The Accelerator Driven Systems, a 21st Century Option for Closing Nuclear Fuel Cycles and Transmuting Minor Actinides

Closing the nuclear fuel cycle and transmuting Minor Actinides (M.As) can be considered as an application of the duty of care principlel principle which says that, “before the final disposal of any waste, any possible chemical and/or physical treatment has to be applied in order to reduce the waste’s toxicity, provided the treatment does not convey unacceptable risks or unacceptable costs”. Forty years of complex research and development has shown that Accelerator Driven Systems could provide a solution to the challenge posed by spent nuclear fuels, by enabling the ability to considerably decrease their radiotoxicity lifetime burden and volume. In particular, a multilateral strategy of treatment of the MAs could be a commendable solution for both the countries phasing out the exploitation of nuclear energy and for those pursuing and developing this exploitation. The pre-industrial assessment of the technical and financial feasibility for industrialization is the next step. This applies to the four R&D and Demonstration building blocks: advanced separation, MAs’ loaded fuel fabrication, dedicated transmuters demonstration (MYRRHA) and provision for MAs’ fuel loaded processing. A global vision of the process leading to a sustainable option is proposed.

A Review of the Surface Modifications for Corrosion Mitigation of Steels in Lead and LBE

The review paper starts with the applications of liquid metals and then concentrates on lead and lead–bismuth eutectic used in Gen IV nuclear reactors and accelerator-driven systems. Key points of degradation modes of austenitic stainless steels and ferritic-martensitic steels, candidates for the structural components, are briefly summarized. Corrosion and liquid metal embrittlement are critical issues that must be overcome. Next, the paper focuses on the strong efforts paid to the mitigation of corrosion and reviews the different solutions proposed for the protection of steels in lead and lead–bismuth eutectic. There exist promising solutions based on protection by deposition of protective coatings or protection by “natural” oxidation resulting from optimized chemical composition of the steels. However, the solutions have to be confirmed especially by longer-term experiments and by additional mechanical testing.

FUEL-CYCLE SCENARIO TO REDUCE RADIOACTIVE WASTE FROM LIGHT-WATER REACTOR

It is important to reduce the amount of trans-uranium (TRU) produced from the existing nuclear power plants to realize sustainable nuclear energy since the some TRU nuclides remain for a long time and have high radioactivity and radiotoxicity. One of the promising solutions is to transmute the TRU nuclides to those with lesser radioactivity and radiotoxicity in the existing nuclear reactors. In the current scheme, the TRU nuclides are transmuted in fast reactors and/or accelerator-driven-systems, however, this scenario seems unpromising in Japan: after the Fukushima Daiichi accident, it is required to reduce the production of TRU nuclides from the light-water reactors. In the previous studies, a concept of FORSETI was investigated, and a nuclear-fuel cycle simulation code ATRUNCYS was developed to study the low TRU production scenario. The FORSETI concept consists of two types of fuels: 1) UO2 fuels with high-assay low-enriched-uranium, and 2) MOX fuels with highly fissile concentrated plutonium reprocessed from the FORSETI-UO2 fuels. The current paper focuses on the following two scenarios: a) once-recycled scenario with the current fuel design, and b) once-recycled scenario with the FORSETI concept. The two scenarios were compared by using the ATRUNCYS code where the simulation studies showed that the amount, radioactivity, and radiotoxicity of resulting waste can be decreased in the FORSETI concept: In the case 1), the production of TRU nuclides decreased in the UO2 fuel; In the case 2), the fission rate increased and neutron-capture reactions of 240Pu and 241Pu decreased in the MOX fuels.

Study of the 232Th(n,f) Cross section at NCSR ‘Demokritos’ using Micromegas Detectors

The accurate knowledge of neutron-induced fission cross sections in actinides, is of great importance when it comes to the design of fast nuclear reactors, as well as accelerator driven systems. Specifically for the 232Th(n,f) case, the existing experimental datasets are quite discrepant in both the low and high energy MeV regions, thus leading to poor evaluations, a fact that in turn implies the need for more accurate measurements.In the present work, the total cross section of the 232Th(n,f) reaction has been measured relative to the 235U(n,f) and 238U(n,f) ones, at incident energies of 7.2, 8.4, 9.9 MeV and 14.8, 16.5, 17.8 MeV utilizing the 2H(d,n) and 3H(d,n) reactions respectively, which generally yield quasi-monoenergetic neutron beams. The experiments were performed at the 5.5 MV Tandem accelerator laboratory of N.C.S.R. “Demokritos”, using a Micromegas detector assembly and an ultra thin ThO2 target, especially prepared for fission measurements at n_ToF, CERN during its first phase of operations, using the painting technique. The masses of all actinide samples were determined via α-spectroscopy. The produced fission yields along with the results obtained from activation foils were studied in parallel, using both the NeusDesc [1] and MCNP5 [2] codes, taking into consideration competing nuclear reactions (e.g. deuteron break up), along with neutron elastic and inelastic scattering with the beam line, detector housing and experimental hall materials. Since the 232Th(n,f) reaction has a relatively low energy threshold and can thus be affected by parasitic neutrons originating from a variety of sources, the thorough characterization of the neutron flux impinging on the targets is a prerequisite for accurate cross-section measurements, especially in the absence of time-of-flight capabilities. Additional Monte-Carlo simulations were also performed coupling both GEF [3] and FLUKA [4] codes for the determination of the detection efficiency.

Measurement of the 234U(n,f) cross section in the energy range between 14.8 and 17.8 MeV using Micromegas detectors

Neutron induced fission cross sections of actinides present special interest, since they lead to the design optimization of new generation reactors (Generation IV) as well as Accelerator Driven Systems (ADS). In the present work, the 234U(n,f) cross section was measured for which only a few available discrepant data exist in literature leading to poor evaluations. More specifically, four irradiations were performed at the 5.5 MV Tandem Accelerator Laboratory of NCSR “Demokritos” using quasi-monoenergetic neutrons produced by the 3H(d,n)4He reaction in the 14.8-19.2 MeV energy range. The 234U(n,f) cross section was measured relatively to the 235U(n,f) and 238U(n,f) reference ones and in order to perform the in-beam measurements for each of the actinide targets (234U, 238U, 235U), a Micromegas detector was used to record the fission fragments. The target-detector pairs were placed in an Al chamber filled with a Ar:CO2 (in 80:20 volume fraction) gas mixture at atmospheric pressure and temperature. The efficiency of the Micromegas detectors was estimated by Monte-Carlo simulations using the GEF and FLUKA codes. In addition, a detailed study of the neutron energy spectra was carried out by coupling both NeuSDesc and MCNP5 codes in order to take into account and correct for the contribution of low energy parasitic neutrons in the fission yields.

Theoretical calculation and evaluation of neutron inducedreactions on Pu isotopes

The nuclear data on n+239,240,242,244Pu reactions for the incident energy up to 200 MeV are consistently calculated and evaluated in order to meet the design requirements of Generation-IV reactors and accelerator driven systems. The optical model, the distorted wave Born approximation theory, the Hauser-Feshbach theory, the fission model, the evaporation model, the exciton model and the intranuclear cascade model are used in the calculation, and new experimental data are taken into account. Our data are compared with experimental data and the evaluated data from JENDL-4/HE and TENDL. In addition, the variation tendency of reaction cross sections related to the target mass numbers is obtained, which is very important for the prediction of nuclear data on neutron-actinides reactions because the experimental data are lacking.

Validation of Neutron Evaluated Data Based on The Experimental Reactivity Worth of Tungsten Target in CiADS

Measurement of a cylindrical tungsten target reactivity worth has been performed on the light water zero-power reactor of VENUS-II at China Institute of Atomic Energy (CIAE) in order to verify the neutron evaluated data related to the engineering design of Chinese initiative Accelerator Driven Systems (CiADS). The reactivity worth of the tungsten target was measured and processed as -1.234±0.114mk by a period method. The experimental result was compared with the simulation ones calculated by MCNP with five different libraries, i.e., ENDF/B-VII.0, ENDF/B-VII.1, JENDL-4.0, CENDL-3.1 and JEFF-3.2. By comparing the results of experiment and simulation, the simulated results from ENDF/B-VII.0, JENDL-4.0 and JEFF-3.2 are higher than the experimental result, however that from CENDL-3.1 is lower. The result from ENDF/B-VII.1 library shows better agreement with the experiment one and the relative deviation is less than 2%. Through the analysis of the differences of the results, non-tungsten elements cross sections in the ENDF/B-VII.1 mainly affect the tungsten radiation capture and elastic scattering reaction rates in the energy range of 10-9-10-7 MeV, which results in a better simulated tungsten target reactivity worth value. Therefore, it is recommended that the tungsten target reactivity worth should be calculated with the ENDF/B-VII.1.

Thermal-hydraulic Assessment of Seven Wire-wrapped rod Bundle

Lead bismuth eutectic (LBE) is one of the most potential materials for coolant and spallation target for Accelerator Driven Systems (ADS). Thermal-hydraulic behavior of LBE in fuel assembly is a key issue for development of the systems. To get a deeper understanding on the complex thermal-hydraulic features of wire-wrapped rod bundle cooled by upward LBE, an electrically bundle with 7 rods wrapped with helical wire was developed in KYLIN-II thermal-hydraulic forced circulation loop. The flow resistance, thermal entrance characteristic and heat transfer coefficient were investigated. As for the entrance characteristics, during the full heating length (exceeding 140 times the hydraulic diameter), the thermal field did not reach a fully developed and stable condition which is contrary to the ducted flows. The experimental heat transfer coefficient showed that the hexagonal shell has a great influence on the heat transfer coefficient in rod bundle geometry. For this reason the application of empirical correlation should be kept cautious in rod bundle analysis.

ANALYSIS OF NEUTRON FIELDS GENERATED IN SPALLATION TARGETS OF B-URAN EXPERIMENTAL ASSEMBLY USING MONTE CARLO METHOD

The aim of this paper is to introduce experimental assembly B-URAN and the results of Monte Carlo simulations of neutron fields, which will be generated by using various spallation targets. This experimental assembly was constructed in Joint Institute of Nuclear Research in Dubna, Russian Federation, in order to study accelerator driven systems fundamental characteristics. Beam of 660 MeV protons should be used for that purpose. The MCNP model of such set-up has been developed at Brno University of Technology, Czech Republic. The goal is to get data needed for prediction of reaction rates in detectors placed in B-URAN experimental channels. Such data will be experimentally validated later. Furthermore, simulations of radiation exposure around this xperimental assembly were performed.