Accepted Manuscripts

Ali Salah Omar Aweimer, Hakim A. Bouzid and Mehdi Kazeminia
ASME J of Nuclear Rad Sci   doi: 10.1115/1.4040493
Leaking valves has forced shutdown in many nuclear power plants. The myth of zero leakage or adequate sealing must give way to more realistic maximum leak rate criterion in design of nuclear flange joints and valve stem packing. It is well established that the predicting leakage in these pressure vessel components boxes is a major engineering challenge to designers and end users. This is particularly true in nuclear valve due to the different working conditions and material variations. The prediction of the leak rate through packing rings is not a straightforward task to predict. This work presents a study on the ability of micro channel flow models to predict leak rates through packing rings made of soft materials such as flexible graphite. A methodology based on the experimental characterization of the porosity parameters is developed to predict leak rates at different compression stress levels. Three different models are compared to predicate the leakage, where the diffusive and second order flow models are derived from Naiver-Stokes equations and incorporate the boundary conditions of an intermediate flow regime to cover the wide range of leak rate levels. The lattice model is based on porous media of packing rings as packing bed. The flow porosity parameters of the micro channels assumed to simulate the leak paths present in the packing are obtained experimentally. The predicted leak rates from different gasses are compared to those measured experimentally, in which the set of packing rings is mainly subjected to different gland stresses and pressures.
TOPICS: Packing (Shipments), Valves, Packings (Cushioning), Leakage, Flow (Dynamics), Porosity, Microchannels, Experimental characterization, Flanges, Design, Porous materials, Pressure vessels, Sealing (Process), Stress, Boundary-value problems, Compressive stress, Graphite, Microchannel flow, Nuclear power stations
Jianfeng Mao, Shiyi Bao, Zhiming Lu, Lijia Luo and Zengliang Gao
ASME J of Nuclear Rad Sci   doi: 10.1115/1.4040494
The RPV integrity is significantly influenced by the crust. A strong, coherent crust anchored to the RPV walls could allow the yet-molten corium to fall away from the crust as it erodes the RPV, therefore thermally decoupling the melt pool from the coolant and sharply reducing the cooling rate. Due to the thermal resistance of the crust layer, it somewhat prevents further attack of melt pool from the RPV. In the present study, the effect of crust on RPV structural behaviors was examined under multi-layered crust formation conditions with consideration of detailed thermal characteristics, such as high temperature gradient across the wall thickness. Thereafter, systematic finite element analyses (FEA) and subsequent damage evaluation with varying parameters were performed on a representative reactor pressure vessel (RPV) to figure out the possibility of high temperature induced failures with the effect of crust layer.
TOPICS: Accidents, Failure, High temperature, Finite element analysis, Cooling, Coolants, Reactor vessels, Damage, Thermal resistance, Wall thickness
Qiang Zhao, Yang Li, Zheng Zhang and Xiaoping Ouyang
ASME J of Nuclear Rad Sci   doi: 10.1115/1.4040495
The sputtering of graphite due to the bombardment of hy- drogen isotopes is one of the critical issues in successfully using graphite in the fusion environment. In this work, we use molecular dynamics method to simulate the sputtering by using the LAMMPS. Calculation results show that the peak values of the sputtering yield are located between 25 eV to 50 eV. After the energy of 25 eV, the higher incident energy cause the lower carbon sputtering yield. The temperature which is most likely to sputter is about 800 K for hydrogen, deuterium and tritium. Before the 800 K, the sputtering rates increase when the temperature increase. After the 800 K, they decrease with the temperature increase. Under the same temperature and energy, the sputtering rate of tritium is big- ger than that of deuterium, the sputtering rate of deuterium is bigger than that of hydrogen. When the incident energy is 25 eV, the sputtering yield at 300 K increases before the incident angle at 30? and remains steady after that.
TOPICS: Isotopes, Sputtering (Irradiation), Graphite, Hydrogen, Molecular dynamics simulation, Temperature, Molecular dynamics methods, Carbon
Valentyn Tsisar, Carsten Schroer, Olaf Wedemeyer, Aleksandr Skrypnik and Jürgen Konys
ASME J of Nuclear Rad Sci   doi: 10.1115/1.4040422
Effect of structural state (solution annealed and after 40% cold work) and surface finishing (turning, grinding, polishing) on the corrosion behavior of austenitic 1.4970 (15 15 Ti) steel in flowing (2 m/s) Pb-Bi eutectic containing 1E-7 mass% dissolved oxygen at 400°C and 1E-6 mass% O at 500°C is investigated. At 400°C for ~13000 h, the corrosion losses are minor for steel in both structural states and for surfaces finished by turning and grinding - a thin Cr-based oxide film is formed. In contrast, the polished surface showed initiation of solution-based corrosion attack with the formation of iron crystallites and preferential propagation along the grain boundaries. The depth of corrosion attack does not exceed 10 µm after ~13000 h. At 500°C for 2000 h, the samples in both structural states showed general slight oxidation. Cold-worked steel underwent a severe groove-type and pit-type solution-based attack of 170 µm in maximum depth while the solution annealed sample showed only sporadic pit-type corrosion attack to the depth of 45 µm in maximum.
TOPICS: Steel, Corrosion, Oxygen, Surface finishing, Grinding, Polishing, Grain boundaries, Iron, oxidation
Shengli Chen and Cenxi Yuan
ASME J of Nuclear Rad Sci   doi: 10.1115/1.4040423
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. The Partitioning-Transmutation method is supposed to efficiently treat the long-lived radionuclides. Accordingly, the transmutation of long-lived Minor Actinides (MAs) is significant 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 neutron transport method. Two types of MAs are homogeneously incorporated into MOX fuel assembly with different mixing ratios. In addition, two types of design of semi-homogeneous loading of $^{237}$Np in MOX fuels are studied. The results indicate an overall nice efficiency of transmutation in PWR with MOX fuel, especially for $^{237}$Np and $^{241}$Am, which are primarily generated in the current UOX fuel. In addition, the transmutation efficiency of $^{237}$Np is excellent, while its inclusion has no much influence on other MAs. The flattening of power and burnup are achieved by semi-homogeneous loading of MAs. The uncertainties of Monte Carlo method are negligible, while those due to nuclear data change little the conclusions of the transmutation of MAs. The transmutation of MAs in MOX fuel is expected to be an efficient method for spent fuel management.
TOPICS: Manufacturing, Pressurized water reactors, Fuels, Spent nuclear fuels, Radioisotopes, Neutrons, Sustainable development, Transmutation (Nuclear physics), Nuclear fuel cycle, Design, Monte Carlo methods, Nuclear power, Uncertainty
Hidekazu Takazawa, Kazuma Hirosaka, Katsumasa Miyazaki, Norihide Tohyama and Naomi Matsumoto
ASME J of Nuclear Rad Sci   doi: 10.1115/1.4040432
A new Japanese nuclear regulation requirement prepares estimating the possible damage to plant structures due to intentional aircraft impact. Aircraft impact needs to be considered in existing nuclear power plants. The structural damage and fuel dispersion behavior after aircraft impact into plant structures can be estimated using finite element analysis (FEA). FEA needs validated experimental data to determine the reliability of results. In this study, an analysis method was validated using a simple model such as a cylindrical tank. Numerical simulations were conducted to estimate the impact and dispersion behavior for a water-filled cylindrical tank. The simulated results were compared with the test results of the VTT Technical Research Centre of Finland. Simulations were carried out using a multipurpose FEA code LS-DYNA®. The cylindrical tank was modeled using a shell element, and filled water was modeled using a smoothed particle hydrodynamics (SPH) element. First, two analysis models were used to estimate the effect of the number of SPH elements. One was generated with about 300,000 SPH elements. The other was generated with 37,000 SPH elements. The cylindrical tank ruptured in the longitudinal direction after impact into a rigid wall, and the filled water dispersed. Few differences emerged in the simulated results using different numbers of SPH elements. The impact test of the VTT was simulated with an arbitrary Lagrangian-Eulerian (ALE) element to consider the air-drag. The analytical dispersion pattern and history of dispersion velocity ratio agreed well with the impact test results.
TOPICS: Water, Finite element analysis, Aircraft, Impact testing, Damage, Nuclear power stations, Shells, Hydrodynamics, Particulate matter, Fuels, Computer simulation, Drag (Fluid dynamics), Reliability, Simulation, Engineering simulation
Feng Wang, Ziqiang Yang, Long Wang and Qiang Wen
ASME J of Nuclear Rad Sci   doi: 10.1115/1.4040377
Performance of methane steam reforming reactor heated by helium for hydrogen production has been studied by numerical method. Results show with the increasing of reactant gas inlet velocity, temperature in the reactor drops, leading to the decreasing of methane conversion and hydrogen production rate. Methane conversion, hydrogen production and hydrogen production rate rise with the increasing of reactant gas inlet temperature, while the increasing degree of system thermal efficiency reduces. Besides, with helium inlet velocity rising, temperature in the reactor increases and reaction in the reactor becomes more sufficient. Therefore, methane conversion and hydrogen production also increase when helium inlet temperature of rises, but its influence is weaker compared to that of helium inlet velocity. In the process of methane steam reforming heated by high temperature gas cooled reactor (HTGR) for hydrogen production, lower reactant gas inlet velocity, suitable inlet temperature, higher inlet velocity and higher HTGR outlet temperature of helium are preferable
TOPICS: Catalysts, Helium, Hydrogen production, Methane, Steam reforming, High temperature, Computational fluid dynamics, Temperature, Very high temperature reactors, Thermal efficiency, Gas cooled reactors, Numerical analysis
Mohammad Islam, Douglas Janssen, Carlos A. Romero-Talamás, Dan Kostov, Wanpeng Wang, Zhongchi Liu, Narsingh Singh and Fow-Sen Choa
ASME J of Nuclear Rad Sci   doi: 10.1115/1.4040364
Plants exhibit complex responses to change in environmental conditions such as radiant heat flux, water quality, airborne pollutants, soil contents. We seek to utilize natural chemical and electrophysiological response of plants to develop novel plant-based sensor networks. Our present work focuses on plant responses to nuclear radiation - with the goal of monitoring plant responses as benchmarks for detection and dosimetry. In our study, we used plants includeing Cactus, Arabidopsis, Dwarf mango (pine), Euymus, Azela, and Arborvitae. We demonstrated that these plants' Chlorophyll-a (F680) to Chlorophyll-b (F735) ratio can be changed according to the radiation dose amount. The recovery processes and speed are different for different plants.
TOPICS: Nuclear radiation, Pine (Wood product), Dosimetry, Radiant heat, Sensor networks, Soil, Pollution, Water pollution, Radiation (Physics)
Yayu M Hew, Kevin J Schillo, Akansha Kumar, Kurt Harris and Steven D Howe
ASME J of Nuclear Rad Sci   doi: 10.1115/1.4040370
This paper presents a power management and distribution system for a growing Martian colony. The colony is designed for a 15-year operation lifetime, and will accommodate a population that grows from 6 to 126 crewmembers. To provide sufficient power, a nuclear fission surface power system is proposed with a total capacity of 1 MWe. The system consists of three 333 kWe fission reactors. DC transmission with 2000 VDC is found to provide the best power density and transmission efficiency for the given configuration. The grounding system consists of grounding rods, grounding grids, and a soil-enhancement plan. A regenerative fuel cell using a propellant tank recycled from the lander was found to have the best energy density and scalability among all the options investigated. The thermal energy reservoir, while having the worst storage efficiency, can be constructed through in-situ resource utilization, and is a promising long-term option. A daily load following a 12-hr cycle can be achieved, and the power variation will be less than 10% during normal operation. Several main load-following scenarios are studied and accommodated, including an extended dust storm, nighttime, daytime, and transient peak power operation. A contingency power operation budget is also considered in the event that all of the reactors fail. The system has a power distribution efficiency of 85%, a storage efficiency of 50%, and a total mass of 13 Mt.
TOPICS: Nuclear fission, Storage, Stress, Transients (Dynamics), Fuel cells, Cycles, Nuclear reactors, Propellants, Rods, Soil, Storms, Power density, Density, Power systems (Machinery), Dust, Thermal energy, Reservoirs
Hui Li and Guangxin Zhang
ASME J of Nuclear Rad Sci   doi: 10.1115/1.4040369
The control function for process, HVAC and electrical systems in nuclear power plant (NPP) are represented by control logic diagram. To develop Distributed Control System (DCS), the designer and supplier should complete the activities of control logic configuration, testing and verification which are based on control logic diagram. Design Verification is an effective method to ensure the correctness of control logic design. This paper represents a system which is capable of implementing control logic design verification automatically for NPP I&C system, as well as an overview of the procedure and some examples by using this system. With the design data (including control requirements and control logic diagrams in computer-readable format) and simulation technology, this system automatically performs design verification based on different rules and confirms the design outputs meet the inputs - the control requirements of plant's systems. Finally, a conclusion about the design verification system and future scenarios is given.
TOPICS: Design, Nuclear power stations, Control systems, Simulation, HVAC equipment, Testing, Computers, Electronic systems
Arnold Gad-Briggs, Pericles Pilidis and Theoklis Nikolaidis
ASME J of Nuclear Rad Sci   doi: 10.1115/1.4040371
The Intercooled Cycle (IC) is a simplified novel proposal for Generation IV Nuclear Power Plants (NPP) based on studies demonstrating efficiencies of over 45%. As an alternative to the Simple Cycle Recuperated (SCR) and the Intercooled Cycle Recuperated (ICR), the main difference in configuration is no recuperator, which reduces its size. It is expected that the components of the IC will not operate at optimum part power due to seasonal changes in ambient temperature and grid prioritisation for renewable sources. Thus the ability to demonstrate viable part load performance becomes an important requirement. The main objective of this study is to derive Off-Design Points (ODPs) for a temperature range of -35 to 50°C and COTs between 750 to 1000°C. The ODPs have been calculated using a tool designed for this study. Based on results, the intercooler changes the mass flow rate and compressor pressure ratio. However, a drop of ~9% in plant efficiency, in comparison to the ICR (6%) was observed for pressure losses of up to 5% . The reactor pressure losses for IC has the lowest effect on plant cycle efficiency in comparison to the SCR and ICR. Characteristic maps are created to support first order calculations. It is also proposed to consider the intercooler pressure loss as a handle for ODP performance. The analyses brings attention to the IC an alternative cycle and aids development of cycles for Generation IV Nuclear Power Plants specifically Gas Cooled Fast Reactors (GFRs) and Very High Temperature Reactors (VHTRs)
TOPICS: High pressure (Physics), Design, Gas turbines, Cycles, Helium, Nuclear power stations, Pressure, Temperature, Very high temperature reactors, Fast neutron reactors, Compressors, Stress, Flow (Dynamics)
Deeksha Gupta, Edita Bajramovic, Holger Hoppe and Antonio Ciriello
ASME J of Nuclear Rad Sci   doi: 10.1115/1.4040372
Companies involved in the nuclear domain, like component and platform manufacturers, have well established yearly trainings on Nuclear Safety Culture. These trainings are typically covered as part of the annual quality assurance-related refresher trainings, introductory courses for new employees. Gradually, security awareness trainings are also addressed on a regular basis, typically with a focus on IT, test bay or construction site work environment, and privacy-related topics. Due to emerging national nuclear regulation, steadily but surely, specialized cybersecurity trainings are foreseen for integrator and utilities. Beyond these safety, physical security and cybersecurity specific trainings, there is a need to address the joint part of these disciplines, starting from the planning phase of a new NPP. The engineers working on safety, physical protection and cybersecurity, must be aware of these interrelations to jointly elaborate a robust I&C architecture and a resilient security architecture. This paper provides more in-depth justification of when and where additional training is needed, due to the ubiquitous deployment of digital technology in new NPPs. Additionally, for existing NPPs, the benefits of conveying knowledge by training on specific interfaces between the involved disciplines, will be discussed. Furthermore, the paper will address the need of focused training of management stakeholders, as eventually, they must agree on the residual risk. The decision-makers are in charge of facilitating the inter-disciplinary cooperation in parallel to the allocation of resources, e.g. on security certifications of products, extended modeling-based safety and security analyses and security testing coverage.
TOPICS: Safety, Computer security, Security, Nuclear power stations, Engineering disciplines, Modeling, Testing, Engineering education, Engineers, Public utilities, Construction, Resource allocation, Security analysis, Risk
Bingbing Liang, Xu Zhang, Haifeng Yin and Yang Dai
ASME J of Nuclear Rad Sci   doi: 10.1115/1.4040368
Accumulative test data indicates that the effects of the light water reactor (LWR) environment could cause the fatigue resistance of primary pressure boundary components materials to be significantly reduced. EAF is the abbreviation of the environmentally assisted fatigue effects. In 2007, NRC issued RG. 1.207. It was updated in 2014. And it requires that the effects of LWR environment on the fatigue life reduction of metal components should be considered for new design plants. And it suggests to use environmental correction factor (Fen) to account for EAF. Fen=Nair/Nwater (N is occurrences). NUREG/CR-6909 [1] presents the detail Fen calculation formula. Fen is a function of temperature, strain rate, dissolved oxygen level in water, and sulfur content of the steel. Accordingly, Fen calculation will present a comparatively conservative result. Depends on the experience of the primary pressure boundary piping transient operation, Fen varies during each transient. More uncertainty and confusion are raised during the application of the Fen method. The research work in this article includes: first, the typical character of piping thermal transient is derived based on the existing experience. Second, small specimen EAF tests are conducted depend on the above derived combined loading characters. Then effort is taken to improve the application of the Fen method for the combined multi-transient loading conditions. And the results are compared with that of the lowest instantaneous Fen method and equalization of the weighted Fen method. Finally, a designed test plan is presented.
TOPICS: Electric arc furnaces, Stress, Transients (Dynamics), Pipes, Light water reactors, Pressure, Fatigue, Temperature, Steel, Design, Metalwork, Uncertainty, Fatigue life, Oxygen, Sulfur, Water
Ming Wang, Modi Lin and Jinkai Wang
ASME J of Nuclear Rad Sci   doi: 10.1115/1.4040367
Spent fuel pool stores fuel assemblies removed from the reactor over the years. In the event of an earthquake, spent fuel pool and its accident mitigation measures may fail at the same time, which can cause serious accident consequences. Based on the design and operation characteristics of typical CPR1000 nuclear power plant, this paper uses probability safety assessment method to quantitatively evaluate the risk of spent fuel pool caused by seismic events, and identify the weaknesses in the design and operation of the nuclear power plant. Quantitative analysis results show that spent fuel pool damage frequency caused by seismic events is lower than core damage frequency caused by seismic events. The main risks come from the collapse of fuel building and structural damage of the spent fuel pool. In addition, seismic events affect both reactor core and spent fuel pool, Water vapor and radioactive materials from spent fuel pool accident have an adverse effect on mitigation of core accident.
TOPICS: Risk analysis, Earthquakes, Spent nuclear fuels, Accidents, Damage, Design, Fuels, Nuclear power stations, Probability, Collapse, Safety, Radioactive substances, Water vapor, Risk
Technical Brief  
Tulis Jojok Suryono and Akio Gofuku
ASME J of Nuclear Rad Sci   doi: 10.1115/1.4040366
In an emergency condition of nuclear power plant, operators have to mitigate the accident in order to remove the decay heat and to prevent the release of radioactive material to the environment following the emergency operating procedures. The action of operators on a component, for example changing the parameter level of a component, which is described in a procedure step, will impact other components of the plant and the plant behavior. Nowadays, the advanced main control rooms have been equipped with the computer-based procedures (CBPs) which provide some features and benefits which are not available in paper-based procedures (PBPs). However, most of CBPs do not provide information of the impact of the counteractions on each procedure step (components influenced and future plant behavior) although it is useful for operators to understand the purpose of the procedure steps before making decisions and taking the actions. This paper discusses the functional information and the method to generate the information using multilevel flow modeling (MFM) model of operator actions on some procedure steps of a simplified emergency operating procedure (EOP) of PWR plant, as an example. Keywords: emergency operating procedures, functional information, components influenced, future plant behavior, steam generator tube rupture, PWR plant
TOPICS: Emergencies, Pressurized water reactors, Flow (Dynamics), Heat, Radioactive substances, Control rooms, Accidents, Boilers, Modeling, Computers, Decision making, Nuclear power stations, Rupture
Yewang Bie, Donghui Zhang, Wenbin Xiong, Li Huwei, Mingyu Wu and Xinzhe Wang
ASME J of Nuclear Rad Sci   doi: 10.1115/1.4040029
As the first fast reactor of China, the safety of CEFR is extremely important, and will decide the future of Chinese fast reactor project. The failed fuel detection system of CEFR provides surveillance and protection for the first barrier-fuel cladding of CEFR, so it is one of the most important systems for the safety of CEFR. As tag gas method is an important method for fuel-failure location in fast reactor, CEFR has a medium-term and long-term plan of using this method to locating failed fuel assemblies. This article introduces the main principle of tag gas method, summarizes the application of this method, and compares the advantages and disadvantages of each fuel failure location method. Combining the design characteristics of China Experimental Fast Reactor, this work analyzes the selection principle of tag gas isotopes and the effects on heat transfer capability of fuel element while tag gas filled in. Meanwhile, according to the detection ability of mass spectrometer and the foreign advanced utilization experiences of tag gas method, some suggestions are provided.
TOPICS: Fuels, Manufacturing, China, Fast neutron reactors, Failure, Safety, Heat transfer, Isotopes, Mass spectrometers, Cladding systems (Building), Design, Surveillance
Wei Gao, Guofeng Tang, Jingyu Zhang and Qinfang Zhang
ASME J of Nuclear Rad Sci   doi: 10.1115/1.4039967
Shanghai Nuclear Engineering Research and Design Institute (SNERDI) has been studying seismic risk analysis for nuclear power plant for a long time, and completed Seismic Margin Analysis (SMA) for several plants. After Fukushima accident, seismic risk has drawn an increasing attention worldwide, and the regulatory body in China has also required the utilities to conduct a detailed analysis for seismic risk. So we turned our focus on a more intensive study of Seismic Probabilistic Safety Assessment (PSA/PRA) for nuclear power plant in recent years. Since quantification of seismic risk is a key part in Seismic PSA, lots of efforts have been devoted to its research by SNERDI. The quantification tool is the main product of this research, and will be discussed in detail in this paper. First, a brief introduction to Seismic PSA quantification methodology is presented in this paper, including fragility analysis on system or plant level, convolution of seismic hazard curves and fragility curves, and uncertainty analysis as well. To derive more accurate quantification results, the binary decision diagram (BDD) algorithm was introduced into the quantification process, which effectively reduces the deficiency of the conventional method on coping with large probability events and negated logic. Finally, this paper introduced the development of the seismic PSA quantification tool based on the algorithms discussed in this paper. Tests and application has been made for this software based on a specific nuclear power plant seismic PSA model.
TOPICS: Industrial research, Probabilistic safety assessment, Earthquake risk, Nuclear power stations, Algorithms, Design, China, Computer software, Probability, Nuclear engineering, Public utilities, Fukushima nuclear disaster, Japan, 2011, Uncertainty analysis
Muhammad/Q Awan, Liangzhi Cao, Hongchun Wu and Chuanqi Zhao
ASME J of Nuclear Rad Sci   doi: 10.1115/1.4039971
Use of FCM fuel in LWRs is an attractive option for existing and future generations of these reactors to make them accident tolerant in nature. Present work focuses on the neutronic study of the use of burnable material in various configurations to control the excess reactivity and keeping the MTC feedback negative for entire cycle length. Erbia and gadolinia, two conventional materials are used in three different configuration including QUADRISO, BISO and Matrix Mix forms. Results obtained from the implicit random treatment of the double heterogeneity of TRISO, QUADRISO and BISO particles show that the erbia is the best material to be used in QUADRISO and Matrix Mix configurations with lowest reactivity swing for the life cycle and residual poison well below 0.5 %. Gadolinia is usable in FCM environment only in the BISO form where enhanced self-shielding controls the depletion performance of the material. The gadolinia has almost zero residual poison at EOC; however, it has relatively large reactivity swing, which will need more micromanagement of the control rods during the plant operations. At BOC, erbia loaded assemblies have shown an increase in negative value of MTC compared with reference due to presence of resonance peak in erbium near 1 eV. The finally recommended material-configuration combinations have shown the excess reactivity containment in desired manner with good depletion performance and negative feedback of the MTC for life cycle.
TOPICS: Fuels, Manufacturing, Pressurized water reactors, Cycles, Feedback, Rods, Light water reactors, Accidents, Resonance, Particulate matter, Containment
Kei Sugihara, Hirotaka Sakai, Kanako Hattori, Genki Tanaka, Mitsunobu Hayashi, Toshiaki Ito and Naotaka Oda
ASME J of Nuclear Rad Sci   doi: 10.1115/1.4039968
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.
TOPICS: Sensors, Radiation measurement, Simulation, Design, Modeling, Nuclides, Radiation (Physics), Radioactive substances, Beta particles
Yankun Dou, xinfu He, Dongjie Wang, Wu Shi, Lixia Jia and Wen Yang
ASME J of Nuclear Rad Sci   doi: 10.1115/1.4039969
In order to study the contribution of Mn atoms in Cu precipitates to hardening in bcc Fe matrix, the interactions of a (111){110} edge dislocations with nanosized Cu and Cu-Mn precipitates in bcc Fe have been investigated by using of molecular dynamics. The results indicate that the critical resolved shear stresses of the Cu-Mn precipitates are larger than that of Cu precipitates. Meanwhile, the critical resolved shear stresses of the Cu-Mn precipitates show a much more significant dependence on temperature and size, compared to Cu precipitates. Mn atoms exhibit strong attractive interaction with <111> crowdion and improve the fraction of transformed atoms from body centered cubic (bcc) phase to face centered cubic (fcc) phase for big size precipitates. Those all lead to the higher resistance to the dislocation glide. The increasing temperature can assist the Cu atoms rearrange back towards a bcc structure, resulting in the rapid decline of the critical resolved shear stresses. Eventually, these features are confirmed that the appearance of Mn atoms in Cu precipitates greatly facilitates the hardening in bcc Fe matrix.
TOPICS: Hardening, Atoms, Shear stress, Dislocations, Temperature, Molecular dynamics

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