Propagation of nuclear data uncertainties in fuel cycle calculations using MONTE-CARLO Technique

The uncertainty propagation in fuel cycle calculations due to Nuclear Data (ND) is a important important issue for : issue for : • Present fuel cycles (e.g. high burnup fuel programme) • New fuel cycles designs (e.g. fast breeder reactors and ADS) Different error propagation techniques can be used: • Sensitivity analysis • Response Response Surface Method Surface Method • Monte Carlo technique Then, p p , , in this paper, it is assessed the imp y pact of ND uncertainties on the decay heat and radiotoxicity in two applications: • Fission Pulse Decay ( y Heat calculation (FPDH) • Conceptual design of European Facility for Industrial Transmutation (EFIT)

Thermal diffusion profile by NRA of Deuterium implanted in solid breeder blanket materials

This paper studies the relationship between aging, physical changes and the results of non-destructive testing of plywood. 176 pieces of plywood were tested to analyze their actual and estimated density using non-destructive methods (screw withdrawal force and ultrasound wave velocity) during a laboratory aging test. From the results of statistical analysis it can be concluded that there is a strong relationship between the non-destructive measurements carried out, and the decline in the physical properties of the panels due to aging. The authors propose several models to estimate board density. The best results are obtained with ultrasound. A reliable prediction of the degree of deterioration (aging) of board is presented. Breeder blanket materials have to produce tritium from lithium while fulfilling several strict conditions. In particular, when dealing with materials to be applied in fusion reactors, one of the key questions is the study of light ions retention, which can be produced by transmutation reactions and/or introduced by interaction with the plasma. In ceramic breeders the understanding of the hydrogen isotopes behaviour and specially the diffusion of tritium to the surface is crucial. Moreover the evolution of the microstructure during irradiation with energetic ions, neutrons and electrons is complex because of the interaction of a high number of processes.

Present status of the Liquid Breeder Validation

One of the objectives of IFMIF (International Fusion Materials Irradiation Facility), as stated in its specifications, is the validation of breeder blanket concepts for DEMO design. The so-called Liquid Breeder Validation Module (LBVM) will be used in IFMIF to perform experiments under irradiation on functional materials related to liquid breeder concepts for future fusion reactors. This module, not considered in previous IFMIF design phases, is currently under design by CIEMAT in the framework of the IFMIF/EVEDA project. In this paper, the present status of the design of the LBVM is presented.

New simulations to qualify eutectic lithium-lead as breeder material

Pb17Li is today a reference breeder material in diverse fusion R&D programs worldwide. One of the main issues is the problem of liquid metals breeder blanket behavior. The knowledge of eutectic properties like optimal composition, physical and thermodynamic behavior or diffusion coefficients of Tritium are extremely necessary for current designs. In particular, the knowledge of the function linking the tritium concentration dissolved in liquid materials with the tritium partial pressure at a liquid/gas interface in equilibrium, CT =f(PT ), is of basic importance because it directly impacts all functional properties of a blanket determining: tritium inventory, tritium permeation rate and tritium extraction efficiency. Nowadays, understanding the structure and behavior of this compound is a real goal in fusion engineering and materials science. Atomistic simulations of liquids can provide much information; not only supplementing experimental data, but providing new tests of theories and ideas, making specific predictions that require experimental tests, and ultimately helping to a deeper understanding

On the Extension of the Analytic Nodal Diffusion Solver ANDES to Sodium Fast Reactors

Within the framework of the Collaborative Project for a European Sodium Fast Reactor, the reactor physics group at UPM is working on the extension of its in-house multi-scale advanced deterministic code COBAYA3 to Sodium Fast Reactors (SFR). COBAYA3 is a 3D multigroup neutron kinetics diffusion code that can be used either as a pin-by-pin code or as a stand-alone nodal code by using the analytic nodal diffusion solver ANDES. It is coupled with thermalhydraulics codes such as COBRA-TF and FLICA, allowing transient analysis of LWR at both fine-mesh and coarse-mesh scales. In order to enable also 3D pin-by-pin and nodal coupled NK-TH simulations of SFR, different developments are in progress. This paper presents the first steps towards the application of COBAYA3 to this type of reactors. ANDES solver, already extended to triangular-Z geometry, has been applied to fast reactor steady-state calculations. The required cross section libraries were generated with ERANOS code for several configurations. The limitations encountered in the application of the Analytic Coarse Mesh Finite Difference (ACMFD) method –implemented inside ANDES– to fast reactors are presented and the sensitivity of the method when using a high number of energy groups is studied. ANDES performance is assessed by comparison with the results provided by ERANOS, using a mini-core model in 33 energy groups. Furthermore, a benchmark from the NEA for a small 3D FBR in hexagonal-Z geometry and 4 energy groups is also employed to verify the behavior of the code with few energy groups.

An educational initiative between the Universidad Politécnica de Madrid and Spanish Young Generation in Nuclear (Jóvenes Nucleares): The Seminar of Nuclear Safety in Advanced Reactors

Jóvenes Nucleares (Spanish Young Generation in Nuclear, JJNN) is a non-profit organization and a commission of the Spanish Nuclear Society (SNE). The Universidad Politécnica de Madrid (Technical University of Madrid, UPM) is one of the most prestigious technical universities of Spain, and has a very strong curriculum in nuclear engineering training and research. Finishing 2009, JJNN and the UPM started to plan a new and first-of-a-kind Seminar in Nuclear Safety focused on the Advanced Reactors (Generation III, III+ and IV). The scope was to make a general description of the safety in the new reactors, comparing them with the built Generation II reactors from a technical point of view but simple and without the need of strong background in nuclear engineering to try to be interesting for the most number of people possible.

Effects of feeding strategy, fiber source of the diet, and crude protein content on productive performance of broiler breeder hens

A 12-wk experiment was conducted to investigate the effect of feeding program, dietary fiber, and CP content of the diet on productive performance of Ross broiler breeder hens (41 wk of age). There were 12 treatments arranged factorially with 2 levels of CP (14.5 vs. 17.4%), 3 fiber sources (0 vs. 3% inulin vs. 3% cellulose), and 2 levels of feed intake (160 vs. 208 g/d) that corresponded to restricted (R) or ad libitum (AL) feeding systems. The experimental diets contained 2,800 kcal ME with either 0.65 (14.5% CP) or 0.78% Lys (17.4% CP).

Simulación atomística de metales líquidos y aleaciones para reactores de fusión nuclear = Atomistic simulation of liquid metals and alloys for nuclear fusion reactors

La presente tesis comprende un estudio de metales líquidos, Li, Pb y eutéctico Li17Pb en el ámbito de la tecnología de fusión nuclear. Uno de los problemas fundamentales en futuros reactores de fusión es la producción y extracción de tritio en la denominada envoltura regeneradora (blanket en inglés). Dicho blanket tendrá dos propósitos, la extracción del calor generado por las reacciones de fusión para su posterior conversión en energía eléctrica así como la producción de tritio para realimentar el proceso. Dicha producción se realizará mediante el “splitting” del Li con los neutrones provenientes de la fusión. Esta reacción produce tritio y helio por lo que la interacción del T y el He con el metal líquido, con los materiales estructurales así como con el He es un problema fundamental aun no bien entendido y de gran importancia para futuros diseños. Los capítulos 1 2 y 3 presentan una introducción a dichos problemas. El capítulo 1 introduce al lector en la tecnología de fusión nuclear. El segundo capítulo explica en mayor detalle el uso de metales líquidos en reactores de fusión, no solo en blankets sino también como primera pared, divertor etc, lo que se denomina en general “plasma facing materials”. Por último se ofrece una breve introducción a las técnicas de dinámica molecular clásica (CMD) y un breve resumen de los potenciales más usados. El estudio se ha llevado a cabo utilizando simulación atomística mediante potenciales semi-empíricos del tipo átomo embebido (EAM). La Tesis consta de 4 partes bien definidas. En primer lugar se verificó la idoneidad de los potenciales tipo EAM para simular las propiedades de los metales Li y Pb en fase líquida. Dicho estudio se detalla en el Capítulo 4 y en su extensión, el Apéndice 1, en el que se estudia los límites de validez de esta aproximación. Los resultados de dicho estudio han sido publicados y presentados en diversos congresos internacionales. Un resumen de la metodología seguida fue publicado como capítulo de libro en Technofusión 2011. Los resultados se presentaron en diversos congresos internacionales, entre ellos ICENES 2011, (Artículo en ICENES Proceedings) ICOPS-SOFE 2011, en una presentación oral etc. El trabajo ha sido aceptado recientemente en Journal of Nuclear Materiales (Fraile et al 2012). La segunda parte y más importante comprende el desarrollo de un potencial para el estudio de la mezcla de ambos metales. Éste es el trabajo más novedoso e importante dado que no existía en la literatura un potencial semejante. Se estudiaron dos aproximaciones distintas al problema, un potencial tipo EAM/cd y un potencial EAM/alloy. Ambos potenciales dan resultados satisfactorios para la simulación del eutéctico (y concentraciones de Li menores que el 17%). Sin embargo el sistema LiPb en todas las concentraciones es un sistema que se aparta enormemente de una solución ideal y dicho potencial no daba buenos resultados para mezclas PbLi con concentraciones de Li grandes. Este problema fue solventado mediante el desarrollo de un segundo potencial, esta vez tipo EAM/alloy (segunda parte del Capítulo 5). Dicho trabajo será enviado a Physical Review Letters o a Phys. Rev. B, y una extensión junto con un estudio detallado de las propiedades del eutéctico de acuerdo con nuestras simulaciones se enviará a continuación a Phys. Rev. B. En tercer lugar se estudió el problema de la difusividad del H en metales líquidos aprovechando distintos potenciales existentes en la literatura. El problema del H en metales líquidos es importante en metalurgia. En dicho capítulo se estudió la difusividad del H en Pd, Ni y Al con potenciales tipo EAM, y también con un potencial más sofisticado que tiene en cuenta la dependencia angular de las interacciones (ADP por sus siglas en inglés). De este modo disponemos de un estudio detallado del problema con diferentes modelos y diferentes metales. La conclusión apunta a que si se compara con los resultados experimentales (muy escasos) los resultados obtenidos mediante CMD dan valores bajos de la difusividad del H. Las razones de dicho desacuerdo entre simulación y experimentos se detallan en el Capítulo 6. Este trabajo ha sido presentado en una presentación oral en el reciente congreso internacional “Trends on Nanothecnology” TNT 2012 celebrado en Madrid. El trabajo será publicado en un futuro próximo. Por último, como se dijo anteriormente, el estudio del He, la formación de burbujas en metales líquidos, su difusión nucleación y cavitación es otro problema deseable de ser estudiado mediante técnicas atomísticas. Para ello es necesario el desarrollo de diversos potenciales, He-Li, He-Pb y un potencial ternario Pb-Li-He. Para ello se han realizado simulación ab initio de los sistemas Pb+He y Li+He. Dicho estudio pretende calcular las fuerzas entre los átomos del metal (Pb o Li) con intersticiales de He. De este modo aplicaremos el “force matching method” (FMM) para el desarrollo de dichos potenciales. En el Capítulo 7 se detallan los resultados obtenidos referidos a las posiciones más favorables de las impurezas de He dentro de redes cristalinas de Pb y Li así como el efecto de tener en cuenta el acoplo spin-orbita (SOC en inglés). El análisis de los resultados en términos de transferencia de carga y análisis de las densidades electrónicas, así como la creación de los potenciales mencionados está en progreso. En conjunto la tesis presenta un estudio de los diversos problemas relacionados con el uso de metales líquidos en reactores de fusión y representa un primer paso en la determinación de parámetros de gran importancia para el diseño de blankets y sistemas de primera pared. Con la simulación MD de dichos problemas mediante, importante, potenciales realistas, valores de difusión, solubilidad etc de especies ligeras, H (o sus isotopos) y He en metales líquidos podrá ser calculada complementando así la base de datos que presenta enormes incertidumbres.

Plasma–wall interaction in laser inertial fusion reactors: novel proposals for radiation tests of first wall materials

Dry-wall laser inertial fusion (LIF) chambers will have to withstand strong bursts of fast charged particles which will deposit tens of kJ m−2 and implant more than 1018 particles m−2 in a few microseconds at a repetition rate of some Hz. Large chamber dimensions and resistant plasma-facing materials must be combined to guarantee the chamber performance as long as possible under the expected threats: heating, fatigue, cracking, formation of defects, retention of light species, swelling and erosion. Current and novel radiation resistant materials for the first wall need to be validated under realistic conditions. However, at present there is a lack of facilities which can reproduce such ion environments. This contribution proposes the use of ultra-intense lasers and high-intense pulsed ion beams (HIPIB) to recreate the plasma conditions in LIF reactors. By target normal sheath acceleration, ultra-intense lasers can generate very short and energetic ion pulses with a spectral distribution similar to that of the inertial fusion ion bursts, suitable to validate fusion materials and to investigate the barely known propagation of those bursts through background plasmas/gases present in the reactor chamber. HIPIB technologies, initially developed for inertial fusion driver systems, provide huge intensity pulses which meet the irradiation conditions expected in the first wall of LIF chambers and thus can be used for the validation of materials too.

Plasma–wall interaction in laser inertial fusion reactors: novel proposals for radiation tests of first wall materials

Dry-wall laser inertial fusion (LIF) chambers will have to withstand strong bursts of fast charged particles which will deposit tens of kJ m−2 and implant more than 1018 particles m−2 in a few microseconds at a repetition rate of some Hz. Large chamber dimensions and resistant plasma-facing materials must be combined to guarantee the chamber performance as long as possible under the expected threats: heating, fatigue, cracking, formation of defects, retention of light species, swelling and erosion. Current and novel radiation resistant materials for the first wall need to be validated under realistic conditions. However, at present there is a lack of facilities which can reproduce such ion environments. This contribution proposes the use of ultra-intense lasers and high-intense pulsed ion beams (HIPIB) to recreate the plasma conditions in LIF reactors. By target normal sheath acceleration, ultra-intense lasers can generate very short and energetic ion pulses with a spectral distribution similar to that of the inertial fusion ion bursts, suitable to validate fusion materials and to investigate the barely known propagation of those bursts through background plasmas/gases present in the reactor chamber. HIPIB technologies, initially developed for inertial fusion driver systems, provide huge intensity pulses which meet the irradiation conditions expected in the first wall of LIF chambers and thus can be used for the validation of materials too.

The seminar of nuclear safety in advanced reactors and the seminar of nuclear fusion: two formative initiatives from Universidad Politécnica de Madrid and spanish young Generation in Nuclear (Jóvenes Nucleares)

panish Young Generation in Nuclear (Jóvenes Nucleares) is a commission of the Spanish Nuclear Society (SNE), whose main goals are to spread knowledge about nuclear energy among the society. Following this motivation, two Seminars have been carried out with the collaboration of the Technical University of Madrid: The Seminar of Nuclear Safety in Advanced Reactors (SRA) and the Seminar of Nuclear Fusion (SFN). The first one, which has been celebrated every year since 2010, aims to show clearly the advances that have been obtained in the section of safety with the new reactors, from a technical but simple point of view and without needing great previous nuclear engineering knowledge. The second one, which first edition was held in 2011, aims to give a general overview of the past, present and future situation of nuclear fusion technology, and was born as a result of the increasing interest of our Spanish Young Generation members in this technology.

Nanostructured tungsten as a first wall material for the future nuclear fusion reactors

The lack of materials able to withstand the severe radiation conditions (high thermal loads and atomistic damage) expected in fusion reactors is the actual bottle neck for fusion to become a reality. The main requisite for plasma facing materials (PFM) is to have excellent structural stability since severe cracking or mass loss would hamper their protection role which turns out to be unacceptable. Additional practical requirements for plasma facing materials are among others: (i) high thermal shock resistance, (ii) high thermal conductivity (iii) high melting point (iv) low physical and chemical sputtering, and (v) low tritium retention.

Analysis of a Spanish energy scenario with Generation IV nuclear reactors

The advantages of fast-spectrum reactors consist not only of an efficient use of fuel through the breeding of fissile material and the use of natural or depleted uranium, but also of the potential reduction of the amount of actinides such as americium and neptunium contained in the irradiated fuel. The first aspect means a guaranteed future nuclear fuel supply. The second fact is key for high-level radioactive waste management, because these elements are the main responsible for the radioactivity of the irradiated fuel in the long term. The present study aims to analyze the hypothetical deployment of a Gen-IV Sodium Fast Reactor (SFR) fleet in Spain. A nuclear fleet of fast reactors would enable a fuel cycle strategy different than the open cycle, currently adopted by most of the countries with nuclear power. A transition from the current Gen-II to Gen-IV fleet is envisaged through an intermediate deployment of Gen-III reactors. Fuel reprocessing from the Gen-II and Gen-III Light Water Reactors (LWR) has been considered. In the so-called advanced fuel cycle, the reprocessed fuel used to produce energy will breed new fissile fuel and transmute minor actinides at the same time. A reference case scenario has been postulated and further sensitivity studies have been performed to analyze the impact of the different parameters on the required reactor fleet. The potential capability of Spain to supply the required fleet for the reference scenario using national resources has been verified. Finally, some consequences on irradiated final fuel inventory are assessed. Calculations are performed with the Monte Carlo transport-coupled depletion code SERPENT together with post-processing tools.

Core physics and safety analysis of Generation-IV Sodium Fast Reactors using existing and newly developed computational tools

El futuro de la energía nuclear de fisión dependerá, entre otros factores, de la capacidad que las nuevas tecnologías demuestren para solventar los principales retos a largo plazo que se plantean. Los principales retos se pueden resumir en los siguientes aspectos: la capacidad de proporcionar una solución final, segura y fiable a los residuos radiactivos; así como dar solución a la limitación de recursos naturales necesarios para alimentar los reactores nucleares; y por último, una mejora robusta en la seguridad de las centrales que en definitiva evite cualquier daño potencial tanto en la población como en el medio ambiente como consecuencia de cualquier escenario imaginable o más allá de lo imaginable. Siguiendo estas motivaciones, la Generación IV de reactores nucleares surge con el compromiso de proporcionar electricidad de forma sostenible, segura, económica y evitando la proliferación de material fisible. Entre los sistemas conceptuales que se consideran para la Gen IV, los reactores rápidos destacan por su capacidad potencial de transmutar actínidos a la vez que permiten una utilización óptima de los recursos naturales. Entre los refrigerantes que se plantean, el sodio parece una de las soluciones más prometedoras. Como consecuencia, esta tesis surgió dentro del marco del proyecto europeo CP-ESFR con el principal objetivo de evaluar la física de núcleo y seguridad de los reactores rápidos refrigerados por sodio, al tiempo que se desarrollaron herramientas apropiadas para dichos análisis. Efectivamente, en una primera parte de la tesis, se abarca el estudio de la física del núcleo de un reactor rápido representativo, incluyendo el análisis detallado de la capacidad de transmutar actínidos minoritarios. Como resultado de dichos análisis, se publicó un artículo en la revista Annals of Nuclear Energy [96]. Por otra parte, a través de un análisis de un hipotético escenario nuclear español, se evalúo la disponibilidad de recursos naturales necesarios en el caso particular de España para alimentar una flota específica de reactores rápidos, siguiendo varios escenarios de demanda, y teniendo en cuenta la capacidad de reproducción de plutonio que tienen estos sistemas. Como resultado de este trabajo también surgió una publicación en otra revista científica de prestigio internacional como es Energy Conversion and Management [97]. Con objeto de realizar esos y otros análisis, se desarrollaron diversos modelos del núcleo del ESFR siguiendo varias configuraciones, y para diferentes códigos. Por otro lado, con objeto de poder realizar análisis de seguridad de reactores rápidos, son necesarias herramientas multidimensionales de alta fidelidad específicas para reactores rápidos. Dichas herramientas deben integrar fenómenos relacionados con la neutrónica y con la termo-hidráulica, entre otros, mediante una aproximación multi-física. Siguiendo este objetivo, se evalúo el código de difusión neutrónica ANDES para su aplicación a reactores rápidos. ANDES es un código de resolución nodal que se encuentra implementado dentro del sistema COBAYA3 y está basado en el método ACMFD. Por lo tanto, el método ACMFD fue sometido a una revisión en profundidad para evaluar su aptitud para la aplicación a reactores rápidos. Durante ese proceso, se identificaron determinadas limitaciones que se discutirán a lo largo de este trabajo, junto con los desarrollos que se han elaborado e implementado para la resolución de dichas dificultades. Por otra parte, se desarrolló satisfactoriamente el acomplamiento del código neutrónico ANDES con un código termo-hidráulico de subcanales llamado SUBCHANFLOW, desarrollado recientemente en el KIT. Como conclusión de esta parte, todos los desarrollos implementados son evaluados y verificados. En paralelo con esos desarrollos, se calcularon para el núcleo del ESFR las secciones eficaces en multigrupos homogeneizadas a nivel nodal, así como otros parámetros neutrónicos, mediante los códigos ERANOS, primero, y SERPENT, después. Dichos parámetros se utilizaron más adelante para realizar cálculos estacionarios con ANDES. Además, como consecuencia de la contribución de la UPM al paquete de seguridad del proyecto CP-ESFR, se calcularon mediante el código SERPENT los parámetros de cinética puntual que se necesitan introducir en los típicos códigos termo-hidráulicos de planta, para estudios de seguridad. En concreto, dichos parámetros sirvieron para el análisis del impacto que tienen los actínidos minoritarios en el comportamiento de transitorios. Concluyendo, la tesis presenta una aproximación sistemática y multidisciplinar aplicada al análisis de seguridad y comportamiento neutrónico de los reactores rápidos de sodio de la Gen-IV, usando herramientas de cálculo existentes y recién desarrolladas ad' hoc para tal aplicación. Se ha empleado una cantidad importante de tiempo en identificar limitaciones de los métodos nodales analíticos en su aplicación en multigrupos a reactores rápidos, y se proponen interesantes soluciones para abordarlas. ABSTRACT The future of nuclear reactors will depend, among other aspects, on the capability to solve the long-term challenges linked to this technology. These are the capability to provide a definite, safe and reliable solution to the nuclear wastes; the limitation of natural resources, needed to fuel the reactors; and last but not least, the improved safety, which would avoid any potential damage on the public and or environment as a consequence of any imaginable and beyond imaginable circumstance. Following these motivations, the IV Generation of nuclear reactors arises, with the aim to provide sustainable, safe, economic and proliferationresistant electricity. Among the systems considered for the Gen IV, fast reactors have a representative role thanks to their potential capacity to transmute actinides together with the optimal usage of natural resources, being the sodium fast reactors the most promising concept. As a consequence, this thesis was born in the framework of the CP-ESFR project with the generic aim of evaluating the core physics and safety of sodium fast reactors, as well as the development of the approppriated tools to perform such analyses. Indeed, in a first part of this thesis work, the main core physics of the representative sodium fast reactor are assessed, including a detailed analysis of the capability to transmute minor actinides. A part of the results obtained have been published in Annals of Nuclear Energy [96]. Moreover, by means of the analysis of a hypothetical Spanish nuclear scenario, the availability of natural resources required to deploy an specific fleet of fast reactor is assessed, taking into account the breeding properties of such systems. This work also led to a publication in Energy Conversion and Management [97]. In order to perform those and other analyses, several models of the ESFR core were created for different codes. On the other hand, in order to perform safety studies of sodium fast reactors, high fidelity multidimensional analysis tools for sodium fast reactors are required. Such tools should integrate neutronic and thermal-hydraulic phenomena in a multi-physics approach. Following this motivation, the neutron diffusion code ANDES is assessed for sodium fast reactor applications. ANDES is the nodal solver implemented inside the multigroup pin-by-pin diffusion COBAYA3 code, and is based on the analytical method ACMFD. Thus, the ACMFD was verified for SFR applications and while doing so, some limitations were encountered, which are discussed through this work. In order to solve those, some new developments are proposed and implemented in ANDES. Moreover, the code was satisfactorily coupled with the thermal-hydraulic code SUBCHANFLOW, recently developed at KIT. Finally, the different implementations are verified. In addition to those developments, the node homogenized multigroup cross sections and other neutron parameters were obtained for the ESFR core using ERANOS and SERPENT codes, and employed afterwards by ANDES to perform steady state calculations. Moreover, as a result of the UPM contribution to the safety package of the CP-ESFR project, the point kinetic parameters required by the typical plant thermal-hydraulic codes were computed for the ESFR core using SERPENT, which final aim was the assessment of the impact of minor actinides in transient behaviour. All in all, the thesis provides a systematic and multi-purpose approach applied to the assessment of safety and performance parameters of Generation-IV SFR, using existing and newly developed analytical tools. An important amount of time was employed in identifying the limitations that the analytical nodal diffusion methods present when applied to fast reactors following a multigroup approach, and interesting solutions are proposed in order to overcome them.

D-depth profiling in as-implanted and annealed Li-based breeder blanket ceramics

In future power plants (i.e. DEMO), the nuclear fusion of hydrogen isotopes will be used for energy production. The behaviour of hydrogen isotopes in lithium-enriched ceramics for breeder blankets (BBs) is one of the most important items to be understood. In this paper we present the chemical, microstructural and morphological features of Li4SiO4, Li2TiO3 and a third ceramic candidate with a higher Li:Si proportion (3:1), implanted with D at an energy of 100 keV and at room temperature at a fluence of 1 × 1017 cm−2. The D depth-profile in as-implanted and annealed ceramics (at T ⩽ 200 °C) was characterised by Resonance Nuclear Reaction Analysis (RNRA). The RNRA data indicate that the total amount of D is retained at room temperature, while annealing at 100 °C promotes D release and annealing at T ⩾ 150 °C drives D to completely desorb from all the studied ceramics. D release will be discussed as a function of the microstructurural and morphological features of each material.