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At least 541 records · Page 30

$χ$-$MeRA$: Computationally efficient adaptive mesh refinement of Monte Carlo mesh based tallies

Here, the reactor physics community is always focused on reducing the computational time and memory required for simulations. $χ$-$MeRA$, which stands for flux-based-($χ$)-Mesh tally Refinement Adaptively, was built to reduce the computational time and memory required to solve the neutronics side of a multiphysics problem when compared to traditional methods for mesh based tallies in Monte Carlo (MC) simulations. $χ$-$MeRA$ couples a MC code with an adaptive mesh refinement (AMR) algorithm to take advantage of the accuracy of a MC code and the efficiency of an AMR algorithm. Also developed within $χ$-$MeRA$ was a set of metrics to assess the effects of the refinement on various parameters in the simulation space. For a plutonium sphere, $χ$-$MeRA$ shows a reduction in memory usage and computation time when compared to a fully refined mesh by a factor of 14.7 and 6.7, respectively. When compared to an unstructured mesh, improvement of 1.3 and 4.8 was achieved for memory usage and computation time. The development of $χ$-$MeRA$ helps solve the neutronics side of a multiphysics problem in a faster, more computationally efficient manner than traditional methods, and the final mesh created contains accurate results that can be passed onto the next physics code.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗

Dirac $R$-matrix calculations for the electron-impact excitation of neutral tungsten providing noninvasive diagnostics for magnetic confinement fusion

Neutral tungsten is the primary candidate as a wall material in the divertor region of the International Thermonuclear Experimental Reactor (ITER). The efficient operation of ITER depends heavily on precise atomic physics calculations for the determination of reliable erosion diagnostics, helping to characterize the influx of tungsten impurities into the core plasma. The following paper presents detailed calculations of the atomic structure of neutral tungsten using the multiconfigurational Dirac-Fock method, drawing comparisons with experimental measurements where available, and includes a critical assessment of existing atomic structure data. We investigate the electron-impact excitation of neutral tungsten using the Dirac R-matrix method, and by employing collisional-radiative models, we benchmark our results with recent Compact Toroidal Hybrid measurements. The resulting comparisons highlight alternative diagnostic lines to the widely used 400.88-nm line.

74 ATOMIC AND MOLECULAR PHYSICS↗

Developing A Fully Implicit Compressible Flow Valve Model in RELAP-7 and Its Application in Short Term BWR Station Black-Out Analyses

This paper presents a set of dynamical models for SRV simulation and their implementation in the RELAP-7 code, which is a new reactor system analysis code being developed at INL. RELAP-7 is based on the MOOSE (the Multi-Physics Object-Oriented Simulation Environment) framework. The compressible flow SRV model covers three different sets of models: fully closed case, subsonic case, and choking case. The RELAP-7 is a fully implicit code and the preconditioned Jacobian-free Newton-Krylov (JFNK) method is used to solve the discretized nonlinear problem. The JFNK method generally needs a good preconditioning scheme to be effective and efficient. There are different ways to precondition the JFNK method. In RELAP-7, two main options are used to obtain the preconditioning matrix: (1) FDP (Finite Difference Preconditioning) option which uses the default finite differencing method provided in the PETSc package. PETSc is the numerical solver software developed by Argonne National Laboratory and used by MOOSE as the numerical solution engine; (2) SMP (Single Matrix Preconditioning) option which requires code developers to provide approximate analytical Jacobian inputs for the important diagonal and off-diagonal terms. Due to the strong discontinuity in the residual vector space when the compressible flow valve model switches from one set of model to another, FDP option cannot guarantee convergence. Instead, accurate analytical Jacobain terms are provided for the compressible flow valve model. With the SMP option selected, the compressible flow valve model can be solved efficiently under different cases. The developed compressible flow valve model has been used for SRV component simulation in several simplified BWR SBO (Station Black-Out) simulations with RELAP-7. This paper only presents cases for short-term SBO in which all the safety systems fail to start and function properly when SBO occurs. Consequently, only the SRVs automatically open and close periodically to control the primary system pressure and discharge the high temperature and high pressure steam into the suppression pool. Although simplified, this case is quite similar to what happened in Fukushima Daiichi Unit 1. Unit 1 had no RCIC system, while the isolation condenser system was believed to be non-functioning, or only available for a very short period of time during the accident. Therefore, the Unit 1 reactor core reached a fuel failure temperature only a few hours into the accident. Two short-term SBO cases will be compared: one with single SRV discharge line which represents all the SRV lines and another case with multi-group SRVs which represent each group of SRV line with different setting points.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

A Backward Euler Doubling Feasibility Study Based on Thorium Series Cascade

The solution to the Bateman equations of radioactive decay are central to burnup calculations for operational nuclear reactors. There are several excellent methods of solving the equations of radioactive decay, with the CRAM algorithm as one of the best, at least for constant flux burnup. In this work, we attempt to investigate the feasibility of a relatively straight forward algorithm based simply on finite difference and doubling and apply the resulting benchmark to a recently developed decay chain formulation. A simple Backward Euler Doubling Finite Difference (BEDFD) method for solving first order ordinary differential equations is developed with enhanced precision through the convergence acceleration. Numerical results of two analytically solvable test problems clearly demonstrated that this newly proposed method is robust and accurate. Therefore, this method could be promising in solving larger depletion problems-our next effort in our comparisons with TTA. (authors)

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗

Development of a Griffin model of the advanced test reactor

In the pursuit of a higher fidelity deterministic simulation capability of the Advanced Test Reactor, it is important to have a fast yet accurate deterministic neutronics model. Here, to achieve this, we employed an advanced two-step method. The first step involves generating homogenized cross sections using OpenMC, a cutting-edge Monte Carlo neutron transport code. OpenMC offers excellent modular capabilities, allowing for easy component integration and flexibility in incorporating new designs into the model. The second step involves deterministic transport calculations, which are performed using Griffin, a reactor physics application based on the Multiphysics Object-Oriented Simulation Environment (MOOSE). To ensure the accurate spatial resolution and assignment of material cross sections, a Cubit-generated mesh for the Advanced Test Reactor is utilized as an intermediate step between the OpenMC and Griffin models; Griffin utilizes the mesh for its finite element solution, while OpenMC material identifications are written to the mesh file to be used in Griffin material assignments. Additionally, a Python-based script converts the cross sections generated by OpenMC into the ISOXML format required by Griffin. Initial comparisons using the Griffin diffusion solver indicated good agreement between the neutron multiplication factors obtained from the standalone OpenMC model and the Griffin model, with differences of less than 10 pcm in the 2D geometry configuration; it was later determined that this agreement was likely due to compensating effect and was more likely on the order of –700 pcm relative to the OpenMC solution. However, in three-dimensional calculations, an unacceptably large error (almost 8,000 pcm) was found in the Griffin solution with the diffusion solver. Subsequent calculations using Griffin’s discrete ordinates solver demonstrated substantially improved agreement, within 116 pcm of the OpenMC solution used to generate the cross sections for Griffin. Building on this capability, future work will seek to perform more detailed validation calculations. The ultimate goal is to evaluate both transient and multiphysics simulations of the reactor.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

Thermochemical reactor systems and methods

Thermochemical reactor systems that may be used to produce a fuel, and methods of using the thermochemical reactor systems, utilizing a reactive cylindrical element, an optional energy transfer cylindrical element, an inlet gas management system, and an outlet gas management system.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗

Irradiation test about oxidation-resistant graphite in WWR-K research reactor - 18533

Graphite materials are used for the in-core components of High Temperature Gas-cooled Reactor (HTGR). It is desirable to enhance oxidation resistance of graphite to keep much safety margin. SiC coating is the candidate method for this purpose. JAEA and four Japanese graphite companies, Toyo Tanso, Ibiden, Tokai Carbon and Nippon Techno-Carbon are studying to develop oxidation-resistant graphite for the in-core components. Neutron irradiation test with two capsules was carried out by WWR-K research reactor of Institute of Nuclear Physics of the Republic of Kazakhstan (INP) through the International Science and Technology Center (ISTC) partner project. The total irradiation cycles of WWR-K operation was 10 cycles and the total duration was 200 days. Irradiation temperature about 1473 K would be attained. The maximum fast neutron fluence (E>0.18 MeV) for the gas-swept capsule irradiated at a central irradiation hole was preliminary calculated as 1.2x10{sup 25} m{sup -2}, and for the sealed capsule at a peripheral irradiation hole as 4.2x10{sup 24} m{sup -2}. Dimension and weight of the irradiated specimens were measured, and outer surface of the specimens were observed by optical microscope. For the irradiated oxidation resistant graphite, out-of-pile oxidation test will be carried out at an experimental laboratory. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

Performing k eff Validation of As-Loaded Criticality Safety Calculations Using UNF-ST&DARDS: Sensitivity Calculations

The general method for performing validation of as loaded criticality safety calculations using UNF ST&DARDS is presented in a paper by Clarity, which includes a description of the UNF-ST&DARDS system. Proof-of-principle analyses were performed in the summer of 2019 for MPC-32 dual purpose canisters (DPCs) containing pressurized water reactor (PWR) fuel assemblies. Summaries of these results are presented in this and a companion paper for this conference. The current paper describes the TSUNAMI-3D calculations performed to generate sensitivity data, and the companion paper discusses the selection of critical experiments applicable for validation of the 11 MPC-32 DPCs considered. The generation of sensitivity data for as-loaded spent nuclear fuel (SNF) DPCs is a challenge given the detailed model of the fuel compositions generated by UNF ST&DARDS. Each fuel assembly is modeled with its own irradiation history in 18 axial nodes, unless the fuel assembly is damaged and thus considered as fresh by design basis. This results in a set of 576 fuel compositions, each of which must be processed separately in a multigroup (MG) calculation. Therefore, a continuous-energy (CE) TSUNAMI-3D method was chosen to alleviate this challenge. Two CE TSUNAMI-3D methods are available in SCALE 6.2.3: the iterated fission probability (IFP) and contribution-linked eigenvalue sensitivity/uncertainty estimation via track-length importance characterization (CLUTCH). Since the IFP method is not feasible because of memory requirements associated with its implementation in SCALE, the CLUTCH method was selected for these calculations. CLUTCH has been implemented in SCALE in parallel, allowing long calculations to be performed in reasonable timeframes. The two primary user inputs necessary for CLUTCH calculations are the F*(r) mesh and the number of latent generations used in determining the F*(r) function. This F*(r) function is used as the importance function for fission chains originating in a given volume element (voxel), and it is calculated using the IFP method in the skipped generations. A large number of skipped generations is thus required to ensure accurate calculation of this importance function. In these calculations, 500 generations were used to calculate the F*(r) function. For more information regarding the calculation of F*(r), see Jones [4]. The remainder of this paper is focused on the selection of the F*(r) mesh and the number of latent generations.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗

Stochastically estimated covariance matrices for independent and cumulative fission yields in the ENDF/B-VIII.0 and JEFF-3.3 evaluations

A Monte-Carlo method for the generation of correlation and covariance matrices for independent and cumulative fission yields has been developed. The method uses a constrained Monte-Carlo resampling structure in order to vary evaluated fission yield libraries in a way that meets basic conservation principles. This results in the generation of correlation/covariance matrices with limited model bias and uncertainty; the matrices are primarily reflective of the evaluated fission yield uncertainties and correlations that arise from the evaluation process. This method has been applied to generate correlation and covariance matrices for all of the fissioning systems of the ENDF/B-VIII.0 and JEFF-3.3 evaluations, marking the first time such matrices have been generated for all of these systems. These covariance matrices have been published online for immediate public use. These correlation and covariance matrices can be used to improve uncertainty estimation in calculations of reactor antineutrino emission rates, decay heat problems, and nuclear forensics.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗

Branch-and-Bound algorithm applied to uncertainty quantification of a Boiling Water Reactor Station Blackout

Evaluation of the impacts of uncertainty and sensitivity in modeling presents a significant set of challenges in particular to high fidelity modeling. Computational costs and validation of models creates a need for cost effective decision making with regards to experiment design. Experiments designed to validate computation models can be used to reduce uncertainty in the physical model. In some cases, large uncertainty in a particular aspect of the model may or may not have a large impact on the final results. For example, modeling of a relief valve may result in large uncertainty, however, the actual effects on final peak clad temperature in a reactor transient may be small and the large uncertainty with respect to valve modeling may be considered acceptable. Additionally, the ability to determine the adequacy of a model and the validation supporting it should be considered within a risk informed framework. Low fidelity modeling with large uncertainty may be considered adequate if the uncertainty is considered acceptable with respect to risk. In other words, models that are used to evaluate the probability of failure should be evaluated more rigorously with the intent of increasing safety margin. Probabilistic risk assessment (PRA) techniques have traditionally been used to identify accident conditions and transients. Traditional classical event tree methods utilize analysts’ knowledge and experience to identify the important timing of events in coordination with thermal-hydraulic modeling. These methods lack the capability to evaluate complex dynamic systems. In these systems, time and energy scales associated with transient events may vary as a function of transition times and energies to arrive at a different physical state. Dynamic PRA (DPRA) methods provide a more rigorous analysis of complex dynamic systems. Unfortunately DPRA methods introduce issues associated with combinatorial explosion of states. This study presents a methodology to address combinatorial explosion using a Branch-and-Bound algorithm applied to Dynamic Event Trees (DET), which utilize LENDIT (L – Length, E – Energy, N – Number, D – Distribution, I – Information, and T – Time) as well as a set theory to describe system, state, resource, and response (S2R2) sets to create bounding functions for the DET. The optimization of the DET in identifying high probability failure branches is extended to create a Phenomenological Identification and Ranking Table (PIRT) methodology to evaluate modeling parameters important to safety of those failure branches that have a high probability of failure. The PIRT can then be used as a tool to identify and evaluate the need for experimental validation of models that have the potential to reduce risk. Finally, in order to demonstrate this methodology, a Boiling Water Reactor (BWR) Station Blackout (SBO) case study is presented.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

Collision Tracking in OpenMC: Methods and Applications in Neutron Noise, Neutron Imaging, Time-of-Flight, and Multiplicity Counting

We present the development and application of a collision tracking feature within the OpenMC Monte Carlo particle transport code, designed for diverse applications such as neutron spectroscopy, scatter camera system, neutron noise, and multiplicity counting simulations. This feature enables the tracking of individual particle collisions, with potential applications in nuclear nonproliferation, reactor physics, and nuclear security. Additionally, the feature holds potential for the calibration of neutron detectors, specifically in converting light output into energy deposited within the detectors. The implementation consists of a set of filters—such as reaction type, energy, cell, and material—that constrain the set of collisions that are tracked, extensions to the Python API to enable simple input specification, and support for writing either OpenMC’s native HDF5-based format or the Monte Carlo particle list format. This feature was added to the official OpenMC release in version 0.15.3. In this work, the feature will be applied to showcase scenarios such as time-of-flight simulations, scatter-camera imaging for neutron source localization, neutron-noise analysis to extract integral kinetic parameters such as the prompt decay constant α, and multiplicity counting to estimate the mass of special nuclear materials. Ultimately, this feature aims to expand the application scope of open-source Monte Carlo particle transport codes such as OpenMC.

Monte Carlo code↗

A Monte Carlo volumetric-ray-casting estimator for global fluence tallies on GPUs

A Monte Carlo fluence estimator has been designed to take advantage of the computational power of graphical processing units (GPUs). This new estimator, termed the volumetric-ray-casting estimator, is an extension of the expectation estimator. It can be used as a replacement of the track-length estimator for the estimation of global fluence. Calculations for this estimator are performed on the GPU while the Monte Carlo random walk is performed on the central processing unit (CPU). This method lowers the implementation cost for GPU acceleration of existing Monte Carlo particle transport codes as there is little modification of the particle history logic flow. Three test problems have been evaluated to assess the performance of the volumetric-ray-casting estimator for neutron transport on GPU hardware in comparison to the standard track-length estimator on CPU hardware. Evaluation of neutron transport through air in a criticality accident scenario showed that the volumetric-ray-casting estimator achieved 23 times the performance of the track-length estimator using a single core CPU paired with a GPU and 15 times the performance of the track-length estimator using an eight core CPU paired with a GPU. Finally, simulation of a pressurized water reactor fuel assembly showed that the performance improvement was 6 times within the fuel and 7 times within the control rods using an eight core CPU paired with a single GPU.

74 ATOMIC AND MOLECULAR PHYSICS↗

Experimental data based combinatorial kinetic simulations for predictions of synergistic catalyst mixtures

Physical mixtures of catalysts can sometimes show synergistic activity which exceeds the sum of the activities of the same amount of catalysts employed separately. However, identifying such synergistic mixtures is non-trivial. Not all mixtures display synergy, and the number of combinations that are possible (even for binary mixtures of equal portions) scales very rapidly, and thus they would be costly to screen experimentally. In this work, we show that it is possible to predict synergistic mixtures using combinatorial kinetic simulations based on experimental data collected on individual catalysts. The data was first collected for conditions-of-interest under low conversions (also called near differential conditions, such that each condition approximates a small volume in a non-differential conditions reactor) to build a library of kinetic models (one model for each catalyst). This data was then used for combinatorial kinetic simulations of non-differential reactor scale conversions, and successfully predicted the qualitative behavior of two synergistic physical mixtures. The capability was utilized in the present work in the context of converting CO, C 3 H 6 , and NO species in simulated car exhaust, but is a general approach. We provide equations to estimate the costs of the screening method as well as the combinatorial kinetics method and show that the costs of identifying synergistic physical mixtures become much lower with the combinatorial kinetics method even before ten catalysts of interest. The cost of using prediction by the combinatorial kinetics method continue to become lower (relative to screening) when more catalysts are added -- and even more so if mixtures beyond two components, or consideration of multiple catalyst ratios, is of interest.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗

A Simplified Quantitative Model for Fissile Material Production from Fast Spallation Neutrons

The Accelerator Driven System (ADS) as a reactor to produce fissile material, or for partitioning & transmutation purposes, has potential as an inherently safe reactor since it can be shut down by simply turning off the accelerator. Multiple studies have focused on the intranuclear and internuclear physics involved in the spallation process for high-energy projectiles impacting high-Z targets relevant to ADS applications. To quantify spallation neutronics, and thereby fissile material production, generally requires numerical methods such as MCNP, and such calculations are specific to the geometry under consideration. This study uses published cross sections from the ENDF database and an assumed generic cylindrical geometry for the ADS target and fertile blanket to derive analytic expressions for the production of Pu-239 and U-233, without needing to run MCNP codes, yet matching published data.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

Characterization of melted fuel by neutron resonance spectroscopy

Neutrons can be used as a tool to study properties of materials and objects. An evolving activity in this field focusses on neutron induced reaction cross sections. The probability that a neutron interacts with nuclei strongly depends on the energy of the neutron. The cross sections reveal the presence of resonance structures, the energy and width of which are isotope specific. As such, these resonance structures can be used as fingerprints to determine the elemental and isotopic composition of materials and objects. They are the basis of two analytical methods which have been developed at Institute for Reference Materials and Measurements of the European Commission's Joint Research Centre (EC-JRC-IRMM): Neutron Resonance Capture Analysis (NRCA) and Neutron Resonance Transmission Analysis (NRTA). The first technique is based on the detection of gamma rays emitted during a neutron capture reaction in the sample being studied; the latter determines the fraction of neutrons transmitted through a sample positioned in a neutron beam. In the past both techniques have been applied to determine the composition of archaeological objects and to characterize nuclear reference materials. More recently a combination of NRTA and NRCA is being studied as a non-destructive method to determine the heavy metal content of particle-like debris of melted fuel that is formed in severe nuclear accidents such as the one which occurred at the Fukushima Daiichi nuclear power plant in Japan. This study is part of a collaboration between the Japan Atomic Energy Agency (JAEA) and ECJRC- IRMM and is a spin-off from the core activity of IRMM, i.e. the production of nuclear data for nuclear technology applications. This contribution focusses on a newly developed NRTA measurement station that has been set up recently at one of the flight paths of the neutron time-of-flight facility GELINA at the EC-JRC-IRMM. The basic principles of NRTA and first results of measurements at the new set up will be discussed. (authors)

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗

Measurement of neutron and gamma-ray noise in large subcritical condition for criticality safety - 14485

Various experimental methods have been proposed for subcriticality measurement and neutron noise analysis method is one of important experimental techniques for this purpose. The variance to mean value method, called Feynman-alpha method, is one of well-known noise analysis methods. In this method, variation of neutron count rate distribution from the Poisson distribution caused by fission chain reactions, which is called Y-value, is measured, however, it is difficult to measure the Y-value in large subcritical state such as nuclear fuel facility, because number of emitted neutrons by fission chain reaction is small, which causes large statistical error. We developed new experimental technique based on Feynman-alpha method combined with pulsed neutron source, called Delayed Neutron Noise Analysis (DNNA) method, to monitor the change of subcriticality. In this new method, neutron and gamma-ray counts measured only appeared in the delayed neutron region after injection of pulsed neutron are analyzed, and the prompt neutron region is completely ignored. The advantage of DNNA compared with conventional method is that spatial dependency of detector positions on measured results can be greatly decreased, which result in increase of accuracy of measured subcriticality. The validity of the DNNA method was experimentally investigated in the Kyoto University Critical Assembly (KUCA) at Kyoto University Research Reactor Institute. In the experiments, pulsed neutrons by D-T reactions are injected into the core and we used {sup 3}He detectors to measure neutron noise signals and BGO scintillation detectors to measure gamma-ray noise signals, and all data in the delayed neutron region were analyzed based on Feynman-alpha method. From the experimental results, it was found that the Y-values are clearly larger than zero in the case of existence of fuel assemblies and this method can be applied to detection of hidden special nuclear materials for example in a cargo container. (authors)

42 ENGINEERING↗