Three Mile Island

Accident

Outlines

Description of the event of the accident

Date of accident

Schematic diagram of TMI UNIT2

The sequence of events(SENARIO)

Lessons learned

Simulation of the accident

TMI-2 analysis

Mitigation of the accident

Timeline

References

Description of the Event

Location of the accident

The Three Mile Island accident was a partial

core meltdown in Unit 2 (a pressurized water

reactor manufactured by Babcock & Wilcox)

of the Three Mile Island Nuclear Generating

Station in Dauphin County, Pennsylvania near

Harrisburg. The plant was owned and operated

by General Public Utilities and the

Metropolitan Edison Co. Metropolitan Edison

pleads guilty to falsifying reactor leak rates

right before the emergency. In fact, if the reactor was shut down for repairs as per

regulations, the partial meltdown would not have occurred at that time. It was the

most significant accident in the history of the American commercial nuclear power

generating industry, resulting in the release of up to 481 PBq (13 million curies) of

radioactive gases, but less than 740 GBq (20 curies) of the particularly dangerous

iodine-131.

Date of accident

In March 1979, an event occurred at the Three Mile Island Unit 2 that resulted in

the first case of melted fuel in a full scale commercial nuclear power plant. There

had been prior cases of small scale fuel melting, e.g. the Fermi 1 reactor near

Monroe, Michigan. TMI-2 was a Babcock & Wilcox unit with a vertical once-

through steam generator. In the event a valve in the secondary system closed and

initiated the sequence of events.

Schematic diagram of the TMI unit 2

10. Pumps

1. Reactor building 11. Steam A. Primary circuit

2. Reactor core 12. Turbine

3. Reactor vessel 13. Alternator B. Secondary circuit

4. Control rods 14. Transformer

5. Pressurizer 15. Condenser C. Condenser circuit

6. Relief valve 16. Water

7. Block valve 17. Cooling tower

8. Drain tank

9. Steam generator

As shown in the last fig. the type of THI unit 2 is the PWR with initiating

event of LOCA(loss of coolant accident). due to inadequate training and

human factors, such as industrial design errors relating to ambiguous

control room indicators in the power plant's user interface.

The sequence of events (SENARIO) was

1- A valve in the condensate system

(between the condenser and the pump

on the secondary side) failed closed,

which reduced the amount of water

being supplied to the steam generator;

the main feed water pumps and the

turbine tripped within seconds.

2- The design of the vertical one-through steam generator is such that

there is not much water on the secondary (non- radioactive) outer side

of the steam generator tubes that will boil to steam when the plant is at

full power and the reactor continues to put out full power; thus all the

water on the secondary side was rapidly converted to steam within

minutes. The emergency feed water pumps, which started as expected,

were unable to inject water into the steam generators because several

valves in the system were closed.

3- The reactor continued to heat the reactor coolant. The reactor coolant

pumps continued circulating the water to the steam generators,

however no heat could be removed by the secondary side since there

was no water in the steam generators. The reactor coolant system

started to heat up.

4- Pressure rose in the reactor cooling system until the reactor shutdown.

A power operated relief valve opened in the line between the

pressurizer and the quench tank. This valve failed to reclose when it

was supposed to - after pressure dropped below the setpoint for

closure. This relief valve continued to discharge to the quench tank.

The fact that the valve was open allowed steam to continue

discharging to the quench tank. Pressure dropped in the reactor

cooling system because the valve was still open (however, due to poor

control board design and a failure to indicate the valve position

properly, the operators did not know the valve was open). The quench

tank has a rupture disc that opens at about 10-12 pounds per square

inch. When this happened, the steam was released to the containment.

5- The pressurizer is normally at about 650F. As pressure dropped in the

reactor cooling system, eventually water in the upper-most area of the

reactor (about 10-15 feet above the fuel) flashed to steam. The

indicated water level in the pressurizer stayed high (the relationship

between the pressurizer and the reactor was like a manometer).

6- The operators turned off the emergency water injection pumps

because they thought there was still water in the pressurizer.

7- The operators turned off the reactor cooling pumps because they were

concerned about damage due to potential excessive vibration.. This

resulted in a steam void forming in the reactor coolant loop. In

addition, a steam bubble formed in the upper part of the reactor above

the fuel. Eventually as the fuel heated, this void expanded. Eventually,

the fuel cladding material overheated. It is likely that some hydrogen

was produced by a chemical reaction between the zircaloy clad and

the steam in the reactor. In addition, the hydrogen normally present in

the reactor cooling system (used to reduce the presence of oxygen and

subsequent corrosion in the system) was released to the containment

through

8- At one point, containment pressure rapidly spiked to 28 pounds per

square inch; then rapidly dropped. This was most likely due to the

chemical reaction of hydrogen with the oxygen in the containment.

9- Water was added to the reactor cooling system and the level raised in

the pressurizer until cooling of the reactor was assured.

1. Reactor

2. Once-through Vertical Steam

Generator

3. Pressurizer

4. Quench Tank or Pressurizer Relief

Tank

Green identifies the Reactor Coolant

System flow path.

Blue on right shows feed water going to

and in the secondary side of the steam

generator

Blue in bottom of containment shows

containment sump.

Blue in upper left shows the Quench Tank. Note steam leaving.

Five lessons learned:

1-Operator training needed to be improved.

Operators a better understanding of both the theoretical and practical

aspects of plant operations and a sound basis for evaluating and

responding to unfamiliar situations. Licensed reactor operator training

today is conducted on full-scale replica simulators of actual plants.

These simulators permit operators to practice and be tested in all kinds

of accident scenarios, making them more proficient and knowledgeable

reactor operators.

Practice emergency plans / operations.

2-Sharing of industry knowledge needed to be more effective

TMI led to the establishment of the Atlanta-based Institute of

Nuclear Power Operations (INPO) and its National Academy for

Nuclear Training. These two industry organizations have been

effective in promoting excellence in the operation of nuclear

plants and accrediting their training programs.

INPO has had a profound impact on the way nuclear plants are

managed and operated. The proof is the steady improvement in

plant performance in the 30 years since TMI. Plant capacity

factors (the ratio of a power plant's average production to its

rated capability) have increased to 91.8 percent in 2007 from

58.4 percent in 1979. Meanwhile, the industry average of

significant events has decreased from an average of 0.9 per year

in 1989 to 0.01 per year in 2006.

3-Fission products don't escape in the real world.

The accident at TMI yielded insight into the "source term"--the

amount of radioactive fission products released in the event of a

major accident. From TMI data we learned that the release of

volatile fission products was three to four orders of magnitude

smaller than that provided for in the 1962 federal licensing

criteria.

This knowledge, that strict leak-tightness of the containment

wasn't a significant factor in reducing fission product leakage to

the biosphere, led to the downward revisions by the NRC toward

a more realistic source term in 1995.

Since that time, numerous experiments have examined the

timing, magnitude, and controlling processes for fission product

releases from the fuel, the primary system, and containment.

Today, the magnitude of the source term available for release in

an accident has been reduced significantly.

4-Control rooms were complex, poorly organized, and did not provide

important information.

Improve design of control room

Improve surveillance and instrumentation of critical

systems required to cool the reactor and stop the escape

of radio nuclides.

Control rooms in the TMI generation of plants weren't designed

with the needs of operators in mind. Craig Faust, one of the

control room operators during the event stated, "I would have

liked to have thrown away the alarm panel.

It wasn't giving us any useful information." The operators were

overwhelmed and unnerved from the "alarm avalanche."

Necessary information wasn't readily available in a convenient

and understandable form. After the event, important safety

system modifications were made to detect and mitigate

inadequate core cooling and post-accident conditions. The next

generation of reactors will have control rooms designed with

human factors in mind and with computer technology that

prioritizes the information operators receive.

5- The consequences of a nuclear accident were less than we thought.

A postulated "worst accident" happened--the TMI-2 core

melted. Yet, there was no "China syndrome." And in spite

of operator errors, there weren't thousands of casualties.

Similarly, the casualties from Chernobyl were largely

limited in number to first responders and, except for

seldom fatal thyroid cancers, far lower than what was

predicted. While these lessons have been learned, we must

not return to pre-TMI complacency.

TMI supervisors aril control roam operators confer in the Unit 2 control

room during the accident.

To simulate the accident

The TMI-2 accident provides a unique opportunity to assess the

capability of codes to simulate a severe accident on a full scale nuclear

power plant. The first two phases of TMI-2accident have been calculated

with ASTEC. The Phase 1 of the transient was characterized by loss of

primary coolant through the PORV until shutdown of all primary pumps.

The Phase 2started with core uncover and involved core heat up and

melting until core refold was initiated by restart of one primary pump.

The overall primary system behavior was well predicted by ASTEC

during both Phase1 and 2. The primary pressure history was well

reproduced (Fig.); it was just slightly underestimated towards the end of

Phase 2. Pressurizer level behavior, which played a key role in the

accident evolution, was very well simulated. Hot leg gas temperature

increase following core uncover and heat up in Phase 2 was reasonably

well predicted by ASTEC.The residual water level in the core at the end

of Phase 2 is in good agreement with TMI-2observations, as inferred

from bottom crust location in the central core ring.

TMI2 primary system pressure

The best simulation of TMI-2 core melt progression in Phase 2 was

obtained using standard clad failure criteria (T > 2300 K and oxide shell

thickness < 300 μm), BEST-FIT correlation (recommended in ICARE2

code) for zircaloy oxidation, and lowering the UO2-ZrO2 ceramic

melting temperature down to 2550 K, according to the interpretation of

some tests. Core degradation and total molten mass calculated by ASTEC

at the end Phase 2 (Fig.) are consistent with the TMI-2 core configuration

hypothesized just before molten core relocation in the lower plenum of

the vessel. The total mass of hydrogen produced in Phase 2 (300 kg) was

very well predicted by ASTEC.

TMI2 core degradation as calculated by ASTEC (end of Phase 2)

Extension of the TMI-2 accident analysis to Phase 3 and 4 including core

reflood and molten material relocation into the lower head of the vessel is

foreseen with the ASTEC V2code version under development provided

with ICARE2 debris-magma model for late phase simulation and

improved models for degraded core reflood.

Final state:

Prompted Severe Accident Research

TMI-2 analysis

Early phase Late phase (corium in core) Late phase (corium in lower head)

ICARE/CATHARE Simulation (IRSN, M. Zabiego,2002)

Mitigation: denotes all measures taken to limit the radiological

consequences of an accident, including: limiting release into containment;

limiting release from the facility; reducing public radiation exposure by

sheltering, evacuation, off-site cleanup, etc. A narrower term, release

mitigation, refers only to measures taken to limit the release of

radioactive material from the facility.

The accident precursor program should have the following

characteristics:

1. The program should be owned by a recognized authority in the

industry and should be driven by consistent, robust goals and objectives

that address the needs of the future. Operational events should be

considered precursors to more serious events; from these precursors, the

program should provide insights into improving safety in the future.

2. The program must be supported by an infrastructure that can sustain it.

A system must be in place for gathering appropriate operational data and

providing access to data providers when more detailed information is

needed. Barriers to full and honest disclosure, such as proprietary

information and fear of repercussions, must be addressed. Also, industry

members must have incentives (either voluntary or by regulatory action)

for participating.

3. The program should provide a trending and tracking system to

correlate changes in industry design and practices with changes in the

occurrence and nature of observed precursors. The system should also be

able to distinguish between changes in trends that reflect real progress in

the field and changes attributable to maturing of the process and program.

The program could then provide excellent feedback to the industry on the

real impact of the precursor program.

4. Systems and methods should be sensitive enough to identify an

operational event as a precursor without generating too many “false

detects” of events of little interest. The event-reporting requirements and

event screening and selection criteria and processes must remain

consistent over time to support trending and analysis.

5. Risk assessment in the industry must be mature enough to instill

confidence that potential accident sequences have been identified and that

the models used to assess events are sufficient and only need changes that

reflect the configurations and operating practices of specific facilities.

Risk models must be updated to reflect improvements in facilities, but

these changes should be made in a way that does not change the level of

detail or the scope of coverage. This will facilitate trending and

comparison over the years.

6. Analysis should be performed on a continual basis by a consistent team

of analysts to ensure the timeliness and consistency of results.

Timeline

Date Event

March 1979 TMI operators are falsifying reactor leaks rates.

March 1979

TMI accident occurred. Containment coolant and unknown

amounts of radioactive contamination released into

environment.

April 1979 Containment steam vented to the atmosphere in order to

stabilize the core.

July 1980 Approximately 1591 TBq (43,000 curies) of krypton were

vented from the reactor building.

July 1980 The first manned entry into the reactor building took place.

Nov. 1980

An Advisory Panel for the Decontamination of TMI-2,

composed of citizens, scientists, and State and local officials,

held its first meeting in Harrisburg, PA.

July 1984 The reactor vessel head (top) was removed.

Oct. 1985 Defueling began.

July 1986 The off-site shipment of reactor core debris began.

Aug. 1988

GPU submitted a request for a proposal to amend the TMI-2

license to a "possession-only" license and to allow the

facility to enter long-term monitoring storage.

Jan. 1990 Defueling was completed.

July 1990 GPU submitted its funding plan for placing $229 million in

escrow for radiological decommissioning of the plant.

Jan. 1991 The evaporation of accident-generated water began.

April 1991 NRC published a notice of opportunity for a hearing on

GPU's request for a license amendment.

Feb. 1992 NRC issued a safety evaluation report and granted the

license amendment.

Aug. 1993 The processing of accident-generated water was completed

involving 2.23 million gallons.

Sept. 1993 NRC issued a possession-only license.

Sept. 1993 The Advisory Panel for Decontamination of TMI-2 held its

last meeting.

Dec. 1993 Post-Defueling Monitoring Storage began.

Oct. 2009 TMI-1 license extended from April 2014 until 2034 without

a public hearing

References:

G. Guillard et al., “ASTEC V1 code: DIVA physical modeling,”

Report ASTECV1/DOC/06-17 (2006).

http://www.nucleartourist.com/events/tmi.htm

http://www.efmr.org/edu/nuclear2009.pdf

http://en.wikipedia.org/wiki/Three_Mile_Island_accident#Lessons_learne

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