The European Journal of Physics N (EPJ-N)
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Technical note: stable and unstable reactors
No full textIt is well known that a reactor is stable if the core reactivity decreases with the core power. This is the case for many types of reactors, including the PWR. However, this was not the case for the RBMK (Reaktor Bolshoy Moshchnosti Kanalniy) which could be unstable at low power. What does it mean precisely? By using a 2 × 2 system of non-linear ordinary differential equations we show that naturally (i.e. without using the control rods), with the same reactivity injection, if the initial power is lowered, then the final power may be higher, which is a rather unusual behaviour
REactor Safety Analysis ToolboX RESA-TX
No full textThe REactor Safety Analysis ToolboX RESA-TX is a software and data package in development that combines the automatisation of all established procedures for deterministic safety analysis (DSA), the integration of expert know-how and a large database including the most relevant information required for conducting a DSA. In the current state of the art, DSA is a complex and thus error-prone process that is highly time-consuming and repetitive. The reliability of the result is strongly dependent on the availability of plant data and expert know-how. The idea of developing the RESA-TX toolbox arose at GRS to cope with these conditions. The innovative approach proposes an automated and standardised procedure, supported by a large database of plant design characteristics, plant behaviour, regulatory rules and DSA expert knowledge incorporated within the tool. Its application allows the end user to automatically generate and verify an input deck, as well as conduct design basis accident (DBA) calculations for a certain design with highly reduced manual intervention. The databases can be extended depending on available information or other boundary conditions. A heuristic approach is integrated into the model generation process, where users often suffer from a lack of information about the facility under consideration. These heuristics can be replaced when higher information quality is available or enhanced over time, which can lead to more reliable results with increasing usage of the tool. As a result, the application of RESA-TX could highly increase the efficiency of the DSA process, reducing both repetitiveness as well as user-induced errors. This in return will lead to an improvement in the quality of the analysis and reliability of the results. In consequence, RESA-TX will allow for a DSA to be conducted more frequently in situations where time or budget was a limitation before, thereby contributing to an increase in reactor safety
Templates of expected measurement uncertainties for average prompt and total fission neutron multiplicities
No full textIn this paper, we provide templates of measurement uncertainty sources expected to appear for average prompt- and total-fission neutron multiplicities,
and
, for the following measurement types: absolute manganese-bath experiments for
, absolute and ratio liquid-scintillator measurements for
. These templates also suggest a typical range of these uncertainties and their correlations based on a survey of available experimental data, associated literature, and feedback from experimentalists. In addition, the information needed to faithfully include the associated experimental data into the nuclear-data evaluation process is provided
Non-destructive verification of materials in waste packages using QUANTOM
No full textThe nuclear and non-nuclear industry has produced a considerable amount of low and intermediate-level radioactive wastes during the last decades. The material characterization of waste packages recently became more and more important in order to dispose of these waste packages in a final underground repository. Material characterization remains an indispensable criterion to prevent pollution of the groundwater with toxic materials and is usually required by the national licensing and supervisory authorities. Information on the nature of waste materials can be obtained based on existing documentation or, if the documentation is insufficient, on further destructive or non-destructive analysis. Non-destructive methods are to be preferred to minimize radiation exposures of operating personnel as well as costs. Existing non-destructive techniques (Gamma scanning, X-ray, active/passive neutron counting, muon tomography) do not allow the identification of non-radioactive hazardous substances. An innovative non-destructive measurement system called QUANTOM® (QUantitative ANalysis of TOxic and non-toxic Materials) has been developed. It is based on the prompt and delayed gamma neutron activation analysis (P&DGNAA). This technology is able to identify and quantify the elemental composition (Cd, Cu, B, Pb, Hg, Fe, Al, …) in radioactive packages such as 200-l radioactive drums. This information helps waste producers verify the content of their radioactive wastes, especially regarding the presence of hazardous substances. Different reference materials have been analysed by means of the same technology (P&DGNAA) at the research reactor of BUDAPEST. A comparison of those results for five reference materials is presented. The results show a very good agreement between QUANTOM® and standardized reference analyses
Templates of expected measurement uncertainties for (n, xn) cross sections
No full textA template is provided for evaluating experimental uncertainties for neutron elastic and inelastic scattering cross sections and γ-ray production cross sections from (n, xn) measurements at laboratories with monoenergetic or white neutron sources. A typical range of uncertainties is presented for experiments detecting the scattered neutrons or the resulting de-excitation γ rays based on a survey of available data and input from many experimentalists and theorists with extensive knowledge in the field. Models commonly used to evaluate the resulting cross-sections are also discussed. Suggestions are made regarding what experimental and uncertainty information is needed for data evaluations and should be included when reporting experimental (n, xn) cross sections. Uncertainty values and correlations are recommended if these values cannot be estimated for past data from the literature
Templates of expected measurement uncertainties for prompt fission neutron spectra
No full textIn this paper, we provide templates of uncertainty sources expected to appear for three measurement types of prompt fission neutron spectra (PFNS): (1) shape measurements, (2) clean-ratio shape, that is the monitor PFNS are measured in nearly exactly the same surrounding as the PFNS of interest, and (3) indirect ratios, where the detector efficiency is backed out from PFNS monitor measurements. Information is also listed that is needed to faithfully include PFNS in nuclear data evaluations to guide experimenters on how to best report data and metadata for their measurements. These templates also suggest a typical range of pertinent uncertainty values and their correlations in case realistic uncertainties cannot be estimated from information on the measurement itself. The templates were based on a literature review, information found in EXFOR for 252Cf, 235, 238U, and 239Pu PFNS, and enhanced by expertise from experimenters contributing to these PFNS templates
Templates of expected measurement uncertainties
No full textThe covariance committee of CSEWG (Cross Section Evaluation Working Group) established templates of expected measurement uncertainties for neutron-induced total, (n,γ), neutron-induced charged-particle, and (n,xn) reaction cross sections as well as prompt fission neutron spectra, average prompt and total fission neutron multiplicities, and fission yields. Templates provide a list of what uncertainty sources are expected for each measurement type and observable, and suggest typical ranges of these uncertainties and correlations based on a survey of experimental data, associated literature, and feedback from experimenters. Information needed to faithfully include the experimental data in the nuclear-data evaluation process is also provided. These templates could assist (a) experimenters and EXFOR compilers in delivering more complete uncertainties and measurement information relevant for evaluations of new experimental data, and (b) evaluators in achieving a more comprehensive uncertainty quantification for evaluation purposes. This effort might ultimately lead to more realistic evaluated covariances for nuclear-data applications. In this topical issue, we cover the templates coming out of this CSEWG effort–typically, one observable per paper. This paper here prefaces this topical issue by introducing the concept and mathematical framework of templates, discussing potential use cases, and giving an example of how they can be applied (estimating missing experimental uncertainties of 235U(n,f) average prompt fission neutron multiplicities), and their impact on nuclear-data evaluations
Reflections on achievements, activities, and emerging issues of strategic nature within the current and future EURATOM RTD programme on radioactive waste management
No full textThis paper provides a review of and some reflections on radioactive waste management activities (including disposal) at the strategic level in connection with the ongoing ‘European Joint Programme on Radioactive Waste management – EURAD’ and the European ‘PREDIS’ project on pre-disposal issues. The review took advantage of the large number of contributions made during the FISA 2022/EURADWASTE ’22 conference. The paper addresses the key characteristics of EURAD and PREDIS and highlights some of the key strengths of Joint Programming in supporting the member states with implementing waste management activities. Then, it discusses topics of strategic importance for waste management and the contributions of EURAD and PREDIS to these topics. This includes a summary of waste management strategies, the current status of implementing disposal solutions, the importance of knowledge management (taking the long duration of disposal programmes into account) and the importance of societal support of ongoing and future waste management activities. Finally, some remarks are made about issues of importance when organizing future joint activities on radioactive waste management at the European level
Coupling the GUARDYAN code to subchanflow
No full textGUARDYAN is a GPU-based dynamic Monte Carlo code developed at the Budapest University of Technology and Economics, Hungary. Dynamic Monte Carlo computes the neutron population evolution by calculating the direct time dependence of the neutron histories in multiplying systems. Some well-established Monte Carlo codes have DMC versions with coupling to Thermal-Hydraulic solvers. GUARDYAN has the computational advantage of applying GPUs, thus calculation burden can be carried by commonly available hardware, and is capable of handling power plant size systems for kinetics problems. GUARDYAN has been recently coupled to the subchannel thermal-hydraulics code SUBCHANFLOW in order to carry out dynamic calculations with TH feedback. This paper describes some convergence studies regarding reaching the initial equilibrium state. A literature-suggested set of TH input settings and high sample numbers resulted in very low statistical errors of the power estimates and stable global measures (L2) of power release, fuel, and coolant temperatures for both static and dynamic convergence. Dynamic mode low-sample simulations provided surprisingly stable global L2 measures, correct fuel temperatures, and power release, while coolant temperatures were off, without any indication of the incorrectness of the result. Static convergence showed an alternating, fluctuating L2 behavior that did not affect the final stable state
Preservation of kinetics parameters generated by Monte Carlo calculations in two-step deterministic calculations
No full textThe generation of accurate kinetic parameters such as mean generation time Λ and effective delayed neutron fraction βeff via Monte Carlo codes is established. Employing these in downstream deterministic codes warrants another step to ensure no additional error is introduced by the low-order transport operator when computing forward and adjoint fluxes for bilinear weighting of these parameters. Another complexity stems from applying superhomogenization (SPH) equivalence in non-fundamental mode approximations, where reference and low-order calculations rely on a 3D full core model. In these cases, SPH factors can optionally be computed for only part of the geometry while preserving reaction rates and K-effective, but the impact of such approximations on kinetics parameters has not been thoroughly studied. This paper aims at studying the preservation of bilinearly-weighted quantities in the Serpent–Griffin calculation procedure. Diffusion and transport evaluations of IPEN/MB-01, Godiva, and Flattop were carried out with the Griffin reactor physics code, testing available modeling options using Serpent-generated multigroup cross sections and equivalence data. Verifying Griffin against Serpent indicates sensitivities to multigroup energy grid selection and regional application of SPH equivalence, introducing significant errors; these were demonstrated to be reduced through the use of a transport method together with a finer energy grid