bluuebunny/crossref_metadata_2025 huggingface dataset:
Validation of CTF Void Predictions Using the BFBT Database
Authors: Porter Nathan W. , Avramova Maria N. , Mousseau Vincent A.
Published: October 2018

A quantified validation of CTF void fraction and equilibrium quality predictions is performed using the NUPEC BFBT database. Standard code modeling options are used, and there is no calibration to the experimental data. Each of the 392 steady state BFBT void distribution experiments has a specified geometry, power distribution, and spacer grid design. All experiments are modeled in CTF, but some cases that have oscillations are excluded from the final analyses. Three quantities of interest are compared between the experiment and the code predictions: exit equilibrium quality, average exit void fraction, and subchannel void fractions. The results indicate that, in general, CTF accurately predicts equilibrium quality but overpredicts void content. By splitting the subchannels into different groups, it is shown that subchannels near unheated surfaces are the least accurately modeled, especially for cases which have a high concentration of unheated surfaces. Though the BFBT database is the most complete source of experimental data for fine-mesh void distributions, the measurements are asymmetric even for symmetric geometry and boundary conditions. A method is proposed to approximate the experimental uncertainties which led to these asymmetries.
Welded Joint Evaluation for Chromium Controlled Carbon Steel Piping to Improve FAC Resistance
Authors: Uemoto Yoshio , Kawabe Takahiro , Shibata Hiroyuki , Tarasawa Shoh , Asano Hiroshi , Kaneda Junya
Published: October 2018

For condensate and feed water piping in nuclear power plants, it is desired to mitigate the pipe wall thinning risk due to Flow-Accelerated Corrosion (FAC). In aspect of material selection, low alloy steels are generally applied to improve FAC resistance. However, low alloy steels are inferior to carbon steels from the point of material cost and construction efficiency due to requirement of post weld heat treatment (PWHT). On the other hand, chromium is known as the most effective element to improve FAC resistance, and it is reported that a certain improvement of FAC resistance is also expected for carbon steels by increasing chromium content to over 0.10 wt%.

Such chromium controlled carbon steels are manufactured within the chemical composition range specified by material code of carbon steels, such as ASME B&PV Code Sec.II. Therefore, the amount of alloy content is lower than those for low alloy steels. The authors expect that PWHT can also be exempted for a certain thickness range of chromium controlled carbon steels, according to the exemption condition for normal carbon steels by ASME B&PV Code Sec.III. Furthermore, the chromium controlled carbon steels are generally cheaper than low alloy steels for base materials such as pipe and plate.

However, since chromium content of normal welding materials for carbon steels is generally lower than 0.05 wt%, chromium controlled carbon steel welding materials are specially-produced material. It makes the procurability worse compared to normal carbon steel welding materials. Additionally it should be confirmed if the increased chromium content affects the soundness of welded joint.

From the above reasons, it is necessary to decide the appropriate welding materials and methods for the chromium controlled carbon steel piping, considering the procurability of welding materials and the soundness of welded joint. In this study, the authors prepared the test pieces which simulate the assumed circumferential butt welded joints, then conducted the mechanical test such as tensile, impact, bend and hardness test to evaluate the soundness of welded joints. Furthermore, the authors evaluated the chromium content distribution of welded joints by using the Electro Probe Micro Analyzer (EPMA), in order to confirm if the chromium content is maintained over 0.10 wt% within the whole expected area.
Research of γ-Ray 3D Scanning Detection and Analysis System
Authors: Chen Chen , Hou Su-xia , Zhang Quan-hu , Su Xiang-hua , Zuo Wen-ming
Published: October 2018

With the development of nuclear industry, nuclear facilities have accumulated a large amount of pending nuclear waste during operation. Usually, they are stored in a closed container, and their physical and chemical forms are complicated and unevenly distributed. γ-ray 3D scanning detection and analysis system using non-destructive analysis of γ-ray analysis method to achieve a closed container of radionuclide recognition, positioning, quantitative measurement and analysis. The system has broad application prospects in the fields of nuclear safety, nuclear material control and accounting, radioactive recyclables and non-destructive measurement of nuclear waste. γ-ray 3D scanning detection and analysis technology is a comprehensive high-tech integrated nuclear science, nuclear radiation detection, radiation imaging, image processing, mechanical design, automatic control, computer technology and other disciplines.
Severe Accident Analysis With Spatial Discretized Model by MAAP: Part 2 — Parametric Study on Fukushima-Daiichi Unit-3
Authors: Nishi Yoshihisa , Kanda Kenichi , Abe Kazuma , Nishimura Satoshi , Nakamura Koichi , Furuya Masahiro , Ui Atsushi
Published: October 2018

Accident analyses of unit 3 in the Fukushima-Daiichi nuclear power station were performed with MAAP (Modular Accident Analysis Program) 5.03. In the analysis, the operations of RCIC, HPCI, SRV were simulated and assuming the operation of alternative water injection to reproduce the measured pressure and temperature. As a result of parametric evaluation, analysis results consistent with measured values are obtained. In this paper, the results of the sensitivity analysis are reported. Simultaneously, attempts were made to analyze the transient and deposition amount of fission product (FP) in the reactor building (R/B) and a model dividing each room of R/B in detail was created to confirm the FP behavior. Based on these analyses, the deposition amount of the primary containment vessel (PCV) (drywell (D/W) and wetwell (W/W)) and R/B could be estimated using detailed model of the R/B dividing into nodes in the MAAP simulation.
Development of Techniques for RCP Performance Verification Test and Optimization of Flow Stability
Authors: Kim Seok , Bae Byoung-Uhn , Cho Yun-Je , Jeon Woo-Jin , Kim Yeon-Sik , Cho Seok
Published: October 2018

Korea Atomic Energy Research Institute (KAERI) has designed and constructed a test facility for reactor coolant pumps (RCPs). In RCP test facility (RCPTF), it is possible to perform a type test for a newly-developed RCP as well as a production test for the Advanced Power Reactor 1400 (APR1400) RCP before its installation in nuclear power plants. The thermal hydraulic and electric capability of the RCPTF covers up to 18.5 MPa, 343 °C, 11.7 m3/s, and 14.0 MW for the design pressure, temperature, flow rate, and maximum electric power, respectively. In 2013, a commissioning test was performed to verify its designed capability, followed by several modifications in the RCPTF including signal processing and control logic to enhance the verification and evaluation capability of the RCP performance. During the commission test period, the technical activity for RCP performance verification test and optimization of flow stability have also been performed. The developed techniques can be divided into four subjects; 1) Development of flow stabilization technique for high flow condition test facility, 2) Improvement on the accuracy of performance verification factor and development of evaluation method, 3) Evaluation of test facility vibration characteristics and optimization of measurement system, and 4) RCP coast-down data production. KAERI has completed a full set of technique developments, a prerequisite for the RCP performance test.
LBB Analysis on Nuclear-Class Pipes of Floating Nuclear Power Stations
Authors: Bai Fan , Liu Yong , Lao Xingsheng , Xiao Qi , Zhao Zhenxing , Wu Jun , Ma Can , Wei Wang
Published: October 2018

Floating nuclear power stations are vessels with nuclear reactors, designed to power offshore oil and gas drilling, island development and remote areas. The safety of the facilities is an important issue. The Leak-Before-Break (LBB) assessment is essential to the design and evaluation of nuclear power plants against a sudden double-ended guillotine break of pipes. This paper describes the LBB assessment procedure applicable to the nuclear-class pipe of the floating nuclear power station. The loads considered in the analysis include variation of temperature/pressure, swing and underwater impact. Circumferential cracks are postulated at the dangerous positions of the pipe. The fatigue propagation of the surface crack is predicted based on the fracture mechanics and the finite element method to determine the time that pipe leaking happens. The critical length of the through-wall crack is calculated by the J integral–tearing modulus approach, and then compared with the minimum detectable crack length for the purpose that unstable fracture would not be happened before the leakage detected. According to the results of the analysis it could conclude that the pipe fulfils the LBB requirements.
Validation of a CFD Code for the Analysis of Hydrogen Behaviors and Thermal Hydraulics in Containments
Authors: Chen Meilan , Zheng Zeming
Published: October 2018

During the process of core melt-down accident in light water reactors, large quantities of hydrogen generated by drastic water-metal reaction are released to the containment. Subsequently, hydrogen-rich layer may be formed under the dome of the containment, threatening the integrity of nuclear Power Plants (NPPs). In the framework of a China national R&D project, China Nuclear Power Research Institute (CNPRI) has developed a three dimensional CFD Code for the assessment of hydrogen behaviors and relative thermal hydraulics in containment. The code solves the time-dependent Navier-Stokes Equations with multi-gas species. Validation with International Standard Problems (ISP) and other test data based on a Phenomena Identification and Ranking Table (PIRT) has been undergoing together with the development of this code. In this paper, the test cases of HYJET, COPAIN and TOSQAN 101 Test are validated. Stratification, buoyancy induced mixing in gases, convection heat transfer and condensation on surface are evaluated in the former two cases, while gas entrainment and mixing by spray droplets in the later one. Excellent agreements between experimental data and model predictions are obtained. In order to meet the requirements for application of the code in practical NPP design and safety analysis, further validations of other phenomena in PIRT should be performed in the near future.
Research on the Passive Residual Heat Removal System of Floating Nuclear Plants
Authors: Cong Wang , Wang Jue , Chen Hu , Yi Liao , Lei Chen
Published: October 2018

For electric energy generation, the design organizations of China have developed variants of floating nuclear plants. Depending on the customer requirement one or more power plants could operating in the same sea area at a time. To achieve a high degree of reactor safety, the passive core cooling subsystems is utilized widely on the floating nuclear plants. This paper conducted a preliminary design of the passive residual heat removal system (PRHRS), which is very important for decay hear removal under the ship blackout accident, used on floating nuclear plants. The limitations due to ocean environment including condition of sea wave and wind, space availability and weight limitations have be considered in the design stage. The computation model of PRHRS was established by use RELAP program. The transient behaviors of PRHRS are simulated. The results showed that PRHRS of floating nuclear plants could achieve the function of remove decay heat and keep reactor core safe under the ship blackout accident with the impact of ocean environment.
The Characteristics Study of Helium-Xenon Mixture in Closed Brayton Cycle for Space Nuclear Reactor Power
Authors: Yang Xie , Shi Lei
Published: October 2018

Differing from the adoption of helium as working fluid of closed Brayton cycle (CBC) for terrestrial high temperature gas cooled reactor (HTGR) power plants, helium-xenon mixture with a proper molar weight was recommended as working fluid for space nuclear reactor power with CBC conversion. It is essential to figure out how the component of helium-xenon mixture affects the net system efficiency, in order to provide reference for the selection of appropriate cycle working fluid. After a discussion of the physical properties of different helium-xenon mixtures, the related physical properties are studied to analyze their affection on the key parameters of CBC, including adiabatic coefficient, recuperator effectiveness and normalized pressure loss coefficient. Then the comprehensive thermodynamics of CBC net system efficiency is studied in detail considering different helium-xenon mixtures. The physical properties study reveals that at 0.7 MPa and 400 K, the adiabatic coefficient of helium-xenon mixture increases with increased molar weight, from 0.400 (pure helium) to 0.414 (pure xenon), while recuperator effectiveness firstly increases and then decreases with the increase of molar weight, and the normalized pressure loss coefficient increases monotonically with molar weight increases. The thermodynamic analysis results show that the adiabatic coefficient has less effect on the net system efficiency, while the net system efficiency increases with increased recuperator effectiveness, and the net system efficiency decreases with normalized pressure loss coefficient increases. Finally, the mixture of helium-8.6% xenon was adopted as working fluid, instead of pure helium, for ensuring less turbine mechanicals (turbine and compressor) stages, and resulting maximum recuperator effectiveness. At the given cold / hot side temperature of 400 / 1300 K, the net system efficiency can reach 29.18% theoretically.
Design and Safety Analysis of a Kind of UO2 Pellets Transport Container
Authors: Sun Shutang , Meng Dongyuan , Li Guoqiang , Sun Hongchao , Zhang Jiangang , Zhuang Dajie , Chen Lei , Lian Yiren
Published: October 2018

In this paper, a kind of transport container we designed was introduced. This container was designed to transport nuclear fuel pellets in different enrichment of U-235. The weight of this package is about 400kg, including the contents. One protect shell and two sealed border were designed in this container, which can ensure the contents were intact and the package has no criticality risk after transport accidents. During the design and safety analysis process, finite element analysis methods were used to improve the structure and analyze the safety performances of the container. In addition, we will test the safety performances of this container through a series of experiments in the future, including 9m drop (or crush), puncture, fire, water immersion and so on. Now, the calculate results show that this container was fit to the safety requirements in the transport accidents.
Experimental Research on the Fluid Induced Forces of Clearance Flow in Canned Motor Reactor Coolant Pump
Authors: Xu Rui , Hu Yaoyu , Long Yun , Yin Junlian , Wang Dezhong
Published: October 2018

Reactor coolant pump is one of the key equipment of the coolant loop in a pressurized water reactor system. Its safety relies on the characteristics of the rotordynamic system. For a canned motor reactor coolant pump, the liquid coolant fills up the clearance between the metal shields of the rotor and stator inside the canned motor, forming a clearance flow. The fluid induced forces of the clearance flow in canned motor reactor coolant pump and their effects on the rotordynamic characteristics of the pump are experimentally analyzed in this work. A vertical experiment rig has been established for the purpose of measuring the fluid induced forces of the clearance. Fluid induced forces of clearance flow with various whirl frequencies and various boundary conditions are obtained through the experiment. Results show that clearance flow brings large mass coefficient into the rotordynamic system and the direct stiffness coefficient is negative under the normal operating condition. The rotordynamic stability of canned motor reactor coolant pump does not deteriorate despite the existence of significant cross-coupled stiffness coefficient from the fluid induced forces of the clearance flow.
Analysis on the Dynamic Buckling Behavior of the Spacer Grid Structure
Authors: Guo Yan , Gu Chenglong , Tian Wei , Li Weicai
Published: October 2018

The spacer grid is a key element of the fuel assembly used in the Pressurized water reactor. Due to its structural complexity, the analysis and the design of the spacer grid structure is difficult. This paper discusses the 5 × 5 cell size partial grid analysis including the detailed grid structural elements, through which, the impact force, the rebound velocity and the time history of acceleration and as well as other mechanical properties of grid under different initial impact velocity were obtained. This paper carried out the dynamic buckling criterion studies, and determined the dynamic buckling load of the 5 × 5 cell size partial spacer grid. Based on assuming the impact process is simple harmonic vibration, the method to determine the dynamic stiffness of the spacer grid was proposed. The experiments were also performed for the comparison with the analytical results. It is found that the analytical results are in good agreement with the experimental results. As a result, we can conclude that the analysis model including detailed grid elements is able to yield accurate analytical results.
Reactor Baffle Cooling CFD Framework for Swelling Assessment
Authors: Filonova Yuliia , Filonov Vladislav , Dubyk Yaroslav
Published: October 2018

This Paper presents an improved estimation of reactor core baffle temperature distribution, during operation, at the nominal power level to address swelling problems of the reactor internals. Swelling is the main limiting factor in the reactor core internals long term operation of VVER-1000 nuclear units. The material irradiation-induced swelling and creep models are very sensitive to temperature distribution in metal, thus a more detailed analysis of the core baffle metal thermohydraulic cooling characteristics is required. A framework for CFD analysis of VVER-1000 reactor baffle cooling is presented. Firstly, an analytical model was developed to obtain boundary conditions and simplify CFD analysis, i.e. the real geometry of the cooling channels was replaced by equivalent elements, the core was presented as porous body with the appropriate characteristics. Secondly, the CFD analysis was performed using 60–degree symmetry, which included: core, baffle and core barrel, it is limited by the height of the baffle. Core is simplified as a homogeneous body with considering of spatial volumetric energy release. Core baffle is presented as monolithic body with considering of gamma-ray heat generation. Model includes a simplified geometry of connecting studs, considering cooling flow of the coolant through the nuts grooves. Calculated convection coefficient and temperature are in good agreement with analytical model, and give a more convenient result comparing to RELAP5/mod3.2. Obtained results were used to estimate baffle swelling process. Due to the less conservative results in temperature distribution swelling and creep deformations significantly decreased.
Virtual Fixture Augmentation of Operator Selection of Non-Contact Material Reduction Task Paths
Authors: Sharp Andrew , Petlowany Christina , Pryor Mitch
Published: October 2018

Decommissioning of nuclear systems and consolidation of resulting waste remains a major hurdle for the industry. The radioactive nature of nuclear waste makes manual handling unfeasible, while teleoperation and inflexible automation also have drawbacks.

This effort utilizes virtual fixtures to augment path creation for non-contact material reduction tasks. It builds on previous Nuclear and Applied Robotics Group work with Variable Normal Surface Virtual Fixtures, which generate surfaces offset from task surfaces. Offset surfaces can then provide poses at a set orientation to task surfaces. Once the Variable Normal Surface Virtual Fixture is generated from task information, operators build a list of desired path poses. The Robot Operating System Descartes package takes the pose list and plans a smooth trajectory for task execution. Planar and cylindrical demonstrations based on experimental studies at the United Kingdom’s Sellafield site were performed. This methodology augments waste reduction by allowing flexible laser cutting routes.
Coupled Calculation on Fluid Structure Interaction in Plate-Type Fuel Element
Authors: Yu Yiqi , Merzari Elia , Solberg Jerome
Published: October 2018

In nuclear reactors that use plate-type fuel, the fuel plates are thermally managed with coolant flowing through channels between the plates. Depending on the flow rates and sizes of the fluid channels, the hydraulic forces exerted on a plate can be quite large. Currently, there is a worldwide effort to convert research reactors that use highly enriched uranium (HEU) fuel, some of which are plate-type, to low-enriched uranium (LEU). Because of the proposed changes to the fuel structure and thickness, a need exists to characterize the potential for flow-induced deflection of the LEU fuel plates. In this study, as an initial step, calculations of Fluid-Structure Interaction (FSI) for a flat aluminum plate separating two parallel rectangular channels are performed using the commercial code STAR-CCM+ and the integrated multi-physics code SHARP, developed under the Nuclear Energy Advanced Modeling and Simulation program. SHARP contains the high-fidelity single physics packages Diablo and Nek5000, both highly scalable and extensively validated. In this work, verification studies are performed to assess the results from both STAR-CCM+ and SHARP. The predicted deflections of the plate agree well with each other as well as exhibiting good agreement with simulations performed by the University of Missouri utilizing STAR-CCM+ coupled with the commercial structural mechanics code ABAQUS. The study provides a solid basis for FSI modeling capability for plate-type fuel element with SHARP.
An Effect of Bismuth Ion on the Reduction of Terbium Ion in Molten LiCl-KCl Eutectic Salt
Authors: Kim Beom Kyu , Park Byung Gi , Han Hwa Jeong , Park Ji Hye , Kim Won Ki
Published: October 2018

A salt waste generated from the pyroprocess contains residual actinides and needs to be purified for recycling of the salt and waste conditioning. A co-reduction process could be considered for removal of residual actinides from the salt waste, which contains lanthanides and residual actinides. In the study, specifically, an effect of Bi(III) ion on the electrochemical reaction of Tb(III) ion was investigated in the molten LiCl-KCl eutectic with BiCl3 and TbCl3 at 773 K using electrochemical techniques of cyclic voltammetry, square wave voltammetry and open circuit chronopotentiometry. Tb(III) has a single redox couple without Bi(III). However, the cyclic voltammograms obtained at tungsten electrode in LiCl-KCl-BiCl3-TbCl3 showed four redox couples. The square wave voltammogram in same condition also showed five reduction peaks. Cyclic voltammogram and square wave voltammogram was resolved to find the accurate peaks for redox reaction. Each peak indicates the formation of Tb-Bi intermetallic compound except Tb(III) reduction peak. From the phase diagram of Tb-Bi, it is inferred that each peak corresponds to TbBi2, TbBi, Tb4Bi3, and Tb5Bi3. The open circuit chronopotentiometry was conducted to estimate Gibbs free energy of formation of Tb-Bi intermetallic compound.

The experimental results obtained from three kind of the electrochemical techniques showed that Tb-Bi intermetallic compounds were electrochemically formed under potential of Tb(III) reduction potential by co-reduction of Bi(III) and Tb(III). These results indicate that underpotential deposition by co-reduction could be used for Tb(III) removal from the salt waste with Bi(III).
Numerical Simulation of Debris Bed Relocation Behavior in Sodium-Cooled Fast Reactor
Authors: Teng Chunming , Zhang Bin , Shan Jianqiang
Published: October 2018

For the core disruptive accident (CDA) of sodium-cooled fast reactor (SFR), the molten fuel or steel is solidified into debris particles which form debris bed in the lower plenum. When the boiling occurs inside debris bed, the flow of coolant and vapor makes debris relocated and flattened, which called debris relocation. The thickness of debris bed has great influence to the cooling ability of fuel debris in low plenum. To ensure the effective implementation of the in-vessel retention (IVR), it’s very necessary to evaluate the transient changes of shape and thickness in relocation behavior for CDA simulation analysis. To simulate relocation behavior, a debris relocation model based on COMMEN code was developed in this paper. The debris relocation model was established based on the extrapolation of the shear strength mechanism, which was originally proposed and widely applied in soil mechanics filed. Shear strength is a function of the particles’ density and position. Debris bed is fluidized only when the shear stress in particle unit is larger than shear strength of debris particles. By integrating the debris relocation model into the COMMEN code, the transition process of the bed in depressurization experiments was simulated and compared against the experimental results. Good agreement shows that the debris relocation model presented in this paper can reasonably simulate the relocation behavior.
Experimental and Numerical Analysis of Steam Generator Tube Rupture Event for MYRRHA Reactor in CIRCE Facility With SIMMER-IV Code
Authors: Pesetti Alessio , Tarantino Mariano , Forgione Nicola
Published: October 2018

The Steam Generator Tube Rupture (SGTR) postulated event constitutes one of the most hazardous safety issues for Gen IV pool reactors, cooled by heavy liquid metals. This accidental scenario is characterized by quick water flashing when in contact with primary coolant liquid metal, causing pressure wave propagation, cover gas pressurization in the reactor main vessel as well as possible tube rupture propagation, vapour dragged through the core, oxides precipitation and consequent slugs and plugs formation.

The design phase of Gen IV MYRRHA reactor addressed the SGTR scenario issues in the framework of MAXSIMA project, supported by the European Commission. This research activity was fully executed at ENEA CR Brasimone, where a new test section was designed, assembled, instrumented and implemented in the large scale pool facility CIRCE. It was supported by the execution of preliminary and detailed pre-tests analysis performed adopting SIMME-III and -IV code, respectively.

This paper details the test section main features, able to host four full scale portions (each one constituted by 31 tubes) of the MYRRHA Primary Heat eXchanger (PHX), for carrying out four independent SGTR experiments. A couple of tests investigated the tube rupture at middle position between two spacer grids of the bundle. The other two tests analysed instead the rupture near the bottom tube plate. Auxiliary systems were adopted for reaching primary (Lead Bismuth Eutectic alloy, LBE) and secondary (water) coolant initial conditions in accordance with MYRRHA design. Water was injected at 16 bar and 200°C in LBE at 350°C under an argon cover gas at about atmospheric pressure. The experimental results of the first test (middle rupture), in terms of CIRCE vessel pressurization, vapour flow path through tube bundle and tubes deformation, are presented.

The post-test analysis was performed by SIMMER-IV code adopting the 3D Cartesian code version. The whole main vessel of CIRCE facility and implemented test section were modelled conserving heights and flowing areas. The experimental initial conditions were successfully matched by numerical results as well as the vessel pressurization and temperature time trends in the tube bundle following the SGTR. An important engineering feedback, for MYRRHA designer, was the evidence of rupture propagation absence. Moreover, the effectiveness of implemented safety devices, rupture disks, was evaluated and characterized for pressure relief feedbacks.

A wide series of high quality measured data (pressure, temperature, strain and mass flow rate) was acquired and constitutes a database enlargement for future codes validation and possible new model development.
Application of Vorticity Method in Auxiliary Impeller Optimization of HTR-PM Main Helium Fan
Authors: Fei Mo , Youjie Zhang
Published: October 2018

The helium circulator is an important part of HTR-PM. In order to avoid the heat which produced by the helium calculator of HTR-PM cumulating in the motor cavity, the heat should be taked out by the way of forced convection. According to the the spatial structure of channel, we need to design a single stage axial flow compressor here which has the technical characteristics with small power, small flow and high pressure. Let it meet the cooling demand of the specific heat of the motor, and also reduce the power consumption of cooling system itself, so as to improve the overall efficiency of the system.

To sum up, in the present study, through the design path of the positive problem, the conventional variable method is used to solve the flow field, that is, the flow field is solved by the velocity and pressure terms, and then the vortex in the flow field is deduced according to the velocity distribution According to the distribution of vorticity in the flow field, it provides the direction for the optimal design of the impeller. In the end, we hope to use this optimized design method to further optimize the blades and runner of the auxiliary impeller of helium circulator for HTR-PM.
Creep Deformation Analysis of a Pipe Specimen Based on Creep Damage Evaluation Method
Authors: Katsuyama Jinya , Yamaguchi Yoshihito , Li Yinsheng
Published: October 2018

After the Fukushima Daiichi nuclear power station (1F) accident, it has become more important to develop methods for evaluating the failure behavior of the nuclear components under severe conditions. We are, therefore, researching prediction methods for creep deformation and failure behavior of nuclear components under elevated temperature conditions based on finite element analysis. In this study, as part of a project called Components and Structures under Severe Accident Loading (COSSAL), we perform failure analysis of a large-scale pipe experiment to validate our prediction methods based on a creep damage evaluation method. The analytical results obtained from the conventional Norton and modified Garofalo constitutive laws, as well as the Kachanov-Rabotnov law considering material damage due to creep deformation, were compared with experimental results. We conclude that the Kachanov-Rabotnov law allows for the most accurate analysis among the three creep constitutive laws considered herein.
Stress and Fatigue Analysis of Thermal Shock for Cladding of Center Measurement Column Lower Head in the Fast Reactor Under Shutdown Condition
Authors: Zheng Shu , Lu Daogang , Cao Qiong , Liu Chao , Ding Yunlong , Huang Yi
Published: October 2018

The center measurement column lower head, which is under long-term influence of the liquid sodium, locates just above the core outlet in the fast reactor. Under the shutdown condition, the sudden drop of coolant temperature will cause the center measurement column lower head to be affected by the strong thermal shock that will threaten the integrity of the lower head. Cladding is always added to lower head to protect it from thermal shock damage. Therefore, the calculation and analysis of stress and fatigue damage for cladding is significant. Based on the built complete model and shutdown temperature curve of reactor core outlet, using ASME fatigue evaluation method, the stress and fatigue damage of thermal shock in cladding is calculated. And the influence of different layers of cladding on the von Mises equivalent stress distributions is analyzed in the paper. The results show that under a given shutdown curve, the maximum von Mises stress emerges at the joints of cladding and the sleeve. As the number of layers increases, the von Mises equivalent stress of each layer decreases. Due to the existence of thermal resistance, the von Mises equivalent stress has similar distributions between two designs of layer and different distributions between layers. The study can provide design reference on cladding for relevant engineering design.
The Canadian Nuclear Safety Commission: Readiness Activities to Regulate Small Modular Reactors
Authors: Lee Kevin
Published: October 2018

Over the course of the last several years the Canadian Nuclear Safety Commission (CNSC) has engaged with numerous vendors and potential licenses of small modular reactors (SMR) technology. This paper describes why Canada, and the CNSC, is of such interest to the international SMR community for prelicensing engagement and potential licensing of SMRs. It discusses what an SMR is and what potentially differentiates them from standard nuclear power plants (NPP). Readiness activities for the potential licensing of SMRs are described as well as modifications being made to the CNSC’s existing regulatory framework to facilitate the same, without reducing safety. The role of the CNSC’s discussion paper (DIS-16-04, Small Modular Reactors: Regulatory Strategy, Approaches and Challenges) and how feedback received on it helped confirm the CNSC’s modifications to be undertaken to the regulatory framework, as well as areas requiring further clarity, are highlighted. Finally, The CNSC Vendor Design Review (VDR) process is described as well as its part in ensuring a state of readiness to evaluate a licence application.
Transient Analysis of Fire Protection System at a Nuclear Power Plant Using Computer Code USLAM
Authors: Arastu Asif H. , Tom Eugene
Published: October 2018

A new Method Of Characteristics (MOC) based computer code, USLAM, has been used to analyze the Fire Protection System (FPS) of a Nuclear Power Plant. Some unique features of this system are that it has a complex geometry, has many operating modes, is connected to an adjacent nuclear power plant FPS, and has a significant number of high elevation standpipes or risers. In a typical FPS, firefighting water is pumped from a low elevation reservoir at atmospheric pressure to the whole system including higher elevations regions where water hammer due to water column separation & rejoining may occur. A low capacity keep-full system is employed to keep the system pressurized during standby conditions. A loss of system pressure due to the opening of the pre-action or deluge valves can lead to void formation at high elevations whose collapse can result in severe water hammer. A catastrophic valve failure due to a water hammer event at a nuclear power plant (Arastu, et al, 1999) disabled the entire FPS.

The analysis presented in this paper is to evaluate the possibility of such a water hammer event and mitigate it in an effective manner. The mitigation method employed is the use of multiple vacuum breakers. As many as 11 vacuum breakers are used for this purpose. The paper discusses the analysis results without and with vacuum breakers for all operating modes. The basic methodology of the USLAM code is also discussed.
Facile synthesis of ultrasmall polydopamine-polyethylene glycol nanoparticles for cellular delivery
Authors: Harvey Sean , Ng David Yuen Wah , Szelwicka Jolanta , Hueske Lisa , Veith Lothar , Raabe Marco , Lieberwirth Ingo , Fytas George , Wunderlich Katrin , Weil Tanja
Published: October 2018

Very small polydopamine (PDA) polyethylene glycol (PEG) crosslinked copolymer (PDA-PEG) nanoparticles have been prepared following a convenient one-step procedure in aqueous solution. Particle sizes and colloidal stabilities have been optimized by varying PEG in view of chain length and end group functionalities. In particular, amine-terminated PEG3000 [PEG3000(NH2)2] reacted with polydopamine intermediates so that very small, crosslinked PDA-PEG nanoparticles with sizes of less than 50 nm were formed. These nanoparticles remained stable in buffer solution and no sedimentation occurred. Chemical functionalization was straight-forward as demonstrated by the attachment of fluorescent dyes. The PDA-PEG nanoparticles revealed efficient cellular uptake via endocytosis and high cytocompatibility, thus rendering them attractive candidates for cell imaging or for drug delivery applications.
Experimental study of field emission from ultrasharp silicon, diamond, GaN, and tungsten tips in close proximity to the counter electrode
Authors: Lenk Claudia , Lenk Steve , Holz Mathias , Guliyev Elshad , Hofmann Martin , Ivanov Tzvetan , Rangelow Ivo W. , Behzadirad Mahmoud , Rishinaramangalam Ashwin K. , Feezell Daniel , Busani Tito
Published: October 2018

The patterning process in field-emission scanning probe lithography (FE-SPL), a high-resolution and cost-effective method for nanofabrication, is based on the field emission of electrons from ultrasharp tips in close proximity to a sample (distances below 100 nm). Thereby, the emitted electrons expose directly an ultrathin resist film. The field enhancement at the tip apex is crucial for the field emission current, which follows the Fowler–Nordheim theory. Despite the success of FE-SPL in nanofabrication, systematic experimental studies of the field-emission process, including the determination of the tip radius and tip-to-sample distance during the measurement, for these small tip-to-sample distances and different tip materials are lacking. To resolve this issue, experimental measurements of the field-emission current for tip–sample proximity distances below 100 nm were performed. For this purpose, the developed AFM in SEM system was modified,1,2 which enables one to monitor the tip–sample distance with a high accuracy using SEM while simultaneously recording the field-emission current. The authors present experimental results of the dependence of the field-emission current on the tip shape, tip material, applied voltage, and tip–sample distance. Therefore, the emission characteristics of silicon, diamond, GaN, and tungsten tips are shown. The knowledge about the field-emission process for small tip-to-sample distances will help to understand and improve the current FE-SPL, regarding also the choice of tip material. Furthermore, these measurements enable the detailed comparison with current FE models beyond state-of-the-art since all necessary parameters (voltage, current, tip diameter, and tip-to-sample distance) could be measured and controlled during the FE experiment due to the unique experimental system.
Atomic force microscope integrated with a scanning electron microscope for correlative nanofabrication and microscopy
Authors: Rangelow Ivo W. , Kaestner Marcus , Ivanov Tzvetan , Ahmad Ahmad , Lenk Steve , Lenk Claudia , Guliyev Elshad , Reum Alexander , Hofmann Martin , Reuter Christoph , Holz Mathias
Published: October 2018

More than 40 years after its invention, the atomic force microscopy (AFM) can be integrated with scanning electron microscope (SEM) instruments as an increasingly capable and productive characterization tool with sub-nanometer spatial resolution. The authors have designed and developed an AFM instrument capable to be integrated into any SEM or in a combination of SEM with a focused ion-beam (FIB) tool. The combination of two or more different types of techniques like SEM, energy dispersive x-ray spectroscopy, and AFM is called correlative microscopy because analytical information from the same place of the sample can be obtained and correlated. For the first time, they introduced to the SEM/FIB tool correlative nanofabrication methods like field-emission scanning probe lithography, tip-based electron beam induced deposition, and nanomachining. The combination of all these methods provides a completely new nanotechnology instrument, which should be seen as a tool for correlative nanofabrication and microscopy. Thus, it provides for the first time the capabilities of a stand-alone instrument with the capabilities of nondestructive three-dimensional tip-based metrology and nanofabrication into the combined SEM/FIB tool. In this article, the authors describe all these methods in detail and present a brief example of the results obtained. They demonstrate that the self-sensing, self-actuating cantilevers (called active cantilevers) equipped with Diamond tip are a versatile toolkit for fast imaging and emerging nanofabrication. The AFM integrated into SEM is using active cantilevers that can characterize and generate nanostructures all in situ without the need to break-vacuum or contaminate the sample.
Pathogen blockade of TAK1 triggers caspase-8–dependent cleavage of gasdermin D and cell death
Authors: Orning Pontus , Weng Dan , Starheim Kristian , Ratner Dmitry , Best Zachary , Lee Bettina , Brooks Alexandria , Xia Shiyu , Wu Hao , Kelliher Michelle A. , Berger Scott B. , Gough Peter J. , Bertin John , Proulx Megan M. , Goguen Jon D. , Kayagaki Nobuhiko , Fitzgerald Katherine A. , Lien Egil
Published: October 2018

Caspase-8 is a player in pyroptosis

The activation of certain pattern-recognition receptors by pathogen-associated molecular patterns results in the formation of inflammasome complexes. Inflammasome complexes can initiate both the maturation of inflammatory cytokines and pyroptotic cell death via the caspase-mediated cleavage of gasdermin D (GSDMD). As of now, the only known regulators of GSDMD in macrophages are caspase-1 and caspase-11. Orning
et al.
report an additional pathway controlling GSDMD processing. YopJ, an effector molecule produced by
Yersinia
(the causative agent of plague), inhibits TAK1–IκB kinase signaling. This, in turn, results in caspase-8–directed cleavage of GSDMD, pyroptosis, and the release of interleukin 1β (IL-1β) and IL-18. Thus, in the arms race between host and pathogen, the host recognizes signaling disturbances as pathogenic and counters with inflammation and cell death.

Science
, this issue p.
1064
Structural basis of latent TGF-β1 presentation and activation by GARP on human regulatory T cells
Authors: Liénart Stéphanie , Merceron Romain , Vanderaa Christophe , Lambert Fanny , Colau Didier , Stockis Julie , van der Woning Bas , De Haard Hans , Saunders Michael , Coulie Pierre G. , Savvides Savvas N. , Lucas Sophie
Published: October 2018

Visualizing TGF-β1 regulation by GARP

Regulatory T cells (T
regs
) can suppress immune responses through a variety of mechanisms. One such mechanism involves the activation of a surface-bound latent form of the cytokine transforming growth factor–β1 (TGF-β1). Within the cell, newly synthesized pro-TGF-β1 homodimers form disulfide bonds with the transmembrane protein GARP, which acts to chaperone and orient the cytokine for activation at the cell surface. Liénart
et al.
reveal how GARP interacts with TGF-β1, using a crystal structure in which the complex was stabilized using a Fab fragment from a monoclonal antibody (MHG-8) that binds to the complex. In so doing, they also demonstrate how MHG-8 prevents membrane-associated TGF-β1 release. These structural and mechanistic insights may inform treatments of diseases with altered TGF-β1 functionality and dysfunctional T
reg
activity, including cancer immunotherapy.

Science
, this issue p.
952
Immunotopological Analysis of theTreponema denticolaMajor Surface Protein (Msp)
Authors: Godovikova Valentina , Goetting-Minesky M. Paula , Timm John C. , Fenno J. Christopher
Published: October 2018

Treponema denticolais among a small subset of the oral microbiota contributing to severe periodontal disease. Due to its relative genetic tractability,T. denticolais a model organism for studyingTreponemaphysiology and host-microbe interactions.T. denticolaMsp is a highly expressed outer membrane-associated oligomeric protein that binds fibronectin, has cytotoxic pore-forming activity, and disrupts intracellular regulatory pathways. It shares homology with the orthologous group ofT. pallidumTpr proteins, one of which is implicated inT. pallidum in vivoantigenic variation. The outer membrane topologies of both Msp and the Tpr family proteins are unresolved, with conflicting reports on protein domain localization and function. In this study, we combined empirical immunological data derived both from diverseT. denticolastrains and from recombinant Msp expression inE. coliwithin silicopredictive structural modeling ofT. denticolaMsp membrane topology, to move toward resolution of this important issue inTreponemabiology.
Minute Virus of Canines NP1 Protein Interacts with the Cellular Factor CPSF6 To Regulate Viral Alternative RNA Processing
Authors: Dong Yanming , Fasina Olufemi O. , Pintel David J.
Published: October 2018

The
Parvovirinae
are small nonenveloped icosahedral viruses that are important pathogens in many animal species, including humans. Unlike other parvoviruses, the bocavirus genus controls expression of its capsid proteins via alternative RNA processing, by both suppressing polyadenylation at an internal site, termed the proximal polyadenylation (pA)p site, and by facilitating splicing of an upstream adjacent intron. This regulation is mediated by a small genus-specific protein, NP1. Understanding the
cis
-acting targets of NP1, as well as the cellular factors with which it interacts, is necessary to more clearly understand this unique mode of parvovirus gene expression.
Murine Leukemia Virus Glycosylated Gag Reduces Murine SERINC5 Protein Expression at Steady-State Levels via the Endosome/Lysosome Pathway to Counteract SERINC5 Antiretroviral Activity
Authors: Li Sunan , Ahmad Iqbal , Shi Jing , Wang Bin , Yu Changqing , Zhang Lixin , Zheng Yong-Hui
Published: October 2018

MLV glycoGag not only enhances MLV replication but also increases HIV-1 infectivity similarly as Nef. Recent studies have discovered that both glycoGag and Nef antagonize a novel host restriction factor Ser5 and promote viral replication. Compared to Nef, the glycoGag antagonism of Ser5 is still poorly understood. MLV glycoGag is a transmembrane version of the structural Gag protein with an extra 88-amino-acid leader region that determines its activity. We now show that glycoGag interacts with Ser5 in live cells and internalizes Ser5 via receptor-mediated endocytosis. Ser5 is polyubiquitinated and relocalized to endosomes and lysosomes for massive destruction. In addition to the previously identified tyrosine-based sorting signal, we find two more important residues for Ser5 relocalization and downregulation. We also find that the Ser5 sensitivity to glycoGag is conserved in the SERINC family. Together, our findings highlight the important role of endosome/lysosome pathway in the enhancement of viral replication by viral proteins.