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Pre-conceptual design studies for a European Demonstration Fusion Power Plant (DEMO) have been in progress since 2014. At this stage, while a range of design options are being considered, assessments of the safety and environmental impact of these options are carried out. This is to ensure that the DEMO plant is optimized for safety performance, and that it will demonstrate the favourable safety and environmental characteristics of fusion energy as part of its mission. To this end, safety studies have been under way since the start of the project, to set clear safety objectives and requirements, to analyse the response of the plant to off-normal events, to assess hazardous inventories, to develop strategies to minimize them, and to identify the main potential contributors to environmental releases and to occupational radiation exposure. Development of computer codes and models for safety analysis is accompanied by selected experimental activities focused on improving their validation, and these models have been used for initial studies of postulated accident scenarios, selected by a formal methodology. Studies of key aspects of waste management are also performed, to minimize the waste burden of DEMO and of fusion power plants that will follow.

This paper aims to compare three Monte Carlo (MC) burnup based codes, i.e. MCNP6, Monteburns and Serpent on a future prototype reactor, named ALLEGRO, based on Gas cooled Fast Reactor (GFR) technology. GFR reactors are one of the proposed Generation-IV fast reactors; ALLEGRO facility is scheduled to be built in Europe as a GFR demonstrator, so its deepened simulation can help in its future development. The present study follows other researches already performed and aims to exhibit the different approaches in burnup calculations applied to a gas cooled fast reactors, i.e. this paper would like to show and to compare some results concerning nuclear parameters as keff and flux spectra, as well as the mass inventories versus burnup for some nuclides evaluated with different Monte Carlo codes. From obtained results, it seems to exist some differences in evaluation of nuclear parameters, mainly in effective multiplication factor and in mass inventories. The remaining differences are mainly related to calculation time: indeed between the fastest, that is SERPENT, and the slowest, that are MCNP6 and MONTEBURNS, the differences are about one order of magnitude. As far as precision is concerned, it was considered the standard only for effective multiplication factor and it seems that all codes are in good agreement.

An important issue for the Fusion Reactors is the hydrogen explosion hazard assessment. For this reason, in the Work Package of Safety Analyses and Environment (WPSAE) of the EUROFusion consortium, a task was established to identify the potential for hydrogen production in DEMO vessel in accident situations and to investigate the possible solutions which can prevent the risk of hydrogen and dust explosion. The hydrogen in the Vacuum Vessel (VV) can lead to combustion progressing in deflagration and detonation. Besides, the tungsten dust could enhance the effects of H2 reaction as demonstrated by experiments, however, in this first set of accidents, only the H2 generated by the chemical reaction is accounted for. For these reasons, the solutions to avoid or limit the H2 production and/or to control safely the risk of H2 accumulation and explosion need to be screened starting from the solutions adopted in the fission technology, and later on, adapted and assessed in the Pressure Suppression System (PSS) foreseen in DEMO. One promising idea seems to install passive catalytic recombiners in the PSS. The aim of this paper is firstly to set a range of operating parameters for the PAR functioning and intervention suitable in DEMO PSS configuration. Secondly to verify if the present criteria and parameters work efficiently in mitigating the H2 accumulation a PSS in accidental conditions. To test the theoretical effectiveness of the PAR intervention, two reference accidents have been analyzed: an in-VV Loss Of Coolant Accident (LOCA) in the frame of the Design Basis Accident (DBA) and a Loss Of Flow Accident (LOFA) without active plasma shutdown in the context of the Beyond Design Basis Accidents (BDBA). The results seem promising if the performances of the PAR will be confirmed in sub atmospheric and saturated conditions.

Energy Well (EW) is a concept of a small modular reactor, designed as a source of energy for places with low infrastructure. Its design combines TRISO fuel particles, graphite moderator and molten salt coolant. One of key features is that the active core is small and therefore easy to transport. Residual heat estimation of EW fuel assembly is conducted using two different computation codes, SCALE and SERPENT. The obtained residual heat power is then used as an input in TRACE model of the EW to obtain the development of residual temperature in the coolant in the core inlet and outlet and also maximum temperature on the outer side of a fuel slab. The resulting temperatures are below the operational temperature.

The safety features of the future nuclear fusion reactors are one of the key issues for their attractiveness if compared with the fission plants. In fusion devices, accidents with high release of radioactive materials have low probabilities because the most part of abnormal transients lead to passive plasma shutdown. It does not mean that radiological source terms such tritium and activated dust are not generated and released, but their inventory does not increase during abnormal events. Therefore, the source term inventory has to be assessed during normal operation and traced when accidents occur. For this reason, a study for qualification and quantification of the tritium and dust source term (DTS) was established with the aim to understand their production, deposition, and penetration in the vacuum vessel (VV) and in the breeding blanket (BB). The main concern is source term release during the main accident scenarios to comply with a future licensing process. In case of abnormal event scenarios, the source term inventory involved in the release changes and requires a different confinement approach and mitigation. For the estimation of the source term in the DEMOnstration Fusion Power Station (DEMO), a methodology was developed. The methodology scales the tritium and DTS inside the VV from the International Thermonuclear Experimental Reactor, the European Power Plant Conceptual Study, and reports the tritium generated inside the breeder blanket from data quantified in other studies for DEMO. In this article, the methodology was updated and tritium and DTS for DEMO 2016 design were estimated. Moreover, the tritium and dust release pathways were highlighted according to different accidental scenarios. These results were obtained for all blanket concepts, which are analyzing in the ongoing DEMO EUROFusion project. The values estimated in this article will be used in the safety analyses to evaluate releases or to quantify the operational limits starting from values postulated in International Thermonuclear Experimental Reactor.

The severe accident integral codes such as Methods for Estimation of Leakages and Consequences of Releases (MELCOR) are complex tools used to simulate and analyse the progression of a severe accident from the onset of the accident up to the release from the containment. For this reason, these tools are developed in order to simulate different phenomena coupling models which can simulate simultaneously the ThermoHydraulic (TH), the physics and the chemistry. In order to evaluate the performance in the prediction of those complicated phenomena, several experimental facilities were built in Europe and all around the world. One of these facilities is the PHEBUS built by Institut de Radioprotection et de Sûrete Nucléaire (IRSN) in Cadarache. The facility reproduces the severe accident phenomena for a pressurized water reactor (PWR) on a volumetric scale of 1:5000. This paper aims to continue the assessment of the MELCOR code from version 2.1 up to version 2.2 underlying the difference in the fission product transport. The assessment of severe accident is an important step to the sustainability of the nuclear energy production in this period where the old nuclear power plants are more than the new reactors. The analyses presented in this paper focuses on models assessment with attention on the influence of B4C oxidation on the release and transport of fission products. Such phenomenon is a concern point in the nuclear industry, as was highlighted during the Fukushima Daiichi accident. Simulation of the source term is a key point to evaluate the severe accident hazard along with other safety aspects.

The topic of Nuclear Safety Culture touches several different aspects with contributions from the main organizations involved in nuclear projects and belonging to vendors, utility and regulators. Two nuclear safety directives issued by the European Commission emphasize the fundamental principle of national responsibility for nuclear safety and are implemented in each member country’s legislation. An example of fission implementation is highlighted, referring to the Czech Republic legislation, an example of application in fusion technology is the implementation of the Nuclear Safety Culture in the ITER project, located in Cadarache, in the south of France. The aim of the paper is to highlight the importance of this field, pointing out the cross reference between fission and fusion technology as applied in two countries, with concrete experiences and future prospects for nuclear technologies.

Nowadays nuclear power is the only greenhouse-free source that can contribute appreciably to the solution of the increasing world-wide energy demand problem. The use of Thorium in the nuclear energy production may offer some unique advantages to accomplish this task. Extensive research and development on the thorium fuel cycle has been conducted in many countries around the world. Starting from the current nuclear waste policy as well as from the (V)HTR good capabilities of Pu burning, the EU PUMA project focus on core physics investigations, in order to further demonstrate the potential benefits of using the HTR core as a Pu/MA transmuter. In this paper the following aspects have been analyzed: – The state of art about the studies focusing on the use of Th in different core concepts – The use of Th in HTRs, with a particular emphasis on Th-Pu fuel cycles – An original assessment of Th-Pu fuel cycles in HTR, on the basis, the energy produced totally, per initial Pu and the ratio between discharged and loaded Pu mass Some aspects related to Thorium exploitation were outlined, with a particular emphasis on its suitability for working in pebble-bed HTR as a fertile element in a Th-Pu fuel cycle, that is a promising kind of fuel considered in the framework of the EU PUMA project. Particularly, the influence of the Th/Pu weight fraction at BOC in a typical HTR pebble was analyzed as far as the reactivity trend vs. burn-up, the energy produced per Pu unit mass, and the Pu isotopic composition at EOC are concerned. The obtained results suggest us that there is not a best composition for the PBMR Pu/Th fuel. However deeper investigations need to be performed in order to draw final conclusions. Thus, at the moment it is possible to state that some optimized Th percentage in the initial Pu/Th fuel could be suggested on the basis of the aim we are trying to reach.

The rod ejection (RE) is a design basis accident in accordance with NUREG-0800 and usually studied using point kinetics. In this paper a methodology and a 3D kinetic model is prepared (PARCS), coupled with a thermal hydraulic system code (TRACE) for simulating this accident scenario for general VVER 1000 technology.

The conceptual design studies for a European DEMO, include safety studies and analyses, to help guide the design process and start preparation for eventual licensing. A safety approach has been adopted that is expected to satisfy a nuclear regulator, but at this early stage it is difficult to anticipate the safety concerns that may be raised. Potential safety issues for DEMO that have been listed by a regulatory technical body include decay heat removal following an accidental loss of cooling or during blanket module replacement, the minimization of occupational radiation exposure, the comprehensive identification of postulated accident scenarios, the environmental release of gaseous tritium during normal operation, and the management of radioactive waste, particularly where contaminated with tritium. Here we explain how these issues are being addressed in the safety programme for a EU DEMO, and are taken into account in the design from the beginning of its conception.

The problem of source term qualification is one of the most important topics in order to predict possible releases of the Activation Products (APs) and Tritium from the DEMO Fusion reactor. The prevention of any possible consequence, which can affect the environment and the population, is the mission of Fusion technology. In the frame of the EUROfusion Work Package of Safety and Environment (WPSAE) a methodology to scale and evaluate the source terms assessed for ITER and the Fusion Power Plant Conceptual Study (PPCS) has been studied. This paper refers to the activity currently conducted for the DEMO source terms assessment and the preliminary results obtained. During activities in the task, the methodology was developed for the evaluation of Tritium and APs concentration inventory. The methodology is explained in detail for the prediction of the Tritium and APs concentration in Vacuum Vessel (VV) and in the Breeding Blanket (BB) starting from the DEMO current design data and the inventories assumed in ITER, PPCS and SEAFP programs. These results refer to the Helium Cooled Lead Liquid (HCLL) and the Helium Cooled Pebble Bed (HCPB) concepts. The approach is based on the foundations, set in the fission technology safety analysis of the Design Basis Accidents (DBA), Design Extension Conditions (DEC) and Beyond Design Basis Accident (BDBA).