The present paper describes a first step taken at the Paul Scherrer Institut in the development of a new multi-physics platform for reactor analysis. Such platform is based on the finite-volume software OpenFOAM and aims at a tightly coupled description of neutron transport, thermal mechanics and fluid dynamics. For this purpose, a steady-state 3-D discrete ordinates/thermal-mechanics solver was first developed in collaboration with the Politecnico di Milano. The present work briefly discusses such solver and its preliminary validation, which will be described in detail in parallel publications. It then focuses on its extension to time-dependent simulations. The solver is first tested by simulating different step-wise reactivity insertions in a critical configuration constituted by an infinite slab of highly enriched uranium. Subsequently, a super-prompt-critical power burst in the Godiva reactor has been simulated. Godiva was a spherical assembly of highly enriched uranium built and operated at the Los Alamos National Laboratory (US) during the Fifties. A prompt-critical transient in such system configures as a quick power excursion (up to ∼10 GW), which causes a temperature rise, and a subsequent reactivity reduction via expansion of the sphere. The overall transient lasts for few fractions of a millisecond. Results obtained with the newly developed model have been compared to experimental results, showing a relatively good agreement.

The Minor Actinides (MA) generated by nowadays PWRs fleet has significant impact on environment and biosphere. Inert Matrix Fuels (IMF) is a possible way to reduce the production and hazard of MA in recent. From neutronic aspect, using the MCNP code with temperature related continuous neutron data, the present paper analyses the isotopic contributions to the Doppler Coefficients of certain types IMF fuels. It is concluded that, the Doppler Coefficients of Al 2 O 3 +ZrO 2 +MgO and ZrO 2 based IMF fuels are much smaller than those containing ThO 2 , since the low neutron absorptions and lacking of resonance broadening of Al, Zr, Mg and O elements. For the same Inert Matrix, Reactor Grade Plutonium (RG-Pu) IMF fuels have more negative Doppler Coefficients than Weapon Grade Plutonium (WG-Pu) IMF fuels, which induce by the more abundance of resonance isotopes 240 Pu, 242 Pu in RG-Pu. Since the different neutron absorption cross-section profiles, the Er 2 O 3 burnable poison has negative contribution to the Doppler Coefficient, however 10 B, a typical 1/v absorber, is on the contrary way.

The aim of this study is to investigate the availability and accuracy of the cross section data for 233 U to perform the calculations of the critical system. Two evaluated data libraries are available, U.S. data bank (ENDF) and the Japanese data bank (JENDL), by using BAYES method for resonance parameters available in SAMMY code and weighted least square method with nonlinear regression by using FITWR computer code. Evaluation of the 233 U has been investigated by using of SAMMY code, in order to generate a useful data base for critical calculations, the computer code FITWR for experimental - experimental data fitting show same results obtained from Bayes method included within SAMMY code, with a slight deference in the results at the evaluated cross sections due to different mathematical methods have different results.

An anticipated transient without scram (ATWS) is an anticipated operational occurrence (AOO) followed by failure of the automatic reactor trip function of the reactor protection system. The failure of the reactor to shut down during the certain AOOs can lead to increase in reactor coolant system (RCS) pressure and decrease in departure from nucleate boiling ratio (DNBR) margin for a pressurized water reactor (PWR). Japanese standard PWRs are equipped with ATWS mitigation system which consists of a diverse mitigation system which is independent from the reactor trip system. The ATWS mitigation system automatically initiates isolation of the main steam line flow and the auxiliary feed water system under condition indicative of an ATWS. Mitsubishi Heavy Industries, Ltd. (MHI) applies 3D coupled code, SPARKLE-2 [1] [2], to the ATWS evaluation. SPARKLE-2 is a 3D coupled code developed by MHI and consist of the PWR system transient analysis code M-RELAP5, the 3D neutron kinetics code COSMO-K [3] and the 3D core thermal-hydraulics code MIDAC [4]. SPARKLE-2 implements the 3D characteristics such as local moderator feedback and change in 3D power distribution during transient. Thanks to gain from the 3D calculation, the analysis results show that the plant transients are effectively mitigated by the ATWS mitigation system and the RCS pressure and the minimum DNBR meet the safety criteria. These results also show that operational margins are increased, which enables more flexible design of the reload core.

GPUs have gradually increased in computational power from the small, job-specific boards of the early 90s to the programmable powerhouses of today. Compared to CPUs, they have a higher aggregate memory bandwidth, much higher floating-point operations per second (FLOPS), and lower energy consumption per FLOP. Because one of the main obstacles in exascale computing is power consumption, many new supercomputing platforms are gaining much of their computational capacity by incorporating GPUs into their compute nodes. Since CPU optimized parallel algorithms are not directly portable to GPU architectures (or at least without losing substantial performance gain), transport codes need to be rewritten in order to execute efficiently on GPUs. Unless this is done, we cannot take full advantage of these new supercomputers for reactor simulations. In this work, we attempt to efficiently map the Monte Carlo transport algorithm on the GPU while preserving its benefits, namely, very few physical and geometrical simplifications. Regularizing memory access and introducing parallel-efficient search and sorting algorithms are the main factors in completing the task.

In recent years, the core physics code ANDREA has been significantly improved and its capabilities were vastly extended. The code implementation has been overhauled to more modern, object-oriented and modular architecture. The code structure was adapted to allow use of multiple different neutronics solvers and to tackle various spatial and energy discretization models. The data library formats and processing workflow have been completely generalized, and different transport codes (e.g. HELIOS, SERPENT or SCALE) can be used to prepare several-group cross-section libraries for ANDREA. The new version of the code has gone through extensive validation both on the benchmarks and experimental data. The modular architecture of ANDREA code allow for its ongoing development. Currently we focus on coupling of ANDREA code with TRANSURANUS code.

Analytical or semi-analytical benchmarks for the neutron transport equation are relatively infrequent. Some may argue they are no longer necessary because of the enormous computing power and computational technology that is now available. While to some extent true, they can still provide valuable code verification and also serve to teach theoretical and numerical transport methods not taught by executing MATLAB, MAPLE or MATHEMATICA programs or Monte Carlo simulations. The focus of this presentation is on a new analytical solution technique for the solution of the 1D, monoenergetic Green’s function for neutron transport. In this formulation, we consider the analytical solution to a three–term recurrence for flux moments resulting in a semi–analytical benchmark. We then apply the benchmark to assess the accuracy of the PN approximation leading to a rather unexpected result.

For the project of the Chinese Initiative Accelerator Driven Sub-critical system (CIADS), the Lead-Bismuth-Eutectic (LBE) spallation target is one of the two alternatives, which has high good thermal performance, mature technology, and other advantages. The physical design of the spallation target determines the neutron yield and the utilization of the neuron source, as well as the performance of the sub-critical reactor and other key issues. Based on the Monte Carlo program MCNPX, we did the preliminary design of spallation target coupled with the reactor with a k eff about 0.95. The energy deposition density distribution of the target and the window were calculated. In the mean time, the neutron flux density, the neutron energy spectrum, and the power amplification factors were calculated. By changing the positions of the target, the radii of the beam pipe and the thickness of target, we studied the variation of the neutronic parameters mainly mentioned above. The energy deposition density distribution was used as the heat source of the thermal-hydraulics analyses. From the neutronic parameters, we found that to get the maximum power amplification factor, the target window should be put at the positions 11.4 cm above the center of the core. Actually, when the target was put above the center of the core, from 0cm to 22cm, the maximum differences of the power amplification factor is less than 4.0%, which means the position will have little influences in this range. When the target window was put at the center, increasing of the window’s thickness will lead the decreasing of the power amplification factor. The enlargement of the beam pipe radii will decrease the maximum that the amplification factors can reach. Meanwhile, the increasing of the beam radii will enlarge the power amplification factor slightly. The physics analysis of the LBE target coupled with the reactor can give more information to the optimization of the target structure and the sub-critical reactor for CIADS.

The general equivalence theory (GET) and the superhomogenization method (SPH) are widely used for equivalence in the standard two-step reactor physics calculation. GET has behaved well in light water reactor calculation via nodal reactor analysis methods. The SPH was brought up again lately to satisfy the need of accurate pin-by-pin core calculations. However, both of the classical methods have their limitations. The super equivalence method (SPE) is proposed in the paper as an attempt to preserve the surface current, the reaction rates and the reactivity. It enhances the good property of the SPH method through reaction rates based normalization. The concept of pin discontinuity factors are utilized to preserve the surface current, which is the basic idea in the GET technique. However, the pin discontinuity factors are merged into the homogenized cross sections and diffusion coefficients, thus no additional homogenization parameters are needed in the succedent reactor core calculation. The eigenvalue preservation is performed after the reaction rate and surface current have been preserved, resulting in reduced errors of reactivity. The SPE has been implemented into the Monte Carlo method based homogenization code MCMC, as part of RMC Program, under developed in Tsinghua University. The C5G7 benchmark problem have been carried out to test the SPE. The results show that the SPE method not only suits for the equivalence in Monte Carlo based homogenization but also provides improved accuracy compared to the traditional GET or SPH method.

A space-time nodal transport code, DAISY, was developed to evaluate dynamic neutron behavior in innovative nuclear system. The steady transport process is based on an arbitrary triangles-z mesh nodal method which can treat complicated geometry configuration with enough precision and acceptable calculated quantity. This code employs the improved quasi-static method for neutron kinetics with a predictor-corrector scheme to improve computational efficiency. The direct method and the point approximation for neutron kinetics are also implemented into DAISY to evaluate the precision and efficiency of this predictor-corrector scheme. This code was verified by several transient benchmarks. It shows that the predictor-corrector scheme in DAISY can greatly reduce the computational time with enough precision.

A Monte Carlo burn-up code system, MCADS, which combines the Monte Carlo code MCNP and the depletion code LITAC, was developed for evaluation the capabilities of fuel burn-up in reactors. The code can be extremely useful for analysts who perform isotope production, material transformation, and depletion and isotope analyses on complex, non-lattice geometries, and uniform and non-uniform lattices. Three well-documented benchmarking problems including IAEA-ADS, OECD/NEA-PWR and fast reactor were used to validate MCADS. The discrepancy between the MCADS calculation and other codes fell into a reasonable range.

This paper reports the results of a comparison among JEFF and ENDF/B datasets when used by SERPENT and MONTEBURNS codes on a GFR-like configuration. Particularly, it shows a comparison between the two Monte Carlo based codes, each one adopting three different cross sections dataset, namely JEFF-3.1, JEFF-3.1.2 and ENDF/B-VII.1. Calculations have been carried out on the Allegro reactor, i.e. an experimental GFR-like facility that should be built in EU as GFR demonstrator. Results concern nuclear parameters as effective multiplication factor and fluxes, as well as the atomic densities for some important nuclides versus burnup.

In this paper, a newly developed hybrid subgroup decomposition method is tested in a 1D problem characteristic of g as c ooled thermal r eactors (GCR). The new method couples an efficient coarse-group eigenvalue calculation with a set of fine-group transport source iterations to unfold the fine-group flux. It is shown that the new method reproduces the fine-group transport solution by iteratively solving the coarse-group quasi transport equation. The numerical results demonstrate that the new method applied to 1D GCR problem is capable of achieving high accuracy while gaining computational efficiency up to 5 times compared to direct fine-group transport calculations.

An efficient grid depression reconstruction model on axial assembly power distribution was developed for MHI nuclear design code system GalaxyCosmo-S. The objective of this paper is to present the background, methodology and its application of the new model in GalaxyCosmo-S. In order to consider the grid depression effect to the homogeneous axial power distribution obtained from 3D nodal core calculation, the new model employs the concept of the pin-power reconstruction model widely used in modern core design codes. In the new model, axial heterogeneous assembly power distribution is calculated by synthesizing the grid form function to the axial homogeneous power distribution by nodal calculation. The form function is pre-produced by fitting the local grid depression data processed from the measured axial thimble reaction rate in the grid position. By incorporating the measured data, the form function can reflect the precise grid depression information. According to the present study, it was shown that the form function has a burnup dependency for its depth, and it is prepared for each fuel type and axial grid position. In order to confirm the applicability of the present method to the existing PWRs, the predicted axial power distribution by GalaxyCosmo-S was compared with the measured data by the movable detector (M/D). As a result, the good agreements were confirmed without any specific trends for burnup condition. In addition, the difference of axial power distribution between predicted and measured data was statistically analyzed for multiple plants, cycles and burnup conditions. From the results, it is confirmed that the systematic over- or under-estimations of the power distribution observed in the grid homogenized model are reduced by the grid depression model. So this model is suitable for the 3D power distribution analysis and F Q uncertainty evaluation.

Research Centre Rez solves several safety related projects dealing with safety of Czech NPPs, some of which require fully functioning Three Dimensional (3D) model of the reactor core. While in a number of safety analysis of various accident scenarios it is sufficient to use one point reactor kinetics, there are selected types of accidents in which it is useful to model the space (3D) neutron kinetics, in particular control rod ejections, boron dilution scenarios, including transitions from design basis to beyond design basis accidents. This paper is focused to analyze the present model of the core of VVER1000/V320 reactor. Which is applicable for 3D modeling of neutron kinetics in selected design and beyond design basis accidents. The model is based on a cross-sections library created by SCALE 6.1.2/TRITON simulations. PARCS 3.2 code uses homogenized cross-sections libraries to calculate neutronic and other core parameters of the PWR reactors. Similar model is prepared with MCNP6 for comparison between deterministic (Pn spherical-harmonics method used in PARCS) and the stochastic (Monte Carlo) approach (used in MCNP6). Such comparison will serve as a demonstration of the capability of the PARCS code for VVER1000/V320 analyses.

The reactor core of an accelerator driven sub-critical system has been physically analyzed by the MCNP code. Neutron flux density of different area within the reactor has been calculated, and the influence on its distribution has also been analyzed. Results show that there exists higher fast neutron flux variation at different element layer in fast region, and relatively lower thermal neutron flux variation at different element layer in thermal region. The calculated neutron flux meets the general design requirements in the reflector and shielding layer. Neutron multiplication factor is remarkable in the fast neutron spectrum area, and it realizes the energy amplification in the thermal spectrum area. The statistical particle number of code can influence the accuracy of the calculation and variation of the core design parameters can change the neutron flux distribution in the reactor core.

During the operating of nuclear reactor, the applicable condition of point reactor neutron kinetics is changed by control rods bulk insertion and reactor lattice deviating from the critical state, The measurement of reactivity, with the increase of subcriticality, shows that the results impact on the kinetic distortion effect, along with prompt neutron flux strongly deteriorated. According to the diffusion theory about prompt neutron and delayed neutron, the theoretical analysis and application of point reactor neutron kinetics have been carried out to quantify the kinetic distortion correction factors in subcritical systems, and these indicate that prompt neutron distributions are strongly affected by kinetic distortion. With the self-developed Pulsed Neutron Source measurement system, subcriticality measurement in different configuration of control rods in the zero power reactor of Nuclear Power Institute of China was carried out. In this paper the reactivity of the deep subcritical system are obtained by experiment of Pulsed Neutron Source method, meanwhile, the bias between experiment results of the areas-ratio method and the characteristic decay constant method are analyzed by comparing the condition of the experiment and the theory model, the main factors inducing the bias are found, which supplies helpful reference for other similar design and experiment.

Monte Carlo (MC) simulation, especially suitable for large and complex nuclear systems, can become computationally expensive due to the large number of neutrons which must be simulated for statistically accurate and precise estimates. It is generally understood that a sample estimate will converge to the population mean when a ‘large’ sample size is taken. The term ‘large’ is usually based on a guess and hence MC simulation is understood to be both an art and a science. Considerable work has been done to analyze convergence of MC results and develop posterior diagnostic tools. This paper addresses the convergence of MC simulation for two problems viz (i) a fixed-source non-multiplying system, and (ii) a critical system represented by Godiva. A traditional approach is used in the first part of the work while a ‘new’ approach essentially following Signals and Systems techniques from Digital Signal Processing gives ‘orginality’ to the analysis as it provides insight into the convergence of didactic problems in neutron transport simulation. The methods used are (i) comparison of MC flux with exact transport and diffusion solutions and relative entropy, with the Kullback-Leibler (KL) divergence, to quantify the convergence of estimates for flux as a function of sample size in Monte Carlo simulations, (ii) the effect of ‘skip cycles’ on the k eff estimate, and (iii) a system identification approach based on the ARX (Auto Regressive Exogenous Source) method to determine the correlation between generations. The latter can be incorporated in Monte Carlo codes leading to a priori rather than to a posteriori diagnostic tools for establishment of convergence. The main findings of this work for simple one-group problems are that a Kullback Leibler ε∼10 −3 can be specified a priori for the convergence criteria of a fixed source problem while a system-identification approach for a simple Godiva simulation would need a large number of data points to build an accurate ARX model and hence would be more difficult to include as an a priori tool; so it would essentially serve a purpose similar to the FOM which gives a quality metric only after the simulation is completed.

In-hospital Neutron Irradiator (IHNI) was specially designed for Boron Neutron Capture Therapy (BNCT), the rated power of IHNI is 30kW, corresponding to the neutron flux density 1×10 12 n·cm −2 ·s −1 in reactor core. IHNI is an undermoderated reactor of pool-tank type, and UO 2 with enrichment of 12.5% as fuel, light water as coolant and moderator, and metallic beryllium as reflector. The fission heat produced by the reactor is removed by the natural convection. On the both sides of the reactor core, there are two neutron beams, one is thermal neutron beam, and the other opposite to the thermal beam, is epithermal neutron beam. A small thermal neutron beam is specially designed for the measurement of blood boron concentration by the Prompt Gamma Neutron Activation Analysis (PGNAA). The decay constants and shares of six group of ordinary delayed neutron and nine group of photoneutron were obtained by WIMS code. Based on that, the relationship between the reactivity and the reactor period was calculated through the inhour equation. In this way, the excess reactivity and the reactivity worthies of the components (control rod, water, etc) in the core are obtained by periodic method during the startup of the reactor. The six test experiments were completed during startup, The test results show that the maximum continuous operation time at full power is 12h; the excess reactivity at cold clean state of the core is 4.2mk; The radiation levels at technical rooms are within the specified values at full power operation. When the positive reactivity with 4.2 mk is inserted into the reactor suddenly, the power will be increased to peak power, and then, it will turn to the normal value due to the negative temperature effect, this experiment shows the inherent safety of IHNI.

This paper mainly studies the frequency-dependent response of neutron noise to perturbation in general case in a full-sized cavity type Molten Salt Reactor (MSR). The previous works mainly discussed the situation in one dimension, for its simplexes to be calculated and can give some qualitative analysis, but it’s not enough to analysis the real situation in the MSR. Hence a 2-D cylindrical coordination is utilized in this paper. We also considered the influence of different fluid velocity and different height-diameter ratio on the static and dynamic condition. For simplification, here we use diffusion equations with one-group 2-D neutron flux and precursor, and the perturbation is generated in the axial line along the flow direction, the fluid materials set as homogeneous. In a full-sized system, since the coupling is tighter, closed-form results can’t be found, we can only give out the theoretical expression or numerical results.