The lack of plant-side energy storage analysis to support nuclear power plants (NPP), has setup this research endeavor to understand the characteristics and role of specific storage technologies and the integration to an NPP. The paper provides a qualitative review of a wide range of configurations for integrating the energy storage system (ESS) to an operating NPP with pressurized water reactor (PWR). The role of ESS technologies most suitable for large-scale storage are evaluated, including thermal energy storage, compressed gas energy storage, and liquid air energy storage. The methods of integration to the NPP steam cycle are introduced and categorized as electrical, mechanical, and thermal, with a review on developments in the integration of ESS with an operating PWR. By adopting simplified off-design modeling for the steam turbines and heat exchangers, the results show the performance of the PWR steam cycle changes with respect to steam bypass rate for thermal and mechanical storage integration options. Analysis of the integrated system characteristics of proposed concepts for three different ESS suggests that certain storage technologies could support steady operation of an NPP. After having reviewed what have been accomplished through the years, the research team presents a list of possible future works.

Since the nuclear data forms a vital component in reactor physics computations, the nuclear community needs processing codes as tools for translating the Evaluated Nuclear Data Files (ENDF) to simulate nuclear-related problems such as an ACE format that is used for MCNP. Errors, inaccuracies or discrepancies in library processing may lead to a calculation that disagrees with the experimentally measured benchmark. This paper provides an overview of the processing and preparation of ENDF/B-VIII.0 incident neutron data with NECP-Atlas and NJOY codes for implementation in the MCNP code. The resulting libraries are statistically inter-compared and tested by conducting benchmark calculations, as the mutualcomparison is a source of strong feedback for further improvements in processing procedures. The database of the benchmark experiments is based on a selection taken from the International Handbook of Evaluated Criticality Safety Benchmark Experiments (ICSBEP handbook) and those proposed by Russell D. Mosteller. In general, there is quite good agreement between the NECP-Atlas1.2 and NJOY21^(1.0.0.json) results with no substantial differences, if the correct input parameters are used.

The objective of this research is to the use of americium (AmO₂) as a burnable absorber effectively instead of conventional gadolinium (Gd₂O₃) for VVER-1200 reactor by analyzing its impacts on reactivity, power peaking factor (PPF), safety factor, and quality of the spent fuel. The assembly is burned to 60 GWd/t by using SRAC-2006 code and JENDL-4.0 data library for finding the optimum amount and effective way of using AmO₂ as a burnable absorber. From these studies, it is found that AmO₂ can decrease the excess reactivity like Gd₂O₃ without changing the criticality life span and enrichment of ²³⁵U. A homogeneous mixture of the 0.20% AmO₂+ 4.95% enriched UO₂ fuel rod (model MF-4) decreases the PPF than the reference assembly. The use of AmO₂+UO₂ in the integral burnable absorber (IBA) rod or the outer layer could also decrease the PPF up to 10 GWd/t but increases rapidly after 30 GWd/t, which could be a safety threat. The fuel temperature coefficient and void coefficient of the model MF-4 are the same as the reference assembly. In addition, 22% of initially loaded Am are burning effectively and contributing to the power production.

Direct-contact heat transfer of a single saturated droplet upon colliding with a heated wall in the regime of film boiling was experimentally investigated using high-resolution infrared thermometry technique. This technique provides transient local wall heat flux distributions during the entire collision period. In addition, various physical parameters relevant to the mechanistic modelling of these phenomena can be measured. The obtained results show that when single droplets dynamically collide with a heated surface during film boiling above the Leidenfrost point temperature, typically determined by droplet collision dynamics without considering thermal interactions, small spots of high heat flux due to localized wetting during the collision appear as increasing We_n. A systematic comparison revealed that existing theoretical models do not consider these observed physical phenomena and have lacks in accurately predicting the amount of direct-contact heat transfer. The necessity of developing an improved model to account for the effects of local wetting during the direct-contact heat transfer process is emphasized.

To improve safety analysis technology for a nuclear reactor containment considering an interaction between a reactor coolant system (RCS) and containment, this study aims at an experimental investigation on the integrated simulation of the RCS and containment, with an integral effect test facility, ATLAS-CUBE. For a realistic simulation of a pressure and temperature (P/T) transient, the containment simulation vessel was designed to preserve a volumetric scale equivalently to the RCS volume scale of ATLAS. Three test cases for a steam line break (SLB) transient were conducted with variation of the initial condition of the passive heat sink or the steam flow direction. The test results indicated a stratified behavior of the steam-gas mixture in the containment following a high-temperature steam injection in prior to the spray injection. The test case with a reduced heat transfer on the passive heat sink showed a faster increase of the P/T inside the containment. The effect of the steam flow direction was also investigated with respect to a multi-dimensional distribution of the local heat transfer on the passive heat sink. The integral effect test data obtained in this study will contribute to validating the evaluation methodology for mass and energy (M/E) and P/T transient of the containment.

The wall film-wise condensation plays an important role in the heat transfer processes of heat exchangers, refrigerators, and air conditioner. In the field of nuclear engineering, steam condensation is often utilized in safety systems to remove the core decay heat under both transient and accident conditions. In particular, passive containment cooling system (PCCS), are designed to ensure containment safety under severe accident conditions. A computational fluid dynamics (CFD) scale analysis has been conducted to calculate the heat transfer rate of the PCCS. However, despite the increase in computing power, there are challenges in the long-term transient simulation of containment using CFD scale codes. In this study, a heat structure coupling between the CFD and system analysis codes was performed to efficiently analyze PCCS. In addition, the component unstructured program for interfacial dynamics (CUPID) was improved to analyze the condensation behavior of ternary gas mixtures. Thereafter, the condensation heat transfer on the primary side was calculated using the improved CUPID and CFD code, whereas that on the secondary side was simulated using MARS. Both the coupled codes were validated against the CONAN facility database. Finally, conjugate heat transfer simulations with wall condensation in the presence of non-condensable gases were appropriately performed.

Passive containment cooling systems (PCCSs) have been actively studied to improve the inherent safety of nuclear power plants. Hered, we present two concepts, open-loop PCCS (OL-PCCS) and closed-loop PCCS (CL-PCCS), applicable to the PWR with a concrete-type containment. We analyzed the heat-removal performance and flow instability of these PCCS concepts using the GOTHIC code. In both cases, PCCS performance improved when a passive containment cooling heat exchanger (PCCX) was installed in the lower part of the containment building. The OL-PCCS was found to be superior in terms of heat-removal performance. However, in terms of flow instability, the OL-PCCS was more vulnerable than the CL-PCCS. In particular, the possibility of flow instability was higher when the PCCX was installed in the upper part of the containment. Therefore, the installation location of the OL-PCCS should be restricted to minimize flow instability. Conversely, a CL-PCCS can be installed without any positional restriction by adjusting the initial system pressure within the loop, which eliminates flow instability. These results could be used as base data for the thermo-hydraulic evaluation of PCCS in PWR with a large dry concrete-type containment.

A valve assembly used in nuclear power plants must be qualified and supervised. New technical standards such as ASME QME-1 2007 particularly require detailed qualification using experiment and analysis. Particularly, diagnostic tests and engineering studies are required for qualification of ASME QME-1 2007. Among these studies, the research on the measurement of friction coefficient and packing stress is important. The irregular change of packing stress along the stroke distance occurs because of the abnormal phenomenon, which must be found and studied with quantitative methods. Packing stress should be analyzed conservatively through experimentation and analysis. In this study, various formulas were applied to measure and calculate coefficient of friction and packing stress. This study can be used in relation to qualification and supervision of packing materials. And the calculation using static diagnosis test can be used to find the packing frictional force in dynamic diagnosis test with flow pressure in a pipe. This study has made it possible to reliably consider packing frictional force generated in a valve body. And so, it is believed that more margin can be secured when evaluating the capacity of valve actuator by applying the accurate frictional force generated in the valve assembly.

In the last decade, the best estimate plus uncertainty methodologies in nuclear technology and nuclear power plant design have become a trending topic in the nuclear field. Since BEPU was allowed for licensing purposes by the most important regulator bodies, different uncertainty assessment methods have become popular, overall non-parametric methods. While non-parametric tolerance regions can be well stated and used in uncertainty quantification for licensing purposes, the propagation of the uncertainty through different codes (multi-scale, multiphysics) in cascade needs a better depiction of uncertainty than the one provided by the tolerance regions or a probability distribution. An alternative method based on the parametric or distributional probability boxes is used to perform uncertainty quantification and propagation regarding statistic uncertainty from one code to another. This method is sample-size independent and allows well-defined tolerance intervals for uncertainty quantification, manageable for uncertainty propagation. This work characterizes the distributional p-boxes behavior on uncertainty quantification and uncertainty propagation through nested random sampling.

Nuclear emergency preparedness and response is an essential part to ensure the safety of nuclear power plant (NPP). Key support technologies of nuclear emergency decision-making usually consist of accident diagnosis, source term estimation, accident consequence assessment, and protective action recommendation. Source term estimation is almost the most difficult part among them. For example, bad communication, incomplete information, as well as complicated accident scenario make it hard to determine the reactor status and estimate the source term timely in the Fukushima accident. Subsequently, it leads to the hard decision on how to take appropriate emergency response actions. Hence, this paper aims to develop a method for rapid source term estimation to support nuclear emergency decision making in pressurized water reactor NPP. The method aims to make our knowledge on NPP provide better support nuclear emergency. Firstly, this paper studies how to build a Bayesian network model for the NPP based on professional knowledge and engineering knowledge. This paper presents a method transforming the PRA model (event trees and fault trees) into a corresponding Bayesian network model. To solve the problem that some physical phenomena which are modeled as pivotal events in level 2 PRA, cannot find sensors associated directly with their occurrence, a weighted assignment approach based on expert assessment is proposed in this paper. Secondly, the monitoring data of NPP are provided to the Bayesian network model, the real-time status of pivotal events and initiating events can be determined based on the junction tree algorithm. Thirdly, since PRA knowledge can link the accident sequences to the possible release categories, the proposed method is capable to find the most likely release category for the candidate accidents scenarios, namely the source term. The probabilities of possible accident sequences and the source term are calculated. Finally, the prototype software is checked against several sets of accident scenario data which are generated by the simulator of AP1000-NPP, including large loss of coolant accident, loss of main feedwater, main steam line break, and steam generator tube rupture. The results show that the proposed method for rapid source term estimation under nuclear emergency decision making is promising.

The frequency of reactor coolant leakage is expected to increase over the lifetime of a nuclear power plant owing to degradation mechanisms, such as flow-acceleration corrosion and stress corrosion cracking. When loss of coolant accidents (LOCAs) occur, several parameters change rapidly depending on the size and location of the cracks. In this study, leak flow during LOCAs is predicted using a deep fuzzy neural network (DFNN) model. The DFNN model is based on fuzzy neural network (FNN) modules and has a structure where the FNN modules are sequentially connected. Because the DFNN model is based on the FNN modules, the performance factors are the number of FNN modules and the parameters of the FNN module. These parameters are determined by a least-squares method combined with a genetic algorithm; the number of FNN modules is determined automatically by cross checking a fitness function using the verification dataset output to prevent an overfitting problem. To acquire the data of LOCAs, an optimized power reactor-1000 was simulated using a modular accident analysis program code. The predicted results of the DFNN model are found to be superior to those predicted in previous works. The leak flow prediction results obtained in this study will be useful to check the core integrity in nuclear power plant during LOCAs. This information is also expected to reduce the workload of the operators.

Metaheuristic algorithms can work well in solving or optimizing problems, especially those that require approximation or do not have a good analytical solution. Particle swarm optimization (PSO) is one of these algorithms. The response quality of these algorithms depends on the objective function and its regulated parameters. The nonlinear nature of the pressurized light-water nuclear reactor (PWR) dynamics is a significant target for PSO. The two-point kinetics model of this type of reactor is used because of fission products properties. The proportional-integral-derivative (PID) controller is intended to control the power level of the PWR at a short-time transient. The absolute error (IAE), integral of square error (ISE), integral of time-absolute error (ITAE), and integral of time-square error (ITSE) objective functions have been used as performance indexes to tune the PID gains with PSO. The optimization results with each of them are evaluated with the number of function evaluations (NFE). All performance indexes achieve good results with differences in the rate of over/under-shoot or convergence rate of the cost function, in the desired time domain.

The neutronic analysis of Helium Cooled Ceramic Reflector (HCCR) breeding blankets has been performed using the 3D Monte Carlo code MCNPX and ENDF nuclear data library. This study aims to reduce ⁶Li percentage in the breeder zones as much as possible ensuring tritium self-sufficiency. This work is devoted to investigating the effect of ⁶Li percentage on the HCCR breeding blanket's neutronic parameters, such as neutron flux and spectrum, Tritium Breeding Ratio (TBR), nuclear power density, and energy multiplication factor. In the ceramic breeders at the saturated thickness, increasing the enrichment of ⁶Li reduces its share in the tritium production. Therefore, ceramic breeders typically use lower enriched Li from 30% to 60%. The investigation of neutronic analysis in the suggested geometry shows that using 60% ⁶Li in Li₂TiO₃ can yield acceptable TBR and energy deposition results, which would be economically feasible.

The exchange behaviors of hydrogen isotopes between protonated lithium metal compounds and deuterated water or tritiated water were investigated. The various protonated lithium metal compounds were prepared by acid treatment of lithium metal compounds with different crystal structures and metal compositions. The protonated lithium metal compounds could more effectively reduce the deuterium concentration in water compared with the corresponding pristine lithium metal compounds. The H⁺ in the protonated lithium metal compounds was speculated to be more readily exchangeable with hydrons in the aqueous solution compared with Li⁺ in the pristine lithium metal compounds, and the exchanged heavier isotopes were speculated to be more stably retained in the crystal structure compared with the light protons. When the tritiated water (157.7 kBq/kg) was reacted with the protonated lithium metal compounds, the protonated lithium manganese nickel cobalt oxide was found to adsorb and retain twice as much tritium (163.9 Bq/g) as the protonated lithium manganese oxide (69.9 Bq/g) and the protonated lithium cobalt oxide (75.1 Bq/g) in the equilibrium state.

The thermal stability and crystallization behaviors of La₂O₃ containing B₂O₃-CaO-Al₂O₃ glass waste forms were investigated to evaluate the stability of waste form during emergencies in deep geological disposal. For glasses containing 15% La₂O₃, LaBO₃ phases were observed as major crystals from 780 ℃ and exhibited needlelike structures. Al, Ca, and O were homogeneously distributed throughout the entire specimen, while some portions of B and La were concentrated in some parts. By differential thermal analysis at various heating rates, the activation energy for grain growth and the crystallization rate of LaBO₃ were calculated to be 12.6 kJ/mol and 199.5 kJ/mol, respectively. These values are comparable to other waste forms being developed for the same purpose.

The high-energy milling is one of the most extended techniques to produce Oxide dispersion strengthened (ODS) powder steels for nuclear applications. The consequences of the high energy mill process on the final powders can be measured by means of deformation level, size, morphology and alloying degree. In this work, an ODS ferritic steel, Fe-14Cr-5Al-3W-0.4Ti-0.25Y₂O₃-0.6Zr, was fabricated using two different mechanical alloying (MA) conditions (M_std and M_act) and subsequently consolidated by Spark Plasma Sintering (SPS). Milling conditions were set to evidence the effectivity of milling by changing the revolutions per minute (rpm) and dwell milling time. Differences on the particle size distribution as well as on the stored plastic deformation were observed, determining the consolidation ability of the material and the achieved microstructure. Since recrystallization depends on the plastic deformation degree, the composition of each particle and the promoted oxide dispersion, a dual grain size distribution was attained after SPS consolidation. M_act showed the highest areas of ultrafine regions when the material is consolidated at 1100 ℃. Microhardness and small punch tests were used to evaluate the material under room temperature and up to 500 ℃. The produced materials have attained remarkable mechanical properties under high temperature conditions.

The effect of thermal aging at 475 ℃ and 750 ℃ of Z3CN20.09M cast duplex stainless steel (CDSS) on microstructure, mechanical and intergranular corrosion properties were investigated by transmission electron microscope (TEM), nano indenter, scanning electron microscope (SEM) and corrosion fatigue test system. The result indicated that the spinodal decomposition and G precipitated were occurred after aged at 475 ℃, as well as sigma precipitated at 750 ℃. The microstructure degeneration of ferrite was saturated after aged for 2000h and 200 h at 475 ℃ and 750 ℃ respectively. The mechanical properties, intergranular corrosion resistance and corrosion fatigue lives were continuing deteriorated with increasing the aging time at both temperatures. The difference of the degeneration mechanisms of Z3CN20.09M CDSS aged at 475 ℃ and 750 ℃ was analyzed.

Low cycle fatigue behaviors of SA508-3 low-alloy steel were investigated in room-temperature air, high-temperature air and in light water reactor (LWR) water environments. The fatigue mean curve and design curve for the low-alloy steel are developed based on the fatigue data in room-temperature and high-temperature air. The environmental fatigue model for low-alloy steel is developed by the environmental fatigue correction factor (F_en) methodology based on the fatigue data in LWR water environments with the consideration of effects of strain rate, temperature, and dissolved oxygen concentration on the fatigue life.

The prediction of irradiation-induced transition temperature shift for RPV steels is an important method for long term operation of nuclear power plant. Based on the irradiation embrittlement data, an irradiation-induced transition temperature shift prediction model is developed with machine learning method XGBoost. Then the residual, standard deviation and predicted value vs. measured value analysis are conducted to analyze the accuracy of this model. At last, Cu content threshold and saturation values analysis, temperature dependence, Ni/Cu dependence and flux effect are given to verify the reliability. Those results show that the prediction model developed with XGBoost has high accuracy for predicting the irradiation embrittlement trend of RPV steel. The prediction results are consistent with the current understanding of RPV embrittlement mechanism.

The results of effective irradiation swelling in a wide range of burnup levels are numerically obtained for an inert matrix fuel, which are verified with DART model. The fission gas swelling of fuel particles is calculated with a mechanistic model, which depends on the external hydrostatic pressure. Additionally, irradiation and thermal creep effects are included in the inert matrix. The effects of matrix creep strains, external hydrostatic pressure and temperature on the effective irradiation swelling are investigated. The research results indicate that (1) the above effects are coupled with each other; (2) the matrix creep effects at high temperatures should be involved; and (3) ranged from 0 to 300 MPa, a remarkable dependence of external hydrostatic pressure can be found. Furthermore, an explicit multi-variable mathematic model is established for the effective irradiation swelling, as a function of particle volume fraction, temperature, external hydrostatic pressure and fuel particle fission density, which can well reproduce the finite element results. The mathematic model for the current volume fraction of fuel particles can help establish other effective performance models.

Within the frame of the EMPIrE test, four monolithic mini-plates were irradiated in the ATR reactor. In two of them, the monolithic U(Mo) foil had been PVD-coated with Zr before the plate manufacturing. Extensive microstructural characterizations were performed on a fresh archive mini-plate, using Optical Microscopy (OM), Scanning Electron Microscopy (SEM) combined with Energy Dispersive Spectroscopy (EDS), Electron Backscattered Diffraction (EBSD) and Focused Ion Beam (FIB)/Transmission Electron Microscopy (TEM) with nano EDS. A particular attention was paid to the examination of the U(Mo) foil, the PVD coating, the cladding/Zr and Zr/U(Mo) interfaces. The Zr coating has a thickness around 15 ㎛. It has a columnar microstructure and appears dense. The cohesion of the cladding/Zr and Zr/U(Mo) interfaces seems to be satisfactory. An almost continuous layer with a thickness of the order of 100-300 nm is present at the cladding/Zr interface and corresponds to an oxidized part of the Zr coating. At the Zr/U(Mo) interface, a thin discontinuous layer is observed. It could correspond to locally oxidized U(Mo). This work provides a basis for interpreting the results of characterizations on EMPIrE irradiated plates.

We report the development of a gamma-ray imaging device, named Large-Area Hybrid Gamma Imager (LAHGI), featuring high imaging sensitivity and good imaging resolution over a broad energy range. A hybrid collimation method, which combines mechanical and electronic collimation, is employed for a stable imaging performance based on large-area scintillation detectors for high imaging sensitivity. The system comprises two monolithic position-sensitive NaI(Tl) scintillation detectors with a crystal area of 27 × 27 cm² and a tungsten coded aperture mask with a modified uniformly redundant array (MURA) pattern. The performance of the system was evaluated under several source conditions. The system showed good imaging resolution (i.e., 6.0-8.9° FWHM) for the entire energy range of 59.5-1330 keV considered in the present study. It also showed very high imaging sensitivity, successfully imaging a 253 µCi ¹³⁷Cs source located 15 m away in 1 min; this performance is notable considering that the dose rate at the front surface of the system, due to the existence of the ¹³⁷Cs source, was only 0.003 µSv/h, which corresponds to ~3% of the background level.

A rectangular-shaped PET system with an adjustable gantry (AGPET) has been developed for imaging small animals. The AGPET system employs a new depth of interaction (DOI) method using a depth dependent reflector patterns and a new digital time pickoff method based on the pulse reconstruction method. To evaluate the performance of the AGPET, timing resolution, intrinsic spatial resolution and point source images were acquired. The timing resolution and intrinsic spatial resolution were measured using two detector modules and Na-22 gamma source. The PET images were acquired in two field of view (FOV) sizes, 30 mm and 90 mm, to demonstrate the characteristic of the AGPET. As a result of in the experiment results, the timing resolution was 0.9 ns using the pulse reconstruction method based on the bi-exponential model. The intrinsic spatial resolution was an average of 1.7 mm and the spatial resolution of PET images after DOI correction was 2.08 mm and 2.25 mm at the centers of 30 mm and 90 mm FOV, respectively. The results show that the proposed AGPET system provided higher sensitivity and resolution for small animal imaging.

The gamma-ray exposure doses in decontamination operation after a nuclear accident were evaluated with a consideration of various geometrical conditions and specific gamma-ray energies. The calculation domain is organized with three residence types and each form is divided into two kinds of geometrical arrangements. The position-wise air KERMA values were calculated with an assumption of evenly distributed gamma-ray source based on Monte Carlo radiation transport analysis using the MCNP code. The radioactivity is initially set to be unity to be multiplied by the deposition value measured in the actual accident condition. The workforce data set depending on the target object was determined by modifying the Fukushima report. The external exposure doses for decontamination workers were derived from the calculated KERMA values and the workforce analysis. These results can be used to efficiently determine the workforce required by the characteristics of the area and the structure to be decontaminated within the dose limits.

Gamma radiation shielding features for three series of binary alloys identified as (Pb-Sn), (Pb-Zn), and (Zn-Sn) have been investigated. The mass attenuation coefficients (µ/ρ) for the selected alloys were simulated using the MCNP-5 code in the energy range between 0.01 and 15 MeV. Moreover, the (µ/ρ) values were computed using WinXCOM database in the same energy range to validate the simulation results. Results reveal a good agreement between the simulated and computed values. The half value layer (HVL), mean free path (MFP), effective atomic number (Z_eff) and exposure buildup factor (EBF) were evaluated for the selected binary alloys. Results showed that the PS1, PZ1, and ZS2 alloys have the best shielding parameters and better than the commercially standard and available radiation shielding materials. Therefore, the investigated alloys can be used as effective radiation shielding materials against gamma ray with energies between 0.01 and 15 MeV.

Valve is widely used in the various industry areas to adjust and control the flow. Cavitation frequently takes place and sometimes is inevitable in various types of valve to cause the erosion damage. Therefore, how to control and minimize the effect of cavitation is still an important topic. This study numerically investigates the cavitating flow in a control valve with a perforated cage. The effects of some parameters on the cavitation are discussed. It also discusses to use the throttling steps to govern the cavitating flow. The results show that the opening degree of valve and the length of downstream divergent connection both influence the cavitation. The increase of the divergent length reinforces the cavitation. And the larger the opening of valve is, the intenser the cavitation is and the more vapor is present. The more throttling steps are helpful to decrease the cavitation.

This paper aims to evaluate the structural dynamic responses and damage/failure of the nuclear fuel reprocessing plant under the free drop impact of spent fuel cask (SFC) and fuel assembly (FA) during the on-site transportation. At the present Part I of this paper, the large-scale SFC model free drop test and the corresponding numerical simulations are performed. Firstly, a composite target which is composed of the protective structure, i.e., a thin RC plate (representing the inverted U-shaped slab in the loading shaft) and/or an autoclaved aerated concrete (AAC) blocks sacrificial layer, as well as a thick RC plate (representing the bottom slab in the loading shaft) is designed and fabricated. Then, based on the large dropping tower, the free drop test of large-scale SFC model with the mass of 3 t is carried out from the height of 7 m-11 m. It indicates that the bottom slab in the loading shaft could not resist the free drop impact of SFC. The composite protective structure can effectively reduce the damage and vibrations of the bottom slab, and the inverted U-shaped slab could relieve the damage of the AAC blocks layer dramatically. Furthermore, based on the finite element (FE) program LS-DYNA, the corresponding refined numerical simulations are performed. By comparing the experimental and numerical damage and vibration accelerations of the composite structures, the present adopted numerical algorithms, constitutive models and parameters are validated, which will be applied in the further assessment of drop impact effects of full-scale SFC and FA on prototype nuclear fuel reprocessing plant in the next Part II of this paper.

The purpose of this study is to investigate the efficiency of various structural modeling schemes for evaluating seismic performances and fragility of the reactor containment building (RCB) structure in the advanced power reactor 1400 (APR1400) nuclear power plant (NPP). Four structural modeling schemes, i.e. lumped-mass stick model (LMSM), solid-based finite element model (Solid FEM), multi-layer shell model (MLSM), and beam-truss model (BTM), are developed to simulate the seismic behaviors of the containment structure. A full three-dimensional finite element model (full 3D FEM) is additionally constructed to verify the previous numerical models. A set of input ground motions with response spectra matching to the US NRC 1.60 design spectrum is generated to perform linear and nonlinear time-history analyses. Floor response spectra (FRS) and floor displacements are obtained at the different elevations of the structure since they are critical outputs for evaluating the seismic vulnerability of RCB and secondary components. The results show that the difference in seismic responses between linear and nonlinear analyses gets larger as an earthquake intensity increases. It is observed that the linear analysis underestimates floor displacements while it overestimates floor accelerations. Moreover, a systematic assessment of the capability and efficiency of each structural model is presented thoroughly. MLSM can be an alternative approach to a full 3D FEM, which is complicated in modeling and extremely time-consuming in dynamic analyses. Specifically, BTM is recommended as the optimal model for evaluating the nonlinear seismic performance of NPP structures. Thereafter, linear and nonlinear BTM are employed in a series of time-history analyses to develop fragility curves of RCB for different damage states. It is shown that the linear analysis underestimates the probability of damage of RCB at a given earthquake intensity when compared to the nonlinear analysis. The nonlinear analysis approach is highly suggested for assessing the vulnerability of NPP structures.

The Multi-Purpose Overload Robot (CMOR) is a key subsystem of China Fusion Engineering Test Reactor (CFETR) remote handling system. Due to the long cantilever and large loads of the CMOR, it has a large rigid-flexible coupling deformation that results in a poor position accuracy of the end-effector. In this study, based on the Levenberg-Marquardt algorithm, the spatial grid, and the linearized variable load principle, a variable parameter compensation model was designed to identify the parameters of the CMOR's kinematics models under different loads and at different poses so as to improve the trajectory tracking accuracy. Finally, through Adams-MATLAB/Simulink, the trajectory tracking accuracy of the CMOR's rigid-flexible coupling model was analyzed, and the end position error exceeded 0.1 m. After the variable parameter compensation model, the average position error of the end-effector became less than 0.02 m, which provides a reference for CMOR error compensation.

The core shroud is one of the most important internal components of the reactor vessel internals because it meets the neutron fluence directly emitted by the nuclear fuel. In particular, dynamic effects for an earthquake should be evaluated with respect to the neutron irradiation flux. As a prerequisite to this study, simplified and detailed finite element models are developed for the core shroud using the ANSYS Design Parametric Language. Using the El Centro earthquake, seismic analyses are performed for the simplified and detailed core shroud models. Modal characteristics are obtained and their results are used for a time history analysis. Response spectrum analyses are also performed to access the degree of seismic excitation. The results of these analyses are compared to investigate the response characteristics between the simplified and detailed core shroud models from the time history and response spectrum analyses.

The purpose of this study is to reduce seismic responses of an actual nuclear piping system using a tuned mass damper (TMD) device. A numerical piping model was developed and validated based on shaking table test results with actual nuclear piping. A TMD for nuclear piping was newly devised in this work. A TMD shape design suitable for nuclear piping systems was conducted, and its operating performance was verified after manufacturing. The response reduction performance of the developed TMD under earthquake loading on actual piping was investigated. Results confirmed that, on average, seismic response reduction rates of 34% in the maximum acceleration response, 41% in the root mean square acceleration response, and 57% in the spectral acceleration response were shown through the TMD application. This developed TMD operated successfully within the seismic response reduction rate of existing TMD optimum design values. Therefore, the developed TMD and dynamic interpretation help improve the nuclear piping's seismic performance.

The environmental effects of China's nuclear energy consumption in a dynamic framework of the pollution haven hypothesis are examined. This study uses a dynamic autoregressive distributed lag simulation approach. Empirical evidence confirms that the pollution haven hypothesis does not exist for China; i.e., foreign direct investment plays a promising role in influencing environmental outcomes. Furthermore, empirical results concluded positive contribution of nuclear energy in pollution mitigation. From the results it is expected that encouraging foreign investment to increase generation of nuclear energy would benefit environmental quality by reducing CO₂ emissions.

This paper proposes using sodium-cooled fast reactor technologies for use in hydrogen vapor methane (SMR) modification. Using three independent energy rings in the Russian BN-600 fast reactor, steam is generated in one of the steam-generating cycles with a pressure of 13.1 MPa and a temperature of 505 ℃. The reactor's second energy cycles can increase the gas-steam mixture's temperature to the required amount for efficient correction. The 620 ton/hr 540 ℃ steam generated in this cycle is sufficient to supply a high-temperature synthesis current source (700 ℃), which raises the steam-gas mixture's temperature in the reactor. The proposed technology provides a high rate of hydrogen production (approximately 144.5 ton/hr of standard H₂), also up to 25% of the original natural gas, in line with existing SMR technology for preparing and heating steam and gas mixtures will be saved. Also, exergy analysis results show that the plant's efficiency reaches 78.5% using HTR heat for combined hydrogen and power generation.

A method for preparing "reagent" for radioactive methyliodide vapors production using an isotopic exchange reaction has been developed. Based on the obtained data of the isotopic exchange efficiency and hydraulic resistance, white fused alumina (700-840 ㎛) was selected as the carrying agent material for "reagent" production. The radioiodine isotopic exchange dependences on such parameters as temperature, gas flow velocity, and the methyliodide concentration in it were determined. Optimal conditions have been selected to achieve 85% of the isotopic exchange rate in 1 h of the experiment. The obtained data allowed to develop an approach to the test of iodine filters for nuclear power plants and to determine their efficiency.

We used the GEANT4 Monte Carlo MC Toolkit to simulate carbon ion beams incident on water, tissue, and bone, taking into account nuclear fragmentation reactions. Upon increasing the energy of the primary beam, the position of the Bragg-Peak transfers to a location deeper inside the phantom. For different materials, the peak is located at a shallower depth along the beam direction and becomes sharper with increasing electron density NZ. Subsequently, the generated depth dose of the Bragg curve is then benchmarked with experimental data from GSI in Germany. The results exhibit a reasonable correlation with GSI experimental data with an accuracy of between 0.02 and 0.08 cm, thus establishing the basis to adopt MC in heavy-ion treatment planning. The Kolmogorov-Smirnov K-S test further ascertained from a statistical point of view that the simulation data matched the experimentally measured data very well. The two-dimensional isodose contours at the entrance were compared to those around the peak position and in the tail region beyond the peak, showing that bone produces more dose, in comparison to both water and tissue, due to secondary doses. In the water, the results show that the maximum energy deposited per fragment is mainly attributed to secondary carbon ions, followed by secondary boron and beryllium. Furthermore, the number of protons produced is the highest, thus making the maximum contribution to the total dose deposition in the tail region. Finally, the associated spectra of neutrons and photons were analyzed. The mean neutron energy value was found to be 16.29 MeV, and 1.03 MeV for the secondary gamma. However, the neutron dose was found to be negligible as compared to the total dose due to their longer range.