ترغب بنشر مسار تعليمي؟ اضغط هنا

A domain-specific analysis system for examining nuclear reactor simulation data for light-water and sodium-cooled fast reactors

255   0   0.0 ( 0 )
 نشر من قبل Jay Billings
 تاريخ النشر 2014
والبحث باللغة English




اسأل ChatGPT حول البحث

Building a new generation of fission reactors in the United States presents many technical and regulatory challenges. One important challenge is the need to share and present results from new high-fidelity, high-performance simulations in an easily usable way. Since modern multiscale, multi-physics simulations can generate petabytes of data, they will require the development of new techniques and methods to reduce the data to familiar quantities of interest (e.g., pin powers, temperatures) with a more reasonable resolution and size. Furthermore, some of the results from these simulations may be new quantities for which visualization and analysis techniques are not immediately available in the community and need to be developed. This paper describes a new system for managing high-performance simulation results in a domain-specific way that naturally exposes quantities of interest for light water and sodium-cooled fast reactors. It describes requirements to build such a system and the technical challenges faced in its development at all levels (simulation, user interface, etc.). An example comparing results from two different simulation suites for a single assembly in a light-water reactor is presented, along with a detailed discussion of the systems requirements and design.

قيم البحث

اقرأ أيضاً

An open-source, Python-based Temporal Analysis of Products (TAP) reactor simulation and processing program is introduced. TAPsolver utilizes algorithmic differentiation for the calculation of highly accurate derivatives, which are used to perform sen sitivity analyses and PDE-constrained optimization. The tool supports constraints to ensure thermodynamic consistency, which can lead to more accurate parameters and assist in mechanism discrimination. The mathematical and structural details of TAPsolver are outlined, as well as validation of the forward and inverse problems against well-studied prototype problems. Benchmarks of the code are presented, and a case study for extracting thermodynamically-consistent kinetic parameters from experimental TAP measurements of CO oxidation on supported platinum particles is presented. TAPsolver will act as a foundation for future development and dissemination of TAP data processing techniques.
145 - M. Osipenko , M. Ripani , G. Ricco 2015
In this paper we describe the development and first tests of a neutron spectrometer designed for high flux environments, such as the ones found in fast nuclear reactors. The spectrometer is based on the conversion of neutrons impinging on $^6$Li into $alpha$ and $t$ whose total energy comprises the initial neutron energy and the reaction $Q$-value. The $^6$LiF layer is sandwiched between two CVD diamond detectors, which measure the two reaction products in coincidence. The spectrometer was calibrated at two neutron energies in well known thermal and 3 MeV neutron fluxes. The measured neutron detection efficiency varies from 4.2$times 10^{-4}$ to 3.5$times 10^{-8}$ for thermal and 3 MeV neutrons, respectively. These values are in agreement with Geant4 simulations and close to simple estimates based on the knowledge of the $^6$Li(n,$alpha$)$t$ cross section. The energy resolution of the spectrometer was found to be better than 100 keV when using 5 m cables between the detector and the preamplifiers.
This paper develops a Nearly Autonomous Management and Control (NAMAC) system for advanced reactors. The development process of NAMAC is characterized by a three layer-layer architecture: knowledge base, the Digital Twin (DT) developmental layer, and the NAMAC operational layer. The DT is described as a knowledge acquisition system from the knowledge base for intended uses in the NAMAC system. A set of DTs with different functions is developed with acceptable performance and assembled according to the NAMAC operational workflow to furnish recommendations to operators. To demonstrate the capability of the NAMAC system, a case study is designed, where a baseline NAMAC is implemented for operating a simulator of the Experimental Breeder Reactor II during a single loss of flow accident. When NAMAC is operated in the training domain, it can provide reasonable recommendations that prevent the peak fuel centerline temperature from exceeding a safety criterion.
The CONNIE experiment is located at a distance of 30 m from the core of a commercial nuclear reactor, and has collected a 3.7 kg-day exposure using a CCD detector array sensitive to an $sim$1 keV threshold for the study of coherent neutrino-nucleus e lastic scattering. Here we demonstrate the potential of this low-energy neutrino experiment as a probe for physics Beyond the Standard Model, by using the recently published results to constrain two simplified extensions of the Standard Model with light mediators. We compare the new limits with those obtained for the same models using neutrinos from the Spallation Neutron Source. Our new constraints represent the best limits for these simplified models among the experiments searching for CE$ u$NS for a light vector mediator with mass $M_{Z^{prime}}<$ 10 MeV, and for a light scalar mediator with mass $M_{phi}<$ 30 MeV. These results constitute the first use of the CONNIE data as a probe for physics Beyond the Standard Model.
116 - X.B. Ma , F.Lu , L.Z. Wang 2014
Reactor antineutrino experiment are used to study neutrino oscillation, search for signatures of nonstandard neutrino interaction, and monitor reactor operation for safeguard application. Reactor simulation is an important source of uncertainties for a reactor neutrino experiment. Commercial code is used for reactor simulation to evaluate fission fraction in Daya Bay neutrino experiment, but the source code doesnt open to our researcher results from commercial secret. In this study, The open source code DRAGON was improved to calculate the fission rates of the four most important isotopes in fissions, $^{235}$U,$^{238}$U,$^{239}$Pu and $^{241}$Pu, and then was validated for PWRs using the Takahama-3 benchmark. The fission fraction results are consistent with those of MITs results. Then, fission fraction of Daya Bay reactor core was calculated by using improved DRAGON code, and the fission fraction calculated by DRAGON agreed well with these calculated by SCIENCE. The average deviation less than 5% for all the four isotopes. The correlation coefficient matrix between $^{235}$U,$^{238}$U,$^{239}$Pu and $^{241}$Pu were also studied using DRAGON, and then the uncertainty of the antineutrino flux by the fission fraction was calculated by using the correlation coefficient matrix. The uncertainty of the antineutrino flux by the fission fraction simulation is 0.6% per core for Daya Bay antineutrino experiment. The uncertainties source of fission fraction calculation need further to be studied in the future.
التعليقات
جاري جلب التعليقات جاري جلب التعليقات
سجل دخول لتتمكن من متابعة معايير البحث التي قمت باختيارها
mircosoft-partner

هل ترغب بارسال اشعارات عن اخر التحديثات في شمرا-اكاديميا