ALBERT

Description: The high-temperature gas-cooled reactor pebble-bed module (HTR-PM) nuclear power plant consists of two nuclear steam supply system modules, each of which drives the steam turbine by the superheated steam flow and is fed by the heated-up water flow. The shared steam/water system induces mutual effects on normal operation conditions and transients of the nuclear power plant, which is worthy of safety concerns and intensive study. In this paper, a coupling code package was developed with the TINTE and vPower codes to understand how the HTR-PM operated. The TINTE code was used to analyze the reactor core and primary circuit, while the vPower code simulated the steam/water flow in the conventional island. Two TINTE models were built and coupled to one vPower model through the data exchange in the steam generator models. Using this code package, two typical transients were simulated by decreasing the primary flow rate or introducing the negative reactivity of one module. Important parameters, including the reactor power, the fuel temperature, and the reactor inlet and outlet helium temperatures of two modules, had been studied. The calculation results preliminarily proved that this code package can be further used to evaluate working performance of the HTR-PM.

Description: The paper presents a conceptual design of a 10 MW multipurpose nuclear research reactor (MPRR) loaded with the low-enriched uranium (LEU) VVR-KN fuel type. Neutronics and burnup calculations have been performed using the REBUS-MCNP6 linkage system code and the ENDF/B-VII.0 data library. The core consists of 36 fuel assemblies: 27 standard fuel assemblies and 9 control fuel assemblies with the uranium density of 2.8 gU/cm3 and the 235U enrichment of 19.75 wt.%. The cycle length of the core is 86 effective full-power days with the excess reactivity of 9600 and 1039 pcm at the beginning of cycle and the end of cycle, respectively. The highest power rate and the highest discharged burnup of fuel assembly are 393.49 kW and 56.74% loss of 235U, respectively. Thermal hydraulics analysis has also been conducted using the PLTEMP4.2 code for evaluating the safety parameters at a steady state of the hottest channel. The maximum temperatures of coolant and fuel cladding are 66.0°C and 83.0°C, respectively. This value is lower than the design limit of 98°C for cladding temperature. Thermal fluxes at the vertical irradiation channels and the horizontal beam ports have been evaluated. The maximum thermal fluxes of 2.5 × 1014 and 8.9 ×1013 n·cm−2·s−1 are found at the neutron trap and the beryllium reflector, respectively.

Description: Although many types of simulated radionuclides have been widely used as a substitute for actual nuclear waste in the studies of nuclear waste solidification, the understanding of the applicability and validity of simulated radionuclides is still insufficient. In particular, the selection and use of simulated radionuclides, which can play a decisive role in the accuracy of the experimental results, still lack unified or integrated references. This paper provides a critical review on the selection, experimental methods, and applicability of the most commonly studied simulated radionuclides, followed by a careful discussion and recommendation of simulated radionuclides suitable for different solidified bodies. The main factors (e.g., temperature, pH, and atmosphere) affecting the choice of simulated radionuclides were analyzed in detail. This work helps to integrate the selection and use of simulated radionuclides, and it will be beneficial for improving the effectiveness of nuclide solidification research.

Description: The thermal hydraulic and neutronics coupling analysis is an important part of the high-fidelity simulation for nuclear reactor core. In this paper, a thermal hydraulic and neutronics coupling method was proposed for the plate type fuel reactor core based on the Fluent and Monte Carlo code. The coupling interface module was developed using the User Defined Function (UDF) in Fluent. The three-dimensional thermal hydraulic model and reactor core physics model were established using Fluent and Monte Carlo code for a typical plate type fuel assembly, respectively. Then, the thermal hydraulic and neutronics coupling analysis was performed using the developed coupling code. The simulation results with coupling and noncoupling analysis methods were compared to demonstrate the feasibility of coupling code, and it shows that the accuracy of the proposed coupling method is higher than that of the traditional method. Finally, the fuel assembly blockage accident was studied based on the coupling code. Under the inlet 30% blocked conditions, the maximum coolant temperature would increase around 20°C, while the maximum fuel temperature rises about 30°C. The developed coupling method provides an effective way for the plate type fuel reactor core high-fidelity analysis.

Description: Electric valves have significant importance in industrial applications, especially in nuclear power plants. Keeping in view the quantity and criticality of valves in any plant, it is necessary to analyze the degradation of electric valves. However, it is difficult to inspect each valve in conventional maintenance. Keeping in view the quantity and criticality of valves in any plant, it is necessary to analyze the degradation of electric valves. Thus, there exists a genuine demand for remote sensing of a valve condition through nonintrusive methods as well as prediction of its remaining useful life (RUL). In this paper, typical aging modes have been summarized. The data for sensing valve conditions were gathered during aging experiments through acoustic emission sensors. During data processing, convolution kernel integrated with LSTM is utilized for feature extraction. Subsequently, LSTM which has an excellent ability in sequential analysis is used for predicting RUL. Experiments show that the proposed method could predict RUL more accurately compared to other typical machine learning and deep learning methods. This will further enhance maintenance efficiency of any plant.

Description: The geometries, adsorption energies, and electronic structures of Cs, Sr, and Ag atoms on matrix graphite surface with point defects were calculated and analyzed using the density functional theory (DFT) and the Perdew–Burke–Ernzerhof (PBE) formulation of the generalized gradient approximation (GGA). Three different types of point defects, i.e., single vacancy and “bridge” and “spiro” interstitials are considered using approximate van der Waals (vdW) correction methods. The results of adsorption energies show that the metal fission products of Cs, Sr, and Ag are more stable on single vacancy defects than “bridge” or “spiro” interstitial defects. This is further confirmed by the analysis of electronic structures, such as charge density difference (CDD) and density of state (DOS). All these results indicate that dangling bonds play an important role in the adsorption behaviors of metallic fission products on matrix graphite.

Description: Electrorefining is a key step in pyroprocessing. The solid cathode processing is necessary to separate the salt from the cathode of the electrorefiner since the uranium deposit in a solid cathode contains electrolyte salt. Moreover, it is very important to increase the throughput of the salt separation system due to the high uranium content of the spent nuclear fuel and high salt fraction of uranium dendrites. Therefore, in this study, the effect of deposit on the evaporation of the adhered salt in a uranium deposit was investigated by using the samples of salt in the uranium deposit and salt in the deposit of the surrogate material for the effective separation of the salt. It was found that the salt evaporation rate is dependent on the deposit type and bulk density in the crucible. Additionally, the evaporation rate was found to be lower when the deposit structure is complex; the rate also decreases as the bulk density of the deposit is increased owing to the retardation of the salt vapour transport process. It was concluded that the mass transfer of the salt vapour is an important parameter for the achievement of a high throughput performance in the salt distillation process.

Description: This paper focuses on fluid forces acting on a confined cylinder subjected to axial flow in application to fuel assembly dynamic behavior. From the literature, it is difficult to estimate the damping induced by the flow. Therefore, it is proposed to study numerically the damping fluid forces on a cylinder for various parameters. It has been observed that it increases with the smaller confinement and with the presence of an obstacle and decreases when the Reynolds number increases. Larger values correspond to a greater contribution of pressure forces. Dynamic simulations are compared to the steady ones and give different values, but the order of magnitude and general trend remain the same. Therefore, steady simulations are suitable to have a rough estimation of drag coefficients in dynamics.

Description: For nuclear reactor physics, uncertainties in the multigroup cross sections inevitably exist, and these uncertainties are considered as the most significant uncertainty source. Based on the home-developed 3D high-fidelity neutron transport code HNET, the perturbation theory was used to directly calculate the sensitivity coefficient of keff to the multigroup cross sections, and a reasonable relative covariance matrix with a specific energy group structure was generated directly from the evaluated covariance data by using the transforming method. Then, the “Sandwich Rule” was applied to quantify the uncertainty of keff. Based on these methods, a new SU module in HNET was developed to directly quantify the keff uncertainty with one-step deterministic transport methods. To verify the accuracy of the sensitivity and uncertainty analysis of HNET, an infinite-medium problem and the 2D pin-cell problem were used to perform SU analysis, and the numerical results demonstrate that acceptable accuracy of sensitivity and uncertainty analysis of the HNET are achievable. Finally, keff SU analysis of a 3D minicore was analyzed by using the HNET, and some important conclusions were also drawn from the numerical results.

Description: The new Structural Seismic Isolation System (SSIS) intends to provide high safety for important structures such as nuclear power plants, offshore oil platforms, and high-rise buildings against near-fault and long-period earthquakes. The presented SSIS structure foot base and foundation contact surfaces have been designed as any curved surfaces (spherical, elliptical, etc.) depending on the earthquake-soil-superstructure parameters, and these contact surfaces have been separated by using elastomeric (lead core rubber or laminated rubber bearings with up to 4-second period) seismic isolation devices. It would allow providing inverse pendulum behavior to the structure. As a result of this behavior, the natural period of the structure will possess greater intervals which are larger than the predominant period of the majority of the possible earthquakes including near-fault zones. Consequently, the structure can maintain its serviceability after the occurrence of strong and long-period earthquakes. This study has investigated the performance of the SSIS for the nuclear containment (SSIS-NC) structure. The finite element model of SISS-NC structure has been developed, and nonlinear dynamic analysis of the model has been conducted under the strong and long-period ground motions. The results have been presented in comparison with the conventional application method of the seismic base isolation devices for nuclear containment (CAMSBID-NC) and fixed base nuclear containment (FB-NC) structures. The base and top accelerations, effective stress, and critical shear stress responses of the SSIS-NC structure are 48.67%, 36.70%, and 32.60% on average lower than those of CAMSBID-NC structure, respectively. The result also confirms that the SSIS-NC structure did not cause resonant vibrations under long-period earthquakes. On the other hand, there is excessive deformation in the isolation layers of CAMSBID-NC structure.

Description: We are developing a practical-scale mechanical decladder that can slit nuclear spent fuel rod-cuts (hulls + pellets) on the order of several tens of kgf of heavy metal/batch to supply UO2 pellets to a voloxidation process. The mechanical decladder is used for separating and recovering nuclear fuel material from the cladding tube by horizontally slitting the cladding tube of a fuel rod. The Korea Atomic Energy Research Institute (KAERI) is improving the performance of the mechanical decladder to increase the recovery rate of pellets from spent fuel rods. However, because actual nuclear spent fuel is dangerously toxic, we need to develop simulated spent fuel rods for continuous experiments with mechanical decladders. We describe procedures to develop both simulated cladding tubes and simulated fuel rod (with physical properties similar to those of spent nuclear fuel). Performance tests were carried out to evaluate the decladding ability of the mechanical decladder using two types of simulated fuel (simulated tube + brass pellets and zircaloy-4 tube + simulated ceramic fuel rod). The simulated tube was developed for analyzing the slitting characteristics of the cross section of the spent fuel cladding tube. Simulated ceramic fuel rod (with mechanical properties similar to the pellets of actual PWR spent fuel) was produced to ensure that the mechanical decladder could slit real PWR spent fuel. We used castable powder pellets that simulate the compressive stress of the real spent UO2 pellet. The production criteria for simulated pellets with compressive stresses similar to those of actual spent fuel were determined, and the castables were inserted into zircaloy-4 tubes and sintered to produce the simulated fuel rod. To investigate the slitting characteristics of the simulated ceramic fuel rod, a verification experiment was performed using a mechanical decladder.

Description: This paper focuses on fluid forces acting on a confined cylinder subjected to axial flow in application to fuel assembly dynamic behavior. From the literature, it is difficult to estimate the damping induced by the flow. Therefore, it is proposed to study numerically the damping fluid forces on a cylinder for various parameters. It has been observed that it increases with the smaller confinement and with the presence of an obstacle and decreases when the Reynolds number increases. Larger values correspond to a greater contribution of pressure forces. Dynamic simulations are compared to the steady ones and give different values, but the order of magnitude and general trend remain the same. Therefore, steady simulations are suitable to have a rough estimation of drag coefficients in dynamics.

Description: Coupling supercritical carbon dioxide (S-CO2) Brayton cycle with Gen-IV reactor concepts could bring advantages of high compactness and efficiency. This study aims to design proper simple and recompression S-CO2 Brayton cycles working as the indirect cooling system for a mediate-temperature lead fast reactor and quantify the Brayton cycle performance with different heat rejection temperatures (from 32°C to 55°C) to investigate its potential use in different scenarios, like arid desert areas or areas with abundant water supply. High-efficiency S-CO2 Brayton cycle could offset the power conversion efficiency decrease caused by low core outlet temperature (which is 480°C in this study) and high compressor inlet temperature (which varies from 32°C to 55°C in this study). A thermodynamic analysis solver is developed to provide the analysis tool. The solver includes turbomachinery models for compressor and turbine and heat exchanger models for recuperator and precooler. The optimal design of simple Brayton cycle and recompression Brayton cycle for the lead fast reactor under water-cooled and dry-cooled conditions are carried out with consideration of recuperator temperature difference constraints and cycle efficiency. Optimal cycle efficiencies of 40.48% and 35.9% can be achieved for the recompression Brayton cycle and simple Brayton cycle under water-cooled condition. Optimal cycle efficiencies of 34.36% and 32.6% can be achieved for the recompression Brayton cycle and simple Brayton cycle under dry-cooled condition (compressor inlet temperature equals to 55°C). Increasing the dry cooling flow rate will be helpful to decrease the compressor inlet temperature. Every 5°C decrease in the compressor inlet temperature will bring 1.2% cycle efficiency increase for the recompression Brayton cycle and 0.7% cycle efficiency increase for the simple Brayton cycle. Helpful conclusions and advises are proposed for designing the Brayton cycle for mediate-temperature nuclear applications in this paper.

Description: Atmospheric dispersion modelling and radiological safety analysis have been performed for a postulated accident scenario of a generic VVER-1000 nuclear power plant using the HotSpot Health Physics code. The total effective dose equivalent (TEDE), the respiratory time-integrated air concentration, and the ground deposition concentration are calculated considering site-specific meteorological conditions. The results show that the maximum TEDE and ground deposition concentration values of 3.69E – 01 Sv and 3.80E + 06 kBq/m2 occurred at downwind distance of 0.18 km from the release point. This maximum TEDE value is recorded within a distance where public occupation is restricted. The TEDE values at distances of 5.0 km and beyond where public occupation is likely to be found are far below the annual regulatory limits of 1 mSv from public exposure in a year even in the event of worse accident scenario as set in IAEA Safety Standard No. GSR Part 3; no action related specifically to the public exposure is required. The released radionuclides might be transported to long distances but will not have any harmful effect on the public. The direction of the radionuclide emission from the release point is towards the north east. It is observed that the organ with the highest value of committed effective dose equivalent (CEDE) appears to be the thyroid. It was followed by the bone surface, lung, red marrow, and lower large intestine wall in order of decreasing CEDE value. Radionuclides including I-131, I-133, Sr-89, Cs-134, Ba-140, Xe-133, and Xe-135 were found to be the main contributors to the CEDE.

Description: A method for solidifying spent tributyl phosphate and kerosene (TBP/OK) organic liquids in a phosphate acid-based geopolymer (PAG) was investigated. The TBP/OK emulsion containing tween 80 (T80), TBP/OK organic liquids, and H3PO4 was prepared. The TBP/OK emulsion was mixed with metakaolin to obtain solidified TBP/OK forms (SPT). The compressive strength of the SPT was up to 59.19 MPa when the content of TBP/OK was 18%. The loss of compressive strength of SPT was less than 10% after immersion and less than 25% after freeze-thaw treatment. The final setting time was 40.0 h, and the shrinkage of SPT was nearly 3%. The leaching test indicated that the release of TBP/OK from hardened SPT was limited. Characterization of SPT suggested that solidification of TBP/OK using PAG occurred by physical encapsulation.

Description: The maximum fuel temperature under accident condition is the most important parameter of inherently safe characteristics of HTR-PM, and the DLOFC accident may lead to a peak accident fuel temperature. And there are a variety of uncertainty sources in the maximum fuel temperature calculations, and thus the contributions of these uncertainty sources to the final calculated maximum fuel temperature should be quantified to check whether the peak value exceed the technological limit of 1620°C or not. Eight uncertainty input parameters are selected for inclusion in this uncertainty study, and their associated 2 standard deviation uncertainties and probability density functions are specified. Then, the DLOFC thermal analyses and uncertainty analysis are performed with the home-developed ATHENA and CUSA. The numerical results indicate that the pebble-bed effective conductivity and the decay heat contribute the most of the uncertainty in the DLOFC maximum fuel temperature while this peak fuel temperature is most sensitive to the initial reactor power and the decay heat. In short, uncertainties in these selected eight parameters lead to the two standard deviation (2σ) uncertainty of ±77.6°C (or 5.2%) around the mean value of 1493°C for the maximum fuel temperature under DLOFC accident of HTR-PM. At the same time, the LHS-SVDC method of CUSA is recommended to propagate uncertainties in inputs and 100–200 model simulations seem to be sufficient to get an uncertainty prediction with full confidence.

Description: Best-Estimation Plus Uncertainty (BEPU) analysis method can provide more information to improve the reliability of calculation results than the safety analysis with conservative assumption. And the statistical sampling-based uncertainty and sensitivity analysis methods are widely used in practical applications of the multiphysics, multiscale coupling nuclear reactor system. In this paper, a novel and efficient sampling method for inputs with normal and uniform distributions is introduced and a systematic theory for uncertainty and sensitivity analysis is established based on the classical statistical theory. Then the Code of Uncertainty and Sensitivity Analysis (CUSA) is updated based on these new strategies. For applications, the explicit and implicit effects for resonance and nonresonance isotopes are studied in depth, and a simple UO2 pin cell is considered to examine the performance of CUSA and the total uncertainty and sensitivity analysis abilities. The numerical results indicate that the implicit sensitivity analysis model and the uncertainty quantification functions developed in CUSA are correct and can be used for sensitivity and uncertainty analysis in nuclear reactor calculations. Even more important, the LHS-SVDC is recommended to propagate the uncertainties in multigroup cross sections.

Description: Passive containment cooling system (PCCS) is an important passive safety facility in the large advanced pressurized water reactor. Using the physical laws, such as gravity and buoyancy, the water film/air countercurrent flow is formed in the external annular channel to keep inside temperature and pressure below the maximum design values. Due to the large curvature radius of the annular channel, one of the short arc segments is taken out, as a rectangular channel, to analyze the main water film evaporation heat transfer characteristics. Two numerical methods are used to predict the water film evaporative mass flow rate during the heat transfer process in the large-scale rectangular channel with asymmetric heating when the water film temperature is not saturated. At the same time, these numerical simulation results are validated by the experiment which is set up to study water film/air countercurrent flow heat transfer on a vertical back heating plate with 5 m in length and 1.2 m in width. It is shown that the maximum deviation between numerical simulation and experiment is 30%. In addition, the influences on these parameters, such as heat flux, evaporative mass flow rate, and water film thickness, are evaluated under the different tilted angles of the rectangular channel and horizontal plane, water/air inlet flow rates, water/air inlet temperatures, heating surface temperatures, and air inlet relative humidities. All these results can provide a good guidance for the design of PCCS in the future.

Description: Rapid 3D radiation field evaluation is the key point of occupational dose optimization for design and operation of nuclear power plant. Based on the requirement analysis from designers and operators of nuclear power plant, three key technical issues are identified and solved through the development of the RPOS system, which are rapid calculation of 3D radiation field, reconstruction of the calculated 3D radiation field based on measured data, and occupational dose optimization based on 3D radiation field. Operational measurements of dose rate from in-service nuclear power plants are used to test the RPOS system, which shows that accurate 3D radiation field can be rapidly generated by the RPOS system and effectively used on the occupational dose optimization for on-site workers. The applications of the established rapid 3D radiation field evaluation technique on HPR1000 unit design provide evidence on its feasibility in a large scale, the improvement of radiation protection design efficiency and the enhancement of ALARA assessment and justification for nuclear power plants.

Description: The prediction of flow and heat transfer characteristics of liquid sodium with CFD technology is of significant importance for the design and safety analysis of sodium-cooled fast reactor. The accuracies and uncertainties of the CFD models should be evaluated to improve the confidence of the numerical results. In this work, the uncertainties from the turbulent model, boundary conditions, and physical properties for the flow and heat transfer of liquid sodium were evaluated against the experimental data. The results of uncertainty quantization show that the maximum uncertainties of the Nusselt number and friction coefficient occurred in the transition zone from the inlet to the fully developed region in the circular tube, while they occurred near the reattachment point in the backward-facing step. Furthermore, in backward-facing step flow, the maximum uncertainty of temperature migrated from the heating wall to the geometric center of the channel, while the maximum uncertainty of velocity occurred near the vortex zone. The results of sensitivity analysis illustrate that the Nusselt number was negatively correlated with the thermal conductivity and turbulent Prandtl number, while the friction coefficient was positively correlated with the density and Von Karman constant. This work can be a reference to evaluate the accuracy of the standard k-ε model in predicting the flow and heat transfer characteristics of liquid sodium.

Description: The source term for the JRTR research reactor is derived under an assumed hypothetical severe accident resulting in generation of the most severe consequences. The reactor core is modeled based on the reactor technical design specifications, and the fission products inventory is calculated by using the SCALE/TRITON depletion sequence to perform burnup and decay analyses via coupling the NEWT 2-D transport lattice code to the ORIGEN-S fuel depletion code. Fifty radioisotopes contributed to the evaluation, resulting in a source term of 3.7 × 1014 Bq. Atmospheric dispersion was evaluated using the Gaussian plume model via the HOTSPOT code. The plume centerline total effective dose (TED) was found to exceed the IAEA limits for occupational exposure of 0.02 Sv; the results showed that the maximum dose is 200 Sv within 200 m from the reactor, under all the weather stability classes, after which it starts to decrease with distance, reaching 0.1 Sv at 1 km from the reactor. The radiation dose plume centerlines continue to the exceed international basic safety standards annual limit of 1 mSv for public exposure, up to 80 km from the reactor.

Description: In pebble-bed high temperature gas-cooled reactor, gaps widely exist between graphite blocks and carbon bricks in the reactor core vessel. The bypass helium flowing through the gaps affects the flow distribution of the core and weakens the effective cooling of the core by helium, which in turn affects the temperature distribution and the safety features of the reactor. In this paper, the thermal hydraulic analysis models of HTR-10 with bypass flow channels simulated at different positions are designed based on the flow distribution scheme of the original core models and combined with the actual position of the core bypass flow. The results show that the bypass coolant flowing through the reflectors enhances the heat transfer of the nearby components efficiently. The temperature of the side reflectors and the carbon bricks is much lower with more side bypass coolant. The temperature distribution of the central region in the pebble bed is affected by the bypass flow positions slightly, while that of the peripheral area is affected significantly. The maximum temperature of the helium, the surface, and center of the fuel elements rises as the bypass flow ratio becomes larger, while the temperature difference between them almost keeps constant. When the flow ratio of each part keeps constant, the maximum temperature almost does not change with different bypass flow positions.

Description: CAP1400 nuclear island structure is an advanced and novel nuclear power plant structure. In order to explore the seismic response characteristics of CAP1400 nuclear island structure on soft rock sites, a three-dimensional refined nonlinear seismic response analysis model was established for a soft rock foundation-nuclear island structure system using ABAQUS software. The influences of the input ground motion intensity and the frequency spectrum characteristics on the acceleration, relative displacement, and floor response spectrum, as well as the critical shear wave velocity of nonbedrock sites for CAP1400 nuclear island structure, were proposed. The results suggested that the increasing amplitude of the peak acceleration and relative displacement of nuclear island structure decreased as the soft rock site entered a nonlinear state, and the high-frequency components of the input ground motion became more abundant. Specifically, the earthquake response was the largest at the cooling water tank on the top of the shield building, which was the focus of the seismic research on nuclear island structure. Due to the influence of the ground motion frequency spectrum characteristics and the nonbedrock site effect, the peak acceleration, peak relative displacement, and acceleration response spectrum of the nuclear island structure showed different changing trends for the near-field and far-field ground motions. Based on the influence of the site shear wave velocity on the seismic response of nuclear island structure, it was recommended that the critical shear wave velocity of nonbedrock sites for CAP1400 nuclear island structure should be 1250 m/s, and the nuclear island structure-foundation dynamic interaction could be ignored at this time. The research conclusions could provide some technical support and theoretical basis for the construction and seismic performance research of CAP1400 and other nuclear power plants.

Description: The experiments are carried out in a three-dimensional channel with a screw conveyor, which plays the role of granular drives for the granular flow system and determines the injection of granular in the test target section. The jam-to-dense transition of granular flow is studied with the different inclination angle. The results show that, with a fixed diameter of hopper orifice and initial filling position, there is a change from jam to dense when the inclination angle larger than 22°. Variation of the flow rate with elevated frequency of the screw conveyor is further studied. The flow pattern is changed from dilute to dense with increasing rotation frequency of the screw rod. When the rotation frequency is larger than 5 Hz, the flow is dense. The dynamic balance of the interface between dilute to dense granular is observed in the main target section. We further research the dynamic interface by measuring the highest and lowest location with time and also simulate the gravity flow rate and screw conveyor flow rate with EDEM. From the results, we find that the interface between dilute flow and dense flow is influenced by the combined action of crew conveyor flow and dense gravity flow.

Description: Uncertainty analyses of fission product yields are indispensable in evaluating reactor burnup and decay heat calculation credibility. Compared with neutron cross section, there are fewer uncertainty analyses conducted and it has been a controversial topic by lack of properly estimated covariance matrix as well as adequate sampling method. Specifically, the conventional normal-based sampling method in sampling large uncertainty independent fission yields could inevitably generate nonphysical negative samples. Cutting off these samples would introduce bias into uncertainty results. Here, we evaluate thermal neutron-induced U-235 independent fission yields covariance matrix by the Bayesian updating method, and then we use lognormal-based sampling method to overcome the negative fission yields samples issue. Fission yields uncertainty contribution to effective multiplication factor and several fission products’ atomic densities at equilibrium core of pebble-bed HTGR are quantified and investigated. The results show that the lognormal-based sampling method could prevent generating negative yields samples and maintain the skewness of fission yields distribution. Compared with the zero cut-off normal-based sampling method, the lognormal-based sampling method evaluates the uncertainty of effective multiplication factor and atomic densities are larger. This implies that zero cut-off effect in the normal-based sampling method would underestimate the fission yields uncertainty contribution. Therefore, adopting the lognormal-based sampling method is crucial for providing reliable uncertainty analysis results in fission product yields uncertainty analysis.

Description: Tremendous work has been done in the Light Water Reactor (LWR) Modelling and Simulation (M&S) uncertainty quantification (UQ) within the framework of the Organization for Economic Cooperation and Development (OECD)/Nuclear Energy Agency (NEA) LWR Uncertainty Analysis in Modelling (UAM) benchmark, which aims to investigate the uncertainty propagation in all M&S stages of the LWRs and to guide uncertainty and sensitivity analysis methodology development. The Best-Estimate Plus Uncertainty (BEPU) methodologies have been developed and implemented within the framework of the LWR UAM benchmark to provide a realistic predictive simulation capability without compromising the safety margins. This paper describes the current status of the methodological development, assessment, and integration of the BEPU methodology to facilitate the multiscale, multiphysics LWR core analysis. The comparative analysis of the results in the stand-alone multiscale neutronics phase (Phase I) is first reported for understanding the general trend of the uncertainty of core parameters due to the nuclear data uncertainty. It was found that the predicted uncertainty of the system eigenvalue is highly dependent on the choice of the covariance libraries used in the UQ process and is less sensitive to the solution method, nuclear data library, and UQ method. High-to-Low (Hi2Lo) model information approaches for multiscale M&S are introduced for the core single physics phase (Phase II). In this phase, the other physics (fuel and moderator), providing feedback to neutronics M&S in a LWR core, and time-dependent phenomena are considered. Phase II is focused on uncertainty propagation in single physics models which are components of the LWR core coupled multiphysics calculations. The paper discusses the link and interactions between Phase II to the multiphysics core and system phase (Phase III), that is, the link between uncertainty propagation in single physics on local scale and multiphysics uncertainty propagation on the core scale. Particularly, the consistency in uncertainty assessment between higher-fidelity models implemented in fuel performance codes and the rather simplified models implemented in thermal-hydraulics codes, to be used for coupling with neutronics in Phase III is presented. Similarly, the uncertainty quantification on thermal-hydraulic models is established on a relatively small scale, while these results will be used in Phase III at the core scale, sometimes with different codes or models. Lastly, the up-to-date UQ method for the coupled multiphysics core calculation in Phase III is presented, focusing on the core equilibrium cycle depletion calculation with associated uncertainties.

Description: Self-irradiation can affect durability of glasses used to immobilize high-level nuclear waste (HLW). The stability of glasses can also be indirectly affected by the radiolytic changes in contact water leading to decrease in its pH although this is unlikely to occur for disposal systems where the interaction of groundwater with glass is delayed to times when radiation dose rates are decreased to levels insignificant to cause such effects. Besides, interaction of the water influenced by radiation with other repository protective elements (container and bentonite) will suppress the radiolytic changes. Literature analysis shows practical absence or very weak effect of self-irradiation on structure and characteristics of borosilicate glasses with typical content of nuclear waste. Data for aluminophosphate glass used in Russia have showed that, after γ-irradiation with a dose of 6.2·107 Gy, the leaching rates of elements were decreased approximately twice relatively to pristine samples.

Description: This study was carried out to examine the removal of rare earth (RE) elements by electrodeposition for the purification and reuse of LiCl-KCl salt after electrorefining and electrowinning. The electrochemical behavior of RE elements (Dy and Gd) in LiCl-KCl-DyCl3-GdCl3 at 500°C was investigated using the cyclic voltammetry (CV) technique using Mo and Mg electrodes. It was observed that the reduction potential of the RE elements shifted at the Mg electrode owing to the alloy formation with Mg (RE-Mg alloy). Subsequently, a series of potentiostatic electrolysis tests were conducted to remove the RE elements in the salt and check the formation of deposits at the Mg and Mo electrodes. The scanning electron microscopy-energy dispersive X-ray spectroscopy (SEM/EDS) technique was used to confirm that the reduced RE metals were deposited on the surface of the Mg electrode. However, no significant deposit on the Mo electrode was observed, and a mud-like deposit was found on the bottom of the electrochemical cell. The salt analysis performed by employing the inductively coupled plasma-optical emission spectrometry (ICP-OES) indicated that the removal efficiency of Dy3+ and Gd3+ through electrodeposition was 83.5∼95.2 and 91.6∼95.2%, respectively.

Description: In this study, the Best Estimate Plus Uncertainty (BEPU) approach is developed for the systematic quantification and propagation of uncertainties in the modelling and simulation of lead-cooled fast reactors (LFRs) and applied to the demonstration LFR (DLFR) initially investigated by Westinghouse. The impact of nuclear data uncertainties based on ENDF/B-VII.0 covariances is quantified on lattice level using the generalized perturbation theory implemented with the Monte Carlo code Serpent and the deterministic code PERSENT of the Argonne Reactor Computational (ARC) suite. The quantities of interest are the main eigenvalue and selected reactivity coefficients such as Doppler, radial expansion, and fuel/clad/coolant density coefficients. These uncertainties are then propagated through safety analysis, carried out using the MiniSAS code, following the stochastic sampling approach in DAKOTA. An unprotected transient overpower (UTOP) scenario is considered to assess the effect of input uncertainties on safety parameters such as peak fuel and clad temperatures. It is found that in steady state, the multiplication factor shows the most sensitivity to perturbations in 235U fission, 235U ν, and 238U capture cross sections. The uncertainties of 239Pu and 238U capture cross sections become more significant as the fuel is irradiated. The covariance of various reactivity feedback coefficients is constructed by tracing back to common uncertainty contributors (i.e., nuclide-reaction pairs), including 238U inelastic, 238U capture, and 239Pu capture cross sections. It is also observed that nuclear data uncertainty propagates to uncertainty on peak clad and fuel temperatures of 28.5 K and 70.0 K, respectively. Such uncertainties do not impose per se threat to the integrity of the fuel rod; however, they sum to other sources of uncertainties in verifying the compliance of the assumed safety margins, suggesting the developed BEPU method necessary to provide one of the required insights on the impact of uncertainties on core safety characteristics.

Description: Early core degradation determines the amount of hydrogen generated by cladding oxidation as well as the temperature, the mass, and the composition of corium that further relocates into the lower head of reactor pressure vessel (RPV), which is essential for the effectiveness analysis of in-vessel retention (IVR) and hydrogen recombiners. In this paper, the mechanisms of controlling phenomena in the early phase of core degradation are analysed at first. Then, numerical models adopted to calculate (1) core heating up, (2) cladding oxidation, (3) dissolution between molten zirconium and fuel pellets, and (4) formation of a molten pool in the core active section are presented. Compared with integral codes for severe accident analysis (such as MAAP and MELCOR), the models in this paper are established at the fuel pin level and the calculation is performed in 3D, which can capture the detail local phenomena during the core degradation and eliminate the average effect due to equivalent rings used in integral codes. In addition, most of the control equations in this paper are calculated by implicit schemes, which can improve the accuracy and stability of the calculation. In the simulation, the calculation oxidation is calculated by using the oxygen diffusion model, while the dissolution is calculated with Kim, Hayward, Hofmann, and IBRAE models to perform uncertainty analysis. For the validation, the cladding oxidation model is verified by Olander theoretical cases in the conditions of both steam-rich and steam-starved. The dissolution models are validated by the RIAR experiment. The code is overall verified by Phebus FPT0 on the integral phase of core early degradation. According to the simulation results, it can be inferred that the dissolution reaction between the molten zirconium and fuel pellets is the main reason for the melting of UO2 at low temperature. In the case of starved steam, part of the fuel pellets can melt down even at 2248 K and relocate to the bottom of the core, which is much lower than the melting point of UO2 (3113 K).

Description: With the wide application of sea-based reactors, the impact of ocean conditions on the safety performance of reactors has gradually attracted attention. In this paper, by establishing the thermal hydraulic transient analysis model and the critical heat flux (CHF) model of natural circulation system, the CHF characteristics in the rectangular channel of natural self-feedback conditions under ocean conditions are studied. The results show that the additional acceleration field generated by ocean conditions will affect the thermal hydraulic parameters of the natural circulation system, that is, the external macroscopic thermal hydraulic field. On the other hand, the boiling crisis mechanism will be affected, that is, the force on the bubble and the thickness of the liquid film. Within the parameters of the study, ocean conditions have a great impact on CHF of natural circulation, and the maximum degradation of CHF is about 45%. The obtained analysis results are significant to the improvement of design and safety operation of the reactor system.

Description: The luminescence of Kr-Xe, Ar-Kr, and Ar-Xe mixtures was studied in the spectral range 300–970 nm when excited by 6Li (n, α)3 H nuclear reaction products in the core of a nuclear reactor. Lithium was deposited on walls of experimental cell in the form of a capillary-porous structure, which made it possible to measure up to a temperature of 730 K. The temperature dependence of the radiation intensity of noble gas atoms, alkali metals, and heteronuclear ionic noble gas molecules was studied. Also, as in the case of single-component gases, the appearance of lithium lines and impurities of sodium and potassium is associated with vaporization during the release of nuclear reaction products from the lithium layer. The excitation of lithium atoms occurs mainly as a result of the Penning process of lithium atoms on noble gas atoms in the 1s states and subsequent ion-molecular reactions. Simultaneous radiation at transitions of atoms of noble gases and lithium, heteronuclear ion molecules of noble gases allows us to increase the efficiency of direct conversion of nuclear energy into light.

Description: This paper proposes to design security measures based on the radioactive material package as the basic unit. The principle of four-layer defense in depth is put forward. Based on the concept of self-security intelligence, combined with out-of-vehicle monitoring, in-vehicle monitoring, and Beidou positioning technology, a security system for transport of radioactive materials was designed. It realized the perception, early warning, delay, and alarm functions and greatly improved the security.

Description: Multiphysics coupling of neutronics/thermal-hydraulics models is essential for accurate modeling of nuclear reactor systems with physics feedback. In this work, SCALE/TRACE coupling is used for neutronic analysis and spent fuel validation of BWR assemblies, which have strong coolant feedback. 3D axial power profiles with coolant feedback are captured in these advanced simulations. The methodology is applied to two BWR assemblies (2F2DN23/SF98 and 2F2D1/F6), discharged from the Fukushima Daini-2 unit. Coupling is performed externally, where the SCALE/T5-DEPL module transfers axial power data in all axial nodes to TRACE, which in turn calculates the coolant density and temperature for each of these nodes. Within a burnup step, the data exchange process is repeated until convergence of all coupling parameters (axial power, coolant density, and coolant temperature) is observed. Analysis of axial power, criticality, and coolant properties at the assembly level is used to verify the coupling process. The 2F2D1/F6 benchmark seems to have insignificant void feedback compared to 2F2DN23/SF98 case, which experiences large power changes during operation. Spent fuel isotopic data are used to validate the coupling methodology, which demonstrated good results for uranium isotopes and satisfactory results for other actinides. This work has a major challenge of lack of documented data to build the coupled models (boundary conditions, control rod history, spatial location in the core, etc.), which encourages more advanced methods to approximate such missing data to achieve better modeling and simulation results.

Description: In order to enhance the ability of severe accident mitigation for Pressurised Water Reactor (PWR), different kinds of severe accident mitigation strategies have been proposed. In-Vessel Retention (IVR) is one of the important severe accident management means by External Reactor Vessel Cooling. Reactor cavity would be submerged to cool the molten corium when a severe accident happens. The success criterion of IVR strategy is that the heat flux which transfers from the corium pool must be lower than the local critical heat flux (CHF) of the reactor pressure vessel (RPV) outside wall and the residual thickness of the RPV wall can maintain the integrity. The residual thickness of RPV is determined by the heat flux transfer from the corium pool and the cooling capability of outer wall of the RPV. There are various factors which would influence the CHF and the cooling capability of outer wall of the RPV. In order to verify the optimized design which is beneficial to the heat transfer and the natural circulation outside the actual reactor vessel, a large-scale Reactor Vessel External Cooling Test (REVECT) facility has been built. A large number of sensitivity tests were carried out, to study how these sensitivity factors affect CHF value and natural circulation. Based on the test results, the structure of the test section flow channel has an obvious effect on the CHF distribution. The flow channel optimized can effectively enhance the CHF value, especially to enhance the CHF value near the “heat focus” region of the molten pool. The water level in the reactor pit has also a great impact on the natural circulation flow. Although natural circulation can be maintained with a low water level, it will lead to a decrease of the cooling capacity. Meanwhile, some noteworthy test phenomena have been found, which are also essential for the design of the reactor pit flooding system.

Description: A new radioactive liquid waste cementation facility was under commissioning recently in the Institute of Nuclear and New Energy Technology of Tsinghua University, which is designed to simultaneously process multiple intermediate-level radioactive waste drums. Therefore, the multiple volume sources and the scattering effect becomes a key issue in its radiation protection. For this purpose, the Monte Carlo program FLUKA code and experimental measurement were both adopted. In the FLUKA simulation, five different scenarios were considered, i.e., one drum, two drums, four drums, six drums, and eight drums. For the multiple volume sources, the source subroutine code of FLUKA was rewritten to realize the sampling. The complex shielding also leads to a deep penetration problem; hence, the optimization algorithm and variance reduction techniques were adopted. During the measurement, two scenarios, outdoor and indoor, were carried out separately representing the dose field when only one drum is considered and when the scattering effect is considered. A comparison between the experiments and calculations shows very good agreement. From both of the Monte Carlo simulation and the experimental measurement, it can be drawn that, in the horizontal direction, with the increase of the drum number, the dose rate increases very little, while in the vertical direction, the increase of the dose rate is very obvious with the increase of the drum number. The complicated source term sampling methods, the optimization algorithm and variance reduction techniques, and the experimental verification can provide valuable references for the similar scattering problem in radiation protection and shielding design.

Description: For nuclear reactor physics, uncertainties in the multigroup cross sections inevitably exist, and these uncertainties are considered as the most significant uncertainty source. Based on the home-developed 3D high-fidelity neutron transport code HNET, the perturbation theory was used to directly calculate the sensitivity coefficient of keff to the multigroup cross sections, and a reasonable relative covariance matrix with a specific energy group structure was generated directly from the evaluated covariance data by using the transforming method. Then, the “Sandwich Rule” was applied to quantify the uncertainty of keff. Based on these methods, a new SU module in HNET was developed to directly quantify the keff uncertainty with one-step deterministic transport methods. To verify the accuracy of the sensitivity and uncertainty analysis of HNET, an infinite-medium problem and the 2D pin-cell problem were used to perform SU analysis, and the numerical results demonstrate that acceptable accuracy of sensitivity and uncertainty analysis of the HNET are achievable. Finally, keff SU analysis of a 3D minicore was analyzed by using the HNET, and some important conclusions were also drawn from the numerical results.

Description: Coupling supercritical carbon dioxide (S-CO2) Brayton cycle with Gen-IV reactor concepts could bring advantages of high compactness and efficiency. This study aims to design proper simple and recompression S-CO2 Brayton cycles working as the indirect cooling system for a mediate-temperature lead fast reactor and quantify the Brayton cycle performance with different heat rejection temperatures (from 32°C to 55°C) to investigate its potential use in different scenarios, like arid desert areas or areas with abundant water supply. High-efficiency S-CO2 Brayton cycle could offset the power conversion efficiency decrease caused by low core outlet temperature (which is 480°C in this study) and high compressor inlet temperature (which varies from 32°C to 55°C in this study). A thermodynamic analysis solver is developed to provide the analysis tool. The solver includes turbomachinery models for compressor and turbine and heat exchanger models for recuperator and precooler. The optimal design of simple Brayton cycle and recompression Brayton cycle for the lead fast reactor under water-cooled and dry-cooled conditions are carried out with consideration of recuperator temperature difference constraints and cycle efficiency. Optimal cycle efficiencies of 40.48% and 35.9% can be achieved for the recompression Brayton cycle and simple Brayton cycle under water-cooled condition. Optimal cycle efficiencies of 34.36% and 32.6% can be achieved for the recompression Brayton cycle and simple Brayton cycle under dry-cooled condition (compressor inlet temperature equals to 55°C). Increasing the dry cooling flow rate will be helpful to decrease the compressor inlet temperature. Every 5°C decrease in the compressor inlet temperature will bring 1.2% cycle efficiency increase for the recompression Brayton cycle and 0.7% cycle efficiency increase for the simple Brayton cycle. Helpful conclusions and advises are proposed for designing the Brayton cycle for mediate-temperature nuclear applications in this paper.

Description: Best-Estimation Plus Uncertainty (BEPU) analysis method can provide more information to improve the reliability of calculation results than the safety analysis with conservative assumption. And the statistical sampling-based uncertainty and sensitivity analysis methods are widely used in practical applications of the multiphysics, multiscale coupling nuclear reactor system. In this paper, a novel and efficient sampling method for inputs with normal and uniform distributions is introduced and a systematic theory for uncertainty and sensitivity analysis is established based on the classical statistical theory. Then the Code of Uncertainty and Sensitivity Analysis (CUSA) is updated based on these new strategies. For applications, the explicit and implicit effects for resonance and nonresonance isotopes are studied in depth, and a simple UO2 pin cell is considered to examine the performance of CUSA and the total uncertainty and sensitivity analysis abilities. The numerical results indicate that the implicit sensitivity analysis model and the uncertainty quantification functions developed in CUSA are correct and can be used for sensitivity and uncertainty analysis in nuclear reactor calculations. Even more important, the LHS-SVDC is recommended to propagate the uncertainties in multigroup cross sections.

Description: The maximum fuel temperature under accident condition is the most important parameter of inherently safe characteristics of HTR-PM, and the DLOFC accident may lead to a peak accident fuel temperature. And there are a variety of uncertainty sources in the maximum fuel temperature calculations, and thus the contributions of these uncertainty sources to the final calculated maximum fuel temperature should be quantified to check whether the peak value exceed the technological limit of 1620°C or not. Eight uncertainty input parameters are selected for inclusion in this uncertainty study, and their associated 2 standard deviation uncertainties and probability density functions are specified. Then, the DLOFC thermal analyses and uncertainty analysis are performed with the home-developed ATHENA and CUSA. The numerical results indicate that the pebble-bed effective conductivity and the decay heat contribute the most of the uncertainty in the DLOFC maximum fuel temperature while this peak fuel temperature is most sensitive to the initial reactor power and the decay heat. In short, uncertainties in these selected eight parameters lead to the two standard deviation (2σ) uncertainty of ±77.6°C (or 5.2%) around the mean value of 1493°C for the maximum fuel temperature under DLOFC accident of HTR-PM. At the same time, the LHS-SVDC method of CUSA is recommended to propagate uncertainties in inputs and 100–200 model simulations seem to be sufficient to get an uncertainty prediction with full confidence.

Description: A method for solidifying spent tributyl phosphate and kerosene (TBP/OK) organic liquids in a phosphate acid-based geopolymer (PAG) was investigated. The TBP/OK emulsion containing tween 80 (T80), TBP/OK organic liquids, and H3PO4 was prepared. The TBP/OK emulsion was mixed with metakaolin to obtain solidified TBP/OK forms (SPT). The compressive strength of the SPT was up to 59.19 MPa when the content of TBP/OK was 18%. The loss of compressive strength of SPT was less than 10% after immersion and less than 25% after freeze-thaw treatment. The final setting time was 40.0 h, and the shrinkage of SPT was nearly 3%. The leaching test indicated that the release of TBP/OK from hardened SPT was limited. Characterization of SPT suggested that solidification of TBP/OK using PAG occurred by physical encapsulation.

Description: The new Structural Seismic Isolation System (SSIS) intends to provide high safety for important structures such as nuclear power plants, offshore oil platforms, and high-rise buildings against near-fault and long-period earthquakes. The presented SSIS structure foot base and foundation contact surfaces have been designed as any curved surfaces (spherical, elliptical, etc.) depending on the earthquake-soil-superstructure parameters, and these contact surfaces have been separated by using elastomeric (lead core rubber or laminated rubber bearings with up to 4-second period) seismic isolation devices. It would allow providing inverse pendulum behavior to the structure. As a result of this behavior, the natural period of the structure will possess greater intervals which are larger than the predominant period of the majority of the possible earthquakes including near-fault zones. Consequently, the structure can maintain its serviceability after the occurrence of strong and long-period earthquakes. This study has investigated the performance of the SSIS for the nuclear containment (SSIS-NC) structure. The finite element model of SISS-NC structure has been developed, and nonlinear dynamic analysis of the model has been conducted under the strong and long-period ground motions. The results have been presented in comparison with the conventional application method of the seismic base isolation devices for nuclear containment (CAMSBID-NC) and fixed base nuclear containment (FB-NC) structures. The base and top accelerations, effective stress, and critical shear stress responses of the SSIS-NC structure are 48.67%, 36.70%, and 32.60% on average lower than those of CAMSBID-NC structure, respectively. The result also confirms that the SSIS-NC structure did not cause resonant vibrations under long-period earthquakes. On the other hand, there is excessive deformation in the isolation layers of CAMSBID-NC structure.

Description: Atmospheric dispersion modelling and radiological safety analysis have been performed for a postulated accident scenario of a generic VVER-1000 nuclear power plant using the HotSpot Health Physics code. The total effective dose equivalent (TEDE), the respiratory time-integrated air concentration, and the ground deposition concentration are calculated considering site-specific meteorological conditions. The results show that the maximum TEDE and ground deposition concentration values of 3.69E – 01 Sv and 3.80E + 06 kBq/m2 occurred at downwind distance of 0.18 km from the release point. This maximum TEDE value is recorded within a distance where public occupation is restricted. The TEDE values at distances of 5.0 km and beyond where public occupation is likely to be found are far below the annual regulatory limits of 1 mSv from public exposure in a year even in the event of worse accident scenario as set in IAEA Safety Standard No. GSR Part 3; no action related specifically to the public exposure is required. The released radionuclides might be transported to long distances but will not have any harmful effect on the public. The direction of the radionuclide emission from the release point is towards the north east. It is observed that the organ with the highest value of committed effective dose equivalent (CEDE) appears to be the thyroid. It was followed by the bone surface, lung, red marrow, and lower large intestine wall in order of decreasing CEDE value. Radionuclides including I-131, I-133, Sr-89, Cs-134, Ba-140, Xe-133, and Xe-135 were found to be the main contributors to the CEDE.

Description: This paper focuses on fluid forces acting on a confined cylinder subjected to axial flow in application to fuel assembly dynamic behavior. From the literature, it is difficult to estimate the damping induced by the flow. Therefore, it is proposed to study numerically the damping fluid forces on a cylinder for various parameters. It has been observed that it increases with the smaller confinement and with the presence of an obstacle and decreases when the Reynolds number increases. Larger values correspond to a greater contribution of pressure forces. Dynamic simulations are compared to the steady ones and give different values, but the order of magnitude and general trend remain the same. Therefore, steady simulations are suitable to have a rough estimation of drag coefficients in dynamics.

Description: We are developing a practical-scale mechanical decladder that can slit nuclear spent fuel rod-cuts (hulls + pellets) on the order of several tens of kgf of heavy metal/batch to supply UO2 pellets to a voloxidation process. The mechanical decladder is used for separating and recovering nuclear fuel material from the cladding tube by horizontally slitting the cladding tube of a fuel rod. The Korea Atomic Energy Research Institute (KAERI) is improving the performance of the mechanical decladder to increase the recovery rate of pellets from spent fuel rods. However, because actual nuclear spent fuel is dangerously toxic, we need to develop simulated spent fuel rods for continuous experiments with mechanical decladders. We describe procedures to develop both simulated cladding tubes and simulated fuel rod (with physical properties similar to those of spent nuclear fuel). Performance tests were carried out to evaluate the decladding ability of the mechanical decladder using two types of simulated fuel (simulated tube + brass pellets and zircaloy-4 tube + simulated ceramic fuel rod). The simulated tube was developed for analyzing the slitting characteristics of the cross section of the spent fuel cladding tube. Simulated ceramic fuel rod (with mechanical properties similar to the pellets of actual PWR spent fuel) was produced to ensure that the mechanical decladder could slit real PWR spent fuel. We used castable powder pellets that simulate the compressive stress of the real spent UO2 pellet. The production criteria for simulated pellets with compressive stresses similar to those of actual spent fuel were determined, and the castables were inserted into zircaloy-4 tubes and sintered to produce the simulated fuel rod. To investigate the slitting characteristics of the simulated ceramic fuel rod, a verification experiment was performed using a mechanical decladder.

Description: A modified multiobjective self-adaptive differential evolution algorithm (MMOSADE) is presented in this paper to improve the accuracy of multiobjective optimization design in the nuclear power system. The performance of the MMOSADE is tested by the ZDT test function set and compared with classical evolutionary algorithms. The results indicate that MMOSADE has a better performance in convergence and diversity. Based on the MMOSADE, a multiobjective optimization design platform for the nuclear power system is proposed, and the application of which is carried out. The evaluation program of the PRHR-HX in AP1000 is developed, and its reliability is verified. The optimal design schemes of PHHR-HX are obtained by utilizing the multiobjective optimization design platform. The results show that the optimal design schemes can envelop the prototype design scheme. This conclusion proves that the optimization design platform proposed in this paper is effective and feasible.

Description: COMSOL Multiphysics has been used to conduct thermal-hydraulic analysis in multiple nuclear applications. The aim of this study is to benchmark the prediction accuracy of COMSOL Multiphysics in performing thermal-hydraulic analysis of TRIGA (Training, Research, Isotopes, General Atomics) reactors such as the Geological Survey TRIGA Reactor (GSTR) by comparing its predictions with RELAP5 (a widely used code in nuclear thermal-hydraulic analysis) results and experimental data. The GSTR type is Mark I with a full thermal power of 1 MW, and it resides at the Denver Federal Center (DFC) in Colorado. The numerical investigation of the present work is carried out by developing single-subchannel thermal-hydraulic models of the GSTR utilizing RELAP5 and COMSOL codes. The models estimate the temperatures (fuel, outer clad, and coolant) and water flow patterns in the core as well as fuel element powers at which void starts to form within the coolant subchannels. Then, these models’ predictions are quantitatively evaluated and compared with the measured data.

Description: A practical scale mechanical decladder that can slit spent nuclear fuel rod-cuts (hulls + pellets) of several tens of kg HM/batch is being developed to supply UO2 pellets to a voloxidation process. The mechanical decladder is an apparatus for separating and recovering fuel material and cladding tubes by horizontally slitting the cladding tube of a fuel rod and a defective irradiated fuel rod. In this study, we address the engineering design of the mechanical decladder for the pretesting of rod-cut slitting. To obtain the requirements of the mechanical decladder, we first manufactured a slitter for testing based on the decladding and shearing conditions of hulls and pellets. The performance test of the testing device for decladding was carried out using a 2-CUT blade module and a 3-CUT blade module. We evaluated the decladding methods for the mechanical decladder and selected the 3-CUT blade module based on the results. A buckling measurement instrument was used to perform a buckling verification test according to the length of a rod-cut and to determine decladder dimensions. The optimum decladding rod-cut length for buckling prevention was calculated. Furthermore, we analyzed the decladding mechanism for various slitting methods. Design/fabrication and preliminary tests of the practical scale mechanical decladder were also performed. For this purpose, we constructed the main mechanism by utilizing the SolidWorks modeling and analysis program and fabricated a new mechanical decladder. Based on the derived requirements, a mechanical decladder with three main modules was designed and fabricated for testing. Simulated rod-cuts of zircaloy were also manufactured to test the basic performance of the decladder, and a data acquisition system was constructed using RSC 232 to measure decladding force and velocity. In the basic test, the rod-cut was completely sectioned into three evenly spaced locations by the new mechanical decladder.

Description: CIPS is a shift in the axial power towards the bottom half of the core, also known as axial offset anomaly (AOA), which results from the deposited of corrosion products during an operation. The main reason of CIPS is the solute particles especially boron compounds concentrated inside the porous deposit. The impact of CIPS is that the axial power distribution control may be more difficult and the shutdown margin can be decreased simultaneously. Besides, it also requires estimated critical condition (ECC) calculations to account for the effects of AOA. In this article, thermal-hydraulic subchannel code and boron deposit model have been combined to analyze the CIPS risk. The neutronics codes deal with the generation of homogenized neutron cross section as well as the calculation of local power factor. A simple rod assembly is analyzed with this combined method and simulation results are presented. Simulation results provide the boron hideout amount inside crud deposits and power shapes. The obtained results clearly show the power shape suppression in regions where crud deposits exist, which is a clear indication of CIPS phenomenon. And the CIPS effects on CHF have also been investigated. Result shows a margin of DNBR decrease in the crud case.

Description: After the severe accident (SA) occurred at the Three-Miles Island Nuclear Power Plant (NPP), important efforts on the investigation of the different phenomena during this kind of accidents have been started. Several experimental campaigns investigating one phenomenon at time or the combination of two or more phenomena have been performed. Today, the Phébus experimental campaign is probably the most important activity on the evaluation of the coupling among different phenomena. Four out of five tests investigated the degradation of an intact Pressurized Water Reactor (PWR) fuel bundle and the subsequent transport of Fission Products (FP) and Structural Materials (SM) through the primary circuit and into the containment, while the fifth test was only the degradation of a bed of PWR fuel bundle debris. These tests were performed between 1990 and 2010 at the CEA Cadarache laboratories (France) in a 5000:1 scaled facility. The main four tests varied the employed control rod materials, the fuel burn-up, and the oxidizing conditions of the atmosphere (strongly or weakly). The outcomes of this experimental campaign created a solid base for the understanding of the involved phenomena and allowed the development of models and software codes capable of simulating the evolution of a SA in a real NPP. ASTEC and MELCOR were two of the main SA codes profiting from the results of this Phébus campaign. These two codes were further improved in the latest years to account for the findings obtained in more recent experimental campaigns. A continuous verification and validation work is then necessary to check how the newer code’s versions reproduce the tests performed in these older experimental campaigns such as Phébus one. The present work is intended to be the final step of a series of publications covering the activities carried out at University of Pisa with the ASTEC and the MELCOR SA codes on the four Phébus tests employing an intact PWR fuel bundle. Because of the complexity and the extent of these tests, only the containment aspects were considered in the precedent works, i.e., only the thermal-hydraulics transient and its coupling with the FP and SM behavior. Then, general conclusions based on the outcomes of these precedent works are summarized in this work.

Description: For sequentially collected data, this paper introduces a lag-one differencing method to estimate the random error standard deviation and then uses the estimate 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: This study presents the time-dependent analyses of transmutations of long-lived fission products (LLFPs) and medium-lived fission products (MLFPs) occurring in thermal reactors in a conceptual helium gas-cooled accelerator-driven system (ADS). In accordance with this purpose, the CANDU-37 and PWR 15 × 15 spent fuels are separately considered. The ADS consists of LBE-spallation neutron target, subcritical fuel zone, and graphite reflector zone. While the considered ADS is fueled with the spent nuclear fuels extracted from each thermal reactor without the use of additional fuel, fission products extracted from same thermal reactor are also placed into transmutation zone in graphite reflector zone. The LLFP transmutation performance of the modified ADS is analyzed by considering three different spent fuels extracted from the thermal reactors. Spent fuels are extracted from CANDU-37 in case A, from PWR-15 × 15 in case B, and from CANDU-37 fueled with mixture of PWR 15 × 15 spent fuel and 46% ThO2 in case C. The LBE target is bombard with protons of 1000 MeV. The proton beam power is assumed as 20 MW, which corresponds to 1.24828·1017 protons per second. MCNPX 2.7 and CINDER 90 computer codes are used for the time-dependent burn calculations. The ADS is operated under subcritical mode until the value of keff increases to 0.984, and the maximum operation times are obtained as 3400, 3270, and 5040 days according to the spent fuel cases of A, B, and C, respectively. The calculations bring out that in the modified ADS, LLFPs and MLFPs, which are extracted from thermal reactors, can be transformed to stable isotopes in significant amounts along with energy production.

Description: Thermal reactors have been considered as interim solution for transmutation of minor actinides recycled from spent nuclear fuel. Various studies have been performed in recent decades to realize this possibility. This paper presents the neutronic feasibility study on transmutation of minor actinides as burnable poison in the VVER-1000 LEU (low enriched uranium) fuel assembly. The VVER-1000 LEU fuel assembly was modeled using the SRAC code system, and the SRAC calculation model was verified against the MCNP6 calculations and the available published benchmark data. Two models of minor actinide loading in the LEU fuel assembly have been investigated: homogeneous mixing in the UGD (Uranium-Gadolinium) pins and coating a thin layer to the UGD pins. The consequent negative reactivity insertion by minor actinides was compensated by reducing the gadolinium content and boron concentration. The reactivity of the LEU assembly versus burnup and the transmutation of minor actinide nuclides were examined in comparison with the reference case. The results demonstrate that transmutation of minor actinides as burnable poison in the VVER-1000 reactor is feasible as minor actinides could partially replace the functions of gadolinium and boric acid for excess reactivity control.

Description: As key equipment in nuclear power plant, the reactor power control system is adopted to strictly control and regulate the reactor power of a PWR (pressurized water reactor) in a nuclear power plant. A well-optimized predictive control algorithm based on SDMC (stepped dynamic matrix controller) is developed and introduced in this paper and applied to the power regulation of a reactor power model. In addition, the test and verification of this application is conducted by two different methods and devices: the virtual verification platform and the physical DCS (digital control system). The result of the verification suggests that the application of SDMC gains a better performance in the maximum dynamic deviation, adjustment time, overshoot, and so on.

Description: In this study, we first examined the sorption of Pd on MX-80 in Na-Ca-ClO4 solution as a function of (3–9) and ionic strength (0.1 M–4 M) and confirmed that the experimentally derived values could be fitted by a 2-site protolysis nonelectrostatic surface complexation and cation exchange (2SPNE SC/CE) model using three binary surface complexation constants previously estimated. Then, we investigated the sorption of Pd on MX-80 in Na-Ca-Cl-ClO4 solution as a function of (3–9) and molar concentration ratio (0–∞) at the ionic strength = 4 M. We found that the sorption of Pd on MX-80 in Na-Ca-Cl-ClO4 solution could be simulated only by the three binary and one ternary surface complexations (). This suggests that the contribution of other ternary surface complexations such as ≡S-OH ≡ ( = 1, 2 and 3) to Pd sorption in Na-Ca-Cl-ClO4 solution with ionic strength = 4 M was negligibly small.

Description: To ensure that the outside dose rate of waste package is below the limitation of national laws and regulations, based on the standard 200L drum, a new drum with inner shielding was proposed for intermediate-level radioactive waste (ILW) storage. For comparison, FLUKA and QAD-CGA were used to verify the shielding design of the ILW storage drums produced in INET with multiple inner shielding layers. The flux and dose were calculated and analyzed for four different cases. In QAD-CGA calculation, it was found that different buildup factors can lead to the considerably different results. A weighted algorithm was proposed to correct QAD-CGA for multilayer shielding cases. In FLUKA calculation, parameter optimization and tailored variance reduction technique (VRT) were used. Quantitative efficiency evaluation of different FLUKA settings using the FOM factor was carried out. The differences in the calculated dose rates results between the FLUKA and QAD-CGA programs are within one order of magnitude. The results of QAD-CGA are generally higher than those of FLUKA. The analysis shows that appropriate corrections in QAD-CGA can make the trend of the calculation results more consistent with the theory. In FLUKA calculation, with optimized setting and VRT adopted, the calculation efficiency can be improved more than 20 times. The results of this study provide not only suggestions for the design of the ILW storage drums but also useful references for other similar work.

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: 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: The Critical Heat Flux (CHF) prediction under high pressure condition, even close to the vicinity of the critical pressure of water, is an important issue. Although there are many empirical CHF correlations, most of them have covered the pressure under 15MPa. In this study, based on the CHF experiment database of upflow boiling in vertical round tube from 15MPa to the vicinity of the critical pressure of water, the Katto, Bowring, Hall-Mudawar, Alekseev correlations, and Groeneveld LUT-2006 are comparatively studied. With an error analysis of the predicted CHF to the experiment database, the prediction capability and the applicability of these correlations are evaluated and the parametric trends of CHF varying with pressure from 15MPa to critical pressure are proposed. Simultaneously, according to the characteristics of Departure from Nucleate Boiling (DNB) type CHF under high pressure condition, the constitutive correlations of Weisman & Pei model are proposed. The prediction results of three entrainment and deposition correlations of Kataoka, Celata, and Hewitt corresponding to the Dry-Out (DO) type CHF are analyzed. Based on the two improved models above, a comprehensive CHF mechanistic model under high pressure condition combining the DNB and DO type CHF is established. The verification based on the experiment database of upflow boiling in vertical round tube and the parametric trends analysis of CHF varying with thermal-hydraulic and geometric parameters are carried out. Findings of this study have a positive effect on further development of CHF prediction method for universal CHF mechanism, especially under high pressure region.

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: For the proposed novel procedure of immobilizing HLW with magnesium potassium phosphate cement (MKPC), Fe2O3 was added as a modifying agent to verify its effect on the solidification form and the immobilization of the radioactive nuclide. The results show that Fe2O3 is inert during the hydration reaction. It slows down the hydration reaction and lowers the heat release rate of the MKPC system, leading to a 3°C-5°C drop in the mixture temperature during hydration. Early comprehensive strength of Fe2O3 containing samples decreased slightly while the long-term strength remained unchanged. For the sintering process, Fe2O3 played a positive role, lowering the melting point and aiding the formation of ceramic structure. CsFe(PO4)2, or CsFePO4, was generated by sintering at 900°C. These products together with the ceramic structure and absorption benefit the immobilization of Cs+. The optimal sintering temperature for heat treatment is 900°C; it makes the solidification form a fired ceramic-like structure.

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: The management of spent nuclear fuel assemblies of nuclear reactors is a priority subject among member states of the International Atomic Energy Agency. For the majority of these countries, the destination of such fuel assemblies is a decision that is yet to be made and the “wait-and-see” policy is thus adopted by them. In this case, the irradiated fuel is stored in on-site spent fuel pools until the power plant is decommissioned or, when there is no more racking space in the pool, they are stored in intermediate storage facilities, which can be another pool or dry storage systems, until the final decision is made. The objective of this study is to propose a methodology that, using optimization algorithms, determines the ideal time for removal of the fuel assemblies from the spent fuel pool and to place them into dry casks for intermediate storage. In this scenario, the methodology allows for the optimal dimensioning of the designed spent fuel pools and the casks’ characteristics, thus reducing the final costs for purchasing new Nuclear Power Plants (NPP), as the size and safety features of the pool could be reduced and dry casks, that would be needed anyway after the decommissioning of the plant, could be purchased with optimal costs. To demonstrate the steps involved in the proposed methodology, an example is given, one which uses the Monte Carlo N-Particle code (MCNP) to calculate the shielding requirements for a simplified model of a concrete dry cask. From the given example, it is possible to see that, using real-life data, the proposed methodology can become a valuable tool to help making nuclear energy a more attractive choice costwise.

Description: Nordic Boiling Water Reactors (BWRs) employ ex-vessel debris coolability as a severe accident management strategy (SAM). Core melt is released into a deep pool of water where formation of noncoolable debris bed and ex-vessel steam explosion can pose credible threats to containment integrity. Success of the strategy depends on the scenario of melt release from the vessel that determines the melt-coolant interaction phenomena. The melt release conditions are determined by the in-vessel phase of severe accident progression. Specifically, properties of debris relocated into the lower plenum have influence on the vessel failure and melt release mode. In this work we use MELCOR code for prediction of the relocated debris. Over the years, many code modifications have been made to improve prediction of severe accident progression in light-water reactors. The main objective of this work is to evaluate the effect of models and best practices in different versions of MELCOR code on the in-vessel phase of different accident progression scenarios in Nordic BWR. The results of the analysis show that the MELCOR code versions 1.86 and 2.1 generate qualitatively similar results. Significant discrepancy in the timing of the core support failure and relocated debris mass in the MELCOR 2.2 compared to the MELCOR 1.86 and 2.1 has been found for a domain of scenarios with delayed time of depressurization. The discrepancies in the results can be explained by the changes in the modeling of degradation of the core components and changes in the Lipinski dryout model in MELCOR 2.2.

Description: The analysis of the thermal condition of spent FA (fuel assembly) of BN-350 reactor in a six-place cask for dry storage is presented. Simulation of the thermal condition of the cask is conducted with finite elements method using ANSYS software. Calculations of fuel temperature, fuel cladding, and assembly structural elements are the part of the safety analysis for storage of spent FA. In conclusion, the results of the thermal calculations in the cases of filling cask with argon and atmospheric air are given when the thickness of the insulation cask with concrete is 0.5 and 1 m. As a result of the calculated studies, the parameters of SNF (spent nuclear fuel) storage are determined, under which the fuel temperatures will have minimum and maximum values.

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: Two-phase flow instability may occur in nuclear reactor systems, which is often accompanied by periodic fluctuation in fluid flow rate. In this study, bubble rising and coalescence characteristics under inlet flow pulsation condition are analyzed based on the MPS-MAFL method. To begin with, the single bubble rising behavior under flow pulsation condition was simulated. The simulation results show that the bubble shape and rising velocity fluctuate periodically as same as the inlet flow rate. Additionally, the bubble pairs’ coalescence behavior under flow pulsation condition was simulated and compared with static condition results. It is found that the coalescence time of bubble pairs slightly increased under the pulsation condition, and then the bubbles will continue to pulsate with almost the same period as the inlet flow rate after coalescence. In view of these facts, this study could offer theory support and method basis to a better understanding of the two-phase flow configuration under flow pulsation condition.

Description: The Nuclear Material Accounting (NMA) system is one of the main safeguards measures to detect the existence of nuclear material diversion. It has become more important for large reprocessing facilities to apply Near Real Time Accountancy (NRTA) system based on NMA and statistical techniques to meet quantitative and timeliness goals. It is also important to quantitatively evaluate the performance of NMA system including NRTA from the standpoints of Safeguards and Security by Design (SSBD) prior to construction of nuclear-material-handling facilities. Such evaluation improves safeguards effectiveness and efficiency. Modeling and Simulation (M&S) work is a good way to evaluate performance for various NMA systems and to determine the optimal one among different options. For these purposes, in the present study, the PYroprocessing Material flow and MUF Uncertainty Simulation+ (PYMUS+) code, which uses evaluation algorithms to calculate many safeguards factors such as MUF uncertainty, detection probability, and others, was developed. According to a previous report, the PYMUS code, the predecessor of PYMUS+, can calculate MUF uncertainties only for a fixed model having 10 tHM/year, whereas the PYMUS+ code can additionally calculate detection probabilities according to diverse nuclear diversion scenarios as well as MUF uncertainties. The most important feature of the PYMUS+ code is its capability to evaluate many process and NMA system model options that a user wants to evaluate. Furthermore, a user can make a static process model having simplicity and a matching NMA model based on the PYMUS+ code regardless of facility throughput and is not even required to have professional programming knowledge. In the present work, some intercomparative studies were conducted to verify the M&S techniques applied in this code. It is expected that this code will be a useful tool for evaluation of NRTA system of pyroprocessing and other reprocessing facilities.

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: Interest in evaluation of severe accidents induced by extended station blackout (ESBO) has significantly increased after Fukushima. In this paper, the severe accident process under the high and low pressure induced by an ESBO for a small integrated pressurized water reactor (IPWR)-IP200 is simulated with the SCDAP/RELAP5 code. For both types of selected scenarios, the IP200 thermal hydraulic behavior and core meltdown are analyzed without operator actions. Core degradation studies firstly focus on the changes in the core water level and temperature. Then, the inhibition of natural circulation in the reactor pressure vessel (RPV) on core temperature rise is studied. In addition, the phenomena of core oxidation and hydrogen generation and the reaction mechanism of zirconium with the water and steam during core degradation are analyzed. The temperature distribution and time point of the core melting process are obtained. And the IP200 severe accident management guideline (SAMG) entry condition is determined. Finally, it is compared with other core degradation studies of large distributed reactors to discuss the influence of the inherent design characteristics of IP200. Furthermore, through the comparison of four sets of scenarios, the effects of the passive safety system (PSS) on the mitigation of severe accidents are evaluated. Detailed results show that, for the quantitative conclusions, the low coolant storage of IP200 makes the core degradation very fast. The duration from core oxidation to corium relocation in the lower-pressure scenario is 53% faster than that of in the high-pressure scenario. The maximum temperature of liquid corium in the lower-pressure scenario is 134 K higher than that of the high-pressure scenario. Besides, the core forms a molten pool 2.8 h earlier in the lower-pressure scenario. The hydrogen generated in the high-pressure scenario is higher when compared to the low-pressure scenario due to the slower degradation of the core. After the reactor reaches the SAMG entry conditions, the PSS input can effectively alleviate the accident and prevent the core from being damaged and melted. There is more time to alleviate the accident. This study is aimed at providing a reference to improve the existing IPWR SAMGs.

Description: In a fast spectrum reactor, the fuel rod bundle is mainly positioned radially by the wire which can make contact with the adjacent fuel rods, and then it is inevitable that flow-induced vibration (FIV) will cause fretting wear and vibration fatigue of the fuel cladding at the contact position. Therefore, the computational model of fretting wear and fatigue life about the fuel rod bundle caused by FIV will be studied in this paper. Based on the random vibration model of the fuel rod bundle, Hertz contact theory, and Archard wear theory, the fretting wear life computational model and the fatigue life computational model of the wire-to-adjacent fuel rod (WAFR) contact have been established. Finally, taking CEFR design parameters as an example, the fretting wear life and vibration fatigue life of the cladding are calculated, and it is found that fatigue affects the service life of the fuel rod more seriously than fretting wear. The calculation model and method lay a foundation for further accurate prediction and analysis of the fuel rod life.

Description: For the new nuclear power plants, the hazard of liquefaction due to earthquakes should be excluded by appropriate site selection or eliminated by engineering measures. An important question is how to define a quantitative criterion for negligibility of the liquefaction hazard. In the case of operating plants, liquefaction can be revealed as a beyond-design-basis event. It is important to learn whether the liquefaction hazard has a safety relevance and whether there is a sufficient margin to the onset of liquefaction. The use of pseudoprobabilistic method would be practicable for the definition of probability of liquefaction, but it could result in overconservative results. In this paper, the applicability of the pseudoprobabilistic procedure is demonstrated for the sites in diffuse seismicity environment and for low hazard levels that are typical for nuclear safety considerations. Use of the procedure is demonstrated in a case study with realistic site-plant parameters.

Description: The electrophoretic deposition (EPD) technique was used to create a uniform SiO2 thin film coating on boiling plates, 4 mm in width and 9 mm in length. Significant enhancement in critical heat flux (CHF), for the hydrophilic surfaces generated by this anodic EPD method, has been observed. In order to increase the coating strength, the plates were sintered at various temperatures. To find the thickness and uniformity of the coatings, the SEM images were captured. The captured images showed that the coating thickness uniformly increased up to 90 nm for 0.5% nanofluid percentage by the EPD method. The results show that the hydrophilic and super-hydrophilic surfaces have different boiling heat transfer (BHT) coefficients and CHF behaviors. Also, the results showed an increase of 160% in the CHF value by sintering compared to a bare surface. However, because of the setup simplicity, the shape independency, the particle-coating uniformity, and thickness controllability, the EPD technique can be an appropriate option for modification of the surface and coating on the nuclear fuel cladding.

Description: A new data-driven sampling-based framework was developed for uncertainty quantification (UQ) of the homogenized kinetic parameters calculated by lattice physics codes such as TRITON and Polaris. In this study, extension of the database for the delayed neutron data (DND) is performed by exploring more delayed neutron experiments and adding additional isotopes/actinides to the data libraries. Afterwards, the framework is utilized to obtain a deeper knowledge of the kinetic parameters’ sensitivity and uncertainty. The kinetic parameters include precursor-group-wise delayed neutron fraction (DNF) and decay constant. Input uncertainties include nuclear data (i.e., cross-sections) and DND (i.e., precursor group parameters and fractional delayed neutron yield). It is found that kinetic parameters, especially DNFs, have large uncertainties. The DNF uncertainty is driven by the cross-section uncertainties for LWR designs, while decay constant uncertainty is dominated by the DND uncertainties. The usage of correlated U-235 thermal DND in the UQ process significantly reduces the DND uncertainty contribution on the kinetic parameters. Large void fraction and presence of neutron absorber (e.g., control rod) increase the DNF uncertainty due to the hardening of neutron spectrum. High correlation between the DNF groups () is observed, while the decay constant groups () show weak correlation to each other and also to DNF groups. The DNF uncertainties of the dominant precursor group 4 for PWR, BWR, and VVER are about 7.5%, 9.4%, and 7.6%, respectively. The DNF uncertainty grows to larger values after fuel burnup. Kinetic parameters’ values and uncertainties provided here can be efficiently used in subsequent core calculations, point reactor kinetics, and other applications.

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: Data analyses of radioactive contamination of the RBMK-1500 reactor’s steam pipelines (SP) and components of high pressure rings (HPR) are presented in this paper. Also, modelled results of the SP-HPR system are compared to the results of other RBMK-1500 systems at Ignalina NPP Unit 1. Characteristics of SP-HPR components, thermal-hydraulic conditions of the coolant, and system operational regimes were evaluated employing the computer code LLWAA-DECOM (Tractebel Energy Engineering, Belgium). The presented results complement radiological characterization activities and facilitate the decommissioning process of nuclear facilities with RBMK type reactors. Analysis of the modelled results showed that the spread of radioactive contamination is very uneven between different components of the SP-HPR. The overall activity level of deposits of the SP-HPR is mostly determined by activated corrosion products and is lower than the activity level in the main circulation circuit (MCC) and nonpurified water subsystem activity of the purification and cooling system (PCS).

Description: The fuel safety and performance of high-temperature gas-cooled reactor (HTGR) are dependent on the integrity and geometric parameter of Tri-structural Isotropic (TRISO) coated particle. Micro X-ray computed tomography (CT) was used for nondestructive testing and three-dimensional measurement of the particle components which are composed of kernel, buffer layer, inner pyrolytic carbon layer (IPyC), silicon carbide (SiC) layer, and outer pyrolytic carbon (OPyC) layer. The thickness distribution and volume of kernel and coating layers are obtained by constructing 3D volume rendering of TRISO particle. Mean thickness of each layer is calculated for comparison with design value. A comparison between two-dimensional and three-dimensional measurement results is also made. It is found that the thickness distribution of all layers approximately obeys Gaussian distribution. Deviation of the thickness of kernel and coating layers between 3D measurement result and design value is 7.88%, -25.63%, -45.50%, 13.87%, and 14.73%, respectively. The deviation will affect the failure probability of TRISO particle. Obvious difference of the OPyC mean thickness between 3D measurement and 2D measurement is found, which proves that the proposed 3D measurement provides comprehensive information of the particle. However, 2D and 3D measured thickness of the kernel and IPyC layer tend to be similar.

Description: The heat and mass transfer processes of natural convective condensation with noncondensable gases are very important for the passive containment cooling system of water cooled reactors. Numerical simulation of natural convective condensation with noncondensable gases was realized in the Fluent software by adding condensation models. The scaled AP600 containment condensation experiment was simulated to verify the numerical method. It was shown that the developed method can predict natural convective condensation with noncondensable gases well. The velocity, species, and density fields in the scaled AP600 containment were presented. The heat transfer rate distribution and the influences of the mass fraction of air on heat transfer rate were also analyzed. It is found that the driving force of natural convective condensation with noncondensable gases is mainly caused by the mass fraction difference but not temperature difference. The natural convective condensation with noncondensable gases in AP1000 containment was then simulated. The temperature, species, velocity, and heat flux distributions were obtained and analyzed. The upper head of the containment contributes to 35.1% of the total heat transfer rate, while its area only takes 25.4% of the total condensation area of the containment. The influences of the mass fraction of low molecular weight noncondensable gas (hydrogen) on the natural convective condensation were also discussed based on the detailed species, density, and velocity fields. The results show that addition of hydrogen (production of zirconium-water reaction after severe accident) will weaken the intensity of natural convection and the heat and mass transfer processes significantly. When hydrogen contributes to 50% mole fraction of the noncondensable gases, the heat transfer coefficient will be reduced to 45%.

Description: After the September 11 attack, the resistant capability of containments against aircraft impacts is required to be assessed for newly constructed nuclear power plants (NPPs). In this paper, the crash of a commercial airplane Boeing 767-200ER on the reinforced concrete containment building of an NPP is analyzed using the missile-target interaction method. Two plane models with the same total weight but different fuel distribution are analyzed. The force-time history obtained by FEA (finite element analysis) is compared with the one calculated by the Riera function. In the integral analysis, the mesh sensitivity of the reinforced concrete containment model is studied, and recommendations are provided on the modelling of containment. The impact phenomenon and damage on the containment are investigated through the validated model. The fuel distribution in the aircraft is found to have strong influence on the damage of the containment, which indicates that the load distribution in the transverse direction is critical in the analysis of aircraft impact. The classic load-time function analysis is unable to incorporate this factor and may not be adequate to provide satisfactory results. For this reason, the application of an integral analysis is advantageous in the safety assessment of aircraft impact.

Description: A fracture criterion is newly proposed to evaluate fracture behavior and predict fracture initiation of metal materials in different complicated stress states for four different fracture mechanisms including quasicleavage fracture, normal fracture with void, shear fracture with void, and shear fracture without void. The dominant factors of these four different mechanisms are distinct, so it is impossible to capture all features of fracture initiation under different stress states with a single criterion, and different functions are necessary to predict fracture initiation of different mechanisms. In the new fracture criterion, different branches of the fracture criterion have been proposed corresponding to different fracture mechanisms. Quasicleavage fracture and normal fracture with void are described as a function of the principal stress, shear fracture with void is a function of the stress triaxiality and maximal shear stress, and shear fracture without void is only controlled by the maximal shear stress. The new fracture criterion is applied to predict the fracture initiation site and the fracture direction of nodular cast iron QT400-15 in combined tension-torsion tests. Predicted results are compared with experimental results to validate the performance of the new criterion in the intermediate stress triaxiality between 0 and 1/3. The new criterion is also applied to predict the crack initiation site and the direction of crack initiation of LY12 aluminium alloy and HY130 mild steel in mixed mode fracture tests to validate the performance of the new criterion in the high stress triaxiality. The new fracture criterion gives consistent results for these materials in a wide stress triaxiality range. It is shown that the new fracture criterion is a better supplement to the deficiency of fracture mechanics and also a better amendment to traditional strength theory in complicated stress states. Therefore, the new fracture criterion is recommended to be utilized to evaluate the fracture initiation of metal structures in nuclear waste storage and other engineering applications.

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: 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: 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: Spent fuel pools are used as temporary storage for spent fuel assemblies in nuclear power plants and are filled with coolant which removes the decaying heat from spent fuel assemblies. Sloshing of the coolant can occur if an earthquake occurs in the area. It may produce additional forces on the pool or inner structure and cause overflow of the coolant. It is therefore critical to investigate the phenomenon of sloshing in a seismic assessment of the spent fuel pool. The size of an actual spent fuel pool is excessive for carrying out an experimental study; thus, a scale model is necessary for experimentation. In this study, a scaling law was defined for test conditions using a scale model to understand sloshing behavior, and the results were validated via computational fluid dynamic analysis. Because sloshing is resonant in a fluid and the first mode natural frequency of a fluid is dominant in sloshing behavior, the test condition could be obtained based on the natural frequency of the fluid. In the model, which is scaled with a factor of “,” the scale factors “,” “,” “,” and “” were used for displacement, acceleration, excitation frequency, and excitation time, respectively. Approximately 5% difference in maximum sloshing height between two models was predicted in the only case that 1/8 and 1/4 models (1/8 and 1/4 scaled down from an actual spent fuel pool) were excited with 10 Hz and 7.071 Hz, respectively, but the same sloshing height and pressure were predicted in other cases. The results of this study support the idea that the Froude scaling law can be used when using a scale model for a seismic assessment of spent fuel pools to investigate sloshing behavior.

Description: Accidental release of gaseous or liquid effluents is a critical issue and of a greater concern to the nuclear industry when it comes to the protection of the public and the environment. The emphasis becomes paramount when the release involves particulate of radiation particles. This paper provides a comprehensive insight report on an account of a research investigation carried out in addressing a radiological safety issue of Ghana’s Miniature Neutron Source Reactor (MNSR) during its core conversion project. The amounts of Strontium-90 (Sr-90) and Krypton-85 (Kr-85) effluents presumably released from the reactor hall to the surroundings and the consequential emission radiation to the working area within a 200 m radius were analyzed for a six-month working period. The objective was to estimate specifically the approximate total effective dose equivalent (TEDE) of Sr-90 and Kr-85 by considering a conjectural accident scenario using a well-recognized and user-friendly known atmospheric dispersion model before the preparatory period. The maximum TEDE value recorded at a ground deposition value of 4.6E − 01 kBq/m2 was approximately 1.80E − 02 mSv and 4.90E − 4 mSv for Sr-90 and Kr-85, respectively, at a maximum distance of 0.1 km from the source. The estimated dose values recorded were found to be within the recommended regulatory safety limits of 50 mSv for onsite workers and 1 mSv for the general public. No adverse effect was experienced with respect to human health and the environment.

Description: COMSOL Multiphysics has been used to conduct thermal-hydraulic analysis in multiple nuclear applications. The aim of this study is to benchmark the prediction accuracy of COMSOL Multiphysics in performing thermal-hydraulic analysis of TRIGA (Training, Research, Isotopes, General Atomics) reactors such as the Geological Survey TRIGA Reactor (GSTR) by comparing its predictions with RELAP5 (a widely used code in nuclear thermal-hydraulic analysis) results and experimental data. The GSTR type is Mark I with a full thermal power of 1 MW, and it resides at the Denver Federal Center (DFC) in Colorado. The numerical investigation of the present work is carried out by developing single-subchannel thermal-hydraulic models of the GSTR utilizing RELAP5 and COMSOL codes. The models estimate the temperatures (fuel, outer clad, and coolant) and water flow patterns in the core as well as fuel element powers at which void starts to form within the coolant subchannels. Then, these models’ predictions are quantitatively evaluated and compared with the measured data.

Description: A practical scale mechanical decladder that can slit spent nuclear fuel rod-cuts (hulls + pellets) of several tens of kg HM/batch is being developed to supply UO2 pellets to a voloxidation process. The mechanical decladder is an apparatus for separating and recovering fuel material and cladding tubes by horizontally slitting the cladding tube of a fuel rod and a defective irradiated fuel rod. In this study, we address the engineering design of the mechanical decladder for the pretesting of rod-cut slitting. To obtain the requirements of the mechanical decladder, we first manufactured a slitter for testing based on the decladding and shearing conditions of hulls and pellets. The performance test of the testing device for decladding was carried out using a 2-CUT blade module and a 3-CUT blade module. We evaluated the decladding methods for the mechanical decladder and selected the 3-CUT blade module based on the results. A buckling measurement instrument was used to perform a buckling verification test according to the length of a rod-cut and to determine decladder dimensions. The optimum decladding rod-cut length for buckling prevention was calculated. Furthermore, we analyzed the decladding mechanism for various slitting methods. Design/fabrication and preliminary tests of the practical scale mechanical decladder were also performed. For this purpose, we constructed the main mechanism by utilizing the SolidWorks modeling and analysis program and fabricated a new mechanical decladder. Based on the derived requirements, a mechanical decladder with three main modules was designed and fabricated for testing. Simulated rod-cuts of zircaloy were also manufactured to test the basic performance of the decladder, and a data acquisition system was constructed using RSC 232 to measure decladding force and velocity. In the basic test, the rod-cut was completely sectioned into three evenly spaced locations by the new mechanical decladder.

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: This study presents the time-dependent analyses of transmutations of long-lived fission products (LLFPs) and medium-lived fission products (MLFPs) occurring in thermal reactors in a conceptual helium gas-cooled accelerator-driven system (ADS). In accordance with this purpose, the CANDU-37 and PWR 15 × 15 spent fuels are separately considered. The ADS consists of LBE-spallation neutron target, subcritical fuel zone, and graphite reflector zone. While the considered ADS is fueled with the spent nuclear fuels extracted from each thermal reactor without the use of additional fuel, fission products extracted from same thermal reactor are also placed into transmutation zone in graphite reflector zone. The LLFP transmutation performance of the modified ADS is analyzed by considering three different spent fuels extracted from the thermal reactors. Spent fuels are extracted from CANDU-37 in case A, from PWR-15 × 15 in case B, and from CANDU-37 fueled with mixture of PWR 15 × 15 spent fuel and 46% ThO2 in case C. The LBE target is bombard with protons of 1000 MeV. The proton beam power is assumed as 20 MW, which corresponds to 1.24828·1017 protons per second. MCNPX 2.7 and CINDER 90 computer codes are used for the time-dependent burn calculations. The ADS is operated under subcritical mode until the value of keff increases to 0.984, and the maximum operation times are obtained as 3400, 3270, and 5040 days according to the spent fuel cases of A, B, and C, respectively. The calculations bring out that in the modified ADS, LLFPs and MLFPs, which are extracted from thermal reactors, can be transformed to stable isotopes in significant amounts along with energy production.

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 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: For the new nuclear power plants, the hazard of liquefaction due to earthquakes should be excluded by appropriate site selection or eliminated by engineering measures. An important question is how to define a quantitative criterion for negligibility of the liquefaction hazard. In the case of operating plants, liquefaction can be revealed as a beyond-design-basis event. It is important to learn whether the liquefaction hazard has a safety relevance and whether there is a sufficient margin to the onset of liquefaction. The use of pseudoprobabilistic method would be practicable for the definition of probability of liquefaction, but it could result in overconservative results. In this paper, the applicability of the pseudoprobabilistic procedure is demonstrated for the sites in diffuse seismicity environment and for low hazard levels that are typical for nuclear safety considerations. Use of the procedure is demonstrated in a case study with realistic site-plant parameters.

Description: The Critical Heat Flux (CHF) prediction under high pressure condition, even close to the vicinity of the critical pressure of water, is an important issue. Although there are many empirical CHF correlations, most of them have covered the pressure under 15MPa. In this study, based on the CHF experiment database of upflow boiling in vertical round tube from 15MPa to the vicinity of the critical pressure of water, the Katto, Bowring, Hall-Mudawar, Alekseev correlations, and Groeneveld LUT-2006 are comparatively studied. With an error analysis of the predicted CHF to the experiment database, the prediction capability and the applicability of these correlations are evaluated and the parametric trends of CHF varying with pressure from 15MPa to critical pressure are proposed. Simultaneously, according to the characteristics of Departure from Nucleate Boiling (DNB) type CHF under high pressure condition, the constitutive correlations of Weisman & Pei model are proposed. The prediction results of three entrainment and deposition correlations of Kataoka, Celata, and Hewitt corresponding to the Dry-Out (DO) type CHF are analyzed. Based on the two improved models above, a comprehensive CHF mechanistic model under high pressure condition combining the DNB and DO type CHF is established. The verification based on the experiment database of upflow boiling in vertical round tube and the parametric trends analysis of CHF varying with thermal-hydraulic and geometric parameters are carried out. Findings of this study have a positive effect on further development of CHF prediction method for universal CHF mechanism, especially under high pressure region.

Description: After the September 11 attack, the resistant capability of containments against aircraft impacts is required to be assessed for newly constructed nuclear power plants (NPPs). In this paper, the crash of a commercial airplane Boeing 767-200ER on the reinforced concrete containment building of an NPP is analyzed using the missile-target interaction method. Two plane models with the same total weight but different fuel distribution are analyzed. The force-time history obtained by FEA (finite element analysis) is compared with the one calculated by the Riera function. In the integral analysis, the mesh sensitivity of the reinforced concrete containment model is studied, and recommendations are provided on the modelling of containment. The impact phenomenon and damage on the containment are investigated through the validated model. The fuel distribution in the aircraft is found to have strong influence on the damage of the containment, which indicates that the load distribution in the transverse direction is critical in the analysis of aircraft impact. The classic load-time function analysis is unable to incorporate this factor and may not be adequate to provide satisfactory results. For this reason, the application of an integral analysis is advantageous in the safety assessment of aircraft impact.

Description: A new data-driven sampling-based framework was developed for uncertainty quantification (UQ) of the homogenized kinetic parameters calculated by lattice physics codes such as TRITON and Polaris. In this study, extension of the database for the delayed neutron data (DND) is performed by exploring more delayed neutron experiments and adding additional isotopes/actinides to the data libraries. Afterwards, the framework is utilized to obtain a deeper knowledge of the kinetic parameters’ sensitivity and uncertainty. The kinetic parameters include precursor-group-wise delayed neutron fraction (DNF) and decay constant. Input uncertainties include nuclear data (i.e., cross-sections) and DND (i.e., precursor group parameters and fractional delayed neutron yield). It is found that kinetic parameters, especially DNFs, have large uncertainties. The DNF uncertainty is driven by the cross-section uncertainties for LWR designs, while decay constant uncertainty is dominated by the DND uncertainties. The usage of correlated U-235 thermal DND in the UQ process significantly reduces the DND uncertainty contribution on the kinetic parameters. Large void fraction and presence of neutron absorber (e.g., control rod) increase the DNF uncertainty due to the hardening of neutron spectrum. High correlation between the DNF groups () is observed, while the decay constant groups () show weak correlation to each other and also to DNF groups. The DNF uncertainties of the dominant precursor group 4 for PWR, BWR, and VVER are about 7.5%, 9.4%, and 7.6%, respectively. The DNF uncertainty grows to larger values after fuel burnup. Kinetic parameters’ values and uncertainties provided here can be efficiently used in subsequent core calculations, point reactor kinetics, and other applications.

Description: In this study, we first examined the sorption of Pd on MX-80 in Na-Ca-ClO4 solution as a function of (3–9) and ionic strength (0.1 M–4 M) and confirmed that the experimentally derived values could be fitted by a 2-site protolysis nonelectrostatic surface complexation and cation exchange (2SPNE SC/CE) model using three binary surface complexation constants previously estimated. Then, we investigated the sorption of Pd on MX-80 in Na-Ca-Cl-ClO4 solution as a function of (3–9) and molar concentration ratio (0–∞) at the ionic strength = 4 M. We found that the sorption of Pd on MX-80 in Na-Ca-Cl-ClO4 solution could be simulated only by the three binary and one ternary surface complexations (). This suggests that the contribution of other ternary surface complexations such as ≡S-OH ≡ ( = 1, 2 and 3) to Pd sorption in Na-Ca-Cl-ClO4 solution with ionic strength = 4 M was negligibly small.

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: 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: A modified multiobjective self-adaptive differential evolution algorithm (MMOSADE) is presented in this paper to improve the accuracy of multiobjective optimization design in the nuclear power system. The performance of the MMOSADE is tested by the ZDT test function set and compared with classical evolutionary algorithms. The results indicate that MMOSADE has a better performance in convergence and diversity. Based on the MMOSADE, a multiobjective optimization design platform for the nuclear power system is proposed, and the application of which is carried out. The evaluation program of the PRHR-HX in AP1000 is developed, and its reliability is verified. The optimal design schemes of PHHR-HX are obtained by utilizing the multiobjective optimization design platform. The results show that the optimal design schemes can envelop the prototype design scheme. This conclusion proves that the optimization design platform proposed in this paper is effective and feasible.