The work deals with the comparison of voltage dividers with different wiring of resistors and capacitors. The voltage divider recommended by the manufacturer of the used photomultiplier (ET Enterprises, GB) was used as a reference. The aim was to assess the influence of different voltage dividers together with the large volume scintillation detector on detection parameters such as the efficiency of the measurement and dose rate linearity using 241Am, 137Cs, and 60Co gamma sources. The experiments showed relatively great differences between 15_10(_C) voltage divider (R1 and R2-R11, 15 MW and 10 MW) and the rest. Furthermore, it was confirmed that the voltage divider recommended by the manufacturer of the used photomultiplier showed the best results, but some of the measured dividers exhibited similar results and therefore can be used in radiation portal monitors as well.
{"title":"Comparison of the response of different voltage dividers to low-level measurement using a large plastic scintillator intended for radiation portal monitors","authors":"L. Fiserova, J. Janda, Pavel Skotak","doi":"10.2298/ntrp2103249f","DOIUrl":"https://doi.org/10.2298/ntrp2103249f","url":null,"abstract":"The work deals with the comparison of voltage dividers with different wiring of resistors and capacitors. The voltage divider recommended by the manufacturer of the used photomultiplier (ET Enterprises, GB) was used as a reference. The aim was to assess the influence of different voltage dividers together with the large volume scintillation detector on detection parameters such as the efficiency of the measurement and dose rate linearity using 241Am, 137Cs, and 60Co gamma sources. The experiments showed relatively great differences between 15_10(_C) voltage divider (R1 and R2-R11, 15 MW and 10 MW) and the rest. Furthermore, it was confirmed that the voltage divider recommended by the manufacturer of the used photomultiplier showed the best results, but some of the measured dividers exhibited similar results and therefore can be used in radiation portal monitors as well.","PeriodicalId":49734,"journal":{"name":"Nuclear Technology & Radiation Protection","volume":"1 1","pages":""},"PeriodicalIF":1.2,"publicationDate":"2021-01-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"68546427","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":4,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Busra Durmaz, Gizem Bakir, Bugra Arslan, H. Yapici
This paper presents analyses of enrichments of uranium taken out from Canada Deuterium Uranium and pressurized water reactors spent fuels and fissile fuel breeding from thorium in two different helium cooled-accelerator driven system designs, DESIGN A and DESIGN B. In the beginning, the 235U percentages in the uranium fuels taken out from the reactors spent fuels are 0.17% and 0.91%, respectively. Both system cores are fuelled with two different minor actinides compositions extracted from PWR-MOX spent fuels. The DESIGN A has one transmutation zone (enrichment zone) surrounding the fuel core and containing thorium or spent uranium fuels, while DESIGN B has a second transmutation zone (fissile fuel breeding zone) surrounding the first transmutation zone and containing only thorium fuel. In brief, a total of ten cases formed by the combinations of accelerator driven system designs, minor actinides components, and spent uranium with thorium fuels are analysed, which are six in DESIGN A containing one transmutation zone and four in DESIGN B containing two transmutation zones. Lead-bismuth eutectic alloy, a liquid heavy metal, consisting of 45% lead and 55 % bismuth is used as target material in the investigated accelerator driven system. It is assumed that the target is bombarded with 1.2383?1017 protons per second and that the energy of each proton is 1000 MeV. This means a proton beam power of 20 MW. The 3-D and time-dependent neutronic analyses are conducted by using the MCNPX 2.7 and CINDER 90 nuclear code. Both accelerator driven system designs are operated until the values of keff rise to 0.985 to determine the longest operation times that are the effective burn times in all cases. Depending on the design, minor actinide composition, and fuel type (spent UO2 and ThO2), the results obtained at the end of cycle exhibit the effective burn times vary from 300 days to 2050 days, the fuel enrichments can reach up to 2.49-4.23% and the values of gain reach up to 10.8-25.1.
{"title":"Neutronic analysis of an ads fuelled with minor actinide and designed for spent fuel enrichment and fissile fuel production","authors":"Busra Durmaz, Gizem Bakir, Bugra Arslan, H. Yapici","doi":"10.2298/ntrp2104299d","DOIUrl":"https://doi.org/10.2298/ntrp2104299d","url":null,"abstract":"This paper presents analyses of enrichments of uranium taken out from Canada Deuterium Uranium and pressurized water reactors spent fuels and fissile fuel breeding from thorium in two different helium cooled-accelerator driven system designs, DESIGN A and DESIGN B. In the beginning, the 235U percentages in the uranium fuels taken out from the reactors spent fuels are 0.17% and 0.91%, respectively. Both system cores are fuelled with two different minor actinides compositions extracted from PWR-MOX spent fuels. The DESIGN A has one transmutation zone (enrichment zone) surrounding the fuel core and containing thorium or spent uranium fuels, while DESIGN B has a second transmutation zone (fissile fuel breeding zone) surrounding the first transmutation zone and containing only thorium fuel. In brief, a total of ten cases formed by the combinations of accelerator driven system designs, minor actinides components, and spent uranium with thorium fuels are analysed, which are six in DESIGN A containing one transmutation zone and four in DESIGN B containing two transmutation zones. Lead-bismuth eutectic alloy, a liquid heavy metal, consisting of 45% lead and 55 % bismuth is used as target material in the investigated accelerator driven system. It is assumed that the target is bombarded with 1.2383?1017 protons per second and that the energy of each proton is 1000 MeV. This means a proton beam power of 20 MW. The 3-D and time-dependent neutronic analyses are conducted by using the MCNPX 2.7 and CINDER 90 nuclear code. Both accelerator driven system designs are operated until the values of keff rise to 0.985 to determine the longest operation times that are the effective burn times in all cases. Depending on the design, minor actinide composition, and fuel type (spent UO2 and ThO2), the results obtained at the end of cycle exhibit the effective burn times vary from 300 days to 2050 days, the fuel enrichments can reach up to 2.49-4.23% and the values of gain reach up to 10.8-25.1.","PeriodicalId":49734,"journal":{"name":"Nuclear Technology & Radiation Protection","volume":"1 1","pages":""},"PeriodicalIF":1.2,"publicationDate":"2021-01-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"68547298","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":4,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Vuk Gajić, Ivica Vujčić, G. Dražić, J. Milovanović, S. Mašić
The Kolubara river pollutes the coastal land in the river basin and makes it unsuitable for agricultural activities in that area. Also, contaminated land poses a risk to the environment. Different methods can be used for soil decontamination. These methods include biological treatment/bioremediation, chemical oxidation, soil stabilization, physical methods, such as soil leaching, or treatment with high-energy ionizing radiation. Gamma irradiation of soil is a well-known method of inhibiting microbial activity. This paper investigated the influence of different doses and dose rates of gamma irradiation on microorganisms' decontamination of coastal soil, in the Kolubara river basin. The irradiation effects on reducing the total number of microorganisms and removing mold and pathogenic bacteria from soil samples were examined. Gamma radiation affects the soil's organic matter, causing the formation of free reactive radicals, which act as reducing and oxidizing agents, cleaving C-C bonds, and depolymerizing carbohydrates. It was found that a dose of 3 kGy of gamma radiation, neutralizes all pathogenic bacteria, a dose of 5 kGy deactivates mold in soil samples, and a dose of 10 kGy is optimal to kill all microorganisms in the samples and sterilize exposed soil. The research showed that the dose rate does not significantly affect microbiological decontamination of soil using gamma irradiation. The content of heavy metals in soil was determined, and the obtained values were compared with the remediation limit values prescribed by the regulations. It was concluded that the content of heavy metals in the analyzed soil samples is below the limit of remediation values. The only exception is the slightly increased copper content in one sample. The result of this research is the conclusion that the coastal land from the Kolubara basin can be decontaminated by gamma radiation treatment. This advanced soil treatment technology is available in Serbia because there is an industrial plant for gamma radiation treatment within the Vinca Institute.
{"title":"Use of high-energy ionizing radiation for microbiological decontamination of coastal soil in the Kolubara river basin, Serbia","authors":"Vuk Gajić, Ivica Vujčić, G. Dražić, J. Milovanović, S. Mašić","doi":"10.2298/ntrp2103261g","DOIUrl":"https://doi.org/10.2298/ntrp2103261g","url":null,"abstract":"The Kolubara river pollutes the coastal land in the river basin and makes it unsuitable for agricultural activities in that area. Also, contaminated land poses a risk to the environment. Different methods can be used for soil decontamination. These methods include biological treatment/bioremediation, chemical oxidation, soil stabilization, physical methods, such as soil leaching, or treatment with high-energy ionizing radiation. Gamma irradiation of soil is a well-known method of inhibiting microbial activity. This paper investigated the influence of different doses and dose rates of gamma irradiation on microorganisms' decontamination of coastal soil, in the Kolubara river basin. The irradiation effects on reducing the total number of microorganisms and removing mold and pathogenic bacteria from soil samples were examined. Gamma radiation affects the soil's organic matter, causing the formation of free reactive radicals, which act as reducing and oxidizing agents, cleaving C-C bonds, and depolymerizing carbohydrates. It was found that a dose of 3 kGy of gamma radiation, neutralizes all pathogenic bacteria, a dose of 5 kGy deactivates mold in soil samples, and a dose of 10 kGy is optimal to kill all microorganisms in the samples and sterilize exposed soil. The research showed that the dose rate does not significantly affect microbiological decontamination of soil using gamma irradiation. The content of heavy metals in soil was determined, and the obtained values were compared with the remediation limit values prescribed by the regulations. It was concluded that the content of heavy metals in the analyzed soil samples is below the limit of remediation values. The only exception is the slightly increased copper content in one sample. The result of this research is the conclusion that the coastal land from the Kolubara basin can be decontaminated by gamma radiation treatment. This advanced soil treatment technology is available in Serbia because there is an industrial plant for gamma radiation treatment within the Vinca Institute.","PeriodicalId":49734,"journal":{"name":"Nuclear Technology & Radiation Protection","volume":"1 1","pages":""},"PeriodicalIF":1.2,"publicationDate":"2021-01-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"68547340","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":4,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Satisfactory discrimination between the neutron and gamma components in a mixed neutron-gamma field is one of the most important objectives of neutron dosimetry. One of the common techniques for estimating gamma and neutron dose components in mixed neutron-gamma fields is the two peak method. This method has been applied using dosimeters such as LiF:Mg,Ti, but in the present work, a 6LiF:Mg,Cu,P dosimeter has been used, whose thermoluminescence sensitivity is much higher than the LiF:Mg,Ti dosimeter, and therefore, if appropriate results are achieved, it can drastically reduce the dose estimation threshold. Applicability of 6LiF:Mg,Cu,P for estimation of the gamma dose using the two peak method in a mixed thermal neutron-gamma radiation field was studied. The ratio of the area underneath the high temperature thermoluminescence glow peak to dosimetry peak of this phosphor in an Am-Be neutron field is 0.127, while this ratio in a pure gamma ray field of 137Cs is 0.039. The calibration curves were obtained by separately irradiating 6LiF:Mg,Cu,P chips with known gamma and neutron doses. Results show that 6LiF:Mg,Cu,P can be used to estimate the contributions of neutron and gamma doses in a mixed neutron-gamma field by using the two peak method.
{"title":"Neutron-gamma mixed field dosimetry using a 6LiF:Mg,Cu,P thermoluminescent dosimeter","authors":"E. Sadeghi, M. Zahedifar, Parasto Rezaii","doi":"10.2298/ntrp2104346s","DOIUrl":"https://doi.org/10.2298/ntrp2104346s","url":null,"abstract":"Satisfactory discrimination between the neutron and gamma components in a mixed neutron-gamma field is one of the most important objectives of neutron dosimetry. One of the common techniques for estimating gamma and neutron dose components in mixed neutron-gamma fields is the two peak method. This method has been applied using dosimeters such as LiF:Mg,Ti, but in the present work, a 6LiF:Mg,Cu,P dosimeter has been used, whose thermoluminescence sensitivity is much higher than the LiF:Mg,Ti dosimeter, and therefore, if appropriate results are achieved, it can drastically reduce the dose estimation threshold. Applicability of 6LiF:Mg,Cu,P for estimation of the gamma dose using the two peak method in a mixed thermal neutron-gamma radiation field was studied. The ratio of the area underneath the high temperature thermoluminescence glow peak to dosimetry peak of this phosphor in an Am-Be neutron field is 0.127, while this ratio in a pure gamma ray field of 137Cs is 0.039. The calibration curves were obtained by separately irradiating 6LiF:Mg,Cu,P chips with known gamma and neutron doses. Results show that 6LiF:Mg,Cu,P can be used to estimate the contributions of neutron and gamma doses in a mixed neutron-gamma field by using the two peak method.","PeriodicalId":49734,"journal":{"name":"Nuclear Technology & Radiation Protection","volume":"1 1","pages":""},"PeriodicalIF":1.2,"publicationDate":"2021-01-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"68547437","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":4,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
The objective of this study was to analyze the dependence of the neutron dose from the geometry of the second band of the maze using dosimetric measurements of neutrons and Monte Carlo simulations, and based on those results to design a novel radiotherapy room layout. Measurements of the neutron dose at a two-band maze therapy room were performed for a 15 MeV photon beam only. Monte Carlo simulations were performed using the GEANT4 toolkit. In order to obtain the geometry dependence, we were changing the second band angle while we kept the length, height, and width the same as in reality. Results show that the highest calculated dose was obtained for the 60? angle of the second maze. It is 17 % higher than for standard 0? angle. For 30? it was 30 % smaller and for 90? was 10% smaller. Although the lowest dose was obtained for 30? band angle with calculations, it is not very practical for clinical use. Clinically the most interesting would be the 90? angle which is practically a short three-band maze, which could be promising from the perspective of neutron radiation protection since it could offer a compact constructional solution, and better optimization of the available space.
{"title":"Novel design of radiotherapy room suggestion - three-band maze","authors":"A. Toth, M. Marjanović, I. Gencel, B. Petrovic","doi":"10.2298/ntrp2104371t","DOIUrl":"https://doi.org/10.2298/ntrp2104371t","url":null,"abstract":"The objective of this study was to analyze the dependence of the neutron dose from the geometry of the second band of the maze using dosimetric measurements of neutrons and Monte Carlo simulations, and based on those results to design a novel radiotherapy room layout. Measurements of the neutron dose at a two-band maze therapy room were performed for a 15 MeV photon beam only. Monte Carlo simulations were performed using the GEANT4 toolkit. In order to obtain the geometry dependence, we were changing the second band angle while we kept the length, height, and width the same as in reality. Results show that the highest calculated dose was obtained for the 60? angle of the second maze. It is 17 % higher than for standard 0? angle. For 30? it was 30 % smaller and for 90? was 10% smaller. Although the lowest dose was obtained for 30? band angle with calculations, it is not very practical for clinical use. Clinically the most interesting would be the 90? angle which is practically a short three-band maze, which could be promising from the perspective of neutron radiation protection since it could offer a compact constructional solution, and better optimization of the available space.","PeriodicalId":49734,"journal":{"name":"Nuclear Technology & Radiation Protection","volume":"1 1","pages":""},"PeriodicalIF":1.2,"publicationDate":"2021-01-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"68547724","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":4,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
The capabilities of electret ion chambers to measure non-target absorbed dose for distances greater than 20 cm from the irradiated volume during radiotherapy treatment was investigated for the first time. During radiotherapy, nontarget doses can be classified as one of three approximate dose levels: high doses, intermediate doses and low doses. Low doses (<5 % of the prescription dose) are not generally considered during treatment planning, due to the fact that is difficult to measure, characterize, or model them in the planning system. In this work were performed measurements with electret ion chambers of absorbed dose outside the treated volume (<5 % of the prescription dose), during external photon radiation therapy in an Elekta Infinity Linear Accelerator of ?Theagenio? Cancer Hospital of Thessaloniki, Greece. The absorbed dose values for distances greater than 20 cm from the irradiated volume varied from 0.3 to 17 mGy which corresponds to 0.01% up to 0.6% of the prescription dose (2660 mGy). Near the irradiation volume the absorbed dose values were greater than the upper detection limit of the electret ion chambers (threshold 40 mGy). The results are compared with the calculated ones by the Monaco Treatment Planning System (Elekta Monaco 5.11.03). In the non-target radiation region where Monaco Treatment Planning System calculates rather precisely (within uncertainties of less than 10%) the absorbed dose, measured and calculated doses are the same within experimental uncertainties. On the contrary, when leakage radiation becomes the dominant source of out-of-field dose the differences are up to 31%.
{"title":"Measurement with electret ion chambers of absorbed dose outside the treated volume, during external-photon radiation therapy","authors":"A. Clouvas, A. Makridou, M. Chatzimarkou","doi":"10.2298/ntrp201113015c","DOIUrl":"https://doi.org/10.2298/ntrp201113015c","url":null,"abstract":"The capabilities of electret ion chambers to measure non-target absorbed dose for distances greater than 20 cm from the irradiated volume during radiotherapy treatment was investigated for the first time. During radiotherapy, nontarget doses can be classified as one of three approximate dose levels: high doses, intermediate doses and low doses. Low doses (<5 % of the prescription dose) are not generally considered during treatment planning, due to the fact that is difficult to measure, characterize, or model them in the planning system. In this work were performed measurements with electret ion chambers of absorbed dose outside the treated volume (<5 % of the prescription dose), during external photon radiation therapy in an Elekta Infinity Linear Accelerator of ?Theagenio? Cancer Hospital of Thessaloniki, Greece. The absorbed dose values for distances greater than 20 cm from the irradiated volume varied from 0.3 to 17 mGy which corresponds to 0.01% up to 0.6% of the prescription dose (2660 mGy). Near the irradiation volume the absorbed dose values were greater than the upper detection limit of the electret ion chambers (threshold 40 mGy). The results are compared with the calculated ones by the Monaco Treatment Planning System (Elekta Monaco 5.11.03). In the non-target radiation region where Monaco Treatment Planning System calculates rather precisely (within uncertainties of less than 10%) the absorbed dose, measured and calculated doses are the same within experimental uncertainties. On the contrary, when leakage radiation becomes the dominant source of out-of-field dose the differences are up to 31%.","PeriodicalId":49734,"journal":{"name":"Nuclear Technology & Radiation Protection","volume":"120 1","pages":""},"PeriodicalIF":1.2,"publicationDate":"2021-01-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"68545311","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":4,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
In order to enhance the defense in depth for nuclear safety after the Fukushima nuclear accident, U.S. Nuclear Energy Institute put forward the concept of Diverse and Flexible Coping Strategies and the corresponding FLEX support guidelines for the special scenarios of Extended Loss of Alternating current Power and Loss of Ultimate Heat Sink caused by Beyond-Design-Basis External Event. Subsequently, the idea of the FLEX strategy was widely accepted and spread widely. The introduction of the concept of FLEX strategy into the defense in depth was the biggest improvement for nuclear safety in the recent decade. This paper has reviewed the concept of traditional defense in depth and its weakness that led to the Fukushima nuclear accident, which led to the development motivation for the FLEX strategy. The research progress of the FLEX strategy in different countries in the past ten years has been reviewed. Based on the literature, and the above-mentioned review, some recommended future work has been given.
{"title":"A review on the defense-in-depth concept and the flex strategies in different countries after Fukushima accident","authors":"Hong Xu, Baorui Zhang","doi":"10.2298/ntrp210128013x","DOIUrl":"https://doi.org/10.2298/ntrp210128013x","url":null,"abstract":"In order to enhance the defense in depth for nuclear safety after the Fukushima nuclear accident, U.S. Nuclear Energy Institute put forward the concept of Diverse and Flexible Coping Strategies and the corresponding FLEX support guidelines for the special scenarios of Extended Loss of Alternating current Power and Loss of Ultimate Heat Sink caused by Beyond-Design-Basis External Event. Subsequently, the idea of the FLEX strategy was widely accepted and spread widely. The introduction of the concept of FLEX strategy into the defense in depth was the biggest improvement for nuclear safety in the recent decade. This paper has reviewed the concept of traditional defense in depth and its weakness that led to the Fukushima nuclear accident, which led to the development motivation for the FLEX strategy. The research progress of the FLEX strategy in different countries in the past ten years has been reviewed. Based on the literature, and the above-mentioned review, some recommended future work has been given.","PeriodicalId":49734,"journal":{"name":"Nuclear Technology & Radiation Protection","volume":"1 1","pages":""},"PeriodicalIF":1.2,"publicationDate":"2021-01-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"68546290","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":4,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Jihyun Ahn, Junsung Park, Hayoung Sim, Geunyoeng An, Hee Seo
To ensure the peaceful use of nuclear energy, nuclear safeguards are applied in member states of the International Atomic Energy Agency the Non-Proliferation Treaty. The two main goals of nuclear safeguards are effectiveness and efficiency. The International Atomic Energy Agency has a great interest in using a containment and surveillance technology to maintain continuity of knowledge. A representative means of a containment and surveillance technology is a sealing system to alert the user to tampering. The existing sealing systems used by the International Atomic Energy Agency are of limited utility for real-time verification purposes. To address this limitation, the present study analyzed the design requirements of a sealing system proposed by various institutions including the International Atomic Energy Agency, the U.S. Nuclear Regulatory Commission, a number of national laboratories, and companies. Then, we identified the appropriate design requirements of this system for real-time verification. The next step is to develop a real-time verification sealing system based on the design requirements identified and discussed herein. Such a system is expected to significantly enhance the efficiency of nuclear safeguards.
{"title":"Design requirements of safeguards sealing system for real-time verification","authors":"Jihyun Ahn, Junsung Park, Hayoung Sim, Geunyoeng An, Hee Seo","doi":"10.2298/ntrp2104376a","DOIUrl":"https://doi.org/10.2298/ntrp2104376a","url":null,"abstract":"To ensure the peaceful use of nuclear energy, nuclear safeguards are applied in member states of the International Atomic Energy Agency the Non-Proliferation Treaty. The two main goals of nuclear safeguards are effectiveness and efficiency. The International Atomic Energy Agency has a great interest in using a containment and surveillance technology to maintain continuity of knowledge. A representative means of a containment and surveillance technology is a sealing system to alert the user to tampering. The existing sealing systems used by the International Atomic Energy Agency are of limited utility for real-time verification purposes. To address this limitation, the present study analyzed the design requirements of a sealing system proposed by various institutions including the International Atomic Energy Agency, the U.S. Nuclear Regulatory Commission, a number of national laboratories, and companies. Then, we identified the appropriate design requirements of this system for real-time verification. The next step is to develop a real-time verification sealing system based on the design requirements identified and discussed herein. Such a system is expected to significantly enhance the efficiency of nuclear safeguards.","PeriodicalId":49734,"journal":{"name":"Nuclear Technology & Radiation Protection","volume":"1 1","pages":""},"PeriodicalIF":1.2,"publicationDate":"2021-01-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"68547745","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":4,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
This paper presents results of the development of a small-sized free release monitor designed for the release of materials, various hand tools, equipment and instruments of nuclear enterprises and laboratories staff that weight up to 50 kg, from radiation control. To increase the registration sensitivity of controlled radionuclides, 12 scintillation units based on a 3"x3" sized NaI (Tl) crystal were used as gamma-radiation detector. Volume of the measuring chamber of the monitor amounted to 200 liters, the thickness of the low-background shielding was chosen 50 mm. The values of the minimum detectable activity of the designed monitor for the point sources 123I, 131I, 99mTc, 18F were better than 100 Bq with measurement time not exceeding 60 s.
{"title":"Improvement of the minimum detectable activity of a free release monitor for small articles","authors":"S. Pohuliai, Igors Krainukovs","doi":"10.2298/ntrp210222018p","DOIUrl":"https://doi.org/10.2298/ntrp210222018p","url":null,"abstract":"This paper presents results of the development of a small-sized free release monitor designed for the release of materials, various hand tools, equipment and instruments of nuclear enterprises and laboratories staff that weight up to 50 kg, from radiation control. To increase the registration sensitivity of controlled radionuclides, 12 scintillation units based on a 3\"x3\" sized NaI (Tl) crystal were used as gamma-radiation detector. Volume of the measuring chamber of the monitor amounted to 200 liters, the thickness of the low-background shielding was chosen 50 mm. The values of the minimum detectable activity of the designed monitor for the point sources 123I, 131I, 99mTc, 18F were better than 100 Bq with measurement time not exceeding 60 s.","PeriodicalId":49734,"journal":{"name":"Nuclear Technology & Radiation Protection","volume":"1 1","pages":""},"PeriodicalIF":1.2,"publicationDate":"2021-01-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"68545942","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":4,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Sabahudin Hadrović, I. Čeliković, Jelena Krneta-Nikolic, M. Rajačić, D. Todorovic
Forests, with a large coverage of lands area, represent important ecosystem. They have greater ability to absorb atmospheric pollutant, including radionuclides compared to other vegetation types. Thus it is important to monitor radionuclides concentration in forest ecosystems. The results of the first gamma-spectrometric measurements in 16500 ha large region of South-western Serbia are presented. It is investigated activity concentrations of 40K, 137Cs and 210Pb in different deciduous and evergreen trees in the region. For all investigated isotopes, it was observed tendency that the smallest activity concentrations in average were found in tree stem, than in leaves, while the highest ones were in soil. Statistical analysis did not show any differences between activity concentrations of leaves and needles, showing that both leaves and needles could be equally well used as a biomonitors.
{"title":"Radionuclides’ content in forest ecosystem located in southwestern part of Serbia","authors":"Sabahudin Hadrović, I. Čeliković, Jelena Krneta-Nikolic, M. Rajačić, D. Todorovic","doi":"10.2298/ntrp210112014h","DOIUrl":"https://doi.org/10.2298/ntrp210112014h","url":null,"abstract":"Forests, with a large coverage of lands area, represent important ecosystem. They have greater ability to absorb atmospheric pollutant, including radionuclides compared to other vegetation types. Thus it is important to monitor radionuclides concentration in forest ecosystems. The results of the first gamma-spectrometric measurements in 16500 ha large region of South-western Serbia are presented. It is investigated activity concentrations of 40K, 137Cs and 210Pb in different deciduous and evergreen trees in the region. For all investigated isotopes, it was observed tendency that the smallest activity concentrations in average were found in tree stem, than in leaves, while the highest ones were in soil. Statistical analysis did not show any differences between activity concentrations of leaves and needles, showing that both leaves and needles could be equally well used as a biomonitors.","PeriodicalId":49734,"journal":{"name":"Nuclear Technology & Radiation Protection","volume":"1 1","pages":""},"PeriodicalIF":1.2,"publicationDate":"2021-01-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"68545469","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":4,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
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