• Nuclear Engineering and Technology
• /
• v.52 no.5
• /
• pp.956-963
• /
• 2020

Dose optimization for Radioactive Occupational Personal (ROP) is an important subject in nuclear and radiation safety field. The geometric environment of a nuclear facility is complex and the work area is radioactive, so traditional navigation model and radioactive data field cannot form an effective environment model for dose assessment and dose optimization. The environment model directly affects dose assessment and indirectly affects dose optimization, this is an urgent problem needed to be solved. Therefore, this paper focuses on an environment model used for Dose Assessment and Dose Optimization (DA&DO). We designed a multi-layer radiation field coupling modeling method, and then explored the influence of the environment model to DA&DO by virtual simulation. Then, a simulation test is done, the multi-layer radiation field coupling model for nuclear facilities is demonstrated to be effective for dose assessment and dose optimization through the experiments and analysis.

• Nuclear Engineering and Technology
• /
• v.51 no.3
• /
• pp.825-836
• /
• 2019

One of the most challenging safety precautions for workers in dynamic, radioactive environments is avoiding radiation sources and sustaining low exposure. This paper presents a sampling-based algorithm, DL-RRT*, for minimum dose walk-path re-planning in radioactive environments, expedient for occupational workers in nuclear facilities to avoid unnecessary radiation exposure. The method combines the principle of random tree star ($RRT^*$) and $D^*$ Lite, and uses the expansion strength of grid search strategy from $D^*$ Lite to quickly find a high-quality initial path to accelerate convergence rate in $RRT^*$. The algorithm inherits probabilistic completeness and asymptotic optimality from $RRT^*$ to refine the existing paths continually by sampling the search-graph obtained from the grid search process. It can not only be applied to continuous cost spaces, but also make full use of the last planning information to avoid global re-planning, so as to improve the efficiency of path planning in frequently changing environments. The effectiveness and superiority of the proposed method was verified by simulating radiation field under varying obstacles and radioactive environments, and the results were compared with $RRT^*$ algorithm output.

• Journal of Radiation Protection and Research
• /
• v.44 no.3
• /
• pp.118-126
• /
• 2019

Background: Government conducts environmental radioactivity surveillance for verification purpose around nuclear facilities based on the Nuclear Safety Law and issues a surveillance report every year. This study aims to evaluate the short and the long-term fluctuation of radionuclides detected above MDC and their origins using concentration ratios between these radionuclides. Materials and Methods: Sample media for verification surveillance are air, rainwater, groundwater, soil, and milk for terrestrial samples, and seawater, marine sediment, fish, and seaweed for marine samples. Gamma-emitting radionuclides including $^{137}Cs$, $^{90}Sr$, Pu, $^3H$, and $^{14}C$ are evaluated in these samples. Results and Discussion: According to the result of the environmental radioactivity verification surveillance in the vicinity of nuclear power facilities in 2017, the anthropogenic radionuclides were not detected in most of the environmental samples except for the detection of a trace level of $^{137}Cs$, $^{90}Sr$, Pu, and $^{131}I$ in some samples. Radioactivity concentration ratios between the anthropogenic radionuclides ($^{137}Cs/^{90}Sr$, $^{137}Cs/^{239+240}Pu$, $^{90}Sr/^{239+240}Pu$) were similar to those reported in the environmental samples, which were affected by the global fallout of the past nuclear weapon test, and Pu atomic ratios ($^{240}Pu/^{239}Pu$) in the terrestrial sample and marine sample showed significant differences due to the different input pathway and the Pu source. Radioactive iodine ($^{131}I$) was detected at the range of < $5.6-190mBq{\cdot}kg-fresh^{-1}$ in the gulfweed and sea trumpet collected from the area of Kori and Wolsong intake and discharge. A high level of $^3H$ was observed in the air (Sangbong: $0.688{\pm}0.841Bq{\cdot}m^{-3}$) and the precipitation (Meteorology Post: $199{\pm}126Bq{\cdot}L^{-1}$) samples of the Wolsong nuclear power plant (NPP). $^3H$ concentration in the precipitation and pine needle samples showed typical variation pattern with the distance and the wind direction from the stack due to the gaseous release of $^3H$ in Wolsong NPP. Conclusion: Except for the detection of a trace level of $^{137}Cs$, $^{90}Sr$, Pu, and $^{131}I$ in some samples, anthropogenic radionuclides were below MDC in most of the environmental samples. Overall, no unusual radionuclides and abnormal concentration were detected in the 2017's surveillance result for verification. This research will be available in the assessment of environment around nuclear facilities in the event of radioactive material release.

• Fire Science and Engineering
• /
• v.25 no.5
• /
• pp.54-61
• /
• 2011

The study is to analysis of fire protection regulations for the nuclear power plants (NPP) in the United States, Japan, the UAE (United Arab Emirates), and Korea with the intention of exporting NPP to the UAE. Fire protection regulations for NPP for these countries permit the fire protection design and facilities in accordance with the evaluation of the potential fire hazards. However, in Korea, the NPP is a part of power generation facilities in Korea fire protection law, and the atomic energy act classifies them as the reactor and related nuclear facilities. The fire protection law and atomic energy act are different to the criteria for the fire protection of NPP. To maintain the leading position as a nuclear exporting country, the performance-based fire hazard analysis should be reasonably incorporated in the design of the fire protection system. It was suggested that the integrated requirements of the fire protection for NPP should be incorporated to the construction article for the fire protection facilities specified in paragraph 2 of Act II, being classified into the special objects to be protected against fire, which requires a performance-based design in order to incorporate the specific requirements for NPP.

• Nuclear Engineering and Technology
• /
• v.50 no.5
• /
• pp.738-744
• /
• 2018

Fuel failure due to debris is a major cause of failure in pressurized water reactors. Fuel vendors have developed various filtering devices to reduce debris-induced failure and have evaluated filtering performance with their own test facilities and methods. Because of the different test facilities and methods, it is difficult to compare filtering performances objectively. This study presents an improved filtering test and an efficiency calculation method to fairly compare fuel-filtering efficiency regardless of the vendor's filtering features. To enhance the reliability of our evaluation, we established requirements for the test method and had a facility constructed according to the requirements. This article describes the debris specimens, the amount of debris, and the replicates for the proposed test method. A calculation method of comprehensive debris-filtering efficiency using a weighted mean is proposed. The test method was verified by repeated tests, and the tests were carried out using the PLUS7 and 17ACE7 test fuels to calculate the comprehensive debris-filtering efficiencies. The evaluation results revealed that the filtering performance of PLUS7 is better than that of 17ACE7. The proposed method can be used on any kind of debris-filtering devices and is appropriate for use as a standard.

• Nuclear Engineering and Technology
• /
• v.53 no.7
• /
• pp.2277-2288
• /
• 2021

To regulate the safety protocols in nuclear facilities, radioactive aerosols have been extensively researched to understand their health impacts. However, most measured particle-size distributions remain at low resolutions, with the particle sizes ranging from nanometer to micrometer. This study combines the high-resolution detection of 500 size classes, ranging from 6 nm to 10 ㎛, for aerodynamic diameter distributions, with a regional lung deposition calculation. We applied the new approach to characterize particle-size distributions of aerosols generated during the plasma arc cutting of simulated non-radioactive steel alloy wastes. The high-resolution measured data were used to calculate the deposition ratios of the aerosols in different lung regions. The deposition ratios in the alveolar sacs contained the dominant particle sizes ranging from 0.01 to 0.1 ㎛. We determined the distribution of various metals using different vapor pressures of the alloying components and analyzed the uncertainties of lung deposition calculations using the low-resolution aerodynamic diameter data simultaneously. In high-resolution data, the changes in aerosols that can penetrate the blood system were better captured, correcting their potential risks by a maximum of 42%. The combined calculations can aid the enhancement of high-resolution measuring equipment to effectively manage radiation safety in nuclear facilities.

• Nuclear Engineering and Technology
• /
• v.53 no.7
• /
• pp.2229-2236
• /
• 2021

In nuclear-related facilities, such as nuclear power plants, research reactors, accelerators, and nuclear waste storage sites, radiation detection, and mapping are required to prevent radiation overexposure. Sensor network systems consisting of radiation sensor interfaces and wxireless communication units have become promising tools that can be used for data collection of radiation detection that can in turn be used to draw a radiation map. During data collection, malfunctions in some of the sensors can occasionally occur due to radiation effects, physical damage, network defects, sensor loss, or other reasons. This paper proposes a reproduction strategy for radiation maps using a U-net model to compensate for the loss of radiation detection data. To perform machine learning and verification, 1,561 simulations and 417 measured data of a sensor network were performed. The reproduction results show an accuracy of over 90%. The proposed strategy can offer an effective method that can be used to resolve the data loss problem for conventional sensor network systems and will specifically contribute to making initial responses with preserved data and without the high cost of radiation leak accidents at nuclear facilities.

• Proceedings of the Korea Society for Energy Engineering kosee Conference
• /
• 1999.11a
• /
• pp.119-124
• /
• 1999

A selection methodology of liquid levelmeters, especially, level sensors in non-nuclear category, to be installed in nuclear energy facilities is developed using linguistic fuzzy approaches such as fully-linguistic and semi-linguistic methods. Depending on defuzzification techniques, the linguistic fuzzy methodology leads to either linguistic (exactly, fully-linguistic) or cardinal (i.e., semi-linguistic) evaluation. For the linguistic method, for each alternative, fuzzy preference index is converted to linguistic utility value by means of a similarity measure determining the degree of similarity between fuzzy index and linguistic ratings. For the cardinal method, the index is translated to cardinal overall utility value. According to these values, alternatives of interest are linguistically or numerically evaluated and a suitable alternative can be selected. Under given selection criteria, the suitable selections out of some liquid levelmeters for nuclear facilities are dealt with using the linguistic fuzzy methodology proposed. Then, linguistic fuzzy evaluation results are compared with qualitative result available in the literature. It is found that as to a suitable option the linguistic fuzzy selection is in agreement with the qualitative selection. Additionally, the comparative study shows that the fully-linguistic method using adequate scale system facilitates linguistic interpretation regarding evaluation results.

• Nuclear Engineering and Technology
• /
• v.54 no.10
• /
• pp.3788-3794
• /
• 2022

The possibility of denuclearization of the Korean Peninsula has been continuously debated, and the initiative participation of the Republic of Korea has necessitated preemptive measures against neighboring countries. In this study, we present a proposal for formulating a site survey plan when the amount of site information provided is insufficient and the accuracy of the information is not guaranteed. Considering a case wherein "a soil sample analysis is used to determine the presence or absence of nuclear activity" in a radiochemical laboratory, which is a typical key facility for denuclearization, the optimal soil sample collection plan is designed based on international guidelines and public information. In the event of denuclearization, a scenario that is not based on the expertise of the sample collector is set, and the data quality objective (DQO) process is applied to ensure reality. Consequently, the primary sample collection points can be derived in consideration of accessibility, and the sample collection scale can be adjusted according to the cost. By applying the DQO process to ensure sample representativeness and reality, reliable and resource-efficient soil sample collection can be achieved in radiochemical laboratories and other denuclearization facilities.

• Proceedings of the Korean Radioactive Waste Society Conference
• /
• 2018.11a
• /
• pp.275-276
• /
• 2018