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 for the PORV rather than separate open and close indicators. 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 fulfil 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 Electric Company AP-1000 ("AP" standing for "Advanced Passive") is a design reported to use passive safety components. In the event of an accident, no operator action is required for 72 hours.

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 was a molten salt reactor run by the Oak Ridge National Laboratory. It was a fluoride salt cooled reactor in which the fuel molecules function also as a molten fluoride salt coolant. It featured thermochemical freeze valves in which the molten salt was actively cooled to freezing point by air in flattened sections of the Hastelloy-N salt piping to block flow. If the reactor vessel developed excessive heat or if electric power was lost to the air cooling, then the fuel and coolant could thermochemically penetrate the valve into drain tanks away from the neutron reflector becoming sub-critical enroute for passive or active water cooling. During testing, it was observed that about 6–10% of the calculated 54 Ci/day (2.0 TBq/day) production of tritium diffused out of the fuel system into the containment cell atmosphere and another 6–10% reached the air through the heat removal system. Inhalation of 70 GBq of tritium is equivalent to an adult human dose of 3 Sv in which 50% of cases would be expected to die within 30 days. The fluoride salt molecular bond passive safety component failed to prevent tritium production from fission thus presenting a proliferation risk. The fluoride salt molecular bonds did not prevent tritium from leaking into the containment.

The fleet of BWRs and PWRs operating within the last 10 years in the United States have reported on 42 occasions a quarterly average daily tritium emission level of more than 22 mCi/day (70 GBq/day) from a power plant. During the first quarter of 2001 Palo Verde Unit 1 released on average 9 Ci/day (333 GBq/day) tritium gas. The passive safety component of water as neutron moderator failed to prevent excessive tritium gas (hydrogen with 2 neutrons) from being released from the plant as gas for dilution with air rather than water diluted tritiated water. Inhalation of tritium is absorbed at almost twice the rate as ingested tritium.

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