99 accidents at nuclear power plants from 1952 to 2009 (defined as incidents that either resulted in the loss of human life or more than US$50,000 of property damage, the amount the US federal government uses to define major energy accidents that must be reported), totaling US$20.5 billion in property damages. Fifty-seven accidents have occurred since the Chernobyl disaster, and almost two-thirds (56 out of 99) of all nuclear-related accidents have occurred in the USA. There have been comparatively few fatalities associated with nuclear power plant accidents.
|
DATE |
LOCATION |
DESCRIPTION |
DEATHS |
I-131 RELEASE IN 1,000 CI |
COST (IN MILLIONS 2006 $US) |
INES LEVEL |
|
January 3, 1961 |
Idaho Falls, Idaho, US |
Explosion at National Reactor Testing Station |
3 |
0.08 |
22 |
|
|
February 22, 1977 |
Jaslovské Bohunice, Czechoslovakia |
Severe corrosion of reactor and release of radioactivity into the plant area, necessitating total decommission |
0 |
|
1,700 |
4 |
|
March 28, 1979 |
Middletown, Pennsylvania, US |
Loss of coolant and partial core meltdown, see Three Mile Island accident and Three Mile Island accident health effects |
0 |
0.017 |
2,400 |
5 |
|
September 15, 1984 |
Athens, Alabama, US |
Safety violations, operator error, and design problems force six year outage at Browns Ferry Unit 2 |
0 |
|
110 |
|
|
March 9, 1985 |
Athens, Alabama, US |
Instrumentation systems malfunction during startup, which led to suspension of operations at all three Browns Ferry Units |
0 |
|
1,830 |
|
|
April 11, 1986 |
Plymouth, Massachusetts, US |
Recurring equipment problems force emergency shutdown of Boston Edison’s Pilgrim Nuclear Power Plant |
0 |
|
1,001 |
|
|
April 26, 1986 |
Pripyat, Ukraine |
Steam explosion and meltdown (see Chernobyl disaster) necessitating the evacuation of 300,000 people from Kiev and dispersing radioactive material across Europe (see Chernobyl disaster effects) |
53 |
7000 |
6,700 |
7 |
|
May 4, 1986 |
Hamm-Uentrop, Germany |
Experimental THTR-300 reactor releases small amounts of fission products (0.1 GBq Co-60, Cs-137, Pa-233) to surrounding area |
0 |
0 |
267 |
|
|
March 31, 1987 |
Delta, Pennsylvania, US |
Peach Bottom units 2 and 3 shutdown due to cooling malfunctions and unexplained equipment problems |
0 |
|
400 |
|
|
December 19, 1987 |
Lycoming, New York, US |
Malfunctions force Niagara Mohawk Power Corporation to shut down Nine Mile Point Unit 1 |
0 |
|
150 |
|
|
March 17, 1989 |
Lusby, Maryland, US |
Inspections at Calvert Cliff Units 1 and 2 reveal cracks at pressurized heater sleeves, forcing extended shutdowns |
0 |
|
120 |
|
|
November 24, 1989 |
Greifswald, East Germany |
Electrical error causes fire in the main trough that destroys control lines and five main coolant pumps |
0 |
|
443 |
|
|
February 20, 1996 |
Waterford, Connecticut, US |
Leaking valve forces shutdown Millstone Nuclear Power Plant Units 1 and 2, multiple equipment failures found |
0 |
|
254 |
|
|
September 2, 1996 |
Crystal River, Florida, US |
Balance-of-plant equipment malfunction forces shutdown and extensive repairs at Crystal River Unit 3 |
0 |
|
384 |
|
|
September 30, 1999 |
Ibaraki Prefecture, Japan |
Workers at the Tokaimura uranium processing facility try to save time by mixing uranium in buckets, killing two and exposing one more to radiation levels above permissible limits |
2 |
|
54 |
4 |
|
February 16, 2002 |
Oak Harbor, Ohio, US |
Severe corrosion of control rod forces 24-month outage of Davis-Besse reactor |
0 |
|
143 |
3 |
|
August 9, 2004 |
Fukui Prefecture, Japan |
Steam explosion at Mihama Nuclear Power Plant kills 5 workers and injures dozens more |
5 |
|
9 |
1 |
Boiling water reactor (BWR) safety systems are nuclear safety systems constructed within boiling water reactors in order to prevent or mitigate environmental and health hazards in the event of accident or natural disaster.
Like the pressurized water reactor, the BWR reactor core continues to produce heat from radioactive decay after the fission reactions have stopped, making a core damage incident possible in the event that all safety systems have failed and the core does not receive coolant. Also like the pressurized water reactor, a boiling water reactor has a negative void coefficient, that is, the neutron (and the thermal) output of the reactor decreases as the proportion of steam to liquid water increases inside the reactor.
However, unlike a pressurized water reactor which contains no steam in the reactor core, a sudden increase in BWR steam pressure (caused, for example, by the actuation of the main steam isolation valve (MSIV) from the reactor) will result in a sudden decrease in the proportion of steam to liquid water inside the reactor. The increased ratio of water to steam will lead to increased neutron moderation, which in turn will cause an increase in the power output of the reactor. This type of event is referred to as a "pressure transient".
The BWR is specifically designed to respond to pressure transients, having a "pressure suppression" type of design which vents overpressure using safety relief valves to below the surface of a pool of liquid water within the containment, known as the "wetwell" or "torus". There are 11 safety overpressure relief valves on BWR/1-BWR/6 models (7 of which are part of the ADS) and 18 safety overpressure relief valves on ABWR models, only a few of which have to function to stop the pressure rise of a transient. In addition, the reactor will already have rapidly shut down before the transient affects the RPV (as described in the Reactor Protection System section below.
Because of this effect in BWRs, operating components and safety systems are designed to ensure that no credible scenario can cause a pressure and power increase that exceeds the systems' capability to quickly shutdown the reactor before damage to the fuel or to components containing the reactor coolant can occur. In the limiting case of an ATWS (Anticipated Transient Without Scram) derangement, high neutron power levels (~ 200%) can occur for less than a second, after which actuation of SRVs will cause the pressure to rapidly drop off. Neutronic power will fall to far below nominal power (the range of 30% with the cessation of circulation, and thus, void clearance) even before ARI or SLCS actuation occurs. Thermal power will be barely affected.
In the event of a contingency that disables all of the safety systems, each reactor is surrounded by a containment building consisting of 1.2–2.4 m (4–8 ft) of steel-reinforced, pre-stressed concrete designed to seal off the reactor from the environment.
However, the containment building does not protect the fuel during the whole fuel cycle. Most importantly, the spent fuel resides long periods of time outside the primary containment. A typical spent fuel storage pool can hold roughly five times the fuel in the core. Since reloads typically discharge one third of a core, much of the spent fuel stored in the pool will have had considerable decay time. But if the pool were to be drained of water, the discharged fuel from the previous two refuelings would still be "fresh" enough to melt under decay heat. However, the zircaloy cladding of this fuel could be ignited during the heatup. The resulting fire would probably spread to most or all of the fuel in the pool. The heat of combustion, in combination with decay heat, would probably drive "borderline aged" fuel into a molten condition. Moreover, if the fire becomes oxygen-starved (quite probable for a fire located in the bottom of a pit such as this), the hot zirconium would rob oxygen from the uranium dioxide fuel, forming a liquid mixture of metallic uranium, zirconium, oxidized zirconium, and dissolved uranium dioxide. This would cause a release of fission products from the fuel matrix quite comparable to that of molten fuel. In addition, although confined, BWR spent fuel pools are almost always located outside of the primary containment. Generation of hydrogen during the process would probably result in an explosion damaging the secondary containment building. Thus, release to the atmosphere is more likely than for comparable accidents involving the reactor core.
A spent fuel pool accident releasing radioactive material to the atmosphere happened in a Mk-1 type BWR reactor in Fukushima, Japan, on March 14, 2011.
The Reactor Protection System (RPS) is a system, computerized in later BWR models, that is designed to automatically, rapidly, and completely shut down and make safe the Nuclear Steam Supply System (NSSS – the reactor pressure vessel, pumps, and water/steam piping within the containment) if some event occurs that could result in the reactor entering an unsafe operating condition. In addition, the RPS can automatically spin up the Emergency Core Cooling System (ECCS) upon detection of several signals. It does not require human intervention to operate. However, the reactor operators can override parts of the RPS if necessary. If an operator recognizes a deteriorating condition, and knows an automatic safety system will activate, they are trained to pre-emptively activate the safety system.
If the reactor is at power or ascending to power (i.e. if the reactor is supercritical; the control rods are withdrawn to the point where the reactor generates more neutrons than it absorbs) there are safety-related contingencies that may arise that necessitate a rapid shutdown of the reactor, or, in Western nuclear parlance, a "SCRAM". The SCRAM is a manually triggered or automatically triggered rapid insertion of all control rods into the reactor, which will take the reactor to decay heat power levels within tens of seconds. Since ~ 0.6% of neutrons are emitted from fission products ("delayed" neutrons), which are born seconds/minutes after fission, all fission can not be terminated instantaneously, but the fuel soon returns to decay heat power levels. Manual SCRAMs may be initiated by the reactor operators; while automatic SCRAMs are initiated upon:
1. Turbine stop-valve or turbine control-valve closure.
1. If turbine protection systems detect a significant anomaly, admission of steam is halted. Reactor rapid shutdown is in anticipation of a pressure transient that could increase reactivity.
2. Generator load rejection will also cause closure of turbine valves and trip RPS.
2. Loss of offsite power (LOOP)
1. During normal operation, the reactor protection system (RPS) is powered by offsite power
1. Loss of offsite power would open all relays in the RPS causing all rapid shutdown signals to come in redundantly.
2. would also cause MSIV to close since RPS is fail-safe; plant assumes a main steam break is coincident with loss of offsite power.
3. Neutron Monitor Trips – the purpose of these trips are to ensure an even increase in neutron and thermal power during startup.
1. Source range monitor (SRM) / intermediate-range monitor (IRM) upscale:
1. The SRM, used during instrument calibration, pre-critical, and early non-thermal criticality, and the IRM, used during ascension to power, middle/late non-thermal, and early/middle thermal stages, both have trips built in that prevent rapid decreases in reactor period when reactor is intensely reactive (e.g. when no voids exist, water is cold, and water is dense) without positive operator confirmation that such decreases in period are their intention. Prior to trips occurring, rod movement blocks will be activated to ensure operator vigilance if preset levels are marginally exceeded.
2. Average power range monitor (APRM) upscale:
1. Prevents reactor from exceeding pre-set neutron power level maxima during operation or relative maxima prior to positive operator confirmation of end of startup by transition of reactor state into "Run".
3. Average power range monitor / coolant flow thermal trip:
1. Prevents reactor from exceeding variable power levels without sufficient coolant flow for that level being present.
4. Low reactor water level indicative of:
1. Loss of coolant contingency (LOCA)
2. Loss of proper feedwater (LOFW)
3. etc.
5. High drywell (primary containment) pressure
1. Indicative of potential loss of coolant contingency
6. Main steam isolation valve closure (MSIV)
1. Redundant backup for turbine trip
2. Indicative of potential main steam line break
7. High RPV pressure:
1. Indicative of MSIV closure.
2. Decreases reactivity to compensate for boiling void collapse due to high pressure.
3. Prevents pressure relief valves from opening.
4. Serves as a backup for several other trips, like turbine trip.