Simulation of the burnup in cell calculation using the wimsd-5b code considering different nuclear data libraries

Comparison of FANT results using the ENDF/B-VII.1, JEFF-3.3 and TENDL2017 nuclear data libraries

Nodal Integral method for multi-group neutron diffusion equation in three dimensional cylindrical coordinate system

In the present work, we discuss the multi-group neutron diffusion equation solution in a homogeneous cylinder geometry case. We construct two kinds of solutions for the related problem considering two energy groups and a steady-state case. Upon applying the finite Hankel transform to the diffusion equation, one obtains a generalized system for the both neutron fluxes (fast and thermal), which may be solved in a closed form solution and depends on the specific source terms. As an equivalent approach we solve the problem by the infinite Hankel transform. Comparison of the two results shows that both methods provide acceptable solutions which are comparable to the ones in the literature. Applications for four parameter sets allow to conject that the infinite Hankel transform is the simpler approach although approximate, but providing solutions comparable to the exact solution by the finite Hankel transform.

In this paper, the standard multigroup neutron diffusion equations are derived as an asymptotic approximation to the multigroup neutron transport equations. The asymptotic analysis employs a scaling that (1) is suggested by the multigroup neutron diffusion equations themselves and (2) generalizes the long-known asymptotic scaling for monoenergetic transport problems. Two other asymptotic scalings of the multigroup transport equations are also considered, both of which lead to a new “group-collapsed” (monoenergetic) “equilibrium” diffusion approximation. The standard multigroup and equilibrium diffusion approximations are shown to preserve certain nonasymptotic properties of the multigroup transport equations. Generalizations of the analyses in this paper, and possible practical applications, are discussed.

Polynomial relaxation in the quasi-static approach and its implementation in nonlinear reactor transient analyses

Characterization of the neutron spectra in three irradiation channels of the JSI TRIGA reactor using the GRUPINT spectrum adjustment code

SummaryThe multigroup neutron diffusion equations (an approximation of the neutron transport equation) are widely used for studying the motion of neutrons and their interactions with stationary background materials. Solving the multigroup neutron diffusion equations is challenging because the unknowns are tightly coupled through scattering and fission events, and solutions with high spatial resolution of full reactor cores in multiphysics environments are frequently required. In this paper, we focus on the development of a scalable, parallel preconditioner for solving the system of equations arising from the finite‐element discretization of the multigroup neutron diffusion equations in space and an implicit finite‐difference scheme in time. The parallel preconditioner (here referred to as the “fully coupled Schwarz preconditioner”) is constructed by monolithically applying the overlapping domain decomposition method together with a smoothed aggregation‐based coarse space to the coupled system. Our approach is different from the traditional block Gauss–Seidel sweep method that applies the preconditioner from the fast group to the thermal group sequentially, and we demonstrate that it provides significant improvements in terms of both the number of iterations required and the total compute time for a system of equations with millions of unknowns on a large supercomputer.

PARCS code multi-group neutron diffusion constants generation using Monte Carlo method

This work is aimed to present the main differences of nuclear data uncertainties among three different nuclear data libraries: EAF-2007, EAF-2010 and SCALE-6.0, under different neutron spectra: LWR, ADS and DEMO (fusion). To take into account the neutron spectrum, the uncertainty data are collapsed to onegroup. That is a simple way to see the differences among libraries for one application. Also, the neutron spectrum effect on different applications can be observed. These comparisons are presented only for (n,fission), (n,gamma) and (n,p) reactions, for the main transuranic isotopes (234,235,236,238U, 237Np, 238,239,240,241Pu, 241,242m,243Am, 242,243,244,245,246,247,248Cm, 249Bk, 249,250,251,252Cf). But also general comparisons among libraries are presented taking into account all included isotopes. In other works, target accuracies are presented for nuclear data uncertainties; here, these targets are compared with uncertainties on the above libraries. The main results of these comparisons are that EAF-2010 has reduced their uncertainties for many isotopes from EAF-2007 for (n,gamma) and (n,fission) but not for (n,p); SCALE-6.0 gives lower uncertainties for (n,fission) reactions for ADS and PWR applications, but gives higher uncertainties for (n,p) reactions in all applications. For the (n,gamma) reaction, the amount of isotopes which have higher uncertainties is quite similar to the amount of isotopes which have lower uncertainties when SCALE-6.0 and EAF-2010 are compared. When the effect of neutron spectra is analysed, the ADS neutron spectrum obtained the highest uncertainties for (n,gamma) and (n,fission) reactions of all libraries.\n

In this paper, we developed a new approach of analytic function expansion nodal (AFEN) method to solve the multi-group and multi-dimensional neutron diffusion equation in reactor cores with hexagonal fuel assembly. This method represents a multidimensional intra nodal flux distribution in terms of analytic basis functions at any points in the node. New types of boundary conditions have been considered that constrain the intranodal flux distributions in the hexagonal-z node, which include twelve radial surface-averaged partial currents and two axial surface-averaged partial currents. We utilized the coarse group rebalancing (CGR) method to increase the speed of code calculations. The computer code takes a few-groups cross sections produced by a lattice code and calculates the effective multiplication factor (keff ), flux in multi-group energy, reactivity, and the relative power density at each fuel assembly. Finally, the solution accuracy is tested for two well-known benchmark problems. The numerical results demonstrate that the new AFEN method is an accurate method for calculating keff and power density distribution in hexagonal-z geometries.

Comparative studies of thorium fuel cycles in boiling water reactor (BWR) with JENDL 3.2 and JEF 2.2 nuclear data libraries have been performed. The once through cycle (OTC) and the all heavy metals (HMs) confining scenarios were evaluated. In this study, we have utilised the time independent fuel burnup scheme, that we called equilibrium burnup. OTC case with JEF 2.2 results in the higher required U-233 concentration for criticality compared to that of JENDL 3.2. In contrast, the required U-233 concentration for criticality becomes smaller for the all HMs confining case with JEF 2.2 compared to that of JENDL 3.2. Conversion ratio increases with the boosting of void fraction for both scenarios of the two nuclear data libraries. JENDL 3.2 gives the harder neutron spectra compared to that of JEF 2.2 for both scenarios.

In this paper, a discontinuous Bubnov-Galerkin isogeometric analysis discrete ordinate (DBG-IGA-SN) form of the subgroup projection and resonance spectrum expansion (RSE) methods within the DRAGON5 code is presented. The DBG-IGA-SN method uses non-uniform rational B-spline (NURBS) basis functions to spatially discretize the neutron transport equation. The NURBS basis functions, which are used to represent both the solution field and the geometry, are able to exactly represent pin cell geometries even on the coarsest spatial representation. This is not possible with traditional Lagrangian finite elements. This is the first instance of a DBG-IGA-SN discretization of the RSE method. We compared the numerical accuracy of the DBG-IGA-SN method to the existing collision probability (CP) method within DRAGON5 against Serpent and CP ultrafine-group reference solutions for a series of uranium-oxide and mixed-oxide Rowlands pins. In DRAGON5, the CP method is frequently used for resonance self-shielding calculations. We use it in this work as a deterministic method standard for comparison to assess DBG-IGA-SN performance. We found that our implementation of the DBG-IGA-SN method achieved similar numerical accuracy compared to the CP method overall. The infinite multiplication factor errors were no greater than 107 pcm for the uranium-oxide cases and 152 pcm for the mixed-oxide cases with a Serpent reference, with a variation of about 10 pcm between DBG-IGA-SN and CP. For the CP ultrafine-group method reference cases, the infinite multiplication factor errors did not surpass 108 pcm and 127 pcm for the uranium-oxide and mixed-oxide cases, respectively. In addition, DBG-IGA-SN had lower infinite multiplication factor numerical errors than CP in all cases, in one case by 21 pcm. We also found that the RSE method had higher errors in the multiplication factors and reaction rates than the subgroup projection method for both the uranium-oxide and mixed-oxide cases when using the CP ultrafine-group method reference. Future work will investigate the relative performance of the DBG-IGA-SN for larger, more complex cases, where strengths of the method, such as its scalability and ability to model exact geometries and anisotropic scatter, may make it desirable for production resonance self-shielding calculations.

Neutronic experiments with fluorine rich compounds at LR-0 reactor

Work on the analysis of the 238U Livermore Pulsed Sphere experiments relevant for the validation of the CIELO evaluation is presented. The aim of the work described in this report was to validate the new ENDF/B-VIII.0 nuclear data library by comparing calculations to the benchmark experiments. The pulsed sphere experiments were performed at the Lawrence Livermore National Laboratory (LLNL). Different references describing the experimental configuration were reviewed. The evaluated experiments with their characteristics are listed. Calculations were performed using Stephanie C. Frankle reference benchmark model received from Denise Neudecker (LANL). The calculated neutron spectra with a reference model were compared with the experimental values for different experimental configurations and various deviations were observed. Due to the inconsistent results of different detectors the possible reasons for discrepancies (e.g. flight path length, detector angle, etc.) were evaluated. It was concluded that the uncertainty in the flight path length and detector angle have negligible effect on the detected neutron spectra, however the effect of not explicitly modelling experimental geometry can not be neglected. To evaluate the bias due to the not explicitly modelled experimental geometry a detailed computational model with Monte Carlo neutron transport code MCNP was made. In addition, it was found out that the neutron source included in the MCNP model does not take into account the additional neutron peak at 2.8 MeV and the calculated results are considered to be less reliable below 4 MeV.

This paper presents the experimental results of neutron spectra measurements with organic scintillators at the copper beam dump of the 100-MeV proton linear accelerator in the Korea Multi-Purpose Accelerator Complex (KOMAC). For the measurement of the fast neutrons generated by the proton-copper nuclear reaction, two organic scintillators made of a stilbene crystal and liquid EJ-309 combined with photomultipliers and fast digitizers were utilized. Both detectors were installed inside radiation shielding structures facing the conical-shaped 100-MeV proton beam dump. The neutron transport was calculated using two Monte-Carlo codes, MCNP6 and GEANT4, with a proper physical model and a nuclear data library, along with the appropriate consideration of the detailed geometrical factors in the vicinity of the beam-dump and the neutron detectors. Each pulse signal measured in the detectors was analyzed to identify whether it was induced by a neutron or a gamma-ray in a mixed radiation field by using pulse shape discrimination technique. For operation of beam current of 5 mA and duty of 0.2%, the neutron yield was measured to be approximately 9.0E+10 n/s, which was 23% smaller than the numerically estimated value. In addition, both the measured and the estimated neutron energy spectra were compared.