ITER Lifetime Research Articles

Verification of lifetime and mechanical load aspects for ITER diagnostic pressure gauges based on ZrC emitter.

Hot cathode ionization gauges will measure the neutral gas pressure in the vacuum vessel of ITER. Overall, 52 gauge heads based on the concept of the ASDEX pressure gauge but using novel ZrC emitters are located in the divertor, in equatorial ports and in pumping ducts. The initial lifetime tests of the novel gauge design have raised the concern that the pyrolytic graphite used to indirectly heat the ceramic ZrC emitter erodes too fast during operation. This effect might limit the lifetime of the gauges below the limits acceptable according to the inherent availability (ignoring operational delays) requirement of 99.8% for the diagnostic system over the lifetime of ITER. Long term operation tests have been performed using 0.1mm thick W, Ir, and Mo foils as an interlayer between the pyrolytic graphite and the ZrC within a constant atmosphere of 20Pa H2. The longest lifetime of 860h has been achieved with the Ir foil. Subsequent vibration tests applying accelerations on a shaker with magnitudes typical for seismic and ITER disruption loads demonstrated that the emitter can survive such demanding loads even after far-progressed erosion of the pyrolytic graphite. Additional ON/OFF cycle tests demonstrated that the gauge is still fully operational. In combination with a revision and consolidation of the operational plan and the reliability, accessibility, maintainability, and inspectability (RAMI) analysis, ITER's inherent availability requirement for the diagnostic system could be demonstrated successfully.

Remote maintenance scenario of the caesium ovens and beam source of the ITER heating neutral beam

Maintenance of the Heating Neutral Beam (HNB) injectors that are located in the ITER Neutral Beam (NB) cell needs Remote Handling (RH) operations because the NB cell is a radiation controlled area. This paper describes the remote maintenance sequences of the caesium ovens and the Beam Source (BS) of the ITER HNB by using the Neutral Beam Remote Handling System (NBRHS). Caesium oven maintenance is a regular maintenance that needs to be performed every Long-Term Maintenance (LTM) period as caesium is a consumable to produce caesium vapor to enhance negative ion production. The BS maintenance is a corrective maintenance to do if it fails during the lifetime of ITER. Maintenance of the caesium ovens and the BS involves many RH operations that require collaboration of the NBRHS subsystems and RH tools. Moreover, the maintenance operation needs to be done without spreading contamination from the HNB vessel to the NB cell. The developed maintenance scenario relies on the current NB cell environment and the existing interfaces in the plant system.

Simulation and developments for large pellet formation and acceleration for shattered pellet injection of the ITER DMS

Amongst the technological developments needed for ITER, the disruption mitigation system (DMS) is one of the most challenging since it has to mitigate the consequences of high-energy plasma disruptions to ensure the lifetime of ITER's structure and in-vessel components. ITER DMS is using Shattered Pellet Injection, whose technology needs to be developed for ITER specific requirements.In the framework of the “Technology Fundamental Studies for DMS” contract with ITER Organization, as part of the ITER DMS Task Force activities, and CEA, the DSBT (Low Temperature Systems Department) designed and operates a pellet injection test bench to produce and accelerate large cryogenic pellets with diameters from 10 to 28.5 mm and length/diameter ratios up to two. The objective of this activity is to study and to optimise the pellet formation and acceleration process using pure Protium, Deuterium, Neon and mixtures of these gases.Analytic and numerical studies were used to define the geometrical and cryogenic parameters for the pellet desublimation in the in-situ condensation cell of the cryostat. An analytical model has also been developed for the pellet acceleration in the single stage gas gun to validate the propellant valve design aiming at achieving velocities of up to 500 m/s for 28.5 mm protium pellets and 200 m/s for neon pellets.The complete test bench is composed of the gas propulsion system, the cryostat with its key component the pellet formation cell, the flight tube and the target chamber with viewing ports. The mechanical assembly is completed with specific instrumentation and fast data acquisition systems for the observation of the pellet formation and acceleration. The pellet integrity and trajectory are characterised by shadowgraphy with high resolution, high-speed cameras located at 1 m and 3 m after the barrel exit and with an accelerometer and a target foil in the target chamber.This document details the simulation studies, the selected test bench design and first obtained results.

Overview of recent ITER TBM Port Plug R&D activities

The objective of ITER project is the demonstration of the scientific and technological feasibility of fusion energy. Tritium fuel is required for the fusion reaction in DEMOnstration reactors and future commercial reactors but it is not available in nature because of its instability. To investigate tritium breeding self-sufficiency, two equatorial ports in ITER are dedicated to operating test blanket modules (TBMs) using the TBM Port Plug (TBM-PP). The TBM-PP design should be investigated with the physical mock-ups to validate the leak tightness in the primary vacuum boundary and to ensure TBM replacement and refurbishment compatible with Remote Handling (RH) operations in the ITER Hot Cell Facility during the ITER lifetime. This paper summarizes the achievements on the two most relevant R&D topics as described herewith. Several TBM-PP components (i.e., back parts of the TBM-Frame, the Dummy-TBMs, the TBM-Shields, and the TBM feeding pipes) are part of the vacuum boundary implying that these components are Vacuum Quality Classification (VQC) 1A. To test and develop high vacuum sealing for the TBM-Frame flange and the TBM-Set or Dummy-TBM flange, the ITER Large Seal Test Rig (LSTR), and metallic gasket seals for the TBM application have been manufactured and installed in the ITER site. The helium leak tests were performed at different operating temperatures to investigate the vacuum performance of metallic gasket seals with VQC 1A. The TBM-PP design shall guarantee the feasibility of a rapid replacement and refurbishment compatible with ITER Remote Handling (RH) operations by using RH tools in the Hot Cell Facility during the ITER's lifetime. Therefore, an experimental program has been performed to demonstrate the feasibility of the critical RH refurbishment tasks including the insertion and removal of TBM-Set/Dummy-TBM and the RH bolting.

Remote maintenance scenario of the beam line components of the ITER neutral beam system

The ITER neutral beam (NB) cell contains the neutral beam injectors (NBI) that provide heating power for fusion reaction in the tokamak machine. The NB cell is a radiation-controlled area where human worker access is strictly limited or prohibited depending on the NB cell configurations. Establishing the maintenance processes in the early design phase of a RH system is crucial in order to guarantee that the overall system configuration satisfies the maintenance requirements in given conditions and constraints. This paper describes the remote maintenance scenario of the beam line components (BLC) of the NB system by using the neutral beam remote handling system (NBRHS). This operation is a corrective maintenance operation that could happen multiple times during lifetime of ITER. If one of the BLCs fails, the failed NBI is isolated by closing the absolute valve to continue the tokamak operation. After that when the long-term maintenance campaign starts, the failed BLC needs to be replaced. The maintenance sequences involve human worker hands-on operations in the hot cell complex (HCC) and in the NB cell to prepare the RH operations, deployment operations of the NBRHS and the RH tools, removal/opening/transportation/storage operations to access to the failed component, and to remove and replace the failed component. Many RH tools are required during these processes such as unbolting/bolting, connector disconnection/connection, pipe cutting/welding/inspection, heavy and large component handling etc. Moreover, the operations need to be done without or minimizing spreading of contamination from the NBI vessels/components to the NB cell as the NB cell shall remain human accessible area during the long-term maintenance periods. The maintenance sequence study shows that the developed RH sequence is applicable in the current NB cell environment and is compatible with the defined interfaces in the plant systems. The developed scenario needs to be elaborated further detail in the next design phases together with the detail design of the RH equipment and the RH tools.

Investigation of plasma wall interactions between tungsten plasma facing components and helium plasmas in the WEST tokamak

ITER will operate with a tungsten divertor, a material featuring surface morphology changes when exposed to helium plasmas, in particular the formation of the so called tungsten fuzz under specific conditions. Investigating interactions between tungsten plasma facing components and helium plasmas in a tokamak environment is therefore a key point to consolidate predictions for the ITER divertor performance and lifetime. To this end, a dedicated helium campaign was performed in the full tungsten WEST tokamak, cumulating ∼2000 s of repetitive L mode discharges. It is shown that conditions for tungsten fuzz formation, as derived from linear devices experiments (incident helium energy E inc > 20 eV, helium fluence >1024 He/m2, surface temperature T surf > 700 °C), were met in the outer strike point (OSP) area of the inertially cooled tungsten divertor elements of WEST. Preliminary inspection of the components after the campaign did not show visible signs of surface modification, but points to significant net erosion in the OSP area. An extensive post mortem analysis is now ongoing to confirm these first indications. These results underline that the complex balance between erosion/redeposition (in particular linked to impurities) and tungsten fuzz formation needs to be taken into account in tokamak conditions.

RAMI analysis for the ITER In-Vessel Coils System

The ITER project is the largest international collaboration in the scientific field ever set up. It forms the heart of a unique collaborative agreement to build the first experimental nuclear fusion device designed to prove the scientific and technological feasibility of sustained fusion power generation. Among hundreds of systems, the In-Vessel Coils System plays a crucial role in ITER: ensuring and maintaining stable plasma operations. ITER has an ambitious inherent availability target for plasma production. To be able to achieve this objective, the potential technical risk that may affect the machine operation, was assessed by means of RAMI (Reliability, Availability, Maintainability and Inspectability) approach. The RAMI analysis for the IVC system was performed on the system design, following the Staged Approach Configuration adopted by ITER for Construction and Operation phases. A functional breakdown was prepared, resulting in the system being divided into 3 main functions described using the IDEFØ method. A qualitative analysis was performed applying the Failure Modes, Effects and Critical Analysis technique in order to identify the potential effects due to different anomalies of the IVC components. Criticality chart highlights the distribution of the Failure Modes across three risk zones with regard to their probability of occurrence and the impact on the availability. In relation to the operation of the IVC System, the analysis identified 32 risks, none of whom were classified as major. To estimate the reliability and availability of IVC system and to assess the compliance to the requirements, a quantitative analysis was performed using the Reliability Block Diagrams Analysis. The inherent availability of the IVC system during DT phase was calculated over the 20 years of ITER lifetime. The availability of the ITER IVC system results to be compliant with the inherent availability requirement for DT phase.

Radiation resistance of single-mode optical fibres with view to in-reactor applications

Single-mode optical fibres (SMFs) are required for ITER in-vessel applications as transport fibres to deliver the signal at wavelength λ = 1.55 µm from/to optical fibre sensors. The paper describes the first comparison study of radiation resistance of six pure-silica-core SMFs of different manufacturers performed in the process of fission-reactor irradiation in the conditions corresponding to the whole ITER lifetime (fast-neutron fluence, flux, γ-dose and dose rate of up to 1.8·1020n/cm2, 1.08·1014n/(cm2·s), 2.32 GGy and 1.39 kGy/s, respectively, temperature of 170–190 °C and vacuum pumping). Polyimide- and acrylate-coated SMFs failed mechanically during the irradiation; therefore, only metal-coated fibres can be considered for the in-vessel applications. Induced optical loss in all three metal-coated SMFs (copper- and aluminum-coated ones) at the fast-neutron fluence of 1·1020n/cm2 was found to lie in the range ~1.5–4 dB/m, the lower value of this range allowing the employment of at least 10-m-long transport fibre lengths in the in-vessel applications, assuming the dynamic range of 30 dB. The temperature-dependent microbending optical loss in metal-coated SMFs is discussed, of which the contribution to the total induced loss may be comparable in value to the radiation-induced absorption of light. Neither radiation-induced luminescence, nor Cerenkov emission was detected in the SMFs in the near-infrared range.

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.

Progress of ITER-India activities for ITER deliverables—challenges and mitigation measures

The responsibilities of ITER-India include a mix of precision, heavy, R&D intensive and interface intensive systems, under the built-to-print and functional systems categories. In several systems, the components are the first or largest of their kind. The uniqueness of component specifications and adherence to the desired quality standards considering the harsh nuclear environment in which the components need to operate and survive the lifetime of ITER leads to a challenging situation—namely that neither the existing laboratories nor potential suppliers have ever done or encountered such a scale-up (either in size, capacity, precision, quality etc) and in many cases do not have even the research and development (R&D) infrastructure to match ITER’s requirements. To bridge this gap, ITER-India adopted a two-step approach of first manufacturing prototypes as per ITER-specific standards and then subjecting them all to critical evaluation before launching bulk production. Some of the key areas of development include demonstrations of the following: (a) 1.5 MW radiofrequency power in the 35–65 MHz frequency range under Diacrode and Tetrode tube technologies for the ion cyclotron resonance heating system; (b) a low loss multi-process pipe cryogenic transfer line and the development of a 3.3 kg s−1 cold circulator for the cryoline and cryo-distribution system; (c) an angled accelerator grid with precision in positioning of apertures within 50 μm, as a first of its kind development, and special, ITER grade, copper alloy development for the diagnostic neutral beam (DNB) system. Full scale test facilities have been built for the performance assessment of the DNB, the electron cyclotron and ion cyclotron sources, cryogenic transfer lines and diagnostics systems. A summary is presented of the technical features of the major areas of procurements, notable achievements in R&D and manufacturing and their application in the Indian R&D related to fusion.

R&D status of the Indian test facility for ITER diagnostic neutral beam characterization

The Indian test facility (INTF) is a R&D facility under development at the Institute for Plasma Research, Gandhinagar, as part of the neutral beam development from negative ion source program. The major advantage of the INTF, besides developing beams from large ion sources, is to characterize and benchmark the ITER diagnostic neutral beam (DNB) to the desired specifications over transport lengths of ~21 m, a unique feature of this test bed. Such a study will establish the expected power to be delivered by DNB into the ITER plasma, an important parameter to estimate the S/N ratio expected from the He ash measurements by charge exchange recombination spectroscopy. It may be noted that DNB is a procurement package for India as an in-kind contribution to ITER. Extensive physics and thermomechanical calculations have been performed to finalize the component design and layout. Adequate choice of materials, and manufacturing and jointing processes compatible to ITER safety standards have been made in order to make the components adhere to the safety and quality classification, thereby ensuring that the components survive the ITER lifetime while operating in harsh nuclear environments. The components are currently in various phases of manufacturing, and the first operations of the INTF are anticipated in Q4 of 2019.

Irradiation test of fiber optics for H-alpha diagnostics in ITER

The reliability of the optical diagnostics in ITER critically depends on a radiation resistance of the fiber optics for a transmission of the plasma light to remote detectors. The design of the H-alpha diagnostics includes the fiber bundles 60 m long, the first 10 m of that run through the gamma-neutron fields. This part of the bundle will accumulate the absorbed dose of no more than 10 kGy (for silica) and the total neutron fluence of 1014 n/cm2 in the ITER lifetime. The modern types of the radiation resistant silica-based fibers can be applicable in the visible range under such condition. The convenient way to select the fiber type is in-situ gamma irradiation test. The test has to be performed at a dose rate comparable to the radiation conditions in the ITER facility. The report is devoted to the irradiation tests of three types of the silica based fibers in the Co-60 gamma source at the dose rate of 0.015 Gy/s. The transmittance losses of the fibers were measured in-situ in the spectral range of 450–700 nm. The conclusion was made about the applicability of the tested fibers for the H-alpha diagnostics in ITER.

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.

Specific Features of Fiber Optics Application in ITER

Silica-based optical fibers have high transmission in the visible range and are widely used for transmitting light from plasma to detectors in modern fusion devices. Pure silica-core/F-doped silica-clad fibers are mainly applied for plasma diagnostics. The possible problems that can be encountered with fiber optics in ITER diagnostics are discussed. The neutron fluence to be accumulated in the bundles during the ITER lifetime is estimated. The results of the irradiation tests of the optical fibers at the fission reactor and the 14 MeV neutron generator are used for analysis of radiation resistance of the optical fibers. The results of radiation tests and neutron calculations prove the requirement to apply silica fibers with high radiation resistance in ITER. It is recommended to place the fiber bundles in the port cell behind the biological shielding, where the neutron flux from the plasma is attenuated to a value of ~107 n/(cm2 s). The conclusion is made that it is possible to use the modern optical fibers in the plasma optical diagnostics (in a spectral range of 450–700 nm), the optical fibers being capable of surviving the radiation load during the whole ITER lifetime.

Design advances of the Core Plasma Thomson Scattering diagnostic for ITER

The Core Plasma Thomson Scattering (CPTS) diagnostic on ITER performs measurements of the electron temperature and density profiles which are critical to the understanding of the ITER plasma. The diagnostic must satisfy the ITER project requirements, which translate to requirements on performance as well as reliability, safety and engineering. The implications are particularly challenging for beam dump lifetime, the need for continuous active alignment of the diagnostic during operation, allowable neutron flux in the interspace and the protection of the first mirror from plasma deposition. The CPTS design has been evolving over a number of years. One recent improvement is that the collection optics have been modified to include freeform surfaces. These freeform surfaces introduce extra complexity to the manufacturing but provide greater flexibility in the design. The greater flexibility introduced allows for example to lower neutron throughput or use fewer surfaces while improving optical performance. Performance assessment has shown that scattering from a 1064 nm laser will be sufficient to meet the measurement requirements, at least for the system at the start of operations. Optical transmission at λ < 600 nm is expected to degrade over the ITER lifetime due to fibre darkening and deposition on the first mirror. For this reason, it is proposed that the diagnostic should additionally include measurements of TS 'depolarised light' and a 1319 nm laser system. These additional techniques have different spectral and polarisation dependencies compared to scattering from a 1064 nm laser and hence provide greater robustness into the inferred measurements of Te and ne in the core.

Nuclear responses in the ITER IVVS port cell

The ITER In-Vessel Viewing System (IVVS) consists of six identical units located at the B1 level of the Tokamak complex. In this work relevant nuclear quantities in the Port Cell (PC) have been calculated by means of the MCNP Monte Carlo code in a full 3-D geometry, including the IVVS and its shielding blocks. A comprehensive MCNP model of the PC has been developed including a detailed description of the Bioshield plug, pipes, penetrations, cask rails and PC door. The neutron and gamma sources needed to perform the nuclear analyses include both contributions from the radiation streaming through the Lower Port and activated water. Monte Carlo calculations have been performed to assess the radiation field inside the PC through maps of neutron and gamma fluxes. Absorbed dose on sensitive components over the ITER lifetime has been calculated in the PC area, in order to estimate the nuclear loads that the installed equipment has to withstand. Moreover, the shutdown dose rate distribution at 1day after shutdown has been calculated.

Review of LIBS application in nuclear fusion technology

Nuclear fusion has enormous potential to greatly affect global energy production. The next-generation tokamak ITER, which is aimed at demonstrating the feasibility of energy production from fusion on a commercial scale, is under construction. Wall erosion, material transport, and fuel retention are known factors that shorten the lifetime of ITER during tokamak operation and give rise to safety issues. These factors, which must be understood and solved early in the process of fusion reactor design and development, are among the most important concerns for the community of plasma–wall interaction researchers. To date, laser techniques are among the most promising methods that can solve these open ITER issues, and laser-induced breakdown spectroscopy (LIBS) is an ideal candidate for online monitoring of the walls of current and next-generation (such as ITER) fusion devices. LIBS is a widely used technique for various applications. It has been considered recently as a promising tool for analyzing plasma-facing components in fusion devices in situ. This article reviews the experiments that have been performed by many research groups to assess the feasibility of LIBS for this purpose.

Heating Neutral Beams for ITER: Present Status

The heating neutral beam (HNB) systems at ITER are designed to inject a total of 33 MW of either 1 MeV $\text{D}^{0}$ or 870 keV $\text{H}^{0}$ beams into the ITER plasma using two injectors with a possible addition of a third injector later to increase the injected power to $\sim 50$ MW. The injectors operate in a radioactive environment and should survive the life time of ITER, placing thereby stringent requirements on material and manufacturing choices. To ensure a smooth operational phase of neutral beams at ITER, a neutral beam test facility is under construction at Consorzio RFX, Padova, (hereinafter referred to as RFX), and consists of two test beds. The 100-kV SPIDER test bed will be used to optimize the source operation for H and D beams. The 1-MV MITICA test bed is essentially a full scale ITER prototype injector. The manufacturing and operational experiences at MITICA will not only establish the manufacturing processes of ITER HNB components but will also allow validation of the operational space of the injectors for ITER HNB. Operation of the two facilities is expected to begin in 2016 and 2019, respectively. Currently, the experiments on the ELISE facility, IPP Garching, with a half ITER sized RF beam source are underway. The ITER relevant parameters for the H beams have been achieved. Efforts are underway to optimize the same with D beams. The experimental database from ELISE will be an important input for establishing the SPIDER operation. This paper discusses the present status of the design and development of the injectors for ITER and the progress on the test facilities.

Erosion corrosion of CuCrZr specimens exposed for simulated ITER operational conditions

Corrosion is one of the main factors that determine the operational lifetime of any type of thermal power station. The ITER In-Vessel design includes coolant water interfaces with the copper alloy CuCrZr that are exposed to high flow velocities. These interfaces are near the plasma and thus subject to oxidants such as H2O2 generated through irradiation-induced radiolysis of the coolant water when the plasma is active.The erosion corrosion and release rates of CuCrZr and CuCrZr/316L(N)-IG joints were derived from weight change data and metallographic examinations of specimens after autoclave exposures at 110°C, 150°C and 250°C at reducing-, oxidizing- and cyclic redox water chemistry conditions. Reducing water chemistry represents periods considered as the nominal off-plasma operational conditions while the oxidizing environment simulates situations when the plasma is active. The cyclic redox conditions represent periods with shorter cycles simulating plasma activations with subsequent periodical variation of the water chemistry from reducing to oxidizing. The erosion corrosion rates for CuCrZr at active plasma conditions were 20 and 40µm/year at 110 and 150°C. At 250°C the corresponding rate was much higher. This result gives important information on what may happen if, for example, a First Wall panel is exposed to an unexpectedly high heat flux. Under reducing conditions the erosion corrosion rates were 3µm/year and 20µm/year at 110°C and 250°C respectively. The results at 250°C under off-plasma conditions reveal that the effect of erosion corrosion also has to be taken into account during baking. Cyclic conditions with respect to oxidant content turned out to be the most demanding environment (more demanding than pure oxidizing conditions) for the CuCrZr-alloy. Erosion corrosion rates of 90µm/year and 370µm/year at 110 and 150°C were recorded respectively. The highest temperature 250°C was not tested. This raise the question whether measures should be introduced that renders the system either oxidizing or reducing. In summary, the erosion corrosion rates recorded for CuCrZr under simulated ITER coolant water conditions are high, especially during plasma operation. For comparison it can be mentioned that corrosion rates of structural materials in the primary loop of light water reactors generally are considerably lower than 1µm/year. The estimated rates for CuCrZr are then 1–3 orders of magnitude higher, however the short total duration of plasma operation during ITER life time must be considered in this context. Erosion corrosion of CuCrZr can potentially cause problems for the ITER coolant systems and the chemical control systems.

Characterization of the radiation field and evaluation of the nuclear responses in the ITER cryopump port cell

The ITER main vacuum system consists of six torus exhaust pumps integrated in the Cryostat through dedicated housings at the building B1 level. The Port Cell area outside the cryopump plug is affected by neutrons streaming through the housing structure and diagnostics penetrations. The aim of the study presented in this paper is to perform a complete assessment of the nuclear responses in the Cryopump Port Cell #12 by means of the MCNP-5 Monte Carlo code in a full 3-D geometry. The results of the neutronic analyses provide guidelines for the design and maintenance of the embedded components, ensuring their structural integrity and proper operation.Radiation transport calculations have been carried out to determine the radiation field inside the Port Cell through 3-D neutrons and gamma maps. Nuclear heating induced by neutrons and photons and the absorbed dose during the ITER lifetime on steel and silicon have been estimated in the Port Cell area, in order to assess the nuclear loads that the diagnostics and related electronic components have to withstand. Furthermore, the impact of the gamma-rays emitted by neutron-activated water circulating in the Primary Heat Transfer System have been evaluated on the Port Cell environment: 3-D maps of the gamma flux, nuclear loads on steel/silicon during plasma operation are provided.