Blanket Shield Research Articles

Investigation of material homogenisation on the nuclear heating estimates of the electron cyclotron-heating upper launcher blanket shield module

The electron cyclotron-heating upper launcher (ECHUL) will be installed in four upper ports of the ITER tokamak thermonuclear fusion reactor. Each ECHUL is able to deposit 8 MW power into the plasma for plasma mode stabilization via microwave beam lines. These beam lines and other components needed for the guidance of the microwaves are mounted into the upper port plug. In order to protect these components and to mitigate radiation leaving the reactor, several shield blocks are also included in the port plugs design. The ECHUL components which are closest to the fusion plasma are heated by the gamma and neutron radiation and therefore will need to be actively cooled. This paper reports on the influence of material homogenization by comparing results obtained in the neutronic analyses performed for the cooling system of the ECHUL blanket shield module (BSM). A small reduction in shielding performance of the BSM was found in the case of a heterogeneous description of the BSM compared to the standard approach of homogenizing the materials of the BSM by removing pipes and cooling circuits for the neutronic analyses.

Electro-Magnetic Analysis of the ITER Upper Visible Infrared Wide Angle Viewing System

The Upper Visible Infrared Wide Angle Viewing System (UWAVS) is a diagnostic used in five upper ports of ITER. Each UWAVS provides visible and infrared views of various sections of the divertor. A single UWAVS is designed in three main sections: in-vessel, interspace and port cell assemblies. Each assembly utilizes multiple steering and relay mirrors to direct the in-vessel light out of the tokamak to the port cell camera sensors.For the in-vessel components, the transient electro-magnetic (EM) environment resulting from the ITER magnet operation and plasma events induces design driving Lorentz forces. As such, all in-vessel systems require detailed electro-magnetic finite element analysis (FEA) to derive the resulting time dependent Lorentz loads.ANSYS Maxwell software was used to perform transient electro-magnetic simulations of the UWAVS in ITER upper port 14. A 20 degree sector, cyclic symmetric model was employed and included, inner and outer vacuum vessel, blanket shield modules, diagnostic fist wall (DFW) and shield module (DSM), upper port plug structure, DSM shield blocks, and a detailed model of the UWAVS in-vessel assembly.The resulting data includes eddy current density and vector plots along with force and moment summation for various UWAVS components. Front end optical components are specifically reported as these components have significant EM loads.

Modeling of ITER TF cooling system through 2D thermal analyses and enthalpy balance

The winding pack of the ITER Toroidal Field (TF) coils is composed of 134 turns of Nb3Sn Cable in Conduit Conductor (CICCs) wound in 7 double pancakes and cooled by supercritical helium (He) at cryogenic temperature. The cooling of the Stainless Steel (SS) case supporting the winding pack is guaranteed by He circulation in 74 parallel channels.A 2D approach to compute the temperature distribution in the ITER TF winding pack is here proposed. The TF is divided in 32 poloidal segments, for each segment the corresponding 2D model is built and a thermal analysis is performed applying the corresponding nuclear heating computed with MCNP code considering the latest design updates, such as thickness increase of the blanket shield module. The Heat Transfer coefficient (HTC) of the He flowing in the CICC and in the cooling channels of the SS case is computed with Dittus Boelter correlation at the nominal inlet pressure of 6bar. The He is assumed to enter the coil at 4.5K in the lower terminal junction, and then the bulk temperature in all the CICCs in each of the 32 segments is calculated by means of enthalpy balance between segments, considering the actual direction of He circulation, i.e. clockwise or counter-clockwise in neighboring pancakes. The He properties needed to compute the HTC, such as viscosity, specific heat and thermal conductivity, are also varied using the same strategy. With these assumptions, He temperatures close to 5.7K are computed, due to the high values of nuclear heating (which is estimated as high as 21.58kW in the 18TF). In the paper, the methodology is presented and the results are discussed in detail. Further parametric analyses are also presented to show the impact of the inlet temperature and of the nuclear heating on the temperature distribution.

Progress on the design development and prototype manufacturing of the ITER In-vessel coils

Abstracrt ITER is incorporating two types of In-Vessel Coils (IVCs): ELM Coils to mitigate Edge Localized Modes and VS Coils to provide a reliable Vertical Stabilization of the plasma. Strong coupling with the plasma is required in order that the ELM and VS Coils can meet their performance requirements. Accordingly, the IVCs are mounted on the Vacuum Vessel (VV) inner wall, in close proximity to the plasma, just behind the Blanket Shield Modules (BSM). Due to high radiation environment, mineral insulated copper conductors enclosed in a stainless steel jacket have been selected. The reference design and prototype work provided a good basis for the development of radiation resistant conductor capable of operating within the harsh conditions in ITER vacuum chamber. However, this effort identified shortcomings in achieving satisfactory manufacturing solution, and most significantly, difficulties in brazing the brackets onto the ELM coil conductor. Since this process has not proven successful, alternative designs are under development and prototyping. Prototype manufacturing on the alternative designs has been completed at ICAS, Italy and ASIPP, China. The aim was to eliminate the need for internal coil joints, to prove the principle of longer conductor length manufacturing, and to perform bending and welding trials on two different conductor cross-sections: circular and square. The procurement of the IVCs and their conductors will be done via direct call for tender from the ITER Organization. This paper will give an overview of the alternative design and prototype manufacturing of the ITER In-Vessel coils.

Irradiation tests of radiation hard materials for ITER blanket remote handling system

The ITER Blanket Remote Handling System (BRHS) will handle the blanket first walls (FWs) and shield blocks (SBs), which can weigh up to 4.5ton and be larger than 1.5m, stably and with a high degree of positioning accuracy. When the ITER has stopped plasma operations for maintenance, the BRHS will be installed in the vacuum vessel, whose components are radioactive, to remove and install the FWs and SBs. Therefore, the BRHS will be operated in a high radiation environment (up to 250Gy/h) having an estimated total dose of 5MGy during maintenance (maximum of two years). The radiation hardness requirement for ITER BRHS components is 1MGy total dose. As components may degrade by gamma irradiation, some equipment is expected to malfunction which causes delays in the in-vessel maintenance schedule. Therefore, failure mode and effects analyses (FMEA) were performed on the BRHS components and FMEA results suggest trouble with the power supply and signals due to degradation of the insulation of electrical components, and malfunctioning motors due to degradation of the lubricants of mechanical components. In this study, material property tests for O-rings and coating materials were performed to verify radiation hardness after the irradiation with gamma rays up to 5MGy.

Neutronics analysis of the in-vessel components of the ITER plasma-position reflectometry system on the low-field side

The ITER Plasma Position Reflectometry (PPR) system will be used to estimate the distance between the position of the magnetic separatrix and the first-wall at four pre-defined locations. For gap 4, the antennas are to be installed in-vessel between two blanket shield modules. The antennas and adjacent waveguides are directly exposed to the plasma through cut-outs in the blanket modules and are therefore subjected to high radiation doses from neutrons and photons, which may cause irradiation-induced changes in the material properties and compromise the integrity of these components. Here, we report on the neutronics analysis of the PPR system of gap 4 performed with the Monte Carlo simulation code MCNP6, with the objective of estimating the thermal loads in the system. The nuclear loads obtained in the neutronics analysis will be used as input in the thermal analysis of the in-vessel components of the ITER PPR system.

Development of the Advanced Neutronic Analysis Model for the K-DEMO With MCNP Code

Neutronic assessment is the important foundation of Korean fusion demonstration reactor (K-DEMO) study. Previously, preliminary neutronic calculations using a simplified model have been performed [1] , and optimized structure of blanket module and shield material has been derived. Based on the previous calculation results, the advanced K-DEMO neutronic analysis model has been developed to perform more accurate and reliable neutronic assessments with the detailed configuration of vacuum vessel (VV), toroidal field (TF) coil, blanket, and shield. The advanced K-DEMO neutronic analysis model was imported into the integrated interface program by the help of Monte Carlo automatic modeling system to produce input files for the Monte Carlo neutron photon transport code analyses. The neutronic calculation on neutron wall loading (NWL) distribution, tritium breeding ratio (TBR), and nuclear heating in tokamak components was conducted for the newly developed advanced K-DEMO neutronic analysis model. In the NWL calculation, the peak was indicated at the inboard (IB) and the outboard (OB) blanket first wall with the value of 3.81 and 3.93 MW/ $\text{m}^{2}$ . The different types of models by changing gaps between the blanket modules and the thickness of separated borated steel shield blocks are calculated to check global TBR and the shield performance. The global TBR of 1.017 was achieved with the 49.7- and 76.2-cm-thick IB and OB blanket modules. The model with 1-cm-thick assembly gap between blanket modules, 22-cm-thick IB shield, 50-cm-thick OB shield, and 34-cm-thick divertor shield are considered as a reasonable shield model with the total nuclear heating in 16 TF coils and VV of 9 and 32 kW.

Neutronics Analysis of Water-Cooled Ceramic Breeder Blanket for CFETR**supported by the National Magnetic Confinement Fusion Science Program of China (Nos. 2013GB108004, 2014GB122000, and 2014GB119000), and National Natural Science Foundation of China (No. 11175207)

In order to investigate the nuclear response to the water-cooled ceramic breeder blanket models for CFETR, a detailed 3D neutronics model with 22.5° torus sector was developed based on the integrated geometry of CFETR, including heterogeneous WCCB blanket models, shield, divertor, vacuum vessel, toroidal and poloidal magnets, and ports. Using the Monte Carlo N-Particle Transport Code MCNP5 and IAEA Fusion Evaluated Nuclear Data Library FENDL2.1, the neutronics analyses were performed. The neutron wall loading, tritium breeding ratio, the nuclear heating, neutron-induced atomic displacement damage, and gas production were determined. The results indicate that the global TBR of no less than 1.2 will be a big challenge for the water-cooled ceramic breeder blanket for CFETR.

Acceleration of calculation of nuclear heating distributions in ITER toroidal field coils using hybrid Monte Carlo/deterministic techniques

Because the superconductivity of the ITER toroidal field coils (TFC) must be protected against local overheating, detailed spatial distribution of the TFC nuclear heating is needed to assess the acceptability of the designs of the blanket, vacuum vessel (VV), and VV thermal shield. Accurate Monte Carlo calculations of the distributions of the TFC nuclear heating are challenged by the small volumes of the tally segmentations and by the thick layers of shielding provided by the blanket and VV. To speed up the MCNP calculation of the nuclear heating distribution in different segments of the coil casing, ground insulation, and winding packs of the ITER TFC, the ITER Organization (IO) used the MCNP weight window generator (WWG). The maximum relative uncertainty of the tallies in this calculation was 82.7%. In this work, this MCNP calculation was repeated using variance reduction parameters generated by the Oak Ridge National Laboratory AutomateD VAriaNce reducTion Generator (ADVANTG) code and both MCNP calculations were compared in terms of computational efficiency and reliability. Even though the ADVANTG MCNP calculation used less than one-sixth of the computational resources of the IO calculation, the relative uncertainties of all the tallies in the ADVANTG MCNP calculation were less than 6.1%. The nuclear heating results of the two calculations were significantly different by factors between 1.5 and 2.3 in some of the segments of the furthest winding pack turn from the plasma neutron source. Even though the nuclear heating in this turn may not affect the ITER design because it is much smaller than the nuclear heating in the turns that are closer to the plasma source, it is our recommendation to adopt the utilization of ADVANTG in similar calculations to increase the reliability of MCNP calculations of detailed distributions of nuclear responses.

Electromagnetic analysis of the Korean helium cooled ceramic reflector test blanket module set

Korean helium cooled ceramic reflector (HCCR) test blanket module set (TBM-set) will be installed at equatorial port #18 of Vacuum Vessel in ITER in order to test the breeding blanket performance for forthcoming fusion power plant. Since ITER tokamak has a set of electromagnetic coils (Central Solenoid, Poloidal Field and Toroidal Field coil set) around Vacuum Vessel, the HCCR TBM-set, the TBM and associated shield, is greatly influenced by magnetic field generated by these coils. In the case of fast transient electromagnetic events such as major disruption, vertical displacement event or magnet fast discharge, magnetic field and induced eddy current results in huge electromagnetic load, known as Lorentz load, on the HCCR TBM-set. In addition, the TBM-set experiences electromagnetic load due to magnetization of the structural material not only during the fast transient events but also during normal operation since the HCCR TBM adopts Reduced Activation Ferritic Martensitic (RAFM) steel as a structural material. This is known as Maxwell load which includes Lorentz load as well as load due to magnetization of structure material. This paper presents electromagnetic analysis results for the HCCR TBM-set. For analysis, a 20° sector finite model was constructed considering ITER configuration such as Vacuum Vessel, ITER shield blankets, Central Solenoid, Poloidal Field, Toroidal Field coil set as well as the HCCR TBM-set. Three major disruptions (operational event, likely event and highly unlikely event) were selected for analysis based on the load specifications. ANSYS-EMAG was used as a calculation tool. The results of EM analysis will be used as input data for the structural analysis.

Progress in fusion technology at SWIP

The fusion research activities at Southwestern Institute of Physics (SWIP) include the HL-2A & HL-2M tokamak programs, fusion reactor design and materials, along with key fusion technologies including R&D on ITER procurement packages. This paper presents the progress of fusion technology at SWIP, including the ITER first wall and blanket, Chinese helium cooled ceramic breeder test blanket module (HCCB–TBM) for ITER, gas injection system and gas discharge cleaning system, as well as the recent activities on reactor materials and R&D related to advanced divertor. The final design for ITER first wall and blanket shielding blocks allocated to SWIP have been completed, and were validated by recent tests. Major manufacturing technologies, such as forging, deep drilling, explosion bonding and deep laser welding, have been successfully demonstrated. Furthermore, the conceptual design of CN–HCCB–TBM has been completed, the related materials’ preparation, mock-up manufacturing and tests have been implemented. The tungsten divertor has been studied with various bonding and coating technologies. Meanwhile, highlights of functional material for TBM, oxides and carbides dispersion strengthened (ODS, CDS) reduced activation ferritic/martensitic (RAFM) steel, vanadium and tungsten alloys are also presented.

Investigation of effect of post weld heat treatment conditions on residual stress for ITER blanket shield blocks

The blanket shield block (SB) shall be required the tight tolerance because SB interfaces with many components, such as flexible support keypads, First Wall (FW) support contact surfaces, FW central bolt, electrical strap contact surfaces and attachment inserts for both FW and Vacuum Vessel (VV). In order to fulfil the tight tolerance requirement, stress relieving shall be performed for dimensional stability after cover welding operation.In this paper, effect of Post Weld Heat Treatment (PWHT) conditions, temperature and holding time, was investigated on the residual stress and hardness. The 316L Stainless Steel (SS) was prepared and welded by manual TIG welding by using filler material with 2.4mm of diameter. Welded 316L SS plate was machined to prepare the specimen for PWHT. PWHT was implemented at 250, 300, 400°C for 2 and 3h (400°C only) and residual stress after relaxation were determined. The evaluation of residual stress and hardness for each specimen was carried out by instrumented indentation technique. The residual stress and hardness were decreased with increasing the heat treatment temperature and holding time.

Comparative evaluation of structural integrity for ITER blanket shield block based on SDC-IC and ASME code

The ITER blanket Shield Block is a bulk structure to absorb radiation and to provide thermal shielding to vacuum vessel and external vessel components, therefore the most significant load for Shield Block is the thermal load. In the previous study, the thermo-mechanical analysis has been performed under the inductive operation as representative loading condition. And the fatigue evaluations were conducted to assure structural integrity for Shield Block according to Structural Design Criteria for In-vessel Components (SDC-IC) which provided by ITER Organization (IO) based on the code of RCC-MR.Generally, ASME code (especially, B&PV Sec. III) is widely applied for design of nuclear components, and is usually well known as more conservative than other specific codes. For the view point of the fatigue assessment, ASME code is very conservative compared with SDC-IC in terms of the reflected Ke factor, design fatigue curve and other factors. Therefore, an accurate fatigue assessment comparison is needed to measure of conservatism. The purpose of this study is to provide the fatigue usage comparison resulting from the specified operating conditions shall be evaluated for Shield Block based on both SDC-IC and ASME code, and to discuss the conservatism of the results.

Verification of dimensional stability on ITER blanket shield block after stress relieving

The tight tolerance requirement is one of key issue to manufacture the ITER blanket shield blocks (SBs) which have many interfaces with the First Wall (FW) and Vacuum Vessel (VV). Manufactured SB shall be satisfied with general tolerances (Class “C” of ISO 2768-1 and “L” of ISO 2768-2) and specific tolerance in 2D general assembly drawings. In order to fulfill the tight tolerance requirements in the final stage of SB, stress relieving after welding operations in the manufacturing process shall be performed. Hot helium leak test, Post Welding Heat Treatment (PWHT) and three-dimensional inspection before and after heat treatment were implemented by using the Full Scale Prototype (FSP) of SB in the framework of domestic R&D activities. The hot He leak test was performed at 250°C for 30min, and the result was satisfied the requirements. PWHT was carried out at 400°C for 24h by brazing furnace with test chamber. The deformation value before and after was measured by contact type coordinate measuring machine. The objective of this study is to verify dimensional stability of SB after stress relieving. The results will support to determine the machining allowance prior to welding process.

Experimental and numerical study of the pressure drop for ITER blanket shield block

The blanket shield block (SB) is located inside the ITER vacuum chamber, and the main function is to provide the thermal and nuclear shielding to the vacuum vessel and external components. The SB is foreseen to undergo a significant heat load which is a body load throughout the whole thickness of the SB under normal operation conditions. Therefore, the cooling configuration in SB should be designed very carefully based on the various experiences.The pressure drop in the cooling design is one of the most important factors to balance a water distribution of overall blanket cooling system. In order to verify the pressure drop characteristic and validate the design methodology of SB, experiment and numerical analysis are performed and compared their results. These results would be a benchmarking of the numerical results with experimental results to assess the gap between calculations and experiments.

Shutdown dose rate analysis of European test blanket modules shields in ITER Equatorial Port #16

‘Fusion for Energy’ (F4E) is designing, developing, and implementing the European Helium-Cooled Lead-Lithium (HCLL) and Helium-Cooled Pebble-Bed (HCPB) Test Blanket Systems (TBSs) for ITER (Nuclear Facility INB-174). An essential element of the Conceptual Design Review (CDR) of these TBSs is the demonstration of capability of Test Blanket Modules (TBM) and their shields to fulfil their function and comply with the design requirements. One of the TBM shields highly relevant design aspects is the project target for shutdown dose rates (SDDR) in the interspace. We investigated two functions of the TBMs and TBM shields—the neutron flux attenuation along the shields, and the reduction of the activation of the components contributing to SDDR. It is shown that TBMs and TBM shields reduce significantly the neutron flux in the port plug (PP). In terms of neutron flux attenuation, the TBM shield provides sufficient neutron flux reduction, being responsible for 5×106n/cm2s at port interspace, while the EPP gaps and BSM gaps are responsible for 5×107n/cm2s each. When considering closed upper, lower and lateral neighbour equatorial ports (thus, excluding the cross-talk between ports), a SDDR of 121μSv/h averaged near the port closure flange was obtained, out of which, only 4μSv/h are due to the activation of TBMs and TBM shields. Maximum SDDR in the range of 300–350μSv/h were observed in the interspace. Thus, although the SDDR are found above the project target the good performance of the TBM shields design was demonstrated. Two main reasons for high SDDR were identified in this work—the neutron streaming through the equatorial PP gaps and the neutron penetration through the blanket shield modules and vacuum vessel.

A Possible Shielding Blanket Module for CFETR Reactor

The basic function of shielding blanket (SB) is to provide the main thermal and nuclear shielding to the vessel and external machine components. CFETR SB modules will be operated in steady mode with continuous neutron radiation and heat source. The SB module is currently operated in pulse mode, which indicates the SB module for CFETR should be designed to fit with the new operation conditions. In this paper, the SB module structural materials are investigated, and the reduced activation ferritic/martensitic steel is selected. The cooling structure of SB module is established and a new cooling system is put forward to ensure the temperature of SB module in an allowable range. The hydraulic and the thermo-mechanical performance are analyzed by finite element method. The analysis results indicate that the new cooling structure is well in the heat removal under the steady heat of CFETR.

Preliminary Neutronics Analysis of the ITER Toroidal Interferometer and Polarimeter Diagnostic Corner Cube Retroreflectors

ITER is an international project under construction in France that will demonstrate nuclear fusion at a power plant-relevant scale. The Toroidal Interferometer and Polarimeter (TIP) Diagnostic will be used to measure the plasma electron line density along 5 laser-beam chords. This line-averaged density measurement will be input to the ITER feedback-control system. The TIP is considered the primary diagnostic for these measurements, which are needed for basic ITER machine control. Therefore, system reliability & accuracy is a critical element in TIP’s design.There are two major challenges to the reliability of the TIP system. First is the survivability and performance of in-vessel optics and second is maintaining optical alignment over long optical paths and large vessel movements. Both of these issues greatly depend on minimizing the overall distortion due to neutron & gamma heating of the Corner Cube Retroreflectors (CCRs). These are small optical mirrors embedded in five first wall locations around the vacuum vessel, corresponding to certain plasma tangency radii. During the development of the design and location of these CCRs, several iterations of neutronics analyses were performed to determine and minimize the total distortion due to nuclear heating of the CCRs. The CCR corresponding to TIP Channel 2 was chosen for analysis as a good middle-road case, being an average distance from the plasma (of the five channels) and having moderate neutron shielding from its blanket shield housing. Results show that Channel 2 meets the requirements of the TIP Diagnostic, but barely. These results suggest other CCRs might be at risk of exceeding thermal deformation due to nuclear heating.

Electromagnetic analysis on Korean Helium Cooled Ceramic Reflector (HCCR) TBM during plasma major disruption

Korean Helium Cooled Ceramic Reflector (HCCR) Test Blanket Module (TBM) will be installed at the #18 equatorial port of the Vaccum Vessel in order to test the feasibility of the breeding blanket performance for forthcoming fusion power plant in the ITER TBM Program. Since ITER tokamak contains Vaccum Vessel and set of electromagnetic coils, the TBM as well as other components is greatly influenced by magnetic field generated by these coils. By the electromagnetic (EM) fast transient events such as major disruption (MD), vertical displacement event (VDE) or magnet fast discharge (MFD) occurred in tokamak system, the eddy current can be induced eventually in the conducting components. As a result, the magnetic field and induced eddy current produce extremely huge EM load (force and moment) on the TBM. Therefore, EM load calculation is one of the most important analyses for optimized design of TBM. In this study, a 20-degree sector model for tokamak system including central solenoid (CS) coil, poloidal field (PF) coil, toroidal field (TF) coil, vaccum vessel, shield blankets and TBM set (TBM, TBM key, TBM shield, TBM frame) is prepared for analysis by ANSYS-EMAG tool. Concerning the installation location of the TBM, a major disruption scenario is particularly applied for fast transient analysis. The final goal of this study is to evaluate the EM load on HCCR TBM during plasma major disruption.

Current status of final design and R&D for ITER blanket shield blocks in Korea

The main function of the ITER blanket shield block (SB) is to provide nuclear shielding and support the first wall (FW) panel. It needs to accommodate all the components located on the vacuum vessel (in particular the in-vessel coils, blanket manifolds and the diagnostics). The conceptual, preliminary and final design reviews have been completed in the framework of the Blanket Integrated Product Team. The Korean Domestic Agency has successfully completed not only the final design activities, including thermo-hydraulic and thermo-mechanical analyses for SBs #2, #6, #8 and #16, but also the SB full scale prototype (FSP) pre-qualification program prior to issuing of the procurement agreement. SBs #2 and #6 are located at the in-board region of the tokamak. The pressure drop was less than 0.3 MPa and fully satisfied the design criteria. The thermo-mechanical stresses were also allowable even though the peak stresses occurred at nearby radial slit end holes, and their fatigue lives were evaluated over many more than 30 000 cycles. SB #8 is one of the most difficult modules to design, since this module will endure severe thermal loading not only from nuclear heating but also from plasma heat flux at uncovered regions by the FW. In order to resolve this design issue, the neutral beam shine-through module concept was applied to the FW uncovered region and it has been successfully verified as a possible design solution. SB #16 is located at the out-board central region of the tokamak. This module is under much higher nuclear loading than other modules and is covered by an enhanced heat flux FW panel. In the early design stage, many cooling headers on the front region were inserted to mitigate peak stresses near the access hole and radial slit end hole. However, the cooling headers on the front region needed to be removed in order to reduce the risk from cover welding during manufacturing. A few cooling headers now remain after efforts through several iterations to remove them and to optimize the cooling channels. The SB #8 FSP was manufactured and tested in accordance with the pre-qualification program based on the preliminary design, and related R&D activities were implemented to resolve the fabrication issues. This paper provides the current status of the final design and relevant R&D activities of the blanket SB.