Pyotr B. Baskov, G. V. Marichev, V. Sakharov, V. A. Stepanov
In the design of nuclear-optical converters (NOC) for detecting intense neutron fields (fluxes over 1015 cm–2·s–1), it is proposed to use hybrid gas ionization chambers (IC), in which electrical and optical neutron detecting methods are combined. For hybrid ICs, a technology is proposed for obtaining radiation-resistant and mechanically strong radiator materials capable of operating at temperatures of up to 1000 °C. This technology is based on solid-phase boron diffusion saturation of steel. It is shown that, at thermal neutron fluxes of 1×1010 n/(cm2·s) and higher, the integral intensity of argon luminescence as a result of ionization by α-particles and 7Li ions from layers of boride phases is sufficient for detection. The combination of optical and radiation properties of multicomponent fluoride glasses makes it possible to use them as condensed active substances of NOCs. Choosing the elemental and isotopic composition, it becomes possible to use fluoride glasses for multichannel neutron detection as well as to significantly simplify the procedure for separating gamma and neutron components of radiation under conditions of intense radiation fluxes. It has been experimentally shown that in irradiation with a neutron flux of 1×1017 n/(cm2·s), the intensity of Nd IR luminescence in glasses based on zirconium fluoride (ZBLAN) increases in the presence of Gd, which interacts with neutrons.
{"title":"Nuclear-optical converters for detecting intense neutron","authors":"Pyotr B. Baskov, G. V. Marichev, V. Sakharov, V. A. Stepanov","doi":"10.3897/nucet.8.82558","DOIUrl":"https://doi.org/10.3897/nucet.8.82558","url":null,"abstract":"In the design of nuclear-optical converters (NOC) for detecting intense neutron fields (fluxes over 1015 cm–2·s–1), it is proposed to use hybrid gas ionization chambers (IC), in which electrical and optical neutron detecting methods are combined. For hybrid ICs, a technology is proposed for obtaining radiation-resistant and mechanically strong radiator materials capable of operating at temperatures of up to 1000 °C. This technology is based on solid-phase boron diffusion saturation of steel. It is shown that, at thermal neutron fluxes of 1×1010 n/(cm2·s) and higher, the integral intensity of argon luminescence as a result of ionization by α-particles and 7Li ions from layers of boride phases is sufficient for detection.\u0000 The combination of optical and radiation properties of multicomponent fluoride glasses makes it possible to use them as condensed active substances of NOCs. Choosing the elemental and isotopic composition, it becomes possible to use fluoride glasses for multichannel neutron detection as well as to significantly simplify the procedure for separating gamma and neutron components of radiation under conditions of intense radiation fluxes. It has been experimentally shown that in irradiation with a neutron flux of 1×1017 n/(cm2·s), the intensity of Nd IR luminescence in glasses based on zirconium fluoride (ZBLAN) increases in the presence of Gd, which interacts with neutrons.","PeriodicalId":100969,"journal":{"name":"Nuclear Energy and Technology","volume":"53 1","pages":""},"PeriodicalIF":0.0,"publicationDate":"2022-03-17","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"91314360","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}
V. V. Fomichev, D. A. Pakholik, O. Kochnov, N. Kuznetsov, M. V. Kharitonov, Vyacheslav V. Nichugovsky
The demand for the use of radioactive isotopes in medicine is increasing with each coming year necessitating the increased output of radionuclide products. One of the most widely spread radionuclides used in medicine is technetium-99m (99mТс) (Feasibility of producing molybdenum-99 2015, NEA 2012, The Supply of Medical Radioisotopes 2015). The very short 99mТс life (6-hour half-life) requires its production directly on the site of medical treatment. This is achieved using molybdenum-technetium generators (Kodina and Krasikova 2014, Technical Reports No. NF-T-5.4. 2013, Technetium-99 Generator 2021) loaded with molybdenum-99 (99Мо), which uninterruptedly decays (half-life of 66 hours) yielding 99mTc. Close attention must be paid in the course of production of molybdenum-technetium generators to radiation safety during transportation of 99Мо on the territory of the manufacturing facility. The main measure for ensuring radiation safety during transportation of 99Мо is the application of special packaging kits. The Karpov Institute of Physical Chemistry JSC uses a wide range of packaging kits of types A and B for transportation of radioactive materials on the territory of the manufacturer with design features providing the required level of radiation safety. In particular, the KL-15 shipping cask loaded/unloaded from the top is used for onsite transportation of 99Мо for charging molybdenum-technetium generators. The maximum permissible activity of 99Мо is not specified in the passport of the KL-15 cask. Planned construction of a radionuclide production shop in accordance with GMP requirements will require the increase of output of target radionuclides by several times. The above considerations necessitated the evaluation of the maximum permissible activity of 99Мо planned to be transported in KL-15 casks. No other type of standard casks can be used because of their outside dimensions prohibiting the unloading of 99Мо inside the “hot” chamber. Calculation and experimental evaluation of permissible 99Мо activity during transportation inside the KL-15 cask was performed. The paper presents the calculated evaluation of the maximum permissible activity of 99Мо in a KL-15 cask to ensure the radiation exposure of personnel of group A working with the cask not exceeding the established level at the enterprise (80 μSv per shift) and not requiring the use of additional measures and means of protection. The results of the work allow us drawing the conclusion that the KL-15 cask ensures the required level of radiation safety with up to 241 Ki of 99Мо loaded in the cask.
在医学中使用放射性同位素的需求每年都在增加，因此必须增加放射性核素产品的产量。医学中使用最广泛的放射性核素之一是锝-99m (99mТс)(生产钼-99的可行性2015,NEA 2012，医用放射性同位素供应2015)。极短的99mТс寿命(6小时半衰期)要求其直接在医疗现场生产。这是使用钼-锝发生器实现的(Kodina和Krasikova 2014，技术报告号:nf - t - 5.4。2013年，锝-99发电机2021)装载钼-99 (99Мо)，它不间断地衰变(半衰期为66小时)，产生99mTc。钼-锝发生器在生产过程中，必须密切关注99Мо在生产设施境内运输过程中的辐射安全问题。确保99Мо在运输过程中的辐射安全的主要措施是特殊包装套件的应用。卡尔波夫物理化学研究所联合公司在制造商境内运输放射性物质时，广泛使用a类和B类包装包，其设计特点提供了所需的辐射安全水平。特别是，从顶部装卸的KL-15运输桶用于99Мо的现场运输，用于给钼-锝发生器充电。KL-15木桶的护照上没有规定99Мо的最大允许活动。按照GMP要求计划建设的放射性核素生产车间，需要将目标放射性核素的产量提高数倍。出于上述考虑，有必要对计划用KL-15桶运输的99Мо的最大允许活度进行评估。不能使用其他类型的标准桶，因为它们的外部尺寸禁止在“热”腔内卸载99Мо。对KL-15桶内运输过程中允许的99Мо活性进行了计算和实验评估。本文提出了KL-15木桶中99Мо的最大允许活性的计算评价，以确保a组工作人员与木桶的辐射暴露不超过企业规定的水平(每班80 μSv)，不需要使用额外的措施和保护手段。工作结果使我们能够得出结论，KL-15桶确保所需的辐射安全水平，桶内加载高达241 Ki 99Мо。
{"title":"Evaluation of the permissible 99Mo activity in the KL-15 cask in the design of transportation and process scheme","authors":"V. V. Fomichev, D. A. Pakholik, O. Kochnov, N. Kuznetsov, M. V. Kharitonov, Vyacheslav V. Nichugovsky","doi":"10.3897/nucet.8.82239","DOIUrl":"https://doi.org/10.3897/nucet.8.82239","url":null,"abstract":"The demand for the use of radioactive isotopes in medicine is increasing with each coming year necessitating the increased output of radionuclide products. One of the most widely spread radionuclides used in medicine is technetium-99m (99mТс) (Feasibility of producing molybdenum-99 2015, NEA 2012, The Supply of Medical Radioisotopes 2015). The very short 99mТс life (6-hour half-life) requires its production directly on the site of medical treatment. This is achieved using molybdenum-technetium generators (Kodina and Krasikova 2014, Technical Reports No. NF-T-5.4. 2013, Technetium-99 Generator 2021) loaded with molybdenum-99 (99Мо), which uninterruptedly decays (half-life of 66 hours) yielding 99mTc.\u0000 Close attention must be paid in the course of production of molybdenum-technetium generators to radiation safety during transportation of 99Мо on the territory of the manufacturing facility. The main measure for ensuring radiation safety during transportation of 99Мо is the application of special packaging kits. The Karpov Institute of Physical Chemistry JSC uses a wide range of packaging kits of types A and B for transportation of radioactive materials on the territory of the manufacturer with design features providing the required level of radiation safety.\u0000 In particular, the KL-15 shipping cask loaded/unloaded from the top is used for onsite transportation of 99Мо for charging molybdenum-technetium generators. The maximum permissible activity of 99Мо is not specified in the passport of the KL-15 cask. Planned construction of a radionuclide production shop in accordance with GMP requirements will require the increase of output of target radionuclides by several times. The above considerations necessitated the evaluation of the maximum permissible activity of 99Мо planned to be transported in KL-15 casks. No other type of standard casks can be used because of their outside dimensions prohibiting the unloading of 99Мо inside the “hot” chamber. Calculation and experimental evaluation of permissible 99Мо activity during transportation inside the KL-15 cask was performed.\u0000 The paper presents the calculated evaluation of the maximum permissible activity of 99Мо in a KL-15 cask to ensure the radiation exposure of personnel of group A working with the cask not exceeding the established level at the enterprise (80 μSv per shift) and not requiring the use of additional measures and means of protection.\u0000 The results of the work allow us drawing the conclusion that the KL-15 cask ensures the required level of radiation safety with up to 241 Ki of 99Мо loaded in the cask.","PeriodicalId":100969,"journal":{"name":"Nuclear Energy and Technology","volume":"1 1","pages":""},"PeriodicalIF":0.0,"publicationDate":"2022-03-17","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"89614548","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}
S. Fesenko, N. Sanzharova, Yevgeny I. Karpenko, N. Isamov, V. Kuznetsov, Aleksey V. Panov, P. N. Tsygvintsev
The article presents methodological approaches to the organization of radioecological monitoring in the regions where nuclear power plants are located. The analysis of the monitoring results at the Beloyarsk, Kursk, Leningrad and Rostov NPPs showed that the contribution of the natural radiation background to the public exposure dose is within a narrow range from 3.13 to 4.16 mSv per year, and the dose from the existing technogenic contamination varies from 0.47 μSv (Rostov NPP) up to 150 μSv per year (Beloyarsk NPP). The variability of the exposure doses is determined by the influence of natural climatic conditions and by differences in characteristics of contamination sources, including differences in electricity generation technologies. The technogenic radiation background in the area of the Beloyarsk NPP is determined by environmental contamination as a result of previous activities, whereas in the areas of the Leningrad NPP and the Kursk NPP it is associated with Chernobyl fallout (91 and 14 μSv per year, respectively). The contribution of NPPs to the existing technogenic radiation background varies from 1% (Rostov NPP) to 10–11% (Kursk and Beloyarsk NPPs).
{"title":"Radioecological monitoring and its role in ensuring the safety of nuclear power plants","authors":"S. Fesenko, N. Sanzharova, Yevgeny I. Karpenko, N. Isamov, V. Kuznetsov, Aleksey V. Panov, P. N. Tsygvintsev","doi":"10.3897/nucet.8.82619","DOIUrl":"https://doi.org/10.3897/nucet.8.82619","url":null,"abstract":"The article presents methodological approaches to the organization of radioecological monitoring in the regions where nuclear power plants are located. The analysis of the monitoring results at the Beloyarsk, Kursk, Leningrad and Rostov NPPs showed that the contribution of the natural radiation background to the public exposure dose is within a narrow range from 3.13 to 4.16 mSv per year, and the dose from the existing technogenic contamination varies from 0.47 μSv (Rostov NPP) up to 150 μSv per year (Beloyarsk NPP). The variability of the exposure doses is determined by the influence of natural climatic conditions and by differences in characteristics of contamination sources, including differences in electricity generation technologies. The technogenic radiation background in the area of the Beloyarsk NPP is determined by environmental contamination as a result of previous activities, whereas in the areas of the Leningrad NPP and the Kursk NPP it is associated with Chernobyl fallout (91 and 14 μSv per year, respectively). The contribution of NPPs to the existing technogenic radiation background varies from 1% (Rostov NPP) to 10–11% (Kursk and Beloyarsk NPPs).","PeriodicalId":100969,"journal":{"name":"Nuclear Energy and Technology","volume":"86 1","pages":""},"PeriodicalIF":0.0,"publicationDate":"2022-03-17","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"80127833","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}
G. V. Arkadov, V. Pavelko, V. Povarov, M. T. Slepov
The insufficiently studied issues of acoustic standing waves (ASW) in the main circulation circuits of the VVER reactor plants are considered. For a long time no proper attention has been given to this phenomenon both by the researchers and NPP experts. In general, generation of ASWs requires the acoustic inhomogeneities of the medium in the planes perpendicular to the direction of propagation of the longitudinal wave, in which a jump in acoustic resistance occurs, this is shown by the authors based on an example of the wave equation solution (D’Alembert equation) for a certain function of two variables. The ASW classification has been developed based on the obtained experimental material, 6 ASW types have been described, and their key parameters have been specified. The amplitude distributions have been plotted for all major ASW types proceeding from the phase relations of signals from the pressure pulsation detectors and accelerometers installed on the MCC pipelines. The nature of these distributions is general and they are valid for all VVER types. For the first time the globality of all lowest ASW types is identified. Four attribute properties of the ASWs have been formulated. The first attribute is the regular ASW temperature dependences, which is the source of the diagnostic information in the process of heating/cooling of the VVER unit. The linear experimental dependences of the ASW frequencies on coolant temperature have been obtained. The frequencies, at which the MCC resonant excitation due to coincidence of the ASW frequencies with the RCP rotational frequency harmonics, have been found experimentally. The ASW energy, which origin has resulted from the RCP operation, is estimated. The RCP operation can be presented as continuous generation of pressure pulsations, which fall onto the acoustic path inhomogeneities in the form of a traveling wave and generate a standing wave after reflection from them.
{"title":"Phenomenology of acoustic standing waves as applied to the VVER-1200 reactor plant","authors":"G. V. Arkadov, V. Pavelko, V. Povarov, M. T. Slepov","doi":"10.3897/nucet.8.82755","DOIUrl":"https://doi.org/10.3897/nucet.8.82755","url":null,"abstract":"The insufficiently studied issues of acoustic standing waves (ASW) in the main circulation circuits of the VVER reactor plants are considered. For a long time no proper attention has been given to this phenomenon both by the researchers and NPP experts. In general, generation of ASWs requires the acoustic inhomogeneities of the medium in the planes perpendicular to the direction of propagation of the longitudinal wave, in which a jump in acoustic resistance occurs, this is shown by the authors based on an example of the wave equation solution (D’Alembert equation) for a certain function of two variables. The ASW classification has been developed based on the obtained experimental material, 6 ASW types have been described, and their key parameters have been specified. The amplitude distributions have been plotted for all major ASW types proceeding from the phase relations of signals from the pressure pulsation detectors and accelerometers installed on the MCC pipelines. The nature of these distributions is general and they are valid for all VVER types. For the first time the globality of all lowest ASW types is identified. Four attribute properties of the ASWs have been formulated. The first attribute is the regular ASW temperature dependences, which is the source of the diagnostic information in the process of heating/cooling of the VVER unit. The linear experimental dependences of the ASW frequencies on coolant temperature have been obtained. The frequencies, at which the MCC resonant excitation due to coincidence of the ASW frequencies with the RCP rotational frequency harmonics, have been found experimentally. The ASW energy, which origin has resulted from the RCP operation, is estimated. The RCP operation can be presented as continuous generation of pressure pulsations, which fall onto the acoustic path inhomogeneities in the form of a traveling wave and generate a standing wave after reflection from them.","PeriodicalId":100969,"journal":{"name":"Nuclear Energy and Technology","volume":"34 1","pages":""},"PeriodicalIF":0.0,"publicationDate":"2022-03-17","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"84305326","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}
M. Hossain, Y. Akter, Mehraz Zaman Fardin, A. S. Mollah
A handful of computational benchmarks that incorporate VVER-1000 assemblies having low enriched uranium (LEU) and the mixed oxide (MOX) fuel have been put forward by many experts across the world from the Nuclear Energy Agency. To study & scrutinize the characteristics of one of the VVER-1000 LEU & MOX assembly benchmarks in different states were considered. In this work, the VVER-1000 LEU and MOX Assembly computational-benchmark exercises are performed using the OpenMC software. The work was intended to test the preciseness of the OpenMC Monte Carlo code using nuclear data library ENDF/B-VII.1, against a handful of previously obtained solutions with other computer codes. The kinf value obtained was compared with the SERPENT and MCNP result, which presented a very good similarity with very few deviations. The kinf variation with respect to burnup upto 40 MWd/kgHM was obtained for State-5 by using OpenMC code for both the LEU and MOX fuel assembly. The depletion curves of isotope concentrations against burnup upto 40 MWd/kg/HM were also generated for both the LEU and MOX fuel assembly. The OpenMC results are comparable with those of benchmark mean values. The neutron energy vs flux spectrum was also generated by using OpenMC code. Based on the OpenMC results such as kinf, burnup, isotope concentrations and neutron energy spectrum, it is concluded that the OPenMC code with ENDF/B-VII.1 nuclear data library was successfully implemented. It is planned to use OpenMC code for calculation of neutronics and burnup of the VVER-1200 reactor to be commissioned in Bangladesh by 2023/2024.
{"title":"Neutronics and burnup analysis of VVER-1000 LEU and MOX assembly computational benchmark using OpenMC Code","authors":"M. Hossain, Y. Akter, Mehraz Zaman Fardin, A. S. Mollah","doi":"10.3897/nucet.8.78447","DOIUrl":"https://doi.org/10.3897/nucet.8.78447","url":null,"abstract":"A handful of computational benchmarks that incorporate VVER-1000 assemblies having low enriched uranium (LEU) and the mixed oxide (MOX) fuel have been put forward by many experts across the world from the Nuclear Energy Agency. To study & scrutinize the characteristics of one of the VVER-1000 LEU & MOX assembly benchmarks in different states were considered. In this work, the VVER-1000 LEU and MOX Assembly computational-benchmark exercises are performed using the OpenMC software. The work was intended to test the preciseness of the OpenMC Monte Carlo code using nuclear data library ENDF/B-VII.1, against a handful of previously obtained solutions with other computer codes. The kinf value obtained was compared with the SERPENT and MCNP result, which presented a very good similarity with very few deviations. The kinf variation with respect to burnup upto 40 MWd/kgHM was obtained for State-5 by using OpenMC code for both the LEU and MOX fuel assembly. The depletion curves of isotope concentrations against burnup upto 40 MWd/kg/HM were also generated for both the LEU and MOX fuel assembly. The OpenMC results are comparable with those of benchmark mean values. The neutron energy vs flux spectrum was also generated by using OpenMC code. Based on the OpenMC results such as kinf, burnup, isotope concentrations and neutron energy spectrum, it is concluded that the OPenMC code with ENDF/B-VII.1 nuclear data library was successfully implemented. It is planned to use OpenMC code for calculation of neutronics and burnup of the VVER-1200 reactor to be commissioned in Bangladesh by 2023/2024.","PeriodicalId":100969,"journal":{"name":"Nuclear Energy and Technology","volume":"185 1","pages":""},"PeriodicalIF":0.0,"publicationDate":"2022-03-14","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"80565797","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}
A procedure has been developed to determine the geometrical parameters of fuel assemblies (FA) by an ultrasonic pulse-echo technique used for all types of light-water reactor FAs. The measurement of geometrical parameters is achieved through the pairwise installation of ultrasonic transducers opposite the FA spacer grid faces at a distance of not more than a half of the transducer acoustic field near-region length such that the acoustic axes of the pairwise transducers are parallel to each other. The advantages of the presented technique is that it enables monitoring of any FA modifications, including the VVER reactor assemblies with a different number of spacer grids. The paper presents a mathematical model of the acoustic path developed in a geometrical acoustics approximation and its verification results. The model was used for computational and experimental studies of the ultrasonic test technique, and engineering formulas have been developed to calculate the errors of the transducer-measured distance to the FA surface. A code has been developed to simulate the FA form change monitoring and can be used to design new monitoring systems. The developed technique to determine the VVER-1000 FA geometrical parameters was introduced at units 1 and 2 of the Temelin NPP, the Czech Republic, for the TVSA-T FA form change monitoring. The successful use of the proposed technique makes it possible to recommend it for use in inspection benches at other NPPs.
{"title":"Ultrasonic monitoring of the VVER-1000 FA form change","authors":"A. Voronina, S. Pavlov, Sergey V. Amosov","doi":"10.3897/nucet.8.89350","DOIUrl":"https://doi.org/10.3897/nucet.8.89350","url":null,"abstract":"A procedure has been developed to determine the geometrical parameters of fuel assemblies (FA) by an ultrasonic pulse-echo technique used for all types of light-water reactor FAs. The measurement of geometrical parameters is achieved through the pairwise installation of ultrasonic transducers opposite the FA spacer grid faces at a distance of not more than a half of the transducer acoustic field near-region length such that the acoustic axes of the pairwise transducers are parallel to each other. The advantages of the presented technique is that it enables monitoring of any FA modifications, including the VVER reactor assemblies with a different number of spacer grids.\u0000 The paper presents a mathematical model of the acoustic path developed in a geometrical acoustics approximation and its verification results. The model was used for computational and experimental studies of the ultrasonic test technique, and engineering formulas have been developed to calculate the errors of the transducer-measured distance to the FA surface. A code has been developed to simulate the FA form change monitoring and can be used to design new monitoring systems.\u0000 The developed technique to determine the VVER-1000 FA geometrical parameters was introduced at units 1 and 2 of the Temelin NPP, the Czech Republic, for the TVSA-T FA form change monitoring. The successful use of the proposed technique makes it possible to recommend it for use in inspection benches at other NPPs.","PeriodicalId":100969,"journal":{"name":"Nuclear Energy and Technology","volume":"34 1","pages":""},"PeriodicalIF":0.0,"publicationDate":"2022-01-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"82631113","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}
Aleksander S. Zevyakin, V. Kolesov, A. Sobolev, O. Kochnov
Thermal-hydraulic calculations have been conducted with respect to the active part of the MAK-2 loop facility of the VVR-ts research reactor for the 99Mo production. The computational studies were undertaken both for the case of using a highly 235U enriched target and for a low-enriched target. The calculation was performed for the actual technical characteristics of the research channel. The power density for the two simulated cases was obtained in the course of a preliminary neutronic calculation and selected for the most heated channel. The problem is solved for the steady-state mode of the channel coolant flow and takes into account the dependence of the thermophysical parameters of materials on temperature. The volumetric temperature distribution in the channel was obtained in the process of the calculation. The calculation results present the maximum temperatures of the target materials for the 99Mo production. An analysis of the obtained results has shown that the maximum temperatures of the aluminum sleeve and the target filling materials do not exceed the critical values. For the analyzed calculation cases, the maximum coolant temperature is localized at a point near the sleeve wall surface and does not reach the boiling temperature for a given pressure. The study has therefore shown that it is possible to reduce the 235U enrichment of the target filling fissile material to 19.7%, provided the average density of the mixture and the amount of 235U in the target remain the same. At the same time, the amount of the medicinally important 99Mo generated will not practically change, which will lead to reduced capital costs for a highly enriched mixture of the target matrix.
{"title":"Possibility for using a low-enriched target to produce 99Mo in the MAK-2 research channel of the VVR-ts reactor","authors":"Aleksander S. Zevyakin, V. Kolesov, A. Sobolev, O. Kochnov","doi":"10.3897/nucet.8.89351","DOIUrl":"https://doi.org/10.3897/nucet.8.89351","url":null,"abstract":"Thermal-hydraulic calculations have been conducted with respect to the active part of the MAK-2 loop facility of the VVR-ts research reactor for the 99Mo production. The computational studies were undertaken both for the case of using a highly 235U enriched target and for a low-enriched target. The calculation was performed for the actual technical characteristics of the research channel. The power density for the two simulated cases was obtained in the course of a preliminary neutronic calculation and selected for the most heated channel.\u0000 The problem is solved for the steady-state mode of the channel coolant flow and takes into account the dependence of the thermophysical parameters of materials on temperature. The volumetric temperature distribution in the channel was obtained in the process of the calculation.\u0000 The calculation results present the maximum temperatures of the target materials for the 99Mo production. An analysis of the obtained results has shown that the maximum temperatures of the aluminum sleeve and the target filling materials do not exceed the critical values. For the analyzed calculation cases, the maximum coolant temperature is localized at a point near the sleeve wall surface and does not reach the boiling temperature for a given pressure. The study has therefore shown that it is possible to reduce the 235U enrichment of the target filling fissile material to 19.7%, provided the average density of the mixture and the amount of 235U in the target remain the same. At the same time, the amount of the medicinally important 99Mo generated will not practically change, which will lead to reduced capital costs for a highly enriched mixture of the target matrix.","PeriodicalId":100969,"journal":{"name":"Nuclear Energy and Technology","volume":"29 1","pages":""},"PeriodicalIF":0.0,"publicationDate":"2022-01-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"78142671","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}
The objectives of the article are (1) to show the nuclear and physical causes of hard γ-quanta in the U-232 decay chain, (2) to propose tactics for handling uranium containing U-232, and (3) to assess the efficiency of its protective γ-barrier against uncontrolled proliferation. The authors show the general picture of the decay chains of U-232 nuclide transformations, on which the protection of uranium from its uncontrolled proliferation is based. During the decay of nuclei, their emission of α- or β-particles is only the first stage of the most complex process of rearrangement of both the internal structure of the nucleus itself, which consists in the rearrangement of the neutron and proton shells and the levels of its excitation, and in the rearrangement of the electron shells of the atom. As a rule, the daughter nucleus is in a highly excited state, which is removed by the emission of hard γ-quanta and internal conversion electrons. After the second case, the remaining excitation of the atom is removed by the emission of characteristic γ-quanta and Auger-electrons with characteristic γ-quanta. In addition, explanations are given for the quantum-mechanical reasons for the hard γ-radiation of Tl-208 and Bi-212, which complete the U-232 decay chain. The authors also proposed a tactic for handling uranium containing uranium-232. Since the hard γ-quanta of Tl-208 and Bi-212 appear only at the end of the U-232 decay chain, after its chemical purification from its decay products, U-232 itself does not pose a radiation hazard; therefore, at this time it is advisable to conduct all necessary operations for transporting the material to the plant, fabricating uranium-based fuel containing U-232, and transporting this fuel to the nuclear facility where it will be used.
{"title":"Proliferation protection of uranium due to the presence of U-232 decay products as intense sources of hard gamma radiation","authors":"G. Kulikov, A. Shmelev, V. Apse, E. Kulikov","doi":"10.3897/nucet.8.87814","DOIUrl":"https://doi.org/10.3897/nucet.8.87814","url":null,"abstract":"The objectives of the article are (1) to show the nuclear and physical causes of hard γ-quanta in the U-232 decay chain, (2) to propose tactics for handling uranium containing U-232, and (3) to assess the efficiency of its protective γ-barrier against uncontrolled proliferation.\u0000 The authors show the general picture of the decay chains of U-232 nuclide transformations, on which the protection of uranium from its uncontrolled proliferation is based. During the decay of nuclei, their emission of α- or β-particles is only the first stage of the most complex process of rearrangement of both the internal structure of the nucleus itself, which consists in the rearrangement of the neutron and proton shells and the levels of its excitation, and in the rearrangement of the electron shells of the atom. As a rule, the daughter nucleus is in a highly excited state, which is removed by the emission of hard γ-quanta and internal conversion electrons. After the second case, the remaining excitation of the atom is removed by the emission of characteristic γ-quanta and Auger-electrons with characteristic γ-quanta.\u0000 In addition, explanations are given for the quantum-mechanical reasons for the hard γ-radiation of Tl-208 and Bi-212, which complete the U-232 decay chain.\u0000 The authors also proposed a tactic for handling uranium containing uranium-232. Since the hard γ-quanta of Tl-208 and Bi-212 appear only at the end of the U-232 decay chain, after its chemical purification from its decay products, U-232 itself does not pose a radiation hazard; therefore, at this time it is advisable to conduct all necessary operations for transporting the material to the plant, fabricating uranium-based fuel containing U-232, and transporting this fuel to the nuclear facility where it will be used.","PeriodicalId":100969,"journal":{"name":"Nuclear Energy and Technology","volume":"12 1","pages":""},"PeriodicalIF":0.0,"publicationDate":"2022-01-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"88195181","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}
To timely detect failed fuel elements, a reactor plant should be equipped with a fuel cladding tightness monitoring system (FCTMS). In reactors using a heavy liquid-metal coolant (HLMC), the most efficient way to monitor the fuel cladding tightness is by detecting gaseous fission products (GFP). The article describes the basic principles of constructing a FCTMS in liquid-metal-cooled reactors based on the detection of fission products and delayed neutrons. It is noted that in a reactor plant using a HLMC the fuel cladding tightness is the most efficiently monitored by detecting GFPs. The authors analyze various aspects of the behavior of fission products in a liquid-metal-cooled reactor, such as the movement of GFPs in dissolved and bubble form along the circuit, the sorption of volatile FPs in the lead coolant (LC) and on the surfaces of structural elements, degassing of the GFPs dissolved in the LC, and filtration of cover gas from aerosol particles of different nature. In addition, a general description is given of the conditions for the transfer of GFPs in a LC environment of the reactor being developed. Finally, a mathematical model is presented that makes it possible to determine the calculated activity of reference radionuclides in each reactor unit at any time after the fuel element tightness failure. Based on this model, methods for monitoring the fuel cladding tightness by the gas activity in the gas volumes of the reactor plant will be proposed.
{"title":"Features of methods for monitoring the fuel cladding tightness in lead-cooled fast breeder reactors","authors":"A. V. Dragunova, M. Morkin, V. Perevezentsev","doi":"10.3897/nucet.7.78372","DOIUrl":"https://doi.org/10.3897/nucet.7.78372","url":null,"abstract":"To timely detect failed fuel elements, a reactor plant should be equipped with a fuel cladding tightness monitoring system (FCTMS). In reactors using a heavy liquid-metal coolant (HLMC), the most efficient way to monitor the fuel cladding tightness is by detecting gaseous fission products (GFP).\u0000 The article describes the basic principles of constructing a FCTMS in liquid-metal-cooled reactors based on the detection of fission products and delayed neutrons. It is noted that in a reactor plant using a HLMC the fuel cladding tightness is the most efficiently monitored by detecting GFPs. The authors analyze various aspects of the behavior of fission products in a liquid-metal-cooled reactor, such as the movement of GFPs in dissolved and bubble form along the circuit, the sorption of volatile FPs in the lead coolant (LC) and on the surfaces of structural elements, degassing of the GFPs dissolved in the LC, and filtration of cover gas from aerosol particles of different nature.\u0000 In addition, a general description is given of the conditions for the transfer of GFPs in a LC environment of the reactor being developed. Finally, a mathematical model is presented that makes it possible to determine the calculated activity of reference radionuclides in each reactor unit at any time after the fuel element tightness failure. Based on this model, methods for monitoring the fuel cladding tightness by the gas activity in the gas volumes of the reactor plant will be proposed.","PeriodicalId":100969,"journal":{"name":"Nuclear Energy and Technology","volume":"8 1","pages":""},"PeriodicalIF":0.0,"publicationDate":"2021-12-17","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"89212468","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}
The paper presents a review of modern studies on the potential types of coolant flow instabilities in the supercritical water reactor core. These instabilities have a negative impact on the operational safety of nuclear power plants. Despite the impressive number of computational works devoted to this topic, there still remain unresolved problems. The main disadvantages of the models are associated with the use of one simulated channel instead of a system of two or more parallel channels, the lack consideration for neutronic feedbacks, and the problem of choosing the design ratios for the heat transfer coefficient and hydraulic resistance coefficient under conditions of supercritical water flow. For this reason, it was decided to conduct an analysis that will make it possible to highlight the indicated problems and, on their basis, to formulate general requirements for a model of a nuclear reactor with a light-water supercritical pressure coolant. Consideration is also given to the features of the coolant flow stability in the supercritical water reactor core. In conclusion, the authors note the importance of further computational work using complex models of neutronic thermal-hydraulic stability built on the basis of modern achievements in the field of neutron physics and thermal physics.
{"title":"Analysis of numerical studies on the thermal-hydraulic and neutronic thermal-hydraulic stability of supercritical water reactors","authors":"Artavazd M. Sujyan, V. I. Deev, V. Kharitonov","doi":"10.3897/nucet.7.78368","DOIUrl":"https://doi.org/10.3897/nucet.7.78368","url":null,"abstract":"The paper presents a review of modern studies on the potential types of coolant flow instabilities in the supercritical water reactor core. These instabilities have a negative impact on the operational safety of nuclear power plants. Despite the impressive number of computational works devoted to this topic, there still remain unresolved problems. The main disadvantages of the models are associated with the use of one simulated channel instead of a system of two or more parallel channels, the lack consideration for neutronic feedbacks, and the problem of choosing the design ratios for the heat transfer coefficient and hydraulic resistance coefficient under conditions of supercritical water flow. For this reason, it was decided to conduct an analysis that will make it possible to highlight the indicated problems and, on their basis, to formulate general requirements for a model of a nuclear reactor with a light-water supercritical pressure coolant. Consideration is also given to the features of the coolant flow stability in the supercritical water reactor core. In conclusion, the authors note the importance of further computational work using complex models of neutronic thermal-hydraulic stability built on the basis of modern achievements in the field of neutron physics and thermal physics.","PeriodicalId":100969,"journal":{"name":"Nuclear Energy and Technology","volume":"14 1","pages":""},"PeriodicalIF":0.0,"publicationDate":"2021-12-17","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"78449668","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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