scholarly journals Nuclear data sensitivity and uncertainty analysis of effective neutron multiplication factor in various MYRRHA core configurations

2017 ◽  
Vol 101 ◽  
pp. 330-338 ◽  
Author(s):  
P. Romojaro ◽  
F. Álvarez-Velarde ◽  
I. Kodeli ◽  
A. Stankovskiy ◽  
C.J. Díez ◽  
...  
Keyword(s):  
Uncertainty Analysis ◽  
Nuclear Data ◽  
Multiplication Factor ◽  
Neutron Multiplication ◽  
Effective Neutron ◽  
Sensitivity And Uncertainty Analysis ◽  
Data Sensitivity

scholarly journals Sensitivity and uncertainty analysis of βeff for MYRRHA using a Monte Carlo technique

2018 ◽  
Vol 4 ◽  
pp. 42 ◽  
Author(s):  
Hiroki Iwamoto ◽  
Alexey Stakovskiy ◽  
Luca Fiorito ◽  
Gert Van den Eynde
Keyword(s):  
Monte Carlo ◽  
Uncertainty Analysis ◽  
Nuclear Data ◽  
Ratio Method ◽  
Delayed Neutron ◽  
Continuous Energy ◽  
A Value ◽  
Transport Code ◽  
Sensitivity And Uncertainty Analysis ◽  
Data Sensitivity

This paper presents a nuclear data sensitivity and uncertainty analysis of the effective delayed neutron fraction βeff for critical and subcritical cores of the MYRRHA reactor using the continuous-energy Monte Carlo N-Particle transport code MCNP. The βeff sensitivities are calculated by the modified k-ratio method proposed by Chiba. Comparing the βeff sensitivities obtained with different scaling factors a introduced by Chiba shows that a value of a = 20 is the most suitable for the uncertainty quantification of βeff. Using the calculated βeff sensitivities and the JENDL-4.0u covariance data, the βeff uncertainties for the critical and subcritical cores are determined to be 2.2 ± 0.2% and 2.0 ± 0.2%, respectively, which are dominated by delayed neutron yield of 239Pu and 238U.

Sensitivity and uncertainty analysis on keff due to nuclear data in the KRITZ-2:19 – Comparison between JENDL-4.0 and ENDF/B-VII.1

2019 ◽  
Vol 129 ◽  
pp. 308-315
Author(s):  
Abdulaziz Ahmed ◽  
H. Boukhal ◽  
T. El Bardouni ◽  
M. Makhloul ◽  
E. Chakir ◽  
...  
Keyword(s):  
Uncertainty Analysis ◽  
Nuclear Data ◽  
Sensitivity And Uncertainty Analysis

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.

Enhancement of Stochastic Sampling Capability in RMC Code

2018 ◽  
Author(s):  
Guanlin Shi ◽  
Yishu Qiu ◽  
Kan Wang
Keyword(s):  
Uncertainty Analysis ◽  
Sampling Method ◽  
Nuclear Data ◽  
Data Uncertainty ◽  
Monte Carlo Code ◽  
Tabular Data ◽  
Stochastic Sampling ◽  
First Order ◽  
Fission Spectrum ◽  
Sensitivity And Uncertainty Analysis

As people pay more attention to nuclear safety analysis, sensitivity and uncertainty analysis has become a research hotspot. In our previous research, we had developed an integrated, built-in stochastic sampling module in the Reactor Monte Carlo code RMC [1]. Using this module, we can perform nuclear data uncertainty analysis. But at that time the uncertainty of fission spectrum was not considered. So, in this work, the capability of computing the uncertainty of keff induced by the uncertainty of fission spectrum, including tabular data form and formula form, is implemented in RMC code based on the stochastic sampling method. The algorithms and capability of computing keff uncertainty induced by uncertainty of fission spectrum in RMC are verified by comparison with the results calculated by the first order uncertainty quantification method [2].

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.

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