Atomic Energy

A. V. Zrodnikov

G. I. Toshinskii

O. G. Grigor'ev

Yu. G. Dragunov

V. S. Stepanov

N. N. Klimov

I. I. Kopytov

V. N. Krushel'nitskii

A. A. Grudakov

SVBR-75/100 Multipurpose Modular Low-Power Fast Reactor with Lead–Bismuth Coolant

SVBR-75/100 多功能模块化低功率铅铋冷却快堆

E. O. Adamov

A. V. Kaplienko

V. V. Orlov

V. S. Smirnov

A. V. Lopatkin

V. V. Lemekhov

A. V. Moiseev

Brest Lead-Cooled Fast Reactor: From Concept to Technological Implementation

快中子反应堆：从概念到技术实现

ConclusionsThe neutrino method is essentially a remote method and does not require contact with the elements of the reactor. The accuracy and reproducibility of the neutrino procedure in measurements of the power and power output of the VVÉR-540 reactor are close to the data obtained with thermal measurements. For other types of reactors the neutrino method can be more accurate than the traditional methods. The neutrino method makes it possible to monitor plutonium production directly during the plutonium production process. It is possible to develop a method of neutrino tomography for monitoring the spatial nonuniformity of energy release in the fuel core of the reactor as well as methods for monitoring in the case of accidents.Neutrino installations are now portable and compact, and they can be further improved.We are not offering the neutrino method as an alternative to other methods of monitoring, which have been checked over a period of many years, but rather we regard it as a method for extracting additional information in order to obtain a more complete picture of reactor operation.We thank A. G. Vershinskii for providing the electronics for the experiment, and I. N. Machulin for helpful discussions and for calculating the composition of the fuel core, and the management and personnel of the nuclear power plant for assistance.

Yu. V. Klimov

V. I. Kopeikin

L. A. Mikaélyan

K. V. Ozerov

V. V. Sinev

Neutrino method remote measurement of reactor power and power output

反应堆功率和功率输出的中微子方法远程测量

O. M. Saraev

N. N. Oshkanov

V. M. Poplavskii

Yu. M. Ashurko

M. V. Bakanov

B. A. Vasiliev

Yu. L. Kamanin

V. N. Ershov

M. N. Svyatkin

A. S. Korolkov

Yu. M. Krasheninnikov

Operating experience and prospects for future development of sodium-cooled fast reactors

钠冷快堆的运行经验和未来发展前景

M. N. Belonogov

I. A. Volkov

N. D. Dyrda

G. N. Rykovanov

I. V. Sannikov

P. A. Sannikova

R. R. Fazylov

D. V. Khmelnitsky

V. A. Shelan

Liquidus temperature study of LiF–NaF–KF-based melts simulating the fuel salt of a molten-salt reactor for Np, Am, Cm transmutation

基于LiF-NaF-KF熔盐的液相线温度研究：模拟熔盐反应堆燃料盐中Np、Am、Cm的嬗变

M. V. Antipov

M. A. Uvakin

A. L. Nikolaev

I. V. Makhin

E. V. Sotskov

An analysis of machine learning for the safety justification of VVER reactors

VVER反应堆安全验证的机器学习分析

Development of a neural-network methodology for the safety justification of VVER reactors in manoeuvring modes

VVER反应堆操纵模式安全验证的神经网络方法论发展

E. E. Ostashkina

A. E. Savkin

Yu. T. Slastennikov

Development and testing of methods for extracting and conditioning accident elimination sorbents in the basements of the Fukushima Daiichi NPP

福岛第一核电站地下室事故消除吸附剂提取与处理方法的开发与测试

A. V. Shishkin

V. Yu. Shishkin

P. N. Mushnikov

Yu. P. Zaikov

Formation of intermetallics during the metallization of model nuclear fuel based on uranium dioxide containing oxides of rare earth metals and palladium

含稀土金属和钯氧化物的模拟核燃料二氧化铀金属化过程中金属间化合物的形成

Abbreviation	At Energy
Journal Impact	0.44
Quartiles(Global)	NUCLEAR SCIENCE & TECHNOLOGY(Q4)

ISSN	1063-4258, 1573-8205
h-index	22

Publication Information	Publisher: Springer GmbH & Co, Auslieferungs-Gesellschaf，Publishing cycle: Monthly，Journal Type: journal，Open Access Journals: No
Basic data	Year of publication: 1992，Proportion of original research papers: 100.00%，Self Citation Rate:0.00%， Gold OA Rate: 0.00%
Average review cycle	网友分享经验：偏慢,4-8周
Average recruitment ratio	网友分享经验：容易

Atomic Energy

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SVBR-75/100 Multipurpose Modular Low-Power Fast Reactor with Lead–Bismuth Coolant

Brest Lead-Cooled Fast Reactor: From Concept to Technological Implementation

Neutrino method remote measurement of reactor power and power output

Operating experience and prospects for future development of sodium-cooled fast reactors

Liquidus temperature study of LiF–NaF–KF-based melts simulating the fuel salt of a molten-salt reactor for Np, Am, Cm transmutation

Liquidus temperature study of LiF–NaF–KF-based melts simulating the fuel salt of a molten-salt reactor for Np, Am, Cm transmutation

An analysis of machine learning for the safety justification of VVER reactors

Development of a neural-network methodology for the safety justification of VVER reactors in manoeuvring modes

Development and testing of methods for extracting and conditioning accident elimination sorbents in the basements of the Fukushima Daiichi NPP

Formation of intermetallics during the metallization of model nuclear fuel based on uranium dioxide containing oxides of rare earth metals and palladium