Area Monitor Research Articles

Digitization of Microfinancing Operational Procedures

This study delves into the crucial theme of enhancing income-generating capacity, recognizing its pivotal role in meeting basic human needs and catalyzing prosperity within communities. The motivation behind this research lies in addressing this significant issue through the lens of microfinancing, a viable solution to the identified problem. Focusing on a group of microfinancing units strategically located in North Cotabato, Philippines, this study follows the experiences of personnel within these units. The integration of digitization into their operational procedures aimed to elevate service efficiency, providing a contemporary solution to the challenges faced. The research methodology employed was qualitative, utilizing purposive sampling to gain in-depth insights. Observations, interviews, and informal conversations with key informants, including loan officers, branch managers, and an area monitor, were integral to the inquiry. Both physical and electronic documents played a crucial role in enriching the data collected. Specifically, the study aimed to identify both digitized and non-digitized operational procedures, elucidate the experiences of microfinancing employees concerning these procedures, observe the effects of digitization, and pinpoint immediate improvements resulting from digitization. Key conclusions emerged from the study. Notably, loan processing was identified as the sole procedure yet to undergo digitization. The digitized procedures, encompassing data management, loan disbursement, repayment tracking, and financial reporting, demonstrated notable advantages, characterized by increased speed and reduced errors. Importantly, personnel satisfaction with digitized operational procedures witnessed a positive upturn. Identified improvements that underscore these findings include digitizing loan processing, optimization of loan officers' client-handling capacity, and fortifying institutional data security.

Assessment the Air Pollution at the some Areas in Salahaddin Governorate - Iraq

The study was conducted in Salah al-Din Governorate - northern Iraq, during the period from 1/11/2021 to 30/3/2022, with the aim of evaluating the level of pollutants in the atmospheric air for three areas: Al-Boutama (rural area), Baiji refinery (oil industry area), and the city of Tikrit (a densely populated area), the concentrations of toxic gases were measured: sulfur dioxide (SO2), nitrogen oxide (NO), nitrite (NO2), hydrochloric acid (HCl), hydrogen fluoride (HF), carbon dioxide (CO2) and carbon monoxide. (CO), using the Wolf Pack Modular Area Monitor. The results were as follows: Significant differences were found between the study sites for all the variables tested, and Baiji refinery recorded the highest values for each of the variables SO2, NO, NO2, HCl and HF, which amounted to 2.54, 8.28, 6.4, 10.62 ppm and 0.42 mg/ m3, respectively. November 2021 recorded the highest concentrations of HCl and CO compounds with averages of 7.6 ppm and 18.37 mg/m3, respectively, while January 2022 recorded the highest averages of NO2, HF and CO2, which amounted to 5.27 ppm and 0.4 mg/m3 and 791 ppm, respectively, the interaction between Baiji refinery site and the months of November, December and January recorded the highest concentrations of pollutants amounting to 4.4 ppm for SO2 concentration, 11.2 ppm for NO concentration, 7.9 ppm for NO2 concentration, 13.8 ppm for HCl concentration, and 0.7 mg/m3 for HF concentration, and 16.5 ppm for VOCs concentration.

Neutrons produced in a 12 MV LINAC working in electron mode

The electroneutron spectrum was estimated with Monte Carlo methods using two source terms. In addition, the electroneutron spectrum was measured using a passive Bonner sphere spectrometer. With estimated spectra were calculated, per history, the total neutron fluence, the ambient dose equivalent, the ambient dose, the mean energy, and the spectrum-averaged neutron fluence-to-dose conversion coefficients. With the measured spectrum were calculated, per monitor unit, the total neutron flux, the ambient dose equivalent; additionally, were calculated the mean energy and the spectrum-averaged neutron fluence-to-dose conversion coefficients. Doses were calculated using the fluence-to-dose conversion coefficients from the ICRP 74, LB 6411 neutron area monitor and those of the ambient dose, recently defined by the ICRP/ICRU joint publication. To 1 m from the isocenter calculations and measurements were carried out around a 12 MV linear accelerator for medical using with a solid water phantom.

Computational spectrometry and dosimetry, of neutrons and γ rays, outside a subcritical nuclear reactor with water and polyethylene moderators

An external neutron source is necessary to start up and run a subcritical thermal nuclear reactor. The external neutron source can be a neutron generator, a radionuclide having spontaneous fission or an isotopic neutrons source. Turning off the neutron generator or removing the neutron source shuts down the reactor. In subcritical reactors for teaching, fuel tubes and items for irradiation can be inserted or removed even when the reactor is running because are radiological safe. The aim of this work was to estimate, using Monte Carlo methods, the neutron and γ-ray spectra at five sites outside a Subcritical thermal neutron reactor Nuclear Chicago model 9000 using water and polyethylene moderators. In addition, the ambient dose equivalent and the ambient dose, for neutrons and γ-rays were calculated. For neutrons, the ambient dose equivalent were calculated with the neutron fluence-to-ambient dose equivalent coefficients from the ICRP 74 and the neutron fluence-to-ambient dose response of the neutron area monitor LB 6411; while, the ambient dose was calculated using the ICRU 95 neutron fluence-to-ambient dose coefficients. For γ-rays, was calculated the ambient dose equivalent using the ICRP 74 photon fluence-to-ambient dose equivalent coefficients; while, the ambient dose was calculated with the photon fluence-to-ambient dose coefficients from the ICRU 95. Outside the reactor, neutron spectra have fast, epithermal and thermal neutrons, and gamma ray spectra have γ rays induced in the inelastic scattering and capture of neutrons with the reactor materials. Neutron and γ-ray fluence rates and dose rates are larger in the water-moderated reactor. The ambient dose rates, due to neutrons and γ-rays, are smaller than the ambient dose equivalent, because the ambient dose resembles the maximum effective dose.

Neutron area monitor for ambient dose equivalent (H*(10)) and effective dose (EAP) using prompt gamma rays induced in high density polyethylene with metallic external layer

Estimations of neutron ambient dose equivalent, H∗(10) and effective dose (anterior-posterior), EAP using measured prompt gamma intensities are theoretically investigated for several prompt gamma generating detection systems, each consisting of a cylindrical high density polyethylene (HDPE) covered with a layer of different materials. The neutron induced prompt gamma intensities are measured using a NaI(Tl) detector with lead as shielding. Materials like cadmium, iron, nickel, zinc, copper and lead are used as an outer layer covering the HDPE cylinder. A linear combination of the distributions of the prompt gamma intensities emitted from these materials at different incident neutron energies are subjected to multiple linear regression analyses to fit the energy dependent neutron fluence to dose conversion coefficients (DCC) as provided by the International Commission on Radiological Protection (ICRP). The aim of the present work is to explore the suitability of other elements to replace boron in the earlier used borated HDPE in the system to overcome certain practical difficulties. Different neutron energy distributions in the energy range from about 0.01 eV to 20 MeV are selected to validate the dose estimation. Among the materials used, iron, nickel, zinc and copper show promising results for the estimation H∗(10) as well as EAP as a replacement of boron for neutron energy distributions with predominant high energy components.

Conceptual design of a new multi-purpose, passive and flexible neutron area monitor with two configurations

In proton therapy facilities, high-energy neutron fields are generated as a consequence of (p,Xn) nuclear interactions. UPM has designed and manufactured, for these facilities, a new flexible and passive neutron area monitor with two configurations: extended-range and conventional. The conceptual design phase of this monitor is reported in this work, presenting the main functional and technical features of the monitor, including the response functions of its two configurations. These response functions are compared with those of some widely-used monitors.

Comparison of PHITS2.88 and MCNP6.1 for the characterization of a LUPIN-II neutron area monitor

The use of different nuclear data libraries and physics models can be a source of discrepancies in neutron transport simulation. Different Monte Carlo simulation toolkits can be used to characterize neutron monitors, these codes usually employ by default different nuclear data libraries and physics models. This work presents, for the first time, a comparison of MCNP and PHITS for the characterization of a LUPIN-II neutron rem-meter. The most significant discrepancies between the codes have been found around 100 MeV.

Determination of energy correction coefficient for neutron albedo dosimeter. Three approaches

To accurately determine dose, albedo TLD dosimeters require application of a neutron energy correction factor (NECF). In this paper, the results from three different methods were used to determine NECFs. In the first method, NECF values were calculated as a ratio of the known “conventional true” dose and the neutron dose reported by albedo dosimeter(s) irradiated on a phantom. The results from this direct method were utilized for verification of two other methods. The second method used a specially designed device called the Neutron Area Monitor (NAM) model 5. The third method of NECF determination was based on the response of a BF3 tube under different levels of moderation. The results of this work showed that the Navy-developed Neutron Area Monitor (NAM) model 5 measures NECF values correctly when irradiated with 252Cf (bare and D2O-moderated). This was an important result, because it validated the use of the NAM-5 in nuclear power and marine propulsion reactors, as their neutron energy spectrum is approximated by the field of a Cadmium-coated 30 cm diameter steel sphere of D2O moderated 252Cf. In the case of a broad neutron energy spectrum the 6Li response to the small amount of slow neutrons outweighs its response to the fast neutrons overwhelmingly responsible for the neutron dose rate. Based on the obtained results we can conclude that NAM-5 provides correct NECF determination if the fraction of slow neutrons in the neutron dose rate is more than 20%. A satisfactory correlation between the NECF determined by NAM-5 and the multi-moderated BF3 tube method was observed, but the multi-moderated BF3 tube values were ∼30% lower than those obtained by the NAM-5 at the same locations for broader neutron energy spectrum locations.

Application of an algebraic methodology for the combination of Berthold LB6411 and WENDI-II for neutron area monitoring in D-T neutron generators and fusion facilities

Neutron area monitors do not often have a good adjustment of their dose response functions to the ICRP74 neutron fluence-to-H*(10) conversion function between 10 and 20 MeV. The objective of this work is to establish a methodology to combine the dose response functions of Berthold LB6411 and WENDI-II, adjusting this combined function to the ICRP74 conversion function: this combination shows an almost perfect adjustment between 0.5 and 20 MeV. Thus, this article presents an easy and cheap alternative to the recalibration in D-T generators.

Application of area monitors and scintillating detectors in the development of CBRN defense reconnaissance vehicles

This study describes the use of ionizing radiation detectors in a Brazilian Army reconnaissance vehicle (RV), used in area monitoring actions based on the occurrence of a radiological event. Two scintillating detectors, NaI and LaBr3, were used to carry out field measurements of surveying for different distances from a simulated detector position in the vehicle, in order to compare the energy spectra obtained in each measurement system. The Brazilian Army Technological Center (Centro Tecnológico do Exército - CTEx) region was chosen to carry out the measurements. The meteorological information was annotated to verify the conditions into which the experimental apparatus would be submitted. The results suggest that both the NaI and LaBr3 scintillating detector could be used in radiological emergency response RV, offering satisfactory responses in the gamma radiation detection. However, the NaI detector was chosen considering the wide network of technical assistance and its low operational cost. The relevance of this investigation shows the importance of planning responses in emergency situations and the influence of efficient instrumentation in the measurement processes.

Enhanced denosing autoencoder aided bad data filtering for synchrophasor-based state estimation

Synchrophasor-based linear state estimation (LSE) attracts great interests in the past decade due to its high accuracy and straightforward computation and has been the foundation of many wide area monitor system (WAMS) applications. However, an increasing number of data quality issues has been reported, among which bad data can significantly undermine the performance of LSE and many other WAMS applications it supports. In practice, bad data filtering can be very challenging due to a variety of factors including constrained computation time, nonuniform and changing patterns, etc. This paper presents an enhanced denosing autoencoder (DA) aided bad data filtering scheme to pre-process PMU measurements for LSE. Bad data are firstly identified by a classifier module of the proposed DA and then recovered by autoencoder module of the proposed DA. The proposed methodology has two distinct features: 1) it not only identifies bad data based on adaptive learning but also recovers them, including the ones on critical measurements; 2) the algorithm is lightweight and can be deployed on individual PMU level to achieve maximum parallelism and high efficiency, making it suitable for real-time processing. Performance of the proposed methodology is validated and demonstrated through numerical experiments conducted using both simulated and real-world phasor measurements.

HOLDER DESIGN EVALUATION OF GAMMA AREA MONITOR IN SOLID RADIOACTIVE WASTE INTERM STORAGE ROOM OF RADIOMETALURGY INSTALLATION

HOLDER DESIGN EVALUATION OF GAMMA AREA MONITOR IN SOLID RADIOACTIVE WASTE INTERM STORAGE ROOM OF RADIOMETALURGY INSTALLATION . Radiometalurgy Installation (RMI) is one of Non-Reactor Nuclear Installation where post-irradiation examination (PIE) take place, which give gamma radiation ranged from one µSv/h to the order of mSv/h. The PIE activity produce high-activity radioactive waste as a by-product that stored temporarily in RMI Building. The temporary solid radioactive waste storage room is equipped with in situ gamma area monitor (GAM) as a mean of external radiation protection for radioactive waste operator. This research is conducted to understand the characteristic of GAM Detector and evaluate the positioning of detector, the design of detector holder, and the threshold setting of radiation detection alarm in The temporary solid radioactive waste storage room. The research is conducted by measuring a standard radiation source Cs-137 with the activity of 1 µCi using GAM Detector with variation of position in 4 point, which are in the middle-front of detector, the front-side of the detector vessel, the middle of detector vessel, and back-end of detector vessel, along with variation of source to detector distance, which are 0 cm, 10 cm, 20 cm, 30 cm, 40 cm, and 50 cm. The measurement done in the middle-front of the detector give the highest dose of gamma radiation measured which are 5.63, 4.20, 0.58, 0.28 μSv/h. The result of gamma radiation dose measurement by the detector shows that the closer the source to middle-front part of the detector the higher the radiation measured. The holder design of the gamma radiation detector used in the current GAM device provides information on the gamma radiation dose rate of 74.26% of the actual dose rate. The detector holder currently positioned the middle-front side of the detector upwards therefore the design of detector holder needs to be modified in order for the detector to measure the dose more effectively and give the true number of dose measured

Development of risk mitigation guidance for sensor placement inside mechanically ventilated enclosures – Phase 1

Guidance on Sensor Placement was identified as the top research priority for hydrogen sensors at the 2018 HySafe Research Priority Workshop on hydrogen safety in the category Mitigation, Sensors, Hazard Prevention, and Risk Reduction. This paper discusses the initial steps (Phase 1) to develop such guidance for mechanically ventilated enclosures. This work was initiated as an international collaborative effort to respond to emerging market needs related to the design and deployment equipment for hydrogen infrastructure that is often installed in individual equipment cabinets or ventilated enclosures. The ultimate objective of this effort is to develop guidance for an optimal sensor placement such that, when integrated into a facility design and operation, will allow earlier detection at lower levels of incipient leaks, leading to significant hazard reduction. Reliable and consistent early warning of hydrogen leaks will allow for the risk mitigation by reducing or even eliminating the probability of escalation of small leaks into large and uncontrolled events. To address this issue, a study of a real-world mechanically ventilated enclosure containing GH2 equipment was conducted, where CFD modeling of the hydrogen dispersion (performed by AVT and UQTR, and independently by the JRC) was validated by the NREL Sensor laboratory using a Hydrogen Wide Area Monitor (HyWAM) consisting of a 10-point gas and temperature measurement analyzer. In the release test, helium was used as a hydrogen surrogate. Expansion of indoor releases to other larger facilities (including parking structures, vehicle maintenance facilities and potentially tunnels) and incorporation into QRA tools, such as HyRAM is planned for Phase 2. It is anticipated that results of this work will be used to inform national and international standards such as NFPA 2 Hydrogen Technologies Code, Canadian Hydrogen Installation Code (CHIC) and relevant ISO/TC 197 and CEN documents.

CubBonner – A novel passive neutron area monitor

Responses of a novel passive neutron area monitor, named CubBonner, were calculated. The responses were estimated for sixty monoenergetic neutrons, from 10−9 to 20 MeV, with the MCNP5 code. The CubBonner is a cubic polyethylene moderator and a gold foil as thermal neutron detector. The ambient dose equivalent response was calculated for three cubes (5″, 8″ and 10” side) with the gold foil at the cube's centers. The moderator cube having the best ambient dose equivalent response was used to estimate the neutron fluence and the 197Au(n,γ) responses per history. The ambient dose response per unit mass of gold was compared with the response of the Berthold LB 6411 active neutron area monitor, and the response for the (n,γ) reaction in the gold foil was compared with the evaporation photo neutrons produced in linear accelerators for radiotherapy.

Design of IoT-based Radiation Monitor Area for Nuclear and Radiological Emergency Preparedness System in Yogyakarta Nuclear Area

Yogyakarta Nuclear Area (YNA) is one of BATAN’s nuclear facility that has many nuclear facilities such as nuclear reactor, accelerator, irradiator, and waste management system. Due to nuclear accident is potentially occur, YNA has nuclear emergency response program and there is a preparedness system to support this program. Radiation monitoring system is one of utilization to acquire and record nuclear radiation exposure in the facility and its environment. An Internet of Thing based radiation area monitor was developed to improve the system not only to acquire the radiation exposure, but also to distribute the acquired data to the cloud server through the internet network. The distributed radiation data can be utilized to analyze nuclear emergence potential in the nuclear emergency response and preparedness system. The area monitor system was designed as sensor network that consists of Geiger Muller detector and high voltage power supply, signal conditioning system, and Arduino as counter and data processor. Data is acquired by a detector and it is transmitted to the server through wireless network using the node MCU communication module. The receiver station system was developed to acquire data from database server to display the radiation exposure in the environment and to identify radiological emergency status. Based on the chi square test stability method, the developed device has good stability which the probability of x2 was obtained 0.75. The validation testing also was done using Sr-90 and x-ray spectrometer as radiation source, and the developed device has been compared using a standard survey-meter. The mean square error of this validation testing was 0.37 and 0.24 respectively for Sr-90 at two difference source distance, while for x-ray spectrometer source the mean square error value was 2.48. For further development, the system will be combined with a smart meteorological system to build the integrated data acquisition system for Nuclear and Radiological Emergency Preparedness System in Yogyakarta Nuclear Area.

Energy Efficient and Secured Key Based Management in Area Monitor by WSN

Popular Technology for monitoring Environment, observation of battle field, caring of health, tracking of object, managing during various types of attack is by Wireless sensor networks. Whenever, area which is strictly monitoring by WSNs it has to handle all the critical situation. That is, when area affected by radiation creates node outage problems with in the critical areas monitoring by WSNs. For that solution should be needed when communication problem arises between the nodes. Reliable Transferring of information from nodes to the base station, less utilization of energy to improve the network life time and also along with these secure mechanisms are also needed to save the sensed information from hostile environment. In this paper clustering process with security mechanisms are used. Key based security approach used to validate data inside cluster and outside the cluster during data reaches to the base station. RSA encoding of the data to hide data from external persons. A security mechanism of our scheme shows that our protocol is effective in defending against radiation attack and intentionally node attacked hackers. Key based security approach implemented in Linux OS and simulate it using NS2 simulator to assess its time, energy, communication, and memory performance..

DOSE RATE IN A DEACTIVATED URANIUM MINE

The Ore Treatment Unit is a deactivated uranium mine and milling situated in Caldas, MG, BR. Although disabled, there are still areas considered controlled and supervised from the radiological point of view. In these areas, it is necessary to keep an occupational monitoring program to ensure the workers' safety and to prevent the dispersion of radioactive material. For area monitoring, the dose rate, in µSv∙h-1, was measured with Geiger Müller (GM) area monitors or personal electronic monitors type GM and thermoluminescence dosimetry (TLD), in mSv∙month-1, along the years 2013 to 2016. For area monitoring, 577 samples were recorded; for personal dosimeters monitoring, 2,656; and for TLD monitoring type, 5,657. The area monitoring showed a mean dose rate of 6.42 µSv∙h-1 associated to a standard deviation of 48 µSv∙h-1 with a maximum recorded value of 685 µSv∙h-1. 96 % of the samples were below the derived limit per hour for workers (10 µSv∙h-1). For the personal electronic monitoring, the average of the data sampled was 15.86 µSv∙h-1, associated to a standard deviation of 61.74 µSv∙h-1. 80 % of the samples were below the derived limit and the maximum recorded was 1,220 µSv∙h-1. Finally, the TLD showed a mean of 0.01 mSv∙h-1 (TLD detection limit is 0.2 mSv∙month-1 equivalent to 0.28 µSv∙h-1), associated to a standard deviation of 0.08 mSv∙h-1. 98% of the registered values were below 0.2 mSv∙month-1 and less than 2 % of the measurements had values above the limit of detection. The samples show areas with low risk of external exposure, as can be seen by the TLD evaluation. Specific areas with greater risk of contamination have already been identified, as well as operations at higher risks. In these cases, the use of the individual electronic dosimeter is justified for a more effective monitoring. Radioprotection identified all risks and was able to extend individual electronic monitoring to all risk operations, even with the use of the TLD.

Determination of the delivered dose of nanoparticles in the trachea-bronchial and alveolar regions of the lung

Particle dosimetry is a necessary step for the risk assessment especially with regard to toxicological effects of nanoparticles. In many cases, dosimetry is linked to exposure via the surface area of the nanoparticles. Due to interface dependent reactions of particles and the lung tissue poorly soluble particles (PSP) are particularly in the focus of interest. In recent years the measurement the geometric particle surface area concentration, GSA, in the air and determination of the lung deposited surface area (LDSA) based on lung deposition models gained interest. The effective particle concentration in the Lung (LDSA) cannot be measured directly. The maxima of many common airborne particle surface area distributions occur in the particle size range below 400 nm. High alveolar and trachea-bronchial deposition can be encountered in this size range.In general, it is very difficult to measure the particle surface area directly. To measure particle surface area concentration in aerosols, online instruments were developed (e.g. nanoparticle surface area monitor - NSAM). These instruments are based on unipolar particle charging. To determine the charge level of single particles as a function of the particle size, the current of monodisperse particles is divided by the aerosol number concentration. In order to relate the total measured current of the entire size-distributions, the instrument response function has to have the same slope as the size-dependent particle surface area. Calibration allows putting these curves on top of each other at one specific particle size (no error). Any deviations of the two curves at other sizes cause errors.The first instrument based on this principle was NSAM. Its response function for the measurement of spherical particles shows a diameter dependency of d1.16. It was found that if the geometric surface area (GSA) of a particle is multiplied with the alveolar deposited fraction (alveolar deposition curve of the ICRP-model) the resulting curve has almost the same slope as the NSAM response function. In principle, this concept also works for the comparison with the deposition in the trachea-bronchial region. To date, only the LDSA (alveolar or trachea-bronchial) of spherical particles in human lungs based on the ICRP-model can be determined by the available instruments.To extend the application range of NSAM to other lung deposition models, e.g. the Multiple-Path Particle Dosimetry model (MPPD) which also provides the possibility to investigate particle deposition in animal lungs, we defined model-specific calibration factors. This allowed the correct determination of the surface area concentration at the calibration point (100 nm) and also facilitated the measurement at other particle diameters.We assessed whether the slope differences between the modified instrument responses and the calculated LDSA in the MPPD deposition models led to tolerable errors. First, we described the calculation of these errors as a function of the particle size. Using these size-dependent error functions the total error (for the measurement of an entire particle size distribution) can be determined if e.g. a known particle number size distribution is transferred into a surface area distribution. To present a possibility to estimate the errors without performing the described error calculations, we performed a systematic study. We determined the total error concerning the measurement of LDSA for two ambient real aerosols (combustion of diesel and firewood). For these very different particle size distributions and models for different animals, the errors were below 20%, which seemed tolerable. Therefore NSAM and other instruments with a similar working principle can be used to determine the LDSA with regard to particle deposition in human and animal lungs.

The temperature process analysis and control on laser-assisted milling of nickel-based superalloy

Nickel-based alloys of Inconel 718 are well-known for their excellent properties of high-temperature hardness, mechanical strength, and wear resistance and are currently popular and applied in aerospace industry, such as the manufacturing of aero-engine turbine disks and compressor disk. Laser-assisted milling (LAML), which uses laser to locally heat the workpiece to improve the machinability of the material, has the advantage of increasing machining efficiency, reducing tool wear, and improving machining quality. In LAML, laser preheating temperature is the most important parameter influencing laser heating effect; however, constant laser energy commonly used in experiments cannot provide a stable preheating temperature. Therefore, temperature feedback control method is used in LAML and composite simulation models are established to simulate the temperature field and cutting process. The composite simulation models contain laser preheating temperature model, temperature feedback model, temperature difference prediction model, and cutting process model. Finite element method and neural network calculation method are used in building the model, which can be used to get the temperature difference, and then through monitoring to control the temperature of the laser heating. The simulation results show that the laser energy and the laser moving speed are the main factors affecting the laser preheating temperature, the difference between cutting area temperature and monitor temperature. A LAML experiment system is set up; cutting force and cutting chips are obtained through experiments. The simulation results express a good agreement with the experimental results, thus demonstrating the feasibility of the temperature-monitoring laser heating system and capabilities of LAML in fabricating difficult-to-machine parts for industrial implementation.

GSM module for wireless radiation monitoring system via SMS

A customised Global System for Mobile communication (GSM) module is designed for wireless radiation monitoring through Short Messaging Service (SMS). This module is able to receive serial data from radiation monitoring devices such as survey meter or area monitor and transmit the data as text SMS to a host server. It provides two-way communication for data transmission, status query, and configuration setup. The module hardware consists of GSM module, voltage level shifter, SIM circuit and Atmega328P microcontroller. Microcontroller provides control for sending, receiving and AT command processing to GSM module. The firmware is responsible to handle task related to communication between device and host server. It process all incoming SMS, extract, and store new configuration from Host, transmits alert/notification SMS when the radiation data reach/exceed threshold value, and transmits SMS data at every fixed interval according to configuration. Integration of this module with radiation survey/monitoring device will create mobile and wireless radiation monitoring system with prompt emergency alert at high-level radiation.