As drone technology and industry develop around the world, the use of drones are increasing in number and expanding to different fields. On the other hand, illegal flight and terrorist incidents using drones are also increasing day by day. In Korea, it is reflected in the “Design Basis Threat (DBT)”, which is the standard for designing and evaluating the physical protection system of nuclear power plants in accordance with the “Act on Physical Protection and Radiological Emergency”, that nuclear power plants continue to establish physical protection against drone threats. A total of 141 drone attacks or incidents have occurred around the world since 2015. Cases related to the Russian-Ukraine war, in which so many cases occurred, were excluded. There were 112 cases (79%) of terrorism or suspected terrorism using a single drone. There were 4 cases of terrorism using more than 5 drones, and a total of 20 drones were used to attack an oil facility in Yemen (2019). By region, a total of 111 incidents occurred in Middle East & North Africa. By country, there were 49 cases in Iraq, 35 cases in Saudi Arabia, and 8 cases in Syria. Among major countries, three cases occured in Korea, five in the United States, two in England, Canada, and Italy, and one in Japan and France. Since 2021, there have been 15 drone attacks. Multiple drones were used in attacks targeting military or large-scale Important National Facilities such as the Saudi oil refinery, Indian Kashmir air base, and reconnaissance of Iranian Natanz nuclear and surrounding military facilities. Also in 89% of the cases, the drones were loaded with explosives in order to cause large-scale damage. Accordingly, nuclear power plants, which are important national facilities, need to establish a system that can detect and respond to multiple drones. Furthermore, additional protective measures are needed for areas that are expected to be severely damaged which can be established by evaluating the impact of explosives on major points in the plant. In additionthere is a high possibility of terrorism by organizations aiming for national turmoil rather than individual terrorists. So it is important to identify signs of terrorism in advance and prepare through cooperation with related agencies.

UAVs (Unmanned Aerial Vehicle) are a rising threat to national facilities due to their cheap price and accessibility. Incidents such as the terrorism attack in Saudi Arabia’s oil facilities and the paralysis of the airport system in England’s Gatwick airport shows the need for integrating CUAS (Counter- Unmanned Aerial Systems) in important national facilities. Recently efforts have been made to evaluate the technical performance of the CUAS. Especially SNL (Sandia National Laboratory) modified the methodology used for PPS (Physical Protection Systems) to develop a performance metrics for CUAS. The performance metrics can be used to effectively analyze the facilities capability of countering drone attacks in a probabilistic way. In this study, we managed to derive the safety boundary of a reference nuclear power plant model based on its current CUAS and protection capabilities with a simplified methodology. Based on the outermost boundary of the model, the time table of the UAS consist of 4 variables which are the assessment time, transmission time, neutralization time and the maximum vehicle velocity. Dividing the maximum velocity to the net time derived, we estimated the minimum sensing point of the CUAS which is the minimum safety boundary of the facility to safely manage the UAV attack. Two practice cases were evaluated with the methodology which is based on the UAV groups classified by the United States DOD (Department Of Defense) that matches the classification of the UAV in Korea. Each variable was assumed to fit the process of a realistic nuclear power plant. Using the variables, we calculated the minimum safety boundary of the facility. With the methodology introduced in this study, regulators and stakeholders can easily evaluate the capability of the facilities CUAS for a design basis UAV attack. Also it can be used as a simple tool to analyze the facilities vulnerability for specific UAV specifications and a guideline to check the protective procedures of the facility.

The purpose of the present research is to verify the design characteristics of the SMART facility for the application of the IAEA’s safeguards-by-design (SBD) concept to small modular reactor (SMR) and to establish a foundation for SBD to be faithfully implemented as early as possible from the design stage. International Atomic Energy Agency (IAEA) is planning to facilitate the verification activities of inspectors by developing a safeguards approach to the reactor as early as possible and preparing a safeguards technical report (STR) before commercial operation of SMR begins. To this end, the IAEA is developing various approaches to the application of SBD to SMR with countries such as Republic of Korea, Russian Federation, China, the United States, and Canada through the Member State Support Program (MSSP). In order to review the unique design information of SMART facilities, the only deployable SMR in Korea, and to establish safeguards from the early design stages of SMART, it is necessary to carry out the task through cooperation with the Korea Atomic Energy Research Institute (KAERI) and Korea Institute of Nuclear Nonproliferation and Control (KINAC). IAEA agreed with the KINAC and KAERI to the direction of the project and to prepare both the Design Information Questionnaire (DIQ) and the Safeguards Technical Report (STR) for SMART facilities sequentially. The DIQ is a collection of questions to understand the characteristics of the reactor facilities that must be considered in applying safeguards. The STR is a document referenced by IAEA inspectors when verifying safeguards. Those draft versions were prepared and submitted to the IAEA. After review opinions were received, additional revision was conducted. In 2022, the IAEA holds the consultancy meeting on SBD for SMART. The purpose of the meeting is to review the draft DIQ and STR prepared by designers and discuss the future work plan of the task with designer and the task point of contact in order to safeguards can be considered at the early stage of the design. The results will be beneficial to the efficient safeguards verification activities of IAEA inspectors in the future.

The acoustic emission (AE) is proposed as a feasible method for the real-time monitoring of the structural damage evolution in concrete materials that are typically used in the storage of nuclear wastes. However, the characteristics of AE signals emitted from concrete structures subjected to various environmental conditions are poorly identified. Therefore, this study examines the AE characteristics of the concrete structures during uniaxial compression, where the storage temperature and immersion conditions of the concrete specimens varied from 15℃ to 75℃ and from completely dry to water-immersion, respectively. Compared with the dry specimens, the water-immersed specimens exhibited significantly reduced uniaxial compressive strengths by approximately 26%, total AE energy by approximately 90%, and max RA value by approximately 70%. As the treatment temperature increased, the strength and AE parameters, such as AE count, AE energy, and RA value, of the dry specimens increased; however, the temperature effect was only minimal for the immersed specimens. This study suggests that the AE technique can capture the mechanical damage evolution of concrete materials, but their AE characteristics can vary with respect to the storage conditions.

With the rapid development of China's economy and the gradual maturity of high-speed rail technology, China's high-speed rail tourism is also developing. With the slow recovery of the COVID-19 epidemic and the slow recovery of the tourism industry, the high-speed rail tourism industry has also continued to grow, and the combination of high-speed rail and tourism has been studied in depth to meet market demand. In this diverse society, the development of high-speed rail tourism will be promoted, employment will be increased, and consumption will be stimulated, thereby promoting a better development of the society. This paper explains the current situation of high-speed rail tourism and the specific impact of high-speed rail on tourism, as well as the current status of high-speed rail tourism, in order to better promote the development of high-speed rail tourism, thereby promoting national economic and social development.

Human intestinal flora is very diverse; with lactic acid bacteria (LAB) existing as part of the most vital gut microbes that improve host health. The application of LAB as a whole organism or its metabolite in the case of probiotic and bacteriocin respectively is extensive. Thus, the need to always bio-prospect for newer strains of LAB is essential. This study focused on isolating LAB from kunun-zaki and kindirmo, a fermented non-alcoholic beverage of non-dairy and dairy sources respectively, explored their physiological and biochemical properties, antibiotics sensitivity pattern and identified based on their 16S rRNA sequencing. A total of eighty isolates were selected sixty-six from kunun-zaki and fourteen from kindirmo in which 93.7% were bacilli and 6.3% were spherical in shape having 68.75% and 30% homofermentative and heterofermentative pathway respectively. All isolates have the ability to utilize glucose to produce lactic acid while their tolerance to pH 3 and salt concentration at 2%, 4% and 6.5% varied widely. Thirty-four isolates based on their physiological and biochemical properties were selected for molecular identification to ascertain their genera and species. Limosilactobacillus fermentum (68%); Lactiplantibacillus plantarum (6%) and Weissella confusa (3%) were confirmed species isolated. Thus, it was concluded that traditional fermented foods such as kunun-zaki and kindirmo are a good source to bio-prospect for LAB for product development, starter culture and probiotic study.

Electroanalytical study for the rotating cylinder electrode in molten LiCl-KCl eutectic salt (58– 42mol%) containing MgCl2 (0.1wt%) at 600°C is conducted. The researches of rotating cylinder electrode have been widely conducted for the century. The advantage of the electrode is that it can mitigate the unintended natural convection by providing a controlled diffusion boundary layer thickness. However, the experimental data for the high temperature molten salts is barely existed. The study adopts the electrochemical techniques such as cyclic voltammetry for the static cell and linear sweep voltammetry for the dynamic cell to calculate the diffusion coefficient. The peak current density and limiting current density are measured according to the scan rate. In order to evaluate the mass transfer under hydrodynamic flow condition, the revolution speeds of cylindrical electrode are varied from 10 rpm to 500 rpm which are corresponded to the Reynolds number of 4 and 185 respectively. The flow regime covers from the laminar to semi-turbulent regime (transient) as the critical Reynolds number Recrit is 200. The limiting current density shows a linear trend with the revolution speed and agrees well with the existing mass transfer correlations. For the extended flow regime, a new mass transfer correlation is suggested as the relation of non-dimensional numbers (Sh = aRebScc) based on the dimensionless analysis.

Radioactive wastes that are generated as a result of operating NPPs, contain 63Ni and 59Ni that should be analyzed in accordance with the notice of Nuclear Safety and Security Commission (NSSC) for the acceptance of Korea Radioactive Waste Agency (KORAD). Analyzing 63Ni and 59Ni has few challenges to determine activities of each nuclide in radioactive waste sample that contains both nuclides. As is well known, 63Ni can be analyzed by liquid scintillation counter (LSC) detecting its emitted beta rays, however, beta rays emitted from 59Ni are overlapped on the spectrum. Therefore, to discriminate those two nuclides, spectrum channel should be divided according to its dedicating part of the spectrum. For instance, 59Ni contribute to spectrum channel 30–250, on the other hand, 63Ni contributes to spectrum channel 30–450. In other word, 63Ni solely can be analyzed on the channel from 260 to 450. To analyze both 63Ni and 59Ni using this channel division method, detection efficiency must be measured in advance; efficiency of 63Ni and 59Ni at ch. 30–250, and efficiency of 63Ni at ch. 260–450, then the activity can be calculated using the corresponding efficiency. In this study, for verifying the feasibility of channel division method, 5 simulated samples were prepared with different ratio of 63Ni/59Ni. The ratio varies as 1, 2, 10, 20 and 100 spiking standard source of 63Ni and 59Ni. Each sample was mixed with scintillation cocktail and detected for 90 minutes by LSC (300SL, Hidex) after the stabilization of solutions. As a result, calculated 63Ni activities for all sample were averaged as 97% of spiked activity. However, calculated 59Ni activity were 101%, 103%, 128%, 140%, 260%, respectively. The result indicates that 59Ni cannot be discriminated by channel division method when it exists in the sample with high 63Ni over 10 times then 59Ni such as radioactive waste sample. However, the results also show that the channel division method for analyzing 63Ni activity was successful verifying it can determine the activity of 63Ni regardless of the affect of 59Ni on the spectrum.

Measurement of the physical properties of high-temperature molten salts is important for the efficient design and operation of molten salt reactors (MSR) in which the reactor coolant and nuclear fuel are in a homogeneous liquid state. Although some crucial physical properties such as viscosity, thermal conductivity, density, etc., have been drawing much attention, relative data, especially for molten chloride salts, are scarce. Thus, it is urgent to prepare the viscosity data as one of the key transport properties in thermal hydraulics analysis. However, it is not an easy task to measure the molten salt viscosity with high accuracy due to end effect, a small gap between the chamber and spindle, thermal expansion of the chamber and spindle at high temperatures in a rotational viscometer. Additionally, molten salt temperatures inside furnace are not uniform due to the large temperature gradient inside the chamber, and therefore the assumption of laminar condition can be violated. In this study, geometric factors, which can be a major interference in the torque measurement, were considered for the accurate determination of the viscosity. We established a high-temperature molten salt viscosity measurement system with Brookfield rotational viscometer. KNO3 molten salt was used as a model substance at a temperature range of 650–773 K. In-house designed spindles and chambers were made of corrosion-resistant alumina. Thermal expansion has a significant influence on the size and shape of the chamber and spindle. The effect of thermal expansion on the conventional correction method was examined with temperature variation and distribution. Gap size variation was also investigated in order to improve the accuracy.

A molten salt reactor (MSR) that uses molten salt mixtures as nuclear liquid fuel has recently received much attention due to its inherent safety. Various fluoride and chloride salt mixtures are considered as fluid fuel for MSRs. Among those, NaCl-MgCl2-UCl3 system is the one of the most promising candidates for molten salt fast reactor. The comprehensive information on thermo-physical properties such as density, viscosity, heat capacity and thermal conductivity are fundamental to MSR design development, but experimental data for NaCl-MgCl2-UCl3 system are unknown to the best of our knowledge. In this study, we estimated the thermophysical properties of NaCl-MgCl2-UCl3 system. The properties were calculated by mole fraction additive method using reliable experimental data from pure salt system. Other methods, such as rule of additivity of molar volume for density, modified Dulong-Petit method for heat capacity, and Rao-Turnbull prediction and Ignatieve-Khokolve correlation for thermal conductivity, have also been applied. Estimated values for the properties were compared with each other as well as available binary experimental data.

Spent nuclear fuel is a very complex material because various elements such as fission products, transuranium elements and activation products are produced from initial fresh UO2 fuel after irradiation. These elements exist in UO2 with various forms and can change the structure and of physicochemical properties of UO2. These changes could provide the surface activation site that could enhance chemical reactions and corrosion processes, and would significantly affect the storage environment for long-term disposal of spent nuclear fuel. Therefore, it can be important to understand the characteristics of spent nuclear fuel to design reliable and safe geological repositories. However, it is too hard to study the characteristics of spent nuclear fuel, because it is a very complex material by itself and not easy to handle due to its radioactivity, and it is also difficult to independently understand the effects of each element. Therefore, a simulated spent nuclear fuel containing an element that forms a solid solution and epsilon particle was manufactured to understand the change in characteristics of each element. Most of the elements that form solid solutions are lanthanides or actinides and can change the structure of the UO2 lattice itself. The epsilon particles exist as metals at the grain boundaries of UO2. In this study, structural changes were measured using XRD, SEM, and Raman spectroscopy, and physical and chemical properties were also identified by measuring electrical conductivity and electrochemical properties. The results were summarized, and the effects of solid solution elements and epsilon particles on the structure and properties of UO2 matrix were compared and discussed.

Polyoxometalates (POMs) are nanoclusters composed of transition metals with high oxidation states. Owing to their redox properties and structural diversity, POMs have been applied to broad fields, such as catalysis, materials, and medicine. Among various fields of application, POMs play an important role in radiochemistry. POMs can form complexes with tri- and tetravalent lanthanides and actinides (radioactive elements), which may be good sequestrators or agents for separating nuclear wastes. Among the most prominent POM structures, Anderson-type POMs with a general formula of [Hy(XO6)M6O18]n− (y=0–6, n=2–8, M=addenda atom, X=heteroatom) represent one of the basic topological structures of the POM family. An important feature of Anderson type POMs is incorporating a large number of various heteroatoms with different size and oxidation states, which can lead to tune chemical properties. Interestingly, no example of Anderson type POMs with early transition metal ions in the heteroatom site has been reported to date. Herein, we discovered that the Anderson POM Na2K6Ti0.92W6.08O24·12H2O, which consists of pure inorganic framework built from a central Ti core supported by six WO6 inorganic scaffold, and the crystal structure was confirmed and refined using single-crystal X-ray diffraction (SC-XRD). In addition, structural characterizations, including, Fourier-transform infrared spectroscopy (FT-IR), X-ray photoelectron spectroscopy (XPS), and Inductively coupled plasma-optical emission spectroscopy (ICP-OES) were performed.

According to Article 4 and 5 of the Nuclear Safety and Security Commission (NSSC) Notice No. 2020-6, radioactive waste packages should be classified by radioactive levels, and finally permanently shipped to underground or surface disposal facilities. The level of the radioactive waste package is determined based on the concentrations of the radionuclides suggested in Article 8 of NSSC Notice No. 2021-26. Since most of the radionuclides in radioactive wastes are beta nuclides, chemical separation and quantification of the target nuclides are essential. Conventional methods to classify chemically non-volatile radionuclides such as Tc-99, Sr-90, Nb- 94, Fe-55 take a lot of time (about 5 days) and have low efficiency. An automated non-volatile nuclide analysis system based on the continuous chemical separation method of radionuclides has been developed to compensate for this disadvantages of the conventional method in this study. The features of the automated non-volatile nuclide separation system are as follows. First, the amount of secondary waste generated during the chemical separation process is very small. That is, by adopting an open-bed resin column method instead of a closed-bed resin column method, additional fittings and connector are unnecessary during the chemical separation. In addition, because the peristaltic pump is supplied for the sample and solution respectively, it is great effective to prevent cross-contamination between radioactive samples and the acid stock solution for analysis. Second, the factors that may affect results, such as solution amount, operating time and flow rate, are almost constant. By mechanically controlling the flow rate precisely, the operating time and additional factors required during the separation process can be adjusted and predicted in advance, and the uncertainty of the chemical separation process can be significantly reduced. Finally, it is highly usable not only in the continuous separation process but also in the individual separation process. It can be applied to the individual separation process because the user can set the individual sequence using the program. As a result of the performance evaluation of the automation system, recovery rates of about 80–90% and reproducibility within 5% were secured for all of the radionuclides. Furthermore, it was confirmed that the actual work time was reduced by more than 50% compared to the previous manual method. (It was confirmed that the operation time required during the separation process was reduced from 6 days to 3 days.) Based on these results, the automation system is expected to improve the safety of workers in radiation exposure, reduce human error, and improve data reliability.

Appropriateness of the minimum detectable activity in the analysis of gamma radionuclides is very important. This is reason determine the time factor among the conditions of the analysis when it is rationally determined has the advantage that radioactivity analysis can be performed accurately and quickly. In this study, 100 mL of an unknown sample was diluted in Marinelli Beaker 1L to obtain, review data on gamma radiation analysis results and minimum detectable activity for each measurement time. The measurement was used High Purity Germanium detector, target nuclides are Co-57, Co-58, Y-88 and Cs-137. Since the radioactivity analysis sample will be expected to be the waste subject to selfdisposal or less during the radioactive waste classification, the minimum detectable activity standard was set based on the detection of less than the permissible activity for self-disposal for each nuclide. The measurement methods were measured by classifying it into seven categories: 1000 seconds, 3600 seconds, 10000 seconds, 30000 seconds, 80000 seconds, 100000 seconds, and 150000 seconds. The radioactivity from this measurement are Co-57 2.89 Bq·g−1, Co-58 0.19 Bq·g−1, Y-88 0.20 Bq·g−1, Cs-137 0.15 Bq·g−1, the measurement results under all conditions were similar. On the other hand, the minimum detectable activity showed values above the allowable activity for self-disposal in not but Co-58 at 1000 and 3600 seconds. Only after taking the measurement time of 10000 seconds, the result was derived Co-57 0.0095 Bq·g−1, Co-58 0.0068 Bq·g−1, Y-88 0.0052 Bq·g−1, Cs-137 0.0062 Bq·g−1, which was confirmed to less than the allowable activity for self-disposal by nuclide. Reasonably determining the measurement time in gamma radionuclide analysis is a very important issue in terms of economy of time and accuracy of measurement. Although this study cannot be said to be able to determine a reasonable measurement time for all gamma radionuclide analysis, it is hoped that research on various samples will be made to contribute to the efficient measurement of gamma radioactivity.

Density of chloride molten salts is an essential physical property in the reactor core design and thermal-hydraulic design simulation, especially in molten salt reactor (MSR) design currently under development in Korea. NaCl-MgCl2-UCl3 pseudo-ternary system is one of the various candidate chloride-based salt mixtures because it has relatively-low melting point, very low vapor pressure, high thermal conductivity, etc. However, to the best of our knowledge, the density data of NaCl-MgCl2- UCl3 have not yet been measured or published worldwide, and therefore the ballpark figures of the density should be given for the preliminary reactor design. In our present study, the density estimation of NaCl-MgCl2-UCl3 based on the pseudo-binary data, i.e., NaCl-MgCl2, MgCl2-UCl3, and NaCl- UCl3, reported in the literature previously were performed using the Redlich-Kister model. Binary interaction parameter for MgCl2-UCl3 was higher than that for NaCl-MgCl2 and lower than that for NaCl-UCl3. As an example, calculated density of 0.62 NaCl: 0.18 MgCl2: 0.20 UCl3 at 873 K was 2.578 g·cm−3. In our further study, the methodology using Redlich-Kister model will be applied to more complex multicomponent systems and to other physical properties such as viscosity, thermal conductivity, surface tension, etc.

Radioactivity of radiostrontiums, Sr-89 and Sr-90, which are both pure beta-emitters, are generally measured via Cherenkov counting. However, the determination of Cherenkov counting efficiencies of radiostrontiums requires a complicated procedure due to the presence of Y-90 (also a pure betaemitter) which is the daughter nuclide of Sr-90. In this study, we have developed a machine learning approach using a linear regression model which allows an easier and simultaneous determination of the Cherenkov counting efficiencies of the radiostrontiums. The linear regression model was employed because total net Cherenkov count (Ct) from the three beta-emitters at time t after the separation of Y- 90, can be expressed as a linear combination of their respective time-varying radioactivities with their respective coefficients (parameters) being their counting efficiencies: Ct = εSr-90[ASr-90·exp(–λSr-90·t)] + εSr-89[ASr-89·exp(–λSr-89·t)] + εY-90[ASr-90·exp(1–λSr-90·t)], where ε is a counting efficiency, A is an initial activity, λ is a decay constant and t is time after the separation of Y-90, Thus, if we train the model with multiple Cherenkov counts measured from the three beta emitters, then we can obtain their estimates for counting efficiencies (so-called parameters) straightforward. For this, the model has been trained by two methods: Ordinary Least Squares (OLS) and Bayesian linear regression (BLR), for which two software packages, PyMC3 and Stan were employed to compare their performances. The results showed that the accuracy of the OLS was worse than that of the BLR. Particularly, the counting efficiency of Sr-90 was estimated to be smaller than 0, which is an unrealistic value. On the other hand, the estimates of the BLR gave realistic values which are close to the true values. Additionally, the BLR was able to provide a distribution for each counting efficiency (so-called “posterior”) from which various types of inference can be made including median and credible interval in the Bayesian statistics which is analogous to, but different from confidence interval in the Frequentist statistics. In the results of the BLR, the Stan package gave more accurate estimates than the PyMC3 package. Therefore, it is expected that counting efficiencies of the radiostrontiums including radioyttrium can be determined at the same time, more easily and accurately, by using the BLR with the Stan package and that the activities of radiostrontium also can be determined more easily by using the BLR if we know their counting efficiencies in advance. It is worth noting that the usage of the linear regression model in this study was different from the usual one where the trained model is used to predict a response value (count) from a set of unseen regressor values (activities).

Once a radioactive material is released from the nuclear power plant (NPP) by accident, it is necessary to understand the behavior of radioactive plume to protect residents adequately. For this, it is essential to measure the radiation dose rate around NPPs at important locations. Our previous study developed a movable radiation detector that can be installed quickly in an accident to measure gamma dose rate in areas where environmental radiation monitoring system is not installed. The data measured by the detector are transmitted to the server in real-time through LoRA wireless communications. There are two methods to use LoRA communications; one is self-network, and the other is the network provided by the mobile carrier. A signal receiver, called a gateway, should be equipped near the installation location of radiation detectors to use a self-network without using the mobile carrier’s system. In other words, the movable radiation detectors we made can function if there should be any gateway near them. The distance capable of communication between gateway and detector is about 8 km in an open area without significant obstacles. Korea has many significant obstacles, such as mountains around most NPPs. Thus, the gateways could be installed in the proper position before the accident to operate the movable radiation detectors without problems. If the gateway is located at a high position like a mountain top, it could cover a wide area. In this study, the elevation database in the area around the NPPs was collected and analyzed to determine where gateways should be installed. The analysis range is limited in the urgent protective action planning zone. The optimization was also performed to minimize the number of gateways.

A radioactive complex disaster refers to a situation in which large-scale natural and social disasters such as earthquakes, fires, and chemical incidents occur in addition to nuclear power plants accident simultaneously. For the safe evacuation of residents, protective equipment must be prepared appropriately and sufficiently. We presented effectiveness, sufficiency, and versatility as selection criteria for choosing the essential protective equipment. The results of the survey of residents and local public officials suggested that masks, radioactivity meters, protective buses, and air purifiers have high priorities. Finally, through consultation with central government officials, masks, measuring instruments, and adsorption filters in protective buses were presented as prototypes. These items can effectively protect the people even in the event of a radioactive complex disaster.

For the final disposal of radioactive waste generated during the operation of nuclear power plants, concentrations of 14 radionuclides including gross alpha have to be determined to meet nuclear regulatory requirements. In order to determine the gross alpha radioactivity in radioactive waste, the sample must be preprocessed into a solution which is usually a strong acid. When this solution is used to prepare the gross alpha measurement sample, it produces a lot of salt, which makes an accurate measurement difficult. Also it causes corrosion of a planchet, which causes problems in the disposal of waste in the future. For these reasons, an acid treatment of the solution was added to the existing preprocess procedure, which is also expected to improve the measurement error. Although the gross alpha measurement is known to be easy to perform and able to give rapid results, it cannot be used for quantitative analysis. This is because the energies emitted by the individual alpha nuclides are assumed to be produced from a single alpha emitted by the individual alpha nuclides are assumed to be produced from a single alpha emitter used as the standard calibration source. Also, due to self-absorption of alpha particles a counting rate depends on the thickness (or weight) of the residues on the planchet. In this study, we compared gross alpha radioactivity with and without an acid treatment to prepare gross alpha measurement samples. The weights of the treated samples increased by at most 5% after about 12 hours of evaporation to dryness, and then saturated or slightly decreased, while the weights of the untreated samples increased up to 20% over time. In addition, the radioactivities of the untreated samples were about two times those of the treated samples. This is considered to be due to differences both in the geometric shapes of the samples and the weights of their residues which resulted from whether acid treatment was applied or not. The results of this study showed that an acid treatment was beneficial in reducing both production corrosion and salts which could result in more reliable and constant measurements of gross alpha activity. The results showed that acid treatment was beneficial in reducing corrosion and measurement errors.

In Malaysia, there are several industries processing mineral ores generate residues containing naturally occurring radioactive material (NORM) with activity concentrations above the control limits established by the Malaysian Atomic Energy Licensing Board (AELB). These industries use mineral ores or concentrated ores as their feed materials to produce or extract valuable sand minerals or rare earth compounds for use in another industries. The control limits for activity concentrations of Uranium-238 (U-238) and Thorium-232 (Th-232) and their decay series is 1.0 Becquerel per gram (Bq·g−1) while activity concentration of Potassium 40 (K-40) is 10.0 Bq·g−1. The management of residue containing NORM radioactivity above the control limits must be done in accordance with current rules and regulations including proper handling, storage, transportation and/or disposal. Where possible, appropriate mixture process with other non-radiological material would reduce the activity concentrations to below the control limits. Depending on specific characteristics of residue, appropriate approach to reuse or recycle should be encouraged as part of special waste management. For this case, an exemption to release it from radiological controls can be applied but require scrutiny review and approval process by AELB. In addition, the health and safety aspects and environmental issues should be assessed which to be done in accordance with the relevant rules and regulations. As a last resort, a disposal of residue containing NORM radioactivity shall be done at the landfill disposal facility approved by AELB and other relevant Authorities.