P. Batistoni

ITER Nuclear Analysis Strategy and Requirements

Abstract 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 interdependence within these categories. This paper reviews the status of the methods applied to date and recommends the future strategy which ITER should adopt to address the continuing requirements and responsibilities. This is done by consideration of the application of radiation transport codes, the creation of an ITER reference neutronics model, the application of a neutronics results database, and the management tools which will be required. Areas in which new codes and techniques need to be developed will be identified.

Absolute experimental and numerical calibration of the 14 MeV neutron source at the Frascati neutron generator

The absolute neutron yield of the 14 MeV Frascati neutron generator (FNG) is routinely measured by means of the associated alpha‐particle method with a silicon surface barrier detector (SSD). This paper describes the work carried out to characterize the neutron source in terms of absolute intensity and angle‐energy distribution of the emitted neutrons. The development of the measuring setup and the assessment of the measurement results are also reported. A complementary calibration procedure for validating the SSD results, based on the use of fission chambers and the activation technique, is also reported. An accurate analysis of the system has been performed via the Monte Carlo neutron and photon MCNP transport code. A detailed model of the neutron source that includes ion slowing down has been inserted into the MCNP code to permit a numerical calibration of the neutron source for comparison with the experimental results. The resulting agreement among the various methods is very good considering the unce...

On the possibility of neutron spectrometry for determination of fuel ion densities in DT plasmas

The fuel density (given by nd+nt and nd/nt) is one of the important parameters to determine and control in burning DT plasmas. Here, we propose that nd and nt be deduced from data on the dd and dt neutron energy distributions to be recorded with a neutron spectrometer having energy selective response to the plasma neutron flux. The potential diagnostic capabilities of the magnetic proton recoil spectrometer system are tested for the measuring conditions to be presented by the next step tokamaks such as ITER.

Neutronics experiment on a mock-up of the ITER shielding system at the Frascati neutron generator (FNG)

Abstract 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.

Triton burnup measurements in JET using a neutron activation technique

The confinement and slowing-down of fast tritons in JET deuterium plasmas have been investigated by measuring the 14.1 MeV to 2.45 MeV neutron production ratio which is a direct measure of the triton burnup fraction. The 14.1 MeV neutron production was measured by means of the activation of copper samples, and the 2.45 MeV neutron production was obtained from fission chambers. Data were obtained for different plasma conditions giving burnup fractions in the range of 5 × 10−3 to 13 × 10−3. Comparison with calculations based on classical triton confinement and slowingdown in the JET plasma shows that the data are generally 20% lower than calculated values but still within the overall uncertainties involved

Benchmark analysis of neutronics performances of a SiC block irradiated with 14 MeV neutrons

Abstract 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.

Experimental validation of shut down dose rates calculations inside ITER cryostat

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.

Benchmark Experiment for the Validation of Shut Down Activation and Dose Rate in a Fusion Device

A neutronics experiment has been performed at the 14 MeV Frascati Neutron Generator (FNG) and using a proper experimental set-up, with the objective to validate the calculations of shut down dose rate outside the ITER vacuum vessel. In the experiment, a shield mock up was irradiated for sufficiently long time that the level of induced activation was followed after shut down by dosemeters for a cooling time as required for allowing personal access. The measurements were analysed, using a rigorous, standard two-step method (MCNP transport / FISPACT inventory codes), and a new one-step method with an ad hoc modified version of MCNP code, developed within the ITER Project to provide a more straightforward tool for dose rate analysis outside the vessel. The results of the experiment are presented together with the analysis performed with both approaches and with different nuclear data.

Neutron streaming experiment at FNG: results and analysis

Abstract 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.

Design of the neutron multicollimator for Frascati tokamak upgrade

The design of a six‐channel one‐dimensional neutron/x‐ray camera (multicollimator) for the Frascati tokamak upgrade (FTU) machine is described only with regard to the aspects concerning the neutron measurements. The multicollimator, viewing a poloidal cross section of FTU, provides the radial profile of neutron emission from D–D plasmas. The ion temperature profile can be derived in ohmic plasmas if independent data on density are available, while energetic ion tails can be located in the presence of additional rf heating. Information on the position and shape of FTU circular plasmas can also be obtained. The constraints imposed on the collimation and shielding by the compactness of FTU and the expected neutron fluxes are analyzed in detail. The detection system and the performance of the multicollimator are also discussed.