bluuebunny/crossref_metadata_2025 huggingface dataset:
Improving Public Acceptance to Nuclear Power: Policy, Practice and Experience With Public Communication
Authors: Xiaofeng Zhang , Feng Zhao , Rongxu Zhu , Zongzhen Yang , Zhihong Shangguan
Published: October 2017

With the development of public awareness on environmental protection, especially after the Fukushima nuclear accident, the opposition to nuclear power due to NIMBY (not in my back yard) effect begins to hinder the rapid development of Chinese nuclear industry. For example, in recent years several large-scale mass incidents with appealing to stop the siting and construction of nuclear facilities in China have put related projects (including nuclear power plant and nuclear fuel cycle facility) into termination, resulting in certain financial loss and unnecessary social unstabilization, thus causing more and more concern from administrative authority, research institution and nuclear industry.

To strengthen public acceptance on nuclear power, related enterprises such as CGN and CNNC have made great efforts in information disclosure to eliminate mysterious feelings towards nuclear power and expect to build new impression as clean energy. Domestic institutions and universities carry out plenty of work on methods to help public correctly perceive nuclear risk and present strategies for effective public communication. Administrative authority also issued detailed guidance on public communication required to be fulfilled during plant’s siting phase, which provided explicit provisions on the responsibility and job content of different entities.

Here we will take one public communication practice of one nuclear power project located in south Zhejiang region as an example. In this scenario, we face more difficulty than other projects, such as doubt from local government, complexity of public types, and large amount of stakeholders. In this paper, we will make summary on endeavors to improve public acceptance, such as large amount of NPP visits, comprehensive scientific popularization, direct communication with stakeholders and integration development between local society and nuclear industry. And we will discuss the feasibility of innovative practice, combining several similar tasks needed in different subjects, such as environmental impact assessment and social stabilization assessment, to fulfill at once. To achieve this goal, we design specific questionnaire and use it to survey the opinion of more than 800 people in the fairly large region across different provinces, covering 30km radius area of site, which gains satisfactory results.

By comparing outcomes of opinion surveys carried out before and after the practice, we will put forward to the considerable effect of public communication in improving public acceptance to nuclear power, and analysis the pros and cons of this example. Moreover, we also expect the good experience in practice can be promoted to overall processes of nuclear power plant, including siting, construction, commission and life extension, helping nuclear power gain more public acceptance.
Operating Reliability Assessment of FPGA-Based NPP I&C Systems: Approach, Technique and Implementation
Authors: Babeshko Eugene , Bakhmach Ievgenii , Kharchenko Vyacheslav , Ruchkov Eugene , Siora Oleksandr
Published: October 2017

Operating reliability assessment of instrumentation and control systems (I&Cs) is always one of the most important activities, especially for critical domains like nuclear power plants (NPPs). Intensive use of relatively new technologies like field programmable gate arrays (FPGAs) in I&C which appear in upgrades and in newly built NPPs makes task to develop and validate advanced operating reliability assessment methods that consider specific technology features very topical. Increased integration densities make the reliability of integrated circuits the most crucial point in modern NPP I&Cs.

Moreover, FPGAs differ in some significant ways from other integrated circuits: they are shipped as blanks and are very dependent on design configured into them. Furthermore, FPGA design could be changed during planned NPP outage for different reasons. Considering all possible failure modes of FPGA-based NPP I&C at design stage is a quite challenging task. Therefore, operating reliability assessment is one of the most preferable ways to perform comprehensive analysis of FPGA-based NPP I&Cs. This paper summarizes our experience on operating reliability analysis of FPGA based NPP I&Cs.
SCORPIO-VVER: Two Decades of Experience and Enhancements in Reactor Core Monitoring and Surveillance in Central Europe
Authors: Molnar Jozef , Pecka Marek , Kment Jaroslav
Published: October 2017

During the years 1970–80’ in the satellite countries of the former Soviet Union more than 24 new reactor builds were started. In the former Czechoslovakia, the new builds were realized with a high degree of input from the local engineers and the local industry. This territory up to today has an indigenous nuclear industry, machinery and nuclear engineering background.

Starting with the year 1972, on the territory of Czechoslovakia (now on the territory of the Czech and Slovak Republic) 12 new units of the VVER-440 type of reactors were started to build. Nowadays the 2 oldest units were already shutdown in Slovakia, 4+4 units are still operating in both countries, and 2 units of VVER-440 model V213 are still under construction in Slovakia. The reactors designed lifetime in original configuration are 30 years.

During these units’ operation period, wide range of modifications and upgrades were performed to strengthen the nuclear safety and the reactors operability. In 2015/16, activities related to the unit operation licenses extension were carried out in both of countries.

In scope of strengthening the reactor’s core monitoring and surveillance, at Dukovany NPP (CZ, 1998) and at Bohunice NPP (SK, 2001) the original Russian VK3 computation system was completely replaced with an alternative advanced Core Monitoring and Surveillance System (CMS) SCORPIO-VVER. In Hungary, a locally developed “Verona”, and on the units under construction in Slovakia the Russian “Kruiz” CMS is used.

Nowadays the SCORPIO-VVER CMS presents a nuclear fuel type and fuel vendor independent, advanced computer based reactor core monitoring system with an open and flexible framework, including the latest achievements in the fields of N/F and T/H for reliable and safe reactor operation with high efficiency of fuel cycle. The system’s framework governing the know-how and knowledge of 5 European institutes with proven experiences with reactor operation, fuel pattern and fuel campaign design and with utilizing the existing unit’s project reserves to increase the reactor operation and fuel campaign efficiency.

Since the first installation the SCORPIO-VVER CMS system has a remarkable operating history and experience. More than 18 years of experiences at 6 units of VVER-440 type of reactors in two different countries helps to put the system to a very high level of usability and reliability. Even if the SCORPIO-VVER is installed only on VVER-440 reactors, it could be adapted to the needs of other VVER type of reactors and to needs of education and training centers too.
Investigation on Heat Transfer Characteristics in the Test Section With Non-Uniform Heat Flux Distribution Under Natural Circulation
Authors: Wang Qiang , Gao Puzhen , Chen Xianbing , Wang Zhongyi , Huang Ying
Published: October 2017

Natural circulation served as an indispensable part of nuclear, attracted much more attentions in recent years. It does not need a pump to provide power. The operating principle of natural circulation caused its complexity in analysis process. It was still a difficult issue to reveal the law of natural circulation accurately. Many experiments and calculations had to be conducted to study the basic physical regulation. This paper concentrated upon the heat transfer characteristics in the test section with two different types of heat flux distribution. The two types of heating flux distribution in the test section were linear and chopped cosine along axial direction. Based on a natural circulation experimental facility, physical models and mathematic models were established. RELAP5 code was used to calculate the thermal hydraulic state of natural circulation loop. The variation of heat transfer coefficient along flow direction was different. It was tightly related to heat flux. Some relevant experiments were conducted in many different conditions and steady sate experimental data were achieved to verified theoretical calculations. Experimental data, such as water temperature, wall temperature and flow rate were recorded when the system is stable. The heat transfer coefficients were calculated according to the experimental data. The factors that affected the heat transfer characteristics of natural circulation were analyzed by comparing the heat transfer coefficient under different conditions. The heat transfer coefficient was calculated according to the empirical correlations as well.

After a series of analysis, the results indicated heat transfer coefficient had an obvious difference, which influenced ability of natural circulation. Comparing with experimental data, the evaluation of different empirical correlations was conducted in two test sections. Some empirical correlations turned out to be suitable for the estimation of heat transfer in experiment facility. The increase of heat flux could enhance heat transfer process in the two test section under low pressure. Average heat transfer coefficient increased with the decrease of inlet subcooling degree. The system pressure effected the heat transfer characteristics of natural circulation as well. The increase of mass flux would promote heat transfer while the level was different. RELAP5 had a great agreement with experimental data in single phase flow. Natural circulation ability was influenced by the position of average heat source center, which was slightly different in the research objects. The research would lend strong empirical support to the guideline of experiment and subsequence study in natural circulation.
Turbulent Mixing Characteristics in the Converging Region of Five-Parallel Jets
Authors: Zhuang Nailiang , Yang Bonan , Wang Lianfa , Yuan Hongsheng , Tan Sichao
Published: October 2017

Plate-type reactor fuel is getting increasing attentions as it features excellent heat transfer ability and compact structure. Turbulence mixing accompanied with momentum and mass exchange is typical phenomena in the converging region downstream of parallel plane fuels. To deepen understanding turbulence mixing characteristics and obtain more benchmark data for CFD, an experimental study was conducted by 2D particle image velocimetry (2D-PIV). Ensemble average profile of velocity, vorticity, turbulence intensity, as well as Reynolds stress are analyzed, respectively.

Results reveal that two kinds of merging points (mp) were found, i.e., flows from two side jets merge at y/d = 3 (mp1), and flows from middle three jets merge at y/d = 7.6(mp2). The decay of vertical velocity decreases fast primarily until y/d = 6, then slightly increases until y/d = 8.8, and finally decreases gradually. Decay of velocity magnitude of present study decreases sharply until y/d = 6.4, then decreases gradually. This tendency of five-parallel jets is similar to the result of two-parallel jets. From flow visualization, the vortex scale decreases once the vortex formed. The vorticity reaches its maximum at about y/d = 1.3 and then decreases gradually. Consistently with the indicated of velocity distribution, the vorticity distribution tends to mild with the flow developing downstream. However, the location of vortex center is affected by spanwise momentum exchange. Vortices started moving closer around the mp2 indicating a strong combination of activities. Spanwise turbulence intensity shows strong fluctuations existed around mp1 implying that the main momentum transfer happened in the merging region.
Experimental Study on the Steam-Water Co-Current Flow Friction Multiplier of the AP1000 Surge Line
Authors: Yu Jiangtao , Zhang Dalin , Shi Leitai , Wang Zhiwei , Tian Wenxi , Su Guanghui , Qiu Suizheng
Published: October 2017

The occurrence of Steam-Water Co-Current Flow (SWCOCF) in the AP1000 Pressurizer Surge Line (PZR SL) is an important phenomenon during the depressurization stage of the Small Break Loss-of-Coolant Accidents (SB-LOCAs), and the predictions of SWCOCF pressure drop in PZR SL are important issues in the AP1000 reactor safety analyses. To investigate the pressure drop characteristics of SWCOCF in the PZR SL assembly of AP1000, a PZR SL test model made of stainless steel is designed with scaling ratio of 1:9.15, and a test loop is also established to supply SWCOCF conditions for the experiments. The test section is mainly composed of a long inclined tube with a short acrylic visual section, a whole surge line tube, and a vertical section. The SWCOCF pressure drop experiments are conducted with pressure ranging from 0.1MPa to 0.24MPa and vapor mass fraction from 0.02 to 0.5. The research results show that: the whole surge line two-phase friction multiplier increases with the increasing of steam fraction, and decreases with the increasing of pressure. A prediction model of two-phase friction multiplier for the PZR SL is developed based on the Martinelli-Nelson method, which can predict the tests data very well.
Bubble Formation Characteristic of Submerged Single-Hole Orifice in Inorganic Solution
Authors: Wen Jiming , Gu Haifeng , Sun Qiunan , Sun Zhongning , Zhou Yanmin , Li Yingzhi
Published: October 2017

Bubble column is regarded as a kind of wet filtration method. The solution used in filter containment vented system is composed of a large amount of inorganic salt including sodium hydroxide and sodium thiosulfate. Solution with high inorganic salt concentration is more viscous and has stronger surface tension than distilled water. This property has significant effect on bubble size and bubble deformation during formation process. Besides bubble coalescent is suppressed with inorganic concentration increase. This phenomenon is found to possess influence on bubble formation regime when bubble formation regimes belong double, paring and injection regime. The existence of aerosol in solution is another factor that has obvious effect on bubble formation regime and the transition velocity between regimes. This paper apply high speed camera to study the influence rule of inorganic salt and aerosol concentration on bubble size during formation and formation regimes transition.
Development of Security and Safety Fuel for Pu-Burner HTGR: Part 3 — Preliminary Reactor Safety Analyses Under Depressurization Accidents
Authors: Nakano Masaaki , Ohashi Kazutaka , Okamoto Koji
Published: October 2017

Preliminary reactor safety analyses were performed for the Pu-burner HTGR (High Temperature Gas-cooled Reactor) to investigate plutonium fuel influence on reactor thermal and kinetic behavior for the depressurization accident which is design base accident, and the depressurization accident without scram, which is beyond design base accident. The analytical results of depressurization accident show that the reactor thermal power is restricted lower than the reference reactor with uranium fuel, because the core width is wider for better Pu-burning performance. The analytical results of depressurization accident without scram show that the re-criticality is delayed compare to the reference reactor with uranium-fueled core, due to its temperature reactivity coefficient. The delay of re-criticality brings that operators have more margins for accident management, for example, manual insertion of shutdown devices to the reactor core.
Analysis of Primary Coolant Corrosion Product Effect on Solid Waste Source Term
Authors: Xiaoxia Wang , Zhuoran Li , Liying Zhang , Wei You , Song Liu , Aijun Mi , Yawei Mao , Guiling Gao
Published: October 2017

Corrosion product in the primary coolant is formed mainly by structure material corrosion and activated corrosion product. Under normal condition, source term of corrosion product maintains at a low level while its radioactive concentration reaches a peak value after the primary is injected with hydrogen peroxide during shutdown oxygenation operation. Following the procedure of corrosion product elimination, analysis has been made on effect which accumulated corrosion product exerts on RCV pre-filter and mixed bed demineralizer. Different primary coolant source terms and the cold shutdown peak value of corrosion product have been considered in the analysis. Under the assumption that dissolved and particle dissolved product has a fixed ratio of 1:1, the surface dose rate of RCV pre-filter varies as the different primary coolant varies. Nuclides proportion has been listed and charted based on the calculation. If cold shutdown condition has been taken in primary coolant source term calculation, then the RCV pre-filter surface dose rate will reach 345Sv/h; If primary coolant design value has been taken in source term calculation, then the RCV pre-filter surface dose rate will reach 42.2Sv/h; If 1/3 primary coolant design value has been taken in source term calculation, then the RCV pre-filter surface dose rate will reach 14.1Sv/h. For the waste resin in RCV mixed bed demineralizer, the corrosion product leaves a small contribution (less than 5%) if the cold shutdown is not considered in primary coolant source term calculation; and a contribution of over 90% if the cold shutdown is considered. The analysis provides a radiation safety analysis basis for NPP solid waste.
Experimental Study on Chugging Condensation Regime of Steam Direct Contact Condensation in Water Flow in a Tee Junction
Authors: Jie Wang , Tao Lu , Bo Liu
Published: October 2017

The phenomenon of direct contact condensation (DCC) is encountered in both the nuclear industry and conventional industry and it is intimately associated with safety operation of relevant equipment. Studies on steam jet discharged directly into a stagnant water pool have been studied extensively; however, they are sparse for direct contact condensation of steam in sub-cooled water in pipes. In this paper, the experiments were performed to study the direct contact condensation of low mass flux saturated steam discharged into sub-cooled water flow in a Tee junction. By using high-speed camera, we investigated chugging condensation characteristics including two-phase interface movements. It is can be seen from the experiment results that, for chugging, steam plume and surrounding sub-cooled water undergo the following process: steam plume grows to the maximum, then steam plume becomes smaller due to condensation, and finally, water enters the vertical branch pipe and steam ejects from the vertical branch pipe again.
Methodology and Application of Human Reliability Analysis in CAP1400 Nuclear Power Plant
Authors: Qiu Yongping , He Jiandong , Hu Juntao , Zhuo Yucheng , He Jie
Published: October 2017

It is well recognized that humans play an important role in the safety operation of nuclear power plants (NPPs). Usually three types of human interactions (HIs) are defined in the human reliability analysis (HRA) of probabilistic safety assessment (PSA) for NPPs, i.e., pre-initiating event HIs, initiating event-related HIs, and post-initiating event HIs. In this paper, a brief introduction of the HRA methodology for CAP1400 nuclear power plant is first presented, including internal events and external events (mainly internal fire and flooding) HRA. Next, the CAP1400 human failure event quantification content is given with a typical example, and some insights and proposals based on CAP1400 PSA/HRA results are discussed. Finally, the application of HRA in human factor engineering design of CAP1400 is described. The human actions (HAs) most important to safety are identified via a combination of probabilistic and deterministic analyses, and then addressed when conducting the human factor engineering program. The CAP1400 HRA is one of the most important PSA elements and provides fundamental support for CAP1400 PSA and the relevant applications.
Application of Internal Fire PRA in Elimination of Fire Common Modes
Authors: Liu Xiang , Huang Lihua , Liu Zeyu
Published: October 2017

During the design and construction phases of nuclear power plants (NPPs), fire protection design principle requires that the safety classified equipment within the nuclear island, the service water pump building and galleries should meet the safety goal whenever a design basis fire occurs. The design of passive facilities, such as structure of buildings and physical separation of equipment, should prevent one single fire from failing all redundant equipment of the same safety function. When one single fire fails all redundant equipment of the same function, it is called a fire common mode. By performing fire vulnerability analysis of fire areas or fire zones, analysts can identify the possible fire common modes, and confirm if they are acceptable or not through further analysis. For NPPs in operation, the vulnerability analysis can be performed by examining the details of plant specific cables and equipment. The confirmed unacceptable common modes should be modified or protected through changing cable routing or adding fire coating, etc. Moreover, by employing internal fire probabilistic risk analysis (PRA), analysts can categorize the identified fire common modes according to severity, assess the changes in the risk of internal fire after elimination, and evaluate the benefit of performing elimination, which provides reference for engineering modification by showing the order of priority of each modification. This article introduces the process of identifying fire common modes, performing internal fire PRA and providing suggestions of modification, and illustrates the application of internal fire PRA in eliminating fire common modes and enhancing the safety level of NPPs.
Development of Security and Safety Fuel for Pu-Burner HTGR: Part 4 — Study on 3S-TRISO Fuel Fabrication
Authors: Saiki Yohei , Honda Masaki , Takahashi Masashi , Ohira Koichi , Okamoto Koji
Published: October 2017

In order to develop 3S-TRISO fuel for Pu-burner High Temperature Gas Reactor (HTGR), we conducted lab. scale experiments such as preparation test of simulated fuel kernel; CeO2-YSZ particle, and coating pre-test with simulated kernel. In the preparation test, based on the actual achievement of manufacturing fuel for High Temperature Engineering Test Reactor (HTTR)[1], we tried to fabricate some CeO2-YSZ particles through external gelation process. As a result, we successfully obtained the manufacturing parameters that can prepare good particles. In addition, we carried out some parametric coating test with fluidized-bed equipment and ZrO2 particle as simulated ZrC coated fuel kernel, and obtained the prospect of the possibility to coat the layer having desired thickness.
Development of a Multiphysics Approach With ANSYS CFX and PARCS for Local Analysis of Neutron Oscillation
Authors: Du Zhuoqi , Seidl Marcus , Macián-Juan Rafael
Published: October 2017

Neutron noise analysis has been done over the decades to predict fuel assembly vibrations and to evaluate safety related issues. Neutron noise occurs due to several reasons: the vibration of the fuel rods, flow obstacles such as rod bending and crud deposition, the moderator temperature and time dependent changes caused by varying flow distributions within a fuel assembly, etc. In order to have a better insight of the neutron noise, a fluid mechanics, structural and neutronics coupled code is developed to perform detailed multiphysics simulations at the level of the fuel rods inside a fuel assembly. In this paper the coupling routine of both steady state and transient calculation is described and the outcome is discussed under several scenarios to understand the influence of rod vibration, moderator temperature and flow distribution on the neutronic field. This paper presents the methodology to couple the multiphysics Computational Fluid Dynamics (CFD) code ANSYS-CFX 16.0 with the 3D neutron diffusion code PARCS v3.0. The model for a 16×16 Pressurized Water Reactor (PWR) fuel assembly is set up for ANSYS-CFX. A sensitivity analysis is carried out to obtain the optimal mesh parameters which results in a good accuracy, as well as a small need for computation capability. Transient cases are studied on a quarter fuel assembly applying oscillating moderator inlet boundary conditions in which the inlet moderator temperature and the inlet moderator velocity are varying over time. In order to simulate the vibration of the fuel rod, the fuel rod part is implemented as immersed solid in ANSYS-CFX. Different vibration modes are applied to both cases: individual single rods of the fuel assembly, and all rods of the fuel assembly. The results of each case are shown in this paper giving a better understanding of how axial power distribution develops with varying flow conditions and vibrating fuel rods.
Evaluation on CAP1400 Passive Containment Cooling System Capability
Authors: Di Zhang , Xin Liu , Zhe Wang , Guodong Wang , Chenxiao Ni , Shengjie Wei , Benxue Hu , Zhangli Wang
Published: October 2017

CAP1400 is the large advanced passive demonstration plant, which is one of the National Science and Technology Major Projects. CAP1400 is an innovative design and development based on AP1000 advanced design philosophy with passive safety features. While keeping the same safety level as AP1000, CAP1400 aims to improve plant economy by means of enlargement of reactor core, increase of unit power output and optimization of plant overall design. The CAP1400 nuclear power plant uses the passive containment cooling system (PCS) to remove heat from the containment after accidents when high mass and energy fluid is released from primary or secondary systems. On account of the more mass and energy release and the internal structures design changes of CAP1400, the operating parameters may exceed the range of AP600 and AP1000 containment tests. In order to further study the performance of the CAP1400 PCS system, a series of verification tests have been carried out in the National PWR Project.

This paper describes the background of the test setup, the really focused issues of the experiment and the main test results. The experimental data from separate effect tests and large scaled test were analyzed and evaluated in depth. The performance of the PCS was studied, including water film coverage outside the containment shell, water film evaporation and heat transfer characteristics, air natural circulation and convection outside the containment vessel, convection and condensation of the air / steam mixture on the inner surface of the containment vessel, as well as the overall performance of PCS systems. The results show that the important correlations applied in CAP1400 containment pressure and temperature response analysis could predict the heat and mass transfer reasonably and the envelope factors used for safety analysis are conservative. Parameters affecting the containment pressure and temperature have been investigated, and the reasonability and conservation of containment thermal and hydraulic analysis code has been evaluated by comparing the calculated results with the measured results. The results show that containment analysis code could simulate the important heat transfer phenomena of the test containment well and predict the temperature and pressure response reasonably with some conservation.

Containment pressure and temperature response analysis after typical design basis accident analysis have been preformed, and the results show that the CAP1400 PCS is sufficient to remove the heat from the containment which ensure the safety of the plant. All of the results provide firm basis for the development of CAP1400 and sufficient support for the safety review. This work also lays the foundation for the further development of relevant analytical tools and the development of advanced and passive technologies with higher power levels.
Development of Security and Safety Fuel for Pu-Burner HTGR: Part 2 — Design Study of Fuel and Reactor Core
Authors: Goto Minoru , Ueta Shohei , Aihara Jun , Inaba Yoshitomo , Fukaya Yuji , Tachibana Yukio , Okamoto Koji
Published: October 2017

JAEA (Japan Atomic Energy Agency) has conducted feasibility studies of the fuel and of the reactor core for the plutonium-burner HTGR (High Temperature Gas-cooled Reactor). The increase of the internal pressure, which is caused by generations of CO gas and stable noble gases, is considered to be the one of the major causes of TRISO (TRI-structural ISO-tropic) fuel failure at high burn-up. The CO gas is generated by the chemical reaction of the graphite making up the buffer layer with the free-oxygen released from the fuel kernel by fission. The stable noble gases, which are fission products, are also released from the fuel kernel. Although it is considered very difficult to suppress the increase of the partial pressure of the stable noble gases because of its chemically inert nature, the increase of the CO gas partial pressure can be suppressed by reducing the free-oxygen mole concentration using a chemical reaction. ZrC acts an oxygen getter, which reduces the free-oxygen generated with fission reaction. An increase of the CO gas partial pressure with burn-up in a TRISO fuel is expected to be suppressed by coating ZrC on a fuel kernel. A PuO2-YSZ (Yttria Stabilized Zirconia) fuel kernel with a ZrC coating, which enhances safety, security and safeguard, namely: 3S-TRISO fuel, was proposed to introduce to the plutonium-burner HTGR. In this study, the efficiency of the ZrC coating as the free-oxygen getter under a HTGR temperature condition was examined based on a thermochemical calculation. A preliminary feasibility study on the 3S-TRISO fuel that enables to attain a high burn-up around 500 GWd/t was also conducted focusing on a fuel failure caused by an increase of the internal pressure. Additionally, a preliminary nuclear analysis was conducted for the plutonium-burner HTGR with a fuel shuffling in the radial direction. As a result, the thermochemical calculation result showed that all the amount of the free-oxygen is captured by a thin ZrC coating under 1600°C condition. The plutonium-burner HTGR will be designed to suppress fuel temperature to be lower than 1600°C under severe accident conditions, and hence it was confirmed that coating ZrC on the fuel kernel is very effective method to suppress the internal pressure. The internal pressure the 3S-TRISO fuel at 500 GWd/t is calculated to be lower than 60 MPa, which allows to prevent the fuel failure, and hence the feasibility of the 3S-TRISO fuel was also confirmed. Additionally, the results of the whole core burn-up calculations showed that the fuel shuffling in the radial direction allows to achieve the high burn-up around 500 GWd/t. It also showed that the temperature coefficient of reactivity is negative value during the rated power condition through the operation period.
Integrity Analysis of Electrical Penetration With Intense Radiation Heat Transfer by Hydrogen Diffusion Flame
Authors: Huang Gaofeng , Zhang Kun , Wang Jiayun
Published: October 2017

While hydrogen release into large compartment from confined compartment, hydrogen diffusion flame is easy to occur. There is intense heat radiation effect on electrical penetration from diffusion flame. Aim to evaluate the influence of diffusion flame on electrical penetration, systematic method is constructed, including computing view factor of electrical penetration, assessment of hot spot of containment vessel and research of heat transfer for electrical penetration. Research results give theory basis for determining location of venting which can generate hydrogen diffusion flame. The method can be extended to use in the influence evaluation of personnel hatch and equipment hatch in the containment vessel.
The Evaluation Method of Uranium Hexafluoride Leakage Accident Release Source Term
Authors: Lei Chen , Ning Wang
Published: October 2017

Uranium hexafluoride is the intermediate material of uranium fuel enrichment process, which is widely used in uranium conversion plant, uranium enrichment plant and nuclear fuel element plant[1]. Because of its active chemical properties and its radioactive and chemical toxicity, great importance should be attached to the uranium hexafluoride release accident. This paper describes the possible leakage scenarios for uranium hexafluoride accident. And the general step of the evaluation for uranium hexafluoride leakage accident release source term is given, as well as an application example for the feed facility of a gaseous diffusion plant.
CFD Simulation of Thermal-Hydraulic Characteristics of Water Seal Structure in PWR Pressurizer
Authors: Wang Xi , Wu Y. W. , Tian W. X. , Qiu S. Z. , Su G. H.
Published: October 2017

The pressurizer is the fundamental equipment in nuclear power plant, maintaining the pressure in the primary side. A U-shaped tube filled with water, as the water seal structure, was installed in front of the safety valve with the purpose of reducing the non-condensable gas leaked through the safety valve. When the safety valve opens, water slug in the U-shaped tube moves through the safety valve, causing a large number of thermal-hydraulic loads on the safety valve and downstream line. In order to reduce the thermal-hydraulic loads on the safety valve system, the U-shaped tube water seal structure was placed inside the pressurizer dome and replaced by a scoop-shaped structure. Thermal-hydraulics characteristics of the water seal structure are simulated based on the geometric model with a scale ratio of 1:1 to investigate the condensation and formation of water seal. The key parameters in the water seal formation process are investigated under different pressure and heat dissipating capacity. The species transport model is utilized to describe the impact of non-condensable gas under various mass fractions. Three-dimensional distributions of pressure and temperature are obtained from the calculation by using the CFD code ANSYS FLUENT. The water seal formation time is calculated by using the condensation rate and the geometric model. The result reveals that water seal formation can be completed within the required time, even under a high mass fraction of non-condensable gas. Water seal formation time reduces when the system pressure increases. The temperature difference across the water seal is lower than 20K.
Application of Probabilistic Safety Assessment in the Design of Instrumentation and Control Systems for Nuclear Power Plants
Authors: Peng Pengyi , Liu Weidong , Yang Zhichao
Published: October 2017

Instrumentation and control (I&C) systems in nuclear power plants (NPPs) have the ability to initiate the safety-related functions necessary to shut down the plants and maintain the plants in a safe shutdown condition. I&C systems of low reliability will bring risks to the safe operation of NPPs. A sufficient level of redundancy and diversity of I&C design to ensure the safety is a major focus when designing a new reactor. Usually multiple signal paths are included in an I&C system design. Meanwhile, besides the protection and safety monitoring system (PMS), other sub-systems of I&C such as the diverse actuation system (DAS) will be included as a diverse backup of PMS to perform the functions of reactor trip and engineered safety features actuation systems (ESFAS). However, the construction costs increase as the level of system redundancy and diversity grows. In fact, from the perspective of deterministic theory, an I&C system of only two chains can meet the single failure criterion. So how to obtain the balance of safety and economy is a challenging problem in I&C system designing. Probabilistic Safety Assessment (PSA) is the most commonly used quantitative risk assessment tool for decision-making in selecting the optimal design among alternative options. In this paper, PSA technique was used to identify whether the I&C system design offers adequate redundancy, diversity, and independence with sufficient defense-in-depth and safety margins in the design of a new reactor. Firstly, detailed risk assessment criteria for I&C design were studied and identified in accordance with nuclear regulations. Secondly, different designs were appropriately modeled, and the risk insights were provided, showing the balance of safety and economy of each design. Furthermore, potential design improvements were evaluated in terms of the current risk assessment criterion. In the end, the optimal design was determined, and uncertainty analyses were performed. The results showed that all four designs analyzed in this paper were met the safety goals in terms of PSA, but each design had a different impact on the balance of risk. As the support systems of the NPP we analyzed were relatively weak, loss of off-site power and loss of service water were two main risk contributors. The common cause failure of reactor trip breakers and the sensors of containment pressure were risk-significant. After identifying the major risk factors, the I&C design team can perform subsequent optimizations in the further design based on the PSA results and achieve an optimal balance between safety and economy.
The Numerical Simulation of the Detachment Characteristic of Bubble in Narrow Channels With Different Width
Authors: Jie Yang , Wang Jianjun , Gao Puzhen , Ying Huang
Published: October 2017

The plate-type fuel element is widely used in ship-based nuclear reactors. The typical coolant channel in such reactor is a narrow rectangular one with a relatively large aspect ratio. The thermal-hydraulic characteristics in narrow rectangular channels may be different from that in conventional pipe. It could be expected that the geometry of narrow channels has some impact on the bubble behaviors in subcooled flow boiling.

In this paper, the bubble behaviors were simulated by using CFD software FLUENT in narrow rectangular channel with different widths under both stagnation conditions and flow conditions. The main concern was placed on the parameters of bubble departure time, departure diameter and departure velocity.

In the model setup process, the VOF model and PISO algorithm were selected. The results were compared against the experimental data and showed the models good capability of capturing the interface between gas and liquid. Furthermore, for stagnation cases, the width of the channel played an important role for bubble departure diameter. In flow cases, the width also had a great impact on departure diameter and departure velocity as well. The departure time and departure velocity showed strong relevance with each other under different contact angle conditions.
Hydraulic Investigations on Free Surface Flow of ADS Windowless Target
Authors: Hu Chen
Published: October 2017

The formation and control of free surface flow is one of the most essential parts in the studies of windowless target in acceleration driven sub-critical system (ADS). Water model experiments and 360° full scale three dimensional simulations are conducted in this page. The free surface motion is significantly affected by the inlet flow velocity and outlet pressure. The length of free surface decrease in the second order with the inlet flow velocity, while it decreases linearly with the outlet pressure. The structure and feature of flow field are investigated. The results show that the free surface is vulnerable to the vortex movement. Transient simulations are performed with volume of fluid (VOF) method, large eddy simulation (LES) and the pressure implicit with splitting of operators (PISO) algorithm. The simulation results agree qualitatively well with the experimental data related to both free surface flow and flow field. These simulation models and methods are proved to be practicable in the hydraulic simulations of liquid heavy metal target.
Microstructure and Properties Analysis of Inner Cladding Tube and End Plug Girth Weld of Annular Fuel Element
Authors: Pengbo Ji , Hanghang Lu , Zhaosong Wang , Zhoufeng Zhang , Bin Liu
Published: October 2017

Annular fuel element is a kind of new double-sided fuel element, and the welding between the inner cladding tube and end plug of it belongs to a new kind of welding. Actually the conception was first proposed by Massachusetts Institute of Technology and the China Institute of Atomic Energy of Reactor Engineering makes the physical design of annular fuel element product. Through this research, we now basically master the method of welding annular fuel element. In this research, we designed a girth-welding fixture for welding the inner cladding tube and end plug of annular fuel element. The influences of the power input on the weld penetration and morphology have been obtained. The metallurgical performance of welded joints was analyzed through optical microscope, hardness testing and scanning electron microscope (SEM). The mechanical tests results indicate that the tensile properties of the welded joints are closely related to the microstructure. And the welding joints are also tested in the autoclave. The research shows that the micro hardness along the longitudinal section of the inner cladding tube appears to be the trend: firstly gradually reduced, and then stayed. The highest hardness is in the welding zone. And the heat-affected zone has great impacts on the micro hardness: the far the area is away from the welding line, the lower the micro hardness becomes. The grains in both of the weld zone and heat-affected zone are obviously grown up, but the grain growth is more obvious in the weld zone. The tensile fracture of the welding joint all occurs in the welding zone, the tensile strength is larger than that of the bar, which is used for processing the end plug. And the maximum force belongs to the shear stress fracture. In microstructure picture, the weld fracture appears to be dimple-shaped. And some of the dimples showed equiaxial and some of the dimples showed the elongated dimples. And the surface-welding zone is coated with the uniform and compact black oxide film without the white and brown corrosion. All the results and study in the paper will be of guidance for the further processing of the annular fuel element used in the pressurized water reactor (PWR).
Fundamental R&D Work on Methods for State Monitoring of Transport and Storage Containers for Spent Fuel and Heat-Generating High-Level Radioactive Waste on Prolonged Intermediate Storage
Authors: Fiß Daniel , Schmidt Sebastian , Reinicke Sebastian , Kratzsch Alexander
Published: October 2017

The continuing search for a long-term storage for highly-active nuclear waste in Germany requires a prolonged intermediate storage period of spent fuel in dry storage casks at the power plant sites. Currently, it is not sufficiently clear if there might be a loss of integrity of the fuel rods within such long periods, e.g. due to rising pressure from gaseous products of nuclear decay. Regarding a final evaluation, extrapolative modelling of the radiochemical and thermomechanical material behavior is challenging and not suitable for predictions on the condition of storage container inventory after the intermediate storage period. Therefore, it is of public interest to find measurement principles or methods which can provide information about the condition of the storage container inventory.

In line with a cooperative project (project partners: Technical University Dresden, Zittau/Görlitz University of Applied Sciences) different measurement principles and methods (radiation emission, muon transmission, thermography, acoustical spectrometry) for non-invasive condition monitoring of the storage container inventory in case of prolonged intermediate storage are going to be investigated and evaluated. The results shall help to determine suitable methods for the identification of both changes of the spent fuel and inner container structure over long periods without opening the container and would be a significant contribution for the long-term safety of intermediately stored highly radioactive waste. Furthermore, suitable methods would provide information about the transport and conditioning ability of the waste before transfer to the repository.

This paper deals with the content of the subproject of Zittau/Görlitz University of Applied Sciences as well as with the approach for project realization. A further main part of this paper is the development of experimental infrastructure to support the investigations.
Study on the Application of CDAG Methodology in EPR Nuclear Power Plant
Authors: Yuhao Yin , Yichao Huang , Feng Zhao
Published: October 2017

The Westinghouse Owners Group Core Damage Assessment Guidance (CDAG), which has been authorized by the NRC staffs, is now used by licensee emergency response organization staff for estimating the extent of core damage that may have occurred during an accident at a Westinghouse nuclear power plant. On the other hand, EPR is a 3rd generation nuclear power plant, which applies the advanced European nuclear power technology.

This paper introduced Core Damage Assessment Guidance methodology in detail. The CDAG methodology is then attempted to apply to the EPR nuclear power plant. Detailed calculations have been performed for the setpoints of containment radiation monitors (CRM) and core exit thermocouples (CETs) with EPR design characteristics, which are the two main methods for estimation core damage amount. This paper also focuses the discussion on the reasons of difference of core damage estimating results between CRM method and CETs method; based on the discussion, several advices are provided when the two methods show a reasonable discrepancy in conclusions.

Several conclusions can be made from the discussions in this article that 1)the Westinghouse Owners Group CDAG methodology proved to be reasonable when applied to EPR power plant for core damage assessment under severe accident; 2) the CDAG methodology which reflect the latest understanding of fission product behavior, is very simple and timely for core damage assessment based on NPP (nuclear power plant) real-time parameters; 3) conservative calculation results of setpoints on CRM and CETs based on EPR design show a reasonable trend and range; 4) it is concluded that several factors such as the releasing way, RCS fission product retention, fuel burnups might have great impact on the estimating results, when the results from two main indications (CRM and CETs) show an unexpected response.