In this research, the design and evaluation of reactor physics experiments in support of MSR development are discussed. The reference MSR is a 600 MW (thermal), fast neutron spectrum reactor using a mixture of NaCl–CaCl2–UCl3–PuCl3 as fuel. An MSR core was designed with the composition of the fuel and the core geometry based on several references. The reactor physical characteristics were calculated for the core using the ERANOS code system. The results show that the reactor can be critical with a compact size and has high passive safety with negative temperature feedback coefficients. A control rod bundle made of stainless steel without the need for strong neutron absorbers was designed. It is expected that reactor physics experiments are needed for MSR development. To design reactor physics experiments for the MSR, the current research target is to use the existing critical assembly in Japan, KUCA, to its maximum potential. Several configurations are proposed and the representativity factor of the experiments to the MSR was evaluated. The results show that the designed configurations have a relatively low representativity. Therefore, a configuration based on the ZPPR using plutonium fuel and with a large size was designed. The results demonstrate that the experimental configuration is very similar to the MSR. This suggests that to design reactor physics experiments for MSR, a critical facility like the ZPPR should be built in the future.

• The published paper by Rusov et al. contains several fundamental errors. • The work is based on physically impossible assumptions and has wrong conclusions. • Simulations of Rusov et al. are based on an incorrect depletion chain. • There is no "nuclear burning wave" in the work of Rusov et al.

Various research groups and private interprises are pursuing the design of a Molten Salt Reactor (MSR) as one of the Generation-IV concepts. In the current work a fast neutron MSR using chloride fuel is analyzed, specially analyzing the power production and neutron flux level in the Intermediate Heat Exchanger (IHX). The neutronic analysis in this work is based on a chloride-fuel MSR with 600 MW thermal power. The core power density was set to 100 MW m −3 with a core H/D [[EQUATION]] 1.0 amd four Intermediate Heat Exchanger (IHX). This leads to a power of 150 MW per IHX; this power is also comparable to the IHX proposed in the SAMOFAR framework. In this work, a preliminary design of a 150 MW helical-coil IHX for a chloride-fueled MSR is prepared and the fission rate, capture rate, and inelastic scatter rate are evaluated.

The capability of beta(eff) /? obtained by lambda-mode and omega-mode calculations is examined for target subcriticality in the accelerator-driven system through a comparison with that obtained in the pulsed-neutron source (PNS) experiments at the Kyoto University Critical Assembly. Directly measured results of beta(eff) /?, alpha and rho($) in the PNS experiments are validated by varying the experimental conditions: the external neutron source, detector position, detector type and subcriticality ranging between 500 and 7500 pcm. The numerical analyses of beta(eff) /? are conducted by using MCNP6.1 together with ENDF/B-VII.1 for lambda-mode calculations and PARTISN (SCALE6.2.2 with ENDF-VII.1 for the effective cross sections) for both the lambda-mode and the omega-mode calculations. The comparison between calculated and measured beta(eff) /? with varying subcriticality shows good agreement between the experiments and the omega-mode calculations, although a difference is observed between the experiments and the lambda-mode calculations.

In experimental benchmarks of the accelerator-driven system (ADS) conducted at the Kyoto University Critical Assembly (KUCA), the prompt neutron decay constant was measured using two types of pulsed neutron sources, i.e. a D-T neutron source and a spallation neutron source driven by a 100-MeV proton beam. The measurement results of are useful information to validate the numerical results predicted by the prompt -eigenvalue calculation. In this study, the numerical analysis of using a multi-energy group S-N neutron transport code was carried out for the uranium-lead zoned experimental cores. To reduce the discretization error owing to the deterministic code, the KUCA geometry was modelled in detail as a three-dimensional heterogeneous plate-by-plate geometry, and an improved variant of EON quadrature was utilized. In addition, the sensitivity coefficients of with respect to nuclear data were efficiently evaluated by first-order perturbation theory, followed by nuclear data-induced uncertainty quantification based on the 56 neutron-energy group SCALE covariance library. Consequently, the numerical results of were validated successfully by the experimental results of the pulsed neutron source method, compared with the range of the nuclear data-induced uncertainties.

Uncertainty quantification is conducted for the criticality of excess reactivity and control rod worth obtained at the Kyoto University Critical Assembly (KUCA). By combining SRAC2006 and MARBLE code systems, the sensitivity coefficients of the cross sections for aluminum-27 (Al-27) comprising mainly of core components are large in the solid-moderated and -reflected cores (A cores) at KUCA. Also, the uncertainty is dominant in the uranium-235 isotope by the covariance data of JENDL-4.0, and a quantitative value is about 150 pcm induced by the JENDL-4.0 data library in the KUCA A cores, whereas the covariance data of Al-27 are not prepared in JENDL-4.0. Moreover, the effect of decreasing uncertainty is obtained by applying the cross-sectional adjustment method to the uncertainty analyses. From the results, a series of uncertainty quantifications is expected to clarify the uncertainty of sub-criticality in accelerator-driven system experiments with spallation neutrons in the KUCA A cores.

Nuclear data-induced uncertainties of neutronics parameters (neutron multiplication factor k(eff), one-point kinetics parameters and prompt neutron decay constant alpha) are quantified for lead-bismuth zoned accelerator-driven system experiments at the Kyoto University Critical Assembly, in order to contribute validation for subcritical core analysis. The random sampling technique using SCALE6.2.1/Sampler/NEWT/PARTISN is utilized for the validation and the uncertainty quantification, because the random sampling technique is applicable for a problem which is not easy to apply the perturbation theory. Consequently, it is confirmed that the numerical results of alpha reasonably agree with the experimental ones, compared with the nuclear data-induced uncertainties. In addition, it is clarified that the nuclear data-induced correlations between alpha and k(eff) and between alpha and neutron generation time Lambda are strongly negative and positive, respectively. This fact implies that the numerical predictions of k(eff) and Lambda can be improved by the data assimilation technique using subcritical experimental results of alpha, which can be directly measured even for a deep subcritical system.

Within the Coordinated Research Project (CRP) initiated by the International Atomic Energy Agency (IAEA) for investigating Shutdown Heat Removal Tests (SHRT) at Experimental Breeder Reactor II (EBR-II), an optional neutronics benchmark has been defined for providing reactivity feedback coefficients for the thermal hydraulic analysis of SHRT-45R. Several institutes participated in this benchmark, including: Karlsruhe Institute of Technology (KIT), University of Fukui, Paul Scherrer Institute (PSI), Argonne National Laboratory (ANL) and the Italian National Agency ENEA. Several stochastic and deterministic codes were used for this purpose. The results obtained in general were in good agreement. Remained discrepancies have been underlined and discussed in the present paper.

A new capability of calculating fuel burnup sensitivity with the generalized perturbation theory is implemented into a deterministic reactor physics code system CBZ, which is under development at Hokkaido University. This capability is well verified through comparisons with reference sensitivities obtained by numerical differentiation. Nuclear data-induced uncertainties of two neutronics parameters, keff and coolant void reactivity, of an accelerator-driven system designed by the Japan Atomic Energy Agency are quantified using sensitivities calculated with the new version of CBZ, and effect of burnup term in the sensitivities is also quantified. On both keff and coolant void reactivity, cancellation between static and burnup components in nuclear data-induced uncertainties is observed; uncertainties become small if the burnup component is taken into account.

Because of the uncertainties associated with the definition of Critical Heat Flux (CHF), the best estimate approach should be used. In this paper the application of best-estimate approach for the analysis of CHF phenomenon in the boiling water reactors is presented. At first, the nodalization of RBMK-1500, BWR-5 and ABWR fuel assemblies were developed using RE-LAP5 code. Using developed models the CHF and Critical Heat Flux Ratio (CHFR) for different types of reactors were evaluated. The calculation results of CHF were compared with the well-known experimental data for light water reactors. The uncertainty and sensitivity analysis of ABWR 8 x 8 fuel assembly CHFR calculation result was performed using the GRS (Germany) methodology with the SUSA tool. Finally, the values of Minimum Critical Power Ratio (MCPR) were calculated for RBMK-1500, BWR-5 and ABWR fuel assemblies. The paper demonstrate how, using the results of sensitivity analysis, to receive the MCPR values, which covers all uncertainties and remains best estimated.

Accelerator-driven system experiments with spallation neutrons (100MeV protons and Pb-Bi target) are carried out in the 235U-fueled and Pb-Bi-zoned core at the Kyoto University Critical Assembly, under a subcritical state ranging between 1160 and 11,556pcm. In these experiments, measurement of the prompt neutron decay constant and the subcriticality is conducted by the pulsed neutron source (PNS) method and the Feynman-α method with the use of optical fiber detectors. The experimental results successfully validate the prompt neutron decay constant and the subcriticality through the deduction of kinetic parameters by both the PNS and the α-fitting methods. The detector position dependency, neutron spectrum and subcriticality measurement methods still remain, however, in these experiments. For onward studies, the experimental benchmarks obtained from these experiments are expected to be involved in the numerical verification of subcriticality on-line monitoring, in the analysis of subcriticality uncertainty and in the deterministic approach to kinetic parameters.

This paper discusses our calculations of the neutronics benchmark of the EBR-II Shutdown Heat Removal Test 45R (SHRT-45R). The SHRT-45R experiment was performed about 30 years ago in the EBR-II and involved an Unprotected Loss of Flow (ULOF). The experiment is now evaluated as a benchmark in a Coordinated Research Program of the IAEA. The SHRT-45R neutronics benchmark was analyzed with the ERANOS v2.0 code using cross sections based on JENDL-4.0. Calculated results include the multiplication factor, several feedback coefficients, as well as kinetics parameters and the power distribution throughout the core. Where possible, the uncertainty of the calculated parameters is also calculated. Results obtained with ERANOS v2.0 are generally in line with results from other benchmark participants. The energy deposition due to gamma rays in the core is not well calculated. Results with ERANOS v2.0 are generally satisfactory. Recommendations are included.