First Wall Mock-ups Research Articles

Post-Test Numerical Analysis of a Helium-Cooled Breeding Blanket First Wall under LOFA Conditions with the MELCOR Fusion Code

The validation of numerical tools employed in the analysis of incidental transients in a fusion reactor is a topic of main concern. KIT is taking part in this task providing both experimental data and by performing numerical analysis in support of the main codes used for the safety analyses of the Helium Cooled Pebble Bed (HCPB) blanket concept. In recent years, an experimental campaign has been performed in the KIT-HELOKA facility to investigate the behavior of a First Wall Mock-Up (FWMU) under Loss Of Flow Accident (LOFA) conditions. The aim of the experimental campaign was twofold: to check the expected DEMO thermal-hydraulics conditions during normal and off-normal conditions and to provide robust data for code validation. The present work is part of these validation efforts, and it deals with the analysis of the LOFA experimental campaign with the system code MELCOR 1.8.6 for fusion. A best-estimate methodology has been used in support of this analysis to ease the distinction between user’s assumptions and code limitations. The numerical analyses are here described together with their goals, achievements, and lesson learnt.

Comparison between measurement and calculations for a 14 MeV neutron water activation experiment

The nuclear heat loads due to gamma rays emitted from the decay of 16N and delayed neutrons from17N, generated by the activation of water in cooling circuits, are critical for ITER design. The assessment of nuclear heating from activated water is complex; it requires temporal and spatial dependent transport and activation calculations taking into account variation of irradiation, water flow conditions and cooling circuits’ parameters. A water activation experiment has been recently conducted at the14 MeV Frascati Neutron Generator (FNG) in order to validate the methodology for water activation assessment used for ITER and to reduce the safety factors applied to the calculation results, which have a large impact on the schedule, commissioning and licensing. Water circulating inside an ITER First Wall (FW) mock-up was irradiated with 14 MeV neutrons and then measured using a large CsI scintillator detector. The system consists of a closed water loop where the cooling water, transiting through an ITER FW mock-up, is irradiated by FNG. The induced 16N activity via 14 MeV neutrons interactions with 16O via the 16O(n,p)16N reaction is measured in a dedicated counting station via an expansion volume. The water then passes to a much larger holding delay tank, and after several 16N half-lives decay time, it is then recirculated and exposed again to neutrons in the ITER First Wall (FW) mock-up. The measured 16N activity is obtained measuring the emitted characteristic 6.13 and 7.12 MeV gamma-rays. Calculations were performed in an accurate model of the FW mock-up using the MCNP Monte Carlo code and FENDL-3.1 nuclear data library to obtain the predicted flux impinging on the water. The EASY-2007 inventory code was used to predict the 16N activity. In this work, a comparison between measurements and calculations is reported together with associated uncertainty analysis.

Thermo-mechanical stress analysis of the water cooled DEMO First Wall mock-up components

The European DEMOnstration (DEMO) reactor precedes the development of the fully commercial fusion power plant. One of the key components of the DEMO is the first wall (FW). The FW is acted upon by extreme particle and radiation loads, which results in high heat fluxes varying across the reactor vessel. This paper attempts to evaluate two DEMO FW water cooling concepts from the stress point of view. The heat sink material chosen for the FW is Eurofer97 with thermal conductivity of ∼28 W/mK at 20 °C; the cooling conditions opted are those of pressurized water reactor (PWR), i.e. 285 °C and 15.5 MPa.In the present work it is attempted to perform the stress analysis of the two different heat sink channel configurations, these are circular pipes with countercurrent flow and channel with hypervapotron configuration. The thermo-hydraulic optimization of these components has been performed in one of the author’s previous work. The stress in the components is caused by the pressure of the fluid and increase in the temperature of the material. To accomplish the work, commercial finite element method software ANSYS is used. In a case of unsatisfactory results, further recommendations are given. Simulations are performed with P91, which is material being commonly used in power engineering and will be also used for manufacturing of the discussed configuration of DEMO FW mock-ups which will be afterwards tested in high heat flux facility HELCZA.

Fabrication feasibility studies for first wall of Indian LLCB TBM

The conceptual design of the Indian Lead Lithium Ceramic Breeder (LLCB) Test Blanket Module (TBM) First Wall (FW) presents a single U-shaped component made of IN-RAFM steel having 65 square channels of 20 mm × 20 mm running along the toroidal and radial direction. There are 13 circuits; each circuit consisting of 5 channels with inlet on the 1st channel of one side and the outlet on the 5th channel of other side. The conventional machining route with innovations in the existing technique was explored to develop the method to make channel mock-ups for FW application. This developed technique includes initial pre-machining and milling of higher thickness plates, making square channels by gun drilling followed by wire-cut EDM and final milling on the external surface and then hot bending into U-shape. Modified 9Cr-1Mo steel was the structural material used for this mock-up which has two flow circuits (each circuit of 5 channels). The other dimensions of this mock-up were chosen in such a way that functional requirements of full-sized FW would be fulfilled. During manufacturing, this mock-up was subjected to a series of inspection and non-destructive tests such as helium leak test and hydro-test at 150 bar pressure as a part of in-stage and final inspections. This paper describes the main activities carried out for the fabrication and factory acceptance tests of this U-shaped 10-channels TBM FW mock-up.

Characterization of the cooling channels of First Wall Mock-up dedicated to the HCPB TBM qualification

The First Wall (FW) of the EU Helium Cooled Pebble Bed (HCPB) Test Blanket Module (TBM) faces the fusion plasma and experiences high heat fluxes; therefore its cooling channels design is a key R&D task for qualifying the HCPB TBM for the fusion reactors ITER and DEMO. Within the manufacturing and qualification activities performed in KIT for the HCPB TBM, a First Wall Mock-up (FWM) was designed and manufactured. The objective of this study is to characterize the hydrodynamic behaviour of the FWM cooling channels by investigating the coolant pressure drop for each channel. The FWM has a shape of rectangular prism (710mm×405mm×45mm) with 10 U-shaped cooling channels which have a square cross section (15mm×15mm) with rounded corners of 4mm radius. The FWM was integrated into a gas loop that has the relevant instruments and piping system for measuring and controlling the air (6bar abs.) flow parameters. This paper presents experimental measurements of the pressure drop across the FWM cooling channels. The present results support the qualification of the HCPB TBM mock-ups which will be tested in the Helium Loop Karlsruhe (HELOKA) facility with fusion-relevant heat flux and helium cooling.

Simulation and Experimental Study of Forming Technology on First Wall Mock-Up for Fusion Reactor Blanket

Blanket, as a plasma-facing component, is one of the most critical functional component for fusion reactor operation. It will withstand very high heat load and work in very complicated condition. The first wall (FW) is one of the most important components of the blanket. The conventional design of the FW for the fusion reactor is based on an actively cooled system under high surface heat flow from the nuclear heat to make sure the temperature and stress of the FW are lower than the design limits, at the same time the process technology of the FW must be taken into account. This paper describes the forming technology of FW mock-up for fusion reactor. The simulation of the FW forming technology has been performed, the main calculation steps for the mock-up are described. As well as the test of the forming technology of FW mock-up was carried out in order to verify the feasibility of the FW fabrication scheme.

First wall fabrication of 1/3 scale china dual functional lithium lead blanket

The dual functional lithium lead blanket is chosen as one of the candidate blankets for China fusion reactor, for its advantages of tritium breeding and good heat exchange performance. As one of the most important components of the blanket, the first wall (FW) is assembled with China low activation martensitic (CLAM) rectangular tubes and plates by hot isostatic pressing (HIP)–diffusion bonding (DB). In this work, the rectangular tube fabrication and FW assembly were carried out in order to verify the feasibility of the FW fabrication scheme. The mechanical property and dimensional accuracy of CLAM rectangular tubes were tested, the microstructure observation and non-destructive detection revealed the sound of the FW mock-up, and the reliability of the FW mock-ups is under evaluation.

Experimental and numerical assessment of normal heat flux first wall qualification mock-ups under ITER relevant conditions

The ITER first wall (FW) panel consists of beryllium in the form of tiles covering its surface, high strength copper alloy as the heat sink material and stainless steel as the structural material. Small-scale normal heat flux FW mock-ups, provided by Fusion for Energy, are tested in the electron beam facility JUDITH 2 at Forschungszentrum Jülich to determine the performance of this design under thermal fatigue. The mock-ups are loaded cyclically under a surface heat flux of 2 MW m−2 with ITER relevant water coolant conditions. In this study, three-dimensional finite element method thermo-mechanical analyses are performed with ANSYS to simulate the thermal fatigue behaviour of the mock-ups. The temperature results indicate that the beryllium surface temperature is below the maximum allowed temperature (600 °C) of beryllium to be tested. The thermal mechanical results indicate that copper rupture and debonding between Be and copper are the drivers of the failure of a mock-up. In addition, the experimental data, e.g. the surface temperature measured using an infrared camera and the bulk temperature measured using thermocouples, are reported. A comparative study between experimental and simulation results is performed.

Fabrication and Integrity Testing of a Korean ITER TBM FW Mock-Up in Preparation for High Heat Flux Test

Korea (KO) has developed and participated in the Test Blanket Module (TBM) program in the ITER, in which Ferritic Martensitic (FM) steel was used as the structural material for the TBM first wall (FW). To develop the fabrication method for the TBM FW and verify its integrity, a half-scale sub-module mock-up was fabricated and integrity test has been prepared; A dimension of it is a 444 mm height, 260 mm width, and 435 mm depth. A mock-up was assembled by HIPping of the previous fabricated components between the welded front and back plates, and then machining these plates to form the completed sub-module FW. To investigate the deformation of the cooling channels after the fabrication of the mock-up, neutron radiography was carried using neutron radiography facility. Pressure and He leak tests were successfully performed without any leak and failure. The flow rates in each channel were measured with the conventional ultrasonic sensor but it shows non-uniform flow distribution at each channel differently from the estimation by ANSYS-CFX. HHF test conditions were evaluated through the ANSYS-CFX analysis considering the above measured flow rates in each channel and it shows non-uniform temperature distribution of the FW mock-up. We will re-assemble or fabricate the manifold and perform the flow test before HHF test in the near future.

Two Different Medium-Scale First Wall Fabrication Experiments at KIT for ITER Helium-Cooled Pebble Bed Test Blanket Module

Subcomponent manufacturing and assembly concepts for the fabrication of the helium-cooled pebble bed test blanket module (TBM) for ITER have been developed over more than one decade at KIT, in particular the first wall (FW), which is a key element for the TBM fabrication. The design of this subcomponent foresees the manufacturing of a large U-bended plate of EUROFER with built-in channels for helium cooling. Manufacturing technologies developed at KIT are based on diffusion welding of two half-plates as the most promising option. This paper deals with the manufacturing of two medium-scale TBM FW mock ups according to two different industrial processes: a uni-axial diffusion welding process realized in a mechanic press at high temperature and a hot isostatic pressing process applied to a canned assembly at relatively low pressure.The qualification of the welds produced is described, and the results are compared to previous small- and medium-size scale experiments. The results of the recent FW fabrication mock ups are presented with regard to material data (e.g., ultimate strength, ductile-brittle transition temperature) and TBM-relevant parameters (e.g., deformation of cooling channels). The paper concludes with an overview of the strategy to evolve from 1/8th-scale experiments to TBM-relevant dimensions.

Thermal fatigue testing devices for fusion-related projects

Paper introduces thermal fatigue testing devices, developed and operated under fusion-related projects with focus on testing of ITER First Wall (FW) mock-ups. In frame of EFDA tasks, FW mock-up testing device was developed and put into operation in Centrum Výzkumu Řež s.r.o. (CV Rez) research centre, and one testing device was modified for thermal fatigue tests in Forsungszentrum Jülich (FZJ) research centre. The FW mock-ups were tested for several parties, where the key role was played by Fusion for Energy (F4E), the European Union's Joint Undertaking for ITER and the Development of Fusion Energy.

Thermo-hydraulic testing and integrity of ITER Test Blanket Module (TBM) First Wall mock-up in JAEA

Fluid flow and heating tests on the full-scale Test Blanket Module (TBM) First Wall (FW) mock-up under the TBM-relevant operating condition were carried out in an ion beam irradiation system with a pressurized high temperature water loop in JAEA to verify the fabrication process and design of the TBM First Wall and demonstrate its heat removal capability. The TBM FW mock-up is made of reduced activation ferritic/martensitic (RAFM) steel, F82H, and its parts such cooling channels and plates were assembled with Hot Isostatic Pressing (HIP) method. In the fluid flow test, the cooling water at room temperature is supplied to 15 parallel flow paths in the mock-up. The flow distribution in each path inside the First Wall mock-up is compared with a numerical simulation. This result shows that no severe cross-sectional deformation of the entire flow path in the mock-up takes place during the fabrication process. In the heating test, the mock-up was exposed to repetitive hydrogen ion beam irradiation with the maximum heat flux of 0.5 MW/m 2. No indication of joint defect of the HIP joint like a hot spot was observed during 80 irradiation cycles.

Status of the EU R&D programme on the blanket-shield modules for ITER

A research and development (R&D) programme for the ITER blanket-shield modules has been implemented in Europe to provide input for the design and the manufacture of the full-scale production components. It involves in particular the fabrication and testing of mock-ups and full-scale prototypes of shield blocks and first wall (FW) panels. This paper summarises the main achievements obtained so far and presents the latest results of this R&D programme. In particular, it reports the status of the shield fabrication development programme with the manufacture of a full-scale shield prototype. It also reports the latest results of high heat flux and thermal fatigue tests of FW mock-ups. It describes the preparation of irradiation experiments of Be coated FW mock-ups. Finally, it presents the outline of a possible qualification programme that each contributing participant teams should pass prior to the procurement of the blanket-shield modules for ITER.

EU R&D on the ITER First Wall

This paper presents the last results obtained in the ITER First Wall (FW) development work programme performed in Europe on beryllium (Be), copper (Cu) alloys and 316L(N) stainless steel (SS) joining by Hot Isostatic Pressing (HIPping). It reports first the performances under high heat flux testing of CuCrZr/316L(N) SS mock-ups and compares them with those obtained earlier with CuAl25/316L(N) SS mock-ups. It reports then the results of the last developments done on Be/Cu alloy HIPped joints. In particular it presents the results of high heat flux testing of Be/CuCrZr FW mock-ups and compares them with those obtained earlier with Be/CuAl25 FW mock-ups. It gives the status of on-going thermal fatigue tests of Be/Cu alloy FW mock-ups and presents the result of a neutron irradiation experiment performed on Be/CuAl25 mock-ups. The results obtained so far with FW mock-ups made from CuCrZr alloys have exceeded those obtained earlier with CuAl25 alloys.

Manufacturing and testing of large-scale mock-ups of ITER plasma facing components in Russia

Various ITER plasma facing components (PFC) are described, such as first wall (FW), start-up limiter and divertor vertical target (VT). Manufacturing and testing of large-scale mock-ups is the final stage on the long way of structure optimization and selection of the most reliable and cheap manufacturing technologies. For the FW and limiter mock-ups fast brazing of Be tiles to CuCrZr heat-sink by brazing alloy STEMET-1101 was successfully applied. Testing of the FW large-scale mock-up up to 5000 cycles at 1 MW/m 2 did not reveal any damages of BeCu joints. For CuCrZr–SS bimetallic heat-sink/support structure two alternative joining technologies (HIP—for the limiter mock-up and CuCrZr casting onto solid SS—for the FW mock-up) were used. For divertor target mock-up the following technologies were used: pure Cu casting onto W tiles; fast brazing of WCu bimetallic tiles onto CuCrZr heat-sink; explosion bonding and CuCrZr casting as two alternatives for manufacturing of hermetic CuCrZr–SS structure of hypervapotron type; laser welding for attachment of heat-sink elements to support plate (SSSS welds). Results of thermocycling testing of all mock-ups at operational heat loads are presented. Recommendations for further improvement and characterization of above-mentioned technologies for ITER application are given.

Fabrication and Thermal Cycle Testing of a First Wall Mock-Up for ITER Baffle Blanket Module

The objective of this study is to develop joining technology of a plasma facing CFC armor and DSCu heat sink by means of Ag-free brazing and to evaluate its performances under repeated high heat flux. For fabrication of a baffle first wall (FW) composed of dispersion strengthened copper (DSCu) as heat sink and carbon fiber composite (CFC) as amor material, a two-step brazing has been proposed. CFC armor tiles were joined to an oxygen free copper (OFCu) plate in the first step and the OFCu plates joined, in turn, to the DSCu heat sink in the second step. After screening tests, brazing materials of Cu-Mn and Al were selected for the first and the second steps, respectively. A small scale baffle FW mock-up of 400 mm in length, 49 mm in width and 52 mm in thickness was successfully fabricated. By destructive test of a piece with the same materials bonded under the same condition as the mock-up, bondability at each brazing interface was confirmed. Thermal cycle tests were performed with the fabricated mock-up. Test conditions were selected based on thermo-mechanical analyses to simulate the temperature at the brazing interface under ITER operation condition and also higher temperatures with higher heat fluxes for acceleration of thermal cycle tests. The mock-up withstood more than 4600 cycles with heat fluxes of 5–10 MW/m2. From results of the thermal cycle tests, the integrity of CFC/DSCu interface was demonstrated.

Ratchetting of first-wall elements under combined thermo-mechanical and electro-mechanical stresses

During the operation of ITER a temperature difference in the range of some 100°C occurs between the heat-loaded outer surface and the wall of the cooling channels. Due to this inhomogeneous temperature field secondary stresses are produced. In addition high electromagnetic forces can occur caused by disruptions or vertical plasma movements. The resulting stresses may lie beyond the yield stress of the structural material. From the interaction of thermal and mechanical forces, a complex stress state is achieved which may lead to premature failure or to progressive plastic deformations (ratchetting). In the present paper the conditions for the occurrence of ratchetting are studied using simplified models of the first wall. For the experimental study of ratchetting phenomena an testing rig has been erected. In this rig, bending forces can be achieved on first-wall mockups with rise times of ∼ 30 ms combined with cyclic temperature gradients.