Pub Date : 2020-06-30DOI: 10.7733/jnfcwt.2020.18.2.217
Taewoong Lee, Seongmin Jo, Changyeon Yoon, Nakjeom Kim
Concrete waste accounts for approximately 70~80% of the total waste generated during the decommissioning of nuclear power plants (NPPs). Based upon the concentration of each radionuclide, the concrete waste from the decommissioning can be used in the determination of the clearance threshold used to classify waste as radioactive. To reduce the cost of radioactive concrete waste disposal, it is important to perform decontamination before self-disposal or limited recycling. Therefore, it is necessary to estimate the internal radioactivity distribution of radioactive concrete waste to ensure effective decontamination. In this study, the performance metrics of various Compton reconstruction algorithms were compared in order to identify the best strategy to estimate the internal radioactivity distribution in concrete waste during the decommissioning of NPPs. Four reconstruction algorithms, namely, simple back-projection, filtered back-projection, maximum likelihood expectation maximization (MLEM), and energy-deconvolution MLEM (E-MLEM) were used as Compton reconstruction algorithms. Subsequently, the results obtained by using these various reconstruction algorithms were compared with one another and evaluated, using quantitative evaluation methods. The MLEM and E-MLEM reconstruction algorithms exhibited the best performance in maintaining a high image resolution and signal-to-noise ratio (SNR), respectively. The results of this study demonstrate the feasibility of using Compton images in the estimation of the internal radioactive distribution of concrete during the decommissioning of NPPs.
{"title":"Comparison of Compton Image Reconstruction Algorithms for Estimation of Internal Radioactivity Distribution in Concrete Waste During Decommissioning of Nuclear Power Plant","authors":"Taewoong Lee, Seongmin Jo, Changyeon Yoon, Nakjeom Kim","doi":"10.7733/jnfcwt.2020.18.2.217","DOIUrl":"https://doi.org/10.7733/jnfcwt.2020.18.2.217","url":null,"abstract":"Concrete waste accounts for approximately 70~80% of the total waste generated during the decommissioning of nuclear power plants (NPPs). Based upon the concentration of each radionuclide, the concrete waste from the decommissioning can be used in the determination of the clearance threshold used to classify waste as radioactive. To reduce the cost of radioactive concrete waste disposal, it is important to perform decontamination before self-disposal or limited recycling. Therefore, it is necessary to estimate the internal radioactivity distribution of radioactive concrete waste to ensure effective decontamination. In this study, the performance metrics of various Compton reconstruction algorithms were compared in order to identify the best strategy to estimate the internal radioactivity distribution in concrete waste during the decommissioning of NPPs. Four reconstruction algorithms, namely, simple back-projection, filtered back-projection, maximum likelihood expectation maximization (MLEM), and energy-deconvolution MLEM (E-MLEM) were used as Compton reconstruction algorithms. Subsequently, the results obtained by using these various reconstruction algorithms were compared with one another and evaluated, using quantitative evaluation methods. The MLEM and E-MLEM reconstruction algorithms exhibited the best performance in maintaining a high image resolution and signal-to-noise ratio (SNR), respectively. The results of this study demonstrate the feasibility of using Compton images in the estimation of the internal radioactive distribution of concrete during the decommissioning of NPPs.","PeriodicalId":17456,"journal":{"name":"Journal of the Nuclear Fuel Cycle and Waste Technology","volume":"1 1","pages":"217-225"},"PeriodicalIF":0.0,"publicationDate":"2020-06-30","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"89710675","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 : 2020-06-30DOI: 10.7733/jnfcwt.2020.18.2.207
S. Paek, C. Lee, D. Yoon, Sung-Jai Lee
The feasibility of rare earth (RE) removal process via oxidation reactions with UCl3 was investigated using the HSC Chemistry code to reduce the concentrations of RE in transuranic (TRU) products. The composition and thermodynamic data of TRU and RE elements contained in the reference spent fuel were examined. The reactivity was evaluated by calculating equilibrium data considering oxidation reactions with UCl3. Both RE removal rate and TRU recovery rate were evaluated for the two cases, wherein TRU products with different RE concentrations were used. When TRU products were reacted with UCl3, selective oxidation was driven by the difference in the Gibbs free energy of each element. The calculation results imply that the TRU/RE ratio of the final product can be increased by removing RE elements while maintaining the maximum recovery rate of TRU, which is accomplished by controlling the amount of UCl3 injected. Since the results of this study are based on thermodynamic equilibrium data, there are many limitations to apply to the actual process. However, it is expected to be used as an important data for the process design to supply the TRU product of pyroprocessing to SFR’s fuel demanding low RE concentrations.
{"title":"Simulation of Rare Earth Elements Removal Behavior in TRU Product Using HSC Chemistry Code","authors":"S. Paek, C. Lee, D. Yoon, Sung-Jai Lee","doi":"10.7733/jnfcwt.2020.18.2.207","DOIUrl":"https://doi.org/10.7733/jnfcwt.2020.18.2.207","url":null,"abstract":"The feasibility of rare earth (RE) removal process via oxidation reactions with UCl3 was investigated using the HSC Chemistry code to reduce the concentrations of RE in transuranic (TRU) products. The composition and thermodynamic data of TRU and RE elements contained in the reference spent fuel were examined. The reactivity was evaluated by calculating equilibrium data considering oxidation reactions with UCl3. Both RE removal rate and TRU recovery rate were evaluated for the two cases, wherein TRU products with different RE concentrations were used. When TRU products were reacted with UCl3, selective oxidation was driven by the difference in the Gibbs free energy of each element. The calculation results imply that the TRU/RE ratio of the final product can be increased by removing RE elements while maintaining the maximum recovery rate of TRU, which is accomplished by controlling the amount of UCl3 injected. Since the results of this study are based on thermodynamic equilibrium data, there are many limitations to apply to the actual process. However, it is expected to be used as an important data for the process design to supply the TRU product of pyroprocessing to SFR’s fuel demanding low RE concentrations.","PeriodicalId":17456,"journal":{"name":"Journal of the Nuclear Fuel Cycle and Waste Technology","volume":"29 1","pages":"207-215"},"PeriodicalIF":0.0,"publicationDate":"2020-06-30","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"82833654","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 : 2020-06-30DOI: 10.7733/jnfcwt.2020.18.2.195
J. Moon
Although the national radiological emergency response system has been improved by incorporating lessons from the Fukushima nuclear power plant accident and recent domestic natural disasters, it has not fully incorporated these lessons. In addition, it cannot deal with a variety of aftermath of the radiological disaster. Even for the same disaster, the national emergency response system should comply with multiple domestic laws in our country. Furthermore, there are a few discrepancies between the articles of the domestic laws that the national radiological emergency response system should address. Therefore, this study investigates the characteristics of radiological disasters, examines articles on the domestic laws related to the national radiological emergency response system, and analyses the Japanese government’s responses to the Fukushima nuclear power plant accident. Based on the results of the review, suggestions for the improvement of the national radiological emergency response system in terms of response organization and framework have been proposed in this study.
{"title":"Suggestions to Improve the Effectiveness of National Radiological Emergency Response System","authors":"J. Moon","doi":"10.7733/jnfcwt.2020.18.2.195","DOIUrl":"https://doi.org/10.7733/jnfcwt.2020.18.2.195","url":null,"abstract":"Although the national radiological emergency response system has been improved by incorporating lessons from the Fukushima nuclear power plant accident and recent domestic natural disasters, it has not fully incorporated these lessons. In addition, it cannot deal with a variety of aftermath of the radiological disaster. Even for the same disaster, the national emergency response system should comply with multiple domestic laws in our country. Furthermore, there are a few discrepancies between the articles of the domestic laws that the national radiological emergency response system should address. Therefore, this study investigates the characteristics of radiological disasters, examines articles on the domestic laws related to the national radiological emergency response system, and analyses the Japanese government’s responses to the Fukushima nuclear power plant accident. Based on the results of the review, suggestions for the improvement of the national radiological emergency response system in terms of response organization and framework have been proposed in this study.","PeriodicalId":17456,"journal":{"name":"Journal of the Nuclear Fuel Cycle and Waste Technology","volume":"39 1","pages":"195-206"},"PeriodicalIF":0.0,"publicationDate":"2020-06-30","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"79855837","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 : 2020-06-30DOI: 10.7733/jnfcwt.2020.18.2.113
B. Westphal, D. Tolman, K. Tolman, S. M. Frank, S. Herrmann, S. Warmann, K. Marsden, M. Patterson
An options study was performed for the treatment of residual elemental sodium in driver plenums following the chopping operation during the pyroprocessing of used nuclear fuel. Given the pending availability of a multi-function furnace for distillation and consolidation operations in the Fuel Conditioning Facility, the furnace was considered for the processing of driver plenums. Although two options (oxidation and distillation) could be performed in the multi-function furnace, neither option has been developed sufficiently to date to warrant the use of the furnace for treatment operations. Thus, it was decided to defer the treatment of elemental sodium from driver plenums in the multi-function furnace until more developed technologies and/or furnaces become available. In the interim, storage of the plenums and characterization efforts are recommended.
{"title":"Options Study for the Neutralization of Elemental Sodium During the Pyroprocessing of Used Nuclear Fuel","authors":"B. Westphal, D. Tolman, K. Tolman, S. M. Frank, S. Herrmann, S. Warmann, K. Marsden, M. Patterson","doi":"10.7733/jnfcwt.2020.18.2.113","DOIUrl":"https://doi.org/10.7733/jnfcwt.2020.18.2.113","url":null,"abstract":"An options study was performed for the treatment of residual elemental sodium in driver plenums following the chopping operation during the pyroprocessing of used nuclear fuel. Given the pending availability of a multi-function furnace for distillation and consolidation operations in the Fuel Conditioning Facility, the furnace was considered for the processing of driver plenums. Although two options (oxidation and distillation) could be performed in the multi-function furnace, neither option has been developed sufficiently to date to warrant the use of the furnace for treatment operations. Thus, it was decided to defer the treatment of elemental sodium from driver plenums in the multi-function furnace until more developed technologies and/or furnaces become available. In the interim, storage of the plenums and characterization efforts are recommended.","PeriodicalId":17456,"journal":{"name":"Journal of the Nuclear Fuel Cycle and Waste Technology","volume":"55 1","pages":"113-118"},"PeriodicalIF":0.0,"publicationDate":"2020-06-30","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"79755720","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 : 2020-03-30DOI: 10.7733/jnfcwt.2020.18.1.83
Jinseok Han, Junkyung Jang, Hyun Chul Lee
The Graphite Isotope Ratio Method (GIRM) can verify non-proliferation of nuclear weapon by estimating the total plutonium production in a graphite-moderated reactor. Using the reactor, plutonium is generated and accumulated through the 238U neutron capture reaction, and impurities in the graphite are converted to nuclides due to the nuclear reaction. Therefore, the amount of plutonium production and concentration of the impurities are correlated. However, the plutonium production cannot be predicted using only the absolute concentration of the impurities. It can only be predicted when the initial concentration of the impurities is obtained because the concentration, at a certain time, depends on it. Nevertheless, the ratios of the isotopes in an element are known regardless of the impurity of an element in the graphite moderator. Thus, the correlation between the isotope ratio and amount of plutonium produced helps predict plutonium production in a graphite-moderated reactor. Boron, Lithium, Chlorine, Titanium, and Uranium are known as indicator elements in the GIRM. To assess whether the correlation between the indicator isotope and amount of plutonium produced is independent of the initial concentration of the impurities, four different impurity compositions of graphite were used. 10B/11B, 36Cl/35Cl, 48Ti/49Ti, and 235U/238U had a consistent correlation with the cumulative plutonium production, regardless of the initial impurity concentration of the graphite, because these isotopes were not generated through the nuclear reaction of other elements. On the other hand, the correlation between 6Li/7Li and plutonium production depended on the initial concentration of the impurities in graphite. Although 7Li can be produced through the neutron capture reaction of 6Li, the (n, α) reaction of 10B was the major source of 7Li. Therefore, the initial concentration of 10B affected the production of 7Li, making Li unsuitable as an indicator element for the GIRM.
{"title":"A Suitability Study on the Indicator Isotopes for Graphite Isotope Ratio Method (GIRM)","authors":"Jinseok Han, Junkyung Jang, Hyun Chul Lee","doi":"10.7733/jnfcwt.2020.18.1.83","DOIUrl":"https://doi.org/10.7733/jnfcwt.2020.18.1.83","url":null,"abstract":"The Graphite Isotope Ratio Method (GIRM) can verify non-proliferation of nuclear weapon by estimating the total plutonium production in a graphite-moderated reactor. Using the reactor, plutonium is generated and accumulated through the 238U neutron capture reaction, and impurities in the graphite are converted to nuclides due to the nuclear reaction. Therefore, the amount of plutonium production and concentration of the impurities are correlated. However, the plutonium production cannot be predicted using only the absolute concentration of the impurities. It can only be predicted when the initial concentration of the impurities is obtained because the concentration, at a certain time, depends on it. Nevertheless, the ratios of the isotopes in an element are known regardless of the impurity of an element in the graphite moderator. Thus, the correlation between the isotope ratio and amount of plutonium produced helps predict plutonium production in a graphite-moderated reactor. Boron, Lithium, Chlorine, Titanium, and Uranium are known as indicator elements in the GIRM. To assess whether the correlation between the indicator isotope and amount of plutonium produced is independent of the initial concentration of the impurities, four different impurity compositions of graphite were used. 10B/11B, 36Cl/35Cl, 48Ti/49Ti, and 235U/238U had a consistent correlation with the cumulative plutonium production, regardless of the initial impurity concentration of the graphite, because these isotopes were not generated through the nuclear reaction of other elements. On the other hand, the correlation between 6Li/7Li and plutonium production depended on the initial concentration of the impurities in graphite. Although 7Li can be produced through the neutron capture reaction of 6Li, the (n, α) reaction of 10B was the major source of 7Li. Therefore, the initial concentration of 10B affected the production of 7Li, making Li unsuitable as an indicator element for the GIRM.","PeriodicalId":17456,"journal":{"name":"Journal of the Nuclear Fuel Cycle and Waste Technology","volume":"6 1","pages":"83-90"},"PeriodicalIF":0.0,"publicationDate":"2020-03-30","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"83647390","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 : 2020-03-30DOI: 10.7733/jnfcwt.2020.18.1.51
Jun-Ha Kim, Jea-Hak Cheong, Sangbum Hong, B. Seo, B. Lee
This study was conducted to develop a method for depth assessment of embedded sources using gamma-spectrum ratio and for the evaluation of field applicability. To this end, Peak to Compton and Peak to valley ratio changes were evaluated according to 137Cs, 60Co, 152Eu point source depth using HPGe detector and MCNP simulation. The effects of measurement distance of PTV and PTC methods were evaluated. Using the results, the source depth assessment equation using the PTC and PTV methods was derived based on the detection distance of 50 cm. In addition, the sensitivity of detection distance changes was assessed when using PTV and PTC methods, and error increased by 3 to 4 cm when detection distance decreased by 20 cm based on 50 cm. However, it was confirmed that if the detection distance was increased to 100 cm, the effects of detection distance were small. And PTV and PTC methods were compared with the two distance measurement method which evaluates the depth of source by the change of net peak counting rate according to the detection distance. As a result of source depth assessment, the PTV and PTC showed a maximum error of 1.87 cm and the two distance measurement method showed maximum error of 2.69 cm. The results of the experiment confirmed that the accuracy of the PTV and PTC methods was higher than two distance measurement. In addition, Sensitivity evaluation by horizontal position error of source has maximum error of less than 25.59 cm for the two distance measurement method. On the other hand, PTV and PTC method showed high accuracy with maximum error of less than 8.04 cm. In addition, the PTC method has lowest standard deviation for the same time measurement, which is expected to enable rapid measurement.
{"title":"Study on Development of Embedded Source Depth Assessment Method Using Gamma Spectrum Ratio","authors":"Jun-Ha Kim, Jea-Hak Cheong, Sangbum Hong, B. Seo, B. Lee","doi":"10.7733/jnfcwt.2020.18.1.51","DOIUrl":"https://doi.org/10.7733/jnfcwt.2020.18.1.51","url":null,"abstract":"This study was conducted to develop a method for depth assessment of embedded sources using gamma-spectrum ratio and for the evaluation of field applicability. To this end, Peak to Compton and Peak to valley ratio changes were evaluated according to 137Cs, 60Co, 152Eu point source depth using HPGe detector and MCNP simulation. The effects of measurement distance of PTV and PTC methods were evaluated. Using the results, the source depth assessment equation using the PTC and PTV methods was derived based on the detection distance of 50 cm. In addition, the sensitivity of detection distance changes was assessed when using PTV and PTC methods, and error increased by 3 to 4 cm when detection distance decreased by 20 cm based on 50 cm. However, it was confirmed that if the detection distance was increased to 100 cm, the effects of detection distance were small. And PTV and PTC methods were compared with the two distance measurement method which evaluates the depth of source by the change of net peak counting rate according to the detection distance. As a result of source depth assessment, the PTV and PTC showed a maximum error of 1.87 cm and the two distance measurement method showed maximum error of 2.69 cm. The results of the experiment confirmed that the accuracy of the PTV and PTC methods was higher than two distance measurement. In addition, Sensitivity evaluation by horizontal position error of source has maximum error of less than 25.59 cm for the two distance measurement method. On the other hand, PTV and PTC method showed high accuracy with maximum error of less than 8.04 cm. In addition, the PTC method has lowest standard deviation for the same time measurement, which is expected to enable rapid measurement.","PeriodicalId":17456,"journal":{"name":"Journal of the Nuclear Fuel Cycle and Waste Technology","volume":"12 1","pages":"51-62"},"PeriodicalIF":0.0,"publicationDate":"2020-03-30","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"90885519","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 : 2020-03-30DOI: 10.7733/jnfcwt.2020.18.1.73
Chung-Kyun Park, Tae-Jin Park, Seung-Yup Lee, Jae-Kwang Lee
Strontium (90Sr) and nickel (59Ni) have been considered as key radionuclides in the safety assessment of radioactive waste disposal. Through various efforts to impede the migration of radioactive nuclides underground, it has been established that some minerals generated from the corrosion of the waste containers have a positive chemical interaction with these radionuclides. Among these minerals we selected mackinawite (FeS), an iron and sulfur compound, and performed a sorption experiment for the Sr and Ni in FeS under anoxic and alkaline conditions by reflecting deep underground environments. The effects of pH on sorption were likewise investigated in the pH range of 8 ~ 12. As a result, it was found that strontium failed to exhibit a good sorption capacity in a weak alkaline range, while nickel showed a noticeably higher sorption affinity over the entire experimental pH range. Moreover, we determined that as the pH increased in the solution, the distribution coefficients (Kd) were increased for both nuclides, which reflects when an alkalinity increses, the surface of the mineral charges much negatively by detaching the hydrogen or cations on the mineral surface. Thus, it can be concluded that the cationic nuclides of Sr and Ni can attach easily to the mineral under strong alkalinity.
{"title":"Sorption Characteristics of Strontium and Nickel on Mackinawite According to pH Variations in Alkaline Conditions","authors":"Chung-Kyun Park, Tae-Jin Park, Seung-Yup Lee, Jae-Kwang Lee","doi":"10.7733/jnfcwt.2020.18.1.73","DOIUrl":"https://doi.org/10.7733/jnfcwt.2020.18.1.73","url":null,"abstract":"Strontium (90Sr) and nickel (59Ni) have been considered as key radionuclides in the safety assessment of radioactive waste disposal. Through various efforts to impede the migration of radioactive nuclides underground, it has been established that some minerals generated from the corrosion of the waste containers have a positive chemical interaction with these radionuclides. Among these minerals we selected mackinawite (FeS), an iron and sulfur compound, and performed a sorption experiment for the Sr and Ni in FeS under anoxic and alkaline conditions by reflecting deep underground environments. The effects of pH on sorption were likewise investigated in the pH range of 8 ~ 12. As a result, it was found that strontium failed to exhibit a good sorption capacity in a weak alkaline range, while nickel showed a noticeably higher sorption affinity over the entire experimental pH range. Moreover, we determined that as the pH increased in the solution, the distribution coefficients (Kd) were increased for both nuclides, which reflects when an alkalinity increses, the surface of the mineral charges much negatively by detaching the hydrogen or cations on the mineral surface. Thus, it can be concluded that the cationic nuclides of Sr and Ni can attach easily to the mineral under strong alkalinity.","PeriodicalId":17456,"journal":{"name":"Journal of the Nuclear Fuel Cycle and Waste Technology","volume":"35 1","pages":"73-81"},"PeriodicalIF":0.0,"publicationDate":"2020-03-30","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"73416777","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 : 2020-03-30DOI: 10.7733/jnfcwt.2020.18.1.43
Seok Yoon, S. Park, Min-Seop Kim, Geon-Young Kim, Seung-Rae Lee
An engineered barrier system (EBS) for the geological disposal of high-level radioactive waste (HLW) consists of a disposal canister packed with spent fuel, buffer material, backfill material, and gap-filling material. The buffer material fills the space between the canister and the near-field rock, thus serving to restrain the release of radionuclides and protect the canister from groundwater penetration. Furthermore, as significant amounts of heat energy are released from the canister to the surrounding rock, the thermal conductivity of the buffer plays an important role in maintaining the safety of the entire disposal system. Therefore, given the high levels of heat released from disposal canisters, this study measured the thermal conductivities of compacted bentonite buffers from Gyeongju under temperature variations ranging 25 to 80~90℃. There was a 5~20% increase in thermal conductivity as the temperature increased, and the temperature effect increased as the degree of saturation increased.
{"title":"Thermal Conductivity Evaluation of Compacted Bentonite Buffers Considering Temperature Variations","authors":"Seok Yoon, S. Park, Min-Seop Kim, Geon-Young Kim, Seung-Rae Lee","doi":"10.7733/jnfcwt.2020.18.1.43","DOIUrl":"https://doi.org/10.7733/jnfcwt.2020.18.1.43","url":null,"abstract":"An engineered barrier system (EBS) for the geological disposal of high-level radioactive waste (HLW) consists of a disposal canister packed with spent fuel, buffer material, backfill material, and gap-filling material. The buffer material fills the space between the canister and the near-field rock, thus serving to restrain the release of radionuclides and protect the canister from groundwater penetration. Furthermore, as significant amounts of heat energy are released from the canister to the surrounding rock, the thermal conductivity of the buffer plays an important role in maintaining the safety of the entire disposal system. Therefore, given the high levels of heat released from disposal canisters, this study measured the thermal conductivities of compacted bentonite buffers from Gyeongju under temperature variations ranging 25 to 80~90℃. There was a 5~20% increase in thermal conductivity as the temperature increased, and the temperature effect increased as the degree of saturation increased.","PeriodicalId":17456,"journal":{"name":"Journal of the Nuclear Fuel Cycle and Waste Technology","volume":"48 1","pages":"43-49"},"PeriodicalIF":0.0,"publicationDate":"2020-03-30","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"79761870","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 : 2020-03-30DOI: 10.7733/jnfcwt.2020.18.1.103
Lee Bong Jae, Yong-Kyu Kwon, Changseob Hong, Dong Won Lee, Kyongnam Min
Pulling-type cutting devices, which use a diamond wire saw, have been used generally for cutting concrete structures. In this study, a pushing-type cutting device with a collection cover was developed by overcoming the disadvantages of pulling-type devices. In this device, dry or liquid methods can be selected to cool frictional heat. Operation and leakage tests of the dust generated during the dismantling of a concrete structure were carried out, confirming the suitable operation of the fabricated cutting device; the leakage rate was approximately 1.7%. For a conservative evaluation, the internal dose of workers was estimated in dismantling the core center part of biological shield concrete with a specific activity of 99.5 Bq·g-1. The committed effective dose per worker was 0.25 mSv. The developed cutting device contributed to reducing radioactive concrete waste and minimizing worker exposure due to its easy installation. Therefore, it can be utilized as a cutting apparatus for dismantling not only reinforced concrete structures but also radioactive biological shield concrete in nuclear power plant decommissioning efforts.
{"title":"Development of the Pushing Type Cutting Device to Dismantle Concrete Structure for Decommissioning of Nuclear Power Plant","authors":"Lee Bong Jae, Yong-Kyu Kwon, Changseob Hong, Dong Won Lee, Kyongnam Min","doi":"10.7733/jnfcwt.2020.18.1.103","DOIUrl":"https://doi.org/10.7733/jnfcwt.2020.18.1.103","url":null,"abstract":"Pulling-type cutting devices, which use a diamond wire saw, have been used generally for cutting concrete structures. In this study, a pushing-type cutting device with a collection cover was developed by overcoming the disadvantages of pulling-type devices. In this device, dry or liquid methods can be selected to cool frictional heat. Operation and leakage tests of the dust generated during the dismantling of a concrete structure were carried out, confirming the suitable operation of the fabricated cutting device; the leakage rate was approximately 1.7%. For a conservative evaluation, the internal dose of workers was estimated in dismantling the core center part of biological shield concrete with a specific activity of 99.5 Bq·g-1. The committed effective dose per worker was 0.25 mSv. The developed cutting device contributed to reducing radioactive concrete waste and minimizing worker exposure due to its easy installation. Therefore, it can be utilized as a cutting apparatus for dismantling not only reinforced concrete structures but also radioactive biological shield concrete in nuclear power plant decommissioning efforts.","PeriodicalId":17456,"journal":{"name":"Journal of the Nuclear Fuel Cycle and Waste Technology","volume":"53 1","pages":"103-111"},"PeriodicalIF":0.0,"publicationDate":"2020-03-30","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"77881820","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 : 2020-03-30DOI: 10.7733/jnfcwt.2020.18.1.63
E. Jung, Tae-Hyeong Kim, Yongheum Jo, Hee-Kyung Kim, Hye-Ryun Cho, W. Cha, M. Baik, J. Yun
Dissolved hexavalent uranium can exist in the form of several different chemical species. Furthermore, species distributions depend on the pH value of the aqueous solution. Representatively, UO2 2+, UO2OH+, (UO2)2(OH)2 2+, and (UO2)3(OH)5 + species coexist in solutions at acidic and circumneutral pH values. When amorphous silica particles are suspended in an aqueous solution, the dissolved chemical species are easily adsorbed onto silica surfaces. In this study, it was examined whether the species distribution of the adsorbed U(VI) on a silica surface followed that of the dissolved U(VI) in an aqueous solution. Time-resolved luminescence spectra of three different dissolved species (UO2 2+, UO2OH+, and (UO2)3(OH)5 +) and two different adsorbed species (≡SiO2UO2, ≡SiO2(UO2)OH–, or ≡SiO2(UO2)3(OH)5–) were measured in the pH range 3.5−7.5. The spectral shapes of these chemical species were compared by changing the pH value; consequently, it was confirmed that the species distribution of the adsorbed U(VI) species was different from that of the dissolved U(VI) species.
{"title":"Study on the Species Distributions of Dissolved U(VI) and Adsorbed U(VI) on Silica Surface","authors":"E. Jung, Tae-Hyeong Kim, Yongheum Jo, Hee-Kyung Kim, Hye-Ryun Cho, W. Cha, M. Baik, J. Yun","doi":"10.7733/jnfcwt.2020.18.1.63","DOIUrl":"https://doi.org/10.7733/jnfcwt.2020.18.1.63","url":null,"abstract":"Dissolved hexavalent uranium can exist in the form of several different chemical species. Furthermore, species distributions depend on the pH value of the aqueous solution. Representatively, UO2 2+, UO2OH+, (UO2)2(OH)2 2+, and (UO2)3(OH)5 + species coexist in solutions at acidic and circumneutral pH values. When amorphous silica particles are suspended in an aqueous solution, the dissolved chemical species are easily adsorbed onto silica surfaces. In this study, it was examined whether the species distribution of the adsorbed U(VI) on a silica surface followed that of the dissolved U(VI) in an aqueous solution. Time-resolved luminescence spectra of three different dissolved species (UO2 2+, UO2OH+, and (UO2)3(OH)5 +) and two different adsorbed species (≡SiO2UO2, ≡SiO2(UO2)OH–, or ≡SiO2(UO2)3(OH)5–) were measured in the pH range 3.5−7.5. The spectral shapes of these chemical species were compared by changing the pH value; consequently, it was confirmed that the species distribution of the adsorbed U(VI) species was different from that of the dissolved U(VI) species.","PeriodicalId":17456,"journal":{"name":"Journal of the Nuclear Fuel Cycle and Waste Technology","volume":"83 1","pages":"63-72"},"PeriodicalIF":0.0,"publicationDate":"2020-03-30","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"89024674","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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