Results 1 - 10 of 670
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[en] The results of measurements βeff and βeff/Λ and calculation results based on various sets of evaluated six-group delayed neutron parameters for the coupled fast-thermal system HERBE are shown in this paper
[en] There is a growing interest for the use of some actinide elements as nuclear fuels. In this paper we predict the total delayed neutron yields as well as the energy components of the delayed activities. This prediction is given for actinide elements which are important in the nuclear energy field, and do not have experimental data
[en] The delayed neutron parameters and methods used for calculation in reactor safety studies are verified by measurement of the effective neutron fraction in the coupled fast-thermal system 'HERBE'. (authors). 13 refs., 1 tabs., 3 figs
[en] The purpose of this paper is to report a series of measurements of the energy feedback coefficient. The same measurements were used to determine the ratio of effective delayed neutron fraction to mean neutron lifetime. The data were also used to determine whether any feedback mechanisms exist which are proportional to power level rather than to energy
[en] The applicability of a 3-dimensional reactor kinetics model to the rod ejection accidents was examined in the view of the enthalpy rise in the fuel rod. PARCS code was used for the 3-dimensional reactor kinetics mode. As a result of the various parametric analysis, the values of 'maximum enthalpy rise' for 'ρrod(ejected rod worth)- β(delayed neutron fraction)' were obtained, and it could be expressed as a linear curve for the complicated and various reactor design and operation conditions. If the theoretical basis of this linear curve can be verified for all the loading patterns and the operation conditions, it will be an index in the regulatory evaluation for the validation of the 3-dimensional reactor kinetics analysis for the rod ejection accidents
[en] New methods are proposed to estimate the effective delayed neutron fraction βeff in Monte Carlo calculations: the eigenvalue methods jointly used with the differential operator sampling and correlated sampling techniques. In particular, the eigenvalue method with the differential operator sampling technique has a distinct feature that it theoretically gives an exact βeff value. To verify the proposed methods, Monte Carlo calculations are performed for several systems with simple geometry. It is found that the results obtained with the proposed methods agree with the reference deterministic results within sufficiently small statistical uncertainties. The indirect perturbed source effect must be taken into account to estimate an exact βeff value.
[en] 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 α introduced by Chiba shows that a value of α = 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. (authors)
[en] Highlights: • A capability of computing adjoint-weighted kinetic parameters is developed in the Reactor Monte Carlo code RMC. • Three algorithms of adjoint-weighted kinetic parameters based on the IFP method have been investigated. • The adjoint-weighted kinetic parameters computed by these algorithms in RMC agree well with MCNP6 and experimental results. - Abstract: In this work, the capability of computing adjoint-weighted kinetic parameters, including effective delayed neutron fraction and neutron generation time, was implemented in the Reactor Monte Carlo (RMC) code based on the iterated fission probability (IFP) method. Three algorithms, namely, the Non-Overlapping Blocks (NOB) algorithm, the Multiple Overlapping Blocks (MOB) algorithm and the superhistory algorithm, were implemented in RMC to investigate their accuracy, computational efficiency and estimation of variance. The algorithms and capability of computing kinetic parameters in RMC were verified and validated by comparison with MCNP6 as well as experimental results through a set of multi-group problems and continuous-energy problems.