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The tokamak cooling water system (TCWS) is the primary cooling system of the ITER tokamak machine, providing cooling water to the vacuum vessel and in-vacuum vessel components. In addition, it prov...

The Water Cooled Lithium Lead (WCLL) blanket is based on the eutectic liquid alloy Pb16Li as breeder material and neutron multiplier, and pressurised water as coolant. The liquid breeder flows at few mm/s in the blanket module while the pressurised water is circulated inside double-wall tubes. In spite of the adoption of double-wall tubes for the coolant, the probability of a water large leak because of a tube rupture accident cannot be considered negligible. As a consequence, the Pb16Li/water interaction due to a large break in one or more cooling tubes still remains one of the biggest concerns for this blanket concept. This paper reports the results of three experimental tests on Pb16Li/water interaction carried out at ENEA-Brasimone operating the LIFUS 5 facility. Water was injected into the reaction tank, containing Pb16Li at 330 C, at a pressure of 155 bar with different values of sub-cooling and with different free volumes in the reaction system. In addition, post test analyses with SIMMER III code are presented in order to compare the pressure evolution measured during the experiments with that calculated by the code. © 2013 Elsevier B.V.

A crucial issue for the design of HCLL (Helium Cooled Lead Lithium) Test Blanket Module of ITER and HCLL, WCLL, DCLL Breeder Blanket of DEMO is to efficiently characterise the tritium inventory inside the blanket and the permeation of tritium into the coolant in order to reduce as much as possible the radiological hazard towards the external environment. A fast and reliable sensor is required both for tritium management at the several steps of the reactor fuel cycle and for the development of tritium processing systems, e.g. the TES (Tritium Extraction System).

The HCPB (Helium Cooled Pebble Bed) Test Blanket Module (TBM), developed in EU to be tested in ITER, adopts a ceramic containing lithium as breeder material, beryllium as neutron multiplier and helium at 80 bar as primary coolant. In HCPB-TBM the main function of Tritium Extraction System (TES) is to extract tritium from the breeder by gas purging, to remove it from the purge gas and to route it to the ITER Tritium Plant for the final tritium processing. In this paper, starting from a revision of the so far reference process considered for HCPB-TES and considering a new modeling activity aimed to evaluate tritium concentration in purge gas, an updated conceptual design of TES is reported.

The knowledge of the tritium transport parameters in lead lithium is fundamental for the design of the HCLL (helium cooled lead lithium) blanket. In fact, the inventory of tritium in fusion reactors blankets and the permeation of tritium into the blanket coolant, with the consequent leaks toward the environment, are strongly depending on its solubility and diffusivity in the lead alloy PbLi. Several experiments, devoted to investigate the function linking the tritium solubilised in lead lithium with the corresponding tritium partial pressure at equilibrium, were carried out in the past, but significant uncertainties still remain. A detailed analysis of the past experimental works is carried out in this paper with the aim to investigate the main problems occurred in the facilities used to measure the tritium solubility in PbLi that caused such a big spread in the achieved results. On the basis of this analysis, a new a multipurpose laboratory scale apparatus has been designed. The apparatus is able to measure the tritium solubility and diffusivity in PbLi in the range of temperature 300–550 °C and it will be operated with hydrogen partial pressure in the range 10 2 –10 4 Pa. The facility can work with desorption and absorption technique. Moreover, the apparatus has been designed to allow the testing of H/D concentration sensors in Pb–15.7Li in operative conditions relevant to the HCLL–TBM and the characterisation of hydrogen permeation barrier.

The Coolant Purification System (CPS) together with the Tritium Extraction System (TES) and Helium Cooling System (HCS) are the principal auxiliary circuits of HCLL and HCPB TBMs to be tested in IT...

The coolant purification system (CPS) together with the tritium extraction system (TES) and helium cooling system (HCS) are the principal auxiliary circuits of helium-cooled-lithiium–lead (HCLL) and helium-cooled-pebble-bed (HCPB) test blanket modules (TBMs). To extract heat from TBMs, Helium is used as primary coolant. CPS is used to extract tritium from the helium primary circuit as well as to guarantee removal of impurities which could interact with structural material. The reference process proposed for CPS is composed of 3 main successive steps. Step 1 consists in oxidation of Q 2 and CO to Q 2 O and CO 2 using a copper oxide bed ( Q represents either: H, D or T). Step 2 is dedicated to the removal of water which is adsorbed together with CO 2 on molecular sieve bed. Step 3 will remove residual impurities using a heated getter. Based on the operating conditions of CPS (pressure, flowrate, temperature) and on an estimation of the impurities foreseen, this paper presents a design of the oxidising bed which fulfils all requirements in terms of efficiency and lifespan. The design is obtained using a phenomenological approach taking into account competition between oxidation of CO and Q 2 on the metal oxide. The model was implemented in matlab software. A column of 0.41 m large and 2 m long containing 480 kg of CuO is proposed to assure complete oxidation of Q 2 for 16 months long.

Extraction of tritium from liquid lead lithium eutectic alloy is a key topic for the feasibility of any PbLi based tritium breeding blanket (BB). Particularly in DEMO, high tritium extraction efficiency will be required in order to keep low the tritium concentration in the Pb-16Li loop. This is essential to minimize tritium release into the environment and tritium permeation from BB into the primary cooling system. In addition, the tritium extraction process needs to be highly reliable in order not to impact negatively on the operation of the whole fusion reactor, ITER or DEMO.In the present paper, a critical review of the main candidate technologies for tritium extraction from Pb-16Li, particularly gas liquid contactors and vacuum permeators, is accomplished. The intrinsic limits and possible advantages of these technologies are presented and discussed, in the light of considerations coming directly from mathematical models describing their behaviour as well as from the experimental results so fa...

The design of the auxiliary systems for the two European Test Blanket Modules is an important engineering task that will allow the successful integration of the two DEMO Blanket mock-ups into the ITER machine, and, through these experiments, demonstrate the technological feasibility of a blanket module for a future fusion power plant. These auxiliary systems are mainly circuits devoted to the removal of thermal power and tritium recovery from the blanket modules, as follows: • The Helium Cooling System (HCS); • The Coolant Purification System (CPS); • The Tritium Extraction System, divided in two subsystems in case of Helium Cooled Lithium Lead (HCLL), the Tritium Extraction Unit and the Tritium Removal System; • The Lead Lithium Loop (HCLL). Starting from the preliminary design developed in the past, the optimization and the engineering design of these subsystems was performed in the first year of activity of the TBM Consortium of Associates. The input parameters were reviewed and, when possible, DEMO relevant solutions were proposed. The present design takes in account both the TBM operational requirements as well the ITER operation, safety, and spatial constrains. In this paper the current loops design with accent on the integration into the ITER machine are presented.

Chronical releases of tritium from the helium primary coolant into the water secondary coolant is a fundamental safety issue in the design of a fusion reactor steam generator. It is well known that the steam/water circuit of a fusion reactor would be considered not relevant from a radiological point of view, while if a strong permeation of tritium will be present it will be released together with incondensable gases in the condenser. The permeation of hydrogen isotopes through candidate steam generator materials in different conditions was studied in the past. Further experiments demonstrated that nickel alloys of nuclear interest are always covered by a thin and adherent oxide layer able to reduce permeation of orders of magnitude. The major objective of this work is the evaluation of the permeated flux through nickel alloys, when exposed to pure hydrogen and to an oxidant gas stream, to verify the real permeability of these materials in conditions close to those foreseen in the helium side of the steam generator.

In the frame of the IP-EUROTRANS Project, an experimental program, focused on studying the LBE/water interaction has been performed using the LIFUS 5 facility available at ENEA-Brasimone. The physical effects and the possible consequences of this interaction have been evaluated over a wide range of different conditions. Besides the experimental activities, a numerical simulation activity has been performed with SIMMER code in order to better investigate the thermo-hydraulic phenomena involved in the interaction and to confirm the capabilities of the code to simulate this kind of phenomena. The experimental and the calculated results in terms of pressure and temperature evolutions in the system show a good agreement.

European test blanket modules (TBM) for both European blanket concepts, namely the Helium Cooled Lithium Lead (HCLL) and the Helium Cooled Pebble Bed (HCPB), and their auxiliary systems have been designed in the last years. This paper presents the experimental facilities available in Europe for the qualification tests of European TBMs. In particular, devices designed to analyze specific technological issues, for example corrosion under strong magnetic fields or tritium management and extraction, as well as large scale facility to perform integral tests are described. A preliminary overview on the qualification strategy is presented and discussed.

The HCPB (Helium Cooled Pebble Bed) and HCLL (Helium Cooled Lithium Lead) Test Blanket Modules (TBMs), developed in EU to be tested in ITER, adopt helium at 80 bar as primary coolant. This paper contains a conceptual design of the TBMs Coolant Purification System (CPS) based on the need to remove permeated tritium and gas impurities. The following steps have been considered: identification of CPS design requirements; review of the purification systems developed for Helium Cooled Fission Reactors and proposed for DEMO Fusion Reactor; selection of the most promising technologies for CPS; indications on instrumentation and procedures for tritium balance. The proposed solution is a three-stage process constituted by an oxidiser to convert Q 2 and CO to Q 2 O and CO 2 , an adsorption step, performed on molecular sieve at room temperature to remove Q 2 O and CO 2 , and a final step performed on a heated getter to remove residual impurities.

The interaction between heavy liquid metal, such as LBE, and pressurized water has been analyzed at ENEA Brasimone Research Centre, in order to investigate the evolution of this phenomenon over a wide range of conditions. The study includes an experimental campaign on LIFUS 5 facility and a numerical simulation activity performed with SIMMER III code. The first test of the experimental program was carried out injecting water at 7 MPa and 235 °C in a reaction vessel containing LBE at 350 °C. A pressurization up to 8 MPa was observed in the test section during the short term (about 2 s) of the transient. In the post-test analysis performed with SIMMER III code, two different geometrical models were developed in order to reproduce in the best manner the experimental results and, therefore, to confirm the code's capabilities of reproducing the phenomenology of the LBE–water interaction. The data calculated through both models agreed in a good way with the experimental results, despite the necessary simplifications adopted in the models due to the 2D features of the code.

Ignitor is a nuclear fusion experiment aimed at studying Deuterium-Tritium plasmas. If European proposed waste management strategies were applied, all Ignitor radioactive materials could be...

One of the most challenging issues for testing the different Test Blanket Modules (TBM) concepts in ITER is the demonstration of the ability to correctly and efficiently manage the bred tritium. For all TBM concepts the main auxiliary systems involved in this task are the Tritium Extraction System (TES), whose main purpose is to extract tritium from the breeder, making it available to the Tokamak Exhaust Processing system in the ITER Tritium Plant, and Coolant Purification System (CPS) which extracts the permeated tritium from He-coolant and keeps controlled the chemistry of the primary cooling circuit. Both these systems have to be compatible with: - the foreseen requirements from the TBM experimental campaign; - the ITER space requirements; - the requirements in terms of interface with the ITER Tritium Plant. In this paper, the composition and flow-rate of the gas stream to be processed by the tritium extraction system and coolant purification system in the Helium Cooled Pebble Bed Test Blanket Module is given for the high duty DT phase of ITER. Moreover, in the light of the operative conditions planned for the experimental campaign on the HCPB-TBM, a preliminary design of the TES and CPS is presented and discussed, with a first estimation of the required space for their installation.

The objective of the European 6th framework project EUROTRANS is to demonstrate the technical feasibility of transmutation of high level nuclear waste using accelerator driven systems. Within this objective the design of a European experimental ADS should demonstrate the technical feasibilities to transmute a sizeable amount of waste and to operate an ADS safely. This ADS will be a subcritical reactor system having liquid lead–bismuth eutectic as coolant. The liquid lead–bismuth eutectic is also intended to serve as target material for the spallation reaction which forms a crucial part to the subcritical reactor core. Since lead–bismuth eutectic is used as core coolant and spallation material, knowledge of its thermal hydraulic behaviour is essential. Within the DEMETRA domain of the EUROTRANS project, basic thermal hydraulic studies in order to support the design and safety analysis of XT-ADS components and the development of measurement techniques have been started.

A low-activation glass–ceramic based on silica, alumina and yttria has been designed and tested as joining material for 2D fusion grade SiC/SiC. Neutron-induced radioactivity of elements present in the glass has been simulated by European Activation System EASY-2007 code package. The mechanical strength of the joined SiC/SiC has been tested by 4-point bending on three different kinds of joined samples. Bending strength higher than 120 MPa has been measured at room temperature, with composite failure in most cases.

The interaction between heavy liquid metal (HLM) and water is a safety concern for the preliminary designs of lead fast reactor (i.e. LFR) and of subcritical transmutation system prototypes (i.e. XT-ADS). Current pool-type configurations have steam generators (SG) inside the reactor vessel. This implies that the primary to secondary leak (e.g. steam generator tube rupture) shall be considered as a postulated initiating event. The issue is addressed for CIRCE facility in ICE (Integral Circulation Experiment) configuration. CIRCE facility is a large pool system aimed at studying key operating principles of Lead Bismuth Eutectic (and Lead) systems. The configuration ICE was carried out to perform integral experiments, simulating the coupling between a high-performance heat source (electrically heated fuel bundle) and the heat exchanger, which was representative of the preliminary design of the XT-ADS heat exchanger. A Failure Mode and Effect Analysis (FMEA) is applied in order to get a complete picture of all the failure modes pertaining to this system, to determine their effects and to classify them according to their severity. The outcome of the analysis has identified as major hazard, relative to the CIRCE facility in the ICE configuration, the risk related to the LBE/water reaction, although with a very low probability, with the potential for a suddenly and dangerous pressurization (beyond the failure threshold) within the main vessel. A SIMMER-III code model of the system has been setup to provide deterministic results of the scenario. The results are supported by means of a LBE/water interaction experiment executed in LIFUS5 facility. LIFUS5 is a separate effect test facility dedicated to the investigation of LBE/water interaction. SIMMER-III code pre-test and post-test analyses are performed to define the boundary conditions of the experiment and to demonstrate the reliability of the code in simulating the phenomena of interest. The activity contributes to solving the safety issue raised for the operation of CIRCE facility and it provides a sample approach for addressing the safety studies needed in the development of the lead fast reactor and of the subcritical transmutation system.

When dealing with activities related to the design and optimisation of the H elium C ooled L ithium L ead (HCLL) blanket, doubts arose in the past about the technical feasibility of the tritium processing system. The questionable feasibility was due to the heavy performance requirements for different tritium processing subsystems, mainly consequence of the high tritium permeation rate from the breeding blanket into the main cooling circuit, which was claimed as an intrinsic characteristic of such a blanket concept. This was directly related to the supposed very low tritium solubility in the liquid lead lithium in eutectic composition. Recent data on the hydrogen-PbLi Sieverts’ constant achieved at ENEA CR Brasimone are one order of magnitude higher than those so far assumed as reference. Because of the consequent much lower tritium permeation rate from the liquid metal into the primary cooling circuit, the performance requirements for the main tritium processing subsystems/technologies in HCLL blanket for PPCS reactor are significantly relaxed. Nevertheless, further efforts have to be focused in lowering the required C oolant P urification S ystem (CPS) feed flow-rate which still remains the most critical tritium processing subsystems.

The possibility of clearance of vanadium–chromium–titanium (V–Cr–Ti) alloys is analysed. These alloys after their service in fusion power plants, have the potential to reach clearance if they are purified from activation products. The extraction part of the technological scheme for radiochemical separation of components of irradiated V–Cr–Ti alloy and their purification from metallic activation products, developed earlier, was tested for the first time in laboratory conditions using activated alloy specimens. The replacement of the acid reextraction of V with peroxide and of acid reextraction of Cr with alkaline improved characteristics of the extraction reprocessing. Duration of the V and Cr reextraction was shortened by about an order of magnitude, the output of these alloy components was increased, V purification from rare-earth metals became two times as great, and Cr decontamination from Co increased by two orders of magnitude. Activation of Be contaminated with trace quantities of uranium is an issue: estimation of Be activation in the blanket of the Power Plant Conceptual Study (PPCS) has suggested that traces of U impurity in Be should be removed – or substantially reduced – prior to use.

The first and main step of the HCLL (helium cooled lithium lead) blanket fuel cycle consists of tritium extraction from the liquid breeder. In DEMO reactor a dedicated system, called TES (tritium extraction system), will be devoted to accomplish this task. Different technologies, belonging to the families of gas–liquid contactors, getters and tritium permeators have been studied in the last years, some of them mainly theoretically, others with much more experimental effort. Gas–liquid contactors, which have been more experimentally studied, showed contradictory results. Plate, spray and bubble columns resulted in a very low-extraction efficiency. On the contrary, packed columns gave better results, although some doubts arise about the accuracy of the experiments performed in the past. However, packed columns seem to be the most attractive technology and for this reason was decided for a further optimisation and testing of their performances. In order to accomplish this task, a dedicated facility called TRIEX (tritium extraction) was designed and built and is now available in ENEA-Brasimone. In TRIEX the extraction efficiency of different packed columns will be tested in a systematic way paying particular attention to the hydrogen monitoring system in gas and liquid metal phase. The design and the main instrumentation of TRIEX facility, as well as the operating modes and the planned experimental campaign are presented in this paper.

To develop a sensor for measurements of hydrogen (and its isotopes) concentration in liquid Pb–16Li, two permeating capsules made with niobium and pure iron have been tested. The experimental results demonstrated that the Nb-sensor is not suitable for the foreseen application. The Fe-sensor cannot be operated in equilibrium mode because of the very long time necessary to reach the hydrogen pressure equilibrium. However, the equilibrium mode cannot be definitively excluded. In fact, by means of a substantial optimisation of the sensor geometry, the pressurisation rate could be fast enough to reach acceptable response times. In this context, different optimised solutions have been developed and good performance demonstrated by simulations. The present Fe-sensor can be successfully operated in dynamic mode, since it is possible to link reliably the permeation fluxes to the hydrogen concentration in the liquid metal.

In the helium cooled lithium lead (HCLL) blanket for DEMO the breeding material is represented by eutectic lead lithium alloy flowing in poloidal–radial direction in the breeding module. The low solubility of hydrogen isotopes in the alloy generates an high partial pressure and, as a consequence, an high tritium permeation rate from the breeder into the He coolant and a considerable tritium inventory in the structural materials. Several experiments, devoted to determine the hydrogen solubility in lead lithium, have been performed in the past, giving contradictory results. To find a reliable value to be used in the safety analysis and in the design of both the tritium extraction system and the coolant purification system, the solubility of hydrogen (SOLE) experiment was launched in ENEA Brasimone in 2002. The SOLE device is based on the use of the absorption technique. The experimental results are in good agreement with the ones obtained in the past with similar techniques, varying from 1 × 10 −2 mol m −3 Pa −1/2 up to 7 × 10 −2 mol m −3 Pa −1/2 . The experimental device and the obtained results are presented and discussed in this paper.

In this paper, we present computations linking the tritium release rate to the characteristics of lithium–lead and helium cooling circuits. Impacting component performances are evaluated such as tritium permeation towards the He coolant in the blanket modules, lithium–lead circulation rate, tritium extraction unit efficiency, tritium permeation in steam generator, helium coolant leak rate, helium purification unit maximum flow rate and efficiency. Safety considerations are also taken into account. A finite element model (FEM) for tritium permeation was developed considering various phenomena such as tritium transport by convection and diffusion in lithium–lead, MHD effects on liquid metal flows, tritium permeation in structures with temperature gradients. Other sub-system performances, like He leak rate and efficiency of tritium extraction systems, are discussed via an engineering approach. The results show that a reasonable compromise among the various requirements can be found, leading to technologically achievable requirements for tritium permeation barriers, tritium extraction systems both from Pb–17Li and He, and leak rates from the He cooling system.

The reduction of tritium permeation from the Pb–17Li, or plasma, into the coolant is of crucial importance in order to reduce the radiological hazard in the steam generator vault as well as in the turbine/condenser area and to optimise the tritium balance in the reactor. The use of aluminium rich coatings has been selected as reference solution for the water cooled lithium lead (WCLL) blanket in order to produce reliable tritium permeation barriers (TPB). TPB qualification activities performed in the past allowed the selection of two reference deposition techniques, the chemical vapour deposition (CVD) process developed on laboratory scale by CEA, and the hot dipping (HD) process developed by FZK. On the basis of the results obtained in the past with the Corelli I–II devices, a new apparatus named Vivaldi was designed to perform comparative tests on two hollow cylindrical specimens in the same operating conditions. The performance of alumina coating on EUROFER 97 steel has been tested in gas and liquid metal phase. The obtained results in terms of permeated fluxes and permeation reduction factors (PRF) are herein presented and discussed. A post experiment examination of coatings was performed by use of optical and SEM microscopy.

The interaction between pressurised water and liquid Pb–17Li is a topical issue for the WCLL blanket concept and needs to be studied in detail because of the potentially significant effect on the reliability and safety of the blanket. Particularly, it seems important from the design point of view to deeply investigate the consequences of coolant micro-leaks in the blanket module generated by micro-cracks with a size in the order of 10−3 mm2. This kind of interaction has been studied during an extensive experimental campaign on the RELA loops in the ENEA Research Centre at Brasimone. The results showed a reduction in the liquid metal flow-rate in the circuit and a deterioration of the heat exchange properties between breeder and coolant due to the formation and growth of solid reaction products generated by the chemical reaction between lithium and water. The molar ratio between the recovered hydrogen recovered and water injected was in the range 0.3–0.5, confirming that, in the test conditions, the main solid reaction product is lithium hydroxide (LiOH). The possible impact of these results on the TBM-ITER design is also presented and discussed.

Management of activated waste from fusion power reactors is part of the long-term action of the European Fusion Programme. The options of recycling , the reuse in novel reactors for in-vessel materials, and clearance , the declassification to non-active waste, for ex-vessel materials, were analysed with the aim to reduce the amount of radioactive waste. A further step is the attempt to reduce the long-term radioactivity of selected materials to levels allowing clearance also of in-vessel structures. Clearance can be an alternative option also for in-vessel materials. Vanadium-based alloys belong at the moment to the materials with sufficiently low-activation constituting elements. Attainment of clearance conditions in activated in-vessel V–4Cr–4Ti structures would require very low concentrations of impurities, not achievable with the current purification methods and in some cases below the present detection limits, and the development of methods to reprocess the activated alloy for extracting radiotoxic nuclides.

The next generation of fusion power plants (DEMO) should rely on a breeding blanket (BB) able to efficiently convert the neutrons kinetic energy into heat, to ensure the tritium self-sufficiency and to adequately shield the toroidal field coils from neutrons and gamma rays. The eutectic lithium-lead alloy is a consolidate liquid blanket material, which simultaneously includes the breeder (Li) and the neutron multiplier (Pb). The assessment of the tritium inventory inside the blanket and its environmental release requires knowledge of the hydrogen isotopes interactions with blanket materials, in particular the hydrogen solubility in lithium-lead, which is defined by means of the hydrogen Sievert's constant. Several experiments, aiming to determine the hydrogen isotopes solubility in lithium-lead, have been performed in the past giving values of the temperature-dependent Sieverts' constant, KS , distributed in a wide range (covering about two orders of magnitude on the Arrhenius plot). Starting from a literature review of the $K-{S}$ values, this paper provides a parametric analysis for the most influencing parameters related to tritium migration in helium-cooled and water-cooled lead-lithium DEMO BBs. This analysis has been performed using the computational code FUS-TPC for several operative scenarios and considering the different KS values. This paper demonstrates that the tritium Sieverts' constant in Pb-17Li has a great impact on the assessment of tritium losses, whose value can spread in more than one order of magnitude. Furthermore, the analysis suggests suitable permeation reduction factors to be adopted in the different scenarios as well as the need of addressing new accurate experiments on the solubility constant. © 2014 IEEE.

Tritium mobility through breeding blanket (BB) and steam generator heat transfer areas is a crucial aspect for the design of the next generation DEMO fusion power plants. Tritium is generated inside the breeder, dissolves in and permeates through materials, thus leading to a potential hazard for the environment. For this reason, it is important to carry out the tritium migration analysis for a specific DEMO blanket configuration to predict the released amount of tritium during the plant operation. Unfortunately, tritium assessments are often affected by several uncertainties implying very important modeling and parametric issues. In this paper, the main permeation issues are identified and possible solutions are discussed to address the modeling issues and the parametric uncertainties affecting the T migration assessments for the two DEMO helium-cooled BBs: 1) helium-cooled pebble beds and 2) helium-cooled lithium-lead. For these two helium-cooled blanket concepts various tritium migration analyses will be carried out by means of the computational tool FUS-TPC to define proper and feasible tritium mitigation techniques, which are needed to keep the tritium losses lower than the allowable environmental release (i.e., 20 Ci/d). © 1973-2012 IEEE.

Tritium permeation through Breeding Blanket and Steam Generator heat transfer areas is a crucial aspect for the design of the next generation DEMO fusion power plants. Tritium is generated inside the breeder, dissolves in and permeates through materials, thus leading to a potential hazard for the environment. For this reason it is important to carry out the tritium migration analysis for a specific DEMO blanket configuration in order to predict the released amount of tritium during the plant operation. Unfortunately, tritium assessments are often affected by several uncertainties implying very important modelling and parametric issues. In this study the main permeation issues are identified and possible solutions are discussed to face the modelling issues and the parametric uncertainties affecting the T migration assessments for the two DEMO helium-cooled breeding blankets, i.e.: 1) Helium-Cooled Pebble Beds (HCPB) and 2) Helium-Cooled Lithium-Lead (HCLL). For these two helium-cooled blanket concepts various tritium migration analyses will be carried out by means of the computational tool FUS-TPC in order to define proper and feasible tritium mitigation techniques which are needed to keep the tritium losses lower than the allowable environmental release (i.e. 20 Ci/d). © 2013 IEEE.

Hydrogen dissolves in and permeates through most materials, thus it is important to understand the permeation, diffusion and dissolution phenomena of hydrogen and its isotopes in materials. We address the problem of tritium transport in Helium Cooled Lead-Lithium (HCLL) DEMO blanket from lead lithium breeder through different heat transfer surfaces to the environment by developing a computational code (FUS-TPC). The main features of the code are briefly described and a parametric study is performed in order to identify the most influencing parameters in terms of tritium releases into the environment and of tritium inventories. The results showed that the results are strongly affected by the Lithium and the efficiency of permeation barriers

In the breeding areas a significant fraction of the tritium produced is extracted out from the Breeding Zone by the He gas purging the breeding ceramic in the helium cooled pebble bed (HCPB) blanket concept or transported in solution by the flowing liquid metal alloy in the helium cooled lead lithium (HCLL) blanket concept and then extracted outside the blanket box. Tritium produced inside the breeder can enter the metal structures and can be lost by permeation to the environment. It should therefore be kept under close control throughout the fusion reactor lifetime. In this study the problem of tritium transport in HCPB DEMO blanket has been studied and analysed by means of the computational code FUS-TPC, which has been originally developed to study the tritium transport in HCLL blanket and it has the main goal to provide the total tritium losses into the environment and the tritium inventories inside several HCPB blanket locations (e.g. purge gas and main coolant loops). © 2013 Elsevier B.V.

The interaction between heavy liquid metal (HLM) and water is a safety concern for the preliminary designs of lead fast reactor and of subcritical transmutation system prototypes. Current pool-type configurations have the steam generators inside the reactor vessel. This implies that the primary to secondary leak (e.g. steam generator tube rupture) shall be considered as a safety issue in the design, but also in the preliminary safety analysis, of this reactor types. This requires availability of qualified experiments carried out with initial and boundary conditions representative of the reactor prototype. The objective is to support the development and to demonstrate the reliability of computer codes in simulating the phenomena of interest. LIFUS5/Mod2 is a separate effect test facility, designed for investigating the interaction between heavy liquid metal (HLM) and water, installed at ENEA CR Brasimone. LIFUS5 test facility is refurbished based on the operating experience acquired during the previous experimental campaigns and the support of numerical tools (i.e. SIMMER-III, -IV and, with some extent, RELAP5). The new design involves a new facility configuration, test section, instrumentation and control systems. This paper presents the new LIFUS5/Mod2 facility, the planned experimental campaign and code calculations devoted to support the design modifications, the instrumentation and the final test matrix. The activities are performed in the framework of the EC funded THINS Project and thanks the support of the Italian Ministry for the Economic Development.Copyright © 2012 by ASME

Hydrogen dissolves in and permeates through most materials, thus it is important to understand the permeation, diffusion and dissolution phenomena of atomic hydrogen in materials in which hydrogen and its isotopes are present. In this work the problem of tritium transport from lead–lithium breeder through different heat transfer surfaces to the environment has been studied and analyzed by means of a computational code. The code (FUS-TPC) is a new fusion-devoted version of the fast-fission one called Sodium-Cooled Fast Reactor Tritium Permeation Code (SFR-TPC). The main features of the model inside the code are described. A simulation, using the code, was performed by adopting the configuration of the European configuration of the Helium Cooled Lead Lithium (HCLL) blanket for DEMO.

In the framework of the research activities of the EURATOM FP6 project named ELSY (European Lead-cooled System), aimed at demonstrating the possibility of designing a competitive and safe fast critical reactor based on the Generation IV Lead Fast Reactor (LFR) concept, the study of the lead-water interaction following an incidental SGTR (Steam Generator Tube Rupture) event is an important issue to address. To simulate such event, an experimental test has been carried out on the LIFUS 5 facility at the ENEA Brasimone Research Centre, in order to assess the physical effects and the possible consequences connected to this kind of interaction. The experiment has been conducted by injecting water at the pressure of 185 bar and with a temperature of 300 °C into a volume of 80 l of Lead Bismuth Eutectic (LBE) kept at atmospheric pressure and at a temperature of 400 °C. The experimental facility has been suitably modified in order to reproduce as close as possible the operating conditions of the ELSY Steam Generator Unit (SGU), in which a free volume of cover gas (argon) is foreseen at the top of the system, with the objective to dampen the pressure waves inside the SGU itself. The experimental test has been supported through a numerical modelling campaign performed at the University of Pisa by means of the SIMMER code within both 2-D (SIMMER III) and 3-D (SIMMER IV) models. Pre-test simulations have been carried out to aid the design of the new facility configuration and to select the test conditions which could better reproduce the behaviour expected for ELSY. In addition, a post-test analysis has also been accomplished, allowing to compare the numerical and experimental results, so as to validate and assess the performance of the code when employed for this kind of applications.Copyright © 2010 by ASME

Ignitor is a proposed compact high-magnetic field tokamak aimed at studying plasma burning conditions in Deuterium-Tritium plasmas. Localisation of this experiment in Italy has seen growing attention during the last years. The determination to carry out this project in Italy has driven to taking into account several candidate sites. Recently, due to several technical reasons, it was identified a new site, the Caorso Site, in northern Italy, a nuclear site were a nuclear power reactor was sited. Preliminary Radiological Impact analyses for the siting of Ignitor in Caorso are presented in this paper. Te Ignitor radiological impact turns out to be negligible both during normal operation and in case of accident.

This paper concerns the study of the phenomena involved in the interaction between LBE and pressurised water which could occur in some hypothetical accidents in accelerator driven system type reactors. The LIFUS 5 facility was designed and built at ENEA-Brasimone to reproduce this kind of interaction in a wide range of conditions. The first test of the experimental program was carried out injecting water at 70 bar and 235 °C in a reaction vessel containing LBE at 1 bar and 350 °C. A pressurisation up to 80 bar was observed in the test section during the considered transient. The SIMMER III code was used to simulate the performed test. The calculated data agree in a satisfactory way with the experimental results giving confidence in the possibility to use this code for safety analyses of heavy liquid metal cooled reactors.

One of the most challenging issues for the TBM (Test Blanket Module) testing campaign foreseen in ITER is the operation of TES (Tritium Extraction Systems). This is essential not only to prove the ...

The paper examines the neutronics and activation of Candor, a proposed design evolution of Ignitor in the direction of a device using a DHe fuel cycle. The main characteristics of the Candor machine are described. Activation calculations were done with two codes: ANITA and FISPACT. A comparison of the results for the two codes is carried out. Finally, the paper draws some conclusion on the radioactive waste management options for Candor. No Candor spent material will need to be disposed of as permanent waste in underground repositories. All materials may be recycled, if convenient, after a short interim decay. Concerning clearance, the structure may be declassified after less than 50 years of decay. For magnets, about 100 years of interim decay are necessary. Candor does reach the zero-waste option, without the need of any materials selection, low-activation materials, or further shielding. The final design of a fusion power plant with DHe3 plasmas has yet to be conceived. However, the zero-waste option is a reachable goal for such reactors, as the results for Candor have indicated.

The feasibility of reprocessing a vanadium alloy after its use as a structural material in a fusion reactor, in order to enable its subsequent hands-on recycling within the nuclear industry, has been determined. For less neutron-exposed components, clearance of materials has also been considered. A conceptual model for the radiochemical processing of the alloy has been developed and tested experimentally. Using di-2-ethyl-hexyl-phosphoric acid it is possible to purify the components of the V–Cr–Ti alloy after its exposure in a fusion reactor down to the required level of activation product concentrations.

The reduction of long-term radioactivity is analysed here in V–Cr–Ti alloys, one of the proposed structural materials for fusion power plants. In particular is explored the possibility of recycling within the nuclear industry and of clearance, that is declassification to non-active material. Vanadium alloys have the potential to reach the dose rate hands-on recycling limit when used in a blanket, if noxious radioactive products coming from impurities activation are eliminated. Clearance is also possible in principle, but only if further separation of activation products of titanium is carried out after service.

Ignitor is a proposed compact high-magnetic field tokamak aimed at studying plasma burning conditions in Deuterium–Tritium plasmas. The determination to carry out this project in Italy has driven to taking into account several candidate sites. Recently, due to several technical reasons, it was identified a new site, the ENEL–TERNA plant in Rondissone (TO). This paper deals with two main aspects of the Ignitor safety: accident analysis with dose assessment to personnel and public, and radioactive waste management evaluations. Concerning the first point, accidents with a frequency higher than 10 −7 per year bring to a maximum dose for the most exposed individual (MEI) below 1 mSv. We have applied the Italian waste management regulations to the Ignitor experiment radioactive materials: none of them should be classified in the high level waste (HLW) category and most materials are classified as low level waste (LLW). If European proposed waste management strategies were applied, all Ignitor radioactive materials could be recycled.

In the helium cooled lithium lead (HCLL) and water cooled lithium lead (WCLL) blanket concepts for DEMO correct and reliable management of tritium is of basic importance, both for safety and fuel cycle reasons. To develop a sensor for measurements of hydrogen (and its isotopes) concentration in liquid Pb-17Li, a permeable capsule of niobium was chosen. Different simulations with a mathematical model have been performed, and then the sensor was designed, constructed and tested. The first experimental results in gas phase showed a permeating flux much lower than the predicted one, probably due to the formation of an oxide layer on the capsule surface or to the formation of niobium hydrides. To solve this problem different solutions are presented.

The reduction of tritium permeation through blanket structural materials and cooling tubes has to be carefully evaluated to minimise radiological hazards. A strong effort has been made in the past to select the best technological solution for the realisation of tritium permeation barriers (TPB) on complex structures not directly accessible after the completion of the manufacturing process. The best solution was identified in aluminium rich coatings, which form Al2O3 at their surface. Two technologies were selected as reference for the realisation of coating in the WCLL blanket concept: the chemical vapour deposition (CVD) process developed on laboratory scale by CEA, and the hot dipping (HD) process developed by FZK. The results obtained during three years of tests on CVD and HD coated specimens in gas and liquid metal phase are summarised and discussed.

National regulations on waste management do not seem adequate to cope with the fusion case. The paper shows the relevance of this problem, showing some aspects of application of national radioactive waste management practices and regulations to fusion. In particular, the case of Italian waste management regulations is considered. The waste management strategy proposed in the SEAFP and PPCS studies, based upon recycling and clearance techniques, is compared to Italian national regulations. If those regulations were applied to fusion, a relevant part of the fusion radioactive materials should be classified in the Italian High Level Waste category. Also in the case of other national regulations, fusion waste would be rated mostly in the local `high-level' category. An evolution of those regulations in the future, in order to take into account the special characteristics of fusion radioactive materials, is recommended.

Since several years ENEA is involved in experimental activities concerning the interaction between molten lithium lead alloy, in eutectic composition, and pressurised water in conditions relevant for DEMO fusion reactor, in order to predict the behaviour of a WCLL blanket module in case of an in blanket LOCA. In this ambit LIFUS 5 apparatus was designed and constructed to carry out the experimental campaign on water large leaks into liquid Pb–17Li, with the final aim to provide data for the validation of the mathematical modelling of the related phenomena. After two tests performed to qualify all mechanical components and the data acquisition system, LIFUS 5 was operated at the ENEA C.R. Brasimone for tests nos. 3 and 4. Water was injected into the reaction tank at a pressure of 155 bar with two different values of sub-cooling. The initial liquid metal temperature was fixed to 330 °C. The first pressure peak due to the water vaporization and jet expansion was clearly recognized together with the subsequent pressure increase due to further water injection and hydrogen generation. In the performed experiments the maximum pressure peak, as detected in both reaction and expansion vessels, never overcame the value of the injected water. Moreover, in these two tests a significant temperature increase in the reaction vessel occurred, strictly connected to the amount of injected water.

Fusion is seen as a much ‘cleaner’ energy than fission, and a resource for clean energy in the far future. However, being a form of nuclear energy, fusion shares with fission most of its safety problems. This study concerns the assessment of both the short term and the long term hazards associated with fusion, compared with the same figures for fission reactors. Fission data derive from well-known PWR safety assessments, and in particular from the Italian project Progetto Unificato Nucleare (PUN). Fusion data derive from the latest findings of the European Safety and Environmental Assessment of Fusion Power (SEAFP) programme. Concerning inadvertent intrusion in a radioactive waste disposal site and long-term degradation for fusion, this analysis was performed in the frame of the SEAFP studies, while for fission data are taken from typical European waste disposal sites. In all relevant cases, evaluation of associated risks is carried out, with a comparison of the obtained results.