bluuebunny/crossref_metadata_2025 huggingface dataset:
The Investigation in the Oscillation Modes of In-Phase and Out-of-Phase in a Two-Phase Forced Circulation Loop With Multiple Nuclear-Coupled Boiling Channels
Authors: Lee Jin Der , Chen Shao Wen , Pan Chin
Published: October 2017

This study develops the nonlinear dynamic model for a multi-channel nuclear-coupled boiling forced circulation loop. Stability maps and nonlinear dynamics of a three nuclear-coupled boiling channel forced circulation loop are investigated. The possible mechanisms of in-phase and out-of-phase oscillations, and their distributions along with the stability boundaries are also discussed. For the pure thermal-hydraulic system without considering the nuclear coupling, the thermal-hydraulic feedback gain tends to dominate the system with different radial heat flux ratios to present out-of-phase mode of oscillations in the boundary states. The difference in heat fluxes among channels would enhance the channel interactions to destabilize the system and cause the phase difference among channel oscillations. The effects of neutron interaction and void-reactivity feedback have a great influence on the oscillation mode among three channels in the nuclear-coupled boiling forced circulation loop system. For the asymmetric heating system with a reference void-reactivity coefficient, the case with a relatively weak neutron interaction remains presenting out-of-phase mode with phase difference along with the stability boundary. However, the system with a relatively strong neutron interaction, the in-phase mode with phase difference distribute over the boundary states in the medium to high subcooling number region while the neutronic feedback gain may be dominant. On the other hand, it would favor out-of-phase mode with phase difference at the boundary states of low subcooling numbers if the thermal-hydraulic feedback gain is stronger and the two-phase frictional pressure drop is dominant in all channels. Moreover, the out-of-phase mode among unequal-heating channels in the weak neutron interaction system may turn into in-phase mode as strengthening the neutronic feedback gain through the reference void-reactivity coefficient doubled.
The Measuring of Sodium Droplet Size Distribution and its Impact on the Consequence of Sodium Spray Fire
Authors: Guozhi Wang , Haiou Du , Rongdong Wang , Wentao Shi , Jun Piao
Published: October 2017

The sodium droplet size distribution has significant impact on the consequence of sodium spray fire. And it is fundamental input data for the validation and application of sodium spray fire code. The experiments were carried out in a closed vessel to measure the sodium droplet size distribution. Liquid sodium of 250 °C was sprayed downward into the vessel in the form of sodium droplets through a nozzle with a diameter of 2.4mm. The vessel was inerted by argon gas to prevent the sodium droplets from burning. A real time spray droplet sizing system based on ensemble diffraction technique was used to measure the size of the droplets. And the laser beam was passed through two glass windows on the wall of the vessel to reach the sodium droplets. The tests showed that the sizes of the sodium droplets ranged from 184μm to 1000μm at the nozzle pressure of 0.15MPa. And median diameter was 532μm. The sodium spray fire code named NACOM was used to evaluate the impact of particle size distribution on sodium fire. The measured sodium droplets size distribution and the Nukiyama-tanasama drop size distribution were divided into 11 groups to be used as input data for the NACOM code. A comparison showed that 23% of particles in Nukiyama-tanasama drop size distribution were over 1000μm, while the largest size of particles in the measured sodium droplets was 1000μm. The calculation by NACOM code showed that the trend and value of temperature and pressure in the vessel were similar, so to some extent Nukiyama-tanasama drop size distribution is a good approximation of the real sodium droplet size distribution. However, Nukiyama-tanasama drop size distribution may be unsuitable for application in sodium fire safety analysis, because the temperature and pressure calculated from which was lower than that of the real droplet size distribution.
Comparison and Analysis on Two Kinds of Passive Residual Heat Removal System Designs Under Station Blackout Accident for Integral Small Modular Reactor
Authors: Li Fei , Shen Feng , Bai Ning , Meng Zhaocan
Published: October 2017

Small Modular Reactor has gained much attention in recent years. The passive residual heat removal system (PRHRS) is designed to increase the inherent safety features of the Integral Small Modular Reactor. There are many differences on the design of PRHRS. To get a comprehensive understanding of the PRHRS design in ISMRs, two simplified simulation models of ISMRs with different PRHRS design are built by the use of thermal hydraulic system code Relap5/Mod3.2 in this paper. A blackout accident is introduced to study the different performance between two PRHRS design models. The calculation results show that both two cases can successfully remove decay heat from the core. But there are still some differences between two cases in aspects of primary and PRHRS coolant parameters. Comparisons of the results from two cases are conducted in this paper, and the differences are carefully analyzed too.
Nuclear Power Plant Flexible Power Operation Capability Analysis
Authors: Cheng Ye , Minglu Wang , Zhongming Qiu , Yong Wang
Published: October 2017

Nuclear power plants are used extensively as base load sources of electricity. This is the most economical and technically simple mode of operation. In this mode, power changes are limited to frequency regulation for grid stability purposes and shutdowns for safety purposes. However for countries with high nuclear shares or desiring to significantly increase renewable energy sources, the question arises as to the ability of nuclear power plants to follow load on a regular basis, including daily variations of the power demand. This paper give a detailed analysis of the technical and economic aspects of load-following with nuclear power plants, and summaries the impact of load-following on the operational mode.
The R&D Activities and Future Plan of 4MW Hot Cathode Ion Source for EAST Neutral Beam Injector
Authors: Xie Yahong , Hu Chundong , Liu Sheng , Li Jun , Xie Yuanlai , Liang Lizhen , Jiang Caichao , Wei Jianglong , Xu Yongjian , Tao Ling , Gu Yuming , Chen Shiyong , Sheng Peng , Liu Zhimin
Published: October 2017

The Experimental Advanced Superconducting Tokamak (EAST) is one of the fully superconducting tokamak, its aim at the long-pulse operation (1000s) to study the physics of steady-state operation for nuclear fusion sciences. In order to support the steady-state operation and physical research, the high power neutral beam injection (NBI) system need to be employed on the EAST for the plasma heating and current driving. According to the scientific study schedule of the EAST, the designed NBI system includes two beam lines which will be constructed in two phases. Each beam line will deliver a deuterium neutral beam with beam energy of 50–80 keV with beam duration of 10–100 s. Each beam line has the maximum beam power of 4MW.

The high current ion source is one of the most important parts in the high power NBI. A hot cathode positive ion based source was developed for EAST-NBI, which shown in Fig. 1. The ion source contains a bucket hot cathode arc chamber with 650 mm long, 260 mm width and 300 mm depth. There are 32 hairpin filaments with diamond of 1.5 mm and 160 mm long to supply primary electrons. A tetrode type accelerator with slit type used to extract the ions from the plasma and accelerated to the desired energy. The beam extraction area is 120 mm × 480 mm (can be changed) with beam transmittance of 60 %. The designed beam species is deuterium with beam power of 2–4 MW and beam energy of 50–80 keV and beam pulse length of 10–100 s.

The ion source needs to be conditioning before operation on the EAST-NBI. An ion source test bed was designed and developed for the ion source performance tests and the optimization verification. The characteristics of ion source were tested with hydrogen beam, and each ion source should achieve 4MW ion beam with beam energy of 80 keV. The optimum beam perveance and arc efficiency were analyzed too. The optimum beam perveance was 2.8 μp with beam energy of 50 keV and the arc efficiency was 0.55A/kw.

Long pulse operation was one of the requests of EAST ion source. The real-time feedback control method was employed and can got stable plasma and ion beam. The beam extraction was tested to achieve 100 s on the test bed. Consider the high power deposited on the calorimeter, the beam was modulated with suitable frequency and duty ratio. When the conditioning finished, the ion sources were moved to the EAST-NBI. The deuterium beam was extracted and injected into the EAST plasma. Details of the performance of positive ion source on the test bed and EAST-NBI will be presented.
Numerical Simulation of the Cavitation Flow Inside the Jet Pump
Authors: Xiao Qi , Wang Junrong , Zhao Zhenxing , Bai Fan , Wei Zhiguo , Li Shaodan
Published: October 2017

Jet pumps have the advantages in reliable operation, none of moving components and simple construction etc, which makes it applicable in a wide range of industry situations. The cavitation easily occurs within jet pumps and it turns to be a bottleneck for the jet pump design. In this paper, the cavitation flow inside the jet pump would be investigated by 2-D axisymmetric CFD simulations. Firstly, the single liquid phase flow in the jet pump would be simulated and the grid independence test would be carried out. The simulation results show that both the pressure ratio (h) and the efficiency (n) would reach to convergence values with 143,484 meshes, which would be adopted in the following simulations. By comparing with the experimental data of single phase flow in literatures, it was found that the Reynolds stress model offered better predictions. Then the cavitation flow in the jet pump was simulated. It could be found that the cavitation phenomenon trends to occur in the shear layer and the wall detached regions of the throat. At last, the influence of saturated vapor pressure model, the outlet pressure and the nozzle-to-throat clearance would be evaluated. It was found that different from the single phase flow, the optimal nozzle-to-throat clearance would vary with the operation conditions.
Research on Natural Circulation Flow Characteristics of Floating Nuclear Power Plant in Heaving Motion
Authors: Ren Li , Minjun Peng , Genglei Xia , Yanan Zhao
Published: October 2017

The FNPP (Floating Nuclear Power Plant) expanded the application field of Integrated Pressurized Water Reactor (IPWR) in the movable marine platform, it is necessary to study the natural circulation flow characteristics in heaving motion on the ocean. From the characteristics of FNPP, by means of THEATRe code which was based on the two-phase drift flux model and was modified by adding module calculating the effect of heaving motion, the simulation model in heaving motion was built. Using the models developed, the natural circulation operating characteristics of natural circulation in heaving motion and the transitions between forced circulation and natural circulation are analyzed. In the case of amplitude limited, the periods of mass flow rate are equal to periods of heaving motion. The oscillation amplitude of mass flow rate increases with the heaving amplitude increase. In the case of period limited, the natural circulation flow rate oscillating amplitude increases with the heaving period increases. The result obtained are not only evaluating FNPP design behavior properly but also pointing out the direction to further optimum design to ensure FNPP operating safety in heaving motion.
Preliminary Investigation of Pool Sloshing Behavior in a Two-Dimensional Experiment
Authors: Shuo Li , Songbai Cheng , Kejia Li , Shaopeng Lin , Nan Zhang
Published: October 2017

Studies on the pool sloshing behavior are of crucial importance for the improved assessment of energetic potential of a large whole-core-scale molten fuel pool that might be formed in the transition phase of a Core Disruptive Accident (CDA) of Sodium-cooled Fast Reactors (SFR). In this work, to clarify the mechanisms underlying this behavior, a series of simulated experiments was performed by injecting nitrogen gas into a two-dimensional rectangular water pool from a nozzle positioned at the center of pool bottom. To achieve a comprehensive understanding, various experimental parameters, including the nitrogen gas pressure, nozzle size, initial water depth, as well as the initial particle bed height, particle size, particle density and the particle shape, have been planned. From the preliminary analyses, it is found that the gas pressure, initial water depth and particle bed height are observable to have a remarkable impact on the sloshing characteristics (e.g. maximum magnitude of water level in the center and peripheries of water pool). Due to a much-reduced residence time of the injected gas in the pool, mitigation effect of gas pressure, which could restrain the intensity of sloshing motion in the fuel pool, is observable. However, as to the sloshing motion with a certain kind of particles, as the initial particle bed height increases, the intensity of sloshing motion is confirmable to become stronger and stronger. Knowledge and fundamental data from this work might be utilized for the predicative-model studies as well as verifications of SFR severe accident analysis codes in China in the future.
Study on Final Depth Under Hydraulic Coolant Penetration Condition
Authors: Zhou Yuan , Chen Jingtan , Huang Qingyu , Gong Houjun
Published: October 2017

Coolant injection mode of molten fuel and coolant interaction (FCI) is a key issue during steam generator tube rupture accident (SGTR) in liquid metal reactors. In the present study the hydraulic breakup behavior of coolant injection mode is investigated using experimental and numerical approaches. Visual experiments are conducted using low-density gasoline as coolant jet and high-density water as denser liquid. The gasoline jet is released into a transparent water tank through a nozzle. The jet breakup behaviors are captured by a DSLR (digital single-lens relax) camera. The images of jet behaviors and the data of gasoline jet penetration depth are obtained and analyzed. By changing nozzle diameter and nozzle height, the parameter effect of jet diameter and jet inlet velocity on final penetration depth are studied. Based on FLUENT15.0, the hydraulic breakup behaviors of gasoline jet are simulated. A 3D axisymmetric model is built and Volume of Fluid (VOF) method is used. The numerical simulation results agree with the experimental results quantitatively and qualitatively. These experimental images and data are helpful to substantiate the understanding of the coolant jet breakup behavior and the pattern of jet penetration depth.
Scaling Analysis of Passive Heat Exchanger Under Station Blackout Accident
Authors: Yu-sheng Liu , Chao Xu , Shao-xin Zhuang , Cong-xin Li , Pan Zhang
Published: October 2017

Station blackout accidents increasingly become the focus of research in the field of nuclear safety after Japan’s Fukushima nuclear plant accident in March 2011. Core decay heat under station blackout condition will be transferred by natural circulation occurring between core and passive heat exchanger for the nuclear plants incorporated passive safety design concept such as AP1000 or CAP1400. As a result, response of safety systems will differ in accident sequence and kind between passive safety plant and traditional plant. What is more, cooling capacity of passive heat exchanger (PHX) which takes on heat sink has significant effect on performance of natural circulation in passive safety system.

The safety need that characteristics of passive safety plant should be verified through integral experiment facility makes scaling analysis important in design or modification of experiment facility. Furthermore, scaling analysis of natural circulation phenomena under station blackout accident plays an important role in design verification, safety review verification or thermo-hydraulic program development. It not only determines the similar similarity criteria between the nuclear power plant prototype and test facility, but also provides technical basis for selecting different experiment schemes.

As a part of scaling analysis on natural circulation phenomena for station blackout, the cooling capacity of PHX in test facility should be scaled properly and reasonably with conservatism. Therefore, scaling of passive residual heat removal (PRHR) heat exchanger under station blackout accident is investigated analytically in this paper. The analytical model for natural circulation in passive heat exchanger is established based on the performance characteristics of PRHR system in passive plant. By proper hypothesis and simplification, the governing equations for PHX are normalized using steady-state solutions, initial or boundary conditions. The similarity criteria that should be preserved between PHXs in test facility and prototype are finally obtained from non-dimensionalized equations. Furthermore, the distortion analysis for PHE design is also investigated based on the similarity criteria for selected scaling factors and parameters. The safety analysis based on models of nuclear power plant prototype and test facility is conducted on transient performance of designed PHX with PHX of prototype. The results show that: heat source number is the dominant similarity criteria for PHXs design under SBO condition. Requirements of Richardson number and friction number could be satisfied by resistance adjusting on test loop. The performance of PHX designed following heat source number requirement can better represent the transient response characteristics of prototype under SBO condition.
An Experimental Study on Critical Heat Flux in Rectangular Channel With Different Angle of Inclination
Authors: Zhang Ming , Zhao Yufeng , Gao Tianfang , Hou Fangxin , Chen Peipei
Published: October 2017

It is a hot research field to study the Critical Heat Flux (CHF) occurrence mechanism in boiling heat transfer. Although lots of researchers have studied on it, no unified conclusion has been achieved up to now. The proposed CHF occurrence mechanism is also not widely accepted. Because the void fraction close to the heating surface is larger when the heat flux approaches CHF, it is difficult to make visual observations of the boiling heat transfer on the heating surface. Usually the CHF mechanism is based on certain assumptions, and then confirmed by testing. Therefore, it is meaningful to study the occurrence of CHF by experimental methods.

Based on the system design of the test system for CHF in a rectangular channel, an experimental facility was set up. The main test section consists of a rectangular flow channel with the copper heating surface downwards mounted into one of the channel walls. The fluid is deionized water. Fluid subcooling is 15 K. The entire test section is mounted on a rotating arm which can be set at different inclination angles from 0° (horizontal) to 90° (vertical). By adjusting the loop, the natural circulation and forced circulation test conditions can be achieved from the experimental facility. Through test research, visual observations are acquired about the bubble growth characteristics in the rectangular channel in the range of 0 ° to 90 °, and the test data about the CHF is also attained.

It is found that under the condition of forced circulation and natural circulation, the CHF values is increased effectively with the improved mass flow rate and the increased inclination angles of the test section. Through visual observation, it is found that in forced circulation conditions, CHF occurs first at the entrance location of the test surface, rather than the maximal heat flux location of the test surface (test surface center) or the maximal local void fraction of the test facility (exit location of the test facility). In the natural circulation conditions, CHF occurs first at the exit location of the test facility. This phenomenon may imply that the mechanisms of CHF are different in forced circulation and natural circulation. In forced circulation, the flow plays a main role. In natural circulation, the local void fraction plays a main role.

There are some differences in the experimental phenomena compared with the traditional CHF theory, like the bubble crowding theory and the micro fluid layer theory. Through the experiment research, the complexity of the flow boiling heat transfer was found. The flow and boiling heat transfer affect each other. It is can help us to research the CHF theory in flow boiling heat transfer.
Development of a Sea-State Database in the Northwest Pacific Ocean by Long-Term Numerical Hindcast Modelling
Authors: Shi Jian , Joly Antoine , Zheng Jinhai , Zhang Jisheng , Chen Tian , Chen Guofei
Published: October 2017

In the Northwest Pacific zone, all the active nuclear power plants are located in coastal areas. The extreme typhoon waves, which occur in summer and autumn, pose great risks on the nuclear facilities and related coastal infrastructures. To facilitate the assessment of the impact of typhoon waves over the whole Chinese coastal line, detailed data of the sea state over this region is necessary. For this reason, a numerical sea state data base, which includes the North-West Pacific Ocean is under construction. It will range over several decades, to provide long-term wave data in this area. An oceanic wave model with a triangular mesh covering the whole Pacific Ocean has been set up by using the third generation spectral wave model TOMAWAC. The model is calibrated by comparing the simulation results over the period of January 1999 and during typhoon Muifa (1109) to buoy data. The numerical result of significant wave height (SWH) shows an excellent agreement with the oceanic buoy observations. The minimum values of the average bias between prediction and measurement is around 2–3 cm. The prediction of mean period (MP), however, is less accurate than the results of SWH. The model underestimates the mean period with amplitudes of around 0.6–1.2 s. Five different configurations of the source terms in TOMAWAC were tested during the calibration to find the best results. The comparisons show that the BAJ configuration can provide most accurate results of SWH. The predictions for configurations JKD and JWD show best results of MP. However, the SWD configuration provides the worse results for both of SWH and MP.
Neutronics Analyses of Small Compact Prismatic Nuclear Reactors for the Preliminary Design
Authors: Yang Xie , Shi Lei , She Ding
Published: October 2017

Due to the advantages of small volume, light weight and long-time running, nuclear reactor can provide an idea energy source for submarines, ships and even space crafts. In this paper, two small compact prismatic nuclear reactors with different core block material are presented, which have a thermal power of 5 MW for 10 years of equivalent full power operation. These two reactors use Mo-14%Re alloy or nuclear grade graphite IG110 as core block material, loaded with high enriched uranium nitride fuel and cooled by helium, whose inlet/outlet temperature of the reactor and operational pressure are 850/1300 K and 2 MPa respectively. High temperature helium flowing out of the reactor can be used as the working medium for Closed Brayton Cycle (CBC) power conversion to generate at least 1 MW electricity due to the high efficiency of CBC. Neutronics analyses of reactors for the preliminary design in this paper are performed using Reactor Monte-Carlo (RMC) code developed by Tsinghua University. Both the two reactors have enough initial excess reactivity to ensure 10 years of full power operation without refueling, which have at least $1 reactivity shutdown margin, and remains at least $1 subcritical in the submersion accident as well as one control drum failed accident. Finally, the optimization design is determined after comparing the U-235 mass and the total reactor mass of the above two prismatic reactors.
The Shutdown PSA Model of the TQNPP CANDU Reactor and the Outage Risk Management
Authors: Peng Xiaochun , Fu Qiang , Han Yu , Cao Yong
Published: October 2017

This paper presents the characteristic of Outage Nuclear Safety Management Guideline (ONSMG) for Qinshan CANDU plant and the introduction of shutdown and low-power PSA (SPSA) of TQNPP. The SPSA model is used to assess the outage configuration risk of OT109. The result and its risk insight are discussed about the two different outage working plan. In conclusion, the risk of nuclear safety and the economic are discussed based on the risk assessment result.
CFD Analysis Technique Development for Wire Contact Effect of Fuel Rods Within Next Generation FBR in Japan
Authors: Nakamura Hironori , Oyama Kazuhiro , Hayakawa Satoshi , Ogata Tomoaki , Maeda Seiichiro , Ozawa Takayuki
Published: October 2017

Generally, the inside of a fuel subassembly of sodium cooled fast reactor (SFR) has complex geometry and this complexity makes it difficult to conduct computational fluid dynamics (CFD) utilizing cyclic boundary condition in a radial direction. However, present work has created a new modeling method which enables us to model only one rod with one wire (1-rod model) to represent periodically installed rods with wires, like an actual fuel subassembly. The axial geometry of the 1-rod model has one wire pitch length and has two extra buffer zones to reduce the influence of inlet and outlet boundary conditions. Since this 1-rod model requires low calculation load to represent flow field of a center rod in the fuel subassembly, the model allows the choice of increased number of mesh in both radial and axial directions. With this 1-rod model, three-dimensional (3D) thermal hydraulic simulation has been conducted to evaluate a peak local temperature caused by contact of a wire with the fuel rod cladding. As the coefficient of wire contact effect, the simulation obtained a value of 2.4, which means peak coefficient of fuel cladding temperature.
Study on Purification of U Scraps in Continuous Countercurrent Extractor
Authors: He Fei , Wan Qiang
Published: October 2017

In this paper, the feasibility of continuous countercurrent extractor (CCE) applied in extraction of U scraps rather than mixer-settler was studied. Effects of rotation Reynolds number, residence time on extraction and reverse extraction efficiency were investigated respectively. During the process, impurity content in raw material had a slight effect on extraction. Under the optimum conditions of O/A phase ratio of 1.2 and rotation Reynolds number of 13824, extraction degree of scrap pellet solution and alkalescence dreg lixivium were good with the residence time of 19 min and 26 min respectively.

Under the best reverse extraction conditions of A/O phase ratio of 1.2, rotation Reynolds number of 23128 and residence time of 19 min, two kinds of raw material liquid mentioned as above could get better reverse extraction efficiency, while uranium concentration of raffinate phase and impurity content of reverse extract phase satisfied the technological requirement. This study verified the feasibility of CCE developed for purification of U scraps, identified the operation parameter, and provided an experimental basis for industrial expansion in the future.
Performance and Application of Internal Flood Probabilistic Safety Assessment in Nuclear Power Plant Design
Authors: Huang Lihua , Liu Xiang , Wang Lijuan
Published: October 2017

An internal flood probabilistic safety assessment (IFPSA) identifies risk-significant flood areas, components and scenarios, and evaluates the risk that a nuclear power plant (NPP) faces under internal flood conditions. This paper draws a comparison between NPPs in operation and those in the design stage with regards to the performance of IFPSA, and the main differences are shown in such tasks as flood area identification, flood source and SSC identification, plant walkdown, flood scenario characterization, flood frequency quantification and flood consequence analysis. Specific cases are presented in order to illustrate the application of IFPSA in NPP design to assess risks, identify potential weak links and provide insights into possible design modifications regarding internal flood protection.
Generation and Distribution of Tritium in HTGRs and Review on the Tritiated Water Treatment Technologies
Authors: Jia Fuming , Xie Feng , Li Hong , Cao Jianzhu
Published: October 2017

Tritium (H-3) can discharge mainly in the form of HT and/or HTO as gaseous and/or liquid waste into the environment from the nuclear power plant, and participate in the cycle among hydrosphere, atmosphere and biosphere, which would lead to the long-term radiological effects on organisms. Thus, in the daily operation of the nuclear power station, tritium became one of the most concerned nuclides in the source term analysis. In high temperature gas-cooled reactors (HTGRs), tritium was mainly generated by the ternary fission reaction of heavy nuclei in the fuel and neutron activation reaction of impurities like lithium-6 (Li-6), lithium-7 (Li-7) and boron-10 (B-10) in the graphite matrix, carbon bricks, etc. Tritium would be resorted completely in the intact tristructural-isotropic (TRISO) coated particles, while tritium in the graphite can diffuse into the primary loop depending on the local temperature. In the helium purification system of a typical HTGR, the molecular adsorber can adsorb the tritium in the primary coolant, and then the tritiated water was formed from the regeneration and desorption process of the molecular adsorber. Meanwhile, since the high permeability of tritium at a high temperature, it can permeate into the secondary loop through stainless steel heat exchange tube from the primary loop, and entered into the environment with leakage of the secondary water. Therefore, it was very important to analyze the production, transport and release mechanism of tritium for the estimation of the inventory and distribution of tritium in a nuclear power plant. With the rapid development of nuclear energy and the commercial application of HTGRs, tritiated water treatment technologies attracted more attention in the field of radioactive nuclear waste. Current paper will introduce and summarize general tritiated water treatment technologies, including water distillation, tritium sorbent process, palladium membrane reaction (PMR), and combined electrolysis catalytic exchange (CECE).
The Activation Sources of Reactor Internals During Decommissioning
Authors: Han Jianchun , Zhou Yan , Li Hui , Mei Qiliang
Published: October 2017

As China’s first nuclear power plant connected to the grid, the first Qinshan nuclear power plant is approaching the decommissioning period. Other nuclear power plants also turn into the preparation phase of decommissioning in succession. In order to facilitate decommissioning, source survey is conducted during the pre-decommissioning phase, which can provide radioactive inventory, contamination distribution, species and quantities of nuclides.

The internals of the reactor work under the most severe radiation environment. During the reactor operation, the materials of internals are irradiated by high-energy neutrons. So activated nuclides are generated due to the neutron capture reaction, which are the main radioactive waste to be treated during decommissioning. In this paper, the neutron irradiation and the generated activation source of the internals for pressurized water reactors (PWR) are studied and analyzed.

Firstly, core modeling was carried out, and the neutron transport calculation is performed to obtain three-dimensional distribution of the neutron flux.

Secondly, according to the three-dimensional distribution of the material composition and the neutron flux rate of the reactor, the activation calculation is carried out to obtain the activation source.
A Method for Online Calibration of MAAP Simulations in a Severe Accident Management System Database
Authors: Xu Weifeng , Yang Fangqing , Chen Peng , Liao Yehong
Published: October 2017

During a nuclear plant accident, five accident events are usually considered, including core uncovery, core outlet temperature arrived at 650 °C, core support plate failure, reactor vessel failure and containment failure. In accident emergency aspect, when an accident happens, the initial event can be utilized in the severe accident management system which is based on MAAP to simulate the long process of the accident, so as to provide support for operators to take actions. However, in MAAP, many sensitivity parameters exist, which reflect phenomenological uncertainty or models uncertainty and will influence the happening time of the five accident events above. In this paper, based on MAAP5 and LOCAs, the CPR1000 is simulated to analyze the influences of MAAP5’s sensitivity parameters reflecting phenomenological uncertainty on the accident process, which is aimed to find out the sensitivity parameters associated to the five important accident events and build the database between these sensitivity parameters and five accident events’ happening time. Then, based on the research above, a preliminary approach to optimize the MAAP5’s accidents simulation is introduced, which is realized by adjusting sensitivity parameters. Finally, the application of this research will be showed in a severe accident management system developed by us. The research results offer great reference significance for the severe accident simulation and prediction in MAAP5.
Research of the Node Ordering Technique According to the Shortest Fault Tree
Authors: Tang Guofeng , Zhang Jingyu , Gao Wei , Zhang Qinfang
Published: October 2017

Zero-suppressed Binary Decision Diagram (ZBDD) algorithm is an advanced method in fault tree analysis, which is developing quickly in recent years and being used in the development of the Probabilistic Safety Assessment (PSA) Quantification Engine. This algorithm converts a fault tree to a ZBDD structure, solves the minimal cut sets and calculates the top node unavailability. The ordering of the basic events and logical gates is the core technique of the ZBDD algorithm, which determines the efficiency of the ZBDD conversion and the size of the ZBDD structure. A variable ordering method based on the structure of the fault tree is developed in this paper, which gives a better basic events order by compressing the fault tree; meanwhile, the method offers a logical gates order. The nodes order derived from this method can accelerate the ZBDD conversion obviously.
The Research of Function Reliability in Passive System Based on Fuzzy Theory
Authors: Shi Qi , Tao Zhou , Dong-liang Ma , Yu Li , Guang-ming Jiang
Published: October 2017

A new method has been presented in this study, for evaluating the functional reliability of passive systems based on the fuzzy method. A case study has been presented, in order to analyze the passive residual heat removal system (PRHRS) of AP1000. The RELAP5 is used to analyze the thermal-hydraulic behavior. However, the failure criterion may not be suitable, since the local saturated boiling temperature is not only related to the pressure, but also related to the local surface conditions. In addition, it is not reasonable to categorize the failure, based on a single numerical value. There are no essential differences among these values which are near the critical numerical value. Hence, it is different to use the classical theory for calculating the failure probability. However, the fuzzy method has been founded to be appropriate for such a case. The failure criteria parameter is regarded as a fuzzy number, which can be described by a membership function. Finally, the failure probability of PRHRS is obtained as 7.26×10−8 which is well within the range of classical failure probability.
Numerical Simulation of Flow Instability in Vortex Diodes
Authors: Wang Junrong , Xiao Qi , Ke Hanbing , Hu Xu , Li Shaodan , Wei Zhiguo
Published: October 2017

A vortex diode is used as a highly reliable check-valve in nuclear applications, where it mainly benefits from the intrinsic properties of no moving parts and no leakage. Its basic principle is similar to the diode in an electric circuit. The typical structure of a vortex diode consists of a chamber with axial and tangential ports. When the fluid is injected through the axial port, a simple radial flow in the chamber leads to a relatively low flow resistance. On the other hand, in the reverse flow mode, a strongly swirling vortex can be set up in the chamber, resulting in a very high flow resistance.

Several experimental studies found vortex-induced vibration of a vortex diode in the reverse flow mode, where it indicated that the flow was unstable in the vortex diode. This phenomenon may affect the reliability of the vortex diode. However, the mechanism has not been investigated systematically and profoundly. In this paper, 3-D simulations are carried out to help understand the related flow characteristics in the vortex diode. Standard k-ε model was selected for forward flow, while Reynolds stress model was selected for reverse flow. We have found that the results from transient simulations are in good agreement with experimental data. The transient simulations also capture the periodic pressure fluctuation in the vortex diode. Vortex diodes with different structures and geometrical parameters are simulated at different Reynolds number conditions. It is found that the characteristics of the pressure fluctuation are determined by the structure parameters and working conditions of the vortex diode. The flow instability is mainly caused by the asymmetry of the vortex diode. The work presented in this paper will be useful to give better understanding of flows in vortex diodes and to provide some guidance for optimizing the vortex diode.
Accident Tolerance Cladding Thermal Hydraulic Simulation and Oxidation Kinetic Sensitivity Analysis
Authors: Wang Jun , Mccabe Mckinleigh , Haskin Troy Christopher , Wu Yingwei , Su Guanghui , Corradini Michael L.
Published: October 2017

Since the Fukushima accident, an increasing number of studies have considered Accident Tolerant Fuels and cladding properties that can improve the safety performance during an accident in a nuclear reactor. In a research project funded by the US Department of Energy, a group at the University of Wisconsin-Madison focused on modeling the fuel performance for beyond design basis accident conditions. The MELCOR 1.8.6-YR program was used to model the SURRY power station for Zircaloy compared to FeCrAl alloy cladding. The model used an oxidation model for FeCrAl developed by Oak Ridge National Lab (ORNL). The ORNL oxidation model was also used to analyze the fuels’ performance, and these results were compared with the results obtained with the Idaho National Lab (INL) oxidation model for FeCrAl. Results showed that the ORNL model predicted a larger amount of hydrogen generated than the INL model. More experimental data may be needed to better understand the FeCrAl oxidation kinetic mechanisms and perfect oxidation models for this alloy.
Component and Integration Test of an FPGA-Based PWR Protection Sub-System Using UVM
Authors: Shui Xuanxuan , Wu Yichun , Zhou Junyi , Cai Yuanfeng
Published: October 2017

Field programmable gate arrays (FPGAs) have drawn wide attention from nuclear power industry for digital instrument and control applications (DI&C), because it’s much easier and simpler than microprocessor-based applications, which makes it more reliable. FPGAs can also enhance safety margins of the plant with potential possibility for power upgrading at normal operation. For these reasons, more and more nuclear power corporations and research institutes are treating FPGA-based protection system as a technical alternative.

As nuclear power industry requires high reliability and safety for DI&C Systems, the development method and process should be fully verified and validated. For this reason, to improve the application of FPGA in NPP I&C system, the specific test methods are critical for the developers and regulators. However, current international standards and research reports, like IEC 62566 and NUREG/CR-7006, which have demonstrated the life circle of the development of FPGA-based safety critical DI&C in NPPs, but the specific test requirements and methods which are significant to the developers are not provided.

In this paper, the whole test process of a pressurized water reactor (PWR) protection sub-system (Primary Coolant Flow Low Protection, Over Temperature Delta T Protection, Over Power Delta T Protection) is described, including detail component and integration tests. The Universal Verification Methodology (UVM) based on System Verilog class libraries is applied to establish the verification test platform. All these tests are conducted in a simulation environment. The test process is driven by the test coverage which includes code coverages (i.e., Statement, Branch, Condition and Expression, Toggle, Finite State Machine) and function coverage. Specifically, Register Transaction Level (RTL) simulation is conducted for Component tests, while RTL simulation, Gate Level simulation, Timing simulation and Static timing analysis are conducted for the integration test. The issues (e.g., the floating point calculation, FPGA resource allocation and optimization) arose in the test process are also analyzed and discussed, which can be references for the developers in this area.

The component and integration tests are part of the Verification and Validation (V&V) work, which should be done by the V&V team separated from the development team. The testing method could assure the test results reliable and authentic. It is practical and useful for the development and V&V of FPGA-based safety DI&C systems.
Experimental Study of Uniform and Non-Uniform Transverse Heat Flux Distribution Effect on the Onset of Nucleate Boiling
Authors: Al-Yahia Omar S. , Kim Taewoo , Jo Daeseong
Published: October 2017

An experiment study is conducted to investigate the effect of transverse power distribution on the Onset of Nucleate Boiling (ONB) through a one-side heated narrow rectangular channel. Two test section are used to perform the experiment; uniform and non-uniform heated suction. The demineralized water is flowing in upward direction through the coolant channel with a thickness of 2.35 mm, a width of 54 mm, and a length of 300 mm. The experiment is carried out under different thermal power (0.5 kW – 6.5 kW) for the both test section. As well as, a wide variety of inlet subcooling and flow velocity are used as; 65−35 °C and 0.1–1.0 m/s, respectively. The wall temperature distribution of the heated plate is measured by 10 TCs for the uniformly heated test section, and 20 TC for the non-uniformly heated section. On the other hand, the ONB location is visualized via high speed camera, in which the ONB occurs near the edges for the non-uniformly heated section and occurs at the center of the heated surface for the uniformly power distribution. The results of the ONB heat flux and temperature in the non-uniformly heated section are compared against the one in the uniformly heated power. The results show the variety of the ONB location, ONB heat flux with the different power distribution. With the increase of the power, the ONB is shifted toward the inlet. On the other hand, the ONB for the non-uniform power distribution occurs near the edges at power lower than that the one in the uniformly power distribution. Also, the results are compared against the available correlations, such as Bergles and Rohsenow (1965), Jens and Lottes (1951), and Thom et al. (1965), as well as other experimental results done by several research institutes.
New Exploration on TMSR: Redesign of the TMSR Lattice
Authors: Zhao Jinkun , Si Shengyi , Chen Qichang , Bei Hua
Published: October 2017

Molten Salt Reactor (MSR) has been recognized as one of the Next Generation Nuclear Power systems. Most MSR concepts are the variants evolved from the ORNL’s Molten-Salt Breeder Reactor (MSBR) which employs Molten-Salt as both fuel and coolant, and normally graphite is used as moderator. Many evaluations have revealed that such concepts have low breeding ratio and might present positive power coefficient. Facing these impediments, TMSR (Thorium Molten Salt Reactor) with redesigned lattice is proposed in this paper. Based on comprehensive investigation and screening, important lattice parameters including molten salt fuel composition, solid moderator material, lattice size, structure and lattice P/D ratio (lattice pitch to channel diameter) are redesigned.

In this paper, new composition of fuel salt without BeF2, which is also recommend for Molten Salt Fast Reactor (MSFR), is employed instead of LiF-BeF2-ThF4-UF4 adopted in the design of single fluid MSBR. The new fuel composition makes TMSR to benefit from the increased solubility for actinides (e.g. Th4, UF4). Moreover, due to the decent slowing-down power and neutron multiplication effect by (n,2n) reaction of beryllium, BeO is employed as moderator to improve neutron economy instead of graphite. To avoid corrosion on the one hand, Ceramic cladding (e.g. SiC) is introduced to separate the flowing liquid fuel and fixed solid moderator. More importantly, ceramic cladding is capable of maintaining a stable flow channel and supporting the core structure on the other hand. Concerning neutron spectrum, P/D ratio is an important parameter indicating the volume fraction of fuel in the lattice. In order to obtain a suitable spectrum for better breeding and safety features, lattice size and P/D ratio have been optimized for TMSR. Furthermore, since online reprocessing capability and refueling control are key parameters influencing depletion behavior which concerns the sustainability of the reactor system, these issues are also discussed in this paper.

Simulation of the redesigned TMSR system is performed to evaluate the outcomes of the lattice parameters optimization. SONG/TANG-MSR codes system is applied in the simulation, which is independently developed by Shanghai Nuclear Engineering Research & Design Institute (SNERDI). A traditional core model with LiF-BeF2-ThF4-UF4 fuel and graphite moderator is also evaluated by the codes for reference. Thanks to the optimized lattice parameters and as consequences of the redesigned lattice, TMSR has achieved a high breeding ratio close to 1.13. With a proper reprocessing and refueling strategy, the doubling time of TMSR can be shortened to about 15 years. Meanwhile a negative power coefficient is still maintained. Based on this lattice design, TMSR will have excellent performance on safety and sustainability.
Control Strategy of Soluble Boron-Free Operation for a Small PWR
Authors: Bi Guangwen , Tang Chuntao , Yang Bo
Published: October 2017

Elimination of soluble boron will be a challenge to reactor operation for PWR. This paper is to promote a control strategy of soluble boron-free operation for a small PWR, through selection of burnable poison (BP), BP loading and control rod loading, based on the reactivity balance and manage requirement. The analysis for on-power operation and shutdown condition indicated that this strategy could be suitable for long-term and short-term reactivity and power distribution control for soluble boron-free operation.