Commercial Fission Research Articles

Comparative analysis of different FeCrAl alloys in pressurized water reactors

Discussing and researching issues related to their safety is essential to guarantee the continuity of fission systems in generating energy for existing and new electrical matrices. Commercial fission reactors currently in operation have undergone a series of redesigns. One aspect much discussed in recent years is related to research about new fuels and cladding materials. Therefore, in this work, the feasibility of replacing Zircaloy-4 with FeCrAl-based alloys (C06M, C35M, C36M, Kanthal APMT) was evaluated considering these materials' physical–chemical, thermal, and neutronic properties. Moreover, thermal–hydraulic and neutronic simulations have been presented to compare the cladding behavior in a PWR core in steady state and transient operation conditions. The results showed, in a general way, that the analyzed alloys have equivalent properties from a physical–chemical and thermal point of view. Despite the similarity, the Kanthal APMT alloy presented the worst performance, and the C35M alloy showed more advantageous characteristics. Considering the similarities in the composition of the C06M, C35M, and C36M alloys, which have the same elements in their constitution, the alloy with the best thermal performance has the lowest aluminum content. Considering the neutronic behavior, the Kanthal APMT alloy, with more Cr percentile, presents the lowest neutron effective multiplication factor, keff. Axial and radial power distribution did not change significantly after the cladding replacement. Nevertheless, modifications must be made to keep the criticality during the fuel cycle burnup for all FeCrAl alloys.

Phase-field simulation of high-temperature corrosion of binary Zr-2.5Sn alloy

Due to the small neutron absorption cross section and excellent thermal creep performance, zirconium alloy is one of the most important cladding materials for fuel rods in commercial fission reactors. However, quantitative analysis of the effects of temperature and grain boundaries on the corrosion microstructure evolution of zirconium alloys is still needed. The establishing of a phase field simulation for the corrosion process of polycrystalline zirconium alloy and the systematical investigating of the thermodynamic influence are both very important. In this study, the phase field model of the corrosion process in zirconium alloys is developed by combining corrosion electrochemistry through calculating the interfacial energy at the metal-oxide and oxide-fluid boundaries. Then the model is used to investigate the uniform corrosion behavior on the surface of Zr-2.5Sn alloy, which demonstrates that the corrosion kinetic curve follows a cubic rule. Subsequently, the influence of temperature on the corrosion thickening curve of zirconium alloy is examined, and good agreement between simulation and experimental results is achieved. It is observed that during early stage of oxide layer formation, there is a high growth rate with minimal temperature dependence; however, as the oxide layer thickness increases, temperature becomes a significant factor affecting its growth rate, with higher temperatures resulting in faster corrosion rates. Furthermore, the effect of polycrystalline zirconium alloy matrices on corrosion rate is investigated, revealing that the grain boundaries accelerate oxide layer thickening due to enhanced oxygen diffusion rates. At metal-oxide interface, O<sup>2–</sup> bands are formed in areas with higher O<sup>2–</sup> concentration along these grain boundaries towards the metal matrix, which mainly influences oxidation-corrosion rate during the initial oxidation stage.

Development and characterization of fission chamber neutron detectors in Isfahan miniature neutron source reactor

In this research, two fission chamber neutron detectors have been designed and constructed, and their performance has been evaluated in the radiation field of the Isfahan miniature reactor neutron beam line. Each detector consists of two coaxial cylinders called anode and cathode made of stainless steel. The outer layer of the anode is covered with a very thin layer of uranium-235 oxide as fissile material, and a mixture of argon and nitrogen gas is used as the filling gas. The thickness of the deposited layer on the anode is different, so that on one of them, named FC400A, the thickness of the uranium oxide layer was more than 5 μm, and on the other one, named FC400B, it was less than 1.5 μm. A spectrum collection software has also been designed and developed in LabVIEW. The developed detectors are irradiated in the neutron field resulting from the neutron radiography beamline of the miniature neutron source reactor in the range of neutron flux of about 2E+02 to 2E+05 n.cm−2.s−1. In order to determine the sensitivity of the detection response, the results of these two detectors are also compared with a commercial fission chamber neutron detector. The results showed that the working voltage range of these detectors is in the range of 400–600 V and both of these detectors have a suitable linear response in the mentioned neutron flux range. Also, the sensitivity of FC400A and FC400B detectors is 0.1 and 1 counts/1E+04 neutrons, respectively.

Concept of accident tolerant fuel in nuclear reactors

The excellent behavior of zirconium alloy under normal operating state as well as significant advantage of neutron economy have resulted in the use of zirconium alloy as the cladding material. However, extensive metal-water reaction in the event of severe accidents and subsequent hydrogen generation can be considered to be a big disadvantage for zirconium alloy. The Fukushima Daiichi Nuclear Plant accident has brought challenges to the nuclear industry and renewed attention about the idea of accident tolerant fuels (ATFs). Concept of ATFs is to mitigate the devastating interaction of steam at high temperature with fuel clad in off-normal condition of the reactor. These materials have higher resistance to core degradation, lower oxidation kinetics and can withstand against high temperature for a longer duration of time. Being in the direct contact with the coolant the accident tolerant fuel cladding plays an important role at high temperature in off-normal conditions. Many studies have been initiated worldwide to find an advanced oxidation-resistant alternative cladding material to enhance fuel performance and safety, specifically under accident conditions. One of the candidate materials for fuel cladding in LWRs is oxidation-resistant iron alloys. The stainless steel technology and the material science have been studied extensively and several new generations of increasingly higher performance steels are now commercially available that may offer significant performance improvements over the relatively simple austenitic steels that were utilized as cladding in some early commercial fission reactors. One study has been performed to find the reactivity impact of SiC and stainless steel as clad material on various parameters. This paper presents the results of a study that has been made to understand the reactor physics implications of using SiC and stainless steel as the cladding material.

Comparative Maps of Safety Features for Fission and Fusion Reactors

Abstract The differences between nuclear fission and fusion have been discussed widely in the literature. However, little has been done to investigate the key differences in safety designs and regulatory requirements between the nuclear reactor types. In this study, an innovative methodology was successfully developed to map nuclear safety features to the fundamental safety principles set out by the nuclear regulators. Three safety cases were assessed in the mapping study, a research fusion reactor (Joint European Torus), a research fission reactor (Tsing Hua Open-pool Reactor) and a commercial fission reactor (Hinkley Point C). The graphical representation allowed a comparative analysis of the safety features and fundamental principles which revealed differences between the hazard profiles of fission and fusion reactors and provided important insights for the creation of a similar map for a future commercial fusion device.

Failure mode and reliability study for dynamic confinement and conditioning of the IFMIF-DONES radiological areas

The reference engineering design of the IFMIF-DONES nuclear Heating, Ventilation, and Air Conditioning system (HVAC) has been analyzed in terms of potential failure modes and consequences as well as life-cycle reliability and availability. The HVAC system performs the safety function of preventing lithium fire and uncontrolled releases of radioactive materials from the radiological areas. The results of the Reliability, Availability, Maintainability and Inspectability (RAMI) analysis are of high practical importance for future activities on the system optimization and safety demonstration. This paper proposes a unified approach to the Failure Mode and Effects Analysis (FMEA) that can be applied to the plant systems evaluation in terms of the failure frequency, severity, and detectability. The reliability and availability calculation has been performed for two sets of data, the U.S. NRC collection representing the industry standards of fission power plants, and the fusion-specific data gathered from several sources and integrated by Monte Carlo-based method. Inherent availability of above 99.4% has been reached for the HVAC system based on the fusion-specific data. The use of reliability data from the commercial fission plants increased the system availability to nearly 99.8%, thus indicating the potential for allocation of higher availability requirements to the HVAC functions.

High temperature zirconium alloys for fusion energy

This review considers current Zr alloys and opportunities for advanced zirconium alloys to meet the demands of a structural material in fusion reactors. Zr based materials in the breeder blanket offer the potential to increase the tritium breeding ratio above that of Fe, Si and V based materials. Current commercial Zr alloys might be considered as a material in water-cooled breeder blanket designs, due to the similar operating temperature to fission power plants. For breeder blankets designed to operate at higher temperatures, current commercial Zr alloys will not meet the high temperature strength and thermal creep requirements. Hence, Zr alloys with an operational temperature capability beyond that of current commercial fission alloys have been reviewed, specifically: binary Zr alloy systems Zr-Al, Zr-Be, Zr-Cr, Zr-Nb Zr-Ti, Zr-Si, Zr-Sn, Zr-V and Zr-W; as well as higher order Zr alloys Zr-Mo-Ti, Zr-Nb-Ti, Zr-Ti-Al-V and Zr-Mo-Sn. It is concluded that, with further work, higher order Zr alloys could achieve the required high temperature strength, alongside ductility, while maintaining a low thermal neutron cross-section. However, there is limited data and uncertainty regarding the structural performance and microstructural stability of the majority of advanced Zr alloys for temperatures 500–700 °C, at which they would be expected to operate for helium- and liquid metal-cooled breeder blanket designs.

Do proposed regulatory approaches for commercialfusion energy jeopardize successful technologydeployment?

Fusion energy has long been touted as an energy source capable of producing large amounts of clean energy without the fuel, pollution, siting, and safety constraints of other energy sources. Despite this promise, fusion energy has not come to fruition after six decades of research and development due to continuing scientific and technical challenges. Significant private investment in commercial fusion start-ups signals a renewed interest in the prospects of near-term development of fusion technology. Successful deployment of fusion energy, however, will require an appropriate regulatory framework to ensure public safety and economic viability. Initial discussions and proposals from fusion start-ups (e.g., Commonwealth Fusion Systems, General Fusion, TAE Technologies) and industry associations (e.g., Fusion Industry Association)have advocated the use of risk-informed regulations as the basis for the licensing of commercial fusion technology in the United States. These proposals are based, in part, on regulatory methods used for licensing commercial fission technology that developed in parallel with the commercial fission industry. Risk-informed regulations incorporate risk information from probabilistic safety analyses to ensure that regulations are appropriate for the actual risk of an activity. Proponents of risk-informed regulation believe that these methods can improve the economics and operation of nuclear facilities by focusing regulatory resources on addressing credible accident scenarios that are most likely to pose threats to worker and public safety. In addition to interest from industry, the U.S. Nuclear Regulatory Commission (NRC)has begun an initiative to develop a new risk-informed regulatory framework for the licensing of advanced nuclear fission power plants, but it is unclear if this framework would be applied to fusion. Despite the benefits of adopting a risk-informed framework for a mature fission industry, use of risk-informed regulations for the licensing of first-generation commercial fusion technology could be detrimental to the goal of economic near-term deployment of fusion. Commercial fusion technology has an insufficient operational and regulatory experience base to support the rapid and effective use of risk-informed regulations. More conservative regulatory analyses could instead be used for first-generation fusion facilities to facilitate more rapid deployment, and enable collection of operating experience to support future use of risk-informed regulations. A hybrid regulatory pathway that incorporates seven decades of lessons learned in commercial fission would enable the implementation of regulatory requirements that evolve with the development of commercial fusion technology, from a first-of-a-kind demonstration plant to a mature low-carbon energy source.

IRDFF-II: A New Neutron Metrology Library

High quality nuclear data is the most fundamental underpinning for all neutron metrology applications. This paper describes the release of version II of the International Reactor Dosimetry and Fusion File (IRDFF-II) that contains a consistent set of nuclear data for fission and fusion neutron metrology applications up to 60 MeV neutron energy. The library is intended to support: a) applications in research reactors; b) safety and regulatory applications in the nuclear power generation in commercial fission reactors; and c) material damage studies in support of the research and development of advanced fusion concepts. The paper describes the contents of the library, documents the thorough verification process used in its preparation, and provides an extensive set of validation data gathered from a wide range of neutron benchmark fields. The new IRDFF-II library includes 119 metrology reactions, four cover material reactions to support self-shielding corrections, five metrology metrics used by the dosimetry community, and cumulative fission products yields for seven fission products in three different neutron energy regions. In support of characterizing the measurement of the residual nuclei from the dosimetry reactions and the fission product decay modes, the present document lists the recommended decay data, particle emission energies and probabilities for 68 activation products. It also includes neutron spectral characterization data for 29 neutron benchmark fields for the validation of the library contents. Additional six reference fields were assessed (four from plutonium critical assemblies, two measured fields for thermal-neutron induced fission on 233U and 239Pu targets) but not used for validation due to systematic discrepancies in C/E reaction rate values or lack of reaction-rate experimental data. Another ten analytical functions are included that can be useful for calculating average cross sections, average energy, thermal spectrum average cross sections and resonance integrals. The IRDFF-II library and comprehensive documentation is available online at www-nds.iaea.org/IRDFF/. Evaluated cross sections can be compared with experimental data and other evaluations at www-nds.iaea.org/exfor/endf.htm. The new library is expected to become the international reference in neutron metrology for multiple applications.

An assessment of the available alternatives for fusion relevant neutron sources

The discovery of void swelling in neutron-irradiated stainless steels by Cawthorne and Fulton in 1966 showed that radiation effects would seriously affect the lifetime of fission reactors. A few years later, in the early 1970s, serious damage levels were observed in core components of the first commercial fission reactors that had been in operation for one decade. Driven by the harder neutron spectrum, the fusion community was impelled to explore the technical possibilities to make available a fusion relevant neutron source to anticipate the difficulties faced by the fission reactors community and ensure the long-term operation of a fusion reactor. In 1976, the conclusions of the first review published announced the assets and drawbacks of the different possibilities, which today despite the four decades passed and the sound technological advancements, continue being basically the same. A suitable neutron spectrum can be theoretically obtained both from plasma-based or accelerators based facilities with either gas, liquid or solids targets. Needed facilities performances, available irradiation volumes that the concept allow and cost are at stake. In the past, the cost of the most promising facility based on stripping the neutron from a deuteron on a lithium target was misleadingly peered with that of a fusion reactor. A tokamak with sufficient volume to irradiate equipment under fusion relevant conditions will be likely available in the future once commercial fusion reactors have become a reality; unfortunately, this concept does not match the needs timely, and is unrealistic w.r.t. its costs nowadays with worldwide efforts prioritizing ITER. In this article, we will focus on the technical aspects and assess the present maturity of different existing possibilities based on nowadays technologies to count with neutrons at a suitable flux and spectrum to characterize and qualify fusion materials to be timely ready with our world fusion roadmaps.

Materials research for fusion

Fusion materials research started in the early 1970s following the observation of the degradation of irradiated materials used in the first commercial fission reactors. The technological challenges of fusion energy are intimately linked with the availability of suitable materials capable of reliably withstanding the extremely severe operational conditions of fusion reactors. Although fission and fusion materials exhibit common features, fusion materials research is broader. The harder mono-energetic spectrum associated with the deuterium–tritium fusion neutrons (14.1 MeV compared to <2 MeV on average for fission neutrons) releases significant amounts of hydrogen and helium as transmutation products that might lead to a (at present undetermined) degradation of structural materials after a few years of operation. Overcoming the historical lack of a fusion-relevant neutron source for materials testing is an essential pending step in fusion roadmaps. Structural materials development, together with research on functional materials capable of sustaining unprecedented power densities during plasma operation in a fusion reactor, have been the subject of decades of worldwide research efforts underpinning the present maturity of the fusion materials research programme. For achieving proper safety and efficiency of future fusion power plants, low-activation materials able to withstand the extreme fusion conditions are needed. Here, the irradiation physics at play and fusion materials research is reviewed.

Neutron Multiplicity Measurements with 3He Alternative: Straw Neutron Detectors

Counting neutrons emitted by special nuclear material (SNM) and differentiating them from the background neutrons of various origins is the most effective passive means of detecting SNM. Unfortunately, neutron detection, counting, and partitioning in a maritime environment are complex due to the presence of high-multiplicity spallation neutrons (commonly known as “ship effect”) and to the complicated nature of the neutron scattering in that environment. A prototype neutron detector was built using 10B as the converter in a special form factor called “straws” that would address the above problems by looking into the details of multiplicity distributions of neutrons originating from a fissioning source. This paper describes the straw neutron multiplicity counter (NMC) and assesses the performance with those of a commercially available fission meter. The prototype straw neutron detector provides a large-area, efficient, lightweight, more granular (than fission meter) neutron-responsive detection surface (to facilitate imaging) to enhance the ease of application of fission meters. Presented here are the results of preliminary investigations, modeling, and engineering considerations leading to the construction of this prototype. This design is capable of multiplicity and Feynman variance measurements. This prototype may lead to a near-term solution to the crisis that has arisen from the global scarcity of 3He by offering a viable alternative to fission meters.This paper describes the work performed during a 2-year site-directed research and development (SDRD) project that incorporated straw detectors for neutron multiplicity counting. The NMC is a two-panel detector system. We used 10B (in the form of enriched boron carbide: 10B4C) for neutron detection instead of 3He. In the first year, the project worked with a panel of straw neutron detectors, investigated its characteristics, and developed a data acquisition (DAQ) system to collect neutron multiplicity information from spontaneous fission sources using a single panel consisting of 60 straws equally distributed over three rows in high-density polyethylene moderator. In the following year, we developed the field-programmable gate array and associated DAQ software. This SDRD effort successfully produced a prototype NMC with ˜33% detection efficiency compared to a commercial fission meter.

Reactor neutrino background at the proposed LAGUNA sites

The electron antineutrino background from reactors is estimated in proposed LAGUNA sites using commercial fission reactors operational in 2009. Calculations are made for an input to background studies like e.g. geoneutrino measurement uncertainty estimation.

00/01881 Evaluation of neutron energy spectrum effects and RPV thru-wall attenuation based on molecular dynamics cascade simulations

Displacement cascade formation in iron has been investigated by the method of molecular dynamics (MD) for cascade energies up to 40 keV. The results of these simulations have been used in the SPECOMP code to obtain effective, energy-dependent cross sections for two measures of primary damage production: (1) the number of surviving point defects expressed as a fraction of the those predicted by the standard secondary displacement model by Norgett, Robinson, and Torrens (NRT), and (2) the fraction of the surviving interstitials contained in clusters that formed during the cascade event. The primary knockon atom spectra for iron obtained from the SPECTER code have been used to weight these MD-based damage production cross sections in order to obtain spectrally-averaged values for several locations in commercial fission reactors and test reactors. An evaluation of these results indicates that neutron energy spectrum differences between the various environments do not lead to significant differences between the average primary damage formation parameters. In particular, spectrum-averaged defect production cross sections obtained for PWR and BWR neutron spectra were not significantly different. A representative application of the new defect production cross sections is provided by examining how they vary as a function of depth into the reactor pressure vessel wall. A slight difference was noted between the damage attenuation in a PWR vessel and a BWR vessel. This observation could be explained by a subtle difference in the energy dependence of the neutron spectra. Overall, the simulations indicate that spectrum-averaged defect production cross sections do not vary much among the various environments in light-water moderated fission reactors. As such, the results support the use of dpa as a damage correlation parameter and provide guidance for choosing the primary damage source term in kinetic embrittlement models.

Evaluation of neutron energy spectrum effects and RPV thru-wall attenuation based on molecular dynamics cascade simulations

Displacement cascade formation in iron has been investigated by the method of molecular dynamics (MD) for cascade energies up to 40 keV. The results of these simulations have been used in the specomp code to obtain effective, energy-dependent cross sections for two measures of primary damage production: (1) the number of surviving point defects expressed as a fraction of the those predicted by the standard secondary displacement model by Norgett et al., 1975 [M.J. Norgett, M.T. Robinson, and I.M. Torrens, Nucl. Eng. Des. 33, (1975) 50–54] and (2) the fraction of the surviving interstitials contained in clusters that formed during the cascade event. The primary knock-on atom (PKA) spectra for iron obtained from the specter code have been used to weight these MD-based damage production cross sections in order to obtain spectrally-averaged values for several locations in commercial fission reactors and test reactors. An evaluation of these results indicates that neutron energy spectrum differences between the various environments do not lead to significant differences between the average primary damage formation parameters. In particular, spectrum-averaged defect production cross sections obtained for PWR and BWR neutron spectra were not significantly different. A representative application of the new defect production cross sections is provided by examining how they vary as a function of depth into the reactor pressure vessel wall. A slight difference was noted between the damage attenuation in a PWR vessel and a BWR vessel. This observation could be explained by a subtle difference in the energy dependence of the neutron spectra. Overall, the simulations indicate that spectrum-averaged defect production cross sections do not vary much among the various environments in light-water moderated fission reactors. As such, the results support the use of atomic displacements per atom (dpa) as a damage correlation parameter and provide guidance for choosing the primary damage source term in kinetic embrittlement models.

Observation of uranium photo-fission by 16N decay gamma rays from water activated by D-T fusion neutrons

Radioactive 16N is generated in pure water (H 2O) which has been exposed to D-T fusion neutrons with energies near 14 MeV. The principal neutron-activation reaction is 16O(n,p) 16N. Decay by β emission (100%) of the 7.13±0.02 s 16N activity is accompanied by production of energetic gamma rays (most of them with energies>6 MeV) that are associated with the prompt radiative de-excitation of various elevated energy levels of 16O populated through the process 16N( β-) 16O. The dominant gamma-ray emission branch energies (intensities) are 6.129 MeV (67.0±0.6%) and 7.115 MeV (4.9±0.4%). A particular given intensity refers to the percent of 16N decays that produce the indicated gamma ray. Both photons possess energies that are above the photo-fission threshold for the naturally occurring isotopes of uranium, namely, 234U, 235U, and 238U. This paper reports on an experiment that demonstrated 16N decay gamma-ray induced photo-fission in a commercial fission detector that contained 0.14 g of depleted uranium (predominantly 238U with a small trace of 235U). This detector was placed near a tubing system containing circulating water that had been activated by neutron bombardment as it passed near to the target of an intense D-T fusion neutron generator. The fission detector was situated in a shielded location remote from the source of primary D-T neutrons. Other possible mechanisms leading to the generation of the observed detector events were considered. Among these were natural background radiation (e.g., cosmic rays), delayed neutrons from the decay of 17N and 18N also produced in water by D-T neutrons, secondary neutrons from ( γ,n) reactions in materials near the detector, and electronic noise or pulse pileup. All of these alternative possibilities were ultimately rejected as significant contributors to the experimental detector counts by a detailed examination of the pertinent nuclear data and by specific measurements carried out in the present study. Photo-fission of 238U remained as the predominant source of the recorded fission-detector events after elimination of these other possibilities as significant contributors of observed counts. The anticipated photo-fission yield was calculated based on consideration of the experimental setup, tabulated cross sections, and other pertinent nuclear data. This analysis yielded results which were found to be consistent with the present experimental observations to within the combined — rather large — uncertainties. This outcome further supports the conclusion that photo-fission events were indeed observed in this experiment. Some engineering implications of this investigation, and potential applications for the observed phenomenon in contemporary nuclear fusion technology, are also discussed.

Subcascade formation in displacement cascade simulations: Implications for fusion reactor materials

Displacement cascade formation in iron has been investigated by the method of molecular dynamics (MD) for cascade energies up to 40 keV, corresponding to PKA energies up to 61 keV. The results of these simulations have been used in the SPECOMP code to obtain effective, energy-dependent cross sections for two measures of primary damage production: (1) the number of surviving point defects expressed as a fraction of the those predicted by the standard secondary displacement model by Norgett, Robinson, and Torrens (NRT), and (2) the fraction of the surviving interstitials contained in clusters that formed during the cascade event. The primary knockon atom spectra for iron obtained from the SPECTER code have been used to weight these MD-based damage production cross sections in order to obtain spectrally averaged values for several locations in commercial fission reactors, materials test reactors, and a DT fusion reactor (ITER) first wall. An evaluation of these results indicates that neutron energy spectrum differences between the various environments do not lead to significant differences between the average primary damage formation parameters. This conclusion implies that the displacement damage component of radiation damage produced in a DT fusion reactor should be well simulated by irradiation in a fission reactor neutron spectrum, and that differences in nuclear transmutation production may be the primary source of uncertainty in the prediction of material performance at high doses in DT fusion reactors.

From Molecular Dynamics to Kinetic Rate Theory: A Simple Example of Multiscale Modeling

AbstractRadiation damage formation in iron has been investigated using the method of molecular dynamics simulation. The MD simulations have been used to determine primary defect production parameters for cascade energies up to 50 keV at temperatures from 100 to 900K. The energy dependence of these parameters has been used to determine appropriate neutron-energy-spectrum averaged damage production cross sections for various irradiation environments. Two applications of these effective cross sections are discussed. The first is an evaluation of neutron energy spectrum effects in commercial fission reactor pressure vessels. The second example deals with the use of these cross sections in the source term of a kinetic model used to predict void swelling and microstructural evolution. The simulation of the primary damage event by MD involves times less than 100 ps and a size scale of a few tens of nm, while the kinetic simulation encompasses several years and macroscopic sizes. This use of the MD results to develop an improved source term for rate theory modeling provides a simple example of multiscale modeling.

From Molecular Dynamics to Kinetic Rate Theory: A Simple Example of Multiscale Modeling

AbstractRadiation damage formation in iron has been investigated using the method of molecular dynamics simulation. The MD simulations have been used to determine primary defect production parameters for cascade energies up to 50 keV at temperatures from 100 to 900K. The energy dependence of these parameters has been used to determine appropriate neutron-energy- spectrum averaged damage production cross sections for various irradiation environments. Two applications of these effective cross sections are discussed. The first is an evaluation of neutron energy spectrum effects in commercial fission reactor pressure vessels. The second example deals with the use of these cross sections in the source term of a kinetic model used to predict void swelling and microstructural evolution. The simulation of the primary damage event by MD involves times less than 100 ps and a size scale of a few tens of nm, while the kinetic simulation encompasses several years and macroscopic sizes. This use of the MD results to develop an improved source term for rate theory modeling provides a simple example of multiscale modeling.

Applied electromagnetics research and application

Applied electromagnetics researches have been done in the laboratory as one of off-pile research activities. Applied electromagnetics is very important for the design and analysis of magnetic confinement fusion devices and nondestructive evaluation of nuclear fission reactors. Especially, we have developed many numerical analytic methods to solve electromagnetomechanical coupled phenomena in fusion first walls. The numerical results have been validated by the experiments. On the other hand, magnetic nondestructive evaluation plays one of the most important inspections of commercial fission reactor. We are proceeding this magnetic nondestructive evaluation research based on many advanced numerical analysis techniques and the state-of-art sensor technology. Here we describe the analysis of eddy current and magnetic damping of the fusion first wall and several new numerical and experimental techniques in the magnetic nondestructive evaluation for nuclear engineering.