Core Simulation Research Articles

Methodology for CPR estimations of BWR cycle specific transient reload analyses using CTF sub-channel code

For BWRs, computations of the minimum critical power ratio (CPR) during transients constitute one of the main components for core design and reload licensing. The primary objective of this is to establish the operating limit minimum CPR (OLMCPR) that will prevent occurrence of dryout during both normal operation and for the most limiting transients. To this aim, state-of-the-art industrial methods combine plant system transient simulations with hot-channel analyses using fuel-design specific CPR correlations in order to compute the margins to boiling crisis. To overcome the complexity of CPR correlations, an alternative approach based on combining the full core transient simulator SIMULATE-3 K (S3K) with the COBRA-TF (CTF) sub-channel code along with generic CHF correlations was established at PSI. The objective of this paper is to present the concepts of this computational scheme and to provide an assessment of its applicability for independent audits of core reload licensing analyses. A part of this, experimental validation of each component of the scheme is presented, including a first-of-a-kind validation of CTF against available CPR tests for a modern BWR fuel design employed in Switzerland. For steady-state conditions, a numerical verification of the CTF performance is then performed by comparing the results to those obtained by the SIMULATE 3-D core simulator with implemented CPR correlations. For transient evaluations, an assessment of the S3K/CTF scheme is presented by comparing the estimated OLMPCRs against results from a real reload licensing application submitted by a vendor for a recent Swiss BWR cycle. The qualitative as well as quantitative agreement between S3K/CTF and the vendor results using CPR correlations is presented for the complete matrix of transients including in the reload licensing submittal. Some systematic trends observed in the CTF CPR results are pointed out, e.g. the fact that CTF tends to underpredict the CPRs by around 10%, and discussed in the perspective of further enhancements of the methodology.

Validation of PWR Neutronics Code Package TORCH V2.0 With Nuclear Power Plant Measurements

This article presents the verification and validation (V and V) of PWR neutronics code package TORCH V2.0 with nuclear power plant (NPP) measurements. The advanced nuclear power engineering design software, TORCH V2.0, was developed by the Nuclear Power Institute of China (NPIC), China National Nuclear Corporation (CNNC). Based on the two-step calculation scheme, TORCH V2.0 mainly contains lattice physics code for assembly homogenization, link calculation code for few-group constant parameterization, and core simulation code for few-group core calculation. The calculation modules of each code were already verified against various benchmark problems, whereas this article focuses on the V and V of linked code system. The measured values of the reactor startup physics test and NPP operation from six PWR NPPs (Daya Bay NPP, Ling Ao NPP, Fangjiashan NPP, Qinshan NPP, Hainan Changjiang NPP, and Fuqing NPP) were utilized to perform the comparison and analysis of V and V. Compared parameters of the reactor startup physics test include critical boron concentration, control rod integral value, boron differential value, and isothermal temperature coefficient. Compared parameters of the NPP operation contain critical boron concentration, assembly-wise power distribution, hot spot factor, and nuclear enthalpy rise factor. The results show that the software TORCH V2.0 has reliable calculation ability and can be applied in the PWR nuclear power engineering design which is based on square fuel assembly.

A nodal method based on CMFD for pin-by-pin SP3 calculation

A nodal method based on Coarse Mesh Finite Difference (CMFD) is proposed for pin-by-pin core simulation with SP3 approximation. The Laplace operators of the 0th moment flux of SP3 equations are treated using CMFD, while the Laplace operators of the 2nd moment flux are treated using fine mesh finite difference (FMFD). Correction factor is determined by solving local two-node problems. Transverse 0th moment flux is expanded to second-order Legendre polynomials. SP3 equations are then transformed to a generalized eigenvalue problem which is solved using Krylov subspace methods, including Jacobi-Davidson method, Generalized Davidson method, and Krylov-Schur method. Standard nonlinear iterative strategy is carried out to obtain converged correction factor. A prototype code CORCA-PIN is developed. Numerical results show that solver option of CMFD using 1 × 1 radial mesh per cell and Jacobi Davidson iteration method is suitable for pin-by-pin whole core calculations with SP3 approximation.

Optimization of the loading pattern of the PWR core using genetic algorithms and multi-purpose fitness function

Abstract The study demonstrates an application of genetic algorithms (GAs) in the optimization of the first core loading pattern. The Massachusetts Institute of Technology (MIT) BEAVRS pressurized water reactor (PWR) model was applied with PARCS nodal-diffusion core simulator coupled with GA numerical tool to perform pattern selection. In principle, GAs have been successfully used in many nuclear engineering problems such as core geometry optimization and fuel configuration. In many cases, however, these analyses focused on optimizing only a single parameter, such as the effective neutron multiplication factor (k eff), and often limited to the simplified core model. On the contrary, the GAs developed in this work are equipped with multiple-purpose fitness function (FF) and allow the optimization of more than one parameter at the same time, and these were applied to a realistic full-core problem. The main parameters of interest in this study were the total power peaking factor (PPF) and the length of the fuel cycle. The basic purpose of this study was to improve the economics by finding longer fuel cycle with more uniform power/flux distribution. Proper FFs were developed, tested, and implemented and their results were compared with the reference BEAVRS first fuel cycle. In the two analysed test scenarios, it was possible to extend the first fuel cycle while maintaining lower or similar PPF, in comparison with the BEAVRS core, but for the price of increased initial reactivity.

NeuroGPU: Accelerating multi-compartment, biophysically detailed neuron simulations on GPUs

BackgroundThe membrane potential of individual neurons depends on a large number of interacting biophysical processes operating on spatial-temporal scales spanning several orders of magnitude. The multi-scale nature of these processes dictates that accurate prediction of membrane potentials in specific neurons requires the utilization of detailed simulations. Unfortunately, constraining parameters within biologically detailed neuron models can be difficult, leading to poor model fits. This obstacle can be overcome partially by numerical optimization or detailed exploration of parameter space. However, these processes, which currently rely on central processing unit (CPU) computation, often incur orders of magnitude increases in computing time for marginal improvements in model behavior. As a result, model quality is often compromised to accommodate compute resources. New MethodHere, we present a simulation environment, NeuroGPU, that takes advantage of the inherent parallelized structure of the graphics processing unit (GPU) to accelerate neuronal simulation. Results & comparison with existing methodsNeuroGPU can simulate most biologically detailed models 10–200 times faster than NEURON simulation running on a single core and 5 times faster than GPU simulators (CoreNEURON). NeuroGPU is designed for model parameter tuning and best performs when the GPU is fully utilized by running multiple (> 100) instances of the same model with different parameters. When using multiple GPUs, NeuroGPU can reach to a speed-up of 800 fold compared to single core simulations, especially when simulating the same model morphology with different parameters. We demonstrate the power of NeuoGPU through large-scale parameter exploration to reveal the response landscape of a neuron. Finally, we accelerate numerical optimization of biophysically detailed neuron models to achieve highly accurate fitting of models to simulation and experimental data. ConclusionsThus, NeuroGPU is the fastest available platform that enables rapid simulation of multi-compartment, biophysically detailed neuron models on commonly used computing systems accessible by many scientists.

The evaluation of TRACE/PARCS model for BWR-4 nuclear power plant by startup test transient analyses

Abstract Generally, the thermal hydraulic (TH) codes need the results of Neutron Kinetics (NK) codes providing the reactivity properties to calculate neutron flux. Then the TH codes perform the safety analyses obtaining the responses of pressure, temperature, or water level. Two kinds of different codes calculate different physical behaviors sequentially and separately. Simultaneously computing thermal hydraulic and neutron kinetics behaviors can enhance the accuracy of the analysis. Hence, it is crucial to develop the TH-NK coupled model. This study presents the capability of the TH-NK coupled model, developed by TRACE (TRAC/RELAP Advanced Computational Engine) and PARCS (Purdue Advanced Reactor Core Simulator), for the BWR-4 nuclear power plant. The establishment of the TRACE/PARCS model presented the nodal and component modeling methodologies. This model was used to simulate two startup tests of high power level system transients. Principal system responses, calculated by the TRACE/PARCS model, were compared with the measured data in startup tests and the results of the point kinetic calculation of the TRACE (TRACE/PK) to evaluate the model. The evaluation shows that the TRACE/PARCS model can simulate the interaction between thermal hydraulic and neutron kinetics phenomena and predict the transients suitably. Through the comparison, the TRACE/PARCS model can be confident doing the analyses of normal and abnormal operational transients to predict the transient responses.

Application of Electrical Impedance Measurements’ Method for Studies of Bubble Flows

One of the important tasks in carrying out a computational justification of the reliability and safety of equipment that is part of the projected nuclear power plants today is the modeling of the bubbly regime of the coolant flow. In this regard the aim of this work is the use of extended methods of using matrix conductometric systems which are widespread in research practice for study of gas-liquid flows.The work uses a method of primary processing of experimental data aimed at eliminating of excess conductivity in the cells of the developed wire mesh sensor which makes it possible to obtain the values of the true volumetric gas content in the investigated area.Subsequent analysis of the possibilities to estimate the volumes of registered gas bubbles by the gradient method as well as the size of the interface in the sensor cells which plays a key role in modeling the interfacial heat and mass transfer.Comparison of readings values with the control instruments cues showed a good agreement. The presented work is an adaptation of the use of a conductometric measuring system for the study of multicomponent flows with the aim of further application for the study of two-component flows in the channels of the core simulator using wire mesh sensors.

A New Method of Central Axis Extracting for Pore Network Modeling in Rock Engineering

Characterizing internal microscopic structures of porous media is of vital importance to simulate fluid and electric current flow. Compared to traditional rock mechanics and geophysical experiments, digital core and pore network modeling is attracting more interests as it can provide more details on rock microstructure with much less time needed. The axis extraction algorithm, which has been widely applied for pore network modeling, mainly consists of a reduction and burning algorithm. However, the commonly used methods in an axis extraction algorithm have the disadvantages of complex judgment conditions and relatively low operating efficiency, thus losing the practicality in application to large-scale pore structure simulation. In this paper, the updated algorithm proposed by Palágyi and Kuba was used to perform digital core and pore network modeling. Firstly, digital core was reconstructed by using the Markov Chain Monte Carlo (MCMC) method based on the binary images of a rock cutting plane taken from heavy oil reservoir sandstone. The digital core accuracy was verified by comparing porosity and autocorrelation function. Then, we extracted the central axis of the digital pore space and characterize structural parameters through geometric transformation technology and maximal sphere method. The obtained geometric parameters were further assigned to the corresponding nodes of pore and throat on the central axis of the constructed model. Moreover, the accuracy of the new developed pore network model was measured by comparing pore/throat parameters, curves of mercury injection, and oil-water relative permeability. The modeling results showed that the new developed method is generally effective for digital core and pore network simulation. Meanwhile, the more homogeneity of the rock, which means the stronger “representative” of binary map the rock cutting plane, the more accurate simulated results can be obtained.

Hunting for the correct pressure drop in a scaled reactor pool: Effect of geometry, mesh resolution, turbulence model and mass flow

The E-SCAPE (European SCAled Pool Experiment) facility at the Belgian nuclear research center SCK•CEN is a thermal hydraulic 1/6th-scale model of the MYRRHA reactor with an electrical core simulator and cooled by LBE. The E-SCAPE facility provides experimental feedback to the designers on the forced and natural flow patterns, flow mixing and stratification during operational and accidental conditions. In addition, it provides valuable data for validation of the computational techniques for their use with LBE and in the design of MYRRHA. In terms of momentum field, one of the most important integral quantity to reproduce is the pressure drop in the system which has a direct consequence on the evolution of the mass flows and determines the free-surface levels in transient scenarios. In this paper, a detailed CFD study is performed on the flow in the E-SCAPE core region with the aim on predicting the correct pressure drop. Different numerical representation of the core geometry have been modeled and subjected to validation against experimental results from E-SCAPE. Special attention has been addressed to the choice of geometric simplifications and numerical models for accurate representation of the core of the E-SCAPE pool facility.

Evaluation of PBMR-400 Core Design Steady State Condition with Serpent and AGREE

The significant recent advances in computer speed and memory have made possible an increasing fidelity and accuracy in reactor core simulation with minimal increase in the computational burden. This has been important for modeling some of the smaller advanced reactor designs for which simplified approximations such as few groups homogenized diffusion theory are not as accurate as they were for large light water reactor cores. For narrow cylindrical cores with large surface to volume ratios such the Ped Bed Modular Reactor (PBMR), neutron leakage from the core can be significant, particularly with the harder neutron spectrum and longer mean free path than a light water reactor. In this paper the core from the OECD PBMR-400 benchmark was analyzed using multigroup Monte Carlo cross sections in the HTR reactor core simulation code AGREE. Homogenized cross sections were generated for each of the discrete regions of the AGREE model using a full core SERPENT Monte Carlo model. The cross sections were generated for a variety of group structures in AGREE to assess the importance of finer group discretization on the accuracy of the core eigenvalue and flux predictions compared to the SERPENT full core Monte Carlo solution. A significant increase in the accuracy was observed by increasing the number of energy groups, with as much as a 530 pcm improvement in the eigenvalue calculation when increasing the number of energy groups from 2 to 14. Significant improvements were also observed in the AGREE neutron flux distributions compared to the SERPENT full core calculation.

Off-chip prefetching based on Hidden Markov Model for non-volatile memory architectures.

Non-volatile memory technology is now available in commodity hardware. This technology can be used as a backup memory for an external dram cache memory without needing to modify the software. However, the higher read and write latencies of non-volatile memory may exacerbate the memory wall problem. In this work we present a novel off-chip prefetch technique based on a Hidden Markov Model that specifically deals with the latency problem caused by complexity of off-chip memory access patterns. Firstly, we present a thorough analysis of off-chip memory access patterns to identify its complexity in multicore processors. Based on this study, we propose a prefetching module located in the llc which uses two small tables, and where the computational complexity of which is linear with the number of computing threads. Our Markov-based technique is able to keep track and make clustering of several simultaneous groups of memory accesses coming from multiple simultaneous threads in a multicore processor. It can quickly identify complex address groups and trigger prefetch with very high accuracy. Our simulations show an improvement of up to 76% in the hit ratio of an off-chip dram cache for multicore architecture over the conventional prefetch technique (g/dc). Also, the overhead of prefetch requests (failed prefetches) is reduced by 48% in single core simulations and by 83% in multicore simulations.

Implementation of two-phase gas transport into VERA for molten salt reactor analysis

Molten salt reactors (MSRs) are a class of next-generation nuclear reactors that have received recent industrial and research interest. A generalized species transport solver was implemented in the Virtual Environment for Reactor Applications (VERA) computing suite to extend this tool to analyze liquid-fueled MSRs. This core simulator has been extended to model the transport of fission product gases into a collection of circulating gas bubbles with the purpose of removing the gases. This paper presents the governing species transport equation, along with various nuclear source terms. Development of the source term for phase migration is discussed, along with a simplified interfacial area tracking method. Finally, a case study on a simplified MSR loop is presented in which modeling parameters were varied to assess their impact on gas removal. The steady state results show that parameters such as bubble diameter, gas injection rate and mass transfer coefficient have a low to moderate effect on the fraction of xenon in the core region. Removal efficiency has the greatest effect on the fraction in the core region. After the pump bowl, bubble diameter has a minor effect on the fraction of xenon in the gas void. These results point out that increasing parameters such as mass transfer coefficient, gas injection rate, and removal efficiency drives the xenon into the circulating gas void, while decreasing bubble diameter also drives xenon into the gas void by increasing interfacial area.

Static and transient analyses of Advanced Power Reactor 1400 (APR1400) initial core using open-source nodal core simulator KOMODO

The United Arab Emirates is currently building and operating four units of the APR-1400 developed by a South Korean vendor, Korea Electric Power Corporation (KEPCO). This paper attempts to perform APR-1400 reactor core analysis by using the well-known two-step method. The two-step method was applied to the APR-1400 first cycle using the open-source nodal diffusion code, KOMODO. In this study, the group constants were generated using CASMO-4 fuel transport lattice code. The simulation was performed in Hot Zero Power (HZP) at steady-state and transient conditions. Some typical parameters necessary for the Nuclear Design Report (NDR) were evaluated in this paper, such as effective neutron multiplication factor, control rod worth, and critical boron concentration for steady-state analysis. Other parameters such as reactivity insertion, power, and fuel temperature changes during the Reactivity Insertion Accident (RIA) simulation were evaluated as well. The results from KOMODO were verified using PARCS and SIMULATE-3 nodal core simulators. It was found that KOMODO gives an excellent agreement.

One-step coupled calculations (Serpent-OpenFOAM) for a fuel rod of the IPR-R1 triga reactor

In this work, a single step of coupled calculations for a fuel rod of IPR-R1 TRIGA was performed. The used me-thodology allowed to simulate the fuel pin behavior in steady-state mode for different power levels. The aim of this paper is to present a practical approach to perform coupled calculations between neutronic (Monte Carlo) and thermal-hydraulic (CFD) codes. For this purpose, is necessary to evaluate the influence of the water thermal-physical properties temperature variations on keff parameter. Besides that, Serpent Nuclear Code was used for the neutronics evaluation, while OpenFOAM was used for thermal-hydraulics. OpenFOAM si- mula-tions were made by using a modified chtMultiRegionFoam solver, developed to read Serpent output correctly. The neutronic code was used without any modifications. The results shows that this coupled calculations were consistent and that leads to encouraging further methodology development and its use for full core simulation. Also, the results shows good agreement with calculations performed using other version of OpenFOAM and Milonga as neutronic code.

Optimisation design of double-column multi-gap reactor core structure for noise reduction

The noise from a double-column multi-gap reactor is very loud and should be controlled. Based on different gap configurations, the gap configuration design schemes of the double-column multi-gap reactor to noise reduction are proposed. According to the positions of the gaps, the 20 gaps of the actual reactor are simplified into five gaps. Then, 31 reactor finite element simulation models are constructed respectively according to the gap position or length. A three-dimensional (3-d) electromagnetic–mechanical weak-coupled model is used to study the average surface vibration velocity of each reactor core simulation model with different gap configurations. The 3-d anisotropy of magnetisation curves (BH curves) and magnetostriction curves of laminated Grain-oriented steel plates are considered. Some gap configuration design schemes are proposed on the basis of the results of simulations. Two double-column multi-gap reactor models are used for comparative testing. The test results show that the gap configuration design schemes can reduce the sound power level of the reactor by about 6.836 dB. Finally, gap configuration design schemes for noise reduction of a double-column 20-gap reactor are proposed.

Direct whole core modeling and simulation of the SPERT III E-Core experiments by nTRACER

The SPERT III E-core experiments are modeled and simulated by a direct whole core calculation code nTRACER as an effort to validate its dynamics calculation capability. The control rod ejection experiments performed for the E-core of the Special Power Excursion Reactor are faithfully modeled by explicitly representing all the complex components such as fuel cans and the cruciform transient rod. The unknown experiment conditions such as the initial position of the ejected rod is estimated rigorously to yield the proper rod worth. The simulations are performed in the one-step manner such that the entire calculation for rod ejection is carried out with a single input deck without any prior calculation. All the kinetics data such as the fractions and decay constants of the delayed neutron precursors are directly obtained from the suitable evaluated nuclear data files. Doppler feedback is modeled by the direct estimation of the fuel temperature profiles inside each pellet. The transient results for the five ejection tests are compared with the measured data such as the peak time and magnitude of the power pulse, the net power release up to the power peak, and the post-burst power level. The good agreement between the nTRACER results and measured data demonstrates the efficacy of direct whole core dynamics calculation as well as the accuracy of nTRACER. In addition, a comparison is made with a two-step calculation to demonstrate the advantage of the direct whole core transient calculation.

Core Modeling and Simulation of Peach Bottom 2 Turbine Trip Test 2 Using CASMO5/TRACE/PARCS

The Purdue Advanced Reactor Core Simulator (PARCS) three-dimensional neutron kinetics code and the TRACE nuclear systems analysis code were interfaced. This provides a best-estimate coupled code system for performing transient plant calculations with reactivity feedback from a detailed core model, significantly contributing to nuclear power plant safety analyses. This study performed steady-state and transient simulations of Peach Bottom 2 Turbine Trip Test 2 (PB2 TT2) using the CASMO5/TRACE/PARCS coupled code. Consequently, CASMO5/TRACE/PARCS simulates the rapid positive reactivity addition caused by the sudden closure of the turbine stop valve. Specifically, the discrepancy in the maximum total power during the transient condition was within 3% compared with the PB2 TT2 experimental data. Furthermore, the sensitivity of the thermal-hydraulic channel (CHAN) component modeling in the coupled CASMO5/TRACE/PARCS code revealed that the number of CHAN components influenced the assembly radial power peaking factor in the PB2 TT2 transient calculation.

Transitioning Spiking Neural Network Simulators to Heterogeneous Hardware

Spiking neural networks (SNN) are among the most computationally intensive types of simulation models, with node counts on the order of up to 10 11 . Currently, there is intensive research into hardware platforms suitable to support large-scale SNN simulations, whereas several of the most widely used simulators still rely purely on the execution on CPUs. Enabling the execution of these established simulators on heterogeneous hardware allows new studies to exploit the many-core hardware prevalent in modern supercomputing environments, while still being able to reproduce and compare with results from a vast body of existing literature. In this article, we propose a transition approach for CPU-based SNN simulators to enable the execution on heterogeneous hardware (e.g., CPUs, GPUs, and FPGAs), with only limited modifications to an existing simulator code base and without changes to model code. Our approach relies on manual porting of a small number of core simulator functionalities as found in common SNN simulators, whereas the unmodified model code is analyzed and transformed automatically. We apply our approach to the well-known simulator NEST and make a version executable on heterogeneous hardware available to the community. Our measurements show that at full utilization, a single GPU achieves the performance of about 9 CPU cores. A CPU-GPU co-execution with load balancing is also demonstrated, which shows better performance compared to CPU-only or GPU-only execution. Finally, an analytical performance model is proposed to heuristically determine the optimal parameters to execute the heterogeneous NEST.

Seepage Mechanism of Tight Sandstone Reservoir Based on Digital Core Simulation Method

Recently, tight sandstone oil has played an increasingly important role in the energy strategies of countries around the world. However, the understanding of a microscopic mechanism is still not clear enough, which has been affecting the improvement of the recovery of tight sandstone oil. In this article, a digital core model was established to simulate the pore network of a physical core with CT scan and difference equations were verified by Fourier counting. Then, a combination of orthogonal tests and cubic digital cores was used to experimentally investigate various parameters including pressure, length, permeability, viscosity, and time. By combining the physical experiments with the digital core methods, it can be observed that the state of the micro-crack affects the conductivity of the core, which may be the decisive reason for changing the pressure gradient. The orthogonal test showed that the sensitivity of the parameters was pressure, length, permeability, time, and viscosity in order. The results of the numerical simulations showed that this method can reveal the seepage mechanism of a tight sandstone reservoir, greatly shortening the experimental time and improving flexibility.

An Experimental Study into the Hydrodynamics of the Loop Coolant Flows’ Mixing in the Nuclear Reactor Downcomer

This work is an experimental study into the hydrodynamics of the coolant flow in the in-vessel pressure duct of a pressurized water reactor. In the article, an experimental testbench and a nuclear reactor model under investigation are described, the measurement methods are set forth, and the operating variables at which the study was conducted and the obtained results are provided. In the experiments, the mixing of the loop coolant flows inside the model of the nuclear reactor downcomer was simulated. The study was conducted on the high-pressure aerodynamic testbench of Alekseev State Technical University in Nizhny Novgorod. The scale model of the nuclear reactor had the structural components characteristic of loop-type reactor units, such as the annular downcomer and the bottom pressure vessel. The experiments were conducted at Reynolds numbers within the 20 000–50 000 range measured in the annular gap of the downcomer of the model. The axial velocity field at the inlet to the reactor core simulator was investigated using a pneumometric probe. The temperature field was recorded in the experiment by the impurity diffusion method, i.e., by introducing of a contrast tracer into one of the loops of the model. The degree of mixing the flows was estimated by the admixture (tracer) concentration at the inlet to the core simulator. Propane was used as the contrast admixture. The study has yielded the spatial distribution of the tracer in the coolant flow in the annular downcomer and in the bottom pressure vessel. The data on the distribution of the contrast admixture are presented in the form of charts. The swirling of the coolant flow in the in-vessel pressure duct has been analyzed. It has been shown that the mixing intensity in the bottom pressure vessel is affected by the central vortex with the central axis. The parameters of mixing the admixture in the model of the in-vessel pressure duct have been estimated.