Reactor Site Research Articles

Ultrasonic degradation of selected dyes using Ti3C2Tx MXene as a sonocatalyst.

MXene, a new family of two dimensional materials, was utilized as a sonocatalyst in an ultrasonic treatment (US) process for removal of methylene blue (MB) and acid blue 80 (AB). The physico-chemical properties of MXene were characterized using scanning electron microscopy, transmission electron microscopy, porosimetry, and a zeta potential analyzer. Degradation of dyes by US was systemically investigated under several experimental conditions including: power density of US (45, 90, 135, and 180WL-1), frequency of US (28 and 970kHz), pH of dye solution (3.5, 7, and 10.5), solution temperature (293, 303, and 313K), and addition of hydroxyl radical promotor (H2O2) and scavenger (t-BuOH) to concentrations of 25mM. Based on the experimental results, the quantity of H2O2, which was used as an indicator of hydroxyl radical concentration, was an important factor in determining the degradation rate of MB and AB in this US study. Additionally, synergetic indices for removal of both dyes were higher than 1.0 in all cases, indicating the outstanding efficiency of MXene as a sonocatalyst in the US reactor for removal of both, due to an increase in both (i) the quantity of H2O2 in the US reactor and (ii) active sites for adsorbates from dispersion effects. A stability test on MXene in the US process was conducted using X-ray diffraction and five-cycle recycling performance tests. Based on our experimental data, MXene can be utilized as a sonocatalyst in the US process for a high removal rate for dyes (e.g., MB).

Comparison of two methods to determine 129I in charcoal cartridge samples by AMS

To investigate the short- and long-term accumulated emissions from a nuclear facility, charcoal air filter cartridges were exposed at a reactor site for seven days. The short-lived 131I adsorbed in these filters was measured by gamma counting immediately after the samples were retrieved. The very long-lived 129I was analyzed later by AMS. For this purpose, two methodologies were developed: an extraction method and direct analysis, i.e., inserting an aliquot of the filter carbon directly into the target piece. The former method used a system designed to avoid sample loss resulting from the high I2 species volatility. Iodine-125 was used as a tracer to monitor the efficiency of this extraction process. The direct method was based on our experience with charcoal in the analysis of 14C in Cs sputter ion sources. In this direct analysis, charcoal samples were first spiked with a known quantity of stable iodine (127I), in an excess over the quantity naturally present in the filter, and then pressed into targets. In this way, 129I concentrations from all the charcoal samples were quickly determined. In both methods, it was found that the atom concentration of 129I in these samples exceeded that of the 131I by 2–3 orders of magnitude. These studies demonstrate the effectiveness of the charcoal capture method as a diagnostic tool for monitoring the emissions from nuclear facilities. This approach may be extended for the detection of other long-lived radioactive isotopes, which would further expand the scope of such monitoring studies.

Light-activated catalyst makes syngas greener

Synthesis gas, or syngas, is a feedstock for multiple industrial processes, including the Haber-Bosch process for making ammonia. Syngas is a mixture of carbon monoxide, carbon dioxide, and hydrogen. Gasification plants, which make syngas from coal, natural gas, and other sources, use catalysts to break apart fuels at around 800 °C. Naomi Halas and coworkers at Rice University have now come up with a greener way to make syngas (Nat. Energy 2020, DOI: 10.1038/s41560-019-0517-9). Using light to activate a nanoparticle catalyst made of copper antennas and single ruthenium atom reactor sites, the team was able to make syngas at around 200 °C. Their system is stable up to 50 h with 99% selectivity. Current industrial syngas catalysts are less robust because they’re susceptible to coking, a C–C bond-forming process that leads to the growth of a carbonaceous material that can block active sites and poison the catalyst. “By having isolated

Advanced Nuclear Physical Analytical Techniques

Among the modern analytical methods, the nuclear analytical techniques, and especially neutron activation analysis, pay an important role in accurate and sensitive analysis. The capabilities of this technique and its application range have broadened lately with the development and employment of a combination of special more or less novel sub-techniques. These techniques are (a) bss-free counting gamma-ray spectrometry, (b) radioactive decay compensation and (c) repeated activation and measurement. The first technique permits short-time measurements at high count rates in multielement activation analysis with a wide nuclide concentration and half-life range, while by the other two techniques the counting statistics can be improved considerably especially in short-lived nuclide analysis. Thus, because of the high throughput rate, more customers can be served even far from the reactor site by mailing the samples and getting the analytical results back.

Amphiphilic Dendrimer as Reverse Micelle: Synthesis, Characterization and Application as Homogeneous Organocatalyst

The core and surface terminal groups are the two main catalytic sites in a dendrimer. In most of the reported examples, the catalytic sites in dendritic catalysis are the surface terminal functional groups. This perspective article concerned with the dendrimer based catalysis, involving these two catalytic sites and the dendrimer cavities. The interior cavities provide the nanoscale reactor sites, by creating reverse micelle like appearance for catalysis. In exploring the significant achievements in this area, a low generation PAMAM dendrimer with amphiphilic nature, having a polymeric core with large number of pendent amino groups was synthesized and concentrated its catalytic activity. The key features with respect to positive and/or negative catalytic activity was highlighted by synthesizing various aryl and heteroaryl 2-substituted benzimidazoles. The synthesized dendritic organocatalyst was proved to be amazingly reactive and gave high yield of products within a few minutes at room temperature with low catalyst loading. Here, a new stable hemiaminal, the species rarely been detected and much less isolated in bulk, was obtained during the synthesis of benzimidazoles. Moreover, this is the first reported method for the synthesis of benzimidazoles, using the homogeneous PAMAM dendrimer as a basic organocatalyst.

Neutron capture cross sections of 74, 76, 78, 80, 82Se

The cross sections for radiative neutron capture by the 5 stable even-mass isotopes of natural selenium have been measured by observation of the \(\gamma\) rays emitted by the 7 radioactive isomeric and ground states produced following neutron capture. By irradiating the Se samples in reactor sites having different compositions of thermal and epithermal neutrons, consistent values of the thermal cross sections and resonance integrals have been obtained. Systematic variations in the deduced cross sections have been minimized by observing in each sample the decays of several different Se nuclides, thus providing overlapping and redundant values for comparison. The half-life of 79mSe has been measured to be 3.884(9)h. Ratios of cross sections of isomeric and ground states in 77, 79, 81, 83Se are compared with a statistical model.

Human factors engineering and participatory design in the project of a nuclear research reactor control centre

The Human Factors Engineering (HFE) program is an essential aspect for the design of nuclear installations. The overall aim of the HFE program is the improvement of the operational reliability and safety of plant operation. The HFE program main purpose is to ensure that human factor practices are incorporated into the plant design, emphasizing man-machine interface issues and design improvement of the nuclear reactor Control Centre. The Control Centre of nuclear reactor is a combination of control rooms, control suites and local control stations, which are functionally connected and located on the reactor site. The objective of this paper is to present a participatory design approach, including human factor issues, for the Control Centre of a nuclear research reactor used to produce radioisotopes and for nuclear research. The design approach is based on participatory design principles, using human factor standards, ergonomic guidelines, and the participation of a multidisciplinary team during all design phases. Using the information gathered, an initial sketch 3D of the Control Centre was developed.

A study of the protective actions for a hypothetical accident of the Bushehr nuclear power plant at different meteorological conditions.

In this work, protective actions have been studied assuming a hypothetical severe accident of the Bushehr nuclear power plant at different meteorological conditions. Simulations of the atmospheric dispersion of accidental airborne releases were performed using the RASCAL code. Total effective dose equivalent (TEDE) and thyroid dose received by members of the public living within a radius of 40km around the reactor site were calculated for various atmospheric stability classes and weather conditions. According to the results of the dose assessment and by following the protective action guide of the Environmental Protection Agency (EPA), the critical zone and appropriate protective actions were determined depending on various metrological conditions. It was found that, for atmospheric stability class F and calm weather conditions, the maximum distance from the site of release for which TEDE is greater than the corresponding dose limit and for which sheltering or evacuation response actions are required, is 11km. For the same weather conditions, the corresponding maximum distance for which iodine prophylaxis is required is 32km. Based on the present simulations, it can be concluded that the metrological condition has a great influence on the radionuclide atmospheric dispersion and, consequently, on the critical zone where protective actions are required after the assumed accident condition.

Przesiedleńcy ze strefy wykluczenia po katastrofie atomowej w Czarnobylu: problemy z zadomawianiem się w nowej przestrzeni

After the Chernobyl nuclear disaster, the so-called Chernobyl Exclusion Zone, covering the area within the 30 km radius from the nuclear reactor site, was established (also in Belarus). All people were evacuated from the zone and displaced to “clean” territory. For the purpose of the current paper, reports of witnesses from the documental prose, dialectal texts, publications and Belarus Archives of Oral History were analysed in an attempt to find in their narratives the answers to the following questions: who the people called “chernobyltsy” (literally: the Chernobyl ones) are and in what way they were and are seen by other people, especially just after the disaster; what the reaction of the people to the process of evacuation was; what the indigenous people during the evacuation took with them; what they left at their homes and why; what their attitude towards new comfortable houses and flats was; in what way they tried to adapt themselves to new environment; where they buried the dead; and whether they successfully settled in their new places of residence. Having analysed the reports and arranged them according to the phenomenology of the area based on the report of H. Buczynska-Garewicz, the category of “rootedness” of S. Weil and J. Tischner, A. van Gennep’s theory of rites of passage”, the author concludes that a great number of “chernobyltsy” are deeply rooted in their motherland; an approach that excludes the possibility of expanding the definition of “home” and “the sense of settlement”. Those people usually suffer at their new places of residence and sometimes return home. The only strategies favouring their acceptance of a new place that have been observed are focusing on work, especially on working the land (an allotment), or focusing on the health of their children.

Study of Dispersion Hazard Potential of The LPG Stations Around the RDE Site in Rainy and Dry Season

STUDY OF DISPERSION HAZARD POTENTIAL OF THE LPG STATIONS AROUND THE RDE SITE IN RAINY AND DRY SEASON. There are two LPG station (SPPBE) which are the depot of filling, storage and distribution of Liquid Petroleum Gas (LPG) namely PT. BM and PT. ISR which the distance each are 2,995 and 4,141 km from Experimental Power Reactor (RDE) site with capacity 15 and 30 tons. LPG station is a stationary source, which is one aspect of the external human induced events that need to be analyzed in the preparation of site evaluation reports to obtain site permits. Hazard potential that may occur from the depot LPG are fire, explosion and dispersion of hazardous and toxic gas. The release of LPG due to valve leakage which is then dispersed at a certain dose has potentially harmful to health, even death to the population around the RDE site. The purpose of the study was to know the effect of seasons (rainy and dry) to the potential hazard of LPG dispersion from LPG truck tank valve to the around RDE site. The method of study are collection the atmospheric data such as wind direction and speed, temperature and humidity, collection the station LPG characteristic, such as mass of gas, diameter and length of tank, and valve diameter, etc. The atmospheric data was obtained from Pondok Betung Climatology Station, in dry, transition, and rainy seasons, furthermore data was analyzed using ALOHA software version 5.4.5. The results show dispersion from LPG release due to valve leakage from PT. BM and PT. ISR around the RDE site, in the dry season (April), the transition (January and July) as well as the rainy season (October) does not hazardous to the RDE site. Maximum threat zone occurs in dry season at April (wind speed 1.54 m/s), which reaches radius 179 m with airborne LPG concentration 5500 ppm, radius 111 m with concentration 17000 ppm and radius 71 m with concentration 53000 ppm.

Performance of a full scale prototype detector at the BR2 reactor for the SoLid experiment

The SoLid collaboration has developed a new detector technology to detect electron anti-neutrinos at close proximity to the Belgian BR2 reactor at surface level. A 288 kg prototype detector was deployed in 2015 and collected data during the operational period of the reactor and during reactor shut-down. Dedicated calibration campaigns were also performed with gamma and neutron sources. This paper describes the construction of the prototype detector with a high control on its proton content and the stability of its operation over a period of several months after deployment at the BR2 reactor site. All detector cells provide sufficient light yields to achieve a target energy resolution of better than 20%/√E(MeV). The capability of the detector to track muons is exploited to equalize the light response of a large number of channels to a precision of 3% and to demonstrate the stability of the energy scale over time. Particle identification based on pulse-shape discrimination is demonstrated with calibration sources. Despite a lower neutron detection efficiency due to triggering constraints, the main backgrounds at the reactor site were determined and taken into account in the shielding strategy for the main experiment. The results obtained with this prototype proved essential in the design optimization of the final detector.

Design Considerations for Secure Wireless Sensor Communication Systems in Harsh Electromagnetic Environments of Nuclear Reactor Facilities

Wireless sensors can potentially play a significant role in safety, efficiency, and reliability of the instrumentation and control process in current and next generation nuclear power reactors. While conventional narrowband wireless sensors have shown a certain level of success in some nuclear power plants (NPPs), the radio frequency (RF) propagation challenges posed by the heavy metallic and cluttered environment of NPPs has prevented their widespread use in such operations. These challenges include RF wave propagation in harsh (reflective, absorptive, cluttered) environments, data security issues, and RF interference to and from other devices in the vicinity of a nuclear reactor core. In this paper, first we address how ultrawideband (UWB) RF technology can complement the narrowband (i.e., WiFi) solutions that have been used in some NPPs by providing an alternative solution in addressing the signal propagation issues in such electromagnetically harsh environments. Second, we discuss and present the UWB software simulation results on multipath harsh environments, and then address the data security issues. In the final sections of the paper, we present the experimental results of using UWB signaling in a representative harsh environment conducted at the Massachusetts Institute of Technology research reactor site. We plan to develop the UWB communications hardware based on the results of this paper and report on its performance in the field with emphasis on the security aspects of the system in a subsequent paper.

Parameters Affecting 137Cs Migration within Soil Profile

SEVERAL studies have been carried out on the adsorption, distribution and migration of 137Cs within soil profile in the vicinity of the Nuclear Research Center, Atomic Energy Authority, Egypt, in addition to Ismailia Canal areas. The soil physicochemical and mineralogical characteristics were carried out and indicated that the soil samples consist mainly of sand (quartz) and silt fractions. The kinetics of caesium adsorption and its related adsorption isotherms for the tested soils were also studied. The sorption of 137Cs by soil minerals markedly affects its migration rate. The natural background of both locations under study indicated that the 137Cs concentration in the reactor site were found to be 30.82 Bq/kg while that at the canal site was found to be 0.87 Bq/kg. 137Cs in the reactor site may be originated from the fallout and from the external contamination which affected the background level. The vertical distribution and the migration rate of 137Cs have been studied for the soil profiles in both locations. These rates were found to be 0.056 and 0.031 cm/year for the reactor and canal site respectively.

Rainfall flood hazard at nuclear power plants in India

ABSTRACTFlood hazard evaluation is an important input for Nuclear Power Plants external events safety studies. In the present study, flood hazard at various nuclear sites in India due to rainfall has been evaluated. Hazard estimation is a statistical procedure by which rainfall intensity versus occurrence frequency is estimated from historical records of rainfall data and extrapolated with asymptotic extreme value distribution. Rainfall data needed for flood hazard assessment are daily annual maximum rainfall (24 h data). The observed data points have been fitted using Gumbel, power law and exponential distribution, and return period has been estimated. To study the stationarity of rainfall data, a moving window estimate of the parameters has been performed. The rainfall pattern is stationary in both coastal and inland regions over the period of observation. The coastal regions show intense rainfall and higher variability than inland regions. Based on the plant layout, catchment area and drainage capacity, the prototype fast breeder reactor (PFBR) site is unlikely to be flooded.

Radiological effect of cask-drop accident in open-pool spent fuel storage

This study investigates the radiological risk that may occur during transport of cobalt device after irradiation in open-pool-type reactor from the reactor core to the cobalt cell. The cobalt transport process depends on using heavy shielded cask of 3500 kg weight under the water surface of the spent fuel storage pool which may cause a load drop accident. The load drop accident in the spent fuel storage pool would result in damage of 48 spent fuel elements (FE) maximally. Conservative evaluation for the amount of fission products release from the damage of the spent FE was considered in this study by assuming that all spent FE was recently stored at the time of accident. Consequently, the resulting radiation dose distribution was calculated around the reactor building depending on the meteorological data of the reactor site. GENII-2 code was used to estimate the individual effective dose distribution around the reactor. The result shows that the receptor who located at 1500 m from the reactor building in the south-east direction will receive the maximum individual effective dose of 0.11 Sv.

STUDY OF WATER QUALITY AT WEST SUNTER RESERVOIR, NORTH JAKARTA BASED ON PHYSICAL AND CHEMICAL PARAMETERS

Aim: This study is to determine effect of contamination level of population settlement activity on water quality in the reservoir at west Sunter by measurements of reservoir water quality in the form of physical and chemical parameters including the identification of the kinetics of COD. Methodology and Result: This research was conducted in April to August 2016 with 11 sampling locations. Research indicated that the reservoir had experienced heavy pollution with DO value ranging between 1.05 mg/L-1.52 mg/L with a minimum quality standard of 3 mg/L, Ammonia 0.13 mg/L-0.38 mg/L with a quality standard of 0.02 mg/L, surfactant (1.74 mg/L-4.63 mg/L) quality standard of 0.2 mg/L and phosphate (0.8 mg/L-1:19 mg/L) to the quality standard of 1 mg/L. Reservoir also were polluted with heavy loads inclusion of organic content with COD values ranging between 112.58 mg/L - 196.39 mg/L. The source of pollution in the west Sunter reservoir was derived from domestic sewage. Based on the average constant value it takes 16.96 hours to reduce the COD from 120 mg/L to 40 mg/L. Natural Retention time at West Sunter Reservoir is 13.6 days. Conclusion, significance and impact study: Domestic waste has obviously contaminated West Sunter reservoir from its physically green color, high organic content with high COD values, low DO, high oil and fat content, and high phosphate levels. These polluted compounds must be removed before spreading to the next water body. If the reservoir is considered as a reactor site, then the reservoir must be able to remove the contaminants before disposal.

Neutron imaging and tomography with MCPS

A neutron imaging detector based on neutron-sensitive microchannel plates (MCPS) was constructed and tested at beamlines of thermal and cold neutrons. The MCPS are made of a glass mixture containing 10B and natural Gd, which makes the bulk of the MCP an efficient neutron converter. Contrary to the neutron-sensitive scintillator screens normally used in neutron imaging, spatial resolution is not traded off with detection efficiency. While the best neutron imaging scintillators have a detection efficiency around a percent, a detection efficiency of around 50% for thermal neutrons and 70% for cold neutrons has been demonstrated with these MCPS earlier. Our tests show a performance similar to conventional neutron imaging detectors, apart from the orders of magnitude better sensitivity. We demonstrate a spatial resolution better than 150 μm. The sensitivity of this detector allows fast tomography and neutron video recording, and will make smaller reactor sites and even portable sources suitable for neutron imaging.

Predicting the cost of the consequences of a large nuclear accident in the UK

Abstract Nuclear accidents have the potential to lead to significant off-site effects that require actions to minimise the radiological impacts on people. Such countermeasures may include sheltering, evacuation, restrictions on the sale of locally-grown food, and long-term relocation of the population amongst others. Countries with nuclear facilities draw up emergency preparedness plans, and put in place such provisions as distributing instructions and iodine prophylaxis to the local population. Their plans are applied in simulated exercises on a regular basis. The costs associated with emergency preparedness and the safety provisions to reduce the likelihood of an accident, and/or mitigate the consequences, are justified on the basis of the health risks and accident costs averted. There is, of course, only limited actual experience to indicate the likely costs so that much of the costing of accidents is based on calculations. This paper reviews the methodologies used, in particular the approach that has been developed in the UK, to appraise the costs of a hypothetical nuclear accident. Results of analysing a hypothetical nuclear accident at a fictitious reactor site within the United Kingdom are discussed in relation to the accidents at Three Mile Island 2, Chernobyl and Fukushima Dai-ichi.

Determining Optimal Used Fuel Allocation Strategies

A key challenge in fulfilling the U.S. federal government’s obligations under the Nuclear Waste Policy Act is in the transition of used nuclear fuel (UNF) storage away from at-reactor storage and to a consolidated interim storage facility (CISF). The default strategy (Standard Contract) for the U.S. Department of Energy is to use the oldest fuel first (OFF) allocation strategy, which would entail the federal government prioritizing UNF shipments based on fuel discharge date with the option to prioritize shutdown sites. This may not be the most cost-efficient model given the extensive amount of UNF already at reactor sites. Currently, there is no way to preemptively remove fuel from sites that may be close to shutdown or have a higher storage or potential storage cost. As wet storage pools at reactors continue to fill to capacity at operating reactors, the backlog of UNF shipments to the CISF places additional pressure on operators to expand at-reactor dry storage capacity, thus adding to total system costs.An essential aspect of this transition is in developing appropriate analytical tools to evaluate the effect of factors such as fuel shipment prioritization, logistics, and associated expenses. Examples of this would include evaluating fuel offloading prioritization strategies (OFF versus shutdown sites first), strategies to minimize transfer of UNF to dry storage (i.e., through direct shipment from cooling pools to the CISF), etc.By applying integer programming techniques, it is possible to make a rigorous analytical determination of a UNF removal allocation strategy that minimizes the total number of shutdown reactor years (SRYs). Our findings indicate that an optimal unloading strategy can result in a threefold reduction in total system SRYs compared with an OFF-based queue, for a systemwide savings of about $8 billion.

Assessment of secular equilibrium and determination of natural and artificial radionuclide concentrations in the zone surrounding the site of the first nuclear reactor in Jordan

High-resolution gamma-spectrometry and ICP-MS measurements were utilized to confirm the validity of secular equilibrium among the identified natural radioactive progeny of 238U and 232Th series. The measurements of 238U, 232Th, and 40K concentrations, in the soil within 2 km range around the first nuclear reactor in Jordan, were close to the worldwide average levels. Among artificial radionuclides, only 137Cs was detected but with very low traces. The dose rate and radiological hazard parameters were found to be close to worldwide average values and below the recommended limits. Our results indicate that secular equilibrium is unperturbed within and around the uncontaminated reactor site.