scholarly journals SUPER-HISTORY METHODS FOR ADJOINT-WEIGHTED TALLIES IN MONTE CARLO TIME EIGENVALUE CALCULATIONS

2021 ◽  
Vol 247 ◽  
pp. 04008
Author(s):  
F. Filiciotto ◽  
A. Jinaphanh ◽  
A. Zoia
Keyword(s):  
Monte Carlo ◽  
Eigenvalue Problems ◽  
Neutron Transport ◽  
Computation Time ◽  
Nuclear Data ◽  
Analytic Solutions ◽  
Benchmark Problems ◽  
Probability Method ◽  
Fission Probability ◽  
Transport Problems

Time eigenvalues emerge in several key applications related to neutron transport problems, including reactor start-up and reactivity measurements. In this context, experimental validation and uncertainty quantification would demand to assess the variation of the dominant time eigenvalue in response to a variation of nuclear data. Recently, we proposed the use of a Generalized Iterated Fission Probability method (G-IFP) to compute adjoint-weighted tallies, such as kinetic parameters, perturbations and sensitivity coefficients, for Monte Carlo time (or alpha) eigenvalue calculations. With the massive use of parallel Monte Carlo calculations, it would be therefore useful to trade the memory burden of the G-IFP method (which is comparable to that of the standard IFP method for k-eigenvalue problems) for computation time and to rely on history-based schemes for such adjoint-weighted tallies. For this purpose, we investigate the use of the super-history method as applied to estimating adjoint-weighted tallies within the α-k power iteration, based on previous work on k-eigenvalue problems. Verification of the algorithms is performed on some simple preliminary tests where analytic solutions exist. In addition, the performances of the proposed method are assessed by comparing the super-history and the G-IFP methods for the same sets of benchmark problems.

scholarly journals Uncertainty propagation based on correlated sampling technique for nuclear data applications

2020 ◽  
Vol 6 ◽  
pp. 8 ◽  
Author(s):  
Axel Laureau ◽  
Vincent Lamirand ◽  
Dimitri Rochman ◽  
Andreas Pautz
Keyword(s):  
Monte Carlo ◽  
Cross Section ◽  
Neutron Transport ◽  
Uncertainty Propagation ◽  
Computation Time ◽  
Sampling Technique ◽  
Nuclear Data ◽  
Correlated Sampling ◽  
Enriched Uranium ◽  
The Impact

A correlated sampling technique has been implemented to estimate the impact of cross section modifications on the neutron transport and in Monte Carlo simulations in one single calculation. This implementation has been coupled to a Total Monte Carlo approach which consists in propagating nuclear data uncertainties with random cross section files. The TMC-CS (Total Monte Carlo with Correlated Sampling) approach offers an interesting speed-up of the associated computation time. This methodology is detailed in this paper, together with two application cases to validate and illustrate the gain provided by this technique: the highly enriched uranium/iron metal core reflected by a stainless-steel reflector HMI-001 benchmark, and the PETALE experimental programme in the CROCUS zero-power light water reactor.

scholarly journals DIAGNOSIS OF THE UNRESOLVED DOMAIN TREATMENT IN MONTE CARLO TRANSPORT CALCULATIONS THROUGH THE IDENTIFICATION AND MODELLING OF CRITICALITY SAFETY EXPERIMENTS

2021 ◽  
Vol 247 ◽  
pp. 10003
Author(s):  
N. García-Herranz ◽  
J. Rodríguez ◽  
A. Jiménez-Carrascosa ◽  
O. Cabellos
Keyword(s):  
Monte Carlo ◽  
Neutron Transport ◽  
Resonance Region ◽  
Nuclear Data ◽  
High Fidelity ◽  
Continuous Energy ◽  
Critical Facilities ◽  
Nuclear Systems ◽  
Monte Carlo Transport ◽  
Criticality Safety

Monte Carlo neutron transport codes can be used for high-fidelity predictions of the performance of nuclear systems. However, validation against experiments is required in order to establish the credibility in the results and identify the inaccuracies due to the used calculation scheme and associated databases. The International Handbook of Evaluated Criticality Safety Benchmark Experiments (ICSBEP) contains criticality safety benchmarks derived from experiments that have been performed at various nuclear critical facilities around the world and are very valuable for validation purposes. The main objective of this work is the identification and modelling of experimental benchmarks included at ICSBEP in support of the validation of Monte Carlo neutron transport calculations when applied to fast systems, and in particular, KENO-VI and associated AMPX-formatted continuous-energy libraries from SCALE package. In such systems, the predicted k-eff values can be very sensitive to the treatment of nuclear data in the Unresolved Resonance Region (URR). Consequently, benchmarks with intermediate and fast spectra are identified and modelled with KENO-VI. Then, calculated results with and without probability tables in the URR are compared with each other in order to identify the most sensitive configurations to the URR. As a result of the proposed study, recommendations are given about the benchmarks that should be modelled and analysed to qualify the processed continuous-energy libraries before their use in Monte Carlo transport codes for practical fast reactor applications.

scholarly journals INVESTIGATION ON DETERMINISTIC TRUNCATION TO CONTINUOUS ENERGY MONTE CARLO NEUTRON TRANSPORT CALCULATION

2021 ◽  
Vol 247 ◽  
pp. 04023
Author(s):  
Inhyung Kim ◽  
Yonghee Kim
Keyword(s):  
Monte Carlo ◽  
Figure Of Merit ◽  
Solution Method ◽  
Energy Transport ◽  
Neutron Transport ◽  
Computing Time ◽  
Computational Cost ◽  
Benchmark Problems ◽  
Continuous Energy ◽  
Transport Calculation

This paper presents the application and evaluation of a deterministic truncation of Monte Carlo (DTMC) solution method in a whole core reactor problem based on a continuous energy transport calculation. The DTMC method has been studied and developed as a systematic way to truncate the high-fidelity Monte Carlo (MC) solution to reduce the computational cost without compromising the essential reliability of the solution. Its fea-sibility and capability were preliminarily validated in several benchmark problems using a multi-group energy MC code. In this paper, further study has been conducted in the more practical application with the continuous-energy based MC calculation. The con-cept of the DTMC method is briefly described. Improvements to enhance the numerical stability and efficiency are specified in details. The DTMC method is applied to an SMR problem, in which reactor parameters are estimated to characterize the numerical per-formance and are compared to the standard MC method. Last, the computing time and corresponding figure-of-merit are evaluated.

scholarly journals Monte Carlo Alpha Iteration Algorithm for a Subcritical System Analysis

2015 ◽  
Vol 2015 ◽  
pp. 1-7 ◽  
Author(s):  
Hyung Jin Shim ◽  
Sang Hoon Jang ◽  
Soo Min Kang
Keyword(s):  
Monte Carlo ◽  
Iteration Method ◽  
System Analysis ◽  
Neutron Transport ◽  
Analytic Solutions ◽  
Source Distribution ◽  
Iteration Algorithm ◽  
Driven System ◽  
Accelerator Driven System ◽  
Subcritical System

Theα-kiteration method which searches the fundamental mode alpha-eigenvalue via iterative updates of the fission source distribution has been successfully used for the Monte Carlo (MC) alpha-static calculations of supercritical systems. However, theα-kiteration method for the deep subcritical system analysis suffers from a gigantic number of neutron generations or a huge neutron weight, which leads to an abnormal termination of the MC calculations. In order to stably estimate the prompt neutron decay constant (α) of prompt subcritical systems regardless of subcriticality, we propose a new MC alpha-static calculation method named as theαiteration algorithm. The new method is derived by directly applying the power method for theα-mode eigenvalue equation and its calculation stability is achieved by controlling the number of time source neutrons which are generated in proportion toαdivided by neutron speed in MC neutron transport simulations. The effectiveness of theαiteration algorithm is demonstrated for two-group homogeneous problems with varying the subcriticality by comparisons with analytic solutions. The applicability of the proposed method is evaluated for an experimental benchmark of the thorium-loaded accelerator-driven system.

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