Monte Carlo Codes for Neutron Buildup Factors

The point-kernel method is a widely used practical tool for gamma-ray shielding calculations. However, application of that method for neutron transport simulations is very limited. The accuracy of the method strongly depends on the accuracy of buildup factors used in the calculations. Buildup factors are usually obtained using appropriate computer codes, either based on discrete ordinates transport method or Monte Carlo approach. Since these codes put strong demands on computer resources, they are applied on a limited number of shielding configurations and an attempt is made to use these results and formulate an empirical expression for buildup factors estimation. Due to high physical complexity of neutron transport through shielding material it is very hard to perform parameterisation in order to establish adequate empirical formula. Existing formulas are very limited and are usually applicable to a narrow neutron energy range for few commonly used shielding materials, mostly in monolayer configuration. Recently, a new approach has been proposed for determination of gamma ray buildup factors for mono-layer, as well as multi-layer shielding configurations covering a wide gamma ray energy range. The new regression model is based on support vector machines learning technique, which has theoretical background in statistical learning theory. Development of named regression model required a large number of experimental data obtained by Monte Carlo computer code. More than 7000 Monte Carlo runs were required. Due to physical complexity neutron transport is likely to require even more experimental data in order to generate a model of reasonable accuracy. Therefore, the choice of appropriate Monte Carlo code is a very important question. One has to take into account the accuracy as well as the time required for input preparation and running the code. What also has to be considered is the possibility of the code to be incorporated in an algorithm for automated generation of experimental data. In this paper three Monte Carlo codes are analysed, namely SCALE4.4 code package (SAS3 sequence), SCALE6.0 code package (MAVRIC sequence), and MCNP5. Two simple experimental setups based on a point isotropic source in spherical and slab-like shield are modelled, and the codes are examined on previously mentioned issues. The comparison results show that each one of the examined codes has potential to be used for neutron buildup factor model generation. However, some aspects of their utilization require further analysis prior to final selection

Characterization of the GBC-32 Fuel Assembly Source Terms

This paper presents burnup/depletion calculations of the typical Westinghouse 17x17 fuel assembly to be used as a radioactive waste package in a Generic Burnup Credit cask benchmark problem with 32 elements (GBC-32). This first phase is addressing spent fuel source terms calculation while evaluation of the shielding performance of the GBC-32 cask is planned for the second phase. The TRITON-NEWT methodology of the SCALE6.1.3 program package was used in a tandem with ORIGEN-S code for deterministic 2D calculation of the GBC-32 fuel assembly neutron multiplication factor, providing spatial-temporal fluxes and isotopic concentration change. The burnup simulation was done up to 60 GWd/tU with sensitivity analysis of relevant physical parameters influenced by the working cross-section library. This approach also allowed generation of the specific user-defined collapsed cross-section libraries as a function of fuel enrichment and burnup level. Calculation of isotopic concentrations, decay heat, neutron-gamma spectra and major actinides activity for different fuel assembly cooling periods was performed using ORIGEN-ARP module

Spent Fuel Pool Dose Rate Calculations Using Point Kernel and Hybrid Deterministic-Stochastic Shielding Methods

This paper presents comparison of the Krško Power Plant simplified Spent Fuel Pool (SFP) dose rates using different computational shielding methodologies. The analysis was performed to estimate limiting gamma dose rates on wall mounted level instrumentation in case of significant loss of cooling water. The SFP was represented with simple homogenized cylinders (point kernel and Monte Carlo (MC)) or cuboids (MC) using uranium, iron, water, and dry-air as a bulk region materials. The pool is divided on the old and new section where the old one has three additional subsections representing fuel assemblies (FAs) with different burnup/cooling time (60 days, 1 year and 5 years). The new section represents the FAs with the cooling time of 10 years. The time dependent fuel assembly isotopic composition was calculated using ORIGEN2 code applied to the depletion of one of the fuel assemblies present in the pool (AC-29). The source used in Microshield calculation is based on imported isotopic activities. The time dependent photon spectrum with total source intensity from Microshield multigroup point kernel calculations was then prepared for two hybrid deterministic-stochastic sequences. One is based on SCALE6.2b3/MAVRIC (Monaco and Denovo) methodology and another uses Monte Carlo code MCNP6.1.1b and ADVANTG3.0.1. code. Even though this model is a fairly simple one, the layers of shielding materials are thick enough to pose a significant shielding problem for MC method without the use of effective variance reduction (VR) technique. For that purpose the ADVANTG code was used to generate VR parameters for the MCNP fixed-source calculation using continuous energy transport. ADVATNG employs a deterministic forward-adjoint transport solver Denovo which implements CADIS/FWCADIS methodology. Denovo uses a structured, Cartesian-grid SN solver based on the Koch- Baker-Alcouffe parallel transport sweep algorithm across x-y domain blocks. This was our first application of ANDVANTG/MCNP hybrid sequence for this type of calculation and the results where compared to SCALE/MAVRIC sequence which we regularly use for similar calculations. The comparison of gamma dose rates on different point detector locations (central above pool and at the top of pool periphery) showed a good agreement between Microshield (point-kernel) and deterministic-stochastic shielding methodologies for the cylindrical approximation of the pool geometry. More complicated cases for model with multi-source option and for cuboids showed very good agreement between SCALE/MAVRIC and ANDVANTG/MCNP calculations

Analysis of possibilities for linking land registers and other official registers in the Republic of Croatia based on LADM

AbstractWeak or non-existing linkage of official registers in the Republic of Croatia and the data redundancy as an inevitable outcome of such a state are the causes of various unwanted consequences for the relevant public authorities, as well as for citizens and companies as the end-users of that data.In this paper we present the results of an analysis of the status of the redundancy within the Croatian land administration-related registers. Following the analysis, suggestions are given on how the effectiveness of the analyzed registers can be increased by introducing a linking based on the Land Administration Domain Model (LADM). The proposed linkages were created by extending the Unified Modelling Language (UML) object classes of the LADM. The compliance analysis between registers and the LADM was conducted by using schema matching. Schema matching is a set of techniques used for comparing schemas (usually referred to as data models), and is well known within the database research domain. The results of the analysis were used to determine in which direction to go with extending of the LADM.All of the outputs of this research can be used to create a strategy for improving the effectiveness of the overall system of registers, which in turn should result in an overall economic progress of the country

Solinski svećenici i Sokolsko društvo – govor pri blagoslovu Sokolane i zastave

Turbulentna povijest sokolstva ugniježđena u previranjima između hrvatskoga nacionalnog društva i jugoslavenskoga unitarističkog pokreta tema je nebrojenih radova posvećenih sokolskomu pitanju. Solinska pak epizoda u svojoj cjelini iščekuje temeljito predanje istraživanju i nije temom ovoga rada, nego je riječ tek o njezinu jednom segmentu, povijesnoj epizodi koja se rađala i kulminirala otvaranjem i blagoslovom Sokolskoga doma (Sokolane) u Solinu i društvene zastave. Todobni je naime solinski župnik don Mate Mihanović unatoč izričitim preporukama i smjernicama mjesnoga Ordinarija prilikom svečanosti uz blagoslov održao govor političko-društveno kontroverznoga sadržaja. Objava njegova govora u dnevnom listu Novo doba, reakcija i pritužba vranjičkoga župnika don Ante Braškića kao i Mihanovićevo očitovanje biskupu daju novo svjetlo na pastoralni rad jednoga od najistaknutijih i najzaslužnijih solinskih župnika. Radi stjecanja šire slike i objektivnosti, odnos partikularne Crkve prema sokolskim društvima na širem solinskom području (Vranjic i Klis) dodatno je ilustriran drugim izvorima u kojima se uz aktualna društveno-politička pitanja »sokolske« provenijencije susreće mnoštvo drugih podataka o pastoralnoj aktivnosti, ćudorednom životu kao i unutarcrkvenim odnosima

Characterization of Fast Neutron Transmission Through an Iron Shield

In this paper we give an analysis of the neutron transmission through an iron sphere using Monte Carlo and transport theory methods based on ENDF/B-VII.1 general purpose library. The motivation for this investigation comes from a well-known deficiency in the iron inelastic data from the older library evaluation (ENDF/B-V), giving a concern for a fast neutron flux underestimation within the reactor pressure vessels. In order to benchmark the next-generation ENDF/B-VI iron data, the U.S. Nuclear Regulatory Commission and the former Czechoslovakian National Research Institute have jointly preformed several experiments in 1990s, addressing neutron leakage spectra obtained for a 252Cf fission source in a centre of an iron sphere. It was shown that the ENDF/B-VI iron cross section, containing several improvements over previous evaluations, will not entirely resolve the neutron spectrum discrepancies observed at high neutron energies. Since safety analyses of reactor pressure vessel embrittlement are often based on neutron transport calculations using specific multigroup cross section libraries, simulation of this benchmark was performed using a hybrid shielding methodology of ADVANTG3.0.3 and MCNP6.1.1b codes. Comparison of calculated and referenced dosimeter activation rates are presented for several "standard" nuclear reactions, often used in reactor pressure vessel dosimetry. For that purpose, the new IRDFF-II special library from the IAEA Nuclear Data Services was used as a reference source of dosimetry cross sections. The MCNP6.1.1b code was used for calculation of reaction rates, which were also compared with previous IRDFF-1.05 special library and general purpose ENDF/B-VII.1 library

Bibliografija Časopisa za solinske teme Tusculum (broj 1-10)

Tusculum donosi znanstvene i stručne priloge ponajprije iz humanističkoga područja, a otvoren je i srodnim područjima. Glavni su ciljevi časopisa objavljivanje rezultata znanstvenih istraživanja vezanih uz Solin, njegovu bogatu prošlost, sadašnjost pa i promišljanje budućnosti. Donosimo bibliografiju radova objavljenih od prvoga do desetoga broja, tj. od godine 2008. do 2017

Solinski svećenici i Sokolsko društvo – govor pri blagoslovu Sokolane i zastave

Turbulentna povijest sokolstva ugniježđena u previranjima između hrvatskoga nacionalnog društva i jugoslavenskoga unitarističkog pokreta tema je nebrojenih radova posvećenih sokolskomu pitanju. Solinska pak epizoda u svojoj cjelini iščekuje temeljito predanje istraživanju i nije temom ovoga rada, nego je riječ tek o njezinu jednom segmentu, povijesnoj epizodi koja se rađala i kulminirala otvaranjem i blagoslovom Sokolskoga doma (Sokolane) u Solinu i društvene zastave. Todobni je naime solinski župnik don Mate Mihanović unatoč izričitim preporukama i smjernicama mjesnoga Ordinarija prilikom svečanosti uz blagoslov održao govor političko-društveno kontroverznoga sadržaja. Objava njegova govora u dnevnom listu Novo doba, reakcija i pritužba vranjičkoga župnika don Ante Braškića kao i Mihanovićevo očitovanje biskupu daju novo svjetlo na pastoralni rad jednoga od najistaknutijih i najzaslužnijih solinskih župnika. Radi stjecanja šire slike i objektivnosti, odnos partikularne Crkve prema sokolskim društvima na širem solinskom području (Vranjic i Klis) dodatno je ilustriran drugim izvorima u kojima se uz aktualna društveno-politička pitanja »sokolske« provenijencije susreće mnoštvo drugih podataka o pastoralnoj aktivnosti, ćudorednom životu kao i unutarcrkvenim odnosima

Influence of Spacer Grids Homogenization on Core Reactivity and Axial Power Distribution

The paper presents the influence of spacer grid homogenization during cross section generation on core reactivity and axial power distribution. Homogenization calculation was performed at fuel assembly level using FA2D code. The first approach is to smear uniformly all centrally located spacer grids along 120 inches of fuel assembly and carry out 2D transport calculation. The second approach is to smear spacer grid within 6 inches of fuel assembly and perform homogenization calculation. That composition is then assigned to closest 6 in axial subdivision of the core calculation. The last analysed option is to do additional localization of spacer grids and carry out homogenization within 2 inches of fuel assembly height. The additional subdivision is afterward performed of the closest regular axial core subdivision in nodal core calculation. The core calculation was performed using modified PARCS 2.5 code for NPP Krško cycle 29. The normalized axial power distributions obtained by PARCS for three different ways of spacer grid homogenization are then compared to quantify the influence of modelling. Similar comparison was performed for critical boron concentration. As expected larger influence is present for axial power distribution (more details for fine localization), with some influence on axial power offset and global reactivity

Spent Fuel Pool Dose Rate Calculations Using Point Kernel and Hybrid Deterministic-Stochastic Shielding Methods

This paper presents comparison of the Krško Power Plant simplified Spent Fuel Pool (SFP) dose rates using different computational shielding methodologies. The analysis was performed to estimate limiting gamma dose rates on wall mounted level instrumentation in case of significant loss of cooling water. The SFP was represented with simple homogenized cylinders (point kernel and Monte Carlo (MC)) or cuboids (MC) using uranium, iron, water, and dry-air as a bulk region materials. The pool is divided on the old and new section where the old one has three additional subsections representing fuel assemblies (FAs) with different burnup/cooling time (60 days, 1 year and 5 years). The new section represents the FAs with the cooling time of 10 years. The time dependent fuel assembly isotopic composition was calculated using ORIGEN2 code applied to the depletion of one of the fuel assemblies present in the pool (AC-29). The source used in Microshield calculation is based on imported isotopic activities. The time dependent photon spectrum with total source intensity from Microshield multigroup point kernel calculations was then prepared for two hybrid deterministic-stochastic sequences. One is based on SCALE6.2b3/MAVRIC (Monaco and Denovo) methodology and another uses Monte Carlo code MCNP6.1.1b and ADVANTG3.0.1. code. Even though this model is a fairly simple one, the layers of shielding materials are thick enough to pose a significant shielding problem for MC method without the use of effective variance reduction (VR) technique. For that purpose the ADVANTG code was used to generate VR parameters for the MCNP fixed-source calculation using continuous energy transport. ADVATNG employs a deterministic forward-adjoint transport solver Denovo which implements CADIS/FWCADIS methodology. Denovo uses a structured, Cartesian-grid SN solver based on the Koch- Baker-Alcouffe parallel transport sweep algorithm across x-y domain blocks. This was our first application of ANDVANTG/MCNP hybrid sequence for this type of calculation and the results where compared to SCALE/MAVRIC sequence which we regularly use for similar calculations. The comparison of gamma dose rates on different point detector locations (central above pool and at the top of pool periphery) showed a good agreement between Microshield (point-kernel) and deterministic-stochastic shielding methodologies for the cylindrical approximation of the pool geometry. More complicated cases for model with multi-source option and for cuboids showed very good agreement between SCALE/MAVRIC and ANDVANTG/MCNP calculations