Monte Carlo Neutron Transport Code Research Articles

Impact of new evaluated nuclear data libraries on core characteristics of innovative reactor designs

The impact of new evaluated nuclear data libraries (ENDF/B, JENDL and JEFF) on core characteristics of thermal and fast spectra innovative reactor designs was investigated. The innovative reactor design with thermal neutron spectrum is represented by small-sized, long-life CANDLE (Constant Axial shape of Neutron flux, nuclide densities and power shape During Life of Energy producing reactors) HTGRs (High Temperature Gas-cooled Reactors) with uranium and thorium fuel cycles, while the one with the fast neutron spectrum is represented by compact, sodium-cooled B&BRs (Breed-and-Burn Reactors). The CANDLE HTGR core characteristics (i.e. kinf, keff, discharge burnup, burning region moving velocity, core life time and axial power peaking factor) were evaluated by a dedicated deterministic neutronics and depletion code based on few-group neutron diffusion theory in 2-D RZ geometry, while the required microscopic cross sections were prepared by using SRAC2006 with JENDL-4.0, ENDF/B-VI and JEFF-3.1 based new SRAC2006 libraries. The B&BR core characteristics were evaluated using a continuous energy Monte Carlo neutron transport code, SERPENT, with JENDL-4.0, ENDF/B-VII.0, ENDF/B-VII.1 and JEFF-3.1.1 nuclear data based new libraries. The impact of new evaluated nuclear data libraries for innovative CANDLE HTGR core characteristics was found significantly larger for thorium fuel cycle than the one for uranium fuel cycle, especially in the discharge burnup, burning region moving velocity which affected the core life time significantly. The findings indicated the needs of more accurate nuclear data libraries for nuclides involved in thorium fuel cycle. The impact of new evaluated nuclear data libraries for innovative B&BR core characteristics was found in a slightly large variation for keff evolution during burnup which in turn affected the estimated core life time. However, a good agreement within the statistical error on the integral kinetic parameters, Doppler reactivity coefficients, coolant density reactivity coefficients, axial and radial expansion reactivity coefficients, power density distributions can be observed among the libraries. In addition, the buildup of fission products and minor actinides were also similar.

Transient Fission Matrix: Kinetic calculation and kinetic parameters [formula omitted] and [formula omitted] calculation

In this paper, we present a time dependent version of the fission matrix method called Transient Fission Matrix (TFM) that we developed to perform accurate spatial kinetic calculations with a reduced computational cost. The matrices are generated once using the Monte Carlo neutron transport MCNP prior to the kinetic calculations. This article is devoted to the description of this approach with two different sets of equations depending on the aimed precision. We tested the TFM approach in several instances. In the calculation of the kinetic parameters such as βeff and Λeff on the Flattop and Jezebel experiments, the results obtained are compatible with adjoint weighted calculations performed with the Monte Carlo neutron transport code SERPENT. Likewise, our calculated value of the αRossi observable is also compatible with the experimental measurement. Finally a time–space prompt neutron propagation induced by a neutron pulse on Flattop has been calculated using the TFM approach and compared to a reference time dependent SERPENT calculation. Again, the results obtained demonstrate the good characterization of the time–space system response by the TFM approach.

Optimizations of the energy grid search algorithm in continuous-energy Monte Carlo particle transport codes

In this work we propose, implement, and test various optimizations of the typical energy grid-cross section pair lookup algorithm in Monte Carlo particle transport codes. The key feature common to all of the optimizations is a reduction in the length of the vector of energies that must be searched when locating the index of a particle’s current energy. Other factors held constant, a reduction in energy vector length yields a reduction in CPU time. The computational methods we present here are physics-informed. That is, they are designed to utilize the physical information embedded in a simulation in order to reduce the length of the vector to be searched. More specifically, the optimizations take advantage of information about scattering kinematics, neutron cross section structure and data representation, and also the expected characteristics of a system’s spatial flux distribution and energy spectrum. The methods that we present are implemented in the OpenMC Monte Carlo neutron transport code as part of this work. The gains in computational efficiency, as measured by overall code speedup, associated with each of the optimizations are demonstrated in both serial and multithreaded simulations of realistic systems. Depending on the system, simulation parameters, and optimization method employed, overall code speedup factors of 1.2–1.5, relative to the typical single-nuclide binary search algorithm, are routinely observed.

Estimate of the reactivity of the VENUS-F subcritical configuration using a Monte Carlo MSM method

The GUINEVERE project was launched in 2006 in order to study the feasibility of transmutation in Accelerator Driven subcritical Systems (ADS). This facility, hosted at the Belgian Nuclear Research Center SCK·CEN, couples the fast subcritical lead reactor VENUS-F with an external neutron source provided by the GENEPI-3C accelerator.In order to test on-line subcriticality monitoring techniques, an independent measurement of the reactivity of VENUS-F was achieved using the Modified Source Multiplication Method (MSM). The unknown reactivity of the subcritical configuration of interest was determined by comparing detector count rates driven by an external neutron source in this configuration with those obtained in a slightly subcritical configuration whose reactivity was determined by rod drop experiments. To account for spatial effects, MSM correction factors were calculated for all the detectors using the Monte Carlo neutron transport code MCNP. The corrected reactivity values of all the detectors were found to be consistent. This led to a final estimate of −5.28±0.13$ for the reactivity of VENUS-F.The MSM factors calculated with MCNP were found to be insensitive to the assumptions made for modeling the reactor and the detectors. Thus the MSM method appears to be a robust technique for measuring large subcriticality values with good accuracy.

Advanced calculation methodology for manufacturing and technological parameters’ uncertainties propagation at arbitrary level of lattice elements grouping

A new methodology, referred to as manufacturing and technological parameters uncertainty quantification (MTUQ), is under development at Paul Scherrer Institut (PSI). Based on uncertainty and global sensitivity analysis methods, MTUQ aims at advancing state-of-the-art for the treatment of geometrical/material uncertainties in light water reactor computations, using the MCNPX Monte Carlo neutron transport code. The development is currently focused primarily on criticality safety evaluations (CSE). In that context, the key components are a dedicated modular interface with the MCNPX code and a user-friendly interface to model functional relationship between system variables. A unique feature is an automatic capability to parameterize variables belonging to so-called “repeated structures” such as to allow for perturbations of each individual element of a given system modelled with MCNPX. Concerning the statistical analysis capabilities, these are currently implemented through an interface with the ROOT platform to handle the random sampling design. This paper presents the current status of the MTUQ methodology development and a first assessment of an ongoing organisation for economic cooperation and development/nuclear energy agency benchmark dedicated to uncertainty analyses for CSE. The presented results illustrate the overall capabilities of MTUQ and underline its relevance in predicting more realistic results compared to a methodology previously applied at PSI for this particular benchmark.

Integral test of International Reactor Dosimetry and Fusion File on graphite assembly with DT neutron at JAEA/FNS

We perform an integral test of a new library of dosimetry cross section data, International Reactor Dosimetry and Fusion File release 1.0 (IRDFF 1.0), which has been released from the International Atomic Energy Agency recently. A pseudo-cylindrical graphite assembly is used with a DT neutron source at the fusion neutronics source facility of Japan Atomic Energy Agency. A lot of foils for measurements of dosimetry reaction rates are placed in the small space along the central axis in the assembly. After the DT neutron irradiation, forty dosimetry reaction rates are measured and are compared with calculated values, which are obtained by using a Monte Carlo neutron transport code MCNP5-1.40 and a nuclear data library ENDF/B-VII.1 with the IRDFF-v.1.03 as the response functions for the dosimetry reactions. The calculated results generally show good agreements with the experimental ones. The cross section data of the 68Zn(n,np)67Cu and 68Zn(n,d)67Cu reactions and the 113In(n,γ)114mIn reaction, which are not included in IRDFF-v1.03, are needed for the analysis of the 67Zn(n,p)67Cu reaction and the 115In(n,2n)114mIn reaction as the competitive reactions, respectively.

Minor actinides impact on basic safety parameters of medium-sized sodium-cooled fast reactor

Abstract An analysis of the influence of addition of minor actinides (MA) to the fast reactor fuel on the most important safety characteristics was performed. A special emphasis was given to the total control rods worth in order to describe qualitatively and quantitatively its change with MA content. All computations were performed with a homogeneous assembly model of modified BN-600 sodium-cooled fast reactor core with 0, 3 and 6% of MA. A model was prepared for the Monte Carlo neutron transport code MCNP5 for fresh fuel in the beginning-of-life (BOL) state. Additionally, some other parameters, such as Doppler constant, sodium void reactivity, delayed neutron fraction, neutron fluxes and neutron spectra distribution, were computed and their change with MA content was investigated. Study indicates that the total control rods worth (CRW) decreases with increasing MA inventory in the fuel and confirms that the addition of MA has a negative effect on the delayed neutron fraction.

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

Integral neutronics experiments have been investigated at Institute of Modern Physics, Chinese Academy of Sciences (IMP, CAS) in order to validate evaluated nuclear data related to the design of Chinese Initiative Accelerator Driven Systems (CIADS). In the present paper, the accuracy of evaluated nuclear data for Tungsten has been examined by comparing measured leakage neutron spectra with calculated ones. Leakage neutron spectra from the irradiation of D-T neutrons on Tungsten slab sample were experimentally measured at 60° and 120° by using a time-of-flight method. Theoretical calculations are carried out by Monte Carlo neutron transport code MCNP-4C with evaluated nuclear data of the ADS-2.0, ENDF/B-VII.0, ENDF/B-VII.1, JENDL-4.0 and CENDL-3.1 libraries. From the comparisons, it is found that the calculations with ADS-2.0 and ENDF/B-VII.1 give good agreements with the experiments in the whole energy regions at 60°, while a large discrepancy is observed at 120° in the elastic scattering peak, caused by a slight difference in the oscillation pattern of the elastic angular distribution at angles larger than 20°. However, the calculated spectra using data from ENDF/B-VII.0, JENDL-4.0 and CENDL-3.1 libraries showed larger discrepancies with the measured ones, especially around 8.5–13.5MeV. Further studies are presented for these disagreements.

Monte Carlo MSM correction factors for control rod worth estimates in subcritical and near-critical fast neutron reactors

The GUINEVERE project was launched in 2006, within the 6th Euratom Framework Program IP-EUROTRANS, in order to study the feasibility of transmutation in Accelerator Driven subcritical Systems (ADS). This zero-power facility hosted at the SCK·CEN site in Mol (Belgium) couples the fast subcritical lead reactor VENUS-F with an external neutron source provided by interaction of deuterons delivered by the GENEPI-3C accelerator and a tritiated target located at the reactor core center. In order to test on-line subcriticality monitoring techniques, the reactivity of all the VENUS-F configurations used must be known beforehand to serve as benchmark values. That is why the Modified Source Multiplication Method (MSM) is under consideration to estimate the reactivity worth of the control rods when the reactor is largely subcritical as well as near-critical. The MSM method appears to be a technique well adapted to measure control rod worth over a large range of subcriticality levels. The MSM factors which are required to account for spatial effects in the reactor can be successfully calculated using a Monte Carlo neutron transport code.

Algorithmic choices in WARP – A framework for continuous energy Monte Carlo neutron transport in general 3D geometries on GPUs

In recent supercomputers, general purpose graphics processing units (GPGPUs) are a significant faction of the supercomputer’s total computational power. GPGPUs have different architectures compared to central processing units (CPUs), and for Monte Carlo neutron transport codes used in nuclear engineering to take advantage of these coprocessor cards, transport algorithms must be changed to execute efficiently on them. WARP is a continuous energy Monte Carlo neutron transport code that has been written to do this. The main thrust of WARP is to adapt previous event-based transport algorithms to the new GPU hardware; the algorithmic choices for all parts of which are presented in this paper. It is found that remapping history data references increases the GPU processing rate when histories start to complete. The main reason for this is that completed data are eliminated from the address space, threads are kept busy, and memory bandwidth is not wasted on checking completed data. Remapping also allows the interaction kernels to be launched concurrently, improving efficiency. The OptiX ray tracing framework and CUDPP library are used for geometry representation and parallel dataset-side operations, ensuring high performance and reliability.

Combining Total Monte Carlo and Benchmarks for Nuclear Data Uncertainty Propagation on a Lead Fast Reactor's Safety Parameters

Analyses are carried out to assess the impact of nuclear data uncertainties on keff for the European Lead Cooled Training Reactor (ELECTRA) using the Total Monte Carlo method. A large number of 239Pu random ENDF-formatted libraries generated using the TALYS based system were processed into ACE format with NJOY-99.336 code and used as input into the Serpent Monte Carlo neutron transport code to obtain distribution in keff. The mean of the keff distribution obtained was compared with the major nuclear data libraries, JEFF-3.1.1, ENDF/B-VII.1 and JENDL-4.0. A method is proposed for the selection of benchmarks for specific applications using the Total Monte Carlo approach. Finally, an accept/reject criterion was investigated based on χ2 values obtained using the 239Pu Jezebel criticality benchmark. It was observed that nuclear data uncertainties in keff were reduced considerably from 748 to 443 pcm by applying a more rigid acceptance criteria for accepting random files.

Benchmark experiment on titanium with DT neutron at JAEA/FNS

Titanium is contained in lithium titanate which is a tritium breeding material candidate. In the nuclear design, accurate nuclear data are needed. However, few benchmark experiments had been performed for titanium. We performed a benchmark experiment with a titanium assembly and a DT neutron source at JAEA/FNS. The titanium assembly was covered with Li2O blocks in order to reduce background neutrons. Dosimetry reaction rates were measured with niobium, indium and gold foils inside the assembly. And fission rates of 235U were measured by using micro fission chambers. This experiment was analyzed by using the Monte Carlo neutron transport code MCNP5-1.40 with recent nuclear data libraries of ENDF/B-VII.0, ENDF/B-VII.1, JEFF-3.1.2, JENDL-4.0 and JENDL-4.0u1. The calculation results were compared with the measured one in order to validate the nuclear data libraries of titanium. The calculated results with ENDF/B-VII.1 agreed with the measured one the best because the (n,2n) and (n,n′cont) reaction cross section data and resonance parameters were improved.

Accelerated sampling of the free gas resonance elastic scattering kernel

In this work, we present the derivation and investigation of a new Doppler broadening rejection sampling approach for the exact treatment of resonance elastic scattering in Monte Carlo neutron transport codes. Implemented in OpenMC, this method correctly accounts for the energy dependence of cross sections when treating the thermal motion of target nuclei in elastic scattering events. The method is verified against both stochastic and deterministic reference results in the literature for 238U resonance scattering. Upscatter percentages and mean scattered energies calculated with the method are shown to agree well with the reference scattering kernel results. Additionally, pin cell and full core keff results calculated with this implementation of the exact resonance scattering kernel are shown to be in close agreement with those in the literature. The attractiveness of the method stems from its improvement upon a computationally expensive rejection sampling procedure employed by an earlier stochastic resonance scattering treatment. With no loss in accuracy, the accelerated sampling algorithm is shown to reduce overall runtime by 3–5% relative to the Doppler broadening rejection correction method for both pin cell and full core benchmark problems. This translates to a 30–40% reduction in runtime overhead.

Validation the Monte Carlo code RMC with C5G7 benchmark

RMC (Reactor Monte Carlo code) is a new 3D Monte Carlo neutron transport code being developed by Department of Engineering Physics in Tsinghua University. The current version of RMC is a β version. In this paper, based on 2D and 3D benchmark of C5G7, the criticality calculation capacity of RMC was verified. Comparisons were made between the benchmark eigenvalues and those outputs by the RMC code. The RMC-generated eigenvalues were within two standard deviations of the benchmark and MCNP values in all cases. Additionally, the flux was compared pin by pin between MCNP and RMC. The flux tallies generated by RMC were found to be in well agreement with those from MCNP.

Development of a MCNP–ORIGEN burn-up calculation code system and its accuracy assessment

An MCNP–ORIGEN burn-up calculation code system, named MCORE (MCNP and ORIGEN burn-up Evaluation code), is developed in this work. MCORE makes use of the Monte Carlo neutron and photon transport code MCNP4C and nuclides depletion and decay calculation code ORIGEN2.1. MCNP and ORIGEN are coupled by data processing and linking subroutines. In MCORE, a so called “modified predictor corrector” approach is used. MCORE provides the capability of using different depletion calculation schemes and simulating fuel shuffling. Total nuclide density changes in active cells are considered in MCORE. The validity and applicability of the developed code are tested by investigating and predicting the neutronic and isotopic behavior of a “VVER-1000 LEU Assembly Computational Benchmark” at lattice level and a “Physics of Plutonium Recycling” fast reactor at core level (OECD-NEA).The comparison results show that the MCORE code predicts the nuclide composition within 5% accuracy and k∞ within 800pcm at the end of the burn-up for LEU assembly (40MWD/kg HM). For a fast reactor, the results obtained by MCORE are in the range of reported results except for 243Am. In general, MCORE results show a good agreement with the benchmark values.

Multi-core performance studies of a Monte Carlo neutron transport code

Performance results are presented for a multi-threaded version of the OpenMC Monte Carlo neutronics code using OpenMP in the context of nuclear reactor criticality calculations. Our main interest is production computing, and thus we limit our approach to threading strategies that both require reasonable levels of development effort and preserve the code features necessary for robust application to real-world reactor problems. Several approaches are developed and the results compared on several multi-core platforms using a popular reactor physics benchmark. A broad range of performance studies are distilled into a simple, consistent picture of the empirical performance characteristics of reactor Monte Carlo algorithms on current multi-core architectures.

Multiscale simulation of neutron induced damage in tritium breeding blankets with different spectral shifters

A multiscale material defect simulation established to evaluate neutron induced damages on metals is applied to an estimation of material degradation in helium cooled molten lithium blankets in which four different spectral shifter materials are examined as a means of maximizing the tritium breeding ratio through proper shaping of the neutron spectrum. The multiscale system consists of a Monte Carlo neutron transport code, a recoil spectrum generation code, a molecular dynamics code, a high energy cascade breakup model, an object kinetic Monte Carlo code, and a simple formula as the shear stress estimator. The average recoil energy of the primary knock-on atoms, the total concentration of the defects, average defect sizes, and the increase in shear stress after a certain irradiation time are calculated for each spectral shifter. Among the four proposed materials of B4C, Be, Graphite and TiC, B4C reveals the best shielding performance in terms of neutron radiation hardening. The result for the increase in shear stress after 100 days of irradiation indicates that the increased shear stress is 1.5GPa for B4C which is about 40% less than that of the worst one, the graphite spectral shifter. The other damage indicators show consistent trends.

MCNPX–BUCAL1 code to code verification through burnup analysis

The availability of accurate burnup data is an essential first step in any systematic approach to enhancement of economics, safety and performance of a research reactor. This first step requires the utilization of a well verified burnup code system. In this work a newly home-developed burnup code called BUCAL1 is presented. The code provides the full capabilities of the Monte Carlo neutron and photon transport code MCNP (version 5c). BUCAL1 has the capability of using several depletion calculation schemes that do not exist in several other burnup code systems such as: shuffling, refueling and multicycles burnup calculation, in an automatic way.The accuracy and precision of BUCAL1 were tested for U-Zrh fuels, by a code to code verification with MCNPX2.7, by predicting the burnup parameters of the 2MW TRIGA Mark II Moroccan research reactor. Continuous energy cross section data from the more recent nuclear data evaluation ENDF/B-VII.0 as well as S(α, β) thermal neutron scattering functions distributed with the MCNP code were used. Analysis of the verification results shows that BUCAL1 is enough accurate to be used in burnup calculations.

Toward a new polyethylene scattering law determined using inelastic neutron scattering

Monte Carlo neutron transport codes such as MCNP rely on accurate data for nuclear physics cross-sections to produce accurate results. At low energy, this takes the form of scattering laws based on the dynamic structure factor, S(Q,E). High density polyethylene (HDPE) is frequently employed as a neutron moderator at both high and low temperatures, however the only cross-sections available are for ambient temperatures (∼300K), and the evaluation has not been updated in quite some time. In this paper we describe inelastic neutron scattering measurements on HDPE at 5 and 294K which are used to improve the scattering law for HDPE. We review some of the past HDPE scattering laws, describe the experimental methods, and compare computations using these models to the measured S(Q,E). The total cross-section is compared to available data, and the treatment of the carbon secondary scatterer as a free gas is assessed. We also discuss the use of the measurement itself as a scattering law via the one phonon approximation. We show that a scattering law computed using a more detailed model for the Generalized Density of States (GDOS) compares more favorably to this experiment, suggesting that inelastic neutron scattering can play an important role in both the development and validation of new scattering laws for Monte Carlo work.

三维中子-光子输运的蒙特卡罗程序MCMG

三维中子-光子输运的蒙特卡罗程序MCMG