ML13333B323
| ML13333B323 | |
| Person / Time | |
|---|---|
| Site: | San Onofre  |
| Issue date: | 11/26/1985 |
| From: | Martin T NRC - INCIDENT INVESTIGATION TEAM |
| To: | Dircks W NRC OFFICE OF THE EXECUTIVE DIRECTOR FOR OPERATIONS (EDO) |
| References | |
| NUDOCS 8512180433 | |
| Download: ML13333B323 (8) | |
Text
November 26, 1985 MEMORANDUM FOR:
William 3. Dircks Executive Director of Operations FROM:
Thomas T. Martin, Leader San Onofre Incident Investigation Team
SUBJECT:
INVESTIGATION OF THE NOVEMBER 21, 1985, EVENT AT SAN ONOFRE UNIT 1
Investigation of the November 21, 1985, San Ono-fre Unit 1 plant trip continues.
The licensee and INPO are conducting parallel investigations.
The Incident Investigation Team (1IT) is working well. together and all NRC offices have been very supportive.
The licensee and onsite licensee staff have also been supportive, although the latter is somewhat apprehensive.
No written material has been shared with the licensee or INPO, but the substance of our frequent contacts with the licensee has certainly conveyed the majority of our initial conclusions.
The status of the investigation along with a preliminary sequence of events is provided in ATTACHMENT 1.
ATTACHMENT 2 is the prime hypothesis developed by and being pursued by the ITT, which to date, seems to best explain the significant events of which we are aware.
I request you limit the dissemination of that hypothesis to minimize unproductive speculation.
The event of greatest safety significance during the San Onofre plant trip appears to be the "B" Steam Generator Feedwater Pipe water hammer.
Lessons learned from the investigation will cause this licensee to modify equipment and associated maintenance programs, and may be of generic importance to other licensees.
Although the licensee's action plans do not schedule completion of certain equipment troubleshooting until as late as February, 1986, the ITT is on schedule to produce a quality report within the 45 day deadline.
Thomas T. Martin, Leader San Onofre Incident Investigation Team Attachments:
- 1)
PNO-IIT-85-2B
- 2) Prime Hypothesis
,/cc:
J. Heltemes, AEOD 8512180433 851126 DESIrTITATED ORIGINAL PDR ADOCK 05000206 Certified By d
S PDR
DATE: 11/26/85 PRELIMINARY NOTIFICATION OF EVENT OR UNUSUAL OCCURRENCE--PNO-IIT-85-2B This preliminary notification constitutes EARLY notice of events of POSSIBLE safety or public interest significance. The information presented is preliminary, requires further evaluation and is basically all that is known by the IIT on this date.
FACILITY: Southern California Edison Company Emergency Classification Unit 1 X Notification of Unusual Event Docket No. 50-206 Alert Site Area Emergency General Emergency Not Applicable
SUBJECT:
Status Report from NRC Incident Investigation Team The Incident Investigation Team (IIT) remains onsite gathering data, conducting interviews, inspecting equipment, meeting with the licensee, concurring in licensee action plans and analyzing facts. A preliminary sequence of events has been developed by the IIT and is attached. A set of preliminary hypotheses explaining the significant events has been developed by the IT and are being investigated.
All interviews should be completed on November 27, 1985. All licensee action plans for further troubleshooting and uncovering remaining event related information should be finalized on November 28, 1985. Assuming the combination of information possessed by the IIT and the licensee action plans to uncover additional facts appear adequate to project closure of significant open issues, the IIT intends to depart the site by December 1, 1985, and to reassemble in Bethesda, Maryland.
A final status report will be issued prior to the IIT's departure from the site.
CONTACT: T. Martin W. Lanning 714-492-2641 714-492-2641 DISTRIBUTION H St.
MNBB Phillips E/W Willste Air Rights Mail:
Chairman Pallidino EDO NRR IE NMSS ADM:DMB Comm. Zech PA R. Dudley OIA RES DOT:Trans Only Comm. Bernthal MPA AEOD Comm. Roberts ELD Comm. Asselstine Regions:
CA (Reactor Licensees)
PDR Resident Inspector
11/26/85 Rev 2 PRELIMINARY SEQUENCE OF EVENTS INCIDENT INVESTIGATION TEAM San Onofre Nuclear Generating Station Unit 1 Plant Trip Initial Conditions, November 21, 1985 Unit operating at 60% power Saltwater leak into main condenser, south circulating water pump stopped and numbers 1 and 4 water boxes removed from service Block valves for PORVs closed Troubleshooting of ground on 4160V bus 1C in progress Bus 1C removed from X winding of auxiliary transformer C and being supplied by bus 1A which was tied to auxiliary transformer A supplied by main generator.
Steam generator blowdown ongoing about 100 gpm per generator Transient Initiator Transformer C differential relay actuates to de-energize transformer 04:51 Circuit breakers 4032 and 6032 open to isolate transformer C from switchyard Circuit breaker 12CO2 opens to isolate transformer C from bus 2C Bus 2C and vital bus 4 de-energized Systems Response/Operator Actions to Loss of Power The following equipment de-energized:
East feedwater pump (G-3A)
Southeast condensate pump (G-1B)
Northeast condensate pump (G-1A)
East heater drain pump (G-36A)
North turbine plant cooling water pump Switchyard supply South saltwater coolng pump Auxiliary saltwater cooling pump Diesel generator number 2 starts automatically, and per design it does not load 04:51+
Operators manually trip reactor (per procedure), turbine trips Operators push unit trip button (opens main generator output breaker and trips turbine)
Turbine-driven auxiliary feedwater pump starts on low steam generator level and begins three minute warm-up cycle 04:51+
Circuit breakers 4012 and 6012 open resulting in loss of all
remaining 4160 busses (IA and 1C, 1B), 480 busses and the 120V utility bus All in-plant AC power lost Control room emergency lighting on East feedwater pump shaft seal drain trap vents blowing water Some card readers fail/operators bypass for access Diesel generator number 1 starts automatically, and per design does not load A loud "bang" is heard unknown Low pressure feedwater heater ruptured 04:53+
Automatic sequencer fails to realign power from busses IC and 2C to busses 1A and 2B.
Operator takes actions to realign breakers unknown Foam fire suppression system actuates near lube oil reservoir 04:54+
Turbine-driven AFW pump delivering 100-150 gpm to each steam generator. (Time estimated to be three minutes after reactor trip.)
unknown Operator unable to close circuit breaker 4012 04:55+
Circuit breaker 6012 closed by operator on second attempt Operators began to re-energize busses 1A and 1C, lB, and 2C 04:56+
Motor-driven auxiliary feedwater pump starts automatically after busses re-energized. Total flow about 135-150 gpm to each steam generator.
04:57 All 4160 and 480V busses energized Operators respond to safety injection actuation annunicator.
Operator verified no actuation occurred or required.
Operators begin to re-energize equipment.
Cooldown of reactor coolant system more rapid than expected.
Low pressurizer level (5%) and pressure (1900 psig)
Operator started south charging pump
North charging pump starts automatically Suction of both charging pumps switched automatically to RWST Operator stops AFW flow by closing throttle valves Operator increased AFW flow to about 25 gpm to each steam generator Plant equipment operator dispatched to close main steam block valves unknown Steam generator blowdown re-established to 100 gpm (radiation monitors reset)
Emergency response by fire brigade 05:06 Unusual event declared Prompt Notification Report made to NRC A loud "bang" was heard. Plant equipment operator heard water hammer and observed steam on turbine building mezzanine 05:08 Circuit breaker 4012 closed by operator 05:15 Started reactor coolant pump "B" Reactor cooling pump "B" high thrust bearing temperature alarm annunciated 05:23 Started reactor coolant pumps "A" and "C" Wide range level indication off-scale low in all three steam generators 05:25 Operators stopped reactor coolant pump "B" Flow increased to steam generators "A" and "C" from 25 gpm to 40 gpm each 5:30 Blowdown from steam generators secured Wide range water level indication on-scale in A and C steam generators Personnel wear steam suits in three attempts to identify source of feedwater leak unknown Feedwater leak identified on "B" steam generator from check valve FWS-378 06:00 Reactor coolant system cooled down to Mode 4
unknown Sandbags are placed at entrance to chemical feed room 08:35 Turbine-driven auxiliary feedwater pump secured 09:10 RHR suction valve MOV-813 interlock fails to clear although RCS pressure below 400 psig 09:12 Containment entry made to isolate hot leg injection 09:18 MOV-813 opened 09:20+
RHR flow established 9:40 Unusual event terminated, both RHR pumps in service 10:45 Feedwater leakage manually isolated 15:08 Entered Mode 5 unknown Containment entry found damaged pipe supports and insulation
ATTACHMENT 2 Prime Hypothesis Being Investigated Ground fault on "C"
Auxiliary Transformer Transformer trip deenergizes "East" Main Feedwater (MFW)
Pump "West" MFW Pump remains energized, due to alternate power source alignment from the Unit Main Electric Generator "East" MFW Pump discharge-check-valve fails to seat properly "West" MFW Pump over-pressurizes "East" Feedwater Heater-Condensate Train "East" #4-5 Feedwater Heater ruptures, due to over-pressurization Both Main Turbine "East" Rupture Discs fail Operator trips unit, deenergizing "West" MFW Pump Feedwater (FW) flow to all steam generators stop "B" Feedwater Regulating Valve discharge-check-valve fails to seat properly Electric Auxiliary Feedwater (AFW) Pump gets start signal, but lacks electric power to start Steam AFW Pump gets start signal, but takes three minutes to start delivering flow, due to.automatic warmup cycle "A",
"B" and "C" Steam Generator steam spaces are interconnected and begin blowing "B" Feedwater Line dry, thru the stuck open check valves ahead of the "B" Feedwater Regulating Valve and the "East" MFW pump Steam AFW Pump begins to deliver relatively cold Auxiliary Feedwater to all three Main Feedwater lines The Auxiliary Feedwater to the "B" Steam Generator Feedwater Line flows to both the malfunctioning check valves and to the long horizontal run of FW pipe in the primary containment Operators reduce AFW flow to minimize rate of cooldown of primary system Steam in "B"
Feedwater Line finally encounters cool water laying in the horizontal pipe, resulting in rapid condensation
Water slug forms in "B" Feedwater Line in containment, is accelerated by steam flow toward the point of condensation, encounters multiple pipe turns, and damages associated "B" Feedwater Line and Supports Water hammer pressure pulse stretches "B" Feedwater Regulating Valve Bypass Line check-valve top-hat-balots, extrudes gasket and produces substantial steam-water leak "B"
Steam Generator boils dry with all "B"
AFW -flow being carried by steam out check-valve leak