Monte Carlo Neutron Transport Code Research Articles

Heat pipe failure accident analysis in megawatt heat pipe cooled reactor

The heat pipe cooled reactor is an advanced reactor design which is nearly solid-state with an array of in-core heat pipes for passive heat transfer. The local heat pipe failure probability is certainly high over the reactor lifetime. Therefore, the reactor design must include consideration of heat pipe failure accidents which needs to be explored in-depth. The Monte Carlo neutron transport code RMC was used with the general finite element analysis software ANSYS Mechanical to analyze the thermal/mechanical characteristics of a megawatt heat pipe cooled reactor for both normal and heat pipe failure conditions. The simulations were verified against a previous study with the predicted temperature distributions being consistent with each other with an absolute error of less than 3 K. Further analyses show that a 2-D model simulation can also well represent the 3-D model for heat pipe failure accident analyses. Sterbentz et al. (2017) simplified the heat pipe model with a constant temperature boundary condition along the evaporator wall of 950.15 K regardless of the heat load or heat pipe operating temperature. Therefore, a heat pipe subroutine was added to the codes to predict the heat pipe temperatures during a heat pipe failure accident. The results show that the heat pipe failure will cause larger temperature rises and stress concentrations. The heat pipe wall temperatures in the core differ from each other with the differences between the heat pipe loads during normal heat pipe operation and with failures reflecting the local effect of the heat pipe failures. The operating conditions far from the failure area are little affected (<20 K temperature change) by a single heat pipe failure. The effects of the heat pipe failure on the reactor core reactivity and local power distribution were also analyzed with a reactivity change of less than 5 pcm and a local power change of less than 0.1%. Therefore, the heat pipe failure can be analyzed in steady-state simulations with unchanged power distributions.

Development and verification of code IMPC-Depletion for nuclide depletion calculation

Simulation of nuclide inventory evolution in the reactor during the entire irradiation history is of significance to the design of fuel component and core layout scheme and nuclear waste disposal. To perform accurate burnup calculations for fast reactor systems such as accelerator-driven sub-critical reactors, a nuclide depletion code IMPC-Depletion has been developed. Depletion of reactor material is described by thousands of coupled first-order ordinary differential equations, which is the stiff ODEs system. For IMPC-Depletion, algorithms including Chebyshev Rational Approximation Method (CRAM) and Transmutation Trajectory Analysis method (TTA) are implemented to solve Bateman equations. The parameter-extended XML file format is used as standard input. Decay, constant flux and power modes are supported. To verify IMPC-Depletion code, decay of Ra228, fixed-flux irradiation of Zircaloy-4 and 2.4% UO2, fixed-power irradiation of 3.2% UO2 are tested. Results are basically consistent with those of ORIGEN2.1. In addition to independent nuclides inventory analysis, IMPC-Depletion can also be coupled with Monte Carlo neutron transport code to solve burnup problem. Here, simple MOX pin cell and OECD/NEA fast reactor burnup benchmarks are employed to verify the Out-coupling Monte Carlo burnup code system IMPC-Burnup2.1, which couples OpenMC and IMPC-Depletion. Result parameters such as keff, reactivity changes and nuclides inventory values are reliable when compared with other participants.

Benchmark of Serpent-2 with MCNP: Application to European DEMO HCPB breeding blanket

Abstract Magnetic confinement fusion reactors such as the European fusion demonstration reactor (EU DEMO) have a very complex geometry, which makes the modelling of the nuclear analyses rather tedious and time consuming. The Monte Carlo neutron transport code Serpent-2 developed by VTT in Finland, is able to directly import CAD geometry, then is considered an attractive alternative to the Monte Carlo N-Particle Transport code (MCNP) for its application on fusion reactors. In the first step, a comparison activity is carried out to benchmark the Serpent-2 with MCNP calculations. Both Serpent-2 and MCNP are used to perform the 3D Monte Carlo neutron and photon coupled transport making use of the EU-DEMO BL2017 model and a grid discrete volume source. A very useful conversion tool (called csg2csg) developed by UKAEA researchers was used to translate the MCNP input model to the Serpent-2 input model. The nuclear performance of the EU DEMO Helium Cooled Pebble Bed breeding blanket, such as tritium breeding ratio (TBR), neutron and photon flux spectra, and nuclear heating are analyzed, and the corresponding results are compared between Serpent-2 and MCNP. This study shows that Serpent-2 is suitable for the application of nuclear analysis for the EU DEMO HCPB breeding blanket.

Optimization of tritium breeding ratio in ARC reactor

Abstract Affordable Robust Compact reactor is a conceptual design for a Tokamak conceived by Massachusetts Institute of Technology (MIT) researchers. The design of this tokamak is under development and update. One of the key parameters for fusion reactor power plants is the tritium breeding ratio (TBR), which has to guarantee the tritium self-sufficiency. The tritium inventory circulating in a fusion power plant must be minimized. In the meantime, to enhance plant’s economics, the amount of tritium generated and stored should be maximized, since it would be used to startup new reactors. Both of the aforementioned trends meet their best in a TBR as high as possible. In this work, ARC tritium breeding ratio is studied and optimized. Taking advantage of Monte Carlo neutron transport codes, several configurations of ARC’s blanket and vacuum vessel have been analyzed in order to find the most effective one for a high TBR. The study takes into account different materials for the structure, such as Inconel718, V-15Cr-5Ti and Eurofer97. Moreover, it scans different width of coolant’s channels and evaluates the effect of lithium-6 enrichment in the blanket looking for the best configuration in terms of TBR.

A new Monte Carlo approach for solution of the time dependent neutron transport equation based on nodal discretization to simulate the xenon oscillation with feedback

In this paper a probabilistic methodology based on core nodalization is proposed to estimate the core power in the presence of xenon oscillation. A time-dependent Monte Carlo neutron transport code named MCSP-NOD is developed for dynamic analysis in arbitrary 3D geometries to simulate xenon oscillations as well as sub-critical condition with feedbacks. The new code is based on the approach adopted in MCNP-NOD which was previously introduced as a tool for core transient analysis using the MCNPX platform. As before, the core is divided into nodes of arbitrary dimensions, and all terms of the transport equation e.g. interaction rates, leakage ratio are estimated using the MC techniques. However, as a new option the concentration of iodine and xenon are also estimated which enables us to predict the oscillatory behavior of reactor power following poison oscillation. These quantities are then employed within the time-dependent neutron transport equation for each node independently to compute the neutron population. Simulations prove the robustness of the method.

In-core dosimetry for the validation of neutron spectra in the CROCUS reactor

The present article describes the preliminary validation study of simulated in-core and reflector n eutron spectra in preparation of oncoming experimental programs in the zeropower reactor CROCUS at EPFL. For this purpose, a set of activation foils were irradiated at three characteristic positions in the CROCUS reactor, and the subsequent activities were analyzed via γ spectrometry. The experimental setup was then modeled with the Monte Carlo neutron transport code Serpent2 and associated with an analysis tool to include the effect of the reactor power history during experiments. The comparison of calculated and measured reaction rates (C/E) indicates a general consistency (at 2σ) between calculated and measured spectra. However, offsets of C/E values were observed in (n, γ) reactions, up to 18% for 115In and 8% for 63Cu dosimeters. This could be caused by an unexpected isotopic composition, uncertainties in nuclear data, or the spectrometry analysis. In addition, a 100-groups spectrum unfolding was performed using the experimentally determined reaction rates and the Serpent2 spectra as the prior knowledge. The unfolded spectra were mainly adjusted in the thermal and fast ranges, while few modifications w ere m ade i n t he e pithermal r egion d ue t o the low contribution of epithermal neutrons in activation processes. Moreover, within energy groups where the capture reactions show resonant behavior, flux depletion (up to 38% as compared to the prior spectra) is observed due to the absence of self-shielding effect in the unfolding process. For this purpose, an unfolding method based on energy groups weighting is developed and tested.

Reactor physics evaluation of the TRIGA LEU fuel in the 20 MW NIST research reactor

This paper performs a neutronics evaluation of the General Atomics UZrHx low enriched uranium fuel – TRIGA fuel – in the National Bureau of Standards Reactor (NBSR) at the National Institute of Standards and Technology (NIST). The objective of this study is to examine the accountability and sustainability of the TRIGA fuel on neutronics aspects when applying it to the NBSR conversion. A feasibility scoping study was previously undertaken with considerations on various fuel dimensions, fuel rod layout configurations, and structure material selections, identifying the best option of deploying the TRIGA fuel to NBSR. Continuing with these efforts, an equilibrium NBSR core using the identified fuel was generated, and a well-round physics assessment was carried out by examining key neutronics performance characteristics of the core. All calculations were completed with MCNP-6, a 3-D Monte Carlo neutron transport code. The same fuel management scheme and fuel cycle length as the existing NBSR was adopted in the equilibrium core generation adopts to retain performance consistencies. The effectiveness of the fuel was examined at four representative burnup states of the fuel cycle. Neutronics performances of the equilibrium core was characterized by the fast and thermal neutron flux level as well as power distribution in the core. Reactor safety related parameters such as kinetics parameters and power peaking factors were also evaluated in the study. All results were compared against the current NBSR fueled with HEU for justifications. The findings in this research prove the viability of the TRIGA fuel for the NBSR conversion, and provide supporting data for future investigations on this subject.

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.

Measurement of leakage neutron spectra for zirconium with D-T neutrons and validation of evaluated nuclear data

Integral neutronics experiments have been the powerful tool and method for examining the reliability of the evaluated nuclear data. In present paper, the accuracy of evaluated nuclear data for zirconium has been validated by comparing measured leakage neutron spectra with calculated ones. Leakage neutron spectra from the irradiation of D-T neutrons on zirconium samples were experimentally measured at 60° and 120° by using a time-of-flight method. Theoretical calculations are carried out by the Monte Carlo neutron transport code MCNP-4C with the evaluated nuclear data of the ENDF/B-VII.0, ENDF/B-VIII.0, JENDL-4.0 and CENDL-3.1 libraries. From the comparisons, it is found that the calculations with JENDL-4.0 gives better agreements with the experiments at 60° while a large discrepancies are observed in the 4–13.5 MeV energy range at 120°. However, the calculated spectra using the data from ENDF/B-VII.0, ENDF/B-VIII.0 and CENDL-3.1 libraries showed some discrepancies with the measured ones, especially caused by the difference of the elastic and inelastic angular distributions and the secondary neutron energy distributions in the continuum inelastic scattering.

Comparative analysis of coupling schemes of Monte Carlo burnup calculation in RMC

As wrapping tools for coupling Monte Carlo (MC) neutron transport code with depletion solver, four different coupling schemes are implemented in Reactor Monte Carlo Code (RMC) developed at Tsinghua University, and verified in two assembly testing cases derived from VERA international benchmark. In this paper, accuracy and time consumption are comparatively analyzed, as well as memory requirement which is significant in large-scale burnup calculation. Results illustrate that beginning of step approximation, as the basic and simplest coupling method, performed with relatively poor accuracy. With nearly the same efficiency, two predictor-corrector methods provide identical accuracy better than the beginning of step approximation, but require different computer memories. High order predictor-corrector combined with sub-step method (HSPC) has the best performance and requires the most memory. By means of hybrid parallelism, domain decomposition and parallel depletion solver in RMC, HSPC is able to be applied in large-scale burnup problem.

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.

The first application of modified neutron source multiplication method in subcriticality monitoring based on Monte Carlo

Abstract The control rod drive mechanism needs to be debugged after reactor fresh fuel loading. It is of great importance to monitor the subcriticality of this process accurately. A modified method was applied to the subcriticality monitoring process, in which only a single control rod cluster was fully withdrawn from the core. In order to correct the error in the results obtained by Neutron Source Multiplication Method, which is based on one point reactor model, Monte Carlo neutron transport code was employed to calculate the fission neutron distribution, the iterated fission probability and the neutron flux in the neutron detector. This article analyzed the effect of a coarse mesh and a fine mesh to tally fission neutron distributions, the iterated fission probability distributions and to calculate correction factors. The subcriticality before and after modification is compared with the subcriticality calculated by MCNP code. The modified results turn out to be closer to calculation. It's feasible to implement the modified NSM method in large local reactivity addition process using Monte Carlo code based on 3D model.

On data assimilation with Monte-Carlo-calculated and statistically uncertain sensitivity coefficients

Sensitivity coefficients from Monte Carlo neutron transport codes have uncertainties that can affect nuclear data adjustments with integral experiments. This paper presents an extended version of Generalized Linear Least Squares (GLLS), called xGLLS, that accounts for these uncertainties. With very large sensitivity uncertainties, xGLLS constrains the nuclear data adjustments so that the posterior biases and uncertainties are larger than with GLLS. However, for the range of sensitivity uncertainties realistically encountered, xGLLS does not produce adjustments different from GLLS. This indicates that sensitivity uncertainties are not important compared to experimental, modeling, methodological, and nuclear data uncertainties. To balance a simulation’s accuracy with its computational cost, we recommend stopping a simulation once the uncertainty of a calculated integral parameter, caused by modeling and methodologies and by the sensitivities, is an order of magnitude smaller than that caused by nuclear data.

Multiphysics approach to plasma neutron source modelling at the JET tokamak

A novel multiphysics methodology for the computation of realistic plasma neutron sources has been developed. The method is based on state-of-the-art plasma transport and neutron spectrum calculations, coupled with a Monte Carlo neutron transport code, bridging the gap between plasma physics and neutronics. In the paper two JET neutronics tokamak models are used to demonstrate the application of the developed plasma neutron sources and validate them. Diagnostic data for the record JET D discharge 92436 are used as input for the TRANSP code, modelling neutron emission in two external plasma heating scenarios, namely using only neutral beam injection and a combination of the latter and ion cyclotron resonance heating. Neutron spectra, based on plasma transport results, are computed using the DRESS code. The developed PLANET code package is employed to generate plasma neutron source descriptions and couple them with the MCNP code. The effects of using the developed sources in neutron transport calculations on the response of JET neutron diagnostic systems is studied and compared to the results obtained with a generic plasma neutron source. It is shown that, although there are significant differences in the emissivity profiles, spectra shape and anisotropy between the neutron sources, the integral response of the time-resolved ex-vessel neutron detectors is largely insensitive to source changes, with major relative deviations of up to several percent. However it is calculated that, due to the broadening of neutron spectra as a consequence of external plasma heating, larger differences may occur in activation of materials which have threshold reactions located at DD neutron peak energies. The PLANET plasma neutron source computational methodology is demonstrated to be suitable for detailed neutron source effect studies on JET during DT experiments and can be applied to ITER analyses.

Processing of the multigroup cross-sections for MCNP calculations

Stochastic Monte Carlo (MC) neutron transport codes are widely used in various reactorphysics applications, traditionally related to criticality safety analyses, radiation shielding and validation of deterministic transport codes. The main advantage of Monte Carlo codes lies in their ability to model complex and detail geometries without the need of simplifications. Currently, one of the most accurate and developed stochastic MC code for particle transport simulation is MCNP. To achieve the best real world approximations, continuous-energy (CE) cross-section (XS) libraries are often used. These CE libraries consider the rapid changes of XS in the resonance energy range; however, computing-intensive simulations must be performed to utilize this feature. To broaden ourcomputation abilities for industrial application and partially to allow the comparison withdeterministic codes, the CE cross section library of the MCNP code is replaced by the multigroup (MG) cross-section data. This paper is devoted to the cross-section processing scheme involving modified versions of TRANSX and CRSRD codes. Following this approach, the same data may be used in deterministic and stochastic codes. Moreover, using formerly developed and upgraded crosssection processing scheme, new MG libraries may be tailored to the user specific applications. For demonstration of the proposed cross-section processing scheme, the VVER-440 benchmark devoted to fuel assembly and pip-by-pin power distribution was selected. The obtained results are compared with continues energy MCNP calculation and multigroup KENO-VI calculation.

Benchmarking of evaluated nuclear data for iron by a TOF experiment with slab samples

In order to validate the evaluated nuclear data of iron, neutron leakage spectra in the range from 0.8 to 15 MeV from natural iron slab samples were measured by time-of-flight technique at 60° and 120°, the measurements were performed at China Institute of Atomic Energy (CIAE) using a D-T neutron source, the slab thicknesses were chosen to be 5 cm, 10 cm and 15 cm, corresponding to 1.08, 2.16 and 3.24 mean free paths for 14.5 MeV neutrons, respectively. Theoretical calculations were carried out by Monte Carlo neutron transport code MCNP-4C, using the evaluated nuclear libraries of iron from the CENDL-3.1, ENDF/B-VIII.0, JENDL-4.0, JEFF-3.3, and TENDL-2017. The experimental results were compared with the corresponding calculated ones. From the comparison, it was found that: (1) In the 12–15 MeV range, the results with the ENDF/B-VIII.0 library well reproduce the experimental ones at both 60°and 120°, but those with the JENDL-4.0, JEFF-3.3 and TENDL-2017 libraries are underestimated, especially at 120°; (2) In the 8.5-12 MeV energy range, the calculated neutron spectra with the TENDL-2017 library are underestimated by more than 35%, but those with the JENDL-4.0 library are overestimated around 20%.

A GPU-based direct Monte Carlo simulation of time dependence in nuclear reactors

A novel 3D Monte Carlo (MC) neutron transport code, GUARDYAN, was developed to simulate direct time dependence in nuclear reactors. GUARDYAN (GpU Assisted Reactor DYnamic ANalysis) addresses the huge computational need by exploiting massive parallelism available on modern Graphics Processing Units (GPUs). While the code is still under development, transient analysis on large scale problems is already obtainable. The implementation is verified via comparison of differential and integral quantities to MCNP6 results, including several criticality safety benchmarks. Unlike most conventional MC codes GUARDYAN is intentionally designed for time-dependent calculations supporting parallel scalability on state-of-the-art high performance computing platforms. The methodology of transport simulation thus differs in many aspects: generation-by-generation tracking is replaced by a time step method; branching of neutron histories, neutron banking is eliminated by statistical weight manipulations; a robust delayed neutron treatment is implemented. These concepts, along with advanced acceleration techniques for improving the performance of point-in-cell search routine and the delta tracking method, resulted in an efficient MC tool that seems to outperform existing methods for kinetic MC simulation. Transient analysis was performed on an LWR core demonstrating that simulation of one second of a transient requires around 50 h on a single GeForce GTX 1080 GPU. The power evolution produced by GUARDYAN during this transient was also compared to experimental data; remarkably close agreement was found despite the uncertainties in the MC model.

Reactivity monitoring of the accelerator driven VENUS-F subcritical reactor with the “current-to-flux” method

In this article, we evaluate the quality and robustness of a method envisaged for the on-line monitoring of the subcriticality of an ADS, called the “current-to flux” (CTF) method. For this evaluation, we performed a dedicated experiment at the GUINEVERE facility. It is hosted at the SCK-CEN and consists of the subcritical VENUS-F reactor coupled to a continuous external neutron source provided by the GENEPI-3C accelerator. During this experiment, the reactor control rods were moved in various patterns, and the subsequent dynamical evolutions of the reactor reactivity were monitored using nine fission chambers (FCs). The space-energy effects that bias the reactivity values are corrected using a procedure based on simulations computed with the Monte Carlo neutron transport code MCNP. We investigate the precision of this correction procedure by comparison with a reactivity value extracted with the beam interruption technique and we demonstrate its insensitivity to the simplifications made on the VENUS-F reactor modeling and to a simulation key parameter such as the boron carbide density filling the control rods.

On the upper limit for the energy of epithermal neutrons for Boron Neutron Capture Therapy

The upper limit of the energy for epithermal neutrons is usually fixed to 10 keV based on the experience with spectra from reactors used until now for Boron Neutron Capture Therapy. The future use of accelerators with this therapy will provide different neutron spectra that require a more detailed study of this upper limit. Using our own Monte Carlo neutron transport code, we have calculated figures of merit on the dose depth profile to find this upper limit. Our results indicate that this upper limit could be incremented up to 15–20 keV under these new conditions.