Diagnostic Shield Module Research Articles

Mechanical design of ITER radial neutron camera Ex-Port system

The Radial Neutron Camera is an ITER diagnostic designed to measure the un-collided 14 MeV and 2.5 MeV neutrons from deuterium-tritium (DT) and deuterium-deuterium (DD) fusion reactions, through an array of detectors covering a poloidal plasma section along collimated Lines Of Sight (LOS). It is composed by two fan-shaped collimating structures viewing the plasma radially through vertical slots in the diagnostic shielding module of ITER Equatorial Port 1: the In-Port RNC, devoted to plasma edge coverage, and the Ex-Port RNC, devoted to the plasma core coverage. This paper presents an overview of the mechanical design of the Ex-Port RNC at the Preliminary Design Review (PDR) stage. The Ex-Port RNC is located in the Port Interspace and consists of a massive shielding structure hosting the detector units and two sets of collimators lying on different toroidal planes. The Ex-Port RNC design is presented both from the point of view of functional requirements (e.g. LOS positions and angles, radiation shielding, weight limitations) and of manufacturability. Finally, the Ex-port RNC structural integrity is assessed, and its design validated against the main loads and load combinations.

Nuclear analyses for the integration of ITER equatorial Port 2

The present work is devoted to nuclear analyses in support of the ITER diagnostic Equatorial Port 2 (EP#2) integration. ITER EP #2 is a diagnostic port based on the long-modular Diagnostic Shielding Module (DSM) housing the following systems: Disruption Mitigation System (DMS) in DSM#1 and #3 and X-Ray Crystal Spectroscopy Core (XRCS-Core) in DSM#2. Ensuring adequate radiation shielding is a major challenge since the diagnostic systems require several apertures from the Vacuum Vessel (VV) through the Port Interspace (PI) and up to the Port Cell (PC). In the present study, a three-dimensional MCNP model of EP#2 has been developed, starting from the latest design available from Preliminary Design Review stage (PDR), and successively integrated into the reference 40° ITER C-Model. Comprehensive nuclear analyses have been carried out employing the D1SUNED v3.1.4 code based on the MCNP Monte Carlo transport code. Relevant nuclear quantities during and at the end of plasma operations have been evaluated: i.e., neutron and gamma fluxes and energy spectra along the port from the Diagnostic First Wall (DFW) up to the Bio-Shield Plug (BP), nuclear heating, neutron damage, helium and tritium production, and shutdown dose rate. This analysis allowed the identification of potentially critical areas, and therefore, the implementation of additional shielding options aimed at reducing the neutron streaming and mitigating of the radiation field in the PI region. In this work, the results of the analyses are presented and discussed. Some solutions to mitigate nuclear loads and to improve the shielding in PI area are proposed and their impact has been assessed. Finally, some recommendations for the optimization of the design of EP#2 are provided as well.

Computationally efficient methodology of 3D thermal-hydraulic analysis of Diagnostic Shielding Module #2 of ITER Equatorial Port #11

The process of design justification of Diagnostic Shielding Modules (DSMs) of port plugs of the International Thermonuclear Experimental Reactor (ITER) includes a stage of thermal-hydraulic computational analysis. The ITER tokamak operates in a pulsation mode that means the nuclear heating affecting DSMs is unsteady, and transient analysis is required. Three-dimensional (3D) approaches for fully coupled thermal-hydraulic analysis, which imply simultaneous computations for fluid and solid parts of a 3D DSM model, lead typically to “heavy” numerical models being very expensive from the computational point of view, especially for transient simulations. In order to significantly reduce the consumption of computational resources, the present 3D thermal-hydraulic analysis of DSM #2 of ITER Equatorial Port #11 was performed using one-way coupling. Firstly, steady-state CFD computations of turbulent flow were performed for the front part of DSM cooling system, which is a high-density network of channels of complex geometry. Obtained in CFD analysis map of wall shear stress on surface of channels was then converted into a map of local heat transfer coefficient using a simple and conservative approach based on known empirical correlations for turbulent flow in a circular pipe. For the back part of DSM cooling system, which basically consist of many relatively long channels, an individual (single) value of heat transfer coefficient for each of the channels was directly evaluated using empirical correlation. The heat transfer coefficient maps obtained for both parts of DSM cooling system were used finally as a boundary condition in transient thermal analysis of the solid part of the DSM model. Details and justifications of the described computationally-efficient methodology are presented.

Nuclear Analyses for the Assessment of the Loads on the ITER Radial Neutron Camera In-Port System and Evaluation of Its Measurement Performances

The radial neutron camera (RNC) is a key ITER diagnostic system designed to measure the uncollided 14- and 2.5-MeV neutrons from deuterium–tritium (DT) and deuterium–deuterium (DD) fusion reactions, through an array of detectors covering a full poloidal plasma section along collimated lines of sight (LoS). Its main objective is the assessment of the neutron emissivity/ <inline-formula xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink"> <tex-math notation="LaTeX">$\alpha $ </tex-math></inline-formula> source profile and the total neutron source strength, providing spatially resolved measurements of several parameters needed for fusion power estimation, plasma control, and plasma physics studies. The present RNC layout is composed of two fan-shaped collimating structures viewing the plasma radially through vertical slots in the diagnostic shielding module (DSM) of ITER Equatorial Port 1 (EP01): the ex-port subsystem and the in-port one. The ex-port subsystem, devoted to the plasma core coverage, extends from the Port Interspace to the Bioshield Plug: it consists of a massive shielding unit hosting two sets of collimators lying on different toroidal planes, leading to a total of 16 interleaved LoS. The in-port system consists of a cassette, integrated inside the port plug DSM, containing two detectors per each of the six LoS looking at the plasma edges. The in-port system must guarantee the required measurement performances in critical operating conditions in terms of high radiation levels, given its proximity to the plasma neutron source. This article presents an updated neutronic analysis based on the latest design of the in-port system and port plug. It has been performed by means of the Monte Carlo MCNP code and provides nuclear loads on the in-port RNC during normal operating conditions (NOC) and inputs for the measurement performance analysis.

Nuclear Analysis for the Upper Ports in the NB Cell in ITER

In International Thermonuclear Experimental Reactor (ITER), 14 upper ports (UPs) are allocated for diagnostic systems mounted into the port plug. The upper port plug (UPP) consists of the port plug (PP) structure, diagnostic shield module (DSM), and diagnostic first wall (DFW). One of the main functions of the UPP is to provide the neutron shielding to reduce the radiation level in port interspace and the nuclear loads on the superconducting coils. In this work, the nuclear analysis of the UPs #4–#7 that are located in the neutral beam cell is presented. Two configurations have been analyzed with Monte Carlo N-Particle (MCNP) code: the generic UPP design of full shielding (22 tons) and a light version (18 tons) with reduced shielding. The UPP MCNP models have been developed on the basis of latest design files and integrated in a 120° neutral beam injection (NBI) sector extracted from E-lite. Calculations have been performed to evaluate the neutron flux at the port interspace, the nuclear heating of the PP and toroidal field (TF)/poloidal field (PF) coils (TFC/PFC), the neutron flux, and damage for the electrical feedthrough installed on the closure plate with D1SUNED v3.1.4 code. The impact of crosstalk with NBI and neighboring ports has been assessed for neutron flux and shutdown dose rate (SDDR) at 12 days after shutdown. The results of this analysis showed that the impact of the additional shielding of the 22 tons model compared to the 18 tons model is small and limited in space: the fluxes and damage at the electrical feedthrough are about 40%–50% smaller with additional shielding. Analyses enhanced a large radiation crosstalk with NBI ports and a large streaming through the gaps of UPPs.

A new method to reduce the toroidal field ripple effect due to the ferromagnetic materials in updated CFETR

The toroidal field ripple (TFR) is one of the critical factors affecting the long-pulse high-performance discharges for the tokamak device. The TFR is not only determined by the toroidal field coils (TFCs) but is also affected by the ferromagnetic materials in the plasma facing components such as diagnostic system and blanket system. In the phase Ⅱ design of the China Fusion Engineering Test Reactor (CFETR), the inner ferromagnetic constituent and the poloidal segmentation of water-cooled ceramic breeder (WCCB) blanket are modified and the diagnostic system also has major change compared with the phase Ⅰ design, therefore it is necessary to investigate and reduce the TFR effect of the updated blanket system and diagnostic system for the phase Ⅱ designed CFETR. Firstly, the effect of the blanket system on the TFR is evaluated systematically, considering the diagnostic shielding modules (DSMs) of the diagnostic system at the equatorial ports (EPs). Moreover, a new TFR reducing method by adjusting the distribution of ferromagnetic materials filling ratios in the toroidal direction is proposed. According to the results, the local TFR nearby the plasma separatrix line already exceeds the design specification limit with the blanket system, and this phenomenon will be worse due to installing the DSMs and the EP. After adjusting the percentage of the ferromagnetic materials in the outboard-side (OBs) segment, the plasma TFR is reduced obviously and within the 0.5% limit.

Progress in ITER ECE diagnostic design and integration

The ITER electron cyclotron emission (ECE) diagnostic system has primary roles in providing measurements of the core electron temperature profile and the electron temperature fluctuation associated with the neoclassical tearing modes. The ITER ECE system includes a radial and oblique line-of-sight. Four 43-meter long low-loss transmission lines (TLs) are designed to transmit millimeter 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. The measurement instrumentation includes two Fourier transform spectrometer (FTS) systems and two radiometer systems. The Indian Domestic Agency (IN-DA) and United States Domestic Agency share the responsibility. The IN-DA scope excluding instrumentation and control has passed its preliminary design review and is progressing towards the final design review (FDR). In parallel, the diagnostic integration in different areas is ongoing. Several captive components for the TLs have passed FDR and will be manufactured for installation in the tokamak building soon. A peer review meeting has been held on the prototype hot calibration source, and its integration and new thermal analysis in the diagnostic shield module are continuing. A prototype TL is being tested. A prototype polarizing Martin-Puplett type FTS, operating in the frequency range 70–1000 GHz, features an in-vacuo fast scanning mechanism and a cryo-cooled dual-channel THz detector system. Its performance has been assessed in detail against ITER requirements.

Modeling of the stress-strain state of ITER upper ports №2 & №8 constructions’ components under thermal, electromagnetic and inertial loads

This paper presents the results of calculations of the upper port plug (UPP) of the ITER upper ports №2 and №8. Detailed finite-element models of the UPP structure were developed taking into account nonlinear contact interaction between the diagnostic shield module with extension frame and the UPP structure. In the structural analysis under design loads the stress-strain state (SSS) taking into account bolts pretension was obtained. Besides, the FE model includes cooling system pipes and boron carbide shielding. In the thermal analysis of the UPP for baking and normal operation modes the neutron analysis data was taken into account. The UPP cooling systems’ hydrodynamic characteristics were calculated including the flow-pressure characteristics. Subsequently, the whole UPP construction flow and pressure characteristics were evaluated taking into account the presence of throttling devices in the UPP hydraulic circuit. Structural analysis of the UPPS during seismic and electromagnetic events was carried out using the response spectrum method based on the response spectra at the UP point of the ITER building and interface electromagnetic loads due to the vacuum vessel movements respectively. The FE model includes port extension from the generic upper port plug model. Except this, the response spectra at the tenant attachment points were calculated using power spectral density analysis. Laplace force distribution in the UPP under different scenarios of plasma disruption was calculated. The electromagnetic model was significantly refined and includes the extension frame trays. Based on the electromagnetic analysis data, the structural analysis of the UPP under electromagnetic loads was carried out.

Advanced design of the ITER core CXRS shutter and integration into the diagnostic shield module of the Upper Port Plug No. 3

For the ITER core CXRS diagnostic, a shutter is mandatory to protect the first mirror when no measurement takes place. With the continuing design development of the diagnostic and the knowledge obtained from previous shutter prototypes, the frictionless fast shutter did undergo some design changes and manufacturing improvements without degrading the performance (opening/closing <1 s). In this paper, the major modifications of the shutter design are presented. An adjustment system was developed to allow compensation of manufacturing deviations and precise positioning of the protective blade according to the final aperture position, which is not known before the final assembly with the as-built dimensions of the port plug and diagnostic shield module (DSM). The shutter components and their adjustment systems are implemented into the shutter‘s actively cooled base structure, which in turn simplifies the remote handling procedure. For the helium circuit an activity study is presented, which allows to access the nuclear pressure equipment classification and to determine the number of required purges during the ITER lifetime. With the heat loads obtained from the known heat flow on the blade and from the neutronics analysis, a thermal FE-analysis was performed on the full shutter model by which the expected heat flow at the mechanical interfaces can be given.

Neutronic analysis for the Edge Charge Exchange Recombination Spectroscopy in Equatorial Port of ITER

• Neutronics for Edge Charge Exchange Recombination Spectroscopy (CXRS-Edge). • CXRS-Edge Diagnostics installed in Equatorial Port #3 (EP3) of ITER. • New MCNP model of CXRS-Edge Diagnostic system in EP3 was created. • Personnel access to the Inter-Space Structure (ISS) the EP3 is evaluated. • Maps of local shutdown dose rate (SDDR) are shown in EP3 ISS with CXRS-Edge. The neutronic analysis results presented in this paper support the design development of the Edge Charge Exchange Recombination Spectroscopy (CXRS-Edge) system to be installed in Equatorial Port #3 (EP3) of ITER. That is an active optical diagnostic system with a dogleg labyrinth design of two optical pathways and a system of mirrors arranged in Diagnostic Shield Modules #1 and #2 (DSM #1 and #2) of EP3. The shielding performance of the CXRS-Edge system and the radiation effects of its mirror materials have been evaluated. To this end, a neutronic 3D model of the CXRS-Edge system has been created in the format of the MCNP code. The MCNP model of CXRS has been converted from the underlying CAD engineering model using the SuperMC code. Then the model has been integrated with the actual standard MCNP model of the ITER tokamak called “C-Model R181031 (Revision 190110)”. With this C-Model, neutron and photon fluxes, nuclear heating distributions and neutron damage in molybdenum, steel, and silicon carbide (SiC) materials have been calculated. Assessing the possibility of personnel access to the Inter-Space Structure (ISS) of EP3 with CXRS-Edge installed, a map of shutdown dose rate (SDDR) has been calculated using the local approach, without any other diagnostics systems shared by EP3. In conclusion, the suitability of CXRS Edge doglegs for necessary attenuations of neutron and photon fluxes, nuclear heating, and SDDR is demonstrated. As design work is progressing, this paper presents an initial contribution to the complete neutronic analysis.

Progress on final design of ITER radial X-ray camera

The design of ITER Radial X-ray Camera (RXC) is nearly finished. The design of internal camera was updated using gas detector modules and considering the mechanical, electrical and gas interfaces with Diagnostic Shielding Module (DSM). The design of camera structure was optimized for the purpose of easy maintenance, safety and simplification, including support flange, Be window flange, boron carbide shielding module, external camera chamber, back flange assembly, cooling and feedthrough. The design of camera electronics was also optimized. The pre-amplifiers and mid-amplifiers are more resistant to disturbances. The chassis facilitates the maintenance and provides efficient cooling to the components inside.

Preliminary Design of DSMs for ITER Upper Ports #02 and #08 Integration

Since 2014, Budker Institute of Nuclear Physics (BINP) SB RAS has been working on the design of ITER upper port-plugs (UPPs) #02 and #08. The design of upper diagnostic shielding modules (DSMs) for UP #02 and UP #08 has been performed using CATIA v5.0. DSMs represent a welded structure made of stainless steel SS316L(N)-IG. Cooling water channels to be performed use the unique gundrilling technology. The main conception of structural design of DSMs is in good agreement with ITER recommendations. The engineering calculations of DSM structure were done to demonstrate the sufficient strength of the considered design under thermal (normal operation and baking), electromagnetic (vertical displacement event (VDE) II), and seismic (SL-2) loads. To prove the satisfaction of upper DSMs design to ITER requirements, the load combination dead weight (DW) + normal operation (NO) + SL-1 + VDE II numerical simulations were performed. Further studies of the modular design will be optimized for the use of standardized boron carbide shielding trays.

Engineering Calculations and Preparation for Manufacturing of ITER Equatorial Port #11

The current integration state of ITER Equatorial Port #11 (EP#11) is described. This port is the first plasma port and thus the integration phase is close to the final stage. The new modular design of neutron shielding is implemented in the Diagnostic shielding modules’ (DSM) construction to achieve the maximum shielding efficiency within the weight restrictions and keeping fully operable the tenant diagnostics systems. The same approach was applied to Interspace structure frame design to allow the safe hand-on maintenance near the highly activated components. The current iteration of the engineering calculations, including the neutron transport and shutdown dose rate (SDDR), thermohydraulics, electromagnetics, seismic, and structure integrity calculations, showed that the current design of EP#11 satisfy all ITER requirements. The special metalworking, welding, and various other technologies are tested during the process of preparation for the manufacturing of ITER EP#11. Production technologies of the full-size DSM mock-up and various elements of its infrastructure were tested. The multizoned clean area assembly hall for the final assembly is close to getting completed and the testing of vacuum items has to be sent to ITER.

Overview of optical designs of the port-plug components for the ITER Equatorial Wide Angle Viewing System (WAVS)

The equatorial visible and infrared Wide Angle Viewing System (WAVS) for ITER is one of the key diagnostics for machine protection, plasma control and physics analysis. To achieve these objectives, the WAVS will monitor the surface temperature of the Plasma Facing Components (PFCs) by infrared (IR) thermography (3–5 μm range) and will image the edge plasma emission in the visible range. It will be composed of 15 lines of sight installed in four equatorial ports (no. 3, 9, 12 and 17) in order to survey at least 80% of the overall area of the vacuum vessel. This paper presents the optical design of port-plug which has to cope with both challenging performances (wide field of view, wide temperature range 200–2000 °C, high measurement accuracy required lower than 20%) and severe constraints such as harsh nuclear environment, complex interfaces. Indeed the optical components are embedded within the ITER Diagnostic Shielding Module, DSM, which is not usual for diffraction limited instruments. It is shown that nominal optical performance of the Port-Plug are limited by the diffraction only. In operation, this design has demonstrated its superiority on the former ones since it is much less sensitive to tolerance variations.

Standardized integration of ITER diagnostics Equatorial Port Plugs

The Diagnostic Shielding Modules (DSM) are secondary containers of the ITER diagnostic Port Plugs where shielding and diagnostics components have to be integrated together. For the development of Equatorial Port DSMs several key requirements have to be met by the Port Integrators. First of all, the total dry Plug weight must not exceed the allowable weight (currently 48 ton) in order to guarantee the consistency with the specification of other interfacing systems involved in the Port Plugs maintenance. Secondly, a shielding capability able to limit Shut-Down Dose Rate in the Interspace Zone (man access corridors) below 100 μSv/h needed to allow the human accessibility for the Port Plug sealing disassembly, the inspection and the maintenance of diagnostic components behind the Plug closure plate and in the interspace. Finally, Remote Handling (RH) compatibility for refurbishment of the Port Plug and maintenance of systems integrated is necessary as well. These operations involve activated components that are processed in the Hot Cell. As a result of these demands, the ITER Port Integration team has developed the so called Modular Equatorial DSM able to answer the weight/shielding/RH requirements described above following the ALARA principle and offering a standard infrastructure for the common DSM integration problems.

Thermal Loads and Cooling Design for ITER in-Port Low Field Side Reflectometer Diagnostic System

The current design for the ITER Low Field Side Reflectometer (LFSR) diagnostic system contains six circular waveguides that function as both launch and receive antennas. The front end of LFSR system, interfacing with the plasma facing DFW and its supporting diagnostic shielding Modules (DSM), is integrated into Equatorial Port Plug 11(EPP#11). In-vessel LFSR configuration is mainly divided into two sections, i.e. Front Antenna Unit (FAU) and in-vessel Transmission Line (TL) components. During plasma normal operation, the 14 MeV neutrons from D-T fusion reactions penetrate into the port plug and LFSR components. Neutronics code ATTILA is used for assessing the volumetric heating of in-vessel components. The maximum normal heat load due to radiation and the charge exchange from plasma is 0.35 MW/m2, which has to be applied on those surfaces affected by a viewing factor consistent with the aperture surface position and orientation. The front antenna unit (FAU) needs the active cooling to handle the radiant and nuclear heating from the plasma. Due to the specific viewing aperture layout at the front-end of in-vessel LFSR, the cooling channel design is challenging. The thermal-hydraulic iterations are performed to achieve the desirable temperature distribution of the water cooled front antenna units. Finally, the progressive ratcheting effects of the FAU subjected to cyclic thermal loading are examined by the preliminary non-linear elastic-plastic analysis.

Estimation of neutral fluxes on the first mirror of H-alpha diagnostics in ITER

The first mirrors of all optical diagnostics in ITER will be exposed to the fluxes of neutrals, mainly D, T and Be. This can lead to formation of erosion and deposition zones on the mirror surface. In this work, the fluxes of neutral D, T and Be on the first mirror of H-alpha diagnostics in the equatorial ports were analyzed with Zemax OpticStudio software. Only stationary modes of ITER were considered. The geometry of viewing cutouts in diagnostic shielding modules (DSM) and the model of the first mirror units (FMU) were taken from the ENOVIA database. It is shown that the net erosion is dominating in major part of the first mirror zone with erosion rate ≈100 nm/year. The net deposition zone is located on the edges of the mirror, where deposition rate reaches up to 900 nm/year. Being made of single crystal molybdenum, the mirrors will keep their optical properties under erosion over the whole ITER lifetime. The contaminated edges are not important for the light collection and imaging of the SOL emission in the areas of interest at the ITER first wall within the selected field-of-view. The DC plasma discharge mirror cleaning system will be used to recover FM performance in case of eventual contamination.

The ITER radial neutron camera in-port system: Nuclear analyses in support of its design development

The ITER Radial Neutron Camera (RNC) is a multichannel detection system hosted in the Equatorial Port Plug 1 (EPP 1). It is designed to measure the uncollided neutron flux from the plasma, providing information on the neutron emissivity profile and total strength. The RNC structure consists of two sub-systems based on fan-shaped arrays of cylindrical collimators: the ex-port system, covering the plasma core with 2 sets of lines of sight lying on different toroidal planes, and the in-port system, enclosed in a dedicated cassette within the EPP1 diagnostic shielding module, for the measurement of neutrons generated in the plasma edge. Due to the harsh environment in which it has to operate, the design of the in-port RNC system is particularly critical both from the measurements point of view (low signal to noise ratio induced by the high level of scattered neutrons at the detector positions) and from the structural point of view. The paper presents the results of the neutronic analyses performed with the MCNP Monte Carlo code with the aim of optimizing the in-port RNC design in order to enhance the diagnostic measurement performance and evaluating the nuclear loads that have to be withstand by its structural elements, detectors and associated components.

Neutronic effects of diagnostic shield module length on radiation environment of ITER diagnostic generic upper port plug

Newly developed 3D CAD-based MCNP neutronics model of ITER Diagnostic Generic Upper Port Plug (DGUPP) is a baseline MCNP model of the ITER tokamak Upper Port Plug. This paper presents 3D neutron flux and Shut-Down Dose Rate (SDDR) maps for the DGUPP integrated into the MCNP C-lite V2 global model. The SDDR assessments have been performed with the R2Smesh approach and focused on the UPP Inter-Space Structure (ISS), where personnel access is planned. Following the ALARA principle, SDDR in the ISS should be minimized, for that reason the length of the Diagnostic Shielding Module (DSM) was extended by filling all the UP structure till the UP Closure Plate with the DSM shielding material. The corresponding MCNP model was called DGUPP “long-DSM”, in contrast to the original CAD-based “short-DSM”. The results of the DSM extension indicated that the strongest, up to 3 times reduction of fluxes & SDDR could be reached at the UP Closure Plate, while inside the maintenance corridors the SDDR reduction is 11%. Found possibility to reduce the SDDR emphasizes the importance of this work.

Design and Analysis Progress of ITER Diagnostic Equatorial Port #09

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.