Recent activities in the field of reactor physics

Abstract

As the basis and fundamentals of nuclear technology, reactor physics has an important role to play; recent requirements for reliability and accountability to realize a higher level of safety have been encouraging researchers and engineers to study and develop more advanced and sophisticated numerical methods and calculation codes. Many of the outstanding research and developments are presented in scientific journals, including the Journal of Nuclear Science and Technology. Some topics have been covered in this summary from the latest activities in the field of reactor physics.

Keywords:

• reactor physics
• reactor analysis method
• Monte Carlo method
• deterministic method
• sensitivity analysis
• uncertainty analysis

As the computational cost is being drastically reduced by virtue of the significant advancements of computer technology, first-principle calculations based on the Monte Carlo method are widely used for solving the Boltzmann equations for neutral particle transport. Recently, there have been some significant advancements/developments in the major Monte Carlo codes. For example, the MCNP6 code has been released as a merger of MCNP5 and MCNPX capabilities, and high confidence in the code is based on its performance with the verification and validation test suites [1]. Furthermore, a new functionality of calculating k-eigenvalue sensitivity coefficients has been successfully implemented in MCNP6 [2]. Modern computing environments enable us to apply the Monte Carlo method to a variety of applications such as reactor kinetics calculations [3], estimation of vessel neutron fluence and ex-core chamber response [4], etc. However, the Monte Carlo method is not almighty; it is important to consider autocorrelation and bias characteristics in the Monte Carlo calculations under extreme conditions [5].

For the purpose of design calculations for advanced reactor concepts or determination of operational conditions for nuclear power reactors, deterministic methods are essential from the viewpoint of the balance between accuracy of the solutions and computational speed. In order to meet this necessity, various approaches have been studied in research and development of reactor analysis methods. For example, new resonance self-shielding methods based on the equivalence theory or the subgroup model have been proposed to improve the accuracy of the transport calculations [6,7]. The generalized equivalence theory and the superhomogeneisation (SPH) method, which have been widely used in current core calculations with the advanced nodal method to mitigate cell/assembly homogenization errors, have been applied to the integro-differential transport equation to further improve the accuracy of whole-core calculations [8,9]. Improvements on robustness, efficiency and accuracy of transport calculations based on the method of characteristics [10–12] and calculations of the space-dependent kinetic equation [13] are expected to enlarge the range of applicability of those advanced methods in realistic problems.

One direction of improvement in the reactor analysis method is realization of so-called pin-by-pin calculations in three-dimensional fine-mesh systems in order to directly treat heterogeneity inside fuel assemblies where multi-group transport equations are solved instead of the two-group diffusion equation that is widely used in industrial applications. With the success of the approach for pressurized water reactors (PWRs) [14], applications to boiling water reactor (BWR) problems have also been studied; modeling of the pin-cell homogenized cross section [15] and its energy structure [16] are examined since there is strong intra-/inter-assembly heterogeneity in BWR configurations. Another approach is to enhance calculation methods for pin power distribution within the framework of the advanced nodal method; a method using pin-wise cross sections [17] has been proposed. Databases such as a stylized three-dimensional PWR benchmark problem [18] and experimental data for BWR fuel assemblies [19] are valuable for the verification and validation of pin power calculation methods.

Sensitivity/uncertainty analyses and uncertainty reduction approaches, such as the bias factor method and the cross section adjustment method, are important subjects in research and development of the best-estimate methods and codes. Those subjects have been well studied, especially for fast reactors (FRs) [20] followed by recent research works on light water reactors [21–23] and accelerator-driven systems (ADSs) [24]. With recent improvements in the computing environment, new approaches for sensitivity analysis and cross section adjustments have been studied; stochastic methods such as the “Total Monte Carlo” or “Random Sampling” approach have been proposed for adjustment of cross section data [25,26] and uncertainty analysis [27]. A reduced-order modeling based on the subspace method [28] as a “generalized perturbation theory (GPT) free” approach has been proposed for the problems where GPT does not work because many responses are required.

Reactor physics has also contributed to the development of advanced nuclear reactors such as FRs and ADSs. Advanced concepts of FR cores have been proposed by several authors [29,30]. On ADSs, some fundamental research works have been carried out using an experimental facility [31,32].