bluuebunny/crossref_metadata_2025_split huggingface dataset:
New Requirements of Defence in Depth and Extension of its Application
Authors: Chengcheng Wang 
Published: October 2020

Abstract
As states in the IAEA SSR-2/1 (Rev.1) and IAEA TECDOC-1791, the defence in depth concept is not to be understood as merely limited to the request for the implementation of a number of consecutive barriers and protection levels, but is to be understood as the main general principle that leads to the formulation of safety requirements including requirements necessary to achieve the quality and reliability expected for the barriers and for systems ensuring their integrity. Thus, the application of defence in depth concept is not only limited in the reactor core design, but also can extend to a wider range in nuclear power plant design.
In this paper, the application of defence in depth concept, which is defined in IAEA latest requirements, in the irradiated fuel water pool storage, electrical system, internal hazards and external hazards of the nuclear power plant design are reviewed. Base on the study of the newest standard and codes, the design requirements of each level of defence in depth in the above systems are confirmed. The analysis of defence in depth design features of an advanced nuclear power plant in China are also show the application value of defence in depth in these extended regions.
Seawater Desalination Using a High-Temperature Gas-Cooled Reactor (HTR): Overview of Regulatory and Safety Considerations
Authors: Al-Saud Mishari , Chao Fang 
Published: October 2020

Abstract
High-Temperature Gas-Cooled Reactor (HTR) is a promising Gen IV reactor technology that has a wide range of applications. Saudi Arabia expressed interest in using HTR as an energy source for seawater desalination. A pre-feasibility study showed that HTR-Desalination is economically competitive and feasible. Yet, the application of HTR power and process heat in the desalination industry faces some technical, conceptual, and regulatory challenges. These challenges are mainly because the reactor and desalination plant are co-located and share common systems and facilities. Moreover, there is a risk of radioactivity and brine discharge impact, since both plants share the water source and discharge location. All these issues challenge the reliability and safety of both plants. Therefore, it is essential to develop effective regulatory frameworks. The basic regulatory and infrastructural requirement for the HTR is like any other nuclear power plant. This study reflects on the typical operational issues and influence of accidents in both plants and their impact on the other. Concluded with regulatory recommendations with an effort to find common interfaces between the regulatory aspects of the nuclear power and desalination industries, which aim at providing a more holistic view on a more comprehensive regulatory framework for nuclear desalination.
The Solenoid Valve for Main Steam Relief Valve Adapted Under Severe Accident
Authors: Kojima Nobuo , Nishino Koji , Sakemura Keisuke , Kobayashi Kengo 
Published: October 2020

Abstract
MSSRVs in Boiling Water Reactor (BWRs) can be mandatory opened through a solenoid valve, when a nuclear reactor raises unusual pressure. The solenoid valve consists of a valve main part which forms a flow channel, and a pilot part which controls a flow direction.
In the accident of the Fukushima Daiichi nuclear power plant, MSSRVs ware over the design specification of these solenoid valves, and were not able to operate. One of the reason, there is degradation of a sealant and a coil of a solenoid valve.
As one of the measures for rebooting the BWRs, the development and verification of a solenoid valve which were applied to the SA condition are required. Since we developed and verified it applied to the SA.condition, we report here.
Static Application of Transient Hydrodynamic Loads on Vessel Internal Structures As a Result of Pulse Jet Mixer Overblow: High-Frequency Load Acoustic Event
Authors: Fant Brian , Garcilazo Rafael , Blevins Robert 
Published: October 2020

Abstract
At the Hanford Waste Treatment and Immobilization Plant (WTP), various vessels are designed to be agitated with internal pulse jet mixers (PJMs) in order to provide a means of mixing with no moving parts local to the vessel. PJMs are operated by use of an applied vacuum to draw liquid in followed by motive air to force liquid out (while not completely discharging all the liquid). This continual operation results in mixing of the vessel contents. In off-normal conditions, PJMs may completely discharge resulting in air rapidly injected into the vessel (PJM overblow).
An evaluation is complete to determine the statically applied transient acoustic event loads resulting from Pulse Jet Mixer (PJM) overblow on a vessel’s internal submerged structures. The high-frequency acoustic load on internal structures is determined via analysis of overblow hydrophone test data, vessel modal and harmonic analysis, determination of an overblow forcing function through a Fourier analysis of test data compared to modal analysis of the vessel, application of the overblow forcing function in a finite element analysis model with acoustic fluid elements, use of a displacement-response spectra, and Hooke’s Law. Compared to previous analyses, this improved method can account for changes to PJM cavity pressure, PJM nozzle diameter, and more accurately represents the loads because it considers geometric constraints.
Guidance for Developing Fuel Design Limit of High Temperature Gas-Cooled Reactor
Authors: Sato Hiroyuki , Aoki Takeshi , Ohashi Hirofumi 
Published: October 2020

Abstract
The present study aims to propose a guidance that facilitates to determine fuel design limits of commercial HTGR on the basis of licensing experience through the high temperature engineering test reactor (HTTR) construction. The guidance consists of a set of figure of merits (FOMs) and a process to determine their evaluation criteria. The FOMs are firstly identified to satisfy safety requirements and a basic concept of safety guides established in a special committee under the Atomic Energy Society of Japan with the support of the Research Association of High Temperature Gas Cooled Reactor Plant. The development process for the evaluation criteria takes into account not only the top-level regulatory criteria but also design dependent constraints including the performance of fission product containment in physical barriers other than fuel, fuel qualification criteria, design specifications of an instrumentation and control system. As a result, a comprehensive and transparent procedure for designers of prismatic-type commercial HTGR has been developed.
This paper provides details on the development procedure for fuel design limit. methods to derive the limits are also presented.
A Comparative Study of Constrained and Unconstrained Melting Inside a Sphere
Authors: Kothari Rohit , Revankar Shripad T. , Sahu Santosh K. , Kundalwal Shailesh I. 
Published: October 2020

Abstract
Present study is focused on the computational analysis of melting of PCM inside the spherical capsule. Both unconstrained and constrained melting is analyzed for the constant PCM volume and similar initial and boundary conditions. RT27 is chosen as the PCM for this study. Air is considered at the top of PCM inside the spherical capsule. Results are validated with the existing experimental and computational results and found to be in good agreement. Results obtained from present study are compared for the melting fraction, pattern and time. Composite diagrams are presented for the streamline and temperature contours.
On-the-Fly Treatment of Discrete Representation Thermal Neutron Scattering Data in RMC Code
Authors: Zheng Lei , Li Kaiwen , Wang Kan 
Published: October 2020

Abstract
A proper treatment of thermal neutron scattering data is required for the high-fidelity neutronics calculation of thermal reactors. Monte Carlo codes typically use an S(α, β) treatment to describe scattering events in the thermal region if the S(α, β) data is available for the material. The S(α,β) model stores a large majority of scattering physics and can handle thermal scattering process accurately. In neutronic-thermohydraulic coupling calculations, the temperature effect on nuclear data must be treated properly. The on-the-fly sampling method or the on-the-fly interpolation method are typically used in thermal region. In this paper, the on-the-fly interpolation method for the discrete representation S(α,β) data was introduced. The two-dimensional linear-linear interpolation was used to calculate the scattering cross sections and the secondary information for inelastic scattering, coherent elastic scattering and incoherent elastic scattering. The implemented on-the-fly capability was tested by a series of benchmarks that contain various thermal materials, including light water, beryllium and beryllium oxide. The integral kinf eigenvalues, the efficiency and the fine energy spectra of the on-the-fly treatment capacity were compared with those of the references. Results show that the on-the-fly treatment capability has high accuracy, and the computational time increases up to 20%.
Study on Eutectic Melting Behavior of Control Rod Materials in Core Disruptive Accidents of Sodium-Cooled Fast Reactors: Thermophysical Properties of Eutectic Mixture Containing of High Concentration Boron in a Solid State
Authors: Takai Toshihide , Furukawa Tomohiro , Yamano Hidemasa 
Published: October 2020

Abstract
In a core disruptive accident scenario, boron carbide, which is used as control rod material, may melt below the melting temperature of stainless steel due to the eutectic reaction with it. Produced eutectic mixture is assumed to relocate widely in the degraded core, and this behavior plays an important role to reduce the neutronic reactivity of the degraded core materials significantly. However, these behaviors have never been simulated in the severe accident computer codes, and reducing the uncertainty is important for reasonable assessment. To contribute improvement of the core disruptive accident analysis code to handle these eutectic melting and relocation behavior, authors had been carried out the evaluation of the thermophysical properties of stainless steel containing boron carbide, which needed as a basic data for cord improvement. Since the solubility range of boron against iron is expected to be wide, the crystalline phase of eutectic mixture may change according to boron concentration in the eutectic mixture. And this may affect the thermophysical properties themselves.
In this work, the density and specific heat of stainless steel containing 17 mass% boron carbide in a solid state are obtained and compared with these of stainless steel containing 0 and 5 mass% boron carbide. By adding 17 mass boron carbide to stainless steel type 316L, the density decreased approximately 24% and the specific heat increased approximately 25% at 293 K. The density of stainless steel containing boron carbide tended to decrease almost linearly depending on the amount of boron carbide added, none the less for difference of crystalline phase. On the other hand, increasing trend of the specific heat of stainless steel containing 17 mass% boron carbide accompanying elevating temperature showed different behavior from that of stainless steel containing 0 and 5 mass% boron carbide. This difference in the trend of the specific heat was considered to be caused the difference in the crystalline phase.
Interfacial Area Transport Model for Bubbly Flow System in Vertical Rod Bundle With Spacer Grids
Authors: Liu Hang , Lai Jianyong , Li Yi , Zhang Yulong , Liu Minghao , Pan Liangming 
Published: October 2020

Abstract
In relation to the modeling of the one-dimensional interfacial area transport equation for an adiabatic bubbly flow in a vertical rod bundle, some existing models of sink and source terms were reviewed and evaluated. Based on the reviewed interaction mechanisms of bubbles and turbulent eddies, a new interfacial area transport model has been proposed. Two important impacts on bubble interaction have been taken into account in the new model: the effects of spacer grids with mixing vanes and the impacts of geometry structure. The spacer grids breakup large bubbles into small bubbles resulting in enhanced bubble random collision at the downstream of the spacer grids. Void transport is the main contribution between spacer grids. The new interfacial area transport model has been evaluated against the obtained experimental data in 17 bubbly flow conditions. The results indicate that the new model can predict the interfacial area concentration with the relative error of 19.9%. It is recognized that the proposed model is promising for predicting the interfacial area concentration for a bubbly flow in a rod bundle.
Experimental Study on Natural Convection Heat Transfer Outside Tube Bundle in Space Under Low Temperature Difference
Authors: Xu Junxiu , Ding Ming , Yan Changqi , Fan Guangming 
Published: October 2020

Abstract
The Passive Residual Heat Removal System (PRHRS) is very important for the safety of the heating reactor after shutdown. PRHRS is a natural circulation system driven by density difference, therefore, the heat transfer performance of the Passive Residual Heat Removal Heat Exchanger (PRHR HX) has a great impact to the heat transfer efficiency of PRHRS. However, the most research object of PRHR HX is the C-shape heat exchanger at present, which located in In-containment Refueling Water Storage Tank (IRWST). This heat exchanger is mainly used for the PRHRS of nuclear power plants. In the swimming pool-type low-temperature heating reactor (SPLTHR), the PRHR HX is placed in the reactor pool, which the pressure and temperature of the reactor pool are relatively low, and the outside heat transfer mode of tube bundle is mainly natural convection heat transfer. In this study, a miniaturized single-phase pool water cooling system was built to investigate the natural convective heat transfer coefficient of the heat exchanger under the large space and low temperature conditions. The experimental data had been compared with several correlations. The results show that the predicted value of Yang correlation is the closest to the experimental data, which the maximum deviation is about 11%.
Finite Element Method for Thermal Design of Radioactive Material Transport Packages
Authors: Meng Dongyuan , Sun Shutang , Sun Hongchao , Li Guoqiang 
Published: October 2020

Abstract
In this paper, a simplified typical radioactive material transport container model is used to change the thickness of the container insulation layer and the thickness of the heat sink respectively. The finite element method is used to simulate the heat transfer, and a quantitative optimization of the heat transfer performance of the radioactive material transport container is given.
Experimental Study of Thermal and Hydraulic Characteristics of Pressurizer-Type Heating Container During Blowdown Process
Authors: Ke Bingzheng , Gao Puzhen , Wang Bo , Chen Bowen , Li Ru , Tian Ruifeng , Cheng Kun , Lu Chuan 
Published: October 2020

Abstract
Blowdown process is an important process in LOCA. Studying the internal thermal hydraulic characteristics of the pressurizer during LOCA blowdown process can help to accurately analyze the process of the whole LOCA, which is of great significance to reduce the hazard of accidents and improve the safety of nuclear power plants. An experiment system of blowdown process under low pressure conditions is built, and it uses a pressure gauge and a differential pressure level gauge to measure the pressure and liquid level in the pressurizer-type heating container, and uses a thermocouple to measure the temperature in the pressurizer-type heating container at different heights and distances from the centerline. The changes of pressure, liquid level and temperature characteristics during the blowdown process were obtained. The research results show that the higher the heating power, the greater the slope of pressure change in the heating container, and the pressure changes from negative to positive growth; The greater the heating power is, the faster the liquid level decreases, and the liquid level decreases sharply after the blowdown inflection point. The greater the heating power is, the greater the rate of temperature change, from negative growth to positive growth. The temperature change rate after the blowdown inflection point is reduced. As the height increases, the temperature inside the container first increases and then decreases; the further away from the container’s centerline, the higher the measured temperature.
Method for Prognosis of LBLOCA Initiated Emergency Condition
Authors: Ning Wang , Renze Wang , Yapeng Yang , Zongyang Feng , Linsheng Jia , Yining Liu , Boning Liang 
Published: October 2020

Abstract
Fukushima accident shows again that the probability of reactor accident exists even though it is extremely small. In case of emergency in nuclear power plant, emergency condition of the reactor plays an important role in decision making. During emergency response, especially early stage of severe accident with large release of radioactive nuclides, decision making for protection actions should be based on emergency condition in NPPs. If emergency condition could be prognosed, more time could be bought for decision making and emergency response.
In this paper, method for prognosis of large break loss of coolant (LBLOCA) initiated severe accident progression was established based on transient analysis for M310 reactor. Mass and energy conservation equations are the basis of the method. Separated flow model is used for prognosis of emergency condition for large break loss of coolant accident initiated severe accident. These conservation equations are solved approximately in order to significantly increase calculation speed. The active core is divided into 4 radial rings and 10 axial levels, which means there are 40 cells. Heat transfer calculation in the core is done using four experimental correlations.
Based on the method established in this paper, a code using for prognosis of LBLOCA initiated severe accident emergency condition was developed. Research on method for prognosis of other severe accidents are being conducted.
Implementation of a Particle Resuspension Model in a CFD Code: Application to an Air Ingress Scenario in a Vacuum Toroidal Vessel
Authors: Gelain Thomas , Ricciardi Laurent , Gensdarmes François 
Published: October 2020

Abstract
During a loss of vacuum accident (LOVA), dust particles that will be present in the future tokamak ITER are likely to be resuspended, inducing a risk for explosion and airborne contamination. Evaluating the particle resuspension/deposition and resulting airborne concentration in case of a LOVA is therefore a major issue and it can be investigated by using a CFD code. To this end, this article presents the implementation of a resuspension model in a CFD code (ANSYS CFX) and its application to an air ingress in a vacuum toroidal vessel with a volume comparable to ITER one.
In the first part of the article, the Rock’n Roll model and its operational version with the Biasi’s correlation is presented.
The second part of the article will be devoted to the implementation of the Rock’n’Roll model in ANSYS CFX for constant friction velocities and its adaptation to non-constant friction velocities.
Finally, the paper presents the simulations obtained on the particle resuspension for an air ingress scenario in a large vacuum vessel. This case is particularly interesting and non-intuitive because as the initial pressure is reduced, the particle behavior is different from that at atmospheric pressure. Further, a competition between airflow forces and gravitational force occurs, due to the low pressure environment, potentially restricting the resuspension, and the pressure influence also has to be taken into account in the particle transport and deposition (Nerisson, 2011).
Three particle diameters were studied allowing to show the evolution of the resuspension with this parameter and to calculate dust resuspension rates and airborne fractions during the air ingress.
Uranium Radioactivity Measurement for Low-Level Wastewater Based on Ultra-Low Level Liquid Scintillation Analyzer
Authors: Lv Jinxu , Lv Ning , Guo Huiping , Hu Zeting , Song Geng , Cai Yinying , Zhang Heling , Li Daji 
Published: October 2020

Abstract
In this paper, a method based on ultra-low level liquid scintillation analyzer for measuring uranium radioactivity in low-level radioactive wastewater is proposed. It can easily and quickly measure the radioactivity level of uranium in samples, and even distinguish the main isotopes of uranium.
In low-level wastewater samples, if the uranium isotope has the same or similar composition and ratio to the natural uranium isotope, Comparing to the results in units of mass concentration given by laser-fluoremetry method, the liquid scintillation method can give the results in units of activity concentration, and these two results with different units can interconvert to each other. However, interconversion of results on above two measurement methods cannot be applied in low-level wastewater after uranium enrichment, because the proportion of isotopes with high specific activity in above wastewater, such as 234U and 235U, is significantly higher than that without enrichment. The measurement results in units of mass concentration will underestimate the uranium radioactivity level of the sample.
The liquid scintillation method directly gives the results in units of radioactivity activity concentration, and it is more convenient to compare with relevant national standards to determine whether the emission standards are met. The lowest limit of detection of this method is 0.014 Bq L−1 within 600 min.
A New Data Driven Method for Monitoring a Large Number of Process Values and Detecting Anomaly Signs With a Two-Stage Model Composed of a Time Window Autoencoder and a Deviation Autoencoder
Authors: Naito Susumu , Taguchi Yasunori , Kato Yuichi , Nakata Kouta , Miyake Ryota , Nagura Isaku , Tominaga Shinya , Aoki Toshio 
Published: October 2020

Abstract
In a large-scale plant such as a nuclear power plant, thousands of process values are measured for the purpose of monitoring the plant performance and the health of various systems. It is difficult for plant operators to constantly monitor all of the process values. We present a new data-driven method to monitor many process values and to enable early detection of anomaly signs including unknown events with few false detections. In order to accurately predict the process values in the normal state, we created a two-stage model composed of a time window autoencoder and a deviation autoencoder. The two-stage model handles a large number of process values, their rapid changes of the process values such as an operation mode change, changes of the process values in both the steady and the transient states, and the external disturbances such as exogenous noise, atmospheric temperature, etc. The time window autoencoder examines time correlations of time series process values while the deviation autoencoder treats correlations of variation due to external factors. We evaluated a predicting ability of the rapid changes, detection performances in the transient state, and detection performances under noisy conditions with simulated process values of a nuclear power plant, a 1,100 MW Boiling Water Reactor having 3,100 analog process values. The two-stage model clearly showed a good anomaly detection performance with zero or few false detections. The two-stage model would be an effective solution for plant monitoring and early detection of anomaly signs.