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In this work, new multi-recycling options of TRU nuclides using PWR fuel assemblies comprised of MOX and FCM (Fully Ceramic Micro Encapsulated) fuels are suggested and neutronically analyzed. These options do not use a fully recycling of TRU but a partial recycling where TRUs from MOX fuels are recycledwhile the ones fromFCM fuels arenotrecycled due totheirhigh consumptionrateresulted from high burnup. In particular, additional external TRU feed in MOX fuels for each cycle was considered to sign

In this paper, a new multi-group neutron-gamma transport calculation code system STRAUM-MATXST for complicated geometrical problems is introduced and its development status including numerical tests is presented. In this code system, the MATXST (MATXS-based Cross Section Processor for SN Transport) code generates multi-group neutron and gamma cross sections by processing MATXS format libraries generated using NJOY and the STRAUM (SN Transport for Radiation Analysis with Unstructured Meshes) code

New 80-MW (electric) ultra-long-life sodium cooled fast reactor core having inherent safety characteristics is designed with heterogeneous fuel assemblies comprised of driver and blanket fuel rods. Severaloptions using upper sodium plenum and SSFZ (Special Sodium Flowing Zone) for reducing sodium voidreactivity are neutronically analyzed in this core concept in order to improve the inherent safety of thecore. The SSFZ allowing the coolant flow from the peripheral fuel assemblies increases the ne

The discrimination of the source for xenon gases’ release can provide an important clue for detecting the nuclear activities in the neighboring countries. In this paper, three machine learning techniques, which are logistic regression, support vector machine (SVM), and k-nearest neighbors (KNN), were applied to develop the predictive models for discriminating the source for xenon gases’ release based on the xenon isotopic activity ratio data which were generated using the depletion codes, i.e.,

This paper presents the validation of the MCS code for critical safety analysis with burnup credit for thespent fuel casks. The validation process in this work considers five critical benchmark problem sets,which consist of total 80 critical experiments having MOX fuels from the International Criticality SafetyBenchmark Evaluation Project (ICSBEP). The similarity analysis with the use of sensitivity and uncertaintytool TSUNAMI in SCALE was used to determine the applicable benchmark experiments c

This study addresses the development and verification of a pin-by-pin core multigroup SP3 solver CTRP-Clouds that employs NEM (Nodal Expansion Method) and three simplified NEM methods within a unified formulation for simultaneously solving the coupling 0th and 2nd SP3 equations. In this work, the solver using this unified formulation does not only include the original NEM and its simplifications but also the EFEN (Exponential Function Expansion Nodal) method and FDM (Finite Difference Method) fo

The control rod worth is a key safety parameter required to be measured in commercial pressurized water reactors (PWRs). Conventionally, the control rod worth is measured after reaching the critical state, which occupies the considerable time in the zero-power physics test. In this study, an efficient control-rod worth measurement technique has been proposed based on the improved neutron-source multiplication method, which can be implemented with the source-range detector count rates in the subc

This paper introduces a new point depletion-based source term calculation code named BESNA (Bateman Equation Solver for Nuclear Applications), which is aimed to estimate nuclide inventories and source terms from spent nuclear fuels. The BESNA code employs a new modified CE/CM (Constant Extrapolation e Constant Midpoint) predictor-corrector scheme in depletion calculations for improving computational efficiency. In this modified CE/CM scheme, the decay components leading to the large norm of the

Since H2O is the predominant material utilized as both a moderator and coolant in nuclear reactor operations, accurately describing the neutron moderation process within the H2O is crucial for reactor design and analysis. This work investigates the production of thermal neutron scattering data for H2O using molecular dynamics (MD) simulations with various water models. Conventional thermal neutron scattering models, such as the GA(General Atomics) and IKE(Institut für Kernenergetik and Energiesy

Discrete ordinates (SN) method with unstructured meshes is highly appropriate for high-fidelity modeling and simulation of radiation shielding problems with complicated geometries. However, the large number of unknowns resulting from discretization of the transport equation on the spatial, angular, and energy variables, necessitates the use of parallel computing to achieve efficient solutions. In this work, a GPU-based SN transport sweep algorithm combined with a GPU-parallel multi-group Krylov

In this work, the source-terms such as nuclide inventories, radioactivities, neutron and gamma spectra, and decay heats were evaluated using new code BESNA in which the burnup dependent one-group cross sections were evaluated based on MCNP6. This work started with the selection of the reference spent fuels through a statistical analysis of all the accumulated PWR spent fuels in South Korea for safety analysis of a geological repository, which led to the two reference spent fuels for each of 16x1

With South Korea increasingly focusing on nuclear energy, the management of spent nuclear fuel has attracted considerable attention in South Korea. This study established a novel procedure for selecting safety-relevant radionuclides for long-term safety assessments of a deep geological repository in South Korea. Statistical evaluations were performed to identify the design basis reference spent nuclear fuels and evaluate the source term for up to one million years. Safety-relevant radionuclides

This paper gives an extensive analysis on the characterization of xenon isotopic ratios for various nuclear reactors and nuclear explosions through neutronic depletion codes. The results of the characterization can be used for discriminating the sources of the xenon isotopes’ release among the nuclear explosions and nuclear reactors. The considered sources of the xenon radionuclides do not only include PWR, CANDU, and nuclear explosions using uranium and plutonium bombs, but also IRT-200 and 5MW

This study presents a novel method for estimating the masses of uranium-235 and total plutonium in spent nuclear fuel assemblies using only Differential Die-Away (DDA) analysis results and cooling time without relying on passive or delayed neutron measurements. A database of DDA signals was generated through MCNP6 simulations based on spent nuclear fuel compositions derived from ORIGAMI depletion calculations under a wide range of initial enrichments, burnups, and cooling times. The correlations

A new burnable absorber called GdN-CBA with Cr-coating is applied to the i-SMR (Innovative Small Modular Reactor) for boron-free operation. The burnable absorber considered in this study employs a gadolinium nitride coating burnable absorber with a 20 μm chromium protective layer to preserve coating integrity and reactivity control over long cycles. Furthermore, an optimized control rod (CR) pattern consisting of 20 Inconel-625 fingers and 4 Ag-In-Cd fingers (24 fingers in total), is introduced