ML19309H790
| ML19309H790 | |
| Person / Time | |
|---|---|
| Site: | Crystal River  |
| Issue date: | 03/05/1980 |
| From: | Igne E Advisory Committee on Reactor Safeguards |
| To: | Advisory Committee on Reactor Safeguards |
| References | |
| ACRS-SM-0181, ACRS-SM-181, NUDOCS 8005190722 | |
| Download: ML19309H790 (49) | |
Text
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- 3 ~.f T h fCLEAR REGULATORY COMMISSIOr 3 T
E ADEuRY COMMITTEE ON REACTOR sAFECUA..,*4 '~ WA&MINGTON, D. C. D655 f {g 2g
- ., * * " * /
March 5,1980 i ACRS Members ACRS Technical Staff ACRS Fellows i
SUBJECT:
CRYSTAL RIVEP 3 INCIDENT A meeting on the status of the Crystal River 3 incident was held at the Holiday Inn in Bethesda on March 4,1980. In attendance were NRC, Florida Power Corporation (FPC) Babcock & Wilcox and Babcock & Wilcox owners, and members of the public. 26,1980 was A synopsis of the events that occurred at 14:23 on February presented FPC. The handout of the sequence (as of 2300,3/1/80),Rev.5 is attached. It is interesting to note that at the onset of the incident, because of erroneous signals from the loss of 24 VDC supply, the spurious signals modified the reactor demand such that the control rods were withdrawn to increase T and reactor power. The power increase was teminated at 103% by th$y$CS. The erroneous signals also caused a loss of heat sink to the reactor initiating an excessively high RC pressure condition of 2300 psi which tripped the reactor on the turbine. ~ The core exit incore themocouples indicated th'e highest core outlet 0 0 temperature to be 560 F. The subcooling margin at this time was 100 F. The minimum subcooling margin for the entire transient was 80F. It is postulated that some localized boiling occurred in the core as indicated by the self-powered neutron detectors. 'It was stated at the meeting that one of the reasons that the Emergency Response Center at Bethesda was acti.vated was the report.that the contain-ment pressure was at 18 psig. FPC stated that this infomation was relayed to NRC as 18 psig or about 3 psi above ambient pressure. This was misinter-preted by the Staff to mean 18 psi above ambient pressure. Without the misunderstanding, the Center may not have been activated. FPC also di.:ussed the equipment deficiencies and equipment that remained unaffected by the incident. Some of D.e systems that perfomed adequately are as follows: Reactor Protec)ive System Engineered Safeguards Activation Control Rod Drive System A description of equipment deficiencies are attached. As more infomation on the above matter is recieved, you will be kept infomed. ( p E.'Isce "S "2'" Attac..ent: As stated l THIS DOCUMENT CONTAINS Y P00RQUAUTYPAGES c.. T
7:32 1 ,4 i gEQUINCE (AS OF 2300 3/1/80) l 26 February Transient CR-3 s. EVErr STNCPSIS At 14:23 on February 26, 1980 Crystal River -3 Nuclear Station experienced a reactor trip from approximately 100% full power. A synopsis of key events ard parameters was obtained from the plant computer's post-trip review and plant alarm su=ary, the sequence of events monitor, control room strip charts, and the Shift Superviser's log. The reactom was operating me approximately 100 %. full power with ' Integrated Centrol System (ICS) in automatic. No tests were in progress and minor main-tenance was being performed in the Non-Nuclear Instrumentation (MTI) cabinet "Y". Time Event Cause/ Comments I 14:23:00 The following is a summary of plant conditions prior to the trip Fluz 98.6% RC Pressure 2157 psig PZ1 level 202 inches MU tank level 71 inches "A" 599'T. "B" 600*F. C "A" 55 7
- F.
B" 556*F. 1 T Rb Flow "A" 73 2106 lbs/hr EC Flow "B" 73 I 106 lbs/hr Letdown Flow 48 gym CISG "A" Ivl {0P) 67: OTSG "B" 171 (CP) 65% OTSG "A" FRLV 242 inches OTSG "B" FELY 254 inches CISG "A" Pressure 911 psis OTSG "B" pressure 909 psig Main Steam Pressure 894 psig Main Steam Temp. 589'F. Cendenser Vacuum 1.76 Generated MR 834 DFT level 12.7 ft. 6 Feed F1bw "A" 5 I 10 lbs/hr Feed Flow "B" 5 I J.0 lbs/hr Feed Pressure "A" 970 psig Feed Pres ure "B" 968 psig 14:23:21 +24 Volt Bus failure CGI Cause still unknown. Apparently, power loss'."I" supply) the posit 1ve 24 VDC bus shorted dragging the bus voltage down to a .M"* _g, go m.
i / O O > => 2 .. Time Event Caus,/Conuments l low voltoge trip condition. There is a built-in k to second delay l at which time all power supplies Till trip. There was no trip indication on negative (-) voltage. I This event was missed by the annunciator. Following the RNI power failure, such of the contr01 room indication was lost. Of th instrum-nntation that remained ope able transient, conditions made thdr indic-cation questionable to the operators. 14:23:21 PORY and Spray Open When the positive 24 TDC supply was lost due to the sequence discussed above the signal monitors in RNI changed state c ' sing PORV/ Spray valves to open. The PLAY circuitry is desigced to seal in upon actuation and did so. The resultant loss of the negative 24 VDC halted spray valve motor operator and prevented 70R7 seal in from clearing on low pressure. It is postulated that the POR7 opened fully and the spray valve stroked for approximately second. The 4C% open 6 indication on spray valve did not actuate, therefore, the spray valve did act exceed 40% open. 1 14:23:21 Reduction in Feedwater As a result of the "I" power supply failure many pri: nary plant control signals responded erroneously, Teoid failed to 570*F (normal indication was 557*7) producing several spurious alar =s. Tave failed to 570*F (decreased). The resultant Tave error modified the reacto: demand such that control rods were ) withdrawn $ yto incrase Tave and reactor power. The power increase was terminated at 103% by the ICS and a "Rasetor Demand Eigh Limit" alarm was received. That failed to g70*7 (low) and RC flow failed to 40 I 10 lbs/hr in each loop (low). Both these failures created a BTU alarm and limit on feedwater which reduced e J e feedwater flow to both OTSG's to ~ essentially zero. Turbine Header Pressur failed to 900 psig (high) which caused the turbine valves to open slightly to
Bav. 5 A l v Pcs2 3 Cause/ Comments Event. Time regulate header pressura thus increasing generated messumets. These combined failures resulted in a loss of heat sink to the reactor initiating an excessively high RC pressure condition. Ex trip esoaed by high RCS pressure at 2300 ps 14:23:35 taaetor Trip / Turbine Turbine was tripped by the reactor. Trip 14:24:02 El Pressure Inj. This was a computer printout and indicates <50* subcooling.* See attached graph of RC tag. (.Tlag) Pressure / Temp. vs. Time. This graph is based
- on Post Trip data and actual incere thermo-couple data. Trom the reactor trip point (14:2 to 14:33, core exit temperature data was obtained by extrapolation and calculated data.
This is supported by.tvo alarm 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 8'? for a very short period of time. 850TE: This computer program was initiated as a result of the TMI incident. Suspect condensate pump trippeli due to high 14:24:02 I,oes of Both DPT. level. This is verified by 1777 printed condensate Pumps by computer, indicatidg the level instr *msent was over ranged as well as a lov flow indication in the gland steam condenser as als indicated by computer. At this time a high RC Drain Tank leve!.' alarm 14:25:50 PORY Isolated was received. This was resultant from th P0R7 remaining open and was positive ind:'.catic that the PORV was open. At this time, the operator closed the PORY block valve due to RCS pressure decreasing and high RCDT level. EPI initiated automatically due to lov RCS 14:26:41 EPI Auto Initiation pressure of 1500 psig. The low pressure condition was resultant from the PORY remaini: full open while the plant was tripped. Full IPI was initiated with 3 pumps resulting in approximately 1100 gym flow to the ICS. At i this time, all remaining non-essential 1.3. isolation valves 4 o' g 1 i i i i l
mov.
- 0 Paga 4 Time Event Cause/ Concents were closed per TMI Lessons rearned Guidelines.
r e. g 14:26:54 RC Pumps shutdevn Operator turned RC pumps off as required by the applicable emergency procedure and 3 & W small bresh ruidelines. 14:27:20 13 Pressure Increasing This is first indication that RCDT ruptura disc had Yuptured. 13 pressure increase data was obtained from Post Trip Review and Strip Chart indication. 14:31:32 RB Fressure High This alarm was initiated by 2 psig in 13. This is attributed to steam release from RCDT. Cod 6 safeties had not opened at this time based upo- . tail pipe temperatures recorded at 14:32:03 (Computer). 14:31:49 CSTC "A" Rupture Matrix This occurred due to <600 psig in OTSG "A. The low pressure was caused by CTSG "A" boilin: Actuation dry which was resultant from the BTU limit and failed OTSG 1evel transmitter. This resulted in the closure of all feedvater and steam bloc'. r ,L valves which servica OTSG "A". 14:31:59 Main feedvater Pump 1A Caused by' suction valve shutting due to Tripped metrix actuation in previous step. 14:32+14:41 IS A/B Bypass Manually bypassed and EPI balanced between all 4 nozzles (Total flow approximately 1100 gp:::
M break operating guidelines).
14:32:35 started Steam Driven Started by operator to ensure feedvater 'was Emergency Feedvater 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 560*7. RCS pressure was 2353 psis atthis ti=e 'therefore, the subcooling margin at this ti=e was 100*T. winimum subcooling margin for the entire transient was 8 F. It is postulated that some localized boiling occurred in the core at this point as indicated by the self powered neutron detectors. 14:3 F14:44 Started Motor Driven Emer-Same discussion as " Started Steam Driven Emer-( sency Teodwater Pump gency Feedvater Pump." 1 14:33:30 RC Pressure High (2395 psig) At this point, pressurizer is solid and code safety lif ts (RCV-8). This is the highest RCS pressure as recorded en Post
- rip Review.
Apparently, RCV-8 lif ted ear'.y due to seat
A Q Rev. 5 tsg2 5 e Cause/ Comments Time Event leakag's prior to the transient and RCV-9 did not lift. 14:34:23 RB Dome Ri Rad ieval RHG-19 alarmed at this point. Righest level indicated during course of incident was 50 1/hr. Righ radiation levels in RB caused by release of non-condensabia gases in thm press-urizer and coolant. 14:35:33 Attempted NNI Repower With-This resulted in spikes o5 served en de-ener-out success sized strip charts. 14:36:50 Computer overload caused by overload of buffer. Resulting in no further computer data until buffer catches up with printout. This valve was closed to prevent overfeeding 14:38:15 PWV-34 Closed OTSG "B" beyond 100I indicated Operating Range 14:44:12 RNI 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 556*F (Loop A & 3 average), and RC average temperature indi-cated 532*7 (Loop A & 3). The highest core er thermocouple temperature at this time was 531* RSC pressure was 2400 psig (saturation temp. a-this pressure is 662*T.). This data verified natural circulation was in preeress and the l plast subcooling margin was 131*7. (based on core exit thermocouples). 14:44:31 13 Isolation and cooling I At this time, 13 pressure increased to 4 psig Actuation and initiated 13 Isolation. The operator verified all insnediate actions occurred proper for BPI, LPI, and R3 Isolation and Cooling. T increasing R3 pressure was resultant from RC'.'- passine, HPI at this time. j 14:46:10 Bypassed HPI, LPI and 13" These "ZS" sysf.ess 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.). m. e se ese e em_ S* e6 e og m ---e .g, e. e ee $ 6D '
h h hcg 3 6 a Cause/Censsents Time Ev4nt The actuation was resultant from a des-14:51:57 Rupture Matrix Actuation on radation of OTSG-B pressure. Cold emer-OTSG-B gency feed was being injected into the OTSG This matrix actuation isolated' at this time. all feedwater.and steam block valves so the 5-0TSG and tripped the "B" sain W pump. Both Emergency W pumps were already in operation at this time. 5-CTSG 1evel at this time was 70% (operation Range). At this time, the==w mum core exit thermo-16:52 EPI Throttled and RCS i Pressure Reduced to 2300 couple temperature was 515'F, RCS pressure was 2390 psig. Therefore, the subcooling psig margin was 147'F. Natural circulation was in effect as verified previously. All con-ditions had been satisfied to throttle EPI. Therefore, flow was throttled down to approx-l instely 250 syn 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 14:53 Reestablished Letdown to establish RCS pressure control via nor=al RC makeup and letdown, This was done to assure the MU pungs would 14:56 Opened EU Fump Recire. u4n4=um flow at all times to, prevent have Yalves possible pump damage. Feedwater was slowly admitted 14:56:43 Bypassed the A-OTSG Rupture to the A-CTSG which was dry up te this poin:. f Matrix and Reestablished - Feedvater was admitted through the Auxiliary Feed to the A-CTSG N header via the E W bypass valves. The feedrate was very slow in order to M a4=fze thermal shock to the OTSG and resultant depres-surization of the RCS. RCS pressure control. was very unstable at this time. It 18 postulcte that some localized' boiling.occured in core stithis point *as' indicated by Jelf neutron detectors. i d e L
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- m. 5 Pcan 7 d'ause / Comments Event _
g This was done to regain N centrol of the l 14:57:09
- Bypassed the 5-0TSG 3H7TSG. Level was still high in this OTSG (apprr rf utely 65: Operating Range). Therefore, j
Bupture Matrix feed was not necessary at this time. The Main Steam Isolation valves were open. in preparation fcr iypass valve operation (when necessary). i This was done in preparation for a RCP start 16:57:15 Established RC Pump (when necessary) and to adnd=42a pump seal Seal Esturn degradation. This verified feedvater was being admitted to 15:00:09 Reestablished Level the OTSG and made '.c available for core In A-CTSG cooling via naturr.1 circulation. Feed to this generator was ceutinued with the intent of proceeding to 55% on the Operating Range. 15:00:09 77*J Subcooled "A" I4cp This value was based upon "A" RCS loop parameters at this ti=. The "A" loop was being cooled down at this time by the A-0TSG fill and the operator was attempting to equalize loop temperatures. 'At this time, loop temperatures were nearing 15:15 23*7 Delta-T/ Manned the equalization. This delta-T was calculated Technical Support Center from loop A & 3 T 's and core exit thermo-e couples. This was dene based on the fact there was a 15:17 Declared Class "3" Emergency loss of coolant through RC7-8 in the ~ ~ containment and HPI had been initiated. All non-essen,gial CR$ 1 personnel were directed to wicuate sad;c'entact off-site agencies'be- 'gan. Survey" teem was'sent to' Auxiliary Building k this point the A-CTSG 1evel was increasing 15:19 opened Emergency N Eleck and the decision was made to commence filling to B-CTSG the 3-0TSG simultaneously. The intent was to go 95% on both OTSG's without exceeding RCS cooldown limits (10d>F/hr) while maintaining RCS pressure control. i e e 4 I e e
l. O i5 7:33 8 Time Event Cause/ Comments s 15:26 to Level Alarm in Sodium This was recultant from the tank supply valve Bydroxide Tank opening when the 4 psig R3 isolation and cool-ins signal actuated. The sodium hydroxide was released to both LPI traf"<
- Sodium Hydro' zide iras admitted to the RCS via. HPI from.the BWST.
(ApProxi'astely 2.pps injected into the RCS.) TerminatedhPI At this time, all conditions had been satis-15:50 find (per small break operatir.g -uidelines) to terminate HPI. RCS pressure control had been established using normal makaup and letdown. HPI was terministed and essentially all releases to the RB vere discontinued. 16:00 Commenced Pressurizar At this time, RCS pressure and te=perature were well under control. Natural circulation Eastup was functioning as designed (approvisstaly 23*y delta-T). RCS temperatura was being maintained 6t approximately 450*. RCS pressure was approx imately 2300 psig. The decision was made at this point to connance pressurt:er heatup 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 offsita vera above back-ground. 16:0^ :04 Shutdown Steam Drive Emergency W Pump The motor driven Emergency W pump was running, therefore, the steam driven pump was not needed. The plant remained in this condition for app-roximately 2 hours, while heating up the press-urizar to satur-cion temperatura for 1800 psig. 16:15 Press Release Media was notifie.1 of plant status. 18:05 Established Steam Space Pressurizar At this point, pressurizer temperature was approximately 620*P. Pressurizer level was brought back on scale by incrassing letdown. From this point pressurizar level was reduced i to normal operating level and normal pressuru
- was established via pressure heaters.
15:30 Terminated Class 3 Emergency State and Teder h Agencias notified. e
O O R., s F2gn 9 Cause/ comments Event .w The decision ves made to re-establish forced 3 Forced Flow Initiated flow cooling in the RCS at this time. ~ B&W in RCS and NRC vare consulted. RCP-13 and ID vere started. At this point,,RCS parameters were .e stabilized and maintained at RC pressure-2000 psig, RCS temperature-420*F. Pressurizer level-235 inches. The plant was considered in a normal configuration. e e e J 9 6 e 1.. eme me e. .ee e , se e .we emo s eee se suese
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'I' DESCRIPTION 07 EQUIP'ENT DEFICIDICIES l I I The following equipment deficiencies were noted during an in depth review of the Crystal River Unit 3 transient on 7ebruary 26, 1980. ( Loss of the Non Nuclear Instrumentation (NNI) System (I) DC 1. power supplies. 1 2. Power operated relief valve (PORV) failed open. 3. Pressurizer spray valve failed open. 4. Loss of indication of major plant parametern. Integrated control system '(ICS) response to 3NI (I) f ailure. 5. A. Auto initiation of Emergency Peedvatar. 3. Reactor trip on loss of feedvater. 6. 1.s loss of 120 v AC for 2 secoods. 7. Events recorder degradation. 8. Computer alarm buffer overflow. J 9. Systems that performed adequately. A. Rasctor Protective System. 3. Engineered Safeguards Activation. Control Rod Drive 5y' stem. C. D. Steam Rupture Matriz. 1. Secondary Plant Equipment. F. Radiation Monitoring System. e O l ha s e ge4 1 ee
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