ALBERT

Beschreibung: This paper reports two numerical simulation methods for modeling displacement instabilities around a surface groove in a metal substrate used in nuclear power plant. The amplitude change in the groove, the downward displacement at the base node, and the groove displacement at the periphery were simulated using ABAQUS to compare the results from two methods, as well as the tangential stress in the elements at the groove base and periphery. The comparison showed that for the tangential stress two methods were in close agreement for all thermal cycles. For the amplitude change, the downward displacement, the groove displacement, and the stress distribution, the two methods were in close agreement for the first 3 to 6 thermal cycles. After that, inconsistency increased with the number of thermal cycles. It is interesting that the thermal cycle at which the discrepancy between the two methods began to occur corresponded to a thermally grown oxide (TGO) thickness of 1 μm, which showed the accuracy of the present work over the classic method. It is concluded that the present work’s numerical simulation scheme worked better with a thinner TGO layer than the classic method and could overcome the limitation of TGO thickness by simulating any thickness.

Beschreibung: Instrumental neutron activation analysis (INAA) is a powerful technique for trace element determination in rocks. Nine alabaster samples were collected from Wadi El-Nakhil located at the intersection of lat. 26°10′50′′N and long. 34°03′40′′E, central Eastern Desert, Egypt, for investigation by INAA and Energy Depressive X-Ray Fluorescence (EDXRF). The samples were irradiated by thermal neutrons at the TRIGA Mainz research reactor at a neutron flux of 7 × 1011 n/cm2·s. Twenty-two elements were determined, namely, As, Ba, Ca, Co, Cr, Sc, Fe, Hf, K, Mg, Mn, Na, Rb, U, Zn, Zr, Lu, Ce, Sm, La, Yb, and Eu. The chemical analysis of alabaster indicated having high contents of CaO and MgO and LOI and low contents of SiO2, Al2O3, Na2O, K2O, MnO, and Fe2O3.

Beschreibung: This paper reports the results of sensitivity analysis of the multidimension, multigroup neutron diffusion NODAL3 code for the NEACRP 3D LWR core transient benchmarks (PWR). The code input parameters covered in the sensitivity analysis are the radial and axial node sizes (the number of radial node per fuel assembly and the number of axial layers), heat conduction node size in the fuel pellet and cladding, and the maximum time step. The output parameters considered in this analysis followed the above-mentioned core transient benchmarks, that is, power peak, time of power peak, power, averaged Doppler temperature, maximum fuel centerline temperature, and coolant outlet temperature at the end of simulation (5 s). The sensitivity analysis results showed that the radial node size and maximum time step give a significant effect on the transient parameters, especially the time of power peak, for the HZP and HFP conditions. The number of ring divisions for fuel pellet and cladding gives negligible effect on the transient solutions. For productive work of the PWR transient analysis, based on the present sensitivity analysis results, we recommend NODAL3 users to use radial nodes per assembly, axial layers per assembly, the maximum time step of 10 ms, and 9 and 1 ring divisions for fuel pellet and cladding, respectively.

Beschreibung: The double-ended guillotine break (DEGB) of the horizontal coaxial gas duct accident is a serious air ingress accident of the high temperature gas-cooled reactor pebble-bed module (HTR-PM). Because the graphite is widely used as the structure material and the fuel element matrix of HTR-PM, the oxidation analyses of this severe air ingress accident have got enough attention in the safety analyses of the HTR-PM. The DEGB of the horizontal coaxial gas duct accident is calculated by using the TINTE code in this paper. The results show that the maximum local oxidation of the matrix graphite of spherical fuel elements in the core will firstly reach  mol/m3 at about 120 h, which means that only the outer 5 mm fuel-free zone of matrix graphite will be oxidized out. Even at 150 h, the maximum local weight loss ratio of the nuclear grade graphite in the bottom reflectors is only 0.26. Besides, there is enough time to carry out some countermeasures to stop the air ingress during several days. Therefore, the nuclear grade graphite of the bottom reflectors will not be fractured in the DEGB of the horizontal coaxial gas duct accident and the integrity of the HTR-PM can be guaranteed.

Beschreibung: Gamma-ray measurements in various research fields require efficient detectors. One of these research fields is mass attenuation coefficients of different materials. Apart from experimental studies, the Monte Carlo (MC) method has become one of the most popular tools in detector studies. An NaI(Tl) detector has been modeled, and, for a validation study of the modeled NaI(Tl) detector, the absolute efficiency of 3 × 3 inch cylindrical NaI(Tl) detector has been calculated by using the general purpose Monte Carlo code MCNP-X (version 2.4.0) and compared with previous studies in literature in the range of 661–2620 keV. In the present work, the applicability of MCNP-X Monte Carlo code for mass attenuation of concrete sample material as building material at photon energies 59.5 keV, 80 keV, 356 keV, 661.6 keV, 1173.2 keV, and 1332.5 keV has been tested by using validated NaI(Tl) detector. The mass attenuation coefficients of concrete sample have been calculated. The calculated results agreed well with experimental and some other theoretical results. The results specify that this process can be followed to determine the data on the attenuation of gamma-rays with other required energies in other materials or in new complex materials. It can be concluded that data from Monte Carlo is a strong tool not only for efficiency studies but also for mass attenuation coefficients calculations.

Beschreibung: Considering dangerous environmental conditions, maintenance of radioactive equipment can be performed by remote handling maintenance (RHM) system. The RHM system is a sophisticated man-machine system. Therefore, human factors analysis is an inevitable aspect considered in guaranteeing successful and safe task performance. This study proposes an approach for integrated analysis of human factors in RHM so as to make the evaluating process more practical. In the approach, indicators of accessibility, health safety, and fatigue are analyzed using virtual human simulation technologies. The human error factors in the maintenance process are analyzed using the human error probability (HEP) based on the success likelihood index method- (SLIM-) analytic hierarchy process (AHP). The psychological factors level of maintenance personnel is determined with an expert scoring. The human factors for the entire RHM system are then evaluated using the interval method. An application example is present, and the application results show that the approach can support the evaluation of the human factors in RHM.

Beschreibung: Based on the special application of 90-degree elbow pipe in the HTR-PM, the large eddy simulation was selected to calculate the instantaneous flow field in the 90-degree elbow pipe combining with the experimental results. The characteristics of the instantaneous turbulent flow field under the influence of flow separation and secondary flow were studied by analyzing the instantaneous pressure information at specific monitoring points and the instantaneous velocity field on the cross section of the elbow. The pattern and the intensity of the Dean vortex and the small scale eddies change over time and induce the asymmetry of the flow field. The turbulent disturbance upstream and the flow separation near the intrados couple with the vortexes of various scales. Energy is transferred from large scale eddies to small scale eddies and dissipated by the viscous stress in the end.

Beschreibung: Korea Atomic Energy Research Institute (KAERI) started the experimental research on moderator circulation as one of a the national research and development programs from 2012. This research program includes the construction of the moderator circulation test (MCT) facility, production of the validation data for self-reliant computational fluid dynamics (CFD) tools, and development of optical measurement system using the particle image velocimetry (PIV). In the present paper we introduce the scaling analysis performed to extend the scaling criteria suitable for reproducing thermal-hydraulic phenomena in a scaled-down CANDU- (CANada Deuterium Uranium-) 6 moderator tank, a manufacturing status of the 1/4 scale moderator tank. Also, preliminary CFD analysis results for the full-size and scaled-down moderator tanks are carried out to check whether the moderator flow and temperature patterns of both the full-size reactor and scaled-down facility are identical.

Beschreibung: Sensors health monitoring is essentially important for reliable functioning of safety-critical chemical and nuclear power plants. Autoassociative neural network (AANN) based empirical sensor models have widely been reported for sensor calibration monitoring. However, such ill-posed data driven models may result in poor generalization and robustness. To address above-mentioned issues, several regularization heuristics such as training with jitter, weight decay, and cross-validation are suggested in literature. Apart from these regularization heuristics, traditional error gradient based supervised learning algorithms for multilayered AANN models are highly susceptible of being trapped in local optimum. In order to address poor regularization and robust learning issues, here, we propose a denoised autoassociative sensor model (DAASM) based on deep learning framework. Proposed DAASM model comprises multiple hidden layers which are pretrained greedily in an unsupervised fashion under denoising autoencoder architecture. In order to improve robustness, dropout heuristic and domain specific data corruption processes are exercised during unsupervised pretraining phase. The proposed sensor model is trained and tested on sensor data from a PWR type nuclear power plant. Accuracy, autosensitivity, spillover, and sequential probability ratio test (SPRT) based fault detectability metrics are used for performance assessment and comparison with extensively reported five-layer AANN model by Kramer.

Beschreibung: A combined cycle that combines AWM cycle with a nuclear closed Brayton cycle is proposed to recover the waste heat rejected from the precooler of a nuclear closed Brayton cycle in this paper. The detailed thermodynamic and economic analyses are carried out for the combined cycle. The effects of several important parameters, such as the absorber pressure, the turbine inlet pressure, the turbine inlet temperature, the ammonia mass fraction, and the ambient temperature, are investigated. The combined cycle performance is also optimized based on a multiobjective function. Compared with the closed Brayton cycle, the optimized power output and overall efficiency of the combined cycle are higher by 2.41% and 2.43%, respectively. The optimized LEC of the combined cycle is 0.73% lower than that of the closed Brayton cycle.

Beschreibung: In sodium-cooled fast reactors (SFR), grid plate is a critical component which is made of 316 L(N) SS. It is supported on core support structure. The grid plate supports the core subassemblies and maintains their verticality. Most of the components of SFR are made of 316 L(N)/304 L(N) SS and they are in contact with the liquid-metal sodium which acts as a coolant. The peak operating temperature in SFR is 550°C. However, the self-welding starts at 500°C. To avoid self-welding and galling, hardfacing of the grid plate has become necessary. Nickel based cobalt-free colmonoy 5 has been identified as the hardfacing material due to its lower dose rate by Plasma Transferred Arc Welding (PTAW). This paper is concerned with the measurement and investigations of the effects of the residual stress generated due to thermal cycling on a scale-down physical model of the grid plate. Finite element analysis of the hardfaced grid plate model is performed for obtaining residual stresses using elastoplastic analysis and hence the results are validated. The effects of the residual stresses due to thermal cycling on the hardfaced grid plate model are studied.

Beschreibung: Spent fuel rack is the key equipment for the storage of spent fuel after refueling. In order to investigate the performance of the spent fuel rack under the earthquake, the phenomena including sliding, collision, and overturning of the spent fuel rack were studied. An FEM model of spent fuel rack is built to simulate the transient response under seismic loading regarding fluid-structure interaction by ANSYS. Based on D’Alambert’s principle, the equilibriums of force and momentum were established to obtain the critical sliding and overturning accelerations. Then 5 characteristic transient loadings which were designed based on the critical sliding and overturning accelerations were applied to the rack FEM model. Finally, the transient displacement and impact force response of rack with different gap sizes and the supporting leg friction coefficients were analyzed. The result proves the FEM model is applicable for seismic response of spent fuel rack. This paper can guide the design of the future’s fluid-structure interaction experiment for spent fuel rack.

Beschreibung: In water-cooled reactor, the dominant radioactive source term under normal operation is activated corrosion products (ACPs), which have an important impact on reactor inspection and maintenance. A three-node transport model of ACPs was introduced into the new version of ACPs source term code CATE in this paper, which makes CATE capable of theoretically simulating the variation and the distribution of ACPs in a water-cooled reactor and suitable for more operating conditions. For code testing, MIT PWR coolant chemistry loop was simulated, and the calculation results from CATE are close to the experimental results from MIT, which means CATE is available and credible on ACPs analysis of water-cooled reactor. Then ACPs in the blanket cooling loop of water-cooled fusion reactor ITER under construction were analyzed using CATE and the results showed that the major contributors are the short-life nuclides, especially Mn-56. At last a point kernel integration code ARShield was coupled with CATE, and the dose rate around ITER blanket cooling loop was calculated. Results showed that after shutting down the reactor only for 8 days, the dose rate decreased nearly one order of magnitude, which was caused by the rapid decay of the short-life ACPs.

Beschreibung: Falling water film on an inclined plane is studied by shadowgraphy. The ranges of inclination angle and the film Reynolds number are, respectively, up to 21° and 60. Water is used as working fluid. The scenario of wave regime evolution is identified as three distinctive regimes, namely, initial quiescent smooth film flow, two-dimensional regular solitary wave pattern riding on film flow, and three-dimensional irregular wave pattern. Three characteristic parameters of two-dimensional solitary wave pattern, namely, inception length, primary pulse spacing, and propagation velocity, are examined, which are significant in engineering applications for estimation of heat and mass transfer on film flow. The present experimental data are well in agreement with the Koizumi correlations, the deviation from which is limited to 20% and 15%, respectively, for primary pulse spacing and propagation velocity. Through the scrutiny of the present experimental observation, it is concluded that wave evolution on film flow at the moderate Reynolds number is controlled by gravity and drag and the Rayleigh-Taylor instability that occurred on the steep front of primary pulse triggers the disintegration of continuous two-dimensional regular solitary wave pattern into three-dimensional irregular wave pattern.

Beschreibung: There is a large water storage tank installed at the top of containment of AP1000, which can supply the passive cooling. In the extreme condition, sloshing of the free surface in the tank may impact on the roof under long-period earthquake. For the safety assessment of structure, it is necessary to calculate the impact pressure caused by water sloshing. Since the behavior of sloshing impacted on the roof is involved into a strong nonlinear phenomenon, it is a little difficult to calculate such pressure by theoretical or numerical method currently. But it is applicable to calculate the height of sloshing in a tank without roof. In the present paper, a simplified method was proposed to calculate the impact pressure using the sloshing wave height, in which we first marked the position of the height of roof, then produced sloshing in the tank without roof and recorded the maximum wave height, and finally regarded approximately the difference between maximum wave height and roof height as the impact pressure head. We also designed an experiment to verify this method. The experimental result showed that this method overpredicted the impact pressure with a certain error of no more than 35%. By the experiment, we conclude that this method is conservative and applicable for the engineering design.

Beschreibung: Long-term immersion was adopted to explore the damage deterioration and mechanical properties of granite under different chemical solutions. Here, granite was selected as the candidate of parent rocks for nuclear waste storage. The physical and mechanical properties of variation regularity immersed in various chemical solutions were analyzed. Meanwhile, the damage variable based on the variation in porosity was used in the quantitative analysis of chemical damage deterioration degree. Experimental results show that granite has a significant weakening tendency after chemical corrosion. The fracture toughness , splitting tensile strength, and compressive strength all demonstrate the same deteriorating trend with chemical corrosion time. However, a difference exists in the deterioration degree of the mechanical parameters; that is, the deterioration degree of fracture toughness is the greatest followed by those of splitting tensile strength and compressive strength, which are relatively smaller. Strong acid solutions may aggravate chemical damage deterioration in granite. By contrast, strong alkaline solutions have a certain inhibiting effect on chemical damage deterioration. The chemical solutions that feature various compositions may have different effects on chemical damage degree; that is, ions have a greater effect on the chemical damage in granite than ions.

Beschreibung: The acceptable accuracy for simulation of severe accident scenarios in containments of nuclear power plants is required to investigate the consequences of severe accidents and effectiveness of potential counter measures. For this purpose, the actual capability of CFX tool and COCOSYS code is assessed in prototypical geometries for simplified physical process-plume (due to a heat source) under adiabatic and convection boundary condition, respectively. Results of the comparison under adiabatic boundary condition show that good agreement is obtained among the analytical solution, COCOSYS prediction, and CFX prediction for zone temperature. The general trend of the temperature distribution along the vertical direction predicted by COCOSYS agrees with the CFX prediction except in dome, and this phenomenon is predicted well by CFX and failed to be reproduced by COCOSYS. Both COCOSYS and CFX indicate that there is no temperature stratification inside dome. CFX prediction shows that temperature stratification area occurs beneath the dome and away from the heat source. Temperature stratification area under adiabatic boundary condition is bigger than that under convection boundary condition. The results indicate that the average temperature inside containment predicted with COCOSYS model is overestimated under adiabatic boundary condition, while it is underestimated under convection boundary condition compared to CFX prediction.

Beschreibung: Two tests related to a new safety system for a pressurized water reactor were performed with the ROSA/LSTF (rig of safety assessment/large scale test facility). The tests simulated cold leg small-break loss-of-coolant accidents with 2-inch diameter break using an early steam generator (SG) secondary-side depressurization with or without release of nitrogen gas dissolved in accumulator (ACC) water. The SG depressurization was initiated by fully opening the depressurization valves in both SGs immediately after a safety injection signal. The pressure difference between the primary and SG secondary sides after the actuation of ACC system was larger in the test with the dissolved gas release than that in the test without the dissolved gas release. No core uncovery and heatup took place because of the ACC coolant injection and two-phase natural circulation. Long-term core cooling was ensured by the actuation of low-pressure injection system. The RELAP5 code predicted most of the overall trends of the major thermal-hydraulic responses after adjusting a break discharge coefficient for two-phase discharge flow under the assumption of releasing all the dissolved gas at the vessel upper plenum.

Beschreibung: A neural network-direct inverse control (NN-DIC) has been simulated to automatically control the power level of nuclear reactors. This method has been tested on an Indonesian pool type multipurpose reactor, namely, Reaktor Serba Guna-GA Siwabessy (RSG-GAS). The result confirmed that this method still cannot minimize errors and shorten the learning process time. A new method is therefore needed which will improve the performance of the DIC. The objective of this study is to develop a particle swarm optimization-based direct inverse control (PSO-DIC) to overcome the weaknesses of the NN-DIC. In the proposed PSO-DIC, the PSO algorithm is integrated into the DIC technique to train the weights of the DIC controller. This integration is able to accelerate the learning process. To improve the performance of the system identification, a backpropagation (BP) algorithm is introduced into the PSO algorithm. To show the feasibility and effectiveness of this proposed PSO-DIC technique, a case study on power level control of RSG-GAS is performed. The simulation results confirm that the PSO-DIC has better performance than NN-DIC. The new developed PSO-DIC has smaller steady-state error and less overshoot and oscillation.

Beschreibung: The effect of BWR fuel assembly 3D model on void reactivity coefficient (VRC) estimation is investigated. VRC values were calculated for different BWR assembly models applying deterministic T-NEWT and Monte Carlo KENO-VI functional modules of SCALE 6.1 code package. The difference between deterministic T-NEWT and Monte Carlo KENO-VI simulations is negligible (0.18 pcm/%). The influence of the assumed more detailed coolant density profile was estimated as well. VRC increases with the application of a larger number of coolant density values across fuel assembly height. It was shown that the coolant density profile described by 6 values per height could be considered sufficient from prospect of VRC estimation, as a more detailed density profile has impact below 1% on total assembly void effects. VRC values were decomposed to values for individual nodes and isotopes, since decomposition provides useful insights to describe the overall behaviour of VRC in detail.

Beschreibung: The impact of thermal agitation on Doppler coefficient for Gd-bearing fuel was analyzed. It was found through the analysis that the impact increases when a small amount of Gd2O3 is added to pure UO2 fuel although the impact decreases for a large amount of Gd2O3. This tendency was discussed with the usage of simplified expression for the difference of Doppler coefficient. The simplified expression was used to consider the tendency, and it was revealed that the tendency mainly comes from the rapid decrement of multiplication factor and the relatively slow decrement of the magnitude of sensitivity coefficient of U-238 capture cross section at low Gd2O3 concentration. Similar tendency which shows a maximum impact on Doppler coefficient at interior concentration is expected for other UO2 fuel with a slight content of strong absorber. This indicates that Doppler coefficient of UO2 fuel system with low content of strong absorber should be analyzed carefully by considering thermal agitation in epithermal range.

Beschreibung: In nuclear power plant construction scheduling, a project is generally defined by its dependent preparation time, the time required for construction, and its reactor installation time. The issues of multiple construction teams and multiple reactor installation teams are considered. In this paper, a hierarchical particle swarm optimization algorithm is proposed to solve the nuclear power plant construction scheduling problem and minimize the occurrence of projects failing to achieve deliverables within applicable due times and deadlines.

Beschreibung: Pellet-clad interaction (PCI) is one of the major issues in fuel rod design and reactor core operation in water cooled reactors. The prediction of fuel rod failure by PCI is studied in this paper by the method of radial basis function neural network (RBFNN). The neural network is built through the analysis of the existing experimental data. It is concluded that it is a suitable way to reduce the calculation complexity. A self-organized RBFNN is used in our study, which can vary its structure dynamically in order to maintain the prediction accuracy. For the purpose of the appropriate network complexity and overall computational efficiency, the hidden neurons in the RBFNN can be changed online based on the neuron activity and mutual information. The presented method is tested by the experimental data from the reference, and the results demonstrate its effectiveness.

Beschreibung: Particle-induced X-ray emission is used for determining the elemental composition of materials. This method uses low-energy protons (of several MeV), which can be obtained from high-energy (of tens MeV) accelerators. Instead of manufacturing an accelerator for generating the MeV protons, the use of a PET cyclotron has been suggested for designing the beamline for multipurpose applications, especially for the PIXE experiment, which has a dedicated high-energy (of tens MeV) accelerator. The beam properties of the cyclotron were determined at this experimental facility by using an external beamline before transferring the ion beam to the experimental chamber. We measured the beam profile and calculated the emittance using the pepper-pot method. The beam profile was measured as the beam current using a wire scanner, and the emittance was measured as the beam distribution at the beam dump using a radiochromic film. We analyzed the measurement results and are planning to use the results obtained in the simulations of external beamline and aligned beamline components. We will consider energy degradation after computing the beamline simulation. The experimental study focused on measuring the emittance from the cyclotron, and the results of this study are presented in this paper.

Beschreibung: The latest versions of RELAP5-3D© code allow the simulation of thermodynamic system, using different type of working fluids, that is, liquid metals, molten salt, diathermic oil, and so forth, thanks to the ATHENA code integration. The RELAP5-3D© water thermophysical properties are largely verified and validated; however there are not so many experiments to generate the liquid metals ones in particular for the Lead and the Lead Bismuth Eutectic. Recently, new and more accurate experimental data are available for liquid metals. The comparison between these state-of-the-art data and the RELAP5-3D© default thermophysical properties shows some discrepancy; therefore a tool for the generation of new properties binary files has been developed. All the available data came from experiments performed at atmospheric pressure. Therefore, to extend the pressure domain below and above this pressure, the tool fits a semiempirical model (soft sphere model with inverse-power-law potential), specific for the liquid metals. New binary files of thermophysical properties, with a detailed mesh grid of point to reduce the code mass error (especially for the Lead), were generated with this tool. Finally, calculations using a simple natural circulation loop were performed to understand the differences between the default and the new properties.

Beschreibung: With the adoption of new technologies, more risk is introduced into modern society. Important decisions about new technologies tend to be made by specialists, which can lead to a mismatch of risk perception between citizens and specialists, resulting in high social cost. Using contingent valuation methods, this paper analyzes the relationship between willingness to pay (WTP) and the factors expressed through people’s image of nuclear power plants (NPP), their perception of NPP safety, and how these can be affected by their scientific background level. Results indicate that groups with a high scientific background level tend to have low risk perception level, represented through their image and safety levels. Further, the results show that mean WTP is dependent on scientific background and image levels. It is believed that these results could help decision makers address the mismatch of trust between the public and specialists in terms of new policy.

Beschreibung: A hot gas duct is an indispensable component for the nuclear-process heat applications of the Very-High-Temperature Reactor (VHTR), which has to fulfill three requirements: to withstand high temperature, high pressure, and large pressure transient. In this paper, numerical investigation of pressure transient is performed for a hot gas duct under rapid depressurization. System depressurization imposes an imploding pressure differential on the internal structural elements of a hot gas duct, the structural integrity of which is susceptible to being damaged. Pressure differential and its imposed duration, which are two key factors to evaluate the damage severity of a hot gas duct under depressurization, are examined in regard to depressurization rate and insulation packing tightness. It is revealed that depressurization rate is a decisive parameter for controlling the pressure differential and its duration, whereas insulating-packing tightness has little effect on them.

Beschreibung: This study presents the medical radioisotope production performance of a conceptual accelerator driven system (ADS). Lead-bismuth eutectic (LBE) is selected as target material. The subcritical fuel core is conceptually divided into ten equidistant subzones. The ceramic (natural U, Pu)O2 fuel mixture and the materials used for radioisotope production (copper, gold, cobalt, holmium, rhenium, thulium, mercury, palladium, thallium, molybdenum, and yttrium) are separately prepared as cylindrical rods cladded with carbon/carbon composite (C/C) and these rods are located in the subzones. In order to obtain the flattened power density, percentages of PuO2 in the mixture of UO2 and PuO2 in the subzones are adjusted in radial direction of the fuel zone. Time-dependent calculations are performed at 1000 MW thermal fission power () for one hour using the BURN card. The neutronic results show that the investigated ADS has a high neutronic capability, in terms of medical radioisotope productions, spent fuel transmutation and energy multiplication. Moreover, a good quasiuniform power density is achieved in each material case. The peak-to-average fission power density ratio is in the range of 1.02–1.28.

Beschreibung: U-10wt%Zr spherical particles for use as particulate fuel were prepared by centrifugal atomization and subjected to pressureless sintering, which is one of the simplest powder processing techniques. At sintering temperature of 1100°C for 30 or 60 min, all samples ranging from +50 to −325 mesh showed no apparent bonding between the particles. However, at 1150°C (80 min), all samples formed a bulk body and the microstructures showed apparent sintering stages. Particularly, sample B (50–70 mesh) and sample C (70–100 mesh) showed pore characteristics suitable for a particulate fuel. The results suggest that pressureless sinterability for U-10Zr particulate fuel can be improved by adding small-size (–325 mesh) particles.

Beschreibung: This paper presents the results of the statistical analysis of the loss of offsite power events (LOOP) registered in four reviewed databases. The reviewed databases include the IRSN (Institut de Radioprotection et de Sûreté Nucléaire) SAPIDE database and the GRS (Gesellschaft für Anlagen- und Reaktorsicherheit mbH) VERA database reviewed over the period from 1992 to 2011. The US NRC (Nuclear Regulatory Commission) Licensee Event Reports (LERs) database and the IAEA International Reporting System (IRS) database were screened for relevant events registered over the period from 1990 to 2013. The number of LOOP events in each year in the analysed period and mode of operation are assessed during the screening. The LOOP frequencies obtained for the French and German nuclear power plants (NPPs) during critical operation are of the same order of magnitude with the plant related events as a dominant contributor. A frequency of one LOOP event per shutdown year is obtained for German NPPs in shutdown mode of operation. For the US NPPs, the obtained LOOP frequency for critical and shutdown mode is comparable to the one assessed in NUREG/CR-6890. Decreasing trend is obtained for the LOOP events registered in three databases (IRSN, GRS, and NRC).

Beschreibung: Due to increased global demand for clay, the present work involves the use of INAA for elemental analysis and pollutants concentration in clay. The samples were collected from Aswan in South Egypt. The samples were irradiated using the thermal neutrons “at the TRIGA Mainz research reactor” and at a neutron flux “of 7 × 10 n/cm s”. Twenty-six elements quantitatively and qualitatively were specified for the first time upon studying the samples. The elements determined are U, Th, Ta, Hf, Lu, Eu, Ce, Ba, Sn, Nb, Rb, Zn, Co, Fe, Cr, Sc, Sm, La, Yb, As, Ga, K, Mn, Na, Ti, and Mg. The concentrations of natural radionuclides 232Th, 226Ra, and 40K were also calculated. Based on these concentrations, to estimate the exposure risk for using clay as raw materials in building materials, the radiation hazard indices such as radium equivalent activities, effective doses rate, and the external hazard indices have been computed. The obtained results were compared with analogous studies carried out in other countries and with the UNSCEAR reports.

Beschreibung: A simple and analytical formula is suggested to solve the problems of the local burnup and the isotope distributions. The present method considers two extreme conditions of neutrons penetrating the fuel rod. Based on these considerations, the formula is obtained to calculate the reaction rates of 235U, 238U, and 239Pu and straightforward the local burnup and the isotope distributions. Starting from an initial burnup level, the parameters of the formula are fitted to the reaction rates given by a Monte Carlo (MC) calculation. Then the present formula independently gives very similar results to the MC calculation from the starting to high burnup level but takes just a few minutes. The relative reaction rates are found to be almost independent of the radius (except of  238U) and the burnup, providing a solid background for the present formula. A more realistic examination is also performed when the fuel rods locate in an assembly. A combination of the present formula and the MC calculation is expected to have a nice balance between the numerical accuracy and time consumption.

Beschreibung: A common measure of deformation between atomic scale simulations and the continuum framework is provided and the strain tensors for multiscale simulations are defined in this paper. In order to compute the deformation gradient of any atom , the weight function is proposed to eliminate the different contributions within the neighbor atoms which have different distances to atom , and the weighted least squares error optimization model is established to seek the optimal coefficients of the weight function and the optimal local deformation gradient of each atom. The optimization model involves more than 9 parameters. To guarantee the reliability of subsequent parameters identification result and lighten the calculation workload of parameters identification, an overall analysis method of parameter sensitivity and an advanced genetic algorithm are also developed.

Beschreibung: Small modular reactor (SMR) has drawn wide attention in the past decades, and Lead cooled fast reactor (LFR) is one of the most promising advanced reactors which are able to meet the safety economic goals of Gen-IV nuclear energy systems. A small modular natural circulation lead cooled fast reactor-100 MWth (SNRLFR-100) is being developed by University of Science and Technology of China (USTC). In the present work, a 3D CFD model, primary heat exchanger model, fuel pin model, and point kinetic model were established based on some reasonable simplifications and assumptions, the steady-state natural circulation characteristics of SNCLFR-100 primary cooling system were discussed and illustrated, and some reasonable suggestions were proposed for the reactor’s thermal-hydraulic and structural design. Moreover, in order to have a first evaluation of the system behavior in accident conditions, an unprotected loss of heat sink (ULOHS) transient simulation at beginning of the reactor cycle (BOC) has been analyzed and discussed based on the steady-state simulation results. The key temperatures of the reactor core are all under the safety limits at transient state; the reactor has excellent thermal-hydraulic performance.

Beschreibung: Gamma-ray measurements in various research fields require efficient detectors. One of these research fields is mass attenuation coefficients of different materials. Apart from experimental studies, the Monte Carlo (MC) method has become one of the most popular tools in detector studies. An NaI(Tl) detector has been modeled, and, for a validation study of the modeled NaI(Tl) detector, the absolute efficiency of 3 × 3 inch cylindrical NaI(Tl) detector has been calculated by using the general purpose Monte Carlo code MCNP-X (version 2.4.0) and compared with previous studies in literature in the range of 661–2620 keV. In the present work, the applicability of MCNP-X Monte Carlo code for mass attenuation of concrete sample material as building material at photon energies 59.5 keV, 80 keV, 356 keV, 661.6 keV, 1173.2 keV, and 1332.5 keV has been tested by using validated NaI(Tl) detector. The mass attenuation coefficients of concrete sample have been calculated. The calculated results agreed well with experimental and some other theoretical results. The results specify that this process can be followed to determine the data on the attenuation of gamma-rays with other required energies in other materials or in new complex materials. It can be concluded that data from Monte Carlo is a strong tool not only for efficiency studies but also for mass attenuation coefficients calculations.

Beschreibung: Falling water film on an inclined plane is studied by shadowgraphy. The ranges of inclination angle and the film Reynolds number are, respectively, up to 21° and 60. Water is used as working fluid. The scenario of wave regime evolution is identified as three distinctive regimes, namely, initial quiescent smooth film flow, two-dimensional regular solitary wave pattern riding on film flow, and three-dimensional irregular wave pattern. Three characteristic parameters of two-dimensional solitary wave pattern, namely, inception length, primary pulse spacing, and propagation velocity, are examined, which are significant in engineering applications for estimation of heat and mass transfer on film flow. The present experimental data are well in agreement with the Koizumi correlations, the deviation from which is limited to 20% and 15%, respectively, for primary pulse spacing and propagation velocity. Through the scrutiny of the present experimental observation, it is concluded that wave evolution on film flow at the moderate Reynolds number is controlled by gravity and drag and the Rayleigh-Taylor instability that occurred on the steep front of primary pulse triggers the disintegration of continuous two-dimensional regular solitary wave pattern into three-dimensional irregular wave pattern.

Beschreibung: In water-cooled reactor, the dominant radioactive source term under normal operation is activated corrosion products (ACPs), which have an important impact on reactor inspection and maintenance. A three-node transport model of ACPs was introduced into the new version of ACPs source term code CATE in this paper, which makes CATE capable of theoretically simulating the variation and the distribution of ACPs in a water-cooled reactor and suitable for more operating conditions. For code testing, MIT PWR coolant chemistry loop was simulated, and the calculation results from CATE are close to the experimental results from MIT, which means CATE is available and credible on ACPs analysis of water-cooled reactor. Then ACPs in the blanket cooling loop of water-cooled fusion reactor ITER under construction were analyzed using CATE and the results showed that the major contributors are the short-life nuclides, especially Mn-56. At last a point kernel integration code ARShield was coupled with CATE, and the dose rate around ITER blanket cooling loop was calculated. Results showed that after shutting down the reactor only for 8 days, the dose rate decreased nearly one order of magnitude, which was caused by the rapid decay of the short-life ACPs.

Beschreibung: Based on the special application of 90-degree elbow pipe in the HTR-PM, the large eddy simulation was selected to calculate the instantaneous flow field in the 90-degree elbow pipe combining with the experimental results. The characteristics of the instantaneous turbulent flow field under the influence of flow separation and secondary flow were studied by analyzing the instantaneous pressure information at specific monitoring points and the instantaneous velocity field on the cross section of the elbow. The pattern and the intensity of the Dean vortex and the small scale eddies change over time and induce the asymmetry of the flow field. The turbulent disturbance upstream and the flow separation near the intrados couple with the vortexes of various scales. Energy is transferred from large scale eddies to small scale eddies and dissipated by the viscous stress in the end.

Beschreibung: There is a large water storage tank installed at the top of containment of AP1000, which can supply the passive cooling. In the extreme condition, sloshing of the free surface in the tank may impact on the roof under long-period earthquake. For the safety assessment of structure, it is necessary to calculate the impact pressure caused by water sloshing. Since the behavior of sloshing impacted on the roof is involved into a strong nonlinear phenomenon, it is a little difficult to calculate such pressure by theoretical or numerical method currently. But it is applicable to calculate the height of sloshing in a tank without roof. In the present paper, a simplified method was proposed to calculate the impact pressure using the sloshing wave height, in which we first marked the position of the height of roof, then produced sloshing in the tank without roof and recorded the maximum wave height, and finally regarded approximately the difference between maximum wave height and roof height as the impact pressure head. We also designed an experiment to verify this method. The experimental result showed that this method overpredicted the impact pressure with a certain error of no more than 35%. By the experiment, we conclude that this method is conservative and applicable for the engineering design.

Beschreibung: Pellet-clad interaction (PCI) is one of the major issues in fuel rod design and reactor core operation in water cooled reactors. The prediction of fuel rod failure by PCI is studied in this paper by the method of radial basis function neural network (RBFNN). The neural network is built through the analysis of the existing experimental data. It is concluded that it is a suitable way to reduce the calculation complexity. A self-organized RBFNN is used in our study, which can vary its structure dynamically in order to maintain the prediction accuracy. For the purpose of the appropriate network complexity and overall computational efficiency, the hidden neurons in the RBFNN can be changed online based on the neuron activity and mutual information. The presented method is tested by the experimental data from the reference, and the results demonstrate its effectiveness.

Beschreibung: The acceptable accuracy for simulation of severe accident scenarios in containments of nuclear power plants is required to investigate the consequences of severe accidents and effectiveness of potential counter measures. For this purpose, the actual capability of CFX tool and COCOSYS code is assessed in prototypical geometries for simplified physical process-plume (due to a heat source) under adiabatic and convection boundary condition, respectively. Results of the comparison under adiabatic boundary condition show that good agreement is obtained among the analytical solution, COCOSYS prediction, and CFX prediction for zone temperature. The general trend of the temperature distribution along the vertical direction predicted by COCOSYS agrees with the CFX prediction except in dome, and this phenomenon is predicted well by CFX and failed to be reproduced by COCOSYS. Both COCOSYS and CFX indicate that there is no temperature stratification inside dome. CFX prediction shows that temperature stratification area occurs beneath the dome and away from the heat source. Temperature stratification area under adiabatic boundary condition is bigger than that under convection boundary condition. The results indicate that the average temperature inside containment predicted with COCOSYS model is overestimated under adiabatic boundary condition, while it is underestimated under convection boundary condition compared to CFX prediction.

Beschreibung: Small modular reactor (SMR) has drawn wide attention in the past decades, and Lead cooled fast reactor (LFR) is one of the most promising advanced reactors which are able to meet the safety economic goals of Gen-IV nuclear energy systems. A small modular natural circulation lead cooled fast reactor-100 MWth (SNRLFR-100) is being developed by University of Science and Technology of China (USTC). In the present work, a 3D CFD model, primary heat exchanger model, fuel pin model, and point kinetic model were established based on some reasonable simplifications and assumptions, the steady-state natural circulation characteristics of SNCLFR-100 primary cooling system were discussed and illustrated, and some reasonable suggestions were proposed for the reactor’s thermal-hydraulic and structural design. Moreover, in order to have a first evaluation of the system behavior in accident conditions, an unprotected loss of heat sink (ULOHS) transient simulation at beginning of the reactor cycle (BOC) has been analyzed and discussed based on the steady-state simulation results. The key temperatures of the reactor core are all under the safety limits at transient state; the reactor has excellent thermal-hydraulic performance.

Beschreibung: Helically coiled tube Once-Through Steam Generator (H-OTSG) is one of the key equipment types for small modular reactors. The flow instability of the secondary side of the H-OTSG is particularly serious, because the working condition is in the range of low and medium pressure. This paper presents research on density wave oscillations (DWO) in a typical countercurrent H-OTSG. Based on the steady-state calculation, the mathematical model of single-channel system was established, and the transfer function was derived. Using Nyquist stability criterion of the single variable, the stability cases were studied with an in-house computer program. According to the analyses, the impact law of the geometrical parameters to the system stability was obtained. RELAP5/MOD3.2 code was also used to simulate DWO in H-OTSG. The theoretical analyses of the in-house program were compared to the simulation results of RELAP5. A correction factor was introduced to reduce the error of RELAP5 when modeling helical geometry. The comparison results agreed well which showed that the correction is effective.

Beschreibung: In Korea, R&D on a sodium-cooled fast reactor (SFR) was begun in 1997, as one of the national long-term nuclear R&D programs. As one fuel option for a prototype SFR, a metallic fuel, U-Zr alloy fuel, was selected and is currently being developed. For the fabrication of SFR metallic fuel rods, the end plug welding is a crucial process. The sealing of the end plug to the cladding tube should be hermetically perfect to prevent a leakage of fission gases and to maintain a good reactor performance. In this study, the welding technique, welding equipment, welding conditions, and parameters were developed for the end plug welding of SFR metallic fuel rods. A gas tungsten arc welding (GTAW) technique was adopted and the welding joint design was developed. In addition, the optimal welding conditions and parameters were established. Based on the establishment of the welding conditions, the GTAW technique was qualified for the end plug welding of SFR metallic fuel rods.

Beschreibung: Current radiation protection methodology offers abundant experiences on light-water reactors, but very few studies on high temperature gas-cooled reactor (HTR). To fill this gap, a comprehensive investigation was performed to the radiation protection practices in the helium circulator maintenance of the Chinese 10 MW HTR test module (HTR-10) in this paper. The investigation reveals the unique behaviour of HTR-10’s radiation sources in the maintenance as well as its radionuclide species and presents the radiation protection methods that were tailored to these features. Owing to these practices, the radioactivity level was kept low throughout the maintenance and only low-level radioactive waste was generated. The quantitative analysis further demonstrates that the decontamination efficiency was over 89% for surface contamination and over 34% forγdose rate and the occupational exposure was much lower than both the limits of regulatory and the exposure levels in comparable literature. These results demonstrate the effectiveness of the reported radiation protection practices, which directly provides hands-on experience for the future HTR-PM reactor and adds to the completeness of the radiation protection methodology.

Beschreibung: An integration design platform is under development for the design of the China Fusion Engineering Test Reactor (CFETR). It mainly includes the integration physical design platform and the integration engineering design platform. The integration engineering design platform aims at performing detailed engineering design for each tokamak component (e.g., breeding blanket, divertor, and vacuum vessel). The vacuum vessel design and analysis module is a part of the integration engineering design platform. The main idea of this module is to integrate the popular CAD/CAE software to form a consistent development environment. Specifically, the software OPTIMUS provides the approach to integrate the CAD/CAE software such as CATIA and ANSYS and form a design/analysis workflow for the vacuum vessel module. This design/analysis workflow could automate the process of modeling and finite element (FE) analysis for vacuum vessel. Functions such as sensitivity analysis and optimization of geometric parameters have been provided based on the design/analysis workflow. In addition, data from the model and FE analysis could be easily exchanged among different modules by providing a unifying data structure to maintain the consistency of the global design. This paper describes the strategy and methodology of the workflow in the vacuum vessel module. An example is given as a test of the workflow and functions of the vacuum vessel module. The results indicate that the module is a feasible framework for future application.

Beschreibung: Instrumental neutron activation analysis (INAA) is a powerful technique for trace element determination in rocks. Nine alabaster samples were collected from Wadi El-Nakhil located at the intersection of lat. 26°10′50′′N and long. 34°03′40′′E, central Eastern Desert, Egypt, for investigation by INAA and Energy Depressive X-Ray Fluorescence (EDXRF). The samples were irradiated by thermal neutrons at the TRIGA Mainz research reactor at a neutron flux of 7 × 1011 n/cm2·s. Twenty-two elements were determined, namely, As, Ba, Ca, Co, Cr, Sc, Fe, Hf, K, Mg, Mn, Na, Rb, U, Zn, Zr, Lu, Ce, Sm, La, Yb, and Eu. The chemical analysis of alabaster indicated having high contents of CaO and MgO and LOI and low contents of SiO2, Al2O3, Na2O, K2O, MnO, and Fe2O3.

Beschreibung: This paper reports two numerical simulation methods for modeling displacement instabilities around a surface groove in a metal substrate used in nuclear power plant. The amplitude change in the groove, the downward displacement at the base node, and the groove displacement at the periphery were simulated using ABAQUS to compare the results from two methods, as well as the tangential stress in the elements at the groove base and periphery. The comparison showed that for the tangential stress two methods were in close agreement for all thermal cycles. For the amplitude change, the downward displacement, the groove displacement, and the stress distribution, the two methods were in close agreement for the first 3 to 6 thermal cycles. After that, inconsistency increased with the number of thermal cycles. It is interesting that the thermal cycle at which the discrepancy between the two methods began to occur corresponded to a thermally grown oxide (TGO) thickness of 1 μm, which showed the accuracy of the present work over the classic method. It is concluded that the present work’s numerical simulation scheme worked better with a thinner TGO layer than the classic method and could overcome the limitation of TGO thickness by simulating any thickness.

Beschreibung: The graphite oxidation of fuel element has obtained high attention in air ingress accident analysis of high temperature gas-cooled reactor (HTR). The shape function, defined as the relationship between the maximum and the average of the oxidation, is an important factor to estimate the consequence of the accident. There are no detailed studies on the shape function currently except two experiments several decades ago. With the development of computer technology, CFD method is used in the numerical experiment about graphite oxidation in pebble bed of HTR in this paper. Structured packed beds are used in the calculation instead of random packed beds. The result shows the nonuniform distribution of oxidation on the sphere surface and the shape function in the condition of air ingress accident. Furthermore, the sensitive factors of shape function, such as temperature and Re number, are discussed in detail and the relationship between the shape function and sensitive factors is explained. According to the results in this paper, the shape function ranges from 1.05 to 4.7 under the condition of temperature varying from 600°C to 1200°C and Re varying from 16 to 1600.

Beschreibung: This study presents the power flattening and time-dependent neutronic analysis of a conceptual helium gas cooled Accelerator Driven System (ADS) loaded with TRISO (tristructural-isotropic) fuel particles. Target material is lead-bismuth eutectic (LBE). ThO2, UO2, PuO2, and CmO2TRISO particles are used as fuel. PuO2and CmO2fuels are extracted from PWR-MOX spent fuel. Subcritical core is radially divided into 10 equidistant subzones in order to flatten the power produced in the core. Tens of thousands of these TRISO fuel particles are embedded in the carbon matrix fuel pebbles as five different cases. The high-energy Monte Carlo code MCNPX 2.7 with the LA150 library is used for the neutronic calculations. Time-dependent burnup calculations are carried out for thermal fission power (Pth) of 1000 MW using the BURN card. The energy gain of the ADS is in the range of 99.98–148.64 at the beginning of a cycle. Furthermore, the peak-to-average fission power density ratio is obtained between 1.021 and 1.029 at the beginning of the cycle. These ratios show a good quasi-uniform power density for each case. Furthermore, up to 155.1 g233U and 103.6 g239Pu per day can be produced. The considered system has a high neutronic capability in terms of energy multiplication, fissile breeding, and spent fuel transmutation with thorium utilization.

Beschreibung: Maintaining the structural integrity of the reactor pressure vessel (RPV) is a critical concern related to the safe operation of nuclear power plants. To estimate the structural integrity over the designed lifetime and to support analyses for a potential plant life extension, an accurate calculation of the fast neutron fluence (E〉1.0 MeV orE〉0.1 MeV) at the RPV is significant. The discrete ordinates method is one of the main methods to solve such problems. During the calculation process, many factors will affect the results. In this paper, the deviations introduced by different differencing schemes and mesh sizes on the AP1000 RPV fast neutron fluence have been studied, which are based on new discrete ordinates code ARES. The analysis shows that the differencing scheme (diamond difference with or without linear zero fix-up, theta weighted, directional theta weighted, and exponential directional weighted) introduces a deviation within 4%. The coarse mesh (4 × 4 cm meshes inXYplane) leads to approximately 23.7% calculation deviation compared to those of refined mesh (1 × 1 cm meshes inXYplane). Comprehensive study on the deviation introduced by differencing scheme and mesh size has great significance for reasoned evaluation of RPV fast neutron fluence calculation results.

Beschreibung: Due to safety reasons, the exact number and location of people working in an underground tunnel need to be known all the time. This work introduces the development and implementation of an RFID-based access monitoring system for the ONKALO nuclear waste storage facility. This system was taken into use in 2010 and was systematically monitored for one year. The system principle and the used equipment are presented in this paper together with the reliability evaluation results of the implemented system. According to the field use evaluation of the ready system, the reading reliability at the end of the monitoring period was 100%. In addition, even after the successful monitoring period, the system has been updated and new features for safety improvement have been created based on fire department guidelines and achieved user experience. In the future, the RFID system has been planned to be used also in the final depositing of the used nuclear fuel and buffer materials.

Beschreibung: Commercial 41Cr4 (ISO standard) steel was treated by a composite technique. An intermediate layer was introduced firstly at the 41Cr4 steel surface by traditional carburizing and nitriding. Then a hard Cr coating was brush-plated on the intermediate layer. Finally, the coating layer was modified by high current pulsed electron beam (HCPEB), followed by quenching and subsequent tempering treatment. The microstructure, mechanical properties, and fracture behavior were characterized. The results show that a nanocrystalline Cr coating is formed at the 41Cr4 steel surface by the treatment of the new composite technique. Such nanocrystalline Cr coating has acceptable hardness and high corrosion resistance performance, which satisfies the demands of the gears working under high speed and corrosive environment. The composite process proposed in this study is considered as a new prospect method due to the multifunction layer design on the gear surface.

Beschreibung: The moderator system of CANDU, a prototype of PHWR (pressurized heavy-water reactor), has been modeled in multidimension for the computation based on CFD (computational fluid dynamics) technique. Three CFD codes are tested in modeled hydrothermal systems of heavy-water reactors. Commercial codes, COMSOL Multiphysics and ANSYS-CFX with OpenFOAM, an open-source code, are introduced for the various simplified and practical problems. All the implemented computational codes are tested for a benchmark problem of STERN laboratory experiment with a precise modeling of tubes, compared with each other as well as the measured data and a porous model based on the experimental correlation of pressure drop. Also the effect of turbulence model is discussed for these low Reynolds number flows. As a result, they are shown to be successful for the analysis of three-dimensional numerical models related to the calandria system of CANDU reactors.

Beschreibung: Due to high radioactivity and significant content of medium- and long-lived radionuclides, different operations with spent nuclear fuels (e.g., handling, transportation, and storage) shall be accompanied by suitable radiation protections. On the other hand, determination of radioactive source specification is the initial step for any radiation protection design. In this study, radioactive source specification of the spent fuels of Bushehr nuclear power plant, which is a VVER-1000 type pressurized water reactor, was determined. For the depletion and decay calculations, ORIGEN code was utilized. The results are presented for burnups of 30 to 49 GWd/MTHM and different cooling times up to 100 years. According to these results, total activity of a spent fuel assembly with initial enrichment of 3.92%, burnup of 49 GWd/MTHM, and cooling time of 3 years is 1.92 × 1016 Bq. The results can be utilized specifically in transportation/storage cask design for spent fuel management of Bushehr nuclear power plant.

Beschreibung: For the China Initiative Accelerator Driven System (CIADS), the energy of the protons is 250 MeV, and the current intensity will reach 10 milliamperes. A new concept of a dense granular spallation target is proposed for which the tungsten granules are chosen as the target material. After being bombarded with the accelerated protons from the accelerator, the tungsten granules with high-temperature flow out of the subcritical reactor and the heat is removed by the heat exchanger. One key issue of the target is to remove the 2.5 MW heat deposition safely. Another one is the heat exchange between the target and the subcritical reactor. Based on the model of effective thermal conductivity, a new thermal code is developed in Matlab. The new code is used to calculate the temperature field of the target area near active zone and it is partly verified by commercial CFD code Fluent. The result shows that the peak temperature of the target zone is nearly 740°C and the reactor and the target are proved to be uncoupled in thermal process.

Beschreibung: The passive safety systems of AP1000 are designed to operate automatically at desired set-points. However, the unavailability or failure to operate of any of the passive safety systems will change the accident sequence and may affect reactor safety. The analysis in this study is based on some hypothetical scenarios, in which the passive safety system failure is considered during the loss of coolant accidents. Four different cases are assumed, that is, with all passive systems, without actuation of one of the accumulators, without actuation of ADS stages 1–3, and without actuation of ADS stage 4. The actuation of all safety systems at their actuation set-points provides adequate core cooling by injecting sufficient water inventory into reactor core. The LOCA with actuation of one of the accumulators cause early actuation of ADS and IRWST. In case of LOCA without ADS stages 1–3, the primary system depressurization is relatively slow and mixture level above core active region drops much earlier than IRWST actuation. The accident without ADS stage 4 actuation results in slow depressurization and mixture level above core active region drops earlier than IRWST injection. Moreover, the comparison of cladding surface temperature is performed in all cases considered in this work.

Beschreibung: Spent fuel rack is the key equipment for the storage of spent fuel after refueling. In order to investigate the performance of the spent fuel rack under the earthquake, the phenomena including sliding, collision, and overturning of the spent fuel rack were studied. An FEM model of spent fuel rack is built to simulate the transient response under seismic loading regarding fluid-structure interaction by ANSYS. Based on D’Alambert’s principle, the equilibriums of force and momentum were established to obtain the critical sliding and overturning accelerations. Then 5 characteristic transient loadings which were designed based on the critical sliding and overturning accelerations were applied to the rack FEM model. Finally, the transient displacement and impact force response of rack with different gap sizes and the supporting leg friction coefficients were analyzed. The result proves the FEM model is applicable for seismic response of spent fuel rack. This paper can guide the design of the future’s fluid-structure interaction experiment for spent fuel rack.

Beschreibung: The effect of BWR fuel assembly 3D model on void reactivity coefficient (VRC) estimation is investigated. VRC values were calculated for different BWR assembly models applying deterministic T-NEWT and Monte Carlo KENO-VI functional modules of SCALE 6.1 code package. The difference between deterministic T-NEWT and Monte Carlo KENO-VI simulations is negligible (0.18 pcm/%). The influence of the assumed more detailed coolant density profile was estimated as well. VRC increases with the application of a larger number of coolant density values across fuel assembly height. It was shown that the coolant density profile described by 6 values per height could be considered sufficient from prospect of VRC estimation, as a more detailed density profile has impact below 1% on total assembly void effects. VRC values were decomposed to values for individual nodes and isotopes, since decomposition provides useful insights to describe the overall behaviour of VRC in detail.

Beschreibung: Thermally grown oxide (TGO), commonly pure α-Al2O3, formed on protective coatings acts as an insulation barrier shielding cooled reactors from high temperatures in nuclear energy systems. Mixed zone (MZ) oxide often grows at the interface between the alumina layer and top coat in thermal barrier coatings (TBCs) at high temperature dwell times accompanied by the formation of alumina. The newly formed MZ destroys interface integrity and significantly affects the displacement instabilities of TGO. In this work, a finite element model based on material property changes was constructed to investigate the effects of MZ on the displacement instabilities of TGO. MZ formation was simulated by gradually changing the metal material properties into MZ upon thermal cycling. Quantitative data show that MZ formation induces an enormous stress in TGO, resulting in a sharp change of displacement compared to the alumina layer. The displacement instability increases with an increase in the MZ growth rate, growth strain, and thickness. Thus, the formation of a MZ accelerates the failure of TBCs, which is in agreement with previous experimental observations. These results provide data for the understanding of TBC failure mechanisms associated with MZ formation and of how to prolong TBC working life.

Beschreibung: A three-dimensional, multigroup, diffusion code based on a high order nodal expansion method for hexagonal-zgeometry (HNHEX) was developed to perform the neutronic analysis of hexagonal-zgeometry. In this method, one-dimensional radial and axial spatially flux of each node and energy group are defined as quadratic polynomial expansion and four-order polynomial expansion, respectively. The approximations for one-dimensional radial and axial spatially flux both have second-order accuracy. Moment weighting is used to obtain high order expansion coefficients of the polynomials of one-dimensional radial and axial spatially flux. The partially integrated radial and axial leakages are both approximated by the quadratic polynomial. The coarse-mesh rebalance method with the asymptotic source extrapolation is applied to accelerate the calculation. This code is used for calculation of effective multiplication factor, neutron flux distribution, and power distribution. The numerical calculation in this paper for three-dimensional SNR and VVER 440 benchmark problems demonstrates the accuracy of the code. In addition, the results show that the accuracy of the code is improved by applying quadratic approximation for partially integrated axial leakage and four-order approximation for one-dimensional axial spatially flux in comparison to flat approximation for partially integrated axial leakage and quadratic approximation for one-dimensional axial spatially flux.

Beschreibung: The capabilities of the SCALE6.1/MAVRIC hybrid shielding methodology (CADIS and FW-CADIS) were demonstrated when applied to a realistic deep penetration Monte Carlo (MC) shielding problem of a full-scale PWR containment model. Automatic preparation of variance reduction (VR) parameters is based on deterministic transport theory (SNmethod) providing the space-energy importance function. The aim of this paper was to determine the neutron-gamma dose rate distributions over large portions of PWR containment with uniformly small MC uncertainties. The sources of ionizing radiation included fission neutrons and photons from the reactor and photons from the activated primary coolant. We investigated benefits and differences of FW-CADIS over CADIS methodology for the objective of the uniform MC particle density in the desired tally regions. Memory intense deterministic module was used with broad group library “v7_27n19g” opposed to the fine group library “v7_200n47g” used for final MC simulation. Compared with CADIS and with the analog MC, FW-CADIS drastically improved MC dose rate distributions. Modern shielding problems with large spatial domains require not only extensive computational resources but also understanding of the underlying physics and numerical interdependence betweenSN-MC modules. The results of the dose rates throughout the containment are presented and discussed for different volumetric adjoint sources.

Beschreibung: Subsurface damage could affect the service life of structures. In nuclear engineering, nondestructive evaluation and detection of the evaluation of the subsurface damage region are of great importance to ensure the safety of nuclear installations. In this paper, we propose the use of circumferential horizontal shear (SH) waves to detect mechanical properties of subsurface regions of damage on cylindrical structures. The regions of surface damage are considered to be functionally graded material (FGM) and the cylinder is considered to be a layered structure. The Bessel functions and the power series technique are employed to solve the governing equations. By analyzing the SH waves in the 12Cr-ODS ferritic steel cylinder, which is frequently applied in the nuclear installations, we discuss the relationship between the phase velocities of SH waves in the cylinder with subsurface layers of damage and the mechanical properties of the subsurface damaged regions. The results show that the subsurface damage could lead to decrease of the SH waves’ phase velocity. The gradient parameters, which represent the degree of subsurface damage, can be evaluated by the variation of the SH waves’ phase velocity. Research results of this study can provide theoretical guidance in nondestructive evaluation for use in the analysis of the reliability and durability of nuclear installations.

Beschreibung: This paper reports the results of sensitivity analysis of the multidimension, multigroup neutron diffusion NODAL3 code for the NEACRP 3D LWR core transient benchmarks (PWR). The code input parameters covered in the sensitivity analysis are the radial and axial node sizes (the number of radial node per fuel assembly and the number of axial layers), heat conduction node size in the fuel pellet and cladding, and the maximum time step. The output parameters considered in this analysis followed the above-mentioned core transient benchmarks, that is, power peak, time of power peak, power, averaged Doppler temperature, maximum fuel centerline temperature, and coolant outlet temperature at the end of simulation (5 s). The sensitivity analysis results showed that the radial node size and maximum time step give a significant effect on the transient parameters, especially the time of power peak, for the HZP and HFP conditions. The number of ring divisions for fuel pellet and cladding gives negligible effect on the transient solutions. For productive work of the PWR transient analysis, based on the present sensitivity analysis results, we recommend NODAL3 users to use2×2radial nodes per assembly,1×18axial layers per assembly, the maximum time step of 10 ms, and 9 and 1 ring divisions for fuel pellet and cladding, respectively.

Beschreibung: NPP’s power supply system is repairable and there is common cause failure between the components. The repair rate is introduced and total signaling is considered in the improved GO-FLOW method, aimed at reliability analysis for NPP’s power supply system. Traditional GO-FLOW operators’ algorithms are improved. Comprehensively considering the effect of total signaling flow in the power supply system, the equivalent reliability parameter model and common cause failure probability model of multimodal repairable components are constructed. The improved GO-FLOW model of NPP’s power supply system is set up. Based on the proposed model, components’ reliability parameters are computed. The failure probability time-varying trend in thirty years, respectively, of NPP’s offsite power source and power supply system, is simulated and analyzed. Compared with calculation results of dynamic fault tree analysis method, the validity and the simplicity of the improved GO-FLOW method are verified. The effectiveness and applicability of the improved GO-FLOW model for NPP’s power supply system are proved by simulation examples.

Beschreibung: Sensors health monitoring is essentially important for reliable functioning of safety-critical chemical and nuclear power plants. Autoassociative neural network (AANN) based empirical sensor models have widely been reported for sensor calibration monitoring. However, such ill-posed data driven models may result in poor generalization and robustness. To address above-mentioned issues, several regularization heuristics such as training with jitter, weight decay, and cross-validation are suggested in literature. Apart from these regularization heuristics, traditional error gradient based supervised learning algorithms for multilayered AANN models are highly susceptible of being trapped in local optimum. In order to address poor regularization and robust learning issues, here, we propose a denoised autoassociative sensor model (DAASM) based on deep learning framework. Proposed DAASM model comprises multiple hidden layers which are pretrained greedily in an unsupervised fashion under denoising autoencoder architecture. In order to improve robustness, dropout heuristic and domain specific data corruption processes are exercised during unsupervised pretraining phase. The proposed sensor model is trained and tested on sensor data from a PWR type nuclear power plant. Accuracy, autosensitivity, spillover, and sequential probability ratio test (SPRT) based fault detectability metrics are used for performance assessment and comparison with extensively reported five-layer AANN model by Kramer.

Beschreibung: Neutron production methods are an integral part of research and analysis for an array of applications. This paper examines methods of neutron production, and the advantages of constructing a radioisotopic neutron irradiator assembly using252Cf. Characteristic neutron behavior and cost-benefit comparative analysis between alternative modes of neutron production are also examined. The irradiator is described from initial conception to the finished design. MCNP modeling shows a total neutron flux of 3 × 105 n/(cm2·s) in the irradiation chamber for a 25 μg source. Measurements of the gamma-ray and neutron dose rates near the external surface of the irradiator assembly are 120 μGy/h and 30 μSv/h, respectively, during irradiation. At completion of the project, total material, and labor costs remained below $50,000.

Beschreibung: In order to investigate single bubble evolution, a boiling phase change model in subcooled flow boiling is proposed in this paper, and VOF model combined with phase change model is adopted to simulate the single bubble growth and movement. The effects of flow velocity, liquid subcooling, wall superheat, and vapor-liquid contact angle are considered in this model. The predicted bubble growth curve agrees well with the experimental result. Based on the analysis of bubble shape evolution and temperature field, it is found that the average bubble growth rate, flow velocity, and dynamic contact angle have significant effect on the bubble shape evolution during the bubble growth and movement while the temperature gradient in superheated liquid does not change with bubble growing. The character of dynamic contact angle during bubble growth and movement is also obtained in different working condition.

Beschreibung: The fabrication technology for metallic fuel has been developed to produce the driver fuel in a PGSFR in Korea since 2007. In order to evaluate the irradiation integrity and validate the in-reactor of the starting metallic fuel with FMS cladding for the loading of the metallic fuel, U-10 wt.%Zr fuel rodlets were fabricated and evaluated for a verification of the starting driver fuel through an irradiation test in the BOR-60 fast reactor. The injection casting method was applied to U-10 wt.%Zr fuel slugs with a diameter of 5.5 mm. Consequently, fuel slugs per melting batch without casting defects were fabricated through the development of advanced casting technology and evaluation tests. The optimal GTAW welding conditions were also established through a number of experiments. In addition, a qualification test was carried out to prove the weld quality of the end plug welding of the metallic fuel rodlets. The wire wrapping of metallic fuel rodlets was successfully accomplished for the irradiation test. Thus, PGSFR fuel rodlets have been soundly fabricated for the irradiation test in a BOR-60 fast reactor.

Beschreibung: A combined cycle that combines AWM cycle with a nuclear closed Brayton cycle is proposed to recover the waste heat rejected from the precooler of a nuclear closed Brayton cycle in this paper. The detailed thermodynamic and economic analyses are carried out for the combined cycle. The effects of several important parameters, such as the absorber pressure, the turbine inlet pressure, the turbine inlet temperature, the ammonia mass fraction, and the ambient temperature, are investigated. The combined cycle performance is also optimized based on a multiobjective function. Compared with the closed Brayton cycle, the optimized power output and overall efficiency of the combined cycle are higher by 2.41% and 2.43%, respectively. The optimized LEC of the combined cycle is 0.73% lower than that of the closed Brayton cycle.

Beschreibung: Considering dangerous environmental conditions, maintenance of radioactive equipment can be performed by remote handling maintenance (RHM) system. The RHM system is a sophisticated man-machine system. Therefore, human factors analysis is an inevitable aspect considered in guaranteeing successful and safe task performance. This study proposes an approach for integrated analysis of human factors in RHM so as to make the evaluating process more practical. In the approach, indicators of accessibility, health safety, and fatigue are analyzed using virtual human simulation technologies. The human error factors in the maintenance process are analyzed using the human error probability (HEP) based on the success likelihood index method- (SLIM-) analytic hierarchy process (AHP). The psychological factors level of maintenance personnel is determined with an expert scoring. The human factors for the entire RHM system are then evaluated using the interval method. An application example is present, and the application results show that the approach can support the evaluation of the human factors in RHM.

Beschreibung: Korea Atomic Energy Research Institute (KAERI) started the experimental research on moderator circulation as one of a the national research and development programs from 2012. This research program includes the construction of the moderator circulation test (MCT) facility, production of the validation data for self-reliant computational fluid dynamics (CFD) tools, and development of optical measurement system using the particle image velocimetry (PIV). In the present paper we introduce the scaling analysis performed to extend the scaling criteria suitable for reproducing thermal-hydraulic phenomena in a scaled-down CANDU- (CANada Deuterium Uranium-) 6 moderator tank, a manufacturing status of the 1/4 scale moderator tank. Also, preliminary CFD analysis results for the full-size and scaled-down moderator tanks are carried out to check whether the moderator flow and temperature patterns of both the full-size reactor and scaled-down facility are identical.

Beschreibung: Two tests related to a new safety system for a pressurized water reactor were performed with the ROSA/LSTF (rig of safety assessment/large scale test facility). The tests simulated cold leg small-break loss-of-coolant accidents with 2-inch diameter break using an early steam generator (SG) secondary-side depressurization with or without release of nitrogen gas dissolved in accumulator (ACC) water. The SG depressurization was initiated by fully opening the depressurization valves in both SGs immediately after a safety injection signal. The pressure difference between the primary and SG secondary sides after the actuation of ACC system was larger in the test with the dissolved gas release than that in the test without the dissolved gas release. No core uncovery and heatup took place because of the ACC coolant injection and two-phase natural circulation. Long-term core cooling was ensured by the actuation of low-pressure injection system. The RELAP5 code predicted most of the overall trends of the major thermal-hydraulic responses after adjusting a break discharge coefficient for two-phase discharge flow under the assumption of releasing all the dissolved gas at the vessel upper plenum.

Beschreibung: A methodology to preliminarily evaluate the size of the suppression tank and the relief pipes for a Vacuum Vessel Pressure Suppression System, to be adopted in a fusion reactor based on a water cooled blanket, is presented. The volume of the ST depends on the total energy of the water cooling system and it can be sized based on a required final pressure at equilibrium, by a simple energy balance. The pressure peak in the VV depends mainly on break area and the flow area of the relief pipes and some suggestions about the method for a preliminarily evaluation of their size are discussed. The computer code CONSEN has been used to perform a parametric study and to verify the methodology.

Beschreibung: Atmospheric dispersion modeling and radiation dose calculations have been performed for a hypothetical AP1000 SGTR accident by HotSpot code 3.03. TEDE, the respiratory time-integrated air concentration, and the ground deposition are calculated for various atmospheric stability classes, Pasquill stability categories A–F with site-specific averaged meteorological conditions. The results indicate that the maximum plume centerline ground deposition value of1.2E+2 kBq/m2occurred at about 1.4 km and the maximum TEDE value of1.41E-05 Sv occurred at 1.4 km from the reactor. It is still far below the annual regulatory limits of 1 mSv for the public as set in IAEA Safety Report Series number 115. The released radionuclides might be transported to long distances but will not have any harmful effect on the public.

Beschreibung: Generally the piping system of a nuclear power plant (NPP) has to be designed for normal loads such as dead weight, internal pressure, temperature, and accidental loads such as earthquake. In the proposed paper, effect of Stockbridge damper to mitigate the response of piping system of NPP subjected to earthquake is studied. Finite element analysis of piping system with and without Stockbridge damper using commercial software SAP2000 is performed. Vertical and horizontal components of earthquakes such as El Centro, California, and Northridge are used in the piping analysis. A sine sweep wave is also used to investigate the control effects on the piping system under wide frequency range. It is found that the proposed Stockbridge damper can reduce the seismic response of piping system subjected to earthquake loading.

Beschreibung: Long-term immersion was adopted to explore the damage deterioration and mechanical properties of granite under different chemical solutions. Here, granite was selected as the candidate of parent rocks for nuclear waste storage. The physical and mechanical properties of variation regularity immersed in various chemical solutions were analyzed. Meanwhile, the damage variable based on the variation in porosity was used in the quantitative analysis of chemical damage deterioration degree. Experimental results show that granite has a significant weakening tendency after chemical corrosion. The fracture toughnessKIC, splitting tensile strength, and compressive strength all demonstrate the same deteriorating trend with chemical corrosion time. However, a difference exists in the deterioration degree of the mechanical parameters; that is, the deterioration degree of fracture toughnessKICis the greatest followed by those of splitting tensile strength and compressive strength, which are relatively smaller. Strong acid solutions may aggravate chemical damage deterioration in granite. By contrast, strong alkaline solutions have a certain inhibiting effect on chemical damage deterioration. The chemical solutions that feature various compositions may have different effects on chemical damage degree; that is,SO42-ions have a greater effect on the chemical damage in granite thanHCO3-ions.

Beschreibung: This paper presents the results of the statistical analysis of the loss of offsite power events (LOOP) registered in four reviewed databases. The reviewed databases include the IRSN (Institut de Radioprotection et de Sûreté Nucléaire) SAPIDE database and the GRS (Gesellschaft für Anlagen- und Reaktorsicherheit mbH) VERA database reviewed over the period from 1992 to 2011. The US NRC (Nuclear Regulatory Commission) Licensee Event Reports (LERs) database and the IAEA International Reporting System (IRS) database were screened for relevant events registered over the period from 1990 to 2013. The number of LOOP events in each year in the analysed period and mode of operation are assessed during the screening. The LOOP frequencies obtained for the French and German nuclear power plants (NPPs) during critical operation are of the same order of magnitude with the plant related events as a dominant contributor. A frequency of one LOOP event per shutdown year is obtained for German NPPs in shutdown mode of operation. For the US NPPs, the obtained LOOP frequency for critical and shutdown mode is comparable to the one assessed in NUREG/CR-6890. Decreasing trend is obtained for the LOOP events registered in three databases (IRSN, GRS, and NRC).

Beschreibung: Particle-induced X-ray emission is used for determining the elemental composition of materials. This method uses low-energy protons (of several MeV), which can be obtained from high-energy (of tens MeV) accelerators. Instead of manufacturing an accelerator for generating the MeV protons, the use of a PET cyclotron has been suggested for designing the beamline for multipurpose applications, especially for the PIXE experiment, which has a dedicated high-energy (of tens MeV) accelerator. The beam properties of the cyclotron were determined at this experimental facility by using an external beamline before transferring the ion beam to the experimental chamber. We measured the beam profile and calculated the emittance using the pepper-pot method. The beam profile was measured as the beam current using a wire scanner, and the emittance was measured as the beam distribution at the beam dump using a radiochromic film. We analyzed the measurement results and are planning to use the results obtained in the simulations of external beamline and aligned beamline components. We will consider energy degradation after computing the beamline simulation. The experimental study focused on measuring the emittance from the cyclotron, and the results of this study are presented in this paper.

Beschreibung: U-10wt%Zr spherical particles for use as particulate fuel were prepared by centrifugal atomization and subjected to pressureless sintering, which is one of the simplest powder processing techniques. At sintering temperature of 1100°C for 30 or 60 min, all samples ranging from +50 to −325 mesh showed no apparent bonding between the particles. However, at 1150°C (80 min), all samples formed a bulk body and the microstructures showed apparent sintering stages. Particularly, sample B (50–70 mesh) and sample C (70–100 mesh) showed pore characteristics suitable for a particulate fuel. The results suggest that pressureless sinterability for U-10Zr particulate fuel can be improved by adding small-size (–325 mesh) particles.

Beschreibung: The latest versions of RELAP5-3D©code allow the simulation of thermodynamic system, using different type of working fluids, that is, liquid metals, molten salt, diathermic oil, and so forth, thanks to the ATHENA code integration. The RELAP5-3D©water thermophysical properties are largely verified and validated; however there are not so many experiments to generate the liquid metals ones in particular for the Lead and the Lead Bismuth Eutectic. Recently, new and more accurate experimental data are available for liquid metals. The comparison between these state-of-the-art data and the RELAP5-3D©default thermophysical properties shows some discrepancy; therefore a tool for the generation of new properties binary files has been developed. All the available data came from experiments performed at atmospheric pressure. Therefore, to extend the pressure domain below and above this pressure, the tool fits a semiempirical model (soft sphere model with inverse-power-law potential), specific for the liquid metals. New binary files of thermophysical properties, with a detailed mesh grid of point to reduce the code mass error (especially for the Lead), were generated with this tool. Finally, calculations using a simple natural circulation loop were performed to understand the differences between the default and the new properties.

Beschreibung: A simple and analytical formula is suggested to solve the problems of the local burnup and the isotope distributions. The present method considers two extreme conditions of neutrons penetrating the fuel rod. Based on these considerations, the formula is obtained to calculate the reaction rates of235U,238U, and239Pu and straightforward the local burnup and the isotope distributions. Starting from an initial burnup level, the parameters of the formula are fitted to the reaction rates given by a Monte Carlo (MC) calculation. Then the present formula independently gives very similar results to the MC calculation from the starting to high burnup level but takes just a few minutes. The relative reaction rates are found to be almost independent of the radius (except(n,γ)of  238U) and the burnup, providing a solid background for the present formula. A more realistic examination is also performed when the fuel rods locate in an assembly. A combination of the present formula and the MC calculation is expected to have a nice balance between the numerical accuracy and time consumption.

Beschreibung: Emergency planning zones (PAZ and UPZ) around the Karachi-2 and Karachi-3 nuclear power plants (K-2/K-3 NPPs) have been realistically determined by employing Gaussian puff model and Gaussian plume model together for atmospheric transport, diffusion, and deposition of radioactive material using onsite and regional data related to meteorology, topography, and land-use along with latest IAEA Post-Fukushima Guidelines. The analysis work has been carried out using U.S.NRC computer code RASCAL 4.2. The assumed environmental radioactive releases provide the sound theoretical and practical bases for the estimation of emergency planning zones covering most expected scenario of severe accident and most recent multiunit Fukushima Accident. Sheltering could be used as protective action for longer period of about 04 days. The area about 3 km of K-2/K-3 NPPs site should be evacuated and an iodine thyroid blocking agent should be taken before release up to about 14 km to prevent severe deterministic effects. Stochastic effects may be avoided or minimized by evacuating the area within about 8 km of the K-2/K-3 NPPs site. Protective actions may become more effective and cost beneficial by using current methodology as Gaussian puff model realistically represents atmospheric transport, dispersion, and disposition processes in contrast to straight-line Gaussian plume model explicitly in study area. The estimated PAZ and UPZ were found 3 km and 8 km, respectively, around K-2/K-3 NPPs which are in well agreement with IAEA Post-Fukushima Study. Therefore, current study results could be used in the establishment of emergency planning zones around K-2/K-3 NPPs.

Beschreibung: A common measure of deformation between atomic scale simulations and the continuum framework is provided and the strain tensors for multiscale simulations are defined in this paper. In order to compute the deformation gradient of any atomm, the weight function is proposed to eliminate the different contributions within the neighbor atoms which have different distances to atomm, and the weighted least squares error optimization model is established to seek the optimal coefficients of the weight function and the optimal local deformation gradient of each atom. The optimization model involves more than 9 parameters. To guarantee the reliability of subsequent parameters identification result and lighten the calculation workload of parameters identification, an overall analysis method of parameter sensitivity and an advanced genetic algorithm are also developed.

Beschreibung: The double-ended guillotine break (DEGB) of the horizontal coaxial gas duct accident is a serious air ingress accident of the high temperature gas-cooled reactor pebble-bed module (HTR-PM). Because the graphite is widely used as the structure material and the fuel element matrix of HTR-PM, the oxidation analyses of this severe air ingress accident have got enough attention in the safety analyses of the HTR-PM. The DEGB of the horizontal coaxial gas duct accident is calculated by using the TINTE code in this paper. The results show that the maximum local oxidation of the matrix graphite of spherical fuel elements in the core will firstly reach3.75⁎104 mol/m3at about 120 h, which means that only the outer 5 mm fuel-free zone of matrix graphite will be oxidized out. Even at 150 h, the maximum local weight loss ratio of the nuclear grade graphite in the bottom reflectors is only 0.26. Besides, there is enough time to carry out some countermeasures to stop the air ingress during several days. Therefore, the nuclear grade graphite of the bottom reflectors will not be fractured in the DEGB of the horizontal coaxial gas duct accident and the integrity of the HTR-PM can be guaranteed.

Beschreibung: This study presents the medical radioisotope production performance of a conceptual accelerator driven system (ADS). Lead-bismuth eutectic (LBE) is selected as target material. The subcritical fuel core is conceptually divided into ten equidistant subzones. The ceramic (natural U, Pu)O2fuel mixture and the materials used for radioisotope production (copper, gold, cobalt, holmium, rhenium, thulium, mercury, palladium, thallium, molybdenum, and yttrium) are separately prepared as cylindrical rods cladded with carbon/carbon composite (C/C) and these rods are located in the subzones. In order to obtain the flattened power density, percentages of PuO2in the mixture of UO2and PuO2in the subzones are adjusted in radial direction of the fuel zone. Time-dependent calculations are performed at 1000 MW thermal fission power (Pth) for one hour using the BURN card. The neutronic results show that the investigated ADS has a high neutronic capability, in terms of medical radioisotope productions, spent fuel transmutation and energy multiplication. Moreover, a good quasiuniform power density is achieved in each material case. The peak-to-average fission power density ratio is in the range of 1.02–1.28.

Beschreibung: In nuclear power plant construction scheduling, a project is generally defined by its dependent preparation time, the time required for construction, and its reactor installation time. The issues of multiple construction teams and multiple reactor installation teams are considered. In this paper, a hierarchical particle swarm optimization algorithm is proposed to solve the nuclear power plant construction scheduling problem and minimize the occurrence of projects failing to achieve deliverables within applicable due times and deadlines.

Beschreibung: A neural network-direct inverse control (NN-DIC) has been simulated to automatically control the power level of nuclear reactors. This method has been tested on an Indonesian pool type multipurpose reactor, namely, Reaktor Serba Guna-GA Siwabessy (RSG-GAS). The result confirmed that this method still cannot minimize errors and shorten the learning process time. A new method is therefore needed which will improve the performance of the DIC. The objective of this study is to develop a particle swarm optimization-based direct inverse control (PSO-DIC) to overcome the weaknesses of the NN-DIC. In the proposed PSO-DIC, the PSO algorithm is integrated into the DIC technique to train the weights of the DIC controller. This integration is able to accelerate the learning process. To improve the performance of the system identification, a backpropagation (BP) algorithm is introduced into the PSO algorithm. To show the feasibility and effectiveness of this proposed PSO-DIC technique, a case study on power level control of RSG-GAS is performed. The simulation results confirm that the PSO-DIC has better performance than NN-DIC. The new developed PSO-DIC has smaller steady-state error and less overshoot and oscillation.

Beschreibung: Due to increased global demand for clay, the present work involves the use of INAA for elemental analysis and pollutants concentration in clay. The samples were collected from Aswan in South Egypt. The samples were irradiated using the thermal neutrons “at the TRIGA Mainz research reactor” and at a neutron flux “of 7 × 10 n/cm s”. Twenty-six elements quantitatively and qualitatively were specified for the first time upon studying the samples. The elements determined are U, Th, Ta, Hf, Lu, Eu, Ce, Ba, Sn, Nb, Rb, Zn, Co, Fe, Cr, Sc, Sm, La, Yb, As, Ga, K, Mn, Na, Ti, and Mg. The concentrations of natural radionuclides232Th,226Ra, and40K were also calculated. Based on these concentrations, to estimate the exposure risk for using clay as raw materials in building materials, the radiation hazard indices such as radium equivalent activities, effective doses rate, and the external hazard indices have been computed. The obtained results were compared with analogous studies carried out in other countries and with the UNSCEAR reports.

Beschreibung: In sodium-cooled fast reactors (SFR), grid plate is a critical component which is made of 316 L(N) SS. It is supported on core support structure. The grid plate supports the core subassemblies and maintains their verticality. Most of the components of SFR are made of 316 L(N)/304 L(N) SS and they are in contact with the liquid-metal sodium which acts as a coolant. The peak operating temperature in SFR is 550°C. However, the self-welding starts at 500°C. To avoid self-welding and galling, hardfacing of the grid plate has become necessary. Nickel based cobalt-free colmonoy 5 has been identified as the hardfacing material due to its lower dose rate by Plasma Transferred Arc Welding (PTAW). This paper is concerned with the measurement and investigations of the effects of the residual stress generated due to thermal cycling on a scale-down physical model of the grid plate. Finite element analysis of the hardfaced grid plate model is performed for obtaining residual stresses using elastoplastic analysis and hence the results are validated. The effects of the residual stresses due to thermal cycling on the hardfaced grid plate model are studied.

Beschreibung: A hot gas duct is an indispensable component for the nuclear-process heat applications of the Very-High-Temperature Reactor (VHTR), which has to fulfill three requirements: to withstand high temperature, high pressure, and large pressure transient. In this paper, numerical investigation of pressure transient is performed for a hot gas duct under rapid depressurization. System depressurization imposes an imploding pressure differential on the internal structural elements of a hot gas duct, the structural integrity of which is susceptible to being damaged. Pressure differential and its imposed duration, which are two key factors to evaluate the damage severity of a hot gas duct under depressurization, are examined in regard to depressurization rate and insulation packing tightness. It is revealed that depressurization rate is a decisive parameter for controlling the pressure differential and its duration, whereas insulating-packing tightness has little effect on them.

Beschreibung: With the adoption of new technologies, more risk is introduced into modern society. Important decisions about new technologies tend to be made by specialists, which can lead to a mismatch of risk perception between citizens and specialists, resulting in high social cost. Using contingent valuation methods, this paper analyzes the relationship between willingness to pay (WTP) and the factors expressed through people’s image of nuclear power plants (NPP), their perception of NPP safety, and how these can be affected by their scientific background level. Results indicate that groups with a high scientific background level tend to have low risk perception level, represented through their image and safety levels. Further, the results show that mean WTP is dependent on scientific background and image levels. It is believed that these results could help decision makers address the mismatch of trust between the public and specialists in terms of new policy.

Beschreibung: The impact of thermal agitation on Doppler coefficient for Gd-bearing fuel was analyzed. It was found through the analysis that the impact increases when a small amount of Gd2O3is added to pure UO2fuel although the impact decreases for a large amount of Gd2O3. This tendency was discussed with the usage of simplified expression for the difference of Doppler coefficient. The simplified expression was used to consider the tendency, and it was revealed that the tendency mainly comes from the rapid decrement of multiplication factor and the relatively slow decrement of the magnitude of sensitivity coefficient of U-238 capture cross section at low Gd2O3concentration. Similar tendency which shows a maximum impact on Doppler coefficient at interior concentration is expected for other UO2fuel with a slight content of strong absorber. This indicates that Doppler coefficient of UO2fuel system with low content of strong absorber should be analyzed carefully by considering thermal agitation in epithermal range.

Beschreibung: The cold neutron source (CNS) system of the Open Pool Australian Light-Water (OPAL) reactor is a 20 L cryogenically cooled liquid deuterium thermosiphon system. The CNS is cooled by forced convective helium which is held at room temperature during stand-by (SO) mode and at ~20 K during normal operation (NO) mode. When helium cooling stops, the reactor is shut down to prevent the moderator cell from overheating. This computational fluid dynamics (CFD) study aims to determine whether the combined effects of conduction and natural convection would provide sufficient cooling for the moderator cell under the influence of reactor decay heat after reactor shutdown. To achieve this, two commercial CFD software packages using an identical CFD mesh were first assessed against an experimental heat transfer test of the CNS. It was found that both numerical models were valid within the bounds of experimental uncertainty. After this, one CFD model was used to simulate the thermosiphon transient condition after the reactor is shut down. Two independent shutdown conditions of different decay-heat power profiles were simulated. It was found that the natural convection and conduction cooling in the thermosiphon were sufficient for dissipating both decay-heat profiles, with the moderator cell remaining below the safe temperature of 200°C.