bluuebunny/crossref_metadata_2025 huggingface dataset:
An Improved Algorithm for Potential Accident Consequence Assessment at NPPs Based on PAVAN
Authors: Qian Chen , Liang Zhai , Yuhao Yin , Yunliang Tao
Published: October 2017

PAVAN is an atmospheric dispersion program for evaluating design basis accidental releases of radioactive materials from nuclear power plant. It was developed by Pacific Northwest Laboratory on the basic of the atmospheric dispersion models described in RG 1.145 by NRC (U.S. Nuclear Regulatory Commission). Using the joint frequency of wind direction, wind speed and atmospheric stability, the atmospheric relative concentration values for the exclusion area boundary and outer LPZ boundary of nuclear power plant are calculated and given by the program. Once the program was introduced, it has been widely used in the radioactive accident consequence assessment, especially in the FSAR (Final Safety Analysis Report) and Report of EIA (Environmental Impact Assessment) of NPPs in China.

The theory basis and general method of PAVAN is introduced in this paper. And specialty of the X/Q points based on joint frequency data is discussed. The envelope algorithm of PAVAN is also introduced and discussed.

The paper presents an improved algorithm based on PAVAN which uses the hourly meteorological data as input instead of joint frequency data. In this algorithm, the size of X/Q points is related to the quantity of the hourly meteorological data. When the quantity is large enough, e.g. 17520 sets of hourly meteorological data in two years, the envelope curve for X/Q points fit more exactly than PAVAN.

Using the observed meteorological data, the improved algorithm is compared with PAVAN. The result proves that the former is more accurate. In general, the improved algorithm is relatively conservative. In some situation, the conservativeness is not certain. The factors which result in the uncertainty are deeply discussed.

Further optimized are performed by the algorithm. The number of points to seek in envelope curve fitting is set to be dynamic and be a quarter of total number of X/Q points to be fitted. The result shows that increasing the number of points to seek in the iteration process of envelope curve fitting will lead to more conservative X/Q values.

Additionally, the optimized algorithm provides X/Q value of 50% probability level for overall site. The value is not relatively conservative. From the standpoint of statistical probability, it is more realistic and is acceptable for potential accident consequence assessment. Especially, when X/Q value of 95% probability level for overall site is too conservative to accept, the value of 50% probability level can be used to replace the conservative value.
Thermodynamic Performance of 5MW Space Nuclear Power System Based on Liquid Metal Rankine Cycle
Authors: Cheng Xian-Wei , Zhang Hao-Chun , Li Xiao-Qi , Zhang Chen-Xu , Zhao Guang-Bo
Published: October 2017

In this paper, the losses of energy and exergy of 5WM Rankine space nuclear power system were analyzed and compared through thermodynamic analysis. The power system includes a liquid metal Rankine power conversion system coupled with a lithium cooled fast reactor. Meanwhile, a reheat process is contained to enhance the efficiency of the system. The aim of this paper is to evaluate the sites with largest losses of energy and exergy upon the thermodynamic analysis of the system. Hence, this study focus on the positions and loss types of the efficient energy in such a liquid metal Rankine space nuclear power system. In addition, the effects of varied space environment state on this analysis was presented through investigating the influences of ambient temperature on the energy losses and efficiencies in every parts. At last, the characteristics of four Rankine space nuclear power systems with different liquid metals were studied.
Study of Leakage and Explosion of Hydrogen and Blast Wall Failures in an Offshore Platform
Authors: Kumar Chenthil , Rajendran Vishnu , Kumar Anil , Tripathi Amita
Published: October 2017

Study of gaseous explosions and their effects on structures is helpful in designing offshore platforms. Specifically, reliable methods for the prediction of overpressures in offshore explosions are highly useful and are extensively researched. The selection and/or development of means of prevention, control and mitigation of explosions often depends on the comprehensive analysis of their probability of incidence and damage potential. This involves a number of factors, such as explosive gas leak size, location, composition, wind direction, and characteristics of probable ignition. This paper presents a 3D transient CFD based analysis tool for such purposes and the results of some simulations done using it. The first set of simulations is a validation exercise, which involves hydrogen leakage and explosion, and the computational results are compared with the experimental data. The second set of calculations involved simulation of a hydrogen gas leakage scenario on an offshore platform, followed by explosion studies for different scenarios to find the effect of various guidelines for the initial conditions in the reactive cloud. These results show that, the maximum explosion pressure occurs when stoichiometric initial mixture conditions are applied in the dispersed flammable region. The worst case explosion scenario thus observed has maximum over pressures and maximum blast wall displacement of about 18 to 20 times higher than the base case explosion.
The Need for Integrated Cybersecurity and Safety Training
Authors: Gupta Deeksha , Bajramovic Edita , Hoppe Holger , Ciriello Antonio
Published: October 2017

Companies involved in the nuclear energy domain, like component and platform manufacturers, system integrators and utilities, have well established yearly trainings on Nuclear Safety Culture. These trainings are typically covered as part of the annual quality assurance-related refresher trainings, introductory courses for new employees, or indoctrinations of temporary staff. Gradually, security awareness trainings are also addressed on a regular basis, typically with a focus on IT, the daily office work, test bay or construction site work environment, and some data protection and privacy-related topics. Due to emerging national nuclear regulation, steadily but surely, specialized cybersecurity trainings are foreseen for integrators and utilities.

Beyond these safety, physical security and cybersecurity specific trainings, there is a need to address the joint part of these disciplines, starting from the planning phase of a new Nuclear Power Plant (NPP). The engineers working on safety, physical protection and cybersecurity, must be aware of these interrelations to jointly elaborate a robust I&C architecture (defense-in-depth, design basis events, functional categorization and systems classification) and a resilient security architecture (security by design, security grading, zone model or infrastructure domain, security conduits, forensic readiness, Security Information and Event Management).

This paper provides more in-depth justification of when and where additional training is needed, due to the ubiquitous deployment of digital technology in new NPPs. Additionally, for existing NPPs, the benefits of conveying knowledge by training on specific interfaces between the involved disciplines, will be discussed.

Furthermore, the paper will address the need of focused training of management stakeholders, as eventually, they must agree on the residual risk. The decision-makers are in charge of facilitating the inter-disciplinary cooperation in parallel to the allocation of resources, e.g. on security certifications of products, extended modeling-based safety and security analyses and security testing coverage.
Diffusion of Fission Gas in Uranium Dioxide: A First-Principles Study
Authors: Zhao Qiang , Zhang Zheng , Li Yang , Ouyang Xiaoping
Published: October 2017

Uranium dioxide (UO2) is the typical fuel that is used in nuclear fission reactor, fission gas are produced during and after the reactor operation, and the fission gas have a significant impact on the performance of UO2 in reactor. In this paper, we investigated the effects of the fission gas on the performance of UO2 by using the first-principles calculation method based on the density functional theory. The results are that, the volume of UO2 increased when there is a fission gas atom enter in UO2 supercell; fission gas prefer to occupy the octahedral interstitial site over the uranium vacancy site and the oxygen vacancy site, and the oxygen vacancy site is the most difficult occupied site due to the formation of an oxygen vacancy is more difficult than that of the uranium vacancy; our results of the UO2 elastic constants are in good agreement with other simulation results and experimental data, and the fission gas atoms make the ductility of UO2 decreased. Our works may shed some light on the development of the UO2 fuel and the spent fuel reprocessing.
Research and Development of Protective Action Analysis System for the Public Based on Operation Intervention Level in Reactor Accident
Authors: Xu Xiaoxiao , Zhang Jiangang , Yang Yapeng , Feng Zongyang , Jia Linsheng
Published: October 2017

Operation intervention level (OIL) is a type of action level that can be used immediately and directly (without further assessment) to determine the appropriate protective actions and other response actions on the basis of an environmental measurement. OIL is used to implementation actions to protect the public due to the severe conditions at a nuclear power plant. This paper presents an integrated solution that integrates remote access and remote support over the Internet. The system real-time acquire the data of source term, meteorological, off-site monitoring points, monitoring vehicles, execute operation intervention level calculation, protective action decision-making to public based on data from database or input by user. The system simultaneous input multiple monitor data, and protection of the public action based on the results of multiple OIL decisions. The system display off-site monitoring points, protection of the public area of operations and administrative village using WebGIS.
Void Reactivity Effect Research on Liquid Metal Cooled Fast Reactor
Authors: Lyu Yang , Liang Xiao
Published: October 2017

In the fourth generation of advanced nuclear power systems, the liquid metal cooled fast reactor plays a more and more important role, such as SFR, LFR and ADS system with LBE coolant. Void reactivity effect means bubbles produced in the core area will induce the change of reactivity. And this reactivity will affect the safety of the reactor. Through investigation and comparison of several liquid metal cooled fast reactors in the nuclear industry, this paper studies bubbles in different positions and partial voiding of the active zone inside the core and fuel assemblies with Monte Carlo core physics calculation method and then concludes the main influencing factors of void reactivity coefficient. The results can provide reference for the follow-up research and development of new type liquid metal fast reactor core design.
Implementation of Fast Searching Method for the Cross-Section Usage in HTGR Simulators
Authors: Auve Benjamin , Wei Chunlin , Sui Zhe , Sun Jun
Published: October 2017

The modular High-Temperature Gas-cooled Reactor (HTGR) is one of the six generation IV advanced nuclear reactors. With the final purpose of operator training and licensing, the engineering simulation system (ESS) has been studied to model the pebble-bed type reactor core and has been successfully implemented into the full scope simulator of HTR-PM.

As stated in corresponding industrial standards, one important feature of the nuclear power plant simulator is real-time calculation, and the other one is simulation results with high fidelity (compared to design parameters or operational data in different stages). In ESS, each macro cross-section was in the form of polynomial by function of several variables (like burn-up, buckling, temperatures), the expression of which was finalized by multivariate regression analysis from large scattered database generated by the VSOP. Since the polynomial is explicit and prepared in advance, the macro cross-sections are quickly calculated in running ESS. However, some variables (such as temperature) in HTGR are in larger scope so that the polynomial is not easy to meet full range accuracy. One normal idea is to optimize the expression of polynomial, while another means was proposed and tested in present paper.

Other than focusing on the polynomials, a new method, called the fast searching, was described to significantly improve the accuracy of macro cross-section calculation while it was also fast to maintain the real-time feature. Instead of setting up a regression polynomial from the large cross-section database, the fast searching method treated the database as scatted points in the multi-dimension space, and aimed to locate the target position of unknown macro cross-section by fast searching and interpolating. Searching was to find the neighbouring database points around the target point in the multi-dimension space, which naturally improved the accuracy. While interpolating was to predict the macro cross-section of target point based on those neighbouring database points. To keep the searching and interpolating fast, the original database of macro cross-sections was analysed. A series of searching and interpolating methods have been described, programmed, tested and compared to find appropriate methods to calculate all the macro cross-sections in limited time cost. Finally, the fast searching method and its program was implemented into ESS to show better performances.
A Simulation of Small Break Loss of Coolant Accident (SB-LOCA) in AP1000 Nuclear Power Plant Using RELAP5-MV
Authors: Yousif Eltayeb , Zhijian Zhang , Zhao-fei Tian , Mustafa A. M.
Published: October 2017

To ensure effective operation of nuclear power plants, it is very important to evaluate different accident scenarios in actual plant conditions with different codes. In the field of nuclear safety, Loss of Coolant Accident (LOCA) is one of the main accidents. RELAP-MV Visualized Modularization software technology is recognized as one of the best estimated transient simulation programs of light water reactors, and also has the options for improved modeling methods, advanced programming, computational simulation techniques and integrated graphics displays. In this study, transient analysis of the primary system variation of thermo-hydraulics parameters in primary loop under SB-LOCA accident in AP1000 nuclear power plant (NPP) is carried out by Relap5-MV thermo-hydraulics code. While focusing on LOCA analysis in this study, effort was also made to test the effectiveness of the RELAP5-MV software already developed. The accuracy and reliability of RELAP5-MV have been successfully confirmed by simulating LOCA. The calculation was performed up to a transient time of 4,500.0s. RELAP5-MV is able to simulate a nuclear power system accurately and reliably using this modular modeling method. The results obtained from RELAP5 and RELAP5-MV are in agreement as they are based on the same models though in comparison with RELAP5, RELAP5-MV makes simulation of nuclear power systems easier and convenient for users most especially for the beginners.
Research on Resistance Characteristic in Blanket Assembly of China Experimental Fast Reactor by Experiment and Numerical Simulation
Authors: Bing Yang
Published: October 2017

Experiment investigation and numerical calculation were carried out on frictional resistance characteristics of single-phase flow for China Experiment Fast Reactor (CEFR) blanket assembly.By using the CFD analysis, the suitable size of Labyrinth throttling at the base pin to meet the requirements of design pressure drop were selected.Then resistance characteristic experiment were carried on with the suitable size. It is show that numerical results agrees well with experiment results.The flow field was greatly affected by the labyrinth structure of the throttling element in CEFR blanket assembly and resistance are mainly fouces on the CEFR’s Labyrinth throttling at the base pin. It is show that using this method can effectively reduce experiment times, shorten research time and reduce the cost..
Considerations on a Validation Matrix for System Analysis Codes of Pebble Bed Modular HTRs
Authors: Gou Feng , Chen Fubing , Dong Yujie
Published: October 2017

The modular High Temperature Reactor (HTR) experienced substantial development over the past two decades in China. At present, the 10 MWth experimental reactor HTR-10 is under operation and the 200 MWe demonstration power plant HTR-PM is in its commissioning phase. Both of the HTR-10 and the HTR-PM are pebble bed HTRs designed by the institute of nuclear and new energy technology (INET) of Tsinghua University.

In the design and analysis of modular HTRs, system analysis codes play an important role in predicting the reactor behaviour under both normal operation and accident conditions. According to China’s nuclear safety regulations, codes and models employed in the HTR system analysis must be verified and validated in conformance with a specified procedure. Thus, a flowchart is recommended in this paper for the validation of system analysis codes. As the first step of the flowchart, a phenomena identification and ranking table (PIRT) process has been performed on the subject of the HTR-PM thermal hydraulics and accident analysis. In this paper, the PIRT process is introduced and the PIRT results are presented. With the purpose of developing a validation matrix, test data have been collected mainly from the operation history and various tests of the HTR-10. On such basis, some considerations on developing a validation matrix are discussed to benefit the code validation work.
Natural Convection: A Case of Simple Harmonics
Authors: Usman Shoaib
Published: October 2017

Natural convection has been an area of intense research since it was discovered over 100 years ago. However, a phenomenological explanation for the onset of natural convection is still not available. The role of surface roughness in various thermal hydraulic phenomena is widely accepted; however, the scale of roughness is missing from Rayleigh number (Ra) which is commonly used to predict the onset of natural convection. Using the mass-spring analogy, an analytical derivation is presented to accurately characterize the thermal system and the interplay of gravity, viscosity and scale of roughness responsible for the onset of natural convection. The necessary conditions for the onset of natural convection are given in terms of a new dimensionless number βgΔTεH2α2 which captures the effect of surface roughness. Using the mass-springs analogy, the natural frequency of the convection system is shown to depend on thermal and kinematic property of the fluid and the scale of roughness. These results are applicable to many natural system and engineering designs.
ROSA/LSTF Test on Nitrogen Gas Behavior During Reflux Cooling in PWR and RELAP5 Post-Test Analysis
Authors: Takeda Takeshi , Ohtsu Iwao
Published: October 2017

An experiment focusing on nitrogen gas behavior during reflux cooling in a pressurized water reactor (PWR) was performed with the rig of safety assessment/large scale test facility (ROSA/LSTF) at Japan Atomic Energy Agency. The test conditions were made such as the constant core power of 0.7% of the volumetric-scaled PWR nominal power and the primary pressure of lower than 1 MPa, unlike a previous related test with the LSTF. The steam generator (SG) secondary-side collapsed liquid level was maintained at a certain liquid level above the SG tube height. Nitrogen gas was injected stepwise into each SG inlet plenum at a certain constant amount. The primary pressure and the SG U-tube fluid temperatures were greatly dependent on the amount of nitrogen gas accumulated in the SG U-tubes. Nitrogen gas accumulated from the outlet towards the inlet of the SG U-tubes. Non-uniform flow behavior was observed among the SG U-tubes with nitrogen gas ingress. The RELAP5/MOD3.3 code indicated remaining problems in the predictions of the primary pressure and the SG U-tube fluid temperatures after nitrogen gas inflow.
Development of Numerical Simulation Method to Evaluate Heat Transfer Performance of Air Around Fuel Debris: Part 1 — Effect of the Debris Shape
Authors: Yamashita Susumu , Uesawa Shinichiro , Yoshida Hiroyuki
Published: October 2017

Fuel debris retrieval in atmospheric conditions is proposed as one candidate in the Fukushima Daiichi NPS since it can suppress generation of the contaminated water. However, there are unknowns and uncertainties to evaluate the air cooling performance of the fuel debris. Solving the problems, we have been developing the air-cooling analysis method for accumulated debris in the pedestal. A numerical simulation code JUPITER, which is based on a CFD technics and treats a multi-phase, multi-component thermal-hydraulics, is used as the basis of the method. We preliminary evaluated the effect of debris shapes and heat source conditions on free convection and heat transfer in the simplified structure of the pedestal region using the JUPITER. As a result, it was confirmed that the flow pattern of the free convection in the pedestal clearly differs with debris shapes and a correlation exists between the temperature distribution and the velocity fluctuation around the debris surface.
Analysis and Application of Trace H2 and CH4 in Cycling Process Gas of Fusion Reactors by Gas Chromatography
Authors: Li-ling Yang , Hong-guang Yang , Zhen-xing Liu , Ti-rui Xia
Published: October 2017

It’s necessary to establish a high-precision analytical method of Gas Chromatography (GC) for achieving the rapid detection of trace hydrogen and methane in cycling process gas of fusion reactors. The analysis of H2, CH4 with concentrations from 0.5ppm to 5000ppm were detected through the application of Discharge Ionization Detector (DID) in high purity helium and the column of molecular sieve, whose parameters was 15m×0.53mm×50μm.

The studies have shown that the retention time of H2, CH4 were 36s and 71s, respectively, their relatively standard deviation (RSD) of area responses were severally below 1.0% and 0.5%, which illustrated good repeatability. Separately, the detection limit (DL) were no more than 10ppb and 2ppb of H2 and CH4 with high column efficiency, while the column temperature was 40°C and the flow rate was 15mL/min. Meanwhile their values of R2 were all larger than 0.999, which included a better linearity relationship between area response and concentration. This method has already applied to the on-line measurement of molecular sieve adsorption, the analysis of electrochemical hydrogen pump and methane decomposition, etc.
Effect of Fuel Meat Thickness on the Non-Uniform Irradiation-Induced Thermo-Mechanical Behavior in Monolithic UMo/Al Fuel Plates
Authors: Yan Feng , Zhao Yunmei , Ding Shurong
Published: October 2017

Monolithic UMo/Al fuel plates have a promising prospect in the advanced research and test reactors because of their high equivalent uranium density and stable irradiation performance. They will undergo complicated in-pile thermo-mechanical behavior, which may affect their lifetime and the safety of nuclear reactors. It is necessary to capture the effect of fuel meat thickness on in-pile thermo-mechanical behavior evolution in the fuel plates in order to realize their optimized design and control their service safety. In this study, considering a non-uniform irradiation condition, several 3D finite element models are built to simulate the in-pile behavior in different-thickness UMo/Al plates. The user subroutines are programmed based on the thermo-mechanical constitutive relations and stress update algorithms of the constituent materials. The influences of fuel meat thickness on the temperature field, the main deformations and the interfacial normal stresses are numerically investigated based on the obtained results.
Heat Transfer Characteristics of Downward Facing Hot Horizontal Surfaces Using Mist Jet Impingement
Authors: Yadav Ashutosh Kumar , Sharma Parantak , Sharma Avadhesh Kumar , Modak Mayank , Nirgude Vishal , Sahu Santosh K.
Published: October 2017

Impinging jet cooling technique has been widely used extensively in various industrial processes, namely, cooling and drying of films and papers, processing of metals and glasses, cooling of gas turbine blades and most recently cooling of various components of electronic devices. Due to high heat removal rate the jet impingement cooling of the hot surfaces is being used in nuclear industries. During the loss of coolant accidents (LOCA) in nuclear power plant, an emergency core cooling system (ECCS) cool the cluster of clad tubes using consisting of fuel rods.

Controlled cooling, as an important procedure of thermal-mechanical control processing technology, is helpful to improve the microstructure and mechanical properties of steel. In industries for heat transfer efficiency and homogeneous cooling performance which usually requires a jet impingement with improved heat transfer capacity and controllability. It provides better cooling in comparison to air. Rapid quenching by water jet, sometimes, may lead to formation of cracks and poor ductility to the quenched surface. Spray and mist jet impingement offers an alternative method to uncontrolled rapid cooling, particularly in steel and electronics industries. Mist jet impingement cooling of downward facing hot surface has not been extensively studied in the literature. The present experimental study analyzes the heat transfer characteristics a 0.15mm thick hot horizontal stainless steel (SS-304) foil using Internal mixing full cone (spray angle 20 deg) mist nozzle from the bottom side. Experiments have been performed for the varied range of water pressure (0.7–4.0 bar) and air pressure (0.4–5.8 bar). The effect of water and air inlet pressures, on the surface heat flux has been examined in this study. The maximum surface heat flux is achieved at stagnation point and is not affected by the change in nozzle to plate distance, Air and Water flow rates.
Hierarchical Modeling of GO-FLOW Models for Online Risk Monitoring of Nuclear Power Plants
Authors: Xinyu Dai , Ming Yang , Bowen Zou , Hongxing Lu
Published: October 2017

As a clean and safe energy, nuclear power plays an important role in solving the increasingly prominent environmental problems in the world. Researches on the reliability analysis and risk assessment for the safety operation of Nuclear Power Plants (NPPs) have been paid much more attention due to several accidents in the past decades.

GO-FLOW is a methodology for analyzing the reliability of dynamic systems with common cause failures (CCFs), man-machine failures, and time-dependent mission issues. The traditional GO-FLOW methodology presents a set of operators each of which stands for a relatively basic reliability logic. Complex behavior of a physical component may be simply described by one operator or a combination of operators. Therefore the GO-FLOW chart of a system is much more similar to its physical structure diagram than other reliability models such as Fault Tree.

In a previous study by authors, a Living PSA technology based on GO-FLOW was proposed which utilizes a group of interrelated GO-FLOW operators to describe the potential change of states of a physical component between operation, standby, failure, preventive and corrective maintenance. The proposed Living PSA technology is intended to provide a convenient way for the safety engineers to update the PSA models for risk assessment. However, it will make the PSA models complex, quite different from their physical structure diagram and therefore difficult to check.

This paper presents a hierarchical modeling method of GO-FLOW. The GO-FLOW model of a system will be organized at different layers, including system layer, structure layer and component layer. The proposed Living PSA technology is only applied at component layer and different level of layers are interrelated by introduction of new GO-FLOW operators and equivalent relations. Thus, the GO-FLOW models at the higher layers are more similar to their physical structures of system for understanding and checking. A bottom-up calculation algorithm for the reliability analysis is also presented. An Auxiliary Feedwater System (AFWS) of PWR plant is taken as a study case to show to modeling and analyzing a system for reliability and risk assessment based on this new technology.
Research Progress of Red Oil Explosion Accidents in Nuclear Fuel Reprocessing Plant
Authors: Zhang Chunlong , He Hui , Zhao Shangui , Song Fengli , Liu XinHua
Published: October 2017

Since Westinghouse Savannah River Company (WSRC) of America first applied PUREX process in 1954, PUREX process is always the top priority in nuclear fuel reprocessing plant. And this process is based on liquid to liquid extraction with TBP as the extractant. TBP is irreplaceable in the development of PUREX process in nuclear fuel reprocessing, its advantages are well recognized. However TBP does have some disadvantages such as formation of red oil, which will appear in the system of high nitric acid concentration and heavy metal nitrate, once the red oil forms, it can lead a exothermic runaway decomposition in reasonable conditions, such as exceeding a certain temperature (typically 130°C) or high acid concentration. If gas products and energy released from the decomposition reaction could not be exported in time, it will lead to vessel overpressure and caused violent explosion accidents. By now, it has happened 6 times so-called red oil explosion accidents worldwide, resulting in different degrees of equipment and construction damage and environmental contamination. From 1953 to now, research related to red oil has never stopped. WSRC, Hanford Company, Oak Ridge National Laboratory and Los Alamos National Laboratory of America have conducted many studies, as well as some research institutions from Russia, UK, France and India. Defense Nuclear Facilities Safety Board of America issued a technical report in 2003, preventive measures for red oil explosion were established in this report, and these measures provided good practice experience and reference for other countries, and the temperature condition (⩽130°C)and nitric acid concentration (⩽10M)for preventing red oil explosion are employed in some countries which has built the reprocessing plant. Nevertheless, research conclusions and knowledge of red oil vary from country to country. Especially, Kumar and Smitha etc. conducted several experiments in adiabatic condition in recent years, and investigation on stability of TBP - nitric system was made, the results indicated that the red oil runway reaction will happen even in lower temperature and lower nitric acid concentration in contrast with the reported value, and they thought it would need a further study to assess the validity of present preventive measures, and to rebuild the safety limits for preventing red oil explosion in the operation of nuclear fuel reprocessing plants. In this paper, related research results of red oil explosion accidents were combed, and the characters of study work of different periods were summarized, and definition, formation conditions of red oil, pathway of runaway reaction, control and preventive measures for preventing red oil explosion of different countries were analyzed and compared, as well as the new viewpoints of recent literatures. And some research ideas for future investigation based on present work were also proposed.
A Research Reactor Core Design for Advance Neutron Source
Authors: Wu Zeyun
Published: October 2017

This paper presents the recent neutronics analysis results of a proposed LEU-fueled research reactor. The main goal of the research reactor is to provide advanced neutron source with a particular emphasis on high intensity cold neutron sources. A tank-in-pool type reactor with an innovative horizontally split compact core was developed in order to maximize the yield of the thermal flux trap in the reflector area. The reactor was designed with 20 MW thermal power and 30-day operating cycle. For non-proliferation purposes, the LEU fuel (U3Si2-Al) with 19.75 wt.% enrichment was used. The estimated maximum thermal flux of the reactor is ∼5×1014 n/cm2-s. The total peaking factor of the start-up (SU) core is ∼2.5. The calculated brightness of the cold neutron source (CNS) demonstrates the superiority of the cold neutron performance of the design.
Analysis of Cyber Security Incident in South Korean Nuclear Power Plant and Cyber Security Solutions for Nuclear Power Industrial Control System
Authors: Bian Yufang , Li Jing , Yin Baojuan
Published: October 2017

South Korean nuclear power plant information leakage incident occurred in 2014. A hacker on the Internet issued four compressed files of Korean nuclear power plant, and threatened the nuclear power plant to stop related units. This paper reviews the cyber security incident and analyses the cause of the event. Then, it analyzes the security risks of industrial control system in nuclear power plants and proposes some suggestions to strengthen nuclear power cyber security.
Development of the Severe Accident Evaluation Method on Second Coolant Leakages From the PHTS in a Loop-Type Sodium-Cooled Fast Reactor
Authors: Yamada Fumiaki , Imaizumi Yuya , Nishimura Masahiro , Fukano Yoshitaka , Arikawa Mitsuhiro
Published: October 2017

The loss-of-reactor-level (LORL) is a typical accident condition causing a severe accident (SA) in sodium-cooled fast reactors (SFRs). In the loop-type SFR Monju, important cause of the LORL is the second coolant leakage at the low elevation of the primary heat transport system (PHTS), which occurs in cold standby in a different loop from that of the first coolant leakage in rated power operation because of excessive declining of the sodium level. This study developed an evaluation method for the LORL with considering countermeasures to prevent LORL: i.e., pumping sodium up into reactor vessel (RV) and interruption of sodium flowing out by siphon effect.

To evaluate the effectiveness of the countermeasures, the transient behavior in the RV sodium level was analytically studied in representative accident sequences. The representative sequences with lowest sodium level were selected by considering combinations of possible coolant leakage positions.

The evaluation result clarified that the LORL can be prevented by conducting the above-mentioned countermeasures to maintain the RV sodium level sufficient for the operation of decay heat removal system, even after the second coolant leakage of PHTS.
Uncertainty Analysis for Source Term Evaluation of High Temperature Gas-Cooled Reactor Under Accident Conditions: Identification of Influencing Factors in Loss-of-Forced Circulation Accidents
Authors: Honda Yuki , Sato Hiroyuki , Nakagawa Shigeaki , Ohashi Hirofumi
Published: October 2017

One of the key elements in probabilistic risk assessment is identification and characterization of uncertainties. This paper aims to suggest a procedure to identify influencing factors for uncertainty in source term evaluation, which are important to risk of public dose. We propose the following five steps for the identification: (1) derivation of variable parameters by expansion of dynamic equation for the system and scenario to be investigated, (2) extraction of uncertainties in variable factors, (3) selection of influencing factors for uncertainty among variable parameters by analysis of existing knowledge and sensitivity study, (4) identification of influencing factors for uncertainty using expert judgement, and (5) integration of selected factors in the aforementioned steps and determination of input parameters in the uncertainty evaluation of the source term evaluation tools relevant to selected factors. The approach is applied to select factors for a risk dominant accident scenario in direct cycle High Temperature Gas-cooled Reactor (HTGR) plant. As a first step, this approach is tested to evaluation of fuel temperature in a depressurized loss-of-forced circulation (DLOFC) accident and failure of mitigation systems such as control rod systems in a representative HTGR plant from the view point of reactor dynamics and thermal hydraulic characteristics. The procedures (1) and (2) are trial to investigate in this paper. As a result, the variable parameters and the relevant uncertainties are successfully extracted in accordance with the suggested procedure. The result will be used for determination of input parameter and uncertainty distribution of these parameters in the uncertainty analysis of source term evaluation.
Design and Simulation of 10kW Faraday Cup for Ion Beam Current Measurement
Authors: Rawat Bharat Singh , Vala Sudhirsinh , Abhangi Mitul , Kumar Ratnesh , Chauhan Sanket
Published: October 2017

Faraday Cups (FCs) have been used since a long time to measure the beam current in particle accelerators. The charge collected by the beam stopper is used to measure the intensity of the beam current by an ammeter. Here, at IPR (Institute for Plasma Research) the Faraday cup is used to measure the deuteron beam current in 14 MeV neutron generator. The FC presently used is for 300 KeV beam and 500μA beam current whereas the new design is proposed for deuteron beam ranging from 300–400 KeV and 20mA beam current. The energy of Ion beam is usually very high in comparison to the work function of the material resulting in the emission of the secondary electrons which can escape from the cup and effect the current measurement. Simulations for the suppressor voltage required to suppress all the secondary electrons has been carried out using SIMION8.0 and LORENTZ. Materials like Copper, Molybdenum, Tantalum etc are commonly used in manufacturing of FCs and the present paper shows the inter-comparison between the use two FC material namely Copper and Molybdenum. The heat load deposited from the beam has been analytically simulated and proper cooling system has been suggested. Simulations for selection of FC material and heat load are needed to be carried out for the new FC and its cooling system, using analysis tools like SRIM, ANSYS WORKBENCH etc. The results of these analysis are reported in this paper. The final fabrication of the FC will be based on these simulations and analysis.
Scaling Analysis of Scrubber Pool and Cyclone for Filtered Containment Venting System Testing
Authors: Revankar Shripad T. , Song JinHo
Published: October 2017

During a severe accident when a corium is ejected from the reactor vessel into the containment, one of the mitigations strategy to contain radioactivity from discharging in to environment and to keep containment under design pressure is to use containment flirted venting system (CFVS). In this paper the key phenomena related to the operation and performance in the components of the CFVS system, pool venture scrubber, cyclone separator, particulate filter, and molecular sieve filters were identified. Based on these phenomena a scaling analysis was performed that was based on system level and local phenomena level scaling. For system level scaling boundary flow, mass flux, pressure and temperature were preserved. On the local phenomena scaling various phenomena were considered. Scaling analysis was also carried out for scrubber system, cyclone and filter. Utilizing a scaled down model from 50 nozzles in the prototype to three nozzles in the scaled model, the flow parameters for the model facility scrubber were obtained. Using these parameters, the governing non-dimensional parameters were obtained for prototype and model facility. The scaling ratios for all the relevant parameters are summarized in this paper.
Crack Growth Safety Assessment of Overlay Weld on the Ω Seal for Control Rod Drive Mechanism
Authors: Luo Jiacheng , Zhang Yong , Li Pengzhou , Luo Juan , Sun Lei
Published: October 2017

As a key device, the Control Rod Drive Mechanism (CRDM) is of great important for the safety operation of the nuclear reactor, where the pressure shell of CRDM is exposed to the primary loop pressure, and is installed on the head of the reactor pressure vessel by the Ω seal. During the operation cycle of the reactor, cracks may be induced on the Ω seal weld by primary water stress corrosion cracking which is the major factor leading to leakage of coolant and impacting the safety of the reactor. Therefore, the Ω seal is required to be repaired to maintain structural integrity and safety by using the evaluated overlay weld technology. Based on fracture mechanics, this paper investigates the crack growth safety assessment of the overlay weld structure on the Ω seal of control rod drive mechanism, induced by the operation fatigue and stress corrosion, and the evaluation is completed by the standard code. The calculation method and assessment results can be a reference for the overlay weld structure design and safety evaluation of the Ω seal in CRDM.
Passive System Reliability Analysis Based on Improved Support Vector Machine
Authors: Ding Hao , Cai Qi , Zhang Yongfa , Jiang Lizhi
Published: October 2017

The desire for improved safety and simplified designs has led to the increased consideration and rapid development of passive systems. Passive system reliability analysis and evaluation is essential, especially in innovative nuclear reactor designs. Uncertainty related with functional failure should be considered into assessment of passive system reliability based on thermal hydraulics. The low efficiency of thermal hydraulic simulation codes cannot meet the needs of practical application and require a surrogate method to obtain the instant failure probability of passive systems.

In this paper, improved Support Vector Machine (SVM) method with Latin Hypercube Sampling (LHS) is adapted to set an approximation. This method, which can fit the nonlinear elements in thermal hydraulic equation, could determine the failure criterion, a surface classifying operation and failure mode, precisely. Then the calculation of an implicit function is trained and accomplished through the characteristics and features existing in samples. A comparison of polynomial form and neural network method is presented. According to the analysis of parameters effecting the probabilistic assessment, the inputs are adjusted to retrain the implicit function and obtain a better solution.
Study on Resistance Characteristics of Trilobal Throttling Element in Narrow Annular Channel
Authors: Youguang Wu , Weihua Li , Haijun Jia
Published: October 2017

An experimental study was performed to investigate the resistance characteristics of the trilobal throttling element in narrow annular channel, while corresponding flow fields analyzed using Fluent software with RNG k-ε turbulence model based on the structural grid for the trilobal throttling element with different placket angle. A relationship between the packet angle θ and the local resistance coefficient ξ is obtained. Results show that the local resistance coefficient obtained by numerical simulation of the trilobal throttling element in narrow annular channel have a good agreement with the experimental data, and their relative error is less than 8%. This also fully proves that the RNG k-ε turbulence model is suitable for the simulation of the trilobal throttling element in the narrow annular channel. The local resistance coefficient of the trilobal throttling element in narrow annular channel increases with rising packet angle θ, and the relationship is similar to the power function change.
The Study on High Efficiency Solidification Technology of Radioactive Liquid Waste Containing Boron
Authors: Wang Yaguang , Zhang Jinsong , Chen Yunming , Li Bing , Cao Qi
Published: October 2017

The technology of radioactive liquid wastes containing boron is a hot spot of radioactive waste disposal. The radioactive borate waste is solidified by conventional cement solidification in our country. In order to dispose the radioactive boron-containing waste more securely and efficiently, this work focuses on the development of high-efficiency cement solidification. In this work, the borate in radioactive liquid wastes containing boron is changed into polyborate and solidified by high efficiency formula to increase the waste containing rate. The high efficiency cement solidification formula was gotten in lab, its indicators are as follows: The containment of boron is 49.5%;The fluidity is 375mm;The compressive strength is 22MPa;The resistance of leachability (Sr, Cs,Co) is better than the requirement of national standard.
Study on Kernel Functions of Pressure Oscillation in MPS Method
Authors: Yue Zhu , Chen Sun , Shengyao Jiang , Xingtuan Yang , Riqiang Duan
Published: October 2017

In this paper,we studied the characteristic of four kernel functions of Moving Particle Semi-implicit method (MPS).In order to find the dependence of pressure oscillation on kernel functions, the dam break problem was selected for as a benchmark.Results showed that proper value of a kernel function was required in the compact support region to smooth the pressure distribution. The value of the kernel function should approach zero when the particle distance was close to the edge of compact support region to avoid large replusive force.When the distance between particles was close to zero,the value of kernel function should be large enough to avoid particle clustering. At the same time,the kernel function should be simple to stabilize the simulation process and decrease the computational time.
Verifying Calculation of the Parameters of AP1000 Core Design Based on MCNP Model
Authors: Wen-bin Xiong , Xiu-ping Yan , Bo Wang , Zhan-yong Wang , Ye-wang Bie , Bo-ping Zhang , Ming Yang , Ci Chen , Ning Cai
Published: October 2017

This study investigates the reactor core physical properties of the AP1000, which applies the MCNP4a program to model the AP1000 reactor core with the parameters and data from the DCD (19th Edition) of the AP1000 Nuclear Power Plant, which has been submitted to the NRC. The model is applied to calculate and verify the physical parameters of AP1000 core design. The results match well with the design values in the DCD of the AP1000 nuclear power plant. The design values have been calculated by the KENO-Va program, which proves the correctness of the MCNP model. The model will be improved and applied for safety review and verification analysis of AP1000 nuclear power plant in the future.
Simulation Calculation of the Core Physical Characteristics of China Experimental Fast Reactor With MCNP
Authors: Wen-bin Xiong , Jian Cao , Qin Xie , Zhan-yong Wang , Chuang Wang , Hu-wei Li , Sen-gai Yang , Huan Mao , Jing Wang
Published: October 2017

The whole core model of CEFR is established according to the parameters of China Experimental Fast Reactor which are given by IAEA-TECDOC-1531, and the physical parameters of CEFR are simulated with the MCNP4a program. The calculation results are compared with the data contained in the safety analysis report of CEFR. The calculation results are consistent with the design value, which successfully demonstrates the reliability of this MCNP model. The MCNP model will be further refined and applied for nuclear safety review of the CEFR in the future.