ML17334B323
| ML17334B323 | |
| Person / Time | |
|---|---|
| Site: | Cook  |
| Issue date: | 08/16/1989 |
| From: | Alexich M INDIANA MICHIGAN POWER CO. (FORMERLY INDIANA & MICHIG |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| AEP:NRC:1105, NUDOCS 8908240332 | |
| Download: ML17334B323 (25) | |
Text
ACCEIZRATZD D3 BU'DON DEMONSTR ATION SYSTEM REGULATORY INFORMATION DISTRIBUTION SYSTEM (RIDS)
ACCESSION NBR:8908240332 DOC.DATE: 89/08/16 NOTARIZED: NO DOCKET FACIL:50-316 Donald C.
Cook Nuclear Power Plant, Unit 2, Indiana 05000316 AUTH.NAME AUTHOR AFFILIATION ALEXICH,M.P.
(formerly Indiana
& Michigan Ele RECIP.NAME RECIPIENT AFFILIATION Document Control Branch (Document Control Desk)
SUBJECT:
Forwards summary of util planned response to 890814 reactor trip,per 890815 telcon w/Region III,NRR & AEPSC.
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TITLE: General (50 Dkt)-Insp Rept/Notice of Violation Response NOTES RECIPIENT ID CODE/NAME PD3-1 PD INTERNAL: AEOD AEOD/TPAD LOIS, ERASMIA NRR/DEST DIR NRR/DOEA DIR ll NRR/DREP/RPB 10 NUDOCS-ABSTRACT OGC/HDS1 RES MORISSEAU,D EXTERNAL: LPDR NSIC COPIES LTTR ENCL 1
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AEOD/DEIIB DEDRO NRR SHANKMAN,S NRR/DLPQ/PEB NRR/DREP/EPB 10 NRR/PMAS/ILRB12 OE~~
ERMAN,J EG FIQE 02 GN3 FILE 01 NRC PDR COPIES LTTR ENCL 1
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Indiana Michigan Power Company P.O. Box 16631 Columbus, OH 43216 Z
INDIANA MICHIGAhf POWER AEP: NRC: 1105 Donald C.
Cook Nuclear Plant Unit 2 Docket No.
50-316 License No.
DPR-74 EVALUATION OF AUGUST 14, 1989 UNIT 2 REACTOR TRIP U.
S. Nuclear Regulatory Commission Attn:
Document Control Desk Washington, D,
C.
20555 Attn:
A. B. Davis August 16, 1989
Dear Mr. Davis:
A conference call was held on August 15, 1989 among NRC Region III,
- NRR, and AEPSC representatives to address the August 14, 1989 Unit 2 reactor trip.
The attachment to this letter formally documents actions being taken or that are planned in response to the reactor trip. All of the indicated actions will be completed prior to the restart of Unit 2 with the exception of item 3.
This document has been prepared following Corporate procedures that incorporate a reasonable set of controls to ensure its accuracy and completeness prior to signature by the undersigned.
Sincerely, MD P. Al xich Vice President MPA/eh cc:
D.
H. Williams, Jr.
W
~
G. Smith, Jr.
- Bridgman R.
C. Callen G. Charnoff NFEM Section Chief NRC Resident Inspector
- Bridgman
ATTACHMENT TO AEP:NRC:1105 Page 1
The following information is provided to formally document actions being taken or that are planned as a result of the recent failure of a Unit 2 120 VAC control room instrumentation distribution (GRID) power supply and subsequent Unit 2 reactor trip.
1)
Root Cause Evaluation A root cause analysis of the GRID failure will be performed.
2)
Surveillance of Equipment Affected by GRID IV Failure Before returning to power operation, the necessary surveillances will be performed to provide assurance that equipment affected by the incident is reliable.
3)
Wide Range Steam Generator Level Indication Compliance with Regulatory Guide 1.97 Our position regarding this and other accident monitoring channels was documented in submittal AEP:NRC:0773AB dated October 5, 1988.
That submittal includes the regulatory basis for the Donald C.
Cook Nuclear Plant steam generator wide range level instrumentation.
We will, however, reconsider our position in conjunction with the NRC's SER associated with our submittal.
4)
Rod Bottom Lights The root cause of the rod bottom light failure will be evaluated to determine whether the failure is related to the GRID IV electrical failure.
5)
GRID Common Mode Failure Potential Once the root cause of the GRID IV failure has been established, an evaluation of the potential for that root cause to lead to common mode failure of the other GRID invertors will be conducted including consideration for bus transfer failures and necessary testing.
6)
Feedwater Check Valve Failure A root cause evaluation of the check valve failure will be performed including an examination of recent maintenance history.
7)
NRC Briefing Prior to Startup The NRC will be notified regarding the status of each of the above commitments prior to plant startup (except item 3).
Indiana Michigan Power Company ATTACHl'1ENT 82 IR 889025/89025 INDIANA NlCHlGAN POWER AEP:NRC:1105 Donald C.
Cook Nuclear Plant Unit 2 Docket No.
50-316 License No.
DPR-74 EVALUATION OF AUGUST 14, 1989 UNIT 2 REACTOR TRIP U. S. Nuclear Regulatory Commission Attn:
Document Control Desk Washington, D.
C.
20555 Attn:
A.
B. Davis August 16, 1989
Dear Mr. Davis:
A conference call was held on August 15, 1989 among NRC Region III, NRR, and AEPSC representatives to address the August 14, 1989 Unit 2 reactor trip,.
The attachment to this letter formally documents actions being taken or that are planned in response to the reactor trip.
All of the indicated actions will be completed prior to the restart of Unit 2 with the exception of item 3.
This document has been prepared following Corporate procedures that incorporate a reasonable set of controls to ensure its accuracy and completeness prior to signature by the undersigned.
Sincerely, M. P. Al xich Vice President MPA/eh cc:
D.
H. Williams, Jr.
W.
G. Smith, Jr.
- Bridgman R.
C. Callen G. Charnoff NFEM Section Chief NRC Resident.
Inspector
- Bridgman
ATTACHMENT NO.
3 IR No. 50-315/89025 50-316/89025 D.
C.
COOK UNIT 2 REACTOR TRIP 4:01 P.M.
AUGUST 14, 1989 SE UENCE OF EVENTS Time 3:40 p.m.
3:45 p.m.
3:45 p.m.
3:55 p.m.
4:01:24 p.m.
Event Power Range N-44 fuse (5 amp) blows; CRIO-IV transfers to alternate supply; 4
SSPS output relays "chatter" briefly N-44 bistables placed in "trip" per TS N-44'efused completely CRID-IV checkout for transfer to normal power CRID-IV pushbutton actuated to transfer power; Reactor trip-FIRST OUT: Reactor Coolant Pump 4 breaker "Open" relay dropout-Reactor Trip Note SUBSEQUENT:
. RTB "B" open RTB "A" open Turbine Trip - left channel Turbine Trip right channel Main feedwater Trip RTB undervoltage; "A" and "B" EHC system trip 0 seconds 4:01:26 4:01:26 4:01:28"30 4:01:30 (approx) 4:01:36-.38 Turbine stop valves closed Control rods in 3 "bottom" lights not illuminated; rod H-8 light came in a few
, seconds later, leaving two (B-4 and C-7) failed Auxiliary feedwater pump autostarts Steam generator low levels received-all channels Operators commence E-0 "Reactor Trip or Safety Injection" Steam generator low-low levels 1 second 2 seconds MFP trip
4:01:52 4:Ol:55 Main generator output breaker opens normally Operator manual turbine (solenoid) trip, followed by manual AMSAC trip All times hereafter are approx 4 02 Steam generator levels offscale low 4:02 4:03 4:03 Operators verify no SI required, commence ES-O.l, "Reactor Trip Response" Onshift STA reviews "Status Trees" no emergency indications Operator performs emergency boration (pumps already running) by opening boration valve 4 05 Control room vent fan started to clear out smell of smoke 4'10 4
15 4:15 4:05-4:15 Emergency boration secured Steam generator narrow-range levels back onscale Both source range NIs reenergize normally Operations in manual control:
letdown/charging fl'ow letdown pressure control CCW to letdown heat exchanger steam generator PORVs 4:16 Turbine AFW pump secured; RCS cooldown stopped at 535 degrees fahrenheit 4:20-4:30 4:30 4:30-4:35 4:36 Reactor and secondary plants stablilized in MODE 3 Turbine turning gear motor secured reported smoking RCS cooldown below 541 degrees fahrenheit to secure RCP No.
4 RCP No.
4 stopped due to absence of pump monitoring instrumentation 4:46 CRID-IV transferred to lighting panel power supply CRP-3, restoring normal voltage 4:46-5:30 CRID-IV instrumentation and control loads individually restored with minor exceptions; appropriate controls placed in automatic to maintain plant in standby status pending startup or cooldown decision
ATTACHMENT NO.
4 E UIPMENT AFFECTED BY GRID NO.
4 The following is a list of affected safety and nonsafety-related instrumentation and control components that were loss as the result of the opening of various breakers and fuses:
Breaker No.
Instrument Function Rack 12 (RPS)
MPP-212 MPP-242 8LP-110 SG No.
1 Pressure Channel SG No.
4 Pressure Channel SG Loop No.
1 Narrow Range Level Channel BLP-120 BLP-130 SG Loop No.
2 Narrow Range Level Channel SG Loop No.
3 Narrow Range Level Channel BLP-140 SG Loop No.
4 Narrow Range Level Channel NPS-153 PPP-300 PZR Pressure Transmitter Lower Containment Pressure Channel ELS-951 FFI-240 RWST Level Channel Auxiliary Feedwater Flow to SG No.
4 Channel
Breaker No.
Rack 13 (RPS)
Cab 22-CG4 (Control Rack)
Instrument NTP-141 NTP-140 NTP-241 NTP-240 FFI-241 ILA-131 ILA-141 Function Reactor Coolant Loop No.
4 RTD Thot Reactor Coolant Loop No.
4 RTD Thot Reactor Coolant Loop No.
4 RTD Tcold Reactor Coolant Loop No.
4 RTD Tcold Feed Water Flow to SG No.
4 Channel Accumulator Tank No.
3 Level Channel Accumulator Tank No.
4 Level Channel IPA-131 IPA-141 QTI-240 NTA-252 Accumulator Tank No
~
3 Pressure Channel Accumulator Tank No ~
4 Pressure Channel RCP Loop No.
4 Low Bearing Temperature RCP Loop No.
4 Seal No.
1 Temperature PZR Vapor Temperature
Breaker No.
Instrument Function IFI-54 QFA-240 QDA-40 Reactor Coolant= Loop No.
4 Cold Injection Flow Channel Seal Water Injection Flow Channel.
RCP Loop No.
4 Seal Water Flow Cab 23-CG4 (Control Rack)
ITR-311 QTC-302 NTA-152 IPA-310 QPC-301 IPA-250 IFI-310 QLC-452 IFI-.311 QRV-303 Residue Heater No.
1 RTD Letdown Heat Exchanger RTD PRZR Relief Discharge Temperature RHR Pump No.
1 Di scharge Pressure Channel Letdown Heater Low Pressure Channel Boron Injection Tank Pressure Channel Residue Heater No.
2 Outlet Flow Volume Control Tank Level Channel Residue Heater No.
2 Outlet Flow Letdown to CVT Diversion Valve
Breaker No.
Cab 24-CG4 (Control Rack)
Instrument FRV-240 CRV-470 QRV-301 TY-412C TY-422C TY-432C TY-442C TY-5050 TY-411D TY-421D TY-4310 TY-441D NPT-411 NPT-421 NPT-431 NPT-441 Function Loop No.
4 Feedwater Control Valve II II II II II I I II II I I II I I II II II SG Loop No.
1 Wide Range Level Channel SG Loop No.
2 Wide Range Level Channel SG Loop No.
3 Wide Range Level Channel SG Loop No.
4 Wide Range Level Channel Letdown Heat Exchanger CCW Valve Letdown Heater Control Valve Delta Temperature/TAVG Current to Current Converters
Breaker No.
Cab 25-CG4 (Control Rack)
Instrument Function Control Rod Bank A Position Control Rod Bank B Position Control Rod Bank C Position Control Rod Bank D Position Rod Insertion Recorder Bank A Limit Bank A Position Bank B Limit Bank B Position Bank C Limit Bank C Position Bank D Limit Bank D Position Cab B
Demultiplexer (ckt.
14)
K-0704 K-0705 K-0706 K-0707 Average Power Quadrant I Quadrant 2
Quadrant 3
Quadrant 4
Rod Control Automatic Rods In Automatic Rods Out Rod Speed Demand Incore Thermocouples Turbine Stop Valve Status Light Relays
Instrument Fuses Re laced Train B No.
2 48Y Power Supply Train B No.
2 15V Power Supply NR-44 SG-14 FRV-210 Function Solid State Protection System Solid State Protection System Power Range NI Recorder.
Overpower Recorder SG Water Level Control Valve Auto/Manual Station NRV-164 PZR Water Spray Valve Auto/Manual Station QRV-450 Boric Acid Transfer Pump Tank No.
2 Recirculation Manual Station Feedwater Differential Pressure Controller PZR Safety and Relief Valve Flow Monitor SG-31 Incore Thermocouple Train B
Recorder GRV-341 Nitrogen Supply to Accumulator Vent Valve Controller N-44 Power Range Control Power Fuses Comparator - Rate Drawer Control Power Fuses Audio Count Rate Drawer Audio Channel PWR Timer Sealer PWR
ATTACHHEQT NO.
5 IR No.50-315/89025; 50-316/89025 PROCEDURES REVIEMED 02-OHP 4023.E-O, "Reactor Trip or Safety'njection."
02-OHP 4023.ES-O. 1, "Reactor Trip Response."
- "2-OHP 4021.082.008, "Operation of CRID Power Supplies."
2-OHP 4022.013.004, "Power Range Malfunction."
"*2-OHP 4022.013.006, "Tripping of Protection Set Bistables."
2-OHP 4024.207, "Drops"81-100, "Reactor Coolant Annunciator."
2-OHP 4024.208, "Drops" 9, 34, 37, "Pressurizer Annunciator."
2-OHP 4024.219, "Drops" 30, "Station Auxiliary AB Annunciator."
2-OHP 4024.206, "Drops" 18, 19, 23, 24, "Residual Heat Removal Annunciator."
2-QHP 4024.213, "Drops" 4, 34, "Steam Generator 1 and 2 "
2-OHP 4024.214, "Drops" 4, 34, "Steam Generator 3 and 4."
2-OHP 4024.205, "Drops" 32, 34, "Containment Spray Annunciator."
OHI-5030 Attachment No. 2, Test No. 95, "Unit 2 Operations 5030 Surveillance."
- ~12 MHP 5021.001.071, "Inspection and Repair of Atwood and Morrill Swing Check Valves (except 2-CS-321)."
Mr. A. B. Davis AEP:NRC:1099 ACTIONS TO IMPROVE WATER CHEMISTRY TRENDING In February
- 1989, a task force was assembled to evaluate alternatives to the CMCP for use in long term trending of Cook Nuclear Plant water chemistry parameters.
The task force consists of personnel from the Cook Nuclear Plant Technical Physical Sciences Chemistry Section, AEPSC Chemical Engineering and Performance
- Section, and AEPSC Information Systems Department.
The result of the task force's efforts will be to identify a computer based program for long term trending of water chemistry parameters that is more conducive to fulfillingboth Cook Nuclear Plant and AEPSC needs for trend information.
Final recommendations for changes to the long term trending program will be made by, the task force on or before September 15, 1989.
In the interim period while the task force recommendations are being finalized and action to modify the long term water chemistry program is initiated, Cook Nuclear Plant will resume use of the CMCP for long term trending.
Long term trending (e.g.,
one year) of secondary water chemistry parameters using the CMCP will be performed on a monthly basis by Cook Nuclear Plant Technical Physical Sciences Chemical Supervisors and on a quarterly basis by the AEPSC Chemical Engineering and Performance Section.
The CMCP can produce trend graphs either on a terminal screen or by hard copy and is accessible at both Cook Nuclear Plant and Corporate offices.
These measures are considered adequate to provide the necessary long term trend information on an interim basis.
This document has been prepared following Corporate procedures that incorporate a reasonable set of controls to ensure its accuracy and completeness prior to signature by the undersigned.
Sincerely, M. P.
A exich
%See President MPA/eh cc:
D. H. Williams, Jr, W.
G. Smith, Jr.
- Bridgman R.
C. Callen G. Charnoff NFEM Section Chief NRC Resident Inspector
- Bridgman