Atomic Energy最新文献_第5页

Neutron generators with an external target at the remote end of the neutron tube ion guide 在中子管离子波导的远端有一个外部目标的中子发生器

IF 0.4 4区工程技术

Atomic Energy Pub Date : 2024-11-25 DOI: 10.1007/s10512-024-01152-1

S. V. Syromukov, E. P. Bogolyubov, D. I. Yurkov

引用次数: 0

Verification of the TARUSA-9 software for measuring the activity of fission products in the coolant of a pressurized water reactor 验证用于测量压水堆冷却剂中裂变产物活性的TARUSA-9软件

IF 0.4 4区工程技术

Atomic Energy Pub Date : 2024-11-25 DOI: 10.1007/s10512-024-01148-x

A. V. Lopatkin, A. V. Nikitin, A. V. Mikhaylov, A. A. Gordeev, V. A. Vasilenko, B. A. Gusev, M. N. Baev

{"title":"Verification of the TARUSA-9 software for measuring the activity of fission products in the coolant of a pressurized water reactor","authors":"A. V. Lopatkin, A. V. Nikitin, A. V. Mikhaylov, A. A. Gordeev, V. A. Vasilenko, B. A. Gusev, M. N. Baev","doi":"10.1007/s10512-024-01148-x","DOIUrl":"10.1007/s10512-024-01148-x","url":null,"abstract":"<div><p>The paper presents the results of the verification of the TARUSA‑9 software, which uses an exact analytical solution to the system of linear differential equations of mass transfer, for measuring the activity of fission products in the water coolant of a pressurized water reactor. A comparison of the calculation and measurement results was performed for three typical reactor operating modes: operation at a variable power with the disabled reactor water purification system, at a constant power with variable purification water consumption, and at a variable power and purification water consumption. The activity of 14 fission product isotopes in 39 reactor water samples was compared. The discrepancy between the results of calculations and measurements does not exceed the measurement error in more than 90% of cases. For the entire set of comparison points, the average relative difference between calculations and measurements is equal to 0 at a confidence probability of 0.95, standard deviation of 1%, and confidence interval of ± 2%. For individual nuclides, these values vary from −10 to 7%, from 0 to 12%, and from −23 to 24%, respectively.</p></div>","PeriodicalId":480,"journal":{"name":"Atomic Energy","volume":"136 3-4","pages":"171 - 179"},"PeriodicalIF":0.4,"publicationDate":"2024-11-25","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142798381","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

A study of the coolant flow at the inlet to the fuel assembly of the RITM nuclear icebreaker reactor RITM核破冰船反应堆燃料组件入口冷却剂流动研究

IF 0.4 4区工程技术

Atomic Energy Pub Date : 2024-11-11 DOI: 10.1007/s10512-024-01143-2

S. M. Dmitriev, T. D. Demkina, A. A. Dobrov, D. V. Doronkov, A. N. Pronin, A. V. Ryazanov

{"title":"A study of the coolant flow at the inlet to the fuel assembly of the RITM nuclear icebreaker reactor","authors":"S. M. Dmitriev, T. D. Demkina, A. A. Dobrov, D. V. Doronkov, A. N. Pronin, A. V. Ryazanov","doi":"10.1007/s10512-024-01143-2","DOIUrl":"10.1007/s10512-024-01143-2","url":null,"abstract":"<div><p>The article presents the results of an experimental study connected with the features of a coolant flow at the inlet to the fuel assembly (FA) in the cartridge core of a RITM nuclear icebreaker reactor. The study is focused at the non-uniformity of the axial flow velocity at the FA inlet. To this end, a series of experiments was carried out using a scale model including structural elements of the inlet section between the orifice plate and the unit attaching fuel rods to the diffuser, as well as the section of the fuel rod bundle up to the second spacer grid. In addition, the propagation of a non-uniform axial flow along the length of the fuel rod bundle was assessed during the experiments. The studies were carried out using the pneumometric method in characteristic cross-sections along the length of the model. Features of the coolant flow are visualized by cartograms of the axial velocity for the working medium flow. The results of experiments were used for hydraulic profiling of the FA inlet section. The obtained experimental database can be used to validate the LOGOS CFD software and clarify the methods for thermal-hydraulic calculations of cores in a cell approximation.</p></div>","PeriodicalId":480,"journal":{"name":"Atomic Energy","volume":"136 3-4","pages":"141 - 145"},"PeriodicalIF":0.4,"publicationDate":"2024-11-11","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142798532","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

Decommissioning of fuel elements at the reactor of the “Akademik Lomonosov” floating thermal power plant “罗蒙诺索夫院士”号浮动热电厂反应堆燃料元件退役

IF 0.4 4区工程技术

Atomic Energy Pub Date : 2024-11-11 DOI: 10.1007/s10512-024-01144-1

G. V. Kulakov, Yu. V. Konovalov, A. V. Vatulin, A. A. Kosaurov, S. A. Ershov, E. V. Mainikov, V. I. Sorokin, A. V. Kozlov

引用次数: 0

Application of metal-ceramic uranium-molybdenum fuel in an aluminum matrix for research reactors 金属陶瓷铀钼燃料在研究堆铝基中的应用

IF 0.4 4区工程技术

Atomic Energy Pub Date : 2024-11-11 DOI: 10.1007/s10512-024-01145-0

A. M. Savchenko, Y. V. Konovalov, G. V. Kulakov, B. A. Tarasov, S. A. Ershov, E. V. Mainikov, D. A. Bubenschikov

引用次数: 0

Natural air circulation in an open circuit 开路中的自然空气循环

IF 0.4 4区工程技术

Atomic Energy Pub Date : 2024-11-11 DOI: 10.1007/s10512-024-01140-5

V. I. Solonin, V. G. Krapivtsev, S. I. Getya, P. V. Markov, S. G. Kutychkin

{"title":"Natural air circulation in an open circuit","authors":"V. I. Solonin, V. G. Krapivtsev, S. I. Getya, P. V. Markov, S. G. Kutychkin","doi":"10.1007/s10512-024-01140-5","DOIUrl":"10.1007/s10512-024-01140-5","url":null,"abstract":"<div><p>The paper presents experimental data on the formation of natural air circulation in an open circuit with the length of 22.4 m and a pipe diameter of 120 mm, which is located in a room connected to the atmosphere, in a case of the emergency cooldown of a lead-cooled reactor. Natural circulation without heating was ensured by the difference between room and atmospheric temperatures, while the air heated to the temperature of 170–470 °C by the electric heater with a power of 7–12 kW circulated due to the reduction in its density at the outlet. A model is proposed for calculating the pressure of natural circulation using generalized experimental data obtained in an open-air circuit under the conditions of a quasi-stationary flow, temperature, and static pressure. The generalization error does not exceed 30% with a total error of 20%. Hydraulic losses in the circuit and heat losses through the circuit insulation were determined. Atmospheric turbulence is demonstrated to cause pressure pulsations in the circuit with an amplitude close to the flow velocity head of 8.5–20 Pa in the duct. When the atmospheric air pressure at the inlet to the circuit exceeds that at the outlet, the amplitude increases to 280 Pa.</p></div>","PeriodicalId":480,"journal":{"name":"Atomic Energy","volume":"136 3-4","pages":"117 - 126"},"PeriodicalIF":0.4,"publicationDate":"2024-11-11","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142798535","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

Numerical simulation of plutonium measurements using an active well coincidence counter (AWCC) 利用有源阱符合计数器（AWCC）测量钚的数值模拟

IF 0.4 4区工程技术

Atomic Energy Pub Date : 2024-11-11 DOI: 10.1007/s10512-024-01146-z

D. A. Vladimirov, V. Yu. Rogozhkin, A. Yu. Gorbunova, T. B. Aleeva, P. A. Pugachev

引用次数: 0

Calculation analysis of a severe accident at a nuclear power plant with a sodium-cooled fast reactor 某核电站钠冷快堆严重事故的计算分析

IF 0.4 4区工程技术

Atomic Energy Pub Date : 2024-11-08 DOI: 10.1007/s10512-024-01138-z

A. A. Peregudov, N. V. Solomonova, L. A. Schekotova, S. V. Zabrodskaya, M. V. Levanova, O. O. Peregudova, I. V. Buryevskiy, D. V. Dmitriev, E. P. Averchenkova

{"title":"Calculation analysis of a severe accident at a nuclear power plant with a sodium-cooled fast reactor","authors":"A. A. Peregudov, N. V. Solomonova, L. A. Schekotova, S. V. Zabrodskaya, M. V. Levanova, O. O. Peregudova, I. V. Buryevskiy, D. V. Dmitriev, E. P. Averchenkova","doi":"10.1007/s10512-024-01138-z","DOIUrl":"10.1007/s10512-024-01138-z","url":null,"abstract":"<div><p>The article represents an approach used at the IPPE JSC to the computational analysis of a severe accident caused by a complete cessation of the system and emergency power supply at a nuclear power plant with a sodium-cooled fast reactor. The analysis starts with the initial event and ends with the calculation of the radiation dose to the population. The used codes and methods, the development of the accident, and its consequences are described. The COREMELT code simulating thermal-hydraulic and neutron-physical processes in fast reactors was used to determine the scale of the core destruction. The accumulation of fission products in the fuel is calculated by the SKIF 1.0 code; their release to the gas volumes of fuel elements, to the coolant during the depressurization of fuel elements, and then to the gas cavity of the reactor is determined methodically (in the future, it is planned to use the Alpha‑M code). The transfer through the gas system is calculated by the COREMELT gas module. The activity of fission products released into the environment is methodically assessed (in the future, it is planned to use the KUPOL-BR code). The VYBROS-BN code was used to determine the radiation dose to the population. According to the performed calculation analysis, despite the significant destruction of the core, the dose load to the population at the border of the zone will not exceed the standards.</p></div>","PeriodicalId":480,"journal":{"name":"Atomic Energy","volume":"136 3-4","pages":"101 - 109"},"PeriodicalIF":0.4,"publicationDate":"2024-11-08","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142798349","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

Key lessons from major radiation accidents for emergency response in agriculture 重大辐射事故对农业应急反应的重要启示

IF 0.4 4区工程技术

Atomic Energy Pub Date : 2024-11-05 DOI: 10.1007/s10512-024-01153-0

S. V. Fesenko

引用次数: 0

Development of a method for measuring the effective fraction of delayed neutrons using a 240Pu spontaneous fission source 利用 240Pu 自发裂变源开发测量延迟中子有效部分的方法

IF 0.4 4区工程技术

Atomic Energy Pub Date : 2024-10-11 DOI: 10.1007/s10512-024-01137-0

V. A. Grabezhnoy, V. A. Dulin, G. M. Mikhailov

引用次数: 0