Neutron Transport and Reactor Physics Analysis

Neutron transport and reactor physics analysis form the theoretical and computational foundation for the design, operation and safety assessment of nuclear reactors. At its core lies the neutron transport equation, which describes the spatial, angular and energy distribution of neutrons as they scatter, absorb or induce fission in reactor materials. Two main solution strategies prevail: deterministic methods, which discretise the transport equation in space, angle and energy to produce a system of algebraic equations; and stochastic Monte Carlo approaches, which simulate the random paths of individual neutrons to estimate macroscopic quantities such as flux and reaction rates. Reactor physics analysis couples these neutron distributions with thermal-hydraulic feedbacks, fuel depletion and structural material behaviour to predict core performance over time. Accurate nuclear data libraries, encompassing cross sections and fission yields, underpin these calculations. Advances in computational power and algorithms have enabled high-fidelity multiphysics simulations, integrating neutron transport with fluid flow, heat transfer and fuel cycle modelling. This integration is essential for next-generation reactor concepts, waste transmutation strategies and real-time operational support, offering enhanced safety margins and reduced environmental impact.

Research from Nature Portfolio

Recent studies have demonstrated innovative methods for in-core transmutation of long-lived fission products using fast-spectrum reactors without chemical separation. By introducing a novel yttrium-deuteride moderator material into peripheral blanket regions, researchers have successfully softened the neutron spectrum leaking from the core. This adjustment greatly increases capture rates in target nuclides, reducing their effective half-lives by orders of magnitude and achieving support ratios close to unity for several problematic isotopes. The work highlights a practical route to integrate radioactive waste reduction directly into reactor operations, offering a paradigm shift in sustainable reactor design and waste management.

Neutron Transport and Reactor Physics Analysis publication trend

The graph below shows the total number of articles in neutron transport and reactor physics analysis across all publications each year (not limited to Nature Index journals).

Technical terms

Neutron transport equation: A balance equation describing the change in neutron density due to streaming, collisions, fission and external sources as a function of position, direction and energy.

Monte Carlo simulation: A stochastic method that tracks individual neutrons through probabilistic interactions to estimate macroscopic reactor parameters without spatial or angular discretisation.

Neutron cross section: A measure of the probability of interaction (scattering, absorption or fission) between a neutron and a target nucleus, typically expressed in barns.

Nuclear data library: A structured collection of evaluated nuclear quantities, including cross sections, fission yields and decay data, essential for accurate reactor and shielding calculations.

Reactivity coefficient: The change in reactor reactivity per unit change in a physical parameter (temperature, void fraction or moderator density), indicating the feedback effect on core behaviour.

References

1. Upsampling Monte Carlo neutron transport simulation tallies using a convolutional neural network. Energy and AI (2023).
2. Japanese evaluated nuclear data library version 5: JENDL-5. Journal of Nuclear Science and Technology (2023).
3. NCrystal: A library for thermal neutron transport. Computer Physics Communications (2020).
4. Method to Reduce Long-lived Fission Products by Nuclear Transmutations with Fast Spectrum Reactors. Scientific Reports (2017).