• Nuclear Engineering and Technology
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• v.56 no.6
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• pp.2388-2394
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• 2024

In our previous works, two different spherical burnable poison particles (BPPs) as B₄C and Gd₂O₃ in pin-in block type HTGR core had utilized to suppress the excess reactivity and to control long-term reactivity during the burnup period. In the present work, we performed the neutronic analysis of a prismatic HTGR operating at 850 ℃ with thermal power of 100 MW containing spherical and cylindrical BPPs and then studied the self-shielding effect of BPPs and shape effect. The calculations were performed when the surface area (1) or volume (2) of cylindrical BPPs equals to that of the spherical BPPs. The calculations showed that the neutronic parameters were slightly better for the second case than the first one, such as the excess reactivity of the reactor core at the beginning of the cycle were more suppressed, the core lifetime were more extended, and the fuel-burning were more efficiently. The neutron spectrum in each region of the cylindrical BBPs slightly differs than that of the spherical BPPs. Therefore, the self-shielding effect of BPPs on reactor core performance depends on the particle's geometrical shape.

• Nuclear Engineering and Technology
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• v.51 no.1
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• pp.1-12
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• 2019

One of proposed Accident Tolerant Fuel (ATF) concept is fully ceramic micro-encapsulated fuel (FCMF). FCMF using uranium mononitride (UN) has better safety aspects than $UO_2$ pellet fuel although it might not have a better neutronic performance due to the presence of matrix and high neutron-induced interaction of $^{14}N$. Before implementing UN-FCMF technology in Small-PWR, further research must be taken place to make sure the proposed design of fuel assembly has inherent safety features and maintain the fuel performance. This study focusses on the neutronic analysis of UN-FCMF based fuel assembly using Serpent and SCALE codes. It is shown in the proposed fuel assembly design has inherent safety features with respect to the fuel temperature reactivity coefficient, void reactivity coefficient, and moderator temperature reactivity coefficient. It is noted that the use of FCMF leads to a lower ratio of burnup to $^{235}U$ enrichment ratio compared to the $UO_2/Zr$ fuel.

• Nuclear Engineering and Technology
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• v.56 no.7
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• pp.2641-2649
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• 2024

Effective management of slightly used nuclear fuel (SUNF) is crucial for both technical and public acceptance reasons. SUNF management, radiotoxicity risk, and associated financial investment and technological capabilities are major concerns in nuclear power production. Reducing the volume of SUNF can simplify its management, and one possible solution is utilizing small modular reactors (SMR) and advanced fuel designs like those with thorium. This research focuses on studying the neutronic performance and radionuclide inventory of three different thorium fuel configurations. The mass of fissile material in thorium-based fuel significantly impacts Kinf, burn-up, and neutron energy spectrum. Compared to uranium, thorium as a fuel produces far fewer transuranic elements and less long-lived fission products (LLFPs) at the end of the core cycle (EOC). However, certain fission product elements produced from thorium-based fuel exhibit higher radioactivity at the beginning of the core cycle (BOC). Physical separation of thorium and uranium in the fuel block, like seed-and-blanket units (SBU) and duplex fuel designs, generate less radioactive waste with lower radioactivity and longer cycle lengths than homogeneous or mixed thorium-uranium fuel. Furthermore, the SBU and duplex feel designs exhibit comparable neutron spectra, leading to negligible differences in SUNF production between the two.

• Nuclear Engineering and Technology
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• v.56 no.1
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• pp.167-179
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• 2024

This study explores the feasibility of employing (U, Th)-based accident tolerant fuels (ATFs), specifically (_0.8UO₂, _0.2ThO₂), (_0.8UN, _0.2ThN), and (_0.8UC, _0.2ThC). The investigation assesses the overall performance of these proposed fuel materials in comparison to the conventional UO₂, focusing on deep neutronic and thermal-hydraulic (Th) analyses. Neutronic analysis utilized the MCNPX code, while COMSOL Multiphysics was employed for thermal-hydraulic analysis. The primary objective of this research is to overcome the limitations associated with traditional UO₂ fuel by exploring alternative fuel materials that offer advantages in terms of abundance and potential improvements in performance and safety. Given the limited abundance of UO₂, long-term sustainable nuclear energy production faces challenges. From a neutronic standpoint, the U-Th based fuels demonstrated remarkable fuel cycle lengths, except (_0.8UN, _0.2ThN), which exhibited the minimum fuel cycle length and, consequently, the lowest fuel burn-up. Regarding thermal-hydraulic performance, (_0.8UN, _0.2ThN) exhibited outstanding performance with significant margins against fuel melting compared to the other materials. Overall, when considering the integrated performance, the most favourable results were obtained with the use of the (_0.8UC, _0.2ThC) fuel configurations. This study contributes valuable insights into the potential benefits of (U, Th)-based ATFs as a promising avenue for enhanced nuclear fuel performance.

• Nuclear Engineering and Technology
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• v.52 no.8
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• pp.1603-1610
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• 2020

Due to ever-growing advancements in computers and relatively easy access to them, many efforts have been made to develop high-fidelity, high-performance, multi-physics tools, which play a crucial role in the design and operation of nuclear reactors. For this purpose in this study, the neutronic Monte Carlo and thermal-hydraulic sub-channel codes entitled MCNP and COBRA-EN, respectively, were applied for external coupling with each other. The coupled code was validated by code-to-code comparison with the internal couplings between MCNP5 and SUBCHANFLOW as well as MCNP6 and CTF. The simulation results of all code systems were in good agreement with each other. Then, as the second problem, the core of the VVER-1000 v446 reactor was simulated by the MCNP4C/COBRA-EN coupled code to measure the capability of the developed code to calculate the neutronic and thermohydraulic parameters of real and industrial cases. The simulation results of VVER-1000 core were compared with FSAR and another numerical solution of this benchmark. The obtained results showed that the ability of the MCNP4C/COBRA-EN code for estimating the neutronic and thermohydraulic parameters was very satisfactory.

• Nuclear Engineering and Technology
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• v.52 no.12
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• pp.2709-2716
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• 2020

Compositions of large nuclear cores (e.g. boiling water reactors) are highly heterogeneous in terms of fuel composition, control rod insertions and flow regimes. For this reason, they usually lack high order of symmetry (e.g. 1/4, 1/8) making it difficult to estimate their neutronic parameters for large spaces of possible loading patterns. A detailed hyperparameter optimization technique (a combination of manual and Gaussian process search) is used to train and optimize deep neural networks for the prediction of three neutronic parameters for the Ringhals-1 BWR unit: power peaking factors (PPF), control rod bank level, and cycle length. Simulation data is generated based on half-symmetry using PARCS core simulator by shuffling a total of 196 assemblies. The results demonstrate a promising performance by the deep networks as acceptable mean absolute error values are found for the global maximum PPF (~0.2) and for the radially and axially averaged PPF (~0.05). The mean difference between targets and predictions for the control rod level is about 5% insertion depth. Lastly, cycle length labels are predicted with 82% accuracy. The results also demonstrate that 10,000 samples are adequate to capture about 80% of the high-dimensional space, with minor improvements found for larger number of samples. The promising findings of this work prove the ability of deep neural networks to resolve high dimensionality issues of large cores in the nuclear area.

• Nuclear Engineering and Technology
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• v.55 no.11
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• pp.4048-4056
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• 2023

The feasibility of using the Monte Carlo code MCS to generate multigroup cross sections for nodal diffusion simulations RAST-F of liquid metal fast reactors is investigated in this paper. The performance of the MCS/RAST-F code system is assessed using steady-state simulations of the ANTS-100e core. The results show good agreement between MCS/RAST-F and MCS reference solutions, with a keff difference of less than 77 pcm and root-mean-square differences in radial and axial power of less than 0.5% and 0.25%, respectively. Furthermore, the MCS/RAST-F reactivity feedback coefficients are within three standard deviations of the MCS coefficients. To validate the internal thermal-hydraulic (TH) feedback capability in RAST-F code, the coupled neutronic/TH1D simulation of ANTS-100e is performed using the case matrix obtained from MCS branch calculations. The results are compared to those obtained using the MARS-LBE system code and show good agreement with relative temperature differences in fuel and coolant of less than 0.8%. This study demonstrates that the MCS/RAST-F code system can produce accurate results for core steady-state neutronic calculations and for coupled neutronic/TH simulations.

• Nuclear Engineering and Technology
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• v.50 no.7
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• pp.1060-1067
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• 2018

A feasibility of reusing option of spent nuclear fuel in a fusion-fission hybrid system without partitioning was checked as an alternative option of pyro-processing with critical reactor system. Neutronic study was performed with MCNP 6.1 for this option, direct reuse of spent PWR fuel (DRUP). Various options with DRUP fuel were compared with the reference design concept; transmutation purpose blanket with (U-TRU)Zr fuel loading connected with pyro-processing. Performance parameters to be compared are transmutation performance of transuranic (TRU) nuclides, required fusion power and tritium breeding ratio (TBR). When blanket part is loaded only with DRUP, initial $k_{eff}$ level becomes too low to maintain a practical subcritical system, increasing the required fusion power. In this case, production rate of TRU nuclides exceeds the incineration rate. Design optimization is done for combining DRUP fuel with (U-TRU)Zr fuel. Reactivity swing is reduced to about 2447 pcm through fissile breeding compared to (U-TRU)Zr fuel option. Therefore, a required fusion power is reduced and tritium breeding performance is improved. However, transmutation performance with TRU nuclides especially $^{241}Am$ is degraded because of softening effect of spectrum. It is known that partitioning and transmutation should be accompanied with fusion-fission hybrid system for the effective transmutation of TRU.

• Nuclear Engineering and Technology
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• v.52 no.3
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• pp.633-646
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• 2020

Imaging and scattering techniques using thermal neutrons allow to analyze complex specimens in scientific and industrial researches. Owing to this advantage, there have been a considerable demand for neutron facilities in the industrial sector. Among neutron sources, an accelerator driven compact neutron source is the only one that can satisfy the various requirements-construction budget, facility size, and required neutron flux-of industrial applications. In this paper, a target, moderator, and reflector (TMR) system for low-energy proton-accelerator driven compact thermal neutron source was designed via Monte Carlo simulations. For 10-30 MeV proton beams, the optimal conditions of the beryllium target were determined by considering the neutron yield and the blistering of the target. For a non-borated polyethylene moderator, the neutronic properties were verified based on its thickness. For a reflector, three candidates-light water, beryllium, and graphite-were considered as reflector materials, and the optimal conditions were identified. The results verified that the neutronic intensity varied in the order beryllium > light water > graphite, the compacter size in the order light water < beryllium < graphite and the shorter emission time in the order graphite < light water < beryllium. The performance of the designed TMR system was compared with that of existing facilities and were laid between performance of existing facilities.

• Proceedings of the Korean Nuclear Society Conference
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• 1996.11a
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• pp.289-294
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• 1996

Feasibility study on conceptual design tool for liquid metal reactor has been conducted to optimize the thermohydraulic and neutronic design parameters. To accomplish this task the neutronic code PRISM, fuel performance code and scaling method have been included into the conceptual design support system. ALMR(PRISM 303MWe) has been adopted as the reference plant and principally according to the power level, conceptual design parameters are optimized so that energy balance and neutronics balance seem to be satisfied. This paper presents only the results of optimization on primary system including the IHX system.