bluuebunny/crossref_metadata_2025_split huggingface dataset:
Thermal Analysis of the Sensor in a Radiation Budget Instrument
Authors: Pfab Jonathan F. , Vick Brian , Mahan J. Robert , Priestley Kory 
Published: October 2017

Earth radiation budget instruments (RBI) are devices designed to study global climate change. These instruments use telescopes mounted on low earth orbit satellites to measure emitted and reflected solar radiation from the earth. Radiation is measured by virtue of temperature changes caused by absorbed radiation from the earth scans on the surface of a delicate gold-black detector. In this paper a thermal model of the detector in a typical radiation budget instrument is formulated.
A numerical solution is developed using a complex model building procedure. The idea is to split complex physical processes into simpler, individual physical processes. The basic procedure is to solve for each individual physical process in a numerically stable and efficient manner, and then assemble these processes in a cascading sequence to form a complete numerical solution of a complex model. The major advantage is that complicated mathematical models can be solved as a sequence of much simpler and less computationally intensive processes.
Parameter studies are performed on the numerical accuracy, conduction effects, initial conditions, boundary conditions due to contact with adjoining surfaces, radiation exchange with surfaces optically visible to the detector, and the volumetric heat source due to absorbed radiation from the earth scene. Extremely accurate temperature predictions are required due to the low signal to noise ratio. It is found that predictions are most sensitive to the amount and distribution of irradiation on the detector surface, which is computed independently using a Monte Carlo ray trace method.
Convective Heat Transfer of Ethanol/Polyalphaolefin Nanoemulsion Inside Circular Minichannel Heat Exchanger
Authors: Zewede Fana , Argaw Henok , Tran Thanh , Xu Jiajun 
Published: October 2017

This work experimentally studied the convective heat transfer characteristics of a novel nanostructured heat transfer fluid: “Ethanol/Polyalphaolefin(PAO) nanoemulsion fluids” flowing through a heat exchanger made of twelve circular minichannels. Ethanol/PAO nanoemulsion fluid is a thermodynamically stable system formed by dispersing ethanol into a mixture of PAO and surfactants, in which the ethanol added inside forms self-assembled nanodroplets of tens of nanometers in diameter. These ethanol nanodroplets can serve as pre-seed boiling nuclei at elevated temperature. The Reynolds number was varied between 140 and 1200 to maintain the entire range of flow regime remained at laminar flow for both single- and two-phase convective heat transfer experiments. Pure PAO was also tested under same conditions and used as baseline data for comparison.
It is found that: for single phase flow, there is no significant increase in Nusselt number of Ethanol/PAO nanoemulsion compared to that of PAO fluid in laminar flow regime. However, when the nucleation of ethanol nanodroplets inside the nanoemulsion fluid was initiated, it showed a substantial increase in heat transfer coefficient compared to that of PAO fluid: a 75% enhancement can be achieved under current test conditions. While its mechanism is not completely clear yet, it is believed that such an effect is likely related to the latent heat carried by ethanol bubbles, as well as the increased turbulence and mixing generated during the two-phase flow of nanoemulsion which can increase the heat transfer rate.
Numerical Investigation on the Gas Filling Process of Hydrogen-Air-Steam Mixtures in a Semi-Confined Experimental Pipe
Authors: Li Min , Xu Youyou , Wen Xiaojian , Liu Songlin , Luo Guangnan 
Published: October 2017

Hydrogen gathering in the containment may occur followed by a severe accident in a nuclear power plant. A flammable mixture can be formed when hydrogen is mixed with air. The ignition of the gas mixture could threaten the integrity of the containment. In order to provide technology base and experiment data for optimization of hydrogen safety technology of Chinese advanced pressurized water reactor CAP1400, a major project regarding hydrogen safety research of pressurized water reactor containment is underway. As an important part of the project, an experimental facility (A4Q-DH) for the study of hydrogen combustion will be built. Gas displacement method is used to filling the premixed hydrogen-air-steam mixtures into the experimental pipe. Flow behaviors of the gases in the pipe are complicated because fluid flow can be disturbed by the built-in obstacles and gas density can be changed with variation of gas composition concentration. Therefore, it is necessary to evaluate the effectiveness of the gas filling method.
In this paper, gas filling processes for the experimental pipe with different obstacles and gas composition concentrations were simulated using computational fluid dynamics software ANSYS Fluent. The results indicated that hydrogen-air-steam mixtures can be uniformly distributed in the experimental pipe within tens of seconds. The obstacles with modest blockage ratio in the pipe are conducive to shorten the required gas filling time. The hindering effect of annular obstacles is greater than the one of circular and square obstacles. The time required for air to achieve uniformly distribution increases with the increase of the inlet concentration of steam and hydrogen. However, the time required for hydrogen and steam to be evenly distributed in the pipe are relatively close regardless of the shape and blockage ratio of obstacle and the inlet gas concentration.
A New Numerical Method for Fuel Temperature Calculation
Authors: Xin Jin , Liu Xiaohan , Wei Xiaoyan 
Published: October 2017

For most fuel rod codes, the time independent heat conduction equation, which is a steady heat conduction equation, is applied in fuel temperature calculation. However, it can affect the fuel temperature prediction in II condition, which the linear power has much change in some seconds. For improving the fuel temperature prediction in II condition, this paper gives a new numerical method, which combines classical thermal conduction integration method and the difference applied in time partial derivative. For guaranteeing the numerical method’s stability and convergence rate, the multi-dimension Newton-Raphson procedure are applied in fuel temperature calculation. This paper describes the theoretical deduction of the numerical method, and Halden fuel thermal conductivity model applied in fuel temperature calculation.
In order to verify new numerical method’s correctness, stability and convergence rate, the comparison between numerical solution and analytic solution is performed in 4 hypothetical conditions that the power transient duration is respectively 3s, 15s, 30s and 120s, the linear power changes from 15kW/m to 45 kW/m, and the fuel pellet surface temperature changes from 400 degree to 750 degree. And fuel density, specific heat and thermal conductivity are assumed as constants so that there exists analytic solution in this condition. The 4 hypothetical conditions have covered the worst II condition. According to the results in 4 hypothetical conditions, the fuel centerline temperature relative difference between numerical solution and theoretical solution is less than 0.6%, and the iterations are less than 5. So the numerical method possesses excellent correctness, stability and convergence, and this method has much potential in application in fuel rod code.
Study of Minor Actinides Transmutation in PWR MOX Fuel
Authors: Chen Shengli , Yuan Cenxi , Wu Jingxia , Zou Yaolei 
Published: October 2017

The management of long-lived radionuclides in spent fuel is a key issue to achieve the closed nuclear fuel cycle and the sustainable development of nuclear energy. Partitioning-Transmutation is supposed to treat efficiently the long-lived radionuclides. Accordingly, the study of transmutation for long-lived Minor Actinides (MAs) is a significant work for the post-processing of spent fuel.
In the present work, the transmutations in Pressurized Water Reactor (PWR) Mixed OXide (MOX) fuel are investigated through the Monte Carlo based code RMC. Two kinds of MAs are incorporated homogeneously into two initial concentrations MOX fuel assembly.
The results indicate an overall nice efficiency of transmutation in both initial MOX concentrations, especially for two MAs primarily generated in the UOX fuel, 237Np and 241Am. In addition, the inclusion of 237Np has no large influence on other MAs, while the transmutation efficiency of 237Np is excellent. The transmutation of MAs in MOX fuel depletion is expected to be an efficient nuclear spent fuel management method.
Analysis of the Containment Thermal-Hydraulic Behavior of a Small Reactor During Loss of Coolant Accident
Authors: Lin Qian , Zhang Weizhong 
Published: October 2017

The containment thermal hydraulics of a small reactor during loss of coolant accident (LOCA) is studied by a lumped parameter one-dimensional model and a three-dimensional model. The capability of a kind of heat exchanger type passive containment cooling system (PCCS) is analyzed by the one-dimensional model. The calculation results show that, the decay heat can be removed and the containment pressure can be decreased by the proposed PCCS.
The steam and non-condensable gas (the air) distribution in the containment is investigated, the mixing and stratification behaviors are analyzed for several different cases, in which the PCCS and condenser are located at higher, base or lower position. The sensitivity analysis of the PCCS elevation shows that, in despite of the different gas stratification, the containment pressures are nearly the same. Similar conclusions can be obtained by the one-dimensional model and three-dimensional model. The preliminary results may indicate that, the designed PCCS and condenser can be located at a lower part, which will be benefit for the economy of the small reactor or meet other requirements.
Application of a 3D Discrete Ordinates-Monte Carlo Coupling Method on CAP1400 Cavity Streaming Calculation
Authors: Zheng Zheng , Li Hui , Wang Mengqi 
Published: October 2017

Neutrons and photons produced from reactor core during operation pass through the pressure vessel, reach the reactor cavity, and form the reactor cavity streaming. Reactor cavity streaming dose rates calculation during normal operation is important for the evaluation and control of the equipment dose rates in the nuclear power plant. Because reactor is great in dimension and complex in geometry, neutrons and photons fluence rates declined by several orders from reactor core to outside. Cavity streaming calculation is a deep penetration calculation with heavy computation load which is difficult to converge.
Three dimensional Discrete Ordinates and Monte Carlo (SN-MC) coupling method combines the advantage of the SN method with high efficiency and the MC method with fine geometrical modeling. The SN-MC coupling method decreases the tally errors and increases the efficiency of the MC method effectively by using MC surface source generated by the SN fluence rates. In this paper, the theoretical model of the 3D SN-MC coupling method is presented. In order to fulfill the coupling calculation, a 3D Discrete Ordinates code is modified to output angular fluence rates, a link code DO2MC is developed to calculate cummulative distribution functions of source particle variables on surface source, and a source subroutine is written for a 3D Monte Carlo code. The 3D SN-MC coupling method is applied on the calculation of the CAP1400 cavity streaming neutron and photon dose rates. Numerical results show that the 3D SN-MC coupling codes are correct, the relative errors of the results are less than 20% compared with those of the MC bootstrapping method, and the efficiency is greatly enhanced.
The Preliminary Shielding Analysis of CAP1400 Based on JMCT-S Code
Authors: Zhou Yan , Zheng Zheng , Mei Qiliang 
Published: October 2017

JMCT-S[1] 3-D high resolution Monte Carlo neutron and photon transport code is proposed for shielding design due to its significant advantages such as complex geometry modeling, accurate calculation method, high efficient parallel computing and visible mass data analysis. To demonstrate its feasibility and benefits in engineering, this paper presents the verification of JMCT-S code based on primary shielding analysis for CAP1400 nuclear power station.
Firstly, by employing JMCT-S code this paper completes the modeling and simulations of CAP1400 nuclear power station with full-core pin-by-pin and reactor pressure vessel internals inside of primary shield wall[2]. Secondly, neutron fluence rates are calculated based on the equilibrium cycle information. Particularly, the full-core neutron source is calculated from 3-D power distribution, fissionable nuclide fractions, fission spectrum, particle numbers and energy released per fission. Finally, JMCT-S code is verified by comparing results with those of the MCNP[3] code.
According to the calculation results, JMCT-S code has similar accuracy to the MCNP code. The relative error of neutron fluence rate in active core is within ±1.5%, and deviation of neutron fluence rate for fast neutron fluence rate on inner surface of reactor pressure vessel is less than ±10%. Consequently, JMCT-S code is reliable in shielding design of reactor plant which lays the foundation for usage of JMCT-S code.
Application of Monte-Carlo Simulation to Design of Sampler and Detector in Radiation Monitoring System
Authors: Sugihara Kei , Sakai Hirotaka , Hattori Kanako , Tanaka Genki , Hayashi Mitsunobu , Ito Toshiaki , Oda Naotaka 
Published: October 2017

In this study, the applicability of Monte-Carlo code PHITS(1) to the equipment design of sampler and detector in the radiation monitoring system was evaluated by comparing calculation results with experimental results obtained by actual measurements of radioactive materials.
In modeling a simulation configuration, reproducing the energy distribution of beta-ray emitted from specific nuclide by means of Fermi Function was performed as well as geometric arrangement of the detector in the sampler volume. The reproducing and geometric arrangement proved that the calculation results are in excellent matching with actual experimental results. Moreover, reproducing the Gaussian energy distribution to the radiation energy deposition was performed according to experimental results obtained by the multi-channel analyzer.
Through the modeling and the Monte-Carlo simulation, key parameters for equipment design were identified and evaluated.
Based on the results, it was confirmed that the Monte-Carlo simulation is capable of supporting the evaluation of the equipment design.
Radiation Shielding Analysis of a Spent Fuel Dry Storage Cask for FA300 Spent Fuel Assemblies
Authors: Wang Mengqi , Pan Nan , Li Hui , Jia Baojun 
Published: October 2017

Spent fuel dry storage technology is one of the most important intermediate storage technologies for spent fuel, because of its high security, good economic and easy to expand the scale. This article aims at designing a spent fuel dry storage cask which can contain 21 FA300 spent fuel assemblies.
The spent fuel dry storage cask is designed as concrete cask structure, which has the advantages of low manufacturing cost and simple manufacturing technology. Ventilation channels are designed for heating transfer, because the concrete is not a good thermal conductivity material. And labyrinth structure is designed for the ventilation channel to reduce the cavity streaming.
Radiation sources in spent fuel assemblies are mainly produced from fission products, actinides and their daughters located inside the effective fuel region, and other activation products in structure materials, which are calculated by ORIGEN.
The source and geometry of this problem are complex, and this is a real world deep penetration and streaming problem. Discrete ordinate method has great advantage in solving the deep penetration problem. Based on three-dimensional discrete ordinate code TORT, radiation shielding design method for spent fuel dry storage cask is studied, including main shield cask, cover lid, and ventilation channel. The results show that this spent fuel dry storage cask containing 21 FA300 spent fuel (cooling time: 10 years) assemblies can satisfy the requirement of dose rate limits in GB18871.
CFD Study on Specifics of Flow and Heat Transfer in Vertical Bare Tubes Cooled With Water at Supercritical Pressures
Authors: Zvorykin Alexander , Fialko Nataliia , Julii Sherenkovskyi , Aleshko Sergey , Meranova Natalia , Hanzha Mark , Bashkir Ivanna , Stryzheus Svitlana , Voitenko Artem , Pioro Igor 
Published: October 2017

The paper presents results of a study on flow and temperature fields in bare tubes cooled with SuperCritical Water (SCW). This study is based on a Computational Fluid Dynamics (CFD) simulation with the FLUENT code for upward flows in vertical tubes with heated length of 4 m and an inner diameter of 10 mm. Operating conditions were: Mass flux – G ≈ 500 and 1000 kg/m2s; heat flux – q = 189 – 826 kW/m2; and inlet coolant temperature – Tin = 320–360°C.
CFD predictions were compared with experimental data in this study. All three heat-transfer regimes: 1) normal heat transfer; 2) improved heat transfer; and 3) deteriorated heat transfer; were considered. The obtained results show that within normal and improved heat transfer CFD predicts experimental values reasonably well. However, within conditions of deteriorated heat transfer CFD predictions are less satisfactory.
The CFD outcomes of the heat flux effect on the flow and heat transfer of SCW are presented. Specifics of flow within the pseudocritical region (i.e., approximately ±25°C around a pseudocritical point) are discussed. The buoyancy effect is investigated by axial velocity profiles at the medium mass flux of 500 kg/m2s and heat flux of 287 kW/m2.
Experimental Research on the Influence of Inlet Conditions of Corrugated Plate Dryer
Authors: Feng Mao , RuiFeng Tian , BoWen Chen , Wei He 
Published: October 2017

Corrugated plate dryer is one of the key components of natural circulation steam generator in the pressurized water reactor nuclear power plant. It play a decisive role in the water separation system to making steam become a good quality after steam flow out of a vertical steam generator. With the increase of the reactor power in the future design, it requires that the performance of corrugated plate dryer must be increased in the characteristics of steam load, separating efficiency, volume and resistance to make sure that steam generators still produce qualified steam at higher steam loads and higher cycle magnitudes.
In this paper, we have done some experimental studies on the corrugated plate dryer. And we concentrated on airflow parameters on the separation performance and resistance performance of dryer.
(1) First of all, we have determined the experimental research program about this study, after this, we finished the design works of the test bench, and then we have completed the construction work based on the design works.
(2) Afterwards, the flow and separation mechanism of the wave plate dryer was experimentally studied by means of parameter measurement and analysis combined with observation and recording. The effects of different inlet humidities and inlet velocity on separation efficiency, pressure drop and re-entrainment were studied.
The above study shows that:
(1) The separation efficiency of corrugated plate dryer is related to inlet humidity and inlet airflow velocity. The separation efficiency of corrugated plate dryer is increase with inlet humidity goes higher, and increased to reach a highest point then goes down when inlet airflow velocity increased.
(2) The re-entrainment is related to inlet humidity. When inlet humidity increased, the inlet airflow critical velocity which lead to re-entrain would decrease.
(3) The pressure drop of dryer is related to the inlet airflow velocity. When the inlet wind velocity increases, the pressure drop increases.
The Application of Piping Material Code in the CAP1400 Nuclear Plant Project
Authors: Jie Shen , Zhen-ning Zhang , Yu Liu 
Published: October 2017

Nuclear industry differs from most other industries in the characters of large scaling, long period and multiple collaborating institutions [1]. Traditionally, the material codes are compiled respectively by collaborating institutions of nuclear power plant projects, so the whole code system lacks unified management and planning, causing many defects such as the incompleteness in information coverage, the inconformity in classification and description of materials and the confusion in commodity codes. So, it is of great significance in setting up a standard nuclear power material code system to effectively enhance the efficiency of design and management, and to ensure the schedules of the projects.
This article introduces in detail the entire process of the setting up and application of the material code system of State Nuclear Power Technology Company (SNPTC) by Shanghai Nuclear Engineering Research & Design Institute (SNERDI) for the CAP1400 nuclear power plant project. The application of the material code system in the CAP1400 project remarkably simplifies the work of design, material take-off and purchase, improving the project’s quality.
Preliminary Safety Analysis of a Compact Small Reactor: Feedwater Line Break and Small-Break LOCA Accident Analysis
Authors: Li Linsen , Shen Feng , Xing Mian , Liu Zhan , Qi Zhanfei 
Published: October 2017

A small Pressurized Water Reactor (PWR) with compact primary system and passive safety feature, which is called Compact Small Reactor (CSR), is under pre-conceptual design and development. For the purpose of preliminary assessment of the primary coolant system and capability evaluation of the passive safety system, a detailed thermal-hydraulic (T-H) system model of the CSR was developed.
Several design-basis accidents, including feedwater line break, double ended direct vessel injection line break (one of the small-break Loss Of Coolant Accidents, LOCA) and etc, are selected and simulated so as to evaluate and further optimize the design of passive safety systems, especially the passive core cooling system.
The results of preliminary accident analysis show that the passive safety systems are basically capable of mitigating the accidents and protecting the reactor core from severe damage. Further research will be focused on the optimization of pre-conceptual design of the thermal-hydraulic system and the passive core cooling system.