Pub Date : 2021-03-31DOI: 10.7733/JNFCWT.2021.19.1.133
Gi-Woong Kang, Rin-Ah Kim, Dho Hoseog, K. T. man, C. Cho
This is an Open-Access article distributed under the terms of the Creative Commons Attribution Non-Commercial License (http://creativecommons.org/licenses/ by-nc/3.0) which permits unrestricted non-commercial use, distribution, and reproduction in any medium, provided the original work is properly cited This study evaluates the radioactivity of concrete waste that occurs due to large amounts of decommissioned nuclear wastes and then determines the surface dose rate when the waste is packaged in a disposal container. The radiation assessment was conducted under the presumption that impurities included in the bio-shielded concrete contain the highest amount of radioactivity among all the concrete wastes. Neutron flux was applied using the simplified model approach in a sample containing the most Co and Eu impurities, and a maximum of 9.8×104 Bq·g−1 60Co and 2.63×105 Bq·g−1 152Eu was determined. Subsequently, the surface dose rate of the container was measured assuming that the bio-shield concrete waste would be packaged in a newly developed disposal container. Results showed that most of the concrete wastes with a depth of 20 cm or higher from the concrete surface was found to have less than 1.8 mSv·hr−1 in the surface dose of the new-type disposal container. Hence, when bio-shielded concrete wastes, having the highest radioactivity, is disposed in the new disposal container, it satisfies the limit of the surface dose rate (i.e., 2 mSv·hr−1) as per global standards.
{"title":"A Study on the Evaluation of Surface Dose Rate of New Disposal Containers Though the Activation Evaluation of Bio-Shield Concrete Waste From Kori Unit 1","authors":"Gi-Woong Kang, Rin-Ah Kim, Dho Hoseog, K. T. man, C. Cho","doi":"10.7733/JNFCWT.2021.19.1.133","DOIUrl":"https://doi.org/10.7733/JNFCWT.2021.19.1.133","url":null,"abstract":"This is an Open-Access article distributed under the terms of the Creative Commons Attribution Non-Commercial License (http://creativecommons.org/licenses/ by-nc/3.0) which permits unrestricted non-commercial use, distribution, and reproduction in any medium, provided the original work is properly cited This study evaluates the radioactivity of concrete waste that occurs due to large amounts of decommissioned nuclear wastes and then determines the surface dose rate when the waste is packaged in a disposal container. The radiation assessment was conducted under the presumption that impurities included in the bio-shielded concrete contain the highest amount of radioactivity among all the concrete wastes. Neutron flux was applied using the simplified model approach in a sample containing the most Co and Eu impurities, and a maximum of 9.8×104 Bq·g−1 60Co and 2.63×105 Bq·g−1 152Eu was determined. Subsequently, the surface dose rate of the container was measured assuming that the bio-shield concrete waste would be packaged in a newly developed disposal container. Results showed that most of the concrete wastes with a depth of 20 cm or higher from the concrete surface was found to have less than 1.8 mSv·hr−1 in the surface dose of the new-type disposal container. Hence, when bio-shielded concrete wastes, having the highest radioactivity, is disposed in the new disposal container, it satisfies the limit of the surface dose rate (i.e., 2 mSv·hr−1) as per global standards.","PeriodicalId":17456,"journal":{"name":"Journal of the Nuclear Fuel Cycle and Waste Technology","volume":"9 1","pages":"111-118"},"PeriodicalIF":0.0,"publicationDate":"2021-03-31","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"74657522","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 2021-03-31DOI: 10.7733/JNFCWT.2021.19.1.113
Park Sang June, Jihyang Byon, Lee, Jun-Yeop, Seokyoung Ahn
{"title":"A Study About Radionuclides Migration Behavior in Terms of Solubility at Gyeongju Low- and Intermediate-Level Radioactive Waste (LILW) Repository","authors":"Park Sang June, Jihyang Byon, Lee, Jun-Yeop, Seokyoung Ahn","doi":"10.7733/JNFCWT.2021.19.1.113","DOIUrl":"https://doi.org/10.7733/JNFCWT.2021.19.1.113","url":null,"abstract":"","PeriodicalId":17456,"journal":{"name":"Journal of the Nuclear Fuel Cycle and Waste Technology","volume":"24 1","pages":"113-121"},"PeriodicalIF":0.0,"publicationDate":"2021-03-31","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"83875591","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 2021-03-31DOI: 10.7733/JNFCWT.2021.19.1.51
Seo-yeon Cho, Byong-Woo Kim, Bang, Youngsuk, Keon-Yeop Kim
This is an Open-Access article distributed under the terms of the Creative Commons Attribution Non-Commercial License (http://creativecommons.org/licenses/ by-nc/3.0) which permits unrestricted non-commercial use, distribution, and reproduction in any medium, provided the original work is properly cited Chemical decontamination of primary systems in a nuclear power plant (NPP) prior to commencing the main decommissioning activities is required to reduce radiation exposure during its process. The entire process is repeated until the desired decontamination factor is obtained. To achieve improved decontamination factors over a shorter time with fewer cycles, the appropriate flow characteristics are required. In addition, to prepare an operating procedure that is adaptable to various conditions and situations, the transient analysis results would be required for operator action and system impact assessment. In this study, the flow characteristics in the steady-state and transient conditions for the chemical decontamination operations of the Kori-1 NPP were analyzed and compared via the MARS-KS code simulation. Loss of residual heat removal (RHR) and steam generator tube rupture (SGTR) simulations were conducted for the postulated abnormal events. Loss of RHR results showed the reactor coolant system (RCS) temperature increase, which can damage the reactor coolant pump (RCP)s by its cavitation. The SGTR results indicated a void formation in the RCS interior by the decrease in pressurizer (PZR) pressure, which can cause surface exposure and tripping of the RCPs unless proper actions are taken before the required pressure limit is achieved.
{"title":"Flow Characteristics Analysis for the Chemical Decontamination of the Kori-1 Nuclear Power Plant","authors":"Seo-yeon Cho, Byong-Woo Kim, Bang, Youngsuk, Keon-Yeop Kim","doi":"10.7733/JNFCWT.2021.19.1.51","DOIUrl":"https://doi.org/10.7733/JNFCWT.2021.19.1.51","url":null,"abstract":"This is an Open-Access article distributed under the terms of the Creative Commons Attribution Non-Commercial License (http://creativecommons.org/licenses/ by-nc/3.0) which permits unrestricted non-commercial use, distribution, and reproduction in any medium, provided the original work is properly cited Chemical decontamination of primary systems in a nuclear power plant (NPP) prior to commencing the main decommissioning activities is required to reduce radiation exposure during its process. The entire process is repeated until the desired decontamination factor is obtained. To achieve improved decontamination factors over a shorter time with fewer cycles, the appropriate flow characteristics are required. In addition, to prepare an operating procedure that is adaptable to various conditions and situations, the transient analysis results would be required for operator action and system impact assessment. In this study, the flow characteristics in the steady-state and transient conditions for the chemical decontamination operations of the Kori-1 NPP were analyzed and compared via the MARS-KS code simulation. Loss of residual heat removal (RHR) and steam generator tube rupture (SGTR) simulations were conducted for the postulated abnormal events. Loss of RHR results showed the reactor coolant system (RCS) temperature increase, which can damage the reactor coolant pump (RCP)s by its cavitation. The SGTR results indicated a void formation in the RCS interior by the decrease in pressurizer (PZR) pressure, which can cause surface exposure and tripping of the RCPs unless proper actions are taken before the required pressure limit is achieved.","PeriodicalId":17456,"journal":{"name":"Journal of the Nuclear Fuel Cycle and Waste Technology","volume":"128 1","pages":"51-58"},"PeriodicalIF":0.0,"publicationDate":"2021-03-31","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"83416117","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 2021-03-31DOI: 10.7733/JNFCWT.2021.19.1.75
Hang-Lo Lee, Jin-Seop Kim, Chang-Ho Hong, Ho-Young Jeong, D. Cho
{"title":"Data-Driven Modelling of Damage Prediction of Granite Using Acoustic Emission Parameters in Nuclear Waste Repository","authors":"Hang-Lo Lee, Jin-Seop Kim, Chang-Ho Hong, Ho-Young Jeong, D. Cho","doi":"10.7733/JNFCWT.2021.19.1.75","DOIUrl":"https://doi.org/10.7733/JNFCWT.2021.19.1.75","url":null,"abstract":"","PeriodicalId":17456,"journal":{"name":"Journal of the Nuclear Fuel Cycle and Waste Technology","volume":"2 1","pages":"75-85"},"PeriodicalIF":0.0,"publicationDate":"2021-03-31","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"82191174","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 2021-03-31DOI: 10.7733/JNFCWT.2021.19.1.9
J. John, P. Bartl, Kateřina Čubová, M. Němec, M. Semelová, F. Šebesta, Tereza Šobová, J. Šuľaková, A. Vetešník, D. Vopálka
This is an Open-Access article distributed under the terms of the Creative Commons Attribution Non-Commercial License (http://creativecommons.org/licenses/ by-nc/3.0) which permits unrestricted non-commercial use, distribution, and reproduction in any medium, provided the original work is properly cited As an example of research activities in decontamination for decommissioning, new data are presented on the options for corrosion layer dissolution during the decommissioning decontamination, or persulfate regeneration for decontamination solutions re-use. For the management of spent decontamination solutions, new method based on solvent extraction of radionuclides into ionic liquid followed by electrodeposition of the radionuclides has been developed. Fields of applications of composite inorganic-organic absorbers or solid extractants with polyacrylonitrile (PAN) binding matrix for the treatment of liquid radioactive waste are reviewed; a method for americium separation from the boric acid containing NPP evaporator concentrates based on the TODGA-PAN material is discussed in more detail. Performance of a model of radionuclide transport, developed and implemented within the GoldSim programming environment, for the safety studies of the LLW/ILW repository is demonstrated on the specific case of the Richard repository (Czech Republic). Continuation and even broadening of these activities are expected in connection with the approaching end of the lifespan of the first blocks of the Czech NPPs.
{"title":"Development of New Processes for the Decommissioning Decontamination and for Treatment and Disposal of the Secondary Low- and Intermediate-Level Radioactive Waste","authors":"J. John, P. Bartl, Kateřina Čubová, M. Němec, M. Semelová, F. Šebesta, Tereza Šobová, J. Šuľaková, A. Vetešník, D. Vopálka","doi":"10.7733/JNFCWT.2021.19.1.9","DOIUrl":"https://doi.org/10.7733/JNFCWT.2021.19.1.9","url":null,"abstract":"This is an Open-Access article distributed under the terms of the Creative Commons Attribution Non-Commercial License (http://creativecommons.org/licenses/ by-nc/3.0) which permits unrestricted non-commercial use, distribution, and reproduction in any medium, provided the original work is properly cited As an example of research activities in decontamination for decommissioning, new data are presented on the options for corrosion layer dissolution during the decommissioning decontamination, or persulfate regeneration for decontamination solutions re-use. For the management of spent decontamination solutions, new method based on solvent extraction of radionuclides into ionic liquid followed by electrodeposition of the radionuclides has been developed. Fields of applications of composite inorganic-organic absorbers or solid extractants with polyacrylonitrile (PAN) binding matrix for the treatment of liquid radioactive waste are reviewed; a method for americium separation from the boric acid containing NPP evaporator concentrates based on the TODGA-PAN material is discussed in more detail. Performance of a model of radionuclide transport, developed and implemented within the GoldSim programming environment, for the safety studies of the LLW/ILW repository is demonstrated on the specific case of the Richard repository (Czech Republic). Continuation and even broadening of these activities are expected in connection with the approaching end of the lifespan of the first blocks of the Czech NPPs.","PeriodicalId":17456,"journal":{"name":"Journal of the Nuclear Fuel Cycle and Waste Technology","volume":"44 1","pages":"9-27"},"PeriodicalIF":0.0,"publicationDate":"2021-03-31","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"85419973","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 2021-03-31DOI: 10.7733/JNFCWT.2021.19.1.61
Young-Min Kim, Chang-Lak Kim
This is an Open-Access article distributed under the terms of the Creative Commons Attribution Non-Commercial License (http://creativecommons.org/licenses/ by-nc/3.0) which permits unrestricted non-commercial use, distribution, and reproduction in any medium, provided the original work is properly cited A paradigm shift in the government’s energy policy was reflected in its declaration of early closure of old nuclear plants as well as cancellation of plans for the construction of new plants. To this end, unit 1 of Kori Nuclear Power Plant was permanently shut down and is set for decommission. Based on these changes, the off-site transport of spent fuels from nuclear power plants has become a critical issue. The purpose of this study is to develop an optimized method for transportation of spent fuels from Kori Nuclear Power Plant’s units 1, 2, 3, and 4 to an assumed interim storage facility by simulating the scenarios using the Flexsim software, which is widely used in logistics and manufacturing applications. The results of the simulation suggest that the optimized transport methods may contribute to the development of delivery schedule of spent fuels in the near future. Furthermore, these methods can be applied to decommissioning plan of nuclear power plants.
{"title":"Application of Logistic Simulation for Transport of SFs From Kori Site to an Assumed Interim Storage Facility","authors":"Young-Min Kim, Chang-Lak Kim","doi":"10.7733/JNFCWT.2021.19.1.61","DOIUrl":"https://doi.org/10.7733/JNFCWT.2021.19.1.61","url":null,"abstract":"This is an Open-Access article distributed under the terms of the Creative Commons Attribution Non-Commercial License (http://creativecommons.org/licenses/ by-nc/3.0) which permits unrestricted non-commercial use, distribution, and reproduction in any medium, provided the original work is properly cited A paradigm shift in the government’s energy policy was reflected in its declaration of early closure of old nuclear plants as well as cancellation of plans for the construction of new plants. To this end, unit 1 of Kori Nuclear Power Plant was permanently shut down and is set for decommission. Based on these changes, the off-site transport of spent fuels from nuclear power plants has become a critical issue. The purpose of this study is to develop an optimized method for transportation of spent fuels from Kori Nuclear Power Plant’s units 1, 2, 3, and 4 to an assumed interim storage facility by simulating the scenarios using the Flexsim software, which is widely used in logistics and manufacturing applications. The results of the simulation suggest that the optimized transport methods may contribute to the development of delivery schedule of spent fuels in the near future. Furthermore, these methods can be applied to decommissioning plan of nuclear power plants.","PeriodicalId":17456,"journal":{"name":"Journal of the Nuclear Fuel Cycle and Waste Technology","volume":"52 9","pages":"61-74"},"PeriodicalIF":0.0,"publicationDate":"2021-03-31","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"91480516","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 2021-03-31DOI: 10.7733/JNFCWT.2021.19.1.1
S. Nagasaki
This is an Open-Access article distributed under the terms of the Creative Commons Attribution Non-Commercial License (http://creativecommons.org/licenses/ by-nc/3.0) which permits unrestricted non-commercial use, distribution, and reproduction in any medium, provided the original work is properly cited The pH dependence of sorption distribution coefficient (Kd) of Np(IV) on MX-80 in Ca-Na-Cl type solution with the ionic strength of 0.3 M, which was similar to one of the reference groundwaters in crystalline rock, was experimentally investigated under the reducing conditions. The overall trend of Kd on MX-80 was independent of pH at 5 ≤ pH ≤ 10 but increased as pH increased at pH ≤ 5. The 2-site protolysis non-electrostatic surface complexation and cation exchange model was applied to the experimentally measured pH dependence of Kd and the optimized surface complexation constants of Np(IV) sorption on MX-80 were estimated. The values of surface complexation constants in this work agreed relatively well with those in the Na-Ca-Cl solution previously evaluated, suggesting that compared to Na+, the competition of Ca2+ with Np(IV) for surface complexation on MX-80 was not much strong in Ca-Na-Cl solution. The sorption model well predicted the pH dependence of Kd values but slightly overestimated the sorption at the low pH region.
这是一篇基于知识共享归属非商业许可(http://creativecommons.org/licenses/ by-nc/3.0)的开放获取文章，该许可允许在任何媒介上不受限制地进行非商业使用、分发和复制，前提是正确引用原始作品。在离子强度为0.3 M的Ca-Na-Cl型溶液中，Np(IV)对MX-80的吸附分布系数(Kd)的pH依赖性。与结晶岩中的一种参考地下水相似，在还原条件下进行了实验研究。在5≤pH≤10时，MX-80上Kd的总体变化趋势与pH无关，但在pH≤5时，Kd随pH的增大而增大。将2位元解非静电表面络合和阳离子交换模型应用于实验测量Kd对pH的依赖，并估计了MX-80吸附Np(IV)的最佳表面络合常数。本文计算的表面络合常数值与前人评价的Na- ca - cl溶液中的表面络合常数值吻合较好，表明与Na+相比，Ca-Na-Cl溶液中Ca2+与Np(IV)对MX-80表面络合的竞争并不强。该吸附模型较好地预测了Kd值对pH的依赖关系，但对低pH区吸附的估计略高。
{"title":"Sorption of Np(IV) on MX-80 in Ca-Na-Cl Type Reference Water of Crystalline Rock","authors":"S. Nagasaki","doi":"10.7733/JNFCWT.2021.19.1.1","DOIUrl":"https://doi.org/10.7733/JNFCWT.2021.19.1.1","url":null,"abstract":"This is an Open-Access article distributed under the terms of the Creative Commons Attribution Non-Commercial License (http://creativecommons.org/licenses/ by-nc/3.0) which permits unrestricted non-commercial use, distribution, and reproduction in any medium, provided the original work is properly cited The pH dependence of sorption distribution coefficient (Kd) of Np(IV) on MX-80 in Ca-Na-Cl type solution with the ionic strength of 0.3 M, which was similar to one of the reference groundwaters in crystalline rock, was experimentally investigated under the reducing conditions. The overall trend of Kd on MX-80 was independent of pH at 5 ≤ pH ≤ 10 but increased as pH increased at pH ≤ 5. The 2-site protolysis non-electrostatic surface complexation and cation exchange model was applied to the experimentally measured pH dependence of Kd and the optimized surface complexation constants of Np(IV) sorption on MX-80 were estimated. The values of surface complexation constants in this work agreed relatively well with those in the Na-Ca-Cl solution previously evaluated, suggesting that compared to Na+, the competition of Ca2+ with Np(IV) for surface complexation on MX-80 was not much strong in Ca-Na-Cl solution. The sorption model well predicted the pH dependence of Kd values but slightly overestimated the sorption at the low pH region.","PeriodicalId":17456,"journal":{"name":"Journal of the Nuclear Fuel Cycle and Waste Technology","volume":"98 1","pages":"1-7"},"PeriodicalIF":0.0,"publicationDate":"2021-03-31","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"76148668","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 2021-03-31DOI: 10.7733/JNFCWT.2021.19.1.141
Eun-Jin Sim, Sun-Kee Lee, Chang-Lak Kim, Tae-Man Kim
{"title":"Preparation of the Applicable Regulatory Guideline on Mixed Waste in Korea Based on the Analysis of US Laws and Regulations","authors":"Eun-Jin Sim, Sun-Kee Lee, Chang-Lak Kim, Tae-Man Kim","doi":"10.7733/JNFCWT.2021.19.1.141","DOIUrl":"https://doi.org/10.7733/JNFCWT.2021.19.1.141","url":null,"abstract":"","PeriodicalId":17456,"journal":{"name":"Journal of the Nuclear Fuel Cycle and Waste Technology","volume":"15 1","pages":"141-160"},"PeriodicalIF":0.0,"publicationDate":"2021-03-31","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"85504710","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 2021-03-31DOI: 10.7733/JNFCWT.2021.19.1.123
Jiseon Jang, Tae-Man Kim, C. Cho, D. Lee
This is an Open-Access article distributed under the terms of the Creative Commons Attribution Non-Commercial License (http://creativecommons.org/licenses/ by-nc/3.0) which permits unrestricted non-commercial use, distribution, and reproduction in any medium, provided the original work is properly cited Radiological impact analyses were carried out for a near-surface radioactive waste repository at Gyeongju in South Korea. The RESRAD-ONSITE code was applied for the estimation of maximum exposure doses by considering various exposure pathways based on a land area of 2,500 m2 with a 0.15 m thick contamination zone. Typical influencing input parameters such as shield depth, shield materials’ density, and shield erosion rate were examined for a sensitivity analysis. Then both residential farmer and industrial worker scenarios were used for the estimation of maximum exposure doses depending on exposure duration. The radiation dose evaluation results showed that 60Co, 137Cs, and 63Ni were major contributors to the total exposure dose compared with other radionuclides. Furthermore, the total exposure dose from ingestion (plant, meat, and milk) of the contaminated plants was more significant than those assessed for inhalation, with maximum values of 5.5×10−4 mSv‧yr−1 for the plant ingestion. Thus the results of this study can be applied for determining near-surface radioactive waste repository conditions and providing quantitative analysis methods using RESRAD-ONSITE code for the safety assessment of disposing radioactive materials including decommissioning wastes to protect human health and the environment.
{"title":"Radiological Safety Assessment for a Near-Surface Disposal Facility Using RESRAD-ONSITE Code","authors":"Jiseon Jang, Tae-Man Kim, C. Cho, D. Lee","doi":"10.7733/JNFCWT.2021.19.1.123","DOIUrl":"https://doi.org/10.7733/JNFCWT.2021.19.1.123","url":null,"abstract":"This is an Open-Access article distributed under the terms of the Creative Commons Attribution Non-Commercial License (http://creativecommons.org/licenses/ by-nc/3.0) which permits unrestricted non-commercial use, distribution, and reproduction in any medium, provided the original work is properly cited Radiological impact analyses were carried out for a near-surface radioactive waste repository at Gyeongju in South Korea. The RESRAD-ONSITE code was applied for the estimation of maximum exposure doses by considering various exposure pathways based on a land area of 2,500 m2 with a 0.15 m thick contamination zone. Typical influencing input parameters such as shield depth, shield materials’ density, and shield erosion rate were examined for a sensitivity analysis. Then both residential farmer and industrial worker scenarios were used for the estimation of maximum exposure doses depending on exposure duration. The radiation dose evaluation results showed that 60Co, 137Cs, and 63Ni were major contributors to the total exposure dose compared with other radionuclides. Furthermore, the total exposure dose from ingestion (plant, meat, and milk) of the contaminated plants was more significant than those assessed for inhalation, with maximum values of 5.5×10−4 mSv‧yr−1 for the plant ingestion. Thus the results of this study can be applied for determining near-surface radioactive waste repository conditions and providing quantitative analysis methods using RESRAD-ONSITE code for the safety assessment of disposing radioactive materials including decommissioning wastes to protect human health and the environment.","PeriodicalId":17456,"journal":{"name":"Journal of the Nuclear Fuel Cycle and Waste Technology","volume":"11 1","pages":"123-132"},"PeriodicalIF":0.0,"publicationDate":"2021-03-31","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"89784786","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 2021-03-31DOI: 10.7733/JNFCWT.2021.19.1.29
R. Trtílek, V. Havlová, J. Podlaha, K. Svoboda, Tomas Otcovský
{"title":"Status of Czech Low and Intermediate Radioactive Waste Management in the Context of European Development","authors":"R. Trtílek, V. Havlová, J. Podlaha, K. Svoboda, Tomas Otcovský","doi":"10.7733/JNFCWT.2021.19.1.29","DOIUrl":"https://doi.org/10.7733/JNFCWT.2021.19.1.29","url":null,"abstract":"","PeriodicalId":17456,"journal":{"name":"Journal of the Nuclear Fuel Cycle and Waste Technology","volume":"23 1","pages":"29-38"},"PeriodicalIF":0.0,"publicationDate":"2021-03-31","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"90945862","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
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