New measurement of the B-11(p,alpha(0))Be-8 bare-nucleus S(E) factor via the Trojan horse method

A new measurement of the B-11(p,alpha(0))Be-8 has been performed applying the Trojan horse method (THM) to the H-2(B-11,alpha Be-8(0))n quasi-free reaction induced at a laboratory energy of 27 MeV. The astrophysical S(E) factor has been extracted from similar to 600 keV down to zero energy by means of an improved data analysis technique and it has been compared with direct data available in the literature. The range investigated here overlaps with the energy region of the light element LiBeB stellar burning and with that of future aneutronic fusion power plants using the B-11+p fuel cycle. The new investigation described here confirms the preliminary results obtained in the recent TH works. The origin of the discrepancy between the direct estimate of the B-11(p,alpha(0))Be-8 S(E)-factor at zero energy and that from a previous THM investigation is quantitatively corroborated. The results obtained here support, within the experimental uncertainties, the low-energy S(E)-factor extrapolation and the value of the electron screening potential deduced from direct measurements.

Impact radiologique liés au transport de matières radioactives par route en Espagne / Radiological impact associated with the transport by road of radioactive material in Spain

Questions relating to the transport of radioactive materials are very much an issue of current interest due to the increasing mobility of the materials involved in the nuclear fuel cycle, commitment to the environment, the safety and protection of persons and the corresponding regulatory legal framework. The radiological impact associated with this type of transport was assessed by means of a new data-processing tool that may be of use and serve as complementary documentation to that included in transport regulations. Thus, by determining the level of radiation at a distance of one metre from the transport vehicle and by selecting a route, the associated impacts will be obtained, such as the affected populations, the dose received by the most highly exposed individual, the overall radiological impact, the doses received by the population along the route and the possible detriment to their health. The most important conclusion is that the emissions of ionising radiation from the transport of radioactive material by road in Spain are not significant as regards the generation of adverse effects on human health, and that their radiological impact may be considered negligible.

Uncertainty methods inactivation and inventory calculations

For a number of important nuclides, complete activation data libraries with covariance data will be produced, so that uncertainty propagation in fuel cycle codes (in this case ACAB,FISPIN, ...) can be developed and tested. Eventually, fuel inventory codes should be able to handle the complete set of uncertainty data, i.e. those of nuclear reactions (cross sections, etc.), radioactive decay and fission yield data. For this, capabilities will be developed both to produce covariance data and to propagate the uncertainties through the inventory calculations.

Performance of a transmutation advanced device for sustainable energy application

Preliminary studies have been performed to design a device for nuclear waste transmutation and hydrogen generation based on a gas-cooled pebble bed accelerator driven system, TADSEA (Transmutation Advanced Device for Sustainable Energy Application). In previous studies we have addressed the viability of an ADS Transmutation device that uses as fuel wastes from the existing LWR power plants, encapsulated in graphite in the form of pebble beds, cooled by helium which enables high temperatures (in the order of 1200 K), to generate hydrogen from water either by high temperature electrolysis or by thermochemical cycles. For designing this device several configurations were studied, including several reflectors thickness, to achieve the desired parameters, the transmutation of nuclear waste and the production of 100 MW of thermal power. In this paper new studies performed on deep burn in-core fuel management strategy for LWR waste are presented. The fuel cycle on TADSEA device has been analyzed based on both: driven and transmutation fuel that had been proposed by the General Atomic design of a gas turbine-modular helium reactor. The transmutation results of the three fuel management strategies, using driven, transmutation and standard LWR spent fuel were compared, and several parameters describing the neutron performance of TADSEA nuclear core as the fuel and moderator temperature reactivity coefficients and transmutation chain, are also presented

Analysis of a Spanish energy scenario with Generation IV nuclear reactors

The advantages of fast-spectrum reactors consist not only of an efficient use of fuel through the breeding of fissile material and the use of natural or depleted uranium, but also of the potential reduction of the amount of actinides such as americium and neptunium contained in the irradiated fuel. The first aspect means a guaranteed future nuclear fuel supply. The second fact is key for high-level radioactive waste management, because these elements are the main responsible for the radioactivity of the irradiated fuel in the long term. The present study aims to analyze the hypothetical deployment of a Gen-IV Sodium Fast Reactor (SFR) fleet in Spain. A nuclear fleet of fast reactors would enable a fuel cycle strategy different than the open cycle, currently adopted by most of the countries with nuclear power. A transition from the current Gen-II to Gen-IV fleet is envisaged through an intermediate deployment of Gen-III reactors. Fuel reprocessing from the Gen-II and Gen-III Light Water Reactors (LWR) has been considered. In the so-called advanced fuel cycle, the reprocessed fuel used to produce energy will breed new fissile fuel and transmute minor actinides at the same time. A reference case scenario has been postulated and further sensitivity studies have been performed to analyze the impact of the different parameters on the required reactor fleet. The potential capability of Spain to supply the required fleet for the reference scenario using national resources has been verified. Finally, some consequences on irradiated final fuel inventory are assessed. Calculations are performed with the Monte Carlo transport-coupled depletion code SERPENT together with post-processing tools.

Application of noise analysis for sensor surveillance in Trillo NPP

It is a known fact that noise analysis is a suitable method for sensor performance surveillance. In particular, controlling the response time of a sensor is an efficient way to anticipate failures and to have the opportunity to prevent them. In this work the response times of several sensors of Trillo NPP are estimated by means of noise analysis. The procedure applied consists of modeling each sensor with autoregressive methods and getting the searched parameter by analyzing the response of the model when a ramp is simulated as the input signal. Core exit thermocouples and in core self-powered neutron detectors are the main sensors analyzed but other plant sensors are studied as well. Since several measurement campaigns have been carried out, it has been also possible to analyze the evolution of the estimated parameters during more than one fuel cycle. Some sensitivity studies for the sample frequency of the signals and its influence on the response time are also included. Calculations and analysis have been done in the frame of a collaboration agreement between Trillo NPP operator (CNAT) and the School of Mines of Madrid.

Nuclear waste forms for actinides

The disposition of actinides, most recently 239Pu from dismantled nuclear weapons, requires effective containment of waste generated by the nuclear fuel cycle. Because actinides (e.g., 239Pu and 237Np) are long-lived, they have a major impact on risk assessments of geologic repositories. Thus, demonstrable, long-term chemical and mechanical durability are essential properties of waste forms for the immobilization of actinides. Mineralogic and geologic studies provide excellent candidate phases for immobilization and a unique database that cannot be duplicated by a purely materials science approach. The “mineralogic approach” is illustrated by a discussion of zircon as a phase for the immobilization of excess weapons plutonium.

Light water breeder reactor program.

"June 1976."

Commercial nuclear power.

"DOE/EIA-0438."

Effect of aspect ratio on thermal convection in open and closed systems

Internally heated fluids are found across the nuclear fuel cycle. In certain situations the motion of the fluid is driven by the decay heat (i.e. corium melt pools in severe accidents, the shutdown of liquid metal reactors, molten salt and the passive control of light water reactors) as well as normal operation (i.e. intermediate waste storage and generation IV reactor designs). This can in the long-term affect reactor vessel integrity or lead to localized hot spots and accumulation of solid wastes that may prompt local increases in activity. Two approaches to the modeling of internally heated convection are presented here. These are based on numerical analysis using codes developed in-house and simulations using widely available computational fluid dynamics solvers. Open and closed fluid layers at around the transition between conduction and convection of various aspect ratios are considered. We determine optimum domain aspect ratio (1:7:7 up to 1:24:24 for open systems and 5:5:1, 1:10:10 and 1:20:20 for closed systems), mesh resolutions and turbulence models required to accurately and efficiently capture the convection structures that evolve when perturbing the conductive state of the fluid layer. Note that the open and closed fluid layers we study here are bounded by a conducting surface over an insulating surface. Conclusions will be drawn on the influence of the periodic boundary conditions on the flow patterns observed. We have also examined the stability of the nonlinear solutions that we found with the aim of identifying the bifurcation sequence of these solutions en route to turbulence.

Desenvolvimento e validação de metodologia analítica para quantificação de urânio em compostos do ciclo do combustível nuclear por espectroscopia no infravermelho com transformada de Fourier (FTIR)

Tese (Doutorado em Tecnologia Nuclear)

Composite electrode for unitized regenerative proton exchange membrane fuel cell with improved cycle life

To improve the cycle life of unitized regenerative fuel cells (URFCs), an electrode with a composite structure has been developed. The cycle life and polarization curves for both fuel cell and electrolysis modes of URFC operation were investigated. The cycle life of URFCs was improved considerably and the performance was fairly constant during 25 cycles, which illustrates that the composite electrode is effective in sustaining the cyclic performance of URFCs. It shows the URFCs with such an electrode structure are promising for practical applications. (C) 2004 The Electrochemical Society.