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RISS 인기검색어
- 검색키워드
- 저자 : Pautz, A. (검색결과 4 건)
- 무료
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Mala, Petra,Pautz, Andreas Korean Nuclear Society 2020 Nuclear Engineering and Technology Vol.52 No.11
Currently, the pin-by-pin homogenized solvers are a very active research field as they can, unlike the nodal codes, directly predict the local power, while requiring significantly less computational resources than the heterogeneous transport codes. This paper presents a recently developed pin-by-pin diffusion/SP3 solver Tortin, its spatial discretization method and the reflector treatment. Regarding the spatial discretization, it was observed that the finite difference method applied on pin-cell size mesh does not properly capture the big flux change between MOX and uranium fuel, while the nodal expansion method is more accurate but too slow. If the finite difference method is used with a finer mesh in the outer two pin rows of the fuel assembly, it increases the required computation time by only 50%, but decreases the pin power errors below 1% with respect to lattice code reference solutions. The paper further describes the coupling of Tortin with a microscopic depletion solver. Several verification tests show that the SP3 pin-by-pin solver can reproduce the heterogeneous transport solvers results with very good accuracy, even for fuel cycle depletion of very heterogeneous core employing MOX fuel or inserted control rods, while being two orders of magnitude faster.
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Uncertainty Analyses with Nuclear Covariance Ddata in Reactor Core Ccalculations
W. Zwermann,B. Krzykacz-Hausmann,L. Gallner,A. Pautz,M. Mattes 한국물리학회 2011 THE JOURNAL OF THE KOREAN PHYSICAL SOCIETY Vol.59 No.23
To investigate the influence of nuclear data uncertainties on the core power distributions systematically, the uncertainty and sensitivity software package SUSA developed at GRS was extended for the use with nuclear covariance data. Varied nuclear data are generated randomly corresponding to the uncertainty information from the covariance matrices. After performing a large number of calculations with these data, the results are statistically evaluated; this can be done not only with integral, but also differential output quantities like the assembly power distribution. The method is applied to multi-group Monte Carlo calculations for the PWR MOX/UO<sub>2</sub> benchmark core. The uncertainty in the resulting k-eff turns out to be of the same order as for pin cell calculations. Unexpectedly large uncertainties result for the normalized radial power distribution. The relative 1σ-uncertainties in the central and peripheral assembly powers are of the order of 4%. The dominating contribution is the uncertainty in the number of neutrons per fission of ^(239)Pu; therefore, the effect is particularly pronounced in a UO_2/MOX core with a high amount of ^(239)Pu in the isotopic composition.
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Investigation on the effect of eccentricity for fuel disc irradiation tests
Scolaro, A.,Van Uffelen, P.,Fiorina, C.,Schubert, A.,Clifford, I.,Pautz, A. Korean Nuclear Society 2021 Nuclear Engineering and Technology Vol.53 No.5
A varying degree of eccentricity always exists in the initial configuration of a nuclear fuel rod. Its impact on traditional LWR fuel is limited as the radial gap closes relatively early during irradiation. However, the effect of misalignment is expected to be more relevant in rods with highly conductive fuels, large initial gaps and low conductivity filling gases. In this paper, we study similar characteristics in the experimental setup of two fuel disc irradiation campaigns carried out in the OECD Halden Boiling Water Reactor. Using the multi-dimensional fuel performance code OFFBEAT, we combine 2-D axisymmetric and 3-D simulations to investigate the effect of eccentricity on the fuel temperature distribution. At the same time, we illustrate how the advent of modern tools with multi-dimensional capabilities might further improve the design and interpretation of in-pile separate-effect tests and we outline the potential of such an analysis for upcoming experiments.
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Uncertainty analyses of spent nuclear fuel decay heat calculations using SCALE modules
Shama, Ahmed,Rochman, Dimitri,Pudollek, Susanne,Caruso, Stefano,Pautz, Andreas Korean Nuclear Society 2021 Nuclear Engineering and Technology Vol.53 No.9
Decay heat residuals of spent nuclear fuel (SNF), i.e., the differences between calculations and measurements, were obtained previously for various spent fuel assemblies (SFA) using the Polaris module of the SCALE code system. In this paper, we compare decay heat residuals to their uncertainties, focusing on four PWRs and four BWRs. Uncertainties in nuclear data and model inputs are propagated stochastically through calculations using the SCALE/Sampler super-sequence. Total uncertainties could not explain the residuals of two SFAs measured at GE-Morris. The combined z-scores for all SFAs measured at the Clab facility could explain the resulting deviations. Nuclear-data-related uncertainties contribute more in the high burnup SFAs. Design and operational uncertainties tend to contribute more to the total uncertainties. Assembly burnup is a relevant variable as it correlates significantly with the SNF decay heat. Additionally, burnup uncertainty is a major contributor to decay heat uncertainty, and assumptions relating to these uncertainties are crucial. Propagation of nuclear data and design and operational uncertainties shows that the analyzed assemblies respond similarly with high correlation. The calculated decay heats are highly correlated in the PWRs and BWRs, whereas lower correlations were observed between decay heats of SFAs that differ in their burnups.
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