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Fukushima nuclear accident

From Wikipedia, the free encyclopedia

https://en.wikipedia.org/wiki/Fukushima_nuclear_accident

Fukushima nuclear accident
Part of the 2011 Tōhoku earthquake and tsunami
The four damaged reactor buildings (from left: Units 4, 3, 2, and 1) on 16 March 2011. Hydrogen-air explosions in Units 1, 3, and 4 caused structural damage.
Date	11 March 2011; 13 years ago
Location	Ōkuma and Futaba, Fukushima, Japan
Coordinates	37°25′17″N 141°1′57″E
Outcome	INES Level 7 (major accident)
Deaths	1 suspected from radiation (lung cancer, 4 years later), and up to 2,202 from evacuation-related stress among the elderly.
Non-fatal injuries	6 with cancer or leukemia, 16 with physical injuries due to hydrogen explosions. 2 workers hospitalized with radiation burns
Displaced	+164,000 local residents

The Fukushima nuclear accident was a major nuclear accident at the Fukushima Daiichi nuclear power plant in Ōkuma, Fukushima, Japan which began on March 11, 2011. The proximate cause of the accident was the 2011 Tōhoku earthquake and tsunami, which resulted in electrical grid failure and damaged nearly all of the power plant's backup energy sources. The subsequent inability to sufficiently cool reactors after shutdown compromised containment and resulted in the release of radioactive contaminants into the surrounding environment. The accident was rated seven (the maximum severity) on the INES by NISA, following a report by the JNES (Japan Nuclear Energy Safety Organization).

No adverse health effects among Fukushima residents or power station workers have been documented that are directly attributable to radiation exposure from the accident, according to the United Nations Scientific Committee on the Effects of Atomic Radiation. Insurance compensation was paid for one death from lung cancer, but this does not prove a causal relationship between radiation and the cancer. 6 other persons have been reported as having developed cancer or leukemia. 2 workers were hospitalized because of radiation burns, and several other people sustained physical injuries as a consequence of the accident. Criticisms have been made about the public perception of radiological hazards resulting from accidents and the implementation of evacuations (similar to the Chernobyl nuclear accident), as they were accused of causing more harm than they prevented. Following the accident, at least 164,000 residents of the surrounding area were permanently or temporarily displaced (either voluntarily or by evacuation order). This response resulted in at least 51 fatalities, with more in following estimates, mainly attributed to subsequent stress or fear of radiological hazards, among older people for the most part.

Investigations faulted lapses in safety and oversight, namely failures in risk assessment and evacuation planning. Controversy surrounds the disposal of treated wastewater once used to cool the reactor, resulting in numerous protests in neighboring countries.

Background

The Fukushima Daiichi Nuclear Power Plant consisted of six General Electric (GE) light water boiling water reactors (BWRs). Unit 1 was a GE type 3 BWR. Units 2-5 were type 4. Unit 6 was a type 5. During the 12-year construction of the power station, improvements in technology and design allowed for improvements to be made in the reactors which were constructed sequentially (beginning with unit 1, ending with unit 6).

At the time of the Tōhoku earthquake on 11 March 2011, units 1–3 were operating. However, the spent fuel pools of all units still required cooling.

Materials

Many of the internal components and fuel assembly cladding are made from a zirconium alloy (Zircaloy) for its low neutron cross-section. At normal operating temperatures (~300 °C (572 °F), it is inert. However, above 1,200 degrees Celsius (2,190 °F), Zircaloy can be oxidized by steam to form hydrogen gas or by uranium dioxide to form uranium metal. Both of these reactions are exothermic. In combination with the exothermic reaction of boron carbide with stainless steel, these reactions can contribute to the overheating of a reactor.

Isolated cooling systems

In the event of an emergency situation, reactor pressure vessels (RPV) are automatically isolated from the turbines and main condenser and are instead switched to a secondary condenser system which is designed to cool the reactor without the need for pumps powered by external power or generators. The isolation condenser (IC) system involved a closed coolant loop from the pressure vessel with a heat exchanger in a dedicated condenser tank. Steam would be forced into the heat exchanger by the reactor pressure, and the condensed coolant would be fed back into the vessel by gravity. Each reactor was initially designed to be equipped with two redundant ICs which were each capable of cooling the reactor for at least 8 hours (at which point, the condenser tank would have to be refilled). However, it was possible for the IC system to cool the reactor too rapidly shortly after shutdown which could result in undesirable thermal stress on the containment structures. To avoid this, protocol called for reactor operators to manually open and close the condenser loop using electrically operated control valves.

After the construction of unit 1, the following units were designed with new open-cycle reactor core isolation cooling (RCIC) systems. This new system utilized the steam from the reactor vessel to drive a turbine which would power a pump to inject water into the pressure vessel from an external storage tank to maintain the water level in the reactor vessel and was designed to operate for at least 4 hours (until the depletion of coolant or mechanical failure). Additionally, this system could be converted into a closed-loop system which draws coolant from the suppression chamber (SC) instead of the storage tank, should the storage tank be depleted. Although this system could function autonomously without an external energy source (besides the steam from the reactor), DC power was needed to remotely control it and receive parameters and indications and AC power was required to power the isolation valves.

In an emergency situation where backup on-site power was partially damaged or insufficient to last until a grid connection to off-site power could be restored, these cooling systems could no longer be relied upon to reliably cool the reactor. In such a case, the expected procedure was to vent both the reactor vessel and primary containment using electrically or pneumatically operated valves using the remaining electricity on site. This would lower the reactor pressure sufficiently to allow for low-pressure injection of water into the reactor using the fire protection system in order to replenish water lost to evaporation.

On-site backup power

In the event of a loss of off-site power, emergency diesel generators (EDG) would automatically start in order to provide AC power.Two EDGs were available for each of units 1–5 and three for unit 6. Of the 13 EDGs, 10 were water-cooled and placed in the basements about 7–8 m below the ground level. The coolant water for the EDGs was carried by a number of seawater pumps placed on the shoreline which also provide water for the main condenser. These components were unhoused and only protected by the seawall. The other three EDGs were air-cooled and were connected to units 2, 4, and 6. The air-cooled EDGs for units 2 and 4 were placed on the ground floor of the spent fuel building, but the switches and various other components were located below, in the basement. The third air-cooled EDG was in a separate building placed inland and at higher elevation. Although these EDGs are intended to be used with their respective reactors, switchable interconnections between unit pairs (1 and 2, 3 and 4, and 5 and 6) allowed reactors to share EDGs should the need arise.

The power station was also equipped with backup DC batteries kept charged by AC power at all times, designed to be able to power the station for approximately 8 hours without EDGs. In units 1, 2, and 4, the batteries were located in the basements alongside the EDGs. In units 3, 5, and 6, the batteries were located in the turbine building where they were raised above ground level.

In the late 1990s, three additional EDGs were placed in new buildings located inland and at higher elevation to comply with new regulatory requirements, but the switching stations that connected the EDGs to units 1–5 were located in the turbine buildings. Only the switching station for unit 6 was inside of the reactor building.

Fuel inventory

The units and central storage facility contained the following numbers of fuel assemblies:

Location	Unit 1	Unit 2	Unit 3	Unit 4	Unit 5	Unit 6	Central storage
Reactor fuel assemblies	400	548	548	0	548	764	N/A
Spent fuel assemblies	292	587	514	1331	946	876	6375
Fuel type	UO ₂	UO ₂	UO ₂/MOX	UO ₂	UO ₂	UO ₂	UO ₂
New fuel assemblies	100	28	52	204	48	64	N/A

In September 2010, Reactor 3 was partially fueled by mixed-oxides (MOX). There was no MOX (mixed oxide) fuel in any of the cooling ponds at the time of the incident.

Earthquake tolerance

The original design basis was a zero-point ground acceleration of 250 Gal and a static acceleration of 470 Gal, based on the 1952 Kern County earthquake (0.18 g, 1.4 m/s², 4.6 ft/s²). After the 1978 Miyagi earthquake, when the ground acceleration reached 0.125 g (1.22 m/s², 4.0 ft/s²) for 30 seconds, no damage to the critical parts of the reactor was found. In 2006, the design of the reactors were reevaluated with new standards (which included vertical acceleration and differentiated E/W and N/S motion) which found the reactors would withstand accelerations ranging from 412 Gal to 489 Gal.

Accident

Earthquake

The 9.0 M_W earthquake occurred at 14:46 on Friday, 11 March 2011, with the epicenter off of the east coast of the Tōhoku region. It produced maximum ground g-forces of 0.56, 0.52, 0.56 at units 2, 3, and 5 respectively. This exceeded the seismic reactor design tolerances of 0.45, 0.45, and 0.46 g for continued operation, but the seismic values were within the design tolerances at units 1, 4, and 6.

Upon detecting the earthquake, all three operating reactors (units 1, 2, and 3) automatically shut down. Due to expected grid failure and damage to the switch station as a result of the earthquake, the power station automatically started up the EDGs, isolated the reactor from the primary coolant loops, and activated the emergency shutdown cooling systems.

Tsunami and loss of power

The largest tsunami wave was 13–14 m (43–46 feet) high and hit approximately 50 minutes after the initial earthquake, overtopping the seawall and exceeding the plant's ground level, which was 10 m (33 ft) above the sea level.

The waves first damaged the seawater pumps along the shoreline, disabling the 10 water cooled EDGs. The waves then flooded all turbine and reactor buildings, damaging EDGs and other electrical components and connections located on the ground or basement levels at approximately 15:41. The switching stations that provided power from the three EDGs located higher on the hillside also failed when the building that housed them flooded. One air cooled EDG, that of unit 6, was unaffected by the flooding and continued to operate. The DC batteries for units 1, 2, and 4 were also inoperable shortly after flooding.

As a result, units 1–5 lost AC power and DC power was lost in units 1, 2, and 4. In response, the operators assumed a loss of coolant in units 1 and 2, developing a plan in which they would vent the primary containment and inject water into the reactor vessels with firefighting equipment. TEPCO notified authorities of a "first-level emergency".

Two workers were killed by the impact of the tsunami.

The dry cask storage building was also flooded, causing some concerns about possible damage.

Reactors

Unit 1

The isolation condenser (IC) was functioning prior to the tsunami, but the DC-operated control valve outside of the primary containment had been in the closed position at the time to prevent thermal stresses on the reactor components. This status was uncertain at the time due to a loss of indications in the control room, who had correctly assumed loss of coolant (LOC). 3 hours later, the plant operators attempted to manually open the control valve, but the IC failed to function, suggesting that the isolation valves were closed. Although they were kept open during IC operation, the loss of DC power in unit 1 (which occurred shortly prior to the loss of AC power) automatically closed the AC-powered isolation valves in order to prevent uncontrolled cooling or a potential LOC. Although this status was unknown to the plant operators, they correctly interpreted the loss of function in the IC system and manually closed the control valves. The plant operators would continue to periodically attempt to restart the IC in the following hours and days, but it did not function.

The plant operators then attempted to utilize the building's fire protection (FP) equipment, operated by a diesel-driven fire pump (DDFP), in order to inject water into the reactor vessel. A team was dispatched to the reactor building (RB) in order to carry out this task, but the team found that the reactor pressure had already increased significantly to 7 MPa, which was many times greater than the limit of the DDFP which could only operate below 0.8 MPa. Additionally, the team detected high levels of radiation within the RB, indicating damage to the reactor core, and found that the primary containment vessel (PCV) pressure (0.6 MPa) exceeded design specifications (0.528 MPa). In response to this new information, the reactor operators began planning to lower the PCV pressure by venting. The PCV reached its maximum pressure of 0.84 MPa at 02:30, after which it stabilized around 0.8 MPa. Venting of the PCV was completed later that afternoon at 14:00.

At the same time, pressure in the reactor vessel had been decreasing to equalize with the PCV, and the workers prepared to inject water into the reactor vessel using the DDFP once the pressure had decreased below the 0.8 MPa limit. Unfortunately, the DDFP was found to be inoperable and a fire truck had to be hooked up to the FP system. This process took about 4 hours, as the FP injection port was hidden under debris. The next morning (March 12, 04:00), approximately 12 hours after loss of power, freshwater injection into the reactor vessel began, later replaced by a water line at 09:15 leading directly from the water storage tank to the injection port to allow for continuous operation (the fire engine had to be periodically refilled). This continued into the afternoon until the freshwater tank was nearly depleted. In response, injection stopped at 14:53 and the injection of seawater, which had collected in a nearby valve pit (the only other source of water), began.

Power was restored to unit 1 (and 2) using a mobile generator at 15:30 on March 12.

At 15:36, a hydrogen explosion damaged the secondary confinement structure (the RB). The cause was unknown to the workers at the time, most of whom evacuated shortly after the explosion. The debris produced by the explosion damaged the mobile emergency power generator and the seawater injection lines. The seawater injection lines were repaired and put back into operation at 19:04 until the valve pit was nearly depleted of seawater at 01:10 on the 14th. The seawater injection was temporarily stopped in order to refill the valve pit with seawater using a variety of emergency service and JSDF vehicles. However, the process to restart seawater injection was interrupted by another explosion in the unit 3 RB at 11:01 which damaged water lines and prompted another evacuation. Injection of seawater into unit 1 would not resume until that evening, after 18 hours without cooling.

Subsequent analysis in November suggested that this extended period without cooling resulted in the melting of the fuel in unit 1, most of which would have escaped the reactor pressure vessel (RPV) and embedded itself into the concrete at the base of the PCV. Although at the time it was difficult to determine how far the fuel had eroded and diffused into the concrete, it was estimated that the fuel remains within the PCV.

In November 2013, Mari Yamaguchi reported for Associated Press that there are computer simulations that suggest that "the melted fuel in Unit 1, whose core damage was the most extensive, has breached the bottom of the primary containment vessel and even partially eaten into its concrete foundation, coming within about 30 cm (1 ft) of leaking into the ground" – a Kyoto University nuclear engineer said with regard to these estimates: "We just can't be sure until we actually see the inside of the reactors."

TEPCO estimated for Unit 1 that "the decay heat must have decreased enough, the molten fuel can be assumed to remain in PCV (primary containment vessel)".

In February 2015, TEPCO started the muon scanning process for Units 1, 2, and 3. With this scanning setup it was possible to determine the approximate amount and location of the remaining nuclear fuel within the RPV, but not the amount and resting place of the corium in the PCV. In March 2015 TEPCO released the result of the muon scan for Unit 1 which showed that no fuel was visible in the RPV, which would suggest that most if not all of the molten fuel had dropped onto the bottom of the PCV – this will change the plan for the removal of the fuel from Unit 1.

Unit 2

Unit 2 was the only other operating reactor which experienced total loss of AC and DC power. Prior to blackout, the RCIC was functioning as designed without the need for operator intervention. The safety relief valve (SRV) would intermittently release steam directly into the PCV suppression torus at its design pressure and the RCIC properly replenished lost coolant. However, following the total blackout of unit 2, the plant operators (similar to unit 1) assumed the worst-case scenario and prepared for a LOC incident. However, when a team was sent to investigate the status of the RCIC of unit 2 the following morning (02:55), they confirmed that the RCIC was operating with the PCV pressure well below design limits. Based on this information, efforts were focused onto unit 1. However, the condensate storage tank from which the RCIC draws water from was nearly depleted by the early morning, and so the RCIC was manually reconfigured at 05:00 to recirculate water from the suppression chamber instead.

On the 13th, unit 2 was configured to vent the PCV automatically (manually opening all valves, leaving only the rupture disk) and preparations were made to inject seawater from the valve pit via the FP system should the need arise. However, as a result of the explosion in unit 3 the following day, the seawater injection setup was damaged and the isolation valve for the PCV vent was found to be closed and inoperable.

At 13:00 on the 14th, the RCIC pump for unit 2 failed after 68 hours of continuous operation. With no way to vent the PCV, in response, a plan was devised to delay containment failure by venting the reactor vessel into the PCV using the SRV in order to allow for seawater injection into the reactor vessel.

The following morning (March 15, 06:15), another explosion was heard on site coinciding with a rapid drop of suppression chamber pressure to atmospheric pressure, interpreted as a malfunction of suppression chamber pressure measurement. Due to concerns about the growing radiological hazard on site, almost all workers evacuated to the Fukushima Daini Nuclear Power Plant.

In February 2017, six years after the accident, radiation levels inside the Unit 2 containment building were crudely estimated to be about 650 Sv/h. The estimation was revised later to 80 Sv/h. These readings were the highest recorded since the accident occurred in 2011 and the first recorded in that area of the reactor since the meltdowns. Images showed a hole in metal grating beneath the reactor pressure vessel, suggesting that melted nuclear fuel had escaped the vessel in that area.

In February 2017, TEPCO released images taken inside unit 2 by a remote-controlled camera that show a 2 m (6.5 ft) wide hole in the metal grating under the pressure vessel in the reactor's primary containment vessel, which could have been caused by fuel escaping the pressure vessel, indicating a meltdown/melt-through had occurred, through this layer of containment. Ionizing radiation levels of about 210 sieverts (Sv) per hour were subsequently detected inside the Unit 2 containment vessel. Undamaged spent fuel typically has values of 270 Sv/h, after ten years of cold shutdown with no shielding.

In January 2018, a remote-controlled camera confirmed that nuclear fuel debris was at the bottom of the Unit 2 PCV, showing fuel had escaped the RPV. The handle from the top of a nuclear fuel assembly was also observed, confirming that a considerable amount of the nuclear fuel had melted.

Unit 3

Although AC power was lost, some DC power was still available in unit 3 and the workers were able to remotely confirm that the RCIC system was continuing to cool the reactor. However, knowing that their DC supply was limited, the workers managed to extend the backup DC supply to about 2 days by disconnecting nonessential equipment, until replacement batteries were brought from a neighboring power station on the morning of the 13th (with 7 hours between loss and restoration of DC power). At 11:36 the next day, after 20.5 hours of operation, the RCIC system failed. In response, the high pressure coolant injection (HPCI) system was activated to alleviate the lack of cooling while workers continued to attempt to restart the RCIC. Additionally, the FP system was utilized to spray the PCV (mainly the SC) with water in order to slow the climbing temperatures and pressures of the PCV.

On the morning of the 13th (02:42), after DC power was restored by new batteries, the HPCI system showed signs of malfunction. The HPCI isolation valve failed to activate automatically upon achieving a certain pressure. In response, the workers decided to switch off HPCI and begin injection of water via the lower pressure firefighting equipment. However, the workers found that the SRV did not operate to relieve pressure from the reactor vessel in order to allow water injection by the DDFP. In response, workers attempted to restart the HPCI and RCIC systems, but both failed to restart. Following this loss of cooling, workers established a water line from the valve pit in order to inject seawater into the reactor alongside unit 2. However, water could not be injected due to RPV pressures exceeding the pump capability. Similarly, preparations were also made to vent the unit 3 PCV, but PCV pressure was not sufficient to burst the rupture disk.

Later that morning (9:08), workers were able to depressurize the reactor by operating the safety relief valve using batteries collected from nearby automobiles. This was shortly followed by the bursting of the venting line rupture disk and the depressurization of the PCV. Unfortunately, venting was quickly stopped by a pneumatic isolation valve which closed on the vent path due to a lack of compressed air, and venting was not resumed until over 6 hours later once an external air compressor could be installed. Despite this, the reactor pressure was immediately low enough to allow for water injection (borated freshwater, as ordered by TEPCO) using the FP system until the freshwater FP tanks were depleted, at which point the injected coolant was switched to seawater from the valve pit.

Cooling was lost once the valve pit was depleted, but was quickly resumed two hours later (unit 1 cooling was postponed until the valve pit was filled). However, despite being cooled, PCV pressure continued to rise and the RPV water level continued to drop until the fuel became uncovered on the morning of the 14th (6:20), as indicated by a water level gauge, which was followed by workers evacuating the area out of concerns about a possible second hydrogen explosion similar to unit 1.

Shortly after work resumed to reestablish coolant lines, an explosion occurred in the unit 3 RB at 11:01 on March 14, which further delayed unit 1 cooling and damaged unit 3's coolant lines. Work to reestablish seawater cooling directly from the ocean began two hours later, and cooling of unit 3 resumed in the afternoon (approximately 16:00) and continued until cooling was lost once more as a result of site evacuation on the 15th.

In August 2014, TEPCO released a new revised estimate that reactor 3 had a complete melt through in the initial phase of the accident. According to this new estimate within the first three days of the accident the entire core content of reactor 3 had melted through the RPV and fallen to the bottom of the PCV. These estimates were based on a simulation, which indicated that reactor 3's melted core penetrated through 1.2 m (3 ft 11 in) of the PCV's concrete base, and came close to 26–68 cm (10–27 in) of the PCV's steel wall.

Unit 4

Unit 4 was not fueled at the time, but the unit 4 spent fuel pool (SFP) contained a number of fuel rods.

On 15 March, an explosion was observed at the unit 4 RB during site evacuation. A team later returned to the power station to inspect unit 4, but were unable to do so due to the present radiological hazard. The explosion damaged the fourth floor rooftop area of Unit 4, creating two large holes in a wall of the RB. The explosion was later found to be caused by hydrogen passing to unit 4 from unit 3 through shared pipes.

The following day, on the 16th, an aerial inspection was performed by helicopter which confirmed there was sufficient water remaining in the SFP. On the 20th, water was sprayed into the uncovered SFP, later replaced by a concrete pump truck with a boom on the 22nd.

Unit 5

Unit 5 was fueled and was undergoing a RPV pressure test at the time of the accident, but the pressure was maintained by an external air compressor and the reactor was not otherwise operating. Removal of decay heat using the RCIC was not possible, as the reactor was not producing sufficient steam. However, the water within the RPV proved sufficient to cool the fuel, with the SRV venting into the PCV, until AC power was restored on March 13 using the unit 6 interconnection, allowing the use of the low-pressure pumps of the residual heat removal (RHR) system.

Cold shutdown was achieved in the afternoon on the 20th.

Unit 6

Unit 6 was not operating, and its decay heat was low because it had been in an outage since August 2010.

All but one EDG was disabled by the tsunami, allowing unit 6 to retain AC-powered safety functions throughout the incident. However, because the RHR was damaged, workers decided to activate the make-up water condensate system to maintain the reactor water level until the RHR was restored on the 20th.

Cold shutdown was achieved on the 20th, less than an hour after unit 5.

Central fuel storage areas

On 21 March, temperatures in the fuel pond had risen slightly, to 61 °C (142 °F), and water was sprayed over the pool. Power was restored to cooling systems on 24 March and by 28 March, temperatures were reported down to 35 °C (95 °F).

Evacuation

In the initial hours of the accident, in response to station blackout and uncertainty regarding the cooling status of units 1 and 2, a 2 km radius evacuation of 1,900 residents was ordered at 20:50.However, due to difficulty coordinating with the national government, a 3 km evacuation order of ~6,000 residents and a 10 km shelter-in-place order for 45,000 residents was established nearly simultaneously at 21:23. The following morning (05:44), this evacuation radius was expanded to 10 km by local authorities in response to the unit 1 core damage and plans to vent the PCV later that day. The evacuation radius was further revised at 18:25 to 20 km, involving a total of 78,000 residents, in response to the hydrogen explosion at unit 1. However, miscommunication of this final evacuation order resulted in those within 20 km to shelter in place. Additionally, many municipalities independently ordered evacuations ahead of orders from the national government due to loss of communication with authorities; at the time of the 3 km evacuation order, the majority of residents within the zone had already evacuated.

Due to the multiple overlapping evacuation orders, many residents had evacuated to areas which would shortly be designated as evacuation areas. This resulted in many residents having to move multiple times until they reached an area outside of the final 20 km evacuation zone. 20% of residents who were within the initial 2 km radius had to evacuate more than six times.

Additionally, a 30 km shelter in place order was communicated on the 15th, although some municipalities within this zone had already decided to evacuate their residents. This order was followed by a voluntary evacuation recommendation on the 25th, although the majority of residents had evacuated from the 30 km zone by then. The shelter in place order was lifted on April 22, but the evacuation recommendation remained.

Fatalities

Of an estimated 2,220 patients and elderly who resided within hospitals and nursing homes within the 20 km evacuation zone, 51 fatalities are attributed to the evacuation. There was one suspected death due to radiation, as one person died 4 years later of a lung cancer possibly triggered by it. According to one estimation, more than 2,200 deaths are to be attributed to evacuation-related stress, the vast majority of whom were over the age of 65.

Radionuclide release

The predominant mechanism by which fission products can leave the core during core melt is through vaporization, thus only relatively volatile nuclides mix with the vaporized coolant and can be transported by the flow of gas. This gas can then exit the RPV and into the PCV through small leak paths in imperfections in the RPV, but in a situation in which the RCIC is used, this gas flows through the RCIC system and into the suppression pool, where some of the vaporized or suspended fission products are condensed or captured (scrubbed) by the SC, although some remainder (notably, radioactive noble gasses) will remain vaporized or suspended inside of the PCV. From the PCV, similar to the RPV, some small quantity inevitably leaks through small imperfections in the structure, but the predominant designed path for the escape of suspended radionuclides is through venting of the PCV where they are dispersed by the vent stack. However, if the PCV is compromised, the gas will be released directly into the secondary containment, and the potential loss of the SC function would also increase the concentration of unwanted fission products in the gas.

The fraction of releases associated to certain events is debated, as some of the detected fluctuations in the environment do not strongly correlate with events at the power station.

Comparison of radiation levels for different nuclear events

Once released into the atmosphere, those which remain in a gaseous phase will simply be diluted by the atmosphere, but some which precipitate will eventually settle on land or in the ocean. Thus, the majority (90~99%) of the radionuclides which are deposited are isotopes of iodine and caesium, with a small portion of tellurium, which are almost fully vaporized out of the core due to their low vapor pressure. The remaining fraction of deposited radionuclides are of less volatile elements such as barium, antimony, and niobium, of which less than a percent is evaporated from the fuel.

Quantities of the released material are expressed in terms of the three predominant products released: caesium-137, iodine-131, and xenon-133. Estimates for atmospheric releases range from 7–20 PBq for Cs-137, 100–400 PBq for I-131, and 6,000–12,000 PBq for Xe-133.

Approximately 40–80% of the atmospheric releases were deposited over the ocean.

In addition to atmospheric deposition, there was also a significant quantity of direct releases into groundwater (and eventually the ocean) through leaks of coolant which had been in direct contact with the fuel. Estimates for this release vary from 1 to 5.5 PBq. Although the majority had entered the ocean shortly following the accident, a significant fraction remains in the groundwater and continues to mix with coastal waters.

According to the French Institute for Radiological Protection and Nuclear Safety, the release from the accident represents the most important individual oceanic emissions of artificial radioactivity ever observed. The Fukushima coast has one of the world's strongest currents (Kuroshio Current). It transported the contaminated waters far into the Pacific Ocean, dispersing the radioactivity. As of late 2011 measurements of both the seawater and the coastal sediments suggested that the consequences for marine life would be minor. Significant pollution along the coast near the plant might persist, because of the continuing arrival of radioactive material transported to the sea by surface water crossing contaminated soil. The possible presence of other radioactive substances, such as strontium-90 or plutonium, has not been sufficiently studied. Recent measurements show persistent contamination of some marine species (mostly fish) caught along the Fukushima coast.

Consequences

Evacuation

In January 2015, the number of residents displaced due to the accident was around 119,000, peaking at 164,000 in June 2012. In terms of months of life lost, the loss of life would have been far smaller if all residents had done nothing at all, or were sheltered in place, instead of evacuated.

In the former Soviet Union, many patients with negligible radioactive exposure after the Chernobyl accident displayed extreme anxiety about radiation exposure. They developed many psychosomatic problems, including radiophobia along with an increase in fatalistic alcoholism. As Japanese health and radiation specialist Shunichi Yamashita noted:

We know from Chernobyl that the psychological consequences are enormous. Life expectancy of the evacuees dropped from 65 to 58 years – not because of cancer, but because of depression, alcoholism, and suicide. Relocation is not easy, the stress is very big. We must not only track those problems, but also treat them. Otherwise people will feel they are just guinea pigs in our research.

A 2012 survey by the Iitate local government obtained responses from approximately 1,743 evacuees within the evacuation zone. The survey showed that many residents are experiencing growing frustration, instability, and an inability to return to their earlier lives. Sixty percent of respondents stated that their health and the health of their families had deteriorated after evacuating, while 39.9% reported feeling more irritated compared to before the accident.

Summarizing all responses to questions related to evacuees' current family status, one-third of all surveyed families live apart from their children, while 50.1% live away from other family members (including elderly parents) with whom they lived before the disaster. The survey also showed that 34.7% of the evacuees have suffered salary cuts of 50% or more since the outbreak of the nuclear disaster. A total of 36.8% reported a lack of sleep, while 17.9% reported smoking or drinking more than before they evacuated.

Stress often manifests in physical ailments, including behavioral changes such as poor dietary choices, lack of exercise, and sleep deprivation. Survivors, including some who lost homes, villages, and family members, were found likely to face mental health and physical challenges. Much of the stress came from lack of information and from relocation.

A 2014 metareview of 48 articles indexed by PubMed, PsycINFO, and EMBASE, highlighted several psychophysical consequences among the residents in Miyagi, Iwate, Ibaraki, Tochigi and Tokyo. The resulting outcomes included depressive symptoms, anxiety, sleep disturbance, social functioning, social isolation, admission rates, suicide rates and cerebral structure changes, radiation impacting food safety, maternal anxiety and lowered maternal confidence. The rates of psychological distress among evacuated people rose fivefold compared to the Japanese average due to the experience of the accident and evacuation. An increase in childhood obesity in the area after the accident has also been attributed to recommendations that children stay indoors instead of going outside to play.

Worldwide media coverage of the incident has been described as "ten years of disinformation", with media and environmental organizations routinely conflating the casualties of the earthquake and tsunami, with casualties of the nuclear incident. The incident dominated media coverage while the victims of the natural disasters were "ignored", and a number of media reports incorrectly describing thousands of victims of tsunami as if they were victims of the "nuclear disaster".

Energy policy

Part of the Seto Hill Windfarm in Japan, one of several windfarms that continued generating without interruption after the 2011 earthquake and tsunami and the Fukushima nuclear accident

Prior to the accident, over 25% of domestic electricity generation in Japan utilized nuclear power and Japan had set a fairly ambitious GHG reduction target of 25% below 1990 levels by 2020, which involved increasing the share of nuclear power in electricity generation from 30% to 50%. However, this plan was abandoned and target was quickly revised to a 3% emissions increase by 2020 following the accident, alongside a focus on reducing dependence on nuclear power in favor of improved thermal efficiency in fossil fuel energy use and increasing the share of "renewables". The contribution of nuclear energy dropped to less than a percent following the accident and all nuclear reactors in the country were shut down by 2013. This resulted in an increase in the share of fossil fuel energy use, which had increased to ~94% by 2015 (the highest of any IEA member state, with the remaining ~6% produced by renewables, an increase from 4% in 2010). The required fossil fuel imports in 2011 resulted in a trade deficit for the first time in decades which would continue in the following decade.

In the immediate aftermath, nine prefectures served by TEPCO experienced power rationing. The government asked major companies to reduce power consumption by 15%, and some shifted their weekends to weekdays to smooth power demand. As of 2013, TEPCO and eight other Japanese power companies were paying approximately 3.6 trillion JPY (37 billion USD) more in combined imported fossil fuel costs compared to 2010 to make up for the missing power.

Elections

On 16 December 2012, Japan held its general election. The Liberal Democratic Party (LDP) had a clear victory, with Shinzō Abe as the new Prime Minister. Abe supported nuclear power, saying that leaving the plants closed was costing the country 4 trillion yen per year in higher costs. The comment came after Junichiro Koizumi, who chose Abe to succeed him as premier, made a statement to urge the government to take a stance against using nuclear power. A survey on local mayors by the Yomiuri Shimbun newspaper in 2013 found that most of them from cities hosting nuclear plants would agree to restarting the reactors, provided the government could guarantee their safety. More than 30,000 people marched on 2 June 2013, in Tokyo against restarting nuclear power plants. Marchers had gathered more than 8 million petition signatures opposing nuclear power.

Previously a proponent of building more reactors, Prime Minister Naoto Kan took an increasingly anti-nuclear stance following the accident. In May 2011, he ordered the aging Hamaoka Nuclear Power Plant closed over earthquake and tsunami concerns, and said he would freeze building plans. In July 2011, Kan said, "Japan should reduce and eventually eliminate its dependence on nuclear energy".

International Impact

In the aftermath, Germany accelerated plans to close its nuclear power reactors and decided to phase the rest out by 2022 (see also Nuclear power in Germany). Belgium and Switzerland have also changed their nuclear policies to phase-out all nuclear energy operations. Italy held a national referendum, in which 94 percent voted against the government's plan to build new nuclear power plants. In France, President Hollande announced the intention of the government to reduce nuclear usage by one third. However, the government earmarked only one power station for closure – the aging Fessenheim Nuclear Power Plant on the German border – which prompted some to question the government's commitment to Hollande's promise. Industry Minister Arnaud Montebourg is on record as saying that Fessenheim will be the only nuclear power station to close. On a visit to China in December 2014 he reassured his audience that nuclear energy was a "sector of the future" and would continue to contribute "at least 50%" of France's electricity output. Another member of Hollande's Socialist Party, the MP Christian Bataille, said that Hollande announced the nuclear curb to secure the backing of his Green coalition partners in parliament.

China suspended its nuclear development program briefly, but restarted it shortly afterwards. The initial plan had been to increase the nuclear contribution from 2 to 4 percent of electricity by 2020, with an escalating program after that. Renewable energy supplies 17 percent of China's electricity, 16% of which is hydroelectricity. China plans to triple its nuclear energy output to 2020, and triple it again between 2020 and 2030.

New nuclear projects were proceeding in some countries. KPMG reports 653 new nuclear facilities planned or proposed for completion by 2030. By 2050, China hopes to have 400–500 gigawatts of nuclear capacity – 100 times more than it has now. The Conservative Government of the United Kingdom is planning a major nuclear expansion despite some public objection. So is Russia. India is also pressing ahead with a large nuclear program, as is South Korea. Indian Vice President M Hamid Ansari said in 2012 that "nuclear energy is the only option" for expanding India's energy supplies, and Prime Minister Modi announced in 2014 that India intended to build 10 more nuclear reactors in a collaboration with Russia.

In the wake of the accident, the Senate Appropriations Committee requested the United States Department of Energy “to give priority to developing enhanced fuels and cladding for light water reactors to improve safety in the event of accidents in the reactor or spent fuel pools”. This brief has led to ongoing research and development of Accident Tolerant Fuels, which are specifically designed to withstand the loss of cooling for an extended period, increase time to failure, and increase fuel efficiency. This is accomplished by incorporating specially designed additives to standard fuel pellets and replacing or altering the fuel cladding in order to reduce corrosion, decrease wear, and reduce hydrogen generation during accident conditions. While research is still ongoing, on 4 March 2018, the Edwin I. Hatch Nuclear Power Plant near Baxley, Georgia has implemented “IronClad” and “ARMOR” (Fe-Cr-Al and coated Zr claddings, respectively) for testing.

Radiation effects in humans

Radiation exposure of those living in proximity to the accident site is expected to be below 10 mSv, over the course of a lifetime. In comparison, the dosage of background radiation received over a lifetime is 170 mSv. Very few cancers are expected as a result of accumulated radiation exposures and residents who were evacuated were exposed to so little radiation that radiation-induced health effects were likely to be below detectable levels. There is no increase in miscarriages, stillbirths or physical and mental disorders in babies born after the accident.

Outside the geographical areas most affected by radiation, even in locations within Fukushima prefecture, the predicted risks remain low, and no observable increases in cancer above natural variation in baseline rates are anticipated.
— World Health Organization, 2013

Estimated effective doses outside Japan are considered to be below (or far below) the levels regarded as very small by the international radiological protection community. The Integrated Fukushima Ocean Radionuclide Monitoring project (InFORM) failed to show any significant amount of radiation and as a result its authors received death threats from supporters of a "wave of cancer deaths across North America" theory.

Thyroid cancer

The overwhelming majority of thyroid growths are benign growths that will never cause symptoms, illness, or death, even if nothing is ever done about the growth. Autopsy studies on people who died from other causes show that more than one third of adults technically have a thyroid growth/cancer. As a precedent, in 1999 in South Korea, the introduction of advanced ultrasound thyroid examinations resulted in an explosion in the rate of benign thyroid cancers being detected and needless surgeries occurring. Despite this, the death rate from thyroid cancer has remained the same.

There is a statistically significant correlation between external radiation dose and thyroid cancer in those under the age of 18. However, this is mostly attributable to the early detection of non-symptomatic disease cases by the screening-effect. Rates of thyroid cancer in children controlled for examination frequency showed no association between the nuclear accident/radiation exposure and thyroid cancer.

As of 2020, research into the correlation between air-dose and internal-dose and thyroid cancers remains ongoing. Further research is necessary in understanding the dose-response relationship and the prevalence of incident cancers.

Thyroid cancer is one of the most survivable cancers, with an approximate 94% survival rate after first diagnosis. That rate increases to a nearly 100% survival rate if caught early. Cancer may spread to another part of the body, however, and in cases where the thyroid must be removed, the resulting hormonal deficiencies are terminal. In January 2022, six such patients who were children at the time of the accident sued TEPCO for 616 million yen after developing thyroid cancer.

Infant/fetal cancer risk

Evacuated infant girls, the most radiation-sensitive demographic, have an estimated increased lifetime risk of developing thyroid cancer of 1.25% (compared to 0.75% background risk), with the increase being slightly less for males. The risks from a number of additional radiation-induced cancers are also expected to be elevated. There is an estimated 7% higher relative risk of leukemia in males exposed as infants and a 6% higher relative risk of breast cancer in females exposed as infants. In total, an overall 1% higher lifetime risk of developing cancers of all types is predicted for infant females, with the risk slightly lower for males. The fetuses, depending on their sex, would have the same elevations in risk as the infant groups.

Linear no-threshold models (LNT)

LNT models estimate that the accident would most likely cause 130 cancer deaths.However, LNT models have large uncertainties and are not useful for estimating health effects from radiation especially when the effects of radiation on the human body are not linear, and with obvious thresholds. Producing a statistically useful estimate would require an impractically large number of patients, and LNT models have been described as "junk science". In September 2018, one cancer fatality was the subject of a financial settlement, to the family of a former nuclear station workman.

Radiation effects in non-humans

On 21 March, the first restrictions were placed on the distribution and consumption of contaminated items. However, the results of measurements of both the seawater and the coastal sediments led to the supposition that the consequences of the accident, in terms of radioactivity, would be minor for marine life as of autumn 2011. Despite caesium isotopic concentrations in the waters off of Japan being 10 to 1000 times above the normal concentrations prior to the accident, radiation risks are below what is generally considered harmful to marine animals and human consumers.

As of March 2012, no cases of radiation-related ailments had been reported.

Fisheries

Organisms that filter water and fish at the top of the food chain are, over time, the most sensitive to caesium pollution. It is thus justified to maintain surveillance of marine life that is fished in the coastal waters off Fukushima. Migratory pelagic species are also highly effective and rapid transporters of pollutants throughout the ocean. Elevated levels of Cs-134 appeared in migratory species off the coast of California that were not seen prior to the accident.

In April 2014, studies confirmed the presence of radioactive tuna off the coasts of the Pacific U.S. Researchers carried out tests on 26 albacore tuna caught prior to the 2011 power plant accident and those caught after. However, the amount of radioactivity is less than that found naturally in a single banana. Cs-137 and Cs-134 have been noted in Japanese whiting in Tokyo Bay as of 2016. "Concentration of radiocesium in the Japanese whiting was one or two orders of magnitude higher than that in the sea water, and an order of magnitude lower than that in the sediment." They were still within food safety limits.

In June 2016, the political advocacy group "International Physicians for the Prevention of Nuclear War", asserted that 174,000 people have been unable to return to their homes and ecological diversity has decreased and malformations have been found in trees, birds, and mammals. Although physiological abnormalities have been reported within the vicinity of the accident zone, the scientific community has largely rejected any such findings of genetic or mutagenic damage caused by radiation, instead showing it can be attributed either to experimental error or other toxic effects.

In February 2018, Japan renewed the export of fish caught off Fukushima's nearshore zone. According to prefecture officials, no seafood had been found with radiation levels exceeding Japan safety standards since April 2015. In 2018, Thailand was the first country to receive a shipment of fresh fish from Japan's Fukushima prefecture. A group campaigning to help prevent global warming has demanded the Food and Drug Administration disclose the name of the importer of fish from Fukushima and of the Japanese restaurants in Bangkok serving it. Srisuwan Janya, chairman of the Stop Global Warming Association, said the FDA must protect the rights of consumers by ordering restaurants serving Fukushima fish to make that information available to their customers, so they could decide whether to eat it or not.

In February 2022, Japan suspended the sale of black rockfish from Fukushima after it was discovered that one fish from Soma had 180 times more radioactive Cesium-137 than legally permitted. The high levels of radioactivity led investigators to believe it had escaped from a breakwater at the accident site, despite nets intended to prevent fish from leaving the area. A total of 44 other fish from the accident site show similar levels.

Remediation and recovery

To assuage fears, the government enacted an order to decontaminate over a hundred areas where the level of additional radiation was greater than one millisievert per year. This is a much lower threshold than is necessary for protecting health. The government also sought to address the lack of education on the effects of radiation and the extent to which the average person was exposed.

In 2018, tours to visit the accident area began. In September 2020, The Great East Japan Earthquake and Nuclear Disaster Memorial Museum was opened in the town of Futaba, near the power plant. The museum exhibits items and videos about the earthquake and the nuclear accident. To attract visitors from abroad, the museum offers explanations in English, Chinese and Korean.

Tokyo Electric Power Company (TEPCO) is going to remove the remaining nuclear fuel material from the plants. TEPCO completed the removal of 1535 fuel assemblies from the Unit 4 spent fuel pool in December 2014 and 566 fuel assemblies from the Unit 3 spent fuel pool in February 2021. TEPCO plans to remove all fuel rods from the spent fuel pools of Units 1, 2, 5, and 6 by 2031 and to remove the remaining molten fuel debris from the reactor containments of Units 1, 2, and 3 by 2040 or 2050. An ongoing intensive cleanup program to both decontaminate affected areas and decommission the plant will take 30 to 40 years from the accident, plant management estimated.

Treating contaminated water

As of 2013, about 400 metric tons (390 long tons; 440 short tons) of cooling water per day was being pumped into the reactors. Another 400 metric tons (390 long tons; 440 short tons) of groundwater was seeping into the structure. Some 800 metric tons (790 long tons; 880 short tons) of water per day was removed for treatment, half of which was reused for cooling and half diverted to storage tanks. Ultimately the contaminated water, after treatment to remove radionuclides other than tritium, has to be discharged into the Pacific. TEPCO decided to create an underground ice wall to block the flow of groundwater into the reactor buildings. A $300 million 7.8 MW cooling facility freezes the ground to a depth of 30 meters. As of 2019, the contaminated water generation had been reduced to 170 metric tons (170 long tons; 190 short tons) per day.

In February 2014, NHK reported that TEPCO was reviewing its radioactivity data, after finding much higher levels of radioactivity than was reported earlier. TEPCO now says that levels of 5 MBq (0.12 millicuries) of strontium per liter (23 MBq/imp gal; 19 MBq/U.S. gal; 610 μCi/imp gal; 510 μCi/U.S. gal) were detected in groundwater collected in July 2013 and not the 900 kBq (0.02 millicuries) (4.1 MBq/imp gal; 3.4 MBq/U.S. gal; 110 μCi/imp gal; 92 μCi/U.S. gal) that were initially reported.

On 10 September 2015, floodwaters driven by Typhoon Etau prompted mass evacuations in Japan and overwhelmed the drainage pumps at the stricken power plant. A TEPCO spokesperson said that hundreds of metric tons of radioactive water entered the ocean as a result. Plastic bags filled with contaminated soil and grass were also swept away by the flood waters.

As of October 2019, 1.17 million cubic meters of contaminated water was stored in the plant area. The water is being treated by a purification system that can remove radionuclides, except tritium, to a level that Japanese regulations allow to be discharged to the sea. As of December 2019, 28% of the water had been purified to the required level, while the remaining 72% needed additional purification. However, tritium cannot be separated from the water. As of October 2019, the total amount of tritium in the water was about 856 terabecquerels, and the average tritium concentration was about 0.73 megabecquerels per liter.

A committee set up by the Japanese Government concluded that the purified water should be released to the sea or evaporated to the atmosphere. The committee calculated that discharging all the water to the sea in one year would cause a radiation dose of 0.81 microsieverts to the local people, whereas evaporation would cause 1.2 microsieverts. For comparison, Japanese people get 2100 microsieverts per year from natural radiation. IAEA considers that the dose calculation method is appropriate. Further, the IAEA recommends that a decision on the water disposal must be made urgently.

Despite the negligible doses, the Japanese committee is concerned that the water disposal may cause reputational damage to the prefecture, especially to the fishing industry and tourism.

Tanks used to store the water are expected to be filled in 2023. In July 2022, Japan's Nuclear Regulation Authority approved discharging the treated water into the sea. A US State Department spokesperson supported the decision. South Korea's foreign minister and activists from Japan and South Korea protested the announcement. In April 2023, fishers and activists held protests in front of the Japanese embassy in the Philippines in opposition to the planned release of 1.3 million tons of treated water into the Pacific Ocean. On 22 August, Japan announced that it would start releasing treated radioactive water from the tsunami-hit Fukushima nuclear plant into the Pacific Ocean in 48 hours, despite opposition from its neighbours. Japan says the water is safe, many scientists agree, and the decision comes weeks after the UN's nuclear watchdog approved the plan; but critics say more studies need to be done and the release should be halted. On 24 August, Japan begun the discharge of treated waste water into the Pacific Ocean, sparking protests in the region and retaliation from China, who said it would block all imports of seafood from Japan.

Other radioactive substances created as a byproduct of the contaminated water purification process, as well as contaminated metal from the damaged plant, have drawn recent attention as the 3,373 waste storage containers for the radioactive slurry were found to be degrading faster than expected.

Compensation and government expenses

Initial estimates of costs to Japanese taxpayers were in excess of 12 trillion yen ($100 billion). In December 2016 the government estimated decontamination, compensation, decommissioning, and radioactive waste storage costs at 21.5 trillion yen ($187 billion), nearly double the 2013 estimate. By 2021 12.1 trillion yen had already been spent, with 7 trillion yen on compensation, 3 trillion yen on decontamination, and 2 trillion yen on decommissioning and storage. Despite concerns, the government expected total costs to remain under budget.

The amount of compensation to be paid by TEPCO is expected to reach 7 trillion yen.

In March 2017, a Japanese court ruled that negligence by the Japanese government had led to the Fukushima accident by failing to use its regulatory powers to force TEPCO to take preventive measures. The Maebashi district court near Tokyo awarded ¥39 million (US$345,000) to 137 people who were forced to flee their homes following the accident. On 30 September 2020, the Sendai High Court ruled that the Japanese government and TEPCO are responsible for the accident, ordering them to pay $9.5 million in damages to residents for their lost livelihoods. In March 2022, Japan's Supreme Court rejected an appeal from TEPCO and upheld the order for it to pay damages 1.4 billion yen ($12 million) to about 3,700 people whose lives were harmed by the accident. Its decision covered three class-action lawsuits, among more than 30 filed against the utility.

On 17 June 2022, the Supreme Court acquitted the government of any wrongdoing regarding potential compensation to over 3,700 people affected by the accident.

On 13 July 2022, four former TEPCO executives were ordered to pay 13 trillion yen ($95 billion) in damages to the operator of power plant, in the civil case brought by Tepco shareholders.

Equipment, facility, and operational changes

A number of nuclear reactor safety system lessons emerged from the incident. The most obvious was that in tsunami-prone areas, a power station's sea wall must be adequately tall and robust. At the Onagawa Nuclear Power Plant, closer to the epicenter of 11 March earthquake and tsunami, the sea wall was 14 meters (46 ft) tall and successfully withstood the tsunami, preventing serious damage and radioactivity releases.

Nuclear power station operators around the world began to install passive autocatalytic recombiners ("PARs"), which do not require electricity to operate. PARs work much like the catalytic converter on the exhaust of a car to turn potentially explosive gases such as hydrogen into water. Had such devices been positioned at the top of the reactor buildings, where hydrogen gas collected, the explosions would not have occurred and the releases of radioactive isotopes would arguably have been much less.

Unpowered filtering systems on containment building vent lines, known as Filtered Containment Venting Systems (FCVS), can safely catch radioactive materials and thereby allow reactor core depressurization, with steam and hydrogen venting with minimal radioactivity emissions. Filtration using an external water tank system is the most common established system in European countries, with the water tank positioned outside the containment building. In October 2013, the owners of Kashiwazaki-Kariwa nuclear power station began installing wet filters and other safety systems, with completion anticipated in 2014.

For generation II reactors located in flood or tsunami prone areas, a 3+ day supply of back-up batteries has become an informal industry standard. Another change is to harden the location of back-up diesel generator rooms with water-tight, blast-resistant doors and heat sinks, similar to those used by nuclear submarines. The oldest operating nuclear power station in the world, Beznau, which has been operating since 1969, has a 'Notstand' hardened building designed to support all of its systems independently for 72 hours in the event of an earthquake or severe flooding. This system was built prior to Fukushima Daiichi.

Upon a station blackout, similar to the one that occurred after back-up battery supply was exhausted, many constructed Generation III reactors adopt the principle of passive nuclear safety. They take advantage of convection (hot water tends to rise) and gravity (water tends to fall) to ensure an adequate supply of cooling water to handle the decay heat, without the use of pumps.

As the crisis unfolded, the Japanese government sent a request for robots developed by the U.S. military. The robots went into the plants and took pictures to help assess the situation, but they couldn't perform the full range of tasks usually carried out by human workers. The accident illustrated that robots lacked sufficient dexterity and robustness to perform critical tasks. In response to this shortcoming, a series of competitions were hosted by DARPA to accelerate the development of humanoid robots that could supplement relief efforts. Eventually a wide variety of specially designed robots were employed (leading to a robotics boom in the region), but as of early 2016 three of them had promptly become non-functional due to the intensity of the radioactivity; one was destroyed within a day.

Criticism

Prior safety concerns

On 5 July 2012, the NAIIC found that the causes of the accident had been foreseeable, and that the plant operator (TEPCO) had failed to meet basic safety requirements such as risk assessment, preparing for containing collateral damage, and developing evacuation plans. At a meeting in Vienna three months after the accident, the IAEA faulted lax oversight by the Japanese Ministry of Economy, Trade and Industry, saying the ministry faced an inherent conflict of interest as the government agency in charge of both regulating and promoting the nuclear power industry. On 12 October 2012, TEPCO admitted that it had failed to take necessary measures for fear of inviting lawsuits or protests against its nuclear plants.

Unit 1 EDG disabled by flooding in 1991

On 30 October 1991, one of unit 1's EDGs failed as a result of a condensate coolant leak in the turbine building, as reported by former employees in December 2011. A TEPCO report in 2011 detailed that the room was flooded through a door and some holes for cables, but the power supply was not cut off by the flooding. An engineer was quoted as saying that he informed his superiors of the possibility that a tsunami could damage the generators.

In response, TEPCO installed doors to prevent water from leaking into the generator rooms. The JNSC stated that it would revise its safety guidelines and would require the installation of additional power sources.

Tsunami studies

In 1991, the U.S. Nuclear Regulatory Commission warned of a risk of losing emergency power in 1991 (NUREG-1150) and NISA referred to that report in 2004, but took no action to mitigate the risk.

In 2000, an in-house TEPCO report recommended safety measures against seawater flooding, based on the potential of a 50 foot (15 m) tsunami. TEPCO did not act due to concerns about creating anxieties over the safety of the nuclear power plant.

In 2002, the government earthquake research headquarters estimated that a tsunami up to 15.7 meters (52 ft) could hit the power station.

In 2004, the cabinet office warned that tsunamis taller than the maximum of 5.6 meters (18 ft) forecast by TEPCO and government officials were possible.

In 2008, another in-house study identified an immediate need to better protect the facility from flooding by seawater which cited the 15.7 meters (52 ft) estimate from the 2002 study.

In 2009, the Active Fault and Earthquake Research Center urged TEPCO and NISA to revise their assumptions for possible tsunami heights upwards, based on his team's findings about the 869 Sanriku earthquake, but this was not seriously considered at the time.

Communications

Many criticisms have been made which claim that the evacuation zone should have been further expanded, namely regarding the limited proliferation of data which governments of less affected areas may have acted upon. The national government only sent data from the SPEEDI network to the Fukushima prefectural government and was later criticized for delaying the communication of data to the U.S. military. Additionally, the U.S. military produced a detailed map using aircraft and provided it to the Ministry of Economy, Trade and Industry (METI) on 18 March and to the Ministry of Education, Culture, Sports, Science and Technology (MEXT) two days later, but no new evacuation plans were made a week after the accident. The data was not sent to the prime minister or the Nuclear Safety Commission, for which the government was criticized, but was made accessible to the public on the 23rd.

Record-keeping

The Japanese government did not keep records of key meetings during the crisis. Emails from NISA to the Fukushima prefectural government, including evacuation and health advisories from 12 March 11:54 PM to 16 March 9 AM, went unread and were deleted.

On 14 March 2011 TEPCO officials were instructed not to use the phrase "core meltdown" at press conferences.

International reaction

U.S. Navy humanitarian flight undergoes radioactive decontamination.

The international reaction to the accident was diverse and widespread. Many inter-governmental agencies immediately offered help, often on an ad hoc basis. Responders included IAEA, World Meteorological Organization and the Preparatory Commission for the Comprehensive Nuclear Test Ban Treaty Organization.

In May 2011, UK chief inspector of nuclear installations Mike Weightman traveled to Japan as the lead of an International Atomic Energy Agency (IAEA) expert mission. The main finding of this mission, as reported to the IAEA ministerial conference that month, was that risks associated with tsunamis in several sites in Japan had been underestimated.

In September 2011, IAEA Director General Yukiya Amano said the Japanese nuclear disaster "caused deep public anxiety throughout the world and damaged confidence in nuclear power". Following the accident, it was reported in The Economist that the IAEA halved its estimate of additional nuclear generating capacity to be built by 2035.

Investigations

TEPCO released estimates of the state and location of the fuel in a November 2011 report. The report concluded that the Unit 1 RPV was damaged during the accident and that "significant amounts" of molten fuel had fallen into the bottom of the PCV. The erosion of the concrete of the PCV by the molten fuel after the core meltdown was estimated to stop at approx. 0.7 m (2 ft 4 in) in depth, while the thickness of the containment floor is 7.6 m (25 ft). Gas sampling carried out before the report detected no signs of an ongoing reaction of the fuel with the concrete of the PCV and all the fuel in Unit 1 was estimated to be "well cooled down, including the fuel dropped on the bottom of the reactor". Fuel in Units 2 and 3 had melted, however less than in Unit 1. The report further suggested that "there is a range in the evaluation results" from "all fuel in the RPV (none fuel fallen to the PCV)" in Unit 2 and Unit 3, to "most fuel in the RPV (some fuel in PCV)". For Unit 2 and Unit 3 it was estimated that the "fuel is cooled sufficiently". According to the report, the greater damage in Unit 1 (when compared to the other two units) was due to the longer time that no cooling water was injected in Unit 1. This resulted in much more decay heat accumulating, as for about 1 day there was no water injection for Unit 1, while Unit 2 and Unit 3 had only a quarter of a day without water injection.

Three investigations into the accident showed the man-made nature of the catastrophe and its roots in regulatory capture associated with a "network of corruption, collusion, and nepotism." A New York Times report found that the Japanese nuclear regulatory system consistently sided with, and promoted, the nuclear industry based on the concept of amakudari ('descent from heaven'), in which senior regulators accepted high paying jobs at companies they once oversaw.

In August 2011, several top energy officials were fired by the Japanese government; affected positions included the Vice-minister for Economy, Trade and Industry; the head of the Nuclear and Industrial Safety Agency, and the head of the Agency for Natural Resources and Energy.

In 2016 three former TEPCO executives, chairman Tsunehisa Katsumata and two vice presidents, were indicted for negligence resulting in death and injury. In June 2017 the first hearing took place, in which the three pleaded not guilty to professional negligence resulting in death and injury. In September 2019 the court found all three men not guilty.

NAIIC

The Fukushima Nuclear Accident Independent Investigation Commission (NAIIC) was the first independent investigation commission by the National Diet in the 66-year history of Japan's constitutional government.

The accident "cannot be regarded as a natural disaster," the NAIIC panel's chairman, Tokyo University professor emeritus Kiyoshi Kurokawa, wrote in the inquiry report. "It was a profoundly man-made accident – that could and should have been foreseen and prevented. And its effects could have been mitigated by a more effective human response." "Governments, regulatory authorities and Tokyo Electric Power [TEPCO] lacked a sense of responsibility to protect people's lives and society," the Commission said. "They effectively betrayed the nation's right to be safe from nuclear accidents. He stated that the accident was "made in Japan", since it was a manifestation of certain cultural traits, saying:

“Its fundamental causes are to be found in the ingrained conventions of Japanese culture: our reflexive obedience; our reluctance to question authority; our devotion to ‘sticking with the program’; our groupism; and our insularity.”

The Commission recognized that the affected residents were still struggling and facing grave concerns, including the "health effects of radiation exposure, displacement, the dissolution of families, disruption of their lives and lifestyles and the contamination of vast areas of the environment".

Investigation committee

The purpose of the Investigation Committee on the Accident at the Fukushima Nuclear Power Stations (ICANPS) was to identify the accident's causes and propose policies designed to minimize the damage and prevent the recurrence of similar incidents. The 10 member, government-appointed panel included scholars, journalists, lawyers, and engineers.It was supported by public prosecutors and government experts and released its final 448-page investigation report on 23 July 2012.

The panel's report faulted an inadequate legal system for nuclear crisis management, a crisis-command disarray caused by the government and TEPCO, and possible excess meddling on the part of the Prime Minister's office in the crisis' early stage. The panel concluded that a culture of complacency about nuclear safety and poor crisis management led to the nuclear accident.

Boiling water reactor safety systems

From Wikipedia, the free encyclopedia

Boiling water reactor 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".

Safety systems

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", "torus" or "suppression pool". All BWRs utilize a number of safety/relief valves for overpressure, up to 7 of these are a part of the Automatic Depressurization System (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 with the intention that no credible scenario can cause a pressure and power increase that exceeds the systems' capability to quickly shut down 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 (3.9–7.9 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.

Reactor Protection System (RPS)

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 or 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:

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.
3. This trip is only active above approximately 1/3 reactor power. Below this amount, the bypass steam system is capable of controlling reactor pressure without causing a reactivity transient in the core.
Loss of off-site power (LOOP)
1. During normal operation, the reactor protection system (RPS) is powered by off-site power
  1. Loss of off-site 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 off-site power.
Neutron monitor trips – the purpose of these trips is to ensure an even increase in neutron and thermal power during startup.
1. Source-range monitor (SRM) or 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 or 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. Oscillation Power Range Monitor
  1. Prevents reactor power from rapidly oscillating during low flow high power conditions.
Low reactor water level:
1. Loss of coolant contingency (LOCA)
2. Loss of proper feedwater (LOFW)
3. Protects the turbine from excessive moisture carryover if water level is below the steam separator and steam dryer stack.
High water level (in BWR6 plants)
1. Prevents flooding of the main steam lines and protects turbine equipment.
2. Limits the rate of cold water addition to the vessel, thus limiting reactor power increase during over-feed transients.
High drywell (primary containment) pressure
1. Indicative of potential loss of coolant contingency
2. Also initiates ECCS systems to prepare for core injection once the injection permissives are cleared.
Main steam isolation valve closure (MSIV)
1. Protects from pressure transient in the core causing a reactivity transient
2. Only triggers for each channel when the valve is greater than 8% closed
3. One valve may be closed without initiating a reactor trip.
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.
Low RPV pressure:
1. Indicative of a line break in the steam tunnel or other location which does not trigger high drywell pressure
2. Bypassed when the reactor is not in Run mode to allow for pressurization and cooldown without an automatic scram signal
Seismic event
1. Generally only plants in high seismic areas have this trip enabled.
Scram Discharge Volume High
1. In the event that the scram hydraulic discharge volume begins to fill up, this will scram the reactor prior to the volume filling. This prevents hydraulic lock, which could prevent the control rods from inserting. This is to prevent an ATWS (Anticipated Transient Without Scram).

Emergency core-cooling system (ECCS)

Diagram of a generic BWR reactor pressure vessel

While the reactor protection system is designed to shut down the reactor, ECCS is designed to maintain adequate core cooling. The ECCS is a set of interrelated safety systems that are designed to protect the fuel within the reactor pressure vessel, which is referred to as the "reactor core", from overheating. The five criteria for ECCS are to prevent peak fuel cladding temperature from exceeding 2200 °F (1204 °C), prevent more than 17% oxidation of the fuel cladding, prevent more than 1% of the maximum theoretical hydrogen generation due the zircalloy metal-water reaction, maintain a coolable geometry, and allow for long-term cooling. ECCS systems accomplish this by maintaining reactor pressure vessel (RPV) cooling water level, or if that is impossible, by directly flooding the core with coolant.

These systems are of three major types:

High-pressure systems: These are designed to protect the core by injecting large quantities of water into it to prevent the fuel from being uncovered by a decreasing water level. Generally used in cases with stuck-open safety valves, small breaks of auxiliary pipes, and particularly violent transients caused by turbine trip and main steam isolation valve closure. If the water level cannot be maintained with high-pressure systems alone (the water level still is falling below a preset point with the high-pressure systems working full-bore), the next set of systems responds.
Depressurization systems: These systems are designed to maintain reactor pressure within safety limits. Additionally, if reactor water level cannot be maintained with high-pressure coolant systems alone, the depressurization system can reduce reactor pressure to a level at which the low-pressure coolant systems can function.
Low-pressure systems: These systems are designed to function after the depressurization systems function. They have large capacities compared to the high-pressure systems and are supplied by multiple, redundant power sources. They will maintain any maintainable water level, and, in the event of a large pipe break of the worst type below the core that leads to temporary fuel rod "uncovery", to rapidly mitigate that state prior to the fuel heating to the point where core damage could occur.

High-pressure coolant injection system (HPCI)

The high-pressure coolant injection system is the first line of defense in the emergency core cooling system. HPCI is designed to inject substantial quantities of water into the reactor while it is at high pressure so as to prevent the activation of the automatic depressurization, core spray, and low-pressure coolant injection systems. HPCI is powered by steam from the reactor, and takes approximately 10 seconds to spin up from an initiating signal, and can deliver approximately 19,000 L/min (5,000 US gal/min) to the core at any core pressure above 6.8 atm (690 kPa, 100 psi). This is usually enough to keep water levels sufficient to avoid automatic depressurization except in a major contingency, such as a large break in the makeup water line. HPCI is also able to be run in "pressure control mode", where the HPCI turbine is run without pumping water to the reactor vessel. This allows HPCI to remove steam from the reactor and slowly depressurize it without the need for operating the safety or relief valves. This minimizes the number of times the relief valves need to operate, and reduces the potential for one sticking open and causing a small LOCA.

Versioning note: Some BWR/5s and the BWR/6 replace the steam-turbine driven HPCI pump with the AC-powered high-pressure core spray (HPCS); ABWR replaces HPCI with high-pressure core flooder (HPCF), a mode of the RCIC system, as described below. (E)SBWR does not have an equivalent system as it primarily uses passive safety cooling systems, though ESBWR does offer an alternative active high-pressure injection method using an operating mode of the Control Rod Drive System (CRDS) to supplement the passive system.

Isolation Condenser (IC)

Some reactors, including some BWR/2 and BWR/3 plants, and the (E)SBWR series of reactors, have a passive system called the Isolation Condenser. This is a heat exchanger located above containment in a pool of water open to atmosphere. When activated, decay heat boils steam, which is drawn into the heat exchanger and condensed; then it falls by weight of gravity back into the reactor. This process keeps the cooling water in the reactor, making it unnecessary to use powered feedwater pumps. The water in the open pool slowly boils off, venting clean steam to the atmosphere. This makes it unnecessary to run mechanical systems to remove heat. Periodically, the pool must be refilled, a simple task for a fire truck. The (E)SBWR reactors provide three days' supply of water in the pool. Some older reactors also have IC systems, including Fukushima Dai-ichi reactor 1, however their water pools may not be as large.

Under normal conditions, the IC system is not activated, but the top of the IC condenser is connected to the reactor's steam lines through an open valve. The IC automatically starts on low water level or high steam pressure indications. Once it starts, steam enters the IC condenser and condenses until it is filled with water. When the IC system is activated, a valve at the bottom of the IC condenser is opened which connects to a lower area of the reactor. The water falls to the reactor by gravity, allowing the condenser to fill with steam, which then condenses. This cycle runs continuously until the bottom valve is closed.

Reactor core isolation cooling system (RCIC)

The reactor core isolation cooling system is not an emergency core cooling system proper, but it is included because it fulfills an important-to-safety function which can help to cool the reactor in the event of a loss of normal heat sinking capability; or when all electrical power is lost. It has additional functionality in advanced versions of the BWR.

RCIC is an auxiliary feedwater pump meant for emergency use. It is able to inject cooling water into the reactor at high pressures. It injects approximately 2,000 L/min (600 gpm) into the reactor core. It takes less time to start than the HPCI system, approximately 30 seconds from an initiating signal. It has ample capacity to replace the cooling water boiled off by residual decay heat, and can even keep up with small leaks.

The RCIC system operates on high-pressure steam from the reactor itself, and thus is operable with no electric power other than battery power to operate the control valves. Those turn the RCIC on and off as necessary to maintain correct water levels in the reactor. (If run continuously, the RCIC would overfill the reactor and send water down its own steam supply line.) During a station blackout (where all off-site power is lost and the diesel generators fail) the RCIC system may be "black started" with no AC and manually activated. The RCIC system condenses its steam into the reactor suppression pool. The RCIC can make up this water loss, from either of two sources: a makeup water tank located outside containment, or the wetwell itself. RCIC is not designed to maintain reactor water level during a LOCA or other leak. Similar to HPCI, the RCIC turbine can be run in recirculation mode to remove steam from the reactor and help depressurize the reactor.

Versioning note: RCIC and HPCF are integrated in the ABWRs, with HPCF representing the high-capacity mode of RCIC. Older BWRs such as Fukushima Unit 1 and Dresden as well as the new (E)SBWR do not have a RCIC system, and instead have an Isolation Condenser system.

Automatic depressurization system (ADS)

The Automatic depressurization system is not a part of the cooling system proper, but is an essential adjunct to the ECCS. It is designed to activate in the event that there is either a loss of high-pressure cooling to the vessel or if the high-pressure cooling systems cannot maintain the RPV water level. ADS can be manually or automatically initiated. When ADS receives an auto-start signal when water reaches the Low-Low-Low Water Level Alarm setpoint. ADS then confirms with the Low Alarm Water Level, verifies at least 1 low-pressure cooling pump is operating, and starts a 105-second timer. When the timer expires, or when the manual ADS initiate buttons are pressed, the system rapidly releases pressure from the RPV in the form of steam through pipes that are piped to below the water level in the suppression pool (the torus/wetwell), which is designed to condense the steam released by ADS or other safety valve activation into water), bringing the reactor vessel below 32 atm (3200 kPa, 465 psi), allowing the low-pressure cooling systems (LPCS/LPCI/LPCF/GDCS) to restore reactor water level. During an ADS blowdown, the steam being removed from the reactor is sufficient to ensure adequate core cooling even if the core is uncovered. The water in the reactor will rapidly flash to steam as reactor pressure drops, carrying away the latent heat of vaporization and providing cooling for the entire reactor. Low pressure ECCS systems will re-flood the core prior to the end of the emergency blowdown, ensuring that the core retains adequate cooling during the entire event.

Low-pressure core spray system (LPCS)

The Core Spray system, or Low-Pressure Core Spray system is designed to suppress steam generated by a major contingency and to ensure adequate core cooling for a partially or fully uncovered reactor core. LPCS can deliver up to 48,000 L/min (12,500 US gal/min) of water in a deluge from the top of the core. The core spray system collapses steam voids above the core, aids in reducing reactor pressure when the fuel is uncovered, and, in the event the reactor has a break so large that water level cannot be maintained, core spray is capable of preventing fuel damage by ensuring the fuel is adequately sprayed to remove decay heat. In earlier versions of the BWR (BWR 1 or 2 plants), the LPCS system was the only ECCS, and the core could be adequately cooled by core spray even if it was completely uncovered. Starting with Dresden units 2 and 3, the core spray system was augmented by the HPCI/LPCI systems to provide for both spray cooling and core flooding as methods for ensuring adequate core cooling. For most BWR models, core spray ensures the upper 1/3rd of the core does not exceed 17% cladding oxidation or 1% hydrogen production during a LOCA when used in combination with the LPCI system.

Versioning note: In ABWRs and (E)SBWRs, there are additional water spray systems to cool the drywell and the suppression pool.

Low-pressure coolant injection (LPCI)

Low-pressure coolant injection is the emergency injection mode of the Residual Heat Removal (RHR) system. LPCI can be operated at reactor vessel pressures below 375 psi. LPCI consists of several pumps which are capable of injecting up to 150,000 L/min (40,000 US gal/min) of water into the reactor. Combined with the Core Spray system, the LPCI is designed to rapidly flood the reactor with coolant. The LPCI system was first introduced with Dresden units 2 and 3. The LPCI system can also use the RHR heat exchangers to remove decay heat from the reactor and cool the containment to cold conditions. Early versions of the LPCI system injected through the recirculation loops or into the down comer. Later versions of the BWR moved the injection point directly inside the core shroud to minimize time to reflood the core, substantially reducing the peak temperatures of the reactor during a LOCA.

Versioning note: ABWRs replace LPCI with low-pressure core flooder (LPCF), which operates using similar principles. (E)SBWRs replace LPCI with the DPVS/PCCS/GDCS, as described below.

Depressurization valve system (DPVS) / passive containment cooling system (PCCS) / gravity-driven cooling system (GDCS)

The (E)SBWR has an additional ECCS capacity that is completely passive, quite unique, and significantly improves defense in depth. This system is activated when the water level within the RPV reaches Level 1. At this point, a countdown timer is started.

There are several large depressurization valves located near the top of the reactor pressure vessel. These constitute the DPVS. This is a capability supplemental to the ADS, which is also included on the (E)SBWR. The DPVS consists of eight of these valves, four on main steamlines that vent to the drywell when actuated and four venting directly into the wetwell.

If Level 1 is not resubmerged within 50 seconds after the countdown started, DPVS fires and rapidly vents steam contained within the reactor pressure vessel into the drywell. This will cause the water within the RPV to gain in volume (due to the drop in pressure) which will increase the water available to cool the core. In addition, depressurization reduces the saturation temperature enhancing the heat removal via phase transition. (In fact, both the ESBWR and the ABWR are designed so that even in the maximum feasible contingency, the core never loses its layer of water coolant.)

If Level 1 is still not resubmerged within 100 seconds of DPVS actuation, then the GDCS valves fire. The GDCS is a series of very large water tanks located above and to the side of the Reactor Pressure Vessel within the drywell. When these valves fire, the GDCS is directly connected to the RPV. After ~50 more seconds of depressurization, the pressure within the GDCS will equalize with that of the RPV and drywell, and the water of the GDCS will begin flowing into the RPV.

The water within the RPV will boil into steam from the decay heat, and natural convection will cause it to travel upwards into the drywell, into piping assemblies in the ceiling that will take the steam to four large heat exchangers – the Passive Containment Cooling System (PCCS) – located above the drywell – in deep pools of water. The steam will be cooled, and will condense back into liquid water. The liquid water will drain from the heat exchanger back into the GDCS pool, where it can flow back into the RPV to make up for additional water boiled by decay heat. In addition, if the GDCS lines break, the shape of the RPV and the drywell will ensure that a "lake" of liquid water forms that submerges the bottom of the RPV (and the core within).

There is sufficient water to cool the heat exchangers of the PCCS for 72 hours. At this point, all that needs to happen is for the pools that cool the PCCS heat exchangers to be refilled, which is a comparatively trivial operation, doable with a portable fire pump and hoses.

Standby liquid control system (SLCS)

The SLCS is a backup to the reactor protection system. In the event that RPS is unable to scram the reactor for any reason, the SLCS will inject a liquid boron solution into the reactor vessel to bring it to a guaranteed shutdown state prior to exceeding any containment or reactor vessel limits. The standby liquid control system is designed to deliver the equivalent of 86 gpm of 13% by weight sodium pentaborate solution into a 251-inch BWR reactor vessel. SLCS, in combination with the alternate rod insertion system, the automatic recirculation pump trip and manual operator actions to reduce water level in the core will ensure that the reactor vessel does not exceed its ASME code limits, the fuel does not suffer core damaging instabilities, and the containment does not fail due to overpressure during high power scram failure.

The SLCS consists of a tank containing borated water as a neutron absorber, protected by explosively-opened valves and redundant pumps, allowing the injection of the borated water into the reactor against any pressure within; the borated water will shut down a reactor and maintain it shut down. The SLCS can also be injected during a LOCA or a fuel cladding failure to adjust the ph of the reactor coolant that has spilled, preventing the release of some radioactive materials.

Versioning note: The SLCS is a system that is never meant to be activated unless all other measures have failed. In the BWR/1 – BWR/6, its activation could cause sufficient damage to the plant that it could make the older BWRs inoperable without a complete overhaul. With the arrival of the ABWR and (E)SBWR, operators do not have to be as reluctant about activating the SLCS, as these reactors have a reactor water cleanup system (RWCS) which is designed to remove boron – once the reactor has stabilized, the borated water within the RPV can be filtered through this system to promptly remove the soluble neutron absorbers that it contains and thus avoid damage to the internals of the plant.

Containment system

The ultimate safety system inside and outside of every BWR are the numerous levels of physical shielding that both protect the reactor from the outside world and protect the outside world from the reactor.

There are five levels of shielding:

The fuel rods inside the reactor pressure vessel are coated in thick Zircaloy shielding;
The reactor pressure vessel itself is manufactured out of 6-inch-thick (150 mm) steel, with extremely high temperature, vibration, and corrosion resistant surgical stainless steel grade 316L plate on both the inside and outside;
The primary containment structure is made of steel 1 inch thick;
The secondary containment structure is made of steel-reinforced, pre-stressed concrete 1.2–2.4 meters (3.9–7.9 ft) thick.
The reactor building (the shield wall/missile shield) is also made of steel-reinforced, pre-stressed concrete 0.3 to 1 m (0.98 to 3.28 ft) thick.

If every possible measure standing between safe operation and core damage fails, the containment can be sealed indefinitely, and it will prevent any substantial release of radiation to the environment from occurring in nearly any circumstance.

Varieties of BWR containments

As illustrated by the descriptions of the systems above, BWRs are quite divergent in design from PWRs. Unlike the PWR, which has generally followed a very predictable external containment design (the stereotypical dome atop a cylinder), BWR containments are varied in external form but their internal distinctiveness is extremely striking in comparison to the PWR. There are five major varieties of BWR containments:

The "premodern" containment (Generation I); spherical in shape, and featuring a steam drum separator, or an out-of-RPV steam separator, and a heat exchanger for low-pressure steam, this containment is now obsolete, and is not used by any operative reactor.

the Mark I containment, consisting of a rectangular steel-reinforced concrete building, along with an additional layer of steel-reinforced concrete surrounding the steel-lined cylindrical drywell and the steel-lined pressure suppression torus below. The Mark I was the earliest type of containment in wide use, and many reactors with Mark Is are still in service today. There have been numerous safety upgrades made over the years to this type of containment, especially to provide for orderly reduction of containment load caused by pressure in a compounded limiting fault. The reactor building of the Mark I generally is in the form of a large rectangular structure of reinforced concrete.

the Mark II containment, similar to the Mark I, but omitting a distinct pressure suppression torus in favor of a cylindrical wetwell below the non-reactor cavity section of the drywell. Both the wetwell and the drywell have a primary containment structure of steel as in the Mark I, as well as the Mark I's layers of steel-reinforced concrete composing the secondary containment between the outer primary containment structure and the outer wall of the reactor building proper. The reactor building of the Mark II generally is in the form of a flat-topped cylinder.
the Mark III containment, generally similar in external shape to the stereotypical PWR, and with some similarities on the inside, at least on a superficial level. For example, rather than having a slab of concrete that staff could walk upon while the reactor was not being refueled covering the top of the primary containment and the RPV directly underneath, the Mark III takes the BWR in a more PWR-like direction by placing a water pool over this slab. Additional changes include abstracting the wetwell into a pressure-suppression pool with a weir wall separating it from the drywell.

Advanced containments; the present models of BWR containments for the ABWR and the ESBWR are harkbacks to the classical Mark I/II style of being quite distinct from the PWR on the outside as well as the inside, though both reactors incorporate the Mark III-ish style of having non-safety-related buildings surrounding or attached to the reactor building, rather than being overtly distinct from it. These containments are also designed to take far more stress than previous containments were, providing advanced safety. In particular, GE regards these containments as being able to withstand a direct hit by a tornado beyond Level 5 on the Old Fujita Scale with winds of 330+ miles per hour. Such a tornado has never been measured on earth. They are also designed to withstand seismic accelerations of .2 G, or nearly 2 meters per second² in any direction.

Containment Isolation System

Many valves passing in and out of the containment are required to be open to operate the facility. During an accident where radioactive material may be released, these valves must shut to prevent the release of radioactive material or the loss of reactor coolant. The containment isolation system is responsible for automatically closing these valves to prevent the release of radioactive material and is an important part of a plant's safety analysis. The isolation system is separated into groups for major system functions. Each group contains its own criteria to trigger an isolation. The isolation system is similar to reactor protection system in that it consists of multiple channels, it is classified as safety-related, and that it requires confirmatory signals from multiple channels to issue an isolation to a system. An example of parameters which are monitored by the isolation system include containment pressure, acoustic or thermal leak detection, differential flow, high steam or coolant flow, low reactor water level, or high radiation readings in the containment building or ventilation system. These isolation signals will lock out all of the valves in the group after closing them and must have all signals cleared before the lockout can be reset.

Isolation valves consist of 2 safety-related valves in series. One is an inboard valve, the other is an outboard valve. The inboard is located inside the containment, and the outboard is located just outside the containment. This provides redundancy as well as making the system immune to the single failure of any inboard or outboard valve operator or isolation signal. When an isolation signal is given to a group, both the inboard and outboard valves stroke closed. Tests of isolation logic must be performed regularly and is a part of each plant's technical specifications. The timing of these valves to stroke closed is a component of each plant's safety analysis and failure to close in the analyzed time is a reportable event.

Examples of isolation groups include the main steamlines, the reactor water cleanup system, the reactor core isolation cooling (RCIC) system, shutdown cooling, and the residual heat removal system. For pipes which inject water into the containment, two safety-related check valves are generally used in lieu of motor operated valves. These valves must be tested regularly as well to ensure they do indeed seal and prevent leakage even against high reactor pressures.

Hydrogen management

During normal plant operations and in normal operating temperatures, the hydrogen generation is not significant. When the nuclear fuel overheats, zirconium in Zircaloy cladding used in fuel rods oxidizes in reaction with steam:

Zr + 2H₂O → ZrO₂ + 2H₂

When mixed with air, hydrogen is flammable, and hydrogen detonation or deflagration may damage the reactor containment. In reactor designs with small containment volumes, such as in Mark I or II containments, the preferred method for managing hydrogen is pre-inerting with inert gas—generally nitrogen—to reduce the oxygen concentration in air below that needed for hydrogen combustion, and the use of thermal recombiners. Pre-inerting is considered impractical with larger containment volumes where thermal recombiners and deliberate ignition are used. Mark III containments have hydrogen igniters and hydrogen mixers which are designed to prevent the buildup of hydrogen through either pre-ignition prior to exceeding the lower explosive limit of 4%, or through recombination with Oxygen to make water.

The safety systems in action: the Design Basis Accident

The Design Basis Accident (DBA) for a nuclear power plant is the most severe possible single accident that the designers of the plant and the regulatory authorities could reasonably expect. It is, also, by definition, the accident the safety systems of the reactor are designed to respond to successfully, even if it occurs when the reactor is in its most vulnerable state. The DBA for the BWR consists of the total rupture of a large coolant pipe in the location that is considered to place the reactor in the most danger of harm—specifically, for older BWRs (BWR/1-BWR/6), the DBA consists of a "guillotine break" in the coolant loop of one of the recirculation jet pumps, which is substantially below the core waterline (LBLOCA, large break loss of coolant accident) combined with loss of feedwater to make up for the water boiled in the reactor (LOFW, loss of proper feedwater), combined with a simultaneous collapse of the regional power grid, resulting in a loss of power to certain reactor emergency systems (LOOP, loss of offsite power). The BWR is designed to shrug this accident off without core damage.

The description of this accident is applicable for the BWR/4.

The immediate result of such a break (call it time T+0) would be a pressurized stream of water well above the boiling point shooting out of the broken pipe into the drywell, which is at atmospheric pressure. As this water stream flashes into steam, due to the decrease in pressure and that it is above the water boiling point at normal atmospheric pressure, the pressure sensors within the drywell will report a pressure increase anomaly within it to the reactor protection system at latest T+0.3. The RPS will interpret this pressure increase signal, correctly, as the sign of a break in a pipe within the drywell. As a result, the RPS immediately initiates a full SCRAM, closes the main steam isolation valve (isolating the containment building), trips the turbines, attempts to begin the spinup of RCIC and HPCI, using residual steam, and starts the diesel pumps for LPCI and CS.

Now let us assume that the power outage hits at T+0.5. The RPS is on a float uninterruptible power supply, so it continues to function; its sensors, however, are not, and thus the RPS assumes that they are all detecting emergency conditions. Within less than a second from power outage, auxiliary batteries and compressed air supplies are starting the Emergency Diesel Generators. Power will be restored by T+25 seconds.

Let us return to the reactor core. Due to the closure of the MSIV (complete by T+2), a wave of backpressure will hit the rapidly depressurizing RPV but this is immaterial, as the depressurization due to the recirculation line break is so rapid and complete that no steam voids will likely collapse to liquid water. HPCI and RCIC will fail due to loss of steam pressure in the general depressurization, but this is again immaterial, as the 2,000 L/min (600 US gal/min) flow rate of RCIC available after T+5 is insufficient to maintain the water level; nor would the 19,000 L/min (5,000 US gal/min) flow of HPCI, available at T+10, be enough to maintain the water level, if it could work without steam. At T+10, the temperature of the reactor core, at approximately 285 °C (545 °F) at and before this point, begins to rise as enough coolant has been lost from the core that voids begin to form in the coolant between the fuel rods and they begin to heat rapidly. By T+12 seconds from the accident start, fuel rod uncovery begins. At approximately T+18 areas in the rods have reached 540 °C (1,004 °F). Some relief comes at T+20 or so, as the negative temperature coefficient and the negative void coefficient slows the rate of temperature increase. T+25 sees power restored; however, LPCI and CS will not be online until T+40.

At T+40, core temperature is at 650 °C (1,202 °F) and rising steadily; CS and LPCI kick in and begins deluging the steam above the core, and then the core itself. First, a large amount of steam still trapped above and within the core has to be knocked down first, or the water will be flashed to steam prior to it hitting the rods. This happens after a few seconds, as the approximately 200,000 L/min (3,300 L/s, 52,500 US gal/min, 875 US gal/s) of water these systems release begin to cool first the top of the core, with LPCI deluging the fuel rods, and CS suppressing the generated steam until at approximately T+100 seconds, all of the fuel is now subject to deluge and the last remaining hot-spots at the bottom of the core are now being cooled. The peak temperature that was attained was 900 °C (1,650 °F) (well below the maximum of 1,200 °C (2,190 °F) established by the NRC) at the bottom of the core, which was the last hot spot to be affected by the water deluge.

The core is cooled rapidly and completely, and following cooling to a reasonable temperature, below that consistent with the generation of steam, CS is shut down and LPCI is decreased in volume to a level consistent with maintenance of a steady-state temperature among the fuel rods, which will drop over a period of days due to the decrease in fission-product decay heat within the core.

After a few days of LPCI, decay heat will have sufficiently abated to the point that defueling of the reactor is able to commence with a degree of caution. Following defueling, LPCI can be shut down. A long period of physical repairs will be necessary to repair the broken recirculation loop; overhaul the ECCS; diesel pumps; and diesel generators; drain the drywell; fully inspect all reactor systems, bring non-conformal systems up to spec, replace old and worn parts, etc. At the same time, different personnel from the licensee working hand in hand with the NRC will evaluate what the immediate cause of the break was; search for what event led to the immediate cause of the break (the root causes of the accident); and then to analyze the root causes and take corrective actions based on the root causes and immediate causes discovered. This is followed by a period to generally reflect and post-mortem the accident, discuss what procedures worked, what procedures didn't, and if it all happened again, what could have been done better, and what could be done to ensure it doesn't happen again; and to record lessons learned to propagate them to other BWR licensees. When this is accomplished, the reactor can be refueled, resume operations, and begin producing power once more.

The ABWR and ESBWR, the most recent models of the BWR, are not vulnerable to anything like this incident in the first place, as they have no liquid penetrations (pipes) lower than several feet above the waterline of the core, and thus, the reactor pressure vessel holds in water much like a deep swimming pool in the event of a feedwater line break or a steam line break. The BWR 5s and 6s have additional tolerance, deeper water levels, and much faster emergency system reaction times. Fuel rod uncovery will briefly take place, but maximum temperature will only reach 600 °C (1,112 °F), far below the NRC safety limit.

According to a report by the U.S. Nuclear Regulatory Commission into the Fukushima Daiichi nuclear disaster, the March 2011 Tōhoku earthquake and tsunami that caused that disaster was an event "far more severe than the design basis for the Fukushima Daiichi Nuclear Power Plant". The reactors at this plant were BWR 3 and BWR 4 models. Their primary containment vessels had to be flooded with seawater containing boric acid, which will preclude any resumption of operation and was not anticipated in the DBA scenario. In addition, nothing similar to the chemical explosions that occurred at the Fukushima Daiichi plant was anticipated by the DBA.

Prior to the Fukushima Daiichi disaster, no incident approaching the DBA or even a LBLOCA in severity had occurred with a BWR. There had been minor incidents involving the ECCS, but in those circumstances it had performed at or beyond expectations. The most severe incident that had previously occurred with a BWR was in 1975 due to a fire caused by extremely flammable urethane foam installed in the place of fireproofing materials at the Browns Ferry Nuclear Power Plant; for a short time, the control room's monitoring equipment was cut off from the reactor, but the reactor shut down successfully, and, as of 2009, is still producing power for the Tennessee Valley Authority, having sustained no damage to systems within the containment. The fire had nothing to do with the design of the BWR – it could have occurred in any power plant, and the lessons learned from that incident resulted in the creation of a separate backup control station, compartmentalization of the power plant into fire zones and clearly documented sets of equipment which would be available to shut down the reactor plant and maintain it in a safe condition in the event of a worst-case fire in any one fire zone. These changes were retrofitted into every existing US and most Western nuclear power plants and built into new plants from that point forth.

Notable activations of BWR safety systems

General Electric defended the design of the reactor, stating that the station blackout caused by the 2011 Tōhoku earthquake and tsunami was a "beyond-design-basis" event which led to Fukushima I nuclear accidents. According to the Nuclear Energy Institute, "Coincident long-term loss of both on-site and off-site power for an extended period of time is a beyond-design-basis event for the primary containment on any operating nuclear power plant".

The reactors shut down as designed after the earthquake. However, the tsunami disabled four of the six sets of switchgear and all but three of the diesel backup generators which operated the emergency cooling systems and pumps. Pumps were designed to circulate hot fluid from the reactor to be cooled in the wetwell, but only units 5 and 6 had any power. Units 1, 2 and 3 reactor cores overheated and melted. Radioactivity was released into the air as fuel rods were damaged due to overheating by exposure to air as water levels fell below safe levels. As an emergency measure, operators resorted to using firetrucks and salvaged car batteries to inject seawater into the drywell to cool the reactors, but only achieved intermittent success and three cores overheated. Reactors 1–3, and by some reports 4 all suffered violent hydrogen explosions March 2011 which damaged or destroyed their top levels or lower suppression level (unit 2).

As emergency measures, helicopters attempted to drop water from the ocean onto the open rooftops. Later water was sprayed from fire engines onto the roof of reactor 3. A concrete pump was used to pump water into the spent fuel pond in unit 4.

According to NISA, the accident released up to 10 petabecquerels of radioactive iodine-131 per hour in the initial days, and up to 630 PBq total, about one eighth the 5200 PBq released at Chernobyl.

Habitability of yellow dwarf systems

From Wikipedia, the free encyclopedia
https://en.wikipedia.org/wiki/Habitability_of_yellow_dwarf_systems

Habitability of yellow dwarf systems defines the suitability for life of exoplanets belonging to yellow dwarf stars. These systems are the object of study among the scientific community because they are considered the most suitable for harboring living organisms, together with those belonging to K-type stars.

Yellow dwarfs comprise the G-type stars of the main sequence, with masses between 0.9 and 1.1 M☉ and surface temperatures between 5000 and 6000 K, like the Sun. They are the third most common in the Milky Way Galaxy and the only ones in which the habitable zone coincides completely with the ultraviolet habitable zone.

Since the habitable zone is farther away in more massive and luminous stars, the separation between the main star and the inner edge of this region is greater in yellow dwarfs than in red and orange dwarfs. Therefore, planets located in this zone of G-type stars are safe from the intense stellar emissions that occur after their formation and are not as affected by the gravitational influence of their star as those belonging to smaller stellar bodies. Thus, all planets in the habitable zone of such stars exceed the tidal locking limit and their rotation is therefore not synchronized with their orbit. The Earth, orbiting a yellow dwarf, represents the only known example of planetary habitability. For this reason, the main goal in the field of exoplanetology is to find an Earth analog planet that meets its main characteristics, such as size, average temperature and location around a star similar to the Sun. However, technological limitations make it difficult to find these objects due to the infrequency of their transits, a consequence of the distance that separates them from their stars or semi-major axis.

Characteristics

Yellow dwarf stars correspond to the G-class stars of the main sequence, with a mass between 0.9 and 1.1 M☉, and surface temperatures between 5000 and 6000 K. Since the Sun itself is a yellow dwarf, of type G2V, these types of stars are also known as solar analogs. They rank third among the most common main sequence stars, after red and orange dwarfs, with a representativeness of 10% of the total Milky Way. They remain in the main sequence for approximately 10 billion years. After the Sun, the closest G-type star to the Earth is Alpha Centauri A, 4.4 light-years away and belonging to a multiple star system.

All stars go through a phase of intense activity after their formation due to their rotation, which is much faster at the beginning of their lives. The duration of this period varies according to the mass of the object: the least massive stars can remain in this state for up to 3 billion years, compared to 500 million for G-type stars.Studies by the team of Edward Guinan, an astrophysicist at Villanova University, reveal that the Sun rotated ten times faster in its early days. Since the rotation speed of a star affects its magnetic field, the Sun's X-ray and UV emissions were hundreds of times more intense than they are today.

The extension of this phase in red dwarfs, as well as the probable tidal locking of their potentially habitable planets with respect to them, could wipe out the magnetic field of these planets, resulting in the loss of almost all their atmosphere and water to space by interaction with the stellar wind. In contrast, the semi-major axis of planetary objects belonging to the habitable zone of G-type stars is wide enough to allow planetary rotation. In addition, the duration of the period of intense stellar activity is too short to eliminate a significant part of the atmosphere on planets with masses similar to or greater than that of the Earth, which have a gravity and magnetosphere capable of counteracting the effects of stellar winds.

Habitable area

The habitable zone around yellow dwarfs varies according to their size and luminosity, although the inner boundary is usually at 0.84 AU and the outer one at 1.67 in a G2V class dwarf like the Sun. In a G5V class dwarf -smaller- of 0.95 R☉ the habitable zone would correspond to the region located between 0.8 and 1.58 AU with respect to the star, while in a G0V type — larger — it would be located at a distance of between 1 and 2 AU from the stellar body. In orbits smaller than the inner boundary of the habitable zone, a process of water evaporation, hydrogen separation by photolysis and loss of hydrogen to space by hydrodynamic escape would be triggered. Beyond the outer limit of the habitable zone, temperatures would be low enough to allow CO₂ condensation, which would lead to an increase in albedo and a feedback reduction of the greenhouse effect until a permanent global glaciation would occur.

The size of the habitable zone is directly proportional to the mass and luminosity of its star, so the larger the star, the larger the habitable zone and the farther from its surface. Red dwarfs, the smallest of the main sequence, have a very small habitable zone close to them, which subjects any potentially habitable planets in the system to the effects of their star, including probable tidal locking. Even in a small yellow dwarf like Tau Ceti, of type G8.5V, the locking limit is at 0.4237 AU versus the 0.522 AU that marks the inner boundary of the habitable zone, so any planetary object orbiting a G-class star in this region will far exceed the locking limit, and will have day-night cycles like Earth.

In yellow dwarfs, this region coincides entirely with the ultraviolet habitability zone. This area is determined by an inner limit beyond which exposure to ultraviolet radiation would be too high for DNA and by an outer limit that provides the minimum levels for living things to carry out their biogenic processes. In the solar system, this region is located between 0.71 and 1.9 AU with respect to the Sun, compared to the 0.84-1.67 AU that mark the extremes of the habitable zone.

Life potential

Given the length of the main sequence in G-type stars, the levels of ultraviolet radiation in their habitable zone, the semi-major axis of the inner boundary of this region and the distance to their tidal locking limit, among other factors, yellow dwarfs are considered to be the most hospitable to life next to K-type stars.

One goal in exoplanetary research is to find an object that has the main characteristics of our planet, such as radius, mass, temperature, atmospheric composition and belonging to a star similar to the Sun. In theory, these Earth analogs should have comparable habitability conditions that would allow the proliferation of extraterrestrial life.

Based on the serious problems for planetary habitability presented by red dwarf systems and stellar bodies of type F or higher, the only stars that might offer a bearable scenario for life would be those of type K and G. Solar analogs used to be considered as the most likely candidates to host a solar-like planetary system, and as the best positioned to support carbon-based life forms and liquid water oceans. Subsequent studies, such as "Superhabitable Worlds" by René Heller and John Armstrong, establish that orange dwarfs may be more suitable for life than G-type dwarfs, and host hypothetical superhabitable planets.

However, yellow dwarfs still represent the only stellar type for which there is evidence of their suitability for life. Moreover, while in other types of stars the habitable zone does not coincide entirely with the ultraviolet habitable zone, in G-class stars the habitable zone lies entirely within the limits of the latter. Finally, yellow dwarfs have a much shorter initial phase of intense stellar activity than K-type stars, which allows planets belonging to solar analogs to preserve their primordial atmospheres more easily and to maintain them for much of the main sequence.

Discoveries

Most of the exoplanets discovered have been detected by the Kepler space telescope, which uses the transit method to find planets around other systems. This procedure analyzes the brightness of stars to detect dips that indicate the passage of a planetary object in front of them from the perspective of the observatory. It is the method that has been most successful in exoplanetary research, together with the radial velocity method, which consists of analyzing the vibrations caused in the stars by the gravitational effects of the planets orbiting them. The use of these procedures with the limitations of current telescopes makes it difficult to find objects with orbits similar to the Earth's orbits or higher, which generates a bias in favor of planets with a short semi-major axis. As a consequence, most of the exoplanets detected are either excessively hot or belong to low-mass stars, whose habitable zone is close to them and any object orbiting in this region will have a significantly shorter year than the Earth.

Planetary bodies belonging to the habitable zone of yellow dwarfs, such as Kepler-22b, Kepler-452b or Earth, take hundreds of days to complete an orbit around their star. The higher luminosity of these stars, the scarcity of transits and the semi-major axis of their planets located in the habitable zone reduce the probabilities of detecting this class of objects and considerably increase the number of false positives, as in the cases of KOI-5123.01 and KOI-5927.01. The ground-based and orbital observatories projected for the next ten years may increase the discoveries of Earth analogs in yellow dwarf systems.

Kepler-452b

Kepler-452b lies 1400 light-years from Earth, in the Cygnus constellation.^[45] Its radius of about 1.6 R_⊕ places it right on the boundary separating telluric planets from mini-Neptunes established by the team of Courtney Dressing, a researcher at the Harvard-Smithsonian Center for Astrophysics (CfA). If the planet's density is similar to Earth's, its mass will be about 5 M_⊕ and its gravity twice as great. A G2V-type yellow dwarf like the Sun belongs to Kepler-452, with an estimated age of 6 billion years (6 Ga) versus the solar system's 4.5 Ga.

The mass of its star is slightly higher than that of the Sun, 1.04 M_☉, so despite the fact that it completes an orbit around it every 385 days versus 365 terrestrial days, it is warmer than the Earth. If it has similar albedo and atmospheric composition, the average surface temperature will be around 29 °C.

According to Jon Jenkins of NASA's Ames Research Center, it is not known whether Kepler-452b is a terrestrial planet, an ocean world or a mini-Neptune. If it is an Earth-like telluric object, it is likely to have a higher concentration of clouds, intense volcanic activity, and is about to suffer an uncontrolled greenhouse effect similar to that of Venus due to the constant increase in the luminosity of its star, after having remained throughout the main sequence in its habitable zone. Doug Caldwell, a SETI Institute scientist and member of the Kepler mission, estimates that Kepler-452b may be undergoing the same process that the Earth will undergo in a billion years.

Tau Ceti e

Tau Ceti e orbits a G8.5V-type star in the constellation Cetus, 12 light-years from Earth. It has a radius of 1.59 R_⊕ and a mass of 4.29 M_⊕, so like Kepler-452b it lies at the separation boundary between terrestrial and gaseous planets. With an orbital period of only 168 days, its temperature assuming an Earth-like atmospheric composition and albedo would be about 50 °C.

The planet is located just at the inner edge of the habitable zone and receives about 60% more light than Earth. Its size may also imply a higher concentration of gases in its atmosphere, making it a super-Venus type object. Otherwise, it could be the first thermoplanet discovered.

Kepler-22b

Kepler-22b is at a distance of 600 light-years, in the Cygnus constellation. It completes one orbit around its G5V-type star every 290 days. Its radius is 2.35 R_⊕ and its estimated mass, for an Earth-like density, would be 20.36 M_⊕. If the planet's atmosphere and albedo were similar to Earth's, its surface temperature would be around 22 °C.

It was the first exoplanet found by the Kepler telescope belonging to the habitability zone of its star. Because of its size, considering the limit established by Courtney Dressing's team, its probability to be a mini-Neptune is very high.

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Friday, April 5, 2024

Fukushima nuclear accident

Background

Materials

Isolated cooling systems

On-site backup power

Fuel inventory

Earthquake tolerance

Accident

Earthquake

Tsunami and loss of power

Reactors

Unit 1

Unit 2

Unit 3

Unit 4

Unit 5

Unit 6

Central fuel storage areas

Evacuation

Fatalities

Radionuclide release

Consequences

Evacuation

Energy policy

Elections

International Impact

Radiation effects in humans

Thyroid cancer

Infant/fetal cancer risk

Linear no-threshold models (LNT)

Radiation effects in non-humans

Fisheries

Remediation and recovery

Treating contaminated water

Compensation and government expenses

Equipment, facility, and operational changes

Criticism

Prior safety concerns

Unit 1 EDG disabled by flooding in 1991

Tsunami studies

Communications

Record-keeping

International reaction

Investigations

NAIIC

Investigation committee

Boiling water reactor safety systems

Safety systems

Reactor Protection System (RPS)

Emergency core-cooling system (ECCS)

High-pressure coolant injection system (HPCI)

Isolation Condenser (IC)

Reactor core isolation cooling system (RCIC)

Automatic depressurization system (ADS)

Low-pressure core spray system (LPCS)

Low-pressure coolant injection (LPCI)

Depressurization valve system (DPVS) / passive containment cooling system (PCCS) / gravity-driven cooling system (GDCS)

Standby liquid control system (SLCS)

Containment system

Varieties of BWR containments

Containment Isolation System

Hydrogen management

The safety systems in action: the Design Basis Accident

Notable activations of BWR safety systems

Habitability of yellow dwarf systems

Characteristics

Habitable area

Life potential

Discoveries

Kepler-452b

Tau Ceti e

Kepler-22b

Introduction to entropy