Code MCS Research Articles

Coupled Neutronics–Thermal-Hydraulic Simulation of BEAVRS Cycle 1 Depletion by the MCS/CTF Code System

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-hydraulic code CTF as a thermal-hydraulic solver in the Monte Carlo neutron transport code MCS. MCS has been developed by the Computational Reactor Physics and Experiment Lab group at the Ulsan National Institute of Science and Technology for the full-core analysis of large-scale commercial light water reactors with high fidelity at the engineering level. With the high-fidelity performance of MCS, the quarter-core pinwise depletion simulation for the BEAVRS Cycle 1 benchmark has been conducted with thermal-hydraulic feedback including fuel temperature, coolant temperature, and coolant density. Moreover, the MCS internal one-dimensional thermal-hydraulic solver TH1D (MCS/TH1D) has been utilized as the reference. On one hand, the simulated results of the criticality boron concentration and axially integrated assemblywise detector signals were compared with measured data. On the other hand, the comparisons of power, fuel temperature, coolant temperature, and density are also presented in this paper.

Fuel performance analysis of BEAVRS benchmark Cycle 1 depletion with MCS/FRAPCON coupled system

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-state fuel behavior prediction code FRAPCON as a thermal-hydraulic (T/H) and FP solver. MCS is developed by the Computational Reactor Physics and Experiment Lab of Ulsan National Institute of Science and Technology for the high-fidelity full-core analysis of large-scale commercial light water reactors. Results of power, fuel temperature, coolant temperature and coolant density distributions are presented and analyzed for a quarter-core pin-wise depletion simulation of the BEAVRS Cycle 1 benchmark with T/H and FP feedback (10 axial meshes per pin, 180,870 depletion cells in total). For code-code comparison purpose, the depletion simulation is also conducted with the MCS/TH1D (internal one-dimensional T/H solver) and MCS/CTF (external sub-channel 3D T/H solver) coupled code systems. The dependence to the burnup of the power, fuel temperature, coolant temperature, and coolant density distributions is analyzed by comparison between the three coupled systems. Validation is performed against BEAVRS measured data for the calculated boron letdown curve and calculated distributions of axially-integrated assembly-wise detector signal. Finally, unique distributions of parameters that can only be obtained by FP analysis are examined to illustrate the advanced analysis capability of MCS/FRAPCON.

Validation of UNIST Monte Carlo code MCS using VERA progression problems

Abstract 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 core phenomena in an operating nuclear power reactor. The MCS solutions for the Consortium for Advanced Simulation of Light Water Reactors (CASL) Virtual Environment for Reactor Applications (VERA) core physics benchmark progression problems 1 to 5 were verified with KENO-VI and Serpent 2 solutions for geometries ranging from a single-pin cell to a full core. MCS was also validated by comparing with results of reactor zero-power physics tests in a full-core simulation. MCS exhibits an excellent consistency against the measured data with a bias of ±3 pcm at the initial criticality whole-core problem. Furthermore, MCS solutions for rod worth are consistent with measured data, and reasonable agreement is obtained for the isothermal temperature coefficient and soluble boron worth. This favorable comparison with measured parameters exhibited by MCS continues to broaden its validation basis. These results provide confidence in MCS's capability in high-fidelity calculations for practical PWR cores.

Simulations of BEAVRS benchmark cycle 2 depletion with MCS/CTF coupling system

Abstract 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 previously applied in the BEAVRS benchmark Cycle 1 full-core simulation. The Cycle 2 depletion has been performed with T/H feedback based on the spent fuel materials composition pre-generated by the Cycle 1 depletion simulation using refueling capability of MCS code. Meanwhile, the MCS internal one-dimension T/H solver (MCS/TH1D) has been also applied in the simulation as the reference. In this paper, an analysis of the detailed criticality boron concentration and the axially integrated assembly-wise detector signals will be presented and compared with measured data based on the real operating physical conditions. Moreover, the MCS/CTF simulated results for neutronics and T/H parameters will be also compared to MCS/TH1D to figure out their difference, which proves the practical application of MCS into the BEAVRS benchmark two-cycle depletion simulations.

Determination of the activity inventory and associated uncertainty quantification for the CROCUS zero power research reactor

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 Carlo code MCS developed at UNIST.Even though the depletion capabilities of MCS were pre-existing to this work, no verification has been documented so far. A comparison of MCS and SERPENT results for the determination of the activity inventory for the CROCUS fuel is presented; both codes agree within the 1% for the major isotopes contributing to both inhaled and ingested doses. The source term of CROCUS is calculated under a postulated accident. Eight isotopes, 90Sr, 91Y, 131I, 137Cs, 140Ba, 140La, 144Ce, and 239Pu produce the largest contributions to the effective dose to the public.For uncertainty quantification, nuclear data uncertainties, specifically cross sections and fission yields are considered. Three stochastic sampling methods are implemented in the MCS code namely, TMC, fast-TMC and fast-GRS methods. The performances of the two fast methods have been analyzed when applied to the CROCUS reactor. All three methods produce consistent uncertainty estimates and the two fast methods showed a reduced computational cost compared to the original TMC method. The fission yield uncertainty is the leading factor for the determination of uncertainty of the activity for 90Sr, 91Y, 131I, 137Cs, 140Ba, 140La and 144Ce isotopes. On the other hand, the cross section uncertainty is the leading factor for the uncertainty of 239Pu activity.Finally, the modeling of the irradiation history for CROCUS is simplified to reduce the computational cost. It is demonstrated that the activities for the short lived isotopes (131I and 144Ce) are very sensitive to the irradiation history specifications. Nonetheless, the irradiation history used for the determination of the nominal fuel inventory activity is conservative as it overestimates the activity of the short lived isotopes.

Application of TULIP/STREAM code in 2-D fast reactor core high-fidelity neutronic analysis

Abstract 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-IV concept, it was decided that the code should be upgraded for the analysis of fast neutron spectrum reactors. This paper presents a coupled code - TULIP/STREAM, developed for the fast reactor assembly and core calculations. The TULIP code produces self-shielded multi-group cross-sections using a one-dimensional cylindrical model. The generated cross-section library is used in the STREAM code which solves eigenvalue problems for a two-dimensional assembly and a multi-assembly whole reactor core. Multiplication factors and steady-state power distributions were compared with the reference solutions obtained by the continuous energy Monte-Carlo code MCS. With the developed code, a sensitivity study of the number of energy groups, the order of anisotropic PN scattering, and the multi-group cross-section generation model was performed on the keff and power distribution. The 2D core simulation calculations show that the TULIP/STREAM code gives a keff error smaller than 200 pcm and the root mean square errors of the pin-wise power distributions within 2%.

Performance analysis of random sampling algorithms on GNU Octave

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 when the probability density function is a constant function throughout the simulation. For this purpose, the runtime of the inverse-transform method and Vose aliasing method are compared for increasing input size with scripts developed on GNU Octave. To compare the execution time, the initialization and generation time of both methods are determined and discussed.

Coupling of FRAPCON for fuel performance analysis in the Monte Carlo code MCS

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 code and the thermal–hydraulics code. However, there are still some issues that cannot be considered precisely when predicting fuel thermal conductivity, and the gap thermal conductance between fuel pellets and cladding with the increase of burnup. Therefore, the FRAPCON has been chosen in this paper to be coupled with the MCS code to increase the accuracy of fuel temperature calculations and the corresponding fuel temperature feedback. A fixed-point iteration scheme is adopted for the coupling interface, which has been verified by a single rod case. In addition, the paper also depicts the application of the MCS/FRAPCON coupling system to the BEAVRS quarter core benchmark, by comparison with MCS internal one-dimension T/H solver — TH1D. The results clearly explain the necessity for considering the fuel performance in multi-physics coupling analysis and demonstrate the capability of the MCS/FRAPCON coupling system.

MCS based neutronics/thermal-hydraulics/fuel-performance coupling with CTF and FRAPCON

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 systems, which suffered respectively from the approximations in CTF on the fuel thermal conductivity and gap conductance, and from the approximations due to the single closed channel enthalpy model in FRAPCON. Thus, compared to the previous coupling systems, the new full coupling system benefits from the transverse cross flow between neighboring sub-channels in CTF, the burnup-dependent fuel thermal conductivity formulation in FRAPCON, and the iteratively determined fuel pellet-cladding gap thermal conductance in FRAPCON. Two applications of the full coupling systems are presented. First, a single fuel rod case is tested to verify the accuracy and efficiency of the new coupled system. Then, the simulation of the BEAVRS whole core model with three-dimensional (3D) pin-by-pin power density and T/H feedbacks exchange between different solvers is performed using the MCS based multi-physics coupling system. The obtained results demonstrate the practical capability of the Monte Carlo based steady-state multi-physics coupling code system for large-scale high-fidelity LWR analysis.

Verification of photon transport capability of UNIST Monte Carlo code MCS

A photon transport capability has been implemented and verified in the Monte Carlo code MCS of Ulsan National Institute of Science and Technology for the purpose of radiation shielding studies. The MCS photon fixed-source mode simulates the transport of photons between 1 keV and 20 MeV for all elements from hydrogen to fermium. The specific physics for the main four photo-atomic reactions (Rayleigh scattering, Compton scattering, photoelectric effect and pair production) and three secondary processes of photon production (positron–electron annihilation, atomic relaxation and electron/positron bremsstrahlung) are reviewed. Verification results against Monte Carlo codes MCNP6.1 and SERPENT2.1.29 are presented. The verification cases include the comparison of energy distributions of photon flux in an infinite medium, of spatial distributions of photon flux in a cylinder, of the spatial distribution of photon body-equivalent dose in a spent nuclear fuel transport cask, and of photon KERMA (Kinetic Energy Released per MAss) in photon detector calibration geometries. Good calculation/calculation agreement is observed overall, with some marked differences in the detailed photon flux comparison at given energy ranges traced back to differences in photon physics implementation. MCS can from now on be applied for the purpose of advanced photon studies and corresponding validation against experimental shielding benchmarks will follow in the future.

Preliminary coupling of the Thermal/Hydraulic solvers in the Monte Carlo code MCS for practical LWR analysis

A Thermal/Hydraulic (T/H) coupling interface has been developed in the Monte Carlo code MCS for coupling with different T/H solvers. The single channel T/H solver TH1D and the Sub-Channel T/H solver COBRA-TF (CTF) both have been coupled to MCS through this interface. A fixed-point iteration scheme is applied in this coupling interface to enable the T/H feedback to be fully updated for Monte Carlo neutron simulation. This paper describes in detail the single rod case to demonstrate the accuracy and efficiency of these two coupled T/H solvers, with the discussion of the results for fuel temperature, coolant temperature and coolant density distributions. On the other hand, the paper also depicts the preliminary application of MCS T/H coupling in BEAVRS whole core benchmark to demonstrate the feasibility of this coupling system for use in practical LWRs multi-physics coupling analysis.

Validation of UNIST Monte Carlo code MCS for criticality safety analysis of PWR spent fuel pool and storage cask

This paper presents the validation of the continuous-energy Monte Carlo neutron-transport code MCS with the ENDF/B-VII.0 neutron cross-section library for the criticality safety analysis of PWR spent fuel pools and storage casks. The MCS code is developed by the COmputational Reactor physics and Experiment laboratory (CORE) of Ulsan National Institute of Science and Technology (UNIST) for the analysis of Pressurized Water Reactors (PWRs) with high fidelity and high performance. The validation is conducted with 279 selected critical benchmarks from the International Criticality Safety Benchmark Evaluation Project (ICSBEP). The 279 validation cases are representative of PWR spent fuel pools and storage casks with 235U enrichment ranging from 2.35 wt% to 4.74 wt%, pin pitch ranging from 1.075 cm to 2.540 cm, moderator to fuel ratio (H/U) ranging from 0.4683 and 11.5398, Energy of the Average Lethargy causing Fission (EALF) ranging from 0.0109 eV to 1.0600 eV, without soluble boron and with soluble boron concentration ranging from 0.015 g/L to 5.030 g/L. The calculation of the effective neutron multiplication factor by MCS is validated by the comparison between experiment and calculation for the selected critical benchmarks. The Upper Safety Limit (USL) of the MCS code is established in accordance to the NUREG/CR-6698 guideline recommended by the NRC (US National Regulatory Commission). The full validation process and determination of USL based on the selected critical benchmarks was also repeated with the MCNP6.1 and SERPENT2.1.27 codes in order to compare the performances of MCS with other reactor analysis codes. This paper demonstrates the capability of the MCS code for the criticality safety analysis of PWR spent fuel pools and storage casks.

Multi-physics steady-state analysis of OECD/NEA modular high temperature gas-cooled reactor MHTGR-350

ABSTRACTThis paper presents the application results of MCS/GAMMA+ to multi-physics analysis of OECD/NEA modular high temperature gas-cooled reactor (MHTGR) benchmark Phase I Exercise 3. It is a part of international R&D efforts lead by the Next Generation Nuclear Plant (NGNP) US project to improve the neutron-physics and thermal-fluid simulation of (high temperature gas-cooled reactors) HTGRs, one of the next generations of safer nuclear reactors. Accurate and validated analysis tools are indeed a crucial requirement for safety analysis and licensing of nuclear reactors. To guide this effort, a numerical benchmark on the MHTGR was created by the NGNP project and formally approved in 2012 for international participation by the OECD/NEA. The benchmark defines a common set of exercises and the comparison of solutions obtained with different analysis tools is expected to improve the understanding of simulation methods for HTGRs. The coupled neutronics/thermal-fluid solution presented in this paper was obtained with the neutron transport Monte Carlo code MCS developed by Ulsan National Institute of Science and Technology and the thermal-fluid code GAMMA+ developed by Korean Atomic Energy Research Institute. The purpose of this paper is to present the GAMMA+/MCS coupled system, the calculation methodology, and the obtained solutions.

Play It Again: Teaching Statistics With Monte Carlo Simulation

ABSTRACTMonte Carlo simulations (MCSs) provide important information about statistical phenomena that would be impossible to assess otherwise. This article introduces MCS methods and their applications to research and statistical pedagogy using a novel software package for the R Project for Statistical Computing constructed to lessen the often steep learning curve when organizing simulation code. A primary goal of this article is to demonstrate how well-suited MCS designs are to classroom demonstrations, and how they provide a hands-on method for students to become acquainted with complex statistical concepts. In this article, essential programming aspects for writing MCS code in R are overviewed, multiple applied examples with relevant code are provided, and the benefits of using a generate–analyze–summarize coding structure over the typical “for-loop” strategy are discussed.

Assembly dynamics of PML nuclear bodies in living cells

The mammalian cell nucleus contains a variety of organelles or nuclear bodies which contribute to key nuclear functions. Promyelocytic leukemia nuclear bodies (PML NBs) are involved in the regulation of apoptosis, antiviral responses, the DNA damage response and chromatin structure, but their precise biochemical function in these nuclear pathways is unknown. One strategy to tackle this problem is to assess the biophysical properties of the component parts of these macromolecular assemblies in living cells. In this study we determined PML NB assembly dynamics by live cell imaging, combined with mathematical modeling. For the first time, dynamics of PML body formation were measured in cells lacking endogenous PML. We show that all six human nuclear PML isoforms are able to form nuclear bodies in PML negative cells. All isoforms exhibit individual exchange rates at NBs in PML positive cells but PML I, II, III and IV are static at nuclear bodies in PML negative cells, suggesting that these isoforms require additional protein partners for efficient exchange. PML V turns over at PML Nbs very slowly supporting the idea of a structural function for this isoform. We also demonstrate that SUMOylation of PML at Lysine positions K160 and/or K490 are required for nuclear body formation in vivo.We propose a model in which the isoform specific residence times of PML provide both, structural stability to function as a scaffold and flexibility to attract specific nuclear proteins for efficient biochemical reactions at the surface of nuclear bodies.MCS code: 92C37