Best-Estimate for System Codes (BeSYC): A New Software to Perform Best-Estimate Plus Uncertainty Analyses with Thermal-Hydraulic and Safety System Codes for Both Fusion and Fission Scenarios

The development and the validation of old and new software in relevant DEMO reactor conditions have been exploited in the latest years within the EUROfusion Consortium. The aim was to use—if possible—the software already validated for fission reactors and to fill the gaps with new ad-hoc software. As contribution to this effort, the Karlsruhe Institute of Technology (KIT) developed and tested a novel software to apply the Best-Estimate Model Calibration and Prediction through Experimental Data Assimilation methodology to the system codes RELAP5-3D, MELCOR 1.8.6, and MELCOR 2.2. This software is called Best-estimate for SYstem Codes (BeSYC), and it is developed as a MATLAB App. The application is in charge of applying the mathematical framework of the methodology, writing and executing the code runs required by the methodology, and printing the obtained results. The main goal of BeSYC is to wrap up the methodology in a software suitable to be used by any user through a simple graphical user interface. Albeit developed in the fusion research context, BeSYC can be applied to any reactor/scenario type supported by the specific system code. The goals of BeSYC, the mathematical framework, the main characteristics, and the performed verification and validation activities are described in this paper.

Improving the Quantification of Deuterium in Zirconium Alloy Atom Probe Tomography Data Using Existing Analysis Methods

Zirconium alloys are common fuel claddings in nuclear fission reactors and are susceptible to the effects of hydrogen embrittlement. There is a need to be able to detect and image hydrogen at the atomic scale to gain the experimental evidence necessary to fully understand hydrogen embrittlement. Through the use of deuterium tracers, atom probe tomography (APT) is able to detect and spatially locate hydrogen at the atomic scale. Previous works have highlighted issues with quantifying deuterium concentrations using APT due to complex peak overlaps in the mass-to-charge-state ratio spectrum between molecular hydrogen and deuterium (H2 and D). In this work, we use new methods to analyze historic and simulated atom probe data, by applying currently available data analysis tools, to optimize solving peak overlaps to improve the quantification of deuterium. This method has been applied to literature data to quantify the deuterium concentrations in a concentration line profile across an α-Zr/deuteride interface.

Development of neutron scattering kernels for cold neutron reflector materials

The newest neutron scattering applications are highly intensity-limited techniques that demand reducing the neutron losses between source and detectors. In addition, the nuclear industry demands more accurate data and procedures for the design and optimization of advanced fission reactors, especially for the treatment of fuel and moderator materials. To meet these demands, it is necessary to improve the existing calculation tools, through the generation of better models that describe the interaction of neutrons with the systems of interest. The Neutron Physics Department at Centro Atomico Bariloche (CNEA, Argentina) has been developing over the time new models for the interaction of slow neutrons with materials, to produce scattering kernels and cross section data in the thermal and cold neutron energy region. Besides the studies carried out on neutron moderators, we have recently begun looking at materials that could serve as efficient neutron reflectors over those energy ranges. In this work we present the results of transmission and scattering experiments on diamond nanopowder and magnesium hydride, carried out simultaneously at the VESUVIO spectrometer (ISIS, UK), and compare them with newly generated cross-section libraries.

Nuclear energy on the 21st century in climate limits context

Nuclear energy system development in the current century is considered in the article. The importance of this issue both for the future and for the choice of ways to solve current energy problems is high, which is largely due to the increasing environmental restrictions. Nuclear energy using promises negative impact of the energy system on enviroment minimizing. However, the danger of radiative pollution of the environment is a significant deterrent to the large scale industry development. The article discusses the problem of the nuclear power system development in the variant of the coordinated use of fission reactors for energy production and hybrid thermonuclear reactors for production of artificial fuel from thorium raw materials for fission reactors. Estimates are presented to substantiate the preferences of the structure of the nuclear power system under consideration.

Preliminary evaluations of the environmental impact for the production of 99Mo by fusion neutrons

ENEA is developing an accelerator-driven 14 MeV neutron source exploiting the deuterium–tritium fusion reaction to produce 99Mo medical radioisotope as an alternative production route not based on fission reactors. It is expected that, during normal operation, a number of radionuclides, generated by means of neutron irradiation on the raw material (natural Molybdenum), will be produced and managed. The present manuscript, as foreseen by national law, discusses a hypothetical scenario to test the environmental screening models, in turn evaluating the mechanisms and parameters which affect and control the path of liquid effluents potentially released during normal operation of the facility. The aim is to estimate the amount of radioactivity to be operated and the fraction potentially discharged in this hypothetical scenario, so as to ensure that the radioactive material can be managed without any risk for the population and the environment, according to national regulations and thoroughly fulfilling the international guidelines.

CEA-JSI Experimental Benchmark for validation of the modeling of neutron and gamma-ray detection instrumentation used in the JSI TRIGA reactor

Constant improvements of the computational power and methods as well as demands of accurate and reliable measurements for reactor operation and safety require a continuous upgrade of the instrumentation. In particular, nuclear sensors used in nuclear fission reactors (research or power reactors) or in nuclear fusion facilities are operated under intense mixed neutron and gamma-ray fields, and need to be calibrated and modeled to provide selective and accurate neutron and gamma-ray measurements. The French Atomic Energy and Alternative Energies Commission (CEA) and the Jožef Stefan Institute (JSI) have started an experimental program dedicated to a detailed experimental benchmark with analysis using Monte Carlo particle transport calculations and a series of neutron and gamma-ray sensor types used in the JSI TRIGA Mark II reactor. CEA has setup a simplified TRIPOLI-4® modeling scheme of the JSI TRIGA reactor based on the information available in the IRPhEP benchmark in order to facilitate analysis of future neutron and gamma-ray measurements. These allow the CEA to perform a TRIPOLI-4 instrumentation calculation scheme benchmarked with the JSI MCNP model. This paper presents the main results of this CEA calculation scheme application and the analysis of their comparison to the JSI results obtained in 2012 with the MCNP5 & ENDF/B-VII.0 calculation scheme. This paper will conclude with some information about the new experimental program to be carried out in 2022 in the TRIGA reactor core.

COMPARISON OF WASTE DUE TO IRRADIATED STEELS IN THE ESFR AND DEMO

For either nuclear fusion or generation IV fission reactors to be viable as a commercial energy source the decommissioning and waste disposal solutions must be considered during the design. A multi-step simulation process combining Monte Carlo Neutron Transport simulations with inventory simulations have been performed to estimate the activation of steels in key reactor components of the European Sodium-cooled fast Reactor (ESFR). Waste classifications based on UK waste disposal regulations have been applied to the key components to estimate the expected masses of low level and intermediate level waste. The use of reduced activation steels, EUROFER and F82H, in reactor components external to the core results in a factor of 10 reduction in the percentage of waste classified as Intermediate Level Waste (ILW). Waste estimates are compared to existing waste estimates for the European Demonstration fusion reactor (DEMO). The ESFR has a lower percentage of ILW per total reactor mass due to irradiated steels compared to DEMO. However, there is no Higher Activity Waste (HAW) associated with DEMO, compared with arisings from the ESFR spent fission fuel.

A Review of In Situ Isotopic Studies of the Oklo and Bangombé Natural Fission Reactors Using Microbeam Analytical Techniques

Isotopic analyses of elements in the natural reactor materials have often been performed to understand the distribution behaviors of the fission products and to evaluate the function of nuclear reactions since the first discovery of a natural reactor in 1972. Several types of unique microminerals, including significant amounts of fission products, have been found in and around the Oklo and the Bangombé natural reactors. In the past two decades, microbeam techniques using ion and laser probe facilities have been effectively applied for the in situ isotopic analyses of individual microminerals to investigate the migration behaviors of fissiogenic radioisotopes produced in the reactors. This paper presents a review of interpretations of the isotopic results of microminerals found in and around the natural reactors.