Fukushima Daiichi crisis | Accident analysis

How would US units fare?

7 December 2011

The United States is home to almost half of the boiling water reactors in the world, many of them the same era as the stricken Fukushima units. This article discusses how US BWRs might have fared under severe conditions similar to those experienced in Japan. By Salomon Levy

On 11 March 2011, six boiling water reactors (BWRs) at Fukushima Daiichi site were subjected to an earthquake and particularly tsunami attacks well above their design bases. The results were incredible flooding and damage, very poor work conditions, extended station blackout (SBO), loss of normal and emergency cooling systems, reactor core melting, and radioactive releases from the operating units 1, 2, and 3.

While final evaluations of the Fukushima Daiichi crisis are still ongoing, several assessments have been published, including the US Nuclear Regulatory Commission (NRC) Near Term Task Force Review (NTTTR) and the IAEA Mission Report (IMR). They have helped establish the five important factors contributing to the Fukushima accident and its very serious consequences:

1. Earthquake and especially tsunami which caused most of the damage at Fukushima

2. SBO, its duration, and inadequate water addition to the reactor core

3. Hydrogen release and fuel melt

4. Venting of BWR suppression containments and ensuing hydrogen explosions and associated fires

5. Severe accident management.

These five contributing factors were applied to US BWRs to determine how they would fare under Fukushima US applicable severe conditions. The findings and conclusions are discussed in detail later.

In summary, US BWRs would have fared much better. Fukushima BWRs faced extremely difficult and unpredicted circumstances. They deserve considerable praise for their performance and for limiting the radiation exposure to the high population density beyond the site.

Generic BWR behaviour

In order to better understand the comments about US BWRs and the Fukushima timeline we will first discuss generic BWR behaviour during a station blackout.

Due to the absence of steam generators, BWRs use a reduced volume pressure suppression containment, which consists of a drywell and a wetwell. In the Mark I containment design used in Fukushima units 1, 2, and 3 the drywell consists of an inverted lamp-shaped steel vessel that houses the reactor vessel (see www.neimagazine.com/

markI for graphic). The lower wetwell is a steel torus half filled with water. The drywell and wetwell are connected by pipes partly submerged in the wetwell water. Both chambers are inerted to avoid burning or explosion of hydrogen formed during a loss of coolant accident.

BWRs are equipped with an isolation condenser (IC) or a reactor core isolation cooling (RCIC) system to handle decay heat during SBOs. IC relies upon reactor steam to rise and be condensed in a heat exchanger located above the core and immersed in a tank of water. The condensed water is returned by gravity to the reactor. Fukushima unit 1 had an IC capable of removing decay heat for 8 hours. By adding water to IC tank, IC could operate beyond 8 hours and handle SBOs of longer duration. RCIC uses reactor steam to drive its turbine and pump water into the reactor from the condensate storage tank or the suppression chamber of the containment where the turbine exhaust is condensed and increases the wetwell water temperature. RCIC requires control power from batteries and the batteries have limited lives. Also, the RCIC performance may degrade or fail due to increased suppression pool temperature, lack of suppression pool cooling, and presence of hydrogen. Extending battery life or having spare batteries would extend RCIC operating time; this is one area of potential improvement to severe accident management.

During extended SBOs, the wetwell water is the primary heat sink available. Using the containment cooling system to withdraw wetwell water and replacing it with cooler water or improvised ways to cool wetwell water in place or using containment spray may deserve emergency consideration. RCIC degradation would also be delayed.

The high pressure cooling injection (HPCI) provided in BWRs can also add water to BWR reactor cores. It operates like RCIC except that it can deliver large flows of water from the condensate storage tank to the core to recover it during a loss of coolant accident. It was used early at Fukushima unit 1 before the tsunami destroyed the condensate tank. The result of its use was a sharp decrease in reactor pressure and water temperature beyond the allowable rate, and operators correctly turned off the HPCI. Interestingly, the use of HPCI provided external water to the reactor and helped delay its core damage. If HPCI flow had been controlled to stay below the specified limit for water temperature decrease, it would have been able to refill the core until the water level would have tripped it. The impact could be an hour’s delay in need to add external water to the reactor. HPCI was also used as backup to RCIC at Fukushima. It was operating with wetwell water and at a point far away from its design basis and its effectiveness needs validation.

As long as IC and RCIC are working, the reactor conditions will not change noticeably. With IC, the containment pressure and temperature will remain constant while they will rise during RCIC operation. When IC or RCIC degenerate or stop working, the decay heat will continue to produce steam and the reactor vessel pressure will rise and its relief valves will open to allow steam to escape and reach the wetwell water where it is condensed. The relief valves will re-close to restore the vessel pressure to its normal level. This process will repeat itself, delivering water from reactor to wetwell until the core is uncovered and hydrogen starts forming.

In conclusion, with no external water addition, the reactor water level cannot rise and will decline continuously. Its measurement becomes less reliable partly due to its reduced level, repeated opening and closing of relief valves, and the presence of hydrogen. That possibility was not recognized at Fukushima. The reactor and containment pressures and the wetwell water temperature would be superior alternate indicators. For example, the early and repeated opening and closing of relief valves indicate degradation and failure of IC and RCIC. The reactor and the containment pressures will rise faster when hydrogen is produced. Increased reactor and containment pressure rates and wetwell temperature rises confirm accelerated core melt.

Simplified heat and water volume balances have been carried out by Dr James Healzer to determine the amount of time for the reactor core to be uncovered (for full breakdown of these calculations please see www.neimagazine.com/bwrcorecalcs). Using the integration of decay heat and the knowledge of water content with height in the reactor vessel, it was found that after four hours of RCIC operation, the wetwell temperature would rise by about 56°C (90°F). If the RCIC arbitrarily stops working at that time, the water level would drop to the top of the reactor in less than one hour, and to the core midpoint shortly thereafter. In conclusion, that simplified assessment confirms the urgency of adding water to the reactor core after IC or RCIC stop working and the value of carrying such on-site predictions with time.

The timelines provide graphic descriptions of how the lack of lighting, frequent evacuations and suspensions, as well as the late arrival of equipment delayed Fukushima management to recognize need for early addition of water to reactor cores. They are good accounts of the severe accident’s progress versus time, including times for water addition to the reactor cores, venting the containments, and the ensuing explosions.

Application to US BWRs

This section discusses the principal contributing factors to the events at Fukushima Daiichi and applies them to US BWRs, where applicable.

1. Earthquake and tsunami

According to NTTR, “current NRC regulations and associated regulatory guidance provide a robust regulatory approach for evaluation of site hazards associated with natural phenomena.” Furthermore, the NRC requires that “safety-significant structures, systems, and components be designed to take into account even rare and extreme seismic and tsunami events.” As new information becomes available, the implications of natural events are reevaluated at appropriate intervals.

No current or future US BWR will face tsunamis at the Fukushima level as discussed in NRC News Release 11-053 of 19 March 2011. According to that release, the only subduction zone similar to the Fukushima site is the Cascades region off the California, Oregon, and Washington coasts. Only the Columbia BWR is located there, but it is more than 200 miles from the coast. This would preclude the possibility of a tsunami reaching that BWR. Finally, safety margins are provided with respect to designed earthquake resistance in all BWRs. These margins were confirmed by the Fukushima plants' resistance to earthquake forces that exceeded their design basis. In other words the considerable damages sustained at Fukushima cannot happen at US BWRs. The conclusion is that US BWRs would have undergone normal shutdown and recovery under applicable Fukushima-type conditions in the USA.

2. SBO and water addition

The US NRC has rules that require US BWRs to keep the reactor core covered with water and to maintain containment integrity for a prescribed SBO duration. That time relies upon evaluating the redundancy of the on-site emergency power system and its reliability, the anticipated loss of off-site power and the time to react to it. Based upon that assessment and the anticipated availability time of IC or RCIC, an overall strategy is developed and reviewed by NRC to assure that BWRs will be maintained in a safe condition.

US SBOs are very rare and of short duration. When the SBO extends beyond the capability of provided systems to remove decay heat, then reactor depressurization, other available water sources, piping connections to BWRs, mobile power, and special equipment are used to assure continued water addition. Due to site damage, these techniques were not possible at Fukushima. Since tsunami of the degree of Fukushima have been ruled out for the USA (see first point above), in my opinion there is no apparent reason to change the current US SBO duration to an arbitrary 72 hours proposed in NTTR. What is most needed is to confirm that the preferred current strategy of depressurization and addition of water to the reactor can be carried out to preserve safety.

My recommendation is that available US SBO strategies be revisited to confirm their success by actual training at plant simulators. The use of simulator training provides the ability to turn off IC or RCIC and to recognize their failures by observing reactor pressure and wetwell temperature. It also permits operators to switch their focus from water level to those variables when the level is decreasing. Also, measures outlined above could be tried: adding water to IC, providing reserve battery power to RCIC, and cooling the wetwell water. Site probabilistic estimates of SBO occurrence and its duration may also need to be reevaluated.

3. Hydrogen release

Hydrogen is released shortly after BWR reactor water level falls below the midpoint of active fuel. The amount of hydrogen formed accelerates as more fuel is exposed to steam and as considerable energy is added from the chemical reaction of steam with Zircaloy fuel cladding.

It is well-known that without water addition, the time to core melt is short. The computer code MELCOR in NUREG/CR 5850, assuming RCIC operation for 5 hours and no depressurization to add water, predicts Zircaloy oxidation 76 minutes ahead of core melt. Melt relocation is predicted to happen two hours later.

The use of the Modular Accident Analysis Program (MAAP) at Fukushima unit 1 was reported in IMR. That prediction showed that top of active fuel was reached about three hours after plant trip and core damage started four hours after the trip. Unfortunately, that prediction, which is likely correct, was changed subsequently by relying upon improper water level data. Rising rates of reactor and containment pressures and containment water temperature would have yielded more reliable information than reactor water level even in the darkened control rooms of Fukushima.

The preceding analyses confirm the need to have on-site capability to predict hydrogen formation and to compare such predictions with plant data for containment pressure and wetwell water temperature. The conclusion is that formation of hydrogen and the acceleration in the rate of its formation need to be forecasted and detected to shift top priority to reactor water addition and to assure its success.

4. Venting and explosions

All US Mark I containments vent through hardened pipes connected to stack or to another high point outside the reactor building in order to avoid hydrogen explosions and fires within that building.

It is important to note that BWR containment venting is necessary only under the extremely low probability of not being able to add water to the reactor core. Early venting is preferred when the containment pressure and hydrogen concentration are low and not prone to explosions and fires. Additional periodic venting with addition of inert gas may be necessary until the reactor is recovered with water.

At Fukushima, although the venting pipes did terminate at the stack, a Japanese regulation prohibited venting until containment pressure was twice the design pressure. This allowed considerable hydrogen to build up in the containment and increased the chances of leakage within the reactor building and the chances of hydrogen explosions.

Since venting strategies may vary at US BWRs, formalizing them and verifying their practice during US BWR SBO simulator training is urged. The concept of early venting followed by subsequent periodic vents with adding nitrogen could be checked during the SBO training on plant simulators. Also, it would give the chance to discuss ways to add water to IC, to use containment spray, and to cool the wetwell water in order to reduce containment pressure and wetwell water temperature and improve RCIC performance.

5. Accident management

Control and command during a severe accident in the USA rests at the site. Severe Accident Management Guidelines (SAMGs) are available to develop a proper response.

At Fukushima, severe accident strategies were mostly improvised due to degraded site and reactor conditions. Fukushima had SAMGs but they did not anticipate the extensive damage caused by the tsunami. They did not account for darkened control rooms and lack of control and instrumentation information. The site flooding and debris did not help. It was difficult to get needed equipment and support. Too many tasks had to be carried out simultaneously and the operator had to get the approval of four different ministries before proceeding with water addition and venting. Overall, the Fukushima site staff can only be praised.

Although US BWRs are unlikely to be hindered by a Fukushima-size tsunami in implementing SAMGs, some benefits can be gained from Fukushima experience. First, US BWRs’ SAMGS should be reviewed to revalidate them. Consideration of ‘what if’ circumstances and lessons learned at Fukushima may be useful. Second, prioritization of potential tasks would be beneficial, with top priority given to reactor water addition. Site availability of essential batteries for RCIC and control room lighting, timely delivery of mobile power, fire, and water supply equipment deserve rechecking. The use of on-site models to predict severe accident progress and the recognition of improper water level data are essential. Finally, contingency plans to cool wetwell water could be considered. In my opinion, it is essential that full control and command remains at the sites due to urgency of taking action during SBOs.

6. Spent fuel pool cooling

Concern arose about spent fuel pool cooling at Fukushima unit 4 because of its unloading of the full reactor core during the refueling outage, which significantly increases pool decay heat. It was enhanced by detection of radioactivity leakage from unit 3. The observation of that pool when the radiation level was low would have avoided the diversion of personnel away from more important tasks. The use of helicopters to drop water was unsuccessful because little of that water reached its target and worsened the site flooding and the eventual cleanup. Potential splashing of water has also been mentioned. Splashing at US BWRs will not happen due to their reduced earthquake levels. Improvised ways to add water to US BWRs spent fuel pools need to be adequate and periodic observations of them by visits or cameras would avoid false alarms.


With respect to the Fukushima crisis it can be concluded that the tsunami was responsible for the severe accidents, reactor melts, releases of radioactivity, and the inability to reduce their consequences at the Fukushima Daiichi site.

At Fukushima, top priority should have been given to reactor water addition and less emphasis put on unreliable reactor water level data. Increased focus should have been put upon containment pressure and especially its water temperature.

In my view, there is great urgency to upgrade the Japanese tsunami model and to apply it to other subduction zones along Japan coastlines in order to reduce the risks and to protect the population in those areas. This point deserves top priority, which I feel is not emphasized enough in IMR.

With respect to US BWRs, we can conclude that a high safety level is preserved because they are not subject to Fukushima tsunamis levels. Nevertheless, review of SBO, SAMGS, and containment venting strategies are recommended for US BWRs. Other suggestions have been offered about confirming reactor water level accuracy, using early venting, emergency availability of equipment, and cooling containment water.

This article was first published in the November 2011 issue of Nuclear Engineering International

Author Info:

Dr. Salomon Levy, Honorary Member ASME, Member NAE, 3425 South Bascom Avenue, Suite 225, Campbell, CA, 95008. USA. Dr. Levy was the manager responsible for General Electric (GE) BWR heat transfer and fluid flow and the analyses and tests to support their nuclear fuel cooling during normal, transient, and accident analyses from 1959 to 1977.

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Recommendations based on timeline of events

Several comments are in order based upon the Fukushima timelines. They should not be seen as criticism of Fukushima personnel. The advice would be:

- Carry out water addition to reactor as soon as possible, as was the case at unit 1 at 17:30 on 11 March.

- Recognize unreliable water level data and avoid their use to delay urgent actions.

- Presume that early indication of radioactivity indicates start of core damage.

- Recognize RCIC deterioration with increased wetwell water temperature and consider cooling containment water.

- Consider early venting rather than waiting for containment pressure to reach or exceed design pressure.

- Realize that salt water addition is preferable to core melt progress. While salt water presence is undesirable due to stress corrosion, this phenomena can occur only at reduced pressures and temperatures, when its impact will be diminished.

There are many lessons yet to be learned from Fukushima, including the apparent poor performance of IC at unit 1, the inability to open several valves, and many others yet to be reported. It is hoped that times for reactor pressure valve opening, pressure and temperature containment values versus time could be made available to assess RCIC and HPCI performances during severe accidents.

FilesSimplified Fukushima Daiichi timelines

External weblinks
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Diagram of Mark I containment
Full breakdown of BWR core calculations by Dr. James Healzer

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