Alireza Haghighat

A Three-Dimensional Block-Oriented Hybrid Discrete Ordinates and Characteristics Method

Abstract In this paper, we present a hybrid formulation/algorithm to solve the linear Boltzmann equation, specifically for application to problems containing regions of low scattering. The hybrid approach uses the characteristics method in low scattering regions, while the remaining regions are treated with the discrete ordinates method (SN). A shared scattering kernel allows an arbitrary order of anisotropic scattering in both block-oriented solvers. A new three-dimensional transport code (TITAN) has been developed based on the hybrid approach. TITAN divides a problem model into coarse meshes (blocks) in the Cartesian geometry. The block-oriented structure allows different fine-meshing schemes (or characteristic ray densities) and angular quadrature sets for different coarse meshes. Angular and spatial projection techniques are developed to transfer angular fluxes on the interfaces of the coarse meshes. We have tested the performance and accuracy of the new hybrid algorithm within the TITAN code for a number of benchmark problems. The results of a computed tomography model and the Kobayashi benchmark problems are presented in this paper. It is demonstrated that while preserving high-level accuracy as compared to reference Monte Carlo simulations, the hybrid algorithm achieves significant computation efficiency as compared to the SN method only.

Monte Carlo transport calculations and analysis for reactor pressure vessel neutron fluence

The application of Monte Carlo methods for reactor pressure vessel (RPV) neutron fluence calculations is examined. As many commercial nuclear light water reactors approach the end of their design lifetime, it is of great consequence that reactor operators and regulators be able to characterize the structural integrity of the RPV accurately for financial reasons, as well as safety reasons, due to the possibility of plant life extensions. The Monte Carlo method, which offers explicit three-dimensional geometric representation and continuous energy and angular simulation, is well suited for this task. A model of the Three Mile Island unit 1 reactor is presented for determination of RPV fluence; Monte Carlo (MCNP) and deterministic (DORT) results are compared for this application; and numerous issues related to performing these calculations are examined. Synthesized three-dimensional deterministic models are observed to produce results that are comparable to those of Monte Carlo methods, provided the two methods utilize the same cross-section libraries. Continuous energy Monte Carlo methods are shown to predict more (15 to 20%) high-energy neutrons in the RPV than deterministic methods.

Angular parallelization of a curvilinear S sub n transport theory method

In this paper a parallel algorithm for angular domain decomposition (or parallelization) of an r-dependent spherical S{sub n} transport theory method is derived. The parallel formulation is incorporated into TWOTRAN-II using the IBM Parallel Fortran compiler and implemented on an IBM 3090/400 (with four processors). The behavior of the parallel algorithm for different physical problems is studied, and it is concluded that the parallel algorithm behaves differently in the presence of a fission source as opposed to the absence of a fission source; this is attributed to the relative contributions of the source and the angular redistribution terms in the S{sub s} algorithm. Further, the parallel performance of the algorithm is measured for various problem sizes and different combinations of angular subdomains or processors. Poor parallel efficiencies between {approximately}35 and 50% are achieved in situations where the relative difference of parallel to serial iterations is {approximately}50%. High parallel efficiencies between {approximately}60% and 90% are obtained in situations where the relative difference of parallel to serial iterations is {lt}35%.

Evaluation of the uncertainties in the source distribution for pressure vessel neutron fluence calculations

The methodology used to prepare the source for neutron fluence calculation at the reactor pressure vessel is examined, and its effect on the calculated cavity dosimeter reaction rate is evaluated. ...

Determination of the field of view of a boiling water reactor in-core detector

The field of view of a boiling water reactor (BWR) in-core detector is evaluated via a combined R-Z transport and X-Y-Z diffusion theory model. A cell homogeneous X-Y-Z diffusion theory model is sufficient for the evaluation of the adjoint function distribution around an in-core BWR detector. All types of subchannels (i.e., side, center, and corner) contribute to the detector signal fluctuations, and the side subchannels are the dominant contributor. Finally, the measured flow velocity via the cross-correlation method in a BWR is an averaged quantity rather than a localized quantity.

Iterative schemes for parallel Sn algorithms in a shared-memory computing environment

Several two-dimensional spatial domain partitioning S{sub n} transport theory algorithms are developed on the basis of different iterative schemes. These algorithms are incorporated into TWOTRAN-II and tested on the shared-memory CRAY Y-MP C90 computer. For a series of fixed-source r-z geometry homogeneous problems, it is demonstrated that the concurrent red-black algorithms may result in large parallel efficiencies (>60%) on C90. It is also demonstrated that for a realistic shielding problem, the use of the negative flux fixup causes high load imbalance, which results in a significant loss of parallel efficiency.

Novel Investigation of Iron Cross Sections via Spherical Shell Transmission Measurements and Particle Transport Calculations for Material Embrittlement Studies

Abstract We have carried out a multifaceted research project to improve our knowledge of the iron nonelastic scattering cross sections. Spherical shell transmission measurements were made using time-of-flight techniques with neutrons from the 15N(p,n)15O and D(d,n)3He source reactions. For the 15N(p,n)15O work, measurements were made with a proton energy of 5.1 MeV. Measurements were made from 3 to 7-MeV deuteron energy for the D(d,n)3He work. For both source reactions, the angular range was as large as 15 to 135 deg. Two shell thicknesses were used. Comparisons are given between Monte Carlo predictions and experimental data. Utilizing a new tallying option, the estimated total iron cross sections at energies corresponding to the peak of the spectra for the 0-deg experiments were calculated to within 1% of the data in the ENDF/B-VII library. A processing code was developed to adjust ENDF format files to obtain closer agreement between measurements and calculations. Sensitivity analyses were performed at energies corresponding to the 0-deg beam angle neutrons. Using cross sections where the nonelastic and elastic cross sections were adjusted while constraining the total cross section to be constant, differences between experiment and calculation were reduced by ~;40% for a pressure vessel calculation. Such fluence calculations with adjusted cross sections indicate possible underestimation of neutron fluence, and therefore more material damage.

Development of INSPCT-S for Inspection of Spent Fuel Pool

In this paper, we discuss an accurate and fast software tool (INSPCT-S, Inspection of Nuclear Spent fuel-Pool Calculation Tool, version Spreadsheet) developed for calculation of the response of fission chambers placed in a spent fuel pool, such as Atucha-I. INSPCT-S is developed for identification of suspicious regions of the pool that may have missing or substitute assemblies. INSPCT-S uses a hybrid algorithm based on the adjoint function methodology. The neutron source is comprised of spontaneous fission, ({alpha}, n) interactions, and subcritical multiplication. The former is evaluated using the ORIGEN-ARP code, and the latter is obtained with the fission matrix (FM) formulation. The FM coefficients are determined using the MCNP Monte Carlo code, and the importance function is determined using the PENTRAN 3-D parallel Sn code. Three databases for the neutron source, FM elements, and adjoint flux are prepared as functions of different parameters including burnup, cooling time, enrichment, and pool lattice size. INSPCT-S uses the aforementioned databases and systems of equations to calculate detector responses, which are subsequently compared with normalized experimental data. If this comparison is not satisfied, INSPCT-S utilizes color coding to identify the suspicious regions of a spent fuel pool. (authors)

Final Report - Novel investigation of iron cross sections via spherical shell transmission measurements and particle transport calculations for material embrittlement studies.

OAK B204 We have been pursuing a multi-year project, funded by the U.S. Department of Energy, to study neutron scattering interactions in iron. The principal objective of this work is to investigate the well-known deficiency that exists for reactor pressure vessel neutron fluence determinations. Specifically, we are using the spherical-shell transmission method, employing iron shells with different thicknesses, and neutron time-of-flight (TOF) measurements of the scattered neutrons, in an effort to precisely determine specific energy regions over which deficiencies in the non-elastic scattering cross section for neutron scattering in iron appear to exist.

The RAPID Fission Matrix Approach to Reactor Core Criticality Calculations

Abstract The Real-time Analysis for Particle transport and In-situ Detection (RAPID) code uses a unique, extremely fast, fission matrix–based methodology to compute the eigenvalue, and three-dimensional, pinwise fission source distribution for reactor, spent fuel pool, and spent fuel cask problems. In this paper, the RAPID fission matrix method is described and analyzed for application to several large pressurized water reactor problems, based on the Organisation for Economic Co-operation and Development/Nuclear Energy Agency Monte Carlo Performance Benchmark problem. In the RAPID methodology, fission matrix coefficients precalculated using the Serpent Monte Carlo code, are then coupled together and solved for different core arrangements. A boundary correction method is used to obtain more accurate fission matrix values near the radial and axial reflectors. Eigenvalues and fission source distributions are compared between RAPID and Serpent reference calculations. In most cases, the eigenvalue differences between methods are less than 10 pcm. For a uniform core model, pinwise fission distributions between the methods differ by a root-mean-square value of , compared to a Serpent uncertainty of . The pinwise, axially dependent (100 axial levels) differences are , compared to a similar Serpent uncertainty of . To achieve these levels of uncertainty, the RAPID calculations are over 2500 times faster than Serpent, not counting the precalculation time.