Shield Block Research Articles

Impact of nuclear heating distribution on the structural behavior of ITER first wall panel 11

The ITER blanket system safeguards the Tokamak's Vacuum Vessel and ex-vessel components from thermal and nuclear loads induced by the plasma. It comprises two parts: the First Wall panel facing the plasma and the Shield Block that provides the bulk of the shielding. The First Wall panels consist of Beryllium tiles on a CuCrZr layer, supported by stainless-steel structures which are cooled using a pressurized water circuit at 40 bar with an inlet temperature of 70 °C.Fusion for Energy has updated the thermal and mechanical analyses of ITER First Wall Panel 11 in preparation for the serial manufacturing. The panel will face heat flux up to 2 MW/m2 from plasma radiation, in addition to the volumetric nuclear heating, and the electromagnetic loads. A model with an enhanced detailed 3D geometry and heterogeneous materials indicates a 25% reduction of the deposited heating compared to the previous 2D model, from 625.4 kW (2D) to 455.4 kW (3D).The study embraces numerous thermal and mechanical analyses using Ansys software v19.2, incorporating 2D and 3D nuclear heating data. Responses are then compared and thus the possible impact of the two distributions on the thermal and mechanical behavior of the Fist Wall Panel 11 is assessed.

Key technologies of manufacturing and testing for the ITER blanket shield block

The SB is one of the main in-vessel components manufactured with SS316L(N)-IG to play a major role in neutron shielding, to support the FW panels and to supply the FW panel with cooling water. Cooling channels are located inside of the SB to remove the neutron heat deposition while keeping the structure temperature to acceptable levels. The water coolant is routed first through the plasma-facing first wall panel which is attached to the SB and then through the SB. A number of deep slits are introduced into the SB to disrupt eddy current loops and reduce the Electro-Magnetic (EM) loads on the support system and VV. Water headers are machined on the side of the module with 7–10 mm welded cover plates. A Procurement Arrangements (PAs) for the manufacturing and testing of the SBs have been signed with Korea (50 %) and China (50 %).As part of the PA in Korean Domestic Agency (KODA), Process Qualification (PQ) to ensure the manufacturing and testing ability of the SB was completed successfully through manufacture of a Full Scale Prototype (FSP) as an essential step before starting series production. This paper reports key technologies, such as machining, welding, NDE and hot helium leak testing, of manufacturing and testing for the ITER SB series production in Korea toward successful completion of the missions.

An Evaluation of the Symposion Scene Relief and Shield Block in Kayabaş (Korkuteli)

Antikçağ kültürlerinde mezar sahibinin kimliğini, özelliklerini, mesleğini veya üstün bir yönünü vurgulamak amacıyla mezar yapıları üzerinde, yazıt ve figürsel anlatımlara rastlanır. Kayabaş’ta görülen symposion sahnesi ile mezar sahibinin şanını gösteren anlatım; Mezopotamya kültürleri, Anadolu ve Antik Hellen dünyasında da sevilerek uygulanmıştır. Kayabaş Mahallesi’ndeki ana kaya kütlesi üzerinde yer alan genel figüratif anlatıların yerel bir örneği olan symposion kabartmasının sol tarafında yer alan, tepeleri oval formlu, tabanları düz, birbirine bitişik ve aynı zemini paylaşan iki kabartma, ana sahneyle bütünlük oluşturuyor görünmektedir. Kabartmaların güneyinde ve zeminde tahrip olmuş iki adet lahit teknesi, büyük ölçekli yuvarlak bir taş blok vardır. Yakın çevrede taş ocağı, 2 ostothek yuvası, 4 khamosorion, biri sağlam, biri kırık, diğeri tamamen ufalanmış 3 lahit daha bulunmaktadır. Söz konusu kalıntılar bu alanın bir nekropol olduğunu ortaya koymaktadır. Bunun yanı sıra kabartma ve alanda bağımsız olarak yer alan yuvarlak bir taş blok incelenmiş ve öneriler geliştirilmiştir.

Construction of hot helium leak test facility for ITER blanket shield block

The ITER Blanket Shield Block (SB) is one of the main in-vessel components that provides nuclear shielding and removes nuclear heating for the Vacuum Vessel (VV) and external vessel components. The SB, which contains internal cooling channels, is manufactured from 316L(N)-IG stainless steel, and the internal channels are designed to provide the required thermal-hydraulic performance while aiming for a water/steel ratio that results in good nuclear shielding with headers and welded cover plates. Therefore, the soundness of all SB products' welded joints must be confirmed to eliminate the risk of leakage during manufacturing and the Factory Acceptance Test (FAT). However, a test facility capable of testing large components at high temperatures has never been constructed anywhere. The ITER Korea Domestic Agency (KODA) has therefore successfully completed the construction and commissioning of two Hot Helium Leak Test (HHLT) facilities, and all SB series products are being entirely tested under RT and an elevated temperature of 250 °C with a maximum operational pressure of 5.0 MPa. This paper provides information on the design features of the two HHLT facilities and lessons learned from the commissioning test.

China's progress on hot helium leak test of ITER shield blocks

China Domestic Agency (DA) will provide ITER with 220 shield blocks (SB) which are important, large, and actively water-cooled components in ITER vacuum vessel, their leakage rate shall not exceed 2.69×10−10Pa·m3/s in hot helium leak tests (HHLT) at 250±20 °C. The HHLT work started with developing a HHLT facility and manufacturing a SB full-scale prototype (FSP). The result showed the FSP could satisfy ITER's requirement at 235 °C. H2 outgassing from stainless steel surfaces is the main factor that impacts both the background leakage rate and sensitivity of the testing system in hot testing cases. Hydrogen-removal measures were taken to improve the facility performance, and a new facility was built. A better facility performance came out at 250 °C rather than 235 °C, the background leakage rate decreased by one order of magnitude while the system sensitivity increased by two orders of magnitude. The background is at 10−10 Pa.m3/s and 10−9 Pa.m3/s level at room temperature (RT) and 250 °C, respectively, while the system sensitivity is at 10−12 Pa.m3/s and 10−11 Pa.m3/s level at the two temperatures. With the two facilities, series HHLTs have been carried out on 50 SBs covering two types, with all results complying with ITER requirements.

Зарубежный опыт и подходы к извлечению графита из уран-графитовых реакторов

The article explores the best practices for graphite removal from the stacks of uranium-graphite reactors (GLEEP, WAGR, BGRR, HTGR) and reactors with graphite reflectors (DIORIT, DR 1) operated abroad. Based on several case studies, it considers different approaches, in particular: graphite crushing; underwater stack demolition by reactor shaft flooding and upper biological shield dismantling; dismantlement of the upper biological shield and layer-bylayer graphite block removal. It discusses current trends, existing challenges and long-term plans for the practical implementation of advanced approaches under international projects.

The other side of tropical forest drought: do shallow water table regions of Amazonia act as large-scale hydrological refugia from drought?

Tropical forest function is of global significance to climate change responses, and critically determined by water availability patterns. Groundwater is tightly related to soil water through the water table depth (WT), but historically neglected in ecological studies. Shallow WT forests (WT < 5 m) are underrepresented in forest research networks and absent in eddy flux measurements, although they represent c. 50% of the Amazon and are expected to respond differently to global-change-related droughts. We review WT patterns and consequences for plants, emerging results, and advance a conceptual model integrating environment and trait distributions to predict climate change effects. Shallow WT forests have a distinct species composition, with more resource-acquisitive and hydrologically vulnerable trees, shorter canopies and lower biomass than deep WT forests. During 'normal' climatic years, shallow WT forests have higher mortality and lower productivity than deep WT forests, but during moderate droughts mortality is buffered and productivity increases. However, during severe drought, shallow WT forests may be more sensitive due to shallow roots and drought-intolerant traits. Our evidence supports the hypothesis of neglected shallow WT forests being resilient to moderate drought, challenging the prevailing view of widespread negative effects of climate change on Amazonian forests that ignores WT gradients, but predicts they could collapse under very strong droughts.

Implementation of the full-scale prototype pipe option design into non-standard first wall panel

Abstract The internal surface of the vacuum vessel of ITER is covered with 440 Blanket Modules (BM). Each BM consists of the First Wall (FW) panel and the Shield Block (SB). JSC NIKIET and JSC NIIEFA are responsible for manufacturing and procurement of the FW panels. Each FW panel includes plasma-facing components (FW fingers), FW beam, mechanical attachment system and electrical connection system with the SB. The FW fingers are mechanically attached to the FW beam and at the same time, hydraulic connection of the FW beam to the FW fingers is realized by welded connectors. In order to demonstrate manufacturability and to qualify major technological processes, the Full-Scale Prototype (FSP), was designed and approved by IO. The design of non-standard FW panels is to be based on the FSP design. Specialist from JSC NIKIET and JSC NIIEFA attempted to develop the design of non-standard FW panel 14 type NE, based on the FSP design. Considering the complex shape of the FW panel 14 type NE, cooling circuit modifications of the FW panel was necessary. JSC NIKIET specialists conducted thermal and hydraulic analysis of the FW fingers in order to present a design solution. This paper presents the design of the Enhanced Heat Flux (EHF) FW panel 14 type NE, based on the FSP design, and results of hydraulic and thermal analysis of the FW fingers.

Fabrication and characteristic of the 316L(N)-IG forged block and rolled plate for application to ITER blanket shield block

• The entire fabrication process of forged and rolled 316L(N)-IG production for ITER blanket Shield Block. • The false indication was observed upon UT of austenitic stainless weldment. • The microstructure of weldment was observed by EBSD analysis to find out the cause of false UT signal. • The result was found the sudden change of grain size in the fusion zone can affect to the UT signal. The austenitic stainless steel 316L(N)-IG (ITER Grade) is mainly employed for the structural material of the core machine considered for the in-vessel components of the ITER, which require good resistance to corrosion, weldability and mechanical properties at elevated temperature. The ITER blanket Shield Block (SB) is one of the in-vessel components and, it has to be made with 316L(N)-IG. The SB has a rather complex internal cooling channel to remove the nuclear heat from cooling water. For this structural reason, it is necessary to weld with dozens of small cover plates after manufacturing the cooling channel to make a closed circuit. This could cause a risk of possible water leakage due to technical difficulties and/or limitations of volumetric non-destructive inspection on small welded cover plates with geometrical constraints. Therefore, it is important to have verification of the Ultrasonic Testing (UT) for more accurate defect detection. The present study describes the entire fabrication process and characteristic of the 316 L(N)-IG, which are different fabrication types of the forged block and rolled plate. The study also provides the microstructure on the welded joint between the forged block and rolled plate using the Electron BackScattered Diffraction (EBSD) analysis. Based on this results, we dicussed that the effects of change of grain size on the UT signal in the weldment.

Study of decontamination and maintenance for vehicle manipulator of ITER blanket remote handling system

Abstract The ITER blanket remote handling system (BRHS) will be operated in the ITER vacuum vessel (VV) after plasma operation to exchange the blanket first walls (FWs) and shield blocks (SBs), which can weigh up to 4.5 t and be larger than 1.5 m. The most important operation for the BRHS is the two-year maintenance campaign in which all 440 FWs must be replaced. The in-vessel components will have been activated by neutrons produced by the D-T reaction, with the total dose estimated to be 5 MGy. Moreover, activated dust will have accumulated in the VV and will contaminate the BRHS. The BRHS must first be transported to the red zone of the hot cell facility where it will be decontaminated remotely. The BRHS will then be transported to the hands-on area to be decontaminated manually and undergo maintenance. The ITER Organization has defined the dose rate criteria in the hands-on area as 100 μSv/h, thus, decontamination and maintenance activities must be performed in compliance with these criteria. In this paper, the parts that will require replacement in the two-year maintenance campaign are estimated based on the results of past irradiation experiments studies; areas of the BRHS that could not be decontaminated remotely are specified; and the dose rate to workers standing near the BRHS are calculated by using Monte Carlo N–Particle Transport Code analysis to establish the BRHS maintenance plan.

Analyses of the shielding options for HCPB DEMO blanket

Abstract The enhancement of the HCPB DEMO blanket shielding capability is a challenging issue that assumes complex studies involving numerous neutronics analyses based on the Monte Carlo simulations with MCNP code. To this end very detailed, fully heterogeneous MCNP geometry models were developed for the latest designs of the HCPB and WCLL DEMO blankets to compare their shielding efficiency. This enables to perform rigorous calculations of the nuclear responses in the blanket structure and also in the vacuum vessel that entitles to assess the shielding capability of the blankets. The basic response, tritium breeding ratio (TBR), was computed for both DEMOs and appeared to be 1.20 and 1.10 for HCPB and WCLL DEMOs respectively. Several materials were proposed for the shielding purposes including metal hydrides, tungsten compounds, boron carbide and pressurized water. Two options for the shield arrangement in the HCPB blanket were verified: inside and outside the back supporting structure (BSS) of the blanket. The most effective options appeared to be the 18 cm thick shield block made of TiH2, ZrH1.6 and WB and placed outside the BSS that can provide the dpa accumulation in the inner wall of the vacuum vessel (VV) below the results obtained for the WCLL DEMO. Activation analyses performed for both concepts revealed the challenging problem with the activity of the VV exceeding low-level waste limit.

Crustal and uppermost mantle shear wave velocity structure beneath the Middle East from surface wave tomography

SUMMARY We have constructed a 3-D shear wave velocity (Vs) model for the crust and uppermost mantle beneath the Middle East using Rayleigh wave records obtained from ambient-noise cross-correlations and regional earthquakes. We combined one decade of data collected from 852 permanent and temporary broad-band stations in the region to calculate group-velocity dispersion curves. A compilation of &amp;gt;54 000 ray paths provides reliable group-velocity measurements for periods between 2 and 150 s. Path-averaged group velocities calculated at different periods were inverted for 2-D group-velocity maps. To overcome the problem of heterogeneous ray coverage, we used an adaptive grid parametrization for the group-velocity tomographic inversion. We then sample the period-dependent group-velocity field at each cell of a predefined grid to generate 1-D group-velocity dispersion curves, which are subsequently inverted for 1-D Vs models beneath each cell and combined to approximate the 3-D Vs structure of the area. The Vs model shows low velocities at shallow depths (5–10 km) beneath the Mesopotamian foredeep, South Caspian Basin, eastern Mediterranean and the Black Sea, in coincidence with deep sedimentary basins. Shallow high-velocity anomalies are observed in regions such as the Arabian Shield, Anatolian Plateau and Central Iran, which are dominated by widespread magmatic exposures. In the 10–20 km depth range, we find evidence for a band of high velocities (&amp;gt;4.0 km s–1) along the southern Red Sea and Arabian Shield, indicating the presence of upper mantle rocks. Our 3-D velocity model exhibits high velocities in the depth range of 30–50 km beneath western Arabia, eastern Mediterranean, Central Iranian Block, South Caspian Basin and the Black Sea, possibly indicating a relatively thin crust. In contrast, the Zagros mountain range, the Sanandaj-Sirjan metamorphic zone in western central Iran, the easternmost Anatolian plateau and Lesser Caucasus are characterized by low velocities at these depths. Some of these anomalies may be related to thick crustal roots that support the high topography of these regions. In the upper mantle depth range, high-velocity anomalies are obtained beneath the Arabian Platform, southern Zagros, Persian Gulf and the eastern Mediterranean, in contrast to low velocities beneath the Red Sea, Arabian Shield, Afar depression, eastern Turkey and Lut Block in eastern Iran. Our Vs model may be used as a new reference crustal model for the Middle East in a broad range of future studies.

Hot Helium Leak Test of ITER blanket shield block

To ensure successful operation of the International Thermonuclear Experimental Reactor (ITER), the blanket shield block modules are required to demonstrate the vacuum performance by Hot Helium Leak Test (HHLT), which is performed at the relevant ITER operational conditions and is one of the Factory Acceptance Tests (FAT) of the product. A specific facility is developed in China to perform these demanding tests, which has a 10 m3 vacuum chamber to accommodate the large size of shield blocks and achieved an overall system leak detect sensitivity of 7.19 × 10−11Pa. m3/s at 235 °C. A full-scale proto-type of ITER shield block was tested in this facility, which showed maximum helium leak rate is less than 1.74 × 10-10Pa. m3/s under 4.2 MPa pure helium gas pressure at 235 °C. The HHLT was performed with cyclic pressure and temperature process and the results were given in this paper. The behavior of the background leak rate during these processes were also monitored and discussed.

Image-guided total-body irradiation with a movable electronic portal imaging device for bone marrow transplant conditioning

IntroductionTo prevent radiation pneumonitis following total body irradiation (TBI) clinicians usually use lung shield blocks. The correct position of these shields relative to the patient's lungs is usually verified via mega-voltage imaging and computed radiographic (CR) films. In order to improve this time-consuming procedure, we developed in our department a dedicated, movable, real-time imaging system for image-guided TBI. Material & MethodsThe system consists of an electronic portal imaging device (EPID) mounted on a dedicated support whose motion along a rail can be controlled from the linac console outside the bunker room. Images are acquired online using a stand-alone console. To test the system efficacy we retrospectively analyzed data of lung blocks positioning from two groups of 10 patients imaged with EPID or CR-films, respectively. ResultsThe median number of portal images per fraction was 2 (range 1-5) and 1 (range 1-2) for the EPID and the CR-film system, respectively. The minimum time required for an EPID image acquisition, without interpretation and no need of patient position correction in the bunker, was 20seconds against 214seconds for the CR-film. Lung shielding positioning in the right-left and superior-inferior directions was improved using the EPID system (p<0.01). ConclusionsCompared to CR-films, our movable real-time imaging EPID system is a simple technical solution able to reduce the minimum imaging time for lung shielding by a factor of 10. With the increased possibility to acquire more images as compared to CR-film system the EPID system has the potential to improve patient alignment, as well as patient's comfort and overall setup time.

Optimization and adjustment of impact set-up for testing of insulated pads of ITER blanket module connectors and first wall

Insulated pads are used on ITER blanket module connectors and the first wall; their main insulating function is to break any current loop between the shield block and vacuum vessel and/or between the first wall panel and shield block. The design of the pads consists of a cylindrical or prismatic body manufactured from NiAl-bronze, a ceramic insulating coating which is applied on interfacing surfaces of the pad body. The pads work in ultra-high vacuum conditions and at elevated temperature 100 – 300оC. Static and dynamic loads up to 2 MN can act on the insulated pads during various plasma events in ITER.The qualification program for the insulated pads consists of cyclic and impact tests. The tests are performed for two categories of loading anticipated from plasma events. The impact tests are performed on a weight drop test bench, which has been designed and manufactured in JSC “NIKIET”.The description of the pad design and the impact test set-up, the results of analysis for justification of the set-up parameters, and the results of the test bench commissioning are presented in this paper.

Experimental validation of enhanced heat flux first wall panel mechanical attachment system

JSC NIKIET is a main supplier of ITER in-vessel components and its responsibility includes the manufacture of the FW beam, finger bodies, and the mechanical attachment system of the Enhanced Heat Flux (EHF) First Wall (FW) Panels in the framework of Procurement Arrangement 1.6.P1А.RF.01 dated 14.02.2014. The mechanical attachment system comprises the central bolt, threaded barrel and system washers. This FW mechanical attachment system allows the mounting of the FW onto the installed Shield Block (SB) while accommodating all external forces acting during plasma disruptions and normal operation.The design of EHF FW mechanical attachment system has been developed and optimized through a collaboration between the ITER Organization Central Team (IO-CT) and Russian specialists in order to provide the possibility to use it for all the FW panels. Full scale mockups of the FW attachment system components have been manufactured in JSC NIKIET in 2015–2016 and mechanical tests have been performed to demonstrate operation of these mock-ups under a 1 MN cyclic force.This paper summarizes the design description and experimental results of the FW attachment system.

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.

Nuclear analysis of the ITER In-Vessel viewing system

The In-Vessel Viewing System (IVVS) is a diagnostic tool in ITER to perform in-vessel inspections between plasma pulses or during shutdown. During plasma pulses, the IVVS is parked inside a dedicated port extending from the vacuum vessel (VV) to the port cell (PC). The IVVS design must fulfill specific requirements related to the service lifetime of sensitive parts and to the radiation effects on neighbouring components and the PC.Analyses were performed to determine nuclear responses in the IVVS and neighbor components at plasma operation and after shutdown. The neutron flux map shows improvements due to improved shielding in the neighbouring lower ports and gives insight to further possibilities of reducing nuclear loads to IVVS sensitive components. The activation analysis reveals that the major contributors to the decay photon source are the movable shield block counterweights.

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.

Radiation In-Port Cross-Talks for ITER Port Diagnostics

This paper emphasizes the need of estimation of the mutual influence, called “cross-talk,” for neutronic analyses of neighboring diagnostics systems shared by the same ITER port. Using examples of several diagnostic systems inserted inside the ITER Equatorial and Upper Port Plugs, we have demonstrated this mutual influence. Cross-talk effects have been shown by examining the radiation environment inside the port plug in terms of neutron energy spectra and Shut-Down Dose Rate (SDDR) inside the Port Interspace (PI) area. In-port cross-talk was investigated for the diagnostic systems deployed in two Equatorial Port Plugs (EPP) #17 and #8, and for the components of Upper Port Plug (UPP) #3. One example of in-port cross-talks is a gamma shadow effect of the Tritium and Deposit Monitor (TDM) shield block, which affects the SDDR inside the PI of EPP#17. Where the gamma radiation originated from the dominant radioactive sources of the irradiated structures of Core-Imaging X-ray Spectrometer (CIXS) is blocked by the TDM shield. Another example is an influence of neutron streaming along the Fast Ion Loss Detector (FILD) channel on the neutron energy spectra calculated in the Tangential Neutron Spectrometer (TNS) in EPP#8. For the example of UPP#3 with Charge eXchange Recombination Spectroscopy (CXRS-core), performed neutronic analysis identified excessive neutron streaming along the CXRS shutter, which must be reduced by further design iterations.