Characterization of crud deposited on fuel rods under HWC environment in Kuosheng Nuclear Power Plant

Under normal water chemistry conditions, the oxygen and hydrogen peroxide produced by water radiolysis in the coolant of boiling water reactors(BWRs) can lead to intergranular stress corrosion cracking in the constituent materials of plant components. This fact has led to the wide-scale adoption of hydrogen water chemistry(HWC) in the nuclear industry to counteract these effects.This study seeks to characterize the metallic composition and the surface properties of the constituent materials of plant components in order to determine their effects on the accumulation of chalk river unidentified deposits(crud) on fuel rods in the BWR Unit-1 of the Kuosheng Nuclear Power Plant in Taiwan. Inductively coupled plasma-atomic emission spectroscopy was used to calculate the concentrations of surface crud and gamma spectrometry was used to determine the radioactivity of the corrosion products, as well as their axial distribution across the surface of the fuel rods. X-ray diffraction analysis and scanning electron microscopy/energy-dispersive X-ray spectroscopy were used to identify the crystalline phase and morphology of the crud as irregular shapes and flakes. The amount of crud deposited during the fourth fuel cycle exceeded that of the third fuel cycle due to extended burn-up time. Our analytical results indicate that the implementation of HWC had no significant effect on the characteristics of subsequent crud.

Development of a 1200 fine group nuclear data library for advanced nuclear systems

Accurate and reliable nuclear data libraries are essential for calculation and design of advanced nuclea systems. A 1200 fine group nuclear data library Hybrid Evaluated Nuclear Data Library/Fine Group(HENDL/FG with neutrons of up to 150 Me V has been developed to improve the accuracy of neutronics calculations and anal ysis. Corrections of Doppler, resonance self-shielding, and thermal upscatter effects were done for HENDL/FG Shielding and critical safety benchmarks were performed to test the accuracy and reliability of the library. The dis crepancy between calculated and measured nuclea parameters fell into a reasonable range.

Development and testing of the code for automatic generating of multi-temperature continuous-energy neutron cross section libraries

In the nuclear reactor design, a code for automatically generated multi-temperature continuous-energy neutron cross section data library, which is called AMTND for short, was designed and developed to meet the need of the reactor core design coupled with thermal-hydraulic design. The code can provide a point-wise crosssection at any temperature for a Monte Carlo neutron transport program, such as MCNP. In ensuring that the nuclear data produced by AMTND meets the testing of critical benchmark experiments, the time-consumed by the nuclear data generating of AMTND compared with NJOY's was carried out and the result shows the code's excellence. In order to test the accuracy of the code, the Doppler coefficient test benchmark was also carried out and the results verified the code preliminarily.

Neutronics analysis of a subcritical blanket system driven by a gas dynamic trap-based fusion neutron source for ^(99)Mo production

Gamma-emitting radionuclide ^(99m)Tc is globally used for the diagnosis of various pathological conditions owing to its ideal single-photon emission computed tomography (SPECT) characteristics.However,the short half-life of ^(99m)Tc (T_(1/2)=6 h)makes it difficult to store or transport.Thus,the production of ^(99m)Tc is tied to its parent radionuclide ^(99)Mo (T_(1/2)=66 h).The major production paths are based on accelerators and research reactors.The reactor process presents the potential for nuclear proliferation owing to its use of highly enriched uranium (HEU).Accelerator-based methods tend to use deuterium–tritium(D–T) neutron sources but are hindered by the high cost of tritium and its challenging operation.In this study,a new ^(99)Mo production design was developed based on a deuterium–deuterium (D–D) gas dynamic trap fusion neutron source (GDT-FNS) and a subcritical blanket system (SBS) assembly with a low-enriched uranium (LEU) solution.GDT-FNS can provide a relatively high-neutron intensity,which is one of the advantages of ^(99)Mo production.We provide a Monte Carlo-based neutronics analysis covering the calculation of the subcritical multiplication factor (k_(s)) of the SBS,optimization design for the reflector,shielding layer,and ^(99)Mo production capacity.Other calculations,including the neutron flux and nuclear heating distributions,are also provided for an overall evaluation of the production system.The results demonstrated that the SBS meets the nuclear critical safety design requirement (k_(s)<0.97) and maintained a high ^(99)Mo production capacity.The proposed system can generate approximately 157 Ci ^(99)Mo for a stable 24 h operation with a neutron intensity of 1×10^(14) n/s,which can meet 50%of China’s demand in 2025.

Stage-efficiency of centrifugal extractor used in nuclear industry

The stage efficiency of a single stage prototype ( d 70) centrifugal extractor and the cascade is tested by HNO 3 Nd 3+ and 30%TRPO kerosene system. The experimental results of the single stage centrifugal extractor show that the carryover of the two phases decreases with increasing ratio of the two flow rate and rotation rate and the stage efficiency increases with not only decreasing total flow rate but also increasing rotor speed. However, the experimental results of the cascade show that the average stage efficiency of the cascade increases with not only decreasing total flow rate but also increasing rotor speed in both three stage mode and two stage mode.

Energy and Exergy Analysis of a New Small Concentrating Solar Power Plant

A new small concentrating solar power plant which is suitable for urban area is presented, and a theoretical framework for the energy and exergy analysis in the overall power plant is also constructed. The framework can be used to evaluate the energy and exergy losses in each component. Furthermore, the energy and exergy efficiencies have also been computed and compared for the individual components as well as for the overall plant.

Numerical simulation of tritium behavior under a postulated accident condition for CFETR TEP system

China Fusion Engineering Test Reactor(CFETR)is China's self-designed and ongoing next-generation fusion reactor project.Tritium confinement systems in CFETR guarantee that the radiation level remains below the safety limit during tritium handling and operation in the fuel cycle system.Our tritium technology team is responsible for studying tritium transport behavior in the CFETR tritium safety confinement systems of the National Key R&D Program of China launched in 2017,and we are conducting CFETR tritium plant safety analysis by using CFD software.In this paper,the tritium migration and removal behavior were studied under a postulated accident condition for the Tokamak Exhaust Processing system of CFETR.The quantitative results of the transport behavior of tritium in the process room and glove box during the whole accident sequence(e.g.,tritium release,alarm,isolation,and tritium removal)have been presented.The results support the detailed design and engineering demonstration-related research of CFETR tritium plant.

Design and R＆D Progress of China Lead-Based Reactor for ADS Research Facility

In 2011,the Chinese Academy of Sciences launched an engineering project to develop an acceleratordriven subcritical system(ADS)for nuclear waste transmutation.The China Lead-based Reactor(CLEAR),proposed by the Institute of Nuclear Energy Safety Technology,was selected as the reference reactor for ADS development,as well as for the technology development of the Generation IV lead-cooled fast reactor.The conceptual design of CLEAR-I with 10 MW thermal power has been completed.KYLIN series lead-bismuth eutectic experimental loops have been constructed to investigate the technologies of the coolant,key components,structural materials,fuel assembly,operation,and control.In order to validate and test the key components and integrated operating technology of the lead-based reactor,the lead alloy-cooled non-nuclear reactor CLEAR-S,the lead-based zero-power nuclear reactor CLEAR-0,and the lead-based virtual reactor CLEAR-V are under realization.

Recovery of RE from Baotou rare earth concentrate with chlorination roasting

A process for recovery of RE from Baotou rare earth concentrate was developed by fixing the fluorine and chlorinating RE with ammonium chloride in the ore. The optimum conditions were determined as follows: fixing the fluorine of the ore 80 min with the MgO dosage m (ore) / m (MgO) =3∶1 at 600 ℃; chlorinating the fixed fluorine calcine 80 min, with NH 4Cl dosage m (NH 4Cl) / m (ore) =2∶1 at 500 ℃. The RE recovery reaches 85% under optimum conditions. The products of fixing fluorine with MgO were determined by X ray diffraction, and the mechanism of fixing fluorine was also discussed.

Effect of Calcium Leaching on the Properties of Cement-based Composites

Leaching is one of the major factors that alter the mechanical properties of cement-based composites.This study is aimed to investigate the effect of leaching on the properties of cement-based composites.Specimens with two water/cementitious ratios and two mineral admixtures were tested.An electrical potential was applied to accelerate the leaching process.Compressive strength test,scanning electronic microscopy,thermogravimetric analysis and X-ray diffraction analysis were conducted.Test results demonstrated that the calcium leaching reduced compressive strengths of concrete specimens,and such effect was prominent on the specimens without mineral admixtures.The leaching resistance increased with a decrease in water/cementitious ratio and an increase in amount of mineral admixtures.The mineral admixtures would reduce the amount of calcium hydroxide and refine the pore structure through pozzolanic reactions.A fair relationship was found between the calcium leaching and the compressive strength.

Performance evaluation of ultra-long lithium heat pipe using an improved lumped parameter model

Lithium heat pipes have broad applications in heat pipe cooling reactors and hypersonic vehicles owing to their ultra-high working temperature.In particular,when the length of the lithium heat pipe is ultra-long,the flow and heat transfer characteristics are more complex.In this study,an improved lumped parameter model that considers the Marangoni effect,bending effect,and different vapor flow patterns and Mach numbers was developed.Thereafter,the proposed model was verified using the University of New Mexico’s Heat Pipe and HTPIPE models.Finally,the verified model was applied to simulate the steady-state operation of an ultra-long lithium heat pipe in a Heat PipeSegmented Thermoelectric Module Converters space reactor.Based on the results:(1)Vapor thermal resistance was dominant at low heating power and decreased with increasing heating power.The vapor flow inside the heat pipe developed from the laminar to the turbulent phase,whereas the liquid phase in the heat pipe was always laminar.(2)The vapor pressure drop caused by bending was approximately 22–23%of the total,and the bending effect on the liquid pressure drop could be ignored.(3)The Marangoni effect reduced the capillary limit by hindering the liquid reflux,especially at low vapor temperatures.Without considering the Marangoni effect,the capillary limit of the lithium heat pipe was overestimated by 9%when the vapor temperature was 1400 K.(4)The total thermal resistance of the heat pipe significantly increased with increasing adiabatic length when the vapor temperature was low.Further,the wick dryness increased with increasing adiabatic length at any vapor temperature.Such findings improve on current knowledge for the optimal design and safety analysis of a heat pipe reactor,which adopts ultra-long lithium heat pipes.

Alpha radiolysis of nitric acid aqueous solution irradiated by 238Pu source

Alpha radiolysis of nitric acid aqueous solution by a ^(238)Pu source is investigated experimentally and theoretically.The time dependence of the nitrous acid yield on dose rate,nitric acid concentration,and nitrate ion concentration is studied.A novel kinetic model for the α-radiolysis of nitric acid aqueous solution is established,by considering the direct and indirect effects.The simulation results agree well with the experimental data,indicating the validity of our model to treat the reaction paths for generation and consumption of nitrous acid.It is shown that the redox reactions involving Pu cannot be neglected in theα-radiolysis of the solution.The results provide a better understanding of the α-ray radiolysis of aqueous nitric acid.

Flow characteristics of natural circulation in a lead–bismuth eutectic loop

Lead and lead-alloys are proposed in future advanced nuclear system as coolant and spallation target.To test the natural circulation and gas-lift and obtain thermal-hydraulics data for computational fluid dynamics(CFD) and system code validation, a lead–bismuth eutectic rectangular loop, the KYLIN-Ⅱ Thermal Hydraulic natural circulation test loop, has been designed and constructed by the FDS team. In this paper, theoretical analysis on natural circulation thermal-hydraulic performance is described and the steady-state natural circulation experiment is performed. The results indicated that the natural circulation capability depends on the loop resistance and the temperature and center height differences between the hot and cold legs. The theoretical analysis results agree well with,while the CFD deviate from, the experimental results.

Adsorption and desorption of hydrogen on/from single-vacancy and double-vacancy graphenes

Adsorption and desorption of hydrogen on/from single-vacancy and double-vacancy graphenes were studied by means of first-principles calculations. The structure and stability of continuous hydrogenation in single vacancy were investigated. Several new stable structures were found, along with their corresponding energy barriers. In double-vacancy graphene, the preferred sites of H atoms were identified, and H2 molecule desorption and adsorption of from/on were calculated from the energy barriers. This work provides a systematic and comprehensive understanding of hydrogen behavior on defected graphene.

Diffusion characteristics of HTO and 99TcO_4^- in compacted Gaomiaozi(GMZ) bentonite

The characteristics of diffusion are essential to the transport of radionuclides through buffer/backfill materials, such as bentonite, which are commonly found in waste repositories. This study used through-diffusion techniques to investigate the diffusion behavior of HTO and ^(99)TcO_4^- on GMZ bentonite of various densities. Diffusion rates were calculated by measuring the diffusion coefficients(De, Da), plotting breakthrough curves and interpreting experiment data. The apparent and effective diffusion coefficients of HTO ranged from(1.68 ± 0.40) 9 × 10^(-11) to(2.80 ± 0.62) 9 × 10^(-11) m^2/s and from(4.61 ±1.28) 9 × 10^(-12) to (16.2 ± 2.50) 9 × 10^(-12) m^2/s, respectively.The apparent and effective diffusion coefficients of^(99)TcO_4^-ranged from(5.26 ± 0.16) 9 × 10^-12to(7.78 ± 0.43) 9× 10^-12m^2/s and from(1.49 ± 0.002) 9 × 10^(-12) to(4.16 ±0.07) 9 × 10^(-12) m^2/s, respectively. The distribution coefficients of HTO and^(99)TcO_4^-ranged from(0.70 ± 0.12) 9× 10^(-2) to(1.36 ± 0.53) 9 × 10^(-2) mL/g and from(1.12 ±0.06) 9 × 10^(-2) to(5.79 ± 2.22) 9 × 10^(-2) mL/g, respectively.The Deand Kdvalues were shown to decrease with an increase in the bulk dry density of compacted bentonite. Our results show that HTO and ^(99)Tc could be considered nonsorbent radionuclides. The data obtained in this studyprovide a valuable reference for the safety assessment of waste repositories.

A method for direct conversion of EPID images to incident fluence for dose reconstruction

A direct incident fluence measurement method based on amorphous silicon electronic portal imaging device(a-Si EPID) has been developed for pretreatment verification of intensity-modulated radiation therapy(IMRT).The EPID-based incident fluence conversion method deconvolves EPID images to the primary response distribution based on measured lateral scatter kernels in the EPID detector using Conjugate Gradient algorithm.The primary response is converted to the incident fluence based on measured fluence conversion matrix which corrects for off-axis position dependence of the a-Si EPID response and the "horn" beam profile caused by flatting filter. To verify feasibility and accuracy of this method, square fields of various sizes and two IMRT plans were delivered. The dose distributions computed based on EPID-derived incident fluence were compared with the measurement data. For all square field sizes except the smallest field(2 cm), the mean dose differences in cross-line dose profiles were within 1% excluding the penumbra region, and gamma passing percentages with a 2%/2 mm criterion were about 99%. For two IMRT plans, the least gamma passing percentage for all eight IMRT fields was 98.14% with 2%/3 mm criteria. It can be concluded that our direct EPID-based incident fluence conversion method is accurate and capable of being applied to pretreatment dose verification in clinical routines.

Preparation of ZrO_2 nano-particles by the hydrolysis of ZrOCl_2 solution in the reverse micelles

Zirconia nano-particles have been produced by the hydrolysis of ZrOCl2 solution in the reverse micelles of a liquid-liquid two-phase system, in which sodium bis（2-ethylhexyl） sulfosuccinite （AOT） and toluene were chosen as the surfactant and organic phase, respectively. The reverse micelles prevented the aggregation of primary particles and reduced the diameters of zirconia nanoparticles. Superfine zirconia powders soft-aggregated by the zirconia nano-particles were obtained. The diameters of zirconia nanoparticles were influenced by the quantity of the surfactant.

The performance of Pt/air oxygen sensors in stagnant Pb-Bi eutectic at high temperatures

Oxygen control technology is a critical issue for compatibility of candidate structural materials with liquid lead-bismuth eutectic(LBE) in accelerator driven systems. Performances of a self-developed Pt/air sensor and another one from Karlsruher Institute of Technology(KIT) were tested in stagnant oxygen-saturated liquid LBE.Calibrations showed that the trend and values of corrected electromotive force(EMF) of the self-developed sensor, with a bias voltage of 20 mV, were consistent with theoretical results above 425℃, and similar results were obtained in cross-calibration test with EMF value of KIT sensor as reference. In stability test at 450℃ for 100 hours, the KIT sensor performed better than the self-developed one, which showed signal fluctuations.Both sensors exhibited quick response to temperature variations in the responsiveness test.

Influence of graphite particle size and its shape on performance of carbon composite bipolar plate

Bipolar plates for proton exchange membrane fuel cell (PEMFC) where polymer is used as binder and graphite is used as electric filler were prepared by means of compression molding technology. Study on the effects of graphite particle size and shape on the bipolar plate performance, such as electrical conductivity, strength, etc. showed that with decrease of graphite particle size, bulk electrical conductivity and thermometric conductivity decreased, but that flexural strength was enhanced. After spherical graphite occurrence in flake-like form, the flexural strength of the bipolar plate was enhanced, electrical conductivity increased but thermal conductivity decreased in direction paralleling pressure direction, and both electrical conductivity and thermometric conductivity reduced in direction perpendicular to pressure direction.