ML19275A964
| ML19275A964 | |
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|---|---|
| Site: | Crane  |
| Issue date: | 10/31/1979 |
| From: | PRESIDENT'S COMMISSION ON THE ACCIDENT AT THREE MILE |
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| NUDOCS 7910310253 | |
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Text
.
TECHNICAL STAFF ANALYSIS REPORT ON
SUMMARY
SEQUENCE OF EVENTS TO PRESIDENT'S COMMISSION ON THE ACCIDENT AT THREE MILE ISLAND ADVANCE COPY NOT FOR PUBLIC RELEASE BEFORE AMs, WEDNESDAY, OCTOBER 31, 1979 1184 161
'J 910 3 \\ 07S3
TECHNICAL STAFF ANALYSIS REPORT ON
SUMMARY
SEQUENCE OF EVENTS BY Jasper L. Tew Technical Assessment Task Force OCTOBER 1979 WASHINGTON, D.C.
I184 162
This document is solely the work of the Commission staff and does not necessarily represent the views of the President's Commission or any member of the Commission.
This pre-publication copy is a final document and will be subject only to minor editorial changes in its published form.
1184 163
TABLE OF CONTENTS Introduction Sequence of Events Summary A.
Plant Conditions Prior to the Initiating Event........................................
1 2
B.
The Initiating Event.........................
2 C.
Initial Plant Response.......................
D.
Declaration of Emergency and Stabilization of the Plant.................................
12 Note:
The figures used in the summary are reproduced from work done by the Electric Power Research Institute, in their report NSAC-1, July 1971, the General Public Utilities Sequence of Events, dated by Babcock and Wilcox, published April 30, 1979.
The data used to compile the figures was extracted from the data output from the B&W Reactimeter installed in T lI-2.
The reactimeter data source has been validated by the Commission staff.
Figures 1.
Feedwater Flow...............................
14 2.
Reactor Coolant Pressure and Pressurizer Level........................................
15-16 3.
Reactor Coolart Drain Tank Pressure..........
17-20 4.
Steam Generator A - Water Level and Pressure 21-24 5.
Steam Gerierator B - Water Level and Pressure 25-28 6.
Reactor Coolant Loop Hot Leg (Narrow Range) and Cold Leg (Wide Range) Temperatures.......
29-30 7.
Reactor Building Temperature and Pressure....
31 8.
Out-of-Core Nuclear Instrument Readings......
32-33 9.
Reactor Coolant Loop Flow....................
34-35 10.
Reactor Coolant Loop Hot Leg Temperature 36 (Wide Range).................................
1184 164 1nA i /7 U 't IUJ
Appendix A Decay Heat Removal Methods -
TMI-2.......
A-1 Figures:
Note:
The figures in this section are the same as those above.
Appendix B Significant Equipment Problems...........
B-1 Appendix C Summary of Incorrect Operational Actions.
C-1 1184 165
1 Introduction This report summarizes the technical sequence of events of the TMI-2 accident of March 28, 1979.
It also provides some commentary and or explanation of critical events.
Except where otherwise noted, the commentary and explanation <4 are based upon analysis by the technical staff.
The summar'.
includes a statement of the conditions of the plant at the start of the accident and carries through approximately the first 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br /> at which time forced cooling of the reactor was reestablished placing the plant in a relatively stable condition.
The events herein are extracted from the much longer Catalog of Events available in the Commission files.
Three appendices are attached which provide further discussion of:
A.
Decay Heat Removal Methods B.
Significant Equipment Problems C.
Incorrect Operational Actions The Sequence of Evente Summary has been separated into the time periods where significant events occurred.
A.
Plant Conditions Prior To The Initiating Event -
Three Mile Island Unit Two was at 97% (o f 2') 7 2 MWT) power with the Integrated Control system in full automatic.
Rod groups one through five were fully withdrawn, rod groups six and seven were 95% withdrawn and rod group eight was 27%
withdrawn.
Reactor Coolant System total flow was approxi-mately 138 million pounds per hour and the Reactor Coolant System pressure was 2155 psig.
Reactor Coolant Makeup Pur.sp B (MU-P-1B ) was in service supplying makeup and Reactor Coolant Pump seal injection flow.
Normal Reactor Coolanc System letdown flow was approximately 70 gpm.
The Reactor Coolant System boron concentration was approximately 1030 parts per million.
The pressurizer Spray Valve (RC-V1) and the pressurizer heater bank number 4 was in manual control while spraying the pressurizer to equalize boron concen-trations between the pressurizer and the remainder of the Reactor Coolant System.
The pressurizer relief valve and safety valves discharge header temperatures were high enough to give periodic alarms (at about 200*F) and continuously indicated temperatures in the range of 180 F to 200'F.
These temperatures are significantly above the maximum allowable operating temperature of 130 F.
The Steam Generator and associated secondary plant
- aditions prior to the accident were as follows:
1184 166
s 2
Steam Generator A Steam Generator B Feedwater Flow 5.7459 MPPH*
5.7003 MPPH*
Operating Range Level 56.0%
57.4%
Startup Range Level 159.8 Inches 163.4 inches Steam Pressure 910 psig 889 psig Feedwater Temperature 462.'F 462.'F Steam Temperature 595'F 594*F
- Million Pounds Per Hour Main Feedwater Pumps (FW-P-1A and FW-P-1B), Condensate Boocter Pumps (CO-P-2A and CO-P-2B) and Condensate Pumps (CO-P-1A and CO-P-1B) were in service.
B.
The Initiating Event For approximately 11 hours1.273148e-4 days <br />0.00306 hours <br />1.818783e-5 weeks <br />4.1855e-6 months <br /> prior to 0400 March 28, 1979 TMI-2 operating personnel had been attempting to transfcr spent resin from at isolated condensate polisher unit to the resin regeneration system.
The resin was apparently clogged in the outlet of the polisher and the operators were inject-i'ng a water and air mixture into the polisher to break up the clogged resin.
The air system is isolated only by a check valve (Tag. No. R-I-50 ) while performing the operation and water can leak into the service and instrument air system through the check valve.
Water entering the instru-ment air system can restrict air flow to the control valves for the polisher outlet valves causing them to shut.
This is the most probable cause of the outlet valve cicsure and loss of feedwater.
A few seconds before 0400:37 March 28, 1979 the con-densate polisher out]et valves shut causing the Condensate Booster Pumps to trip due to low suction pressure which caused the Main Feedwater Pumps to trip from low suction pressure at 0400:37.
(Unit 2 Emergency precedures 2202.2-2
" Loss of Main Feedwater to Both OTSG's, Section 2.B states in Manual Action 1. "If loss of FW is due to loss of both feed pumps: a. Trip the Reactor".
The operator did not manually trip the reactor, although it tripped on high pressure 8 seconds later).
The polisher outlet valves had been accidentally shut from water c:tering the air system on at least one other occasion.
(See Tecnnical Staff Report on Condensate Polisher).
C.
Initial Plant Response To the Accident.
(time 0400:37 to approximately 0656) 0400:37-Both Main Feed Pumps trip (figure 1) automatically (00:00:00) tripping the main turbine.
Three Emergency Feed Pumps start automatically.
With the reactor still operating the primary coolant 1184 167
3 began to heat up because the turbine was no longer extracting heat from the system.
Reactor coolant system pressure increases 0400:40-(00:00:03) to 2255 psig, (figure 2) opening the PORV as designed.
The pressure in the Reactor.'oolant drain tank began to increase (figure 3).
Pressure reached 2355 psig and the reactor 0400:45-tripped on a high pressure signal as designed (00:00:08)
(figure 2).
After the reactor tripped the plant began to cool down due to heat rejection through the steam generator relief valves, which had lifted, and the turbine bypass valve, decreasing the plant pressure, (figure 4,
5, and 6).
Reactor coolant system pressure was reduced 0400:49-(00:00:12) to 2205 psig, where the PORV should have closed (the PORV did not close) (figure 3).
into The expected insurge of reactor coolant 260 inches the pressurizer peaked at about and began tc decrease.
The operators, as specified in the operating procedures, stopped letdown flow and started another makeup pump (lA) to compensate for the expected pressuriz-er out-surge as the plant continued to cool-cown.
PORV and one pressurizer safety valve high 0401:07-(00:00:30) outlet temperature alarm were received (temperatures were 239.5'F and 203.5'F respectively).
The operators were aware that the PORV had lifted but thought the valve had closed because the valve position indicator light was extinguished.
This indicator only indicated that power was applied to the pilot solenoid for the PORV and did not indicate valve stem position.
The high temperature was assumed to be a result of the temporary opening of the POPV and an existing leak in either the PORV or a code safety valve.
Although the TMI-2 procedures indicate the PORV will open on a severe transient (Abnormal Procedure 2203-2.2 Turbine Trip, Section 2.0 automatic action A.3 states " Pressurizer Pilot Operated Relief Valve Open".) none of the operating procedures required the operators to make positive checks to ensure the PORV had closed after an increasing pressure transient that approached or exceeded the PORV set point.
1i81 168
4 0401:07-Both steam generators reached the water (00:00:30) level control set point of 30 inches, through (figure 4 and 5), where the Emergency 0401:10-Feedwater Control Valves EF-V-ll A and 11 B (00:00:33) opened.
No water was added to the steam generators because the downstream block valves EF-V-12A and 12B were closed.
The operators were not aware that the block valves were closed.
The question of when the EF-V-12A and 12B valves were closed is address-ed in a separate paper included in the staff technical report.
0401:25-With two makeup pumps (1A and 1B) running (00:00:48) the rate of pressurizer level decrease was reduced and after reaching a minimum of about 160 inches it began to increase (figure 2).
04)1:37-A second pressurizer safety valve high outlet 00:01:00) temperature alarm was recieved.
The indicated temperature was 294.5*F.
The safety valve outlet temperature increase was probably due to the hot reactor coolant being discharged through the PORV which increased the temp-erature of the safety valve outlet piping.
0402:22-Both steam generators boiled dry and effective (00:01:45) heat transfer from the reactor coolant system approx.
to the secondary system stopped, (figures 4, 5 and 6).
0402:38-The open PORV continued to reduce reactor (00:02:01) coolant pressure to 1640 psig where Engineered Safeguards Features (ESF) for High Pressure Injection (HPI) Activated.
ESF Actuation automatically stopped makeup pump 1B, started makeup pump 1C (makeup pump 1A was started previously at 0400:49)and fully opened the makeup valves providing a total injection flow rate of about 1000 gpm to the reactor coolant system.
1184 169
5 0403:"s The HPI portion of the Engineered Safety Features (00:0;:13) was bypassed.
(TMI-2 Emergency Procedure 2202-1.3 Loss of Reactor Coolant / Pressure requires the operator'to
" Bypass" the Engineered Safety Features and throttle the valves to prevent pump runout.)
Note:
" Bypass" only returns HPI to manual control-this action does not change any valve settings (see action at 0405:15) 0403:50-Reactor coolant drain tank relief lifts at (00:03:13) about 122 psig - (figure 3).
0404:03-Reactor coolant drain tank high temperature (00:03:26) alarm occurred.
0404:05-Pressurizer high coolant alarm occurred (260 inches)
(00:03:28)
(Figure 2).
Note:
Over 100 alarms occurred during the first few minutes of the accident.
Note:
TMI-2 operating procedure 2103-1.3 " Pressurizer Operation"- requires the operator to maintain level between 45 and 385 inches, see page 7, para.
2.2.7.
In addition, page 5 para.
2.1.8.
States, "the pressurizer /RC System must not be filled with coolant to solid conditions (400 inches) at any time except as required for system Hydrostatic tests."
0405:15-Operator stopped makeup pump 1C and throttled (00:04:38) the high pressure injection valves.
(Operators had previously bypassed HPI at 0403:50)
Note: TMI-2 Alarm Procedure 2201-13 Alarm 13.A2 (Engineered Safeguards Features Actuation) states that the cause for the ESFA alarm actuation (other than test or channel fai:ure) is "LOCA"
..nd requires followup action with TMI-2 Emergency Procedure 2202-1.3 " Loss of Reactor Coolant / Reactor System Pressure".
Prior to Automatic initiation of HPI at about 0402 the plant response to the turbine trip appeared to be normal and the operators probably had no reason to suspect a Reactor Coolant system leak.
After the automatic initiation of HPI the rapidly decreasing reactor coolant pressure with constant Reactor coolant system temperature was an unambiguous symptom that coolant was leaking from the system.
Subsequent to overriding and reducing the HPI flow the operators increased letdown flow to its maximum value (about 160 gpm) in response to high pressurizer level, which further exacerbated the loss of coolant.
1184 170
6 During the period HPI was activated from about two minutes to about four minutes the reactimeter traces indicate there was no net heat up of the Reactor coolant System (figure 6) indicating that the plant had achieved a heat rejection rate equal to the decay heat and reactor coolant pump heat input.
0406:07-The indicated reactor coolant system hot leg (00:05:30) temperature and pressure reached saturation conditions of 582*F and 1340 psig (Figures 2 and 6) As steam bubbles formed in the Reactor Coolant System they took control of the plant pressure increasing the pressurizer level as the bubble expanded.
0406:28-Pressurizer level indication was off scale (00:05:51) high (greater than 400 inches) (Figure 2) 0408:06-Reactor building sump pump 2A started auto-(00:07:29) matically at a water level of 38 inches.
0408:37-The operators discovered that the Emergency Feedwater (00:08:00)
Block Valves EF-V-12A and 12B were shut and began opening the valves.
Addition of the cold feedwater to the steam generators sub-cooled the reactor coolant system over the next 15 minutes and the system pressure followed saturation temperature (figure 2 and 6).
Since the pressurizer could not regain control of plant pressure, due to flow out of the open PORV, it appears that the eight minute delay in providing feedwater to the steam generators did not aaterially affect the outcome of the accident.
0410:56-Reactor building sump pump 2B started automatically (00:10:19) at a water level of 53 inches.
0411:25-Reactor building sump high level alarm occurred.
(00:10:48)
This alarm is one of the symptoms of a loss of coolant shown in TMI-2 Procedure 2202-1.3 Loss of Reactor Coolant / Reactor System Pressure.
0415:25-The Reactor Coolant Drain Tank (RCDT) rupture (00:14:48) disc failed as designed when pressure increased to about 191 psi (figure 3).
The Reactor Building ambient temperature began to increase rapidly as a result of released steam, (figure 7).
0415:27-Reactor Coolant pump alarms occurred.
(00:14:50)
Reactor Coolant System pressure was about 1275 psig and the temperature was about 570 F at this time (figures 2 and 6).
These conditions are very close to the lower limit for operating the reactor coolant pumps.
The pumps were apparently vibrating due to the voids being formed in the reactor coolant.
i184 i7i
7 0420:37-The out-of-core Neutron Instrument Flux levels (00:20:00) on the source range began to increase, (figure 8). The Reactor Coolant System contained significant steam voids at this time and the source range nuclear instru-ments located outside the resetor vessel and measuring the radiation levels as attenuated by any water in the reactor, were responding to this decrease in density.
Electric Power Research Institute report H5AC-1 dated July 1979 Appendix CI pages 8 through 15 provides an analysis of the out-of-core neutron detectors and their response.
The operators depressed the reactor manual trip pushbutton at 0422;54 as a precautionary action.
0423:21-The steam generator A water level reached 30 (00:22:44) inches and could have been used for heat transfer.
However, the turbine bypass valve control was set to automatically control the steam generators pressure at a value about equal to saturation pressure of the primary coolant.
Therefore, the valve was not effectively used to remove heat from the Reactor Coolant System.
0425:35-The operators requested PORV outlet temperature.
(00:24:58)
The PORV outlet temperature was 285.4 F.
0427:03-Plant status information requested by the (00:26:26) operator was printed out by the utility typewriter:
Reactor coolant loop A hot leg temperature-551.9 F Reactor coolant loop B hot leg temperature-550.9*F Reactor coolant loop A cold leg temperature-541.1 F Reactor coolant loop A cold leg temperature-547.0 F Reactor coolant loop B cold leg temperature-547.0*F Reactor coolant loop B cold leg temperature-546.8 F Reactor coolant loop A pressure-1040 psig Reactor coolant loop B pressure-1043 psig 0430:00-Reactor building temperature and pressure (00:29:23) were increasing rapidly.
The operators responded by starting the reactor building emergency cooling booster pumps and switching all 5 reactor building cooling fans to high speed.
The rate of pressure increase in the reactor building slowed down as a result of these actions (Figure 7).
0433:13-In-core thermocouple 10 R read greater than (00:32:36) 700 F which is the highest reading the computer soft-ware was programmed to record.
The significance of this reading is still not understood since the core was ccvered and being cooled at this time.
1184 172
8 0438:47-Both Reactor Building sump pumps were stopped.
(00:38:10)
Based on the run time and the pumping capacity of to these pumps.
They could have transferred as much as 0438:48 8100 gallons of water out of the reactor building.
The (00:38:11) pumpa were apparently aligned to discharge to the auxiliary building sump tank (which had a failed rupture disc) instead of the miscellaneous waste hold up tank (the level of this tank did not change during the March 28, 1979 operations).
The sump pump, by procedural guidance, could be aligned to either tank.
0440:37-The source 7.ange out-of-core nuclear instruments (00:40:00) continued to show an increasing count rate through (Figure 8) due to the continuing decrease 0500:37 in reactor coolant density.
(01:00:00)
Increases in the Reactor Building background radiation level were shown on the Reactor Intermediate Closed Cooling System Letdown Monitor (lC-R-1092).
Reactor coolant flow had decreased from a normal rate of about 69 million pounds per hour to less than 50 million pounds per hour (Figure 9).
0514:00-Reactor coolant pumps 1B and 2B were stopped because of (01:13:23) the vibration readings and because the plant conditions (approx)
(temperature and pressure) were outside the specified range for pump operation.
0514:00-The out-of-core nuclear instruments, both source and (01:15:23) intermediate range, increased their readings as the to reactor coolant density continued to decrease (Figure 8).
0541:00 (01:40:23)
The PORV discharge line temperature remained at about 283*F.
Reactor coolant flow continued to decrease (Figure 9). There were some momentary indications of steam flow from steam generator A.
The feed flow rate to steam generator B was increased.
Steam generator A boiled dry and a few minutes later feedwater flow to steam generator A was increased (Figures 4 and 5) and was apparently used effectively for a few minutes to remove heat from reactor coolant loop A (Figures 6).
0515:00-Intermediate Closed Cooling System Radiation Monitor (01:44:23)
(lC-R-1092) began increasing from 3500 counts through per minute.
The monitor reached its alarm point of 0518:00 5000 counts per minute at 0518.
(01:17:25) i184 173
9 0518:00-The reactor building air particulate monitor (01:17:23)
HP-R-227 (P) reached its alarm point of 50,000 counts per minute.
Due to the fact that the reactor coolant system had been below the required pressure conditions for fuel rod compression for some time and the core temperature was increasing above normal (at least one core exit thermocouple was reading off-scale high at 0433:03) it is inferred that these radiation monitor readings indicate that fuel cladding was being ruptured mechanically by internal pressure.
The ruptures at this time were probably small but did allow some of the fission gases accumulated in the fuel-to-cladding gap to escape into the reactor coolant system.
0541:22-Reactor coolant pumps lA and 2A were stopped (01:40:45)
(figure 9).
Forced cooling of the core (approx) was terminated.
The source and intermediate range nuclear instruments decreased significantly as the cooler water being held up in coolant loop A hot leg fell back into the core, temporarily increasing the coolant density in the core.
Reactor coolant loop A hot and cold leg temperature both decreased for about 12 minutes.
Then the hot leg temperatures indicated in the control room began to rise rapidly going off scale high (greater than 620*F) within 38 minutes.
The loop A cold leg temperature continued to decrease slowly over the next hour (Figure 6). Reactor coolant loop B hot leg temperature continued to decrease until about 0605 at which time it began to increase rapidly and the indicated temperature in the control room went off scale high (greater than 620 F) within about 25 minutes.
The loop B cold leg temperature continued to decrease (Figure 6).
Figure 10 data, which was not available to the operators in the control room, indicates that superheated conditions existed in both the A and B loop hot legs from about 0615 to about 1430.
About 2 minutes after the reactor coolant pumps stopped, the out-of-core nuclear instruments, both source and intermediate range, began to increase rapidly, indicating boil off of the reactor vessel inventory.
Reactor system pressure at the time the A loop reactor coolant pumps were stopped was about 1000 psi and it continued to decrease rapidly (Figure 2).
1184 174
10 0602:00-Analysis of a reactor coolant sample showed (02:01:23) the gross beta-gamma activity to be 4 micro (approx) curies per milliliter which is about 10 times the normal expected reading.
This sample is a further indication that some mechanical damage to the fuel cladding had been sustained and fission products had been released into the reactor coolant from the fuel to cladding gap space.
~
0615:00-The self powered neutron detectors (SPND's)
(02:14:23) installed in the core began responding to (approx) high temperatures (see Electric Power Research Institute report NSAC-1 July 1979, Appendix C1 page 19-22 for a description and temperature response of these devices) indicating that the water level in the reactor vessel was below the top of the active core.
The response of the SPND's is consistent with the sharp rise in the reactor coolant loop hot leg temperatures which started at about 0615. (Figures 6 and 10).
0622:37-The PORV Block Valve was shut.
Reactor system (02:22:00) pressure began to increase (Figure 2) and the reactor building pressure began to decrease (figure 7) indicating that the PORV was the source of coolant leakage from the system.
0624:00-The Reactor Building Air Particulate Sample (02:23:23)
Monitor (HO-R-227 (P)) reached its alarm point of 50,000 counts per minute for the second time.
0626:00-The area monitor in the reactor building on (02:25:23) the 347 ft. level reached its alarm point (approx) of 50 mr/hr.
0630:00-General radiation levels in the auxiliary (02:24:23) building increased and ranged from about to 10 mr/hr to more than 5 R/hr at the purification 0700:00-valve room door.
(03:59:23) 0643:00-Analysis of a reactor coolant sample taken (02:42:23) at this time shcwed gross beta-gamma activity of 140 microcuries per milliliter.
The area monitor for Unit One Sample Room (RM-G3), which contained the Unit 2 sample lines, reached its alarm set point of 2.5 mr/hr.
1184 175
11 0648:00-Unit 1 Hot Machine Shop area monitor RM-G4 (02:47:23) reached the alarm set point of 2.5 mr/hr.
A survey of the reactor coolant sample line running through this area read 1.5 R/hr.
Note:
For further details of various radiation alarms and information see the NRC Radiological Sequence of events starting on page II-A-I of NUREG 0600.
The reactor vessel inventory continued to boil Summary:
off after the re.
- or coolant pumps were stopped at 0541 and by 0615 are is evidence of super heated steam in the coolant loap hot legs.
Subsequent to closing the PORV ulock valve at about 0622 no significant heat was removed from the core until the block valve was again opened at 0712.
By 0700 the temperature of the hot leg was at least 750*F in the B loop.
The A loop temperature was about 775'F.
No significant change in make up flow to the loops is evident until almost 0720 one hour after the PORV was closed.
Response c- "e SPND's at 0648 indicates that the reactor vessel water level may have been 8 to 9 fret below the top of the active core.
Radiation monicoring instruments and reactor coolant sample analyses had previously indicated that some mechanical damage (ruptured cladding from internal pressure) to the core was occurring about 0602.
By 0650 the high radiation levels indicated by the radiation monitoring instruments indicate that severe core damage was taking place.
It is unclear why the operators,or engineers and supervisors who were present, did not immediately start high pressure injection when the PORV block valve was closed and plant pressure began to increase indicating that the PORV was the source of the leak.
0648:23-The operators managed to get the condensate (02:47:50) system to function automatically by 0650 (see section B of this report).
The difficulty to with this system was found to be a broken air line which 0655:37-The valve supplies operating air to the valve air operator.
(02:55:00) was then opened manually and the system began to control the condenser hot well level automatically. The operating air line was apparently broken during the transient since the hot well level control was functioning norma'.ly prior to the turbine trip.
1184 176
12 Af,ter jumpering interlocks in the control circuits Reactor Coolant Pump 2B was started at 0654:46 and allowed to run for 19 minutes.
The reactor out-of-core nuclear instruments showed a sharp decrease in level as the colder water trapped in the reactor coolant loop cold legs and the B steam generator was transferred into the reactor vessel.
Radiation level increases and alarms in several areas of the plant including Reactor Building Atmospheric Sample monitors and the Hot Machine Shop Area Radiation Monitor led the Shift Supervisor and the Unit 2 Super-intendent, Technical Support to decide to declare a Site Emergency at approximately 0656.
D.
Declaration of Emergency and Stabilization of the Plant (Time 0656 to approximately 1950) 0656:00-A Site Emergency was declared.
(02:55:23)
(approx) 0656:00-Radiation levels continued to increase in the (02:55:23)
Reactor Building, the Auxiliary Building and to in the Fuel handling Building.
0700:00-0705:00-The TMI Station Superintendent arrived in the (03:04:23)
Unit 2 Control Room and assumed the role of Emergency Director.
0705:00-Radiation levels throughout the plant continued (03:04:25) to increase.
The PORV block Valve was opened at to 0713 and closed at 0717.
High pressure injection 0724:00 was manually initiated at 0720.
(03:23:23) 0724:00-A General Emergency was declared by the TMI (03:23:23)
Station Superintendent.
The radiation monitor in the dome of the Reactor Building had reached a reading of 8R/hr which is a specified condition in the TMI-2 Emergency Plan that requires declaration of a General Emergency.
0724:00-The operation of plant systems and components (03:23:23) during this period are summarized as follows:
to 1950:00-A.
From 0712 to 1108 a combination of high (15:49:23) pressure injection flow into the loop and flow out of tha PORV was the principal means of cooling the core.
Based on the out-of-core nuclear instrument readings the re: tor vessel inventory appears to have been recovered to a level above the active core by about 1100 (figure 8).
1184 177
13 Starting at about 1140 a prolonged depressurization B.
of the reactor coolant system began, with a relatively low high pressure injection flow, which may have resulted in some core uncovery as indicated by the out-of-core nuclear instruments (figure 8).
C.
Even though substantial quantities of steam were discharged into the reactor building through the open PORV until it was isolated at 0622, the operators precluded a reactor building pressure rise to 4 psig (the actuation set point for Engineered Safeguards Features Actuation for reactor building isolation) by manipulation of the reactor building ventilation system.
However, a prolonged discharge out of the PORV which started at 0740 caused the first reactor building isolation to occur at 0756.
A pressure spike of at least 28 psig occurred in D.
the reactor building at about 1350.
The pressure spike was apparently the result of a hydrogen burn caused by flammable concentrations of hydrogen which were generated by the zirconium / water reactions that took place during the times the core was uncovered and overheated (See Technical Staff Analysis Report on Chemistry).
Continued operation of the letdown system and E.
other systems such as the waste gas decay system after the core was damaged contributed significantly to the escape of radiation to the environment.
A leak in the waste gas header system was an important factor in this.
(See Technical Staff Analysis Report on Containment).
Plant repressurization was started at 1508 and at F.
1950 forced circulation was reestablished when Reactor Coolant Pump 1A was started and run continuously placing the reactor in a stable cooling mode.
1184 178
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A-1 Appendix A.
Decay Heat Removal Methods - TMI-2
Background
Subsequent to a reactor shutdown decay heat must be extracted f rom the reactor core to preclude overheating.
While the decay heat, due to the energy produced by radioactive decay of fission byproducts, af te r shutdown is only a fraction of the heat produced by the core at full this heat source of 168 megawatts at shutdown drops
- power, to 13 megawatts in one day and to 0.14 megawatt in one year (LA8041MS).
Consequently, this heat must be removed from the core and rejected out of the system or the core will overheat and be damaged.
Three basic rules must be followed in order to protect the reactor core in the event of an upset of normal operating conditions.
These rules are (1) stop the nuclear reaction; (2) keep water over the core; and (3) remove the heat generated by the core.
Stopping the nuclear reaction occurs by actuation of a separate system and is not included in the discussions below.
Keeping the core covered with water and removal of the decay heat are di cussed in the succeeding paragraphs.
Normal Decay Heat Removal Methods When the reactor is shut down, after power operation, by a normal planned shutdown or an event that causes a reactor trip, the decay heat generated in the core is transferred to the reactor coolant and trat. sported to the steam generators as the reactor coolant is circulated by the reactor coolant pumps.
The heat in the reactor coolant is given up to the water in the secondary side of the steam generator which becomes heated and produces steam.
The heat from the secondary water is removed from the steam generator in the form of steam and transported to the condenser through a flow patt.ontrolled and regulated by the turbine bypass valve.
The steam is condensed to water in the condenser, with the heat in the steam being given up to the condenser cooling water system which rejects its heat to the environment.
The condensed water or condensate in the condenser is returned to the steam genrator through the condensate and feedwater system where the cycle starts over.
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A-2 The rate of cooldown of the reactor coolant or the rate of decay heat removal is dependent on the rate of steam flow out of the steam generators.
The emergency feedwater sytem is a backup to the main feedwater system that can use the flow path described for the normal condensate and feedwater sytem or it can provide feedwater separately from the condensate storage tanks.
In the e' vent that the condenser is not available, steam can be dumped directly to the atmosphere through atmospheric dump valves tied into the main steam lines.
Water level in the steam generators is controlled automatically, under normal decay heat removal conditions, as is the rate of heat rejection (steam flow) from the steam generators.
The automatic control system will normally reject only the amount of decay heat being generated, thus keeping the reactor coolant very close to its normal operating temperature.
This designed mode of operation facilitiates returning the plant to service af ter a reactor shutdown.
Should the plant operators desire to cool the plant to a lower temperature they can adjust the controls to maintain a constant cool down rate to the desired temperature.
The reactor coolant system is intact and pressurized during normal decay heat removal and water is added to the loop by the makeup pumps as required during plant cooldown.
When the reactor coolant temperature and pressure has been reduced to a level consistent with the operational capability of the low pressure injection system (about 200'F and 350 psig) it is used in its companion role of a decay heat removal system.
Backup Decay Heat Removal Methods The major backup system for decay heat removal is the Emergency Core Cooling System, comprises (1) a water supply (the borated water storage tank which holds 472,000 gallons),
(2) the High Pressure Injection system composed of pumps and appropriate piping which can inject water into the system up to a rate of 1000 gpm at 1600 psig, (3) the Low Pressure Injection system composed of high flow rate (3000 gpm) pumps and appropriate piping, (4) the core flood tanks which can discharge directly to the reactor vessel if a sufficiently 1184 203
A-3 low pressure is reached, and (5) the Reactor Building Ventilation system to reject heat dumped into the building to the environment.
The Emergency Core Cooling System is designed to function automatically if plant pressure decreases to a preset value of about 1600 psig.
At this predeterm:.ned pressure the High Pressure Injection Pumps start automatically, appropriate valves open, and water is injected into the reactor coolant system.
The Low Pressure Injectio's pumps also start upon receiving the low pressure signal, but do not add water to the system at this time.
The Core Flood tanks do not actuate unless system pressure decreases to a lue lower than the nitrogen pressure in the tanks.
Should plant pressure continue to decrease to the appropriate value, the Low Pressure Injection system and the core flood system will be used to inject water into the reactor coolant system to keep the core covered.
Heat is removed from the core, during periods when the various components of the Emergency Core Cooling System are designed to function, by water flowing into the reactor vessel absorbing heat from the core and flowing out a hole in the reactor coolant system as water or steam.
The design of the Emergency Core Cooling System is based on the ability to keep the core covered under serious accidents postulated to ue caused by ruptures that mightThe occur in the reactor coolant system pressure boundary.
system design is analyzed to show that it is adequate to perform its function for a wide range of bgeak sizes from about.04 ft 2 up to and including a 14.lf t split in the reactor coolant system hot leg.
The Emergency Core Cooling System design and operation presume that a hole in the reactor coolant system pressure boundary is available as a place to reject the heat.
An additional assumption concerning the analysis of the emergency core cooling system design is that there is no middle ground between availability and use of the normal decay heat removal systems an. the need for emergency core cooling.
However, a leak in the coolant system pressure boundary which can be compensated for by running the normal makeup pump, but which is too small to remove all the decay heat generated by the core requires that the normal steam generator decay heat removal path be functional; otherwise, 1184 204
A-4 the decay heat energy of the core will neat up the reactor coolant, expanding it and increasing the system pressure.
Thic process will continue until the PORV and/or one or both of the code safety valves on the pressurizer opens and makes a " hole" in the system to remove decay heat.
The High Pressure Injection system under these conditions should be operated in a manner that maintains the reactor coolant inventory constant (i.e., as much water must be charged into the system as is rejected out through the hole and/or relief valve).
This postulated leak is not analyzed in the TMI-2 FSAR.
Various parts of the emergency core cooling system are designed for multiple use.
Parts of the High Pressure Injection system are used as a normal makeup system, to supply water for Purification System letdown flow, Reactor Coolant Pump seal cooling, and to make up losses due to small leaks, and the Low Pressure Injection system is used as a normal decay heat removal system.
When the normal decay heat removal path through the steam generators is lost and there is no rupture of the reactor coolant pressure boundary the High Pressure Injection system used in this manner requires the operators to create a hole in the reactor coolant system pressure boundary by opening the pilot operated relief valve (PORV) and charging into the reactor coolant system an amount of water equal to that rejected through the relief valve.
This event is not analyzed in the TMI-2 FSAR because total loss of feedwater is not considered to be a credible accident.
Decay Heat Removal During the TMI-2 Accident During the period of time between trip of the main turbine and the reactor trip at 8 seconds into the accident some of the heat generated by the reactor operating at power was removed by (1) the turbine bypass line discharging to the condenser (2) the main steam safety valves in the secondary loop opening and discharging steam to the at mospere and (3) the (PORV) on the pressurizer which opened t about 3 seconds into the transient.
Af ter the reactor tripped 8 seconds into the transient, the steam generators began to cooldown the reactor coolant and the main steam alief valves closed.
The turbine bypass system continued to remove heat until the steam generators boiled dry at about 1 minute 45 seconds into the accident.
The steam generators boiled dry because the emergency feedwater block valves were left shut improperly.
The open PORV i184 205
A-5 continued to reject heat and mass from the reactor coolant system until flow through it was stopped af ter about 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and 22 minutes by closing the block valve down stream of the relief valve.
From 1 minute 45 seconds to 2 minutes the discharge of the PORV, together with modest makeup flow, probably on the order of 300 gpm, was the heat removal method and was clearly not suf ficient as the plant continued to heat up (figure 6).
High Pressure Injection was initiated automatically at about 2 minutes, delivering on the order of 10000 gallons per minute flow to the ractor coolant system.
This high injection flow rate in conjunction with the continuing flow out of the open relief valve removed an amount of heat equal to the decay heat at that time.
Figure 6 shows that the reactor coolant temperature rise stopped after High Pressure Injection commenced.
It then leveled of f and was essentially constant when high pressure injection was terminated at about 4 minutes and 38 seconds.
Terminating High Pressure Injection at 4 minutes and 38 seconds upset the equilibrium of this decay heat removal mode snd the reactor coolant system started to heat up (figure 6).
The plant continued to heat up until feedwater was added to the steam generators (figures 4 and 5), after at about 8 opening the emergency feedwater block valves, minutes.
Between 8 minutes and 30 minutes the reactor coolant system temperature was reduced from about 597 F to 550 *F by use of the steam generators (figure 6) and the open During this period the makeup flow rate was very low, PORV.
probably less than 100 gpm.
Letdown flow of about 160 gallons per minute was started at about 4 minutes 38 seconds (to reduce pressurizer level) further increasing the rate of coolant loss from the reactor coolant system.
The reactor coolant system pressure continued to decrease due to flow out the open relief valve while the loop was being subcooled and reached 1100 psig at 18 When feedwater was rapidly added to the steam minutes.
generators starting at about 18 minutes pressure was reduced to about 1050 psig but returned 1100 psig with a few minutes as the added feedwate.r heated up.
A fairly constant heat balance was maintained from about 18 minutes until the B loop reactor coolant pumps were stopped.
During this period the A and B steam generators, 1184 206
A-6 together with the boil off through the open relief valve, were removing essentially all the decay heat generated by the core and the heat input of the reactor coolant pumps (Sigure 6).
The B loop reactor coolant pumps were stopped at 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> and 14 minutes and both the A & B loop temperatures began to increase (figure 6) indicating that the heat balance was upset by the sharply reduced forced circulation in the reactor coolant system.
The atmospheric dump valves were opened on the secondary side of the A Steam Generator at about 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> and 31 minutes as evidenced by sharply decreasing steam generator pressure.
The combination of increased feed flow and possible opening of the B loop atmospheric dump valves, at I hour and 14 minutes appears to account for the rapidly decreasing B steam generator pressure (figures 4 and 5).
It is concluded that there was flow in both loops until about 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> and 31 minutes when the B steam generator appears to have been isolated and the A steam generator boiled dry. The conclusion is based on the fact that the average temperatures and the differential temperature across the steam generators were essentially equal in both loops.
For average loop temperatures and delta temperatures across the steam generators to be equal each generator must have been removing equal amounts of heat.
The temperature in both loops had been reduced to about 530 F just prior to stopping the A loop reactor coolant pumps at 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> and 41 minutes.
The reactor coolant system pressure followed the decrease in saturation temperature indicating that this heat removal mode was capable of subcooling the system (figure 2 and 6).
When the A loop reactor coolant pumpn were stopped, followed closely by a rapid feeding of the A steam generator at about 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> and 42 minutes, the A and B loop hot leg temperatures began to diverge.
The B loop hot le'g temperatures began to increase and then stabilized for a few minutes.
The A loop hot leg temperature decreased until 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> and 52 minutes at which time it began to increase rapidly from 530*F and was greater than 800 by about 5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />.
The E loop temperature began to decrease at about 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> and 55 minutes and reached 620*F at 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and 22 minutes.
It then began to increase rapidly and reached about 790 *F by about 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> and 15 minutes.
(figures 6 and 10).
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A-7 The extremely rapid heat up of the reactor coolant loop hot legs af ter the last reac 3r coolant pumps were stopped indicates that any fluid t' had been in the loops collapsed leaving only ste in the hot legs which achieved superheated conditions within a few minutes.
In order for superheated steam to be present a heat source with a temperature greater than saturation temperature had to be available to heat the steam, thus it is concluded that a portion of the active core was exposed (uncovered) between 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> and 55 minutes and 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.
Af ter flow in the loops was stopped at I hour and 40 minutes boiling of f the water inventory in the reactor vessel was apparently an ef fective heat removal method until loss of mass began to expose (uncover) the active core.
Steam flow with its low heat transfer coef ficient, apparently was inadequate to remove the heat generated in the exposed fuel; since the fuel began a rapid heat up during this period.
This mode continued until the PORV was shut at 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and 22 minutes.
While it is not possible to show the precise water level in the core, from 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> and 30 minutes to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and 22 minutes, the level can be inferred by use of data from the out-of-core neutron detectors and the response of the Self Powered Neutron Detectors (SPND'S) located at various elevations in the core as indicators of what parts of the core were covered at various time.
These instruments indicate that the water level at 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and 22 minutes could have been as low as 8 to 9 feet below the top of the active core.
The Electric Power Research Institute report NSAC-1 dated July 1979 Appendix Ci pages 1622 provides a description and an interpretation of indications from these instruments.
No effective cooling. occurred between 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and 22 minutes (Block valve shut) and 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> and 19 minutes when high pressure injection was started and maintained for 18 minutes.
This was a period that produced large quantities of hydrogen f rom the zirconium-water reaction at high temperature and significant damage to the core (see the staff reports on chemistry and Core Dama.Je).
Sustained High P:: essure Injection Flow was started at 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> and 26 minutes and maintained until about 9 hours1.041667e-4 days <br />0.0025 hours <br />1.488095e-5 weeks <br />3.4245e-6 months <br /> and 4 minutes.
The core appears to have been recovered by about 6 hcurs and 30 minutes as indicated by the out-of-core nuclear instrument readings.
Heat removal during this period was by ( 1) letdown flow which is believed to have 1184 208
A-8 been close to n.aximum level of 160 gpm, (2) periodic opening of the PORV (open 5 minutes stari e, at 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> and 12 minutes, open 98 minutes startirr 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> and 40 minutes, periodic cycling (about 30 cycle minutes and 7 hours8.101852e-5 days <br />0.00194 hours <br />1.157407e-5 weeks <br />2.6635e-6 months <br /> and 38 minute ween 5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> and 43 (3) and some steam refluxing :3 the steam generator' Reactor coolant pump 2B was restarted a.t 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and 54 minutes and some of the water trapped in the steam generator B and loop B cold legs was returned to the reactor vessel.
See Electric Power Research Institute Reporc NSAC-1 dated July 1979 Appendix Th pages 60-63 for an explanation of the reaction of the loops to the pumps start.
Temperature measurements from the core exit thermo-couples readings obtained between 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> and 5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> and 30 minutes show at least 9 temperatures above 2000 *F with the highest being 2580'F and several above 1000 *F.
The Electric Power Research Institute report NSAC-1 dated July 1979, Appendix CI, pages 16-17 provides an interpretation of this da ta'; the staff agrees with EPRI's interpretation.
Their Appendix CI figure CI-10 is a core map of the observed readings.
During this period high pressure injection flow into the reactor vessel and out through the loop A hot leg then via the surge line o'it the pressurizer when the PORV is open is the cooling flow path.
The continuous depressurization of the Reactor Coolant system, in an attempt to dump the core flood tanks into the system, which was started at about 7 hours8.101852e-5 days <br />0.00194 hours <br />1.157407e-5 weeks <br />2.6635e-6 months <br /> and 38 minutes may have uncovered the core again as evidenced by the increasing out-of-core neutron instruments.
The High Pressure Injection flow during this period was very low and cooling was accomplished by boil off of the core inventory.
Since the conditions of this depressurization and possible uncovery is similar to the earlier uncovery it must be assumed that conditions for zirconium-water reactions also existed during this period.
The continuous depressurization allowed the Itydrogen generated by tl.e zirconium water reaction to be expelled into the reactor building.
This new hydrogen combined with any hydroger. generated and expelled during the earlier uncovery reached a level sufficient to support combustion.
A 28 lb pressure spike shown on the reactor building pressure instrument at about 9 hours1.041667e-4 days <br />0.0025 hours <br />1.488095e-5 weeks <br />3.4245e-6 months <br /> and 50 minutes probably indicates a burn of the hydrogen in the reactor building atmosphere (figure 7).
(See Staff report on Chemistry.)
i184 209
A-9 The depressurization attempt was terminated at 11 hours1.273148e-4 days <br />0.00306 hours <br />1.818783e-5 weeks <br />4.1855e-6 months <br /> and 8 minutes with the closure of the PORV Block Valve.
For the next 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> and 35 minutes there was very little heat removed from the syctem.
The PORV Block Valve was opened for two periods of about 10 minutes each, there was no forced circulation flow and high pressure injection flow was sporadic until about 13 hours1.50463e-4 days <br />0.00361 hours <br />2.149471e-5 weeks <br />4.9465e-6 months <br /> and 23 minutes.
It can be inferred from the actions during this period diat substantial continued core heat up occurred at least until the system was finally filled and reactor coolant pump 1A was started and remained running at 15 hours1.736111e-4 days <br />0.00417 hours <br />2.480159e-5 weeks <br />5.7075e-6 months <br /> and 50 minutes.
Af ter the 1A main coolant pump was started a slow cooling trend was later established and the plant was placed in the natural (no pumps running) circulation mode of cooling on 27 April 1979.
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B-1 Appendix B.
SIGNIFICANT EQUIPMENT PROBLEMS Discussion During the course of the accident there were several equipment problems that may have drawn the operators' attention away from those principle actions necessary to protect the reactor core.
Some of the more important of these problems are discussed below; 1.
Condensate System The initiating event (condensate polisher outlet valves shutting) at a few seconds before 0400:37 blocked the condensate discharge path from the condenser hot well to the section of the condensate booster pumps.
(The condensate polisher is further discussed in the Staff Report on the subject.)
The Polisher tank bypass valve does not open auto-matically in the TMI-2 plant and apparently could not be opened by the operators in the control room, probably due to high differential pressure across the valve.
This valve was apparently opened manually by the auxiliary operators about 59 minutes into the event.
Hot well level control could not be maintained auto-matically due to broken air operating lines to the air These air operator for the condensate reject valve Co-C057.
lines apparently broke at the start of the transient since the system was operating normally prior to turbine trip.
An additional problem was an excessive seal leakage problem that developed on condensate pump 2A which started after the turbine trip.
The actions required to reestablish condenser hot well level control lasted from about 0405 to about 0650.
See Electric Power Research Institute report NSAC-1 dated July 1979 appendix C/FDW for a description of the system.
The following sequence of events illustrates the extensive efforts devoted to the condensate problem.
Note that the Unit 2 shift supervisor left the control room at about 0420 (a critical time) and spent about 45 minutes in the turbine building trying to fix the condensate system (The Unit 1 shif t supervisor was in charge in the Unit 2 control room during this time.)
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B-2 Sequence of events related to the condensate system 0400:30 Condensate pump 1A tripped.
0400:37 Main feedwater pumps lA & 1B tripped.
0400:50 Condenser hotwell how level alarm.
0401:05 Condenser hotwell low level alarm cleared.
0401:50 Condenser high level alarm.
0405:52 Condensate pump 1A started.
0405:52 Condensate boster pump trip signal received three times between 0405:52 and 0407:36 indicates operators were trying to reestablish flow in the condensate system.
0409:35 Condensate pump 1A tripped.
0409:50 Condensate booster pump suction header pressure low alarm.
0414:04 Condensate booster low suction pressure alarm cleared.
0416:20 Condensate booster pump low discharge pressure alarm.
0416:94 Condensate booster pump low suction pressure alarm.
0420:00 Unit 2 shift supervisor leaves control room to go approx.
to turbine building.
0459:00 Condensate booster pump low suction pressure alarm cleared.
0459:00 Condensate polisher bypass valve opened manually.
0459:58 Condensate high temperature alarm.
0500:00 Unit 2 shift supervisor returns to control room approx.
0653:07 Condenser hotwell high level alarm cleared.
0655:00 Condensate reject valve Co-v-57 opened manually.
approx.
0656:00 Condenser hotwell low level alarm.
0701:33 Condenser hotwell level was off scale low.
0701:48 Condenser storage tank low level alarm.
0703:04 The condensate storage tank low level alarm cleared.
0711:47 Condenser hotwell low level alarm cleared.
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B-3 2.
Pressurizer level The TMI-2 operating procedure 2101-1.1 Nuclear Plant Limits and Precautions requires that the pressurizer level not be taken solid, defined as greater than 400 inches indicated level, 9.xcept for hydrostatic tests.
The Technical Specification requires that pressurizer level be maintained between 45 inches and 385 inches for normal operation and not allowed to exceed 385 inches any time the plant is in operation mode 1, 2 or 3.
Consequently the operaters were extremely concerned with pressurizer level during the course of the accident.
The following sequence of events indicates the changes in Pressurizer Level observed by the operators and their responses to these indicates.
Time Sequence of events related to pressurizer level 0401:25 Pressurizer level reached minimum level of 158 inches.
0404:05 High pressurizer level alarm.
Set point is 260 inches.
The operators responded by stopping HPI flow at 0404:38 even though plant pressure continued to decrease.
0405:37 Pressurizer level reached 377 inches, decreased Momentarily, then continued to rise.
0406:28 Pressurizer level off scale high (greater than 400 inches).
The operator responded by initiating maximum letdown flow, then reduced it at 0407:35.
0410:52 Pressurizer level came back on scale and remained between 350 inches and about 390 inches until about 0733.
0733:33 Pressurizer high level alarm.
The level was 271 inches but it went off scale high within a few minutes.
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B-4 0800:00 Pressurizer level was 380 inches with a reactor coolant system pressure of 1500 psig.
1510 Pressurizer level begins a rapid decreasu.
1511 Pressurizer high level alarm clears.
1519 Pressurizer level low level alarm.
Operator responded by starting a makeup pump.
1529 Pressurizer level beginning a steady increase.
1544 Pressurizer low level alarm clears ab about 206 inches.
1554 Pressurizer high level alarm, level is 260 inches.
1622 Pressurizer level reaches 400 inches.
3.
Pressurizer heaters.
Throughout the sequence the operators experienced trouble with the pressurizer heaters tripping.
This tripping could be attributed to grounding due to the moisture being injected into the reactor building during the course of the accident.
The following sequence indicates the magnitude and persistance of this problem.
TIME 0400:45 Pressurizer heater groups 1 throvin 5 off.
These heaters placed in automatic control by the operators at the start of the event and because pressure is high at this time the automatic control has the heaters turned off.
0401:00 Pressurizer heater groups 1-5 are automatically energized.
0654:56 Pressurizer heater groups 1-5 tripped.
0824:00 Pressurizer heater groups 1-5 on.
All heaters approx.
are operable at this time.
0831:00 Pressurizer heater group 10 tripped.
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B-3 0846:00 Pressurizer heater groups 4 and 5 tripped.
0930:00 Pressurizer heater group 3 trips and remains off throughout the remainder of the sequence.
1014:16 Pressurizer heater groups 1 and 2 tripped.
1014:43 Pressurizer heater groups 1 and 2 on.
1144:21 Pressurizer heater groups 1 and 2 off, came on again in 2 seconds.
1150:53 Pressurizer heater groups 1 and 2 tripped.
1355:47 Pressurizer heater group 8 trips.
1406:02 Pressurizer heater groups 1 and 2 on.
1407:56 Pressurizer heater groups 1 and 2 off.
1432:13 Pressurizer heater groups 1 and 2 on.
1439:34 Pressurizer heater groups 1 and 2 off.
1440:28 Pressurizer heater groups 1 and 2 on.
1529:29 Pressurizer heater groups 1 and 2 off.
1545:54 Pressurizer heater groups 1 and 2 on.
1726:46 Pressurizer heater groups 1 and 2 off.
1826:03 Pressurizer heater groups 1 and 2 on.
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C-1 Appendix C Summary of Incorrect Operational Actions (During the TMI-2 Accident on March 28, 1979.)
The purpose of this section is to summarize and eval-uate those actions taken by the plant which were not correct and either cont ributed to the onset of the accident or worsened the final outcome.
It should be understood that this paper does not attempt to show causes or attempt to explain why these actions were taken.
Control room operators, shift foremen, shift supervisors, duty officers, engineers and managers were all involved to various degrees in manipu-lations or decision making partinent to the accident.
Also highlighted are inactions which affected the outcome.
Not mentioned in this analysis are the actions and decisions which were correct.
With respect to errors committed before the accident transient three are important:
permitting the emergency feedwater block valves to be shut during operations, not shutting the pilot operated relief valve block valve and permitting water to enter the condensate polisher control air system.
The emergency feedwater block valves, EF-V-12A and 12B were almost certainly closed at the commencement of the transient and were most probably not reopened following a surveillance test on March 26, 1978.
(Although this is not proven and there are other possible courses.)
In any event shift reliefs were not conducted in such a fashion as to ensure the relieving control room operator was appraised of plant status as required by Administrative Procedure 1012,
" Shift Relief and Log Entries."
Not noting that indicating lights on the front control panel showed the block valves to be out of position suggests a lack of rigor regarding reactor operators attention to the conditions of the control panel indicators.
(See staf f report on these valves for further details.)
For a lengthly period before 28 March temperature detection in the discharge piping downstream of the pilot operated relief valve (PORV) and the code safety valves indicated that one of these valves was leaking.
The temp-erature of the PORV tailpiece nas nearly 200'F; periodi-cally, safety valve discharge header temperature alarms, which occur at 200 F, had been received.
The PORV isolation valve, RC-V2, had not been shut as Emergency Procedure 2202-1.5, " Pressurizer System Failure," requires it to be when the relief valve discharge line temperature exceeds 130 F.
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C-2 Immediately prior to commencement of the transient operators were attempting to transfer spent resin from an isolated condensate polisher unit to the resin regeneration system.
The resin was apparently clogged in the outlet of the polisher and the operators were injecting a water and air mixture into the polisher to break up the clogged resin.
The air system is isolated only by a check valve during this operation and water can leak into the service and instrument air systems. Water entering the instrument air system can restrict air flow to the control valves for the polisher outlet valves causing them to sense a requirement to shut the valves.
Inadvertent isolation of the condensate polishers had occurred during other similar spent resin transfers.
Prccedures for the clearing of resin from the system failed to take this into account.
The operators could have been instructed to bypass and remove the polishers from service prior to attempting this evolution to avoid the problem.
Following isolation of the condensate polishing system on March 28 there followed other operational errors which will be discussed below.
Upon the loss of both feedwater pumps the immediate actions of Emergency Procedure 2202-2.2, " Loss of Steam Generator Feed," require the operator to manually trip the reactor regardless of whether an automatic trip took place The operator did not immediately trip the reactor or not.
as required by the procedure.
Following a turbine trip Abnormal Procedure 2202-2.2 states that the PORV will open indicating that this is an expected phenomenon.
The valve should be verified shut following actuation.
Operators relied upon an inadquate panel indication that the valve had been commanded to close but failed to recognize symptoms which indicated that the valve was open.
Th'ese symptons included a rapidly decreasing reactor coolant system pressure to saturation, reactor coolant drain tank pressure increase, PORV discharge high temperature alarm, pressurizer safety valve high temperature alarm, reactor coolant drain tank high temperature alarm, Reactor Building radiation alarms, Reactor Building sump high level alarm and Reactor Building temperature and pressure increase.
There was sufficient information in the control room to know that the pilot operated relief valve was stuck open.
The operators failed to verify that the emergency feedwater pumps were not only running as automatically
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C-3 required by the transient but that they were also delivering feedwater to the steam generator as required by Abnormal Procedure 2203-2.2, " Turbine Trip" and Emergency Procedure 2202-2.2, " Loss of Steam Generator Feed."
The lack of emergency feedwater flow resulted in the steam generator boiling dry.
^
Neither the reactor operators, senior reactor opera-tors, the duty officer, unit superintendent or station manager recognized that a loss of coolant accident was in progress.
Emergency Procedure 2202-1.3, " Loss of Reactor Coolant / Reactor Coolant System Pressure" lists ten symptoms of a loss of reactor coolant, of these the following were decreasing Reactor present and recognized by the operators:
Coolant System pressure, high radiation levels, high Reactor Building temperature, high Reactor Building sump level, high Reactor Building pressure.
One important symptom, decreasing pressurizer level, was not present.
This latter sympton confused the situation and they concluded incorrectly that they had a steam leak.
Also confusing them was the fact that one of the ten symptoms, high Reactor Building pressure, is listed as common to both a loss of coolant or a steam leak.
It is not clear what the operators considered as the cause for rapidly decreasing reactor coolant system pressure.
This could only be due to a loss of coolant as is recognized by the title of the emergency procedure.
The control room operators bypassed the high pressure injection portion of the Safety Features Actuation System (SFAS) less than five minutes after the commencement of the accident.
This was in violation of Operating Procedure 2105-1.3, " Safety Features Actuation System" which requires that SFAS be fully enabled except during maintenance or it was consonant with the requirements of testing.
- However, Emergency Procedure 2202-1.3, (Loss of Reactor Coolant),
Section B, paragraph 3.4 which stipulates that the Safety Injection Channels should be bypassed in order to protect (prevent pump runout) the High Pressure Injection Pumps (HPI).
Statements from the operators suggest that they were concerned about the abnormal pressurizer level increase rather than HPI pump limits and therefore they considered the previsions of Emergency Procedure 2202-1.3 were of lessor priority in guiding their actions.
Apparently all those present in the control room were concerned about the high pressurizer level indication.
They were confused and did not understand the significance of 1184 218
c-4 this phenomenon.
However, they realized that the system was not reacting as if it were solid.
"We knew that we weren't solid."l/ The apparent principal concern was that a full pressurizer and a non-solid system indicated the presence of a bubble in the reactor coolant system other than in the pressurizer.
"We were sitting there trying to figure out how the heck we were going to cool this thing down--g(d that thing [ pressurizer bubble] back-without aggravating our problem."2/
The Shift Foreman (Schiemann) when interviewed immediately after the accident indicated that when the pressurizer "was up to the top" he was not concerned prin-cipally about pressurizer level ger se but rather where the water was coming from.
"We were also watching all our other tanks to see where we were getting the water from.
We couldn't find any other water source coming in and we contin-ued maximum letdown"3/
The increased letdown flow coupled with the reduced HPI flow hastened core uncovering.
Af ter emergency feedwater restored steam generator levels approximately 23 minutes into the transient the steam generators were not used effectively to remove decay heat.
The turbine bypass valve control was set to automatically control the steam generator pressure at a value about equal to saturation pressure of the primary coolant.
Therefore the valve could not open to remove heat from the reactor coolant system.
If the plant operators had used the steam generators as an effective heat sink cooling of the reactor coolant system would have taken place ameliorating plant saturation conditions and, as a minimum, would have delayed core uncovering.
It was not recognized that heat was not being effect-ively removed from the reactor early in the accident.
When Mr. Faust was interviewed on March 30 he offered some insight concerning removal of decay heat.
"As far as I'm concerned, once the turbine is down I don't have a source of steam going out there-so I'm safe there as far as pulling any more heat off or too much heat out of the core.
The reactor starts cooling herself, so the idea is just to stabilize out down at saturation for about 547*F temperature (in the secondary system). O.K.?"5/
The plant status information requested from the computer by the operator 27 minutes into the accident indeed showed that the primary reactor coolant system was also at conditions of about 547 F and 1040 psig indicating that no heat was being removed and that the primary system was also at saturation conditions.
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C-5 The increased readings on the out-of-core source range and then the 1: termediate range nuclear instruments, as would be expected when the coolant density decreased signi-ficantly or as the core becomes uncovered, were not understood.
These readings were misinterpreted to indicate that the reactor was starting up (becoming critical) although control rods were inserted.
Thirty minutes after the commencement of the accident Reactor Building high temperature alarms were received.
This was also not recognized as an important symptom of a loss of coolant accident.
In response the operators merely started the Reactor Building emergency cooling booster pumps and placed the Reactor Building cooling fans in high speed.
This slowed the rate of building pressure increase and delayed containment actuation.
When plant conditions degraded to outside the range specified for reactor coolant pump operation, and the coolant contained steam-produced voids causing cavitation and vibration the operator tripped the pumps.
However, they o f pumps,
failed to recognize that the cause for these degraded condi-tion was a loss of coolant.
The act of tripping the pumps was in response to the equipment protection provisions of Operating Procedure OP 2103-1 4,
" Reactor Coolant Pump Operations."
The operators did not take appropriate action to correct the condition causing the pump operating limits to be reached, that is, a loss of coolant.
After the PORV block valve was shut at about 0622 ne significant heat was removed from the core until about O'/12 when the block valve was again opened.
It is not clear why the operator did not immediately start high pressure inject-ion when shutting the PORV block resulted in a sharp pressure increase and clearly indicated that there had been a loss of coolant for over two hours through the open PORV.
As hot leg temperatures increased to offscale (high) values there is no evidence that the operator recognized the significance of this indication.
Rather, it appears that they inferred that average coolant temperature had stabi-lized near the top of the indicated range.
The Station Manager was informed of the turbine trip and reactor trip about two minutes after these events occur-red.
He issued no order because he assumed the Unit Superin-tendent would.6/
1184 220
C-6 the Duty Officer of conditions in the plant.Approximately 0500 the S Althtugh he was disturbed, Subsequently the Station Manager set up a conference callthe Statio including the Vice President, Generation, the B&W On Site Representative, and the Duty Officer in the Control Room.
No instructions were given to the Duty Officer as a result of this lengthy telephone conversation.8/
Approximately three hours after the beginning of the accident the Station Manager arrived in the control room.
A site emergency had been declared and the Station Manager assumed the role of Emergency Director.
ing operation of the plant.9/Again, he issued no direction concern-The Unit 2 Superintendent was called shortly after 0400 and was informed only that there had been a turbine trip and a reactor trip.
He assumed that the Duty Officer would be called.
When the Unit Superintendent arrived in the Control Room at about 0545 he tried to ascertain what was taking place.
He stated that he did not pay attention to reactor coolant system temperature.
He did not issue any order concerning reinitiating high pressure injection.
When the Station Manager arrived the Unit Superintendent was placed in charge of ensuring the Emergency Plan steps were carried out.10/
The Duty Officer, who was not licensed on Unit 2 arrived in the Control Room at about 0450 and was the first engineer on site.
He was briefed on the situatienll/,
considered that he was not sufficiently familiar with the plant and directed that additional technical and operation support be called in.12/
He offered no specific guidance to the operator either before, during or after the conference call with the Station Manager, et al.
In summary key management personnel expected the opera-tors to keep them informed of the situation but they did not in turn provide the Shift Supervisor with instructions which would place the plant in a safe condition.
Operational errors, then, appear to be a significant factor in the accident.
It can be argued that overall actions taken worsened the sit;ation during the first 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br />.
The core was damaged because adequate cooling was not provided to remove decay heat.
The operators apparently did not understand that they were not providing adequate cooling to the core.
They were confused by the phenomenon of high pressurizer level and low reactor coolant system pressure 1184 221
C-7 which they had not experienced before.
Nevertheless, they believed that high pressurizer level indicated a fr.il reac-tor coolant system.
The procedures used by the operator did not recognize the phenomenon of high pressurizer level and low reactor coolant system pressure could occur.13/
There-fore, the operating procedures did not provide adequate guidance and reactor safety was dependent on the operator's ability to comprehend the significance of key parameters such as reactor coolant systeri pressure and temperature.
The early throttling of high pressure injection for a period of time is understandable; operators had con altted the same error at Davis-Besse 1 in 1977.
However, in the face of continuing indications that the core was being hazarded serious errors were committed in continuing letdown and not injecting water into the core.
1i84 222 GPO e6262