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

Fig2

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

List of publications

Lye at al. (2025) Physics-enhanced machine learning for probabilistic risk assessment in nuclear safety: An overview, recent developments, and perspectives 10.1016/j.anucene.2025.111562

Lye et al. (2025) A Hellinger distance-based stochastic model updating framework for the accreditation validation of a material thermal property under limited data, In the Proceedings of the 35th European Safety and Reliability & 33rd Society for Risk Analysis Europe Conference 10.3850/978-981-94-3281-3_ESREL-SRA-E2025-P4632-cd

Lye et al. (2024) Advances in the Reliability Analysis of Coherent Systems under Limited Data with Confidence Boxes https://doi.org/10.1061/AJRUA6.RUENG-1380

Lye et al. (2024) An Overview of Probabilistic Safety Assessment for Nuclear Safety: What Has Been Done, and Where Do We Go from Here? https://doi.org/10.3390/jne5040029

Lye, Ferson, Xiao (2024) Distribution-free stochastic model updating for the Physics-guided reliability analysis of a material thermal property under limited data https://iapsam.org/PSAM17/program/Papers/PSAM17&ASRAM2024-1002.pdf

Lye, Ferson, Xiao (2024) Stochastic Model Updating Using The Jenson-Shannon Divergence For Calibration and Validation Under Limited Data

Tan et al. (2024) Preliminary MELCOR Input Deck for Steady State and Turbine Trip Analysis of the Nuscale US600 SMR https://inis.iaea.org/records/zmh11-gf291

Lye et al. (2024) Comparison between Distance Functions for Approximate Bayesian Computation to Perform Stochastic Model Updating and Model Validation under Limited Data https://doi.org/10.1061/AJRUA6.RUENG-1223

Choo & Xiao (2023) Criticality Analysis of HTR-10 Using the High-Temperature Gas-Cooled Reactor Code Package https://inis.iaea.org/records/9cr77-d5t43

Guo & Xiao (2022) Steady-state thermal hydraulic modelling and turbine trip transient simulation of the NuScale integral pressurised water reactor using the ASTEC code https://doi.org/10.3389/fenrg.2022.1036142

Than & Xiao (2022) lp-CMFD acceleration schemes in multi-energy group 2D Monte Carlo transport https://doi.org/10.3389/fenrg.2022.1035797

Guo & Xiao (2022) Preliminary Steady-State Thermal Hydraulic Modelling of the NuScale Integral Pressurised Water Reactor Using the CESAR Module of ASTEC https://www.ans.org/pubs/proceedings/article-51644/

Than & Xiao (2022) lp-CMFD Acceleration Schemes in 2D Monte Carlo Transport https://www.ans.org/pubs/transactions/article-51427/

Chan & Xiao (2021) Implementation and performance study of the lp-CMFD acceleration scheme for Monte Carlo method based k-eigenvalue neutron transport calculation in 1D geometry

Tiang & Xiao (2021) Long-term reactivity control of accident tolerant fuel loaded marine small modular reactor using particle-type burnable poisons https://doi.org/10.1016/j.anucene.2021.108177

Chan & Xiao (2021) A linear prolongation CMFD acceleration for two-dimensional discrete ordinate k-eigenvalue neutron transport calculation with pin-resolved mesh using discontinuous Galerkin finite element method https://doi.org/10.1016/j.anucene.2020.108103

Chan & Xiao (2021) Numerical stability analysis of lp-CMFD acceleration for the discrete ordinates neutron transport calculation discretized with discontinuous Galerkin Finite Element Method https://doi.org/10.1016/j.anucene.2020.108036

Krishna, Yap, Xiao (2020) Assembly design of a fluoride salt-cooled high temperature commercial-scale reactor: Neutronics evaluation and parametric analysis https://doi.org/10.1016/j.anucene.2019.107288

Chan & Xiao (2020) Implementation and performance study of lpCMFD acceleration method for multi-energy group k-eigenvalue neutron transport problem in hexagonal geometry https://doi.org/10.1016/j.anucene.2019.107220

Chan & Xiao (2019) Theoretical Convergence Study of lpCMFD for Fixed Source Neutron Transport Problems in 2D Cartesian Geometry http://dx.doi.org/10.13182/T30701

Chan & Xiao (2019) A Linear Prolongating Coarse Mesh Finite Difference Acceleration of Discrete Ordinate Neutron Transport Calculation Based on Discontinuous Galerkin Finite Element Method https://doi.org/10.1080/00295639.2020.1752045

Chan & Xiao (2019) Convergence study of variants of CMFD acceleration schemes for fixed source neutron transport problems in 2D Cartesian geometry with Fourier analysis https://doi.org/10.1016/j.anucene.2019.06.021

Chan & Xiao (2019) Convergence study of CMFD and lpCMFD acceleration schemes for k-eigenvalue neutron transport problems in 2-D Cartesian geometry with Fourier analysis http://dx.doi.org/10.1016/j.anucene.2019.05.035