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The Research Center Řež (CVŘ) is developing a methodology to simulate nuclear power plants under accidental conditions. A particular effort is focused in the severe accident phenomenology where hydrogen deflagration carries a critical issue for the containment integrity, such as Fukushima Daiichi accident. For this purpose, the Thermal-hydraulics, hydrogen, aerosol and iodine (THAI) experimental campaigns are chosen due to the several tests involved in different conditions. THAI containment test facility is used to open questions concerning the behavior of hydrogen, iodine, and aerosols in the containment of water-cooled reactors during severe accidents.The Fukushima Daiichi Accident demonstrates that the hydrogen deflagration could lead to a significant containment damage. For this reason, a particular attention is given to the hydrogen deflagration scenario. All simulations are prepared and modeled in melcor 2.1. The results obtained showed a strong influence related to some factors such as: the nodalization pattern, control volumes (CVs) number, flow paths (FPs) number, and time-step. In order to assess the THAI model with the THAI final reports, a sensitivity analysis focused on those parameters was performed, which developed four different models with different nodalization.

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.

Supercritical water (SCW) has advantages like high thermal efficiency and can operate at high temperature and pressure. At the same time, however, these properties bring up related issues, such as material compatibility and corrosion resistance. In an effort to fully investigate the operating conditions, and solutions to these issues, test facilities are being built by many research organizations. One such organization, the Research Center Řež (CVR) located in the Czech Republic, has developed an experimental supercritical water loop (SCWL). The purpose of this loop is to provide experimental data from material testing in various conditions, including operating under the neutron field. This will be achieved by inserting a test channel into the existing experimental reactor light water reactor 15 (LVR-15), which will require a license from the state nuclear regulator (State Office for Nuclear Safety (SUJB)). Part of the licensing documentation is the safety analysis, which combines results from developed models using the thermohydraulic code ATHLET 3.1 A patch 1, as well as the experimental out of pile data. Among the postulated scenarios, an abnormal sequence (labeled A2—Loss of power in the loop) was analyzed in order to provide a preliminary benchmark. This scenario is similar to the postulated in-pile A2 and it was used for the benchmark activity. The aim of this paper is to present this activity including the adopted assumptions in the model. In particular, the paper presents, how these assumptions influenced the results indicating the discrepancies obtained in the first part of the transient.

After the Fukushima DAIICHI accident, new safety requirements were imposed in order to reduce risk of severe accident. One of the principles that have been adopted is the introduction of emergency action levels resulting from the expected consequences. They cover a wide range of component and system malfunctions resulting in emergency, incident, and/or accident conditions. To evaluate those emergency action levels, thermal hydraulic (TH) analyses simulating these malfunction/incident/accident conditions are required. This paper describes the simulation of a real operational incident scenario using a standard thermal hydraulic model of the power plant in the TRACE code that was originally intended for simulation of design basis accidents such as large break coolant accident or loss of flow accident. Special attention was given to the methodology, addressing a long duration of an incident with corrective actions of the operators, and to computational issues leading to model modifications caused by a long duration of the incident along with the necessary conservatisms in the estimated results of the simulated radioactivity release.

In this paper is described the concept of a 20 MW microreactor cooled by molten salts with the registered trademark of Energy Well. Energy Well reactor is designed to supply either electricity or heat, in remote areas or into populated areas. The description includes context, main design features, safety systems, thermohydraulic analysis in RELAPS 3D, logistic, and transport of the reactor.

Research reactors, together with the computational codes, represent a valuable tool combination to investigate over phenomena which can take place in nuclear power plants. In this framework, the Research Centre Řež (CVŘ), which owns two research reactors, Lehkovodni Reaktor 0 (LR-0) and Lehkovodni Reaktor 15 (LVR-15), is moving towards the creation of a code portfolio to be used to relicense them in front of the State Office of Nuclear Safety (SUJB). LVR-15 is a light water tank type research reactor, adopting a sandwich tube type fuel (IRT4M). The first step of this research activity is the preparation of a working model of LVR-15 using SCALE and PARCS codes: SCALE is used to estimate the Cross Sections (XS) based on the real geometry of the reactor and PARCS is used to perform calculation of transients. The results obtained show that the model is adequately accurate for steady state, as soon as the calculation reaches the criticality. The power calculated during the transient calculations, instead, stabilize after about 1000 s and become comparable, within a 2% bias, with the experimental data, extrapolated from the reactor operation. In conclusion, this demonstrates that PARCS code is capable of modelling our research reactor fuel assemblies geometry in a satisfactory way, although further developments would improve the accuracy of the adopted methodology.

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.

This paper focuses on the TRACE code assessment for helium-cooled systems thermal-hydraulic analysis. In the frame of the GoFastR (gas cooled fast reactor) European Collaborative Project, ENEA has offered some selected experimental data for the organization of a benchmark exercise aimed at the validation of the system and CFD codes for the gas reactor transient analyses. One of the Research Center Řež teams participated in it with a CFD code application. Now, the experimental data are used in order to assess the TRACE code for the ongoing high-temperature helium loop (HTHL-2) licensing process. The results of the TRACE calculations agreed very well with the experimental measurements (often within the experimental uncertainties) data provided by the He-FUS3 facility, indicating that the code, despite developed for water coolant applications, if an appropriately tuned input is adopted, it can also be suitable for reasonably accurate gas technology thermo-hydraulic simulations.

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.

This paper shows some results obtained through simulating a VVER-1000 mock-up installed within the LR-0 research reactor in the Research Centre Řež (Czech Republic): the simulations have been performed by using the Serpent 2 and MCNP6 Monte Carlo codes. The mock-up is analysed in 6 different critical configurations, obtained with variation of coolant level and boron concentration. Some important nuclear parameters are evaluated in different positions of the mock-up, in order to know how the level of coolant influences the flux, the fission density, the reaction rate and the neutron spectrum changes, although the reactor remains with k eff around 1.0 to less than 300 pcm of reactivity.

The Fukushima Daiichi NPP (Nuclear Power Plant) accident pointed out the hydrogen explosion issue as one of the main problems that can affect the NPP containment integrity. During a severe accident scenario, the hydrogen combustion can occur and lead to containment integrity failure, since it generates local and global pressure and heat spikes. Such topic was analysed in several research programs addressed all around the world. An important series of test campaigns was done in OECD/NEA WGAMA (Organisation for Economic Co-operation and Development/Nuclear Energy Agency Working Group on Analysis and Management of Accidents) program called THAI (Thermal-hydraulics, Hydrogen, Aerosol and Iodine). The THAI goal is to simulate several phenomena related on hydrogen and Fission Product behaviour in the containment to obtain data relevant for the code benchmarking and validation. Therefore, theoretical analyses are needed, in order to obtain a reliable prediction of the accidental scenario. The facility allows to investigate safety relevant effects under thermal-hydraulics conditions of severe accidents. The experiments performed cover from hydrogen deflagration to iodine and aerosol behaviour under different thermal-hydraulics conditions. Three representative experiments were chosen from the THAI campaign to be modeled and simulated using the MELCOR code with versions 2.1 and 2.2 and compare the results with the experimental ones. This work aims to assess the MELCOR code capability pointing out on the limitation in simulating the hydrogen deflagration and underling possible method to reduce their effect on the simulate results. The benchmarks were addressed with old version of MELCOR however the new version presented slightly different results due to the modification in the parametric model and the default sensitivity coefficients.

Centrum Výzkumu Řež has a vast R&D program which was financed through the SUStainable Energy project. Several state-of-the-art facilities were designed in order to cover the demand of the energy community and to face new challenges for such advanced technologies. Two new experimental loops are to be inserted inside of the LVR-15 Research Reactor. Their purpose is to analyze cladding and structural material behaviour under different pressures, temperatures and coolant media conditions in comparison to the Light Water Reactors. The loops in question are the High Temperature Helium Loop and the Super Critical Water Loop. This paper focuses solely on the supercritical water loop. The first task of the Safety Research Section within the CVŘ was to select and to assess a code in order to simulate the media used in the loops. The second task was to use the selected codes to prepare the facilities models in order to simulate the possible consequences of the accident scenarios on the LVR-15 reactor. Based on the criteria set by the authors of this paper, through a comprehensive selecting procedure the ATHLET 3.1A code was chosen for the super critical water simulation. The paper is presenting the methodology followed in assessing the code’s capability for simulation of SCW.

The Czech Republic National Radiation Protection Institute (SURO) provides technical support to the Czech Republic State Office for Nuclear Safety, providing safety analysis and reviewing of the technical documentations for Nuclear Power Plants (NPPs). For this reason, several computational models created in SURO were prepared using different codes as tools to simulate and investigate the design base and beyond design base accidents scenarios. This paper focuses on the creation of SCALE and PARCS neutronic models for a proper analysis of the VVER-440 reactor analysis. In particular, SCALE models of the VVER-440 fuel assemblies have been created in order to produce collapsed and homogenized cross sections necessary for the study with PARCS of the whole VVER-440 reactor core. The sensitivity study of the suitable energy threshold to be adopted for the preparation with SCALE of collapsed two energy-group homogenized cross sections is also discussed. Finally, the results obtained with PARCS core model are compared with those reported in the VVER-440 Final Safety Report.

This paper presents the comparison between two Monte Carlo based burnup codes: SERPENT and MONTEBURNS. Monte Carlo codes are fully and worldwide adopted to perform analyses on nuclear facilities, also in the frame of Generation IV advanced reactors simulations. Thus, faster and most powerful calculation codes are needed with the aim to analyze complex geometries and specific neutronic behaviors. Burnup calculations are an efficient tool to test neutronic Monte Carlo codes: indeed these calculations couple transport and depletion procedures, so that neutronic reactor behavior can be simulated in its totality. Comparisons have been performed on a configuration representing the Allegro MOX 75 MWth reactor proposed by the European GoFastR (Gas-cooled Fast Reactor) Project in the frame of the 7th Euratom Framework Program. Although in burnup and criticality comparisons the codes used in simulations show different calculation times and some differences in amounts and in precision (in term of statistical errors), a reasonably good agreement between them exists.

In the Czech Republic, as a follow-up, a consortium of research organizations and universities has decided to simulate selected stress tests’ scenarios, in station blackout (SBO) and the loss of ultimate heat sink (LoUHS), with the aim to verify the national stress report and to analyze time response of respective source term releases. These activities are carried out in the frame of the project prevention, preparedness, and mitigation of consequences of severe accident (SA) at Czech NPPs in relation to lessons learned from stress tests after Fukushima, financed by the Ministry of Interior. The Research Centre Rez has been working on a methods for estimation of leakages and consequences of releases (MELCOR) model for VVER1000 nuclear power plant (NPP) starting with a plant systems nodalization. The aim was to benchmark the MELCOR model with the validated TRAC/RELAP advanced computational engine (TRACE) model, first comparing the steady state and continuing in a long-term SBO plus another event until the beginning of the SA. The presented work is based on the previous paper from the ICONE 23rd Conference hosted in Japan. It focuses mainly on the comparison of the thermohydraulics of the two models created in MELCOR and TRACE codes as outcome of the “Fukushima project.” After that, preliminary general results of the SA progression showing the hydrogen production and the relocation phenomena will be shortly discussed. This scenario is considered closed after some seconds to the break of the lower head. It is important to note that this paper is a substantial update of a previous one (Mazzini et al., 2015, “Analyses of SBO Sequence of VVER1000 Reactor Using TRACE and MELCOR Codes,” 23rd International Conference on Nuclear Engineering (ICONE23)) and although it contains the same descriptive sections, all the results are new.

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.

This paper reports the results of a comparison among JEFF and ENDF/B data sets when used by SERPENT and MONTEBURNS codes on a gas-cooled fast reactor (GFR)-like configuration. Particularly, it shows a comparison between the two Monte Carlo-based codes, each one adopting three different cross-section data sets, namely, JEFF-3.1, JEFF-3.1.2, and ENDF/B-VII.1. Calculations have been carried out on the Allegro reactor, i.e., an experimental GFR-like facility that could be built in the European Union as a GFR demonstration. Results include nuclear parameters, such as the effective multiplication factor and fluxes, as well as the atomic densities for some important nuclides versus burn-up.

Research Centre Rez solves several safety related projects dealing with safety of Czech NPPs, some of which require fully functioning Three Dimensional (3D) model of the reactor core. While in a number of safety analysis of various accident scenarios it is sufficient to use one point reactor kinetics, there are selected types of accidents in which it is useful to model the space (3D) neutron kinetics, in particular control rod ejections, boron dilution scenarios, including transitions from design basis to beyond design basis accidents. This paper is focused to analyze the present model of the core of VVER1000/V320 reactor. Which is applicable for 3D modeling of neutron kinetics in selected design and beyond design basis accidents. The model is based on a cross-sections library created by SCALE 6.1.2/TRITON simulations. PARCS 3.2 code uses homogenized cross-sections libraries to calculate neutronic and other core parameters of the PWR reactors. Similar model is prepared with MCNP6 for comparison between deterministic (Pn spherical-harmonics method used in PARCS) and the stochastic (Monte Carlo) approach (used in MCNP6). Such comparison will serve as a demonstration of the capability of the PARCS code for VVER1000/V320 analyses.Copyright © 2014 by ASME

Due to the fact that Fukushima accident occurred, the European Commission decided to evaluate all European NPPs. The Research Centre Rez plans to simulate accident scenarios of NPP Temelin and NPP Dukovany in accordance with the stress tests conditions. While in a number of safety analysis of various accident scenarios is sufficient enough to use point kinetics, there are selected types of events in which is useful to model the 3D space behaviour of neutron kinetics. In particular control rod ejections, boron dilution scenarios, including transition between design basis and beyond design basis accidents. The main purpose of the presented work is to create a model of the active core of VVER1000/V320 reactor applicable for 3D modelling of neutron kinetics and its coupling with the thermo-hydraulic system codes, such as RELAP5/TRACE. Cross-section data was prepared with computation code SCALE 6.1.2/TRITON. PARCS 3.2 code uses homogenized cross-sections libraries to calculate core parameters of the PWR reactors.

This paper reports the results of a comparison among JEFF and ENDF/B datasets when used by SERPENT and MONTEBURNS codes on a GFR-like configuration. Particularly, it shows a comparison between the two Monte Carlo based codes, each one adopting three different cross sections dataset, namely JEFF-3.1, JEFF-3.1.2 and ENDF/B-VII.1. Calculations have been carried out on the Allegro reactor, i.e. an experimental GFR-like facility that should be built in EU as GFR demonstrator. Results concern nuclear parameters as effective multiplication factor and fluxes, as well as the atomic densities for some important nuclides versus burnup.Copyright © 2014 by ASME

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).