bluuebunny/crossref_metadata_2025 huggingface dataset:
Towards Optimal In-Core Fuel Management of Thorium-Plutonium-Fuelled PWR Cores
Authors: Zainuddin Nurjuanis Z. , Lindley Benjamin A. , Parks Geoffrey T.
Published: February 2014

Plutonium is a significant proliferation concern as well as a major contributor to the long-term toxicity of nuclear waste. Partial incineration in PWRs with uranium-MOX fuel is often considered to mitigate these concerns. Thorium-MOX is an alternative fuel with superior material properties and higher plutonium destruction rates, as shown in multiple feasibility studies. However, the core performance and operational characteristics (e.g. discharge burn-up, feasibility of controlling the core) are ultimately dependent on the core loading pattern (LP) and burnable poison (BP) design. In this paper, the LP for Th-Pu fuel of various compositions is optimized for (1) discharge burn-up, (2) radial form factor (RFF), (3) cycle length, (4) moderator temperature coefficient (MTC), and (5) reactivity swing over cycle. Maximizing the cycle length makes the discharge burn-up and reactivity swing worse due to placement of once- and twice-burnt fuel near the core periphery. It also makes the MTC less negative. The harder neutron spectrum of Th-Pu fuel compared to conventional U fuel favours the use of distributed integral burnable poisons to control the reactivity swing over the cycle. This leads to a significant amount of dissimilarity between LPs with relatively similar performance measures, and between optimal LPs for different Pu loadings in the fuel. The RFF can vary throughout the cycle but a careful placement of the assemblies can mitigate this. The cycle reactivity swing is controlled using enriched soluble boron, which makes the MTC worse, and this constrains feasibility for high Pu loading in the fuel.
Preliminary Study on Human Resources Guarantee in China’s Nuclear Safety Supervision
Authors: Zhang Qiong , Chen Jinrong , Jin Huixin
Published: February 2014

This paper illustrated the necessity of nuclear power development and the importance of nuclear safety supervision. Firstly, this paper briefly introduced the current situation of nuclear power development worldwide, nuclear safety supervision authorities in the developed countries and the situation of the human resources guarantee. Secondly, the paper highlighted the current situation of nuclear safety supervision in China, existing problems and what required to be urgently solved. Finally, this paper layout emphasis on the people-oriented idea and the necessity of improving nuclear safety supervision, and put forward several suggestions on the human resources guarantee in the nuclear safety supervision.
Sensitivity Analysis and Design Improvement of Contra-Motion Check Valve
Authors: Yang Zhi-da , Chen Bing-de , Huang Wei , Han Wei-shi
Published: February 2014

This paper presents the design and development of a Contra-Motion check valve for use in pipework of size in the order of 270mm inside diameter. The paper describes the operation of the valve, the fundamental motive force being generated by creating a pressure difference across a piston head, the head being linked by a sensing tube to downstream pressure. This type of valve can provide benefits over the common types of check valves such as: swing, double disc and lift, in that damping is intrinsic to its operation thereby reducing impact on valve seats and eliminating water hammer effects that can be problematic with the other types. Valve characteristics are presented which were generated using Flow Field analysis techniques and also rig testing covering steady state flow conditions. The key factors that affect performance are covered such as: the streamline shape of the structure projecting into the flow, the area difference between the head and piston, and fluid leakage between clearances of parts.
Study on Medical Emergency Rescue for Nuclear Accident on the Sea
Authors: Wang Chuan , Zhang Jian-guo
Published: February 2014

As a floating nuclear plant on the sea, the nuclear submarine accidents on the sea is much more possible with the service life increasing year after year because of marine mission and harsh environment. In case of nuclear accident, it has a serious result and is difficult comparatively to medical rescue and guarantee. This paper introduces the sorts and basic principles of medical rescue for nuclear accident, especially analyses the characteristic of medical emergency rescue within the submarine, scene disposition points and radiation protection measures of emergency work personnel when an accident occurred on the sea. All that have been discussed has guide meaning and practical significance to medical emergency rescue for nuclear accident on the sea in our country in future.
Experimental Research of Flow Instability Onset Conditions in a Natural Circulation System With Subcooled Boiling
Authors: Pan Liang-ming , Li Jia , Chen De-qi
Published: February 2014

With water as the working medium, this paper has investigated into the factors which affect the onset points of flow instability in the natural circulation system with subcooled boiling. Study finds that the increase of entrance subcoolling and pressure in the system can improve system stability. Increasing threshold heat flux can make the subcooling boundary area of density wave oscillation bigger. Both entrance subcooling and system pressure have significant influence on the position where ONB appears at stability boundary. According to dimensionless analysis of two-phase flow differential equations, this paper has put forward an semi-empirical formulae that can forecast the onset conditions of density wave oscillation.
Study of Graphite Dust Releasing Behavior in a Depressurization Accident of HTR
Authors: Peng Wei , Zhen Ya-nan , Yang Xiao-yong , Yu Su-yuan
Published: February 2014

The behavior of the graphite dust has an important effect on the safety analysis of High Temperature Gas-cooled Reactors. The graphite dust with the large size will deposit at the bottom of the reactor core by gravity, while the graphite dust with the small size will flow in the primary circuit by carrying of helium. These suspended dusts can deposit in primary loop surface and influence the surface’s feature such as fouling resistance. Moreover, fission products released by fuel elements would enter primary loop and combine with dust, resulting in making the maintenance and repair of steam generator difficult. On the other hand, in the loss-of-coolant accident, when the discharge pipe of fuel pellets was broken, helium will be rejected with the speed of sound. The graphite dust with radioactive contamination will be carried by helium into the environment. The present study experimentally investigates the resuspension of graphite dust in the process of depressurization accident. The effect of both the initial pressure and the velocity of the gas on the resuspension of graphite dust were studied.
The Criticality Safety Analysis of AP1000 Fuel Assembly Transport Package
Authors: Wang Xuexin , Zhuang Dajie , Sun Hongchao , Li Guoqiang , Xu Xiaoxiao , Zhang Qun , Cao Fangfang
Published: February 2014

Criticality safety needs to be considered carefully during the transport of nuclear fuel assemblies. Several factors should be taken in to account, such as package arrays, container damages under hypothetical accident condition, water moderation conditions and so on. In this work, criticality safety analysis has been carried out for a Westinghouse XL shipping container loaded with AP1000 fuel assemblies. The fuel assembly and transport container have been modeled and simulated using the MCNP code. The results of MCNP show that the Westinghouse XL package is subcritical, if the number of packages meets the requirements in the approval of the competent authority.
The Influences of Significant Nuclear Events on Public Attitude and Acceptance
Authors: Liu Mingye
Published: February 2014

In this paper, the influence of significant nuclear events on public attitude, and the reaction, i.e., public attitude on the development of nuclear industry have been studied. Though it has been widely known that nuclear is a type of clean, green and effective source of energy, due to the characteristics of nuclear, for most common people, they seldom pay attention to the nuclear power plants (NPPs) or the entire nuclear industry except NPPs are built near the residence of their own, or when some catastrophic failures occur. This fact leads to that when the public’s attention is attracted to nuclear, the effects on their attitude to nuclear are often negative. Even if there is positive news about nuclear, in most cases, the public will still be worried and prudent. That is one of the reasons why though the nuclear power related techniques has been developing rapidly, in some countries the usage of nuclear power is still quite limited.

In order to carry the development of nuclear power forward, to improve the public acceptance is as important as to improve nuclear related science and technology. This paper focuses on how the significant events related to nuclear influence the public acceptance, which will have direct or indirect effects on the development and/or policy of nuclear industry in a country, even the whole world. Additionally, this paper discusses possible and proper solutions to improve the public acceptance to NPPs and nuclear related techniques.
Very Slow Pebble Flow: A New Research Field
Authors: Jiang Shengyao , Yang Xingtuan , Gui Nan , Tu Jiyuan
Published: February 2014

As well known, the very slow pebble flow (VSPF) exists in the pebble bed high temperature gas cooled reactor, which is regarded as the 4th generation of nuclear power plant, one of the most promising reactors in the 21th century. Unfortunately, we still know very little on the fundamentals of the VSPF, a specific type of granular flows.

In general, a granular flow is a collection of a large number of discrete but closely packed solid particles. A granular flow may behave as either an elastic solid or a fluid, depending on the local stress conditions. A rapid granular flow behaves much more like a fluid rather than a solid collection, and sometimes can even be analogically analyzed by the kinetic theory. However, the very slow granular flow (VSGF) is showing its elastic solid behavior at any instantaneous time and fluid-like behavior in a long time. It can shear like fluid, dissipate energy by collision, and also it can support loads and form finite piling slopes like solid. It is more complicated than the rapid flow and cannot be easily analyzed solely by the fluid or kinetic theories.

We regard the VSPF as a special VSGF. In the pebble bed under gravity, the pebbles flow out of the bed very slowly, almost one-by-one. Thus, the flow-out process is an intermittent or discrete process in time. We can regard each flow-out process as an “original disturbance” (OD), and the transfer process of OD throughout the bed as a “subsequent disturbance” (SD). The SD is decaying in time as the pebble collision is dissipative. Thus, the VSPF can be specifically defined as a special case in which the OD and SD processes are not overlapped. In other words, the effect of SD has already been dissipated almost completely before the next OD occurs.

In this study, we just show some examples for the interesting phenomenon of VSPF both experimentally and numerically. Based on the OD and SD theory, we show the typical processes and basic characteristics of OD and SD to validate their existences and justify the definitions of VSPF by the OD and SD.
ITER Shield Block Cooling System Design and Analysis
Authors: Kang Weishan , Yuan Tao , Zhang Xiujie
Published: February 2014

A steady-state thermo-hydraulic analysis of SB was done to get pressure drop and heat transfer coefficient in the cooling pipes, then a transient thermal analysis was performed to get temperature field of the SB with time in Inductive plasma operational scenario. Finally static structural analysis of the SB based on the temperature fields obtained with the transient analyses at specific time points were done to get the stress and strain distribution of the SB. The numerical analysis results were investigated by rules from ITER System Design Criteria for In-vessel Component to further verify the design. The final results of analyses indicate that the design of the SB meets with the ITER design requirements.
Analysis of Temporary Cooling Strategy for CPR1000 Power Plant Under Situation Similar to Fukushima Severe Accident
Authors: He Dongyu , Chen Xing , Lin Jiming
Published: February 2014

The Fukushima Daiichi nuclear accident was a series of equipment failures, nuclear meltdowns, and releases of radioactive materials at the Fukushima I Nuclear Power Plant, following the Tōhoku earthquake and tsunami on 11 March 2011. According to the cause and process of Fukushima severe accident, several possible scenarios for CPR1000 nuclear power plant were analyzed in this paper, under the assumed situation similar to Fukushima severe accident, with the severe accident analysis code MAAP4. According to the analysis results, several weak points of CPR1000 power plant under situation similar to Fukushima accident were found. The electric power and cooling ability for CPR1000 power plant appear to be most important factors under such accident. Then, several temporary cooling strategies for CPR1000 power plant were suggested, including ASG water supply strategy, temporary injection strategy for primary loop, temporary injection strategy for secondary loop, which would improve the safety of CPR1000 power plant under the situation similar to Fukushima accident. At the last, assessments of effectiveness for these strategies were performed, and the results were compared with analysis without these strategies. The comparisons showed that correctly actions of these strategies would effectively prevent the accident process of CPR1000 power plant under situation similar to Fukushima accident.
Several Constraint Methods for Avoiding Thermal Stress Distortion on Structure Simulating Calculation
Authors: Dai Shoutong , Yin Hao
Published: February 2014

When the conventional thermal stress analysis is made using simulation software, the thermal expansion is regarded to be completely limited on the fixed place. The zero displacement constraint is so applied. But in fact the structure can still expand in some degree, the boundary conditions of the simulation model are so strict seriously that the calculated stress will be far greater than the actual values, which is called “thermal stress distortion” in this paper.

In order to reduce the conservative degree some of the structure constraints should be set free according to the actual structure to make the structure extend appropriately. At the same time the integrity of the applied constraint should be noticed specially. The rigid motion shouldn’t be expected to exist because of the inadequate constraints.

The paper presents several constraint methods for avoiding the thermal stress distortion according to the typical structure of nuclear engineering: “extension constraint”, “column coordinates constraint” and “segmentation and alternate constraint”.

The “extension constraint” method is suitable for the structure simulation which distortion locates near the constraint place. But it can only eliminates parts of the distortion. The “column coordinates constraint” and “segmentation and alternate constraint” methods both can eliminate all of the distortion. The former is only applicable for the rotational structure, while the latter is suitable for frame and rotation structure, what is important is that the latter also provides an idea for structure design to eliminate the thermal stress distortion: setting expansion gap.
Role of Irradiation Embrittlement in RPV Lifetime Assessment
Authors: Brumovský Milan , Kytka Miloš
Published: February 2014

Integrity and lifetime of reactor pressure vessels are practically determined by their material resistance against fast/non-ductile failure and consequently by their radiation damage resulting in irradiation embrittlement and hardening. Mechanism of irradiation embrittlement of RPV materials depends on selected materials and operation conditions but their values depend in a large extent on reactor design, i.e. on neutron flux/fluence depending on RPV wall. Generally, new RPV design allows smaller neutron fluences but absolute value of irradiation embrittlement still depends on a choice of RPV material. Even though radiation damage (especially irradiation embrittlement) is important for RPV behavior, integrity and lifetime depends, in principle, on final value of applied fracture mechanics parameter — transition temperature. Thus, its initial value as well as its shift due to irradiation embrittlement is of interest but only the embrittlement can be affected during operation.

The paper discusses absolute and relative effect of both these parameters for integrity and lifetime assessment for different type of reactors — of old and new as well as the newest design. Then, role of surveillance specimen programs for the evaluation of radiation embrittlement of RPV materials is discussed, first of all from point of view of existing codes and standards. Additionally, some recommendations are given for decreasing the effect of irradiation embrittlement on final integrity and lifetime evaluation.
The Failure Analysis and Processing of Digital Reactor Protection System
Authors: Zhao Jun , Zhou Xing , Hu Jin , Yu Yanling
Published: February 2014

The Qinshan Nuclear Power Plant phase 1 unit (QNPP-1) has a power rating of 320 MWe generated by a pressurized water reactor that was designed and constructed by China National Nuclear Corporation (CNNC). The TELEPERM XS I&C system (TXS) is to be implemented to transform analog reactor protection system (RPS) in QNPP-1.

The paper mainly describes the function, structure and characteristic of RPS in QNPP-1. It focuses on the outstanding features of digital I&C, such as strong online self-test capability, the degradation of the voting logic processing, interface improvements and CPU security.

There are some typical failures during the operation of reactor protection system in QNPP-1. The way to analyze and process the failures is different from analog I&C. The paper summarizes typical failures of the digital RPS in the following types: CPU failure, communication failure, power failure, Input and output (IO) failure. It discusses the cause, risk and mainly processing points of typical failure, especially CPU and communication failures of the digital RPS. It is helpful for the maintenance of the system.

The paper covers measures to improve the reliability of related components which has been put forward effective in Digital reactor protection system in QNPP-1.

It will be valuable in nuclear community to improve the reliability of important components of nuclear power plants.
Research on Fault Diagnosis Approach Based on the Signed Directed Graph Model for Nuclear Power Plant
Authors: Liu Zhen , Liu Yong-kuo , Wu Xiao-tian , Zeng Xian
Published: February 2014

Signed Directed Graph (SDG) is a kind of fault diagnosis method based on a qualitative model. SDG can not only efficiently express the interrelations among variables of complex systems and has strong completeness, flexible reasoning ways and effective reasoning algorithm, but also shows faults propagation paths and faults detailed explanations. In this paper, SDG is applied to fault diagnosis in nuclear power plant which is a large-scale and complex system, and diagnosis results are transformed into a series of expert rules. The case studies show this method can find faults propagation paths and fault source can be determined effectively and improve the efficiency of fault diagnosis. The application of SDG in nuclear power plant can provide operation support to operators, so as to guarantee the safe and stable operation of the system.
Self-Breeding and Radiotoxicity Analysis on Thorium-Based Molten Salt Reactors
Authors: Zou Chun-yan , Chen Jin-gen , Cai Xiang-zhou , Yu Cheng-gang , Jiang Da-zhen , Guo Wei , Zou Yang , Yu Xiao-han , Chen Kun
Published: February 2014

As one of the candidates in the Generation IV reactors program., the molten salt reactor (MSR) has the properties of online refueling and fuel salt reprocessing, MSR is especially attractive for the Thorium fuel cycle, which is very ideal for nuclear non-proliferation, radiotoxicity and nuclear energy sustainability. Therefore, the “Thorium-based Molten Salt Reactor (TMSR) nuclear system” project has been proposed as one of the “Strategic Priority Research Program” of Chinese Academy of Science (CAS). In this paper, we mainly investigated the influence on the breeding ratio and waste radiotoxicity with different reprocessing schemes. By considering the key parameters mentioned above, the aim is to choose an efficient reprocessing scheme for TMSR to reach self-breeding with Th/U fuel cycle and minimize the radioactive waste production of the molten salt.
The Principle and Preliminary Validation of a New Subchannel Code
Authors: Bai Ning , Zhu Yuanbing , Ren Zhihao , He Haibo , Lu Haoliang , Chen Jun , Li Wenhuai , Li Jinggang
Published: February 2014

Following China’s road map of nuclear technology development, the development of self-reliant nuclear design codes becomes one of the most significant steps in the plan. Among the nuclear design codes, thermal-hydraulic analysis code is indispensable because it is the foundation of reactor safety analysis and reactor design. Recently, China Guangdong Nuclear Power Group has launched a series of R&D projects of reactor design code development. The sub-channel analysis code-LINDEN becomes one of the key subprojects. Since the sub-channel code is developed for thermal-hydraulic design and safety analysis of pressurized water reactors (PWRs), the basic requirements for the LINDEN code are reliability and stability. Therefore, the mathematical model and numerical method developed in the code are based on the matured approaches that have been used in various industrial applications. These models and methods includes: four-equation drift framework model of two-phase flow; the classical heat transfer model and fuel rod model (Poisson equation) as well as the constitutive relations; explicit formulation and stepping algorithms for equation solutions. The solver module of the code is programmed using object-oriented C/C++ language with the structural design.. With all these features, the code was developed to be stable, scalable and compatible. The code’s applicability has been further improved after model improvement and design optimization according to characteristics of China’s proprietary type of reactor. COBRA-IV and LINDEN have been used to conduct the thermal-hydraulics analysis for the Daya bay unit 1 and 2 nuclear plants at the steady-state conditions. The results demonstrate that the two codes agree well with each other. The preliminary tests show that the LINDEN code should be suitable for thermal-hydraulics analysis of large PWRs.
Re-Evaluation of Pulsed Neutron Experiment on Fast-Thermal Coupled Sub-Critical Reactor VENUS-1#
Authors: Xie Jinsen , Yu Tao
Published: February 2014

The Pulsed Neutron Source (PNS) experiment on the Fast-thermal Coupled sub-critical core VENUS-1# has been performed in CIAE with 14MeV accelerator neutron source. The preliminary experimental data indicated spatial dependence of prompt neutron attenuation constant α, which is conflict to traditional theories. To re-evaluate the experiment and probe possible special features of the Fast-thermal Coupled core, a Monte-Carlo simulation of PNS experiment has been performed in this paper. Considering Contamination of harmonics mode of prompt neutrons, a data-processing technique, e.g. harmonics filtering, is used to obtain the fundamental mode decay. The simulated results indicate that, the influence of harmonics and external source are mainly in the time region 0∼500μs after pulsed neutron injection; by harmonics filtering, spatially independent α values are obtained and verified. By the method above, the experimental data are re-processed, however, the problem of the spatially dependence of α is not improved at all, the α value indicated by detector R is still quite different from others’. The works performed in this paper suggest that, some future PNS experiments should be deployed to confirm the experimental phenomenon.
Influence of Helium Cluster Structure on Diffusion Behavior
Authors: Chen Min
Published: February 2014

The migration processes of He cluster in Ti have been simulated in detail using molecular dynamics. It is observed that He2 has no preferred molecular orientation in the migration process. He2 is observed to migrate quickly across the Ti atom layer when its molecular orientation is in the direction of perpendicularity to [001] axis. And it is difficult for He2 to migrate across the Ti atom layer when its orientation is along the [001] axis. He2 only induce a slight rearrangement of its nearby Ti atoms as it migration. As He3 significantly deviates from the orientation of (001) plane, it migrates quickly in the direction along [001] axis. As the deviation of the orientation of He3 from the (001) plane is small, He3 tends to embed in a (001) monolayer and dislocation will be induced at this condition. The molecular orientation of He3 tends to lay in the (001) plane for most of the time. He trimer consequently diffuses slower than He atom and dimer do. Results demonstrate that the details of the migration process of Hen are obviously different for different n.
Thermal Hydraulics Analysis on a Simplified Model of Spent Fuel Pool
Authors: Liang Min , Lu Daogang , Xia Huining , Zhang Yuhao
Published: February 2014

The spent fuel still has massive decay heat and is temporarily stored in the spent fuel pool, after unloaded from the reactor core; it is cooled by the circulating water in the spent fuel pool. The present study generally agreed that the spent fuel pool is relatively safe when the cooling water works, even if under accident conditions. It is generally considered that the thermal-hydraulic process in the spent fuel pool is very slow, and does not endanger the safety of nuclear power plant. But the boiling in the pool has been given more and more attention after the Fukushima nuclear accident; especially after the failure of core safety injection and the cooling water in the spent fuel pool is lost.

School of Nuclear Science and Engineering, North China Electric Power University intends to perform overall test bench to simulate the AP1000 spent fuel pool, the experimented data can be used to validate the results of the COSINE software calculations. In this paper, the numerical simulation is performed with CFD software to calculate the temperature, velocity, pressure field of the scaled spent fuel pool experimental bench; it can provide the guideline for the selection of the test section measuring instruments and the arrangement of the measuring points. Also, it’s essential to verify the COSINE software and guide the experimental process.
An Experimental Investigation on Molten Behaviors for a New Type Core Catcher
Authors: Lu Donghua , Xu Haiyan
Published: February 2014

A new core catcher concept was presented and experimentally investigated in the paper. The new core catcher composed by many vertical cells. Parallel steel plats protecting by the ceramic layer form a vertical channel which can accommodate core molten and cool it from both sides. One cell model was constructed and experimentally studied in the paper. The experiment proved the molten can be solidified and cooled in a very short time.
Analysis of RESCUE External Reactor Cooling Experiments Using RELAP5 Mod3.3 and ASTEC V2.0 Codes
Authors: Kubic Martin , Bachraty Milan , Barnak Miroslav , Matejovic Peter , Guenadou David , Verloo Eric
Published: February 2014

Safety margin of the in-vessel retention strategy is given by the difference between thermal load acting on inner reactor surface and coolability limit (in terms of critical heat flux) on outer reactor surface. In order to study the two-phase flow in external reactor vessel cooling loop and heat transfer from curved reactor wall, the RESCUE-2 (Representative loops for External System Cooling Understanding Experiments) experimental facility was erected in CEA Cadarache in France. The facility consists of electrically heated simulator of reactor vessel with an ellipsoidal lower head and cooling loop that enables natural circulation of coolant around the reactor wall. In the frame of SARNET 2 project (Severe Accident Research Network of Excellence, 7th EU Framework Programme) several experiments devoted to external reactor vessel cooling phenomena with certain relevance to VVER-440/V213 reactors, were performed on this facility. The heat flux profile generated by electrical heaters in these experiments was based on the results that were obtained by using ASTEC code for this reactor design. The results of two RESCUE-2 experiments are used in this paper for benchmarking of RELAP 5 Mod.3.3 code and ASTEC V2.0 code.
Validation of TRACE Code Against ROSA/LSTF Test for SBLOCA of Pressure Vessel Upper-Head Small Break
Authors: Xing Mian , Hu Xiao , Chen Yaodong , Li Liangxing , Ma Weimin
Published: February 2014

OECD/NEA ROSA/LSTF project tests are performed on the Large Scale Test Facility (LSTF). LSTF is a full-height, full-pressure and 1/48 volumetrically-scaled two-loop system which aims to simulate Japanese Tsuruga-2 Westinghouse-type 4-loop PWR. ROSA-V Test 6-1 simulates a pressure vessel (PV) upper-head small break loss-of-coolant accident (SBLOCA) with a break size equivalent to 1.9% of the volumetrically scaled cross-sectional area of the reference PWR cold leg.

By building a TRACE calculation model of LSTF and PV upper-head, the paper dedicated to assess the effect of different modeling options and parameters on simulating thermal hydraulic behaviors of TRACE code. The results show that TRACE code well reproduce the physical phenomena involved in this type of SBLOCA scenarios. Almost all the events in the experiment are well predicted by the model based on TRACE code. In addition, the sensitivity of different models and parameters are investigated. For example, the code slightly overestimated the break mass flow from upper head which could affect the accuracy of the results significantly. The rising of core exit temperature (CET) is significantly influenced by the bypass flow area between downcomer and hot leg.
Exploration of China Nuclear Power Standard System Development
Authors: Yan Jinquan , Gu Shenjie , Tian Lin
Published: February 2014

Based on investigation and analysis of the status of code & standard of China’s nuclear industry, as well as the challenges the industry is facing, a series of suggestions are proposed for the system development, including a hierarchical structure of projected code & standard system to accommodate diversity, coherence and coordination, authority building, prospect and internalization and operation mechanism.
TRACE Validation Against LOCA Transients Performed on FIX-II Facility
Authors: Hu Xiao , Xing Mian , Ma Weimin
Published: February 2014

As a latest developed computational code, TRACE is expected to be useful and effective for analyzing the thermal-hydraulic behaviors in design, licensing and safety analysis of nuclear power plant. However, its validity and correctness have to be verified and qualified before its application into industry. Loss-of-coolant accident (LOCA) is a kind of transient thermal hydraulic event which has been emphasized a lot as a most important threat to the safety of the nuclear power plant. The FIX-II experiments were performed to produce experimental data for understanding the initial stage of LOCA and so as to verify the computational codes. In the present study, based on FIX-II LOCA tests, simulation models for the tests of No. 3025, No. 3061 and No. 5052 which correspond to different LOCA cases were developed to validate the TRACE code (version 5.0 patch 2). The predictions of the TRACE code including the pressure in the primary system, the mass flow rate in certain key parts, and the temperature in the core were compared with the experimental data. The results show that TRACE model can well reproduce the transient thermal-hydraulic behaviors under different LOCA situations. In addition, sensitivity analysis are also performed to investigate the influence of particular models and parameters, including counter current flow limitation (CCFL) model and choked flow model on the results, which show that both the models have significant influence on the outcome of the model.
The Impact of the High Temperature on the Performance of Critical Components in the HTR-PM Control Rod Drive Mechanism
Authors: Chen Cheng , Diao Xingzhong
Published: February 2014

Control rod drive mechanism is very important to the operation and safety of reactors, therefore, requiring good mechanical properties and a high degree of reliability. The control rod drive mechanism of high-temperature-reactor pebble-bedmodule (HTR-PM) is powered by permanent magnet stepping motor and damped by the permanent magnetic damper. Therefore, the performance of both motor and damper have important impact on the control rod drive mechanism. However, the performance of permanent magnets is very sensitive to high-temperature. This paper describes the impact of the high temperature on the performance of critical components in the HTR-PM control rod drive mechanism. A Full-size thermal test bench is used to determine the performance of the key components within the scope from room temperature (20 ° C) to 250 ° C, with these data we can calculate the emergency insertion time. And further work plan is introduced too.
Recent Knowledge From an Experimental Investigation on Self-Leveling Behavior of Debris Bed
Authors: Cheng Songbai , Yamano Hidemasa , Suzuki Tohru , Tobita Yoshiharu , Nakamura Yuya , Taketa Syohei , Nishi Sinpei , Zhang Bin , Matsumoto Tatsuya , Morita Koji
Published: February 2014

Studies on the self-leveling behavior of debris bed are crucial in the assessment of core-disruptive accident (CDA) that could occur in sodium-cooled fast reactors (SFR). To clarify the mechanisms of this behavior, several series of experiments were elaborately designed and performed in recent years under the collaboration between Japan Atomic Energy Agency (JAEA) and Kyushu University (Japan). This paper presents the recent knowledge obtained from the newly developed large-scale experiments using gas-injection to simulate coolant boiling. Compared to previous investigations, it can cover a much wider range of gas velocities (presently up to a flow rate of around 300L/min). Based on the experimental data obtained, influence of various experimental parameters, including gas flow rate (∼ 300 L/min), water depth (180 mm and 400mm), bed volume (5L, 7L), particle size (2 ∼ 6 mm), particle density (beads of alumina, zirconia and stainless steel) along with particle shape (spherical and irregularly-shaped) on the leveling was checked and compared. In addition, the status of developing empirical model to predict the self-leveling over current setup was also presented. This work, which gives a large palette of favorable data for a better understanding and an improved estimation of CDAs in SFRs, is expected to benefit future analyses and verifications of computer models developed in advanced fast reactor safety analysis codes.
Research on Nodular Corrosion Resistance of N36 Zirconium Alloy at 500°C
Authors: Chen Liang , Yang Zhongbo , Miao Zhi , Dai Xun , Sun Chao
Published: February 2014

The nodular corrosion tests of improved Zr-4 tubes and two types of N36 tubes were carried out in 500 °C, 10.3Mpa super-heated steam. The results shown corrosion resistance of N36 tubes was much better than Impr.Zr-4 tubes. Characteristics of the second-phase particles and the oxide layer formed on N36 tubes were analyzed by scanning electronic microscope. The volume fraction of t-ZrO2 in oxide was calculated by Laser-Raman spectroscopic analysis. In connection with corrosion kinetics, the effect of SPPs size and volume fraction of t-ZrO2 in oxide related to corrosion resistance was discussed. It was pointed out that nodular corrosion resistance of N36 tubes can be improved by SPPs.
Development of Numerical Simulation for Jet Breakup Behavior in Complicated Structure of BWR Lower Plenum: 2 — Flow Observation With Visualized Experimental Apparatus
Authors: Saito Ryusuke , Abe Yutaka , Kaneko Akiko , Suzuki Takayuki , Yoshida Hiroyuki , Nagase Fumihisa
Published: February 2014

In order to improve the safety of Boiling Water Reactor (BWR), it is required to know the behavior of the plant when an accident occurred as can be seen at Fukushima Daiichi nuclear power plant accident. Especially, it is important to estimate the behavior of molten core jet in the lower part of the containment vessel at severe accident.

In the BWR lower plenum, the flow characteristics of molten core jet are affected by many complicated structures, such as control rod guide tubes (CRGT), instrument guide tubes and core support plate. However, it is difficult to evaluate these effects on molten core jet experimentally. Therefore, we considered that multi-phase computational fluid dynamics approach is the best way to estimate the effects on molten core jet by complicated structure.

The objective of this study is to develop the evaluation method for the flow characteristic of molten core jet including the effects of the complicated structures in the lower plenum. So we are developing a simulation method to estimate the behavior of molten core jet falling down through the core support plate to the lower plenum of the BWR. The method has been developed based on interface tracking method code TPFIT (Two Phase Flow simulation code with Interface Tracking). To verify and validate the applicability of the developed method in detail, it is necessary to obtain the experimental data that can be compared with detailed numerical results by the TPFIT. Thus, in this study, we are carrying out experimental works by use of multi-phase flow visualization technique.

For the experiment works, we constructed two experimental apparatuses, one is single-channel experimental apparatus and the other is multi-channel experimental apparatus. The single-channel experimental apparatus simply simulate single flow channel between four CRGTs. The multi-channel experimental apparatus is a 1/10 planar type model of a lower plenum of BWR. In the experiments, we will obtain the jet breakup behavior and surrounding velocity profiles of the jet by using LIF and PIV.

In this paper, the outline of two-experimental apparatuses are shown. And the results of the single-channel experimental apparatus with/without modelled complicated structures are also shown. In the results, it was confirmed that the complicated flow channel affects the jet injection and breakup behavior. And it was also confirmed that the complicated structures restrain diffusion of fragments of the jet.
Analysis of Flow Stability Boundaries of ERVC System
Authors: Li Fei , Jin Di , Liu Xiao Jing , Cheng Xu
Published: February 2014

ERVC is widely adopted as a part of in-vessel retention (IVR) in severe accident management strategies. In this paper, two-phase flow instability in natural circulation loops of external reactor vessel cooling (ERVC) system in a large size power PWR (CAP 1700 with a thermal power 5000 MWt) is simulated and evaluated by the RELAP5 code. Under certain conditions, flow instability of ERVC system are obtained. It is a kind of density wave oscillation that occurs in non-equilibrium boiling in the heat section and void flashing in the riser at low equilibrium quality and low system pressure. The calculation results show such oscillation course clearly. And several parameters affecting the flow stability are discussed.
ATHLET Simulation of OECD/NEA ROSA-2 Test-2 About Cold Leg Intermediate-Break LOCA
Authors: Yu Nan , Fu Xiaoliang , Du Zheng , Liu Lifang , Cao Zhen , Yang Yanhua
Published: February 2014

Experiment about intermediate-break loss-of coolant accident with 17% break at cold leg was performed in OECD/NEA ROSA-2 project on Large Scale Test Facility (LSFT). Safety injection was assumed single failure and only injected into intact loop. Before the loop seal clearing, the liquid level dropped obviously and the core dryout took place. ATHLET Mod 2.1 Cycle A was used to do the post-test calculations of this test. The major calculated parameters were compared with the test data. The trend of the prediction results fit well with that of the test data, and the cause of the deviations was analyzed.
Factors Promoting Embrittlement of T91 Steel in Liquid Lead-Bismuth Eutectic Alloy
Authors: Ye Changqing , Vogt Jean-Bernard , Proriol-Serre Ingrid
Published: February 2014

The mechanical behavior of the T91 martensitic steel has been studied in liquid lead-bismuth eutectic (LBE) and in inert atmosphere. Several conditions were considered to point out the most sensitive embrittling factor. The Small Punch Test technique has been employed by using smooth specimens.

T91 appeared in general as a ductile material, and became brittle in the considered conditions only at least, but not necessarily, if the test was performed in LBE.

It turns out that the loading rate appeared as a critical parameter for the occurrence of liquid metal embrittlement of T91 in LBE. For standard heat treatment condition, loading the T91 very slowly instead of rapidly in low oxygen LBE resulted in brittle fracture.
Numerical Study of the Gas-Steam Transient Behavior in the Integrated Pressurizer
Authors: Wu Lei , Jia Haijun , Liu Yang
Published: February 2014

The integrated gas-steam pressurizer stabilizes the pressure by compressing the gas and steam mixture. It has attracted much attention because of its simple structure, eliminating heating and spraying of equipment, and preventing the liquid boiling. The NHR series developed by Institute of Nuclear and New Energy Technology in Tsinghua University uses the integrated gas-steam pressurizer. The major loop thermal parameters in NHR series increased progressively, which made it suitable for heating, industrial steam supply and seawater desalinization. In order to ensure the safety of the NHR series major loop system and guarantee the natural circulation capability of the system under high temperature and pressure, the researches on the gas-steam transient characteristics of the integrated gas-steam pressurizer is needed. This paper is mainly about study on transient characteristics of the gas-steam typed pressurizer using the Relap5 code. The classic experiment on the pressure behavior of gas-steam pressurizer during the in-surge performed at MIT is considered as reference objects, and the analysis model is established by using Relap5 code. By comparing the computing results with the MIT experiment data about pressure-time, the applicability of Relap5 code for forecasting the transient behavior of the gas-steam (nitrogen) pressurizer has been verified. The results show that Relap5 code can effectively track the transient behavior of the pressure in the gas-steam pressurizer. In addition, the transient characteristics of the integrated gas-steam pressurizer in the NHR series have been studied. It is founded that the pressure and the liquid temperature adjoining to the pressurizer lag behind the power change in natural circulation loop with integrated gas-steam pressurizer, and the liquid temperature adjoining to the pressurizer and the liquid volume under the pressurizer are the main factors determining the pressure change.
Dimensional Analysis for the Experiment of PM2.5 Thermophoretic Deposition in the Narrow Rectangular Channel
Authors: Lin Daping , Zhou Tao , Yang Xu
Published: February 2014

A 1000mm×20mm×20mm narrow rectangular channel is built and the particles used belong to the PM2.5. Particles flow in the channel and they will deposit on the wall because of the temperature difference between the wall and the flow. The measurements of thermophoretic velocity of the particles are performed by the phase doppler anemometry (PDA). The radial velocity of the particles changes as the temperature of the flow or the mass rate of the flow is changing. Dimensional analysis is used to do with the experiment data, to get the relationship between the data. The difference between the result of the experiment and the fitting formula is less than 30%, so the formula is exact. The conclusion is of great importance for the further experiment and research and provides a new method for the PM2.5 research.
Computational Study on a Single Flow Element in a Nuclear Thermal Rocket
Authors: Zhao Xiang , Montgomery Trent , Zhang Sijun
Published: February 2014

The nuclear thermal rocket is one of the candidate propulsion systems for future space exploration including traveling to Mars and other planets of the solar system. Nuclear thermal propulsion can provide a much higher specific impulse than the best chemical propulsion available today. A basic nuclear propulsion system consists of one or several nuclear reactors that heat hydrogen propellant to high temperatures and then allow the heated hydrogen and its reacting product to flow through a nozzle to produce thrust. This paper presents computational study on a single flow element in a nuclear thermal rocket. The computational results provide both detailed and global thermo-fluid environments of a single flow element for thermal stress estimation and insight for possible occurrence of mid-section corrosion.
Preliminary Development of System Analysis Code for Traveling-Wave Reactor
Authors: Guo Chao , Lu Daogang , Huang Hao
Published: February 2014

Traveling wave reactors (TWR) are high-burnup nuclear reactors designed to operate indefinitely using fertile material such as natural or depleted uranium (DU), light-water reactor (LWR) waste, or thorium as reload fuel. After an initial partial loading of fissile fuel, the TWR will operate for decades without needing reprocessing or further enrichment, until the lifetime of major structural components is reached. For the safety analysis of this kind of reactor, it is necessary to develop a system analysis code. In the present paper, a system code (named as “TRAVELER”) was developed for a conceptually designed traveling wave reactor. In this code, some important models were established in detail, such as core neutron kinetics model, key equipment models (including centrifugal pump, check valve, lower plenum, upper plenum, fuel assembly, to utilize fuel efficiently without uranium enrichment or reprocessing, instead directly using depleted uranium, natural uranium, thorium, spent fuel removed from light water reactors, or some combination of these materials. The name refers to the fact that fission does not occur throughout the entire TWR core, but remains confined to a boundary zone that slowly advances through the core over time. TWRs could theoretically run, bypass flow channel, pipes connecting these major components) and calculation logic under steady state and transient operation condition. The travelling wave reactor system analysis code currently for the thermal-hydraulic analysis of primary heat transport system was finished. It is expected that the present code could be used for the design and safety analysis of travelling wave reactor to be developed in China.
AP1000 Digital Rod Position Indication Detector Malfunction Issues and Analytical Processing
Authors: Chen Feng , He Fanfan
Published: February 2014

AP1000 Digital Rod Position Indication (DRPI) detector is used to locate the control rods’ position in the core to within 6 steps. Without rod position information, the plant is not allowed to remain at power. Constant and accurate rod position is necessary due to the importance of control rods and their effect on reactor power control. This paper briefly introduced the function of the AP1000 DRPI detectors and its significance to the plant, and then described the malfunction issues of DRPI detectors discovered at Sanmen Unit 1 during inspection and installation, and potential risks/effect detailed as well as the RCA (Root Cause Analysis) and Remediation /Corrective Actions. Finally, summed up experiences and feedback, provide a reference for the follow-up units and other nuclear power plants.