[en] A Resonance calculation using energy Spectral Expansion (RSE) method has been applied to a heterogeneous pin cell geometry and its validity is confirmed through comparison with the ultra-fine group calculation. The RSE method utilizes expansion of angular flux by orthogonal basis on continuous energy. The singular value decomposition (SVD) and the low-rank approximation (LRA) are used to construct the orthogonal basis from ultra-fine group spectrum calculations in the homogeneous geometry. The calculation results show the validity of the present method. (author)

[en] High-fidelity multi-physics reactor core simulation includes the pin-by-pin neutronics, thermal-hydraulics, thermal-mechanics and fuel performance physical models. Every run of a high-fidelity code consumes a lot of computation time, memory and usually requires an expansive multi-core computational cluster. All output data is to be saved for further analysis if one does not want to repeat the expansive calculations. The memory amount of high-fidelity output data can be too large, e.g. in the case of uncertainty analysis. Therefore, the data compression algorithms are needed for reduction of the memory without accuracy loose. In this paper we suggest such algorithm based on Principle Component Analysis (PCA) adopted for multi-physics data. (author)

[en] This study presents a Monte Carlo based analysis for CANDLE burning reactor. In this procedure, neutronic calculation was performed by MVP code with JENDL-4.0 library and burnup calculation was performed by MVP-BURN code with the detailed burnup chain. To simulate the fuel movement process, an auxiliary program was developed by Python language. To demonstrate the feasibility of this procedure, a trial calculation was performed for a reference core design. The results showed that this Monte Carlo based procedure was developed successfully. (author)

[en] We investigate the effect of minor actinide recycling on nuclear wastes from fast reactors by using an updated version of a FRBurner module developed for fast reactor whole-core burn-up calculations in a code system CBZ. The update of CBZ enables calculation of inventory of transuranic wastes composed of hulls and end-pieces, so calculations of decay heat of high-level wastes and transuranic wastes become possible. Decay heat of cobalt-60 from hulls and end-pieces is as large as decay heat of fission products. It is turned out that minor actinide recycling conditions and cooling time changes are not important for decay heat of transuranic wastes in comparison with that of high-level wastes. (author)

[en] In order to improve the convergence behavior of the fixed-point iteration (Picard iteration) for neutronics/thermal-hydraulics coupled problems, Anderson acceleration is implemented in a pin-wise whole core analysis code nTER/ESCOT. The fixed-point map of Anderson acceleration is established for serially coupled whole core transport code and subchannel code. The performance of Anderson acceleration is examined with single assembly problems having low and high boron concentrations. The Anderson scheme shows the comparable convergence behavior to that of the optimum fixed-point iteration with under-relaxation factor. (author)

[en] Functional expansion tallies (FETs) are implemented in the Monte Carlo code MCS to improve the fidelity of multi-physics (MP) solution of light water reactor (LWR) problems. Neutronics (N) parameters (spatial flux or power) are expanded in a set of orthogonal basis functions (Legendre and/or Zernike polynomials) and the coefficients of the functional expansions are determined from the Monte Carlo (MC) simulations. With the coefficients, the required distribution is reconstructed continuously in space. Moreover, thermal-hydraulic (TH) quantities (fuel temperature, coolant density) obtained from the TH calculations of MCS are integrated into polynomial functions and their coefficients are calculated. Coupled MP simulation results are presented in this paper using a three-dimensional (3D) PWR fuel pin. FETs coefficients of power are transferred to the TH solver. Then the power distribution is reconstructed for the TH calculations. Subsequently, the coefficients of fuel temperature/coolant density are returned as feedback, and the fuel temperature/coolant density profiles are reconstructed for the next neutronic cycle. The results obtained with the coefficients demonstrates promise of the FETs for high fidelity MP coupled calculations. (author)

[en] This paper aims to clarify that the subcritical measurement at the solid moderated KUCA core is useful to reduce nuclide data-induced uncertainties of the effective neutron multiplication factors k_eff in other light water moderated systems. For this purpose, we pay attention to the prompt neutron decay constant, α, which was measured by the Feynman-α method with the moving block bootstrap method for the deep subcritical KUCA core under the shutdown state without any external neutron source. The sensitivity analyses of α for the subcritical KUCA core and of k_eff for various light water moderated critical cores were carried out based on the first-order perturbation theory. Consequently, it is demonstrated that the nuclear data-induced uncertainties of k_eff can be reduced by the bias factor method using the measurement value of α, if the target k_eff value has a strong correlation with the measured α value in the subcritical KUCA core. (author)

[en] In order to perform reactor core analysis with high accuracy, appropriate resonance treatment method plays a very important role within the whole calculation procedures. On the other hand, the fast reactor designs are the best choices to solve future energy problem. Therefore, the in-house code STREAM, which is initially developed for light water reactor calculation, is decided to upgrade for fast reactor analysis. Current work implements the Improved Tone's Method into the STREAM code to perform 2-D pin-cell/assembly calculation. Several verification tests show the accuracy of new STREAM code, with less than 50 pcm difference on k_eff value. (author)

[en] This paper presents passive safety device called FAST (Floating Absorber for Safety at Transient) for the improved inherent safety of sodium-cooled fast reactors (SFR). FAST is designed to cause negative reactivity feedback against coolant temperature rise compensating the positive reactivity induced by positive coolant temperature reactivity feedback. This study investigates the transient response of FAST in MOX-loaded SFR considering three ATWS (Anticipated Transient Without Scram) scenarios, which are ULOF (Unprotected Loss of Flow), ULOHS (Unprotected Loss of Heat Sink) and UTOP (Unprotected Transient Overpower). The ATWS simulations are performed with an in-house computer code taking into account all possible reactivity feedback components including the FAST movement and core expansion coefficients, and EOL (End-Of-Life) condition of the core with most positive CTC (Coolant Temperature Coefficient) is considered for the conservative analysis. Temperatures of core components and reactor power are considered as main factors of interest and it is confirmed that consequence of ULOF, ULOHS and UTOP in a MOX-loaded SFR core an be noticeably mitigated by the application of FAST device. (author)

[en] Critical experiment data of Toshiba Nuclear Critical Assembly were used to benchmark for FRENDY and NJOY codes, both with three neutron libraries: JENDL-4, ENDF/B-VII.1, and ENDF/B-VIII.0. The results obtained by using ACE-libraries processed by FRENDY and NJOY were in good agreements with the each other. (author)