bluuebunny/crossref_metadata_2025_split huggingface dataset:
Boring Analysis in Solving Out-of-Tolerance of Wall Thickness of Surge Line in Nuclear Power Project
Authors: Zhu Huadong 
Published: October 2017

The nuclear reactor coolant surge lines are installed in the reactor building to connect the pressurizer and hot section of RCL (Reactor Coolant Loop). After being adjusted through heating or spraying, the pressure in the pressurizer is transmitted to the main loop through the surge lines so as to maintain stable working pressure inside the main loop. The weld of surge line is required to have in-service inspection. According to the requirements for the inner diameter counter bore indicated in surge line groove machining drawings, the design standard states that “if the weld is subject to service inspection, the length of the counter bore shall be 2Tmin for pipe and Tmin for components and fittings. After the surge lines of XXX project arrived on site, it was found that the back wall thickness of internal boring exceeded tolerance by analyzing the minimum wall thickness in the boring processing area of the surge lines. In order to solve the problem of back wall thickness of these borings less than the designed wall thickness, this paper carries out specific analysis and gives solutions. As for the problem of out-of-tolerance of minimum wall thickness in the boring processing area inside the grooves in the middle section of surge lines; it is recommended to modify the diameter of supposed internal boring of the groove to 367.0.00mm to satisfy the requirement of minimum wall thickness. As for the out-of-tolerance of minimum wall thickness in the boring processing area on the inner wall of groove of the connection-line nozzle in both hot section of main pipe of surge lines and in the pressurizer, it is recommended to modify the angle of inner boring of groove of surge lines. Specific solutions include:
Solution 1:
Modify the internal groove form from being perpendicular to the end face as shown in the design drawing to being 0°–4° included angle between the internal boring face and the end face.
Solution 2:
Correct the inner diameter of pipe section S001F to the designed inner diameter, in order to satisfy the requirement of processing inner diameter and minimum wall thickness. Analyze and calculate the wall thickness based on the corrected inner diameter to obtain the data of wall thickness on the equipment interface end of the surge lines after groove preparation.
Solution 1 and Solution 2 were used simultaneously, which is to modify the groove design on the one hand and ask the manufacturer to correct the inner diameter on the other hand so as to meet the design scope. By applying the above-mentioned combined solutions, we have solved the out-of-tolerance of wall thickness of surge lines. At present, the groove processing technology of surge lines has been approved by Administration and the installation has already been completed on the whole, which testifies to been a successful practice.
Enhancement of RPV Critical Heat Flux by Injection Flow
Authors: Chen Wei , Sun Canhui , Geng Jun 
Published: October 2017

Under severe accidents, the reactor pressure vessel is flooded with water and the residual heat is removed by two-phase natural circulation through the flow channel between the reactor vessel and thermal insulation. If the heat flux of the outer wall is lower than local critical heat flux, the residual heat can be removed, and if the heat flux of the outer wall is higher than local critical heat flux, the reactor pressure vessel should be molten. For AP-type reactors, like AP1000 and CAP1400, critical heat flux of the reactor pressure vessel is the heat transfer limits of residual heat under severe accidents. Previous studies indicate that after severe accident a two-layer molten pool can be formed, namely metallic layer and oxide layer. Compared with oxide layer, in metallic layer, the heat flux more easily exceeds the heat transfer limits due to its low thermal resistance.
In this study, an approach was proposed to enhance local critical heat flux. This approach is expected to be used in local area around reactor pressure wall, like metallic layer, to increase the reactor pressure vessel intact probability under severe accidents. In this new approach, injection flow channels are added to the present flow channel by adding simple flow pipes from insulation near 60 to 80 degree where exceeding critical heat flux is most likely to happen. The fluid flow under external reactor vessel cooling (ERVC) condition is divided into two parts: one part is from downward (0 degree) to upward (90 degree) along the curved reactor pressure vessel and the other part is from injection pipe (about 70 degree) to upward. The fluid temperature from injection pipe is lower than that from downward due to residual heat from the reactor pressure wall. And hence, the local critical heat flux is likely to increase because of inject turbulence and low fluid temperature.
An experimental facility is conducted to study the mechanism of injection influence on critical heat flux under normal pressure condition. There are two main loops in this facility: one is main loop while the other is injection loop. The test section is an inclined downward heated rectangular channel with its inclined angle varied from 0 degree to 90 degree. Flow and thermal conditions are listed: in main loop, mass flow velocity ranging from 100kg/m2s to 600kg/m2s with fluid temperature from 90 °C to 105 °C; In injection loop, mass flow velocity ranging from 0 to 600 kg/m2s with fluid temperature from 85 °C to 105 °C. Under the above condition, with and without injection flow, critical heat flux experiments were conducted. It indicates that injection velocity has great effect on critical heat flux, while injection subcooled has little effect. The critical heat flux can be increased by 0.07MW/m2 to 0.33MW/m2 depending on various injection velocities and main loop conditions.
Study on Water Hammer in the Overpressure Discharge Pipe of Nuclear Power Plant
Authors: Xu Qingmu , Cai Kun , Qin Jie , Yuan Junkai , Li Juan 
Published: October 2017

Water hammer phenomenon is a significant pressure wave in pipe system caused by momentum change when the moving fluid is forced to stop or change direction instantaneously. Common causes of water hammer are sudden valve closing at the end of a pipeline system, pump failure, check valve slam etc. The steam transportation pipeline system may also be vulnerable to water hammer when it confronts with the situation where liquid and steam co-exist. Water hammer often occurs when steam condenses into water in a horizontal section of steam piping. Then steam “picks up” water to form a high-velocity “slug” and create extra stress to pipe. When steam is trapped into sub-cooled water, the collapse of vapor cavity can lead to collision of two columns of liquid, resulting in a large rise in pressure which will damage pipes, supporting structures and hydraulic machinery. Nuclear power plant is composed of complex equipments and piping systems, lots of which contain both liquid and steam. Hence, there is a potential threat of occurrence of water hammer to the normal operation of systems. Thus, this phenomenon needs to be well investigated and prevented with some effective methods.
For the purpose of overpressure relief under severe accidents, the spent fuel pool cooling system of CAP1000 series nuclear power plant provides a discharge passage from containment to spent fuel pool. When the containment pressure exceeds the control value, valve is opened to discharge high-temperature and high-pressure steam until the pressure drops to a safety value. During this process, serious water hammer happens, causing pressure rise beyond the design pressure and further leading to damages to pipes and structures. Therefore, water hammer of overpressure discharge pipeline in CAP1000 plant is studied in this work. On the basis of verification of the capabilities of computational code RELAP5/MOD3.3, hydraulic transient of water hammer is simulated under different conditions. It is indicated that after steam discharge stops, residual steam in pipe condenses because of contact with sub-cooled water in spent fuel pool. Subsequently, the rapid backflow and vapor cavity lead to a severe water hammer. The detailed analysis has shown that water temperature of spent fuel pool has a decisive influence on the mechanism of water hammer phenomenon, including collision of liquid column to valve disc and cavity collapse in the horizontal pipe. The collision and separation of liquid column result in relatively lower pressure amplitude.
Modeling the Natural Circulation Influenced by Earthquake
Authors: Zhang Yinxing , Gao Puzhen , Zhou Jian , Chen Hanying , Ji Peng 
Published: October 2017

More and more attentions have been paid to the impact of the earthquake on the reactor thermal hydraulics after the Fukushima accident. In order to figure out the seismic effect on the flow in primary loop, the natural circulation in primary loop of CPR1000 influenced by earthquake is modeled by learning from the study on natural circulation under ocean conditions. Natural circulation ability at earthquake, which includes longitudinal earthquake and transversal earthquake, is dynamically simulated by a program written with MATLAB, respectively. Besides, the safety of the system is discussed based on the calculated results.
The research results show that, once the longitudinal earthquake occurs, in the earlier stage of earthquake, the peak and valley values of the coolant mass flow rate decrease gradually and the amplitude will be a constant as time goes on because of flow fluctuation in seismic condition. On the contrary, as for transversal earthquake, the peak and valley values of the coolant mass flow rate increase gradually in the earlier stage of earthquake. However, the amplitude will be a constant as time goes on, just like that in the longitudinal earthquake.
For both of the two conditions, they have the same characteristics. The constant mentioned above grows up with the increase of the earthquake acceleration peak value and the period; the fluctuation period of mass flow rate is generally the same as the period of earthquake acceleration and the mass flow rate changes steadily; the average mass flow rate decreases slightly with the increase of the earthquake acceleration peak value and the period.
Both of the vibration period and the peak of seismic wave have disturbing effect on reactor coolant system natural circulation. And the increase of the peak of the earthquake acceleration as well as the period of the earthquake acceleration will reduce the natural circulation ability. In the range of research parameters, the system is considered to be safe according to the calculation results of core coolant temperature.
A New Low Enrichment Uranium Core Design of MNSR
Authors: Mengjiao Wang , Yiguo Li , Xiaobo Wu , Dan Peng , Jingyan Hong , Qian Hao , Jinhua Zhang , Jin Lu 
Published: October 2017

The Miniature Neutron Source Reactor (MNSR) is a low-power research reactor, which uses 90% high enriched uranium (HEU) fuel. However, due to the nuclear safety risk, and according to the principle of nuclear non-proliferation, MNSR must be gradually converted from HEU to low enriched uranium (LEU), which means the LEU fuel with U-235 enrichment less than 20% should be used.
The prototype MNSR of China Institute of Atomic Energy has completed the transformation, but other commercial MNSRs have not finished, which is different with the prototype in the application and structure. Therefore, using MCNP code to simulate, calculate and optimization design LEU core has been done in this issue. Firstly, UO2 with U-235 enrichment of 12.5% was selected as the fuel pellet of LEU core, keeping the rest of the core unchanged. The Φ, excess reactivity and the worth of the central control rod are calculated and analyzed. The results show that the commercial MNSR of LEU conversion is feasible. Secondly, in this paper, through changing the fuel elements and the arrangement method, the new low enriched uranium (NLEU) core was designed to improve Φ/P ratio of the core, the proportion of thermal neutrons and the worth of the control rod. UO2 with U-235 enrichment of 19.75% was selected as the fuel pellet of the NLEU, NLEU not only meets the design parameters, but in many parameters, NLEU is better than LEU. The fuel element quantity is reduced by 43%, from original 344 to 196; reducing the amount of U-235 loading; improving the Φ/P ratio and the thermal neutron fraction is increased. The results show that the NLEU optimizes some parameters, simplifies the core structure, saves the construction cost, improves the nuclear safety and is more suitable for the application of MNSR.
Local Single-Phase Heat Transfer Research of Natural Circulation in Narrow Rectangular Channel
Authors: Zhu Zhiqiang , Cao Xiaxin , Yan Changqi , Tian Chunping 
Published: October 2017

In order to explore and analyze the heat transfer characteristics in narrow rectangular channel, experiments on local single-phase heat transfer of natural circulation in a one-side heating narrow rectangular channel have been conducted under vertical and inclined condition. The thermotechnical parameters such as inlet temperature, heat flux and inclination angle varies during the experiments. The width of the flow channel is 40 mm and the narrow gap is 2 mm. It is heated from one side with a homogeneous and constant heat flux and the working medium is deionized water.
Based on the experimental results, under vertical condition, the driving force in the loop goes up and the Reynolds number also increases when the inlet temperature is elevated, which causes an increase in local Nusselt number. When the heat flux rises, the local Nusselt number increases and the heat transfer temperature difference increases. The local Nusselts number is influenced by entrance effect and the entrance region length is computed for laminar and turbulent flow. Under inclined condition, with the inclination angle from −30° to 30°, it is found that when the inclination angle is positive, the local Nusselt number in fully developed region is larger than that under vertical condition and increases with the angle value, even though the Reynolds number decreases by the effect of incline. This phenomenon is explained by giving an analysis of the natural convection, which is characterized by the normal Grashof number, in the direction perpendicular to the heating plat. Moreover, the variation of heat transfer is also interpreted on the basis of field coordination principle. However, when the inclination angle is negative, the heat transfer shows no obvious difference between vertical condition and inclined condition.
Small Modular Pressurized Water Reactors Combined With Conventional Thermal Power Plant
Authors: Jiang Huijing , Bai Ning , Zhan Huanfen , Shen Feng , Gao Bin , Wang Xuee 
Published: October 2017

Small modular reactor is investigated worldwide with the advantages of lower initial investment and short construction period. Generally, the economy of small modular pressurized water reactor (PWR) is not as good as large PWR, so various applications of small PWR are investigated, such as marine reactor, heat supply and sea water desalination. Limited to the parameters of steam generator, the generating efficiency for the pressurized water reactor nuclear power plant is about 33%, while the steam temperature of supercritical fossil power plant can exceed 600°C and generating efficiency is more than 45%. The essence of a hybrid power plant is to use a fossil fuel to superheat wet steam in an outer steam superheating device, after the steam generator to improve the parameters of working fluid. On one hand, the innovative hybrid nuclear power plant which combines nuclear reactor with conventional thermal energy can improve the efficiency of small PWR. On the other hand, this hybrid power plant has lower carbon emission compared with traditional thermal power plant. This paper describes two different coupling schemes of small pressurized water reactor combined with supercritical thermal power plant using steam turbine.
Efficiency of hybrid power plant is influenced by the coupling scheme, steam parameter of the superheating device outlet, the proportion of nuclear energy, efficiency of assemblies and so on. The plant efficiency becomes higher with the improvement of parameter of the superheating device’s outlet steam, and it is higher when the proportion of nuclear energy becomes lower. Take the 660MWt integrated small pressurized water reactor as an example, when the proportion of nuclear energy accounts for 48%, the thermal efficiency of this innovative hybrid power plant is about 43%,while the net efficiency is 41%, that is much higher than the efficiency of traditional pressurized water reactor, improving about 24 percent. As to the carbon emission, it depends on the coal consumption rate of power supply. The coal consumption rate of this hybrid power plant is 158g /kWh, while the consumption rate of thermal hybrid power plant is 280g/kWh, reducing about 44 percent. Also, the fundamental solutions of technical problems for this innovative hybrid power plant are discussed in the paper. Furthermore, several useful outcomes and suggestions for key equipments are put forward, such as the scheme of a superheating device and high-temperature steam turbine, and the possibility of using a lava boiler as the superheating device to improve the steam parameters after the steam generator is analyzed.
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.