[en] A general solution strategy of the modified power iteration method for calculating higher eigenmodes has been developed and applied in continuous energy Monte Carlo simulation. The new approach adopts four features: 1) the eigen decomposition of transfer matrix, 2) weight cancellation for higher modes, 3) population control with higher mode weights, and 4) stabilization technique of statistical fluctuations using multi-cycle accumulations. The numerical tests of neutron transport eigenvalue problems successfully demonstrate that the new strategy can significantly accelerate the fission source convergence with stable convergence behavior while obtaining multiple higher eigenmodes at the same time. The advantages of the new strategy can be summarized as 1) the replacement of the cumbersome solution step of high order polynomial equations required by Booth's original method with the simple matrix eigen decomposition, 2) faster fission source convergence in inactive cycles, 3) more stable behaviors in both inactive and active cycles, and 4) smaller variances in active cycles. Advantages 3 and 4 can be attributed to the lower sensitivity of the new strategy to statistical fluctuations due to the multi-cycle accumulations. The application of the modified power method to continuous energy Monte Carlo simulation and the higher eigenmodes up to 4th order are reported for the first time in this paper. -- Graphical abstract: -- Highlights: •Modified power method is applied to continuous energy Monte Carlo simulation. •Transfer matrix is introduced to generalize the modified power method. •All mode based population control is applied to get the higher eigenmodes. •Statistic fluctuation can be greatly reduced using accumulated tally results. •Fission source convergence is accelerated with higher mode solutions.

[en] The noise propagation matrix method (NPMM) has been extended to get higher mode solutions. Previous studies show that the NPMM can be used to compute the dominance ratio of a system. It is essentially the same as the Coarse Mesh Projection Method (CMPM), both of which use the noise propagation matrix (NPM) to determine the dominance ratio, either after finishing the Monte Carlo simulation or on-the-fly during the simulation. Since only the fundamental fission source information is explicitly utilized while the higher mode information is implicitly contained in the statistical noises, the NPMM can usually only give an approximate estimation of the dominance ratio after thousands of cycles. In this study, the NPMM is extended by simulating the higher modes explicitly, so that the dominance ratio estimation can be more accurate and efficient. Besides, the higher mode solutions can be obtained at the same time with good accuracy and efficiency.

[en] This paper presents the preliminary result of BEAVRS Hot Full Power (HFP) solution at Beginning of Cycle (BOC). It is solved by in-house Monte Carlo code which is being developed at Ulsan National Institute of Science and Technology (UNIST). The code employs simple 1-dimensional Thermal Hydraulic (TH) module and multipole based On-The- Fly (OTF) cross section generation module. In this paper, fission reaction rate, fuel temperature, moderator density, moderator temperature, fuel temperature, and xenon number density distributions are presented. This paper presented preliminary solution of BEAVRS HFP state at BOC by Monte Carlo code which is being developed at UNIST. The five quantities were presented and all looks reasonable: Fission reaction rate, fuel temperature, xenon number density, moderator density, moderator temperature

[en] Multigroup cross section (MG XS) generation by the UNIST in-house Monte Carlo (MC) code MCS for fast reactor analysis using nodal diffusion codes is reported. The feasibility of the approach is quantified for two sodium fast reactors (SFRs) specified in the OECD/NEA SFR benchmark: a 1000 MW_th metal-fueled SFR (MET-1000) and a 3600 MW_th oxide-fueled SFR (MOX-3600). The accuracy of a few-group XSs generated by MCS is verified using another MC code, Serpent 2. The neutronic steady-state whole-core problem is analyzed using MCS/RAST-K with a 24-group XS set. Various core parameters of interest (core k_eff, power profiles, and reactivity feedback coefficients) are obtained using both MCS/RAST-K and MCS. A code-to-code comparison indicates excellent agreement between the nodal diffusion solution and stochastic solution; the error in the core k_eff is less than 110 pcm, the root-mean-square error of the power profiles is within 1.0%, and the error of the reactivity feedback coefficients is within three standard deviations. Furthermore, using the super-homogenization-corrected XSs improves the prediction accuracy of the control rod worth and power profiles with all rods in. Therefore, the results demonstrate that employing the MCS MG XSs for the nodal diffusion code is feasible for high-fidelity analyses of fast reactors

[en] The Coarse Mesh Finite Difference Method (CMFD) has been widely used to accelerate the convergence of deterministic methods. It was shown that the CMFD acceleration technique is very effective for fission source convergence. It was expected that CMFD is also effective on active cycle by reducing inter-cycle correlation. However it turns out CMFD also has inter-cycle correlation. In order to reduce the inter-cycle correlation, well known technique superhistory method was adopted. In this paper, the tally performance of CMFD with superhistory method was studied with 1D homogeneous problem. The effect of CMFD with superhistory method on the MC was studied. It was shown that the inter-cycle correlation was reduced dramatically with CMFD and superhistory method. The magnitude of RMS real STD increases as the number of generations for superhistory increases. And the magnitude decreases with CMFD

[en] A possible technique to get the even higher modes is suggested, but it is difficult to be applied practically. In this paper, a general solution strategy is proposed, which can extend Tom Booth's modified power method to get the higher Eigenmodes and there is no limitation about the number of Eigenmodes that can be obtained with this method. In this paper, a general solution strategy is proposed, which can extend Tom Booth's modified power method to get the higher Eigenmodes and there is no limitation about the number of Eigenmodes that can be obtained with this method. It is more practical than the original solution strategy that Tom Booth proposed. The implementation of the method in Monte Carlo code shows significant advantages comparing to the original power method

[en] At the end of this work, case test results are presented for both PWR and VHTR system. The library generation process and resonance treatment method in the deterministic transport code in UNIST has been summarized. The code shows reasonable results of k_eff compared to Monte Carlo results for pin cell problems. Deterministic transport code has been developed at UNIST. The code uses the method of characteristics (MOC) for neutron transport analysis. The MOC can provide accurate numerical solutions when accurate multi-group cross section library is given. In this work, nuclear library generation process and implemented resonance treatment method are presented. The code can cover PWR and VHTR system. NJOY code system and RMET21 are used to generate self-shielded cross section and resonance parameter. Equivalence theory is adopted in the code as a resonance treatment method. Additionally, Doubly-heterogeneous (DH) self-shielding method is implemented in the code to consider double heterogeneity of fuel in VHTR

[en] Monte Carlo neutron transport code named MCS is under development at UNIST for the advanced reactor design and research purpose. This MC code can be used for fixed source calculation and criticality calculation. Continuous energy neutron cross section data and multi-group cross section data can be used for the MC calculation. This paper presents the overview of developed MC code and its calculation results. The real time fixed source calculation ability is also tested in this paper. The calculation results show good agreement with commercial code and experiment. A new Monte Carlo neutron transport code is being developed at UNIST. The MC codes are tested with several benchmark problems: ICSBEP, VENUS-2, and Hoogenboom-Martin benchmark. These benchmarks covers pin geometry to 3-dimensional whole core, and results shows good agreement with reference results

No abstract available

[en] A review of the documentation and an interpretation of the NEA-1517/74 and NEA-1517/80 shielding benchmarks (measurements of photon leakage flux from a hollow sphere with a central 14 MeV neutron source) from the SINBAD database with the Monte Carlo code MCS and the most up-to-date ENDF/B-VIII.0 neutron data library are conducted. The two analyzed benchmarks describe satisfactorily the energy resolution of the photon detector and the geometry of the spherical samples with inner beam tube, tritium target and cooling water circuit, but lack information regarding the detector geometry and the distances of shields and collimators relatively to the neutron source and the detector. Calculations are therefore conducted for a sphere model only. A preliminary verification of MCS neutron-photon calculations against MCNP6.2 is first conducted, then the impact of modelling the inner beam tube, tritium target and cooling water circuit is assessed. Finally, a comparison of calculated results with the libraries ENDF/B-VII.1 and ENDF/B-VIII.0 against the measurements is conducted and shows reasonable agreement. The MCS and MCNP inputs used for the interpretation are available as supplementary material of this article