bluuebunny/crossref_metadata_2025 huggingface dataset:
Characteristic Analysis of Supercritical Water-Cooled Reactor During Uncontrolled CR Withdrawal at Normal Operation Transient
Authors: Cheng Wanxu , Zhou Tao , Liu Liang , Song Mingqiang , Xia Bangyang
Published: February 2014

Uncontrolled CR withdrawal at normal operation is one of transients of the supercritical-pressure light water cooled reactor. Based on the physical and thermal analysis models, the related parameters are analyzed at the condition of Uncontrolled CR withdrawal at normal operation. The results show that both the reactivity of Doppler feedback decreases and the reactivity of density feedback decreases at first, the net reactivity changes slightly, which results in a modest increase in core power, the main coolant flow rate increases with the reactor power due to operation of the main steam temperature control system, the maximum cladding temperature increases, but it is below the criterion.
A Study and Development of Primary Circuit Simulation Code for China Experimental Fast Reactor
Authors: Wang Jin , Zhang Donghui , Hu Wenjun , Ren Lixia
Published: February 2014

A fast reactor is one of recommended candidates of Generation IV nuclear energy systems, which would meet wide requirements such as sustainability, safety and economics for nuclear energy development. To be the China’s first fast reactor, China Experimental Fast Reactor (CEFR) typical technical options are following: 65 MW thermal power and 20 MW electric power, three circuits of sodium-sodium-water, integrated pool type structure for the primary circuit.

To establish modular simulation system for sodium fast reactor, the code which simulated the thermal-hydraulic behavior of primary circuit was developed. The physical models include reactor core, reactor vessel cooling channel, pumps, protection vessel, intermediate heat exchangers, ionization chamber cooling channel, cold sodium pool, hot sodium pool, inlet plenum, and pipes, etc. The code could compute coolant pressures, flow rates, and temperatures in the primary circuit. This module was designed for analysis of a wide range of transients. Although based on CEFR, it can treat an arbitrary arrangement of components.
Study of Mixing Characteristics Over a Spacer Grid With Sloping Channels Based on Numerical Simulations in a PWR 5×5 Rod Bundle Using CFD Codes
Authors: Chen Xi , Zhang Hong
Published: February 2014

Spacer grids are important components of fuel assemblies for Pressurized Water Reactors (PWR). The presence of spacer grid promotes local heat transfer adjacent to the rod wall downstream by inducing swirl and cross flows within and between sub-channels to increase thermal hydraulic safety margin. Recent years, Computational Fluid Dynamics (CFD) methodologies are widely adopted to designs of spacer grids. This paper presents results of numerical simulations with commercial code CFX 12.0 in a PWR 5 × 5 rod bundle including a spacer grid with sloping channels. Based on a combined mesh generation approach of structured and unstructured mesh, distributions of velocity fields, temperature and pressure fields downstream the spacer grid were analyzed. The results indicate that cross flows caused by the spacer grid are uniform in circumference inducing no thermal hydraulic deterioration, but mass exchange between central hot fluid and external cold fluid appears insufficient for the new style grid.
Experimental Study on Oxidation of Graphite for HTR
Authors: Liu Yang , Sun Ximing , Dong Yujie
Published: February 2014

In the high temperature gas-cooled reactor, the oxidation of graphite is inevitable as a result of impurities in helium coolant. As an air or water ingress accident would cause graphite components to oxidize more seriously, thereby it would affect the reactor normal operation and safety. Oxidation velocity and oxidation product of a selected graphite (excess material from 10MW High Temperature Gas-cooled Reactor) were studied, samples are oxidized between 400°C to 1200 °C with gas flow rates ranging from 125 to 500 ml/min. Relationship between oxidation conditions and surface properties of oxidized graphite is also elaborated by means of gas chromatography and scanning electron microscopy for scanning of graphite surface.
Flow-Induced Grid-to-Rod Fretting Test and Simulation for NFI PWR Fuel Assembly
Authors: Okui Shota , Kubo Yuichiro , Kakinoki Shumpei , Lu Roger Y. , Karoutas Zeses , Ikeno Tsutomu
Published: February 2014

A long-term flow-induced vibration and wear test was performed for a full-scale 17×17 PWR fuel mockup, and the test results were compared with numerical simulations. The flow-induced vibration on a fuel assembly or fuel rods may cause Grid-to-Rod Fretting (GTRF) and result in the leakage of fuel rods in PWRs. GTRF involves non-linear vibration of a fuel rod due to the excitation force induced by coolant flow around a fuel rod. So, the numerical simulation is performed by VITRAN (Vibration Transient Analysis Non-linear) and Computational Fluid Dynamics (CFD). VITRAN code was developed by Westinghouse to simulate fuel rod flow induced vibration and GTRF. In this paper, it was confirmed that the code can reproduce GTRF wear for NFI fuel assembly. CFD calculation is performed to obtain the axial and lateral flow velocity around the fuel rods, reflecting detailed geometries of fuel assembly components like bottom nozzle, spacer grids. The numerical simulation reasonably reproduced the vibration and wear test for NFI fuel assembly.
Out-of-Pile Heat Transfer Test Device for Low Pressure Irradiation Capsule
Authors: Ma Yongzhe , Zhang Peisheng
Published: February 2014

In-pile irradiation capsules are used for fuel and material irradiation at research reactor. Design of irradiation capsule should be based on thermal analysis to determine the key structural parameters. A set of out-of-pile heat transfer test device is innovated for the design validation. Heat transfer performance of in-pile irradiation capsule can be simulated, which includes heat conduction, convection and radiation. Electric heating rod is used as the heat source for the device, which simulates linear power or heat flux of in-pile irradiation capsule. Temperature different of internal components is simulated by adjusting helium gap. Thermocouples are fixed to measure wall temperature and internal temperature of the components in the device. Some out-of-pile heat transfer tests have been carried out for irradiation capsule development for China Advanced Research Reactor (CARR) and China Experimental Fast Reactor (CEFR) at China Institute of Atomic Energy (CIAE). The test verifies surface and internal temperature of the capsules and conclusions of thermal analysis. A finished in-pile test verifies the capsule design at rated power, which also proves that results of the out-of-pile heat transfer test are conservative. The article describes the function, structure, test process and other aspects of the technical requirements of the heat transfer test device.
Structural Integrity of a Reactor Vessel Lower Head Under In-Vessel Steam Explosion Loads
Authors: Xue Guohong , He Yinbiao , Cao Ming , Yu Hao , Gao Yongjian
Published: February 2014

Passive nuclear power plants emphasize the “In vessel retention” idea such that, after a postulated severe accident event, the reactor vessel wall, flooded with emergency cooling water, will maintain its structural integrity and consequently keep the molten core inside the reactor vessel. However, steam explosion may still occur when the melting core or molten metal is mixed with cooling water. The huge pressure pulses from the steam explosion may be a threat to the structural integrity of the reactor vessel lower head and the potential failure may make the situation difficult to control. This paper presents a detailed analysis on the structural integrity of a reactor vessel lower head. First, a mathematical model is built to relate the equivalent plastic strain in the lower head under explosive loads based on the law of conservation of energy. Then a finite element model, using the computer code ABAQUS, is built and the material’s yield strength as a function of strain rate was simulated using the Bodner-Symonds methodology. With this model, the dynamic response and the structural integrity of the reactor vessel lower head is studied, considering the effect of the magnitude, the shape and the duration of the pressure pulses. The method used in this paper is believed to be applicable to other types of devices containing potential explosive materials and thus could provide guiding significance to similar problems.
Conceptual Studies of Design Scheme for Steam Generator in FBR
Authors: Zhang Zhenxing , Yu Huajin , Wu Zhiguang , Zhu Lina
Published: February 2014

Steam generator (SG) is one of the most important equipment in the heat transport system of China experimental fast reactor (CEFR). As per design requirements for improving economics and enhancing safety, it is necessary to make optimization analysis on various key aspects of SG. Many basic parameters were discussed in this paper, such as the design input of SG, tube-length, tube specification and the number of tubes, etc. The optimization analysis provides valuable suggestions about the number of tubes, tube-length, the heat transfer area of the margin and tube specification for the SG in its design life.
Ongoing Research of Water Chemistry Effect on SCC Properties of Candidate Materials for SCWR
Authors: Gong Bin , Huang Yan-ping , Jang E , Liu Jin-Hua , Xia Xiao-jiao , Zhao Yong-Fu
Published: February 2014

One key issue during development of Super Critical Water Cooled Reactor (SCWR) is selection of water chemistry operating and control methods for core coolant. Under strong corrosive effect of supercritical water (SCW), the stress corrosion cracking (SCC) sensitivity of materials increases. Most of materials reliably used in PWR or BWR may not be competent for SCWR. Modified austenitic steel and nickel base alloy are hopeful due to low general corrosion rate but failures happen caused by SCC when meet incompatible chemistry environment. The possible options of water chemistry specifications for SCWR should be qualified carefully on SCC effect. This paper introduces research status of SCW chemistry by Nuclear Power Institute of China (NPIC). The SCC behaviors in oxidation and reduction SCW are under study and the testing data will be collected to screen out water chemistry operating strategy for million kilowatts SCWR (CSR1000) under development in China.
Fretting Wear Behavior of NC30Fe in Distilled Water and Hydrazine Solution
Authors: Guo Hong , Ren Pengdi , Zhang Xiaoyu , Chen Guangxiong , Ye Yi
Published: February 2014

In this study, fretting wear behavior of NC30Fe tubes against 1Cr13 stainless steel cylinder in distilled water and hydrazine solution was investigated at room temperature. Wear scar was analyzed systematically by NanoMap500DLS dual-mode profilometer, SEM, EDX. The results showed that the fretting logs were running in slip regime. The hydrazine solution may lead to corrosion. The friction coefficient in aqueous medium was lowered obviously, but it in distilled water was lower than that in hydrazine solution due to the effect of hydrazine solution. The wear was lightened in aqueous medium significantly, but it was more severe in hydrazine solution than that in distilled water for the reason of hydrazine solution. The abrasive wear, friction oxidation and delamination were the main wear mechanisms of NC30Fe in ambient. However, in aqueous medium the abrasive wear and delamination were the main wear mechanisms of NC30Fe.
Tightness Analysis of Pressured Equipment in Experimental Test on Steam Condensation on the Cold Surface
Authors: Yang Lin , Han Liyong
Published: February 2014

To maintain the containment within the design limits of pressure and temperature, the advanced pressurized water reactor (APWR) designed by Westinghouse uses a passive safety system to transfer the heat from inner containment to outside. The passive containment cooling system (PCCS) includes many natural phenomena mechanisms. Steam condensation is one of the most important phenomena. Most heat is removed by steam condensation on inside surface of the containment during the postulated design basic accidents (DBA). It is very significant for engineering designing and code developing to study the mechanism of steam condensation on cold surface. There was an experiment made by University of Wisconsin on it. In this paper, the structure pressure of the pressured equipment is calculated and the tightness is also analyzed.
Neutron Absorbers in CANDU Natural Uranium Fuel Bundles to Improve Operating Margins
Authors: Chan Paul K. , Paquette Stéphane , Bonin Hugues W. , French Corey , Pant Aniket
Published: February 2014

Safety margins are particularly tight in natural uranium-fuelled CANDU reactors which are refueled on-power. During on-power refueling, the insertion of xenon-free fresh fuel bundles into the reactor core affects the reactor’s excess reactivity in such a way that this could lead to temporary power derating. It is desirable from a fuel management perspective, and to maintain safety margins to eliminate this xenon-free effect and any other power ripples such as the subsequent plutonium reactivity peak. A redesign of the CANDU NU fuel bundle with an appropriate combination of elements, with some including neutron-absorbers, could well address the issue of the xenon-free initial portion of the bundle’s irradiation and also lower the plutonium-peak that occurs shortly thereafter. This may improve the fuel utilization (by further optimizing the fuelling strategy) and provide improved safety margins (by lowering the maximum channel and bundle powers).

The use of neutron-absorbers in fuel design and manufacturing has been a regular practice in Light Water Reactor fuels for more than three decades. In CANDU applications, neutron absorbers have also been considered for the conceptual Advanced CANDU Reactor and the Low Void Reactivity fuel designs, for which the fissile content is made of low enriched uranium (LEU) or MOX fuels. The application to CANDU natural uranium (NU) fuel, however, especially as burnable poisons, is a relative novel approach. The reason for this is that the neutron economy in natural uranium-fuelled CANDU reactors is a prime concern, thus the addition of extra neutron absorbers is generally shunned. In our proposed application of burnable poisons to existing CANDU NU fuel design, because of low excess reactivity for NU fuel, the amount of neutron-absorber is expected to be restricted to small quantities and in a manner whereby the poison effect is restricted to the initial period of excess reactivity of a newly inserted fuel bundle. This implies that the impact on neutron economy would be relatively minimal, but the fuel performance would be significantly improved.

Small amounts and appropriate mixtures of neutron absorbers were selected (approximately 500 mg of absorbers in a CANDU fuel bundle having a nominal weight of 24 kg). Preliminary results indicate that the fuelling transient and the subsequent reactivity peak can be lowered to improve the reactor’s operating margins. A parametric study using the Los Alamos National Laboratories’ MCNP 5 and Atomic Energy of Canada Limited’s WIMS-AECL 3.1 codes is presented in this paper. Details of this project and future work are also to be discussed.
Method to Select the Countries and Scenarios More Appropriate for the Deployment of SMR
Authors: Locatelli Giorgio , Mancini Mauro , Belloni Pietro
Published: February 2014

The evolution of nuclear technologies has led to the development of a large number of reactor designs. In particular there is significant and growing interest towards Small Modular Reactors (SMRs), i.e. nuclear reactor of a size between 40MWe and 700 MWe. Several heterogeneous countries are potentially interested in their deployment and approximately twenty advanced SMR designs are under development all over the world. However, both the market dimensions for SMRs and where they may be deployed remain unclear. This paper presents a method for the identification of countries which could be interested in the construction of SMRs and which factors foster their deployment. The presented method, grounded on a comprehensive literature review, is composed of three screenings and comprises the strategic consequences of the Fukushima accident. The first screening selects all the countries of the world that the International Atomic Energy Agency (IAEA) and World Nuclear Association (WNA) indicate as possible market for nuclear reactor in general. The second screening selects countries based on their potential interest in SMR development in the short to medium term. The third screening identifies countries where SMRs are a particularly suitable choice. To perform this last “scenario based” selection the strengths and weakness of the SMRs have been analyzed to define their strategic match with the potential countries. The findings of this investigation reveal the strategic factors promoting the deployment of SMR and provide the basis for a ranking of countries in which these factors could create the market for SMR.
An Experimental Study on Liquid Film Dynamics and Interfacial Wave of Air-Water Two-Phase Flow in a Horizontal Channel
Authors: Zhang Youjia , Ma Weimin , Gong Shengjie
Published: February 2014

This study is concerned with liquid film dynamics and stability of annular flow, which plays an important role in understanding film rupture and dryout in boiling heat transfer. The research work starts from designing and making a test facility which enables the visualization and measurement of liquid film dynamics. A confocal optical sensor is applied to track the evolution of film thickness. A horizontal rectangular channel made of glass is used as the test section. Deionized water and air are supplied into that channel in such a way that an initial stratified flow forms, with the liquid film on the bottom wall. The present study is focused on characterization of liquid film profile and dynamics in term of interfacial wave and shear force induced film rupture under adiabatic condition. Based on the experimental data and analysis, it is found that given a constant water flowrate, the average thickness of water film decreases with increasing air flowrate, while the interfacial wave of the two-phase flow is intensified. As the air flowrate reaches a critical value, a localized rupture of the water film occurs.
Structure Design of Jordan Subcritical Reactor
Authors: Yao Chengzhi , Fan Yuerong , Zhang Jinshan , Han Haifen , Yi Dayong , Zhang Zhanli
Published: February 2014

The subcritical reactor is a kind of small nuclear facility, which may realize the chain reactor reaction through introducing an external neutron source to meet with the production, research, teaching, testing and training purposes. Meanwhile, the subcritical reactor will not reach the critical status, which has advantages of simple structure, safety, reliability. Jordan subcritical reactor is a light water moderated reactor, it employs the Pu-Be as external neutron source, and the neutron multiplication factor in the range of 0.94～0.95. Jordan subcritical reactor mainly consists of fuel elements, reactor core and reactor core vessel assembly, operation platform, water loop system, neutron source driving system, which can be used for the purposes of teaching, training, testing and research. This paper reviews the international application history of subcritical reactor and its status of research and development, describes the design purpose and requirement of Jordan subcritical reactor. The detailed structure design of Jordan subcritical reactor is illustrated. Furthermore, the structure design characteristics and some difficulties of Jordan subcritical reactor are mentioned in detail, such as the choose of structure materials, lay out of Jordan subcritical reactor, control of machining precision and sealing of pipes. Then, the possible solutions of these problems are presented. Now, the manufacture and installation of Jordan subcritical reactor has been completed, which fulfill the anticipated design requirement.
Design of a Low Vulnerability Electronic Fiber Optic Seal System
Authors: Di Yuming , Zhang Quanhu , Zuo Guangxia , Han Feng , Chen Ronghua , Huo Yonggang , Xu Peng
Published: February 2014

Electronic seal is an important technology in nuclear disarmament verification, which is also widely used in nuclear safeguards, nuclear nonproliferation and other fields. This paper proposes a low vulnerability electronic fiber optic seal scheme according to the requirements of performance and low vulnerability for the seal. The seal system mainly consists of two major components, including the main control circuit and the secondary control circuit. The main control circuit includes the main microcontroller unit, the optical measuring unit, the temperature measuring unit, the switch measuring unit, the acceleration measuring unit and a data storage unit, mainly to complete multiple tamper indicating. In the main control circuit, the main microcontroller unit is the core of the system, mainly to implement the timing trigger, the timing adjustment, system fault diagnosis, fiber-optic monitoring, temperature monitoring, switch module monitoring, acceleration monitoring, and so on. The secondary control circuit includes a USB microcontroller unit and the encryption chip, mainly to complete data encryption and authentication. In the secondary control circuit, the USB microcontroller is mainly to implement the communication with external verification PC, legitimate users identification, encrypted-data transmission, starting/stopping the seal, clock synchronization calibration, and so on. In this way, the designed electronic fiber optic seal has been provided with multiple anti-tampering capabilities and lower vulnerability.
Research on Digital Instrumentation and Control System in Nuclear Power Plant Using Mixed Redundant Model
Authors: Ke Zhiwu , Li Xianling , Yu Rui , Xiang Rui
Published: February 2014

Reliability is the most important factor in instrumentation and control (I&C) system for nuclear power plant (NPP). Now in order to reach high reliability, static redundant and dynamic redundant technique is in common use. Static redundant method can keep out the failure of one module, but could not detect the mistake. Dynamic redundant mean would examine the mistake of the working unit, but can not ward off the failure. In this study, our main purpose is to develop the advanced redundant way which includes the advantage of static redundancy and dynamic redundancy means by mixed redundant model (MRM). Utilizing logic order in the relationship between unit and system using reliable diagram, analyzing the reliability of MRM by the way of probability, the result has obvious effect upon improving the system reliability compared with above two models. MRM can be applied for safety critical I&C.
Study on Primary and Secondary Heat-Transport Systems for Sodium-Cooled Fast Reactor
Authors: Dragunov Alexey , Saltanov Eugene , Pioro Igor , Harvel Glenn , Ikeda Brian
Published: February 2014

One of the current engineering challenges is to design next generation (Generation IV) Nuclear Power Plants (NPPs) with significantly higher thermal efficiencies (43–55%) compared to those of current NPPs to match or at least to be close to the thermal efficiencies reached at fossil-fired power plants (55–62%). The Sodium-cooled Fast Reactor (SFR) is one of the six concepts considered under the Generation IV International Forum (GIF) initiative.

The BN-600 reactor is a sodium-cooled fast-breeder reactor built at the Beloyarsk NPP in Russia. This concept is the only one from the Generation IV nuclear-power reactors, which is actually in operation (since 1980’s). At the secondary side, it uses a subcritical-pressure Rankine-steam cycle with heat regeneration. The reactor generates electrical power in the amount of 600 MWel. The reactor core dimensions are 0.75 m (height) by 2.06 m (diameter). The UO2 fuel enriched to 17–26% is utilized in the core.

There are 2 loops (circuits) for sodium flow. For safety reasons, sodium is used both in the primary and the intermediate circuits. Therefore, a sodium-to-sodium heat exchanger is used to transfer heat from the primary loop to the intermediate one. In this work major parameters of the reactor are listed. The actual scheme of the power-conversion heat-transport system is presented; and the results of the calculation of thermal efficiency of this scheme are analyzed. Details of the heat-transport system, including parameters of the sodium-to-sodium heat exchanger and main coolant pump, are presented.

In this paper two possibilities for the SFR in terms of the power-conversion cycle are investigated: 1. a subcritical-pressure Rankine-steam cycle through a heat exchanger (current approach in Russian and Japanese power reactors); 2. a supercritical-pressure CO2 Brayton gas-turbine cycle through a heat exchanger (US approach).

With the advent of modern super-alloys, the Rankine-steam cycle has progressed into the supercritical region of the coolant and is generating thermal efficiencies into the mid 50% range. Therefore, the thermal efficiency of a supercritical Rankine-steam cycle is also briefly discussed in this paper.

According to GIF, the Brayton gas-turbine cycle is under consideration for future nuclear power reactors. The supercritical-CO2 cycle is a new approach in the Brayton gas-turbine cycle. Therefore, dependence of the thermal efficiency of this SC CO2 cycle on inlet parameters of the gas turbine is also investigated.
Real-Time Simulation Platform for the Design and Verification of the Operation Strategy of the HTR-PM
Authors: Dong Zhe , Huang Xiaojin
Published: February 2014

Because of the inherent safety feature, the modular high temperature gas-cooled reactor (MHTGR) has been seen as one of the best candidates for the next generation of nuclear power plants. Power-regulation is a key technique to provide safe, stable and efficient plant operation for the multi-reactor-module power plants. Based on the successful design and operation of Chinese test high temperature gas-cooled reactor HTR-10, the high temperature gas cooled reactor pebble-bed module (HTR-PM) project is proposed. Since the HTR-PM is composed of two interconnected nuclear steam supply system (NSSS), the operation strategy of the multi-reactor-modular HTR-PM is different from that of those one-reactor-modular high temperature reactors such as HTR-10. In order to design and verify the operation strategy for HTR-PM, it is necessary to establish a real-time simulation platform. The design and development of this platform is presented in this paper.
Stress Corrosion Cracking of Nickel Weld Metals in PWR Primary Water
Authors: Xiong Ru , Zhao Yuxiang , Liu Guiliang
Published: February 2014

This discussion reviews the occurrence of stress corrosion cracking (SCC) of Alloys 182 and 82 weld metals in primary water of pressurized water reactors (PWR) from both operating plants and laboratory experiments. Results from in-service experience show that more than 340 Alloy 182/82 welds have sustained SCC, and Alloy 182 with lower Cr have more failures than Alloy 82. Most of these cases have been attributed to the presence of high residual stresses produced during the manufacture aside from the inherent tendency for Alloy 182/82 to sustain SCC. The affected welds were not subjected to a stress relief heat treatment with adjacent low alloy steel components. Results from laboratory studies indicate that time-to-cracking of Alloy 82 (with Cr up to 18%–22%) was a factor of 4 to 10 longer than that for Alloy 182. SCC depends strongly on surface conditions, surface residual stresses and surface cold work, which are consistent with the results of in-service failures. Improvements in the resistance of advanced weld metals, Alloys 152 and 52, to SCC are discussed.
Study on Mixing Process of Two Component Gases in a Stable Stratified Fluid Layer: Effectiveness on Natural Convection
Authors: Takeda Tetsuaki , Funatani Shumpei
Published: February 2014

This study is to investigate an effect of natural convection or natural circulation on a transport process by molecular diffusion in a stratified fluid layer consisting of two component gases. There are many experiment and analysis regarding natural convection or natural circulation in the vertical slot. However, there are few studies on natural convection or circulation and molecular diffusion in the stratified fluid layer consisting of two component gases. It was confirmed that these phenomena appear when the depressurization accident occurs in the very high temperature reactor (VHTR). Therefore it is important to evaluate the transport and mixing processes during the depressurization accident of the VHTR.

The experiment has been performed regarding the combined phenomena of molecular diffusion and natural convection or circulation in a two parallel vertical slots filled with two component gases. The vertical slot consists of the one side heated wall and the other side cooled wall. The other slot consists of the two cooled walls. The dimension of heated wall is 500mm×200mm and thickness is 3mm. The width of the slot is 20mm and the aspect ratio is 25. Combination of nitrogen (N2)/argon(Ar), neon(Ne)/Ar, helium(He)/N2 and He/Ar was used as the first step of the experiment of the two component gases. The density change of the gas mixture and the gas temperature distribution in the slots were obtained. The mixing process of the heavier gas from the bottom side of the slot filled with the lighter gas was discussed in this paper.

The experimental results showed that the transport phenomena by the molecular diffusion were influenced by the localized natural convection or circulation of the gas mixture in the stratified fluid layer. From the experimental results, it was found that the mixing process by molecular diffusion was affected significantly by the natural convection or circulation induced by the slight temperature difference between both vertical walls.
Structure and Erosion Characteristics of Underexpanded Inert-Gas Jets Impinging on a Single Cylinder Submerged in Liquid Sodium
Authors: Kudoh Hideyuki , Sugiyama Ken-ichiro , Narabayashi Tadashi , Ohshima Hiroyuki , Kurihara Akikazu
Published: February 2014

When a heat transfer tube wall in a steam generator of a sodium-cooled fast reactor fails, high-pressure steam leaks into low-pressure liquid sodium side. Then the chemical reaction between steam and liquid sodium causes the erosive and corrosive reaction jet, and the secondary failure of a neighboring heat transfer tube is expected. The objectives of the present experimental study are to understand the structure of an underexpanded impinging gas jet which is injected into liquid sodium and to grasp a relationship between the gas jet velocity and the erosion characteristics of liquid sodium droplets which are entrained into the gas jet.

In the present study, the structure of an underexpanded inert gas jet impinging on a single cylinder within liquid sodium was clarified using the visualization method, which has been constructed by the authors. Then the erosion tests were carried out at the most erosive geometry, which was confirmed by the visualization experiments.

From the visualization, the underexpanded impinging gas jet in liquid sodium could be classified into four regions: core region, droplet flow region, liquid film flow region, and separation region. The erosion phenomenon on the cylinder surface was observed only when the droplet flow region was impinged. In this region, the liquid droplet impingement type erosion was confirmed. The erosion rate, the ratio of the volume loss of specimens to the duration time, was correlated as the function of the gas jet velocity at the end of the external expansion region.
Hydrogen Absorption Behavior of Titanium Alloys by Cathodic Polarization
Authors: Ishijima Yasuhiro , Motooka Takafumi , Ueno Fumiyoshi , Yamamoto Masahiro , Uchiyama Gunzo , Sakai Jun’ichi , Yokoyama Ken’ichi , Tada Eiji , Tsuru Tooru , Nojima Yasuo , Fujine Sachio
Published: February 2014

Titanium and Ti-5mass%Ta alloy has been utilized in nuclear fuel reprocessing plant material because of its superior corrosion resistance in nitric acid solutions. However, Ti alloy have been known to high susceptibility of hydrogen embrittlement. To evaluate properties of hydrogen absorption and hydrogen embrittlement of Ti alloys, cathodic polarization tests and slow strain rate tests (SSRT) under cathodic polarization were carried out. Results show titanium hydrides covered on the surface of metals and hydrides thickness were within 10μm. But hydride did not observed at inner part of metals. Ti and Ti-5%Ta did not show hydrogen embrittlement by SSRT under cathodic charging. These results suggested that Ti and Ti-5%Ta could absorb hydrogen. But hydrogen did not penetrate inner portion of the metals more than 10μm in depth because titanium hydrides act as barrier of hydrogen diffusion. It is considered that retardation of hydrogen diffusion hindered hydrogen embrittlement of Ti and Ti-5%Ta alloys.
Preliminary Study on Thermal-Load Fluctuation Test Using HTTR
Authors: Tochio Daisuke , Shimazaki Yosuke , Iigaki Kazuhiko , Nakagawa Shigeaki , Takada Shoji , Sakaba Nariaki , Sawa Kazuhiro
Published: February 2014

JAEA has designed and developed a commercial very high temperature reactor hydrogen cogeneration system named GTHTR300C. JAEA has investigated that a hydrogen production facility is treated as not a reactor grade facility but a general industrial plant grade facility. The reactor plant system should be designed absorbing the disturbance under AOO and becoming to steady state in the case that thermal-load of the facility is fluctuated or lost. Also, high-accuracy plant dynamics code should be developed to design the GTHTR300C.

In order to demonstrate continuing the reactor operation in if the AOO is occurred in the hydrogen production facility and to obtain the validation data for high-accuracy plant dynamics code, the thermal-load fluctuation tests were planned using the HTTR in JAEA. In this study, it is concluded that the thermal-load fluctuation tests can be carried out without reactor scram and without modification of the HTTR facility and that the test data for validation of the code can be obtained sufficiently.
A Comparison in the Accuracy of Mapping Nuclear Fallout Patterns Using DELFIC and Several Modified Models Based on DELFIC
Authors: Shang Jian-bo , Shi Quan-lin , Wang Qun-shu
Published: February 2014

There are several codes or software used to predict the distribution of fallout after nuclear explosion, such as Hazard Prediction and Assessment Capability (HPAC), Hybrid Single-Particle Lagrangian Integrated Trajectory (HYSPLIT) and Defense Land Fallout Interpretive Code (DELFIC). For the purpose of predicting on-site dose contours, DELFIC has better performance for its computation starts in an earlier time after explosion. Five nuclear fallout prediction codes are studied to determine which code predicts the most accurate fallout dose-rate contours with low computation time. These codes consist of DELFIC and several modified models based on DELFIC, such as Diffusive Transport Module modified with wafer thrower, Cloud Rise Module modified with Hill vortex model and Diffusive Transport Module modified to explain effect of 2-D stabilized cloud. These codes were compared with the Defense Nuclear Agency’s (DNA) DNA 1251-1-EX, Compilation of Local Fallout Data from Test Detonations 1945-1962 Extracted from DASA 1251, using NAD method. In the comparing of small boy test, it appears that there are few differences between different codes. Efforts will be made to find which model predicts less false negative or less false positive and the reasons of these faults will be studied.
Experimental Research of Supercritical Water Heat Transfer in Different Channels
Authors: Li Hong-bo , Zhao Meng , Gu Han-yang , Wang Fei , Zhang Jian-min , Zhang Yong , Yang Jue
Published: February 2014

The experimental research of supercritical water heat transfer has been performed on the supercritical water multipurpose test loop (SWAMUP) with tube, annular channel, and bundles. The normal heat transfer, heat transfer deterioration (HTD) and heat transfer enhancement were observed; and the heat transfer experimental data were obtained. The experimental results show that: the first kind of HTD caused by buoyancy effect only occurs with low mass flow velocity and high heat flux when the fluid temperature is below pseudo-critical point in all the tested channels and the second kind of HTD caused by acceleration effect always occurs when the fluid temperature reaches pseudo-critical point in tube and annular channel; the heat transfer enhancement occurs when the fluid temperature reaches pseudo-critical point with low mass flow velocity in tube; and the heat transfer enhancement in bundles is caused by the space grids. It is concluded that the heat transfer in bundles is better than in other tested channels.
Steam Generator Secondary Side Chemical Cleaning During Reactor Plant Cooling Down
Authors: Xia Xiaojiao , Wen Juhua , Ma Weigang
Published: February 2014

For safe and reliable operation of NPP steam generators, it is required to remove the sludge from heat exchanging tubes and steam generator (SG) volume in due time. Chemical cleaning technology of SG secondary side during NPP cooling down will be used in Tianwan NPP, which was used to remove iron and copper oxides from steam generators secondary side in Russia NPPs and to resume the heat exchange capacity of heat exchange tubes. To validate and evaluate the effectiveness and safety of the SG cleaning formula during NPP cooling down provided by Russia (RF) for Tianwan NPP, cleaning effective tests and safety tests were done in autoclave with the chemical cleaning process parameters simulated. To compared with RF, cleaning effective tests and safety tests were done with A3B1 under the same condition. Cleaning effective test results showed that the simulated sludge for Tianwan NPP can be more effectively dissolved and removed with A3B1 than with RF. Cleaning safety test results showed that the general corrosion amount of 0Cr18Ni10Ti stainless steel was very low but the general corrosion amount of SA508-III steel was high both with A3B1 formula and the formula provided by Russia.
Study of Pulsating Laminar Flow Field and Resistance Characteristics in Rectangular Channel
Authors: Liu Yusheng , Zhang Chunming , Tan Sichao , Wen Lijing , Ma Shuai
Published: February 2014

Pulsating laminar flow field and resistance characteristics in rectangular channel are investigated analytically and experimentally. Mathematical model is established. The influences of flow fluctuation period on the flow field in rectangular channel are analyzed. The velocity gradient in the pulsating laminar flow is compared with that in the steady laminar. Analytical results are verified by experiment.

The results show that flow pulsation causes phase difference between the velocity and pressure drop. Compared with steady flow filed, the flow field changes a lot under pulsating condition due to the oscillating velocity component. The velocity gradient in the pulsating laminar flow is changed periodically by the oscillating velocity component, which leads some change in friction coefficient.
Preliminary Research and Design in the Nuclear Island Systems of 1500MWt Standing-Wave Reactor
Authors: Duan Tianying , Zhou Peide , Long Bin , Hu Yun , Liu Yizhe , Huang Chen , Yu Huajin , Sun Gang , Zhang Yuanyuan , Yu Chunli , Feng Weiwei , Liu Haojie
Published: February 2014

As one kind of fast reactor, the Traveling-Wave Reactor (TWR) utilizes depleted uranium with a small amount of enriched uranium/ plutonium which is used to kick off the chain reaction. The TWR can run for decades without refueling or removing any used fuel from the reactor. The most challenging issues on TWR are fuel design, structural material for fuel cladding, core physics process analysis and core physics design. Based on the present technology of fuel and structural material, a new concept named Standing-Wave Reactor (SWR) which is the preliminary stage of the TWR is proposed. The wave of fission would move through the depleted uranium core by fuel transfer in SWR. According to the concept of SWR and the published data of fuel and material, the R&D works on 1500MWt SWR have been performed, which cover the reactor core, reactor structure, process system et.al. The preliminary results confirm the feasibility of SWR. Meanwhile, the design of reactor core and the main systems which is based on the technologies of available pool sodium-cooled fast reactor has been accomplished.
Study on Seismic Fragility of Air Heat Exchanger for CEFR
Authors: Fu Zhiwei , Zhang Donghui , Zhang Tianyi , Zhang Chunming , Zuo Jiaxu , Chen Yan , Song Wei , Wang Zhe
Published: February 2014

Seismic Fragility Analysis is studied, then the basic fragility analysis variables and their quantification methods which are used to analyze and calculate the seismic fragility of the air heat exchanger of passive accident heat removal system for China Experiment Fast Rector (CEFR) are introduced. Next, the median response spectrum in NUREG0098 is used to calculate median response factors of both structure and equipment, and finite element analysis is used to determine the capacity factor for the equipment. Finally, the results show that the median seismic capacity of the air heat exchanger is 3.53g which has the inherent randomness about the median and the uncertainty in the median value, respectively, 0.36 and 0.47, and the high confidence low probabilistic failure (HCLPF) capacity is 0.9g, so the air heat exchanger has a better seismic capacity.
Effect of Physical Properties on Gas Entrainment Rate From Free Surface by Vortex
Authors: Naosuke Ohte , Koizumi Yasuo , Kamide Hideki , Ohno Shuji , Ito Kei
Published: February 2014

Gas entrainment rate into liquid by a vortex formed on free surface was examined experimentally. Four kinds of test fluid were used; water at 25 °C, water at 60 °C, 20 cSt silicone oil and kerosene. Gas was air. The flow state of gas entrainment was visually observed by using a high speed video camera. The gas entrainment rate into liquid was measured. When liquid velocity was low, bubble-type gas entrainment occurred. As the liquid velocity increased, the gas entrainment type turned from the bubble type to a vortex type and gas entrainment rate considerably increased. The relation between gas entrainment rate and liquid velocity was mainly affected by the viscosity of liquid. As viscosity became large, higher liquid velocity was required to get the same gas entrainment rate. The effect of surface tension on the gas entertainment rate was minor or little. No systematic trend by the surface tension was noticed in the gas entrainment rate As liquid velocity increased, vortex became deep in the test vessel without gas entrainment occurrence and eventually a vortex tip reached a bottom outlet. After the vortex tip reached the bottom outlet, the tip penetrated into an outlet piping. Vortex growing speed becomes slow. Then, bubble-type gas entrainment was initiated from the vortex tip. A further increase in liquid velocity resulted in a transition from the bubble-type gas entrainment to vortex-type gas entrainment. By assuming that liquid flow was free falling film flow in an outlet pipe, liquid velocity was derived. The Kelvin-Helmholtz instability wave length was calculated for this liquid velocity. The wavelength that was observed at the vortex tip interface in the condition that bubbles were torn off from the tip was close to the Kelvin-Helmholtz instability wave length in both bubble-type and vortex-type gas entrainment.
Numerical Simulation of Supercritical Water Heat Transfer in the Vertically Heated Tube
Authors: Wang Feng , Cui Bo , Zhang Shijun , Qin Xue
Published: February 2014

The heat transfer of water in a vertically heated tube at 24.52 MPa is numerically simulated by computational fluid dynamics software of FLUENT. The IAPWS-IF97 formulation is applied to obtain the water properties, which vary substantially at supercritical condition. The two-dimensional axi-symmetric model using RNG k-ε turbulence model with enhanced wall treatment gives fine prediction of wall temperature and heat transfer coefficient. The mesh size near the wall adapted smaller when at high heat fluxes for the accuracy of computed results. The wall heat fluxes were set to be 233, 698, 930 and 1100 kW/m2 to match the simulation with experiment performed by Yamagata. It is found that k-ε turbulence model with enhanced wall treatment can give outstanding prediction of heat transfer enhancement and heat transfer deterioration. The heat transfer coefficient value reaches a maximum near the pseudocritical point and it decreases with increase of heat flux.
The Design of a Small Annular Linear Induction EM Pump for the Transportation of Liquid Sodium in the SFR
Authors: Kim Hee Reyoung
Published: February 2014

The linear annular induction electromagnetic (EM) pump was designed for the purpose of transporting the electrically conducting liquid sodium that is used as a coolant in the Sodium cooled Fast Reactor (SFR) in the high temperature of 550 °C. The EM pump was designed by using an equivalent circuit method which is commonly employed to the design of the induction machines. The mathematical equations on the developing pressure and efficiency was found out by using Laithewaite’s standard design formula. The main geometrical parameters for the design were a pump core length, a diameter of the pump and an annular gap size. The dominant electromagnetic variables were the input frequency, magnetic pole pitch and number, and coil turns. Especially, it was shown that the developing pressure and efficiency were maximized at the range of the low frequency less than 60 Hz of the commercial frequency. Also, the size of an annular gap was suitably selected considering the hydraulic frictional loss at the narrow annular channel. The P-Q characteristic was theoretically predicted according to the change of the input current, voltage and power. The functional and structural components of the pump consisted of the material compatible with the high temperature and chemical reactivity of the liquid sodium. The designed pump was predicted to have the nominal flowrate of 10 L/min and the developing pressure of 4 bar.
Numerical Simulations and Experiments on Gaseous Flow and Heat Transfer in Multiple Narrow Channels in Transition Region
Authors: Chen Yuming , Klein Christine , Arbeiter Frederik , Ibaceta German
Published: February 2014

The International Fusion Materials Irradiation Facility (IFMIF) is designated to generate a materials irradiation database for the future fusion reactors. The High Flux Test Module (HFTM) houses the test specimens which will be cooled with helium flows passing through several parallel narrow channels. The Reynolds numbers for individual sub-channels range from about 1000 to 8000. For most sub-channels the flow is in the transition region, which cannot be accurately predicted by using laminar or turbulence models. As a part of the HFTM validation activities, the thermo-hydraulic tests of a 1:1 single irradiation rig were carried out. The experiments also provide data for validating the CFD codes for using in the simulations of the whole HFTM. In this paper, the CFD simulations and experiments will be compared for helium flows at a Reynolds number between 3000 and 8000. The simulations were carried out with the commercial CFD code Ansys-CFX V14 using several turbulence models. It is found that, the CFX Gamma model with the transition onset Reynolds number of about 180 provides the best predictions for the transition flows.