Results 1 - 3 of 3
Results 1 - 3 of 3. Search took: 0.014 seconds
|Sort by: date | relevance|
[en] In the two-step CANDU core analysis, the lattice physics code WIMS-IST is used for generation of the few-group diffusion theory constants (hereafter, few-group constants), and the neutronics design parameters such as the effective multiplication factor (keff), the power distribution, reactivity coefficients, etc. of the reactor core are calculated by the diffusion theory code RFSP-IST. In addition, three dimensional (3-D) supercell calculations are conducted to take into account the effect of the reactivity devices perpendicular to the horizontal fuel channels by DRAGON-IST. Recently, the Monte Carlo (MC) few-group constant generation method has been successfully applied for the two-step reactor core analysis. In this paper, the CANDU reactor core analysis is performed with the two-group constants generated by the Seoul National University MC code, McCARD
[en] In this study, the EPBM is implemented in Seoul National university Monte Carlo (MC) code, McCARD which has the k uncertainty evaluation capability by the adjoint-weighted perturbation (AWP) method. The implementation is verified by comparing the sensitivities of the k-eigenvalue difference to the microscopic cross sections computed by the DPBM and the direct subtractions for the TMI-1 pin-cell problem. The uncertainty of the coolant void reactivity (CVR) in a CANDU fuel lattice model due to the ENDF/B-VII.1 covariance data is calculated by its sensitivities estimated by the EPBM. The method based on the eigenvalue perturbation theory (EPBM) utilizes the 1st order adjoint-weighted perturbation (AWP) technique to estimate the sensitivity of the eigenvalue difference. Furthermore this method can be easily applied in a S/U analysis code system equipped with the eigenvalue sensitivity calculation capability. The EPBM is implemented in McCARD code and verified by showing good agreement with reference solution. Then the McCARD S/U analysis have been performed with the EPBM module for the CVR in CANDU fuel lattice problem. It shows that the uncertainty contributions of nu of "2"3"5U and gamma reaction of "2"3"8U are dominant
[en] Highlights: • An efficient MC method for the sensitivity calculation of reactivity coefficients is developed. • The sensitivity of reactivity coefficient is calculated by MC second-order perturbation techniques. • Its effectiveness is examined in a two-group homogeneous problem and Godiva. • S/U analyses are performed for MDC of a LWR pin cell and FTC of a CANDU 6 lattice model. - Abstract: The uncertainty quantification of the reactivity coefficients such as the fuel temperature coefficient (FTC) and the moderator density coefficient (MDC) is crucial for the nuclear reactor safety margin evaluation. This paper proposes a continuous-energy MC second-order perturbation (MC2P) method as a new way to estimate efficiently the sensitivity of reactivity coefficients to nuclear cross section data. The proposed MC2P method takes into account the second-order effects of the fission operator and the fission source distribution. The effectiveness of the MC2P method implemented in a Seoul National University MC code, McCARD, is demonstrated in a Godiva 235U density coefficient problem via comparison of its results with direct subtraction MC calculation. It is shown that the new method can predict the cross section sensitivities of the reactivity coefficient more accurately even with much smaller number of MC history simulations than the direct subtraction MC method. It is also shown that the proposed method is applicable for quantifying the uncertainties of the MDC of a LWR pin cell problem and the FTC of a CANDU 6 lattice cell problem due to the uncertainties of the nuclear cross section input data represented by nuclear cross section covariance data.