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Nuclear power safety method that does not require electrical power nor intervention

Passive nuclear safety is a design approach for safety features, implemented in a nuclear reactor, that does not require any active intervention on the part of the operator or electrical/electronic feedback in order to bring the reactor to a safe shutdown state, in the event of a particular type of emergency (usually overheating resulting from a loss of coolant or loss of coolant flow). Such design features tend to rely on the engineering of components such that their predicted behaviour would slow down, rather than accelerate the deterioration of the reactor state; they typically take advantage of natural forces or phenomena such as gravity, buoyancy, pressure differences, conduction or natural heat convection to accomplish safety functions without requiring an active power source. Many older common reactor designs use passive safety systems to a limited extent, rather, relying on active safety systems such as diesel powered motors. Some newer reactor designs feature more passive systems; the motivation being that they are highly reliable and reduce the cost associated with the installation and maintenance of systems that would otherwise require multiple trains of equipment and redundant safety class power supplies in order to achieve the same level of reliability. However, weak driving forces that power many passive safety features can pose significant challenges to effectiveness of a passive system, particularly in the short term following an accident.

Three Mile Island Unit 2 was unable to contain about 480 PBq of radioactive noble gases from release into the environment and around 120 kL of radioactive contaminated cooling water from release beyond the containment into a neighbouring building. The pilot operated relief valve at TMI 2 was designed to shut automatically after relieving excessive pressure inside the reactor into a quench tank. However the valve mechanically failed causing the PORV quench tank to fill, and the relief diaphragm to eventually rupture into the containment building. The containment building sump pumps automatically pumped the contaminated water outside the containment building. Both a working PORV with quench tank and separately the containment building with sump provided two layers of passive safety. An unreliable PORV negated its designed passive safety. The plant design featured only a single open/close indicator based on the status of its solenoid actuator, instead of a separate indicator of the PORV's actual position. This rendered the mechanical reliability of the PORV indeterminate directly, and therefore its passive safety status indeterminate. The automatic sump pumps and/or insufficient containment sump capacity negated the containment building designed passive safety. The notorious RBMK graphite moderated, water cooled reactors of Chernobyl Power Plant disaster were designed with a positive void coefficient with boron control rods on electromagnetic grapples for reaction speed control. To the degree that the control systems were reliable, this design did have a corresponding degree of active inherent safety. The reactor was unsafe at low power levels because erroneous control rod movement would have a counter intuitively magnified effect. Chernobyl Reactor 4 was built instead with manual crane driven boron control rods that were tipped with the moderator substance, graphite, a neutron reflector. It was designed with an Emergency Core Cooling System (ECCS) that depended on either grid power or the backup Diesel generator to be operating. The ECCS safety component was decidedly not passive. The design featured a partial containment consisting of a concrete slab above and below the reactor – with pipes and rods penetrating, an inert gas filled metal vessel to keep oxygen away from the water cooled hot graphite, a fire proof roof, and the pipes below the vessel sealed in secondary water filled boxes. The roof, metal vessel, concrete slabs and water boxes are examples of passive safety components. The roof in the Chernobyl Power Plant complex was made of bitumen – against design – rendering it ignitable. Unlike the Three Mile Island accident, neither the concrete slabs nor the metal vessel could contain a steam, graphite and oxygen driven hydrogen explosion. The water boxes could not sustain high pressure failure of the pipes. The passive safety components as designed were inadequate to fulfill the safety requirements of the system. The General Electric Company ESBWR (Economic Simplified Boiling Water Reactor, a BWR) is a design reported to use passive safety components. In the event of coolant loss, no operator action is required for three days. The Westinghouse AP1000 ("AP" standing for "Advanced Passive") uses passive safety components. In the event of an accident, no operator action is required for 72 hours. Recent versions of the Russian VVER have added a passive heat removal system to the existing active systems, utilising a cooling system and water tanks built on top of the containment dome. The integral fast reactor was a fast breeder reactor run by the Argonne National Laboratory. It was a sodium cooled reactor capable of withstanding a loss of (coolant) flow without SCRAM and loss of heatsink without SCRAM. This was demonstrated throughout a series of safety tests in which the reactor successfully shut down without operator intervention. The project was canceled due to proliferation concerns before it could be copied elsewhere. The Molten Salt Reactor Experiment (MSRE) was a molten salt reactor run by the Oak Ridge National Laboratory. It was nuclear graphite moderated and the coolant salt used was FLiBe, which also carried the uranium 233 fluoride fuel dissolved in it. The MSRE had a negative temperature coefficient of reactivity: as the FLiBe temperature increased, it expanded, along with the uranium ions it carried; this decreased density resulted in a reduction of fissile material in the core, which decreased the rate of fission. With less heat input, the net result was that the reactor would cool. Extending from the bottom of the reactor core was a pipe that lead to passively cooled drain tanks. The pipe had a "freeze valve" along its length, in which the molten salt was actively cooled to a solid plug by a fan blowing air over the pipe. If the reactor vessel developed excessive heat or lost electric power to the air cooling, the plug would melt; the FLiBe would be pulled out of the reactor core by gravity into dump tanks, and criticality would cease as the salt lost contact with the graphite moderator. The General Atomics HTGR design features a fully passive and inherently safe decay heat removal system, termed the Reactor Cavity Cooling System (RCCS). In this design, an array of steel ducts line the concrete containment (and hence surround the reactor pressure vessel) which provide a flow path for air driven natural circulation from chimneys positioned above grade. Derivatives of this RCCS concept (with either air or water as the working fluid) has also been featured in other gas cooled reactor designs, including the Japanese High temperature engineering test reactor, the Chinese HTR 10, the South African PBMR, and the Russian GT MHR. While none of these designs have been commercialized for power generation research in these areas is active, specifically in support of the Generation IV initiative and NGNP programs, with experimental facilities at Argonne National Laboratory (home to the Natural convection Shutdown heat removal Test Facility, a 1/2 scale air cooled RCCS) and the University of Wisconsin (home to separate 1/4 scale air and water cooled RCCS).