Water-cooled Pb 17Li Research Articles

04/01298 On the nuclear response of the water-cooled Pb-17Li test blanket module for ITER-FEAT: Chiovaro, P. et al. Fusion Engineering and Design, 2003, 69, (1–4), 469–477

04/01298 On the nuclear response of the water-cooled Pb-17Li test blanket module for ITER-FEAT: Chiovaro, P. et al. Fusion Engineering and Design, 2003, 69, (1–4), 469–477

Further improvements of the water-cooled Pb–17Li blanket

The water-cooled lithium–lead (WCLL) blanket is based on reduced-activation ferritic–martensitic steel as the structural material, the liquid alloy Pb–17Li as breeder and neutron multiplier, and water at typical PWR conditions as coolant. It was developed for DEMO specifications and shall be tested in ITER. In 1999, a reactor parameter optimization was performed in the EU which yielded improved specifications of what could be an attractive fusion power plant. Compared to DEMO, such a power reactor would be different in lay-out, size and performance, thus requiring to better exploit the potential of the WCLL blanket concept in conjunction with a water-cooled divertor. Several new approaches are currently under evaluation. This paper outlines several specific modifications, it highlights progress made on various issues and outlines the R&D work which is still required to define an improved reference design for the WCLL concept.

ITER test-blanket safety assessment

One of the missions of the International Thermonuclear Experimental Reactor (ITER) project is to serve as a test bed for various test blanket concepts. Part of the design for these test modules is to assess their impact on the safety of ITER. This paper presents a set of safety analyses, which were performed on a consistent basis of analysis assumptions and safety requirements. The four basic concepts investigated are a helium-cooled ceramic breeder, a water-cooled ceramic breeder, a water-cooled Pb-17Li breeder, and a liquid lithium self-cooled concept. The analysis performed so far has confirmed the expectation of little safety impact from the test modules, except in a few areas. The use of liquid lithium poses well-known risks of chemical reactivity, which need to be further addressed. For the water-cooled Pb-17Li concept the need was confirmed for further R&D on reactivity of Pb-17Li with water to quantitatively assess bounding scenarios. For the water-cooled ceramic test module the analysis has identified the issue of excessive Be-steam reactions, which needs further investigations.

On the use of tin–lithium alloys as breeder material for blankets of fusion power plants

Tin–lithium alloys have several attractive thermo-physical properties, in particular high thermal conductivity and heat capacity, that make them potentially interesting candidates for use in liquid metal blankets. This paper presents an evaluation of the advantages and drawbacks caused by the substitution of the currently employed alloy lead–lithium (Pb–17Li) by a suitable tin–lithium alloy: (i) for the European water-cooled Pb–17Li (WCLL) blanket concept with reduced activation ferritic–martensitic steel as the structural material; (ii) for the European self-cooled TAURO blanket with SiC f/SiC as the structural material. It was found that in none of these blankets Sn–Li alloys would lead to significant advantages, in particular due to the low tritium breeding capability. Only in forced convection cooled divertors with W-alloy structure, Sn–Li alloys would be slightly more favorable. It is concluded that Sn–Li alloys are only advantageous in free surface cooled reactor internals, as this would make maximum use of the principal advantage of Sn–Li, i.e., the low vapor pressure.

Potential and limits of water-cooled Pb–17Li blankets and divertors for a fusion power plant

Blankets and divertors are key components of a fusion power plant. They have a large impact on the overall plant design, its performance and availability, and on the cost of electricity. The water-cooled Pb–17Li (WCLL) blanket uses reduced activation ferritic–martensitic steel as structural material. It was previously validated under numerous aspects such as TBR, mechanical and thermo-mechanical stability, thermal–hydraulics, MHD, safety and others. This was done assuming the specifications for a European DEMOnstration reactor which were fixed back in 1989. A WCLL blanket would best be combined with a water-cooled divertor so that a single coolant could be used for the entire reactor. Several divertor designs were proposed recently. This paper investigates the applicability of the WCLL blanket concept and a water-cooled divertor in attractive power reactors with increased power densities compared with DEMO.

Hydrogen and deuterium permeation measurements on the double-wall tubes material for the water-cooled Pb–17Li DEMO blanket

The aim of this study is to measure the hydrogen/deuterium transport and inventory parameters in the double-wall tubes (DWT) material for the water-cooled Pb–17Li DEMO blanket. Measurements are conducted using a time dependent permeation method over the temperature range 473–723 K, with hydrogen and deuterium driving pressures in the range 1–10 5 Pa (0.01–1000 mbar). The samples are also characterised using optical microscopy, SEM, microhardness measurements and mechanical tests. An effective permeability of about 12 times lower than 2.1 mm of bare F82H steel at 573 K and three times lower than 2.1 mm of bare F82H steel at 723 K have been measured. This result is close to the theoretical prevision (multi-layer model).

Objectives feasibility assessment of the water-cooled lithium–lead mock-up testing in ITER

This paper presents a short description of the latest design version of the water-cooled Pb–17Li test-blanket module system and the proposal for the test program to be performed during the ITER Basic Performance Phase. The present R&D program foresees to start the test on the first day of ITER operation. The test-blanket module is expected to use all technologies required for the corresponding DEMO blanket (e.g. same structural material, double-wall tubes, permeation barriers) presently under development. Main testing objectives are the demonstration of the system functionality and the validation of the predictions obtained from the theoretical analyses. The feasibility of such objectives is discussed from the point of view of instrumentation and interpretation.

Hydrogen isotope permeation through and inventory in the first wall of the water cooled Pb–17Li blanket for DEMO

The hydrogen isotope transport through the first wall between plasma and first wall coolant has been investigated. The time dependence of hydrogen isotope permeation through the first wall from plasma into water coolant has been calculated. The hydrogen isotope inventory (both mobile and trapped hydrogen isotopes) in the first wall has been evaluated. The influence of temperature gradient, surface conditions, isotopic effect and trapping in the ion- and neutron-induced defects on the hydrogen isotope permeation and inventory has been considered.

Design development and manufacturing sequence of the European water-cooled Pb-17Li test blanket module

In 1996, the European Community started the development of a water-cooled Pb-17Li blanket test module for ITER. First tests are currently scheduled to start with the beginning of the Basic Performance Phase prior to d–t operation. The test module is designed to be representative for a DEMO breeding blanket and relies on the liquid alloy Pb-17Li as both tritium breeder and neutron multiplier material, and water at PWR pressure and temperature as coolant. The structural material is martensitic steel. The straight, box-like structure of this blanket confines a pool of liquid Pb-17Li which is slowly circulated for ex-situ tritium extraction and lithium adjustment. The box and the Pb-17Li pool are separately cooled, the former with toroido-radial tubes, the latter with a bundle of double-walled U-tubes, equally made of martensitic steel and equipped with a permeation barrier. This paper presents the latest design and three manufacturing schemes with different degrees of technology. Advanced techniques such as solid or powder HIP are proposed to provide design flexibility. With a 3D neutronics analysis, the power and tritium generation were determined.

Neutronic and photonic analysis of the single box water-cooled lithium lead blanket for a DEMO reactor

The water-cooled Pb–17Li demonstration plant (DEMO) breeding blanket line was selected in 1995 as one of the two EU lines to be further developed in the next decade. In this paper the results of a neutronic and photonic analysis of the `single box' concept is presented. A full three-dimensional model, including the whole assembly and many of the DEMO reactor components, has been developed, together with a three-dimensional neutron source. A tritium breeding ratio (TBR) value of 1.16, with no ports and a Li 6 enrichment of 90%, has been obtained and a further analysis has been performed to determine Li 6 enrichment that would still ensure tritium breeding self-sufficiency. Selected power densities, calculated for the different materials and zones, are also presented. Material damage through displacements per atom and helium gas production has been investigated. Some shielding capability considerations are presented too.

Development of the EU water-cooled Pb-17Li blanket

The reference concept of the EU water-cooled Pb-17Li DEMO blanket is essentially formed by a directly-cooled steel box having the function of Pb-17Li container and by a double-wall U-tube bundle, immersed in the liquid metal, in which the pressurised water-coolant flows. The structural material is martensitic steel. All blanket performances satisfy DEMO requirements, such as tritium breeding self-sufficiency, capability of the box-structures to withstand water-coolant pressure under faulted conditions, limitation of the Pb-17Li/steel interface temperature to 480°C, power conversion efficiency of about 35%, and acceptable increase of temperature in case of out-of-vessel LOCA. A preliminary design of the test module and of the associated circuit components to be tested in ITER has been performed taking into account ITER specifications for test ports and ITER pulsed working conditions. Test-module DEMO-relevance has been one of the leading criteria for the module design. Preliminary module manufacturing sequences have been defined. The paper recalls the associated R and D activities which have to be finalised prior to the ITER testing and mainly involve tritium permeation barrier development, tritium extraction from Pb-17Li, advanced manufacturing techniques, definition of water/Pb-17Li interaction counter-measures, and Li-concentration control techniques.

LIBRETTO-3: modelling tritium extraction/permeation and evaluation of permeation barriers under irradiation

Permeation barriers are required in order to limit the size and cost of the detritiation plants for future fusion reactor blankets of the water-cooled Pb 17Li type. The LIBRETTO irradiations were performed to evaluate the efficiency of permeation barriers under high flux reactor (HFR) conditions. Tritium extraction and permeation characteristics from Pb 17Li under variable temperatures 553–723 K, H 2 doping (0–1 vol%) and purge gas flow rates 20–100 scc/min were tested in LIBRETTO-3. An external TiC coating, an internal (TiC + Al 2O 3), both produced by chemical vapour deposition (CVD), and an internal Al 2O 3 produced by pack cementation (PC) on AISI 316L steel were tested as permeation barriers. The release mechanisms, experimental uncertainties and method for permeation barriers qualification are presented. As a result permeation reduction factors (PRF) at 0.1 dpa of 17 and 34 were obtained for the CVD-Al 2O 3 at 498 K and for the PC-Al 2O 3 at 508 K, respectively. These values were confirmed by a residence time analysis and are higher than in a preliminary analysis [1].

Tritium Confinement Requirements for the Water-Cooled Pb-17Li Blanket for DEMO

In the water-cooled liquid Pb-17Li blanket concept for DEMO the limitation of tritium permeation from the breeder material into the cooling water will be required. In order to find out on what conditions this tritium permeation remains within reasonable limits, a 1-d FEM code was developed which evaluates the tritium partial pressure in Pb-17Li, the tritium inventory in the blanket material, and the tritium permeation from the Pb-17Li into the cooling water as a function of the permeation reduction factor of a barrier and the efficiency of the tritium extraction from Pb-17Li. With a parametric study the conditions were identified which allow a permeation rate of as little as 1 g.d -1 without pushing the requirements for permeation barriers and extraction efficiencies excessively far. An example is a barrier with a permeation reduction factor of 75 together with an extractor efficiency of approximately 83%. In these conditions the expected tritium inventory is attributed to approximately one third to the martensitic blanket structure (some ten grams) while two thirds will be found in the Pb-17Li. These inventory values are two orders of magnitude lower than in solid breeder blankets and are thus not considered a critical issue.

Analysis of tritium behaviour and recovery from a water-cooled Pb17Li blanket

Analysis of tritium behaviour and recovery from a water-cooled Pb17Li blanket

Analysis of tritium behaviour and recovery from a water-cooled Pb17Li blanket

Abstract The question of the tritium recovery in water-cooled Pb 17Li blankets has been under investigation for several years at JRC Ispra. The method which has been more extensively analysed is that of slowly circulating the breeder out from the blanket units and of extracting the tritium from it outside the plasma vacuum vessel by helium gas purging or vacuum degassing in a suited process apparatus. A computerized model of the tritium behaviour in the blanket units and in the extraction system was developed. It includes four submodels: (1) tritium permeation process from the breeder to the cooling water as a function of the local operative conditions (tritium concentration in Pb 17Li, breeder temperature and flow rate); (2) tritium mass balance in each breeding unit; (3) tritium desorption from the breeder material to the gas phase of the extraction system; (4) tritium extraction efficiency as a function of the design parameters of the recovery apparatus. In the present paper, on the basis of this model, a parametric study of the tritium permeation rate in the cooling water and of the tritium inventory in the blanket is carried out. Results are reported and discussed in terms of dimensionless groups which describe the relative effects of the overall resistance on tritium transfer to the cooling water (with and without permeation barriers), circulating Pbl7Li flow rate and extraction efficiency of the tritium recovery unit. The parametric study is extended to the recovery unit in the case of tritium extraction by helium purge or vacuum degassing in a droplet spray unit.

European Reference Design of the Water-Cooled Lithium-Lead Blanket for a Demonstration Reactor

The water-cooled Pb-17Li blankets represent one of the blanket lines selected within the European Union for DEMO-relevant design and R&D activities. This paper gives a presentation of the reference...

Influence of a magnetic field on the corrosion of austenitic and martensitic steels by semi-stagnant Pb17Li

The influence of a magnetic field on the compatibility of 316L austenitic and 1.4914 martensitic steels with Pb17Li has been studied in conditions simulating the special features of the water-cooled Pb17Li blanket (low Pb17Li velocity, significant radial thermal gradient and short distances between hot and cold zones). In the 420–475°C temperature range, the results show an increase of the corrosion rate in the presence of a magnetic field. This increase is about 50% for 316L steel and 30% for 1.4914 martensitic steel. Moreover the magnetic field induces a loss of symmetry in the deposition process: the amount of recovered deposit is greater in the direction parallel to the magnetic field than in the perpendicular one.

LIPSIE device: Pb-17Li irradiation in water loop with on-line tritium measurements

Due to the low solubility and high diffusivity of tritium in Pb-17Li, tritium permeation through the walls of a water-cooled Pb-17Li blanket has to be investigated. This has been done in Lipsie irradiation performed by CEA in the OSIRIS reactor. The Isabelle 4 PWR loop, in the OSIRIS reactor pool, has been adapted to allow us to irradiate in demineralized water at 15 MPa and 300°C a 316L stainless steel capsule of 0.025 m diameter, containing about 0, 450 kg of Pb-17Li manufactured with natural lithium. Two tritium measurements systems have been specially developed around radioactivity scintillation counting devices: one for tritium in cooling water out of the capsule and one for tritium in helium gas after sweeping or sparging in Pb-17Li inside the capsule. This paper describes the loop used for this irradiation and the tritium measurement systems. The experiment will enable us to estimate tritium permeation rate through the capsule wall in Demo-representative conditions of tritium in Pb-17Li and also study the effect of oxidation of the external capsule wall on tritium permeation by oxygen injection near the capsule.

European Research and Development Programme for Water-Cooled Lithium-Lead Blankets Present Status and Future Work

The European R&D programme in support of the development of water-cooled Pb-17Li blankets for DEMO aims at improving the data base concerning tritium behaviour and compatibility between blanket materials. The four main areas of the experimental programme are structural material corrosion by Pb-17Li, tritium extraction and permeation control, Pb-17Li physico-chemistry, and water/Pb-17Li interaction. This paper describes the most significant results obtained to date in the various experiments performed in Europe and the future programme required to complete the data base by 1994.

Water-Cooled Lithium-Lead Blanket Design Studies for DEMO Reactor: Definition and Recent Developments of the Box-Shaped Concept

AbstractWithin the framework of the European DEMO blanket study programme, this paper describes the design work performed on the “box-shaped” concept for water-cooled Pb-17Li blankets in the past two years. The first part of the paper describes the general design, and the thermo-mechanical and neutronical analyses of the main segment structure leading to the optimization of the thickness of the various component. The second part describes specific analysis of particular segment regions, like water-and Pb-17Li-headers, box bottom and support, which have required detailed design development. All results indicate that the “box-shaped” concept is a very promising candidate for DEMO reactor blankets.