ML20085L468
| ML20085L468 | |
| Person / Time | |
|---|---|
| Site: | Dresden  |
| Issue date: | 02/22/1972 |
| From: | COMMONWEALTH EDISON CO. |
| To: | |
| Shared Package | |
| ML20085L460 | List: |
| References | |
| NUDOCS 8310310090 | |
| Download: ML20085L468 (24) | |
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t .L l g:6.4 vitrto.d 2 ~ 2 2' 7 A 4 DRESDEN NUCLEAR POWER STATION UNIT 3 SPECIAL REPORT INCIDENT OF DECEMBER 8, 1971
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1 1 REPORT ON DRESDEN 3 INCIDENT OF DECEMBER 8, 1971 .. Table of Contents Susanary A. CHRONOLOGICAL DISCUSSION . . . . . .. .. .......... 1 B. DAMAGE ASSESSMENT ...................... 6 C. OPERATIONS ASSESSMENT ..................... 6 D. EVALUATION OF EVENTS . . . . . . . ... ........... 8
- 1. Transient Effects on Fuel . . .. . ........... 8
- 2. Transient Effects on Reactor Internals . . . . . . . . . . 8
- a. Pressure induced lo' ads
; b. Thermal stresses , 3.- Effects of Pressure, Temperature, and Steam Impingement on Primary Containment Integrity . . . . . . . 9 r
- 4. Effect on the Reactor Vessel of the Differential Temperatures Experienced during Transient and
- Post-incident Recovery ... . . . . . . . . . . . . . . . . 10
. 5. Performance of the Pressure Suppression System . . . . . . 10
- 6. Safety Valve Operation . . . . . . . . . . . . . . . . . . 11' .
- 7. DC Grounds in Drywell . . . . ... ........... 11 E. CORRECTIVE ACTIONS . . .... . . ....... ....... 13 i-F. CCHCLUSIONS ......................... .
15. i , G. CHRONOLOGY ' ...........-.............. . 16 a lb FIGURES s W
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M . Figure 1 - Drywell Pressure Transient .-. . . .
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;@ Figure 2 - Torus Temperature ... .. . ' . ' . . . . .:._.....-. -
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09- 9e ' . REPORT ON DRESDEN 3 INCIDENT OF DECDiBER 8,1971 Summary At 1413 on December 8, 1971, as a result of a condensate booster pump trip on Dresden Unit 3, the reactor feed pumps tripped on low suction pressure, initiating a reactor water level transient. The transient resulted I in an abnormally high reactor water level, causing water to enter the main steam lines, and opening a safety valve for approximately one and one-half minutes which resulted in pressurization of the drywell. The transient was terminated approximately 13 minutes af ter the loss of the condensate booster pump. During the transient, the maxt=nm drywell pressure reached was 20 psig. The maximum and minimum reactor pressures were 1050 psig and 795 _ psig respectively; and the reactor water level reached a minimum of minus 20 inches and a maximum of plus 130 inches. All safety systems functioned as designed and no significant radioactivity was released to the environment as a result of the incident. During post incident recovery, both the primary system and the primary I containment were maintained in an isolated condition until analysis of reactor water and containment atmosphere could be made. Damage was limited to Local Power Range Monitor (LPRM) cabling, one electromatic relief valve solenoid operator and piping insulation. I The following report contains a sequence of events, damage report,
! corrective actions and conclusions.
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f D/R0 KNUTH AD/RDiP KUlILMAN AD/C0 GRIER " COBB ~ FAULKENBERRY ' GILBERT MPPB WERNER FIB HILDRETH GREHER CATAUlA MCDEPJiOTT REEDER MAY NOLAN SOLEM PERANICH ~ SNIEZEK DD/F0 DAVIS FS/EB THORNBURG REP ROY BRYAN BIDINGER APDB REINMUTH DREHER BROCKETT POTAPOVS ELLIS METZGER TRIPP COWER UHITESELLf PAULUS 3 7,s-- y, CUNNINGHAM FRIESS
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1 _l V ., , A. CHRONOLOGICAL DISCUSSION .~ A complete description of the events is presented below in chron-ological order with discussion where pertinent. Item
- 1. On December 8, 1971, Dresden Unit 3 was operating in a base loaded I condition at 792 MWe and 2300 MWt. The base load condition was being maintained for a test which was in progress for determination )
of chimney monitor sampling system plateout. Other significant parametric values prior to the incident included: Steam flow 9 x 106 lb/hr Feedwater flow 9 x 10 lb/hr Reactor pressure 990 psig , teactor water level , plus 29 inches Equipment in service included "3B", "3C", and "3D" condensate-booster pumps ("3A" pump was out of service to check alignment , and for repairs to a gland seal), "3A".and "3C" reactor feed - pumps ("3B" pump in standby) . "3B" feedwater regulating valve was in service in the automatic "3 element" control mode ("3A" feeduster regulating valve in standby) and the low flow feed-water regulating valve was open approximately 20%. The reactor containment was inerted and at a normal pressure of - approximately 0.25 psig. Chimney off-gas release rate was approximately 3,090 uCi/sec.
- 2. At 14:13:08, "3C" condensate booster ptusp tripped. Commonwealth Edison Company and General Electric Company personnel in the area heard noises coming from "3C" pump just prior to the trip, which *
- indicated possible mechanical trouble within the pump. However, complete teardown and inspection of both the "3C" condensate pump and booster ptamp revealed no damage. A subsequent search of asso- '
cisted recorder charts revealed no ~ anomalous behavior. The pump
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is tripped automatically on undervoltage or overcurrent. A check of the pump motor breaker showed no trip target.
- 3. Loss of the "3C" condensate-booster pump resulted in a low feed-p
- water pump suction pressure condition, since two condensate booster
_ pumps are inadequate to supply the required water at the existing h(,N' ," power level. The two operating reactor feed pumps "3A" and "3C"
. _ .~ tripped on low suction pressure (120 psig). . '
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p , y 6 4. Standby reactor feed pump "3B" started automatically when the oper-ating feed pumps tripped. It was demonstrated during the startup j test program, that the standby pump will reach full flow in approx-imately six seconds. It is surmised that the feedwater control system went into a " runout" or " flow control" mode at this time.
- 5. Following the decrease in feedwater input, the reactor water low level trip scrammed the reactor at 14 seconds.
- 6. Reactor water level continued to decrease and reached a low point of approximately minus 20 inches (approximately 123 inches above the top of the fuel), as observed by the operator. Level then began to increase and the operator took manual action between 1 minus 20 inches and minus 12 inches in anticipation of a. rapid level increase. He reduced the " master feedwater controller" i set point, to close the feedwater regulating valve, reduced the l manual . output control potentiometers on the " manual / auto" controllers i to zero, and transferred from " auto" to " manual" on the " manual / )
auto" controllers. He then reduced the manual output control potentiometer on the. low flow feedwater regulating valve to zero. When reactor water level reached minus 12 inches, it , appeared to hesitate at that point, and the operator. l increased the manual output control potentiometer _on the flow feedwater regulating valve in an attempt to increase level. Reactor water level began to increase rapidly, and as soon as the operator verified that level was increasing, he again reduced the manual output l control potentiometer to zero. As reactor water level came through zero, the operator started closing the feed-water regulating valve motor operated isolation valve, - again in anticipation of a rapid level increase. As the valve closed, feedwater flow was reduced from 5.7 x 10 6
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~~ of this valve, it stalled due to high differential pressure across the gate.
During the initial feedwater system transient the operator observed i
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the following conditions:
$ a. The " Flow On" light on the benchboard, indicating a runout con- '
( , dition, blinked on and off. n.
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Q b. The Feedwater Pump Max Capacity annunciator came up. This is
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- c. The " Loss of Air to F.W. Reg. Valve A" and " Loss of Air to h @- "
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'whea 'the -standby reac' tor feed pump . st"arted, T ' -
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[ went into a " runout" condition. (Flow control mode). This is ^J
- substantiated by the fact that the pump flow corresponded to a O runout flow control point of approximately 5.7 x 10t' lb/hr. and the observation of the annunciator which indicates a " runout" i condition. The operator stated that the " Flow On" light on the benchboard, which would also indicate a runout condition, only blinked on and then off. (The " Flow On" light should have been on steady for the runout condition. Subsequent testing disclosed no malfunction of the " Flow On" light.) At some time during the initial feedwater transient "3B" feedwater regulating valve locked out in an open position.
Subsequent testing has shown that the lockout condition was probably caused by low air pressure which resulted from rapid moven.ent of the valve.
- 7-8. Reactor pressure decreased following the pump trip due to
- shrinkage from the cold feedwater, and initially, loss of steam to the turbine. At one minute, five seconds, main steam line low pressure of 850 psig' initiated _a Group I isolation signal and the main steam isolation valvesclosed.
- 9. Reactor pressure continued to decrease due to shrinkage from continued input of reactor feedwater, and reached a low point of 795 psig. At that point, feedwater input and decay heat caused reactor pressure to increase.
- 10. Raactor water level continued to increase from feedwater input and the low reactor water level trip automatically reset at one minute, 21 seconds.
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- 11. Reactorwaterlevelcontiguedtoincreasefromafeedwaterinput of approximately 5.7 x 10 lb/hr. and reached the high water level trip point of plus 48 inches, at which point the turbine stop valves tripped. The control valves had already closed while attempting to maintain pressure.
- 12. Reactor pressure continued to increase from feedwater input and the operator manually cut in t:se isolation condenser. At this time,
, reactor water level was above ;ha isolation condenser supply nozzle,
{ . Since the condenser was essenti. ally oparating as a water to water [ heat exchanger, it had little effect on reactor pressure. 1 . 4 13. Reactor water level continued to increase from the feedwatar input 1 and reached the' level of.the main steam lines at approximately two {_ o minutes, 45 seconds, and began to fill them. u., .14. Safety valve "3F" lifted. Reactor pressure had reached 1020 psig - g \#4D$L ,
- and decreased rapidly when the valve opened. At this point water had probably filled most of'the steam lines. It is believed that
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p - i - E 15. Primary system water released to the drywell through the safety valve flashed to steam, an'2 pressurized the drywell. Drywell pressure reached two psig at five minutes, 11 seconds, and the reactor containment high pressure trip was actuated. He reactor containment high pressure trip started both diesel generators, the core spray pumps, and LPCI pumps automatically. he HPCI received an initiation signal, but tripped on high reactor water level. He reactor recirculation pumps tripped and a containment isolation (Group II) was initiated. All ECOS subsystems functioned as designed.
- 16. Drywell pressure increased to five psig at five minutes, 31 seconds.
- 17. Drywell pressure continued to increase, due to input from "3F" safety valve, and peaked at 20 psig approximately six minutes, 30 seconds into the incident. It is estimated that the safety valve remained open for approximately one and one-half minutes plus or minus 30 seconds, based on drywell pressure information and reactor vessel pressure data.
- 18. At seven minutes, suppression cooling was initiated via recirculation of the torus water through the containment cooling heat exchangers.
He torus sprays were not placed in service.
- 19. At eight minutes, the first indications of local power range monitor (LPRM) failure were observed. At this point the shutdown condition of the reactor had already been observed.
- 20. At 13 minutes, five seconds the operator tripped the operating reactor feed pump "3B". At this time reactor water level was approx-imately 130 inches as indicated on the wide range reactor water level indicator. Reactor pressure peaked at 1025 psig and dropped to 980 .
psig when the reactor feed pump was tripped. Reactor pressure then began to increase again due to decay heat input.
- 21. At approximately 22 minutes, drywell pressure had decreased to 13.5 psig due to pressure suppression operation and was continuing te decrease.
22-23-24. At 22 minutes, 45 seconds, the "0" reactor high pressure sensor tripped and reactor pressure peaked at 1044 psig. Reactor pressure then decreased due to loss of water inventory when an unsuccessful attempt was made to place the cleanup domineralizer system in service. n - Pressure dropped to 950 psig and again began to increase from decay
!. . heat input. A check of the "C" reactor high pressure trip calibration
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I 25-26-29. Containment pressure continued to decrease due to pressure suppression system operation. At approximately three hours, 47 minutes containment [ pressure had decreased to 4.5 psig. . i 27-28-30. At approximately four hours, a jumper was installed to permit open-ing the reactor water sample isolation valves so that a reactor water sample could be collected. This was necessary, since reactor blowdown to reduce reactor water level could not be established without first collecting a water sample. But, high water level isolates the sample line. The sample was collected and the jumper removed, thus reestablishing isolation. Analysis of the sample ~ indicated normal activity and reactor water blowdown via the cleanup system was established at six hours. Vessel water level had increased to approx-imately plus 145 inches at this time due to continued input of water from the control rod drive system.
. 31-32-33. At approximately 12 hours, drywell pressure had decreased to 1.75 psig and the containment high pressure trips automatically reset.
ECCS systems no longer had an initiation signal and these systems were returned to standby condition and the drywell coolers were restarted manually. At approximately 17 hours, drywell pressure had decreased to 0.63 psig. 34-36. At approximately 13 hours, reduction of reactor water level to below the main steam lines was initiated and at approximately 17 hours, reactor water level had been decreased to plus 76 inches on the wide range indicator.
. 35. Analysis of the first sample of containment atmosphere indicated
,' radioactivity slightly above normal.
- 37. Controlled cooldown of the primary system continued using the cleanup system. At approximately 33 hours reactor pressure had been reduced to 180 psig and the shutdown cooling system was placed in service.
38-39-40-41. Additional drywell atmosphere sample analysis indicated reduced radioactivity levels and at approximately 40 hours, drywell deinerting) 4 according to established procedures, was instituted. At 44 hours an initial ^ entry was made for atmospheric sampling. Radioactivity levels were low and the first entry for drywell inspection was made
, approximately 46 hours after the incident.
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B. DAMAGE ASSESSMENT . Inspection in the drywell following the incident revealed damage to the following equipment.
- l. The rupture discs on 4 of the 8 safety valves were ruptured. This nay not be related to the incident s'ince this condition has been encountered previously
, on normal shutdown. Additionally "3F" safety valve rupture disc was completely blown out.
- 2. The "3A" electromatic valve solenoid operator was damaged by the steam jet from the "F" safety valve. One steam discharge
" rams horn" on the "F" safety valve was directed towards the electromatic valve. The cover on the solenoid assembly of this valve was blown off. The holding coil circuit of the ..
- solenoid assembly was found open. Wiring to a position indicating limit switch was also damaged, rendering the position indication inoperative.
- 3. Miscellaneous thermal piping insulation was damaged.
- 4. The top coat of paint on the containment wall over an area about 3 x 3 foot was removed by the steam jet from the "F" safety valve impinging on the interior surface of the containment.
. 5. Sections of ventilating duct in the vicinity of the staan jet were dislodged.
- 6. gasentially all of the LPRM cables were found damaged.
- 7. One containment cooling fan motor was found to have a ground -
caused by moisture in the containsment. The other'six cooling fan motors were found to be in good condition, a C. _0_PERATIONS ASSESSMENT During the incident and post incident recovery period, the primary
- coolant system and primary containment system were maintained in an
. " isolated" condition until samples of water and at.mosphere could be collected and analyzed. It was necessary to install a " jumper" in
. order to permit opening of the reactor recirculation system sample
!" isolation valves to obtain a sample of reactor water. Following sample
' collection the " jumper"jwas removed, thus reestablishing isolation,
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Analysis of tho' reactor water indicated normal radioactivity.- g; m g- > - The cleanup system was placed in service and blowdown to the
- t. main condenser was established.
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- reset, and a sample of the drywell atmosphere was collected and analyzed.
Based on results of the analysis, the drywell was purged according to established operating procedures in preparation for entry and damage assessment.
- 1he ECCS systems which were initiated during the incident were not secured until the high drywell pressure initiation signal had cleared.
Although cable damage caused most of the LPRM flux monitors to fail, there was never any question about the reactivity status of the reactor. It was more than eight minutes af ter shutdown before any of 7 the flux monitors began to fail, and the shutdown status of the reactor was verified in=nediately af ter the scram and before the damage occurred. t Operator response was in accordance with operating procedures throughout the incident and post incident recovery with two exceptions. The operator did not reset the feedwater regulating valve lockout 4 condition when it occurred, and he did not trip the feed pump when the water level reached plus 60 inches. Had he done so, the incident i 3- It is important to place these actions in s may have been preventen. a proper prospective and it should be emphasized that he did take a number of steps to control feedwater input to the vessel. The operator's I actions were: - E
- 1) He reduced the anster controller set point to minimize the I error signal between the actual level and set point level. ;
This response was previously established on shift by General Electric during the startup program to compensate for the . ,; known overshoot which has been experienced following scrams. IAsile not specifically called for by the station operating 1 procedure 600-AN1; it is consistent with the intent of the y procedure to keep the level on scale. _ h; 2) He rejduced the manual output control potentiometer i on Pche main flow feedwater valve in an attempt to j 1.c miriimize feedwater input to the reactor vessel. [ 3) He_ reduced the manual output control potentiometer on the J " manual / auto" controller to zero and transferred them to manual, k By doing this, he thought he had closed the feedwater regulator, kh thus terminating feedwater input. m p, - During the period of the transient, operation of the shift was
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V%W - .under the direction of a licensed shift supervisor,'who was present, w - - - jdh%.g in the control room. -
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O' -O r. u r' pt . a , , r . j/i -s-tr . Gj D. EVALUATION OF EVENTS ) e
- 1. Transient Effects on Fuel !
t An evaluation based on total core flow, core pressure, core pressure drop and the calculated heat flux decay following a scram has been studied as related to the December 8 event.
- he results indicate that mechanical design limits were not exceeded and the critical heat flux condition did not occur. No heat flux increase was experienced. The MLHGR of 14.4 kw/ft.
and MCHFR of 2.98 which existed prior to the transient were not exceeded, and no Technical Specification limit violation on heat flux occurred.
- 2. Transient Effects ou Reactor Internals A.
- a. Fressure induced loads An svaluation has been made of the loads imposed-on the reactor vessel internal structures by'the incident. It was
- concluded that in no case did the maximum loads approach
- r. design values. No internals damsge could thus have occurred.
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; The design values of the reactor vessel internals loadings are arrived at by postulating the simultaneous failure of a i steam line and the occurrence of the design basic earthquake.
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; _The steam line break generates the maximum gross pressure ,
. differentials across the internals. The magnitude of these I differentials is determined by a combination of the initial value of the differentials and the increase associated with the rapid vessel depressurization. For this plant, the depressurization rate produced by a steam line break would be 30 psi /sec. An examination of the pressure transient during
.' the incident indicates that the maximum depressurization rate was less than 4 psi /sec. Since there was also no earthquake, 4
{ it can be concluded that the pressure induced loads on the i vessel internals were considerably less than design values. , k
- h. thermal stresses l r$.
[ 1he secondary stresses produced by thermal gradients in the l { core structure during the incident can be shown to have been i within the 3S, limit defined in Section III, para. NB-3222.2 of lf the A.S.M.E.' Boiler and Pressure Vessel Code. .W .. - - QWt..;: . a - 1. . ~ e, pi BIN, -.
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Calculations were based upon the fact that during the incident the feedwater flow went down to approximately 22% of rated flow after a relatively short time. he feedwater temperature at this flow rate was assumed to be 100 0 F. (lowest measured temperature during the incident). His loading is less severe than the thermal loading during High Pressure Coolant Injee. tion (HPCI). During HPCI condition, water of 400 F. is pumped through the feedwater spargers at a flow rate of 407.. he Reactor Internals
. loading diagram (G.E. Drawing 731E621) shows that the core structure tesperature during this condition is 3000 F. Based upon the loading diagram it was assumed that the shroud well inside tem-perature was 5460 F. and the outside wall temperature was 3000 F.
He calculated stress intensity of approximately 50,000: psi, which is produced by At=546-300=246 F., is secondary. his stress intensity is less than the 3s, limit of 55,200 psi at an average wall temperature of 4230 F. H e thermal transient condition imposed on the jet pumps and their supporting hardware is considered to have had an insignificant effect upon their design life. h is thermal condition most severely affects the jet pump adapter at its attachment to the shroud support shell and the jet pump riser
> brace. For a one time occurrence of lowering the jet pumps to 240* F. while assuming that the vessel and shroud support shelf remsja at 5300F. the. fatigue usage factors are estimated to be less than .04 for the jet pump adapter, and less than .01 for tha riser brace.
T
- 3. Effects of Pressure. Temperature, and Steam ImpinRement on Primary Containment Integrity He primary containment vessel was subjected to a pressure- ,
4 temperature transient which was rucorded by instrussentation as a maximum local temperature of 295 F. at 20 psi.
- This is less severe than the 320' F. at 20 psi calculated for the June 5,1970, incident at D-2. herefore, the drywell integrity-
< analysis as defined in the D-2 report and its supplement more than
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adequately apply to the D-3 incident.
! It was also visually observed that the area of the containment L. vessel impinged upon by the steam and water jet from the safety 8 valve was not affected -(damaged.) to the same extent and magnitude
}.e at D-3 as comparable areas at D-2.
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y .. s I. 4. Effect on the Reactor Vessel of the Differential Temperatures 3 Experienced During Transient and Post-Incident Recovery Available pressure and temperature curves of various parts of the reactor pressure vessel during the incident have been reviewed. Rose data were compared to the certified stress report for the Dresden III vessel prepared by the Babcock & Wilcox Company, the vessel vendor, and to the special report, also prepared by the Babcock & Wilcox Company, of the Loss of Feedwater Pumps Event in the Dresden II vessel.
- i This comparison leads to the conclusion that the thermal transients i r were no more severe than transients analyzed in the above reports.
he usage factor for the vessel for the subject incident was increased by less than 0.011. Bis conclusion was reached by
; comparing the transients of the subject incident with those of the more severe loss of feedwater pumps event. For this latter event, the increase in vessel usage is 0.011 for each occurrence of the event.
- 5. Performance of the Pressure Suppression System i.
All evidence available indicates the containment system functioned
;- as predicted. Containment pressure and torus water temperature are
. presented in Figures 1 and 2. Since the exact size and duration of j the steam release to the drywell is not known, a detailed containment response calculation is not possible. However, from the containment I, pressure data .available (Figures 1 and 2), it is inferred that the release was of sufficient size and duration to carry most of the non-T condensible gas from the drywell into torus free vol me. Based on ,
I information available, maximum containment pressure was calculated i at 20 psig. h e maximum pressure observed was 20 peig which agrees
; with the value calculated.
e Isumediately following the incident the drywell contained primarily
; a steam atmosphere and the torus free volume contained the non-con-t densibles from the drywell. Under these conditions, initiation of f -. torus cooling would have negligible effect on containment pressure.
k Significant depressurization due to condensing of the steam on cool j drywell surfaces would be expected. I( From this it.is concluded that the torus performance was satis-4 factory; steam condensing was complete and non-condensibles were
. stored in the air space exactly as expected. Mechanically the-inspection conducted by CECO station, G.E. site, S & L engineering, ,
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- 6. Safety Valve Operation "3F" safety valve was tested following removal. Prior to dis-assembly the valve was popped three times. The popping point was approximately.50 psig below the nameplate value. All " pops" were
" clean" and " sharp", and the valve rescated leaktight follocing the first two tests. Follouing the third " pop" the valve leaked upon rescating, probably due to dirt on the seat from the previous
" pops". j The, valve was then disassembled. Nothing abnormal was noted during disassembly. During subsequent reassembly, difficulty was experienced in fitting the parts, and it was discovered that the valve stem was bent approximately. 080 inches. It is believed that the stem was bent during reassembly of the valve.
A theoretical basis for the exact mechanism which caused the safety valve to lift has not been determined. It appears that the problem cannot be solved strictly on an analytical basis as there are too many unknowns. There are several mechanisms which could cause a valve to lif t, all of which are related to the flooding of the steamline with scmewhat subcooled water. Among'. these possible mechanisms are: I 1. Pressure pulses originating from sudden water velocity decreases either caused by rapid steam condensation or . quantities of water falling upon the surface of a partially filled steamline.
- 2. Actual contact of slightly subcooled water against the nozzle or disk of the safety valve causing warping of the nozzle or disk.
The cause of the safety valve lift, regardless of the exact mechanis= of lift, is obviously related to the flooding of the steam-lines at elevated reactor pressures. i-
. 7. DC Crounds in Drywell
- h." During the incident, 125 Volt DC grounds were experienced in l p,
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- g. CORRECTIVE ACTIONS a
- 1. All defective LPRM cabling was replaced.
- 2. All damaged thermal inculation was repaired and/or replaced.
- 3. He "F" safety valve was replaced with a previously tested valve.
- 4. Operability of "3A" electromatic was satisfactorily tested during startup.
- 5. He "3F" safety valve discharge " rams horn" was rotated one " bolt hole" to redirect the opening away from the "3A" electromatic relief valve solenoid operator.
- 6. The feedwater control system calibration was checked and regulating valve response to control signals was verified. ,
- 7. Torque setting of the feedwater regulator isolation valve was increased.
- 8. Operating procedures fer handling water level transients were 4 revised and emphasized.
- 9. Dye penetrant inspection of the " safety valve to main steam line" welds was conducted on the "3F" safety valve. No indications were found.
- 10. All electrical penetrations were tested and found to be well within allowable leakage limits. The bellows on the main steam lines were visually inspected and pressure checked for leakage. Essults were satisfactory.
- 11. All motors in the drywell were visually inspected and mesgered. The one containment cooling fan motor found to have a ground was repaired.
- 12. The damsged paint on the drywell wall was repaired.
- 13. A functional test of all equipment exposed to the drywell incident
{ environment was performed eatisfactorily. I f 14. During startup an inspection was made in the drywell to verify piping I freedom of movement. No anomalies were noted. x 7 15. The primary system was hydrostatically tested at 1000 psig prior to *
?%A startup. .
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- 16. Installation of a 10 ft accumulator and larger air supply piping on the feedwater regulating valve. Following installation a test was conducted to demonstrate that the feedvater regulating valves would not lockout. No lockout occurred during the test.
- 17. Additional modifications to the feedwater control system are being evaluated and will be implemented where considered appropriate.
Possible modifications include the following.
- a. A circuit to automatically trip the reactor feedwater pumps on signal of high reactor water icvel.
- b. Elimination of as much as possible piping vibration in the area of the feedwater regulating valves. This might allow higher gain settings of the control loop and faster response.
- c. Installation of an alternate type of valve positioner,
- d. Change the amplifier section of the master level controller to -
the type used in GE/MAC factory style 2 controllers which provide high and low limits. 'this would prevent the controller from going into saturation af ter a large drop in water level as has been experienced after reactor scram, and would thus allow faster response in closing feedwater regulating valves upon rising water level.
- a. Installation of valve position indication in the main control room for the feedwater regulating valves.
t
- f. Elimination of the run out circuitry. Runout protection circuitry is required for plants which rely on continued ,
feedwater pump operation as an emergency core cooling system. i Because this protection is not required.for Dresden Units 2 and 3, it can be deleted without affecting ECCS or reactor protection capability. Deletion of unnecessary circuitry and associated equipment tends to improve reliability.
- 18. A program to evaluate the phenomons which causes premature safety valve operation is in progress.
- 19. An evaluation of the feasibility of eliminating the reactor recir-culation system sample valve isolation is in progress.
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$. r L - l F. CONCLUSIONS The following conclusions have been reached regarding the Dresden 3 incident of December 8, 1971.
- l. There was no significant release to the envirorument from the incident.
- 2. No compromise to the health and safety of the public resulted from the incident.
- 3. All operations during the incident and post incident recovery period were within Technical Specifications.
- 4. All safety systems functioned as designed, including'High Pressure Coolant Injection (HPCI), Low Pressure Coolant Injection (LPCI), Core Spray, Main Steam and Containment Isolations, Standby Gas Treatment System, Pressure Suppression System, Standby Diesel Generators.
- 5. Feedwater control system performance during the transient was deficient, in that, the control system locked out on low air pressure, probably during rapid valve movement.
Previous experience has demonstrated the inability of the feedwater control system to automatically control level below the high water level trip point ~ for main steam line isolation during a system transient. This was the prianary reason for the need to take operator action.
- 6. Operator response was in accordance with operating procedures .
throughout the incident and post incident recovery with two exceptions. The operator did not reset the feedwater regulating valve lockout condition when it occurred, and he did not trip the feed pump when the water level reached plus 60 inches. \.
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G. CHRONOLOGY The following is the sequence of events from initiation of the incident until the first containment entry for inapection. I Item Event Real Time Elapsed Time From Time Zero . 1 Operating base loaded at 792 MWe for a Dec . 8, 1970 --- test. 2 "3C" condensate-booster pump tripped. 14:13:08 Time Zero 3 operating reactor feed rumps "3A" and 14:13:09 1 second "3C" tripped on low suction pressure. 4 Standby reactor feed pump "3B" started 14:13:10 2 seconds automatically. 5 Reactor scrammed on low reactor water 14:13:32 14 seconds level. 6 Reactor water level reached a low 14:33:33* 25 seconds
- point of approximately minus 20 inches and began to increase.
7 Main steam line low pressure trip - 14:14:13 1 min. 5 sec. Group 1 isolation signal. 8 Main steam isolation valves reached 14:14:14 1 min. 6 sec. 107 closed. . 9 Reactor pressure reached low point 14:14* 1 minute
- of 795 psig.
10 Low reactor water level trip reset. 14:14:29-32 1 min. 21 sec. 11 Turbine tripped on high reactor water 14:15:15* 2 min. 7 sec.* level. . 12 Operator put isolation condenser in 14:15* 2 minutes , service. 13 Reactor wacer level reached 130 14:14:53* 2 min. 45 sec.
. , inches and began filling main
? g., steam lines.
- f. . x , 14 Safety valve "3F" lifted. Reactor 14:18:08* 5 minutes Me*C pressure peaked at 1020 peig. ~
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1 og l 'f w i. os . . 4 ~ Item Event Real Time Elapsed Time From Time Zerg Diesel Generators "3" and "2/3" started automatically. Core Spray pumps st.arted automatically. Low Pressure Coolant and Injection (LPCI) pumps started automatically. High , Pressure Coolant Injection (HPCI) received an initiation signal but tripped on itigh reactor water level. Containment Isolation (Group II) 16 Drywell, pressure reached 5 psig 14:18:39 5 min. 31 sec. 17 Containment pressure peaked at 14:19* 6 min-. 30 sec. approx. 20 psig. 18 Suppression chamber cooling placed 14:20 7 min.
. in service via recirculation through the containment cooling heat exchangers 19 First Local Power Range Monitors (LPRM) 14:21 8 min.
failed.
- 20. Operator manually tripped reactor 14:26:13 13 min. 5 sec.
feed pump "3B". Reactor water level
. at approximately 130 inches on wide ,
,- range indicator) 21 Drywell pressure decreased to 13.5 14:35* 22 min. * .
psig. 22 "C" reactor high pressure trip 14:35:53 22 min. 45 sec. 23 Reactor pressure peaked at 1044 psig ( 24 "C" reactor high pressure trip reset. 14:38:47 25 min. 39 sec. 25 Drywell pressure decreased to 10.5 15:17 1 hour 4 min. psig. j i 26 Drywell pressure decreased to 4.5 18:00* 3 hrs. 47 min.* A e psig. 1 A
- f. L , 27 Installed " jumper" to permit opening -18:00* 3 krs. 47 min.* ,
jMc reactor water sample valves for l
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iindicated ' activity normal J2.7'.~N-m +lM1i UW" < '%i gph . n ' chn.:K,u,,:, ,n .x'10y.pci/1.) - +
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O fm., y J* v+ - g. l 18-1 Item _Even t_ Real Time _ Elapsed Time Frcum Time Zero 29 Drywell pressure had decreased to 20:00* 5 hrs. 47 min. 4.0 psig. 30 Reactor water blowdown established 20:00* 6 hrs. 47 min.* via the cleanup system. Vessel water level at plus 145 inches. 31 Drywell pressure at 1.75 psig. Drywell 12/9/71 12 hours high pressure trips reset. at 03:15 ECCS systems returned to standby condition.
; 32 Restarted Drywell coolers. ,
33 Drywell pressure at 0.63 psig. 08:00* 16.8 hours 34~ Concenced lowering reactor water 04:10 l'3hours level via blowdown to below main steam lines. 35 First sample of containment atmosphere 4:10 13 hours indicated. 1131 - 5.7 x 10-9 uCi/cc Alpha 2.0 x 10-11 u Ci/cc. 36 Reactor water level at 76 inches on 08:00* 16.8 hours
- range indicator.
37 Reactor pressure at 180 psig. 12/10/71* 33 hours . Shutdown cooling placed in service. 00:00 38 Drywell atmosphere sample analysis 6:00 39 hours it.dicated I131 - 1.3 x 10-9 u C1/cc Gross activity Beta ga g 6.3 x 10-11 u Ci/cc Alpha 1.1 x 10 u Ci/cc 59 Initiated drywell deinerting. 07:15 40 hours Reactor water temp. 1860F. 40 Initial drywell entry for atmos- 11:00 43.8 hours phere samples. Results of analysis indicated: . 1131 - 8.5 x 10 0 u Ci/cc . I ., Beta gn=nm - 2.3 x 10-11 u Ct./cc Alpha - 1.8 x 10-12 u C1/cc
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From Time zero 45.8 hours
- 41 Drywell entry for inspection. 13:00*
- Approximate time 9 9
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