | |
| LICENSEE EVENT REPORT (LER) TEXT CONTINUATION FACILITY NAME (1) DOCKET NUMBER (2) LER NUMBER (6) Pace (3)
| |
| Year /// Sequential // Revision fj//j
| |
| / Nurter /jj/
| |
| j// Number Byron. Unit 1 0l510l010l41Sl4 8l8 -
| |
| Ol0l1 - 010 01 2 0F 0l3 fTEXT Energy industry Identification System (E!!S) codes are identified in the text as [xx]
| |
| A. PLANT CONDITIONS PRIOR TO EVENT:
| |
| Event Date/ Time 2-9-88 / 0103 Unit 1 MODE 1 - Power Operation Rx iower 98% RCS [AB] Temperature / Pressure Nomal Operatino Unit 2 MODE 1 - Power Operation Rx Power 80% RCS [AB] Temperature / Pressure Nomal Oper0 ting B. DESCRIPTION OF EVENT:
| |
| On February 9, 1988, at 0103 with Byron Unit 1 in power operation (Mode 1) at 98% reactor power and Byron Unit 2 in power opeastion (Mode 1) at 80% reactor power, process radiation monitor OPR33J (Main Control Room Outside Air Intare 'B') (PR) [IL] alarmed CHANNEL IN HIGH ALARM on the main control room radiation monitoring display console. This alem was due to a spike on the gas channel. The spike also caused an interlock signal which aukunstically transferred the main control room ventilation system (VC) [VI] to it's Engineered Safety Features (ESF) configuration. Radiation Chemistry pulled a grab sangle which verified no radia. ion was present. No plant systems or conponents were pretiously inoperable that contributed to this event. The plant was maintained in a stable condition during this event. All operator actions taken were correct. This event is reportable per 10CFR50.73(a)(2)(iv).
| |
| C. CAUSE OF EVENT: g The cause of the gas channel spike on process radiation monitor OPR33J is unknown. It was detemined from y/ the duration of the spike that the spike was probably electronically induced noise. A strip chart recorder was connected to process radiation monitors OPR33J and OPR34J to monitor the 120VAC supply voltage and the gas detector high voltage power supply on each monitor. This was done to aid in determining the cause of the spike. No cause was found, and the c.ause has been deemed indeterminable. There was no personnel error involved in this event.
| |
| D. SAFETY ANALYSIS:
| |
| There was no affect on plant and public safety. The transfer of the Main Control Room Ventilation System to the makeup mode of operation is an ESF actuation. The actuation establishes a safer plant condition by allowing for the filtering of radioactive contamination from the air that is supplied to the main control room. At no time was the filtering capability required since no airborne activity existed during this event.
| |
| E. CORRECTIVE ACTIONS:
| |
| The Instrument Maintenance Department visually inspected the detector connectors and connected the strip chart recorder to 0FR33J and OPR34J. The monitors did not spike while the strip chart recorder was connected and observation of the trend displays on the radiation monitoring display console revealed steady trends. The strip chart recorder was disconnected on February 18, 1988, and the VC system was returned to its normal operating configuration. No further corrective *.;tions are warrented at this tirne.
| |
| l O
| |
| l 1 (1915M/0206M) 4581 8/88 L_-____-_- . _ .
| |
| | |
| LICENSEE EVENT REPORT (LER) TEXT CONTINUATION FACILITY NAME (1) DOCKET NUMBER (2) LER NUMBER (6) Page (3)
| |
| Year /// Sequential // Revision fj/j
| |
| // Number j/
| |
| /j/j
| |
| / Nunber 8vron. Unit 1 0 l 5 1 0 1 0 l 0 l Al 51 4 8l8 -
| |
| 010l1 - 0l0 01 3 0F 01 3 DEXT Energy Industry Identification System (EII5) codes are identified in the text as [xx]
| |
| F. PREVIOUS OCCURRENCES:
| |
| The following LER's. detailed previous control room ventilation actuations due to spikes on radiation nonitors. This large number of events in 1984, 1985 and 1986 was attributed to construction activities in the plant. Construction activities have now ended and noise suppression circuitry has been added to the radiation nonitors, thus reducing the number of spurious ventilation actuations.
| |
| LER NUPSER 84-027-00 84-026-00 84-033-00 84-038-00
| |
| .85-010-00 85-088-01 85-099-01 -
| |
| 86-002-01 86-007-01 86-024-00 G. COMPONENTFAILUREgTA:
| |
| a) MANUFACTURER NOMENCLATURE MODEL NUPSER MFG PART NUPBER Not Appilcable b) RESULTS OF NPRDS SEARCH
| |
| * Not Applicable l
| |
| C) l l (1915M/0206M) 4581 8/88 u_________-_______ _ _ _ _ _ _ _ _ . - - -
| |
| | |
| " /Q Lt.h Il E R A I I"400 D-OPE R DVb LATE .
| |
| ACT1.ON ITEM 20-i-80-00500 ' w;E ?3 lEri NO : 456-200-88-00500 01HER UNIl NO:
| |
| ITEM CA TE - 01/08/88 .BCHEDULAR CAT IEST CONDITLON i MODE: 5 COMMITMEt!i 10:
| |
| CURRENT LOC (DEPF A PERGON: CENTRAL FIL.ES DATE SENT: ')2/17/88 >
| |
| SUBJECT DVR 20-1-80-005; 1 RAIN A t'ONTROL ROOM RADIATION MONITORING INOPERABLE DUE TO NOISY PRESSURE SWITCHES TYPE: DEVIATION SEVERITY LEVEL: LER NO: 88-001 CRITERION- SUP ORG CAUSING ITEM: BW RESPONSE DUE DATE INSPECTOR HRS:
| |
| ORIG ORG/ PERSON: OP /NALEWAJKA r0 BY SET BY S(STEM ZZ VI RESP DEPT /SUPV . TSEL/ROUNTREE CORRECTIVE ACT COG PERSON / R/F OUTAGE COG. PERSON / PRIORITY COG PERSON / TR: BD7IM BY/BW F'ROCEDURE$
| |
| TRANSMII: BW: BY: D: L: Q: Z: NOD: DNS: ESS: PTC: NFS:
| |
| R
| |
| , '9WER: 0 ORIGINAL DUE DATL 02/07/88 STATUS: COMPLETE I
| |
| lTERIM REPORT:
| |
| * RDY FOR CLOSURE: 01/19/83
| |
| ' INTERIM REPORT: 10CFR50.73(A)(2)(IV) ORIG EXIT DATE 02/02/38 CLOSING REPORT: ORIG CLOSED 02/06/88 SIGNATURE K.L.KOFRON DATE COMPLETED 02/i7/88 REFER: DVR 20-1-88-005 DESCRIPTION:
| |
| AT 0950 ON 01/08/88, A SPIKE OCCURRED ON OPR032 RAD MONITOR (THE GAS CHANNEL) WHICH CAUSED AN ESF ACTUATION TO OCCUR ON THE OA VC TRAIN.
| |
| THE DA VC TRAIN CHARCOAL ABSORBER BYPASS DAMPER FAIL.E0 TO CLOSE. NO EP0XY PAINTING WAS DONE IN THE AREA DURING THE PREVIOUS 24 HOURS.
| |
| 10CFR50.72 N RED PHONE NOTIFICATION MADE AT 1116. 01/06/88 OPERATING EN ER'S COMMENTS:
| |
| 4 NONC. ROBERT J. UNGERAN 01/09/88 NOTIFICATION: RESIDENT INSPECTOR, NRC REGION III, 01/11/88, 1000
| |
| * T. J. MAIMAN/D. P. GALLE, VP/NSD, 01/11/88, 1000 7_.s.
| |
| 30 DAY REPORTABLE /10CFR50.73(A)(2)(IV)
| |
| \ i i
| |
| | |
| f lbRNIDWOOD-GfiER DVR i:( ' .- 1 :., :.
| |
| [ JACTIUN ITEM 20-1-00-00700 ' P4W
| |
| '( CON T ,
| |
| [jEM''NO: .456-200-88-00500'
| |
| [M' DESCRIPTION (CONT):
| |
| h
| |
| * t LER.NUMBERi 88-001
| |
| . ACTION
| |
| | |
| ==SUMMARY==
| |
| A.- PLANT CONDITIONS PRIOft TO EVENT:
| |
| OCCURRENCE 1.
| |
| - . UNI 1 BRAIDWOOD is' EVENT DATE: JANUARY 8, 1988; EVENT TIME 0950 MODE: 5 COLD SHUTDOWN; RX POWER: 0%; RCS'(AB)
| |
| -TEMPERATURE / PRESSURE: 107 DEGREES F/0 PSIG x.
| |
| OCCURRENCE 2 ~
| |
| UNIT': BRAIDWOOD 1;' EVENT DATE: JANUARY 11, 1988; EVENT TIME ~ 1230 MODE: 5 - COLD. SHUTDOWN; RX POWER: 0%; RCS (AB)
| |
| TEMPERATURE / PRESSURE: 103 DEGREES.F/0 PSIG 0CCURRENCE 3 UNIT: PRAIDWOOD 13 EVENT DATE: JANUARY 12, 1998; EVENT' TIME: '1024 MODE: 5jrOLD.. SHUTDOWN; RX POWER: 0%; RCS (AB)
| |
| TEMPER ATURE7 PRESSURE : 101 DEGREES F/31 PSIG I
| |
| * B. DESCRIPTION OF EVENT:
| |
| .THE CONTROL-ROOH VENTILATION SYSTEM (VI) WAS DECLARED INOPERABLE A~
| |
| 2130 ON JANUARY 1, 1988 FOR UNRELATED SCHEDULED WORK ACTIVITIES.
| |
| THERE WERE NO OTHER SYSTEMS OR COMPONENTS INOPERABLE AT: THE BEGINNING
| |
| 'OF THE EVENT WHICH CONTRIBUTED.TO THE SEVERITY OF ANY OF THE OCCURRENCES LISTED BELOW.
| |
| OCCURRENCE 01 AT 0950 ON JANUARY 8, 1988, IT WAS OBSERVED VIA THE RADIATION MONITORING SYSTEM CONTROL CONSOLE (RM-11) (IL) THAT THE TRAIN A CONTROL ROOM AIR INTAKE RADIATION MONITOR OPR32J (IL) SPURIOUSLY SPIKED INTO HIGH RADIATION ALARM, CAUSING lAN AUTOMATIC ACTUATION OF THE TRAIN A CONTROL ROOM VENTILATION SYSTEM'tVI) TO ITS EMERGENCY M AKEUP ' MODE DE .0PERATION. DURING THE OCCURRENCE, THE CHARCOAL ABSORBER BYPASS DAMPER OVC43Y FAILED TO CLOSE. IT WAS DETERMINED THAT THE ACTUATION WAS-SPURIOUS, AND THE MONITOR WAS IMMEDIATELY' RETURNED TO SERVICE. 'A NUCLEAR WORK REQUEST WAS WRITTEN TO INVESTIGATE THE DAMPER FAILURE.
| |
| * THE APPROPRIATE NRC NOTIFICATION VIA THE ENS PHONE SYSTEM WAS MADE AT 1116 ON JANUARY 8. 1988, PURSUANT TO iOCFR50(B)(2)(II).
| |
| THIS OCCURRENCE IS BEING REPORTED PURSUANT TO 10CFR50.73(A)(2)(IV) -
| |
| y ANY EVENT OR CONDITION THAT RESULTED IN MANUAL OR AUTOMATIC ACTUATION
| |
| | |
| BRAIDOODD-GPER OW g. i , i, -
| |
| :ACTTON ITEM- 20 88 -:)o:5 0') e . GE -
| |
| En NC: 456-200-38-00500 CCONi)
| |
| ACTION
| |
| | |
| ==SUMMARY==
| |
| (CONT.):
| |
| OI'ANY ENGINEERED SAFETY FEATURE, INCLUDING THE RFACTOR PROTEC ION.
| |
| SYSTEM.
| |
| c.,
| |
| OCCURRENCE #2 AT 1230 ON JANUARY 11, 1988, IT WAS OBSERVED VIA CONTROL ROOM ANNUNCIATION THAT THE TRAIN A CONTROL ROOM AIR INTAKE RADIATION MONITOR OPR31J (IL) GENERATED A NOISE SPIKE, SENDING Ils MEASURED RADIATION LEVEL INTO HIGH ALARM. THIS CAUSED AN AUTOMATIC Ac run fICri 0F,THE TRAIN A CONTROL ROOM VENTILATION SYSTEM TO ITS EMERGENCY MANEUF MODE OF OEPRATION. DURING THE OCCURRENCE, CHARCOAL ADSORBER DYPA3S-DAMPER OVC43f AGAIN FAILED TO CLOSE. MONITOR WAS RETURNED TO SERVICE AT i338-ON JANUARY 16, 1988.
| |
| HE APPROPRIATE NRC NOTIFICATION VIA THE ENS PHONE SYSTEM WAS MADE Al-1309 ON JANUARY i1, 1988, PURSUANT TO 10CFR50(B)(2)(II).
| |
| THIS OCCURRENCE IS BEING REPORTED PURSUANT TO 10CFR50.73(A)(2)(IV) -
| |
| ANY EVENT OR CONDITION THAT RESULTED IN MANUAL OR AUTOMATIC ACTUATION OFANYENGIgEEREDSAFETYFEA'IURE,.INCLUDINGTHEREACTORPROTECTION.
| |
| SYSTEM.
| |
| \ OCCURRENCE #3 AT 1024 ON JANUARY-12, 1988, THE OPR32J MONITOR SENT ANOTHER SPIAE WHICH PROPAGATED AN ENGINEERED SAFETY FEATURE ACTUATION SIGNAL TO CAUSE THE TRAIN A CONTROL ROOM VENTILATION SYSTEM TO SHIFf: TO Ils EMERGENCY' MAKEUP MODE OF OPERATION. AGAIN-, DAriPER OVC43Y FAILED TO CLOSE. FOLLOWING THIS OCCURRENCE, TRAIN A 0F CONTROL ROOM VENTIL.Ai10b WAS MAINTAINED IN ITS EMERGENCY MAKEUP MODE.
| |
| n THE APPROPRIATE NRC NOTIFICATION VIA THE ENS PHONE-SYSTEM WAS MADE AT 132 't ON JANUARY 12, 1988, PURSUANT TO 10CFR50.72(B)(2)(II).
| |
| THIS OCCURRENCE IS BEING REPORTED PURSUANT TO 10CFR50.73(A)(2)(IV) -
| |
| l ANY EVENT OR CONDITION THAT RESULTED IN MANUAL OR AUTOMATIC ACTUATION OF ANY ENGINEERED SAFETY FEATURE, INCLUDING THE REACTOR PROTECTION
| |
| ~
| |
| SYSTEM.
| |
| * y OPERATOR ACTION NEITHER INCREASED NOR DECREASED THE SEVERITY OF THE EVENT. PLANT CONDITIONS REMAINED STABLE THROUGHOUT THE EVENT.
| |
| C. CAUSE OF EVENT: a n
| |
| THE CAUSE OF THE RADIATION MONITOR FAILURES WAS DUE TO AN INTERM1Tiri*
| |
| PROBLEM WITH THE MONITOR'S PRESSURE SWITCHES. DURING HODULATION CF THE MONITORS' FLOW CONTROL VLAVES, THE SWITCHES EMIT ELECTRICAL NOIf1
| |
| | |
| ' BPAIDWOOD-OPER DVR Dr i- .; -
| |
| ACTION ITEM ''' 0 - 1 ~ 8 E - G M 0 0 e 5, t '
| |
| EN NO: 456-200-38-00500 (ConTi tic 1ID'N
| |
| | |
| ==SUMMARY==
| |
| -~ (CONT)
| |
| '5PIVES WHICH.. UL. TIM ATELY ARE IN TERF RETED AS R ADI ATION PROPAGA rliD CURRENT PULSEG BY THE MOMITURS' PREAMP' CIRCUITRY.
| |
| THE CAUSE OF "THE CHARCOAL ADSORDER DYPASO DAMPER OVC43Y FAILURE IS
| |
| 'STILL UNDER INVESTIGATION. LT 13. SUSPECTED THAT THE CAUSE IS RELATED TO'A MFCHANICAL-INTERFERENCE IN THE AC'10ATOR. SHOULD ANY ADDITIONAL INFORMATION REGARDING THE MODE OF FAILURE BE DETERMINED, IT WILL 0E DOCUMENTED'IN A SUPPLEMENT.TO 1HIS'RFPORT.
| |
| A k
| |
| D. SAFETY ANALYSIS:
| |
| -THERE WAS NO IMPACT ON PLANT OR PUBLIC SAFETY AS THERE WAS NO ACTUAL RADIOACTIVITY-PRESENT DURING-THE DURATION OF THE EVENT. CGNTROL ROOM-VENTILATION HAD BEEN INOPERABLE FOR SCHEDULED MAINTENANCE ON THE CHILLERS SINCE JANUARY i, 1908 AND THE ASSOCIATED TECHNICAL '
| |
| SPECIFICATION ACTION REQUIREMENTS HAVE BEEN COMPLIED WITH AT ALL TIMES. HAD THIS EVENT OCCU,RRED UNDER. WORST CASE CONDITIONS'0F ACTilAL liADIDACTIVigTY PRESENT WITH THE UNIT IN POWER OPERATIONS, TRAIN B 0F-PERFORM CONTROL ROOM ISOALTION, THEREFORE, MITIGATING THE CONSEUQEWJE3 O'..CONTROLROOM.VENTILATIONWASAVAILABLETOBE OF THIS TYPE OF EVENT, AND NOT COMPROMISING THE SAFETY OF THE PLAN T OR PUBLIC.
| |
| * E. CORRECTIVE ACTIONS:
| |
| x THE IMMEDIATE CORRECTIVE ACTION WAS TO DETERMINE THAT THE SOURCE UF THE' ACTUATION WAS SPURIOUS IN NATURE AND NOT DUE TO ACUTAL RADI0 ACTIVITY.
| |
| 4 ACTION TO PREVENT-RECURRENCE INCLUDES ADDING NOISE ATTENUATING FILTEPS TO THE PPESSURE SWITCH CIRCUITRY TO ELIMINATE THE FALSE TRIPS. THESE FILTERS WILL BE USED AS NECESSARY TO ELIMINATE ANY FURTHER SPIKING PROBLEMS WHICH MAY OCCUR.
| |
| HE CHARCOAL A5SORBER DAMPER OVC43Y ACTUATOR WILL BE REPAIRED DURING
| |
| .THE. CURRENT U441T GUTAGE. THE ACTUATOR HAS BEEN DISASSEMBLED IN AN ATTEMPT TO DETERMINE'THE MODE OF FAILURE. SHOULD ANY ADDITIONAL INFORMATION REGARDING THE MODE OF FAILURE DE DETERMINED, IT WILL BE DOCUMENTED IN A SUPPLEMENT TO IHIS REPORT. THIS WIL.L BE TRACKED 10 COMPL.ETf0N BY ACTION ITEM- 456-200-88-00501.
| |
| K F. PREVIOUS OCCURRENCES:
| |
| n O " "" -
| |
| | |
| 9;; A 1.DWOO D-OPER I V!'. i '
| |
| ACTION I(EM 20-i h3 -- O O''i O O 5
| |
| ' AI r . :?
| |
| 'E ci NO- 456i200--88-00500 (CCHT)
| |
| ACTION
| |
| | |
| ==SUMMARY==
| |
| (CONT) y G. COMPONENT FAILUPE DATA:
| |
| M.nNUFACTURER NOMENCLATURE MODEL NUMBER MFC PAR 1 (JUMBER x
| |
| GENERAL ATOMICS PRESSURE SWITCH x
| |
| ITT DARION ACTUATOR NH95 NH95G2602Pi03 4
| |
| .- - - E N D - --
| |
| ?
| |
| N m
| |
| .y$ , ,
| |
| | |
| . p - 17
| |
| . B P A I I;t.,0 0 0 ~ G *E R O (..-
| |
| ACFION ITEh 'O- '
| |
| N JOO ,
| |
| (hM N '.' . C 6 -200 ;8-- J 2 300 U a -i n 4 U sIi NU:
| |
| ITEM DAfE: 04/10/89 ZilFDUtAR CAT TEST COND(f1GN i MODE 5 CONhlTMENI 10:
| |
| C UR R!"N T LUC ( U T 'r o PERSON: CENTRAL FILES DniE SLNT 05 '95G
| |
| * SURJ'IC1 DVR 20-1-08-C03, CUNTROL ROOM VEN T [1.a f ION TO iiAREUP MODE OF OPERATION F()R n A HIGH R AD T.aTI()N ALARM DUE TU INCORRECi SETPOINr f Y F fi - DEVIATION SEVERIT( LEVEL LER NO: SG-010 CNITERION ORG CAUSING ITEM: BW RESPONSE DUE DATE INSPECTOR llR S -
| |
| ORIG ORG/PERCON: OP /UALRATH TO BY SE1 Bf SYSTF'1 VI V1
| |
| 'HESP DEPT /SUPV TSEL/STANCZAK CORRECTIVE ACT COG PERSON / R/F Ou t.".G E COG' PERSON / PRIORITY COG PERSON / TR: F01IM BY/BW FROCEDURE!
| |
| TRANSMIT PW: Pg : D: L: Q. 'Z - NOD: DNS: ESS: PTC: NFS:
| |
| ( '?WER: 0 ORIGINAL DUE DATE: 05/10/88 STATUS. Colwt . TE V '.'4TERIM REPORT
| |
| * RDY FOR CLOSURE: 91/17'00 INTERIM REPORT iOCFR50.73(A)(2)(1V) ORIG EXIT DATE 04/20<59 CLOSING REPORT: ORIG CLOSED O'5 / 0 3 / U D SIGNATORE R.E. OUERIO DATE COMPLETED C$/20'89 RFFER: DVR 20-1-08-083 DESCRIPTION:
| |
| ON 04/10/80 AT 0103, THE OB CONTROL ROOM VENTILATION SYSTFM AUTO SWAPPED TO THE tiAKE-UP/ADSORBER MODE DUE TO A HIGH RAD SIGNAL RECIEVED cROM OPRO33J PARTICULATE CHANNEL FOR THE CONTROL AIR IN f AKE MONITORIrtC SYSTEM. AFTER FURTHER INVESTIGATING AND DISCUSSIONS WITH TECH STM F ,
| |
| q IT WAS DISCOVERED THAT THE SETPOINT FOR OPR033J PARTICULATE MONITOR WAS INCORRECT. THE SETPOINT SHOULD HAVE BEEN 1,00E+10 BUT WAS 1,00E-10. THIS WAS IN THE COiJSE RVAT] VE DIRECTION AND IHE PLANI A.% {
| |
| PUBLIC WERE IN NO DANGER. A! 0242, ANOTHER ACTUATION Ci(GNAL MAS )
| |
| RECEIVED AS THE SETPOINT POINT FOR THE MONITOR WAS NOT CHANGED BACl; DUE TO COMMUNICATION PROBLEMS ON THE TWO PM-ti PANELS. fME SE1Pulttr WAS CHAPCEO SUCCESSFULLY AT 0252 AND THE RAD CHANNEL WILL D E h 0 N J i s'. i F f' FOR ANY FUR'IHER PROBLEMS.
| |
| i CCFR50.7 NRC RED PHONE NO11FIC AT ION tiADE , 04/10/09, 0305 j OPER A11 NG ENGINEER ' S ConnENTS :
| |
| []
| |
| V l
| |
| 1 1
| |
| I
| |
| | |
| 7 I. F aI PWDO C - Ci1R DVR L, .
| |
| AC T i till ITEM ./.) 1;U - X ^ J ^ v., r ,
| |
| )Ed i'O ~ 456-200 '33-00300
| |
| ! ( C C d .~ )
| |
| DESCOIH ION (CONT).
| |
| + .,
| |
| ESF r.CTU:41 T ON '# S D UIL 10 .;N IMPRUPER GL f11 NG IN THE DA1A F:ASE H. J. LEGNER 04/11/C8 NOTIFICATIONS: RFSIDErlT INSPECTOR, NRC REGION III, 04/11/38, s .0
| |
| * T. J. MAInAN/D. P. GALLE, VP/NSD, 04/11/88, 1: 90 30 DAY REPOR1ABLE/10CFR50.73(A)(2)(IV)
| |
| LER NUMBER 88-010 ACTION SUMi1ARY:
| |
| A. PL ANT CONDITIONS PRIOR TO EVENT:
| |
| n UNIT BRAIDWOOD 1 ; EVENT DATE MARCH 21, 1908 ;
| |
| EVENT TIME: 1955 MODE: 1 - POWER OPERATION; RX POWER: 23%;
| |
| RCS (AB) TEMPERATilRE/ PRESSURE: 564 DEGREES F/2235 PSIG W
| |
| b v
| |
| F.
| |
| DESCRIPTION OF EVENT:
| |
| UNIT i NUCLEAR STATION OPERA 10R (NSO) REPORTED TO THE COGNIZANT TECHNICAL STAFF ENGINEER THAT COMPONENT COOLING (CC)tCC) SURCE TANK (1CCCIT) LEVEL INDICATION DTD NOT CORRESPOND TO DRAINING OPERATIONS AT THE TANK. TECHNICAL STAFF (TS) INVESTIGATION OBSERVED 1 HAT WHEN THE '1B' PORTION OF iCCoiT WAS BEING DRAINED, THAT THE '1A' MAIN CONTROL DOARD (MCB) INDICATOR, iLI-670, WAS SHOWING A LEVEL DECREASE. THE iCC0iT IS A ' SPLIT' TANK WITH ONE PORTION SUPPLYING THE 'iA' PUMP AND THE OTHER THE 'iB' POMP. THE TANK IS SPLIT BY A DAFFLE PLATE THAT EXTENDS TO THE 40% LEVEL IN THE TANK (SEE ATrACHMENT 11).
| |
| n THL TS INVESTIGATION CONFIRhED THE NSO CONCERNS BY DETERMINING THAT iLIT-670 (LEVEL TRANSMITTER), LOCATED ON 10 SIDE OF i C C 0 1 T. , WHICH FEEDS iLI-670, IS L ABELLED AS THE 'iA*
| |
| CC SURGE TANK LEVEL. THE 1 LIT-676, LOCATED ON IHE 'iA' 91DE OF 1CC01T, WHICH FEEDS iLI-676, IS LABEL. LED AS THE 'iB' CC SURCE TANK LEVEL. CAUTION CARDS WERE PLACED IN MAIN CONTROL ROOfi AND AT iCCOiT ON 3-17-88 TO ENSURE PROPER DRAIN VALVES ARE OPENED AND (1AIN CONIROL BOARD LEVEL INDICATORS ARE ODSCRVED, FURTHER INVESTIGATION ON 3-21 -88 REVEA' ED THAT THE i LI T-670 AND 1 L I T-6'76 INSTRUMENT POWER SUPPLIES A5tE POWEPED UFF OF THE f>PPOSITE ELEC1RICAL DIVISION THAN THE ASSOCIATED
| |
| /' CC PUtiP WHICH IS RECEIVING SURGE TANK LEVEL IND'. CATION
| |
| \ %,)
| |
| | |
| ~
| |
| 9 ORAIDWOu0cOiG -
| |
| DVR %E i iACTTONLIIEh 20-1y89 OH MO cW . E 3 WN. . . . _ ,
| |
| ( EM.NQ: '456-2000 0$~9300 .CONTi
| |
| (
| |
| +
| |
| ! C, TION
| |
| | |
| ==SUMMARY==
| |
| MCONT)
| |
| (EKAnPlE: ' iiL I TE670 IS ACTUALLY THE '1B' I.EVEL INDICATION BUT IS -SOWERED FPOc1 Tt:E 'iA' TRAIN POWER. SUPPLY). THE LOW LEVEL PUMP TRIP SN170Hi4 WERE F OUND TO. BE WIRED /LABELI ED CORPECIL Y (SEE ATTACllMENT fr 2 ) STABLE PLANT CONDI TIONS WERE M/itN TAINED THROUGHOUT THE EVENT.
| |
| n
| |
| .u C.-CAUSE OF EVENT +
| |
| n THE ROOT CAUSE OF-THE EVENT WAS A DESIGN ERROR DY THE ARCHITECT ENGINEER (AE),: TEGTING ERRORS BY THE BRAIDWOOD-TESTING' ORGANIZATION CONTRIDU1ED TO THE EVENT.' -THE FOLLOWING EVEN TS LED' TO THIS EVENT: WESTINGHOUSE LETTER CAW-4151/CDW-3396 WAS ISSUED,ON MARCH 24, 1982 PERTAINING TO THE ELEC TRICAL POWER SUPPLIES FOR THE. BYRON /BRAIDWOOD UNIT 1, 2 CC PUMPS. THE LETTER INDICATED THAT THE ARRANGEMENT OF CC
| |
| , PUMPS;WAS UNACCEPTABLE DURING POST-ACCIDENT ALIGNMENTS. THE LETTER REQUIRED THE *iA' AND '1B' CC PUMPS 10.HAVE'THEIR RESPECTIVE g4160 VOLT' POWER SUPPLIES INTERCHANGED. NO
| |
| .. MOVEMENT OF PUMPS AND PIPING WAS REOUIRED. THIS CHANGE MADE THE '1A' CC PUhP BECOME THE '1B' CC PUNP ELECTRICALLY AND THE
| |
| 'iB' BECOME THE~'1A' PUPP ELECTRICALLY.. CHANGING OF PUMP LADELLING, VALVE TAGS AltD INSTRUMENT NUMBERS DUE TO THE CHANGE WAS TO BE .HANDL ' D BY THE AE THROUGH APPROPRI ATE DESIGN CHANGES. .ALSO. LOW LEVEL 1CC01T PUMP TRIP SWITCHES WERE~
| |
| I NST ALL E') ALONG WITH DESIGN CHANGES REQUIRED DUE TO THE PUMP CHANGE.
| |
| THE DESIGN CHANGES ISSUED BY.THE AE WERE PROPERLY INSTALLED IN THE FIELD, HOWEVER, THE AE CHANGES DID NOT TAKE INTO ACCOUNT-1HE INSTRUMENT POWER SUFPLIES FOR iLIT-670 AND
| |
| :1 LIT ~676. THE MCB LADELLING. CHANGES FOR 1LI-670 AND iLI-676, THE-HIGH/ LOW LEVEL ALARMS, OR TFE SIGNIFICANT EVENTS RECORDER ALARM MESSAGES NECESSARY TO PROPERLY CONFIGURE THE CC SURGE 1ANK.
| |
| THE DESIGN CHANGES OMISSIONS WERE NOT DISCOVERED DURING F3 ELD
| |
| ' VERIFICATIONS OF INSTALLATION. THE TESTING PHASE PERFORNED BY THE STATION TESTING ORGANIZATION INCLUDED A PRETEST REVIEW, INSTRUMENT CALIBRATIONS, TEST PERFORMANCE AND f*0ST-TEST REVIEW OF CC SYSTEM FUNCTIONS AS SPECIFIED ON El.ECTRICAL AND INSTRUMENT SCHEMATICS. THE TESTING OF THE
| |
| .fN$1RUMEN15 iL.I, LIT-6?O/676 DID NOT UNCOVER THE OMISSIONS BECAUSE THE 1EST WAS WRITTEN TO CORRESPOND TO THE INSTRUMENT NUMDERS AND Al. ARM IFPUTS 1 HAT EXIS1ED PRIOR 10 ISSUANCE OF WEST INGHOU3tE
| |
| * LE TTER . THE INSTRUMENT NUMDERS AND ALARM INPUT
| |
| _=_-_ _ _ _ _ _ -
| |
| | |
| l ORA [ DWOOD- Of'Lt' 7 '/P .ii' ACT10N !1EM '0-1 a v810 ,
| |
| g-( )En NO: 4 ~.,6 - 2 0 0 - 0 8 - 0 9 3 0 0 ( C th ! ! ) !
| |
| i l
| |
| ACTION
| |
| | |
| ==SUMMARY==
| |
| (CONT) ;
| |
| l rltaNGES NERE.NOT T NCMPOR AT E D IN THE DS. SIGN CHANCES, T!:US thFY WERE UOT DISCOVERED DURING THE CC PRE-GPERAT10NAL 1 ES T. l Thr lESTING DID UNCOVER DR AT N VALVE NUMf:FR ERRORS , BUT DID j NOT INVESTIGATE ANY FURIHER S[dCE fME TEST STEPS WERE l O THDlW I SE COMPLETED AS CRIGINAfLY WRITTEN. CONTRIBUTING i FACTORS TO fiiE L ACK UF DISCOVERY APE THE LOCATION OF THE iCC01T RFtATIVE TO THE UC PUMPS. lHE iCC0il PIPE ROUTING LOCATION TO RESPECTIVE CC PUMP SUCTIONS, AND THE LACK GF LABELLING UN r.ND AROUND THE 1CC01T. THE REVIEWS CONDUCTFD Bi CONSTRUCTION AND SIATION PERSONNEL PRIOR TO RELEASING SYSTEM FOR OPERATION A!ST' DID N01 UNCOVER THE DFSIGN AND TFSTING EPRORS.
| |
| AN ADDITIONAL CAUSE OF THE EVENT WAS THAT THE BRAIDWOOD-1 CC PRE-OPERATIONAL TEST WAS ESSENTIALLY A DUPLICATE OF THE BYRON-1 TEST. THE BYRON-1 TEST WAS COMPLETED PRIOR TO THE ISSUANCE UF DESIGN CHANGES AND HAD ONLY A FEW TEST DEFICIENCIES THAT REFERENCED THE DESIGN CHANGES. 1HE CHANGES WERE INADVERTENTLY NOT INCORPORATED DURING BRAIDWOOD TEST p)
| |
| (
| |
| LJ PREPARATION.
| |
| D. SAFETY ANALYSIS:
| |
| THERE WERE NO SAFETY CONSEQUENCES ASSOCIA1ED WITH THE EVENT DURING NORMAL PLANT OPE.R A r t ONS . THE CC SYSTEM IS OPERATED WITH BOIH TRAINS AND BOTH PORTIONS 1CC01T CROSSTIED DURING NORMAL OPERATIONS. SHOULD A LARGE ENOUGH LEAK OCCUR, GREATER THAN MAKEUP CAPACITY, THEN THE REACTOR WOULD BE TRIPPED, REACTOR CDCLANT PUMPS SHUTDOWN, AND THE UNIT WOULD BE COOLED TO A HOT SHUTDOWN CONDITION WITHOUT THE AID OF CC FLOW. THE CC SYSTEM LEAK COULD BE ISOLATED AND THE CC SYSTEM COULD BE RESTARTED. UNDER WORST-CASE CONDilIONS, THE CC SYSTEM WOU L.D DE OPERATED IN A SPLIT-TRAIN ALIGNMENT TO MEET POST-ACCIDENT, DESIGN BASIS LQSS OF COOLANT ACCIDENT REQUIREMENTS. IN THIS CONDITION, IF THE *iA' TRAIN HAD DEVELOPED A LEAK THEN THE
| |
| 'iB' MCB LEVEL INDICATOR WOULD SHOW A LEVEL DECREASE. THIS CGNDITION MAY PROMPT THE UNIT i NSu TO SHUTDOWN THE '1B' CC PUMP WHILE THE '1A' CC PUhP WOULD TRIP DN LOW-LOW CC SURGE TANK LEVEL IF BRAIDWOOD ABNORMAL OPERATING PROCEDURE, iDWOA PRI-6 " COMPONENT COOLING t%LFUNCTION*, WAS FOLLOWED. 1HUS, A POTENTIAL LOS9 OF ALL UNIT 1 CC WAS POSSIBLE DUE TO THE EVENT FOR INDEIEPMINATE PE.RIOD OF TIME DURING A POSl-ACCIDENT ALIGNMENT.
| |
| ,m
| |
| | |
| y BRAIDWOOD-OPER DVR oc :. W,4 w ACT10N ITEM 20 -i - 03 ')P 300 P .1 6 E - M IEM NO: 456-200-88-08300 ( Corrr >
| |
| ACTION
| |
| | |
| ==SUMMARY==
| |
| (CONT)
| |
| E. CORRECTIVE ACTIONS:
| |
| * 1. THE UNIT i SURCE LANK MCB AND LOCAL INDIC ATIOr45 PLUS i *- DRAIN VALVES HAVE CAUTION CANDS ATTACHED TO ALERT l
| |
| * OPERATING PERSONNEL TO 1HE PROPER INSTRUMENTATION TO
| |
| * MONITOR DURING TANK l.EVEL CHANGE EVOLUTIONS.
| |
| * 2. MODIFICATION M20-1-88-021 HAS BEEN COMPT.ETED TO REDESIGN
| |
| * LEVEL INDICATION CABLES AND MCB LABELLING.
| |
| * 3. 1CC01T WILL BE LABELLED AS 1HE 'iA* AND '1B' PORTIONS OF
| |
| * THE CC SURGE TANK. THIS WILL BE TR4,CKED TO COMPLETION
| |
| * BY ACTION ITEM 456-200-88-07301.
| |
| F. PREVIOUS OCCURRENCES: ,.
| |
| NONE .
| |
| }
| |
| s
| |
| * g
| |
| ) G. COMPONENT FAILURE DATA:
| |
| 'n J u.
| |
| NONE h
| |
| ***END***
| |
| .sg-fY _ ,y o .
| |
| ,r'
| |
| \ b v
| |
| . - - ~ - - _ _ . - - . -
| |
| | |
| y
| |
| \
| |
| &X*I)
| |
| ~
| |
| . ?
| |
| . 'l . . ha I, %
| |
| . ACTION ITE.M 2 i - i " O S --'A B C 0 F , . C ':
| |
| .EM tiU: 456-200-88-08800 0 f MEf' UNff NO: - -
| |
| ITEM DATE 04/15/88 SCHEDULAR CAT T E t:,1 CONfeITION .
| |
| tiODE: 5 COMti t TMEN T 10:
| |
| CURPENT LOC (DEPT PAC PERSON: BERRY PATE SENT: 06- < o ., ) ,,
| |
| SUHJECT- DVR 20-1-80-008; CONTROL ROOM VENTILATION SHIFT TO EMERGENCY ,
| |
| li AK E UP (10 DE DUE TO SPURIOUS RADIATION MONITOR NOISE SPIh. -
| |
| !YPE: DEVI A TION SEVERITY L.EVEL: LER NO: 88-01i CRIICRION:
| |
| ORG CAUSING ITEM: BW 'ESPONSE DUE DATE INSPECTOR HRD.
| |
| OR .l G ORG/ PERSON: CP /WALRATH TO BY SET BY SYSTEM VI VI RESP DEP T/SUPV TSEL/STANCZAK CORRECTIVE ACT COG PERSON A /DE R/F OUTAGE COG' PERSON / PRIORITY COG PERSON / TR: XIEEL4IM BY/BW PROCEDURE:
| |
| TRANSMIT: DW: B(: D: L* Q: I: NOD: DNS: ESS: PTC: NFS:
| |
| WER: O ORIGINAL DUE DATE: 05/15/88 STATUS: COMPIETE
| |
| .TERIM REPUR T :
| |
| * r.DY FOR CLOSURE: 04/26/09
| |
| .1NTERIM REPORT: 10CFR50.7J(A)(2)(IV) ORIG EXIT DATE 05/11,'08 CLOSING REPOR T : ORIG CLOSED 05/13/88 SIGNATURE D.E. O'BRIEN DATE COMPLETED 06/17/66 1
| |
| REFER: DVR 20-1-88-088 DESCRIPTION-ON 04/15/88 AT 0132, THE OB CONTAROL ROOM VENTILATION SYSTEM RECEIVED A HIGH RAD SIGNAL FROM OPRO33J GAS CHANNEL AND SWAPPED TO THE MAKEUP MODE OF OPERATION. RADCHEM WAS NOTIFIED TO PULL THE FILTERS AND CARTRIDGE FOR ANALYSIS. NO ABNORMAL AMOUNTS OF RADIATION WERE DETECTED. THE MONITOR APPEARED TO HAVE HAD A SPIKE. THE MONITOR OPERATED PROPEjLY BOTH BEFORE AND AFTER THE EVENT. AN ENE PHONE CALL WAS COMPLETED PER BUAP 1350 REQUIREMENTS.
| |
| * .m THi" RCT THAT CtlANGED OUT THE FILTER REPORTED THAT OPR5313 UAS FOUND OPEN WHICH IS AN INCORRECT POSITION. THIS SHOULD NOT '' AVE AFFECl EO THE ADILITY OF THE MONITOR TO PERFORM IT'S FUNCTION.
| |
| THE ONLY OTHER EVENT THAT MAY HAVE HAD AN EFFECT ON THE RAD MONITOF WAS A OB VC CHILLER TRIP A T 2326 ON 04/14/88. THE CHILLER WAS OFF FOR ; ,
| |
| ADOUT TEN MINUIES WHICH WOULD HAVE CAUSED A SLIGHT TEMPERATURE T MAN'31EN T ON THE SYGIEM.
| |
| | |
| l-i;R AID 9100D- OPER DVP 00rE- ' i.
| |
| ACrTON ITEM :
| |
| .:6 03 - 3:;. 30 ; Ja,E +
| |
| 'EM NC c56-200 :38-M300 ( Cot! r
| |
| .DESCRfPr!GN (Cot 1T) i (OCCR50.?2 t1RC PFD PHOtJE NO TI FIC AT10td r1 ADE , 0400. 04,'15/3G n
| |
| 9 OFERATING ENCINEEP'S COMt1ENIS: 4
| |
| * I tt0NE. BARRY MCCUE 04/15/80 NOrJFICATIONS: RESILENT INSPECTOR, NRC REGION III, 04/15/08, 1600 j
| |
| * T. J. MAIMAN/D. P. GALLE, VP/NSD, 04/i5/88, 1600 j 30 DAY REPORTABLE /iOCFR50.73(A)(2)(1V) n LER NUMBER: 88-011 ACTION
| |
| | |
| ==SUMMARY==
| |
| | |
| lt
| |
| ~
| |
| O
| |
| | |
| BRAIDWOOD SIMULATOR MALFUNCTION
| |
| | |
| ==Title:==
| |
| Leak in WS Comon Header ID: AUK-18 NO: 6.3.4.1.18
| |
| | |
| == Description:==
| |
| Leak occurs on the comon discharge i header of the WS System Variati7ns: Date: 4/11/87 Rev: 0 Selectable Steps Inputs Conuments
| |
| : 1. Select leak rate 0-100,000 spa .
| |
| : 2. Select ramp time 0-99,999 see Brief Plant Response: (IC-17, 100%, ali systems in automatic)
| |
| A large leak will cause t.he stby WS pump to auto start and WS header pressure to decrease. Components relying on WS for cooling will heat up due to lack of WS flow. The first annunciator received will be Wu HDh ?RESSURE LOW.
| |
| Suggested Instructor Action: -
| |
| None.
| |
| Events:
| |
| None l
| |
| l l
| |
| l 303M/83M/22 4/87
| |
| | |
| l I BRAIDWOOD SIMULATOR-MALFUNCTION LISTING-COMPONENT COOLING WATER
| |
| .;f.y.
| |
| 1) r CCW Letdown Heat Exchanger Tube Leak CCW-2 CCW Heat Exchanger's Loss of Coolant Flow' CCW-4: Loss of CCW to Letdown Heat. Exchanger CCW-5 Loss of CCW to RHR Heat Exchanger (CCW Outlet Valve)
| |
| CCW-6 CCW Pump Trip CCW-7 Seal Water HK Tube Leak CCW RHR Heat Exchanger Tube Leak CCW-9 RCP Thermal Barrier Leak CCW-11 CCW Heat Exchanger Tube Leak CCW-13 CCW Surge Tank "A" Leak CCW-14 CCW Surge Tank "B" Leak
| |
| \
| |
| s/ . .
| |
| . C.W-15 CCW System Leak 638M/263M/2 8/87 L - _ _ _ _ -____ ___-----___--___--
| |
| | |
| BRAIDWOOD SIMULATOR MALFUNCTION SIMULATOR MALFUNCTION
| |
| | |
| ==Title:==
| |
| Letdown Heat Exchanger Tube Leak ID: CCW-1 NO: 6.3.4.2.1
| |
| | |
| == Description:==
| |
| Leakage from CVCS to CCW system.
| |
| Ol Q.,
| |
| Date 8/4/86 -
| |
| Variations:
| |
| Rev: 4 Selectable Steps Inputs Comesnts
| |
| : 1. Select heat exchanger CCW-1A 1A = HK 1A CCW-1B 1B = HK 1B
| |
| : 2. Select leak rate O to 100 percent Percent of letdown flow -
| |
| leak rate will depend upon the amount of letdown flow
| |
| : 3. Select reap time 0-9.~999 sec.
| |
| Brief Plant Response:
| |
| The letdown heat exchanger tube leak will pass selected percentage of letdown flow into the CCW system. CCW surge tank level will slowly increase se will the CCW system radiation level. The increase in CCW radiation will eventually
| |
| [') cauce an auto closure of the surge tank vent valve. Letdown flow to the VCT V will decrease, causing VCT level to drop. The reactor make-up water system will respond to the decrease in the VCT level. Pressurizer level will ,
| |
| decrease, causing charging flow to increase.
| |
| Letdown isolation on low pressurizer level could occur.
| |
| 1 1
| |
| l u) 318M/83M/2 5/89 1
| |
| ~ ~ ~ ' ~ * ~ ~ ~ ~ ~ ~~ ~
| |
| | |
| .. . .. . .. - : , ; .u- : ,. = ..: -. .= r-_= -. . . u. - . . .; . - . . _ . .-. - . - - . - . . ; .
| |
| ' BRAIDWOOD SIMUIATOR MALFUNCTION SIMULATOR MALFUNCTION
| |
| )
| |
| 3
| |
| | |
| ==Title:==
| |
| Letdown Heat Exchanger Tube Leak ID: CCW-l' Suggested' Instructor Action:
| |
| When' requested to. isolate the faulty heat exchanger, use the applicable LOA's and place the' standby HK in service, isolate the faulted HX.
| |
| Events:-
| |
| 4 None 9
| |
| 0 318M/83M/3 :/89
| |
| | |
| BRAIIWOOD SIMULATOR MALFUNCTION l
| |
| l1-
| |
| | |
| ==Title:==
| |
| CCW Heat Exchanger's Loss of Coolant Flow ID: CCW-2 NO: 6.3.4.2.2 N
| |
| | |
| == Description:==
| |
| Selected CCW heat' exchanger's ESW inlet or l -(- [~j ' outlet isolation valve closes due to a l failure.
| |
| l:
| |
| Variations: Date: 8/4/86 Rev: 3 Selectable Stepe Inputs Comunents
| |
| : 1. Select faulty heat 1-f 1 = Unit I heat exchanger exchanger and valve inlet 1SX 004
| |
| ' failure 2 = Unit 1 heat exchanger outlet ISX 007-1 3 = Commnon heat exchanger inlet ISX 005 4 = Comunon heat exchanger outlet ISX 007-2 Brief Plant Response:
| |
| When the selected valve closes ESW flow through the heat exchanger drops to
| |
| ; zero. CCW beat. exchanger outlet temperature increases rapidly, causing
| |
| ( various temperature alarms on the components cooled by that train of CCW.
| |
| 318M/83M/4 5/89 l
| |
| l
| |
| '~
| |
| ~ ~ ~ ~ ~~~
| |
| ' ~ ~ ~ ~ ~ ~ ' ~
| |
| | |
| BRAIDWOOD SIMUIATOR MALFUNCTION i
| |
| .e,4 -
| |
| Titl'e:''CCW Heat. Exchanger's Loss of Coolant'Tiow ID:' CCW-2'-
| |
| I' . Suggested Instructor Action:
| |
| None.
| |
| Events:
| |
| l' None -
| |
| s 4
| |
| I
| |
| {
| |
| l l
| |
| 318M/83M/5 5/89 ;
| |
| I o ~
| |
| ' ~ ' ~
| |
| '~
| |
| d
| |
| | |
| BRAIDWOOD SIMULATOR MALFUNCTION i
| |
| | |
| ==Title:==
| |
| Loss of CCW to Letdown Heat Exchanger ID: CCW-4 No: 6.3.4.2.4 l r
| |
| | |
| == Description:==
| |
| Failure of 130A or 130B to selectable J (Q) position with or without manual control available. i Variations: Date: 8/4/86 i Rev: 3 Selectable Steps Inputs Conunents -
| |
| : 1. Select fail position 0-100 percent Percent - open
| |
| : 2. Select type of failure 1-3 1 - Manual control possible simulate failure of auto controller input
| |
| - No manual' control possible of TCV-130A (I/P failure) 3 - No manual control of TCV-130B (I/P failure)
| |
| Brief Plant Response:
| |
| O h' When ICC130A or ICC130B goes shut, letdown temperature starts increasing.
| |
| When letdown temperature increases to the high teraperature divert point, valve ICV 129 will divert letdown flow around the letdown demineralizers (including O
| |
| 318M/83M/7 5/89
| |
| -------_-__---------____.m___ _ _ _ _
| |
| | |
| BRAIDWOOD SIMUIATOR MAIJUNCTION
| |
| | |
| ==Title:==
| |
| Loss of CCW to Letdown Heat Exchanger ID: CCW-4
| |
| [
| |
| Brief Plant Response (continued):
| |
| BTRS) to the VCT.. VCT temperature will start increasing as the hot letdown is added. Seal injection temperature increases slowly, causing a high temperature alarm.
| |
| Suggested Instructor Action:
| |
| Clear the malfunction when the students request that repairs be made on controller I/P converter f~o r valve.
| |
| Events:
| |
| None O
| |
| 318M/83M/8 5/89
| |
| ~ ~ ~ - ~ ~ ~ - ~ ~ ~ - " ' ' - - - - - - ~ - - - ~ ' - - - ~ ~ - - - ' ~ - -
| |
| ~~TT 1__ --___-- -_ Ti---_ __
| |
| | |
| . - . . . . . . . ~ . . .
| |
| BRAIDWOOD SIMULATOR MALFUNCTION
| |
| | |
| ==Title:==
| |
| -- Loss of CCW to RER Heat Exchanger (CCW Outlet Valve) ID: CCW-5 NO: 6. 3. 4. 2.5 L ID
| |
| | |
| == Description:==
| |
| Selected CCW outlet valve to RHR heat k,) exchanger closes when switch released to neutral position due to faulty switch contact Variations: Date: 1/4/88 Rev: 5 Selectable Steps Inputs Comunents
| |
| : 1. Select Ex with faulty 1 or 2 1 = RHR Ex 1A outlet valve valve ICC9412A 2 = RHR Ex 1B outlet valve ICC9412B Brief Plant Response:
| |
| Assume RCS cooldown in progress. When the selected RRR heat exchanger's cooling water outlet valve closes, the heat sink is lost. RCS cooldown would directly stop and RCS temperature would start increasing. No annunciators associated with malfunction are received, however high flow to the RCP thermal barrier may cause those associated annunciators to alarms.
| |
| (O ,) Suggested Instructor Action:
| |
| Clear the taalf.metion when requested by the students to repair the faulty switch.
| |
| Events:
| |
| None
| |
| )
| |
| 318M/83M/9 5/89
| |
| =L_rL___ ___ . _--_
| |
| | |
| BRAIDWOOD SIMUIATOR MALFUNCTION Titles' CCW Pump Trip ID: CCW-6 NO: 6.3.4.2.6 f3
| |
| | |
| == Description:==
| |
| Selected CCW pump trips or fails to start due to a breaker failure.
| |
| ~
| |
| ()
| |
| Variations: Date: 1/4/88 Rev: 5 Selectable Steps Inputs Comunents
| |
| : 1. Select pump CCW-6A 6A = 1A CCW -6B 6B = 1B CCW.6C 6C = 0 Brief Plant Response:
| |
| When the only operating CCW pump trips, the standby pump will auto start on low discharge pressure. The first annunciators received include CC PUMP TRIP, CC PUMP AUTO START and CC DISCH PRESS LOW.
| |
| Suggested Instructor Action:
| |
| When requested to repair the faulty CCW pump, clear the malfunction.
| |
| (O f Events:
| |
| None l
| |
| l
| |
| ..q) l 31BM/83M/10 3/89 f
| |
| _a.---_-_--a__---- - _ _
| |
| | |
| BRAIDWOOD SIMULATOR MAIJUNCTION
| |
| | |
| ==Title:==
| |
| Seal Water Ex hbe Leak ID: CCW-7 NO: 6.3.4.2.7
| |
| - (]
| |
| | |
| == Description:==
| |
| Selectrale size seal water heat
| |
| \j exchanger tube leak.
| |
| Variations: Date: 1/4/88 Rev: 4 Selectable
| |
| -Steps Inputs Comments-
| |
| : 1. Select leak rate O to 100 percent Percent of system flow at normal system pressures.
| |
| : 2. Select ramp time 0-99,999 sec.
| |
| Brief Plant Response:
| |
| Note: Letdown pressure at the seal water heat exchanger during normal operations is =15 - 40 psig.
| |
| CCW pressure is a55 - 70 psig. During operation with excess letdown in service, letdown pressure could increase to a pressure > CCW pressure, thus causing the leak to be into the CCW system instead of from the CCW system.
| |
| i l
| |
| k 318M/83M/11 5/89
| |
| | |
| BRAIDWOOD SIMULATOR MALFUNCTION i f
| |
| | |
| ==Title:==
| |
| Seal H2O Ex Tube Leak ID: CCW-7 Brief Plant Response (continued): !
| |
| . .During normal operation, the CCW 1eak into the seal water heat exchanger will cause an increase in the No. 1 seal leakoff back pressure, causing a reduction ,
| |
| in No. 1 seal leakoff flow. VCT level will increase and the CCW surge tank l supplying the common CCW train will slowly decrease. An RCS dilution will also occur. No annunciators are received.
| |
| Suggested Instructor Action:
| |
| LOA's: ICV 8400,'1CV8398A, ICV 8398B, CC9449A, and CC9449B.
| |
| (CVC22), (CVC19), (CVC20), (CCW22) and (CCW23)
| |
| Events:
| |
| None
| |
| %/
| |
| 318M/83M/12 5/8P
| |
| ~ ~ ~ ~ ~ ~ ~ " ~ ~ ~ ~ ~ ~ ' ~~~~~
| |
| ~ ~ ' - ~ ~ ' - ' ~ ~ ~ ~ ~ " ~ ~ - ~ ~ ~ "
| |
| _ _ _ _ _ . _E .__ ___
| |
| | |
| BRAIDWOOD SIMULATOR MALFUNCTION
| |
| | |
| ==Title:==
| |
| - RHR Heat Exchanger Tube Leak ID: CCW-8 ,
| |
| NO: 6.3.4.2.8
| |
| | |
| == Description:==
| |
| Leak between the RCS and CCW systems, i
| |
| (( .
| |
| direction of leak dependent upon the L-relative pressure in the two systems.
| |
| Variations: Date: 3/13/89-Rev: 4 Selectable Steps -Inputs Conuments
| |
| : 1. Select leaky RHR heat CCW-8A 8A = 1A RH HX exchanger CCW-8B 8B = 1B RH EX
| |
| : 2. Select leak rate 0-100 gym Maximum leak rate is based upon RRR heat exchanger inlet pressure of 150 psig.
| |
| At higher pressures, leakage rates will change proportionally.
| |
| : 3. Select ramp time 0-99,999 sec.
| |
| O Brief Plant Response: (Mode 5, plant systems in automatic) ,
| |
| td (Based on RCS pressure of =150 psig.]
| |
| O 318M/83M/13 5/89
| |
| - -.--_____________._s . _ _ . . . _ _ . . . . - _ _ _ _ _. ._. _ . . _ ._ _ _ _ _ _ _ _ _ _ _
| |
| | |
| 7 .
| |
| BRAIDWOOP SIMULATOR MALFUNCTION ID: CCW-8 RER pressure > CCW pressure -
| |
| Level in the affected CC Surge Tank will slowly increase. Pressurizer level drops slowly, and if pressurizer is solid, pressurizer pressure drops rapidly. CCW radiation level slowly increases to the point of causing auto closure of the CC Surge Tank vent valve. The first annunciator received is CC SURGE TANK LEVEL HIGH/ LOW.
| |
| CCW pressure > RHR pressure -
| |
| CC Surge Tank level will slowly decrease. Pressurizer level and/or VCT level will increase. Possible VCT high level divert to the holdup tanks could occur.
| |
| Suggested Instructor Action:
| |
| Isolate the faulty RHR heat exchanger using LOA's CCW28 & 46 for RH HX 1A, CCW 29 & 47 for RH HX 1B when requested by the students.
| |
| Events:
| |
| None O
| |
| O 31BM/83M/14 5/89
| |
| | |
| BRAIDWOOD SIMULATOR MALFUNCTION
| |
| | |
| ==Title:==
| |
| RCP Ther.aal Barrier Leak ID: CCW .
| |
| NO: 6.3.4.2.9 C
| |
| \,1
| |
| | |
| == Description:==
| |
| Leak between the RCS and CCW systems through the selected RCP thermal barrier. Direction '
| |
| of. leak depends on relative pressures in the two systems.
| |
| Variations: Date: 3/13/89 Rev: 4 l
| |
| Selectable i Steps' Inputs Cosunents
| |
| : 1. Select faulty RCP CCW-9A 9A - RCP 1A CCW-9B 9B - RCP 1B CCW-9C '
| |
| 90 - RCP IC CCW-9D 9D - RCP 1D
| |
| : 2. Select leak rate 0-200 gpm Leak rate based upon RCS pressure of 2235 psig
| |
| : 3. Select ramp time 0-99,999 sec.
| |
| Brief Plant Response: (IC-17, 100%, all systems in automatic)
| |
| ~0 -
| |
| O RCS pressure > CCW pressure -
| |
| Level in the CC Surge Tank slowly increases. Pressurizer level decreases slowly, causing charging flow to increase to help maintain the programmed level. VCT level decreases to the auto make-up setpoint. Radiation in the CCW system slowly increases, actuating the auto closure of the surge tank's vent valve. If the leak is large enough, and closure of CC 685 (CCW return isolation valve from thermal barriers) occurs, the leak flow into the CCW system will be stopped. The first annunciators received include RCP SEAL WTR INJ FLTR DP HIGH and RCP THERM BARR CC WTR TEMP HIGH.
| |
| 318M/83M/15 5/89
| |
| - _ = _ _ , _ _ _ _ _ _ _ _ _ _ , , _ _ , ,_ _ _ _ _ . . . . _ .
| |
| .--------..._.________m._m_._._______m.
| |
| | |
| BRAIDWOOD SIMUIATOR MALFUNCTION
| |
| | |
| ==Title:==
| |
| RCP Thermal Barrier Leak ID: CCW-9 O -Brief Plant Response (continued):
| |
| V CCW pressure > RCS pressure -
| |
| , CCW surge tank level will slowly decrease. Pressurizer level and/or.
| |
| VCT level will increase. Possible VCT high level divert to the holdup tanks could occur.
| |
| Suggested Instructor Action:
| |
| When proper action has been taken to isolate the leak, clear the malfunction.
| |
| LOA CCW-13 for 1A RCP (CC9496A)
| |
| LOA CCW-14 for 1B RCP (CC9496B)
| |
| LOA CCW-15 for 1C RCP (CC94960),
| |
| LOA CCW-16 for ID RCP (CC9496D)
| |
| Events:
| |
| None k
| |
| 318M/83M/16 5/89 i
| |
| ~~~ ' ~ ' ~ ~
| |
| * * ~ ~ ~ ~ ~ ~ ~ ' ~ ~ ~ ~ ~ " ~ ' " ~ ~ ^~~ ' ~ ~
| |
| TT--_____"_T~_'____-___ .
| |
| o
| |
| | |
| .n._.--. . - ..-. . - . - - -..--- ..- - . . - , .- :-- - ;. .-- . . :. . =
| |
| 'BRAIDWOOD SIMULATOR MALFUNCTION J i
| |
| | |
| ==Title:==
| |
| -CCW Heat Exchanger Tube Leak ID: .CCW-11 NO: 6.3.4.2.11
| |
| -- (
| |
| | |
| == Description:==
| |
| Leak from CCW system into the essential A - :. service water system. ,
| |
| Date: 8/4/86 Variations:
| |
| Rev: 4 Selectable Steps Inputs Comments
| |
| : 1. Select heat exchanger CCW-11A 11A = ICC01A (Unit 1)
| |
| CCW-11B 115 = OCC01A (Common)
| |
| : 2. Select leak ate 0-100 gym
| |
| : 3. Select ramp time 0-99,999 sec.
| |
| Brief Plant Response:
| |
| The leak in the CCW Ex will allow the potentially radioactive and chemically contaminated water of the CCW system to leak,into the essential service water.
| |
| sy's tem. The control room will receive a High/ Low Surge Tank alarm as the surge tank level decreases.
| |
| k Suggested Instructor Action:
| |
| After determination that the CCW Ex is leaking, isolate affected Ex.
| |
| Events:
| |
| Nont 1
| |
| Lj 318M/83M/18 5/89 f
| |
| l I' . . _ . , . . .. . . . u .. .
| |
| | |
| ll BRAIDWOOD SIMUIATOR MALFUNCTION l
| |
| h
| |
| | |
| ==Title:==
| |
| CCW Surge Tank "A" Leak ID: CCW-13
| |
| .% - NO: 6.3.4.2.13 I
| |
| | |
| == Description:==
| |
| Lo level due to CCW out-leakage.
| |
| Variations: CCW Surge Tank "B" Leak (See CCW-14) Date: 12/22/86 Rev: 4 Selectable Sceps Inputs Coassents
| |
| : 1. Select leak rate 0-150 gym
| |
| : 2. Select ramp time 0-99,999.sec.
| |
| Brief Plant Response:
| |
| Alarm sounds in Main Control Room. The 1A CC pump trips on lowering level as well as the "0" CC pump if running and powered from bus 141.
| |
| Suggested' Instructor Action:
| |
| Insure student locates source of trouble and isolates it. When told to repair the fault, clear the malfunction.
| |
| f
| |
| ( Events:
| |
| None 318M/83M/20 5/89
| |
| | |
| BRAIDWOOD SIMULATOR MAIJUNCTION i'
| |
| | |
| ==Title:==
| |
| . CCW Surge Tank "B" leak ID: CCW-14
| |
| []
| |
| %)
| |
| | |
| == Description:==
| |
| Lo level due to CCW out-leakage.
| |
| Variations: CCW Surge Tank "A" Leak (See CCW-13) Date: 12/22/86 Rev: 1 Selectable Steps Inputs Conunents
| |
| : 1. Select leak rate 0-150 gpa
| |
| : 2. Select ramp time 0-99,999 sec.
| |
| Brief Plant Response:
| |
| Alarm sounds in Main Control Room. The IB CC ptump trips on lowering level as well as the "0" CC pump if running and powered from bus 142.
| |
| Suggested Instructor Action:
| |
| Insure student locates source of trouble and isolates it. When told to repair -
| |
| the fault, clear the malfunction.
| |
| [ . Events:
| |
| None I
| |
| l f%
| |
| l.j 318M/83M/21 5/89
| |
| | |
| BRAIIMOOD SIMULATOR MALFUNCTION
| |
| | |
| ==Title:==
| |
| CCW System Leak ID: CCW-15 NO: 6.3.4.2.15
| |
| | |
| == Description:==
| |
| CCW 1eakage lowers both sides of i - surge tank.
| |
| - Variations: CCW Surge Tank "A" Leak (See CCW-13) Date: 1/4/88 CCW Surge Tank "B" Leak (See CCW-14) Rev: 2 l'
| |
| Selectable Steps Inputs Comments
| |
| : 1. Select leak rate 0-150 gym
| |
| : 2. Select ramp time 0-99,999 sec.
| |
| Brief Plant Response:
| |
| Alarm sounds in Main Control Room. All CC pumps trip on lowering level. CCW is lost to all components.
| |
| Suggested Instructor Action:
| |
| Insure student locates source of leak and isolates it. When told to repair the fault, clear the malfunction.
| |
| Events:
| |
| : 1. LER 06-02-86-001: Faulty Relief Valve Drains CC System
| |
| : 2. LER 06-01-87-012: Two Trains of CC Inoperable Due to CC System Leak, (Override PA1152 B & D = 2, clear MALF to simulate stopping leak).
| |
| : 3. LER 20-01-87-011: Loss of CC Surge Tank Level causing Loss of Both Trains of CC.
| |
| 318M/83M/22 5/89 I
| |
| __-__..-_._m.m_ _-____ N
| |
| | |
| { LICENSE (l GV(lN7 OGD007 (LGO) !
| |
| /a c t i t t y came ( t ) Occket Numeer (2) 8r e ' ' '
| |
| Rvron. Unit 2 of$1ofofofsiil5 i l9 ! 3s l MAINS OF CGWGNENT COOLING INOPt#4RLE Det 70 a Pt#50NNtl EPDCR IN a Erlftr Valy[ stffrNG
| |
| #ecort Date (7) other rae=1 t'et t w o t . etj,,31 snt Date it) ttR Number (O Month rae,1,tv Namet i Decket N peer!;)
| |
| /// $equentia) /// Revision Day Year j Month Day Year Year i fff Number fff Number I
| |
| NONE of si of of of 8 l l 11 1 flo a la al 6
| |
| ~~
| |
| o I o 11
| |
| ~~~
| |
| of a l 12 11 9 816 of si of of 91 i t THIS REPORT 15 $USMtifED PUR5UANT TO THE REQUIREMENTS OF 10CFR fCback one or mere of the f ollowinn t ftti 1 20.aO2(b) _, 20.405(c) . _ _ 50.73(alt tit iv l _ 73.~ttet POWER , 20.405(a)(1)( t l 50.36(c)(1) 50.73(a)(21(v) 73.7t(c) i 20.405(a)(1)(tt) 50.36(c)(2) .,1L. 50.73( a )( 2 )( v t il Other licecify )
| |
| LEVEL ,__.
| |
| lN la t101 , 20.405(a)(litttt) 50.73(a)(2)(t) _ 50.73(a)(2l(v oit (al in abst oct
| |
| ////////////////////////// 20.405(al(ll(tv) ___. 50.73(al(2)(it) _, 50.73( a )( 2)( v i t t )(B) celew and in
| |
| ////////////////////////// ,,,,,,,,, 20.405( a)( 1)(v) _ 50.73(a)(2)(it t) _ 50.73(a)(2)(n) teit) tiethitt CONF af f FOR THtE Ltt (121 ,,,
| |
| rtttPwcNE Nupete came aPts C00t aan arendia. ts2 rinerat tne Ene tnear tit. 221s aiiIs 21 11 41 I 11 48 tlL caMPLtit ont LINE roa (ACM COMP NT F AILURE Ot1Q1lQ.Jg.,LMI5.ERORf f IU CAust $Y5ftM COMPoNtut MAnvFAC. REPotfastr / CAUSE Sf5 FEM COMPONENT Mamurat. 8tPC'f**Lt ;
| |
| fuRER 70 NPtos / tunga tq,,gsf701 *;y'
| |
| /
| |
| I I I I I I l / _
| |
| l i I L_ l LL .
| |
| a eie I la tv cl 71 il o y f H I I I L . _ L_ L. L_.. ... .
| |
| SUPPLEMENTAL REPotf EXPtettD f lal Espected MQn.Ln. 1 0a'.! ' W Suteission j
| |
| Date (151 L Net fir yet. comniate EXPttito tumMrsst0m cafft 71No , l l LBSTRACT (Limit to 1400 spaces, t.e. approstmately f trteen single-space typewritten lines) (16)
| |
| On November 20,1906 at 1026 Component Coeling (CC) ,
| |
| Pumps 2A and 20 tripped, the Unit Ocerator bad Trst shut down the ZA CC Pump because it was no longer needed to support plant operations. The shutdown e tu e1 a pressure sotke,which Ilfted a retter valve. The ret tef valve did not reseat and partially drained the CC System, the levei in the CC surge Tank fell below the low level interlock which trtpoed the 28 CC Pur D.
| |
| The 2A CC Pump started but also trtpped on low level. The unit is in initial fuel load and precrittent stage and therefore there is no decay heat 1 cad. The CC Pumps were not needed for any safety related lo es b:cause of this conditton, therefore safety was not affected. The retter valve was tsolated. CC Surge ryk Level restored, and the 28 CC Pump was restarted at 1038. The cause of the event was a personeel error m the initial setting of the retter valve. The rettef valve was repatred and re. installed. Necessary procedures will be revised to caution operators of the possittlity of this event. Other CC retter estves en bcth units will be benen tested. Yhts is the first occurrence of this type.
| |
| -l V
| |
| (1135M)
| |
| | |
| ..:I%E! : .E v 3 E s;a- .ta. .,
| |
| Fac t 11ty tame ( !) %: vet sumcer ,:) 2n.
| |
| brenUm't 1 91 11 3t c1 SI sf ?! aIot ! al 3ls Title (4) ggggFgTY RELATE 3 COMPONEhf COOLING I40PER40LE QUE TO LOSS OF WATER invent:RY CAUSED
| |
| ( Event Date !st LER quemar f 61 eeeert Data f 71 l Strer Far111ttet favelves f t)
| |
| ( Month Day Year Year $4guenttal Revtston Month Day Year Factitty Names I recket Numterfst g/// 4~ r g/// wunear wont of El of Di of i f 314 ole el[ .a f 7
| |
| ~~
| |
| of 1 12 ole of5 ole al7 of st of el of I f
| |
| \ THIS REPORT TTED PutsuAuf TO THE REQUIREMENT 5 0F 10CFR Tru - m re of the en11awinnt fit)
| |
| E 20.402(b) , 20.405(c) 50.73(a)(2)(tv) 73.71(b)
| |
| POWER 20.405(a)(1)(1) 50.36(c)(1) 50.73(a)(2)(v) . _ 7f.71(c)
| |
| LEVEL 20.405(a)(1)(11) 50.36(c)(2) .1. 50.73(a)(2)(vtt) 0!o !e Other (specify 1101 20.405(a)(1)(ttt) 50.73(4)(2)(t) 50.73(a)(2)(vit t)( A) in abstract 20.405(a)(1)(iv) _ 50.73(a)(2)(tt) 50.73(a)(2)(viti)(B) below and in
| |
| _ 20.405(a)(1)(v) 50.73(a)(2)(it1) 50.73(a)(2)(a) Test) tfEtattr com7Aff Foe THft tre f121 C*ame TfLEPMonf NUMafe AREA CODE T. Schutter. AttittARt Technical staff tunaryttar frt. 22aB eijI$ 21 1l al l sl Al al ctmpttTE ONE Lfht F0e EAEH COMPOS T Fattuer OftfatetD im TM11 RIPoef f13)
| |
| CAUSE SYSTEM COMPOWENT MAeuFAC- REPORTA8LE / CAUSE SY5 FEM COMP 0eENT MAmurAC- REPORTABLE /
| |
| /
| |
| Tuare to hPeet / tuare to NPent [
| |
| I I I I I I I '
| |
| / I I I I i t I f I l I l l I I f '/ I I I I f f I I SUPPLEMENTAL REPORY EXPfffED f141 Expected Manth l Day I Year Submission
| |
| ! lyet fit vet. ennelate txPEtite mawittfoe DAfri YIme " I f !l l ASSTRACT (tiett to 1400 spaces. i.e. approutmately fif teen single-space typewritten Itnes) (16)
| |
| On april 8. 1987, at approximately 1725. a contracted maintenance crew began work on the Limttorque motor operator of the *1 A" Residual Heat esmoval (RH) Heat Eschanger Component Cooling Water Outlet Isolation Valve. ICC9412A. Thts valve was a potnt of isolation for work on the RH Heat Exchanger, whten reoutred it to be drained of Component Cocitng Water (CC). shift Operating personnel granted parmtssion. with the understanding that if it became necessary for the crch to stroke the valve. they would obtain avtnorttation. The maintenance crew stroked the valve in order to release torous on the matcr gear set.
| |
| It is unclear unether they actually received authorization or not. This allowed Comoonent Cooling Water to back flow through ICC9412A to the Heat Exchanger and out the dratn. This caused the (CC) surge tank to reach the low level CC Pump Trip. The "1A" CC Pump tripped at 1726 on April 8. 1987. The surge tank ts common to both trains. Consequently, both trains of Component Cooling were inoperable. The leak was discovered and isolated. The system was then re-filled. and the *1 A' CC Pump re-started. Total time both trains were inoperable was 17 minutes. The cause of the event was a conrqunication breakttown tetween the maintenance crew and Shtft Operating personnel. Corrective actions util regutre the contracted maintenance crew to 00tain written authertzation prior to manipulating a valve for work on the valve's operator. In odottton. a modification has been initiated to provide automatic makeup watt,r to the Component Cooling system in the event of a leak. The safety significance was sintmal. RC5 Temperature never exceeced 85'F.
| |
| There was one similar previous occurrence reported in LER 455/86-01.
| |
| I
| |
| (
| |
| (1370M/016aM)
| |
| | |
| m_. _ _
| |
| _ . .. a -
| |
| _._r.._ _ _ _ . _ _ _ _ _ _ _ .
| |
| mt*stt I,tv stoco nt* > te w:wm FACILITY cAME (1) CCCKET 4"SER (2) it# wuwerf is! ! pm ,
| |
| rear pq Secuential l/; s evision fri mue er !!i numtee Evron. unit 1 o f E 1 o I e f s f al El a 817 . of I fr - olo 9tr ar os l TEXT Energy Industry Identification tystem (E!!$) coces are toentified in the text as (sx]
| |
| i A. PLANT CcNDITIONs Pefot TO EVENT:
| |
| Byron Untt 1 Event Date/ Time 04/08/17 / 172s MODE L - e'afueline tu Power L RCS (As) Temperature / Pressure as*r /de-nreiturund B. DESCRIPTION OF EVENT:
| |
| The *1A* Residual Heat Renoval (RM)[tP) Heat ENChanger was Out of Service (OCS) for gasket replacement.
| |
| The shell side, conststing of Component Cooling water (CC)(CC] was isolated and drained. One of the potnts of isolation of the 005 for CC was the RH Heat Exchanger CC Outlet Isolation valve. 1CC9412A.
| |
| The grease in Limitorque valve noter operators was scheduled to be changed out duttng the refueling outage. This activity was being handled by contracted maintenance personnel. supervised by utility management. Since this involved numerous Liettorques, the Work Supervisor developed a plan with Operating Management that each valve would be only taken Out of Service electrically for personnel protection. If it became necessary to machentes11y stroke the valve the maintenance crew foreman would ask the Shtf t Engineer for specifte authertratten. This plan was consistent with Station's work practices and programs.
| |
| At approstmetely 0842. on April 6.1987 the Work Supervisor (uttlity non-licensed) for the RH Heat Enchanger gasket replacement requested a temporary lif t of the mechantent partton of the Out of Service on ICC9412A in orcer to perform the grease gnange and gear box flush on the motor operator of ICC9412A, The Operating Shtft Foreman (Itcensed) granted permisston with the espitett agreement that work was only to be performed on the motor and that the valve was not to be stroked open for any reason.
| |
| At approximately 1725. on Aprtl 3,1987 the contracted maintenance crew (non-Itcensed) began work on the valve. During the course of their work it became necessary to release the torque on tu motor gear set which required stroking ICC9412A approntmately half open. They stroked the valvt. This allowed Component Cooling water to back flow to the RH Heat Eschanger, f111 the empty Heat Encnanger and pass through the open crain valve. The CC surge tant level dropped to the low level CC Pump Trip setpotnt. The "1A" CC Pump, which was running to support plant operattens, tripped at 1726. The CC Surge Tank is cocinon to both CC Trains. consequently both tratns were inoperable at this time.
| |
| Shif t Operations. in response to the "lA* CC Pump Trip and tow Surge Tank level, dispatched an cperator to investigate. He ouickly determined that CC was draining into and out of the RH Heat Eachanger. He then closed *1A" RH Heat Exchanger Component Cooling Outlet throttle valve. 1CCg507A. to isolate the leak.
| |
| Water was then restored to the surge tank and the *1 A CC" Pump re-started. The total time both tratns of CC were inoperable was 17 mtnutes. There were no safety system actuations.
| |
| A Generating Station Emergency Plan Alert was declared and appropriate notifications mace.
| |
| This report is required pursuant to 10CFR(a)(2)(vit).
| |
| O d
| |
| (1270M/0164M)
| |
| | |
| trt=rtt Est=? eterer ta_ 'r4+ : w+;=ta.::=
| |
| Fa(IL17Y rtAret 11) 00Cutf NUMBER (2) __g s so.ete !4! am :
| |
| Year /// 5ecu:ntial /// 8evision
| |
| // Number / Mumter arran. Unit t 0 I E I $ 1 0 l D I al si a al7 - of 1 1r - oto ofs er ete
| |
| '('N Energy Industry identtftcation System (E!!$) codes are identified in the tent as (un)
| |
| \),TEKT
| |
| \ C, CAtM OF cVEkT:
| |
| The root cause of this event was a communts:ation breakdown between the contracted maintenance crew perfarmtng the work and Operating ShifL porsennel.
| |
| The contract maintenance personnel had been instructed to always receive permissten from the Shif t Engineer prior te mantpuisting the valve they are working on. The maintenance crew was Interviewed and insist they did receive verbal permissten to strose ICC9412A, However, they de not remember who they talked to. $htft operating personnel maintata that they never gave such permissten. There was no reeutrement to document thts permissten in wrtttng. Neither verston could be corroborated.
| |
| D. SAFETY aEALY1tI:
| |
| Plant and Public safety were not affected. Loss of a heat sink for the Reactor Coelant System (RCS),
| |
| without loss of circulatten, has a negligible effect for the short perted of time the less occurred.
| |
| neacter costant tenversture was metatatned at approntmately as degrees fahrenheit through+out the event and RCS forced re-circulation was maintained, via the operating RM tratn. The water level in the reactor cavity was grector than 23 feet, provtalng suff tctent heat stek during the loss of Coupenent Cooling. It would have provided suff tetent heat sink and cooling for an extended period of time if RM flow had been lost.
| |
| E. CORRICTIVE ACTIDMS:
| |
| a Communications and proper work coordination between station maintenance personnel and Operating Shift
| |
| , personnel has been ef fective and does not warrant concern, consequently, corrective actions are focused on
| |
| ( centracted maintenance personnel.
| |
| Contracted maintenance personnel have been re-informed of the reeutrement to obtain shift Engineer authertration prior to stroking any valve they are working on durtng the Limitorque motor operator grease changeout. In adattton, they are reeutred to obtain this authorization in writing to document that Operating Management has given permissten and is aware of the valve manipulation. This reeutrement will be extended to all stattar work activities.
| |
| A edification to the Component Coeling System has been initiated to provies automatte makeup water to maintain surge tank level. This would attenet to maintain water inventory in the event of a leak. This is balng tracked by a Action Item Record 6-87-113.
| |
| This report will be placed in the Licensed Operator reeutred reading program. In addttton, this report wt11 be distributed to Station Departments to be disseminated to respective cepartment personnel.
| |
| I l
| |
| r (1379M/0164M) l
| |
| | |
| _I!!wt!! I,E4? 8E8*pf '_ t o ? tt- *:% :w.,2 *;w FACILITY 8:AM[ (I) DOCKET ELP'8EE (2) _fD N M ES :6) an, -
| |
| j vear / Se aver,t t al // sevision s . . . , ,s -.,
| |
| avrm unu i oisioioIo ai si 4 aw - as i er - ovo or, er 2,. - )
| |
| rw itxt Energy Industry reenttrication system (strs) cones are toentified in the text as [ns) b r. Parvinus occusarmers:
| |
| tra utsent 1 ELL 455-86~-001 Both Trains or Component Cooling Inoperable Oue to Personnel Error in a Relief Valve Setting.
| |
| G. cDMPoufML ' ES&IA: G a) MAmurkYuang MontacLATURE tglDik, NLEGER MFG PARY NUMRit met Applicatie b) tisuLis or thgD1 LEARcH:
| |
| met Applicable is 4
| |
| g O.
| |
| e es a
| |
| l I 4
| |
| f (1370M/Ol64M) m--.-_---_-_--_--___--_-______.__________---________ _ _ _ _ _ _ . - _ _ _ _ _ . _ . _ _ _ - _ _ _ _ . _ _ . _ _ _ _ _ _ _ _ _ . - . - _ . . _ _ _ _ _ _ - ,
| |
| | |
| , LICENSEE EVENT REPORI (LER) dellity Name (1). Docket humber (2) 8ade (11 Braidwood. Un'tt 1 Of El 01 01 01 41 51 6 1 e'l 0 1 Title (4) Loss of Residus) Heat Removal due to loss of Component Cooling as result of a leaking Component Coci eng
| |
| ~
| |
| Inlet Valve. _
| |
| l hentCate(El LER Number f61 Reoort Date f71 Other Facilities Involved (8)
| |
| Sequential Revtston Month Day Year Facility Names Docket Numberts) anth Day Year Year //
| |
| /p,//, /j/j
| |
| /
| |
| j///
| |
| / Number Number NONE 01 51 01 Ol 0! I I of 1- 21 1 al 7 al 7
| |
| ~
| |
| 01 11 1
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| ~
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| 01 0 0t2 21. e al 7 .
| |
| Di El of 01 01 1 1 THIS REPORT 11 SUBMITTED PURSUANT u, .ne REQUIREMENTS OF 10CFR (Check one or more of the followinni fil) 20.402(b) 20.405(c) _ 50.73(a)(2)(tv) _ 73.71(b)
| |
| POWER _ 20.4::5(a)(1)(1) _, 50.36(c)(1) _L $0.73(a)(2)(v) _ 73.71(c)
| |
| LEVEL __ 20.405(a)(!)(it) _ 50.36(c)(2) ,__ 50.73(a)(2)'(vii) _
| |
| Other (specify (101 0! o! O _ 20.40$(a)(1)(itt) _ 50.73(a)(2)(1) ,,. 50.73(a)(2)(vitt)(A) in Abstract
| |
| ////////////////////////// 20.405(a)(1)(tv) _ 50.73(a)(2)(it) _,,_ 50.73(a)(2)(vitt)(S) below and in
| |
| ////////////////////////// _ 20.405(a)(1)(v) ,
| |
| 50.73(a)(2)(111)L_,,,_ 50.73(a)(2)(x ) Text)
| |
| LICENSEE CONTACT FOR T415 LER (121 TELEPPONE NUMBER Name AREA CODE Nov ard .f ames. Tech Staf f Enoineer. Ext. 24R1 ai1 l5 41 51 al -l 21 El 0 1 COMPLETE ONE LINE FOR EACH COMPON N FAILURE DESCRIBE 0 IN THIS REPORT (131 CAU$E SYSTEM COMPONENT MANUFAC- REPORTABLE CAUSE SYSTEM COMPONENT MANUFAC- REPORTABLE /
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| TURER TO NPPD$,,, TURED TO NPROS /
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| e C iC Ils IV la yhL[018 N I I l i I l 1 -
| |
| /
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| I I I I _ J__l I j } I I I I J, I $
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| SUPPLEMEGAL REPORLDffCTED (141 Expected Menth ! Day I Year Submisston
| |
| (/
| |
| Date (15) l l lyes (If yes. comolete EXPGi(Q3EM1HigL0affl X l No ! L _.l.- L_.1 ABSTRACT (Limit to 1400 spaces.1.e. appr o.imately fif teen $1ngle-space typewritten lines) (16)
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| The IB Residual Heat Removal (RHR) Heat Exchanger (Hu) was out of service with the tube side drained.
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| Preparations were made f or draining the Component Ccoling Water (CC) she1151de of the Hs to allow replacement of a flaage gasket. At 1745 draining of the shell side of the Hx was started. At 1802 the la CC oump tripped due to low 1cvel in the CC Surge Tank The Low Level Alarm on the Main Control Board did net annunciate although the sequence of events recorder did indicate a low level. The draining was stopped. the $ urge Tank refilled.
| |
| and the Isolation valves were checked. At 1816 the 1 A CC pump was restarted and the S/ stem was restored to normal.
| |
| The cause was the CC Inlet Isolation Valve 12aking and contributing was the f ailure of the CC Surge Tank High/ Low Level Alarm to annunctate on the Main Control Board, Additionally, the CC Motor Operated outlet <t1ve on the IB RHR Hu was found 6 turns off its seat.
| |
| The leaking valve has been repatred. the limits for a Motor Operated Valve were adjusted. the Main Control Board alarm was troubleshoot and the symptoms could not be d-uplicated, work ts in progress to check the calibration and scaling on the CC Surge Tank Instrument Loop.
| |
| 1585m(021987)/1102A/15
| |
| | |
| 1 3~ - - _ .
| |
| LIQ!$Q,{yQL gRT f LE01 TFrT CCNLINUaTION FaC!LITV NaME (1) DOCKET NUMBER (2) LER NUMBIP (6) P30e ( 3)
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| Year /p//* Sequent $al // Revision
| |
| * /// Number /p/// Number traidwood. Unit 1 0l510 1 0 1 0 l 41 El 6 al 7 - Ol 11 1 - Of 9 01 2 0F Of 3 XT Energy Industry Identification System (E!!S) codes are identtf ted in the text as [xx}
| |
| A. plant CONDITIONS PRIOR _TO LyiNJ.;
| |
| Moos 5 - Cold shutdown, Rx Power Q%. Reactor Coolant System (RB] Temperature / Pressure: 165'F/371 oQ.g i 9. gescrietion of Event:
| |
| The 18 Residual Heat Removal (RHR) (BP] Heat Exchanger (Hz) was out of service with the tube side dratned.
| |
| Preparations were made for dratning the Component Cooling Water (CC) SC] shell side of the 18 RHR Hr. The A Train of RHR was in service and the IC Reactor Coolant Pump (RCP) (A8] was running. The CC system was in its normal operating configuration with the 1A CC pump running and the 18 CC pump in standby.
| |
| , At 1745 on Jsnuary 21, 1987, draining of the shell side of the 18 RHR Hx was started by opening the shell stG l
| |
| drain valve IRH0028 to allow replacement of the Hx flange galket. Tt.4 Ur.1% 1 Nuclear station coerator (NS01 verifted that CC Surge Tank Level on the Main Control Board was nct dropping. The Unit 1 NSO then went to the other sice of the Control Room to perform an unrelated evolution.
| |
| At approximately 1751 the Sequence of Events Recorder (SER) indicated a low level on the A-side of the CC Surge Tank (setootnt 35%). The Main Control Board alarm, which receivas the same signal that actuates the SER did not annunctate.
| |
| At approximately 1755 the SER todicated a low level on the B side of the CC Surge Tank. Once again, the Main Control Board Alarm did not annunctate. The' Unit 1 NSO had completed the unrelated evolution approu tmately one minute prior to this cccurring.
| |
| At approximately 1802 the IA CC pump tripped on Low Surge Tank Level (this comes from a separate level indicating switch, ses, point 131). A low pressure signal aas indicated on the SER and the Main Control Beard as a result of the IA CC pump tripping. Tnts caused the IB CC pump to auto start, however. the CC Surge Tark Level was less than 13*. and tripped the pump. This occurred two more times before the pump was manually started at the direction of the Station Control Room Engineer (SCRE) who noted CC Surge Tank Level at 0% on the A-side and 20% or. the b-51de. The IB CC pump tripped af ter 4 seconds. The SCRE had the control switche? f or both CC pumps placed tn pull to lock and directed the NSO to stop the IC RCP. Operating personnel intnediately began ref t11 ng the Surge Tank and closed the IB RHR Hx shell side drain valve. They also checked the CC Isolation Valves to the IB RHR Hx and found the Inlet Manual Isolation Valve ICC9504B valve fully closed and the Motor Operated Outlet Valve. NOV 1CC94128 valve 6 turns of f its seat. The A-side and B-stde CC Surge Tank Low Level Alarms were cleared by 1806.
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| at 1816 the 1A CC pump was restarted and the system w&S restored to normal operation thus ending the evert.
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| This event is being reported under 10 CFR 50.73( A)(2)(V) - any event or condition that alone could have prevented the fulfillment of the safety function of structures or systems that are needed to remove residual heat.
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| 1585m(021987)/1102A/16
| |
| | |
| ~
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| i tl((M(UMNT #EPFT f lea 1 TEXT CONTIWaTION
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| ./ACILIIf NAME (1) DOCKET NUMBER (2)- LER NUMRER f6) hee (3)
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| Year /// Secuential /// Revision
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| . hhh Number hhh Nurnbe r Braidwood. Unit 1 01510l0 1 0 1 41 El 6 al 7 - 01 Il 1 - 01 0 01 3 0F O
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| _ j.XT. Energy Industry identif tcation System (EIIS) codes are identified in the text as (xx]
| |
| C. Cause of Event:
| |
| The root cause of this event was leakage past the seat on ICC9504B inlet valve to the 18 RHR Hr. The outlet motor operated valve was checked at the L1me of the event and was six turns of f its fully closed position.
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| A contributing cause of this event can be attributed to the f ailure of the CC Surge Tank Level High/ Low alarm to annunciate on the main control board. According to the Unit 1 N50. SCRE. and an additional NSO on shif t. there was no alarm indicating Low Surge Tank Level although the SER. which has no audible alarm, output showed a . low level alarm condition was present. Had the alarm sounded. Operators could have taken prompt action to restore CC Surge Tank Level before the 1A CC pump trip.
| |
| There were no unusual characteristics in the work location that contrib9ted to this event.
| |
| D .' .$Arety analsvis:
| |
| Since the reactor has not yet been taken critical.- there is no residual heat in the RCS and no spent fuel in the fuel pool. Therefore no safety consequences resulted f rom the temporary loss of CC incident. Had the event occurred under more 1tmiting conditions with residual heat in the RCS and the spent fuel pool full of spent fuel, plant safety would not have been compromised during the short terra (14 minutes) while CC Surge Tank level and CC flow was being restored. The RCS would have a 15'F temperature rise (worst case) which would not result in a loss of sub-cooltng. The fuel pool would take 4.5 hours for bolling to occur (worst Case). Additionally.
| |
| a minimum of two steam generators were avatlable to remove heat as required by the Technical Specifications.
| |
| Corrective Action:
| |
| 1, The valve trody for ICC9504B was repaired to alle* 100% seating of the disc and it has been vertfwed that tne
| |
| ~ leakage past the disc and seat has been stopped.
| |
| : 2. The limits for Motor Operated valve ICC94128 have been adjusted to ensure complete closure when the valve 15 actuated remotely.
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| : 3. The main control board annunciator was troubleshoot and the symptoms could not be cup 11cated.
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| : 4. The calibration and scaling of the entire CC Surge Tank tevel Instrumentation Loop is in progress. (Action Item 456-200-87-02901)
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| F. (revious occurry,,g15:
| |
| NONE G. Comeonent Failure 041:
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| Manuf 3g.13tCIC Nomenclature Serial Number Valve IQ Numter Velan 12" Cast 78G804 12G32 Bolted Bonnet No Model Number Gate Valve 1585m(021987)/1102A/17
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| | |
| -BRAIDWOOD SIMULATOR MALFUNCTION LISTING CONDENSER j
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| ,r -- s )
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| 4
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| 's
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| )
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| CND-1 Loss of Condenser Vacuum CND-2 Hotwell Level Controller Failure CND-3 Circulating Water Tube Leak in Condenser
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| ?'
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| W i
| |
| n
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| (,,/
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| 638M/263M/3 8/87 u_____________
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| | |
| ~
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| :-._..___..._.............._._ . _ , . . . _ . . _ _ _ _ . _ . _ E~_~ E ;
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| :BRAIDWOOD SIMULATOR MALFUNCTION
| |
| -Title: Loss of Condenser Vacutaa ID: CND-1 NO: 6.3.4.3.1
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| | |
| == Description:==
| |
| Main condenser exhsust boot fails causing increased air in-leakage.
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| Variations: Date: 1/6/88 Rev: 6 Selectable Steps Inputs Comments None Brief Pla?.t Responses (IC-17, 100%, all systems in automatic)
| |
| With the exhaust boot failed, air in-leakage increases causing a slow loss of main condenser vacuim (5-8 minutes). The throttle valves will open to maintain desemded load as vacuum is lost. As condenser pressure incttases, a turbine trip will be actuated. The reactor will trip if the plant is above P-8. The first annunciator received will be CNDSR PRESS LOW.
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| Suggested Instructor Action If ' desired, this malfunction can be cleared to allow main condenser vacuum to be regained if another initiating event is simulated.
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| Events:
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| : 1) DVR 06-02-87-092: Delta T Runback Due to Loss of Vacuum O 306M/83M/2 5/89
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| ^
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| a--1-_-_:_=________= :_------:-----------~~- ---
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| C ' .. - f DEVIATION INVEST!0ATION REPORT 4
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| PACE DELTA T pUNBACK DUE TO LOS5 OF VACUUM AT 2C FEEDWATER PUMP 1 !Orl 0 12 DIR NUMBER DEPORT DATE
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| // SEQUENTIAL // REVISION
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| .dNTH DAY YEAR STA UNIT YEAR NUMBER NUMBER MONTE 4 DAY YEAR 1 POWER
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| .01 9 LEVEL ,
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| Il E Rf 7 01 6 01 2 al 7 -
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| 0 19 l 2 ""*
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| 0l0 I10 3 !0 8 l7 gg 99 4 CONTACT FOR THIS DIR NAME TELEPHONE NUMBER AREA C002 T. Didier. Deeratina Enoineer Ext. U17 COMPLETE ONE LINE FOR EACH COMPONE 8 l 1 1E 211 l4 l-lE l4 l4 l 1l LORE DERCRIBED IN THft REPORT CAUSE SYSTEM COMPONENT MANUFAC- REPORTABLE l
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| CAUSE SYSTEM COMPONENT MANUFAC- REPORTABLE l TURER 70 NPRDS TURER TO NPPOS I I I I I I ! l i 1 i i l i l i I I I l i i !
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| l I l l 1 SUPPLEMENTAL REPORT EXPECTED MONTH DAY YEAR SUBMISSION j Ji$ f if ves. comolete EXPECTED SUBMIS$f 0N DATE) Xl NO
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| {
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| !sXT A. PL ANT CONDITIONS PRIOR TO EVENT:
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| Event Date/ Time 9/1E/97 / 111s Hrs Unit 1 MODE 1 - Power doeration Rx Power 941 RCS [AB) Temperature / Pressure Normal coeratino B. DESCRIPTION OF EVENT:
| |
| On 9/15/tf. at 1115 hours. Byron Unit
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| * was cperating in Mode 1 at 94 percent power. The 2C Turbine Driven Feedwater Purg (FW)[5J) was out-of-ser$1ce at this time for maintenance. The pump draining incorporated the use of condenser vacuum through the pumps recirculation line. . The pumps casing drain to the floor drains was opened slightly to aid draining. In additton, Mechanical Maintenance personnel removed a leaking section of line. opening the ptre casing to atmosphere, thus, opening the condense- to additional air inleakage. This additional opening to the main condenser through the pumps recirculation Itnes caused a rapid loss of vacuum. The rapidly decreasing condenser vacuum caused an increase in Nuclear Power (51) due to the decreased efficiency. An Over Power Delta Temperature (0 PAT) runback of 70 megawatts and a one minute delta I penalty resulted. The Unit 2 Hogging vacuum Pump was started and the recirculation line on the pump was isolated.
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| C. 311,E OF EVENT:
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| The loss of vacuum was due to the manual recirculation valve being slightly opened. The pump casing drain to the floor drains was throttled open causing air inleakage. In addition. Mechanical Maintenance was repatring a leak and removed a section of line that was connected to the pump castnj causing another source of air in1takage. The volume of air inleakage caused a loss of vacuum. The loss of vacuum caused a decrease in efficiency, but the load remained the same so Nuclear Power increased 5% in order to maintain load. An OP AT runback of 70 megawatts occurred and a one minute delta I penalty resulted. There was a lack of communication and awareness between shif t personnel and Mechanical Maintenance at the point when the work actually started on the pump.
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| %J g (1677M/0199M)
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| | |
| ./
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| DEVIATION INVEST!GATICN REPORT TEXT CONTINUATION i DIR NUMBER PaGE SEQUENTIAL REVISION
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| . JP A T RUNBACK OUE TO LOSS OF VACUUM STA UNIT YEAR NUMBER NUMBER 01 6 01 2 81 7 -
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| 0l9l2 -
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| 010 2 0F 0 l2
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| ' (T D. SAFETY ANALYSIS:
| |
| Th2 plant or public safety was not affected by the incident. Safety functions operated per design on the Reactor Protection System. Starting the Unit 2 Hogging Vacuum pump stopped the loss of vacuum on the main condenser.
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| .E. CORRECTIVE ACTIONS:
| |
| The Operating staff had communications established with the Maintenance people during the pumps draining.
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| Th2 Operating staff had started the scenario by creating a minor leak to the main condenser through the
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| . pumps recirculation line. Maintenance people remDved a leaking section of line on the pump causing additional air inleakage. This Deviation Report will be required reading for both departments foreman in crder. to stress the need for attention to condenser vacuum not only on the Feedwater Pumps but on all p ssible air inleakage Components.
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| F. PREVIOUS OCCURRENCES:
| |
| LER NUMBER IIILE NONE h i1MPONENT FAILURE DATA:
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| V a) MANUFACTURER NOMENCLATURE MCDEL NUMBER MFG PART NUMBER Not Applicable b) RESULTS cF NPROS SEARrH Not Applicable i /"
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| i i
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| '% g 677M/0199M)
| |
| | |
| ~ -~ ~~--~- - - - - --
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| BRAIDWOOD SIMULATOR MALFUNCTION
| |
| | |
| ==Title:==
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| Hotwell Level Controller Failure ID: CND-2 NO: 6.3.4.3.2
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| / ' _
| |
| | |
| == Description:==
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| Selected hotwell level controller fails to L
| |
| selected position.
| |
| Variations: Date: 1/4/88 Rev 4 Selectable-Steps Inputs Comments
| |
| : 1. Select failed controller Ctm2A 2A = uc037 (emergency overflow) f 2.?'8,B 2B = LC038 (normal overflow)
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| CND2C 2C = LC039 (normal makeup)
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| CND2D 2D = LC040 (emergency makeup)
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| : 2. Select fail position 0 to 100 percent Percent of controller output
| |
| : 3. Select ramp time 0-99,999 sec.
| |
| Brief Plant Responses (Normal makeup Failed on)
| |
| Normal overflow actuates to stabilize hotwell level slightly higher than normal. No annunciators are received.
| |
| Suggested Instructor Action:
| |
| LOA FWM-1: Emergency Makeup Isolation (CD028)
| |
| LOA FWM-2: Normal Makeup Isolation (CD031)
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| LOA FWM-29: Emergency Overflow Isolation (CD140) l LOA FWM-30: Normal Overflow Isolation (CD143)
| |
| Events:
| |
| None f
| |
| Ot 306M/83M/3 5/89
| |
| ~ * - ~ ~ ~ ~ * * * ~ ~ ~ ~ ~ = = ~ - * ~ ~ ' - ~ ~ ~ ' ~ ~
| |
| = _ _ _ _ _ _ ._ *7=
| |
| | |
| 1____ _ . _ _ _ . . . . _ _ _ . , _ . _ . . _ . _ _ _ _ _ _ . , _ _ . _ _ _ . _ . _ . _ _ _ _ __
| |
| BRAIDWOOD SIMULATOR MALFUNCTION
| |
| | |
| ==Title:==
| |
| Circulating Water Tube Leak in Condenser ID: CND-3
| |
| .N0: 6.3.4.3.3
| |
| | |
| == Description:==
| |
| Variable size leak into the H.P. condenser-ph
| |
| * tube bundle A from the circulating water system.
| |
| Variations: Date: 3/13/89 c Revi 4 Selectable Steps Inputs Comments
| |
| : 1. Select leak rate 0 to 1,000 spa
| |
| : 2. Select ramp time 0-99,999 sec.
| |
| Brief Plant Response: (IC-17, 100%, all systems in autonatic)
| |
| When the condenser tube leak occurs, condenser hotwell level starts increasing. This will cause the condensate normal overflow valve to open, sending vater back to the condensate storage tank. .The increase in rejection flow to the condensate storage tank will cause its level to increase eventually actuating a high level alare. No annunciators are received upon initiating this malfunction.
| |
| Suggested Instructor Attion:
| |
| None.
| |
| Events:
| |
| : 1) DVR 20-02-88-177: Waterbox 2A tube rupture l 2) DVR 06-01-89-002: Waterbox 18 tube leak
| |
| +
| |
| i O 306M/83M/4 5/89
| |
| . . . . _ . . - - _ . . . - . .. . ~ . - .
| |
| | |
| CMb DEVIATION INVESTIGAT!O2 REPORT (JIR) Form sev 2.0 patE Facility Name ' ICFI 0 l2 Bratswood Unit 2
| |
| / \ Title Waterbox 2A Tube Rupture' Que to Improper Design k
| |
| Oft NUMBER REPORT DATE EVENT DATE OPERATING'
| |
| // SEQUENTIAL // REVISION NUMeER NUMBER MONTH DAY YEAR MONTH DAY YEAR STA UNIT YEAR POWER LEVEL al a 21 o of 2 --
| |
| 1 I 71 7 --
| |
| oI o 11 o 21 1 al a of 2f 3 of e il e al a CONTACT FOR THft DIR
| |
| ** TELEPHONE NUMBER NAME AREA CODE Ext. 2477 ai1 iE 4 1EIa1 12 Ia!O I1 Dan Stroh Teen staff Enaineer COMPLETE ONE LINE FOR EACH COMPONE URE DEtCRfRED IN THi% REPORT CAUit SYSTEM COMPONENT MANUFAC. REPORTABLE CAttSE SYSTEM COMPONENT MANUFAC- REPORTABLE' TURER TO NPRDS TURER TO NPRDS I i i I i 1 I I l i 1 I I I I i 1 f I I I I I I I I MONTH DAY YEAR SUPPLEMENTAL REPORT EXPECTED SUBMISSION
| |
| -- DATE I I i XI NO I f r
| |
| ! YES fit vet _ camelete EXPECTED CURMf t1tDN DATE)
| |
| TEXT Energy Industry Identification System (EIIS) codes are identif ted in the text as [XX]
| |
| A. PLANT CONDif f 0NS PRf 0R TO EVENT:
| |
| Unit: Braidwood 2; Event Date: September 30. 1988; Event Time: C600 f'
| |
| k Mode: 1 - Power Operation ax Power: 23% RC5(A8] Temperature / Pressure: 557'F/2230 psig
| |
| : 8. DESCRIPTION OF' EVENT:
| |
| On September 30, 1988 at 0600. the chemistry department reported that steau gene *stor water chemistry At 0630 power revealed that sodium levels were at 600 ppb and conductivity was at 40 y"#05 'ar Unit 2.
| |
| was ramped down f rom 275 MW to 175 MW as water chemistry action level 1.1 was entered. Water chemistry continued to degrace and at 1246 the main turbine was tripped. On 10-1 .; '. 0206 Unit 2 entered moce
| |
| : 3. Waterbox 2A was crained and the cir:ulating water tubes were inscected. A single severed tube was discovered with the break in the B rone. On Monday. 10-3-88. personnel entered the condenser and removed the severed tute. A hole of 4' x 5' was also discovered in the f alse bottom directly below the severed tube. The tube was removed back to the nearest support plates and the tube was plugv G in the water boxes. The nole in the f alse bottom was not regatt ed; however, the damaged portion of the f alse Dittom was removed to prevent further damage. On 10-5-88. Unit 2 went critical and continued power ascensten.
| |
| l l
| |
| t 236&m(110188)/2 l l
| |
| l
| |
| ____m=___:t___2_m.__- _ .._.
| |
| | |
| sr DEVIATIO1 INVESTIGATION RETORT TEXT CONTINUATION term ieu 2.0 DfR NUMeER PACE l FACILITY NAME "
| |
| Braidwood Unit 2 j,^j STA UNIT of 2 YEAR al n -
| |
| NUMeER i I7I7 -
| |
| NUMeER of a 2 or oi2 V .
| |
| 21 o (EXT C. CAutt 0F EVENT:
| |
| Initial investigation into the cause of tube failure showed that tne break occurred just below a condenser penetration used for feedwater (FW) ($J) recirculation flow. This penetration was installed as a modification several years ago. It was evident from the large hole' ripped in the condenser false bottom, that the penetration was improperly designed for the high flow-high temperature conditions of the FW recirculation flow. Because of the clean break of the tube, it is assumed at this time that the mechanism of tube f ailure was due to high frequency fatigue caused by large scale flashing and flow induced vibration by the FW recirculation water. The tube has been sent to .ECo system materials analysis department for a complete report on the failure mechanism.
| |
| ,D. RAFETY ANALY111:
| |
| Yhere were no direct safety consequences to the untt as a result of this event. If this event had happened under the most Itmiting condition of 100% power operation. there still would have been no impact on plant safety as the unit would have been brought down in a safe and controlled manner per procedure.
| |
| The condenser was always available during the event. Even if the condenser had become unavailable. it is not required for the safe shutdown of the plant. Long term results would include degradation of the stram generator metallurgy and a poss,1ble increase in the rate of steam generator tube failures.
| |
| j% E. CORRECTIVE ACTfcN!
| |
| t b
| |
| Inunediate action was 10 bring down the unit in a controlled manner and begin cleanup of the water using the condensate polishers. After the tube was found and removed, the tube ends were plugged to prevent further inleakage. Sargent and Lundy engineering will review the casign of the FW recirculation penetration and possibly oesign/ install a new sparger to handle the high temperature recirculation flow, as well as evaluate the required repatrs to the false bottom.
| |
| F. #3 V Mtp SCEtlPRENrES:
| |
| Nere G. COMPONENT Fa! LURE DAT[:
| |
| Manufacturer NomenclAhtg Medal Number Mfe. Part Number l' Stain 1*st steel N/A N/A Foster Wheeler circulating wate?
| |
| condenser tube
| |
| .J
| |
| , 236Bm(1,10188)/3 I_______________________________ __ __ - . -
| |
| | |
| 4e '
| |
| udb cu b.s .
| |
| DEVIATION INVESTIGATION REPOR,7 (OIR)
| |
| Facility Name PAGE Byron Nuclear Power Station 1 10 0 ~
| |
| t Title Condensar Tube Leak EVENT DATE ~ DIR NuleER REPORT DATE
| |
| // SEQUENTIAL // REVISION M) NTH DAY YEAR STA UNIT YEAR / NUfeER f/ NUfeER #CNTH DAY YEAR 1 POWER LEVEL o 11 o 17 a 19 0 16 0 11 s 19 -
| |
| o lo 1 2 -
| |
| eio 0: 1 0l9 8l9 o g,lg CONTACT FOR THIS DIR nag TELEPHONE NUPRER AREA CODE T. Tulon Assistant Sumerint:nd:nt Omerations Ext. 2213 811I5 21314l-l5l414l1 COMPLETE ONE LINE FOR EACH COMPONEN FA tar DE$catarn IN THIS REPORT MANUFAC- REPORTABLE - CAUSE SYSTEM COMPONENT MANUFAC- REPORTABLE CAUSE SYSTEM COMPONENT Tiere TO NPRDS TURER TO NPRDS l l l 1 l i I I I l l I l l 1 I I I I I I I I I I I SUPPLEpKNTAL REPORT EXPECTED _
| |
| MONTH DAY YEAR SUBMISSION x 1 YES fif ven. enmolate EXPECTED timMO$!ON DATE) I ND D _l40l1 9 0 TEXT Energy Industry Identificatti.14 System (EII5) codes are identified in the text as [XX) f A. PLANT CopeITIONS PRIOR TO EVENT:
| |
| Event Date/ Time 01-07-89 / 2129 Unit 1 MODE 1 - Power Goeration Rx Power . 98% RCS (A8) Temperature / Pressure Normal Qaeratino.
| |
| B. DESCRIPTION OF EYENT:
| |
| On January 7, 1989 at 2320 hours Byron Station Unit I was at 98 percent power when the station chemistry department notified the Shif t Engineer that steam generator catioe; conductivity was elevated.
| |
| Testing by the Chemistry Department determined that the '8' Circulating W terbox on the Main Condenser had a tube leak. A condensate polisher was used to compensate for the leak which .?t estimated at 2 gallons per hour (gph). On January 9,1989, the decision was made to ramp down to 75 percent power at 2300 hours in order to remove the 18 Circulating Waterbox from service. On January 10, 1989 the leaking tube was found using a Helium Leak Detector. The tube was plugged and the waterbox was back in service by 0600 hours on January 11, 1989.
| |
| O (0227R/0027R)
| |
| | |
| DEVIATION INVESTIGATION REPORT TEXT CONTINUATION Form'Rev 2.0 !
| |
| DIR NLDSER PAGE TACILITY NANE
| |
| ~p SEQUENTIAL REVISION STA UNIT YEAR NUPRER NUpeER Buran Nuclear M r Station 0 f6 0 11 8 19 -
| |
| Ol012 -
| |
| 0 10 2 0F 0l _
| |
| TEXT Energy Industry Identification System (E!!S) codes are identified in the text as [XX]
| |
| , C. CAUSE OF EVENT:
| |
| The cause of the event was a tube leak in the 1d Circulating Waterbox. ' The leak was adequately controlled by using a condensate polisher, however system loads made, the January 10, 1989 waterbox outage possible.
| |
| The cause of the tube leak could not be determined. The tube will be pulled and analyzed in an outage of sufficient duration. All of the leaks found on, Unit I since 1987 have been attributed to mechanical 1
| |
| damage. There have been no indications of corrosion in the stainless steel tubes.
| |
| D. 5AFETY ANALYSIS:
| |
| The Noin Condenser tube leak did not cause any plant safety concern. The tube leak did not contribute to any other event important to plant or public safety. In the worst case of a large tube leak the condensate polishing system would be unable to maintain chemistry limits, which would require the Unit to ramp down to 25 percent Power.
| |
| E. CORRECTIVE ACTIONS:
| |
| The leak did not require immediate repair due to its ses11 size. The operation of one condensate polisher was required to maintain steam generator. chemistry within. operating limits. The tube was plugged when the Unit was ramped down to 75 percent power.
| |
| h
| |
| : j. The tube wl11 be pulled during the next outage of sufficent duration in orifer to determine the cause of the condenser tube failure. Byron Unit I third refueling outage would be the next scheduled outage. Action Item Record 89-411 tracks this work.
| |
| F. RECURRING EVENTS SEARCH Als ANALYSIS:
| |
| a) ,
| |
| EVENT SEARCH (DIR. LER)
| |
| DIR NL9BER IIILI, 6-1.86-152 Waterbox leakage b) ItWUSTRY SEARCH (OPEK's NPRD51 Not in NPR05 data base. No relevent information found.
| |
| c) !flE See attached Itst of plugged tubes 1
| |
| d) ANALYSIS No corrosion problem apparent.
| |
| (0227R/0027R) l
| |
| -. - - L . -. . . ..
| |
| - mm _m:__.._m-_____m.___m._.-_-_. __
| |
| | |
| 7:n --- c
| |
| 'BRAIDWOOD SIMULATOR MALFUNCTION. LISTING'
| |
| : g. ,
| |
| REACTOR CONTROL y'.- , f~M A* \
| |
| : sl.! .<
| |
| j
| |
| .(-
| |
| 4 j s. CRF-1 Rods Fail'to Move
| |
| .CRF-3 Improper Bank Overlap iCRF-4 Dropped Rod l
| |
| [CRF-5 Dropped Rods
| |
| .CRF-6 Rod Ejection:
| |
| 'CRF-7 Uncontrolled Rod Motion
| |
| 'CRF-8 Auto Rod Speed Controller Failure CRF-9 Tref Failure (Rod Control)
| |
| CRF-10 DRPI - Loss of Voltage CRF-11 DRPI - Open or Shorted Coil CRF-13 Stuck Rod -
| |
| CRF Rod Control System Failure
| |
| ~
| |
| j '.
| |
| CRF-15' Power Cabinet Urgent Failure CRF-16. . Logic Ca'o inet Urgent or Non-Urgent Failure CRF-17 Rod Bank Misalignment CRF-18 Rod Stops Fall.
| |
| CRF-19 ' Urgent Failure in Logic Section of Shutdown Banks C, D and E CRF-20 Reactor Trip Failuto CRF-21 Manual Reactor Trip Failure CRF-23 OT AT Setpoint Failure CRF-24 0F AT Setpoint Failure CRF-25 Under-Voltage on RCP Buses 'I CRF-26 Under-Frequency on RCP Buses O
| |
| 1 638M/263M/4 8/87 j
| |
| _ _ _ - - - -- __ i
| |
| | |
| BRAIDWOOD SIMULATOR MALFUNCTION 1,
| |
| | |
| ==Title:==
| |
| Rods Tail to Move ID: CRF-1 NO: 6.3.4.4.1
| |
| | |
| == Description:==
| |
| Selected type of failure prevents various banks or groups of rods from moving.
| |
| Variations: Date: 8/12/86 Rev: 3 Selectable Steps Inputs Comments
| |
| : 1. Select type of failure 1-3 1= Auto-Failure of auto. cir-cuit summing unit.
| |
| (Affects auto only).
| |
| 2= Man-Open in line supply-ing power to the in-out relays. -(Affects auto,
| |
| ( man, and all modes of bank select except S/D banks C,D and E).
| |
| 3= Group-Failure of the slave cycler signal to reach the power cabinet.
| |
| ' (Affects all modes).
| |
| : 2. Select cabinet. LAC Use only when input of 2AC "3" used above. 1BD 2BD SCD
| |
| , v 519M/197M/2 '5/89
| |
| | |
| BRAIDWOOD SIMULATOR MALFUNCTION D
| |
| | |
| ==Title:==
| |
| Rods Fail to Move ID: CRF-1
| |
| -[V Brief Plant Response: (Based on fault occurring to the controlling bank during a power reduction with no operator action)
| |
| Auto and Man - When the control rods do not move following the start of the load reduction, Tavg will increase, causing Tavg-Tref deviation and high Tavg alarms. The increased Tavs will help decrease nuclear power, but will cause pressurizer level and pressure, and steam pressure to increase. Charging flow, steam flow and feed flow will all decrease. The variable heaters will turn off and later the backup heaters will turn on due to a +5 per-cent pressurizer level deviation. As pressurizer pressure increases, pressurizer spray valves will cycle to control pressure. With the Tavs increasing, the OTAT turbine runback will occur, worsening the problem. This will be accompanied by
| |
| : p. an alarm and may be followed shortly by the OTAT reactor trip.
| |
| Group - When the turbine load starts decreasing, only one of the controlling banks group will insert. This will cause rod deviation alarms, and possible rod insertion limit alarms. Fluz tilts, associated alarms and indication will also occur.
| |
| Suggested Instructor Action:
| |
| None.
| |
| Events:
| |
| None O 519M/197M/3 5/89
| |
| | |
| BRAIDWOOD SIMULATOR MALFUNCTION
| |
| | |
| ==Title:==
| |
| Improper Bank Overlap ID: CEF-3 Q NO: 6.3.4.4.3
| |
| | |
| == Description:==
| |
| Failure of the control banks to properly over-lap due to the bank overlap unit counter being mispositioned.
| |
| Variations: Date: 9/19/88 Rev: 5 Selectable Steps Inputs Consents f
| |
| : 1. Select BOU counter BOU - bank overlap unit error i 1-5 steps The BOU counter can be set high or low from 1 to 5 steps.
| |
| -s Brief Plant Response: (Based on failure occurring during a reactor startup)
| |
| As rods are withdrawn, the selected bank starts withdrawing too early or too late depending on the polarity of BOU counter error.
| |
| 519M/197M/5 5/89
| |
| - - - - - - -~ -- - - - -
| |
| | |
| BRAIDWOOD SIMUIATOR MALFUNCTION
| |
| | |
| ==Title:==
| |
| Improper Bank Overlap ID: CRF-3 Suggested Instructor Action:
| |
| Simulate rasetting the counter to its proper value when requested, by setting MCRFOC (BOU counter) to its proper setting.
| |
| Events: 1) DVR 06-0.2-88-048: Improper BOU Setting.
| |
| o .
| |
| O 519M/197M/6 5/89 1
| |
| -- - - - ~ --
| |
| r_-r__-_-______-_T_____-_________________r_:-__~~T~~----------
| |
| | |
| cEF-3 DEVIATION INVESTIGATION REPORT TITLE PACE IMPROPER CONTROL R0D mag cytRLAp k h DIR NUMBER REPORT DATE EVENT DATE SEQUENTIAL REVISION MONTH DAY YEAR STA UNIT YEAR y// WlDGER g/7 NUPRER MONTH DAY YEAR 1 POWER <
| |
| ol E cl 1 al a 01 6 of 2 al a -
| |
| o i 41 a -
| |
| 0Io O6i O9i 8,8 LEVEL. ; ,g ,
| |
| CONTACT FOR THIS DIR N.WE TELEPHONE NUMBER AREA C00E Free Hornbank. Kanier Staff Encineer Ext. 2R22 8l 1lE 2l 1 l4 l-IE I4I4l 1 COMPLETE out LINE FOR ear D IN THit REPORT j CAUSE SYSTEM COMPONENT MANUFAC- REPORTABLE CAUSE $YSTEM COMPONENT MANUFAC- REPORTABLE TURER TO NPRDS Tuare TO hPRDS I I I I I I I / I I I I l l l l l l l l 1 l _
| |
| '/ l 1 l l l
| |
| $ SUPPLEMENTAL REPORT EXPECTED MONTH DAY YEAR SUOMIS$10N I YES fif yet. conolate EXPECTED U mMIt$10N GATE 1 i NO TEXT A. PLANT CWIDITIONS PRIOR 70 EVENT:
| |
| h Event Date/ Time E/1/as / 210e I d Unit 1 MODE 1 - Pawar Gnarations Rx Power 981 RCS (AB) Temperature / Pressure Normal Oneratina Unit 2 MODE 1 - Power onaration Rx Power _ 111. RCS [AB) Temperature / Pressure _ Normal Onaratina
| |
| : 8. DESCRIPTION OF EVENT:
| |
| At 2300 hours on 5/1/88, a
| |
| * Computer Rod Deviation sequence
| |
| * Alarm was received by Unit 2. m ile investigating the cause of the alarm, the rod bank select switch (AA) was put into manual at 2314 to adjust delta I to its target value (per station policy). A
| |
| * Rod Sank C 1.ow Insertion" alarm was received while inserting rods. At this time Control Sank C (CBC) position was observed to be at 222 steps on Digital Rod Position Indication (DRPI) (AA) and 225 steps on CBC group step counters. The Shift Engineer was imediately notified. At 2317 the rod bank select switch was placed in the *CBC" position and CBC was withdrawn to 228 steps, clearing both the alarms. The Bank Overlap Unit (BOU) counter was observed to be indicating 455 total steps withdrawn when it should have been indicating 546. At this time the station i Nuclear Engineer was notified. Upon the advice of the Technical Staff Engineer the 500 was incremented 91 steps to 546, corresponding to Control Bank 0 at 201 steps withdrawn. No systems were inoperable prior to this event that contributed to the event. No operator actions either increased or decreased the severity of this event. There were no manual or automatic safety system actuations. and stable plant conditions existed throughout the event.
| |
| C. CAUSE OF EVENT:
| |
| The intermediate cause of this event was the BOU being 91 steps low. The root cause of this event is j l g indeterminate. A possible cause of this is the failure of one of the SOU counter circuit cards. If this j event should reoccur, further investigation will be pursued and a supplemental report will be issued.
| |
| (
| |
| I l
| |
| (0014R/0003R)
| |
| L______________
| |
| | |
| OEVIATICM INVESTIGATION REPORT TEXT CONT!;:UATIO1 TITLE DIR NUMBER PACE l $EQUENTIAL REVISION STA UNtf YEAR NUMBER NQ l l g rER costnot Roo RANK OVERLAP- 0 16 0 12 a la -
| |
| 0l4i B -
| |
| 0 l0 2 0F 0I2 TEXT D. SAFETY ANALYS!$:
| |
| There were no safety consequences as a result of this event. The health and safety of the public were at no time adversely affected or threatened. The rods remained capable of being tripped into the core at all times during this event. If this event had occurred under a more severe set of circumstances there would have been no safety consequences as the rods would still have been capable of being tripped.
| |
| E. CMRECTIVE actions:
| |
| The imediate corrective action was to step rods out to clear the alanus. The BOU was incremented to prevent the problem from reoccurring and rods moved in to verify the problem was cleared.
| |
| F. PREVIOUS OCCURRENCES:
| |
| There has been one previous occurrence of this type of failure where the Logic Cabinet BOU counter was 50 steps below actual rod position. The cause for this event was unknown.
| |
| DVR N M ER IIILE 06-02-87-075 Control Bank C Below RIL/ Bank Overlap Unit Error D) 6
| |
| \__/ G . COMPONENT FAILURE DATA:
| |
| a) MANUFACTURER NOMENCLATURE MODEL NLDRER MFG PART NlseER.
| |
| Not Applicable b) RESULTS OF NPRD5 SEARCH:
| |
| Not Applicable c) RESULTS OF the SEARCH:
| |
| Not Applicable C\
| |
| I l
| |
| (0014R/ coo 3R)
| |
| | |
| .y , ,
| |
| BRAIDWOOD' SIMULATOR MALFUNCTION p- l
| |
| | |
| ==Title:==
| |
| Dropped' Rod ID: CRF-4 d' NO: '6.3.4.4.4 p
| |
| | |
| == Description:==
| |
| Failure of either the movable or the stationary gripper circuit, causing the selected rod to-fall-into the core.
| |
| Trouble.is due to a blown fuse.
| |
| Variations: Date: 3/15/89
| |
| -Rev: 6 Selectable Steps Inputs Comments
| |
| : 1. Select malfunction ~ CRF4A number CRF45
| |
| : 2. Select rod H6, M12, See attached core map for-O -
| |
| F2, etc. rod locations
| |
| : 3. Select faulty gripper 1 or 2 1 = Movable gripper 2 = Stationary gripper 4
| |
| Brief Plant Response: (Based on plant being at 100 percent power when rod drops, no operator action)
| |
| Note: In this malfunction, the indication and alarms received will differ from case to case, depending upon which rod was dropped and its proximity to the excore neutron detectors. Basically, it will be as .
| |
| l follows:
| |
| When the control rod drops, the digital rod position for the affected rod will indicate the rod on bottom. The first annunciators received include ROD AT BOTTOM and R0D DEV POWER RNG TILT.
| |
| 519M/197M/7 5/89
| |
| ~~~ .-- , - - -- ~ ~ .
| |
| _ - K - __ L__-__'_L _ ___ 2 '_*i_.___ ._____ ~ *~~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ - ~ - ~ - - - ~ ~ - ~
| |
| | |
| BRAIDWOOD SIMULATOR MALFUNCTION
| |
| .j/-
| |
| | |
| ==Title:==
| |
| Dropped Rod ID: CRF-4 V
| |
| Brief Plant Response (continued):
| |
| 1 1
| |
| The power range meter nearest the dropped rod will decrease and the
| |
| -meter across the core will increase.
| |
| Tavs will decrease and cause the pressurizer level and pressure to decrease. The backup heaters will come on, accompanied by a low pressurizer pressure deviation alarm. The decreasing pressurizer level could cause a low isvel deviation alarm and increase charging flow. The decreasing Tavs will cause the Tavs-Tref deviation and possibly low Tavs alarms.
| |
| Suggested Instructor Action:
| |
| : 1. To make a recovery of the dropped rod, the Instructor Essi first clear the malfunction.
| |
| : 2. If the rod is in the control bank, the P/A converter for the affected group must be set to zero during recovery, (MCRFPA(x)).
| |
| Events: 1) DVR 06-02-87-004: Control Rod D-2 dropped
| |
| : 2) DVR 20-01-87-144: Control Rod K-14 dropped i
| |
| l i
| |
| l 519M/197M/8 5/89 !
| |
| I i
| |
| W*''* * ' * * '****** ~*"*e
| |
| "* __I___'_'"_""O_*'_
| |
| | |
| g, g,g, , yg ,
| |
| l ncr
| |
| .TLE 1 !0FI 3 1 2 CONT 2OL R00 0-2 DROP 7ED DUE TO BLOWN FUSE REPORT DATE // ///
| |
| l__ EVENT DATE DIR NUMEER SEQUENTIAL REVISION NUMBER- MONTH DAY YEAR 2
| |
| '4 NTH DAY YEAR STA UNIT YEAR NUMEER POWER f , f, /
| |
| /
| |
| 0 11 3? I al 7 01 6 01 2 81 7 ~~
| |
| '01 01 4 0l0 01 3 11 6 al 7 0 10 11 ,f . fif CONTACT FOR THis 01R TELEPHONE NUMBER NAME AREA CODE
| |
| : f. Schuster. Assistant Technical Staff Suoervisor Ext. 244 8l 1 15 2 l 3 l 4 l - l 5 l4 I4 l 1 COMPLETE ONE LINE FOR EACH COMPONEN F LURE DEtCRIBE0 IN THIS REPORT CAUSE SYSTEM COMPONENT HANUFAC- REPORTABLE CAUSE SYSTEM COMPONENT MANUFAC- REPORTABLE TURER TO NPROS TURER TO NPROS _
| |
| X AIA FI UI I a 11 15 16 Y I l l 1 l l I 1 l l l 1 I l 1 1 l 1 i MONTH DAY YEAR SUPPLEMENTAL REPORT D PECTED SUBMISSION l YES fif yet. comalete EXPECTED $UBMISSION DATE) X l NO TEXT ,
| |
| A. PLANT CONDITIONS PRIOR TO EVENT:
| |
| HODE 2 - startuo Rx Power M _ RCS [AB] Temperature / Pressure 558'F/2235 osia
| |
| /'' B. DESCRIPTION 0F EVENT:
| |
| At 0942 hours on 01-31-87 while in Mode 2 at 3% power the Shutdown Bank A control rod (RO)(AA) corresponding to core location 0-2 fell from 228 steps (fully withdrawn) to RB (rod at bottom, zero steps) as indicated by the Olgttal Rod Position Indication (P!) display (RO). It was determined that the control rod actually dropped (not a PI r splay error) based on power decrease to 1%. The unit NSO stabilized the unit at 1% power and troubleshooting of the RD system was started via Nuclear Work Request B41040. There were no components or systems that were inoperable at the beginning of the event which contributed to the event.
| |
| C. CAUSE OF EVENT:
| |
| Fuse FU7 in the IAC Rod Control Power Cabtnet 2R006J was found blowh. This fuse supplies Phase A The Stationary Gripper power for power cabinet rod group C (corresponds to Shutdown Bank A. group 1).
| |
| transient change in current to the stationary gripper cot 1s following the loss of phase A power allowed rod 0-2 to drop but was not severe enough to drop the other three rods in Shutdown Bank a, group 1. The cause of the blown fuse is unknown, the electrical conditions at the phase A. B. and C fuses were essentially identical following fuse replacement.
| |
| I e ,
| |
| m 1225M/0152M/9) i I
| |
| | |
| 1TLE. OIR nut *Mp nCE SEQUENT!al REVISION I STA UNff YEAR NUMBER. NUMBER l
| |
| CONTROL ROD 0-2 DROPPED DUE TO BLOWN FUSE 016 01 2 BI 7 -
| |
| 01 0 14 -
| |
| O l 0 2 0F 0 f2 I
| |
| 1 ,XT
| |
| %_ f D. SAFETY ANaLY$f5:
| |
| There were no adverse safety consequences resulting from this event. The dropped rod added negative reactivity to the 1: ore and did not tmpair the abtitty of the other rods to drop tf required. The dropped rod following a partial loss of coti power %s a f ailure in the safe direction. Due to the ver,y low power level at the time of the event (3T.) there were no adverse af fects on core power dtstributtons. The short duration of the rod misalignment resulted in no significant changes in core Xenon distribution. If the rod had dropped at a higher power level, there would have been a significant reactor turbine power mismatch and core reactivity, redistributions. This is an analyzed event,and is predicted to not result in fuel damage of any kind.
| |
| E. CORRECTIVE ACTIONS:
| |
| ruse FU7 was replaced and the control rod retrieved per BOA R00-4 within 54 minutes of the time the rod dropped. Shutdown Bank A was moved to steps in and out and was declared operable. Fuses do occasionally blow and due to the low frequency of this type of event no actions were taken to try to prevent recurrence. j F. PREVIOUS OCCURRENCES:
| |
| None. however DVR 6-1-85-133 was written for a blown pha d B stationary gripper fuse that did not result in a dropped rod.
| |
| G. COMPONENT FAILURE DATA:
| |
| * f
| |
| \
| |
| a) MANUFACTURER NOMENCLATURE MODEL NUMBER MFC PART NUMBER Shawnut "Amptrap" 30 amp A60 x 30 Type i N/A 600V Fust b) RESULTS OF NPROS SEARCH:
| |
| Not Applicable I .
| |
| D .
| |
| l 225M/0152Mn 0 )
| |
| _ _ _ - - _ _ - . _m_-.______ - .- _---____.__._ -
| |
| | |
| DEVIATION INVESTIGATION REPO:tf TITLE: Control Rod Oropped to Zero Steps Due to Blown Fuse DACF t 10rl i 2 tytNT DATE D!R WOMBER REP 0pf DAff f//
| |
| SEQUEN';! AL (h j y//
| |
| REVISION MONTH DAY YEAR STA UNIT YEAR NUMBER // MONTH DAV vtAR NUMB [R_ 4 / t POLER I
| |
| .31 4 21 2 al 7 21 0 of 1 al 7 -
| |
| 11 41 4 -
| |
| Of D of 5 21 0 RI 7 of of 0 CONTACT FOR THIS OfR NAME T[LEPHONE NUMBER AREA CODE FRANK W. TRIKUR Tech Staf f fncineer Ert. 2495 81t l$ 4 I5 lR I-I2l8 i0 It COMPLET[ CNE LINE FOR EACH COM60NENT FA.tLURE Ct3CRIBED IN fwis stPORT CAUSE SYSTEM COMPONENT MANUFAC- REPORTABLE CAUSE SYSTEM COMPONENT MANUFAC- REPORTABLE TurtR 70 Neocs TuptR To NPRas a A Ia Fl ul *l
| |
| * xi 91 91 9 N 1 l l 1 l l l l I I I I I I / I l i I I tuPPLEMENial REPORT EXPECTED MONfH DAY vtap SUBM!$510N l YES fir ves. comolete EXPECTED $URMIS$f0N DATE1 X! NO l
| |
| ' TEXT A. PL ANT CONDf f f 0N1 PRIOR TO EVENT: .
| |
| Unit: Braidwood 1; Event Date: Aortl 22. 1997: Event Time: gilg MODE .4. - Hot shutdown Rx Power 1 RCS (AB) Temperature / Pressure 337 r/343 PsIG i
| |
| B. DESCRIPTION OF EVENT: .
| |
| At 0430 with the plant in normal Mode 4 condition 18wc510.5-1. Rev. St. *SPECIAL TEST EXCEPTION RCD POSITION INDICATION SYSTEM DA!LY SURVEILLANCE PRIOR TO AND AFTER ROD DROP TESTING" was intttated. There were no systems or components inoperable at the beginning of the event that contributed to the event.
| |
| At 0530. Rod K14 in Conrol Bank B dropped to zero steps. The rematntng rods in control Bank B were manually inserted to zero steps. At 0537 a Manual Turbine Trtp was initiated and at 0539 the Reactor Trip Breakers =ere opened. At 0540 Feedwater Isolation was manually reset and 18woS 10.5-1 was finally exited at 0541.
| |
| The plant was stable throughout the enttre event.
| |
| C. cAust CF tytNT:
| |
| The root cause of the event was the f ailure of Fuse FUIS in Power Cabinet 180. This prevented the stationary gripper cotl from remaining energized. thus dropping rod K14 D. SAFETY ANALYSIS:
| |
| There were no safety consequences of this event. At no time was the health or safety of the public adversely affected. The blown fuse only prevented roc KI4 f rom being withdrawn. The rods were trippable at all times.
| |
| The consequences of this event under a more severe set of intttal conditions, reactor critical and end of core Itfe, would be acceptable stnce th FSAR. Section 15.4.3.2. "One or More Dropped Rod Cluster control Assemoly .
| |
| states that the Departure from Nucleate Boiling Ratto will not fall below the limit value, k
| |
| 1694m( 052187)/33 1204A
| |
| __ ________________.__a__ - _ _ - . _ _ _ _-
| |
| | |
| DEV!ATION INVI$f!GATION REPORf TEXT CONTINUATION l;> TITLE; Control and Dropped to Zero Steps Due to Blown' .,, DIR NUMBER PACE l.
| |
| : y. .Fust SEQUENTIAL REVISION i y f. -; $7A UNIT YEAR - flUMBER tluMBER fl 0 01 1 B1.7 *""
| |
| I la14 -
| |
| 'O l 9 '2 0F 0 l2 iTEXT' E. CCRRECTIVE Actions;
| |
| ^
| |
| -Fuse Full in Power Cabinet 100 has been replaced. System design allows for such fattures without system degradation and provides detection capatt11ttes fer this type of failure, "FfPREVtoutDCtuRRENCit;
| |
| . NONE ,
| |
| . G. COMP 0hENT FAttuef DAT A:
| |
| M FACTURER -NOMENCLATURE N0 DEL NUMBER MFC PAef NUMBER j Bussman ' fuse 2432559 BUSS N/A st. Louis Semiconductor 104
| |
| , 250V
| |
| (
| |
| 9 O
| |
| 1694m(051487)/34 1204A.
| |
| | |
| N s. h N3, o LOOP 2 N .N .L . ... 0 ...
| |
| LOOP 3 b MMMMMMM MEM @ M @ M @ MEBM b
| |
| =
| |
| MMM E M B E B E 8 ENE MMM E X E E E ME MB i
| |
| MM E M M N N M E ME
| |
| =
| |
| N @ N E @ M e M @M E @M r EM B E 88f H H E8 EM -
| |
| .0 .
| |
| M e E s M e M eM e s s E e m no-
| |
| -MM E M E M E E B MW
| |
| = M-@ M M @ M e 88s @ 8 81 M @ M MM E R H M 888 8 81 B MM BM I!!M E E E X MEMME
| |
| 's MMM 8 58 8 M B M E B88EM O .. EsM e n e s e MsE O EmmMEMM O l N44 Net LOOP i LOOP +
| |
| N.,oN,. ;
| |
| 0 CONTROL 200 PATTERE Coat statows (* Coat ogTECTORS CONTROL BANK A h ~
| |
| , SOURCE RANFE a N51 a N31
| |
| '"I'""""'''' * *** * "3" c0NTRot sawK s @ 2 % 2.4 3a PC.tR RANGE o N41.ket.N43,N44 CONTROL SANK C h SOURCES CeNr.0L .- O g ,Ni.. .
| |
| l
| |
| ..ur00.N .- i g ><coaa= X l iMur00.N . - . g sNor00.. .- C g ,
| |
| - sNvTee.= us. . 8 j sNur00 N sank E E l
| |
| | |
| I L
| |
| V) BRAIDWOOD SIMULATOR MALFUNCTION Titles Dropped Rods ID: CRF-5 NO: 6.3.4.4.5
| |
| | |
| == Description:==
| |
| Failure of power to stationary gripper coils thyristors for a group of rods due to firing circuit card failure Variations: Date: 9/19/88 Rev: 6 Selectable Steps Inputs Comments
| |
| : 1. Select faulty group ISA-2CD ISA (4) ICB (4)-
| |
| ISB (4) ICC (4)
| |
| ~
| |
| 2SA (4) 1CD (3) 2SE (4) 2CA (2) 1EC (4) 2CB (4)
| |
| ISD (4) 2CC (4)
| |
| ISE (4) 2CD (2)
| |
| ICA (2)
| |
| I g, & (4) Bank selected )
| |
| ' Type selected (i.e.,
| |
| l Shutdown / Control)
| |
| Group selected I
| |
| No. of rods in group Brief Plant Responses (Based on plant at high power when multiple rod drop occurs) 519M/197M/9 5/89 l
| |
| | |
| . BRAIDWOOD SIMULATOR MALFUNCTION
| |
| ;i 4% .
| |
| | |
| ==Title:==
| |
| . Dropped Rods ID: CEF-5 Brief Plant Response (continued):
| |
| When the group of rods selected drops, if power is greater than 5 percent, a reactor trip will occur due to negative rate and cause the reactor trip breakers to open.
| |
| Suggested Instructor Action:
| |
| None.
| |
| Events:
| |
| : 1. July, 1985 Byron Station Trouble Report.
| |
| : 2. LRR 06-01-86-028: Power Cabinet 2AC Failure.
| |
| ()
| |
| (/ 3. LER 20-01-88-016: Rx Trip'Due to Negaf;ive Rate.
| |
| : 4. LER 20-02-88-009: Multiple Rod Drops.
| |
| : 5. LER 06-01-88-002: Multip1'e Rod Dror During Manual Rod Motion.
| |
| : 6. LER 06-02-88-006: Multiple Red Drops.
| |
| O 519M/197M/10 5/89 l~
| |
| l
| |
| - _ _ _ _ : _ ______-__ ---~~ - - ~ ~ ~ - - - --- - ^
| |
| ~---
| |
| | |
| I L /
| |
| A ly 15, 1985 -
| |
| l l
| |
| f) .
| |
| preliminary Root Regulatory Implications cause Assessment l Ex_ l _No t e Monc W !!ance l l _l Personnel Error l l_l.,_ Level TV Noncor:pliance l l,x_x l Eculpment Failure l l_,,1 L,1 cense violetton l l _,,,l Pr ocedu r a l . l l I Possible Enforcement Action l l l Other l TROUBLE REPORT l
| |
| BYRCW STATION l
| |
| On Saturday *l/13/85 at 0450 hours. Byron Unit 1 experienced an unplanned automatic reactor scram from 60 MWe. At the time of the scram. Byron station was in the process of removing Unit 1 from service due to operational problems with the number P-8 reactor control rod drive mechanism. During the shutdown process. Byron Station was struck by lightning. The lightning caused the failure of multiple control room instruments and resulted in the automatic insertion of 3 reactor control rods. The insertion of the control rods scrammed the reactor on negative neutron flux rate.
| |
| (j The station in conjunction with Station Electrical Engineering
| |
| * department (SEED) and Operation Analysis Department (OAD) is investigating the details of the lightning strike to assess what follow-up actions are required. In addition to the lightning strike problem. the station is working with Westinghouse to repair the original problems with the P-8 reactor control rod drive mechanism. A return to service date is not available at this time..
| |
| A Red Phone call wis made to the NRC. and Communication Services-was notified of the event, e Li) t 73 Stephen E. petrowski s
| |
| O 0606B/3
| |
| | |
| -___,__..n_
| |
| LICENSIE E Waf aE M2T ;tE81 Occuet Numoer (2) 8m 1 l F ac t i t t/ Cl ''* ( 1 )
| |
| 01 11 01 01 01 41,_,11 4i i l pl 9 l3 l Byron. Unit 1
| |
| ' Msquat Rf AcfoR TRIP Dtf TO 900 CROP C AU'E BY FAUL TY CIRCUIT C ARDS Other Factitttes Involveer g31
| |
| ,d Pecert Date f7)
| |
| - LER m_=ner f61 Factitty wames I cocket Numeerts, i k ) EveM Data ts) 5equential /// Revision Month cay year Month ' Day Year Year /// 8 NI Numeer If'/
| |
| / Numeer NONE Of El OI Of Of I I i
| |
| ~~ ~"
| |
| O I O 11 0 21 9 816 Of sf 01 Ol Of l l 11 0 of 2 al 6 al 6 012 Ia TH15 REPCRT 15 $UBMITTED PUR5UANT TO THE REQUIREMENTS OF 10CFR ,
| |
| fcheck one or mera of the followinal (111 73.71(c) l 20.40$(c) .1_ 50.73(a)(2)(tv) 2 20 402(b) ,,,,
| |
| 73.71(c) 20.405(a)(1)(t) 50.36(c)(3) 50.73(a)(2)(v) ._._
| |
| POWER _,,,
| |
| 50.73(a)(2)(vit) Other (St,ecify 20.405( a )( 11( t t ) 50.36(c)(2) ,
| |
| LEVEL in Abstract 50.73(a)(2)(vitt)(A) 0 0l 1 20.405(4)(1)(111) _ 50.73(a)(I)(1) flot ,_
| |
| 50.73(a )(2 )(vi t t )(8) below and in 20.405(a)(1)(tv) _ 50.73( a )(it)( t t )
| |
| ////////////////////////// _
| |
| 50.73(a)(2)(u) Text) 20.40$(a)(1)(v) _,. 50.73(a)(2)(tit) .,_
| |
| //////////////////////////
| |
| * ' LICENSEE CONTACT FOR THis tfE (121 fftEPHONE NUMeER tame AAEA CODE 101E 3l 1 1E 2l 31 4l .l 51 4 Alt G. Brindia. U2 Oaarattna Ensinaar. Eut. T F Att t1EE' OEstif RED IN THIf REPORT f 111 COMPLETE ONE LfNt FOR EACH r0M SYSTEM COMPONENT MANUFAC. REPORTa8LE' /
| |
| MAhWF AC- REPORTA8LE CAUSE CAUSE SYSTEM CCMPONENT Tutta TO NPeDi I TURER TO NPRos X Al A l Il FI 7 WI11210 Y f X AI A I XI FI 7 WI 11 21 0' Y WI 11.21 0 Y I Y AI A l Yi Fl 7 X AI A I Kl FI 7 WI II 21 0 Y Espected Month ! D.iv I vez tuPP g MTAL REPatt EXPECTED (141 Submission
| |
| \
| |
| * l l
| |
| * "'"" Date (15) l ! I l.
| |
| Y l No 7 lYas f f f vas. comalate EXPffft0 tuaMf tston DAffi ASSTRACT (Limit to 1490 spaces.1.e. approutmately fif teen single-space typewritten Itnes) (16) at 195 and 80 steps withdrawn.
| |
| On 10/2/86 at 0650 during a reactor startup with control bank "C" and *0" Limiting Condition respectively, alarm window l-10-C06 *R00 CONTROL URGENT FAILURE * [ AA) (RO) annunciated.
| |
| for Operation Action Response 1805 1.3.1-la was entered into, due to "All Wss Detng inoperable Dut trtppaele'. Troubleshooting was performed and ctrcutt cards in the 2AC roe artve power catteet were being dropped at 1150. The Shif t Engineer then ordered replaced by Maintenance when 4 rods in contrel bank "C" the reactor to be manually tripped. The reactor was manaully tripped, and all rematntng rods dropped.
| |
| Further troubleshooting of the system was performed, the circuit cards responstele for the rod drop ,ere "C" was successfully exercised to vertfy resolution of the replaced and the alarm cleared. Control bank problem. The trip recovery was normal. and the start up resumed. There was one prevtous occurrence of rod artve problems causing a reactor trts. This was described in LER 85-42.
| |
| o -
| |
| I (1077M)
| |
| | |
| LICENSEE EVENT REPORT (LER) -
| |
| CRF Facility Name (1) Braidwnd Unit 1 Docket Numb 2r (2) , Dace (3) 015101010IdiSl6 1lof!0l3 O Title (4) Reactor Trip Due Negative Rate Trip as a Resu?t of Improper Administrative Controls
| |
| _ Event Date (5) LER Number (6) Report Date (7) Other Facilities Involved (8)
| |
| Month Day Year Year /// Sequential / Revision Pbath Day Year Facility
| |
| * fames Docket._fbmberfs)
| |
| //jj f
| |
| / Number j//j/j/
| |
| / Number None 01 51 01 01 01 l I 01 8 11 1 81 8 81 8
| |
| ~
| |
| 0l116
| |
| ~
| |
| n01 0 0l8 21 3 81 8 01 Sl of 01 Of I I THIS REPORT IS SUBMITTED PURSUANT TO THE REQUIREMENTS (# 10CFR p
| |
| (Cheek one or mere of the followino) fill 20.402(b) 20.405(c) JL 50.73(a)(2)(iv) 73.71(b) 1 _
| |
| POWER __ 2h405(a)(1)(1) _ 50.36(c)(1) _ 50.73(a)(2)(v) __ 73.71(:)
| |
| LEVEL _ 20.405(a)(1)(ti) g..__ 50.36tc)(2) ,_ 50.73(a)(2)(vil) ._,_ Other (Speci fy g
| |
| (10) II o1 0 _ 20.405(a)(1)(iii) ._,._ 50.73(a)(2)(1) _ 50.13(a)(2)(viii)(A) in Abstract
| |
| / /,/ / / /,/ / / // / // / /,/,/,/,/,// ///,/, ,,__ 20.405(a)(1)(tv) _ 50.73(a)(2)(ti) _ 50.73(a)(2)(vill)(B) below and in 20.405(a)(1)(v) 50.73(a)(2)(li t ) 50.73(a)(2l tm) Texti
| |
| /}////j/////////}'jj'/' ////j/'j
| |
| / /
| |
| /f/jj / _ _ __ _, _
| |
| LICENSEE CONTACT FOR THIS LER f121 _
| |
| Name TELEPHONE nut 9ER AREA CODE l Crale (tutpa. Shif t Enoineer Ext. 2202 8 l 1 1 5 i il SI 81 l 21 81 01 COMPLETE ONE LINE FOR EACH COMP 0NE FAILURE DESCRIBED IN THIS REPORT f13)
| |
| SYSTEH COMPONENT MANUFAC- REPORTABLE CAV5E SYSTEM COMPONENT MANUFAC. 'REPOR1ABLE CAUSE TURER TO NPRDS TURER TO HPRDS l l l X AI A FI Ul l $1 Il 51 6 N .
| |
| l l l l l 1 l l l I I I I I I l l l
| |
| : p. SUPPLEMENTAL REPORT EXPECTED f14) Expected Month i Day 1 Year Submission lyes (If ven. complete EXPECTED SUBtQ$$ ION DATE) X l NO I l l ABSTRACT (Limit to 1400 spaces, i.e approximately fif teen single-space typewritten lines) (16)
| |
| On August 11, 1968, the Rod Control Urgent Alarm annunciated three times. Also, the C-11 Rod Stop alarm was annunciating prir,r to 223 wers on Bank D control rods. Preparations were made to check the power cabinet ISO fuses. Separate discussions were held to discuss the details of checking the fuses. The unit operator was instructed to place rod control in MANUAL and to perform no rod movement during the time the power disconnect switch was open. ,The unit opeestor was occupied with normal duties and observed that the C-11 Rod Stop alarm had annunciated again and attempted to clear it by manually moving rods. When rod motion was demanded, the stationary gripper coils doenergized and Control Bank D Group 1, rods dropped. This caused an automatic reactor trip on High Flux Negative Rate. Contributing to this event was a lack of administrative controls to indicate an off-normal system status. The cause of the Rod Urgent alarms is attributed to two bad fuses in cabinet 1W0 which were replaced. This event has been reviewed with the individuals involved.
| |
| Additional administrative controls will be developed to aid the operator in identif ying of f-normal equipment status. This report will be included in the licensed operator Required Re6 ding Program. There have been no previous occurrences.
| |
| 0 2261m(082688)/6 l
| |
| | |
| i ' -i: LICENSEE EVENT REPORT fLER) TEXT CONTINUATION j FACILITY.NAME (1) DOCKET NUteER (2) LER NUteER (6) , Pane O)
| |
| " ' '""' fff "
| |
| 'Braidwood Unit 1 /// Number /// Nuaber
| |
| . fm j 0 1 5 1 0 1 0 1 0 1 41 51 6 8l8 - 0l116 - 01 0 01 2 0F 01 3 TEXT Energy Industry Identification System (E!!$) codes are Mentified in the text as (xx)
| |
| A. PLANT CONDITIONS PRIOR TO EVENT:
| |
| Unit: Braidwe'od 1 Event Date: Auaust 11. 1988 Event Time: 1316 Reactor Mode: = .,l_ Mode Name: Power Oseration. power Level: 100%
| |
| RC$ (A8) Temperature / pressure: NOT/NDP
| |
| : 8. DESCRIPTION OF EVENT:
| |
| There were no systems cc components inoperable at the beginning of the event which contributed to the severity of the event.
| |
| On August 11, 1988, during normal operations on the day shif t the Rod Control Urgent Alarm (AA) annunciated three times. It was determined that the prc':lem was a phase failure on the movable gripper coli in power cabinet 180. It was decided to check and if necessary replace the fuses in the cabinet.
| |
| During the day shif t and on previous shif ts, the C-11 Rod Stop alarm was annunciating prior to 223 steps on Bank D control rods. The alarm was also clearing at the wrong setpoint.
| |
| At approximately 1300, preparations were made with the Station Control Room Engineer (SCRE), the unit Nuclear $tation Operator (NS0), an extra NSO and the Technical Staff to check the power cabinet 180 7
| |
| V fuses. Separate dicussions were held by the SCRE, prior to the start of the work; one discussion with the unit NSO and the other discussion with the extra NSO and the Technical Staff to discuss the details of checking the fuses.
| |
| At approximately 1311, the unit NSO was instructed to place rod control in MANUAL and to perform no rod movement during the time the power disconnect switch was open to replace the movable gripper coil fuses. The extra NSO and the Technical Staff Engireer met the Shif t Engineer ($E) by the 180 power cabinet. The Technical Staff Engineer showed the SE what the job entailed and the SE lef t. Ths extra NS0 opened the disconnect switch at power cabinet IBD and proceeded to take voltage readings across the fuses to ensure it was de-energized. The unit NSO was occupied with normal duties and ol' served that the C-11 Rod $ top alarm had annur:iated again.
| |
| At 1316, while the ertra NSO was taking the voltage reading across the third fuse, the unit NSO attempted to clear the C-11 Rod $ top alarm by moving manually moving rods.
| |
| When rod motion was demanded the stationary gripper coils deenergized. Since power was removed f rom the movable gripper coils for cabinet 18D, they could not energize.
| |
| As a result, Control Bank 0 Group 1, rods D4 and M12 dropped. This caused an automatic reactor trip on High Flux Negetive Rate.
| |
| The appropriate NRC notification via the ENS phone system was made at 1458 on August 11, 1988, pursuant to 10CFR50.72(b)(2)(ll).
| |
| This event is being reported pursuant to 10CFR50.73(a)(2)(iv) - Any event or condition that resulted in manual or automatic actuation of any engineered safety feature, including the reactor protection system.
| |
| 'h
| |
| ,Q 2261m(082988)/7
| |
| - - . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ f
| |
| | |
| i I
| |
| LICENSEE EVENT REPORT (LER1 iEXT CONTINUATION FACILITY NAME (1) DOCKET NUMBER (2) LER NUteER (6) Pace On Braidwood Unit 1 fff
| |
| ///
| |
| N"'"
| |
| Number fff
| |
| ///
| |
| Number
| |
| [ ,
| |
| 0 1 5 1 0 1 0 1 0 1 41 51 6 819 - Ol116 - 01 0 01 3 0F 01 3
| |
| ( M TEXT Energy Industry Identification System (E!!5) codes are identified in the text as (xx)
| |
| C. CAUSF OF EVENT:
| |
| The root cause of the reactor trip was a cognitive personnel error by the licensed unit NSO.
| |
| Contributing to this event was a lack of administrative controls to indicate an off-normal system status.
| |
| The cause of the C-11 Rod $ top alars annuerlating prematurely on day shif t was the failure to reset the Pulse / Analog converter when resetting the Aod Urgent Failure alarms which had occurred earlier in the day.
| |
| The cause of the Rod Urgent alarms is attributed to two bad fuses in cabinet 180.
| |
| D. SAFETY ANALY$IS:
| |
| There was no effect on plant or public safety as all systems operated as designed in response to the High Flux Negative Rate. This event is enveloped by the Final Safety Analysis Report (F$AR). A dropped Rod Control Cluster Assembly (RCCA) bank typically results in a reactlytty insertion greater than SE-3 delta K/K which will be detected by the power range negative neutron flux rate trip circuitry. The reactor is tripped within approximately 2.5 seconds following the drop of a RCCA bank. The core is not adversely affected during this period, since power is decreasing rapidly.
| |
| . E. CORRECTIVE ACTIONS:
| |
| (
| |
| The immediate corrective actions were to estabitsh stable plant conditions following the reactor trip.
| |
| The fuses in caninet 180 were replaced and the system line-up was returned to normal.
| |
| Action to prevent recurrence include:
| |
| : 1. This event has been reviewed with the individuals involved.
| |
| : 2. Additional administrative controls will be developed to aid the operator in identif, g and maintaining off-normal equipment status. The cuntrols being considered are fabrication of a movable guard and additional guidance on pre-job briefing and communication requirements. This will be tracked to completion by Action Item 456-200-88-18201.
| |
| : 3. This report will be includ d in the Licensed Operator Required Reading program. This will be tracked by Action Item 4$6-200-88-18202.
| |
| F. PREVIOUS OCCURRENCES:
| |
| There have been eight previous reactor trips due to personnel error. In each case corrective actions were implemented addressing both root and contributing causes. However, the cognitive personnel error j in this event is different in that it involved a conditioned response to an alarw. Previous corrective actions are not appilcable to this event.
| |
| G. COMPONENT FAILURE DATA:
| |
| ( Manufacturer Nomenclature Model Number MFG part Number Chase-Shwout Fuse Type 1 30 amp A60X30 2261m(083188)/8 j
| |
| | |
| h[~ {
| |
| LICENSEE EVENT REPORT (LER)
| |
| Facility Name (1) Dock 2t Numb 3r (2) Page L' Braidwood.' nit 2 01 51 01 01 01 41 51 7 1 lofl0 3
| |
| 'Itle (4) Manual iteactor Trip Due to Inoperable Rod Control System Q Event Date (5)
| |
| Year LER Number (6)
| |
| Sequential Revision Recort Date (7) Other Facilities Involved (8)
| |
| Facility Names Docket Numberft)
| |
| Month Day Year ///
| |
| fj/, /j/j f
| |
| / Month Day Year
| |
| // Number /// Number NONE 01 51 01 01 01 I l
| |
| ~~~ ~~~
| |
| 01 5 31 0 81 8 Bl' 8 0l0l9 010 0l6 11 4 Bl B of Sl 010101 l l THIS REPORT IS SUBMITTED PURSUANT TO THE REQUIREMENTS OF 10CFR OPGM (Check one or more of the followinal fil) 20.402(b) _ _. 20.405(c) _E_ 50.73(a)(2)(iv) _ 4.71(b)
| |
| POWER ___ 20.405(a)(1)(1) ___. r0.36(c)(1) _ 50.73(a)(2)(v) _ 73.71(c)
| |
| LEVEL ___ 20.405(a)(1)(41) ___ 50.36(c)(2) _ 50.73(a)(2)(vil) _ Other (Specify (101 0l0!O _ 20.405(a)(1)(lit) _ 50.73(a)(2)(1) _ 50.73(a)(2)(vill)(A) in Abstract
| |
| // //// /// /// /// ////
| |
| /
| |
| ___ 20. 405 ( a ) ( 1 ) ( i v ) _ 50.73(a)(2)(li) _ 50.73(a)(2)(viii)(B) below and
| |
| $%,/,,/,,S,H,,/,,S,,/,M,,,/,,,/,M,,,S,/,
| |
| SH H H SHH H **'(*H n'"> 50 ' (*H2H*> '**U
| |
| * **'!* n ' "" > -
| |
| LICENSEE CONTACT FOR THIS LER f12)
| |
| Name TEtEPHONE NUPSFR Christopher M. Wiegand, Nuclear Engineer, Ext. 2492 g g gg _g g g g CnMPLETE ONE LINE FOR EACH COMPONEN FAILURE DLSCRIBED IN THIS REPORT f13l CAUSE SYSTEM COMPONENT MANUFAC. REPORTABLE CAUSE SYSTEM COMPONENT MANUFAC. REPORTABLE TUprt TO NPRDS TURER TO NPRDS_.
| |
| x vIL SI 11 al - wi il 21 0 N I I l l l l 1 l I I I l l l l l 1 1 I I l SUPPLEMENTAL REPORT EXPECTED (14) Expected Month i Day l Year p SubeissIon lyes (If ven_ comolete EXPECTED SUBMISSION DATE) X l N0 ak W ; lg lg ABSTRACT (Limit to 1400 spaces, i.e. approximately fif teen single-space typewritten lines) (16)
| |
| At 0348 on May 30, 1988, whlie withdrawing Control Bank Rods, an urgent and non-urgent alarm occurred followed by the release of rods in Shutdown Banks C, D E, and Group 2 Rods in Shutdown Bank A and Control Banks A and C. At 0406 the reactor was manually tripped. The root cause of this event was the failure of a miscellaneous Electric Room (MER) Ventilation Fan. shis failure caused the temperature of the Rod Control power cabinets to increase to the Thermal Overload Protection setpoint. Actuation of the Thermal Overload Protection de-energized the power supplies which resulted in the rods being released.
| |
| Temporary cooling fans were installed in the HER until the ventilation fan repairs were completed. Procedural revisions are being processed to specify the personnel to be notified should the ventilation system become inoperable. This should allow appropriate actions to be taken in a timely manner to maintain HER Ambient Temperature within the limits of the power supplies. Additionally, an analysis and evaluation of the HER Ventilation System *.o verify its adequacy in cooling durie g the summer month = will be performed. There have been no previous occurrences.
| |
| l l
| |
| p .
| |
| U 2160m(061388)/42 L-___________-______-- _ _ - _ _ _ _ _ _ _ _ __ _
| |
| | |
| LICENSEE EVENT REPORT fLER) TEXT CONTINUATION FACILITY NAME (1) DOCKET NUPSER (2) LER NUPBER (6) Pace (3)
| |
| Year j/jj
| |
| // Sequential //j/ Revision ff
| |
| . /// Number /// Number Braidwoo W nit 2' 0 l 5101010141517 8l8 - 01019 - 01 0 01 2 0F 01 3 TEXT. Enerny Industry Identification System (EIIS) codes are identified in the text as [xx)
| |
| A. PLANT COPE)ITIONS PRItm TO EVENT:
| |
| Unit: Braidwood 2': Event Date: Mav 30. 1988 ; Event Time: 0406 HDDE: _Z_ .Startuo ; Rx Fewer:_H_; RCS [A8) Temperature / Pressure: 558 Dearnes F/2233 osia
| |
| : 8. DESCRIPTION OF EVENT:
| |
| The Miscellaneous E10ctric Room (MER) ventilation fan 2VE01C [VL] was inoperable at the beginning of this event and which contributed to the severity bf the event.
| |
| 'At 0338 on May'30, 1988, mode chanje checklist 28wGp 100-2T2, was completed. All shutdown banks were fully i withdrawn, Boron Concentration wat, 884 ppe, and the estimated critical posttion was control bank D at 105 steps.
| |
| At 0343 rod control was placed in manual mode and control banks'were withdrawn in overlap. At 0348 control bank A was fully withdrawn,' control tank 8 was being withdrawn to 116 steps, and control bank C was being withdrawn to step 1. Although the uni", was administrative 1y in Mode 2. It was actually Subcritical with a K,gy cggy, of approximately 0.98.
| |
| At 0348, ahlle withdrawing control bank rods, an urgent and non-urgent alare occurred followed by the release of the rods on shutdown banks C, D, E, and the rods in Group ?? of Shutdown Sank A and the rods in group 2 of g control banks A and C.
| |
| \ At 0406 the reactor was manually tripped due to the operat'onal condition of the Rod Control System. All rods were fully inserted, and safe shutdown was accomplished.
| |
| Operator actions neither increased nor decreased the severity of the event. !
| |
| The appropriate NRC notification via the ENS phone system was made at 0432 on May 30, 1988 pursuant to
| |
| '10CFR50.72(b)(2)(ll).
| |
| This event is being reported pursuant to 3CCFR50.73(a)(R)(iv) - any event or condition that resulted in manual or automatic actuation of any engineered safety feature, including the reactor ptotection system.
| |
| C. CAUSE OF THE EVENT The intermediate 'cause of the event was a trip of the Main and Auxiliary Rod Control power Supplies [AA) of the affected power cabinets. The root cause of the event was the inoperability of the 2VE01C Ventilation Fan which resulted in high ambient temperatures in the MER, where the rod control power cabinets are located. The affected power supplies are rated for a maximum ambient temperature of 104 degrees (deg) F. The room temperature at the time of the event was approximately 100 deg T. It is estimated that inside the cabinet, and at the level of the power supplies, the temperature was as high as 110 deg F. The power supplies are equipped with thermal and electrical overload protection. Both of these circuits will trip when the temperature exceeds a preset value. The thermal circuit will reset by itself when the temperature drops to an acceptable level, and ,
| |
| , the electrical trip is reset by removing power to the supply. I When the electrical trips were reset and the temperature had cooled to below 95 deg F, the system was restored
| |
| , to normal.
| |
| l 21CDe(062188)/43 L- -_-_:_-__
| |
| | |
| LICENSEE EYENT REPORT (LER) TEXT CONTINUATION FACILITY NAME (1) DOCKET NUMBER (2) _._LER NUMBER (6) Pace #3)
| |
| Year Sequential Revision
| |
| /{/{/{ Number , //{ Number
| |
| / ) Braidwand. Unit 2 0 1 5 1 0 1 0 1 0 1 41 51 7 8I8 - 010l9 - 01 0 01 3 0F 01 3 TEXT Energy Industry Identification System (EIIS) codes are identified in the text as [xx)
| |
| D. SAFETY ANALYSIS:
| |
| There were no safety consequences as a result of this event. All equipment operated as designed. The manual trip was a consorsative action. Under worst case conditions, operating at 1001. power will all rods fully withdrawn, there wou'4 still have been no safety consequences as the reactor would have automatically tripped and safe shutdown would have been accomplished using plant procedures. Additionally, this event is described in section 15.4.3 of the Final Safety Analysis Report, " Rod Cluster Control Assembly Misoperation".
| |
| E. CORRECTIVE ACTIONS:
| |
| Imediate corrective actions were to repair ventilation fan 2VE01C and install temporary cooling fans in the HER.
| |
| Actions taken to prevent recurrence include revising Operating Rounds Procedures, Sw0P 199-A53 and Bw0P 199-A41, which monitor HER Temperu ure. The purpose of this change is to inform the Technical Staff Nuclear Group whenever the MER Temperature exceeds 95 deg F. This will allow appropriate actions to be taken in a timely manner to maintain MER ambient temperature within the design limits of the power supplies. These procedure revisions will be tracked to completion by action item numbers 457-200-88-08701, and 457-200-88-06102, respectively.
| |
| In addition, procedures BwAP 0-34-A',, and BwAR 0-31-A3, will be revised to include notifying the Technical Staff Heating Ventilation and Air Conditioning Group (HVAC) System Test Engineer whenever the 2VE01C fan becomes inoperable. These will be trecked to completion by item numbers 457-200-88-08703, and 457-200-88-08704, respectively.
| |
| Although the 2VE01C fan was responsible for the high ambient temperature in the MER that lead to this event, a similar event could happen even though the fan was operating properly. The MER is cooled by outside air, and therefore the room can be only as cool as the outside temperature. With the possibility of summertime temperatures reaching 95 deg or higher, the trip setpoint of the power supplies could again be reached.
| |
| Therefore, an evaluation of the MER Ventilation System to verify its adequacy in cooling the MER during the summer months will be performed. This will be tracked to completion by item number 457-200-88-08705.
| |
| F. PREVIOUS OCCURRENCES:
| |
| There have been no previous occurrences of a reactor trip as the result of excessive ambient temperatures in the HER.
| |
| G. COMPONENT FAILURE DATA:
| |
| MANUFACTURER NDPENCLATURE MODEL NUteER MFG PART NUPeER
| |
| (
| |
| Westinghouse Relay, Overload None AA33A lO 2160m(061388)/44
| |
| | |
| UF- T LICENSEE EVENT REPORT (LER)
| |
| Fccility Name (1) Docket Number (2) Pace (3)
| |
| Byron. Unit 1 01Sl01010141514 1lof!0l5 l REACTOR TRIP DUE TO R0P. DROP DURING MANUAL CONTROL ROD MOTION Event Date (5) LER Nur.ber (6) Report Date (7) Other Facilities Involved (8)
| |
| Month Day Year Year // Sequential /j// Revision Month Day Year Facility Names l Docket Number (s)
| |
| ,/j/
| |
| //
| |
| f Number ff
| |
| /// Number I NONE 0151010101 I i
| |
| ~ ~
| |
| 014 1 IB BIB 8l8 0IOl2 0l0 015 1l3 8I 8 015101Of01 l l OPNM THIS REPORT IS SUBMITTED PURSUANT TO THE REQUIREMENTS OF 10CFR (Check one or more of the followinal (11) 1 20.402(b) _ 20.405(c) .2_ 50.73(a)(2)(lv) _ 73.71(b)
| |
| POWER _ 20.435(a)(1)(1) ___ 50.36(c)(1) _ 50.73(a)(2)(v) _ 73.71(c)
| |
| LEVEL 20.405(a)(1)(ll) 50.36(c)(2) 50.73(a)(2)(vil) Other (Specify (101 0l9!8 ___
| |
| _ 20.405(a)(1)(iii)
| |
| _ 50.73(a)(2)(1)
| |
| _ 50.73(a)(2)(viii)(A) in Abstract
| |
| / /// /,/,/,/,/,/,/,/,/,/,/,/,/,//,/,/,/,/,/,/,/, _ 20.40ha)(1)(iv) _ 50.73(a)(2)(li) __. 50.73(a)(2)(viii)(B) below and in
| |
| //////' /' /' /j/' /' /' /jj/j' / (j/j///'/ y/j y // /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 NUPEER AREA CODE Lee Suws. Assistant Superintendent Technical Services Extension 2214 8l115 2131dl-l5141di COMPLETE ONE LINE FOR EACH COMPONEN FAILURE DESCRIBED IN THIS REPORT (13)
| |
| CAUSE SYSTEM COMPONENT MANUFAC- REPORTA8LE CAUSE SYSTEM COMPONENT MANUFAC- REPORTABLE l TURER TO NPRDS TURER TO NPRDS I l I I I I I I I I I I I I i l l I I I I I I I I I I I I '/ l SUPPLDENTAL REPORT EXPECTED (14) Expected Month l Day l Year Submission lyes (If ves. comelete EXPECTED SUBMISSION DATE) X l NO l l l /
| |
| ABSTRACT (Limit to 1400 spaces, i.e, approximately fifteen single-space typewritten lines) (16) l At 2120 on April 18, 1988, a licensed reactor operator manually inserted the controlling bank of control rods one step to adjust Axial Flux Difference. The " Power Range Flux Rate High Reactor Trip" annunciator actuated and the reactor trip breakers opened. Control room operators entered and compiled with " Reactor Trip or Safaty Injection Unit 1 Emergency Procedure". The Auxiliary Feedwater Pumps started due to low-low steam generator levels that resulted free the trip at high power. Stable plant conditions were achieved in Hot Standby at 2230 on April 18, 1988.
| |
| The intermediate cause of the reactor trip was the dropping of one or more control rods into the reactor core, which resulted in the flux high negative rate reactor trip. Troubleshooting efforts failed to determine a root cause of the dropped rods. It is believed that an intermittent component failure in the rod control system caused the event, but the component did not remain in the failed mode following the reactor trip.
| |
| Extensive troubleshooting was conducted to locate discrepancies that may have caused the rod drop. A number of loose electrical connections were identified in the rod control power cabinets and all were repaired. Movable gripper coil and lif t coil power bridge thyristors were replaced in the 280 power cabinet due to their likelihood of causing this event. On April 21, 1988, all approvals required to start up the plant following a reactor trip for which no root cause has been detemined, were obtained and the plant entered the Startup operational mode.
| |
| l Previous occt rences of reactor trips caused by dropped rods are documented in Licensee Event Reports
| |
| [V] 85-042, 85-063 and 86-028.
| |
| ! [
| |
| (2061H/0226M)
| |
| E
| |
| | |
| LICENSEE EVENT REPORT (LER) TEXT CONTINUATION FACILITY NAME (1) DOCKET NUMBER (2) LER NUMBER (6) Pace (3)
| |
| Sequential Year /// // Revision fj/j/
| |
| / Number /j//j f
| |
| / Number Byron. Unit 1 015101010141514 8iB - 0l012 - 0 10 0 12 0F 0I!
| |
| l TEXT Energy Industry Identification System (E!IS) codes are identified in the text as [xx)
| |
| A. PLANT CONDITIONS PRIOR TO EVENT:
| |
| Event Date/ Time 018/88 / 2120 Unit 1 N)DE 1 - Power Ooerations Rx Power 98% . RCS [A8] Temperature /Pressurs Normal Doeratino
| |
| : 8. DESCRIPTION OF EVENT:
| |
| There were no systems or components inoperable at the beginning of this event that contributed to this event.
| |
| As a result of routine monitoring of main centrol board instrumentation, a licensed reactor ooerator noted that indicated Axial Flux Difference (AI), which is derived from the Power Range Nuclear Instruments
| |
| [IG), was not on the desired target value. Although the Byron Tec;mical Specifications permit unrestricted plant operation as long as AI is maintained within a prescribed band on either side of the target valve.
| |
| It is Byron operating policy to maintain AI at the target value, if plant conditions allow. The indicated AI could be corrected to the target value in this case by partially inserting the Control Bank D rods from their withdrawn position of 219 steps. At 2120 on April 18, 1988, the reactor operator moved the Rod Drive [AA) Selector Switch from the AUTOMATIC position to the MANUAL position, and pushed the Rod Drive In-Hold-Out switch to drive the Control Bank D Grcup 2 rods in one step. " Rod at Bottom" and
| |
| " Digital Rod Position Indication Urgent Failure" printed on the Sequence of Events Recorder. Power Range
| |
| ,q Nuclear Instrumentation measured a high negative flux rate condition on two of four rka*mels. The " Power
| |
| { Range Flux Rate High Reactor Trip" annunciator alarmed and a reactor trip occurred. ine control room operators entared and complied with " Reactor Trip or Safety snjection Unit 1 Emergency Procedure" (IBEP-0). A Rod Control urgent failure alarm followed the opening of the reactor trip breakers due to expected voltage reguistion failures in all Rod Drive [AA) power cabinets, however, no rod control urgent failure alarms actuatsd prior to the reactor trip. The lack of urgent failure alarms prior to the rod drop suggested blown rod control power cabinet gripper fuses as a possible cause. Operators were dispatched to ir.spect for blown fuses, but all fuses were determined to be intact.
| |
| At 2122 the 1A and 18 Auxiliary feedwater Pumps (BA) (AFP) started automatically due to low-low steam 9enerator levels. A Feedwater Isolation Signal (FWIS) occurred due to the Reactor Trip and. low Reactor Coolant average temperature (Tayg). These Engineered Safety Feature (ESF) actuations are expected following a Reactor Trip from high pwer level. At 2148, the Start-up Feedwater Pump [SJ) was started, the j FWIS was reset, and a flowpath was aligned to the steam generators from the Start-up Feedwater Pump. At 2157 the IB AFP was stopped and at 2230 the 1A AFP was stopped. All operator actions taken during this event were correct and contributed to the safe conclusion of this event. Stable plant conditions were achieved in Hot Standby (Mode 3) by 2230 on April 18, 1988.
| |
| This Licensee Event Report (LER) is submitted in accordance with 10CFR50.73(a)(2)(iv) due to the automatic ESF actuations.
| |
| l l
| |
| 0 N.,Y l
| |
| (20MM/0226M) !
| |
| | |
| LICENSEE EVENT REPORT (LER) TEXT CONTINUATION FACILITY NAME (1) DOCKET NUNIER (2) LER NUMBER (6) pace (3)
| |
| Year /// Sequential lU/' Revision fff Number ff Number Byron. Unit 1 0 1 5 1 0 l 0 1 0 1 41 51 4 8l8 - Ol6I2 - 0 10 01 3 0F 015 l TEXT Energy Industry Identification System (EIIS) codes are identified in the text as btx]
| |
| C. CAUSE OF EVENT:
| |
| The intermediate cause of the reactor trip was the dropping of one or more control rods into the reactor core, which resulted in the flux high negative rate reactor trip. It could not be determined how many rods dropped Or specifically which rods dropped due to the rapid progression of the event and the unavailability ;
| |
| of the Process Computer (ID] Alarm Typer, which was mechanically Janused. The Alarm Typer may have indicated specific rod misalignments in the early stages of the event. This information would have focused the troubleshooting effort following the reactor trip. Extensive troubles %ing eibts failed to detemine a root cause of the dropped rods, however, minor rod control s / stem discrepancies were corrected as documented in the Corrective Actions section of this report. It is Lelieved that an iMermittent component failure in the rod control system caused the event, but the component did not rema?n in the failed mode following the reactor trip.
| |
| D. SAFETY ANALYSIS:
| |
| All ESF systems actuated and performed as designed. The Reactor Protection System (JG) responded properly to a power range flux high negative rate condition sensed by two of the four power range nuclear instruments. The AFP's automatically started due to low-low steam generator levels that resulted following the reactor t~ rip from high power. Plant /pubile safety were not affected by this event.
| |
| E. CORRECTIVE ACTIONS:
| |
| p
| |
| ( Immediately following the reactor trip on April 18, 1988, the rod control power cabinet gripper fuses were C inspected for blown fuse indicators and the Process Computer Alam Typer jam was cleared. No blown fuse indicators were actuated. By 2300 on April 18, 1988, rod drive system conditions at the time of the reactor trip had been fully researched and documented, and detailed troubleshooting efforts were consnenced by Technical Staff engineers and Instrument Maintenance (IM) technicians. Rod drive power cabinet circuit cards were tested and no failures were identified. The voltage outputs of all Direct Current (DC) power supplies in each power cabinet were checked and determined to be satisfactory. At 0205 on April 19, 1988, IM technicians checked all stationary gripper coil fuses to determine if any had failed without triggering the appropriate blown fuse indicator. All fuses were intact.
| |
| On the afternoon of April 19, 1988, Byron Station management requested technical assistance from the Braidwood Station Technical Staff and the Braidwood Westinghouse Site Engineering Team due to their recent experience in rod control system problems. Also, Byron Station's management requested that Westinghouse provide a rod control system technical expert to assist in the investigation.
| |
| At 2320 on April 19, 1988, alarm circuitry was tested to determinte if a rod control urgent failuri condition could exist in a power cabinet and not generate the associated alarw. All alarm circuitry functioned properly during the test.
| |
| l At 0115 on April 20, 1988, a temporary change to the " Checkout of the Bank Overlap Unit Surveillance Procedure" (IBVS XPT-2) was performed. This modified procedure permitted exercising of the control rods while strip chart recorders monitored the follcwing information:
| |
| : 1. Vref (reference voltage) '
| |
| , 2. V,, (sat.uration voltage) f 3. Verr (error voltage!
| |
| d 4. Vrip (ripple voltage) l
| |
| : 5. Stationary gripper coil voltage - mechanism #1 1
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| (2061M/0226M) l l
| |
| | |
| LICENSEE EVENT REPORT (LER) TEXT CONTINUATION FACILITY NAME (1) DOCKET NUtWER (2) LER NUteER (6) Pace (3)
| |
| Year /j/j/ Sequential Revision
| |
| /j///j/j/ Nu*nba r j//
| |
| / Number Byron. Unit 1 0 1 5 1 0 1 0 1 0 1 41 51 4 8l8 - 0l0l2 - 010 014 0F 0l5l l TEXT. Energy Industry Identification System (EIIS) codes are identified in the text as (xx]
| |
| E. CORRECTIVE ACTIONS: (Continued)
| |
| : 6. Lift coil voltage - mechanism #1
| |
| : 7. Moving coil voltage - mechanism #1 l
| |
| : 8. Moving call voltage - mechanism #2 {
| |
| : 9. Moving coil voltage - mechanism #3 '
| |
| : 10. Moving Phase Control, Phase A
| |
| : 11. Moving Phase Control, Phase B
| |
| : 12. Moving Phase Control, Phase C The tendified IBVS XpT-2 was approved by station management with the understanding that this was an unusual system configuration that presented a high risk for rod drops during the testing. While testing power cabinet 2BD (power to Control Bank D Group 2 rods) at 0400 during the first execution of the modified 1BVS XPT-2, a rod control urgent failure alarm actuated due to a moving phase failure. IM technicians pulled the circuit card frame out of the 280 power cabinet to inspect for faulty electrical connections on the frame. While inspecting, two of the DC power supplies were shorted to ground, and the partially withdrawn -
| |
| control rods powered by the 250 power cabinet dropped to the bottom of the core. The power cabinet alignment was restored and the rods were again withdrawn. Attempts to repeat the rod control urgent failure were unsuccessful.
| |
| At 1200 on April 20, 1988, the voltages for the lif t colla, stationary gripper coils and movable lif t coils
| |
| , supplied by the 280 power cabinet were determined to be acceptable. The line voltages of the rod drive
| |
| ( motor-generator setc were checked and were acceptable. At 1830 movable gripper coil currents were measured k in all power cabinets by forcing the bridge thyristors to conduct. All currents measured normal and steat!y. At 2047 all power cabinets were deenergized, and Electrical Maintenance (EM) technicians checked .
| |
| all electrical connections in the power cabinets for tightness. A total of forty-eight loose connections .
| |
| were located and tightened or repaired. power cabinet electrical connections in the bus ducts were also .
| |
| checked.
| |
| At 2120 on April 20, 1988, all movable gripper coil and lif t call power bridge thyristors were removed from the 280 power cabinet and replaced with thyristors from spare parts. This action was taken due to the dropped rod symptoms and the potential for thyristor intermittent failure. Sampilng resistor connections in all power cabinets were checked. IM technicians cleaned and tightened the circuit card edge connectors in the 280 power cabinet. At 0930 on April 21, 1988, the power cabinets were entrgized and strip chart traces of the twelve parameters tabulated previously in this report section were analyzed. No abnormal conditions were indicated.
| |
| On April 21, 1988, all approvals required to start up the plant following a reactor trip for which no root cause has been determined, were obtained and the plant entered the Startup Operational Mode (Mode 2) at 1149. {
| |
| The " Operating Shif t Turnover and Relief Administrative Procedure" (BAP 335-1) will be revised to include a check for normal condition of the process Computer Alarm Typer by the NSO. Implementation of this revision is tracked by Action Item Record 454-225-88-0102.
| |
| 1 1 %
| |
| I -
| |
| - l (2061M/0226M)
| |
| | |
| m .
| |
| 1 LICENSEE EVENT REPORT (LER) TEXT CONTINUATION FACILITY NAME (1) DOCKET NUPSER (2) LER NUPSER (6) . Paoe (3)
| |
| Year Sequential Revision I
| |
| ((/{/
| |
| / Number /{/{/
| |
| / Number ]
| |
| Byron. Unit 1 0 1 5 1 0 l 0 l 0 l 41 51 4 8l8 - 0i0l2 - 0l0 01 5 0F 01 5 TEXT Energy Industry Identification System (EIIS) codes are identified in the text as (xx] j F. PREVIOUS OCCURRENCES:
| |
| Reactor trips due to rod dreps have occurred previously as described in the following LER's:
| |
| LER NUPSER I1111 85-042 (Unit 1) Reactor Trips Due to Dropped Rods 85-063 (Unit 1) Reactor Trip Due to Turbine Trip Above P-7 86-028 (Unit 1) Manual Reactor Trip Due to Rod Drop Caused by Faulty Circuit Cards Root causes were determined in all of these dropped rod incidents and appropriate corrective actions were taken.
| |
| G. COMPONENT FAILURE DATA:
| |
| a) MANUFACTURER NDPENCLATURE MODEL N M MFG PART NUPSER Not Applicable b) RESULTS OF NPRDS SEARCH:
| |
| / Not Applicable N],/
| |
| l l
| |
| 1 ;
| |
| (2061M/0226M)
| |
| | |
| c LP 5^
| |
| LICENSEE EVENT REPORT (LER)
| |
| F:cility Name (1) Docket Number (2) Pace (3)
| |
| Svron. Unit 2 01 51 01 01 01 di 51 5 1 of!0!4 Title (4) Reactor Trip Due to Control Rod Drop Caused by Intermittent Component Failure in the Rod Drive System Event Date (5) LER Number (6) Report Date (7) Other Facilities Involved (8)
| |
| Month Day Year Year /// Sequential f/j/j/ Revision Month Day Year Facility Names l Docket Number (s) fH
| |
| / Number /// Number '
| |
| NONE 01510101Of l I
| |
| ~ ~~~
| |
| 01 6 0 12 8l8 81 8 0 1 0 16 010 0l6 2l1 81 8 01 51 01 01 of I l THIS REPORT IS SUBMITTED PURSUANT TO THE REQUIREMENTS OF 10CFR (Check one or more of the followine) (11) 1 20.402(b) __ 20.405(c) _L 50.73(a)(2)(lv) __._ 73.71(b)
| |
| POWER _ 20.405(a)(1)(1) .___ 50.36(c)(1) _ 50.73 s)(2nv) _ 73.7nc)
| |
| LEVEL 20.t05(a)(1)(li) 50.36(c)(2) 50.73(a)(2)(vil) Other (Specify (101 0!9 l4 __._
| |
| _ 20.405(a)(1)(lii)
| |
| _._ 50.73(a)(2Hi)
| |
| _ 50.73(a)(2)(viii)(A) in Abstract ffffffffffffffffffffffffff _ 20.405 a n n o v) _ w.73an2nu) _ 50.73(a)(2)(viii>(B) beiow and in 3HHHHHHf6HnHHH 28 d'54"'"v) 50 73 4 "2" " " 5' 73(*" "") 'a u I LICENSEE CONTACT FOR THIS LER (12)
| |
| Name TELEPHONE NUPBER AREA CODE Don Brindle. Operatino Enaineer Ext. 2218 8l1I5 213141 lSl'4141 COMPLETE ONE LINE FOR EACH COMPONENT FAILURE DESCRIBED IN THIS REPORT (13)
| |
| CAUSE SYSTEM COMPONENT MANUFAC- REPORTABLE CAUSE SYSTEM COMPONENT MANUFAC. REPORTABLE TURER TO NPRDS TURER TO NPRDS I I I 1 l l I I I I I I I I l I I I I I I I I I I I I l i
| |
| / SUPPLEMENTAL REPORT EXPECTED (14) Expected Month l Day I Year
| |
| ( Submission lyes (If ves. comelete EXPECTED SUBMISSION DATE) X l ND l ll ll ABSTRACT (Limit to 1400 spaces, i.e, approximately fifteen single-spa e typewritten lines) (16) {
| |
| At 0640 on June 2,1988, with the reactor at 94% power and the Rod Control Selector Switch in AUTOMATIC.
| |
| the " Power Range Flux Rate High Reactor Trip" annunciator actuated and the reactor tripped. Control Room operators entered and compiled with " Reactor Trip or Safety Injection Unit 2 Emergency Procedure." The Auxl11ery Feedwater Pumps started due to low-low steam generator levels that resulted from the trip at high power. Stable plant conditions were achieved in Hot Standby at 0730 on June 2, 1988.
| |
| The Intermediate cause of the reactor trip was the dropping of control rods into the reactor core, which resulted in the high flux negative rate reactor trip. Troubleshooting efforts failed to determine a root cause of the dropped rods. It is believed that an intermittent component failure in the rod drive system cevsed the event, but the component did not remain in the failed mode following the reactor trip.
| |
| Troubleshooting was conducted to locate component failures that may have caused the rod drop. A number of degraded fuses were identified in the rod drive power cabinets and aH were replaced. On June 3, 1988. all approvals required to start up the plant following a reactor trip for which no root cause has been l determined were obtained and the unit entered the Startup operational mode.
| |
| l Previous occurrences of reactor trips caused by dropped rods on Unit I are documented in Licensee Event Reports 85-042, 85-063, 86-028 and 88-002.
| |
| I !
| |
| (0029R/0002R)
| |
| | |
| LICENSEE EVENT REPORT (LER) TEXT CONTINUATION FACILITY NAME (1) DOCKET NUMBER (2) LER NUtBER (6) Pace Cl)
| |
| Year // Sequential / ,p/
| |
| / Revision fj/j/
| |
| // Number f
| |
| /// Number Byron. Unit 2 0 l S I 0 1 0 1 0 1 41 51 5 8l8 - 0Ln16 - 0l0 O! 2 0F 01 4 l TEXT Energy Industry Identification System (EIIS) codes are identified in the text as (xx)
| |
| A. PLANT CONDITIONS PRIOR TO EVENT:
| |
| Event Date/ Time 6'2/88
| |
| / / 0640 Unit 2 MODE 1 - Power Operations Rx Power 94% RCS (AB] Temperature / Pressure Normal Oneratino B. Q11CRIPTION OF EVENT:
| |
| The Rod Control Selector Switch was in the AUTOMATIC position and no red motion was demanded by the rod control system, Shif t relief was in progress and the oncoming Nuclear Station Operator (NS0) (licensed reactor operator) was clearing a mechanical jam of the Process Computer Alarm Typer. At 0640 on June 2, 1988, with Unit Two operating at 94 percent reactor power, the Power Range Nuclear Instruments (IG) sensed a high negative flux rate condition indicative of dropped control rods. The " Power Range Flux Rate High Reactor Trip" annunciator actuated in the Main Control Room, and the reactor trip breakers opened. The licensed control room operators entered and compiled with " Reactor Trip or Safety Injection Unit 2 Emergency Procedure" (2BEP-0). A Feedwater Isolation occurred due to the Reactor Trip and low Reactor Coolant Average Temperature (Tayg), which is an expected Engineered Safety Features (ESF) actuation following a reactor trip. The Auxiliary Feedwater Pumps (AFP) (BA) automatically started in response to low-low steam generator levels caused by indicated level shrink after the trip. No annunciators related to the Rod Drive System (RD) (AA) actuated prior to the trip to indicate a possible cause. Expected post trip rod drive Urgent Failure alarms were actuated for voltage regulation failures due to the opening of the
| |
| , reactor trip breakers. The lack of urgent alarms prior to the trip suggested blown fuses as a possible cause for a rod drop event that would cause th6 negative rate reactor trip.
| |
| At 0655 the Startup Feedwater Pump (50 ras started. At 0656 the NSO manually reenergized the Source Range InstrumentsbecauseIntermediateRangethannelI35appearedtobeundercompensatedandwoulddelaythe automatic reenergization. At 0707 the Feedwater Isolation signal was reset and a feedwater flow path from the Startup Feedwater Pump to the steam generators was established. At 0725 the diesel-driven AFP was stopped and at 0728 the motor-driven AFP was stopped. At approximately 0730 stable plant conditions were achieved with Unit Two in Hot Standby (Mode 3).
| |
| All Operator actions taken during this event were correct and contributed to its safe conclusion. This Licensee Event Report (LER) is submitted in accordance with 10CFREO.73 (a)(2)(iv) due to the automatic ESF actuations.
| |
| C. CAUSE OF EVENT:
| |
| The intermediate cause of the reactor trip was the dropping of control rods into the reactor core, which resulted in the power range flux negative rate high reactor trip. It was determined that rods dropped from the 1AC rod drive power cabinet, initiating the event. Three Stationary Gripper Phase fuses were blown during the event. Other Stationary Gripper Phase fuses remained intact and were capable of maintaining all control rods in their withdrawn positions. Therefore, the three blown fuses alone could not have caused multiple dropped rods. Fuses that conduct current to the Control Rod Drive Mechar. ism (CRDM) coils from the RD power cabinets were resistance checked using a Digital Low Resistance Ohanieter, and several fuses had resistance readings slightly higher than permitted by acceptance criteria. The slight degradatina of these
| |
| : l. fuses may have contributed to the control rod drop by limiting current flow to the CRDM coils, however, the j minor nature of the fuse degradation does not provide conclusive evidence for designating these fuses as l A the root cause. Troubleshooting efforts failed to determine a root cause for the dropped rods. It is
| |
| ('')
| |
| ~
| |
| believed that an intermittent component failure in the rod drive system caused the event, but the component failure did not continue following the trip.
| |
| I I (0029R/0002R)
| |
| | |
| l j
| |
| LICENSEE EVENT REPORT fLER) TEXT CONTINUATION FACILITY NAME (1) DOCKET NureER (2) LER NUMBER (6) Pace (3)
| |
| Year ///j Sequential // Revision l jj/
| |
| // Number /p/j/j
| |
| / Number Byron. Unit 2 0 l 5 1 0 1 0 1 0 1 41 51 5 8I8 -
| |
| 0l0l6 - O l0 01 3 0F 01 4 TEXT Energy Industry Identification System (EIIS) codes are identified in the text as (xx)
| |
| C. CAUSE OF EVENT: (Continued)
| |
| Intermediate Range Channel N35 was determined to be properly compensated. The P-6 permissive that automatically reenergizes the Source Range Instruments actuated shortly af ter the manual reenergization, within the expected 15 to 18 minute time frame following the reactor trip.
| |
| j 1
| |
| D. SAFETY ANALYSIS:
| |
| All ESF systems actuated and performed as designed. The Reactor Protection System (JG) responded properly to a power range flux high negative rate condition sensed by two of the four power range nuclear instruments. The Auxiliary Feedwater Pumps automatically started due to low-low steam generator levels ;
| |
| that resulted following the reactor trip from high power. At no time during this event was the health or safety of the public adversely affected or threatened.
| |
| E. CORRECTIVE ACTIONS:
| |
| Innedtately following the reactor trip or. June 2,1986, the rod drive power cabinet fuses were inspected for blown fuse indicators and the Process Computer Alam Typer jam was cleared. Blown fuse indicators were actuated on the following three fuses:
| |
| Eust Cabinet Phase A Stationary Grippers Group 1 2RD03J 2BD f, Phase B Stationary Grippers Group 3 2RD02J SCDE b [
| |
| Phase B Stationary Grippers Group I 2RD06J 1AC ov The rod drive motor-generator (MG) sets' protective relays were checked for targets indicating trouble, and none were found. By 0830 the rod drive conditions at the time of the trip had been documented. At 1130 the RD bus duct was checked for shorts, both phase to phase and phase to ground, but no shorts were found.
| |
| At 1200 the three fuses associated with the blown fuse indicators were verified to be bad. The balance of the stationary phase fuses were checked using a Digital Low Resistance Ohmmeter (DLRO). By visual and DLR0 checks, 23 fuses were found to be either cracked or electrically unacceptable, so all stationary phase fuses in all power cabinets were replaced. From 1530 to 2150 the fuses leading out to the Control Rod Drive Mechani2m's (CRDMs) coils were resistance checked utilizing " CONTAINMENT PENETRATION CONDUCTOR PROTECTIVE DEVICES 260 VAC RCD POWER (FUSES) Electrical Surveillance Procedure" (2BHS 8.4.1.a.3-1) and the fuse clips tightened. Although 14 fuses were found to be outside acceptance criteria for the surveillance, none had degraded enough to have caused rods to drop. All fuses failing the acceptance criteria were replaced, and the replaumont fuses were checked before installation. In parallel with the fuse checks, the cables to the stationary coils on the reactor head package from RD power cabinet 1AC were checked for circuit shorts, and none were found. On June 3, 1988 at 030'/, the RD system was energized and reset, but an urgent alarm remained actuated on RD power cabinet SCDE. The card edge lights on the error detector circuit cards indicated a stationary gripper voltage regulation failure. The voltages into the thyristors and out of the power supplies were checked and found to be acceptable. After troubleshooting and swapping out the associated firing circuit card and the signal processing circuit card, the urgent alarm was cleared and determined to be due to a loose card edge connector on the stationary gripper firing circuit card. A cracked solder connection on a capacitor was fixed and the card edge connectors tightened. The rods were exercised from 0915 to 1115 to verify proper operations. Approval for startup following a reactor trip for j which no root cause has been detemined was obtained and the plant entered the Startup operational mode at
| |
| ( 1227 on June 3, 1988.
| |
| k I
| |
| I (0029R/0002R)
| |
| | |
| LICENSEE EVENT REPORT (LER) TEXT CONTINUATION FACILITY NAME (1) DOCKET NUMBER (2) LER NUteER (6) Pace 431 Year ,/,/j/ Sequential //j ff
| |
| / Revision
| |
| /// Number Number
| |
| ///
| |
| Bvron. Unit 2 0 l 5 I O I 0 1 0 1 41 51 5 8I8 -
| |
| 0l0l6 - Ol0 01 4 0F 01 4 l TEXT Energy Industry Identification System (EIIS) codes are identified in the text as [xx]
| |
| E. CORRECTIVE ACTIONS: (Continued) ,
| |
| As a result of tho' fuse discrepancies discovered on Unit 2, fuses on Unit I were resistance checked on June 9, 1988. One fuse was determined to be broken and was replaced with a new fuse.
| |
| Personnel from Cominonwealth Edison's Pressurize) Water Reactor facilities and Corporate Offices have joined to form a task force, whose goal is to reduce reactor trips caused by rod drive system problems.
| |
| F. PREVIOUS OCCURRENCES:
| |
| LER NUPSER TITLE 85-042 (Unit 1) Reactor Trips Due to Dropped Rods 85-063 (Unit 1) Reactor Trip Due to Turbine Trip Above P-7 86-028 (Unit 1) Manual Reactor Trip Due to Rod Drop Caused by Faulty Circuit Cards 88-002 (Unit 1) Reactor Trip Due to Rod Drop During Manual Control Rod Motion G. COMPONENT FAILURE DATA:
| |
| a) MANUFACTURER NDPENCLATURE MODEL NUPSER mfg PART NUPSER Not Applicable O
| |
| b) RESULTS OF NPRDS SEARCH!
| |
| Not Applicable I
| |
| l l
| |
| l o
| |
| I I
| |
| (0029R/0002R)
| |
| L_____-_____--__ - -
| |
| | |
| BEAIDWOOD SIMULATOR MALFUNCTION f
| |
| | |
| ==Title:==
| |
| Rod Ejection ID: CRF-6
| |
| -Q NO: 6.3.4.4.6
| |
| | |
| == Description:==
| |
| Shear of the CRDM housing, causing a small LOCA and rod ejection.
| |
| Variations: Date 8/12/86 Rev: 2 Selectable Steps Inputs Comments
| |
| : 1. Select rod H6, M12, etc. See MALF CRF-4 grid map for rod location.
| |
| 0 to 2000* spa
| |
| * Flow is based on RCS pres-
| |
| : 2. Select size of leakage through CRDM housing - sure at 2235 psig.
| |
| Note Hole will be partly blocked by the rod, thua
| |
| (
| |
| reducing the maximum flow through the hole.
| |
| Brief Plant Response: (Response based on plant being at full power)
| |
| LOCA will occur, causing a loss of pressurizer pressure and level. The affected rod, if in the controlling bank, will be ejected out of the core with the DRPI and IR rate indication responding accordingly. If the rod were inserted into the core far enough, a positive rate trip is possible. If not, the reactor will trip due to the LOCA. Pressurizer pressure and level both decrease. Charging flow will increase. Pressurizer low pressure reactor trip and ;sfety injection will occur, actuating the safeguards systems.
| |
| O 519M/197M/11 5/89
| |
| - - ' - -~ - - - -
| |
| ~ '' ~ ~ ~ ~ ~ ~ ' ' ~- -
| |
| _T___ _ . E l__ __
| |
| | |
| 9 t
| |
| -BRAIDWOOD SIMULATOR MALFUNCTION.
| |
| f., t
| |
| ,. \, Title Rod Ejection. 'ID: CRF-6
| |
| . Suggested Instructor Actions -
| |
| None.
| |
| Events:
| |
| . None l
| |
| l i
| |
| l l
| |
| I i
| |
| 519M/197M/12 5/89 l
| |
| -- __1*'_~__ "T_"'._"1_~~********-*s'+
| |
| +- ~' ---- -- * -*' ~ ~ ~ .----...e---+~~.- - - - - - - - . - - - . . . . . .
| |
| | |
| BRAIDWOOD SIMULATOR MALFUNCTION hD
| |
| ' V" Titles.-Uncontrolled Rod Motion ID:. CRF-7 50: 6.3.4.4.7
| |
| | |
| == Description:==
| |
| Uncontrolled rod motion in direction selected.
| |
| Automatic mode-failure of UT 412's output.
| |
| ; All mode failure of in relay (K-16) or out relay (K-15).
| |
| ' Variations: -Date: 4/11/87 Rev: 3 Selectable Steps Inputs Comments
| |
| : 1. Select mode 1, 2 or 3- 1 = (Auto) In Automatic mode only, BE31 select
| |
| . direction below.
| |
| 2 = In Manual Mode only, rods move in~ direction of.
| |
| .( last rod motion. ' Starts when initiating signal exists.
| |
| 3 = In both modes, last
| |
| ~ direction that the rods move will cause continuous rod motion..
| |
| : 2. Select direction 1 or 2 Use only when " Auto" (1) selected above 1 = In 2 = Out l
| |
| Brief Plant Responses (IC-16, 75%, all systems in automatic)
| |
| O 519M/197M/13 5/89
| |
| | |
| BRAIDWOOD SIMULATOR MALFUNCTION O \
| |
| Title Uncontrolled Rod Motion ID: CRF-7 (d
| |
| Brief Plant Response (continued):
| |
| Asto - When rod motion is called for, the rods will start movement in the direction selected. The rods will continue to move even after the initiating signal clears. Reactor power and Tavs will respond according to the direction of rod motion. (If rods placed in
| |
| " manual," rod motion will stop.) A reactor trip could occur due to overpower, AT trips, pressurizer pressure, etc., if no action is taken. First out annunciator varies with initial plant conditions and direction of failure.
| |
| All - Rod motion will start in direction of last movement when initiating signal exists. Failure will affect rods in all modes of control except bank select for SCD, SDD, SDE. Reactor power and Tavs will respond according to the direction of rod motion. A reactor trip i
| |
| \s. could occur due to overpowor, AT trips, pressurizer pressure, etc.
| |
| First out annunciator varies with initial plant condition and direction of failure.
| |
| Suggested Instructor Action:
| |
| None.
| |
| Events:
| |
| l l
| |
| None l
| |
| l l .
| |
| [
| |
| %.)
| |
| 519M/197M/14 5/89
| |
| '** ~ ~ ' ~ ' ~ ' ~ ~ ~ '~
| |
| ~ '
| |
| _-_i_J T_' ' ^'
| |
| | |
| FRAIDWOOD SIMULATOR MALFUNCTION r
| |
| ,4 Title ' Auto Rod Speed Controller Failure ID: CRF-8 NO: 6.3.4.4.8
| |
| | |
| == Description:==
| |
| Failure of the rod speed programmer (SY 412A) to a selected value Variations: Date: 8/12/86 Rce 2 Selectable Steps Inputs Comments
| |
| : 1. Select speed 0-80 steps / min.
| |
| Brief Plant Responses (Based on plant being at 50 percent power)
| |
| If Rod Sneed >0 When auto rod motion is required, the rods will move at the selected speed until the actuating signal clears. This could cause both over control and under' control of Tavs (large Tavs swings). Rods will respond properly in
| |
| " manual" control.
| |
| If Rod Speed = Os no auto or manual rod movement.
| |
| Suggested Instructor Action:
| |
| None.
| |
| Events:
| |
| None 519M/197M/15 5/89
| |
| __ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ ___--___._r__---:
| |
| | |
| BRAIDWOOD SINEATOR MALFUNCTION
| |
| \
| |
| i ym
| |
| | |
| ==Title:==
| |
| Tref Failure (Rod Control) ID: CRF-9 NO: 6.3.4.4.9 Descriptions -Failure of the Tref signal processor TY 505A to a selectable value. j Variations: Date: 1/4/88 Rev 4 Selectable Steps Inputs Comments
| |
| : 1. Select failed value 550' to 630'F
| |
| : 2. Select ramp time O to 99,999 sec.
| |
| Brief Plaht Response: (Based on failure occurring while plant is at 100 per-cent power)
| |
| Failure will cause the rod control system to respond (if in automatic rod control) to control Tavs to the " failed Tref" value. Pressure level and pressure will respond to the changing Tavg, as will steam generator pressure.
| |
| The first an~..anciator received is TAVG CONT DEV HIGE.
| |
| Suggested Instructor Action:
| |
| None.
| |
| Events:
| |
| None O >
| |
| 519M/197M/16 5/E9
| |
| ~ ~ - ~ ~ - - " --
| |
| -_ill_________li ' T 1__ _
| |
| | |
| BRAIDWOOD SIMULATOR MALFUNCTION
| |
| | |
| ==Title:==
| |
| DRPI - Loss of Voltage ID: CRF-10 j.
| |
| NO: 6.3.4.4.10
| |
| {-
| |
| | |
| == Description:==
| |
| Voltage failure of either coil Group A or B.
| |
| Variations: Date 1/5/88 Rev 4 Selectable Steps Inputs Comunents
| |
| : 1. Select failure location 1-3 1 = Group A 2 = Group B 3 = Both Brief Plant Responses O A or B - Loss of power to one coil group will result in an ROD CONT NON-URGENT FAILURB alarm, accompanied by Data A (or B) failure 1, 2, 3 lights, general warning (GW) lights for all rods, with the system indicating at half accuracy (every 12 steps).
| |
| Suggested Instructor Action:
| |
| Clear the malfunction when requested to repair power supplies.
| |
| Events:
| |
| None O 519M/197M/17 5/89
| |
| -< ~ -- - - - -- - - - -
| |
| --_______-_-_-r__
| |
| | |
| BRAIDWOOD SIMULATOR MALFUNCTION Titles DRPI - Open or Shorted Coil ID: CRF-11 NO: 6.3.4.4.11
| |
| | |
| == Description:==
| |
| Open or shorted coil in Data "A" or Data "B", or both.
| |
| Variations: Date 1/5/88 Rev 4 Selectable Steps Inputs Comments
| |
| : 1. Select rod B6, M12, etc. See MALF CRF-4 core map for rod location and group ,
| |
| 2.- Select faulty coil 1-3 1 = Coil in Data "A" location >
| |
| 2 = Coil in Data "B" 3 = Coils in both Data "A" t'~) and Data "B" -
| |
| (/
| |
| m j
| |
| Brief Plant Response:
| |
| A or B - Fault in a coil in either Data "A" or "B" circuit will cause ROD CONT NON-URGENT FAILURE alarm, general warning (GW) light for selected rod, and a Data "A" (or "B") Failure light. Indication for affected rod will go to half securacy (indicates only every 12 steps now).
| |
| Both - Fault in coils of both Data "A" and Data "B" circuits will cause ROD CONT URGENT FAILURE, ROD CONT NON-URGENT FAILUi alarms, and general warning (GW), and rod bottom (RB) lights for the selected rod. " Data A Failure" and " Data B Failure" light will also come on.
| |
| O 519M/197M/18 5/39
| |
| - ~~ ' - ~ ~ - - -- - - ---
| |
| _ : r :L ~ _- L
| |
| | |
| - - -____-_ _7 BRAIDWOOD SIPETLATOR MALFUNCTION!-
| |
| J
| |
| | |
| ==Title:==
| |
| LDRPI - Open or Shorted Coil ID: CRF-11 f:'(
| |
| 8taggested Instructor Action:'
| |
| Kone.
| |
| Events:- None. .
| |
| e 9
| |
| 4 A
| |
| 519M/197M/19 5/89 i
| |
| ~ '~
| |
| ' 2 T_" ~ - _._??- * ---- ?_"???----__l_?.
| |
| | |
| BRAIDWOOD SI.'sULATOR MALFUNCTION cx .
| |
| | |
| ==Title:==
| |
| Stuck Rod ID: CRF-13
| |
| )- NO: 6.3.4.4.13
| |
| ~
| |
| | |
| == Description:==
| |
| Selected rod will not mov~4 due no mechanical failure.
| |
| Variations: Date: 3/15/89 Rev: 6 Selectable Steps Inputs Comments
| |
| : 1. Select malfunction CRF13A CRF13B
| |
| : 2. Select stuck rod H6, M12, etc. See EALF CRF-4 core map for rod locations.
| |
| 7 >
| |
| Brief Plant Responses (Based on rod in control bank "D" being stuck 3 i When control bank D is required to enae, all rods in the bank will move except for the one selected to stick. DRPI will indicate the failure. If the misalignment becomes large, indications of flux shifts can be observed on the nuclear instrumentation. Annunciators received include ROD CONT LO INSERTION LIMIT, ROD CONT LO-2 INSERTION LIMIT and R0D DEV POWER RANGE TILT.
| |
| Suggested Instructor Actions i
| |
| 01 ear the malfunction before attempting to re-align it with its group.
| |
| Events: None 1
| |
| i U
| |
| 519M/197M/21 5/89 i
| |
| ~
| |
| :L_ __~* .___-~"~~__L_ T^
| |
| | |
| n L .
| |
| BRAIDWOOD SIMULATOR MALFUNCTION
| |
| ~
| |
| l i
| |
| | |
| ==Title:==
| |
| Rod Control Syster Yailure ID: CRF-14
| |
| ~
| |
| NO: 6.3.4.4.14 j
| |
| | |
| == Description:==
| |
| The rate comparator fails in the rod control system resulting in erroneous. demands for rod movement in the automatic mode.
| |
| Variations:~ Date: 1/4/88 Rev: 5 Selectable Steps Inputs Comments None.
| |
| Brief Plant Reryonse (IC-17, 100%, all sya. cess in automatic)
| |
| .L/
| |
| Rods move in erratically with rod control in " auto". No annunciators are received innaediately, however numerous annunciators will be received due to the lowering Tase.
| |
| Suggested Instructor Action:
| |
| None.
| |
| Events: Rone.
| |
| O 519M/197M/22 5/89
| |
| . ~ . _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _
| |
| | |
| L ! .
| |
| 1!
| |
| h BRAIDWOOD SIMULATOR MALFUNCTION
| |
| ,(7 ID: ,CEF-15 l()
| |
| | |
| ==Title:==
| |
| Power Cabinet Urgent Failure NO: 6.3.4.4.15
| |
| | |
| == Description:==
| |
| Multiplexing error causes two groups in same power cabinet to attempt to move together.
| |
| Variations: Dates 3/15/89 Rev: 6 Selectable Steps Inputs Co;mnents
| |
| : 1. Select ca. binet 1,-4 1 = 1AC 2 = 2AC 3 = 1BD 4 = 2BD Brief Plant Responses (IC-17, 100%, all systems in automatic)
| |
| Power cabinet failures prevent rod motion for the rods in the affected power cabinet. Rod motion does not occur for any rods in manual or automatic positions of the bank select switch. The first annunciator received is ROD CONT URGENT FAILURE.
| |
| Suggested Instructor Action:
| |
| Clear malfunction when cause is determined.
| |
| Events: 1. DVR 6-01-86-163: Power Cabinet 2AC Urgent Failure
| |
| : 2. DVR 20-01-87-267: Power Cabinet 1AC Urgent Failure
| |
| : 3. DVR 06-01-87-100: Power Cabinet SCD2 Urgent Failure
| |
| : 4. LER 20-01-87-032: Power Cabinet 1AC Urgent Failure O
| |
| 519M/197M/23 5/89
| |
| | |
| CW~I f DEVIATION INVESTIGATION REPORT TITLE pace R00 CONTp0L URCENT at-etMS DUE TO CIRCUf f C ARD F AILURES 0 hlVENT DATE DIR NUMBER GEFORT DATE
| |
| // SEQUENTIAL // REVI5!0N 1 MONfH DAY YEAR STA UNIT YEAR NUMBER -
| |
| NUMBER MONTH DAY YEAR 2 F0WER LEVEL A _cir 816 016 011 816 1 I sil 3 "
| |
| O IO 111 114 81 6 g, gg f CONTACT FOR THIS OfR CAME TELEPHONE NUMBER AREA CODE Don trindle. U2 Ortag,p.tino Eneineer Ext. 2018 8 I1 iE 21114 l -15 Iala11
| |
| , COMPLETE ONE LfME FOR EACH COMPONE LURE DESCRIBED IN THft RFFORT.
| |
| CAUSE SYSTEM (CP1P00ii MANUFAC- REPORTABLE CAUSE SY$ TEM COMPONENT MANUFAC- REPORTABLE TURER TO NPRD5 TURER TO NRfDS X RfD 1X IF 17 .WI11210 Yes l I kl l l !
| |
| I i 1 -l I j l I I 1 I I I I SUPPLEMENTAL REPORT EXPECTED MONTH DAY ' YEAR SUBM15510N I YEs fit ves. comolete EXPECTED suRMISSION DATEt i NO TEXT A. PLANT COND.[ HONE FRIOR TO EVENT:
| |
| ( MODE 2 - startuo _ Rx Power 0 __ RC5 [A8] temperature / Pressure fiermal Oceratino B. 02$CRfRif0N Of EVENT: -
| |
| No systems were incperable at the bgginning of this event which contttbuted to this event. On 10/02/86 at 0155 durtng a reactor start up with control !aank *B' at 115 steps withdrami, alarm window 1-10-C06 "R00 i CONTROL yRGENT FAILURE
| |
| * annunctateh IBCA R00-2 was initiated and Limiting Condition for Operation Action l
| |
| Respense (LCOAR) 18051.3.1-la was ehtared into due to "All rods betng inoperable but trippable". Ctreutt caras in the 2AC power tantnet were replaced by Instrument maintenance and tha alarm was cleared. LCOAR l 18051.3.1-la was exited at 0550 on 10/42/80. Start up was resuwd untti a secono urgent alarm occurred at 0650. A rod drop and manual reactor trip occurred during the second urgent alaria. Additional corrective action was tak-sn and start up was resumed again. For furtfier information see LER 86-028-00.
| |
| C. ftUSE OF EVENT:
| |
| Initially the root cause of the event was thought to be failures in the following cards in the 2AC power cabinet; regulation circuit cards for the stationary, moving and Inf t cotls, and the firing circutt card for the moving coils. It was later discovered that a combination of an intermittently failing Alarm Ctreult Card may ' dave contributed to the urgent alarm. For further information see LER 86-028-00.
| |
| O. idEff ANALYSI1:
| |
| At all times all r?ds remalnad trippable and at no time was the health or safety of the pubite adversely affacted.
| |
| f (09COM)
| |
| | |
| ;E/IAT!:' ITvt3fI0At:;9 RE8:RT C G- ! T T17tt Fai bre For Roos In The 1AC Power *,acinet fa Move L e [J {
| |
| To a Failed Firina Card i 1 E EVENT DATE DIR ND*BER DEpcRf DATE OPERATING '//'
| |
| l' .', <
| |
| SEQUENTIAL / REVISION // , //' ,
| |
| MONTH DAY YEAR STA UNIT YEAR NUMGER f NUMEER MONTH DAY YEAR 2 / /'/ ' h/ / .
| |
| 'o"c" Shmmu l Of 8 Ol 9 Bi 7 21 0 01 1 81 7 ~
| |
| 2 l 61 7 -
| |
| 0 IO 01 8 ll 1 21 7 01013II'bhfIN[ /
| |
| CCNTACT FOR THIS DIR _
| |
| NAME TELEPHONE NUMBER _
| |
| AREA Cr0E Frank W. Trikur. Technical Staff Encineer Ext. 2495 B11 15 4l 5 lR l-l 2 Ia10 l1
| |
| . J 0MPLETE ONE LINE FOR EACM COMPONEN URE D*SLRIBED IN T C REPORT CAUSE SYSTEM -COMPONENT MANUFAC- REPORTABLE CAU5E SY S T'.M COMPONENT MANUFAC- REPORTABLE TUPER . TO NPDD$ TURER TO NP905 a .A IA al al al
| |
| * wi 11 21 0 Y . I i l i i 1 1 I I I I I I I /f '/ I I I I I SUPPLEMENTAL RENRf EXPECTE0 ,
| |
| MONTH DAY VEAP SUBMIS$1CN
| |
| ~""
| |
| ~
| |
| DATE I I l I YE1 fif ves. e g M t EXPECTED SUBMISSION DATE) Xi NO I l !
| |
| TEXT A. PLANT CONDITIONS PRIOR TD EVENT:
| |
| Unit: _An tgweed 1.; Event Date: Auoust 9. 1987 : Event Time: 1830 MODE: _1. - starten ; Rx Power:_2L: RCS (AB] Temperature /Pressurs: 5599 /12235 osia 9 B. DESCRIPTION OF EVENT:
| |
| At 1830 with the plant in Mode 2. an URGENT ALAkM on Ptwer Cabinet I AC was received and LC0AR 1.3.1 1a was entered because control rods did not move upon demand. There were no systems or components inoperable 3t the beginning of the event that contributed to the event. Control Bank C rods were positioned at 228 steps (fully withdrawn) and the reactor was being controlled with Control tank D selected. At 1900 hours, Technical Staff and the Instrument Maintenance Department began troubleshooting. At 2240 the Firing Card for the moveable gripper coils in Power Capinet 1 AC was found to be bad and replaced. The rod control system was tested satisf actortly, per 15was 1.3.1.2-1. HoveaDie Contr01 AssemD11es Monthly Surveillance, and LC0AR 1.3.1-14 was exited. The plant was stab 1g throughout the entire event.
| |
| C. CAusE OF EVENT:
| |
| The root cause of the event was the f ailure of the Firing Card for the Moveable Gripper Colls in the 1 AC Power Cabinet. This prevented the bridge thyristors from being fired properly which produced an urgent alarm due to a phase failure. The control rods in Power Cabinet'1AC locked up and would not move.
| |
| O 1801m(082787)/24
| |
| | |
| '~
| |
| ?'
| |
| * D E /I A T IO*4 IN'/ E S LAT MN 4 EF;R T
| |
| * E A T - QT;U M '*i TITLE g yyg s 's,g_,_ ,
| |
| Fatture For Rods in The 1AC Power Cabinet To Move Due
| |
| #E
| |
| * To A Falled Firing Card fl 0 Ol 1 91 7 ~
| |
| 2 l6 17 -
| |
| 0l0 2 0F 0 fL
| |
| . G rEXT
| |
| ' O. iAFETY aNaLYSiE:
| |
| There were no safety consequences as a re: ult of this event. The Action Statement _in Technical Specification 3/4.1.3.13 was entered and the system was repaired within the specified 6 hour time Itmit. There was no reduction in the margin of safety since control rods remained trippable throughout the event. Under a riore severe set of initial conditions such as being at full power and end of core Itfe with one Rod Cluster Controt Assembly (RCCA) fully inserted, the Departure from Nucleate Botling Ratto (DNBR) would not f all below its limiting value. The ability of the Reactor Coolant to remove heat f rom the fuel rods would not be reduces. and the peak te9perature resulting from the mtsaligned RCCA would be below that welch would cause fuel damage. An event of this type it bounded by FSAR Acetdent Analysts as presented in Chapter 15.4.3.2.a and reautres'ocerstar action per SwoA Rod-2. Failure of Rods to Nove - Unit 1.
| |
| E. CORRECTIVE.ACT10NS:
| |
| The Firing Circuit. Card for the Moveable Grtpper Coils in Power Cabinet IAC has been replaced. System design allows for such f ailures without system degradation and provides detection capabilities for this type of railure.
| |
| F. PREV 10US OCCURRENCES:
| |
| None G. CONPONENT FAILURE DATA:
| |
| MANUFACTURER NOMENCLATUGi MODEL NUMEER MFG PART NUMBER l Westittghouse Electric Firing Circutt Card N/A 6050012G01
| |
| : l. Corporation l
| |
| l l
| |
| i l
| |
| ( l 1801m(082787)/25
| |
| | |
| ciF.
| |
| DEVIATION INVESTIGATION REPORT
| |
| ]
| |
| TITLE- R00 CRIVE URGENT FAILURE ALARM IN 580 POWER CABINET PAGE DUE TO FAILED SOLDZR JOINT ON FIRING CARD 1 IAFl 0 12 EVENT DATE DIR NUMBER PEPORT DATE
| |
| // SEQUENTIAL // REVISION HONTH DAY YEAR STA UNIT YEAR / NUMBER NUMBER t!QETH DAY YEAR 1 POWER 01 87 LEVEL
| |
| . g i 21 a el 7 01 6 Of I si 7 -
| |
| 1 10 l0 -
| |
| 010 Of9 1 1 1 91 a CONTACT FOR THIS DIR NAME TELEPHONE NUMBER AREACODEl T. Hionins. Doeratina Enaineer Ext. 2215 Bl1 l5l2l 3 14 l 1514 14 l1 COMPLETE ONE LINE FOR EACH COMPONEN A URE DESCRIBED IN THIS REPORT CAUSE SYSTEM COMPONENT MANUFAC- REPORTABLE CAUSE SYSTEM COMPONENT MANUFAC - REPORTABLE TURER TO NPROS TURER TO NPRD$
| |
| X A1A El Cl Bl D Wll 12 10 Y l l I I I ( l I i i i l i l i I I I I SUPPLEMENTAL REPORT EXPECTED MONTH day YEAR
| |
| $UBMIS$10N l YES fif vet. comolete EXPECTED SdPN11110N DATE) [I NO TEXT A. PLANT CONDITIONS PRIOR 70 EVENT:
| |
| Event Date/ Time _Z/J8/87 / 0815 h
| |
| Reactor Power RCS (AB] Temperature / Pressure Normal Ooeratina
| |
| ~
| |
| Unit 1 MODE 1 - Rx Power 987, Unit 2 MODE 1 - Reactor Power Rx Power 4 91. .. RCS (AB] Temperature / Pressure Normal Oceratina B. OEstRIPTION OF EVENT:
| |
| At 0815. on 7/28/87.while performing IBOS 1.3.1.2-1 " MOVABLE CONTROL ASSEMBLIES MONTHLY SURVEILLANCE" the
| |
| " ROD CONT. URGENT FAILURE" window annunciated. It was determined that the Rod Drive (AA] System was the source of the alarm (both the Rod Drive and the Digital Rod Position Indication (AA) systems feed this alarm). Technical Staff and Instrument Maintenance departments were . contacted and consnenced troubleshooting the system. A lt,gic error light and multiplex error lights were found lit in the SCOE Rod Drive Power Cabinet. From this it was determined that either the firing card or the regulation card for the Shutdown Bank B stationary gripper circuit was at fault. These cards were replaced and the alarm was cleared at 1335 on 7/28/87. No systems were declared inoperable. No manual or automatic safety system actuations occurred. No operator actions were taken that either increased or decreased the severity of the event.
| |
| C. CAUsE or EVENT:
| |
| The intermediate cause of this event was a f ailed solder joint on a gripper firing card. leabing to an urgent alarm being generated. The root cause of the failure is indeterminate, with multiple possible causes, including vibration, fatigue or manufacturing process.
| |
| u ,
| |
| (1619M/0185M)
| |
| | |
| DEV!ATION INVESTIGATION REPCRT TEXT CCNTINUATICN TITLE c1R asuMsER pact SEQUENTIAL REVISION RDO ORIVE URGENT FAILURE ALARM IN 580 POWER' STA UNIT YEAR MttlSQ i
| |
| _ N MR_
| |
| SINET DUE TO FAILED SDLDER J0!NT ON FIRING CARD l6 d1 07 -
| |
| *Ic) o 3 Io ;b J
| |
| . TEXT oI; D. SAFETY ANA(liig:
| |
| There were no adverse safety consequences of this event. At all times all rods remained trippable. There would have been no adverse safety consequences had thic event occurred under a more severe set of initial conditions. as the rods would still have remained trippable.
| |
| E. CORRECTIVE ACTI0at:
| |
| The inmediate action was to have the Technical Staff and Instrument Maintenance departments perform troubleshooting and replace the suspect circuit cards. Since.this is an isolated incident the only action to prevent recurrence is to observe the cabinet in the future for further f ailures of this type.
| |
| F. PREVIOUS OCCURRENCES:
| |
| There were no previous occurrences of this failure in this power cabinet.
| |
| G. CfDtPONENT FAILURE DATA:
| |
| a) MANUFACTURER NIDEMCLATURE M EL NL3GER MFG PART NL3GER
| |
| . Westinghouse Electric Co. 6050D12G01
| |
| ') Nuclear InstrLauentation and Control Department b) REtuLTS QF kPRDS 1EARCH!
| |
| The NPROS Search yielded:
| |
| 1 Regulator lift firing caro failure 2 Card failures due to faulty solder , joints in the Rod Drive Power Cabinets l
| |
| l l
| |
| (1619M/0185M)
| |
| | |
| ;,*CE 5 E Eh ahn g[ c -y - [
| |
| Dock et %mce- 2 f .fnt .
| |
| Facilit/ Name (1) 31 51 31 01 1' al l' 6 L;d-Bra'cwood. Unit i
| |
| ' wamual Reacter Trio Due To Leese Conrecticas iQ Ead Fuses In 4 #cd Control Discweet in'icn y;. g*
| |
| Desert Date f71 Other racilit$es N e1<ed :41
| |
| (
| |
| Event Sate (5) LER Number (6)
| |
| Year racility N aw s Oceket utr;i l Year Year Sequential /f/, Revision Month Day k Month Day g// f// Number
| |
| / Number NONE 01 51 01 01 O! I f
| |
| ~
| |
| 013 l2 010 0 17 21 2 al 7 Ci El dl dl of I L_
| |
| 01 7 01 1 al 7 al 7 THIS REPORT !$ SUBMITTED PURSUANT TO THE REQUIREMENTS OF 10CFR p
| |
| fCheck one e' more of the followino) fili 20.405(c) L. 50.73(a)(2)(iv) 73.71(b) 2 20.40?(b) __.
| |
| 50.36(c)(1) 50.73(a)(2)(v) 73.71(c)
| |
| POWER _ 20.405(a)(1)(1) _ __._
| |
| Other (specify LEVEL 20.4C5(a)(1)(tt) 50.36(c)(2) _ 50.73(a)(2)(vit) 0l1 50.73(a)(2)(vii11(A) in Abstract (101 0 . 20.405(a)(1)(tit) L 50.73(a)(2)(1) 50.73(a)(2)(it) 50.73(a)(2)(vitt)(B) below and in
| |
| ////////////////////////// 20.405(a)(1)(iv) _
| |
| 50.73(a)(2)(111) 50.73(4)(2)(x: Text)
| |
| ////////////////////////// 20.405(a)(1)(v) _ _
| |
| LICENSEE CONTACT FOR THIS LER (121 TELEPHONE NUMBER Name AREA CODE fxt. 24M ai1 15 el 51 51 -l 21 al 1 Frank W. Trikur. Tech Staf f Enoineer COMPLETE ONE LINE FOR EACH CDM Faltuh(RRIBE0 IN THIS REPORT f til '
| |
| CAUSE SYSTEM C OMPON(NT MANUFAC- REPORTABLE CAUSE SYSTEM COMP 0NEMT MANUFAC- REPORTABLE TURER TO NPED5 TURER TO NPRD$
| |
| xl 91 91 9 I .I l l I l l a ait FI ul al a N I i i i 1 1 1 I i 1 i I L.
| |
| Espected Month I Day l vgg WPPLEMENTAL REPORT EXPECTED (141 Submtssion lyes (If yes. comelete EXPECTED $UBMI5510N DafEi X l NO I L I ABSTRACT (Limit to 1400 soaces, i.e. approximately fif teen single-space typewritten lines) (16)
| |
| Tecnnical On A.re 29.1987 at 0710 hours a Rod Control Ur2ent Fatkre Alarm was receive ( in the Control Poom.
| |
| Staff found a phase failure in the Rod Control System Power Cabinet 1AC. A movable gripper Coll fuse was l
| |
| rep 14ced and rods were moved af ter the alarm was reset. At 1352 hours another Urgent rat 1ure Alarm was received. The same f ailure was indicated. Technical Specification Action Requirement 3/4.1.3. was entered.
| |
| All fuses were checked and were found to be good. The firing card and phase f ailure card -ere replaced. Rods again moved after.the alarm was reset. Rods were restored to an operable status af ter being tested. On June
| |
| : 30. 1987 while taking: traces on the 1 AC Power Cabinet, the phase f ailure Rod Control Urgent Alarm annunciated.
| |
| Rod Control was declared inoperable at 1850 and the reactor was tripped c,n July 1,1987 at 3022 hours in accordance with Tech Spec 3/4.1.3.1. j The causes of the event were loose connections and bad fuses in the 1AC movable disconnect switch cabinet located above the 1AC power cabinet.
| |
| The 30 Ampere fuses were replaced and all loose connections tightened. All other Pcwer Cabinet Ducennect Switch Cabinets were inspected and loose connections tightered.
| |
| There have been no previous occurrences of this problem.
| |
| l l
| |
| 1744m(072387)/1249A/33 w___--_______-_______. _ _ _ _ _ _ _ . - _ _ _ _ _ _ _ _ _ . _ _ _ . _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
| |
| | |
| t
| |
| * ttNs tt t,t's' #F8 9 t ftta! '!<*C M L:Q _ _ . _ _
| |
| FaC!LITY cAME:(1) 00CKET V 8E8 (2)- _U v aEt Al i 8 va. f
| |
| ' Ye4* l 5et.emt141 qf eeva s on
| |
| .$q// kumqtv -ir/ Nteter j j Braid = cod. tJnit 1 0l 5 l34 0'1 0 I'al El 6 91' - 3l1' 2 - 0 l3 Of N Cut
| |
| [ TEXT Energy Industry Identification System (E!!$) codes are identifted 'n the test as (sx) ,
| |
| v A. PL ANT CONDITIONS PRIOR 70 EVENT:
| |
| July 1. 1987; Event Time: .0022..
| |
| Unit: .ir a i t' woo d 1 : Event Date:
| |
| - Moct _L, - itartuo Rs Power 1 RC5 (AB) Temperature / Pressure 557+e/2230 stio-
| |
| : 8. Of ttRIPTION OF EV@
| |
| The Rod Control System was inoperable at the beginning of the event.
| |
| At 0710 on June 29. 1987. 4 Rod Control Urgent Alarm was received. Troubleshooting by the Technical Staff revealed a phase f ailure for the movable gripper coilt on Card 12 in the Red Control System Power Cantret 1AC. ~
| |
| in addition. Fuse Fbt at cabinet location A67 indicated blown. The fuse was replaced. the alarm was reset. and control rods were made operable and remained trippable.
| |
| At 1352 while maintaining 17. power in manual controle a Rod Control Urgent Alarm was again received. The same indications were found and Technical Specification Action Statement 3/4.1.3.1 was entered. The 1AC power cabtnet was checked and the movable gripper disconnect switch fuses, esternal to the power cabinet, were found to be good. The Firing Card and Phase Control Card were replaced per the vendor manual. Rods again were moved.
| |
| tested. ar}d declared operable.
| |
| On June 30.1987 at 1515 with rods in manual Gntrol, a Rod Control Urgent Alarm again annunciated during rod J_ ' motion. At 1850 the Rod Control System was declared inoperable due to the inability to initiate rod m'otion. At-approximately 0022 on July 1, 1987 the reactor was manually tripped per Technical Specification 3/ U .3.1.
| |
| f'\
| |
| The plant conditions remained stable throughout the entire event. Operator action taken neither increasec ner decreased the severity of the event.
| |
| This evert is being eported under 10CFR50.73(a)(2)(1)(a). the completion of any plant shutdown required by the plants Technical ,,ccifications, and 10C7R 50.73(a)(2)(iv). any event or condition that resulted in marval er automatic actuation of any Engineered Safety Feature (ESF). including the Reactor Protection System (RPS).
| |
| C. Cautt or EVENT: .
| |
| The root cause of the Rod Control Urgent Alarm was a combination of loose connections and blown fuses 'n the lac power cabinets' movable gripper disconnect switch cabinet.
| |
| A 17aam(072387)/1249A/34
| |
| | |
| L MENsE!"E, W sf,,8:a* *_!a 'p m':% a '_;;w _ _ , _ _ _ _
| |
| f 00CK t * %**BIE (2) lee bwPE4 s 84-M _
| |
| FAC!.!TY e >E (t)
| |
| Year ' /// iecuta!'al /.i #te's'en,
| |
| %n": e
| |
| * Mumter ,
| |
| 0 lt!OI0 I3 1 at $l 6 8 f7 - Of1 7 - 310 01 35 CF !! j!
| |
| Braidwood. Unit 1 f
| |
| } T[XT [nergy Industry Identificat1on System (E!!5) coces are identified in the tent as (sv} i-
| |
| 'v. D. 1AFETY ANALV111:
| |
| There were no safety consequences as a resuit of this event. The limiting Coedition f or Operation (LCD) in Technical Specification 3/4.1.3.1 was adhered to and the plant was p1 Aced in *ct Standby within 6 nours f rom the time rods were declared inoperable. There was no reduction in the margin of safety as defined of Technical Specification 3/a.1.3.1. Since adequate shutdown margin was maintained and rods remained trippable at all times. Under a more severe set of initial conditions. such as 1 Rod Cluster Control Assemoly (RCCA) fully inserted at f ull power, the Departure f rom Nucleate Boiling Ratio (DNf6R) would not f all below its limiting value. The ability of the reactor toolant to remove heat from the fuel rods would not be reduced, and the peak An fuel temperature resulting f rom the nsaligned RCCA would be well below that which would cause fuel damage.
| |
| event of this type.is bounded by F5AR Accident Analysis as presented in chapter 15.a.3.2.a and requires operator action per BwCA Rod-2 Failure of Rods to Move - Unit 1.
| |
| E. CoARECTivE ACTIONS:
| |
| The fuses f or the movable gripper fused disconnect switches for power cabinet 1 AC were replaced. All connections were tightened. In addition, the rema0nteg rod drive power cabinet otsconnect switch bones were inspected and loose wires tightened. No aoditional corrective action is warranted at this time.
| |
| F. PREVIOUS OCCURRENCES:
| |
| DVR/tER NUMBER TITLE O -'
| |
| G. COMPONENT FAILURE CATA:
| |
| NOMENCLATURE N0LEL NUPBER HFG PART NUMBEk MANUFACTURER Bussmann Mfg.' Division 600 Volts KBC 30 N/A 30 Amperes Rectifier Fuse l
| |
| l I
| |
| l l
| |
| [
| |
| (
| |
| 17aam(0723EF)/1149A/35
| |
| | |
| c-I BRAIDWOOD SIMULATOR MALFUNCTION l'.
| |
| | |
| ==Title:==
| |
| Logic Cabinet Urgent or Ison-Urgent Failure ID: CRF-16 lf'
| |
| -- Nos 6.3.4.4.16
| |
| | |
| == Description:==
| |
| Logic cabinet failure causing urgent or non-urgent failure.
| |
| Variations: See CEF-19 Date: 3/15/89 Rev: 5 Selectable
| |
| $teps Inputs Comments f
| |
| : 1. Select failure 1, 2 1 = Urgent failure 2 = Non-urgent failure Brief Plant Responses (IC-17, 100%, all systems in automatic)
| |
| Urgent failure in logic cabinet causes all rod motion in manual automatic and bank select to stop. The first annunciator received is ROD CONT URGENT FAILURE.
| |
| If failure is non-urgent, rod motion is not affected. The first annunciator received is R0D CONT NON-URGENT FAILUE2.
| |
| Suggested Instructor Action:
| |
| clear malfunction when cause is determined by students.
| |
| Events:
| |
| : 1) DVR 06-02-88-056: Logic Cabinet Non-Urgent Alarm.
| |
| 519M/197M/24 5/89
| |
| , i n . . . - . . . . . . . . .
| |
| m._-1_____.___.___._m____________________.m _ . _ _ .
| |
| | |
| c4F-/G DEVIATION INVESTIGATION REPORT TITLE p,g NON-URGENT ALARM DJE TO OVERVOLTAGE PROTECTION FAILURE 0 2
| |
| ~
| |
| C. EVENT DATE __ DIR NUMBER REPORT DATE OPERATING
| |
| // SEQUENTIAL // REVISION M) NTH DAY YEAR STA UNIT YEAR
| |
| /f
| |
| // NUMER ff NUN ER MONTH DAY YEAR 1 POWER 0 15 2 10 8 18 0 16' 0 12 8 18 -
| |
| 0 15 l 6 -
| |
| 010 O6 g 2g4 88i LEVEL 01 91 4 l
| |
| l~ CONTACT FOR THIS DIR NAME TELEPHONE NUM ER AREA CODE Don Brindle. Doeratina Enoineer Ext. 2218 8l115 21314l-l51414l 1 COMPLETE ONE LINE FOR EACH COMPONEN A URE DESCRIBED IN THIS REPORT CAUSE SYSTEM COMPONENT MANUFAC- REPORTABLE CAUSE SYSTEM COMPONENT MANUFAC. REPORTABLE TURER TO NPRDS TURER TO NPROS X AlA l 15 l 9 L 10 14 15 Y l l 1 l l l l l l l l l l l l l l l 1 SUPPLEMENTAL REPORT EXPECTED MONTH DAY YEAR p
| |
| SUBMISSION DATE I I l X l YES M f ven. comolate EXPECTED SUBMISSION DATE1 --l NO 01 9 01 9 81 8 TEG A. PLANT COM)ITIONS PRIOR 70 EVENT:
| |
| Event Date/ Time 5/20/88 / 1755
| |
| .[
| |
| Unit 1 MODE 1 - Power Operation Rx Power 97.5% RCS [AB) Temperature / Pressure Normal Oeeratiny Unit 2 MDDE 1 - Power Goeration Rn Power 94% RCS [AB) Temperature / Pressure Normal Oeeratina B. DESCRIPTION OF EVENT:
| |
| At 1755 hours on 5/20/88, a " ROD CONT NON-URGENT FAILURE" alarm was received at the Unit 2 Main Control Board. Investigation revealed a 15 VDC power supply (PSS) in the Rod Drive Logic Cabinet had failed. When the overvoltage protection devices on power supplies of this type trip, it causes the power supply te heat up. The power supply's fuses were removed to allow the power supply to cool down. At 1930 the fuses were re-installed, which reset the overvoltage protection, and the power supply verified to have normal output, and the alarm was cleared. This non-urgent failure also occurred on 5/21/88 at 1641 hours, 5/25/88 at 2113 hours. 5/28/88 at 1922 hsurs, and 5/30/88 at 1838. with similiar sequences of events. Previous to this DVR, two related occurrences of power supply failures were documented: on 4/4/88 (N6-0-88-051), and on 5/1/88 (N6-0-88-082).
| |
| C. CAUSE OF EVENT:
| |
| The cause of this event was spurious actuation of the overvoltage protection circuitry on the output of the power supply. Due to these sporadic occurrences and due to the potential for causing more Rod Drive cabinet malfunctions during replacement of this supply, removal and installation of a new power supply was delayed untti an outage. Several more spurious alarms occurred during this period. The power supply was replaced on 6/2/88. The power supply will be bench tested to determine the root cause of the failure.
| |
| When the root cause of the failure is determined, a supplemental report will be issued.
| |
| 9 I I (0032R/0004R)
| |
| | |
| L.- .p DEVIATION INVESTIGATION REPORT TEXT CONTINUATION TITLE DIR NumER PAGE f
| |
| y'j STA UNIT YEAR SEQUENTIAL.
| |
| NUteER REVISION NUteER j
| |
| l
| |
| 'u J NDN. URGENT ALARN DUE TO OVERVOLTAGE PROTECTION FAILURE 01 6 01 2 8 18 -
| |
| 0l5l6 -
| |
| 0 I0 2 0F 012 TEXT D. SAFETY ANALYSIS:
| |
| There were no safety consequences as a result of this event. There was no effect on the health and safety of the public. The control rods remained capable of being tripped and remained capable of full movement at all times during this event. If this event had occurred under a more severe set of circumstan'ces there would have been no safety consepences as the rods would still be capable of being moved or trippad.
| |
| E. CORRECTIVE ACTIONS:
| |
| The insiodiate corrective action was tc pull the fuses on the tripped power supply to allow it to cool down and to reset the overvoltage protection circuitry. The long term corrective action was to replace the failing overvoltage protection circuitry and power supply pSS itself.
| |
| F. PREVIOUS OCCURRENCES:
| |
| There have been no prev' 4 4 occurrences of failures of overvoltage protection on the tambda power supplies.
| |
| DVR NUPRER IllLE NONE
| |
| (
| |
| l G. CONPONENT FAILURE DATA:
| |
| N a) MANUFACTURER N0tENCLATURE MODEL NUPSER tfG PART NUteER Lambda Electronics Overvoltage LM-0V-2 Corporation h) RESULTS OF NPRDS SEARCH!
| |
| There have been no reported failures of overvoltage protection on Lambda power supplies.
| |
| c) RESULTS OF NWR SEARCH!
| |
| Not Applicable i a (0032R/0004R)
| |
| | |
| BRAIDWOOD SIMULATOR MALFUNCTION l
| |
| l
| |
| (~'3
| |
| | |
| ==Title:==
| |
| Rod-Bank Misalignment ID: CRF-17 e a
| |
| , 'v/ NO: 6.3.4.4.17
| |
| | |
| == Description:==
| |
| Failure of first pulse control circuit to maintain groups in a bank within one step of one another.
| |
| Variations: Date: 3/15/88 Rev 4 Selectable Steps Inputs Comments
| |
| : 1. Select rod bank SA1, SA2, SB1, 1st Letter = Shutdown (s) or SB2, CA1, CA2, control (c) bank.
| |
| CB1, CB2, CC1, 2nd Letter = bank number.=
| |
| f CC2, CD1, CD2 group number f%
| |
| i )
| |
| v Brief Plant Responses (Based on fault occurring to a bank at power)
| |
| Yailure of first pulse control circuit will cause a gradual misalignment of the rods in a group. No transient would be expected. ROD DEV POWER RNG TILT annunciator will occur when deviation reaches i 12 steps.
| |
| Suggested Instructor Action:
| |
| None.
| |
| Events: None 519M/197M/25 5/89
| |
| . - . ~ . . . . . . . . - .. . . .
| |
| -_m ______m._. - - _ . . _ _ _
| |
| | |
| BRAIDWOOD SIMULATOR MALFUNCTION 1
| |
| l l
| |
| ')
| |
| | |
| ==Title:==
| |
| Rod Stops Fail ID: .CRF-18
| |
| \~-1 NO: 6.3.4.4.18
| |
| | |
| == Description:==
| |
| Appropriate rod stop fails to stop outward rod motion.
| |
| Variations: Date: 3/15/89 Rev: 5 Selectable I Steps Inputs Comments
| |
| : 1. Select rod stop 1-5, 11 1 = Intermediate Range Red Stop C-1 2 = Power Range Rod Stop C-2 3 = OTAT Rod Stop C-3 4 = OPAT Rod Stop C-4 5 = Low Power Rod Stop C-5 11 = Bank D Rod Withdrawal Stop C-11
| |
| ('~h V (5 and 11 for auto mode only)
| |
| Brief Plant Response: (IC-17, 100%, all systems in automatic)
| |
| Failure of C1, 2, 3, 4 to stop rods is based on up power transient with rods in auto. Failure to initiate a rod stop will result in a reactor trip on OTAT, OPAT, IR HI FLUX or PR HI FLUX.
| |
| Failure of C5 or C11 is applicable only in automatic mode. C-11 failure would allow bank D to step out to top of core and cause mismatch between bank demand and rod posicion indication causing errors in bank overlap and insertion limits. The first annunciator received is ROD DEV POWER RNG TILT.
| |
| r~
| |
| l x-_
| |
| 519M/197M/26 5/89
| |
| | |
| {
| |
| l t
| |
| l BRAIDWOOD SIMULATOR MALFUNCTION
| |
| ,n.
| |
| ( ) . Title Rod Stops Fail ID: CRF-18
| |
| \__/
| |
| l Brief Plant Response (continued):
| |
| C-5 failure would result in automatic reactor control at less than 15 percent power where the NSSS instruments are less accurate and plant is not stable enough for accurate control. Could result in erratic and inaccurate rod motion and the resulting transients could cause a reactor trip.
| |
| Suggested Instructor Action:
| |
| None.
| |
| Events:
| |
| : 1) e D.VR 06-02-87-072: Failure of C-11 O 519M/197M/27 5/89
| |
| ****ee *- + , we e 4 *w a es,e.,- og -
| |
| | |
| DEVIATION INVESTIGATION REFCRT TITLE pace.
| |
| FAILURLOF C-11 AUTO RCD STOP 1 10Fl 0 l 2
| |
| 'ENT DATE DIR NUMBER I REPORr DATE '
| |
| { ) // SEQUENTIAL // REVISIGN pj
| |
| ^ I" n0 NTH DAY YEAR STA UNIT YEAR // NUMBER // NUMBER MONTH ' DAY YEAR 1 POWER of 7 11 a al 7 of 6 of 2 al 7 -
| |
| oID2 -
| |
| o10 018 24 1 871 LEVEL 01 91 a CONTACT FOR THIS DIR NAME TELEPHONE NUMsER AREA CODE W. Kouba. Asst. Tech Staff Sunervisor Ext. 2274 al 1 l5 2 13 l4l-l 5 l4 14l 1 COMPLETE ONE LINE FOR EACH COMPONE T URE DESCRIBED IN THIS REPORT CAUSE SYSTEM COMPONENT MANUFAC- REPORTABLE CAUSE SYSTEM COMPONENT MANUFAC- REPORTABLE TURER TO NPRDS TURER TO NPRDS X AlA Al Ml Pl Wl11210 Y l l l l l l l l 1 I I I I I I I I l l SUPPLEMENTAL REPORT EXPECTED MONTH DAY YEAR SUBMISSION l YES fif ves. comolete EXPECTED $URHISSION DATE1 Xl N3 DATE l l l TEXT A. PLANT CONDITIONS PRIOR 70 EVENT:
| |
| Event Date/ Time 7/ta/s7 / 1as4 Unit 1 MODE - NA Rx Power NA RCS (AB] Temperature / Pressure NA Unit 2 MODE 1 - Power Operation Rx Power 981 RCS [AB) Temperature / Pressure SB4*F / 2200 osio B. DESCRIPTION OF EVENT:
| |
| On 7/18/87 at 1834 hours. Startup Test 2.$2.87. 10% Load Swing was in progress. Turbine generator power had been increased frten 881 to 98% at a rate of 2001/ minute per the test procedure. As a result of this power ramp Tave decreased and Tref increased. The rod control system was in automatic and attempted to restore Tave to match Trer by auto withdrawal of Control Bank D. Control Bank D was at 180 steps prior to the load increase and withdrew to an indicated 230 steps when the Unit 2 N50 placed rod control in manual.
| |
| stopping Control East D motion. The C-11 interlock was suppose to block automatic rod withdrawal at 223 steps on Control Bank D but f ailed to do so. Subsequently. Control Bank 0 was returned to its normal operating range and the P/A converter, group demand step counters, bank overlap unit, and process computer rod superv$sion program indications were corrected to proper rod position within one hour. No safety system actuations resulted from this event, nor were any supposed to.
| |
| C. tAuff O' EVEMI:
| |
| During troufaleshooting under Nuclear Work Request 847(41 Zener Diode Z11-1 on sununing amplifier 2ZY-442A was found to be failed. The diode was distorted ir, shape and was scorched. This diode serves to limit output voltage frne 191s amplifier to +10 VDC. In its damaged state it was limiting output voltage to +
| |
| 7.93 VOC. Yhts had the effect of clipping the Control Bank D position signal which feeds C-11. at 182 steps. Any actual find position exceeding 182 steps would be input as 182 steps to the C-11 voltage corgarator 2AB-442C. which has an interlock setpoint of 223 steps (+9.697 VDC). Therefore, the C-11 O
| |
| r l
| |
| setpoint was1d not he reached at any Control Bank D position. The cause of the diode failure is unknown and is attributed to a weak component.
| |
| (160)M/01894)
| |
| _I
| |
| | |
| CEVIATION INVESTIGATION REPORT TEXT CONTINUATION I T!TLE DIP NUMBER PAGE SEQUENTIAL REVISION STA UNIT YEAR NUMBER NUMBER
| |
| .altuRE OF c-11 AUTO ROD STOP 01 6 of 2 Bl 7 -
| |
| 0 l712 -
| |
| 0 l O' 2 0F 0 l2 9XT D. SAFETY ANALYSIS:
| |
| There were no adverse safety consee.uences as a result of this event. The reactor trip breakers were operable at all times, and all control rods were available to be tripped in if needed. Both automatic and manual control rod insertion capability was available at all times to control reactor power if needed. The f i
| |
| reactivity worth of Control Bank D from the 223 step C-11 auto rod stop position to the fully wit Mrawn position is less than 1 pcm (1 x 10-Sk/K) which had a negligible effect on reactor power resfonse, The 2ZY -332A amplifier also feeds the Control bank D Low and Low-2 Rod Insertion Limit (RIC) ..sim.. th that the amplifier was found to work properly up to +7.93 VDC (182 steps) there was no adverse ef fect on RIL alarms. The highest RIL alarm occurs at 1007, power at 171 steps on Control Bank 0, which is below the '
| |
| 182 steps voltage clipping point. The resetting of the bank overlap unit counter ensured that proper overlap and other banks RILs would be met. Because rod travel was stopped at 230 steps the overlap would be off by no more than two steps prior to the reset. The difference it physical rod pesition at 230 steps (as much as 231 steps is mechanically available) versus the 228 steps used for rod drop timing has a '
| |
| negligible impact because all Control Bank D rods have at least 0.99 sec. margin available with respect to the 2.40 see. drop time test requirement.
| |
| E. CORRECTIW. ACTI0i'S:
| |
| * Under Nuclear Work Request 847441 amplifter board 2ZY-442A was replaced. The new amplifier was calibrated per specification and the full range of proper output verified. The C-11 interlock at 223 steps on Control Bank D was also verified to properly function.
| |
| , . PREVIOUS OctuRRENCE$:
| |
| LER NUMBER HILL NONE G. COMPONENLEM11LBE D.81&:
| |
| a) MANUFACTURER NOMENCLATURE MODEL NUMBER MFG PART NUMBER Westinghouse NSA Summing Amplifier 2837A14G01 Z11-1 (5tyle 743A403H11 Card (7300 Series) 10V Zener Diode) b) RESULTS OF NPRDt SEARCH!
| |
| No Zener Diode failures were found.
| |
| w- .I (1603M/0188M)
| |
| | |
| a-BRAIDWOOD SIMULATOR MALFUNCTION Title Urgent Failure In Logic Section of Shutdown ID: CRF-19
| |
| 'v' Banks C, D and E NO: 6.3.4.4.19
| |
| | |
| == Description:==
| |
| Urgent alarm energizes on SCDE slave cycler failure.
| |
| Variations: See CRF-16. Date 3/15/89 Rev: 5 Selectable Steps Inputs Comments None.
| |
| Brief Plant Response: (IC ',, prior to Rx S/U)
| |
| -~
| |
| The R0D CONT URGENT FAILURE annunciator energizes in the control room but due to being in logic section of SCDE does not affect the other shutdown banks A and B or control banks A, B, C and D.
| |
| Suggested Instructor Action:
| |
| None.
| |
| Events: None 1
| |
| J 519M/197M/28 5/89
| |
| ________...___.__._.__m_
| |
| _ _ - . _ _ _ _ _ . _ _ . I
| |
| | |
| BRAIDWOOD SIMULATOR MALFUNCTION
| |
| /''N
| |
| | |
| ==Title:==
| |
| Reactor Trip Failure ~ ID: CRF-20 NO: 6.3 4.4.20
| |
| | |
| == Description:==
| |
| Failure of reactor trip breakers to open, thus preventing a reactor trip.
| |
| Variations: Date: 3/15/89 Rev: 4 Selectable Steps Inpt''.s Comunents -
| |
| : 1. Select breaker 1-3 1 = "A" Breaker (RTA) 2 = "B" Breaker (RTB) 3 = Both
| |
| : 2. Select inode 1-3 1 (auto) = Rx trip breakers do not ,
| |
| 'I open on a AUTO trip signal but can be opened by a
| |
| '- MANUAL Ex trip.
| |
| 2 (manual) = Rx trip breakers do not open on a MANUS trip signal but can be opened by an AUTO signal.
| |
| 3 (both) = Trip breakers do not open under any condition.
| |
| i Brief Plant Response: (IC-17, 100%, all systems in automatic)
| |
| Transient that caused trip to be initiated in protective system will continue. Plant should be tripped manually immediately. If breakers still do j not trip, turbine should be tripped and auxiliary feed pumps started, f l
| |
| Possibility of pressurizer relief valves lifting, high pressurizer level, high l
| |
| RCS pressure, rapid RCS cooldown depending on initiating action or transient. J l
| |
| LO l-519M/197M/20 5/89 j
| |
| 1 --
| |
| l .
| |
| .m .. , c.-..
| |
| | |
| BRAIDWOOD SIMULATOR MALFUNCTION.
| |
| I '
| |
| I l
| |
| i .- .
| |
| IMT
| |
| | |
| ==Title:==
| |
| Reactor Trip Failure ID: CRF-20
| |
| - Brief Plant Response'(continued):
| |
| I i
| |
| t
| |
| 'd
| |
| ' If only one breaker fails to open, other breaker will cause Rx trip to occur.
| |
| Suggested Instructor Action:
| |
| None.
| |
| Events:
| |
| : 1. OPEK 85-52 (RTB failed to auto open due to failed transistor in SSPS).
| |
| O h.
| |
| 519M/197M/30 5/89
| |
| ----_m..__________,.__m_._ _ _ _ ___m_________ _ , _ _ _ _ __ . _ _ _ . _
| |
| | |
| we e, m - _
| |
| }
| |
| (stg to-gy l NUCLEAR NETUORK INFORMATION
| |
| , q ,
| |
| D IS 491 GILLISPIE (INPO) 01-MAR-8513:53 PT 10-85 i
| |
| | |
| ==Subject:==
| |
| INPO SIGNIFICANT EVENT REPORT (SER): j
| |
| | |
| ==SUBJECT:==
| |
| REACTOR TRIP BREAKER FAILURE CAUSED BI IMPROPER TEST EQUIPMENT SETUP
| |
| , UNIT (TYPE): SEQUOYAH 2.(PWR) l DOC N0/LER NO: 50-328/85002 EVENT DATE: 1-12-85 NSSS/AE: WESTINGHOUSE /TVA 1
| |
| | |
| ==SUMMARY==
| |
| DURING A PLANT TRIP, ONE REACTOR TRIP BREAKER FAILED TO OPEN AUTOMATICALLY BECAUSE OF A FAILED TRANSISTOR IN THE SOLID STATE PROTECTION SYSTEM. THE TRANSISTOR FAILURE WAS CAUSED BY AN ERROR IN SETUP OF TEST EQUIPMENT DURING SURVEILLANCE TESTING.
| |
| DESCRIPTION:
| |
| SEQUOYAH 2 WAS OPERATING AT 96-PERCENT POWER WHEN A FEEDWATER PUMP TRIP AND SUBSEQUENT TURBINE RUNBACK CAUSED A LOW-LOW STEAM GENERATOR LEVEL CONDITION. THE REACTOR TRIPPED AS DESIGNED, BUT THE CONTROL ROOM OPERATOR OBSERVED THAT THE "A" REACTOR TRIP BREAKER FAILED
| |
| ' vf) OPIN AUTOMATICALLY. HE IMMEDIATELY OPENED THE BREAKER M *, NU A L L Y FR_
| |
| TH: CONTROL ROOM IN ACCORDANCE WITH PROCEDURES.
| |
| AN INVESTIGATION OF THE SOLID STATE PROTECTION SYSTEM REVEALED THAT THE "A" REACTOR TRIP BREAKER UNDERVOLTAGE (UV) COIL REMAINED ENERGIZED DURING THE EVENT BECAUSE OF A FAILED OUTPUT TRANSISTOR WHICH CONTINUED TO MAINTAIN VOLTAGE TO THE COIL. THE CIRCUIT BOARD WAS REPLACED, AND BOTH THE CIRCUITRY AND THE REACTOR TRIP BREAKER WERE SUBSEQUENTLY TESTED SEVERAL TIMES WITH SATISFACTORY RESULTS.
| |
| A SIMILAR CIRCUIT BOARD FAILURE HAD OCCURRED PREVIOUSLY AT UNIT 2, BUT '
| |
| THE CAUSES OF THE FAILURE HAD NOT BEEN IDENTIFIED. AFTER 1HE RECENT EVENT, IT WAS DETERMINED THAT ERRORS DURING TESTING ACTIVITIES COULD SUBJECT THE TRANSISTORS TO HIGH CURRENTS AND POTENTIAL DAMAGE.
| |
| IN 1983, 1 HE SURVEILLANCE TEST METHOD WAS REVISED TO SEPARATELY ACTUATE THE UV AND SHUNT TRIP CIRCUITS. THE TEST REQUIRES INSTALLATION OF A JUMPER IN THE MANUAL TRIP CIRCUIT AND USE OF A VOLT-AMMETER TO VERIFY THAT THE UV COIL REMAINS ENERGIZED WHEN THE SHUNT TRIP DEVICE IS ACTUATED.
| |
| FOLLOWING A REACTOR TRIP IN LATE DECEMBER 1984, A VOLT-AMMETER WAS!
| |
| CONNECTED ACROSS THE "A" ,UV COIL IN ACCORDANCE WITH THE TEST PROCEDURE, BUT THE METER WAS INADVERTENTLY SET TO MEASURE CURREN*
| |
| [ RATHER THAN VOLTAGE. THE AMMETER FUNCTION CREATEO A LOh RESISTANL PATH (SHORT CIRCUIT) AROUND THE UNDERVOLTAGE COIL AND ALLOWED ABNJRMALLY HIGH CURRENT TO PASS THROUGH THE TRANSISTOR CAUSING IT TO FAIL.
| |
| _ __ _ __ _ 1
| |
| | |
| i' p.
| |
| i .
| |
| SINCE THE UV TRIP- PORTION OF THE TEST HAD BEEN SUCCESSFULLY COMPLETC!
| |
| PREVIDUSLY, THE TRANSISTOR FAILURE WAS NOT' DETECTED, AND THE PLANT WA!
| |
| L RETURNED TD SERVICE ' UNTIL THE JANUARY 12 EVENT. THE FAILURE W OULf l
| |
| HAVE BEEN DETECTED BY THE NEXT SURVEILLANCE TEST SCHEDULED FOR JANUAR' 18.
| |
| COMMENTS:
| |
| : 1. THIS EVENT IS SIGNIFICANT BECAUSE UNDETECTED FAILURE UF T H{
| |
| ACTUATING . SIGNAL TO ONE OF THE TWO REACTOR TRIP BREAKER SUBSTANTIALLY REDUCED RELIABILITY O F- THE REACTOR PROTECTIO (
| |
| SYSTEH. IT IS CONCEIVABLE THAT ERRORS COULD DISABLE BOTH REACTO!
| |
| -TRIP BREAKERS. IT SHOULD BE NOTED THAT THE SHUNT TRIP FUNCTION I:
| |
| UNAFFECTED BY UNDERVOLTAGE CARD FAILURES, AND THE CONTROL R001
| |
| ' OPERATOR COULD OPEN THE BREAKERS IF REQUIRED (THE CIRCUITRY F0tA' AUT0HATIC SHUNT TRIP ACTUATION HAS NOT YET BEEN INSTALLED SEQUOYAH).
| |
| : 2. THE NORMAL TESTS OF THE SOLID STATE PROTECTION SYSTEM WOULD DETCC' TRANSISTOR FAILURES OF THIS TYPE AND WOULD LIMIT THE PERIOD O!
| |
| POTENTIAL BREAKER INOPERABILITY TO A MAXIMUM OF THIRTY DAYS.
| |
| HAS EXPERIENCED SIMILIAR TRANSISTOR FAILURES 6
| |
| : 3. AN3THER PLANT O THREE OCCASIONS (ALL WERE. DETECTED BY TESTING BEFORE THE UNIT w
| |
| \J STARTED UP). EACH OF THESE FAILURES APPEARS TO'HAVE RESULTED FR3t 6REAKE6 MAINTENANCE OR MODIFICATION ACTIVITIES ON REACTOR TRIP CIRCUITRY RATHER .THAN TESTING ACTIVITIES.
| |
| : 4. IT WOULD BE ADVISABLE TO TEST THE SOLID STATE PROTECTION FOLLOWING SYSTEC ant USING- THE BUILT-IN, SEMI-AUTOMATIC TESTER MAINTENANCE OR TEST THAT COULD AFFECT THE UNDERVOLTAGE CARD. ,
| |
| : 5. FOR ELECTRONIC. SYSTEMS SUBJECT TO SUCH DAMAGE, TEST PROCEDURE!
| |
| SHOULD CONTAIN CAUTIONS REGARDING CORRECT SETUPS AND USE OF TES1 EQUIPMENT. IT WOULD ALSD BE ADVISABLE TO M0DIFT CIRCUITRY SO THAT ROUTINE TESTS CAN BE PERFORMED WITHOUT LIFTING LEADS, JUMPERINh OR INSTALLING TEMPORARY METERS.
| |
| INPO'S EVALUATION OF THIS EVENT IS COMPLETE.
| |
| LIMITED DISTRIBUTION NUCLEAR POWER OPERATIONS. ALL COPYRIGHT 1985 BY THE INSTITUTE OF NOT FOR SALE. UNAUTHORIZED REPRODUCTION IS 4 RIGHTS RESERVED.
| |
| VIOLATION Or APPLICABLE LAW.
| |
| REPRODUCTION OF NOT NORE THAN TEf4 COPIES- BYTHE EACH RECIPIENT NORMAL COURSE FOR IT4 e INTERNAL USE OR USE BY ITS CONTRACTORS IN THIS REPORT SHOULD NOT BE OTHERWIbs BUSINESS IS PERMITTED.
| |
| TRANSFERRED OR DELIVERED TO ANY THIRD PERSON, AND ITS CONTENT S SHOULC NOT BE MADE PUBLIC, WITHOUT THE PRIOR AGREEMENT OF INPO.
| |
| DAVID HEMBREE, INPO, 404/953-7651 Information
| |
| | |
| ==Contact:==
| |
| | |
| _q.
| |
| BRAIDWOOD SIMULATOR MALFUNCTION
| |
| | |
| ==Title:==
| |
| Manual Reactor Trip Failure- ID: CRF-21
| |
| . _ ~ ~
| |
| * _-
| |
| NO: 6.3.4.4.21
| |
| [. ' -
| |
| Descriptions. Reactor trip breakers fail to open when operator initiates trip. Problem is due
| |
| 'to a faulty switch.
| |
| Variations . Date: 8/12/86 Rev: 4 u
| |
| Selectable Steps Inputs Comments
| |
| : 1. Select switch 1-3 1 = RD001 (Reactor trip switch on 5 panel) 2 = RD002 (Reactor trip switch on 6 panel) 3 = Both
| |
| / ,
| |
| Brief Plant Response:
| |
| Transient will continue. Operator should trip turbine and drive all rods in.
| |
| Power will decrease as rods are driven in.
| |
| Suggested Instructor Action:
| |
| j Use LOA CRF1 (RTA) and CRF2 (RTB) to have EO go to aux. electric room and trip
| |
| .the Rx trip' breakers.
| |
| Events: None l
| |
| .i 519M/197M/31 5/89 i.
| |
| l I
| |
| ( _ _-.-.-- . - -
| |
| | |
| 5P
| |
| .BRAIDWOOD SIMULATOR MALFUNCTION ,
| |
| Title OTAT Satpoint Failure ID:~ CRF-23 NO: 6.3.4.4.23
| |
| | |
| == Description:==
| |
| OTAT.setpoint fails due to summer failure.
| |
| Variations: Date: 9/19/88 Rev: 5 Selectable Steps . Inputs Comments
| |
| : 1. Select setpoint CRF23A 23A = Loop A Setpoint CRF235 23B = Loop B Setpoint CRF23C 23C = Loop C Setpoint CRF23D 23D = Loop D Setpoint
| |
| : 2. Select setpoint 0-150 percent
| |
| : 3. Select ramp time 0-99,999 sec.
| |
| Brief Plant Response:
| |
| OTAT setpoint will rise or fall as selected over a period of time giving the operator time to respond. A lack of corrective measures by the operator will result in overtemperature delta T alarm and/or trip. A turbine runback is initiated prior to a trip and outward rod motion is blocked at 3 percent below the trip setpoint. ,
| |
| Suggested Instructor Action:
| |
| ' Allow malfunction to continue as an option or clear malfunction when recognized by students and ordered corrected.,
| |
| 1 .
| |
| I Events: 1) DVR.06-02-88-036: OTAT Setpoint Failure High.
| |
| 519K/197M/33 5/89 0
| |
| | |
| C/cF- 2 3 I
| |
| DEVIATION INVESTIGATION REPORT PAGE LOOP 2C OTAT SETPOINT FAILURf HIGM G EVENT DATE DIR NUPSER _EEPORT DATE OPERATING I
| |
| 1 0 1
| |
| // SEQUENTIAL ,/,/ REVISION
| |
| /f MONTH DAY YEAR. STA UNIT YEAR // Nt#eER // NUPSER MONTH DAY YEAR i '
| |
| POWER 01 3 21 4 81 8 01 6 01 2 8 18 -
| |
| 0 13 1 6 -
| |
| 0IO O g5 063 883 LEVEL
| |
| , gg 5 i
| |
| CONTACT FOR THIS DIR TELE" HONE NUPEER AREA CODE Lee Sues. Anst. Superintendent Technical Services Ext. 2214 8l1l5 2I314l-I514I411 COMPLETE ONE LINE FOR EACH_COMPONEN A URE DESCRIBED IN THIS REPORT _
| |
| CAUSE SYSTEM COMPONENT MANUFAC- REPORTABLE CAUSE SYSTEM COMPONENT MANUFAC- REPORTABLE i TURER TO NPRDS TURER TO NPRDS X AIB El Cl Bl D RI11315 Y l I l I I l l l 1 1 I I I I 1 l l 11 SUPPLEMENTAL REPORT EXPEt'TED MDNTH DAY YEAR p
| |
| SUBMISSION
| |
| ^
| |
| I YES fif ves. conelete EXPECTED SUBMISSION DATE) 51 NO l l l TEXT A. PLANT CQPWITIQNS PRIOR TO EVENT:
| |
| Event Date/Ttoe_ V24/88 /.1516 Unit 1 MODE lf/A - N/A Rx Power _M/A__ RCS [AB) Temperature / Pressure N/A 4
| |
| Unit 2 M00E 1 - Power OserA11gna Rx Power 93% RCS (AB} Temperature / Pressure Nomal Ooeratine B. DESCRIPTION OF EVENT:
| |
| At 1516 on March 24, 1988, during the shif tly/ daily Surveillance 2005 C.1-1.2.3, the Loop 2C OTAT ;
| |
| Setpoint Indicator 2TI-431C was found in a failed high condition. This was confirmed at the OTAT Setpoint Pen Recorder 2TR 411. Abnormal procedure 200A INST-2 was entered and the blstables were placed in ;
| |
| a tripped condition. Limiting Condition for Operation Action Requirement (LC0AR) 3.1-la was entered and Nuclear Work Request (NWR) 854937 was initiated to troubleshoot and correct the failure.
| |
| C. CAUSE OF EVENT:
| |
| The root cause of the high setpoint indications, AT,p, on 2TI-431C and 2TR 411 was the failure of NLL l Card 2TY-0432A. NLL Card 2TY-0432A provides the temperature input to the AT,p summing and 2TI-411L.
| |
| AT sp is esiculated using the following equation and is electronically produced in this loop.
| |
| I+T1s (AT,p = K; -K2 3+T2s (T-588.4) 4 K3 (P-2235) -fi (Aq))
| |
| l I
| |
| $1nce it failed low, the -588.4 ters with the -K2 term suwuned to a higher setpoint value. The cause of
| |
| [ the failure is indeterminate. No further action is to be taken to determine the cause of the failure.
| |
| I I (2054M/0224M) J
| |
| | |
| DEVIATION INVESTIGATION REPORT TEXT CONTINUATION j TITLE DIR NUPEER PAGE
| |
| .(N SE')UENTIAL REVISION l LOOP 2C 6 TAT SETp0 INT FAILURE HIGH A MT YEAR NM mER.__ !
| |
| 01 6 01 2 81 8 -
| |
| Ol316 -
| |
| 0l0 2 0F 0l2 TEXT D. 14((TY ANALYSIS:
| |
| The abnormal procedurs 2BM INST-2 was entered and the bistables were placed in a tripped condition. Due to the redundancy of the Reactor Protection System (RPS), the tripped bistables resulted in a 1/3 coincident logic for RPS scluation which is conservative. The normal logic is 2/4 channels coincidence for the RPS actuation. Thus, there was no effect in the health and safety of the public.
| |
| E. CORRECTIVE _ ACTIONS:
| |
| NLL Card 2TY-0432A has been replaced and the affected loop recalibrates. LC0AR 3.1-la was talted on 3/25/88 at 0420 and the channel was declared operable.
| |
| F. EgylDU1.,9ff,UltRENCES:
| |
| There have been no previous occurrences of this particular failed component as documented by a DVR though an RTD failed for the 2C loop as documented by the following DVR. No trend has been identified.
| |
| W R NUPS ER TITLE ,
| |
| 6-2-67-003 (87-001) Reactor Trip Due to 2 of 4 Logic on Over Temperature Delta Temperature r
| |
| 'N G. COPf0NENT FAILURE DATA:
| |
| a) MANUFACTURER NDPENCtATURE MDDEL NUPSER mig PART NUPSER Westinghouse 7300 Circuit Card NL L 819579 b) RESutTS Cr NPRDS SEARCH:
| |
| Search shows that s'1though this type of card has failed before, there is nothing consistent (or trending) in the types of failures. 1TY-0421C at Byron, which is a simliar loop, had previously l f ailed due to a defective relay (see DVR above), however, most card failures are not normally tracked to this degree of discreteness.
| |
| c) RESULTS OF mR SEARCH:
| |
| None found i a (2054M/0224N)
| |
| | |
| y_ _
| |
| . LRAIDWOOD SIMULATOR MALFUNCTION rm, . Title Overpower Delta T Setpoint Failure ID: CRF-24
| |
| 'v) R0: 6.3.4.4.24
| |
| | |
| == Description:==
| |
| Overpower Delta T setpoint gradually fails to desired value due to summer failure.
| |
| Variations: Late 7/8/87 Rev: 5 Selectable Steps Inputs Comments
| |
| : 1. Select loop CRF24A 24A - OPAT setpoint in loop A CEF24B 248 - OPAT setpoint in loop B CEF24C 24C - OPAT setpoint in loop C CRF24D 24D - OPAT setpoint in loop d
| |
| : 2. Select setpoint 0-150 0-150 percent i
| |
| (,
| |
| \*
| |
| : 3. Select ramp time 0-99,999 sec.
| |
| Brief Plant Response:
| |
| As setpoint changes toward failed value, overpower delta T high alarms will be initiated along with rod withdrawal block and turbine runback. A trip will occur on two out of four coincidence if more Lian one loop setpoint is affected. The first annunciators received are OPAT TRIP and C-4 OPAT ROD STOP.-
| |
| Suggested Instructor Action:
| |
| None.
| |
| Evruts: None
| |
| '~ 519M/197M/34 5/89 ,
| |
| E
| |
| ~ _ . _ . _ , . . . . - ~ ~ _ _ _ . . . - . . _ . . .. - . - .-
| |
| -A_____.:_-_____---____..-._a ____________-__m._
| |
| | |
| BRAIDWOOD SIMULATOR MALFUNCTION i
| |
| Title Undervoltage Reisy Failure on RCP Buses ID: CRF-25 V NO: 6.3.4.4.25
| |
| | |
| == Description:==
| |
| Indicated voltage. ,n RCP buses gradually decrease until setpoint is reached and
| |
| - reactor trip occurs. This is due to a potential transformer problem.
| |
| Variations: Date: 8/12/86 Rev: 3 Selectable
| |
| $teps Input Comments
| |
| : 1. Select RCP bus 1, 2, 3, 4 1 = RCP A Bus 2 = RCP C Bus
| |
| . 3 = RCP D Bus .
| |
| 4 = RCP Bus A and B O
| |
| Brief Plant Response:
| |
| Undervoltage on RCP buses is an anticipatory trip prior to low flow conditions to prevent exceeding DNB limits. The >ndervoltage causes a reactor trip on two of four buses. One bus undericitage will trip the appropriate RCP )
| |
| resulting in three loop operations which result in the operator inserting a menual trip. Will get reactor trip on low flow if power >30 percent power.
| |
| Suggested Instmetor Action:
| |
| . . None.
| |
| Events: Ncne 519M/197M/35 5/89 i
| |
| ___2._ .-____
| |
| | |
| g.
| |
| BRAIDWOOD SIMULATOR MALFUNCTION
| |
| | |
| ==Title:==
| |
| Underfrequency Relay Failure on RCP Buses ID: CEF-26
| |
| ?v' NO: 6.3.4.4.26
| |
| | |
| == Description:==
| |
| RCP buses U/F relays failure on train A.
| |
| Variations: Date: 2/15/83 Revt 2 Selectable Steps Inputs Comments
| |
| : 1. Select RCP bus 1, 2, 3, 4 1 = RCP A Bus 2 = RCP C Bus 3 = RCP D Lus 4 = Bus A and B O
| |
| Brief Plant Response:
| |
| Underfrequency on two -f four RCP buses will cause a reactor trip. Trip signal should be generated prior to low flow trip. Only the appropriate RCP will trip on one bus underfrequency.
| |
| Suggested Instructor Action:
| |
| None.
| |
| U 519M/197M/36 5/89 j I
| |
| ,m =%g -e .~ % ,,
| |
| MN#reem+ i eae, e v .a. . .,
| |
| *N W
| |
| ---- -m._____ _ _ _ ___. _ _ _ _ _ . , _ , _ _ _ _ , _ _
| |
| | |
| BRAIDWOOD SIMULATOR' l MALFUNCTION LISTING j_, CHEMICAL AND VOLUME CONTROL SYSTEM 1
| |
| Q)
| |
| CVC-1 Boric Acid Flow Transmitter (FT-110) Failure CVC-2 VCT Divert Valve Failure (112A)
| |
| CVC-3 PCV 131 Control Failure CVO-4 Charging Header HCV-182 Control Failure CVC-5 Failure of Positive Displacement Charging Pump Speed Control CVC-6 RCP Seal Pailure CVC-14 Boric Acid Transfer Pump Trip CVC-15 Plugged CVCS Filters CVC-16 Letdown Line Leak Inside Containment CVC-17 Charging Line Leak Inside Containment CVC-18 Loss of Charging Pump r'~'T CVC-19 Primary Water Make-up Pump Trio
| |
| %.Y CVC-20 CVCS Various Valve Failure CVC-21 VCT Level Malfunction CVC-23 Make-up Control Failure CVC-26 CVCS Demineralizers Depletion CVC-27 Letdown Leak Outside Containment CVC-28 Seal Injection Leak
| |
| ,m
| |
| ~~ /
| |
| 638M/263M/S 8/87
| |
| | |
| BRAIDWOOD SIMULATOR MALTUNCT. 9 N
| |
| | |
| ==Title:==
| |
| Boric Acid Flow Trans:nitter (PT-110) Failure ID: CVC-1 50: 6.3.4.5.1
| |
| | |
| == Description:==
| |
| Boric acid flow transmitter is miscalibrated.
| |
| Variations: Date: 4/11/87 Rev: 4 Selectable Steps Inputs Comments 1 Select calibration -100 percent to -100 percent 0 span range +100 percent +100 percent 2X normal span
| |
| : 2. Select ramp tir.a 0 to 99,999 sec.
| |
| Brief Plant Response: [ Based on system operating for auto make-up.]
| |
| O Failure of the boric acid transmitter high, causing a lower than required boric acid concentration in the blended flow.
| |
| : 1. Slow dilution of RCS will occur.
| |
| : 2. Control rods (if in auto) will intermittently insert to control T,, .
| |
| : 3. First annunciator received will be AI LIMITS EXCEEDED or LO-2 INSERTION LIMIT depending on initial power level.
| |
| : 4. If actuated during plant startup, reactor could achieve criticality at a lower rod height.
| |
| Failure of the boric acid transmitter 1qr, causing a higher than required boric acid concentration in the blended flow. (IC-17, 100%, all systems in
| |
| , automatic)
| |
| : 1. Control roda (if in auto) will withdraw to maintain T,yg.
| |
| : 2. First annunciator received will be BANK D ROD STOP C-11.
| |
| O 0645D:4 883M/2 5/89
| |
| ~ - - - - . - - - - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
| |
| | |
| BRAIDWOOD, SIMULATOR MALFUNCTION-l l~
| |
| 1,l k
| |
| | |
| ==Title:==
| |
| Boric Acid Flow Transmitter.'(FT-110) Failare -ID: CVC-1
| |
| %'j D. .. Brief Plant Respons's (continued):
| |
| : 3. If actuated during plant startup, criticality at a higher rod hei8ht-or'perhaps operator will not be able to take reactor critical.
| |
| Suggest4d Instructor A: tion:
| |
| None.
| |
| Events: Nons kJ d 0645D 4 883M/3 5/89 C-*----------_mL._.__--..._.___________.________._
| |
| | |
| BRAIDWOOD SIMULATOR MALFUNCTION l
| |
| Titles VCT Divert Valve Failure (112A) ID: CVC-2 l NO: 6.3.4.5.2
| |
| -V Description . Failure of I/P converter (LY 112A) for LCV 112A to selectable position.
| |
| Variationsi Date: 4/11/87 Rev: 4-Selectable Steps Inputs Comments
| |
| : 1. Select divert 0, 1 0 - VCT position to hold 1 - HUT up tank
| |
| ]
| |
| : 2. Select ramp time O to 99,999 sec.
| |
| iQ
| |
| () (IC-17, 100%, all systems in automatic)
| |
| Brief Plant Response:
| |
| Note: Manual override possible to VCT or HUT.
| |
| Failure of full flow to the VCT will usually only be a significant problem if initiated during a plant cooldown or during an RCS dilution. VCT level will increase, with first annunciator received being either VCT LEVEL HIGH or VCT PRESSURE HIGH depending on initial VCT level.
| |
| Failure of full flow to the HUT during normal power operations will result in a decrease in VCT level with actuation of automatic make-up (if selected) and possible change-over to RWST suction for the charging pumps. No annunciators received with RMCS in automatic.
| |
| -(
| |
| 0645D:4 883M/4 5/89 l
| |
| ~~~-~~ ~ ~ ' ~ ~ ~ ~ ~ ~ ~ ~
| |
| ~~ ~
| |
| i_E- - - -[ ~ ~
| |
| | |
| BRAIDWOOD SIMULATOR MALFUNCTION-
| |
| -Title: L-VCT Divert Valve Failure.(112A) ID: CVC-2 L( ,
| |
| . Suggested' Instructor Action:
| |
| ' Clear malfunction when repairs are complet:4d .
| |
| Event: None i
| |
| O O 0645D:4 883M/5 5/89 p.-
| |
| b-_-__._________________ _ _ _ _ __
| |
| | |
| BRAIDWOOD SIMULATOR MALFUNCTION
| |
| | |
| ==Title:==
| |
| PCV 131 Control Failure IDt CVC-3'
| |
| _ ,q NO: 6.3.4.5.3 l
| |
| | |
| == Description:==
| |
| Failure of I/P converter (PY 131) for PCV 131 to'selectable position. (NO MANUAL CONTROL POSSIBLE)-
| |
| L Variations: Date: 4/11/87 Rev: 4 Selectable Stept Inputs Comments
| |
| ,1. Select valve fail 0, 1 0 - close position 1 - open
| |
| : 2. Select ramp time O to 99,999 sec.
| |
| O- Brief Plant Responses (IC-17, 100%, all systems in automatic)
| |
| Failure of PCV 131 full open - will result in reduced pressure ir, the letdown line and a corresponding increase in letdown flow with an increase in charging flow.
| |
| Failure of PCV 131 full closed - will result in a loss of letdown, increased pressure in the letdown line, and lifting of the letdown relief valves, with the first annunciators received being LETDOWN EX OUTLET PRESSURE HIGH. LP LETDOWN RELIEF TEMP HIGH annunciator will also alarm.
| |
| Suggested Instructor Action:
| |
| Use LOA CVC-1 (CV8408A) to isolate PCV-131, CVC-2 (CV8409) to bypass PCV-131.
| |
| Events: None 0 0645D:4 883M/6 5/89 a __ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ -
| |
| | |
| BRAIDWOOD SIMULATOR MALFUNCTION
| |
| | |
| ==Title:==
| |
| Charging Header HCV-182 Control Failure ID: CVC-4
| |
| -/O NO: 6,3.4.5.4 h
| |
| | |
| == Description:==
| |
| I/P cunverter (HY 182) fails to a selectable position. (NO MANUAL CONTROL POSSIBLE)
| |
| Variations: Date: 4/11/87 Rev: 4 Selectable Steps Inputs Consents .
| |
| : 1. Select valve fail 0, 1 0 - cloae position 1 - open
| |
| : 2. Select ramp time 0 to 99,999 sec.
| |
| (IC-17, 100%, all systems in automatic)
| |
| O]
| |
| Q.
| |
| Brief Plant Responses Failure of HCV-182 full closed - will result in an increase in seal injection flow and a no charging flow with the first annunciator received being RCP SEAL INJ FLTR AP HIGH. Other annunciators received include: 1) LETDOWN EX OUTLET PRESSURE HIGH, 2) REGEN HK LETDOWN TEMP HIGH, and 3) LP LETDOWN RELIEF TEMP HIGH.
| |
| Failure of HCV-182 full open - will result in a lower seal injection flow.
| |
| (No annunciators received.)
| |
| Suggested Instructor Action:
| |
| Use LOA CVC-10 (8402B) to isolate HCV-182, CVC-11 (8403) to bypass around RCV-182.
| |
| Events: None 0645D:4 883M/7 5/89 m .
| |
| E_i___._____________ _ . _ _ . _
| |
| | |
| BRAIDWOOD SIMULATOR MALFUNCTION
| |
| | |
| ==Title:==
| |
| Failure of Positive Displacement' Charging Pump 'ID: CVC-5 Speed Control NO: 6.3.4.5.5
| |
| [(
| |
| | |
| == Description:==
| |
| Mechanical failure of P. D. pump speed changer (NO MANUAL CONTROL POSSIBLR).
| |
| Variations .Date: 4/11/87 Rev: 4 l Selectable Steps Inputs Comments
| |
| : 1. Select P. D. pump 12 - 98 spa output
| |
| : 2. Select ramp time O to 99,999 sec.
| |
| V Briet ~1 ant Response: (IC-17, 100%, all systems in automatic)
| |
| If the ptamp speed fails low, then the RCP SEAL WTR INJ FLOW LOW and CHG LINE FLOW LOW annunciators are received first and a gradual loss of pressurizer level occurs with eventus1 letdown isolation.
| |
| If the pump speed fails high, a gradual pressurizer level fueresse will occur. The first annunciator received will be PZR LEVEL CONT. DEV. HIGH HTRS ON.
| |
| Suggested Instructor Action:
| |
| Clear ms1 function when repairs are completed.
| |
| Events: None 0645D:4 883M/8 5/89 l
| |
| 1
| |
| --.r g- - - - - - - - . .
| |
| A
| |
| | |
| BRAIDWOOD SIMULATOR MALFUNCTION
| |
| | |
| ==Title:==
| |
| ECP Seal Failure ID: CVC-6 NO: 6.3.4.5.6
| |
| /)
| |
| | |
| == Description:==
| |
| . Number 1 seal fails in selected FC?.
| |
| Variations: Date: 4/11/87 Rev: 5 Selectable Steps Inputr. Comments
| |
| : 1. Select pump CVC-6A CVC-6A = Loop 1 CVC-6B CVC-6B = Loop 2 CVC-6C CVC-6C = Loop 3 CVC-6D CVC-6D = Loop 4
| |
| : 2. Select leak size 0 - 50 spa (Based on normal parameters)
| |
| : 3. Select ramp time 0 to 99,999 sec'.
| |
| Q)
| |
| Brief Plant Response This will result in an increase in flow in the seal return line, a decrease in seal flow to the other three RCPa, a decrease in pressurizer level with possible letdown isolation. The first annunciator received are RCP SEAL LEAKOFF FLOW HIGH and RCP SEAL WTR INJ FLOW LOW. ;
| |
| )
| |
| Sugfested Instructor Action:
| |
| i None.
| |
| Events: None
| |
| (
| |
| %/
| |
| 0645D:4 883M/9 5/89 1 l l
| |
| L___________._________ _ _ . _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ .
| |
| | |
| BRAIDWOOD SIMULATOR MALFUNCTION
| |
| | |
| ==Title:==
| |
| ' Boric Acid Transfer Pump Trip -ID: CVC-14 p,
| |
| / NO: 6.3.4.5.14
| |
| | |
| == Description:==
| |
| Selected boric acid transfer pump trips due to overload (bad bearings).
| |
| Variations: Date: 3/14/89 Rev: 5 Selectable Steps Inputs Comments
| |
| : 1. Select BA pump CVC-14A CVC-14A = 1A Pump CVC-14B CVC-14B = Common Pump Brief Plant Responses (IC-17; 100% with boration in progress).
| |
| (Q -Selected pump trips a M boric acid flow stops. The first annunciator received is BA PUMP TRIP.
| |
| Suggested Instructor Action:
| |
| To sway to the common BA pump, the following LOA's are needed:
| |
| ~
| |
| - CVC-50 = F
| |
| - CVC-49 = T
| |
| - CVC-43 = 1
| |
| - CVC-74 = 1
| |
| - CVC-76 = 1 c Events:
| |
| : 1) DVR 20-01-86-053: BA Transfer Pump Failure 0645D:4 883M/21 5/89 i
| |
| ___-___.__-_____--_-_______-_______m
| |
| | |
| _ _ _ {Y 0_
| |
| OE/!ATICN INvt3TICA f {3N DE*':# f TLE: FAILURE OF "0" BORIC ACIO TRAh5FER PUMP 3EaL DUE TO DEACHEADING
| |
| * IGE 1 10F3 S' 2 l fENTDAf(_ DtR NUMBER R CRf DATE // r // .',.
| |
| // SEQUENTIAL '' REVI5!0N
| |
| # ' }/ f '/ / ,- '
| |
| MONTH DAY YE STA UNIT YEAR / JLLM(Q,, ,fgM.g1 *!Q!!M ,,0 A_Y_ J_ EAR 5
| |
| ///fIIIIN '
| |
| P0YER /t ''h
| |
| // ''t i o
| |
| ///t' or
| |
| . ?! 1 21 2 BI 6 21 0 01 1 81 6 --
| |
| Of El 1 -
| |
| El 0 of 1 1 10 al 7 01 01 0 l f/f/f''i CONTACT FOR THI1 klR __ ___ _ _ . _ _ _
| |
| NAME TELEPHONE NUMiER AREA CODE JENNY 0. TOLAR. TECH STAFF ENGINEER Ett. 2484 COMPLfff DNE Lfht F0e EACH COMPONEN FA LORE Df1Ce!BfD IN THit rep 0RT 811 l $ 4 J. E lB I-12 19 Lo I i '
| |
| CAUSE SYSTEM COMPONENT MANUFAC- REPORTABLE CAUSE
| |
| / SYSTEM COMPONENT MANUFAC- REPORfaB'[
| |
| TURER TO NPPOS /, TURED f3 NFD0i I A Al B Pl *l *l
| |
| * Gl21010 N / l l I l 'l 1 l _
| |
| l i I I I I I '
| |
| / I I I I I SUfft.EMENTAL REPORT EXPECTED , MONTH DAY V!AR SUBM15510N f YES fif ves. comolete EXPECT 7.0 submiss 10N D ATE) . Jl 40 l, TEXT O. PL ANT CONDIT0NS DRIOR TO EVENT:
| |
| Mod) 1 Cold Shutdown RCS Temperature / Pressure 100*F/0 ns to G
| |
| )JESCRIPTIONOFEVQ{;
| |
| Durtng Shift !!! on 11-21-86. operating was transferring the contents of the Boric Acid Eatching Tar'k (CB] t) I the Unit 1 Boric Actd Tank (per BwCP CV-25) using the Unit "0" Boric Acid Tr41sf er Pump (OAB03Pl. Aft.e* the transfer process was complete the system was lined up per Swop AB-10 to recir:ulate the Barte tc'1 fin. :Es*
| |
| using 0 ABO)P so chemistry could cbtain samples. After starting the purro on rec'rcu14 tion (3t 2'19). D e r p. . ~,.
| |
| personnel vertfted a pump discharge pressure of 115 ostg. '/alues of flo a to *Ne BAf or oressure trov,3 tre filter were not noted, l
| |
| Ouring 11-22-16 Shif t I excessive leakage f rom the p4cking of DAB 03P a4 s ioentified. At 1105, the cumo =4; I stopped, isolated (valves IAB8a65 and 1ABSJ68 were closed). .tnd pump IABG3P was lineup per SwCP 18-10 tnc started to continue the recirculation mode. During the process of vertffing the lineup for IA313P. Opet it W identified that valves IMM45g (BAT ractre. valve) and 1 AB84a6A (fitter outlet valvel were closed. Thus. tne Unit "0" Boric Acid TranMer Pump had been running =1tnout i recirculation 04th.
| |
| I I
| |
| I a
| |
| y
| |
| /%
| |
| i i
| |
| 1557m(0128B7)/108EA/24 l
| |
| t-_-__. __
| |
| | |
| 3Ev!Af0N [N/EST! W IC4 8EFCE" iEef CurtwATIO a 3 E: FaltuRE OF "0" BCRIC ACI3 !RAN5fER PUttP SEAu C!p *stPEEP c:*E,,_
| |
| DUE TO CEADHEADING SE%ENTIAL D E'/ ! 5 ! 0's j j
| |
| ,,1T A UNIT vfAR NUMEER
| |
| * _,,&PSER !
| |
| I G ,,m 2l 0 el 1 81 6 -
| |
| of si : -
| |
| A L_ J ._.?ict ) i :
| |
| ; )
| |
| ' L j' i C. tam t or tvtNf-I The root cause of this inc'tcent is that valves 14E8359 and IAB8486A were closed causing oump 0ABC3P to ce derdhted and the seal f ailure. Per cpertting perso'inel, the valves were procerly positioned Dr cr to enter rg BwCP A8-10 to rec trculate tne BAf. If this were the case, uriauthor tred personrel reposit tened 'cloieo) the f
| |
| valves while the pump was running, another D1;sibility newever, is tnit the valve; -ere trrorocerlf posittored j l by operating prior to starting CA803P on recirculation (BwCP AB-10). Based upcn a review of the pump curve f or OAB037 with the system conditions that eitsted it t'ie ttee of the incidert , this is possible. Swop C/-25 requires valve 1 AB4459 to be closed while transf erring coric 4cid f r om the B,iten fink to tre S AT. After complettng Sw0P CV-25 and prior to start'rq 8 0P AB 10 s alve I AE8459 should have' ocen ocened oer B-0P AB N .
| |
| which is a prereoutstte to BwCP AB 'O. *he saive (1488459) could have eeen missaj or impreperif reposittoneo of uperating prior to entering BwCP AB.!0 ' alve I A68446A 18 to be open oer BwAP aB-Mt and is 90t to of cl3 Ho during Bw0P CV-25 or SWAP AB.10. 'h t '. e4twe (t A88446A) c2uld h4ve been cloted in error b,y opertting or ter to entering BwCP AB-10. In both cases, personnel error caused the improper postttoning of the v41.es.
| |
| : 1. 5AFETY ANALYs!s:
| |
| At the present plant conditions, deadheading 0AB03P while attempt tng to rectrCul4te the BAI does not crel*.! m ,[
| |
| adverse safety consequences. Though this may dam.lge the pump, the plant condition is nat ;eporotzed. If tre pup *o is damaged beyond use the Unit "I'' or Unit"2" purnps may be used to rectreulate tne Sat. Since the Soric Acid fransf er Pumps are used to borate the Reactor Coolant System (RCS) {A8l adver,e l1 ret, tenaitions coule result *1f this event (ded@teading the pump) were to occur under the worst case cond'tions (using 0AB03P to torate,the RCS at full power). The Tacnesical Spec 1ftcations (Tech Specs) recuire 4t letst 2 of 3 boron
| |
| ( njection flow paths to be operible. Thus, depanding on the availabt11ty of other plant ijstets the c::rren e (d of tats event could result in a fech Spec violation.
| |
| E. Q8LRJfJ.1Y.L.AGION S :
| |
| ine Shif L Engincer s . nill at.cus 5 ..ite tne er x . cent
| |
| * tre rgori wce of eoll w$ng orocecu a . :t:.ai, iri 4ssuring proper salve timeuu . inis w' te *.* ica e1 ty Ite"t sa56-200-M-05301 ;
| |
| )
| |
| F. P? 1Y1DillEC1LRAIEnli None G. (OW OMENT F&ILuft l&IA; t' A'IUF AC TUR E8 N_QPENCLA{Qf( t,$,(j L Q , t[,$ P&Pf N'LM,$1![
| |
| Gould Pumps Inc. Boric Acid !!965 T N7386092-1-2-3 ffansfer Pump i ,
| |
| O i i
| |
| t
| |
| \ $
| |
| %.)
| |
| 1557 bi S12387)/ lM?,A/2 $ )
| |
| _ _ _ _ - - - - - - - - - - - 1
| |
| | |
| ( '
| |
| BRAIDWOOD SIMULATOR MALFUNCTION ID: CVC-15 j..j.
| |
| | |
| ==Title:==
| |
| Plugged CVCS Filters NO: 6.3.4.5.15 i J.
| |
| | |
| == Description:==
| |
| Selected ffiters become blocked to selected value.
| |
| Variations: Date: 3/14/89 Rev: 5 k i
| |
| Selectable Steps Inputs Comments
| |
| : 1. Select filter CA* 'f 4
| |
| * C CVC-15A = Boric Acid Filter CVC-15B = Reactor Coolant Filter CVC-15C = Seal Injection Filter (1)
| |
| CVC-15D = Seal Injection Filter (2)
| |
| CVC-15E = Seal Return Filter
| |
| : 2. Select blockage 0 - 100 percent 0 percent = no blockage V 100 percent = full blockage
| |
| : 3. Select ramp time 0 to 99,999 sec.
| |
| Brief Plant Response: (IC-17, 100%, all systems in automatic).
| |
| BA Filter - BA flow will be lost. THe first annunciator received is BA FLOW DEVIATION, RC Filter - not noticeable during normal operations, however during dilution, letdown will bypass CV-112A and fill the VCT.
| |
| Seal Inj Filter - loss of seal injection flow to the RCP's. THe first I annunciator received 16 RCP SEAL WTR INJ FLTR AP EIGH.
| |
| Seal Return Filter - seal return flow lowers. THe first annunciator received is RCP SEAL LEAKOFF FLOW LOW.
| |
| O k.
| |
| 0645D:4 883M/22 5/89 l
| |
| p.
| |
| | |
| BEAIDWOOD SIMULATOR MALFUNCTION
| |
| ['
| |
| .I jb. Title 3- Plugged CVCS Filters (Cont.) ID: CVC-15 NO: 6.3.4.5.15 l_. _
| |
| }
| |
| j Suggested Instructor Action: 1 BA Filter - when requested, bypass filter by using LOA CVC-42 and isolate ,
| |
| i.
| |
| filter by using LOA-CVC-48.
| |
| RC Filter - when requested, bypass filter by using LOA CVC-3.
| |
| Seal Inj. Filter - when requested, swap to standby filter by using LOA CVC-51/52 (A/B).
| |
| Seal Return Filter - when requested, bypass filter by using LOA CVC-21 and isolate filter by using LOA CVC-18.
| |
| Events:
| |
| : 1) DVR 06-02-87-017: .Ex Coolant Filter Plugged j i
| |
| s !
| |
| 'l O 883M/23 5/89
| |
| , 0645D:4
| |
| . . . _ .. . . ~ . . . . . . . _ .
| |
| | |
| DEv!ATION INVESTIGATION REPORT TITLE paGE FAILURE OF CATION DJNINERALIZER 01 forl 021 v EVENT DATE DIR NUMBER REPORT DATE SEQUENTIAL REVISION /
| |
| NTH DAY YEAR STA UNIT YEAR g/
| |
| / NUMSER NUMBER MONTH DAY YEAR 1 POWER /
| |
| //g 012 214 af7 016 012 air -
| |
| of 117 -
| |
| 010 0 14 0 19 817 LEVEL
| |
| , ,, q
| |
| / ,ff
| |
| /,
| |
| CONTACT FOR THIt DIR NAME TELEPHONE-NUHeER AREA CODE F. Herrhaak Tech traff tunarvisar Ext. 2243 8i t l E 211 l a l' - l 514 l4 l 1 COMPLETE ONE LINE FOR EACH COMPONEN URE DESCRIBED IN THIS REPORT CAUSE SYSTEM COMPONENT MANUFAC- REPORTABLE CAUSE SYSTEM COMPONENT MANUFAC. REPORTA8LE TURER 70 NPRDS TURER TO NPROS r Cla i FI 01 M WI 11 210 N l i i ! l l 1 1 I I I I I I I I I I I I SUPPLEMENTAL REPORT EXPECTED MONTH DAY YEAR p
| |
| SUBMISSION l YEt f if van. conn 1sta EXPECTID_1UBP111110N DATE1 l NO O7 tl5 87 TEXT A. PLANT CONDITIONS PRIOR TO EVENT:
| |
| MODE 1 - Power Onoratinn Rx Power E RCS (AB] Temperature / Pressure Normal OP$
| |
| .O B. DESCRIPTION OF EVENT:
| |
| On 02/24/87 at 1300, the Unit 2 Cation Domineralizer was valved into the letdown flowpath. Shortly thereaf ter, volume cwitrol tank level was observed to be decreasing and letdown line pressure was increasing, signifying a flow blockage. A blind flange downstream of the domineralizer resin fill / flush valve 2CV8515 began leaking contaminated liquid. Letdown was isolated stopping the leakage. The reactor coolant filter was found to be plugged and was replaced. After flushing the line to the hold-up tank.
| |
| letdown was re-established. No safety actuations occurred. No safety systems were declared inoperable.
| |
| C. CAUSE OF EVENT:
| |
| When the Catton Dominera112er was valved in, the effluent contained loose resins which caused a flow blockage at the reactor coolant filter. At this time it is unknown whether the resin was loosely packed in the vessel or if the lateral structure was damaged. Further inv#,stigation will be conducted during an outage of sufficient duration. A supplemental report will be issued at that time.
| |
| I I i a
| |
| O
| |
| =
| |
| (1325M/0156M) l
| |
| | |
| DEVIATIO4 INVESTIGATION REPORT TEXT CCNTINUAf!ON TITLE DIR NUMBER FACE ;
| |
| SEQUENTIAL REV!5!ON JTA UNIT YEAR NUMBER NUMBER f .
| |
| \ .
| |
| FafLURE OF CAff0N DENINERALI2ER Ol6 012 817 ~
| |
| Of 1 17 -
| |
| Ol0 02 0F 02 l TEXT l
| |
| 0, $AFETY ANALYSI1:
| |
| The event had no safety consequences. Reactor coolant lithium concentration limits were never exceeded.
| |
| The isolation of letdown did not affect any safety related functions of the chemical volume control system. The leak was isolated and the contaminated area was contained and decontaminated.
| |
| \
| |
| E. CORRECTIVE ACT10mE:
| |
| Nuclear Work Re@ests are intttated to repair the leaking valve and to investigate and repair the problem with the Catton Demineralized for interis operation. 4 cubic feet of cation resins were added to the 28 l mixed bed domineralizer to control lithium concentration. AIR 6-87-2033 18 tracking the congletion of this {
| |
| investigation.
| |
| F. PREV!aus OCCURRENCES-0.Yt.JM Sit 1111.1 NONE COMPONENT FAILURE DATA:
| |
| d a) MANUFACTURER NOMENCLATURE MODEL MulGER MFG PART NUMBER Westinghouse Resin Domineralizer DMCB b) RESULTS GF NPRDS SEARCH:
| |
| Not Appitcable
| |
| : 1. 2 1
| |
| (_%
| |
| (1325M/0156M)
| |
| | |
| _j- BRAIDWOOD SIM'JLATOR MALFUNCTION; ID: CVC-16'
| |
| -Title: Letdown Line' Leak Inside Containment -]
| |
| i
| |
| ' NO : 6.3.4.5.16
| |
| ~
| |
| | |
| == Description:==
| |
| Leak between CV-459 and the inlet to the regen HK.
| |
| Variations: Date 3/14/89-Rev 4 1,
| |
| 4' Selectable Steps Inputs Comments
| |
| : 1. Select leak rate 0 - 1000 spa Flow based on normal operating pressure of 2235 peig
| |
| : 2. Select ramp time 0 - 99,999 sec.
| |
| O Brief Plant Responses (IC-16, 75%, all systems in automatic)
| |
| Pressurizer level and pressure will decrease in proportion to the selected leak rate. Charging flow will increase to try to maintain pressurizer level.
| |
| When pressurizer level decreases to the low level,-heater cutoff / letdown isolation (17 percent), the leak is automatically isolated.
| |
| i Containment parameters will indicate symptoms of a small LOCA. (Pressure and temperature increasing, radiation levels increasing, and sump levels increasing). The first annunciators received are PZR LEVEL CONT DEV LOW and PZR PRESS CONI DEV LOW HTRS OR.
| |
| Suggested Instructor Action:
| |
| None.
| |
| Events: None 0645D:4 883M/24 5/89 L . . . _ . _ . . . . . _ . . _ _ . . _ . . . . . . . _ , _ . . . . . . _ . . _ . . . . , . . _ . _ _ _ _ , . . . -
| |
| ^ ----- m- - - - - , - - - - - - - - - - - - - _ _ , . , _ , . _ _ . , _ _ _ , , , _ _ _ , , _ _ _ _ _ _ _ _ _ , _ _ _ _ _ _ _ _ _ , _
| |
| | |
| 1
| |
| . BRAIDWOOD SIMULATOR MALWNCTION i
| |
| ! Titles ~ Charging Line Leak Insida Containment ID: CVC-17
| |
| , . f' - NO: 6,". 4.5.17
| |
| {
| |
| l
| |
| | |
| == Description:==
| |
| Leak between ICV 8378A and 1CV8378'B.
| |
| Variations: Date: 3/14/89 Rev: 5 Selectable Steps Inputs Comments
| |
| : 1. Select leak rate 0 - 100 percent Percent of total charging flow (based on normal operating pressure)
| |
| : 2. Select ramp time 0 - 99,999 sec.
| |
| C Brief Plant Responses (IC-17,100%, A:4 SYSTEMS IN AUTOMATIC)
| |
| Chargins flow will increase, letdown tempera 6 re will decrease in proportion to the magnitude of the charging flov, pressurizer level vill decrease and possible letdown isolation could occur. RCP seal flow will increase, and containment parameters will be indicative of an RCS leak. The first annunciators received are CENT DRAIN LEAK DETECT FLOW HIGH and CHG LINE .40W
| |
| ''-- HIGH LOW.
| |
| Suggested Instructor Action: j Clear malfunction when repairs are ccapleted.
| |
| Events: None 0645D:4 883M/25 5/89 i . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ . _ . _ _ _ . . _ . . _ _ , . _ . _ _ . _ _ _ _ _ ._
| |
| | |
| BRAIDWOOD SIMULATOR MALFUNCTION i.
| |
| ~
| |
| | |
| ==Title:==
| |
| Loss of Charging Ptump ID: CVC-18 NO: 6.3.4.5.18.
| |
| | |
| == Description:==
| |
| ' Selected charging pump trips due to overload (bad bearing).
| |
| L Variations:- Date: 3/14/89 Rev: 5 s.
| |
| Selectable Steps Inputs Comunents
| |
| : 1. Select ptany CVC-18A CVC-18A - Positive displacement psamp CVC-18B CVC-188 - Centrifugal charging ptany A CVC-18C CVC-18C - Centrifugal charging ptany B Brief Plant Response (IC-17, 10015, all systems in automatic)
| |
| When the selected charging pump trips, charging and seal injection flow drops to zero. Pressurizer level starts slowly decrassing. Due to the loss of cooling to the regenerative heat exchanger, letdown temperature will
| |
| . increase.- VCT level will increase and letdown diversion to the holiup tanks vill, occur. Letdown isolation will occur as the pressurizer level decreases.
| |
| The first annunciators received are CHG PUMP TRIP and CHG LINE FLOW HIGH LOW.
| |
| Suggested Instructor Action:
| |
| None.
| |
| Events: SER 86-38/0PEX 86-3800 i
| |
| > Charging ptanp problems due to plugged Lube Oil Coolers on the SX l ''
| |
| sidie. Loss of SX flow to the toolers caused increased lube oil temperatures resulting is numerous temperature alarms and requiring shutdown of the charging pumps. .
| |
| i ..
| |
| i l
| |
| O
| |
| * 0645D:4 883M/26 5/89
| |
| | |
| x ... ..-.:.:....:
| |
| ... .............. . .M-B -
| |
| :IS:633 FORSYTHL(INPO).26-NOV-86.11:17 PT
| |
| | |
| ==Subject:==
| |
| SERl38-86, POTENTIAL' LOSS OF CHARGING FLOW'
| |
| ~
| |
| [~
| |
| | |
| ==SUBJECT:==
| |
| POTENTIAL' LOSS OF CHARGING FLOW DUE TO LUBE OIL HEAT EXCHANGER PLUGGING'
| |
| . UNIT (TYPE): FARLEY 1 (PWR) ]
| |
| DOC.NO/LER NO: 50-348/(LATER) '
| |
| EVENT DATE: S/1/86-
| |
| 'NSSS/AE! WESTINGHOUSE /BECHTEL j i
| |
| SUhMARY:
| |
| SEDIMENT IN THE SERVICE WATER SYSTEM PLUGGED THE LUBE OIL HEAT EXCHANGERS FOR THE TWO AVAILABLE CENTRIFUGAL CHARGING PUMPS. i THIS CAUSED THE LUBE OIL. SYSTEMS FOR THE GEAR DRIVES OF THE PUMPS
| |
| :TO OVERHEAT._ ONE CHARGING PUMP WAS MAINTAINED.IN SERVICE BY
| |
| . USING ABNORMAL METHODS TO COOL ITS LUBE OIL HEAT EXCHANGER. THIS ALLOWED ENOUGH TIME TO. CLEAN THE PLUGGED HEAT EXCHANGER ON THE OTHER CHARGING-PUMP AND RETURN IT TO SERVICE.
| |
| THIS EVENT IS SIGNIFICANT BECAUSE IT INVOLVED A COMMON-MODE FAILURE MECHANISM THAT CAN DISABLE ALL CHARGING CAPABILITY. THE CHARGING SYSTEM AT THIS PLANT IS REQUIRED FOR EMERGENCY CORE COOLING DURING CERTAIN SMALL BREAK LOSS OF COOLANT ACCIDENTS.
| |
| DESCRIPTION:'
| |
| WITH THE PLANT OPERATING AT 100% POWER.,.ONE CHARGING PUMP (lA)
| |
| ~
| |
| WAS OUT 0F SERVICE FOR MAINTENANCE. A SECOND CHARGING PUMP (lB)
| |
| WAS IN NORMAL SERVICE. A THIRD CHARGING PUMP (lC) IS A SWING PUMP. IT WAS LINED UP TO THE OUT-OF-SERVICE TRAIN AND WAS AVAILABLE FOR USE. ALL THREE ARE CENTRIFUGAL PUMPS.
| |
| AT 1230, INCREASING LUBE OIL TEMPERATURES INITIATED A CHARGING
| |
| :PUMP HIGH-TEMPERATURE ALARM. THE LUBE OIL TEMPERATURE FOR THE IB CHARGING PUMP GEAR DRIVE WAS 145 DEGREES FAHRENHEIT. THIS INCREASED TO 155 DEGREES FAHRENHEIT BY 1238. THE 1C SWING CHARGING PUMP WAS STARTED, AND THE 1B PUMP WAS SHUT DOWN SO THAT
| |
| ~
| |
| MAINTENANCE PERSONNEL COULD CLEAN THE LUBE OIL HEAT EXCHANGER.
| |
| * WHILE THIS MAINTENnWCE WORK PROCEEDED THE LUBE OIL TEMPERATURE l FOR THE 1C PUMP WAS INCREASING AND INITIATED THE HIGH-TEMPERATURE ALARM (145 DEGREES FAHRENHEIT) AT 1310. IN ANTICIPATION OF LOSING THIS PUMP AND CONSISTENT WITH TECHNICAL SPECIFICATION RESTRICTIONS, REACTOR POWER WAS DECREASED. TO MAINTAIN THE 1C PUMP IN SERVICE AS THE LUBE OIL TEMPERATURE ROSE TO 150 DEGREES FAHRENHEIT, FANB, DEMINERALI2ED WATER, AND ICE WERE USED TO COOL THE EXTERIOR SURFACE OP THE LUBE OIL HEAT EXCHANGER. THE CLEANING AND FLUSHING OF THE LUBE OIL HEAT EXCHANGER FOR THE IB PUMP WAS COMPLETED, AND THE' PUMP.WAS RESTARTED BY 1523. THE 1C PUMP WAS SUBSEQUENTLY SHUT DOWN, AND ITS LUBE OIL HEAT EXCHANGER O WAS CLEANED AND PLUSHED.
| |
| | |
| g__________ _ _
| |
| f BOTH 1UBE[ OIL HEAT EXCHANGERS HAD'BEEN OBSTRUCTED BY. ACCUMULATED SEDIMENT, I.E , MUD AND SILT. .(A FEW CLAM SHELLS'WERE ALSO FOUND RECENT BUT.WERE MINOR CONTRIBUTORS TO THE OBSTRUCTION.)
| |
| SURVEILLANCE TESTS OF-THE SERVICE WATER SYSTEM, INVOLVING SOME UNUSUAL SYS. TEM LINEUPS, HAD APPARENTLY. CREATED FLOW AND PRESSURE =
| |
| (x ; '
| |
| TRANSIENTS THAT LOOSENED THE ACCUMULATED SEDIMENT IN:THE~
| |
| j l
| |
| THIS SEDIMENT PREFERENTI' ALLY RESETTLED INTO THE CHARGING tt. SYSTEM.
| |
| PUMP LUBE OIL. HEAT EXCHANGERS BECAUSE THEY ARE LOCATED AT LOW o- ' POINTS IN.THE SYSTEM. I l
| |
| BOTH TRAINS OF SERVICE WATER ARE SUPPLIED'FROM A. COMMON POND AND I WET PIT AND HAVE SIMILAR PIPING, COMPONENTS, AND FLOW RATES.
| |
| .THEREFORE, THE' PROBABILITY FOR INTRODUCING SEDIMENT FROM THE POND INTO SERVICE WATER COOLED COMPONENTS IS THE SAME FOR BOTH' TRAINS. ALL THREE CHARGING PUMP LUBE OIL COOLERS HAVE CONTINUOUS SERVICE WATER COOLING FLOW EVEN.WHEN THE PUMPS ARE OUT OF.
| |
| SERVICE. THE ONLY INDICATION TO THE OPERATOR OF CLOGGING IN A CHARGING. PUMP LUBE OIL COOLER IS EXCESSIVE THEREFORE, IT ISLUBE OIL TEMPERATURE POSSIBLE-FOR ,
| |
| RISE WHEN THE PUMP IS RUNNING. ;
| |
| CLOGGING TO OCCUR-IN THE LUBE OII. COOLER OF AN OUT-OF-SERVICE j PUMP WITHOUT ANY-INDICATION.TO THE OPERATOR.
| |
| SEDIMENT ACCUMULATION IN THE SERVICE WATER SYSTEM HAS BEEN AN ONGOING PROBLEM AT FARLEY. HOWEVER THE PREVIOUS PROBLEMS HAVE l
| |
| BEEh'MUCH LESS SERIOUS THAN THOSE IN THIS EVENT.
| |
| THE PLANT IS CONSIDERING USING COMPONENT COOLING WATER TO SUPP THESE COOLERS TO ELIMINATE THE POTENTIAL FOR SEDIMENTATION.
| |
| COMMENTS:
| |
| : 1. OPERATION OF THE SERVICE WATER SYSTEM WITH UNUSUAL FLOW CONDITIONS, SUCH AS SELDOM USED PUMP COMBINATIONS OR. VALVE LINEUPS, CAN DISLODGE.AND REDISTRIBUTE ACCUMULATED SEDIMENT. LOW POINTS AND LOW-FLOW AREAS IN.THE SYSTEM ARE PARTICULARLY VULNERABLE'TO RAPID ACCUMULATION OF SEDIMENT.
| |
| THESE AREAS SHOULD BE CHECKED FOR SEDIMENT BUILDUP AFTER SYSTEM TESTS;OR-OTHER UNUSUAL FLOW CONDITIONS.
| |
| : 2. SOME HEAT EXCHANGERS AND PIPING CAN ACCUMULATE LARGE AMOU OF SEDIMENT BEFORE THE CONDITION.BECOMES PERIODIC VISUAL INSPECTIONS,APPARENT FROM PARTICULARLY PERFORMANCE DATA.
| |
| OF SMALL HEAT EXCHANGERS, ARE NECESSARY TO IDENTIFY DEVELOPING SEDIMENT PROBLEMS BEFORE PERFORMANCE RAPIDLY DEGRADES. TRENDING OF SEDIMENT DEPOSITION, FLOW RATES, TEMPERATURE DIFFERENCES, AND OTHER PERTINENT DATA CAN BE A USEFUL TECHNIQUE FOR ANTICIPATING PROBLEMS THAT DEVELO RAPIDLY.
| |
| : 3. PERIODIC USE OF LOW-POINT DRAINb M REMOVE SEDIMENT CAN
| |
| - PREVENT PERFORMANCE DEGRADATION DUE TO CLOGGING.
| |
| IT IS RECOMMENDED THAT PLANT OPERATORS, THE OPERATIONS MANAGER.
| |
| THE TECHNICAL SUPPORT MANAGER, AND THE MAINTENANCE MANAGER BS INCLUDED IN THE DISTRIBUTION OF THIS SER.
| |
| (
| |
| ^ ^ ^ ~ - - - - - - - - - - _ _ _ _ _ . _ _ _ _ _ _
| |
| | |
| m ;_a _ ; ---- - .;, m. _ .. . . m m_. . _ _ . _. , _ .
| |
| BRAIDWOOD SIMLTLATOR MALFUNCTION i
| |
| | |
| ==Title:==
| |
| Primary Water Make-Up Pump Trip ID: CVC-19 g.
| |
| NO: 6.3.4.5.19
| |
| },
| |
| | |
| == Description:==
| |
| Selected W Pump trips due to overload (bad bearing).
| |
| Variations: Date: 3/14/89 Rev: 5 Selectable Steps Inputs Comments
| |
| : 1. Select W Pump 1, 2, 3 1 = OA N Pump 2 = OB W Pump 3 = W Pumps ( L+ h Brief Plant Responses (IC-14, 50%, all systems in automatic) k,e Loss of selected W Pump occurs. The first annunciator received is W Pump TRIP OR AUTO START.
| |
| Suggested Instructor Action:
| |
| i None.
| |
| Events: None O 0645D:4 883M/27 5/89
| |
| | |
| q BRAIDWOOD SIMULATOR MALFUNCTION 1
| |
| I
| |
| | |
| ==Title:==
| |
| CVCS Various Valve Failure ID: CVC-20
| |
| ,- q -
| |
| NO: 6.3.4.5 20
| |
| ]
| |
| | |
| == Description:==
| |
| Selected valve goes to its " fail" position due to a loss of air supply, i
| |
| Variations: Date: 8/14/86 Rev: 3 3 I
| |
| I Selectable Steps Inputs Commenta
| |
| : 1. Select valve 1 - 12 1-LCV 459 (F.C.) 7-C7 8142 (F.C.)
| |
| 2-HV 8149B (F.C.) 8-HCV 123 (F.C.)
| |
| 3-HV 8160 (F.C.) 9-HV 8145 (F.C.)
| |
| 4-TCV 129(F. Bypass)10-FCV 110B (F.C.)
| |
| 5-FCV 111B (F.C.) 11-FCV 111A (F.C.)
| |
| 6-FCV-121 (F.0.) 12-HV 8147 (F.O.)
| |
| r.
| |
| T u
| |
| Brief Plant Response:
| |
| 1-LCV 459 - Causes Icss of letdows, letdown orifice isolation valves close.
| |
| 2-HV 81498 - Letdown orifica valve: loss of letdown.
| |
| 3-HV 8160 - Letdown containment isolation valve: loss of letdown.
| |
| 4-TCV 129 - Temp divert valve bypassing deminst diverts flow around letdown demins And BTRS.
| |
| 5-FCV 111B - Boric acid to VCT inlett no make-up possible to inlet of VCT.
| |
| 6-FCV 121 - Centrifugal charging pumps flow control valve: if centrifugal pumps running, charging flow increases to maximum; pressurizer level increases.
| |
| 7-CV-8142 - No. I seal return bypass closes, high temp alama on RCP lower radial bearing and low flow alarms.
| |
| 0645D:4 883M/28 5/89
| |
| ,__.__1 ._. _ . _ _ . _ _ _ _ _ _ . . . _ _ _ _ . . _ . _
| |
| | |
| L.
| |
| I BRA 1DWOOD SIMULATOR MALFUNCTION l
| |
| \
| |
| | |
| ==Title:==
| |
| CVCS Variors Valve Failure ID: CVC-20 n
| |
| I-i l
| |
| Brief Plant Response (Continued) 8-ECV 123 - Excess letdown flow control: isolates excess letdown system, stopping flow.
| |
| 9-HV 8145 - Auxiliary spray valves _ aux spray flow stops.
| |
| 10-FCV 110B - Boric acid to VCT discharge no make-up possible to outlet of l
| |
| VCT.
| |
| 11-FCV lilA - Make-up water supply to blenders no make-up water available to reactor make-up water system.
| |
| 12-CV-8147 - Loop A alternate charging isolation valve opens.
| |
| Suggested Instructor Action:
| |
| Clear malfunction when repairs are completed.
| |
| Events: None ,
| |
| O 0645D:4 883M/29 5/89
| |
| | |
| BRAIDWOOD SIMULATOR MALFUNCTION
| |
| | |
| ==Title:==
| |
| VCT Level Malfunction ID: CVC-21 i NO: 6.3.4.5.21
| |
| | |
| == Description:==
| |
| Failure of detector LT-112 or LT-185 caused by mechanical malfunction.
| |
| l Date: 3/14/89 Variations:
| |
| Rev: 5 Selectable Steps Inputs Comments
| |
| : 1. Select level transmitter CVC-21A CVC-21A = Level Channel 112 CVC-21B CVC-21B , Level Channel 185
| |
| : 2. Select value 0 - 100 percent
| |
| : 3. Select ramp time 0 - 99,999 sec.
| |
| Brief Plant Response:
| |
| LT-112 - Fails Hiah - at 95 percent a high level alarm sounds.
| |
| - at 73 percent LCV 112A opens to divert to HUT.
| |
| Fails Low - at 37 percent auto make-up comes on; at 20 percest-low level alarm; at 5 percent (and if LT-185 is also below 5 percent) LCV 112 D/E will open and LCV 112 B/C will cSose.
| |
| 0645D:4 883M/30 5/89 I
| |
| -._- ________.m_2 .____m___ _ _ _ _ , _ _ _ _ _ _ _ _ _ _ _ , . _ _ _ . _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
| |
| | |
| om-- -_.%._u..__...._-._.__ . _ _ .. _ . _ _ . _ _ _ __._ _ _ _.
| |
| BRAIDWOOD SIMULATOR MALFUNCTION
| |
| -Title: VCT Level Malfunction ID: CVC-21
| |
| .. \m/
| |
| Brief Plant Response (continued):
| |
| LT 185 - Fails Himh - at 95 percent LCV 112A full divert to HUT and high level alarm.
| |
| Fails Low -- at 20 percent low level alarm.
| |
| - at 5 percent (and if LT-112 is also below 5 percent) LCV-112 D/E will open and LCV-112 B/C will close.
| |
| Note: LCV-112B will auto close only if both level channels are at 5 percent and LCV-112D'is open.
| |
| LCV-112C will auto close only if both level channels are at 5 percent and LCV-112E is open. .
| |
| Suggested Instructor Action:
| |
| l''
| |
| When told to repair the detector, clear the malfunction.
| |
| Events:
| |
| : 1) DVR 06-02-89-002: Level Channel 112 failure l
| |
| O 0645D 4 883M/31 5/89
| |
| ~ . - * * . . . . . . . . . . . . . . _ _ . . _ _ _ . . . . , _ , _ , _
| |
| | |
| ~
| |
| l
| |
| - DEVIATION REPORT Cv c.-v
| |
| :( DVR NO.
| |
| 06 - 02 - 89 - 002 l ,
| |
| STA UNIT YEAR NO. Form Rev 2.0. J l( PART1l TITLE OF DEVIATION VCT LOW LEVEL SUCTION SWITCHOVER OCCURRED 01-05-89 1007 l TO RWST DUE TO FAILED CIRCUIT CARD AND INADE00 ATE PROCEDURE DATE TIME SYSTEM AFFECTED PLANT STATUS AT TIME OF EVENT TESTING g g g7 CV H0DE 1 POWER (%) 31% WORK REQUEST t*L i X l l l DESCRIPTION OF EVENT -
| |
| Investigations into a Rx, makeup controller NWR were in progress that disconnected the output of VCT 1evel transmitter 112. When these leads were lif ted the centrifugal charging pump suction valve realigned to the RWST which should only happen if both VCT level channels 112 and 185 indicated a low level. Further investigations revealed that the 185 channel was already in the low level tripped r.,ndition due to a defective circuit card. All indications on the 185 channel read normal VCT level. ,
| |
| NO POTENTIALLY SIGNIFICANT EVENT PER NSD DIRECTIVE A-07 g YES g,g ___
| |
| 10CFR50.72 NRC RF.D PHONE l l NOTIFICATION MADE l l l X l DATE TIME RESPONSIBLE SUPERVISOR PART2l OPERATING ENGINEER'S COPMENTS There is no indication of coincidence for this function. Valves were immediately realigned.
| |
| iJ g
| |
| NON REPORTABLE EVENT NOTIFICATION 30 DAY REPORTABLE /10CFR REGION III DATE TIME g g
| |
| " ~
| |
| 5 DAY REPORT PER 10CFR21 I l l NSD DATE TIME g g ANNUAL /SPECIAL REPORT REQUIRED A.I.R. #
| |
| TELECOPY
| |
| (,g,,, ,
| |
| CECO CORPORATE OFFICER DATE TIME PRELIMINARY REPORT COMPLETED AND REVIEWED D. Brindle 01-06-89 ,
| |
| s OPEkATING ENGINEER DATE INVESTIGATION REPORT & RESOLUTION 1} M[ b~M ' j ACCEPTED BY STATION REVIEW f ,
| |
| hY$4/f1 R UTION 2 )
| |
| STATION MANAGER DATE 86-5176 (Form 15-52-1) 11-20-85 DOCUMENT ID U
| |
| (0237R/0029R)
| |
| I l
| |
| ee me-s -e one w-ew = * * '=''
| |
| e * *
| |
| * p ruea,- e -----. meme e - .e en a +ma.esae ge =s
| |
| -.-_--__m
| |
| | |
| BRAIDWOOD SIMULATOR MALFUNCTION
| |
| | |
| ==Title:==
| |
| Make-Up Control Failure ID: CVC-23
| |
| /~T _ NO: 6.3.4.5.23
| |
| '. (]
| |
| | |
| == Description:==
| |
| Failure of make-up control system in any mode due to various mechanical switch problems.
| |
| Variations: Date 8/14/86 Rev 3 Selectable Steps Inputs Comments
| |
| : 1. Select mode 1. Auto ,
| |
| : 2. Dilute
| |
| : 3. Alt dilute
| |
| : 4. Borate
| |
| : 5. Manual
| |
| %)
| |
| Brief Plant Responses Mode failures:
| |
| : 1. Automatic - Failure of make-up control system to initiate make-up at the selected boric acid concentration during normal operations. Could lead to I low VCT level. If make-up is added without boric acid, dilution would take place. If make-up is added with too much boric acid, negative reactivity would be added to the core.
| |
| : 2. Dilution - Failure to dilute would result in no change in RCS boron concentration or if too much dilution, positive reactivity addition to the core is possible.
| |
| : 3. Alternate dilution - Same as 2.
| |
| 4 Borate - Cannot borate the RCS. Student must emergency borate.
| |
| : 5. Manual - Failure of manual mode could lead to any condition 1-4 above.
| |
| l l
| |
| 3645D:4 883M/34 5/89 l
| |
| i j
| |
| .- -- _ ..- . . _ _ _ _ . . - .__ ._ j j
| |
| | |
| ' BRAIDWOOD SIMULATOR MALFUNCTION-
| |
| )
| |
| | |
| ==Title:==
| |
| Make-Up Control Failure . ID:' CVC-23 Suggested Instructor Action:
| |
| Insure operator takes appropriate action.
| |
| Events: None l
| |
| l
| |
| \
| |
| 0645D:4 883M/35 5/89 te i egg g , -gh. g 9,9 ,, ,,gy, ,,
| |
| ' M %@6 WeSO1W M M M.h4ee
| |
| | |
| BRAIDWOOD SIMULATOR MALFUNCTION i
| |
| | |
| ==Title:==
| |
| CVCS Domineralizers Depletion ID: CVC-26.
| |
| NO: 6.3.4.5.26
| |
| ::(\s:
| |
| | |
| == Description:==
| |
| ' Selected. demineralized is depleted. .
| |
| Variations: Date: 8/14/86 Re?: 2 Selectable 5 tops Inputa Comments
| |
| : 1. Select domineralizer 1-3 1 = "A" Mixed bed 2 = "B" Mixed bed 3 = Cation domin
| |
| : 2. 3 elect percent deplaced 1-5 1 = 20 percent 2 = 40 percent 3 = 60 percent 4 = 80 percent
| |
| '(_/ 5 = 100 percent
| |
| : 3. Select delay time
| |
| , , Brief Plant Response:
| |
| l Failure of Demins (1 -3) should only be detected in the control room by a high rr.diation alarm in the reactor coolant filter or sample taken by Rad. Chem.
| |
| 1 Suggested Instructor Action:
| |
| None.
| |
| Events: None O 0645D:4 883M/38 5/89 {
| |
| | |
| _ . . . . . _ . . ..e._ , ____ . _ . - _ . . _ _ _ _ _ . _ . _ _ - . . . _ . _ . . _ _ . _ _ _
| |
| BRAIDWOOD SIMULATOR MALFUNCTION
| |
| | |
| ==Title:==
| |
| Letdown Leak Outside Containment ID: 'CVC-27 NO: 6.3.4.5.27 Q
| |
| 1 i Descriptions Letdown Leak outside containment at the outlet of "A" Letdown l Heat Exchanger.
| |
| l Variations: None Date: 8/14/86 Rev 1 Selectable Steps Inputs Comments
| |
| : 1. Select leak size 0 - 200 spa Flow based on nozinal operating pressure of 2235 Psig.
| |
| : 2. Select ramp time 0 - 99,999 sec l,
| |
| Brief Plant Response Slow decrease in pressurizer level with indications of increased letdown flow but indicated letdown flow drops to zero.
| |
| Suggested Instructor Action: When requested, isolate "A" L/D HK and place "B" L/D HK in service.
| |
| Events: None 0645D:4 883M/39 5/89
| |
| | |
| 1 BRAIDWOOD SIMULATOR MALFUNCTION-e
| |
| | |
| ==Title:==
| |
| ' ' seal Injection Leak ID: CVC-28'-
| |
| NO: 6.3.4.5.28 f t .
| |
| | |
| == Description:==
| |
| | |
| ~ Leak in' individual seal injection lines inside containment or-i leak in common seal injection line outside containment.
| |
| - Variations: None Date: 4/21/88 Rev 2 Selectable Steps Inputs Comments-
| |
| : 1. -Select option .CVC-28A 1. Leak inside containment on RCP A seal injection line.
| |
| CVC-285- 2. Leak inside containment on RCP B seal injection line.
| |
| CVC-28C 3. Leak inside containment on RCP C seal injection line.
| |
| CVC-28D 4. Leak inside containment on RCP D seal injection line.
| |
| ,CVC-285 5. Leak outside .
| |
| containment in common injection line.
| |
| O
| |
| : 2. Select leak size 0-100 percent
| |
| : 3. Select ramp time 0-99,999 see Brief Plant Response: (IC-17, 1001, all systems in automatic)
| |
| For selections A-D (leak is in ennt. prior to the missile barrier), seal injection flow to the specified ptmp increases. Charging flow also increases and containment radiation increases.
| |
| For selection E (leak is downstream of seal injection filters), seal injection o flow to all RCP's decrease and charging flow increases while pressurizer level
| |
| - decreases.
| |
| The first annunciators received are RCP SEAL WTR INJ FLTR AP HIGH and RCP SEAL WTR INJ FLOW LOW.
| |
| Suggested Instructor Action: None Events: 1) DVR 06-02-88-028: Seal Injection Filter Leak O 0645D:4 883M/40 5/89
| |
| --a_,-,-.-_.,__m ---- - - -- - - _ _ - -
| |
| | |
| DEVIATION INVESTIGATION REPORT .
| |
| TITLE PAGE 28 SEAL INHCTION FILTER LE AK DUE TO FAILED 0 DING INDUCED BY FOREIGN MATIQQL l i
| |
| /ENT DATE DIR NUMBER REPORT DATE l
| |
| ^
| |
| / SEQUENTIAL // REVISION j
| |
| MONTH day VFdP.
| |
| a STA UNIT YEAR NUMEER / NUMBER MONTH DAY YEAR 1 )
| |
| POWER 0, 4 1, 2 8, 8 LEVEL of 2 21 7 81 8 01 6 of 2 81 8 -
| |
| 0 l ?! 8 010 l 11 6 CONTACT FOR THIS DIR NAME TELEPHONE NUMBER AREA CODE {
| |
| Qsle St. Clair. Asst. Scot. Work Plannine Ext. 2888 8!1 l 5 2 l1 l4 I-1S l4 14 l 1 COMPLETE ONE LINE FOR EfrH COMPONE URE DESCRIBED IN THIS REPORT l CAUSE SYSTEM COMPONENT MANUFAC- REPORTABLE CAUSE SYSTEM COMPONENT MANUFAC- REPORTABLE TURER TO NPROS TURER TO NPRDS I 1 I I I i l I I i 1 I I I I I I I I I I I I l.) 1 I SUPPLEMENTAL REPORT EXPECTED MONTH DAY YEAR p
| |
| SUBMISSION I YES fif ves. comolete EXPECTED SUBMISSION DATE) Xl NO TEXT A. PLANT CONDITIONS PRIOR TO EVENT:
| |
| IN Event Oate/ Time 2/26/88 / 2000 v
| |
| )
| |
| Unit 1 MODE 1 - Power Deerations Rx Power 08 RCS [AB) Temperature / Pressure Nermal Oeeratino Unit 2 MODE 1 fewer Ocerations Rx Power 16 RCS [AB) Temperature / Pressure . Normal Oceratino.
| |
| B. DESCRIPTION OF EVENT:
| |
| On 2/26/PS. at 2000, the 2A Seal Injection Filter (2CV01FA) was taken off-line due to a high OP of 23#
| |
| during a power decrease from approximately 95% power to 16% power on Unit 2. The 22 Seal Injection Filter (2CV01FB) was valved in to maintain seal injection flow. At 2140 the 120 gpm letdown flow was isolated due to low pressurizer level. An Equipment Attendant (EA) was sent to look for possible leakage in the vicinity of the 2B filter. The EA surinoned a Shif t Foreman (SF) and both of them identified a spraying sound beneath the 2B filter plug (401') and leakage was observed coming through a plugged vent / drain pipe penetration for the 2B filter (383'). At 2730 the 28 filter was taken off-line and administrative 1y removed from service. The 2A filter had not yet been changed and <as put back on-line with a measured 308 DP. Mechanical Maintenance personnel attritNted the 2B filter leakage to a split 0-ring on the filter cartridge. A new filter cartridge containing a new 0-ring was installed under NWR B53455. No OVR was written for this initial 0-ring f ailure.
| |
| At 0143. on 2/27/88. the 2B filter out-of-service was temporarily lif ted and the 2B filter put back on-line while the 2A filter was taken off-line for filter cartridge replacement. At 2039 on 2/27/88 alarms annunciated for seal injection low flow and pressurizer level was again observed to be decreasing. The 2B seal injection filter was suspected and EA's were dispatched to the vicinity of the 28 filter and valves.
| |
| Leakage was again observed coming through the same plugged vent / drain pipe penetration for the 2B filter (383'). At 2102 the 2B filter was isolated and the 2A filter was placed back on-line. Seal injection low j flow alarms cleared as seal injection flow returned to normal and stable plant conditions were resumed.
| |
| I I
| |
| (1978M/0221M)
| |
| | |
| r I
| |
| DEVIATION INVESTIGATION REPORT TEXT CONTINUATION TITLE DIR NUMEER PAGE SEQUENTIAL REVISION
| |
| ( SEAL IN.1ECT10N FILTER LEAK DUE TO FAI;rD STA UNIT YEAR NUMBER NUMBER
| |
| ( ,,hG' INDUCED BY FOREIGN MATERIAL .
| |
| g g gg -
| |
| Ol2l9 -
| |
| O l0 2 0F 0 l3 TEXT
| |
| ' ;B. DESCRIPTION OF EVENT: (Continued)
| |
| ~
| |
| i Mechanical Matutenance personnel inspected the .B filter housing sealing surfaces per directives specified {
| |
| in the filter cartridge replacement procedure BMP 3100-10 Rev. 3. 10/5/87 which had been revised to add prerequisites and steps for verifying that dirt problems and seal area damage both in the filter and the g;neral' vicinity are identified and corrected. Another split 0-ring was found and the licensed Senior Reactor Operator (SRO) initiated this DVR.
| |
| The filter housing inspection did not identify any obvious mechanism for 0-ring f ailure; however. in the more inaccessible lower portion of the filter housing. a piece of hard plastic (similar to model plastic) approximately 1/16" x 1" x 2" was found insise the housing. Personnel involved with the removal of the filter cartridge and plastic did not know the origin of the plastic or how tt got inside the filter h:using. The plastic and cartridge were disposed of as radwaste. A new 28 filter cartridge was installed. On 3/1/88 the 2B filter was pressurized and Tech Staff jersonnel performed a visual leakage examination through the 28 filter boroscopic floor plug opening and no leakage was observed. This exam completed testing for 28 filter NWR's B54355 and 853226. The 2B filter had been replaced on 2/29/88 due to high DP and had been restored with no apparent leakage problems. No systems were made inoperable during the sequence of 2A and 28 filter changeouts and stable plant conditions had been obtained at 2102 on 2/27/88.
| |
| N CAlfSE OF EVENT:
| |
| The root cause of the failed 2B 0-ring is postulated to be from the piece of plastic cutting the 0-ring.
| |
| Th3 most probable source of the plastic is that it entered the filter housing during the previous 2B filter 1 changeout on 2/20/88 (NWR 853226). This is considered an isolated event.
| |
| D. SAFETY ANALYSIS: ,
| |
| 4 o
| |
| Th1re were no adverse safety consequences associated with this event because it was an isolated event in a dual train system. Under a more severe set of initial conditions where the plastic left the filter housing and entered the RCP or where both filter 0-rings failed simultaneously abnormal operating procedures 2 BOA RCP-1. RCP Seal Failure and/or 2 BOA RCP-2. Loss of Seal Injection, would be implemented as required.
| |
| E. CORDECTIVE ACTIONS:
| |
| Between July 1984 and February 1988. the four CV seal injection filters at Byron Units 1 and 2 have been i changed a total of 88 times. Eighty seven (87) of these replacements were due to normal filter changeout in response to high DP. One of these was replaced due to a crimped 0-ring (7/10/85) and no DVR had been written at that time. Since this is the first occurrence of a split seal injection filter 0-ring at Byron, and existing procedures contain steps to preclude such failure, no further corrective action is deemed nicessary at this time.
| |
| O I I
| |
| (1978M/0221M)
| |
| | |
| .Ll DEVIATICNIhVESTIGATION'REPORTTEXTCCNTINUATION DIR NUMBER PAGE
| |
| [ TITLE; SEQUENTIAL REVISION STA UNIT YEAR NUMBER NUMBER EAL' INJECTION. FILTER LEAK DUE TO FAILED f '"hG INDUCED BY FOREIGN MATERIAL0 16 0 12 a la -
| |
| 0 l 2 l' R -
| |
| 0 10 1 0F 0 li
| |
| ? EXT- ,
| |
| F. PREVIOUS OCCURRENCES:
| |
| There have been no previous occurrences documented by DVR's.
| |
| DVR NUMBER IIILE NONE G. COMPONENT FAILURE DATA:
| |
| MODEL NUMBER MFG PART NUMj[1 a). MANUFACTURER NOMENCLATURE
| |
| -052 l Pall-Trinity (P050) ' Filter SEHD10602-22SEC32 SESC10670-
| |
| ' b) RESULTS OF NPRDS SEARCH!
| |
| l Three previous similar occurrences of failed 0-rings were found at the following locations:
| |
| 2/25/85 Davis-Beese 1
| |
| .p' 7/10/85 Byron 1 V.C. Sumer 1 5/23/86 The broken or split 0-ring failures at Davis-Besse 1 and V.C. Sumner I were attributed to wear and aging of the 0-ring. The crimped 0-ring failure at Byron could not be specifica11y' identified and was attributed to either plant operation or filter cartridge installation. The current filter l
| |
| cartridge procedure BMP 3100-10, Rev. 2 6/5/83, in use at the time, did not contain any precautions or steps to assure a clean area and adequate sealing surface.
| |
| l RESULTS OF NWR SEARCH I
| |
| . C)
| |
| See explanation in Corrective Action, Section E.
| |
| I
| |
| .I 1
| |
| l a 1 Q a
| |
| {
| |
| 1 (1578M/022 1)
| |
| I. . __---__ -.- _--
| |
| | |
| 7~-
| |
| BRAIDWOOD SIMULATOR MALFUNCTION LISTING
| |
| .~ j' ELECTRICAL POWER SYSTEMS LJ EPS-1 Loss of Off Site Power EPS-2 ' Main Generator Voltage Regulator Failure EPS-3 345 KV Switchyard Bus Trip EPS-4 Loss of Non ESF Bus EPS-5 Loss of ESF Bus EPS-6 Diesel Generator Failure EPS-7 Loss of D.C. Distribution Bus EPS-8 Loss of 120 VAC ESF Instrument Bus EPS-9 Load Rejection EPS-10 Loss of 6.9 KV Bus EPS-11 Loss of Feed to 480 Volt ESF Bus or MCC
| |
| ,y
| |
| () EPS-13 Failure of System Aux Transformer (SAT)
| |
| EPS-14 Failure of Unit Aux Transformer (UAT)
| |
| EPS-17 120 VAC Instrument Bus Inverter Failure EPS-18 Main Generator Exciter Failure EPS-20 Loss of SX Cooling to Diesel Generator 1
| |
| EPS-21' Safeguard Shutdown Relay Failure EPS-22 Diesel Generator Failure to Flash Generator Field
| |
| ~
| |
| f
| |
| '\
| |
| o 638M/263M/6 8/87
| |
| | |
| BRAIDWOOD SIMULATOR MALFUNCTION J'[
| |
| | |
| ==Title:==
| |
| Loss of Off Site Power ID: EPS-1 NO: 6.3.4.6.1 !
| |
| t, v).
| |
| | |
| == Description:==
| |
| ' A sequence of breakers tripping due to weather conditions, causing a loss of site power.
| |
| Variations: See event 1 for variation in Loss of Date: 4/21/88 Offsite Power, LER-06-02-87-019. Rev: 5 Selectable Steps Inputs Comments None Brief Plant Responses (IC-17, 100%, all systems in automatic):
| |
| I
| |
| 'b The cause of this malfunction is a series of events, all of which can be attributed to the weather.
| |
| Note: This malfunction will take approx. 10 minutes to initiate.
| |
| This time span has been programmed to add realism to the fault condition. ,
| |
| : 1. LaSalle Bkrs. 1-3 and 1-8 trip.
| |
| Note: This failure could have been caused by e line fault. The first annunciators received are CB 1-3 TRIP, ACB 1-8 TEIP and LINE 0104 TRIP.
| |
| : 2. After about five minutes, all power to both units from the E.
| |
| Frankfort lines are lost and their transmission lines isolated I in the switchyard (breakers 3-4, 9-15 and 14-15). )
| |
| l I
| |
| i
| |
| | |
| ^
| |
| BRAIDWOOD SIMULATOR MALFUNCTION
| |
| | |
| ==Title:==
| |
| Loss of. Off. Sits ' Power ID: ..EPS-1.
| |
| f Brief Plant' Response' (continued):
| |
| Note: This failure could have been caused by a plane running into the. transmission'line towers, a tornado crossing the transmission lines, lightning striking them, or icing and high.
| |
| v0 winds during winter, causing them to " arc over" and. trip.
| |
| '3.- Approximately five minutes later, the line from Davis Creek isolate
| |
| -(break'rs e 4-7 and 7-g open), causing a loss of generator load (excluding in-house loads) and a trip of the turbine.
| |
| Note: The reason for this fault. could also be any of those listed for the failure of the other transmission lines.
| |
| When the turbine trips, this will cause a loss of power to the service buses.- A reactor trip could be generated by loss of flow (RCP's trip),. turbine trip-reactor trip,.or pressurizer high pressure.
| |
| When this occurs, a loss of site power signal will be generated,-
| |
| stripping the ESF buses, starting the diesel generators, re-energizing the ESF buses, and sequencing the loads onto the diesels. RCS temperature increases and is controlled by the lifting of the steam generator reliefs.--Auxiliary feedwater system will supply make-up water to the steam generators. Natural circulation flow will i J
| |
| establish'in the RCS, cooling the core.
| |
| Suggested Instruction Action: ..
| |
| ~
| |
| Clear the malfunction prior to sttempting the electrical system recovery. Use i LOA EPS-1, 67 and 80 to allow AC restoration.
| |
| Events:
| |
| : 1) LER 06-02-87-019: Loss of Offsite Power (Use LOA EPS-65 to open U-1 SAT disconnect).
| |
| O 881M/3 5/89
| |
| ~ ~ ' ' '
| |
| | |
| .: . -a. . .. ~. x = . . :.:
| |
| , =. w =;
| |
| . .. -.: ? : -- c------------
| |
| BRAIDWOOD SIMULATOR MALFUNCTION
| |
| : n. i 7
| |
| | |
| ==Title:==
| |
| Main Generator Voltage Risulator Failure ID: EPS-2
| |
| , )$ ' NO: 6.3.4.6.2
| |
| . Description ' Failure of the'AC (automatic) regulator, -
| |
| -causing a change in the generator's 1 voltage output.. Manual control possible in manual mode (base adjuster). j f
| |
| Variations:. Date: 4/21/83 .
| |
| Rev: 5 l
| |
| Selectable Steps Inpute Cossents i
| |
| : 1. Select percent of 0-200 percent 100 percent = normal output' !
| |
| voltage output change voltage
| |
| : 2. Select ramp time 0-99,999 sec.
| |
| 4 Brief Plant Responses v
| |
| When the failure occurs, the automatic itaulator will cause the exciter field current and output voltage to change, afracting the generator output voltage. Depending on the-direction of the failure, the generator's MVAR load will .hange in proportion to the amount of under or over excitation.
| |
| Various protection circuits can be actuated, possibly tripping the generator output breakers and/or exciter field breaker. Possible annunciators received due.to this malfunction include GENERATOR V')LT REG TRIP, MAIN XIMR OVER EXCITATION, and GEN VOLT REG TROUBLE.
| |
| Suggested Instructor Action:
| |
| None.
| |
| Events:
| |
| : 1) LER 20-1-87-052: Main Power Transformer Overexcitation Relay Actuation (Set JGENVH1=T) 881M/4 5/89
| |
| | |
| EM -l LICENSEE Evimf REP 0af (Ltt)
| |
| Facility same (1) 7 N 0:cket Kumeer (2) Paam t u avran unit 2> al il al al al al 11 1 t lmrlsl Title (4) g nei rfm TRILEtapfii!-F STEAM GEeERATOR LEVEL & SUSSE00ENT Lost 0F 0FFSITE PtadER AS A RESULT
| |
| () tvant nata fit year tra - -- tai ammart aate f71 other Factittima invalvan rat l Month Day Year #1 Seguentt41 /// Revisten Month Oay Year rae111ty m-e i nackat m m rs1 I
| |
| //,f , , -
| |
| {f/, m, -
| |
| l
| |
| % an=r el 11 al al al I i 11 a al 2 a 7 al 7 al 1 Ie aIa 1i0 3Oi 817 al si al al al i I OPERATING \ THIS REPORT IS SUBMITTED PUGUANT TO THE REQUIREMENTS OF 10CFR rch.,a, ans er mara aF ther Fc11aminal f 111 MODE (0) % '
| |
| ... w - , , , , 20. ass (c) L. 50.73(a)(2)(tv) ,, 73.71(b)
| |
| PohdER 20.at$(a)(1)(1) ,,, 50.34(c)(1) ,,. $4.73(a)(2)(v) 73.71(c) !
| |
| LEVEL 20.405(a)(1)(11) 58.34(c)(3) 54.73(a)(2)(vit) Other (Specify
| |
| !1 !1 .,, _ I ftal 20.40$(a)(1)(111) , 50.73(a)(2)(1) ,,,, 50.73(a)(2)(rttt)(A) in Abstract M ,
| |
| 20.405(a)(1)(iv) 20.at$(a)(1)(v) 54.73(a)(2)(11) 54.73(a)(2)(111) tffruift contiff Fim Twit Lia f121 54.73(a)(2)(vitt)(B)
| |
| St.73(a)(2)(s) below and in Test) m fate - stessa ASEA CINE 3
| |
| : 7. tamanat ar . Aantatame fee _tateal tenff *"- Ntaar Ent. 22aa Al 1 lE 21 11 al -l El al etBSLEff int Lfur Fim emu flusement Parties artestata in tutt Rtstaf f111 CAUSE SYSTEM Cips'enENT ManUFAC-7888.
| |
| REPORTABLE To meGat f//
| |
| /
| |
| f CAUSE SYSTEM l ClpegmEnf ManuPAC-fumes rep 0RTABLE te meget {/
| |
| I I i l I i I / f I I I I I I I I i l I I i i i I '/ l 1 l l l t I I supptrufufAL REeGRT ElptfftB f ?al (Spectog taanth l Day I Yea O ~lynn t f r vna. -tata ruptffra ma==restam natti YIma Sutatssion Date (18) , lg l, ASSTRACT (Liett to 1a00 spaces.1.e. apprestestely f tfteen single-space typewritten Itnes) (16)
| |
| On October 2.1987. at 0446. Unit I was returning to service, naten Unit 2 was synchrentand to the grid.
| |
| the Steam Generator (SG) levels increased and caused a NI-2 S/6 Level Trtp. The Mt-2 $4 level was reached en SE 2C ese to escessive " leak by* of the 2ft434 valve. The high S/S level caused a tureine trip and a subsequent reacter trip because reacter gemer was ateve 191. An equipment operator (EO) was instructed to realign the switchyard ring aus after the trip. The E0 opened the System Aus Transformer disconnects instead of the main power transformer itscannects. The safety related aKY tuses were doenerStaed causing l the emergency diesel generators to start, reenergtas the buses, and seguence the safe shutdeun leads. The rest cause of the less of offstte peuer was due to personnel error. The E0 opened the wrong discoanect.
| |
| The corrective actions are as fellems:
| |
| * Otsciplinary aatten was taken with the EO.
| |
| me switchyard operatten will te performed without a second tridividual present.
| |
| Temperary latels were placed on MPT & SAT switchyard dtscene.ects, untch ut11 eventually became permanent
| |
| * A walk through of the switchyard with Otvtsten Superintendent of Power Supply to demonstrate proper operettens and ceumantcations was conducted.
| |
| e The SAT disconnects are locked with unique 1ecks for each untt.
| |
| e A checklist has been developed that fernalizes checks te be made prior te disconnect operation.
| |
| Annual high voltage switching regualtftcatten will include actual switching operattens.
| |
| * Modificattens were tosta11ed in 1985 that reduced the undervettage trip setpoint en all process and area radtatten sentters fres 100 VAC to 30 VAC. Past expertence indicates that the setpeint modificatten has reduced the mentter's sonsttivity te voltage transtents caused by large pung starts j O o and most grid disturbances. No further corrective actten will be taken since this is constdered an ts01sted occurrence, Mata Feesmater Regulating Valve 27Wl30 was repatred et teinating the " leak by* prehlen.
| |
| l 1
| |
| (issm/01,0M)
| |
| | |
| LfffN1ff fVfkT #ffdRT fLER) ffrf cantiquatics FaCIt!TT mant (1) . DOCKET NUMBER (2) Lft Nupart fs) sue g39 Year /,/p/ ' Sequential Revision j/// u ~ --
| |
| ,///
| |
| /// an. - e A arram_ tan t t 2 o i s I e I o I o i 41 si s el7 -
| |
| oIiIe - oIo of 2 or of Q YEXT Energy Industry teentification System (E!!S) codes are teentified in the test as (um]
| |
| A. PLAAT Comofff0Ns PRIDE To fvtuT:
| |
| Event Date/ Time lo/2/e7 / osaa Unit 2 M00E 1 -
| |
| Pa==* naarat tnan Rx Power 12L., RCS (A8] Teneerature/Prassure marnat naarattne
| |
| : 8. atletIPTION OF EVENT:
| |
| On Octeter 2.1987, at Saad. Unit 2 was returntng to service following a seacter Trip which ec:urred en Octoser 1. An Equipment Operator (50) and a shift Forman (SF) went out to the switchyard te verify that l i
| |
| 4113 phases of Air Ctreutt treaker (nCS) 10-11 and 011 Ctreutt Breaker (OCO) 11-12 were opened, tsolating the meta generater in preparatten for synchrentaatten. In additten, the Main Power Transformer (MPT) disconnect for Unit 2 was to be closed by the EG. After completing these mentpulattens the E0 and the Shift Foruman sent to the relay house and completed the restred leg book entries.
| |
| In the Matn Centrol Rees (MCR). one Reactor Operater was maintaining staae generater levels with the 2C Stese Generator in manual level control. A second Reactor Operator was preparing to synchrentae the generator to the grid.
| |
| As the Unit was synchrentaed the steam generator levels started te quickly increase. The three steam generators that were in auteestic level control statt11aed at appreatmately 64 percent. The 2C Steam Generater level, which mes in manual. continued to increase. The Operator closed the IPWB39C valve and opened the Startup Fee heter Pump Recirculatten valve to try to reduce the water supplied to the 2C Steam Generater. tefore these valves cwid fully stroke the steam generator reached P-14. the Mi-2 Turtine Trip Setpoint. Stace the Unit was gre)ter than le percent power. P-7. a Reacter Trip occurred at 0444.
| |
| The Emergency Procedures were entered and properly performed. Per these precedures, the Et and the Shift foreman were called en the radle te verify that all 3 phases of aC810-11 and OCS 11-12 were opened. The nest res tred step was to realign the ring tus. The EO was then called en the radle to open the Unit 2 345KV Meta Pomer Transformer disconnect. Throughout all the radte cessamicattens the E0 repeated back all eessages te verify proper and cespirM caemmicatten. The EO understood the instructions. Both the E0 and the Shif t Foremen approached the 16atten Auxiliary Transformer (SAT) manual disconnect unen the Shift Foremen was paged. He returned to the relay house. The Egipment Operator continued on to the 547 maneal disconnect and opened it at g681. The E0 mistakenly opened the wrong transformer disconnect caustng the less of effstte power.
| |
| The Shtft Foremen heard sonst sounded Itke an emplesien and went out to where the EO was standing. The Eeutpment Operator mes unharmed.
| |
| In *.he Main Centrol Rees, the Itphts went out the Unit 2 Stesel Generators automatically started and the safe shutdows loads semenced en the diesel generators. The $20 entered Byron Operating Annormal Procedure (80A) Elec-4. Loss of Offsite Power For Mode 3 or 4 All steps were properly performed. The SRO in the Centrol Room used stP 0.2. Natural Circulation Coolsewn, to cael down the primsry loop.
| |
| Due to the less of offstte power at 0581. a poner fluctuation was felt en unit 1. Both the IRT-ARO11 and IRT-ARe12 radiatten monitors went into alarm cenettien, causing a Contattement ventilatten Isolatten en unit
| |
| : 1. In additten, the Unit 1 Process Coguter and the Unit 1 DEM computer stopped operating.
| |
| The l%tt 1 Process Computer was restored te normat operation at 07a8. However, the seetnistrative
| |
| ( j restrument to report s less of Untt I emergency assessment capattlity within I hour of a 2 hour loss of (theProcessCamputerwasnotmet.
| |
| 1 (iss7M/siteM)
| |
| | |
| [
| |
| tfffh1Ef fVfhY Rfr0RT ftER) ffri CONTINU4tt0N FACILITY maME (1) DOCKET CUMBER (2) tra Ntmera t an paa. fn Year 5:euent1al g/j Revis1on
| |
| = ~ - fff ug avean. ur>tt 2 a I a i e I o 1 e i al sf a 'a f 7 .
| |
| ei1 ie' -
| |
| aIe ah or a1 g TEXT'
| |
| \ Energy Industry Identif.tcatten System (1115) codes are toontif tes in the test a; (su)
| |
| Due to recovery actions already in progress, action a of Limiting Condittens for Operation (LCO) 3.4.1.1 was not complied within the I hour specified.
| |
| * normal and Alternate Offstte AC Power Availattttty*
| |
| Surveillance 1/2005 8.1.1.1.a-1 were performed initially at 0721. uhen they should have been completed by 0601 An unusual'twent was declared at 0538. An examinatten of the Unit 2 SAT disconnect showed ne significant damage to the etsconnect. The Unit 2 SAT disconnect was reclosed, the Unit 2 SAT was reenerglaed at 1323 and the unusual twent was terminated at 1417.
| |
| C. cAMEE GF EVERT:
| |
| The rest cause of the Less of Offstte Power was due to a personnel error by the Egipmunt Operator eme opened the wrong transformer disconnect.
| |
| A contributing facter was the confusing disconnect labeltng. The MPT disconnect was labeled *Ta2 3aSKV TR O!SC'. The SAT disconnect was labeled *TR242 345KV message from the lead dispatcher and the center desk operator was to open the Mats Power Transformer 34SKV disconnect.
| |
| The lateltng en the disconnects was not censistent with the nomenclature used to identify the discannec&.
| |
| The rest cause of the Reactor Trip was the " leak by* of the Mati feeheter Regulating Valve 2FWE30 for the 2C Steam Generater. ,
| |
| The *1eak by* of this valve allemed oncess water into the 2C Steam Generater causing !
| |
| the Nigh-2 Steen Generator level candition.
| |
| The rest cause of the Unit 1 Centalmeent ventilatten Iseletten was the weltage transtant felt by the IRT-Amell and 187-4A412 radiatten annitors caused by the Unit 2 less of offstte pomer. noten gemer to the radiatten monitors is interrupted, an alare cenettien is generated. This alare generates a Containment Venttlatien Iselatien signal.
| |
| The rest cause of the less of the Unit 1 Process famouter was the weltage transient caused by the Unit 2 less of offsite power. The Unit 1 Process Conguter is currently se *dtriy* power. Its normel inverter peuer supply is inepe. 11e awetting replacement parts. Caspletten is being tracked on the daily plant {i status meetings.
| |
| O. 1&FETY am&LY111:
| |
| There was no impact on the plant er puelle safety. The Unit 2 Otesel Generators teth started as designed.
| |
| All safe shutdemn leads more sogenced en the huses supplied by the diesel generators as designed. The Unit was cooled dome en natural circulation. The operators properly vertfted all natural circulatten coeldown parameters per precedure.
| |
| E. CORRECTIVE Actimit The corrective actlens are as fellows:
| |
| * Olsetplinary action was taken with the (0, e
| |
| es switchyard operation will he performed without a second Individual present.
| |
| Temporary labels were placed en wf & SAT switchyard disconnects. unich wt11 evalually become I periesnent A usik through of the switchyard with Otvisten Superintendent of Pomer Supply te demonstrate proeer operattens and communications was conducted.
| |
| The SAT disconnects are locked with unique locks for each untt.
| |
| (
| |
| * A checklist has been developed that ferime113es checks to De mees prier te disconnect operation.
| |
| Annual high voltage switchtng regualificatten will include actual switching operattens.
| |
| l ossavoipeni _.
| |
| _ _ - _ _ - - _ - _ _ _ - _ l
| |
| | |
| titENstt rytwr streef ftrat irrf ccuttwuatics FACILITY caME (1) DOCKET NUMBER (2) tri uuMatt ist _
| |
| saae f it Year 5eguential j///
| |
| llj jj as, Revision ps e
| |
| ['
| |
| \
| |
| avran. unit 2 a i E l e I e I e I al si s ai7 -
| |
| ait i e -
| |
| o!e at 4 or oi
| |
| . TEXT Energy Industry Identificatten System (t!!s) codes are toontified in the text as (mm)
| |
| Modifications were installed in 1905 tnat reduced the underveltage trip setpotnt en all process and area radiation mentters fren 100 VAC to 9e VaC. Past expertence indicates that the setpoint modification has_ reduces the mentter's sonsttivity te weltage transients caused by large pump starts and most grid disturbances. No further corrective action will be taken since this is censidered an isolated occurrence.
| |
| e Main Feedwater Regulating Valve 2FWE30 was repaired. eliminating the " leak by* pretten.
| |
| locause of the delay in reporting the less of the Unit i Process Computer, the method of tracktag -
| |
| degraded eeutament status during emergency plant events will be reytened.
| |
| No corrective actten will he perferned as a result of the delayed performance of Surveillance 1/1905 8.1.1.1.a-1. In conjunction with the recovery actions and reporting activtttes, the sw. vet 11ance was perferund as seen as pesstkle.
| |
| F. M tylout atttmatK ts:
| |
| Ltf blaatt I,gkg most G. CtBR MEET FAILimi RATA:
| |
| a) MAENAffLEER utBEKLAT1EE peMfL Insett DWS PaAT EseER met Applicatie b) EfilE 71 GF RPRAS 1EAtcM!
| |
| O ~ ~ "'''''
| |
| n Vl -
| |
| (1647M/019dM)
| |
| | |
| _ - - ~ ~ __-_ ___
| |
| l
| |
| | |
| ==Title:==
| |
| M2in G:nr.r ter Voltsge R gulstor Failura ID: EPS-2 NO: 6.3.4.6.2
| |
| ''i -
| |
| | |
| == Description:==
| |
| Failure of the AC.(automatic) regulator,
| |
| ~
| |
| \' ,
| |
| causing'a change in the generator's voltage output. Manual control possible in manual mode (base adjuster).
| |
| , Variations: Date: 4/21/88 Rev: 5 Selectable Steps Inputs Comments
| |
| : 1. Select percent of 0-200 percent 100 percent = normal output voltage output change voltage
| |
| : 2. Select ramp time 0-99,999 sec.
| |
| Brief Plant Response:
| |
| fs When the failure occurs, the automatic regulator will cause the exciter field current and output voltage to change, affecting the generator output voltage.. Depending on the direction of the failure, the generator's MVAR load will change in proportion to the amount of under or over excitation.
| |
| Various protection circuits can be actuated, possibly tripping the generator output breakers and/or exciter field breaker. Possible annunciators received due to this malfunction include GENERATOR VOLT REG TRIP, MAIN XFMR OVER EXCITATION, and GEN VOLT REG TROUBLE.
| |
| Suggested Instructor Action:
| |
| None.
| |
| Events:
| |
| : 1) LER 20-1-87-052: Mein Power Transformer Overexcitation Relay Actuation (Set JGENVH1=T)
| |
| O 0112w:4 309M/84M/4 1/88
| |
| (
| |
| | |
| ~
| |
| ,, ,,i,.., ~]
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| + >
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| e n i e , +. 3. 3.fi o i:ir p !ip i : ioi EF5-2
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| :f;.i p . t.i ti nP ti,i irni s'upp1'l J or! 2
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| i . tinsii.ilip lis 10-V>,,t1 ' .: o: 41is. ' f A l 'L' R G ull : I 6: N I'MI l" I l . ' * . Dn I1 '
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| If i 8 I' l I / i 'I< :':I 'i 5
| |
| . f .s : .. t. .:87-336; 1G'' 118 6( l il l P li' 88; i o i; ili' d all.II. D 'l lu.NSI'O H t1E R i .w ? - . . - i,. i o,i RC LiW AC iiIA I J Or8 l'oR ik? 'i 'hi r 4 f.Eii ,Od '
| |
| i . . i y Of - S E 'v l R I l s LEVEL i..Lh riu : 8/-052 CR I TERIUi! :
| |
| '.. .- I J. r p i l i ';i - T..: u itE gr UDGt:. OllE DafE li PJPIXI UR lilMi :
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| ., c. y; ..! pqi;;. pp .gi, ttO y n 10 y: , try py 1 y t1 I E ri 27 I l'
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| ,, i .r. v. r rp ip / T ciiilc' DOR TCH COkRECl1VE ACT l
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| . . U t: ! .rt / P/F OU l ar;E 1
| |
| <ar. \.s , I n i nn I l'Y Ilt . 'c/ F.XN S E E
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| , <n iu+8 9 /
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| IJ'r / H!l PHOCEDURE 1 f o i , r; r: Y : D: L 4 2- UUD: OtPi- ER9 P1C flFG :
| |
| i Oft 1 G I r4 AL DiiE DATC iOc24'a/ f; I A'I U S : COi1PI Ell' Li Li in. -
| |
| RD( l Of t CLiiSilltE : 10/07/G~l i
| |
| nei , a. r 70, /3 ( A ) ( ' y ( i: y y up[G ig t j' polt i o ,1 % > : ,-
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| , w; p v r a ,'yy ... o g ij 0:< 1 c ,;i .u 7,g p 3 0. co co ;
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| . , .n a p r 3. PEDf'ORhED I:Wi-S - '3. r.l. L J. i q l f t h, k l'S Poe l D E D r'initiii>Lt i
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| .. .. lir r P C (- 1 ili etsr. i o l i J F .l l? A i I Oil o..i :: , O'< ^)/ni.
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| 't* t Ie Ifl tt J'*I', j( (- [' ) l)[;,i j } ( f l 'jf I [ l I I,[ll. ( !l' l ' !?e I, l f Ipy J[j, es.
| |
| ),s / [3 [ , j, no.
| |
| D. P G ALil ' t . J. It/ i tiinis . i ii,1> / VP , i n ' ?R 87, Oho .
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| ' - y PF 1- c.!:I . 10Ci 9 50 7 :i( A ) ' ,' ) ' fV)
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| )
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| w '
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| ; y ;p : W. ,< ; 4 4 r=-q. sop . , t c on f y
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| .r i i" . < (3ip 1; y x , , ' ? ' f,5;2
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| +
| |
| +, 4 ch. .p .
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| U E ;i e is S M 'i C* D i p '; T 7 0t+9 ' PR IOR 'i n li /E H T -
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| . a_.
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| m . prf,1< uqon j:i rytp1 pn 1 :: : ggVItriBEP 24, I90/ EVEdi 1.l nE 2
| |
| ' ?I v HU!;E : 1 - POWER'OPERATIUH4 RX POHER: 130 BCC '. A U )
| |
| 1 p U.i tWf Us"E 'PREGGURE : 5'5 7 DE6f TEE'3 1 '2?11 PSJG
| |
| : 5. 4
| |
| .4 )
| |
| 8, FCNCPIPT10hl OF EVENTJ x-
| |
| . I d p r- HER i4! Li*t S I EMS OR tlOhPDr'EN I 9 iFOl Flo SLE AI- 1HL IJEG 1 HN It!E OF 'lilE EVENT IMi1C!! CON TR11:UTED ' 10 ltlE SEVLR11 Y . HF 'f HE EVffiJ I .
| |
| : v. .
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| . 7,1 ,4io ice :FI'TLiiOE P 24, i ?f r; , IAlt.I' 600 0..
| |
| PUMI R MtT UR1 Hr8 %
| |
| U. 1.)'d,I149_LH PROGRESS, DURING IHE PP9Ul!S9 Ol' ' 9YNCHROH lZ I NG
| |
| . t i t: . tinIN GiTHERol0H (TB) IO !HE GYSTEM GRID, lHE HA,fN I R AtF FORhE l!
| |
| l OVEftEXCITnTION ftELAY ACTUATED. I'll19 ACT lVA I ED Ti!E EIGHTY-G1X 8utIWUT HELAi IJHICH CAUSED THE Hol:4 GEHERA10R, HAIN POWEP- i T P.A p c S O D iiE R. tEL1- AND UNIT AUXILIARY IRAllGFfjRHEF (EA) IO 15E j l- 1. t ' ) o I t' ELL Y I SOLATED FROM Tl1E S(S FEM Gh10. ACTIVATIDH OF.
| |
| Ei MI7 E IClll Y-SIX LOCKOUT RELAY 6LGO INIT.I ATES A GENERATOR '
| |
| * li i n . TURPJHE 19 [P t.'JJ) , AND RE AC TOtt IRIP fJG) TO PROfEC1-e4t".rni.itiHENT FPOM POSSIBLE ELECTRICAL FAULT DAMAGE. nLL
| |
| ' t: 1 Y SYETLh9 OPERATED AG DESIGNED. G i n D L E Pl..A U T C O N D i~I .t O N D
| |
| ~ #ir r>r ESTnPl.;T GH'iD AT 0100 ON SEPTEMBER 25,'iYO7.
| |
| q f i t esfc [*lff* C,9ff!''EL RE8.tPDHDED 6PPROPilI All!LY AHD EN I ERED PP hr7? D11P E iPHEPa0, REACTOR TRIP OR 9AFETY J H JEC1 ION.
| |
| 4
| |
| ! I]8 "l'['l?'''cj'ld$jb ll if) lfd } [f (CA I1hf 0 \*' [ g ) I l l(( l.[ f } f[. (*llbidk h I S I k. H S :. ;;/Al ' n 1 '30 ? ON SEPIEhl:CR ?1, 190;' I't lP S Udll I IO ire,cpy;,oik,(a)(I.g),
| |
| i N 'i). l OklhN N((f Ol{ I f I) l 1ll M ,llNe 1 I lQ l+H~ I [ U , s' I I O I i. / I \lhA
| |
| . t.. V 'lil i 119 Calt!n.i.1 1 0 t> 'I l l o r l' t~. 91 !! I ! . 9 14 liv iF!''ril 8.*It tilJ f 0Hre l (U ra dia cl f oi! rl AN f i.:NGINEEUL'O SAFEi Y 8;F A I UP E , 'l Mr l IID I NG THE
| |
| : i. .$1 TOI! PRO 1rCTION SYSTEM.
| |
| { 's l - 1I ,
| |
| j t I il : (.- 1 - '.,t 4 :: li E li VF t f 1 I f3 1 til!' r l: '.isi f . (11 I:' B., ! ] t.'ti !.it... fil)Al 't ff. I G l
| |
| . . _ . _ _ _ - _ _ _ _ _ _ _ . - _ _ - _ _ - _ _ _ _ _ _ _ _ - - _ _ _ - _ - - _ - _ _ . . - - . - - - _ _ _ . _ . _ _ . ._ -- __. _. _ J
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| ..~,_, x . e ne... l . . m . m m . <,1 m, _ .m 1E ,,
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| -, T r! ~ D . r. lie 8'ob f ilVESTIGAT Ttiu til!Ly f l El. bEb A l. oieqE Cotti'EC 1 totJ
| |
| * - t io imc LL CutnACT PLDCK . LOCn~IED 0,4 1HE GECOs tD ARY SL0E OF liiE
| |
| ': i lVr 9'' ;"i MV.RECULATING POTENTIAL TP AN9620kMER. fHis FAJLURE
| |
| : t. i '- . '0011; .VO it'F f. VEti T . HOWEVER, II t.lAS i'01 ~t !!'i itOD I Ch8)GE
| |
| - 09 fl DC 1HE ':UPSE OllEN T PL Ai!T ' 3TARTUP A JD l
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| * flE R . 6 9 f. E tJ S I Oil , OAD i
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| 4 .o o .1. y a n a r,9: 11IHUED 10'MOHIlOR THE o si!D ". P!lA9E I:'U 1 Ei' f 1 OL P A H ; O M E !! S .
| |
| IiJC.LUDED IN THIS COhl RCHE!!G1 VE lion! 10RIt8G PROF Psh i
| |
| *n o r - HE ObF.REXC1TAllON PELAY, UllICil ~ nC IL.t V41!!D BOP TilG THjS 1
| |
| ~ IOEHY, Ai#D Tile VOLTAGE REGULAIOR. 'Ili'l G T! 51 T HG REVFALED IHal- !
| |
| ( ' 601HG: . (:.4 C P E N O R M A L .
| |
| ( pt r cir N! UL THA1 1HIS WAS OPERATOR-ERROR 11:; RULED 001, AS THE 91 rq e TO",9 t. E RF" nDSERVED EXECI)TIflG 1IIE E V OI lli 1 0t! IU ACCul! DANCE It6 :'e l'p e ii, E s it...r< E puGit 100-.3. 11 ht e% I AL% s !t 1 f81 H 1 0 D Olo f '11t o l
| |
| 'sflj P Tout' -99tS NOT REQUIRE SI'ECIAL T R A.lillHG OR P R A C I t C E..
| |
| !''I II I.E I f1 . '
| |
| .l f9 .I i l ! , 9 i
| |
| Vi14 EVEH1' l.:6!.i;0hE AVAIL ABLE , Illifi REPOR1 8.81 LL !!E S UPPL Elli:.M I E D .
| |
| * rr . nji, L. T p j ..; EVLHT REEtJR, n MEL8 REPoltT 48 T1. I UE :I S Si lli D , l
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| ' is , hatElY AiI ALY 31 S :
| |
| O a
| |
| 4: "'
| |
| im a t: .DO GATETY CONSEQUENCES ASSOf?IAILD IJ t ill IHIS EVENI
| |
| ^1 6tL RYMrt'MS-OPERnTED AS DEC'':GNED. UNDER WORSE CASE
| |
| , i''h95f30iC OF A fif 0!IER Ir4I11 AL POWEP. LEVEL, file SYS T ENS diULD !!&VE RESP 0tlDED THE SAME AS TIIf 9 EVE 41 AllD lilEP EF OP E ,
| |
| lia y pr, ult:. OE . NO C(.i ETY CONSEQUEpr;Es . 1 tis i :t E r APABTL T 1 (
| |
| . sip i sip i .-y,(.f E l Y RELATED PtitiES 10 Uni-Elf PEfnlED fillSEG ON llHi l' e ,0.- ' ' E F .,. AVOIlAPLE 'IllROUCHOUT THE EVFa11.
| |
| a.
| |
| p 8
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| . I
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| < . ,13 r e. c.. iitw r,r1libag ung ro R j;Te l i .iq li n; u plu t ,1.
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| .;...c> s e or sie ;,up PLA(:F 1HE IIDIi 'l ? ! e. > i GI'll f PHI Ttiti'h I lin'.
| |
| +f.. ! el tick .' i r i t r,. o O.
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| -3 i'
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| * 1. I P Ar ".;F OR i1E PM . O V Pif ''C'! I A ri D;' 611.Ar,
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| ,.t' i ty (llE 4
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| o> c? auniop u!lng Hop yi vi i. p in ! ORT 'd I:ikil PLIIEH il VE
| |
| !1 ! -
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| 7 ?iE H E TURI! TO SEP VILI til UUf1 1 1113. L Hl it i lleur.iftG
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| . ' . . - <> y ,0 p ip ntn, pui,sy.}t nqcrp:;1oje pr,vi:f,iy p inq nphpIl OPAL
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| . m- .
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| I t t! l. r : op p , sin 4)* at'
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| * t h /,l. . d. ) ! O j 8 9 ARE Cl)f4] EiiPl . A TED t t i i}llN I I H f. ,
| |
| 1 O
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| ; .;- , , , --T-
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| 'i~ sjp; , . Qp. .,p- [.its : ;;8 .
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| .1 4
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| > t ' ,, a
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| ,1 X c;,j --' y- 4 a , Y; r
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| |
| ~
| |
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| |
| g 'i _
| |
| ; h : $ ', .,' #4*#;
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| * E; 4 **
| |
| .g,' ,, h )I, , k I)}th h
| |
| ;.*f t ti'~MJ i .(JC( / i.. I f t s T ) , ,
| |
| f 1- --' '
| |
| * E V -I l e' t OC C UhMEl(CCG :
| |
| itm.,
| |
| u iff t .i ,';,;Vf f sh thluPRilllCES OF OVERE 9 Ci l(ii 10i'' H El n'( Af;fUAT!Of!. '
| |
| 1
| |
| (. COWOtEili' 4 FAILUPE I>ATA:
| |
| ., +
| |
| i . f; a iUr. ,1,' f URE P. NI)flENGLATURE itot'EL ;; ilFG pal (I
| |
| . ' ',;/n JI .h!I U . COHfACT, S E C O P ') A F Y , 2S l' V it E G U A L i l i4 G 23 250H >.1 A llD lt F.C C P T A C L E (RAl'SFOPMElt 5
| |
| IO 1,
| |
| 1' i
| |
| l 1
| |
| t I..
| |
| 1
| |
| | |
| BRAIDWOOD SIMULATOR MALFUNCTION
| |
| | |
| ==Title:==
| |
| 345 KV Switchyard Bus Trip ids EPS-3
| |
| / N0: 6.3.4.6.3 Descriptions Selected transmission line trips due to line fralt, causing possible loss of power rv;ely or load.
| |
| Variations: Dates 6/13/87 Rev: 3 Selectable Stept; Inputs Comunents
| |
| : 1. Select faulty line 1, 2 or 3 1 = Davis Creek (2002) 2 = East Frankfort (2001) 3 = LaSalle (0104) t k- Brief Plant Responst: (IC-17, 100%, all systems in automatic)
| |
| : 1. Switchyard breakers 4-7 and 7-8 open. First annunciators received are CB 4-7 TRIP, CB 7-8 TRIP and LINE 2002 TRIP.
| |
| : 2. Switchyard breakers 3-4 and 1-3 open. First annunciators receivr.d are CB 1-3 TRIP, CB 3-4 TRIP and LINE 2001 TRIP.
| |
| : 3. Switchyard breakers 1-3 and 1-8 open. First annunciators recieved are CB 1-3 TRIP, ACB 1-8 TR.IP and LINE 0104 TRIP.
| |
| Suggested Instructor Action:
| |
| Malfunction must first be cleared and LOA's EPS-1 and 67 used before the tripped breakers can be reciesed.
| |
| Events: None O 881M/5 5/89 l
| |
| l.
| |
| L_===______.___________ _ _ _ _ . __
| |
| | |
| BRAIDWOOD SIMUI.JLTOR MALFUNCTION
| |
| | |
| ==Title:==
| |
| Loss of Non ESF Bus ID: EPS-4
| |
| - NO: 6.3.4.6.4
| |
| | |
| == Description:==
| |
| Selected 4160 service bus supply breakers trip causing a loss of power to that bus' loads. Trip is due to a bus fault.
| |
| Variations: Date 6/13/87 Rev: 3 Selectable Steps Inputs Comments
| |
| : 1. Select faulted bus 1 or 2 1 = 143 2 = 144 Brief Plant Responses (IC-14, 50%, all systems in automatic)
| |
| Loss of Non-ESF bus causing loss of various non-essential equipment. The.
| |
| first annunciators received include BUS FD BKR TRIP, BUS VOLT LOW and FD BKR TRIP from 480 volt buses.
| |
| Note: Loss of bus 143 at high turbine power will cause a turbine trip /Rx trip due to losing 2 Cire. Water Pumps.
| |
| Suggested Instructor Action:
| |
| Clear the malfunction before attempting to reenergize the bus. Use of LOA CRF-8 will swap DRPI power supplies.
| |
| Events: None O 881M/6 5/89
| |
| ' 4-----------~___. _m______ ___ __ _
| |
| | |
| BRAIDWOOD SIMULATOR MALFUNCTIO?!
| |
| | |
| ==Title:==
| |
| ' Loss of ESF Bus -
| |
| ID: EPS-5 7 NO: 6.3.4.6.5
| |
| \
| |
| | |
| == Description:==
| |
| Selected 4160 ESF bus supply breakers trip causing a loss of power to that. bus' loads.
| |
| Trip is due to a bus fault.
| |
| Date 6/13/,7 Variations:
| |
| Rev: 4 Selectable Steps Inputs Comments
| |
| : 1. Select faulty bus 1 or 2 1 = Bus 141 2 = Bus lar f ,
| |
| Brief Plant Responses (IC-17, 100%, all systems in automatic) .'
| |
| l Loss of ESF bus causing loss of ESF equipment. First annunciators received-include BUS FEED BKR TRIP, BUS OVERLOAD OR VOLT LOW and various PUMP TRIP alarms.
| |
| Suggested Instruction Action: i i
| |
| Clear the malfunction before attempting to reenergize the bus. f Events: None 1
| |
| l 1 O ,,1M,, ,,,,
| |
| 1
| |
| | |
| BRAIDWOOD SIMULATOR f14LFUNCTION
| |
| )
| |
| js. '
| |
| | |
| ==Title:==
| |
| Diesel Generator Failure ID: EPS-6
| |
| '[
| |
| 6.3.4.6.6 NO:
| |
| (/
| |
| | |
| == Description:==
| |
| Selected diesel trips or fails to start due to a mechanical failure of the fuel rack's control mechanism.
| |
| Variations: Date: 9/20/88 Rev 5 Selecc40? e Steps Inputs Comments
| |
| : 1. Select faulty diesel 1 7$
| |
| 1 = 1A diesel 2 = 1B diesel .
| |
| Brist Plant Response: (IC-17, 100%, all systems in automatic)
| |
| .O If the diesel were running and loaded, it would trip, causing a loss of power to its ESF bus. If the diesel were off when the malfunction was actuated, nothing would occur until the diesel was required to start (manually or automatically), then it would start " cranking", then trip. The first annunciator received is DG TROUBLE / FAIL TO START.
| |
| Suggested Instructor Actfon:
| |
| Clear the malfunction to restore the diesel to an operable status.
| |
| Events: 1) DVR 06-01-86-175: IB D/G high Jacket water temperature trip.
| |
| : 2) LER 06-02-88-003: 2B D/G inoperable.
| |
| \ 881M/8 5/89 I
| |
| : r___ ~:" - ' - - ~ ' -- -- - -
| |
| /
| |
| | |
| OEVIATION INVEST!0AT!04 REPORT T--Q 18 O!ESEL GENERATOR HIGH JACKET WATER TEPFERATURE TRIP A'ID Pact i k.-)INCOMPLETESTART I [QF l EVENT DATE DIR NUMBM . REPORT DATE
| |
| )
| |
| l $EQUENTIAL REVISION
| |
| ' MONTH DAY YEAR STA UNIT YEAR MODE NUMQ13_ NUH8ER, MJN TH DAY YEAR _
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| POWER //
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| 110 Ils als als O l'1 sls -
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| 117 is -
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| o ! q_ 11 2 01 1 81 6
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| 'EVI'
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| ['/
| |
| ol 91.1 CONTACT FOR THf5 DIR NAME
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| _ E PHONE NUMt3R AREA CUO!
| |
| Bill Knuba Ext _ 2274 ei1 Ii 211 14 I-t5l4 1411
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| , COMPLETE ONE LINE FOR EACH (OMPONE 7 UPE DESCRIBE 0 IN TMf1 REPORT CAU$E SYSTEH COMPONENT MANUFAC- REPORTABLE CAU$E SYSTEH COMPONENT MANUFAC- PEPORTABLE TURER 70 NPRDS TURER TO NPOD$
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| X Elf Tlt I l 11 il 4l E Me l l l l l l l X Elf FlD l i WI21910 Yes l l i l l SUPPLEMENTAL REPORT EXPECTED gl DAY YEAR
| |
| $UBMIS$10N l YES fif vet. camelete EXPECTED $UBMIK1fDN DATE) l NO OE Of 1 87 3XT A. PLANY CONDfff0NS PRfCR TO EVEMT:
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| (
| |
| (/ MODE 1 - Power ceeration Rx Power 9H RCS [AB) Temperature / Pressure Normal Oceratina B. DESCRIPTION OF EVENT: '
| |
| At 1317 on 13/18/85 the 15 diesel generator (OG) {EK) was started per the monthly operability surveillance.
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| 180$ 8.1.1.2.a-2. At 1346 the diesel tripped on high jacket water temperature. The actual jacket water tenp;rature at the time of the trip was 170 F w,hich is normal. Limiting Condition for Operation Action Requirement (LC0AR) 1805 8.1.1-la was entered. At 1721 the diesel was restarted in an effort to exit the LCOAR. The diesel failed to reach rated speed (600 RPH). The ecuipment operator verified settings on the governor with the local operator aid. A normal stop $?gnal was initiated af ter two minutes with the diesel at $35 RPM. The diesel failed to go into cooldown and stopped ininediately. At 0956 on 10/19/86 the 18 diesel was started per the monthly operabtitty surveillance tn order to exit the LCOAR. At 1215 on 10/19/86 LCOAA 180$ 8.1.1- a was exited. There were no systems or components inoperable prior to the event that contributed to the event.
| |
| C. CAUSE OF EVENT:
| |
| The root cause for the jacket water high ternperature trip was a loose wire on the temperature switch. The roIt cause for the diesel's failure to reach rated speed was a failed motortzed speed potentiometer on the governor. The cause of the failed potentiometer has not been determined yet. A supplemental report will be issued af ter the cause has beeri investigated. The cause for the failure of the engine to go into croldown is indeterminate.
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| 3 d I
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| )930*1)
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| - - _ - _ _ _ _ _ - _ _ - _ _ _ _ _ _ _ _ _ _ - - _ _ _ _ _ _ - - - _ . _ _ _ _ _ _ _ - ._ _ ______-_b
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| | |
| DEV!ATION INVESTIGAr!ON REPORT TEXT CONTINUAf!CN
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| ! TITLE QIpjuPRER I
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| _ _ FtGE KEVISION il - " l SEQUENTIAL IB DIESEL GENERATOR flIGH JACKET WATER TEMPERATURE l
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| ~~l 1
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| TRIP AND INCOMPLETE ST&?T 1{$_ 01 1 916 ~~
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| 11 7 IE ~~
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| 01JL,[_2 0F 0 12 TEXT D. SAFETY ANALYSIS:
| |
| There were no safety consequences as a result of this event. Both failures wcJ1d not have impatreo the operation of the er.gine if the engine was operatthg in the emargency mode. The high jacket water temperature trip is bypassed in the emerge. icy mode. In aedition the trip was consinered an intermittent f ailure of the temperature switch. At no time was the engine subjected to any adverse temoeraturec. The failure of the speed potentiometer did rot impair the safety function of the diesel becaase tne potentiometer is not used in the energency mode. In ar. en.er2 enc / condition the governcr operates in the isochronous mode. In this mode the governor is given a constant speed reference resistance thro 9gh a separate rasistor. In the testing mode the motorited potentiometer allows the reference resistance to be changed in order to adjust engine speed and generator load. The failure of the engine to go into the cocidown has no effect on the operability of the diesel.
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| E. CDRAECTIVE ACTIONS
| |
| * The loose wire on the temperature switch was tightened and the switch calibration was verified. The failed potentiometer was renoved and replaced.
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| F. PREVIOUS OCCURRENCES:
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| DVR NUMBER I.Eg NONE G. COMPONENT FAILURE DATA:
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| 4 MAMPFACTURER 30MENCLATUDE MODEL NUMBER' PFG PaRT NUMBER Woodward Motorized Potentiometer 37716 Square 0 Temperature switch 9012-B00-1 U
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| l .
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| i (0900M)
| |
| | |
| papQ LICENSEE EVENT REPORT ILER)
| |
| Facility Name (1) Docket Number (2) pace e3)
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| Byron. Unit 2 01SI01010141515 1 lof!0l4 U LI' I#I M h t g C g C gI g g g g g NT NOT SATISFIED DURING UNINTENDED 28 DIESEL GENERATOR Event Data f5) LER Mumber (61 Renart Date (7) Other Facilities Involved (8)
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| Honth Day Year Year /// Seguential //j/ Revision Month Day Year facilitv Names I Decket Nhaber(s) fff ff
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| /// Number /// Number I NONE Of 5l Of 01 Ol I l
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| ~ -
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| 0l3 2 19 Bl8 B1 8 0.l,013 _ Q ,1.,0., I l I of 51 01 01 01 l l THIS REPORT !$ SUOMITTEI) PURSUANT TO THE REQUIREMENTS OF 10CFR (Check one er more of tLe followinal (111 1 20.402(b) '20.405(c) 50.73(a)(2)(iv) ,_ 73.71(b)
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| POWER , 20.405(a)(I H t ) 50.36(c)(1) 50.73(aH2)(v) 73.71(c)
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| LEVEL 20.405(aH I H il) 50.36(c)(2) 00.73(a)(2)(vil) Other (Speelfy f10) 0l9 3 ., 20.405(aH 1)(lli) _1_ 50.73(a)(2)(1) 50.73(aH 2)(vill H A) In Abstract
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| /////////////////////////,/ 20.405(a H 1)(lv) 50.73(a)(2)(ll) 54.73(aH2)(vill HO) below and in
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| /////////////////////////' /
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| f
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| ,,, 20.405(a)(1)(v) _ 50.73(a)(2)(lit) 50.73(aH2H z) Text) tirrwtFF CGNTAff FOR THIS LER f121 Name TFtFeNaur paamre AREA CODE D. Brindle. Omarattne Ensiname. Entan=lan 2218 al115 2131di.1514141 CmWLETE fME TIN Fu ram cmpnaamT FAltiMF DElfEfAFB IN THIS REPMT (131 .
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| CAUSE SYSTEM COMPONENT MANUFAC- REPORTAett / CAUSE SYSTEM C0pr0NDIT l MANUFAC- REPORTABLE Tianas TB gympg ,/ Tiers To gegDS l I I I I i i I I I I I I I
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| /',
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| I I I I I I I / I I I I I I I SlipptEMMTAL REPMT DPECTED f141 Empected Manth l Dav i Year
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| ~
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| Substseien lYan (If ven. e===lete DPf 9mm **- ^111m BATE) Yl IS ** I !I l A85 TRACT (Limit to 1400 spaces, f.e. apprealmstely fifteen single-space typewritten lines) (16)
| |
| On March 29,1988, the 28 Dieset Generator (DE) lef t bank starting air system was removed from service to repair a leaking valve. Upstream and downstream isolation points were chosen from a pip ng and Instrumentation Orowing (p610) to perelt the maintenance. The p4!D lacerrectly represented the actual piping arrangement in the plant, therefore, the right bank of the starting air system was aise isolated during the intended (seletten of the left bank. This condition resulted in the f eelation of all starting air from the 28 06, thus, making it inoperable. The Inoperability was identified on March 31, 1986 when an Equipment Operator noticed that the "theit Avellable for Emergency" indicating Ilght was extinguished, and the Technical Specificatten tietting Canditten for Operation Action Requirement was implemented.
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| The following corrective actions have been er are tioing takca:
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| : 1. Caetien ca.-se and labels emplaining the piping discrepancies have been hung locally and in the mein control rees.
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| : 2. The P & 10 will be corrected to represent actual plant condittens.
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| : 3. DG auntilary egulpment labels will have Byron part nebers.
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| ~ 4. The operating rounds procedure will require periodic checks of the " Unit Available for Emergency
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| * indicating Ilght.
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| : 5. Training programs will address the Indicating Ilght.
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| \/
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| Similar events have not occurred previews 1J .
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| I 8
| |
| (2015M/0206M)
| |
| | |
| 7 i
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| LICENSEE EVENT REPORT (LER) TEXT CONTINUATION FACILITY NAME (1) DOCKET NUreER (2) LER NUDSER (6) Pace !D Year /// Sequential /j/j/ Revision
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| /,/,/j f
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| _ Number /// Number By,tyn. Unit 2 0 l 510 l 0 l 014l Sl 5 8lB - 0l013 - 0 l 0 of 2 0F Of TEXT Energy Industry Identification System (EII5) codes are identified in the text as [xx]
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| A. PLANT C0peITIONS PRIOR TO EVENT:
| |
| Event Date/ Time' 3/29/88 / 0815 Unit 2 MODE 1 - Power Omeratiana Rx Power 931 RCS (AO) Temperature / pressure Nermal Oneratino B. DESCRIPTION OF EVENT On March 29, 1988, preparations were in progress to remove the 28 Diesel Generator (DG)[EK] left bank starting air system from service in order to repair the left bank :meisture separator drain valve (25A141D) which had been leaking by its seat. A Ilconsed reacter operator reviewed the appitcable Piping and i Instrumentation Drawing (p & 10) 554 sheet 48 to determine upstrees and downstream isolation points to {
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| safely permit the maintenance. By closing the left bank air receiver evtlet valve 25A1400, the left bank air receiver (20G0158-TO) would be isolated from the work ares. By closing the left bank starting air valve, 2DG51823, the left bank starting air system would be isolated frcm the work area. Diesel Generator operability requires er.ly one of the two starting air banks to be in service, therefore, no Technical Specification Lietting Conditten for Operation Action 2eguirement (LC0AR) was igtemented prior to .
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| Isolating the left bank. At 2236 en March 29, 1908, a non Itcensed Eguipment Operator (ES) closed valves 25A1400 and 20051828.
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| On March 31, 1988 at apprestastely edge, an E0 liformed the Itcensed Senter Reactor Operator Shif t Control Room Engineer (SCRE) that the " Unit Available for Emergoney" indicating Ilght at the 29 06 local control panel was not illuminated. There was ne procedural regelrement to perledically check the status of this Ilght, but the E0 noted this abnormel condition during a general Inspection of the DE leccl control panel.
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| This indicatirp Ilght is illuminated if one of two starting air banks is pressurized, and one of two direct current (DC) power supplies is energized. The SCRE tamediately Initiated a nuclear work request to have the problem investigated. The initial tavestigation was condweted by Electrical Maintenance technicians and indicated that the right starting air bank pressure switch, which provides an input to the " Unit Available for Emergency" indicating Ifght, was defective. The left starting air bank was known to be out of service and depressurized for maintenance. The cou61 nation of a defective right bank pressure switch
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| 'and a lef t bank out of service conditten emplained the doenergized Indicating Ilght. Therefore, the LC0AR was not entered, since the problem appeared to be an indicatten fallere unrelated to 06 operability.
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| Subsequent investigetten deterurined the pressere switch to be in satisfactory working condittsn.
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| On March 31. Iges as ST39, a non-Itcensed Technical Staf f Engineer jelped the investigetten. Following a pressere check of both starting air banks on the 20 DE, the Technical Staff Engineer concluded that both starting air banks were depressurtzed to atmospheric pressure. At 0815 the engineer notified the $CRE of this conditten. The SCE tamediately initiated "LC0AR Electrical power Systems AC Sovrees Tech Spec LCD
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| .8.1.1 Operating procedure" (LCent 2005 8.1.1-la) for the 28 DG Insperablitty.
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| Investigetten continued to determine the cause of the depresserlaatten of both starting air banks. The investigation revealed that the actus1 air bank piping arrangement did not agree with the P & 10 piping arrangement. Specifically, the p 4 10 shows valve 2DG81829 isolating air from the left bank starting air system (20G0135-T0) at the engine. In actuality, valve 2DG61828 isolates the right bank starting air system at the engine. Therefore, when the E0 closed 20C51828 en March 29, he actually isolated right bank starting air from the 06. When he closed 2$A1400, he isolated lef t bank starting air from the DG. At this point the 28 DG became inoperable, since neither bank of starting air was available to start the DG.
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| Oi (20lSM/0206M)
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| | |
| LICENSEE EVENT REPORT fLER) TEXT CONTINUATION FACILITY NAME (1) DOCKET NUPSER (2) LER NUPSER f 6) pace (3)
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| Year /// Sequential /// Revision fff Number fff N+r Buran. Unit 2 0 1 5 1 0 1 0 1 0 1 41 51 5 8l8 - 010l1 - 0l0 01 3 0F Of _A TEXT Energy Industry Identification System (E!!S) codes are identified in the text as (xx]
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| B. DESCRIPTION OF EVENT 3 (Continued)
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| At 1315 on March 31, 1988, the actual air start piping arrangement had been verified, the right bank had been restored to service, the 28 OG had been declared operable and LCOAR 200$ 8.1.1-la was esited. There were no other systems or components inoperable prior to this event that contributed to the event.
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| From 2236 on March 29, 1980, until 0815 on March 31,1988, the 28 OG was inoperable and the appropriate !
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| Technical Specificatter. LC0AR was not satisfied. This event is reportable in accordance with 10CFR50.73(a)(2)(1)(8) as a violetten of the plant's Technical Specifications.
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| C. CAUSE OF EVENT:
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| l The intermodlate cause of this event was an incorrect representation of the starting air piping on p & ID 4 5d sheet 48. The rest cause of the errors in the drawing is indeterminate. The subject p & 10 shows the left bank starting air receiver (2DG0l&TD) feeding starting air valve 20051828 and the right bank starting air receiver (2DG01ETC) feeding starting air valve 20G61815. In the actual pfplag arrangement, 20001 5 79 feeds valve 20051835, and 2000l & TC feeds valve 20081828. Sectlens of both the left and.right banks of air start piping are buried in concrete, therefore, the E0 had no way te verify the connections between the recelvers and the engine. The p & 10 was used by the licensed reacter operator to determine I maintenance Isolatten points that resulted in inoperablitty of the 23 08.
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| D. SAFETY AlthtY111:
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| There were no safety consequences as a result of this event. The 2A diesel generator was fully operable i during this avant and could have supplied emergency electrical power to Unit 2 If required. In addltion, both Unit I diesel generators were fully operable and cov1d have been electrically crosstled to Unit 2 If regutred.
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| Pinintenance personnel were not endangered during this event, because the werb area was completely isolated free sources of pressurized air.
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| E. CORRECTIVE ACTIfMIS:
| |
| The following corrective actless were Labsn er are planned to prevent reoccurrence of this event.
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| : 1. The IA, IS, and 2A d{esel generator starting air systeen, were reviewed for sieller discrepancies.
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| No problems were identified on the Unit I diesel generators. The 2A diesel generator starting air ;
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| piping was found f e be siellarly misrepresented on p & ID 454 sheet de.
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| : 2. Cautten cards owplaining the discrepancy between the actual starting air piping, and the p & ID -ere hung on the OG local centrol panel swttches, the main control board switches, and the air receivers ]
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| for the 2A and 2B diesel generators.
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| : 3. Labels were placed on all Unit 1 and Unit 2 diesel generater air receivers with the correct Equipment Part Numbers (EpNs), and a statement emplaining which side of the engine each receiver feeds. I I
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| I (2915M/0200M) 1 l
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| | |
| LICENSEE EVENT REPORT (IER) TEXT CONTINUATION FACILITY NME (1) DOCKET NUPSER (2) LER NUPSER f 61 Pace (3)
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| Year / Sequential / Revision
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| /,/,#/
| |
| [ /, Number ,/,/,/ Ntabe r
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| //
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| Bvenn. Unit 2 0 1 5 1 0 1 0 1 0 l 41 51 5 8l8 . Ol0l3 - 0 10 01 4 0F Di ( '
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| . TEXT Energy Industry Identification System (E!!5) codes are identified in the text as (xx).
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| E. CORRECTIVE ACTIONS: - (Continued) 4 A design change request has been submitted to change the p & 10 to match the actual routing of the starting air piping. Action Item Record (AIR) M54-225-86-0078 will track completion of the changes.
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| : 5. Control switch labels for all auxillary equipment on the DG tocal control panels will be changed to indicate the Byron EpNs. Currently, the labels have numbers assigned by the egulpment supplier, while the p & ids for the avulliary egulpment show Oyron EpNs. This corrective action will prevent any equipment confuelen when future out of services are implemented on DG aux 111ery egulpment. AIR M54 225-48-0000 will track completion of this iten.
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| : 6. A temporary procedure change to the E0 rounds precedure has been implemented to direct the E0s to check the status of the " Unit Available for Emergency" Ilght on a shif tly basis. A permanent procedure change, which includes this addition, has been submitted. AIR M54-225-85-407g will ensure the permanent procedure change is completed.
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| : 7. A training revisten roguest will be subeltted to include an explanetten of the " Unit Avellable for Emergency" Ilght in the E0 and the ifconse training programs to ensure operator understanding of the meaning of the status light. A required reading package will be issued to all licensed and.
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| Egulpment Operator personnel to ensure an understanding of the indicatten on a short ters hasis.
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| AIR M.54-225 48-4077 will track the completion of this ites.
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| F. pttEVilE11 EC1EAEEEtt LG IRBMI ILT1C NONE G. CINIMElli FAILlat DATAt a) NMEMACTtMER MBEE1AftM MML IRBMI W E PART IRSAER Not Applicable I
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| b) Im_ GElli Met App 11caMe -
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| c) REBE.TI fir M S SEAM Mr Not Appilcable )
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| i l
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| O
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| .I l (2elsN/etosN)
| |
| | |
| BRAIDWOOD SIMULATOR MALFUNCTION
| |
| , Tit.le s Loss of D.C. Distribution Bus ID: EPS-7 NO: 6.3.4.6.7
| |
| | |
| == Description:==
| |
| Supply breakers for selected D.C.
| |
| bus trip due to bus fault.
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| Variations: Dates 6/28/87 Rev: 4 Selectable Steps Inputs Comments
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| : 1. Selections EPS-7A EPS-7B
| |
| : 2. . Select faulty bus 1, 2, 3, 4 1 - 125 vde bus 111 (EST) or 5 2 - 125 vde bus 112 (ESF) 3 - 125 vde bus 113 (Non EST) 4 - 125 vde bu,s 114 (Non EST)
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| N~/ 5 - 250 yde bus 123 (Non EST)
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| Brief Plant Response: (IC-17, 100%, all systems in automatic)
| |
| ~1 or 2 - Loss of the 111 or 112 vde bus will cause an auto switchover to the A.C. input to the instrument inverters and loss of control power to all ESF buses, RCPs, 6.9 KV buses, and D.G. control power. Also loss of control power to Rx trip switchgear and sequencing cabinets. The first DC related annunciators rectived include: 125 VDC BATT MAIN BRKR TRIP, 125 VDC BUS VOLT LOW, (EST) BUS CONT PWR FAILURE, (6.9 KV)
| |
| BUS CONT PWR FAILURE and ,(NON-EST) EUS CONT PW1 FAILURE.
| |
| 881M/9 5/89 l
| |
| m--_____ _ _ _ _ _ _ . _ _ _ _ _ _ _ . _
| |
| | |
| --:_ _ - --un.; = u -
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| __x, - a. _.. _ w:- ---...r_-
| |
| BRAIDWOOD SIMULATOR MALFUNCTION I
| |
| ,f]
| |
| | |
| ==Title:==
| |
| Loss of D.C. Distribution Bus ID: EPS-7
| |
| .g Brief Plant Response (continued):
| |
| 3 or 4 - Loss of miscellaneous D.C. equipment. The First DC related annunciators received include: 125 VDC BUS VOLT LOW, 125 VDC DISTR
| |
| ~ PANEL FD BRKR OPEN, (NON-ESF) CONT PWR FAILURE, (6.9 KV) CONT PWR FAILURE, GENERATOR LOCKOUT RELAY DC FAILURE and FW PMP TURB TRIP CKT PWR FAILURE.
| |
| 5 - Loss feed to computer, E0Ps for main and feedvater turbines and the vacuum breaker.
| |
| Suggested Instructor Action:
| |
| - Clear the malfunction before attempting to resnergize the bus.
| |
| Events: None O 881M/10 5/89
| |
| ~~ ~ '~~ -- -
| |
| -~'rrrrz__z___ ____________~_r_r:1_._______: 3 21_r___ __ _ _ -
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| | |
| i BRAIDWOOD SIMULATOR MALFUNCTION Title Loss of 120 VAC ESF Instrument Bus ID: EPS-8
| |
| ~/~] '' NO: 6.3.4.6.8
| |
| | |
| == Description:==
| |
| Breakers supplying the selected instrument bus trip due to a bus internal fault. l j
| |
| Variations: See Malfunction EPS-17 Date: 3/16/89 Rev 6 Selectable Steps Inputs Coraments
| |
| : 1. Selections EPS-8A EPS-8B i
| |
| : 2. Select instrument bus 1, 2, 3 or 4 1 = Bus 111 with fault 2 = Bus 112 3 = Bus 113 4 = Bus 114 Brief Plant Response: Loss of various instrumentation in the Control Room including one section of the NI's. First annunciators received include PROCESS I & C CAB PWR SUP FAILURE, SOLID STATE PROT CAB GENERAL WARNING and BUS INVERTER TROUBLE.
| |
| Suggested Instructor Action:
| |
| Clearing the malfunction will reenergize the bus.
| |
| Events:
| |
| : 1) DVR 06-01-86-114: Loss of Inverter 114.
| |
| 881M/11 5/89
| |
| - .__,.__--_-----_-....-_-_---_..Lw-. . _ _ -_.. --__,___.._----._._x_ . _ - . _ _ _ _ - . _ _ _ . - - - - - _ _ _ _ _ _ _ - . _ . _
| |
| | |
| $f5$ ~ $
| |
| I f DEVIATION INVESTIGATION REPORT l
| |
| l l TITL( PAGE Lost OF INSTRUMENT INVERTER 114 00E TO FAILED RESfsTOR 1 10Fl 0 l 2 VENT DATE DIR NUMBER REPORT DATE SEQUENTIAL REVISION MONTH DAY YEAR STA UNIT YEAR y// NUMBER // NUMBER MONTH DAY YEAR 1 POWER LEVEL ,, g 3 ol 6 11 2 al 6 of 6 of 1 al 6 -
| |
| 1 1 11 4 -
| |
| oI o 0 17 1 18 8 16 CONTACT FOR THIS DIR NAME TELEPHONE NUMBER AREA CODE Terry Schuster. Assistant Technical Staff tunervitar Ext. 2244 Al 1 lE 2l1 l4l-l 514 l 4 l 1 COMPLETE ONE LING FOR EACH COMPONEN URE DESCRIRED IN THIS REPORT CAUSC SYSTEM COMP 0 RENT MANUFAC- REPORTA8LE CAUSE SYSTEM COMPONENT MANUFAC- REPORTABLE TURER TO NPRD$ TURER TO NPRDS l X EIr Il N1 Vl Y Wl Il fl o l i l l l l l I l l I l l i l l l 1 I SUPPLEMENTAL REPORT EXPECTED MONTH DAY YEAR SUBMISSION !
| |
| l YES tif vet. connlate EXPECTED SUBMIssf 0N r; ATE) i No f f 1
| |
| TEXT I A. PLANT CONDITIONS PRIOR TO EVENT:
| |
| MODE 1 - Power Doeration Rx Power . 731 RCS [A8] Temperature / Pressure Normal Oeeratina
| |
| /
| |
| \' 8. DESCRIPTION OF EVENT:
| |
| At 1030 on June 12. 1986 normal AC power feed to Instrument Inverter 114 (IP) [EF] was lost. Automatic switchover to DC Power supply also failed. This resulted in a de-energized Instrument Bus 114. The appropriate action requirement was followed and power to the tus was restored by switching to its reserve AC power source. In addition the action requirements for an inoperable Power Range (NI) [IG) channel, which resulted from Sus 114 being de-energtred, was also f 3110wed. Power to the Bus was restored prter to shutdown initiation. There were no systems or components inoperable that contributed to the cause of this event. Plant condition remaieed stable throughout the event.
| |
| C. CAUSE OF EVENT:
| |
| The root cause of this event was an electrical component f ailure. The Instrument Inverter failed wnen a 200 ohm /100 watt resistor. in series with the inverter gating control logic, f ailed open. This de-energized the electronic gating control circutt which converts the DC power within the inverter, to 120VAC output. This prevented both the AC & DC inverter power supplies from supplying 120VAC to the instrument bus.
| |
| D. SAFETY. ANALYSIS:
| |
| l There was no impact on plant or pubite safety. Reactor protection and safeguard actuations were always l available throughout this event. No Technical Specification Limits were esteeded.
| |
| l
| |
| . 1 O
| |
| v I
| |
| hmm, I
| |
| | |
| DEVIATION INVESTIGATION REPORT TEXT CONTINUAT!0N TITLE DIR NUMBER PAGE
| |
| .,_ SEQUENTIAL REV!5!0N
| |
| .[ ;
| |
| STA UNIT YEAR NLMBER NUMBER oss OF fusTRUMEuf INVERTER 114 DUE TO FAILED REsf$ TOR O! 6 01 ~ 1 B1 6 -
| |
| 1l 1 14 ""
| |
| Ol 0 2 0F: 0 l2 TEXT E. CORRECTIVE ACTIONS:
| |
| The defective registor was replaced and Instrument Inverter 114 was returned to service. Although other Instrument Inverter component failures have occurred it is felt this problem is an isolated incident as stattar resistor failures have not previously occurred. No further action is required.
| |
| F. PREVIOUS GCCtRRENCES:
| |
| DVR Nt9EER TITLE NONE G. CfDrtulENT FAILURE DATA:
| |
| MANUFACTURER NOMENCLATURE MODEL Nt3GER HFG PART NtDGER Westinghouse Resistor 443A325N11 f%
| |
| l )
| |
| v
| |
| )
| |
| 1 l
| |
| l l
| |
| l s
| |
| i I
| |
| (0900Mr
| |
| | |
| -~;x. n z..;_._.;.L..u... ._..__m... c ._. _--
| |
| 5-..- _ . . .
| |
| .BRAIDWOOD SIMULATOR MALFUNCTION I
| |
| -l Iitles- Load Rejection IDt EPS-9
| |
| (]
| |
| NJ NO: 6.3.4.6.9 Descriptions Loss of part of all of the load being supplied by the LaSalle line due to a grid distribution failure j t
| |
| Variations: Date: 1/5/88 )
| |
| Rev: 4 Selectable Steps Inputs Coastents
| |
| : 1. Select percent of load 0-100 percent 100 percent = 1175 MW that is lost
| |
| : 2. Ramp time 0-99999 sec
| |
| .\
| |
| Brief Plant Response:
| |
| When the fault occurs, generator load will drop rapidly. RCS temperature and pressure will incre.ase, control rods will insert at maximum speed, steam dumps will arm a d open. Depending on initial generator load, a reactor trip could occur due to pressurizer pressure or steam generator level. Various annunciators will alarm depending on size of load reject and initiating power level.
| |
| Suggested Instructor Action:
| |
| When it is desired to increase load again, first clear the malfunction.
| |
| A Events: None
| |
| . ;O .
| |
| l 881M/12 5/89
| |
| | |
| p , -- - . . ;-- . ; . ; :. : . :=;..-. - : .u;--- x- ; -
| |
| :; :a = :-. - _ - = . ;.== =
| |
| =.;-===w==--=-=--.-=----==p h
| |
| BRAIDWOOD SIMULATOR MALNNCTION L
| |
| fy
| |
| | |
| ==Title:==
| |
| ~ Loss of 6.9 KV Bus ID: EPS-10 M NO: 6.3.4.6.10
| |
| | |
| == Description:==
| |
| Selected 6.9 KV bus feed breakers trip.
| |
| Trip is due to a bus fault.
| |
| Variations: Date: 1/5/88 -
| |
| 1 Rev: 4
| |
| ]
| |
| li Selectable Steps Inputs Cosaments
| |
| ]
| |
| : 1. Selections EPS-10A i EPS-10B
| |
| : 2. Select bus 1, 2, 3 or 4 1 - 6.9 KV Bus 156 2 - Bus 157 3 - Bus 158 A .
| |
| 4 - Bus 159 Brief Plant Response:
| |
| Loss of 6.9 KV bus causes loss of RCP and associated loads from that bus. The reactor trips if above P-8. The first annunciators received are BUS VOLT LOW and BUS FD BRKR TRIP.
| |
| Suggested Instructor Action:'
| |
| Clear the malfunction before attempting to reenergize the bus.
| |
| l Events: None O 881M/13 5/89 a . . _ . . _, . . . _ . . ._
| |
| t- -- - . . _ . . . _ _ . . _ _ .
| |
| w__..-----..__-__. - . - . - . - _ _ _ _ _ . . . - . . _ . . _ - . . - - - - . . . - . - - . . - . - - . - - . _ - - _ - . - - - . - .
| |
| | |
| r~;;G:2:..w: = --~ . - - . - - L.-....-- _ . . a. ; , . .- - . - ,
| |
| BRAIDWOOD SIMULATOR MALFUNCTION t
| |
| 7
| |
| | |
| ==Title:==
| |
| .Lsas of Feed to 480 Volt ESF Bus or MCC ID: EPS-11 f-(
| |
| AM NO: 6,3.4.6.11
| |
| | |
| == Description:==
| |
| Selected 480V ESF bus or MCC feed breaker trips.
| |
| Variations: Date: 1/5/88 Rev: 4 i
| |
| if Selectable -
| |
| Steps Inputs Cossnents
| |
| : 1. Selections EPS-11A, EPS-11B, EPS-11C, EPS-11D 3
| |
| : 2. Select bus 1-10 1 - Bus 131X 6 - MCC 131X4 2 - Bus 132X 7 - MCC 132X1 3 - MCC 131X1 8 - MCC 132X2
| |
| (~ 4 - MCC 131X2 9 - MCC 132X3-
| |
| \ 5 - MCC 131X3 10 - MCC 132X4 Brief Plant Response:
| |
| 1 or 2 - Loss of the major 480 vac ESF buses will cause cooling problems in containment, control room,.and aux, building. Also a loss of your l
| |
| > normal source of power to the A.C. instrument inverters. If coupled with a loss of a vital D.C. bus could result in a Rx trip if one instrument had already failed.
| |
| 3 More individual with no loss of HVAC but possible Rx trip due to above reason.
| |
| Note: Nunerous annunciators are received associated with a loss of power.
| |
| O 881M/14 5/89 i
| |
| w-- __ - --- - -- - - - - - - - - - - --
| |
| | |
| , ,_ --.,n.. .- _ . -- _ _ . _ . _ _ __. -- . . _ - . _--- .
| |
| ' b -! : .Y) ' l f,''(,
| |
| 'BRAIDWOOD SIMULATOR MALFUNCTION-4 9 4
| |
| ' } ,l ID: EPS-11 p;
| |
| | |
| ==Title:==
| |
| Loss' of Feed' to 480 Volt ESF Bus or MCC V
| |
| t E Suggested. Instructor. Action .
| |
| ~ None.
| |
| s Events: None t
| |
| +
| |
| e k
| |
| 4 0
| |
| l '
| |
| l' 1
| |
| O 881M/15 5/89 l
| |
| ==&:L- ___z_2_____-_r.-______.-___..______-.___________.__.__,____________ _ __ _
| |
| | |
| BRAIDWOOD SIMULATOR MALFUNCTION
| |
| | |
| ==Title:==
| |
| Failure of System AUZ Transformer (SAT) ID EPS-13 NO: 6.3.4.6.13
| |
| | |
| == Description:==
| |
| Fire in the SAT causes loss of vital electrical feed.
| |
| Variations: Date: 3/16/89 Revt 5 Selectable Steps Inputs Conunents
| |
| : 1. Select transformer 1, 2 1 = SAT 142-1 2 = SAT 142-2 Brief Plant Response (IC-17,' 100%, all systems in automatic)
| |
| ('
| |
| N Loss of either SAT causes the other SAT to trip. Both DG's start, all safety cquipment sequences on. First annunciators received include SAT LOCKOUT RELAY TRIP, SAT TROUBLE, 6.9 KV BUS LOW VOLTS, AND 4.1 KV BUS LOW VOLTS.
| |
| NOTE: Malfunction assumes SAT is on the line.
| |
| Suggested Instructor Action:
| |
| l
| |
| : 1) LOA EPS-1 (EPS Breaker Reset)
| |
| : 2) LOA EPS-67 (Switchyard Breaker Reset)
| |
| : 3) LOA EPS-80 (SAT Lockout Relay Reset)
| |
| : 4) LOA's EPS 82,83 (SAT Regets)
| |
| Events:
| |
| None 1
| |
| 881M/17 5/89
| |
| ~ '~ ~
| |
| TT __ ___ _____ _L_- -_
| |
| * ugy- . . .
| |
| BRAIDWOOD SIMULATOR MALFUNCTION b
| |
| P^=
| |
| | |
| ==Title:==
| |
| = Failure of Unit AUX Transformer (UAT) ID: EPS-14 :
| |
| 'l 9
| |
| =
| |
| NO: 6.3.4.6.14 {
| |
| I
| |
| | |
| == Description:==
| |
| Fire in UAT.
| |
| Variations:, Date: 3/16/89-Rev: 5 Selectable -i Stops. Inputs Comments
| |
| . 1. ; elect transformer 1 or 2 1 - UAT 141-1 2 - UAT 141-2 i
| |
| Brief Plant Response
| |
| ~
| |
| The generator will trip due,to UAT differential / Gen Trip. If reactor power is
| |
| %/ greater than 30%, the reactor will trip duc to the turbine / generator trip.
| |
| All loads should transfer to SAT's for the 6.9 KV buses and the 4160 volt non-ESF loads. Loss of one UAT will cause loss of other UAT. First annunciators include UAT FEEDER BREAKER TRIP, UAT OIL FLOW LOW / TEMP HIGH AND U!? DIFFERENTIAL GENERATOR TRIP.
| |
| Suggested Instructor Action:
| |
| None.
| |
| Events: None
| |
| 'O 881M/18 5/89
| |
| = 2 ~2L-_ __ _ _ _- L_- ___ _ LL __ _ _ __ _- _ __ _ _ _ -
| |
| | |
| . . ~ . . . .. . . . .; . .-
| |
| = . . - . -. ; - - . ~. . z::..: - : : : ---- -. a ..::r-- :::- - ;=-
| |
| =
| |
| BRAIDWOOD SIMULATOR MALFUNCTION r%
| |
| | |
| ==Title:==
| |
| 120 VAC Instrument Bus Inverter Failure ID: EPS-17 4 )
| |
| K0: 6.3.4.6.17 U-
| |
| | |
| == Description:==
| |
| Lose selected inverter due to blown fuse.
| |
| Variations: See Malfunction EPS-8. Date: 3/16/89 Rev: 5 Selectable Steps Inputs Comments
| |
| : 1. Selections EPS-17A, EPS-17B
| |
| : 2. -Select inverter 1-4 1 - inverter 111 2 - inverter 112 3 - inverter 113 4 - inverter 114 Brief Plant Responses (IC-17, 100%, all systems in automatic) e g'~%
| |
| Loss of various instrumentation in the Control Room including one section of the NI's. FirstannunciatorsreceivedincludePROCESSI/CCABPWRSUP FAILURE, SOLID STATE PROT CAB GENERAL' WARNING and BUS INVERTER TROUBLE.
| |
| Suggested Instructor Action:
| |
| To reenergize the instrument bus, uses LOA EPS 21 - Bus _11 EPS 22 - Bus 112 EPS 23 - Bus 113 EPS 24 - Bus 114 Events:
| |
| : 1) LER 06-02-87-007: Rx Trip Due to Loss of Inverter and Failed PR.
| |
| : 2) Byron Trouble Report: Loss of Inst. Bus 211.
| |
| : 3) DVR 06-01-87-147: Loss of Inst. Bus 111.
| |
| : 4) LER 20-01-87-010: Loss of Inst. Eus 111.
| |
| : 5) LER 20-02-88-008: Loss of Inst. Bus 212.
| |
| O 8tt1M/22 5/89 e _ _____:_r __ L_ _::_ - - _ = _- - _ :T _ :L _ - _ ~ _ _ _ _ . _
| |
| ^
| |
| | |
| ! . . ( *. .i i b
| |
| * Jia N- ,.
| |
| A^
| |
| Fac11t*f N ee (f: %: net V :ee :2' sin Ber e U"'t 2 Of 51 01 St ]l a! S' ? Of p- .)
| |
| ' bMLb bT H H C L E O 5k E Event Date (5) LER Numeer (6) Pecort Date (7) i
| |
| {})
| |
| y Montn Day year year g// /
| |
| Secuential
| |
| ,/g/
| |
| Revision Month Day Year Other Fact 1*ttet Involved f!)
| |
| racility Names i e m et Nunterf s )
| |
| Numcer /1 Numbat
| |
| ~~~~ ~
| |
| ~--.s NONE of si of cl 31 I l als' ole al sl7 of of 7 ole als 012 sit of El al of of I I
| |
| \
| |
| o m ATING N' THIS REPORT !$ SUEMITTED PUR5UANT TO THE REQUIREMENTS OF 10CFR oer er d re of tem fo11awinal fitt 1
| |
| 20.402(b) __ 20.405(c) _.L. 50.73(a)( 2)( tv ) 7 3. 71( b )
| |
| POWER __ 20.405(a)(1)(1) 50.36(c)(1) ,, 50.73(a)(2)(v) 73.71(c)
| |
| LEVEL 20.405(a)(1)(11) 50.36(c)(2) 50.73(a)(2)(vit) Other (Scecify f101 o! a! e 20.405(a)(1)(111)
| |
| _ 50.73( a)( 2 )( t ) 50.73(a)(2)(viit)( A) in Abstract
| |
| ,/ _ 20.405(a)( 1)( t v) 50.73(a)(2)(tt) _ 50.73(a)(2)(v111)(8) below and in
| |
| /,
| |
| __ 20.405(a)(1)(v) 50.73(a)(2)(111) 50.73(a)(2)(u) Text)
| |
| LICENSEE CONTACT FOR THf t LER f f 21 Name f ELEPHONE NUPatER AREA CODE T. Joyce. 4tititant tunerintendent tunnart Services Ext. 2214 8l 1 l 1 ?! 11 41 -l Si al a COMPLETE ONF LINE FOR rJCW COM80N NT Faf LL'RE DESCRf eED IN THIS PEPORT f il)
| |
| CAUSE SYSTEM COMPONENT MANUFAC- REPORTABLE j CAUSE SYSTEM COMPONENT MANUFAC- REPORTABLE ,
| |
| TURER 70 NPeOS TURER TO NPe01 l K Elr el El Cl T WI Il 21 o Y l l l l l l l :
| |
| X elf i Cl Al P WI fi 2I o Y f f I l l l 1 ! '
| |
| SUPPLEMENTAL REPORT EXPECTED (14) Expected Month i Day i Ven Submission
| |
| [m lYet fff yet. camniate EXPECTED sueMftsf0N DATE) K l NO Date (15) l I I f i f I V l A85 TRACT (Ltmit to 1400 spaces.1.e. approximately fif teen single-space typewritten 11nes) (16)
| |
| On May 4, 1987, at 0644 the Unit 2 Reactor tripped during the performance of the Quarterly Power Range Calibration Survet11ance on char.nel N1-43. Untt 2 was in Mode 1 - Power Operations at 881; Fower. As a reoutrement of the surveillance, the Power Range channel's b1 stables were tripped. During the concuct of the surveillance, the instrument bus 213 breaker tripped. This de-energized channel N41 and tripped its 01 stables. Th,ts satisfied the 2 out 4 logic coincidence for High Neutron Flus Reactor Trip, as a normal result, the Steam Generator level reached the low level automatic Aust11ary Feedwater ( AF) Pump start setpoint. The 28 Diesel Driven AF Pump started as expected. The 2A AF Pump did not start as a result of the instrument Dus 211 Trip. The Untt was recovered consistent with Emergency Procedures without incident.
| |
| The cause of the the instrument bus 211 trtp is due to a f ailure of a Stitcon controlled rectif ter and a capacitor in the Inverter. The capacitor failure caused the rectifter fatlure. The fatlure mechanism of the capacitor could not be determined and is attr1Dutable to normal wearout. The components were replaced and the bus re-energtred. There were no adverse safety consequences.
| |
| l i
| |
| (1410M/0164M)
| |
| __ . _ - _ . .__-_-___-_-.-_-_--_-_-_---_--_------------------------A
| |
| | |
| . :EUEE *,E V :!:19' .ia E.- m *:9 ;- :y ,
| |
| i
| |
| ~ ~
| |
| FACILITY NAldE (1) ;CC(Et U SER t2) .28 N M ED et , :.a (car J., sesenttal f 9 ae,i;3:n IV '
| |
| Nuecer 'i numter Evren. Unit 2 O Is101010 l al si s 3l7 .
| |
| 0! of 7 .l O!0 012 er 3!3
| |
| )
| |
| [V TEXT A.
| |
| Energy Industry Identification System (EII5) coces are identifted in the test as [ss]
| |
| PLANT CONDITIONS Pe!OR TO EVENT:
| |
| I Event Date/ Time 0E/04/a7 / 06aa Unit 2 h00E 1- - Power onorattant Ax Power E81 RC5 [AS) Temperature / Pressure germal ocerstma
| |
| : 5. Q[ifflPTION OF TVEht:
| |
| On May 4 1967. at 0644. Unit 2 Reactor trtoped during the perforftence of the Quarterly Power Range Callbratton Surveillance on channel NI-43[IG). The Reactor Trto signal resulted from a 2 out of 4 logic coincidence on High Neutron Flux.
| |
| The high flus btstable of channel N!-43 was tripped as a reautrement of the Power Range Calibration 5'arve t1 tance. During the conduct of the survat11ance, the instrument bus 211 [EF) tripped. This de-energized Power Range Channel N!-41 and tripped the channel *$ bistables, including the high neutron flux. This satisfted the 2 out of 4 logic coincidence and generated the Reactor Trip Stenal on a spurtous High Neutron Flus conettion.
| |
| The normal resulting transient on the Steam Generator water levels cause the low level setpoint to be reached and a automatic Aust11ary Feedwater (AF)[BA) start signal to be generated. The 28 Diesel Crtvan AF Pump star ced as designed. The 2A Motor Driven AF Pump failed to start because it also 1 cst control power from the instrument bus 211 trip. Operators (Licensed) manually started the 2A AF Pump in accordance with emergency proceJures. There were no other systems or components inoperable that contributed to this event. The Unit responded as designed and the operators recovered the Unit consistent with Unit Emergency.
| |
| [ '
| |
| Procedures. Nc other safety systems actuated. A Reactor Protection actuation and an Engineered Safeguard ''-
| |
| actuation are reprtable pursuant to 10CFR 50.73(a)(2)(1v).
| |
| C. cAust OF EVENT:
| |
| The intermediate cause of this event was the tripping of the instrument bus 211 breaker coincident with the performance of the Power Range Calibration Survet11ance on channel p!.43. Instrument bus 211 breaker
| |
| * tripped due to the fatlure of the Inverter. The Inverter fatlure was due to the failure of the Silicon Controlled Rectif ser (SCR) number 1 and capacitor C11. The SCR failure was a street result of the capacitor C11 failure. The root cause of why the capacitor f ailed could not be determined. The failure of the capacitor is attributable to normal wearout.
| |
| D. $AFETY ANALY1fs:
| |
| There were no adverse safety consequences as a result of this event. The Reactor Protection System functioned as designed and placed the Unit in a safe condition. Desptte the 2A AF Pump fatture to automatically start. Auxiliary Feedwater was provided from the 28 AF Pump automatic start and subsequently by the 2A Pump which was manually started. Failure of Instrument power to one enannel while another channel is in test is the most severe set of condtttons this could happen in.
| |
| I O v (1410M/0164M)
| |
| | |
| _ :tss!! :.t9f Es:cf _t=r 'ti' ::N :s;-- y raCILIrv saet (1) 00CtEf NureER (2) _F# Ntmsta es. J ia Year //r seguential /, aevtsion
| |
| ///
| |
| t!< numeer &// i! n ume tt_ g
| |
| ,s arron. unit 2 0 l 5 1 0 l 0 1 0 l al SI E El' - Of of 7 - 010 011 or ?!1 J Energy Industry Identification Systen (EIIS) coces are toentified in the text as (un)
| |
| { v l TEXT E. CORRECTivt ACTION 5; Inverter 211 was repaired by replactng SCR number 1 and capacitor C11. The Inverter was tested and returned to service.
| |
| These particular capacitors had been previously identified as regutring changeout because of their approach to the end of their normal Itfe expectancy. Unit I capacitors have already bes.t changed. , Unit 2 capacitors are scheoulac to be changed at the neut outage of sufficient duration. This is tracked by Action Item 6-87-2037. ,
| |
| F. PREVIGus OCCURRENCES:
| |
| LER NL9GER TlILL Nowt G. COMPntfMT Fa!Luet DATA:
| |
| a) MANUFACTURER NOMENCLATURE MODEL NUMBER mfg PART NUMBER Westinghouse $CR N/A 153BA73H10 Westinghouse Capacitor N/A 1815A70HZ3 20 MFD
| |
| (- .'t-k b) RESULTE OF NPRDS SEARCH' No useful information was found on failure of these Components A,
| |
| e O
| |
| emp-
| |
| .h (141DM/0164M) k b
| |
| | |
| d?5 '7 i
| |
| August 11, 1987 L
| |
| n I Preliminary Root j Regulatory Impli::ations Cause Assessment lgl Not a NonconD11ance l l_l Personnel Error l l_l Ievel IV Noncompliance l lgl EGuipment Failure l l_l Tech. Spec. Violation l l_l Procedural Inadequacy l l _l Possible Enforcement Action l l I Unknown l l l Possible Noncompliance l r
| |
| TROUBLE REPORT.
| |
| BYRON STATION A GSEP " Unusual Event" was declared at Byron Unit 2 on Tuesday, August 11, 1987 at 1157 hours. Unit 2 was ut approximately 98% reactor power (1140 We) prior to the " Unusual Event" being declared.
| |
| This " Unusual Event" was initiated as a result of the inoperability of instrument bus 211 inverter. The Technical Specification Limiting condition for operation requires the inverter to be capable of reenergizing the A.C.
| |
| instrument bus within 24 hours of being declared inoperable or be in Hot Standby (Mode 3) within the next 6 hcurs.
| |
| The unit is currently ramping down in power at a rate of 4 We per minute. At this time, it is not known whether or not the inverter will be operable prior to the 6 hour clock expiration.
| |
| A Red Phone call was made to the NRC. The State of Illinois along with communication Services were notified of the event.
| |
| T L. [tutVc /Amy D. Erich Wurz s, 5092B/3 l l
| |
| | |
| OEVIATICN INbESTIGAT13N REPCRT q ,.
| |
| : v. ~ >
| |
| TITLE LCSS OF INSTRUMENT INVERTER 111 pace 1 10Fl 0 l 2 EVENT DATE DIR NUMBER REPORT DATE
| |
| / // $EQUENTIAL // REV!510N
| |
| ( 40 NTH DAY YEAR STA UNIT YEAR f/ fMSER ff NUMBER MONTM' CAY YEAR POWER 11 1 01 4 al 7 al 6 of 1 al 7 -
| |
| 1 1 41 7 -
| |
| 0 Io 11 2 il 4 al 7 _ _ .
| |
| O f 91 7 CONTACT FOR THIS Oft NAME TELEPHONE NUMBER AREA CODE D. Brindle. Onaratine Encinaar. Ext. 201R al 1 1E 2 l114 1-l 6 14 l4 I1 COMPLETE Saf LINE FOR EACH COMPONE T FAI URE DEtrafern IN THf t REPORT CAUSE SYSTEM COMP 0NED.T MANUFAC- REPORTA8LE / CAUSE SYSTEM COMPONENT MANUFAC- REPORTA8LE TUREE TO NPRDS , TURER TO NPRDS r EIr 11 NI VI T W l 11 21 o Y I I I l l I I I I I 8 I I . '! . . I i 1 1 1
| |
| _'JPLEMENTAL REPORT [1PECTED MONTH ' DAY YEAR SUBMI5510N l YES fif vet _ enmelate EXPECTED suRMISSION OATE) l No TEXT A. f'LANT C0 EDITIONS PRIOR TO EVENT.
| |
| Event Date/ Time 11/4/a7 / nata Unit 1 MODE 1 - Fmsar eneration Rx Power .97 RCS [AE] Temperature / Pressure Marinal comratica Unit 2 MODE 1 - Power eneration Rx Power L RCS (A8] Temperature / Pressure Normal eeeratten L. DESCRIPTION OF EVENT:
| |
| No systems were inoperable at the beginning of this event wilich contributed to the severity of this event.
| |
| l At 0858 on 11/4/87, with Unit 1 at $7% power. the Instrument Inverter til failed. Since both normal AC feed to the instrument invertcr 111 and auto switchover to DC power supply f ailed. Instrurent bus 111 was de-energtred. Limiting Condition for Operating Action Requirements (LCOAR) 3.1 1-a 3.2.1-a and 8.3.1-la were entered. At 0908 the Instrisment Bus was placed on its reserve power supply. LC0AR's 3.1.1-a and l 3.2.1-a were exited. The Electrical Maintenance Department cetermined that a fuse and resistor had f ailed open. ho other failed components were found. The defective resistor and fuse were replaced and instrument inverter 111 tested satisf actorily. The Instrument Inverter was re-energized and connected to the 111 Instrument Sus and LCCAR 8.3.1-1a was extted at 1830 on 11-4-87. Plant conditions rematr.ed stable throughout the event.
| |
| C. CAUSF QF EVENT:
| |
| The root cause of the event was electrical component failures. The Instrument Inverter failed when a 2000hm/100 watt resistor, in series with the inverter gating control logic, f ailed open and the fuse blew.
| |
| i This de-energized the electronte gating control circuit which converts the DC power within the inverter. to 120 SAC output, thus disabling the inverter output. The instrument bus til was thus de-energized. The cause of the resistor failure is unknown. A previously formed task force headed by PWR Engineering Department at Syron and Braidwood Statiorts is independently investigating instrument bus reliability.
| |
| l (1750M/0205M) 1
| |
| _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ . _ _ _ _ . . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . . _ .J
| |
| | |
| CEVIATION I%VESTIGATICN cEFCRT TEXT C03t!NUATIO:8 TITLE CIR NUMBER PACE SEQUE3TIAL REV!510N STA UNIT YEAR NUMBER NLSeER 01 6 DI 1 al 7 ""
| |
| 1 14 17 F 6 IO 2 CF 0 j ,1, f ))
| |
| TEXT D. 1METY ANALYSIS' There was no impact on plant or public safety. Reactor protection and safeguard actuations were always available throughout this event. No Technical Specifications Limits were exceeded. LC0ARs IBOS 8.3.1-la.
| |
| 1905 3.1-la, and 1905 3.2-la were entered. The instrument bus 111 was placed on backup feed at 0908 on 11-4-87. The instrument bus 111 was energized from its inverter at 1830 on 11-4-87. Thus the instrument bus power and operability were restored within Tech Spec Action Statement time requirements. The other three instrument busses were operable throughout the event. Thus reactor protection was available if under a more severe set of initial conditions the protection system had been required.
| |
| E. CORRECTIVE ACTIONS:
| |
| The defective resistor and fuse were replaced and Instrument Inverter 111 was returned to service. Voltage level, frequency and waveform were checked and found acceptable. The instrument inverter was determined operable and LCDAR 1805 8.3.1-la was exited at 1830 on 11-4-87.
| |
| F. PREYious OCCURRENCES:
| |
| DVR NLDSER TITLE NONE on instnament inverter 111.
| |
| A preytous occurrence occurred on instrument inverter 114. DVR 06-1-86-114 on 6-12-St.
| |
| G. CSFONENT FAILURE DATA:
| |
| a) MANUFACTURER NW1ENCLATURE tM EL NtteER WG PART NLDRER Westinghouse Resistor attA325H11 b) RESULTS OF NPRDS SEARCH: .
| |
| The NPRDS search yieldad no information.
| |
| \
| |
| V, i (1750M/0205M)
| |
| | |
| wr ~ ,
| |
| .1 ,
| |
| s+-
| |
| 4 g:c Q.Q 456-200 - >>f3 :. O -
| |
| 1 1M - ''i ll f -DO :
| |
| W si Dh fE: 02/09/07' ;HEDULAR CnT T DT COi!Di ' i.Od i
| |
| '' mnn DE. : 5 t aintTHEHT TO:
| |
| pef (Sui 4 : DATE. SEPT ?
| |
| -CURiiEN! LOC (1.'EPT; QO:ilEC T . DVR 20-1 707-943s U!ADVERTANT LOSS-OF POWER TO IDSTRUMENT bus iii RESULTING IN A REACTOR TRIP DUE TO PERSOHNEL ERROR - COr!1RACTOR SEVERITY LEVEL- LER NO: G7-010 CRITERION:
| |
| Ct i?t.? DEVIATION t RESPONSE DUE DATE '1 H S P E C T O R lik S ;t OuG CAUSING ITEMi DW- EF
| |
| '10 B Y. SET !!Y GYSTEM
| |
| .OR M ORG/PERGON: OP / COOPER CORfiEC1IVE ACT RfGP DEPT /SUPV : TSEL/llILL. R/F OUTACE
| |
| 'UOL PER'60N /
| |
| PRIORITY COG PERSON' , /
| |
| C;' l 'EP sot 4 /,
| |
| TP: AWilSCI4 lif /DW PROCE!>URC :
| |
| D: L: Q: Z: HDD: DNS: EGS: PTC: NFS:
| |
| T f K 4.4 Cri1T - DW : C' . UY_ :
| |
| OR IG I:4 AL. DuE DATE- 03/10/07- 3TATUS: Coni'L E'I E l' t:OL ti : 9'.
| |
| '1HIERlh REPORT:
| |
| * RDY FOR CLO3URE: 02/20/a7 "f!! WRIri REPORT - 10LFRSO .~/3 C A > ( ? ) ( II ) ORIG EXIr DATE - OL'07/G7 ORIG CLQi,ED D3/1r T7 Os1HG' REPORT: IR 454/97-007 i WiTUl(E '
| |
| 'E.E. FITZPJ.TOICK DniE C04iPLEfcD -
| |
| iuTt"R: DVR '.20-1 0 43 li 'J 1* f PTIOrt :
| |
| 12 EVAC'.1NGi RuiiEttr GUS iii WAS LOST buE ~10 n CONTRACTOR DPOIPld6 A-BnG ONTO THE J HVERTER OUTPl!T BREAKER. THE BUS WAS DOWH FOR ArPROXIMAIE8_Y-15' SECONDS. LOSS OF THjS DUS GENERATES A REACFOR TRIP-FROM SOURCE AND IHIERNEDIATE RAMGE lIIGH FLUX. -THESE TRIPS uEpF ALSO, ALbEADY IN DUE TO H-32 AND N-36 INOPERADILITY DOWN POWER.
| |
| DOROtJ DILUTION PREVENTION SYST EM WAS ACTUATED.
| |
| x 10CFR50.72 NRC RED PHONE NOrJFICATION MADE AT 1632 ON 02/09/B7.
| |
| e n.
| |
| = OPERnTING ENGINEER' S COnddr! f 9 CONS 1RUCTION WILL BE FiADE AWARE OF TifIS EVENT AT THE h0RiiING iiEE11NG ON 02/10/87.
| |
| SO' DAY REPORTABL.E/iOCFRSO.73rA>(.?)(II)
| |
| | |
| . ., o 4
| |
| ,f .
| |
| y , ,1q .1 p :10F1)OH t CON T E V:
| |
| Lt! R l!UHsER : DF-OiO I
| |
| l ACTION SINIMARY' A. PL,iHT CON.OETIONS PRIOR 10 EVENT siUDE 5 - COLD SHUTDOWN, RX l'O W E R 0 % ,' 14 C S' ( A D ) TEMPERATURE / PRESSURE: 1O'7 DEGREES F/370 PSIG. . j n
| |
| 4 B. .DESLkIPI'10N OF EVENT, .DRAIDWOOD UNIT OHE WAS IN COLD SHUTDOWN. )
| |
| HERE EERE NO SYSTEMS INOPERABLE WHICH CONTR1DurED TO IHE EVENT. AT 1JT ON FEDRUnRY 9, THERE WAS A LOSS OF POWER 10 INSIRUNENT BUS (EFi !
| |
| 111. THIS' CAUSED'A DUS 111 INVERTER IROUDLE ALARM (IB) 1H THE COirl R OL. ROOM. THE LOSS OF INSTRUNEllT. BUS iii ALSO CAUSED A REAC10R
| |
| ^
| |
| 1RtP SIGNAL (JG) DUE TO THE LOSS OF col 41ROL POWER TO SUURCE RANGli.
| |
| (SP) (IG ) ' tl-31 AND INTERMEDIATE RANGE (IG) (IR) N-35. L.OSS OF CUlf!ROL PUWER TO THE SOURCE RANGE ALSO GoVE A BORON D1LUIION PRutECI'IOM SYGTEri (DDPS) (IG) S I G N Al. .
| |
| Al i ! ! ! '.i PO .l il i , A L1 CENSED NUCLEAR STAIIhr! OPERolOR (NGO) AND A "b" UPLi< ATOR OF.RI. D'l CP Al Lill? D Tel 11(E I D S f R I H lf:ilT 141 W I NVI R 4 I"R . IHE 110 0 R .U MI E D l lin i C tHSIRUCTION CRAF1 FERBudHEL UFRf: U UP P. l i hi IU lllE 6Rif A OF TME '';HVER'I ER AT THE TIME OF Tile BREAKER TRIP. Tile Hs0 UUEST100ED lhl JUNSiRUC1 ION CRAFT FERGOdHEL, AND fHEY JfA1ED f ile 1 6 DeiG WAS L O UE RE.D 1H FkUNT OF THE INVERFER. THEY ALSO SinlED siinT THEY (00K Ni' PCIION GELAIIVE fu THE EQUIPhENT. THE HSO At!D THE "B" NnN g INSPECTED 1HE INVERTER AND FOUND ALL UREAKERS 1H TilEIR PROPER l*001Y10N.
| |
| ,y.
| |
| PL IM I COr!vil f CNS WERE STADLE IHROUGilOUT THE EVEtt1. A REVIEW OF LHE SEblUENCE UF EVENI'S RECORDER (SER) (IN) REVEALED THAT INSTRUNENT 1013 ili WnS-REf..NERGIZED ONE MINUIE AND FOUli 9ECUNDS AFTER THE (RLP.
| |
| UPERAllONS VERIF1ED THE PROPER LIHL-UP FOR lilE INVER'IER 10 Ils NORMAL POULR SOURCE. OPERAIOR ACTION flEllllER INCREASED OR i FCREACI o (HE SLVER11 Y OF THE EVENT.
| |
| K 1:110. EVEN! JS RCPORIABLE UNDER i OCFR 50. 7J ( A ) ( .' > ( 1 V ) - nif f EVEUI UP l COHbiTTON THAT RESULTS IN MANUAL OR AU1OHAT1C ACTUATION OF ANY ENGINEERED SAFETY FEATURE (ESF), INCLUDING TIIE REACTOR PRO 1 EC110tf
| |
| 'iY S T EM (RPS).
| |
| 1 C. CAUSE OF EVEHT THE RuOI CAUSE OF THid EVENI IS PflMG nTTRIDUTED TO PERSONNEL ERROR DY A CONSTRUCTION CRAFT PERSON. IllE UiEIATIONS PERSOth4EL INVES11 GATING TH.l S EVEu r WERE UNABLE 10 DE1ERMlNE WilIGH OF THE SEVERAL CONSTRUCTION CRAFT PEOPLE l'N lilE nicEn NER IiW OL M A, TH0:Y WCrtE ALSO UNABL E 10 J;ETERhTHE IlOW THE INVER'IER WAh REENERG1ZEv. APPARENTLY, Tile CUf f1 R AC I OR UHO i'R I P I E D fME liR C Al'.l 9 0
| |
| | |
| .. t t 'a '
| |
| ~sC . ,b , , . , >~ .. u .
| |
| s>
| |
| T1Url . GUhhARY L(CONT) -
| |
| Ric A L .l i.:ED 1.HF LRROR AND RiiHf l li tu ' DRT:.W ER.
| |
| +
| |
| D. SAFEIY At4ALYSIS: 1HIS EVEH1 ilAD 00 EFFEC T OH 1HE GnFEli ul- si!F PLANT OR Pul!LIC SlHCE THE UNIT HAS NOT~ DEEN TAl(Er4 CR1TICAL. AND NO RADIUnCTIVE EFFLUENT HAD DEEN PRODUCED. It1 ADDITION, THF REhCTult TR1P - DilE AKERG WEllE IN. Tile 'IRIPPED (OPEN) POSITION. UNDER UOR9E.
| |
| C Af,C COr4D1 f100S , IJ1FH THE UllIT AT FULL POWER , THE BREAKER IJOlJLD liAVE ikiPPED AND THE llHIT , lJOULD ElluT DuldiJ AS DESIGNED. Tile COLD 3TnitT VOL I AGE 'IR ANSF URMER AND FhERGENC Y 125V DC BATTERIES, INGlRUNEdr UlW
| |
| )1i HACK-UP. POWER SUPPLIES, UERE AVAILABLE THROUGHOU'I THE EVENT.
| |
| ALL t:SF EQU I PHErl'l FUt1LT10tlED AS DEGIGNED.
| |
| M
| |
| : n. COPPLCf1VE nC'ITOHC; illE 1: LCC1RICAL LTHE-UP TO INSTitudEllr DVD 111 1.!A S V E R 1 0 l E D .
| |
| M litt (.005 : P1 A 1 'ON CR Al 1 1T R$0Hi l2L l TH f HE AltEA IJERE ORALl_ r 1 H S T P UC 1 tl0 00 'l llL AL T ) Die 3 TO Tol E IJilErl 1:TtU I PMENT IN THE nREA IIAS DEEll lhPACICD.
| |
| n BH ?.tF1C 1 rtu 1 V .l D U A L C O P P E C 1 't V E .)CIIOtt COULD HIJI f.E InhEN DUE TO filli lilADILITY TO l' REC 1GELY I DENT 16'I 1hVOLVED PART[ES.
| |
| a(.
| |
| Ti;t' bE COr*CE itriS llAVE DEEN HiG!iLIGillED TO CONSTRUCTION HANAGEnFN L , A'O G 1 NLREAGED nT'l EHI IOil 15 bE J r;G G I VEiJ TO Curls I R(JC T I ON IMPACl OH Ui'l 1 i Ol't.. Ital 10HS. lHIE 11 5 CV lLt? r!t:Lu 8Y Tite W Oii t I N G OF t At E R 9 0 0 , 0 <. 0 MANilOURS SINCE lilE LAST EVEi!T nF A '91 HILAR NATURF..
| |
| Jt F. I 'P E V T 011G OI:011RRElJCF S :
| |
| LilR NUhl.ER I1iLL w
| |
| m /06-00;-00 Cort i A1HnEl8 f PilkCE 1butAYTOtl FRun A Vui,i ACE
| |
| * TR6M31EHI CollSED BY LOi4S IRUCT TON AC11VI f Y.
| |
| * PEhCONNEL ERROR STriGIOG.
| |
| $ VU6-01i-Of CONTAlhhfilT VEllTILATION ISOLATION DUE 10 9 l'It 'I t h "
| |
| l
| |
| * SIGNAL - PERSOHHEL ERROR, CONTRACTOR ALTTV111E, l
| |
| * SUSPECIED.
| |
| l
| |
| * es/u?-OO2-OO CON T Al tfMFH f VFtffILATIuu IbulAT100 DUE lu n LnP 4 OF PflWER 10 A IIADIATION NONITOR - PERSOdHEL
| |
| * EltROR , Sti..sPil C I COtfiRnCIOR X
| |
| n G. C OriP DrlE H F Fn:l LUPE isAT A : unt'E .
| |
| O
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| h j
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| '. f.'
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| , 'g : - -.f d. !! I y is '
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| ..' . .. t , y , , ,
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| yy. i 4
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| -<j-
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| ; 3 -. . . ' ' .
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| * '', ;.; l ''I
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| ' ^
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| ') _' ' *3-f . ',1l' l \ - :
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| j ''^g'. y g' _'y y
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| ~
| |
| 1 4
| |
| 4 .
| |
| 9=
| |
| * b, , : a-. .[" N [p 6
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| ?
| |
| 4..
| |
| I i
| |
| - J .-
| |
| 1_>
| |
| l l It' l
| |
| ~
| |
| | |
| LICENSEE EVENT 8E83RT ! LED: g g.gg i Flctitty Name (1) Occatt Numter (2) 83ce f3)
| |
| Braiewcos. unit 2 ci si of 11 ci al si d 1 lcfl0 , 3 l Title (4) Inaceguate Capacitor Connection Results in Degraded Instrument tus Voltage and b3scouent~ Reacter Trio Event Data fE) LER Number (61 Reecrt Date (7) Other Facilittet Involved fB1 anth Day Year Year /
| |
| p/ Sequential /// Revtsion Month Day Year _ Facility Namet cocket Numeer(s)
| |
| I
| |
| / Number /// Nuiener l
| |
| NONE of El c.I 01 of l I 0 0 9 . l 01 2 ft o al a al a oi 14 0fG 0l3 Il 7 al a u of 51 el al al l I TMIS REPORT :S SU8MITTED PURSUA4T TO THE REQUIREMENTS OF 10CFR (Check one er mere c' the follow,mel (11) 1 20.402(b) 20.a05(c) J. 58.73( a H 2 )( t v) _ 73.71(D)
| |
| POWER _ 20.405(a)(1)(13 _. 50.36(LM11 ,,_. 50.73(a)(2)(v) 73.7Hc)
| |
| LEVEL _ 23.405(a )( 1)( u ) _ 50.36(cH 2) _ 58.73( a H 2)( vi t ) _ Other (Specify f101 0l0!O _ 23.405(a H1 Hili) 50.73( a H 2)( t ) _ 58.73(a)( 2 )( v t H )( a ) in Abstract 2 8 'a " * > " " " > 5 '"'" 2 n " ' 5' '2(aH2Hn n usi eeio ano in
| |
| #fffffMHffHfSHfMHf -
| |
| ** *0"* n '"*
| |
| 'o'"*n2n""
| |
| s''1"n nn '" u
| |
| %%%%%%'H%%'i%%% i t f CENSEE CObf aCT FOR T4f $ LER (121 0'ame TELEPwCNE NUMBER AREA CODE Harold M911. Tecttmical Staf f Erele.eer Es t . 2425 911 1g al gl gi .I 21 gj of 1 COMPLETE ONE tihr FOR EaCH COMPO 7 F AILORE DEstRIBEL IN THf t REPORT f t 3)
| |
| CAUSE SYSTEM COMP 0 KENT MusuFAC. REPORTABLE
| |
| /,/
| |
| CAUSE SYSTEM J0MP0hENT MAftOF AC- REPORTABLE
| |
| //g//
| |
| ' j TURER TU NPR05 / TURER TO NPRDS '///
| |
| t aI a ci al Pl a el of al 0 NO S I i 1 1 I l I Hk/
| |
| l l 8 l z l l I f ( l l I l i ! ! IIIII(
| |
| SUPPREeTAL REPOWf EXPECTEB f 14% Espected Mortn i Day i vear Submiss 1on
| |
| ( lYet (If yet. comolete EXPECTED SUBMf 5SION DaTE1 ASSTRACT (L1mit to 1400 spaces, t.e. approximately fif teen single-space typewritten lines) (16) i NO l i i g
| |
| At 0626 on February 23. 1988, durtng the performance of startup test BwSU RD-70. there was a loss of power to Instrument Bus 212. Thts resulted in a reactor trip s1gnal betng generated. anc caused the reactor trip t'reakers to open. Titts lors of power also caused a boren dilutton prctection system actuation. An equipment coerator was sent to the bus and he re-etiergized it front its constant voltage transformer. Acticn to prevent recurrence stil be to conduct an inspection of all " Fast-on-Con tectors' f or neat damage to the same connections for each Inverter on oath units.
| |
| There have been no previous occurrences.
| |
| p 19ada(0318C81/6
| |
| | |
| l ticENSEE EvtNT REPORT f LE#1 f E x? CONTNuaTI'JN l LER NUMBER f6) a ne 3_,
| |
| 8 ' FaCTLITY NAME (1) ' 00CKET CUMBER (2) ,
| |
| fear
| |
| ,/,/ p/
| |
| Sequential /,/ / Rev i sion
| |
| /// Number /// Numcar__
| |
| 0IoI5 0l e of 2 CF 01 1
| |
| ; eraidweed unit 2 01510 10 1 O l al 51 7 aia - -
| |
| j.T Energy Industry Identification System (E!!s) ccces are Identified in the text as (xx]
| |
| A. &LANT CONDITIDNs PRfoe TO [ VENT:
| |
| Untt: Braidwood 2 . Event Date: rebruary 20. 198R : Eyed Time: 0626 MODE: _1. -- Hot Standby j Rx Power:E .: RCS (A5} Temperature / Pressure: 557'F/2M5 mig E. c[5tRfFffon DF EVENT:
| |
| There wer; no systems or components inoperable at the Degtnetng of the event editch contrtouted to the severtty i of the event.
| |
| at 0626 om February 20. 1988. during the perf ormance of BWSU 40-70. Control Rod Ortve Mechanssm Operational Test, there was a loss of power to Instrument Bus (EF] 212. Thts caused a tus 212 Trouble alann (18] in the etntrol room. The loss af Instrument Bus 212 also caused the Reactor Trip Breakers (JG) to spen cue to tne loss cf control power to source range N-32 and intermediate range N-36 (!G}. toss of control power to the sCurce ranga also resulted in 4 Boron Otlutton Protection System (EDPS) signal.
| |
| An ecutament operator was dtspatched to inverter 212 and found the inverter output voltage had degraded to 50 volts. Inverter 212 is the normal f eed to Instrument Bus 212. The input AC and DC voltages were within their spectfied ranges. The inverter was 5t.ut down by the operater.
| |
| The equtament operator attempted to re-energtze the bus frem the Constant voltage Transformer (CVT) (EA] but was
| |
| 'D_ unsuccS5ful 45 its output breaker trtpped. The startup precedure for the CVT was repeated by the operator and the bus was energized. At 0651 on February 20. 1988. the plant was returned to a stable conettton.
| |
| s Operator settons neither increased nor decreased the severtty of the event.
| |
| The appropetate NRC nettf tcatten via the ENS phone systent was made at OTt! on February 20, 1988, pursuant to 10CFR50.72(c)( 2)( t t ) .
| |
| This event is treing reportes pursuant to 10CFR50.7'J(a)(2)(tv) - Any event or condition that resulted in manual or automatic actuation of any Engineered Safety Feature. including the Reactor Protection $ystem.
| |
| C. [ AUSE OF EVENT: )
| |
| The roct cause of this event was a pre-servtce insta11at1on error by the vendor which resulted in a bad connection on a capscitor for the auto transformer in 2tP06E. The improper connection ytelded excessive resistance which produced heat and caused it to burn of f. This produced an $mbalance in one phase of the transformer and a degraded output voltage condition.
| |
| The cause of the constant voltage transformer output breaker trtoptng ts indeterminate as the symotoms would not ter' .
| |
| Should this recar, then it will be addressed in a new report.
| |
| D. SAFETY amaLYSIS:
| |
| Thts event had no ef fect on the saf ety of the plant or pubite stnce the unit had not yet been taken critical and no raritoactive ef fluent had been produced. All Enginetred Saf ety Feature equipment f unctioned as designeo.
| |
| ; i tJnder worst case condtttons of the unit being at full power, the unit would have responded in tne same manner as in this event. The CVT and emergency 125 VOC batteries used f or instrument aus nackup power suppites were available throughout the event.
| |
| )
| |
| l 199am(0321881/7 L____ __
| |
| | |
| - - . . . _ , . . . -~ . - _ . , _ _ . . . . - . . . - - _ _ . - _ _ . - , . _ , , . , _ .
| |
| 7 _-
| |
| LICE 41EE EVEgf #EP0ef f LEh Text "awtiweaticN j FACILITY RAME (1) DOCKET NUMata ( 2) ,
| |
| LER NUMBER fu sne > ? >
| |
| l Year Sequent 141 //,p/ Pevision j g//
| |
| / Nurnbe r j// / Number l
| |
| i . arateces. Unit 2 o i 5 i o I o I o I el El 7 aIa - oIe!a -
| |
| oI o of 3 er of 3
| |
| * EXT Energy Industry Identification System (E!!S) cooes are scentttted in the text as (o]
| |
| E. CDRRECTIVE ACTIONS:
| |
| - The inenedtate corrective action was to restore the testrument bus to the CVT. The inverter was repaired by replacing the capacitor and connecttng leads.
| |
| Action to prevent recurrence includes a full inspection of all f ast-un connections and an inspection for neat damage to the sa 4 connections for each taverter on Octh untts. Yhts will ne trackva to completton ny Action i Item 457-200-80-02401.
| |
| , There are no corrective actions proposed for the output artaker of the Cvf since the symptoms could not be repeated. Should tMs recur, then it wt11 De investigated anc a new report will De submitted, i
| |
| Fi PR[vrous occuREElecFK: !
| |
| There have treen no prevtous occurrences of inverter capsetter or Teac fattures regardless of cause.
| |
| C. COMP 0hENT FAILUEE DATA:
| |
| MANUFACTURER NOME 4ClaTURE N00EL NUMBER NFC PART NUMBEe General Electrie 13uf d Trtnsetng 770036 23L60 %
| |
| f] Capacitor N)
| |
| O L-)
| |
| ts94m( 0371881/8
| |
| | |
| - 1 BRAIDWOOD SIMULATOR MALFUNCTION l
| |
| | |
| ==Title:==
| |
| Main Generator Exciter Failure ID: EPS-18 NO: 6.3.4.6.18
| |
| 'd'
| |
| | |
| == Description:==
| |
| Faulty signal ca.uses a trip of the
| |
| -PMG output breaker (41M).
| |
| Variations: Date: 3/16/89 Rev: 5 Selectable Steps Inputs Comunants None Brief Plant Responses (IC-M, 100% all systems in autor atic)
| |
| Trip of PMG output breaker causes a turbine trip. (Rx trip if above P-8).
| |
| (% The first annunciators received are GENERATOR LOSS OF TIELD GEN TRIP'and PMG k OUTPUT BRKR TRIP.
| |
| Suggested Instructor Actions None, Events: None n
| |
| ~
| |
| U 881M/23 5/89
| |
| | |
| ; ~. w n . .- .--~ - . , . - -
| |
| BRAIDWOOD TIMULATOR MALFUNCTION
| |
| | |
| ==Title:==
| |
| Loss of SX' Cooling to Diesel Generator ID: EPS-20
| |
| ,,-~.q
| |
| <ls,_,) . NO: 6.3.4.6.20
| |
| | |
| == Description:==
| |
| The-SK cooling supply valve fails to open on a DG dtart signal or close when the DG is shut down.
| |
| Variations: Date: 8/15/86 Rev 4 Selectable Steps Inputs Comments
| |
| : 1. Select DG 1 or 2 1 - 1A DG valve (SK169A) 2 - 1B DG SX valve (SK169B) e~' Brief Plant Response:
| |
| j ,
| |
| \
| |
| If DG is running in " test" or manual start scia, the diesel will shutdown on high temperature.
| |
| If DG is started in emergency mode, the DG trouble light will alarm in control room but will not shutdown until it burns itself up.
| |
| Suggested Instructor Action:
| |
| None.
| |
| Drents: None l
| |
| 1
| |
| \- 881M/25 5/89 1
| |
| * ..w..
| |
| | |
| BRAIDWOOD SIMULATOR MALFUNCTIO4 Title Safeguards Shutdown Relay Failure ID: EPS-21 (j- NO: 6.3.4.6.21
| |
| | |
| == Description:==
| |
| Safeguards shutdown relay fails to operate on a loss of offsite power (blackout).
| |
| Variations: Date: 10/4/88 Rev: 5 Selectable Steps Inputs Comments
| |
| : 1. Select malfunction EPS21A, EPS21B
| |
| : 2. Select relay 1-2 1-SDRA (Safeguards shtdwn relay trn A) 2-SDRB (Safeguards shtdwn relay trn B)
| |
| : 3. Select failure time 0-40 sec. See Note 1 on next page.
| |
| Brief Plant Response:
| |
| Relay 1 or 2 selected: that train's ESF equipment doesn't auto start depending on failure time selected.
| |
| 881M/26 5/89
| |
| _ _m_____ _ _ _ _ _ _ _ . - . _ _ _ _
| |
| | |
| NOTE lt DIESEL GENERATOR LOADING SEQUENCE LOAD EAFEGUARDS SHUTDOWN
| |
| .e my Centrifugal Charging Pumps O sec OSX063's O see Control Roon Chilled Water Pumps O see Safety Injection Pumps Residual Heat Removal Pumps Control Roon Chiller 15 see Containment Spray Pumps 15 - 18 see Component Cooling Pumps 20 see Essential Service Water Pumps 25 setc Auxiliary Feedwater Pump 35 see Containment Spray Pumps 40 see NOTE 2 A loss of offsite power with an SI will not be affected by this malfunction.
| |
| (
| |
| N Suggested Instructor Action:
| |
| None.
| |
| Events: None l
| |
| O
| |
| -- 881M/27 5/89 4
| |
| ~ '
| |
| | |
| BRAIDWOOD SIMULATOR MALFUNCTION k
| |
| . . . ~
| |
| | |
| ==Title:==
| |
| Diesel Generator Failure to Flash Generator Fleid ID: EPS-22 NO: 6.3.4.6.22 l:
| |
| l'
| |
| | |
| == Description:==
| |
| D.G. field flash circuit fails to energize the exciter circuit.
| |
| Variations: Date: 8/15/86 Rev: 3 Selectable
| |
| * Steps Inputs Comments
| |
| : 1. Select D.C. 1 or 2 1 - 1A D/G 2 = 1B D/G Brief Plant Response:
| |
| i D.G. should not produce any electrical output or be very erratic and slow down
| |
| - energ'zation of safeguard equipment.
| |
| Suggested Instructor Action:
| |
| None.
| |
| t Evente: None
| |
| ./
| |
| 881M/28 5/89
| |
| | |
| l: BRAIDWOOD SIMULATOR l MALFUNCTION LISTING CONDENSATE AND FEEDWATER
| |
| 'l '
| |
| }
| |
| .%f
| |
| 'FWM-1 Main Feedwater Pvnp Trip FWi-2 Condensate Pum,. ~, ip FWM-3 Steam Generator Feedwater Control Valve Failure FWM-4 Feedwater Flow Transmitter Failure FWM-5 Turbine Driven Feedwater Pump Speed Control Failure
| |
| 'FWM-6 Heater Drain Pump Trip FWM-7 Feedwater Heater Bypass Valve Failure FWM-8 Feed Line Break inside Containment FWM-9 Feed Line Break Outside Containment FWM-10 Feedwater Pump Speed Control Oscillates
| |
| -FWM-11 Feedwater Heater Tube Break FWM-12 Auxiliary Feedwater Pump Trip FWM-13 Loss of Feed Water Pump Speed Control FWM-14 Main Feedwater Oil Pump Failure FWM-18 Feedwater Regulation Bypass Valve Failure FWM-19 Feedwater Preheater Bypass Malfunction FWM-20 Aux Feedwater Valve Malfunction ;
| |
| FWM-21 Aux Feedwater Line Rupture FWM-24 Start-up Feed Pump Trip FWM-26 Steam Generator Tempering Line Break FWM-28 Leak in CST h>
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| 63BM/263M/7 8/87
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| )
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| | |
| - x ;. _ .. ;.: :.:. . .:. ..
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| 7 :--- i BRAIDWOOD SIMULATOR MALFUNCTION I u
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| 1 ID: FWM-1 W- .
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| | |
| ==Title:==
| |
| ' Main Feedwater Pump Trip j 6.3.4.7.1 k] Feedwater pump trips due to inadvertent NO:
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| l :
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| | |
| == Description:==
| |
| | |
| I local trip.
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| Variations: All MFP'S trip on faulty relay K621
| |
| .(train A). Date: 9/20/88 Revs 8 Selectable Steps Inputs Comments
| |
| : 1. Select pump FWM-1A FWM-1A = 1A MFP
| |
| -FWM-1B WM-1B = 1B MFP FWM-1C FWM-1C = IC MFP
| |
| : 2. Select delay time '0-99,999 sec.
| |
| O Brief Plant Responses (IC-17, 100%, all systems in automatic) 1, 2, or 3 - Selected pump trips, causing a reduction in feedwater flow. Drop in feed flow causes RCS temperature to increase, which in turn causes the control rods to insert. Steam generator level decreases and eventually will cause a reactor trip on low-low steam generator level. First annunciators received include: FW PUMP TRIP and S/G ILOW MISMATCH W FLOW LOW.
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| m 0110w:4 882M/2 5/89 1
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| :. m - ,. u - .- - - . . w .- . _ y . _ . _ - . . . . _ . . __, ,. . .. . . . , _ . _ _ , , _
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| "0BRAIDWOOD SIMULATOR MALFUNCTION-
| |
| | |
| ==Title:==
| |
| Main Feedwater Pump Trip ID: 'FWM-1
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| ?
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| .i
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| ; Suggested Instructor Action:
| |
| . None.
| |
| . Events:
| |
| .1) STR (Eion/5/06, Inadvertent Pump Shutdown
| |
| : 2) DVR 6-1-85-191/LER 86-01-85-06: EA Microphone Cord Pulls Local Trip Lever.
| |
| 3). Arkansas Ones Broken Governor Assembly Shaft Causes Overspeed Trip
| |
| : 4) LER 06-02-87-009 Manual Rx Trip Due To MFP Trip
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| : 5) LER 06-01-87-018: Ex Trip Due.To MFP Trip
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| : 6) LER 06-02-87-018: Ex Trip Due To MFP Trip
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| ,.7).LER 06-02-88-001: Ex Trip Due To MFP Trip
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| : 8) LER 06-02-88-004: Rx Trip Due To MFP Trip
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| : 9) LER 06-01-88-004: Rx Trip DUe to MFP Trip l
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| l l
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| l 4
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| 0110w:4 882M/3 5/89 u - , . - . . . . . . . . . . . _ . _ . ...
| |
| | |
| c I I May 20, 1986 Preliminary Root
| |
| . Regulatory Implications Cause Assessment I lgl Not a Noncompliance l lgl Personnel Error l l_l Level IV Noncompliance l l _l Ecuipment Failure l ,
| |
| l_l Tech. Spec. Violation l l_l Procedural Inadequacy l l l Possible Enforcement Action l l l Unknown l TROUBLE REPORT ZION STATION on Monday, 5/1' '86, at 111~1 hours, Zion Station Unit 2 experienced an unplanned reactor tr p while at 1090 mwe.
| |
| Unit 2 tripped when an instrument mechanic, while working on a modification on the 2c feedwater pump, inadvertently shut down the power supply to the 28 feedwater pump rather than the 2c feedwater pump. The unit' tripped on low level in the steam generators'Eien~thiilB feedwater pump lost O its power supply.
| |
| ' \,.)
| |
| The root cause for the instrument mechanics error is not known at this Unit 2 returned to service at 0412 hours today and is holding at SOC power for conductivity.
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| A Red Phone call was made to the NRC, and P. communication Services was notified of the event.
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| : i. E
| |
| / Mark C. Stoddard 0606B/3
| |
| * a O
| |
| . - - - _ - _ _ _ - - _ _ _ _ _ _ - _ . _ - _ . . - - _ . _ . _ . _ . _ _ _ _ . _ . . _ _Y
| |
| | |
| DEVIAT10H *EPORT Ccmm:nwGalth Edis:n ,
| |
| * in em C6 - 1 - 85 -
| |
| F a n . j ~T ;.g :r :g.iAT.;N
| |
| :~' ~'- 't'-
| |
| .;;.4aE: l 6-24-85 1:26 RX TRIP .. . , , .
| |
| [O pg. v aF:; ig; PLANT STATLS AT TivE :F EVENT aca. n ; Ii No.
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| E b .:e 3.
| |
| uc0E____1____ . *Cate N _____9_7_%_
| |
| RP 3 acR oT;;s 2, E,Es- 1 I
| |
| RX trip from 971 power on low low S/G level on lA S/G. The low level was caused by l
| |
| IC FW Pp trip,' which was actuated by the Equip. Att. while taking local readings at the pu.p. His radio microphone cord caught the overspeed test 1r or -- tripping the I
| |
| pump. ]
| |
| Tne turb load was run back, and t.he 1A FW Pp was started, but the recovery attempt was not successful.
| |
| POTENTIALLY SIGNiF0 CANT EkENT PCR NSO DIRECTIVE A.07 YES No 1
| |
| #C;78$0. 72 *.RC RED PMChE I NOLR NOTirtCAriCN waot 4 woua 02:05 No P. Allen 6-24-85 e at FCPON51 ALE * *ERVisep OATE PART 2l PERATING ENGINEE4'$ COMMENTS O
| |
| Station will investigate. Will also look at other eauirmant for centroin which I
| |
| are easily ooerated or easy to stad NMwhile workino around th. .cu i ~.e . ww * . j 1
| |
| 1 written to provide guards on FW Pps. ]
| |
| acN atPCRTAs.E ESEst NOTIFICAT10N REGION ..; CATE TW 30 OAv SEPC4"B L E ' I CC F R , JQ ,,7,3, (a)(2)(iv)
| |
| S DAY ALPORT PER t CCF R 2 8 Nso CATE i 'E ANNLAL.5PEC:AL REPORT REQutRED CECO CORPCRATE NOTIF8CAfl0N MAOE IF ABOVE NOTIFICATION t$ PER 10CFR21 7gggcepy D.P. Galle 6-24-95 1050 cec 0 ConP0aATE crr,cEn cATE ix L E.= = 85,-061-00 ,
| |
| i PRELiu,.An =E*:at D. Brindle 6-24-85 i Ccvpt,E E: AND AEsiE6ED Jari CPEDATiNG ENGINEER A 5 g, Y $T5* O E 1 i,
| |
| (~' /1 AtC~
| |
| p .i
| |
| >h+lc
| |
| .t
| |
| ,y' j ',, r ar ,/2 vlecI Q f
| |
| / //fh GEi:L,TtCN Anti:n' \D h "}>
| |
| Aa-e< :E3 ,c. :.5 % T.eN :- 5?AT6CN 5LPERtNTENOEN* OATE
| |
| ,i,;;. ....6 scaw 4. s ;. .4
| |
| | |
| ~ ~ _ _ _ _
| |
| t!CENSEE E'/EMT REPCat t'.ER) racility Name (1) Oceket Numcer (2) he > !
| |
| byren. Unit 1 S! ?! S! of of al !! a i l pl 3 l7 ,
| |
| " II I DEAcTCR TRIP Date (5) LER Nunser (6) eeeert Date (7) Other Fic11$ ties !avolves rg}
| |
| l l e . .. Day Year Year / Sequential /j/j/ Revision Month Day Year ric,1 sty Names I cecret Numrerfs3 f' j// '
| |
| Nunear / Numeer None 0! Sl Of of 01 I t
| |
| ~~~ "'"
| |
| of 6 21 a al s al 1 0 l6I1 0 iO O I7 21 a al 1 of El O! 01 of ! f THIS REPORT 15 5UBMITTED PURSUANT TO THE REQUIREMENTS OF 10CFR 1 (Check one or more of the followinal f111 '
| |
| 20.402(b) ._,_ 20.405(c) _L, 50.73(a)(2)(tv) 73.71(b)
| |
| POWER ,,,, 20.405(a)(I H i) , 50.36(c)(1) , 50.73(a)(2)(v) 73.71(c)
| |
| LEVEL 20.405(aH1)(11) 50.36(c)(2) 50.73(a)(2)(vii) Other (Specify f101 0 l9 7 .,,,,,. 20.405(a)(1)(111) 50.73(a)(2)(1) 50.73(aH2)(vii1HA) in abstract .
| |
| / / / /// / // / //// _
| |
| f 20.405(a)(1)( tv) 50.73(aH2)(ii) ,_ 50.73(a)(2)(viii)(8) below and in ,
| |
| /
| |
| %% jj//jj%j/j$jjj/j/j/jjjj//jj//jj/jjjj/
| |
| H%%%S%%%%%
| |
| f
| |
| ** **'(''0 Mn '* '" *M*" " " '' '" *M'M " ''*U 1 LICENTEE CONTACT FOR THis LER (121
| |
| *ame TELEPHONE NUMBER f AREA CODE I Richard M. Williams. Systent Test Eneineer. Ext. 2185 R i1 lE fl 11 al -l El al COMPLETE ONE LINE FOR EACH COM ON T FAILURE DEttRfRED IN THft REPORT (111 CAUSE SYSTEM COMPONENT MANUFAC- REPORTABLE CAUSE SYSTEM COMPONENT MANUFAC- REPORTABLE Tuere TO NPROS TURER TO NPRD1 a siJ l 1 I I I l N I l l l l "
| |
| l l 1 1 I i f f I l i i f SUPPLEMENTAL REPORT EXPECTED f141 Expected P*on th I Car ! Year Subetssion
| |
| (
| |
| tres fff Yet. Comolate EXPECTfD $UBMf15fCN DATEl I NO 11 0 l311 91 5 ABSTRACT (Limit to 1400 spaces, i.e approximately fifteen single-space typewritten Itnes) (16) j While operating in Mode 1 at a reactor power of 97%, a reactor trip occurred due to IA Steam Generator LO-2 level. The 1cw Steam Generator level was caused by the trip of the IC Feedwater Pump. af ter an Equipment Attendant inadvertently activated the local overspeed trip bar. Following the IC Feedwater Putnp trip, the main turetne/ gene-ator load was run back and the 1A Motor Driven Feedwater Pump was started. However. the recovery attempt was not tuccessful. This was an isolated incident. The trip lever handles were removed to prevent recurrence. An inspection will be initiated to identify other equipment with similar proolems, 1
| |
| s O l V i (0639M) l 1
| |
| l
| |
| | |
| - _ _ - ~.
| |
| ~
| |
| _P y.
| |
| %J
| |
| - ARKANSAS I -
| |
| .3 SCRAM. CAUSE: UNKNOWN. DESCRIPTION: HIGH REACTOR COOLAhT S'rSTEM PRESSURE REACTOP TRIP FROM 97% POWER FOLLOWING A TRIF CF.
| |
| THE "B" MAIN FEED FUMP (MFP). THE REASON FOR THE TRIP OF THE "B" MFP IS UNDER INVESTIGATION. ACCORDING TO THE FLANT, IT IS NORMAL TO EXPERIENCE A REACTOR TRIP FOLLOWING YHE LOSS OF A SINGLE MFP UNDER FULL POWER OPERATING CONDITIONS. ALL SYSTEMS
| |
| )
| |
| FERFORMED THEIR DESIRED FUNCTION IN THE REOUIRED MANNER.
| |
| ** UPDATE AT 0430 EDT ON 09/00** A BROKEN SHAFT IN THE GOVERNOR
| |
| ~
| |
| ASSEMBLY CAUSED.THE "B" MAIN FEEDWATER PUMP TO TRIP ON AN C OVERSPEED CONDITION. SATE OF EVENT: 09/02/95.' TIME OF r^s EVENT: 1902 EDT.
| |
| i 1 wJ II. SIGNIFICANT EVENT (AS DEFINED IN 10 CFR 50.72) REPORTS FILED IN THE PAST 96 HOURS e
| |
| b o
| |
| 0 I
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| e l
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| /
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| $ ) e v
| |
| t
| |
| | |
| Oacility Nariwt (1) Occket Numcer (2) _f.ap e ( 3 )
| |
| Byron. Unit 2 01510101O!41515 01 ofl Ol1 Ottle (4) gN Ag g,Tg R g N g T g Ef 0 N0 k b N kP b Ob D Oh A'It_ flat e f E 1 _
| |
| LER Number (6) Reoort Date (7) Other Facilities involved f8)
| |
| U9 Cay Year Year /// Sequential /// Revision Month Day Year Facility Names l Docket Number (s)
| |
| I jf f ff N*r Number NONE .Q.l El 01 al of I I
| |
| ~~ ~~~
| |
| 016 219 817 817 01 01 9 010 017 219 817 of El of 01 of i l THIS REPORT IS SUBMITTED PURSUANT TO THE REQUIREMENTS OF 10CFR (Check one or more of the followino) (111 1 20.402(b) _ 20.405(c) _X_ 50.73(a)(2)(iv) _,, 73.71(b) l POWER ,_ 20.405(a)(1)(1) ,,,,_ 50.36(c)(1) , 50.73(a)(2)(v) ,_ 73.71(c) l LEVEL , 20.a05(a)(1)(11) _,,,,_ 50.36(c)(2) _ 50.73(a)(2)(v11) _,,,,_ Other (Specify 4 f101 0 9 8 _ , , 20.40$(a)(1)(111) _,,,,_ 50.73(4)(2)(1) _,,_,, 50.73(a)(2)(viii)(A) in Abstract
| |
| ///////j/////////////////// 20.405(a)(1)(iv) .,__ 50.73(a)(2)(11) ,_ 50.73(a)(2)(viii)(B) below and in l//////f/////////////////// ,_ 20.405(a)(1)(v) ._,_. 50.73(a)(2)(tii) 50.73(a)(2)(x) Text)
| |
| LICENSEE CONTACT FOR TWft LER f12)
| |
| , Name TELEPHONE NUMBER I
| |
| AREA CODE l W. Walter. Assistant Technical Staff tunervisor ext. 2160 9 l1 15 21 31 41 -l El 41 41 COMPLETE OWE LINE FOR EACH CDM T FAftuRE DESCRIBED IN THIS REPORT (131 4AUSE SYSTEM COMPONENT MANUFAC- REPORTABLE CAU$E SYSTEM COMPONENT MANUFAC- REPORTABLE j/ /p/
| |
| TURER TO kPRDS / TURER TO NPRDS X Sl3 $l Cl Gl Wl11fl0 Y / l l I I i l I f I I I l l l I
| |
| / l l I 1 1 I l f
| |
| $ SUPPLEMENTAL REPORT EXPECTED (141 .
| |
| Expected Month I Oav l Yeg Submission XJYes f f f ves. complete EXPECTED SUBMISSION DATE) l No 019 011 817
| |
| , ACT (Limit to 1400 spaces i.e. approximately fifteen single-space typewritten lines) (16)
| |
| At 0921 on June 29. 1987 a Manual Unit 2 Reactor Trip was initiated due to the 2C (Turbine Driven) Main Feedwater Pump (FW) trip. The level in all four steam generators was below the Lo level alarm setpoint and tr nding toward the Lo-Lo Reactor Trip 5etpoint. The Shift Engineer (Licensed) ordered the Nuclear $tation Operator (Licensed) to manually trip the reactor in accordance with good operating practices. At the Lo-Lo Steam Generator level setpoint the Auxiliary Feedwater pumps (AF) automatically started as designed. The Unit was stab 111 red in Mode 3 - Hot Standby. The cause of the 2C Main Feedwater Pump trip was a cold solder joint on a capacitor in the speed control feedback loop. The solder joint was repaired and tested.
| |
| In addition, selected parameters will be monitored during subsequent operations to ensure proper functioning. Results of this monitoring will be reported in a supplunental report. There have been no prey 1ous occurrences.
| |
| 1 3
| |
| (V C1552M/0175M)
| |
| | |
| )FACILITYNAME(1) DOCKET NUMBER (2) LER NUMBER (M P ue (31 Year //
| |
| p/p, Sequential / Revision
| |
| /// Number j////jj/ Number _
| |
| < avron. Unit 2 0151010 l o l 41 51 5 817 - 01 Ol 9 - 010 Ol2 0F 013 g TEXT- Energy Industry Identification System (EIIS) codes are identified in the text as (xx]
| |
| f , PLANT CONDITIOPS PRIOR TO EVENT:
| |
| 3 V
| |
| Event Date/ Time 06/29/87/ 0922 Unit 2 H00E 1 - Power daaration Rx Power 981 RCS [AS) Temperature / Pressure Normal Doeratino
| |
| : 8. DESCRIPTION OF EVENT:
| |
| On 06/29/87 at 0922 hours Byron Unit 2 was in Mode 1 at 98% power. At this time the 2C (Turbine Driven) Main Feedwater (FW) (5J] pump tripped. In accordance with Station Operating Abnormal procedures, the Nuclear Station Operator (N50) (Licensed) attempted to recover from the loss of the Feedwater Pump. however levels in all four (4) Steam Generators contirJed to decrease rapidly. The N50 manually tripped the reactor as the Steam G:nerator levels trended below the Lo level setpcint and towards the Lo-Lo level setpoint. Stable plant conditions were achieved and the unit was maintained in Mode 3 - Hot Standby pending a root cause investigation. There were no systems or components inoperable that contributed to this event. A Reactor Prgtection Actuation is reportable pursuant to 10CFR 50.73(a)(2)(iv).
| |
| C. CAUSE OF EVENT:
| |
| Th2 cause of the 2C Main Feedwater Pump trip was a cold solder joint on a capacitor in the speed control fccdback loop. This cold solder joint caused a high speed demand with no electronic speed limiting. The Fccdwater Pump reached the mechanical overspeed trip setpoint, and tripped. The cold solder joint in the speed control feedback loop is considered an isolated incident.
| |
| SAFETY ANALYSIS:
| |
| Th2 plant or public safety was not affected by the trip of the Main Feedwater Pump or the ensuing Reactor Trip. All Engineered Safety Features were available for operation and would have automatically actuated if the N50 had not manually tripped the Reactor. The decision to manually trip the reactor was in keeping with good cperating practice to anticipate the plant conditions which would cause an automatic safety system initiation.
| |
| The Auxiliary Feedwater Pumps (AF) {BA), did automatically start on Lo-Lo Steam Generator level as designed.
| |
| Loss of a Feedwater pump at full power or near full power is the most severe set of credible initial conditions.
| |
| E. CORRECTIVE ACTIONS:
| |
| ,The Operational Analysis Department in conjunction with the Instrument Maintenance Department found and rcpaired the cold solder joint on the speed control card. The circuit was tested and verified to be working ctrrectly. In addition selected parameters of the Feedwater regulating circuitry will be monitored during operations to ensure proper functioning. Results of this monitoring will be reported in a supplemental report. This is being tracked by Action Item Record 6-87-178.
| |
| I I v
| |
| (1552M/0175M)
| |
| | |
| "MBN W R ..
| |
| : FACILITY NAME (1)- DOCKET NUMBER (2)' LER NUMBER (6) Pine f3)
| |
| Year // Siguential Revision
| |
| ,/,/ Number Number
| |
| 'avran. Unit 2 1 o I E 1 o I o I o I al El E al7 - of t 1_ g
| |
| - ole of3 0F 013 Energy Industry Identification System (E!!s) codes are identified in the text as [xx]
| |
| PREVIOUS OCCURRENCES:
| |
| LER NLEGER llILL NONE l G. COMPONENT FAILURE DATA:
| |
| a) MANUFACTURER NOMENCLATURE MODEL NUPRER MFG PART NUPRER Westinghouse speed Reference Adjust N/A Card #4734046G02 And Speed Controller Card, b) RUuLTS OF NPRDS SEARCH!
| |
| Not Applicable -
| |
| e l I I l
| |
| n l (1552M/0175M) i i
| |
| 1 l'
| |
| | |
| r -
| |
| %--- n F ,s cA - / )
| |
| LICENSEE EVENT REPORT (LER)
| |
| Name (1)
| |
| Docket Numcer (2) _ Pace (3)
| |
| Byron. Unit 1 OI El 01 O! Of al El__t_ 1 i ofl0 1 j lE OR TRIP CAUSED BY MAIN FEEDW'.TER PUMP T ' ate f t) fpIP 00E TO a BRCKEN WIRE IN THE THPUST BEARING WEAR CIRCUITRY LER Number (6) Reecrt Date (7) l_
| |
| , Oay Year Year Other Facilities Involved fbi
| |
| ,/p/p/ 5equential Revision Month Day ' Year
| |
| /// Number j//g// Number racility Names ! Docket Num3erft)
| |
| _01 s 11 i ~
| |
| NONE 01 El 01 01 01 I I el 7 al 7 0 11 Ia
| |
| ~
| |
| 0IO O 19 013 81 7 pg, Of s1 01 01 of I I THIS REPORT IS SUBMITTED PUR5UANT TO THE REQUIREMENTS OF 10CFR (Check one or more of the fo11ewinal f111 1
| |
| 20.402(b) 20.405(c) .JL 50 73(a)(2)(iv)
| |
| POWER 2C.405(a)(1)(1)
| |
| __., 73.7)(b)
| |
| LEVEL 50.36(c)(1) 50.73(a)(2)(v) 73.71(c)
| |
| {
| |
| 20.405(a)(1)(11) j (10) _9 7
| |
| '10.36(c)(2) ,
| |
| 50.73(a)(2)(vit) Other (Specify !
| |
| 20.AN(a)(1)(tii) 50.73(a)(2)(i) 50.73(a)(2)(viii)(A)
| |
| '///////////////////////// 20.4.5(a)(1)(iv) in Abstract 50.73(a)(2)(ti) 50.73(a)(2)(viii)(B) t'elow and in
| |
| :///////////////////////// 20.405(a)(1)(v) 50.73(a)(2)(111) ,,,,_
| |
| 50.73(a)(2)(x? Text)
| |
| Name LICENSEE CONTACT FCR THIS LER (12)
| |
| TELEPHONE NUMBER AREA CODE Tom Hiocins. Oceratina Enoineer Ext. 2215 ei1 l E 2! 11 41 -l El al 4 COMPLETE ONE LINE FOR EACH COM N AUSE SYSTEM i COMPONENT FAILURE DESCRIBED IN THIS REPORT (til MANUFAC- REPORTABLE / CAUSE SYS?Pt COMPONENT MANUFAC-TURER 70 NPROS j/ REPORTABLE s 1 .1 TURER X 1 I IP WI1121O Y f I i 1 1 I I I TO NPDDS l I I I I I l I I I I I I I I SUPPLEMENTAL REPORT EXPECTED (141 p Expected Menth ! Day I Year Submission 1 / tIf ves. comolete EXPECTED _5UBMISSION DATE) X l NO I I I sB5 TRACT (1.imit to 1400 spaces, i.e. approximately fifteen single-space typewritten lines) (16)
| |
| On August annunciated. 11 Byron Unit One was at 97T. power when an alarm for the 1B Feedwater Pump Thrust Bearin Operators went out locally and checked the affected pump. A high frequency noise indicated the alarm was valid.
| |
| pump tripped. The unit load was reduced to f acilitate taking the feedwater' pump off itne. Then the steam generatorAlevel. load reduction to 50% power was intttated, but not in' time as the Reactor tripped on low All plant safety systems responded as expected. T avg following the trip. was stabilized 45 minutes Level. A fet0 water isolation signal was generated by a High-2 level in the 1C Steam Generater The actuation alerted the operators to the problem, and they isolated flow to the affected steam generator and restored the level to normal.
| |
| The thr9st bearing wear circuitry was checked and a broken wire to one of the proximity probes was found The wire wn repaired and the instrument calibrated. .
| |
| play was conte. No excessive wear was detected.
| |
| To ensure the pump was cperable, a check of the shaft fcund to have no operational problems. During the startup the feedwater pump was monitored and
| |
| \
| |
| O
| |
| . I 17M/0187M)
| |
| +
| |
| | |
| LICENSEE EVEgf SEPcpT (Lt9) Text test 1NuaT;cN Ty N AME ( 1 )
| |
| 00tsti MdMBER (2) _ LED NUMBER 61 p3ee t3)
| |
| Year / Sequential ///
| |
| ,//p/ Revision
| |
| /p/
| |
| Number ///
| |
| Number _
| |
| vron. Unit i D1E 1010 l0 l 41 El 4 9 l 7 -
| |
| O liiB -
| |
| 010 01 2 1,_j d Energy Industry Identification System (E!IS) codes are identified in the text as (xh]
| |
| 2.
| |
| FL ANT CONDITIONS PRIOR TO FVENT:
| |
| Event Date/ Time B/11/97 / 1015 Unit 1 MODE 1 - Pswer eneration Rx Pewer 97 RCS [AB) Temperature / Pressure 524*F / 2238 otic B. DESCRIPTfDN OF EVENT:
| |
| At 1011 hours an alarm annunciated for thrust bearing wear on the IB Feedwater Pump . The {$J) operators started to reduce load in anticipation of removing the affected feedwater pump from service.
| |
| At 1013 hours the IB Feedwater Pump tripped. A load reduction was initiated down to 559when MWe, the target load was reached steam generator levels were steady at 461 - 51%. The steam ger.erator Power Op; ratedpressure g:nerator Relief Valves increased.were lifting to remove the excess heat generated, and when they closed
| |
| . steam At 1915 a Reactor Trip occurred due to the pressure increase in the steam generators vels to causing le d; crease below the Lo-2 Reactor Trip setpoint of 41%.
| |
| 1tw level. Steam generator levels and Reactor Coolant avg System TThe Av.iliary Feedwcter System [BA) stabilized at 1100 hours.
| |
| At 1179 hours a Feedwater Isolation occurred due to a Hi-2 level (81. %) inThis ev;nt is unrelated to the initiating event as it occurred after the piant was stabilized.
| |
| . the 1C Ste FY '5E OF EVENT:
| |
| '~ )
| |
| The Main Feedwater Pump tripped on thrust bearing wear due to a broken wire on the proximity . This probe caused the wear detector to indicate a large axial displacement and picked up the trip circuit.
| |
| l Th3 subsequent Reactor Trip was due to the load reduction not being started as soon as the Feedw tripped.
| |
| The operator had to program the load rate and endpoint in the DEH computer. This delay was sufficient to allow the Steam Generator levels to decrease to the Reactor Trip Setpoint.
| |
| The feedwater isolation occurred I hour and 14 minutes af ter the Reactor Trip and was due to a Hig in the IC Steam Generator. -
| |
| l The level increased in the 1C Steam Generator to the High-1 level setpoint where an annunciator should have alarmed warning the operator of increasing Steam Generator level )
| |
| alarm was received. . No The level continued to increase without an alarm condition until the High-2 level was reached where the Feedwater Isolation occurred. The cause of the Feedwater Isolation was a cognative error ;
| |
| on the part of the Unit Operator.
| |
| D. SAFETY ANALYSit: !
| |
| Th] loss of a Main Feedwater Pump at power initiated the event.
| |
| Trip was mitigated by the safety systems. The subsequent toad Rejection and Reactor at no time was the safety of public, p?snt personnel or plant equipment endangered. All safety systems operat$d as expteted.
| |
| n}
| |
| l 17M/0187M)
| |
| | |
| t f CENTEE EVENT DES 007 f tts) TEXT CCNTfNUaticN mANC (1)
| |
| DOCKET NUMBER (2) LER NUweER 161 Year Pace (31
| |
| / 5:qu ntial
| |
| //,/ ,/// Revision i
| |
| // Mumber /,/ p
| |
| , vron.' Unit i / Number
| |
| [ix'~T o I t I o i o I o I di El 4 sl 7 -
| |
| oI1Ia o1o j; i 9ergy Industry Identification system (E!!5) codes are identified i oi3 0F 013
| |
| !. V M TfvE ACTIONS: n the text as [xx]
| |
| The 18 Feedwater Pump supervisory instrumentation was repaired and recalibrates . Furthermore, the automatic trips on thrust bearing wear on all four Feedwater Pumps, Unit 1 andIn2. have been cddition. It was recommended that Braidwood Station also remove their trips .
| |
| on thrust be A work requ:st was initiated to correct the problem of not receiving the High-1 Steam Generator level ala m.
| |
| PREVIOUS OCCURRENCES:
| |
| LER NUMBER lilLE NONE COMPONENT FAILURE DATA:
| |
| a) MANUFACTURER NOMENCLATURE E L NUMBER MF0 PART NUMB {g Not Applicable b)
| |
| Fr$ULTS OF NPRDS SEARCH 1
| |
| ,. There have been 3 occurrences where Thrust Bearing Failures have caused Main Feedwater P ump Trips.
| |
| (
| |
| l
| |
| \.
| |
| )") t 187M)
| |
| * l
| |
| | |
| LICt95ft t twf atP st .gR g.f Facility Mame (1)
| |
| Docket tumb3r (2) pace f t)
| |
| Byron. Unit 2 _
| |
| gl gj Of Of Sl 4l 5l $ 1 of $ l2 REAff0f TM45-toht-TTE.AP CfurRATCR LEVEL $[N TWf 2B Mafk FEfDWaffR PUMP T_ RIPPED Def TO PERtCNNEL f ffet Event Data ($1 L E E M' "** P f61 ISBOPL Oata ff) Other Facilitiet Involved (R)
| |
| Month Day Year Year Sequential /// Revision Month Day Year raciitty gaman i necket womearf o V u, 2r hhh u , -- e I r mont of si of el of f i 11 o of I ald al 7 oi1 Ia oIo I10 21 3 817 of gl el of al l i AI N, '
| |
| THIS REP 0af !$ $USMITTED PUR$UANT TO THE REQUIREMENTS OF 10CFR ffhack ane or more af the followinal fill 1
| |
| 20.402(b) 20.405(c) 1. 50.73(a)(2)(tv) ,_ 73.71( b) )
| |
| POWER .,,,,,,, 20.405(a)(1)(t) _ 50.36(c)(1) 50.73(a)(2)(v) 73.71(c)
| |
| LivtL 20.40$(a)(1)(11) 50.34(c)(2) 50.73(a)(2)(vit)
| |
| (101 eie l2 ,,,,,,. 20.405(a)(1)( $ 41) 50.73(a)(2)(1) 50.73(a)(2)(vitt)(A)
| |
| ,,,,_,. Other ( $pec t f y in Abstract
| |
| ////////////////////////// 20.405(a)(1)(tv) 50.73(a)(2)(11) 50.73(a)(2)(vtit)(B) below and in
| |
| ////////////////////////// _ 20.405(a)(1)(v) ,. 50.73(a)(2)(itt) .,, 50.73(a)(2)(u) Text)
| |
| LIEEkstf CONTACT FfE TNft LER f121 Came TELEPhout NUMRff AREA CODE D_ 11. Clair. Assistant hPintenannt idark PlanntRS [If_ IRAI EI1 lI fl 31 41 *l Il 41 4 COMPLfft ont Lits Foe FAfW CfDromfuf Faf t ter DEtrataro IN TMf1 ttPM T f111 CAU$t $YSTEM CtprontNT MANUFAC- REP 0tTA8LE / CAU$t Tiera 70 RPeet
| |
| /
| |
| j
| |
| $YSTEM COMPONENT MAmuFAC. REPORTA8tt l/
| |
| / / Tiera 70 kPent ,/
| |
| l I l l l l l / l l l l l l l l l l l l l l / / l l l I l l l /
| |
| 10PPLEMMTAL ttPMT EXPtfT[0 f 141 _ Expected Month l Day l Vaar
| |
| $441B1ssien g lYan f f f vas _ commlete EXPECTED ttmatf1110E GAffi I 20 l l I
| |
| ( jj AS$ TRACT (Limit to 1400 spaces, i.e. approstaately ftfteen stegle-space typewritten lines) (16)
| |
| Byron Unit 2 was operating at g21 power when the 28 Main Feedmater Pump tripped. The Unit mees ramped back to 50% power. The reactor tripped due to a low-2 Staam Generator Level in one steam generator. All safety systems responded as designed.
| |
| The cause was the inadvertent actuation of the overspeed trip plunger by a contractor working on the 28 Feedwater Pump high pressure stop valve.
| |
| To prevent recurrence the operating and maintenance personnel will be required to read this report. The installation of a guard on the trip plunger will be investigated.
| |
| l I
| |
| 1 V
| |
| (1660M/0194M) {
| |
| I l
| |
| l
| |
| | |
| LfEENSEE EVENT DEPSRT fLER) TEXT CONfMUaffSN FACILITY CaME (1) DOCKET CUMBER (2) LER nuMars t61 Pm f H Ycar
| |
| //l/l 5:0uent1a1 ///
| |
| l//
| |
| R:vtston l u-- Il/t t u. e l avenn unit 2 e i s I o I o 1 e i 41 st 1 af7 oIt Ia eIe el 2 or e_12 l TEXT Energy Industry Identification System (E!!5) coees are teentif ted in the text as (un}
| |
| O4 A. PLAnfcounivfamePetanToEVreh:
| |
| Event Date/Ttme to/1/a7 / 1121 Unit 2 MODE 1 - Pa=ar enern e tant Ru Power _12L RCS [AS) Temperature / Pressure earmat anaratins B. Dt1CRIPTION OF EVtai:
| |
| l On 10/1/87 Byron Unit 2 was in Nede 1 operating at 92 percent power. No systems were inoperable prior to the event which contrit" se to the event, :
| |
| A contractor was working on a High Pressure $ top Valve steam leak. Ne was under the supervision of a maintenance foreman. He was fully aware of the sensitive areas around the Main Feedwater Pump.
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| l At 1323 hours the 23 Fegeseter Pump (FW)[$J) tripped. The Unit Reactar Operator (licensed) rassed the turbine to $$g megawatts from 1985 megawatts at 2000 megawatts / minute. in accordance with Station procedures. The reacter subsequently tripped due to a low-2 Steam Generater leve1> on the 20 steam Generater, soth Aust11ary Feeesater Puses (AF)[BA) autgestically started te resters the Steam Generator levels. The plant was stattlized at 13ss hours. This event is reportatie per 10cFRst.73(a)(2)(tv).
| |
| : c. caust or LytmT: -
| |
| Inadvertent contact with the everspeed trip plunger by the contracter working se, the High Pressure Stop valve was the cause of the reeduster Puse Trip. The as Facesster furtine manual er.7 speed trip plunger is located next to the high pressure step valve. The plunger was found in a tripped state by cperating C personnel. There was ne other evidence of my other Feedmater Pump trt ss.
| |
| D. SAFETY ANALYSI1:
| |
| The plant or public safety was not affected by the pump trip and subsequent reactor trip. All safety systems functioned properly, including the Austitary Feedwater pumps which autenatically started to restore the steam Generator levels.
| |
| E. CgggCTIVE ACTIost:
| |
| The. Licensed Operators and Maintenance Fo non will te required to read this Deviation Report in an effort to make all respensttle personnel aware of the manual overspeed trip plunger on the peup. The installation of a guard over the manual overspeed trip plunger is being investigated and is being tracked by AIR 87-0256.
| |
| F. PatVIGut attuttlact$2 LER lRaett 111LL 85-41-01 Reactor Trip on Low steam Generator Level Due to Inadvertent Feed Pump Trip G. CmMLET FAILiitt QATA:
| |
| a) MAEuFACTinti INDEhrLAfter feDEL NLBett MFG PART InSSER Not Applicable b) RisuLTE OF NPRBS SEARCH:
| |
| not Applicable d164EM/8194M) l
| |
| ___ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _________J
| |
| | |
| i
| |
| ; LICENSEE EVENT REPORT (LER) p g g. g Facility Name (1) Docket Number (2) Pace (3)
| |
| Byron. Unit 2 015101010141Sl5 1Iofl0l3 ti2 (4) gggg2gg ggigg ggEQ{y0 A FEEDWATER PUMP TRIP AND FAILURE OF DIGliAL Event Date (5) LER Nunber (6) Report Date (1) Other Facilities involved (8)
| |
| Month
| |
| * Day Year Year fjj/
| |
| // Sequential /jj/j/ Revision Month Day Year Facility Names l Docket Number (s) 8
| |
| /// Number /// Nunber WONE 0151010101 l l 012 11 2 81 8 8I8 0 I O II 010 l I I 01 51 of 01 01 I I THIS REPORT IS SUBMITTED PURSUA;JT TO THE REQUIREMENTS OF 10CFR (Check one or more of the followino) (11)
| |
| MODE W I 20.402(b) _ 20.405(c) _x_ 50.73(a)(2)(iv) _ 73.71(b)
| |
| POWER _ 20.405(a)(1)(1) _ 50.36(c)(1) _ 50.73(a)(2)(v) _ 73.7)(c)
| |
| LEVEL _ 20.405(a)(1)(ll) _ 50.36(c)(2) _ 50.73(a)(2)(vii) _
| |
| Other (Specify (10) 0l9l4 _ 20.405(a)(1)(lit) _ 50.73(a)(2)(1) _,_ 50.73(a)(2)(viii)(A) in Abstract
| |
| ///// //////////////////// _ 20.405(a)(1)(iv) _ 50.73(a)(2)(li) ,,,_ 50.73(a)(2)(viii)(B) below and in
| |
| ////////////////////////// ,,_ 20.405(a)(1)(v) ,,_, 50.73(a)(2)(lii) ,_,,,50.73(a)(2)(x) Text)
| |
| LICENSEE CONTACT FOR THIS LER (12)
| |
| Name TELtpuoNE NUMBER AREA CODE T. Joyce. Assistant Superintendent Operatino Ext. 2213 8l115 21 31 41 -l 51 41 All COMPLETE ONE LINE FOR EACH COMPONE,N FAILURE DESCRIBED IN THIS REPORT (13)
| |
| CAUSE SYSTEM COMPONENT MANUFAC- REPORTABLE CAUSE SYSTEM COMPONENT MANUFAC- REPORTABLE TURER TO NPROS TURER TO NPRDS. .
| |
| X JlJ l l IV Ml4 12 13 Y l l l l l l l l I I I I I I I I I I I I I q g SUPPLEMENTAL REPuRT EXPECTED (14) Expected Month i Day l Year Submission x lyes (If yes, canelete EXPECTED SUBMISSION DATE) i NO 110l311l818 ABSTRACT (Limit to 1400 spaces, i.e, approximately fifteen single-space typewritten lines) (16)
| |
| On February 12, 1980, at 1804 hours with Byron Unit 2 in power operation (Mode 1) at 94% power, the 2C feedwater pump tripped on overspeed. Efforts to shed load were unsuccessful due to failure of the digital electrohydraulic control system to respond properly in the automatic mode. This resulted in loss of inventory in the steam generators and a reactor trip on low steam generator level. All safeguard actuation features functioned as designed. The feedwater pung trip was due to a failed servovalve which allowed the feedwater pump turbine high pressure governor valve to fall open. This caused the pump overspeed and trip. The defective servovalve was replaced. There have been previous reactor trips due to feedwater pump trips.
| |
| b d
| |
| l I ;
| |
| (1917M/0206M) 4591 8/88
| |
| | |
| llCEN5fE EVENT REPORT (LER) TEXT CONTINUAT!Cil FACILITY NAME (1) DOCKET NUMBER (2) LER NUMBER (6) Pace (3)
| |
| Year /jjj// Sequential //jj/ Revision j//
| |
| l i
| |
| /// Nunt>er / Nunt>er l Byron. Unit 2 0l5l01010ldl515 8l8 -
| |
| 010l1 - 0l0 01 2 0F 01 3 l TEXT Energy Industry Identification Systone (Ell 5) codes are identified in the test as [xx]
| |
| A. PLANT CONOTTIONS PRIOR TO EVENT:
| |
| Event Date/ Tine 2-12-88 /_,,_1f404 Unit 2 MODE 1 - Power Coerations,, Rx Power 941 RCS [A8) Temperature / Pressure Nome) Operatino
| |
| : 8. DESCRIPTION OF EVENT:
| |
| On February 12, 1988, at 1804 hours, Byron Unit 2 was in power operation (Mode 1) at 94 percent power. The Digital Electrohydraulic (DEH)[TC) eontrol systun was maintaining the turbine generator at 1070 MWe in Auto with Inpulse and Speed feedback loops "IN" and Magawatt Feedbick loops "OUT". At this time the PC Turbine Driven Feedwater pung (FW)[5J) tripp3d due to on 3rspeed. The Nuclear Station Operator (WSO, ticensed) correctly initiated a runback of the Main Turbine Generator (TG)[T8]. The rang was programmed for 2000 PWe/ min to 559 MWe per Byron Operating Procedures. The DEH Conputer did not execute the runback properly and load only dropped 60 'We and held as 1014 f%e. The operator depressed the " HOLD" button and the ramp was re-initiated in the manual control mooe. The Turbine Generator runback was not sufficient and a " low low" level in the 2C Steam Generator caused a reactor trip. The Unit was maintained in Mode 3, Hot Standby, until initial investigations of the turbine runback failure and feedwater pung trip were conducted. The NRC was notified at 1853 hours on 2/12/88.
| |
| Unit 2 was brought back on line using the 28 Turbine Driven Feedwater punp in place of the 2C Feedwater punp. The high pressure governor on the 2C Feedwater pung was manually isolated upstream to allow O i monitoring of the 2C Feedwater pung governor valve without affecting pung operation. The governor valve's b servo-actuator valve was replaced and was Deing monitored at seven different points by a strip chart recorcer.
| |
| All safety systems responded as required. No o*her systeru or components were inoperable prior to this event which contributed to this event. All operator actions were correct. This event is reportable pursuant 10CFR50.73 (a)(2)(iv).
| |
| On 2/22/88, at 0130 hours Unit 2 was in Mode 1 at 86 percent power when the High Pressure Governor valve f on the 2C Turbine Driven Feedwater Pump opened. The valve was teeing monitored following the replacement of )
| |
| the servo-actuator valve, and the steam supply was manually isolated. The Unit experienced no adverse I affects from this occurrence, l C. CAUSE OF EVENT:
| |
| The cause of the 2C Feedwater puno trio on February 12, 1988, was found to be a failure of the servovalve on the High Pressure Governor Valve. The High Pressure Governor Valve oa the Feedwater Punes are only used at startup and stdtdown. When the servovelve failed the High Pressure Governor Valves failed open causing the Feedwater Turcine to overspeed and trip. The servovalve has been sent to the manufacturer for a failure analysis to detemine the root cause of the failure. A supplewntal report will be issued when the results of this anslysis are known.
| |
| \
| |
| l I j (1917M/0206M) 4591 8/88 l
| |
| l
| |
| | |
| LICENSEE EVENT REPORT (LER) TEXT CONTINUATION FACILITY NAME (1) DOCKET NUMBER (2) l_lfR_ NUMBER (6) Pa9? (3)
| |
| Year /// Sequential // Revision fjj/
| |
| // Nunter /j/j/
| |
| f
| |
| / Nunter Byron. Unit 2 0 l 5 1 0 1 0 1 0 l 41 51 5 8l8 -
| |
| 0l0l1 -
| |
| 0 IO O! 3 0F 01 3 l TEXT Energy Industry Identification System (E!!S) codes are identified in the text as [xx]
| |
| C. CAUSE OF EVENT: (Continued)
| |
| The Feedwater pump trip required the Unit 2 NSO (licensed) to initiate a turbine rLnback which did not occur as plaaned. Previously, changes were made per Westinghouse instructions and in accordance with the Statiot.'s software control program to the DEH computer, in order to fine tune the main turbine governor valve operations. These changes were made tc minimize load swings during governor valve testing. One of the change; made to the DEH computer involved reducing the deadband for the impulse pressure Feedback loop by adjusting some of the conputer gains. Basically, this loop looks at a calculated inpulse pressure. If the difference between the calculated inpulse pressure and the actual inpulse pressure is too great the DEH computerrejectstheimpulsepressureloopandstopstheunitatthatpowerlevel. The turbine runback, initiated after the 2C Feedwater Pump trip, dropped electrical output approximately 60 MWe when the gain value between calculated and actual impulse pressure was exceeded, halting the runback. The runback was co.npleted manually.
| |
| On 2/22/88 the failure of the replacement servovalve was detennined to be the cause of the High Pressure Governor l'alve opening. In this event there were ro adverse affects to the Unit due to the fact that steam to the High Pressure Governor Valve was isolated. The servovalve was found to have a defective coil with high internal resistance. In the first event the cause of the serwovalve failure was not apparent.
| |
| D. SAFETY ANALYSIS:
| |
| All plant safety systems actuated and performed as designed. The reactor tripped on Low-2 level on the 2C Steam Generator. The Manual Turbine Generator runback a s still available to runback the Turbine.
| |
| E. CORPECTIVE ACTIONS:
| |
| The servov21ve was replaced on the 2C Feedwater Pump High Pressure Governor Valve after the 2/12/83 occurrence. The second High Pressure Governor Valve failure a s being monitored and examination of the strip chart recording: showed that the servovalve was again malfu.nctioning but with a different nrade cf failure. The serwovalve was again replaced after tte 2C Feedwater Pung was trAen off line, nd the 2A Motor Driven pep was put into service. The 2C pucp was mnnitored following conpenent replacement, and will continue to be monitored when it is returned to operation to ensure proper operation. The DEH problem was resubmitted to Westinghouse Corporation for reevaluation. In the interim the gains in the DEH computw will be returned to their previous values. Subsequent valve tests have been conducted with satisfactory results.
| |
| F. PREVIOUS _0CCURRENCES:
| |
| Prcvious ractor trips due t9 Feedwater pupp trips were roerted in the following LER's.
| |
| LEA HU g 3 TJILE,,
| |
| 454/85-061-01 454/S7-018-00 455/87-00 M O
| |
| : 2. COMPC/NENT
| |
| ~
| |
| FAILURE D#TA:
| |
| a) MANUFACT'JRER NOMENCLATURE FoDELNUMBfj PFGPARTNUMQQ (
| |
| Moog Servovalve A076-IB5 1161 (
| |
| l p Fag Model 7's0 l ( b) RESULTS OF N??OS SEARCH:,
| |
| l No pertinent infonnation found during NPRJS search.
| |
| I I (1917M/0206M) 4591 8/88
| |
| | |
| F~W M -l LICENSEE EVENT REPORT (LER)
| |
| Fccility N ee (1) Docket Number (2) Pace (31 rw Byron. Unit 2 01 51 01 01 01 41 51 5 1lof!0l4
| |
| . Iill' I4) NM*N'fe$r gDue toi Improper Isolation of Electrohydraulic Control Fluid Supply Event Date (5) LER Number (6) Report Date (7) Other Facilities Involyed (B)
| |
| Month Day Year Year /// Sequential ///, Revision Month Day Year Facility Names l Docket Number (s) fj/j
| |
| // Number ff
| |
| /// Number I NONE 01 Sl 01 01 01 l l
| |
| ~ ~
| |
| d5 0 16 8iB 81 8 0 I O 14 010 0l6 01 3 El 8 0151010101 I l THIS REPORT IS SUBMITTED PUR$UANT TO THE REQUIREMENTS OF 10CFR OPERATING (Check one or more of the followinal (111 MODE (M 1 20.402(b) ___ 20.405(c) .JL 50.73(a)(2)(iv) _l 73.71(b)
| |
| POWER __ 20.405(a)(1)(1) 50.36(c)(1) _ 50.73(a)(2)(v) _ 73.71(c)
| |
| LEVEL _ 20.405(a)( 1)( H ) ___ 50.36(c)(2) _ 50.73(a)(2)(vii) _ Other (Specify
| |
| ( 101 0l9l4 _ , _ 20.405(a)(1)(lii) _ 50.73(a)(2)(1) _ 50.73(a)(2)(viii)(A) in Abstract
| |
| .g////
| |
| /f/// {///////////////// _ 20.405(a)(1)(iv) _ 50.73(a)(2)(li) _ 50.73(a)(2)(vill)(B) below and in
| |
| /jf
| |
| /
| |
| j/.//////////j//////f//
| |
| f
| |
| / _ _ 20.405(a)(1)(v) _ _ 50.73(a)(2)(iii) 50.73(a)(2)(x) Text)
| |
| LICENSEE CONTACT FOR THIS LER (12)
| |
| Name TELEPHONE NUteER AREA CODE l
| |
| M. Snow. Peaulatory Assurance Sunervisor Extension 2280 81115 2131dl-l5141dl I COMPLETE ONE LINE FOR EACH COMPONENT FAILURE DESCRIBED IN THIS SEPORT (13)
| |
| CAUSE SYSTEM COMPONENT MANUFAC- LREPORTABLE CAUSE SYSTEN COMPONENT MANUFAC- REPORTABLE l TURER TO NPRDS TURER TO NPRDS ;
| |
| l i I I I I I I I i l I I I l l t i I I I I I I I I i 1 l l 9 -
| |
| SUPPLEMENTAL REPORT EXPECTED (14) Expected Month i Day l YearI Submission Date (15) lg l, lyes (If ven. complete EXPECTED SUBMISSION DATE) YlNO ,
| |
| ABSTRACT (Limit to 1400 spaces, i.e approximately fifteen single-spare typewritten lines) (16)
| |
| On April 27, 1988, an out-of-service contittion on the 2C Hain Feedwater Pump (6tFP) was temporarily 1Hted to permit operation of the NFP, and the puup was started and placed in service. On May 6 a licensed reacter operator (NS0) noted that the teeporary lif t was fue to expire en that c'ay t.nd requested a disposition f rom a licensed senior reacttr operator (SCRElm The SCRE directed the WSO to terminate the temporary itf t by retarning the equipment to its original out-of-service condition. Both the SCRE and the NSO incorrectly believed that returning the valve IIsted on the temporery lif t paperwork to its out-of-sttrvice c1csed pcsition would not affect thw operoilun of the 2C MFP. At 1714 on May 6 with Unit 2 at 94 percent power as Equipmeat Operatir closed the vatve, which iselated electrohyeraglic IEH) fluid supply to the 2C M9. At 1215 the 2C MFP tripped due to low EH fluto pressure. Steam peserstor levels lowered rapidly and the NSO manaally tripped the reacter in anticipation of an automatic trip. Operator 6ctions teken following the reacter trip were correct, and stable plant u,nditions were achieved in Hot Standby at 1330.
| |
| Several causes contributed to the irproper closbre of the EH valre. The NSO and the SCRE committod copattive persoenel errers by falling in recognize the consequen;es of the return to out-of-servico condition. Both individuals ir.ade incorrest assuertions regarding system design without reference to system drawings. The administrative procedere fer control of temporary lif ts contributed to the persone.e1 errors.
| |
| The Operating Department personnel involved in tb event have been interviewed and specific Forf ormance weaknesses heve been discussed. Administrative procedures will bs revised appropriately to mirdmize i
| |
| n recurrence.
| |
| ! (b There have been no previous similar occurrences of this event.
| |
| I
| |
| {
| |
| (0011R/0002R)
| |
| | |
| LICENSEE EVENT REPORT (LER) TEXT CONTINUATION FACILITY NAME (1) DOCKET NUMBER (2) LER NUtgG (6) Pace (3)
| |
| Year /// Sequential- /,// Revision fff
| |
| / /// Number f// Number
| |
| \ $
| |
| V Byron. Unit 2 0 1,$,1 0 1 0 l 0 1 41 51 5 8l8 - 0l014 - 0 10 0l2 0F 0 14 TEXT Energy Industry Identification System (E!!S) codes are identitled in the text as (xx)
| |
| A. PLANT CONDITIONS PkIOR TO EVENT:
| |
| Event Date/ Time 5/6/88 / 1216 Unit 2 M00E 1 - Power Operation Rx Power 947. RCS (AB) Temperature / Pressure Normal Ooergling_
| |
| : 8. DESCRIPTION OF EVENT:
| |
| On April C,1988, at 1152 ths 2C turbine driven Main Feedwater Pump (MFP) ($J) was removed from service in order to replace the servo valve on the Low Pressure Governor Valve (2E450498). The position of 2ES504nB had been oscillating abnormally, and its servo valve was believed to t+ the cause. Establishment of the maintenance out-of-service boundary was accomplished by closing the Electrohydraulic Control Fluid Supply Valve (2EH-50588) (JJJ. The servo valve was replaced but could not be tested to verify satisfactory operation because the internal components of 2EH-50498 had been damaged during the valve's oscillations.
| |
| Therefore, the out-of-service condition could not be administratively cleared by completing the required valve testing. Complete repair of 2EH-50498 will require that the main condenser be at atmospheric pressure.
| |
| Due to feedwater piping vibration and flow control problems associated with the operation of the 2A motor driven MFP, the starting of the 2C turbine driven MFP was pursued. At 1730 on April 27. a Temporary Lift on the out-of-service was authorized by the Unit 2 Shift Foreman (licensed Senior Reactor Operator) to
| |
| ,3 pemit operation of the 2C MFP using stees supplied from the main steam header (SB) via the High 'rsssure
| |
| '(\ Governor Valve. Valve 2EW50588 was opened and the 2C MFP was started and aligned to supply feedweter to the steam generators. .The ZA motor driven MFP, which had been supplying feedwater, was stopped. The 28 turbine driven MFP continued to operate with steam supplied to its turbine from a Moisture Separator Reheater via its Low Pressure Governor Valve, which is the normal steam supoly for the MFP turbines at high power levels.
| |
| On ftay 6,1996, a Unit 2 licensed reactor operrtor Nuclear Statior. Ope-ator (NS0) noted that the Temporary Lift on the out-of-service for valve 2E W 50496 was due to expire on t. net day. The N$0 delhered the Teeporary Alft par,erwork to a licensed senior reacter operator Shif t Control Room Enginer.r (SCRE) to cbtain a decisiou cs to whether the Tamperary Lif4 should be extended or- terminated by restoring the out-of-service. The SCRE returteed the Temporary lif t paperwork to the NSO and directed him to terminate the Tamporary Lif t by returning the otoipr. ant to a.n out-of-service evndition. Both the SCRE and the hSO belleted tttst the out-of-service would only affset ZEW50498 by closing 2EW50588, however, f r. actuai,ity the closing of 2EA50531 also isolutn fluid to the High Pressure Governor Valve which was supplying steem to ts.e 2f M"P turbine at the ties. Tlie NSO directed an equipe;ent oparator (LO) to restore the cut- o f -s e rvi ce.
| |
| #t 121;l en May 6, 1988, with Unit 2 at 94 percent prtwer, t.he E0 closed 2EW 50568 as instructed. Shortly thtresiter the "Feedvate- Pump furtine Oil Pressure tom a serarciator acteated in the Cuatrel Room. This annunciator alarms due to either low Ivbricaticg eli pr,tsrure or low electrohydrar.Ile (EH) fluid pressure.
| |
| The NSO conta:te6 the EO by radio 'auridiately to notify hir cf the aleming conditier.. Simcitaneously the EO heard the 2C MFP turt,ine speet: decreasing and tried to opsen 2ES50588.
| |
| 4 l [
| |
| l $'
| |
| I (0011R/0002A)
| |
| J U _- _ __-_- ______-___ _____ j
| |
| | |
| r l I LICENSEE EVENT REPORT (LER) TEXT CONTINUATION FACILITY NAME (1) DOCKET NUMBER (2) LFR NUMBER (6) Pace (3) {
| |
| Year Sequential ; Revision p (({
| |
| /// Number {/{f
| |
| // Number i
| |
| !bj Byron. yg h 2 0l51010 8 0 1 41 51 5 8l8 - 0l014 - 0 l0 01 3 0F 01 4 TEXT Energy Industry Identification System (EIIS) codes are identified in the text as [xx]
| |
| B. DESCRIPTION OF EVENT: (Continued)
| |
| At 1215 the "Feedweter Pump 2C Trip" annunciator actuated. The NSO initiated a preset main turbine [TB) runback to 559 Megewatts-electric (HWe) at 175 MWe/ minute. At 1216 narrow range levels in all steam generators ($/G's) had decreased to the low alarm setpoint. The SCRE selected manual governor valve fast i action to more rapidly runback the main turbine because the preset runback did not seem to be operating as expected. Narrow range levels in all S/G's continued to decrease. The SCRE directed the NSO to manually trip the reactor when narrow range S/G 1evels dropped to 20 percent. At 1216 the NSO manually tripped the reactor and an automatic turbine trip followed. The Control Room operators entered and compiled with
| |
| " Reactor Trip or Safety Injection Unit 2 Emergency procedurs" (TBEP-0). The ZA and 29 Auxiliary Feedwater Pumps (AFP's) [BA) automatically started due to the low low S/G 1evels resulting from the feedwater-steam flow mismatch and Indicated level shrink on the trip. At 1217 a Feedwater Isolation occurred dee to the expected decrease in Average Reactor Coolant Temperature (T,yg) to its low setpoint coincident with the reactor trip. The 2C S/G Power Operated Relief Valve (PORV) opened fully and remained open until the NSO placed its controller in manual and fully closed it at 1219.
| |
| At 1245 the Feedwater Isolation signal was reset, the Startup feedwater purp was started, and a flow path
| |
| , from the Startup Feedwater pump to the S/G's was established. At 1308 the 2B AFP was stopped end at 7323 the ZA AFP was stopped. Stable plant conditions were achieved at 1330 with Unit 2 in Hot Standby (Mode 3). Operator actions taken following the reactor trip were correct and contributed to the safe conclusion of the event.
| |
| O t This event is reportable in accordance with 10CFR50.73(a)(2)(iv) due to the manual actuation of the Reactor Protection System.
| |
| C. CAM ElyM:
| |
| The cause of the reactor trip was the manuel actuation of the reactor trip switch on the main control board by the NSO. The NSO manually actuated the trip due to downward trending low narrow range S/G 1evels in anticipatit:n of an autteatic reactor trip. The low r. arrow range S/G 1evels were caused by the tripping of the 2C MFP while the plart was at 94 percent power, which resulted in a steam flow-feed flow mismatch.
| |
| Contributing to the low levels was indicated Icvr1 snrink caused by the operator initiated main turbine rurback. The 2C MFP trip was caused by the LO when he closed 2EH-50588, which isolated the EH control fluid s9pply from the Hiph Pressure Governor Valve. This caused the governor valve to c1cse and bicck all stenra flow to the 2C MFP turbine. The cause for the improper opening of the 2C PORV was not determined af ter exter-sive troubleshooting by Maintenance department technicians, and the valve was declared operable ori Kay 11, 1908. The main turbine runback was determined to have been respending properly to the event.
| |
| Sever,1 causes contributed to the improper closure of 2EH-50588. The licensed NS0 and SCRE committed cognitive personns1 errors by failing to recognise the consequences of the retver; to out-of-service condition. The NSO and SCRE cirected the operation of plant equipment without fuliy understanding thE impact of that operation. Both ind!viduals 1,elieved that the closing of valve 2EH-50585 woult' only isolate the EH fluid suppiv from the Low Pressure Governer Valve, and that the High Pressure Goverco- Valve would reain unaffected and permit vainterrupted operation of the 2C MfP. Thdr belief wes based on an incorrect ass-uption that the affected portio 3 of the MrP EH system is designed siellarly to the Hein Turbir.e EH system where individual EH isolation valves are provided for each governor valve. Neither operetor consulted pipiry system drawings to veriff that the return to cut-cf-service could be pyrformed without
| |
| [
| |
| (
| |
| seriously impacting plant operation. There were no unusual cha-acteristics of the plant environment that contributed to the persor.nel errtrs.
| |
| I I
| |
| (0011R/0002R)
| |
| | |
| 1 l LICLN$E_E EVENT REPORT (LER) TEXT CONTINUATION FACILITY NAME (1) DOCKET NUMBER (2) LER NUM ER (6) Paer (3) p Year Sequential Revision Numbor Number Byron. Unit 2 0 1 5 1 0 1 0 1 0 1 41 51 5 Bl8 - 01014 - 0 10 01 4 0F 01 4 TEXT Energy Industry Identification System (EIIS) codes are identified in the text as [xx]
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| C. CAUSE OF EVENT: (Continued)
| |
| Byron Administrative Procedure " Administrative Requiremer.ts for Temporarily Lif ting 005 Cards and/9r Placing Equipment in Test" (BAP 331-1) provides direction and responsibility for the initiation and termination of Temporary Lifts. It does not require licensed senior reactor operator ($RO) involvement for the termination of a Temporary Lift. Althouoh the N50 involved the SCRE during this event, the SCRE still was not required to docunc t his agproval of the proposed action. Additionally BAP 331-1 does not require a thorough review of the effects that a return to out-of-service condition may have on operating plant equipment, as would be required for the initial out-of-service per " Station Equipment Out-of-Service Procedure" (BAP 330-1).
| |
| D. SAFETY ANALY!IS:
| |
| Neither plant safety nor public safety were affected by this event. All Engineered Safety Feature (ESF) systems operated properly to minimize the consequences of the plant trip. Although the 2C S/G PORV opened and remained open until placed in manual and closed, this caused only a minor Reactor Coolant System cooldown to approximately 551'F and briefly delayed the achievement of stable plant conditions. The more severe condition of a MFP trip at 100% reactor power would only have accelerated the pace of events, and plant /public safety would have remained unaffected.
| |
| E. CORRECTIVE ACTIONS:
| |
| In order to permit continued operation of the 2C MFP using the High Pressure Governor V&1ve, an Onsite Review was completed to clear the out-of-service that required the closing of 2DI-50588. This action eliminated the need to operate the 2C MF* with a temporary lift condition in effect.
| |
| The 2C 5/G PORV was initially it.oleted by closing its manual isolation valve, and the at.sociated Technical Specification Limiting Condition for Operation Action Requirement was satisfies. Whers troubleshooter.g efforts failed to identify any compenent failures, valve ope-ability was verified and the valve w%
| |
| declared operable at 1212 on May 11, 1988.
| |
| The 1,perating Department personnel involved in the event have been interviewed and specific performance .
| |
| weainesses have been discussed. I f
| |
| The SAP 331-1 will be revised to include: )
| |
| 6
| |
| : 1. iRD responsibility for the termination cf Temporary Lif ts.
| |
| L A cautiors statement to ensure the conduct of a thorough technical revievt prior to retu<ning temporarily lif ted equipment to an out-ni-service condition.
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| : 3. A require. rent for Onsite Review of To:nporary Lif tc whose duration exceeds five working days.
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| l The ' Operating Shift Turnover ar.d Retter Adminiu rative Procedure" (BAP 335-1) will be revised to require Shift Engineer (licensed $lt0! review of Tempo:ary Lif t parhge; each shif t. Ctepletion of the procedere revisions is tracked by Actic,a Ite'r, Record 45/-225-06 4 117.
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| 1 F. Pyf,J @ $ OCCURREV fS: !
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| G. COMPONENT FAILURE DATA:
| |
| Not Applicable l l (0011R/000ZR)
| |
| | |
| l FW "' I LICENSEE EVENT REPORT (LER) l Facility Name (1) Docket Number (2) Paes (31 Byron Unit 1 01 51 01 01 01 41 51 4 1lof!0l3 Tachometer Failure Caused Oversneed Trio of Main Feed Puno Resultino in Reacter Trio Event Date (5) ,
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| LER Number (6) Resort Date (7) Other racilities Involved (8)
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| Month Day Year Year Sequential //j/j' Revision Month Day Year Facility M n l Docket Number (s) f i Number /// Number NONE 01 El 01 01 01 l I
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| ~ """'"
| |
| Ol 7 11 6 Bj 8 81 8 0 l 0 14 010 0 i 8 .1 10 8l8 01 51 01 01 01 l l OPERATING THIS REPORT !$ $UBMITTED PURSUANT TO THE REQUIREMENTS OF 10CFR
| |
| "*k ""* "" ""'" " * #"" ""'"" ""
| |
| MODE (9) 1 20.402(b) __ 20.405(c) L 50.73(a)(2)(iv) 73.71(b)
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| POWER ,_ 20.405(a)(1)(1) 50.36(c)(1) _ 50.73(a)(2)(v) _.__ 73.71(c)
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| LEVEL __ 20.405(a)(1)(ii) _ 50.36(c)(2) 50.73(a)(2)(vil) Other (Specify (101 0l9l8 _ 20.405(a)(1)(lii) 50.73(a)(2)(1)
| |
| _. 50.73(a)(2)(viii)(A) in Abstract
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| / ///
| |
| f/jjj/jj/jj/j/j//jj/jjjjj/u/////u/ //// / / - ' - 20.405(a)(1)(iv) _ 50.73(a)(2)(li) _ 50.73(a)(2)(viii)(B) below and in
| |
| ////
| |
| HfHf###HHf6fHHH n dos (*HiH > 50 73(*H2H"O 50 73(a)(2)(=) 7'=t>
| |
| LICENSFF CONTACT FOR THIS LER (12)
| |
| Name TFtFPHONE NUEER MU CWE i T. Tulon. Antt Superintendent Doeratina Extenulon 2213 81115 213141-l51414l1 COMPLETE ONE LINE FOR EACH COW ONENT FAftuRE DESCRIBED IN THIS REPORT f13) !
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| CAU$E SYSTEM COMPONENT MANUFAC- REPORTA8LE / CAUSE SYSTEM COMPONENT MANUFAC- REPORTABLE I TURER Td NPRDS j/ TURER TO NPRDS I
| |
| X 5l J l il Al C A l Il 21 3 Y / l I l l l l l 1 I I I I I I I I I I I I i !
| |
| SUPPLEMENTAL REPORT EXPECTED (14) Month l Day 1 Year O- Expected Subaission lYet f f f vor. comolate EXPECTED SUBMISSION DATE) 71 NO Date (15) , l,l9 ,
| |
| ABSTRACT (Limit to 1400 spaces, i.e. approximately fif teen single-space typewritten lines) (16) l Unit 1 was at 98 percent reactor power at 0431 on July 16, 1966, when the IB Main Feedwater Pump (MFP) tripped. Steam Generator (5/G) levels decreased due to the feedwater flow-steam flow mismatch. In spite of licensed operator actions to reduce steam flow and increase feed flow,10 $/G 1evel decreased to the low-low reactor trip setpoint at 0434. An automatic reactor trip occurred and both Auxiliary Feedwater Pumps automatically started. The licensed operators complied with emergency operating procedures and brought the plant to a stable condition in Hot Stantby at 0530. This report is submitted in accordance with 10CFR50.73 (aH2)(iv) due to the automatic safety system actuations.
| |
| The 18 HFP's precision tachometer failed. The tachbunter transmitted a constant (Ecrease speed output signal to the turbine's automatic speed control circuitry. Turbine speed increased until it reached tne averspeed turbine trip setpotnt and tripped.
| |
| The tachometer was repaired and 18 MFP operation was monitored during the subsequent Unit startup. Ti.e pump was returns 11 to service without incident.
| |
| A sitilar previour, occurrence was reported in Unit 2 Licensee Event Repers 67.009.
| |
| O i i (0078R/0008R)
| |
| ._-______-._-_.m_m_ _ _ _ _ . _ . .-
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| | |
| LICENSEE EVENT REPORT fLER) TEXT CONTINUATION
| |
| - FACILITY NAME (1) DOCKET NUPSER (2) LER NLDGER (6) Pane f*h ]
| |
| [ Year M fff Sequential M Num6er fff Revision Number .-
| |
| 8venn. Unit 1 0 l 5 1 0 .3 0 1 0 1 41 El 4 8I8 - 0l014 - 0 l0 01 2 0F of 3 TEXT Energy Industry Identification System (EII5) codes are identified in the text as [xx]
| |
| A. PLANT Copelff0NS PRIOR TO EVENT:
| |
| Event Date/ Time 7/16/88 /. 0434 Unit 1 M00E 1 - Powe70neration Rx Power .281._, RC$ [AB) Temperature / Pressure Normal Onoratina
| |
| : 8. DESCRIPTION OF EVENT: ~'
| |
| There were no systems or components inoperable at the beginning of this event that contributed to the event. Unit I was at 98 percent reactor power at 0431 on July 16, 1988, when the 18 Main feedwater Pump (MFP) [$J) turbine thrust bearing wear and the 18 MFP high discharge flow annunciators actuated in the main control room. The 18 MFP tripped and steam generator (5/G) levels decreased due to the feedwater flow-steam flow mismatch. - The Nuclear Station Operator (NS0) (licensed reactor operator) initiated a Turbine Generator [TB) runback to 599 Megawatts-electric (MWe) at a rate of 175 We per minute and maximized feedwater flow-tate by increasing IC MFP speed and starting an additional Condensate / Condensate Booster Pump [50). In spite of these actions 5/G 1evels continued to decrease slowly and at 0434 1D S/G 1evel dropped to the low-low level reactor trip setpoint (40.8%). An automatic reactor trip occurred and the 1A and 18 Auxiliary Teodwater Pumps (AFP) [BA) automatically started. A normal post reactor trip Feedwater Isolation occurred when average reactor coolant temperature (Tay '
| |
| the reactor trip breakers open. The Itcensed operators entered and complib) decreased with " Reactor Trip or Selow Safety 564*F w Injection - Unit 1 Emergency Operating Procedure" (IBEP-.0) and " Reactor Trip Response - Unit 1 Emergency Q Operating Procedure" (IBEP ES-0,1). It 0436 the N50 manually isolated chemical and Volume Control System
| |
| (/ [CB) letdown flow due to-T avg decreasing below the no load value and the corresponding decrease in pressurizer level. Auxiliary feedwater flow rate was reduced and the T,yg reduction was stopped at approximately 550"F. By 0450 T ayg returned to its no load value and letdown flow was established.
| |
| At 0451 the Feedwater Isolation signal was reset and the Startup Feedwater Pump was started and eligned to supply feedwater flow to the $/G's. At 0523 the 14 AFP was stopped and 4.t 0527 t b 1A AFP waw stoppe(,
| |
| since the pumps were no longer needed to malatt.in 5/G 1evels. Stable plant conditions were achiever; f r. Hot Standby at 0530.
| |
| This Licensee Event eeport (LE't! is sulaitted in accordance with 10fJR50.73 (a)(2)(iv) due to the automatic 1 Reactor Protection Syster and Engfr, sered Safety Fertures Systeas actuations. i i
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| C. rAUSE OF EVENT:
| |
| l Th cause of the svent was the Inss of one Turbine Driven Feedwater Pump. Ihv 1B Freewetsr Turbine trippe(
| |
| due to an overspmj co# tice. The reedwatsr Turbine's Tsch-par, serf ee 630 Precision T6chnmeter was fcuad to be defwettve. The tachometer transm'itted a r.onstant hereste speed signal to the tu' tine's spud .
| |
| contrul circuitry. Turkin's Speed incraesed untti it rsached the everspewd tvbine tria setroi7t at whit % l time the turbine tripped. The tachneter failure was caused by the electdcP starting of a divoo, D. #dUlf.ELy1H:
| |
| Neither phnt for public safety were af fected by the event. All safety Systemt actuated as designed. The AFP's a:tuated and provided feedwater flow to the 5 team Generators as designed. The plant was stabilized in Hot $tandby for investigation of the MfP trip. '
| |
| /
| |
| I I (0078R/0008R) f ,
| |
| l
| |
| * ^ ~~
| |
| _-__-___?_.__l_T_.
| |
| | |
| ;,- -,; 7-.,- -. - .-_---- ;; , -
| |
| _ , - ; ;,--.y ; -----;_ , _ _ _ , , _ _, ._
| |
| b.-
| |
| I LICrutrr EVENT REPORT (LER) TEXT CONTINUATION FACILITY NAME (1) DOCKET NUPEClt (2) LER NLDera (6) Pane (3)
| |
| -(~
| |
| Year g// Sequential f
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| /
| |
| j///'fj/
| |
| Revision
| |
| '\ -
| |
| /// Numhar "A r
| |
| avren. Unit 1 _
| |
| 0 l 5 I O l 0 I o i 4I SI 4 aiB OIOI4 - Ol0 of 3 0F DI 3 TEXT Energy Industry Identification System (E!!$) codes are identified in the text as [xx)
| |
| E. CORRECTIVE ACTIONS:
| |
| The tachometer was repaired by replacing two failed diodes and a resister and monitored for proper operation. The 1A Motor Driven Main reedwater Pump was operated to conduct a Unit startup while allowing the IB MFP to be monitored. The monitoring indicated proper operation of the 18 MFP and it was returned to ..
| |
| service without incident. !
| |
| No further corrective action is planned at this time.
| |
| F. PREVIOUS OCCURRENCES:
| |
| I LER Number LER Title '
| |
| 87409 (Unit 2) Manual Reactor Trip in Response to Decreasing Steam Generator Levels I Resulting from a Feedwater Pump Trip Due to a Defective Speed Control Feedback Loop G. COWGNENT FAILURE DATA:
| |
| a) MAMJFACTURER N0fENCLATlRE PCDEL NLDRER MFG PART NUpBER
| |
| /* A(rPax Electronic Tack Pac Series 600 990-000-815 -
| |
| k Controls Division Precision Tachometer -
| |
| l l
| |
| I 1
| |
| i l
| |
| l l
| |
| 1 I I (0078R/0000R)
| |
| :- :__-_ : ' _ L_~_-.
| |
| | |
| BRAIDWOOD SIMULATOR MALFUNCTION I
| |
| l-
| |
| ~
| |
| ,A:
| |
| | |
| ==Title:==
| |
| Condensate Pump Trip ID: IVM-2 NO: 6.3.4.7.2
| |
| | |
| == Description:==
| |
| Selected condensate pump trips due to
| |
| ~
| |
| ' bearing failure.
| |
| Variations: None. Date: 6/28/87 Rev: 5 l
| |
| Selectable Steps Inputs Comments
| |
| : 1. Select pump IWM-2A FWM-2A = 1A Cond/Cond Booster Pump FWM-2B FWM-2B = 1B Cond/Cond Booster Pump IVM-2C FWM-2C = IC Cond/Cond Booster Pump FWM-2D FWM-2D = 1D Cond/Cond Booster Pump
| |
| : 2. Select delay time 0-99,999 sec.
| |
| Brief Plant Response (IC-17, 100%, all systems in automatic)
| |
| Loss of selected condensate pump will cause a drop in condensate header pressure and flow. This, in turn, will decrease the svetion pressure of the feedwe.ttr pumpe, initially causing their discharge pressure ed flow tu drop.
| |
| IT at high power levci, this could cause a dt.cp in steam generator level and a possibla rescur trip.
| |
| The first annunciators received includet CD/CB PUMP TRIP, IV PUMP NPSH LOW, CD/CB PUMP AUTO START and CD/CB PUMP SELECTOR POSITION WRONG.
| |
| Suggested Instractor Actions None.
| |
| Events: None r
| |
| 0110w:4 882M/4 5/89 l
| |
| : v. .- .. .
| |
| BRAIDWOOD S!!L?ATOR MALFUNCTION jr~
| |
| | |
| ==Title:==
| |
| Steam Generator Feedwater Control Valve Failure Its: WM-3
| |
| %- NO: 6.3.4.7.3 Descriptions- Selected feedwater control valve fails to selected position due to a faulty valve. controller feed controller card, manual control available.
| |
| Date: 3/17/89 Variations: None.
| |
| Rev: 6 Selectable Steps Inputs Comments
| |
| : 1. Select S/G with faulty 'FWM-3A FWM-3A = W510.(1A S/G) control valve FWM-3B FWM-3B = W 520 (1B S/G)
| |
| FWM-3C FWM-3C = W 530 (1C S/G)
| |
| FWM-3D FWM-3D = W 540 (10 S/G) r t .
| |
| k E. Select fail position 0-100 percent open
| |
| : 3. Select ramp time 0-99,999 sec.
| |
| I Brief Flant Response: (IC47,100%, all . systems in automatic)
| |
| Il valve fails oper.- Fted flow to affected steam generdtor increasers, causing that loops AT to increase and its Tavs to decrease. Overall Tavs decreases and the u,ntrol rods tsar withdraw. Stcais flow for ths affected steam f generator vill decrease as tha'; generator's steaa preaante drops. 1.evel ou j that generator initially decreases (shrink) and then increases to the hi-hi ?
| |
| c~
| |
| turbine j
| |
| ( i O 882M/5 5/89 0110w:4
| |
| ===== == _ _ _
| |
| | |
| ___g_ ___ __ _ _. __. _... _._ _ . _._ _ _ _ __ _, _ ._
| |
| BRAIDWOOD SIMULATOR MALESic.! ION l l 1
| |
| l
| |
| (''N,
| |
| | |
| ==Title:==
| |
| Steam Generator Feedwater Control Valve Failure ID: FWM-3 l
| |
| .O Brief Plant Response (continued):
| |
| I trip setpoint. The turbine trip actuates a reactor trip and permissive P-14 causes a feedwater isolation and trips the feedwater pumps. The first annur.ciator received is S/G FLOW MISMATCH STM FLOW LOW.
| |
| l l
| |
| If valve fails closed - Feed flow to the affected steam generator decreases to zero, causing that loops AT to decrease and its Tava to increase. Steam flow from that generator increases due to its increase in steam pressure.
| |
| Overall Tavs increases and control rods insert. The affected steam generator's level initially increases (swell, then decreases to the lo-lo level reactor trip setpoint. The first annunciator received is S/G FLOW MISMATCH FEED FLOW LOW. .
| |
| Suggested Instructor Action:
| |
| None.
| |
| Events:
| |
| : 1) DVR 20-02-88-170: Spiking W Reg. Valve 0110w:4 882M/6 5/89
| |
| _=_=_:__-_=_=_:-_=__-__-_=_-:--r_=___z_- _-__x _ _ _ _ _. -. _ _ _ ________ __ __
| |
| | |
| g (to M .3 DEVIATION INVESTIGATION REPORT (DIR) g Facility Name PAGE Braid _::f 2 1 2 (p) s,s Title Feedwater Flow Perturbation due to a Personnel Error l EVENT DATE DIR NUPSER REPORT DATE SEQUENTIAL REVISION MONTH DAY YEAR STA UNIT YEAR {/{/ NUPRER {/{/ NLDSER PONTH DAY YEAR POWER 11 3 01 7 al B 21 0 01 2 81 8 --
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| 1 I 71 0 -
| |
| 0IO 11 0 21 5 81 8 01715
| |
| , , , i CONTACT FOR THIS DIR NAPE TELEPHONE NUteER AREA CODE David Ibrahim. Technical Staff Ennineer Ext. 2402 8l115 4l5lBl ,12lBl0I1 COMPLETE ONE LINE FOR EACH COMPONEN FAILURE DEstRf8ED IN THf$ REPORT CAU$E SYSTEM COMPONENT MANUFAC- REPORTA8LE CAUSE SYSTEM COMPONENT MANUFAC- REPORTABLE TURER TO NPRDS TURER TO NPRDS I l I I I l l 1 I l l I I l 1 1 I I I I l l I I I i
| |
| $UPPLEfENTAL REPORT EWETED MONTH DAY YEAR p
| |
| SUBMISSION I YES fif ven. enmalete EXPECTED StRNIS$f0N DATE) l ND TEXT Energy Industry Identification System (EIIS) codes are identified in the text as [XX)
| |
| A. PLANT CD'eITIONS PRIOR TO EVENT:
| |
| Unit: Braidwood 2: Event Date: 10-7-88; Event Time: 1440; Mode: 1 - Power Generation; Rx Power: 757.;
| |
| RCS [A8) Temperature / Pressure: N.0.T./N.O.P.
| |
| B. DESCRIPTION OF EVENT:
| |
| On 10-7-88 at 1300 the Shift Test Director ($TD) requested an Instrument Maintenance (IM) foreman, ef ter getting the Shift Control Asem Engineer's (SCRE's) permission, tg install test instrumentatico on Feesinter Flow Transmitters ZFT-0510, 2FT-0520, 2FT-0530 and 2FT-0540 to be used during sta.rtup tests FW75 and TG90.
| |
| The $TD requested from the IM foremen that the test transmitters should be backfilled with demineralized water, connes,ted to 1he tast taps and ngt valved in. The IM foreman asked if the techniciso should talk to the SCRE aga'h Imt the ST!? telt him that it would not be necessary since he already had the SCRE's permissio9 aud these instrteents werv going to be used for startup tests that 6(d ect require his signature.
| |
| At approximately 1440 the feedwater flow sta7ted spiking, the fee % water reguisting valve 2FW520 opened by 153 and the few4 water regulating valve 2FW530 almost closed and required the Huclear Station Operator (NS0) t3 take manual control of the valve. The STD contacted the IM technician over the page and learned that the technician was fillirg the transmitters by cracking open the process isolation valves on 2FT-0520 and 2FT-0530 transmitters. The STO asked tM technician to stop. A few minutes later all the parameters were
| |
| .. stabilised.
| |
| Subsequently the STO discovered that the IM foreman did not specify to the I+1 technician to backfill the transmitters with demineralized water and agi to valve in the transmitters.
| |
| , 2335m(112188)/18 1
| |
| . e
| |
| . . . . _ ~ ~
| |
| | |
| U - _1_. . - ._ -
| |
| DEVIATION INVESTIGATION REPORT TEXT CONTINUATION l
| |
| * rorm p,v r a ,
| |
| I .
| |
| FACILITY NAPE DfR NUPSER fjEE SEQUENTIAL REVIWW 8 Braidwood 2 UNff
| |
| ,g. O) ,
| |
| STA YEAR NUPSER NUPSER 21 o of 2 si s -
| |
| 1I7Io -
| |
| oI o 2 or oi2 TEXT C. CAUSE OF EVENT:
| |
| Hiscommunication led the IM technician to nel follow the STO's instructions, filling the test transmitters by cracking open the process isolation valves let the 2FT-0520 and 2FT-0530 process transmitt(.rs artificially feel a change in feedwater flow in each of their respective loops. As a result, the controller for each loop compensated for that change by sending the'" proper" signal to its feedwater regulating valve. $1nce the sensed change in feedwater flow did not reflect the actual situation in 8 & C loops the control signal to the regulating valve of each loop was not the proper one and caused perturbations in feedwater flow and steam generator (S/G) level in each of the respective loops.
| |
| , D. SAFETY ANALYSIS:
| |
| There were no safety consequences in this event. The change in plant parameters did not require any automatic Safeguard Engineering Features (E5F) actuation.
| |
| In the worst case scenario the reactor would trip and an ESF actuation would take place.
| |
| E. CORRECTIVE ACTIONS:
| |
| : 1) Immediate Actions:
| |
| ,7 The NSO took manual control of the ZFW530 valve and stabilised the plant parameters.
| |
| I L 2) Leno-tava Actions:
| |
| Clear communications between maintenance personnel and the control room staff should be empnastred. This will be tracked to completion by Action Item 457-200-88-17001.
| |
| F. PREVIOUS OCCURRENCES:
| |
| None G. COMPOPFNT FAILURE DATA:
| |
| None 0
| |
| h V 2335m(112188)/1g e
| |
| ..__, . . . . - . . . . . ~ . . . . . . . . . . _ - . . _ . .. ~ ....~.-- -. ... _
| |
| | |
| + _ . . . ._ _ . . _ ,
| |
| .. ._ s. . . ..
| |
| BRAIDWOOD SIMULATOR MALFUNCTION l r~" .
| |
| | |
| ==Title:==
| |
| Feedwater Flow Transmitter Failure ID: FWM-4 1.1 NO: 6.3.4.7.4 l
| |
| | |
| == Description:==
| |
| Selected feed flow transmitter fails to selected setpoint due to transmitter failure.
| |
| Variations: None. Date: 1/7/88 Rev: 5 Selectable Steps Inputs ' Comments
| |
| : 1. Select faulty FWM-4A - FWM-4A = FTS10 for "A" S/G transmitter FWM-4H FWM-4B = FT511 for "A" S/G FWM-4C = FT520 for "B" S/G WM-4D = FT521 for "B" S/G
| |
| 'FWM-4E = FT530 for "C" S/G WM-4F = FT531 for "C" S/G
| |
| ' FWM-4G = FT540 for "D" S/G FWM-4H = FT541 for "D" S/G 6
| |
| : 2. Selected fail value 0 to 5 x 10 lb/hr.
| |
| : 3. Select ramp time 0-99,999 sec.
| |
| ~
| |
| Brief Pls.nt Response: (IC-17, 100%, all systems in automatic) j 71 ant response will depend upon which flow transmitter is used for SGWLC and
| |
| {
| |
| vhich one is failed. j l
| |
| p
| |
| (
| |
| 0110v:4 882M/7 5/89
| |
| )
| |
| i
| |
| -- =1:4 ?- W m =___' h _---T.2***? + ~
| |
| --* * * *- - _ . _ ^ " ~ " '. 'L'-
| |
| | |
| l' BRAIDWOOD SIMULATOR MALFUNCTION
| |
| , i r'
| |
| k-)\
| |
| | |
| ==Title:==
| |
| Feedwater Flow Transmitter Failure ID: FWM-4 l
| |
| Brief Plant Response (continued):
| |
| l l Transmitter nnt used for SGWLC - transmitter will go to selected value, l causing alarms and protective bistables whose setpoints are exceeded to actuate.
| |
| Transmitter used for SGWLC - indication on that channel will go to the selected value. If high, it will cause that generator's feed flow valve to close. .is, in turn, will cause that steam generator's level to decrease '
| |
| eventually to the lo-lo level reactor trip setpoint. The first annunciator received is S/G FLOW MISMATCH STM FLOW LOW. If low, it will cause that generator's feed flow valve to open. This will cause the level to increase to the hi-hi level turbine trip (P-14 setpoint), causing a reactor trip if reactor power is greater than P-8. The first annunciator received includes S/G FLOW MISMATCH W FLOW LOW and NPSH LOW.
| |
| Suggested Instructor Action:
| |
| Place the affected channels' bistables in the tripped mode when requested by the students.
| |
| P Eventst None f
| |
| J 0110w:4 80tM/8 5/89
| |
| | |
| BRAIDWOOD SIMULATOR MALFUNCTION L
| |
| | |
| ==Title:==
| |
| Turbine Driven Feedvater Pump Speed Control Failure ID: FWM-5 D NO: 6.3.4.7.5
| |
| | |
| == Description:==
| |
| Feed pump speed control fails to specified value. Failure of FC-SKSO9A affecting all three pumps, failure of TC-SK509B or FC-SK509C for individual-pumps. Manual control available in both cases. (Failure is in ,
| |
| control card.)
| |
| Variations: None. Date: 7/05/87 Rev: 4 Selectable Steps Inputs Comments I
| |
| i
| |
| : 1. Select pump with faulty FWM-5A A = FC-SK5098, Pump 1B speed controller WM-5B B = FC-SK509C, Pump IC WM-5C C = FC-SK509A, Pumps A, B and C l
| |
| 1
| |
| : 2. Select failed speed 0 to 125 0 = 0 rpm percent 3700 rpm = 54 percent 5200 rpm = 100 percent 5720 rpm = Overspeed trip (110 percent) 6500 rpm = Upper range on tach (125 percent) l 3 .. Select ramp time 0-99,999 )
| |
| sec.
| |
| O I N.]
| |
| 882M/9 5/89 0110w:4 1
| |
| ,__,_n ~ . . - . - . . . -
| |
| . - - ~ _ - - _ - _ . --~ ~ m - - _ _ _ _ _ __7_______ ________,___L______________-___
| |
| | |
| BRAIDWOOD SIMULATOR MALFUNCTION )
| |
| 3
| |
| | |
| ==Title:==
| |
| Turbine Driven Feedwater Pump Speed Control Failure ID: FWM-5
| |
| [V Brief Plant Responle: [ Based on plant at full load when a feed pump speed controller fails.]
| |
| Fails high - Affected pump's speed increases to specified value. The other pump decreases its speed in an attempt to maintain the proper feed header pressure. Feed flow increases to all steam generators, causing their levels to increase slightly and their feed control valves to close down. The first annunciators received include FW PUMP DISCH FLOW HIGH and NPSH LOW.
| |
| Fails low - Affected pump's speed decreases to specified value.- The other pump increases its speed in an attempt to maintain the proper feed header pressure. Feed flow decreases to all steam generators, causing their levels to decrease slightly and their feed control valves to open. The first annunciator received is S/G LEVI.L DEV HIGH/ LOW.
| |
| Suggested Instructor Action:
| |
| None.
| |
| Events: None l-l l
| |
| O 0110w:4 882M/10 5/89
| |
| | |
| BRAID'400D SIMULATOR MALFUNCTION
| |
| | |
| ==Title:==
| |
| Heater Drain Pump Trip ID: FWM-6 (N
| |
| k} NO: 6.3.4.7.6
| |
| | |
| == Description:==
| |
| Selected heater drain pump trips or fails to start due to faulty breaker.
| |
| Variations: None. Date: 7/05/87 Rev: 4 Selectable Steps Inputs Comments
| |
| : 1. Select faulty heater Nd-6A WM-6A = 1A HD Pump drain pump WM-68 FWM-6B = 1B HD Pump FWM-6C FWM-6C = IC HD Pump
| |
| : 2. Select delay time ,
| |
| 0-99,999 see Brief Plant Response: (IC-17, 100%, all systems in automatic)
| |
| When the selected heater drain pump trips, level in the heater drain tank increases. The operating pump's discharge valves will open fully in an i I
| |
| attempt to control the tank's level. The heater drain tank emergency overflow valve opens. The steam generator feed control valves open to maintain levels. Flow from the condensate pumps increase due to a lower discharge head. Overall feedvater temperature decreases, causing nuclear power to increase to maintain the same turbine load. The first annunciator received is HD PUMP TRIP.
| |
| I
| |
| -[
| |
| 0110w:4 882M/11 5/89
| |
| __-m.. . . -_. . .. ._ -
| |
| | |
| 7; ,
| |
| BRAIDWOOD SIMULATOR MALFUNCTION
| |
| | |
| ==Title:==
| |
| Heater Drain Pump Trip ID: FWM-6 3
| |
| ,w/
| |
| Suggested Instructor Action:
| |
| j.
| |
| Clear the malfunction when repairs to the faulty pump have been requested by l the students.
| |
| ic Events: None 1
| |
| 0 -
| |
| 5/89 0110w:4 882M/12
| |
| | |
| P, .
| |
| 1 L
| |
| : j. BRAIDWOOD SIMULATOR MALFUNCTION-l I
| |
| (7 -Title: Feedwater Heater Bypass Valve Failure ID: FWM-7 l
| |
| . (-/- NO: 6.3.4.7.7
| |
| | |
| == Description:==
| |
| Feedwater heater bypass valve fails l-open due to a failed control switch.
| |
| Variations: None. Date ~ 1/7/88 Rev: 6
| |
| )i Selectable Steps Inputs Comments i
| |
| : 1. Select heater string 1, 2 or 3 1 - High pressure heaters.17A and B bypass valve IMM-005 2 - L.P. heater bypass around 15 and 16 heaters 3 25, bypass around 11 to 14 heaters bLJ
| |
| : 2. Select delay time 0-99,999 sec.
| |
| Brief Plant Response: (IC-17, 100%, all sywtems in automatic)
| |
| : 1. When the H.P. heater bypass valve fai3s open, feedwater flow through the H.P. heater strings decreases, Feel header temperature decreases slightly. The colder feed temperature causes Tavg to decrease slightly and the control rods may step out. This increases nuclear power with the same turbine load, thus lowering plant efficiency.
| |
| : 2. About the same effect as "1" with more decrease in feedwater temperature and Tavg.
| |
| O <
| |
| 0110w 4 882M/13 5/89
| |
| | |
| BRAIDWOOD SIMULATOR MALFUNCTION j l
| |
| ,/m r
| |
| | |
| ==Title:==
| |
| Feedwater Heater Bypass Valve Failure ID: FWM-7 1
| |
| -N))
| |
| Brief Plant Response (continued):
| |
| : 3. The bypassing of the low pressure heater string 11-14 will cause the most severe increase in Ex power. If at a high power level, you could receive, a Rx trip. The first annunciators received are various DRAIN TANK LEVEL l 1
| |
| HIGH/ LOW.
| |
| Suggested Instructor Action:
| |
| Clear the malfunction when requested to reposition the valve locally.
| |
| Events:
| |
| : 1. DVR 6-1-85-284: 11C W Heater Hi Level Opens CD025.
| |
| 1
| |
| (,) 2. DVR 6-1-86-148: Generator Load Reduction Due to Heater String ,
| |
| Icolatiou 0
| |
| 0110w:4 882M/14 5/89
| |
| | |
| CEVIATION INVESTIGATION REPCRT TITLE pact FAILURE OF 1CB025 To CLOSE i r a 1 O
| |
| EvtNT DATE DIR NUMBER REPORT DATE </ ,
| |
| SEQUENTIAL REVI510N ;
| |
| MONTH DAY YEAR STA UNIT YE AR NUMBER NUMBER MONTH DAY YEAR 1 l POWER l of e of 7 al s of 6 Of I al 6 -
| |
| 2 I al a --
| |
| oI o i I I of El o CONTACT FOR THIS DIR NAME TELEPHONE NUMBER AREA CODE Lee Wehner Ext. 22Aa R 11 l t 2 l 1 Ia1-l 5 l 4 ial 1 COMPLETE ONE LINE FOR EACH COMPONE URE DESCRfaED IN THIS REPORT CAUSE SYSTEM COMPONENT MANUFAC- REPORTA8tE CAUSE SYSTEM COMP 0NENT MANUFAC- REPORTABLE TURER TO NPRDE TURER TO NPeOS x sIo / f/ l/ I v MI 11 21 0 Y l l l l l l l 1 1 I I I I l 1 ! l l l SUP8LEMENTAL REPORT EXPECTED MONTHI DAY vEAR SUSMISSION l YEs ftr vet. camelete EXPECTED susMisstDN DATE) l No i l l TEXT WHAT HAPPENED?
| |
| While operating at 50". power on 9-7-85. the Control Room operators shutdown the 18 feedwater pump. 09ening of s the FW pump rectreulation valve caused a high level in the 811C FW heater. This actuated the heater bypass valve IC8025 open. After the high heater level condition was reset, the b3 pass valvt would not reclose.
| |
| WHAT WAS T.4E 9007 f.sjg1 The root cause for this geviation is indeterminate. The valve has been stroked several times since the deviatsen occurred. f40 abnorea9tties were found. The Electrical Maintenance department checked the soler,oid valve which controls the position of the bypass valve. The solenoid and control circuit were found to be operating as designed.
| |
| HOW OID f1 AFFECT PLANT nNS/OR PUBLfc SAFETY?
| |
| This deviation did not affect plant or public safety. Turbine load was decreased manually because of the lowering of 1sv . 9 e plant is designed to mitigate the consequences of excessive heat removal by the secondary plant. Continued low feedwater temperatures wculd eventually cause a reactor trip on High Neutrun Flux Power Range. overpower AT. or Overtemperature AT.
| |
| HAs if HAPPENED REF0e.E?
| |
| This deviation has not happened before.
| |
| WHAT WAt DONE 70 enmarcT THE CONDITION AND HOW ARE WE f>0fMO TO PRfVENT eECURRENCE?
| |
| Af ter the high level alarm was cleared, the Control Roog operator could not close the heater typass valve.
| |
| Operators were dispatched to the valve and f ailed the instrument air to the valve actuator. The valve then O stroked closed. AIR 6-85-372 has teen generated to further investigate into the operation of the electrical solenota valve.
| |
| k uosoru
| |
| | |
| 4 WT - 1 DEVIATION INVESTIGATION REPORT
| |
| ' PaOE TITLE
| |
| *RATOR LOAD REOUCTION DUE TO TEATER STRING !$0LATION k DIR NUMBER REPORT DATE
| |
| /ENT DATE SEQUENTIAL REVISION UNIT YEAR y// NUMBER
| |
| {/
| |
| // NUMBER _ MONTH DAY YEAR 1 MONTH DAY YEAR STA POWER 01 9 11 E al 6 01 6 of 1 al 6 -
| |
| 1 1 41 8 -
| |
| 0 l0 110 9I4 Rl A of 91 4 CONTACT FOR THIS DIR CAME TELEPHONE NUMBER AREA CODE W. Kauha. Assistant Technical-Staff tunervisor Ext. 2244 Rl1 lE 2 l1 14 1 I$Ia14 11 COMPLETE ONE LINE FOR EACH COMPONEN LORE DESCRIBED IN THIS REPORT MANUFAC- REPORTABLE CAUSE SYSTEM COMPONENT MANUFAC- REPORTABLE CAUSE SYSTEM COMPONENT TURER 70 NPROS TURER TO NPROS X l i I I I I I N 1 l I I I I I I I I I I I I I l i I I SUPPLEMENTAL REPORT EXPECTED MONTH DAY YEAR SUSMISSION l YEE fif ves. comnlete (yfECTED SUEMIS$10N DATEl l NO l f l TEXT A. PLANT CONDITIONS PRIOR TO EVENT:
| |
| MODE 1 - Power Onaration RN Power 941 RCS (AB) Temperature / Pressure Normal Oeeratino B. DESCRIPTION OF EV!jf,I:
| |
| On 9/15/86, whilt Unit I was operating at steady statu in mode 1, the 11 through If "C" string of low pressure f eewattr heaters were automatically bypassed and isolated as a result of receis tv) the lit feedwater heater level htgh-high alare, Wtwh the turbine governor valves previously positioned 100% open and a decrease in plant efficienc/ as a retult of the het,tc.- string isolation, the electrical output decreased -trom 1090 to 1970 PWe. The alarm cleared shortly thersaf ter, and with heater levels indicating normal the haatar string was restored to normal and electrical output increased.
| |
| C. CAvif 0F EVENT:
| |
| The cause of the heater string isolation was the actuation of the 11C feedwater heater high-high Sevel twt (ch, However. circaustances surrounding the event suggests that an actual hip level did not occur. It appear's that flashirvJ torn plata inside the level switch housing, which resulted in a spurious signal.
| |
| Several other alarms shoul4 have annunciated had an actual high-htgh level in the heater existed. i.e. 11C feedwater heater level high,11C flashtan's leval high 11C flashtank itval high-high. 11C flashtank emergency drain valve open. Additionally, the plant was operating at steady state prior to the event wnich implies that an actual high level did not occur.
| |
| l l
| |
| s l i
| |
| O (0900H)
| |
| | |
| DEVIATION INVESTIGATION REPORT TEXT CONTINUATION DIR NUMBER PAGE TITLE SEQUENTIAL REVISION RATOR LOAD REDUCTION DUE TO HEATER STRING ~
| |
| 0 12
| |
| [(,]/t 7f0N Al 6 0 l 0 0F 01 6 01 1 I l4 l 8 2 TEXT
| |
| : 0. SAFETY ANALYSTS:
| |
| This event did not affect plant or public safety. The only consequence was a decrease in plant efftetency and ultimately a decrease in electrical output. Additionally, the occurrence of this deviation did not interfere with the safe shutdown of the plant. Byron Unit 1 is currently operating at lower steam supply pressures due to a THOT reduction program. The Unit is restricted to 94% power e tch is the maximum steam flow available (Turbine Govenor values 100% open). Had this event occurred prior to the THOT reduction program, reactor power would have increased requiring a manual load reduction. Even under these circumstances this deviation would not have af fected plant or public safety because it would not interfere with safe shutdown of the plant.
| |
| E. CORRECTTVE ACTIONS:
| |
| The magnetrol which actuated the 11C feedwater heater level high-high alarm was removed. inspected, and calibrated. This magnetrol has dual switches; one actuates the high alarm and the other the high-high alarm in conjunction with the interlock. Both switches were found within calibration, however, a washer was found fixed to the switch magnet which actuates the high level alarm. This washer interfered with the switch magnet operation by not allowing it to engage, which explains why the high level alarm was not received. The plant was in normal steady state operation prior to the heater string isolation. No other '
| |
| L.P. heater alarms or automatte actions wh'tch would occur in association with an actual high-high level in g the 11C heater took place. This deviation has not occurred prior to or since this event. Therefore, this deviation appears to be an isolated incident and no further action is requirett. ;
| |
| (
| |
| F. FEyJ0US 0QQDgli: 1 D'IRJuMBER UILE i
| |
| 9CNE G. COMP 0NENT FafLURE DATA:
| |
| NW4ENCLATURE MODEL NUMBER MFC PARY NUMeER MANUFACTURER l
| |
| Not Applicable I
| |
| l
| |
| ' j O ,
| |
| (0900M)
| |
| | |
| BRAIDWOOD SIMULATOR MALFUNCTION 1
| |
| ID: FWM-8
| |
| [^) Title Feed'Line Break Inside Containment N3: 6.3.4.7.8
| |
| | |
| == Description:==
| |
| Breck in the selected feedwater line downstream of the check valve inside containment (break is at inlet nozzle).
| |
| Variations: None. Date: 1/5/88 Rev: 4 Selectable Steps Inputs Comments
| |
| : 1. Select S/G with IMM-8A IMM-8A = Loop A feedlise break FWM-8B FWM-8B = Loop B INM-8C FWM-8C = Loop C IMM-8D FWM-8D = Loop D O
| |
| 5 2. Select leak rate , O to 6 x 10 6
| |
| Normal feedwater flow rate Ib/hr at 100 percent is l 6
| |
| 3.875 x 10 lb/hr l
| |
| : 3. Select ramp time 0-99,999 see I
| |
| l Brief Plant Response: (IC-17, 100%, 6E6 break)
| |
| The affected steam generator will indicate a higher feeiwater flow while the othe; three steam generator's feed flow decreases, j i
| |
| )
| |
| l J
| |
| O 0110w:4 882M/15 5/89 i
| |
| - - _ _ _ _ _ _ _ _ _ _ _ _ _ ,__. _ ._ _ _ _ _ __ _ _ _ ___ _ }
| |
| | |
| . _ ~ . . . . . - . . . . .
| |
| BRAIDWOOD SIMULATOR MALFUNCTION
| |
| | |
| ==Title:==
| |
| Feed Line Break Inside Containment ID: FWM-8 Brief Plant Response (continued):
| |
| Containment humidity, pressure, and temperature increase. Reactor trip will occur due to low-low level or containment pressure safety injection. First annunciator received is SG LEVEL DEV HIGH LOW.
| |
| Suggested Instructor Action:
| |
| None.
| |
| Events: None O .
| |
| ! )
| |
| i l
| |
| l g 1 l 1 li !
| |
| l l
| |
| l O
| |
| 0110w:4 832M/16 5/89
| |
| - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ __ _ _ __ _ . 3
| |
| | |
| . . - - - _-- m_....-. . . . . . _ . . _ . _ . _ . . _ . . . . . _ _ . . . . _ . _ _ _ . - . _ . __ ._ , . _ _ _
| |
| BRAIDWOOD SIMULATOR MALFUNCTION Os
| |
| | |
| ==Title:==
| |
| Feed Line Break Outside Containment ID: FWM-9
| |
| '~ NO: 6.3.4.7.9
| |
| | |
| == Description:==
| |
| Break in feed line outside containment at selected location.
| |
| Variations: None. Date: 3/17/89 Rev: 6 Selectable Steps Inpu'ts Cossents
| |
| : 1. Select break location FWM-9A FWM-9A = Downstream of feed flow detector FE 530 (S/G C) at FE 530 outlet
| |
| ,FWM-98 FWM-9B = Between feed flow detector FE 520 and flow control valve 1FW 520 (S/G B)
| |
| IWM -9C FWM-9C = Common feed header FWM-9D FWM-9D = Between discharge of feed pump B and IIW 001C
| |
| : 2. Select leak rate O to 16 x 10 6 Normal flow rate to each S/G 6
| |
| lb/hr of full power is 3.875 x 10 lb/hr i
| |
| : 3. Settet ramp time 0-99,999 sac.
| |
| ~
| |
| 0110w:4 882M/17 5/89 1
| |
| ----.n.. .na e,:,.. . . . . .- , -- -- - ~ - .
| |
| | |
| L _ _ . . _ . _ _ _ . _ _ . _ . . . . . - _ _ .
| |
| .__g_._..._......_...._.._..____ . _ _ _ _ . _ _ . . . _ . _ , . . ,
| |
| l BRAIDWOOD SIMULATOR MALFUNCTION
| |
| | |
| ==Title:==
| |
| Feed Line Break Outside Containment ID: FWM-9 Eq']/
| |
| > g, l
| |
| Brief Plant Response: [ Based on plant at full load.)
| |
| : 1. When the break occurs, feed flow to S/G "C" increases rapidly, while.the flow on the other S/Gs decreases. Feed control valve FCV 530 closes down due to the high flow. This will cause the leak to decrease. When S/G "C" level drops, the feed control valve opens again, increasing the leak flow. Reactor trip will occur on low-low S/G 1evel.
| |
| : 2. The plant response for this break will be basically the same as above except that the break is upstream of the feed flow detector and the flow to S/G "B" will decrease, causing the feed control valve to open, increasing the flow through the break. S/G "B" level decreases to the ;
| |
| low-low level reactor trip setpoint.
| |
| : 3. When the break occurs, feed flow to all steam generators decreases. Their 1 feed control valves open in an attempt to maintain steam generator level.
| |
| All levels deer sase and a reactor trip on low-low level will occur.
| |
| : 4. When this break occurs, flow from feed pmap "A" increases, robbing flow from feed pump "B". Feed float decreases to all S/Gs. The feed control valves open in an attempt to r,aintain S/G 1evels. A recetor trip will occur due to low-low leve.l.
| |
| Notet The firot annunciators received ury drepending on lack size.
| |
| Suggenced Instructer Actions None.
| |
| 1 Events l'
| |
| , 1) IE Information Botice Eo. 36-106a Feedvater Line break.
| |
| : 2) DVR 06-02-88-115: MFP Recire Line Leak O
| |
| 0110w:4 882M/18 5/89
| |
| | |
| - ______7_ g,3-SSINS Co.: 6835
| |
| ./ IN 86-106
| |
| /
| |
| *h UNITED STATES j
| |
| NUCLEAR REGut.ATORY COMISSION f OFFICE OF INSPECTION AND ENFORCEMENT WASHINGTON, D.C. 20555 December 16, 1986 IE INFORMA. ION NOTICE NO. 86-106: FEEDWATER LINE BREAK Aadressees:
| |
| All nuclear power reactor facilities holding an operating license or a con-struction permit.
| |
| | |
| ==Purpose:==
| |
| This information notice is to alert addressees of a potentially generic probles with feedwater pipe thinning and other problems related to this event. -
| |
| Recipients are expected to review the information for applicability to their facilities and consider actions, if appropriate, to preclude similar problems occurring at their facilities. However, suggestions contained in this information notice oo not constitute NRC requirements; therefore, no specific action or written response is required.
| |
| Description of Circumstances:
| |
| On Tuesday, December 9,1986, at 2:20 p.m., both units at the Surry Power Station were operating at full power when the 18-inch suction line to the main feedwater pump A for Unit 2 failed catastrophically. Eight workers who were replacing thermal insulation on a nearby line were burned by flashing feedwater.
| |
| All wars transported to area hospitals. Two workers were treated and released. i Four other workers subsequently died, I
| |
| Units 1 and 2 are identical. In each unit, ftedwater flows from a 24-inch header to two 18-inch suction lines that each supply one of two main feedwater pumps. At maximum load under normal conditions, feedwater flow through each pump is 5 million Ib/hr. Feedwater temperature, pressure, and enthalpy are i
| |
| 370*F, 450 psig, and 346 Btu /lb, re.spectively. At these conditions the fluid l
| |
| l is in the single phase, liquid only regime. That is, the piping does not see a mixture of licuid and vapor.
| |
| l The event was initiated by the main steam isolation valve on steam generator C failing closed. Because of the increased pressure in steas generator C that collapsed the voids in the water, the reactor tripped on low-low level in that 1 steam generator. A 2-by-4 foot section of the wall of the suction line to the M main feedwater pump was blown out and came to rest in an overhead cable tray. The break was located in an cibow The lateral in force reactive the 18 inch lineby generated about one foot escaping from the 24-st.ch header.
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| 8612160250 s
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| (
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| - - _ 1
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| | |
| m 1
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| 1 I
| |
| IN 86-106 D
| |
| s Deresber 16, 1986 Page 2 of 3 I
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| l feedwater completely severed the suction Ifne. The free end whipped and came j
| |
| to rest against the discharge line for the other pump.
| |
| l Steam flashing fma the break ard condensing in control cabinets and in open conduit piping apparently causeo the fire suppression systes to actuate, resulting in release of halon and carbon dioxide in the energency switchgear room and in various catie tunnels and vaults and in the cable spreading room.
| |
| cperators isolated Because of high lines carrying the volume of water energy fluids and to areas steam being inundated by released, steam Steam generator water levels were maintained with the auxilbry feedwater systes, and system cooling was provided by actuating atmospheric dump valves as necessary.
| |
| The primary systes responded normally to the loss of load transient with a partial loss of main feedwater. Primary coolant temperature was stabilized at 520*F and pressurizer level was recovered as it reached the low level set point. Primary pressure decreased froc 2235 to 2015 psig following the reactor trip. By 2 a.m. on the following day, reactor temperature had been reduced to the point where the residual heat removal system could be put on line. The unit reached cold shutdown that morning. Daring the recovery offort, the operators and the plant performed as &xpected.
| |
| Discussfor.:
| |
| The pipe material is A-1068 carbon steel and the elbow is 18-inch, extra strong D A-234 grade WP8 carbon steel. Nominal wall thickness of the' suction piping is 0.500 inch. Measurements of the wall fragment demonstrated that the wall had been generally eroded to about 0.25 inch and was one of the causes of the failure. Preliminary examination of the 2-by-4 foot section of pipe blown out during the event shows the thinning to be relatively uniform except for some small localized areas. The thinnest areas are localized and appear to be about 1/16 inch thick. Some corrosion pitting is present. A preliminary micro-examination Indicated that thi pipe surface near the fracture had not been highly strained as with a high stress event, such as a high pressure spike in the system.
| |
| It has not been determined &t this time whether a pressure spike in the system was a contributor to this event. There was no damage evident in the hanger supports to the condensate system.
| |
| Inspection revealed a disabled check valve in the discharge piping of the A main feedwater pump. This check valve was found with its seat displaced and a I hinge pin missing.
| |
| On December 10, the licensee shut down Unit I for examination of the condition of feedwater piping. Inspection of the Unit i feedwater piping shows wall thinning similar to but not as severe as that in Unit 2.
| |
| The NRC dispatched an augmented investigation team (AIT) to the site . The AIT includes a metallurgist and a water hammer analyst.
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| D ~
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| l I
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| ____-_- 1
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| | |
| IN 86-106 December 16, 1986 Page 3 of 3 The NRC will issue additional information as more inspection and analysis is completed.
| |
| No specific action or written response is required by this information notice.
| |
| If you have questions abaut this matter, please contact the Regional khinis-trator of the appropriate NRC regional office or this office.
| |
| G Edward L. Jordan,, Director Olvision of Emergency Preparedness and Engineering Response Office of Inspection and Enforcement Technical
| |
| | |
| ==Contact:==
| |
| Roger Woodruff, IE (301)492-7205 Vincent Panciera, Region !!
| |
| (404) 331-5540
| |
| | |
| ==Attachment:==
| |
| | |
| List of Recently Issued IE Information Notices O -
| |
| | |
| - ..-m - . . . . .
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| __ $"QaJ s -]
| |
| DEVIATION INVESTIGATION REP 03T (DIR)
| |
| . FormRev2.0j '
| |
| FactIity Namn pagt Byron helaar Power Statlan i 3J Title i.
| |
| Or 2B Main Feedwater Pume Racirculatter Line tank EVENT DATE DIR NUleER , REPORT DATE SEQUENTIAL REVISION leNTH DAY YEAR STA UNIT YEAR NLRSER NUDGER IGNTH ' DAY YEAR 1 .
| |
| POWER 1 11 2 14 a 18 o 16 0 12 8 18 -
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| 1 11 I 5 -
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| 0Io 1 I i o 14 18 -
| |
| CONTACT FOR THIS DIR NADE TELEPHONE NUfeER ARLA CODE D. Brindle. Onoratina Enaineer Ext. 2218 8l115 213l4l-.5l4l4l1 COMPLETE ONE LINE FOR EACH C0pr0NENT Af tter DESCRIBED IN THIS REPORT CAUSE SYSTEN COMPONENT MANUFAC- REPORTABLE CAUSE SYSTEM COMPONENT MANUFAC- REPORTA8LE j 1
| |
| Tiante TO NPRDS TURER TO NPRDS 1 I I I I I I I I I I I I I i I I I i i I i l' 1 I I SUPPLEMENTAL REPORT EXPECTED XPECTED
| |
| $USNISSION l
| |
| [l YES (if ves. enmolate EXPECTED SUBMISSION DATE) l ND 0 4 1 5 8 9 TEXT Energy Industry Identificatten System (EIIS) codes are identified in the text as [XX)
| |
| A. PLANT ColeITIONS PRIOR TO EVENT:
| |
| Event Date/ Time 11 E / 2000 Unit 2 N00E 1 - Power Onoratter- Rx Power _jg_ RCS [A8] Temperature / Pressure Normal Onoratina B. QlEJIPTION OF EVENT:
| |
| On 11-24-88 at 2000 hours Unit 2 was at 48 percent power ehen shift por:annel received a call concerning a leak on 401 elevation of the turbine building. Subsequent investigation revealed the source of the leak to be a hole in the 28 Main Feedwater (FW) [5J) Puup recirculation I{ne at the elbow downstream of valve 2FWO128. The pump was taken out-of-service. Only one pump is required at 48 percent power, so a power reduction was not required.
| |
| All operator actions were correct. No other system or consnents . vere inoperable at the beginning of 'this event that contributed to the event. No safety system actuations occurred.
| |
| (0211R/0024R)
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| - _ _ _ _ . __=_- ____ =-
| |
| | |
| .. . . . . ~ ~ - .- m - -
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| ;- - -- -~- ~ ~ ' ~
| |
| 5 DEVIATION INVESTIGATION REPORT TEXT CONTINUATION Form Rev 2.0 C . FACILITY NADE DIR lasare Pard
| |
| ^
| |
| SEI)UENTIAL REV!510N STA UNIT YEAR 188888 IAmara Svron Nucleme Power Station 0 16 0 12 a la -
| |
| 11115- -
| |
| 0 IO 2 0F 0l3 TEXT Energy Industry Identification System (EII5) codes are identified in the text as (XX]
| |
| C. CAUSE OF EVENT; The cause of the event was erosion of the pipe elbow due to leakage through the upstream 28 Feodwater Pump
| |
| . recirculation valve 2FWO129. The problem of elbow erosion downstream of the Nain Feedwater recirculation valve had also been seen on Unit 1. The problem with Unit I was found to be the mis-eppilcation of a flow directional cone on the downstream side of the valve plug. The nose cone directed any valve leakby directly into the elbow immediately downstream. The nose cones on all of the 2FWO12 valves were removed' prior to initial startup in response to the Unit 1 problem. It is possible that the erosion damage on 2FWO12B was done prior to the cone removal, but it seems more Itkely that the valve may have been excessively leaking by the' seat. Unit i elbows are being monitored for pipe thinning and the indications are showing no thinning in the general area of the elbow. Unit 2 elbows downstream of 2FWO12A and 2FWO12C were tested and the results showed that there has been no thinning. The indications showed an average elbow well thickness of 0.82 inches while a new elbow has a nominal well thickness of 0.75 inches.
| |
| Further detal16d investigations will be conducted during the upcoming refueling outage. The final determinations of the causes will be reported in a supplement to this report .
| |
| D. SAFETY ANALYSIS:
| |
| The plant or public safety was not compromised by the failure of the elbow downstream of the 2FWO128. The leak found on the pipe elbow was from a sea 11 hole in one location. The leak was on the condenser vacuum
| |
| ( side of IFWO128. therefore most of the water / steam was drawn into the main condenser steam space. The recirculation line on the main feedwater pump can be easily isolated. The operation of the pump was not immediately affected by the leak. The pump was taken out of service in order to repair the elbow.
| |
| E. CORRECTIVE ACTIONS:
| |
| The pump was taken out of service and the manual isoletten valve was (1 M ed. The remaining feedwater pump recirculation elbows were examined and showed no erosion on the elbow. The damaged pipe elbow will be replaced inn the next refueling outage under Nuclear Work Roguest 862633. The valve internals on 2FW0128 will be inspected to verify proper valve closure under Nuclear Work Roguest 862836. Both Nuclear Work
| |
| : l. Requests will be tracked under Action Item Record 88-0294. Pipe well thickness is being monitored under i j the station's erosion / corrosion program. Relocation of the pipe elbows to reduce erosion potential was f l also considered but determined to not be a cost effective measure. Additional actions will be reported in j l
| |
| a supplement to this report. '
| |
| F. RECURRING EVENTS SEARCH afb ANALYSIS:
| |
| a) EVENT SEARCH (DIR. LER)
| |
| There was one previous occurrence on Byron Unit 1 in early 1986. .
| |
| b) IDE11$TRY SEARCN (OPEX8m NPRD$)
| |
| Problem known; vendor contacted and substantiated problem.
| |
| 50ER 87-03 and 82-11 .
| |
| I (0211R/0024R)
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| - _ n.e ..
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| - _ --- n.. .-
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| | |
| _ _ . ._ .._ u. - - - - - -- -
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| 4
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| - DEVIATION INVESTIGATE 0ti REPORT TEXT CONTINUATION Form Rev 2.0
| |
| ,.A .
| |
| FACILITY NAE DIR NLNSER PAGE j SEQUENTIAL REVISION L \_ - STA UNIT YEAR NUPSER Ni#eER
| |
| ,Ryrgn Nuclear Power Station 0 16 0 12 8 18 -
| |
| 1l1l5 -
| |
| 0 l0 3 0F 013 TEXT Energy Industry Identification System (EIIS) codes are identified in the text as [XX) c) f6R Elbows:
| |
| 827560 1C IfWO79AC 1C Recirculation Line Elbow (4-2-86) 527668 18 1FWO79A818 N PR Recire Line Elbow (4-11-86) 862559 28 2N079A8 28 N PR Recirc Line Elbow (11-24-88)
| |
| Recirc Valves:
| |
| 827552 IFW12A verify proper closure 527554 I W128 ver*fy proper closure d) ANALYSIS l There has been several instances of pipe erosion problems in the industry.
| |
| This event, while significant in that sense, appears to have minimal operating er safety significance itself. Corrective actions appear adequate.
| |
| G. COMPONENT FAILURE DATA:
| |
| MANUFACTURER NDENCLATURE MODEL NUPSER MFG PART NUPEER Control Components Inc. dragvalve 922101036 l
| |
| H. OTHER RELATED DOCUE NTS:
| |
| NRC IE8 87-11 I. EFFECTIVENESS REVIEW: ,
| |
| Scheduled 03-01-90 J. ADDITIONAL DATA:
| |
| a) Affected Technical Specification: None b) Prncedures: None c) Cause Code: XPRP942P91 j l
| |
| d) Equipment involved: W Pump Recirculation Line e) Other: Valve leakby, Pipe erosion v ,
| |
| N.
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| (0211R/0024R)
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| . ~ _ - - ~ . . - _ . . . ~ . . , . . , . - - . .
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| | |
| e 1
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| BRAIDWOOD SIMULATOR MALMNCTION l
| |
| 1 l
| |
| | |
| ==Title:==
| |
| Feedwater Pump Speed Control Oscillates ID: FWM-10 1
| |
| D)\'- NO: 6.3.4.7.10
| |
| | |
| == Description:==
| |
| Selected feedwater pump's speed oscillates at a set frequency due to failure in pump's
| |
| , speed controller or master controller.
| |
| Manual control possible.
| |
| Variations: None. Date: 1/5/88 Rev: 5 Se3ectable Steps Inputs Comments
| |
| : 1. Select faulty pump WM-10A 1 = 1B MFP FWM-10B 2 = 1C MFP
| |
| : 2. Select oscillation 0 to 50 percent Turbine full load speed =
| |
| f amplitude 5200 rps
| |
| : 3. Select ramp time 0-99,999 Brief Plant Response:
| |
| When the selected pump's speed starts oscillating, feed flow to all steam generators also starts oscillating. These oscillating feed flows can cause actuation of various alarms including S/G LEVEL DEV HIGH LOW and FEED FLOW /STM FLOW MISMATCH alarms. A reactor trip and/or turbine trip could occur due to oscillating levels.
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| Suggested Instructor Ac. tion:
| |
| None.
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| Events:
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| None
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| .L 0110w:4 882M/19 5/89
| |
| | |
| 4 #, ,. .
| |
| 'BRAIDWOOD SIMULATOR MAL W NCTION 1
| |
| l
| |
| | |
| ==Title:==
| |
| Feedwater Heater Tube Break ID: FWM-11 1
| |
| f~]
| |
| NO: 6.3.4.7.11 l-
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| | |
| == Description:==
| |
| Sel'cted a feedwater heater experiences a tube break of specified size.
| |
| l
| |
| - Variations: None. Date: 1/5/88 Rev: 5 I
| |
| Selectable Steps Inputs Comments
| |
| : 1. Select faulty heater FWM-11A - FWM-11D FWM-11A = 12A W HTR FWM-118 = 14C W HTR FWM-11C = 15A FW HTR FWM-11D = 17B FW HTR
| |
| : 2. Select leak rate 0 to 2000 spa Based on normal pressure
| |
| ~
| |
| 3.- Select ramp time 0 to 99,999 sec Brief Plant Response:
| |
| When the t'reak occurs,' level in the selected heater shell increases, causing various level alarms including LEVEL HTR HIGH LOW and CB PUMP FLOW HIGH and the dump valve to the condenser to open. It also will cause the extraction steam valve to that heater (if it has one) to close. Depending upon which heater has the tube break, the condensate and feedwater flows will be affected, causing a reduction in steam generator level.
| |
| Suggested Instructor Action:
| |
| None.
| |
| N Events: None 0110w:4 882M/20 5/89
| |
| | |
| 1 BRAIDWOOD SIMULATOR MALFUNCTION
| |
| | |
| ==Title:==
| |
| Auxiliary Feedwater Pump Trip ID: FWM-12
| |
| .{- NO: 6.3.4.7.12
| |
| | |
| == Description:==
| |
| Selected pump fails to start or trips, if running. For motor-driven pump, feeder breaker.from Bus-141 trips. For diesel- J driven pump, engine failure lockout relay K-12 fails to energize.
| |
| Variations: None. Date 1/5/88 Rev: 5
| |
| -Selectable i Steps Inputs Conunents
| |
| : 1. Select pump 1, 2, 3 1 = 1A (Motor-driven pump) 2 = 1B (Diesel-driven pump).
| |
| 3 = Both !
| |
| O '
| |
| V -
| |
| Brief Plant Response: [ Based on plant being in hot standby with the auxiliary feedwater system maintaining S/G 1evel.]
| |
| If actcr-driven pump is tripped, feed flow to All steam generators will decrease or drop to zero.
| |
| If the diesel-driven pump is tripped, feed flow to gli steam generators will decrease or drop to zero.
| |
| In both cases, the affected steam generator levels decrease.
| |
| The first annunciator received is AF pump TRIP.
| |
| Suggested Instructor Action:
| |
| l None.
| |
| Events: 1) DVR 06-02-87-105: 2B Diesel AF Pump Inoperable i 2) DVR 06-01-87-119 1A AF Pump Trip 0110w:4 882M/21 5/89 f
| |
| V -- - - - - .... . -. - . ._
| |
| | |
| . CEV!ATION INVESTIGATION REP 02T
| |
| }
| |
| l l
| |
| TITLE PAGE
| |
| . 28 OfEEEL DRIVEN AUXILIARY FEEDWATER PUMP INDPERABLE EVENT DATE DIR NUMBER REPORT DATE
| |
| ? // SEQUENTIAL // REVISION MONTH DAY YEAR STA UNIT YEAR NUMBER NUMBER MONTH DAY YEAR 3 POWER LIVIL 1l o of 2 al 7 of 6 al tal 7 1 l cl ?
| |
| oIo 11 1 . 11 2 81 7 ol of a CONTACT FOR THIS gg, CAME TELEPHONE NUMBER AREA CODE Lee tuat. 411t. tunarintendent onaratino Ext. 2213 8l 1 lE 213 l4l -1El4 I4 l 1 00MPLETE ONE LINE FOR EACH COMPONE URE DEicRIBE0 IN THIS REPORT CAUSE SYSTEM COMP 0NENT ' MANUFAC- REPORTABLE CAUSE SYSTEM COMP 0 MENT MANUFAC- REPORTABLE TUREE TO NPRDS TUREE TO NPRDS I l l l l i f f j l l l l l 1 l I i l I i i l l i l SUPPLEMENTAL REPORT EXPECTED MONTH DAY YEAR SUBMISSION DATE I l 1 I 'iEt fir van enenlete EXPECTED tumMit110N DATE1 Xi NO I I I TEXT A. PLANT CONDITIONS PRIOR TO EVENT:
| |
| Event Date/ Time 10/2/87 / 1600 ,,,
| |
| f3 Power Onaration RCS (AB} Temperature / Pressure i Unit 1 MODE 1 - Rx Power 97 la7 / 2215 osia Unit 2 MODE 1 - Het Standhv Rx Power o RCS (AB) Temperature / Pressure 42E*F / 2000 otig.,
| |
| : 8. DEttetPTION OF EVENT:
| |
| On 10/2/87, at 1316 hours. during the recovery of the plant following the loss of normal f eed to the Station Auxiliary Transformers, the 28 Diesel Driven Aust11ary Feedwater Pump (AF)[SA) discharge pressure started cycling between 500 and 2000 psig. This occurred at the same time that the ( AF) diesel oil day tank was being refilled. At 1318 hours the AF pump was tripped and placed in Pull to tock and declared inoperable. No other systems were inoperable at the beginning of this event which contributed to this event and all operator actions were correct.
| |
| C. cAust 0F EVENT:
| |
| The operators were in the process of refilling the AF diesel oil day tank when the engine began surging.
| |
| Subsequent investigation showed the fuel system had lost its prime. The combination of a low tank level (f tve inches above the bottom) and the fill piping located directly above the supply piping ts suspected to have caused turbulence in the tank. introducing air into the fuel line.
| |
| D. SAFETY ANALYSI5*
| |
| The loss of the B train of Aux 11' ry Feedwater during this event did not affect the safety of the plant or threaten public safety. The reoun h'l motor driven pump was running and maintained the steam generator levels.
| |
| /^
| |
| \ \
| |
| (1725M/0201M)
| |
| | |
| DEV!ATION INVESTIGATION REFORT TEXT CONTINUATION DIR NUMBER PACE TITLE SEQUENTIAL REVISION
| |
| __m STA UNIT YEAR NUMBER NUMBER
| |
| .B OIEEEL ORIVEN AUXILIARY FEEDWATER PUMP INOPERABLE 01 6 01 2 81 7 ""
| |
| 1 IO l5 ""
| |
| 0 l0 2 0F 0 l2 TEXT E. CORRECTIVE ACTIONS:
| |
| The Auxiliary Feedwater Diesel was reprimed and run. The day tank level gauge calibration was checked and found in tolerance. An Operator af d has been posted at each control switch displaying the tank curve.
| |
| The following procedure revisions will be made and tracked by AIR 87-26b: The Byron Emergency Procedure (BEP) fo16out page will be revised to maintain the day tank level above 50%. Also the Byron Operating Procedure (SOP) for running the AF diesel will be revised to call for maintaining level above 50% and to fill slowly if its level is below 50%. Also the 80P's for filling the Day Tank will be revised to call for tilling the tank slowly if its level is below 50% and the engine is running.
| |
| F. PPEVIOUS OCCURRENCES:
| |
| LER Nt9EER IIILL NONE G. COMPONENT FAILURE DATA-a) MANUFACTURER NOMENCLATURE MODEL NUMBER MFC PART NUMBER Not Applicable v
| |
| 1 b) RESULTS OF NPROS SEARCH:
| |
| Not Applicable I
| |
| (1725M/0201M)
| |
| | |
| r F ~ r^ - 1 DEVIATION INVESTIGATION REFORT "TLE PACE [
| |
| stfrILIARY FEEDWATER M'MP TRIP DUE TO $[NSING LINE CRAIN BEING LEFT OPEN EVENT DATE DIR NUMBER REPORT DATE SEQUENTIAL // REVISION MONTH DAY YEAR STA UNIT YEAR NUMBER NUMBER MONTH DAY YEAR 1 POWER LEVEL of 9 11 4 al 7 of 6 01 1 al 7 --
| |
| 1 1119 --
| |
| O lo 11 0 217 817 , ,, ,
| |
| CONTACT FOR THIS DIR NAME . _ TELEPHONE NUMBER AREA CODE R. Flahive. Rad Cham Suoervisor Ext. 2233 811 1E 2 l3 l 4 l - l E l4 l4 l1 COMPLFTE ONE LINE FOR EACH COMPONEN A LURE DESCRIBED IN THIS REP 03T CAUSE SYSTEM COMP 0NENT MANUFAC- REPORTABLE CAU$E SYSTEM COMPONENT MANUFAC- REPORTABLE TURER TO NPRDS TURER TO NPRDS I I I I I I I I I I l I I I I I I I I I I I l i 1 l SUPPLEMENTAL REPORT EXPECTED MONTH DAY YEAR SUBMISSION 1 YES fif yet. comolete EXPECTED SUBMISSION CATE) l NO TEXT A. PLANT CONDITIONS PRIOR to EVENT:
| |
| \ Event 04te/ Tit.1e 9/14/87 / 1100
| |
| ,V Unit 1 MODE 1 - Power doeration Rx Power 98 RC5 (As) Temperature / Pressure Nonnal Oceratina Unit 2 MODE 1 - Power Doeration Rx Power _jlL__ RCS (A8] Temperature / Pressure Normal Doeratina B. DESCRIPTION OF EVENT:
| |
| At 0900. 9/14/87 a Rad / Chem Technician (RCT) obtained a sample of the 1A Auxiliary Feedwater ( Ar) [8A) pump suction water from the 1A AF Pump suction pressure instrument tap as specified in a prerequisite for the Auxiliary Feedwater Pump Startup Procedure. This was done prior to the starL of the AF pump to verify the water quality in the suction line. Operating opened the instrument valve to tilow a one-half hour purge of the line prior to sampling. The RCT then sampled the line with an inline specific conductivity meter to verify the water was less than the 2.0 micromho/cm guideline limit. The technician obtained a value of 0.87 microm o/cm at the instrument tap, which was reported to the Unit One Operator at 0907.
| |
| After the sample was taken and the water quality was vertfied, the sample valve was inadvertently left open due to a communication problem between the Operator and Rad / Chem T6chnician. The pump was started at 1109 for the performance of surveillance 18V5 7.1.2.1.a-1. and shutdown at 1112. due to a trip on low suction pressure. The pump was restarted at 1119 and shutdown at 1123. At 1128. NWR B48997 was generated to investigate the low suction pressure instrumentation.
| |
| C. CAUSE OF EVENT:
| |
| The drain open on 1PT-AF051 for sampling caused an erroneous low suction pressure of f our pounds, which caused the apparent suction pressure to exceed the low setpoint, causing the pump to trip.
| |
| J l I (1676M/0199M)
| |
| | |
| I l
| |
| I j
| |
| DEVIATIDN.1::YESTIGATIDN REPDRT TEXT CDNTINUATIDN l
| |
| *'TLE DIR NUMeER PAGE SEDUENTIAL REVISIDN l UNIT YEAR NUMBER NUMBER !
| |
| (/1 A AUXILIARY FEEDWATER PLMP TRIP DUE TD _STA SENSING LINE DRAIN BEING LEFT DPEM M6 d1 al7 "'
| |
| 1 11 19 --'
| |
| oIo M of2 TEXT i
| |
| SAFETY ANALYSIS:
| |
| )
| |
| D.
| |
| The pump tripped on low suction pressure with the pump running in recirculation. The transient seen by the pump would have been greater if the pump had been started in its normal mode (Aux Feed to steam generators) and as a result the pug may have tripped. The event did not endanger the health and utcty of the plant or public as the IB Auxiliary Feedwater pump was still operabit, and capable of supplying water to the steam generators if required.
| |
| E, MTIVE ArTIONS:
| |
| Sue to the problem of obtaining a sample locally without affecting a pressure transmitter, the requirement for samp1tng the AF suction prior to the running of the pumps, has been deleted from the startup procedures. The water historically has been of good quality, with no indications of Essential Service Water intrusion into the suction Itne.
| |
| AIR 87-237 was written to develop a program to track the use of local sample points which will include preapproval and verification of proper realignment after sampling.
| |
| AIR 87-241 was written to write a procedure for sampling the Steam Gcnerators during Cold Shutdown. which is done locally but not off of pressure transmitting lines.
| |
| U The tygon lines attached to the drain on 1PT-AF051 have bar removed and the line has been capped.
| |
| F. PREVIOUS OCCURRENCES:
| |
| LER NUMBER IIILE i:cxE G. COMPONENT FAILURE DATA a) MANUFACTURER NOMENCLATURE tgQEL NUptER MFG PART NUPRER Not Applicable b) RESULTS OF NPRDS SEARCH:
| |
| Not AppltCable rh i
| |
| I I
| |
| (1676M/0199M)
| |
| | |
| BRAIDWOOD SIMULATOR MALFUNCTION
| |
| /7
| |
| | |
| ==Title:==
| |
| Loss of Feedwater Pump Speed Control IDt FWM-13 N0! 6.3.4.7.13
| |
| | |
| == Description:==
| |
| Loss of actual governor valve position signal due to loss of LVDT on governor servo-actuator.
| |
| Variations: None. Date: 1/5/88 Rev: 4 Selectable Steps Inputs Comments
| |
| : 1. Select pump with 1,2,3 1 = B-MTP failed input 2 = C-MTP 3 = BOTH
| |
| , 2. Select delay time 0 - 99,999 sec.
| |
| Brief Plant Response: [ Based on fault occurring at full power.]
| |
| Affected feed pump apeed increases to trip setpoint. No manual control sellabis. Tha u .r y-, d: rsassa speed in an attempt to maintain the proper feed header pressure. Feed flow increases to all steam generators causing their levels to increase slightly and their feed control valves to close down. Affected pump trips causing a reduction in feed flow. Drop in feed flow causes RCS temperature to increase which in turn causes the control rods to insert. Steam generator level decreases and eventually causes a reactor trip on low-low steam generator level. The first annunciators received are S/G FEED FLOW /STM FLOW MISMATCH and MFP TRIP. 1 Suggested Instructor Action:
| |
| Clear malfunction.
| |
| Events: None 0110w:4 882M/22 5/89 L - - . - - _- - -- - - . .
| |
| | |
| BRAIDWOOD SI!EiLATOR MALFUNCTION T
| |
| | |
| ==Title:==
| |
| Main Feedwater Oil Pump Failure ID: FWM-14
| |
| '\ % -
| |
| NO: 6.3.4.7.14-Description Operating main oil pump trips and j standby oil pump starts.
| |
| Variations: Both operating and standby Date 1/5/88 oil pumps fail. Rev: 4 Selectable Steps Inputs Comunents i
| |
| )
| |
| : 1. Select MF? FWM-14A FWM-14A = 1B MFP FWM-14B FWM-14B = 1C MFP .
| |
| : 2. Select oil pump 1,2,3 1 = "A" main lube oil pump 2 = "B" main lube oil pump.
| |
| 3 = Both oil pumps 1
| |
| : 3. Select delay time 0 - 99,999 sec. i Brief Plant Response:
| |
| A or B - If you select the operating main oil pump (MOP) to trip, the standby MOP should start at 50 lb. decreasing.
| |
| Both - If you select "both," then MFP should trip on closure of stop valves with the emergency oil pump starting at a bearing pressure of 10 pais.
| |
| Note: No annunciators are received.
| |
| Suggested Instructor Actions None.
| |
| Events: None 0110ws4 882M/23 5/89
| |
| = = = = = = = - - . = _ - . . . ~ ~ - - - - .- - .. . . . . . . -
| |
| }
| |
| | |
| < BRAIDWOOD SIMULATOR MALFUNCTION l
| |
| l'
| |
| ; Titles Feedwater Regulation Bypasa Valve Failure ID: FWM-18 Al NO: 6.3.4.7.18
| |
| | |
| == Description:==
| |
| The' feed reg bypass valve fails to
| |
| , selected position due to a faulty
| |
| . valve controller feed controller card, manual control is available.
| |
| Variations: None. Date: 3/E /89 Rev: 5 Selectable Steps Inputs Comments
| |
| : 1. Select valve- FWM-18A FWM-18A = W 510A FWM-18B FWM-18B = FW 520A FWM-18C FWM-18C : FW 530A FWM-18D FWM-18D = W 540A
| |
| - 2. Select position 0-100 percent
| |
| -3. Select ramp time 0-99,999 sec. {
| |
| i l
| |
| i 1
| |
| I 4
| |
| 0 5/89 0110w:4 882M/29
| |
| )
| |
| n n _-_ . - . , - - - , , - - , - - - , - - - - - - -
| |
| | |
| . g _ .. _.._... ._.2_-. . . . , - . . _ .
| |
| BRAIDWOOD SIMULATOR MALFUNCTION
| |
| | |
| ==Title:==
| |
| Feedwater Regulation Bypass Valve Failure ID: FWM-18 1
| |
| Brief Plant Responses (IC-11, 8%)
| |
| Failure of the feed reg. bypass valves closed during a plant startup causes a loss of feedwater to that S/G. The first annunciator received is S/G LEVEL DEVIATION HIGH LOW.
| |
| Suggested Instructor Action:
| |
| None.
| |
| Ev ents: None O
| |
| ?
| |
| 0110w:4 882M/30 5/89
| |
| | |
| l' g-BRAIDWOOD SIMULATOR MALNNCTION:
| |
| Title .-Feedwater Preheater Bypass Valve Malfunction ID: WM-19
| |
| [ ' NO: 6.3.4.7.19
| |
| | |
| == Description:==
| |
| The preheater bypass valve fails during the plant startup.
| |
| m Variations: Date: 3/17/89 Rev 5 Selectable Steps. Inputs Comments
| |
| : 1. Select S/G W M-19A FWM-19A = FW 039A FWM-19B FWM-19B = FW 039B FWM-19C FWM-19C = FW 039C FWM-19D FWM-19D = FW 039D
| |
| : 2. Select position 0-100 percent 0 = closed i
| |
| 100 = open Brief Plant Response: (IC-11, 8%)
| |
| If the preheater bypass valve fails during low flow and low temperature operation, the S/G affected will lose feedwater flow entirely, the first annunciator received is S/G LEVEL DEVIATION HIGH LOW.
| |
| Suggested Instructor Action:
| |
| None.
| |
| . Events:
| |
| : 1) LER 06-02-88-007: FWO39C Fails To Open.
| |
| : 2) DVR 06-02-88-004: FWO39A Fail to Close.
| |
| 3). DVR 20-02-88-168: FWO39A Fails to Open.
| |
| : 4) LER 06-02-88-009: FWO39B/C Fail to Open.
| |
| O 0110w:4 882M/31 5/89
| |
| _ . _ _ - _ - _ _ - - _ . - - _ . - __ - _ -_-.--_ - - __-__-____=_--_--_---___:__-___ -
| |
| | |
| N <% ~ l 9 l LICENSEE EVENT REPORT (LER)
| |
| Facility Name (1) Docket Number (2) . Pace (35
| |
| .g Byron. Unit 2 01 51 01 01 01 41 51 5 1 lofl0i3
| |
| ( Title (4) Feedwater Isolation Actuations Due to Steam Gent retor Preheater Bypass Valve failure to Open Event Date (5) LER Number (6) Report Date (7) Other facilities Involved (B)
| |
| Facility Naews l Dociret Number (s)
| |
| Honth Day Year Year // Sequential
| |
| /j/j/j j//
| |
| /,/j Revision Month Day Year I
| |
| / Number // Number NONE 01 Sl 0101 of l I
| |
| ^ ~~~
| |
| Ol 6 0 13 BlB Bl B 0 l 0 17 010 016 31 0 81 8 01 51 01 01 of I l i OPERATING
| |
| [ (Check one or more of the followinal (111 l 2 20.402(b) 20.405(c) _K. 50.73(a)(2)(iv) ___ 73.71(b) l POWER ___ 20.405(a)(1)(1) ___ 50.36(c)(1) __ 50.73(a)(2)(v) ,_._ 73.71(c)
| |
| LEVEL 20.405(a)(1)(ll) 50.36(c)(2) 50.73(a)(2)(vil) Other (Specify (101 0l0 l2 _
| |
| 20.405(a)(1)(iii)
| |
| _ 50.73(a)(2)(1)
| |
| _ 50.73(a)(2)(vill)(A) in Abstract
| |
| ___ 20.405(a)(1)(iv) ._._ 50.73(a)(2)(ll) _ 50.73(a)(2)(vill)(B) below and in l
| |
| l
| |
| _ 20.405(a)(1)(v) _ 50.73(a)(2)(lit) _ _ 50.73(a)(2)(x) Text)
| |
| LICENSEE CONTACT FOR THIS LER f12)
| |
| Name TELEPHONE NUpeER D. Brindle Operating Engineer Ext. 2218 AREA CODE Bl115 213141 l514141 C0ffLETE ONE LINE l'OR EACH COMPONENT FAIttar DESCRIBED IN THIS REPORT f131 CAUSE Fv''rM COMPONENT MANUFAC- REPORTABLE CAUSE SYSTEM COMPONENT MANUFAC- REPORTABLE l TURER TO NPRD$ TURER TO NPRDS I I I I I I I I I I I I I I l l 1 I I l I I I I I l l I I SUPPLEK NTAL REPORT EXPECTED f141 Expected Month i Day I Year Submission lyes (If ven. commlete EXPECTED StBMISSION DATEl YI NO l l I
| |
| ]
| |
| ABSTRACT (Limit to 1400 spaces, i.e. approximately fif teen single-space typewritten lines) (16)
| |
| On June 3,1986, Unit 2 was in the Startup operational mode at 2% reactor power. A Nuclear Station l Operator (N50) had established Steam Generator Preheater Bypass feedwater flow to the A, B and D Steam
| |
| ! Generators, but could not open the Preheater Bypass Valve (2FWO39C) for the C Steam Generator. Levels in ;
| |
| the A, B and D 5 team Generators increased while C Steam Generator level decreased. Valve 2FWO39C was i opened by a non-licensed operator locally, but D Steam Generator level reached the high-2 level setpoint of l
| |
| 78.1% which actuated the P-14 permissive and caused an automatic feedwater isolation. While recovering from this feedwater transient another p-14 permissive actuation occurred due to high-2 level in the A Steam Generator at 1338. Levels were restored, the Feedwater Isolation signal was reset, and the plant startup continued without further incident.
| |
| The root cause cf the event was the unexpected failure of the 2FWO39C valve to open when demanded remotely by the N50 in the Main Control Room. The valve failure initiated a feedwater flow control disturbance that resulted in the high-2 steam generator level condition. Subsequent stroking of the 2FWO39C valve remotely ;
| |
| was successful in all attempts.
| |
| N50 actions during the event were in accordance with current operating strategies for dealing with D-5 Steam Generator level control problems. The strategies have been successful during normal plant startups, but were not able to compensate for the level effects caused by the 2FWO39C valve failure. All l'rensed operators will be required to read this Licensee Event Report (LER) to reinforce their understandmg of D-5 Steam Generator level control during low power operations.
| |
| Previous similar occurrences were reported in Unit 2 LER 87-002.
| |
| 5 l l!
| |
| 1 (0037R/0002R) 4
| |
| | |
| l l
| |
| l LICENSEE EVENT REPORT (LER) TEXT CONTINUATION FACILITY NAME (I) DOCKET NUteER (2) LER NUMBER (6) Pace (3)
| |
| ,. Year /// Sequential /// Revision l- - /,//ff Number g//
| |
| / Number pg Byron. Unit 2 0 I 5101010 l dl 515 8l8 01017 0 1 0 01 2 01 3 (O) 0F TEXT Energy Industry Identification System (E!!S) codes are identified in the text as [xx)
| |
| A. PLANT CONDITIONS PRIOR TO EVENT:
| |
| Event Date/ Time 6/3/88 / 1327 Unit 2 MODE 2_._ - Startuo Rx Power 2% RCS (AB) Temperature / Pressure Normal Ooeratino B. DESCRIPTION 0F EVENT:
| |
| On June 3.1988, Unit 2 was in the Startup operational mode at 2% reactor power. There were three Nuclear Station Operators (NS0's) (licensed reactor operators) assigned to the Startup. One had assumed the reactor operations board, the second had the feedwater board, and the third was preparing for a Main Turbine (TB) roll. The NSO at the feedwater board was attempting to initiate feedwater (FW) [SJ) flow through the Steam Generator Preheater Bypass Flow valves. 2FWO39A, B, C D. At 1327 the NSO hay established flow through 2FWO39A. 8 and D and was attempting to open the final Preheater Bypass valve, 2FWO39C, when it failed to open. The feedwater board NS0's attention was diverted from the A, B and D Steam Generator levels to the opening of the 2FWO39C valve. The NSO at the Main Turbine control board ceased his activities there and inanediately assumed a position at the feedwater control panel. Levels in the A. B and D Steam Generators began to rise and C Steam Generator level lowered. Feedwater flow to the A, B and D Steam Generators was manually isolated in an attempt to control the level increases while attempts were made to establish preheater flow to the C Steam Generator as its level continued to lower.
| |
| The 2FWO39C valve was opened by a non-1.icensed operator locally. At 1327 the D Steam Generator level reached a high-2 level setpoint of 78.1% which actuated the F-14 pemissive and caused an automatic feedwater Isolation and a Main Turbine Trip. The NSO's restored all Steam Generator levels to normal and
| |
| -Q C/
| |
| reset the feedwater Isolation signal. At 1338 while attempting to stabilize Steam Generator levels the A Steam Generator level swelled causing another P-14 actuation and Feedwater Isolation. Levels were restored to normal, the feedwater Isolation signal was reset and the startup continued without further incident.
| |
| This Licensee Event Report is submitted in accordance with 10CFR50.73 (a)(2)(iv) due to the two Engineered Safety Feature (ESF) actuations.
| |
| C. CAUSE OF EVENT:
| |
| The intermediate cause of the event is attributed to the atypical response of DS Steam Generator level indications at low power levels. The failure of the 2FWO39C valve to open upset the controlled addition of feedwater through the 2FWO39A, B. D Preheater Bypass Valves which in turn initiated a swell in the Steam Generators. The shrink / swell phenomena are most pr?nounced at low power and Unit 2 was at 2 percent reactor power at the time of the event. One NSO's attention was diverted fror the remaining Steam Generator levels but the duty was immediately assumed by the turbine control operator, therefore, inattention was not a contributor to the event.
| |
| The second level excursion occurred at 1338 hours when NSO's were recovering from the first feedwater transient and were attempting to stabilize Steam Generator levels. The A Steam Generator level swelled and initiated a Feedwater Isolation. The level was restored and the unit startup continued.
| |
| (0037R/0002R)
| |
| | |
| ; LICENSEE EVENT REPORT fLER) TEXT CONTINUATION FACILITY NAME (1) DOCKET NUPSER (2) LER NUteER (6) Pace (3)
| |
| Year /// Sequential // Revision f] /j/jj/ Number /j/j f
| |
| // Numbe r._
| |
| l h Byron. Unit 2 _ 0 l 5 1 0 1 0 1 0 l 41 51 5 8I8 - 010I? - 0 10 013 0F 01 3!
| |
| TEXT- Energy Industry Identification System (EIIS) codes are identified in the text as [xx]
| |
| C. CAUSE OF EVENT: (Continnd)
| |
| The cause of the event is attributed to the unexpected failure of the 2FWO39C to open when demanded remotely by the NSO. This occurrence initiated a disturbance in the controlled feed to all four Steam Generators and resulted in Steam Generator levels shrinking oa C and swelling on A, B, D. The 2FWO39C was opened by a non-licensed operator who momentarily interrupted air flow from the air operated valve's diaphrayn. The valve worked properly during several subsequent strokes. It is speculated that a check valve in the air flow path to 2FWO39C was not properly seated. and this was the root cause of the failure of the valve to open. The non-licensed operator interrupted the air flow, thus permitting the check valve to seat and permit subsequent satisfactory operation of 2FWO39C.
| |
| D. SAFETY ANALYSIS:
| |
| The Feedwater Isolations occurred when the majority of Feedwater Isolation Valves were already closed due to the inw power operating condition. All safety systems actuated as designed. The Main Turbine had been latched in preparation to roll, therefore, there were no signifle. ant effects on the turbine. Steam Generator levels were quickly lowered from the p-14 permissive setpoint and the Feedwater Isolation signals were reset. All ESF equipment functioned as designed. At no time was plant or public safety threatened.
| |
| E. CORRECTIVE ACTIONS:
| |
| All operator actions were correct in responding to indicated process changes. The operator followed the o four main strstegies for controlling D-5 Steam Generator levels:
| |
| : 1. Give 1007 dedicated attention to S/G 1evel control during low power transients.
| |
| : 2. Prior to inducing a planned transient, all plant parameters shall be stable at their nominal valve.
| |
| : 3. Only one parameter and one Steam Generator at a time will be altered unless inaction would result in a protective feature actuation.
| |
| : 4. All changes are made in small increments with time for stabilization between steps.
| |
| The level instabilities induced by the failure of the 2FWO39C valve made control of the Steam Generator levels difficult. The strategies have been successful during normal plant startups, but were not able to compensate for the level effects caused by the 2FWO39C valve failure. All licensed reactor operators and senior reactor operators will be required to read this Licensee Event Report to reinforce their understanding of the care that must be exercised in controlling D-$ Steam Generator levels at low power.
| |
| The 2FW039C valve was successfully stroked several times following its initial local opening by the non-licensed operater. The valve remained in service and functioned normally while the power escalation continued.
| |
| F. PREVIOUS OCCURRENCES:
| |
| LER NU'9ER IllLE l B7-002 Reactor Trips and Feedwater Isolations Due to Operator Difficulty in l Controlling Steam Generator Level Transients at Low Power
| |
| .l ,) G.
| |
| COMPONENT FAILURE DATA:
| |
| i \.J at t%NUFACTURER NOMENCLATURE t!QDLL NUteER MFC- PART NUMBER Not Applicable i 1
| |
| j
| |
| | |
| ,_ _ . . . . - - - . . . . . . . - - ---.y~~-------~ - - - - - -
| |
| p:"W M - l C)
| |
| DEVIATION INVESTIGATION REPORT f~
| |
| TITLE PAGE
| |
| (% FAILURE OF FEEDWATER PREHEATER BYPASS VALVE 2FWo398 TO CLOSE rp0M MAESWITCH 1 10Fl o l 2 EVENT DATE DIR NUMER REPORT DATE OPERATING SEQUENTIAL REVISION l 19 @ DAY YEAR STA UNIT YEAR
| |
| /g/ NL8eER y/
| |
| / NUPRER MONTH DAY YEAR M00E 1
| |
| POWER of a si a LEVR of 1 01 6 of 2 al a -
| |
| o I ol 4 -
| |
| ol1 112 310 8 18 of si o 1 CONTACT FGR THIS DIR NAPE TELEPHONE NureER AREA CODE W. Waltern. Anst Tech Staf f Sumervisor Ext. 2240 81115 2131,4l-l$l4l4II COMPLETE ONE LINE FOR EACH COMPONEN Alttier DESCRIBED IN THIS REPORT CAU$E SYSTEM COMPONENT MANUFAC- REPORTA8LE CAUSE SYSTEM COMPONENT MANUFAC- REPORTA8LE TURER TO NPRDS TtierR TO NPRDS I I I I I I I I l l I I I I I I I I I I I I l l l l SUPPLEMENTAL REPORT EXPECTED MNTH DAY YEAR SUBMISSION I~1YESfifven.enmolateEXPECTEDSUBMISSIONDATE) [l NO l
| |
| +
| |
| TEXT A. PLANT COM ITIONS PRIOR TO EVENT:
| |
| 's 1
| |
| Event Date/ Time 1/4/88 / 1118 hrn Unit 2 MODE 1 Power Onorations Rx Power 801 RCS [AB) Temperature / Pressure Normal Doeratien B. DESCRIPTION OF EVENT:
| |
| On 1/4/88, at 1118. Byron Unit 2 was in Mode 1 operating at 80% power when the reedwater (FW) ISJ)
| |
| Preheater Bypass Valve 2FWO398 failed to close on demand from the Main Control Board. IPM04J. control switch. Subsequently, the air supply to the valve actuator was closed and the valve failed closed. A LCOAR. Limiting Condition For Operation Action Requirement, was entered.
| |
| C. CAUSE OF EVENT:
| |
| The cause of the failure of the 2FWO398 valve to close has been determined to be the inability of the "C" solenoid to vent the air from the valve diaphragm / accumulator. The valve opens properly and would close using the two remaining safety related solenoids.
| |
| D. SAFETY ANALYSIS:
| |
| The plant or public safety was not affected by the failure of the 2FWO398 valve to close from the Main Control Board switch signal. The valve wculd have closed on a Feedwater Isolation Signal due to the two remaining safety related solenoids. The affected solenoid does not receive a safety signal and is used only for manual opening and closing functions on the 2FWO398 valve. The last known operable time of the "C" solenoid was on 12/12/87 when 2Fwo39 was tested under 28V5 6.3.3-20.
| |
| O I
| |
| (1872M/0216M)
| |
| | |
| -- -- -~
| |
| l . _ _ _ . - - . . . . . . . . . . . -
| |
| /
| |
| l l
| |
| DEVIATION INVESTIGATION REPORT TEXT CONTINUATION
| |
| \
| |
| () ' TITLE DIR NUpmER SEQUENTIAL REVISION PAGE FAILURE OF FEEDWATER PREHEATER BYPA$$ STA UNIT YEAR NupeER NUPBER VALVE 2FWO398 TO CLOSE FROM HAND $ WITCH of 6 of 2 81 8 -
| |
| 010I4 -
| |
| 011 2 0F 0I2 TEXT E. CORRECTIVE ACTIONS:
| |
| /
| |
| The Preheater Bypass valve was verified to close by using one of the safety related solenoids. The "C" solenoid, which is non-safety related, was replaced under Nuclear Work. Request 851747, and it was verified that.the solenoid was receiving the signal from the control switch. The problem has been determined to be the failure of the air to vent from the "C" solenoid. A nuclear work request was written to investigate {
| |
| and repair the venting problem. The valve was declared operable when the valve was verifled by temporary procedure 88-2-014 to close in less than 6 seconds on 1/6/88, using one of the safety related solenoids to close the valve. On 7/22/88, the instrument air check valve was replaced. The check valve disc had become disconnected from its stem and stem nut. The valve was successfully tested and returned to operable stat,us.
| |
| F. PREVIOUS OCCURRENCES:
| |
| DVR NUPRER 11111 unng ..
| |
| G. COMPONENT FAILURE DATA:
| |
| ; a) MANUFACTURER . NOMENCLATURE M00CL NUPSER MFG PART nut 9ER Parkee Hennifin Check valve 3/4" b) RESULTS OF NPRDS SEARCH:
| |
| Not Applicable c) 19fK None i
| |
| I (1872M/0216M)
| |
| | |
| , _ . . . . . . _ - - - -- ~ ~ 4 go.$
| |
| I DEVIATION INVESTIGATION REPORT (DIR)
| |
| Form Rev 2.0 Facility Name '
| |
| RAGE Braidwood Unit 2 1 0 0 2 Failure of Feedwater Bypass Isolation Valve to Open due to Mecharical, Binding
| |
| /] Title EVENT DATE - DIR NL3.ER REPORT DATE SEQUENTIAL REVISION P.NTs oAv vEiR STA t u. vsAR {7{
| |
| 7 Nu ER {7{
| |
| 7 Nu,.ER ,.NTw oAv vEAR POWER 1 017 81 8 '21 0 01 2 81 8 -
| |
| Il 61 8 -
| |
| 01 0 Il 1 11 S Bl 8 01910 l . , , CONTACT FOR THIS DIR NAPE TELEPHONE Nup.ER
| |
| ' AREA CODE Jagity Ibrahim. Technical Staf f Enoineer ' Ext. 2402 8I115 4l5l81-l2l810l1 COMPLETE ONE LINE FOR EACH COMPONENT A URE DESCRIBED IN THIS REPORT
| |
| .CAUSE SYSTEM COMPONENT MANUFAC- REPORTABLE CAUSE SYSTEN COMPONENT MANUFAC. REPORTABLE para TO NPRDS TURER TO NPRDS
| |
| . I I I I l I I I l i I I I I l i I I I I I I I I I I SUPPLEMENTAL REPORT EXPECTED MDNTH DAv YEAR p
| |
| SUBNISSION DATE
| |
| -xl YES fif van. enamlete EXPECTED StaMISSION DATE) l No 0 6 0 1 8l9 TEXT Energy Industry Identification System (E!!S) codes are identified in the text as (XX)
| |
| ..~~.
| |
| ~
| |
| : k. PLANT C0tWITIONS PRIOR TO EVENT:
| |
| Unit: Braidwood 2, Event Date: October 7,1988 Event Time: 0607; Node: 1 - Power Generation Rx Power: 90%;
| |
| RCS(AB)Tankrature/ Pressure: N.0.T./N.0.P.
| |
| : 8. DESCRIPTION OF EVENT:
| |
| On 10/7/88 at 0607, during power ascension according to procedure 28wGP 100-3, Power Ascension 5% to 100% the Nuclear Station Operator (NSO) performed step F.59 to open the feedwater isolation valves 2FW039A, 2FWO398 2FWO39C and 2FWO390. He noticed that 2FWO39A valve failed to open. Nuclear Work Request number (NWR #)
| |
| A26043 was issued and Limiting Condition for Operating Action Requirement (LC0AR) 6.3-1A, Containment Isolation Valves Tech Spec LC0 3.6.3, was entered. There were no inoperable systems or components that contributed to the event.
| |
| C. CAUSE OF EVENT:
| |
| The cause of this event is partially due to tight packing resulting in high friction between the valve stem and the packing material. In addition it is also suspected that mechanical or thermal binding may have occurred within the valve internals.
| |
| t 2368m(111788)/13
| |
| ~ - . _ _ . .. _ .. . . , a
| |
| | |
| t DEVIATION INVESTIGATION REPORT TEXT CONTINUATION
| |
| ' Form Rev 2.0 FACILITY NAME ,~ _ DIR NUteER PAGE N ION N'\ Braidwood Unit 2 STA UNIT YEAR NUPRER Nr 3ER
| |
| ' \j,
| |
| ' - 21 0 01 2 BI 8 -
| |
| 116l8 -
| |
| Ol 0 2 0F 012 l TEXT l
| |
| D. $AFETY ANALYSIS:
| |
| l- s., The feedwater bypass line is a means to deliver feedwater to the Steam Generator ($/G) through the uprer I nozzle, instead of the main, lower, nozzle, at low loads to prevent the formation of water hammer in the
| |
| $/G preheater section. The feedwater isolation bypass valve (2FWO39A) safety function is to isolate between the $/G In the Containment Building, and the rest of the feedwater system if there is an Engineering Safeguard Feature (ESF) actuation. The valve was already closed, functioning as an isolation valve, and at no time were the public, personnel or the equipment in danger.
| |
| E. CORRECTIVE ACTIONS:
| |
| {} famadiate Actions:
| |
| NWR #A26043 was issued and LC0AR 63-1A was entered. Packing was adjusted and the stem was lubricated. The valve was operated and passed Technical Staff surveillance SwVS 6.3.3-20.
| |
| ii) Lana fare Actirna:
| |
| : 1) In response to the possible therme1/ mechanical binding problem, BwGP 100-3 will be changed to instruct the N50's, at step F.59, to open the feedwater bypass isolation at 70% power. This P will reduce the pressure differential across the valve and, consequently, will reduce mechanical binding of the valve internals.
| |
| ( .
| |
| : 2) This event may be similar to Byron's events reported in Byron's Licensee Event Reports number 88-007 and 88-009. The applicable procedures will be reviewed for possible inclusion of a step to stroke the subject feedwater bypass valves subsequent to a reactor trip. This is being done as a result of both the event at Braidwood and in response to INPO SER B-88, pressure Locking of Residual Heat Removal Gate Valves.
| |
| All procedures review will be tracked by action item 457-200-88-16801.
| |
| : 3) NWR #A26308 has been issued to inspect the valve internals. Also removing the check valves 2FWO78A, 2FWO788, 2FWO78C and 2FWO780, per modification number N20-2-68-029, will reduce the pressure dif ferential across each of the feedwater bypass isolation valves and will further reduce the internals binding. Results will be tracked by actian item 457-200-88-16802.
| |
| F. PREVIOUS OCCURRENCES:
| |
| 'here have been no previous occurrences of nechanical binding recorded in a Deviation Report.
| |
| G. COMPONENT FAILURE DATA:
| |
| None.
| |
| U 2368e(111988)/14
| |
| | |
| FWm-8 LICENSEE EVENT REPORT (LER) racility Name (1) Docket Number (2) Pane (3)
| |
| Byron. Unit 2 0151010101di515 1 ! of! 0 ! 3
| |
| - v Title (4) gg ging of Steam Generator Prehester Bypass Valvas Resulting in Low Steam Generator Level Event Date (5) LER Number (6) Renort _fhte (7) Other Facilities Involved (8)
| |
| Month Day Year Year /// Sequential /jj/
| |
| / Revision Month Day Year racility Me l Docket u. + rts, j/jj
| |
| // Number j/// Number .,
| |
| 3 N0;K 01 51 01 01 01 l i
| |
| ~~~ ~~
| |
| OI7 11 5 81 8 8l8 010l9 0l0 0l8 0I9 8l8 01 51 01 01 Ol l i OPERATING (Check one or more of the followinal (111 2 20.402(b) _ 20.405(c) .JL 50.73(a)(2)(lv) _ 73.71(b)
| |
| POWER 20.405(a)(1)(1) _ 50.36(c)(1) _ 50.73(a)(2)(v) __ 73.71(c)
| |
| LEVEL g g 20.405(a)(1)(ll) 50.36(c)(2) _ 50.73(a)(2)(vil) ___ Other (Specify (101 010 12 ._ _ 20.405(a)(1)(lit) _ 50.73(a)(2)(1) _ 50.73(a)(2)(vtli)(A) in Abstract
| |
| ////////////,///////j////j// /
| |
| f 20.405(a)(1)(iv) 50.73(a)(2)(ll) 50.73(a)(2)(vill)(8) below and in '
| |
| ///////////' ////u/)/~//}///
| |
| /
| |
| j /j 20.405(a)(1)(v) 50.73(a)(2)(lii) 50.73(a)(2)(x) Text)
| |
| LICENitr CONTACT FOR THIS LER f12)
| |
| Name TrtrPHONE NUteER AREA CODE L. Sues. Assistant Sumerintendant Technical Services Ext. 2214 8l115 213141 l5141di COMPLETE ONE LINE FOR EACH COMPONENT FAltuar DFSCRIBED IN THIS REPORT f13)
| |
| CAUSE SYSTEM COMPONENT MANUFAC- REPORTA8LE / CAUSE SYSTEM COMPONENT MANUFAC. REPORTABLE f
| |
| TURER TO NPRDS / TURER TO NPRDS X S I .1 l l lV A 13 19 11 Y / l l l l l l l f I l i I l l f I l l 1 l l 1 SUPPLEE NTAL REPORT EXPCCTED (14) Expected Month l Day I Year k '
| |
| Subaission lyes (If ven. cannlete EXPECTED StBMISSION DATE) X l ND " l l l A8STRACT (Limit to 1400 spaces, i.e, approximately fifteen single-space typewritten lines) (16) i On July 15, 1988, Byron Unit 2 reactor power was 2 percent. At 0436 a Nuclear Station Operator (NS0) attempted to open the Steam Generator Preheater Bypass Valves (2FWO39A,8,C,0) to feed the Steam Generators (S/G's). The A and 0 valves opened properly, but the 8 and C valves failed to open as demanded. The levels in the 28 and 2C S/G's lowered to the low-low level reactor trip setpoint at which point the reactor automatically tripped. The licensed operators entered and compiled with emergency operating procedures.
| |
| Stable plant conditions were achieved in Hot Standby at 0500.
| |
| The event was caused by the failures of valves 2rWO398 and 2rWO39C to open on demand. These valves need to be open to provide sufficient feedwater flow to the S/G's at I to 2 percent reactor power. The level instabilities induced by the valve failures made level control difficult. Investigation revealed that the valves had become thermally bound following their automatic closure during a reactor trip event on July 14 1988.
| |
| 1 Both valves were opened using hydraulic lif ts and 2rWO39C operated properly, however, 2rWO398 still would {
| |
| not open when demanded by the handswitch. A non. safety related air check valve was replaced and the 2rWO398 valve operated, properly. 1 A previous similar occurrence was reported in Unit 2 Licensee Event Report 88-007.
| |
| d I 1 (0075R/0008R)
| |
| | |
| LICENSEF EVENT REPORT (LER) TEXT CONTINLIATION FACILITY NAME (1) DOCKET NUteER (2) LER NLDeER (6) Paas (3)
| |
| Sequential
| |
| []
| |
| Year Numeer y//
| |
| /
| |
| Revision Number Byren. Unit 2 0 l 5 1 0 1 0 1 0 1 41 El 5 al8 - 0I019 .-_ 0l0 012 or el 3 TEXT Energy Industry Identification System (EIIS) codes are identified in the text as [xx)
| |
| A. PLANT E0mITIONS PRIOR TO EVENT:
| |
| Event Date/ Time 7/15/88 / 0436 Unit 2 MODE .l - _Startus Rx Power 2% . RCS (A8] Temperature / Pressure Na mm1 Onarating_
| |
| : 8. DESCRIPTION OF EVENT:
| |
| 1 There were no systems or components inoperable at the beginning of this event that contributed to the event. On July 15, 1988, Byron Unit 2 was in the Startup Operational Mode (Mode 2) with reactor power at 2 percent. At 0436 a Nuclear Station Operator (N50) (licensed reactor operator) attempted to open the Steam Generator Preheater Oypass Valves (5J) (2rWO39A,8,C,0)'to feed the $tsam Generators (5/G's). The 2rWO39A and 2FWO390 valves opened properly, but the 2rWO398 and 2FWO39C valves failed to open as demanded. The N50 ;
| |
| isolated the blowdown flow paths from the 28 and 2C 5/G's in an effort to slow the rate of level decrease I in the 5/G's. The levels in the 28 and 2C 5/G's continued to lower. As the low-low level reactor trip j setpoint (171) was approached and it was evident that levels could not be restored, a licensed Senior ]
| |
| Reactor Operator directed the N50 to manually trip the reactor, however the Reactor Protection System initiated an automatic reactor trip due to low-low level in the 2C 5/G before the manual trip was accompilshed. The licensed operators entered and complied with the " Reactor Trip or Safety Injection-Unit 2 Emergency Operating Procedure" (28EP-0) and the " Reactor Trip Response Unit 2 Emergency Operating' l Procedure" (28EP ES-0.1). The 2A and 28 Auxiliary Feedwater Pumps (AFP) (BA) automatically started due to p the low-low 5/G 1evel condition as expected. An expected Feedwater Isolation occurred due to the opening l
| |
| 1 of the reactor trip breakers coincident with low average reactor coolant temperature (T,yg) of 564*F.
| |
| At 0446 the 28 AFP was stopped since its operation was not required to maintain 5/G 1evels. At 0451 tre
| |
| ~
| |
| Feedwater Isolation signal was reset. At 0454 the Startup feedwater Pump was started and flow was estabitshed to the 5/G's. At 0455 the ZA AFP was stopped. At approximately 0500 the stable plant )
| |
| conditions were achieved in Hot Standby (Mode 3). Valves 2rWO398 and 2rWO39C were declared inoperable and l
| |
| Technical Specification Limiting Condition for Operati . AstuJ. T.@Rw.d.; (LCSAR) 3.6.3 for the two '
| |
| containment isolation valves was entered and satisfied.
| |
| This Licensee Event Report (LER) is submitted in accordance with 10CFR50.73 (a)(2)(tv) due to the automatic Reactor Protection System and Engineered Safety Features (E5F) actuations. i C. CAUSE OF EVENT:
| |
| The event was caused by the failures of valves 2fWO398 and 2rWO39C to open on demand. At I to 2 percent reactor power the Preheater Sypass Valves must be open to provide sufficient feedwater flow to maintain 5/G 1evels. The Syron Unit 2 5/G's are Westinghouse Model D-5. The shrink / swell phenomena are most pronounced at low power, and Unit 2 was at 2 percent reactor power at the time of the event. The level instabilities !
| |
| induced by the failure of the 2rWO398 and 2FWO39C valves made control of the S/G 1evels difficult. The license $ operators' actions were in accordance with Station Operating Procedures and operating strategies for D-b 5/G 1evel control.
| |
| The 2FWO398 and 2rWO30c volves were raind to be theme 11y bound. The valves had automatically closed w+.en Unit 2 tripped on July 14, 1988 (see tndt 2 LER 83-006). Normally, the valves are manually closed during a controlled shutdown of the plant. A controlled shutdown does not close the Preheater Bypass Valves at such '
| |
| l q high feedwater temperatures.
| |
| l k.j
| |
| ~ I' 1
| |
| (0075R/0008R) j l
| |
| l
| |
| .. - - _ . .- . , ~ . . .. .- -, -- - -
| |
| b
| |
| | |
| LICEN$EE EVENT REPORT (LERI TEXT CONTINuATIDW FACILITY NAPE (1) DOCKET NUPSER (2) LER Nupers (6) Pane (3)
| |
| .(^ Year
| |
| //l/
| |
| lll Sequential l/l/l/
| |
| Revision
| |
| \ jj ua - - jjj y. s _. 7 8vran. Unit 2 0 I B l 0 l 0 l 0 1 4! El 5 8l8 - 01019 - 0l 0 01 3 0F 0l3-TEXT Energy Industry Identifiestion System (E!!$) codes are identified in the text as [xx) i D. $AFETY ANALYSIS:
| |
| The event occurred when the majority of Feedwater Isolation Valves were already closed. All E5F systems actuated and functioned as designed. The 5/G 1evels were reestablished and the Unit stabilised operations in Mode 3. The Failed Preheater Bypass Valves. failed in their safe position. Neither plant nor pubite safety were affected by this event.
| |
| E. CORRECTIVE ACTIONS:
| |
| The 2FWC398 valve was uncoupled from its actuator. The actuator was found to be operating correctly. The
| |
| ' 2FWO398 and 2FWO39C were then opened using hydraulic lif ts. The 2FWO39C stroked properly. The 2FWO398 did not operate properly from the control switch. The problem was isolated to a non-safety related air check valve for the *C" solenoid. The air check valve was replaced and the 2FWO398 stroked properly when demanded from the handswitch. The valves were successfully tested and returned to operable status.
| |
| Technical Specification LC0AR 3.6.3 for the fatted valves was exited at 1836 on July 15, 1988. In addition, applicable operating procedures will be reviewed for the possible inclusion of a step to stroke the Preheater Bypass Valvec subsequent to a reactor trip. This may prevent thenns) binding cf the type that was experienced durlog this event. This proposed corrective action is tracked by Action Item Record 454-225-88-0158.
| |
| The Preheater Bypass Valves have not thermally bound following a normal unit shutdown. No further
| |
| /% corrective action is planned at this time.
| |
| F. PREVIOUS OCCURRENCES:
| |
| LER NupBER IIIlf, 86-007 feedwater Isolation Actuation due to 5/G Preheater Bypass Valve ret 1ure to Open G. COMPOND(T FAILURE DATA:
| |
| a) MANUFACTURER N0fENCLATURE PEDEL laster MFG PART NUPSER 5tratafio Products 1/2-inch NPT Check Valve 9
| |
| O -
| |
| 1 4 (0075R/0008R) i
| |
| | |
| ......._...._.._i1.~'.._...___
| |
| BRAIWOOD SIMULATOR MALFUNCTION l
| |
| | |
| ==Title:==
| |
| Auxiliary Feedwater Valve Malfunction ID: WM-20 NO: 6.3.4.7.20
| |
| | |
| == Description:==
| |
| The~ flow control valve fails to control properly.
| |
| Varittions: None: Date: 3/17/89 Rev: 6 s
| |
| i Selectable I Steps Inputs Comments
| |
| : 1. Select position 0-100 percent 0 = closed 100 = open
| |
| : 2. Select valve 1, 2, 3, 4, 1 - 1AF005A 5 - 1AF005E 5, 6, 7, or 8 2 - 1AF005B 6 - 1AF005F 3 - 1AF005C 7 - 1AF005G 4 - 1AF005D 8 - 1AF005H Brief Plant Response:
| |
| Improper flow of aux. feedwater to S/G. First annunciators include AW FUMP DISCHARGE FLOW HIGH.
| |
| Suggested Instructor Action:
| |
| None.
| |
| Events:
| |
| : 1) DVR 20-02-88-103: 2AF005G Failed Open.
| |
| : 2) DVR 06-02-88-102: 2AF005B Tailed Closed.
| |
| 9 0110w:4 882M/32 5/89
| |
| __,-m___m___==__.. _ . . _ . .
| |
| | |
| l Fw M-L Q i DEVIATION INVESTIGATION REPORT 1
| |
| TITLE Failure of Auxiliary feedwater Throttle Valve Due to Mispositioning of Handwheel by Person PAG Unknown F 2
| |
| ! EVENT DaT" DIR NUMEER pEPORT DATE /
| |
| "O
| |
| // SEQUENTIAL // REVISION f HDNTH DAY YEAR STA UNIT YEAR NUMBER /f NUPEER MONTH DAY YEAR 1 POWER 01 6 21 0 81 8 21 0 01 2 81 8 -
| |
| 1 l 01 3 -
| |
| 010 01 8 01 1 81 8 01 41 0 CONTACT FOR THIS DIR NAME TELEPHONE NUMBER AREA CODE Cheryl A. Melone. Technical Staff Enoineer Ext. 2400 811 l5 4l5181 I2iBl0l1 COMPLETE ONE LINE FOR EACH COMPONEN LURE DESCRIBED IN THIS REPORT CAUSE SYSTEM COMPONENT MANUFAC- REPORTABLE / CAUSE SYSTEM COMPONENT MANUFAC- REPORTABLE TURER TO NPRDS TURER TO NPWDS I I I I I I I l i I I I I I l l l l I I I I I I l i SUPPLEMENTAL REPORT EXPECTED MONTH DAf YEAR SUBMISSION
| |
| ~l YES (if vet. complete EXPECTED SUBMISSION DATEl l NO l TEXT A. PLANT CONDITIONS PRIOR TO EVENT:
| |
| Unit: Braidwoon 2 ; Event Date: June 20. 1988 : Event Time: 1403 MODE: _1_ - Power Operations : Rx Power: 49% ; RCS [AB) Temperature / Pressure:_$55 Deorees F/2235 esto V)
| |
| (
| |
| B. DESCRIPTION Or EVENT:
| |
| During an Auxiliary Feedwater ( AF) (BA] actuation due to Lo-Lo Steam Generator levels, the Nuclear Station Operstor (N50) reduced flows to the steam generators to maintain proper levels. The 2C steam generator level continued to increase af ter the potentiometer for valve 2AF005G AF flow control valve, was set to its minimum setting. With the potentiometer set to its minimum setting, the flow to the generator was still greater than 200 gallons per minute (gpm) when the flow required to cool the steam generators is 160 gpe. The NSO maintained proper steam generator level by throttling the AF steam generator isolation valve. 2AF013G. Plant stability, given the in progress recovery from Le-Lo 2C Steam Generator level condition, die not degrade as a result of this event.
| |
| C. CAUSE OF EVENT:
| |
| Upon investigation it was revealed that the valve handwheel for throttle valve 2AF005G was not in the neutral position. Personnel on shif t were unaware of any prior repositioning of the handwheel. A review of shift logs and of the Unit 2 Component Abnormal Position Log yielded no documentation with regard to its repositioning.
| |
| Thus the root cause of this event is valve mispositioning by person or persons unknown. This did not allow positioning of the throttle valve from the main control board. The purpose of this handwheel is to allow throttling of the flow control valves, on B train AF only, without the use of air or electricity.
| |
| I
| |
| (
| |
| 2211m(0801BB)/24 1
| |
| h
| |
| . _ _ _ _ _ _ _ _ l
| |
| | |
| DEVIATION INVESTIGATION REPORT TEXT CONTINUATION TITLE . Failure of Auxiliary Feedwater Throttle Valve DIR NUMBER PAGE
| |
| . Due to Mispositioning of Handwheel By Person SEQUENTIAL REVISION Unknown STA UNIT YEAR NUMBER NUMBER f
| |
| '# 21 0 01 2 Bl B M110l3 -
| |
| Ol 0 2 0F 0l2 l
| |
| TEXT l D. SAFETY ANALYSIS:
| |
| Since the AF Steam Generator Isolation Valve 2AF013G was throttled to maintain steam generator level, no safety concerns were raised.
| |
| In a worst case scenario with the NSO unable to throttle the isolation valve with an increasing level in the affected steam generator due to excessive AF flow, two actions could restore steam generator level to normal:
| |
| : 1. Trip the affected train pump, 5-train in this case, since both trains of AF were running.
| |
| : 2. Manually throttle the B train AF flow control valve by use of the manual handwheel.
| |
| E. CORRECTIVE ACTIONS:
| |
| The NS0 took prompt action to maintain steam generator level by throttling the AF steam generator isolation valve, ZAF013G.
| |
| Procedure IBw05 7.1.2.1.a-2 will be revise,d to verify that the valves are in '' Neutral" as well as open. This will be tracked to completion by action item 456-200-88-10301.
| |
| F. PREVIOUS OCCURRENCES:
| |
| There have been several occurrences of mispositioning events by person or persons unkonwn.
| |
| G. COMPONENT FAILURE DATA:
| |
| NONE f
| |
| \
| |
| 2211m(080188)/25 l
| |
| | |
| FW ~m DEVIATION INVESTIGATION REPORT (DIR) g 7
| |
| Facility Name PAGE A Byron Nuclear Power Station 1 10 l C _"
| |
| ! ) Title POWER SUPPLY SURGE IN CABINET 2PA333 EVENT DATE DIR NUteER REPORT DATE ,
| |
| SEQUENTIAL REVISION ,
| |
| g ,
| |
| MONTH DAY YEAR STA UNIT YEAR NUteER NUteER MONTH DAY iTAR POWER sf 8 LEVEL 01 9 21 0 01 6 OI 2 el 8 -
| |
| 1 1 01 2 -
| |
| 010 1 11 OI3 81 8 of 61 51 CONTACT FOR THIS DIR NAPE TEtEPHONE NUMBER AREA CODE Tim Tulon. Asst. Superintendent Doeratina Ext. 2213 8l1j5 21314I-l5I4l4l COMetETE ONE LINE FOR EACH COMPONENT FA LURE DESCRIBED IN THIS REPORT CAUSE SYSTEM COMPONENT MANUFAC- REPORTABLE CAUSE SYSTEM COMPONENT MANUFAC- REPORTABLE TURER TO NPRDS _ TURER TO NPoDS X BlA l Cl Nl V Wl11210 Y l 1 l l l l l 1 I I I I I I I I I I I SUPPLEMENTAL REPORT EXPECTED PONTH DAY YEAf p
| |
| SUBMISSION I YES fif ves. comelete EXPECTED SUBMISSION DA*E) XI NJ TEXT Energy Industry Identification System (E!IS) codes are identified in the text as [XX)
| |
| I n A. PLANT CONDITIONS PRIOR TO EVENT:
| |
| \ }\
| |
| Event Date/Timo 09/20/88 / 0346 .
| |
| Unit 2 MDOE 1 - Power Goeration Rx Power _f1}_,_ RCS [AB] Temperature / Pressure Normal Ooeration B. DESCRIPTION OF EVENT:
| |
| At 0346 hrs. on 09/20/B8, a " CONT CAB PWR TROUBLE" alarm was received by the Unit 2 Operator for Panel 2PA33J, Train "A", failure. The flow setpoint signal for flow control to Steam Generator "B", from the "A" Train of the Auxiliary Feedwater (AF) [BA] System failed low due to loss of powor to signal converter 2FY-AF033C. The Train "A", SG "B", flow Control Valve, 2AF0053, logic interpreted the zero setpoint at a requirement for zero flow and subsequently closed the valve. LC0AR 2B05 7.1.2-la was entered based on 1 inoperability of the 2AF0058 valve affecting operability of the "A" Train of the AF system. Concurrently, flow indication at 2FI-C5015 (Not Required by Technical Specifications), CS Eductor 2A flow, was lost due ,
| |
| to failure of the Loop Power Supply, 2FY-CS015A. This card regulates the 26VDC cabinet power for use by I its loop circuitry. l Nuclear Work Requests B59824 and 859825 were written to troubleshoot and repair the 2AF0050 control circuit and loop 2C5015. The 2FY-AF033C card was replaced, the ZAF013 loop was calibrated satisfactorily, and the 2AF00$B valve declared operable. The LC0AR for train "A" of the AF System was exited at 0710 hrs. on 09/21/88. A fuse was replaced on the 2FY-C5015A power supply and indication restored. There were no known components inoperable prior to the occurrence of this event which contributed to the event. All operator actions were correct.
| |
| l ~.a 1
| |
| (0130R/0015R)
| |
| __ _ . . _ _ _ . _ ._._____._________.__J
| |
| | |
| DEVIATION INVESTIGATION REPORT TEXT CONTINUATION Form Rev 2.0 FACILITY NME DIR NUPSER PAGE O SEQUENTIAL REVISION (j STA LBiff YEAR Nupstre NupeER Buren Nuclear Power Station 01 6 01 2 81 8 -
| |
| 1l0l2 -
| |
| 0 10 2 0F 0l2 TEXT Energy Industry Identification System (EII5) codes are identified in the text as [XX)
| |
| C. CAUSE OF EVENT:
| |
| The cause of the event was surge in the 26VDC power supply to the card files in panel 2PA33J. The root cause for the surge in the power supply is indeterminate. The power surge at panel 2PA33J caused the fuse on the 2FY-C$015A to blow and damaged the 2FY-AF033C 51gnal Card. The power supply responsible for -
| |
| supplying the two failed cards is also responsible for supplying 26VDC power, where required, to all of the card racks in 2PA33J. No aiditional failures were seen in other cards in these racks due to this surge.
| |
| Maintenance history indicates that this is not a roccuring problem and therefore should be considered an isolated incident. No further corrective action is planned.
| |
| D. SAFETY ANALYSIS:
| |
| The "B" train of Auxiliary Feedwater,wes available for the duration of the event per the requirements of LC0AR 200$ 7.1.2-la. The "B" Train of Auxiliary feedwater is capable of supplying the flow and head required in the basis for Technical Specification 3/4.7.1.2 for the duration of the event, valve 2AF0058 was capable of being manually manipulated using its pneumatic control circuit at the Remote Shutdown Panel to override the erroneous demand signal .from the flow control loop.
| |
| E. CORRECTIVE ACTIONS:
| |
| .s The damaged signal converter card, 2FY-AF033C, wa:; replaced and calibrated per NWR 859825. Replacement of the fuse of power supply 2FY-C5015A per Nuclear Work Request 859824 resulted in restoration of loop 2C5015 operability. No further corrective action is deemed necessary at this time.
| |
| F. PREVI0lis OCCURRENCES:
| |
| No past occurrences of this or similar events is documented in the Nuclear Work Request history file for i/2PA33J or any of the 2AF-013 loop components.
| |
| G. COMPONENT FAILURE DATA:
| |
| a) MANUFACTURER NDPENCLATURE PODEL NupeER MFG PART NUPEER Westinghouse Signal Converter NSC Card 2837A10608 Card b) ersulTS OF NPRDS SEARCH!
| |
| Not Applicable l
| |
| c) RESut.TS OF NWR SEARCH:
| |
| f See "F" above.
| |
| (0130R/0015R)
| |
| | |
| BRAIDWOOD SIMULATOR MALFUNCTION 1
| |
| i
| |
| | |
| ==Title:==
| |
| -AUX Yeedwater Line Rupture ID: IMM-21 rG i
| |
| .\N) - NO: 6.3.4.7.21
| |
| | |
| == Description:==
| |
| A_ total or partial loss of aux.
| |
| feedwater to the S/Gs.
| |
| Variations: None. Date 8/20/86 Rev: 4 l
| |
| Selectable Steps Inputs Comments
| |
| : 1. Select pump and line FWM-21A -
| |
| FWM-21E (See response below)
| |
| : 2. Select leak rate 0-100 percent 100 percent of norinal flow
| |
| : 3. Select ramp time 0-99,999 sec.
| |
| O .
| |
| Brief Plant Responses A and B - Loss of partial AFW flow, no problem if other train works properly.
| |
| C and D - Loss of AFW flow to designated S/G, minimum flow to other S/G.
| |
| E - Total loss of AIV flow to "B" S/G.
| |
| O 0110w:4 882M/33 5/89
| |
| _:==___:-=_~--=_r-. - m -- - - ~ - - - - - - - - - - - -- - --
| |
| | |
| p, _ .-- - - . - _ - - - - - . _ _ - - - - - - - - _
| |
| L BRAIDWOOD SIffJLATOR MALFUNCTION-I'
| |
| | |
| ==Title:==
| |
| AUX Feedwater Line Rupture' ID: WM-21 l %. '
| |
| i Brief Plant Response (c.r tinued):
| |
| l Comments for Step it A - Discharge of "A" AFW pump (pump flange)
| |
| B - Discharge of "B" AFW pump (pump flange)
| |
| ,C - After flow orifice to C S/G from "A" pump D - After flow orifice to D S/G from "B" pr p E - Downstream of check valve AF 014F Suggested Instruction Action:
| |
| None..
| |
| Events: None O
| |
| O 0110w:4 882M/34 5/89'
| |
| | |
| f-l.
| |
| BRAIDWOOD SIMULATOR MALFUNCTION l
| |
| i
| |
| | |
| ==Title:==
| |
| Start-Up Feed Pump Trip ID: FWM-24 p)
| |
| 'N
| |
| . NO: 6.3.4.7.24
| |
| | |
| == Description:==
| |
| The' start-up feedwater pump fails to l
| |
| start or trips due to faulty breaker.
| |
| Variations: None. Date: 8/20/86
| |
| ' Rev: 3 1
| |
| i Selectable Steps Inputs Conunents i
| |
| : 1. Select delay time 0-99,999 sec.
| |
| Brief Plant Respense Causes a Ex trip on lo-lo steam generator levels if unable to start or put the
| |
| \ motor driven feed pump on line.
| |
| Suggested Instructor Action:
| |
| Kone.
| |
| Events 4 None 0110w:4 882M/38 5/89
| |
| _ _ _ _ _ _ _ _ _ _ - _ :_ _ _ _ - - - - - . _ - - - - ~ ~ ~ - - - - - - - - - - - - - - - - - - - - - - -~ --
| |
| | |
| - ... - - . . . - . . - - . . . - - - .. . - . . . . ~ . . :_.
| |
| BRAIDWOOD SIMULATOR MALFUNCTION
| |
| | |
| ==Title:==
| |
| Steam Generator Tempering Line Break ID: FWM-26
| |
| (
| |
| k NO: 6.3.4.7.26
| |
| | |
| == Description:==
| |
| Break in the feedwater tempering line at steam generator nozzle. !
| |
| ' Variations: None. Date: 8/20/86 Rev: 3 Selectable Steps Inputs Comments
| |
| : 1. Select loop FWM-26A - FWM-26A = Loop 1 FWM '5D FWM-26B = Loop 2 FWM-26C = Loop 3 FWM-26D = Loop 4 ,
| |
| : 2. Select leak rate 0-2500 gym Based on NOP
| |
| : 3. Select ramp time 0-99,999 sec.
| |
| Brief Plant Response:
| |
| None.
| |
| Suggested Instruction Action:
| |
| None.
| |
| 1-Events: None t
| |
| 0110w:4 882M/41 5/89 c- _ ~=----_x-__=____ _- . - - - - = = _ __----:=-_=_ . _ _ _ _ ._ _ _ _ _ _ _ - - _ - _ - _ - _ _ _ _ _ _ ._ . - . _ _ - _ _
| |
| | |
| -+-
| |
| BRAIDWOOD' SIMULATOR MALFUNCTION f ID: FWM-28
| |
| | |
| ==Title:==
| |
| Leak in CST NO: 6.3.4.7.28
| |
| | |
| == Description:==
| |
| Leak occurs in CST at desired rate.
| |
| Variations: None. Date: 6/27/86 Rev: O Selectable Steps Inputs Comments
| |
| : 1. Select leak rate 0-50,000 gym
| |
| : 2. Select ramp time 0-99,999 sec.
| |
| O Brief Plant Response:
| |
| CST level'1owers at desired rate until empty.
| |
| Suggested Instruction Action:
| |
| Malfunction may be cleared at any point at the instructor's discretion to simulate a leak at a given height on the CST.
| |
| Events:
| |
| None.
| |
| O 5/89 0110w:4 882M/43
| |
| % . m
| |
| =.
| |
| e r-e * * * * . * * * - - see. =me =e. am .n ,=
| |
| ?.?_O ??. "
| |
| | |
| BRAIDWOOD SIMULATOR MALFUNCTION LISTING l
| |
| MAIN STEAM SYSTEM l . (,
| |
| l% .
| |
| l
| |
| . MSS-1 Steam Pressure Detector Failure MSS-2 ,
| |
| Steamline Break Inside Containment MSS-3 Steamline Break Outside Containment MSS-5 Main Steam Isolation Valve (s) Failure
| |
| ' MSS-6 Steam Generator Atmospheric Relief Valve Failure MSS-7 Steam Dump Cooldown Valves Control Failure MSS-8 Steam Dump control Failure MSS-9 Stuck Steam Dump Valve MSS-10 Main Steam Header Steam Leak MSS-ll Steam Flow Detector Failure MSS-13 Steam Generator Safety Valve Failure Steam Line Pressure Detector (PT-507) Failure
| |
| -[ MSS-14 MSS MSR Eelief Valve Failure MSS-16 Turbine Driv 4n MFP Control Valve Failure MSS-17 MSIV Bypass Yalve Failure MSS-18 Steam Generator Atmospheric Relief Valve Failure i
| |
| i O
| |
| 638M/263M/8 8/87
| |
| --_.__---_-__-____-___--__-_______-________-______-_________.____________________________O
| |
| | |
| BRAIDWOOD SIMULATOR MALFUNCTION
| |
| | |
| ==Title:==
| |
| Steam Pressure Detector Failure ID: MSS-1 i NO: 6.3.4.8.1
| |
| | |
| == Description:==
| |
| 3pecified steam generator pressure detector fails to selected value due.to mechanical fallure. 3 Variations: None. Date: 7/5/87 l
| |
| Rev: 4 l
| |
| )
| |
| Selectable Steps Inputs Comments
| |
| : 1. Select steam pressure MSS-1A - MSS-1H MSS-1A = 514 transmitter MSS-1B = 515 i MSS-1C = 524 MSS-1D = 525 MSS-1E = 534
| |
| ,O MSS-1F = 535 NY MSS-1G = 544 MSS-1H = 545
| |
| : 2. Select fail value 0-1300 psig
| |
| : 3. Select ramp time 0-99,999 see Brief Plant Response: (IC-17, 100%, all systems in automatic)
| |
| Selected steam pressure channel indication goes in direction of failure. If the steam pressure channel is used for density compensation of a steam flow channel, then that steam flow indication will follow the failed steam pressure channel's response.
| |
| f~ .
| |
| 0125w:4 311M/85M/2 1/87 i
| |
| | |
| BRAIDWOOD SIMULATOR MALFUNCTION
| |
| | |
| ==Title:==
| |
| Steam Pressure Detector Failure ID: MSS-1 m
| |
| Brief Plant Response: (continued)
| |
| If the affected steam flow channel is selected for stemn generator level control, a high failure will cause an increase in steam flow indication, causing feed control to go wide open. The resulting level increase will trip the turbine. The first annunciator received is S/G FLOW NISMATCH FW FLOW LOW. If the failure is low, it will cause a decrease in steam flow, causing a decrease in feed. The resulting level drop will cause a reactor trip on low-low S/G 1evel. The first annunciators received include S/G LOW PRESS STM LINE ISOL ALERT and MS PRESS RATE STM LINE ISOL ALERT.
| |
| Suggested Instructor Action:
| |
| Place the failed channel's protective bistable in the tripped mode when
| |
| - requested by the student.
| |
| -J
| |
| [
| |
| Event: ;
| |
| : 1. DVR 6-1-85-159: Water Hammer Prevention System logic made up
| |
| ;\
| |
| 1 l
| |
| I
| |
| )
| |
| 0125w:4 311M/85M/3 1/87 I
| |
| j
| |
| | |
| E7"A**:N :%E:~ ~ ~ A!!:N *E*; *
| |
| 's ;
| |
| '!"L I .
| |
| I : .
| |
| .Atrs -ta=rs ::: 74**:N i
| |
| ''t*?w it i /1***N v CI2 twEtt ! JE800* SA*E f 32*t i
| |
| * I'' i i JLt/! REV*i*CN t ,
| |
| 'wesal://j$EOUENT ; "CCE
| |
| .q-. a, <t :p s tNiv /'t ' sumo!* eb Newsts i! wc4'ws ?av ef::
| |
| * 1 ll
| |
| ; i
| |
| * } j l80bER t: ?! t i ;- p 3!:
| |
| i
| |
| ' - it ;:-f 1 i ;t o -
| |
| 3 i o ,0!6!!i9l3 ii LEVEL l99-t. :
| |
| e' 50N*:tf TOR *His O!a
| |
| - 9 A84
| |
| *!'EswcNE '.tusta l A AE A r,;0E - i
| |
| ~
| |
| t<t. Tits eI'!1 21 : fsI I' ? Isf af ,
| |
| etea m u. 41 * ?
| |
| * 1m
| |
| *?*Ptt*r !Nr _:st rce Escw ccMPONE Luer 3rtetterd !g *wIs ersOne-MANUFAC- 2EPORTAgLE CAUSE SYSTEM COMPONE!(I MANUFAC- AEPORTABLE USE Sf 57D8l C".MPONEhT I fuefe 70 MPeDE /p/ tuete +3 uppes i
| |
| t J ' e I F f 'JI I '/ f / I/ f / V l I 1 [ l ] l t
| |
| ' I I I I I ! . .
| |
| I t l l' { l tuPetEwfNTAL Rf*0RT EXPEETED MONTW SAV lvE1R SUBMI55!Osl XI NG l l l Ytt f99 <es cavraiete EXPECTED tueMf1110W OAft1 EXT leaf waserqE0? ,
| |
| -'e placing one of the three 1A Steam Generator Loops (Loop 516) in test *te a Byron Instrument Survet11ance on
| |
| ' Fee dater Water Haarner Prevention Systam, the full power feedwater line .D transferred to the low power
| |
| ( Itne-up, necessitating an inanediate drop 'n power.
| |
| WH AT ' 11 twE eccf S AUSES A fuse in anothsr loop feeding the 2 out of 3 logic had bloE thereby maktt'g ap the logic when the 516 loop mas placed in test.
| |
| HOW Of3 IT AFFEc7 *LANT AN0/Ce PUE g SAFETY?
| |
| l This event did. cot re:J1t in a reactor trip. nor did It cause any reactor pr:*.ection or ESF Systems to cecome 1 intperacle. Thos. there was no impact on plant or public safety.
| |
| NAS IT WAPPENED'EEFORf?
| |
| Th3 full power FW line-up has switched tn the low power FW lineup once bef:*4 Out to a blown fuse (DVR 6-1-85-117). Mcwever, in that case, the fuse elew when an IM shortet it out . nile werktng in the cantnet.
| |
| l f
| |
| WHAT WA$ 004f 70 CSeatti twE CONDI-*0N AND HOW Att WE CofqG 70 pee'/ TNT eEC'_ fr g Th3 fuse was replaced and the system restored to proper operating Condtt1Crt To prevent recurrence. 8 C"a9ge in the surveillance procedure was init'ated :.c verify the operactitty of all *:i': inputs before pla:teg :re '' test.
| |
| , i 1
| |
| .0638M)
| |
| | |
| BRAIDWOOD SIMULATOR MALFUNCTION 1
| |
| )
| |
| Titie: !Steamline Break Inside Containment ID: MSS-2
| |
| , NO: 6.3.4.8.2
| |
| [
| |
| | |
| == Description:==
| |
| Non-isolable steamline break (leak) inside containment, sized up to the design basis break on~ selected .
| |
| steam generator.
| |
| l 1
| |
| ; Variations: (Break is downstream of flow restrictor) Date:- 7/5/87 Rev: 4 i
| |
| i Selectable Steps Inputs Cossents
| |
| : 1. Select steam generator. MSS-2A MSS-2A'= Loop A with faulty steam line MSS-2B MSS-2B = Loop B MSS-2C MSS-2C = Loop C -
| |
| MSS-2D MSS-2D = Loop D
| |
| : 2. Select leak rate 0-12 x 10 Leak rate based on no-load Ib/he steam pressure Normal full-load steam flow approx. 15 x 10 lb/hr (4 S/G's)
| |
| : 3. Select ramp time 0-99,999 sec l
| |
| O 0125w:4 311M/85M/4 1/87
| |
| | |
| f 1 BRAIDWOOD SIMULATOR MALFUNCTION i
| |
| j i
| |
| \
| |
| l
| |
| | |
| ==Title:==
| |
| Steamline Break Inside Containment ID: MSS-2 l s I( I
| |
| ! %/
| |
| Brief Plant Response: (IC-17, 100%, all systems in automatic]
| |
| When the steamline break is initiated, steam flow on all steam generators increases. RCS temperature drops, pressurizer pressure and level decrease rapidly. Containment temperature, pressure and humidity increase. Steam pressure drops to < 640 psis, causing steamline isolation and safety injection. When the MSIVs close, steam flow from the non-affected steam generators drops to zero, with the affected steam generator's flow remaining high. The reactor trips, safety injection is initiated, and the containment spray system is actuate 6. Steam flow from the affected generator continues until it boils dry. The first annunciators received include LOOP OT DEV LOW, S/G LEVEL DEVIATION HIGH/ LOW and NP3H LOW on a 2E6, O see ramp, sized leak.
| |
| O V Suggested Instructor Action:
| |
| None.
| |
| Events: None i
| |
| i l
| |
| i l
| |
| l i
| |
| I I
| |
| 0125w:4 311M/85M/S 1/87
| |
| | |
| BRAIDWOOD SIMULATOR MALFUNCTION l
| |
| | |
| ==Title:==
| |
| Steamline Break Outside Containment ID: MSS-3 I 'g
| |
| )
| |
| | |
| == Description:==
| |
| - Non-isolable steamline break (leak)
| |
| No: 6.3.4.8.3 outside containment up to the design basis bre'ak on the'specified loop.
| |
| Variations: (Break is at penetration) Date: 7/11/87 Rev: 4 Selectable Steps Inputs Comments f l
| |
| : 1. Select loop MSS-3A MSS-3A = Loop A j MSS-38 MSS-3B = Loop B MSS-3C MSS-3C = Loop C MSS-3D MSS-3D = Loop D 2'. Select leak rate 0-12 x 10 Leak rate based on no-load Ib/he steam pressure
| |
| )
| |
| Normal full-load steam flow approx. 15 x 10 lb/hr
| |
| : 3. Select ramp time 0-99,999 see
| |
| )
| |
| )
| |
| ?
| |
| 0125w:4 311M/85M/6 1/87
| |
| )
| |
| | |
| . ~ ~ - - - - _ _ _ _ _
| |
| 1-l
| |
| | |
| ==Title:==
| |
| Steamline Break Outside Containment ID: MSS-3 l
| |
| Brief Plant Response: [IC-17, 100%, all systems in automatic, maximum size break]
| |
| When the steamline break is initiated, steam flow on all steam generators increases. RCS temperatures drop, pressurizer pressure and level decrease rapidly. Steam pressure drops to causing steaaline isolat, ion and safety injection. When the MSIVs close, steam flow from the non-affected steam generators drops to zero, with the affected steam r,enerators's flow remaining.
| |
| high. Steam flow from the affected steam generator continues until it boils dry. The first annunciators received include S/G LOW PREb5 STM LINE ISOL ALERT, MS PRESS LOW, S/G FLOW MISMATCH W TLOW LOW AND ST* AM LINE LOW PRESS SI/RX TRIP.
| |
| Suggested Instrtetor Action:
| |
| None.
| |
| Events: None l
| |
| l l
| |
| l l
| |
| i 1
| |
| s 0125w:4 311M/85M/7 1/87 J
| |
| | |
| BRAIDWOOD SIMULATOR MALFUNCTION l i
| |
| | |
| ==Title:==
| |
| Main Steam Isolation valve (s)' Failure ID: MSS-5 lj NO: 6.3.4.8.5 i l
| |
| | |
| == Description:==
| |
| Selected or all S/G isolation valves fail due to a control system failure.
| |
| If gna valve fails, fault is due to individual valve switch contacts.
| |
| If all valves fail, fault is l due to a failure of the steamline (
| |
| isolation Relay K616, Train A.
| |
| Variations: None. Date: 4/9/89 Rev 7 Selectable Steps Inputs Comments
| |
| : 1. Select faulty MSIV(s) MSS-5A MSS-5A = loop A MSIV MSS-5B MSS-5B = loop B MSIV
| |
| [
| |
| * MSS MSS-5C = loop C MSIV MSS-SD MSS-5D = loop D MSIV MSS-5E MSS-5E = all four loops
| |
| : 2. Select valve position -1/1 -1 = Close 1 = Open
| |
| : 3. Select ramp time 0-99,999 see Brief Plant Responses (IC-17, 100%, all systems in automatic)
| |
| If all MSIVs close: Tavs increases, pressurizer level and pressure increase, pressurizer power-operated reliefs lift, reactor trips on high pressure or steam generator low-low level. PRT level, temperature and pressure increase.
| |
| S/G safeties and reliefs lift. First annunciators received include STEAMLINE ISOLATION S/G FLOW MISMATCH FW FLOW LOW, TAVE CONT DEV HIGH and HIGH AUCT TAVE.
| |
| 875M/1 5/89
| |
| .. ..-- --_ .-=-_-_-::-=_==_:-_-_:-_ _ _ _ _ _ - _ _ -
| |
| | |
| *- e e .e-- e.e .emmw . 4 .M..,ew..a.. ...r,.. ,,,m ,, , , , , , _,
| |
| BRAIDWOOD SIMULATOR MALFUNCTION ,1 I
| |
| .2
| |
| .[
| |
| | |
| ==Title:==
| |
| Main Steam Isolation Valve (s) Close ID: MSS-5
| |
| - Brief Plant Response (continued):
| |
| [ Single MSIV closure]
| |
| l.
| |
| Steam flow in affected S/G decreases rapidly, while its steam pressure l
| |
| increases, causing S/G 1evel to shrink. Steam flow in other S/G's increases.
| |
| Affected steam generator relief lifts. Possible reactor trip on S/G low-low level or OPAT. The first annunciators received include STEAMLINE ISOLATION, LOOP TAVE DEV LOW, TAVE CONT DEV HIGH, AUCT TAVE HIGH and S/G FLOW MISMATCH FW FLOW LOW.
| |
| Suggested Instructor Action: )
| |
| None.
| |
| Events:
| |
| : 1) LER 06-01-85-027: Failure of MSIV's to close
| |
| : 2) LER 20-01-88-024: Inopertble MSIV
| |
| : 3) DVR 20-02-88-165: Failure of 2C MSIV to close f
| |
| O 875M/2 5/89
| |
| | |
| _.. -- - - - - ~
| |
| ~~ ' _ _ _ _ _ _ _ - _ - - - _ -
| |
| LICENSEE E C #E E (Li81 F i$-f Fac'11ty ::arne (1) Occret %meer (2) 81ce !11 4rren. Un9t 1 SI 51 of of 01 al si a 1 l efl 3l(
| |
| * OI FAftutt 0F Msiv To CLOSE ON Ms fiataff04 sf CNAL e Event Date (s1 LER utamher f 6) Renert Date (71 other Facilities involved (R)
| |
| .enth Day ' Year Tear Sequential Revtsion Month Day Facil)tv namme k- //), g,, Year
| |
| ,i Docket Numterfs)
| |
| Y .-,
| |
| ~
| |
| i . - . ,
| |
| most of si of of of I I als 114 att als
| |
| ~~~
| |
| of 21 7
| |
| ~
| |
| l ! 017 bli 81 7 of si of of of f I THIS RF*0RT IS SUOMITTEDyd5UANT TO INE REQUIREMENTS OF 10CFR r o.- -.---m the fallawinal f111 1 20.402(b) _ 20.405(c)
| |
| St.73(a)(2)(tv) 73.71(b)
| |
| POWER _ 28.405(a)(1)(1) _ 50.36(c)(1) 58.73(a)(2)(v) .__. 73.71(c)
| |
| Ltyrt te.ac5(a)(1)(11) so.36(c)(2) se.73(a)(2)(vit) .2,. Other ($oecify (1st el c! e zo.405(a)(1)(itt) 5s.73(a)(2)(1) se.73(a)(2)(vtit)(A) in Abstract
| |
| ,_ 20.405(a)(1)(iv) .2 50.73(a)(2)(11) se.73(a)(2)(vtit)(s) below and in
| |
| ., 28.405(a)(1)(v) _ so.73(a)(2)(111) 58.73(a)(2)(x) Text)
| |
| Effruitt CONTAff FOR THIS LER f121 same l Trt rpMaar utmata ARIA COOK Miena.1 1. mantaman - system tant rnainear rut. 21aY ai1 Is 21 11 41 -1 si 41 411 ffbrLETE Det LINE FM EAfM CfDFOMENT Faf ttet Of1ERf RED IM fiffi EEPORT (111 CAUSE SYSTEM COMP 0stNT MamUFAC- REPORTA8tt // CAU$t SYSTEM COMPONENT j MAmufAC- REPORTABLE //
| |
| Timra ,
| |
| TO RPRDS Timaa TO KPADR I
| |
| l I I I l l l I I I l l I g
| |
| n /'l I
| |
| I I l l I I i l 1 i l I I j
| |
| /T/b SUPPLEMtkTAL REPORT EXPf f f tD f 141 Expected Month l Day l Year Substss1on f lYan fIf van. cams,1 ate EXPicTED *1'mMft1 ION DAffl Y l 30 Date (15) l 7 .1t i I l i
| |
| \ A85 TRACT (Limit to 1400 spaces, t.e. approximately fifteen single-space typewritten lines) (le)
| |
| During plant restoration fo,11owing the toss of Offsite Power Startup Test. a safety Injection and Steamline Isolation were initiated. t ain M Steam Isolation Valves (M51V) 18 and IC closed issuedtately as result of the steam 1 ke isolation. However, the IA and 10 M51V's fatted to fully close.
| |
| Subsequent investigations determined _that several of the air supply check valves installed on the M5!v actuators failed to seat when the instrument air supply header slowly depressurized. The ability of the
| |
| , check valves to seat and maintain the air charge stored in the MSIV actuator is required to permit M51Y closure following a loss of air supply pressure, which is an espected result following a loss of offsite power'.
| |
| As a corrective action. the air supply check valves on the Unit One M5!V's have been replaced. Stroke testing of the MSIV's fellowing a slow depressurization of the air supply has demonstrated the adequacy of each of the replacement check valves as well as the abiltty of the M5!V's to close following a loss of supply air pressure.
| |
| The ortgtnally designed check valves wure replaced.with an upgraded brass check valve on a temporary basis untti stainless steel check valves can be procured. M5!Y check valve testing was performed bi-weekly untti July 8. 1985. Six tests were performed and all were successful. Testing will continue on a sont-annual basis through Unit One's second refueling.
| |
| I O
| |
| (1523M/0175M)
| |
| | |
| T -
| |
| L ME45EE EVEN' RE8CRT .LFO) ' Ext MN T!w ?;p FACILITY eaME (1) DOCKET eL*st2 (2) LER 4t#eED f5) em ,3 9 l fear /// Stoutntial // aevision i
| |
| / "r /jj/
| |
| j// eMer Avran. tout 1 0 l s 1 0 l 0 1 0 l al El a ale -
| |
| Gl 21 7 - 013 off 0F Sl2
| |
| [ TEXT Energy Industry Identif G ! ton System (t!!5) codes are identtfted in the tout as (aw)
| |
| On March 14, 1905, at titt hours, uhtle operating in Mode 1 at 12% power, an anticipated' plant trip was initiated as a result of testing being performed within startup test 2.05.30. ' Loss of Offsite Power". As a normal result of a less of offstte power, the Unit he Station Air Compresser (SAC) tripped. A gradual depressertsation of the. instrument air header occurred as a result of the compressor trip. The toss of I Offstte Power test was completed and plant resteretten fellowing the plant trip was beg'n u at 2133 hours. ,
| |
| At 2148 hours the Unit One sac was started. Instrumongairpressurewasfullyrestored'at2216 hours.
| |
| i At 2183 hours. Safety gejectten and Steamline Isolatten were initiated by a law steamline pressure signal. j The cause of the low steamitne pressure is explatned in LER 88-035-06. Main Steam Isolation valves (M5!V) 18 shd IC closed tamedtately as result of the steamitne tselation. Mm ever. the 1A MSIV remained full open and the 10 M5tv anly partially closed. Attempts to manually actuate steam 1tne isolation failed to close the lA and 10 MSIV's. Eventually, the lA and 10 MS!V were both steely closed ustng their air powered hydraulic pumps felleming restoration of instrument air pressure. Full closure of the 1A and 10 MSIV's was
| |
| -completed by 2222 hours.
| |
| Fellemup testing to determine the cause of the MS!V's.FAtlure to close included stroke testing of the 1A and 18 M5tV's with air supply available. Investigattains and testing to determine the effects of a gradual less of air supply pressure, and bench testing of the Unit One M5tV air check valves and the spare parts inventory check valves for acceptability. The results of this testing demonstrated that a gradual depressurizatten of the instrument air supply did not cause adeguate seating of Llw air supply check valves in many cases. Failure of an' air check valves to seat following a less of air supply pressure can allow the valve actuator air reservelr pressure to decay to the point dere the MS!V will not operate.
| |
| The air check valves in the Unit One MS!V's have been replaced with check valves dich passed pressure decay bench testing. Fe11 swing the insta11atten of the bench tested check valves, each MS!V Eas stroke tested under condittens dere the air supply was isolated and a gradual decrease in air supply pressure occurred over the perted of one hour. Each of the MS!V's stroked fully closed in less than two seconds.
| |
| During the first month y test. two check valves failed. These valves were replaced 'with replacement check valves which were bench tested prior to and field tested after insta11atten. After return to power operatten and subsequent completten of a planned Startup Test Program reactor trip etch completed the 30%
| |
| Power Sequence, new upgraded brass check valves were installed in all M51V's. These valves were bench tested prior to insta11atten and field tested after insta11atten by performing a pressure decay test to verify proper operation. Each check valve functioned satisfactorily. A modification was installed to enable us to test the check valves during normal plant operation.
| |
| Testing of the MSIV actuator check valves was done on a ht-weekly bests until July 8.1988. Based on the fact that all eight valves performed successfully the sta tiens they were tested. the testing interval was entended to sent-annually effective July 8.1985. This schedule will continue to Unit One's second refueling. Verhal concurrence was given by Kevin Connaughton NRC Resident Inspector on July $. 19sE.
| |
| In additten to the testing, a meme has been issued to all Licensed Operators stating that untti an acceptable confidence level is attatned for the N5!V operators, the MB!V's are to be closed following a j less of offstte peuer ev,ent.
| |
| 1 (1823M/0178M)
| |
| . _ _ _ . . _ . _ . . _. . . . . . . . . ~
| |
| | |
| ...___.__..7
| |
| - M 5 5 - $~
| |
| LICENSEE EVENT REPORT (LER)
| |
| Form Rev 2.0 Facility Name (1) Docket Number (2) Pace f3) n 01 5l O! 01 01 41 51 6 1lof 0l4l
| |
| * Inoperable Main Steam Isolation Valve Due to Failure of M1 Four Way Solencid valve Event Date f 5) LER Nyghg.t,j6) Resort Date (7) Other Fatilities Involved (8)
| |
| Sequential Revision Month Day Year Facility Names Docket Number (s)
| |
| Month Day Year Year mar Number NONE 01 5l O! 01 Of I I I
| |
| ~ ~~~
| |
| 11 0 31 1 81 8 81 8 0l214 010 111 01 7 81 8 01 51 r,f 01 01 I l THIS REPORT IS SUBMITTED PUR$UANT TO TME REQUIREMENTS OF 10CF::
| |
| OP N M (Check one or more of the followino) (11) 20.402(b) _ 20.405(c) _ 50.73(a)(2)(iv) _ 73.71(b)
| |
| 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 (Speci fy (101 0 !9!7 __ 20.405(a)(1)(iii) _.__ 50.73(a)(2)(1) _ 50.73(a)(2)(vill)(A) in Abstract 20.405(a)(1)(iv) K_ 50.73(a)(2)(ti) _ 50.73(a)(2)(vili)(B) below and in
| |
| _ 20.405(a)(1)(v) _ 50.73(a)(2)(iii) _ 50.73(a)(2)(x) Text)
| |
| LICENSEE CONTACT FOR THIS LER f12)
| |
| Name TELEPHONE NUteER AREA CODE Dan Stroh Technical Staff Enaineer Ext. 2477 8l1 15 415181-l218101 COMPLETE ONE LINE FOR EACH COMPON N FAILURE DESCRIBED IN THIS REPORT (13)
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| CAUSE SYSTEM COMP 0NENT MANUFAC- REPORTABLE CAUSE SYSTEM COMPONENT MANUFAC. REPORTABLE TURER TO NPRDS TURER TO NPRDS X 518 I H1 Cl U Al 31 91 1 Y l l l l l 1 l _
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| l I I I I I I I l l l 1 I l SUPPLEMENTAL REPORT EXPECTED (14) Expected Month l Day l Year w Submission lyes (If ven. conclete EXPECTED SUBMISSION DATE) X l NO l ll l ABSTRACT (Limit to 1400 spaces, i.e approximately fifteen single-space typewritten lines) (16)
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| At 1500 on October 31, 1988 Limiting Condition for Operation Action Requirement (LC0AR) 3.2-la was entered due to the 1A Main Steam Isolation Valve (MSIV) standby hydraulic accumvlstor pressure below 4800 psig. Shortly thereaf ter, the active hydraulic accumulator pressae cecreased to 4800 psig and the hydraulic pump could not maintain hydraulic accumulator pressure. At 1645 the 1 A MSIV was declared inoperable and preparations were made to go into mode 2 c5f. power, per Technical Specification 3/4.7.1.5. Mechanical maintenance began work on the valve by preparing to remove the "N1" 4 way hydraulic valve which was thought to be internally leaking. At 1803 an Event Notification System phone call was made to the NRC per Braidwood Administrative Procedure 1250-6A3. I.15 example holit. At 2043 the source of the hydraulle leak had not been found and a power reduction was begun. Because of forced power reduction per technical specification, an unusual event was declared due to Emergency Action Level 414 at 2115. At 2117, a Nuclear Accident Reporting System phone call was iaade. At 0040 on November 1, 1988, Unit I entered mode 2. and at 0119 the 1A MSIV was closed using only the hydraulic pump. +t 0225, a mechanical block was installed on the 1A MSIV and the event was terminated at 0230. Replacement of the defective "MI" 4 way hydraulic valve was completed by 1000 and the 1A MSIV was tested and returned to service at 1921.
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| l l
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| 2368m(112988)/8 i
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| g i:
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| 1ICENSEE EVENT REPORT (LER) TEXT CONTINUATION Fgg g2A
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| ', DOCKET NUMBER (2) LER NUPSER (6) Psoe M)
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| FACILITY NAME (1).
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| Year /// Sequential //j/ Revision
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| : x. . f7f
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| /// ' Number ff
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| /// Number
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| '/
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| t Y Braldwood Unit I V
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| /; ^
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| 01S1010l0l41516 8I8 - Ol2l4 - Ol 0 01 2 0F Ol' 4 TEXT Energy Industry Identification System (EIIS) codes are identified in the text as (xx)
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| A. PLANT CONDITIONS PRIOR TO EVENT:
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| Unit: Braidwood 1; Event Date: October 31, 1988; Event Time: 1645; Mode: 1 - Power Operation; Rx Power: 98%;
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| RCS (AB) Temperature / Pressure: 557 degrees F/2231 psig B. DESCRIPTION OF EVENT:
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| There were no structures, systems or components inoperable or degraded at the beginning of the event that contributed to the event.
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| At 1400 on October 31, 1988 an operator noted that the hydraulle pump on the 1A Main Steam (MS) (58)
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| -Isolation Valve (MSIV) was pumping continuously. Further investigation revealed that the standby hydraulle accumulator pressure was at 3700 psig. The active accumulator pressure was still at 5000 psig. Because a train is considered inoperable when the hydraulic pressure decreases below 4800 psig, Tec.nnical Specification 3/4.3.2 Action Statement number 23, which requires the inoperable train be restored within 48 hours, was entered at 1500. Mechanical Maintenance (NMD) began to prepare a work package to rep 1 ace the "N1" four way hydraulic valve which was thought to have an internal leak. Shortly after 1630, the pressure on the active IO hydraulic accumulator began to decrease. At 1645 the 1A MSIV was declared inoperable after it became
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| ( apparent that the hydraulic pump could not maintain the required hydraulic pressure on either accum'ulator.
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| This was a result of pump discharge flow being diverted to the leak on the standby system. This prevented the active hydraulic accumulator from being repressurized and in f act allowed the active hydraulic accumulator to depressurize due to normal system losses. Per Technical Specification 3/4.7.1.5 Mode 1 Action Statement, the valve had to be repaired within 4 hours (2045), or the plant be placed in Hot Standby in b hours and in Hot Shutdown within the following 6 hours. Upon entry into Mode 2, this Technical Specification will allow indefinite operation as long as the inoperable MSIV.is closed and maintained closed. A decitinn was made to attempt to repair the valve under the allotted 4 hour time clock and if repairs could not be completed, perform a power reduction to Mode 2 and complete the repairs.
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| At 1930 the 1A MSIV was taken out of service and HMD replaced the "N1" four way hydraulle valve by 2000.
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| However, the leak was not stopped and hydraulic pressure could not be restored. Subsequent bench testing of j
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| the "N1" valve proved that it was functioning properly. Following replacement and testing of the suspect "N1" valve, the Anchor Darling vendor representative was consulted. It was then determined that the only l other cause could be the second valve in the hydraulle circuit (i.e. "M1"). This valve is identical to the "N1" valve. The decision was made to exchange the good *NI" valve that was removed, with the installed "M1" valve. At 2043 a power reduction was begun to satisfy Technical Specification 3/4.71.5. ]
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| At 2115, because of a furced power reduction per Technical Specifications, an Unusual Event was declared per Emergency Action Level (EAL) #14 at 2115.
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| At 2121, the System Power Supply Of fice (SPS0) verified the Generating Stations Emergency Plan (GSEP) classification. The Illinois Emergency Services and Disaster Agency (IESDA) was also contacted.
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| At 2127, the Station Duty Officer (500) and the Nuclear Duty Officer (NDO) were notified of the GSEP event.
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| * Mourly updates were made to IESDA and Illinois Department of Nuclear Safety (IDNS).
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| 2368m(112288)/9 I
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| n----- . _ _ _ .
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| i LICENSEE EVENT REPORT ftER) TEXT CONTINUATION Form Dev 2.0 i FACILITY NAME (1) DOCKET NUMBER (2) LER NUPEER (6) Pace (J) _
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| Year / Sequential /jj// Revision
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| ,/,/p/ /j//
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| [a} Braidwood Unit i // Number Number 0 1 5 1 0 1 0 1 0 l dl 51 6 8lB - 0l2l4 - 01 0 01 3 0F 01 4 TEXT Energy Industry Identification System (E!!S) codes are identified in the text as (XX)
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| B. DESCRIPTION OF EVENT, Continued:
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| At 0040 on November 1,1988, Unit i entered Mode 2, and at 0119 the 1A MSIV was closed using the hydraulic pump.
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| At 0225, a mechanical restraint was installed on the 1A MSIV.
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| At 0230, the Unusual Event was tennineted af ter Technical Specification 3/4.7.1.5 Attion Statement for Mode 2 was fully complied with.
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| Repairs were made on the 1A HSIV, the valve was tested and returned to operable status by 1921 on October 21, 1988.
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| The appropriate NRC notification via the ENS phone system was made at 1803 pursuant to 10CFR50.72(b)(1)(li) and Braidwood Administrative procedure BwAP 1250 - 6A3, I.15 example h.iii. This was 18 minutes late due to the delay in recognizing the one hour time requirement.
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| This event is being reported pursuant to 10CFR50.73(a)(2)(44) - Any event or condition that resulted in the condition of the nuclear power plant, including its principal safety barriers, being seriously degraded, or that resulted in the nucIoar power plant being in an unanalyzed condition that significantly n compromised plant safety.
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| C. CAUSE OF EVENT:
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| The MSIV has two redundant hydraulic accumulators available to close the valve. Each hydraulic accumulator has two identical four way hydraulic valves; an "Ml" and an "N1" valve. The "Ml" valve controls the opening and closing of the MSIV and the "N1" valve controls the charging of the hydraulic accumulator. Because the standby hydraulic accumulator lost pressure, the "N1" valve was first suspected. However, replacement of the "N1" valve did not stop the hydraulic leak. When the "M1" valve was replaceo, the hydraulic system went solid and the hydraulic pump was able to fully pressurize both the active and standby accumulator. The cause of the event was a failure of the "M1" four way valve.
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| When the "M1" four way valve was disassembled, two f ailed 0-rings were found. The failure of these 0-rings allowed hydraulic oli to pass through the valve internals back to the reservoir. This prevented the hydraulic system from building required pressure.
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| D. SAFETY ANALYSIS:
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| Although neither the active or standby accumulator were available to close the 1A MSIV within the required 5 seconds, the hydraulic pump was available and was used to close the valve. Even if the valve had totally failed in the open position, Technical Specification 3/4.7.1.5 could have been complied with and the plant would have been brought into Hot Shutdown within ten hours in a safe and controlled manner. The worst case scenario would have been a main steam line break during Hot Shutdown at zero power which has been analyzed in the FSAR as a uncontrolled cooldown of one steam generator.
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| D 2368m(112188)/10 1
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| n.- .
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| E o LICENSEE EVENT REPORT fLER) TEXT CONTINUATION Form _ REQ DOCKET NUMBER (2) LER NUteER (6) Pace t3)
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| [_ y FACILITY NAME (1)
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| Year Sequential Revision s1
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| {/((
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| // Number g//
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| / Number j
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| ~
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| Braidwood Unit 1 0 l 5101010 l dl 516 8l8 - 0l214 - O l' 0 01 4 Or Q,,g l TEXT Energy Industry Identification System (EIIS) codes are identified in the text as [XX)'
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| l E. CORRECT 1YE ACTIONS:
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| I' The'immediate corrective action involved reduction of power and isolation of the 1A M51V. Nuclear work I Request (NWR) A26531 was written to rep 1' ace the defective four way hydraulic valve ("M1"). Subsequent testing restored both trains of the t4SIV to an operable status.
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| Long term corrective action is being p'ursued by Pressurized Water Reactor Engineering (PWRE). They will be utl11 ring the services of a third party expert to review the operating history of the M5tV's as well as the Identified root cause for this failure. Based on this review. PWRE will be providing recommendations for long term corrective action. .This review will be tracked to completion by Action Item 456-200-88-25501.
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| F. PREVIOUS OCCURRENCES:
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| There have been no previous occurrences of MSIV failure due to 0-ring failures.
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| G. COMPONENT FAILURE DATA:
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| Manufacturer Nomenclature Model Number MFG Part Number Anchor-Darling Four Way N/A 23304 hydraulic directional i
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| slide valve 1
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| l l
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| 1 b
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| 2368m(112188)/11 i
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| . _ _ _ _ _ _ . . . _ .. = . - , . . . .
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| I' m s5-S DEVIATID3 INVESTIGATED 3 REPORT (DIR) ,
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| ' pa0E <
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| Facility Name araidweed 2 1 fCFl 0 1?
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| , g Title Failure of the 2C MSIV to Close Due to Low MydraultC Pump Pressure and Low Accumulator Prectsarge Pressure
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| (
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| Dft NUPGER REPORT Daft I EVtWT Daff
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| ,,~
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| $EQUENTIAL REVISION MODE 3 MONTH cay YEAR iTA UNIT YEAR NUMBER NUMBER MONTH day YEAR f
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| POWER Il e of 1 al a 21 o el 2 al a -
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| Il si s -
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| eIo 11 e 21 a al a el of n a
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| CONTAET FSR Twit Oft TELEPHONE NUMaER NAME '
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| ADEA CODE Dan stroh. Techn9eal Staff Encinaar fat. 2477 a1i!E 4 l E 1a1-12IBIO l1 COMPLftf DNE Lfht FOR [ACH COMPONE LORE Of1ERIBfD IN THft RfP0kT REPORTABLE CAUSE SYSTEM COMPONENT MANUFAC- REPORTABLE CAUSE SYSTEM COMPONENT MANUFAC-TURER 70 NPRDt TURER To NPRDS I I I I I I I I I I I I I I I I I I I I I I I I I I MONTH cay YEAR
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| $Uff1EMENTAL REPORT EXPttT[0
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| $UBM!$$10N I No l l l I Yt1 fit van. enenlate EXPECTED tutMf55f0N catti TEXT Energy Industry ! certification System (E113) codes are identified in the text as [XX]
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| A. Plant tendition5 #rter to Eveet:
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| Unit: Braidwood 2; Event Date: October 1. 1988; Event Ttme: 1600 A
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| Mode: 3 - Hot Standby; Rx Power: 0%
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| RCS (A8) Temperature / Pressure: 558 cegrees F/2238 pstg
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| : 8. Deser4ettsn ef tveat:
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| On DCtoter 1 1988 at 1600 hrs the control swittnes for all four Main Steam Isolatton Valves (MSIV) were taken to the ELOSE position in preparation to take the Unit 2 Concenser out of service. All four valves i
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| .ent full closed with the exception of the 2C valve which showed dual 11gnt incteatton. A 8-man was dispatCPed to the field and ne verified the valve was alnest closed but still cff of the tower limit j i
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| switCft. A maeual Main Steam Isolation was inttiated from the CCntrol Room. This signal uses both tPe active and stancDy accumulator for closteg. When initiated. the 2C M51V went full closed. Nuclear Work Reouest (NWR) #A2$932 was written to investigate and repair tre non-closure problem. Upon investigation.
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| the hydraul1C pressure of the accumulators was found at a100 pstg. it should nave been at $300 pstg.
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| Mechanical Maintenance verified there were no problems with hydraulic or pneumatic pilot valves on the '
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| Unit and both hycraulic and pneumatic accumulators were properly recharged. The valve was returned to service on Tuesday 10-a-88 The plant remained stable throughout the event.
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| i 23a5m(110288)/24
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| - _ - _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ __ ,, _ ]
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| ~
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| p -
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| DEVIAT!03 INVESTIGATED *4 etPORT TEXT CONTINUATION rerm Cev 2.9 ffR NUMets i part FACILITY NAME
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| . jW .
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| Bra 10 wood 2 SE0utNTIALl REVISION Nup4Ee STA UNIT YEAR NUMate v - oI o er a12 21 o el 2 af a -
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| Il si E -
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| 2 TEXT G C. cause et tvant:
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| The cause of the event was two fold; low hydraulic pump press'ra u and low nitrogen precnarge pressure on the
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| ' accumulator. The cause of the low bycraulic pressure was low atr pressure to the cump, the lock nut on the air regulator was loose and apparently the regulator had backed off. The regulator was properly aojusted and the lock nut ttgntened.
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| D. taf et y Anaivtit:
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| Proper actions were initiat'ed by the operator in a timely manner. No unusutl safety concerns resulted from the equipment fatlure since the valve was brought to the Engineered Safety Position (closec) when the problem was ciscoveres. The inactive train was chargea a'na was used to fully close the valve within 5 seconds after receipt of a close signal.
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| [. ferrective Act9an:
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| NWR #25932 was generated to trebblesnoot and recharge both the nitrogen and hygraulic systems. The' air regulator to the hydraulic pump was adjusted and lockea in place.
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| 0
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| 'T .
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| F. Previous occurrence:
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| None G. comonenet Failure Data:
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| None l
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| l l
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| [
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| (
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| l l 2345m(1102881/25 t
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| BRAIDWOOD SIMULATOR MALFUNCTION
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| ==Title:==
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| Steam Generator Atmospheric Relief Valve Failure 1D: MSS-6 NO: 6.3.4.8.6
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| == Description:==
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| Selected steam generator relief valve fails to specified position due to controller failure, with manual control available.
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| Variations': See MSS-18 Date: 4/9/89 Rev: 7
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| ~
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| Selectable Steps Inputs Conunents
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| : 1. Select fail position 0-100 percent 0 percent: = Fully closed 100 percent = Fully open
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| : 2. Select steam generator 1-4 1 = MS018A (A S/G) with faulty relief 2 = MS018B (B S/G) l valve. 3 = MS018C (C S/G) 4 = MS018D (D S/G) l l
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| l Brief Plant Response: [ Based on fault occurring at approx. 5 percent reactor power)
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| Note: When activated at low power levels, plant response to this malfunction is more noticeable.
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| O 8751U3 5/89 L - - --- __ _--__-_______- ___________ _ _ .
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| BRAIDWOOD SIMULATOR MALFUNCTION I [7
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| ==Title:==
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| Steam Generator Atmospheric Relief Valve Failure ID: MSS-6 lXY Brief Plant Response (continued):
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| When the relief var.ve opens, the steam flow from the selected steam generator increases. This ircrease in steam flow will cause a decrease in Tavs and a drop in pressurizer pressure and level. S/G 1evel will initially swell; then decrease. The first annunciator received is T,y, CONT DEV LOW.
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| If activated at full power, the increase in steam flow will cause an increase of approx. 2-3 percent reactor power.
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| Suggested Instructor Action:
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| When requested, locally isolate the faulty relief valve by using:
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| LOA MSS-36: MS019A '
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| MSS-37: MS019B MSS-38: MS019C MSS-39: MS019D Events: ,
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| : 1) SEE 29-86: Inadvertent Rapid Cooldawn and Depressurization.
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| : 2) DVR 06-01-88-074: ID S/G PORV Inoperable.
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| l
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| : 3) DVR 20-02-88-157: Spurious Lifting of 2A S/G PORV.
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| k 875M/4 5/89 I
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| IS 621 FORSYTH (INP'O) 12-AUG-86'08:24 PT
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| ==Subject:==
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| ' SEE 29-86, RAPID COOLDOWN AND DEPRESSURIZATION 0- .
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| ==SUBJECT:==
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| INADVERTENT RAPID COOLDOWN AND DEPRESSURIZATION DURING A REMOTE SHUTDOWN TEST.,
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| UNIT (TYPE)) CATAWBA 2 DOC NO/LER NO: 50-414/86028 L EVENT DATE: 6/27/86 NSSS/AE: WESTINGHOUSE / DUKE POWER COMPANY l
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| ==SUMMARY==
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| -DURING THE PERFORMANCE OF A REQUIRED. POWER ASCENSION TEST (LOSS OF CONTROL-ROOM), THE PRIMARY SYSTEM EXPERIENCED A RAPID COOLDOWN AND DEPRESSURIZATION. AS PART OF THE TEST, THE UNIT WAS TRIPPED AT 24 PERCENT POWER, AND CONTROL WAS TRANSFERRED TO THREE REMOTE SHUTDOWN PANELS. WHEN CONTROL WAS TRANSFERRED, ALL FOUR STEAM GENERATOR POWER-OPERATED RELIEF VALVES (PORVs) OPENED, CAUSING A RAPID DECREASE IN PRIMARY SYSTEM TEMPERATURE AND PRESSURE. AS A RESULT OF THE TEMPERATURE DECREASE, PRESSURIZER LEVEL INDICATION WENT OFFSCALE LOW.- WHEN THE TEST WAS TERMINATED AND CONTROL WAS TRANSFERRED BACK TO THE MAIN CONTROL ROOM, AN AUTOMATIC SAPETY IWJECTION. OCCURRED (PER DESIGN) ON LOW STEAM LINE PRESSURE.
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| THE INITIATING CAUSE OF THIS EVENT WAS INADEQUATE IMPLEMENTATION-OF A DESIGN MODIFICATION TO THE STEAM GENERATOR PORV AN EQUIPMENT MALFUNCTION, IMPROPER LABELING ON THE O CONTROLLERS.
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| REMOTE SHUTDOWN PANELS, AND LACK OF EXPLICIT TEST TERMINATION CRITERIA CONTRIBUTED TO THE EXTENT AND DURATION OF THE EVq -
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| THIS EVENT-IS SIGNIFICANT BECAUSE INADEQUATE DESIGN REVIEW AND CONTROL CREATED A PROBLEM FOR OPERATORS AND IF THE SITUATION HAD REQUIRED AN ACTUAL CONTROL ROOM EVACUATION, EXISTING CONDITIONS COULD HAVE PRECLUDED A SAFE AND ORDERLY UNIT SHUTDOWN.
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| DESCRIPTION:
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| ON 6/27/86, CATAWBA UNIT 2 WAS OPERATING AT 24 PERCENT POWER, AND PREPARATIONS WERE UNDERNAY TO PERFORM A LOSS OF CONTROL ROOM TEST. THIS TEST WAS INTENDED TO VERIFY THE ABILITY TO SHUT DOWN THE PLANT FROM OUTSIDE THE CONTROL ROOM. THE OPERATING SHIFT CONDUCTED A PRETEST BRIEFING AND PROCEDURE WALK-DOWN ON THE PREVIOUS AFTERNOON. NO' PROBLEMS WERE IDENTIFIED. THE TEST A
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| PROVIDED FOR THE NORMAL OPERATING SHIFT TO CONDUCT THE TEST.
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| MINIMUM NUMBER OF OBSERVERS WERE TO REMAIN IN THE' CONTROL ~ ROOM TO MONITOR THE OPERATION OF THE REACTOR COOLANT PUMPS. THESE PUMPS WERE TO REMAIN IN OPERATION TO SIMULATE DECAY HEAT. THE CONTROL ROOM OBSERVERG WERE A SENIOR REACTOR OPERATOR AND A LICENSED OPERATOR WHO WERE TO MAINTAIN COMMUNICATIONS WITH THE OPERATING SHIFT. HOWEVER, THEY WERE ONLY TO COMMUNICATE INFORMATION
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| ~
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| PERTAINING TO THE EQUIPMENT INTENTIONALLY LEFT OPERATING AFTER O TEST INITIATION.
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| O AN EQUIPMENT' MODIFICATION HAD BEEN IMPLEMENTED BETWEEN HOT FUNCTIONAL TESTING, WHEN THIS TEST HAD PREVIOUSLY BEEN PERFORMED, AND UNIT 2 LICENSING. THIS MODIFICATION CHANGED THE FUNCTIONAL CHARACTERIST1CS'BUT NOT THE PHYSICAL APPEARANCE OF THE STEAM-GENERATOR PORV CONTROLS ON THE AUKILIARY FEEDWATER PUMP TURBINE CONTROL PANEL. PRIOR TO THE MODIFICATION, THE CONTROLLERS IN FUNCTIONED AS STEAM GENERATOR PRESSURE SET POINT CONTROLLERS.
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| THIS MODE, THE CONTROLLERS PROVIDED MANUAL ADJUSTMENT OF THE THE PRESSURE SET POINT AT WHICH THE PORVs WOULD OPEN.
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| CONTROLLERS HAD A SINGLE SCALE READING IN UNITS OF PSIG (FULL SCALE BEING 1500 PSIG) AND DUAL POINTERS: ONE POINTER INDICATED STEAM GENERATOR PRESSURE AND THE SECOND INDICATED PRESSURE SET POINT. AFTER THE MODIFICATION, THE CONTROLLERS FUNCTIONED AS DIRECT MANUAL STEAM GENERATOR PORV POSITION DEMAND LOADERS, AND HOWEVER, THE THE SECOND POINTER INDICATED VALVE POSITION DEMAND.
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| SCALE STILL READ IN PSIG UNITS RATHER THAN PERCENT OF VALVE DEMAND. ACCORDING TO THE TEST PROCEDURE, WHICH WAS IN ERROR, THE CONTROLLERS WERE SET AT WHAT WAS BELIEVED TO BE IN REALITY, A STEAM THIS GENERATOR SETTING PRESSURE SET POINT OF 1125 PSIG.
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| PROVIDED A 75 PERCENT OPEN DEMAND SIGNAL TO jHE FOUR STEAM GENERATOR PORVs.
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| A NORMAL. SHIFT TURNOVER OCCURRED AT 0700 ON 6/27/86, AND PREREQUISITES FOR THE TEST WERE COMPLETED BETWEEN 0800 AND 0900. THE TEST WAS INITIATED AT 0941 WHEN THE OPERATIONS PERSONNEL WERE DISPATCHED FROM THE CONTROL ROOM TO THEIR ASSIGNED O STATIONS. AT THAT TIME, THE PRIMARY PRESSURE WAS 2238 PSIG, TEMPERATURE WAS 560 DEGREES FAHRENHEIT, AND PRESSURIZER LEVEL WAS 28 PERCENT. THE STEAM GENERATOR PRESSURE WAS 1030 PSIG.
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| AT 0942, A LICENSED REACTOR OPERATOR TRIPPED THE REACTOR TRIP BREAKERS IN ACCORDANCE WITH THE TEST PROCEDURE. HE THEN PROCEEDED TO THE AUKILIARY FEEDWATER PUMP TURBINE CONTROL PANEL. THIS PANEL IS ONE OF THE THREE AUKILIARY SHUTDOWN PANELS ALSO AT 0942, LOCAL FROM WHICH REMOTE SHUTDOWN IS PERFORMED.
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| THESE CONTROL WAS TAKEN AT AUKILIARY SHUTDOWN PANELS A AND B.
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| TNO PANELS HAVE INDICATIONS AND CONTROLS FOR FUNCTIONS SUCH AS LETDOWN / CHARGING AND SEAL INJECTION. WHEN THIS TRANSFER OF CONTROL WAS PERFORMED, A LETDOWN / CHARGING FLOW MISMATCH
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| ' OCCURRED. THIS LETDOWN / CHARGING MISMATCH, WHICH WAS GREATER THAN ANTICIPATED, RESULTED IN AN INCREASING VOLUME CONTROL TANK LEVEL AND A DECREASING PRESSURIZER LEVEL. BY 0947, THE PRESSURIZER LEVEL HAD DROPPED TO AN INDICATED 18 PERCENT.
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| AT 0943, THE LICENSED OPERATOR WHO HAD TRIPPED THE REACTOR TRIP BREAKERS ARRIVED AT THE AUKILIARY FEEDWATER PUMP TURBINE CONTROL PANEL, AND TOOK CONTROL. THIS PANEL HAS INDICATION AND CONTROLS ASSOCIATED WITH THE STEAM GENERATOR PORVs AND OTHER FUNCTIONS.
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| WHEN THE LOCAL POWER FEEDER BREAKERS FOR THE STEAM GENERATOR PORVs WERE CLOSED AT APPROXIMATELY 0947, ALL FOUR STEAM GENERATOR PORVs OPENED TO 75 PERCENT FULL OPEN, THE PERCENTAGE OF FULL SCALE AT WHICH THE CONTROLLERS H4D BEEN SET.
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| | |
| LTHE'OPERATOK AT THE AUKILIARY FEEDWATER PUMP TURBINE CONTROL PANEL ;IDBEEDIATELY BECAME AWARE OF THE SUDDEN DECREASE IN STEAM Os GENERATOR PRESSURE AND, IN AN ATTEMPT TO ENSURE THE PORVs WERE CLOSED,-MANUALLY-ADJUSTED (WHAT HE THOUGHT TO BE) THE PRESSURE SRT POINT UPWARD. THIS ACTION ACTUALLY CAUSED THE PORVs TO OPEN' EVEN MORE. NO DIRECT INDICATION WAS AVAILABLE TO THE AUKILIARY FEEDWATER PUMP TURBINE CONTROL PANEL OPERATOR OF VALVE POSITION FOR THE STEAM GENERATOR PORVs. THE CONTROL ROOM OBSERVERS DID HAVE POSITION INDICATION IN THE FORM OF RED AND GREEN LIGHTS HOWEVER, THEY WERE RELUCTANT TO COMMUNICATE SUCH INFORMATION TO OPERATORS ON THE REMOTE SHUTDOWN PANELS TO.AVOeD INVALIDATING THE TEST. AT 0950, THE PRESSURIZER PRESSURE HAD DECREASED TO 1845 PSIG AND STEAM LINE PRESSURE TO 725 PSIG, THE SAFETY INJECTION SET POINT. HOWEVER, AUTOMATIC SAFETY INJECTION MAS PARTIALLY BLOCKED (PER DESIGN) BY TRANSFER TO THE REMOTE SHUTDOWN PANELS.
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| AT 0952, THE SRO IN THE CONTROL ROOM ORDERED TEST TERMINATION AND TRANSFER OF CONTROL BACE TO THE MAIN CONTROL ROOM. .UPON TRANSFER, AT 0953, AUTOMATIC SAFETY INJECTION ACTUATION WAS UNBLOCKED AND OCCURRED, THE STEAM GENERATOR-PORVs CLOSED, AND THE BY 0958, THE PRESSURIZER PRESSURE AND LEVEL BEGAN RECOVERING.
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| PRESSURIZER LEVEL AND PRESSURE HAD. RETURNED TO APPROXIMATELY 30 PERCENT AND 1300 PSIG, C.BPECTIVELY. AT THIS POINT, THE SAFETY INJECTION WAS RESET AND REACTOR COOLANT TEMPERATURE WAS STABILIZED AT 468 DEGREES FAHRENHEIT.
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| SUBSEQUENT. INVESTIGATION REVEALED THE FOLLOWING FACTORS THAT CONTRIBUTED TO THE PROGRESSION OF THIS EVENT:
| |
| A. THE DESIGN MODIFICATION TO THE STEAM GENERATOR PORV CONTROL SCHEME DID NOT ADEQUATELY ADDRESS THE CHANGES'NEEDED TO THE AUKILIARY FEEDWATER PUMP TURBINE CONTROL PANEL. FIGURE 2 SHOWS A SIMPLIFIED CONTROL DIAGRAM, BEFORE AND AFTER THE MODIFICATION. THE MOblFICATION REPLACED THE STEAM GENERATOR PORV PRESSURE SET POINT LOADER IN THE CONTROL ROOM WITH A VALVE POSITION DEMAND LOADER AND REMOVED THE PROPORTIONAL l
| |
| CONTROLLER THAT WAS COMMON TO CONTROL ROOM AND AUKILIARY FEEDWATER PUMP TURBINE CONTROL PANEL OPERATION. THIS RESULTED IN THE PRESSURE SET POINT LOADER, LOCATED ON THE AUKILIARY FEEDWATER PUMP TURBINE CONTROL PANEL, FUNCTIONING AS A VALVE POSITION DEMAND LOADER WHEN CONTROL WAS TRANSFERRED TO THIS l PANEL. HOWEVER, THE AUKILIARY FEEDWATER PUMP TURBINE CONTROL l
| |
| PANEL CONTROLLER WAS NOT REPLACED OR RELABELLED. THE NEED i
| |
| FOR RELABELLING WAS NOT IDENTIFIED DURING THE DESIGN MODIFICATION PROCESS.
| |
| O
| |
| | |
| l
| |
| /^ B. THE MODIf! CATION WAS MADE KNOWN TO THE PLANT STAFF, AND INPUT i FROM THE APPROPRIATE DESIGN PERSONNEL WAS SOLICITED BY THE PLANT TO BE USED FOR ANY REQUIRED PROCEDURAL AND TRAINING REVISIONS. HOWEVER, THERE WAS INADEQUATE TRANSFER OF I
| |
| INFORMATION CONCERNING THE CHANGE IN FUNCTION OF THE AUKILIARY FEEDWATER PUMP TURBINE CONTROL PANEL CONTROLLER. ,
| |
| BECAUSE THE APPEARANCE OF THE AUKILIARY CONTROLLER HAD NOT i BEEN ALTERED, THE PLANT HAD THE IMPRESSION THAT THE FUNCTION OF THE CONjROLLER HAD ALSO REMAINED UNCHANGED. THIS RESULTED l IN PROCEDURAL REVISIONS AND OPERATOR TRAINING THAT INCORPORATED THE CONTROL ROOM CONTROLLER PORTION OF THE MODIFICATION IMPACT ON SYSTEM OPERATION, BUT IT DID NOT INCLUDE THE EFFECT ON THE AUKILIARY FEEDWATER PUMP TURBINE {
| |
| l CONTROL PANEL CONTROLLER.
| |
| C. THE LETDOWN / CHARGING MISMATCH THAT OCCURRED AT THE BEGINNING OF THE EVENT WAS CAUSED WHEN LETDOWN PRESSURE CONTROL VALVE 148, DESIGNED TO PROVIDE BACKPRESSURE FOR THE FLOW ORIFICES IN THE LETDOWN LINE, FAILED OPEN DUE TO A FAULTY ELECTRICAL CONNECTION (SEE FIGURE 1). THE SITUATION WAS COMPLICATED BY l THE FOLLOWING:
| |
| PRIOR TO TEST INITIATION, CHARGING FLOW CONTROL VALVE 294 HAD BEEN ADJUSTED TO 32 GPM, AND SEAL INJECTION BACKPRESSURE I CONTROL VALVE 309 WAS CLOSED BY ADJUSTMENT OF AUKILIARY l SHUTDOWN PANEL "A" CONTROLLERS TO LIMIT THE TRANSIENT EFFECT
| |
| (
| |
| ON THE REACTOR COOLANT PUMP (RCP) SEALS DURING THE
| |
| \_ -
| |
| TRANSFER. THE CONTROLLER FOR VALVE 309 ON AUXILIARY SHUTDOWN HOWEVER, PANEL "B" WAS LEFT AT ITS NORMAL OPEN SETTING.
| |
| UNKNOWN TO THE OPERATOR, VALVE 309 WOULD RESPOND TO THE OPEN UPON TRANSFER. THIS RESULTED IN THE VALVE GOING TO AN OPEN i POSITION, INSTEAD OF THE FULL-CLOSED POSITION THAT WAS l DESIRED UPON TRANSFER TO THE AUKILIARY SHUTDOWN PANELS.
| |
| WHEN VALVE 309 OPENED THE OPERATOR ATTEMPTED TO COMPh:tSATE i FOR THE LACK OF FLOW TO SEAL INJECTION BY OPENING VALVE 294; HOWEVER, THE MANUAL VALVE POSITION DEMAND COFtROLLER FOR l VALVE 294 ON'THE AUKILIARY SHUTDOWN PANEL %AS LABELED BACKWARD (INCREASING AND DECREASING DESIGNATIONS WERE TRANSPOSED). THEREFORE, THE OPERATOR'S ATTEMPTS TO OPEN VALVE 294 RESULTED IN CLOSING IT.
| |
| THE PLANT MADE THE NECESSARY EQUIPMENT LABELLING AND PROCEDURAL CHANGES AND CONDUCTED TRAINING TO INCORPORATE THE IMPACT OF THIS MODIFICATION COMPLETELEY. THE LOSS-OF-CONTROL-ROOM TEST WAS ,I SUCCESSFULLY RE-CONDUCTED ON 7/11/86. THE PLANT IS IN THE PROCESS OF REVIEWING ALL MODIFICATIONS PERFORMED BETWEEN !
| |
| t COMPLETION OF HOT FUNCTIONAL TESTING AND LICENSING TO DETERMINE l IF ANY SIMILAR SITUATIONS EXIST.
| |
| a l
| |
| 1 1
| |
| l L-
| |
| | |
| ~
| |
| .CONMENTS:
| |
| IS
| |
| : 1. .AN ESSENTIAL ELEMENT OF THE DESIGN MODIFICATION PROCESS THE= DEVELOPMENT AND DOCUMENTATION OF'A BASIC FUNCTION
| |
| - DESCRIPTION, THIS WOULD COMPLEMENT TECHNICAL AND HARDWARE
| |
| ^ DESCRIPTIONS 0F HOW TOTAL' SYSTEM OPERATION IS AFFECTED BY A O MODIFICATION.- ONLY NITH SUCH A. FUNCTIONAL DESCRIPTION CAN' SV^
| |
| THE SAFETY IMPACT OF.A MODIFICATION SE COMPLETELY
| |
| . EVALUATED. THIS DESCRIPTION SHOULD IDENTIFY NECESSARY REVISIONS TO OPERATING PROCEDURES, MAINTENANCE PROCEDURES, TESTING REQUIREMENTS, AND TRAINING' LESSON PLANS, AND SHOULD BE' INCLUDED AS PART OF THE MODIFICATION PACKAGE DOCUMENTATION. PROCEDURE CHANGES, DRANING UPDATES, AND APPROPRIATE TRAINING SHOULD BE COMPLETED PRIOR TO RETURNING A MODIFIED SYSTEM TO SERVICE.
| |
| : 2. FOLLOWING THE MODIFICATION TO THE STEAM GENERATOR PORV CONTROLS, INDIVIDUAL COMPONENT CHECKS WERE MADE, PUT COMPLETE TO BE COMrsETE, POST-LOOP / SYSTEM TESTING NAS NOT PERFORMED.
| |
| MODIFICATION TESTING SHOULD NOT ONLY VERIFY INDIVIDUAL COMPONENT OPERATION, BUT WHERE APPROPRIATE, SHOULD ALSO '
| |
| VERIFY SYSTEM FUNCTIONAL OPERATION.- ADDITIONALLY, APPROPRIATE COMPONENT / SYSTEM OPERABILITY SHOULD BE VERIFIED PRIOR TO CRITICAL TESTS.
| |
| : 3. WHEN' APPLICABLE, CRITICAL TEST PROCEDURES SHOULD PROVIDE O SPECIFIC' CRITERIA FOR TEST TERMINATION AND SPECIFIC STEPS TO~
| |
| ENSURE TERMINATION IS CONDUCTED IN A SAFE AND ORDERLY
| |
| ,,.i - MANNER. DURING THE CONDUCT OF THE LOSS OF CONTROL ROOM' TEST.
| |
| ' A 0 .d EXPLICIT TEST TERMINATION CRITERIA WERE NOT GIVEN TO THE SRO
| |
| [. OBSERVER IN THE CONTROL ROON. THIS MAY HAVE RESULTED IN INCREASING THE EXTENT AND DURATION OF THE TRANSIENT.
| |
| : 4. IT IS IMPORTANT THAT THE APPROPRIATE PLANT PERSONNEL ARE NELL-TRAINED, PRACTICED, AND HAVE A COMPLETE UNDERSTANDING OF THE PROCESS INVOLVED WITH THE EVACUATION OF THE MAIN CONTROL ROON. AREAS OF IMPORTANCE INCLUDE INDICATIONS AND CONTROLS AVAILABLE AT THE REMOTE SHUTDOWN PANELS AND THE DIFFERENCES THIS BETWEEN A CONTROL R.OOM. SHUTDOWN AND A REMOTE SHUTDOWN.
| |
| IS PARTICULARLY IMPORTANT WITH RESPECT TO CONTROLLING THE PLANT UNDER ABNORMAL CONDITIONS.
| |
| : 5. THE HUMAN PERFORMANCE PROBLEMS THAT OCCURRED DURING THIS EVENT HIGHLIGHT THE IMPORTANCE OF THE APPLICATION OF HUMAN FACTORS CONSIDERATIONS TO ALL PANELS IN THE PLANT, NOT JUST y.
| |
| r THOSE IN THE CONTROL ROOM. THIS IS PARTICULARLY IMPORTANT FOR. PANELS USED DURING INFREQUENT OR OFF-NORMAL CONDITIONS APPLICATION OF HuspgPBACTORS SHOULD SUCH AS REMOTE SHUTDOWN.
| |
| BE AN INTEGRAL PART OF THE DESIGN MODIFICATION PROCESS FOR ALL CONTROLS AND INDICATIONS.
| |
| e
| |
| | |
| r
| |
| _6-l t .
| |
| AS A MINIMUMr THIS SER SHOULD BE' REVIEWED BY PLANT ORGANIZATIONS RESPONSIBI? FOR OPERATIONS, TRAINING, INSTRUMENTATION AND
| |
| (~~)T
| |
| (_ CONTROLS, AND THE DESIGN MODIFICATION PROCESS.
| |
| l l ILLUSTRATIONS, WHICH MAY BE HELPFUL IN UNDERSTANDING THIS SER, ARE BEING TRANSMITTED BY TELECOPY TO THE UTILITY AND PARTICIPANT SEE-IN CONTACTS. RECIPIENTS WHO DO NOT HAVE TELECOPY RECEPTION CAPABILITIES AT THEIR LOCATION CAN OBTAIN A COPY OF THE ILLUSTRATIONS FROM THEIR SEE-IN CONTACT OR JEFF WHEELOCK, INPO, 404/951-4730. RECIPIENTS WITH TELECOPY RECEPTION CAPABILITIES WHO EXPERIENCE PROBLEMS IN RECEIVING ANY TRANSMISSION SHOULD CONTACT SKIP HEEKE, INPO, 404/953-7675.
| |
| I'iPO'S EVALUATION OF THIS EVENT IS COMPLET?.
| |
| LIMITED DISTRIBUTION ALL COPYRIGHT 1986 BY THE INSTITUTE OF NUCLEAR POWER OPERATIONS.
| |
| RIGHTS RESERVED. NOT FOR SALE. UNAUTHORIZED REPRODUCTION IS A VIOLATION OF APPLICABLE LAW.
| |
| REPRODUCTION OF NOT MORE'THAN TEN COPIES BY EACH RECIPIENT FOR ITS INTERNAL USE OR USE EY ITS CONTRACTORS IN THE NORMAL COURSE
| |
| ( - OF BUSINESS IS PERMITTED. THIS REPORT SHOULD NOT BE OTHERNISE
| |
| \ TRANSFERRED OR DELIVERED TO ANY THIRD PERSON, AND ITS CONTENTS SHOULD NOT BE MADE PUBLIC, WITHOUT THE PRIOR AGREEMENT OF INPO.
| |
| Information
| |
| | |
| ==Contact:==
| |
| RICHARD H. REYNOLDS, INPO, 404/953-5392
| |
| | |
| i 7 N\SS-b l DEVIATION INVESTIGATION REPCRT
| |
| ]
| |
| l TITLE 10 STEAM GENERATOR PORV IM501BD INOPERABLE PaGE ,
| |
| 1 10Fl 0 I .l EVENT DATE DIR MER DEPORT DATE
| |
| !'h MONTH DAY YEAR ,lIL LHIT YEAR j/j/ SEQUENTIAL
| |
| // MER
| |
| // REVISION ff MER MONTH DAY YFjR OPERATING POWER ois 111 siti 016 Oli sis -
| |
| 01 71 4 -
| |
| 010 06 2:1 gi e LEVEL 01 91 e ;
| |
| CONTACT FOR THIS DIR NAME TELEPHONE M ER AREA CODE Alex Javorik. Assistant Tech Staf f Suoarvisor Ext. 2106 COMPLETE ONE LINE FOR EACH COMPONEN 8l1I5 A URE DESCRIBED IN THIS REPORT 21314l -I5l4I4l1 l CAUSE SYSTEM COMPONENT MANUFAC-. REPORTABLE CAUSE SYSTEM COMPONENT MANUFAC- REPORTABLE l TURER TO NORDS TURER TO NPRDS x mis I I Pl S Gl 01 el 2 N I l I I I I !
| |
| I I I l l I i 1 1 I I l l
| |
| SUPPLE > ENTAL REPORT EXPECTEL "
| |
| l EXPECTED SUBMISSION
| |
| ~l YES fif ves enen1ste EXPECTED SUBMISSION DATEl ll NO TEXT A. PLANT CONDITIONS PRIOR TO EVENT:
| |
| Event Date/ Time 5/11/88 / 1130 O
| |
| G Unit 1 M00E 1 - Power Ooeration Rx Power 98% RCS [A8) Temperature / Pressure Normal Ooeratina Unit 2 MDDE .2 _ - Power Omeration Rx Powe*r 89% RCS [AB) Temperature / Pressure Normal Ooeratino B. DESCRIPTION OF EVENT:
| |
| On May 11, 1968, while operating at 98% reactor power, a Byron Station Equipment Operator performing rounds noted steam relieving from the 10 Main Steam (MS) [58] Power Operated Relief Valve (PORV). The Unit 1 Operator (licensed) was immediately informed of this anomaly and the ID PORV Manual / Automatic Control Station was placed in the Manual Closed position. With the valve handswitch now in the Closed position, the steam flow coming from the PORV outlet was noted by visual observation to have stopped completely.
| |
| I Limiting Condition for Operation Action Requirement (LCOAR) 180S 6.3-1 A was entered by the Operating Department personnel and at 1155 hours on May 11, 1988, the 1MS018D was isolated by closing the upstream manual isolation valve.
| |
| Nuclear Work Request (NWR) number 855825 was generated to investigate and repair the valve's automatic controls and Instrument Maintenance Department personnel were dispatched to investigate and repair the system. The " drift open" problem being experienced on the PORV was determined to be the result of a newly installed, defective. "Close Pres;ure Switch. 85-1". The function of this component is to ensure sufficient system hydraulic presrure so as to maintain the PORV in its fully closed position. However, with the failure of this pressure switch, coupled with the slight inherent internal leakage of associated positioner components, the system pressure bled down, over time, and was never restored by the hydraulic pump, thus resulting in the valve driftlng partially open.
| |
| 0023R/0003R)
| |
| | |
| M% %-(o
| |
| . OEV!AT10i. INVE$TIGATION REPORT (DIR) Form Pev 2.0 PACE Facility Name .
| |
| Braidwood 2 1 10F1 0 l3 Title Spurias Lifting of 2M5018A (2A Steam Gcnerator Power Operated Relief Valve) due to faulty pressure sw tch
| |
| !]/
| |
| and leaky nitronen fill valve.
| |
| EVENT DATE DIR NUMBER REPORT DATE !
| |
| ; SEQUENTIAL / REVISION STA UNIT YEAR y// NUMBER {//
| |
| f NUMBER MONTH DAY YEAR 3 TENTH DAY YEAR POWER RI a 26 0 01 2 ~
| |
| 11 El 7 -
| |
| 010 11 1 Ol 9 al R 91 01 0 of 9 21 7 Rt I CONTACT FOR THft DIR , , , _
| |
| '' TELEPHONE NUMBER gggg _,
| |
| AR D CODE Dan stroh. Technical staff Enaineer Ext. 2477 _ ,,,1,,L 1 1E 4 It Iaf.I2i B IOI1 COMPLETE ONE LINE FOR EACH COMPONE URE DEt Q(( Q TN THis REPORT CAUSE $75 FEM COMPONENT MANUFAC- REPORTAB'S CAU$E SYSTEM COMPONENT MANUFAC- REPORTABLE TO NPROS TURER 70 NPROS TURER i 11 MI v I I 1 1 1 I I c si a al il sf G Y I I I I I I I I i l I I SUPPLEMENTAL REPORT EXPEETED t9LH T DAY YEAR _
| |
| susMrss10N DATE Y NO l l l l YEs rif yet. eamniate EXPECTED SUBMftsf0N DATED TEXT Energy Industry identification System (E115) codes are toentified in the tent as [XX)
| |
| A. Plant conditi fts Prior to Event:
| |
| Unit: Bratewood 2: Event Date: 9/27/88; Event Time: 0456
| |
| ;O. Mode: 1 - Hot Standby: Rx Power: 0*.
| |
| RC5(AB] Temperature / pressure: 557'C/2238 psig
| |
| : 8. Deter *otion of Event:
| |
| On Segemeer 27. 1988, the shift foreman and a B man were performing rt.it ss .m yections in the main steamline isolation salve room (MSIV room). For no apparent reason the S " tam Generator Power Operated Relief Valve (PORV) lifted momentarily. The Unit 2 Steam Generator pressure was stable at 1085 pst and Reactor Coolant System Temperature and Pressure had been stacle for several hours. The ZA PORV was isolated and declared inoperable and Limiting Condition for Operation Action Requirement (LC0AR) 6.3 lA was entered. Nuclear Work Request (NWR) A25Ba0 was written to troubleshoot and repair the preolem.
| |
| Steam pressure rematned stable througnout the event and the Manual Isolation Valve Upstream of the PORV was closed to isolate the inoperable PORV.
| |
| 2345m(111588)/30
| |
| | |
| j{ u. ~_ ,. . d _ .i . , _ _
| |
| . ?
| |
| DEVIATION !%VESTICAT!DN REPORT TEXT CONTI%UATION Form aev 2.0 Die quMafe #Act FACILITY NAME
| |
| " I Braidwood 2 4 Legate NUMeER j
| |
| f]) c (v
| |
| STA UNIT YEAR cF 0 11 21 o of 2 al a -
| |
| 11 51 7 -
| |
| oI o 2 TEXT .:,
| |
| C. cause er Evect:
| |
| The cause of the event can be traced back to Friday 9-23-88. Alarm 2-15-A10. "2A PORY TROUBLE" came up and LCOAR 6.3-1A was entered. NWR 4A25809 was written to investigate the cause. Because the 2A PORV was needed for the 100% unit trip test Friday night, the Startup Test Engineer (STE) was asked to troubleshoot the valve. He found both nitrogen pressure and hudrat1te ett level good. He concluded that the nitrogen low pressure switch was faulty and giving the alarm. A meeting was held between representatives of the cuality l control (DC). department. operating (ops) department and the startup group to determine a way to repair the valve before the scheduled trip on Friday night. It was agreed that an operability surveillance would be run and if successful, the valve would be placed back in service even though the alarm was still up. The survat11ance was successfully performed. NWR WA25809 was closed out, and LCDAR 6.3 1A was exited at 1323 on
| |
| ~
| |
| 9-23-88. .
| |
| I During the 1007 trip test. the 2A PORV did lif t as reoutred and alarm 2-15-A10 "PORY TROUSLE" reset and stayed clear. The ZA PORV was as a result.1ef t in service and LCOAR 6.3-1A was not re-entered.
| |
| NWR On Tuesday 9 27-88 at Da56. 2A PORV lif ted annunciator 2-15-A10 came in, and LC0AR 6.3-I A was entered.
| |
| 4A258a0 was written to investigate.
| |
| Upon troubleshooting, it was found that the pressure switch was faulty and there was a nitrogen leak at the pneumatic fill valve. Apparently between Saturday 9-2a-88 and Tuesday 9-27-88, the system nitrogen had bled out and the pressure switch had not annunciated. On Tuesday, the PORV lif ted momentarily or puf fed because
| |
| ' /*
| |
| of this low nitrogen pressure condition, and was not recogntred Decause of the faulty pressure switch. Root
| |
| ( cause of event was a f aulty pressure switch in cometnation with a leaking nitrogen fill valve.
| |
| D. safety analysts:
| |
| LCDAR 6.3-1A was entered in a timely manner af ter the trouble alarm came in on both Friday 9-23-88 and
| |
| ' Tuesday 9-27-88. However, when the valve was declared operable on Friday routine surveillance should have been performed on the valve to verify proper pressure wntle the pressure switch was inoperacle. If the valve had been opened anytime between Saturday and Tuesday and had received an emergency Closure signal, there would not have been enough nitrogen pressure in the accumulator to quick close the valve. Because the nyaraulic system was still fully cperante at all times. the valve could have been slow closed using the rnanual/ auto station in the main control room. The additional time reoutred to close the valve using *.ne normal mooulation controller would not have any serious impact on steam 1tne pressure and no unusual safety concerns would have resulted f rom either the f ailed pressure switen or leaking fill valve. Even if the hydraulic system had f ailed, the PORV could have been manually positioned using a locally mounted hand pump which is connected to the PORV actuator. Also the 2A PORV could have been manually isolated at any time.
| |
| (
| |
| b 23a$m(110288)/31 4
| |
| -_-_____mm_.m___ _ _ _ _ - _ _ . _ _ __
| |
| | |
| DEVIATION INVESTIGATION REPORT TEXT CONTINUATION
| |
| -1" Form Rev 2.0 ctR NUMBER PACE
| |
| ,g
| |
| ; FACILITY NAME
| |
| 'j .
| |
| Braidwood 2 STA UNIT YEAR guMaER MUMBER
| |
| . h-} ...
| |
| 21 e of r al a 11 El 7 "-
| |
| 01 0 1 0F 0 I1 TEXT.
| |
| E. Corrective Action:
| |
| Irenadiate corrective action consisted of isolating the ZA PORY and entering LCDAR 6.3-1 A. NWR *A25840 was written to troubleshoot and repair the leaking pneumatic fill valve. NWR #A25874 was written to recal/ repair the nitrogen pressure switch.
| |
| F. Provinut occurrence:
| |
| None G. r - anent Failure nata:
| |
| Manufacturer Nomenclature Model Number Part Number Sorg Warner Schrader Fill Valve N/A 38938000 i
| |
| \
| |
| l 2345m(1102881/32
| |
| | |
| 7.
| |
| .BRAIDWOOD SIMULATOR MALFUNCTION i
| |
| | |
| ==Title:==
| |
| ~ Steam Dump-Cooldown Valves Control Failure , ID: MSS-7 NO: 6.3.4.8.7
| |
| ( ). '
| |
| | |
| == Description:==
| |
| . Input to cooldown valve controller !
| |
| 1PK507, fails.to pre-selected value.
| |
| Manual control possible. (Failure of controller card.)
| |
| Variations: None. Date: 1/6/88-Rev: 3 Selectable Steps Inputs Comments
| |
| : 1. Select fail position 0-100 percent Percent open --
| |
| Malfuntion recognizes.the' interlocks of the steam dump system.
| |
| >3
| |
| (- -) -2. - Select ramp time 0-99,999 see Brief Plant Response: [ Based on plant at hot standby in steam pressure mode on the steam dumps)
| |
| Note: Steam dumps must have an arming signal in order for the malfunction to affect the steam dumps.
| |
| When the failure occurs, the cooldown valves will position as selected in step No. 1 above.
| |
| l 0
| |
| 0125w:4 311M/85M/13 1/88 j 1
| |
| | |
| ==Title:==
| |
| . Steam Dump'Cooldown Valves Control Failure ID: MSS-7 C,
| |
| t
| |
| .N Brief Plant Response (continued):
| |
| If failed cidsed, RCS temperature will increase until the next set of steam dump valves opens to control temperature. Varicus temperature alarms will sound and Tavs will stabilize at a higher temperature on the S/G PORV's.. If failed open, RCS temperature will decrease, causing various low temperature alarms including T,y, CONT DEV HIGH. Pressurizer pressure and level decrease, actuating their c.ontrol systems and alarms. S/G 1evels decrease.
| |
| When Tavg drops below P-12, the steam dump valves close, stopping the cooldown. When RCS temperature increases above P-12, the dump valve re-opens and the cooldown starts again. This cycle will continue until the operator takes manual control.
| |
| Suggested Instructor Action:
| |
| None.
| |
| Events: None O 311M/85M/14 1/88 0125w:4 I 1
| |
| l
| |
| - _ _ = _ _ _ _ - _ _ _ _ _ _ - _. -_ i
| |
| | |
| BRAIDWOOD SIMULATOR MALFUNCTION. ]
| |
| i i
| |
| Titles Steam Dump Control Failure ID: MSS-8
| |
| ' [' NO:_ 6.3.4.8.8 AJ/
| |
| | |
| == Description:==
| |
| Failure of any of the E/I converters for the four groups of steam dump valves.
| |
| Date: 1/6/88
| |
| ' Variations: None.
| |
| Rev: 4 Selectable Steps Inputs Comments 1
| |
| : 1. - Select faulty steam MSS-8A- MSS-8A-Cooldown valves (A, E, J)
| |
| MSS-8B MSS-8B-Group 2 (B, F, K) J MSS-80 MSS-8C-Group 3 (C, G, L)
| |
| MSS-8D MSS-8D-Group'4 (D, H, M)
| |
| : 2. Select fail position 0-100 percent 100 percent - Full open
| |
| : 3. Select ramp time 0-99,999 sec
| |
| 'l Brief Plant Response:
| |
| Note: All steam dump interlocks are functional during this malfunction.
| |
| If failed to zero, selected group will not open when required. RCS temperature vill increase, causing steam pressure to increase and possibly actuate the S/G PORV's.
| |
| O 0125w:4 311M/85M/15 1/88 l
| |
| d_=_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ ___
| |
| | |
| BRAIDWOOD SIMULATOR MALFUNCTION 1
| |
| 'i l-l /~N
| |
| | |
| ==Title:==
| |
| Steam Dump Control Failure ID: MSS-8
| |
| .b.
| |
| Brief Plant Response (continued): )
| |
| 'If failed open, steam flow increases, causing Tavg to decrease, pressurizer pressure and level to decrease, and S/G 1evels to swell. When temperature drops below permissive P-12, the. steam dump valves close, terminating the
| |
| * widown. As RCS temperature increases, P-12 clears, and the steam dumps again open, starting-the cycle all over again. The first annunciator received is T,y, CONT DEV HIGH.
| |
| Suggested Instructor Action:
| |
| None.
| |
| Events: None
| |
| )
| |
| 0125w:4 311M/85M/16 1/88 l
| |
| | |
| BRAIDWOOD' SIMULATOR MALFUNCTION-h[
| |
| v 1r
| |
| | |
| ==Title:==
| |
| Stuck Steam Dump Valve ID: MSS-9.
| |
| ;[ F NO: 6.3.4.8.9 V.
| |
| | |
| == Description:==
| |
| After the valve has opened or closed l
| |
| oto the selected " stuck" position, it
| |
| -will not re-close. This is indicative of a " mechanically" stuck valve.
| |
| . Variations: None Date:- 1/6/88-Rev: 4 Selectable Steps Inputs Comments
| |
| : 1. Select number MSS-9A, MSS-9B, Total of 3 steam dump valves of stuck valves MSS-9C may be stuck.
| |
| : 2. Select sticky valve ~1-12 1-MS004A 5-MS004E. 9-MS004J
| |
| - /~' 2-MS004B 6-MS004T 30-MS004K
| |
| '- 3-MS004C 7-MS004G '11-MS004L.
| |
| 4-MS004D 8-MS004H 12-MS004M
| |
| : 3. Select " stuck" position 0-100 percent Percent open Brief Plant Response: (IC-17 with load rejection)
| |
| Note: The steam dump interlocks will n21 affect the selected valve's position.
| |
| During a transient which requires the use of the steam dumps, when the selected dump valve opens or closes to the specified position, it will not
| |
| -re-close when required to by the control system. This will cause a continued 0125v:4 311M/85M/17 1/88
| |
| | |
| BRAIDWOOD SIMULATOR MALTU:lCTION-
| |
| | |
| ==Title:==
| |
| Stuck Steam Dump Valve ID: MSS-9
| |
| )s.
| |
| i' -
| |
| .Brief_ Plant Response: (continued)-
| |
| .cooldown of the RCS with decreasing Tavg, pressurizer pressure and level, and' possibly S/G 1evel. If at < 15 percent power, nuclear power will increase-approx. 3-4 percent. If at < 15 percent with the rod control system in manual, RCS temperature will continue to decrease, even after permissive F-12 has actuated. A reactor trip on low pressure could occur. No annunciators are received directly related to this malfunction.
| |
| Suggested Instructor Action:
| |
| When instructed by-the student, close the locally-operated isolation valve for the stuck steam dump, using LOA MSS 40-51.
| |
| Events: None ,
| |
| 1-(
| |
| 0125w:4 311M/85M/18 1/88
| |
| : y. . - . .
| |
| q BRAIDWOOD SIMULATOR MALFUNCTION
| |
| /N
| |
| | |
| ==Title:==
| |
| Main Steam Header Steam Leak ID: MSS-10
| |
| ?
| |
| D)-. NO: 6.3.4.8.10
| |
| | |
| == Description:==
| |
| Steam leak of specified size in main steam header (isolable).
| |
| Variations:
| |
| Date: 4/8/89 Rev: 3 Selectable Steps Inputs Comments
| |
| : 1. Select size of steam 0.12 x 10 6 leak Ib/hr 100 percent steam flow 6
| |
| approx. 15 x 10 lb/hr
| |
| : 2. Select ramp time , 0-99,999 see f%
| |
| k Brief Plant Response: [ Based on plant at full power when the break occurs)
| |
| When the steam leak occurs, steam flow from all steam generators increases.
| |
| RCS temperatures drop, pressurizer pressure and level rapidly decrease. Steam pressure drops causing a main steamline isolation and safety injection.
| |
| When the main steam isolation valves close, steam flow from all steam genera-tors decrease to zero, indicating that the leak has been isolated. The first-annunciators received are S/G FLOW MISMATCH W FLOW LOW.
| |
| Note: This malfuntion can also be used to simulate steam leaks in the secondary during a plant startup.
| |
| i Suggested Instructor Action:
| |
| None.
| |
| < I
| |
| 's Events: None 875M/5 5/89 l
| |
| 1 l' _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _
| |
| . . _ . _ , _ . . _ 3
| |
| | |
| BRAIDWOOD SIMULATOR MALMNCTION-3
| |
| | |
| ==Title:==
| |
| Steam Flow Detector Failure ID: MSS-11 I f}
| |
| AI NO: 6.3.4.8.11 Description Detector failure causes'arroneous 1
| |
| signal to be sent to SGWLC.-
| |
| I Variations: None. Dates- 4/8/89 Rev:
| |
| 6 Selectable Steps Inputs Comments a
| |
| : 1. Select transmitter MSS-11A - MSS-11H MSS-11A = FT512 'l MSS-11B = FT513 MSS-11C = FI522 MSS-11D = FT523' MSS-11E = FT532 ,
| |
| MSS-11F = FT533 MSS-11G = FT542
| |
| (
| |
| MSS-11H = FT543 ,
| |
| a l
| |
| ~2. Select failure mode 0-100 percent ]
| |
| l I
| |
| : 3. Select ramp time 0-99,999 sec l
| |
| (
| |
| Brief Plant Response: (IC-17, 100%, all systems in automatic) 4 I
| |
| If detector fails high, it sends a signal to SGWLC demanding an increased feedwater flow. This will result in turbine trip on HI-2 water level. If above P-8 When turbine trips, the Ex trips. The first annunciators received l include S/G FLOW MISMATCH STM/FW FLOW LOW and FW PUMP NPSH LOW.
| |
| i l
| |
| ba 875M/6 5/89 j
| |
| i L .. _ .. _ _ _ _. . _ . . . _ _ - _ . . _ _ . __ .
| |
| | |
| BRAIDWOOD SIMULATOR MALFUNCTION ID: MSS-11 f')
| |
| 9 /
| |
| | |
| ==Title:==
| |
| Steam Flow Detector Failure Brief Plant 11tesponse (continued):
| |
| If detector fails low, it sends a signal to SGWLC demanding a decreased feedwater flow. This will lead to a reactor trip on LO-2 water level in the steam generator. The first annunciators received include S/G FLOW MISMATCH STM/FW FLOW LOW and S/G LEVEL DEVIATION HIGH LOW.
| |
| Suggested Instructor Action:
| |
| When told to repair the detector, clear the malfunction.
| |
| )
| |
| Events: I
| |
| : 1) DVR 20-01-88-128: Loss of FI-532.
| |
| O .
| |
| 4 O 875M/7 5/89 1
| |
| l
| |
| -_m_____ ,.__m . _ . _ , . . , , . , _ , ,
| |
| | |
| ., .o M M- i/
| |
| DEV!ATICN INVESTIGATION REPO2T PACE TITLE Loss of Steam Flow Indication / Control on 1C Steam Generator Due to Personnel Error 6 /
| |
| DIR NUMEER REPORT DATE j EVENT DATE SEQUENTIAL REVISION y// NUMBER y// NUMBER MONTH D AY ' YEAR MONTH DAY YEAR STA UNIT YEAR POWER al a 21 o Of 1 al a -
| |
| 1 1 21 s -
| |
| oIo al 7 01 3 si a of 71 5 of E 21 3 CONTACT FOR THIS DIR TELEPHONE NUMBER NAME AREA CODE Technical staff Enaineer Ent. 2660 Bi 1 15 41EIa1 12lB i 011 J e devle.
| |
| COMPLETE ONE LINE FOR EACH COMPONEN F URE DESCRIBED IN THIS REPORT CAUSE SYSTEM COMPONENT MANUFAC- REPORTABLE CAUSE SYSTEM COMPONENT MANUFAC- REPORTABLE TURER 70 NPPDS TURER TO NPRDS I I i ! I 1 l i I I I I I I
| |
| I I I I l l I I I I I I MONTH DAY YEAR SUPPLEMENTAL REPORT EXPECTED SUBMISSION I vES fif yet. comolete EXPECTED SUBMIS$10N DATEt l NO i TEXT A. PLANT CONDITIONS PRIOR TO EVENT:
| |
| Unit: araidwood 1 : Event Date: May 21. 19aa : Event Time: m MODE: _,L - Power Oceration . Rx Power:_211_; RCS (AB] Temperature / Pressure: 578 Deerees F/2215 esta
| |
| : 8. DESCRIPTION OF EVENT:
| |
| On May 23, 1988 at 0836 with the Plant in Power Operation a loss of IC Steam Generator Steam Flow [58) indication occurred on IFI-532 which was the controlling channel for steam flow / feed flow mismatch. Under the direction of the Station Control Room Engineer (SCRE) the Nuclear Station Operator (N50) took the necessary actions using IBwCA INST-2. 'While stactitzing the plant in manual for feedwater flow the Steam Flow indicator IFI-533 was lost approximately 1 minute after the loss of indication on IFI-532. With the loss of the second indicator all indication for steam flow on 1C Steam Generator was lost. During the loss of indication the Instrument Maintenance Department had been installing a Digital Oscilloscope in the Instrument racks for a retest of Startup Test BwSU FW-31. Calibration of Steam and Feedwater Flow f or Steam Generator IC.
| |
| The connection for the oscilloscope was removed about 1 minute after the loss of indication on IFI-533 occurred. Upon removal of the connector steam flow indication for IC Steam Generator returned. Stable plant conditions were obtained a few minutes afterward. No other operator actions were required.
| |
| C. CAUSE OF EVENT:
| |
| The root cause of the event was determined to be the improper installation of the Digttal Oset11oscope. During the installation of the scope for FW-31 the scope was installed such that the leads from the steam flow transmitters output circuits were connected to the grounded inputs of the scope. This connection grounded the channel causing a false indication of zero on the affected channels. The N50 stopped c.11 work at this time and the oscilloscope was disconnected from the instruments.
| |
| l t
| |
| \
| |
| 2189m(062988)/16 t
| |
| i
| |
| | |
| , DEVIATION INVESTIGATION REPORT TEXT CONTINUATION TITLE DIR Nt#GER PAGE Loss of Steam Flow Indication / Control on 1C Steam SEQUENTIAL REVISION Generator Due to Personnel Error STA UNIT YEAR NLDSEE NLDBER )
| |
| gy 21 0 01 1 al 8 -
| |
| 1 l2l8 -
| |
| 0l0 2 0F 0l2 TEXT D. SAFETY ANALYSIS:
| |
| . There was no effect on plant or public safety. The event was of short duration and stable conditions were obtained very quickly. Under worst case conditions with the plant at 100% power, with no operator intervention. !
| |
| the Steam Generator Water Level Control System (SGWLC) would have drastically reduced feed flow to the IC Steam Generator, ultimately re:sitia.ii ;n Lo-Lo level condition. At this point the Auxiliary Feedwater Pumps would have started and a Reactor Trip would have occurred as per design with no effect on plant or public safety.
| |
| E. CORRECTIVE ACTIONS:
| |
| The oscilloscope was removed from the instrument racks and the leads changed from the grounded inputs to the differential inputs on the scope. A training class was held to acquaint all Instrument Maintenance personnel of ,
| |
| this problem and its solution. No further action is required. ,
| |
| 1 F. PREVIDUS OCCURRENCES:
| |
| NONE G. CONPONENT FAILURF DATA!
| |
| O U
| |
| \'
| |
| t f
| |
| I l
| |
| ( )
| |
| l I
| |
| 2189m(061588)/17 I
| |
| l
| |
| | |
| BRAIDWOOD SIMULATOR MALFUNCTION 1
| |
| ' f~^4
| |
| | |
| ==Title:==
| |
| - Steam Generator Safety Valve Failure ID: MSS-13 NO: 6.3.4.8.13
| |
| | |
| == Description:==
| |
| S/G safety valve fails due to mechanical failure.
| |
| Variations: None. Date: 4/8/89 Rev 5 )i i
| |
| Selectable i Steps Inputs Comments {
| |
| : 1. Select S/G and valve MSS-13A MSS-13A = MS17A (S/G 1A)
| |
| MSS-138 MSS-13B = MS17B (S/G 1B)
| |
| MSS-13C MSS-13C = MS17C (S/G IC)
| |
| I MSS-13D MSS-13D = MS17D (S/G 1D)
| |
| : 2. Select valve position 0 - 100 percent
| |
| : 3. Select ramp time 0-99,999 see Brief Plant Response (3%, plant startup in progress)
| |
| Steam release causes a decrease in RCS pressure and temperature. This causes I a positive reactivitiy addition to the core. Rx power rises due to the increased steam flow. The first annunciator received was T,,, CONT DEV LOW.
| |
| , J 875M/9 5/89 I
| |
| | |
| m- - . . . . . _ _ . .
| |
| BRAIDWOOD SIMULATOR MALFUNCTION
| |
| | |
| ==Title:==
| |
| - Steam Generator Safety Valve Failure ID: MSS-13 Suggested Instructor Action:
| |
| Insure student carries out emergency procedure.
| |
| , Events: None i
| |
| O .
| |
| i
| |
| ~t 875M/10 5/89 I
| |
| _ = _ _ _ - - . . - - _ = _ _ _ _ . - - _ _ . _ _ _ - _ - . _ _ . . --
| |
| | |
| ! BRAIDWOOD. SIMULATOR MALFU!iCTIO:I
| |
| (-
| |
| | |
| ==Title:==
| |
| Steam Line Pressure Detector (PT-507)' Failure ID: MSS-14 NO: 6.3.4'.8.14 lV:
| |
| l
| |
| | |
| == Description:==
| |
| Pressure detector fails due to
| |
| . defective' circuit board in transmitter.
| |
| l' ' Variations: None. Date: 1/7/88 Rev: 3 Selectable Steps Inputs Comments
| |
| : 1. Select failed pressure 0-100 percent 0 percent = 0 psig 100 percent = 1300 psig
| |
| : 2. Select ramp time 0-99,999 sec-D^
| |
| 'U Br'ief Plant' Response:
| |
| Fails hiah -- Feed pump speed increases causing increase S/G water level causes a closing down of FWRV, cold feedwater entering S/G could cause a positive reactivity addition and subsequent S/G swell which could lead to a turbine trip and Rx trip if greater than P-8.
| |
| If in the steam pressure control mode, all 12 dump valves will open causing S/G swell and subsequent turbine and reactor trips.
| |
| Fails low -- Feed pump speed decreases causing S/G water level to decrease causing S/G lo-lo water level reactor trip.
| |
| =,
| |
| 0125w:4 311M/85M/25 1/88
| |
| | |
| BRAIDWOOD SIMULATOR MALFUNCTION' t
| |
| j
| |
| )
| |
| | |
| ==Title:==
| |
| Steam Line Pressure Detector (PT-507) Failure ID: MSS-14
| |
| .lY .
| |
| Suggested Instructor Action:-
| |
| None.
| |
| Events:
| |
| : 1. . DVR 6-1-85-256:. PT 507 Instrument Drift O
| |
| t I
| |
| l 0125w:4 311M/85M/26 8/86
| |
| | |
| 1 i
| |
| DEVIATION INVEST! GAT!0N REPORT
| |
| *act OTITLE Pftn7 natFTins tew 1 lori o 1 2 EVtut DATE Dia htmenta arpear natt
| |
| // SE00ENTIAL // REV!$10N nomfu nay vras tra- unti vtan ff stasate ff mLasate MDm?N BAY vtaa 1 POWER l al a 11 2 al i al s of 1 al 1 -
| |
| 2 la i s -
| |
| eia el e 21 s al i al el 1 !
| |
| coalfAff FOR TNft Die NaME trt r ==r = -- -- m AAEA CODE Mat *k== m - -t tut. 21a1 ai1 1E 211Ia1-1sIeIai1 ensepttra ama t rar aan raras ensepast 7 att unt hRttaftsa 13 Tutt ggpagy CAUSE SYSTEM CowgNENT MANUPAC- REPORTABLE CAUSE SYSTEM COW 0eENT MANUFAC- REPORTAttE v === fa apaat vaare yo apeng i I I I I I I I I I I i f I at tIa i pl al T 11 21 al a von I i i i i sumptissamfat atpaRT Exptrita momTN DAY Ytaa SUOMISSION
| |
| ~
| |
| $1yrtfirvon: emenista Expretta maaerstram nArti l no DATE ,l, ,l, J, TEXT O On August 12. during normal plant operation. Main Steam Pressure Transmitter 5e7 drifted low, affecting the Feetheater pump Master Speed Control, and causing excessive level deviations in the 1 A and 10 steam generators.
| |
| l The transmitter was replaced, but on August 17, the replacement transmitter 547 drifted low, aise, causing the same previously stated condition. Again, on August 19. the same replaced transm1G:;; drifted low. tdhen the Instrument Maintenance Department began their initial investigation, they unscrewed the cap on the transmitter.
| |
| At this peint, the transmitter failed cespletely. In all three instances the operators placed the Feedwater Pues Master Feed Controller in manual and rettered proper steam generator levels.
| |
| 64117 int THE aGAT CAusf7 The first pressure transmitter, which failed on August II, was sent to the manufacturer for a component failure review. ITT Barton, the manufacturer, will issue a report en their findings to Commonwealth Edison. A supplemental report will be filed at that time. The second transmitter which failed on August 17 reeutred the l replacement of the transmitter's circuit board. Instrument Maintenance then placed it back inte service. There I
| |
| were two main causes for the fa11ere editch occurred on August 19. First a loose wtre was detected on the strain page harness which caused the component to drif t low. The second probles was that the pressure transmitter cae l was making contact with a wire en the circuit board. The cap subsequently severed and caused the transmitter to i fati comp 16tely.
| |
| l The Instrument Maintenance Department has concluded that the transmitter cap, by making contact with cournis on the circutt board caused the failures on the second transmitter. The vendor was brought in to inspect N transmitter. The veneer also indicated that the contact of the transatttter cap and the streutt board caused excessive stress on the circuit board resulting.in cong9nent failure. The vender concluend that the contact between the board and the cap was caused by use of the wrong 0-ring between the cap and the transmitter.
| |
| 1 (0638M)
| |
| | |
| 1 l
| |
| \
| |
| l DEVIATION IWESTICATION REPORT TFXT CONTMATION f% }
| |
| (,)IITLE _ DIR ttMBER. __ _PAGE l SEQUENTIAL REVISION STA LNIT YEAR f4MBER f4MBER PT507 DRIFTING LOW . , , , ,,_
| |
| 0,[,1,_0[ 1 81 5 _2,LSj.J -
| |
| 0lL 2 0F, 0 jj TEXT HOW DID IT AFFECT PL#ff SID/OR PUBLIC SAFETY?
| |
| There was no effect on plant and/or public safety since the failuee did not effect the reactor protection systee HAS IT HAPPEMED BEFORE?
| |
| On Neust 8,1985, pressure transseitter 507 drif ted low. The transmitter was recalibrates and placed back into service. On August 10, 1985, PT507 drifted low again. Instrument Maintenance replaced the transmitters circuli:
| |
| board sent then placed the transmitter back into service.
| |
| l_ etat _WAS D(DIE _TD CORRECT THE _Cofet,TI_(Dt MID HERf AltE IdE GOM TO PREVEPg RfCURRENFE7 The proper sized 0-ring was installed on the transmitter and no problems have occurred since. It is believed that this use of the wrenq 0-ring is an isolated occurrence since this problem has not occurred on other transmitters of the same model.
| |
| e t
| |
| k b
| |
| \) - - - -
| |
| _ _=_ _- . . - - - - - _ _.-
| |
| m--- _ - _ _ . . _
| |
| | |
| BPJLIDWOOD SIMULATOR MALFUNCTION l
| |
| /~w..
| |
| | |
| ==Title:==
| |
| MSR Relief Valve Failure ID: MSS-15 NO: 6.3.4.8.15
| |
| | |
| == Description:==
| |
| MSR relief valve failure due to mechanical failure l
| |
| Variations: None. Date: 8/22/86 Rev: 3 .]
| |
| I l
| |
| Selectable Steps Inputs Comments l
| |
| l
| |
| : 1. Select valve -MSS-15A - MSS-ISA = LP relief at 250 lb MSS-15F MSR 1A MSS-ISB = LP relief at 256 lb MSR 1A-MSS-ISC = LP relief at 265 lb MSR 1A MSS-15D = LP relief at 250, Ib MSR 18 MSS-15E = LP relief at 256 lb MSR 1B l MSS-15F = LP relief at 265 lb MSR 1B
| |
| : 2. Select fail position 0-100 percent open .
| |
| : 3. Select ramp time 0-99,999 sec j
| |
| I .
| |
| n\J 0125w:4 311M/85M/27 8/86 l.
| |
| | |
| BRAIDWOOD SIMULATOR MALFUNCTION
| |
| | |
| ==Title:==
| |
| MSR Relief Valve Failure ID: MSS-15
| |
| - (, -- -
| |
| Brief Plant Response:
| |
| -Increased steam flow through MSR with reduced steam flow to LP turbine. Gives up-power reactor transient (appears as a small steam line break).
| |
| Suggested Instructor Action:
| |
| None.
| |
| Events: None O
| |
| ~
| |
| 0125w:4 311M/85M/28 8/86 l
| |
| C. _ _ _ _ _----_------.-.-.-.--- - - _
| |
| | |
| BRAIDWOOD SIMULATOR MALFUNCTION t .
| |
| [ )
| |
| | |
| ==Title:==
| |
| Turbine Driven MFP Control Valve Failure ID: MSS-16 s.
| |
| NO: 6.3.4.8.16
| |
| | |
| == Description:==
| |
| . Partial or total loss of the steam supply to the turbine driven MFPs.
| |
| l Variations: None. Date: 9/20/88 Rev: 4 i-Selectable Steps Inputs Comments
| |
| : 1. Select NFP . MSS-16A - MSS-16A "B" MFP high pressure MSS-16E steam control valve (CV)
| |
| MSS-16B "B" MFP low pressure CV MSS-16C "C" MFP HP CV MSS-16D "C" MFP LP CV MSS-16E - B and C MFP air-operated
| |
| -[ /}
| |
| N ~. check valve
| |
| : 2. Select fail 0-100 percent 0 = closed position 100 = open
| |
| : 3. Select ramp ting 0-99,999 see Brief Plant Response:
| |
| A and C -- Failure of high pressure control valve will cause an entire loss of feedwater flow to S/G causing a lo-lo S/G level Rx trip if at low power. At high power, no effect.
| |
| '0125w:4 311M/85M/29 9/88
| |
| | |
| BRAIDWOOD SIMULATOR MALFUNCTION j''
| |
| | |
| ==Title:==
| |
| Turbine Driven MFP Control Valve Failure ID: MSS-16
| |
| -Brief Plant Responte (continued):
| |
| 4 B and D -- Failure of low pressure control valve will cause a large reduction in feedwater flow, possible Rx trip. Failure open could cause high S/G water level or MFP trip.
| |
| E -- Slow leak off of instrument air pressure will allow check valve to slowly stop LP steam Suggested Instructor Action:
| |
| None.
| |
| Events:
| |
| O 1) DVR 6-1-85-120: EHC Solenoid Valve F..ilure.
| |
| i
| |
| : 2) DVR 06-02-88-044: Low Pressure Governor Valve Oscillations.
| |
| O 0125w:4 311M/85M/30 9/88 t________.______ _ _ _ . .
| |
| | |
| P5.)-/L DEV!ATICN INVEST!0ATICN REPCRT eAct i yrt 3 ~,
| |
| 1 'i
| |
| [iTLEIn FtEDWAftt PUMP WIGN PRE 11uRE STOP VALVE CLOSURE ,
| |
| Oft quMata REPORT DATE l l _df DATE OMRATM j
| |
| // $[00ENTIAL // REVISION t VfAR STA UNIT YEAR WUMe[R NUMAff MONTH DAY YEAR l MONTH DAY POWER
| |
| # I' of a 21 2 al 5 of a of t at t 1 1 z! o -
| |
| ol 1 0 17 118 st 1 of 21 o CONTACT FOR THft oft j TrttPHour quMarn l
| |
| NAME __ ]
| |
| l AREA C00E j nichard M. Williana rut. 21st ai1 l t 2l 1 Ie!-l 1IeIeIt c0MPtrTr our timr For EACM COMPont T tar oftfatarn fu Twit atpagt CAU$E ST$ TEM COMP 0hENT MANUFAC- REPORTA8LE CAU$E SYSTEM COMPONENT MANUFAC- REPORTABLE Tuare 70 NPa01 Tunta To great r aIJ I I al v ut el el e Y I I I I I I I I 1 1 1 I I I I I I I l SUPPLEMENTAL REPORT EXPECTED MONTH DAY YEAR
| |
| $USMIS$10N
| |
| ~
| |
| DATE l Yrs (if ven enmeista trPECTrn e nMittfou CAft1 SI no l l l TEXT WAT HAPPENED 7 1e operating the 18 Turbine Ortven Feedwater Pump, the High Pressure Stop Valve drif ted closed.
| |
| WAT WAS THE Roof CAUSE?
| |
| The EHC Solenoid valve associated with th's 15 Feedwater Pump High Pressure Stop Valve was found to be defective and improperly oriented.
| |
| HOW ofD IT AFFtti PLANT AND/0e puntfC 1AFETY?
| |
| The Closure of this stop valve effectively eliminated the pumcing Capabtlity of this pumo. Another Feedwater Pump was running and prevented a loss of feedwater reactor trip. Thus, there was no impact on plant or public safety.
| |
| HAS ff NAPPthtD REFORE?
| |
| Yes, on two previous occasions the same stop valve has drif ted closed. (LER 885-039-00. DIR #85-098-00)
| |
| WHAT Wat DONE TO CORRECT THE CON 0fff0N AND WOW ARE WE 00fMO TO PREVENT stCUpstNCE?
| |
| The defective solenote was P.ot re-oriented. It was determined that the solenoid spring did not supply enough fcree to maintain valve position. A new solenoid valve with a stif fer spring was installed in the original pisttion. This modification has eliminated the valve drif t and the pump has been run tested without incicent.
| |
| l l
| |
| e s
| |
| e (063tM)
| |
| | |
| lh $ 3 - I fg DEVIATION INVESTIGATION RfPORT TITLE PAG U-2 LOAD REDUCED DUE TO 28 A@ 2C FEEDWATER PUMP LOW PRESSURE GOVERNOR VALVE OSCILLATIONS 0 EVENT DATE DIR NUPEER REPORT DATE ,
| |
| p l ff SEQUENTIAL f REVISION MONTH _DfUlfA!L STA UNIT YEAR '/
| |
| / NUP9ER {/'/ NUPSER W TH DAY YEAR 1 F0WER .
| |
| 0 14 01 8 81 8 01 6 01 2 81 8 -
| |
| 0 14 I 4 -
| |
| 010 0 ,5 23g qE N g,i3 EONTACT FOR THIS DIR NAME TELEPHONE NUPSER AREA CODE Don Brindle. Doeratino Enaineer Ext. 2218 8I115 21314I-l514l4l1 COMPLETE ONE LINE FOR EACH COMPONEN A URE DESCRIBED IN THIS REPORT CAUSE SYSTEM COMPONENT MANUFAC- REPORTABLE CAUSE SYSTEM COMPONENT MANUFAC- REPORTABLE TURE L TO NPRDS TURER TO NPRDS x JlJ l i IV MI 41 21 3 Y l l l l l l l l
| |
| 1
| |
| _ l l l i I l l l l l l l SUPPLEMENTAL REPORT EXPECTED MONTH DAY YEAR p
| |
| SUBMI5$ ION '
| |
| x I YES fif ven. complete EXPECTED SUBMISSION DATE) l ND 0 4 0 1 8 9 TEXT A. PLANT C0teITIONS PRIOR TO EVENT:
| |
| /% Event 1 Dato/ Time 4/8/88 / 1645 d
| |
| Event 2 Date/ Time 4/9/Bd / 0630 Unit 2 MODE (Prior to Event 1) 1 - Power Operations Rx fower 94% RCS (AB)
| |
| Temperature / Pressure Normal Ocaratina Unit 2 MODE (Prior to Event 2) 1 Power Operations Rx Power 80% RCS (AB)
| |
| Temperature / Pressure Normal Doeratino B. DESCRIPTION OF EVENT:
| |
| ErtnL1 - At 1645 on 4/8/88 the 2C Feedwater Pump (FW) ($J) began experiencing low pressure governor valve oscillations. During this event these oscillations increased in intensity, reaching a peak at a frequency of approximately 2 oscillations per second. Each oscillation cycled the valve about 3 inches open and closed causing speed and flow control problems.
| |
| At 1703 a load reduction was coseenced at 2 MW/ Min in the event that a Feedwater Pump trip became necessary. At 1806 with conditions worsening, the load reduction was increased to 4 MW/ Min.
| |
| At 1810 conditions had deteriorated to the point where the 2C FW Pump was unable to maintain speed or load. The load reduction was again increased, this time to 10 MW/ Min. The manual steam isolation was opened for the high pressure governor valve and at 1817 stone was admitted to the 2C FW pump high pressure l
| |
| turbine. The 2C FW pump became stable and began to pick up load. The load reduction was reduced to 3 MW/ Min. The valve oscillations had damaged a control linkage on the low pressure governor valve, so, as a precaution, the steam supply to the low pressure governor valve was manually isolated.
| |
| 1.0002R/0001R)
| |
| | |
| DEVIATION INVESTIGt, TION REPORT TEXT CONTINUATION TITLE DIR NUSER PAGE
| |
| ~
| |
| . SEQUENTIAL REVISION U-2 LOAD REDUCED DUE TO 28 AND 2C FEEDWATER PUMP STA UNIT YEAR NUMER NUMER LOW PRES $UWE GOVERNOR VALVE OSCILLATIONS 01 6 01 2 81 8 -
| |
| 014l4 -
| |
| 0 l0 2 0F 0l1 TEXT B. DESCRIPTION OF EVENT: (Continued)
| |
| At 1831 with plant conditions now stabillied. the load reduction was stopped. Preparations were also made for starting the ZA Motor Driven FW Pump. At 1840 the ZA FW pump was started and the speed was reduced on the 2C FW pump. At 1842 with the 2A FW pump picking up the load, the 2C FW pump was tripped and placed on its turning gear.
| |
| Event 2 - Approximately 15 hours af ter the 2C FW pump failed, the 2B FW experienced oscillations on its low pressure governor valve. At the time of Event 2, the unit was being ramped down at 2 MW/ Min to reach a target value of 585 MWE for Main Turbine repairs and to perform a required valve test.
| |
| At 0645 on 4/9/88 the 2B FW pump low pressure governor valve began oscillating violently. At 0648 the lead reduction was increased to 10 MW/ Min. Attempts were made to stabilize the pump using the high pressure governor valve with little success.
| |
| At 0652 the load reduction was increased to 15 MW/ Min to bring the unit down to approximately 60% power.
| |
| At 0657 the 2B FW pump was tripped and feed flow conditions stabilized. Steam supplies were manually isolated to both governor valves and the 2B FW pump was placed on turning gear.
| |
| C. CAUSE OF EVENT:
| |
| C The cause of events 1 and 2 appear to be failed servo valves on the low pressure governor valves. The 2C FW pump low pressure governor valve was damaged by the inten e valve oscillations. A control linkage was broken and the low pressure governor valve was jammed partially open. The extent of the damage to the governor valve itself has yet to be determined.
| |
| The 2B FW pump had only minor damage consisting of two stripped setscrews on the governor valve linkage.
| |
| The servo valve was replaced the next day as were the stripped setscrews. Operational Analysis Department (OAD) personnel examined the electrical signals from the DEH controller to the servo valve and found them to be satisfactory. Westinghouse personnel were present for linkage rod adjustments as well as pump startup. At 1424 on 4/10/88 the 28 FW pump was placed in service. The pump start was nonnat and the Unit was ramped up during the remainder of the day, without incident.
| |
| D. SAFETY ANALYSIS:
| |
| Thsre were no safety consequences as a result of this event. No safety systems were initiated and the contrdled runbacks precluded reactor trips. All systems performed as designed during these events.
| |
| E. CORRECTIVE ACTIONS:
| |
| 'The failed servo valves have been sent out for analysis by the manufacturer. At the time of this report results of this analysis have not yet been finalized. Action Item Record (AIR) 88-069 is tracking this analysis.
| |
| A supplemental report to this DVR will be written to document cause, if determined, and component,f ailures p after analysis. .
| |
| 'v)
| |
| L0002R/0001R) 1I l
| |
| I E___--_____---- )
| |
| | |
| DEVIATION INVESTIGATION REPORT TEXT CONTINUATION TITLE DIR NUteER '
| |
| PAGE SEQUENTIAL REVISION i U-2 LOAD REDUCED 00E TO 28 AND 2C FEEDWATER PUMP STA UNIT YEAR NUMBER NUteER LOW PRESSURE GOVERNOR VALVE OSCILLATIONS 0 0 8 8 -
| |
| 0l414 -
| |
| 0 10 3 0F 0l3 TEXT F. PREVIOUS OCCURRENCES:
| |
| Previous Servo Valve failures have been documented in the following DVR's.
| |
| DVR HUteER TITLE 6-1-87-106 (LER 87-019) Safety Injection and Reactor Trip from Low Steam Line Pressure due to Failed Main Turbine Throttle Valve During the Throttle Valve to Governor valve Transfer.
| |
| 6-2-88-026 (LER 88-001) Reactor Trip on 2C Steam Generator Low Level Due to a Feedwater Pump Trip and Failure of Digital Electrohydraulic Control System to Runback Turbine 6-2-88-030 Unit 2 Derating Due to Electro Hydraulic (EH) System Pressure Problems G. COMPONENT FAILURE DATA:
| |
| a) MANUFACTURER NDPENCLATURE MODEL NUPEER MFG PART NUPBER Moog Servovalve Moog Model 76 1161
| |
| /'~N b) RESULTS OF NPRDS SEARCH:
| |
| No Moog valve failures were found as they are not usually reportable to NPRDS.
| |
| c) RESULTS OF NWR SEARCH 2 Several Moog Model #76 Servovalve failures have occurred other than those mentioned in this report.
| |
| 1tese valves have been replaced under the following Work Requests.
| |
| 7/15/87 NWR 247189 3/27/88 NWR #54493 8/13/87 NWR 448114 3/27/88 NWR #54495 9/16/87 NWR 249029 3/29/88 NWR #54549 2/22/88 NWR 4 53242 3/29/88 NWR #54551 2/29/88 NWR M 53566 4/09/88 NWR #54884 v
| |
| i L0002R/000tR)
| |
| | |
| BRAIDWOOD SIMULATOR MALFUNCTION >
| |
| p~ ; .
| |
| | |
| ==Title:==
| |
| MSIV Bypass Valve Failure ID: MSS-17 NO: 6.3.4.8,,17
| |
| ():
| |
| | |
| == Description:==
| |
| MSIV bypass, valve sticks mechanically
| |
| - due to scored stem. Valve fails as is.
| |
| Variations: None. Date: 8/22/86 Rev: 2 Selectable Steps Inputs Conunents
| |
| : 1. Select valve 1-4 1-MS 101A-2-MS 1015 3-MS 101C 4-MS 101D O Brief Plant Response:
| |
| No problem until a MS line isolation on signal which should prevent all steam flow from S/G's.
| |
| Suggested Instructor Action:
| |
| None.
| |
| F- ents: None i
| |
| 0125w:4 311M/85M/31 8/86 t-
| |
| | |
| BRAIDWOOD SIMULATOR MALFUNCTION.
| |
| | |
| ==Title:==
| |
| S/G Atmospheric Relief Valve Failure ID: MSS-18
| |
| ' ( ,) NO: 6.3.4.8.18
| |
| | |
| == Description:==
| |
| Selected S/G atmospheric relief valve fails open or closed due to a control circuit malfunction. No manual control available.
| |
| Variations: See MSS-6 for atmospheric relief valve Date: 8/22/86 failure with manual control available Rev: 1 Selectable Steps Inputs Coments
| |
| : 1. Select valve MSS 18A 18A "A" S/G relief MSS 188 188 "B" S/G relief MSS 18C 18C "C" S/G relief MSS 18D 18D "D" S/G relief
| |
| '{
| |
| L
| |
| : 2. Select fail position 0,1 0-Fail closed 1-Fail open
| |
| : 3. Select ramp time 0-99,999 see Brief Plant Response:
| |
| Fail closed - relief valve will not open under any circumstance. Fail open-total steam flow increases, tavs decreases and pressurizer level decreases.
| |
| The increase in steam flow will cause a 2-3% increase in Rx power.
| |
| Suggested Instructor Action:
| |
| 1solate the failed atmospheric relief valve by using the applicable LOA (MSS-36
| |
| * 39) when requested.
| |
| l * ,
| |
| () Events: None 0125w:4 311M/85M/31 8/86
| |
| _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ a
| |
| | |
| .'BRAIDWOOD SIMULATOR-MALFUNCTION LISTING NUCLEAR INSTRUMENTATION
| |
| -b]
| |
| .NIS-1 Source Range Channel Failure NIS-2' ,
| |
| Intermediate Range Channel Failure NIS-3. Power Range Channel Failure
| |
| .NIS Intermediate Range Channel' Gamma Compensation Failure
| |
| ;NIS-7 Power. Range Detector Failure s
| |
| WIS-8 Source Range Channel High Voltage Failure NIS-10 Inadvertent Fuel Loading.
| |
| NIS-11 Improper Detector Overlap
| |
| .NIS-12 Source Range Discriminator Failure
| |
| .NIS-15' Leak Into Cuide Tube for Incore Detector
| |
| . ,/-
| |
| 638M/263M/9 8/87
| |
| | |
| BRAIDWOOD SIMULATOR MALFUNCTION I
| |
| 1
| |
| ' /~'s
| |
| | |
| ==Title:==
| |
| Source Range Channel Failure ID: NIS-1 NO: 6.3.4.9.1
| |
| | |
| == Description:==
| |
| Simulates a failure of the level amplifier.
| |
| 1 Variations: None. Date 4/9/89 Rev: 6 Selectable Steps Inputs Comments
| |
| : 1. Select channel NIS-1A NIS-1A = Ch. 31 ]
| |
| NIS-1B NIS-1B = Ch. 32
| |
| : 2. Select failed 100 -106 eps 105 cps will result in SR channel value high flux trip i
| |
| i
| |
| : 3. Select ramp time 0 to 99,999 see iO Q .
| |
| Brief Plant Responses (IC-8, Rx S/U, channel fails high)
| |
| Failure will result in improper indication on failed source range channel and initially will also affect the corresponding startup rate indication. The first annunciator received is SR HI FLUX RK TRIP.
| |
| Note: Audio countrate circuit and scaler timer are not effected by this malfunction (level amplifier failure).
| |
| Suggested Instructor Action:
| |
| l l Clear the malfunction when channel has been repaired.
| |
| (
| |
| Events:
| |
| l - 1) LER 06-01-86-002: Both SR's Inoperable
| |
| : 2) LER 20-02-88-022: Rx Trip on SR Hi Flux 876M/1 5/89 l
| |
| l l _ . _ _ _ . , _ _ . _ _ , _ _ . _ _ _ _ . . . _ . . . _ . _ _ . . . _ . _ _ . _ _. ;
| |
| | |
| g w-- , ,.,_ _ _ _ _
| |
| -{,gs, y a y ---
| |
| L tCitsili iJEur eiP0k T t LER I Docket Numoer (2) ',l ne :11
| |
| )
| |
| -atitty Name (1) f' Gr1Mdg Unit 1 of 51 01 0[ 21 al 51 6 1 lsf 0 l1
| |
| ' 'All action Requirements of Tech Soecs N2Lfil.1$3]e Both Source Ranoe Channelt Were Inocerable.
| |
| LER 1qumber f61 Recort Date (7) Other Facilities Involved fel
| |
| ,^ gnt Date (5)
| |
| Day Year Facilit y Names Occket Nurece r ( s )
| |
| kji Day Year Year Sequenttal // Revision Montn
| |
| /p/j//j
| |
| // Number
| |
| /jj/
| |
| f
| |
| // Number NONE 01510101of f i 81 6 al 6 0 1012
| |
| ~~
| |
| 010 1 l2 01 .5 al 6 01 El 01 01 01 I l 11 1 11 7 THIS REPORT 15 SUBMITTED PUR5UANT TO THE REQUIREMENTS OF 10CFR (Check one or more of the followinal fill 20.402(b) 20.405(c) 50.73(a)(2)(tv) 73.71(b) 20.405(a)(1)(1) 50.36(c)(1) 50.73(a)(2)(v) 13.71(c)
| |
| POWER _
| |
| 50.73(a)(2)(vit) Other (specify LEVEL 20.405(a)(1)(11) 50.36(c)(2) in Abstract (101 0 0i 0 20.40b(a)(1)(ttt) L $0.73( a)( 2)( t ) 50.73(a)( 2)(vti t )( A) below and in
| |
| ////////////////////////// 20.405(a)(1)(tv) 50.73( a)(2)(i t ) __., 50.73(a)(2)(vitt)(B) 20.40$(a)(1)(v) _,. 50.73(a)(2)( tit) 50.73(a)(2)(x) . Text)
| |
| ////////////////////////// __
| |
| LICENSEE CONTACT FOR THIS LER f12l
| |
| * TELEPHONE NUMBER Name ext. 2495 AREA CODE Ricnard Schitessmann - System Test Engineer l
| |
| BlI l5 41 El BI -l 21 81.01 [
| |
| l COMPLETE ONE LINE FOR EACH COM [ FAILURE DESCRIBE 0 IN THf5 REPORT (13)
| |
| SYSTEH COLPONENT MANUFAC- REPORTABLE //
| |
| SYSTEH COMPONENT MANUFAC. REPORTABLE CAUSE CAUSE
| |
| ' TURER TO NPRDS TUNER TO NPR05 i l i I I I I i l I I I i //
| |
| i I I I l i I I I I i 1 I I ,I #f
| |
| /
| |
| I Expected tmntn.1 Day l YeaC SUPPL [b[NTAL REPOR.L[xPECTELL)(1 ..
| |
| Submiss 1on f Date (15) l I I l .I I i \ yes (If ves. comolete EXPECTED SUBMIMJQ3_DAliL_ ._ J.J NO .
| |
| f_
| |
| ABSTRACT (Limit to 1400 spaces. i.e. approximately fif teen single space typewrittert 1tnes) (16) 1 On 11-17-86 at 0514. Technical Specification action Requirement 3.3.1 was entered to allow blocking tne scurce Range High Flux at Shutdown alarms and Baron Otlution Protection System (BOPS) while performing evolutions m tre 345 l Kv switchyard. The requirement to verif y closed and secured valves ICVill8.1CV8428. ICV 8439. icv 8441 ano IC.8435 to block possible paths of boron attution was not performed.
| |
| It was belteved that credit could be taken for a datly survet11ance that vertf ted four of tne five vilees were Later closed and locked, and that a valve upstream of ICV 8435 was closed and locked, rather than ICV 8435.
| |
| r es tew.
| |
| surveillance verified all valves were secured. This LER will be distributed to a11 licensed personnel f c There have been no previous occurrences.
| |
| o O
| |
| 1458m(112986)/1046A/10 j
| |
| | |
| .n_.-__. - - - - - - - - _ - --- ..-=__.n.-~ _ .-
| |
| LICEMLiviNT o[PQELu(BL TEC CMInfWllM
| |
| /ACILITY NAME (1) DOCKET NUMBER (2) LEP Nug[G W one r 3i Year /
| |
| ,/,p/
| |
| Sequenttal //j / Revision
| |
| /// Number jj//
| |
| / Number Brii h d. Unit 1 0l$l0l0 10 l 41 $1 6 8I6 - Ol0l2 - 0i 0 01 2 0F Of 3
| |
| !T Energy Industry Identification System (EII5) codes are identified in the text as (xx]
| |
| A. PLANT CONOTIONS PRIOR 70 EVENI:
| |
| Mode 5 - Cold shutdown. Reactor Coolant System (RCS) (AB) temperature and pressure are at ambient conditions.
| |
| : 8. DLgRIPTrom Dr E)yM:
| |
| 1 On 11-17-86 at 0514, ,m te,rirforming switching evolutions in the 345 KV switchyard. Technical Specification 3.3.1, Action 5. was etdersd to block both tratns of Boron 011ution Protection System (80PS) (CB] and both source range channels (IG] High Flux at Shutdown alarms. This was performed to prevent expected B0PS actuation and containment evacuation alarms due to source range sptktng that had been previously observed during switchyard activities. The requirement of Action 5 to verify valves ICV 1118, ICV 8428, icv 8439. ICV 8441 and ICV 8435 closed and secured was not performed.
| |
| At 0629,'the High Flux at $hutdown alarms for both channels and 80P1 for both trains were unblocked. and Technical Specification Action Requirement 3.3.1 was exited.
| |
| This event is reportable under 10CFR50.73(a)(2)(1)(B) - any operation or condition prohibited by the plant's Technical Spectftcations.
| |
| C. CAUSE OF EVENT:
| |
| p Valve line-ups were not vertfled due to an interpretation of Techntcal Specifications by the Shif t Control Room 7 Engineer (SCRE) to allow credtt to be taken for a survet11ance that was performed earlier in the shif t.
| |
| Procedure 18w05 XLE-01. Boron Oilution Prevention Locked Valve Daily Surveillance. This surveillance. which is performed to comply witti the 10CFR50.57(c) licensing submittal, vertf tes valves 1CV1118.1CV8428, ICV 8439 and ICV 8441 are locked closed on a daily basis. This surveillance also verifies Closed and locked ICVB453. wnten ts upstream of 1CV8435, and is the only dilution source to ICV 8435. Based on this and the f act that 18 05 xLE-01 is a more comprehensive surveillance than Procedure IBwCS 3.1-la. LCOAR Reactor Trip System Instrumentation, the SCRE felt the intent of the Technical Specification had been met.
| |
| D. SAFETY ANALYSIS:
| |
| There were nu safety consequences resulting from this event. All dilution flow paths had the required valves
| |
| , verified closed and locked per 18wCS XLE-01.
| |
| E. CORRECTIVE ACTIONS:
| |
| : 1) Inunediate wrective Action - On 11-18-86 at 1700 two Senior Reactor Operator Licensed individuals serified ICV 8435 was locked closed. Additionally.18wo$ 9.1.3-1. Refueling Otlution Prevention valve Position Montnly Surveillance, performed on 11-12-86, was reviewed to ensure all five valves had been verified locked closed as of that date.
| |
| : 2) Action to Prevent Recurrence - Copies of this LER will be distributed to all licensed personnel f or review.
| |
| and documentation of this review will be retained by the Tratning Department. Refer to Action Item 456-200-86-05001, which will track this (Jrrective action.
| |
| a J
| |
| 1458m(112986)/1046A/11
| |
| - _ _ _ _ . _ ~ _ _ _ _ _ _ _ _ _ _ - _ _ . _ _ _ _ . _ _ _ . _ ._
| |
| | |
| -% ~ . . . . . m. . _ _ _ _ . _ _ _ _ _ _ , _ __ _ _
| |
| Di i ntastunuttienat<utnMiuantIwmoN _ stre (li DOCKET t4 UMBER (2)- ~'LER 'tsut2E(2 f 41
| |
| . /ACILITY CAME 11)J f ear. . /// . S eque'it i al /// Revision Ls *' .
| |
| ///
| |
| /// Numner g Number' j[" Li;-
| |
| a i 6 _. 0 10l'2 0 1 0 01 1 0F o f 'I
| |
| ;/ " vat u ad:'untt 1 0i G l'0'l'0'l 0 1 41 51 6 - -
| |
| ' Energy Industry Identification system (E!!S) codes sre scentified in the text.as [xx]
| |
| .f r;;pervraus acconarNets:
| |
| - l- Non2, c; eonenurer rattuer narr:
| |
| NIne, p
| |
| 0 L' ~
| |
| .145am(112966)/toa6A/12 L = -- _ _ _ _ _ _ ___
| |
| | |
| LICENSEE EVENT REPORT (LER) Nf3"I f Form Pev 2.0 1 Facility Name (1 1 0;Cket Number (2) Pace f31 l 3
| |
| 0! El 01 01 01 di 51 7 1 of 0 3 Title (4) Rx Trip Que to Loose Connections in 2PH05J (Source Range H1 Flux) w- Event Date th) LER Number (6) Renart Date (7) Other Facilities involved (Ri Month Day Year Year /// Sequential /// Revision Month Day Year Facility Names Docket Numberft)
| |
| ,/j/j/ Number ,/,/j/ Number l
| |
| Braicwood of El 01 Of Gl l l
| |
| ~~ ~~~
| |
| Ol 9 11 9 Bf 8 ' al a 01212 0l 0 1 l0 11 1 al a el El 01 01 01 l l THIS REPORT IS SUBMITTED PUR5UANT TO THE REQUIREMENTS OF 10CFR ftheek one or more of the followino) f111 2 20.402(b) 20.405(c) .,X. 50.73(a)(2)(tv) ,,_ 73.71(b)
| |
| PMR ,,,,,,,, 20.405(a)(1)(1) 50.36(c)(1) ,,,,,,,, 50.73(a)(2)(v) 73.71(c)
| |
| LEVf.L 20.405(a)(1)(11) ,,, 50.36(c)(2) ,, 50.73(a)(2)(vil) _1., other (Specify (101 0l0l 3 20.405(a)(1)(iii) ,,,,,_ 50.73(4)(2)(1) __., 50.73(a)(2)(viti)(A) in Abstract
| |
| ////////////////////////// .,,,_ 20.405(a)(1)(tv) 50.73(a)(2)(11) ,_ 50.73(a)(2)(viti)(B) below and in
| |
| ////////////// /////////// 20.405(a)(1)(v) 50.73(a)(2)( tii) .,,,,,_ 50.73( a)(2)( x) Text) 50.72 LICENSEE CONTACT FOR THIS LER (121 Name TELEPHONE NUMEER AREA CODE Freddie Ramos. Technical staff Encineer Ert. 2487 9 l1 15 41 El 81 -l 21 Bl Of COMPLETE ONE LINE Fct EACH COM 0 T FAILURE DEtCRIBED IN THIS REPORT f131 CAUSE SYSTEM COMPONENT MANUFAC- REPORTABLE / CAUSE SYSTEM COMP 0NENT 'MANUFAC- REPORTABLE -
| |
| TURER TO NPRDs p/ TURER TO NPRDS I i l l l l l /
| |
| l I l I l 1 l i I I I I I I '/ '/ I I I I I I I SUPPLEMENTAL RFPORT EXPEETED (141 Expected Month I Day I Year m Submission i i """"( Ye t fff yet. comolete EXPECTED $UBMIS$f0N DATE)
| |
| Date (15) i i
| |
| ; X l No l' I l I l ABSTRACT (Limit to 1400 spaces. i.e. approximately fif teen single-space typewritten lines) (16)
| |
| At 1800 on September 19. 1988 a reactor trip occurred due to source range channel N31 exceeding its setpoint of 1.0xE5 counts per second (CPS). "A", reactor trip breaker opened automatically. The Nuclear Station Operator initiated a manual trip to open"the "B" reactor trip breaker. The cause of this event was due to a loose connection in main control room panel 2PM05J. which allowed channel N-31 to re-energize. Since reactor power was approstmately 37. the reactor trip occurred. Subsequent investigation revealed that an actuation had only occurred on Train "A" and no failure of Tratn "B" actually occurred. The investigation revealed loose connections at the back of 2PM011 which were associated with the various Nuclear Instrumentation System blocking functions. These connections were tightened to prevent any further breaks in the blocking circuits. Additional terminal strips were checked for loose connections on both units. There have been no previous occurrences of loose connections in the source range resulting in a reactor trip.
| |
| N '
| |
| v 2309m(100788)/17
| |
| . e
| |
| .__----_ _- -- ~ _.-. . . - - . -- - -
| |
| | |
| _. . . . . -- .._ . _ _ . . . - ~ . . - - _ ._. . . - - -
| |
| t.fCENSEE EVENT REPORT (LER) TEXT CONTINUATION 6 -Form Rev 2.0 LER NUMeER f61 Pace f3) )
| |
| , FACILITY NAME (1) DOCKET NUM8ER (2)
| |
| Year Sequential // Revtsion i
| |
| /j/j/j/
| |
| // Number
| |
| /j/j j/ / Number rw Braidwood 2
| |
| ?
| |
| of 2 of 3 C TEXT o I s I o I o I o I el sf 7 a1a -
| |
| Energy Industry Identification System (E!!5) codes are identified in the text as (XX]
| |
| o1212 - oI o cF i
| |
| A. PLANT CON 0!TIONS PRIOR TO EVENT:
| |
| 4
| |
| ., Unit: traidwood 2;~ Event Date: September 19. 1988; Event Time: 1800; Reactor Mode: 2; Mode Name: Startup; Power Level: 3%;
| |
| ]
| |
| l
| |
| ?
| |
| RCS (AE) Temperature / Pressure: 557 degrees F/2240 psig a
| |
| , 8. DESCRIPTION OF EVENT:
| |
| There were no systems or components inoperable at the beginning of the event which contributed to the
| |
| - severity of the event.
| |
| At 1800 on September 19. 1988 a reactor trip occurred on Unit 2. First out annunicator. *5r High Flux Rx Trip", illuminated at the time of the event. Further investigation revealed that source range channel N31 (IG) exceeded its setpoint of 1.0xE5 counts per second (CP5). "A" reactor trip breaker opened automatically. '8" reactor trip breaker did not open automatically. Unit 2 Nuclear Station Operator (N50) initiated a manual, trip to open the "8" reactor trip breaker. Subsequent investigation revealed that an actuation had only occurred on Train "A' and no failure of Train "S" actually occurred.
| |
| Operator actions neither increased or decreased the severity of the event.
| |
| ~
| |
| The appropriate NRC notification via the ENS phone system was made at 1916 on Septemoer 19. 1988, pursuant to i 10CFR50.72(b)(2)(tt).
| |
| This event is being reported pursuant te 10CFR50.73(a)(2)(tv) - Any event or condttton that resulted in manual or automatic actuation of any engineered safety feature. including the reactor protection system.
| |
| C. CAUSE OF EVENT:
| |
| The cause of this event was due to a loose connection in main control room panel 2PM05J. Section 82. Part it, kiser A-2. Terminals 56-1 and 56-2. This loose connection caused a break in the Train "A" source range reset.
| |
| circu- which allowed the source range channel high flux reactor trip associated with channel N-31 to become eble. ed. This allowed channel N-31 to re-energt2e. Since reactor power was approximately 3% the 1.0XC5 CP5 setpoint was exceeded and the reactor trip occurred. The loose connection was disturbed when a Nuclear Station Operator (N50). Itcense reactor operator, was changing the paper on a nearby chart recorcer associated witn the volume control tank level. LR-185. This effect was duplicated during troubleshooting of the source range block circuit. N-32 did not energize because the block / reset and high voltage cutout remained funcional.
| |
| D. 5AFETY ANALYSI5:
| |
| There was no ef fect on the plant or pubitc Safety. The p1 Ant responded per design which is to trip the unit on source range htch flux (i.e.1 out of two coincidence logic). "B" reastor trip breaker did not open i automatica1% oecause only the Train 'A" had its source range unblocked due to loose connectiory in 2PM05J.
| |
| [(/ }
| |
| Under wrst case conditions with the loose connections in 2PM05J being jarred and the plant at 100% power.
| |
| source range high flux would cause a reactor trip to occur per design.
| |
| 2309m(100788)/18
| |
| _ _ mesu-m .1_ m .m4 m_.m ...,t -o , , ~ . . . .
| |
| | |
| _y.:.
| |
| A ^
| |
| -, . ~ ~ . . . . .. . . _ . - - . _.
| |
| l? ..
| |
| . LifENsEE EVENT REPORT fLER) TEXT CONTINUATION % ' Form Rev 2.0 y* FACILITY NAME (1) DOCKET NUMBEP (2)- LEn NuMnER f6) Pane-fil L Year. /// 5equential / . Revision j/j/j/ /j/j/
| |
| j// "ar fN Braiduced 2 M * -r' l'k ,
| |
| o i s I o I o f a l el sf 7 ala - oi212 - oI o of 1 0F of 1
| |
| : l. l TEXT- Energy Industry Identification Systen (EI15) codes are identifteJ in the tent as (XX)
| |
| E. CORRECTIVE ACTIDNS:
| |
| The taunediate corrective act @ by the Unit 2 operator was to trip reactor trip breaker '8*.
| |
| l ' A partial survet11ance on Train 'A' solid state protection system. 2Bw05 3.1.1-23 was performed to determine if the universal cards associated with the source range block circuits were functional. The surveillance did not reveal any abnormalities with $$PS.
| |
| 1' Nuclear Work Request A25642 was written to investigate cause of the source range channel N-31. .The investigation revealed loose connections at the back of 2PM05J which were associated with the various Nuclear Instrumentation System blocking functions. These connections were tightened to prevent any l further breaks in the blocking circuits.
| |
| Adcttional te etnal strips were checked for loose connections on Unit 1 and Unit 2.
| |
| F. PREVIOUS OCCURRENCES:
| |
| There has been previous occurrence of a reactor tirp involving source range monitoring instrumentation.
| |
| The corretive actions were implemented addressing both root and contributing cause. However, the root cause of this event is different in that loose terrinal wiring for the tource rang 2 instrumentation was involved. Previous corrective actions are not app'.1 cable to this event.
| |
| G. COMPONENT FAILURE DATA:
| |
| This event was not cause by component failure. nor did any components fall as a result of this event.
| |
| 4 h
| |
| i 1
| |
| (
| |
| 2309m(100788)/19 l
| |
| . e
| |
| | |
| BRAIDWOOD SIMULATOR MALFUNCTION
| |
| [.
| |
| L p
| |
| f l ,_
| |
| | |
| ==Title:==
| |
| Intermediate Range Channel Failure. ID: NIS-2
| |
| '(y)'
| |
| NO: 6.3.4.9.2 f:
| |
| | |
| == Description:==
| |
| Simulates a failure of the input to the log current amplifier.
| |
| Variations: None. Date: 4/21/88 Rev: 5 l Selectable Steps Inputs Comments l
| |
| : 1. Select channel NIS-2A NIS-2A = Ch. 35 NIS-2B NIS-2B = Ch. 36
| |
| ~ ~
| |
| : 2. Select final value 10 to Failure > 10 amps could
| |
| -3 result in reactor trip 10 amps Select ramp time
| |
| } 3. O to 99,999 see Brief Plant Response: (IC-9, 10-8 ,,p,)
| |
| If channel failure is in the low direction, this will cause erroneous i
| |
| indication only. No annunciators received.
| |
| If failure is in the high direction, a rod stop and/or an intermediate range high flux reactor trip will occur. Also, a premature P-6 permissive status light would occur if failure is introduced while reactor power is operating in the source range. The first annunciators received include IR HIGH FLUX ROD STOP C-1 and IR HIGH FLUX Rx TRIP.
| |
| Intermediate range startup rate indication will also be affected initially by either failure.
| |
| I i
| |
| 0129w:4 312M/85M/3 1/88 !
| |
| _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ ___ _ l
| |
| | |
| _ ~ - _ _ _ _ _ _ _ __
| |
| I- #
| |
| | |
| ==Title:==
| |
| Intermediate Range Channel Failure '
| |
| ID: NIS-2 Suggested Instruction Action:
| |
| Clear the malfunction when channel has been repaired.
| |
| Events:-
| |
| : 1) DVR 06-01-88-033: 'IR Channel Trip Setpoint Inaccurate.
| |
| l l
| |
| l.
| |
| U O
| |
| :0129w:4 312M/85M/4 1/88 l -'
| |
| | |
| r2 OEVIATION INVESTIGATION REPORT ,
| |
| TITLE Intermediate Rang 2 Channel Trip 5 tpoint Inaccuracy Cue to PMF
| |
| -Increased Guadrant Power Tilt Ratic 1 !C 13 EVENT DATE DIR NUMBER DEPORT DATE //
| |
| // SEQUENTIAL // REVISION /
| |
| I DAY YEAR STA UNIT YEAR / NUMBER NUMBER MONTH DAY YEAR POWER / /
| |
| 01 1 21 3 BI 8 01 6 0! 1 81 9 --
| |
| 01 31 3 -"'
| |
| 0 10 l l l l 11 4 CONTACT FOR THI! DIR NAME TELEPHONE NUMBER AREA CODE Pooer Flahive. Tech Staff Suoerviter Ext. 2243 8 l1 15 2 13 14 1-1S Ial4 l 1 COMPLETE ONE LINE FOR EACH COMPONE T URE DE!!RIBED IN THIS REPORT CAUSE SYSTEM COMPONENT MANUFAC- REPORTA8LE CAUSE SfSTEM COMPONENT MANUFAC- REPORTABLE TURER TO NPPDS TURER TO NPRDS I I I I I I I ,,_
| |
| 1 I I I I I I i i i I I I I i _
| |
| l i I I SUPPLEMENTAL REPORT EXPECTED MONTH DAY YEAR EXPECTED SUBMISSION I YES fif ves. cornalete EXPECTED SUBMISSION DATE) l NO l TEXT A. PLANT CONDITIONS PRIDR TO EVENT:
| |
| Event Date/ Time 1/21/ER /
| |
| Ip)
| |
| C/ Unit 1 MODE 1 - Power Ooerations Rx Power 34 RCS (AB) Temperature / Pressure NORMAL CPERATING Unit 2 MODE N/A - N/A Rx Power N/A RCS [AB) Temperature / Pressure N/A
| |
| : 8. DESCRIPTION Or EVENT:
| |
| On January 21. 1988 (prior to this event) a load reduction was performed in order to replace solenoids on two main feedwater isolation valves. This load reduction initiated a quadrant power tilt ratto (CPTR) of Creater than 1.02 (reference OVR 6-1-87-143). As power was increased on January 23, 1988, after the salenoids were replaced. it was noticed that the nuclear instrumentation system (10) intermediate range N36 trip bistable did not clear until approximately 34% power as observed on the power range channels. The expected value was approximately 25% power. That same day, a Nuclear Work Request (NWR) was generated to l the Instrument Maintenance Department to check the calibration of N36. There were no other imediate actions since the Technical Specifications only require the intermediate range trips to be operable in i modes 2 and 1 below P-10. A check of N36. which was done on January 29. 1988. showed the as-found value of j the trip bistable to be the same as the required v&ive (equivalent to 1 X 10~4 amps). )
| |
| l Once a calibration problem had been eliminated. investigation was begun to determine if the intermediate l range / power range overlap response had changed such that the current equivalent of 25% power was no longer 1 X 10~4 amps. During a past reactor startup. 87-05, on August 13. 1987, overlap data had been obtained which verified that 1 X 10~4 amps corresponded to approximately 25% power. On the load drop on January 21, 1988. overlap data showed the N35 trip setpoint to be equivalent to approximately 25.9% power and the '
| |
| N36 trip setpoint to be equivalent to approximately 28.5% power. Both of these are within the allowable value as defined in Technical Specifications. Several hours later during the power ascension through 25%
| |
| [] power, the intermediate range currents were found to have been 8.8 X 10-6 and 8.0 X 10-5 amps for N35
| |
| / and N36. respectively. As power ascension continued. it was found that 1 X 10-4 amps corresponded to about 2b% power for N35 and 35% for N36. The allowable value is 30.9% rated thermal power.
| |
| 1 I l (1985M/0221M/033088)
| |
| | |
| DIR NUMBER pace TITLE
| |
| . Intermediate Range Channel Trip Setpot'nt Inaccuracy SEQUENTIAL REV!510N Du2 to Increased Quadrant Power Tilt Ratio _STA UNIT YEAR NUMBER NUMBER 01 6 Of 1 BI B -
| |
| o1313' -
| |
| 0Io 2 or oi1 TE W
| |
| ~C. CAusE OF EVENT!
| |
| The root cause of this event is indeterminate. It is believed by Technical Staff personnel that this shif t
| |
| .in intermediate range / power range response was due to the shift in power distribution caused by the Quadrant Power Tilt Ratio (QPTR) condition which existed due to the load decrease. The problem of a QPTR l' condition occurring on load drops 'was reported in DVR 6-1-87-143 and work is ongoing to attempt to resolve this (see AIR 454-512-88-0024). In addition, the intermediate range / power range response will be monitored throughout the rest of this Unit 1 fuel cycle to ensure the current trip setpoints are in accordance with the Technical Specification limits. Action Itum Record (AIR) 454-512-88-0060' is tracking this item. At the start of each new fuel cycle for both Unit I and 2. intermediate range / power range overlap data will be obtained to determine proper intermediate range trip setpoints. AIR 454-512-88-0059 is tracking this item.
| |
| l' l D. SAFETY ANALYSIS:
| |
| 1 There was no adverse impact to the plant or public health and safety. The intermediate range trips are l 'only required to'be operable in mode 1 below P-10 and mode 2 per Technical Specifications. Power was never below P-10 during this event. In addition, the intermediate range trips are not assumed to function in any safety analysis as discussed in the Final Safety Antalysis Report. (FSAR). Therefore, under worst cast l conditions of being in mode I below P-10 or mode 2. there would still be no safety impact. During power
| |
| -decreases to below P-10 the intermediate range trip status lights are verified illuminated as reactor power goes below 25% per step 12 in Byron operating Procedure BGP 100-4. This would indicate to the. licensed cperators in the control roort if the intermediate range channel (s) trip setpoint was inaccurate due to QPTR. In addition, shif tly channel checks are done for the intermediate range channels.
| |
| E. CORRECTIVE ACTIONS:
| |
| AIR 454-512-88-0060 was generated to monitor Unit 1 intermediate range / power range response during the l
| |
| remainder of Unit 1 cycle 2 operation. Unit 2 does not have the QPTR problem so the AIR does not include monitoring Unit 2 channels. AIR 454-512-88-0059 was generated to develop a procedure to acquire intermediate range / power range overlap data during initial startup from each refueling and analyze it to determine if a change to the trip setpoints is required. This is applicable for both Units 1 and 2.
| |
| F. PREVIOUS OCCURRENCES:
| |
| l There are no DVRs that document the intermediate range problem induced by QPTR.
| |
| I DVR NUMBER TITLE NONE l
| |
| l l
| |
| . I I l' (1985M/0221M/033088) l
| |
| | |
| . ~ - - - _ - . _ _ - _ _ .
| |
| ;r . - . _ .
| |
| DEVIATION I"YESTIGATION RET' ORT TEXT CONTINUATION TITtt ;. OIR NUMBER PAGE
| |
| . I:.termediate Range Channel Trip Setpoint Inaccuracy SEQUENTIAL REVISION L
| |
| STA UNIT YEAR ' NUMBER NUMBER l Due to. Increased Quadrant Power Tilt Ratio
| |
| ^
| |
| 0 16 0 11 8 la -
| |
| 011 11 -
| |
| o l0 1 0F 0 l1
| |
| ! TEXT
| |
| ' G, COMPONENT FAILURE DATA:
| |
| . c) MANUFACTURER NOMENCLATURE MODEL NUMBER MFG PART NUMBER
| |
| 'Not Applicable bl. .RESULTS OF NPRDS SEARCH!
| |
| .Not Applicable c) gitutTS OF NWR SEARCH Not Applicable bince a LCOAR would have been entered for this problem and a DVR written.
| |
| s*A 1
| |
| i N
| |
| I I (1985M/0221M/033088)
| |
| | |
| - - - ~
| |
| o_ . a _... . . . , _ _ . . . . . . _ _ _ _. _ . . _ . , _--- _ _
| |
| i BRAIDWOOD SIMULATOR MALFUNCTION
| |
| | |
| ==Title:==
| |
| Power Range Channel Failure ID: NIS-3 q
| |
| []
| |
| V NO: 6.3.4.9.3 l
| |
| | |
| == Description:==
| |
| Failure of any power range channel's l summing amplifier's output. l Variations: None. Date: 4/9/89 !
| |
| Rev: 5 l 1
| |
| ' Selectable !
| |
| Steps Inputs Comments !
| |
| i i
| |
| : 1. Select channel NIS-3A - 3D NIS-3A = Ch. 41 NIS-3B = Ch. 42 NIS-3C = Ch. 43 NIS-3D = Ch. 44
| |
| : 2. Select final value 0 to 200 ' percent Failure > 103 percent will result in overpower rod stop f
| |
| : 3. Select ramp time 0 to 99,999 see Brief Plant Response (IC-17%, 100%, all systems in automatic)
| |
| Failure in low direction - will result in erroneous indications and various protection histable indications; will not result in a reactor trip since only one channel will be affected. The first annunciators received are ROD DEV PWR RNG TILT.
| |
| Failure in the high direction will cause insertion of the control rods until:
| |
| : 1) Operator selects manual on rod control, 2) Rate comparator circuit decays, or 3) A reactor trip occurs. Protection bistable actuation also occurs. The first annunciators received include ROD DEV PWR RNG TILT, PWR RNG CHANNEL DEV, PWR ENG HI SETPT RX TRIP ALERT and PWR RNG FLUX HIGH ROD STOP.
| |
| ,~
| |
| 876M/2 5/89
| |
| =__- me m -_ m- -- _ v : - ~ -- - - . - -- -- -- - - - -- - - --
| |
| | |
| . x, -
| |
| BRAIDWOOD SIMULATOR MALFUNCTION l 1
| |
| m.
| |
| | |
| ==Title:==
| |
| Power Range Channel Failure ID: NIS-3 Suggested Instruction Action:
| |
| Trip appropriate bistables when requested. Clear malfunction when channel has .I I
| |
| been repaired.
| |
| Rvents:
| |
| : 1) Byron STR 8/18/85: IM trips 2nd PR channel.
| |
| : 2) DVR 20-02-89-007: pr channel failure.
| |
| )
| |
| l ll
| |
| ( ,
| |
| )
| |
| l I
| |
| l 876M/3 5/89
| |
| | |
| August 8, 1985 ^f'5
| |
| .' f%
| |
| l.
| |
| Preli=tr.ary Root l Regulatory Implications Cause Assessment l*ll Not a Noncompliance l lx,tx,1 Personnel Error l l_l i.evel IV Wencompliance l l _ l Eeuipment Failure l l_,,,,l _ License Violation l l _l Procedural, l l l Possible Enforcement Action l l l Other l TROUBLE REPORT BYRON STATIOW On Wednesday, August 7 at 0916 hours, Byron Unit 1 experienced a reactor scram from 98% power. The scram was attributed to a personnel error which occurred during an instrument surveillance.
| |
| e During the surveillance, one channel (Channel 41) of the Excore Detector System was in the TRIP position. The Instrument Mechanic (IM), in attempting
| |
| ' to disconnect the wires from Channel 41, inadvertently disconnected the wires from Channel 42. G enever 2 out of 4 instrument channels in the Excore f Detector System are out of service, an automatic reactor scram occurs, t
| |
| k Byron has already begun start-up procedures and hopes to be back on line torne time today.
| |
| A Red Phone call was made to the NRC and Communication Services was notified. ,
| |
| /k ti.fq Ww e,-
| |
| Katie Krag 0606B/3
| |
| | |
| f MIS-3 DEVIATION INVEST!GATION REPORT (DIR) 73 PACE l Facility Name 1 10FI O I3 Braid cod 2 Title Failure of Power Range Channel N42 Caused By failure of Ristahla Card Nf 306 DfR NUMBER REPORT DATE EVENT DATE
| |
| // $EQUENTIAL // REVISION DAY YEAR 1 DAY YEAR STA UNIT YEAR NUMBER NUMBER MONTH MONTH POWER RI 9 21 o el 2 BI e - -
| |
| oI o el 2 of B al 9 of of 1 nl 1 21 7 o i el 7 CONTACT FOR THit DIR
| |
| . TELEPHONE NUMBER NAME AREA CODE David R. Lawton. Technical Staff Enaineer Ext. 2402 RI1 l1 4l EIB l-l 219 IoI1 COMPLETE ONE LINE FOR EACH C31 PONE FA LURE DESCRIREO IN THft REPORT CAUSE SfSTEM COMPONENT MANUFAC- REPORTABLE CAUSE SYSTEM COMPONENT MANUFAC- REPORTABLE /j e TU9ER TO NPFDS TURER 10 NPRDS l I 1 l l l l I l l I l I l /
| |
| x t Ic tl El 'l a lwi1121o I I I I I I MONTH DAY ' YEAR SUPPLEMENTAL REPORT EXPECTED p
| |
| $USMI$$10N i NO ! i l l i YEs fif ves. etunclete EXPECTED tuaMISSION DATE1 TEXT Energy Industry Identification $ystem (E!!$) codes are identified in the text as (XX)
| |
| _ A. PLANT CONDITIONS PRIOR TO EVE:'Y: l
| |
| (\J Unit: Braidwood 1: Event Date: 01/27/89: Event Time: 1534: !
| |
| M0da: 1 - Power Operation; b Power: 91%; {
| |
| RCS(AB) Temperature / Pressure: 588 degrees F/2241.5 asig
| |
| : 8. DESCRIPTION OF EVENT:
| |
| At The unit was critical in Mode 1. and had been operating at a constant power level for a number of days.
| |
| 153a on January 27, 1989. the Power Range High Setpotnt Reactor Trip Alert annunciator ( Annunciat3r Window 2-10-A3) actuated. Investigation showed that Trip Status Light 7.2 on Trip Status Light Scard (TSL8) 4 (Power Range High Fluu (High Setpcint) Nt4D) and the Overpower Trip High Range annunciator on the Power Range Nuclear Instrumentation (IG) channel k42 had also actuated. All indications of reactor Percent Full j
| |
| Power that are provided by t.his channdi continued to display approximately 90.4%. which is below the 1087.
| |
| trip setpoint for these alarms. Etnce this alarm had clearly f ailed. Limittng Condition for Operation Action j Requirement (LCDAR) 3.3.1-1A was entered at 1535. While preparations were being made to take channel cf seryice for troubleshooting. the channel's Percent Full Power indications began to intermittently di; play 07.. 28w0A IN$7-1. Nuclear Instrumentation Malfunction procedure, was entered at 1607 because of the fatted channel indication. A Nuclear Work Request (NWR) was generated to troubleshoot and correct the problem.
| |
| Sistable card NC306. which provides the Overpower Trip High Range output from the Power Range, was repliced.
| |
| The appropriate sections of Sw!$ 3.1.1-214 Surveillance Calibration of Nuclear instrumentation $ystem Power l Range Nat, N42. Na3 & N44 were performed to calibrate the new card. Following these corrective actions all functions of Power Range channel M2 performed correctly, therefore at 0645 on January 28. 1989. LCDAR 3.3.1-1A was exited.
| |
| l V
| |
| j 2494m(0215891/2 1
| |
| I
| |
| | |
| ~~ ~ . . - . _ - , - _ -.. _ . .. __ __
| |
| DEVIATIO:) INVESTIGATION REPORT' TEXT CONTINUATION Verm Rev 2.0.
| |
| PACE DIR N_WMBER ,
| |
| . FACILITY NAME
| |
| $EQUENTIAL REVISION
| |
| / stA UNIT YEAR NUNeER NUHe[R
| |
| ! (/
| |
| 21 0 01 2 BI 9 """
| |
| OIOl7 -
| |
| O l 0 2 0F 0 l1 Braidwood 2 TEXT Energy Industry Identification System (E!!$) codes are identified in the text as (XX]
| |
| C. CAUSE OF EVENT:
| |
| The root cause of this event was a fatture of the NC306 Overpower Trip Wigh Range bistable card. The f ailure of this card created the f alse Percent Full Power indications by an undetermined method.
| |
| Possibilities for interaction include depressing the output voltage of the Low Voltage Power Supplies and grounding the output of the Sunning and Level An91tfier card. The behavior of these components, following replacement of the histable card, was tested by performance of sections of Bw15 3.1.1-214. No abnormal indications were noted. indicating that no damage to these components had been caused by the btstable card fatlure.
| |
| There are no procedurf1 inadequacies or personnel errors associated with this event.
| |
| O. SAFETY ANALYSI5:
| |
| There were no safety consequences resulting frors this event because only c3e channel of the Power Range was affected. This had no effect on the operation of the remaining three channels, so no safety functions performed by this equipment were impatred. .
| |
| If more than one Power Range channel had been af fected sinrsitaneously, there still would have been no safety consequences since the Distable f ailed in the tripped condition, thereby inserting the reactor
| |
| [)
| |
| ( trip signal requtted for conservative operation.
| |
| E. CORRECTIVE ACTION 5:
| |
| The inmediate corrective action taten was to declare the Overpower Trip High Range reactor trip function inoperable per Technical Spectfteattons. Section 3.3.1. Following recetot of the erroneous Percent Full Power indications the Power Range channel N42 was removed from service using 2BwCA INST-1.
| |
| Further correct #ve action was accompli $ned by use of the Nuclear Work Request program at this station.
| |
| The histable card was replaced and its trip setpoint was calibrated by use of 4 station-approved procedure. The effect of the bistable card fatlure was analyzed by use of the same procedure. No other failures were taentified.
| |
| This failure was identified during normal operations. The existing survetilance program is deemed sufficient to maintain calibrations of the Power Range protective functions, and to identify faulty components. More extensive surveiling of the system would cause the system to be removed from service for a larger percentage of time than is present practice, and would add no value f or identification of this type of failure.
| |
| A record is presently being maintainee to track the occurrences of failure of this card, and similar !
| |
| carts,in the Nuclear Instrumentation System. If a large number of failures are noted further action {
| |
| will be initiated. No $Uch trend is presently indicated.
| |
| Gt l
| |
| 2494m(021589)/3
| |
| ___.m _ _ _ _ . = _ . . , . _ . _ _ _ . - . _ . . . . , , . .-
| |
| | |
| (. s' DEV!ATION INVEST! GAT!03 REPORT TEXT CONTINUATION
| |
| : l. Form Rev 2.9, I'
| |
| FACILITY NApt ' Dit NUMaER Pact
| |
| $EQUENTIAL t
| |
| ' /N l REVISION TJMBER l- ( ) iTA UNIf YEAR NUMBER 1: %) ~
| |
| arat h d 2 21 e el 2 al e - o1oI7 -
| |
| ol o -1 or oi1 TEXT l Energy Industr> Identification System (E!!5) codes are identified in the text as (XX]
| |
| F. PREVIOUS OCCURA.ENCES:
| |
| No previous Deviation Reports were filed concerning fr. lure of this comepnsnt.
| |
| G. COMPONENT FAILURE DATA:
| |
| Manufacturer Nomenclature Model Number MFG Part Number Westinghouse Bistabit NA 3359C39G01 Electrical Corporation G
| |
| \
| |
| \
| |
| l l
| |
| 1
| |
| )
| |
| i i
| |
| -(
| |
| 2494m(02H991/4 sw_ _ _m_e :_ = -a -s:-~ v - ~ v :- - - -- -- - ~ -- - - ---
| |
| | |
| BRAIDWOOD SIMULATOR MALFUNCTION
| |
| | |
| ==Title:==
| |
| Intermediate Range Channel Camma ID: NIS-4 Q Compensation Failure NO: 6.3.4.9.4
| |
| | |
| == Description:==
| |
| Variation of compensation to either intere,ediate range channel.
| |
| Variations: None. Date: 7/11/87 Rev: 4 Selectable Steps Inputs Comments
| |
| : 1. Select channel NIS-4A NIS-4A = Ch. 35 NIS-4B NIS-4B = ch. 36
| |
| ~
| |
| : 2. Select current value 1 (10 to Positive = undercompensation'
| |
| ~
| |
| , 10 ) amps Negative = overcompensation
| |
| ~
| |
| i If 4 % x 10 or greater is used Iow comp, voltage alarm will actuate
| |
| : 3. Select ramp time 0 to 99,999 sec Briaf Plant Response: (IC-11, 8%, Rx trip occurs)
| |
| Note: IR meter will respond according to the following formult:
| |
| indicated = actual + step 2 value
| |
| : 1. When undercompensated, the IR channel will indicate both neutron and gamma flux, causing it to level off at a higher value (> 10~ ampt) during a e
| |
| 1 0129w:4 312M/85N/7 7/87 u____--_____ _ _ _ _ _ _ _ _ - _ .
| |
| | |
| BRAIDWOOD SIMULATOR MALFUNCTION
| |
| | |
| ==Title:==
| |
| Intermediate Range Channel Gamma ID: NIS-4 Brief Plant Response (continued):
| |
| reactor shutdown. If the failure is high enough, it can prevent automatic reenergization of the source range high voltage. P-6 permissive will not clear . No annunciators received.
| |
| : 2. When overcompensated, the IR channel will indicate lower than actual because part of the neutron flux is also being cancelled with the gamma flux. During a plant shutdown, the affected channel will decrease faster than the unaffected channel. The first annunciator received is IR COMPENSATING VOLT FAILURE.
| |
| Suggested Instruction Action:
| |
| 4 When asked to repair or adjust the affected channel, clear the malfunction.
| |
| }
| |
| E&nts: None l
| |
| j l
| |
| I O
| |
| ~
| |
| 1 0129w:4 312M/85M/8 7/87
| |
| | |
| > -BRAIDWOOD SIMULATOR MALFUNCTION.
| |
| p
| |
| | |
| ==Title:==
| |
| Power Range Detector Failure ID: NIS-7
| |
| [ g g, 3 NO: 6 3.4.9.7
| |
| -y, .
| |
| | |
| == Description:==
| |
| Failure of either the upper or lower detector of a selected p>wer range channel.
| |
| Variations: None. Date: ~1/6/88 Rev: 4 Selectable Steps Inputs Comments
| |
| : 1. Select detector NIS-7A - NIS-7H A = Upper detector 7A = 41A, 41B = 7B B = Lower detector 7C = 42A, 42B = 7D 7E = 43A, 43B = 7F 7G = 44A, 44B = 7H
| |
| : 2. Select final O to 5 ma Note: With rod control in b; current auto, failure in the high direction could cause rod insertion
| |
| : 3. Select ramp time- O to 99,999 see Brief Plant Response:
| |
| The selected channel's indication will respond for the failed detector,'with the associated alarms for abnormal power tilt and AI indication. Possible detector and channel deviation alarms and changes in OTAT and OPAT trip setpoints. .The first annunciators received include PR CHANNEL DEV HIGH and PR
| |
| , DETECTOR FLUX DEV HIGH
| |
| , u)
| |
| I 0129w:4 312M/85M/11 1/88
| |
| | |
| m .A
| |
| ~
| |
| 'BRAIDWOOD SIMULATOR MALFUNCTION.
| |
| 3 1
| |
| a> ,
| |
| '''(p'- i l
| |
| | |
| ==Title:==
| |
| .- Power Range Detector Failure 10: NIs.7,
| |
| ;. 3;
| |
| '1 Suggested' Instructor Ac' ion:
| |
| -c
| |
| ' Clear the malfunction when repairs have been completed.
| |
| Events: None w ,
| |
| r..-
| |
| l l-E s:. ,
| |
| .0129w:4 312M/85M/12 8/86
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| | |
| BRAIDWOOD SIMULATOR MALTUNCIION
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| | |
| ==Title:==
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| .' Source Range Channel High Voltage Failure ID: NIS-8 ll s) - NO: 6.3.4.9.8 l L/. .
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| | |
| == Description:==
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| Failure in high voltage power supply.
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| Variations: ,-None. Date: 1/6/88 Eev: 4 Selectable Steps Inputs Comments
| |
| : 1. Select channel NIS-8A NIS-BA = 31 NIS 9B NIS-8B = 32
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| : 2. Select final detector 0-2500 volts Normal detector voltage
| |
| . voltage approx. 2000 volts.
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| Loss of detector voltage alarm at 100 volts < normal r~
| |
| : 3. Select ramp time O to 99,999 see Brief Plant Response:
| |
| Failure to less than normal voltage - will cause a reduction in the count rate indicated on the selected channel, also a reduction in the startup rate.
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| The first annunciator received is SR HIGH VOLT FAILURE.
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| Failure to greater than normal voltage - will cause a corresponding increase in the count rate and SUR of the affected channel and possibly a source range high flux reactor trip.
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| I 0129w:4 312M/85M/13 1/88
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| A t
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| 'BRAIbWOOD' SIMULATOR' MALFUNCTION
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| | |
| ==Title:==
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| ' Source. Range Channel High Voltage Failure
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| ~
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| ID: NIS-8 A..
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| SuEgested Instruction Action:
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| Clear malfunction When repairs have been completed.
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| Events:
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| 1). DVR 22-01-87-036: Failure of SR Channel 31 to energize.
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| i
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| 'i 1
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| 'k 0129w:4- 312M/85M/14 1/88
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| .]
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| OEVIATION INVESTIGATION REPORT
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| / \
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| / IITLE PAGE l Failure of the Source Range Channel IN31 High Voltage Power Supply to Enercize. 1 J0FI O l EVENT DATE DIR NUMBER REPORT DATE ,
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| SEQUENTIAL REVISION DAY YEAR STA UNIT YEAR y// NUMBER y/
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| / NUMBER MONTH DAY YEAR MODE 2
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| MONTH POWER ',
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| 01 3 11 9 BI 7 212 Ol 1 Bi 7 -
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| 0 13 l 6 -
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| 0! 0 01 7 2l2 Bl 7 Ol 01 0 h CONTACT FOR THIS DIR NAME TELEPHONE NUMBER i
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| AREA CODE Johr F. Kellerhals Ext. 307 3l1l2 7l4l6l-1210IBl COMPLETE ONE LINE FOR EACH COMPONEN URE DESCRIBE 0 IN THIS REPORT CAUSE ST5 TEM COMPONENT MANUFAC- REPORTABLE CAUSE ST5 TEM COMPONENT MANUFAC- REPORTABLI' TURER TO NPROS TURER TO NPR05 ,
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| i X IIG Rl JI XI PI3I2I3 Yes I l l I l I l 8
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| l l I I I I I I I I 1 l juPPLEMENTALREPORTEXPECTED -'
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| EXPECTED SUBMISSION
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| ~ ^
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| l YES (if yes, complete EXPECTED SUBMISSION DATE) ll NO !
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| TEXT Energy Industry Identification System (EIIS) codes are identified in the text as [xx]
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| A. PLANT CONDITION 5' PRIOR TO EVENT:
| |
| O MODE Hot Standby - 2 RX Power 000 RCS [AB] Temperature / Pressure 549 'F/ 2235 psig B. DESCRIPTION OF EVENT:
| |
| During a controlled shutdown of Unit 1 on March 19,1937; at 0045 hours, the tilgh Voltage of Source Range Channel IN31 failed to automatically re-instate when both Intennediate Range Channel indications decreased below SE-11 apps. Source Range Channel IN32 functioned properly throughout the event. The Nuclear Station Operator (N50) att y ted to manually re-instate IN31 via the Start-up Range Manual Reset push buttons on the Main Control Board. However the high voltage on 1N31 did not energize. The N50 then pulled the instrument power fuses on IN31 and after a short period inserted the fuses at which time the high voltage and indication returned for IN31.
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| C. CAUSE OF EVENT:
| |
| The cause of the event was a degradation of the Source Range IN31 Hign Voltage "Crw bar'' circuit.
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| The crow bar circuit is designed to protect the drawer electronics in the event of an instrunent p5er spike. The circuit can also be activated by a momentary loss of power. Once the crow bar circui', i, j activated it must be reset by removing the instrument power from the high voltage power supply for j several seconds. A convenient way to do this is to remove the instrument power fuses.
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| D. SAFETY ANALYSIS:
| |
| l l There was no safety significance to this event since Source Range Chant.e1 IN32 functioned properly x throughout the event.
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| 05340
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| UEVIATION INVE511GATION REPORT TEXT CONTINUATION l.
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| DIR NUFBER # AGE
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| _. TITLE.
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| SEQUENTIAL REVISION Failure of the Source Range Channel IN31 High Voltace' Power Supply to Eneroire 212 01 1 81 7 - 01 31 6 -
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| 0l0 2 0F 012 IEXT Energy industry Identification System (E115) codes are identified in the text as [xx]
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| l E. CORRECT!vE ACTIONS:
| |
| The high voltage power supply which integrally contains the crow bar circuit was replaced with a 1 new high voltage power supply.
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| F. PREVIOUS OCCURRENCES:
| |
| ' on February 15, 1987, during a shutdown of Unit I at 2025 hours the same event occurred on Source Range ' annel IN31. At 2035 hours the same method of cycling the instrument power fuses was used to restore Source Range IN31. A Deviation Report was not written on this event as it was believed to be related to an instrument power spike.
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| Past DVR's include:
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| 22-2-80-36 .
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| 22-1-76-96 O G. . COMPONENT FAILURE DATA:
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| V '
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| MANUFACTURER - NOMENCt.AIURE MODEL NUMBER mfg PARf NUFBER Power Designs High Voltage UPMD-154WM1 Power Supply e
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| 4 05340'
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| BRAIDWOOD SIMULATOR MALFUNCTION
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| ==Title:==
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| Inadvertent Fuel Loading ID: NIS-10
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| ) NO: 6.3.4.9.10
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| == Description:==
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| Fuel assemblies are inaded in the wrong position during fuel loading.
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| Variations: None. Date: 06/25/86 Rev: 3 Selectable Steps Inputs Comments
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| : 1. Select Loading 1,2,3,4 1 = FSAR 15.4-13, E-13 &
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| E-14 swapped.
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| 2 = FSAR 15.4-14, E-7 & E-8 swapped.
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| 3 = FSAR 15.4-16, H-8 & H-9 swapped.
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| ! 4 = FSAR 15.4-17, region 2 in H-8.
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| : 2. Select Delay Time O to 99.999 see Brief Plant Response:
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| 1 Power shapes are more peaked during the reactor startup from refueling.
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| Suggested Instructor Action:
| |
| (
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| Insure student takes incore flut map and/or T/C symmetry check.
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| Events: None e ;
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| o .
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| 0129w:4 312M/85M/16 8/86 L_______________-. _ _ . _ _ i
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| 1 BRAIDWOOD SIMULATOR MALFUNCTION g
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| | |
| ==Title:==
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| Improper Detector Overlap ID: NIS-11 l
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| j^E
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| '% NO: 6.3.4.9.11 i
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| l
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| | |
| == Description:==
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| Failure to indicate proper overlap between SR-IR and IR-PR due to faulty IR circuit.
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| Variations: None. Date: 4/8/89 Rev 4 l-Selectable Steps Inputs Comments
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| : 1. Select channel NIS-11A NIS-11A = IR 35 NIS-11B NIS-11B = IR 36
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| : 2. Select ranges -1 -1 = SR-IR = IR reads low
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| +1 +1 = SR-IR = IR reads high
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| '3. Select amount of overlap 0-3 decades Brief Plant Responses (IC-8, Rx S/U in progress)
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| Either insufficient or excessive overlap is observed during a reactor startup.
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| Suggested Instructor Action:
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| Insure student observes improper overlap and takes appropriate action. When told to repair the circuit, clear the malfunction.
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| Events: None f
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| l L 876M/5 5/89
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| :mr_*r- :_ r - - - - - - " - ~
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| _ ___ __ _ _1 __ r r_:rr r - r--___-- __1-~ - - - - :- - -
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| L BRAIDWOOD SIMULATOR MALFUtlCTION h ,
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| n i x;
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| ==Title:==
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| Source Range Discriminator Failure ID: NIS-12
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| (: y
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| /;
| |
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| == Description:==
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| SR indicates higher or lower than actual NO: 6.3.4.9.12 neutron population due to misadjusted SR discrimin.stor voltage.
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| Variations: None. Date: 8/24/86 Rev: 3 Selectable Steps Inputs Conssents
| |
| : 1. Select channel NIS-12A NIS-12A = 31 NIS-12B NIS-12B = 32
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| : 2. Select discriminator 0-200 percent voltage
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| : 3. Select timo 0-99,999 see Brief Plant Response:
| |
| The affected channel of SR will read significantly different than the other channel. If discriminator voltage is too high, it will read very low. If discriminator voltage'is too low, it will read very high.
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| Suggested Instructor Action:
| |
| When told to adjust the voltage to proper value, clear the malfunction.
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| Events: None a
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| 0129w:4- 312M/85M/18 8/86
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| | |
| BRAIDWOOD' SIMULATOR' MALFUNCTION.
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| .)
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| ,A..
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| i
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| | |
| ==Title:==
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| LeakiInto Guide Tube for Incore Detector- ID: NIS-15 .
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| l ii NO: 6.3.4.9.15
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| ]
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| == Description:==
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| ]
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| d 4
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| Yariations: ,
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| None. Date: 1/7/88
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| .Rev: 4 Selectable Steps. Inputs Comments
| |
| : 1. Select rate 0-10) spa
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| : 2. Select ramp time O to 99,999 sec
| |
| .Brief Plant Response:
| |
| ' if] '
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| Plant would exhibit all characteristics of a primary coolant leak.
| |
| Suggested Instructor Action:
| |
| Insure student attempts to find the source of the leak and follows appropriate procedures.
| |
| Events:
| |
| : 1) NRC Information Notice 87-44: Thiele Tube Thinning l
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| l l
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| l
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| \
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| 0129w:4 312N/85N/22 1/88 p.
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| L _ ____--_._____ __ _
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| A)TS-l5 SSINS No.: 6835 IN 87-44 O
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| [V UNITED STATES NUCLEAR REGULATORY COMMISSION OFFICE OF NUCLEAR REACTOR REGULATION SEP2.1gn,
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| '^
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| WASHINGTON, D C. 20555 September 16, 1987 NRC INFORMATION NOTICE NO. 87-44: THIMBLE TUBE THINNING IN WESTINGHOUSE REACTORS Addr'essees:
| |
| All pressurized water reactor facilities employing a Westinghouse nuclear steam supply system (NSSS) holding an operating license or a construction permit.
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| | |
| ==Purpose:==
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| This information notice is being provided to alert addressees to potential problems resulting from thimble tube thinning in Westinghouse reactot... It is expected that recipients will review the information for applicability to their facilities and consider actions, if appropriate, to avoid similar problems. However., suggestions contained in this information notice do not
| |
| /~T constitute NRC requirements; therefore no specific action or written response C/ is required. -
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| Description of Circumstances:
| |
| During the recent refueling outage at North Anna Unit 1, eddy current (EC) testing identified wall thinning on approximately 23 out of 50 thimble tubes.
| |
| The wall degradation occurred on the thimble tubes just above the lower core plate, between the lower core plate and the fuel assembly guide tubes. Several thimble tubes with greater than 35% wall thinning were identified, with one thimble tube thinned as much as 49%.
| |
| Discussion:
| |
| The movable incore neutron detectors travel within retractable thimble tubes.
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| The thimble tubes normally extend (as indicated in Attachment 1) from a 10 path transfer device, through the seal table, through the bottom of the reactor vessel, and into selected fuel assemblies. The thimble tubes are supported by guide tubes within the lower vessel region and the fuel assemblies, and by high pressure conduits between the reactor vesse_1 and the seal table.
| |
| The thimble tubes are sealed at the leading (reactor) end, but are open at the 10 path transfer device to allow insertion of an incore neutron detector.
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| i
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| {
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| l 8709100056' .
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| ]
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| l l
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| *~
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| IN 87-44 September 16, 1987
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| (_. Page 2 of 3 l
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| V]
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| l Mechanical high pressure seals, iccated at the seal table, are used to seal the area between the thimble tube and the high pressure conduit. This seal serves as- a reactor coolant system (RCS) pressure boundary since the area between the thimble tube and the high pressure conduit is at RCS pressure.
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| Consequently, a leak in a thimble tube results in degradation of the RCS pressure boundary by creating a path for reactor coolant to bypass the mechanical seal. In order to halt the flow of leaking reactor coolant, the manual isolation valve must be closed.
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| As indicated, the thimble tubes ,are . supported over most of their length.
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| However, a small portion'of'.the thimble tube is'directly' exposed 'to RCS"~'
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| - flow.3Jhis exposed portion is between the top of the lower core plate
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| , and tne bottom of the fuel assembly. This region is approximately 18.4
| |
| .to 34.8 mm in length, depending.o,n the, reactor _ type. It'is believed $ hat)
| |
| _ .. flow-induced vibration on this exposed portion causes fretting at~the1 adjacentguidetubjs,.'}
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| ~
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| J Undetected thinniInHfXthimtile tube co.u.1d lead to the development ~of a7
| |
| ~~
| |
| ; non-isolable 1e'ak and a corresponding loss _of reactor coolant! As discussed previously, the manual isolation valve woult' have to be closed to halt the flow of leaking reactor coolant. The leaking coolant may create an environ-ment in the vicinity of the isolation valves too hazardous for personnel to enter.
| |
| Leaking thimble tubes could result in degradation of the incore neutron moni-toring system. If not isolated, reactor coolant from leaking thimble tubes can flow into the 10 path transfer device, allowing coolant to flood the other thimble tubes originating from that device. This could result in rendering inoperable more than just the leaking tube.
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| ~
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| In addition to North Anna Unit 1, incore thimble' tube thinning and leakage has been detected at facilities in France and Belgium. In this country, leaks in thimble tubes are known to have occurred at Salem Unit 1. In Licensee Event Report (LER) 81-023, Public Service Electric & Gas Co. (PSE&G) reported that three incore thimble tubes were known to have developed leaks because of fretting. One of these leaks resulted in the flooding of all six 10 path transfer devices, partially or completely flooding all the thimble tubes in the reactor. In addition, thinning has been detected on the Farley thimble tubes.
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| At North Anna Unit 1, the proposed corrective a:.tiv, was to retract selected thimble tubes approximately 2 inches. This wou?d move the thinned area out of the region of high turbulence. In addition, the thimble tube that experi-enced the most degradation will be taken out of service by closing the corre-sponding isolation valve.
| |
| : l. .:
| |
| L-IN 87-44 f;... September 16, 1987 i
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| Page 3 of 3'
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| :ij No specific action or written response is required by this information notice.
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| If you have any questions.about this matter, please contact the Regional Administrator of the appropriate regional office or this office.
| |
| Charles E. Rossi, Director -
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| Division of Operational. Events Assessment Office of Nuclear Reactor Regulation o Technical
| |
| | |
| ==Contact:==
| |
| Jack Ramsey, NRR (301) 492-9081 Attachments:
| |
| : 1. Typical Westinghouse Incore Neutron Monitoring System
| |
| : 2. List of Recently Issued NRC Information Notices a
| |
| a
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| _ _ _ _ _ _ _ _ _ _ _ _m. _ . __. _
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| 'Bh41DWOOD SIMULATOR-MAL? UNCTION LISTING er 1 PROCESS CONTROL SYSTEM
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| .\ I
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| 'PCS Reactor Trip Breaker Failure PCS-2 First Stage Pressure Transmitter Failure PCS-3 Steam Generator Level Control Failure PCS-4 Unstable Steam Generator Level Controller PCS-5 Inadvertent Safety Injection Actuation PCS-6 Inadvertent Containment Isolation (Phase A)
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| PCS-7 . Inadvertent Containment Isolation (Phase B)
| |
| -PCS-8 Failure of Safety Injection to Actuate PCS-9 Failure of Steam Generator Level Program PCS-10 Failure of Feedwater Isolation to Actuate
| |
| .\
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| 4 638M/263M/10 E/87 l
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| - _ _ _ _ - - _ _ _ - - - - - - - _ - _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ . __ _ .i
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| BRAIDWOOD SIMULATOR MALFUNCTION
| |
| | |
| ==Title:==
| |
| Reactor Trip' Breaker Failure- ID: PCS-1 NO: 6.3.4.10.1 i
| |
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| == Description:==
| |
| Failure of reactor trip breaker l
| |
| H undervoltage coil contacts Variations: None. Date: 7/12/87 i Rev: 3 l
| |
| l.
| |
| Selectable Steps Inputs Comments
| |
| : 1. Select faulty trip 1, 2, 3, or 4 1 = RTA breaker 2 = RTB 3 = BYA 4 = BYB Brief Plant Response: (IC-17.-100%, all systems in automatic)
| |
| The selected reactor trip breaker opens causing all the rods to fall to the bottom of the core. Negative rate trip opens the second resctor trip breaker. The first annunciator received is PR FLUI RATE HIGH RX TRIP.
| |
| Plant response will be a normal reactor trip response.
| |
| Suggested Instructor Action:
| |
| Clear malfunction when operator is ready to commence reactor startup.
| |
| Events: None 319N/85M/2 7/87
| |
| | |
| n- _ - _ _ - _ _ _ _ - _ . _ . _ _ - _ - _ . - _ _ - - __ - . -
| |
| 'BRAIDWOOD SIMULATOR MALFUNCTION 1.',
| |
| | |
| ==Title:==
| |
| First Stage' Pressure Transmitter Failure ID:: .PCS-2
| |
| (( ) NO: 6.3.4.10.2
| |
| | |
| == Description:==
| |
| ~ ' Selected pressure transmitter falls to selected value.
| |
| -Variations: None. Date: '9/20/88 Rev: 61 Selectable
| |
| -Steps. Inputs. Comments k 1. Select. transmitter 'PCS-2A PCS-2A = PT 505 PCS-28 PCS-2B = PT 506 2.
| |
| Select final value .0-120 percent '
| |
| 'Q)
| |
| '("] 3. Select raiap time- 0-99,999 sec Brief Plant Response: (IC-17, 100%, all systems in automatic)
| |
| Note: 1. PT 505 serves the steam dumps as T ref*
| |
| : 2. PT $06 serves the steam dump loss of load interlock (C-7).
| |
| : 3. Both channels feed P-13.
| |
| : 4. Selector switch for input to rod control, S/G level, and C-5 block on auto rod withdrawal.
| |
| PT 505 fails low-control rods insert, pressurizer level and pressure decrease. T59 first annunciators received are TAVE CONT DEV HIGH and the C-5 BYPASS PERM LIGHT.
| |
| PT 505 fails high - control rods withdraw to C-11 setpoint. No annunciator received.
| |
| 319M/85M/3 9/88
| |
| | |
| BRAIDWOOD SIMULATOR MALFUNCTION
| |
| : Tit'le : - First Stage Pressure Transmitter Failure'
| |
| 'ID: PCS-2 ~
| |
| Brief Plant-Response (continued): l I
| |
| PT 506-falls low - The steam dumps are armed'and the only annunciator is the C-7 BYPASS PERM LIGHT. !
| |
| l
| |
| -l
| |
| , .PT 506 fails high - only indication is that the meter fails high. l l
| |
| l 1
| |
| l Suggested Instructor Action: l I
| |
| 1 Clear the malfunction when repairs have been completed.
| |
| I Events: ,1); DVR 06-01-88-062: PT-505 Fails High i
| |
| 2
| |
| :\ ,
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| i i i l:- !
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| l l l
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| i l
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| 1 l
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| 319M/85M/4 -9/88 l
| |
| | |
| f" 1 - 2.
| |
| OEVIATION I;;Vi$TIGATIO3 REPORT TITLE PaGE TURRINE IWULSE P9EttuRE TRANSMITTER PT-50s FAILED HICH 1 10FI O l2
| |
| - EVENT DATE DIR NUMBER REPORT CATE SEQUENTIAL REVISION MONTH DAY YEAR STA UNIT YEAR y// NUMa[R g// NUMa[R MONTH DAY iEAR 1 POWER of 4 11 a al a el 6 -ol 1 a la -
| |
| o 16 1 2 -
| |
| oIo I i l I al 7 CONTACT FOR THft DIE NAME TELEPHONE NUMaER AREA CODE Rocer Flahive. Technical Staff tunervisor Ext. 2241 aI1 l1 2 l 114 l -l t 14I4l 1 COMPLETE ONE LINE FOR EACH COMPoalE fat URE DEttRfaED IN THit REPORT CAUSE SYSTEM COMPONENT MANUFAC- REPORTABLE CAUSE SYSTEM COMPONENT MANUFAC- REPORTABLE TUREE TO NPRD1 TURER TO NPRD1 Y tIa i IP 11 I 12 to 14 Y l l I l l 1 1 1 1 1 I I I I I l l l l SUPPLEMENTAL REPORT EXPECTED MONTH DAY YEAR SUBMISSION I YES fit van. connlete EXPECTED mnMf tsf 0N DATE) [1 No ! ! i TEXT A. PLANT CONDITIONS PRIOR TO EVERT:
| |
| Event Date/ Time 4/la/as / nats Unit 1 MODE 1 - Power nneration Rx Power 87 RC5 [A8] Temperature / Pressure Normal eneration Unit 2 MODE 1 - Power Oneration Rx Power 94 RCS [AB) Temperature / Pressure Normal Onern. tion
| |
| : 8. DESCRIPTION OF EVENT:
| |
| On April 18. 1988 at 0809 hours, while operating at 87% Reactor Power. Unit 1 experienced difficulties with the Turbine Impulse Pressure Transmitter 1PT-505. The problem was discovered during the routine performance of Train "B" Solid State Protection System B1 Monthly Surveillance 1805 3.1.1-21 when the Pressure Transmitter was noted to be failed high.
| |
| The failure of 1PT-505 caused the Tavg - Tref Deviation Alarm to annunciate. and Limiting Condition for Operations Action Requirement (LCDAR) 1805 3.1-la. Action Number 8 was entered at 0816 hours. The licensed personnel in the Control Room insnediately implemented Byron Abrormal Procedure 150A INST-2 as required.
| |
| All affected bistables were verifted to be in their required positions by 0825 hours per 80A INST-2
| |
| " Operation With A Failed Instrument Channel", and Station Nuclear Work Request (NWR) number E55120 was generated to the Instrweent Maintenance Department to investigate and repair the f ailed pressure j transmitter.
| |
| Maintenance activities were completed on April 19, 1988 and post maintenance testing was performed by Operating Department personnel to document continued component operability. LCOAR 1905 3.1-la. Action Number 8. was exited on April 19. 1988, at 2110 hours, with the affected b1 stables being reset at that time.
| |
| There were no systems, subsystems or components considered inoperable at the beginning of this eveent which
| |
| [ would have contributed to or exacerbated this event. No manual or automatic sdfety system actuations k occurred during the event. All operator actions taken throughout the event were Correct. Plant conditions remained stable throughout the event.
| |
| I I
| |
| (0001R/0001R) l .- .
| |
| | |
| r OEv!ATION INVESTIGATION REPORT TEXT CONTINUATION TITLE DIR NUMBER PAGE SEQUENTIAL REVISION
| |
| ["g _ IIA,, jl NIT YEAR NUMBER NUMBER TURafME IMPULSE PRESSURE TOSMlllilt #T-505 FAILED HIC.H 01 6 01 1 al B -
| |
| Ol6 l 2 ~~
| |
| 0l0 2 0F 0f2 TEXT C. CAUSE OF EVENT:
| |
| The root cause of this pressure transmitter failure is indeterminate. However, during maintenance troubleshooting activities, evloence of moisture was noted in the terminal junction box for PT-505. The moisture had no readily apparent source, as all associated conduits leading into it were noted as being dry.
| |
| D. SAFETY ANALYSIS:
| |
| There were no safety consequences resulting from this event which would have adversely impacted plant or public safety. At the time of this occurrence. Operating Department personnel entered the appropriate Dperating Abnormal Procedure. BDA INST-2. and aligned the system for correct and safe interim operation.
| |
| E. CORRECTIVE ACTIONS:
| |
| The moisture noted in the terminal junction box for IPT-505 was dried out and the pressure transmitter internals were also inspected for signs of moisture with none being evidenced.
| |
| Associated 0-Rings for the pressure transmitter were replaced in kind and the calibration of the component was verified to be within tolerance and indicating properly. No further corrective action is planned at this time.
| |
| ! ''g F. PREVIOUS OCCURRENCES:
| |
| D There have been previous Deviation Reports written against failed pressure transmitter channels. However to date. all instances have had a definite mode of failure and none were determined to be of an indeterminate root cause.
| |
| DVR NL9eER 11ILE 06-01-85-353 It S/G Level Loop 529 Failure Due to a Circuit Card Failure 06-01-86-002 1C Steam Generator Loop Failure 06-01-86-101 Failure of Steam Pressure Instrimient 1PI-524A 06-01-86-207 10 Steam Generator Pressure Channel 544 Failure Low 06-01-88-031 5/G 1B Pressure Channel 525 Failure Due to toad / Lag and Multiplier / Divider Card Failures G. CtMPONENT FAILURE DATA:
| |
| a) MANUFACTURER NOMENCLATURE MODEL NLDSER MFG PART NUMBER ITT Barton Pressure Transmitter 753-2706 N/A b) RESULTS OF NPRD1 SEARCH:
| |
| An NPRDS search was not conducted, since the data obtained would not land itself useful in this case where the root cause was determined to be indeterminate.
| |
| C) RESULTS OF NWR $EARCH!
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| A review of,"T.1M History
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| * file for the Unit 1 and 2 Turbine Impulse Pressure Transmitter 1/2 PT-505 indicated only calibration histories of the transmitters t/ 1 (0006R/0001R)
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| BRAIDWOOD SIMULATOR MALFUNCTION L[]
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| ==Title:==
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| Steam Generator Level Control Failure ID: PCS-3 NO: 6.3.4.10.3
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| == Description:==
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| S/G 1evel transmitter failure.
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| -Variations: None. Date: 4/9/89 Rev: 8 Selectable Steps Inputs Comments
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| : 1. Select channel PCS-3A - 3A = LT-519 PCS-3H 3B = LT-556 3C = LT-529 3D = LT-557 3E = LT-539 3F = LT-558 3G = LT-549 O 3R = LT-559 V .
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| : 2. Select level 0-100%
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| : 3. Select ramp time 0-99,999 sec O 879M/1 5/89
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| BRAIDWOOD SIMULATOR MALFUNCTION
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| - (~ E
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| | |
| ==Title:==
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| Steam Generator Level Control Failure ID: PCS-3
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| (
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| Brief Plant Responses (IC-17, 100%, all systems in automatic, S/G 1evel fails high)
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| Note: S/G 1evel selector switch must be selected to proper channel for the malfunction to have the desired effect.
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| Steam generator level increases or decreases as appropriate, as long as level control'is in automatic. Manual control of the feed reg. valve is possible.
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| Possible reactor trip on S/G low-low level.
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| Possible turbine trip, feedwater pump trip on S/G high-high level. The first annunciator received is S/G LVL DEV HIGH/ LOW.
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| Suggested Instructor Action:
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| r' 's G -
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| Clear malfunction after operator has demonstrated proper manual control.
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| Events:
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| : 1) DVR 06-01-85-353: S/G 1evel failure
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| : 2) LER 06-02<-87-015: S/G 1evel detector isolated caused P-14.}}
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