Deokjung Lee

Ulsan National Institute of Science and Technology | UNIST · Department of Nuclear Engineering

This work introduces the novel methods and performance of GREAPMC (GPU-optimized REActor Physics Monte Carlo), a GPU-accelerated Multigroup Monte Carlo (MC) code designed specifically for pressurized water reactor (PWR) simulations. The code incorporates two distinct approaches to optimize history-based neutron tracking for GPUs. The first novel ap...

This paper presents a detailed procedure for implementing the inverted fuel geometry to a fast reactor to improve its safety system and economy. The study is starting from the fuel unit cell, fuel assembly, reactor core, and burnup analysis. A proposed multivariable graph (, , –,) introduced at the fuel unit cell level provides comprehensive therma...

The objective of this research is to develop a deep learning (DL) surrogate model tailored to replace the fuel performance module within the Multi-Physics Core (MPCORE) coupling code developed at Ulsan National Institute of Science and Technology (UNIST)

UNIST has commenced the development of a new GPU-optimized MC code GREAPMC (Gpu-optimzed REActor Physics Monte Carlo) in CUDA C++. This paper presents algorithmic choices, initial results, and future directions for the further development of GREAPMC

The on-the-fly energy release per fission (OTFK) model is implemented in STREAM to continuously update the Kappa values during the depletion calculation. The explicit neutron and photon energy distribution, which has not been considered in previous STREAM versions, is incorporated into the existing on-the-fly model. The impacts of the modified OTFK...

This paper presents a performance analysis of a commercial pressurized water reactor (PWR) core loaded with accident-tolerant fuel (ATF) containing Mo/Cr metallic microcell UO2 pellets and a CrAl coating on the cladding. Metallic microcell UO2 pellets using additive materials, such as Mo and Cr, with coated cladding have been developed by the Korea...

This paper presents the steady state analysis of the Rostov-II benchmark using the conventional two-step approach. It involves the STREAM/RAST-K and CASMO-5/PARCS code systems. This paper documents a comprehensive code-to-code comparison between Serpent 2, CASMO-5, and STREAM at the lattice level for the different fuel assemblies (FAs) loaded in th...

Under the framework of coordinated research activities of the International Atomic Energy Agency (IAEA), the China Institute of Atomic Energy (CIAE) proposed a coordinated research project (CRP) to develop a benchmark based on the start-up tests of the China Experimental Fast Reactor (CEFR). 29 international organizations from 17 countries are part...

The China Institute of Atomic Energy (CIAE) proposed some of the China Experimental Fast Reactor (CEFR) neutronics start-up test data for the IAEA benchmark within the scope of the IAEA’s coordinated research activity. The coordinated research project (CRP) on “Neutronics Benchmark of CEFR Start-Up Tests” was launched in 2018. The benchmark aims to...

A coordinated research project (CRP) entitled "Neutronics Benchmark of CEFR Start-up Tests" has been conducted for improvement of analytical capabilities of fast reactor modelling and simulations. Among the six experiments and two numerical benchmarks, two benchmarks of reaction rate measurements from foil activations and the integral reactivity co...

This paper presents the development of a nodal diffusion code, RAST-V, and its verification and validation for VVER (vodo–vodyanoi energetichesky reactor) analysis. A VVER analytic solver has been implemented in an in-house nodal diffusion code, RAST-K. The new RAST-K version, RAST-V, uses the triangle-based polynomial expansion nodal method. The R...

The propagation of radiation source uncertainties in spent nuclear fuel (SNF) cask shielding calculations is presented in this paper. The uncertainty propagation employs the depletion and source term outputs of the deterministic code STREAM as input to the transport simulation of the Monte Carlo (MC) codes MCS and MCNP6. The uncertainties of dose r...

This study proposes a novel deterministic discrete ordinate method (SN)/method of characteristics (MOC) collaborative method for few-group cross-section generation to realize versatile spectral computation for different types of neutron energy spectra. The method encompasses fast/intermediate/thermal spectra, which facilitates advanced reactor phys...

The MCS code is a computer code developed by the Ulsan National Institute of Science and Technology (UNIST) for simulation and calculation of nuclear reactor systems based on the Monte Carlo method. The code is currently used to solve two main types of reactor physics problems, namely, criticality problems and radiation shielding problems. In this...

In this study, we incorporate an anisotropic scattering scheme involving spherical harmonics into the method of characteristics (MOC). The neutron transport solution in a light water reactor can be significantly improved because of the impact of an anisotropic scattering source with the MOC flat source approximation. Several problems are selected t...

The China Experimental Fast Reactor (CEFR) is a small, sodium-cooled fast reactor with 20 MW(e) of power. Start-up tests of the CEFR were performed from 2010 to 2011. The China Institute of Atomic Energy made some of the neutronics start-up-test data available to the International Atomic Energy Agency (IAEA) as part of an international neutronics b...

STREAM - a lattice transport calculation code with method of characteristics for the purpose of light water reactor analysis - has been developed by the Computational Reactor Physics and Experiment laboratory (CORE) of the Ulsan National Institute of Science and Technology (UNIST). Recently, efforts have been taken to develop a photon module in STR...

CEFR is a small core-size sodium-cooled fast reactor (SFR) using high enrichment fuel with stainless-steel reflectors, which brings a significant challenge to the deterministic methodologies due to the strong spectral effect. The neutronic simulation of the start-up experiments conducted at the CEFR have been performed with a deterministic code sys...

The present manuscript documents the development of a practical calculation scheme to model the response of a diamond detector in the mixed (neutron and gamma) radiation field of the CROCUS experimental reactor at EPFL. The model is shown to perform reasonably well in terms of energy spectra shapes for the limited amounts of irradiations considered...

The pin-based pointwise energy slowing-down method (PSM), which is a resonance self-shielding method, has been refined to treat the nonuniformity of material compositions and temperature profile in the fuel pellet by calculating the exact collision probability in the radially subdivided fuel pellet under the isolated system. The PSM has generated t...

The pin-based pointwise energy slowing-down method (PSM) has been refined through eliminating the approximation for using the pre-tabulated collision probability during the slowing-down calculation. A collision probability table is generated by assuming that material composition and temperature are constant in the fuel pellet using the collision pr...

RAST-F is a new full-core analysis code based on the two-step approach that couples a multi-group cross-section generation Monte-Carlo code MCS and a multi-group nodal diffusion solver. To demonstrate the feasibility of using MCS/RAST-F for fast reactor analysis, this paper presents the coupled nodal code verification results for the MET-1000 and C...

This paper presents a study of uncertainty quantification (UQ) of spent nuclear fuel isotope inventory. Takahama-3 realistic assays were used for calculation to quantify the impact of three different parameter groups, namely, design parameters (i.e., fuel pellet radius, clad outer radius, UO2 density, and UO2 enrichment), operation conditions (i.e....

This work presents a new application of decay heat measurement based on the calibration and inverse uncertainty quantification (IUQ) of modeling parameters of pressurized water reactor (PWR) fuel assemblies. This work (i) solves the problem encountered in forward UQ i.e., the lack of fuel vendor proprietary information (manufacturing tolerances of...

Measured decay heat data of light water reactor (LWR) spent nuclear fuel (SNF) assemblies are adopted to train machine learning (ML) models. The measured data is available for fuel assemblies irradiated in commercial reactors operated in the United States and Sweden. The data comes from calorimetric measurements of discharged pressurized water reac...

A core design of MicroURANUS, a long‐cycle lead‐bismuth‐cooled fast nuclear reactor for marine applications, is presented. It aims to generate a power of 60MWth, which can be regulated during operation. MicroURANUS was designed to achieve a small burnup reactivity swing for 30 effective full‐power years of a lifetime without refueling. To attain th...

This paper addresses the uncertainty quantification and sensitivity analysis of a depleted light-water fuel assembly of the Turkey Point-3 benchmark. The uncertainty of the fuel assembly decay heat and isotopic densities is quantified with respect to three different groups of diverse parameters: nuclear data, assembly design, and reactor core opera...

Multigroup cross section (MG XS) generation by the UNIST in-house Monte Carlo (MC) code MCS for fast reactor analysis using nodal diffusion codes is reported. The feasibility of the approach is quantified for two sodium fast reactors (SFRs) specified in the OECD/NEA SFR benchmark: a 1000 MWth metal-fueled SFR (MET-1000) and a 3600 MWth oxide-fueled...

This paper presents a methodology developed and implemented in the neutron transport code STREAM to perform high-fidelity, multicycle, multiphysics simulations of light water reactor whole-core problems. STREAM uses state-of-the-art methodologies to achieve high accuracy and computational performance. Further, it can compute fine-mesh neutron trans...

This paper presents a verification and validation (V&V) study of the Monte Carlo neutron transport code MCS for the criticality analysis of LWR fuel in transportation and storage casks/packages. A total of 173 critical experiments from the NUREG/CR-6361 report are analyzed with MCS and the ENDF/B-VII.1 cross section library. A preliminary verificat...

This paper presents the verification and validation (V&V) of a calculation module for isotope inventory prediction to control the back-end cycle of spent nuclear fuel (SNF). The calculation method presented herein was implemented in a two-step code system of a lattice code STREAM and a nodal diffusion code RAST-K. STREAM generates a cross section a...

In the BEPU (Best Estimate Plus Uncertainty) framework, uncertainty quantification (UQ) is a requirement to improve confidence and reliability of code predictions. Over the years, a lot of works have been done to quantify uncertainties in code predictions of spent nuclear fuel (SNF) characteristics due to nuclear data uncertainties. The purpose of...

Recent researches have become more interested in the feasibility of using Monte Carlo (MC) code to generate multi-group (MG) cross sections (XSs) for fast reactor analysis using nodal diffusion codes. The current study, therefore, presents a brief methodology for MG XSs generation by the in-house UNIST MC code MCS, which can be compatibly utilized...

Verification and validation (V&V) results of source term calculation capability implemented in the nodal diffusion code RAST-K are presented in this paper. An isotope inventory prediction method is presented in this work which is implemented with RAST-K and the lattice code STREAM. STREAM generates cross-section and provides number density informat...

Uncertainty analysis in Modelling (UAM) for Design, Operation and Safety Analysis of Sodium-cooled Fast Reactors (SFRs) has been formed by OECD/NEA to assess the effect of nuclear data uncertainties on parameters of interest in SFR analysis. In this paper, sub-exercises of a medium 1000 MWth metallic core (MET-1000) and a large 3600 MWth oxide core...

Decay heat (DH) is the heat produced through a radioactive decay of fission products during or after a reactor operation. It is known as the second largest source of power in the core after fission. Being such a strong contributor to reactor power, it should be accurately determined at any time of reactor operation. Currently, there are two main ap...

China Experimental Fast Reactor (CEFR) is a small size sodium-cooled fast reactor (SFR) with a high neutron leakage core fueled by uranium oxide. The CEFR core with 20 MW(e) power reached its first criticality in July 2010, and several start-up tests were conducted from 2010 to 2011. The China Institute of Atomic Energy (CIAE) proposed to release s...

A new reactor core multi-physics system addresses the pellet-to-cladding heat transfer modeling to improve full-core operational transient and accident simulation used for assessment of reactor core nuclear safety. The rigorous modeling of the heat transfer phenomena involves strong interaction between neutron kinetics, thermal-hydraulics and nucle...

The flexible 3D porous structure with a large surface area provides pathways for rapid ion/electron transport and ion diffusion as well as numerous electroactive sites.The wire-shaped supercapacitor exhibits a high energy density of 153.3 Wh kg−1 and a power density of 8810 W kg−1.The hybrid device demonstrates excellent durability under various me...

The International Atomic Energy Agency and the China Institute of Atomic Energy proposed a coordinated research project (CRP) to establish a benchmark, based on the China Experimental Fast Reactor (CEFR) start-up tests which include fuel loading and criticality, measurements of control rod worth and reactivity coefficients, and foil activation meas...

The RAST-K v2, a novel nodal diffusion code, was developed at the Ulsan National Institute of Science and Technology (UNIST) for designing the cores of pressurized water reactors (PWR) and performing analyses with high accuracy and computational performance by adopting state-of-the-art calculation models and various engineering features. It is a th...

The paper deals with the adaptation and application of the data compression algorithms to reactor core multiphysics analysis. Large capacity makes the storage, transfer and processing of multiphysics data cumbersome and expensive. The general purpose lossy compression methods are not able to guarantee the accuracy of compressed information with res...

The criticality analysis of VVER-1000 mock-up benchmark experiments from the LR-0 research reactor operated by the Research Center Rez in the Czech Republic has been conducted with the MCS Monte Carlo code developed at the Computational Reactor Physics and Experiment laboratory of the Ulsan National Institute of Science and Technology. The main pur...

In this paper, we validate the decay heat calculation capability via a two-step method to analyze spent nuclear fuel (SNF) discharged from pressurized water reactors (PWRs). The calculation method is implemented with a lattice code STREAM and a nodal diffusion code RAST-K. One of the features of this method is the direct consideration of three-dime...

Uncertainties are calculated for pressurized water reactor (PWR) spent nuclear fuel (SNF) characteristics. The deterministic code STREAM is currently being used as SNF analysis tool to obtain isotopic inventory, radioactivity, decay heat, neutron and gamma source strengths. The SNF analysis capability of STREAM was recently validated. However, the...

This paper presents the validation of the MCS code for critical safety analysis with burnup credit for the spent 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 Safety Benchmark Evaluation Project (ICSBE...

This paper presents the verification and validation elements of the UNIST in-house Monte Carlo code, MCS, for the multi-cycle and multi-physics analyses of high-fidelity, large-scale commercial pressurized water reactors (PWRs). Analysis on the neutronic performance with thermal/hydraulic (T/H) feedback is the key to detecting the complex behavior...

Many light-water reactors use self-powered neutron detectors (SPNDs) to monitor neutron flux. Some SPNDs suffer from delayed response. To compensate for this inherent delay, many methods such as the direct inversion method (DIM) and the Kalman filter method can be used. In the current investigation, these two methods were applied to the most common...

This paper presents a new conceptual design of soluble‐boron‐free small modular pressurized water reactor (SMPWR) core with the following singular features: long operation cycle, axially heterogeneous adjuster control rods, and ring‐type burnable absorbers (R‐BAs) coated on the outside of cladding materials. The core loads 37 Westinghouse‐type 17 ×...

A new three-dimensional (3D) transport analysis method is developed and implemented in the light water reactor (LWR) whole-core analysis code STREAM. The method is named as 3D method of characteristics/Diamond-difference (MOC/DD). In this method, 3D variables such as flux and source are expressed as a combination of the two-dimensional (2D) radial...

This paper presents a verification and validation (V&V) study of the MCS code for the multi-physics simulation of pressurized water reactor (PWR) multi-cycle operation. MCS is a Monte Carlo (MC) neutron-photon transport code developed by the COmputational Reactor physics and Experiment (CORE) group of Ulsan National Institute of Science and Technol...

The theoretical aspects behind the reactor depletion capability of the Monte Carlo code MCS developed at the Ulsan National Institute of Science and Technology (UNIST) and practical results of this depletion feature for a Material-Testing Reactor (MTR) with plate-type fuel are described in this paper. A verification of MCS results is first performe...

In this work, the depletion capability implemented in Monte Carlo code MCS is investigated, to predict the isotopic compositions of spent nuclear fuel (SNF). By comparison of MCS calculation results to post irradiation examination (PIE) data obtained from one pressurized water reactor (PWR), the validation of this capability is conducted. The deple...

A new Monte Carlo (MC) neutron/photon transport code, called MCS, has been developed at Ulsan National Institute of Science and Technology (UNIST) with the aim of performing the high-fidelity multi-physics simulation of large-scale power reactors, especially pressurized water reactors (PWR). The high-fidelity multi-physics analysis of large-scale P...

The shielding analysis of the spent fuel dry storage cask TN‐32 is carried out using the continuous‐energy Monte Carlo neutron‐ and photon‐transport code MCS developed by the Computational Reactor Physics and Experiment laboratory of the Ulsan National Institute of Science and Technology. The shielding analysis involves the transport simulation of...

A review of the documentation and an interpretation of the NEA-1517/74 and NEA-1517/80 shielding benchmarks (measurements of photon leakage flux from a hollow sphere with a central 14 MeV neutron source) from the SINBAD database with the Monte Carlo code MCS and the most up-to-date ENDF/B-VIII.0 neutron data library are conducted. The two analyzed...

The coupled neutronics-thermal-hydraulic simulation of the Benchmark for Evaluation and Validation of Reactor Simulations (BEAVRS) Cycle 1 depletion has been performed by the Monte Carlo-based multiphysics coupling code system MCS/CTF. MCS/CTF is a cyclewise pi-card iteration-based inner-coupling code system that couples the subchannel thermal-hydr...

A fuel performance (FP) analysis of the BEAVRS (Benchmark for Evaluation and Validation of Reactor Simulations) benchmark Cycle 1 depletion is performed using the MCS/FRAPCON coupled code system. MCS/FRAPCON is a cycle-wise Picard-iteration inner-coupling code system. It is based on the Monte Carlo neutron-transport code MCS and employs the steady-...

This paper presents the verification of the DeCART2D/CAPP code system for the Very High Temperature Gas-Cooled Reactor (VHTR) analysis with the Prismatic Modular Reactor 200 (PMR-200) benchmark. The McCARD Monte Carlo (MC) code is used to obtain the reference solution. The verification has been performed for the effective multiplication factor (kef...

This paper presents the validation of UNIST in-house Monte Carlo code MCS used for the high-fidelity simulation of commercial pressurized water reactors (PWRs). Its focus is on the accurate, spatially detailed neutronic analyses of startup physics tests for the initial core of the Watts Bar Nuclear 1 reactor, which is a vital step in evaluating cor...

The paper describes the source term estimation of CROCUS, the zero power research reactor of EPFL, to be used for dispersion analysis under accidental conditions. To fulfil regulatory requirements, the source term of the CROCUS fuel is estimated through Monte Carlo simulations supplemented by uncertainty quantification, both obtained from the Monte...

This paper presents an improved uncertainty quantification technique for the validation of CASMO-5 on the spent fuel reactivity worth experiments of the LWR-PROTEUS Phase II program. In the program, eleven spent fuel samples manufactured from rods irradiated in a Swiss PWR (discharge burnups of 20–120 MWd/kg) were measured in the PROTEUS research r...

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 pr...

This study presents the development of a new integral-type rack design, characterized by the use of gadolinium (Gd)-containing structure materials that enhances the capacity of spent fuel (SF) pool storage by exploiting the high neutron absorption capability of Gd. Appropriate types and contents of Gd-based neutron-absorbing materials are selected...

Until now, various studies focusing on subcriticality measurements have been conducted using the Rossi-alpha method. However, no guidelines have been provided for the source intensity and time bin size of the Rossi-alpha analysis. In this study, sensitivity analyses were performed to determine an optimized source intensity and time bin size for a n...

This paper presents the performance analysis of two random sampling algorithms, the inverse-transform method and the Vose aliasing method, on GNU Octave. The Monte Carlo code MCS developed by UNIST uses random sampling methods to simulate the physics of neutron and photon transport [1]. The goal is to optimize the sampling time of MCS for cases whe...

The deterministic MOC code STREAM of the Computational Reactor Physics and Experiment (CORE) laboratory of Ulsan National Institute of Science and Technology (UNIST), was initially designed for the calculation of pressurized water reactor two- and three-dimensional assemblies and cores. Since fast reactors play an important role in the generation-I...

A deterministic MOC code STREAM of the Computational Reactor Physics and Experiment (CORE) laboratory of Ulsan National Institute of Science and Technology (UNIST), was initially designed for the calculation of pressurized water reactor two- and three- dimensional assemblies and cores. Since fast reactors play an important role in the generation-IV...

Reduced-moderation pressurized water reactor (RMPWR) utilised thorium (Th) as the fuel in exchange for uranium (U) due to its huge advantages in term of neutronics characteristics, including low void and moderator temperature coefficient, and the potential of breeding into fissile fuel 233 U. In RMPWR, the water to fuel ratio is reduced to 50% of t...

The Kartini Triga Mark II Research Reactor (Kartini TM2RR), one of the three research reactors in Indonesia, is a pool-type reactor with a thermal power of 100 kW that has been operating since 1979 and has legal operation licensing until 2019 from the regulatory body. As of 2018, the reactor is composed of 71 cylindrical fuel elements of height 66...

A Monte-Carlo neutronics/thermal-hydraulics/fuel-performance (N/TH/FP) multi-physics coupling system has been developed based on the MCS code recently for the purpose of large-scale high-fidelity analysis of light water reactors (LWRs). The full N/TH/FP coupling overcomes the drawbacks of the previous N/TH (MCS/CTF) and N/FP (MCS/FRAPCON) coupling...

The authors developed the modified power method (MPM) in previous publications to obtain multiple eigenmodes of an eigenvalue problem at the same time by employing a generalized eigenvalue problem (GEP) of the form of WX = VXK. Special attention has been paid to the Monte Carlo (MC) implementation of the MPM because it always suffers from the inher...

The steady-state fuel behavior prediction code FRAPCON has been coupled with the Monte Carlo code MCS to accomplish fuel performance analysis capability. The Monte Carlo based multi-physics coupling analysis for large-scale light water reactors (LWRs) with high fidelity has mostly focused on the inner coupling of the Monte Carlo neutronics analysis...

This paper presents the verification and validation of the radiation source term calculation capability implemented in the reactor analysis code STREAM for pressurized water reactor (PWR) spent nuclear fuel (SNF) analysis. Activity, decay heat, neutron and gamma source spectra of irradiated PWR fuel assemblies are calculated with STREAM and ORIGEN...

This paper presents a comparative analysis of the VERA depletion benchmark through consistent code-to-code comparisons between four neutronics analysis codes. An optimization of depletion calculation methods has been performed through an extensive sensitivity study in terms of nuclear transmutation equation solution methods and numerical calculatio...

RXSP is a nuclear cross section processing code, which is being developed by Tsinghua University, mainly maintained by joint collaboration with Tsinghua University and Ulsan National Institute of Science and Technology. Recently, RXSP extends its capability of multi-group covariance matrices processing from correlation matrix data stored in Evaluat...

Monte Carlo neutron transport based multi-physics coupling system has been developed rapidly recently, the advanced sub-channel thermal/hydraulic code-CTF from CASL project has been coupled with UNIST in-house Monte Carlo simulator-MCS code [1], which make it practical to analyze PWRs and BWRs with very high fidelity up to directly whole core pin-b...