ALBERT

Description: Passive safety system is the core feature of advanced nuclear power plant (NPP). It is a research hotspot to fulfill the function of passive safety system by improving the NPP natural circulation capacity. Considering that the flow behaviors of stopped pump pose a significant effect on natural circulation, both experimental and computational fluid dynamics (CFD) methods were performed to investigate the flow behaviors of a NPP centrifugal pump under natural circulation condition with a low flow rate. Since the pump structure may lead to different flows depending on the flow direction, an experimental loop was set up to measure the pressure drop and loss coefficient of the stopped pump for different flow directions. The experimental results show that the pressure drop of reverse direction is significantly greater than that of forward direction in same Reynolds number. In addition, the loss coefficient changes slightly while the Reynolds number is greater than 8 × 104; however, the coefficients show rapid increase with the decrease in Reynolds number under lower Reynolds number condition. According to the experimental data, an empirical correlation of the pump loss coefficient is obtained. A CFD analysis was also performed to simulate the experiment. The simulation provides a good accuracy with the experimental results. Furthermore, the internal flow field distributions are obtained. It is observed that the interface regions of main components in pump contribute to the most pressure losses. Significant differences are also observed in the flow field between forward and reverse condition. It is noted that the local flows vary with different Reynolds numbers. The study shows that the experimental and CFD methods are beneficial to enhance the understanding of pump internal flow behaviors.

Description: The chemical forms of important fission products (FPs) in the primary circuit are essential to the source term analysis of high-temperature gas-cooled reactors because the volatility, transfer, and diffusion of these radionuclides are significantly influenced by their chemical forms. Through chemical reactions with gaseous impurities in the primary circuit, these FPs exist in diverse chemical forms, which vary under different operational conditions. In this paper, the chemical forms of cesium (Cs), strontium (Sr), silver (Ag), iodine (I), and tritium in the primary circuit of the Chinese pebble-bed modular high-temperature gas-cooled reactor (HTR-PM) under normal conditions and accident conditions (overpressure and water ingress accident) are studied with chemical thermodynamics. The results under normal conditions show that Cs exists mainly in the form of Cs2CO3 at 250°C and gaseous form at 750°C, and for I and Ag, Ag3I3 and Ag convert to gaseous CsI and AgO, respectively, with increasing temperature, while SrCO3 is the only main kind of compound for Sr. It is also observed that new compounds are generated under accidents: I exists in HI form when a water ingress accident occurs. Regarding tritium, the chemical forms of FPs change little, but compounds need higher temperature to convert. Furthermore, hazard of some FPs in different chemical forms is also discussed comprehensively in this paper. This study is significant for understanding the chemical reaction mechanisms of FPs in an HTR-PM, and furthermore it may provide a new point of view to analyze the interaction between FPs and structural materials in reactor as well as their hazards.

Description: For most of the remote maintenance activities of equipment in a hot cell, replacing breakdown modules is preferred over in situ repair because of insufficient space in the cell and the limited operability of remote handling tools. In such cases, the maintenance operation can be decomposed into transport of the new modules to the failed equipment, replacement of the broken modules with new ones, and then transport of the broken parts to the reserved space for further repair or disposal. In this respect, transfer is the most basic operation during remote maintenance, which is also true for the maintenance of pyroprocessing equipment. Hence, this paper proposes a maintenance automation framework for automated pyroprocessing equipment from the standpoint of module transfer. For the maintenance automation framework, maintenance-related functions and events are defined, and they are integrated with the pyroprocess automation framework. The proposed framework is verified by a case study on the maintenance of a large module through a hardware-in-the-loop simulation.

Description: The stability of W against U, rare-earth (RE) elements, Cd, and various chlorides was evaluated by melting and distillation testing. Three runs were performed with a W crucible to examine its reactivity: (i) RE melting by induction heating, (ii) salt distillation test of U-dendrite and various chlorides, and (iii) Cd distillation test from U–Cd alloy. The W crucible remained stable after the RE melting test using induction melting, exhibiting its applicability for induction heating systems. The salt distillation test with the W crucible at 1050°C exhibited the stability of W against U and various chlorides, showing no interaction. The Cd distillation test with the W crucible at 500°C showed that the crucible was very stable against Cd, maintaining a shiny surface. These results reveal that the W crucible is stable under operation conditions for both salt and Cd distillation, suggesting the high potential utility of W as a crucible material for application in cathode processes in pyroprocessing.

Description: In a fast reactor, we evaluated a new core concept that prevents severe recriticality after whole-scale molten formation in a severe accident. A core concept in which Duplex pellets including neutron absorber are loaded in the outer core has been proposed. Analysis by the continuous energy model Monte Carlo code MVP using the JENDL-4.0 nuclear data library revealed that this fast reactor core has large negative reactivity due to fuel melting at the time of a severe accident, so that the core prevents recriticality. Regarding the core nuclear and thermal characteristics, the loading of Duplex pellets including neutron absorber in the outer core caused no significant differences from the normal core without Duplex pellets.

Description: This paper describes the development of a discrete event simulation model using the FlexSim software to support planning for soil remediation at Korean nuclear power plants that are undergoing decommissioning. Soil remediation may be required if site characterization shows that there has been radioactive contamination of soil from plant operations or the decommissioning process. The simulation model was developed using a dry soil separation and soil washing process. Preliminary soil data from the Kori 1 nuclear power plant was used in the model. It was shown that a batch process such as soil washing can be effectively modeled as a discrete event process. Efficient allocation of resources and efficient waste management including volume and classification reduction can be achieved by use of the model for planning the soil remediation process. Cost will be an important criterion in the choice of suitable technologies for soil remediation but is not included in this conceptual model.

Description: Many tons of porous carbon materials (including BC and IG-110) are contained in HTGR, which are serving as structural material and fuel matrix material. These materials would absorb moisture and other impurities when exposed to the environment, and these impurities (especially moisture) absorbed in the carbon material must be removed before the reactor operation to prevent corrosion reaction at high temperature (more than 500°C). As the pore microscopic structure characteristic is the significant factor affecting the gas adsorption and flow in the porous materials, the detailed 3D pore structures of the carbon materials (BC and IG-110) in HTGR were studied by Micro-XCT and HPMI methods in this paper. These pore structure characteristics include pore geometry, pore size distribution, and pore throat connectivity. The test results show that the pore size distribution of BC material is wide, and the pore diameter is obviously larger than that of IG-110. Pore connections in BC show radial shape connections at some special points, and the pore connectivity in IG-110 is very complex and presents a huge complex 3D pore network.

Description: Reactor pressure vessel (RPV) support is a key safety facility which is categorized as Class 1 in the ASME nuclear safety design. The temperature distribution of RPV support is one of the key considerations for the concrete safety contacting with the bottom of the support. So it is necessary for accurate evaluation on the temperature field characteristics of RPV support, especially the bottom of support. This paper investigates the temperature field characteristics of modified RPV support which will be applied to a large advanced pressurized water reactor. A support entity is manufactured in a ratio of 1:1, and its temperature distribution is measured under simulated reactor operating conditions. Numerical simulation is also used to validate the results by the developed CFD model. The results show that under the operating conditions, of which the inlet cooling air temperature is 35.35°C and the velocity is 6.25 m/s, the temperature distribution of modified RPV support bottom is uneven, and the highest temperature is around 38°C, which is much lower than the demanding design temperature 93.3°C. Therefore, the design of the modified RPV support is reliable. In addition, the results of numerical simulation agree well with the experimental results with the error less than ±4°C, which ensures the reliability of the conclusion. The effects of inlet cooling air temperature and velocity on the RPV support temperature distribution are further studied. Both the temperature decrease and velocity increase can reduce the RPV support temperature. But the effect of inlet cooling air temperature is more obvious than inlet cooling air velocity. So the best way to improve air cooling capacity is to decrease the support inlet cooling air temperature. The results can provide a good guidance to the design of RPV support for the subsequent large advanced pressurized water reactor.

Description: In order to resolve the situations of nonuniform coolant flow distribution and insufficient vortex suppression, the existing Pressurized Water Reactor (PWR) usually adopts complex coolant mixing structures. However, those structures will greatly increase the complexity and maintenance cost of the system. To solve this problem, a trimming-based design method is proposed in this paper for the complex system and applies it to the design process of the PWR coolant flow distribution device. The function model of the coolant flow distribution system is built based on its functional analysis, and, according to the result of the component feature analysis, the columns and part of the basket are trimmed in order to simplify the overall structure of the system. To further solve the technical contradictions occurred in the simplified system, the contradiction solving tools of TRIZ theory are adopted. By setting the stereo flow equalizing plate, which can strengthen the function of flow distribution and vortex suppression, a coolant flow distribution device for PWR based on dome structure is obtained finally. This device owns a simple structure with good effect on coolant flow distribution and vortex suppression, which can achieve the goal of uniform coolant flow distribution of the system effectively.

Description: Evaluation of aerosol deposition in the containment vessel is an important step for the assessment of radioactive material release to the environment. ART Mod 2 is a calculation code that is used for evaluation of aerosol deposition in the containment vessel. The authors modified aerosol deposition models of ART Mod 2, namely, gravitational settling model, Brownian diffusion model, diffusiophoresis model, and thermophoresis model in order to increase potential of capturing the deposition phenomena. This study aims to compare the simulated results of modified ART Mod 2 with aerosol deposition of cesium compounds in the containment vessel of Phébus FPT3 experiment, in order to validate modified ART Mod 2 code. It is found that aerosol deposition using modified ART Mod 2 agrees with Phébus FPT3. Prediction of Brownian diffusion is significantly improved due to the consideration of turbulent damping process. Cesium mass flow rate and aerosol size are factors that can significantly influence the uncertainty of the results. When conditions of single volumes are carefully selected to match those of the Phébus FPT3 experiment, modified ART Mod 2 can predict aerosol deposition in Phébus FPT3 with relative accuracy.

Description: For sequentially collected data, this paper introduces a lag-one differencing method to estimate the random error standard deviation δR and then uses the estimate δ^R to calculate a change detection threshold in a moving window method to detect shifts in the short-term systematic error. Performance results on simulated and real data are presented. Fortunately, the impact of having to perform change detection on the estimated short-term systematic and random error variances is anticipated to be modest or small. The motivating example arises from facilities under nuclear safeguards agreements, where inspector data collected during International Atomic Energy Agency (IAEA) verifications are compared to corresponding operator data. The differences between the operator and inspector values are evaluated using an application of analysis of variance (ANOVA). Typically, it is assumed that short-term systematic errors change across inspection periods, so inspection periods form the groups used in the ANOVA. In some data sets, it appears that the short-term errors have changed at other times, so change detection methods could be used to detect the actual change times.

Description: There are results of long-term thermal aging of samples of irradiated and nonirradiated FA jacket and nonirradiated fuel element cladding at a temperature range from 300 to 550°C in argon, to 600°C in air. Materials have been studied before and after thermal tests. The forecast estimation of expected corrosion damage of barrier material at the radionuclide release from spent fuel assemblies of BN-350 reactor into environment during dry storage for 50 years was carried out.

Description: Because a pool scrubbing is important for reducing radioactive aerosols to the environment for a nuclear reactor in a severe accident situation, many researches have been performed. However, decontamination factor (DF) dependence on aerosol concentration was seldom considered in an aerosol number concentration with limited aerosol coagulation. To investigate an existence of DF dependence on the concentration, DF in a pool scrubbing with 2.4 m water submergence was derived from aerosol measurements by light scattering aerosol spectrometers. It was observed that DF increased monotonically with decreasing particle number concentration in a constant thermohydraulic condition: a gradual increase from 10 to 32 in the range of 1.3×1011 - 8.0×1011/m3 at the inlet and a significant increase from 32 to 77 in the range of 3.6×1010 - 1.3×1011/m3. Two validation experiments were conducted in the range with the gradual DF increase to confirm whether the DF dependence is a real pool scrubbing phenomenon. In addition, characteristics of the DF dependence in different water submergences were investigated experimentally. It was found that the DF dependence became more significant in higher water submergence. Significant DF dependence was observed in the condition of the water submergence higher than 1.6 m and the inlet particle number concentration less than around 1×1011 /m3. It is recommended to perform further analysis for the DF dependence mainly in such condition since it could make a difference to both experiment and model of the pool scrubbing.

Description: After the successful construction and operation experience of the 10 MW high-temperature gas-cooled reactor (HTR-10), a high-temperature gas-cooled pebble-bed modular (HTR-PM) demonstration plant is under construction in Shidao Bay, Rongcheng City, Shandong province, China. An online gross monitoring instrument has been designed and placed at the exit of the helium purification system (HPS) of HTR-PM and is used to detect the activity concentration in the primary circuit after purification. The source terms in the primary loop of HTR-PM and the helium purification process were described. The detailed configuration of the gross monitoring instrument was presented in detail. The Monte Carlo method was used to simulate the detection efficiency of the monitoring system. Since the actual source terms in the primary loop of HTR-PM may be different than the current design values, a sensitivity analysis of the detection efficiency was implemented based on different relative proportions of the nuclides. The accuracy and resolution of the NaI(Tl) detector were discussed as well.

Description: The assessment of hydrogen release, distribution, and mitigation measures in the containment of a nuclear power plant is increasingly based on code calculations. These calculations require state-of-the-art experiments to benchmark the codes against them. Two of these experiments are presented in this paper. These experiments were conducted in the PANDA facility (Switzerland) in the framework of the OECD/NEA HYMERES project. The experiments consider natural circulation flow in a two-room type containment where flow loops can form between the inner and the outer zones. During normal operation these zones are separated and in the case of an accident they become either connected by the opening of rupture disks, convective foils, and dampers or connected by bursting of doors and opening of other connections between compartments. For the experiments considered here one lower PANDA-vessel represents the steam generator (SG) tower and the inaccessible area whereas the other vessel represents the outer room area. The lower vessels are isolated from one another except for a small aperture that represents the damper. The two upper vessels—representing the containment dome—are connected to the lower vessels through tubes. The scenario consisted of four phases. In phase 1, a high steam mass flow rate was injected in the vessel representing the SG tower. After the relaxation phase 2, helium (representing hydrogen) was injected in the same vessel (phase 3). Finally in phase 4 no active interventions were done until the end of the test. Two tests were conducted to evaluate the developing helium transport by the natural circulation flow: one with and one without damper (by closing the aperture). The results showed that a two-room containment (TRC) mixing scenario can be well represented with the PANDA facility. It is found that, with the mixing damper open, a global natural circulation loop develops over all four vessels, whereas with closed damper the natural circulation loop is established only between the three vessels representing the inner zone and the upper dome. It is shown that the presence of the damper has a strong effect on the resulting helium content in the inner zone with 3 times less helium at the end of the test compared with the configuration without damper. The formation of a stable helium stratification in the upper vessels was observed in the presence of the open damper.

Description: The LOCA (loss of coolant accident) is a kind of severe accident in the operation of PHWR (pressurized heavy water reactor) as well as other nuclear facilities, and possible cause of LOCA can be counted on the ballooning of pressure tube (PT) contacted to the outer calandria tube (CT) in the moderator system of CANDU-6 reactors. In the paper, we simulated the 150-kW experimental facility proposed by IAEA/ISCP, modeling the transient creeping behavior of pressurized tube heated with thermal radiation between the gaps of two concentric pipes. The outer boundary is simplified with a switched model that depends on the local temperature. With a multiphysical model supported by a commercial code, COMSOL multiphysics, the unsteady phenomena are simulated with models concerning various kinds of mechanics such as thermodynamics, nonlinear structural dynamics, and two-phase boiling heat transfer models.

Description: BWRs are thus far the simplest energy systems to transform fission energy into electrical power. However, there are still many aspects in their operation that, under certain conditions, may induce BWR unstable behavior. The default indicator to study BWR unstable behavior is the Decay Ratio (DR). However, due to the fact that BWRs show very complex responses under instability and responses that may even be chaotic, the DR might not be a suitable choice to rely on to accommodate for such intricate behavior. In this work a novel methodology based on the Sample entropy (SampEn) and the noise-assisted multivariate empirical mode decomposition (NA-MEMD) is introduced. Such methodology was developed thinking for a real time-implementation of a stability monitor. The proposed methodology was tested with a set of signals that stem from several nuclear power plants in operation today that have experienced in the past unstable events, each one of a different nature.

Description: Vigorously developing nuclear power is the main development direction of current renewable energy. In the nuclear environment, in order to avoid nuclear radiation damage to maintenance personnel and improve the efficiency of nuclear reaction, it is necessary and urgent to realize automatic replacement of vulnerable parts in the electron gun. As the key equipment for the generation and control of nuclear reactions in nuclear reactors, electron guns have been widely used in nuclear power plants of traveling wave reactors. However, the “high-voltage conductive ring” in electron guns is a vulnerable part. It is likely to cause nuclear reactor accidents when the vulnerable part is damaged. Automatic replacement of vulnerable parts is an important part of the entire maintenance equipment. Considering the entire maintenance equipment and the working environment, an innovative design process for vulnerable parts replacement is established. Under the guidance of the process, in order to ensure the continuity of a series of maintenance actions, the technical contradiction resolution theory is first used to conduct the overall analysis of the general direction to obtain the design layout. Then, the contradiction resolution theory and the object-field model analysis are utilized to get and improve the detailed design of the device mechanism. The theory of TRIZ can help us to get the overall mechanical structure design that meets the engineering requirements. The device is designed with a replacement part adjustment scheme to ensure the completion of the maintenance actions. Furthermore, the design provides a solution to the possible jamming phenomenon in the automatic maintenance process and achieves the maximum use efficiency of the storage and replacement of vulnerable parts.

Description: This study was carried out to examine the potential of antimony tri-iodide (SbI3) as a material for radiation detectors that operate at room temperature. SbI3 is a compound semiconductor with an AsI3-type crystal structure, high atomic number (Sb: 51, I: 53), high density (4.92 g/cm3), and a wide band-gap energy (2.2 eV). In addition, crystalline SbI3 is easy to grow by conventional crystal growth techniques from melting phase because the material exhibits a low melting point (171°C) and undergoes no phase transition in the range of its solid phase. In this study, SbI3 crystals were grown by the Bridgman method after synthesis of SbI3 from 99.9999% pure Sb and 99.999% pure I2. The grown crystals consisted of several large grains with red color and were confirmed to be single-phase crystals by X-ray diffraction analysis. SbI3 detectors with a simple planar structure were fabricated using the cleavage plates of the grown crystals, and the pulse-height spectra were recorded at room temperature using an 241Am alpha-particle (5.48 MeV) source. The detector showed response to the alpha-particle radiation.

Description: This paper presents capacity of the passive decay heat removal system (DHRS) operated under the natural circulation conditions to remove decay heat inside the main vessel of the Lead-bismuth eutectic cooled Fast Reactor (LFR). The motivation of this research is to improve the inherent safety of the LFR based on the China Accelerator Driven System (ADS) engineering project. Usually the plant is damaged due to the failure of the main pumps and the main heat exchangers under the Station Blackout (SBO). To prevent this accident, we proposed the DHRS based on the diathermic oil cooling for the LFR. The behavior of the DHRS and the plant was simulated using the CFD code STAR CCM+ using LFR with DHRS. The purpose of this analysis is to evaluate the heat exchange capacity of the DHRS and is to provide the reference for structural improvement and experimental design. The results show that the stable natural circulations are established in both the main vessel and the DHRS. During the decay process, the heat exchange power is above the core decay heat power. In addition, in-core decay heat and heat storage inside the main vessel are efficiently removed. All the thermal-hydraulics parameters are within a safe range. Moreover, the highest temperature occurs at the upper surface of the core. A swirl occurs at the corner of the lateral core surface and some improvements should be considered. And the natural circulation driving force can be further increased by reducing the loop resistance or increasing the natural circulation height based on the present design scenario to enhance the heat exchange effect.

Description: Source term analysis is important in the design and safety analysis of advanced nuclear reactor and also provides a radiation safety analysis basis for Modular High-Temperature Gas-Cooled Reactor (HTR). High-Temperature Gas-Cooled Reactor-Pebble-bed Modules (HTR-PM) design by China is a typical Gen-IV and due to different safety concepts and systems, the implements of source term analysis in light water reactors are not entirely applicable to HTR-PM. To solve this problem, HTR-PM Source Term Analysis Code (HTR-STAC) has been developed and related V&V has been finished. HTR-STAC consists of five units, including LOOP (Primary Circuit Source Term Analysis Code), NORMAL (Normal Condition Airborne Source Term Analysis Code), ARCC (Accident Release Category Calculation code), CARBON (C-14 Source Term Analysis Code), and TRUM (Tritium Source Term Analysis Code). LOOP and NORMAL may be used as calculating primary circuit coolant radioactivity and the release of airborne radioactivity to the environment under normal operating conditions of HTR-PM, respectively. The code ARCC composed of several source term analysis programs in the different typical accidents scenario, including SGTR (Steam Generator Tube Rupture), LOCA (Loss of Coolant Accident), and the Transient Process, is compiled based on the results given by LOOP and NORMAL. CARBON and TRUM are developed to calculate the productions of C-14 and H-3 through a different mechanism. Furthermore, the V&V has been performed and show some positive results.

Description: A 2 inch, cold-leg loss-of-coolant accident (LOCA) in a 900 MWe generic Western PWR was simulated using ASTEC 2.1.1 and MAAP 5.02. The progression of the accident predicted by the two codes up to the time of vessel failure is compared. It includes the primary system depressurization, accumulator discharge, core heat-up, hydrogen generation, core relocation to lower plenum, and lower head breach. The purpose of the code comparison exercise is to identify modelling differences between the two codes and the user choices affecting the results. The two codes predict similar primary system depressurization behaviour until the accumulation injection, confirming similar break flow and primary system thermal-hydraulic response calculations between the two codes. The choice of the accumulator gas expansion model, either isentropic or isothermal, affects the rate and total amount of coolant injected and thereby determines whether the core is quenched or overheated and attains a noncoolable geometry during reflooding. A sensitivity case was additionally simulated by each code to allow comparisons to be made with either accumulator gas expansion models. The two codes predict similar amount of in-vessel hydrogen generated and core quench status for a given accumulator gas expansion model. ASTEC predicts much larger initial core relocation to lower plenum leading to an earlier vessel failure time. MAAP predicts more gradual core relocation to lower plenum, prolonging the lower plenum debris bed heat-up and time to vessel failure. Beside the effect of the code user in conducting severe accident simulations, some discrepancies are found in the modelling approaches in each code. The biggest differences are found in the in-vessel melt progression and relocation into the lower plenum, which deserve further research to reduce the uncertainties.

Description: In order to study the evolution process and hydraulic characteristics of pressurizer insurge in CPR1000 pressurized water reactor (PWR), the computational fluid dynamics (CFD) of three-dimensional unsteady heat transfer was used to capture the temperature and velocity fluctuation intensity of the mixing of the hot and cold. The results show that Realizable k–ε Turbulence Model combined with VOF multiphase heat transfer model can effectively predict the development trend of pressurizer insurge process. The small diameter pressurizer surge line of CPR1000 enhances the intensity of velocity fluctuation. From the essence of flow and heat transfer, it is concluded that buoyancy force can increase the degree of fluctuation and make an accelerated effect on the influx cold fluid. The electric heater inside the pressurizer should be arranged as far as possible in z0.45m; it is beneficial to improve its harsh operating environment. This research can provide reference for the structural design of pressurizer and the layout optimization of the electric heater.

Description: Nonuniform distribution of tri-structural-isotropic (TRISO) fuel particles in a spherical fuel element (SFE) may increase the failure probability of the SFE in the high-temperature gas-cooled reactor, leading to the release of fission products. To evaluate the uniformity of the TRISO particles nondestructively, 3-dimensional cone-beam computed tomography is used to image the SFE, and TRISO particles are segmented. After TRISO particle positions are identified, the Voronoi tessellation and Delaunay triangulation are used to extract several geometric metrics. Results indicate that both the Voronoi volume distribution and the nearest neighbor-distance distribution follow the log-normal distributions, which provide strong evidence that the TRISO particles are approximately randomly uniformly distributed. Further study will be focused on validating the conclusion with more SFE data.

Description: The Fukushima accident has reiterated that the seismic safety is a clear necessity in the design of nuclear power plants. To overcome the weaknesses of the plant design, appropriate measures or interventions have thus to be put in place to improve the nuclear safety. In this study, seismic isolation, widely adopted for conventional constructions, is considered as retrofit measure to provide superior performance of plant itself, even when exceptional events occur. In this paper, we numerically investigate the dynamic behaviour of a Small Modular Reactor (SMR) plant subjected to 0.6g PGA; in doing that time-history analysis has been performed assuming the reactor building with and without isolators. For that purpose, a suitable FEM model has been implemented to provide in-structure response spectra at safety relevant locations and subsystem supports. Adequate steel and concrete properties as well as isolators properties, experimentally determined, have been assumed. Results have shown the benefits of seismic isolation for NPP that can so sustain levels of loading beyond the design input and demonstrated that failure of an isolation system cannot occur before failure of the isolated structure. However, the large horizontal displacements of the structure require appropriate considerations in the layout and interfaces for interconnected systems.

Description: A new method based on shifted Chebyshev series of the first kind is introduced to solve stiff linear/nonlinear systems of the point kinetics equations. The total time interval is divided into equal step sizes to provide approximate solutions. The approximate solutions require determination of the series coefficients at each step. These coefficients can be determined by equating the high derivatives of the Chebyshev series with those obtained by the given system. A new recurrence relation is introduced to determine the series coefficients. A special transformation is applied on the independent variable to map the classical range of the Chebyshev series from to . The method deals with the Chebyshev series as a finite difference method not as a spectral method. Stability of the method is discussed and it has proved that the method has an exponential rate of convergence. The method is applied to solve different problems of the point kinetics equations including step, ramp, and sinusoidal reactivities. Also, when the reactivity is dependent on the neutron density and step insertion with Newtonian temperature feedback reactivity and thermal hydraulics feedback are tested. Comparisons with the analytical and numerical methods confirm the validity and accuracy of the method.

Description: The subcooling effect on pool boiling heat transfer using a copper microporous coating was experimentally studied in water for subcoolings of 10 K, 20 K, and 30 K at atmospheric pressure and compared to that of a plain copper surface. A high-temperature thermally conductive microporous coating (HTCMC) was made by sintering copper powder with an average particle size of 67 μm onto a 1 cm × 1 cm plain copper surface with a coating thickness of ~300 μm. The HTCMC surface showed a two times higher critical heat flux (CHF), ~2,000 kW/m2, and up to seven times higher nucleate boiling heat transfer (NBHT) coefficient, ~350 kW/m2K, when compared with a plain copper surface at saturation. The results of the subcooling effect on pool boiling showed that the NBHT of both the HTCMC and the plain copper surface did not change much with subcooling. On the other hand, the CHF increased linearly with the degree of subcooling for both the HTCMC and the plain copper surface. The increase in the CHF was measured to be ~60 kW/m2 for every degree of subcooling for both the HTCMC and the plain surface, so that the difference of the CHF between the HTCMC and the plain copper surface was maintained at ~1,000 kW/m2 throughout the tested subcooling range. The CHFs for the HTCMC and the plain copper surface at 30 K subcooling were 3,820 kW/m2 and 2,820 kW/m2, respectively. The experimental results were compared with existing CHF correlations and appeared to match well with Zuber’s formula for the plain surface. The combined effect of subcooling and orientation of the HTCMC on pool boiling heat transfer was studied as well.

Description: One of the severe accident management strategies for nuclear reactors is the melted corium retention inside the reactor pressure vessel. The work presented in this article investigates the application of in-vessel retention (IVR) severe accident management strategy in a BWR reactor. The investigations were performed assuming a scenario with the large break LOCA without injection of cooling water. A computer code RELAP/SCDAPSIM MOD 3.4 was used for the numerical simulation of the accident. Using a model of the entire reactor, a full accident sequence from the large break to core uncover and heat-up as well as corium relocation to the lower head is presented. The ex-vessel cooling was modelled in order to evaluate the applicability of RELAP/SCDAPSIM code for predicting the heat fluxes and reactor pressure vessel wall temperatures. The results of different ex-vessel heat transfer modes were compared and it was concluded that the implemented heat transfer correlations of COUPLE module in RELAP/SCDAPSIM should be applied for IVR analysis. To investigate the influence of debris separation into oxidic and metallic layers in the molten pool on the heat transfer through the wall of the lower head the analytical study was conducted. The results of this study showed that the focusing effect is significant and under some extreme conditions local heat flux from reactor vessel could exceed the critical heat flux. It was recommended that the existing RELAP/SCDAPSIM models of the processes in the debris should be updated in order to consider more complex phenomena and at least oxide and metal phase separation, allowing evaluating local distribution of the heat fluxes.

Description: Uncertainty of a severe accident code output needs to be handled reliably considering its use in safety regulation of nuclear industry. In particular, severe accident codes are utilized for probabilistic safety assessment (PSA), where the uncertainty of severe accident progress should be considered carefully due to its influence on human reliability analysis. Therefore, in this study, the uncertainty analysis of severe accident progress was performed using MELCOR code, and a total of 200 data sets of in-vessel uncertainty parameters were generated by Latin hypercube sampling method. The rank regression analysis was also performed to investigate the effect of uncertainty parameters on the severe accident progress. Sensitivity coefficients (SCs) in MELCOR such as molten clad drainage rate and zircaloy melt breakout temperature showed significant influence on relocation time and dryout time of lower plenum. However, the influence of uncertainty parameter diminished as the accident progressed.

Description: The discrete ordinates method (SN) is one of the mainstream methods for neutral particle transport calculations. Assessing the quality of the numerical solution and controlling the discrete error are essential parts of large-scale high-fidelity simulations of nuclear systems. Three error estimators, a two-mesh estimator, a residual-based estimator, and a dual-weighted residual estimator, are derived and implemented in the ARES transport code to evaluate the error of zeroth-order spatial discretization for SN equations. The difference in scalar fluxes on coarse and fine meshes is adopted to indicate the error in the two-mesh method. To avoid zero residual in zeroth-order discretization, angular fluxes within one cell are reconstructed by Legendre polynomials. The error is estimated by inverting the discrete transport operator using the estimated directional residual as an anisotropic source. The inner product of the forward directional residual and the adjoint angular flux is employed to quantify the error in quantities of interest which can be denoted by a linear functional of forward angular flux. Method of Manufactured Solutions (MMS) is adopted to generate analytical solutions for SN equation with scattering and the determined true error is used to evaluate the effectivity of these estimators. Promising results are obtained in the numerical results for both homogeneous and heterogeneous cases. The larger error region is well captured and the average effectivity index for the local error estimation is less than unity. For the series test problems, the estimated goal quantity error can be contained within an order of magnitude around the exact error.

Description: The essential power supply system is important for the nuclear safety and accident mitigation of the currently operating nuclear power plants. This system provides electrical power to the essential instrumentation and control systems of the nuclear power plant when all alternate current power sources are lost. This event is known as station blackout (SBO) event. Operational events of failure or deficiency of the essential power supply system are analyzed in this paper. The relevant events were searched and identified in four databases of operational events. The report includes events identified in French SAPIDE and German VERA operational events records for the time period 1996 to 2015. The International Atomic Energy Agency (IAEA) IRS and Nuclear Regulatory Commission (NRC) LER operational events databases were screened for relevant events that occurred in the period between 2000 and 2016. In total, 308 relevant events are identified in the SAPIDE, 103 in VERA, 56 in LER, and 15 in IRS operational events database. Classification and in-depth analysis were done on the identified events considering the following predefined categories: the plant status during the event, circumstances, affected equipment, cause of the event (direct and root), and implications of the event on plant safety. Main findings from the evaluation of the events are presented. Observations of the causes resulting in the events and potential actions that can decrease the number and consequences of the events are presented.

Description: A steam explosion in a reactor cavity makes a mechanical load of the pressure pulse, which can result in a failure of the containment isolation. To prove the integrity of the containment during the ex-vessel steam explosion, the effects of water conditions on a steam explosion have to be identified, and the impulse of a steam explosion has to be exactly assessed. In this study, the analyses for steam explosions were performed for the conditions of a partially flooded cavity and a submerged-vessel in a pressurized water reactor. The entry velocity of a corium jet for the scale of the test facility was varied to simulate the two plant conditions. The TEXAS-V code was used for simulating the phases of premixing and explosion, and the load of a steam explosion was estimated based on the pressure variation. The impulse of a steam explosion under the condition of a corium jet falling into water without a free-fall height is bigger than that under a free-fall height. The fragmented mass of corium in an explosion phase and the distribution of steam fraction are the main parameters for the total load of the steam explosion. This study is expected to contribute to analyses of a steam explosion for a severe accident management strategy.

Description: Liquid annular seals are primarily used to control the leakage in high-speed turbomachinery, especially in nuclear and petrochemical pumps. In this paper, a theoretical analysis method for dynamic characteristics of liquid seals with herringbone grooves on the stator is proposed based on bulk-flow theory. Steady-state velocities and leakage rates within the upstream and downstream spiral parts and the middle plain part taking account of the pumping effects are figured out first with the inertia term of the fluid within the whole seal. Then, the dynamic characteristics of the whole seal are solved based on Childs’ finite-length solutions and verified by comparing with experimental hydraulic forces. Moreover, characteristic coefficients and instability parameters of the herringbone-grooved teeth-on-stator (TOS) seals and teeth-on-rotor (TOR) seals of the same size under different pressure differences are predicted and compared in detail. The influences of the lengths of constituent parts on the dynamic characteristics and instability parameters of the model seals are theoretically investigated. The results show that the stability of the TOS seal is much better than that of the TOR seal under most operating conditions. And the lengths of the middle plain part significantly affect the dynamic characteristics and the stability parameter.

Description: Recompression supercritical carbon dioxide (SCO2) Brayton Cycle for the Chinese Initiative Accelerator Driven System (CiADS) is taken into account, and flexible thermodynamic modeling method is presented. The influences of the key parameters on thermodynamic properties of SCO2 Brayton Cycle are discussed and the comparative analyses on genetic algorithm and pattern search algorithm are conducted. It is shown that the cycle parameters such as turbine inlet temperature, pressure ratio, outlet temperature at the hot end of condenser, and terminal temperature difference of regenerator 1 and regenerator 2 have significant effects on the cycle thermal efficiency. The calculation results indicate that pattern search algorithm has better optimization performance and quicker calculating speed than genetic algorithm. The result of optimization of the parameters for CiADS with supercritical carbon dioxide Brayton Cycle is 35.97%. Compared with other nuclear power plants of SCO2 Brayton Cycle, CiADS with SCO2 Brayton Cycle does not have the best thermal efficiency, but the thermal efficiency can be improved with the reactor outlet temperature increases.

Description: The code assessment typically comprises basic tests cases, separate effects test, and integral effects tests. On the other hand, the thermal hydraulic system codes like RELAP5/MOD3.3 are primarily intended for simulation of transients and accidents in light water reactors. The plant measured data come mostly from startup tests and operational events. Also, for operational events the measured plant data may not be sufficient to explain all details of the event. The purpose of this study was therefore besides code assessment to demonstrate that simulations can be very beneficial for deep understanding of the plant response and further corrective measures. The abnormal event with reactor trip and safety injection signal actuation was simulated with the latest RELAP5/MOD3.3 Patch 05 best-estimate thermal hydraulic computer code. The measured and simulated data agree well considering the major plant system responses and operator actions. This suggests that the RELAP5 code simulation is good representative of the plant response and can complement not available information from plant measured data. In such a way, an event can be better understood.

Description: The Separation and Safeguards Performance Model (SSPM) uses MATLAB/Simulink to provide a tool for safeguards analysis of bulk handling nuclear processing facilities. Models of aqueous and electrochemical reprocessing, enrichment, fuel fabrication, and molten salt reactor facilities have been developed to date. These models are used for designing the overall safeguards system, examining new safeguards approaches, virtually testing new measurement instrumentation, and analyzing diversion scenarios. The key metrics generated by the models include overall measurement uncertainty and detection probability for various material diversion or facility misuse scenarios. Safeguards modeling allows for rapid and cost-effective analysis for Safeguards by Design. The models are currently being used to explore alternative safeguards approaches, including more reliance on process monitoring data to reduce the need for destructive analysis that adds considerable burden to international safeguards. Machine learning techniques are being applied, but these techniques need large amounts of data for training and testing the algorithms. The SSPM can provide that training data. This paper will describe the SSPM and its use for applying both traditional nuclear material accountancy and newer machine learning options.

Description: A depressurization accident is the design-basis accidents of a gas turbine high temperature reactor, GTHTR300, which is JAEA’s design and one of the Very-High-Temperature Reactors (VHTR). When a primary pipe rupture accident occurs, air is expected to enter the reactor core from the breach and oxidize in-core graphite structures. Therefore, it is important to know a mixing process of different kinds of gases in the stable and unstable density stratified fluid layer. In order to predict or analyze the air ingress phenomena during the depressurization accident, we have conducted an experiment to obtain the mixing process of two component gases and the characteristics of natural circulation. The experimental apparatus consists of a storage tank and a reverse U-shaped vertical rectangular passage. One side wall of the high temperature side vertical passage is heated and the other side wall is cooled. The other experimental apparatus consists of a cylindrical double coaxial vessel and a horizontal double coaxial pipe. The outside of the double coaxial vessel is cooled and the inside is heated. The results obtained in this study are as follows. When the primary pipe is connected at the bottom of the reactor pressure vessel, onset time of natural circulation of air is affected by not only molecular diffusion but also localized natural convection. When the wall temperature difference is large, onset time of natural circulation of air is strongly affected by natural convection rather than molecular diffusion. When the primary pipe is connected at the side of the reactor pressure vessel, air will enter the bottom space in the reactor pressure vessel by counter-current flow at the coaxial double pipe break part immediately. Afterward, air will enter the reactor core by localized natural convection and molecular diffusion.

Description: The paper studies structure, phase composition, and thermophysical properties (TPP) (specific heat capacity, thermal diffusivity, and heat conductivity) of a prototype corium of a fast nuclear reactor (melt of core materials of nuclear reactor produced under out-of-pile conditions). The obtained data will be used to get more accurate understanding of main regularities of actual interaction of core materials of a nuclear reactor under a severe accident.

Description: Design of nuclear power plant shall provide an adequate margin to protect items ultimately necessary to prevent an early large radioactive release in the case of earthquakes exceeding those considered in the design. An essential question is how large the margin should be to be accepted as adequate. In the practice, depending on the country regulation, a plant margin of at least 1.4 or 1.67 times the design basis peak ground acceleration is required to be demonstrated. The catastrophe at the Fukushima Daiichi Nuclear Power Plant revealed the fundamental experience that the plants designed in compliance with nuclear standards can survive the effects of the vibratory ground motion due to disastrous earthquake but may fail due to effects of phenomena accompanying or generated by the earthquakes. Liquefaction is one of those secondary effects of beyond-design basis earthquakes that should be investigated for NPPs at soil sites. However, the question has not been investigated up to now, whether a “margin earthquake”, vibratory effects of which the plant can withstand thanks to design margin, will not induce liquefaction at soil sites and will not result in loss of safety functions. In the paper, a procedure is proposed for calculation of the probability and margin to liquefaction. Use of the procedure is demonstrated on a case study with realistic site-plant parameters. Criteria for probability for screening and acceptable probabilistic margin to liquefaction are proposed. The possible building settlement due to margin earthquake is also assessed.

Description: Three tests were carried out with the ROSA/LSTF (rig of safety assessment/large-scale test facility), which simulated accident management (AM) measures during station blackout transient with loss of primary coolant under assumptions of nitrogen gas inflow and total failure of high-pressure injection system in a pressurized water reactor. As the AM measures, steam generator (SG) secondary-side depressurization was done by fully opening the relief valves in both SGs, and auxiliary feedwater was injected into the secondary-side of both SGs simultaneously. Conditions for the break size and the onset timing of the AM measures were different among the three LSTF tests. In the three LSTF tests, the primary pressure decreased to a certain low pressure of below 1 MPa with or without the primary depressurization by fully opening the relief valve in a pressurizer as an optional AM measure, while no core uncovery took place through the whole transient. Nonuniform flow behaviors were observed in the SG U-tubes under natural circulation (NC) with nitrogen gas depending probably on the gas accumulation rate in the two LSTF tests that the gas accumulated remarkably. The RELAP5/MOD3.3 code predicted most of the overall trends of the major thermal hydraulic responses observed in the three LSTF tests. The code, however, indicated remaining problems in the predictions of the primary pressure, the SG U-tube collapsed liquid levels, and the NC mass flow rate after the nitrogen gas ingress as well as the accumulator flow rate through the analyses for the two LSTF tests, where the remarkable gas accumulation occurred.

Description: Supercritical CO2 Brayton cycle is a good choice of thermal-to-electric energy conversion system, which owns a high cycle efficiency and a compact cycle configuration. It can be used in many power-generation applications, such as nuclear power, concentrated solar thermal, fossil fuel boilers, and shipboard propulsion system. Transient analysis code for Supercritical CO2 Brayton cycle is a necessity in the areas of transient analyses, control strategy study, and accident analyses. In this paper, a transient analysis code SCTRAN/CO2 is developed for Supercritical CO2 Brayton Loop based on a homogenous model. Heat conduction model, point neutron power model (which is developed for nuclear power application), turbomachinery model for gas turbine, compressor and shaft model, and PCHE type recuperator model are all included in this transient analysis code. The initial verifications were performed for components and constitutive models like heat transfer model, friction model, and compressor model. The verification of integrated system transient was also conducted through making comparison with experiment data of SCO2EP of KAIST. The comparison results show that SCTRAN/CO2 owns the ability to simulate transient process for S-CO2 Brayton cycle. SCTRAN/CO2 will become an important tool for further study of Supercritical CO2 Bryton cycle-based nuclear reactor concepts.

Description: The safe introduction of Generation IV (Gen IV) reactor concepts into operation will require extensive testing of their components. This must be performed under neutronic conditions representative of those expected to prevail inside the new reactor cores when in operation. In a thermal Material Testing Reactor (MTR) such neutronic conditions can be achieved by tailoring the prevailing neutron spectrum with the utilization of a device containing appropriate materials. In this work various materials are investigated as candidate components of a device that will be required in case that a thermal MTR neutron energy spectrum must be locally transformed, so as to imitate Sodium cooled Fast Reactor (SFR). Many nuclides have been examined with respect to only their neutronic behavior, providing thus a pool of neutronically appropriate materials for consideration in further investigation, such as regarding reactor safety and fabrication issues. The nuclides have been studied using the neutronics code TRIPOLI-4.8 while the reflector of the Jules Horowitz Reactor (JHR) was considered as the hosting environment of the transforming device. The results obtained suggest that elements with important inelastic neutron scattering could be chosen at a first level as being able to modify the prevailing neutron spectrum towards the desired direction. The factors which are important for an effective inelastic scatterer comprise density and inelastic microscopic cross section, as well as the energy ranges where inelastic scattering occurs. All the above factors have been separately examined in order to suggest potential device materials, able to locally produce SFR neutron spectrum imitation in a thermal MTR.

Description: A large number of carbon materials are adopted in high-temperature gas-cooled reactor (HTGR). These carbon materials mainly include graphite IG-110 and boron-containing carbon material (BC), both of which are typical porous materials and normally absorb moisture. In order to inhibit the chemical corrosion reaction between core internals materials and moisture, the core needs to be strictly dehumidified before the reactor is put into operation. This paper mainly analyzed the moisture transfer mechanism in these carbon materials. Moisture transfer models were developed, and the dehumidification process of HTR-PM core was simulated. In addition, the influence of working temperature and system pressure on dehumidification was studied as well.

Description: Water tank is one important component of passive containment cooling system (PCS) of nuclear island building. The sloshing frequency of water is much less than structure frequency and large-amplitude sloshing occurs easily when subjected to seismic loadings. Therefore, the sloshing dynamics and fluid-structure interaction (FSI) effect of water tank should be considered when the dynamic response of nuclear island building is analyzed. A 1/16 scaled model was designed and the shaking table test was done, in which the hydrodynamic pressure time histories and attenuation data of wave height were recorded. Then the sloshing frequencies and 1st sloshing damping ratio were recognized. Moreover, modal analysis and time history analysis of numerical model were done by ADINA software. By comparing the sloshing frequencies and hydrodynamic pressures, it is proved that the test method is reasonable and the formulation of potential-based fluid elements (PBFE) can be used to simulate FSI effect of nuclear island building.

Description: Multiphase flow measurements have become increasingly important in a wide range of industrial fields. In the present study, a dual needle-contact capacitance probe was newly designed to measure local void fractions and bubble velocity in a vertical channel, which was verified by digital high-speed camera system. The theoretical analyses and experiments show that the needle-contact capacitance probe can reliably measure void fractions with the readings almost independent of temperature and salinity for the experimental conditions. In addition, the trigger-level method was chosen as the signal processing method for the void fraction measurement, with a minimum relative error of −4.59%. The bubble velocity was accurately measured within a relative error of 10%. Meanwhile, dynamic response of the dual needle-contact capacitance probe was analyzed in detail. The probe was then used to obtain raw signals for vertical pipe flow regimes, including plug flow, slug flow, churn flow, and bubbly flow. Further experiments indicate that the time series of the output signals vary as the different flow regimes and are consistent with each flow structure.

Description: A simple method using the time-dependent Monte Carlo (TDMC) neutron transport calculation is presented to determine an effective detector position for the prompt neutron decay constant () measurement through the pulsed-neutron-source (PNS) experiment. In the proposed method, the optimum detector position is searched by comparing amplitudes of detector signals at different positions when their estimates by the slope fitting are converged. The developed method is applied to the Pb-Bi-zoned ADS experimental benchmark at Kyoto University Critical Assembly. The convergence time estimated by the TDMC PNS simulation agrees well with the experimental results. The convergence time map and the corresponding signal amplitude map predicted by the developed method show that polyethylene moderator regions adjacent to fuel region are better positions than other candidates for the PNS measurement.

Description: When light water reactor (LWR) is subject to a cold shutdown, it needs to be cooled with pure water or seawater to prevent the core melting. To precisely evaluate the cooling characteristics in the fuel assembly, a measurement method capable of installing to the fuel assembly structure and determining the temperature distribution with high temporal resolution, high spatial resolution, and in multidimension is required. Furthermore, it is more practical if applicable to a pressure range up to the rated pressure 16 MPa of a pressurized water reactor (PWR). In this study, we applied the principle of the wire-mesh sensor technology used in the void fraction measurement to the temperature measurement and developed a simulated fuel assembly (bundle) test loop with installing the temperature profile sensors. To investigate the measurement performance in the bundle test section, it was confirmed that a predetermined temperature calibration line with respect to time-average impedance was calculated and became a function of temperature. To evaluate the followability of measurement in a transient temperature change process, we fabricated a 16 × 16 wire-mesh sensor device and measured the hot-water jet-mixing process into the cold-water pool in real time and calculated the temperature profile from the temperature calibration line obtained in advance from each measurement point. In addition, the sensors applied to three-dimensional temperature distribution measurement of a complex flow field in the bundle structure. The axial and cross-sectional profiles of temperature were quantified in the forced flow field with nonboiling when the 5×5 bundle was heated by energization.

Description: This article presents the validation of the Code for Thermal-hydraulic Evaluation of Nuclear Reactors with Plate Type Fuels (COTENP), a subchannel code which performs steady-state thermal-hydraulic analysis of nuclear reactors with plate type fuel assemblies operating with the coolant at low pressure levels. The code is suitable for design analysis of research, test, and multipurpose reactors. To solve the conservation equations for mass, momentum, and energy, we adopt the subchannel and control volume methods based on fuel assembly geometric data and thermal-hydraulic conditions. We consider the chain or cascade method in two steps to facilitate the analysis of whole core. In the first step, we divide the core into channels with dimensions equivalent to that of the fuel assembly and identify the assembly with largest enthalpy rise as the hot assembly. In the second step, we divide the hot fuel assembly into subchannels with size equivalent to one actual coolant channel and similarly identify the hot subchannel. The code utilizes the homogenous equilibrium model for two-phase flow treatment and the balanced drop pressure approach for the flow rate determination. The code results include detailed information such as core pressure drop, mass flow rate distribution, coolant, cladding and centerline fuel temperatures, coolant quality, local heat flux, and results regarding onset of nucleate boiling and departure of nucleate boiling. To validate the COTENP code, we considered experimental data from the Brazilian IEA-R1 research reactor and calculated data from the Chinese CARR multipurpose reactor. The mean relative discrepancies for the coolant distribution were below 5%, for the coolant velocity were 1.5%, and for the pressure drop were below 10.7%. The latter discrepancy can be partially justified due to lack of information to adequately model the IEA-R1 experiment and CARR reactor. The results show that the COTENP code is sufficiently accurate to perform steady-state thermal-hydraulic design analyses for reactors with plate type fuel assemblies.

Description: The falling liquid flow rate under flooding conditions is limited at a square top end of a vertical pipe in the pressurizer surge line with the diameter of about 300 mm that consists of a vertical pipe, a vertical elbow, and a slightly inclined pipe with elbows. In this study, therefore, we evaluated effects of diameters on countercurrent flow limitation (CCFL) at the square top end in vertical pipes by using existing air-water data in the diameter range of D = 19-250 mm. As a result, we found that there was a strong relationship between the constant and the slope m in the Wallis-type correlation where the Kutateladze parameters were used for the dimensionless gas and liquid velocities. The constant and the slope m increased when the water level is increased in the upper tank h. CCFL at the square top end of the vertical pipes could be expressed by the Kutateladze parameters with = 1.53±0.11 and m = 0.97 for D ≥ 30 mm. The values were smaller for D = 19-25 mm than those for D ≥ 30 mm.

Description: In the deep geological repository of nuclear waste, the corrosion of waste generates gas, which increases the storage pressure, changes the properties of the rock strata, and affects the stability of nuclear waste repository. Therefore, it is of great importance to understand the gas migration in the engineering barrier and the potential impact on its integrity for the safety assessment of nuclear waste repository. A hydro-mechanical-damage model for analyzing gas migration in sedimentary rocks is established in this paper. On the basis of which, a set of coupled formulas for the coupling of gas migration in rock mass is established. The model considers the characteristics of gas migration in sedimentary rock, especially the microcracks caused by the degradation of elastic modulus and damage, and the coupling between the rock deformation and failure of fractures. The numerical simulation of gas injection test is beneficial to understand the mechanism of gas migration process in sedimentary rock.

Description: At present, the Tokamak has become a mainstream form of the magnetic confinement fusion device. The toroidal field (TF) magnet in the Tokamak system is required to generate a high-steady field to confine and shape the high temperature plasma. To secure high current density and high thermal stability, the no-insulation (NI) winding technique is used in the fabrication of the TF magnet. During plasma operation, heat is generated in the TF magnet caused by the interaction with central solenoid (CS) coils, poloidal field (PF) coils, and the plasma current. The heat generated in NI coils is complex owing to the existence of current flow between adjacent turns. Thus, it is necessary to calculate the thermal problems. Taking into consideration the effect of turn-to-turn contact resistance, this paper presents the thermal behavior of a NI toroidal magnet under different operating conditions. The NI toroidal magnet is composed of 10 double-pancake (DP) coils wound with BSCCO tapes. The analysis procedure combines the finite element method (FEM) with an equivalent circuit model. This analysis has applicability and practical directive to the design of cryogenic cooling system for NI toroidal magnet.

Description: During a hypothesized severe accident, a containment building is designed to act as a final barrier to prevent release of fission products to the environment in nuclear power plants. However, in a bypass scenario of steam generator tube rupture (SGTR), radioactive nuclides can be released to environment even if the containment is not ruptured. Thus, thorough mitigation strategies are needed to prevent such unfiltered release of the radioactive nuclides during SGTR accidents. To mitigate the consequence of the SGTR accident, this study was conducted to devise a conceptual approach of installing In-Containment Relief Valve (ICRV) from steam generator (SG) to the free space in the containment building and it was simulated by MELCOR code for numerical analysis. Simulation results show that the radioactive nuclides were not released to the environment in the ICRV case. However, the containment pressure increased more than the base case, which is a disadvantage of the ICRV. To minimize the negative effects of the ICRV, the ICRV linked to Reactor Drain Tank (RDT) and cavity flooding was performed. Because the overpressurization of containment is due to heat of ex-vessel corium, only cavity flooding was effective for depressurization. The conceptual design of the ICRV is effective in mitigating the SGTR accident.

Description: The molten salt reactor (MSR) is one of the six advanced reactor concepts selected by Generation IV International Forum (GIF) because of its inherent safety and the promising capabilities of TRU transmutation and Th-U breeding. In this study, a three-dimensional thermal-hydraulic model (3DTH) is developed for evaluating the steady-state performance of the graphite-moderated channel type MSR. The coupled code is developed by exchanging the power distribution, temperature, and fuel density distribution between SCALE and 3DTH. Firstly, the thermal-hydraulic model of the coupled code is validated by RELAP5 code. Then, the mass flow distribution, temperature field, , and power density distribution for a conceptual design of the 2MWt experimental molten salt reactor are calculated and analyzed by the coupled code under both normal operating situation and the central fuel assembly partly blocked situation. The simulated results are conductive to facilitate the understanding of the steady behavior of the graphite-moderated channel type MSR.

Description: The containment spray system (CSS) has a significant role in limiting the risk of radioactive exposure to the environment. In this work, the optimal droplet size and pH value of spray water to prevent the fission product release have been evaluated to improve the performance of the spray system during in-vessel release phase. A semikinetic model has been developed and implemented in MATLAB. The sensitivity and removal rate of airborne isotopes with the spray system have been simulated versus the spray activation and failure time, droplet size, and pH value. The alkaline (Na2S2O3) spray solution and spray water with pH 9.5 have similar scrubbing properties for iodine. However, the removal rate from the CSS has been found to be an approximately inverse square of droplet diameter () for Na2S2O3 and higher pH of spray water. The numerical results showed that 450 μm–850 μm droplet with 9.5 pH and higher or the alkaline (Na2S2O3) solution with 0.2 m3/s–0.35 m3/s flow rate is optimal for effective scrubbing of in-containment fission products. The proposed model has been validated with TOSQAN experimental data.

Description: The objective of this study was to evaluate accident-tolerant fuel (ATF) concepts being considered for CANDU reactors. Several concepts, including uranium dioxide/silicon carbide (UO2-SiC) composite fuel, dense fuels, microencapsulated fuels, and ATF cladding, were modelled in Serpent 2 to obtain reactor physics parameters, including important feedback parameters such as coolant void reactivity and fuel temperature coefficient. In addition, fuel heat transfer was modelled, and a simple accident model was tested on several ATF cases to compare with UO2. Overall, several concepts would require enrichment of uranium to avoid significant burnup penalties, particularly uranium-molybdenum (U-Mo) and fully ceramic microencapsulated (FCM) fuels. In addition, none of the fuel types have a significant advantage over UO2 in terms of overall accident response or coping time, though U-9Mo fuel melts significantly sooner due to its low melting point. Instead, the different ATF concepts appear to have more modest advantages, such as reduced fission product release upon cladding failure, or reduced hydrogen generation, though a proper risk assessment would be required to determine the magnitude of these advantages to weigh against economic disadvantages. The use of uranium nitride (UN) enriched in would increase exit burnup for natural uranium, providing a possible economic advantage depending on fuel manufacturing costs.

Description: The Canadian Supercritical Water-cooled Reactor (SCWR), a GEN IV reactor design, is a hybrid design of the well-established CANDU™ and Boiling Water Reactor with water above its thermodynamic critical point. Given the batch fueled design, control rods are used to manage the reactivity throughout the fuel cycle. This paper examines the consequences of a control rod drop accident (CRDA) for the Canadian SCWR. The asymmetry generated by the dropped rod requires an accurate 3-dimensional neutron kinetics calculation coupled to a detailed thermal-hydraulic model. Before simulating the CRDAs, the proper implementation of the 3D reactivity feedback was verified and various sensitivity studies were performed. This work demonstrates that the proposed safety systems for the SCWR core are capable of terminating the CRDA sequence prior to exceeding maximum sheath and centerline temperatures. In one instance involving a rod on the periphery of the core, the proposed trip setpoint (115% FP) was not exceeded and a new steady state was reached. Therefore it is recommended that the design also include provisions for a high-log rate and/or local Neutron Overpower Protection (NOP) trips, similar to existing CANDU designs such that reactor shutdown can be assured for such spatial anomalies.

Description: Integrated Deterministic and Probabilistic Safety Analysis (IDPSA) of dynamic systems calls for the development of efficient methods for accidental scenarios generation. The necessary consideration of failure events timing and sequencing along the scenarios requires the number of scenarios to be generated to increase with respect to conventional PSA. Consequently, their postprocessing for retrieving safety relevant information regarding the system behavior is challenged because of the large amount of generated scenarios that makes the computational cost for scenario postprocessing enormous and the retrieved information difficult to interpret. In the context of IDPSA, the interpretation consists in the classification of the generated scenarios as safe, failed, Near Misses (NMs), and Prime Implicants (PIs). To address this issue, in this paper we propose the use of an ensemble of Semi-Supervised Self-Organizing Maps (SSSOMs) whose outcomes are combined by a locally weighted aggregation according to two strategies: a locally weighted aggregation and a decision tree based aggregation. In the former, we resort to the Local Fusion (LF) principle for accounting the classification reliability of the different SSSOM classifiers, whereas in the latter we build a classification scheme to select the appropriate classifier (or ensemble of classifiers), for the type of scenario to be classified. The two strategies are applied for the postprocessing of the accidental scenarios of a dynamic U-Tube Steam Generator (UTSG).

Description: Hydrogen accumulation in the containment compartments under severe accidents would result in high concentration, which could lead to hydrogen deflagration or detonation. Therefore, getting detailed hydrogen flow and distribution is a key issue to arrange hydrogen removal equipment in the containment compartments. In this study, hydrogen flow behavior in local compartments has been investigated in two horizontal compartments. The analysis model is built by 3-dimensional CFD code in Cartesian coordinates based on the connection structure of the Advanced Pressurized Water Reactor (PWR) compartments. It consists of two cylindrical vessels, representing the Steam Generator compartment (SG) and Core Makeup Tank compartment (CMT). With standard turbulence model, the effects of the connecting pipe size and location on hydrogen concentration distribution are investigated. Results show that increasing the diameter of connection pipe (IP) which is located at 800 mm from 150 mm to 300 mm facilitates hydrogen flow between compartments. Decreasing the length of IP which is located at 800 mm from 1000 mm to 500 mm can also facilitate hydrogen flow between compartments. Lower IP is in favor of hydrogen mixing with air in non-source compartment. Higher IP is helpful for hydrogen flow to the non-source term compartment from source term compartment.

Description: As compared to the two-fluid single-pressure model, the two-fluid seven-equation two-pressure model has been proved to be unconditionally well-posed in all situations, thus existing with a wide range of industrial applications. The classical 1st-order upwind scheme is widely used in existing nuclear system analysis codes such as RELAP5, CATHARE, and TRACE. However, the 1st-order upwind scheme possesses issues of serious numerical diffusion and high truncation error, thus giving rise to the challenge of accurately modeling many nuclear thermal-hydraulics problems such as long term transients. In this paper, a semi-implicit algorithm based on the finite volume method with staggered grids is developed to solve such advanced well-posed two-pressure model. To overcome the challenge from 1st-order upwind scheme, eight high-resolution total variation diminishing (TVD) schemes are implemented in such algorithm to improve spatial accuracy. Then the semi-implicit algorithm with high-resolution TVD schemes is validated on the water faucet test. The numerical results show that the high-resolution semi-implicit algorithm is robust in solving the two-pressure two-fluid two-phase flow model; Superbee scheme and Koren scheme give two highest levels of accuracy while Minmod scheme is the worst one among the eight TVD schemes.

Description: As HANARO has been recently required to support new R&D relevant to future nuclear systems requiring much higher neutron fluence, two types of bottom rod tip of the capsule were preliminarily prepared. The first one is a conventional design made of STS304 and welded using a tungsten inert gas (TIG) welding method. The other is a new design made of STS316L and welded using electron beam (EB) welding to strengthen the fatigue property of the rod tip. During the out-pile testing, they failed after 40 and 203 days, respectively. The fracture surfaces were examined using microscopes and the maximal applied stresses were estimated. The combination of these stresses was proved to be sufficient to cause a fatigue failure of the rod tip of the capsule. Based on the failure analysis, an optimized design of the rod tip of the capsule was made for long-term irradiation testing. It was designed to improve the welding and fatigue properties, to decrease the applied stress on the rod tip, and to fundamentally eliminate the effect of residual stress due to welding. The newly designed capsule was safely out-pile-tested up to 450 days and will be utilized for HANARO irradiation testing.

Description: The one-dimensional two-fluid model approach has been traditionally used in thermal-hydraulics codes for the analysis of transients and accidents in water–cooled nuclear power plants. This paper investigates the performance of RELAP5/MOD3 predicting vertical upward bubbly flow at low velocity conditions. For bubbly flow and vertical pipes, this code applies the drift-velocity approach, showing important discrepancies with the experiments compared. Then, we use a classical formulation of the drag coefficient approach to evaluate the performance of both approaches. This is based on the critical Weber criteria and includes several assumptions for the calculation of the interfacial area and bubble size that are evaluated in this work. A more accurate drag coefficient approach is proposed and implemented in RELAP5/MOD3. Instead of using the Weber criteria, the bubble size distribution is directly considered. This allows the calculation of the interfacial area directly from the definition of Sauter mean diameter of a distribution. The results show that only the proposed approach was able to predict all the flow characteristics, in particular the bubble size and interfacial area concentration. Finally, the computational results are analyzed and validated with cross-section area average measurements of void fraction, dispersed phase velocity, bubble size, and interfacial area concentration.

Description: The Power Burst Facility (PBF) was designed to provide experimental data to determine the thresholds for failure during accident conditions. Thus, the PBF benchmark using severe accidental analysis codes is essential to designing reactor for current directions. This assessment verified and validated that the RELAP/SCDAPSIM/MOD3.4 code can be used to assess the Severe Fuel Damage Scoping Test (SFD-ST) performed in the PBF facility. This study compares the cladding temperatures and hydrogen production results calculated by the RELAP/SCDAPSIM/MOD3.4 code with experimental data and calculated results from the SCDAP/RELAP5/MOD3.2 and SCDAP/RELAP5/MOD3.3 codes. The interested parameters are cladding temperature and hydrogen production since the cladding temperature affects hydrogen production and consequently influences the accident scenario. The calculated cladding temperatures and hydrogen production results from the RELAP/SCDAPSIM/MOD3.4 code are in a good agreement with the experimental data and are generally more reasonable than the calculated results from the SCDAP/RELAP5/MOD3.2 and SCDAP/RELAP5/MOD3.3 codes. There are some discrepancies in the cladding temperature and hydrogen production results but they are expected.

Description: Uncertainty and sensitivity analysis of void reactivity feedback for 3D BWR fuel assembly model is presented in this paper. Uncertainties in basic input data, such as the selection of different cross section library, manufacturing uncertainties in material compositions, and geometrical dimensions, as well as operating data are considered. An extensive modelling of different input data realizations associated with their uncertainties was performed during sensitivity analysis. The propagation of uncertainties was analyzed using the statistical approach. The results revealed that important information on the code predictions can be obtained by analyzing and comparing the codes estimations and their associated uncertainties.

Description: Well-type NaI(Tl) detectors are beneficial for low-level photon activity measurements because of the near 4π solid angle that can be gained with them. The detection efficiency can differ with the source-to-detector system geometries, the absorption of the photon in the detector material, and attenuation layers in front of the detector face. For these purposes, the absolute efficiency and the coincidence corrections of the well-type sodium iodide detector have been measured at 0.121–1.408 MeV energy range (obtained from 152Eu, 137Cs, and 60Co radioactive isotopes). The covenant between the experimental (present work) and the published theoretical values is good, with the high discrepancies being less than 1%.

Description: This article aims to simulate the sudden and fast pressure drop of VVER-1000 reactor core coolant, regarding acoustic phenomenon. It is used to acquire a more accurate method in order to simulate the various accidents of reactor core. Neutronic equations should be solved concurrently by means of DRAGON 4 and DONJON 4 coupling codes. The results of the developed package are compared with WIMS/CITATION and final safety analysis report of Bushehr VVER-1000 reactor (FSAR). Afterwards, time dependent thermal-hydraulic equations are answered by employing Single Heated Channel by Sectionalized Compressible Fluid method. Then, the obtained results were validated by the same transient simulation in a pressurized water reactor core. Then, thermal-hydraulic and neutronic modules are coupled concurrently by use of producing group constants regarding the thermal feedback effect. Results were compared to the mentioned transient simulation in RELAP5 computer code, which show that mass flux drop is sensed at the end of channel in several milliseconds which causes heat flux drop too. The thermal feedback resulted in production of some perturbations in the changes of these parameters. The achieved results for this very fast pressure drop represent accurate calculations of thermoneutronic parameters fast changes.

Description: The development of a turbulent mixing layer at mixing of two horizontal water streams with slightly different densities is studied by the means of numerical simulation. The mixing of such flows can be modelled as the flow of two components, where the concentration of one component in the mixing region is described as a passive scalar. The velocity field remains common over the entire computational domain, where the density and viscosity difference due to the concentration mainly affects the turbulent fluctuations in the mixing region. The numerical simulations are performed with the open source code OpenFOAM using two different approaches for turbulence modelling, Reynolds Averaged Navier Stokes equations (RANS) and Large Eddy Simulation (LES). The simulation results are discussed and compared with the benchmark experiment obtained within the frame of OECD/NEA benchmark test. A good agreement with experimental results is obtained in the case of the single liquid experiment. A high discrepancy between the simulated and the experimental velocity fluctuations in the case of mixing of the flows with the slightly different densities and viscosities triggered a systematic investigation of the modelling approaches that helped us to find out and interpret the main reasons for the disagreement.

Description: An optimized principal component analysis (PCA) framework is proposed to implement condition monitoring for sensors in a nuclear power plant (NPP) in this paper. Compared with the common PCA method in previous research, the PCA method in this paper is optimized at different modeling procedures, including data preprocessing stage, modeling parameter selection stage, and fault detection and isolation stage. Then, the model’s performance is greatly improved through these optimizations. Finally, sensor measurements from a real NPP are used to train the optimized PCA model in order to guarantee the credibility and reliability of the simulation results. Meanwhile, artificial faults are sequentially imposed to sensor measurements to estimate the fault detection and isolation ability of the proposed PCA model. Simulation results show that the optimized PCA model is capable of detecting and isolating the sensors regardless of whether they exhibit major or small failures. Meanwhile, the quantitative evaluation results also indicate that better performance can be obtained in the optimized PCA method compared with the common PCA method.

Description: Neutronics analysis has been performed for the 500 kW Dalat Nuclear Research Reactor loaded with highly enriched uranium fuel using the SRAC code system. The effective multiplication factors, keff, were analyzed for the core at criticality conditions and in two cases corresponding to the complete withdrawal and the full insertion of control rods. MCNP5 calculations were also conducted and compared to that obtained with the SRAC code. The results show that the difference of the keff values between the codes is within 55 pcm. Compared to the criticality conditions established in the experiments, the maximum differences of the keff values obtained from the SRAC and MCNP5 calculations are 119 pcm and 64 pcm, respectively. The radial and axial power peaking factors are 1.334 and 1.710, respectively, in the case of no control rod insertion. At the criticality condition these values become 1.445 and 1.832 when the control rods are partially inserted. Compared to MCNP5 calculations, the deviation of the relative power densities is less than 4% at the fuel bundles in the middle of the core, while the maximum deviation is about 7% appearing at some peripheral bundles. This agreement indicates the verification of the analysis models.

Description: This contribution is aimed at designing the optimal thickness of lead-iron double-layer container to store a radioactive waste releasing the photon energy at 1.3325 MeV and initial radiation intensity at 100 mSv/hr using the optimization design by MATLAB software. This design consisted of three parts of calculations to achieve 1000 times the radiation attenuation of container. The first was the logarithmic interpolation for the mass attenuation coefficient. The second was the bilogarithmic interpolation for the exposure buildup factor. The third was the contour-plotting analytical technique for the optimal thickness of radiation container. The values of mass attenuation coefficient and exposure buildup factor were exactly validated as compared with the standard reference database. Furthermore, we have found that the optimal thickness was 3.2 cm for lead (1st layer) and 17.0 cm for iron (2nd layer). Container weight was 994.30 kg, whilst container cost was 167.30 USD. The benefit of our design can quickly and precisely apply for the radiation safety assessment of the occupational radiation workers who always work in the nuclear reactor area.

Description: The paper presents a method of temperature and stress estimation in pressure components of conventional or nuclear power plants. The proposed algorithm can be applied without the information concerning the thermal boundary condition on the component inner surfaces and it is fast enough to be applied in an online mode. The solution is possible thanks to “measured” temperature histories determined in easily accessible points located on the component outer surface. The presented model has been recently verified analytically, numerically, and experimentally. The proposed algorithm was used to calculate the transient temperature and stress distribution in the outlet header of a steam reheater and the results indicate that the component lower part is loaded the most, but allowable stresses are not exceeded. If the presented heating process was shortened, the calculated stresses would exceed the allowable values. Monitoring the boiler thermal and strength operating conditions makes it possible to identify dangerous loads of the power boiler pressure elements during transient-state operations. The presented method for controlling thermal and pressure-related stresses is suitable for nuclear power plants because it does not require drilling holes for sensors in the pressure element walls.

Description: Nuclear power plants shall be designed to resist the effects of large earthquakes. The design basis earthquake affects large area around the plant site and can cause serious consequences that will affect the logistical support of the emergency actions at the plant, influence the psychological condition of the plant personnel, and determine the workload of the country’s disaster management personnel. In this paper the main qualitative findings of a study are presented that have been performed for the case of a hypothetical 10−4/a probability design basis earthquake for the Paks Nuclear Power Plant, Hungary. The study covers the qualitative assessment of the postearthquake conditions at the settlements around the plant site including quantitative evaluation of the condition of dwellings. The main goal of the recent phase of the study was to identify public utility vulnerabilities that define the outside support conditions of the nuclear power plant accident management. The results of the study can be used for the planning of logistical support of the plant accident management staff. The study also contributes to better understanding of the working conditions of the disaster management services in the region around the nuclear power plant.

Description: Long-term high payload missions necessitate the need for nuclear space propulsion. The National Aeronautics and Space Administration (NASA) investigated several reactor designs from 1959 to 1973 in order to develop the Nuclear Engine for Rocket Vehicle Application (NERVA). Study of planned/unplanned transients on nuclear thermal rockets is important due to the need for long-term missions. In this work, a system model based on RELAP5 is developed to simulate loss-of-flow accidents on the Pewee I test reactor. This paper investigates the radiation heat transfer between the fuel elements and the structures around it. In addition, the impact on the core fuel element temperature and average core pressure was also investigated. The following expected results were achieved: (i) greater than normal fuel element temperatures, (ii) fuel element temperatures exceeding the uranium carbide melting point, and (iii) average core pressure less than normal. Results show that the radiation heat transfer rate between fuel elements and cold surfaces increases with decreasing flow rate through the reactor system. However, radiation heat transfer decreases when there is a complete LOFA. When there is a complete LOFA, the peripheral coolant channels of the fuel elements handle most of the radiation heat transfer. A safety system needs to be designed to counteract the decay heat resulting from a post-LOFA reactor scram.

Description: The objective of this paper is to give an overview of the capabilities of Eulerian bifluid approach to meet the needs of studies for nuclear safety regarding hydrogen risk, boiling crisis, and pipes and valves maintenance. The Eulerian bifluid approach has been implemented in a CFD code named NEPTUNE_CFD. NEPTUNE_CFD is a three-dimensional multifluid code developed especially for nuclear reactor applications by EDF, CEA, AREVA, and IRSN. The first set of models is dedicated to wall vapor condensation and spray modelling. Moreover, boiling crisis remains a major limiting phenomenon for the analysis of operation and safety of both nuclear reactors and conventional thermal power systems. The paper aims at presenting the generalization of the previous DNB model and its validation against 1500 validation cases. The modelling and the numerical simulation of cavitation phenomena are of relevant interest in many industrial applications, especially regarding pipes and valves maintenance where cavitating flows are responsible for harmful acoustics effects. In the last section, models are validated against experimental data of pressure profiles and void fraction visualisations obtained downstream of an orifice with the EPOCA facility (EDF R&D). Finally, a multifield approach is presented as an efficient tool to run all models together.

Description: The geometry of uranium components is one of the key characteristics and strictly confidential. The geometry identification of metal uranium components was studied using 252Cf source-driven correlation measurement method. For the 3 uranium samples with the same mass and enrichment, there are subtle differences in neutron signals. Even worse, the correlation functions were disturbed by scatter neutrons and include “accidental” coincidence, which is not conductive to the geometry identification. In this paper, we proposed an identification method combining principal component analysis and least-square support vector machine (PCA-LSSVM). The results based on PCA-LSSVM showed that the training precision was 100% and the test precision was 95.83% of the identification model. The total precision of the identification model was 98.41%, which indicated that the identification model was an effective way to identify the geometry properties with the correlation functions.

Description: The in-house coupled neutronic and thermal-hydraulic (N/T-H) code of BATAN (National Nuclear Energy Agency of Indonesia), NODAL3, based on the few-group neutron diffusion equation in 3-dimensional geometry using the polynomial nodal method, has been verified with static and transient PWR benchmark cases. This paper reports the verification of NODAL3 code in the NEA-NSC PWR uncontrolled control rods withdrawal at zero power benchmark. The objective of this paper is to determine the accuracy of NODAL3 code in solving the continuously slow and fast reactivity insertions due to single and group of control rod bank withdrawn while the power and temperature increment are limited by the Doppler coefficient. The benchmark is chosen since many organizations participated using various methods and approximations, so the calculation results of NODAL3 can be compared to other codes’ results. The calculated parameters are performed for the steady-state, transient core averaged, and transient hot pellet results. The influence of radial and axial nodes number was investigated for all cases. The results of NODAL3 code are in very good agreement with the reference solutions if the radial and axial nodes number is 2 × 2 and 2 × 18 (total axial layers), respectively.

Description: Flash evaporation of a superheated water droplet in heavy liquid metal coolant (lead) is considered, in application to the analysis of a lead-cooled fast reactor steam generator tube rupture accident. The model is based on thermodynamic equilibrium formulation for the expanding water-steam mixture and inviscid compressible formulation for the surrounding liquid lead, with the interface conditions determined from the solution of the Riemann problem. Numerical solution is performed in the spherically symmetric geometry using a conservative numerical scheme with a moving sharp interface. Transient pressure and velocity profiles in each phase are presented for the parameters typical of the steam generator tube rupture accidents, demonstrating the process of boiling water expansion and pressure wave formation in the coolant. The results obtained are compared with a simplified model which considers the volume-averaged parameters of boiling water droplets and considers coolant as an incompressible liquid. Good agreement between the full and simplified models is demonstrated. Impacts of coolant flow on structures caused by pressure wave propagation and subsequent coolant flow are discussed.

Description: At the beginning, a comparative analysis was made on the oxidation corrosion rate and ash content of A3-3 matrix graphite (MG) pebbles lathed before and after high temperature purification (HTP) treatment. Their oxidation corrosion rate and ash contents were almost identical, which indicated that the HTP process was to purify the entire MG pebbles and not limited on the surfaces. Furthermore, the multiple mechanical and thermal properties of MG treated without and with the treatment of HTP at ~1900°C were compared and their microstructure features were characterized as well. As the crush strength, oxidation corrosion rate, and erosion rate of MG without HTP treatment did not satisfy the specifications, the comprehensive properties and purity of MG with HTP were improved in various degrees through the HTP process so that all performances met the requirements of the A3-3 MG. The improvement of crush strength and erosion rate of MG in the HTP process could be mainly attributed to the upgradation of ordered microstructure and corresponding increase of density. However, the enhancement of oxidation corrosion rate was due to the synergistic effects of microstructural optimization and reduction of impurity elements, especially the transition metal elements of MG in the HTP process.

Description: High thermal neutron fluxes are needed in some research reactors and for irradiation tests of materials. A High Flux Research Reactor (HFRR) with an inverse flux trap-converter target structure is being developed by the Reactor Engineering Analysis Lab (REAL) at Tsinghua University. This paper studies the safety of the HFRR core by full core flow and temperature calculations using the porous media approach. The thermal nonequilibrium model is used in the porous media energy equation to calculate coolant and fuel assembly temperatures separately. The calculation results show that the coolant temperature keeps increasing along the flow direction, while the fuel temperature increases first and decreases afterwards. As long as the inlet coolant mass flow rate is greater than 450 kg/s, the peak cladding temperatures in the fuel assemblies are lower than the local saturation temperatures and no boiling exists. The flow distribution in the core is homogeneous with a small flow rate variation less than 5% for different assemblies. A large recirculation zone is observed in the outlet region. Moreover, the porous media model is compared with the exact model and found to be much more efficient than a detailed simulation of all the core components.

Description: The integral Phébus tests were probably one of the most important experimental campaigns performed to investigate the progression of severe accidents in light water reactors. In these tests, the degradation of a PWR fuel bundle was investigated employing different control rod materials and burn-up levels in strongly or weakly oxidizing conditions. From the results of such tests, numerical codes such as ASTEC and MELCOR have been developed to describe the evolution of a severe accident. After the termination of the experimental Phébus campaign, these two codes were furthermore expanded. Therefore, the aim of the present work is to reanalyze the first Phébus test (FPT-0) employing the updated ASTEC and MELCOR versions to ensure that the new improvements introduced in such codes allow also a better prediction of these Phébus tests. The analysis focuses on the stand-alone containment aspects of this test, and the paper summarizes the main thermal-hydraulic results and presents different sensitivity analyses carried out on the aerosols and fission products behavior. This paper is part of a series of publications covering the four executed Phébus tests employing a solid PWR fuel bundle: FPT-0, FPT-1, FPT-2, and FPT-3.

Description: A voloxidizer with a double reactor capable of processing several tens of kilograms of HM/batch of nuclear spent fuel has been developed for the decladding and voloxidation of rod-cuts into hulls and pellets through the conversion of UO2 pellets to U3O8 powder. In this study, we optimized the engineering design of this voloxidizer to improve its hull-recovery ratio. First, we tested the oxidation performance of the device prototype and evaluated the effectiveness of various mechanical and chemical voloxidizing methods. On the basis of the results, we selected the screw-and-rotation method for the double rotary drum. Next, we derived a theoretical equation for calculating the optimal reactor volume for various rod-cut weights and lengths and then validated the equation using centimeter-scale acryl reactors and hulls. Subsequently, we modularized the main components such as the heater, utility, motor, reactor, valve, and structure. The double reactor was subject to preliminary separation tests of hulls and powder. Moreover, the hull-separation performance of the voloxidizer reactor was tested at a loading of 50 kg HM/batch. Finally, the remote assembling and disassembling possibility of the modules were experimentally optimized.

Description: With a view to providing supportive information for the decision-making on the direction of the future nuclear energy systems in Korea (i.e., direct disposal or recycling of spent nuclear fuel) to be made around 2020, quantitative studies on the spent nuclear fuel (SNF) including transuranic elements (TRUs) and a series of economic analyses were conducted. At first, the total isotopic inventory of TRUs in the SNF to be generated from all thirty-six units of nuclear power plants in operation or under planning is estimated based on the Korean government’s official plan for nuclear power development. Secondly, the optimized deployment strategies are proposed considering the minimum number of sodium cooled-fast reactors (SFRs) needed to transmute all TRUs. Finally, direct disposal and Pyro-SFR closed nuclear energy systems were compared using equilibrium economic model and considering reduction of TRUs and electricity generation as benefits. Probabilistic economic analysis shows that the assumed total generation cost for direct disposal and Pyro-SFR closed nuclear energy systems resides within the range of 13.60~33.94 mills/kWh and 11.40~25.91 mills/kWh, respectively. Dominant cost elements and the range of SFR overnight cost which guarantees the economic feasibility of the Pyro-SFR closed nuclear energy system over the direct disposal option were also identified through sensitivity analysis and break-even cost estimation.

Description: Intentional depressurization and cavity flooding are two important measures in current severe accident management guidelines (SAMGs). An extreme scenario of an extended station blackout (ESBO), when electric power is unavailable for more than 24 hours, actually occurred in the Fukushima Daiichi accident and attracted lots of attention. In an ESBO, the containment spray cannot be activated for condensation, and, thus, cavity flooding will generate a large amount of steam, which, ironically, overpressurizes the containment to failure before the reactor vessel is melted through. Therefore, consideration of these conflicting issues and the ways in which plants operate is crucial for strengthening the strategies outlined in SAMGs. In this paper, the effects of intentional depressurization and cavity flooding in an ESBO for a representative 900 MW second-generation pressurized water reactor (PWR) are simulated with MAAP4 code. Diverse scenarios with different starting times of depressurization and water injection are also compared to summarize the positive and negative impacts for accident mitigation. The phenomena associated with creep ruptures, hydrogen combustion, corium stratification, and cavity boiling are also analyzed in detail to strengthen our understanding of severe accident mechanisms. The results point out the facility limitations of second-generation PWRs which can improve existing SAMGs.

Description: A nuclear power station using gas as a cooling medium has attracted so much attention because it offers high efficiency and greater safety. For a nuclear station that operates at a very high temperature, a gas-cooled reactor is fueled by uranium, moderated by graphite, and customarily cooled by helium. Nevertheless, throughout the operation, the bypass flow might be a result of a change in graphite shape that is caused by neutron damage. Core bypass and cross flows are significant elements to consider since the cross gap set hurdles to the flow field that are capable of diverting sufficient amount of coolant from reactor core location and initiating a possible fuel overheating. However, there is a great need to sufficiently validate this method by carrying out a thorough evaluation based on working environment analysis. Comparing the computed results with the existing data from Groehn’s NHDA PMR-200 study was the only way to validate data. A model simulation was performed on a two-prismatic fuel block with a cross gap to examine the gaping size effect. Finally, the prediction methods for horizontal flow phenomena using a CFD technique and the field investigation results from the VHTR core were verified, and the identification of the horizontal flow behavior played a vital role in investigating the coolant velocity and pressure distribution in the horizontal gap.

Description: The subcooling effect on pool boiling heat transfer using a copper microporous coating was experimentally studied in water for subcoolings of 10 K, 20 K, and 30 K at atmospheric pressure and compared to that of a plain copper surface. A high-temperature thermally conductive microporous coating (HTCMC) was made by sintering copper powder with an average particle size of 67 μm onto a 1 cm × 1 cm plain copper surface with a coating thickness of ~300 μm. The HTCMC surface showed a two times higher critical heat flux (CHF), ~2,000 kW/m2, and up to seven times higher nucleate boiling heat transfer (NBHT) coefficient, ~350 kW/m2K, when compared with a plain copper surface at saturation. The results of the subcooling effect on pool boiling showed that the NBHT of both the HTCMC and the plain copper surface did not change much with subcooling. On the other hand, the CHF increased linearly with the degree of subcooling for both the HTCMC and the plain copper surface. The increase in the CHF was measured to be ~60 kW/m2for every degree of subcooling for both the HTCMC and the plain surface, so that the difference of the CHF between the HTCMC and the plain copper surface was maintained at ~1,000 kW/m2throughout the tested subcooling range. The CHFs for the HTCMC and the plain copper surface at 30 K subcooling were 3,820 kW/m2and 2,820 kW/m2, respectively. The experimental results were compared with existing CHF correlations and appeared to match well with Zuber’s formula for the plain surface. The combined effect of subcooling and orientation of the HTCMC on pool boiling heat transfer was studied as well.

Description: This paper presents capacity of the passive decay heat removal system (DHRS) operated under the natural circulation conditions to remove decay heat inside the main vessel of the Lead-bismuth eutectic cooled Fast Reactor (LFR). The motivation of this research is to improve the inherent safety of the LFR based on the China Accelerator Driven System (ADS) engineering project. Usually the plant is damaged due to the failure of the main pumps and the main heat exchangers under the Station Blackout (SBO). To prevent this accident, we proposed the DHRS based on the diathermic oil cooling for the LFR. The behavior of the DHRS and the plant was simulated using the CFD code STAR CCM+ using LFR with DHRS. The purpose of this analysis is to evaluate the heat exchange capacity of the DHRS and is to provide the reference for structural improvement and experimental design. The results show that the stable natural circulations are established in both the main vessel and the DHRS. During the decay process, the heat exchange power is above the core decay heat power. In addition, in-core decay heat and heat storage inside the main vessel are efficiently removed. All the thermal-hydraulics parameters are within a safe range. Moreover, the highest temperature occurs at the upper surface of the core. A swirl occurs at the corner of the lateral core surface and some improvements should be considered. And the natural circulation driving force can be further increased by reducing the loop resistance or increasing the natural circulation height based on the present design scenario to enhance the heat exchange effect.

Description: Source term analysis is important in the design and safety analysis of advanced nuclear reactor and also provides a radiation safety analysis basis for Modular High-Temperature Gas-Cooled Reactor (HTR). High-Temperature Gas-Cooled Reactor-Pebble-bed Modules (HTR-PM) design by China is a typical Gen-IV and due to different safety concepts and systems, the implements of source term analysis in light water reactors are not entirely applicable to HTR-PM. To solve this problem, HTR-PM Source Term Analysis Code (HTR-STAC) has been developed and related V&V has been finished. HTR-STAC consists of five units, including LOOP (Primary Circuit Source Term Analysis Code), NORMAL (Normal Condition Airborne Source Term Analysis Code), ARCC (Accident Release Category Calculation code), CARBON (C-14 Source Term Analysis Code), and TRUM (Tritium Source Term Analysis Code). LOOP and NORMAL may be used as calculating primary circuit coolant radioactivity and the release of airborne radioactivity to the environment under normal operating conditions of HTR-PM, respectively. The code ARCC composed of several source term analysis programs in the different typical accidents scenario, including SGTR (Steam Generator Tube Rupture), LOCA (Loss of Coolant Accident), and the Transient Process, is compiled based on the results given by LOOP and NORMAL. CARBON and TRUM are developed to calculate the productions of C-14 and H-3 through a different mechanism. Furthermore, the V&V has been performed and show some positive results.

Description: This study was carried out to examine the potential of antimony tri-iodide (SbI3) as a material for radiation detectors that operate at room temperature. SbI3 is a compound semiconductor with an AsI3-type crystal structure, high atomic number (Sb: 51, I: 53), high density (4.92 g/cm3), and a wide band-gap energy (2.2 eV). In addition, crystalline SbI3 is easy to grow by conventional crystal growth techniques from melting phase because the material exhibits a low melting point (171°C) and undergoes no phase transition in the range of its solid phase. In this study, SbI3 crystals were grown by the Bridgman method after synthesis of SbI3 from 99.9999% pure Sb and 99.999% pure I2. The grown crystals consisted of several large grains with red color and were confirmed to be single-phase crystals by X-ray diffraction analysis. SbI3 detectors with a simple planar structure were fabricated using the cleavage plates of the grown crystals, and the pulse-height spectra were recorded at room temperature using an 241Am alpha-particle (5.48 MeV) source. The detector showed response to the alpha-particle radiation.

Description: The essential power supply system is important for the nuclear safety and accident mitigation of the currently operating nuclear power plants. This system provides electrical power to the essential instrumentation and control systems of the nuclear power plant when all alternate current power sources are lost. This event is known as station blackout (SBO) event. Operational events of failure or deficiency of the essential power supply system are analyzed in this paper. The relevant events were searched and identified in four databases of operational events. The report includes events identified in French SAPIDE and German VERA operational events records for the time period 1996 to 2015. The International Atomic Energy Agency (IAEA) IRS and Nuclear Regulatory Commission (NRC) LER operational events databases were screened for relevant events that occurred in the period between 2000 and 2016. In total, 308 relevant events are identified in the SAPIDE, 103 in VERA, 56 in LER, and 15 in IRS operational events database. Classification and in-depth analysis were done on the identified events considering the following predefined categories: the plant status during the event, circumstances, affected equipment, cause of the event (direct and root), and implications of the event on plant safety. Main findings from the evaluation of the events are presented. Observations of the causes resulting in the events and potential actions that can decrease the number and consequences of the events are presented.

Description: Vigorously developing nuclear power is the main development direction of current renewable energy. In the nuclear environment, in order to avoid nuclear radiation damage to maintenance personnel and improve the efficiency of nuclear reaction, it is necessary and urgent to realize automatic replacement of vulnerable parts in the electron gun. As the key equipment for the generation and control of nuclear reactions in nuclear reactors, electron guns have been widely used in nuclear power plants of traveling wave reactors. However, the “high-voltage conductive ring” in electron guns is a vulnerable part. It is likely to cause nuclear reactor accidents when the vulnerable part is damaged. Automatic replacement of vulnerable parts is an important part of the entire maintenance equipment. Considering the entire maintenance equipment and the working environment, an innovative design process for vulnerable parts replacement is established. Under the guidance of the process, in order to ensure the continuity of a series of maintenance actions, the technical contradiction resolution theory is first used to conduct the overall analysis of the general direction to obtain the design layout. Then, the contradiction resolution theory and the object-field model analysis are utilized to get and improve the detailed design of the device mechanism. The theory of TRIZ can help us to get the overall mechanical structure design that meets the engineering requirements. The device is designed with a replacement part adjustment scheme to ensure the completion of the maintenance actions. Furthermore, the design provides a solution to the possible jamming phenomenon in the automatic maintenance process and achieves the maximum use efficiency of the storage and replacement of vulnerable parts.

Description: This article presents the validation of the Code for Thermal-hydraulic Evaluation of Nuclear Reactors with Plate Type Fuels (COTENP), a subchannel code which performs steady-state thermal-hydraulic analysis of nuclear reactors with plate type fuel assemblies operating with the coolant at low pressure levels. The code is suitable for design analysis of research, test, and multipurpose reactors. To solve the conservation equations for mass, momentum, and energy, we adopt the subchannel and control volume methods based on fuel assembly geometric data and thermal-hydraulic conditions. We consider the chain or cascade method in two steps to facilitate the analysis of whole core. In the first step, we divide the core into channels with dimensions equivalent to that of the fuel assembly and identify the assembly with largest enthalpy rise as the hot assembly. In the second step, we divide the hot fuel assembly into subchannels with size equivalent to one actual coolant channel and similarly identify the hot subchannel. The code utilizes the homogenous equilibrium model for two-phase flow treatment and the balanced drop pressure approach for the flow rate determination. The code results include detailed information such as core pressure drop, mass flow rate distribution, coolant, cladding and centerline fuel temperatures, coolant quality, local heat flux, and results regarding onset of nucleate boiling and departure of nucleate boiling. To validate the COTENP code, we considered experimental data from the Brazilian IEA-R1 research reactor and calculated data from the Chinese CARR multipurpose reactor. The mean relative discrepancies for the coolant distribution were below 5%, for the coolant velocity were 1.5%, and for the pressure drop were below 10.7%. The latter discrepancy can be partially justified due to lack of information to adequately model the IEA-R1 experiment and CARR reactor. The results show that the COTENP code is sufficiently accurate to perform steady-state thermal-hydraulic design analyses for reactors with plate type fuel assemblies.

Description: The molten salt reactor (MSR) is one of the six advanced reactor concepts selected by Generation IV International Forum (GIF) because of its inherent safety and the promising capabilities of TRU transmutation and Th-U breeding. In this study, a three-dimensional thermal-hydraulic model (3DTH) is developed for evaluating the steady-state performance of the graphite-moderated channel type MSR. The coupled code is developed by exchanging the power distribution, temperature, and fuel density distribution between SCALE and 3DTH. Firstly, the thermal-hydraulic model of the coupled code is validated by RELAP5 code. Then, the mass flow distribution, temperature field, keff, and power density distribution for a conceptual design of the 2MWt experimental molten salt reactor are calculated and analyzed by the coupled code under both normal operating situation and the central fuel assembly partly blocked situation. The simulated results are conductive to facilitate the understanding of the steady behavior of the graphite-moderated channel type MSR.

Description: When light water reactor (LWR) is subject to a cold shutdown, it needs to be cooled with pure water or seawater to prevent the core melting. To precisely evaluate the cooling characteristics in the fuel assembly, a measurement method capable of installing to the fuel assembly structure and determining the temperature distribution with high temporal resolution, high spatial resolution, and in multidimension is required. Furthermore, it is more practical if applicable to a pressure range up to the rated pressure 16 MPa of a pressurized water reactor (PWR). In this study, we applied the principle of the wire-mesh sensor technology used in the void fraction measurement to the temperature measurement and developed a simulated fuel assembly (bundle) test loop with installing the temperature profile sensors. To investigate the measurement performance in the bundle test section, it was confirmed that a predetermined temperature calibration line with respect to time-average impedance was calculated and became a function of temperature. To evaluate the followability of measurement in a transient temperature change process, we fabricated a 16 × 16 wire-mesh sensor device and measured the hot-water jet-mixing process into the cold-water pool in real time and calculated the temperature profile from the temperature calibration line obtained in advance from each measurement point. In addition, the sensors applied to three-dimensional temperature distribution measurement of a complex flow field in the bundle structure. The axial and cross-sectional profiles of temperature were quantified in the forced flow field with nonboiling when the 5×5 bundle was heated by energization.

Description: The discrete ordinates method (SN) is one of the mainstream methods for neutral particle transport calculations. Assessing the quality of the numerical solution and controlling the discrete error are essential parts of large-scale high-fidelity simulations of nuclear systems. Three error estimators, a two-mesh estimator, a residual-based estimator, and a dual-weighted residual estimator, are derived and implemented in the ARES transport code to evaluate the error of zeroth-order spatial discretization for SN equations. The difference in scalar fluxes on coarse and fine meshes is adopted to indicate the error in the two-mesh method. To avoid zero residual in zeroth-order discretization, angular fluxes within one cell are reconstructed by Legendre polynomials. The error is estimated by inverting the discrete transport operator using the estimated directional residual as an anisotropic source. The inner product of the forward directional residual and the adjoint angular flux is employed to quantify the error in quantities of interest which can be denoted by a linear functional of forward angular flux. Method of Manufactured Solutions (MMS) is adopted to generate analytical solutions for SN equation with scattering and the determined true error is used to evaluate the effectivity of these estimators. Promising results are obtained in the numerical results for both homogeneous and heterogeneous cases. The larger error region is well captured and the average effectivity index for the local error estimation is less than unity. For the series test problems, the estimated goal quantity error can be contained within an order of magnitude around the exact error.

Description: The code assessment typically comprises basic tests cases, separate effects test, and integral effects tests. On the other hand, the thermal hydraulic system codes like RELAP5/MOD3.3 are primarily intended for simulation of transients and accidents in light water reactors. The plant measured data come mostly from startup tests and operational events. Also, for operational events the measured plant data may not be sufficient to explain all details of the event. The purpose of this study was therefore besides code assessment to demonstrate that simulations can be very beneficial for deep understanding of the plant response and further corrective measures. The abnormal event with reactor trip and safety injection signal actuation was simulated with the latest RELAP5/MOD3.3 Patch 05 best-estimate thermal hydraulic computer code. The measured and simulated data agree well considering the major plant system responses and operator actions. This suggests that the RELAP5 code simulation is good representative of the plant response and can complement not available information from plant measured data. In such a way, an event can be better understood.

Description: Multiphase flow measurements have become increasingly important in a wide range of industrial fields. In the present study, a dual needle-contact capacitance probe was newly designed to measure local void fractions and bubble velocity in a vertical channel, which was verified by digital high-speed camera system. The theoretical analyses and experiments show that the needle-contact capacitance probe can reliably measure void fractions with the readings almost independent of temperature and salinity for the experimental conditions. In addition, the trigger-level method was chosen as the signal processing method for the void fraction measurement, with a minimum relative error of −4.59%. The bubble velocity was accurately measured within a relative error of 10%. Meanwhile, dynamic response of the dual needle-contact capacitance probe was analyzed in detail. The probe was then used to obtain raw signals for vertical pipe flow regimes, including plug flow, slug flow, churn flow, and bubbly flow. Further experiments indicate that the time series of the output signals vary as the different flow regimes and are consistent with each flow structure.

Description: As HANARO has been recently required to support new R&D relevant to future nuclear systems requiring much higher neutron fluence, two types of bottom rod tip of the capsule were preliminarily prepared. The first one is a conventional design made of STS304 and welded using a tungsten inert gas (TIG) welding method. The other is a new design made of STS316L and welded using electron beam (EB) welding to strengthen the fatigue property of the rod tip. During the out-pile testing, they failed after 40 and 203 days, respectively. The fracture surfaces were examined using microscopes and the maximal applied stresses were estimated. The combination of these stresses was proved to be sufficient to cause a fatigue failure of the rod tip of the capsule. Based on the failure analysis, an optimized design of the rod tip of the capsule was made for long-term irradiation testing. It was designed to improve the welding and fatigue properties, to decrease the applied stress on the rod tip, and to fundamentally eliminate the effect of residual stress due to welding. The newly designed capsule was safely out-pile-tested up to 450 days and will be utilized for HANARO irradiation testing.

Description: The geometry of uranium components is one of the key characteristics and strictly confidential. The geometry identification of metal uranium components was studied using 252Cf source-driven correlation measurement method. For the 3 uranium samples with the same mass and enrichment, there are subtle differences in neutron signals. Even worse, the correlation functions were disturbed by scatter neutrons and include “accidental” coincidence, which is not conductive to the geometry identification. In this paper, we proposed an identification method combining principal component analysis and least-square support vector machine (PCA-LSSVM). The results based on PCA-LSSVM showed that the training precision was 100% and the test precision was 95.83% of the identification model. The total precision of the identification model was 98.41%, which indicated that the identification model was an effective way to identify the geometry properties with the correlation functions.

Description: At the beginning, a comparative analysis was made on the oxidation corrosion rate and ash content of A3-3 matrix graphite (MG) pebbles lathed before and after high temperature purification (HTP) treatment. Their oxidation corrosion rate and ash contents were almost identical, which indicated that the HTP process was to purify the entire MG pebbles and not limited on the surfaces. Furthermore, the multiple mechanical and thermal properties of MG treated without and with the treatment of HTP at ~1900°C were compared and their microstructure features were characterized as well. As the crush strength, oxidation corrosion rate, and erosion rate of MG without HTP treatment did not satisfy the specifications, the comprehensive properties and purity of MG with HTP were improved in various degrees through the HTP process so that all performances met the requirements of the A3-3 MG. The improvement of crush strength and erosion rate of MG in the HTP process could be mainly attributed to the upgradation of ordered microstructure and corresponding increase of density. However, the enhancement of oxidation corrosion rate was due to the synergistic effects of microstructural optimization and reduction of impurity elements, especially the transition metal elements of MG in the HTP process.