• Nuclear Engineering and Technology
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• 제45권7호
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• pp.951-960
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• 2013

Energy group structure has a significant effect on the results of multigroup transport calculations. It is known that $UO_2-PuO_2$ (MOX) is a recently developed fuel which consumes recycled plutonium. For such fuel which contains various resonant nuclides, the selection of energy group structure is more crucial comparing to the $UO_2$ fuels. In this paper, in order to improve the accuracy of the integral results in MOX thermal lattices calculated by WIMSD-5B code, a swarm intelligence method is employed to optimize the energy group structure of WIMS library. In this process, the NJOY code system is used to generate the 69 group cross sections of WIMS code for the specified energy structure. In addition, the multiplication factor and spectral indices are compared against the results of continuous energy MCNP-4C code for evaluating the energy group structure. Calculations performed in four different types of $H_2O$ moderated $UO_2-PuO_2$ (MOX) lattices show that the optimized energy structure obtains more accurate results in comparison with the WIMS original structure.

• Nuclear Engineering and Technology
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• 제52권10호
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• pp.2387-2393
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• 2020

In this study, a gradient material for shielding neutrons and gamma rays was developed, which consists of epoxy resin, boron carbide (B₄C), lead (Pb) and a little graphene oxide. It aims light weight and compact, which will be applied on the transportable nuclear reactor. The material is made up of sixteen layers, and the thickness and components of each layer were designed by genetic algorithm (GA) combined with Monte Carlo N Particle Transport (MCNP). In the experiment, the viscosities of the epoxy at different temperatures were tested, and the settlement regularity of Pb particles and B₄C particles in the epoxy was simulated by matlab software. The material was manufactured at 25 ℃, the Pb C and O elements of which were also tested, and the result was compared with the outcome of the simulation. Finally, the material's shielding performance was simulated by MCNP and compared with the uniformity material's. The result shows that the shielding performance of gradient material is more effective than that of the uniformity material, and the difference is most noticeable when the materials are 30 cm thick.

• Nuclear Engineering and Technology
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• 제54권2호
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• pp.764-769
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• 2022

The United Arab Emirates is currently building and operating four units of the APR-1400 developed by a South Korean vendor, Korea Electric Power Corporation (KEPCO). This paper attempts to perform APR-1400 reactor core analysis by using the well-known two-step method. The two-step method was applied to the APR-1400 first cycle using the open-source nodal diffusion code, KOMODO. In this study, the group constants were generated using CASMO-4 fuel transport lattice code. The simulation was performed in Hot Zero Power (HZP) at steady-state and transient conditions. Some typical parameters necessary for the Nuclear Design Report (NDR) were evaluated in this paper, such as effective neutron multiplication factor, control rod worth, and critical boron concentration for steady-state analysis. Other parameters such as reactivity insertion, power, and fuel temperature changes during the Reactivity Insertion Accident (RIA) simulation were evaluated as well. The results from KOMODO were verified using PARCS and SIMULATE-3 nodal core simulators. It was found that KOMODO gives an excellent agreement.

• Nuclear Engineering and Technology
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• 제53권7호
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• pp.2151-2161
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• 2021

Nuclear data (ND) sensitivity and uncertainty (S/U) quantification in shielding applications is performed using deterministic and probabilistic approaches. In this paper the validation of the newly developed deterministic program package ASUSD (ADVANTG + SUSD3D) is presented. ASUSD was developed with the aim of automating the process of ND S/U while retaining the computational efficiency of the deterministic approach to ND S/U analysis. The paper includes a detailed description of each of the programs contained within ASUSD, the computational workflow and validation results. ASUSD was validated on two shielding benchmark experiments from the Shielding Integral Benchmark Archive and Database (SINBAD) - the fission relevant ASPIS Iron 88 experiment and the fusion relevant Frascati Neutron Generator (FNG) Helium Cooled Pebble Bed (HCPB) Test Blanket Module (TBM) mock-up experiment. The validation process was performed in two stages. Firstly, the Denovo discrete ordinates transport solver was validated as a standalone solver. Secondly, the ASUSD program package as a whole was validated as a ND S/U analysis tool. Both stages of the validation process yielded excellent results, with a maximum difference of 17% in final uncertainties due to ND between ASUSD and the stochastic ND S/U approach. Based on these results, ASUSD has proven to be a user friendly and computationally efficient tool for deterministic ND S/U analysis of shielding geometries.

• Nuclear Engineering and Technology
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• 제53권3호
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• pp.732-740
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• 2021

Scattering resonance of medium mass nuclides leads complex spectrum in the fast reactor, which requires thousands of energy groups in the spectrum calculation. When the broad-group cross sections are collapsed, reaction rate cannot be completely conserved. To eliminate the error from energy collapsing, the Super-homogenization method in energy collapsing (ESPH) was employed in the fast reactor code SARAX. An ESPH factor was derived based on the ESPH-corrected SN transport equation. By applying the factor in problems with reflective boundary condition, both the effective multiplication factor and reaction rate were conserved. The fixed-source iteration was used to ensure the stability of ESPH iteration. However, in the energy collapsing process of SARAX, the vacuum boundary condition was adopted, which was necessary for fast reactors with strong heterogeneity. To further reduce the error caused by leakage, an additional conservation factor was proposed to correct the neutron current in energy collapsing. To evaluate the performance of ESPH with conservation factor, numerical benchmarks of fast reactors were calculated. The results of broad-group calculation agreed well with the direct full-core Monte-Carlo calculation, including the effective multiplication factor, radial power distribution, total control rod worth and sodium void worth.

• Nuclear Engineering and Technology
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• 제54권12호
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• pp.4731-4742
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• 2022

RBMK fuel assemblies differ from other LWR FA due to a specific arrangement of the fuel rods, the low enrichment, and the used burnable absorber - erbium. Therefore, there is a challenge to adapt modeling tools, developed for other LWR types, to solve RBMK problems. A set of 10 different depletion simulation schemes were tested to estimate the impact on reactivity and spent fuel composition of possible SCALE code options for the neutron transport modelling and the use of different nuclear data libraries. The simulations were performed using cross-section libraries based on both, VII.0 and VII.1, versions of ENDF/B nuclear data, and assuming continuous energy and multigroup simulation modes, standard and user-defined Dancoff factor values, and employing deterministic and Monte Carlo methods. The criticality analysis with burn-up credit was performed for the SFP loaded with RBMK-1500 FA. Spent fuel compositions were taken from each of 10 performed depletion simulations. The criticality of SFP is found to be overestimated by up to 0.08% in simulation cases using user-defined Dancoff factors comparing the results obtained using the continuous energy library (VII.1 version of ENDF/B nuclear data). It was shown that such discrepancy is determined by the higher U-235 and Pu-239 isotopes concentrations calculated.

• Nuclear Engineering and Technology
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• 제55권1호
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• pp.378-390
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• 2023

The SPES (Selective Production of Exotic Species) facility, currently under development at Legnaro National Laboratories of INFN, aims at the production of intense RIB (Radioactive Ion Beams) employing the Isotope Separation On-Line (ISOL) technique for interdisciplinary research. The radioactive isotopes of interest are produced by the interaction of a multi-foil uranium carbide target with a 40 MeV 200 μA proton beam generated by a cyclotron proton driver. The Target Ion Source (TIS) is the core of the SPES project, here the radioactive nuclei, mainly neutron-rich isotopes, are stopped, extracted, ionized, separated, accelerated and delivered to specific experimental areas. Due to efficiency reasons, the TIS unit needs to be replaced periodically during operation. In this highly radioactive environment, the employment of autonomous systems allows the manipulation, transport, and storage of the TIS unit without the need for human intervention. A dedicated remote handling infrastructure is therefore under development to fulfill the functional and safety requirement of the project. This contribution describes the layout of the SPES target area, where all the remote handling systems operate to grant the smooth operation of the facility avoiding personnel exposure to a high dose rate or contamination issues.

• Nuclear Engineering and Technology
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• 제48권5호
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• pp.1273-1279
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• 2016

Production of iodine-131 by neutron activation of tellurium in tellurium dioxide ($TeO_2$) material requires a target that meets the safety requirements. In a radiopharmaceutical production unit, a new lid for a can was designed, which permits tight sealing of the target by using tungsten inert gaswelding. The leakage rate of all prepared targets was assessed using a helium mass spectrometer. The accepted leakage rate is ${\leq}10^{-4}mbr.L/s$, according to the approved safety report related to iodine-131 production in the TRIGA Mark II research reactor (TRIGA: Training, Research, Isotopes, General Atomics). To confirm the resistance of the new design to the irradiation conditions in the TRIGA Mark II research reactor's central thimble, a study of heat effect on the sealed targets for 7 hours in an oven was conducted and the leakage rates were evaluated. The results show that the tightness of the targets is ensured up to $600^{\circ}C$ with the appearance of deformations on lids beyond $450^{\circ}C$. The study of heat transfer through the target was conducted by adopting a one-dimensional approximation, under consideration of the three transfer modes-convection, conduction, and radiation. The quantities of heat generated by gamma and neutron heating were calculated by a validated computational model for the neutronic simulation of the TRIGA Mark II research reactor using the Monte Carlo N-Particle transport code. Using the heat transfer equations according to the three modes of heat transfer, the thermal study of I-131 production by irradiation of the target in the central thimble showed that the temperatures of materials do not exceed the corresponding melting points. To validate this new design, several targets have been irradiated in the central thimble according to a preplanned irradiation program, going from4 hours of irradiation at a power level of 0.5MWup to 35 hours (7 h/d for 5 days a week) at 1.5MW. The results showthat the irradiated targets are tight because no iodine-131 was released in the atmosphere of the reactor building and in the reactor cooling water of the primary circuit.

• Nuclear Engineering and Technology
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• 제52권4호
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• pp.661-673
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• 2020

The quarter-core simulation of BEAVRS Cycle 2 depletion benchmark has been conducted using the MCS/CTF coupling system. MCS/CTF is a cycle-wise Picard iteration based inner-coupling code system, which couples sub-channel T/H (thermal/hydraulic) code CTF as a T/H solver in Monte Carlo neutron transport code MCS. This coupling code system has been previously applied in the BEAVRS benchmark Cycle 1 full-core simulation. The Cycle 2 depletion has been performed with T/H feedback based on the spent fuel materials composition pre-generated by the Cycle 1 depletion simulation using refueling capability of MCS code. Meanwhile, the MCS internal one-dimension T/H solver (MCS/TH1D) has been also applied in the simulation as the reference. In this paper, an analysis of the detailed criticality boron concentration and the axially integrated assembly-wise detector signals will be presented and compared with measured data based on the real operating physical conditions. Moreover, the MCS/CTF simulated results for neutronics and T/H parameters will be also compared to MCS/TH1D to figure out their difference, which proves the practical application of MCS into the BEAVRS benchmark two-cycle depletion simulations.

• Journal of Radiation Protection and Research
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• 제11권1호
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• pp.37-43
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• 1986

ICRU 구(球)를 피사체(被射體)로 하여 300 keV 중성자(中性子)의 방사선량(放射線量) 관계량(關係量)을 평가(評價)하였다. 피사체내(被射體內)의 선량당량(線量當量) 분포(分布)를 직접(直接) 산정(算定)하기 위해 중성자(中性子)-광자(光子)-하전입자(荷電粒子) 결합수송(結合輸送)을 다룰 수 있는 몬테칼로 코드 NEDEP을 사용하였다. 계산결과(計算結果) 얻은 방사선량(放射線量) 관계량(關係量)은 다음과 같다. 심부선량당량지수(深部線量當量指數) $H_{I,d}:1.78{\times}10^{11}\;Sv-cm^2$ 표층선량당량지수(表層線量當量指數) $H_{I,s}:2.08{\times}10^{-1}\;Sv-cm^2$ 주위선량당량(周圍線量當量) $H^*(0.07):1.70{\times}10^{-11}\;Sv-cm^2$ 주위선량당량(周圍線量當量) $H^*(10):1.78{\times}10^{-11}\;Sv-cm^2$ 실효선질계수(實效線質係數) $\bar{Q}^*(10):12.4$