The European Journal of Physics N (EPJ-N)
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UC1 sampling plan, liquid waste storage tanks, JRC Ispra
No full textThe objective of INSIDER work package 3 (WP 3) is to draft a sampling guide for initial nuclear site characterization in constrained environments, based on a statistical approach. In this paper, deliverable 3.4 (D 3.4) is presented for WP 3, where the strategy developed in deliverables 3.1 (D 3.1) to 3.3 (D 3.3) is applied to the first of three reference use cases representative of existing decommissioning scenarios. The present discussion focuses on use case 1 (UC1): the liquid waste storage facility at the JRC site of Ispra (Italy). The proposed characterization strategy developed in D 3.2 is applied in a step by step approach to analyse the pre-existing information (obtained through the use of a pre-sampling questionnaire), and to utilise the available inputs towards the development of a sampling plan sufficient for allowing radiological characterization. The proposed sampling plan follows a three-step approach, i.e. determination of possible elevation in activity concentration by non-destructive testing, biased sampling of layers identified, and finally unbiased sampling after mixing of tank contents
The knowledge management on the design of a generation IV sodium fast reactor project at CEA. The case and methodology applied on the Astrid project
No full textFrom 2010 to 2019, the French Alternative Energies and Atomic Commission (CEA) associated with industrial partners realized the Basic Design of a prototype Sodium Fast Reactor. This project was called ASTRID (ASTRID for Advanced Sodium Technological Reactor for Industrial Demonstration). ASTRID design studies were financed through governmental funds until the end of the basic design. These funds covered also the design studies for the core manufacturing workshop, the refurbishment or construction of large test loops. One year before the term of this Basic Design phase (in 2018), industrial partners, CEA and the French State conducted a review of fast neutrons reactors and fuel cycle strategy. The review which is now translated into the Multiannual Energy Program concluded that the perspective of industrial deployment of Fast Reactors is more distant. Yet it has been concluded to keep this option open, requiring to maintain competences, and to progress on technological barriers and further develop know-how. The strategy for complete closure of nuclear fuel cycle is maintained as a long-term sustainability objective (in the second half of the 21st century). Therefore, as a direct consequence of this decision, the ASTRID project stopped at the end of 2019 at its Basic Design phase. Quickly the question raised on the Knowledge Management (KM) and Know-How capitalization of the huge amount of studies and results realized during ten years (around 23 000 technical documents). Moreover the challenge was to realize this KM process in less than one year, before the ASTRID project team definitive split. The paper is presenting an innovative KM methodology which has been created and specifically performed on the ASTRID project. It is based on a series of interviews and video recordings, all transformed into some New KM tools called “MOOK” (MOOK for Management of Organized Online Knowledge). All these MOOKs considered as “data rich contents” are then inter-connected and linked by the ASTRID Product Breakdown Structure to some fundamental documents, for a comprehensive and quick mapping of the project. They finally form an efficient KM tool recorded in a PLM Software (PLM for Product Lifecycle Management). Thus the ASTRID project team has realized a high level and easy-to-use “GPS” (Global Positioning System) tool to keep the ASTRID history, context, knowledge and know-how for years. This KM methodology can be easily adapted to other nuclear projects and needs
Effect of the [U(IV)]/[U(III)] ratio on selective chromium corrosion and tellurium intergranular cracking of Hastelloy N alloy in the fuel LiF-BeF
No full textEffect of the [U(IV)/U(III)] ratio of fuel salt on selective chromium corrosion and tellurium intergranular cracking (IGC) of Hastelloy N alloy in the LiF-BeF2-UF4 salt mixture was investigated. The chromium corrosion of Hastelloy N alloy is caused by the oxidation of chromium on the alloy surface by reaction with UF4. The tellurium IGC of Hastelloy N alloy is caused by the diffusion of tellurium along the grain boundaries with the formation of unstable tellurides with based metals and alloying additives. Results indicate that the selective chromium corrosion and the tellurium IGC of the Hastelloy N alloy in fuel salt can be controlled by the [U(IV)]/[U(III)] ratio. The tellurium IGC of Hastelloy N alloy exposed in fuel LiF-BeF2-UF4 salt can be avoided. For temperatures up to 760 °C the selective chromium corrosion can be minimized to the acceptable level when the [U(IV)]/[U(III)] ratio of fuel salt is bellow 30–40
Needs of countries with longer timescale for deep geological repository implementation
No full textCountries operating nuclear power plants have to deal with the tasks connected to spent fuel and high-level radioactive waste management. There is international consensus that, at this time, deep geological disposal represents the safest and most sustainable option as the end point of the management of high-level waste and spent fuel considered as waste. There are countries with longer timescale for deep geological repository (DGR) implementation, meaning that the planned date of commissioning of their respective DGRs is around 2060. For these countries cooperation, knowledge transfer, participation in RD&D programmes (like EURAD) and adaptation of good international practice could help in implementing their own programmes. In the paper the challenges and needs of a country with longer implementation timescale for DGR will be introduced through the example of Hungary
Development of a user-friendly guideline for data analysis and sampling design strategy
No full textWithin the H2020 INSIDER project, the main objective of work package 3 (WP3) is to draft a sampling guide for initial nuclear site characterization in constraint environments, before decommissioning, based on a statistical approach. The second task of WP3 aims at developing a strategy for sampling in the field of initial nuclear site characterization in view of decommissioning, with the most important goal to guide the end user to appropriate statistical methods (including, but not limited to those identified during the first overview task) to use for data analysis and sampling design. To aid the end user in applying this strategy, a user-friendly application for guiding the end user through the contents of the strategy and the initial characterization process is also developed
Data assimilation of post-irradiation examination data for fission yields from GEF
No full textNuclear data, especially fission yields, create uncertainties in the predicted concentrations of fission products in spent fuel which can exceed engineering target accuracies. Herein, we present a new framework that extends data assimilation methods to burnup simulations by using post-irradiation examination experiments. The adjusted fission yields lowered the bias and reduced the uncertainty of the simulations. Our approach adjusts the model parameters of the code GEF. We compare the BFMC and MOCABA approaches to data assimilation, focusing especially on the effects of the non-normality of GEF’s fission yields. In the application that we present, the best data assimilation framework decreased the average bias of the simulations from 26% to 14%. The average relative standard deviation decreased from 21% to 14%. The GEF fission yields after data assimilation agreed better with those in JEFF3.3. For Pu-239 thermal fission, the average relative difference from JEFF3.3 was 16% before data assimilation and after it was 12%. For the standard deviations of the fission yields, GEF’s were 100% larger than JEFF3.3’s before data assimilation and after were only 4% larger. The inconsistency of the integral data had an important effect on MOCABA, as shown with the Marginal Likelihood Optimization method. When the method was not applied, MOCABA’s adjusted fission yields worsened the bias of the simulations by 30%. BFMC showed that it inherently accounted for this inconsistency. Applying Marginal Likelihood Optimization with BFMC gave a 2% lower bias compared to not applying it, but the results were more poorly converged
An improved method to evaluate the “Joint Oxyde-Gaine” formation in (U,Pu)O
No full textIn this work, two different thermodynamic softwares, ANG
Probabilistic safety assessment for internal and external events/European projects H2020-NARSIS and FP7-ASAMPSA_E
No full textThe 7th EU Framework programme project Advanced Safety Assessment Methodologies: “Extended PSA” (ASAMPSA_E, 2013–2016) was aimed at promoting good practices to extend the scope of existing Probabilistic Safety Assessments (PSAs) and the application of such “extended PSA” in decision-making in the European context. This project led to a collection of guidance reports that describe existing practices and identify their limits. Moreover, it allowed identifying some idea for further research in the framework of collaborative activities. The H2020 project “New Approach to Reactor Safety ImprovementS” (NARSIS, 2017–2021) aims at proposing some improvements to be integrated in existing PSA procedures for NPPs, considering single, cascade and combined external natural hazards (earthquakes, flooding, extreme weather, tsunamis). The project will lead to the release of various tools together with recommendations and guidelines for use in nuclear safety assessment, including a Bayesian-based multi-risk framework able to account for causes and consequences of technical, social/organizational and human aspects and a supporting Severe Accident Management decision-making tool for demonstration purposes, as well
Uncertainty propagation for the design study of the PETALE experimental programme in the CROCUS reactor
No full textThe PETALE experimental programme in the CROCUS reactor intends to provide integral measurements to constrain stainless steel nuclear data. This article presents the tools and the methodology developed to design and optimize the experiments, and its operating principle. Two acceleration techniques have been implemented in the Serpent2 code to perform a Total Monte Carlo uncertainty propagation using variance reduction and correlated sampling technique. Their application to the estimation of the expected reaction rates in dosimeters is also discussed, together with the estimation of the impact of the nuisance parameters of aluminium used in the experiment structures
Partitioning and transmutation strategy R&D for nuclear spent fuel: the SACSESS and GENIORS projects
No full textProcesses such as PUREX allow the recovery and reuse of the uranium and the plutonium of GEN II/GEN III reactors and are being adapted for the recycling of the uranium and the plutonium of GEN IV MOX fuels. However, it does not fix the sensitive issue of the long-term management of the high active nuclear waste (HAW). Indeed, only the recovery and the transmutation of the minor actinides can reduce this burden down to a few hundreds of years. In this context, and in the continuity of the FP7 EURATOM SACSESS project, GENIORS focuses on the reprocessing of MOX fuel containing minor actinides, taking into account safety issues under normal and mal-operation. By implementing a three-step approach (reinforcement of the scientific knowledge => process development and testing => system studies, safety and integration), GENIORS will provide more science-based strategies for nuclear fuel management in the EU