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Time to radiological equivalence of radioactive waste and natural uranium feedstock at an increasing content of Np, Am, and Cm in long-lived radioactive waste 长寿命放射性废物中Np、Am和Cm含量增加时，放射性废物和天然铀原料的放射性当量的时间
IF 0.4 4区 工程技术
Atomic Energy Pub Date : 2025-01-28 DOI: 10.1007/s10512-024-01161-0
V. K. Ivanov, A. V. Lopatkin, E. O. Adamov, E. V. Spirin, V. M. Solomatin
{"title":"Time to radiological equivalence of radioactive waste and natural uranium feedstock at an increasing content of Np, Am, and Cm in long-lived radioactive waste","authors":"V. K. Ivanov,&nbsp;A. V. Lopatkin,&nbsp;E. O. Adamov,&nbsp;E. V. Spirin,&nbsp;V. M. Solomatin","doi":"10.1007/s10512-024-01161-0","DOIUrl":"10.1007/s10512-024-01161-0","url":null,"abstract":"<div><p>This paper assesses the time to radiation (radiotoxicity) and radiological (radiation risk) equivalence between natural uranium and radioactive waste of thermal and fast reactors at an increasing content of Np, Am, and Cm in radioactive waste. We calculate radiation risk by adapting models of the ICRP and other international organizations for the Russian Federation, taking into account background epidemiological indicators including cancer incidence, cancer mortality, and overall mortality. The radiation risk of internal exposure considers both the time after radionuclide intake and equivalent dose dynamics in human organs and tissues. To determine the time to radiological equivalence, we estimate the lifetime attributable risk of a single intake of natural uranium radionuclides and radioactive waste. An increase in the Np, Am, and Cm content of radioactive waste within 0.1–0.4% causes no effect on the radiological equivalence time, equal to ~100 years. However, a further increase from 0.5–0.8% prolongs the time to 300 years. A problem for optimizing the time to radiological equivalence is posed providing for waste storage costs and content of Np, Am, and Cm in long-lived radioactive waste.</p></div>","PeriodicalId":480,"journal":{"name":"Atomic Energy","volume":"136 5-6","pages":"280 - 287"},"PeriodicalIF":0.4,"publicationDate":"2025-01-28","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143571160","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}
引用次数: 0
Leakage control of HTGR fuel using a weak irradiation technique 利用弱辐照技术控制HTGR燃料泄漏
IF 0.4 4区 工程技术
Atomic Energy Pub Date : 2025-01-28 DOI: 10.1007/s10512-024-01167-8
M. A. Agulnik, A. V. Grol, V. V. Degtyarev, E. S. Kondratieva, P. A. Fomichenko, A. A. Reshetnikov, O. I. Fedin, L. G. Chumak, A. V. Beleevskiy, I. E. Golubev, A. V. Davydov
{"title":"Leakage control of HTGR fuel using a weak irradiation technique","authors":"M. A. Agulnik,&nbsp;A. V. Grol,&nbsp;V. V. Degtyarev,&nbsp;E. S. Kondratieva,&nbsp;P. A. Fomichenko,&nbsp;A. A. Reshetnikov,&nbsp;O. I. Fedin,&nbsp;L. G. Chumak,&nbsp;A. V. Beleevskiy,&nbsp;I. E. Golubev,&nbsp;A. V. Davydov","doi":"10.1007/s10512-024-01167-8","DOIUrl":"10.1007/s10512-024-01167-8","url":null,"abstract":"<div><p>Radiation safety of high temperature gas-cooled reactors (HTGRs) largely depends on the leakage of fission products from the fuel under both normal and emergency modes of operation. Within the development of a pilot industrial fuel technology for an HTGR of the nuclear power engineering plant, we conducted pre-reactor tests of laboratory manufactured fuel to determine the leakage of fission products, as well as to assess the contamination of structural materials with uranium. Bulk TRISO particles, TRISO particle compacts, and blank graphite compacts were studied. The conducted tests used a weak irradiation technique developed by the NRC “Kurchatov Institute”. The technique is based on measuring the leakage of gaseous fission products from weakly irradiated samples during their annealing at temperatures close to the core temperature. A small dose of radiation causes no effect on the basic initial physical and chemical properties of fuel (density, porosity, etc.), as well as the ability to retain fission products. We analyzed the quality of laboratory-manufactured HTGR fuel by comparing the measured relative leakage of reference radionuclides with the established permissible limits, including individual technological stages of production. The study highlights the importance of the technique for quality control of HTGR fuel.</p></div>","PeriodicalId":480,"journal":{"name":"Atomic Energy","volume":"136 5-6","pages":"316 - 325"},"PeriodicalIF":0.4,"publicationDate":"2025-01-28","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143571161","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}
引用次数: 0
Simulation of accidental temperature and humidity in the pipeline rooms of the VVER NPP primary circuit VVER核电站一次回路管路室意外温湿度模拟
IF 0.4 4区 工程技术
Atomic Energy Pub Date : 2025-01-28 DOI: 10.1007/s10512-024-01168-7
E. L. Matveev, M. S. Cherkasova, A. V. Tutukin
{"title":"Simulation of accidental temperature and humidity in the pipeline rooms of the VVER NPP primary circuit","authors":"E. L. Matveev,&nbsp;M. S. Cherkasova,&nbsp;A. V. Tutukin","doi":"10.1007/s10512-024-01168-7","DOIUrl":"10.1007/s10512-024-01168-7","url":null,"abstract":"<div><p>The paper considers temperature and humidity simulation of the primary circuit pipeline rooms located at power units No. 3 and 4 of the Kola NPP upon the occurrence of equipment leakage. Simulation results were used to describe the dependence of the leakage on the temperature and relative humidity recorded by the sensors. Thus, advanced methods of automated coolant leak detection were developed; existing methods were adjusted.</p></div>","PeriodicalId":480,"journal":{"name":"Atomic Energy","volume":"136 5-6","pages":"326 - 331"},"PeriodicalIF":0.4,"publicationDate":"2025-01-28","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143571162","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}
引用次数: 0
Optimal rotor length of a gas centrifuge 气体离心机转子的最佳长度
IF 0.4 4区 工程技术
Atomic Energy Pub Date : 2025-01-27 DOI: 10.1007/s10512-024-01166-9
O. E. Aleksandrov
{"title":"Optimal rotor length of a gas centrifuge","authors":"O. E. Aleksandrov","doi":"10.1007/s10512-024-01166-9","DOIUrl":"10.1007/s10512-024-01166-9","url":null,"abstract":"<div><p>The paper provides an analytical expression to determine the optimal rotor length of a gas centrifuge. The optimum is defined as the maximum separation power per unit volume occupied by the centrifuge. Long centrifuge rotors, typical for European centrifuges, appear economically inappropriate.</p></div>","PeriodicalId":480,"journal":{"name":"Atomic Energy","volume":"136 5-6","pages":"311 - 315"},"PeriodicalIF":0.4,"publicationDate":"2025-01-27","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143571072","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}
引用次数: 0
Justification of the CONV-3D code for simulating natural circulation of a lead-bismuth coolant in a hydraulic circuit 模拟铅铋冷却剂在液压回路中自然循环的convd - 3d代码的论证
IF 0.4 4区 工程技术
Atomic Energy Pub Date : 2025-01-27 DOI: 10.1007/s10512-024-01170-z
V. V. Chudanov, A. E. Aksenova, V. A. Pervichko
{"title":"Justification of the CONV-3D code for simulating natural circulation of a lead-bismuth coolant in a hydraulic circuit","authors":"V. V. Chudanov,&nbsp;A. E. Aksenova,&nbsp;V. A. Pervichko","doi":"10.1007/s10512-024-01170-z","DOIUrl":"10.1007/s10512-024-01170-z","url":null,"abstract":"<div><p>The paper presents the results of simulating circulation of a lead-bismuth coolant in a hydraulic circuit using the CONV‑3<i>D</i> code. The temperature head and flow rate of the coolant are simulated for various heating power values. The developed code satisfactorily simulates the natural convective flow of a lead-bismuth coolant in a hydraulic circuit, thus being appropriate for analyzing these flow types in various applications.</p></div>","PeriodicalId":480,"journal":{"name":"Atomic Energy","volume":"136 5-6","pages":"338 - 341"},"PeriodicalIF":0.4,"publicationDate":"2025-01-27","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143571071","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}
引用次数: 0
Study of alkaline coolant boiling in the fuel assemblies of fast reactors in emergency modes with natural convection 自然对流应急模式下快堆燃料组件中碱性冷却剂沸腾的研究
IF 0.4 4区 工程技术
Atomic Energy Pub Date : 2025-01-27 DOI: 10.1007/s10512-024-01159-8
A. P. Sorokin, N. A. Denisova, Yu. A. Kuzina, G. A. Sorokin
{"title":"Study of alkaline coolant boiling in the fuel assemblies of fast reactors in emergency modes with natural convection","authors":"A. P. Sorokin,&nbsp;N. A. Denisova,&nbsp;Yu. A. Kuzina,&nbsp;G. A. Sorokin","doi":"10.1007/s10512-024-01159-8","DOIUrl":"10.1007/s10512-024-01159-8","url":null,"abstract":"<div><p>The article presents the results of experiments conducted by the IPPE JSC on heat exchange and circulation stability during boiling of alkaline liquid metals in the models of single and parallel fuel assemblies (FAs) under natural circulation conditions. It is shown that the complex structure of liquid metal boiling in FAs is formed under the influence of various factors and is characterized by both stable and oscillatory modes, as well as possible transition to departure from nucleate boiling. The influence of the fuel rod surface roughness on the heat exchange and flow modes of a two-phase liquid metal flow in bundles is shown. The presented results of experiments on the heat exchange during sodium boiling in an FA model with a sodium cavity above the reactor core demonstrate the possibility of long-term cooling of fuel rod simulators in FAs. The dependence on the heat transfer during boiling and a map of the modes for a two-phase liquid metal flow in the FA are presented. The agreement between the results of computational simulation and experiments on alkaline metal boiling in a single model FA and a system of parallel FAs is shown.</p></div>","PeriodicalId":480,"journal":{"name":"Atomic Energy","volume":"136 5-6","pages":"259 - 271"},"PeriodicalIF":0.4,"publicationDate":"2025-01-27","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143571014","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}
引用次数: 0
Hydrogen production at a nuclear power engineering plant with a high-temperature gas-cooled reactor (HTGR NPEP) 采用高温气冷堆（HTGR NPEP）的核电站制氢
IF 0.4 4区 工程技术
Atomic Energy Pub Date : 2025-01-23 DOI: 10.1007/s10512-024-01157-w
I. V. Marov, G. N. Kodochigov, D. S. Birin
{"title":"Hydrogen production at a nuclear power engineering plant with a high-temperature gas-cooled reactor (HTGR NPEP)","authors":"I. V. Marov,&nbsp;G. N. Kodochigov,&nbsp;D. S. Birin","doi":"10.1007/s10512-024-01157-w","DOIUrl":"10.1007/s10512-024-01157-w","url":null,"abstract":"<div><p>In order to structurally diversify the energy sector and transit to carbon-free development, R&amp;D of nuclear hydrogen technologies for the large-scale production and consumption of hydrogen is being carried out including a project of a nuclear power engineering plant with a high-temperature gas-cooled reactor (HTGR NPEP) for obtaining hydrogen and hydrogen-containing products. Based on domestic and foreign experience, steam reforming of methane was selected as the method of the NPEP hydrogen production. A description of the NPEP and its technical characteristics is provided. The main technological processes implemented in the chemical engineering plant are described. A proposal for licensing a NPEP project providing for a single-site location of a nuclear power facility and a hazardous industrial facility with different regulatory requirements is considered. A list of R&amp;D activities, which is necessary to confirm technical solutions and develop materials for licensing the NPEP project in regulatory authorities, is provided.</p></div>","PeriodicalId":480,"journal":{"name":"Atomic Energy","volume":"136 5-6","pages":"245 - 252"},"PeriodicalIF":0.4,"publicationDate":"2025-01-23","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143570972","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}
引用次数: 0
Results of BOR-60 reactor tests and post-reactor studies of experimental fuel rods with mixed uranium-plutonium nitride fuel BOR-60反应堆试验结果及混合铀-钚氮化燃料实验燃料棒的堆后研究
IF 0.4 4区 工程技术
Atomic Energy Pub Date : 2025-01-23 DOI: 10.1007/s10512-024-01158-9
F. N. Kryukov, A. V. Belyaeva, O. N. Nikitin, P. I. Grin, I. Yu. Zhemkov, M. V. Skupov, B. A. Tarasov, L. M. Zabudko
{"title":"Results of BOR-60 reactor tests and post-reactor studies of experimental fuel rods with mixed uranium-plutonium nitride fuel","authors":"F. N. Kryukov,&nbsp;A. V. Belyaeva,&nbsp;O. N. Nikitin,&nbsp;P. I. Grin,&nbsp;I. Yu. Zhemkov,&nbsp;M. V. Skupov,&nbsp;B. A. Tarasov,&nbsp;L. M. Zabudko","doi":"10.1007/s10512-024-01158-9","DOIUrl":"10.1007/s10512-024-01158-9","url":null,"abstract":"<div><p>In the course of the study, various designs of experimental fuel rods with mixed uranium-plutonium nitride fuel and claddings made of various materials were tested as part of the dismountable irradiation devices of the BOR-60 reactor. The experimental fuel rods represent the mock-ups of the BREST-OD-300 and BN-1200 reactor fuel rods, which differ from the real ones in their shorter length, smaller gas collector volume, and fuel column height. Reactor tests and post-reactor studies of 2020–2023 provided new data on the state of fuel, various types of fuel rod claddings, and fuel assembly design elements.</p></div>","PeriodicalId":480,"journal":{"name":"Atomic Energy","volume":"136 5-6","pages":"253 - 258"},"PeriodicalIF":0.4,"publicationDate":"2025-01-23","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143570971","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}
引用次数: 0
Forms and compounds of radioactive iodine in emissions of the JSC Karpov Institute of Physical Chemistry 卡尔波夫物理化学研究所放射物中放射性碘的形式和化合物
IF 0.4 4区 工程技术
Atomic Energy Pub Date : 2025-01-23 DOI: 10.1007/s10512-024-01169-6
A. A. Ekidin, M. E. Vasyanovich, E. I. Nazarov, N. V. Kuznetsov, O. Yu. Kochnov, A. N. Shvalev
{"title":"Forms and compounds of radioactive iodine in emissions of the JSC Karpov Institute of Physical Chemistry","authors":"A. A. Ekidin,&nbsp;M. E. Vasyanovich,&nbsp;E. I. Nazarov,&nbsp;N. V. Kuznetsov,&nbsp;O. Yu. Kochnov,&nbsp;A. N. Shvalev","doi":"10.1007/s10512-024-01169-6","DOIUrl":"10.1007/s10512-024-01169-6","url":null,"abstract":"<div><p>The paper provides the results of determining forms and compounds of radioactive iodine isotopes in emissions of a controlled source during normal operation of the VVR-ts research reactor and production of radiopharmaceuticals. The determined average daily volumetric activity of four iodine isotopes <sup>131</sup>I, <sup>132</sup>I, <sup>133</sup>I, and <sup>135</sup>I equals to 119, 126, 36, and 21 Bq/m<sup>3</sup>, respectively. A 97-% fraction of the first three isotopes has a gaseous form in the gas-air environment of the emission source. Weakly adsorbed compounds predominate among these gases. The isotope of <sup>135</sup>I was identified in a single experiment, which hampers drawing unambiguous conclusions about its forms and compounds in the emissions.</p></div>","PeriodicalId":480,"journal":{"name":"Atomic Energy","volume":"136 5-6","pages":"332 - 337"},"PeriodicalIF":0.4,"publicationDate":"2025-01-23","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143570898","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}
引用次数: 0
Role of the BOR-60 research reactor in the development of fast reactors BOR-60研究堆在快堆发展中的作用
IF 0.4 4区 工程技术
Atomic Energy Pub Date : 2025-01-23 DOI: 10.1007/s10512-024-01156-x
A. A. Tuzov, A. L. Izhutov, Yu. M. Krasheninnikov, I. Yu. Zhemkov, F. N. Kryukov
{"title":"Role of the BOR-60 research reactor in the development of fast reactors","authors":"A. A. Tuzov,&nbsp;A. L. Izhutov,&nbsp;Yu. M. Krasheninnikov,&nbsp;I. Yu. Zhemkov,&nbsp;F. N. Kryukov","doi":"10.1007/s10512-024-01156-x","DOIUrl":"10.1007/s10512-024-01156-x","url":null,"abstract":"<div><p>Being put into operation in December 1969, the BOR-60 reactor has been operating successfully for more than 54 years at design parameters without replacements of the main process equipment and sodium of the primary and secondary cooling circuits. Its operating experience was used in the development of BN-350 and -600 reactors. The experience of the BOR-60 operation and experimental studies was used to develop a concept of a MBIR multi-purpose research fast reactor currently being built at the site of the JSC “SSC RIAR”. The article presents the main results of the BOR-60 operation and experimental studies.</p></div>","PeriodicalId":480,"journal":{"name":"Atomic Energy","volume":"136 5-6","pages":"233 - 244"},"PeriodicalIF":0.4,"publicationDate":"2025-01-23","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143570973","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}
引用次数: 0
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