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site title: Physical Insights An independent scientists observations on society, technology, energy, science and the environment. Modern science has been a voyage into the unknown, with a lesson in humility waiting at every stop. Many passengers would rather have stayed home. Carl Sagan

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n systems for the low pressure coolant injection components of the eccs to operate the pressure within the reactor must be reduced by the depressurization system the automatic depressurization system ads is designed to activate in the event that the reactor pressure vessel is retaining pressure but the water level cannot be maintained using high pressure cooling alone and low pressure cooling must be initiated when the ads activates it rapidly releases pressure from the reactor vessel in the form of steam through pipes that are piped to below the water level in the the torus which is designed to condense the steam released into it bringing the reactor vessel pressure below 32 atmospheres allowing the low pressure components of the eccs to be activated the low pressure eccs systems have extremely large capacities compared to the high pressure systems and are powered by multiple different power sources they will maintain any required water level and in the event of a worst case loca such as a break of a large water pipe feeding into the reactor vessel below core level which could potentially lead to temporary fuel rod uncovery they will rapidly return the water level over the fuel in the core prior to the fuel heating to the point where core damage could occur the low pressure core spray system lpcs is the first of the low pressure eccs components designed to suppress steam generated by a major contingency as such it prevents the reactor vessel pressure from re increasing above the lpci coolant injection pressure 32 atmospheres it activates while the pressure in the reactor is still below 32 atmospheres and delivers approximately 48 000 l min of water in a deluge from the top of the core the low pressure coolant injection system lpci is the final piece of the eccs the heavy artillery in the eccs consisting of 4 pumps driven by diesel engines it is capable of injecting a mammoth 150 000 l min of water into the core combined with the core spray system to keep steam pressure in the core sufficiently low the lpci can suppress all core cooling contingencies by rapidly and completely flooding the core with coolant one should also note that the diesel engine driven pumps that run the lpci are completely independent of off site electrical grid power they are independent of steam power being extracted from the reactor unlike hpci and they are independent of the diesel generators that provide the backup electricity supply for the plant in the event of the loss of offsite power the standby liquid control system the slcs is used in the event of major contingencies as a last ditch measure to prevent core damage it is not intended ever to be used as the rps and eccs are designed to respond to all contingencies even if multiple components of those systems fail but if a complete eccs failure occurs it could be the only thing capable of preventing core damage the slcs consists of a tank containing a large quantity of water loaded with soluble nuclear poisons such as boron protected by explosively opened valves and redundant battery operated pumps allowing the injection of the water into the reactor against any pressure within it this water is fully capable of dissipating heat from the nuclear fuel and the nuclear poisons in this water will send the system fully subcritical even if somehow insertion of the control rods has completely failed which is not the case at present for any of these japanese nuclear power reactors the slcs is a system that is never meant to be activated unless all other measures have failed to maintain integrity of the nuclear fuel in the older generation of existing bwrs its activation could cause sufficient damage to the plant due to the salts used as neutronic poisons causing corrosion and contamination of the whole nuclear steam supply system that it could make the reactor inoperable without a complete overhaul there is now talk of pumping seawater into the reactor building although the information in the press on this subject seems to be vague and confused there is very little good unambiguous information out there are we talking about spraying seawater within the reactor building in order to condense steam and reduce the temperature and pressure that seems to make sense are we talking about spraying seawater within the drywell to help cool the reactor pressure vessel and reduce temperatures within the drywell that also makes sense surely it wouldn t make sense to actually inject seawater within the actual nuclear steam supply system would it this would cause significant problems with regards to contamination and corrosion of the entire nuclear steam supply system which would be difficult time consuming and expensive to rectify why would this ever be considered when the slcs and the eccs systems are designed to perform the same function safely and reliably under adverse emergency conditions without ruining the reactor i do not expect that this is actually what is being planned but again the information that is tricking out through the hysterical mass media is so bad it s hard to tell the emergency core cooling systems and the design basis accident i ve borrowed most of the material for this example scenario illustrated here from here the design basis accident dba for a nuclear power reactor is the most severe possible single accident that the designers of the plant and the regulatory authorities could realistically imagine as a contingency which the operators of the plant must be able to handle 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 the older generations of existing bwrs such as the fukushima bwrs 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 and as such has the makings of a very serious loss of coolant accident or loca the dba scenario combines this large scale loss of coolant with a simultaneous loss of feedwater to make up for the water boiled in the reactor a loss of proper feedwater or lofw combined with a simultaneous collapse of the regional power grid resulting in a loss of power to certain reactor emergency systems or loop the bwr is designed to shrug this accident off without core damage the design basis accident is not directly relevant to what happened to the reactor at fukushima but it is a good example to use to illustrate how the various different layers of the eccs and the reactor protection system work under severe accident conditions which is important background to a good understanding of what happened at fukushima the immediate result of such a large scale pipe break we ll call this time t 0 would be a pressurized stream of water well above boiling point shooting out of the broken pipe into the drywell which is at atmospheric pressure as this water stream flashes into steam the pressure sensors within the drywell will report a pressure increase to the reactor protection system within no more than 300 milliseconds that is by t 0 3 the rps will interpret this pressure increase signal 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 spin up rcic and hpci using the residual steam and starts the diesel driven pumps for lpci and the core spray now let s assume that the loop occurs at this time at t 0 5 the rps is on an uninterruptable power supply so it continues to function the rps immediately detects the loss of offsite power however and already enters a fully defensive state and trips the reactor and the turbine if it has not already 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 remember that at fukushima i the backup diesel generator failed shortly after but there was no real pipe break or loca but never mind that in the scenario we re looking at here in any case remember that many of the eccs sub systems have different redundant energy sources due to the rapid escape of coolant from the reactor core hpci and rcic will fail rapidly due to loss of steam pressure but this is immaterial as the 2 000 l min flow rate of rcic available after t 5 is insufficient to maintain the water level nor would the 19 000 l min flow of hpci available at t 10 be enough to maintain the water level even if it could work without steam in the event of such a serious loca at t 10 the temperature of the reactor core at approximately 285 c at 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 initial loca fuel rod uncovery begins at approximately t 18 parts of the rods have reached 540 c at t 40 the core temperature is at 650 c and rising steadily the lpci and the pressure regulating core spray kick in and begin deluging the steam above the core and then the core itself a large amount of hot 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 fuel this happens after a few seconds as the approximately 200 000 l min of water these systems release begin to cool first the top of the core with the lpci deluging the fuel rods and the core spray suppressing the generated steam until at approximately t 100 seconds when all of the fuel is now subject to this deluge and the last remaining hot spots at the bottom of the core are now being cooled the peak temperature that is attained in the fuel elements in this scenario even with temporary uncovery of the fuel rods is 900 c well below the maximum of 1200 c which is acceptable before fuel damage begins at the bottom of the core which was the last hot spot to be cooled by the water deluge the core is now cooled rapidly and completely by the lpci and following cooling to a reasonable temperature below that consistent with the generation of steam the core spray is shut down and the lpci flow rate is decreased to a level consistent with maintenance of a steady temperature of the fuel rods which will drop over a period of days due to the decay in fission product decay power output within the fuel after a few days decay heat will have sufficiently abated to the point that defueling of the reactor is able to commence following defueling lpci can be shut down completely the explosion on march 12 there was an explosion near the fukushima i 1 reactor building what happened there are about 5 different layers of containment which exist in a power reactor reactor like the ones at fukushima between the people outside and the potentially dangerous radioactive fission products within the nuclear fuel the fuel rods themselves are clad in tubes of zirconium alloy and that represents one such layer that nuclear fuel is inside the reactor pressure vessel which is made of steel six inches thick and that reactor vessel is the next such layer the reactor pressure vessel is within the primary containment vessel the drywell which is made of steel one inch thick and that represents the next such layer the primary containment vessel is within the secondary containment structure which is made of steel reinforced pre stressed concrete between 4 and 8 feet thick the reactor building which is built around the secondary containment structure is the last of these multiple layers of containment and it is also made of steel reinforced pre stressed concrete between 30 cm and 1 m thick if every possible measure standing between safe operation of the plant and severe core damage and melting of the nuclear fuel fails the containment can be sealed indefinitely and it will prevent any significant release of radioactivity to the outside environment occurring under any circumstances now let s look at some diagrams of these structures click through for the full resolution images the outermost layer of the multiple layers of containment the reactor building has walls and a roof made of solid concrete and it s roughly cube shaped on top of the concrete reactor building however there is an additional part of the structure it is not made of concrete but it is made of steel with steel sheets over a steel frame this steel building on top of the reactor building houses the fuel transfer crane and it is built on top of the concrete roof of the reactor building i m referring to the part of the structure above the concrete shield plug and the refueling platform at the top of the concrete reactor building as shown on the first of the diagrams above it is this relatively weak steel structure on top of the reactor building which is not really part of the reactor building proper which seems to have been blown out by a hydrogen explosion the explosion at fukushima i 1 does not appear to have occurred within nor does it appear to have breached any of the fundamental layers of containment structure described above now an explosion has not occured as a result of a release of nuclear energy that is a scenario that is simply outside the laws of reality an explosion can be caused by one of two things a chemical explosion such as an ignition of a hydrogen oxygen mixture or a sudden release of stored gas or steam pressure it appears that the structure has probably been damaged as a result of a hydrogen explosion it s probable that excessive hydrogen generation within the reactor core either radiolytically or chemically by reduction of water in the presence of the zirconium cladding at significantly elevated temperatures has been vented into the torus and as temperatures and pressures have began to rise within the torus steam pressure in the torus has been vented out into the reactor building surrounding the torus from there the hydrogen mixed with that steam and water vapor has risen as hydrogen does and worked its way through the reactor building escaping at the top of the reactor building and accumulating at the top in the area around the fuel transfer crane it appears that the accumulated hydrogen has then mixed with air and exploded radiochemistry and radioactivity releases to the environment when a light water reactor is operating some of the oxygen 16 in the water is activated into radioactive nitrogen 16 by the 16 o n p 16 n reaction 16 n is very short lived with a half life of only 7 seconds but its specific activity is correspondingly very high when a bwr nuclear power station is operating the entire nuclear steam supply system including the turbine hall is a radiological controlled area due to the radioactivity from 1...
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