Yi-Kang Lee - Academia.edu (original) (raw)

Papers by Yi-Kang Lee

In October 2010, a series of benchmark experiments were conducted at the French Commissariat à l'... more In October 2010, a series of benchmark experiments were conducted at the French Commissariat à l'Énergie Atomique et aux Énergies Alternatives (CEA) Valduc SILENE facility. These experiments were a joint effort between the United States Department of Energy Nuclear Criticality Safety Program and the CEA. The purpose of these experiments was to create three benchmarks for the verification and validation

Fusion Engineering and Design, 2016

With the growing interest in using the continuous-energy TRIPOLI-4 ® Monte Carlo radiation transp... more With the growing interest in using the continuous-energy TRIPOLI-4 ® Monte Carlo radiation transport code for ITER applications, a key issue that arises is whether or not the released TRIPOLI-4 code and its associated nuclear data libraries are verified and validated for the D-T fusion neutronics calculations. Previous published benchmark results of TRIPOLI-4 code on the ITER related activities have concentrated on the first wall loading, the reactor dosimetry, the nuclear heating, and the tritium breeding ratio. To enhance the TRIPOLI-4 verification and validation on neutron-gamma coupled calculations for fusion device application, the computational ITER shielding benchmark of M. E. Sawan was performed in this work by using the 2013 released TRIPOLI-4.9S code and the associated CEA-V5.1.1 data library. First wall, blanket, vacuum vessel and toroidal field magnet of the inboard and outboard components were fully modelled in this 1-D toroidal cylindrical benchmark. The 14.1 MeV source neutrons were sampled from a uniform isotropic distribution in the plasma zone. Nuclear responses including neutron and gamma fluxes, nuclear heating, and material damage indicator were benchmarked against previous published results. The capabilities of the TRIPOLI-4 code on the evaluation of above physics parameters were presented. The nuclear data library from the new FENDL-3.0 evaluation was also benchmarked against the CEA-V5.1.1 results for the neutron transport calculations. In general, relevant benchmark results were obtained. Both data libraries can thus be run with TRIPOLI-4 for the fusion neutronics study. This work also demonstrates that the "safety factors" concept is necessary in the nuclear analyses of ITER.

Journal of Nuclear Engineering and Radiation Science, 2020

The ICRP 110 adult male and female voxel phantoms are the official computational models represent... more The ICRP 110 adult male and female voxel phantoms are the official computational models representing the ICRP Reference Male and Reference Female. In 2018 the Working Group 6 (WG6) of European Radiation Dosimetry Group (EURADOS) organized a study on the usage of the ICRP voxel reference phantoms. Organ dose calculation tasks with radiation transport codes were proposed in occupational and environmental dosimetry for nuclear engineering application. The TRIPOLI-4 Monte Carlo radiation transport code has been widely used in radiation shielding, criticality safety and reactor physics fields for supporting French nuclear energy research and industrial applications. To enhance the application fields of TRIPOLI-4, the 2018 EURADOS-WG6 voxel phantom tasks are being taken into account by using different features of TRIPOLI-4 code. In this work, the ICRP reference voxel phantoms were first adapted into TRIPOLI-4. More than 14 million voxels were represented in a mixed lattice geometry includ...

The Proceedings of the International Conference on Nuclear Engineering (ICONE), 2019

Fusion Engineering and Design, 2018

The International Thermonuclear Experimental Reactor (ITER) is currently under construction in Fr... more The International Thermonuclear Experimental Reactor (ITER) is currently under construction in France. From the radiation shielding and radiation protection points of view, an intense and large neutron source with energy around 14.1 MeV will be generated from the D-T plasma zone during the ITER operation and diverse gamma-ray sources from neutron activation of the reactor structure materials and coolant should be considered for the reactor operation and maintenance. To decrease the radiation impacts caused by these neutron and gamma sources, iterative designs and nuclear analyses of ITER components are currently performed with three dimensional Monte Carlo radiation transport calculations. Due to the important dimensions of ITER, the thick tokamak blanket modules, and the diagnostic and functional port openings, variance reduction techniques are essential in these Monte Carlo neutron transport calculations. To verify the reactor components design models and to check the radiation transport calculation results, advanced graphic features of the calculation tool are also necessary. With the growing interest in using the TRIPOLI-4 ® Monte Carlo radiation transport code for ITER applications, the aim of this paper is to study the feasibility to use variance reduction features of TRIPOLI-4 code on a 3D ITER benchmark model which is a 40°toroidal segment including 5796 vol cells. The calculation results reported in this paper include the axial and radial profiles of the inboard TF coil heating and the neutron flux attenuation through the equatorial port plugs and shield. The performance of the TRIPOLI-4 graphic tool under its parallel computing mode was also evaluated.

EPJ Web of Conferences, 2017

Nuclear decommissioning takes place in several stages due to the radioactivity in the reactor str... more Nuclear decommissioning takes place in several stages due to the radioactivity in the reactor structure materials. A good estimation of the neutron activation products distributed in the reactor structure materials impacts obviously on the decommissioning planning and the low-level radioactive waste management. Continuous energy Monte-Carlo radiation transport code TRIPOLI-4 has been applied on radiation protection and shielding analyses. To enhance the TRIPOLI-4 application in nuclear decommissioning activities, both experimental and computational benchmarks are being performed. To calculate the neutron activation of the shielding and structure materials of nuclear facilities, the knowledge of 3D neutron flux map and energy spectra must be first investigated. To perform this type of neutron deep penetration calculations with the Monte Carlo transport code, variance reduction techniques are necessary in order to reduce the uncertainty of the neutron activation estimation. In this study, variance reduction options of the TRIPOLI-4 code were used on the NAIADE 1 light water shielding benchmark. This benchmark document is available from the OECD/NEA SINBAD shielding benchmark database. From this benchmark database, a simplified NAIADE 1 water shielding model was first proposed in this work in order to make the code validation easier. Determination of the fission neutron transport was performed in light water for penetration up to 50 cm for fast neutrons and up to about 180 cm for thermal neutrons. Measurement and calculation results were benchmarked. Variance reduction options and their performance were discussed and compared.

Input Listings-Two-Step Method (US) In order to use the IRDF-2002 cross sections, they must be do... more Input Listings-Two-Step Method (US) In order to use the IRDF-2002 cross sections, they must be downloaded from the IAEA at https://wwwnds.iaea.org/irdf2002/index.htmlx. The cross sections should be downloaded in ACE format in order to use them with MCNP6. The MCNP6 input files presented in this appendix expect the IRDF-2002 cross sections to be in the same directory as the input file and to be named irdf-2002. A.1.1 MCNP6 Eigenvalue Input-The first step of the two-step method is to run an eigenvalue calculation that tallies the spatial and energy dependent distribution of fission neutrons. The MCNP6 input for this simulation is mcnp\source\fineMesh\pulse2. This input contains a cylindrical mesh tally to calculate the spatial distribution of fission neutrons. The results of the mesh tally used in the next step are based on 120,000 histories per batch, 19,070 total batches, and 30 skipped batches. Also included is an MCNP6 input with a PTRAC card, mcnp\source\chi.1000\pulse2, which was used to write the starting energy of all source neutrons. The energies of the source particles saved in the PTRAC file were used to create the energy distribution of fission neutrons. This input uses 5,000 histories per batch, 9,550 total batches, and 30 skipped batches. However, data begin to be written to the PTRAC file after the second batch. Since MCNP cannot run in parallel when writing a PTRAC file, 400 instances of this simulation were run, each using a unique starting random number seed and stride. Therefore, the final analysis of the PTRAC data included 3,819,200 batches of 5,000 particles each. A.1.2 MCNP6 Fixed-Source Input-The second step of the two-step method is to run a fixed-source calculation that tallies all detector responses. In this case the detector responses are neutron activation of foils and absorbed dose deposited in TLDs. The tally of fission neutrons from the first step is converted to a source for the second step. Weight windows were generated for the second step MCNP6 simulations using the ADVANTG a code. The weight window input files (wwinp) are included, and can be used directly in order to skip repeating the ADVANTG calculations. The inputs for the ADVANTG calculations are included in Appendix B. In order to make the MCNP6 simulations more efficient, a separate MCNP6 simulation for collimator A, collimator B, the free field location, and each of the four positions within the scattering box was run. Each MCNP6 simulation had its one set of weight windows and biased source, and therefore its own ADVANTG input. Links to the MCNP6 input files and weight window files are listed below in Table A-1. In each MCNP6 fixed-source simulation 400,000,000 histories were simulated.

Fusion Engineering and Design, 2014

3D Monte Carlo transport codes are extensively used in neutronic analysis, especially in radiatio... more 3D Monte Carlo transport codes are extensively used in neutronic analysis, especially in radiation protection and shielding analyses for fission and fusion reactors. TRIPOLI-4 ® is a Monte Carlo code developed by CEA. The aim of this paper is to show its capability to model a large-scale fusion reactor with complex neutron source and geometry. A benchmark between MCNP5 and TRIPOLI-4 ® , on the ITER A-lite model was carried out; neutron flux, nuclear heating in the blankets and tritium production rate in the European TBMs were evaluated and compared. The methodology to build the TRIPOLI-4 ® A-lite model is based on MCAM and the MCNP A-lite model. Simplified TBMs, from KIT, were integrated in the equatorial-port. A good agreement between MCNP and TRIPOLI-4 ® is shown; discrepancies are mainly included in the statistical error.

Nuclear Science and Engineering

Fusion Engineering and Design, 2011

a b s t r a c t 3D Monte Carlo (MC) transport codes are of first importance for the assessment of... more a b s t r a c t 3D Monte Carlo (MC) transport codes are of first importance for the assessment of breeding blankets neutronic performances. This article supported by the EFDA Goal Oriented Training Program Eurobreed presents the difference in results between the CEA MC code TRIPOLI-4 and MCNP on two fusion neutronics benchmarks, assessing therefore TRIPOLI-4 calculation capabilities on shielding and tritium production rate (TPR). The first selected benchmark, assessing the shielding capability, is the Frascati neutron generator (FNG) ITER bulk shield experiment whereas the second benchmark, assessing the TPR calculation, is the preliminary design of the FNG helium cooled lithium-lead (HCLL) test blanket module (TBM) mock-up. To ensure the consistency of the geometry description, MCAM tool is used for automatic TRIPOLI − MCNP geometry conversions and check. A good coherence between TRIPOLI-4 and MCNP for neutron flux, reaction rates and TPR calculations is obtained. Moreover, it appears that MCAM performs fast, automatic and appropriate TRIPOLI − MCNP geometry conversions and finally that the tabulated FNG neutron source model from KIT is appropriate for TRIPOLI-4 calculations.

Progress in Nuclear Science and Technology, 2011

CRISTAL criticality safety package new version, CRISTALV 1.0, has been developed by the Institut ... more CRISTAL criticality safety package new version, CRISTALV 1.0, has been developed by the Institut de Radioprotection et de Sûreté Nucléaire and the Commissariat à l'Energie Atomique, in collaboration with the COGEMA company, mainly to allow taking into account burn-up credit in criticality safety studies. Meanwhile, the validation database, which was made up of about 500 benchmarks for CRISTAL V0 package, was extended. CRISTAL includes two calculation routes, using nuclear data taken from the JEF2.2 library: a standard route based on multi-group cross-sections (APOLLO2 -MORET 4 or APOLLO2 Sn calculations) and a reference route using the pointwise data (TRIPOLI4 calculations). The CRISTAL V1.0 validation database is made up of 2132 critical experiments mainly taken from the OECD/ICSBEP Handbook or performed in French facilities. Most of them have already been investigated with the different CRISTAL routes. The first trends of the validation work point out that the calculation results are generally in good agreement with the benchmark k eff and that the new calculation schemes improve the validation results.

CRISTAL V0.1, a French calculation package for criticality-safety studies, has been developed and... more CRISTAL V0.1, a French calculation package for criticality-safety studies, has been developed and validated as part of a joint project between IRSN, CEA and COGEMA. This package includes two calculation routes: on the one hand, the "standard route" dealing with the nuclear data library CEA93 (derived from JEF2.2 evaluation), the APOLLO2 cell code and the Monte Carlo MORET4 code, and, on the other hand, the "reference route" using the Monte Carlo TRIPOLI4 code with a JEF2.2 continuous-energy cross-section library. Last years, an extensive validation work has been performed by CEA and IRSN using a large experimental database (more than 500 critical experiments) taking into account most of the different operations encountered in the nuclear fuel cycle. In this framework, a significant effort has been devoted to provide CRISTAL's users a comprehensive and useful synthesis of the different validation studies.

This paper will focus on the calculations of the HTR-10's first criticality. Continuous energ... more This paper will focus on the calculations of the HTR-10's first criticality. Continuous energy Monte Carlo transport code TRIPOLI-4.3, was used for getting the initial fuel loading of the HTR-10. The calculations have been performed on the basis of different treatment of the particles and the pebbles arrangement. The stochastic distribution of the particles in the fuel zone has been simulated, either using point-wise cross-sections associated with cubic or hexagonal lattices or using multi-group cross-sections provided by 1-D spherical transport calculation (APOLLO2 code). Different types of pebble arrangements have also been compared. Cross comparisons of the different hypotheses have been analysed including the impact of the libraries (JEF2 and ENDF/B-VI) and the estimation of the certain homogenisation effects. Therefore, this paper gives information related to the treatment of stochastic geometry in Pebble Bed Reactor (PBR) with the Monte Carlo method.

Power distribution calculations are essential for fuel assembly design and whole core safety anal... more Power distribution calculations are essential for fuel assembly design and whole core safety analysis. In Monte Carlo core power map calculations, both lattice geometry and well-adapted tally functions are necessary. The lattice geometry and lattice tally features of TRIPOLI-4.3 Monte Carlo code for local power map calculation were reported in previous studies. To tally the power distribution on all cells of a whole core in pin-by-pin level, mesh tally can be more efficient on CPU time and more user-friendly on tally input preparation. This paper using the new introduced mesh tally function of TRIPOLI-4.5 interprets the whole core power maps of three PWR critical lattice experiments from LEU-COMP-THERM-008 benchmark. The multiplication factors Keff were calculated by using recent nuclear data libraries, ENDF/B-VII.0 and JEFF3.1, and compared with previous ones using ENDF/B-VI.4 and JEF2.2. The measured pin power distributions of 1/8 central assembly were benchmarked against calculat...

In this updated overview (cf. [1]) of the Monte Carlo transport code TRIPOLI-4, we list and descr... more In this updated overview (cf. [1]) of the Monte Carlo transport code TRIPOLI-4, we list and describe its current main features. The code computes coupled neutron-photon propagation as well as the electron-photon cascade shower. While providing the user with common biasing techniques, it also implements an automatic weighting scheme. TRIPOLI-4 has support for execution in parallel mode. Special features and applications are also presented.