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In this study, an airborne alpha detection system, which consists of a passivated implanted planar silicon (PIPS) detector and an air filter, was developed. A collimator applied to the alpha detection system showed an enhancement in resolution and a degradation in detection efficiency. The resolution and detection efficiency were compared and analyzed to evaluate the performance of the collimator. Thus, the resolution was found to be more important than the efficiency as a determining factor of the detection system performance, from the viewpoint of radionuclide identification. The performance was evaluated on three properties of the collimator: hole shape, hole length, and the ratio between the hole and frame pitches. From the hole shape performance evaluation, a hexagonal collimator showed the highest resolution. Further, the collimator with a hole pitch of 14 mm was found to have the highest resolution while that with a frame pitch of 4–6 mm (i.e., 1.2–1.4 times longer than the hole pitch) showed the highest resolution.

In this study, an electron-scattering device was fabricated to practically use the ultra-high dose rate electron beams for the FLASH preclinical research in Dongnam Institute of Radiological and Medical Sciences. The Dongnam Institute of Radiological and Medical Sciences has been involved in the investigation of linear accelerators for preclinical research and has recently implemented FLASH electron beams. To determine the geometry of the scattering device for the FLASH preclinical research with a 6-MeV linear accelerator, the Monte Carlo N-particle transport code was exploited. By employing the fabricated scattering device, the off-axis and depth dose distributions were measured with radiochromic films. The generated mean energy of electron beams via the scattering device was 4.3 MeV, and the symmetry and flatness of the off-axis dose distribution were 0.11% and 2.33%, respectively. Finally, the doses per pulse were obtained as a function of the source to surface distance (SSD); the measured dose per pulse varied from 4.0 to 0.2 Gy/pulse at an SSD range of 20–90 cm. At an SSD of 30 cm with a 100-Hz repetition rate, the dose rate was 180 Gy/s, which is sufficient for the preclinical FLASH studies.

The review summarizes the published data on the widely applied electron-beam, laser-beam, as well as resistance upset, projection, and spot welding of zirconium alloys for nuclear applications. It provides the results of their analysis to identify common patterns in this area. Great attention has been paid to the quality requirements, the edge preparation, up-to-date equipment, process parameters, as well as post-weld treatment and processing. Also, quality control and weld repair methods have been mentioned. Finally, conclusions have been drawn about a significant gap between the capabilities of advanced welding equipment to control the microstructure and, accordingly, the properties of welded joints of the zirconium alloys and existing algorithms that enable to realize them in the nuclear industry. Considering the ever-increasing demands on the high-burnup accident tolerant nuclear fuel assemblies, great efforts should be focused on the improving the welding procedures by implementing predefined heat input cycles. However, a lot of research is required, since the number of possible combinations of the zirconium alloys, designs and dimensions of the joints dramatically exceeds the quantity of published results on the effect of the welding parameters on the properties of the welds.

This work reported the radiation shielding characteristic of the bismuth titanium vanadium sodium tellurite glass system. The density of the specially-developed glass samples was increased from 2.21 to 4.01 g/cm3 with the addition of Bi2O3, despite the fact the molar volume is decease within 85.43–54.79 cm3/mol. The WinXcom program was used to approximate the effect of Bi2O3 on the gamma radiation shielding parameters of bismuth titanium vanadium sodium tellurite glasses. The μm values decrease with the increase of Bi2O3 concentration. The computed data shows that the glass sample with 20 mol.% of Bi2O3 content has the greatest radiation attenuation performance in comparison to other selected glasses. The Bi2O3–TiO2–V2O5–Na2O–TeO2 glass system shows excellent neutron shielding material with high long-term light transmittance and discharge resistance and could be potentially used as transparent radiation-resistant shielding glass applications.

Although there are still controversial opinions and uncertainty on application of SiCf/SiC composite cladding as next-generation cladding material for its great oxidation resistance in high temperature steam environment and other outstanding advantages, it cannot deny that SiCf/SiC cladding is a potential accident tolerant fuel (ATF) cladding with high research priority and still in the engineering design stage for now. However, considering its disadvantages, such as low irradiated thermal conductivity, ductility that barely not exist, further evaluations of its in-pile behaviors are still necessary. Based on the self-developed code we recently updated, relevant thermohydraulic and mechanical models in FROBA-ATF were applied to simulate the cladding behaviors under normal and accident conditions in this paper. Even through steady-state performance analysis revealed that this kind of cladding material could greatly reduce the oxidation thickness, the thermal performance of UO2–SiC was poor due to its low in-pile thermal conductivity and creep rate. Besides, the risk of failure exists when reactor power decreased. With geometry optimization and dopant addition in pellets, the steady-state performance of UO2–SiC was enhanced and the failure risk was reduced. The thermal and mechanical performance of the improved UO2–SiC was further evaluated under Loss of coolant accident (LOCA) and Reactivity Initiated Accident (RIA) conditions. Transient results showed that the optimized ATF had better thermal performance, lower cladding hoop stress, and could provide more coping time under accident conditions.

Most thickness measurement techniques using X-ray radiation are unsuitable in field processes involving fast-moving organic films. Herein, we propose a Compton scattering X-ray radiation method, which probes the light elements in organic materials, and a new simple, non-destructive, and non-contact calibration-free real-time film thickness measurement technique by setting up a bench-top X-ray thickness measurement system simulating a field process dealing with thin flexible organic films. The use of X-ray fluorescence and Compton scattering X-ray radiation reflectance signals from films in close contact with a roller produced accurate thickness measurements. In a high-thickness range, the contribution of X-ray fluorescence is negligible, whereas that of Compton scattering is negligible in a low-thickness range. X-ray fluorescence and Compton scattering show good correlations with the organic film thickness (R2 = 0.997 and 0.999 for X-ray fluorescence and Compton scattering, respectively, in the thickness range 0–0.5 mm). Although the sensitivity of X-ray fluorescence is approximately 4.6 times higher than that of Compton scattering, Compton scattering signals are useful for thick films (e.g., thicker than ca. 1–5 mm under our present experiment conditions). Thus, successful calibration-free thickness monitoring is possible for fast-moving films, as demonstrated in our experiments.

In this work, we propose and design a GaN-based diode with a p-doped GaN (p-GaN) multi-well structure for high efficiency betavoltaic (BV) cells. The short-circuit current density (JSC) and open-circuit voltage (VOC) of the devices were investigated with variations of parameters such as the doping concentration, height, width of the p-GaN well region, well-to-well gap, and number of well regions. The JSC of the device was significantly improved by a wider depletion area, which was obtained by applying the multi-well structure. The optimized device achieved a higher output power density by 8.6% than that of the conventional diode due to the enhancement of JSC. The proposed device structure showed a high potential for a high efficiency BV cell candidate.

Spacer grid with mixing vane had been widely used in nuclear reactor core. One of the main feather of spacer grid with mixing vane was that strong swirl flow was formed after the spacer grid. The swirl flow not only changed the bubble generation in the near wall field, but also affected the bubble behaviors in the center region of the subchannel. The interaction between bubble and the swirl flow was one of the basic phenomena for the two phase flow modeling in fuel assembly. To obatin better understanding on the bubble behaviors in swirl flow, full three dimension numerical simulations were conducted in the present paper. The swirl flow was assumed in the cylindral calculation domain. The bubble interface was captured by Volume Of Fluid (VOF) method. The properties of saturated water and steam at different pressure were applied in the simulation. The bubble trajectory, motion, shape and force were obtained based on the bubble parameters captured by VOF. The simulation cases in the present study included single bubble with different size, at different angular velocity conditions and at different pressure conditions. The results indicated that bubble migrated to the center in swirl flow with spiral motion type. The lateral migration was mainly related to shear stress magnitude and bubble size. The bubble moved toward the center with high velocity when the swirl magnitude was high. The largest bubble had the highest lateral migration velocity in the present study range. The effect of pressure was small when bubble size was the same. The prelimenery simulation result would be beneficial for better understanding complex two phase flow phenomena in fuel assembly with spacer grid.

This paper presents thermal aging effect on fracture toughness properties of GTAW (gas tungsten arc welding) and SMAW (shielded metal arc welding) of 316L stainless steels, and investigates the applicability of the existing three thermal aging models for CASS (cast stainless steels). Thermal aging was carried out at 350 °C for up to 15,000h and at 400 °C up to 8,000h. After aging, tensile and fracture toughness tests using 0.5T C(T) specimens were carried out at room temperature and at 288 °C. Comparing with the predictions using three (ANL, French and H3T) thermal aging models for CASS show that the predictions can be very non-conservative at operating temperature, and thus that the existing thermal aging models for CASS cannot be applied to the welded stainless steels.

Time-frequency analysis technique is an effective analysis tool for non-stationary processes. In the field of reactor neutron noise, the time-frequency analysis method has not been thoroughly researched and widely used. This work has studied the time-frequency analysis of the reactor neutron noise experimental signals under bubble disturbance and control rod vibration. First, an experimental platform was established, and it could be employed to reactor neutron noise experiment and data acquisition. Secondly, two types of reactor neutron noise experiments were performed, and valid experimental data was obtained. Finally, time-frequency analysis was conducted on the experimental data, and effective analysis results were obtained in the low-frequency part. Through this work, it can be concluded that the time-frequency analysis technique can effectively investigate the core dynamics behavior and deepen the identification of the unstable core process.

The current study relates to RADAR (RAdio Detection and Ranging) application for level measurement of vitrified radioactive liquid nuclear waste. The vitrification of radioactive liquid waste is carried out in special equipment called ‘Melters’. The study is directed towards the design and frequency modulation used in the level measurement of vitrified waste. More specifically, the RADAR design and frequency used for level measurement in a melter. This level measurement technique can also be used for dynamic vitrification process and can be used to measure the level variations without using any external medium/material and using only electromagnetic waves. Also, this technique is durable and accurate even under the high radioactive environment present inside the melter.

The analysis of 41Ca in concrete generated from the nuclear facilities decommissioning is critical for ensuring the safe management of radioactive waste. An analytical method for the determination of 41Ca in concrete is described. 41Ca is a neutron-activated long radionuclide, and hence, for accurate analysis, it is necessary to completely extract Ca from the concrete sample where it exists as the predominant element. The decomposition methods employed were the acid leaching, microwave digestion, and alkali fusion. A comparison of the results indicated that the alkali fusion is the most suitable way for the separation of Ca from the concrete sample. Several processes of hydroxide and carbonate precipitation were employed to separate 41Ca from interferences. The method relies on the differences in the solubility of the generated products. The behavior of Ca and the interfering elements such as Fe, Ni, Co, Eu, Ba, and Sr is examined at each separation step. The purified 41Ca was measured by a liquid scintillation counter, and the quench curve and counting efficiency were determined by using a certified reference material of known 41Ca activity. The recoveries in this study ranged from 56 to 68%, and the minimum detectable activity was 50 mBq g−1 with 0.5 g of concrete sample.

This paper suggests a fracture simulation method for SFR metallic fuel pin under accident condition. Two major failure mechanisms - creep damage and eutectic penetration - are implemented in the suggested method. To simulate damaged element, stress-reduction concept to reduce stiffness of the damaged element is applied. Using the proposed method, the failure size of cladding can be predicted in addition to the failure time and failure site. To verify the suggested method, Whole-pin furnace (WPF) test and TREAT-M test conducted at Argonne National Laboratory (ANL) are simulated. In all cases, predicted results and experimental results are overall in good agreement. Based on the simulation result, the effect of eutectic-penetration depth representing failure behavior on failure size is studied.

The nuclear reactor coolant pump transferring heat energy inherently brings with it the unsteady flow and inevitably threatens to the safe operation of the pump unit, especially with the pressure pulsation induced by the rotor-stator interaction. In this paper, the characteristics of pressure pulsation of the diffuser in a nuclear reactor coolant pump were investigated by the numerical simulation with experimental validation. Pressure pulsation signals measured synchronously from sensors mounted on the radial diffuser of a model pump were analyzed via Welch’s method. Frequency components induced by the rotor-stator interaction can be revealed by the diameter mode analysis method. The pressure pulsation of the diffuser is dominated by the blade passing frequency and its harmonics, which are free from the effect of flow rate and rotational speed while the corresponding amplitudes are easily affected by different operational conditions and measuring positions. The non-uniformity is much more affected by the rotational speed than the flow rate. This research is helpful for further work to reduce the pressure pulsation for the reactor coolant pump.

This paper presents ductile fracture simulation of full-scale cracked pipe for nuclear piping materials using the cohesive zone model (CZM). The main objective of this study is to investigate the applicability of CZM to predict ductile fracture of cracked pipes with various crack shapes and under quasi-static/dynamic loadings. The transferability of the traction-separation (T-S) curve from a small-scale specimen to a full-scale pipe is demonstrated by simulating small- and full-scale tests. T-S curves are calibrated by comparing experimental data of compact tension specimens with finite element analysis results. The calibrated T-S curves are utilized to predict the fracture behavior of cracked pipes. Three types of full-scale pipe tests are considered: pipe with circumferential through-wall crack under quasi-static/dynamic loadings, and with 360° internal surface crack under quasi-static loading. Computational results using the calibrated T-S curves show a good agreement with experimental data, demonstrating the transferability of the T-S curves from small-scale specimen.

During the postulated severe accident of nuclear reactor, eutectic reaction leads to low-temperature melting of fuel cladding and early failure of core structure. In order to model eutectic melting with the moving particle semi-implicit (MPS) method, the eutectic reaction model is developed to simulate the eutectic reaction phenomenon. The coupling of mass diffusion and phase diagram is applied to calculate the eutectic reaction with the uniform temperature. A heat transfer formula is proposed based on the phase diagram to handle the heat release or absorption during the process of eutectic reaction, and it can combine with mass diffusion and phase diagram to describe the eutectic reaction with temperature variation. The heat transfer formula is verified by the one-dimensional melting simulations and the predicted interface position agrees well with the theoretical solution. In order to verify the eutectic reaction models, the eutectic reaction of uranium and iron in two semi-infinite domains is simulated, and the profile of solid thickness decrease over time follows the parabolic law. The modified MPS method is applied to calculate Transient Reactor Test Facility (TREAT) experiment, the penetration rate in the simulations are agreeable with the experiment results. In addition, a hypothetical case based on the TREAT experiment is also conducted to validate the eutectic reaction with temperature variation, the results present continuity with the simulations of TREAT experiment. Thus the improved method is proved to be capable of simulating the eutectic reaction in the severe accident.

High density polyethylene (HDPE) pipe has many advantages over metallic pipe, and has been used in non-safety related application for years in some nuclear power plants (NPPs). Recently, HDPE pipe was introduced into safety related applications. The main difference between safety-related and non-safety-related pipes in NPPs is the design method of extra loadings such as gravity, temperature, and earthquake. In this paper, the mechanical behavior of HDPE pipe under various loads in pipe gallery was studied by finite element analysis (FEA). Stress concentrations were found at the fusion regions on inner surface of mitered elbows of HDPE pipe system. The effects of various factors were analyzed, and the influence of various loads on the damage of HDPE pipe system were evaluated. The results of this paper provide a reference for the design of nuclear safety-related Class 3 HDPE pipe. In addition, as the HDPE pipes analyzed in this paper were suspended in pipe gallery, it can also serve as a supplementary reference for current ASME standard on Class 3 HDPE pipe, which only covers the application for buried pipe application.

For the structural integrity evaluation of pressurized water reactor (PWR) steam generator (SG) tubes subjected to transient hydraulic loading, determination of the tube-to-tube gap velocity and static pressure distributions along the tubes is prerequisite. This paper addresses both computational fluid dynamics (CFD) and analytical approaches for predicting the tube-to-tube gap velocity and static pressure distributions during blowdown following a feedwater line break (FWLB) accident at a PWR SG. First of all, a comparative study on CFD calculations of the transient velocity and pressure distributions in the SG secondary sides for two different models having 30 or no tubes is performed. The result shows that the velocities of sub-cooled water flowing between any adjacent two tubes of a tubed SG model during blowdown can be roughly estimated by applying the specified SG secondary side porosity to those of the no-tubed SG model. Secondly, simplified analytical approximate solutions for the steady two-dimensional SG secondary flow velocity and pressure distributions under a given discharge flowrate are derived using a line sink model. The simplified analytical solutions are validated by comparing them to the CFD calculations.

The current study aims to explore the shielding properties of multi-component borate-based glass series. Seven glass-samples with composition of (80-y)H3BO3–10ZnO–10Na2O–yBaO where (y = 0, 5, 10, 15, 20, 25 and 30 mol.%) were synthesized by melt-quench method. Various shielding features for photons, neutrons, and protons were determined for all prepared samples. XCOM, Phy-X program, and SRIM code were performed to determine and explain several shielding properties such as equivalent atomic number, exposure build-up factor, specific gamma-ray constants, effective removal cross-section (ΣR), neutron scattering and absorption, Mass Stopping Power (MSP) and projected range. The energy ranges for photons and protons were 0.015–15 MeV and 0.01–10 MeV, respectively. The mass attenuation coefficient (μ/ρ) was also determined experimentally by utilizing two radioactive sources (166Ho and 137Cs). Consistent results were obtained between experimental and XCOM values in determining μ/ρ of the new glasses. The addition of BaO to the glass matrix led to enhance the μ/ρ and specific gamma-ray constants of glasses. Whereas the remarkable reductions in ΣR, MSP, and projected range values were reported with increasing BaO concentrations. The acquired results nominate the use of these glasses in different radiation shielding purposes.

The aim of this work is to help inform the decision for choosing a convenient material for the PGAA (Prompt Gamma Analysis Activation) collimator plug to be installed at the tangential channel of the Moroccan Triga Mark II Research Reactor. Two families of materials are usually used for collimator construction: a mixture of high-density polyethylene (HDPE) with boron, which is commonly used to moderate and absorb neutrons, and heavy materials, either for gamma absorption or for fast neutron absorption. An investigation of two different collimator designs was performed using N-Particle Monte Carlo MCNP6.2 code with the ENDF/B-VII.1 and MCLIP84 libraries. For each design, carbon steel and lead materials were used separately as collimator heavy materials. The performed study focused on both the impact on neutron beam quality and the neutron–gamma background at the exit of the collimator beam tube. An analysis and assessment of the principal findings is presented in this paper, as well as recommendations.

In this work, an attempt is made to improve the electrochemical corrosion resistance of Zircaloy-4 by helium ions implantation. For this purpose, the Zircaloy-4 was implanted with 300 keV helium ions of fluences 1 × 1013, 1 × 1015, and 1 × 1016 ions-cm−2 by using Pelletron Accelerator. Electrochemical tests of pristine and ion-implanted samples were performed in NaCl solution and their potentiodynamic polarization curves were obtained. The results showed enhancement of the corrosion resistance of Zircaloy-4 after helium ions implantation. The corrosion rate and current density of the material were significantly reduced by the helium implantation. The decrease in corrosion parameters was attributed to helium ions diffusion inside Zircaloy-4 that reduced the electrons flow from the samples.

In this paper, the design stress intensity values for the plate-type fuel assembly for research reactor are presented. Through a tensile test, the material properties of the cladding (aluminum alloy 6061) and structural material (aluminum alloy 6061-T6), in this case the yield and ultimate tensile strengths, Young’s modulus and the elongation, are measured with the temperatures. The empirical equations of the material properties with respect to the temperature are presented. The cladding undergoes several heat treatments and hardening processes during the fabrication process. Cladding strengths are reduced compared to those of the raw material during annealing. Up to a temperature of 150 °C, the strengths of the cladding do not significantly decrease due to the dislocations generated from the cold work. However, over 150 °C, the mechanical strengths begin to decrease, mainly due to recrystallization, dislocation recovery and precipitate growth. Taking into account the uncertainty of the 95% probability and 95% confidence level, the design stress intensities of the cladding and structural materials are established. The presented design stress intensity values become the basis of the stress design criteria for a safety analysis of plate-type fuels.

Various compositions of CeO2–Ce3Si2 (0, 10, 30, 50, and 100 wt%Ce3Si2) composites were fabricated using conventional sintering and spark plasma sintering. Lower relative density, enhanced interdiffusion of oxygen and silicon, and silicide agglomerations from the congruent melting of Ce3Si2 at 1390 °C were only observed from conventionally-sintered pellets. Thermal conductivity of spark plasma sintered CeO2–Ce3Si2 composites was calculated from the measured thermal diffusivity, specific heat, and density, which exhibited dense (>90 %TD) and homogeneous microstructure. The composite with 50 wt%Ce3Si2 exhibited 55% higher thermal conductivity than CeO2 at 500 °C, and 81% higher at 1000 °C.

Thorium compounds have attracted immense scientific and technological attention with regard to both fundamental and practical implications, owing to unique chemical and physical properties like high melting point, high density and thermal conductivity. Hereby, we investigate the mechanical and thermodynamic stability and report on the structural, electronic and magnetic properties of new silicon-doped cubic ternary thorium phosphides ThSixP1-x (x = 0, 0.25, 0.5, 0.75 and 1). The first-principles density functional theory procedure was adopted within full–potential linearized augmented plane wave (FP-LAPW) method. The exchange and correlation potential terms were treated within Generalized-Gradient-Approximation functional modified by Perdew-Burke-Ernzerrhof parameterizations. The proposed compounds showed mechanical and thermodynamic stable structure and hence can be synthesized experimentally. The calculated lattice parameters, bulk modulus, total energy, density of states, electronic band structure and spin magnetic moments of the compounds revealed considerable correlation to the Si substitution for P and the relative Si/P doping concentration. The electronic and magnetic properties of the doped compounds rendered them non-magnetic but metallic in nature. The main orbital contribution to the Fermi level arises from the hybridization of Th(6d+5f) and (Si+P)3p states. Reported results may have potential implications with regard to both fundamental point of view and technological prospects such as fuel materials for clean nuclear energy.

Analysis of the control rod insertion is important as it is closely related to reactor safety. Previously, the analysis has been carried out in RSG-GAS during static condition, not as a function of the fuel fraction. The RSG-GAS reactor in one cycle is a function of the fuel burn-up. It is necessary to analyze RSG-GAS core reactivity insertion as a function of the fuel burn-up to determine the behavior of the reactor, especially in uncontrolled operations such as continuous pulling of control rods. This analysis is carried out by the computer simulation method using WIMSD-5B and MTR-DYN codes, by observing power behavior as a function of time due to neutron chain reactions in the reactor core. Calculations are performed using point kinetics equation, and the feedback effect will be evaluated using static power coefficient and fuel burn-up function. Analyzes were performed for the core configuration of the core no. 99, by lifting the control rod or inserting positive reactivity to the core. The calculation results show that with the reactivity insertion of 0.5% Δk/k at start-up power of 1 W and 1 MW, safety limit is not exceeded either at the beginning, middle, or end of the cycle. The maximum temperature of the fuel is 135°C while the safety limit is 180°C. The margin from the safety limit is large, and therefore fuel damage is not possible when power excursion were to occur.

In this paper, the single-photon avalanche diodes (SPADs) featuring three different p-well implantation doses ( ∅ p − w e l l ) of 5.0 × 1012, 4.0 × 1012, and 3.0 × 1012 atoms/cm2 under the identical device layouts were fabricated and characterized to evaluate the effects of field enhanced mechanisms on primary dark pulses due to the maximum electric field. From the I–V curves, the breakdown voltages were found as 23.2 V, 40.5 V, and 63.1 V with decreasing ∅ p − w e l l , respectively. By measuring DCRs as a function of temperature, we found a reduction of approximately 8% in the maximum electric field lead to a nearly 72% decrease in the DCR at Vex = 5 V and T = 25 °C. Also, the activation energy increased from 0.43 eV to 0.50 eV, as decreasing the maximum electric field. Finally, we discuss the importance of electric field engineering in reducing the field-enhanced mechanisms contributing to the DCR in SPADs and the benefits on the SPADs related to different types of radiation detection applications.

To establish the exclusive role of hydrogen on burst behaviour of Zircaloy-4 during loss-of-coolant accident transients, an extensive single-rod burst tests were conducted on both unirradiated as-received and hydrogenated Zircaloy-4 cladding tubes at different heating rates and internal overpressures. The visual observations of cladding tubes during bursting as well as post-burst are presented in detail to understand the effect of hydrogen concentration, heating rate, and internal pressure. Impact of hydrogen on burst parameters—burst stress, burst strain, burst temperature—during loss-of-coolant accident transients are compared and discussed. Rupture at multiple locations for hydrogenated cladding at lower internal pressure and higher heating rate is reported for the very first time. A novel burst criterion accounting hydrogen concentration in nuclear fuel cladding is proposed.

The major drawback of zirconia-based materials, in view of their applications as targets for minor actinide transmutation, is their poor thermal conductivity. The addition of MgO, which has high thermal conductivity, to zirconia-based materials is expected to improve their thermal conductivity. On these grounds, the present study aims at phase characterization and thermophysical property evaluation of neodymium-substituted zirconia (Zr0.8Nd0.2O1.9; using Nd2O3 as a surrogate for Am2O3) and its composites with MgO. The composite was prepared by a solid-state reaction of Zr0.8Nd0.2O1.9 (synthesized by gel combustion) and commercial MgO powders at 1773 K. Phase characterization was carried out by X-ray diffraction and the microstructural investigation was performed using a scanning electron microscope equipped with energy dispersive spectroscopy. The linear thermal expansion coefficient of Zr0.8Nd0.2O1.9 increases upon composite formation with MgO, which is attributed to a higher thermal expansivity of MgO. Similarly, specific heat also increases with the addition of MgO to Zr0.8Nd0.2O1.9. Thermal conductivity was calculated from measured thermal diffusivity, temperature-dependent density and specific heat values. Thermal conductivity of Zr0.8Nd0.2O1.9–MgO (50 wt%) composite is more than that of typical UO2 fuel, supporting the potential of Zr0.8Nd0.2O1.9–MgO composites as target materials for minor actinides transmutation.

The power reactor with high-temperature gas-cooled reactor (HTGR) technology uses uranium as the reactor fuel. The energy from fission is converted to electrical energy or used for other needs such as hydrogen production or other research activities at high temperatures of around 700 °C. This operation does not allow the use of metal as the core material for the reactor. The material that fits the requirements as a core structure is graphite. Graphite material has specific characteristics, namely the parameters of the modulus of elasticity, coefficient of thermal expansion, and the volume which changes due to temperature and neutron dose. Because the structure of the reactor core is a vital component in the reactor, this research will develop a method for the design of the reactor core structure with graphite material. The design method is based on Design by Analysis which specifically refers to the strain analysis on each of the reactor core components. The design method developed is based on the finite element method. The object of this research is the side reflector made from the Toyo Tanso IG-110 series graphite. Based on the analysis of heat distribution and heat stress for the material before the effect of neutron exposure, the temperature distribution on the side reflector was found, as well as the displacement and heat stress that occurs. isotropic properties, Young's modulus and Poisson’s ratio values can be verified and estimated. The purpose of this research is to analyze the strain of the reactor core structure by taking into account the uncertainty of the graphite properties.

Efficiency calibration is a fundamental procedure in gamma spectrometric measurement. Experimental technique for efficiency calibration transfer in gamma spectrometer along different geometries and volumes has been developed and validated in this work. The developed technique offers simple and easy procedures to overcome several problems encountered in efficiency calibration of gamma spectrometer such as rate-related correction and different sample volumes. The validation shows that application of the developed technique has a precision of 95%.

Crtical heat flux (CHF) is an important characteristic of nanofluids. The CHF measurements were carried out in nanofluid research at the Center for Applied Nuclear Science and Technology. These measurements are done manually using a variable power supply and a multimeter. However, it was difficult to record the voltage and current due to the sudden break of the wire. In this study, Arduino was used to measure CHF automatically. The voltage is applied to the wire and increases automatically along with the measurement of the voltage and current in the wire. The results of the voltage and current measurements were compared with a multimeter and were not significantly different. It can be concluded that the CHF measurement system using arduino can be used to measure nanofluid CHF.

The free-piston Stirling engine (FPSE) has been widely used in aerospace owing to its advantages of high efficiency, high reliability, and self-starting ability. In this paper, a 20-kW FPSE is proposed by analyzing the requirements of space nuclear power reactor. A code was developed based on an improved simple analysis method to evaluate the performance of the proposed FPSE. The code is benchmarked with experimental data, and the maximum relative error of the output power is 17.1%. Numerical results show that the output power is 21 kW, which satisfies the design requirements. The results show that: a) reducing the pressure shell’s thickness can improve the output power significantly; b) the system efficiency increases with the wire porosity, while the growth of system efficiency decreases when the porosity is higher than 80%, and system efficiency exhibits a linear relationship with the temperatures of the cold and hot sides; c) the system efficiency increases with the compression ratio; the compression ratio increases by 16.7% while the system efficiency increases by 42%. This study can provide valuable theoretical support for the design and analysis of FPSEs for space nuclear power reactors.

The mechanical properties of α-Na3(U0.84(2),Na0.16(2))O4 have been researched using the first-principles calculations combined with the quasi-harmonic Debye model. The obtained lattice parameters agree well with the published experimental data. The results of elastic constants indicate that α-Na3(U0.84(2),Na0.16(2))O4 is mechanically stable. The polycrystalline moduli are predicted. The results show that the α-Na3(U0.84(2),Na0.16(2))O4 exhibits brittleness and possesses obvious elastic anisotropy. The hardness shows that it can be considered a “soft material”. Furthermore, the Debye temperature θD and the minimum thermal conductivity kmin are also discussed, respectively. Finally, the thermal expansion coefficient α , isobaric heat capacity C P and isochoric heat capacity C V are evaluated through the quasi-harmonic Debye model.

This study investigated the creep and creep crack growth (CCG) behavior of the base metal (BM), weld metal (WM), and heat affected zone (HAZ) in a Gr. 91 weldment, which was made by a shield metal arc weld process. A series of tensile, creep, and CCG tests were performed for the BM, WM, and HAZ at 550 °C. Creep behavior of the BM, WM, and HAZ was analyzed in terms of various creep laws; Norton’s power-law, Monkman-Grant relation and damage tolerance factor (λ), and their constants were determined. In addition, each CCGR law for the BM, WM, and HAZ was proposed and compared in terms of a C∗-fracture parameter. The WM and HAZ revealed faster creep rate, lower rupture ductility, and faster CCGRs than the BM, but they showed a similar behavior in the creep and CCG. The CCGRs obtained in the present study exhibited a marginal difference when compared with those of RCC-MRx of currently elevated design code in France. A creep crack path in the HAZ plane progressed towards a weak fine-grained HAZ adjacent to the BM.

Boron–neutron capture therapy (BNCT) is a cancer treatment method that exploits the high neutron reactivity of boron. Monitoring the prompt gamma rays (PGs) produced during neutron irradiation is essential for ensuring the accuracy and safety of BNCT. We investigate the imaging of PGs produced by the boron–neutron capture reaction through Monte Carlo simulations of a gamma camera with a SrI2 scintillator and parallel-hole collimator. GAGG scintillator is also used for a comparison. The simulations allow the shapes of the energy spectra, which exhibit a peak at 478 keV, to be determined along with the PG images from a boron–water phantom. It is found that increasing the size of the water phantom results in a greater number of image counts and lower contrast. Additionally, a higher septal penetration ratio results in poorer image quality, and a SrI2 scintillator results in higher image contrast. Thus, we can simulate the BNCT process and obtain an energy spectrum with a reasonable shape, as well as suitable PG images. Both GAGG and SrI2 crystals are suitable for PG imaging during BNCT. However, for higher imaging quality, SrI2 and a collimator with a lower septal penetration ratio should be utilized.

In view of the complex geometric structure of the rod bundle channel and the limitation of the current CHF visualization experiment technology, it is very difficult to obtain the rod bundle CHF mechanism directly through the phenomenon of the rod bundle CHF visualization experiment. In order to obtain the applicable CHF mechanism assumption for rod bundle channel, firstly, five most representative DNB type round tube CHF mechanistic models are obtained with evaluation and screening. Then these original round tube CHF mechanistic models based on inlet conditions are converted to local conditions and coupled with subchannel analysis code ATHAS. Based on 5 × 5 full-length rod bundle CHF experimental data independently developed by Nuclear Power Institute of China (NPIC), the applicability research of each model for CHF prediction performance in rod bundle channel is carried out, and the commonness and difference of each model are comparatively studied. The CHF mechanism assumption of superheated liquid layer depletion that is most likely to be applicable for the rod bundle channel is selected and two directions that need to be improved are given. This study provides a reference for the development of CHF mechanistic model in rod bundle channel.

The investigation of the utilization of enriched 208Pb as a coolant to enhance the performance of a long-life fast reactor with a Modified CANDLE (Constant Axial shape of Neutron flux, nuclide densities, and power shape During Life of Energy production) burnup scheme has performed. The analyzes were performed on a reactor with thermal power of 800 MegaWatt Thermal (MWTh) with a refueling process every 15 years. Uranium Nitride (enriched 15N), 208Pb, and High-Cr martensitic steel HT-9 were employed as fuel, coolant, and cladding materials, respectively. One of the Pb-nat isotopes, 208Pb, has the smallest neutron capture cross-section (0.23 mb) among other liquid metal coolants. Furthermore, the neutron-producing cross-section (n, 2n) of 208Pb is larger than sodium (Na). On the other hand, the inelastic scattering energy threshold of 208Pb is the highest among Na, natPb, and Bi. The small inelastic scattering cross-section of 208Pb can harden the neutron energy spectrum. Therefore, 208Pb is a better neutron multiplier than any other liquid metal coolant. The excess neutrons cause more production than consumption of 239Pu. Hence, it can reduce the initial fuel loading of the reactor. The selective photoreaction process was developing to obtain enriched 208Pb. The neutronic was calculated using SRAC and JENDL 4.0 as a nuclear data library. We obtained that the modified CANDLE reactor with enriched 208Pb as coolant and reflector has the highest k-eff among all reactors. Meanwhile, the natPb cooled reactor has the lowest k-eff. Thus, the utilization of the enriched 208Pb as the coolant can reduce reactor initial fuel loading. Moreover, the enriched 208Pb-cooled reactor has the smallest power peaking factor among all reactors. Therefore, the enriched 208Pb can enhance the performance of a long-life Modified CANDLE fast reactor.

Knowing a material’s true stress-strain curve is essential for performing a nonlinear finite element analysis to solve an elastoplastic problem. This study presents a simple methodology to determine the true stress-strain curve of type 304 and 316 austenitic stainless steels in the full range of strain from a typical tensile test. Before necking, the true stress and strain values are directly converted from engineering stress and strain data, respectively. After necking, a true stress-strain equation is determined by iteratively conducting finite element analysis using three pieces of information at the necking and the fracture points. The Hockett-Sherby equation is proposed as an optimal stress-strain model in a non-uniform deformation region. The application to the stainless steel under different temperatures and loading conditions verifies that the strain hardening behavior of the material is adequately described by the determined equation, and the estimated engineering stress-strain curves are in good agreement with those of experiments. The presented method is intrinsically simple to use and reduces iterations because it does not require much experimental effort and adopts the approach of determining the stress-strain equation instead of correcting the individual stress at each strain point.

This paper presents numerical estimation of burst pressures of axial cracked U-bend tubes, considering the U-bending process analysis. The validity of the FE simulations is confirmed by comparing with published experimental data. From parametric analyses, it is shown that existing EPRI burst pressure estimation equations for straight tubes can be conservatively used to estimate burst pressures of the U-bend tubes. This is due to the increase in yield strength during the U-bending process. The degree of conservatism would decrease with increasing the bend radius and with increasing the crack depth.

Experiments of vertically upward steam-water two-phase flow have been carried out in single-side heated narrow rectangular channel with a gap of 3 mm. Flow patterns were identified and classified through visualization directly. Slug flow was only observed at 0.2 MPa but replaced by block-bubble flow at 1.0 MPa. Flow pattern maps at the pressure of 0.2 MPa and 1.0 MPa were plotted and the difference was analyzed. The experimental data has been compared with other flow pattern maps and transition criteria. The results show reasonable agreement with Hosler's, while a wide discrepancy is observed when compared with air-water two-phase experimental data. Current criteria developed based on air-water experiments poorly predict bubble-slug flow transition due to the different formation and growth of bubbles. This work is significant for researches on heat transfer, bubble dynamics and flow instability.

In this work, Rhizophora spp. particleboard phantoms were made using SPI-based adhesives, modified with sodium hydroxide and itaconic acid polyamidoamine-epichlorohydrin (0, 5, 10, and 15 wt%). An X-ray computed tomography (CT) imaging system was used to ascertain the CT numbers and density distribution profiles of the particleboards. The SPI-based/NaOH/IA-PAE/Rhizophora spp. particleboard phantoms with 15 wt% IA-PAE addition level had the highest solid content, flexural strength, flexural modulus, and internal bonding strength of 36.06 ± 1.08%, 18.61 ± 0.38 Nmm−2, 7605.76 ± 0.89 Nmm−2, and 0.463 ± 0.053 Nmm−2, respectively. The moisture content, mass density, water absorption, and dimensional stability were 6.93 ± 0.27%, 0.962 ± 0.037 gcm−3, 22.36 ± 2.47%, and 10.90 ± 0.86%, respectively. The results revealed that the mass attenuation coefficients and effective atomic number values within the 16.59–25.26 keV photon energy region, were close to the calculated XCOM values in water, with a p-value of 0.077. Moreover, the CT images showed that the dissimilarities in the discrepancy of the profile density decreased as the IA-PAE concentrations increased. Therefore, these results support the appropriateness of the SPI-based/NaOH/IA-PAE/Rhizophora spp. particleboard with 15 wt% IA-PAE adhesive as a suitable tissue-equivalent phantom material for medical health applications.

This paper presents a numerical prediction of the transient hydraulic loads acting on the tubes and external supports of a pressurized water reactor (PWR) steam generator (SG) during blowdown following a sudden feedwater line break (FWLB). A simplified SG model was used to easily demonstrate the prediction. The blowdown discharge flow was treated as a flashing flow to realistically simulate the transient flow fields inside the SG and the connected broken feedwater pipe. The effects of the SG initial pressure or the broken feedwater pipe length on the intensities or magnitudes of transient hydraulic loads were investigated. Then predictions of the decompression pressure wave-induced impulsive pressure differential loads on SG tubes and the transient blowdown loads on SG external supports were demonstrated and the general aspects of transient responses of such transient hydraulic loads to the FWLB were discussed.

The microstructure and texture of three 316L foils of 25 μm thickness, which were subjected to different manufacturing process, were systematically characterized using advance analytical techniques. Then, the electrochemical property of the 316L foils in simulated pressurized water reactor (PWR) solution was analyzed using potentiodynamic polarization. The results showed that final rolling strain and annealing temperature had evident effect on grain size, fraction of recrystallization, grain boundary type and texture distribution. It was suggested that large final rolling strain could transfer Brass texture to Copper texture; low annealing temperature could limit the formation of preferable orientations in the rolling process to reduce anisotropy. Potentiodynamic polarization test showed that all samples exhibited good corrosion performance in the simulated primary PWR solution.

The analysis of the fluid flow characteristics in reactor pressure vessel is an important part of the hydraulic design of nuclear power plant, which is related to the structure design of reactor internals, the flow distribution at core inlet and the safety of nuclear power plant. The flow distribution and mixing characteristics in the pressurized reactor vessel for the 1000MWe advanced pressurized water reactor is analyzed by using Computational Fluid Dynamics (CFD) method in this study. The geometry model of the full-scaled reactor vessel is built, which includes the cold and hot legs, downcomer, lower plenum, core, upper plenum, top plenum, and is verified with some parameters in DCD. Under normal condition, it is found that the flow skirt, core plate holes and outlet pipe cause pressure loss. The maximum and minimum flow coefficient is 1.028 and 0.961 respectively, and the standard deviation is 0.019. Compared with other reactor type, it shows relatively uniform of the flow distribution at the core inlet. The coolant mixing coefficient is investigated with adding additional variables, showing that mass transfer of coolant occurs near the interface. The coolant mainly distributes in the 90° area of the corresponding core inlet, and mixes at the interface with the coolant from the adjacent cold leg. 0.1% of corresponding coolant is still distributed at the inlet of the outer-ring components, indicating wide range of mixing coefficient distribution.

Dual-cooled annular fuel allows a significant increase in power density while maintaining or improving safety margins. However, the dual-cooled design brings much higher Zircaloy charge in reactor core, which could cause a great threaten of hydrogen explosion during severe accidents. Hence, an innovative fuel combined dual-cooled annular geometry and SiC cladding was proposed for the first time in this study. Capabilities of fuel design and behavior simulation were developed for this new fuel by the upgrade of FROBA-ANNULAR code. Considering characteristics of both SiC cladding and dual-cooled annular geometry, the basic fuel design was proposed and preliminary proved to be feasible. After that, a design optimization study was conducted, and the optimal values of as-fabricated plenum pressure and gas gap sizes were obtained. Finally, the performance simulation of the new fuel was carried out with the full consideration of realistic operation conditions. Results indicate that in addition to possessing advantages of both dual-cooled annular fuel and accident tolerant cladding at the same time, this innovative fuel could overcome the brittle failure issue of SiC induced by pellet-cladding interaction.

A modern approach to nuclear energy involves reprocessing like transmutations of spent nuclear fuel products to reduce their radiotoxicity and time needed for their storage. For this purpose, they are immobilized in inert matrices made of zirconia and can be "burned" in fast neutron reactor or Accelerator Driven System. These matrices in spherical form can be obtained by sol-gel process. The paper presents a method of microspheres fabrication based on the combined Complex Sol-Gel Process and double extraction process consisting in the preparation of zirconium-ascorbate sol and simultaneous extraction of water and nitrates. The procedure allows obtaining gel microspheres with a diameter of 50 μm, which after heat treatment are processed into the final product. The synthesis of zirconia microspheres with Yttrium by internal gelation process is well known for over a decade now. However, the explanation and characterization of synthesis of such material by extraction of water process is rarely found. Parameters such as: pH, viscosity, shape, sphericity and crystal structure have been determined for synthesized products and semi-products. In addition, preliminary research consisting in irradiation of the obtained materials in fast and thermal neutron flux was carried out. The obtained results are presented and described in this work.

Nickel base Alloy X-750, which is used as fastener parts in light-water reactor (LWR), has experienced many failures by environmentally assisted cracking (EAC). In order to improve the reliability of passive components for nuclear power plants (NPP's), it is necessary to study the failure mechanism and to predict crack growth behavior by developing a probabilistic failure model. In this study, The Bayesian inference was employed to reduce the uncertainties contained in EAC modeling parameters that have been established from experiments with Alloy X-750. Corrosion fatigue crack growth rate model (FCGR) was developed by fitting into Paris’ Law of measured data from the several fatigue tests conducted either in constant load or constant ΔK mode. These parameters characterizing the corrosion fatigue crack growth behavior of X-750 were successfully updated to reduce the uncertainty in the model by using the Bayesian inference method. It is demonstrated that probabilistic failure models for passive components can be developed by updating a laboratory model with field-inspection data, when crack growth rates (CGRs) are low and multiple inspections can be made prior to the component failure.

As the High Flux Advanced Neutron Application Reactor (HANARO) has been recently required to support new R&D relevant to future nuclear systems requiring a much higher neutron fluence, the development of irradiation capsule technology for long-term irradiation testing was performed in three steps (3, 5, 10 dpa). At first, several design improvements of a standard capsule were suggested based on a failure analysis of the capsule and successfully applied for irradiation testing at HANARO at up to eight reactor operation cycles equivalent to 3 dpa. Based on a schematic stress analysis of the vulnerable parts of the previous capsule, an optimized design of the capsule was made for 5 dpa irradiation. The newly designed capsule was safely out-pile tested up to 450 days, which was equivalent to 5 dpa irradiation in the reactor. The test results were submitted to the Reactor Safety Review Committee of HANARO and irradiation testing for 5 dpa was approved. The capsule was also successfully out-pile tested to evaluate the possibility of irradiation testing for 10 dpa. For a higher neutron fluence exceeding 10 dpa, new capsule technologies, including a new capsule that has a different bottom design and neutron flux boosting capsule, were also suggested.

Hydrogen peroxide is a radiolysis product of water formed under gamma-irradiation; therefore, its reliable detection is crucial in the nuclear industry for spent fuel management and coolant chemistry. This study proposes an electrochemical sensor for hydrogen peroxide detection. Cysteamine (CYST), gold nanoparticles (GNPs), and horseradish peroxidase (HRP) were used in the modification of a gold electrode for fabricating Au/CYST/GNP/HRP sensor. Each modification step of the electrode was investigated through electrochemical and physical methods. The sensor exhibited strong sensitivity and stability for the detection and measurement of hydrogen peroxide with a linear range of 1–9 mM. In addition, the Michaelis–Menten kinetic equation was applied to predict the reaction curve, and a quantitative method to define the dynamic range is suggested. The sensor is highly sensitive to H2O2 and can be applied as an electrochemical H2O2-sensor in the nuclear industry.

In this work, self-absorption correction factor related to the variation of the composition and the density of soil samples were evaluated using the p-type HPGe detector. The validated MCNP5 simulation model of this detector was used to evaluate its Full Energy Peak Efficiency (FEPE) under the variation of the composition and the density of the analysed samples. The results indicates that FEPE calculation of low gamma ray is affected by the composition and the density of soil samples. The self-absorption correction factors for different gamma-ray energies which was fitted as a function of FEPEs via density and energy and fitting parameters as polynomial function for the logarithm neper of gamma ray energy help to calculate quickly the detection efficiency of detector. Factor Analysis for the influence of the element composition in analysed samples on the FEPE indicates the FEPE distribution changes from non-metal to metal groups when the gamma ray energy increases from 92 keV to 238 keV. At energies above 238 keV, the FEPE primarily depends only on the metal elements and is significantly affected by aluminium and silicon composition in soil samples.