ML19224D201
| ML19224D201 | |
| Person / Time | |
|---|---|
| Site: | Crane  |
| Issue date: | 06/04/1979 |
| From: | Alderman H Advisory Committee on Reactor Safeguards |
| To: | Advisory Committee on Reactor Safeguards |
| Shared Package | |
| ML19224D202 | List: |
| References | |
| ACRS-SM-0135, ACRS-SM-135, NUDOCS 7907110076 | |
| Download: ML19224D201 (8) | |
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UNITED STATES
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NUCLE AR REGULATORY CCM'.!iSS!CN f
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ACRS MEMBERS gjf/ j ACRS TECHNICAL STAFF uf-ACRS FELLOWS O'/ERSIGHT HEARINGS ON THREE MILE ISLAND
'CUSE C:MMITTEE ON INTERICR A1D INSULAR AFFAIRS, CHAIRMAN TEPRESENTATI'/E MORRIS U:ALL, MAY 21, 1973
SUMMARY
OF HEARING Representative Udall opened the Hearin9 We sta*ad +5e +sooics c be covered are:
1.
The March 2Sth incident at Three Mile Island.
2.
The future of commercial nuclear pcwer.
3.
An assessment of risk to the public of the coeration of nuclear reactors.
1 A report from the Task Force on Three Mile Island.
5.
Issues for further review.
5.
The Michelson vmo of 1978.
7.
Status of SRC inquiry into accident.
Representatives ' cleaver and Cheney presented the Task Force Report. The significant aspects of the incident were:
1.
The auxiliary feed water valves were blocked.
2.
The power operated relief ocen procerly but then stuck open allowinc a release until *he blocking valve was closed.
3.
The design of the 3&W system resulted in more drying out of the steam generator.
2 "Se ethoc of core level indication is not reliable.
5.
'he ocerator shut o## the coolant pumcs.
His Tay have been a recuired ocerator action.
5.
- ontainment release to auxiliary building - design er or?
260 113 7 9071100g, m
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2 7.
Conflicting testimony regarding the 23 pound pressure spike in the containment.
8.
Cperators couldn't corralate core temperature fecm the strip charts.
Reactor operators should have been made aware of the ?.0.R.V. sticking open at other. plants and errors in the pressurizer level indication.
The Task Force listed th-i-r;ctant issues as:
1.
Significance of transients at other Babcock and Wilcox plants.
2.
The response (or lack of) to Carl Michelson's letter.
3.
The frequency of the ?.0.R.7. failing to close.
4 The communications between the State and the NRC.
5.
The role of the NRC in decision making during the accident.
6.
The adequacy of radiation monitoring for the area.
The Chariman of the Task Force presented his views:
1.
Instrumentation to measure containment parameters set too low to be of any value.
(under accident conditions ?)
2.
Vulnerability of nuclear power plants to power failure.
3.
Extreme vulnerability of nuclear power plants.
A.
Accident at Three Mile Island could happen again.
5.
"Have we been told the truth"?
6.
The NRC told the Committee 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> after the accident that the accident was serious but not grave.
7 The operators did not know the extent of the accident until one to two days after the accident.
Carl Michelson presented his scenario of the accident.
Mr. Michelson said he was procasing his own views and not that of the ACRS or TVA.
1.
Loss of feedwater to the steam generator.
2.
3.0.2.7. opened about three seconds.
3.
About eight seconds - reactor scrammed.
A.
About nine seconds - relief /alves en steam generator ccened.
260 110
, 5.
About thirteen seconds relief valves supoosed to close (about 2200 pounds).
6.
Thirteen seconds - operator started high pressure makeup.
7.
About thirty seconds, auxiliary feedwater up to pressure, not deliver-ing flow.
3.
Operator observed increasing pressure but decrease in level.
9.
About one minute - increasing level but decreasing pressure.
10.
If it was a loss of coolant accident, the level should be decreasing, 11.
Steam generator proceeding to dry out.
12.
(Missed item 12) 13.
Pressure in steam generator dropping.
14 Two minutes - high pressure injection automatically started.
15.
About four and one-half minutes - pressurized level ;oing out of range.
16.
One pumo tripped and the remaining pumo throttled back to prevent pressurizer from going solid.
17.
Abnut six minutes pressurizer off scale - concerns about going solid.
13.
Hot water in exit of core flashing to steam.
19 Possibility of void formation.
20.
About seven minutes - ocerator started let down cooling - reported maximum rate.
21.
About eight minutes - steam generator not operating as supposed to -
ocened block valves.
22.
About ten minutes - pressurizer level in range.
23.
About 20 minutes - operator increased water level in steam generator.
21 At this coint - exoerienced net loss of water from system - void fo rma ti o n.
25.
About one hour - getting near inventory crisis but not getting indication of it.
26.
Cne to two hours - pumos vibrating and immeter values droccing.
27, One hour and thirteen minutes - one R.C.3 shut o f" 23.
One hour and twenty seven minutes - susoected leak in steam generator -
isolated steam generator.
260 .1
-A-23.
One hour and forty minutes - operating pumps ammeter reading icw -
pumps not effective.
30.
The inventory of a steam cenerator acts like a sychon to the reactor cooling system.
31.
Pressuri:er loop seal maintains an inventory of water in the pressurizer with low coolant level.
The pressurizer had a large trapped volume of water.
32.
Two - three hours - continuing to letdown - trying to maintain water level - actually lowering water level.
33.
Water - bottom of coreysteam heating - uceer core) operator should have known from pressurized level-core uncovered.
34 Steam super heated cladding - good conditions for metal
. vater reaction. Hydrogen accumulated from metal - water reaction.
35.
Radiation alarms - failure of fuel.
36.
Two hours fifty five minutes - site emergency.
37.
Three - four hours - attempt to refill system.
38.
Sixteen hours - purge hydrogen and condense steam - start reactor coolant pump with one steam generator operation.
CUESTIONS AND ANSWERS Q.
Rep. Weaver - At about 100 minutes, the chart showed verv laroe temperature differences between the hot and cold legs. 'elhat is the signi ficance of this ?
A.
At 100 minutes the pumps were tripped - natural circulation was to occur.
Natural circulation did not take place.
The hot leg grew hotter and the cold leg grew colder.
Q.
Rep. Udall - Tell us about the origin of ycur memo.
A.
Michelson - the paper was the result of TVA concerns about sa'ety issues.
O.
To whcm was the :acer directed.
A.
Originally not to anyone.
It was for TVA management.
It was sent to 3&W for review acout April 1973.
C.
Did you hear from them?
A.
3&W did not seem to be paying sufficient attention.
Oeceived 3 real; January 1979.
The reply indicated that the cressuri:er level indication was not a suf#icient indication.
9-LD,
5-Q.
Did 3&W promise to revise the system?
A.
Not in the letter.
Q.
Are difficulties in B&W design unique to B&W systems?
A.
This allcws the pressurizer to remain full with an empty vessel, Peo Cheney Q.
Are B&W instructions regarding pressurizer level adequate for safe operation?
A.
Have not reviewed material in sufficient detail to answer at this time.
Q.
What would have happened if a safety valve opened?
A.
If a safety valve opened and stayed open - could not terminate -
could have same sequence of events.
Q.
Does TVA still anticipate using 3&W reactors?
A.
Yes.
Rec. Singhman Q.
Is there scme technical reason that the water in the core could not be measured directly?
A.
The water level above the core could ce read.
It would be a post accident reading.
During normal operation it would read full.
Q.
Isn't it clear that there should be a means of measuring water level?
A.
Agreed.
Cao. Edwards Do you still feel that~ failure to respond to open relief valve was an operator error, s.
There were four types of problems at Three Mile Island.
1.
Cesign problem - the 1 coo real.
2.
Analysis problem - industry failed to adecuately analyze small breaks.
3.
Procedural orablems - operator folicwed crocedures ' eyed to pressurizer level, 260 117 4
Operator judgment - operator saw relief tank pressure and temoerature going up.
Indicated " stuck-open' relief or safety valve.
Q.
Are you aware of other similar problems?
A.
There are a number of generic safety issues that are being examined by NRC and the ACRS.
The NRC Commissioners were the next witnesses.
There wasn't any prepared tes ti mo ny.
Cuestions and Answers Q.
Rep. Weaver - Why don't we have reactor water level gages ?
A.
Chairman Hendrie - people never expected to see reactor levels drain down under accident conditions. We are looking very hard at this problem.
Q.
Mr. Ebersole received Michelson's analysis.
Has this been handed to anyone in the NRC?
A.
Mr. Mattson received it late 1977 or early 1978.
Mr. Ebersole handed it to a member of the NRC Staff during the Pebble Sorings hearing.
Q.
Wasn't it perceived as significant enough to be acted upon?
A.
It was submitted informally and not cerceived as significant.
G.
March 23, Mr. Miller the shift supervisor recorted core temoeratures about 2400 degrees. The core was believed hot but covered.
Could you explain significance?
A.
There was some doubt about the accuracy of the reading.
Q.
Stello heard of the 2000 degree reading. Mattson had not. When did you hear of this reading and what did you do about it?
A.
I can't answer at this time--suspect it was Friday morning.
Rec. 31ngham to Dr. Hendrie Q.
uow do you feel about NRC procedures and what 3re you going to do about them?
A.
We are going to evaluate end derive significance from coerating reocr s.
We should have evaluated precursor events and made corrections as needed.
We need #urther accident analysis of the transients "olicwing small breaks.
Further empnasis is needed on coerator training and licensing.
2/0 11P d
Iv Same O to Commissioner Kennedy A.
We need re-evaluation and study of generic issues.
We need resolution of unresolved safety issues.
Further work is needed on inspection and licensing. More effort is needed on quality assurance.
Same Q to Commissioner Gilinsky A.
We should pay close attention to operating experience.
Simulator training should be more er. tensive. We should stress emergency planning and procedures.
Our greatest failure is the failure to communicate to industry the level of technical comoetence and diligence required.
Same Q to Commissioner Bradford A.
Communications should be improved.
I would not want to 1ssue more operating licenses and permits until the ful'r significance of Bree Mile Island has been evaluated.
Same Q to Commissioner Ahearne A.
A sense of comolacency has affected licensing.
A change of attitude is necessary.
Q.
What is the time table for the I?tI investigation?
A.
I&E will take this sumer licensing - longer than that over all -
about six months completion about lovember 1, Q.
The containment isolation system differs from plant to plant
'.ihy such variety?
A.
This design is by the A-E.
Each one is different.
The containment isolation for TMI should have been connected to safety injection.
This aspect is under study.
Q.
When was communciation about the pressure soike relayed to NEC?
A.
The containment soike Friday morning.
The pressure soike in the quench tank - after Friday morning.
Q.
What ' as the Commission learned about crisis management?
r A.
The entire process of crisis managemenc is di
' ult. We need crecared plans.
An investigation is going on.
O.
Have you develo::ed directives governing dissimation c# 7I information ?
5 Yes - Centon's ceoole are involved in disseminating this in#crmation.
260 119
.g.
Q.
To Commissioner Gilinsky Would more simulator training for operators be of value?
A.
The simulator training would have to be more complex.
!t would have to stress unusual situations.
Q.
To Commissioner Gilinsky What
- your position regarding " hold up" of new permits and licensing?
A.
We can't continue our business as usual.
I suggest additional operator training and emergency planning.
NLwa Herman Alderman Reactor Operating Experience Engineer s
260 120