bluuebunny/crossref_metadata_2025 huggingface dataset:
Scenario Development for Nuclear Emergency Decision Deduction Training Platform for Radiographers in Developing Countries: Case Study, Ghana
Authors: Oforiwaa Priscilla Obeng , Manchun Liang , Guofeng Su , Ke Li
Published: October 2020

Nuclear Radiological accident is of great threat to the Nuclear Agency aside nuclear terrorism. In countries with little or no major nuclear infrastructural facilities, it is of great importance to educate stakeholders and employees in the handling and emergency decision involved in the usage of these radionuclides, thus training on this platform gives the maximum and technical know-how in cases where it is difficult to procure types of equipment needed for such educational exercise in real life. Ghana is a country that seeks to increase the use of nuclear technological applications. Sealed Radioactive Sources are used in Ghana for research, Agriculture, mining exploration and brachytherapy. Operators need to train immensely in the handling of these sources.

The NEDDTP is a software system to help conduct training and emergency exercises simulation. The NEDDTP is developed by the China Nuclear Security center of excellence with the aim of performing the simulation and training of radiographers and emergency workers in the nuclear industry. This training platform gives a 3D real-life training scenarios, emergency simulation, command and dispatch, Process record and evaluation in exercise with virtual reality display for radiological worker. This paper elaborates on the process for scenario design, types of scenario development for different types of radiological accidents and the stages involved in training with these scenarios.
New Requirements of Defence in Depth and Extension of its Application
Authors: Chengcheng Wang
Published: October 2020

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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.
Experimental Study of the Processes of Gas-Steam Pressurizer Insurge Transients
Authors: Bolong Wang , Weihua Li , Haijun Jia , Jun Li , Wentao Hao
Published: October 2020

Small reactors have received more and more attention for their high safety, reliability, low power density, and short construction period. And the gas-steam pressurizer is widely used in small reactors due to its characteristics of simple structure, saves the heating and spray equipment, and prevents the coolant from boiling. The gas-steam pressurizer is a pressure control equipment for the reactor coolant system, and its characteristic of transient response is an important factor that affect operation stability of nuclear reactor systems. An experimental system was established to study the effect of pressure response for an insurge transient and influence factors were analyzed quantitatively. Experimental investigation shows that for the gas-steam pressurizer, the increase of coolant loading capacity (insurge) can cause system pressure rising. And the change of system pressure has much consistency with the change of liquid level and gas space temperature. The liquid phase exists temperature fluctuations and overall shows a downward trend during the insurge transient. And there exists a temperature gradient from bottom to top in the pressurizer liquid phase region during the insurge transient. The change of water vapor quantity curve is the oscillating curve during the transient and water vapor quantity is in a decreasing trend overall during the insurge transient. What’s more, the experiments also analyzed the pressure response and temperature response during the insurge transient.
Protection of Nuclear Power Plants Against Natural Hazards: Protection Principles Against Rare and Severe Natural Hazards for New Nuclear Power Plants
Authors: Coulon Vincent , Christophe Sébastien , Grammosenis Laurence , Guinard Luc , Cordier Hervé
Published: October 2020

The field of protection against external natural hazards (eg.: rare and severe hazards) has regularly evolved since the design of the first NPPs (Nuclear Power Plants) to take into account the experience feedback. Following the Fukushima Daiichi accident in March 2011, consideration of rare and severe natural hazards has considerably increased in the international context. Taking rare and severe natural hazards into account is a challenge for operating nuclear reactors and a major issue for the design of new nuclear reactors.

In Europe, considering lessons learnt from the Fukushima Daiichi accident, European safety authorities released new reference levels in the framework of WENRA 2013 (Western European Nuclear Regulators Association) standards for new reactors [1] to address external hazards more severe than the design basis hazards. Considering this input, the French and UK nuclear regulators have released specific guidelines (Guide No. 22 related to design of new pressurized water reactors [2] for France and ONR Safety Assessment Principles SAPs [3] for the UK) to describe how to apply those principles in their national context.

To comply with those different guidelines, EDF has developed different approaches, called Beyond Design Basis (BDB) approaches, related to rare and severe natural hazards issue in the French and UK context for nuclear new build projects.

Those two approaches are presented in the present technical paper with the following structure:

- safety objectives;

- hazards to consider;

- SSCs (Structures, Systems, and Components) required to meet safety objectives;

- study rules and assumptions;

- analysis of deterministic or probabilistic nature, thereby including the following:

○ analysis of available margins (margin between 10−4 per annum exceedance frequency of hazard site level or equivalent level of safety and the chosen Design Basis Hazard level also called ‘inherent margin’);

○ Fukushima Daiichi accident Operating Experience feedback;

○ probabilistic safety analyses.

This technical paper highlights common characteristics and differences between the two approaches considering the French and UK regulatory contexts.
Impact of Climate Change on EDF’s Nuclear Facilities: Climate Watch Approach
Authors: Guinard L. , Parey S. , Cordier H. , Grammosenis L.
Published: October 2020

According to the Periodic Safety Review Process, the safety level is re-assessed every ten years, considering national and international operational feedback, evolution of knowledge and best available practices. Protection against natural hazards is part of this safety level re-assessment. In the current global change context, climate change impact has to be integrated in external natural hazards estimations, such as climatic hazards or external flooding.

EDF has consequently implemented a climate watch approach. Undertaken approximately every 5 years, roughly in line with the publication of the assessment reports of the Intergovernmental Panel on Climate Change (IPCC) and with the update of safety licensing basis during Periodic Safety Reviews, this approach is intended to:

- revisit the climatic hazards which present a plausible or certain upward trend, and could lead to an increased reference hazard level,

- monitor the reach of target levels which should trigger a thorough analysis (concept of Major Climate Event) to ensure the robustness of the reference hazard level between two periodic reviews.

This climate watch approach is developed in partnership with the scientific community and is based on the following activities:

- compile and analyze datasets on hazards that are subject to changes with climate change (observed and modelled time series),

- develop knowledge of associated climatic phenomena (models, projections).

The application of this approach is presented in two steps:

- the key implications of the last climate watch exercise carried out in 2015, which identified climatic hazards whose evolution is unfavorable and is plausible or certain for the sites of EDF NPPs:

○ High air and water temperatures (for the “heat wave” hazard)

○ Sea level (for the “external flooding” hazard for coastal or estuary sites)

○ Drought or « low flow » hazard for fluvial sites;

- the results obtained for the 900 MW units, for which EDF started the 4th periodic safety review in 2019.

Such an approach, which is closely linked to periodic reviews, ensures the robustness of nuclear power plants to the climatic hazards through the consideration of the updated hazard levels.
Design of an Innovative Moisture Separator Technology for Use in Nuclear Power Plants: Numerical Approach — Part 1
Authors: Padovan Loris
Published: October 2020

The Moisture Separator Re-heater (MSR) is a key component of Nuclear Power Plants (NPP) both in terms of performance and avoiding erosion and erosion/corrosion damage. Wet steam is usually dried in a MSR by inertial separation using separator elements. Depending on the design of a MSR, the technology of separator elements contributes significantly to its size and performance, hence is seen as a subject for in-depth investigation, improvement and innovation.

Computational Fluid Dynamics (CFD) has been used to understand the working principles of moisture separating devices, in particular the OpenFOAM platform has been utilized for this scope. Eulerian/Lagrangian models, wall-droplet interaction and water film formation models have been adopted to determine the physical phenomena occurring during the moisture separation process. Additional sub-models have been implemented to make a more robust solver and to solve in a comprehensive way all the possible physical processes: in particular a two-layers turbulence model and a film breakup model have been implemented. An out-of-the-box thinking approach was adopted to devise a new proposed chevron vane. Aerodynamic principles were used to design an innovative concept of separator panel, which can entrap the moisture droplets and water rivulets through a subsequent formation of recirculating steam representing artificial slots (hidden pockets) within the separator channel. The control of the steam separation on precise regions of the separator panel wall, helps the drainage of the water film without the utilization of physical obstacles (pockets or drainage channels). To validate the results achieved from the numerical simulation and to characterize separation performance of a new kind of separator technology, a bespoke test rig has been designed, built and put into operation at typical MSR operating conditions [1]. Throttling calorimeter methodology has been adopted to measure, with very good accuracy, the residual moisture content after the separator.

The design developed has shown excellent separation performance. Particularly, this solution will allow improved MSR performance and significantly reduced MSR size. This represents an innovative technology which is a major advance on current technology available within the industry. The novel design features have been patented by General Electric. The first operation of a MSR with this technology is eagerly awaited.
Development of Advanced Fire Zone Model Applicable to Fire PRA for Nuclear Power Plant
Authors: Ji Junghoon , Shirai Koji , Tasaka Koji , Udagawa Toshiko
Published: October 2020

In implementing the fire PRA for nuclear power plants, a highly predictive fire model is required for more realistic fire scenarios and fire risk assessment. The fire simulation zone model BRI2002 developed in Japan has been continuously improved to allow analysis considering the characteristics of a compartment fire.

In this study, a heat feedback phenomenon was introduced in BRI2002, in which combustion of a fire source can be accelerated by radiant heat transfer inside the compartment during a compartment fire. Not only the thermal radiation from the flame and smoke layers, but also radiation from the hot ceiling surface and the ceiling jet flame were considered when the flame impinges with the ceiling.

In addition, in the zone model, the existing model for predicting the oxygen concentration in a compartment was improved so that the oxygen concentration could be predicted considering the vertical location of a fire source (height from the floor).

The prediction results were verified by full-scale compartment fire test results. As a result of the calculation in which the fire source is installed at 2 m above the floor, the prediction results for the burning rate and zone temperature were well consistent with the test results.
Analytical Study on Removal Mechanisms of Cesium Aerosol From a Noble Gas Bubble Rising Through Liquid Sodium Pool
Authors: Miyahara Shinya , Kawaguchi Munemichi , Seino Hiroshi
Published: October 2020

In a postulated accident of fuel pin failure of sodium cooled fast reactor, a fission product cesium will be released from the failed pin as an aerosol such as cesium iodide and/or cesium oxide together with a fission product noble gas such as xenon and krypton. As the result, the xenon and krypton released with cesium aerosol into the sodium coolant as bubbles have an influence on the removal of cesium aerosol by the sodium pool in a period of bubble rising to the pool surface. In this study, cesium aerosol removal behavior due to inertial deposition, sedimentation and diffusion from a noble gas bubble rising through liquid sodium pool was analyzed by constructing a computer program which deals with the expansion and the deformation of the bubble together with the aerosol absorption. In the analysis, initial bubble diameter, sodium pool depth and temperature, aerosol particle diameter and density, initial aerosol concentration in the bubble were changed as parameter, and the sensitivities of these parameters on decontamination factor (DF) of cesium aerosol were investigated. From the results, it was concluded that the initial bubble diameter was most sensitive parameter to the DF of cesium aerosol in the rising bubble due to the inertial deposition. It was found that the sodium pool depth, the aerosol particle diameter and density have also important effect on the DF of cesium aerosol, but the sodium temperature has a marginal effect on the DF.

To meet these results, the experiments for the investigation of cesium aerosol absorption behavior from rising noble gas bubble through sodium pool are under planning to validate the results for the sensitivities of above-mentioned parameters on the DF of cesium aerosol in the analysis.
An Experimental Study of Two-Phase Flow in a Tight Lattice Using Wire-Mesh Sensor
Authors: Zhang Hengwei , Xiao Yao , Gu Hanyang
Published: October 2020

Tight lattice bundle can improve the conversion ratio and the heat transfer coefficient between the fuel bundle and the coolant, which is widely used in the innovative reactor fuel bundle design. The P/D ratio of a tight lattice bundle is usually less than 1.1, which is smaller than that of a conventional rod bundle. In the small-break loss-of-coolant accident (LOCA), the steam-water two-phase flow will occur in the reactor. The investigation of gas-liquid two-phase flow in the tight lattice is very important to the reactor safety analysis. A dual sub-channels tight lattice was designed in this study. The original reference of the channel is the annular fuel bundle, with the fuel diameter of 15.52mm, pitch of 16.51mm, P/D = 1.06. The original reference of working condition is the stream-water two-phase flow under the pressure of 15.5MPa. The experimental condition is the air-water two-phase flow at the normal temperature and pressure. According to the ratio of a critical bubble diameter in the reactor (steam-water) to that in atmospheric conditions (air-water), the channel is zoomed in 2.7 times. The diameter of the rod in the dual sub-channels tight lattice is 42mm and the pitch is 44.6mm. The total length of the dual sub-channels tight lattice is 3m. A self-developed 16 × 32 Wire-mesh sensor (WMS) was used to measure the void fraction distribution of air-water two-phase flow in the dual sub-channels tight lattice channel. The spatial resolution of the WMS is 2.79mm and the temporal resolution is 5000fps. The WMS was installed at a distance of 2.5m from the channel inlet and 0.5m from the outlet, which can avoid the influence of outlet on bubbles. The experimental range of flow condition is 0.921–1.84m/s for the superficial liquid velocity and 0.0884–1.07m/s for the superficial gas velocity. The instantaneous and time-averaged void fraction distributions in the channel was measured. With the increase of superficial gas velocity, the distribution of void fraction distribution changed from the wall peak to the core peak. The characteristics of bubbles in the sub-channel were also discussed in this study.
Numerical Investigation on Turbulence Models in Bubbly Flow Using Euler-Lagrange Approach
Authors: Zhou Yujia , Du Jingyu , Zhao Chenru , Bo Hanliang
Published: October 2020

To accurately simulate the bubble motions in the turbulent flow liquids, two way coupled Euler–Lagrange method is adopted in this work. The continuous phase is solved with the Navier–Stokes equations based on the Euler grid, while the individual bubble is tracked by using the Lagrange frame of reference. Two–way coupling is realized by transferring the interaction forces to the momentum equation of the continuous phase at each fluid flow timestep and in turn influencing the bubble motion. The interaction forces including the drag force, lift force, wall lubrication force and virtual mass force are carefully selected. Turbulent models are of significance to capture the fluid flow conditions. Comparing the different cases treated by the k-ε, PANS and LES models with the experimental data, the results of the PANS with RWM and BIT, and LES can dynamically display the oscillation of the bubble plume and the calculated time averaged axial liquid velocity reasonably agrees with the measured data. In the case of bubble injective flow conditions in large containers, the bubble induced turbulent effect can be almost neglected. However, the RWM describes the fluctuating velocity should be considered applied with the PANS model. In addition, the saturated vapor–liquid system under high pressure of 6.9 MPa is numerically simulated by implementing the PANS with R WM and BIT.
Aerosol Decontamination Behavior in Two Phase Flow During Pool Scrubbing
Authors: Yoshida Kohei , Fujiwara Kota , Nakamura Yuki , Kaneko Akiko , Abe Yutaka
Published: October 2020

In some scenarios in severe accidents (SAs) of nuclear power plants (NPPs), non-condensable gas containing vapor and small particles of fission products (FPs) will be released from the reactor vessel, pass through the wet well of a suppression chamber. In these conditions, water inside the wet well will trap FPs before released to the atmosphere. This trapping effect is called pool scrubbing. Since pool scrubbing is assumed to have a good decontamination effect, researches about the pool scrubbing are important from the viewpoints of NPPs safety. Therefore, it is necessary to understand the two-phase flow behavior in pool scrubbing.

The mechanism of the FPs removal behavior in pool scrubbing is not clearly understood due to the complicated hydrodynamic phenomena. In addition, the evaluation of the validity of physical models used in pre-existing SA analysis codes are not enough. Therefore, a reliable model to evaluate particle decontamination efficiency by pool scrubbing is highly regarded.

Currently, decontamination factor (DF) of pool scrubbing is calculated by analysis code such as MELCOR. In the MELCOR code, some assumptions and empirical formula are used, e.g. bubble diameter and rising velocity. However, the correlation between DF and flow structure characteristics are not understood, and dominant factor of decontamination behavior in two phase flow is required.

Therefore, the aim of this study is to elucidate the dominant factor of decontamination by measuring DF and flow structure of bubbly plume. As to understand two phase flow characteristics during pool scrubbing, we focus on bubble diameter and swarm rise velocity. In Air-particle mixture flow injected into the experimental facility gas velocity, void fraction and bubble diameter distribution were measured by a wire mesh sensor (WMS) in each height. In addition, DF by pool scrubbing was derived from aerosol measurements by an aerosol spectrometer. The measured warm rise velocity were compared with equation of single bubble terminal velocity. The swarm rise velocity is higher than single bubble velocity. Furthermore, Large bubble are identified to be still remained at swarm region defined in MELCOR. In addition, dominant factor of decontamination behavior is defined as particle diameter, neither aerosol solubility nor particle density.
HTGR Power-Level Control Only by Regulating Helium Flowrate
Authors: Dong Zhe
Published: October 2020

Currently, most of the power-level control methods of nuclear reactors are given based upon the scheme of inducing reactivity directly via control rods. However, the reactivity can also be injected indirectly via the negative temperature feedback effect. Motivated by this, a cascaded power-level control of high temperature gas-cooled reactor (HTGR) is proposed by only regulating primary helium flowrate, which can be able to provide globally asymptotical closed-loop stability. This new HTGR power-level control is composed by a helium temperature controller and a neutron flux controller. The former controller, which is located in the inner loop, regulates the primary helium flowrate according to the setpoint of helium temperature. While, the latter one in the outer loop revises the setpoint of helium temperature so as to regulate neutron flux. Numerical simulation results verify the theoretical result while showing the satisfactory performance as well as the influence from the controller parameters.
The Dependence of Pool Scrubbing Decontamination Factor on Particle Number Density: Modeling Based on Bubble Mass and Energy Balance
Authors: Sibamoto Yasuteru , Sun Haomin , Hirose Yoshiyasu , Kukita Yutaka
Published: October 2020

The dependence of pool scrubbing performance on particle number density is studied through numerical simulation of experimental results. The DF values obtained from the authors’ experiments (Sun et al., Sci. Technol. Nucl. Inst., Article ID 1743982, 2019) indicate a sharp decrease with an increase in the inlet particle number density beyond 1011/m3. The mechanisms underlying such dependence is yet to be studied.

In this paper, a simple model is developed to study the factors affecting the experimentally observed dependence of DF. The test results suggest that the condensational growth of particles plays an essential role in the inertial capture. The vapor condensation on the particles has an effect to deplete the vapor supersaturation in the bubble by both lowering the vapor concentration and raising the temperature. This effect will become important at high particle number densities.

The bubble mass and energy balance is calculated to derive the particle growth and the inertial DF as a function of the bubble rise distance through the pool water. The balance is assumed to be quasi-steady, and the vapor concentration and the temperature to be uniform in the bubble. It is shown that the model reproduces the tendency observed in the experimental DF. The model predicts that the degree of supersaturation is affected when particle concentration exceeds 1011/m3, curbing the condensational growth of particles, and thereby retarding the inertial capture.
Selection and Validation of Fast and Synchronous Interface to the Controller of a Space Nuclear Reactor Power System
Authors: Hahn Andrew , Schriener Timothy , El-Genk Mohamed S.
Published: October 2020

Enabling transmission of state variables and controller command signals between physics-based model and physical or emulated PLC requires developing fast and synchronous communication interface. This research investigates four methods of interface communication: (a) reading/writing to text files, (b) serial communication, (c) TCP communication, and (d) shared memory. The shared memory interfacing method provides the best performance, reliability, and fidelity. It is validated by linking a representative PLC controller to a Simulink dynamic simulation model of a fully integrated space reactor power system with multiple closed Brayton cycle loops for energy conversion (DynMo-CBC). Transient performance results of the PLC controlled dynamic simulation model are identical to those obtained using the Simulink model with internal control.