European Fusion File Research Articles

Measurement of Flux Spectra and Tritium Production Rates in an ITER TBM Mock-Up Irradiated with 14-MeV Neutrons

A mock-up of the European Helium-Cooled Pebble Bed TBM was irradiated with DT neutrons in pulsed and continuous mode at the Fusion Neutronics Laboratory of the University of Technology Dresden. The aim was to measure fast neutron and gamma-ray flux spectra as well as time-of-arrival spectra of the slow neutron flux. The results of the experiments were analysed by the Monte Carlo code MCNP and nuclear data from the European Fusion File (EFF-3), and the Fusion Evaluated Nuclear Data Library (FENDL-2.0/2.1). It was found that the calculation of the fast neutron flux above 3 MeV tends to overestimate while the gamma-ray flux and slow neutron flux in two measurement positions in the mock-up was underestimated. The mock-up was also irradiated at FNG/ENEA Frascati to measure tritium breeding rates by means of small Li2CO3 pellet detectors inserted into the breeding layers. The breeding experiment was analysed at FZ Karlsruhe with emphasis on determining sensitivities of the TPR to relevant cross section uncertainties of all materials in the mock-up. It was found that the TPR calculation shows a tendency to underestimate. From the sensitivity analysis it was found that the total TPR is most sensitive to the elastic scattering in Be and the 7Li(n,αT) reaction.

The EFF project status and the NEA nuclear data services

The OECD Nuclear Energy Agency (NEA) Data Bank participates in the production of data and their distribution to users including the collection and validation as well as the distribution of the Joint Evaluated Fission and Fusion (JEFF) library. The JEFF project has evolved from two separate projects, namely the European Fusion File (EFF) and the Joint Evaluated File (JEF), with the release of JEFF-3.1, in May 2005. The NEA also provides tools for the EFF project, such as computer codes for nuclear energy and radiation physics applications. Of special interest for fusion applications are the integral experiments collected in the SINBAD database, with radiation shielding and dosimetry experiments including many fusion neutronics shielding experiments. In this paper an overview will be given of the NEA, and some examples of nuclear data services offered, such as the SINBAD database. JANIS, a display program evaluated and experimental data developed at the NEA will be mentioned briefly. The main emphasis will be given to the EFF project including the contents of the new JEFF-3.1 library, released in May 2005. Examples of recent work are given as well as a discussion on the forth-coming evaluation efforts among the EFF collaborators.

International comparison of measuring techniques of tritium production for fusion neutronics experiments: Status and preliminary results

A neutronics experiment has been launched in the European fusion technology program on a mock-up of a test blanket module with the objective to validate the capability of nuclear data to predict nuclear responses with qualified uncertainties. In the experiment, the mock-up will be irradiated with a 14-MeV neutron source and the tritium production rate will be measured and compared with the calculated value by neutron cross sections from the European Fusion File. Given the difficulty of performing a reliable measurement of tritium production rate in such experiments, an international benchmark comparison on tritium measurement techniques has been carried out in collaboration between ENEA, TUD/VKTA and also JAERI that has already performed such experiments. The main purpose of the comparison is to establish the reference technique of TPR measurements to be used later in the TBM experiment. Three steps of the comparison procedure are presented in the paper as well as the preliminary results. The implications for the TBM experiment are also discussed.

EU Blanket Design Activities and Neutronics Support Efforts

An overview is provided of the design activities and the related neutronics support efforts conducted in the European Union for the development of breeder blankets for future fusion power reactors. The EU fusion programme considers two blanket lines, the Helium-Cooled Pebble Bed (HCPB) blanket with Lithium ceramics pebbles (Li4SiO4 or Li2TiO3) as breeder and beryllium pebbles as neutron multiplier, and the Helium-Cooled Lithium-Lead (HCLL) blanket with the Pb-Li eutectic alloy as breeder and neutron multiplier. The blanket design and the related R&D efforts are based on the use of the same coolant and the same modular blanket structure to minimise the development costs as much as possible. The neutronic support efforts include design analyses for the layout and optimization of the modular HCPB/HCLL blankets based on detailed three-dimensional Monte Carlo calculations as well as underlying neutronics activities conducted in the frame of the European Fusion and Activation File (EFF/EAF) projects to develop qualified nuclear data and computational tools for reliable neutronics design calculations.

A brief overview of the European Fusion File (EFF) Project

The European Fusion File (EFF) Project is a collaborative project with work funded by the European Fusion Development Agreement (EFDA). The emphasis is on the pooling of resources and removal of duplication of effort, leading to the efficient development of two types of nuclear data libraries for use in fusion power plant design and operation studies. The two branches consist of, on the one hand, a general purpose file for modelling and design capabilities and, second, an activation file for the calculation and simulation of dose rates and energy release during operation of a future power plant. Efforts are directed towards a continued improvement of the quality of the nuclear data needed for these analyses. The OECD Nuclear Energy Agency's Data Bank acts as the central repository for the files and all information discussed during twice yearly meetings. It offers its services at no charge to the Project.

Sensitivity and uncertainty analyses of 14 MeV neutron benchmark experiment on silicon carbide

A shielding experiment on a silicon carbide (SiC) composite, a potentially important low activation structural material for the fusion reactor, was performed at the 14 MeV Frascati Neutron Generator (FNG) in 2001 and analyzed using both deterministic and probabilistic transport, sensitivity and uncertainty methods. The objective of the experiment was to verify the nuclear cross section files, as well as the calculational methods, used in the design and shielding calculations for fusion reactor. Neutron activation reaction rates, neutron and photon flux spectra, and nuclear heating rates were measured at several positions inside the SiC block. Point-wise and group cross sections derived from the European Fusion File (EFF-2.4 and -3.0) and the Fusion Evaluated Nuclear Data Library (FENDL-2) were used in the analysis. Results indicate that the calculation based on the recent EFF-3.0 Si-28 nuclear data estimates the measured quantities with a reasonable uncertainty of ±(10–15)% for 1 σ.

Measurement and analysis of neutron and gamma-ray flux spectra in SiC

A silicon carbide assembly of a thickness corresponding to about 10 mean free path for 14 MeV neutrons was irradiated at the Frascati Neutron Generator and the neutron and gamma-ray flux spectra were measured simultaneously, using a NE 213 scintillation spectrometer at four positions inside the block. The computational analysis of the benchmark experiment for the advanced structural material of fusion devices was performed with the Monte Carlo code MCNP-4C and nuclear data taken from the Fusion Evaluated Nuclear Data Library FENDL/MC-2.0, except for 28Si, for which the new data evaluation of the European Fusion File EFF-3.0 was used. Measured and calculated flux spectra are compared and ratios of calculated-to-experimental values are derived for various energy ranges.

Multidimensional Deterministic Nuclear Data Sensitivity and Uncertainty Code System: Method and Application

In the nuclear industry, an increased demand exists to provide modeling results with credible confidence bounds not only for simple but also for refined modeling. With the objective to facilitate and promote such analysis, a user-friendly and complete computer code package was developed comprising the multidimensional cross-section sensitivity and uncertainty code package SUSD3D, the secondary angular distribution (SAD) covariance data-processing module ERRORR34, and the cross-section covariance matrix library ZZ-VITAMIN-J/COVA.The discrete ordinates sensitivity formulation of the first-order perturbation theory is implemented in the SUSD3D code. Much effort was devoted to mitigate some drawbacks of the discrete ordinates-based sensitivity analysis, in particular to allow the analysis of complex geometries and to reduce memory requirements. The SUSD3D code is linked via interface files to a wide range of state-of-the-art transport codes suitable for complex radiation transport and facility design studies (like ANISN, DOT-3.5, DANTSYS, DORT, and TORT) and supports the new cross-section and covariance data formats. The SAD and secondary energy distribution effects can be taken into account. The complete SAD covariance matrices, as available in the European Fusion File evaluations (EFF-2 and -3) can be treated. The computer codes are written in FORTRAN-77 and run under DOS (PC), UNIX (workstations), VMS (VAX), and other compatible operating systems.The code system is extensively used to study fission- and fusion-related problems. The validation and several practical applications of the package are presented, including the reactor pressure vessel surveillance uncertainty studies, and ASPIS-iron, VENUS-3, and FNG benchmark experiment analyses.

Neutron streaming experiment at FNG: results and analysis

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.

Absorbed dose calculations for the Ignitor tokamak magnet coils insulator

The Ignitor experiment is an advanced compact high magnetic field tokamak with cryogenically cooled (30 K) normal conductor magnets. Its main purpose is to produce deuterium–tritium plasma regimes where ignition can take place. From the neutronics point of view, the routine operations with 50% of tritium will lead to a very short (4 s) but intense emission of 14 MeV neutrons (up to 3×10 19 n/s). The capability of the boron-free glass reinforced epoxy resin presently adopted for the insulation of the coils conductor to withstand the severe irradiation condition could be a key issue in the performance, reliability and lifespan of Ignitor. In order to better know the radiation environment in which the insulators will operate, calculations of the absorbed dose are performed using the parallelized version of MCNP-4B Monte Carlo code. A detailed 3D geometry description of the tokamak, an accurate representation of the neutron source distribution together with the more recent version of transport cross-section libraries EFF (European Fusion File) are used in this work. The variation of calculated dose depending on the insulator position inside the device, on the neutron energy spectra and on the total number of neutrons produced over the scheduled machine lifetime are discussed in the present paper.

Neutronics experiment on a mock-up of the ITER shielding blanket at the Frascati Neutron Generator

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.

Benchmark Experiment on Vanadium Assembly with D-T Neutrons -Leakage Neutron Spectrum Measurement-

Benchmark experiments on vanadium and vanadium alloy with D-T neutrons have been done at two angles, 0 degrees and 24.9 degrees, using the slab geometry and the time-of-flight (TOF) method. Data were collected for neutron energies ranging from 50 keV to 15 MeV. For vanadium, measurements were made for three slab thicknesses, i.e., 50.8 mm, 1524 mm, and 254 mm, whereas for the vanadium alloy, measurements were made only for 101.6-mm thickness. The measured neutron spectra were compared with MCNP-4A calculations using evaluated nuclear data from the JENDL-3.2, JENDL Fusion-File(IENDL-FF), FENDL/E-1.0 and European Fusion File veraon-3(EFF-3) libraries. The calculated data show reasonable agreement with the measurement, however, some differences are worth noting. Calculations for a slab thickness of 50.8 mm over the energy range from 0.05 to 0.1 MeV underestimate the measurements by about 40% at an angle of 24.9 degrees, while calculations for the energy range from 0.1 to 1.0 MeV, overestimate the measurements by about 40% at an angle of 0 degrees. Calculations made using the JENDL-FF library show good agreement with measurements for energies greater than 11 MeV. Calculations made using the FENDL/E-1.0 library give smaller results than any of the other three libraries in the energy range from 5 to 11 MeV.

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.

Analysis of integral experiments on stainless steel shields for the validation of the FENDL library

The present paper describes the analysis of integral experiments on stainless steel (SS) and SS/water shielding assemblies, irradiated with 14-MeV neutrons. The analysis is performed with the Monte Carlo transport code MCNP-4A and using the Fusion Evaluated Nuclear Data Library (FENDL). The ratios of calculated over measured quantities are given and discussed for the validation of the FENDL library in view of its use in the design of the International Thermonuclear Experimental Reactor (ITER). The analysis is also done using the European Fusion File (EFF), versions 1 and 2, for comparison. The results show that these libraries can satisfactorily describe the features of the neutron and gamma radiation attenuation in large SS shields. Some trends are, however, found to underestimate the fast neutron flux ( E > ≈ 0.5 MeV) in deep locations, that may lead to a non-correct evaluation of damage parameters for structural components in the reactor design.

FENDL data testing for beryllium, lead, iron and copper

Fusion Evaluated Nuclear Data Library (FENDL) data testing has been performed for beryllium, lead, iron and copper using the multigroup data processed by the International Atomic Energy Agency Nuclear Data Section in the vitamin J group structure (175 neutron and 42 γ groups). The data-testing procedure comprises comparisons of processed data with evaluated data (including data from ENDF/B-VI and the European Fusion File) and measured data, comparisons of transport calculations using different codes ( onetran, antra1 and mcnp ) and data, and comparisons with simple and clean 14 MeV neutron transmission experiments providing measured neutron leakage spectra. It is shown that the analysed lead and copper data need only minor improvements, whereas there are still discrepancies and deficiencies in the beryllium and iron data which may be resolved with the help of integral experiments which are under way.

The bulk shielding benchmark experiment at the Frascati Neutron Generator (FNG)

Abstract In the design of next-step fusion devices such as NET/ITER the nuclear performance of shielding blankets is of key importance in terms of nuclear heating of superconducting magnets and radiation damage. In the framework of the European Fusion Technology Program, ENEA Frascati and CEA Cadarache in collaboration performed a bulk shielding benchmark experiment using the 14 MeV Frascati Neutron Generator (FNG), aimed at obtaining accurate experimental data for improving the nuclear database and methods used in shielding designs. The experiment consisted of the irradiation of a stainless steel block by 14 MeV neutrons. The neutron reaction rates at various depths inside the block have been measured using fission chambers and activation foils characterized by different energy response ranges. The experimental results have been compared with numerical results calculated using both S n and Monte Carlo transport codes and the cross-section library eff.1 (European Fusion File).

Experimental investigation of neutron and photon penetration and streaming through iron assemblies

Nuclear shield parameters of fusion reactors such as ITER are determined by the neutron and photon fluxes within and behind the shield. Iron is the main element of shields. In order to verify calculational tools for the blanket design a solid Fe slab of dimensions 100 cm × 100 cm × 30 cm in thickness was irradiated with 14 MeV neutrons, and the neutron and photon fluxes penetrating the assembly were measured in wide energy ranges. They are compared with calculated fluxes using the three-dimensional Monte Carlo code mcnp and nuclear data of the European Fusion File EFF-1. The calculation underestimates the neutron flux at low and at high energies. It is discussed in connection with the underestimated photon flux. The investigation was extended to Fe assemblies with a gap at several positions.

The stainless steel bulk shielding benchmark experiment at the Frascati Neutron Generator (FNG)

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 à 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).

Three-dimensional neutronic analysis of the European ceramic BIT DEMO blanket

The present study concerns the 3-D geometry evaluations of the tritium breeding ration of the breeder-in-tube helium-cooled ceramic blanket performed at CEA and ENEA in the framework of the recently released DEMONET specifications. For the calculations the nuclear data have all been derived by the European Fusion File (EFF-1), but different Monte Carlo codes, and different design variants and geometrical models have been used. A benchmark calculation has been defined in order to better analyse the results; such an exercise has shown that TBR differences of a few percent could be due to the calculation methods. The determined global TBR ranges from 1.06 in the CEA results to 1.15 in the ENEA results, the expected difference being mainly due to the greater neutron coverage of the ENEA variant of the design. Heat deposition density and He production in beryllium are also given.

Analyses of 14 MeV neutron transport in beryllium

The 14 MeVneutron transport in beryllium is analyzed on the basis of an S N -transport procedure with rigorous treatment of the neutron scattering and double-differential cross-sections (DDX-data) from the nuclear data files ENDF/B-VI and EFF-1 (European Fusion File). The latter one uses the Los Alamos (LANL) data evaluation for the beryllium cross-sections. Comparisons with conventional S N P l transport calculations in one- and two-dimensional geometry show that their quality depends on the way of representing and processing the DDX-data on the files. If the DDX-data are given in the Legendre approximation (EFF-1 file) rigorous S N and approximative S N P l transport calculations agree perfectly. If the DDX data are given in the more appropriate angular representation (ENDF/B-VI file), enabling the DDX data processing without making any use of the Legendre approximation, rigorous S N and approximative S N P l calculations disagree significantly. Calculations of integral beryllium experiments show that it is not possible at present to draw a final conclusion on the quality of the EFF-1 and ENDF/B-VI beryllium date evaluations although they show remarkable differences. There are, however, indications that the 14 MeV (n, 2n) cross-section in ENDF/B-VI is too low and that the secondary-energy distribution of the EFF-1 neutron emission cross-section at low secondary energies is too high. Further integral experiments, especially measurements of neutron leakage spectra, are needed for a reliable assessment of the quality of the beryllium data evaluations.