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At least 1,459 records · Page 81

Modeling Fission Gas Behaviour with the BISON Fuel Performance Code

The behavior of fission gases (Xe, Kr) is of critical importance to nuclear fuel performance, because it determines both the amount of fuel swelling and the quantity of fission gas released to the fuel rod plenum. It follows that accurate models of fission gas behavior need to be developed for the calculation of fission gas swelling and release in fuel performance codes, such as the BISON code developed at Idaho National Laboratory (INL). To this end, the fission gas behavior model included in BISON is being continually extended and improved. In the present contribution, we give an account of the most recent developments and results on modeling fission gas behavior in uranium dioxide fuel with BISON.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

New Release of Fuel Performance Code, FRAPCON-3.5 and FRAPTRAN-1.5

New versions of the USNRC fuel performance codes, FRAPCON-3 and FRAPTRAN are available to members of the code users group. FRAPCON-3.5 and FRAPTRAN-1.5 include model improvements to various material property models such as gaseous swelling, primary creep, pellet relocation, cladding models, cladding ballooning, and ID oxidation. Various new capabilities have been added such as increased numbers of time steps, axial nodes, and radial nodes. The ability to model pellet chamfers has been added. The ability to specify variable axial node lengths and specify different enrichment, gadolinia content, and central hole size in each axial node has been added. Finally the capability to model refabrication of a rodlet following irradiation of a full length rod has been added to FRAPCON-3.5. This paper will describe the changes made in the new versions of FRAPCON-3 and FRAPTRAN.

FRAPCON, FRAPTRAN↗

Using redundant transactions to verify the correctness of program code execution

In the described embodiments, a processor core (e.g., a GPU core) receives a section of program code to be executed in a transaction from another entity in a computing device. The processor core sends the section of program code to one or more compute units in the processor core to be executed in a first transaction and concurrently executed in a second transaction, thereby creating a “redundant transaction pair.” When the first transaction and the second transaction are completed, the processor core compares a read-set of the first transaction to a read-set of the second transaction and compares a write-set of the first transaction to a write-set of the second transaction. When the read-sets and the write-sets match and no transactional error condition has occurred, the processor core allows results from the first transaction to be committed to an architectural state of the computing device.

Gurumurthi, Sudhanva↗

VALIDATION AND DETERMINATION OF SIGNIFICANT SIMULATION PARAMETERS USING THE SMOOTHED PARTICLE HYDRODYNAMIC CODE NEUTRINO

In recent years there has been an increase in flooding concern to nuclear power plants; thus, tools to reduce the cost of mitigation efforts are valuable to industry. Simulating flooding events is achieved through different methods and at varied levels of detail. However, these simulations must be able to match application conditions, which requires code validation as well as knowledge of which parameters affect the accuracy of the simulation. To validate and determine the significant parameters for these tools, a sloshing experiment was conducted at the George Washington University and then simulated with the smoothed particle hydrodynamic code Neutrino. The simulation setup was constructed to match the experiment as closely as possible, and the end-wall pressure results were analysed. The 90% and 50% end-wall pressure bounds were compared along with the pressure impulse. The simulation bounds fell mostly within the experimental bounds, with exceptions in insignificant low-pressure areas, such as at the beginning (~40% difference) and the end of the cycle (~40% difference). Another exception was seen at a point in the peak pressure (~75% difference). The simulated pressure impulse matched the experiments well and was within 10%. Additionally, seven parameters were qualitatively assessed and then investigated to determine their significance. The parameters’ significance was determined by sampling five values across the parameters’ range using the Risk Analysis Virtual Environment, then comparing the average pressure for each run. All seven parameters were identified as having some significance. However, the particle size, interaction-radius to particle-size ratio, and fluid settling seem to have a greater effect due to the larger fluctuation of the results. Now that the significant parameters are identified, further research is being done to quantify the significance of each parameter.

97 MATHEMATICS AND COMPUTING↗

SCALE 6.2 REACTOR PHYSICS CODE ACCURACY ASSESMENT FOR LIGHT WATER REACTOR FUEL

Verification and validation results of the SCALE 6.2 lattice physics codes TRITON/NEWT and Polaris are presented. The SCALE/KENO continuous energy Monte-Carlo code is used as a reference to compare results of more than 2,000 combinations of pressurized water reactor and boiling water reactor lattice designs and state points. Radiochemical assay measurements were also used to assess the accuracy of depletion calculations. Results show that Polaris and TRITON/NEWT using ENDF/B_VII.1 produce accurate lattice physics calculations and cross section data for core simulators.

Mertyurek, Ugur↗

New direction of nuclear code development: artificial intelligence

It has been a long-lasting aspiration of the nuclear power community to apply artificial intelligence (AI) technologies to nuclear power plants (NPPs), as AI is expected to substantially improve the efficiency, reliability, and continuity of NPPs. Along with the recent rise of AI featured with machining learning, there is an ever-increasing interest in utilizing the data-orientated technologies for nuclear power. In this chapter, we provide a brief summary of the history of AI and the previous efforts on introducing AI to nuclear reactor systems in 1980s. These histories suggest that although with a promising future, AI in NPPs also faces unique challenges, which are still not fully addressed and need special attention. We also introduced some recent studies that apply machine learning techniques to nuclear code development.Keywords: Nuclear power plants, code for nuclear power, artificial intelligence, expert system, machine learning

Lin, Lianshan↗

Subchannel codes: CTF and VIPRE-01

In this chapter, two state-of-the-art and well-used subchannel codes are presented: CTF and VIPRE-01. As part of the Consortium for Advanced Simulation of Light Water Reactors, CTF is being used for single- and two-phase flow analysis under light-water reactor normal and anticipated transient operating conditions. Accurate determination of flow distribution, pressure drop, and void content is crucial for predicting margins to thermal crisis and ensuring more efficient nuclear power plant performance. In preparation for the intended applications, CTF is extensively validated against data from experimental facilities. Meanwhile, the licensed commercial code VIPRE-01 is used to generate a baseline set of simulations for the targeted tests and the solution parameters are compared to CTF results. The closure models in CTF that are in the greatest need of improvement are highlighted and discussed.

Zhao, Xingang↗

Revisiting Huffman Coding: Toward Extreme Performance on Modern GPU Architectures

Today's high-performance computing (HPC) applications are producing vast volumes of data, which are challenging to store and transfer efficiently during the execution, such that data compression is becoming a critical technique to mitigate the storage burden and data movement cost. Huffman coding is arguably the most efficient Entropy coding algorithm in information theory, such that it could be found as a fundamental step in many modern compression algorithms such as DEFLATE. On the other hand, today's HPC applications are more and more relying on the accelerators such as GPU on supercomputers, while Huffman encoding suffers from low throughput on GPUs, resulting in a significant bottleneck in the entire data processing. In this paper, we propose and implement an efficient Huffman encoding approach based on modern GPU architectures, which addresses two key challenges: (1) how to parallelize the entire Huffman encoding algorithm, including codebook construction, and (2) how to fully utilize the high memory-bandwidth feature of modern GPU architectures. The detailed contribution is fourfold. (1) We develop an efficient parallel codebook construction on GPUs that scales effectively with the number of input symbols. (2) We propose a novel reduction based encoding scheme that can efficiently merge the codewords on GPUs. (3) We optimize the overall GPU performance by leveraging the state-of-the-art CUDA APIs such as Cooperative Groups. (4) We evaluate our Huffman encoder thoroughly using six real-world application datasets on two advanced GPUs and compare with our implemented multithreaded Huffman encoder. Experiments show that our solution can improve the encoding throughput by up to 5.0× and 6.8× on NVIDIA RTX 5000 and V100, respectively, over the state-of-the-art GPU Huffman encoder, and by up to 3.3× over the multithread encoder on two 28-core Xeon Platinum 8280 CPUs.

Tian, Jiannan↗

Demystify the 1.1 Factor for Ultimate Tensile Strength Above Room Temperature in Development of the Code Stress Tables

As the Boiler and Pressure Vessel Code increasingly gains popularity worldwide, it is not uncommon that many users are mystified by the 1.1 factor required for ultimate tensile strength above room temperature in developing the allowable design stresses of Section II Part D Stress Tables. Questions often arise about the origin and purpose of the factor, the reason for it not applied to the yield strength, its due considerations when evaluating an alloy for acceptance to construction use from the perspective of the alloy’s tested tensile strength value, and more. Not knowing reliable sources for explicit explanations, some users tend to misinterpret the factor, abuse its application, incorrectly define their alloy acceptance or qualification criteria, and mistakenly manipulate the safety margin in structural component design.To help the Code users who struggle with these frequent and confusing issues, particularly those in nuclear industry where rigorous criteria are required for component design and alloy acceptance or qualification, this paper is intended to demystify the 1.1 factor and facilitate knowledgeable interpretation and use of the Section II Part D Stress Tables as well as relevant Mandatory Appendices.A brief review is first given on the background of the 1.1 factor, followed by a summary of its application in the Stress Tables. The provenance and purpose of the factor are then discussed in detail with graphic examples. Finally, current applicability and necessity of the factor are considered through a demonstration using example alloys.

Ren, Weiju↗

Modernization of ORIGEN Library Creation in the SCALE Computer Code

The modernization of SCALE features has been a target for recent SCALE releases. ORIGEN, the activation and depletion code in SCALE, is one such legacy code that has been targeted for modernization. A part of these modernization efforts has been to develop capabilities within ORIGEN to directly read ENDF-formatted data files. In addition, the input format for the COUPLE library creation utility was rewritten to a format more accessible to newer SCALE users. Finally, the ability of COUPLE to create ORIGEN libraries without sublibrary data has been added.KEYWORDS: ORIGEN, depletion, SCALE, API

Hart, Shane↗

3D Source Reconstruction Using Coded Aperture Gamma-Ray Imaging

Recent measurements with coded-aperture imagers demonstrate material mass determination in a holdup setting to an accuracy within a few percent. This capability is of particular interest to the Surplus Plutonium Disposition (SPD) project, which aims to dilute and dispose of surplus plutonium oxide. Gamma-ray imagers can be used to determine holdup without interrupting normal operations. In this work, we examine techniques for 3D source localization and mass determination using gamma-ray imagers. Coded-aperture imagers provide excellent source localization within the 2D image plane; however, multiple imagers operating in tandem are necessary to identify source location in 3D space. A Maximum Likelihood Expectation-Maximization (MLEM) method for fitting detector mappings is a powerful tool for accomplishing this task. MLEM allows 3D source localization to be simultaneously constrained using multiple gamma-ray imagers by constructing the basis for the MLEM fit using detector mappings from different detector locations stitched together. Each of these basis points represents a singular response from a source in 3D space and is generated using Monte Carlo simulations of sources placed individually at different locations throughout the imager’s field of view. Additionally, implementing knowledge of the physical equipment in the simulations of the glovebox used for the SPD project incorporates attenuation effects that are needed to calculate material holdup.

Laminack, Alex↗

Wind Turbine Noise Code Benchmark: A Comparison and Verification Exercise: Preprint

In a number of institutions and companies, researchers and engineers are developing modeling numerical frameworks that are used to predict the aerodynamic noise emission from a wind turbine rotor. The simulation codes vary from empirically-tuned engineering models to high-fidelity computational ones. Their common feature is the fact that they all specifically model the main aerodynamic noise mechanisms occurring at the rotating blades, namely trailing edge noise and turbulent inflow noise. Nevertheless, it can be expected that these different modeling frameworks will produce different results for a same given rotor design, and identical operating conditions. Trailing edge noise is put at the forefront of the study as it is recognized as the main audible source of noise from wind turbines. The present benchmark aims at comparing the results from the different modeling approaches, and drawing some conclusions from these comparisons. This effort, denoted as Wind Turbine Noise Code benchmark, was initiated in 2019 as a joint-activity between the IEA Wind Task 39 (Quiet Wind Turbine Technology) and Task 29 (Detailed Aerodynamics of Wind Turbines, now Task 47). In addition to the investigation of the noise emissions themselves, the rotor aerodynamic characteristics are investigated as they have a significant impact on the noise generation mechanisms. A number of test cases are defined and the aerodynamic and aeroacoustic predictions from the various models are compared. There exist some discrepancies between the different methods, but it is difficult to conclude if one methodology is better than another in order to design a wind turbine with noise as a constraint.

aeroacoustics↗

On-the-fly generation of differential resonance scattering probability distribution functions for Monte Carlo codes

Current Monte Carlo codes use one of three models to model neutron scattering in the epithermal energy range: (1) the asymptotic scattering model, (2) the free gas scattering model, or (3) the S({alpha},{beta}) model, depending on the neutron energy and the specific Monte Carlo code. The free gas scattering model assumes the scattering cross section is constant over the neutron energy range, which is usually a good approximation for light nuclei, but not for heavy nuclei where the scattering cross section may have several resonances in the epithermal region. Several researchers in the field have shown that using the free gas scattering model in the vicinity of the resonances in the lower epithermal range can under-predict resonance absorption due to the up-scattering phenomenon. Existing methods all involve performing the collision analysis in the center-of-mass frame, followed by a conversion back to the laboratory frame. In this paper, we will present a new sampling methodology that (1) accounts for the energy-dependent scattering cross sections in the collision analysis and (2) acts in the laboratory frame, avoiding the conversion to the center-of-mass frame. The energy dependence of the scattering cross section was modeled with even-ordered polynomials to approximate the scattering cross section in Blackshaw's equations for the moments of the differential scattering PDFs. These moments were used to sample the outgoing neutron speed and angle in the laboratory frame on-the-fly during the random walk of the neutron. Results for criticality studies on fuel pin and fuel assembly calculations using these methods showed very close comparison to results using the reference Doppler-broadened rejection correction (DBRC) scheme. (authors)

97 MATHEMATICS AND COMPUTING↗

Application of power time-projection on the operator-splitting coupling scheme of the TRACE/S3K coupled code

This paper reports refinement studies on the temporal coupling scheme and time-stepping management of TRACE/S3K, a dynamically coupled code version of the thermal-hydraulics system code TRACE and the 3D core simulator Simulate-3K. The studies were carried out for two test cases, namely a PWR rod ejection accident and the Peach Bottom 2 Turbine Trip Test 2. The solution of the coupled calculation, especially the power peak, proves to be very sensitive to the time-step size with the currently employed conventional operator-splitting. Furthermore, a very small time-step size is necessary to achieve decent accuracy. This degrades the trade-off between accuracy and performance. A simple and computationally cheap implementation of time-projection of power has been shown to be able to improve the convergence of the coupled calculation. This scheme is able to achieve a prescribed accuracy with a larger time-step size. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

The development and testing of the HELIOS-2 lattice code for use in online cross section generation

This work presents the implementation and testing of the fixed k{sub eff} calculation, with a fixed incoming current boundary condition (fixed k{sub eff} calculation) in the HELIOS-2 lattice physics code. The fixed k{sub eff} numerical algorithm and its implementation in HELIOS-2 are described. The newly implemented feature in the code was tested on a range of test problems that were based on the 2D C5G7 benchmark. Good results were obtained and showed that the implementation performed consistently across the series of test problems. (authors)

97 MATHEMATICS AND COMPUTING↗

Development of a S{sub n} transport code based on discontinuous finite element method and coarse mesh finite difference formulation

The methods and performance of a three-dimensional S{sub n} transport code employing the Discontinuous Finite Element Method (DFEM) and the Coarse Mesh Finite Difference (CMFD) formulation are presented. The mesh generator GMSH and a post processing visualization tool Visit are combined with the code for flexible geometry processing and versatile visualization. The CMFD method for DFEM Sn applications is formulated and the performance of the CMFD acceleration of eigenvalue calculations is demonstrated for a simple set of neutron transport problems. (authors)

97 MATHEMATICS AND COMPUTING↗

Generation of discrete scattering cross sections and demonstration of Monte Carlo charged particle transport in the Milagro IMC code package

A new method for generating discrete scattering cross sections to be used in charged particle transport calculations is investigated. The method of data generation is presented and compared to current methods for obtaining discrete cross sections. The new, more generalized approach allows greater flexibility in choosing a cross section model from which to derive discrete values. Cross section data generated with the new method is verified through a comparison with discrete data obtained with an existing method. Additionally, a charged particle transport capability is demonstrated in the time-dependent Implicit Monte Carlo radiative transfer code package, Milagro. The implementation of this capability is verified using test problems with analytic solutions as well as a comparison of electron dose-depth profiles calculated with Milagro and an already-established electron transport code. An initial investigation of a preliminary integration of the discrete cross section generation method with the new charged particle transport capability in Milagro is also presented. (authors)

97 MATHEMATICS AND COMPUTING↗

Generation of SFR few-group constants using the Monte Carlo code Serpent

In this study, the Serpent Monte Carlo code was used as a tool for preparation of homogenized few-group cross sections for the nodal diffusion analysis of Sodium cooled Fast Reactor (SFR) cores. Few-group constants for two reference SFR cores were generated by Serpent and then employed by nodal diffusion code DYN3D in 2D full core calculations. The DYN3D results were verified against the references full core Serpent Monte Carlo solutions. A good agreement between the reference Monte Carlo and nodal diffusion results was observed demonstrating the feasibility of using Serpent for generation of few-group constants for the deterministic SFR analysis. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗