ML18138A072
| ML18138A072 | |
| Person / Time | |
|---|---|
| Site: | Surry, North Anna  |
| Issue date: | 03/07/1980 |
| From: | James O'Reilly NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | Ferguson J VIRGINIA POWER (VIRGINIA ELECTRIC & POWER CO.) |
| References | |
| NUDOCS 8003140416 | |
| Download: ML18138A072 (15) | |
Text
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Reply Refer To:
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UNITED STATES NUCLEAR REGULATORY COMMISSION REGION II 101 MARIETTA ST., N.W., SUITE 3100 ATLANTA, GEORGIA 30303 MARO 7 1980.
Virginia Electric and Power Company Attn: J. H. Ferguson Executive Vice President-Power P.O. Box 26666 Richmond, VA 23261 Gentlemen:
C?F
. This Information Notice is provided as an early notification of a possible significant matter. It is expected that recipients will distribute this Notice to their operating personnel and will review the information for possible applic-ability to their facilities.
No specific response is requested at this time.
However, we anticipate that further NRC evaluations will result in issuance of an Addendum to IE Bulletin 79-27 in the near future which will recommend or request specific applicant or licensee actions.
If you have questions regarding the matter, please contact the ~irector of the appropriate NRC Regional Office.
Sincerely
~fv*
ames P. O'Reilly
Enclosures:
- 1.
IE Information Notice No. 80-10 w/its Enclosure
- 2.
List of IE Information Notices Recently Issued
-n I
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Virginia Electric and Power Company cc w/encl:
W.R. Cartwright, Station Manager Post Office Box 402 Mineral, Virginia 23117 P. G. Perry Senior Resident Engineer Post Office Box 38 Mineral, Virginia 23117 W. L. Stewart, Manager Post Office Box 315 Surry, Virginia 23883 MAR O 7 1980.,,.
- r.
UNITED STATES NUCLEAR REGULATORY COMMISSION OFFICE OF INSPECTION AND ENFORCEMENT WASHINGTON, D.C.
20555 March 7, 1980 SSINS No.:
6870 Accession No.:
8002280640 IE Information Notice No. 80-10 PARTIAL LOSS OF NON-NUCLEAR INSTRUMENT SYSTEM POWER SUPPLY DURING OPERATION Description of Circumstances:
This notice contains information regarding Crystal River Unit 3 response to a loss of non-nuclear instrumentation (NNI) as a consequence of loss of the
+24 volt power supply to the NNI.
At 2:23 p.m. on February 26 with Crystal River Unit 3 at 100% power, the
+24 volt power supply to the NNI was lost, due to a short to ground.
This initiated a sequence of events (detailed in the enclosure) wherein the PORV opened and stayed open as a direct result of the NNI power supply loss.
HPI initiated as a result of depressurization through the open PORV, and with approximately 70% of NNI inoperable or inaccurate, the operator correctly decided that there was insufficient information available to justify terminating HPI.
Therefore, the pressurizer was pumped solid, one safety valve lifted, and flow through the safety valve was sufficient to rupture the RC Drain Tank rupture disk, spilling approximately forty-three thousand gallons of primary water into containment.
The Crystal River 3 event is closely related to the November 10, 1979 event at Oconee Unit 3 wherein the inverter supplying power to the Integrated Control System (ICS) and to parts of the NNI failed.
That event was the subject of IE Information Notice 79-29 (November 16, 1979) which was followed by IE Bulletin 79-27 (November 30, 1979).
The CR-3 event involved loss of only par_t of the power available from an inverter, rather than the inverter itself, since the +24v supply is only one of several power supplies drawing power from one inverter.
The effects are very similar, however, in that the ICS lost part of its input signals in both events.
The +24 volt power supply short to ground has tentatively been identified by the licensee to have occurred between knife edge connectors of a Bailey Control Company Voltage Buffer Card.
The voltage buffer card was misaligned in its receptacle, and adjacent connectors carrying +24v and "common" were bent such that they contacted one another.
This short circuit cleared itself during subsequent re-energizing of the power supply by burning through the foil on a printed circuit card.
Subsequent review by the licensee identified a second voltage buffer card which was also misaligned but had not caused a short circuit.
The specific circuit cards which were misaligned carried part number 662460911.
The connectors on these cards are slightly thinner and appear to have a somewhat different angle than those found on similar cards elsewhere in the NNI which carry part numbers 6624608Al or 6624609Al.
The 6624609LI cards appear to be more subject to misalignment.
e IE Information Notice No. 80-10 March 7, 1980 Page 2 of 2 The specific shorted voltage buffer card provided the signal to the NNI "x" saturation meter.
Licensees which utilize Bailey Control Company Voltage Buffer Cards are requested to carefully inspect the cards for possible misalignment and take corrective actions if misalignments are identified.
Specific instructions for carrying out these inspections and providing any other information which may be required to define appropriate corrective action is being prepared by Baily Control Company for transmittal to purchasers of this equipment by March 11, 1980.
Initial screening of IE Bulletin No. 79-27 responses indicates a range of responses regarding depth and scope of review.
IE Bulletin No. 79-27 was intended to cause licensees to investigate loss of individual power supplies as well as total loss of an inverter or vital bus.
An addendum to IE Bulletin No. 79-27 is planned to be issued in the near future to reflect the CR-3 event.
This Information Notice is provided to inform licensees of a possibly significant matter.
It is expected that recipients will disseminate the information to all operational personnel working at their licensed facilities.
(A meeting was held with B&W licensees in NRC Headquarters on March 5, 1980 to review the event at Crystal River and to discuss proposed corrective actions. Responses to specific questions have been requested of the B&W licensees.) If you have questions regarding this matter, please contact the Director of the appropriate NRC Regional Office.
No written response to this Information Notice is required.
Enclosure:
Sequence of Events
EVENT SYNOPSIS e
e 26 FEBRUARY TRANSIENT CRYSTAL RIVER UNIT-3 SEQUENCE OF EVENTS At 14:23 on February 26, 1980, Crystal River-3 Nuclear Station experienced a reactor trip.
Nominal full power primary and secondary system parameters were present.
A synopsis of key events and parameters was obtained from the plant computer's post-trip review and plant alarm summary, the sequence of events monitor, control room strip charts, and the Shift Supervisor's log.
The reactor was operating at approximately 100% full power with Integrated Control System (ICS) in automatic.
No tests were in progress.
Time 14:23:21 14:23:21 Event
+24 Volt Bus Failure (NNI power loss "X" supply)
PORV and Pressurizer Spray Valve Open Cause/Comments The positive 24 VDC bus 'shorted dragging the bus voltage down to a low voltage trip condition.
There is a built-in 1/4 to 1/2 second delay at which time all power supplies will trip.
The trip indication on negative(-) voltage was missed by the annunciator.
Following the NNI power failure, much of the control room indication was lost.
Of the instrumentation that remained operable, transient conditions made their indications questionable to the operators.
When the positive 24 VDC supply was lost due to the sequence discussed above the signal monitors in NNI changed state causing PORV/Spray valves to open.
The PORV circuitry is designed to seal in upon actua-tion and did so.
The resultant loss of the negative 24 VDC halted spray valve motor operator, and prevented PORV seal-in from clearing on low pressure. It is postulated that the PORV opened fully and the spral valve stroked for approximately 1/2 second.
The "40% open" indication on spray valve did not actuate, therefore, the spray valve did not exceed 40% open.
q' 26 February Transient CR-3 2 -
Time 14:23:21 14:23:35 14:24:02 14:24:02 Event Reduction in Feedwater Reactor Trip/Turbine Trip Hi Pressure lnj.
Reg. (Flag) loss of Both Condensate Pumps Cause/Comments As a result of the "X" power supply failure many primary plant control signals responded erroneously.
T-cold failed to 570°F (normal indication was 557°F) producing several spurious alarms.
T-ave failed to 570°F (decreased).
The resultant T-ave error modified the reactor demand such that control rods were withd~awn to increase T-ave and reactor power.
The power increase was terminated at 103% by the ICS and a "Reactor Demand High Limit" alarm was received T-hot failed to 570 F (log) and RC flow failed to 40 X 10 lbs/hr in each loop (low).
Both these failures created a BTU alarm and limit on feed-water which reduced feedwater flow to both OTSG's to essentially zero.
Turbine header pressure failed to 900 psig (high) which caused the turbine valves to open slightly to regulate header pressure thus increasing generated megawatts.
These combined failures resulted in a loss of heat sink to the reactor initiating high RCS pressure condition.
Rx trip caused by high RCS pressure at 2300 psi.
Turbine was tripped by the reactor.
This was a computer printout and indicates F 50° subcooling.
(The lowest level of subcooling was 8°F for a very.short period of time, at about 14:30)
Suspect'that the condensate pump tripped due to high De-aeriating feed tank level.
26 February Transient CR-3 Time 14:24:02 (Continued) 14:25:50 14:26:41 14:26:54 14:27:20 Event PORV Isolated HPI Auto Initiation RC Pumps Shutdown RB Pressure Increasing 3 -
Cause/Comments This is verified by a series of questions marks???? printed by computer indicating that the level instrument was over-ranged.
A low flow indication in the gland steam condenser was also indicated by computer.
At this time a high RC Drain Tank level alarm was received.
This was resultant from the PORV remaining open and,was positive indication that the PORV was open.
At this time, the operator closed the PORV block valve due to RCS pressure decreasing and high RCDT level.
HP! initiated automatically due to low RCS pressure of 1500 psig.
The low pressure condition was resultant from the PORV remaining full open while the plant was tripped.
Full HP! was initiated with 3 pumps resulting in approximately 1100 gpm flow to the RCS.
At this time, all remaining non-essential reactor building (RB) isolation valves were closed per TM! Lessons Learned Guidelines.
Operator turned RC pumps off as required by the applicable emergency procedure and B&W small break guidelines.
This is first indication that RCDT rupture disc had ruptured.
RB pressure increase data was obtained from Post Trip Review and Strip Chart indication.
26 February Transient CR-3 4 -
Time 14:31:32 14:31:49 14:31:59 14:32:14:41 14:32:35 14:33 Event RB Pressure High OSTG "A" Rupture Matrix Actuation Main Feedwater Pump IA Tripped ES A/B Bypass Started Steam Driven Emergency Feedwater Pump Core Exit Temp. Verified Cause/Comments This alarm was initiated by 2 psig in RB.
This is attributed to steam release from RCDT.
Code safeties had not opened at this time based upon tail pipe temperatures recorded at 14:32:03 (Computer).
This occurred due to F 600 psig in OTSG "A".
The low pressure was caused by OTSG "A" boiling dry which was resultant from the BTU limit and failed power supply to OTSG "A" level transmitter.
This resulted in the closure of all feedwater and steam block valves which service OTSG "A".
Caused by suction valve shutting due to rupture matrix actuation in previous step.
Manually bypassed and HPI balanced between all 4 nozzles (Total flow approximately 1100 gpm-small break operating guidelines).
Started by operator to ensure feedwater was available to feed OTSG's.
The core exit incore thermocouples indicated the highest core outlet temperature value was 560°F.
RCS pressure was 2353 psig at this time, therefore, the subcooling margin at this time was I00°F.
Minimum subcooling margin for the entire transient was 8°F at 14:30. It is postulated that some localized boiling occurred in the core at this point as indicated by the self powered neutron detectors.
26 February Transient CR-3 5 -
Time 14:33:14:44 14:33:30 14:34:23 14:35:33 14:36:50 14:38:15 14:44:12 Event Started Motor Driven Emer-gency Feedwater Pump RC Pressure High (2395 psig)
RB Dome Hi Rad Level Attempted NNI Repower Without Success Computer Overload FWV-34 Closed NNI Power Restored Successfully e
Cause/Comments Same discussion as "Started Steam Driven Emergency Feed-water Pump."
At this point, pressurizer is solid and code safety lifts (RCV-8).
This is the highest RCS pressure as recorded on Post Trip Review.
Apparently, RCV-8 lifted early due to seat leakage prior to the transient and RCV-9 did not lift.'
RMG-19 alarmed at this point.
Highest level indicated during course of incident was 50 R/hr.
High radiation levels in RB caused by release of non-condensable gases in the pressurizer and coolant.
This resulted in spikes observed on de-energized strip charts.
Caused by overload of buffer.
Resulted in no further computer data until buffer catches up with printout.
This valve was closed to prevent overfeeding OTSG "B" beyond 100% indicated Operating Range.
NNI was restored by removing the "X"-NNI Power Supply Monitor Module.
This allowed the breakers to be reclosed.
At this time, it was observed that the "A" OTSG was dry, the pressurizer was solid (Indicated off-scale high),
RC outlet temperature indicated 556°F (loop A & B average),
and RC average temperature indicated 532°F (Loop A & B).
The highest core exit thermocouple temperature at
26 February Transient CR-3 6 -
Time 14:44:12 (Continued) 14:44:31 14:46: 10 14:51:57 14:52 Event RB Isolation and Cooling Actuation Bypassed HPI, LPI and RB Isolation and Cooling Rupture Matrix Actuation on OTSG-B HPI Throttled and RCS Pressure Reduced to 2300 psig Cause/Comments this time was 531°F.
RCS pressure was 2400 psig (saturation temp. at this pressure is 662°F.).
This data verified natural circulation was in progress and the plant subcooling margin was 131°F (based on core exit thermocouples).
At this time, RB pressu~e increased to 4 psig and initiated RB Isolation.
The operator verified all immediate actions occurred properly for HPI, LPI, and RB Isolation and Cooling.
The increasing RB pressure was resultant from RCV-8 relieving pressure due to continued HPI.
These "ES" systems were bypassed at this time to balance HPI flow and restore cooling water to essential auxiliary equipment (i.e.,
RCP's, letdown coolers, CRDM's etc.).
The actuation was resultant from a degradation of OTSG-B pressure.
Cold emergency feed was being injected into the OTSG at this time.
This matrix actuation isolated all feedwater and steam block valves to the B-OTSG and tripped the "B" main FW pump.
Both Emergency FW pumps were already in operation at this time.
B-OTSG level at this time was 70% (Operation Range).
At this time, the maximum core exit thermocouple temperature was 515°F, RCS pressure was 2390 psig.
26 February Transient CR-3 Time 14:52 (Continued) 14:53 14:56 14:56:43 14:57:09 Event Reestablished Letdown Opened MU Pump Recirc.
Valves Bypassed the A-OTSG Rupture Matrix an9 Reestablished Feed to the A-OTSG Bypassed the B-OTSG Rupture Matrix Cause/Comments Therefore, t};l.e subcooling margin was 147°F.
Natural circulation was in effect as verified previously.
All conditions had been satisfied to throttle HPI.
Therefore, flow was throttled to approximately 250 gpm to reduce RCS pressure to 2300 psig in order to attempt to reduce the flow rate through RCV-8 and into the RB.
At this time, the operator was attempting to establish RCS pressure control via normal RC makeup and letdown.
This was done to assure the MU pumps would have minimum flow at all times to prevent possible pump damage.
Feedwater was slowly admitted to the A-OTSG which was dry up to this point.
Feedwater was admitted through the Auxiliary FW header via the EFW bypass valves.
The feedrate was very slow in order to minimize thermal shock to the OTSG and resultant depressurization of the RCS.
RCS pressure control was very unstable at this time.
This was done to regain FW control of the B-OTSG.
Level was still high in this OTSG (approximately 65% Operating Range).
Therefore, feed was not necessary at this time.
The Main Steam Isolation valves were open in preparation for bypass valve operation (when necessary).
26 February Transient CR-3 8 -
Time 14:57:15 15:00-09 15:00-09 15:15 15 :17 15: 19 Event Established RC Pump Seal Return Reestablished Level in A-OTSG 77°F Subcooled "A" Loop 23°F Delta-T/Manned the Technical Support Center Declared Class "B" Emergency Opened Emergency FW Block to B-OTSG Cause/Comments This was done in preparation for a RCP start (when necessary) and to minimize pump seal degradation.
This verified feedwater was being admitted to the OTSG and made it available for core cooling via natural circulation.
Feed to this generator was continued with the intent of proceeding to 95% on the Operating Range.
This value was based upon "A" RCS loop parameters at this time.
The "A" loop was being cooled down at this time by the A-OTSG fill and the operator was attempting to equalize loop temperatures.
At this time, loop temperatures were nearing equalization.
This delta-Twas calculated from loop A & BT C's and core exit thermocouples.
This was done based on the fact there was a loss of coolant through RCV-8 into the containment and HPI had been initiated. All non-essential CR-3 personnel were directed to evacuate and contact of off-site agencies began.
Survey team was sent to Auxiliary Building.
At this point the A-OTSG level was increasing and the decision was made to commence filling the B-OTSG simultaneously.
The intent was to go 95% on both OTSG's without exceeding RCS cooldown limits (100 F/hr) while maintaining RCS pressure control.
26 February Transient CR-3 Time 15:26 15:50 16:00 16:07 16:08:04 Event Lo Level Alarm in Sodium Hydroxide Tank Terminated HPI Commenced Pressurizer Heatup Survey Team Report Shutdown Steam Driven Emergency FW Pump Cause/Comments This was resultant from the tank supply valve opening when the 4 psig RB isolation and cooling signal actuated.
The sodium hydroxide was released to both LPI trains.
Sodium Hydroxide was admitted to the RCS via HPI from the BWST.
(Approximately 2 ppm injected into the RCS.)
At this time, all conditions had been satisfied (per small break ooerating guidelines) to terminate HPI.
RCS pressure control had been established using normal makeup and letdown.
HPI was terminated and essentially all releases to the RB were discontinued.
At this time, RCS pressure and temperature were well under control.
Natural circulation was functioning as designed (approximately 23°F delta-T).
RCS temperature was being maintained at approximately 450°F.
RCS pressure was approximately 2300 psig.
The decision was made at this point to commence pressurizer heatup in preparation to re-establish a steam space in the pressurizer.
The Emergency Survey Team reported no radiation survey results taken offsite were above background.
The motor driven Emergency FW pump was running, therefore, the steam driven pump was not needed.
The plant remained in this condition for approximately 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, while heating up the pressurizer to saturation temperature for 1800 psig.
26 February Transient CR-3 Time 18:05 18:30 21:07 Event Established Steam Space in Pressurizer Terminated Class B Emergency Forced Flow Initiated in RCS 10 -
Cause/Comments At this point, pressurizer temperature was approximately 620°F.
Pressurizer level was brought back on scale by increasing letdown.
From this point pressurizer level was reduced to normal operating level and normal pressure was established via pressurizer heaters.
State and Federal Agencies notified.
The decision was made to re-establish forced flow cooling in the RCS at this time.
B&W and NRC were consulted.
RCP-IB and ID were started.
At this point, RCS parameters were stabilized and maintained at RC pressure-2000 psig, RCS temperature-420°F.
Pressurizer level-235 inches.
The plant was considered in a normal configuration.
IE Information Notice No. 80-10 March 7, 1980 RECENTLY ISSUED Information Notice No. 80-09 80-08 80-07 80-06 80-05 80-04 80-03 80-02 80-01 79-37 79-36 IE INFORMATION NOTICES Subject Possible Occupational Health Hazard Associated With Closed Cooling Systems For Operating Power Plants The States Company Sliding Link Electrical Terminal Block Pump Shaft Fatigue Cracking Notification of Significant Events Chloride Contamination of Safety Related Piping BWR Fuel Exposure in Excess of Limits Main Turbine Electro-Hydraulic Control System 8X8R Water Rod Lower End Plug Wear Fuel Handling Events Cracking in Low Pressure Turbine Discs Computer Code Defect in Stress Analysis of Piping Elbow Date Issued 3/7/80 3/7 /80 2/29/80 2/27/80 2/8/80 2/4/80 1/31/80 1/25/80 1/4/80 12/28/79 12/31/79 Enclosure Issued To All holders of power reactor OLs and near term CPs All power reactor facilities with,an OL or a CP All Light Water Reactor Facilities holder power reactor OLs and CPs All holders of Reactor OLs and to near term OL applicants All licensees of nuclear power reactor facilities and applicants and holders of nuclear power reactor CPs All BWR' s holding a power reactor OL or CP All holders of power reactor OLs and CPs All BWR Facilities holder power reactor OLs or CPs All holders of power reactor OLs and CPs All power reactor OLs and CPs All power reactor OLs and CPs