Calorimetric measurements on electrochemical cells with Pd-D cathodes

Two series of experiments were performed to determine the conditions of cell operation that produce sufficient excess heat to be useful for the production of energy. In the first series, the results from a differential temperature analysis of identical light- and heavy-water electrochemical cells were too ambiguous and, thus, not suitable for evaluating excess heat effects. In the second series, two Pd-D/LiOD-saturated D₂O/Pt cells were operated at current densities between 12.5–500 mA/cm² in a constant-heatloss-rate twin calorimeter for 460 hours. Water loss measurements during the experiments indicated that the recombination reaction (2D₂ + O₂ 2D₂O) did not occur. The D/Pd ratio was determined gravimetrically during the experiments. No excess heat was found within the sensitivity (0.13 W, 0.082 W/cm³ of Pd, 0.013 W/cm² of Pd) and precision (±0.3 W) of the calorimeter.     

Nuclear Characteristics of Molten-Salt-Cooled D-D Fusion Reactor Blankets

To consecutively decompose ¹⁴CO₂ into carbon (¹⁴C) through its reaction with Hz, an apparatus using microwave discharge and its conditioning were investigated. The reaction produces CO as an intermediate, and proceeds in the two steps of (1) “CO₂+H₂ → CO+H₂O” and (2) “CO+H₂+C_n → C_n+1+H₂O”, where C_n denotes the carbon already deposited on the wall of the discharge tube. Preliminary dispersion of carbon to the wall of the discharge tube by sputtering of a graphite particle was effective to promote the reaction. Two silica discharge tubes (6 mm O.D., 4 mm I.D., and 150 mm length each) were connected in series to proceed the former reaction in the first discharge tube and the latter one in the second one. When a 1:3 mixture of CO₂ and H₂ (total pressure 0.67kPa) was passed through the discharge tubes at a linear gas velocity of approximately 30mm/s and discharged for 60 h under microwave of 30–40 W supplied from two 2,450MHz power generators (200 W each), more than 90% of CO₂ was converted into CO in the 1st tube and about 23% of the CO was then decomposed into carbon in the 2nd tube. However, about 50% of the CO escaped from the tube without being decomposed, and about 0.5% and 1% of the carbon fed were hydrogenated into CH₄ and C₂H₂, respectively. The rest about 25% which was not confirmed was probably evacuated from the 2nd tube as microparticles of carbon. To completely decompose CO₂ into carbon, additional discharge tubes are necessary downstream of the 2nd tube.     

Thermal hydraulic study on a high-temperature gas–gas heat exchanger with helically coiled tube bundles

An experimental study was carried out to investigate flow-induced vibration, heat transfer and pressure drop of helically coiled tubes of an intermediate heat exchanger (IHX) for the HTTR, using a full-size partial model and air as the fluid. The test model has 54 helically coiled tubes separated into three layer bundles, surrounding the center pipe. The vibration of the tube bundles was mainly at the center pipe, and the individual vibrations of the tube bundles were not significant under the operation conditions of the IHX. The heat transfer of the tube outside, due to forced convection, was obtained as a function of Re^0.51Pr^0.3, and the friction factor, depending on the tube arrangement, was proportional to Re^−0.14.     

Thermal Diffusion of Hydrogen in Zircaloy-2 Containing Hydrogen beyond Terminal Solid Solubility

The thermal diffusion of hydrogen is one of causes of uneven hydride precipitation in zircaloy fuel cladding tubes that are used in water reactors. In the diffusion model of hydrogen in zircaloy, the effects of the hydride on the diffusibility of hydrogen has been regarded as negligibly small in comparison with that of hydrogen dissolved in the matrix. Contrary to the indications given by this model, phenomena are often encountered that cannot be explained unless hydride platelets have considerable ostensible diffusibility in zircaloy.

In order to determine quantitatively the diffusion characteristics of hydrogen in zircaloy, a thermal diffusion experiment was performed with zircaloy-2 fuel cladding tubes containing hydrogen beyond the terminal solid solubility. In this experiment, a temperature difference of 20°–30°C was applied between the inside and outside surfaces of the specimen in a thermal simulator.

To explain the experimental results, a modified diffusion model is presented, in which the effects of stress are introduced into Markowitz's model with the diffusion of hydrogen in the hydride taken into account. The diffusion equation derived from this model can be written in a form that ostensibly represents direct diffusion of hydride in zircaloy. The apparent diffusion characteristics of the hydride at around 300°C are D_p = 2.3×10⁵exp(?32,000/RT), (where R: gas constant, T: temperature) and the apparent heat of transport Q _p *=?60,000 cal/mol. The modified diffusion model well explains the experimental results in such respects as reaches a steady state after several hours.     

2-D temperature distribution and heat flux of PFC in 2011 KSTAR campaign

KSTAR has reached a plasma current up to 630 kA, plasma duration up to 12 s, and has achieved high confinement mode (H-mode) in 2011 campaign. The heat flux of PFC tile was estimated from the temperature increase of PFC since 2010. The heat flux of PFC tiles increases significantly with higher plasma current and longer pulse duration. The time-averaged heat flux of shots in 2010 campaign (with 3 s pulse durations and I_p of 611 kA) is 0.01 MW/m² while that in 2011 campaign (with 12 s pulse duration and I_p of 630 kA) is about 0.02 MW/m². The heat flux at divertor is 1.4–2 times higher than that at inboard limiter or passive stabilizer. With the cryopump operation, the heat flux at the central divertor is higher than that without cryopump. The heat flux at divertor is proportional to, of course, the duration of H-mode. Furthermore, a software tool, which visualizes the 2D temperature distribution of PFC tile and estimates the heat flux in real time, is developed.     

An experimental investigation of a passive cooling unit for nuclear plant containment

A set of condensation experiments in the presence of noncondensables (e.g. air, helium) was conducted to evaluate the heat removal capacity of a passive cooling unit in a post-accident containment. Condensation heat transfer coefficients on a vertically mounted smooth tube have been obtained for total pressure ranging from 2.48×10⁵ Pa(abs) to 4.55×10⁵ Pa(abs) and air mass fraction ranging from 0.30 to 0.65. An empirical correlation for heat transfer coefficient (h), has been developed in terms of a parameter group made up of steam mole fraction (Xs), total pressure (P_t), temperature difference between bulk gas and wall surface (dT). This correlation covers all data points within 20%. All data points are also in good agreement with the prediction of the diffusion layer model (DLM) with suction and are approximately 2.2 times the Uchida heat transfer correlation. Experiments with an axial shroud around the test tube to model the restriction on radial flow experienced within a tube bundle demonstrated a reduction of the heat transfer coefficient by a factor of about 0.6. The effect of helium (simulating hydrogen) on the heat transfer coefficient was investigated for helium mole fraction in noncondensable gases (X_He/X_nc) at 15, 30 and 60%. It was found that the condensation heat transfer coefficients are generally lower when introducing helium into noncondensable gas. The difference is within 20% of air-only cases when X_He/X_nc is less than 30% and total pressure is less than 4.55×10⁵ Pa(abs). A gas stratification phenomenon was clearly observed for helium mole fraction in excess of 60%.     

Alpha Induced Reaction Cross Section Calculations of Tantalum Nucleus

The fusion energy is attractive as an energy source because the fusion will not produce CO₂ or SO₂ and so fusion will not contribute to environmental problems, such as particulate pollution and excessive CO₂ in the atmosphere. The fusion reaction does not produce radioactive nuclides and it is not self-sustaining, as is a fission reaction when a critical mass of fissionable material is assembled. Since the fusion reaction is easily and quickly quenched the primary sources of heat to drive such an accident are heat from radioactive decay and heat from chemical reactions. Both the magnitude and time dependence of the generation of heat from radioactive decay can be controlled by proper selection and design of materials. Tantalum is one of the candidate materials for the first wall of fusion reactors and for component parts of irradiation chambers. Accurate experimental cross-section data of alpha induced reactions on Tantalum are also of great importance for thermonuclear reaction rate determinations since the models used in the study of stellar nucleosynthesis are strongly dependent on these rates (Santos et al. in J Phys G 26:301, 2000). In this study, neutron-production cross sections for target nuclei ¹⁸¹Ta have been investigated up to 100 MeV alpha energy. The excitation functions for (α, xn) reactions (x = 1, 2, 3) have been calculated by pre-equilibrium reaction mechanism. And also neutron emission spectra for ¹⁸¹Ta (α, xn) reactions at 26.8 and 45.2 MeV have been calculated. The mean free path multiplier parameters has been investigated. The pre-equilibrium results have been calculated by using the hybrid model, the geometry dependent hybrid (GDH) model. Calculation results have been also compared with the available measurements in literature.     

Fuel Temperature Coefficients of Reactivity for Heavy-Water-Moderated,Cluster-Type Fuel Lattices

Temperature coefficients of reactivity have been measured up to 600°C on cluster-type UO₂ fuel for three kinds of ²³⁵U enrichment and on a hollow cluster of sus-cladding tubes by using a hot He gas loop in a heavy-water-moderated, pressure-tube-type critical assembly. A new experimental method has been developed which accurately eliminates the reactivity disturbance caused by heat leakage in the measurement of an extremely small change in reactivity. The fuel (fuel pellet, cladding land pressure-tube) temperature coefficients of reactivity obtained for the temperature range below 300°C are +1.00±0.04, ?3.48±0.13 and - 6.36±0.25 in the unit of l0^-5% Δk/k.°C for 0.2%, 0.7% and 1.5%²³⁵U enrichment, respectively. In the higher temperature region above 300°C, each coefficient shifts to positive side by about 2x10^-5 Δk/k.°C. Temperature coefficient of reactivity for the hollow cluster of sus-cladding tubes (cladding and pressure-tube) has a large constant value with positive sign, + (6.42±0.26) x 10^-5 Δk/k.°C, all through the temperature range. A calculational model to analyze a hot-loop-type measurement of temperature coefficients with use of WIMS-D code was proposed and could be successfully applied to the present measurement.     

Uranium Isotope Exchange between Uranium Hexafluoride and Uranium Pentafluoride

To aim at a better understanding of the uranium isotope exchange reaction between gaseous UF₆ and solid UF₅ experiments were done with natural UF₆ gas and solid UF₅ containing 3% ²³⁵U under different pressures of UF₆. The experimental results suggest a two-process reaction with an initial rapid increase of ²³⁵UF₆ in the gas phase followed by its slight and gradual increase. A rate equation based on a collision model is given for the two-process reaction which includes a primary exchange reaction on the solid surface and a secondary reaction participated by underlying UF₅ molecules. An analytical solution is provided for both of ²³⁵UF₆ concentration in the gas phase and ²³⁵UF₅ concentration on the solid surface, which is useful for determining the parameters characterizing the exchange reaction. A numerical analysis is also made to evaluate the influence of gas samplings. A remarkable agreement is found between the particle sizes of UF₅ estimated from the reaction parameter and from the direct observation with an electron microscope. The depletion of ²³⁵UF₅ concentration by the exchange reaction is very small when averaged over the whole solid UF₅, because the depletion is virtually limited to the solid surface due to the small reaction probability of underlying UF₅ molecules.     

Potential of a fusion–fission hybrid reactor using uranium for various coolants to breed fissile fuel for LWRs

In this study, neutronic performances of the (D,T) driven hybrid blankets, fuelled with UC₂ and UF₄, are investigated under first wall load of 5 MW/m². The fissile fuel zone is considered to be cooled with three coolants: gas (He or CO₂), flibe (Li₂BeF₄), and natural lithium. The behaviour of the UC₂ and UF₄ fuels are observed during 48 months for discrete time intervals of Δt=15 days and by a plant factor of 75%. At the end of the operation time, calculations have shown that Cumulative Fissile Fuel Enrichment (CFFE) values varied between 5 and 8.5% depending on the fuel and coolant type. The best enrichment performance is obtained in UF₄ fuelled blanket with flibe coolant, followed by gas and natural lithium coolant. CFFE reaches maximum value (8.51%) in UF₄ fuelled blanket (in row #1) and flibe coolant mode after 48 months. The lowest CFFE value (4.71%) is in UC₂ fuelled blanket (in row #8) and natural lithium coolant at the end of the operation period. This enrichment would be sufficient for LWR reactor. At the beginning of the operation, tritium breeding ratio (TBR) values were 1.090, 1.3301 and 1.2489 in UC₂ fuelled blanket and 1.0772, 1.2433 and 1.1533 in UF₄ fuelled blanket for flibe, natural lithium and gas coolant, respectively. At the end of the operation, TBR reach 1.1820, 1.3983 and 1.3138 in UC₂ fuelled blanket and 1.2041,1.3266 and 1.2407 in UF₄ fuelled blanket for flibe, natural lithium and gas coolant, respectively. Nuclear quality of the plutonium increases linearly during the operation period. The isotopic percentage of ²⁴⁰Pu is higher than 5% in UF₄ and UC₂ fuel with flibe coolant, so that the plutonium component in these modes can never reach a nuclear weapon grade quality during the operation period. This is very important factor for safeguarding. The isotopic percentage of ²⁴⁰Pu is lower than 5% in UC₂ fuel with gas and natural lithium coolant. In these modes, operation period must be increased to safeguarding.     

Neutron,Proton and Alpha Emission Spectra of Nickel Isotopes for Proton Induced Reactions

The fusion energy is attractive as an energy source because the fusion will not produce CO₂ or SO₂ and so fusion will not contribute to environmental problems, such as particulate pollution and excessive CO₂ in the atmosphere. The fusion reaction does not produce radioactive nuclides and it is not self-sustaining, as is a fission reaction when a critical mass of fissionable material is assembled. Since the fusion reaction is easily and quickly quenched the primary sources of heat to drive such an accident are heat from radioactive decay and heat from chemical reactions. Both the magnitude and time dependence of the generation of heat from radioactive decay can be controlled by proper selection and design of materials. Nickel (Ni) is an important structural material in fusion (and also fission) reactor technologies and many other fields. So, the working out the reaction cross sections of the Ni isotopes is very important for selection of the fusion materials. In this study, ⁵⁸Ni(p,xn), ⁵⁸Ni(p,xp), ⁶⁰Ni(p,xp), ⁶⁰Ni(p,xα) and ⁶²Ni(p,xp) reactions have been investigated using nuclear reaction models. And also the ⁵⁸Ni(p,xn) reaction has been calculated through a method of offered by Tel et al. The calculated results are discussed and compared with the experimental data taken from EXFOR database.     

Synthesis and Sintering of AlN using Enriched Nitrogen

A high purity Al powder was directly reacted with isotopically enriched ¹⁵N₂ gas at 1,200°C in a vacuum furnace to synthesize Al¹⁵N. In the initial reaction process, a small amount of AlF₃, was mixed with the Al powder to suppress excess temperature rise due to the heat of formation of the nitride. Then, the Al¹⁵N thus formed was used as the diluent for the second run, and the same procedure was repeated. A total of 100g of Al¹⁵N powder with ¹⁵N isotopic content 99.6 at% was formed with average gas efficiency of 75.6%. Then the Al¹⁵N powder was pressureless-sintered at 1,830°C in an argon atmosphere by adding a small amount of Y₂O₃ as the sintering agent. An Al¹⁵N ceramic plate with dimensions 50mm × 50mm × 1.6 mm and relative density 97.3% was obtained with the ¹⁵N isotopic content 99.2 at%. Some physical properties and stability in air of an Al¹⁵N ceramic plate were studied.     

Tests for “cold fusion” in the Pd-D2 and Ti-D2 systems at 40–380 MPa and −196-27°C

Experiments have been conducted on the Pd-D₂ and Ti-D₂ systems at 40–380 MPa and −196-27°C to investigate the possibility that “cold fusion” occurs in palladium and titanium deuterides generated by reaction with high-pressure D₂ gas. The experiments were performed using a 4.8 mm i.d. stainless steel pressure vessel that can be operated routinely at pressures as high as 400 MPa. In experiments completed to date, reactions between high-purity Pd or Ti and D₂ were monitored with: (1) an array of three BF₃ neutron detectors, (2) an internal, type-K sample thermocouple, and (3) an internal, type-K reference thermocouple located approximately 10 cm above the sample thermocouple. Using a²⁵²Cf source, the efficiency of the BF₃ detector array was determined to be approximately 6%. During experimentation, the three neutron detectors were immersed in a water bath thermostated at 27°C. The neutron count rate, D₂ pressure, sample and reference thermocouple readings, and bath temperature were recorded continuously at time intervals ranging from 6 seconds to 10 minutes. Experimental results obtained so far range from negative to potentially significant. No sustained heat production has been observed in any experiment. Thermal pulses that persist briefly after pressurizing Pd with D₂ gas are attributable to small amounts of chemical heat released when Pd and D₂ react to form palladium deuteride. No sustained neutron flux above background was observed in any Pd-D₂ experiment. On the other hand, in a Ti-D₂ experiment just completed, potentially significant results were obtained. During this experiment, there was a period of 5 consecutive hours when count rates rose to approximately 60 counts/hour above the average background rate. This detector count rate corresponds nominally to 1000 neutrons/hour emitted from the Ti-D₂ sample. However, due to several deficiencies in our neutron detection methods and equipment, we cannot demonstrate conclusively that our experimental data are valid. Consequently, we are upgrading our neutron detection equipment in preparation for a second, improved Ti-D₂ experiment.     

Experimental heat transfer of supercritical carbon dioxide flowing inside channels (survey)

A new reactor concept under development at AECL has the main design objective of achieving a 50% reduction in unit energy cost relative to existing reactor designs. The approach builds on using existing operating supercritical water (SCW) experience and turbines in coal-fired power plants.This SCW CANDU^®2 research includes investigating heat transfer and pressure drop at supercritical conditions using carbon dioxide as a modelling fluid as a cheaper and faster alternative to using SCW. Therefore, the objectives are to assess the work that was done with the supercritical carbon dioxide and to understand the specifics of heat transfer at these conditions.Our exhaustive literature search, which included over 450 papers, showed that the majority of experimental data were obtained in vertical tubes, some data in horizontal tubes and just few in other flow geometries.Three modes of heat transfer at supercritical pressures have been recorded: (1) so-called normal heat transfer, (2) improved heat transfer, characterized by higher-than-expected heat transfer coefficient (HTC) values than in the normal heat transfer regime and (3) deteriorated heat transfer, characterized by lower-than-expected HTC values than in the normal heat transfer regime.     

Depth profiling of hydrogen under an atmospheric pressure

Nuclear reaction analysis of hydrogen with a use of the ¹H(¹⁵N,αγ)¹²C reaction was performed under a atmospheric condition. A 100 nm-thick silicon nitride membrane coated with gold of 10 nm was used for the extraction of the ¹⁵N beam into the sample chamber filled with gas molecules. Hydrogen contained in a Y film with a thickness of 80 nm was detected in N₂ of 10⁵ Pa. This nuclear reaction analysis (NRA) setup was also applied to H₂ gas, and the yield curve revealed a plateau feature. The plateau level was, furthermore, found to be constant independent of the H₂ pressure. We show that this plateau intensity can be used to obtain the detection efficiency of a NRA setup.     

Release behavior of bred tritium from LiAlO2

The release behavior of bred tritium to the blanket purge gas is mainly controlled by such bulk phenomena as tritium forming reaction, diffusion of tritium in grain, interaction of tritium with irradiation defects, and absorption together with such surface phenomena as adsorption, isotope exchange reaction between molecular form hydrogen in purge gas and tritium on grain surface (isotope exchange reaction 1), isotope exchange reaction between water vapor and tritium on grain surface (isotope exchange reaction 2), and water formation reaction at addition of hydrogen. Following the observation of the present authors that the isotope exchange reaction 2 is much faster than the isotope exchange reaction 1, the release curve of bred tritium obtained at purge with humidified gas was used for estimation of the effective diffusivity of bred tritium in LiAlO₂. Then, the effective diffusivity of tritium in grain of LiAlO₂ is obtained as D_T = 2.5 × 10⁻⁷exp(−110 [kJ]/RT) [m²/s]. This equation gives the larger diffusivity than any other diffusivity presented so far because the mass transfer resistance at the grain surface is expected to be eliminated in the estimation procedure of this study.     

Transport of nuclear heat by means of chemical energy (nuclear long-distance energy)

Nuclear long-distance energy, i.e. the transportation of chemically bound energy, represents a potential application for process heat plants in which the endothermic reaction takes place at the heat source (high temperature reactor) whereas the exothermic back reaction occurs at the region of heat utilization (consumer). Due to the following criteria, i.e. reversibility of the chemical reaction, sufficiently large reaction enthalpy, favourable temperature region for the forward and back reactions, and the available technology, a combination of the methods of endothermic steam reforming of methane and exothermic methanation is chosen. As well as supplying household and industrial consumers with heating, process steam and electrical energy, an interconnected system with synthesis gas consumers (e.g. methanol production and iron ore reduction plants) is possible. It is shown that the amount of reactor heat which is convertible into long-distance energy depends considerably on the helium temperatures in the high temperature reactor and lies between 60 and 73% of the reactor power. Conceivable circuit schemes for the nuclear steam-reforming plants and the methanation plants are described. Finally, it is demonstrated, with the help of a simple model for cost estimations, that the nuclear long-distance energy system can make heating for households available in competition with oil heating and that due to the lower specific transport costs, for distances larger than 50 km it is also more economical than the hot water supply from the thermal power coupling of steam turbine plants using light water reactors (LWRs) or high temperature reactors (HTRs).     

Design and evaluation of an optimized W/Cu interlayer for W monoblock components

Divertor plasma-facing components of future fusion reactors should be able to withstand heat fluxes of 10-20 MW/m² in stationary operation. Tungsten blocks with an inner cooling tube made of CuCr1Zr, so-called monoblocks, are potential candidates for such water-cooled components. To increase the strength and reliability of the interface between the W and the cooling tube of a Cu-based alloy (CuCr1Zr), a novel advanced W-fibre/Cu metal matrix composite (MMC) was developed for operation temperatures up to 550 °C. Based on optimization results to enhance the adhesion between fibre and matrix, W fibres (W_f) were chemically etched, coated by physical vapour deposition with a continuously graded W/Cu_PVD interlayer and then heated to 800 °C. The W_f/Cu MMC was implemented by hot-isostatic pressing and brazing process in monoblock mock-ups reinforcing the interface between the plasma-facing material and the cooling channel. The suitability of the MMC as an efficient heat sink interface for water-cooled divertor components was tested in the high heat flux (HHF) facility GLADIS. Predictions from finite element simulations of the thermal behaviour of the component under loading conditions were confirmed by the HHF tests. The W_f/Cu MMC interlayer of the mock-ups survived cyclic heat loads above 10 MW/m² without any damage. One W block of each tested mock-up showed stable thermal behaviour at heat fluxes of up to 10.5 MW/m².     

Kinetics of the carbothermic reduction of a ThO2 + UO2 + C mixture to prepare (Th,U)C

Carbothermic reduction of mechanically mixed ThO₂ + UO₂2 + C compact to (Th, U)C has been studied in the temperature range between 1743 and 2043 K with emphasis on reaction kinetics. The rate-limiting step of this reaction was attributed to the diffusion of CO gas in the outermost layer of the reaction products. By X-ray diffraction, ThO₂ and UO₂ were found to react with graphite respectively to produce two nearly separate dicarbide phases, both of which then reacted with residual ThO₂ to form a monocarbide phase. An apparent activation energy of about 320 kJ mol⁻¹ was obtained for this carbothermic reduction. Such a high activation energy can be explained by the CO gas diffusion through micropores in that product layer, by taking into account the standard enthalpy changes of the related reactions to produce CO gas.     

Fusion reactor first-wall cooling for very high energy fluxes

A study has been made of the feasibility of a protective first-wall shield between the plasma and the containment vessel for early experimental controlled thermonuclear fusion machines. The proposed first-wall shield is a water-cooled array of thin-walled tubes designed to take very high local energy fluxes originating from the neutral beam injectors. Detailed computer calculations reveal that heat flux capabilities of 3300 W/cm² are possible with first-wall shield sections made up of tubes 1 m long of Ta-10W alloy (with tubes of 10 mm i.d. and tube wall thickness of 0.5 mm) with a structural safety factor of about four. Required pumping powers on the order of 1 MW/m² of first-wall area exposed to these high energy fluxes are predicted for flow in the non-boiling regime. If operation in the subcooled nucleate boiling regime can be achieved without oscillations or instabilities, the required pumping power is shown to decrease by about an order of magnitude.