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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.

Nuclear analyses have been performed for the ITER Test Blanket Module Port Plug (TBM PP) using the MCNP-5 Monte Carlo Code. A detailed 3D model of the TBM Port Plug with dummy TBM has been integrated into the ITER MCNP model (B-lite v.3). Neutron fluxes, nuclear heating, helium production and neutron damage have been calculated in all the TBM PP components. Global shutdown dose rate calculations have also been performed with Advanced D1S method for different configurations of the TBM PP system. This paper presents the results of these analyses and discusses potential design improvements aiming to further reduce the shutdown dose rate in the port interspace. © 2015 Elsevier B.V. All rights reserved.

The present paper addresses the recent developments and applications of Advanced-D1S to the calculations of shutdown dose rate in tokamak devices. Results of benchmarking with measurements and Rigorous 2-Step (R2S) calculations are summarized and discussed as well as limitations and further developments. The outcomes confirm the essential role of the Advanced-D1S methodology and the evidence for its complementary use with the R2Smesh approach for the reliable assessment of shutdown dose rates and related statistical uncertainties in present and future fusion devices. © 2014 EURATOM-ENEA Association.

The High Resolution Neutron Spectrometer (HRNS) will be one of the largest and most complex neutronic diagnostics in ITER. It will be located in Equatorial Port Plug number 1, in which many other diagnostic devices are placed, increasing the interface problem. HRNS will be able to perform a wide range of measurements of plasma parameters in both DD and DT operations. Since detectors have to be placed in the port cell far away from plasma surface, an efficient collimating system has to be provided. The collimator has to allow an output neutron flux as intense as possible taking into account the ITER requirements in terms of nuclear safety and radioprotection. The collimator has also to control the quality of the spectral response arriving to the detectors, ensuring an acceptable Signal-to-Noise ratio. This study presents a set of parametric analyses performed with MCNP5 using a simple model. With this simple model the effect of geometric key parameters on the response of the collimator can be isolated, without the complexion of the geometry. Parameters such as the aperture of the cone of vision or the materials composing the collimator walls and the Port Plug have a significant effect on neutron flux at collimator output and interspace volume. Finally the results of all those geometric and material variations have been resumed to give a set of recommendations useful for a preliminary design of the collimator.

The ITER Vacuum Vessel has upper, equatorial, and lower ports used for equipment installation, diagnostics, heating and current drive systems, cryo-vacuum pumping, and access inside the vessel for maintenance. At the level of the divertor, the nine lower ports for remote handling, cryo-vacuum pumping and diagnostic are inclined downwards and toroidally located each every 40°. The cryopump port has additionally a branch to allocate a second cryopump. The ports, as openings in the Vacuum Vessel, permit radiation streaming out of the vessel which affects the heating in the components in the outer regions of the machine inside and outside the ports. Safety concerns are also raised with respect to the dose after shutdown at the cryostat behind the ports: in such zones the radiation dose level must be kept below the regulatory limit to allow personnel access for maintenance purposes. Neutronic analyses have been required to qualify the ITER project related to the lower ports. A 3-D model was used to take into account full details of the ports and the lower machine surroundings. MCNP version 5 1.40 has been used with the FENDL 2.1 nuclear data library. The ITER 40° model distributed by the ITER Organization was developed in the lower part to include the relevant details. The results of a first analysis, focused on cryopump system only, were recently published. In this paper more complete data on the cryopump port and analysis for the remote handling port and the diagnostic rack are presented; the results of both analyses give a complete map of the radiation loads in the outer divertor ports. Nuclear heating, dpa, tritium production, and dose rates after shutdown are provided and the implications for the design are discussed.

The principle needs of ITER with regard to nuclear analysis can be divided into the broad categories of safety and licensing, plant operation, and decommissioning although there is much overlap and...

The exploitation of ITER tokamak will require diagnostics for machine protection, inputs to plasma control systems, evaluation and analysis of plasma parameters and performances. The equatorial visible/infrared wide angle viewing system and the radial neutron camera are the two main diagnostics of Procurement Package 11 (PP11), one of the diagnostic procurements under the responsibility of Europe, which also contains Equatorial Port Plug 1 and seven other diagnostics supplied by Europe or other ITER partners and integrated in the same Port Plug. Significant progress has recently been made in the development of the equatorial visible/infrared wide angle viewing system and the radial neutron camera. This paper gives an overview of the major technical achievements on these two diagnostics and points out the urgent needed R&D activities.

Thermo-hydraulic analysis of the equatorial port plug (EPP) is currently being performed by CIEMAT within the framework of the engineering activities related to the ITER port plug diagnostic integration launched by EFDA. This study is focussed on calculating the main hydraulic parameters for the reference cooling circuit, and analyzing different piping configurations, in order to achieve a balanced circuit, this being a key point for obtaining an homogeneous refrigeration of the port plug. It also analyzes the cooling requirements for the Diagnostic Shield Modules (DSM), considering not only the reference circuit design, where the piping cooling circuits run through those parts that need to be cooled, but also a new design proposal which entails a port plug cooling system based on the thermal contact between the DSM and the port plug itself. The overall study represents not only an improvement in the equatorial port plug cooling system design, but also a better understanding of the whole system so as to foresee the weak points that will need to be taken into account in the subsequent detailed design.

The ITER Vacuum Vessel has upper, equatorial and lower port structures. The bottom ports are dedicated to the divertor replacement (five ports) and to vacuum pumping by means of cryopumps (four ports). The latest cryopump port design is more complex as it has a pump with a direct view of the vessel (upper cryopump) and a second pump at the end of a branch port (lower cryopump). 3D neutronic analyses have been performed in order to study the radiation conditions in and around the port system. In detail, nuclear heating on the cryopump has been calculated updating previous analysis performed in 2003 [L. Petrizzi, ITER CTA Detailed Neutronic Analyses, Final Report on contract EFDA/01-633 ENEA ref NE-VV-R-001 April 2003. Also included in Nuclear Analyis Report NAR ITER ref document G 73 DDD 2W 0.2 (v2.0) March 2006]. Calculations have been performed by means of MCNP 5 Monte Carlo code supplied with FENDL 2.1 library. In this work a new 40° model of ITER has been used in which full details of the cryopump system and remote handling ports have been included as well as the updated divertor components. The paper will present the neutronics results. They consist of nuclear heating on cryopump components; a map of dpa and helium production is provided as well. Gamma doses after shutdown have been calculated around the port flange to have an idea of the possible dose to which the eventual operator will be subject and to plan adequately manual operations. The cryopump is located at a distance of almost 5 m from the mouth of the divertor port and it is 3 m long. Calculations of such deep penetration problem are very challenging require special variance reduction techniques with Monte Carlo codes in order to use in an efficient way the computer resources. These will be described.

In the present study a complete neutronic analysis has been performed for the current design of the JT-60SA toroidal field coil (TFC) system. The MCNP5 Monte Carlo code has been used to calculate the nuclear heating, neutron spectra and absorbed dose in the TFC components, assuming a DD neutron emission rate of 1.5 × 1017 n/s (and 1% DT). Nuclear heating of the winding pack is lower than 0.3 mW/cm3 and the maximum nuclear heating of the TFC case is 0.4 mW/cm3. The overall nuclear heating, including the safety margin, is less than 8 kW. Spatial distribution of the nuclear heating has been provided along poloidal, radial and toroidal directions as to be used for thermo-hydraulic analysis and the design of TFC system. The absorbed dose to insulator is as low as to avoid the replacement during the whole life of the machine. Neutron fluxes have been used as input for a preliminary activation analysis performed with FISPACT inventory code. Activity and contact dose rates have been calculated at different cooling times, after 10 years of operations in some representative zone of the winding pack and the case. All the TFC materials can be easily recycled within the first day after shutdown and the hands-on recycling is possible within less than 30 years.

The radial neutron camera (RNC) will provide the spatial distribution and the total strength of the ITER neutron source (emissivity profile and fusion power) by means of collimated neutron measurements. Line-integrated neutron spectral measurements can also provide information on the ion temperature profile. The present design of the RNC consists of two collimating structures for a full coverage of the plasma: 36 collimated lines of sight (LOS) distributed in three different planes view the plasma core (ex-port system) and nine collimated LOS view the plasma edge (in-port system). The RNC design is based on the combined use of the MCNP Monte Carlo code and a software tool performing asymmetric Abel inversion of simulated measured neutron signals (MSST). Neutron and γ-ray transport calculations are performed with MCNP using a 3D RNC model to determine the signal/noise for each RNC channel and the spectra at the detectors. The MSST code is used to check the RNC compliance with the ITER measurement requirements for the neutron emissivity profile. In the present paper the improvement of the hard variance reduction technique applied to the MCNP neutron source (consisting in sampling neutrons only from plasma regions contributing to the detector signal) is presented and the following issues are analyzed: the possibility of reducing the length of the ex-port collimators (resulting in a significant reduction of the overall RNC dimension and weight); options for the reduction of the dose due to the neutron streaming through the RNC cut-outs in the blanket shielding module; the integration of a γ-ray detection system in the RNC by partially filling the collimators with a neutron absorbing material (LiH).

The upgrade of JT-60U to JT-60 Super Advanced (JT-60SA), a fully superconducting tokamak, will be performed in the framework of the Broader Approach (BA) agreement between Europe (EU) and Japan. In particular, the Toroidal Field (TF) system, which includes 18 coils, is foreseen to be procured by France, Italy and Germany. This work covers activities from design and manufacturing to shipping to Japan. The present paper is mainly devoted to the analyses that lead to the conductor design and to the technical specifications of the joints for the JT-60SA TF coils. The conductor geometry is described, which is derived from Cable-In-Conduit concept and adapted to the actual JT-60SA tokamak operating conditions, principally the ITER-like scenario. The reported simulations and calculations are particularly dealing with the stability analysis and the power deposition during normal and off-normal conditions (AC losses, nuclear heating). The final conductor solution was selected through a trade-off between scientific approach and industrial technical orientation. Besides, the TF system connections layout is shown, derived from the industrially assessed twin-box concept, together with the associated thermo-hydraulic calculations ensuring a proper temperature margin.

Neutrons produced by D–D and D–T plasmas induce the activation of tokamak materials and components. The development of reliable methods to assess dose rate is a key issue for maintenance and operation of nuclear machines, in normal and off-normal conditions. In support of ITER design activity an innovative computational tool based upon MCNP-4C Monte Carlo code has been developed to predict the dose rate after shutdown: it is called direct one step method (D1S). In the D1S approach the decay gammas are coupled to neutrons as in the prompt case and are transported in a single step in the same run. To check the capability of D1S method to predict dose rate in a tokamak a dedicated benchmark experiment was proposed for the 2005–2006 experimental campaign of JET. Two irradiation positions were selected for the benchmark: one inner position close to vessel and not far from plasma, called the two upper irradiation end (IE-2), while the second position is just outside a vertical port in an external position (EX). Passive detectors are used for in-vessel measurements: the high sensitivity thermo-luminescent dosimeters (TLDs) GR-200A (natural LiF), while one active detector of Geiger–Mueller (GM) type is used for out of vessel dose rate measurement. Both TLDs and GM detectors were calibrated in a secondary gamma-ray standard facility in terms of air-kerma. In the present work, residual background and dose rate due to early phase of JET 2005–2006 campaign are compared with the same quantities calculated using D1S approach. The impact of geometrical model and materials impurities to the calculated dose rates is discussed as well. The activity has been performed in the frame of the European Technological Tasks for JET JW5-FT5.20, in collaboration with United Kingdom Atomic Energy Agency (UKAEA) and Forschungszentrum Karlsruhe (FZK).

This paper summarizes the work, performed in the frame of various EFDA contracts during 2004–2005, on the design review and upgrade of the ITER radial neutron camera (RNC). The RNC, which should provide information on the spatial distribution and energy spectrum of the neutron emission, consists of an ex-vessel system (fan-like collimator with 12 × 3 lines of sights) and an in-vessel system with further 9 lines for a full coverage of the plasma. A Monte Carlo code (MCNP) has been used for the neutronic calculations. The basic ITER model has been developed from the CATIA drawings to include the RNC with all details relevant for the neutronic analysis. In the model the collimator diameters have been set to 2 and 4 cm, respectively, for the ex-vessel and in-vessel systems. A detailed space dependent fusion neutron source (DD and DT phases in various plasma scenarios) has been used with a consistent ion temperature radial profile. A special variance reduction treatment has been developed so that neutrons reach the far regions in the high collimated neutron beam and score with a satisfying statistical error. Neutron and photon fluxes and spectra have been calculated. Approximately, one neutron out of 1011 emitted in all the plasma reaches a single ex-vessel detector. Therefore, for an emission rate of 1.8 × 1020 n/s (corresponding to 500 MW fusion power) the flux on the detectors is in the range (1–5) × 108 n/(cm2 s) depending on the poloidal orientation. The fraction of scattered neutrons (>1 MeV) is lower than few % of the total. A measurement simulation software tool (MSST) performing asymmetric Abel inversion of simulated measured neutron signals has also been developed for line of sight and design optimization. Combining information from MCNP calculations and MSST, it has been possible to evaluate the performance of the RNC, check whether the present design of the RNC meets the measurement requirements and optimize the RNC design.

Engineering analysis has been carried out on a representative equatorial diagnostic port plug of ITER, outcome of which is given here. A preliminary overview of the work for a prototypical diagnostic upper port plug is also reported. To ensure that the port plug structure satisfies the requirements of the ITER environment, the following analyses have been performed: finite element (FE) analyses of static and dynamic behaviour under electromagnetic loads, FE vibration analysis, FE thermal-structural analysis and nuclear heating and radiation dose studies. The main outcomes from these analyses and consequent design developments are a revision of the stainless steel/water ratio in the plug neutron shielding modules and the incorporation of improved neutron shielding of the port duct, a modification of the port plug top plate arrangement, to increase torsional stiffness of the structure under disruption loads, and improved solutions for cooling arrangements. Manufacturing studies have also been performed, involving the assessment of methods for the production of plug flanges, the analysis of welding techniques for parts assembly and methods for the introduction of cooling features in plug components. Reference design solutions and possible modifications will be presented and discussed.

Neutronics and nuclear data have an essential role in the development programme leading to the realization of a fusion power plant. A well-qualified nuclear database and validated computational tools are required in the nuclear design of fusion devices for reliable neutronics and activation calculations with the associated uncertainties. The EU is conducting a large effort on neutronics and nuclear data for fusion applications. According to the fusion strategy, the efforts focus on the development and qualification of computational tools and nuclear data required for design analyses of ITER, in particular for the test blanket modules (TBM), and of IFMIF. A major effort is devoted to integral experiments with the objective to provide the experimental database for testing the nuclear data and validating neutronics design calculations. The paper provides an overview of the nuclear design activities conducted in the EU on the development of TBMs for ITER and the layout of material test modules for IFMIF. The focus is on the supporting neutronics efforts including the experiments for validating design calculations and nuclear data with associated uncertainties. The recent achievements will be presented and recommendations will be given for further development efforts on neutronics and nuclear data for fusion applications.

A key issue in the long-term programme of European Fusion Development Agreement (EFDA) is the experimental validation of the neutron-induced activation predicted by calculations for structural materials developed for fusion application. Up to now, experiments have been carried out at neutron generators producing mono-energetic 14 MeV neutrons and using experimental assemblies simulating reactor-like neutron flux spectra. Since in October 2003 a Tritium Trace Experiment (TTE) was performed at JET, two samples of EUROFER-97 were exposed to the real DT neutron source produced in the tokamak. The goal was to measure the activity induced by neutrons in this material and to compare it to that calculated using the European Activation System EASY-2003 with FENDL-2/A neutron activation cross-section library. This was the first attempt to compare experimental data obtained in a real DT neutron spectrum with calculation results. The conclusion of this work is that the EASY code and the FENDL-2/A can evaluate the induced activity of medium half-life isotopes with an acceptable uncertainty, within the experimental total error.

The calculation of dose rates after shutdown is an important issue for operating nuclear reactors. A validated computational tool is needed for reliable dose rate calculations. In fusion reactors neutrons induce high levels of radioactivity and presumably high doses. The complex geometries of the devices require the use of sophisticated geometry modelling and computational tools for transport calculations. Simple rule of thumb laws do not always apply well. Two computational procedures have been developed recently and applied to fusion machines. Comparisons between the two methods showed some inherent discrepancies when applied to calculation for the ITER while good agreement was found for a 14 MeV point source neutron benchmark experiment. Further benchmarks were considered necessary to investigate in more detail the reasons for the different results in different cases. In this frame the application to the Joint European Torus JET machine has been considered as a useful benchmark exercise. In a first calculational benchmark with a representative D-T irradiation history of JET the two methods differed by no more than 25%. In another, more realistic benchmark exercise, which is the subject of this paper, the real irradiation history of D-T and D-D campaigns conducted at JET in 1997-98 were used to calculate the shut-down doses at different locations, irradiation and decay times. Experimental dose data recorded at JET for the same conditions offer the possibility to check the prediction capability of the calculations and thus show the applicability (and the constraints) of the procedures and data to the rather complex shutdown dose rate analysis of real fusion devices. Calculation results obtained by the two methods are reported below, comparison with experimental results give discrepancies ranging between 2 and 10. The reasons of that can be ascribed to the high uncertainty on the experimental data and the unsatisfactory JET model used in the calculation. A new dedicated JET benchmark experiment will be performed trying to solve these issues.

In order to validate neutron cross sections for W, a benchmark experiment was carried out at the Frascati Neutron Generator (FNG), that consisted of the irradiation with 14 MeV neutrons of a tungsten block. Neutron flux and gamma heating were measured inside the block. The results were analysed with the Monte Carlo code MCNP-4C using for W, Fe and Ni the cross sections derived from EFF-2.4 and FENDL-2.0. In the neutron flux case, most of the calculations were in agreement with the experimental data within the total uncertainty using EFF-2.4, while they underestimated the fast neutron flux with increasing depth when using FENDL-2 cross sections. A strong discrepancy was found in the photon production data from the EFF-2.4 and FENDL-2 libraries, producing different values for the gamma heating in the two cases. FENDL-2 calculations showed better agreement with the measurements.

A neutronic analysis was performed to assess the activation properties of materials used in two long term conceptual reactors, i.e. the dual coolant liquid lead (DCLL) and the TAURO concepts. Both use advanced structural materials which are under development in the European fusion technology programme. Neutron fluxes were calculated in reference positions in the first wall, breeding blanket and shield for both reactors, that have been assumed to have the same reactor parameters. Activation calculations were performed for several, candidate materials irradiated in the selected reference positions, with different impurity levels: pure materials, real materials as available now, and materials with reasonable reduction of impurities after the necessary R&D. A large amount of data were produced on activation properties of these materials in the different positions and reactors (activity, afterheat, contact, inhalation and ingestion dose, clearance index) from shutdown to 106 years of cooling time, with specification of dominant nuclides and production pathways. The paper presents a few examples of the results concerning, in particular activation properties of SiC/SiC and EUROFER.

Three different nuclear cross section data packages, largely used in fusion studies, were compared through the analysis of a shut-down dose rate experiment in a mock up of the ITER vacuum vessel. These packages are: the European libraries EFF-3.0 (transport) and EAF-2001 (activation), the Japanese libraries JENDL-FF (transport) and JENDL-3.2 (activation), and the international libraries FENDL-2/MC (transport) and FENDL-2/A (activation). The analysis was carried out using a rigorous two-step method using the MCNP transport code and the FISPACT inventory code, coupled by an automated routing of the MCNP neutron flux spectrum distributions to FISPACT and the FISPACT decay gamma source distributions to MCNP. Analysis was performed from shut down, i.e. immediately after the end of mock-up irradiation with 14 MeV neutrons, up to about three months of decay time. All results were found in good agreement with measurements: within 30% at 1 day, and within the ±12% uncertainty in the comparison in the relevant time interval from a few days up to 1 month after shut down. The differences detected between the data packages, of the order of 20% at maximum in the dose rate, were analysed in terms of neutron flux and spectrum, and of activation cross sections.

The ITER machine will be equipped with 6 torus Cryopumps (TCP) that are positioned in their housings (TCPH) and integrated into the cryostat walls at B1 level in the port cells. A comprehensive nuclear analysis of the Cryopump Ports #4 and #12 has been carried out by means of the MCNP-5 Monte Carlo code in a full 3-D geometry, providing guidelines for the design of the embedded components. Radiation transport calculations have been performed in order to determine the radiation field inside the Lower Ports, up the Port Cell: 3-D neutrons and gamma maps have been provided in order to evaluate the shielding effectiveness of the TCPHs. Nuclear heating induced by neutron and photons have been estimated on the TCP and TCPH to assess the nuclear loads during plasma operations. The shutdown dose rate in the maintenance area of the Lower Ports has been assessed with the Advanced D1S method to verify the design limits. © 2015 Elsevier B.V. All rights reserved.

Silicon carbide (SiC) in the form of ceramic matrix is a low activation structural material proposed for fusion reactors. Its development is pursued in the European Fusion Technology Program. A SiC block (457×457×711 mm3), borrowed from JAERI, was irradiated with 14 MeV neutrons at the FNG facility of ENEA Frascati. Activation reaction rates, neutron fluxes and spectra, as well as nuclear heating were measured in four selected experimental positions inside the block. The experimental analysis was performed using the Monte Carlo transport code MCNP-4C and point-wise cross sections derived from FENDL-2.0, EFF-2.4 and EFF-3.0 evaluated nuclear data files. Deterministic transport calculations were also performed using the discrete ordinates code DORT. The sensitivity and uncertainty analysis were performed as well using the SUSD3D code. Results indicate that calculation based on EFF-3.0 nuclear data file estimates the neutron flux and spectra with a reasonable uncertainty which is still lower than ±30% for all measured quantities.

A neutronics experiment has been performed at the 14 MeV Frascati Neutron Generator (FNG) to validate the calculations of shut down dose rates inside the ITER cryostat. A proper experimental set-up, in which a neutron spectrum is generated similar to that occurring in the ITER vacuum vessel, has been irradiated for sufficiently long time to create a level of radioactivity which has been followed, after shut down, by dosemeters for more than 2 months of cooling time. The experiment has been analysed using a rigorous, standard two-step method, i.e. using MCNP-4-B and FISPACT codes, and a new one-step method with an ad hoc modified version of MCNP code used in the nuclear analysis of ITER. FENDL-2 nuclear data libraries are used in both cases. The comparison between experiment and calculations shows a good agreement in both cases.

A new cassette concept based on a welded box structure made of stainless steel plates has been proposed for ITER FEAT. In this paper a summary of the analysis of the design is provided. Due to lack of space all the interesting multidiscipline items cannot be mentioned, but only the main results are presented. A complete report on the activity is in (Final report on Design of Welded Box Divertor Cassette ENEA Report FUS TN DI-NE-R-001, February 2001). The work, which has involved professionals coming from different institutes, shows that all the design requirements are fulfilled in normal operational conditions. The halo currents induce deformations close to the allowable limits. Therefore, some reinforcement of the cassette or some modifications of the attachment system to the Vacuum Vessel (VV) will be studied in the future design activities to reduce the displacement.

The international fusion materials irradiation facility (IFMIF) project deals with the study of an accelerator-based, deuterium–lithium source, producing high energy neutrons at sufficient intensity and irradiation volume to test samples of candidate materials for fusion energy reactors. IFMIF would also provide calibration and validation of data from fission reactor and other accelerator based irradiation tests. This paper describes the activity on neutron/gamma dosimetry (necessary for the characterization of the specimens’ irradiation) performed in the frame of the IFMIF conceptual design evaluation (CDE) neutronics tasks. During the previous phase (conceptual design activity (CDA)) the multifoil activation method was proposed for the measurement of the neutron fluence and spectrum and a set of suitable foils was defined. The cross section variances and covariances of this set of foils have now been used for tests on the sensitivity of the IFMIF neutron spectrum determination to cross section uncertainties. The analysis has been carried out using the LSL-M2 code, which optimizes the neutron spectrum by means of a least-squares technique taking into account the variance and covariance files. In the second part of the activity, the possibility of extending to IFMIF the use of existing on-line in-core neutron/gamma monitors (to be located at several positions inside the IFMIF test cell for beam control, safety and diagnostic purposes) has been studied. A feasibility analysis of the modifications required to adapt sub-miniature fission chambers (recently developed by CEA-Cadarache) to the high flux test module of the test cell has been carried out. The verification of this application pertinence and a gross definition of the in-core detector characteristics are described. The option of using self-powered neutron detectors (SPNDs) is also discussed.

The experimental validation of the available codes and nuclear data to calculate correctly the effects of streaming paths in the shield of a fusion reactor was provided by a streaming experiment carried out at the 14 MeV Frascati Neutron Generator (FNG). The experiment consisted in the irradiation of a mock-up of the shielding blanket (stainless steel and water equivalent material, 1 m thick) and of the toroidal field coil. A void channel with high aspect ratio and a cavity at the end of the channel were realised in the bulk shield. Some nuclear quantities were measured in the cavity at the end of the channel and behind it up to the simulated superconducting magnet. The neutron flux was measured using the activation foil technique while the nuclear heating was measured using thermoluminescent detectors. The experimental data were compared with the results of calculations performed by the Monte Carlo code MCNP-4B, using the Fusion Evaluated Nuclear Data Library (FENDL, version 1.0, 2.0) and the European Fusion File (EFF, versions 3.0, 3.1). The comparison shows that all the used libraries reproduce the neutron flux generally within ±20% total uncertainty both in the void penetration and in the bulk shield behind it. The trend to underestimate the high energy flux ( E >8 MeV) with increasing penetration depth is found again as in previous bulk shield experiment, showing that the relevant transport cross sections at around 14 MeV need further investigation. The comparison between experimental and calculated nuclear heating shows an agreement within ±30% for all tested nuclear data libraries and in all experimental positions. The comparison with the case of bulk shield without penetration is also discussed.

A co-ordinated international benchmark validation task has been conducted to validate the Fusion Evaluated Nuclear Data Library FENDL-1 through data tests against integral 14-MeV neutron experiments. The main objective of this task was to qualify the FENDL-1 working libraries for fusion applications and to elaborate recommendations for further data improvements. A large variety of existing integral 14 MeV benchmark experiments has been analysed with the FENDL-1 working libraries for continuous energy Monte Carlo and multigroup discrete ordinates calculations. This paper reviews the FENDL-1 benchmark analyses and gives an overview on the major findings, conclusions and recommendations for further data improvements to be included in the forthcoming FENDL-2 data library. 21 refs., 3 tabs.

In support of ENEA activity in the European Community Test Programme, neutron analysis has been performed on the two latest blanket designs: helium-cooled ceramic breeder-inside-tube (BIT) (with LiAlO2 and Be multiplier) and water-cooled liquid Li17Pb in cylindrical modules (CM). The powerful MCNP Monte Carlo code was used (version 4.2). A detailed and accurate description of the geometrical model has been performed by inserting the main reactor details and avoiding breeder material dilution inside the modules. The tritium breeding ratio (TBR) performance is low for the solid breeder BIT blanket (with 10 ports 1.011) due mainly to low blanket coverage near the exhaust duct, and this solution should be revised. The CM Li17Pb blanket reaches a sufficient TBR (1.059, with ports) to rely on tritium self-sufficiency. Shielding properties, with respect to the toroidal field coils, have been estimated in a simplified model by means of the ANISN code, supplied with a nuclear data library consistent with that used by MCNP. The analysis suggests that a careful shield thickness/composition design should be used to ensure the shielding capability of the whole blanket plus shield system.

In the framework of the European Technology Program for NET/ITER, ENEA (Ente Nazionale per le Nuove Tecnologie, l'Energia e l'Ambiente), Frascati and CEA (Commissariat a l'Energie Atomique), Cadarache, are collaborating on a bulk shielding benchmark experiment using the 14 MeV Frascati Neutron Generator (FNG). The aim of the experiment is to obtain accurate experimental data for improving the nuclear database and methods used in the shielding designs, through a rigorous analysis of the results. The experiment consists of the irradiation of a stainless steel block by 14 MeV neutrons. The neutron flux and spectra at different depths, up to 65 cm inside the block, are measured by fission chambers and activation foils characterized by different energy response ranges. The γ-ray dose measurements are performed with ionization chambers and thermo-luminescent dosimeters (TLD). The first results are presented, as well as the comparison with calculations using the cross section library EFF (European Fusion File).

Depending on the final decision on the operation time of ITER (International Thermonuclear Experimental Reactor), the Driver Blanket might become a basic component of the machine with the main function of producing a significant fraction (close to 0.8) of the tritium required for the ITER operation, the remaining fraction being available from external supplies. The Driver Blanket is not required to provide reactor relevant performance in terms of tritium self-sufficiency. However, reactor relevant reliability and safety are mandatory requirements for this component in order not to significantly affect the overall plant availability and to allow the ITER experimental program to be safely and successfully carried out. Within the framework of the ITER Conceptual Design Activities (CDA, 1988–1990), a conceptual design of the ITER Driver Blanket has been carried out by ENEA Fusion Dept., in collaboration with ANSALDO S.p.A. and SRS S.r.l., and in close consultation with the NET Team and CFFTP (Canadian Fusion Fuels Technology Project). Such a design has been selected as EC (European Community) reference design for the ITER Driver Blanket. The status of the design at the end of CDA is reported in the present paper.

As a part of the FAST (Fusion Advanced Studies Torus) project a neutronic analysis has been performed aimed to design optimization and radiological safety assessment. The neutron emissivity source foreseen for various FAST scenarios has been calculated and used as input for Monte Carlo calculations. The shielding analysis and nuclear heating calculations have been carried out with MCNP5 using a detailed 3-D model of the machine. The energy and spatial distributions of neutron fluxes have been used to perform activation analysis by means of the FISPACT code for safety assessment. The implications of the results on the design of the machine and on safety issues are presented and discussed.

The DEMO specifications defined so as to ensure the consistency of the various blanket conceptual design studies performed within the framework of the European Test Blanket Program me have been recently updated. A very first attempt has been made to adapt the European Ceramic Breeder Inside-Tube DEMO blanket to these new specifications. Two solutions have been investigated. The first would ensure tritium self-sufficiency of the plant with a large safety margin. The other one, which fully preserves the design simplicity and reliability of the initial design, appears to be somewhat marginal from the tritium breeding capability point of view, but to offer good improvement prospects.

A blanket option featured by helium cooling and lithiated ceramics inside tubes is under development at CEA/ENEA within the framework of the European Test Blanket Program. A first conceptual design...

The helium-cooled lithium lead (HCLL) test blanket module (TBM) is one of the European blanket concepts selected to be tested in ITER. The test is an important step towards the development of DEMO blankets. Nuclear analysis of the TBM has been performed using the MCNP-4C Monte Carlo code, supported with FENDL-2 nuclear data library. Neutron fluxes were calculated with MCNP in LiPb breeder units and in the EUROFER cooling plates in various positions inside the module for a 500-MW fusion power. The activation calculations were performed using FISPACT (EASY 2005.1 package): activity, decay heat and contact dose rate were calculated inside the TBM using as input the neutron fluxes calculated by MCNP. The results of the activation analysis in terms of activity, decay heat and dose rate at different times since shut-down till 10 6 years after at various radial, poloidal and toroidal positions are presented. The effect of impurities on LiPb and EUROFER as well as the impact of the different irradiation scenarios on the activation of TBM has been investigated and discussed.

A multichannel neutron detection system (Radial Neutron Camera, RNC) will be installed on the ITER equatorial port plug 1 for total neutron source strength, neutron emissivity/ion temperature profiles and nt/nd ratio measurements [1]. The system is composed by two fan shaped collimating structures: an ex‐vessel structure, looking at the plasma core, containing tree sets of 12 collimators (each set lying on a different toroidal plane), and an in‐vessel structure, containing 9 collimators, for plasma edge coverage. The RNC detecting system will work in a harsh environment (neutron fiux up to 108–109 n/cm2 s, magnetic field >0.5 T or in‐vessel detectors), should provide both counting and spectrometric information and should be flexible enough to cover the high neutron flux dynamic range expected during the different ITER operation phases. ENEA has been involved in several activities related to RNC design and optimization [2,3]. In the present paper the up‐to‐date design and the neutron emissivity reconstruct...

In the frame of ITER (International Tokamak Experimental Reactor) design program, 3D neutronics calculations have been carried out to assess the system shielding performances in the basic machine configuration by means of Monte Carlo Neutron Photon (MCNP) code (3-B version). The main issue concerns the estimation of the nuclear heat and radiation loads on the toroidal field superconducting coils. “Self generated weight windows” (w.w.) and source biasing technique have been used to treat the deep penetration through the bulk shield and the streaming through the system gaps and openings. The main results are reported together with a discussion of the computing methods, especially of the variance reduction techniques adopted.

Nuclear design for ITER having an impact on components, equipment, and materials, must conform to established requirements, and must be fully traceable to valid data, and capable of withstanding detailed technical reviews. According to the Management and Quality Program that is being established in the ITER project, nuclear data used for neutronic calculations must be verified and validated prior to use in ITER design. Extensive work of validation and verification FENDL-1 and FENDL-2.0 libraries for neutronic calculations was performed during the ITER R&D activities throughad hocbenchmark experiments performed on mock-ups of ITER components at 14MeV neutron generators. In particular, this paper presents the validation of the latest version FENDL-2.1 using the existing benchmark experiments carried out at the FNG 14 MeV neutron generator at ENEA Frascati.

In the ITER (International Thermonuclear Experimental Reactor, [1]) project calculations of the dose rate after shutdown are very important and their results are critical for the machine design. A new method has been proposed [2] which makes use of MCNP [3] also for decay gamma-ray transport calculations. The objective is to have an easy tool giving results affected by low uncertainty due to the modeling or simplifications in the flux shape assumptions. Further improvements to this method are here presented.

In the frame of the studies developed for DEMO breeding blankets, a type of blanket with Pb17Li as breeder and water as coolant has been developed at the JRC-Ispra. It deals with breeder circular modules, internally cooled by water in U-type tubes and arranged poloidally according to rows inside boxes, named segments. A 3-D neutronic analysis of the DEMO reactor with this blanket configuration has been performed, to verify if tritium breeding self-sufficiency could be achieved without the need of additional multipliers. The DEMONET specifications have been applied, and the MCNP-3B Monte-Carlo code has been used. A careful modelling of the complicated blanket layout has been carried out, avoiding where possible homogenization procedures. A total Tritium Breeding Ratio (TBR) of 1.152 is obtained as a final result of the MCNP run. If the presence of the ports is considered, a net TBR of 1.09 is achieved. It is therefore verified that this blanket permits to achieve self-sufficiency without any additional multiplier. The results of this analysis are critically compared with previous ones, which used 1-D models and approximated expressions to estimate the net TBR. A good agreement between the two studies is found out.

The present study concerns the most recent neutronic analyses of two design versions of the european ceramic B.I.T. blanket, jointly developed by ENEA and CEA since few years. The last year developments required a new 3-D geometry evaluations of the global TBR. The results indicated that the ENEA version reaches a global TBR value of 1.13. The CEA version, in a 3-D model using a simplified description of the breeder module layout, reaches a TBR value of 1.12. Nuclear heat deposition density has been determined for all blanket components as a function of the poloidal co-ordinate. Shielding properties of this type of blanket have been analyzed

The radioactive waste Ingestion Toxic Potential IT for neutron activation produced radionuclides (Starfire conditions) has been calculated up to 106 years' decay time for several LAM materials assumed to be potentially used in a fusion reactor. The results have been compared with the IT calculated for toxic heavy metals contained in the burnt coal or fuel oil of thermal power plants. The use of LAM gives a lower waste environmental impact than coal power plants and, after 2–5 years' decay time, than fuel oil thermal power plants.

The advanced ENEA poloidal solid breeder-in-tube (BIT) blanket design for a DEMO fusion reactor is presented. Its main operating conditions and features are max. neutron fluence = 3×1022 n/cm2; max. beryllium vol. swelling = 4%; max. beryllium differential swelling = 3%; module poloidal length = 8.5 m; helium inlet/outlet temperature = 250/500°C; pumping power percentage >4%; max. structure temperature >550°C; ceramic breeder min/max. temperature 450/750°C; purge circuit separated from the coolant circuit; coaxial manifolding system; 3-D value of the t.b. r ≈ 1.20; and thermal cycle efficiency 35%.

Helium-cooled ceramic breeder blanket designs are commonly based on the use of beryllium as both multiplying and moderating material. The possibility of using lead as multiplier and graphite as moderator, instead of beryllium, is investigated by means of a proper optimization code, aiming at maximizing the tritium breeding ratio (TBR) vs the material composition under the constraints of the thermal-hydraulic dimensioning. For a fixed value of the volume percentage of voids (25%, including helium coolant), and for upper and lower limits of ceramic breeder ( 10%), respectively, an optimal blanket configuration is obtained with a 1-D TBR of 1.40 (lithium silicate) and 1.30 (lithium alluminate), with a 75% Li6 enrichment of the breeder material. Results of the optimization code together with their design implications are discussed in the paper. A conceptual design of the optimal blanket configuration is developed, starting from a poloidal breeder-in-tube scheme based on the ongoing ENEA helium-cooled blanket design for DEMO. The main operating conditions and features of the DEMO-relevant ceramic breeder lead/graphite blanket design are max. neutron wall load = 2.6 MW/m2; helium coolant inlet/outlet temperature = 250/500°C; max. structure temperature = 550°C; cooling pumping power percentage = 4%; and helium purge circuit entirely separated from the coolant circuit.

The European Union will supply a number of diagnostics for ITER. Significant progress has been made on the design and integration of these diagnostics. Specific topics, in common for several of the diagnostics, discussed are port-plug integration, simplification of labyrinths against neutron streaming, and design measures to protect and to be able to replace sensitive optical components. Performance analysis to predict the likely capability of the diagnostics to meet the ITER measurement requirements will be an important aspect of the design process. The interpretation of the measurement requirements, specifically of spatial resolution in the case of indirect measurements, is discussed on the basis of two examples, and methods of performance analysis are compared. (c) 2006 American Institute of Physics.

A conceptual study of a helium-cooled solid breeder blanket for a tokamak reactor is presented. Tritium breeding capability together with system reliability are taken as the main design criteria. The blanket consists of tubular poloidal modules made of a central bundle of ceramic rods (γLiAlO 2 ) with a coaxial distribution of the inlet/outlet coolant flow (He) surrounded by a multiplier material (Be) in the form of bored bricks. The Be to γLiAlO 2 volume ratio is 4/1. The He inlet and outlet branches are cooling Be and γLiAlO 2 , respectively. A purge He flow running through small central holes of the ceramic rods is derived from the main flow. Under the typical conditions of a tokamak reactor (neutron wall load = 2 MW/m 2 ), a full coverage tritium breeding ratio of 1.47 is achieved for the following design and operating parameters: outlet He temperature = 570 °C; inlet He temperature = 250 °C; total extracted power = 2700 MW; He pumping power percentage = 2%; minimum/maximum γLiAlO 2 temperature = 400/900 °C; maximum structural temperature = 475 °C; and maximum Be temperature = 525 °C.

An integral bulk shield experiment has been performed at the Frascati Neutron Generator (FNG) with the main objective to validate the ITER shielding system. To this end, a suitable mock-up of the ITER inboard shielding system including first wall, shielding blanket, vacuum vessel and toroidal field coil, had been assembled at ENEA Frascati. Neutron and γ-ray spectra as well as various nuclear responses were measured inside this mock-up when irradiated by 14-MeV neutrons. Measured and calculated data were compared to validate transport codes and nuclear data used in the design of the ITER system. In particular, the measured neutron and γ-spectra, reaction rates and the nuclear heating were compared with the same quantities calculated with the MCNP Monte Carlo code using the Fusion Evaluated Nuclear Data Library (FENDL)-1.0 and the European Fusion File (EFF)-3.0 neutron cross section data. The neutron-induced activation measured in stainless steel was compared with calculations performed with the FISPACT code and using the most recent activation cross section data file FENDL/A-2.0. The paper includes a comparison of calculated over measured quantities (C/E) and a discussion of their relevance to the ITER nuclear design. It is found that, generally, the measured nuclear quantities are predicted by FENDL-1.0 and EFF-3.0 calculations within an uncertainty margin of about ±15% in the shielding blanket and vacuum vessel, and ±30% up to the region of the toroidal field coil.