ML19322E568
| ML19322E568 | |
| Person / Time | |
|---|---|
| Site: | 05000000, Crystal River |
| Issue date: | 03/06/1980 |
| From: | Harold Denton Office of Nuclear Reactor Regulation |
| To: | AFFILIATION NOT ASSIGNED |
| Shared Package | |
| ML19322E559 | List: |
| References | |
| GL-80-18, IEB-79-27, TAC-12961, NUDOCS 8003280395 | |
| Download: ML19322E568 (13) | |
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UNITED STATES NUCLEAR REGULATORY COMMISSION 2
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wAsHtNGTON, D. C. 20656
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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 ally 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 iglication 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 iglications 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 regt!:sted 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 deterraine whether or not your license should be modified, suspended, or revoked.
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This letter confirms the oral request for this infonnation expressed at the March 4,1980 meeting by Mr. Darrell G. Eisenhut, Acting Director, Division of Operating Reactors.
Sincerely, 9 4 ro
. Denton, Director
. Office of Nuclear Reactor Regulation
Enclosures:
As stated 6
9
intiosure 1 Revo 5 Page 1 SIQUENCE (AS OF 2300 3/1/30) 26 February Transient CR-3 EVENT SYNOPSIS At 14:23 on February 26, 1980 Crystal River -3 Nuclear Station experienced a reactor trip from apprnrimately 100% full power. A synopsis of kay events and parameters was obtained from the plant computer's post-trip review and plant alarm summary, the sequence of events monitor, control room scrip chares, and the Shift Supervisor's log.
The reactoritwas operating as approximately 100 % full power +rith ' Integrated Control System (ICS) in automatic. No casts were in progress arid minor sain-tenance was being performed in the Non-Nuclear Instrusettation (NNI) cabinet "Y".
Time Event Cause/ Comments 14:23:00 The following is a su= mary of plant conditions prior to the trip nux 98.6%
RC Pressure 2157 psig PZR level 202 inches E tank level 71 inches
(
"A" 599'F.
"3" 600*7.
C "A" 557'F.
T B" 556*F.
Rb now "A" 73 I 100 lbs/hr g
RC now "3" 73 I 106 lbs/hr Leedown n ow 48 gym OTSG "A" 1v1 (OP) 67 OTSG "B" 1v1 (OP) 65:
OTSG "A" FEL7 242 inches OTSG "3" FRLY 254 inches OTSG "A" Pressure 911 peig OTSG "B" pressure 909 psig
)
Main Steam Pressure 894 psig Main Stesa Temp. 589*F.
Condenser Vacuum 1.76 Generated W 834 DFT level 12.7 fc.
Feed.n'w "A" 5 I 106 o
lbs/hr Feed now "B" 5 I 10 lbs/hr Feed. Pressure "A" 970 psig Feed Pressure "3" 968 peig 14:23:21
+24 Volt Bus Failure (NNI Cause still unkacun. Apparently, power loss'c "I" supply) the positive 24 VDC bus shorted dragging the bus voltage down to a l
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Rev. 5 Page 2 Time Event cause/Cemments
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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 roca indication was lost. Of the I.nstrum-
'entation that remained operable transient conditions made their indic-cation qu'estlonable to the operators.
14:23:21 PORY 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 PORY 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 PORY seal in from clearing on icw pressure. It is postulated that the FORY 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' power supply failure many primary plant control signals responded erroneously. Teoid failed to 570*? (normal indication was 557'F) producing several spurious alarms.
Tave failed to 570'7 (decreased). The resultant Tave error modified the reactor demand such that control rods were Nh withdrawn
- f o incrase Tave and reactor t
9 power. The power increase was terminated h '5\\
at 103% by the ICS and a "Raactor Demand g
\\%
Righ Limit" alarm was received. Thot failed to g70*7 (low) and RC flow failed 9
to 40 % 10 lbs/hr in each loop (low).
Both these failures created a BTU alarm and limit on feedwater which reduced i
feedwater flow to both OTSG's to l
essentially zero. Turbine Header Pressure
(
failed to 900.psig (high) which caused the turbine valves to open slightly to l
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l 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.
1 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 princout and indicates Req. (Flag)
<50* subcooling.* See attached graph of RC Pressure / Temp. vs. Time. This graph is based en Post Trip data and actual incore thermo-couple data. From the r eactor trip point (14:22 to 14:33, core exit tenperature data was obtained by extrapolation and calculated data.
This is supported by.tvo alarm data points j
plotted at 18* and 21* of subcooling during chis period from the computer. It is important to note that lowest level of subcooling was 887 for a very short period of tina.
- 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 DFr level. This is verified by ???? printed by computer, indicatizig the level instrument was over ranged as well as a low flow indication in the gland stesa condenser as alsc indicated by computer.
14:25:50 PORY Isolated At this time a high RC Drain Tank level' alarm was received. This was resultant from the PORY remaining open and vas positive indication that the PORY was open. At this time, the operator closed the PORY bicek valve due to RCS pressure decreasing and high RCDT level.
14:26:41 HPI Auto Initiation HPI 1:21tiated automatically due to low RCS pressure of 1500 psig. The low pressure condition was resultant from the PORY r==4"4"g full open while the plant was tripped. Full HPI was initiated with 3 pumps resulting in approximately 1100 gym flow to the RCS. At this time, all remaining non-essential 1.3.
h isolation valves aDM^6W D
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Rev.5 Page 4 Time Event Cause/ Comments were closed per TMI Lessons Cearned Guidelines.
14:26:54 RC Pumps Shutdown Operator turned RC pumps off as required by the applicable emergency procedure and 3 & W small break guidelines.
14:27:20 R3 Prassure Increasing This is first indication that RCDT rupture disc had yuptured. R3 pressure increase data was obtained frcs Post Trip Review and Strip Chart indication.
14:31:32 R3 Fressure High This alarm was initiated by 2 psig'in R3.
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 CSTG "A" Rupture Matrix This occurred due to <600 psig in OTSG "A.
Actuation The low pressure was caused by CTSG "A" boiling dry which was resultant from the BTU limit and I
failed OTSG 1evel transmitter. This resulted in the closure of all feedwater and steam block valves which service OTSG "A".
14:31:59 Main Fesdt.ater Pump 1A Caused by suction valve shutting due to Tripped matrix actuation in previous step.
14:32+14:41 ES A/3 Bypass Manually bypassed and HPI balanced between all 4 nozzles (Total flow approximately 1100 gym
-small break operating guidelines).
14:32:35 Sentud Steam Driven Started by operator to ensure feedwate was t
Esargency Feedwater Pump available to feed OTSG's.
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14:33 Core "xit Temp. Verified The core exte incore thermocouples indicated l
the highest core outlet temperature value was
$60'7. RCS pressura was 2353 psig atthis time,
'therefore, the subcooling margin at this time was 100*7.
Miniwr= subcooling nargi2 for the l
T entire transient was 8 F.
It is l
g postulated that some localized boiling l
9 y~ p0 occurred in the core at this point as D
indictated by the self powered neutron s
detectors.
14:33+14:44 Started Motor Driven Emer-Sama discussion as " Started Steam Driven Emer.
gency Feedwater Pump gency Feedwater Pump."
14:33:30 RC Pressure High (2395 psig) At this point, pressuriser is solid and ceds safety lifts (RC7-8). This is the highast RCS pressure as recorded on Post Trip Review.
Apparently, RC7-8 lifted early due to seat
e Rev. 5 9 Aoy g
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Time Event Cause/ Comments 1eakage prior to the transient and RC7-9 did not lif t.
14:34:23 RB Dome Hi Rad Lev &1 R.9G-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 coolaat.
I 14:35:33 Attempted WI Repower With-This resulted in spikes c5servad in 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 W 7-34 Closed This valve was closed to prevent overfeeding OTSG "3" b,eyond 100::' indicated 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 reciesed. 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 556*7 (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 $31*F RSC pressure was 2400 psig (saturation temp. at this pressure is 662*F.).
This data verified natural circulation was in orogress and the plant subcooling margin was 131*F. (based on core azit 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 inmediate actions occurred proper 1:
for HPI, LPI, and RB Isolation and Cooling. The increasing RB pressure was resultant from RC7-8 passing HPI at this time.
14:46:10 Bypassed HPI, 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.).
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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-3 pressure. Cold emer-gancy feed was being injected into the OTSG at this time. This matrix actuation isolated all feedvater and steam block valves so the 3-0TSG and tripped the "3" main W pump. Both Emergency W pumps were already in operktion at this time. 3-0TSG 1evel at this time was 70% (Operation Range).
14:52 HPI Throttled and RCS At this time, the maximum core exit ther:.:o-
' Pressure Reduced to 2300 couple temperature was $15'7, RCS pressure psig was 2390 psig. Therefore, the subcooling margin was 147*7. Natural circulation was in effect as verified previously. All con-1 dicions had been satisfied to throtile HPI.
Therefore, flow was throttled down to approx-imately 250 gym to reduce RCS pressure to 2300 psig in order to attempt to reduce the flow rate through RC7-8 and into the R3.
14:53 Reestablished I.eedcwn At this time, the operator was attempting to establish RCS pressure control via normal RC makeup and letdown.
14:56 Opened MU Pump Rectre.
This was done to assure the MU pumps would Valves have =4n4-=_ flow at all times to prevent possible pump damage.
14:56:43 Bypassed the A-0TSG Rupture Feedwater was slowly admitted Matrix and Reestablished -
to the A-0TSG which was dry up to this point.
Feed to the A-0TSG Feedwater was admitted through the Auxiliary W header via the E W bypass valves. The feedrate was very ? low in order to minimize thermal shock to the OTSG -and resultant depres-3 surization of the RCS. RCS pressure control was very unstable at this time. It la Postulatec that some localised' boiling.occured in core ac' this point' as ? indicated by.self. neutron detectors.
e 1
Rev. 5 Page 7 Tire Event Cause/Consnents 14~:57:09
- Sypassed the 3-0TSG This was done to regain W control of the Rupture Matrix 3-0ISG. 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 necessaty).
I 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 feedwatsr was being admitted to In A-0TSG 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 77*.? Subcooled "A" Loop This value was based upon "A" RCS loop parameters at this time. The "A" loop was being cooled down at this cine by the A-0TSG fill and the operator was attempting to equalize loop temperatures.
15:15 23*F' Delta-T/ Manned the At this time, icop temperatures wera nearing Technical Support Center equalization. This delta-T was calculated from loop A & 3 T 's and core exit therno-e couples.
15:17 Declared Class "B" Emergency This was done based on the fact there was a loss of coolant through RC7-8 in the
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contain= ant and HPI had been initiated. All
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non-essential CR91 personnel were directed eo tvicuate 'and;c'on*. set off-site agencies'he-
'gan.
Survey' team was' sine co' Auxiliary Building l
15:19 Opened Emergency N Block Ac this point the A-0TSG 1evel was increasing to 3-0TSG and the decision was made to commence filling the 3-0TSG simultaneously. The intent was to go 95% on both CISG's without exceeding RCS cooldown limits (10@F/hr) while mainem1Mng RCS pressure control.
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Rev. 5 r-Page S Time Event Cause/ Comments 15:26 I.o Iavel Alarm in Sodium This was recultant from the tank supply valva Hydroxide Tank opening when the 4 psig RB isolation and cool-ing signal actuated. The sodium hydroxide was released to both LPI trains. Sodium Hydro'xide Was admitted to the RCS via. HPI from.the B'4ST.
CAPproxibately 2. ppm injected into the RCS.)
15:50 Terminated HPI At this time, all conditions had been satis-fied (per small break operating guidelines) to terminate HPI. RCS pressure control had been established using nor:nal makeup and letdown. HPI was car-Hnisted, and essentially all releases to the R3 were discontinued.
16:00
, Ccamanced Pressurizar At this time, RCS pressura and temperature Heatup were wall under control. Natural circulation was functioning as designed (approximacaly 23*y delta-T). RCS temperature was being maintaiced at approximately 450*.
RCS pressure was approx-imately 2300 psig. The decision was made at this point to comnence pressuri=ar haatup in preparation to re-establish a steam space in the pressurizar.
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 W Pump The motor driven Emergency W pump was running, therefore, the staam driven pump was not needed The plant remained in this condition for app-roximacaly 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 prass-urizer to saturation temperature for 1800 peig.
16:15 Press Release Media was notified of plant status.
18:05 Established Steam Space Pressurizer At this point, pressurizar temperature was approximataly 620*F. Pressurizer level was brought back on scale by increasing latdown.
From this point pressurizer level was reduced 6
to normal operating level and normal pressure was established via pressure heaters.
State and Feder'l Agencias notified.
18:30 Terminated Class 3 Emergency a
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Rev 5 Page 9 Ti:ne Event Cause/Cotanents 21:07 Forced Flow Initiated The decision was made to re-establish forced in RCS flow cooling in the RCS at this time. B&*J and NRC were consulted. RCP-13 and ID were started. At this point,,RCS parameters were stabilized and maintained at RC pressure-2000 psig, RCS temperature-420*?. Pressurizar level-235 inches. The plant was considered in a normal configuration.
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Encicsure 2 PLANNED CORRECTIVE ACTION AT CR-3 Immediate Thorough testing of Nf:I 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 parameters At Next Refueling '(Sectember 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 hNI Remote shutdown is being designed Provide backup AC sources to inverters with automatic transfer.
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Information requested by COB, March 12 and March 17. 1980.
1.
Summarize power upset e..nts 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 cf this expanded review'as discussed on Marcn 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.
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