Why Fukushima's Reactors Failed

Bird's eye view of Fukushima Daiichi Nuclear Power Station before the earthquake and tsunami on Fri. Mar. 11, 2011. Units 1 (right) to 4 (left) are seen in the foreground and 5 (left) and 6 (right) further back. Units 1, 2 and 3 incurred fuel meltdowns (source: TEPCO).

Three of the six reactors at Fukushima Daiichi Nuclear Power Station 120 miles north of Tokyo on Honshu's east coast incurred nuclear fuel meltdowns after the 2011 Tōhoku Earthquake and Tsunami. One reactor had been defueled for inspection. The station's two other reactors could be shutdown safely.

Bird's eye view of Fukushima Daini Nuclear Power Station before the earthquake and tsunami on Fri. Mar. 11, 2011 (unit 1 is on the right). All units could be safely shutdown. (source: TEPCO).

The four operating reactors at Fukushima Daini Nuclear Power Station seven miles away could also be saved.

The meltdowns at Daiichi led to powerful hydrogen explosions, releasing amounts of radioactive material over land and sea only rivaled by the 1986 Chernobyl reactor disaster in the Ukraine. The government imposed an exclusion zone around the stricken station, because radiation doses were deemed too high for habitation. To date, roughly 70,000 former residents cannot return home permanently. According to Shigeru Sato, Tsuyoshi Inajima, Monami Yui and Emi Urabe's report with the title "Japan Mulls Plan for One Operator to Run All Reactors: Energy" published by Bloomberg Oct. 22, 2013, cleanup and compensation are currently projected to cost 112 billion dollars, roughly equaling the cost for two Hurricane Sandies.

A multitude of inquiries have attempted to uncover the causes underlying the Fukushima accident. The operator of both power stations, Tokyo Electric Power Company, tells us what happened in its accident reports. The company's fact compilation points to one essential lynchpin: the residual heat removal system must not fail.

The reactors at Daiichi and Daini are boiling water reactors with General Electric's Mark I or Mark II primary containment systems.

Schematic of a boiling water reactor with a Mark I containment(source: IAEA).

The Mark I containments consist of a pear-shaped drywell housing the reactor pressure vessel and a doughnut-shaped wetwell, also known as torus or pressure suppression chamber, holding a large pool of water. The drywell is connected at the bottom to the wetwell with equally spaced radial vent pipes.

Schematic of a boiling water reactor with a Mark II containment(source: IAEA).

The Mark II containments are shaped differently, but still consist of a drywell with the reactor vessel and a wetwell with a pressure suppression pool.

When nuclear reactors need shutdown in an emergency, control and safety rods are inserted between the fuel rods in milliseconds, disrupting the nuclear chain reaction. Furthermore, the reactor pressure vessel's main steam lines may be isolated from the power-generating turbine and main condenser when potential damage to the pipes is anticipated. Without heat transfer, decay heat builds in the reactor pressure vessel, revving up pressure. Water must be injected into the vessel as coolant and to keep the fuel covered to prevent a meltdown.

Though emergency pumps can inject water into the vessel at high pressure, the pressure must be relieved to prevent damage to the vessel and to facilitate greater coolant injection. Safety relief valves are periodically opened to depressurize the vessel. The valves discharge high-pressure steam into the wetwell's pool, in which the steam is condensed and the pressure is absorbed. If the vessel was breached and high pressure steam bursted into the drywell, the pressure would be relieved into the suppression pool through the pipes that directly connect the drywell to the wetwell.

At Fukushima Daiichi Nuclear Power Station, the quake vibrations tripped the operating reactors automatically. Control and safety rods were inserted and the pressure vessels were isolated. Access to outside power was lost because of quake damage. Emergency diesel generators started up to provide power.

About 40 minutes later, tsunami waves flooded power distribution panels and emergency generators located low in the reactor turbine buildings. Battery banks continued to power emergency core cooling systems. Except for Unit 1 which is equipped with a isolation condenser, two reactor steam-driven pumps, the reactor core isolation cooling system and the high pressure coolant injection system, were available to inject water into the reactor pressure vessels, initially from storage tanks and later from the wetwell pools, while the operators attempted to depressurize the reactors, relieving steam into the pools.

• Eyewitness account of a Daiichi operator:
  “We went into the field in order to open the vent valves. When we were at the near the torus (that is, the wetwell, ed.) room, we heard a large, weird popping sound. The valve is at up high, so I put my foot on the torus to lift myself up. Then, my black rubber boot was melted like butter [source: TEPCO interim report, Dec 2, 2011, page 53].”
• "The decomposition of the rubber-sulphur compound (that is, vulcanized rubber, ed.) takes place to an appreciable extent at the usual temperatures of vulcanization [source: Bureau of Standards Journal of Research, Volume 4 (1930), page 512]."
• "A typical vulcanization temperature for a passenger tire is 10 minutes at 170 °C [source: Vulcanization]."

However, once the pool temperature exceeds 100 °C, absorption of reactor pressure becomes progressively less effective. The steam-driven pumps sucking water from the pool stop working above that temperature. Therefore, removing heat from the wetwell pool is absolutely necessary.

Schematic showing the residual heat removal system on the right (source: NRC).

The residual heat removal system fulfills this function. It consists of electric motor-driven pumps that flush cold service water from a large body of water, known as ultimate heat sink, through heat exchangers, cooling the wetwell pool. In Fukushima, the ocean is the ultimate heat sink, and the seawater pumps of the residual heat removal system were built near the water’s edge 4 meters above sea level.

Exposed seawater pumps at Fukushima Daiichi Nuclear Power Station 2011 (source: Daisuke Tsuda). 

At Daiichi the pumps stood exposed on the dock and the tsunami rendered them irreparable, except one that could be re-powered and jury-rigged to supply units 5 and 6.

By contrast, at Daini the pumps were housed in buildings that can be seen dockside in the areal view above. The structures were not flood resistant, but withstood the brunt of the debris-laden waves. The pump for unit 3 persevered without loss of function, and the others could be repaired before the reactor fuel was uncovered. All reactors at both power stations with functioning residual heat removal systems could be shutdown safely within days.

Therefore, electric power must be available and the residual heat removal system must remain operable for nuclear reactors of the Fukushima type to successfully complete an emergency shutdown. These conditions impose a major constraint on accident recovery, representing a fundamental weakness of the reactor design. The residual heat removal system must remain operable under the conditions of flooding and station blackout.

Thirty reactors of this type currently operate in the United States. A number are sited in flood-prone areas. A moratorium should be imposed on these reactors, until the operators can ascertain that service water pumps are protected against inundation, debris, and loss of power.

Related Posts

• Fukushima: Station Black Out & Delusion
• The Mark I Containment: Recent Insights from Japan

No Cold Shutdown at Fukushima

Steam detected emanating from the refueling floor of Fukushima Dai-ichi Nuclear Power Station Unit 3 in July 2013 (source: TEPCO).

According to Hiroko Tabuchi's post with the headline "Steam Detected at Damaged Fukushima Reactor" published by The New York Times today, the operator of the stricken Fukushima Dai-ichi Nuclear Power Station Tokyo Electric Power Company (TEPCO) noticed steam emanating from the exposed refueling floor of Unit 3, the nuclear fuel of which melted down in the wake of the 2011 Great Tōhoku Earthquake and Tsunami and which was severely damaged by a violent hydrogen explosion.

As I discussed in my essay with the title "Fukushima: Fuel Meltdowns & Cold Shutdowns" posted Feb. 15, 2012, TEPCO assumed having reached temperatures below the boiling point of water, that is 100 °C, 17 months ago, using an improvised open-loop cooling water cycle (see IAEA news by Peter Kaiser with the title "Cold Shutdown Conditions Declared at Fukushima" dated Dec. 18, 2011). Reaching reactor pressure vessel temperatures below the boiling point of water with closed cooling water cycles after the shutdown of an intact nuclear reactor is known as cold shutdown, a mode TEPCO proclaimed to have achieved at the time.

TEPCO speculates the steam detected this week is generated by rainwater seeping into the reactor (see TEPCO's press release in Japanese dated Jul. 18, 2013).

Regardless of the source, steam emanating from the primary containment vessel implies that temperatures inside the vessel must be 100 °C or higher. The discovery of new steam proves the company's assertion of cold shutdown incorrect. TEPCO is preparing countermeasures for the possibility of a nuclear chain reaction. Follow the latest developments on simplyinfo.org.

 Related Posts

• Fukushima: Fuel Meltdowns & Cold Shutdowns

Fukushima: Station Black Out & Delusion

According to a Japan Times editorial with the title "Nuclear power plant collusion" published online Jun. 23, 2012, industry and government experts appointed by the Japanese Nuclear Safety Commission in 1991 to revise safety standards for station blackouts (SBO) concluded in 1993

“that even if an SBO occurs, it would not lead to a severe accident."

Last year's nuclear reactor disaster at Fukushima Dai-ichi Nuclear Power Station (NPS) in the wake of the great Tohoku-Chihou-Taiheiyou-Oki Earthquake and Tsunami can be considered a station blackout that developed into a severe nuclear accident.

According to Cook and others (1981), “a station blackout is defined as the complete loss of all AC (alternating current, ed.) electrical power to the essential and nonessential switchgear buses in a nuclear power plant.” The authors examined a hypothetical station blackout (SBO) and its consequences, using the boiling water reactor (BWR) at Brown's Ferry NPS Unit 1 as model. The unit produces 1152 MW electrical power, and its design resembles closely that of units 1 - 4 at Fukushima Dai-ichi NPS, except Unit 1 at Fukushima was not equipped with an automatic depressurization system and had an isolation condenser instead of a reactor core isolation cooling (RCIC) system. The description of events at Fukushima Dai-ichi NPS on the day of the earthquake provided in this essay is based on the information the station's operator Tokyo Electric Power Company (TEPCO) has released since the accident.

In the study of Cook and others (1981), the loss of offsite AC power trips the main turbine, scrams the reactor, and the emergency diesel generators fail. Battery-supplied DC power continues to support emergency shutdown systems and components important to reactor safety. The authors explore with models various possible event paths after the batteries are exhausted. Though ground motion rather than a loss of power scrammed the reactors at Fukushima Dai-ichi NPS automatically, the authors' predictions on the worst accident progression converge on the chain of events at Fukushima after 4 p.m. on Mar. 11, 2011. Below, Unit 1 at Fukushima Dai-ichi NPS is used as example.

Systems, Structures, and Components Important to Safety
In conjunction with a high pressure coolant injection (HPCI) system, with which all reactors at Fukushima Dai-ichi as well as at Brown's Ferry are outfitted, RCIC system and isolation condenser constitute crucial components designed to keep the nuclear fuel in the reactor pressure vessel (RPV) covered with water, when the electrical power-generating turbine trips, the main steam isolation valves close, and the reactor scrams. HPCI and RCIC system are described in detail in the appendix of the study of Cook and others (1981).

In a scram, all safety control rods are inserted into the fuel core, instantly disrupting the nuclear chain reaction. The closing of the main steam isolation valves disconnects the reactor from the turbine, and the decay heat that the fuel produces cannot be transferred to the main steam condenser. Alternating current is no longer produced onsite. Other sources must supply power needed to operate the reactor. The offsite electrical grid represents the primary alternative source, and should the grid fail, emergency diesel generators are held in reserve onsite. If the emergency generators fail, battery-supplied direct current (DC) is supposed to support important instrumentation and valves for several hours.

Schema of the isolation condenser at Fukushima Dai-ichi NPS Unit 1, labeled here 'emergency steam condenser.' Note subsystem A and subsystem B are both fed by one coolant supply line. The coolant in the condenser is evaporated, while cooling reactor pressure vessel (RPV) steam. The coolant vapor is vented directly into the environment via a pair of exhaust pipes on the west wall of the reactor building. MO - manual-remote actuated motor operated valves (TEPCO).

The isolation condenser at Fukushima Dai-ichi Unit 1 consists of two identical subsystems of water-cooled heat exchangers designed to reduce reactor pressure vessel (RPV) temperature and pressure by steam condensation. The condensate is returned to the RPV to help keep the fuel covered. Similar to Brown's Ferry Unit 1, however, when the RPV's safety/relief valves open to protect the vessel from overpressure, large volumes of steam escape. The valves lift automatically when a pressure setpoint is reached, but can also be actuated remote-manually from the control room with battery-supplied DC power and compressed air. As a result, the water level falls.

Schema of the high pressure coolant injection (HPCI) system at Fukushima Dai-ichi NPS Unit 1. MO - motor operated, HO - hand operated, and AO - air operated valves. Note the reactor steam-driven turbine that powers pump and booster pump. The feedwater for the reactor pressure vessel (RPV) is alternatively provided by the condensate storage tank or the wetwell (suppression chamber) pool (TEPCO).

Low water level automatically starts the high pressure coolant injection (HPCI) system, a reactor steam-driven pump, which can inject coolant rapidly at high pressure into the RPV, keeping the fuel core covered.

HPCI system with reactor steam turbine-driven pump (source: nuclear tourist).
General Electric's Mark I primary containment system encloses the RPVs of units 1-5 at Fukushima Dai-ichi NPS. The system consists of a pear-shaped drywell, housing the RPV, connected with large-diameter pipes like spokes to a hub to a doughnut-shaped water-filled wetwell, also known as suppression chamber. The pipes are designed to relieve pressure building up in the drywell into the wetwell. By contrast, the safety/relief valves relieve RPV pressure into the wetwell through separate smaller diameter pipes.

Schema of the Mark I containment system inside the reactor building of Fukushima Dai-ichi Unit 1. Note the pear-shaped drywell, housing the reactor pressure vessel, connected with 81-inch pipes to the doughnut-shaped wetwell, also known as suppression chamber, filled halfway with water (courtesy: Simplyinfo.org).

The drywell is designed to operate safely up to 138 °C (Cook and others, 1981). When the fuel core in the RPV is uncovered, the core temperature will rise above 1,300 °C, the fuel rods will begin to melt, RPV penetration seals begin to leak, and ambient temperature in the drywell will heat to more than 149 °C, or 300 °F. Above this temperature, valve control solenoids, electrical cable insulation and electrical penetration module seals fail.

Hundreds of kilograms hydrogen are produced at the damaged fuel rods after fuel uncovery, when the super-heated zirconium oxide cladding of the fuel rods reacts with steam. Once the RPV is breached, the hydrogen and gaseous radioactive fission products will escape into the drywell. Once the drywell seals deteriorate in the heat, the gases will vent into the reactor and turbine building. Hence, rising radiation dose rates in these buildings suggest that a fuel meltdown is in progress.

Timeline
☢ 14:46, the earthquake strikes tripping operating units 1, 2 and 3. Offsite AC power service is lost. The emergency diesel generators start up. Two generators supply Unit 1 (TEPCO press release with the title "Release of the Fukushima Nuclear Accidents Investigation Report" dated Jun. 20, 2012).
☢15:27 and ☢15:35, tsunami waves inundate the generator rooms of units 1 to 5. The diesel generators for Unit 1 trip. In the control room wedged between the reactor and the turbine building, charge indicators flash for a brief moment. Alarms ring out and die. The tsunami also floods DC power distribution panels. In rapid succession, control room illumination, displays, indicators, gauges, annunciators and controls dim. On the ground floor below, water stands 80 cm deep.
☢15:30, the RPV pressure recorder for Unit 1 stops. Minutes later, the RPV temperature recorders halt.
☢15:37, control room crew call SBO for units 1 and 2. The government is notified. Crew arriving from the basement of the turbine building confirm that the emergency diesel generators failed and the corridor lights are out.
☢15:50, all DC power is lost. The operators are left with emergency lights on the Unit 1 side of the control room, and in total darkness on the Unit 2 side, unable to read vital reactor parameters and observe the results of their interventions. Unit 1's isolation condenser status is rendered indeterminable. The status indicator for the HPCI system is turned off. The system has fallen into a non-bootable state. The RPV water level displays cease.
☢16:15, the water level recorders for Unit 1 stop.
☢16:25, the shift supervisor declares the status of the emergency core cooling system and the RPV parameters unobtainable.
☢17:56, preparations begin for fire pumps to inject water into Unit 1's RPV.
☢20:07, crew is able to collect gauge readings in the reactor building for the first time since the SBO's inception, concluding the fuel was covered and RPV pressure held at 6.9 MPa.
☢20:47, some control room power is restored with a portable generator. Auxiliary batteries are connected to the electrical RPV water level displays, re-establishing monitoring capability at 21:19.

The observed RPV water level on the gauge indicated less than a foot above the fuel and may have been underestimated by several feet, because the reference leg of the water level gauge, also known as Yarway gauge, is located in the drywell. The reference leg is needed to account for vessel pressure and temperature. The drywell was progressively heating up, evaporating water in the gauge's reference leg and producing lower than actual water levels, when the drywell temperature exceeded design basis (Hodge and others, 1992).

Emergency Cooling
☢14:52, both subsystems of Unit 1's isolation condenser started automatically. Control room crew notes a drop in RPV pressure believed commensurate with the rapid diminution of temperature attributed to the isolation condenser activation.
☢15:03, in compliance with instructions, an operator closes valves of the isolation condenser loop. According to manual, the rate of change in RPV temperature was not permitted to be greater than 55 ℃/h because of the risk of RPV failure.

Line valves were opened to enable HPCI. The control room crew initially found no indication that Unit 1's HPCI system was inoperable. They reassured themselves that the HPCI system was ready to inject water into the RPV, should the safety/relief valves actuate automatically, that is 7.4 MPa for lifting and 6.9 MPa for reseating, resulting in a drop of RPV water level below the threshold and starting HPCI. However, because the RPV water level seemed to persist at sufficiently high levels, keeping the fuel covered, and pressure remained below safety/relief valve lift point, HPCI was not expected to start before the tsunami struck and was disabled afterwards.

The crew continued to control RPV pressure remote-manually with isolation condenser subsystem A into the next day, intermittently opening and closing the steam line valves. TEPCO believes that the isolation condenser was effective in avoiding an excessive rise in RPV pressure, because no evidence has been found that safety/relief valves lifted.

It is important to note that depressurizing the RPV with the help of the safety/relief valves, while adding sufficient coolant with the isolation condenser or the HPCI system, would have been essential for avoiding an immediate fuel meltdown. The operators actuated safety/relief valves at Units 2 and 3, but not at Unit 1.

Simulations
In their SBO simulation study on a BWR-4 boiling water reactor, Park and Ahn (2012) assume that battery power will be available for 6 hours. RCIC and HPCI systems are working, but begin to fail, once the batteries are exhausted. Under these conditions, the author's simulation suggests that, without operator intervention, core uncovery and melt will begin 8 and 9 hours into the SBO, respectively. The drywell will fail after 18 hours. TEPCO's simulation assumes that the drywell temperature beyond which leakage begins is crossed 13 hours into the SBO.

By contrast, the almost concomitant loss of AC and DC power at Fukushima accelerated the progression of events. TEPCO estimates that Unit 1's fuel core began to uncover already about 2 hours into the SBO ("Analysis and evaluation of the operation record and accident record of Fukushima Daiichi Nuclear Power Station at the time of Tohoku-Chihou-Taiheiyou-Oki-Earthquake", TEPCO, May 23, 2011).

Wilkie (2011) suggested that the fuel core in the reactor would uncover and begin to melt in slightly less than one hour, if HPCI, RCIC or the isolation condenser and safety/relief valves failed. Similarly, Cook and others (1981) predict that, without any power and without coolant injection into the RPV, fuel is uncovered in about half-an-hour, the core meltdown begins after two hours, and the drywell electrical penetration modules fail after 4.5 hours, venting radioactive noble gas, cesium, and iodine-based fission products into the reactor building (their Table 9.7.).

Furthermore, Cook and others (1981) conclude that the fuel would uncover twice as rapidly, the fuel rods would begin to melt within less than one hour, and the drywell would begin to vent about 20 minutes earlier, that is within a little more than four hours into the SBO, if a loss of coolant through a small breach in the RPV occurred, also known as small-breach LOCA (loss of coolant accident). The breach could consist of a stuck-open safety/relief valve or a hole in the reactor pressure vessel 13.8 cm in diameter.

Lochbaum (2011) suspected a LOCA might have happened, but could not find any evidence in the data TEPCO released at the time of his evaluation. In accord, Tanaka (2011) hypothesized a LOCA as part of the accident progression of Unit 1, and TEPCO's simulations match the reactor data best, when a LOCA is assumed. Below, I discuss events past 4 p.m. on March 11, 2011, that support the occurrence of a small-breach LOCA.

Small-Breach LOCA
☢17:19 crew that had been sent to Unit 1's reactor building to check on the isolation condenser bilge water level ran into hazardous radiation dose rates at the building entrance, forcing them back ("Fukushima Nuclear Accident Analysis Report (Interim Report)," TEPCO, Dec. 2, 2011).
☢23:00, contamination had reached the turbine building. A reconnaissance team recorded 1.2 mSv/h in front of the north side ground floor airlock, which is roughly 20,000 times greater than the 50 nSv/h ordinarily detected on the premises.

The progressive increase in dose rate, already observed shortly after 5:00 p.m. and corroborated at further distance three hours later, suggests that a fuel meltdown was underway already at that time, consistent with a rapid severe accident development. Without cooling, the drywell must have overheated 90 minutes into the SBO, developed leaks, and allowed substantial radioactive gases to vent into the building, leading to the detected dose rate increases. While, no evidence of a massive RPV water loss has been presented to date, a small RPV breach therefore constitutes a probable cause that would explain fuel uncovery and meltdown within less than one hour.

To date, TEPCO has not reported that safety/relief valves were actuated. However, after the loss of power a valve could have failed open unnoticed. Hydrogen produced by the fuel rod cladding/water reaction may have accumulated in the piping. Deflagrating gas could have blown open a valve.

Certainly, we shall learn one day whether a small-breach LOCA occurred at Fukushima Dai-ichi NPS Unit 1. Without doubt, the meltdown at Unit 1 progressed more rapidly than initially believed. The remarkably narrow window of opportunity for operator intervention between station blackout and fuel meltdown should give all stakeholders pause.

Acknowledgement
I thank the contributors to Simplyinfo.org. This evaluation could not have been accomplished without their input.

Addendum

• Today NHK WORLD aired a video report with the title “Nuclear Watch: Blind Spot in Nuclear Safety“, documenting that an air monitoring station 5.6 km from Fukushima Dai-ichi Nuclear Power Station registered a greater than anticipated spike in radiation on Saturday Mar. 12, 2011, after hardened venting at the station’s Unit 1 had commenced and before a hydrogen explosion damaged that unit. As the NHK WORLD report explains, hardened venting is supposed to reduce primary containment vessel pressure, filtering most radioactive contaminants out by passing the effluent through the primary containment vessel's suppression pool. TEPCO presumed that only 0.01 percent of the initial radioactivity would vent into the atmosphere. However, the radiation dose rates observed at the monitoring station were magnitudes greater than what they should have been according to TEPCO's presumption. NHK WORLD hypothesizes that the greater than expected radiation release resulted from an increase in suppression pool temperature, diminishing the pool's filtering capacity. In a model experiment, the release was increased 500-fold. NHK WORLD suggests that earlier gas and steam releases from the reactor pressure vessel heated the water in the pool via open safety relief valves. According to TEPCO’s reports to date no safety relief valves were opened at Unit 1. A LOCA releasing gas and steam from the reactor pressure vessel into the primary containment vessel represents an ever more likely possibility (04/30/2014).

References

• Cook DH, Greene SR, Harrington RM, Hodge SA, Yue DD (1981) Station Blackout at Browns Ferry Unit One - Accident Sequence Analysis. NUREG/CR-2182, Vol. 1.
• Hodge SA, Cleveland JC, Kress TS, Petck M (1992) Identification and assessment of BWR in-vessel severe accident mitigation strategies. NUREG/CR-5869.
• Lochbaum D (2011) Fukushima Dai-ichi Unit 1: The First 30 Minutes.
• Park S-Y, Ahn K-I (2012) Comparative Analysis of Station Blackout Accident Progression in Typical PWR, BWR, and PHWR. Nuclear Engineering and Technology 44:311-322.
• Tanaka M (2011) What Happened at the Fukushima Daiichi Nuclear Power Plant? In: Statement by Scientists and Engineers Concerning Fukushima Daiichi Nuclear Power Plant (no.3). Our Views of the Accidents at the Fukushima Nuclear Plants after the Earthquake.
• Wilkie D (2011) Analysis: Fukushima first hours. Simplyinfo.