Severe Accident Simulation

Overview of a Typical Pressurized Water Reactor

We aim to develop capabilities to understand reactor designs and failure analysis through simulations of current and advanced reactors. These allow the assessment of the risks and source term releases associated with nuclear incidents within proximity to Singapore.

As commercial nuclear power plants (NPPs) today are predominantly pressurized water reactors (PWRs) or boiling water reactors (BWRs), ASTEC, a PWR severe accident simulator from IRSN, France, has been used to study various accident sequences and estimate the amount of radioactive releases. Additionally, MELCOR, a USNRC sponsored simulation code is also used to model newer generation of reactors, such as the high temperature gas-cooled reactor (HTGR). In the future, in-house developed codes will be used to model molten salt reactors. 

Figure 1 shows the schematic of a typical commercial PWR with the nuclear fuel undergoing controlled fission chain reactions within the reactor pressure vessel, which releases energy to heat the water in the primary coolant loop. The heated primary coolant is circulated through the steam generator where heat is transferred to the secondary coolant loop to produce steam and drive the turbines, which are connected to an electrical generator. Finally, the steam is cooled by the condenser and recirculated into the steam generator.

Schematic

Fig 1. Schematic of a typical Pressurized Water Reactor. Picture credit: United States Nuclear Regulatory Commission website https://www.nrc.gov/reading-rm/basic-ref/students/animated-pwr.html

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Fig 2. Schematic of a typical Boiling Water Reactor. Picture credit: United States Nuclear Regulatory Commission website https://www.nrc.gov/reading-rm/basic-ref/students/animated-bwr.html

Overview of Severe Accident Simulations 

In a core melt accident that resembles the Fukushima Daiichi accident in 2011, in which there was a loss of coolant (water) in the core, and therefore the loss of a heat transfer medium, the core containing the nuclear fuel assemblies is eventually exposed. Subsequently, the cladding breaks and the fuel rods melt. During this process the following occurs:

  1. Hydrogen is formed due to oxidation of zirconium and other metals. 
  2. Corium (molten core + structure) migrates to the lower plenum of vessel
  3. Vessel ruptures if decay heat cannot be removed fast enough.
  4. Corium drops to cavity and interacts with the concrete surface. Hydrogen is also formed in this process.
  5. Containment heats up and pressure builds up if containment spray system fails.
  6. Containment may fail (beyond pressure of 5 atm) – may also be caused by steam and hydrogen explosions or if corium burns through the basemat.
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Fig 3. State of the Reactor Pressure Vessel during a typical loss of cooling accident due to a station blackout (SBO) as time progresses (from left to right).

Animations

Animation 1. Cavity and Vessel Rupture

Animation 2. Core Melt Accident Progression

Further Essential Reading

Bal Raj Sehgal, Perspectives on LWR severe accidents and public risks, Nuclear Engineering and Design, Volume 354, 2019, 110253, ISSN 0029-5493,
https://doi.org/10.1016/j.nucengdes.2019.110253
http://www.sciencedirect.com/science/article/pii/S0029549319302705