Excellence in Research and Innovation for Humanity

International Science Index

Commenced in January 1999 Frequency: Monthly Edition: International Abstract Count: 42820

Nuclear and Quantum Engineering

Investigation of Pu-238 Heat Source Modifications to Increase Power Output through (α,N) Reaction-Induced Fission
The objective of this study is to improve upon the current ²³⁸PuO₂ fuel technology for space and defense applications. Modern RTGs (radioisotope thermoelectric generators) utilize the heat generated from the radioactive decay of ²³⁸Pu to create heat and electricity for long term and remote missions. Application of RTG technology is limited by the scarcity and expense of producing the isotope, as well as the power output which is limited to only a few hundred watts. The scarcity and expense make the efficient use of ²³⁸Pu absolutely necessary. By utilizing the decay of ²³⁸Pu, not only to produce heat directly but to also indirectly induce fission in ²³⁹Pu (which is already present within currently used fuel), it is possible to see large increases in temperature which allows for a more efficient conversion to electricity and a higher power-to-weight ratio. This concept can reduce the quantity of ²³⁸Pu necessary for these missions, potentially saving millions on investment, while yielding higher power output. Current work investigating radioisotope power systems have focused on improving efficiency of the thermoelectric components and replacing systems which produce heat by virtue of natural decay with fission reactors. The technical feasibility of utilizing (α,n) reactions to induce fission within current radioisotopic fuels has not been investigated in any appreciable detail, and our study aims to thoroughly investigate the performance of many such designs, develop those with highest capabilities, and facilitate experimental testing of these designs. In order to determine the specific design parameters that maximize power output and the efficient use of ²³⁸Pu for future RTG units, MCNP6 simulations have been used to characterize the effects of modifying fuel composition, geometry, and porosity, as well as introducing neutron moderating, reflecting, and shielding materials to the system. Although this project is currently in the preliminary stages, the final deliverables will include sophisticated designs and simulation models that define all characteristics of multiple novel RTG fuels, detailed enough to allow immediate fabrication and testing. Preliminary work has consisted of developing a benchmark model to accurately represent the ²³⁸PuO₂ pellets currently in use by NASA; this model utilizes the alpha transport capabilities of MCNP6 and agrees well with experimental data. In addition, several models have been developed by varying specific parameters to investigate their effect on (α,n) and (n,fi ssion) reaction rates. Current practices in fuel processing are to exchange out the small portion of naturally occurring ¹⁸O and ¹⁷O to limit (α,n) reactions and avoid unnecessary neutron production. However, we have shown that enriching the oxide in ¹⁸O introduces a sufficient (α,n) reaction rate to support significant fission rates. For example, subcritical fission rates above 10⁸ f/cm³-s are easily achievable in cylindrical ²³⁸PuO₂ fuel pellets with a ¹⁸O enrichment of 100%, given an increase in size and a ⁹Be clad. Many viable designs exist and our intent is to discuss current results and future endeavors on this project.
Using Habit to Establish the Analysis Methodology for Maahshan Nuclear Power Plant
In this research, we established the HABIT analysis methodology for Maanshan nuclear power plant (NPP) using FSAR, reports, and other data. In order to evaluate the control room habitability under the CO2 storage burst, Maanshan NPP HABIT methodology was used to perform this case analysis. The HABIT results were lower than the R.G. 1.78 failure criteria. This indicates that Maanshan NPP habitability can be maintained. Additionally, the sensitivity study of the parameters such as wind speed, atmospheric stability class, air temperature, and control room intake flow rate was also performed in this study.
Addressing Public Concerns about Radiation Impacts by Looking Back in Nuclear Accidents Worldwide
According to a report of International Atomic Energy Agency (IAEA), there are approximately 437 nuclear power stations are in operation in the present around the world in order to meet increasing energy demands. Indeed, nearly, a third of the world’s energy demands are met through nuclear power because it is one of the most efficient and long-lasting sources of energy. However, there are also consequences when a major event takes place at a nuclear power station. Over the past years, a few major nuclear accidents have occurred around the world. According to a report of International Nuclear and Radiological Event Scale (INES), there are six nuclear accidents that are considered to be high level (risk) of the events: Fukushima Dai-chi (Level 7), Chernobyl (Level 7), Three Mile Island (Level 5), Windscale (Level 5), Kyshtym (Level 6) and Chalk River (Level 5). Today, many people still have doubt about using nuclear power. There is growing number of people who are against nuclear power after the serious accident occurred at the Fukushima Dai-chi nuclear power plant in Japan. In other words, there are public concerns about radiation impacts which emphasize Linear-No-Threshold (LNT) Issues, Radiation Health Effects, Radiation Protection and Social Impacts. This paper will address those keywords by looking back at the history of these major nuclear accidents worldwide, based on INES. This paper concludes that all major mistake from nuclear accidents are preventable due to the fact that most of them are caused by human error. In other words, the human factor has played a huge role in the malfunction and occurrence of most of those events. The correct handle of a crisis is determined, by having a good radiation protection program in place, it’s what has a big impact on society and determines how acceptable people are of nuclear.
Consideration of Failed Fuel Detector Location through Computational Flow Dynamics Analysis on Primary Cooling System Flow with Two Outlets
Failed fuel detector (FFD) in research reactor is a very crucial instrument to detect the anomaly from failed fuels in the early stage around primary cooling system (PCS) outlet prior to the decay tank. FFD is considered as a mandatory sensor to ensure the integrity of fuel assemblies and mitigate the consequence from a failed fuel accident. For the effective function of FFD, the location of them should be determined by contemplating the effect from coolant flow around two outlets. For this, the analysis on computational flow dynamics (CFD) should be first performed how the coolant outlet flow including radioactive materials from failed fuels are mixed and discharged through the outlet plenum within certain seconds. The analysis result shows that the outlet flow is well mixed regardless of the position of failed fuel and ultimately illustrates the effect of detector location.
Seasonal Variation of the Unattached Fraction and Equilibrium Factor of ²²²Rn, ²²⁰Rn
Radon (²²²Rn) and its decay products are the major sources of natural radiation exposure to general population. The activity concentrations of radon, thoron gasses, and their unattached and attached short-lived progeny in indoor environment of the Jaipur and Ajmer districts of Rajasthan had been calculated via passive measurements using the Pinhole cup dosimeter, deposition based progeny sensors (DRPS/DTPS) and wire mesh capped (DRPS/DTPS) progeny sensors. The results of this study revealed that radon and thoron concentrations (CRn, CTn) are highest in the winter season. The variation of the radon and its decay products are observed to vary seasonally, but these environmental parameters seem not to be affecting the thoron and its decay product concentrations in a regular manner. The average values of the radon and its decay products are maximum in winter and minimum in summer. The equilibrium factor for radon is observed to be 0.50, 0.47 and 0.49 in winter, rainy and summer seasons. The annual average value of the unattached fraction of the radon progeny comes out to be 0.34. On the other hand, the average value of thoron (²²⁰Rn) concentration and its equilibrium factor in the studied area comes to be 74, 39, 45 Bq m⁻³ and 0.07, 0.11, 0.07 respectively for the winter, rainy and summer seasons with the annual average value of the unattached fraction of about 0.18. The annual average radiological dose from exposure to indoor radon and thoron progeny comes out to be 0.88 and 0.78 mSv.
Using RADionuclide, Transport, Removal, and Dose Estimation and Symbolic Nuclear Analysis Package to Establish the Analysis Model for Maahshan PWR Plant
In this study, we focus on the establishment of the analysis model for Maanshan PWR nuclear power plant by using RADTRAD and SNAP codes with the FSAR, manuals, and other data. In order to evaluate the cumulative dose at the EAB and LPZ outer boundary, Maanshan NPP RADTRAD/SNAP model was used to perform the analysis of the DBA LOCA case. The analysis results of RADTRAD were similar to FSAR data. The analysis results were lower than the failure criteria of 10 CFR 100.11 (a total radiation dose to the whole body, 250 mSv; a total radiation dose to the thyroid from iodine exposure, 3000 mSv).
The Preparation of Titanate Nano-Materials Removing Efficiently Cs-137 from Waste Water in Nuclear Power Plants
Cs-137, the radioactive fission products of uranium, can be easily dissolved in water during the accident of nuclear power plant, such as Chernobyl, Three Mile Island, Fukushima accidents. The concentration of Cs in the groundwater around the nuclear power plant exceeded the standard value almost 10,000 times after the Fukushima accident. The adsorption capacity of Titanate nano-materials for radioactive cation (Cs+) is very strong. Moreover, the radioactive ion can be tightly contained in the nanotubes or nanofibers without reversible adsorption, and it can safely be fixed. In addition, the nano-material has good chemical stability, thermal stability and mechanical stability to minimize the environmental impact of nuclear waste and waste volume. The preparation of titanate nanotubes or nanofibers was studied by hydrothermal methods, and chemical kinetics of removal of Cs by nano-materials was obtained. The adsorption time with maximum adsorption capacity and the effects of pH, coexisting ion concentration and the optimum adsorption conditions on the removal of Cs by titanate nano-materials were also obtained. The adsorption boundary curves, adsorption isotherm and the maximum adsorption capacity of Cs-137 as tracer on the nano-materials were studied in the research. The experimental results showed that the removal rate of Cs-137 in 0.01 tons of waste water with only 1 gram nano-materials could reach above 98%, according to the optimum adsorption conditions.
Neutron Calculations of 18 MV Electron Linear Accelerator by Using Monte Carlo Method
In this study, the 18 MV medical electron linear accelerator was simulated with the MCNP code, which is calculated by the Monte Carlo method. Dosimetric quality controls were made by comparing simulation results with experimental and theoretical values. Medical linear accelerators are based on the principle of accelerating electrons broken from a metal into the electromagnetic field. Electrons reaching the desired kinetic energy come out of the device as an electron beam for electron therapy. When photon therapy is applied, high-energy electrons are hit on a target, and bremsstrahlung X-rays are produced. The contribution of the linear accelerator to the photoneutron production of the head components was calculated by the Monte Carlo method. The most of the neutron production occurs in the primary collimator, target, and secondary collimator. Neutron fluxes were calculated at the surface of water phantom and the depths of 5 cm, 10 cm and 20 cm in the filtered system for the 10x10 cm² irradiation field. It was determined that there was a steady decrease in fast neutron flux with depth.
X-Ray Diffraction and Precision Dilatometer Study of Neutron-Irradiated Nuclear Graphite Recovery Process up to 1673K
Four kinds of nuclear graphite, IG-110U, ETP-10, CX-2002U and IG-430U were neutron-irradiated at different fluences and temperatures, ranged from 1.38 x 1024 to 7.4 x 1025 n/m2 (E > 1.0 MeV) at 473K, 573K and 673K. To take into account the disorder in the microstructure, such as stacking faults and anisotropic coherent lengths, the X-ray diffraction patterns were interpreted using a comprehensive structural model and a refinement program CARBONXS. The deduced structural parameters show the changes of lattice parameters, coherent lengths along the c-axis and the basal plane, and the degree of turbostratic disorder as a function of the irradiation dose. Our results reveal neutron irradiation effects on the microstructure and macroscopic dimension, which are consistent with previous work. The methodology used in this work enables the quantification of the damage on the microstructure of nuclear graphite induced by neutron irradiation.
Molecular Dynamics Simulation of Irradiation-Induced Damage Cascades in Graphite
Graphite is the matrix, and structural material in the high temperature gas-cooled reactor exhibits an irradiation response. It is of significant importance to analyze the defect production and evaluate the role of graphite under irradiation. A vast experimental literature exists for graphite on the dimensional change, mechanical properties, and thermal behavior. However, simulations have not been applied to the atomistic perspective. Remarkably few molecular dynamics simulations have been performed to study the irradiation response in graphite. In this paper, irradiation-induced damage cascades in graphite were investigated with molecular dynamics simulation. Statistical results of the graphite defects were obtained by sampling a wide energy range (1–30 KeV) and 10 different runs for every cascade simulation with different random number generator seeds to the velocity scaling thermostat function. The chemical bonding in carbon was described using the adaptive intermolecular reactive empirical bond-order potential (AIREBO) potential coupled with the standard Ziegler–Biersack–Littmack (ZBL) potential to describe close-range pair interactions. This study focused on analyzing the number of defects, the final cascade morphology and the distribution of defect clusters in space, the length-scale cascade properties such as the cascade length and the range of primary knock-on atom (PKA), and graphite mechanical properties’ variation. It can be concluded that the number of surviving Frenkel pairs increased remarkably with the increasing initial PKA energy but did not exhibit a thermal spike at slightly lower energies in this paper. The PKA range and cascade length approximately linearly with energy which indicated that increasing the PKA initial energy will come at expensive computation cost such as 30KeV in this study. The cascade morphology and the distribution of defect clusters in space mainly related to the PKA energy meanwhile the temperature effect was relatively negligible. The simulations are in agreement with known experimental results and the Kinchin-Pease model, which can help to understand the graphite damage cascades and lifetime span under irradiation and provide a direction to the designs of these kinds of structural materials in the future reactors.
Guided Energy Theory of a Particle: Answered Questions Arise from Quantum Foundation
This work aimed to introduce a theory, called Guided Energy Theory of a particle that answered questions that arise from quantum foundation, quantum mechanics theory, and interpretation such as: what is nature of wavefunction? Is mathematical formalism of wavefunction correct? Does wavefunction collapse during measurement? Do quantum physical entanglement and many world interpretations really exist? In addition, is there uncertainty in the physical reality of our nature as being concluded in the Quantum theory? We have been able to show by the fundamental analysis presented in this work that the way quantum mechanics theory, and interpretation describes nature is not correlated with physical reality. Because, we discovered amongst others that, (1) Guided energy theory of a particle fundamentally provides complete physical observable series of quantized measurement of a particle momentum, force, energy e.t.c. in a given distance and time.In contrast, quantum mechanics wavefunction describes that nature has inherited probabilistic and indeterministic physical quantities, resulting in unobservable physical quantities that lead to many worldinterpretation.(2) Guided energy theory of a particle fundamentally predicts that it is mathematically possible to determine precise quantized measurementof position and momentum of a particle simultaneously. Because, there is no uncertainty in nature; nature however naturally guides itself against uncertainty. Contrary to the conclusion in quantum mechanics theory that, it is mathematically impossible to determine the position and the momentum of a particle simultaneously. Furthermore, we have been able to show by this theory that, it is mathematically possible to determine quantized measurement of force acting on a particle simultaneously, which is not possible on the premise of quantum mechanics theory. (3) It is evidently shown by our theory that, guided energy does not collapse, only describes the lopsided nature of a particle behavior in motion. This pretty offers us insight on gradual process of engagement - convergence and disengagement – divergence of guided energy holders which further highlight the picture how wave – like behavior return to particle-like behavior and how particle – like behavior return to wave – like behavior respectively. This further proves that the particles’ behavior in motion is oscillatory in nature. The mathematical formalism of Guided energy theory shows that nature is certainty whereas the mathematical formalism of Quantum mechanics theory shows that nature is absolutely probabilistics. In addition, the nature of wavefunction is the guided energy of the wave. In conclusion, the fundamental mathematical formalism of Quantum mechanics theory is wrong.
Thermophysical Properties of Molten Stainless Steel Containing 5 Mass%-B4C
Thermophysical properties of the molten mixture of stainless steel (SS) and control rod material (B4C) are necessary to understand a core degradation mechanism in severe accidents of sodium-cooled fast reactors. Thermophysical property database should be built for the system of SS and B4C. However, the reliable data are scarce for the SS-B4C melts because of the experimental difficulty in measurements. Generally, high-temperature melts may react with container materials, which causes chemical contamination and limits a temperature range for measurements. In addition, convection existing in the melts causes large uncertainty in thermal conductivity measurement. In order to overcome these difficulties, Fukuyama group and their collaborators have developed a noncontact high-temperature thermophysical property measurement system, which is called as PROSPECT (Properties and Simulations Probed with Electromagnetic Containerless Technique). This system consists of an electromagnetic levitator incorporating a superconducting magnet, laser heating system, high-speed CCD camera, data-logging system and gas-controlling system including an oxygen sensor. The levitation technique provides contamination free measurements. A dc magnetic field generated by the superconducting magnet is applied to the levitated droplet, which suppresses the droplet oscillation and also convection in the droplet because of the Lorentz force. Thus, PROSPECT greatly improves uncertainties in measurements of heat capacity, emissivity, density and surface tension of high-temperature molten metals and alloys. Moreover, PROSPECT has a significant technical advantage that the true thermal conductivities can be measured with suppressed convection in the droplet under a dc magnetic field. In this study, the liquidus temperature was first determined by differential scanning calorimetry (DSC) for the stainless steel (SUS316L) containing 5 mass%-B4C. After this, the density, surface tension, normal spectral emissivity, specific heat and thermal conductivity of molten SUS316L - 5 mass%-B4C were measured using PROSPECT. The density was measured by a laser imaging method for the levitated droplet. The surface tension was measured by an oscillating droplet method. The normal spectral emissivity was measured by a direct measurement of radiation from droplet surface. The specific heat and thermal conductivity were measured by a laser modulation calorimetry with controlling a dc magnetic field. All thermophysical properties were successfully determined as a function of temperature over a wide temperature range including a supercooling liquid state. Experimental uncertainty was also evaluated for each thermophysical property. Experimental details and results will be presented in ICNMNS 2017.
Communicating Nuclear Energy in Southeast Asia: A Cross-Country Comparison of Communication Channels and Source Credibility
Nuclear energy is a contentious technology that has attracted much public debate over the years. The prominence of nuclear energy in Southeast Asia (SEA) has burgeoned due to the surge of interest and plans for nuclear development in the region. Understanding public perceptions of nuclear energy in SEA is pertinent given the limited number of studies conducted. In particular, five SEA nations – Singapore, Malaysia, Indonesia, Thailand, and Vietnam are of immediate interest as that they are amongst the most economically developed or developing nations in the SEA region. High energy demands from economic development in these nations have led to considerations of adopting nuclear energy as an alternative source of energy. This study aims to explore whether differences in the nuclear developmental stage in each country affects public perceptions of nuclear energy. In addition, this study seeks to find out about the type and importance of communication credibility as a judgement heuristic in facilitating message acceptance across these five countries. Credibility of a communication channel is a crucial component influencing public perception, acceptance, and attitudes towards nuclear energy. Aside from simply identifying the frequently used communication channels, it is of greater significance to understand public perception of source and media credibility. Given the lack of studies conducted in SEA, this exploratory study adopts a qualitative approach to elicit a spectrum of opinions and insights regarding the key communication aspects influencing public perceptions of nuclear energy. Specifically, the capitals of each of the abovementioned countries - Kuala Lumpur, Bangkok, and Hanoi - were selected, with the exception of Singapore, an island city-state, and Yogyakarta, the most populous island of Indonesia to better understand public perception towards nuclear energy. Focus group discussions were utilized as the mode of data collection to elicit a wide variety of viewpoints held by the participants, which is well-suited for exploratory research. In total, 156 participants took part in the 13 focus group discussions. The participants were either local citizens or permanent residents aged between 18 and 69 years old. Each of the focus groups consists of 8-10 participants, including both male and female participants. The transcripts from each focus group were analysed using NVivo 10, and the text was organised according to the emerging themes or categories. The general public in all the countries was familiar but had no in-depth knowledge with nuclear energy. Four dimensions of nuclear energy communication were identified based on the focus group discussions: communication channels, perceived credibility of sources, circumstances for discussion, and discussion style. The first dimension, communication channels refers to the medium through which participants receive information about nuclear energy. Four types of media emerged from the discussions. They included online and social media, broadcast media, print media, and word-of- mouth (WOM). Collectively, across all five countries, participants were found to engage in different types of knowledge acquisition and information seeking behavior depending on the communication channels used.
Specific Gravity and Specific Heat of Stainless Steel Containing 5mass%-B4C
Control rod materials (B4C: boron carbide) might melt in molten core material mixture pool during severe accident in sodium-cooled fast reactors. In such a situation, the cladding and wrapper tube made of stainless steel (SS) as the structural material would melt by eutectic reaction between B4C and SS below the SS melting temperature, thereby producing a B4C-SS eutectic alloy. To address this material, the development of thermo-physical property models is necessary for severe accident simulation code which could be applied to reactor safety assessment. In the severe accident, the B4C-SS eutectic alloy formed as the melt by the eutectic reaction would freeze by the heat transfer with cold core materials. Therefore, it is necessary to develop the thermophysical property database of the B4C-SS eutectic alloy in the solid phase. This study is intended to measure the specific gravity, and specific heat of the B4C-SS eutectic alloy with emphasis on 5 mass%-B4C and SS. 5mass%B4C-SS standard sample distributed boron and carbon uniformly was synthesized to clarify the thermophysical properties of the B4C-SS eutectic alloy. The homogeneity was confirmed by a chemical analysis and a microstructure observation. The melting start temperature of the sample was obtained by using a thermal gravimetric-differential thermal analyzer (TG-DTA). As the thermophysical properties under solid phase, specific gravity and specific heat were evaluated up to 1000°C. The specific gravity was measured by utilizing the Archimedes method (at room temperature) and the thermal expansion method. The specific heat was evaluated by utilizing the adiabatic calorimetry method and the TG-DTA method. The specific gravity was about 7.4g/cm3 at room temperature and decrease of the specific gravity due to the temperature rise was slow compared with SUS316L. The specific heat was slightly higher than that of the SUS316L, and showed similar temperature dependence up to 800°C.
Investigation of the Effects of Plaster on Radiation Doses Given to Patients
In this work, it is aimed to measure the total attenuation coefficients (TACs) of plaster materials using different geometries of the plaster, field sizes, and photon beam energies. The percent depth dose measurements were made using X-rays at 6 MV and 18 MV from the linear accelerator in order to plan the probable radiation dose to be given to the patient wearing a plaster cast. By changing the thickness of tissue equivalent solid phantom with and without plaster, the linear attenuation coefficients were experimentally determined for (5x5) cm2, (10x10) cm2 and (15x15) cm2 areas and investigated the effect of plaster to dose. It was shown that the TACs are decreasing while energy is increasing. The TAC of plaster was also calculated by using the NIST X-ray program and compared with the measured values. The TAC of solid phantom with plaster was found to be larger than without plaster. This means that dmax is very close to the skin and that in these situations plaster has a bolus effect, increasing the skin's dose.
Approaches for Minimizing Radioactive Tritium and ¹⁴C in Advanced High Temperature Gas-Cooled Reactors
High temperature gas-cooled reactors (HTGRs) are considered as one of the next-generation advanced nuclear reactors, in which porous nuclear graphite is used as neutron moderators, reflectors, structure materials, and cooled by inert helium. Radioactive tritium and ¹⁴C are generated in terms of reactions of thermal neutrons and ⁶Li, ¹⁴N, ¹⁰B impurely within nuclear graphite and the coolant during HTGRs operation. Currently, hydrogen and nitrogen diffusion behavior together with nuclear graphite microstructure evolution were investigated to minimize the radioactive waste release, using thermogravimetric analysis, X-ray computed tomography, the BET and mercury standard porosimetry methods. It is found that the peak value of graphite weight loss emerged at 573-673 K owing to nitrogen diffusion from graphite pores to outside when the system was subjected to vacuum. Macropore volume became larger while porosity for mesopores was smaller with temperature ranging from ambient temperature to 1073 K, which was primarily induced by coalescence of the subscale pores. It is suggested that the porous nuclear graphite should be first subjected to vacuum at 573-673 K to minimize the nitrogen and the radioactive 14°C before operation in HTGRs. Then, results on hydrogen diffusion show that the diffusible hydrogen and tritium could permeate into the coolant with diffusion coefficients of > 0.5 × 10⁻⁴ cm²·s⁻¹ at 50 bar. As a consequence, the freshly-generated diffusible tritium could release quickly to outside once formed, and an effective approach for minimizing the amount of radioactive tritium is to make the impurity contents extremely low in nuclear graphite and the coolant. Besides, both two- and three-dimensional observations indicate that macro and mesopore volume along with total porosity decreased with temperature at 50 bar on account of synergistic effects of applied compression strain, sharpened pore morphology, and non-uniform temperature distribution.
A Case Study on Utility of 18FDG-PET/CT Scan in Identifying Active Extra Lymph Nodes and Staging of Breast Cancer
Breast cancer is the most frequently diagnosed cancer worldwide, and a common cause of death among women. Various conventional anatomical imaging tools are utilized for diagnosis, histological assessment and TNM (Tumor, Node, Metastases) staging of breast cancer. Biopsy of sentinel lymph node is becoming an alternative to the axillary lymph node dissection. Advances in 18-Fluoro-Deoxi-Glucose Positron Emission Tomography/Computed Tomography (18FDG-PET/CT) imaging have facilitated breast cancer diagnosis utilizing biological trapping of 18FDG inside lesion cells, expressed as Standardized Uptake Value (SUVmax). Objective: To present the utility of 18FDG uptake PET/CT scans in detecting active extra lymph nodes and distant occult metastases for breast cancer staging. Subjects and Methods: Four female patients were presented with initially classified TNM stages of breast cancer based on conventional anatomical diagnostic techniques. 18FDG-PET/CT scans were performed one hour post 18FDG intra-venous injection of (300-370) MBq, and (7-8) bed/130sec. Transverse, sagittal, and coronal views; fused PET/CT and MIP modality were reconstructed for each patient. Results: A total of twenty four lesions in breast, extended lesions to lung, liver, bone and active extra lymph nodes were detected among patients. The initial TNM stage was significantly changed post 18FDG-PET/CT scan for each patient, as follows: Patient-1: Initial TNM-stage: T1N1M0-(stage I). Finding: Two lesions in right breast (3.2cm2, SUVmax=10.2), (1.8cm2, SUVmax=6.7), associated with metastases to two right axillary lymph nodes. Final TNM-stage: T1N2M0-(stage II). Patient-2: Initial TNM-stage: T2N2M0-(stage III). Finding: Right breast lesion (6.1cm2, SUVmax=15.2), associated with metastases to right internal mammary lymph node, two right axillary lymph nodes, and sclerotic lesions in right scapula. Final TNM-stage: T2N3M1-(stage IV). Patient-3: Initial TNM-stage: T2N0M1-(stage III). Finding: Left breast lesion (11.1cm2, SUVmax=18.8), associated with metastases to two lymph nodes in left hilum, and three lesions in both lungs. Final TNM-stage: T2N2M1-(stage IV). Patient-4: Initial TNM-stage: T4N1M1-(stage III). Finding: Four lesions in upper outer quadrant area of right breast (largest: 12.7cm2, SUVmax=18.6), in addition to one lesion in left breast (4.8cm2, SUVmax=7.1), associated with metastases to multiple lesions in liver (largest: 11.4cm2, SUV=8.0), and two bony-lytic lesions in left scapula and cervicle-1. No evidence of regional or distant lymph node involvement. Final TNM-stage: T4N0M2-(stage IV). Conclusions: Our results demonstrated that 18FDG-PET/CT scans had significantly changed the TNM stages of breast cancer patients. While the T factor was unchanged, N and M factors showed significant variations. A single session of PET/CT scan was effective in detecting active extra lymph nodes and distant occult metastases, which were not identified by conventional diagnostic techniques, and might advantageously replace bone scan, and contrast enhanced CT of chest, abdomen and pelvis. Applying 18FDG-PET/CT scan early in the investigation, might shorten diagnosis time, helps deciding adequate treatment protocol, and could improve patients’ quality of life and survival. Trapping of 18FDG in malignant lesion cells, after a PET/CT scan, increases the retention index (RI%) for a considerable time, which might help localize sentinel lymph node for biopsy using a hand held gamma probe detector. Future work is required to demonstrate its utility.
Latest Developments of the Gamma and Neutron Radiation Portal Monitor for Radioactive Source Identification
CAEN Sys is developing a new Radiation Portal Monitor (RPM) for the nuclear safety field based on the prototype that INFN developed within the SCINTILLA FP7 program and already qualified in the JRC ITRAP facility. The technology is based on a single detection system to discriminate between neutron and gamma signals, and the modular design combines the scintillating properties of a plastic scintillator with the high neutron absorbing cross-section of the Gadolinium. The prototype was benchmarked using the ANSI N42.35 and IEC 62244 standards demonstrating performances above these requirements. Results confirmed to have a system with higher performances than a primary detection RPM, not only with the ability to discriminate neutron and gamma signals, but also a gamma source categorization typical characteristic of a secondary detection RPM with spectroscopic capabilities. The prototype has been engineered to have a fully integrated system with position and temperature sensors, and a Graphical User Interface (GUI) for end-users, to satisfy commercial requirements. In parallel, the research work continued, and we are exploiting the possibility to improve further the gamma source identification based on the performances obtained with the prototype. The necessity to have a good gamma response over a wide energy range (from tens of keV up to MeV region) and the neutron discrimination were the starting points of the project, but the quality of the results obtained addressed us to study the feasibility to include the identification of a gamma source in real-time. It was firstly explored the sensitivity to gamma radiation, and it was demonstrated the possibility to categorize gamma sources in one of the pre-defined energy ranges. This first categorization helps the users to estimate the energy of the source and give a first indication of the type. The next step was to explore the possibility to distinguish between Special Nuclear Materials (SNM) and NORM (Naturally Occurrence Radioactive Materials). The paper will also describe the study performed, based on simulations of the system. This method was used to develop the strategy for the source identification (ID). The ID algorithm compares spectra obtained with real measurements with simulated or real source spectra. Here the description of all progress made and results obtained during the optimization phase of the RPM system are reported.
Processing and Characterization of Oxide Dispersion Strengthened (ODS) Fe-14Cr-3W-0.5Ti-0.3Y₂O₃ (14YWT) Ferritic Steel
Oxide dispersion strengthened (ODS) ferritic steels are amongst the most promising candidates for large scale structural materials to be applied in next generation fission and fusion nuclear power reactors. This kind of material is relatively stable at high temperature, possess remarkable mechanical properties and comparatively good resistance from neutron radiation damage. The superior performance of ODS ferritic steels over their conventional properties is attributed to the high number density of nano-sized dispersoids that act as nucleation sites and stable sinks for many small helium bubbles resulting from irradiation, and also as pinning points to dislocation movement and grain growth. ODS ferritic steels are usually produced by powder metallurgical routes involving mechanical alloying (MA) process of Y2O3 and pre-alloyed or elemental metallic powders, and then consolidated by hot isostatic pressing (HIP) or hot extrusion (HE) techniques. In this study, Fe-14Cr-3W-0.5Ti-0.3Y₂O₃ (designated as 14YWT) was produced by mechanical alloying process and followed by hot isostatic pressing (HIP) technique. Crystal structure and morphology of this sample were identified and characterized by using X-ray Diffraction (XRD) and field emission scanning electron microscope (FESEM) respectively. The magnetic measurement of this sample at room temperature was carried out by using a vibrating sample magnetometer (VSM). FESEM micrograph revealed a homogeneous microstructure constituted by fine grains of less than 650 nm in size. The ultra-fine dispersoids of size between 5 nm to 19 nm were observed homogeneously distributed within the BCC matrix. The EDS mapping reveals that the dispersoids contain Y-Ti-O nanoclusters and from the magnetization curve plotted by VSM, this sample approaches the behavior of soft ferromagnetic materials. In conclusion, ODS Fe-14Cr-3W-0.5Ti-0.3Y₂O₃ (14YWT) ferritic steel was successfully produced by HIP technique in this present study.
The Mitigation Strategy Analysis of Kuosheng Nuclear Power Plant Spent Fuel Pool Using MELCOR2.1/SNAP
Kuosheng nuclear power plant (NPP) is a BWR/6 plant in Taiwan. There is more concern for the safety of Spent Fuel Pools (SFPs) in Taiwan after Fukushima event. In order to estimate the safety of Kuosheng NPP SFP, by using MELCOR2.1 and SNAP, the safety analysis of Kuosheng NPP SFP was performed combined with the mitigation strategy of NEI 06-12 report. There were several steps in this research. First, the Kuosheng NPP SFP models were established by MELCOR2.1/SNAP. Second, the Station Blackout (SBO) analysis of Kuosheng SFP was done by TRACE and MELCOR under the cooling system failure condition. The results showed that the calculations of MELCOR and TRACE were very similar in this case. Second, the mitigation strategy analysis was done with the MELCOR model by following the NEI 06-12 report. The results showed the effectiveness of NEI 06-12 strategy in Kuosheng NPP SFP. Finally, a sensitivity study of SFP quenching was done to check the differences of different water injection time and the phenomena during the quenching. The results showed that if the cladding temperature was over 1600 K, the water injection may have chance to cause the accident more severe with more hydrogen generation. It was because of the oxidation heat and the "Breakaway" effect of the zirconium-water reaction. An animation model built by SNAP was also shown in this study.
A Real Time Expert System for Decision Support in Nuclear Power Plants
In case of abnormal situations, the nuclear power plant operators must follow written procedures to check the condition of the plant and classify the type of emergency. In this paper, we proposed a Real Time Expert System in order to improve operator’s performance in case of transient or accident with reactor shutdown. The expert system’s knowledge is based on the sequence of event of known accident and two emergency procedures of the Brazilian Pressurized Water Reactor (PWR) Nuclear Power Plant (NPP) and uses two kinds of knowledge representation: rule and logic trees. The results show that the system was able to classify the response of the automatic protection systems, as well as to evaluate the conditions of the plant, diagnosing the type of occurrence, recovery procedure to be followed, indicating the shutdown root cause and classifying the emergency level.
The Study of Ultimate Response Guideline of Kuosheng BWR/6 Nuclear Power Plant Using TRACE and SNAP
In this study of ultimate response guideline (URG), Kuosheng BWR/6 nuclear power plant (NPP) TRACE model was established. The reactor depressurization, low pressure water injection, and containment venting are the main actions of URG. This research focuses to evaluate the efficiency of URG under Fukushima-like conditions. Additionally, the sensitivity study of URG was also performed in this research. The analysis results of TRACE present that URG can keep the peak cladding temperature (PCT) below 1088.7 K (the failure criteria) under Fukushima-like conditions. It implied that Kuosheng NPP was at the safe situation.
Reduction of Plutonium Production in Heavy Water Research Reactor: A Feasibility Study through Neutronic Analysis Using MCNPX2.6 and CINDER90 Codes
One of the main characteristics of Heavy Water Moderated Reactors is their high production of plutonium. This article demonstrates the possibility of reduction of plutonium and other actinides in Heavy Water Research Reactor. Among the many ways for reducing plutonium production in a heavy water reactor, in this research, changing the fuel from natural Uranium fuel to Thorium-Uranium mixed fuel was focused. The main fissile nucleus in Thorium-Uranium fuels is U-233 which would be produced after neutron absorption by Th-232, so the Thorium-Uranium fuels have some known advantages compared to the Uranium fuels. Due to this fact, four Thorium-Uranium fuels with different compositions ratios were chosen in our simulations; a) 10% UO2-90% THO2 (enriched= 20%); b) 15% UO2-85% THO2 (enriched= 10%); c) 30% UO2-70% THO2 (enriched= 5%); d) 35% UO2-65% THO2 (enriched= 3.7%). The natural Uranium Oxide (UO2) is considered as the reference fuel, in other words all of the calculated data are compared with the related data from Uranium fuel. Neutronic parameters were calculated and used as the comparison parameters. All calculations were performed by Monte Carol (MCNPX2.6) steady state reaction rate calculation linked to a deterministic depletion calculation (CINDER90). The obtained computational data showed that Thorium-Uranium fuels with four different fissile compositions ratios can satisfy the safety and operating requirements for Heavy Water Research Reactor. Furthermore, Thorium-Uranium fuels have a very good proliferation resistance and consume less fissile material than uranium fuels at the same reactor operation time. Using mixed Thorium-Uranium fuels reduced the long-lived α emitter, high radiotoxic wastes and the radio toxicity level of spent fuel.
Droplet Entrainment and Deposition in Horizontal Stratified Two-Phase Flow
In this study, the droplet behavior of under horizontal stratified flow regime for air and water flow in horizontal pipe experiments from a 0.24 m, 0.095 m, and 0.0486 m size diameter pipe are examined. The effects of gravity, pipe diameter, and turbulent diffusion on droplet deposition are considered. Models for droplet entrainment and deposition are proposed that considers developing length. Validation for experimental data dedicated from the REGARD, CEA and Williams, University of Illinois, experiment were performed using SPACE (Safety and Performance Analysis Code for Nuclear Power Plants).
Influence of Thermal Ageing on Microstructural Features and Mechanical Properties of Reduced Activation Ferritic/Martensitic Grades
Reduced Activation Ferritic/Martensitic (FM) steels like EUROFER are of interest for first wall application in the future demonstration (DEMO) fusion reactor. Depending on the final design codes for the DEMO reactor, the first wall material will have to function in low-temperature mode or high-temperature mode, i.e. around 250-300°C of above 550°C respectively. However, the use of RAFM steels is limited up to a temperature of about 550°C. For the low-temperature application, the material suffers from irradiation embrittlement, due to a shift of ductile-to-brittle transition temperature (DBTT) towards higher temperatures upon irradiation. The high-temperature response of the material is equally insufficient for long-term use in fusion reactors, due to the instability of the matrix phase and coarsening of the precipitates at prolonged high-temperature exposure. The objective of this study is to investigate the influence of thermal ageing for 1000 hrs and 4000 hrs on microstructural features and mechanical properties of lab-cast EUROFER. Additionally, the ageing behavior of the lab-cast EUROFER is compared with the ageing behavior of standard EUROFER97-2 and T91. The microstructural features were investigated with light optical microscopy (LOM), electron back-scattered diffraction (EBSD) and transmission electron microscopy (TEM). Additionally, hardness measurements, tensile tests at elevated temperatures and Charpy V-notch impact testing of KLST-type MCVN specimens were performed to study the microstructural features and mechanical properties of four different F/M grades, i.e. T91, EUROFER97-2 and two lab-casted EUROFER grades. After ageing for 1000 hrs, the microstructures exhibit similar martensitic block sizes independent on the grain size before ageing. With respect to the initial coarser microstructures, the aged microstructures displayed a dislocation structure which is partially fragmented by polygonization. On the other hand, the initial finer microstructures tend to be more stable up to 1000hrs resulting in similar grain sizes for the four different steels. Increasing the ageing time to 4000 hrs, resulted in an increase of lath thickness and coarsening of M23C6 precipitates leading to a deterioration of tensile properties.
Functional Plasma-Spray Ceramic Coatings for Corrosion Protection of RAFM Steels in Fusion Energy Systems
Nuclear fusion, one of the most promising options for reliably generating large amounts of carbon-free energy in the future, has seen a plethora of ground-breaking technological advances in recent years. An efficient and durable “breeding blanket”, needed to ensure a reactor’s self-sufficiency by maintaining the optimal coolant temperature as well as by minimizing radiation dosage behind the blanket, still remains a technological challenge for the various reactor designs for commercial fusion power plants. A relatively new dual-coolant lead-lithium (DCLL) breeder design has exhibited great potential for high-temperature (>700oC), high-thermal-efficiency (>40%) fusion reactor operation. However, the structural material, namely reduced activation ferritic-martensitic (RAFM) steel, is not chemically stable in contact with molten Pb-17%Li coolant. Thus, to utilize this new promising reactor design, the demand for effective corrosion-resistant coatings on RAFM steels represents a pressing need. Solution Spray Technologies LLC (SST) is developing a double-layer ceramic coating design to address the corrosion protection of RAFM steels, using a novel solution and solution/suspension plasma spray technology through a US Department of Energy-funded project. Plasma spray is a coating deposition method widely used in many energy applications. Novel derivatives of the conventional powder plasma spray process, known as the solution-precursor and solution/suspension-hybrid plasma spray process, are powerful methods to fabricate thin, dense ceramic coatings with complex compositions necessary for the corrosion protection in DCLL breeders. These processes can be used to produce ultra-fine molten splats and to allow fine adjustment of coating chemistry. Thin, dense ceramic coatings with chosen chemistry for superior chemical stability in molten Pb-Li, low activation properties, and good radiation tolerance, is ideal for corrosion-protection of RAFM steels. A key challenge is to accommodate its CTE mismatch with the RAFM substrate through the selection and incorporation of appropriate bond layers, thus allowing for enhanced coating durability and robustness. Systematic process optimization is being used to define the optimal plasma spray conditions for both the topcoat and bond-layer, and X-ray diffraction and SEM-EDS are applied to successfully validate the chemistry and phase composition of the coatings. The plasma-sprayed double-layer corrosion resistant coatings were also deposited onto simulated RAFM steel substrates, which are being tested separately under thermal cycling, high-temperature moist air oxidation as well as molten Pb-Li capsule corrosion conditions. Results from this testing on coated samples, and comparisons with bare RAFM reference samples will be presented and conclusions will be presented assessing the viability of the new ceramic coatings to be viable corrosion prevention systems for DCLL breeders in commercial nuclear fusion reactors.
Degradation of Irradiated UO2 Fuel Thermal Conductivity Calculated by FRAPCON Model Due to Porosity Evolution at High Burn-Up
The evolution of volume porosity previously obtained by using the existing low temperature high burn-up gaseous swelling model with progressive recrystallization for UO2 fuel is utilized to study the degradation of irradiated UO2 thermal conductivity calculated by the FRAPCON model of thermal conductivity. A porosity correction factor is developed based on the assumption that the fuel morphology is a three-phase type, consisting of the as-fabricated pores and pores due to intergranular bubbles whitin UO2 matrix and solid fission products. The predicted thermal conductivity demonstrates an additional degradation of 27% due to porosity formation at burn-up levels around 120 MWd/kgU which would cause an increase in the fuel temperature accordingly. Results of the calculations are compared with available data.
Determination of Rare Earth Element Patterns in Uranium Matrix for Nuclear Forensics Application: Method Development for Inductively Coupled Plasma Mass Spectrometry (ICP-MS) Measurements
During the last 50 years, the worldwide permeation of the nuclear techniques induces several new problems in the environmental and in the human life. Nowadays, due to the increasing of the risk of terrorism worldwide, the potential occurrence of terrorist attacks using also weapon of mass destruction containing radioactive or nuclear materials as e.g. dirty bombs, is a real threat. For instance, the uranium pellets are one of the potential nuclear materials which are suitable for making special weapons. The nuclear forensics mainly focuses on the determination of the origin of the confiscated or found nuclear and other radioactive materials, which could be used for making any radioactive dispersive device. One of the most important signatures in nuclear forensics to find the origin of the material is the determination of the rare earth element patterns (REE) in the seized or found radioactive or nuclear samples. The concentration and the normalized pattern of the REE can be used as an evidence of uranium origin. The REE are the fourteen Lanthanides in addition scandium and yttrium what are mostly found together and really low concentration in uranium pellets. The problems of the REE determination using ICP-MS technique are the uranium matrix (high concentration of uranium) and the interferences among Lanthanides. In this work, our aim was to develop an effective chemical sample preparation process using extraction chromatography for separation the uranium matrix and the rare earth elements from each other following some publications can be found in the literature and modified them. Secondly, our purpose was the optimization of the ICP-MS measuring process for REE concentration. During method development, in the first step, a REE model solution was used in two different types of extraction chromatographic resins (LN® and TRU®) and different acidic media for environmental testing the Lanthanides separation. Uranium matrix was added to the model solution and was proved in the same conditions. Methods were tested and validated using REE UOC (uranium ore concentrate) reference materials. Samples were analyzed by sector field mass spectrometer (ICP-SFMS).
Transient Simulation Using SPACE for ATLAS Facility to Investigate the Effect of Heat Loss on Major Parameters
In order to obtain better initial conditions of ATLAS facility’s transient experiments, heat loss was simulated using SPACE code’s convection correlations and dialing factors. As simulating using a heat loss free input was a common practice; the facility was considered to be totally insulated, and the core power was reduced by the experimentally measured values of heat loss for the primary circuit to compensate the loss of heat in actual experiments, this study, however, takes heat loss into account throughout the simulation. The criterion of spotting the steady state with the newly introduced heat loss input was by having a steady constant water level in the pressurizer; multiple dialing factors were applied to the code’s correlation for the water level to be as prescribed. As it was reached, various other parameters were also to be met, such as the flow rates within the primary circuit, temperatures of various components in the primary and secondary side, and pressurizer pressure and steam generator secondary pressure. In addition to that, the simulated values had to satisfy an enthalpy balance equation of input and output; and so it has. The initial condition which accounts for heat loss yields more realistic outcome; as heat being leaked out of the system throughout a transient will alter many parameters corresponding to temperature and temperature difference in SPACE code simulation. For that; a Main Steam Line Break accident will be simulated using both the insulated system approach and the newly introduced heat loss input of the steady state. This study will show the significance of heat loss consideration in the processes of prevention and mitigation of various incidents and design basis accidents as it will give a better understanding of the behavior of ATLAS facility during both processes of steady state and major transient.
Mathematical Modelling of a Low Tip Speed Ratio Wind Turbine for System Design Evaluation
Vertical Axis Wind Turbine (VAWT) systems are becoming increasingly popular as they have a number of advantages over traditional wind turbines. The advantages are reliability, ease of transportation and manufacturing. These attributes could make these technologies useful in developing economies. The performance characteristics of a VAWT are different from a horizontal axis wind turbine, which can be attributed to the low tip speed ratio operation. To unlock the potential of these VAWT systems, the operational behaviour in a number of system topologies and environmental conditions needs to be understood. In this study, a non-linear dynamic simulation method was developed in Matlab and validated against in field data of a large scale, 8-meter rotor diameter prototype. This simulation method has been utilised to determine the performance characteristics of a number of control methods and system topologies. The motivation for this research was to develop a simulation method which accurately captures the operating behaviour and is computationally inexpensive. The model was used to evaluate the performance through parametric studies and optimisation techniques. The study gave useful insights into the applications and energy generation potential of this technology.