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One of the important aspects of the service life of the first and second optical mirrors (FM and SM) in ITER UWAVS diagnostics is to understand the influence of plasma cleaning on materials re-deposition and optical surface quality after multiple cleaning cycles. The capacitively coupled RF 30-60 MHz system is a candidate for the UWAVS mirror cleaning. It generates ion fluxes of tens of electron-volts, sputtering contaminants as well as construction materials. In the present report, we discuss materials re-deposition after cleaning with 30-60 MHz discharges inside the prototype of the front-end vacuum compartment. Typical ion energies were below 100 eV at the RF electrode and 20-30 eV at the grounded wall. Materials sputtered from the FM were deposited at various locations and on the SM. The deposition rates were estimated as 0.05 nm/hour or lower. The experiments showed advantages of Molybdenum as a plasma facing construction material due to lower sputtering. Traces of Fluorine were found, which may indicate cable damage at higher powers. The single-crystal Mo-mirror mock-up was etched at 0.5-1 nm/hr. rate. The mirror reflectivity showed satisfactory results after 60 h of exposure, with 50-60 nm removed. The overall effect on the reflectivity considering full service time is yet to be investigated.

First metallic mirrors used in optical diagnostics such as the Upper Wide Angle Viewing System (UWAVS) in ITER may undergo contamination with beryllium and tungsten. Contamination levels of 10 nm can significantly degrade mirror performance. A UWAVS first mirror cleaning system prototype uses radio-frequency (RF) gas discharge to remove contaminants. Although tungsten is not expected to be a main contaminant, removing it with ion sputtering may become challenging due to the high atomic mass. Cleaning tests were focused on 10–20 nm tungsten layers removal from various substrates using a tailored RF circuit in He and Ne. W-coatings were also produced on an existing single-crystal (SC) Mo-mirror 105 mm diameter. He and Ne were chosen for many tests due to their effectiveness and suitability for ITER operations. Sputtering conditions were determined through ion energy measurements. Ion energies of 100–150 eV and currents of 0.5–0.8 A/m2 were found to be optimum for 30–40 MHz RF discharges. The cleaning rate in He was 0.3–0.5 nm/hour for W-deposits. Etching rate of 1 nm/hour was observed for the Mo-mirror. No surface roughness degradation was determined with SEM and optical interferometry after mirror material was removed. Overall exposure time of the SC Mo-mirror was 100 h.

In ITER, first mirrors for the Upper Wide Angle Viewing System (UWAVS) are expected to experience contamination with beryllium and tungsten. Contamination levels of 10 nm can degrade mirror performance. In the UWAVS, a first mirror cleaning system will use radio-frequency (RF) gas discharge. A mockup in representative geometry was developed. The tests focused on power delivery, hardware survivability, cleaning uniformity, and removal of contaminants. Using RF discharges without introducing surface damage was confirmed. Sputtering was investigated for 10–20 nm thick Al-coatings used as proxies for Be due to health and safety issues. Helium was chosen for many tests due to its effectiveness and suitability for ITER operations. The most effective cleaning was obtained in Helium at 5 Pa, 36–43 MHz, and 50–150 W. The sputtering conditions were determined through ion energy measurements. Ion energies of 100–150 eV and currents of 0.5–0.8 A/m2 were found to be optimum. The measured cleaning rate in He for Al- and Al-oxide was 1.5–3 nm/hour. Aluminum was used as a proxy material for Be. The setup was operated for 500 h. No significant chamber contamination was identified.

An ITER hard core component (HCC) is defined for each undesirable situation with cliff edge effect consequences. Such structure system components (SCCs) are designed to prevent these situations at nuclear power plants, as well as to return and maintain a safe state in the event such a situation occurs. Only two hard core situations can really lead to cliff edge effect unless HCCs are implemented to limit the consequences: extreme earthquake (SL-3) and multiples fire in tritium building initiated by SL-3. The equipment (window and/or shutter) that closes penetrations from both sides of gallery to outside (Diagnostic building, Tritium building, and Assembly building) are classified as HCCs and SIC-2. Design status of the HCC penetrations belong to US ITER diagnostic systems will be presented in this publication.

ITER is the world’s largest fusion device currently under construction in the South of France with >50 diagnostic systems to be installed inside the port plugs (PPs), the interspace (IS), or the port cell region of various diagnostic ports. The plasma facing diagnostic first wall (DFW) and its supporting diagnostic shield modules (DSM) are designed to protect front-end diagnostics from plasma neutron and plasma radiation, while providing apertures for diagnostic access to the plasma. The design of ITER port plug structures including the DFW and the DSM is largely driven by the electromagnetic loads included on the PP structural components during plasma major disruptions and the vertical displacement events (VDEs). Unlike DFW and DSM, the design of diagnostic system, however, is likely driven by the steady-state thermal loads from plasma volumetric and surface heating and the dynamic response of the in-port components attached to the port-specific DSM or closure plate under transient loads induced on the vacuum vessel and the port extension during asymmetric VDEs. Three tenant diagnostic systems are integrated into the equatorial port 09. The toroidal interferometer/polarimeter, or TIP system, is installed in the left drawer (DSM3, left looking from plasma) for measuring the plasma density so to control the fuel input. The electron cyclotron emission (ECE) system is installed in the middle drawer (DSM2) to provide the high spatial and temporal resolution measurements of electron temperature evolution and the electron thermal transport inferences. The visible/infrared wide angle viewing system is installed in the right drawer (DSM1) to provide visible and infrared viewing and temperature data of the first wall for its protection in support of machine operation. The port plug integration design and multiphysics analyses are performed following port integration requirements including the weight limit (45 tones total), shut down dose rate limits, the cooling/heating and structural integrity validation. Mass distribution for TIP and ECE DSMs has been optimized to minimize the total weight by a new design of the boron carbide shielding pocket. The lightened DSM maintains its front-end load distribution and the structural stiffness with minimum impact to the DFW so to better protect on-board diagnostics; while still provides sufficient front-end stiffness for its structural integrity as well as the diagnostics function requirements.

International thermonuclear experimental reactor (ITER) diagnostic port plugs perform many functions including nuclear shielding and structural support of diagnostic system, while allowing for diagnostic access to the plasma. With design advancing, the in-port diagnostic components are integrated into the port plug structure, and the diagnostic shield modules (DSMs) are customized to accommodate various in-port diagnostic components. This paper summarizes results of transient electromagnetic analysis using Opera 3-D in support of recent design activities for ITER diagnostic upper port plug 14 (UPP14) and upper port plug 11 (UPP11). A complete distribution of disruption loads on each component in the two ports is presented. Impacts of different design features to the electromagnetic loads are discussed.

ITER is the world’s largest fusion device currently under construction in the South of France with >50 diagnostic systems to be installed inside the port plugs (PPs), the interspace (IS), or the port cell (PC) region of various diagnostic ports. Four tenant diagnostic systems will be integrated into the upper port 14 (U14), the upper wide angle viewing system, the disruption mitigation system, the glow discharge cleaning (GDC) system and the plasma position reflectometry, an ex-vessel tenant installed in the U14 IS region. Multiphysics analyses were performed for an in-vessel and ex-vessel component design integration following the ITER port integration requirements and the ITER guideline for load specification. Electromagnetic (EM), inertial, and heating loads of the main U14 components under various conditions for the diagnostic shield module (DSM) and its internal shielding block attachment design have been summarized. The structural design of in-port components is based on an optimization of the EM load distribution to limit the load impact to the diagnostic first wall and the tenant diagnostics and to mitigate the effect of dynamic amplification from asymmetric plasma vertical displacement event inertial loads. The DSM needs to provide a sufficient stiffness for the protection of onboard diagnostics and structural integrity of the port integration under the design-driving EM and inertial loads. The in-port diagnostics and mounting supports, on the other hand, are designed to meet the remote handling requirement, while taking into account the thermal loads due to temperature gradient from nuclear heating, as well as matching the relative stiffness among the structural components within the port assembly. Progress on an integrated design and analysis is reported including a seismic analysis of the IS/PC region for extracting tenant interface loads. The design-driving loads for tenant systems are presented for the attachment structure design as part of the PP engineering tasks.

One of the diagnostic systems being provided by the US is the Upper Wide Angle Viewing System (UWAVS) which provides real-time, simultaneous visible and infrared (IR) images of the ITER divertor regions via optical systems located in five upper ports. 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. The primary design challenge of the in-vessel assembly is maximizing performance of the overall system while surviving the severe electromagnetic and nuclear ITER environment. A first mirror material study was conducted and determined that single crystal molybdenum was the best choice for the first two mirrors in the optical train. A fail open, bellows actuated shutter with cross pivot flexure design was determined to be the most reliable mechanism to protect the first mirror. A geometrically representative glow discharge mirror cleaning system is being designed and will be tested to maximize cleaning effectiveness while minimizing optical degradation of the molybdenum mirrors adjacent to the plasma. The shutter and first mirror assemblies were packaged and designed for replacement via remote handling methods to minimize radiation waste and cost associated with eventual component replacement. The preliminary optical and structural design provides a robust and reliable system while maximizing the field of view. The R&D efforts, the technical challenges and issues, and the design and analysis results are presented.

International Thermonuclear Experimental Reactor (ITER) diagnostic port plugs perform many functions including nuclear shielding, structural support of diagnostic system, while allowing for diagnostic access to the plasma. With design advancing, the in-port diagnostic components are integrated into the port plug structure, and the diagnostic shield modules (DSM) are customized to accommodate the in-port diagnostic components. This technical note summarizes results of transient electro-magnetic analysis using Opera 3d in support of recent design activities for ITER diagnostic equatorial port plug (EPP). A complete distribution of disruption loads on each component in EPP9 is presented. Impacts of different design features, such as the locations of the electrical contact, to the EM loads are discussed, and solutions for improving the port structure design are proposed.

As part of the diagnostic port plug assembly, the ITER Diagnostic Shield Module (DSM) is designed to provide mechanical support and the plasma shielding while allowing access to plasma diagnostics. Thermal and hydraulic analysis of the DSM was performed using a conjugate heat transfer approach, in which heat transfer was resolved in both solid and liquid parts, and simultaneously, fluid dynamics analysis was performed only in the liquid part. ITER Diagnostic First Wall (DFW) and cooling tubing were also included in the analysis. This allowed direct modeling of the interface between DSM and DFW, and also direct assessment of the coolant flow distribution between the parts of DSM and DFW to ensure DSM design meets the DFW cooling requirements. Design of the DSM included voids filled with Boron Carbide pellets, allowing weight reduction while keeping shielding capability of the DSM. These voids were modeled as a continuous solid with smeared material properties using analytical relation for thermal c...

The ITER Electron Cyclotron Emission (ECE) diagnostic is progressing towards its Preliminary Design Review (PDR). In parallel, the diagnostic integration in the Equatorial Port is ongoing. Port Integration has to address the structural integrity to withstand various loads, maintenance and the safety aspects of ECE diagnostic. The ITER ECE system includes radial and oblique lines-of-sight. Recently, a successful peer-review of the in-port plug Hot Calibration Source has taken place and its performance and integration feasibility has been demonstrated. Four 45-meter long low-loss transmission lines are designed to transmit mm-wave power in the frequency range of 70- 1000 GHz in both X- and O-mode polarization from the port plug to the ECE instrumentation room in the diagnostic building. Prototype transmission lines are being tested [1]. A prototype polarizing Martin-Puplett type Fourier Transform Spectrometer (FTS) operating in the frequency range 70-1000 GHz, has a fast scanning mechanism and a cryo-cooled dual-channel THz detector system. Its performance has been tested as per ITER requirements. Assessment of the instrumentation and control requirements, functional and non-functional requirements, operation procedures, plant automation are ongoing for the PDR. The current status of the diagnostic, together with integration activities, is presented.

The generic upper port plug (GUPP) structure in ITER is a 6 m long metal box which deploys diagnostic components into the vacuum vessel. This structure is commonly used for all the diagnostic upper ports. The final design of the GUPP structure, which has successfully passed the final design review in 2013, is described here. The diagnostic port plug is cantilevered to the vacuum vessel with a heavy payload at the front, so called the diagnostic first wall (DFW) and the diagnostic shield module (DSM). Most of electromagnetic (EM) load (∼80%) occurs in DFW/DSM. Therefore, the mounting design to transfer the EM load from DFW/DSM to the GUPP structure is challenging, which should also comply with thermal expansion and tolerance for assembly and manufacturing. Another key design parameter to be considered is the gap between the port plug and the vacuum vessel port. The gap should be large enough to accommodate the remote handling of the heavy port plug (max. 25 t), the structural deflection due to external loads and machine assembly tolerance. At the same time, the gap should be minimized to stop the neutron streaming according to the ALARA (as low as reasonably achievable) principle. With these design constraints, the GUPP structure should also provide space for diagnostic integration as much as possible. This requirement has led to the single wall structure having the gun-drilled water channels inside the structure. Furthermore, intensive efforts have been made on the manufacturing study including material selection, manufacturing codes and French regulation related to nuclear equipment and safety. All these main design and manufacturing aspects are discussed in this paper, including requirements, interfaces, loads and structural assessment and maintenance.

ITER diagnostic first walls (DFWs) and diagnostic shield modules (DSMs) inside the port plugs (PPs) are designed to protect diagnostic instrument and components from a harsh plasma environment and provide structural support while allowing for diagnostic access to the plasma. The design of DFWs and DSMs are driven by 1) plasma radiation and nuclear heating during normal operation 2) electromagnetic loads during plasma events and associate component structural responses. A multi-physics engineering analysis protocol for the design has been established at Princeton Plasma Physics Laboratory and it was used for the design of ITER DFWs and DSMs. The analyses were performed to address challenging design issues based on resultant stresses and deflections of the DFW-DSM-PP assembly for the main load cases. ITER Structural Design Criteria for In-Vessel Components (SDC-IC) required for design by analysis and three major issues driving the mechanical design of ITER DFWs are discussed. The general guidelines for the DSM design have been established as a result of design parametric studies.

We performed numerical simulations of the ITER Diagnostic First Wall (DFW) using ANSYS workbench. During operation DFW will include solid main body as well as liquid coolant. Thus thermal and hydraulic analysis of the DFW was performed using conjugated heat transfer approach, in which heat transfer was resolved in both solid and liquid parts, and simultaneously fluid dynamics analysis was performed only in the liquid part. This approach includes interface between solid and liquid part of the systemAnalysis was performed using ANSYS CFX software. CFX software allows solution of heat transfer equations in solid and liquid part, and solution of the flow equations in the liquid part. Coolant flow in the DFW was assumed turbulent and was resolved using Reynolds averaged Navier-Stokes equations with Shear Stress Transport turbulence model. Meshing was performed using CFX method available within ANSYS. The data cloud for thermal loading consisting of volumetric heating and surface heating was imported into CFX Volumetric heating source was generated using Attila software. Surface heating was obtained using radiation heat transfer analysis. Our results allowed us to identify areas of excessive heating. Proposals for cooling channel relocation were made. Additional suggestions were made to improve hydraulic performance of the coolingmore » system.« less

The Diagnostic First Walls (DFWs) were designed to handle the plasma nuclear and radiant heating along with electro-magnetic loading induced from plasma disruptions. The DFWs also provide custom viewing apertures for the diagnostics within. Consequently, the DFWs contain numerous complex water cooling channels and are designed per ITER SDC-IC for design by analysis.This paper presents the analyses of the Upper Port DFWs proceeding to a final design review. The finite element analyses (FEAs) performed include neutronics, radiative heating, coupled fluid dynamics and heat transfer, and static and transient structural analysis using the combined multi-physics load conditions. Static structural FEAs performed account for the dynamic amplification effects of the transient load. A detailed bolt analysis was also performed per the ITER SDC-IC bolt evaluation based on reaction loads obtained from the mechanical simulations.

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 2more » 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.« less

The Upper port Wide Angle Viewing System (UWAVS) is diagnostic system that provides visible light and InfraRed images of the divertor area in the ITER tokamak. The main functions of the UWAVS are to provide divertor wall and plasma luminance and temperature distributions at high spatial resolution (3–5 mm) and frame rates (500–1000 Hz) for machine protection as well as to enable species influx monitoring (Be, C, O, Ne, Kr, Ar) to the divertor. In addition the UWAVS supports visual inspection as well as various other measurements. For this purpose an optical system is designed to image the divertor area onto four different cameras in the port cell area. Altogether five of these systems are foreseen to provide sufficient coverage of the divertor area (150m2). Each of these systems is equipped with active alignment to compensate for vessel motion, internal transmission monitoring sources to assess mirror reflectivity degradation, as well as calibration sources in the port cell to monitor the camera pixel response, all to guarantee the overall systems performance during operation and over lifetime. All these aspects are combined into an optical design targeting diffraction limited imaging, taking in account the complexities of the harsh ITER environment. This design and its performance are supported by manufacturability analysis, tolerancing analysis and alignment strategy and Structural Thermal Optical Performance (STOP) analysis. The latter is an TNO in-house developed method that provides a direct and reproducible link between the structural thermal analysis for a combined load case comprising nuclear heating, cooling water pressure and temperature, gravity and mounting stresses (in Ansys) and the associated changes of the optical performance, including imaging quality and alignment (in Zemax OpticStudio).

According to Internationa Thermonuclear Experimental Reactor (ITER) integration procurement arrangements, the installment of diagnostics in a port should not increase the shutdown dose rate (SDDR) in the port interspace area by no more than ${\sim}{\rm 50}~\mu{\rm Sv}/{\rm h}$ above the baseline, assuming another 50 $\mu{\rm Sv}/{\rm h}$ is attributed to contribution from the port structure and other ITER in-vessel components, such that the upper SDDR limit of 100 $\mu{\rm Sv}/{\rm h}$ is not exceeded $10^{6}~{\rm s}$ after shutdown. It was found that placing the initial design of the motional stark effect (MSE) and the charge exchange recombination spectroscopy (CXRS) in the equatorial port #3 resulted in an increase in the SDDR that far exceeded the limit. When we follow the optimization process discussed in this paper, substantial reduction in the port interspace SDDR was achieved. The results of this paper show that even when we combine the optimized CXRS and MSE diagnostics with a third glow discharge diagnostic, the excess of the SDDR over the baseline value did not exceed the allowed upper limit. This paper is based on utilizing the 3-D CAD-based ATTILA code for assessing the SDDR.

ITER diagnostic port plugs perform many functions including structural support of diagnostic systems under high electromagnetic loads while allowing for diagnostic access to the plasma. The design of diagnostic equatorial port plugs (EPP) are largely driven by electromagnetic loads and associate responses of EPP structure during plasma disruptions and VDEs. This paper summarizes results of transient electromagnetic analysis using Opera 3d in support of the design activities for ITER diagnostic EPP. A complete distribution of disruption loads on the diagnostic first walls (DFWs), diagnostic shield modules (DSMs) and the EPP structure, as well as impact on the system design integration due to electrical contact among various EPP structural components are discussed.

The upper, equatorial, and lower diagnostics port plugs in ITER will include numerous intermingling labyrinths and many streaming paths whose impact should be carefully studied. For this purpose, the 3-D Discrete Ordinates code, Attila, has been routinely used by PPPL/UCLA to assess the nuclear field in these geometrically complex plugs both during operation and after shutdown. In this paper we describe the calculation procedure followed and the input parameters/assumptions applied to assess the shutdown dose rates (SDDR) everywhere with emphasize on their values inside the generic equatorial port plug (GEPP) and its inter-space extension area. Factors inherent in the Discrete Ordinates method that impact the accuracy of the results (e.g. quadrature sets used, boundary conditions applied, etc.) are discussed. Means to minimize streaming through straight gaps and open channels present in the GEPP are presented in this paper, along with an examination of their effectiveness in reducing the SDDR in t...

The ITER Diagnostic Division is responsible for designing and procuring the First Wall Blankets that are mounted on the vacuum vessel port plugs at both the upper and equatorial levels This paper will discuss the effects of the diagnostic aperture shape and configuration on the coolant circuit design. The DFW design is driven in large part by the need to conform the coolant arrangement to a wide variety of diagnostic apertures combined with the more severe heating conditions at the surface facing the plasma, the first wall. At the first wall, a radiant heat flux of 35W/cm2 combines with approximate peak volumetric heating rates of 8W/cm3 (equatorial ports) and 5W/cm3 (upper ports). Here at the FW, a fast thermal response is desirable and leads to a thin element between the heat flux and coolant. This requirement is opposed by the wish for a thicker FW element to accommodate surface erosion and other off-normal plasma events.

The port-based plasma diagnostic infrastructure on ITER is described, including the port plugs, the interspace support structure and port cell structure. These systems are modular in nature with st...

The operation of ITER requires high-quality estimates of the plasma electron density over multiple regions in the plasma for plasma evaluation, plasma control and machine protection purposes. Although the density regimes of ITER are not very different from those of existing tokamaks (1018–1021 m−3), the severe conditions of the fusion plasma environment present particular challenges to implementing these density diagnostics. In this paper we present an overview of the array of ITER electron density diagnostics designed to measure over the entire ITER domain: plasma core, pedestal, edge, scrape-off layer and divertor. It will focus on the challenges faced in making these measurements, and the technical solutions of the current designs.

The ITER project is considering the inclusion of two sets of in-vessel coils, one to mitigate the effect of Edge Localized Modes (ELMs) and another to provide vertical stabilization (VS). The in-vessel location (behind the blanket shield modules, mounted to the vacuum vessel inner wall) presents special challenges in terms of nuclear radiation (~3000 MGy) and temperature (100oC vessel during operations, 200oC during bakeout). Mineral insulated conductors are well suited to this environment but are not commercially available in the large cross section required. An R&D program is underway to demonstrate the production of mineral insulated (MgO or Spinel) hollow copper conductor with stainless steel jacketing needed for these coils. A preliminary design based on this conductor technology has been developed and is presented herein.

The USITER, through the Princeton Plasma Physics Lab (PPPL), is responsible for the delivery of several fully integrated upper, equatorial and lower port plugs dedicated for the diagnostics in ITER. Each port plug package consists of a generic port plug structure and a set of diagnostics and diagnostic housings. The shielding design of the integrated port plugs calls for maintaining a dose level not to exceed 100 μSv/h inside the interspace of each port; the room behind the port plug where maintenance personnel access the rear of the port. This is set as an upper target design in order to perform routine maintenance 1E6 sec (∼two weeks) following shutdown. Expensive remote handling robots and tooling are required otherwise. In this paper we present results from a parametric study aimed at providing initial assessment of the attainable dose rates in the diagnostics ports and their extension areas in order to properly address the duration time and frequency for the workers to perform the scheduled maintenance. The nuclear analysis is performed using both the serial version and the distributed memory parallel (DMP) version of the ATTILA-7.1.0, 3-D FEM Discrete Ordinates code, along with the FENDL2.1/FORNAX and ANSI/ANS-6.1.1-1977 data bases.

The nuclear engineering infrastructure of port-based diagnostics on ITER is presented, including the equatorial and upper port plug generic designs, the adopted modular concept, the loads and associated load response and the remote handling. A modular approach is adopted for the internal shielding modules as this helps to reduce the dominant electromagnetic loads, reduces interfaces and accelerates remote handling. Extensive use of gun-drilling is employed to produce water channels in the stainless steel plate and forgings used to construct the port plugs and modules. The French Nuclear Code RCC-MR (2007) is employed in the design and the manufacturing. An integrated remote handling scheme is described, including a description of maintenance operations in the Hot Cell Facility.

Neutron transport modeling of the ITER reactor structure including modeling the impact of potential neutron streaming along the divertor cassette requires a detailed 3-D CAD solid model of the ITER...

ITER plasma boundary and first-wall diagnostics are summarized in terms of their physical implementation and physics motivation. The challenge of extracting diagnostic signals while maintaining nuclear shielding is discussed, as well as the problems associated with high levels of erosion and redeposition.

The USITER Project is responsible for providing a generic design of 18 diagnostics upper port plugs (UPPs). These plugs are designed to provide effective nuclear shielding, adequate nuclear heat removal capability, and allowance for personnel accessibility shortly after shutdown. In addition, the design limits for other parameters should not be exceeded during ITER lifetime. This includes the accumulative DPA and helium production in the UPP structure and the vacuum vessel, fast neutron fluence reaching the neighboring TF and PF magnets, and the accumulative local heating in their coil casings. In the present work, we assess these performance characteristics in three generic upper port plug (GUPP) designs of the in-plug shield and labyrinths that span possible configurations ranging from a conservative to worst case. The cases considered are: (1) Case A: “Visible-IR Camera diagnostic” with an intact port-attached blanket shield module (BSM), and (2) Case B: vis/IR system where the port-based BSM is replaced by overlapping adjacent vessel-attached BSM and (3) Case C: a “Large Aperture diagnostic” that is loosely resembles ECH style shielding plugs and aperture. These cases were modeled in a 20-degree sector of ITER. The 3D FEM Discrete Ordinates Code, ATTILA, was used in the analysis along with FENDL2.1 data and the FORNAX activation file.

There is currently extensive effort to develop and benchmark new codes to supplement the accepted computational tools for the nuclear design of ITER. The ITER management and quality program (MQP) officially requested benchmarking the newly developed FEM 3D, discrete ordinates code, ATTILA, to the results obtained by the CAD-based MCNP Monte Carlo codes. CAD-based computational tools are needed to reduce the turnaround time between changes in design and the subsequent nuclear analysis to key design parameters in an iterative process. There is currently extensive effort to develop and use a CAD–MCNP interface for design purposes to facilitate the modeling and analyses. ATTILA uses CAD-based geometrical input and as such it is a potentially quicker alternative to the MCNP code for the neutronics studies contingent on its success in predicting key neutronics parameters in large systems such as ITER. The ITER neutronics community had agreed to use a standard CAD model of ITER (40° sector, denoted as“Benchmark CAD Model”) to compare results for several responses selected for calculation benchmarking purposes to test the efficiency and accuracy of the CAD–MCNP approach developed by each ITER party in addition to ATTILA. In this paper we report such comparison to the results of the CAD-based MCNP developed at the University of Wisconsin for five selected nuclear responses. It is shown that ATTILA's results are within 25–30% of the MCNP results.

Recent advances in radiation transport simulation tools enable an increased fidelity and accuracy in modeling complex geometries in fusion systems. Future neutronics calculations will increasingly be based directly on these 3-D CAD-based geometries, allowing enhanced model complexity and improved quality assurance. Improvements have been made in both stochastic and deterministic radiation transport methodologies and their new capabilities will be compared briefly. A code comparison benchmark exercise has been specified based on a 40° sector of the ITER machine and the analysis results show good agreement. Additional analyses will be discussed, with particular attention to how these new capabilities provide new insights for engineering design of ITER components.

The Upper Wide Angle Viewing System (UWAVS) will be installed on five upper ports of ITER. This paper shows major requirements, gives an overview of the preliminary design with reasons for some design choices, examines self-emitted IR light from UWAVS optics and its effect on accuracy, and shows calculations of signal-to-noise ratios for the two-color temperature output as a function of integration time and divertor temperature. Accurate temperature output requires correction for vacuum window absorption vs. wavelength and for self-emitted IR, which requires good measurement of the temperature of the optical components. The anticipated signal-to-noise ratio using presently available IR cameras is adequate for the required 500 Hz frame rate.

Eddy current induced electromagnetic (EM) loads during plasma events are the design driver for ITER port plug (PP) structure, diagnostic first wall (DFW), shield module (DSM) and diagnostic system supported by the PP structure. Generic models using commercial software OPERA, MAXWELL and ANSYS are developed and benchmarks are performed for global EM analysis to obtain port-specific design driving EM loads. The 20 degree vessel sector models of the upper and equatorial PP structure take the same ITER TF, CS and PF coil and plasma current (15 MA baseline plasma scenario) as input, then solve for eddy current induced on all passive structural components for various DINA disruption cases. The worst load case can be exponential decay or linear decay of plasma current depending on the dimension and location of diagnostic components inside the PPs. Static and transient magnetic fields from generic models are mapped onto an excel datasheet to establish component design load specification. Three levels of modeling effort are suggested. The global model analysis is used to validate the impact of component design to the PP global system response as well as to study its component dynamic effect as a result of the disruption loads on the PP structure. The local sub-model analysis can be used to extract more accurate EM loads on diagnostics and the excel datasheet of static and transient field maps are used as the initial design load specification for in-port components.

There will be 50 diagnostic systems installed on ITER. All have been implemented with great success on experimental fusion reactors around the world. Despite this deep experience, implementation of diagnostic systems on ITER remains very challenging. Structural, nuclear and optical engineering challenges arise from loads that are at least an order of magnitude higher and last for much longer than any previous experiment. Complicating this challenge is an acceleration of the design process because the systems delivered to ITER must be ready for full power D-T operations from day-One. Talented engineering teams around the world are tackling these challenges in many innovative ways. In this paper ITER diagnostic systems from the US will be used to show a selection of examples of how the “D-T Ready” engineering challenges are being answered.

A new diagnostic for the National Spherical Torus Experiment is described whose function is to measure ion rotation and temperature at the plasma edge. The diagnostic is sensitive to C III, C IV, and He II intrinsic emission, covering a radial region of 15 cm at the extreme edge of the outboard midplane. Thirteen chords are distributed between toroidal and poloidal views, allowing the toroidal and poloidal rotation and temperature of the plasma edge to be simultaneously measured with 10 ms resolution. Combined with the local pressure gradient and EFIT reconstructed magnetic field profile, the edge flow gives a measure of the local radial electric field.

According to ITER integration procurement arrangements, the installment of diagnostics in a port should not increase the shutdown dose rate (SDDR) in the port interspace area by no more than ~50 microSv/hr above the baseline, assuming another 50 microSv/hr is attributed to contribution from the port structure and other ITER in-vessel components, such that the upper SDDR limit of 100 microSv/hr is not exceeded 106 s after shutdown. It was found that placing the initial design of the Motional Stark Effect, MSE, and the Charge Exchange Recombination Spectroscopy, CXRS, in the equatorial port #3 (EPP3) resulted in an increase in the SDDR that far exceeded the limit. When we follow the optimization process discussed in this paper, substantial reduction in the port interspace SDDR was achieved. The results of the present work show that even when we combine the optimized CXRS and MSE diagnostics with a third GDC diagnostic, the excess of the SDDR over the baseline value did not exceed the allowed upper limit. The present work is based on utilizing the 3-D CAD-based Attila code for assessing the SDDR.

ITER diagnostic port plugs perform many functions including structural support of diagnostic systems under high electromagnetic loads while allowing for diagnostic access to the plasma. During plasma disruptions, a large amount of induced current flows locally at electrical contacts between diagnostic first walls (DFWs) and the diagnostic shield modules (DSMs). Even a small gap voltage (10-30V) between DFWs, DSMs and supporting rails may trigger local arcing and cause arc damage to the conducting structure. This is particularly true when we consider the ionized gas environment and halo current effect. We perform global electromagnetic analysis with contact details for DFWs and DSMs to quantify the gap voltage and local current transfer effect during plasma disruptions. Electrical contacts between the DFWs and DSMs may also have significant impact on disruption load and thus affect design of the DFW attachment scheme. Large current transfer (>100 kA) between DFWs and DSMs through the attachment keys and tabs during disruption implies local heating and potential welding. This paper reviews the contact current and electrical potential difference between the DFWs, DSMs and the port plug structure. We also assess the impact on the system design itself due to electrical contact among various components.

One of the systems planned for the measurement of electron density in ITER is a multi-channel tangentially viewing combined interferometer-polarimeter (TIP). This work discusses the current status of the design, including a preliminary optical table layout, calibration options, error sources, and performance projections based on a CO2/CO laser system. In the current design, two-color interferometry is carried out at 10.59 μm and 5.42 μm and a separate polarimetry measurement of the plasma induced Faraday effect, utilizing the rotating wave technique, is made at 10.59 μm. The inclusion of polarimetry provides an independent measure of the electron density and can also be used to correct the conventional two-color interferometer for fringe skips at all densities, up to and beyond the Greenwald limit. The system features five chords with independent first mirrors to reduce risks associated with deposition, erosion, etc., and a common first wall hole to minimize penetration sizes. Simulations of performance for a projected ITER baseline discharge show the diagnostic will function as well as, or better than, comparable existing systems for feedback density control. Calculations also show that finite temperature effects will be significant in ITER even for moderate temperature plasmas and can lead to a significant underestimate of electron density. A secondary role TIP will fulfill is that of a density fluctuation diagnostic; using a toroidal Alfvén eigenmode as an example, simulations show TIP will be extremely robust in this capacity and potentially able to resolve coherent mode fluctuations with perturbed densities as low as δn∕n ≈ 10(-5).

Considerable progress has been made on the design of the ITER electron cyclotron emission (ECE) diagnostic over the past two years. Radial and oblique views are still included in the design in order to measure distortions in the electron momentum distribution, but the oblique view has been redirected to reduce stray millimeter radiation from the electron cyclotron heating system. A major challenge has been designing the 1000 K calibration sources and remotely activated mirrors located in the ECE diagnostic shield module (DSM) in the equatorial port plug #09. These critical systems are being modeled and prototypes are being developed. Providing adequate neutron shielding in the DSM while allowing sufficient space for optical components is also a significant challenge. Four 45-meter long low-loss transmission lines transport the 70–1000 GHz ECE from the DSM to the ECE instrumentation room. Prototype transmission lines are being tested, as are the polarization splitter modules that separate O-mode and X-mode polarized ECE. A highly integrated prototype 200–300 GHz radiometer is being tested on the DIII-D tokamak in the USA. Design activities also include integration of ECE signals into the ITER plasma control system and determining the hardware and software architecture needed to control and calibrate the ECE instruments.

The core and edge Thomson systems on Korea Superconducting Tokamak Advanced Research employ two different sets of lens collection optics. Their collection systems are positioned in the front end of a long reentrant cassette for optimum viewing coverage and optical throughput. Both systems collect the scattered light from a single tangential beam of multiple 50-Hz Nd:YAG lasers and image the scattering volume from core to edge with 40 spatial points. In order to obtain a higher resolution of 5 mm, the edge system has more spatial channels than the core system. Pressure-free heat shield windows, which will absorb the radiation heat flux, are mounted in front of large vacuum windows to protect them from the radiation heat load during long-pulse discharges.

ITER will incorporate In Vessel Coils (IVCs) as a method of stabilizing “Edge Localized Modes” (ELM) and providing “Vertical Stabilization” (VS). To meet the ELM and VS Coil requirements strong coupling with the plasma is required so that it is necessary for the coils to be installed in the vessel just behind the blanket shield modules. Due to this close proximity to the plasma the radiation and temperature environment is severe and conventional electrical insulation materials and processes cannot be used. The development of mineral insulated conductor technology has been required in the IVC design to deal with this high radiation and high temperature environment. While mineral insulated conductor technology is not new, building a large magnet with high current carrying capability and a conductor diameter larger than the mineral insulated conductor currently manufactured requires R&D and the extension of existing technologies. A 59mm Stainless Steel Jacketed Mineral Insulated Conductor (SSMIC) using MgO is being developed for this application. The IVC ELM and VS coils design includes both the development of the fabrication techniques for the SSMIC and the design and analysis of the ELM and VS Coil assemblies.

A diagnostic suite has been developed to measure the impurity poloidal flow using charge exchange recombination spectroscopy on the National Spherical Torus Experiment. Toroidal and poloidal viewing systems measure all the quantities required to determine the radial electric field. Two sets of up/down symmetric poloidal views are used to measure both the active emission in the plane of the neutral heating beams and the background emission in a radial plane away from the neutral beams. Differential velocity measurements isolate the line-integrated poloidal velocity from apparent flows due to the energy-dependent charge exchange cross section. Six f/1.8 spectrometers measure 276 spectra to obtain 75 active and 63 background channels every 10 ms. The local measurements from a similar midplane toroidal viewing system are mapped into two dimensions to allow the inversion of poloidal line-integrated measurements to obtain local poloidal velocity profiles. The radial resolution after inversion is 0.6-1.8 cm from the plasma edge to the center.

Neutronics analysis to find nuclear heating rates and personnel dose rates were conducted in support of the integration of diagnostics in to the ITER Upper Port Plugs. Simplified shielding models of the Visible-Infrared diagnostic and of the ECH heating system were incorporated in to the ITER global CAD model. Results for these systems are representative of typical designs with maximum shielding and a small aperture (Vis-IR) and minimal shielding with a large aperture (ECH). The neutronics discrete-ordinates code ATTILA® and SEVERIAN® (the ATTILA parallel processing version) was used. Material properties and the 500 MW D-T volume source were taken from the ITER "Brand Model" MCNP benchmark model. A biased quadrature set equivelant to Sn=32 and a scattering degree of Pn=3 were used along with a 46-neutron and 21-gamma FENDL energy subgrouping. Total nuclear heating (neutron plug gamma heating) in the upper port plugs ranged between 380 and 350 kW for the Vis-IR and ECH cases. The ECH or Large Aperture model exhibited lower total heating but much higher peak volumetric heating on the upper port plug structure. Personnel dose rates are calculated in a three step process involving a neutron-only transport calculation, the generation of activation volume sources at pre-defined time steps and finally gamma transport analyses are run for selected time steps. ANSI-ANS 6.1.1 1977 Flux-to-Dose conversion factors were used. Dose rates were evaluated for 1 full year of 500 MW DT operation which is comprised of 3000 1800-second pulses. After one year the machine is shut down for maintenance and personnel are permitted to access the diagnostic interspace after 2-weeks if dose rates are below 100 μSv/hr. Dose rates in the Visible-IR diagnostic model after one day of shutdown were 130 μSv/hr but fell below the limit to 90 μSv/hr 2-weeks later. The Large Aperture or ECH style shielding model exhibited higher and more persistent dose rates. After 1-day the dose rate was 230 μSv/hr but was still at 120 μSv/hr 4-weeks later. __________________________________________________

Neutronics analysis to find nuclear heating rates and personnel dose rates were conducted in support of the integration of diagnostics in to the ITER Upper Port Plugs. Simplified shielding models of the Visible-Infrared diagnostic and of a large aperture diagnostic were incorporated in to the ITER global CAD model. Results for these systems are representative of typical designs with maximum shielding and a small aperture (Vis-IR) and minimal shielding with a large aperture. The neutronics discrete-ordinates code ATTILA® and SEVERIAN® (the ATTILA parallel processing version) was used. Material properties and the 500 MW D-T volume source were taken from the ITER “Brand Model” MCNP benchmark model. A biased quadrature set equivelant to Sn=32 and a scattering degree of Pn=3 were used along with a 46-neutron and 21-gamma FENDL energy subgrouping. Total nuclear heating (neutron plug gamma heating) in the upper port plugs ranged between 380 and 350 kW for the Vis-IR and Large Aperture cases. The Large Aperture model exhibited lower total heating but much higher peak volumetric heating on the upper port plug structure. Personnel dose rates are calculated in a three step process involving a neutron-only transport calculation, the generation of activation volume sources at pre-defined time steps and finally gamma transport analyses are run for selected time steps. ANSI-ANS 6.1.1 1977 Flux-to-Dose conversion factors were used. Dose rates were evaluated for 1 full year of 500 MW D-T operation which is comprised of 3000 1800-second pulses. After one year the machine is shut down for maintenance and personnel are permitted to access the diagnostic interspace after 2-weeks if dose rates are below 100 µSv/hr. Dose rates in the Visible-IR diagnostic model after one day of shutdown were 130 µSv/hr but fell below the limit to 90 µSv/hr 2-weeks later. The Large Aperture style shielding model exhibited higher and more persistent dose rates. After 1-day the dose rate was 230 µSv/hr but was still at 120 µSv/hr 4-weeks later.

A collaboration is underway between Princeton Plasma Physics Lab and UCLA to develop skill in the use of ATTILA, to benchmark ATTILA against MCNP and to develop Solid Works CAD models for neutronics analysis of the diagnostic ports. MCNP along with the cross section library FENDL 2.1 is the accepted standard tool for neutronics analysis of ITER against which results from ATTILA are being compared. The MCNP community has established a set of benchmark results for a standardized 40 degree CAD model of ITER. These benchmark results create a framework for the acceptance of new applications like A TTILA by the ITER central neutronics, quality assurance and nuclear safety groups. Analysis of the benchmark model with ATTILA also leads to the setting of discrete ordinates solution parameters and model mesh refinement that will help to accelerate the analysis of future diagnostic port design iterations. Flux and heating results in the Divertor, Blanket Shield Modules and Equatorial Port Shielding from the ATTILA benchmarking show good correlation with MCNP results. TF heating results were in error by up to 50% due to poor mesh refinement and boundary condition issues in that area. Detailed models of the Upper and Equatorial ports, port plugs and diagnostics are under development. The detailed port study models will be 40 degree ITER models to preserve the shape of the neutron source loading. These models will include the inner and outer vacuum vessel, inner-wall shielding, blanket shield modules, divertor and cryostat. Models of the OH, TF and PF coils will not be included to save on element count.

A multi-channel Thomson scattering system viewing the edge and core of the KSTAR plasma will be installed at the midplane port Bay-N. An engineering design study was undertaken at PPPL in collaboration with the Korea Basic Science Institute (KB Si to determine the optimal optics and cassette design. Design criteria included environmental, mechanical and optical factors. All of the optical design options have common design features; the Thomson scattering laser, an in-vacuum shutter, a quartz heat shield and primary vacuum window, a set of optical elements and a fiber optic bundle. Neutron radiation damage was a major factor in the choice of competing lens-based and mirror-based optical designs. Both the mirror based design and the lens design are constrained by physical limits of the Bay-N cassette and interference with the Bay-N micro-wave launcher. The cassette will contain the optics and a rail system for maintenance of the optics

A new diagnostic for the National Spherical Torus Experiment (NSTX) is described whose function is to measure ion rotation and temperature at the plasma edge. The diagnostic is sensitive to C III, C IV, and He II intrinsic emission, covering a radial region of 15 cm at the extreme edge of the outboard midplane. Thirteen chords are distributed between toroidal and poloidal views, allowing the toroidal and poloidal rotation and temperature of the plasma edge to be simultaneously measured with 10 ms resolution. Combined with the local pressure gradient and the EFIT code reconstructed magnetic field profile, the edge flow gives a measure of the local radial electric field.

ITER ECE diagnostic [1] needs not only to meet measurement requirements, but also to withstand various loads, such as electromagnetic, mechanical, neutronic and thermal, and to be protected from stray ECH radiation at 170 GHz and other millimeter wave emission, like Collective Thomson scattering which is planned to operate at 60 GHz. Same or similar loads will be applied to other millimetre-wave diagnostics [2], located both in-vessel and in-port plugs. These loads must be taken into account throughout the design phases of the ECE and other microwave diagnostics to ensure their structural integrity and maintainability. The integration of microwave diagnostics with other ITER systems is another challenging activity which is currently ongoing through port integration and in-vessel integration work. Port Integration has to address the maintenance and the safety aspects of diagnostics, too. Engineering solutions which are being developed to support and to operate ITER ECE diagnostic, whilst complying with safety and maintenance requirements, are discussed in this paper.

In this study, the baseline design for the ITER electron cyclotron emission (ECE) diagnostic has entered the detailed preliminary design phase. Two plasma views are planned, a radial view and an oblique view that is sensitive to distortions in the electron momentum distribution near the average thermal momentum. Both views provide high spatial resolution electron temperature profiles when the momentum distribution remains Maxwellian. The ECE diagnostic system consists of the front-end optics, including two 1000 K calibration sources, in equatorial port plug EP9, the 70-1000 GHz transmission system from the front-end to the diagnostics hall, and the ECE instrumentation in the diagnostics hall. The baseline ECE instrumentation will include two Michelson interferometers that can simultaneously measure ordinary and extraordinary mode ECE from 70 to 1000 GHz, and two heterodyne radiometer systems, covering 122-230 GHz and 244-355 GHz. Significant design challenges include 1) developing highly-reliable 1000 K calibration sources and the associated shutters/mirrors, 2) providing compliant couplings between the front-end optics and the polarization splitter box that accommodate displacements of the vacuum vessel during plasma operations and bake out, 3) protecting components from damage due to stray ECH radiation and other intense millimeter wave emission and 4) providing the low-loss broadbandmore » transmission system.« less