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.
Modern reactor designs have focused on increasing the number of passive systems to mitigate risk of compounding human error.
The concrete walls and the steel liner of the vessel exhibit passive safety, but require active systems (valves, feedback loops, external instrumentation, control circuits, etc.)
On the other hand active designs employing variable controls permit much more precise accomplishment of safety functions; this may be particularly desirable under accident management conditions.Nuclear reactor response properties such as Temperature coefficient of reactivity and Void coefficient of reactivity usually refer to the thermodynamic and phase-change response of the neutron moderator heat transfer process respectively.
The feature would only work if it responded faster than an emerging (steam) void and the reactor components could sustain the increased coolant pressure.
Combined inherent and passive safety depends only on physical phenomena such as pressure differentials, convection, gravity or the natural response of materials to high temperatures to slow or shut down the reaction, not on the functioning of engineered components such as high-pressure water pumps.
The Three Mile Island accident exposed this design deficiency: the reactor and steam generator were shut down but with loss of coolant it still suffered a partial meltdown.
[3] Third generation designs improve on early designs by incorporating passive or inherent safety features[4] which require no active controls or (human) operational intervention to avoid accidents in the event of malfunction, and may rely on pressure differentials, gravity, natural convection, or the natural response of materials to high temperatures.
If the reactor overheats, thermal expansion of the metallic fuel and cladding causes more neutrons to escape the core, and the nuclear chain reaction can no longer be sustained.
The large mass of liquid metal also acts as a heatsink capable of absorbing the decay heat from the core, even if the normal cooling systems would fail.
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.
[9] Both a working PORV with quench tank and separately the containment building with sump provided two layers of passive 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 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.
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 General Electric Company ESBWR (Economic Simplified Boiling Water Reactor, a BWR) is a design reported to use passive safety components.
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.
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)[15] and the University of Wisconsin (home to separate 1/4 scale air and water-cooled RCCS).