scholarly journals Experimental methods for determining the effective neutron multiplication factor of the “Yalina-Thermal” subcritical assembly

2020 ◽  
Vol 65 (2) ◽  
pp. 235-242 ◽  
Author(s):  
I. A. Edchik ◽  
T. N. Korbut ◽  
A. V. Kuzmin ◽  
S. E. Mazanik ◽  
V. P. Togushov ◽  
...  
Keyword(s):  
Neutron Flux ◽  
Probabilistic Method ◽  
Nuclear Research ◽  
Fuel Load ◽  
Neutron Density ◽  
Density Level ◽  
Multiplication Factor ◽  
Nuclear Facilities ◽  
Neutron Multiplication ◽  
Effective Neutron

To study the kinetics of subcritical systems and determine the optimal conditions for the transmutation of longlived radioactive waste in the neutron spectrum of ADS-systems the “Yalina” research nuclear facility was created at Joint Institute for Power and Nuclear Research – Sosny (Minsk, Belarus). The main safety indicator of a subcritical system (active zone reactivity) was measured for a “Yalina-Thermal” assembly via three independent methods: inverse multiplication, probabilistic and impulse ones. For the inverse multiplication method, the neutron flux density was monitored during assembly loading. For a fuel load of 285 EK-10 rods the neutron multiplication was M = 22.3±0.6, and the effective neutron multiplication coefficient was keff = 0.9551± 0.0016. The probabilistic method (Feynman-alpha method), based on measuring fluctuations in the neutron density level within a system with a fission chain reaction, gave the ratio of the variance to the average counting rate value D/n = 1.779±0.005, which corresponds to keff = 0.9597 ±0.0003. The pulse method is aimed at studying the neutron flux behavior of after the neutron pulse injection into the breeding system. Measurements were held with the same setup, used in the Feynman-alpha method. The measured decay constant of instantaneous neutrons is α = –670±0.7 1/s, which corresponds to keff = 0.9560±0.0001. The effective multiplication factor keff of the subcritical assembly “Yalina-Thermal”, obtained via three different independent methods, is around average value of keff = 0.9569 ± 0.0018. The methods considered can be used for subcritical level monitoring for ADS-systems and research nuclear facilities.

Burnup Credit in the Criticality Safety Analysis of Spent Fuel in Storage Systems

2014 ◽  
Author(s):  
Vladyslav Soloviov
Keyword(s):  
Isotopic Composition ◽  
Nuclear Fuel ◽  
Beta Decay ◽  
Spent Nuclear Fuel ◽  
Storage Systems ◽  
Multiplication Factor ◽  
Spent Fuel ◽  
Neutron Multiplication ◽  
Initial Fuel ◽  
Effective Neutron

In this paper accounting of spent nuclear fuel (SNF) burnup of RBMK-1000 with actinides and full isotopic composition has been performed. The following characteristics were analyzed: initial fuel enrichment, burnup fraction, axial burnup profile in the fuel assembly (FA) and fuel weight. As the results show, in the first 400 hours after stopping the reactor, there is an increase in the effective neutron multiplication factor (keff) due to beta decay of 239Np into 239Pu. Further, from 5 to 50 years, there is a decrease in keff due to beta decay of 241Pu into 241Am. Beyond 50 years there is a slight change in the criticality of the system. Accounting for nuclear fuel burnup in the justification of nuclear safety of SNF systems will provide an opportunity to increase the volume of loaded fuel and thus significantly reduce technology costs of handling of SNF.

Burnup Credit in the Criticality Safety Analysis of Spent Fuel in Transportation and Storage Systems

2013 ◽  
Author(s):  
Vladyslav Soloviov
Keyword(s):  
Nuclear Fuel ◽  
Beta Decay ◽  
Spent Nuclear Fuel ◽  
Storage Systems ◽  
Multiplication Factor ◽  
Spent Fuel ◽  
Neutron Multiplication ◽  
Initial Fuel ◽  
Effective Neutron ◽  
And Storage

In this paper accounting of spent nuclear fuel (SNF) burnup of RBMK-1000 only with actinides has been performed. The following characteristics were analyzed: initial fuel enrichment, burnup fraction, axial burnup profile in the fuel assembly (FA) and fuel weight. As the results show, in the first 400 hours after stopping the reactor, there is an increase in the effective neutron multiplication factor (keff) due to beta decay of 239Np into 239Pu. Further, from 5 to 50 years, there is a decrease in keff due to beta decay of 241Pu into 241Am. Beyond 50 years there is a slight change in the criticality of the system. Accounting for nuclear fuel burnup in the justification of nuclear safety of SNF systems will provide an opportunity to increase the volume of loaded fuel and thus significantly reduce technology costs of handling of SNF.

Accounting Nuclear Fuel Burnup for Criticality Safety Justification of Spent Nuclear Fuel Storages

2012 ◽  
Author(s):  
Vladislav Solovyov ◽  
Vitaliy Galchenko
Keyword(s):  
Fuel Assembly ◽  
Nuclear Fuel ◽  
Beta Decay ◽  
Spent Nuclear Fuel ◽  
Nuclear Safety ◽  
Fuel Burnup ◽  
Multiplication Factor ◽  
Neutron Multiplication ◽  
Initial Fuel ◽  
Effective Neutron

In this paper we have analyzed accounting spent nuclear fuel (SNF) burnup of VVER-1000 and RBMK-1000 only with actinides. The following characteristics were analyzed: initial fuel enrichment, burnup fraction, axial burnup profile in the fuel assembly (FA) and fuel weight. As the results show, in the first 400 hours after stopping the reactor, there is an increase in the effective neutron multiplication factor (keff) due to beta decay of 239Np into 239Pu. Further, from 5 to 50 years, decrease in keff due to beta decay of 241Pu into 241Am. In the future there is a slight change in criticality of the system. Accounting nuclear fuel burnup in the justification of nuclear safety of SNF storages will provide an opportunity to increase the volume of loaded fuel and thus significantly reduce technology costs of handling of SNF.

scholarly journals Validation of SCALE Model of VVR-M Reactor

2020 ◽  
Vol 18 ◽  
pp. 42-47
Author(s):  
V. I. Borysenko ◽  
◽  
V. V. Goranchuk ◽  
Keyword(s):  
Physical Model ◽  
Nuclear Reactor ◽  
Physical Characteristics ◽  
Multiplication Factor ◽  
The Core ◽  
Fuel Assemblies ◽  
Neutron Multiplication ◽  
Research Nuclear Reactor ◽  
Effective Neutron ◽  
Calculation Code

The peculiarities of development of neutron-physical model of the VVR-M research nuclear reactor in the SCALE calculation code are considered in the article. Models of separate core elements, which influence neutron-physical characteristics of VVR-M, have been developed. Simulation was performed using the CSAS6 control module. Validation of the VVR-M neutron-physical model, built in the SCALE calculation code, has been carried out by comparing the calculated value of the effective neutron multiplication factor with the critical reactor state at the beginning of seven fuel loads with the number of fuel assemblies in the core from 72 to 129. The model is developmed to determine the effective neutron multiplication factor in the reactor, as well as other neutron-physical characteristics, such as neutron spectrum, neutron flux density in various cells of the reactor. Thus, it is possible to conduct numerical experiments to determine the most optimal locations of research channels in the core of the VVR-M, to conduct physical experiments on the irradiation of the research samples, detectors, structural materials, etc. In the article, the simplifications accepted at construction of neutron-physical model of research nuclear reactor VVR-M in SCALE calculation code are presented. The main elements of the model are described: fuel assemblies, beryllium displacer, control rods.

Study of the Space-Time Neutron Multiplication Formula

2010 ◽  
Author(s):  
Jianli Hao ◽  
Wenzhen Chen ◽  
Shaoming Wang ◽  
De Zhang
Keyword(s):  
Neutron Flux ◽  
Discrete Time ◽  
Neutron Transport ◽  
Space Time ◽  
Neutron Density ◽  
Time Process ◽  
Delayed Neutrons ◽  
Multiplication Formula ◽  
Neutron Flux Density ◽  
Neutron Multiplication

The process of neutron multiplication is a discrete-time process, but the neutron transport theory takes neutron multiplication as a continuous neutron source, which ignores the discrete-time process of neutron multiplication, which would take in errors, so it is necessary for describing the process of neutron multiplication as a discrete-time process. “The neutron doubling formula including delayed neutrons” has been established which describes the process of neutron multiplication as a discrete-time process, but it has nothing to do with space. “The neutron doubling formula including delayed neutrons” could not be used to describe the variety of distributing of neutron density in transient process; it also could not be used to deal with the problem of three-dimensional space. In order to solve the problems mentioned above, the space-time neutron multiplication formula is established. Based on the theory of neutron multiplication, the concept of space is introduced to the neutron multiplication formula and the space-time neutron multiplication formula is established by taking into account of neutron transport. The formula can describe the inherent physical process of neutron multiplication in fission chain reaction system. The test of space-time neutron multiplication formula is done, which proves the formula is right. Given the initial neutron density as well as the multiplication factor, the formula can strictly describe the variety of neutron density (neutron flux density) with time. It could be used for setting a standard for estimating error for the measurement of neutron flux density as well as numerical calculation; the space-time neutron multiplication has larger applicability compared with the “neutron doubling formula including delayed neutrons”.

Sensitivity of the neutron multiplication factor to water ingress into a spent fuel cask

Kerntechnik ◽  
2020 ◽  
Vol 85 (1) ◽  
pp. 38-53
Author(s):  
M. J. Leotlela ◽  
I. Petr ◽  
A. Mathye
Keyword(s):  
Multiplication Factor ◽  
Spent Fuel ◽  
Water Ingress ◽  
Neutron Multiplication
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