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New Technologies Development

Generation IV Nuclear Energy Systems (Gen IV)

CVR is involved in the R&D of the Gen IV through participation in several projects and consortia. The GoFastR project is focused on the development of a fast helium-cooled reactor. The MATTER project is dedicated to development and testing of materials for lead-cooled reactors. In the field of R&D on high-temperature gas-cooled reactors, CVR operates the HTHL experimental loop which is, among others, involved in the ARCHER project. Similarly, the SCWL experimental loop is employed for research on the supercritical water reactor. The SCWR-FQT project is specifically dedicated to this issue.

R&D Programmes of the SUSEN Project

The SUSEN project is divided into four parts – research programs:

  • Technological Experimental Circuits (TEO, Technologické a experimentální okruhy)
  • Structural and System Diagnostics (SSD, Strukturní a systémová diagnostika)
  • Nuclear Fuel Cycle (JPC, Jaderný palivový cyklus)
  • Material Research (MAT, Materiálový výzkum)

TEO Programme

TEO is dedicated to the study and detailed description of new coolant media, in particular to their interaction with selected in-vessel components. The use of these coolants is expected for advanced nuclear energy systems (Gen IV). Within the frame of SUSEN, supercritical water, supercritical CO2, helium, and liquid PbLi will be studied. At both sites in Rez and Pilsen, a new experimental infrastructure with several R&D loops will be able to simulate close-to-Gen IV operating conditions with their primary and secondary loops.

SSD Programme

SSD is primarily focused on improvements in the fields of description of the degradation and lifetime of the energy system’s components. This description should lead to assurance of reliable operation of a given facility. It can be achieved through both the mastering of description of degradation mechanisms and the development of advanced non-destructive testing (NDT) methods of given components.

JPC Programme

JPC will participate in the world-wide effort to develop completely novel techniques/methods for RadWaste handling, aiming for a minimisation of its impact on the environment during its final disposal. The programme will be concerned with the improvement of methods of fluoride fraction distillation, as well as the novel methods of waste thermal treatment.

MAT Programme

MAT provides comprehensive support to the research of materials intended for  use in the energy industry. In particular, it concerns testing of the mechanical properties of materials, their microstructure description, as well as the development of novel methods of fusion welding of these materials.

ITER: Test Blanket Modules

Centrum výzkumu Řež is the legal successor of UJV Rez a.s. in the field of fusion R&D activities. The main activities concern: (i) the coordination of  Czech industry involvement within the framework of the Czech Industry for the ITER group and (ii) two technological projects: heat fatigue testing of the First Wall (under Blanket section of ITER) and R&D on the Test Blanket Module (TBM).

The ITER Blanket is a construction component located inside the tokamak vacuum vessel which physically delimits plasma confined by magnetic fields. The Blanket forms a quasi sheathing of the vacuum vessel: it shields radiation, removes thermal power and provides purification of the fusion fuel from ashes. It acts as a primary physical barrier between the fusion plasma and the construction and technological subsystems of the tokamak. Compared with common nuclear fission reactors, the Blanket can be seen like a ‘reactor core baffle’ of VVER, or even like a combination of a core baffle and a reactor basket.

Besides this, the Blanket significantly shields the radiation coming from plasma and thus has a similar role to a reactor pit. Nevertheless, contrary to a reactor pit, the Blanket does not act like a mechanical support – the Blanket itself consists of independent modules, which themselves are attached to dedicated support construction, most often the vacuum vessel of tokamak.

The Blanket has two main parts: First Wall and Diverter. These two parts can be divided into particular construction components, which themselves are composed of various layers of different materials. The First Wall covers almost 80% of the blanket surface, depending on the shape of the chamber. In the case of the most common D-shaped toroid, the First Wall covers about ¾ of the Blanket inner surface. The remaining space of the Blanket is covered with a Divertor. The Divertor is a component which, using a magnetic field, delimits the plasma and the space for collection of impurities. It is located under the so-called X-point of the magnetic field. At this point the magnetic field is bisected and the channels for removal of fusion ash and impurities from the fuel (mixture of deuterium and tritium) are formed here.

According to its primary function, all the components of the Blanket make a physical barrier to delimit plasma, which is then confined in a relatively narrow tunnel inside the toroidal vessel. However, during transition effects leading to generation of Edge Localized Modes (ELM) – intense local increase of plasma parameters at its edge causing huge energy flux deposition on the Blanket wall– a direct contact of plasma with wall materials may occur. Therefore the Blanket has to be built of the most suitable materials combining the following features/properties: high resistance to erosion, low retention of deuterium and tritium (or fusion fuel in general), low magnetic induction, and low production costs. Such materials are called Plasma-Facing Materials (PFM) and components made of them are called Plasma-Facing Components (PFC).

The Primary First Wall and Divertor are both typical PFCs. At present there are several candidates for the PFC materials, such as beryllium, tungsten or CFC (carbon fibre composite). Centrum výzkumu Řež is involved in the Blanket R&D programme by means of the testing of selected material mock-ups of the First Wall. Since the ITER tokamak is designed for pulse operation (with pulse duration of 400 sec.), the First Wall will be continuously facing cyclic heat load in the range of 0–0.5 MW/m2. Beryllium, the most probable PFM candidate for the First Wall, has to be thoroughly tested, as well as the durability of the joint between the CuCrZr alloy support, and Beryllium has to be verified with respect to cyclic changes of the heat flux and the corresponding temperature variations. The joint between Beryllium and CuCrZr support, which serves as a heat sink and removes generated heat power through cooling water, is produced by the Hot Isostatic Pressing (HIP) technique. During this process, the two materials are bonded together by diffusion joint. Such a joint is formed when pressure of tens of MPa and a temperature of about 1100 °C are applied for several hours.

Nevertheless, the durability of this kind of joint is in question and has to be tested. In order to provide the required heat load tests, Centrum výzkumu Řež operates the BESTH device and the TW3 rig. Both devices allow testing of the Beryllium-coated First Wall mock-ups up to the required heat fluxes of 0.5–0.8 MW/m2, with the total duration of tens of thousands of cycles simulating the ITER pulse operation. The main difference between BESTH and the TW3 rig comes from the presence, or absence, of a radiation field.: BESTH is a dedicated water loop, completely out-of-pile, which uses a graphite panel for generation of the required heat flux. The TW3 rig is a similar but smaller in-pile device which is inserted inside the LVR-15 reactors core and thus combines the heat fatigue cycling with the reactor radiation exposure. Centrum výzkumu Řež also actively participates in the development of the Test Blanket Module (TBM), which will provide breeding and separation of tritium in ITER. The outcomes results of the TBM programme are expected to be the crucial ones coming from the ITER R&D.

The TBM will verify the feasibility of tritium breeding, employing several different technical designs. Centrum výzkumu Řež is involved in the R&D programme of HCLL TBM (Helium Cooled Lithium Lead Test Blanket Module), which circulates lead-lithium (PbLi) eutectic through the TBM breeding zone where PbLi melt is irradiated by fusion neutrons, and in this way, tritium and helium are produced. This process depletes lithium from PbLi melt and enriches it with tritium. The melt then flows through the purification section, called “cold trap”, where impurities, such as corrosion products, are removed. After that, tritium is extracted from the PbLi melt with upstream purge gas (He or Ar)