Comparison among MCNP-based depletion codes applied to burnup calculations of pebble-bed HTR lattices

The double-heterogeneity characterising pebble-bed high temperature reactors (HTRs) makes Monte Carlo based calculation tools the most suitable for detailed core analyses. These codes can be successfully used to predict the isotopic evolution during irradiation of the fuel of this kind of cores. At the moment, there are many computational systems based on MCNP that are available for performing depletion calculation. All these systems use MCNP to supply problem dependent fluxes and/or microscopic cross sections to the depletion module. This latter then calculates the isotopic evolution of the fuel resolving Bateman's equations. In this paper, a comparative analysis of three different MCNP-based depletion codes is performed: Montburns2.0, MCNPX2.6.0 and BGCore. Monteburns code can be considered as the reference code for HTR calculations, since it has been already verified during HTR-N and HTR-N1 EU project. All calculations have been performed on a reference model representing an infinite lattice of thorium-plutonium fuelled pebbles. The evolution of k-inf as a function of burnup has been compared, as well as the inventory of the important actinides. The k-inf comparison among the codes shows a good agreement during the entire burnup history with the maximum difference lower than 1%. The actinide inventory prediction agrees well. However significant discrepancy in Am and Cm concentrations calculated by MCNPX as compared to those of Monteburns and BGCore has been observed. This is mainly due to different Am-241 (n,γ) branching ratio utilized by the codes. The important advantage of BGCore is its significantly lower execution time required to perform considered depletion calculations. While providing reasonably accurate results BGCore runs depletion problem about two times faster than Monteburns and two to five times faster than MCNPX. © 2009 Elsevier B.V. All rights reserved.

Numerical analysis of an incremented statistical sampling procedure in MCNP

MCNP has stood so far as one of the main Monte Carlo radiation transport codes. Its use, as any other Monte Carlo based code, has increased as computers perform calculations faster and become more affordable along time. However, the use of Monte Carlo method to tally events in volumes which represent a small fraction of the whole system may turn to be unfeasible, if a straight analogue transport procedure (no use of variance reduction techniques) is employed and precise results are demanded. Calculations of reaction rates in activation foils placed in critical systems turn to be one of the mentioned cases. The present work takes advantage of the fixed source representation from MCNP to perform the above mentioned task in a more effective sampling way (characterizing neutron population in the vicinity of the tallying region and using it in a geometric reduced coupled simulation). An extended analysis of source dependent parameters is studied in order to understand their influence on simulation performance and on validity of results. Although discrepant results have been observed for small enveloping regions, the procedure presents itself as very efficient, giving adequate and precise results in shorter times than the standard analogue procedure. (C) 2007 Elsevier Ltd. All rights reserved.

Análise de algoritmos computacionais utilizados em sistemas de planejamento de braquiterapia de alta taxa de dose: estudo comparativo com o Código MCNP-5C

In radiotherapy, computational systems are used for radiation dose determination in the treatment’s volume and radiometric parameters quality analysis of equipment and field irradiated. Due to the increasing technological advancement, several research has been performed in brachytherapy for different computational algorithms development which may be incorporated to treatment planning systems, providing greater accuracy and confidence in the dose calculation. Informatics and information technology fields undergo constant updating and refinement, allowing the use Monte Carlo Method to simulate brachytherapy source dose distribution. The methodology formalization employed to dosimetric analysis is based mainly in the American Association of Physicists in Medicine (AAPM) studies, by Task Group nº 43 (TG-43) and protocols aimed at dosimetry of these radiation sources types. This work aims to analyze the feasibility of using the MCNP-5C (Monte Carlo N-Particle) code to obtain radiometric parameters of brachytherapy sources and so to study the radiation dose variation in the treatment planning. Simulations were performed for the radiation dose variation in the source plan and determined the dosimetric parameters required by TG-43 formalism for the characterization of the two high dose rate iridium-192 sources. The calculated values were compared with the presents in the literature, which were obtained with different Monte Carlo simulations codes. The results showed excellent consistency with the compared codes, enhancing MCNP-5C code the capacity and viability in the sources dosimetry employed in HDR brachytherapy. The method employed may suggest a possible incorporation of this code in the treatment planning systems provided by manufactures together with the equipment, since besides reducing acquisition cost, it can also make the used computational routines more comprehensive, facilitating the brachytherapy ...

Análise dosimétrica de perfis de distribuições radiais de doses relativas de um aplicador de betaterapia de 90Sr+90Y utilizando o código MCNP-4C e filmes raiocrômicos

Although they are no longer manufactured, the applicators of 90Sr +90Y acquired in the decades of 1990 are still in use, by having half-life of 28.5 years. These applicators have calibration certificate given by their manufacturers, where few have been recalibrated. Thus it becomes necessary to accomplish thorough dosimetry of these applicators. This paper presents a dosimetric analysis distribution radial dose profiles for emitted by an 90Sr+90Y betatherapy applicator, using the MCNP-4C code to simulate the distribution radial dose profiles and radiochromium films to get them experimentally . The results with the simulated values were compared with the results of experimental measurements, where both curves show similar behavior, which may validate the use of MCNP-4C and radiochromium films for this type of dosimetry

Flow Regime Identification Methodology with Mcnp-X Code And Artificial Neural Network

This paper presents flow regimes identification methodology in multiphase system in annular, stratified and homogeneous oil-water-gas regimes. The principle is based on recognition of the pulse height distributions (PHD) from gamma-ray with supervised artificial neural network (ANN) systems. The detection geometry simulation comprises of two NaI(Tl) detectors and a dual-energy gamma-ray source. The measurement of scattered radiation enables the dual modality densitometry (DMD) measurement principle to be explored. Its basic principle is to combine the measurement of scattered and transmitted radiation in order to acquire information about the different flow regimes. The PHDs obtained by the detectors were used as input to ANN. The data sets required for training and testing the ANN were generated by the MCNP-X code from static and ideal theoretical models of multiphase systems. The ANN correctly identified the three different flow regimes for all data set evaluated. The results presented show that PHDs examined by ANN may be applied in the successfully flow regime identification.

Desenvolvimento de um simulador para espectrometria por fluorescência de raios X usando computação distribuída

A Física das Radiações é um ramo da Física que está presente em diversas áreas de estudo e se relaciona ao conceito de espectrometria. Dentre as inúmeras técnicas espectrométricas existentes, destaca-se a espectrometria por fluorescência de raios X. Esta também possui uma gama de variações da qual pode-se dar ênfase a um determinado subconjunto de técnicas. A produção de fluorescência de raios X permite (em certos casos) a análise das propriedades físico-químicas de uma amostra específica, possibilitando a determinação de sua constituiçõa química e abrindo um leque de aplicações. Porém, o estudo experimental pode exigir uma grande carga de trabalho, tanto em termos do aparato físico quanto em relação conhecimento técnico. Assim, a técnica de simulação entra em cena como um caminho viável, entre a teoria e a experimentação. Através do método de Monte Carlo, que se utiliza da manipulação de números aleatórios, a simulação se mostra como uma espécie de alternativa ao trabalho experimental.Ela desenvolve este papel por meio de um processo de modelagem, dentro de um ambiente seguro e livre de riscos. E ainda pode contar com a computação de alto desempenho, de forma a otimizar todo o trabalho por meio da arquitetura distribuída. O objetivo central deste trabalho é a elaboração de um simulador computacional para análise e estudo de sistemas de fluorescência de raios X desenvolvido numa plataforma de computação distribuída de forma nativa com o intuito de gerar dados otimizados. Como resultados deste trabalho, mostra-se a viabilidade da construção do simulador através da linguagem CHARM++, uma linguagem baseada em C++ que incorpora rotinas para processamento distribuído, o valor da metodologia para a modelagem de sistemas e a aplicação desta na construção de um simulador para espectrometria por fluorescência de raios X. O simulador foi construído com a capacidade de reproduzir uma fonte de radiação eletromagnética, amostras complexas e um conjunto de detectores. A modelagem dos detectores incorpora a capacidade de geração de imagens baseadas nas contagens registradas. Para validação do simulador, comparou-se os resultados espectrométricos com os resultados gerados por outro simulador já validado: o MCNP.

Core design options for high conversion BWRs operating in Th- 233U fuel cycle

Several options of fuel assembly design are investigated for a BWR core operating in a closed self-sustainable Th-233U fuel cycle. The designs rely on an axially heterogeneous fuel assembly structure consisting of a single axial fissile zone "sandwiched" between two fertile blanket zones, in order to improve fertile to fissile conversion ratio. The main objective of the study was to identify the most promising assembly design parameters, dimensions of fissile and fertile zones, for achieving net breeding of 233U. The design challenge, in this respect, is that the fuel breeding potential is at odds with axial power peaking and the core minimum critical power ratio (CPR), hence limiting the maximum achievable core power rating. Calculations were performed with the BGCore system, which consists of the MCNP code coupled with fuel depletion and thermo-hydraulic feedback modules. A single 3-dimensional fuel assembly having reflective radial boundaries was modeled applying simplified restrictions on the maximum centerline fuel temperature and the CPR. It was found that axially heterogeneous fuel assembly design with a single fissile zone can potentially achieve net breeding, while matching conventional BWR core power rating under certain restrictions to the core loading pattern design. © 2013 Elsevier B.V. All rights reserved.

BWR fuel design options for self-sustainable Th-233U fuel cycle

In this work, we investigate a number of fuel assembly design options for a BWR core operating in a closed self-sustainable Th-233U fuel cycle. The designs rely on axially heterogeneous fuel assembly structure in order to improve fertile to fissile conversion ratio. One of the main assumptions of the current study was to restrict the fuel assembly geometry to a single axial fissile zone "sandwiched" between two fertile blanket zones. The main objective was to study the effect of the most important design parameters, such as dimensions of fissile and fertile zones and average void fraction, on the net breeding of 233U. The main design challenge in this respect is that the fuel breeding potential is at odds with axial power peaking and therefore limits the maximum achievable core power rating. The calculations were performed with BGCore system, which consists of MCNP code coupled with fuel depletion and thermo-hydraulic feedback modules. A single 3-dimensional fuel assembly with reflective radial boundaries was modeled applying simplified restrictions on maximum central line fuel temperature and Critical Power Ratio. It was found that axially heterogeneous fuel assembly design with single fissile zone can potentially achieve net breeding. In this case however, the achievable core power density is roughly one third of the reference BWR core.

Coupled neutronic thermo-hydraulic analysis of full PWR core with Monte-Carlo based BGCore system

BGCore reactor analysis system was recently developed at Ben-Gurion University for calculating in-core fuel composition and spent fuel emissions following discharge. It couples the Monte Carlo transport code MCNP with an independently developed burnup and decay module SARAF. Most of the existing MCNP based depletion codes (e.g. MOCUP, Monteburns, MCODE) tally directly the one-group fluxes and reaction rates in order to prepare one-group cross sections necessary for the fuel depletion analysis. BGCore, on the other hand, uses a multi-group (MG) approach for generation of one group cross-sections. This coupling approach significantly reduces the code execution time without compromising the accuracy of the results. Substantial reduction in the BGCore code execution time allows consideration of problems with much higher degree of complexity, such as introduction of thermal hydraulic (TH) feedback into the calculation scheme. Recently, a simplified TH feedback module, THERMO, was developed and integrated into the BGCore system. To demonstrate the capabilities of the upgraded BGCore system, a coupled neutronic TH analysis of a full PWR core was performed. The BGCore results were compared with those of the state of the art 3D deterministic nodal diffusion code DYN3D (Grundmann et al.; 2000). Very good agreement in major core operational parameters including k-eff eigenvalue, axial and radial power profiles, and temperature distributions between the BGCore and DYN3D results was observed. This agreement confirms the consistency of the implementation of the TH feedback module. Although the upgraded BGCore system is capable of performing both, depletion and TH analyses, the calculations in this study were performed for the beginning of cycle state with pre-generated fuel compositions. © 2011 Published by Elsevier B.V.

HTGR fuel element depletion benchmark: Stage three results

Recently, a new numerical benchmark exercise for High Temperature Gas Cooled Reactor (HTGR) fuel depletion was defined. The purpose of this benchmark is to provide a comparison basis for different codes and methods applied to the burnup analysis of HTGRs. The benchmark specifications include three different models: (1) an infinite lattice of tristructural isotropic (TRISO) fuel particles, (2) an infinite lattice of fuel pebbles, and (3) a prismatic fuel including fuel and coolant channels. In this paper, we present the results of the third stage of the benchmark obtained with MCNP based depletion code BGCore and deterministic lattice code HELIOS 1.9. The depletion calculations were performed for three-dimensional model of prismatic fuel with explicitly described TRISO particles as well as for two-dimensional model, in which double heterogeneity of the TRISO particles was eliminated using reactivity equivalent physical transformation (RPT). Generally, good agreement in the results of the calculations obtained using different methods and codes was observed.

HTGR fuel element depletion benchmark: Stage one results

In this paper, we reported the results of the first stage of HTGR fuel element depletion benchmark obtained with BGCore and HELIOS depletion codes. The results of the k-inf are generally in good agreement. However, significant deviation in concentrations of several nuclides between MCNP based and HELIOS codes was observed.

Efficient generation of one-group cross sections for coupled Monte Carlo depletion calculations

Coupled Monte Carlo depletion systems provide a versatile and an accurate tool for analyzing advanced thermal and fast reactor designs for a variety of fuel compositions and geometries. The main drawback of Monte Carlo-based systems is a long calculation time imposing significant restrictions on the complexity and amount of design-oriented calculations. This paper presents an alternative approach to interfacing the Monte Carlo and depletion modules aimed at addressing this problem. The main idea is to calculate the one-group cross sections for all relevant isotopes required by the depletion module in a separate module external to Monte Carlo calculations. Thus, the Monte Carlo module will produce the criticality and neutron spectrum only, without tallying of the individual isotope reaction rates. The onegroup cross section for all isotopes will be generated in a separate module by collapsing a universal multigroup (MG) cross-section library using the Monte Carlo calculated flux. Here, the term "universal" means that a single MG cross-section set will be applicable for all reactor systems and is independent of reactor characteristics such as a neutron spectrum; fuel composition; and fuel cell, assembly, and core geometries. This approach was originally proposed by Haeck et al. and implemented in the ALEPH code. Implementation of the proposed approach to Monte Carlo burnup interfacing was carried out through the BGCORE system. One-group cross sections generated by the BGCORE system were compared with those tallied directly by the MCNP code. Analysis of this comparison was carried out and led to the conclusion that in order to achieve the accuracy required for a reliable core and fuel cycle analysis, accounting for the background cross section (σ0) in the unresolved resonance energy region is essential. An extension of the one-group cross-section generation model was implemented and tested by tabulating and interpolating by a simplified σ0 model. A significant improvement of the one-group cross-section accuracy was demonstrated.

Light water breeding with nitride fuel

This scoping study proposes using mixed nitride fuel in Pu-based high conversion LWR designs in order to increase the breeding ratio. The higher density fuel reduces the hydrogen-to-heavy metal ratio in the reactor which results in a harder spectrum in which breeding is more effective. A Resource-renewable Boiling Water Reactor (RBWR) assembly was modeled in MCNP to demonstrate this effect in a typical high conversion LWR design. It was determined that changing the fuel from (U,TRU)O2 to (U,TRU)N in the assembly can increase its fissile inventory ratio (fissile Pu mass divided by initial fissile Pu mass) from 1.04 to up to 1.17. © 2011 Elsevier Ltd. All rights reserved.

Avaliação teórico-experimental da distribuição de dose absorvida em radioterapia em tumores de pele

O trabalho propõe rotinas computacionais usando o Método de Monte Carlo com o Código MCNP-5, para analisar os perfis de dose de radiação liberada nos tratamentos de tumores de pele e otimizar os cálculos radiométricos dos feixes de radiação estudados. Foram realizadas medidas dosimétricas do feixe de radiação, comparando os resultados obtidos com os respectivos valores fornecidos pelo serviço de física médica das instituições, com resultados informados pelo fabricante do equipamento e com as simulações computacionais efetuadas com o Código MCNP-5. A quantificação dos erros relativos percentual entre os resultados simulados e os fornecidos pelo Serviço de Radioterapia (E1), os informados pelo fabricante (E2) e os medidos experimentalmente (E3) são inferiores a 4,0% e validam a metodologia computacional proposta para avaliação do comportamento do feixe de raios-X superficial e do feixe de raios γ da unidade de Cobaltoterapia. A metodologia de análise do espectro energético e da curva de porcentagem de dose profunda (PDP) desenvolvida neste trabalho pode ser estendida para estudos de outros feixes clínicos e subsidiar os dados radiométricos utilizados nos planejamentos e cálculos de dose realizados pelo profissional da física médica na sua rotina nos Serviços de Radioterapia