NRC Generic Letter 1980-18
| ML031350323 | |
| Person / Time | |
|---|---|
| Issue date: | 03/06/1980 |
| From: | Harold Denton Office of Nuclear Reactor Regulation |
| To: | |
| References | |
| BL-79-027 GL-80-018, NUDOCS 8004040005 | |
| Download: ML031350323 (14) | |
REGE N, REUR LATORY noCKET FIL /
ow e4 UNITED STATES copy o
0r NUCLEAR REGULATORYCOMMISION -
WASHINGTON, D. C. 20555 C7L-C' -49 March 6, 1980
TO ALL OPERATING B&W REACTOR LICENSEES
As you know, on February 26, 1980, the Crystal River Unit No. 3 Nuclear Station (CR-3) experienced a reactor trip from approximately 100% full power. The initiating event was a failure in the power supplies for the non-nuclear instrumentation. A discussion of the event was presented by the Florida Power Corporation (FPC) in a meeting attended by representatives of your company in Bethesda, Maryland, on March 4, 1980. FPC also discussed the planned corrective action that would be taken at CR-3. The sequence of events presented by FPC and the planned corrective actions at CR-3 are attached to this letter as Enclosures 1 and 2 respectively.
Representatives from all other B&W operating plants were also present at the March 4, 1980 meeting. Each licensee addressed the history of non- nuclear instrumentation problems at his facility, the susceptibility of his plant(s) to the CR-3 event, and any corrective action that has been, or will be, taken.
On a related matter, Office of Inspection and Enforcement Bulletin 79-27 was issued subsequent to a similar event at the Oconee Nuclear Station, Unit No. 3, on November 10, 1979. This bulletin requested your review of certain matters relative to the Oconee event as they apply to your facility. Our interest in the CR-3 event, and its implication on the operation of your facility, does not relieve you of your responsiblities to provide the information requested by IE Bulletin 79-27.
Because of the implications of the CR-3 event, and potential adverse effects on the public health and safety that could result from future events of this type, we believe that certain information in addition to that requested in IE Bulletin 79-27, should be promptly provided to the NRC concerning your facility. In accordance with 10 CFR 50.54 (f), you are requested to provide us with information in response to Items 1 through 5 of Enclosure 3, submitted under oath or affirmation, no later than close of business March 12, 1980. Information in response to Items 6 and 7 of Enclosure 3 should be submitted no later than close of business March 17, 1980. The information provided in your responses will enable us to determine whether or not your license should be modified, suspended, or revoked.
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- 2- This letter confirms the oral request for this at the March 4, 1980 meeting by Mr. Darrell G. information expressed Director, Division of Operating Reactors. Eisenhut, Acting Sincerely,
- - Harold R. Denton, Director Office of Nuclear Reactor Regulation Enclosures: As stated
K-* ty. -Encli5Dure 1 Rev. 5 Page 1 SEQUENCE (AS OF 2300 3/1/80)
26 February Transient CR-3 EVENT SYNOPSIS
At 14:23 on February 26, 1980 Crystal River -3 Nuclear reactor trip from approximately 100% full power. A synopsis Station experienced a parameters was obtained from the plant computer's post-trip of key events and alarm summary, the sequence of events monitor, control review and plant the Shift Supervisor's log. room strip charts, and The react;r>vas operating at approximately 10O % full power with Integrated Control System (ICS) in automatic. No tests were in progress and minor main- tenance was being performed in the Non-Nuclear Instrumentation cabinet 'Y". (NNI)
Time Event Cause/Comments
14:23:00 The following is a summary of plant conditions prior to the trip Flux 98.6%
RC Pressure 2157 psig PZR level 202 inches MU tank level 71 inches To "SA" 599§F.
"B" 600F.
Tc "A" 5570F.
TC "B" 556F.
d Flow "A" 73 X 106 lbs/hr RC Flow "B" 73 X 106 lbs/hr Letdown Flow 48 gpm OTSG "A" lv1 COP) 67%
OTSG "B" lvl (OP) 65%
OTSG "A" FRLV 242 inches OTSG "B" FRLV 254 inches OTSG "A" Pressure 911 psig OTSG "B" pressure 909 psig Main Steam Pressure 894 psig Main Steam Temp. 589 F.
Condenser Vacuum 1.76 Generated MW 834 DFT level 12.7 ft.
Feed Flow "A" 5 X 106 lbs/hr Feed Flow "B" 5 X 106 lbs/hr Feed Pressure "A" 970 psig Feed Pressure "B" 968 psig
14:23:21 +24 Volt Bus Failure (NNI Cause still unknown. Apparently, power loss' "X"supply) the positive 24 VDC bus shorted dragging the bus voltage down to a
Rev. 5 Page 2 Time Event Cause/Comments low voltage trip condition. There is a built-in k to h second delay at which time all power supplies will trip. There was no trip.
indication on negative (-) voltage.
This event was missed by the annunciator. Following the NNI power failure, much of the control room indication was lost. Of the instrum- tntation that remained operable transient conditions made their indic- cation questionable to the operators.
14:23:21 PORV and Spray Open 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 actuation 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 spray valve stroked for approximately ; second. The 40% open indication on spray valve did not actuate, therefore, the spray valve did not exceed 40% open.
14:23:21 Reduction in Feedwater As a result of the t"IV power supply failure many primary plant control signals responded erroneously. Tcold failed to 5706F (normal indication was
557 6P) producing several spurious alarms.
Tave failed to 570"F (decreased). The resultant Tave error modified the reactor demand such that control rods were
%iithdrawna ito increse Tave and reactor power. The power increase was terminated at 103% by the ICS and a "Reactor Demand High Limit" alarm was received. Thot failed to '7OF(low) and RC flow failed to 40 X 10 lbs/hr in each loop (low).
Both these failures created a BTU alarm and limit on feedwater 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
Rev. 5 Page 3 Time Event Cause/Comments regulate header pressure thus increasing generated megawatts. These combined failures resulted in a loss of heat sink to the reactor initiating an excessively high RC pressure condition.
14:23:35 Reactor Trip/Turbine Rx trip caused by high RCS pressure at 2300 psi Trip Turbine was tripped by the reactor.
14:24:02 Hi Pressure Inj. This was a computer printout and indicates Req. (Flag) <500-subcooling.* See attached graph of RC
Pressure/Temp. vs. Time. This graph is based
.-
on Post Trip data and actual incore thermo- couple data. From the reactor trip point (14:23 to 14:33, core exit temperature data was obtained by extrapolation and calculated data.
This is supported by -two alzrm data points plotted at 18' and 21° of subcooling during this period from the computer. It is important to note that lowest level of subcooling was
8pF for a very short period of time.
- NOTE: This computer program was initiated as a result of the TMI incident.
14:24:02 Loss Of Both Suspect condensate pump tripped due to high Condensate Pumps DFT level. This is verified by ???? printed by computer, indicating the level instrument was over ranged as well as 4 low flow indication in the gland steam condenser as also indicated by computer.
14:25:50 PORV Isolated 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.
14 :26:41 EPI Auto Initiation EPI 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 EPI was initiated with 3 pumps resulting in approximately 1100 gpm flow to the RCS. At this time, all remaining non- essential R.B.
isolation valves
_ ... . A_. .
Rev. S
Page 4 Time Event Cause/Comments were closed per THI Lessons Learned Guidelines.
- . , ..
14:26:54 RC Pumps Shutdown Operator turned RC pumps off as required by the applicable emergency procedure and B & W
small break guidelines.
14:27:20 RB Pressure Increasing This is first indication that RCDT rupture disc had ruptured, RB pressure increase data was obtained from Post Trip Review and Strip Chart indication.
14:31:32 RB Pressure High 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).
14:31:49 OSTG "A" Rupture Matrix This occurred due to <600 psig in OTSG "A.
Actuation The low pressure was caused by OTSG "A" boiling dry which was resultant from the BTU limit and failed OTSG level transmitter. This resulted in the closure of all feedwater and steam block valves which service OTSG "A".
14:31:59 Main Feedwater Pump 1A Caused by suction valve shutting due to Tripped matrix actuation in previous step.
14:32.14:41 ES A/B Bypass Manually bypassed and HPI balanced between all
4 nozzles (Total flow approximately 1100 gpm
-small break operating guidelines).
14:32:35 Started Steam Driven Started by operator to ensure feedwater was Emergency Feedwater Pump available to feed OTSG's.
14:33 Core Exit Temp. Verified The core exit incore thermocouples indicated the highest core outlet temperature value was
560F. RCS pressure was 2353 psig atthis time, therefore, the subcooling margin at this time was 100'F. Minimum subcooling margin for the entire transient was 86 F. It is postulated that some localized boiling occurred in the core at this point as indicated by the self powered neutron detectors.
14:33-*14:44 Started Motor Driven Emer- Same discussion as "Started Steam Driven Emer- gency Feedwater Pump gency Feedwater Pump."
14:33:30 RC Pressure High (2395 psig) 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
Rev. 5 Page 5 Time Event Cause/Comments leakage prior to the transient and RCV-9 did not lift.
14:34:23 RB Dome Hi Rad Level 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 press- urizer and coolant.
14:35:33 Attempted NNI Repower With- This resulted in spikes observed 6n de-ener- out Success gized strip charts.
14; 36:50 Computer Overload Caused by overload of buffer. Resulting in no further computer data until buffer catches up with printout.
14:38:15 FWV-34 Closed This valve was closed to prevent overfeeding OTSG "B" beyond 10OIindicated Operating Range.
14:44:12 NNI Power Restored Success- NNI was restored by removing the X-NNI Power fully Supply Monitor Module. This allowed the breakers to be reclosed. At this time, it was observed that the "A" OTSG was dry, the press- urizer was solid (Indicated off scale high),
RC outlet temperature indicated 556F (Loop A
& B average), and RC average temperature indi- cated 532'F (Loop A & B). The highest core exi thermocouple temperature at this time was 531'F
RSC 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).
14:44:31 RB Isolation and Cooling Actuation At this time, RB pressure increased to 4 psig and initiated RB Isolation. The operator verified all immediate actions occurred properl for BPI, LPI, and RB Isolation and Cooling. Th'
increasing RB pressure was resultant from RCV-8 passing HPI at this time.
14:46:10 Bypassed EPI, LPI and RB These "ES" systems were bypassed at this time Isolation and Cooling to again balance HPI flow and restore cooling water to essential auxiliary equipment (i.e.,
RCP's, letdown coolers, CRDM's etc.).
1/4-~
Rev. .5 Page 6 Time Event Cause/Comments
14:51:57 Rupture Matrix Actuation on The actuation was resultant from a deg- OTSG-B radation of OTSG-B pressure. Cold emer- gency 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 NW pump. Both Emergency FW pumps were already in operation at this time. B-OTSG level at this time was
70% (Operation Range).
14:52 EPI Throttled and RCS At this time, the maximum core exit thermo- Pressure Reduced to 2300 couple temperature was 515'F, RCS pressure psig was 2390 psig. Therefore, the subcooling margin was 147PF. Natural circulation was in effect as verified previously. All con- ditions had been satisfied to throttle EPI.
Therefore, flow was throttled down to approx- imately 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.
14:53 Reestablished Letdown At this time, the operator was attempting to establish RCS pressure control via normal RC makeup and letdown.
14:56 Opened MU Pump Recirc. This was done to assure the MU pumps would Valves have minimum flow at all times to prevent possible pump damage.
14:56:43 Bypassed the A-OTSG Rupture Feedwater was slowly admitted Matrix and Reestablished - to the A-OTSG which was dry up to this point.
Feed to the A-OTSG 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 depres- surization of the RCS. RCS pressure control was very unstable at this time. It is postuleted that some localized'boiling-occured in eore at this point as indicated by .self.neutron detectors.
I
Rev. 5 Page 7 Time Event Cause/Comments
14.57:09 Bypassed the B-OTSG This was done to regain FW control of the Rupture Matrix 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).
14:57:15 Established RC Pump This was done in preparation for a RCP start Seal Return (when necessary) and to minimize pump seal degradation.
15:00:09 Reestablished Level This verified feedwater was being admitted to I; A-OTSG 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.
15:00:09 77F Subcooled "A" Loop 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.
15:15 23 F Delta-T/Marned the At this time, loop temperatures were nearing Technical Support Center equalization. This delta-T was calculated from loop A & B Tcis and core exit thermo- couples.
15:17 Declared Class "B" Emergency This was done based on the fact there was a loss of coolant through RCV-8 in the containment and BPI had been initiated. All non-essential R# 3. personnel wmre- directed to-evacuate and-contact off,.site agencies be- gan. Surtey teem was sent to Auxiliary Building
15:19 Opened Emergency FW Block At this point the A-OTSG level was increasing to B-OTSG 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 (ld0' F/hr) while maintaining RCS pressure control.
I
Rev. 5 Page 8 Time Event Cause/Comments
15:26 Lo Level Alarm in Sodium This was recultant from the tank supply valve Hydroxide Tank opening when the 4 psig RB isolation and cool- ing signal actuated. The sodium hydroxide was released to both LPI trains.- Sodium Hydrdxide was admitttd to the RCS-via. HPI from.the BEWST.
(Approximately 2 ppm injected into the RCS.)
15:50 Terminated EPI At this time, all conditions had been satis- fied (per small break operating guidelines)
to terminate EPI. RCS pressure control had been established using normal makeup and letdown. EPI was terminiated and essentially all releases to the RB were discontinued.
16:00 Commenced Pressurizer At this time, RCS pressure and temperature Heatup were well under control. Natural circulation was functioning as designed (approximately 23'F
delta-T). RCS temperature was being maintained at approximately 450. RCS pressure was approx imately 2300 psig. The decision was made at this point to commence pressurizer heatup in preparation to re-establish a steam space in the pressurizer.
16:07 Survey Team Report The Emergency Survey Team reported no radiation survey results taken offsite were above back- ground.
16:08 :04 Shutdown Steam Drive Emergency FW Pump The motor driven Emergency FW pump was running, therefore, the steam driven pump was not needed The plant remained in this condition for app- roximately 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 press- urizer to saturation temperature for 1800 psig.
16:15 Press Release Media was notified of plant status.
18:05 Established Steam Space Pressurizer 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 pressure heaters.
18:30 Terminated Class B Emergency State and Federal Agencies notified.
Rev 5 Page 9 Time Event Cause/Comments
21:07 Forced Flow Initiated The decision was made to re-establish forced in RCS flow cooling in the RCS at this time. B&W
and NRC were consulted. RCP-lB and lD were started. At this point, RCS parameters were stabilized and maintained at RC pressure-2000
psig, RCS temperature-420F. Pressurizer level-235 inches. The plant was considered in a normal configuration.
,...
Enclosure 2 PLANNED CORRECTIVE ACTION AT CR-3 Immediate
- Thorough testing of NNI system to determine cause of failure
- Modify PORV so that NNI failure closes valve
- Modify pressurizer spray valve so that valve doesn't open on NNI
failure
- Provide positive indication of all three relief or safety valves
- Establish procedural control of NNI Selector switches
- Train all operators in response to NNI failures
- Move 120v ICS "X"power to vital bus
- Initiate more extensive program for events recorder system
- Provide operator with redundant indication of main plant parameter rs At Next Refueling (September 1980)
- Install indication lights on all panels to know if power on panel
- Quick access to fuses is being designed into cabinets
- Modify EFW pump circuit to start pumps on any low steam generator level signal Long Term
- Investigate upgrade of NNI capabilities - total loss of NNI
- Remote shutdown is being designed
- Provide backup AC sources to inverters with automatic transfer.
Enclosure 3 Information requested by COB, March 12 and March 17, 1980.
1. Summarize power upset events on NNI/ICS that have previously occurred at your plant.
2. Specifically review the Crystal River event, and address your plant's susceptibility to it in general.
3. Set forth the information presented by your representative(s)
in the meeting on March 4, 1980.
4. Address information available to the operator following various NNI/ICS power upset events, including a discussion of:
- how the operator determines which information is reliable
- what information is needed to bring the plant to cold shutdown
5. Address the feasibility of performing a test to verify reliable information that remains following various NNI/ICS power upsets.
6. Address each CR-3 proposed corrective action in terms of applicability to your plant.
7. Expand your review under IE Bulletin 79-27 to include the implications of the CR-3 event. Inform us of your schedule for completion of this expanded review as discussed on March 4, 1980.
In addition to the above, Florida Power Corporation should address:
1. Sequence of events for the CR-3 trip.
2. Proposed corrective actions at CR-3.
3. Discuss the impact, whether it be beneficial or detrimental, of NRC
Short Term Lessons Learned and Bulletins and Orders requirements.
Florida Power Corporation cc w/enclosure(s):
Mr. S. A. Brandimore Mr. Robert B. Borsum Vice President and General Counsel Babcock & Wilcox P. 0. Box 14042 Nuclear Power Generation Division St. Petersburg, Florida 33733 Suite 420, 7735 Old Georgetown Road Bethesda, Maryland 20014 Mr. Wilbur Langely, Chairman Board of County Commissioners Citrus County Iverness, Florida 36250 Bureau of Intergovernmental U. S. Environmental Protection Agency Relations
660 Apalacbee Parkway Region IV Office Tallahassee, Florida ATTN: EIS COORDINATOR 32304
345 Courtland Street, N.E.
Atlanta, Georgia 30308 Director. Technical Assessment Division Office of Radiation Programs (AW-459)
U. S. Environmental Protection Agency Crystal Mall #2 Arlington, Virginia 20460
- Crystal River Public Library Crystal River, Florida 32629 Mr. J. Shreve The Public Counsel Room 4 Holland Bldg, Tallahassee, Florida 32304 Administrator Department of Environmental Regulation Power Plant Siting Section State of Florida
2600 Blair Stone Road Tallahassee, Florida 32301 Attorney General Department of Legal Affairs The Capitol Tallahassee, Florida 32304 I