• Energy Research

IFMIF-DONES (International Fusion Materials Irradiation Facility- DEMO Oriented NEutron Source) is a neutron irradiation facility aiming at providing material irradiation data for the construction of DEMOnstration fusion power plants. This work presents nuclear analyses and associated neutronics modeling conducted on an updated design of the High Flux Test Module (HFTM) of IFMIF-DONES. Neutron fluxes, damage dose rates and gradients, gas production and nuclear heating have been calculated for the standard profile in 20 × 5 cm$^{2}$ and a reduced size in 10 × 5 cm$^{2}$ of the deuteron beam with 40 MeV energy and 125 mA current. The results indicate that the reduced beam can provide higher damage dose rates fulfilling the DONES requirement. On the other hand, the reduced beam size results in less irradiation uniformity, higher gas production to damage dose ratios, as well as stronger peak nuclear heating.

IFMIF-DONES (International Fusion Materials Irradiation Facility - DEMO-oriented Neutron Source) is currently being developed in the frame of the EUROfusion Early Neutron Source work package (WPENS), based on the results achieved in the ongoing IFMIF/EVEDA (Engineering Validation and Engineering Design Activities) project by Japan and Europe in the frame of the Braoder Approach (BA) agreement. The neutron source's limited product of “irradiation volume × neutron fluence” on one hand and the large potential test matrix (defined by number of material grades, test types, irradiation- and test temperature levels and damage dose levels) on the other hand, require a careful selection of test conditions by collaborative effort of the communities of DEMO designers, fusion materials science and the irradiation facility designers. This paper describes achievable irradiation conditions in the High Flux Test Module (HFTM) of IFMIF-DONES, characterized by levels and gradients of irradiation temperatures and nuclear responses (dpa, H and He production). The HFTM development focuses on 9%-Cr Reduced Activation Ferritic Martensitic steel irradiations, but also the possibilities of tungsten and copper alloy irradiations are explored. Possible specimen arrangements and issues of the Small Specimens Test Technique are also discussed.

Abstract A new approach is proposed for the calculation of non-equilibrium triton energy distributions in nuclear reactions induced by nucleons of intermediate energies. It combines models describing the nucleon pick-up, the coalescence and the triton knock-out processes. Emission and absorption rates for excited particles are represented by the pre-equilibrium hybrid model. The model of Sato, Iwamoto, Harada is used to describe the nucleon pick-up and the coalescence of nucleons from exciton configurations starting from (2p,1h) states. The contribution of the direct nucleon pick-up is described phenomenologically. Multiple pre-equilibrium emission of tritons is accounted for. The calculated triton energy distributions are compared with available experimental data.

The challenge for any nuclear data evaluation project is to periodically release a revised, fully consistent and complete library, with all needed data and covariances, and ensure that it is robust and reliable for a variety of applications. Within an evaluation effort, benchmarking activities play an important role in validating proposed libraries. The Joint Evaluated Fission and Fusion (JEFF) Project aims to provide such a nuclear data library, and thus, requires a coherent and efficient benchmarking process. The aim of this paper is to present the activities carried out by the new JEFF Benchmarking and Validation Working Group, and to describe the role of the NEA Data Bank in this context. The paper will also review the status of preliminary benchmarking for the next JEFF-3.3 candidate cross-section files.

An acceptable criterion for strong sustainability in the consumption of natural resources is an effective, or virtual, limitlessness of supply, which can be defined, albeit arbitrarily, as corresponding to a few million years. The fuels for nuclear fusion ‐ lithium and deuterium ‐ satisfy this condition because of the abundance of lithium in seawater and of deuterium in all forms of water. The possible use of lithium-ion batteries on a large scale, particularly in the automobile industry, could, however, use up all the known terrestrial reserves and resources of lithium in the next few decades. Little attention has been paid so far to the financial, energetic, and above all, environmental aspects of lithium extraction from seawater. The neutron multipliers foreseen for fusion power plants, in particular beryllium, represent a major supply problem and require that other, sustainable solutions be urgently sought.

During the Engineering Validation and Engineering Design Activities (EVEDA) phase (2007-2014) of the International Fusion Materials Irradiation Facility (IFMIF), an advanced engineering design of the High Flux Test Module (HFTM) has been developed with the objective to facilitate the controlled irradiation of steel samples in the high flux area directly behind the IFMIF neutron source. The development process addressed included manufacturing techniques, CAD, neutronic, thermal-hydraulic and mechanical anal- yses complemented by a series of validation activities. Validation included manufacturing of 1:1 parts and mockups, test of prototypes in the FLEX and HELOKA-LP helium loops of KIT for verification of the thermal and mechanical properties, and irradiation of specimen filled capsule prototypes in the BR2 test reactor. The prototyping activities were backed by several R&D studies addressing focused issues like han- dling of liquid NaK (as filling medium) and insertion of Small Specimen Test Technique (SSTT) specimens into the irradiation capsules. This paper provides an up-todate design description of the HFTM irradiation device, and reports on the achieved performance criteria related to the requirements. Results of the vali- dation activities are accounted for and the most important issues for further development are identified.

Displacement cross-sections for an advanced assessment of radiation damage rates were obtained for a number of materials using the arc-dpa model at neutron incident energies from 10−5 eV to 10 GeV. Evaluated data files, CEM03 and ECIS codes, and an approximate approach were applied for the calculation of recoil energy distributions in neutron induced reactions. Three sets of displacement cross-sections based on the use of low-energy data from JEFF-3.3, ENDF/B-VIII.0, and JENDL-4.0u were prepared. Files contain also cross-sections calculated using the standard NRT model. Special efforts were made to estimate the uncertainty of obtained displacement cross-sections.