Separation and determination of americium in low-level alkaline waste of NPP origin

The aim of this work is to develop a short and cost-saving procedure for the determination of ²⁴¹Am in sludge sample of the alkaline low-level radioactive waste (LL LRAW) collected from Nuclear Power Plant “Kozloduy”. The determination of americium was a part of a complex analytical approach, where group actinide separation was achieved. An anion exchange was used for separation of americium from uranium, plutonium and iron. For the separation of americium extraction with diethylhexyl phosphoric acid (DEHPA) was studied. The final radioactive samples were prepared by micro co-precipitation with NdF₃, counted by alpha and gamma spectrometry. The procedure takes 2 hours. The recovery yield of the procedure amounts to (95 ± 1.5)% and the detection limit is 53 mBq/kg ²⁴¹Am (t=150 000 s). The analytical procedure was applied for actual liquid wastes and results were compared to standard procedure.     

A conceptual high flux reactor design with scope for use in ADS applications

A 100 MWt reactor design has been conceived to support flux level of the order of 10¹⁵ n/cm²/s in selected flux trap zones. The physics design considers high enriched metallic alloy fuel in the form of annular plates placed in a D₂O moderator tank in a hexagonal lattice arrangement. By choosing a tight lattice pitch in the central region and double the lattice pitch in the outer region, it is possible to have both high fast flux and thermal flux trap zones. By design the flux level in the seed fuel has been kept lower than in the high flux trap zones so that the burning rate of the seed is reduced. Another important objective of the design is to maximize the time interval of refueling. As against a typical refueling interval of a few weeks in such high flux reactor cores, it is desired to maximize this period to as much as six months or even one year. This is possible to achieve by eliminating the conventional control absorbers and replacing them with a suitable amount of fertile material loading in the reactor. Requisite number of seedless thorium-aluminum alloy plates are placed at regular lattice locations vacated by seed fuel in alternate fuel layers. It is seen that these thorium plates are capable of acquiring asymptotic fissile content of 14 g/kg in about 100 days of irradiation at a flux level of 8 × 10¹⁴ n/cm²/s. In summary, the core has a relatively higher fast flux in the central region and high thermal flux in the outer region. The present physics design envisages a flat core excess reactivity for the longest possible cycle length of 6 months to one year. It is also possible to modify the design for constant subcriticality for about the same period or longer duration by considering neutron spallation source at the centre and curtailing the power density in the inner core region by shielding it with a layer of thoria fuel loading.      

Neutronic simulation of a research reactor core of (232Th, 235U)O 2 fuel using MCNPX2.6 code

The small reactor design for the remote and less developed areas of the user countries should have simple features in view of the lack of infra-structure and resources. Many researchers consider long core life with no on-site refuelling activity as a primary feature for the small reactor design. Long core life can be achieved by enhancing internal conversion rate of fertile to fissile materials. For that purpose, thorium cycle can be adopted because a high fissile production rate of ²³³U converted from ²³²Th can be expected in the thermal energy region. A simple nuclear reactor core arranged 19 assemblies in hexagonal structure, using thorium-based fuel and heavy water as coolant and moderator was simulated using MCNPX2.6 code, aiming an optimized critical assembly. Optimized reflector thickness and gap between assemblies were determined to achieve minimum neutron leakage and void reactivity. The result was a more compact core, where assemblies were designed having 19-fuel pins in 1.25 pitch-to-diameter ratio. Optimum reflector thickness of 15 cm resulted in minimal neutron leakage in view of economic limitations. A 0.5 cm gap between assembles achieved more safety and 2.2% enrichment requirements. The present feasibility study suggests a thermal core of acceptable neutronic parameters to achieve a simple and safe core.     

Operation of CANDU power reactor in thorium self-sufficient fuel cycle

This paper presents the results of calculations for CANDU reactor operation in thorium fuel cycle. Calculations are performed to estimate the feasibility of operation of heavy-water thermal neutron power reactor in self-sufficient thorium cycle. Parameters of active core and scheme of fuel reloading were considered to be the same as for standard operation in uranium cycle. Two modes of operations are discussed in the paper: mode of preliminary accumulation of ²³³U and mode of operation in self-sufficient cycle. For the mode of accumulation of ²³³U it was assumed for calculations that plutonium can be used as additional fissile material to provide neutrons for ²³³U production. Plutonium was placed in fuel channels, while ²³²Th was located in target channels. Maximum content of ²³³U in target channels was estimated to be ∼13 kg/t of ThO₂. This was achieved by irradiation for six years. The start of the reactor operation in the self-sufficient mode requires ²³³U content to be not less than 12 kg/t. For the mode of operation in self-sufficient cycle, it was assumed that all channels were loaded with identical fuel assemblies containing ThO₂ and certain amount of ²³³U. It is shown that nonuniform distribution of ²³³U in fuel assembly is preferable.      

Natural Transmutation of Actinides via the Fission Reaction in the Closed Thorium–Uranium–Plutonium Fuel Cycle

It is shown for a closed thorium–uranium–plutonium fuel cycle that, upon processing of one metric ton of irradiated fuel after each four-year campaign, the radioactive wastes contain ~54 kg of fission products, ~0.8 kg of thorium, ~0.10 kg of uranium isotopes, ~0.005 kg of plutonium isotopes, ~0.002 kg of neptunium, and “trace” amounts of americium and curium isotopes. This qualitatively simplifies the handling of high-level wastes in nuclear power engineering.     

Fuel quality and operational issues for polymer electrolyte membrane (PEM) fuel cells

Polymer electrolyte membrane (PEM) fuel cells are susceptible to degradation due to the catalyst poisoning caused by CO present in the fuel above certain limits. Although the amount of CO in the fuel may be within the permissible limit, the fuel composition (% CO₂, CH₄, CO and H₂O) and the operating conditions of the cell (level of gas humidification, cell temperature and pressure) can be such that the equilibrium CO content inside the cell may exceed the permissible limit leading to a degradation of the fuel cell performance. In this study, 50 cm² active area PEM fuel cells were operated at 55–60 °C for periods up to 250 hours to study the effect of methane, carbon dioxide and water in the hydrogen fuel mix on the cell performance (stability of voltage and power output). Furthermore, the stability of fuel cells was also studied during operation of cells in a cyclic dead end / flow through configuration, both with and without the presence of carbon dioxide in the hydrogen stream. The presence of methane up to 10% in the hydrogen stream showed a negligible degradation in the cell performance. The presence of carbon dioxide in the hydrogen stream even at 1–2% level was found to degrade the cell performance. However, this degradation was found to disappear by bleeding only about 0.2% oxygen into the fuel stream.     

Mid-infrared laser-absorption diagnostic for vapor-phase fuel mole fraction and liquid fuel film thickness

A novel two-wavelength mid-infrared laser-absorption diagnostic has been developed for simultaneous measurements of vapor-phase fuel mole fraction and liquid fuel film thickness. The diagnostic was demonstrated for time-resolved measurements of n-dodecane liquid films in the absence and presence of n-decane vapor at 25°C and 1 atm. Laser wavelengths were selected from FTIR measurements of the C–H stretching band of vapor n-decane and liquid n-dodecane near 3.4 μm (3000 cm⁻¹). n-Dodecane film thicknesses <20 μm were accurately measured in the absence of vapor, and simultaneous measurements of n-dodecane liquid film thickness and n-decane vapor mole fraction (300 ppm) were measured with <10% uncertainty for film thicknesses <10 μm. A potential application of the measurement technique is to provide accurate values of vapor mole fraction in combustion environments where strong absorption by liquid fuel or oil films on windows make conventional direct absorption measurements of the gas problematic.     

On the optimal void fraction in the thermodynamics of a simple liquid

The free energy of a crystal containing a given void fraction is derived in terms of the truncated interparticle Lennard-Jones potential. The free energy is minimized over the void fraction at constant pressure and temperature. It is shown that for all pressures the optimal void fraction remains less than 1% as the temperature is raised. However, at some temperature it grows suddenly and reaches values of the order of the percolation level for voids in a crystal, ∼0.125. At this point, the crystal transitions to the liquid state takes place. The derived dependence of the transition temperature on the pressure—the melting curve—is in good agreement with experimental data on the melting of solidified inert gases. Zh. éksp. Teor. Fiz. 116, 1375–1385 (October 1999)     

Lower concentration limit of carbon filtration combustion

For mixtures of carbon materials and an inert filler, dependences of the characteristics of the filtration combustion wave on the gaseous oxidizer supply rate at a fuel content in the mixture of less than 7 wt % were obtained. The existence of a lower concentration limit for a steady-state filtration combustion wave was established. It was demonstrated that at a given intensity of heat loss, the concentration limits are determined by the reactivity of the carbon material and the oxidizer supply rate. At the effective coefficient of heat loss α = 8 W/(m² K), effective conductivity of mixture material λ = 2 W/(m K), and air supply rate G = 0.1 m/s, the lowest fraction of carbon in the mixture at which combustion is still possible was 4.5 wt % for carbon-carbon composite, 2.5 wt % for activated birch coal, and 2.0 wt % for birch coal, the most reactive kind of carbon.     

Physics design of a CW high-power proton Linac for accelerator-driven system

Accelerator-driven systems (ADS) have evoked lot of interest the world over because of their capability to incinerate the MA (minor actinides) and LLFP (long-lived fission products) radiotoxic waste and their ability to utilize thorium as an alternative nuclear fuel. One of the main subsystems of ADS is a high energy (∼1 GeV) and high current (∼30 mA) CW proton Linac. The accelerator for ADS should have high efficiency and reliability and very low beam losses to allow hands-on maintenance. With these criteria, the beam dynamics simulations for a 1 GeV, 30 mA proton Linac has been done. The Linac consists of normal-conducting radio-frequency quadrupole (RFQ), drift tube linac (DTL) and coupled cavity drift tube Linac (CCDTL) structures that accelerate the beam to about 100 MeV followed by superconducting (SC) elliptical cavities, which accelerate the beam from 100 MeV to 1 GeV. The details of the design are presented in this paper.     

Study of europium and selected actinides uptake on composite material CMPO-PAN

Modified CMPO-PAN production was used and the resulting composite material was tested for purposes of extraction chromatography. A commercially avaialble extraction agent octyl(phenyl)-N,N’-diisobutylcarbamoylmethylphosphine oxide (CMPO) and polyacrylonitrile (PAN) were used. The europium uptake kinetics by composite material was studied in two different nitric acid solutions. The dependences of weight distribution coefficients (D_g) of europium, americium, plutonium, uranium, and neptunium on nitric acid concentration (0.01–5 mol L⁻¹) in the presence of sodium nitrate (0.1 mol L⁻¹) were determined. High D_g-values were found in 0.1–5 M nitric acid for all elements tested. Increase in europium and americium D_g-values with decrease in nitric acid concentration (bellow 0.1 mol L⁻¹) was observed. This behaviour in diluted nitric acid solutions differs from the behaviour of the similar materials with another support. CMPO-PAN composite material was compared with commercially available TRU Resin. D_g-values of the tested elements in all solutions used were higher for CMPO-PAN than for TRU Resin. The retention of the studied elements on CMPO-PAN in hydrochloric acid was made as a screening study.     

Electrochemical activities of yttrium doped spinel LiMn2O4

It was found for the first time that the catalysis of yttrium doping of spinel LiMn₂O₄ can enhance the electrochemical activities of manganese, leading to both improvement of electrochemical capacity and reactivity with the electrolyte of manganese. A proper amount of doping was 0.5%, and such yttrium-doped sample, Li(Y_0.005Mn_0.995)₂O₄, had an initial capacity of 130 mAh g⁻¹ over that of the undoped one with the capacity retention to reach 92.3% exceeding that of the undoped one at 100th cycle.     

Rapid methods for the determination of long-lived radionuclides in environmental samples by ICP-SFMS and radioanalytical techniques

Two improved sample preparation methods for the determination of americium and plutonium (Method 1) and plutonium (Method 2) from environmental samples by inductively coupled plasma sector field mass spectrometry (ICP-SFMS) and alpha spectrometry are presented. Both procedures involve a rapid CaF₂ co-precipitation step for pre-concentration and matrix removal followed by extraction chromatographic separations. The average recovery after sample preparation was better than 85 % for both americium and plutonium. The method developed also focused on the elimination of possible interferences in the mass spectrometric analysis caused by molecular ions (e.g. ²⁰⁸Pb¹⁶O ₂ ⁺ or ²³⁸U¹H⁺) by employing suitable matrix separation prior to ICP-SFMS analysis of the desolvated sample. Isotopes with alpha energies similar to the analytes that may cause interferences in alpha spectrometric analysis were also separated. For ²³⁹Pu, ²⁴⁰Pu, ²⁴¹Pu and ²⁴¹Am detection limits of 15, 9.2, 14 and 23 fg g⁻¹, respectively were achieved by ICP-SFMS, and 0.1 mBq obtained by alpha spectrometry. The methods developed are especially applicable for monitoring purposes of anthropogenic transuranium elements, as the analysis (sample preparation and ICP-SFMS measurement) can be carried out within 6 hours for one batch of samples.     

The thorium cycle for fast breeder reactors

The role that could be played by liquid metal-cooled fast breeder reactors (LMFBRs) in the utilization of India’s considerable thorium resources is reviewed in this article. Distinct advantages of thorium-based fuels over plutonium-uranium fuels in LMFBRs pertain to a more favourable coolant voiding reactivity coefficient and better fuel element irradiation stability. The poorer breeding capability of thorium-fuelled fast reactors can in principle be overcome by improved core design and development of advanced fuel concepts. The technical feasibility of such advanced thorium fuels and core designs must be established by sustained research and development. It is also necessary to efficiently close the thorium fuel cycle of fast breeder reactors by appropriate development of the fuel reprocessing and refabrication stages. The Fast Breeder Test Reactor (FBTR) at Kalpakkam is expected to be an important tool for development of thorium fuel and fuel cycle technology. A quick look at the economics of the thorium cycle for fast reactors, vis-a-vis the more conventional uranium cycle indicates only a small and acceptable cost disadvantage on account of the need for remote fabrication of recycled thorium fuel. The authors felicitate Prof. D S Kothari on his eightieth birthday and dedicate this paper to him on this occasion.     

Prospects for transmutation of nuclear waste and associated proton-accelerator technology

While a considerable and world-wide growth of the nuclear share in the global energy mix is desirable for many reasons, a major concern or objection is the long-term burden that is constituted by the radiotoxic waste from the spent fuel. The concept of Partitioning & Transmutation, a scientific and technological answer, is therefore of high interest. Its deployment may use dedicated “Transmuter” or “Burner” reactors, using a fast neutron spectrum. For the transmutation of waste with a large content (up to 50%) of (very long-lived) Minor Actinides, a sub-critical reactor, using an external neutron source is a solution of high interest. It is constituted by coupling a proton accelerator, a spallation target and a subcritical core. This promising new technology is named ADS, for accelerator-driven system. The present paper aims at an introduction into the field in order to focus, in its later part, on the development of the required accelerator technology.     

Development and Analysis of the Novel Hybridization of a Single-Flash Geothermal Power Plant with Biomass Driven sCO2-Steam Rankine Combined Cycle

This study investigates the hybridization scenario of a single-flash geothermal power plant with a biomass-driven sCO₂-steam Rankine combined cycle, where a solid local biomass source, olive residue, is used as a fuel. The hybrid power plant is modeled using the simulation software EBSILON^®Professional. A topping sCO₂ cycle is chosen due to its potential for flexible electricity generation. A synergy between the topping sCO₂ and bottoming steam Rankine cycles is achieved by a good temperature match between the coupling heat exchanger, where the waste heat from the topping cycle is utilized in the bottoming cycle. The high-temperature heat addition problem, common in sCO₂ cycles, is also eliminated by utilizing the heat in the flue gas in the bottoming cycle. Combined cycle thermal efficiency and a biomass-to-electricity conversion efficiency of 24.9% and 22.4% are achieved, respectively. The corresponding fuel consumption of the hybridized plant is found to be 2.2 kg/s.     

The optimal path of piston motion for Otto cycle with linear phenomenological heat transfer law

An Otto cycle engine with internal and external irreversibilities of friction and heat leakage, in which the heat transfer between the working fluid and the environment obeys linear phenomenological heat transfer law [q ∝△(T -1)], is studied in this paper. The optimal piston motion trajectory for maximizing the work output per cycle is derived for the fixed total cycle time and fuel consumed per cycle. Optimal control theory is applied to determine the optimal piston trajectories for the cases of with and w...     

Improvement of plasma-sprayed YSZ electrolytes for solid oxide fuel cells by alumina addition

Atmospheric plasma spray is a fast and economical process for deposition of yttria-stabilized zirconia (YSZ) electrolyte for solid oxide fuel cells. YSZ powders have been used to prepare plasma-sprayed thin ceramic films on the metallic substrate employing plasma spray technology at atmospheric pressure. Alumina doping was employed to improve the structural characteristics and electrical properties of YSZ. The effect of alumina addition from 1 to 5 wt.% on the properties of plasma-sprayed YSZ films was investigated. It was found that the gas permeability of the Al-doped YSZ electrolyte layer reached a level of 8.6 × 10⁻⁷ cm⁴ gf⁻¹ s⁻¹, which is a necessary value for the practical operation of solid oxide fuel cells. Alumina doping considerably increased the ionic conductivity of plasma-sprayed YSZ. The open circuit voltage of the alumina-doped YSZ coating was approximately equal to the theoretical value for dense YSZ material.     

Formation of charmonium states in heavy-ion collisions and thermalization of charm

We examine the possibility to utilize in-medium charmonium formation in heavy-ion interactions at collider energy as a probe of the properties of the medium. This is possible because the formation process involves recombination of charm quarks which imprints a signal on the resulting normalized transverse momentum distribution containing information about the momentum distribution of the quarks. We have contrasted the transverse momentum spectra of J/ψ, characterized by 〈p _T ²〉, which result from the formation process in which the charm quark distributions are taken at opposite limits with regard to thermalization in the medium. The first uses charm quark distributions unchanged from their initial production in a pQCD process, appropriate if their interaction with the medium is negligible. The second uses charm quark distributions which are in complete thermal equilibrium with the transversely expanding medium, appropriate if a very strong interaction between charm quarks and medium exists. We find that the resulting 〈p _T ²〉 of the formed J/ψ should allow one to differentiate between these extremes, and that this differentiation is not sensitive to variations in the detailed dynamics of in-medium formation. We include a comparison of predictions of this model with preliminary PHENIX measurements, which indicates compatibility with a substantial fraction of in-medium formation.     

Hybrid fusion–fission reactor with a thorium blanket: Its potential in the fuel cycle of nuclear reactors

Discussions are currently going on as to whether it is suitable to employ thorium in the nuclear fuel cycle. This work demonstrates that the ²³¹Pa–²³²U–²³³U–Th composition to be produced in the thorium blanket of a hybrid thermonuclear reactor (HTR) as a fuel for light-water reactors opens up the possibility of achieving high, up to 30% of heavy metals (HM), or even ultrahigh fuel burnup. This is because the above fuel composition is able to stabilize its neutron-multiplying properties in the process of high fuel burnup. In addition, it allows the nuclear fuel cycle (NFC) to be better protected against unauthorized proliferation of fissile materials owing to an unprecedentedly large fraction of ²³²U (several percent!) in the uranium bred from the Th blanket, which will substantially hamper the use of fissile materials in a closed NFC for purposes other than power production.