bluuebunny/crossref_metadata_2025 huggingface dataset:
Discussion on Application of Accident Sequence Precursor Analysis in China
Authors: Hu Jiang , Zhang Guoxu , Zhang Yushu , Zhang Binbin , Xiao Zhi
Published: October 2017

An accident sequence precursor (ASP) is an initiating event or degraded condition that, when coupled with one or more postulated events, could result in inadequate core cooling or even severe reactor core damage. Through the systematical investigation and evaluation of nuclear power plant (NPP) operating experience, the ASP analysis could provide a comprehensive, risk-informed view of NPP operating experience, a measure for trending core damage risk, a partial validation of the current state of practice in risk assessment, and also a feedback to regulatory activities. This paper firstly gives a brief review on the ASP evaluation process of U.S. and German, and then introduces the current progresses on the establishment of ASP regulatory framework and the development of ASP management platform, which are carrying out by the related nuclear regulation agency in China. Also, this paper gives some insightful discussions on issues such as the ASP application and problems maybe faced.
Innovative Practice of Severe Accidents Measures in CAP1400
Authors: Fang Likai , Liu Xin , Shi Guobao
Published: October 2017

CAP1400 is GenIII passive PWR, which was developed based on Chinese 40 years of experience in nuclear power R&D, construction&operation, as well as introduction and assimilation of AP1000. Severe accidents prevention and mitigation measures were systematically considered during the design and analysis. In order to accommodate high power and further improve the safety of the plant, also considering feedback from Fukushima accident, some innovative measures and design requirements were also applied. Based on the probabilistic&deterministic analysis and engineering judgment, considerable severe accidents scenarios were considered. Both severe accidents initiated at power and shutdown condition were analyzed. Insights were also obtained to decide the challenge to the plant. All known severe accidents phenomena and their treatment were considered in the design. In vessel retention (IVR) was applied as one of the severe accident mitigation measures. To improve the margin of IVR success and verify the heat removal capability through reactor pressure vessel, both design innovative measures and experiments were used. The melt pool behavior and corium pool configuration were also studied by using CFD code and thermodynamic code. Hydrogen risk was mitigated by installation of hydrogen igniters, which were comprised of two serials, and were powered by multiple power sources. To further improve the safety, six extra hydrogen passive recombiners were also added in the containment. Hydrogen risk was analyzed both inside containment and outside containment considering leakage effect. Other severe accident phenomena were also considered by designed or analyzed to show the containment robustness to accommodate it. As one of the Fukushima accident feedback, full scope severe accident management guideline were developed by considering both power condition and shutdown condition, accident management for spent fuel pool was also considered. As the basis of accident management during severe accidents, survivability of equipments and instruments that are necessary in severe accident were assessed and will be further tested and/or analyzed. Such tests will consider severe accident conditions arised from hydrogen combustion.
Multi-Nuclide Separation Using Different Types of Zeolites
Authors: Mimura Hitoshi , Matsukura Minoru , Kurosaki Fumio , Kitagawa Tomoya , Kirishima Akira , Sato Nobuaki
Published: October 2017

Development of selective adsorbents is very important subject for the effective multi-nuclide decontamination related to the severe accident of Fukushima Daiichi Nuclear power Station (Fukushima NPS). In this study, the adsorption properties for nine kinds of zeolites (Zeolite A, Zeolite X, Zeolite Y, Zeolite L, Modified Chabazite, Phillipsite, Erionite, Synthetic Mordenite, Natural Mordenite and Clinoptilolite) are evaluated in the presence of sodium salts, boric acid and seawater. The present study deals with (1) selective adsorption properties for single nuclide ions (Cs+, Sr2+, Eu3+, I−, UO22+, Am3+ and NpO2+), and (2) multi-nuclide adsorption properties of 26 elements (typical elements in Advanced Liquid Processing System (ALPS) in Fukushima NPP-1) for the above zeolites. The distribution coefficient (Kd, ml/g) and uptake (R, %) were estimated by batch method using NaI (Tl) scintillation counter, ICP-AES and AAS.

Zeolites with different crystal structures have the diversity of the adsorption selectivity for various radioactive nuclides. Chabazite, mordenite and clinoptilolite with lantern or tunnel structure were very effective for the adsorption of monovalent Cs+ ions even in real seawater. Zeolite A and X with three-dimensional cage structures were effective for the adsorption of divalent Sr2+ and Co2+ ions under the practical condition (30% diluted seawater). Zeolite L was effective for the adsorption of Eu3+ ions under the practical condition. As for I− adsorption, Ag-zeolites are found to be effective, and the uptake (%) of I− (NaI in pure water) for Ag-zeolites was estimated to be above 98% in pure water. As for actinoid adsorption, the distribution profile, Kdvs pH, had a maximum depending on the hydrolysis pH. Zeolite A, Zeolite L and Zeolite X showed an excellent adsorption property for UO22+, Am3+ and NpO2+, respectively.

Selective adsorption tendencies of different zeolites were evaluated for 26 elements referred to ALPS. Comparing the uptake results for different zeolites, the following tendency of adsorbability was observed. Mordenite had adsorption selectivity for monovalent alkali metal ions of Rb+ and Cs+. Zeolite A and X exhibited relatively high adsorption selectivity for divalent ions of Sr2+ and Co2+. Zeolite L had adsorption selectivity for trivalent lanthanide ions such as Ce3+ and Eu3+. These tendencies were the same as those without boric acid.

Thus, the zeolites with diverse adsorption selectivity are effective for the multi-nuclide decontamination of radioactive contaminated water.
Development and Application of Cobalt Adjuster Rod for 60Co Medical Radioactive Sources Production in CANDU-6 Reactor
Authors: Ding Jie , Dang Yu , Zhu Libing , Tang Chuntao , Li Hui , Qin Huimin
Published: October 2017

60Co radioactive sources are widely used in industrial and medical field. In 2008, Cobalt adjuster rod assemblies were loaded in the core of CANDU-6 reactor in order to produce industrial radioactive sources. Based on successful production and application of 60Co industrial radioactive sources, the program of developing cobalt adjuster rod for 60Co medical radioactive sources production was launched.

Comparing with industrial radioactive sources, medical radioactive sources require higher activity ∼220Ci/g and smaller sources size, which cause challenges to cobalt adjuster rod design and manufacturing. In this paper, several types of cobalt pellets have been developed to match the medical radioactive sources size and the corresponding nuclear design, TH design & safety analysis, shielding design & safety analysis was also completed to prove that the design requirements could be satisfied.

By now the development of cobalt adjuster rod for 60Co medical radioactive sources production has finished and the first batch of cobalt adjuster rod assemblies have been successfully loaded in core. The first batch of 60Co medical radioactive sources is expected to be put into application in 2019.
Modeling and Simulating Supercritical CO2 Brayton Cycle in SMR Using Modelica
Authors: Yang Yinglin , Guo Qixun , Lin Jianshu , Zhang Yaoli , Ye Kai , Bian BoShen , Li Zhuocheng
Published: October 2017

This paper describes the ongoing work on modeling and simulation of energy conversion processes. The efficiency of the supercritical carbon dioxide (SCO2) power cycles are higher than steam Rankine cycle and helium Brayton cycle in the mild turbine inlet temperature region, and the compact structure and the less restraint of the environment make the SCO2 a promising alternative power conversion system for the next-generation nuclear reactors. In this paper, a SCO2 Brayton cycle steady state simulation program in a small modular reactor is developed. The model is implemented in Modelica language and simulated in Openmodelica environment. The studied process is a high-efficient heat transfer system working in the SCO2 Brayton cycle (SBC). And it is used to analyze the performance of recuperated and recompression cycle configurations. The mathematical models for cooler, heat exchanger are formulated by using the mass and energy conservation equations. The CO2 fluid properties adopted the Reference Fluid Thermodynamic and Transport Properties Database distributed by NIST. Using the SCO2 Brayton cycle model, the design-point operation parameter of the SCO2 Brayton cycle was analyzed, including the outlet pressure and the pressure ratio of compression, the pressure drop and temperature of the heat source. In order to study the 10MW SCO2 Brayton cycle, the equipment selection, numerical modeling has been proceeded under the water-cooled and air-cooled condition. The results show that both the two loops are able to achieve the highest thermal efficiency by adjusting the operation. The modeling language Modelica and the CO2-library is given and the modeling of CO2-recuperator is presented.
Improvement of the Total Beta Monitoring Channel of the Radioactive Gaseous Effluent in HTR-10
Authors: Wei Liqiang , Xie Feng , Zheng Jian , Liu Ling , Hu Chuangguo
Published: October 2017

The radiation monitoring for the gaseous and liquid effluents from a nuclear facility has received more and more attention in recent years since it can indicate the discharge characteristic of the radioactive substance which is used for the environmental impact assessment. In the 10MW high temperature gas-cooled reactor (HTR-10), the stationary differential ionization chamber detector and single chip processor controlled circuit board were adopted to monitor the total β activity of the gaseous effluent from the stack. In the current paper, the design principle, structure composition, function declaration and technical characteristic of the total β monitoring channel of the radioactive gaseous effluent in HTR-10 are presented. The problems in operation of this installation are analyzed and discussed in detail. The digitized improvement of the analog processing circuit is implemented. The stability and reliability of the total β monitoring channel of the radioactive gaseous effluent in HTR-10 are enhanced after the improvement and recalibration.
Research and Engineering Application of Repaired Fuel Assembly Reusing Technique
Authors: Wang Lihua , Yang Qingxiang , Yang Ping , Liu Jiazheng , Zhu Libing , Yang Bo , Cao Hong , Wang Zhe , Ding Yang , Ding Jie
Published: October 2017

Due to debris in the coolant against clad, fuel clad wear, fuel handling fault and so on, fuel rods maybe be damaged during the operation of nuclear power plants, in order that the fuel assemblies with damaged fuel rods are discharged before scheduled. If the damaged fuel assemblies are not reloaded into the core of the nuclear power plant, the fuel utilization decreases and the economy of the nuclear power plant is partly lost. For retrieving the loss of the economy, the damaged fuel assemblies can be repaired by replacing damaged fuel rods with dummy rods which don’t include fissile nuclides. Then, the repaired fuel assemblies can be reloaded into the core. As the repaired fuel assemblies are different with the normal fuel assemblies, especially the number of the damaged fuel rods is considerable, a whole quantitative analysis is very necessary to evaluate the effects from the reuse of the repaired fuel assemblies. In this paper, a full scope evaluation of reload design are performed including nuclear design, fuel design, thermal hydraulic design and safety evaluation, and some necessary improvements are done for the software system, design methods and progress which have been used in the normal reload design. As results, an integrated evaluation technique is developed to evaluate the feasibility and safety of reusing the repaired fuel assemblies, and the key effects due to the reuse of the repaired fuel assemblies are extracted, and the different effects are studied for the different materials of the dummy rods which can be used to conduct how to choose the proper material of dummy rods. In addition, this technique has been successfully applied in the engineering and the loss of economy due to the damage of fuel assemblies was retrieved partly. Therefore, the integrated evaluation technique has also important directive to other nuclear power plants if the repaired fuel assemblies are planned to reuse.
Numerical Simulation Research on the Steam Flow and Heat Transfer in Multi-Inlet Condensers
Authors: Liu Zhao-heng , Xue Ruo-jun , Qin Zhi-bo , Sun Ji-lin
Published: October 2017

The multi-inlet condenser has two steam inlets, the main steam inlet of main steam turbine and the steam inlet of auxiliary steam turbine. In this paper, it establishes a two-dimensional model for the multi-inlet condenser. The porous medium concept is used in the simulations. A porosity factor is incorporated into the governing equations to account for the flow volume reduction due to the tube bundles, baffles and other internal obstacles. The mathematical model is based on the fundamental governing conservation equations of mass and momentum, and the air mass fraction conservation equation. Then, the equations are solved by the SIMPLEC algorithm. Research results indicates that : 1) The physical fields, including pressure, velocity, thermal resistances and condensation rate, are obtained by numerical simulation of the multi-inlet condenser. 2) The performance of multi-inlet condenser is better than the performance of single-inlet condenser from the point of view of pressure drop. 3) There is an optimum angle of auxiliary steam inlet to make the minimum pressure drop. The optimum angle of the multi-inlet condenser used in this study is 50.65°.
Lattice Boltzmann Simulation of Jet Breakup and Droplet Formation in Immiscible Liquid-Liquid System
Authors: Saito Shimpei , Abe Yutaka , Kaneko Akiko , Iwasawa Yuzuru , Koyama Kazuya
Published: October 2017

It is essential to understand the fundamental processes between melt jet and coolant during a postulated core-disruptive accident of a sodium-cooled fast reactor. In the present study, jet breakup and droplet formation in immiscible liquid-liquid system were studied numerically. A lattice Boltzmann two-phase model was modified in framework of three-dimensional 27-lattice to enhance the numerical stability. This model was applied to the conditions of jet breakup simulations. The present lattice Boltzmann simulations qualitatively reproduced the characteristic transitions of breakup regimes.
Some Consideration on the Edge Condition of a Finite-Size Horizontal Downward-Facing Plate in Subcooled Pool Film Boiling
Authors: Toyoda Kaoru , Murakami Shintaro
Published: October 2017

Reliability and usability of the analytical results regarding vapor film thickness, vapor velocity, and the mass flow rate of vapor at the plate edge and Nusselt number averaged over the surface are examined to clarify vapor outflow conditions at the plate edge which has not been clarified by conventional experiments. For analysis, we adopt the two-phase boundary layer theory and the integral method which can determine the variation of vapor film thickness in the vapor flow direction to set a velocity profile in a vapor film and a temperature profile in a liquid boundary layer reasonably. In the calculation of the analysis, the inclination angle of the vapor-liquid interface at the plate edge is varied ranging from 0 to 90 degrees. The results of the analysis provide useful guides to future studies in observing the vapor flow conditions in detail and measuring the vapor flow velocity at the plate edge.
Methodology to Construct Burnup Profile of Fuel Pebbles in PBR
Authors: Tang Yushi , Zhang Liguo , Guo Qiuju , Cao Jianzhu , Tong Jiejuan
Published: October 2017

During the normal operation of a pebble bed gas-cooled reactor (PBR), the fuel pebbles undergo a multi-circulation on the basis of online burnup assay. In our last ICONE paper, we proposed a model to describe the relationship between online burnup assay and economy and safety of PBR. It was concluded that improvements on burnup assay accuracy could reduce fuel cost as well as the possibility that excessive burnup of fuel pebble results in unexpected radioactive discharge. However further work was expected on the burnup distribution of pebbles in and out of the core to precisely quantify the relationship. In this paper, the methodology to construct the burnup distribution of fuel pebbles in and out of the core is proposed. Firstly a model for pebble flow circulation is developed to provide a basic simulation framework. Then the irradiation history of fuel pebble could be tracked by combining pebble flow model and burn up calculation. The representative kinematic model and discrete element method (DEM) are introduced to numerically simulate the profiles of pebble flow. The classical batch-tracking methods as well as our newly-introduced DEM-tracking method are presented to perform the time-dependent analysis of pebble burnup. Overall with the burnup data obtained after going through multiple cycles, the burnup distribution of fuel pebbles in and out of the core could be reconstructed through the statistics result according to the pebble circulation model. Finally the quantification of the relationship between the pebble burnup assay and the economy and safety of the PBR would be more precise, thus providing implications on proposing reasonable requirements for accuracy of online burnup assay.
Resonance Self-Shielding Treatment for Fully Ceramic Micro-Encapsulated Fuels With the Embedded Self-Shielding Method
Authors: Li Jikui , Cao Liangzhi , Zu Tiejun , Wu HongChun , He Qingming
Published: October 2017

Fully ceramic micro-encapsulated (FCM) fuels generate double heterogeneity (DH) challenging greatly for classical resonance self-shielding calculation method. New methodologies have been proposed and verified in this research. The target of this study is to provide homogeneous multi-group cross sections reflecting the effect of DH. Embedded Self-Shielding Method (ESSM) [1] was selected to perform resonance self-shielding calculation. Therefore, Monte Carlo code MVP [2] which is capable of well modeling the stochastic dispersed tri-structural isotropic (TRISO) coated fuel particle throughout carbide matrix and method of characteristics (MOC) were chosen to develop the heterogeneous resonance integral (RI) tables for DH problems. Benchmark problems from reference [3] were provided to verify the new methodologies. The results show that ESSM with RI tables from MVP and MOC could well address the resonance calculation for DH problems.
Stress Analysis of Three-Layer SiC Cladding for PWRs
Authors: Lan Weiping , Lin Dachao , Hao Zulong , Niu Fenglei
Published: October 2017

Silicon carbide (SiC) and SiC matrix composites (SiCf/SiC) are being investigated as potential fuel cladding materials for advanced PWRs in order to improve the safety of nuclear power plants. The conceptual design of multi-layered SiC cladding (consisting of a monolithic SiC layer, SiCf/SiC composite layer and a monolithic SiC coating layer) has been investigated to meet the fuel requirements of both the strength and impermeability. A stress distribution model of the triple-layered SiC is developed on the basis of the theory of thermo-elasticity mechanics, taking radial temperature gradient and swelling effects into account as well. The heat transferring behavior of the cladding is investigated by analyzing the temperature distribution under steady conditions. Finite Element Analysis (FEA) code ANSYS is used to obtain the stress and temperature nephogram of multi-layered SiC fuel cladding under simulated steady conditions. Compared with the results of ANSYS, the stress distribution model and temperature distribution is validated.
Experimental Study of the Sub-Channel Flow Regimes in 5×5 Rod Bundles With Simplified Grid Spacer
Authors: Ren Quanyao , Zhou Wenxiong , Pan Liangming , Liu Hang , Yu Bin , Ye Tingpu , Li Songsong
Published: October 2017

Traditionally, the flow regime in two phase flow in rod bundles are considered in a global sense by the visualization. However, a sub-channel flow regime is required to understand and model the two phase flow structures in rod bundles. In this work, a sub-channel impedance meter was designed to get the dynamic feature in the sub-channels, which was applied to identify the sub-channel flow regime by the fast search and finding peaks of the cumulative probability distribution functions (CPDFs) objectively. In the present study, five flow regimes, namely bubbly flow, quasi-cap bubbly flow, quasi-slug flow, cap-turbulent flow and churn-turbulent flow were defined and recognized. The sub-channel flow regimes at the same cross section were compared to each other, which show similar feature with the local flow regimes in pipe. It is possible to identify nine different global flow regime configurations by combining the corner, side and inner sub-channel flow regime at the same cross section, which was drawn in the 2D sub-channel flow regime.
Investigation of the Flow Regime and Flow Induced Vibrations (FIV) of the Control Rod Inside the Guide Channel of PWR
Authors: Ali Muhammad , Daogang Lu , Huimin Zhang , Ahmed Shahroze , Ali Muhammad Zeeshan
Published: October 2017

The phenomenon of Flow induced vibrations (FIV) through the motion induced into the control rod of the nuclear core model apparatus due to high speed internal and external fluid flow. It involves a variety of flow phenomena around the restrictions and over a range of vibration frequencies of the control rod and other related components. The results of a CFD analysis are presented, which are compared with the experimental values obtained from the experimental setup specifically designed for solving and reviewing the fluid induced vibration problems, and thus forming equations of fluid flow regime and vibration modeling. The experimental setup is a model of a nuclear core consists of a fuel assembly without any fuel rod, a control rod held from a rigid screwed support and three fluid mixing plates similar to grid spacers with external and internal fluid flow. There are two inlets, forcing water from below and one of the sides of the apparatus while one outlet located above from both on the opposite side of the second inlet. The results of the experimental setup were detected by laser displacement sensor, to measure the vibration of the control rod, connected with a computer system. This vibration displacement data was measured under different flow conditions rendered due to different pressures through the inlet pumps. The different meshing results give a comparison of 3D modeling under different meshing strategies. Different group of inlet and outlet flow values have been considered according to the designed apparatus including velocities, pressure gradients to form the equations. The procedure followed for numerical method is starting from static fluid flow equations and moving on to the transient equations and thus forming the concluding equation. Similarly, the simulations have been improved from simple steady conditions to the transient solution while for the vibration modeling general modeling rules have been followed and different constants have been taken from the materials libraries available and the values of load from the experimental data and simulation. The results obtained from simulations represent that 3D modeling refinement makes it much complex and as known takes more time and needs heavy computing memories but the results are acceptable to the extent required. So the equations of fluid induced vibrations and fluid flow regime in a vertical nuclear core, by performing simulations of ANSYS CFX and FLUENTS and obtaining the experimental values and calibrating the values with those obtained from simulations as well as numerically solving the problem and finally comparing them all.
The Verification of SONG Library
Authors: Bei Hua , Zhao Jinkun , Chen Qichang , Si Shengyi
Published: October 2017

We have developed the 293-group cross-section library SONGLIB with complete reaction pathes, broad-spectrum solving ability and wide range isotopes, which is adopted in the lattice code SONG, the next generation reactor lattice calculation and analysis code. SONGLIB provides the multi-group data for the transport calculation, resonance calculation and burnup calculation of SONG. So the precision and reliability of data is very important. However, the size of the data is very large and there are still a lot of personal factors in determination of the processing parameters and methods. In order to evaluate the library, it’s necessary to carry out the test work comprehensively.

Amount of test work is then undertaken based on three kinds of reactors such as pressurized water reactor with UO2 fuel, sodium cooled fast reactor and new-type molten salt reactor with none-beryllium core. In order to test the adaptivity of the library, 39 cases are designed totally, changing the fuel composition, the absorber compositon, the burnup depth and the working temperature. Only the results of a few cases are given in details to shorten the length of this article. For comparison, the Monte Carlo code MCNP is chosen and the library source of evaluated nuclear data is ensured to be the same. The model that has been calculated is simple, the cell with “fuel-cladding-modulator” the three-layer structure and total reflection boundary for the aim of minimizing the differences between the codes. From the macroscopic angle, the results of infinite multiplication factor fit well with each other in total. From the microscopic angle, the absorption rate for actinides, fission products and absorbers, the fission rate for fissile isotopes and scattering rate for moderators are compared with that of MCNP whose output is normalized to the actual cell power. The difference between the reaction rates for most isotopes can be neglected. But there still exist some obvious differences for some isotopes whose resonance effect should have been considered, such as Fe56, ZrNat and W, which might affect the macoscopic results to some extent.

As a conclusion, the data of SONGLIB can be expected with high precision and reliability. And SONGLIB is adapted in solving various problems for reactors with different spectrum, depth of burnup, operation condition or fuel cycling. Nevertheless, there still exist some factors that may affect the final results during the process and use of the library, which should be paid attention to while taking a further step in optimizing and updating the library.
Comparison of Heat-Transfer Correlations Obtained in Supercritical-Water-Cooled Bundles
Authors: Xie Hanqing , Maloufi Hakim , Zopf Andrew , Anderson William , Langevin Christian , Pioro Igor
Published: October 2017

SuperCritical Water-cooled Reactor (SCWR) as one of the six Generation-IV nuclear-power-reactor concepts will have increased thermal efficiency compared to that of current Nuclear Power Plants (NPPs) equipped with water-cooled reactors by operating the reactor coolant at supercritical conditions: Coolant pressure of about 25 MPa, inlet temperatures between 300–350°C, and outlet temperatures between 550–625°C. The major flow geometry inside the reactor core is the bundle flow geometry. For safe and efficient operation of an SCWR heat transfer coefficients should be calculated with minimum uncertainties.

Unfortunately, the vast majority of experimental datasets were obtained in vertical bare tubes cooled with SCW. Experiments in a bundle flow geometry are even more complicated and expensive compared to that in bare tubes. Due to this very few experiments have been performed in bundles.

According to the abovementioned, the vast majority of heat-transfer correlations are based on bare-tube data, and only one currently known correlation is based on a 7-element bundle cooled with SCW (the so-called, Dyadyakin and Popov correlation (1977)). Rods in this bundle are equipped with four helical ribs to enhance the heat transfer. However, the authors have not provided any dataset(s) associated with this bundle and correlation.

In the current paper a number of bare-tube heat-transfer correlations obtained in SCW and the Dyadyakin and Popov correlation were compared with two datasets obtained in an annular channel with the heated central rod and 3-element bundle. The central rod in this annular channel and rods in the 3-element bundle have the same heated length as those in the 7-element bundle tested by Dyadyakin and Popov in 1977, and are also equipped with four helical ribs.

The comparison showed that the Jackson correlation (2002) is the most accurate one in predicting Heat-Transfer-Coefficient (HTC) profiles in the annular channel at normal heat-transfer regime. The Dittus and Boelter correlation (1930) is the most accurate in predicting HTC profiles in the 3-element bundle at normal heat-transfer regime. No one correlation is capable to follow closely HTC profiles at the deteriorated heat-transfer regimes in both flow geometries. Aloo, it should be mentioned that bare-tube heat-transfer correlations, which have thermophysical properties based on bulk-fluid and wall temperatures, might have problems with convergence at high heat fluxes, i.e., above the heat flux at which the deteriorated heat-transfer regime starts in bare tubes.
Study on Social Stability Risk Analysis for Nuclear Power Projects
Authors: Rongxu Zhu , Feng Zhao , Haifeng Wang
Published: October 2017

In this paper, the social stability risk analysis of nuclear power projects is carried out, and the working process of social stability risk analysis and risk evaluation index system of social stability is established. The social stability risk analysis of nuclear power plant in Guangdong province was taken as an example to introduce the application in this paper, at the same time for the current nuclear power project should focus on risk points to put forward risk mitigation measures and recommendations.

This study can provide theoretical support and technical support for the social stability risk analysis of nuclear power projects, and provide decision support for the construction of nuclear power projects.
Process Design of Heat Pump Evaporation Technology for Liquid Radioactive Waste Treatment
Authors: Lu Jia , Zhou Yan
Published: October 2017

The Site Radwaste Treatment Facility (SRTF) was to treat the liquid radioactive waste with high concentration of electrolytes in CAP1400 plant. Unlike the forced-circulation or natural-circulation evaporator, the waste was purified by evaporation process using heat pump principle in SRTF. This process combined the benefits of heat pump technology and the separation function of evaporation. This paper gave a brief description of the complete evaporation process. The main equipments of the process included pre-heater, evaporator, demister, steam compressor, and distillate tank. The whole system and each unit operation in the process were elaborately simulated based on energy balance. The results can be further used for equipments selection and design. Through the whole discussion of the heat pump evaporation process, it showed the heat pump evaporation process has remarkable energy efficiency and great advantages in equipment layout compared to traditional evaporation treatment for liquid radioactive waste.
Radiation Dose of Airborne Radioactive Material in Nuclear Power Plant Conventional Operating Conditions
Authors: Zhanyang Liu , Naigui Tao , Yang Chen , Yunliang Tao
Published: October 2017

In this paper, air-immersion, ground deposition, ingestion and inhalation of airborne radioactive effluent released from nuclear power plant under normal operating conditions is studied according to the atmospheric diffusion and ground deposition patterns and parameters that are suitable for the environmental characteristics of the nuclear power plant site, and the public living habits and food chain parameters around the site. Based on the Gaussian plume model, with a radius of 80 kilometers we divide 1, 2, 3, 5, 10, 20, 30, 40,50,60,70,80 km concentric circles around the nuclear power plant site. The 16 compass azimuth axial are the sector center-line, forming a total of 192 sub-regions, atmospheric diffusion of radionuclides is simulated in the assessment area of the region. The annual average atmospheric dispersion factor is calculate by using hourly observation data of wind direction, rainfall and atmospheric stability of the meteorological tower and the ground station, taking into account the ground reflection during transmission, the the decay of the radionuclide, and the loss brought by the wet and dry settling that caused by gravity and rain washing. The airborne radioactive effluent is deposited on the ground or plant surface by dry settling and wet settling in the process of atmospheric environment changing and diffusion. Radioactivity of per unit area brought about by dry settling and rain fall settling is described by the deposition coefficient and deposition speed. The long-term ground deposition factor and ground annual concentration in the evaluation area were calculated under the situation of airborne radioactive effluents in the nuclear power station mixing emission, and the calculated result of radionuclide concentration in the air and soil was compared with the natural background value and the actual monitoring value. Based on the radionuclide deposited on the ground and air through the terrestrial food radioactive transfer mode, together with a large number of environmental surveys data on the population distribution, agriculture, farming, animal husbandry and people’s living and eating habits in the 80km around nuclear station, combing with the actual situation of nuclear power station, the calculation model is amended accordingly. Using reasonable dose mode to calculate the maximum individual and entire public effective dose of the residents in the assessment area, and the results will be compared with other human activities. By comparing the calculated results of radionuclide concentration and radiation dose, it provide quantitative reference information for us understanding the influence of nuclear power station on the surrounding radiation environment, and to meet the requirements of nuclear power plant influence on surrounding environment and people under normal operating conditions.
The Impacts of Scale Effect on Flow Patterns in a Supersonic Steam Injector
Authors: Fujishiro Masaya , Abe Yutaka , Kaneko Akiko
Published: October 2017

From the viewpoint of an importance of safety, the nuclear power plant should be managed to prepare severe accidents. The performance of safety dropped by an accident is strongly to be minimized during the situation of station blackout. The installation of a steam injector (SI) into the nuclear power plant has long been expected.

In the SI, the steam condenses due to the direct contact at the surface of water jet, resulting in the force attracting water. The force drives the circulation of an amount of coolant water. SI also works as a reactor condenser thanks to its high efficient performance during the condensation. Because any external forces to circulate water and steam are not required, SI can be operated without the electric powers. The structure of SI is similar to a convergent-divergent nozzle. After the flow acceleration at a throat, the discharged pressure is expected to exceed the inlet pressure. Owing to its quite simple structure, the reduced cost of installation and maintenance is also expected.

The following previous studies for four cases of throat diameter clarified two-phase flow structures and heat transfer characteristics in water jet and performance of SI: (i) Narabayashi et al. (2000) examined for 5.5 and 6.5 mm in diameter; (ii) Osakabe et al. (2004) for 3.4 mm; (iii) Koizumi et al. (2006) for 4 mm; (iv) Abe et al. (2014) for 4, 6.5, and 8 mm. Although these clarified the operative state which formed a water jet, operative condition was not elucidated. Furthermore, the scale effect for various diameters of SI has not been discussed in detail.

The aim of this study is to clarify scale effect of a test section on operating criteria and performance. Experiment was performed to clarify the scale effect by using three types of throat diameters: 4, 6.5, and 8 mm.

As a result, three formations of a water jet were observed: (i) formation, (ii) incomplete formation, and (iii) no formation. We proposed a classification which enables us to categorize complex flow patterns into five regimes. We clarified the operating criteria of them by comparing water flow rate with steam flow rate. SI did not form a water jet on the condition with low steam flow rate. The suppling water was stopped, and only steam was supplied to the test section for the condition that steam latent heat was larger than subcooled water enthalpy.
Comparison of Test and Analytical Results on CAP1400 Reactor Flow Distributor Designs
Authors: Wei Zhang , Ming Zhang , Hao Yu , Qing Yu , Shaoxuan Lin
Published: October 2017

The CAP1400 reactor internal is going to use a new component termed the “Even Flow Distributor (EFD)”, instead of the existing flow skirt (FS) design, to help distribute the incoming flow more evenly to the fuel assemblies. To verify the effect of the EFD, a scale model of the reactor and internals was built and hydraulic tests of both the EFD and the FS configurations were conducted. In addition, numerical simulations of the flow fields, using CFD, of both designs were also carried out.

From the scale model test results, the overall flow distribution of EFD is better than that of the FS. The core inlet flow distribution taken from the CFD results is slightly better than that from the hydraulic test. The differences between CFD result and test results are less than 3 percent for the most of fuel assemblies, and about 5 percent for a few assemblies. Based on this study, it is concluded that the EFD is a very effective means of controlling core inlet flow distribution in a CAP1400 reactor.
Analysis of CAP1400 Reactor Coolant Hot Leg Flow Field
Authors: Tingting Niu , Chunli Liu , Shuan Xia
Published: October 2017

The CAP1400 reactor coolant main pipe plays a very important role in the plant safety operation, as it transports coolant among the Reactor Vessel, the Steam Generator and the Reactor Coolant Pumps, simultaneously connects the main equipments. The flow resistance of CAP1400 coolant main pipe is calculated by classic flow resistance calculation formula at present, but a big tolerance would be caused by the big size, complex structure and multi system/ instruments interfaces of the main pipe. Otherwise, CAP1400 Reactor Coolant System employs elbow flow instrument to measure coolant flowrate, the big size piping nozzle like ADS 4th and surge line might influence the stability of flow field, which affects the accuracy of the elbow flow instrument. A CFD software STAR-CCM+ is used to model CAP1400 main pipe hot leg in this paper and analyze the detailed flow field to obtain more precise calculation results, and verify if the elbow flow instrument arrangement is reasonable.
Annealing Temperature Effect on Creep Property of CZ Alloy for PWR Modern Fuel Claddings
Authors: Xu Yang , Tan Jun , Chen Liu-tao , Wen Dun-gu , Zou Hong , Gao Chang-yuan , Wang Xu , Dong Si-si , Wang Wei-jun
Published: October 2017

Zirconium alloys remain as the main cladding materials in most water reactors. CZ zirconium alloy is recently developed for advanced PWR fuel assembly (STEP) designed by China General Nuclear Power Group (CGN). There are two kinds of zirconium alloys, designated as CZ1 and CZ2 respectively. It is well known that creep property is one of the most important characteristics to evaluate the integrated performance of new commercial alloys. In this study, we have collected the creep date of CZ alloys heat-treated at different temperatures. The result shows that the creep rates of the zirconium alloys enlarge with the creep stress in the temperature of 375 °C. The creep resistance is improved with the increasing annealing temperature. The creep resistance of CZ-RXA (high temperature) is better than CZ-SRA (low temperature). The creep rate of CZ alloys at relatively lower annealing temperature is much larger than that at higher annealing temperature. At the given stress condition, the creep behaviors of different annealed CZ alloys are found to have almost the same tendency. In addition, the creep rate of CZ1 is smaller than CZ2 at the same annealing temperature.
Numerical Simulations for Super Large-Scale Natural Draft Cooling Tower With an Optimized Non-Uniform Water Distribution
Authors: Li J. Y. , Wang H. , Sheng W.
Published: October 2017

The great spray area of a Super Large-Scale Natural Draft Cooling Tower (SLNDWCT) makes it difficult to achieve an uniform wind field, and non-uniform water spray distributions are adopted in engineering. In this paper, to improve the cooling performance, optimized non-uniform water spray distributions are designed by utilizing network hydraulic calculations and numerical simulations. In the network calculations, the node-formula is applied to figure out the water pressure and flow rate of each spray nozzle, providing more precise data in simulations for the heat and mass transfer. Simulations for operating in summer, Spring/Autumn and winter seasons, which are different in water spray density, have been presented. In the operating in summer, the spray zone is divided into two regions (inner and outer regions), and by adjusting the water spray density and areas of the two regions, an improved water distribution is achieved.
Verification of FPGA Based NPP I&C Systems Considering Multiple Faults: Technique and Automation Tool
Authors: Yasko Alexander , Babeshko Eugene , Kharchenko Vyacheslav
Published: October 2017

Instrumentation and Control (I&C) systems for Nuclear Power Plants (NPP) are exceedingly complicated electronic solutions that include thousands of different components such as microcontrollers, Field-Programmable Gate Arrays (FPGAs), integrated circuits etc. Deployment of such safety-critical systems cannot be performed without complex safety and reliability assessment, verification and validation (V&V) activities that are addressed to exposing of overlooked faults. The examples of such activities are Fault Tree Analysis (FTA), Failure Modes and Effects Analysis (FMEA), Fault Injection Testing (FIT). Due to complexity of NPP I&C systems in most cases the process of assessment is very time consuming and the results mostly depend on experts’ qualification.

Traditional safety and reliability assessment methods are being constantly modified and enhanced so as to comply with increasing demands of national and international standards and guidance, as well as to be applied for I&C systems that contain number of complex components like FPGA.

Although much work related to analysis of FPGA-based systems has been performed, there is a lack of detailed technique for FPGA-based I&C systems failure identification that considers probability of several faults at the same time (multi-faults), development of preventive strategies for controlling or reducing of the risk related to such failures, as well as automation of this technique so as to make it utilizable for real NPP industry tasks.

FIT as verification for Failure Modes, Effects and Diagnostics Analysis (FMEDA) was used during Safety Integrity Level 3 (SIL3) certification process of RadICS NPP I&C platform, while the parts of proposed technique were used as internal verification and validation activities applied on several modules of the platform.
The Research and Comparison of Thermal-Hydraulic Test Facilities for Pressurized Water Reactors (PWRs)
Authors: Lu Yuan , Xiao Changzhi
Published: October 2017

Recently, nuclear power safety draws more attention after Fukushima nuclear accident, for which it is essential to construct a large number of test facilities simulating possible issues occurred in the reactor. The thermal-hydraulic test facility is extensively used to simulate thermal-hydraulic response during a loss of coolant accident (LOCA) or an operational transient which can minimize the nuclear safety accidents. This paper focus on the research of thermal-hydraulic test facilities of PWRs in different countries. All of facilities were designed by scaling analysis method. Meanwhile, a wide range of data comprising of power data, pressure data, volume data, configuration and a series of nuclear safety test data is compared in over ten test facilities. Based on above comparable data and relevant research, the main conclusions are as follows:

Scaling analysis lays a solid foundation for the design and construction of scale-down nuclear reactor thermal-hydraulic test facilities. This would provide the reference for choosing scaling concepts in the reactor integral test facility.
Experiment of CHF Enhancement by Magnetite Nanoparticle Deposition in the Subcooled Flow Boiling Region
Authors: Choi Young Jae , Kam Dong Hoon , Jeong Yong Hoon
Published: October 2017

CHF experiments were conducted in subcooled flow boiling region at atmospheric pressure. Magnetite nanoparticles were deposited sufficiently on the test sections in same deposition process. Then, working fluid was changed to DI water. After the nanoparticle deposition process, the surface became very hydrophilic even after CHF experiment using DI water. The wettable surfaces were observed using static contact angle and SEM image. CHF results of bare stainless surface and nanoparticle-deposited surface were obtained in the subcooled boiling region. Experiments were conducted over a mass flux range from 1,000 kg/m2s to 5,000 kg/m2s and with inlet temperatures of 40, 60, and 80 °C. The CHF enhancement was from 0% to 40 % by the nanoparticle deposition, which is related wettability enhancement. The CHF enhancement increased as the mass flux increased, which lead to exit quality decrement.
Sensitivity Study of LOCA-Generated Debris on the Pressure Drop of Fuel Assembly
Authors: Bhatti Hammad Aslam , Guo Zhangpeng , Zhuo Weiqian , Ahmed Shahroze , Wang Da , Niu Fenglei
Published: October 2017

The coolant of emergency core cooling system (ECCS), for long-term core cooling (LTCC), comes from the containment sump under the loss-of-coolant accident (LOCA). In the event of LOCA, within the containment of the pressurized water reactor (PWR), thermal insulation of piping and other materials in the vicinity of the break could be dislodged. A fraction of these dislodged insulation and other materials would be transported to the floor of the containment by coolant. Some of these debris might get through strainer and eventually accumulate over the suction sump screens of the emergency core cooling systems (ECCS). So, these debris like fibrous glass, fibrous wool, chemical precipitates and other particles cause pressure drop across the sump screen to increase, affecting the cooling water recirculation. As to address this safety issue, the downstream effect tests were performed over full-scale mock up fuel assembly. Sensitivity studies on pressure drop through LOCA-generated debris, deposited on fuel assembly, were performed to evaluate the effects of debris type and flowrate. Fibrous debris is the most crucial material in terms of causing pressure drop, with fibrous wool (FW) debris being more efficacious than fibrous glass (FG) debris.