Demand driven salt clean-up in a molten salt fast reactor - Defining a priority list.

The PUREX technology based on aqueous processes is currently the leading reprocessing technology in nuclear energy systems. It seems to be the most developed and established process for light water reactor fuel and the use of solid fuel. However, demand driven development of the nuclear system opens the way to liquid fuelled reactors, and disruptive technology development through the application of an integrated fuel cycle with a direct link to reactor operation. The possibilities of this new concept for innovative reprocessing technology development are analysed, the boundary conditions are discussed, and the economic as well as the neutron physical optimization parameters of the process are elucidated. Reactor physical knowledge of the influence of different elements on the neutron economy of the reactor is required. Using an innovative study approach, an element priority list for the salt clean-up is developed, which indicates that separation of Neodymium and Caesium is desirable, as they contribute almost 50% to the loss of criticality. Separating Zirconium and Samarium in addition from the fuel salt would remove nearly 80% of the loss of criticality due to fission products. The theoretical study is followed by a qualitative discussion of the different, demand driven optimization strategies which could satisfy the conflicting interests of sustainable reactor operation, efficient chemical processing for the salt clean-up, and the related economic as well as chemical engineering consequences. A new, innovative approach of balancing the throughput through salt processing based on a low number of separation process steps is developed. Next steps for the development of an economically viable salt clean-up process are identified.

OVERVIEW OF NEUTRON MEASUREMENTS IN JET FUSION DEVICE.

The design and operation of ITER experimental fusion reactor requires the development of neutron measurement techniques and numerical tools to derive the fusion power and the radiation field in the device and in the surrounding areas. Nuclear analyses provide essential input to the conceptual design, optimisation, engineering and safety case in ITER and power plant studies. The required radiation transport calculations are extremely challenging because of the large physical extent of the reactor plant, the complexity of the geometry, and the combination of deep penetration and streaming paths. This article reports the experimental activities which are carried-out at JET to validate the neutronics measurements methods and numerical tools used in ITER and power plant design. A new deuterium-tritium campaign is proposed in 2019 at JET: the unique 14 MeV neutron yields produced will be exploited as much as possible to validate measurement techniques, codes, procedures and data currently used in ITER design thus reducing the related uncertainties and the associated risks in the machine operation.

Evaluation of radioactivity in the bodies of mice induced by neutron exposure from an epi-thermal neutron source of an accelerator-based boron neutron capture therapy system.

This study aimed to evaluate the residual radioactivity in mice induced by neutron irradiation with an accelerator-based boron neutron capture therapy (BNCT) system using a solid Li target. The radionuclides and their activities were evaluated using a high-purity germanium (HP-Ge) detector. The saturated radioactivity of the irradiated mouse was estimated to assess the radiation protection needs for using the accelerator-based BNCT system. 24Na, 38Cl, 80mBr, 82Br, 56Mn, and 42K were identified, and their saturated radioactivities were (1.4 ± 0.1) × 102, (2.2 ± 0.1) × 101, (3.4 ± 0.4) × 102, 2.8 ± 0.1, 8.0 ± 0.1, and (3.8 ± 0.1) × 101 Bq/g/mA, respectively. The 24Na activation rate at a given neutron fluence was found to be consistent with the value reported from nuclear-reactor-based BNCT experiments. The induced activity of each nuclide can be estimated by entering the saturated activity of each nuclide, sample mass, irradiation time, and proton current into the derived activation equation in our accelerator-based BNCT system.

Laser-heating and Radiance Spectrometry for the Study of Nuclear Materials in Conditions Simulating a Nuclear Power Plant Accident.

Major and severe accidents have occurred three times in nuclear power plants (NPPs), at Three Mile Island (USA, 1979), Chernobyl (former USSR, 1986) and Fukushima (Japan, 2011). Research on the causes, dynamics, and consequences of these mishaps has been performed in a few laboratories worldwide in the last three decades. Common goals of such research activities are: the prevention of these kinds of accidents, both in existing and potential new nuclear power plants; the minimization of their eventual consequences; and ultimately, a full understanding of the real risks connected with NPPs. At the European Commission Joint Research Centre's Institute for Transuranium Elements, a laser-heating and fast radiance spectro-pyrometry facility is used for the laboratory simulation, on a small scale, of NPP core meltdown, the most common type of severe accident (SA) that can occur in a nuclear reactor as a consequence of a failure of the cooling system. This simulation tool permits fast and effective high-temperature measurements on real nuclear materials, such as plutonium and minor actinide-containing fission fuel samples. In this respect, and in its capability to produce large amount of data concerning materials under extreme conditions, the current experimental approach is certainly unique. For current and future concepts of NPP, example results are presented on the melting behavior of some different types of nuclear fuels: uranium-plutonium oxides, carbides, and nitrides. Results on the high-temperature interaction of oxide fuels with containment materials are also briefly shown.

Temperature oscillations near natural nuclear reactor cores and the potential for prebiotic oligomer synthesis.

Geologic settings capable of driving prebiotic oligomer synthesis reactions remain a relatively unexplored aspect of origins of life research. Natural nuclear reactors are an example of Precambrian energy sources that produced unique temperature fluctuations. Heat transfer models indicate that water-moderated, convectively-cooled natural fission reactors in porous host rocks create temperature oscillations that resemble those employed in polymerase chain reaction (PCR) devices to artificially amplify oligonucleotides. This temperature profile is characterized by short-duration pulses up to 70-100 °C, followed by a sustained period of temperatures in the range of 30-70 °C, and finally a period of relaxation to ambient temperatures until the cycle is restarted by a fresh influx of pore water. For a given reactor configuration, temperature maxima and the time required to relax to ambient temperatures depend most strongly on the aggregate effect of host rock permeability in decreasing the thermal expansion and increasing the viscosity and evaporation temperature of the pore fluids. Once formed, fission-fueled reactors can sustain multi-kilowatt-level power production for 10(5)-10(6) years, ensuring microenvironmental longevity and chemical output. The model outputs indicate that organic synthesis on young planetary bodies with a sizeable reservoir of fissile material can involve more sophisticated energy dissipation pathways than modern terrestrial analog settings alone would suggest.

Nuclear power in the 21st century: Challenges and possibilities.

The current situation and possible future developments for nuclear power--including fission and fusion processes--is presented. The fission nuclear power continues to be an essential part of the low-carbon electricity generation in the world for decades to come. There are breakthrough possibilities in the development of new generation nuclear reactors where the life-time of the nuclear waste can be reduced to some hundreds of years instead of the present time-scales of hundred thousand of years. Research on the fourth generation reactors is needed for the realisation of this development. For the fast nuclear reactors, a substantial research and development effort is required in many fields--from material sciences to safety demonstration--to attain the envisaged goals. Fusion provides a long-term vision for an efficient energy production. The fusion option for a nuclear reactor for efficient production of electricity has been set out in a focussed European programme including the international project of ITER after which a fusion electricity DEMO reactor is envisaged.

Neutron flux characterisation of the Pavia TRIGA Mark II research reactor for radiobiological and microdosimetric applications.

Nowadays the Pavia TRIGA reactor is available for national and international collaboration in various research fields. The TRIGA Mark II nuclear research reactor of the Pavia University offers different in- and out-core neutron irradiation channels, each characterised by different neutron spectra. In the last two years a campaign of measurements and simulations has been performed in order to guarantee a better characterisation of these different fluxes and to meet the demands of irradiations that require precise information on these spectra in particular for radiobiological and microdosimetric studies. Experimental data on neutron fluxes have been collected analysing and measuring the gamma activity induced in thin target foils of different materials irradiated in different TRIGA experimental channels. The data on the induced gamma activities have been processed with the SAND II deconvolution code and finally compared with the spectra obtained with Monte Carlo simulations. The comparison between simulated and measured spectra showed a good agreement allowing a more precise characterisation of the neutron spectra and a validation of the adopted method.

Modification of the radial beam port of ITU TRIGA Mark II research reactor for BNCT applications.

This paper aims to describe the modification of the radial beam port of ITU (Istanbul Technical University) TRIGA Mark II research reactor for BNCT applications. Radial beam port is modified with Polyethylene and Cerrobend collimators. Neutron flux values are measured by neutron activation analysis (Au-Cd foils). Experimental results are verified with Monte Carlo results. The results of neutron/photon spectrum, thermal/epithermal neutron flux, fast group photon fluence and change of the neutron fluxes with the beam port length are presented.

Design and construction of a thermal neutron beam for BNCT at Tehran Research Reactor.

An irradiation facility has been designed and constructed at Tehran Research Reactor (TRR) for the treatment of shallow tumors using Boron Neutron Capture Therapy (BNCT). TRR has a thermal column which is about 3m in length with a wide square cross section of 1.2×1.2m(2). This facility is filled with removable graphite blocks. The aim of this work is to perform the necessary modifications in the thermal column structure to meet thermal BNCT beam criteria recommended by International Atomic Energy Agency. The main modifications consist of rearranging graphite blocks and reducing the gamma dose rate at the beam exit. Activation foils and TLD700 dosimeter have been used to measure in-air characteristics of the neutron beam. According to the measurements, a thermal flux is 5.6×10(8) (ncm(-2)s(-1)), a cadmium ratio is 186 for gold foils and a gamma dose rate is 0.57Gy h(-1).

A feasibility study of the Tehran research reactor as a neutron source for BNCT.

Investigation on the use of the Tehran Research Reactor (TRR) as a neutron source for Boron Neutron Capture Therapy (BNCT) has been performed by calculating and measuring energy spectrum and the spatial distribution of neutrons in all external irradiation facilities, including six beam tubes, thermal column, and the medical room. Activation methods with multiple foils and a copper wire have been used for the mentioned measurements. The results show that (1) the small diameter and long length beam tubes cannot provide sufficient neutron flux for BNCT; (2) in order to use the medical room, the TRR core should be placed in the open pool position, in this situation the distance between the core and patient position is about 400 cm, so neutron flux cannot be sufficient for BNCT; and (3) the best facility which can be adapted for BNCT application is the thermal column, if all graphite blocks can be removed. The epithermal and fast neutron flux at the beginning of this empty column are 4.12×10(9) and 1.21×10(9) n/cm(2)/s, respectively, which can provide an appropriate neutron beam for BNCT by designing and constructing a proper Beam Shaping Assembly (BSA) structure.

Study of plasma meniscus and beam halo in negative ion sources using three dimension in real space and three dimension in velocity space particle in cell model.

Our previous study by two dimension in real space and three dimension in velocity space-particle in cell model shows that the curvature of the plasma meniscus causes the beam halo in the negative ion sources. The negative ions extracted from the periphery of the meniscus are over-focused in the extractor due to the electrostatic lens effect, and consequently become the beam halo. The purpose of this study is to verify this mechanism with the full 3D model. It is shown that the above mechanism is essentially unchanged even in the 3D model, while the fraction of the beam halo is significantly reduced to 6%. This value reasonably agrees with the experimental result.

The FiR 1 photon beam model adjustment according to in-air spectrum measurements with the Mg(Ar) ionization chamber.

The mixed neutron-photon beam of FiR 1 reactor is used for boron-neutron capture therapy (BNCT) in Finland. A beam model has been defined for patient treatment planning and dosimetric calculations. The neutron beam model has been validated with an activation foil measurements. The photon beam model has not been thoroughly validated against measurements, due to the fact that the beam photon dose rate is low, at most only 2% of the total weighted patient dose at FiR 1. However, improvement of the photon dose detection accuracy is worthwhile, since the beam photon dose is of concern in the beam dosimetry. In this study, we have performed ionization chamber measurements with multiple build-up caps of different thickness to adjust the calculated photon spectrum of a FiR 1 beam model.

Determination of critical assembly absolute power using post-irradiation activation measurement of week-lived fission products.

The work presents a detailed comparison of calculated and experimentally determined net peak areas of longer-living fission products after 100 h irradiation on a reactor with power of ~630 W and several days cooling. Specifically the nuclides studied are (140)Ba, (103)Ru, (131)I, (141)Ce, (95)Zr. The good agreement between the calculated and measured net peak areas, which is better than in determination using short lived (92)Sr, is reported. The experiment was conducted on the VVER-1000 mock-up installed on the LR-0 reactor. The Monte Carlo approach has been used for calculations. The influence of different data libraries on results of calculation is discussed as well.

Microdosimetric measurements in the thermal neutron irradiation facility of LENA reactor.

A twin TEPC with electric-field guard tubes has been constructed to be used to characterize the BNCT field of the irradiation facility of LENA reactor. One of the two mini TEPC was doped with 50ppm of (10)B in order to simulate the BNC events occurring in BNCT. By properly processing the two microdosimetric spectra, the gamma, neutron and BNC spectral components can be derived with good precision (~6%). However, direct measurements of (10)B in some doped plastic samples, which were used for constructing the cathode walls, point out the scarce accuracy of the nominal (10)B concentration value. The influence of the Boral(®) door, which closes the irradiation channel, has been measured. The gamma dose increases significantly (+51%) when the Boral(®) door is closed. The crypt-cell-regeneration weighting function has been used to measure the quality, namely the RBEµ value, of the radiation field in different conditions. The measured RBEµ values are only partially consistent with the RBE values of other BNCT facilities.

Radiation protection performance for the dismantling of the WWR-M primary cooling circuit.

The WWR-M is a light-water-cooled and moderated heterogonous research reactor with a thermal output of 10 MW. The reactor has been in operation for >50 y and has had an excellent safety record. A non-hermeticity of the inlet line of the primary cooling circuit (PCC) was found, and the only reasonable technical solution was the complete replacement of the PCC inlet and outlet pipe lines. Such a replacement was a challenging technical task due to the necessity to handle large size components with complex geometries under conditions of high-level radiation fields, and therefore, it required detailed planning aiming to reduce staff exposure. This paper describes the dismantling and removal of the PCC components focusing on radiation protection issues.

Investigation on F/M material aspects of IRT-Sofia NCT channel.

The filter/moderator area of IRT-Sofia BNCT channel was investigated in this study in order to find a higher radiation resistant material as a suitable substitution for the Teflon(®). Two options - Al2O3 and graphite - were investigated. The results show, that both graphite and the Al2O3 can be successfully used as a filter/moderator material at IRT-Sofia. Initial evaluation of the in-phantom performance of the IRT-Sofia BNCT channel was made and merits similar to the best existing ones were found.

Extended set of activation monitors for NCT beam characterization and spectral conditions of the beam after reactor fuel conversion.

Since 2010 the LVR-15 reactor has been gradually converted from highly enriched fuel (36wt% (235)U) to low enriched fuel with the enrichment of 19.75wt% (235)U. Paper presents influence of the core pattern changes on the neutron characteristics of the epithermal beam. The determination of neutron spectrum free in the beam was done with a set of neutron activation monitors. After the reactor conversion the change in neutron spectrum is not provable as differences are in the range of measurement errors.

On the safety of ITER accelerators.

Three 1 MV/40A accelerators in heating neutral beams (HNB) are on track to be implemented in the International Thermonuclear Experimental Reactor (ITER). ITER may produce 500 MWt of power by 2026 and may serve as a green energy roadmap for the world. They will generate -1 MV 1 h long-pulse ion beams to be neutralised for plasma heating. Due to frequently occurring vacuum sparking in the accelerators, the snubbers are used to limit the fault arc current to improve ITER safety. However, recent analyses of its reference design have raised concerns. General nonlinear transformer theory is developed for the snubber to unify the former snubbers' different design models with a clear mechanism. Satisfactory agreement between theory and tests indicates that scaling up to a 1 MV voltage may be possible. These results confirm the nonlinear process behind transformer theory and map out a reliable snubber design for a safer ITER.

Absolute determination of power density in the VVER-1000 mock-up on the LR-0 research reactor.

The work presents a detailed comparison of calculated and experimentally determined net peak areas of selected fission products gamma lines. The fission products were induced during a 2.5 h irradiation on the power level of 9.5 W in selected fuel pins of the VVER-1000 Mock-Up. The calculations were done with deterministic and stochastic (Monte Carlo) methods. The effects of different nuclear data libraries used for calculations are discussed as well. The Net Peak Area (NPA) may be used for the determination of fission density across the mock-up. This fission density is practically identical to power density.

Neutron spectrometry and dosimetry study at two research nuclear reactors using Bonner sphere spectrometer (BSS), rotational spectrometer (ROSPEC) and cylindrical nested neutron spectrometer (NNS).

Neutron spectrometry and subsequent dosimetry measurements were undertaken at the McMaster Nuclear Reactor (MNR) and AECL Chalk River National Research Universal (NRU) Reactor. The instruments used were a Bonner sphere spectrometer (BSS), a cylindrical nested neutron spectrometer (NNS) and a commercially available rotational proton recoil spectrometer. The purposes of these measurements were to: (1) compare the results obtained by three different neutron measuring instruments and (2) quantify neutron fields of interest. The results showed vastly different neutron spectral shapes for the two different reactors. This is not surprising, considering the type of the reactors and the locations where the measurements were performed. MNR is a heavily shielded light water moderated reactor, while NRU is a heavy water moderated reactor. The measurements at MNR were taken at the base of the reactor pool, where a large amount of water and concrete shielding is present, while measurements at NRU were taken at the top of the reactor (TOR) plate, where there is only heavy water and steel between the reactor core and the measuring instrument. As a result, a large component of the thermal neutron fluence was measured at MNR, while a negligible amount of thermal neutrons was measured at NRU. The neutron ambient dose rates at NRU TOR were measured to be between 0.03 and 0.06 mSv h⁻¹, while at MNR, these values were between 0.07 and 2.8 mSv h⁻¹ inside the beam port and <0.2 mSv h⁻¹ between two operating beam ports. The conservative uncertainty of these values is 15 %. The conservative uncertainty of the measured integral neutron fluence is 5 %. It was also found that BSS over-responded slightly due to a non-calibrated response matrix.