The aim of this work is to show a preliminary investigation on the propagation of electromagnetic fields generated by wireless technologies inside a nuclear facility or power plant. First, a survey of currently proposed wireless technologies for nuclear facilities and plants has been carried out. Then, for selected scenarios, the electromagnetic field propagation has been studied by means of electromagnetic simulation tools, and the presence of the nuclear environment has been simulated by properly modeling environmental parameters and engineered barriers. The choice of the proper simulation techniques and tools is mandatory in order to simulate the effect of the realistic environment on the propagation. Accordingly, the feasibility of wireless technologies application at nuclear facilities can be assessed on the basis of results achieved from simulated scenarios. The goal is to analyze, for selected scenarios, possible issues due to the propagation of an electromagnetic field in presence of simplified barriers mimicking the real nuclear environment. This approach can provide indications on how to deploy potential benefits of wireless technologies in a nuclear environment, evaluating pros and cons of the investigated technologies.

With renewed interest in nuclear power to meet the world’s future energy demand, the Generation IV nuclear reactors are the next step in the deployment of nuclear power generation. However, for the potentials of these nuclear reactor designs to be fully realized, its suitability, when coupled with different configurations of closed-cycle gas turbine power conversion systems, have to be explored and performance compared for various possible working fluids over a range of operating pressures and temperatures. The purpose of this paper is to carry out performance analysis at the design and off-design conditions for a Generation IV nuclear-powered reactor in combination with a recuperated closed-cycle gas turbine and comparing the influence of carbon dioxide and nitrogen as working fluid in the cycle. This analysis is demonstrated in GTACYSS; a performance and preliminary design code developed by the authors for closed-cycle gas turbine simulations. The results obtained shows that the choice of working fluid controls the range of cycle operating pressures, temperatures and overall performance of the power plant due to the thermodynamic and heat properties of the fluids. The performance results favored the nitrogen working fluid over CO 2 due to the behavior CO 2 below its critical conditions.

The efficient and accurate burn-up measurement of the spherical fuel element is the key component of the operation of the pebble bed high temperature gas-cooled reactor. The accuracy of the method that determine burnup by the activity of Cs-137 degrades due to operation characteristics of HTR-10. HTR-10, as an test reactor, operated on and off during the past years. It stayed shutdown more than power operation. In order to improve the measurement accuracy of Cs-137 activity and enhance the possibility to detect radionuclides with low activity, which can be used to correct the classic burnup assay method, a new measurement system is now discussed using anti-coincidence technology, which suppresses the Compton plateau. In this paper, Geant4 is used to simulate the anticoincidence measurement process taking high purity germanium γ-ray spectrometer as main detector and plastic scintillator as the annular detector. By analyzing the signal to noise ratio in different detection scenarios with all kinds of shape parameters of the annular detector, the annular detector with the best anti-coincidence effect are optimaized. The above research results provide an important theoretical basis for the construction of online burn-up measurement system based on anti-Compton technology.

The High Temperature Gas-cooled Reactor (HTGR) designed by Tsinghua university is under development in China. The electrical equipment in Pressure Vessel, such as magnetic bearings and helium circulator are operating in the high pressure helium environment. Design research of insulation property of the electrical device in helium under high pressure is necessary. Nevertheless, it is challenging to investigate by experimental technique. We propose a similarity law, converting the high pressure to a lower pressure to simplify the experimental conditions. Similarity theory of gas discharge is that two geometrically similar gaps with scaling coefficients of k , based on the same product number of pressure p and the gas length d , have the similar discharge characteristics. We research the validity of discharge similarity theory by simulation. A fluid model of direct-current discharge in helium atmosphere was established, referencing experimental results. And the discharge models of gaps were solved by finite-element method respectively. Four geometrically similar gaps were designed, the prototype gap is 2cm long and operating at a pressure of 600Pa while the pressure of three similar gaps are 300Pa, 200Pa, and 120Pa, and corresponding to the length of gaps are 4cm, 6cm and 10cm, with scaled-down factor k of 2, 3, and 5 respectively. The simulation results show that as long as the scaled-down factor of pressure k is less than 5 and the reduced length relation meets the condition p 1 d 1 = p 2 d 2 , the discharge characteristics of two geometrically similar gaps are similar. As a result, it is realizable to predict the insulation property of electrical device in helium with similarity law.

In order to study the effect of rolling motion on flow instability of parallel rectangular channels of natural circulation, the natural circulation reactor simulation system is used for physical prototype. And theory analysis model of parallel rectangular channels of natural circulation system under rolling motion is established and coded by Fortran. The results of the program are verified to the experiments, and the results are in good agreement. The flow instability boundaries of different pressure under static and rolling motion are calculated respectively. The results show that: 1) under static condition, with the increase of the pressure, the instability boundary line changes, and the system becomes more stable; 2) under rolling conditions, the heating power of instability boundary decreases comparing to the stable conditions. The instability occurs earlier; 3) the stability of the system decreases with the increasing of rolling amplitude and frequency.

The purpose of tokamak plasma diagnostics is to provide the necessary parameters for device protection, operation, and maintenance. It can also supply parameters for fusion physics research. As one of the main ways to diagnose nuclear fusion plasma, neutron diagnosis focuses on the detection of neutrons, produced by the D-D and D-T fusion reactions, to obtain the physical information of internal plasma. Neutron measurements are widely performed on tokamak to provide the essential information on the neutron yield rate of the plasma that is related to fusion power. Since neutron has no electric charge, neutron can’t be ionized directly by the interaction of electrons in the detection material. The interactions between neutron and nuclei, such as nuclear reaction and nuclear recoil, are used to detect neutrons. According to the front sensitive materials, neutron detectors can be divided into gas detectors, scintillator detectors, semiconductor detectors, ionization chambers and so on. Since the magnetic field surrounding Tokamak can have a magnificent influence on the performance of photo-electronic multiplier tubes (PMTs), it is necessary to employ magnetic shielding in designing detectors, thus guaranteeing the proper operation of detectors within a strong magnetic field. Although the PMTs are equipped with magnetic shielding materials by manufacturers, they can only resist the influence of geomagnetic field. Besides the magnetic shielding and neutron/gamma shielding, neutron detectors should be calibrated before used on the tokamak. Nine similar detectors were assembled and calibrated in this paper. The basic idea of processing calibration data is that we should adjust the resolution and the light response function in order to make experiment spectrum and simulation spectrum fit on the recoil proton edge. A special explication is given to the data processing of neutron calibration, followed by an analysis of its resulting light response function and by comparison with PTB’s results.

Although the neutron mean free path in typical fast reactors is long compared with that of typical thermal reactors, the heterogeneous flux distributions in angular domain and spatial domain are notable due to the strong elastic-scattering resonance and common core-level heterogeneous layout. Full-core multigroup transport analysis becomes inevitable for advanced fast reactor R&D. A fast reactor neutronics analysis system, SARAX is under development at XJTU of China. An existing in-house S N -nodal solver in triangular-Z geometry, DNTR was primarily chosen for its geometric adaptability. However, insufferable problems of computing time and storage were encountered when the solver was on active service. In this paper, the problems are firstly analysed and resolved from the theory and code levels. Then, widely used CMFD acceleration method with some stabilization techniques is implemented for the S N -nodal method in triangular-Z geometry, which can largely reduce the computing time. What’s more, a new acceleration method TCD is proposed and can obtain superior speedups with reasonable accuracy sacrifice. The updated solver, DNTR 1.1 has been developed based on these improvements, which can obtain dozens of speedup with reduced storage compared with DNTR for typical fast reactor simulation on a desktop computer.

In predicting the void fraction in subcooled flow boiling, accurate evaluation of single bubble behaviors is of considerable importance. In particular, bubble lift-off velocity affects the void fraction significantly since the bubble disappear quickly due to heat transfer with subcooled liquid if the lift-off velocity is high. In this study, the process of bubble lift-off was experimentally investigated to develop mechanistic correlations for the bubble lift-off velocity. In the development of the correlations, it was assumed that the bubble lift-off velocity in the horizontal direction is proportional to the bubble growth rate and that in the vertical direction is determined primarily by the local liquid velocity evaluated at the bubble center position. Then, impact of the bubble lift-off velocity on the void fraction was explored through numerical simulations. In the simulations, the bubble lift-off velocity in the lateral direction was parametrically changed. It was shown that the mean void fraction decreases with an increase in the lateral bubble lift-off velocity since the bubble condensation is enhanced. It was therefore confirmed that accurate evaluation of the bubble lift-off velocity is important for high accuracy prediction of the void fraction.

Nuclear reactor simulation is often based on multi-group cross-section libraries. The structure and resolution of these libraries have a strong influence on the accuracy and computational time; hence, number of groups and energy structure must be carefully considered. The relationship between group structures and how they impact generated cross-sections can be a critical parameter. Common energy boundaries shared among major group structures were identified and the relative kinship among those was reconstructed in an effort to build a family tree of major group structures. Stochastic code Serpent2 [1] was employed to generate cross-sections of selected isotopes at different reactor compositions and conditions, using the investigated energy group structures. The impact on their generation was quantified by spectral weighted deviation. The 35 major energy structures were divided into three basic families. The key parameters distinguishing them were their applicability to thermal or fast reactors and their applicability in neutronic or multiphysics investigations. A sensitivity threshold of the generated cross-sections over the group structure resolution was investigated. The aim was to identify a group structure with very low dependency on the actual reactor spectrum.

In this paper CFD analyses of mixed convection in bare tubes cooled with water at supercritical pressure is presented. The study was carried out using the FLUENT code for upward flow in vertical tubes with a heated length of 4 m and an inside diameter of 10 mm at relatively low water mass flux ( G ≈500 kg/m 2 s) and heat fluxes q (from 239 to 310 kW/m 2 ). Various models of turbulence have been tested. The results of the studies demonstrated a reasonable good agreement between CFD predictions and experimental data on the heat transfer coefficient and internal-wall temperature with use the SST turbulence model. Comparison of the CFD simulation data, which correspond to the presence or absence of the buoyancy forces, was performed. The regularities of the influence of these forces on the damping of turbulent transport, the deformation of the radial profiles of velocity and temperature along the channel length, the reduction of the heat transfer coefficients, etc. were studied. The features of the motion of the pseudo-phase transition within various conditions are presented.

When considering the safety of the reactor after the core melted, the external reactor vessel cooling (IVR-ERVC) is regarded as one of the most prominent method and is now widely being studied. And in order to apply this method more efficiently, CHF is the utmost part because it limits the upper threshold of the cooling effect. There has already been a large number of investigation on the CHF both by experiment and simulation. But for the experiment, most of them used the copper for its high thermal conductivity. However, the lower plenum is actually composed of the carbon steel instead of copper since the reactor pressure vessel and some core catchers in nuclear power plants are made of this material. This CHF experiment here used carbon steel plate on a downward face. The carbon steel plate was attached to the copper base by solder. The results of the carbon steel plate experiment were then analyzed. After polishing the surface by sandpapers, the heat flux is being increased until CHF was reached. We found that the surface is somehow changed during the heating process. This process was repeated several times, and the growing tendency of the CHF was found. Detailed images of the heating surface acquired by high speed camera under different heat fluxes were also obtained and analyzed. It was found that the more the oxidization was, the fewer bubbles were generated and the bigger the CHF was. Finally some theories of the CHF mechanism were also analyzed. The authors hope that study could shed some light on surface effect on causing different CHF.

In case of a severe accident scenario in a Sodium cooled Fast Reactor (SFR) such as the ASTRID demonstrator, the fuel might melt and interact with the coolant i.e. liquid sodium. This molten Fuel Coolant Interaction (FCI) can generate an energetic vapor explosion that can jeopardize the reactor structures. The yield of the vapor explosion is strongly dependent on the local distribution of the fragmented melt with respect to the local vapor fractions. The medium is composed of three phases, i.e. corium, liquid sodium and vapor sodium. Thus, a study of the three phase distribution within the system is a key to understand the extent of the explosion. PLINIUS-2, the future large mass experimental platform of CEA Cadarache will be dedicated to conducting experiments to understand the behavior of prototypic corium in case of severe accidents. In order to study these interactions, a high energy X-ray imaging system is being developed. This system consists of a 15 MeV Linear accelerator producing high energy X-rays with significantly high flux, which are attenuated as it passes through the highly dense test section. The transmitted radiation is detected and re-emitted as visible light by the GADOX screen coupled to the CMOS camera. Using this system to study the interaction between corium and sodium is particularly challenging due to the small corium particulates of the size of the order of 1 mm. The qualification of the foreseen radioscopy system on the visualization of such an interaction requires the development of physical phantoms. This paper presents the preliminary simulations of expected images of corium fragments in sodium, vapor bubbles and vapor film around the fragments. The simulations are carried out using a CEA Cadarache in-house tool MODHERATO, which produces radiographic images in satisfactory agreement with the real time imaging. The simulation of particles is based on the knowledge of interaction phenomenology gained from past experiments and on the statistical analysis of the size of corium particles formed. The models which, according to MODHERATO results, qualify to be detected and resolved, help manufacturing physical phantoms to conduct the experiments.

RELAP/SCDAPSIM is a best-estimate nuclear tool designed to analyze the behaviour of reactor systems during normal and accident conditions. Three main versions of RELAP/SCDAPSIM are currently used by program members and licensed users to support a variety of activities. RELAP/SCDAPSIM/MOD3.4 is the version of the code used by licensed users and program members for critical applications such as research reactors and nuclear power plant applications. Even though the code was initially designed for LWRs (Light Water Reactors), Politehnica University of Bucharest demonstrated the applicability of the RELAP/SCDAPSIM code for CANDU (CANada Deuterium Uranium) reactors analyses, by simulating some of the most important postulated accident transients (i.e. large break LOCA, main steam line break, the natural circulation in the heat transport system). Current trends refer to the BEPU (Best Estimate Plus Uncertainty) approach in the safety analysis of nuclear reactors. BEPU is a modern and technically consistent approach has been built upon best estimate methods including an evaluation of the uncertainty in the calculated results. ISS and UPC started the development of an uncertainty evaluation package to RELAP/SCDAPSIM/MOD4.0 code version which is currently implemented in MOD3.4 version of the code also. The uncertainty evaluation capability is implemented as an alternative run mode, the “uncertainty” mode, which allows the automatic execution of an uncertainty analysis based on the probabilistic approach. A complete uncertainty analysis using RELAP/SCDAPSIM/MOD3.4 code requires the execution of three related phases, namely the “setup” phase, the “simulation” phase consisting of several executions, and the “postprocessing” phase. The uncertainty data has to be supplied for the two types of parameters, the “input treatable” and the “source correlation” quantities. The required information is the probability distribution function and its characteristic parameters. This paper is mainly focused on the application of the uncertainty package in CANDU reactors accident analysis and it describes the steps to perform the uncertainty analysis, the uncertainty selected parameters (input treatable and source correlation parameters), the calculation results with RELAP/SCDAPSIM and some conclusions.

As an effective measure to improve the reactor’s inherent safety feature, natural circulation is widely used in current integrated reactor design. The thermal-hydraulic performance of a flashing-driven nature circulation integrated pressurized water reactor (NC-IPWR) is studied, by taking IP100 reactor as reference. The simulation model of the reactor is established by RELAP5 code. A control system is designed based on the operation characteristics of the reactor. Both steady-state and dynamic performance of the reactor are analyzed and the rationality of the control strategy is verified in this work. The results demonstrate the operation characteristics of the IP100 reactor, and the dynamic performance of the reactor during power variation is discussed in detail. The control strategy that keeps the steam pressure and the core outlet temperature constant shows good performance under normal operation conditions. The obtained analysis results are significant for deeper understanding and improving the operation characteristics of the IP100 reactor.

In the pebble-bed high temperature reactor under construction in China, called HTR-PM, the spherical fuel elements continuously flow downward in the cylindrical core. After the discharge, the burnup of each pebble is checked at the core outlet and, according to the achieved burnup level, the pebble might be disposed or reinserted into the upper section of the core, distributing randomly in the radial direction and defining a variable number of passes necessary to achieve the average maximum burnup of 90 MWd/kgU. Discrete Element Method (DEM) simulations have been carried out to achieve a clear understanding of the movement of 420,000 fuel pebbles in the HTR-PM core. At the same time, neutronic properties have been investigated for a single pebble and for the full core with Serpent 2 Monte Carlo code in order to perform a parametrization of the one-group microscopic cross sections at the core-level. The pebble movement has been coupled with the neutronic behavior of a single pebble in a dedicated burnup script called Moving Pebble Burnup (MPB), developed in Matlab. 3,000 single pebble burnup histories were simulated to obtain sufficient statistics and insight on the burnup process in the HTR-PM. The decrease of the average burnup gained per single pass implies that a miss-handling of recirculated fuel elements is unlikely to lead to exceedance of the maximum allowed burnup of 100 MWd/kgU. Furthermore, the core demonstrates a self-compensation effect of burnup, meaning that it always compensates burnup under- or over-runs in the successive passes. Finally, it is possible to conclude that the fuel cycle of the HTR-PM, as it has been laid out, is well-designed and feasible.

The function of Nuclear Power Plants (NPPs) is to produce electricity. However, in order to do that electrical power is also required for the house load for operation and to keep the safety of the plant intact. The Electrical Power Systems (EPSs) are specially designed both for normal plant operation and for particular conditions other than normal operation. In this way, NPP safety is maintained and continuousness of electrical power supplies is available despite temporary interruptions or flows through regular operation and post-shutdown. The power for a plant comes from different and reliable power sources that are physically and electrically isolated. The overall EPSs design intends to assure that any single failure or malfunction will affect only a single source of power supply and the other alternative sources should not be affected. Despite the separation between electrical power sources, malicious cyberattacks [IAEA NSS 8] might disrupt the power flow and interrupt the normal operation of a NPP [IAEA NSS 17]. Therefore, to intact the nuclear safety and security of a nuclear power plant, it is very important to incorporate cybersecurity considerations into the EPS design and implementation. Particularly, threat scenarios for EPS should be developed and evaluated before a cyberattack occurs. This will help to identify possible vulnerabilities, their impact and consequences that we want to prevent. The classification of attacks and the modelling of attacks on critical infrastructure are already described in [1] as part of our SMARTEST R&D project. Here, in this paper, we are describing examples of cyber threat scenarios for the EPSs and EPS interfaces. These are necessary with regard to specifying the anticipated security impact to a physical process controlled by the Instrumentation and Control (I&C). This serves as an input for the security analysis and the closed loop virtual validation via simulation and Fault Tree Analysis (FTA).

Due to the increase of computing efficiency and power, full-resolution two-phase flow simulations have become a practical research tool for model development and analysis of reactor flows. The expansion of state-of-the-art high performance computing (HPC) facilities allows for the use of direct numerical simulation (DNS) coupled with Interface Tracking Methods (ITM) to perform full resolution simulations. Given adequate spatial and temporal resolution, DNS can resolve all relevant turbulent scales, allowing for the extraction of high quality and detailed turbulent and two-phase flow numerical data for use in model development. While larger scale bubbly flow DNS are becoming ever more affordable, it is still computationally expensive due to the requirements of the spatial discretization. This presents the largest obstacle for future applications of DNS. For this reason, mesh adaptation techniques are sought after to reduce the computational expense of bubbly flow simulations in complex geometries. By fully resolving only the areas of specific interest, the computational costs of DNS can be reduced. Grid refinement can be based on the location of the interface between the two phases, area of greatest turbulent intensity, averaged bulk fluid velocity data, or the prediction of bubble movement. Coupled with an advanced bubble tracking algorithm, the path of individual bubbles moving through the computational domain can be predicted, and the computational mesh refined within the path area. This refinement can create tracks of greater resolution for the bubbles to move through in the domain, while keeping the bulk resolution of the mesh coarser. Through these means, the overall cost of the simulation is reduced, while high quality numerical data is still obtainable. This work outlines the enhancement of existing mesh adaptation algorithms to implement the bubble tracking refinement, and its practical applications to full resolution two-phase flow simulations.

UO2-BeO composite fuel may enable Light Water Reactors (LWRs) to have better safety due to its higher thermal conductivity. Much work have been done on the analysis of UO2-BeO fuel performance during LWRs steady state and Loss of Coolant Accident (LOCA) conditions using hypothetical thermal properties and behaviors models, leading to much uncertainty of the results. In this paper, firstly, fuel swelling and densification models for UO2-BeO fuel were developed based on Halden experiment data. Secondly, UO2-BeO fuel thermal properties and behaviors models have been coded in FRAPCON4.0 and FRAPTRAN2.0 after an evaluation of their applicability to UO2-BeO performance simulation. Then, both UO2-BeO composite fuel and traditional UO2 fuel performance during normal conditions and RIA were done in this paper by modified version of FRAPCON4.0 and FRAPTRAN2.0. Finally, comparisons between UO2-BeO and UO2 performance were conducted. The results shows that the peaking temperature of fuel can be reduced about 200K and 150K during normal conditions and RIA by adopting UO2-BeO, respectively. At the same time, the onset of pellet-cladding mechanic interaction (PCMI) can be delayed about 100days during normal conditions and the weakened PCMI effect can be expected during reactivity insertion accidents (RIA) due to the lower thermal expansion coefficient and temperature distribution for UO2-BeO composite fuel. Also, enthalpy stored in UO2-BeO fuel is reduced about 1/5 compared with that of UO2. However, fission gas release ration of UO2-BeO was a bit larger than that of UO2 due to its higher average burnup. And, further experiments stilled required to gain data for UO2-BeO during high burnup, like possibly reduced thermal conductivity and fission gas release threshold.

Betavoltaic cells are nuclear batteries ideal for low-power applications for extended periods of time without maintenance or replacement. Betavoltaics function similarly to photovoltaic (solar) cells where instead of using sunlight, beta particles are used to generate electron-hole pairs within a semiconductor p-n junction to generate current. Even though there have been multiple demonstrations, betavoltaic performance has not been extensively studied. To accurately predict betavoltaic performance, which is important for a device in operation without maintenance for elongated periods, all parameters are required to predict potential fluctuations in cell performance, such as doping densities and resistances for semiconductor variation and absorption coefficients for beta-generated current. However, not all parameters are easily measured, especially when the p-n junction is constantly under irradiation and cannot be separated from the source. Critical parameters were characterized experimentally with the betavoltaic cell by performing capacitance-voltage to determine doping densities and performing current-voltage characterization tests to determine resistances on multiple NanoTritium™ cells, while absorption coefficients were determined from MCNP6 simulations. Experiments indicated that series resistance R s was 1 × 10 6 Ω, while shunt resistance R sh was 2 × 10 8 Ω from I-V characterization, while doping density N D was determined to be 1 × 10 17 cm −3 from C-V characterization. Absorption coefficient α was found to vary with semiconductor material and incoming beta energy and used in conjunction with critical parameters from experimentation to accurately model betavoltaic cell performance similar to experimental results. Both implicit equations and explicit estimations were compared to model betavoltaic cell performance.

The pressurizer is responsible for controlling the pressure and temperature in the primary circuit of the nuclear power plant. It is basically a pressure vessel, filled partially with water and partially with steam at saturation state. The controlling process can be described by two regimes, either by the self-control regime or by the automatic control regime. This paper is describing the simplified automatic control regime on the model of the pressurizer in the primary circuit of VVER 1000 made in Dymola software with the ClaRa library. Reactor power change and corresponding steam generator power change are the actuators in the simulation. The behaviour of the coolant level and pressure in the pressurizer is simulated in the model and it is then compared with data provided by the supplier of VVER 1000.