• Nuclear Engineering and Technology
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• v.55 no.10
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• pp.3732-3753
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• 2023

The Real-time Analysis for Particle-transport and In-situ Detection (RAPID) Code System, developed based on the Multi-stage Response-function Transport (MRT) methodology, enables real-time simulation of nuclear systems such as reactor cores, spent nuclear fuel pools and casks, and sub-critical facilities. This paper presents the application of a novel fission matrix-based burnup methodology to the well-characterized JSI TRIGA Mark II research reactor. This methodology allows for calculation of nuclear fuel depletion by combination and interpolation of RAPID's burnup dependent fission matrix (FM) coefficients to take into account core changes due to burnup. The methodology is compared to experimentally validated Serpent-2 Monte Carlo depletion calculations. The results show that the burnup methodology for RAPID (bRAPID) implemented into RAPID is capable of accurately calculating the k_eff burnup changes of the reactor core as the average discrepancies throughout the whole burnup interval are 37 pcm. Furthermore, capability of accurately describing 3D fission source distribution changes with burnup is demonstrated by having less than 1% relative discrepancies compared to Serpent-2. Good agreement is observed for axially and pin-wise dependent fuel burnup and nuclear fuel nuclide composition as a function of burnup. It is demonstrated that bRAPID accurately describes burnup in areas with high gradients of neutron flux (e.g. vicinity of control rods). Observed discrepancies for some isotopes are explained by analyzing the neutron spectrum. This paper presents a powerful depletion calculation tool that is capable of characterization of spent nuclear fuel on the fly while the reactor is in operation.

• Nuclear Engineering and Technology
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• v.39 no.5
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• pp.603-616
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• 2007

Roy Huddle, having invented the coated particle in Harwell 1957, stated in the early 1970s that we know now everything about particles and coatings and should be going over to deal with other problems. This was on the occasion of the Dragon fuel performance information meeting London 1973: How wrong a genius be! It took until 1978 that really good particles were made in Germany, then during the Japanese HTTR production in the 1990s and finally the Chinese 2000-2001 campaign for HTR-10. Here, we present a review of history and present status. Today, good fuel is measured by different standards from the seventies: where $9*10^{-4}$ initial free heavy metal fraction was typical for early AVR carbide fuel and $3*10^{-4}$ initial free heavy metal fraction was acceptable for oxide fuel in THTR, we insist on values more than an order of magnitude below this value today. Half a percent of particle failure at the end-of-irradiation, another ancient standard, is not even acceptable today, even for the most severe accidents. While legislation and licensing has not changed, one of the reasons we insist on these improvements is the preference for passive systems rather than active controls of earlier times. After renewed HTGR interest, we are reporting about the start of new or reactivated coated particle work in several parts of the world, considering the aspects of designs/ traditional and new materials, manufacturing technologies/ quality control quality assurance, irradiation and accident performance, modeling and performance predictions, and fuel cycle aspects and spent fuel treatment. In very general terms, the coated particle should be strong, reliable, retentive, and affordable. These properties have to be quantified and will be eventually optimized for a specific application system. Results obtained so far indicate that the same particle can be used for steam cycle applications with $700-750^{\circ}C$ helium coolant gas exit, for gas turbine applications at $850-900^{\circ}C$ and for process heat/hydrogen generation applications with $950^{\circ}C$ outlet temperatures. There is a clear set of standards for modem high quality fuel in terms of low levels of heavy metal contamination, manufacture-induced particle defects during fuel body and fuel element making, irradiation/accident induced particle failures and limits on fission product release from intact particles. While gas-cooled reactor design is still open-ended with blocks for the prismatic and spherical fuel elements for the pebble-bed design, there is near worldwide agreement on high quality fuel: a $500{\mu}m$ diameter $UO_2$ kernel of 10% enrichment is surrounded by a $100{\mu}m$ thick sacrificial buffer layer to be followed by a dense inner pyrocarbon layer, a high quality silicon carbide layer of $35{\mu}m$ thickness and theoretical density and another outer pyrocarbon layer. Good performance has been demonstrated both under operational and under accident conditions, i.e. to 10% FIMA and maximum $1600^{\circ}C$ afterwards. And it is the wide-ranging demonstration experience that makes this particle superior. Recommendations are made for further work: 1. Generation of data for presently manufactured materials, e.g. SiC strength and strength distribution, PyC creep and shrinkage and many more material data sets. 2. Renewed start of irradiation and accident testing of modem coated particle fuel. 3. Analysis of existing and newly created data with a view to demonstrate satisfactory performance at burnups beyond 10% FIMA and complete fission product retention even in accidents that go beyond $1600^{\circ}C$ for a short period of time. This work should proceed at both national and international level.

• Nuclear Engineering and Technology
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• v.53 no.5
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• pp.1569-1579
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• 2021

This study aims to provide microstructural characterization for the matrix graphite which molten salt reactors (MSRs) use, and improve resistance to molten salt infiltration of the matrix graphite for fuel elements. Mesocarbon microbeads (MCMB) densified matrix graphite A3-3 (MDG) was prepared by a quasi-isostatic pressure process. After densification by MCMBs with average particle sizes of 2, 10, and 16 ㎛, the pore diameter of A3-3 decreased from 924 nm to 484 nm, 532 nm, and 778 nm, respectively. Through scanning electron microscopy, the cross-section energy spectrum and time-of-flight secondary ion mass spectrometry, resistance levels of the matrix graphite to molten salt infiltration were analyzed. The results demonstrate that adding a certain proportion of MCMB powders can improve the anti-infiltration ability of A3-3. Meanwhile, the closer the particle size of MCMB is to the pore diameter of A3-3, the smaller the average pore diameter of MDG and the greater the densification. As a matrix graphite of fuel elements in MSR was involved, the thermal and mechanical properties of matrix graphite MDG were also studied. When densified by the MCMB matrix graphite, MDGs can meet the molten salt anti-infiltration requirements for MSR operation.

• Nuclear Engineering and Technology
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• v.53 no.2
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• pp.393-398
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• 2021

Current risk assessment of Spent Nuclear Fuel (SNF) transportation has the problem of the incomplete risk factors consideration and the general particle diffusion model utilization. In this paper, the accident frequency calculation and the detailed simulation of the accident consequences are coupled by the integrated risk assessment method. The "man-machine-environment" three-dimensional comprehensive risk indicator system is established and quantified to characterize the frequency of the transportation accidents. Consideration of vegetation, building and turbulence effect, the standard k-ε model is updated to simulate radioactive consequence of leakage accidents under complex terrain. The developed method is applied to assess the risk of the leakage accident in the scene of the typical domestic SNF Road Transportation (SNFRT). The critical risk factors and their impacts on the dispersion of the radionuclide are obtained.

• Proceedings of the Korean Nuclear Society Conference
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• 1999.05a
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• pp.259-259
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• 1999

• Proceedings of the Korean Nuclear Society Conference
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• 2005.10a
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• pp.7-8
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• 2005

• Proceedings of the Korean Nuclear Society Conference
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• 2005.10a
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• pp.19-20
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• 2005

• KSTLE International Journal
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• v.10 no.1_2
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• pp.27-32
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• 2009

In this study, fretting wear tests of nuclear fuel rods have been performed by using two kinds of spacer grid springs with a concave and a convex shape in room temperature dry and distilled water conditions. The objectives were to examine the variation of the wear mechanism with increasing fretting cycles and to evaluate the difference of the wear debris detachment behavior at each test environment. From the test results, the wear volume of each spring condition increased with increasing fretting cycles regardless of the test environments. However, the wear rate did not show a regular tendency and apparently changed with increasing fretting cycles. This is because the formation of the wear particle layer and/or the variation of the contact condition between the fuel rod and spring surfaces could affect a critical plastic deformation for detaching the wear debris. Based on the test results, the relationship between the wear behavior of each spring shape and test environment condition, and the variation of the surface characteristics are discussed in detail.

• Nuclear Engineering and Technology
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• v.40 no.3
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• pp.175-182
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• 2008

The results of process development for the blending of waste salt from the electrorefining of spent fuel with zeolite-A are presented. This blending is a key step in the ceramic waste process being used for treatment of EBR-II spent fuel and is accomplished using a high-temperature v-blender. A labscale system was used with non-radioactive surrogate salts to determine optimal particle size distributions and time at temperature. An engineering-scale system was then installed in the Hot Fuel Examination Facility hot cell and used to demonstrate blending of actual electrorefiner salt with zeolite. In those tests, it was shown that the results are still favorable with actinide-loaded salt and that batch size of this v-blender could be increased to a level consistent with efficient production operations for EBR-II spent fuel treatment. One technical challenge that remains for this technology is to mitigate the problem of material retention in the v-blender due to formation of caked patches of salt/zeolite on the inner v-blender walls.

• Nuclear Engineering and Technology
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• v.54 no.10
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• pp.3864-3877
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• 2022

We propose a new approach called PESA (Prioritized replay Evolutionary and Swarm Algorithms) combining prioritized replay of reinforcement learning with hybrid evolutionary algorithms. PESA hybridizes different evolutionary and swarm algorithms such as particle swarm optimization, evolution strategies, simulated annealing, and differential evolution, with a modular approach to account for other algorithms. PESA hybridizes three algorithms by storing their solutions in a shared replay memory, then applying prioritized replay to redistribute data between the integral algorithms in frequent form based on their fitness and priority values, which significantly enhances sample diversity and algorithm exploration. Additionally, greedy replay is used implicitly to improve PESA exploitation close to the end of evolution. PESA features in balancing exploration and exploitation during search and the parallel computing result in an agnostic excellent performance over a wide range of experiments and problems presented in this work. PESA also shows very good scalability with number of processors in solving an expensive problem of optimizing nuclear fuel in nuclear power plants. PESA's competitive performance and modularity over all experiments allow it to join the family of evolutionary algorithms as a new hybrid algorithm; unleashing the power of parallel computing for expensive optimization.