Monte Carlo-based development of a shield and total background estimation for the COBRA experiment

Energy Technology Data Exchange (ETDEWEB)

Heidrich, Nadine

2014-11-15

The COBRA experiment aims for the measurement of the neutrinoless double beta decay and thus for the determination the effective Majorana mass of the neutrino. To be competitive with other next-generation experiments the background rate has to be in the order of 10{sup -3} counts/kg/keV/yr, which is a challenging criterion. This thesis deals with the development of a shield design and the calculation of the expected total background rate for the large scale COBRA experiment containing 13824 6 cm{sup 3} CdZnTe detectors. For the development of a shield single-layer and multi-layer shields were investigated and a shield design was optimized concerning high-energy muon-induced neutrons. As the best design the combination of 10 cm boron doped polyethylene as outermost layer, 20 cm lead and 10 cm copper as innermost layer were determined. It showed the best performance regarding neutron attenuation as well as (n, γ) self-shielding effects leading to a negligible background rate of less than 2.10{sup -6} counts/kg/keV/yr. Additionally. the shield with a thickness of 40 cm is compact and costeffective. In the next step the expected total background rate was computed taking into account individual setup parts and various background sources including natural and man-made radioactivity, cosmic ray-induced background and thermal neutrons. Furthermore, a comparison of measured data from the COBRA demonstrator setup with Monte Carlo data was used to calculate reliable contamination levels of the single setup parts. The calculation was performed conservatively to prevent an underestimation. In addition, the contribution to the total background rate regarding the individual detector parts and background sources was investigated. The main portion arise from the Delrin support structure, the Glyptal lacquer followed by the circuit board of the high voltage supply. Most background events originate from particles with a quantity of 99 % in total. Regarding surface events a

Self-Shielding Treatment to Perform Cell Calculation for Seed Furl In Th/U Pwr Using Dragon Code

Directory of Open Access Journals (Sweden)

Ahmed Amin El Said Abd El Hameed

2015-08-01

Full Text Available Time and precision of the results are the most important factors in any code used for nuclear calculations. Despite of the high accuracy of Monte Carlo codes, MCNP and Serpent, in many cases their relatively long computational time leads to difficulties in using any of them as the main calculation code. Usually, Monte Carlo codes are used only to benchmark the results. The deterministic codes, which are usually used in nuclear reactor’s calculations, have limited precision, due to the approximations in the methods used to solve the multi-group transport equation. Self- Shielding treatment, an algorithm that produces an average cross-section defined over the complete energy domain of the neutrons in a nuclear reactor, is responsible for the biggest error in any deterministic codes. There are mainly two resonance self-shielding models commonly applied: models based on equivalence and dilution and models based on subgroup approach. The fundamental problem with any self-shielding method is that it treats any isotope as there are no other isotopes with resonance present in the reactor. The most practical way to solve this problem is to use multi-energy groups (50-200 that are chosen in a way that allows us to use all major resonances without self-shielding. In this paper, we perform cell calculations, for a fresh seed fuel pin which is used in thorium/uranium reactors, by solving 172 energy group transport equation using the deterministic DRAGON code, for the two types of self-shielding models (equivalence and dilution models and subgroup models Using WIMS-D5 and DRAGON data libraries. The results are then tested by comparing it with the stochastic MCNP5 code.  We also tested the sensitivity of the results to a specific change in self-shielding method implemented, for example the effect of applying Livolant-Jeanpierre Normalization scheme and Rimman Integration improvement on the equivalence and dilution method, and the effect of using Ribbon

Monte-Carlo simulations of neutron shielding for the ATLAS forward region

CERN Document Server

Stekl, I; Kovalenko, V E; Vorobel, V; Leroy, C; Piquemal, F; Eschbach, R; Marquet, C

2000-01-01

The effectiveness of different types of neutron shielding for the ATLAS forward region has been studied by means of Monte-Carlo simulations and compared with the results of an experiment performed at the CERN PS. The simulation code is based on GEANT, FLUKA, MICAP and GAMLIB. GAMLIB is a new library including processes with gamma-rays produced in (n, gamma), (n, n'gamma) neutron reactions and is interfaced to the MICAP code. The effectiveness of different types of shielding against neutrons and gamma-rays, composed from different types of material, such as pure polyethylene, borated polyethylene, lithium-filled polyethylene, lead and iron, were compared. The results from Monte-Carlo simulations were compared to the results obtained from the experiment. The simulation results reproduce the experimental data well. This agreement supports the correctness of the simulation code used to describe the generation, spreading and absorption of neutrons (up to thermal energies) and gamma-rays in the shielding materials....

Using the Monte Carlo Coupling Technique to Evaluate the Shielding Ability of a Modular Shielding House to Accommodate Spent-Fuel Transportable Storage Casks

International Nuclear Information System (INIS)

Ueki, Kohtaro; Kawakami, Kazuo; Shimizu, Daisuke

2003-01-01

The Monte Carlo coupling technique with the coordinate transformation is used to evaluate the shielding ability of a modular shielding house that accommodates four spent-fuel transportable storage casks for two units. The effective dose rate distributions can be obtained as far as 300 m from the center of the shielding house. The coupling technique is created with the Surface Source Write (SSW) card and the Surface Source Read/Coordinate Transformation (SSR/CRT) card in the MCNP 4C continuous energy Monte Carlo code as the 'SSW-SSR/CRT calculation system'. In the present Monte Carlo coupling calculation, the total effective dose rates 100, 200, and 300 m from the center of the shielding house are estimated to be 1.69, 0.285, and 0.0826 (μSv/yr per four casks), respectively. Accordingly, if the distance between the center of the shielding house and the site boundary of the storage facility is kept at >300 m, approximately 2400 casks are able to be accommodated in the modular shielding houses, under the Japanese severe criterion of 50 μSv/yr at the site boundary. The shielding house alone satisfies not only the technical conditions but also the economic requirements.It became evident that secondary gamma rays account for >60% of the effective total dose rate at all the calculated points around the shielding house, most of which are produced from the water in the steel-water-steel shielding system of the shielding house. The remainder of the dose rate comes mostly from neutrons; the fission product and 60 Co activation gamma rays account for small percentages. Accordingly, reducing the secondary gamma rays is critical to improving not only the shielding ability but also the radiation safety of the shielding house

Applications to shielding design and others of monte carlo method

Energy Technology Data Exchange (ETDEWEB)

Ito, Daiichiro [Mitsui Engineering and Shipbuiding Co., Ltd., Tokyo (Japan)

2001-01-01

One-dimensional or two-dimensional Sn computer code (ANISN, DOT3.5, etc.) and a point attenuation kernel integral code (QAD, etc.) have been used widely for shielding design. Application examples of monte carlo method which could follow precisely the three-dimensional configuration of shielding structure are shown as follow: (1) CASTER cask has a complex structure which consists of a large number of fuel baskets (stainless steel), neutron moderators (polyethylene rods), the body (cast iron), and cooling fin. The R-{theta} model of Sn code DOT3.5 cannot follow closely the complex form of polyethylene rods and fuel baskets. A monte carlo code MORSE is used to ascertain the calculation results of DOT3.5. The discrepancy between the calculation results of DOT3.5 and MORSE was in 10% for dose rate at distance of 1 m from the cask surface. (2) The dose rates of an iron cell at 10 cm above the floor are calculated by the code QAD and the MORSE. The reflected components of gamma ray caused by the auxiliary floor shield (lead) are analyzed by the MORSE. (3) A monte carlo code MCNP4A is used for skyshine evaluation of spent fuel carrier ship 'ROKUEIMARU'. The direct and skyshine components of gamma ray and neutron flux are estimated at each center of engine room and wheel house. The skyshine dose rate of neutron flux is 5-15 times larger than the gamma ray. (M. Suetake)

The biological shield of a high-intensity spallation source: a monte Carlo design study

International Nuclear Information System (INIS)

Koprivnikar, I.; Schachinger, E.

2004-01-01

The design of high-intensity spallation sources requires the best possible estimates for the biological shield. The applicability of three-dimensional Monte Carlo simulation in the design of the biological shield of a spallation source will be discussed. In order to achieve reasonable computing times along with acceptable accuracy, biasing techniques are to be employed and it was the main purpose of this work to develop a strategy for an effective Monte Carlo simulation in shielding design. The most prominent MC computer codes, namely MCNPX and FLUKA99, have been applied to the same model spallation source (the European Spallation Source) and on the basis of the derived strategies, the design and characteristics of the target station shield are discussed. It is also the purpose of the paper to demonstrate the application of the developed strategy for the design of beam lines with their shielding using as an example the target-moderator-reflector complex of the ESS as the primary particle source. (author)

Self-shielding factors for TLD-600 and TLD-100 in an isotropic flux of thermal neutrons

International Nuclear Information System (INIS)

Horowitz, Y.S.; Dubi, A.; Ben Shahar, B.

1976-01-01

The applications of lithium fluoride thermoluminescent dosemeters in mixed n-γ environments, and the dependence of LiF-TL on linear energy transfer are both topics of current interest. Monte Carlo calculations have therefore been carried out to determine the thermal neutron absorption probability (and consequently the self-shielding factor) for an isotropic flux of neutrons impinging on different sized cylindrical samples of LiF TLD-100 and TLD-600. The calculations were performed for cylinders of radius up to 10 cm and heights of 0.1 to 1.5 cm. The Monte Carlo results were found to be significantly different from the analytic calculations for infinitely long cylinders, but, as expected, converged to the same value for (r/h) << 1. (U.K.)

Validation of calculated self-shielding factors for Rh foils

Science.gov (United States)

Jaćimović, R.; Trkov, A.; Žerovnik, G.; Snoj, L.; Schillebeeckx, P.

2010-10-01

Rhodium foils of about 5 mm diameter were obtained from IRMM. One foil had thickness of 0.006 mm and three were 0.112 mm thick. They were irradiated in the pneumatic transfer system and in the carousel facility of the TRIGA reactor at the Jožef Stefan Institute. The foils were irradiated bare and enclosed in small cadmium boxes (about 2 g weight) of 1 mm thickness to minimise the perturbation of the local neutron flux. They were co-irradiated with 5 mm diameter and 0.2 mm thick Al-Au (0.1%) alloy monitor foils. The resonance self-shielding corrections for the 0.006 and 0.112 mm thick samples were calculated by the Monte Carlo simulation and amount to about 10% and 60%, respectively. The consistency of measurements confirmed the validity of self-shielding factors. Trial estimates of Q0 and k0 factors for the 555.8 keV gamma line of 104Rh were made and amount to 6.65±0.18 and (6.61±0.12)×10 -2, respectively.

A new formulation for resonance self-shielding factors

International Nuclear Information System (INIS)

Palma, Daniel A.P.; Martinez, Aquilino S.; Silva, Fernando C. da

2007-01-01

The activation technique allows either absolute or relative very precise neutron intensity measurements. This technique requires the knowledge of the Doppler broadening function to determine resonance self-shielding factors. In the present work a new formulation is proposed for the self-shielding factors where the Doppler broadening function is calculated using the Frobenius's method and compared to the values obtained from the four-pole Pade method. This calculation method is shown to be effective from the point of view of accuracy. (author)

Monte Carlo simulation of determining porosity by using dual gamma detectors

International Nuclear Information System (INIS)

Zhang Feng; Liu Juntao; Yu Huawei; Yuan Chao; Jia Yan

2013-01-01

Current formation elements spectroscopy logging technology utilize 241 Am-Be neutron source and single BGO detector to determine elements contents. It plays an important role in mineral analysis and lithology identification of unconventional oil and gas exploration, but information measured is relatively ld. Measured system based on 241 Am-Be neutron and dual detectors can be developed to realize the measurement of elements content as well as determine neutron gamma porosity by using ratio of gamma count between near and far detectors. Calculation model is built by Monte Carlo method to study neutron gamma porosity logging response with different spacing and shields. And it is concluded that measuring neutron gamma have high counts and good statistical property contrasted with measuring thermal neutron, but the sensitivity of porosity decrease. Sensitivity of porosity will increase as the spacing of dual detector increases. Spacing of far and near detectors should be around 62 cm and 35 cm respectively. Gamma counts decrease and neutron gamma porosity sensitivity increase when shield is fixed between neutron and detector. The length of main shield should be greater than 10 cm and associated shielding is about 5 cm. By Monte Carlo Simulation study, the result provides technical support for determining porosity in formation elements spectroscopy logging using 241 Am-Be neutron and gamma detectors. (authors)

Three-dimensional coupled Monte Carlo-discrete ordinates computational scheme for shielding calculations of large and complex nuclear facilities

International Nuclear Information System (INIS)

Chen, Y.; Fischer, U.

2005-01-01

Shielding calculations of advanced nuclear facilities such as accelerator based neutron sources or fusion devices of the tokamak type are complicated due to their complex geometries and their large dimensions, including bulk shields of several meters thickness. While the complexity of the geometry in the shielding calculation can be hardly handled by the discrete ordinates method, the deep penetration of radiation through bulk shields is a severe challenge for the Monte Carlo particle transport technique. This work proposes a dedicated computational scheme for coupled Monte Carlo-Discrete Ordinates transport calculations to handle this kind of shielding problems. The Monte Carlo technique is used to simulate the particle generation and transport in the target region with both complex geometry and reaction physics, and the discrete ordinates method is used to treat the deep penetration problem in the bulk shield. The coupling scheme has been implemented in a program system by loosely integrating the Monte Carlo transport code MCNP, the three-dimensional discrete ordinates code TORT and a newly developed coupling interface program for mapping process. Test calculations were performed with comparison to MCNP solutions. Satisfactory agreements were obtained between these two approaches. The program system has been chosen to treat the complicated shielding problem of the accelerator-based IFMIF neutron source. The successful application demonstrates that coupling scheme with the program system is a useful computational tool for the shielding analysis of complex and large nuclear facilities. (authors)

A new formulation for resonance self-shielding factors

Energy Technology Data Exchange (ETDEWEB)

Palma, Daniel A.P.; Martinez, Aquilino S.; Silva, Fernando C. da [Universidade Federal, Rio de Janeiro, RJ (Brazil). Coordenacao dos Programas de Pos-graduacao de Engenharia. Programa de Engenharia Nuclear]. E-mail: aquilino@lmp.ufrj.br

2007-07-01

The activation technique allows either absolute or relative very precise neutron intensity measurements. This technique requires the knowledge of the Doppler broadening function to determine resonance self-shielding factors. In the present work a new formulation is proposed for the self-shielding factors where the Doppler broadening function is calculated using the Frobenius's method and compared to the values obtained from the four-pole Pade method. This calculation method is shown to be effective from the point of view of accuracy. (author)

Application of the characteristics method combined with advanced self-shielding models to an ACR-type cell

International Nuclear Information System (INIS)

Le Tellier, R.; Hebert, A.

2005-01-01

In this paper, we present the usage of the method of characteristics (MOC) with advanced self-shielding models for a fundamental lattice calculation on an ACR-type cell i.e. a cluster geometry with light water coolant and heavy water moderator. Comparison with the collision probability method (CP) show the consistency of the method of characteristics as implemented both in flux and self-shielding calculations. Acceleration techniques are tested in the different calculations and prove to be efficient. Comparisons with the Monte-Carlo code Tripoli4 show the advantage of a subgroup approach for self-shielding calculations : the difference in k eff is less than one standard deviation of the Tripoli4 calculation and in terms of total absorption rates, in the resolved resonances group, the maximum relative error is of the order of 3% localised in the most outer region of the central pin. (author)

Monte Carlo simulation using MCNP4B for an optimal shielding design of a 252 Cf source

International Nuclear Information System (INIS)

Silva, Ademir X. da; Crispim, Verginia R.

2001-01-01

This study aim to investigate an optimum shielding design against neutrons and gamma-rays from a source of 252 Cf, using Monte Carlo simulation. The shielding materials studied were: borated polyethylene, borated-lead polyethylene and stainless steel. The Monte Carlo code MCNP, version 4B, was used to design shielding for 252 Cf based neutron irradiator systems. By normalizing the dose equivalent rate values presented to the neutron production rate of the source, the resulting calculations are independents of the intensity of actual 252 Cf source. The results shown what the total dose equivalent rates were reduced significantly by the shielding system optimization. (author)

An ''exact'' treatment of self-shielding and covers in neutron spectra determinations

International Nuclear Information System (INIS)

Griffin, P.J.; Kelly, J.G.

1995-01-01

Most neutron spectrum determination methodologies ignore self-shielding effects in dosimetry foils and treat covers with an exponential attenuation model. This work provides a quantitative analysis of the approximations in this approach. It also provides a methodology for improving the fidelity of the treatment of the dosimetry sensor response to a level consistent with the user's spectrum characterization approach. A library of correction functions for the energy-dependent sensor response has been compiled that addresses dosimetry foils/configurations in use at the Sandia National Laboratories Radiation Metrology Laboratory

Reliability of Monte Carlo simulations in modeling neutron yields from a shielded fission source

Energy Technology Data Exchange (ETDEWEB)

McArthur, Matthew S., E-mail: matthew.s.mcarthur@gmail.com; Rees, Lawrence B., E-mail: Lawrence_Rees@byu.edu; Czirr, J. Bart, E-mail: czirr@juno.com

2016-08-11

Using the combination of a neutron-sensitive {sup 6}Li glass scintillator detector with a neutron-insensitive {sup 7}Li glass scintillator detector, we are able to make an accurate measurement of the capture rate of fission neutrons on {sup 6}Li. We used this detector with a {sup 252}Cf neutron source to measure the effects of both non-borated polyethylene and 5% borated polyethylene shielding on detection rates over a range of shielding thicknesses. Both of these measurements were compared with MCNP calculations to determine how well the calculations reproduced the measurements. When the source is highly shielded, the number of interactions experienced by each neutron prior to arriving at the detector is large, so it is important to compare Monte Carlo modeling with actual experimental measurements. MCNP reproduces the data fairly well, but it does generally underestimate detector efficiency both with and without polyethylene shielding. For non-borated polyethylene it underestimates the measured value by an average of 8%. This increases to an average of 11% for borated polyethylene.

Radiation shielding design for DECY-13 cyclotron using Monte Carlo method

International Nuclear Information System (INIS)

Rasito T; Bunawas; Taufik; Sunardi; Hari Suryanto

2016-01-01

DECY-13 is a 13 MeV proton cyclotron with target H_2"1"8O. The bombarding of 13 MeV protons on target H_2"1"8O produce large amounts of neutrons and gamma radiation. It needs the efficient radiation shielding to reduce the level of neutrons and gamma rays to ensure safety for workers and public. Modeling and calculations have been carried out using Monte Carlo method with MCNPX code to optimize the thickness for the radiation shielding. The calculations were done for radiation shielding of rectangular space room type with the size of 5.5 m x 5 m x 3 m and thickness of 170 cm made from lightweight concrete types of portland. It was shown that with this shielding the dose rate outside the wall was reduced to 1 μSv/h. (author)

Monteray Mark-I: Computer program (PC-version) for shielding calculation with Monte Carlo method

International Nuclear Information System (INIS)

Pudjijanto, M.S.; Akhmad, Y.R.

1998-01-01

A computer program for gamma ray shielding calculation using Monte Carlo method has been developed. The program is written in WATFOR77 language. The MONTERAY MARH-1 is originally developed by James Wood. The program was modified by the authors that the modified version is easily executed. Applying Monte Carlo method the program observe photon gamma transport in an infinity planar shielding with various thick. A photon gamma is observed till escape from the shielding or when its energy less than the cut off energy. Pair production process is treated as pure absorption process that annihilation photons generated in the process are neglected in the calculation. The out put data calculated by the program are total albedo, build-up factor, and photon spectra. The calculation result for build-up factor of a slab lead and water media with 6 MeV parallel beam gamma source shows that they are in agreement with published data. Hence the program is adequate as a shielding design tool for observing gamma radiation transport in various media

Comparison of calculational methods for liquid metal reactor shields

International Nuclear Information System (INIS)

Carter, L.L.; Moore, F.S.; Morford, R.J.; Mann, F.M.

1985-09-01

A one-dimensional comparison is made between Monte Carlo (MCNP), discrete ordinances (ANISN), and diffusion theory (MlDX) calculations of neutron flux and radiation damage from the core of the Fast Flux Test Facility (FFTF) out to the reactor vessel. Diffusion theory was found to be reasonably accurate for the calculation of both total flux and radiation damage. However, for large distances from the core, the calculated flux at very high energies is low by an order of magnitude or more when the diffusion theory is used. Particular emphasis was placed in this study on the generation of multitable cross sections for use in discrete ordinates codes that are self-shielded, consistent with the self-shielding employed in the generation of cross sections for use with diffusion theory. The Monte Carlo calculation, with a pointwise representation of the cross sections, was used as the benchmark for determining the limitations of the other two calculational methods. 12 refs., 33 figs

Monte Carlo simulations of a D-T neutron generator shielding for landmine detection

International Nuclear Information System (INIS)

Reda, A.M.

2011-01-01

Shielding for a D-T sealed neutron generator has been designed using the MCNP5 Monte Carlo radiation transport code. The neutron generator will be used in field for the detection of explosives, landmines, drugs and other 'threat' materials. The optimization of the detection of buried objects was started by studying the signal-to-noise ratio for different geometric conditions. - Highlights: → A landmine detection system based on neutron fast/slow analysis has been designed. → Shielding for a D-T sealed neutron generator tube has been designed using Monte Carlo radiation transport code. → Detection of buried objects was started by studying the signal-to-noise ratio for different geometric conditions. → The signal-to-background ratio optimized at one position for all depths.

URR [Unresolved Resonance Region] computer code: A code to calculate resonance neutron cross-section probability tables, Bondarenko self-shielding factors, and self-indication ratios for fissile and fertile nuclides

International Nuclear Information System (INIS)

Leal, L.C.; de Saussure, G.; Perez, R.B.

1990-01-01

The URR computer code has been developed to calculate cross-section probability tables, Bondarenko self-shielding factors, and self-indication ratios for fertile and fissile isotopes in the unresolved resonance region. Monte Carlo methods are utilized to select appropriate resonance parameters and to compute the cross sections at the desired reference energy. The neutron cross sections are calculated by the single-level Breit-Wigner formalism with s-, p-, and d-wave contributions. The cross-section probability tables are constructed by sampling by Doppler broadened cross-sections. The various self-shielding factors are computer numerically as Lebesgue integrals over the cross-section probability tables

A practical look at Monte Carlo variance reduction methods in radiation shielding

Energy Technology Data Exchange (ETDEWEB)

Olsher, Richard H. [Los Alamos National Laboratory, Los Alamos (United States)

2006-04-15

With the advent of inexpensive computing power over the past two decades, applications of Monte Carlo radiation transport techniques have proliferated dramatically. At Los Alamos, the Monte Carlo codes MCNP5 and MCNPX are used routinely on personal computer platforms for radiation shielding analysis and dosimetry calculations. These codes feature a rich palette of Variance Reduction (VR) techniques. The motivation of VR is to exchange user efficiency for computational efficiency. It has been said that a few hours of user time often reduces computational time by several orders of magnitude. Unfortunately, user time can stretch into the many hours as most VR techniques require significant user experience and intervention for proper optimization. It is the purpose of this paper to outline VR strategies, tested in practice, optimized for several common radiation shielding tasks, with the hope of reducing user setup time for similar problems. A strategy is defined in this context to mean a collection of MCNP radiation transport physics options and VR techniques that work synergistically to optimize a particular shielding task. Examples are offered the areas of source definition, skyshine, streaming, and transmission.

A practical look at Monte Carlo variance reduction methods in radiation shielding

International Nuclear Information System (INIS)

Olsher, Richard H.

2006-01-01

With the advent of inexpensive computing power over the past two decades, applications of Monte Carlo radiation transport techniques have proliferated dramatically. At Los Alamos, the Monte Carlo codes MCNP5 and MCNPX are used routinely on personal computer platforms for radiation shielding analysis and dosimetry calculations. These codes feature a rich palette of Variance Reduction (VR) techniques. The motivation of VR is to exchange user efficiency for computational efficiency. It has been said that a few hours of user time often reduces computational time by several orders of magnitude. Unfortunately, user time can stretch into the many hours as most VR techniques require significant user experience and intervention for proper optimization. It is the purpose of this paper to outline VR strategies, tested in practice, optimized for several common radiation shielding tasks, with the hope of reducing user setup time for similar problems. A strategy is defined in this context to mean a collection of MCNP radiation transport physics options and VR techniques that work synergistically to optimize a particular shielding task. Examples are offered the areas of source definition, skyshine, streaming, and transmission

SCALE Continuous-Energy Monte Carlo Depletion with Parallel KENO in TRITON

International Nuclear Information System (INIS)

Goluoglu, Sedat; Bekar, Kursat B.; Wiarda, Dorothea

2012-01-01

The TRITON sequence of the SCALE code system is a powerful and robust tool for performing multigroup (MG) reactor physics analysis using either the 2-D deterministic solver NEWT or the 3-D Monte Carlo transport code KENO. However, as with all MG codes, the accuracy of the results depends on the accuracy of the MG cross sections that are generated and/or used. While SCALE resonance self-shielding modules provide rigorous resonance self-shielding, they are based on 1-D models and therefore 2-D or 3-D effects such as heterogeneity of the lattice structures may render final MG cross sections inaccurate. Another potential drawback to MG Monte Carlo depletion is the need to perform resonance self-shielding calculations at each depletion step for each fuel segment that is being depleted. The CPU time and memory required for self-shielding calculations can often eclipse the resources needed for the Monte Carlo transport. This summary presents the results of the new continuous-energy (CE) calculation mode in TRITON. With the new capability, accurate reactor physics analyses can be performed for all types of systems using the SCALE Monte Carlo code KENO as the CE transport solver. In addition, transport calculations can be performed in parallel mode on multiple processors.

Determination of self shielding factors and gamma attenuation effects for tree ring samples

International Nuclear Information System (INIS)

Dagistan Sahin; Kenan Uenlue

2012-01-01

Determination of tree ring chemistry using Neutron Activation Analysis (NAA) is part of an ongoing research between Penn State University (PSU) and Cornell University, The Malcolm and Carolyn Wiener Laboratory for Aegean and Near Eastern Dendrochronology. Tree-ring chemistry yields valuable data for environmental event signatures. These signatures are a complex function of elemental concentration. To be certain about concentration of signature elements, it is necessary to perform the measurements and corrections with the lowest error and maximum accuracy possible. Accurate and precise values of energy dependent neutron flux at dry irradiation tubes and detector efficiency for tree ring sample are calculated for Penn State Breazeale Reactor (PSBR). For the calculation of energy dependent and self shielding corrected neutron flux, detailed model of the TRIGA Mark III reactor at PSU with updated fuel compositions was prepared using the MCNP utility for reactor evolution (MURE) libraries. Dry irradiation tube, sample holder and sample were also included in the model. The thermal flux self-shielding correction factors due to the sample holder and sample for were calculated and verified with previously published values. The Geant-4 model of the gamma spectroscopy system, developed at Radiation Science and Engineering Center (RSEC), was improved and absolute detector efficiency for tree-ring samples was calculated. (author)

DEMONR, Monte-Carlo Shielding Calculation for Neutron Flux and Neutron Spectra, Teaching Program

International Nuclear Information System (INIS)

Courtney, J. C.

1987-01-01

1 - Description of problem or function: DEMONR treats the behavior of neutrons in a slab shield. It is frequently used as a teaching tool. 2 - Method of solution: An unbiased Monte Carlo code calculates the number, energy, and direction of neutrons that penetrate or are reflected from a shield. 3 - Restrictions on the complexity of the problem: Only one shield may be used in each problem. The shield material may be a single element or a homogeneous mixture of elements with a single effective atomic weight. Only elastic scattering and neutron capture processes are allowed. The source is a point located on one face of the slab. It provides a cosine distribution of current. Monoenergetic or fission spectrum neutrons may be selected

SHIELD-HIT12A - a Monte Carlo particle transport program for ion therapy research

DEFF Research Database (Denmark)

Bassler, Niels; Hansen, David Christoffer; Lühr, Armin

2014-01-01

. We experienced that new users quickly learn to use SHIELD-HIT12A and setup new geometries. Contrary to previous versions of SHIELD-HIT, the 12A distribution comes along with easy-to-use example files and an English manual. A new implementation of Vavilov straggling resulted in a massive reduction......Abstract. Purpose: The Monte Carlo (MC) code SHIELD-HIT simulates the transport of ions through matter. Since SHIELD-HIT08 we added numerous features that improves speed, usability and underlying physics and thereby the user experience. The “-A” fork of SHIELD-HIT also aims to attach SHIELD....... It supports native formats compatible with the heavy ion treatment planning system TRiP. Stopping power files follow ICRU standard and are generated using the libdEdx library, which allows the user to choose from a multitude of stopping power tables. Results: SHIELD-HIT12A runs on Linux and Windows platforms...

Improved Monte Carlo modelling of multi-energy a-rays penetration through thick stratified shielding slabs

International Nuclear Information System (INIS)

Bakos, G.C.

2001-01-01

This paper deals with the application of Monte Carlo method for the calculation of dose build up factor of, mixed 1.37 and 2.75 MeV, a-rays penetration through stratified shielding slabs. Six double layer shielding slabs namely, 12 A l+Fe, 12 A l+Pb, 6 F e+Al, 6 F e+Pb, 4 P b+Al, 4 P b+Fe were examined. Furthermore, experimental and theoretical results are also presented. The experimental results were taken from the experimental facility installed at the Universities Research reactor Center (Risley, UK). Activated Na2SO3 solution provided a uniform Na-24 disc source of a-rays at both energies (1.37 and 2.75 MeV) with equal intensity. The theoretical results were calculated using the Bowman and Trubey formula. This formula takes into account an exponentially decaying function of the shield thickness (in mfp) to the end point of the multi-layer slab. The experimental and theoretical results were used to evaluate the simulation results produced from a Monte Carlo program (DUTMONCA code) which was developed in Democritus University of Thrace (Xanthi, Greece). The DUTMONCA code was written in Pascal language and run on an Intel PIII-800 microprocessor. The developed code (which is an improved version of an existing Monte Carlo program) has the ability to produce good results for thick shielding slabs overcoming the problems encountered in older version program. The simulation results are compared with experimental and theoretical results. Good agreement can be observed, even for thick layer shielding slabs, although there are some wayward experimental values which are due to sources of error associated with the experimental procedure

Performance of advanced self-shielding models in DRAGON Version4 on analysis of a high conversion light water reactor lattice

International Nuclear Information System (INIS)

Karthikeyan, Ramamoorthy; Hebert, Alain

2008-01-01

A high conversion light water reactor lattice has been analysed using the code DRAGON Version4. This analysis was performed to test the performance of the advanced self-shielding models incorporated in DRAGON Version4. The self-shielding models are broadly classified into two groups - 'equivalence in dilution' and 'subgroup approach'. Under the 'equivalence in dilution' approach we have analysed the generalized Stamm'ler model with and without Nordheim model and Riemann integration. These models have been analysed also using the Livolant-Jeanpierre normalization. Under the 'subgroup approach', we have analysed Statistical self-shielding model based on physical probability tables and Ribon extended self-shielding model based on mathematical probability tables. This analysis will help in understanding the performance of advanced self-shielding models for a lattice that is tight and has a large fraction of fissions happening in the resonance region. The nuclear data for the analysis was generated in-house. NJOY99.90 was used for generating libraries in DRAGLIB format for analysis using DRAGON and A Compact ENDF libraries for analysis using MCNP5. The evaluated datafiles were chosen based on the recommendations of the IAEA Co-ordinated Research Project on the WIMS Library Update Project. The reference solution for the problem was obtained using Monte Carlo code MCNP5. It was found that the Ribon extended self-shielding model based on mathematical probability tables using correlation model performed better than all other models

Gamma self-shielding correction factors calculation for aqueous bulk sample analysis by PGNAA technique

International Nuclear Information System (INIS)

Nasrabadi, M.N.; Mohammadi, A.; Jalali, M.

2009-01-01

In this paper bulk sample prompt gamma neutron activation analysis (BSPGNAA) was applied to aqueous sample analysis using a relative method. For elemental analysis of an unknown bulk sample, gamma self-shielding coefficient was required. Gamma self-shielding coefficient of unknown samples was estimated by an experimental method and also by MCNP code calculation. The proposed methodology can be used for the determination of the elemental concentration of unknown aqueous samples by BSPGNAA where knowledge of the gamma self-shielding within the sample volume is required.

A simple method for correcting the neutron self-shielding effect of matrix and improving the analytical response in prompt gamma-ray neutron activation analysis

International Nuclear Information System (INIS)

Sudarshan, K.; Tripathi, R.; Nair, A.G.C.; Acharya, R.; Reddy, A.V.R.; Goswami, A.

2005-01-01

A simple method using an internal standard is proposed to correct for the self-shielding effect of B, Cd and Gd in a matrix. This would increase the linear dynamic range of PGNAA in analyzing samples containing these elements. The method is validated by analyzing synthetic samples containing large amounts of B, Cd, Hg and Gd, the elements having high neutron absorption cross-section, in aqueous solutions and solid forms. A simple Monte-Carlo simulation to find the extent of self-shielding in the matrix is presented. The method is applied to the analysis of titanium boride alloy containing large amount of boron. The satisfactory results obtained showed the efficacy of the method of correcting for the self-shielding effects in the sample

Monte Carlo simulation of photon buildup factors for shielding materials in diagnostic x-ray facilities

International Nuclear Information System (INIS)

Kharrati, Hedi; Agrebi, Amel; Karoui, Mohamed Karim

2012-01-01

Purpose: A simulation of buildup factors for ordinary concrete, steel, lead, plate glass, lead glass, and gypsum wallboard in broad beam geometry for photons energies from 10 keV to 150 keV at 5 keV intervals is presented. Methods: Monte Carlo N-particle radiation transport computer code has been used to determine the buildup factors for the studied shielding materials. Results: An example concretizing the use of the obtained buildup factors data in computing the broad beam transmission for tube potentials at 70, 100, 120, and 140 kVp is given. The half value layer, the tenth value layer, and the equilibrium tenth value layer are calculated from the broad beam transmission for these tube potentials. Conclusions: The obtained values compared with those calculated from the published data show the ability of these data to predict shielding transmission curves. Therefore, the buildup factors data can be combined with primary, scatter, and leakage x-ray spectra to provide a computationally based solution to broad beam transmission for barriers in shielding x-ray facilities.

Monte Carlo simulation of photon buildup factors for shielding materials in diagnostic x-ray facilities.

Science.gov (United States)

Kharrati, Hedi; Agrebi, Amel; Karoui, Mohamed Karim

2012-10-01

A simulation of buildup factors for ordinary concrete, steel, lead, plate glass, lead glass, and gypsum wallboard in broad beam geometry for photons energies from 10 keV to 150 keV at 5 keV intervals is presented. Monte Carlo N-particle radiation transport computer code has been used to determine the buildup factors for the studied shielding materials. An example concretizing the use of the obtained buildup factors data in computing the broad beam transmission for tube potentials at 70, 100, 120, and 140 kVp is given. The half value layer, the tenth value layer, and the equilibrium tenth value layer are calculated from the broad beam transmission for these tube potentials. The obtained values compared with those calculated from the published data show the ability of these data to predict shielding transmission curves. Therefore, the buildup factors data can be combined with primary, scatter, and leakage x-ray spectra to provide a computationally based solution to broad beam transmission for barriers in shielding x-ray facilities.

Monte Carlo based radial shield design of typical PWR reactor

Energy Technology Data Exchange (ETDEWEB)

Gul, Anas; Khan, Rustam; Qureshi, M. Ayub; Azeem, Muhammad Waqar; Raza, S.A. [Pakistan Institute of Engineering and Applied Sciences, Islamabad (Pakistan). Dept. of Nuclear Engineering; Stummer, Thomas [Technische Univ. Wien (Austria). Atominst.

2017-04-15

This paper presents the radiation shielding model of a typical PWR (CNPP-II) at Chashma, Pakistan. The model was developed using Monte Carlo N Particle code [2], equipped with ENDF/B-VI continuous energy cross section libraries. This model was applied to calculate the neutron and gamma flux and dose rates in the radial direction at core mid plane. The simulated results were compared with the reference results of Shanghai Nuclear Engineering Research and Design Institute (SNERDI).

Neutron shielding for a 252 Cf source

International Nuclear Information System (INIS)

Vega C, H.R.; Manzanares A, E.; Hernandez D, V.M.; Eduardo Gallego, Alfredo Lorente

2006-01-01

To determine the neutron shielding features of water-extended polyester a Monte Carlo study was carried out. Materials with low atomic number are predominantly used for neutron shielding because these materials effectively attenuate neutrons, mainly through inelastic collisions and absorption reactions. During the selection of materials to design a neutron shield, prompt gamma production as well as radionuclide production induced by neutron activation must be considered. In this investigation the Monte Carlo method was used to evaluate the performance of a water-extended polyester shield designed for the transportation, storage, and use of a 252 Cf isotopic neutron source. During calculations a detailed model for the 252 Cf and the shield was utilized. To compare the shielding features of water extended polyester, the calculations were also made for the bare 252 Cf in vacuum, air and the shield filled with water. For all cases the calculated neutron spectra was utilized to determine the ambient equivalent neutron dose at four sites around the shielding. In the case of water extended polyester and water shielding the calculations were extended to include the prompt gamma rays produced during neutron interactions, with this information the Kerma in air was calculated at the same locations where the ambient equivalent neutron dose was determined. (Author)

Development of a shield based on Monte-Carlo studies for the COBRA experiment

Energy Technology Data Exchange (ETDEWEB)

Heidrich, Nadine [Institut fuer Experimentalphysik, 22761 Hamburg (Germany); Collaboration: COBRA-Collaboration

2013-07-01

COBRA is a next-generation experiment searching for neutrinoless double beta decay using CdZnTe semiconductor detectors. The main focus is on {sup 116}Cd, with a decay energy of 2813.5 keV well above the highest naturally occurring gamma lines. The concept for a large scale set-up consists of an array of CdZnTe detectors with a total mass of 420 kg enriched in {sup 116}Cd up to 90 %. With a background rate in the order of 10{sup -3} counts/keV/kg/year, the experiment would be sensitive to a half-life larger than 10{sup 26} years, corresponding to a Majorana mass term m{sub ββ} smaller than 50 meV. To achieve the background level, an appropriate shield is necessary. The shield is developed based on Monte-Carlo simulations. For that, different materials and configurations are tested. In the talk the current status of the Monte-Carlo survey is presented and discussed.

Investigation of Radiation Protection Methodologies for Radiation Therapy Shielding Using Monte Carlo Simulation and Measurement

Science.gov (United States)

Tanny, Sean

The advent of high-energy linear accelerators for dedicated medical use in the 1950's by Henry Kaplan and the Stanford University physics department began a revolution in radiation oncology. Today, linear accelerators are the standard of care for modern radiation therapy and can generate high-energy beams that can produce tens of Gy per minute at isocenter. This creates a need for a large amount of shielding material to properly protect members of the public and hospital staff. Standardized vault designs and guidance on shielding properties of various materials are provided by the National Council on Radiation Protection (NCRP) Report 151. However, physicists are seeking ways to minimize the footprint and volume of shielding material needed which leads to the use of non-standard vault configurations and less-studied materials, such as high-density concrete. The University of Toledo Dana Cancer Center has utilized both of these methods to minimize the cost and spatial footprint of the requisite radiation shielding. To ensure a safe work environment, computer simulations were performed to verify the attenuation properties and shielding workloads produced by a variety of situations where standard recommendations and guidance documents were insufficient. This project studies two areas of concern that are not addressed by NCRP 151, the radiation shielding workload for the vault door with a non-standard design, and the attenuation properties of high-density concrete for both photon and neutron radiation. Simulations have been performed using a Monte-Carlo code produced by the Los Alamos National Lab (LANL), Monte Carlo Neutrons, Photons 5 (MCNP5). Measurements have been performed using a shielding test port designed into the maze of the Varian Edge treatment vault.

Self-shielding characteristics of aqueous self-cooled blankets for next generation fusion devices

International Nuclear Information System (INIS)

Pelloni, S.; Cheng, E.T.; Embrechts, M.J.

1987-11-01

The present study examines self-shielding characteristics for two aqueous self-cooled tritium producing driver blankets for next generation fusion devices. The aqueous Self-Cooled Blanket concept (ASCB) is a very simple blanket concept that relies on just structural material and coolant. Lithium compounds are dissolved in water to provide for tritium production. An ASCB driver blanket would provide a low technology and low temperature environment for blanket test modules in a next generation fusion reactor. The primary functions of such a blanket would be shielding, energy removal and tritium production. One driver blanket considered in this study concept relates to the one proposed for the Next European Torus (NET), while the second concept is indicative for the inboard shield design for the Engineering Test Reactor proposed by the USA (TIBER II/ETR). The driver blanket for NET is based on stainless steel for the structural material and aqueous solution, while the inboard shielding blanket for TIBER II/ETR is based on a tungsten/aqueous solution combination. The purpose of this study is to investigate self-shielding and heterogeneity effects in aqueous self-cooled blankets. It is found that no significant gains in tritium breeding can be achieved in the stainless steel blanket if spatial and energy self-shielding effects are considered, and the heterogeneity effects are also insignificant. The tungsten blanket shows a 5 percent increase in tritium production in the shielding blanket when energy and spatial self-shielding effects are accounted for. However, the tungsten blanket shows a drastic increase in the tritium breeding ratio due to heterogeneity effects. (author) 17 refs., 9 figs., 9 tabs

URR [Unresolved Resonance Region] computer code: A code to calculate resonance neutron cross-section probability tables, Bondarenko self-shielding factors, and self-indication ratios for fissile and fertile nuclides

International Nuclear Information System (INIS)

Leal, L.C.; de Saussure, G.; Perez, R.B.

1989-01-01

The URR computer code has been developed to calculate cross-section probability tables, Bondarenko self-shielding factors, and self- indication ratios for fertile and fissile isotopes in the unresolved resonance region. Monte Carlo methods are utilized to select appropriate resonance parameters and to compute the cross sections at the desired reference energy. The neutron cross sections are calculated by the single-level Breit-Wigner formalism with s-, p-, and d-wave contributions. The cross-section probability tables are constructed by sampling the Doppler broadened cross-section. The various shelf-shielded factors are computed numerically as Lebesgue integrals over the cross-section probability tables. 6 refs

Double-layer neutron shield design as neutron shielding application

Science.gov (United States)

Sariyer, Demet; Küçer, Rahmi

2018-02-01

The shield design in particle accelerators and other high energy facilities are mainly connected to the high-energy neutrons. The deep penetration of neutrons through massive shield has become a very serious problem. For shielding to be efficient, most of these neutrons should be confined to the shielding volume. If the interior space will become limited, the sufficient thickness of multilayer shield must be used. Concrete and iron are widely used as a multilayer shield material. Two layers shield material was selected to guarantee radiation safety outside of the shield against neutrons generated in the interaction of the different proton energies. One of them was one meter of concrete, the other was iron-contained material (FeB, Fe2B and stainless-steel) to be determined shield thicknesses. FLUKA Monte Carlo code was used for shield design geometry and required neutron dose distributions. The resulting two layered shields are shown better performance than single used concrete, thus the shield design could leave more space in the interior shielded areas.

A Modeling of BWR-MOX assemblies based on the characteristics method combined with advanced self-shielding models

International Nuclear Information System (INIS)

Le Tellier, R.; Hebert, A.; Le Tellier, R.; Santamarina, A.; Litaize, O.

2008-01-01

Calculations based on the characteristics method and different self-shielding models are presented for 9 x 9 boiling water reactor (BWR) assemblies fully loaded with mixed-oxide (MOX) fuel. The geometry of these assemblies was recovered from the BASALA experimental program. We have focused our study on three configurations simulating the different voiding conditions that an assembly can undergo in a BWR pressure vessel. A parametric study was carried out with respect to the spatial discretization, the tracking parameters, and the anisotropy order. Comparisons with Monte Carlo calculations in terms of k eff , radiative capture, and fission rates were performed to validate the computational tools. The results are in good agreement between the stochastic and deterministic approaches. The mutual self-shielding model recently introduced within the framework of the Ribon extending self-shielding method appears to be useful for this type of assemblies. Indeed, in the calculation of these MOX benchmarks, the overlapping of resonances, especially between 238 U and 240 Pu, plays an important role due to the spectral strengthening of the flux as the voiding percentage is increased. The method of characteristics is shown to be adequate to perform accurate calculations handling a fine spatial discretization. (authors)

Automated variance reduction of Monte Carlo shielding calculations using the discrete ordinates adjoint function

International Nuclear Information System (INIS)

Wagner, J.C.; Haghighat, A.

1998-01-01

Although the Monte Carlo method is considered to be the most accurate method available for solving radiation transport problems, its applicability is limited by its computational expense. Thus, biasing techniques, which require intuition, guesswork, and iterations involving manual adjustments, are employed to make reactor shielding calculations feasible. To overcome this difficulty, the authors have developed a method for using the S N adjoint function for automated variance reduction of Monte Carlo calculations through source biasing and consistent transport biasing with the weight window technique. They describe the implementation of this method into the standard production Monte Carlo code MCNP and its application to a realistic calculation, namely, the reactor cavity dosimetry calculation. The computational effectiveness of the method, as demonstrated through the increase in calculational efficiency, is demonstrated and quantified. Important issues associated with this method and its efficient use are addressed and analyzed. Additional benefits in terms of the reduction in time and effort required of the user are difficult to quantify but are possibly as important as the computational efficiency. In general, the automated variance reduction method presented is capable of increases in computational performance on the order of thousands, while at the same time significantly reducing the current requirements for user experience, time, and effort. Therefore, this method can substantially increase the applicability and reliability of Monte Carlo for large, real-world shielding applications

Self-shielding coefficient and thermal flux depression factor of voluminous sample in neutron activation analysis

International Nuclear Information System (INIS)

Noorddin Ibrahim; Rosnie Akang

2009-01-01

Full text: One of the major problems encountered during the irradiation of large inhomogeneous samples in performing activation analysis using neutron is the perturbation of the neutron field due to absorption and scattering of neutron within the sample as well as along the neutron guide in the case of prompt gamma activation analysis. The magnitude of this perturbation shown by self-shielding coefficient and flux depression depend on several factors including the average neutron energy, the size and shape of the sample, as well as the macroscopic absorption cross section of the sample. In this study, we use Monte Carlo N-Particle codes to simulate the variation of neutron self-shielding coefficient and thermal flux depression factor as a function of the macroscopic thermal absorption cross section. The simulation works was carried out using the high performance computing facility available at UTM while the experimental work was performed at the tangential beam port of Reactor TRIGA PUSPATI, Malaysia Nuclear Agency. The neutron flux measured along the beam port is found to be in good agreement with the simulated data. Our simulation results also reveal that total flux perturbation factor decreases as the value of absorption increases. This factor is close to unity for low absorbing sample and tends towards zero for strong absorber. In addition, sample with long mean chord length produces smaller flux perturbation than the shorter mean chord length. When comparing both the graphs of self-shielding factor and total disturbance, we can conclude that the total disturbance of the thermal neutron flux on the large samples is dominated by the self-shielding effect. (Author)

Neutron shielding for a {sup 252} Cf source

Energy Technology Data Exchange (ETDEWEB)

Vega C, H.R.; Manzanares A, E.; Hernandez D, V.M. [Unidades Academicas de Estudios Nucleares e Ingenieria Electrica, Universidad Autonoma de Zacatecas, C. Cipres 10, Fracc. La Penuela, 98068 Zacatecas (Mexico); Eduardo Gallego, Alfredo Lorente [Depto. de Ingenieria Nuclear, ETS Ingenieros Industriales, Universidad Politecnica de Madrid, C. Jose Gutierrez Abascal 2, 28006 Madrid (Spain)]. e-mail: fermineutron@yahoo.com

2006-07-01

To determine the neutron shielding features of water-extended polyester a Monte Carlo study was carried out. Materials with low atomic number are predominantly used for neutron shielding because these materials effectively attenuate neutrons, mainly through inelastic collisions and absorption reactions. During the selection of materials to design a neutron shield, prompt gamma production as well as radionuclide production induced by neutron activation must be considered. In this investigation the Monte Carlo method was used to evaluate the performance of a water-extended polyester shield designed for the transportation, storage, and use of a {sup 252}Cf isotopic neutron source. During calculations a detailed model for the {sup 252}Cf and the shield was utilized. To compare the shielding features of water extended polyester, the calculations were also made for the bare {sup 252}Cf in vacuum, air and the shield filled with water. For all cases the calculated neutron spectra was utilized to determine the ambient equivalent neutron dose at four sites around the shielding. In the case of water extended polyester and water shielding the calculations were extended to include the prompt gamma rays produced during neutron interactions, with this information the Kerma in air was calculated at the same locations where the ambient equivalent neutron dose was determined. (Author)

Neutron shielding performance of water-extended polyester

International Nuclear Information System (INIS)

Vega Carrillo, H.R.; Manzanares-Acuna, E.; Hernandez-Davila, V.M.; Vega Carrillo, H.R.; Hernandez-Davila, V.M.; Gallego, E.; Lorente, A.

2006-01-01

A Monte Carlo study to determine the shielding features to neutrons of water-extended polyester (WEP) was carried out. Materials with low atomic number are predominantly used for neutron shielding because these materials effectively attenuate neutrons, mainly through elastic and inelastic collisions. In addition to neutron attenuation properties, other desirable properties for neutron shielding materials include mechanical strength, stability, low cost, and ease of handling. During the selection of materials to design a neutron shield, prompt gamma production as well as radionuclide induced by neutron activation must be considered. In this investigation the Monte Carlo method (MCNP code) was used to evaluate the performance of a water-extended polyester shield designed for the transportation, storage, and use of a 252 Cf isotopic neutron source, for comparison the calculations were extended to water shielding, the bare source in vacuum and in air. (authors)

Neutron shielding performance of water-extended polyester

Energy Technology Data Exchange (ETDEWEB)

Vega Carrillo, H.R.; Manzanares-Acuna, E.; Hernandez-Davila, V.M. [Zacatecas Univ. Autonoma, Nuclear Studies (Mexico); Vega Carrillo, H.R.; Hernandez-Davila, V.M. [Zacatecas Univ. Autonoma, Electric Engineering Academic Units (Mexico); Gallego, E.; Lorente, A. [Madrid Univ. Politecnica, cNuclear Engineering Department (Mexico)

2006-07-01

A Monte Carlo study to determine the shielding features to neutrons of water-extended polyester (WEP) was carried out. Materials with low atomic number are predominantly used for neutron shielding because these materials effectively attenuate neutrons, mainly through elastic and inelastic collisions. In addition to neutron attenuation properties, other desirable properties for neutron shielding materials include mechanical strength, stability, low cost, and ease of handling. During the selection of materials to design a neutron shield, prompt gamma production as well as radionuclide induced by neutron activation must be considered. In this investigation the Monte Carlo method (MCNP code) was used to evaluate the performance of a water-extended polyester shield designed for the transportation, storage, and use of a {sup 252}Cf isotopic neutron source, for comparison the calculations were extended to water shielding, the bare source in vacuum and in air. (authors)

Shielding evaluation of neutron generator hall by Monte Carlo simulations

Energy Technology Data Exchange (ETDEWEB)

Pujala, U.; Selvakumaran, T.S.; Baskaran, R.; Venkatraman, B. [Radiological Safety Division, Indira Gandhi Center for Atomic Research, Kalpakkam (India); Thilagam, L.; Mohapatra, D.K., E-mail: swathythila2@yahoo.com [Safety Research Institute, Atomic Energy Regulatory Board, Kalpakkam (India)

2017-04-01

A shielded hall was constructed for accommodating a D-D, D-T or D-Be based pulsed neutron generator (NG) with 4π yield of 10{sup 9} n/s. The neutron shield design of the facility was optimized using NCRP-51 methodology such that the total dose rates outside the hall areas are well below the regulatory limit for full occupancy criterion (1 μSv/h). However, the total dose rates at roof top, cooling room trench exit and labyrinth exit were found to be above this limit for the optimized design. Hence, additional neutron shielding arrangements were proposed for cooling room trench and labyrinth exits. The roof top was made inaccessible. The present study is an attempt to evaluate the neutron and associated capture gamma transport through the bulk shields for the complete geometry and materials of the NG-Hall using Monte Carlo (MC) codes MCNP and FLUKA. The neutron source terms of D-D, D-T and D-Be reactions are considered in the simulations. The effect of additional shielding proposed has been demonstrated through the simulations carried out with the consideration of the additional shielding for D-Be neutron source term. The results MC simulations using two different codes are found to be consistent with each other for neutron dose rate estimates. However, deviation up to 28% is noted between these two codes at few locations for capture gamma dose rate estimates. Overall, the dose rates estimated by MC simulations including additional shields shows that all the locations surrounding the hall satisfy the full occupancy criteria for all three types of sources. Additionally, the dose rates due to direct transmission of primary neutrons estimated by FLUKA are compared with the values calculated using the formula given in NCRP-51 which shows deviations up to 50% with each other. The details of MC simulations and NCRP-51 methodology for the estimation of primary neutron dose rate along with the results are presented in this paper. (author)

Uncertainty Analysis with Considering Resonance Self-shielding Effect

Energy Technology Data Exchange (ETDEWEB)

Han, Tae Young [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2016-10-15

If infinitely diluted multi-group cross sections were used for the sensitivity, the covariance data from the evaluated nuclear data library (ENDL) was directly applied. However, in case of using a self-shielded multi-group cross section, the covariance data should be corrected considering self-shielding effect. Usually, implicit uncertainty can be defined as the uncertainty change by the resonance self-shielding effect as described above. MUSAD ( Modules of Uncertainty and Sensitivity Analysis for DeCART ) has been developed for a multiplication factor and cross section uncertainty based on the generalized perturbation theory and it, however, can only quantify the explicit uncertainty by the self-shielded multi-group cross sections without considering the implicit effect. Thus, this paper addresses the implementation of the implicit uncertainty analysis module into the code and the numerical results for the verification are provided. The implicit uncertainty analysis module has been implemented into MUSAD based on infinitely-diluted cross section-based consistent method. The verification calculation was performed on MHTGR 350 Ex.I-1a and the differences with McCARD result decrease from 40% to 1% in CZP case and 3% in HFP case. From this study, it is expected that MUSAD code can reasonably produce the complete uncertainty on VHTR or LWR where the resonance self-shielding effect should be significantly considered.

Uncertainty Analysis with Considering Resonance Self-shielding Effect

International Nuclear Information System (INIS)

Han, Tae Young

2016-01-01

If infinitely diluted multi-group cross sections were used for the sensitivity, the covariance data from the evaluated nuclear data library (ENDL) was directly applied. However, in case of using a self-shielded multi-group cross section, the covariance data should be corrected considering self-shielding effect. Usually, implicit uncertainty can be defined as the uncertainty change by the resonance self-shielding effect as described above. MUSAD ( Modules of Uncertainty and Sensitivity Analysis for DeCART ) has been developed for a multiplication factor and cross section uncertainty based on the generalized perturbation theory and it, however, can only quantify the explicit uncertainty by the self-shielded multi-group cross sections without considering the implicit effect. Thus, this paper addresses the implementation of the implicit uncertainty analysis module into the code and the numerical results for the verification are provided. The implicit uncertainty analysis module has been implemented into MUSAD based on infinitely-diluted cross section-based consistent method. The verification calculation was performed on MHTGR 350 Ex.I-1a and the differences with McCARD result decrease from 40% to 1% in CZP case and 3% in HFP case. From this study, it is expected that MUSAD code can reasonably produce the complete uncertainty on VHTR or LWR where the resonance self-shielding effect should be significantly considered

The Monte Carlo method for shielding calculations analysis by MORSE code of a streaming case in the CAORSO BWR power reactor shielding (Italy)

International Nuclear Information System (INIS)

Zitouni, Y.

1987-04-01

In the field of shielding, the requirement of radiation transport calculations in severe conditions, characterized by irreducible three-dimensional geometries has increased the use of the Monte Carlo method. The latter has proved to be the only rigorous and appropriate calculational method in such conditions. However, further efforts at optimization are still necessary to render the technique practically efficient, despite recent improvements in the Monte Carlo codes, the progress made in the field of computers and the availability of accurate nuclear data. Moreover, the personal experience acquired in the field and the control of sophisticated calculation procedures are of the utmost importance. The aim of the work which has been carried out is the gathering of all the necessary elements and features that would lead to an efficient utilization of the Monte Carlo method used in connection with shielding problems. The study of the general aspects of the method and the exploitation techniques of the MORSE code, which has proved to be one of the most comprehensive of the Monte Carlo codes, lead to a successful analysis of an actual case. In fact, the severe conditions and difficulties met have been overcome using such a stochastic simulation code. Finally, a critical comparison between calculated and high-accuracy experimental results has allowed the final confirmation of the methodology used by us

Radiation shielding techniques and applications. 4. Two-Phase Monte Carlo Approach to Photon Streaming Through Three-Legged Penetrations

International Nuclear Information System (INIS)

White, Travis; Hack, Joe; Nathan, Steve; Barnett, Marvin

2001-01-01

solutions for scattering of neutrons through multi-legged penetrations are readily available in the literature; similar analytical solutions for photon scattering through penetrations, however, are not. Therefore, computer modeling must be relied upon to perform our analyses. The computer code typically used by Westinghouse SMS in the evaluation of photon transport through complex geometries is the MCNP Monte Carlo computer code. Yet, geometries of this nature can cause problems even with the Monte Carlo codes. Striking a balance between how the code handles bulk transport through the wall with transport through the penetration void, particularly with the use of typical variance reduction methods, is difficult when trying to ensure that all the important regions of the model are sampled appropriately. The problem was broken down into several roughly independent cases. First, scatter through the penetration was considered. Second, bulk transport through the hot leg of the duct and then through the remaining thickness of wall was calculated to determine the amount of supplemental shielding required in the wall. Similar analyses were performed for the middle and cold legs of the penetration. Finally, additional external shielding from radiation streaming through the duct was determined for cases where the minimum offset distance was not feasible. Each case was broken down further into two phases. In the first phase of each case, photons were transported from the source material to an area at the face of the wall, or the opening of the duct, where photon energy and angular distributions were tallied, representing the source incident on the wall or opening. Then, a simplified model for each case was developed and analyzed using the data from the first phase and the new source term. (authors)

Neutron shielding performance of water-extended polyester

International Nuclear Information System (INIS)

Vega Carrillo, H.R.; Manzanares-Acuna, E.; Hernandez-Davila, V.M.; Vega Carrillo, H.R.; Gallegoc, E.; Lorentec, A.; Hernandez-Davila, V.M.

2006-01-01

A Monte Carlo study to determine the shielding features to neutrons of water-extended polyester was carried out. Materials with low atomic number are predominantly used for neutron shielding because these materials effectively attenuate neutrons, mainly through elastic and inelastic collisions. In addition to neutron attenuation properties, other desirable properties for neutron shielding materials include mechanical strength, stability, low cost, and ease of handling. During the selection of materials to design a neutron shield, prompt gamma production as well as radionuclide induced by neutron activation must be considered. In this investigation the Monte Carlo method (M.C.N.P. code) was used to evaluate the performance of a water-extended polyester shield designed for the transportation, storage, and use of a 252 Cf isotopic neutron source, for comparison the calculations were extended to water shielding, the bare source in vacuum and in air. (authors)

Cross-section fluctuations and self-shielding effects in the unresolved resonance region - International Evaluation Co-operation volume 15

International Nuclear Information System (INIS)

Froehner, F.H.; Larson, Duane C.; Tagesen, Siegfried; Petrizzi, Luigi; Hasegawa, Akira; Nakagawa, Tsuneo; Hogenbirk, Alfred; Weigmann, H.

1995-01-01

A Working Party on International Evaluation Co-operation was established under the sponsorship of the OECD/NEA Nuclear Science Committee (NSC) to promote the exchange of information on nuclear data evaluations, validation, and related topics. Its aim is also to provide a framework for co-operative activities between members of the major nuclear data evaluation projects. This includes the possible exchange of scientists in order to encourage co-operation. Requirements for experimental data resulting from this activity are compiled. The Working Party determines common criteria for evaluated nuclear data files with a view to assessing and improving the quality and completeness of evaluated data. The Parties to the project are: ENDF (United States), JEFF/EFF (NEA Data Bank Member countries), and JENDL (Japan). Co-operation with evaluation projects of non-OECD countries are organised through the Nuclear Data Section of the International Atomic Energy Agency (IAEA). NEA/NSC Subgroup 15 has had the task to assess self-shielding effects in the unresolved resonance range of structural materials, in particular their importance at various energies, and possible ways to deal with them in shielding and activation work. The principal results achieved are summarised briefly, in particular: - New data base consisting of high-resolution transmission data measured at Oak Ridge and Geel; - Improved theoretical understanding of cross-section fluctuations, including their prediction, that has been derived from the Hauser-Feshbach theory; - Benchmark results on the importance of self-shielding in iron at various energies; - Consequences for information storage in evaluated nuclear data files; - Practical utilisation of self-shielding information from evaluated files. Benchmark results as well as the Hauser-Feshbach theory show that self-shielding effects are important up to a 4-or 5-MeV neutron energy. Fluctuation factors extracted from high-resolution total cross-section data can be

Antiproton annihilation physics annihilation physics in the Monte Carlo particle transport code particle transport code SHIELD-HIT12A

DEFF Research Database (Denmark)

Taasti, Vicki Trier; Knudsen, Helge; Holzscheiter, Michael

2015-01-01

The Monte Carlo particle transport code SHIELD-HIT12A is designed to simulate therapeutic beams for cancer radiotherapy with fast ions. SHIELD-HIT12A allows creation of antiproton beam kernels for the treatment planning system TRiP98, but first it must be benchmarked against experimental data. An...

Resonance self-shielding calculation with regularized random ladders

Energy Technology Data Exchange (ETDEWEB)

Ribon, P.

1986-01-01

The straightforward method for calculation of resonance self-shielding is to generate one or several resonance ladders, and to process them as resolved resonances. The main drawback of Monte Carlo methods used to generate the ladders, is the difficulty of reducing the dispersion of data and results. Several methods are examined, and it is shown how one (a regularized sampling method) improves the accuracy. Analytical methods to compute the effective cross-section have recently appeared: they are basically exempt from dispersion, but are inevitably approximate. The accuracy of the most sophisticated one is checked. There is a neutron energy range which is improperly considered as statistical. An examination is presented of what happens when it is treated as statistical, and how it is possible to improve the accuracy of calculations in this range. To illustrate the results calculations have been performed in a simple case: nucleus /sup 238/U, at 300 K, between 4250 and 4750 eV.

Multiconfigurational self-consistent field calculations of nuclear shieldings using London atomic orbitals

DEFF Research Database (Denmark)

Ruud, Kenneth; Helgaker, Trygve; Kobayashi, Rika

1994-01-01

to corresponding individual gauges for localized orbitals (IGLO) results. The London results show better basis set convergence than IGLO, especially for heavier atoms. It is shown that the choice of active space is crucial for determination of accurate nuclear shielding constants.......Nuclear shielding calculations are presented for multiconfigurational self-consistent field wave functions using London atomic orbitals (gauge invariant atomic orbitals). Calculations of nuclear shieldings for eight molecules (H2O, H2S, CH4, N2, CO, HF, F2, and SO2) are presented and compared...

New developments in resonant mixture self-shielding treatment with Apollo code and application to Jules Horowitz reactor core calculation

International Nuclear Information System (INIS)

Coste-Delclaux, M.; Aggery, A.; Huot, N.

2005-01-01

APOLLO2 is a modular multigroup transport code developed by Cea in Saclay. Until last year, the self-shielding module could only treat one resonant isotope mixed with moderator isotopes. Consequently, the resonant mixture self-shielding treatment was an iterative one. Each resonant isotope of the mixture was treated separately, the other resonant isotopes of the mixture being then considered as moderator isotopes, that is to say non-resonant isotopes. This treatment could be iterated. Last year, we have developed a new method that consists in treating the resonant mixture as a unique entity. A main feature of APOLLO2 self-shielding module is that some implemented models are very general and therefore very powerful and versatile. We can give, as examples, the use of probability tables in order to describe the microscopic cross-section fluctuations or the TR slowing-down model that can deal with any resonance shape. The self-shielding treatment of a resonant mixture was developed essentially thanks to these two models. The calculations of a simplified Jules Horowitz reactor using a Monte-Carlo code (TRIPOLI4) as a reference and APOLLO2 in its standard and improved versions, show that, as far as the effective multiplication factor is concerned, the mixture treatment does not bring an improvement, because the new treatment suppresses compensation between the reaction rate discrepancies. The discrepancy of 300 pcm that appears with the reference calculation is in accordance with the technical specifications of the Jules Horowitz reactor

Self-Shielding Of Transmission Lines

Energy Technology Data Exchange (ETDEWEB)

Christodoulou, Christos [Univ. of New Mexico, Albuquerque, NM (United States)

2017-03-01

The use of shielding to contend with noise or harmful EMI/EMR energy is not a new concept. An inevitable trade that must be made for shielding is physical space and weight. Space was often not as much of a painful design trade in older larger systems as they are in today’s smaller systems. Today we are packing in an exponentially growing number of functionality within the same or smaller volumes. As systems become smaller and space within systems become more restricted, the implementation of shielding becomes more problematic. Often, space that was used to design a more mechanically robust component must be used for shielding. As the system gets smaller and space is at more of a premium, the trades starts to result in defects, designs with inadequate margin in other performance areas, and designs that are sensitive to manufacturing variability. With these challenges in mind, it would be ideal to maximize attenuation of harmful fields as they inevitably couple onto transmission lines without the use of traditional shielding. Dr. Tom Van Doren proposed a design concept for transmission lines to a class of engineers while visiting New Mexico. This design concept works by maximizing Electric field (E) and Magnetic Field (H) field containment between operating transmission lines to achieve what he called “Self-Shielding”. By making the geometric centroid of the outgoing current coincident with the return current, maximum field containment is achieved. The reciprocal should be true as well, resulting in greater attenuation of incident fields. Figure’s 1(a)-1(b) are examples of designs where the current centroids are coincident. Coax cables are good examples of transmission lines with co-located centroids but they demonstrate excellent field attenuation for other reasons and can’t be used to test this design concept. Figure 1(b) is a flex circuit design that demonstrate the implementation of self-shielding vs a standard conductor layout.

Recent Improvements in the SHIELD-HIT Code

DEFF Research Database (Denmark)

Hansen, David Christoffer; Lühr, Armin Christian; Herrmann, Rochus

2012-01-01

Purpose: The SHIELD-HIT Monte Carlo particle transport code has previously been used to study a wide range of problems for heavy-ion treatment and has been benchmarked extensively against other Monte Carlo codes and experimental data. Here, an improved version of SHIELD-HIT is developed concentra......Purpose: The SHIELD-HIT Monte Carlo particle transport code has previously been used to study a wide range of problems for heavy-ion treatment and has been benchmarked extensively against other Monte Carlo codes and experimental data. Here, an improved version of SHIELD-HIT is developed...

MPACT Subgroup Self-Shielding Efficiency Improvements

International Nuclear Information System (INIS)

Stimpson, Shane; Liu, Yuxuan; Collins, Benjamin S.; Clarno, Kevin T.

2016-01-01

Recent developments to improve the efficiency of the MOC solvers in MPACT have yielded effective kernels that loop over several energy groups at once, rather that looping over one group at a time. These kernels have produced roughly a 2x speedup on the MOC sweeping time during eigenvalue calculation. However, the self-shielding subgroup calculation had not been reevaluated to take advantage of these new kernels, which typically requires substantial solve time. The improvements covered in this report start by integrating the multigroup kernel concepts into the subgroup calculation, which are then used as the basis for further extensions. The next improvement that is covered is what is currently being termed as ''Lumped Parameter MOC''. Because the subgroup calculation is a purely fixed source problem and multiple sweeps are performed only to update the boundary angular fluxes, the sweep procedure can be condensed to allow for the instantaneous propagation of the flux across a spatial domain, without the need to sweep along all segments in a ray. Once the boundary angular fluxes are considered to be converged, an additional sweep that will tally the scalar flux is completed. The last improvement that is investigated is the possible reduction of the number of azimuthal angles per octant in the shielding sweep. Typically 16 azimuthal angles per octant are used for self-shielding and eigenvalue calculations, but it is possible that the self-shielding sweeps are less sensitive to the number of angles than the full eigenvalue calculation.

A Monte Carlo study for the shielding of γ backgrounds induced by radionuclides for CDEX

International Nuclear Information System (INIS)

Li Lei; Tang Changjian; Yue Qian; Cheng Jianping; Kang Kejun; Li Jianmin; Li Jin; Li Yulan; Li Yuanjing; Ma Hao; Xue Tao; Zeng Zhi; Wong, H.T.

2011-01-01

The CDEX (China Dark matter EXperiment) Collaboration will carry out a direct search for WIMPs (Weakly Interacting Massive Particles) using an Ultra-Low Energy Threshold High Purity Germanium (ULE-HPGe) detector at the CJPL (China JinPing deep underground Laboratory). A complex shielding system was designed to reduce backgrounds and a detailed GEANT4 Monte Carlo simulation was performed to study the achievable reduction of γ rays induced by radionuclides and neutron backgrounds by D(γ,n)p reaction. Furthermore, the upper level of allowed radio purity of shielding materials was estimated under the constraint of the expected goal. Compared with the radio purity reported by other low-background rare-event experiments, it indicates that the shielding used in the CDEX can be made out of materials with obtainable radiopurity. (authors)

A study on optimization of photoneutron shielding in a medical accelerator room by using Monte Carlo simulation

International Nuclear Information System (INIS)

Kim, Yong Nam; Jeong, Kyoungkeun; Kim, Joo Young; Lee, Chang Geol; Seong, Jinsil; Choi, Sang Hyun; Kim, Chan Hyeong

2008-01-01

Medical linear accelerators operating above 10 MV require door shielding for neutrons in addition to photons. A criterion for choice of optimal configuration of lamination of BPE (Borated Polyethylene) and lead is not clear. Moreover, optimal configuration cannot be determined by the conventional method using an analytical formula and simple measurement. This study performs Monte Carlo simulation of radiation field in a commercial LINAC room with 15 MV X-ray sources. Considering two configuration of lamination such as 'lead-BPE' and 'lead-BPE-lead', dose equivalents are calculated by using the MCNPX code and comparative analyses are performed with each other. The obtained results show that there is no significant difference in neuron shielding between both configurations, whereas lead-BPE-lead is more effective for photon shielding. It is also noted that the absolute values of neutron doses are much greater than that of photon doses outside as well as inside the door, by three orders of magnitude. As a conclusion, the laminating of lead-BPE is suggested as the optimal configuration from the viewpoint of simplicity in fabrication and handling, even though it has no significant difference from lead-BPE-lead in terms of total dose equivalent. (author)

New Improvements in Mixture Self-Shielding Treatment with APOLLO2 Code

International Nuclear Information System (INIS)

Coste-Delclaux, M.

2006-01-01

Full text of the presentation follows: APOLLO2 is a modular multigroup transport code developed at the CEA in Saclay (France). Previously, the self-shielding module could only treat one resonant isotope mixed with moderator isotopes. Consequently, the resonant mixture self-shielding treatment was an iterative one. Each resonant isotope of the mixture was treated separately, the other resonant isotopes of the mixture being then considered as moderator isotopes, that is to say non-resonant isotopes. This treatment could be iterated. Recently, we have developed a new method that consists in treating the resonant mixture as a unique entity. A main feature of APOLLO2 self-shielding module is that some implemented models are very general and therefore very powerful and versatile. We can give, as examples, the use of probability tables in order to describe the microscopic cross-section fluctuations or the TR slowing-down model that can deal with any resonance shape. The self-shielding treatment of a resonant mixture was developed essentially thanks to these two models. The goal of this paper is to describe the improvements on the self-shielding treatment of a resonant mixture and to present, as an application, the calculation of the ATRIUM-10 BWR benchmark. We will conclude by some prospects on remaining work in the self-shielding domain. (author)

The resonance self-shielding calculation with regularized random ladders

International Nuclear Information System (INIS)

Ribon, P.

1986-01-01

The straightforward method for calculation of resonance self-shielding is to generate one or several resonance ladders, and to process them as resolved resonances. The main drawback of Monte Carlo methods used to generate the ladders, is the difficulty of reducing the dispersion of data and results. Several methods are examined, and it is shown how one (a regularized sampling method) improves the accuracy. Analytical methods to compute the effective cross-section have recently appeared: they are basically exempt from dispersion, but are inevitably approximate. The accuracy of the most sophisticated one is checked. There is a neutron energy range which is improperly considered as statistical. An examination is presented of what happens when it is treated as statistical, and how it is possible to improve the accuracy of calculations in this range. To illustrate the results calculations have been performed in a simple case: nucleus 238 U, at 300 K, between 4250 and 4750 eV. (author)

Monte Carlo model for a prototype CT-compatible, anatomically adaptive, shielded intracavitary brachytherapy applicator for the treatment of cervical cancer

Energy Technology Data Exchange (ETDEWEB)

Price, Michael J.; Gifford, Kent A.; Horton, John L. Jr.; Eifel, Patricia J.; Gillin, Michael T.; Lawyer, Ann A.; Mourtada, Firas [Department of Radiation Physics, University of Texas M. D. Anderson Cancer Center, 1220 Holcombe Boulevard, Houston, Texas 77030 and Graduate School of Biomedical Sciences, University of Texas-Houston, 6767 Bertner Avenue, Houston, Texas 77030 (United States); Department of Radiation Physics, University of Texas M. D. Anderson Cancer Center, 1220 Holcombe Boulevard, Houston, Texas 77030 (United States); Division of Radiation Oncology, University of Texas M. D. Anderson Cancer Center, 1220 Holcombe Boulevard, Houston, Texas 77030 and Graduate School of Biomedical Sciences, University of Texas-Houston, 6767 Bertner Avenue, Houston, Texas 77030 (United States); Department of Radiation Physics, University of Texas M. D. Anderson Cancer Center, 1220 Holcombe Boulevard, Houston, Texas 77030 and Graduate School of Biomedical Sciences, University of Texas-Houston, 6767 Bertner Avenue, Houston, Texas 77030 (United States); Department of Radiation Physics, University of Texas M. D. Anderson Cancer Center, 1220 Holcombe Boulevard, Houston, Texas 77030 (United States); Department of Radiation Physics, University of Texas M. D. Anderson Cancer Center, 1220 Holcombe Boulevard, Houston, Texas 77030 and Graduate School of Biomedical Sciences, University of Texas-Houston, 6767 Bertner Avenue, Houston, Texas 77030 (United States)

2009-09-15

Purpose: Current, clinically applicable intracavitary brachytherapy applicators that utilize shielded ovoids contain a pair of tungsten-alloy shields which serve to reduce dose delivered to the rectum and bladder during source afterloading. After applicator insertion, these fixed shields are not necessarily positioned to provide optimal shielding of these critical structures due to variations in patient anatomies. The authors present a dosimetric evaluation of a novel prototype intracavitary brachytherapy ovoid [anatomically adaptive applicator (A{sup 3})], featuring a single shield whose position can be adjusted with two degrees of freedom: Rotation about and translation along the long axis of the ovoid. Methods: The dosimetry of the device for a HDR {sup 192}Ir was characterized using radiochromic film measurements for various shield orientations. A MCNPX Monte Carlo model was developed of the prototype ovoid and integrated with a previously validated model of a v2 mHDR {sup 192}Ir source (Nucletron Co.). The model was validated for three distinct shield orientations using film measurements. Results: For the most complex case, 91% of the absolute simulated and measured dose points agreed within 2% or 2 mm and 96% agreed within 10% or 2 mm. Conclusions: Validation of the Monte Carlo model facilitates future investigations into any dosimetric advantages the use of the A{sup 3} may have over the current state of art with respect to optimization and customization of dose delivery as a function of patient anatomical geometries.

Monte Carlo model for a prototype CT-compatible, anatomically adaptive, shielded intracavitary brachytherapy applicator for the treatment of cervical cancer

International Nuclear Information System (INIS)

Price, Michael J.; Gifford, Kent A.; Horton, John L. Jr.; Eifel, Patricia J.; Gillin, Michael T.; Lawyer, Ann A.; Mourtada, Firas

2009-01-01

Purpose: Current, clinically applicable intracavitary brachytherapy applicators that utilize shielded ovoids contain a pair of tungsten-alloy shields which serve to reduce dose delivered to the rectum and bladder during source afterloading. After applicator insertion, these fixed shields are not necessarily positioned to provide optimal shielding of these critical structures due to variations in patient anatomies. The authors present a dosimetric evaluation of a novel prototype intracavitary brachytherapy ovoid [anatomically adaptive applicator (A 3 )], featuring a single shield whose position can be adjusted with two degrees of freedom: Rotation about and translation along the long axis of the ovoid. Methods: The dosimetry of the device for a HDR 192 Ir was characterized using radiochromic film measurements for various shield orientations. A MCNPX Monte Carlo model was developed of the prototype ovoid and integrated with a previously validated model of a v2 mHDR 192 Ir source (Nucletron Co.). The model was validated for three distinct shield orientations using film measurements. Results: For the most complex case, 91% of the absolute simulated and measured dose points agreed within 2% or 2 mm and 96% agreed within 10% or 2 mm. Conclusions: Validation of the Monte Carlo model facilitates future investigations into any dosimetric advantages the use of the A 3 may have over the current state of art with respect to optimization and customization of dose delivery as a function of patient anatomical geometries.

Self-shielding for thick slabs in a converging neutron beam

CERN Document Server

Mildner, D F R

1999-01-01

We have previously given a correction to the neutron self-shielding for a thin slab to account for the increased average path length through the slab when irradiated in a converging neutron beam. This expression overstates the case for the self-shielding for a thick (or highly absorbing) slab. We give a better approximation to the increase in effective shielding correction for a slab placed in a converging neutron beam. It is negligible at large absorption mean free paths. (author)

Radiation shielding calculation using MCNP

International Nuclear Information System (INIS)

Masukawa, Fumihiro

2001-01-01

To verify the Monte Carlo code MCNP4A as a tool to generate the reference data in the shielding designs and the safety evaluations, various shielding benchmark experiments were analyzed using this code. These experiments were categorized in three types of the shielding subjects; bulk shielding, streaming, and skyshine. For the variance reduction technique, which is indispensable to get meaningful results with the Monte Carlo shielding calculation, we mainly used the weight window, the energy dependent Russian roulette and spitting. As a whole, our analyses performed enough small statistical errors and showed good agreements with these experiments. (author)

Radiation shielding of the main injector

International Nuclear Information System (INIS)

Bhat, C.M.; Martin, P.S.

1995-05-01

The radiation shielding in the Fermilab Main Injector (FMI) complex has been carried out by adopting a number of prescribed stringent guidelines established by a previous safety analysis. Determination of the required amount of radiation shielding at various locations of the FMI has been done using Monte Carlo computations. A three dimensional ray tracing code as well as a code based upon empirical observations have been employed in certain cases

The self shielding module of Apollo.II; Module d`autoprotection du code Apollo.II

Energy Technology Data Exchange (ETDEWEB)

Sanchez, R.

1994-06-01

This note discusses the methods used in the APOLLO.II code for the calculation of self shielded multigroup cross sections. Basically, the calculation consists in characterizing a heterogenous medium with a single parameter: the background cross section, which is in then used to interpolate reaction rates from pre tabulated values. Very fine multigroup slowing down calculations in homogenous media are used to generate these tables, which contain absorption, diffusion and production reaction rates per group, resonant isotope, temperature and background cross section. Multigroup self shielded cross sections are determined from an equivalence that preserves absorption rates at a slowing down problem with given sources. This article gives a detailed description of the PIC and ``dilution matrix`` formalisms that are used in the homogenization step, as well as the utilization of Bell macro-groups and the different quadrature formulas that may be used in the calculations. Self shielding techniques for isotopic resonant mixtures are also discussed. (author). 2 refs., 193 figs., 2 tabs.

Nucleonic analysis of a preliminary design for the ETF neutral-beam-injector duct shielding

International Nuclear Information System (INIS)

Urban, W.T.; Seed, T.J.; Dudziak, D.J.

1980-01-01

A nucleonic analysis of the Engineering Test Facility Neutral-Beam-Injector duct shielding has been made using a hybrid Monte Carlo/discrete-ordinates method. This method used Monte Carlo to determine internal and external boundary surface sources for a subsequent discrete-ordinates calculation of the neutron and gamma-ray transport through the shield. The analysis also included determination of the energy and angular distribution of neutrons and gamma rays entering the duct from the torus plasma chamber. Confidence in the hybrid method and the results obtained were provided through a comparison with three-dimensional Monte Carlo results

The problem of resonance self-shielding effect in neutron multigroup calculations

International Nuclear Information System (INIS)

Wang Qingming; Huang Jinghua

1991-01-01

It is not allowed to neglect the resonance self-shielding effect in hybrid blanket and fast reactor neutron designs. The authors discussed the importance as well as the method of considering the resonance self-shielding effect in hybrid blanket and fast reactor neutron multigroup calculations

Self-shielding of hydrogen in the IGM during the epoch of reionization

Science.gov (United States)

Chardin, Jonathan; Kulkarni, Girish; Haehnelt, Martin G.

2018-04-01

We investigate self-shielding of intergalactic hydrogen against ionizing radiation in radiative transfer simulations of cosmic reionization carefully calibrated with Lyα forest data. While self-shielded regions manifest as Lyman-limit systems in the post-reionization Universe, here we focus on their evolution during reionization (redshifts z = 6-10). At these redshifts, the spatial distribution of hydrogen-ionizing radiation is highly inhomogeneous, and some regions of the Universe are still neutral. After masking the neutral regions and ionizing sources in the simulation, we find that the hydrogen photoionization rate depends on the local hydrogen density in a manner very similar to that in the post-reionization Universe. The characteristic physical hydrogen density above which self-shielding becomes important at these redshifts is about nH ˜ 3 × 10-3 cm-3, or ˜20 times the mean hydrogen density, reflecting the fact that during reionization photoionization rates are typically low enough that the filaments in the cosmic web are often self-shielded. The value of the typical self-shielding density decreases by a factor of 3 between redshifts z = 3 and 10, and follows the evolution of the average photoionization rate in ionized regions in a simple fashion. We provide a simple parameterization of the photoionization rate as a function of density in self-shielded regions during the epoch of reionization.

Re-Shielding of Cobalt-60 Teletherapy Rooms for Tomotherapy and Conventional Linear Accelerators using Monte Carlo Simulations

Science.gov (United States)

Çeçen, Yiğit; Yazgan, Çağrı

2017-09-01

Purpose. Nearly all Cobalt-60 teletherapy machines were removed around the world during the last two decades. The remaining ones are being used for experimental purposes. However, the rooms of these teletherapy machines are valuable because of lack of space in radiotherapy clinics. In order to place a new technology treatment machine in one of these rooms, one should re-shield the room since it was designed only for 1.25 MeV gamma beams on average. Mostly, the vendor of the new machine constructs the new shielding of the room using their experience. However, every radiotherapy room has different surrounding work areas and it would be wise to shield the room considering these special conditions. Also, the shield design goal of the clinic may be much lower than the International Atomic Energy Agency (IAEA) or the local association accepts. The study shows re-shielding of a Cobalt-60 room, specific to the clinic, using Monte Carlo simulations. Materials & Methods: First, a 6 MV Tomotherapy machine, then a 10 MV conventional linear accelerator (LINAC) was placed inside the Cobalt-60 teletherapy room. The photon flux outside the room was simulated using Monte Carlo N-Particle (MCNP6.1) code before and after re-shielding. For the Tomotherapy simulation, flux distributions around the machine were obtained from the vendor and implemented as the source of the model. The LINAC model was more generic with the 10 MeV electron source, the tungsten target, first and secondary collimators. The aim of the model was to obtain the maximum (40x40 cm2) open field at the isocenter. Two different simulations were carried out for gantry angles 90o and 270o. The LINAC was placed in the room such that the primary walls were A' (Gantry 270o) and C' (Gantry 90o) (figure 1). The second part of the study was to model the re-shielding of the room for Tomotherapy and for the conventional LINAC, separately. The aim was to investigate the recommended shielding by the vendors. Left side of the room

Monte Carlo simulations for the space radiation superconducting shield project (SR2S).

Science.gov (United States)

Vuolo, M; Giraudo, M; Musenich, R; Calvelli, V; Ambroglini, F; Burger, W J; Battiston, R

2016-02-01

Astronauts on deep-space long-duration missions will be exposed for long time to galactic cosmic rays (GCR) and Solar Particle Events (SPE). The exposure to space radiation could lead to both acute and late effects in the crew members and well defined countermeasures do not exist nowadays. The simplest solution given by optimized passive shielding is not able to reduce the dose deposited by GCRs below the actual dose limits, therefore other solutions, such as active shielding employing superconducting magnetic fields, are under study. In the framework of the EU FP7 SR2S Project - Space Radiation Superconducting Shield--a toroidal magnetic system based on MgB2 superconductors has been analyzed through detailed Monte Carlo simulations using Geant4 interface GRAS. Spacecraft and magnets were modeled together with a simplified mechanical structure supporting the coils. Radiation transport through magnetic fields and materials was simulated for a deep-space mission scenario, considering for the first time the effect of secondary particles produced in the passage of space radiation through the active shielding and spacecraft structures. When modeling the structures supporting the active shielding systems and the habitat, the radiation protection efficiency of the magnetic field is severely decreasing compared to the one reported in previous studies, when only the magnetic field was modeled around the crew. This is due to the large production of secondary radiation taking place in the material surrounding the habitat. Copyright © 2016 The Committee on Space Research (COSPAR). Published by Elsevier Ltd. All rights reserved.

Enhancement of thermal neutron self-shielding in materials surrounded by reflectors

International Nuclear Information System (INIS)

Cornelia Chilian; Gregory Kennedy

2012-01-01

Materials containing from 41 to 1124 mg chlorine and surrounded by polyethylene containers of various thicknesses, from 0.01 to 5.6 mm, were irradiated in a research reactor neutron spectrum and the 38 Cl activity produced was measured as a function of polyethylene reflector thickness. For the material containing the higher amount of chlorine, the 38 Cl specific activity decreased with increasing reflector thickness, indicating increased neutron self-shielding. It was found that the amount of neutron self-shielding increased by as much as 52% with increasing reflector thickness. This is explained by neutrons which have exited the material subsequently reflecting back into it and thus increasing the total mean path length in the material. All physical and empirical models currently used to predict neutron self-shielding have ignored this effect and need to be modified. A method is given for measuring the adjustable parameter of a self-shielding model for a particular sample size and combination of neutron reflectors. (author)

Resonance Self-Shielding Methodologies in SCALE 6

International Nuclear Information System (INIS)

Williams, Mark L.

2011-01-01

SCALE 6 includes several problem-independent multigroup (MG) libraries that were processed from the evaluated nuclear data file ENDF/B using a generic flux spectrum. The library data must be self-shielded and corrected for problem-specific spectral effects for use in MG neutron transport calculations. SCALE 6 computes problem-dependent MG cross sections through a combination of the conventional Bondarenko shielding-factor method and a deterministic continuous-energy (CE) calculation of the fine-structure spectra in the resolved resonance and thermal energy ranges. The CE calculation can be performed using an infinite medium approximation, a simplified two-region method for lattices, or a one-dimensional discrete ordinates transport calculation with pointwise (PW) cross-section data. This paper describes the SCALE-resonance self-shielding methodologies, including the deterministic calculation of the CE flux spectra using PW nuclear data and the method for using CE spectra to produce problem-specific MG cross sections for various configurations (including doubly heterogeneous lattices). It also presents results of verification and validation studies.

SU-E-T-556: Monte Carlo Generated Dose Distributions for Orbital Irradiation Using a Single Anterior-Posterior Electron Beam and a Hanging Lens Shield

International Nuclear Information System (INIS)

Duwel, D; Lamba, M; Elson, H; Kumar, N

2015-01-01

Purpose: Various cancers of the eye are successfully treated with radiotherapy utilizing one anterior-posterior (A/P) beam that encompasses the entire content of the orbit. In such cases, a hanging lens shield can be used to spare dose to the radiosensitive lens of the eye to prevent cataracts. Methods: This research focused on Monte Carlo characterization of dose distributions resulting from a single A-P field to the orbit with a hanging shield in place. Monte Carlo codes were developed which calculated dose distributions for various electron radiation energies, hanging lens shield radii, shield heights above the eye, and beam spoiler configurations. Film dosimetry was used to benchmark the coding to ensure it was calculating relative dose accurately. Results: The Monte Carlo dose calculations indicated that lateral and depth dose profiles are insensitive to changes in shield height and electron beam energy. Dose deposition was sensitive to shield radius and beam spoiler composition and height above the eye. Conclusion: The use of a single A/P electron beam to treat cancers of the eye while maintaining adequate lens sparing is feasible. Shield radius should be customized to have the same radius as the patient’s lens. A beam spoiler should be used if it is desired to substantially dose the eye tissues lying posterior to the lens in the shadow of the lens shield. The compromise between lens sparing and dose to diseased tissues surrounding the lens can be modulated by varying the beam spoiler thickness, spoiler material composition, and spoiler height above the eye. The sparing ratio is a metric that can be used to evaluate the compromise between lens sparing and dose to surrounding tissues. The higher the ratio, the more dose received by the tissues immediately posterior to the lens relative to the dose received by the lens

Buildup factors for multilayer shieldings in deterministic methods and their comparison with Monte Carlo

International Nuclear Information System (INIS)

Listjak, M.; Slavik, O.; Kubovcova, D.; Vermeersch, F.

2008-01-01

In general there are two ways how to calculate effective doses. The first way is by use of deterministic methods like point kernel method which is implemented in Visiplan or Microshield. These kind of calculations are very fast, but they are not very convenient for a complex geometry with shielding composed of more then one material in meaning of result precision. In spite of this that programs are sufficient for ALARA optimisation calculations. On other side there are Monte Carlo methods which can be used for calculations. This way of calculation is quite precise in comparison with reality but calculation time is usually very large. Deterministic method like programs have one disadvantage -usually there is option to choose buildup factor (BUF) only for one material in multilayer stratified slabs shielding calculation problems even if shielding is composed from different materials. In literature there are proposed different formulas for multilayer BUF approximation. Aim of this paper was to examine these different formulas and their comparison with MCNP calculations. At first ware compared results of Visiplan and Microshield. Simple geometry was modelled - point source behind single and double slab shielding. For Build-up calculations was chosen Geometric Progression method (feature of the newest version of Visiplan) because there are lower deviations in comparison with Taylor fitting. (authors)

Buildup factors for multilayer shieldings in deterministic methods and their comparison with Monte Carlo

International Nuclear Information System (INIS)

Listjak, M.; Slavik, O.; Kubovcova, D.; Vermeersch, F.

2009-01-01

In general there are two ways how to calculate effective doses. The first way is by use of deterministic methods like point kernel method which is implemented in Visiplan or Microshield. These kind of calculations are very fast, but they are not very convenient for a complex geometry with shielding composed of more then one material in meaning of result precision. In spite of this that programs are sufficient for ALARA optimisation calculations. On other side there are Monte Carlo methods which can be used for calculations. This way of calculation is quite precise in comparison with reality but calculation time is usually very large. Deterministic method like programs have one disadvantage -usually there is option to choose buildup factor (BUF) only for one material in multilayer stratified slabs shielding calculation problems even if shielding is composed from different materials. In literature there are proposed different formulas for multilayer BUF approximation. Aim of this paper was to examine these different formulas and their comparison with MCNP calculations. At first ware compared results of Visiplan and Microshield. Simple geometry was modelled - point source behind single and double slab shielding. For Build-up calculations was chosen Geometric Progression method (feature of the newest version of Visiplan) because there are lower deviations in comparison with Taylor fitting. (authors)

Revisiting the stamm'ler self-shielding method

International Nuclear Information System (INIS)

Hebert, A.

2004-01-01

The generalized Stamm'ler method is been used in lattice codes such as PHOENIX, WIMS-AECL and DRAGON-IST for computing self-shielded cross sections, prior to the main flux calculation. This method is handicapped by deficiencies, such as its low accuracy and its inability to represent distributed self-shielding effects in a fuel rod or across a fuel bundle. The paper describes improvements that could be made to the generalized Stamm'ler method in order to mitigate these two defects. A validation is presented for the case of 238 U nuclides located in different geometries. The isotopic absorption rates obtained with the proposed numerical scheme are compared with exact values obtained with a fine-group elastic slowing-down calculation in the resolved energy domain. (author)

A Monte Carlo Method for the Analysis of Gamma Radiation Transport from Distributed Sources in Laminated Shields

Energy Technology Data Exchange (ETDEWEB)

Leimdoerfer, M

1964-02-15

A description is given of a method for calculating the penetration and energy deposition of gamma radiation, based on Monte Carlo techniques. The essential feature is the application of the exponential transformation to promote the transport of penetrating quanta and to balance the steep spatial variations of the source distributions which appear in secondary gamma emission problems. The estimated statistical errors in a number of sample problems, involving concrete shields with thicknesses up to 500 cm, are shown to be quite favorable, even at relatively short computing times. A practical reactor shielding problem is also shown and the predictions compared with measurements.

A Monte Carlo Method for the Analysis of Gamma Radiation Transport from Distributed Sources in Laminated Shields

International Nuclear Information System (INIS)

Leimdoerfer, M.

1964-02-01

A description is given of a method for calculating the penetration and energy deposition of gamma radiation, based on Monte Carlo techniques. The essential feature is the application of the exponential transformation to promote the transport of penetrating quanta and to balance the steep spatial variations of the source distributions which appear in secondary gamma emission problems. The estimated statistical errors in a number of sample problems, involving concrete shields with thicknesses up to 500 cm, are shown to be quite favorable, even at relatively short computing times. A practical reactor shielding problem is also shown and the predictions compared with measurements

Shielding benchmark problems

International Nuclear Information System (INIS)

Tanaka, Shun-ichi; Sasamoto, Nobuo; Oka, Yoshiaki; Kawai, Masayoshi; Nakazawa, Masaharu.

1978-09-01

Shielding benchmark problems were prepared by the Working Group of Assessment of Shielding Experiments in the Research Comittee on Shielding Design of the Atomic Energy Society of Japan, and compiled by the Shielding Laboratory of Japan Atomic Energy Research Institute. Twenty-one kinds of shielding benchmark problems are presented for evaluating the calculational algorithm and the accuracy of computer codes based on the discrete ordinates method and the Monte Carlo method and for evaluating the nuclear data used in the codes. (author)

Efficient heterogeneous execution of Monte Carlo shielding calculations on a Beowulf cluster

International Nuclear Information System (INIS)

Dewar, D.; Hulse, P.; Cooper, A.; Smith, N.

2005-01-01

Recent work has been done in using a high-performance 'Beowulf' cluster computer system for the efficient distribution of Monte Carlo shielding calculations. This has enabled the rapid solution of complex shielding problems at low cost and with greater modularity and scalability than traditional platforms. The work has shown that a simple approach to distributing the workload is as efficient as using more traditional techniques such as PVM (Parallel Virtual Machine). In addition, when used in an operational setting this technique is fairer with the use of resources than traditional methods, in that it does not tie up a single computing resource but instead shares the capacity with other tasks. These developments in computing technology have enabled shielding problems to be solved that would have taken an unacceptably long time to run on traditional platforms. This paper discusses the BNFL Beowulf cluster and a number of tests that have recently been run to demonstrate the efficiency of the asynchronous technique in running the MCBEND program. The BNFL Beowulf currently consists of 84 standard PCs running RedHat Linux. Current performance of the machine has been estimated to be between 40 and 100 Gflop s -1 . When the whole system is employed on one problem up to four million particles can be tracked per second. There are plans to review its size in line with future business needs. (authors)

Performance of the improved version of Monte Carlo Code A3MCNP for cask shielding design

International Nuclear Information System (INIS)

Hasegawa, T.; Ueki, K.; Sato, O.; Sjoden, G.E.; Miyake, Y.; Ohmura, M.; Haghighat, A.

2004-01-01

A 3 MCNP (Automatic Adjoint Accelerated MCNP) is a revised version of the MCNP Monte Carlo code, that automatically prepares variance reduction parameters for the CADIS (Consistent Adjoint Driven Importance Sampling) methodology. Using a deterministic ''importance'' (or adjoint) function, CADIS performs source and transport biasing within the weight-window technique. The current version of A 3 MCNP uses the 3-D Sn transport TORT code to determine a 3-D importance function distribution. Based on simulation of several real-life problems, it is demonstrated that A3MCNP provides precise calculation results with a remarkably short computation time by using the proper and objective variance reduction parameters. However, since the first version of A 3 MCNP provided only a point source configuration option for large-scale shielding problems, such as spent-fuel transport casks, a large amount of memory may be necessary to store enough points to properly represent the source. Hence, we have developed an improved version of A 3 MCNP (referred to as A 3 MCNPV) which has a volumetric source configuration option. This paper describes the successful use of A 3 MCNPV for cask neutron and gamma-ray shielding problem

Neutron self-shielding with k0-NAA irradiations

International Nuclear Information System (INIS)

Chilian, C.; Chambon, R.; Kennedy, G.

2010-01-01

A sample of SMELS Type II reference material was mixed with powdered Cd-nitrate neutron absorber and analysed by k 0 NAA for 10 elements. The thermal neutron self-shielding effect was found to be 34.8%. When flux monitors were irradiated sufficiently far from the absorbing sample, it was found that the self-shielding could be corrected accurately using an analytical formula and an iterative calculation. When the flux monitors were irradiated 2 mm from the absorbing sample, the calculations over-corrected the concentrations by as much as 30%. It is recommended to irradiate flux monitors at least 14 mm from a 10 mm diameter absorbing sample.

URR-PACK: Calculating Self-Shielding in the Unresolved Resonance Energy Range

International Nuclear Information System (INIS)

Cullen, Dermott E.; Trkov, Andrej

2016-07-01

This report describes HOW to calculate self-shielding in the unresolved resonance region (URR), in terms of the computer codes we provide to allow a user to do these calculations himself. Here we only describe HOW to calculate; a longer companion report describes in detail WHY it is necessary to include URR self-shielding.

Neutron shielding calculations in a proton therapy facility based on Monte Carlo simulations and analytical models: Criterion for selecting the method of choice

International Nuclear Information System (INIS)

Titt, U.; Newhauser, W. D.

2005-01-01

Proton therapy facilities are shielded to limit the amount of secondary radiation to which patients, occupational workers and members of the general public are exposed. The most commonly applied shielding design methods for proton therapy facilities comprise semi-empirical and analytical methods to estimate the neutron dose equivalent. This study compares the results of these methods with a detailed simulation of a proton therapy facility by using the Monte Carlo technique. A comparison of neutron dose equivalent values predicted by the various methods reveals the superior accuracy of the Monte Carlo predictions in locations where the calculations converge. However, the reliability of the overall shielding design increases if simulation results, for which solutions have not converged, e.g. owing to too few particle histories, can be excluded, and deterministic models are being used at these locations. Criteria to accept or reject Monte Carlo calculations in such complex structures are not well understood. An optimum rejection criterion would allow all converging solutions of Monte Carlo simulation to be taken into account, and reject all solutions with uncertainties larger than the design safety margins. In this study, the optimum rejection criterion of 10% was found. The mean ratio was 26, 62% of all receptor locations showed a ratio between 0.9 and 10, and 92% were between 1 and 100. (authors)

Development and application of the automated Monte Carlo biasing procedure in SAS4

International Nuclear Information System (INIS)

Tang, J.S.; Broadhead, B.L.

1993-01-01

An automated approach for biasing Monte Carlo shielding calculations is described. In particular, adjoint fluxes from a one-dimensional discrete-ordinates calculation are used to generate biasing parameters for a three-dimensional Monte Carlo calculation. The automated procedure consisting of cross-section processing, adjoint flux determination, biasing parameter generation, and the initiation of a MORSE-SGC/S Monte Carlo calculation has been implemented in the SAS4 module of the SCALE computer code system. The automated procedure has been used extensively in the investigation of both computational and experimental benchmarks for the NEACRP working group on shielding assessment of transportation packages. The results of these studies indicate that with the automated biasing procedure, Monte Carlo shielding calculations of spent fuel casks can be easily performed with minimum effort and that accurate results can be obtained at reasonable computing cost. The systematic biasing approach described in this paper can also be applied to other similar shielding problems

Self shielding in cylindrical fissile sources in the APNea system

International Nuclear Information System (INIS)

Hensley, D.

1997-01-01

In order for a source of fissile material to be useful as a calibration instrument, it is necessary to know not only how much fissile material is in the source but also what the effective fissile content is. Because uranium and plutonium absorb thermal neutrons so Efficiently, material in the center of a sample is shielded from the external thermal flux by the surface layers of the material. Differential dieaway measurements in the APNea System of five different sets of cylindrical fissile sources show the various self shielding effects that are routinely encountered. A method for calculating the self shielding effect is presented and its predictions are compared with the experimental results

Theoretical evaluation of self-shielding factors due to scattering resonances in foils

International Nuclear Information System (INIS)

Selander, W.N.

1960-06-01

A semi-analytical method is given for evaluating self-shielding factors for activation measurements which use thin foils having neutron scattering resonances. The energy loss by scattering in the foil is taken into account. The energy-dependent neutron angular distribution is expanded as a double series, the coefficients of which are (energy dependent) solutions of an infinite set of coupled integral equations. These are truncated in some suitable manner and solved numerically. The leading term of the series is proportional to the average, or effective flux in the activation sample. The product of this terra and the neutron capture cross-section is integrated numerically over the resonance to give the resonance self-shielding correction. Figure 4 shows resonance self-shielding factors derived in this mariner for the 132ev resonance in Co-59 and figure 5 shows similar results for the two Mn-55 resonances at 337ev and 1080ev. Self-shielding factors for 1/v capture are not significantly different from unity. (author)

Absorbed dose kernel and self-shielding calculations for a novel radiopaque glass microsphere for transarterial radioembolization.

Science.gov (United States)

Church, Cody; Mawko, George; Archambault, John Paul; Lewandowski, Robert; Liu, David; Kehoe, Sharon; Boyd, Daniel; Abraham, Robert; Syme, Alasdair

2018-02-01

Radiopaque microspheres may provide intraprocedural and postprocedural feedback during transarterial radioembolization (TARE). Furthermore, the potential to use higher resolution x-ray imaging techniques as opposed to nuclear medicine imaging suggests that significant improvements in the accuracy and precision of radiation dosimetry calculations could be realized for this type of therapy. This study investigates the absorbed dose kernel for novel radiopaque microspheres including contributions of both short and long-lived contaminant radionuclides while concurrently quantifying the self-shielding of the glass network. Monte Carlo simulations using EGSnrc were performed to determine the dose kernels for all monoenergetic electron emissions and all beta spectra for radionuclides reported in a neutron activation study of the microspheres. Simulations were benchmarked against an accepted 90 Y dose point kernel. Self-shielding was quantified for the microspheres by simulating an isotropically emitting, uniformly distributed source, in glass and in water. The ratio of the absorbed doses was scored as a function of distance from a microsphere. The absorbed dose kernel for the microspheres was calculated for (a) two bead formulations following (b) two different durations of neutron activation, at (c) various time points following activation. Self-shielding varies with time postremoval from the reactor. At early time points, it is less pronounced due to the higher energies of the emissions. It is on the order of 0.4-2.8% at a radial distance of 5.43 mm with increased size from 10 to 50 μm in diameter during the time that the microspheres would be administered to a patient. At long time points, self-shielding is more pronounced and can reach values in excess of 20% near the end of the range of the emissions. Absorbed dose kernels for 90 Y, 90m Y, 85m Sr, 85 Sr, 87m Sr, 89 Sr, 70 Ga, 72 Ga, and 31 Si are presented and used to determine an overall kernel for the

REPOSITORY LAYOUT SUPPORTING DESIGN FEATURE NO.13 - WASTE PACKAGE SELF SHIELDING

International Nuclear Information System (INIS)

Owen, J.

1999-01-01

The objective of this analysis is to develop a repository layout, for Feature No. 13, that will accommodate self-shielding waste packages (WP) with an areal mass loading of 25 metric tons of uranium per acre (MTU/acre). The scope of this analysis includes determination of the number of emplacement drifts, amount of emplacement drift excavation required, and a preliminary layout for illustrative purposes

Monte Carlo applications to radiation shielding problems

International Nuclear Information System (INIS)

Subbaiah, K.V.

2009-01-01

Monte Carlo methods are a class of computational algorithms that rely on repeated random sampling of physical and mathematical systems to compute their results. However, basic concepts of MC are both simple and straightforward and can be learned by using a personal computer. Uses of Monte Carlo methods require large amounts of random numbers, and it was their use that spurred the development of pseudorandom number generators, which were far quicker to use than the tables of random numbers which had been previously used for statistical sampling. In Monte Carlo simulation of radiation transport, the history (track) of a particle is viewed as a random sequence of free flights that end with an interaction event where the particle changes its direction of movement, loses energy and, occasionally, produces secondary particles. The Monte Carlo simulation of a given experimental arrangement (e.g., an electron beam, coming from an accelerator and impinging on a water phantom) consists of the numerical generation of random histories. To simulate these histories we need an interaction model, i.e., a set of differential cross sections (DCS) for the relevant interaction mechanisms. The DCSs determine the probability distribution functions (pdf) of the random variables that characterize a track; 1) free path between successive interaction events, 2) type of interaction taking place and 3) energy loss and angular deflection in a particular event (and initial state of emitted secondary particles, if any). Once these pdfs are known, random histories can be generated by using appropriate sampling methods. If the number of generated histories is large enough, quantitative information on the transport process may be obtained by simply averaging over the simulated histories. The Monte Carlo method yields the same information as the solution of the Boltzmann transport equation, with the same interaction model, but is easier to implement. In particular, the simulation of radiation

Shielding features of quarry stone

International Nuclear Information System (INIS)

Hernandez V, C.; Contreras S, H.; Hernandez A, L.; Baltazar R, A.; Escareno J, E.; Mares E, C. A.; Vega C, H. R.

2010-10-01

Quarry stone lineal attenuation coefficient for gamma-rays has been obtained. In Zacatecas, quarry stone is widely utilized as a decorative item in buildings, however its shielding features against gamma-rays unknown. The aim of this work is to determine the shielding properties of quarry stone against γ-rays using Monte Carlo calculations where a detailed model of a good geometry experimental setup was carried out. In the calculations 10 pieces 10 X 10 cm 2 of different thickness were utilized to evaluate the photons transmission as the quarry stone thickness is increased. It was noticed that transmitted photons decay away as the shield thickness is increased, these results were fitted to an exponential function were the linear attenuation coefficient was estimated. Also, using XCOM code the linear attenuation coefficient from several keV up to 100 MeV was estimated. From the comparison between Monte Carlo results and XCOM calculations a good agreement was found. For 0.662 MeV γ-rays the attenuation coefficient of quarry stone, whose density is 2.413 g-cm -3 , is 0.1798 cm -1 , this mean a X 1/2 = 3.9 cm, X 1/4 = 7.7 cm, X 1/10 = 12.8 cm, and X 1/100 = 25.6 cm. Having the information of quarry stone performance as shielding give the chance to use this material to shield X and γ-ray facilities. (Author)

Onboard radiation shielding estimates for interplanetary manned missions

International Nuclear Information System (INIS)

Totemeier, A.; Jevremovic, T.; Hounshel, D.

2004-01-01

The main focus of space related shielding design is to protect operating systems, personnel and key structural components from outer space and onboard radiation. This paper summarizes the feasibility of a lightweight neutron radiation shield design for a nuclear powered, manned space vehicle. The Monte Carlo code MCNP5 is used to determine radiation transport characteristics of the different materials and find the optimized shield configuration. A phantom torso encased in air is used to determine a dose rate for a crew member on the ship. Calculation results indicate that onboard shield against neutron radiation coming from nuclear engine can be achieved with very little addition of weight to the space vehicle. The selection of materials and neutron transport analysis as presented in this paper are useful starting data to design shield against neutrons generated when high-energy particles from outer space interact with matter on the space vehicle. (authors)

Electron dose distributions caused by the contact-type metallic eye shield: Studies using Monte Carlo and pencil beam algorithms

Energy Technology Data Exchange (ETDEWEB)

Kang, Sei-Kwon; Yoon, Jai-Woong; Hwang, Taejin; Park, Soah; Cheong, Kwang-Ho; Jin Han, Tae; Kim, Haeyoung; Lee, Me-Yeon; Ju Kim, Kyoung, E-mail: kjkim@hallym.or.kr; Bae, Hoonsik

2015-10-01

A metallic contact eye shield has sometimes been used for eyelid treatment, but dose distribution has never been reported for a patient case. This study aimed to show the shield-incorporated CT-based dose distribution using the Pinnacle system and Monte Carlo (MC) calculation for 3 patient cases. For the artifact-free CT scan, an acrylic shield machined as the same size as that of the tungsten shield was used. For the MC calculation, BEAMnrc and DOSXYZnrc were used for the 6-MeV electron beam of the Varian 21EX, in which information for the tungsten, stainless steel, and aluminum material for the eye shield was used. The same plan was generated on the Pinnacle system and both were compared. The use of the acrylic shield produced clear CT images, enabling delineation of the regions of interest, and yielded CT-based dose calculation for the metallic shield. Both the MC and the Pinnacle systems showed a similar dose distribution downstream of the eye shield, reflecting the blocking effect of the metallic eye shield. The major difference between the MC and the Pinnacle results was the target eyelid dose upstream of the shield such that the Pinnacle system underestimated the dose by 19 to 28% and 11 to 18% for the maximum and the mean doses, respectively. The pattern of dose difference between the MC and the Pinnacle systems was similar to that in the previous phantom study. In conclusion, the metallic eye shield was successfully incorporated into the CT-based planning, and the accurate dose calculation requires MC simulation.

Study on shielding design method of radiation streaming in a tokamak-type DT fusion reactor based on Monte Carlo calculation

International Nuclear Information System (INIS)

Sato, Satoshi

2003-09-01

In tokamak-type DT nuclear fusion reactor, there are various type slits and ducts in the blanket and the vacuum vessel. The helium production in the rewelding location of the blanket and the vacuum vessel, the nuclear properties in the super-conductive TF coil, e.g. the nuclear heating rate in the coil winding pack, are enhanced by the radiation streaming through the slits and ducts, and they are critical concern in the shielding design. The decay gamma ray dose rate around the duct penetrating the blanket and the vacuum vessel is also enhanced by the radiation streaming through the duct, and they are also critical concern from the view point of the human access to the cryostat during maintenance. In order to evaluate these nuclear properties with good accuracy, three dimensional Monte Carlo calculation is required but requires long calculation time. Therefore, the development of the effective simple design evaluation method for radiation streaming is substantially important. This study aims to establish the systematic evaluation method for the nuclear properties of the blanket, the vacuum vessel and the Toroidal Field (TF) coil taking into account the radiation streaming through various types of slits and ducts, based on three dimensional Monte Carlo calculation using the MNCP code, and for the decay gamma ray dose rates penetrated around the ducts. The present thesis describes three topics in five chapters as follows; 1) In Chapter 2, the results calculated by the Monte Carlo code, MCNP, are compared with those by the Sn code, DOT3.5, for the radiation streaming in the tokamak-type nuclear fusion reactor, for validating the results of the Sn calculation. From this comparison, the uncertainties of the Sn calculation results coming from the ray-effect and the effect due to approximation of the geometry are investigated whether the two dimensional Sn calculation can be applied instead of the Monte Carlo calculation. Through the study, it can be concluded that the

A lumped parameter method of characteristics approach and multigroup kernels applied to the subgroup self-shielding calculation in MPACT

International Nuclear Information System (INIS)

Stimpson, Shane G.; Liu, Yuxuan; Collins, Benjamin S.; Clarno, Kevin T.

2017-01-01

An essential component of the neutron transport solver is the resonance self-shielding calculation used to determine equivalence cross sections. The neutron transport code, MPACT, is currently using the subgroup self-shielding method, in which the method of characteristics (MOC) is used to solve purely absorbing fixed-source problems. Recent efforts incorporating multigroup kernels to the MOC solvers in MPACT have reduced runtime by roughly 2×. Applying the same concepts for self-shielding and developing a novel lumped parameter approach to MOC, substantial improvements have also been made to the self-shielding computational efficiency without sacrificing any accuracy. These new multigroup and lumped parameter capabilities have been demonstrated on two test cases: (1) a single lattice with quarter symmetry known as VERA (Virtual Environment for Reactor Applications) Progression Problem 2a and (2) a two-dimensional quarter-core slice known as Problem 5a-2D. From these cases, self-shielding computational time was reduced by roughly 3–4×, with a corresponding 15–20% increase in overall memory burden. An azimuthal angle sensitivity study also shows that only half as many angles are needed, yielding an additional speedup of 2×. In total, the improvements yield roughly a 7–8× speedup. Furthermore given these performance benefits, these approaches have been adopted as the default in MPACT.

A lumped parameter method of characteristics approach and multigroup kernels applied to the subgroup self-shielding calculation in MPACT

Directory of Open Access Journals (Sweden)

Shane Stimpson

2017-09-01

Full Text Available An essential component of the neutron transport solver is the resonance self-shielding calculation used to determine equivalence cross sections. The neutron transport code, MPACT, is currently using the subgroup self-shielding method, in which the method of characteristics (MOC is used to solve purely absorbing fixed-source problems. Recent efforts incorporating multigroup kernels to the MOC solvers in MPACT have reduced runtime by roughly 2×. Applying the same concepts for self-shielding and developing a novel lumped parameter approach to MOC, substantial improvements have also been made to the self-shielding computational efficiency without sacrificing any accuracy. These new multigroup and lumped parameter capabilities have been demonstrated on two test cases: (1 a single lattice with quarter symmetry known as VERA (Virtual Environment for Reactor Applications Progression Problem 2a and (2 a two-dimensional quarter-core slice known as Problem 5a-2D. From these cases, self-shielding computational time was reduced by roughly 3–4×, with a corresponding 15–20% increase in overall memory burden. An azimuthal angle sensitivity study also shows that only half as many angles are needed, yielding an additional speedup of 2×. In total, the improvements yield roughly a 7–8× speedup. Given these performance benefits, these approaches have been adopted as the default in MPACT.

Comparative study on the use of self-shielded packages or returnable shielding for the land disposal of radioactive wastes

International Nuclear Information System (INIS)

Fitzpatrick, J.; Verrall, S.M.

1985-01-01

A comparative study has been carried out on the two philosophies for providing the radiological protection necessary for the transport and handling of packaged intermediate level wastes from their sites of origin to disposal. The two philosophies are self shielding and returnable shielding. The approach taken was to assess the cost and radiological impact differentials of two respective representative waste management procedures. The comparison indicated the merits of each procedure. As a consequence, a hybrid procedure was identified which combines the advantages of each philosophy. This hybrid procedure was used for further comparison. The results of the study indicate that the use of self shielded packages throughout will incur considerable extra expense and give only a small saving in radiological impact. (author)

Computing Moment-Based Probability Tables for Self-Shielding Calculations in Lattice Codes

International Nuclear Information System (INIS)

Hebert, Alain; Coste, Mireille

2002-01-01

As part of the self-shielding model used in the APOLLO2 lattice code, probability tables are required to compute self-shielded cross sections for coarse energy groups (typically with 99 or 172 groups). This paper describes the replacement of the multiband tables (typically with 51 subgroups) with moment-based tables in release 2.5 of APOLLO2. An improved Ribon method is proposed to compute moment-based probability tables, allowing important savings in CPU resources while maintaining the accuracy of the self-shielding algorithm. Finally, a validation is presented where the absorption rates obtained with each of these techniques are compared with exact values obtained using a fine-group elastic slowing-down calculation in the resolved energy domain. Other results, relative to the Rowland's benchmark and to three assembly production cases, are also presented

New improvements in the self-shielding formalism of the Apollo-2 code

International Nuclear Information System (INIS)

Coste, M.; Tellier, H.; Ribon, P.; Raepsaet, V.; Van der Gucht, C.

1993-01-01

One important modelization of a transport code working on a coarse energy mesh is the self-shielding. The French transport code APPOLO 2, developed at the Commissariat a l'Energie Atomique, uses a self-shielding formalism based on a double equivalence. First a homogenization gives the reaction rates in a heterogeneous geometry, and then a multigroup equivalence gives, once the reaction rates are known, the self-shielded cross-sections. The homogenization is a very sensitive part because it is the one which requires physical modelizations. We have added a new model which allows us to treat numerous narrow resonances statistically distributed in the same group of the multigroup mesh. It is important to notice that for a narrow resonance isolated in a group, that new model is equivalent to the previous narrow resonance model (NR)

Radiation monitoring in a self-shielded cyclotron installation

International Nuclear Information System (INIS)

Capaccioli, L.; Gori, C.; Mazzocchi, S.; Spano, G.

2002-01-01

As nuclear medicine is approaching a new era with the spectacular growth of PET diagnosis, the number of medical cyclotrons installed within the major hospitals is increasing accordingly. Therefore modern medical cyclotron are highly engineered and highly reliable apparatus, characterised with reduced accelerating energies (as the major goal is the production of fluorine 18) and often self-shielded. However specific dedicated monitors are still necessary in order to assure the proper radioprotection. At the Careggi University Hospital in Florence a Mini trace 10 MeV self-shielded cyclotron produced by General Electric has been installed in 2000. In a contiguous radiochemistry laboratory, the preparation and quality control of 1 8F DG and other radiopharmaceuticals takes place. Aim of this work is the characterisation and the proper calibration of the above mentioned monitors and control devices

Electron dose distributions caused by the contact-type metallic eye shield: Studies using Monte Carlo and pencil beam algorithms.

Science.gov (United States)

Kang, Sei-Kwon; Yoon, Jai-Woong; Hwang, Taejin; Park, Soah; Cheong, Kwang-Ho; Han, Tae Jin; Kim, Haeyoung; Lee, Me-Yeon; Kim, Kyoung Ju; Bae, Hoonsik

2015-01-01

A metallic contact eye shield has sometimes been used for eyelid treatment, but dose distribution has never been reported for a patient case. This study aimed to show the shield-incorporated CT-based dose distribution using the Pinnacle system and Monte Carlo (MC) calculation for 3 patient cases. For the artifact-free CT scan, an acrylic shield machined as the same size as that of the tungsten shield was used. For the MC calculation, BEAMnrc and DOSXYZnrc were used for the 6-MeV electron beam of the Varian 21EX, in which information for the tungsten, stainless steel, and aluminum material for the eye shield was used. The same plan was generated on the Pinnacle system and both were compared. The use of the acrylic shield produced clear CT images, enabling delineation of the regions of interest, and yielded CT-based dose calculation for the metallic shield. Both the MC and the Pinnacle systems showed a similar dose distribution downstream of the eye shield, reflecting the blocking effect of the metallic eye shield. The major difference between the MC and the Pinnacle results was the target eyelid dose upstream of the shield such that the Pinnacle system underestimated the dose by 19 to 28% and 11 to 18% for the maximum and the mean doses, respectively. The pattern of dose difference between the MC and the Pinnacle systems was similar to that in the previous phantom study. In conclusion, the metallic eye shield was successfully incorporated into the CT-based planning, and the accurate dose calculation requires MC simulation. Copyright © 2015 American Association of Medical Dosimetrists. Published by Elsevier Inc. All rights reserved.

Performance of the improved version of Monte Carlo code A 3MCNP for large-scale shielding problems

International Nuclear Information System (INIS)

Omura, M.; Miyake, Y.; Hasegawa, T.; Ueki, K.; Sato, O.; Haghighat, A.; Sjoden, G. E.

2005-01-01

A 3MCNP (Automatic Adjoint Accelerated MCNP) is a revised version of the MCNP Monte Carlo code, which automatically prepares variance reduction parameters for the CADIS (Consistent Adjoint Driven Importance Sampling) methodology. Using a deterministic 'importance' (or adjoint) function, CADIS performs source and transport biasing within the weight-window technique. The current version of A 3MCNP uses the three-dimensional (3-D) Sn transport TORT code to determine a 3-D importance function distribution. Based on simulation of several real-life problems, it is demonstrated that A 3MCNP provides precise calculation results with a remarkably short computation time by using the proper and objective variance reduction parameters. However, since the first version of A 3MCNP provided only a point source configuration option for large-scale shielding problems, such as spent-fuel transport casks, a large amount of memory may be necessary to store enough points to properly represent the source. Hence, we have developed an improved version of A 3MCNP (referred to as A 3MCNPV) which has a volumetric source configuration option. This paper describes the successful use of A 3MCNPV for a concrete cask neutron and gamma-ray shielding problem, and a PWR dosimetry problem. (authors)

License Application Design Selection Feature Report: Waste Package Self Shielding Design Feature 13

International Nuclear Information System (INIS)

Tang, J.S.

2000-01-01

In the Viability Assessment (VA) reference design, handling of waste packages (WPs) in the emplacement drifts is performed remotely, and human access to the drifts is precluded when WPs are present. This report will investigate the feasibility of using a self-shielded WP design to reduce the radiation levels in the emplacement drifts to a point that, when coupled with ventilation, will create an acceptable environment for human access. This provides the benefit of allowing human entry to emplacement drifts to perform maintenance on ground support and instrumentation, and carry out performance confirmation activities. More direct human control of WP handling and emplacement operations would also be possible. However, these potential benefits must be weighed against the cost of implementation, and potential impacts on pre- and post-closure performance of the repository and WPs. The first section of this report will provide background information on previous investigations of the self-shielded WP design feature, summarize the objective and scope of this document, and provide quality assurance and software information. A shielding performance and cost study that includes several candidate shield materials will then be performed in the subsequent section to allow selection of two self-shielded WP design options for further evaluation. Finally, the remaining sections will evaluate the impacts of the two WP self-shielding options on the repository design, operations, safety, cost, and long-term performance of the WPs with respect to the VA reference design

Neutron shielding properties of a new high-density concrete

International Nuclear Information System (INIS)

Lorente, A.; Gallego, E.; Vega Carrillo, H.R.; Mendez, R.

2008-01-01

The neutron shielding properties of a new high-density concrete (commercially available under the name Hormirad TM , developed in Spain by the company CT-RAD) have been characterized both experimentally and by Monte Carlo calculations. The shielding properties of this concrete against photons were previously studied and the material is being used to build bunkers, mazes and doors in medical accelerator facilities with good overall results. In this work, the objective was to characterize the material behaviour against neutrons, as well as to test alternative mixings including boron compounds in an effort to improve neutron shielding efficiency. With that purpose, Hormirad TM slabs of different thicknesses were exposed to an 241 Am-Be neutron source under controlled conditions in the neutron measurements laboratory of the Nuclear Engineering Department at UPM. The original mix, which includes a high fraction of magnetite, was then modified by adding different proportions of anhydrous borax (Na 2 B 4 O 7 ). In order to have a reference against common concrete used to shield medical accelerator facilities, the same experiment was repeated with ordinary (HA-25) concrete slabs. In parallel to the experiments, Monte Carlo calculations of the experiments were performed with MCNP5. The experimental results agree reasonably well with the Monte Carlo calculations. Therefore, the first and equilibrium tenth-value layers have been determined for the different types of concrete tested. The results show an advantageous behaviour of the Hormirad TM concrete, in terms of neutron attenuation against real thickness of the shielding. Borated concretes seem less practical since they did not show better neutron attenuation with respect to real thickness and their structural properties are worse. The neutron attenuation properties of Hormirad TM for typical neutron spectra in clinical LINAC accelerators rooms have been also characterized by Monte Carlo calculation. (author)

Monte Carlo analysis of helium production in the ITER shielding blanket module

International Nuclear Information System (INIS)

Sato, S.

1999-01-01

In order to examine the shielding performances of the inboard blanket module in the international thermonuclear experimental reactor (ITER), shielding calculations have been carried out using a three-dimensional Monte Carlo method. The impact of radiation streaming through the front access holes and gaps between adjacent blanket modules on the helium gas production in the branch pipe weld locations and back plate have been estimated. The three-dimensional model represents an 18 sector of the overall torus region and includes the vacuum vessel, inboard blanket and back plate, plasma region, and outboard reflecting medium. And it includes the 1 m high inboard mid-plane module and the 20 mm wide gaps between adjacent modules. From the calculated results for the reference design, it has been found that the helium production at the plug of the branch pipe is four to five times higher than the design goal of 1 appm for a neutron fluence of 0.9 MW a m -2 at the inboard mid-plane first wall. Also, it has been found that the helium production at the back plate behind the horizontal gap is about three times higher than the design goal. In the reference design, the stainless steel (SS):H 2 O composition in the blanket module is 80:20%. Shielding calculations also have been carried out for the SS:H 2 O composition of 70:30, 60:40, 50:50 and 40:60%. From the evaluated results for their design, it has been found that the dependence of helium production on the SS:H 2 170 mm will reduce helium production to satisfy the design goal and not have a significant impact on weight limitations imposed by remote maintenance handling limitations. Also based on the calculated results, about 200 mm thick shields such as a key structure in the vertical gap are suggested to be installed in the horizontal gap as well to reduce the helium production at the back plate and to satisfy the design goal. (orig.)

Self-similar regimes of fast ionization waves in shielded discharge tubes

International Nuclear Information System (INIS)

Gerasimov, D.N.; Sinkevich, O.A.

1999-01-01

An analytical self-similar solution to the problem of the propagation of a fast ionization wave (FIW) in a long shielded tube is constructed. An expression determining the influence of the device parameters on the FIW velocity is obtained; the velocity is found to be the nonmonotonic function of the working-gas pressure. The theoretical predictions are compared with the results of experiments carried out with helium and nitrogen. The calculation and experimental results agree within experimental errors

Toolkit for high performance Monte Carlo radiation transport and activation calculations for shielding applications in ITER

International Nuclear Information System (INIS)

Serikov, A.; Fischer, U.; Grosse, D.; Leichtle, D.; Majerle, M.

2011-01-01

The Monte Carlo (MC) method is the most suitable computational technique of radiation transport for shielding applications in fusion neutronics. This paper is intended for sharing the results of long term experience of the fusion neutronics group at Karlsruhe Institute of Technology (KIT) in radiation shielding calculations with the MCNP5 code for the ITER fusion reactor with emphasizing on the use of several ITER project-driven computer programs developed at KIT. Two of them, McCad and R2S, seem to be the most useful in radiation shielding analyses. The McCad computer graphical tool allows to perform automatic conversion of the MCNP models from the underlying CAD (CATIA) data files, while the R2S activation interface couples the MCNP radiation transport with the FISPACT activation allowing to estimate nuclear responses such as dose rate and nuclear heating after the ITER reactor shutdown. The cell-based R2S scheme was applied in shutdown photon dose analysis for the designing of the In-Vessel Viewing System (IVVS) and the Glow Discharge Cleaning (GDC) unit in ITER. Newly developed at KIT mesh-based R2S feature was successfully tested on the shutdown dose rate calculations for the upper port in the Neutral Beam (NB) cell of ITER. The merits of McCad graphical program were broadly acknowledged by the neutronic analysts and its continuous improvement at KIT has introduced its stable and more convenient run with its Graphical User Interface. Detailed 3D ITER neutronic modeling with the MCNP Monte Carlo method requires a lot of computation resources, inevitably leading to parallel calculations on clusters. Performance assessments of the MCNP5 parallel runs on the JUROPA/HPC-FF supercomputer cluster permitted to find the optimal number of processors for ITER-type runs. (author)

Modelling of neutron and photon transport in iron and concrete radiation shieldings by the Monte Carlo method - Version 2

CERN Document Server

Žukauskaite, A; Plukiene, R; Plukis, A

2007-01-01

Particle accelerators and other high energy facilities produce penetrating ionizing radiation (neutrons and γ-rays) that must be shielded. The objective of this work was to model photon and neutron transport in various materials, usually used as shielding, such as concrete, iron or graphite. Monte Carlo method allows obtaining answers by simulating individual particles and recording some aspects of their average behavior. In this work several nuclear experiments were modeled: AVF 65 – γ-ray beams (1-10 MeV), HIMAC and ISIS-800 – high energy neutrons (20-800 MeV) transport in iron and concrete. The results were then compared with experimental data.

Development of neutron shielding concrete containing iron content materials

Science.gov (United States)

Sariyer, Demet; Küçer, Rahmi

2018-02-01

Concrete is one of the most important construction materials which widely used as a neutron shielding. Neutron shield is obtained of interaction with matter depends on neutron energy and the density of the shielding material. Shielding properties of concrete could be improved by changing its composition and density. High density materials such as iron or high atomic number elements are added to concrete to increase the radiation resistance property. In this study, shielding properties of concrete were investigated by adding iron, FeB, Fe2B, stainless - steel at different ratios into concrete. Neutron dose distributions and shield design was obtained by using FLUKA Monte Carlo code. The determined shield thicknesses vary depending on the densities of the mixture formed by the additional material and ratio. It is seen that a combination of iron rich materials is enhanced the neutron shielding of capabilities of concrete. Also, the thicknesses of shield are reduced.

CASIM, High Energy Cascades in Shields of Arbitrary Geometry Using Monte-Carlo Method

International Nuclear Information System (INIS)

Van Ginneken, A.

1987-01-01

1 - Description of problem or function: CASIM is a Monte Carlo program to study the average development of high energy cascades in large targets (shields) of arbitrary geometry and composition. The program is best suited for incident energies in the range 20-1000 GeV. 2 - Method of solution: The simulation makes extensive use of weighting techniques to avoid difficulties encountered in sampling complicated distributions and to allow the user to introduce bias in the sampling. The program uses the particle production model (in the form of a set of inclusive distributions) to compute (a) star densities (i.e. nuclear interaction densities) as a function of location and particle type throughout the target. From these star densities, estimates of a number of quantities of radiobiological interest can be obtained; (b) momentum spectra of particles interacting in the shield also as a function of location and type; (c) energy deposited by the cascade. This quantity is a useful measure of target heating embedded in the target (ionization calorimeter). 3 - Restrictions on the complexity of the problem: The program does not study transport of low momentum particles (less than or equal to 0.3 GeV/c)

Self Shielding in Nuclear Fissile Assay Using LSDS

International Nuclear Information System (INIS)

Lee, Yong Deok; Park, Chang Je; Park, Geun Il; Song, Kee Chan

2012-01-01

The new technology for isotopic fissile material contents assay is under development at KAERI using lead slowing down spectrometer(LSDS). LSDS is very sensitive to distinguish fission signals from each fissile isotope in spent and recycled fuel. The accumulation of spent fuel is current big issue. The amount of spent fuels will reach the maximum storage capacity of the pools soon. Therefore, an interim storage must be searched and it should be optimized in design by applying accurate fissile content. When the storage has taken effect, all the nuclear materials must be also specified and verified for safety, economics and management. Generally, the spent fuel from PWR has unburned ∼1 % U235, produced ∼0.5 % plutonium from decay chain, ∼3 % fission products, ∼ 0.1 % minor actinides (MA) and uranium remainder. About 1.5 % fissile materials still exist in the spent fuel. Therefore, for reutilization of fissile materials in spent fuel at SFR, resource material is produced through pyro process. Fissile material contents in resource material must be analyzed before fabricating SFR fuel for reactor safety and economics. In assay of fissile content of spent fuel and recycled fuel, intense radiation background gives limitation on the direct analysis of fissile materials. However, LSDS is not influenced by such a radiation background in fissile assay. Based on the decided geometry setup, self shielding parameter was calculated at the fuel assay zone by introducing spent fuel or pyro produced nuclear material. When nuclear material is inserted into the assay area, the spent fuel assembly or pyro recycled fuel material perturbs the spatial distribution of the slowing down neutrons in lead and the prompt fast fission neutrons produced by fissile materials are also perturbed. The self shielding factor is interpreted as that how much of absorption is created inside the fuel area when it is in the lead. Self shielding effect provides a non-linear property in the isotopic

Improvement of Monte Carlo code A3MCNP for large-scale shielding problems

International Nuclear Information System (INIS)

Miyake, Y.; Ohmura, M.; Hasegawa, T.; Ueki, K.; Sato, O.; Haghighat, A.; Sjoden, G.E.

2004-01-01

A 3 MCNP (Automatic Adjoint Accelerated MCNP) is a revised version of the MCNP Monte Carlo code, that automatically prepares variance reduction parameters for the CADIS (Consistent Adjoint Driven Importance Sampling) methodology. Using a deterministic 'importance' (or adjoint) function, CADIS performs source and transport biasing within the weight-window technique. The current version of A 3 MCNP uses the 3-D Sn transport TORT code to determine a 3-D importance function distribution. Based on simulation of several real-life problems, it is demonstrated that A 3 MCNP provides precise calculation results with a remarkably short computation time by using the proper and objective variance reduction parameters. However, since the first version of A 3 MCNP provided only a point source configuration option for large-scale shielding problems, such as spent-fuel transport casks, a large amount of memory may be necessary to store enough points to properly represent the source. Hence, we have developed an improved version of A 3 MCNP (referred to as A 3 MCNPV) which has a volumetric source configuration option. This paper describes the successful use of A 3 MCNPV for a concrete cask streaming problem and a PWR dosimetry problem. (author)

Monte carlo calculation of the neutron effective dose rate at the outer surface of the biological shield of HTR-10 reactor

International Nuclear Information System (INIS)

Remetti, Romolo; Andreoli, Giulio; Keshishian, Silvina

2012-01-01

Highlights: ► We deal with HTR-10, that is a helium-cooled graphite-moderated pebble bed reactor. ► We carried out Monte Carlo simulation of the core by MCNP5. ► Extensive use of MCNP5 variance reduction methods has been done. ► We calculated the trend of neutron flux within the biological shield. ► We calculated neutron effective dose at the outer surface of biological shield. - Abstract: Research on experimental reactors, such as HTR-10, provide useful data about potentialities of very high temperature gas-cooled reactors (VHTR). The latter is today rated as one of the six nuclear reactor types involved in the Generation-IV International Forum (GIF) Initiative. In this study, the MCNP5 code has been employed to evaluate the neutron radiation trend vs. the biological shield's thickness and to calculate the neutron effective dose rate at the outer surface. The reactor's geometry has been completely modeled by means of lattices and universes provided by MCNP, even though some approximations were required. Monte Carlo calculations have been performed by means of a simple PC and, as a consequence, in order to obtain acceptable run times, it was made an extensive recourse to variance reduction methods.

Evaluation of the shield calculation adequacy of radiotherapy rooms through Monte Carlo Method and experimental measures

International Nuclear Information System (INIS)

Meireles, Ramiro Conceicao

2016-01-01

The shielding calculation methodology for radiotherapy services adopted in Brazil and in several countries is that described in publication 151 of the National Council on Radiation Protection and Measurements (NCRP 151). This methodology however, markedly employs several approaches that can impact both in the construction cost and in the radiological safety of the facility. Although this methodology is currently well established by the high level of use, some parameters employed in the calculation methodology did not undergo to a detailed assessment to evaluate the impact of the various approaches considered. In this work the MCNP5 Monte Carlo code was used with the purpose of evaluating the above mentioned approaches. TVLs values were obtained for photons in conventional concrete (2.35g / cm 3 ), considering the energies of 6, 10 and 25 MeV, respectively, first considering an isotropic radiation source impinging perpendicular to the barriers, and subsequently a lead head shielding emitting a shaped beam, in the format of a pyramid trunk. Primary barriers safety margins, taking in account the head shielding emitting photon beam pyramid-shaped in the energies of 6, 10, 15 and 18 MeV were assessed. A study was conducted considering the attenuation provided by the patient's body in the energies of 6,10, 15 and 18 MeV, leading to new attenuation factors. Experimental measurements were performed in a real radiotherapy room, in order to map the leakage radiation emitted by the accelerator head shielding and the results obtained were employed in the Monte Carlo simulation, as well as to validate the entire study. The study results indicate that the TVLs values provided by (NCRP, 2005) show discrepancies in comparison with the values obtained by simulation and that there may be some barriers that are calculated with insufficient thickness. Furthermore, the simulation results show that the additional safety margins considered when calculating the width of the primary

Shielding calculations in support of the Spallation Neutron Source (SNS) proton beam transport system

International Nuclear Information System (INIS)

Johnson, Jeffrey O.; Gallmeier, Franz X.; Popova, Irina

2002-01-01

Determining the bulk shielding requirements for accelerator environments is generally an easy task compared to analyzing the radiation transport through the complex shield configurations and penetrations typically associated with the detailed Title II design efforts of a facility. Shielding calculations for penetrations in the SNS accelerator environment are presented based on hybrid Monte Carlo and discrete ordinates particle transport methods. This methodology relies on coupling tools that map boundary surface leakage information from the Monte Carlo calculations to boundary sources for one-, two-, and three-dimensional discrete ordinates calculations. The paper will briefly introduce the coupling tools for coupling MCNPX to the one-, two-, and three-dimensional discrete ordinates codes in the DOORS code suite. The paper will briefly present typical applications of these tools in the design of complex shield configurations and penetrations in the SNS proton beam transport system

Shielding benchmark problems, (2)

International Nuclear Information System (INIS)

Tanaka, Shun-ichi; Sasamoto, Nobuo; Oka, Yoshiaki; Shin, Kazuo; Tada, Keiko.

1980-02-01

Shielding benchmark problems prepared by Working Group of Assessment of Shielding Experiments in the Research Committee on Shielding Design in the Atomic Energy Society of Japan were compiled by Shielding Laboratory in Japan Atomic Energy Research Institute. Fourteen shielding benchmark problems are presented newly in addition to twenty-one problems proposed already, for evaluating the calculational algorithm and accuracy of computer codes based on discrete ordinates method and Monte Carlo method and for evaluating the nuclear data used in codes. The present benchmark problems are principally for investigating the backscattering and the streaming of neutrons and gamma rays in two- and three-dimensional configurations. (author)

Monte Carlo simulation of x-ray buildup factors of lead and its applications in shielding of diagnostic x-ray facilities

International Nuclear Information System (INIS)

Kharrati, Hedi; Agrebi, Amel; Karaoui, Mohamed-Karim

2007-01-01

X-ray buildup factors of lead in broad beam geometry for energies from 15 to 150 keV are determined using the general purpose Monte Carlo N-particle radiation transport computer code (MCNP4C). The obtained buildup factors data are fitted to a modified three parameter Archer et al. model for ease in calculating the broad beam transmission with computer at any tube potentials/filters combinations in diagnostic energies range. An example for their use to compute the broad beam transmission at 70, 100, 120, and 140 kVp is given. The calculated broad beam transmission is compared to data derived from literature, presenting good agreement. Therefore, the combination of the buildup factors data as determined and a mathematical model to generate x-ray spectra provide a computationally based solution to broad beam transmission for lead barriers in shielding x-ray facilities

Shielding Design and Radiation Shielding Evaluation for LSDS System Facility

International Nuclear Information System (INIS)

Kim, Younggook; Kim, Jeongdong; Lee, Yongdeok

2015-01-01

As the system characteristics, the target in the spectrometer emits approximately 1012 neutrons/s. To efficiently shield the neutron, the shielding door designs are proposed for the LSDS system through a comparison of the direct shield and maze designs. Hence, to guarantee the radiation safety for the facility, the door design is a compulsory course of the development of the LSDS system. To improve the shielding rates, 250x250 covering structure was added as a subsidiary around the spectrometer. In this study, the evaluations of the suggested shielding designs were conducted using MCNP code. The suggested door design and covering structures can shield the neutron efficiently, thus all evaluations of all conditions are satisfied within the public dose limits. From the Monte Carlo code simulation, Resin(Indoor type) and Tungsten(Outdoor type) were selected as the shielding door materials. From a comparative evaluation of the door thickness, In and Out door thickness was selected 50 cm

Investigating spatial self-shielding and temperature effects for homogeneous and double heterogeneous pebble models with MCNP

International Nuclear Information System (INIS)

Li, J.; Nuenighoff; Pohl, C.; Allelein, H.J.

2010-01-01

The gas-cooled, high temperature reactor (HTR) represents a valuable option for the future development of nuclear technology, because of its excellent safety features. One main safety feature is the negative temperature coefficient which is due to the Doppler broadening of the (n,y) resonance absorption cross section. A second important effect is the spatial self-shielding due to the double heterogeneous geometry of a pebble bed reactor. At FZ-Juelich two reactor analysis codes have been developed: VSOP for core design and MGT for transient analysis. Currently an update of the nuclear cross section libraries to ENDF/B-VII.0 of both codes takes place. In order to take the temperature dependency as well as the spatial self-shielding into account the absorption cross sections σ (n,y) for the resonance absorbers like 232 Th and 238 U have to be provided as function of incident neutron energy, temperature and nuclide concentration. There are two reasons for choosing the Monte-Carlo approach to calculate group wise cross sections. First, the former applied ZUT-DGL code to generate the resonance cross section tables for MGT is so far not able to handle the new resonance description based on Reich-Moore instead of Single-level Breit-Wigner. Second, the rising interest in PuO 2 fuel motivated an investigation on the generation of group wise cross sections describing thermal resonances of 240 Pu and 242 Pu. (orig.)

FLUKA shielding calculations for the FAIR project

International Nuclear Information System (INIS)

Fehrenbacher, Georg; Kozlova, Ekaterina; Radon, Torsten; Sokolov, Alexey

2015-01-01

FAIR is an international accelerator project being in construction at GSI Helmholtz center for heavy ion research in Darmstadt. The Monte Carlo program FLUKA is used to study radiation protection problems. The contribution deals with general application possibilities of FLUKA and for FAIR with respect the radiation protection planning. The necessity to simulate the radiation transport through shielding of several meters thickness and to determine the equivalent doses outside the shielding with sufficient accuracy is demonstrated using two examples under consideration of the variance reduction. Results of simulation calculations for activation estimation in accelerator facilities are presented.

Ford motor company NDE facility shielding design

International Nuclear Information System (INIS)

Metzger, R. L.; Van Riper, K. A.; Jones, M. H.

2005-01-01

Ford Motor Company proposed the construction of a large non-destructive evaluation laboratory for radiography of automotive power train components. The authors were commissioned to design the shielding and to survey the completed facility for compliance with radiation doses for occupationally and non-occupationally exposed personnel. The two X-ray sources are Varian Linatron 3000 accelerators operating at 9-11 MV. One performs computed tomography of automotive transmissions, while the other does real-time radiography of operating engines and transmissions. The shield thickness for the primary barrier and all secondary barriers were determined by point-kernel techniques. Point-kernel techniques did not work well for skyshine calculations and locations where multiple sources (e.g. tube head leakage and various scatter fields) impacted doses. Shielding for these areas was determined using transport calculations. A number of MCNP [Briesmeister, J. F. MCNPCA general Monte Carlo N-particle transport code version 4B. Los Alamos National Laboratory Manual (1997)] calculations focused on skyshine estimates and the office areas. Measurements on the operational facility confirmed the shielding calculations. (authors)

Ford Motor Company NDE facility shielding design.

Science.gov (United States)

Metzger, Robert L; Van Riper, Kenneth A; Jones, Martin H

2005-01-01

Ford Motor Company proposed the construction of a large non-destructive evaluation laboratory for radiography of automotive power train components. The authors were commissioned to design the shielding and to survey the completed facility for compliance with radiation doses for occupationally and non-occupationally exposed personnel. The two X-ray sources are Varian Linatron 3000 accelerators operating at 9-11 MV. One performs computed tomography of automotive transmissions, while the other does real-time radiography of operating engines and transmissions. The shield thickness for the primary barrier and all secondary barriers were determined by point-kernel techniques. Point-kernel techniques did not work well for skyshine calculations and locations where multiple sources (e.g. tube head leakage and various scatter fields) impacted doses. Shielding for these areas was determined using transport calculations. A number of MCNP [Briesmeister, J. F. MCNPCA general Monte Carlo N-particle transport code version 4B. Los Alamos National Laboratory Manual (1997)] calculations focused on skyshine estimates and the office areas. Measurements on the operational facility confirmed the shielding calculations.

Tax Shield, Insolvenz und Zinsschranke

OpenAIRE

Arnold, Sven; Lahmann, Alexander; Schwetzler, Bernhard

2010-01-01

Dieser Beitrag analysiert den Wertbeitrag fremdfinanzierungsbedingter Steuervorteile (Tax Shield) unter realistischen Bedingungen (keine Negativsteuer; mögliche Insolvenz) für unterschiedliche Finanzierungspolitiken. Zusätzlich wird der Effekt der sogenannten Zinsschranke auf den Wert des Tax Shield ermittelt. Die Bewertung des Tax Shield mit und ohne Zinsschranke findet im einperiodigen Fall auf der Basis von Optionspreismodellen und im mehrperiodigen Fall auf der Basis von Monte Carlo Simul...

Evaluation using Monte Carlo simulations, of the effect of a shielding, called external shielding, for fotoneutrons generated in linear accelerators, using the computational model of Varian accelerator 2300 C/D operating in eight rotation angles of the GA

International Nuclear Information System (INIS)

Silva, Hugo R.; Silva, Ademir X.; Rebello, Wilson F.; Silva, Maria G.

2011-01-01

This paper aims to present the results obtained by Monte Carlo simulation of the effect of shielding against neutrons, called External Shielding, to be placed on the heads of linear accelerators used in radiotherapy. For this, it was used the radiation transport code Monte Carlo N-Particle - MCNPX, in which were developed computational model of the head of the linear accelerator Varian 2300 C/D. The equipment was simulated within a bunker, operating at energies of 10, 15 and 18 MV, considering the rotation of the gantry at eight different angles ( 0 deg, 45 deg, 90 deg, 135 deg, 180 deg, 225 deg, 270 deg and 315 deg), in all cases, the equipment was modeled without and with the shielding positioned attached to the head of the accelerator on its bottom. In each of these settings, it was calculated the Ambient Dose Equivalent due to neutron H * (10)n on points situated in the region of the patient (region of interest for evaluation of undesirable neutron doses on the patient) and in the maze of radiotherapy room (region of interest for shielding the access door to the bunker). It was observed for all energies of equipment operation as well as for all angles of inclination of the gantry, a significant reduction in the values of H * (10) n when the equipment operated with the external shielding, both in the region of the patient as in the region of the maze. (author)

A three-dimensional computed tomography-assisted Monte Carlo evaluation of ovoid shielding on the dose to the bladder and rectum in intracavitary radiotherapy for cervical cancer

International Nuclear Information System (INIS)

Gifford, Kent A.; Horton, John L.; Pelloski, Christopher E.; Jhingran, Anuja; Court, Laurence E.; Mourtada, Firas; Eifel, Patricia J.

2005-01-01

Purpose: To determine the effects of Fletcher Suit Delclos ovoid shielding on dose to the bladder and rectum during intracavitary radiotherapy for cervical cancer. Methods and Materials: The Monte Carlo method was used to calculate the dose in 12 patients receiving low-dose-rate intracavitary radiotherapy with both shielded and unshielded ovoids. Cumulative dose-difference surface histograms were computed for the bladder and rectum. Doses to the 2-cm 3 and 5-cm 3 volumes of highest dose were computed for the bladder and rectum with and without shielding. Results: Shielding affected dose to the 2-cm 3 and 5-cm 3 volumes of highest dose for the rectum (10.1% and 11.1% differences, respectively). Shielding did not have a major impact on the dose to the 2-cm 3 and 5-cm 3 volumes of highest dose for the bladder. The average dose reduction to 5% of the surface area of the bladder was 53 cGy. Reductions as large as 150 cGy were observed to 5% of the surface area of the bladder. The average dose reduction to 5% of the surface area of the rectum was 195 cGy. Reductions as large as 405 cGy were observed to 5% of the surface area of the rectum. Conclusions: Our data suggest that the ovoid shields can greatly reduce the radiation dose delivered to the rectum. We did not find the same degree of effect on the dose to the bladder. To calculate the dose accurately, however, the ovoid shields must be included in the dose model

Application of variance reduction techniques of Monte-Carlo method to deep penetration shielding problems

International Nuclear Information System (INIS)

Rawat, K.K.; Subbaiah, K.V.

1996-01-01

General purpose Monte Carlo code MCNP is being widely employed for solving deep penetration problems by applying variance reduction techniques. These techniques depend on the nature and type of the problem being solved. Application of geometry splitting and implicit capture method are examined to study the deep penetration problems of neutron, gamma and coupled neutron-gamma in thick shielding materials. The typical problems chosen are: i) point isotropic monoenergetic gamma ray source of 1 MeV energy in nearly infinite water medium, ii) 252 Cf spontaneous source at the centre of 140 cm thick water and concrete and iii) 14 MeV fast neutrons incident on the axis of 100 cm thick concrete disk. (author). 7 refs., 5 figs

Self-learning Monte Carlo (dynamical biasing)

International Nuclear Information System (INIS)

Matthes, W.

1981-01-01

In many applications the histories of a normal Monte Carlo game rarely reach the target region. An approximate knowledge of the importance (with respect to the target) may be used to guide the particles more frequently into the target region. A Monte Carlo method is presented in which each history contributes to update the importance field such that eventually most target histories are sampled. It is a self-learning method in the sense that the procedure itself: (a) learns which histories are important (reach the target) and increases their probability; (b) reduces the probabilities of unimportant histories; (c) concentrates gradually on the more important target histories. (U.K.)

A study on the apron shielding ratio according to electromagnetic radiation energy

Energy Technology Data Exchange (ETDEWEB)

Jang, Dong Gun; Lee, Sang Ho; Choi, Hyung Seok; Son, Joo Chul; Yoon, Chang Yong; Ji, Yung Sik; Cho, Yong In; Lee, Hong Je; Yang, Seoung Oh [Dept. of Nuclear Medicine, Dongnam Institute of Radiological and Medical Sciences Cancer Center, Busan (Korea, Republic of)

2014-12-15

The medical institution has been used electromagnetic radiation of various energy. But researchers are divided on whether using apron for radiation shielding will be effective or not. The purpose of present study was to analyze electromagnetic radiation shielding effect of apron by using Monte Carlo simulation. 1 MBq electromagnetic radiation was emitted from 10-500 keV at 10 keV increments in Monte Carlo simulation. Then shielded radiation dose difference was confirmed, when 0.25 mmPb shield use for shielding. As a results, shielding ratio was markedly decreased in high energy electromagnetic radiation. The radiation dose was inversely increased with 0.25 mmPb shielding.

A study on the apron shielding ratio according to electromagnetic radiation energy

International Nuclear Information System (INIS)

Jang, Dong Gun; Lee, Sang Ho; Choi, Hyung Seok; Son, Joo Chul; Yoon, Chang Yong; Ji, Yung Sik; Cho, Yong In; Lee, Hong Je; Yang, Seoung Oh

2014-01-01

The medical institution has been used electromagnetic radiation of various energy. But researchers are divided on whether using apron for radiation shielding will be effective or not. The purpose of present study was to analyze electromagnetic radiation shielding effect of apron by using Monte Carlo simulation. 1 MBq electromagnetic radiation was emitted from 10-500 keV at 10 keV increments in Monte Carlo simulation. Then shielded radiation dose difference was confirmed, when 0.25 mmPb shield use for shielding. As a results, shielding ratio was markedly decreased in high energy electromagnetic radiation. The radiation dose was inversely increased with 0.25 mmPb shielding

Self-shielding effect in unresolved resonance data in JENDL-4.0

International Nuclear Information System (INIS)

Konno, Chikara; Takakura, Kosuke; Ochiai, Kentaro; Sato, Satoshi; Kato, Yoshinari

2012-01-01

At International Conference on Nuclear Data for Science and Technology in 2007 we pointed out that most of unresolved resonance data in JENDL-3.3 have a problem related to self-shielding correction. Here with a simple calculation model we have investigated whether the latest JENDL, JENDL-4.0, was improved for the problem or not. The results suggest that unresolved resonance data in JENDL-4.0 have no problem, but it seems that self-shielding effects for the unresolved resonance data in JENDL-4.0 are too large. New benchmark experiments for unresolved resonance data are strongly recommended in order to verify unresolved resonance data. (author)

Methods for calculating radiation attenuation in shields

Energy Technology Data Exchange (ETDEWEB)

Butler, J; Bueneman, D; Etemad, A; Lafore, P; Moncassoli, A M; Penkuhn, H; Shindo, M; Stoces, B

1964-10-01

In recent years the development of high-speed digital computers of large capacity has revolutionized the field of reactor shield design. For compact special-purpose reactor shields, Monte-Carlo codes in two- and three dimensional geometries are now available for the proper treatment of both the neutron and gamma- ray problems. Furthermore, techniques are being developed for the theoretical optimization of minimum-weight shield configurations for this type of reactor system. In the design of land-based power reactors, on the other hand, there is a strong incentive to reduce the capital cost of the plant, and economic considerations are also relevant to reactors designed for merchant ship propulsion. In this context simple methods are needed which are economic in their data input and computing time requirements and which, at the same time, are sufficiently accurate for design work. In general the computing time required for Monte-Carlo calculations in complex geometry is excessive for routine design calculations and the capacity of the present codes is inadequate for the proper treatment of large reactor shield systems in three dimensions. In these circumstances a wide range of simpler techniques are currently being employed for design calculations. The methods of calculation for neutrons in reactor shields fall naturally into four categories: Multigroup diffusion theory; Multigroup diffusion with removal sources; Transport codes; and Monte Carlo methods. In spite of the numerous Monte- Carlo techniques which are available for penetration and back scattering, serious problems are still encountered in practice with the scattering of gamma rays from walls of buildings which contain critical facilities and also concrete-lined discharge shafts containing irradiated fuel elements. The considerable volume of data in the unclassified literature on the solution of problems of this type in civil defence work appears not to have been evaluated for reactor shield design. In

Modeling of neutron and photon transport in iron and concrete radiation shields by using Monte Carlo method

CERN Document Server

Žukauskaitėa, A; Plukienė, R; Ridikas, D

2007-01-01

Particle accelerators and other high energy facilities produce penetrating ionizing radiation (neutrons and γ-rays) that must be shielded. The objective of this work was to model photon and neutron transport in various materials, usually used as shielding, such as concrete, iron or graphite. Monte Carlo method allows obtaining answers by simulating individual particles and recording some aspects of their average behavior. In this work several nuclear experiments were modeled: AVF 65 (AVF cyclotron of Research Center of Nuclear Physics, Osaka University, Japan) – γ-ray beams (1-10 MeV), HIMAC (heavy-ion synchrotron of the National Institute of Radiological Sciences in Chiba, Japan) and ISIS-800 (ISIS intensive spallation neutron source facility of the Rutherford Appleton laboratory, UK) – high energy neutron (20-800 MeV) transport in iron and concrete. The calculation results were then compared with experimental data.compared with experimental data.

Insufficient self-shielding correction in VITAMIN-B6

International Nuclear Information System (INIS)

Konno, Chikara; Ochiai, Kentaro; Ohnishi, Seiki

2011-01-01

We carried out a simple benchmark calculation test with a multigroup cross-section library VITAMIN-B6 generated from ENDF/B-VI. The model of this test consisted of an iron sphere of 1 m in radius with an isotropic 20 MeV neutron source in the center. Neutron spectra in the sphere were calculated with an Sn code ANISN and VITAMIN-B6 or FENDL/MG-1.1. A calculation with MCNP and ENDF/B-VI was carried out as a reference. The neutron spectra with ANISN and FENDL/MG-1.1 agreed with those with MCNP, while those with ANISN and VITAMIN-B6 were at most 50% different from those with MCNP. We uncovered that the discrepancy came from insufficient self-shielding correction due to the followings; 1) The smallest background cross section of 56 Fe in VITAMIN-B6 is 1. 2) The weighting flux used in generating VITAMIN-B6 is not adequate. VITAMIN-B6 should be revised for adequate self-shielding correction. (author)

A Comparison of Monte Carlo and Deterministic Solvers for keff and Sensitivity Calculations

Energy Technology Data Exchange (ETDEWEB)

Haeck, Wim [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Parsons, Donald Kent [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); White, Morgan Curtis [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Saller, Thomas [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Favorite, Jeffrey A. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

2017-12-12

Verification and validation of our solutions for calculating the neutron reactivity for nuclear materials is a key issue to address for many applications, including criticality safety, research reactors, power reactors, and nuclear security. Neutronics codes solve variations of the Boltzmann transport equation. The two main variants are Monte Carlo versus deterministic solutions, e.g. the MCNP [1] versus PARTISN [2] codes, respectively. There have been many studies over the decades that examined the accuracy of such solvers and the general conclusion is that when the problems are well-posed, either solver can produce accurate results. However, the devil is always in the details. The current study examines the issue of self-shielding and the stress it puts on deterministic solvers. Most Monte Carlo neutronics codes use continuous-energy descriptions of the neutron interaction data that are not subject to this effect. The issue of self-shielding occurs because of the discretisation of data used by the deterministic solutions. Multigroup data used in these solvers are the average cross section and scattering parameters over an energy range. Resonances in cross sections can occur that change the likelihood of interaction by one to three orders of magnitude over a small energy range. Self-shielding is the numerical effect that the average cross section in groups with strong resonances can be strongly affected as neutrons within that material are preferentially absorbed or scattered out of the resonance energies. This affects both the average cross section and the scattering matrix.

Automated importance generation and biasing techniques for Monte Carlo shielding techniques by the TRIPOLI-3 code

International Nuclear Information System (INIS)

Both, J.P.; Nimal, J.C.; Vergnaud, T.

1990-01-01

We discuss an automated biasing procedure for generating the parameters necessary to achieve efficient Monte Carlo biasing shielding calculations. The biasing techniques considered here are exponential transform and collision biasing deriving from the concept of the biased game based on the importance function. We use a simple model of the importance function with exponential attenuation as the distance to the detector increases. This importance function is generated on a three-dimensional mesh including geometry and with graph theory algorithms. This scheme is currently being implemented in the third version of the neutron and gamma ray transport code TRIPOLI-3. (author)

Adjoint electron Monte Carlo calculations

International Nuclear Information System (INIS)

Jordan, T.M.

1986-01-01

Adjoint Monte Carlo is the most efficient method for accurate analysis of space systems exposed to natural and artificially enhanced electron environments. Recent adjoint calculations for isotropic electron environments include: comparative data for experimental measurements on electronics boxes; benchmark problem solutions for comparing total dose prediction methodologies; preliminary assessment of sectoring methods used during space system design; and total dose predictions on an electronics package. Adjoint Monte Carlo, forward Monte Carlo, and experiment are in excellent agreement for electron sources that simulate space environments. For electron space environments, adjoint Monte Carlo is clearly superior to forward Monte Carlo, requiring one to two orders of magnitude less computer time for relatively simple geometries. The solid-angle sectoring approximations used for routine design calculations can err by more than a factor of 2 on dose in simple shield geometries. For critical space systems exposed to severe electron environments, these potential sectoring errors demand the establishment of large design margins and/or verification of shield design by adjoint Monte Carlo/experiment

Importance of self-shielding for improving sensitivity coefficients in light water nuclear reactors

International Nuclear Information System (INIS)

Foad, Basma; Takeda, Toshikazu

2014-01-01

Highlights: • A new method has been developed for calculating sensitivity coefficients. • This method is based on the use of infinite dilution cross-sections instead of effective cross-sections. • The change of self-shielding factor due to cross-section perturbation has been considered. • SRAC and SAINT codes are used for calculating improved sensitivities, while MCNP code has been used for verification. - Abstract: In order to perform sensitivity analyzes in light water reactors where self-shielding effect becomes important, a new method has been developed for calculating sensitivity coefficient of core characteristics relative to the infinite dilution cross-sections instead of the effective cross-sections. This method considers the change of the self-shielding factor due to cross-section perturbation for different nuclides and reactions. SRAC and SAINT codes are used to calculate the improved sensitivity; while the accuracy of the present method has been verified by MCNP code and good agreement has been found

Development and testing of multigroup library with correction of self-shielding effects in fusion-fission hybrid reactor

International Nuclear Information System (INIS)

Zou Jun; He Zhaozhong; Zeng Qin; Qiu Yuefeng; Wang Minghuang

2010-01-01

A multigroup library HENDL2.1/SS (Hybrid Evaluated Nuclear Data Library/Self-Shielding) based on ENDF/B-VII.0 evaluate data has been generated using Bondarenko and flux calculator method for the correction of self-shielding effect of neutronics analyses. To validate the reliability of the multigroup library HENDL2.1/SS, transport calculations for fusion-fission hybrid system FDS-I were performed in this paper. It was verified that the calculations with the HENDL2.1/SS gave almost the same results with MCNP calculations and were better than calculations with the HENDL2.0/MG which is another multigroup library without self-shielding correction. The test results also showed that neglecting resonance self-shielding caused underestimation of the K eff , neutron fluxes and waste transmutation ratios in the multigroup calculations of FDS-I.

Advanced resonance self-shielding method for gray resonance treatment in lattice physics code GALAXY

International Nuclear Information System (INIS)

Koike, Hiroki; Yamaji, Kazuya; Kirimura, Kazuki; Sato, Daisuke; Matsumoto, Hideki; Yamamoto, Akio

2012-01-01

A new resonance self-shielding method based on the equivalence theory is developed for general application to the lattice physics calculations. The present scope includes commercial light water reactor (LWR) design applications which require both calculation accuracy and calculation speed. In order to develop the new method, all the calculation processes from cross-section library preparation to effective cross-section generation are reviewed and reframed by adopting the current enhanced methodologies for lattice calculations. The new method is composed of the following four key methods: (1) cross-section library generation method with a polynomial hyperbolic tangent formulation, (2) resonance self-shielding method based on the multi-term rational approximation for general lattice geometry and gray resonance absorbers, (3) spatially dependent gray resonance self-shielding method for generation of intra-pellet power profile and (4) integrated reaction rate preservation method between the multi-group and the ultra-fine-group calculations. From the various verifications and validations, applicability of the present resonance treatment is totally confirmed. As a result, the new resonance self-shielding method is established, not only by extension of a past concentrated effort in the reactor physics research field, but also by unification of newly developed unique and challenging techniques for practical application to the lattice physics calculations. (author)

Development and application of the automated Monte Carlo biasing procedure in SAS4

International Nuclear Information System (INIS)

Tang, J.S.; Broadhead, B.L.

1995-01-01

An automated approach for biasing Monte Carlo shielding calculations is described. In particular, adjoint fluxes from a one-dimensional discrete-ordinates calculation are used to generate biasing parameters for a three-dimensional Monte Carlo calculation. The automated procedure consisting of cross-section processing, adjoint flux determination, biasing parameter generation, and the initiation of a MORSE-SGC/S Monte Carlo calculation has been implemented in the SAS4 module of the SCALE computer code system. (author)

RZ calculations for self shielded multigroup cross sections

Energy Technology Data Exchange (ETDEWEB)

Li, M.; Sanchez, R.; Zmijarevic, I.; Stankovski, Z. [Commissariat a l' Energie Atomique CEA, Direction de l' Energie Nucleaire, DEN/DM2S/SERMA/LENR, 91191 Gif-sur-Yvette Cedex (France)

2006-07-01

A collision probability method has been implemented for RZ geometries. The method accounts for white albedo, specular and translation boundary condition on the top and bottom surfaces of the geometry and for a white albedo condition on the outer radial surface. We have applied the RZ CP method to the calculation of multigroup self shielded cross sections for Gadolinia absorbers in BWRs. (authors)

RZ calculations for self shielded multigroup cross sections

International Nuclear Information System (INIS)

Li, M.; Sanchez, R.; Zmijarevic, I.; Stankovski, Z.

2006-01-01

A collision probability method has been implemented for RZ geometries. The method accounts for white albedo, specular and translation boundary condition on the top and bottom surfaces of the geometry and for a white albedo condition on the outer radial surface. We have applied the RZ CP method to the calculation of multigroup self shielded cross sections for Gadolinia absorbers in BWRs. (authors)

Efficiency of the delta-tracking technique for Monte Carlo calculations applied to neutron-transport simulations of the advanced Candu reactor design

International Nuclear Information System (INIS)

Arsenault, Benoit; Le Tellier, Romain; Hebert, Alain

2008-01-01

The paper presents the results of a first implementation of a Monte Carlo module in DRAGON Version 4 based on the delta-tracking technique. The Monte Carlo module uses the geometry and the self-shielded multigroup cross-sections calculated with a deterministic model. The module has been tested with three different configurations of an ACR TM -type lattice. The paper also discusses the impact of this approach on the efficiency of the Monte Carlo module. (authors)

Calculation of thermal neutron self-shielding correction factors for aqueous bulk sample prompt gamma neutron activation analysis using the MCNP code

International Nuclear Information System (INIS)

Nasrabadi, M.N.; Jalali, M.; Mohammadi, A.

2007-01-01

In this work thermal neutron self-shielding in aqueous bulk samples containing neutron absorbing materials is studied using bulk sample prompt gamma neutron activation analysis (BSPGNAA) with the MCNP code. The code was used to perform three dimensional simulations of a neutron source, neutron detector and sample of various material compositions. The MCNP model was validated against experimental measurements of the neutron flux performed using a BF 3 detector. Simulations were performed to predict thermal neutron self-shielding in aqueous bulk samples containing neutron absorbing solutes. In practice, the MCNP calculations are combined with experimental measurements of the relative thermal neutron flux over the sample's surface, with respect to a reference water sample, to derive the thermal neutron self-shielding within the sample. The proposed methodology can be used for the determination of the elemental concentration of unknown aqueous samples by BSPGNAA where knowledge of the average thermal neutron flux within the sample volume is required

Photon dose estimation from ultraintense laser–solid interactions and shielding calculation with Monte Carlo simulation

International Nuclear Information System (INIS)

Yang, Bo; Qiu, Rui; Li, JunLi; Lu, Wei; Wu, Zhen; Li, Chunyan

2017-01-01

When a strong laser beam irradiates a solid target, a hot plasma is produced and high-energy electrons are usually generated (the so-called “hot electrons”). These energetic electrons subsequently generate hard X-rays in the solid target through the Bremsstrahlung process. To date, only limited studies have been conducted on this laser-induced radiological protection issue. In this study, extensive literature reviews on the physics and properties of hot electrons have been conducted. On the basis of these information, the photon dose generated by the interaction between hot electrons and a solid target was simulated with the Monte Carlo code FLUKA. With some reasonable assumptions, the calculated dose can be regarded as the upper boundary of the experimental results over the laser intensity ranging from 10 19 to 10 21 W/cm 2 . Furthermore, an equation to estimate the photon dose generated from ultraintense laser–solid interactions based on the normalized laser intensity is derived. The shielding effects of common materials including concrete and lead were also studied for the laser-driven X-ray source. The dose transmission curves and tenth-value layers (TVLs) in concrete and lead were calculated through Monte Carlo simulations. These results could be used to perform a preliminary and fast radiation safety assessment for the X-rays generated from ultraintense laser–solid interactions. - Highlights: • The laser–driven X-ray ionizing radiation source was analyzed in this study. • An equation to estimate the photon dose based on the laser intensity is given. • The shielding effects of concrete and lead were studied for this new X-ray source. • The aim of this study is to analyze and mitigate the laser–driven X-ray hazard.

Unresolved resonance self shielding calculation: causes and importance of discrepancies

International Nuclear Information System (INIS)

Ribon, P.; Tellier, H.

1986-01-01

To compute the self shielding coefficient, it is necessary to know the point-wise cross-sections. In the unresolved resonance region, the parameters of each level are not known; only the average parameters. Therefore the authors simulate the point-wise cross-section by random sampling of the energy levels and resonance parameters with respect to the Wigner law and the x 2 distributions, and by computing the cross-section in the same way as in the resolved regions. The result of this statistical calculation obviously depends on the initial parameters but also on the method of sampling, on the formalism which is used to compute the cross-section or on the weighting neutron flux. In this paper, the authors survey the main phenomena which can induce discrepancies in self shielding computations. Results are given for typical dilutions which occur in nuclear reactors

Evaluation of some resonance self-shielding procedures employed in high conversion light water reactor design

International Nuclear Information System (INIS)

Patino, N.E.; Abbate, M.J.; Sbaffoni, M.M.

1990-01-01

The procedures employed in the treatment of the resonance shielding effect have been identified as one of the causes of the large discrepancies found in the neutronic calculation of high conversion light water reactors (HCLWRs), indicating the need for a revision of the self-shielding procedures employed. In this work some well known techniques applied in HCLWR self-shielding calculations are evaluated; the study involves the comparison of methods for the generation of group constants, the analysis of the impact of considering some isotopes as infinitely diluted and the evaluation of the usual approximations utilized for the treatment of heterogeneities

Radiation shielding techniques and applications. 3. Analysis of Photon Streaming Through and Around Shield Doors

International Nuclear Information System (INIS)

Barnett, Marvin; Hack, Joe; Nathan, Steve; White, Travis

2001-01-01

Westinghouse Safety Management Solutions (Westinghouse SMS) has been tasked with providing radiological engineering design support for the new Commercial Light Water Reactor Tritium Extraction Facility (CLWR-TEF) being constructed at the Savannah River Site (SRS). The Remote Handling Building (RHB) of the CLWR-TEF will act as the receiving facility for irradiated targets used in the production of tritium for the U.S. Department of Energy (DOE). Because of the high dose rates, approaching 50 000 rads/h (500 Gy/h) from the irradiated target bundles, significant attention has been made to shielding structures within the facility. One aspect of the design that has undergone intense review is the shield doors. The RHB has six shield doors that needed to be studied with respect to photon streaming. Several aspects had to be examined to ensure that the design meets the radiation dose levels. Both the thickness and streaming issues around the door edges were designed and examined. Photon streaming through and around a shield door is a complicated problem, creating a reliance on computer modeling to perform the analyses. The computer code typically used by the Westinghouse SMS in the evaluation of photon transport through complex geometries is the MCNP Monte Carlo computer code. The complexity of the geometry within the problem can cause problems even with the Monte Carlo codes. Striking a balance between how the code handles transport through the shield door with transport through the streaming paths, particularly with the use of typical variance reduction methods, is difficult when trying to ensure that all important regions of the model are sampled appropriately. The thickness determination used a simple variance reduction technique. In construction, the shield door will not be flush against the wall, so a solid rectangular slab leaves streaming paths around the edges. Administrative controls could be used to control dose to workers; however, 10 CFR 835.1001 states

Calculation of the electron trajectory for 200 kV self-shielded electron accelerator

International Nuclear Information System (INIS)

Wang Shuiqing

2000-01-01

In order to calculate the electron trajectory of 200 kV self-shielded electron accelerator, the electric field is calculated with a TRAJ program. In this program, following electron track mash points one by one, the electron beam trajectories are calculated. Knowing the effect of grid voltage on electron optics and gaining grid voltage focusing effect in the various energy grades, the authors have gained scientific basis for adjusting grid voltage, and also accumulated a wealth of experience for designing self-shielded electron accelerator or electron curtain in future

Computational methods for high-energy source shielding

International Nuclear Information System (INIS)

Armstrong, T.W.; Cloth, P.; Filges, D.

1983-01-01

The computational methods for high-energy radiation transport related to shielding of the SNQ-spallation source are outlined. The basic approach is to couple radiation-transport computer codes which use Monte Carlo methods and discrete ordinates methods. A code system is suggested that incorporates state-of-the-art radiation-transport techniques. The stepwise verification of that system is briefly summarized. The complexity of the resulting code system suggests a more straightforward code specially tailored for thick shield calculations. A short guide line to future development of such a Monte Carlo code is given

Radiation shielding analysis

International Nuclear Information System (INIS)

Moon, S.H.; Ha, C.W.; Kwon, S.K.; Lee, J.K.; Choi, H.S.

1982-01-01

The theoretical bases of radiation streaming analysis in power reactors, such as ducts or reactor cavity, have been investigated. Discrete ordinates-Monte Carlo or Monte Carlo-Monte Carlo coupling techniques are suggested for the streaming analysis of ducts or reactor cavity. Single albedo scattering approximation code (SINALB) has been developed for simple and quick estimation of gamma-ray ceiling scattering, where the ceiling is assumed to be semi-infinite medium. This code has been employed to calculate the gamma-ray ceiling scattering effects in the laboratory containing a Co-60 source. The SINALB is applicable to gamma-ray scattering, only where the ceiling is thicker than Σsup(-1) and the height is at least twice higher than the shield wall. This code can be used for the purpose of preliminary radiation shield design. The MORSE code has been improved to analyze the gamma-ray scattering problem with on approximation method in respect to the random walk and estimation processes. This improved MORSE code has been employed to the gamma-ray ceiling scattering problem. The results of the improved MORSE calculation are in good agreement with the SINALB and standard MORSE. (Author)

Adjoint acceleration of Monte Carlo simulations using TORT/MCNP coupling approach: A case study on the shielding improvement for the cyclotron room of the Buddhist Tzu Chi General Hospital

International Nuclear Information System (INIS)

Sheu, R. J.; Sheu, R. D.; Jiang, S. H.; Kao, C. H.

2005-01-01

Full-scale Monte Carlo simulations of the cyclotron room of the Buddhist Tzu Chi General Hospital were carried out to improve the original inadequate maze design. Variance reduction techniques are indispensable in this study to facilitate the simulations for testing a variety of configurations of shielding modification. The TORT/MCNP manual coupling approach based on the Consistent Adjoint Driven Importance Sampling (CADIS) methodology has been used throughout this study. The CADIS utilises the source and transport biasing in a consistent manner. With this method, the computational efficiency was increased significantly by more than two orders of magnitude and the statistical convergence was also improved compared to the unbiased Monte Carlo run. This paper describes the shielding problem encountered, the procedure for coupling the TORT and MCNP codes to accelerate the calculations and the calculation results for the original and improved shielding designs. In order to verify the calculation results and seek additional accelerations, sensitivity studies on the space-dependent and energy-dependent parameters were also conducted. (authors)

Verification of effectiveness of borated water shield for a cyclotron type self-shielded; Verificacao da eficacia da blindagem de agua borada construida para um acelerador ciclotron do tipo autoblindado

Energy Technology Data Exchange (ETDEWEB)

Videira, Heber S.; Burkhardt, Guilherme M.; Santos, Ronielly S., E-mail: heber@cyclopet.com.br [Cyclopet Radiofarmacos Ltda., Curitiba, PR (Brazil); Passaro, Bruno M.; Gonzalez, Julia A.; Santos, Josefina; Guimaraes, Maria I.C.C. [Universidade de Sao Paulo (HCFMRP/USP), Sao Paulo, SP (Brazil). Faculdade de Medicina. Hospital das Clinicas; Lenzi, Marcelo K. [Universidade Federal do Parana (UFPR), Curitina (Brazil). Programa de Pos-Graduacao em Engenharia Quimica

2013-04-15

The technological advances in positron emission tomography (PET) in conventional clinic imaging have led to a steady increase in the number of cyclotrons worldwide. Most of these cyclotrons are being used to produce {sup 18}F-FDG, either for themselves as for the distribution to other centers that have PET. For there to be safety in radiological facilities, the cyclotron intended for medical purposes can be classified in category I and category II, ie, self-shielded or non-shielded (bunker). Therefore, the aim of this work is to verify the effectiveness of borated water shield built for a cyclotron accelerator-type Self-shielded PETtrace 860. Mixtures of water borated occurred in accordance with the manufacturer’s specifications, as well as the results of the radiometric survey in the vicinity of the self-shielding of the cyclotron in the conditions established by the manufacturer showed that radiation levels were below the limits. (author)

Unresolved resonance self shielding calculation: causes and importance of discrepancies

International Nuclear Information System (INIS)

Ribon, P.; Tellier, H.

1986-09-01

To compute the self shielding coefficient, it is necessary to know the point-wise cross-sections. In the unresolved resonance region, we do not know the parameters of each level but only the average parameters. Therefore we simulate the point-wise cross-section by random sampling of the energy levels and resonance parameters with respect to the Wigner law and the X 2 distributions, and by computing the cross-section in the same way as in the resolved regions. The result of this statistical calculation obviously depends on the initial parameters but also on the method of sampling, on the formalism which is used to compute the cross-section or on the weighting neutron flux. In this paper, we will survey the main phenomena which can induce discrepancies in self shielding computations. Results are given for typical dilutions which occur in nuclear reactors. 8 refs

Monte Carlo simulations for the optimisation of low-background Ge detector designs

Energy Technology Data Exchange (ETDEWEB)

Hakenmueller, Janina; Heusser, Gerd; Maneschg, Werner; Schreiner, Jochen; Simgen, Hardy; Stolzenburg, Dominik; Strecker, Herbert; Weber, Marc; Westernmann, Jonas [Max-Planck-Institut fuer Kernphysik, Saupfercheckweg 1, 69117 Heidelberg (Germany); Laubenstein, Matthias [Laboratori Nazionali del Gran Sasso, Via G. Acitelli 22, 67100 Assergi L' Aquila (Italy)

2015-07-01

Monte Carlo simulations for the low-background Ge spectrometer Giove at the underground laboratory of MPI-K, Heidelberg, are presented. In order to reduce the cosmogenic background at the present shallow depth (15 m w.e.) the shielding of the spectrometer includes an active muon veto and a passive shielding (lead and borated PE layers). The achieved background suppression is comparable to Ge spectrometers operated in much greater depth. The geometry of the detector and the shielding were implemented using the Geant4-based toolkit MaGe. The simulations were successfully optimised by determining the correct diode position and active volume. With the help of the validated Monte Carlo simulation the contribution of the single components to the overall background can be examined. This includes a comparison between simulated results and measurements with different fillings of the sample chamber. Having reproduced the measured detector background in the simulation provides the possibility to improve the background by reverse engineering of the passive and active shield layers in the simulation.

A code for leakage neutron spectra through thick shields

International Nuclear Information System (INIS)

Nagarajan, P.S.; Sethulakshmi, P.; Raghavendran, C.P.

1975-01-01

An exponential transform Monte Carlo code has been developed for deep penetration of neutrons and the results of leakage neutron spectra of this code have been compared with those of a basic Monte Carlo code for small thickness. The development of the code and optimisation of certain transform parameters are discussed and results are presented for a few thick shields of concrete and water in the context of neutron monitoring in the environs of accelerator and reactor shields. (author)

Monte Carlo computation of Bremsstrahlung intensity and energy spectrum from a 15 MV linear electron accelerator tungsten target to optimise LINAC head shielding

International Nuclear Information System (INIS)

Biju, K.; Sharma, Amiya; Yadav, R.K.; Kannan, R.; Bhatt, B.C.

2003-01-01

The knowledge of exact photon intensity and energy distributions from the target of an electron target is necessary while designing the shielding for the accelerator head from radiation safety point of view. The computations were carried out for the intensity and energy distribution of photon spectrum from a 0.4 cm thick tungsten target in different angular directions for 15 MeV electrons using a validated Monte Carlo code MCNP4A. Similar results were computed for 30 MeV electrons and found agreeing with the data available in literature. These graphs and the TVT values in lead help to suggest an optimum shielding thickness for 15 MV Linac head. (author)

Shielding measurements for a 230 MeV proton beam

International Nuclear Information System (INIS)

Siebers, J.V.

1990-01-01

Energetic secondary neutrons produced as protons interact with accelerator components and patients dominate the radiation shielding environment for proton radiotherapy facilities. Due to the scarcity of data describing neutron production, attenuation, absorbed dose, and dose equivalent values, these parameters were measured for 230 MeV proton bombardment of stopping length Al, Fe, and Pb targets at emission angles of 0 degree, 22 degree, 45 degree, and 90 degree in a thick concrete shield. Low pressure tissue-equivalent proportional counters with volumes ranging from 1 cm 3 to 1000 cm 3 were used to obtain microdosimetric spectra from which absorbed dose and radiation quality are deduced. Does equivalent values and attenuation lengths determined at depth in the shield were found to vary sharply with angle, but were found to be independent of target material. Neutron dose and radiation length values are compared with Monte Carlo neutron transport calculations performed using the Los Alamos High Energy Transport Code (LAHET). Calculations used 230 MeV protons incident upon an Fe target in a shielding geometry similar to that used in the experiment. LAHET calculations overestimated measured attenuation values at 0 degree, 22 degree, and 45 degree, yet correctly predicted the attenuation length at 90 degree. Comparison of the mean radiation quality estimated with the Monte Carlo calculations with measurements suggest that neutron quality factors should be increased by a factor of 1.4. These results are useful for the shielding design of new facilities as well as for testing neutron production and transport calculations

Evaluation of Shielding Performance for Newly Developed Composite Materials

Science.gov (United States)

Evans, Beren Richard

This work details an investigation into the contributing factors behind the success of newly developed composite neutron shield materials. Monte Carlo simulation methods were utilized to assess the neutron shielding capabilities and secondary radiation production characteristics of aluminum boron carbide, tungsten boron carbide, bismuth borosilicate glass, and Metathene within various neutron energy spectra. Shielding performance and secondary radiation data suggested that tungsten boron carbide was the most effective composite material. An analysis of the macroscopic cross-section contributions from constituent materials and interaction mechanisms was then performed in an attempt to determine the reasons for tungsten boron carbide's success over the other investigated materials. This analysis determined that there was a positive correlation between a non-elastic interaction contribution towards a material's total cross-section and shielding performance within the thermal and epi-thermal energy regimes. This finding was assumed to be a result of the boron-10 absorption reaction. The analysis also determined that within the faster energy regions, materials featuring higher non-elastic interaction contributions were comparable to those exhibiting primarily elastic scattering via low Z elements. This allowed for the conclusion that composite shield success within higher energy neutron spectra does not necessitate the use elastic scattering via low Z elements. These findings suggest that the inclusion of materials featuring high thermal absorption properties is more critical to composite neutron shield performance than the presence of constituent materials more inclined to maximize elastic scattering energy loss.

Cosmic Ray Interactions in Shielding Materials

International Nuclear Information System (INIS)

Aguayo Navarrete, Estanislao; Kouzes, Richard T.; Ankney, Austin S.; Orrell, John L.; Berguson, Timothy J.; Troy, Meredith D.

2011-01-01

This document provides a detailed study of materials used to shield against the hadronic particles from cosmic ray showers at Earth's surface. This work was motivated by the need for a shield that minimizes activation of the enriched germanium during transport for the MAJORANA collaboration. The materials suitable for cosmic-ray shield design are materials such as lead and iron that will stop the primary protons, and materials like polyethylene, borated polyethylene, concrete and water that will stop the induced neutrons. The interaction of the different cosmic-ray components at ground level (protons, neutrons, muons) with their wide energy range (from kilo-electron volts to giga-electron volts) is a complex calculation. Monte Carlo calculations have proven to be a suitable tool for the simulation of nucleon transport, including hadron interactions and radioactive isotope production. The industry standard Monte Carlo simulation tool, Geant4, was used for this study. The result of this study is the assertion that activation at Earth's surface is a result of the neutronic and protonic components of the cosmic-ray shower. The best material to shield against these cosmic-ray components is iron, which has the best combination of primary shielding and minimal secondary neutron production.

Study on bulk shielding for a spallation neutron source facility in the high-intensity proton accelerator project

CERN Document Server

Maekawa, F; Takada, H; Teshigawara, M; Watanabe, N

2002-01-01

Under the JAERI-KEK High-Intensity Proton Accelerator Project, a spallation neutron source driven by a 3 GeV-1 MW proton beam is planed to be constructed in a main part of the Materials and Life Science Facility. This report describes results of a study on bulk shielding performance of a biological shield for the spallation neutron source by means of a Monte Carlo calculation method, that is important in terms of radiation safety and cost reduction. A shielding configuration was determined as a reference case by considering preliminary studies and interaction with other components, then shielding thickness that was required to achieve a target dose rate of 1 mu Sv/h was derived. Effects of calculation conditions such as shielding materials and dimensions on the shielding performance was investigated by changing those parameters. By taking all the results and design margins into account, a shielding configuration that was identified as the most appropriate was finally determined as follows. An iron shield regi...

Determination of 210Pb activity concentration in lead shielding

International Nuclear Information System (INIS)

Slivka, J.; Mrdja, D.; Varga, E.; Veskovic, M.

2005-01-01

210 Pb is concentrated during the separation lead from the ore and therefore it is the main pollutant of lead products. The content of this isotope limits the applicability of lead for low-level shielding of gamma spectrometers. In this paper, a new method for the determination of 210 Pb activity concentration in lead shielding from 46.5 keV gamma line intensity is presented. (author) [sr

Wake Shield Target Protection

International Nuclear Information System (INIS)

Valmianski, Emanuil I.; Petzoldt, Ronald W.; Alexander, Neil B.

2003-01-01

The heat flux from both gas convection and chamber radiation on a direct drive target must be limited to avoid target damage from excessive D-T temperature increase. One of the possibilities of protecting the target is a wake shield flying in front of the target. A shield will also reduce drag force on the target, thereby facilitating target tracking and position prediction. A Direct Simulation Monte Carlo (DSMC) code was used to calculate convection heat loads as boundary conditions input into ANSYS thermal calculations. These were used for studying the quality of target protection depending on various shapes of shields, target-shield distance, and protective properties of the shield moving relative to the target. The results show that the shield can reduce the convective heat flux by a factor of 2 to 5 depending on pressure, temperature, and velocity. The protective effect of a shield moving relative to the target is greater than the protective properties of a fixed shield. However, the protective effect of a shield moving under the drag force is not sufficient for bringing the heat load on the target down to the necessary limit. Some other ways of diminishing heat flux using a protective shield are discussed

Using FLUKA to Study Concrete Square Shield Performance in Attenuation of Neutron Radiation Produced by APF Plasma Focus Neutron Source

Science.gov (United States)

Nemati, M. J.; Habibi, M.; Amrollahi, R.

2013-04-01

In 2010, representatives from the Nuclear Engineering and physics Department of Amirkabir University of Technology (AUT) requested development of a project with the objective of determining the performance of a concrete shield for their Plasma Focus as neutron source. The project team in Laboratory of Nuclear Engineering and physics department of Amirkabir University of Technology choose some shape of shield to study on their performance with Monte Carlo code. In the present work, the capability of Monte Carlo code FLUKA will be explored to model the APF Plasma Focus, and investigating the neutron fluence on the square concrete shield in each region of problem. The physical models embedded in FLUKA are mentioned, as well as examples of benchmarking against future experimental data. As a result of this study suitable thickness of concrete for shielding APF will be considered.

Benchmarking study and its application for shielding analysis of large accelerator facilities

Energy Technology Data Exchange (ETDEWEB)

Lee, Hee-Seock; Kim, Dong-hyun; Oranj, Leila Mokhtari; Oh, Joo-Hee; Lee, Arim; Jung, Nam-Suk [POSTECH, Pohang (Korea, Republic of)

2015-10-15

Shielding Analysis is one of subjects which are indispensable to construct large accelerator facility. Several methods, such as the Monte Carlo, discrete ordinate, and simplified calculation, have been used for this purpose. The calculation precision is overcome by increasing the trial (history) numbers. However its accuracy is still a big issue in the shielding analysis. To secure the accuracy in the Monte Carlo calculation, the benchmarking study using experimental data and the code comparison are adopted fundamentally. In this paper, the benchmarking result for electrons, protons, and heavy ions are presented as well as the proper application of the results is discussed. The benchmarking calculations, which are indispensable in the shielding analysis were performed for different particles: proton, heavy ion and electron. Four different multi-particle Monte Carlo codes, MCNPX, FLUKA, PHITS, and MARS, were examined for higher energy range equivalent to large accelerator facility. The degree of agreement between the experimental data including the SINBAD database and the calculated results were estimated in the terms of secondary neutron production and attenuation through the concrete and iron shields. The degree of discrepancy and the features of Monte Carlo codes were investigated and the application way of the benchmarking results are discussed in the view of safety margin and selecting the code for the shielding analysis. In most cases, the tested Monte Carlo codes give proper credible results except of a few limitation of each codes.

Benchmarking study and its application for shielding analysis of large accelerator facilities

International Nuclear Information System (INIS)

Lee, Hee-Seock; Kim, Dong-hyun; Oranj, Leila Mokhtari; Oh, Joo-Hee; Lee, Arim; Jung, Nam-Suk

2015-01-01

Shielding Analysis is one of subjects which are indispensable to construct large accelerator facility. Several methods, such as the Monte Carlo, discrete ordinate, and simplified calculation, have been used for this purpose. The calculation precision is overcome by increasing the trial (history) numbers. However its accuracy is still a big issue in the shielding analysis. To secure the accuracy in the Monte Carlo calculation, the benchmarking study using experimental data and the code comparison are adopted fundamentally. In this paper, the benchmarking result for electrons, protons, and heavy ions are presented as well as the proper application of the results is discussed. The benchmarking calculations, which are indispensable in the shielding analysis were performed for different particles: proton, heavy ion and electron. Four different multi-particle Monte Carlo codes, MCNPX, FLUKA, PHITS, and MARS, were examined for higher energy range equivalent to large accelerator facility. The degree of agreement between the experimental data including the SINBAD database and the calculated results were estimated in the terms of secondary neutron production and attenuation through the concrete and iron shields. The degree of discrepancy and the features of Monte Carlo codes were investigated and the application way of the benchmarking results are discussed in the view of safety margin and selecting the code for the shielding analysis. In most cases, the tested Monte Carlo codes give proper credible results except of a few limitation of each codes

Bonderenko self-shielded cross sections and multiband parameters derived from the LLL Evaluated-Nuclear-Data Library (ENDL)

International Nuclear Information System (INIS)

Cullen, D.E.

1978-01-01

Bonderenko self-shielded cross sections and multiband parameters from the Lawrence Livermore Laboratory Evaluated-Nuclear-Data Library (ENDL) as of July 4, 1978 are presented. These data include total, elastic, capture, and fission cross sections in the TART 175 group structure. Multiband parameters are listed. Bonderenko self-shielded cross section and the multiband parameters are presented on microfiche

Accelerator shield design of KIPT neutron source facility

International Nuclear Information System (INIS)

Zhong, Z.; Gohar, Y.

2013-01-01

Argonne National Laboratory (ANL) of the United States and Kharkov Institute of Physics and Technology (KIPT) of Ukraine have been collaborating on the design development of a neutron source facility at KIPT utilizing an electron-accelerator-driven subcritical assembly. Electron beam power is 100 kW, using 100 MeV electrons. The facility is designed to perform basic and applied nuclear research, produce medical isotopes, and train young nuclear specialists. The biological shield of the accelerator building is designed to reduce the biological dose to less than 0.5-mrem/hr during operation. The main source of the biological dose is the photons and the neutrons generated by interactions of leaked electrons from the electron gun and accelerator sections with the surrounding concrete and accelerator materials. The Monte Carlo code MCNPX serves as the calculation tool for the shield design, due to its capability to transport electrons, photons, and neutrons coupled problems. The direct photon dose can be tallied by MCNPX calculation, starting with the leaked electrons. However, it is difficult to accurately tally the neutron dose directly from the leaked electrons. The neutron yield per electron from the interactions with the surrounding components is less than 0.01 neutron per electron. This causes difficulties for Monte Carlo analyses and consumes tremendous computation time for tallying with acceptable statistics the neutron dose outside the shield boundary. To avoid these difficulties, the SOURCE and TALLYX user subroutines of MCNPX were developed for the study. The generated neutrons are banked, together with all related parameters, for a subsequent MCNPX calculation to obtain the neutron and secondary photon doses. The weight windows variance reduction technique is utilized for both neutron and photon dose calculations. Two shielding materials, i.e., heavy concrete and ordinary concrete, were considered for the shield design. The main goal is to maintain the total

Shielding design of ITER pressure suppression system

International Nuclear Information System (INIS)

Yamauchi, Michinori; Sato, Satoshi; Nishitani, Takeo; Kawasaki, Hiromitsu

2006-01-01

The duct shield from streaming D-T neutrons has been designed for the ITER pressure suppression system. Streaming calculations are performed with the DUCT-III code for the region from the inlet of the pressure relief line to the rupture disk. Next, the neutron permeation through the shield is studied by Monte Carlo calculations with the MCNP code. It is found that 0.15 m thick iron shield is enough to suppress the permeating component from the outside. In addition, it is suggested that the volume of the shield can be reduced by about 30% if the optimized iron shield structure having localized thickness across intense permeation paths is employed to shield the pressure suppression line. (T.I.)

CREST : a computer program for the calculation of composition dependent self-shielded cross-sections

International Nuclear Information System (INIS)

Kapil, S.K.

1977-01-01

A computer program CREST for the calculation of the composition and temperature dependent self-shielded cross-sections using the shielding factor approach has been described. The code includes the editing and formation of the data library, calculation of the effective shielding factors and cross-sections, a fundamental mode calculation to generate the neutron spectrum for the system which is further used to calculate the effective elastic removal cross-sections. Studies to explore the sensitivity of reactor parameters to changes in group cross-sections can also be carried out by using the facility available in the code to temporarily change the desired constants. The final self-shielded and transport corrected group cross-sections can be dumped on cards or magnetic tape in a suitable form for their direct use in a transport or diffusion theory code for detailed reactor calculations. The program is written in FORTRAN and can be accommodated in a computer with 32 K work memory. The input preparation details, sample problem and the listing of the program are given. (author)

Success and prospects for low energy, self-shielded electron beam accelerators

International Nuclear Information System (INIS)

Laeuppi, U.V.

1988-01-01

The advantages of self-shielded, low energy, electron beam accelerators for electron beam processing are described. Applications of these accelerators for cross-linking plastic films, drying of coated materials and printing inks and for curing processes are discussed. (U.K.)

Computation of a Canadian SCWR unit cell with deterministic and Monte Carlo codes

International Nuclear Information System (INIS)

Harrisson, G.; Marleau, G.

2012-01-01

The Canadian SCWR has the potential to achieve the goals that the generation IV nuclear reactors must meet. As part of the optimization process for this design concept, lattice cell calculations are routinely performed using deterministic codes. In this study, the first step (self-shielding treatment) of the computation scheme developed with the deterministic code DRAGON for the Canadian SCWR has been validated. Some options available in the module responsible for the resonance self-shielding calculation in DRAGON 3.06 and different microscopic cross section libraries based on the ENDF/B-VII.0 evaluated nuclear data file have been tested and compared to a reference calculation performed with the Monte Carlo code SERPENT under the same conditions. Compared to SERPENT, DRAGON underestimates the infinite multiplication factor in all cases. In general, the original Stammler model with the Livolant-Jeanpierre approximations are the most appropriate self-shielding options to use in this case of study. In addition, the 89 groups WIMS-AECL library for slight enriched uranium and the 172 groups WLUP library for a mixture of plutonium and thorium give the most consistent results with those of SERPENT. (authors)

AUTOSECOL: an automatic calculation of the self-shielding of heavy isotope resonances

International Nuclear Information System (INIS)

Grandotto-Biettoli, Marc.

The formalism is based on separating both types of resonance effects: local energy effects creating a fine structure in the flux, and bulk effects resulting in a slow variation in the flux. Effective reaction rates are defined that, used as tables in a multigroup calculation of cells with a large pitch in regard to resonance widths, allow an exact account of the dependence of the effective integral upon fast variations in the flux. These tables are used to introduce this phenomenon of resonance self-shielding in the multigroup Apollo program for solving the neutron transport equation, they are derived from nuclear data with using some parameters relating to the physical state of the resonant isotope inside the fuel medium. The AUTOSECOL system provides a library of effective reaction rates for taking account of the resonance self-shielding effect on the neutron flux in nuclear reactor cells. Its versatility in regard to the methods previously used for solving the same problem allows a rapid testing of the consequences of considering the self-shielding effect of new isotope resonances, a following up of the evolution in nuclear data evaluation, and rapidly studying the interest lying in new data. Results obtained with AUTOSECOL are compared with those obtained when using the SECOL code for computing the effective reaction rates of 235 U, 239 Pu, 107 Ag, 109 Ag, and 241 Pu [fr

Monte Carlo simulations for the shielding of the future high-intensity accelerator facility FAIR at GSI.

Science.gov (United States)

Radon, T; Gutermuth, F; Fehrenbacher, G

2005-01-01

The Gesellschaft für Schwerionenforschung (GSI) is planning a significant expansion of its accelerator facilities. Compared to the present GSI facility, a factor of 100 in primary beam intensities and up to a factor of 10,000 in secondary radioactive beam intensities are key technical goals of the proposal. The second branch of the so-called Facility for Antiproton and Ion Research (FAIR) is the production of antiprotons and their storage in rings and traps. The facility will provide beam energies a factor of approximately 15 higher than presently available at the GSI for all ions, from protons to uranium. The shielding design of the synchrotron SIS 100/300 is shown exemplarily by using Monte Carlo calculations with the FLUKA code. The experimental area serving the investigation of compressed baryonic matter is analysed in the same way. In addition, a dose comparison is made for an experimental area operated with medium energy heavy-ion beams. Here, Monte Carlo calculations are performed by using either heavy-ion primary particles or proton beams with intensities scaled by the mass number of the corresponding heavy-ion beam.

Monte Carlo simulations for the shielding of the future high-intensity accelerator facility fair at GSI

International Nuclear Information System (INIS)

Radon, T.; Gutermuth, F.; Fehrenbacher, G.

2005-01-01

The Gesellschaft fuer Schwerionenforschung (GSI) is planning a significant expansion of its accelerator facilities. Compared to the present GSI facility, a factor of 100 in primary beam intensities and up to a factor of 10,000 in secondary radioactive beam intensities are key technical goals of the proposal. The second branch of the so-called Facility for Antiproton and Ion Research (FAIR) is the production of antiprotons and their storage in rings and traps. The facility will provide beam energies a factor of ∼15 higher than presently available at the GSI for all ions, from protons to uranium. The shielding design of the synchrotron SIS 100/300 is shown exemplarily by using Monte Carlo calculations with the FLUKA code. The experimental area serving the investigation of compressed baryonic matter is analysed in the same way. In addition, a dose comparison is made for an experimental area operated with medium energy heavy-ion beams. Here, Monte Carlo calculations are performed by using either heavy-ion primary particles or proton beams with intensities scaled by the mass number of the corresponding heavy-ion beam. (authors)

Multifunctional Shielding and Self-Healing HybridSil Smart Composites for Space, Phase II

Data.gov (United States)

National Aeronautics and Space Administration — NanoSonic has developed revolutionary multifunctional, super lightweight, self-healing and radiation shielding carbon fiber reinforced polymer (CFRP) composites as a...

SELF-ABSORPTION CORRECTIONS BASED ON MONTE CARLO SIMULATIONS

Directory of Open Access Journals (Sweden)

Kamila Johnová

2016-12-01

Full Text Available The main aim of this article is to demonstrate how Monte Carlo simulations are implemented in our gamma spectrometry laboratory at the Department of Dosimetry and Application of Ionizing Radiation in order to calculate the self-absorption within the samples. A model of real HPGe detector created for MCNP simulations is presented in this paper. All of the possible parameters, which may influence the self-absorption, are at first discussed theoretically and lately described using the calculated results.

Accurate Monte Carlo modeling of cyclotrons for optimization of shielding and activation calculations in the biomedical field

Science.gov (United States)

Infantino, Angelo; Marengo, Mario; Baschetti, Serafina; Cicoria, Gianfranco; Longo Vaschetto, Vittorio; Lucconi, Giulia; Massucci, Piera; Vichi, Sara; Zagni, Federico; Mostacci, Domiziano

2015-11-01

Biomedical cyclotrons for production of Positron Emission Tomography (PET) radionuclides and radiotherapy with hadrons or ions are widely diffused and established in hospitals as well as in industrial facilities and research sites. Guidelines for site planning and installation, as well as for radiation protection assessment, are given in a number of international documents; however, these well-established guides typically offer analytic methods of calculation of both shielding and materials activation, in approximate or idealized geometry set up. The availability of Monte Carlo codes with accurate and up-to-date libraries for transport and interactions of neutrons and charged particles at energies below 250 MeV, together with the continuously increasing power of nowadays computers, makes systematic use of simulations with realistic geometries possible, yielding equipment and site specific evaluation of the source terms, shielding requirements and all quantities relevant to radiation protection. In this work, the well-known Monte Carlo code FLUKA was used to simulate two representative models of cyclotron for PET radionuclides production, including their targetry; and one type of proton therapy cyclotron including the energy selection system. Simulations yield estimates of various quantities of radiological interest, including the effective dose distribution around the equipment, the effective number of neutron produced per incident proton and the activation of target materials, the structure of the cyclotron, the energy degrader, the vault walls and the soil. The model was validated against experimental measurements and comparison with well-established reference data. Neutron ambient dose equivalent H*(10) was measured around a GE PETtrace cyclotron: an average ratio between experimental measurement and simulations of 0.99±0.07 was found. Saturation yield of 18F, produced by the well-known 18O(p,n)18F reaction, was calculated and compared with the IAEA recommended

RADSHI: shielding calculation program for different geometries sources

International Nuclear Information System (INIS)

Gelen, A.; Alvarez, I.; Lopez, H.; Manso, M.

1996-01-01

A computer code written in pascal language for IBM/Pc is described. The program calculates the optimum thickness of slab shield for different geometries sources. The Point Kernel Method is employed, which enables the obtention of the ionizing radiation flux density. The calculation takes into account the possibility of self-absorption in the source. The air kerma rate for gamma radiation is determined, and with the concept of attenuation length through the equivalent attenuation length the shield is obtained. The scattering and the exponential attenuation inside the shield material is considered in the program. The shield materials can be: concrete, water, iron or lead. It also calculates the shield for point isotropic neutron source, using as shield materials paraffin, concrete or water. (authors). 13 refs

Neutron radiation shielding properties of polymer incorporated self compacting concrete mixes.

Science.gov (United States)

Malkapur, Santhosh M; Divakar, L; Narasimhan, Mattur C; Karkera, Narayana B; Goverdhan, P; Sathian, V; Prasad, N K

2017-07-01

In this work, the neutron radiation shielding characteristics of a class of novel polymer-incorporated self-compacting concrete (PISCC) mixes are evaluated. Pulverized high density polyethylene (HDPE) material was used, at three different reference volumes, as a partial replacement to river sand in conventional concrete mixes. By such partial replacement of sand with polymer, additional hydrogen contents are incorporated in these concrete mixes and their effect on the neutron radiation shielding properties are studied. It has been observed from the initial set of experiments that there is a definite trend of reductions in the neutron flux and dose transmission factor values in these PISCC mixes vis-à-vis ordinary concrete mix. Also, the fact that quite similar enhanced shielding results are recorded even when reprocessed HDPE material is used in lieu of the virgin HDPE attracts further attention. Copyright © 2017 Elsevier Ltd. All rights reserved.

Self-test Monte Carlo method

International Nuclear Information System (INIS)

Ohta, Shigemi

1996-01-01

The Self-Test Monte Carlo (STMC) method resolves the main problems in using algebraic pseudo-random numbers for Monte Carlo (MC) calculations: that they can interfere with MC algorithms and lead to erroneous results, and that such an error often cannot be detected without known exact solution. STMC is based on good randomness of about 10 10 bits available from physical noise or transcendental numbers like π = 3.14---. Various bit modifiers are available to get more bits for applications that demands more than 10 10 random bits such as lattice quantum chromodynamics (QCD). These modifiers are designed so that a) each of them gives a bit sequence comparable in randomness as the original if used separately from each other, and b) their mutual interference when used jointly in a single MC calculation is adjustable. Intermediate data of the MC calculation itself are used to quantitatively test and adjust the mutual interference of the modifiers in respect of the MC algorithm. STMC is free of systematic error and gives reliable statistical error. Also it can be easily implemented on vector and parallel supercomputers. (author)

Continuous Energy, Multi-Dimensional Transport Calculations for Problem Dependent Resonance Self-Shielding

International Nuclear Information System (INIS)

Downar, T.

2009-01-01

The overall objective of the work here has been to eliminate the approximations used in current resonance treatments by developing continuous energy multi-dimensional transport calculations for problem dependent self-shielding calculations. The work here builds on the existing resonance treatment capabilities in the ORNL SCALE code system. The overall objective of the work here has been to eliminate the approximations used in current resonance treatments by developing continuous energy multidimensional transport calculations for problem dependent self-shielding calculations. The work here builds on the existing resonance treatment capabilities in the ORNL SCALE code system. Specifically, the methods here utilize the existing continuous energy SCALE5 module, CENTRM, and the multi-dimensional discrete ordinates solver, NEWT to develop a new code, CENTRM( ) NEWT. The work here addresses specific theoretical limitations in existing CENTRM resonance treatment, as well as investigates advanced numerical and parallel computing algorithms for CENTRM and NEWT in order to reduce the computational burden. The result of the work here will be a new computer code capable of performing problem dependent self-shielding analysis for both existing and proposed GENIV fuel designs. The objective of the work was to have an immediate impact on the safety analysis of existing reactors through improvements in the calculation of fuel temperature effects, as well as on the analysis of more sophisticated GENIV/NGNP systems through improvements in the depletion/transmutation of actinides for Advanced Fuel Cycle Initiatives.

SUBGR: A Program to Generate Subgroup Data for the Subgroup Resonance Self-Shielding Calculation

Energy Technology Data Exchange (ETDEWEB)

Kim, Kang Seog [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

2016-06-06

The Subgroup Data Generation (SUBGR) program generates subgroup data, including levels and weights from the resonance self-shielded cross section table as a function of background cross section. Depending on the nuclide and the energy range, these subgroup data can be generated by (a) narrow resonance approximation, (b) pointwise flux calculations for homogeneous media; and (c) pointwise flux calculations for heterogeneous lattice cells. The latter two options are performed by the AMPX module IRFFACTOR. These subgroup data are to be used in the Consortium for Advanced Simulation of Light Water Reactors (CASL) neutronic simulator MPACT, for which the primary resonance self-shielding method is the subgroup method.

Deep-penetration calculations in concrete and iron for shielding of proton therapy accelerators

International Nuclear Information System (INIS)

Sheu, Rong-Jiun; Chen, Yen-Fu; Lin, Uei-Tyng; Jiang, Shiang-Huei

2012-01-01

Proton accelerators in the energy range of approximately 200 MeV have become increasingly popular for cancer treatment in recent years. These proton therapy facilities usually involve bulky concrete or iron in their shielding design or accelerator structure. Simple shielding data, such as source terms or attenuation lengths for various proton energies and materials are useful in designing accelerator shielding. Understanding the appropriateness or uncertainties associated with these data, which are largely generated from Monte Carlo simulations, is critical to the quality of a shielding design. This study demonstrated and investigated the problems of deep-penetration calculations on the estimation of shielding parameters through an extensive comparison between the FLUKA and MCNPX calculations for shielding against a 200-MeV proton beam hitting an iron target. Simulations of double-differential neutron production from proton bombardment were validated by comparison with experimental data. For the concrete shielding, the FLUKA calculated depth–dose distributions were consistent with the MCNPX results, except for some discrepancies in backward directions. However, for the iron shielding, if FLUKA is used inappropriately then overestimation of neutron attenuation can be expected as shown by this work because of the multigroup treatment for low-energy neutrons in FLUKA. Two neutron energy group structures, three degrees of self-shielding correction, and two iron compositions were considered in this study. Significant variation of the resulting attenuation lengths indicated the importance of problem-dependent multigroup cross sections and proper modeling of iron composition in deep-penetration calculations.

Protective effect of blue-light shield eyewear for adults against light pollution from self-luminous devices used at night.

Science.gov (United States)

Ayaki, Masahiko; Hattori, Atsuhiko; Maruyama, Yusuke; Nakano, Masaki; Yoshimura, Michitaka; Kitazawa, Momoko; Negishi, Kazuno; Tsubota, Kazuo

2016-01-01

We investigated sleep quality and melatonin in 12 adults who wore blue-light shield or control eyewear 2 hours before sleep while using a self-luminous portable device, and assessed visual quality for the two eyewear types. Overnight melatonin secretion was significantly higher after using the blue-light shield (P light shield (P light shield as providing acceptable visual quality.

Shielding analyses: the rabbit vs the turtle?

International Nuclear Information System (INIS)

Broadhead, B.L.

1996-01-01

This paper compares solutions using Monte Carlo and discrete- ordinates methods applied to two actual shielding situations in order to make some general observations concerning the efficiency and advantages/disadvantages of the two approaches. The discrete- ordinates solutions are performed using two-dimensional geometries, while the Monte Carlo approaches utilize three-dimensional geometries with both multigroup and point cross-section data

Calculation of self-shielding coefficients, flux depression and cadmium factor for thermal neutron flux measurement of the IPEN/MB-01 reactor

International Nuclear Information System (INIS)

Marques, Andre Luis Ferreira; Ting, Daniel Kao Sun; Mendonca, Arlindo Gilson

1996-01-01

A calculation methodology of Flux Depression, Self-Shielding and Cadmium Factors is presented, using the ANISN code, for experiments conducted at the IPEN/MB-01 Research Reactor. The correction factors were determined considering thermal neutron flux and 0.125 e 0.250 mm diameter of 197 Au wires. (author)

Analysis of mixed oxides critical experiments using the Hammer-Technion code with self-shielding treatment by Bondarenko method

International Nuclear Information System (INIS)

Abe, Alfredo Y.; Santos, Adimir dos

1995-01-01

The present work summarizes the verification of the treatment of self-shielding based on Bondarenko method in HAMMER-TECHNION cell code for the Pu O 2 -U O 2 critical system using JENDL-3 nuclear data library. The results obtained are in excellent agreement with the original treatment of self-shielding employed by HAMMER-TECHNION cell code. (author). 9 refs, 1 fig, 9 tabs

Conservative method for determination of material thickness used in shielding of veterinary facilities

International Nuclear Information System (INIS)

Lava, Deise D.; Borges, Diogo da S.; Affonso, Renato R.W.; Moreira, Maria de L.; Guimaraes, Antonio C.F.

2014-01-01

For determination of an effective method for shielding of veterinary rooms, was provided shielding methods generally used in rooms which works with X-ray production and radiotherapy. Every calculation procedure is based in traditional variables used to transmission calculation. The thickness of the materials used for primary and secondary shieldings are obtained to respect the limits set by the Brazilian National Nuclear Energy Commission (CNEN). This work presents the development of a computer code in order to serve as a practical tool for determining rapid and effective materials and their thicknesses to shield veterinary facilities. The code determines transmission values of the shieldings and compares them with data from transmission 'maps' provided by NCRP-148 report. These 'maps' were added to the algorithm through interpolation techniques of curves of materials used for shielding. Each interpolation generates about 1,000,000 points that are used to generate a new curve. The new curve is subjected to regression techniques, which makes possible to obtain nine degree polynomial, and exponential equations. These equations whose variables consist of transmission of values, enable trace all the points of this curve with high precision. The data obtained from the algorithm were satisfactory with official data presented by the National Council of Radiation Protection and Measurements (NCRP) and can contribute as a practical tool for verification of shielding of veterinary facilities that require using Radiotherapy techniques and X-ray production

Radiation shielding for neutron guides

International Nuclear Information System (INIS)

Ersez, T.; Braoudakis, G.; Osborn, J.C.

2005-01-01

Full text: Models of the neutron guide shielding for the out of bunker guides on the thermal and cold neutron beam lines of the OPAL Reactor (ANSTO) were constructed using the Monte Carlo code MCNP 4B. The neutrons that were not reflected inside the guides but were absorbed by the supermirror (SM) layers were noted to be a significant source of gammas. Gammas also arise from neutrons absorbed by the B, Si, Na and K contained in the glass. The proposed shielding design has produced compact shielding assemblies. These arrangements are consistent with safety requirements, floor load limits, and cost constraints. To verify the design a prototype was assembled consisting of 120mm thick Pb(96%)Sb(4%) walls resting on a concrete block. There was good agreement between experimental measurements and calculated dose rates for bulk shield regions. (authors)

Radiation shielding for neutron guides

International Nuclear Information System (INIS)

Ersez, T.; Braoudakis, G.; Osborn, J.C.

2006-01-01

Models of the neutron guide shielding for the out of bunker guides on the thermal and cold neutron beam lines of the OPAL Reactor (ANSTO) were constructed using the Monte Carlo code MCNP 4B. The neutrons that were not reflected inside the guides but were absorbed by the supermirror (SM) layers were noted to be a significant source of gammas. Gammas also arise from neutrons absorbed by the B, Si, Na and K contained in the glass. The proposed shielding design has produced compact shielding assemblies. These arrangements are consistent with safety requirements, floor load limits, and cost constraints. To verify the design a prototype was assembled consisting of 120 mm thick Pb(96%)Sb(4%) walls resting on a concrete block. There was good agreement between experimental measurements and calculated dose rates for bulk shield regions

Dose rate evaluation of body phantom behind ITER bio-shield wall using Monte Carlo method

International Nuclear Information System (INIS)

Beheshti, A.; Jabbari, I.; Karimian, A.; Abdi, M.

2012-01-01

One of the most critical risks to humans in reactors environment is radiation exposure. Around the tokamak hall personnel are exposed to a wide range of particles, including neutrons and photons. International Thermonuclear Experimental Reactor (ITER) is a nuclear fusion research and engineering project, which is the most advanced experimental tokamak nuclear fusion reactor. Dose rates assessment and photon radiation due to the neutron activation of the solid structures in ITER is important from the radiological point of view. Therefore, the dosimetry considered in this case is based on the Deuterium-Tritium (DT) plasma burning with neutrons production rate at 14.1 MeV. The aim of this study is assessment the amount of radiation behind bio-shield wall that a human received during normal operation of ITER by considering neutron activation and delay gammas. To achieve the aim, the ITER system and its components were simulated by Monte Carlo method. Also to increase the accuracy and precision of the absorbed dose assessment a body phantom were considered in the simulation. The results of this research showed that total dose rates level near the outside of bio-shield wall of the tokamak hall is less than ten percent of the annual occupational dose limits during normal operation of ITER and It is possible to learn how long human beings can remain in that environment before the body absorbs dangerous levels of radiation. (authors)

Randomly dispersed particle fuel model in the PSG Monte Carlo neutron transport code

International Nuclear Information System (INIS)

Leppaenen, J.

2007-01-01

High-temperature gas-cooled reactor fuels are composed of thousands of microscopic fuel particles, randomly dispersed in a graphite matrix. The modelling of such geometry is complicated, especially using continuous-energy Monte Carlo codes, which are unable to apply any deterministic corrections in the calculation. This paper presents the geometry routine developed for modelling randomly dispersed particle fuels using the PSG Monte Carlo reactor physics code. The model is based on the delta-tracking method, and it takes into account the spatial self-shielding effects and the random dispersion of the fuel particles. The calculation routine is validated by comparing the results to reference MCNP4C calculations using uranium and plutonium based fuels. (authors)

Monte Carlo determination of heteroepitaxial misfit structures

DEFF Research Database (Denmark)

Baker, J.; Lindgård, Per-Anker

1996-01-01

We use Monte Carlo simulations to determine the structure of KBr overlayers on a NaCl(001) substrate, a system with large (17%) heteroepitaxial misfit. The equilibrium relaxation structure is determined for films of 2-6 ML, for which extensive helium-atom scattering data exist for comparison...

Situations of potential exposure in self-shielding electron accelerators

International Nuclear Information System (INIS)

Rios, D.A.S.; Rios, P.B.; Sordi, G.M.A.A.; Carneiro, J.C.G.G.

2017-01-01

The study discusses situations in the industrial environment that may lead to potential exposure of Occupationally Exposed Individuals and Public Individuals in self-shielding electron accelerators. Although these exposure situations are unlikely, simulation exercises can lead to improvements in the operating procedure as well as suggest changes in production line design in order to increase radiation protection at work. These studies can also be used in training and demonstrate a solid application of the ALARA principle in the daily activities of radiative installations

Shielding design method for LMFBR validation on the Phenix factor

International Nuclear Information System (INIS)

Cabrillat, J.C.; Crouzet, J.; Misrakis, J.; Salvatores, M.; Rado, V.; Palmiotti, G.

1983-05-01

Shielding design methods, developed at CEA for shielding calculations find a global validation by the means of Phenix power reactor (250 MWe) measurements. Particularly, the secondary sodium activation of pool type LMFBR such as Super Phenix (1200 MWe) which is subject to strict safety limitation is well calculated by the adapted scheme, i.e. a two dimension transport calculation of shielding coupled to a Monte-Carlo calculation of secondary sodium activation

CO Self-Shielding as a Mechanism to Make 16O-Enriched Solids in the Solar Nebula

Directory of Open Access Journals (Sweden)

Joseph A. Nuth, III

2014-05-01

Full Text Available Photochemical self-shielding of CO has been proposed as a mechanism to produce solids observed in the modern, 16O-depleted solar system. This is distinct from the relatively 16O-enriched composition of the solar nebula, as demonstrated by the oxygen isotopic composition of the contemporary sun. While supporting the idea that self-shielding can produce local enhancements in 16O-depleted solids, we argue that complementary enhancements of 16O-enriched solids can also be produced via C16O-based, Fischer-Tropsch type (FTT catalytic processes that could produce much of the carbonaceous feedstock incorporated into accreting planetesimals. Local enhancements could explain observed 16O enrichment in calcium-aluminum-rich inclusions (CAIs, such as those from the meteorite, Isheyevo (CH/CHb, as well as in chondrules from the meteorite, Acfer 214 (CH3. CO self-shielding results in an overall increase in the 17O and 18O content of nebular solids only to the extent that there is a net loss of C16O from the solar nebula. In contrast, if C16O reacts in the nebula to produce organics and water then the net effect of the self-shielding process will be negligible for the average oxygen isotopic content of nebular solids and other mechanisms must be sought to produce the observed dichotomy between oxygen in the Sun and that in meteorites and the terrestrial planets. This illustrates that the formation and metamorphism of rocks and organics need to be considered in tandem rather than as isolated reaction networks.

Optimization of a partially non-magnetic primary radiation shielding for the triple-axis spectrometer PANDA at the Munich high-flux reactor FRM-II

CERN Document Server

Pyka, N M; Rogov, A

2002-01-01

Monte Carlo simulations have been used to optimize the monochromator shielding of the polarized cold-neutron triple-axis spectrometer PANDA at the Munich high-flux reactor FRM-II. By using the Monte Carlo program MCNP-4B, the density of the total spectrum of incoming neutrons and gamma radiation from the beam tube SR-2 has been determined during the three-dimensional diffusion process in different types of heavy concrete and other absorbing material. Special attention has been paid to build a compact and highly efficient shielding, partially non-magnetic, with a total biological radiation dose of less than 10 mu Sv/h at its outsides. Especially considered was the construction of an albedo reducer, which serves to reduce the background in the experiment outside the shielding. (orig.)

Electron accelerator shielding design of KIPT neutron source facility

Energy Technology Data Exchange (ETDEWEB)

Zhong, Zhao Peng; Gohar, Yousry [Argonne National Laboratory, Argonne (United States)

2016-06-15

The Argonne National Laboratory of the United States and the Kharkov Institute of Physics and Technology of the Ukraine have been collaborating on the design, development and construction of a neutron source facility at Kharkov Institute of Physics and Technology utilizing an electron-accelerator-driven subcritical assembly. The electron beam power is 100 kW using 100-MeV electrons. The facility was designed to perform basic and applied nuclear research, produce medical isotopes, and train nuclear specialists. The biological shield of the accelerator building was designed to reduce the biological dose to less than 5.0e-03 mSv/h during operation. The main source of the biological dose for the accelerator building is the photons and neutrons generated from different interactions of leaked electrons from the electron gun and the accelerator sections with the surrounding components and materials. The Monte Carlo N-particle extended code (MCNPX) was used for the shielding calculations because of its capability to perform electron-, photon-, and neutron-coupled transport simulations. The photon dose was tallied using the MCNPX calculation, starting with the leaked electrons. However, it is difficult to accurately tally the neutron dose directly from the leaked electrons. The neutron yield per electron from the interactions with the surrounding components is very small, ∼0.01 neutron for 100-MeV electron and even smaller for lower-energy electrons. This causes difficulties for the Monte Carlo analyses and consumes tremendous computation resources for tallying the neutron dose outside the shield boundary with an acceptable accuracy. To avoid these difficulties, the SOURCE and TALLYX user subroutines of MCNPX were utilized for this study. The generated neutrons were banked, together with all related parameters, for a subsequent MCNPX calculation to obtain the neutron dose. The weight windows variance reduction technique was also utilized for both neutron and photon dose

Treatment vault shielding for a flattening filter-free medical linear accelerator

Science.gov (United States)

Kry, Stephen F.; Howell, Rebecca M.; Polf, Jerimy; Mohan, Radhe; Vassiliev, Oleg N.

2009-03-01

The requirements for shielding a treatment vault with a Varian Clinac 2100 medical linear accelerator operated both with and without the flattening filter were assessed. Basic shielding parameters, such as primary beam tenth-value layers (TVLs), patient scatter fractions, and wall scatter fractions, were calculated using Monte Carlo simulations of 6, 10 and 18 MV beams. Relative integral target current requirements were determined from treatment planning studies of several disease sites with, and without, the flattening filter. The flattened beam shielding data were compared to data published in NCRP Report No. 151, and the unflattened beam shielding data were presented relative to the NCRP data. Finally, the shielding requirements for a typical treatment vault were determined for a single-energy (6 MV) linac and a dual-energy (6 MV/18 MV) linac. With the exception of large-angle patient scatter fractions and wall scatter fractions, the vault shielding parameters were reduced when the flattening filter was removed. Much of this reduction was consistent with the reduced average energy of the FFF beams. Primary beam TVLs were reduced by 12%, on average, and small-angle scatter fractions were reduced by up to 30%. Head leakage was markedly reduced because less integral target current was required to deliver the target dose. For the treatment vault examined in the current study, removal of the flattening filter reduced the required thickness of the primary and secondary barriers by 10-20%, corresponding to 18 m3 less concrete to shield the single-energy linac and 36 m3 less concrete to shield the dual-energy linac. Thus, a shielding advantage was found when the linac was operated without the flattening filter. This translates into a reduction in occupational exposure and/or the cost and space of shielding.

Treatment vault shielding for a flattening filter-free medical linear accelerator

Energy Technology Data Exchange (ETDEWEB)

Kry, Stephen F; Howell, Rebecca M; Polf, Jerimy; Mohan, Radhe; Vassiliev, Oleg N [Department of Radiation Physics, University of Texas M. D. Anderson Cancer Center, Houston, TX (United States)], E-mail: sfkry@mdanderson.org

2009-03-07

The requirements for shielding a treatment vault with a Varian Clinac 2100 medical linear accelerator operated both with and without the flattening filter were assessed. Basic shielding parameters, such as primary beam tenth-value layers (TVLs), patient scatter fractions, and wall scatter fractions, were calculated using Monte Carlo simulations of 6, 10 and 18 MV beams. Relative integral target current requirements were determined from treatment planning studies of several disease sites with, and without, the flattening filter. The flattened beam shielding data were compared to data published in NCRP Report No. 151, and the unflattened beam shielding data were presented relative to the NCRP data. Finally, the shielding requirements for a typical treatment vault were determined for a single-energy (6 MV) linac and a dual-energy (6 MV/18 MV) linac. With the exception of large-angle patient scatter fractions and wall scatter fractions, the vault shielding parameters were reduced when the flattening filter was removed. Much of this reduction was consistent with the reduced average energy of the FFF beams. Primary beam TVLs were reduced by 12%, on average, and small-angle scatter fractions were reduced by up to 30%. Head leakage was markedly reduced because less integral target current was required to deliver the target dose. For the treatment vault examined in the current study, removal of the flattening filter reduced the required thickness of the primary and secondary barriers by 10-20%, corresponding to 18 m{sup 3} less concrete to shield the single-energy linac and 36 m{sup 3} less concrete to shield the dual-energy linac. Thus, a shielding advantage was found when the linac was operated without the flattening filter. This translates into a reduction in occupational exposure and/or the cost and space of shielding.

Treatment vault shielding for a flattening filter-free medical linear accelerator

International Nuclear Information System (INIS)

Kry, Stephen F; Howell, Rebecca M; Polf, Jerimy; Mohan, Radhe; Vassiliev, Oleg N

2009-01-01

The requirements for shielding a treatment vault with a Varian Clinac 2100 medical linear accelerator operated both with and without the flattening filter were assessed. Basic shielding parameters, such as primary beam tenth-value layers (TVLs), patient scatter fractions, and wall scatter fractions, were calculated using Monte Carlo simulations of 6, 10 and 18 MV beams. Relative integral target current requirements were determined from treatment planning studies of several disease sites with, and without, the flattening filter. The flattened beam shielding data were compared to data published in NCRP Report No. 151, and the unflattened beam shielding data were presented relative to the NCRP data. Finally, the shielding requirements for a typical treatment vault were determined for a single-energy (6 MV) linac and a dual-energy (6 MV/18 MV) linac. With the exception of large-angle patient scatter fractions and wall scatter fractions, the vault shielding parameters were reduced when the flattening filter was removed. Much of this reduction was consistent with the reduced average energy of the FFF beams. Primary beam TVLs were reduced by 12%, on average, and small-angle scatter fractions were reduced by up to 30%. Head leakage was markedly reduced because less integral target current was required to deliver the target dose. For the treatment vault examined in the current study, removal of the flattening filter reduced the required thickness of the primary and secondary barriers by 10-20%, corresponding to 18 m 3 less concrete to shield the single-energy linac and 36 m 3 less concrete to shield the dual-energy linac. Thus, a shielding advantage was found when the linac was operated without the flattening filter. This translates into a reduction in occupational exposure and/or the cost and space of shielding.

Advances in the development of a subgroup method for the self-shielding of resonant isotopes in arbitrary geometries

International Nuclear Information System (INIS)

Hebert, A.

1997-01-01

The subgroup method is used to compute self-shielded cross sections defined over coarse energy groups in the resolved energy domain. The validity of the subgroup approach was extended beyond the unresolved energy domain by partially taking into account correlation effects between the slowing-down source with the collision probability terms of the transport equation. This approach enables one to obtain a pure subgroup solution of the self-shielding problem without relying on any form of equivalence in dilution. Specific improvements are presented on existing subgroup methods: an N-term rational approximation for the fuel-to-fuel collision probability, a new Pade deflation technique for computing probability tables, and the introduction of a superhomogenization correction. The absorption rates obtained after self-shielding are compared with exact values obtained using an elastic slowing-down calculation where each resonance is modeled individually in the resolved energy domain

Neutron shielding verification measurements and simulations for a 235-MeV proton therapy center

International Nuclear Information System (INIS)

Newhauser, W.D.; Titt, U.; Dexheimer, D.; Yan, X.; Nill, S.

2002-01-01

The neutron shielding at the Massachusetts General Hospital's 235-MeV proton therapy facility was investigated with measurements, analytical calculations, and realistic three-dimensional Monte Carlo simulations. In 37 of 40 cases studied, the analytical calculations predicted higher neutron dose equivalent rates outside the shielding than the measured, typically by more than a factor of 10, and in some cases more than 100. Monte Carlo predictions of dose equivalent at three locations are, on average, 1.1 times the measured values. Except at one location, all of the analytical model predictions and Monte Carlo simulations overestimate neutron dose equivalent

JMCT Monte Carlo simulation analysis of full core PWR Pin-By-Pin and shielding

International Nuclear Information System (INIS)

Deng, L.; Li, G.; Zhang, B.; Shangguan, D.; Ma, Y.; Hu, Z.; Fu, Y.; Li, R.; Hu, X.; Cheng, T.; Shi, D.

2015-01-01

This paper describes the application of the JMCT Monte Carlo code to the simulation of Kord Smith Challenge H-M model, BEAVRS model and Chinese SG-III model. For H-M model, the 6.3624 millions tally regions and the 98.3 billion neutron histories do. The detailed pin flux and energy deposition densities obtain. 95% regions have less 1% standard deviation. For BEAVRS model, firstly, we performed the neutron transport calculation of 398 axial planes in the Hot Zero Power (HZP) status. Almost the same results with MC21 and OpenMC results are achieved. The detailed pin-power density distribution and standard deviation are shown. Then, we performed the calculation of ten depletion steps in 30 axial plane cases. The depletion regions exceed 1.5 million and 12,000 processors uses. Finally, the Chinese SG-III laser model is simulated. The neutron and photon flux distributions are given, respectively. The results show that the JMCT code well suits for extremely large reactor and shielding simulation. (author)

Shielding analysis method applied to nuclear ship 'MUTSU' and its evaluation based on experimental analyses

International Nuclear Information System (INIS)

Yamaji, Akio; Miyakoshi, Jun-ichi; Iwao, Yoshiaki; Tsubosaka, Akira; Saito, Tetsuo; Fujii, Takayoshi; Okumura, Yoshihiro; Suzuoki, Zenro; Kawakita, Takashi.

1984-01-01

Procedures of shielding analysis are described which were used for the shielding modification design of the Nuclear Ship ''MUTSU''. The calculations of the radiation distribution on board were made using Sn codes ANISN and TWOTRAN, a point kernel code QAD and a Monte Carlo code MORSE. The accuracies of these calculations were investigated through the analysis of various shielding experiments: the shield tank experiment of the Nuclear Ship ''Otto Hahn'', the shielding mock-up experiment for ''MUTSU'' performed in JRR-4, the shielding benchmark experiment using the 16 N radiation facility of AERE Harwell and the shielding effect experiment of the ship structure performed in the training ship ''Shintoku-Maru''. The values calculated by the ANISN agree with the data measured at ''Otto Hahn'' within a factor of 2 for fast neutrons and within a factor of 3 for epithermal and thermal neutrons. The γ-ray dose rates calculated by the QAD agree with the measured values within 30% for the analysis of the experiment in JRR-4. The design values for ''MUTSU'' were determined in consequence of these experimental analyses. (author)

3D Space Radiation Transport in a Shielded ICRU Tissue Sphere

Science.gov (United States)

Wilson, John W.; Slaba, Tony C.; Badavi, Francis F.; Reddell, Brandon D.; Bahadori, Amir A.

2014-01-01

A computationally efficient 3DHZETRN code capable of simulating High Charge (Z) and Energy (HZE) and light ions (including neutrons) under space-like boundary conditions with enhanced neutron and light ion propagation was recently developed for a simple homogeneous shield object. Monte Carlo benchmarks were used to verify the methodology in slab and spherical geometry, and the 3D corrections were shown to provide significant improvement over the straight-ahead approximation in some cases. In the present report, the new algorithms with well-defined convergence criteria are extended to inhomogeneous media within a shielded tissue slab and a shielded tissue sphere and tested against Monte Carlo simulation to verify the solution methods. The 3D corrections are again found to more accurately describe the neutron and light ion fluence spectra as compared to the straight-ahead approximation. These computationally efficient methods provide a basis for software capable of space shield analysis and optimization.

Laboratory tests on neutron shields for gamma-ray detectors in space

CERN Document Server

Hong, J; Hailey, C J

2000-01-01

Shields capable of suppressing neutron-induced background in new classes of gamma-ray detectors such as CdZnTe are becoming important for a variety of reasons. These include a high cross section for neutron interactions in new classes of detector materials as well as the inefficient vetoing of neutron-induced background in conventional active shields. We have previously demonstrated through Monte-Carlo simulations how our new approach, supershields, is superior to the monolithic, bi-atomic neutron shields which have been developed in the past. We report here on the first prototype models for supershields based on boron and hydrogen. We verify the performance of these supershields through laboratory experiments. These experimental results, as well as measurements of conventional monolithic neutron shields, are shown to be consistent with Monte-Carlo simulations. We discuss the implications of this experiment for designs of supershields in general and their application to future hard X-ray/gamma-ray experiments...

Automated-biasing approach to Monte Carlo shipping-cask calculations

International Nuclear Information System (INIS)

Hoffman, T.J.; Tang, J.S.; Parks, C.V.; Childs, R.L.

1982-01-01

Computer Sciences at Oak Ridge National Laboratory, under a contract with the Nuclear Regulatory Commission, has developed the SCALE system for performing standardized criticality, shielding, and heat transfer analyses of nuclear systems. During the early phase of shielding development in SCALE, it was established that Monte Carlo calculations of radiation levels exterior to a spent fuel shipping cask would be extremely expensive. This cost can be substantially reduced by proper biasing of the Monte Carlo histories. The purpose of this study is to develop and test an automated biasing procedure for the MORSE-SGC/S module of the SCALE system

Determination of shielding factors for typical buildings in Brazil

International Nuclear Information System (INIS)

Salinas, Isabel Cristina Poquet

2006-10-01

This study presents a methodology for the determination of the air kerma inside buildings due to contamination on the external surfaces and the shielding factors for the construction material to be used on emergency assessment systems for urban areas. The commonly used construction materials were simulated with the MCNP computer code. A special methodology to simulate the bricks with holes were developed, mixing all different regions into a single one, making the simulation easier and faster. The effective density and the attenuation coefficients for the 50-3000 keV energy range were determined. The effective protection for the bricks with no cement cover decreases by 40-50% for energies greater then 300 keV when compared to bricks covered on both sides. With the data made available it was possible to evaluate the influence of the construction materials densities and thickness on the exposure due to external surfaces contamination and to estimate the error on the dose when the shielding factor applied on the calculation differs from the more realistic ones. The shielding factors for three types of walls were determined for a five rooms house. Special protection procedures should be applied for houses built with bricks with no cement cover, because they are due to the double of the dose when compared to houses built with bricks two-sided cement covered. The influence of windows and doors were evaluated too. This work was developed at the IRD in order to provide information on the construction material commonly uses on building in Brazil. (author)

Shielding for neutrons produced by medical linear accelerators

International Nuclear Information System (INIS)

Rebello, Wilson F.; Silva, Ademir X.

2007-01-01

The shielding system called Multileaf Shielding (MLS) was designed in Brazil to be used for protection patients, who undergo radiotherapy treatment, against undesired neutrons produced in the medical linear accelerator heads. During the conceiving of the MLS it was necessary to evaluate its efficiency. For that purpose, several simulations using the Monte Carlo N-particle radiation transport code, MCNP5, were made, in order to evaluate the response of the new shielding system. The results showed a significant neutron dose reduction after the inclusion of the MLS. This work aims to presenting these simulation results. (author)

Evaluation of usability of the shielding effect for thyroid shield for peripheral dose during whole brain radiation therapy

Energy Technology Data Exchange (ETDEWEB)

Yang, Myung Sic; Park, Ju Kyeong; Lee, Seung Hun; Kim, Yang Su; Lee, Sun Young; Cha, Seok Yong [Dept. of Radiation Oncology, Chonbuk National University Hospital, Jeonju (Korea, Republic of)

2014-12-15

To reduce the radiation dose to the thyroid that is affected to scattered radiation, the shield was used. And we evaluated the shielding effect for the thyroid during whole brain radiation therapy. To measure the dose of the thyroid, 300cGy were delivered to the phantom using a linear accelerator(Clinac iX VARIAN, USA.)in the way of the 6MV X-ray in bilateral. To measure the entrance surface dose of the thyroid, five glass dosimeters were placed in the 10th slice's surface of the phantom with a 1.5 cm interval. The average values were calculated by measured values in five times each, using bismuth shield, 0.5 mmPb shield, self-made 1.0 mmPb shield and unshield. In the same location, to measure the depth dose of the thyroid, five glass dosimeters were placed in the 10th slice by 2.5 cm depth of the phantom with a 1.5 cm interval. The average values were calculated by measured values in five times each, using bismuth shield, 0.5 mmPb shield, self-made 1.0 mmPb shield and unshield. Entrance surface dose of the thyroid were respectively 44.89 mGy at the unshield, 36.03 mGy at the bismuth shield, 31.03 mGy at the 0.5 mmPb shield and 23.21 mGy at a self-made 1.0 mmPb shield. In addition, the depth dose of the thyroid were respectively 36.10 mGy at the unshield, 34.52 mGy at the bismuth shield, 32.28 mGy at the 0.5 mmPb shield and 25.50 mGy at a self-made 1.0 mmPb shield. The thyroid was affected by the secondary scattering dose and leakage dose outside of the radiation field during whole brain radiation therapy. When using a shield in the thyroid, the depth dose of thyroid showed 11-30% reduction effect and the surface dose of thyroid showed 20-48% reduction effect. Therefore, by using the thyroid shield, it is considered to effectively protect the thyroid and can perform the treatment.

Basic design of shield blocks for a spallation neutron source under the high-intensity proton accelerator project

Energy Technology Data Exchange (ETDEWEB)

Yoshida, Katsuhiko; Maekawa, Fujio; Takada, Hiroshi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

2003-03-01

Under the JAERI-KEK High-Intensity Proton Accelerator Project (J-PARC), a spallation neutron source driven by a 3 GeV-1 MW proton beam is planed to be constructed as a main part of the Materials and Life Science Facility. Overall dimensions of a biological shield of the neutron source had been determined by evaluation of shielding performance by Monte Carlo calculations. This report describes results of design studies on an optimum dividing scheme in terms of cost and treatment and mechanical strength of shield blocks for the biological shield. As for mechanical strength, it was studied whether the shield blocks would be stable, fall down or move to a horizontal direction in case of an earthquake of seismic intensity of 5.5 (250 Gal) as an abnormal load. For ceiling shielding blocks being supported by both ends of the long blocks, maximum bending moment and an amount of maximum deflection of their center were evaluated. (author)

Basic design of shield blocks for a spallation neutron source under the high-intensity proton accelerator project

CERN Document Server

Yoshida, K; Takada, H

2003-01-01

Under the JAERI-KEK High-Intensity Proton Accelerator Project (J-PARC), a spallation neutron source driven by a 3 GeV-1 MW proton beam is planed to be constructed as a main part of the Materials and Life Science Facility. Overall dimensions of a biological shield of the neutron source had been determined by evaluation of shielding performance by Monte Carlo calculations. This report describes results of design studies on an optimum dividing scheme in terms of cost and treatment and mechanical strength of shield blocks for the biological shield. As for mechanical strength, it was studied whether the shield blocks would be stable, fall down or move to a horizontal direction in case of an earthquake of seismic intensity of 5.5 (250 Gal) as an abnormal load. For ceiling shielding blocks being supported by both ends of the long blocks, maximum bending moment and an amount of maximum deflection of their center were evaluated.

Improvement of top shield analysis technology for CANDU 6 reactor

International Nuclear Information System (INIS)

Kim, Kyo Yoon; Jin, Young Kwon; Lee, Sung Hee; Moon, Bok Ja; Kim, Yong Il

1996-07-01

As for Wolsung NPP unit 1, radiation shielding analysis was performed by using neutron diffusion codes, one-dimensional discrete ordinates code ANISN, and analytical methods. But for Wolsung NPP unit 2, 3, and 4, two-dimensional discrete ordinates code DOT substituted for neutron diffusion codes. In other words, the method of analysis and computer codes used for radiation shielding of CANDU 6 type reactor have been improved. Recently Monte Carlo MCNP code has been widely utilized in the field of radiation physics and other radiation related areas because it can describe an object sophisticately by use of three-dimensional modelling and can adopt continuous energy cross-section library. Nowadays Monte Carlo method has been reported to be competitive to discrete ordinate method in the field of radiation shielding and the former has been known to be superior to the latter for complex geometry problem. However, Monte Carlo method had not been used for radiation streaming calculation in the shielding design of CANDU type reactor. Neutron and gamma radiations are expected to be streamed from calandria through the penetrations to reactivity mechanism deck (R/M deck) because many reactivity control units which are established on R/M deck extend from R/M deck to calandria within penetrations, which are provided by guide tube extensions. More precise estimation of radiation streaming is required because R/M deck is classified as an accessible area where atomic worker can access when necessary. Therefore neutron and gamma dose rates were estimated using MCNP code on the R/M deck in the top shield system of CANDU 6 reactor. 9 tabs., 17 figs., 21 refs. (Author)

Self-shielding flex-circuit drift tube, drift tube assembly and method of making

Science.gov (United States)

Jones, David Alexander

2016-04-26

The present disclosure is directed to an ion mobility drift tube fabricated using flex-circuit technology in which every other drift electrode is on a different layer of the flex-circuit and each drift electrode partially overlaps the adjacent electrodes on the other layer. This results in a self-shielding effect where the drift electrodes themselves shield the interior of the drift tube from unwanted electro-magnetic noise. In addition, this drift tube can be manufactured with an integral flex-heater for temperature control. This design will significantly improve the noise immunity, size, weight, and power requirements of hand-held ion mobility systems such as those used for explosive detection.

Monte Carlo simulation of calibration of shadow shield scanning bed whole body monitor using different size BOMAB phantoms

International Nuclear Information System (INIS)

Bhati, S.; Patni, H.K.; Singh, I.S.; Garg, S.P.

2005-01-01

A shadow shield scanning bed whole body monitor incorporating a (102 mm dia x 76 mm thick) NaI(Tl) detector, is employed for assessment of high-energy photon emitters at BARC. The monitor is calibrated using a Reference BOMAB phantom representative of an average Indian radiation worker. However to account for the size variation in the physique of workers, it is required to calibrate the system with different size BOMAB phantoms which is both difficult and expensive. Therefore, a theoretical approach based on Monte Carlo techniques has been employed to calibrate the system with BOMAB phantoms of different sizes for several radionuclides of interest. A computer program developed for this purpose, simulates the scanning geometry of the whole body monitor and computes detection efficiencies for the BARC Reference phantom (63 kg/168 cm), ICRP Reference phantom (70 kg/170 cm) and several of its scaled versions covering a wide range of body builds. The detection efficiencies computed for different photon energies for BARC Reference phantom were found to be in very good agreement with experimental data, thus validating the Monte Carlo scheme used in the computer code. The results from this study could be used for assessment of internal contamination due to high-energy photon emitters for radiation workers of different physiques. (author)

Shielding calculations using FLUKA

International Nuclear Information System (INIS)

Yamaguchi, Chiri; Tesch, K.; Dinter, H.

1988-06-01

The dose equivalent on the surface of concrete shielding has been calculated using the Monte Carlo code FLUKA86 for incident proton energies from 10 to 800 GeV. The results have been compared with some simple equations. The value of the angular dependent parameter in Moyer's equation has been calculated from the locations where the values of the maximum dose equivalent occur. (author)

ICRS1, Proceedings of the First Radiation Shielding Symposium, Cambridge, UK 1958

International Nuclear Information System (INIS)

Goebelbecker, Hans-Juergen

2008-01-01

Description: The papers of the European Atomic Energy Society Symposium VI-58 on radiation shielding (ICRS1) held at Caius College, Cambridge England from 26 to 29 August 1958 are collected here for the first time in electronic form. This symposium was organised in connection with the Second Atoms for Peace Conference held in Geneva Held in Geneva from 1 to 13 September 1958. The Topics discussed covered gamma rays and neutron radiation; the Methods discussed were analytical approaches, semi-empirical Methods, simple computer codes, Monte Carlo method. Little quality nuclear data for shielding calculations was available and the presentations would concentrate on removal cross-sections and build-up factors. Experimental techniques in support to estimate the effective shielding properties of materials were discussed such as general experimental shielding techniques and experiments on neutron attenuation in different materials and on concrete as shield. Foil detectors for spectra measurements and determination of dose rates were mainly used. The typical issues addressed were gamma-heating, gamma spectra, neutron induced gammas, fission products gamma spectra, skyshine radiation and neutron ducts - streaming. Most participants were researchers from the naval and aeronautics sector

Calculation of self–shielding factor for neutron activation experiments using GEANT4 and MCNP

Energy Technology Data Exchange (ETDEWEB)

Romero–Barrientos, Jaime, E-mail: jaromero@ing.uchile.cl [Comisión Chilena de Energía Nuclear, Nueva Bilbao 12501, Las Condes, Santiago (Chile); Universidad de Chile, DFI, Facultad de Ciencias Físicas Y Matemáticas, Avenida Blanco Encalada 2008, Santiago (Chile); Molina, F. [Comisión Chilena de Energía Nuclear, Nueva Bilbao 12501, Las Condes, Santiago (Chile); Aguilera, Pablo [Comisión Chilena de Energía Nuclear, Nueva Bilbao 12501, Las Condes, Santiago (Chile); Universidad de Chile, Depto. de Física, Facultad de Ciencias, Las Palmeras 3425, Ñuñoa, Santiago (Chile); Arellano, H. F. [Universidad de Chile, DFI, Facultad de Ciencias Físicas Y Matemáticas, Avenida Blanco Encalada 2008, Santiago (Chile)

2016-07-07

The neutron self–shielding factor G as a function of the neutron energy was obtained for 14 pure metallic samples in 1000 isolethargic energy bins from 1·10{sup −5}eV to 2·10{sup 7}eV using Monte Carlo simulations in GEANT4 and MCNP6. The comparison of these two Monte Carlo codes shows small differences in the final self–shielding factor mostly due to the different cross section databases that each program uses.

Evaluation of the shield calculation adequacy of radiotherapy rooms through Monte Carlo Method and experimental measures; Avaliacao da adequacao do calculo de blindagens de salas de radioterapia atraves do metodo de Monte Carlos e medidas experimentais

Energy Technology Data Exchange (ETDEWEB)

Meireles, Ramiro Conceicao

2016-07-01

The shielding calculation methodology for radiotherapy services adopted in Brazil and in several countries is that described in publication 151 of the National Council on Radiation Protection and Measurements (NCRP 151). This methodology however, markedly employs several approaches that can impact both in the construction cost and in the radiological safety of the facility. Although this methodology is currently well established by the high level of use, some parameters employed in the calculation methodology did not undergo to a detailed assessment to evaluate the impact of the various approaches considered. In this work the MCNP5 Monte Carlo code was used with the purpose of evaluating the above mentioned approaches. TVLs values were obtained for photons in conventional concrete (2.35g / cm{sup 3}), considering the energies of 6, 10 and 25 MeV, respectively, first considering an isotropic radiation source impinging perpendicular to the barriers, and subsequently a lead head shielding emitting a shaped beam, in the format of a pyramid trunk. Primary barriers safety margins, taking in account the head shielding emitting photon beam pyramid-shaped in the energies of 6, 10, 15 and 18 MeV were assessed. A study was conducted considering the attenuation provided by the patient's body in the energies of 6,10, 15 and 18 MeV, leading to new attenuation factors. Experimental measurements were performed in a real radiotherapy room, in order to map the leakage radiation emitted by the accelerator head shielding and the results obtained were employed in the Monte Carlo simulation, as well as to validate the entire study. The study results indicate that the TVLs values provided by (NCRP, 2005) show discrepancies in comparison with the values obtained by simulation and that there may be some barriers that are calculated with insufficient thickness. Furthermore, the simulation results show that the additional safety margins considered when calculating the width of the

Automatic generation of biasing parameters for MORSE shielding problems

International Nuclear Information System (INIS)

Hoogenboom, J.E.

1995-01-01

It would be favourable if the biasing functions could be obtained from the Monte Carlo calculation itself. This is discussed in this paper as well as the way to obtain biasing parameters from it for splitting, Russian roulette and path length stretching. The method is demonstrated for a shielding problem solved with the MORSE-SGC/S Monte Carlo code of the SCALE-system. (K.A.)

Elastic removal self-shielding factors for light and medium nuclides with strong-resonance scattering

International Nuclear Information System (INIS)

Nakagawa, Masayuki; Ishiguro, Yukio; Tokuno, Yukio.

1978-01-01

The self-shielding factors for elastic removal cross sections of light and medium weight nuclides were calculated for the parameter, σ 0 within the conventional concept of the group constant sets. The numerical study were performed for obtaining a simple and accurate method. The present results were compared with the exact values and the conventional ones, and shown to be remarkably improved. It became apparent that the anisotropy of the elastic scattering did not affect to the self-shielding factors though it did to the infinite dilution cross sections. With use of the present revised set, the neutron flux were calculated in an iron medium and in a prototype FBR and compared with those by the fine spectrum calculations and the conventional set. The present set showed the considerable improvement in the vicinity of the large resonance regions of sodium, iron and oxygen. (auth.)

Monte Carlo analysis of the effects of a blanket-shield penetration on the performance of a tokamak fusion reactor

International Nuclear Information System (INIS)

Santoro, R.T.; Tang, J.S.; Alsmiller, R.G. Jr.; Barnes, J.M.

1977-05-01

Adjoint Monte Carlo calculations have been carried out using the three-dimensional radiation transport code, MORSE, to estimate the nuclear heating and radiation damage in the toroidal field (TF) coils adjacent to a 28 x 68 cm 2 rectangular neutral beam injector duct that passes through the blanket and shield of a D-T burning Tokamak reactor. The plasma region, blanket, shield, and TF coils were represented in cylindrical geometry using the same dimensions and compositions as those of the Experimental Power Reactor. The radiation transport was accomplished using coupled 35-group neutron, 21-group gamma-ray cross sections obtained by collapsing the DLC-37 cross-section library. Nuclear heating and radiation damage rates were estimated using the latest available nuclear response functions. The presence of the neutral beam injector duct leads to increases in the nuclear heating rates in the TF coils ranging from a factor of 3 to a factor of 196 depending on the location. Increases in the radiation damage also result in the TF coils. The atomic displacement rates increase from factors of 2 to 138 and the hydrogen and helium gas production rates increase from factors of 11 to 7600 and from 15 to 9700, respectively

The computer code system for reactor radiation shielding in design of nuclear power plant

International Nuclear Information System (INIS)

Li Chunhuai; Fu Shouxin; Liu Guilian

1995-01-01

The computer code system used in reactor radiation shielding design of nuclear power plant includes the source term codes, discrete ordinate transport codes, Monte Carlo and Albedo Monte Carlo codes, kernel integration codes, optimization code, temperature field code, skyshine code, coupling calculation codes and some processing codes for data libraries. This computer code system has more satisfactory variety of codes and complete sets of data library. It is widely used in reactor radiation shielding design and safety analysis of nuclear power plant and other nuclear facilities

Neutron shieldings

International Nuclear Information System (INIS)

Tarutani, Kohei

1979-01-01

Purpose: To decrease the stresses resulted by the core bendings to the base of an entrance nozzle. Constitution: Three types of round shielding rods of different diameter are arranged in a hexagonal tube. The hexagonal tube is provided with several spacer pads receiving the loads from the core constrain mechanism at its outer circumference, a handling head for a fuel exchanger at its top and an entrance nozzle for self-holding the neutron shieldings and flowing heat-removing coolants at its bottom. The diameters for R 1 , R 2 and R 3 for the round shielding rods are designed as: 0.1 R 1 2 1 and 0.2 R 1 2 1 . Since a plurality of shielding rods of small diameter are provided, soft structure are obtained and a plurality of coolant paths are formed. (Furukawa, Y.)

GROUPIE2007, Bondarenko Self-Shielded Cross sections from ENDF/B

International Nuclear Information System (INIS)

2007-01-01

1 - Description of problem or function - GROUPIE reads evaluated data in ENDF/B Format and uses these to calculate unshielded group averaged Cross sections, Bondarenko self-shielded Cross sections, and multiband parameters. The program allows the user to specify arbitrary energy groups and an arbitrary energy-dependent neutron spectrum (weighting function). IAEA0849/15: This version include the updates up to January 30, 2007. Changes in ENDF/B-VII Format and procedures, as well as the evaluations themselves, make it impossible for versions of the ENDF/B pre-processing codes earlier than PREPRO 2007 (2007 Version) to accurately process current ENDF/B-VII evaluations. The present code can handle all existing ENDF/B-VI evaluations through release 8, which will be the last release of ENDF/B-VI. 2 - Modifications from previous versions: Groupie VERS. 2007-1 (Jan. 2007): checked against all ENDF/B-VII; increased page size from 120,000 to 600,000 points. 3 - Method of solution: All integrals are performed analytically; in no case is iteration or any approximate form of integration used. GROUPIE reads either the 0 deg. Kelvin Cross sections or the Doppler broadened Cross sections to calculate the self-shielded Cross sections and multiband parameters for 25 values of the 'background' Cross sections (representing the combined effects of all other isotopes and of leakage). 4 - Restrictions on the complexity of the problem: GROUPIE requires that the energy-dependent neutron spectrum and all Cross sections be given in tabular form, with linear interpolation between tabulated values. There is no limit to the size of the table used to describe the spectrum, so the spectrum may be described in as much detail as required. - If only unshielded averages are calculated, the program can handle up to 3000 groups. If self-shielded averages and/or multiband parameters are calculated, the program can handle up to 175 groups. These limits can easily be extended. - The program only uses the

Spiral MRI on a 9.4T Vertical-bore Superconducting Magnet Using Unshielded and Self-shielded Gradient Coils.

Science.gov (United States)

Kodama, Nao; Setoi, Ayana; Kose, Katsumi

2018-04-10

Spiral MRI sequences were developed for a 9.4T vertical standard bore (54 mm) superconducting magnet using unshielded and self-shielded gradient coils. Clear spiral images with 64-shot scan were obtained with the self-shielded gradient coil, but severe shading artifacts were observed for the spiral-scan images acquired with the unshielded gradient coil. This shading artifact was successfully corrected with a phase-correction technique using reference scans that we developed based on eddy current field measurements. We therefore concluded that spiral imaging sequences can be installed even for unshielded gradient coils if phase corrections are performed using the reference scans.

Shielding calculations for the Intense Neutron Source Facility. Final report

International Nuclear Information System (INIS)

Battat, M.E.; Henninger, R.J.; Macdonald, J.L.; Dudziak, D.J.

1978-06-01

Results of shielding calculations for the Intnse Neutron Source (INS) facility are presented. The INS facility is designed to house two sources, each of which will produce D--T neutrons with intensities in the range from 1 to 3 x 10 15 n/s on a continuous basis. Topics covered include the design of the biological shield, use of two-dimensional discrete-ordinates results to specify the source terms for a Monte Carlo skyshine calculation, air activation, and dose rates in the source cell (after shutdown) due to activation of the biological shield

Measurement of the thermal neutron self shielding coefficient in the Syrian miniature neutron source reactor inner irradiation site using the dy soils

International Nuclear Information System (INIS)

Khattab, K.; Khamis, I.

2007-01-01

Measurement of the thermal self shielding coefficient ( Gth ) in the Syrian Miniature Neutron Source Reactor (MNSR) inner irradiation site using Dy foils is presented in this paper. The thermal self shielding coefficient is measured as a function of the foil thickness or numbers. The mathematical equation which calculates the average relative radioactivity (Bq/g) versus the foil number is found as well.

Monte Carlo simulations and measurements for efficiency determination of lead shielded plastic scintillator detectors

Science.gov (United States)

Yasin, Zafar; Negoita, Florin; Tabbassum, Sana; Borcea, Ruxandra; Kisyov, Stanimir

2017-12-01

The plastic scintillators are used in different areas of science and technology. One of the use of these scintillator detectors is as beam loss monitors (BLM) for new generation of high intensity heavy ion in superconducting linear accelerators. Operated in pulse counting mode with rather high thresholds and shielded by few centimeters of lead in order to cope with radiofrequency noise and X-ray background emitted by accelerator cavities, they preserve high efficiency for high energy gamma ray and neutrons produced in the nuclear reactions of lost beam particles with accelerator components. Efficiency calculation and calibration of detectors is very important before their practical usage. In the present work, the efficiency of plastic scintillator detectors is simulated using FLUKA for different gamma and neutron sources like, 60Co, 137Cs and 238Pu-Be. The sources are placed at different positions around the detector. Calculated values are compared with the measured values and a reasonable agreement is observed.

Determination of boron in Jabroc wood used as a shielding material in nuclear reactors

International Nuclear Information System (INIS)

Kamble, Granthali S.; Manisha, V.; Venkatesh, K.

2015-01-01

Jabroc are non-impregnated, densified wood laminates developed commercially for a wide range of industrial applications. Jabroc can be used with other neutron shielding materials such as Lead to form complex shielding structures. Its relative light weight and cleanliness in handling are additional features that make it a suitable candidate for the standard design of neutron shielding equipment. Jabroc can also be impregnated with Boron up to a maximum of 4% to be used in areas where Gamma radiation produced on Neutron capture reaches unacceptable dose rates. Boron impregnated Jabroc wood finds application in TAPS 3 and 4 as a shielding material for the Ion Chambers and the Horizontal Flux Units (HFU). The shielding property of this material is optimized by incorporating requisite amount of boron in wood. Boron content in this material has to be determined accurately prior to its use in the nuclear reactors. In this work a method was standardized to determine boron in Jabroc wood samples to check for conformance to specifications. The wood sample flakes were wetted with saturated barium hydroxide solution and dries under IR. The sample was ashed in a muffle furnace at 600℃ for 2 h

Evaluation of the performance of peridotite aggregates for radiation shielding concrete

International Nuclear Information System (INIS)

Wang, Jinjun; Li, Guofeng; Meng, Dechuan

2014-01-01

Highlights: • Using peridotite rich in crystal water as aggregates of radiation-shielding concrete. • Performance of peridotite concrete is simulated and compared with ordinary concrete. • Performance of concrete samples is tested. • Neutron shielding performance can be significantly enhanced by peridotite aggregates. - Abstract: Peridotite is a kind of material that is rich in crystal water. In this paper, peridotite is used as fine and coarse aggregates for radiation shielding concrete. The transmission data of different concrete thickness and different energy neutron are calculated using Monte-Carlo method. The neutron shielding performance of the peridotite concrete samples are tested using 241 Am-Be neutron source. The results show that the peridotite is an excellent neutron shielding material

Improvement of the neutron flux calculations in thick shield by conditional Monte Carlo and deterministic methods

International Nuclear Information System (INIS)

Ghassoun, Jillali; Jehoauni, Abdellatif

2000-01-01

In practice, the estimation of the flux obtained by Fredholm integral equation needs a truncation of the Neuman series. The order N of the truncation must be large in order to get a good estimation. But a large N induces a very large computation time. So the conditional Monte Carlo method is used to reduce time without affecting the estimation quality. In a previous works, in order to have rapid convergence of calculations it was considered only weakly diffusing media so that has permitted to truncate the Neuman series after an order of 20 terms. But in the most practical shields, such as water, graphite and beryllium the scattering probability is high and if we truncate the series at 20 terms we get bad estimation of flux, so it becomes useful to use high orders in order to have good estimation. We suggest two simple techniques based on the conditional Monte Carlo. We have proposed a simple density of sampling the steps for the random walk. Also a modified stretching factor density depending on a biasing parameter which affects the sample vector by stretching or shrinking the original random walk in order to have a chain that ends at a given point of interest. Also we obtained a simple empirical formula which gives the neutron flux for a medium characterized by only their scattering probabilities. The results are compared to the exact analytic solution, we have got a good agreement of results with a good acceleration of convergence calculations. (author)

A Benchmarking Study of High Energy Carbon Ion Induced Neutron Using Several Monte Carlo Codes

Energy Technology Data Exchange (ETDEWEB)

Kim, D. H.; Oh, J. H.; Jung, N. S.; Lee, H. S. [Pohang Accelerator Laboratory, Pohang (Korea, Republic of); Shin, Y. S.; Kwon, D. Y.; Kim, Y. M. [Catholic Univ., Gyeongsan (Korea, Republic of); Oranj, L. Mokhtari [POSTECH, Pohang (Korea, Republic of)

2014-10-15

In this study, the benchmarking study was done for the representative particle interaction of the heavy ion accelerator, especially carbon-induced reaction. The secondary neutron is an important particle in the shielding analysis to define the source term and penetration ability of radiation fields. The performance of each Monte Carlo codes were verified for selected codes: MCNPX 2.7, PHITS 2.64 and FLUKA 2011.2b.6. For this benchmarking study, the experimental data of Kurosawa et al. in the SINBAD database of NEA was applied. The calculated results of the differential neutron yield produced from several materials irradiated by high energy carbon beam reproduced the experimental data well in small uncertainty. But the MCNPX results showed large discrepancy with experimental data, especially at the forward angle. The calculated results were lower a little than the experimental and it was clear in the cases of lower incident carbon energy, thinner target and forward angle. As expected, the influence of different model was found clearly at forward direction. In the shielding analysis, these characteristics of each Monte Carlo codes should be considered and utilized to determine the safety margin of a shield thickness.

A Benchmarking Study of High Energy Carbon Ion Induced Neutron Using Several Monte Carlo Codes

International Nuclear Information System (INIS)

Kim, D. H.; Oh, J. H.; Jung, N. S.; Lee, H. S.; Shin, Y. S.; Kwon, D. Y.; Kim, Y. M.; Oranj, L. Mokhtari

2014-01-01

In this study, the benchmarking study was done for the representative particle interaction of the heavy ion accelerator, especially carbon-induced reaction. The secondary neutron is an important particle in the shielding analysis to define the source term and penetration ability of radiation fields. The performance of each Monte Carlo codes were verified for selected codes: MCNPX 2.7, PHITS 2.64 and FLUKA 2011.2b.6. For this benchmarking study, the experimental data of Kurosawa et al. in the SINBAD database of NEA was applied. The calculated results of the differential neutron yield produced from several materials irradiated by high energy carbon beam reproduced the experimental data well in small uncertainty. But the MCNPX results showed large discrepancy with experimental data, especially at the forward angle. The calculated results were lower a little than the experimental and it was clear in the cases of lower incident carbon energy, thinner target and forward angle. As expected, the influence of different model was found clearly at forward direction. In the shielding analysis, these characteristics of each Monte Carlo codes should be considered and utilized to determine the safety margin of a shield thickness

The Spallation Neutron Source (SNS) conceptual design shielding analysis

International Nuclear Information System (INIS)

Johnson, J.O.; Odano, N.; Lillie, R.A.

1998-03-01

The shielding design is important for the construction of an intense high-energy accelerator facility like the proposed Spallation Neutron Source (SNS) due to its impact on conventional facility design, maintenance operations, and since the cost for the radiation shielding shares a considerable part of the total facility costs. A calculational strategy utilizing coupled high energy Monte Carlo calculations and multi-dimensional discrete ordinates calculations, along with semi-empirical calculations, was implemented to perform the conceptual design shielding assessment of the proposed SNS. Biological shields have been designed and assessed for the proton beam transport system and associated beam dumps, the target station, and the target service cell and general remote maintenance cell. Shielding requirements have been assessed with respect to weight, space, and dose-rate constraints for operating, shutdown, and accident conditions. A discussion of the proposed facility design, conceptual design shielding requirements calculational strategy, source terms, preliminary results and conclusions, and recommendations for additional analyses are presented

Bulk Shielding Calculation for 90 .deg. Bending Section of RISP

Energy Technology Data Exchange (ETDEWEB)

Oh, J. H.; Jung, N. S.; Lee, H. S. [Pohang Accelerator Laboratory, Pohang (Korea, Republic of); Oranj, L. Mokhtari [POSTECH, Pohang (Korea, Republic of); Ko, S. K. [Univ. of Ulsan, Ulsan (Korea, Republic of)

2014-10-15

The charge state of {sup 238}U beams with maximum intensity was 79+ among multi-charge states of 70+ to 89+, which were estimated by using LISE++ code. The bending section consists of twenty four quadrupoles, two dipoles, two two-cell type superconducting RF cavities and eleven slits. The complicated radiation environment is caused by the beam losses occurred normally during the stripping process and when the produced {sup 238}U beams are transported along the beam line. Secondary radiations generated by {sup 238}U beams irradiation are very important for predicting the prompt and residual doses and the radiation damage at the component. The production characteristics of neutron and photon from thin carbon and thick iron were studied to set up the shielding strategy. The dose estimation was done to the pre-designed the tunnel structure. In these calculations, major Monte Carlo codes, PHITS and FLUKA, were used. The present study provided information of shielding analysis for the 90 .deg. bending section of RISP facility. The source term was evaluated to determine fundamental parameter of the shielding analysis using PHITS and FLUKA codes. And the distribution of the dose rate at the outside of thick shielding wall was presented.

Mood states determine the degree of task shielding in dual-task performance.

Science.gov (United States)

Zwosta, Katharina; Hommel, Bernhard; Goschke, Thomas; Fischer, Rico

2013-01-01

Current models of multitasking assume that dual-task performance and the degree of multitasking are affected by cognitive control strategies. In particular, cognitive control is assumed to regulate the amount of shielding of the prioritised task from crosstalk from the secondary task. We investigated whether and how task shielding is influenced by mood states. Participants were exposed to two short film clips, one inducing high and one inducing low arousal, of either negative or positive content. Negative mood led to stronger shielding of the prioritised task (i.e., less crosstalk) than positive mood, irrespective of arousal. These findings support the assumption that emotional states determine the parameters of cognitive control and play an important role in regulating dual-task performance.

New approximations for the Doppler broadening function applied to the calculation of resonance self-shielding factors

International Nuclear Information System (INIS)

Palma, Daniel A.; Goncalves, Alessandro C.; Martinez, Aquilino S.; Silva, Fernando C.

2008-01-01

The activation technique allows much more precise measurements of neutron intensity, relative or absolute. The technique requires the knowledge of the Doppler broadening function ψ(x,ξ) to determine the resonance self-shielding factors in the epithermal range G epi (τ,ξ). Two new analytical approximations for the Doppler broadening function ψ(x,ξ) are proposed. The approximations proposed are compared with other methods found in literature for the calculation of the ψ(x,ξ) function, that is, the 4-pole Pade method and the Frobenius method, when applied to the calculation of G epi (τ,ξ). The results obtained provided satisfactory accuracy. (authors)

New approximations for the Doppler broadening function applied to the calculation of resonance self-shielding factors

Energy Technology Data Exchange (ETDEWEB)

Palma, Daniel A. [CEFET QUIMICA de Nilopolis/RJ, Rio de Janeiro (Brazil); Goncalves, Alessandro C.; Martinez, Aquilino S.; Silva, Fernando C. [COPPE/UFRJ - Programa de Engenharia Nuclear, Rio de Janeiro (Brazil)

2008-07-01

The activation technique allows much more precise measurements of neutron intensity, relative or absolute. The technique requires the knowledge of the Doppler broadening function psi(x,xi) to determine the resonance self-shielding factors in the epithermal range G{sub epi} (tau,xi). Two new analytical approximations for the Doppler broadening function psi(x,xi) are proposed. The approximations proposed are compared with other methods found in literature for the calculation of the psi(x,xi) function, that is, the 4-pole Pade method and the Frobenius method, when applied to the calculation of G{sub epi} (tau,xi). The results obtained provided satisfactory accuracy. (authors)

Spiral MRI on a 9.4T Vertical-bore Superconducting Magnet Using Unshielded and Self-shielded Gradient Coils

Science.gov (United States)

Kodama, Nao; Setoi, Ayana; Kose, Katsumi

2018-01-01

Spiral MRI sequences were developed for a 9.4T vertical standard bore (54 mm) superconducting magnet using unshielded and self-shielded gradient coils. Clear spiral images with 64-shot scan were obtained with the self-shielded gradient coil, but severe shading artifacts were observed for the spiral-scan images acquired with the unshielded gradient coil. This shading artifact was successfully corrected with a phase-correction technique using reference scans that we developed based on eddy current field measurements. We therefore concluded that spiral imaging sequences can be installed even for unshielded gradient coils if phase corrections are performed using the reference scans. PMID:28367906

Comparison of experimental and calculated shielding factors for modular buildings in a radioactive fallout scenario

DEFF Research Database (Denmark)

Hinrichsen, Yvonne; Finck, Robert; Östlund, Karl

2018-01-01

building used was a standard prefabricated structure obtained from a commercial manufacturer. Four reference positions for the gamma radiation detectors were used inside the building. Theoretical dose rate calculations were performed using the Monte Carlo code MCNP6, and additional calculations were......Experimentally and theoretically determined shielding factors for a common light construction dwelling type were obtained and compared. Sources of the gamma-emitting radionuclides 60Co and 137Cs were positioned around and on top of a modular building to represent homogeneous fallout. The modular...... performed that compared the shielding factor for 137Cs and 134Cs. This work demonstrated the applicability of using MCNP6 for theoretical calculations of radioactive fallout scenarios. Furthermore, the work showed that the shielding effect for modular buildings is almost the same for 134Cs as for 137Cs....

Resonance self-shielding effect analysis of neutron data libraries applied for the dual-cooled waste transmutation blanket of the fusion-driven subcritical system

International Nuclear Information System (INIS)

Liu Haibo; Wu Yican; Zheng Shanliang; Zhang Chunzao

2004-01-01

Based on the Fusion-Driven Subcritical System (FDS-I), the 25 groups, 175 groups and 620 groups neutron nuclear data libraries with/without resonance self-shielding correction are made with the Njoy and Transx codes, and the K eff and reaction rates are calculated with the Anisn code. The conclusion indicates that the resonance self-shielding effect affects the reaction rates strongly. (authors)

Determination of the exposure speed of radiation emitted by the linear accelerator, using the code MCNP5 to evaluate the radiotherapy room shields of ABC Hospital

International Nuclear Information System (INIS)

Corral B, J. R.

2015-01-01

Humans should avoid exposure to radiation, because the consequences are harmful to health. Although there are different emission sources of radiation, generated by medical devices they are usually of great interest, since people who attend hospitals are exposed in one way or another to ionizing radiation. Therefore, is important to conduct studies on radioactive levels that are generated in hospitals, as a result of the use of medical equipment. To determine levels of exposure speed of a radioactive facility there are different methods, including the radiation detector and computational method. This thesis uses the computational method. With the program MCNP5 was determined the speed of the radiation exposure in the radiotherapy room of Cancer Center of ABC Hospital in Mexico City. In the application of computational method, first the thicknesses of the shields were calculated, using variables as: 1) distance from the shield to the source; 2) desired weekly equivalent dose; 3) weekly total dose equivalent emitted by the equipment; 4) occupation and use factors. Once obtained thicknesses, we proceeded to model the bunker using the mentioned program. The program uses the Monte Carlo code to probabilistic ally determine the phenomena of interaction of radiation with the shield, which will be held during the X-ray emission from the linear accelerator. The results of computational analysis were compared with those obtained experimentally with the detection method, for which was required the use of a Geiger-Muller counter and the linear accelerator was programmed with an energy of 19 MV with 500 units monitor positioning the detector in the corresponding boundary. (Author)

Nomogram for Determining Shield Thickness for Point and Line Sources of Gamma Rays

Energy Technology Data Exchange (ETDEWEB)

Joenemalm, C; Malen, K

1966-10-15

A set of nomograms is given for the determination of the required shield thickness against gamma radiation. The sources handled are point and infinite line sources with shields of Pb, Fe, magnetite concrete (p = 3.6), ordinary concrete (p = 2.3) or water. The gamma energy range covered is 0.5 - 10 MeV. The nomograms are directly applicable for source and dose points on the surfaces of the shield. They can easily be extended to source and dose points in other positions by applying a geometrical correction. Also included are data for calculation of the source strength for the most common materials and for fission product sources.

Nomogram for Determining Shield Thickness for Point and Line Sources of Gamma Rays

International Nuclear Information System (INIS)

Joenemalm, C.; Malen, K

1966-10-01

A set of nomograms is given for the determination of the required shield thickness against gamma radiation. The sources handled are point and infinite line sources with shields of Pb, Fe, magnetite concrete (p = 3.6), ordinary concrete (p = 2.3) or water. The gamma energy range covered is 0.5 - 10 MeV. The nomograms are directly applicable for source and dose points on the surfaces of the shield. They can easily be extended to source and dose points in other positions by applying a geometrical correction. Also included are data for calculation of the source strength for the most common materials and for fission product sources

A review of radiation dosimetry applications using the MCNP Monte Carlo code

Energy Technology Data Exchange (ETDEWEB)

Solberg, T.D.; DeMarco, J.J.; Chetty, I.J.; Mesa, A.V.; Cagnon, C.H.; Li, A.N.; Mather, K.K.; Medin, P.M.; Arellano, A.R.; Smathers, J.B. [California Univ., Los Angeles, CA (United States). Dept. of Radiation Oncology

2001-07-01

The Monte Carlo code MCNP (Monte Carlo N-Particle) has a significant history dating to the early years of the Manhattan Project. More recently, MCNP has been used successfully to solve many problems in the field of medical physics. In radiotherapy applications MCNP has been used successfully to calculate the bremsstrahlung spectra from medical linear accelerators, for modeling the dose distributions around high dose rate brachytherapy sources, and for evaluating the dosimetric properties of new radioactive sources used in intravascular irradiation for prevention of restenosis following angioplasty. MCNP has also been used for radioimmunotherapy and boron neutron capture therapy applications. It has been used to predict fast neutron activation of shielding and biological materials. One area that holds tremendous clinical promise is that of radiotherapy treatment planning. In diagnostic applications, MCNP has been used to model X-ray computed tomography and positron emission tomography scanners, to compute the dose delivered from CT procedures, and to determine detector characteristics of nuclear medicine devices. MCNP has been used to determine particle fluxes around radiotherapy treatment devices and to perform shielding calculations in radiotherapy treatment rooms. This manuscript is intended to provide to the reader a comprehensive summary of medical physics applications of the MCNP code. (orig.)

A review of radiation dosimetry applications using the MCNP Monte Carlo code

International Nuclear Information System (INIS)

Solberg, T.D.; DeMarco, J.J.; Chetty, I.J.; Mesa, A.V.; Cagnon, C.H.; Li, A.N.; Mather, K.K.; Medin, P.M.; Arellano, A.R.; Smathers, J.B.

2002-01-01

The Monte Carlo code MCNP (Monte Carlo N-Particle) has a significant history dating to the early years of the Manhattan Project. More recently, MCNP has been used successfully to solve many problems in the field of medical physics. In radiotherapy applications MCNP has been used successfully to calculate the bremsstrahlung spectra from medical linear accelerators, for modeling the dose distributions around high dose rate brachytherapy sources, and for evaluating the dosimetric properties of new radioactive sources used in intravascular irradiation for prevention of restenosis following angioplasty. MCNP has also been used for radioimmunotherapy and boron neutron capture therapy applications. It has been used to predict fast neutron activation of shielding and biological materials. One area that holds tremendous clinical promise is that of radiotherapy treatment planning. In diagnostic applications, MCNP has been used to model X-ray computed tomography and positron emission tomography scanners, to compute the dose delivered from CT procedures, and to determine detector characteristics of nuclear medicine devices. MCNP has been used to determine particle fluxes around radiotherapy treatment devices and to perform shielding calculations in radiotherapy treatment rooms. This manuscript is intended to provide to the reader a comprehensive summary of medical physics applications of the MCNP code. (author)

Guideline of Monte Carlo calculation. Neutron/gamma ray transport simulation by Monte Carlo method

CERN Document Server

2002-01-01

This report condenses basic theories and advanced applications of neutron/gamma ray transport calculations in many fields of nuclear energy research. Chapters 1 through 5 treat historical progress of Monte Carlo methods, general issues of variance reduction technique, cross section libraries used in continuous energy Monte Carlo codes. In chapter 6, the following issues are discussed: fusion benchmark experiments, design of ITER, experiment analyses of fast critical assembly, core analyses of JMTR, simulation of pulsed neutron experiment, core analyses of HTTR, duct streaming calculations, bulk shielding calculations, neutron/gamma ray transport calculations of the Hiroshima atomic bomb. Chapters 8 and 9 treat function enhancements of MCNP and MVP codes, and a parallel processing of Monte Carlo calculation, respectively. An important references are attached at the end of this report.

Automatic welding technologies for long-distance pipelines by use of all-position self-shielded flux cored wires

Directory of Open Access Journals (Sweden)

Zeng Huilin

2014-10-01

Full Text Available In order to realize the automatic welding of pipes in a complex operation environment, an automatic welding system has been developed by use of all-position self-shielded flux cored wires due to their advantages, such as all-position weldability, good detachability, arc's stability, low incomplete fusion, no need for welding protective gas or protection against wind when the wind speed is < 8 m/s. This system consists of a welding carrier, a guide rail, an auto-control system, a welding source, a wire feeder, and so on. Welding experiments with this system were performed on the X-80 pipeline steel to determine proper welding parameters. The welding technique comprises root welding, filling welding and cover welding and their welding parameters were obtained from experimental analysis. On this basis, the mechanical properties tests were carried out on welded joints in this case. Results show that this system can help improve the continuity and stability of the whole welding process and the welded joints' inherent quality, appearance shape, and mechanical performance can all meet the welding criteria for X-80 pipeline steel; with no need for windbreak fences, the overall welding cost will be sharply reduced. Meanwhile, more positive proposals were presented herein for the further research and development of this self-shielded flux core wires.

Shielding considerations for neutral-beam injection systems

International Nuclear Information System (INIS)

de Seynes, X.

1983-03-01

Results of a study on the geometry of an FED-A Neutral Beam Injector beamline duct shield are presented. Also included is a calculation of dose rates, as a function of time, from an activated NBI. The shielding investigations consisted of varying the parameters of the geometry and transporting particles through it using the MCNP Monte-Carlo code. The dose rates were calculated by the ACDOS3 code using realistic MCNP results. A final-to-incident flux ratio of 6.5 x 10 -7 can be achieved through the use of a 65.5 cm reentry duct. This is for a realistic source and pure water shielding material. The activated NBI produced a dose rate of 15.9 mrem/hr two and a half days after shutdown of the reactor

Evaluation of gamma-ray attenuation properties of bismuth borate glass systems using Monte Carlo method

Science.gov (United States)

Tarim, Urkiye Akar; Ozmutlu, Emin N.; Yalcin, Sezai; Gundogdu, Ozcan; Bradley, D. A.; Gurler, Orhan

2017-11-01

A Monte Carlo method was developed to investigate radiation shielding properties of bismuth borate glass. The mass attenuation coefficients and half-value layer parameters were determined for different fractional amounts of Bi2O3 in the glass samples for the 356, 662, 1173 and 1332 keV photon energies. A comparison of the theoretical and experimental attenuation coefficients is presented.

Self-shielding and burn-out effects in the irradiation of strongly-neutron-absorbing material

International Nuclear Information System (INIS)

Sekine, T.; Baba, H.

1978-01-01

Self-shielding and burn-out effects are discussed in the evaluation of radioisotopes formed by neutron irradiation of a strongly-neutron-absorbing material. A method of the evaluation of such effects is developed both for thermal and epithermal neutrons. Gadolinium oxide uniformly mixed with graphite powder was irradiated by reactor-neutrons together with pieces of a Co-Al alloy wire (the content of Co being 0.475%) as the neutron flux monitor. The configuration of the samples and flux monitors in each of two irradiations is illustrated. The yields of activities produced in the irradiated samples were determined by the γ-spectrometry with a Ge(Li) detector of a relative detection efficiency of 8%. Activities at the end of irradiation were estimated by corrections due to pile-up, self-absorption, detection efficiency, branching ratio, and decay of the activity. Results of the calculation are discussed in comparison with the observed yields of 153 Gd, 160 Tb, and 161 Tb for the case of neutron irradiation of disc-shaped targets of gadolinium oxide. (T.G.)

Neutron and gamma ray transport calculations in shielding system

Energy Technology Data Exchange (ETDEWEB)

Masukawa, Fumihiro; Sakamoto, Hiroki [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

1998-03-01

In the shields for radiation in nuclear facilities, the penetrating holes of various kinds and irregular shapes are made for the reasons of operation, control and others. These penetrating holes and gaps are filled with air or the substances with relatively small shielding performance, and radiation flows out through them, which is called streaming. As the calculation techniques for the shielding design or analysis related to the streaming problem, there are the calculations by simplified evaluation, transport calculation and Monte Carlo method. In this report, the example of calculation by Monte Carlo method which is represented by MCNP code is discussed. A number of variance reduction techniques which seem effective for the analysis of streaming problem were tried. As to the investigation of the applicability of MCNP code to streaming analysis, the object of analysis which are the concrete walls without hole and with horizontal hole, oblique hole and bent oblique hole, the analysis procedure, the composition of concrete, and the conversion coefficient of dose equivalent, and the results of analysis are reported. As for variance reduction technique, cell importance was adopted. (K.I.)

The Monte Carlo code MCBEND - where it is and where it's going

International Nuclear Information System (INIS)

Chukas, S.J.; Miller, P.C.; Power, S.W.

1990-05-01

The Monte Carlo method forms a corner stone to the calculational procedures established in the UK for shielding design and assessment. The emphasis of the work in the shielding area is centred on the Monte Carlo code MCBEND. The work programme in support of the code is broadly directed towards utilisation of new hardware, the development of improved modelling algorithms, the development of new acceleration methods for specific applications and enhancements to user image. This paper summarises the current status of MCBEND and reviews developments carried out over the past two years and planned for the future. (author)

Hybrid SN/Monte Carlo research and results

International Nuclear Information System (INIS)

Baker, R.S.

1993-01-01

The neutral particle transport equation is solved by a hybrid method that iteratively couples regions where deterministic (S N ) and stochastic (Monte Carlo) methods are applied. The Monte Carlo and S N regions are fully coupled in the sense that no assumption is made about geometrical separation or decoupling. The hybrid Monte Carlo/S N method provides a new means of solving problems involving both optically thick and optically thin regions that neither Monte Carlo nor S N is well suited for by themselves. The hybrid method has been successfully applied to realistic shielding problems. The vectorized Monte Carlo algorithm in the hybrid method has been ported to the massively parallel architecture of the Connection Machine. Comparisons of performance on a vector machine (Cray Y-MP) and the Connection Machine (CM-2) show that significant speedups are obtainable for vectorized Monte Carlo algorithms on massively parallel machines, even when realistic problems requiring variance reduction are considered. However, the architecture of the Connection Machine does place some limitations on the regime in which the Monte Carlo algorithm may be expected to perform well

Applications of Monte Carlo codes to a study of gamma-ray buildup factors, skyshine and duct streaming

Energy Technology Data Exchange (ETDEWEB)

Hirayama, H. [High Energy Accelerator Research Organization (KEK), Ibaraki (Japan)

2001-07-01

Many shielding calculations for gamma-rays have continued to rely on point-kernel methods incorporating buildup factor data. Line beam or conical beam response functions, which are calculated using a Monte Carlo code, for skyshine problems are useful to estimate the skyshine dose from various facilities. A simple calculation method for duct streaming was proposed using the parameters calculated by the Monte Carlo code. It is therefore important to study, improve and produce basic parameters related to old, but still important, problems in the fields of radiation shielding using the Monte Carlo code. In this paper, these studies performed by several groups in Japan as applications of the Monte Carlo method are discussed. (orig.)

Evaluation of neutron shielding properties of lead glass using bubble detector

International Nuclear Information System (INIS)

Viswanathan, S.; Vishwa Prasad, K.; Srinivasan, T.K.; Ponraju, D.

1999-01-01

Neutron shielding properties of lead glass had been studied using a 241 Am-Be neutron source. Indigenously developed bubble detector was used as neutron detector. Attenuation curves were determined experimentally for the lead glass under the conditions of broad beam geometry. Theoretical calculations were made using Monte Carlo code MCNP3. Measurements were made for polyethylene and concrete to serve as reference. The measured and calculated neutron removal cross sections of lead glass, polyethylene and concrete are reported in this paper. Good agreement is observed between the experimental results and theoretical calculations. (author)

Statistical analysis and Monte Carlo simulation of growing self-avoiding walks on percolation

Energy Technology Data Exchange (ETDEWEB)

Zhang Yuxia [Department of Physics, Wuhan University, Wuhan 430072 (China); Sang Jianping [Department of Physics, Wuhan University, Wuhan 430072 (China); Department of Physics, Jianghan University, Wuhan 430056 (China); Zou Xianwu [Department of Physics, Wuhan University, Wuhan 430072 (China)]. E-mail: xwzou@whu.edu.cn; Jin Zhunzhi [Department of Physics, Wuhan University, Wuhan 430072 (China)

2005-09-26

The two-dimensional growing self-avoiding walk on percolation was investigated by statistical analysis and Monte Carlo simulation. We obtained the expression of the mean square displacement and effective exponent as functions of time and percolation probability by statistical analysis and made a comparison with simulations. We got a reduced time to scale the motion of walkers in growing self-avoiding walks on regular and percolation lattices.

Self-learning Monte Carlo with deep neural networks

Science.gov (United States)

Shen, Huitao; Liu, Junwei; Fu, Liang

2018-05-01

The self-learning Monte Carlo (SLMC) method is a general algorithm to speedup MC simulations. Its efficiency has been demonstrated in various systems by introducing an effective model to propose global moves in the configuration space. In this paper, we show that deep neural networks can be naturally incorporated into SLMC, and without any prior knowledge can learn the original model accurately and efficiently. Demonstrated in quantum impurity models, we reduce the complexity for a local update from O (β2) in Hirsch-Fye algorithm to O (β lnβ ) , which is a significant speedup especially for systems at low temperatures.

MCB. A continuous energy Monte Carlo burnup simulation code

International Nuclear Information System (INIS)

Cetnar, J.; Wallenius, J.; Gudowski, W.

1999-01-01

A code for integrated simulation of neutrinos and burnup based upon continuous energy Monte Carlo techniques and transmutation trajectory analysis has been developed. Being especially well suited for studies of nuclear waste transmutation systems, the code is an extension of the well validated MCNP transport program of Los Alamos National Laboratory. Among the advantages of the code (named MCB) is a fully integrated data treatment combined with a time-stepping routine that automatically corrects for burnup dependent changes in reaction rates, neutron multiplication, material composition and self-shielding. Fission product yields are treated as continuous functions of incident neutron energy, using a non-equilibrium thermodynamical model of the fission process. In the present paper a brief description of the code and applied methods are given. (author)

The influence of Shelter's FCM on the shield efficiency at there of containing

International Nuclear Information System (INIS)

Gorbachev, B.I.

2000-01-01

The reasonable detailed quantitative estimations of the influence of the γ-radiation capture and scattering processes in the Shelter's FCM material on the shield precautions efficiency at there of containing for the further shelf purpose. The Monte-Carlo calculations was carry out by the software Micro Shield 4.00 serial 4.00-00283, which make it possible correctly to account for the radiation shielding geometry, the radiation sources geometry, the radiation sources spectrums and the processes of the gamma-rays multi scattering in 'thick' shielding. Results presented in the tables, which is convenient to use. 3 refs., 18 tab

Status of multigroup cross-section data for shielding applications

International Nuclear Information System (INIS)

Roussin, R.W.; Maskewitz, B.F.; Trubey, D.K.

1983-01-01

Multigroup cross-section libraries for shielding applications in formats for direct use in discrete ordinates or Monte Carlo codes have long been a part of the Data Library Collection (DLC) of the Radiation Shielding Information Center (RSIC). In recent years libraries in more flexible and comprehensive formats, which allow the user to derive his own problem-dependent sets, have been added to the collection. The current status of both types is described, as well as projections for adding data libraries based on ENDF/B-V

Evaluation of gamma-ray attenuation properties of bismuth borate glass systems using Monte Carlo method

International Nuclear Information System (INIS)

Tarim, Urkiye Akar; Ozmutlu, Emin N.; Yalcin, Sezai; Gundogdu, Ozcan; Bradley, D.A.; Gurler, Orhan

2017-01-01

A Monte Carlo method was developed to investigate radiation shielding properties of bismuth borate glass. The mass attenuation coefficients and half-value layer parameters were determined for different fractional amounts of Bi 2 O 3 in the glass samples for the 356, 662, 1173 and 1332 keV photon energies. A comparison of the theoretical and experimental attenuation coefficients is presented. - Highlights: • Radiation shielding properties of bismuth borate glass systems have been reported. • Mass attenuation coefficients increase linearly with increase in Bi concentration. • Half-value layer decreases with increasing concentration of Bi. • Half-value layer decreases with the increase in the sample density.

FOCUS, Neutron Transport System for Complex Geometry Reactor Core and Shielding Problems by Monte-Carlo

International Nuclear Information System (INIS)

Hoogenboom, J.E.

1980-01-01

1 - Description of problem or function: FOCUS enables the calculation of any quantity related to neutron transport in reactor or shielding problems, but was especially designed to calculate differential quantities, such as point values at one or more of the space, energy, direction and time variables of quantities like neutron flux, detector response, reaction rate, etc. or averages of such quantities over a small volume of the phase space. Different types of problems can be treated: systems with a fixed neutron source which may be a mono-directional source located out- side the system, and Eigen function problems in which the neutron source distribution is given by the (unknown) fundamental mode Eigen function distribution. Using Monte Carlo methods complex 3- dimensional geometries and detailed cross section information can be treated. Cross section data are derived from ENDF/B, with anisotropic scattering and discrete or continuous inelastic scattering taken into account. Energy is treated as a continuous variable and time dependence may also be included. 2 - Method of solution: A transformed form of the adjoint Boltzmann equation in integral representation is solved for the space, energy, direction and time variables by Monte Carlo methods. Adjoint particles are defined with properties in some respects contrary to those of neutrons. Adjoint particle histories are constructed from which estimates are obtained of the desired quantity. Adjoint cross sections are defined with which the nuclide and reaction type are selected in a collision. The energy after a collision is selected from adjoint energy distributions calculated together with the adjoint cross sections in advance of the actual Monte Carlo calculation. For multiplying systems successive generations of adjoint particles are obtained which will die out for subcritical systems with a fixed neutron source and will be kept approximately stationary for Eigen function problems. Completely arbitrary problems can

3-dimensional shielding design for a spallation neutron source facility in the high-intensity proton accelerator project

Energy Technology Data Exchange (ETDEWEB)

Tamura, Masaya; Maekawa, Fujio [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

2003-03-01

Evaluation of shielding performance for a 1 MW spallation neutron source facility in the Materials and Life Science Facility being constructed in the High-Intensity Proton Accelerator Project (J-PARC) is important from a viewpoint of radiation safety and optimization of arrangement of components. This report describes evaluated results for the shielding performance with modeling three-dimensionally whole structural components including gaps between them in detail. A Monte Carlo calculation method with MCNPX2.2.6 code and LA-150 library was adopted. Streaming and void effects, optimization of shield for cost reduction and optimization of arrangement of structures such as shutters were investigated. The streaming effects were investigated quantitatively by changing the detailed structure of components and gap widths built into the calculation model. Horizontal required shield thicknesses were ranged from about 6.5 m to 7.5 m as a function of neutron beam line angles. A shutter mechanism for a horizontal neutron reflectometer that was directed downward was devised, and it was shown that the shielding performance of the shutter was acceptable. An optimal biological shield configuration was finally determined according to the calculated results. (author)

Adaptive algorithms for a self-shielding wavelet-based Galerkin method

International Nuclear Information System (INIS)

Fournier, D.; Le Tellier, R.

2009-01-01

The treatment of the energy variable in deterministic neutron transport methods is based on a multigroup discretization, considering the flux and cross-sections to be constant within a group. In this case, a self-shielding calculation is mandatory to correct sections of resonant isotopes. In this paper, a different approach based on a finite element discretization on a wavelet basis is used. We propose adaptive algorithms constructed from error estimates. Such an approach is applied to within-group scattering source iterations. A first implementation is presented in the special case of the fine structure equation for an infinite homogeneous medium. Extension to spatially-dependent cases is discussed. (authors)

Evaluation of backscatter dose from internal lead shielding in clinical electron beams using EGSnrc Monte Carlo simulations.

Science.gov (United States)

De Vries, Rowen J; Marsh, Steven

2015-11-08

Internal lead shielding is utilized during superficial electron beam treatments of the head and neck, such as lip carcinoma. Methods for predicting backscattered dose include the use of empirical equations or performing physical measurements. The accuracy of these empirical equations required verification for the local electron beams. In this study, a Monte Carlo model of a Siemens Artiste linac was developed for 6, 9, 12, and 15 MeV electron beams using the EGSnrc MC package. The model was verified against physical measurements to an accuracy of better than 2% and 2mm. Multiple MC simulations of lead interfaces at different depths, corresponding to mean electron energies in the range of 0.2-14 MeV at the interfaces, were performed to calculate electron backscatter values. The simulated electron backscatter was compared with current empirical equations to ascertain their accuracy. The major finding was that the current set of backscatter equations does not accurately predict electron backscatter, particularly in the lower energies region. A new equation was derived which enables estimation of electron backscatter factor at any depth upstream from the interface for the local treatment machines. The derived equation agreed to within 1.5% of the MC simulated electron backscatter at the lead interface and upstream positions. Verification of the equation was performed by comparing to measurements of the electron backscatter factor using Gafchromic EBT2 film. These results show a mean value of 0.997 ± 0.022 to 1σ of the predicted values of electron backscatter. The new empirical equation presented can accurately estimate electron backscatter factor from lead shielding in the range of 0.2 to 14 MeV for the local linacs.

Revised neutral gas shielding model for pellet ablation - combined neutral and plasma shielding

International Nuclear Information System (INIS)

Houlberg, W.A.; Schuresko, D.D.; Attenberger, S.E.

1986-01-01

The ablation and penetration of pellets in early ORMAK and ISX-A experiments were reliably predicted by the neutral gas shielding model of Milora and Foster. These experiments demonstrated that the principle components of the model - a self-generated shield which reduces the heat flux at the plasma surface - were correct. In more recent experiments with higher temperature plasmas, this model consistently predicts greater penetration than observed in the experiments. Upgarding known limitations of the original model brings the predicted and observed penetration values into agreement. These improvements include: (1) treating the incident electrons as having distribution in energy rather than being monoenergetic; (2) including the shielding effects of cold, dense plasma extending along the magnetic field outside the neutral shield; and (3) modifying the finite plasma, self-limiting incident heat flux so that it represents a collisionless plasma limit rather than a collisional limit. Comparisons are made between the models for a selection of ISX-B Alcator-C, and TFTR shots. The net effect of the changes in the model is an increase in pellet ablation rates and decrease in penetration for current and future experiments

CO Self-Shielding as a Mechanism to Make O-16 Enriched Solids in the Solar Nebula

Science.gov (United States)

Nuth, Joseph A. III; Johnson, Natasha M.; Hill, Hugh G. M.

2014-01-01

Photochemical self-shielding of CO has been proposed as a mechanism to produce solids observed in the modern, O-16 depleted solar system. This is distinct from the relatively O-16 enriched composition of the solar nebula, as demonstrated by the oxygen isotopic composition of the contemporary sun. While supporting the idea that self-shielding can produce local enhancements in O-16 depleted solids, we argue that complementary enhancements of O-16 enriched solids can also be produced via CO-16 based, Fischer-Tropsch type (FTT) catalytic processes that could produce much of the carbonaceous feedstock incorporated into accreting planetesimals. Local enhancements could explain observed O-16 enrichment in calcium-aluminum-rich inclusions (CAIs), such as those from the meteorite, Isheyevo (CH/CHb), as well as in chondrules from the meteorite, Acfer 214 (CH3). CO selfshielding results in an overall increase in the O-17 and O-18 content of nebular solids only to the extent that there is a net loss of CO-16 from the solar nebula. In contrast, if CO-16 reacts in the nebula to produce organics and water then the net effect of the self-shielding process will be negligible for the average oxygen isotopic content of nebular solids and other mechanisms must be sought to produce the observed dichotomy between oxygen in the Sun and that in meteorites and the terrestrial planets. This illustrates that the formation and metamorphism of rocks and organics need to be considered in tandem rather than as isolated reaction networks.

Shielding calculational system for plutonium

International Nuclear Information System (INIS)

Zimmerman, M.G.; Thomsen, D.H.

1975-08-01

A computer calculational system has been developed and assembled specifically for calculating dose rates in AEC plutonium fabrication facilities. The system consists of two computer codes and all nuclear data necessary for calculation of neutron and gamma dose rates from plutonium. The codes include the multigroup version of the Battelle Monte Carlo code for solution of general neutron and gamma shielding problems and the PUSHLD code for solution of shielding problems where low energy gamma and x-rays are important. The nuclear data consists of built in neutron and gamma yields and spectra for various plutonium compounds, an automatic calculation of age effects and all cross-sections commonly used. Experimental correlations have been performed to verify portions of the calculational system. (23 tables, 7 figs, 16 refs) (U.S.)

Design and Shielding of Radiotherapy Treatment Facilities; IPEM Report 75, 2nd Edition

Science.gov (United States)

Horton, Patrick; Eaton, David

2017-07-01

Design and Shielding of Radiotherapy Treatment Facilities provides readers with a single point of reference for protection advice to the construction and modification of radiotherapy facilities. The book assembles a faculty of national and international experts on all modalities including megavoltage and kilovoltage photons, brachytherapy and high-energy particles, and on conventional and Monte Carlo shielding calculations. This book is a comprehensive reference for qualified experts and radiation-shielding designers in radiation physics and also useful to anyone involved in the design of radiotherapy facilities.

Delayed Slater determinant update algorithms for high efficiency quantum Monte Carlo

Science.gov (United States)

McDaniel, T.; D'Azevedo, E. F.; Li, Y. W.; Wong, K.; Kent, P. R. C.

2017-11-01

Within ab initio Quantum Monte Carlo simulations, the leading numerical cost for large systems is the computation of the values of the Slater determinants in the trial wavefunction. Each Monte Carlo step requires finding the determinant of a dense matrix. This is most commonly iteratively evaluated using a rank-1 Sherman-Morrison updating scheme to avoid repeated explicit calculation of the inverse. The overall computational cost is, therefore, formally cubic in the number of electrons or matrix size. To improve the numerical efficiency of this procedure, we propose a novel multiple rank delayed update scheme. This strategy enables probability evaluation with an application of accepted moves to the matrices delayed until after a predetermined number of moves, K. The accepted events are then applied to the matrices en bloc with enhanced arithmetic intensity and computational efficiency via matrix-matrix operations instead of matrix-vector operations. This procedure does not change the underlying Monte Carlo sampling or its statistical efficiency. For calculations on large systems and algorithms such as diffusion Monte Carlo, where the acceptance ratio is high, order of magnitude improvements in the update time can be obtained on both multi-core central processing units and graphical processing units.

Absolute on-line in-pile measurement of neutron fluxes using self-powered neutron detectors: Monte Carlo sensitivity calculations

Energy Technology Data Exchange (ETDEWEB)

Vermeeren, L. [SCK/CEN, B-2400 Mol (Belgium)

2001-07-01

Self-powered neutron detectors (SPND) are well suited to monitor continuously the neutronic operating conditions of driver fuel of research reactors and to follow its burnup evolution. This is of particular importance when advanced or new MTR fuel designs need to be qualified. We have developed a detailed MCNP-4B based Monte Carlo approach for the calculation of neutron sensitivities of SPNDs. Results for the neutron sensitivity of a Rh SPND are in excellent agreement with experimental data recently obtained at the BR2 research reactor. A critical comparison of the Monte Carlo results with results from standard analytical methods reveals an important deficiency of the analytical methods in the description of the electron transport efficiency. Our calculation method allows a reliable on-line determination of the absolute in-pile neutron flux. (author)

Absolute on-line in-pile measurement of neutron fluxes using self-powered neutron detectors: Monte Carlo sensitivity calculations

International Nuclear Information System (INIS)

Vermeeren, L.

2001-01-01

Self-powered neutron detectors (SPND) are well suited to monitor continuously the neutronic operating conditions of driver fuel of research reactors and to follow its burnup evolution. This is of particular importance when advanced or new MTR fuel designs need to be qualified. We have developed a detailed MCNP-4B based Monte Carlo approach for the calculation of neutron sensitivities of SPNDs. Results for the neutron sensitivity of a Rh SPND are in excellent agreement with experimental data recently obtained at the BR2 research reactor. A critical comparison of the Monte Carlo results with results from standard analytical methods reveals an important deficiency of the analytical methods in the description of the electron transport efficiency. Our calculation method allows a reliable on-line determination of the absolute in-pile neutron flux. (author)

Monte Carlo assessment of the dose rates produced by spent fuel from CANDU reactors

International Nuclear Information System (INIS)

Pantazi, Doina; Mateescu, Silvia; Stanciu, Marcela

2003-01-01

One of the technical measures considered for biological protection is radiation shielding. The implementation process of a spent fuel intermediate storage system at Cernavoda NPP includes an evolution in computation methods related to shielding evaluation: from using simpler computer codes, like MicroShield and QAD, to systems of codes, like SCALE (which contains few independent modules) and the multipurpose and multi-particles transport code MCNP, based on Monte Carlo method. The Monte Carlo assessment of the dose rates produced by CANDU type spent fuel, during its handling for the intermediate storage, is the main objective of this paper. The work had two main features: -establishing of geometrical models according to description mode used in code MCNP, capable to account for the specific characteristics of CANDU nuclear fuel; - confirming the correctness of proposed models, by comparing MCNP results and the related results obtained with other computer codes for shielding evaluation and dose rates calculations. (authors)

Radiological Shielding Design for the Neutron High-Resolution Backscattering Spectrometer EMU at the OPAL Reactor

Directory of Open Access Journals (Sweden)

Ersez Tunay

2017-01-01

Full Text Available The shielding for the neutron high-resolution backscattering spectrometer (EMU located at the OPAL reactor (ANSTO was designed using the Monte Carlo code MCNP 5-1.60. The proposed shielding design has produced compact shielding assemblies, such as the neutron pre-monochromator bunker with sliding cylindrical block shields to accommodate a range of neutron take-off angles, and in the experimental area - shielding of neutron focusing guides, choppers, flight tube, backscattering monochromator, and additional shielding elements inside the Scattering Tank. These shielding assemblies meet safety and engineering requirements and cost constraints. The neutron dose rates around the EMU instrument were reduced to < 0.5 µSv/h and the gamma dose rates to a safe working level of ≤ 3 µSv/h.

Radiological Shielding Design for the Neutron High-Resolution Backscattering Spectrometer EMU at the OPAL Reactor

Science.gov (United States)

Ersez, Tunay; Esposto, Fernando; Souza, Nicolas R. de

2017-09-01

The shielding for the neutron high-resolution backscattering spectrometer (EMU) located at the OPAL reactor (ANSTO) was designed using the Monte Carlo code MCNP 5-1.60. The proposed shielding design has produced compact shielding assemblies, such as the neutron pre-monochromator bunker with sliding cylindrical block shields to accommodate a range of neutron take-off angles, and in the experimental area - shielding of neutron focusing guides, choppers, flight tube, backscattering monochromator, and additional shielding elements inside the Scattering Tank. These shielding assemblies meet safety and engineering requirements and cost constraints. The neutron dose rates around the EMU instrument were reduced to < 0.5 µSv/h and the gamma dose rates to a safe working level of ≤ 3 µSv/h.

Neutral and plasma shielding model for pellet ablation

International Nuclear Information System (INIS)

Houlberg, W.A.; Milora, S.L.; Attenberger, S.E.

1987-10-01

The neutral gas shielding model for ablation of frozen hydrogenic pellets is extended to include the effects of an initial Maxwelliam distribution of incident electron energies; a cold plasma shield outside the neutral shield and extended along the magnetic field; energetic neutral beam ions and alpha particles; and self-limiting electron ablation in the collisionless plasma limit. Including the full electron distribution increases ablation, but adding the cold ionized shield reduces ablation; the net effect is a modest reduction in pellet penetration compared with the monoenergetic electron neutral shielding model with no plasma shield. Unlike electrons, fast ions can enter the neutral shield directly without passing through the cold ionized shield because their gyro-orbits are typically larger than the diameter of the cold plasma tube. Fast alpha particles should not enhance the ablation rate unless their population exceeds that expected from local classical thermalization. Fast beam ions, however, may enhance ablation in the plasma periphery if their population is high enough. Self-limiting ablation in the collisionless limit leads to a temporary distortion of the original plasma electron Maxwellian distribution function through preferential depopulation of the higher-energy electrons. 23 refs., 9 figs

Neutron shielding material

International Nuclear Information System (INIS)

Nodaka, M.; Iida, T.; Taniuchi, H.; Yosimura, K.; Nagahama, H.

1993-01-01

From among the neutron shielding materials of the 'kobesh' series developed by Kobe Steel, Ltd. for transport and storage packagings, silicon rubber base type material has been tested for several items with a view to practical application and official authorization, and in order to determine its adaptability to actual vessels. Silicon rubber base type 'kobesh SR-T01' is a material in which, from among the silicone rubber based neutron shielding materials, the hydrogen content is highest and the boron content is most optimized. Its neutron shielding capability has been already described in the previous report (Taniuchi, 1986). The following tests were carried out to determine suitability for practical application; 1) Long-term thermal stability test 2) Pouring test on an actual-scale model 3) Fire test The experimental results showed that the silicone rubber based neutron shielding material has good neutron shielding capability and high long-term fire resistance, and that it can be applied to the advanced transport packaging. (author)

Effect of interpolation error in pre-processing codes on calculations of self-shielding factors and their temperature derivatives

International Nuclear Information System (INIS)

Ganesan, S.; Gopalakrishnan, V.; Ramanadhan, M.M.; Cullan, D.E.

1986-01-01

We investigate the effect of interpolation error in the pre-processing codes LINEAR, RECENT and SIGMA1 on calculations of self-shielding factors and their temperature derivatives. We consider the 2.0347 to 3.3546 keV energy region for 238 U capture, which is the NEACRP benchmark exercise on unresolved parameters. The calculated values of temperature derivatives of self-shielding factors are significantly affected by interpolation error. The sources of problems in both evaluated data and codes are identified and eliminated in the 1985 version of these codes. This paper helps to (1) inform code users to use only 1985 versions of LINEAR, RECENT, and SIGMA1 and (2) inform designers of other code systems where they may have problems and what to do to eliminate their problems. (author)

Effect of interpolation error in pre-processing codes on calculations of self-shielding factors and their temperature derivatives

International Nuclear Information System (INIS)

Ganesan, S.; Gopalakrishnan, V.; Ramanadhan, M.M.; Cullen, D.E.

1985-01-01

The authors investigate the effect of interpolation error in the pre-processing codes LINEAR, RECENT and SIGMA1 on calculations of self-shielding factors and their temperature derivatives. They consider the 2.0347 to 3.3546 keV energy region for /sup 238/U capture, which is the NEACRP benchmark exercise on unresolved parameters. The calculated values of temperature derivatives of self-shielding factors are significantly affected by interpolation error. The sources of problems in both evaluated data and codes are identified and eliminated in the 1985 version of these codes. This paper helps to (1) inform code users to use only 1985 versions of LINEAR, RECENT, and SIGMA1 and (2) inform designers of other code systems where they may have problems and what to do to eliminate their problems

COG10, Multiparticle Monte Carlo Code System for Shielding and Criticality Use

International Nuclear Information System (INIS)

2007-01-01

1 - Description of program or function: COG is a modern, full-featured Monte Carlo radiation transport code which provides accurate answers to complex shielding, criticality, and activation problems. COG was written to be state-of-the-art and free of physics approximations and compromises found in earlier codes. COG is fully 3-D, uses point-wise cross sections and exact angular scattering, and allows a full range of biasing options to speed up solutions for deep penetration problems. Additionally, a criticality option is available for computing Keff for assemblies of fissile materials. ENDL or ENDFB cross section libraries may be used. COG home page: http://www-phys.llnl.gov/N_Div/COG/. Cross section libraries are included in the package. COG can use either the LLNL ENDL-90 cross section set or the ENDFB/VI set. Analytic surfaces are used to describe geometric boundaries. Parts (volumes) are described by a method of Constructive Solid Geometry. Surface types include surfaces of up to fourth order, and pseudo-surfaces such as boxes, finite cylinders, and figures of revolution. Repeated assemblies need be defined only once. Parts are visualized in cross-section and perspective picture views. Source and random-walk biasing techniques may be selected to improve solution statistics. These include source angular biasing, importance weighting, particle splitting and Russian roulette, path-length stretching, point detectors, scattered direction biasing, and forced collisions. Criticality - For a fissioning system, COG will compute Keff by transporting batches of neutrons through the system. Activation - COG can compute gamma-ray doses due to neutron-activated materials, starting with just a neutron source. Coupled Problems - COG can solve coupled problems involving neutrons, photons, and electrons. 2 - Methods:COG uses Monte Carlo methods to solve the Boltzmann transport equation for particles traveling through arbitrary 3-dimensional geometries. Neutrons, photons

Novel hybrid Monte Carlo/deterministic technique for shutdown dose rate analyses of fusion energy systems

International Nuclear Information System (INIS)

Ibrahim, Ahmad M.; Peplow, Douglas E.; Peterson, Joshua L.; Grove, Robert E.

2014-01-01

Highlights: •Develop the novel Multi-Step CADIS (MS-CADIS) hybrid Monte Carlo/deterministic method for multi-step shielding analyses. •Accurately calculate shutdown dose rates using full-scale Monte Carlo models of fusion energy systems. •Demonstrate the dramatic efficiency improvement of the MS-CADIS method for the rigorous two step calculations of the shutdown dose rate in fusion reactors. -- Abstract: The rigorous 2-step (R2S) computational system uses three-dimensional Monte Carlo transport simulations to calculate the shutdown dose rate (SDDR) in fusion reactors. Accurate full-scale R2S calculations are impractical in fusion reactors because they require calculating space- and energy-dependent neutron fluxes everywhere inside the reactor. The use of global Monte Carlo variance reduction techniques was suggested for accelerating the R2S neutron transport calculation. However, the prohibitive computational costs of these approaches, which increase with the problem size and amount of shielding materials, inhibit their ability to accurately predict the SDDR in fusion energy systems using full-scale modeling of an entire fusion plant. This paper describes a novel hybrid Monte Carlo/deterministic methodology that uses the Consistent Adjoint Driven Importance Sampling (CADIS) method but focuses on multi-step shielding calculations. The Multi-Step CADIS (MS-CADIS) methodology speeds up the R2S neutron Monte Carlo calculation using an importance function that represents the neutron importance to the final SDDR. Using a simplified example, preliminary results showed that the use of MS-CADIS enhanced the efficiency of the neutron Monte Carlo simulation of an SDDR calculation by a factor of 550 compared to standard global variance reduction techniques, and that the efficiency enhancement compared to analog Monte Carlo is higher than a factor of 10,000

Radiation dose reduction by water shield

International Nuclear Information System (INIS)

Zeb, J.; Arshed, W.; Ahmad, S.S.

2007-06-01

This report is an operational manual of shielding software W-Shielder, developed at Health Physics Division (HPD), Pakistan Institute of Nuclear Science and Technology (PINSTECH), Pakistan Atomic Energy Commission. The software estimates shielding thickness for photons having their energy in the range 0.5 to 10 MeV. To compute the shield thickness, self absorption in the source has been neglected and the source has been assumed as a point source. Water is used as a shielding material in this software. The software is helpful in estimating the water thickness for safe handling, storage of gamma emitting radionuclide. (author)

76 FR 63942 - Notice of Issuance of Final Determination Concerning a Surgical Mask With a Protective Eye Shield

Science.gov (United States)

2011-10-14

... Determination Concerning a Surgical Mask With a Protective Eye Shield AGENCY: U.S. Customs and Border Protection... country of origin of a Surgical Mask with a Protective Eye Shield. Based upon the facts presented, CBP has concluded in the final determination that Turkey is the country of origin of the Surgical Mask with a...

Present and future problems of radiation shielding for maritime transport of nuclear spent fuels

International Nuclear Information System (INIS)

Ueki, K.; Nariyama, N.; Ohashi, A.

2000-01-01

The transport of spent fuels with casks began in September 1999 by the exclusive spent fuel transport vessel the 'Rokuei Maru'. The casks have been transported to the reprocessing plant at Rokkasho-village in Aomori Prefecture. The 'Rokuei Maru' is approximately 100 m-length, 16.5 m-width and 3,000 gross-tons. The 20 NFT casks can be loaded into 5 holds. At the present time, the NFT casks can carry spent fuels of up to 44,000 MWD/MTU. Serpentine concrete is employed as a neutron shields in the hatch covers, the bulkheads, and the house front of the accommodations except the wheelhouse. Polyethylene covers the side walls in each hold. The neutron shielding ability of serpentine concrete and polyethylene was investigated by a shielding experiment using a 252 Cf-neutron source. The shielding experiment was analyzed with the Monte Carlo code MCNP 4B. In the near future, on-board experiment will be carried out to measure the dose-equivalent rate distributions in the 'Rokuei Maru' and the measured data and the Monte Carlo analysis of it will establish the radiation safety of the ship. (author)

Usefulness of the Monte Carlo method in reliability calculations

International Nuclear Information System (INIS)

Lanore, J.M.; Kalli, H.

1977-01-01

Three examples of reliability Monte Carlo programs developed in the LEP (Laboratory for Radiation Shielding Studies in the Nuclear Research Center at Saclay) are presented. First, an uncertainty analysis is given for a simplified spray system; a Monte Carlo program PATREC-MC has been written to solve the problem with the system components given in the fault tree representation. The second program MONARC 2 has been written to solve the problem of complex systems reliability by the Monte Carlo simulation, here again the system (a residual heat removal system) is in the fault tree representation. Third, the Monte Carlo program MONARC was used instead of the Markov diagram to solve the simulation problem of an electric power supply including two nets and two stand-by diesels

Neutron streaming analysis for shield design of FMIT Facility

International Nuclear Information System (INIS)

Carter, L.L.

1980-12-01

Applications of the Monte Carlo method have been summarized relevant to neutron streaming problems of interest in the shield design for the FMIT Facility. An improved angular biasing method has been implemented to further optimize the calculation of streaming and this method has been applied to calculate streaming within a double bend pipe

Determination of ICRF antenna fields in the vicinity of a 3-D Faraday shield structure

Energy Technology Data Exchange (ETDEWEB)

Ryan, P M; Rothe, K E; Whealton, J H; Shepard, T D [Oak Ridge National Lab., TN (USA)

1990-04-01

A three-dimensional (3-D) magnetostatic analysis developed at Oak Ridge National Laboratory has been used to calculate the electromagnetic transmission properties of representative Faraday shield designs. The analysis uses the long-wavelength approximation to obtain a 3-D Laplace solution for the magnetic scalar potential over one poloidal period of the Faraday shield, from which the complete magnetic field distribution may be obtained. Once the magnetic field distributions in the presence and absence of a Faraday shield are known, the flux transmission coefficient can be found, as well as any change in the distributed inductance of the current strap. The distrbuted capacitance of the strap can be found from an analogous 3-D electrostatic calculation, enabling the phase velocity of the slow-wave structure to be determined. Power dissipation in the shield may be estimated by equating the surface current on a perfect conductor with the surface magnetic field and using this surface current in conjunction with the finite conductivities of the shield materials to obtain the power distribution due to eddy current heating. (orig.).

Monte Carlo modeling of the Fastscan whole body counter response

International Nuclear Information System (INIS)

Graham, H.R.; Waller, E.J.

2015-01-01

Monte Carlo N-Particle (MCNP) was used to make a model of the Fastscan for the purpose of calibration. Two models were made one for the Pickering Nuclear Site, and one for the Darlington Nuclear Site. Once these models were benchmarked and found to be in good agreement, simulations were run to study the effect different sized phantoms had on the detected response, and the shielding effect of torso fat was not negligible. Simulations into the nature of a source being positioned externally on the anterior or posterior of a person were also conducted to determine a ratio that could be used to determine if a source is externally or internally placed. (author)

Simultaneous global calculation of flux and importance with forward Monte Carlo

International Nuclear Information System (INIS)

Deutsch, O.L.; Carter, L.L.

1977-01-01

A procedure is described for obtaining flux and importance globally in one Monte Carlo calculation at small to moderate incremental cost in terms of the time required to process a fixed number of particle histories. The application of this procedure and analysis of results are illustrated for a prototypical controlled thermonuclear reactor (CTR) streaming problem with coolant pipe penetrations through a concrete magnet shield. Our experience indicates that the availability of global information about both flux and importance can help to generate intuition in multidimensional shielding problems and can be of significant value during the early phase of shield design

Shielding data for hadron-therapy ion accelerators: Attenuation of secondary radiation in concrete

CERN Document Server

Agosteo, S; Sagia, E; Silari, M

2014-01-01

The secondary radiation field produced by seven different ion species (from hydrogen to nitrogen), impinging onto thick targets made of either iron or ICRU tissue, was simulated with the FLUKA Monte Carlo code, and transported through thick concrete shields: the ambient dose equivalent was estimated and shielding parameters evaluated. The energy for each ion beam was set in order to reach a maximum penetration in ICRU tissue of 290 mm (equivalent to the therapeutic range of 430 MeV/amu carbon ions). Source terms and attenuation lengths are given as a function of emission angle and ion species, along with fits to the Monte Carlo data, for shallow depth and deep penetration in the shield. Trends of source terms and attenuation lengths as a function of neutron emission angle and ion species impinging on tar- get are discussed. A comparison of double differential distributions of neutrons with results from similar simulation works reported in the literature is also included. The aim of this work is to provide shi...

FIFRELIN - TRIPOLI-4® coupling for Monte Carlo simulations with a fission model. Application to shielding calculations

Science.gov (United States)

Petit, Odile; Jouanne, Cédric; Litaize, Olivier; Serot, Olivier; Chebboubi, Abdelhazize; Pénéliau, Yannick

2017-09-01

TRIPOLI-4® Monte Carlo transport code and FIFRELIN fission model have been coupled by means of external files so that neutron transport can take into account fission distributions (multiplicities and spectra) that are not averaged, as is the case when using evaluated nuclear data libraries. Spectral effects on responses in shielding configurations with fission sampling are then expected. In the present paper, the principle of this coupling is detailed and a comparison between TRIPOLI-4® fission distributions at the emission of fission neutrons is presented when using JEFF-3.1.1 evaluated data or FIFRELIN data generated either through a n/g-uncoupled mode or through a n/g-coupled mode. Finally, an application to a modified version of the ASPIS benchmark is performed and the impact of using FIFRELIN data on neutron transport is analyzed. Differences noticed on average reaction rates on the surfaces closest to the fission source are mainly due to the average prompt fission spectrum. Moreover, when working with the same average spectrum, a complementary analysis based on non-average reaction rates still shows significant differences that point out the real impact of using a fission model in neutron transport simulations.

SP-100 GES/NAT radiation shielding systems design and development testing

International Nuclear Information System (INIS)

Disney, R.K.; Kulikowski, H.D.; McGinnis, C.A.; Reese, J.C.; Thomas, K.; Wiltshire, F.

1991-01-01

Advanced Energy Systems (AES) of Westinghouse Electric Corporation is under subcontract to the General Electric Company to supply nuclear radiation shielding components for the SP-100 Ground Engineering System (GES) Nuclear Assembly Test to be conducted at Westinghouse Hanford Company at Richland, Washington. The radiation shielding components are integral to the Nuclear Assembly Test (NAT) assembly and include prototypic and non-prototypic radiation shielding components which provide prototypic test conditions for the SP-100 reactor subsystem and reactor control subsystem components during the GES/NAT operations. W-AES is designing three radiation shield components for the NAT assembly; a prototypic Generic Flight System (GFS) shield, the Lower Internal Facility Shield (LIFS), and the Upper Internal Facility Shield (UIFS). This paper describes the design approach and development testing to support the design, fabrication, and assembly of these three shield components for use within the vacuum vessel of the GES/NAT. The GES/NAT shields must be designed to operate in a high vacuum which simulates space operations. The GFS shield and LIFS must provide prototypic radiation/thermal environments and mechanical interfaces for reactor system components. The NAT shields, in combination with the test facility shielding, must provide adequate radiation attenuation for overall test operations. Special design considerations account for the ground test facility effects on the prototypic GFS shield. Validation of the GFS shield design and performance will be based on detailed Monte Carlo analyses and developmental testing of design features. Full scale prototype testing of the shield subsystems is not planned

Preliminary shielding analysis of VHTR reactors

International Nuclear Information System (INIS)

Flaspoehler, Timothy M.; Petrovic, Bojan

2011-01-01

Over the last 20 years a number of methods have been established for automated variance reduction in Monte Carlo shielding simulations. Hybrid methods rely on deterministic adjoint and/or forward calculations to generate these parameters. In the present study, we use the FWCADIS method implemented in MAVRIC sequence of the SCALE6 package to perform preliminary shielding analyses of a VHTR reactor. MAVRIC has been successfully used by a number of researchers for a range of shielding applications, including modeling of LWRs, spent fuel storage, radiation field throughout a nuclear power plant, study of irradiation facilities, and others. However, experience in using MAVRIC for shielding studies of VHTRs is more limited. Thus, the objective of this work is to contribute toward validating MAVRIC for such applications, and identify areas for potential improvement. A simplified model of a prismatic VHTR has been devised, based on general features of the 600 MWt reactor considered as one of the NGNP options. Fuel elements have been homogenized, and the core region is represented as an annulus. However, the overall mix of materials and the relatively large dimensions of the spatial domain challenging the shielding simulations have been preserved. Simulations are performed to evaluate fast neutron fluence, dpa, and other parameters of interest at relevant positions. The paper will investigate and discuss both the effectiveness of the automated variance reduction, as well as applicability of physics model from the standpoint of specific VHTR features. (author)

Neutron/photon/electron shielding study for a laser-fusion facility

International Nuclear Information System (INIS)

Thompson, W.L.

1977-01-01

A Monte Carlo shielding study encompassing neutron, photon, and electron transport has been conducted for the High Energy Gas Laser Facility at the Los Alamos Scientific Laboratory. This paper describes the application of the Monte Carlo technique and several variance reduction schemes to the study. The calculations involve a geometry which is complicated in all three dimensions, a very intense 14 MeV neutron source, skyshine and deep penetrations. The facility design with 1.83 m concrete walls and a 1.52 m concrete roof is based on these calculations

Future directions in shielding methods and analysis

International Nuclear Information System (INIS)

Goldstein, H.

1987-01-01

Over the nearly half century history of shielding against reactor radiation, there has been a see-saw battle between theory and measurement. During that period the capability and accuracy of calculational methods have been enormously improved. The microscopic cross sections needed as input to the theoretical computations are now also known to adequate accuracy (with certain exceptions). Nonetheless, there remain substantial classes of shielding problems not yet accessible to satisfactory computational methods, particularly where three-dimensional geometries are involved. This paper discusses promising avenues to approach such problems, especially in the light of recent and expected advances in supercomputers. In particular, it seems that Monte Carlo methods should be much more advantageous in the new computer environment than they have been in the past

Contaminant deposition building shielding factors for US residential structures.

Science.gov (United States)

Dickson, Elijah; Hamby, David; Eckerman, Keith

2017-10-10

This paper presents validated building shielding factors designed for contemporary US housing-stock under an idealized, yet realistic, exposure scenario from contaminant deposition on the roof and surrounding surfaces. The building shielding factors are intended for use in emergency planning and level three probabilistic risk assessments for a variety of postulated radiological events in which a realistic assessment is necessary to better understand the potential risks for accident mitigation and emergency response planning. Factors are calculated from detailed computational housing-units models using the general-purpose Monte Carlo N-Particle computational code, MCNP5, and are benchmarked from a series of narrow- and broad-beam measurements analyzing the shielding effectiveness of ten common general-purpose construction materials and ten shielding models representing the primary weather barriers (walls and roofs) of likely US housing-stock. Each model was designed to scale based on common residential construction practices and include, to the extent practical, all structurally significant components important for shielding against ionizing radiation. Calculations were performed for floor-specific locations from contaminant deposition on the roof and surrounding ground as well as for computing a weighted-average representative building shielding factor for single- and multi-story detached homes, both with and without basement as well for single-wide manufactured housing-unit. © 2017 IOP Publishing Ltd.

Contaminant deposition building shielding factors for US residential structures

International Nuclear Information System (INIS)

Dickson, E D; Hamby, D M; Eckerman, K F

2015-01-01

This paper presents validated building shielding factors designed for contemporary US housing-stock under an idealized, yet realistic, exposure scenario from contaminant deposition on the roof and surrounding surfaces. The building shielding factors are intended for use in emergency planning and level three probabilistic risk assessments for a variety of postulated radiological events in which a realistic assessment is necessary to better understand the potential risks for accident mitigation and emergency response planning. Factors are calculated from detailed computational housing-units models using the general-purpose Monte Carlo N-Particle computational code, MCNP5, and are benchmarked from a series of narrow- and broad-beam measurements analyzing the shielding effectiveness of ten common general-purpose construction materials and ten shielding models representing the primary weather barriers (walls and roofs) of likely US housing-stock. Each model was designed to scale based on common residential construction practices and include, to the extent practical, all structurally significant components important for shielding against ionizing radiation. Calculations were performed for floor-specific locations from contaminant deposition on the roof and surrounding ground as well as for computing a weighted-average representative building shielding factor for single- and multi-story detached homes, both with and without basement as well for single-wide manufactured housing-unit. (paper)

Cloud immersion building shielding factors for US residential structures

International Nuclear Information System (INIS)

Dickson, E D; Hamby, D M

2014-01-01

This paper presents validated building shielding factors designed for contemporary US housing-stock under an idealized, yet realistic, exposure scenario within a semi-infinite cloud of radioactive material. The building shielding factors are intended for use in emergency planning and level three probabilistic risk assessments for a variety of postulated radiological events in which a realistic assessment is necessary to better understand the potential risks for accident mitigation and emergency response planning. Factors are calculated from detailed computational housing-units models using the general-purpose Monte Carlo N-Particle computational code, MCNP5, and are benchmarked from a series of narrow- and broad-beam measurements analyzing the shielding effectiveness of ten common general-purpose construction materials and ten shielding models representing the primary weather barriers (walls and roofs) of likely US housing-stock. Each model was designed to scale based on common residential construction practices and include, to the extent practical, all structurally significant components important for shielding against ionizing radiation. Calculations were performed for floor-specific locations as well as for computing a weighted-average representative building shielding factor for single- and multi-story detached homes, both with and without basement, as well for single-wide manufactured housing-units. (paper)

Application of the personnel photographic monitoring method to determine equivalent radiation dose beyond proton accelerator shielding

International Nuclear Information System (INIS)

Gel'fand, E.K.; Komochkov, M.M.; Man'ko, B.V.; Salatskaya, M.I.; Sychev, B.S.

1980-01-01

Calculations of regularities to form radiation dose beyond proton accelerator shielding are carried out. Numerical data on photographic monitoring dosemeter in radiation fields investigated are obtained. It was shown how to determine the total equivalent dose of radiation fields beyond proton accelerator shielding by means of the photographic monitoring method by introduction into the procedure of considering nuclear emulsions of division of particle tracks into the black and grey ones. A comparison of experimental and calculational data has shown the applicability of the used calculation method for modelling dose radiation characteristics beyond proton accelerator shielding [ru

Summary - COG: A new point-wise Monte Carlo code for burnup credit analysis

International Nuclear Information System (INIS)

Alesso, H.P.

1989-01-01

COG, a new point-wise Monte Carlo code being developed and tested at Lawrence Livermore National Laboratory (LLNL) for the Cray-1, solves the Boltzmann equation for the transport of neutrons, photons, and (in future versions) other particles. Techniques included in the code for modifying the random walk of particles make COG most suitable for solving deep-penetration (shielding) problems and a wide variety of criticality problems. COG is similar to a number of other computer codes used in the shielding community. Each code is a little different in its geometry input and its random-walk modification options. COG is a Monte Carlo code specifically designed for the CRAY (in 1986) to be as precise as the current state of physics knowledge. It has been extensively benchmarked and used as a shielding code at LLNL since 1986, and has recently been extended to accomplish criticality calculations. It will make an excellent tool for future shipping cask studies

Evaluation of radiation shielding performance in sea transport of radioactive material by using simple calculation method

International Nuclear Information System (INIS)

Odano, N.; Ohnishi, S.; Sawamura, H.; Tanaka, Y.; Nishimura, K.

2004-01-01

A modified code system based on the point kernel method was developed to use in evaluation of shielding performance for maritime transport of radioactive material. For evaluation of shielding performance accurately in the case of accident, it is required to preciously model the structure of transport casks and shipping vessel, and source term. To achieve accurate modelling of the geometry and source term condition, we aimed to develop the code system by using equivalent information regarding structure and source term used in the Monte Carlo calculation code, MCNP. Therefore, adding an option to use point kernel method to the existing Monte Carlo code, MCNP4C, the code system was developed. To verify the developed code system, dose rate distribution in an exclusive shipping vessel to transport the low level radioactive wastes were calculated by the developed code and the calculated results were compared with measurements and Monte Carlo calculations. It was confirmed that the developed simple calculation method can obtain calculation results very quickly with enough accuracy comparing with the Monte Carlo calculation code MCNP4C

Radially and azimuthally dependent resonance self-shielding treatment for general multi-region geometry based on a unified theory

International Nuclear Information System (INIS)

Koike, Hiroki; Kirimura, Kazuki; Yamaji, Kazuya; Kosaka, Shinya; Yamamoto, Akio

2018-01-01

A unified resonance self-shielding method, which can treat general sub-divided fuel regions, is developed for lattice physics calculations in reactor physics field. In a past study, a hybrid resonance treatment has been developed by theoretically integrating equivalence theory and ultra-fine-group slowing-down calculation. It can be applied to a wide range of neutron spectrum conditions including low moderator density ranges in severe accident states, as long as each fuel region is not sub-divided. In order to extend the method for radially and azimuthally sub-divided multi-region geometry, a new resonance treatment is established by incorporating the essence of sub-group method. The present method is composed of two-step flux calculation, i.e. 'coarse geometry + fine energy' (first step) and 'fine geometry + coarse energy' (second step) calculations. The first step corresponds to a hybrid model of the equivalence theory and the ultra-fine-group calculation, and the second step corresponds to the sub-group method. From the verification results, effective cross-sections by the new method show good agreement with the continuous energy Monte-Carlo results for various multi-region geometries including non-uniform fuel compositions and temperature distributions. The present method can accurately generate effective cross-sections with short computation time in general lattice physics calculations. (author)

A Shielding Analysis of Hot Cell for a 10 MW Research Reactor

International Nuclear Information System (INIS)

Alnajjar, Alaaddin; Park, Chang Je; Roh, Gyuhong; Lee, Byunchul

2013-01-01

In this paper, a shielding analysis has been performed for the hot cell in a 10 MW research reactor. Two kinds of shielding analysis code systems are used such as MCNPX2.7 and M-Shield7. The first one is Monte Carlo stochastic code and the second one is a deterministic point kernel code. The results are compared in this study. In order to obtain source term, the ORIGEN-S code is used for different kinds of source. Four kinds of sources are taken into consideration. From the simulation, it is also proposed that the proper thickness of shielding material and the maximum source capacity in the hot cell. This study shows preliminary analysis results of hot cell shielding for 10MW research reactor. Total four different source terms are considered such as spent fuel assembly, Ir-192, Mo-99, and I-131. For shielding material, general concrete, heavy concrete, and lead are used. MCNPX code is mainly used for a simplified hot cell model and the result are nearly consistent when compared with M-Shield code. Required shielding thickness and the hot cell capacity are also obtained for various criterion of surface dose rates

Radiation shielding lead shield

International Nuclear Information System (INIS)

Dei, Shoichi.

1991-01-01

The present invention concerns lead shields for radiation shielding. Shield boxes are disposed so as to surround a pipeline through which radioactive liquids, mists or like other objects are passed. Flanges are formed to each of the end edges of the shield boxes and the shield boxes are connected to each other by the flanges. Upon installation, empty shield boxes not charged with lead particles and iron plate shields are secured at first at the periphery of the pipeline. Then, lead particles are charged into the shield boxes. This attains a state as if lead plate corresponding to the depth of the box is disposed. Accordingly, operations for installation, dismantling and restoration can be conducted in an empty state with reduced weight to facilitate the operations. (I.S.)

Application of a dummy eye shield for electron treatment planning

International Nuclear Information System (INIS)

Kang, Sei-Kwon; Park, Soah; Hwang, Taejin; Cheong, Kwang-Ho; Han, Taejin; Kim, Haeyoung; Lee, Me-Yeon; Kim, Kyoung Ju; Oh, Do Hoon; Bae, Hoonsik

2013-01-01

Metallic eye shields have been widely used for near-eye treatments to protect critical regions, but have never been incorporated into treatment plans because of the unwanted appearance of the metal artifacts on CT images. The purpose of this work was to test the use of an acrylic dummy eye shield as a substitute for a metallic eye shield during CT scans. An acrylic dummy shield of the same size as the tungsten eye shield was machined and CT scanned. The BEAMnrc and the DOSXYZnrc were used for the Monte Carlo (MC) simulation, with the appropriate material information and density for the aluminum cover, steel knob and tungsten body of the eye shield. The Pinnacle adopting the Hogstrom electron pencil-beam algorithm was used for the one-port 6-MeV beam plan after delineation and density override of the metallic parts. The results were confirmed with the metal oxide semiconductor field effect transistor (MOSFET) detectors and the Gafchromic EBT2 film measurements. For both the maximum eyelid dose over the shield and the maximum dose under the shield, the MC results agreed with the EBT2 measurements within 1.7%. For the Pinnacle plan, the maximum dose under the shield agreed with the MC within 0.3%; however, the eyelid dose differed by -19.3%. The adoption of the acrylic dummy eye shield was successful for the treatment plan. However, the Pinnacle pencil-beam algorithm was not sufficient to predict the eyelid dose on the tungsten shield, and more accurate algorithms like MC should be considered for a treatment plan. (author)

Neutron shielding properties of boron-containing ore and epoxy composites

International Nuclear Information System (INIS)

Li Zhifu; Xue Xiangxin

2011-01-01

Using the boron-containing iron ore concentrate and boron-rich slag as studying object, the starting materials were got after the specific green ore containing boron dressing in China and blast furnace separation respectively. Monte-Carlo method was used to study the effect of the boron-containing iron ore concentrate and boron-rich slag and their composites with epoxy on the neutron shielding abilities. The reasons that affecting the shielding materials properties was discussed and the suitable proportioning of boron-containing ore to epoxy composites was confirmed; the 14.1 MeV fast neutron removal cross section and the total thermal neutron attenuation coefficient were obtained and compared with that of the common used concrete. The results show that the shielding property of 14.1 MeV fast neutron is mainly concerned with the low-Z elements in the shielding materials, the thermal neutron shielding ability is mainly concerned with boron concentrate in the composite, the attenuation of the accompany γ-ray photon is mainly concerned with the high atom number elements content in the ore and the density of the shielding material. The optimum Janume fractions of composites are in the range of 0.4-0.6 and the fast neutron shielding properties are similar to concrete while the thermal neutron shielding properties are higher than concrete. The composites are expected to be used as biological concrete shields crack injection and filling of the anomalous holes through the concrete shields around the radiation fields or directly to be prepared as shielding materials.(authors)

A comparative study of two digestion methods employed for the determination boron in ferroboron used as an advanced shielding material

International Nuclear Information System (INIS)

Kamble, Granthali S.; Manisha, V.; Venkatesh, K.

2015-01-01

Shielding of nuclear reactor core is an important requirement of fast reactors. An important objective of future Fast Breeder Reactors (FBRs) is to reduce the volume of shields. A large number of materials have been considered for use to reduce the neutron flux to acceptable levels. A shield material which brings down the energy of neutrons by elastic and inelastic scattering along with absorption will be more effective. Ferro boron is identified as one of the advanced shielding materials considered for use in future FBRs, planned to be constructed in India. Ferroboron is an economical and indigenously available material which qualifies as a promising shield material through literature survey and scoping calculations. Experiments have been conducted in KAMINI reactor to understand the effectiveness of prospective shield material Ferro-boron as an in-core shield material for future FBRs. The Ferro boron used in these experiments contained 11.8% and 15% of boron. Precise determination of boron content in these ferro boron samples is very important to determine its effectiveness as a shield material. In this work a comparative study was carried out to determine the boron content in ferro boron samples. In the first method the sample was treated with incremental amounts of nitric acid under reflux (to prevent rigorous reaction and volatalisation of boron). The solution was gradually heated and the solution was filtered through a Whatman Filter paper no. 41. The undissolved ferro boron residue collected in the filter paper after filtration, is transferred to a platinum crucible; mixed with sodium carbonate and is ashed. The crucible is placed over a burner for 1 h to fuse the contents. The fused mass is leached in dilute hydrochloric acid, added to the nitric acid filtrate and made up to pre-determined volume

Shielding design study for the JAERI/KEK spallation neutron source

International Nuclear Information System (INIS)

Maekawa, Fujio; Teshigawara, Makoto; Konno, Chikara; Ikeda, Yujiro; Watanabe, Noboru

2001-01-01

Shielding design for the JAERI/KEK spallation neutron source was studied. Bulk shielding characteristics and optimization of a beam shutter were investigated by using Monte Carlo calculation code NMTC/JAM and MCNP with LA-150 neutron cross-section library. The following remarks were derived. (1) Neutron dose outside of the concrete shield at 6.6 m from the center is ∼10 μSv/hr regardless of angles with respect to the proton beam axis. The neutron dose can be reduced more than a factor of 30 by adding natural boron of 5 wt% in the concrete. (2) When a beam shutter position just outside the void vessel and the shutter length of 2 m are assumed, a shutter made of copper (1.7 m) with polyethylene (0.3 m) is the optimum in terms of shielding performance as well as cost merit. A shutter made of tungsten is not so effective. (3) Further studies are needed for optimization of beam shutter position. (author)

Determination of material and its thickness for Cs-137 gamma source shielding

International Nuclear Information System (INIS)

Tukiman

2008-01-01

Its has been determined the shielding material and its thickness necessarily conducted due to every material will have different half-thickness characteristics, and by the selection a suitable material and its thickness will be obtained. Half-thickness of any material is the ability of the material at a certain thickness to absorb any radiation intensity so that the intensity becomes half of its source. Sample materials to be used are concrete, wood, and lead with their thickness varied. From experiment data and theoretical computation can be concluded that lead is the suitable material for shielding with the value of HVT for gamma radiation 0,732 cm. For wood and concrete will give half-thickness of 11,0 cm and 3,164 cm respectively. (author)

Superhydrophobic coatings on wood substrate for self-cleaning and EMI shielding

Science.gov (United States)

Xing, Yingjie; Xue, Yaping; Song, Jinlong; Sun, Yankui; Huang, Liu; Liu, Xin; Sun, Jing

2018-04-01

A layer of superhydrophobic coating having good electromagnetic shielding and self-cleaning performance was fabricated on a wood surface through an electroless copper plated process. The superhydrophobic property of the wood surface was measured by contact angle (CA) and roll-off angle (RA) measurements. The microstructure and chemical composition of the superhydrophobic coating were analyzed by scanning electron microscopy (SEM), energy dispersive spectrometer (EDS) and X-ray diffraction (XRD). The analysis revealed that the microscale particles were uniformly distributed on the wood surface and the main component of the coating is metallic copper. The as-prepared Cu coatings on wood substrate exhibit a good superhydrophobicity with water contact angle about 160° and rolling angle less than 5°.

The new solid target system at UNAM in a self-shielded 11 MeV cyclotron

International Nuclear Information System (INIS)

Zarate-Morales, A.; Gaspar-Carcamo, R. E.; Lopez-Rodriguez, V.; Flores-Moreno, A.; Trejo-Ballado, F.; Avila-Rodriguez, Miguel A.

2012-01-01

A dual beam line (BL) self-shielded RDS 111 cyclotron for radionuclide production was installed at the School of Medicine of the National Autonomous University of Mexico in 2001. One of the BL’s was upgraded to Eclipse HP (Siemens) in 2008 and the second BL was recently upgraded (June 2011) to the same version with the option for the irradiation of solid targets for the production of metallic radioisotopes.

Electron, electron-bremsstrahlung and proton depth-dose data for space-shielding applications

Science.gov (United States)

Seltzer, S. M.

1979-01-01

A data set has been developed, consisting of depth-dose distributions for omni-directional electron and proton fluxes incident on aluminum shields. The principal new feature of this work is the accurate treatment, based on detailed Monte Carlo calculations, of the electron-produced bremsstrahlung component. Results covering the energy region of interest in space-shielding calculations have been obtained for the absorbed dose (a) as a function of depth in a semi-infinite medium, (b) at the edge of slab shields, and (c) at the center of a solid sphere. The dose to a thin tissue-equivalent detector was obtained as well as that in aluminum. Various results and comparisons with other work are given.

Technology development for radiation shielding analysis

International Nuclear Information System (INIS)

Ha, Jung Woo; Lee, Jae Kee; Kim, Jong Kyung

1986-12-01

Radiation shielding analysis in nuclear engineering fields is an important technology which is needed for the calculation of reactor shielding as well as radiation related safety problems in nuclear facilities. Moreover, the design technology required in high level radioactive waste management and disposal facilities is faced on serious problems with rapidly glowing nuclear industry development, and more advanced technology has to be developed for tomorrow. The main purpose of this study is therefore to build up the self supporting ability of technology development for the radiation shielding analysis in order to achieve successive development of nuclear industry. It is concluded that basic shielding calculations are possible to handle and analyze by using our current technology, but more advanced technology is still needed and has to be learned for the degree of accuracy in two-dimensional shielding calculation. (Author)

A practical neutron shielding design based on data-base interpolation

International Nuclear Information System (INIS)

Jiang, S.H.; Sheu, R.J.

1993-01-01

Neutron shielding design is an important part of the construction of nuclear reactors and high-energy accelerators. Neutron shielding design is also indispensable in the packaging and storage of isotopic neutron sources. Most efforts in the development of neutron shielding design have been concentrated on nuclear reactor shielding because of its huge mass and strict requirement of accuracy. Sophisticated computational tools, such as transport and Monte Carlo codes and detailed data libraries have been developed. In principle, now, neutron shielding, in spite of its complexity, can be designed in any detail and with fine accuracy. However, in most practical cases, neutron shielding design is accomplished with simplified methods. Unlike practical gamma-ray shielding design, where exponential attenuation coupled with buildup factors has been applied effectively and accurately, simplified neutron shielding design, either by using removal cross sections or by applying charts or tables of transmission factors such as the National Council on Radiation Protection and Measurements (NCRP) 38 (Ref. 1) for general neutron protection or to NCRP 51 (Ref. 2) for accelerator neutron shielding, is still very primitive and not well established. The available data are limited in energy range, materials, and thicknesses, and the estimated results are only roughly accurate. It is the purpose of this work to establish a simple, convenient, and user-friendly general-purpose computational tool for practical preliminary neutron shielding design that is reasonably accurate. A wide-range (energy, material, and thickness) data base of dose transmission factors has been generated by applying one-dimensional transport calculations in slab geometry

Comparison and physical interpretation of MCNP and TART neutron and γ Monte Carlo shielding calculations for a heavy-ion ICF system

International Nuclear Information System (INIS)

Mainardi, E.; Premuda, F.; Lee, E.

2004-01-01

Inertial confinement fusion (ICF) aims to induce implosions of D-T pellets to obtain a extremely dense and hot plasma with lasers or heavy-ion beams. For heavy-ion fusion (HIF), recent research has focused on 'liquid-protected' designs that allow highly compact target chambers. In the design of a reactor such as HYLIFE-II [Fus. Techol. 25 (1984); HYLIFE-II Progress Report, UCID-21816, 4.82-100], the liquid used is a molten salt made of F 10 , Li 6 , Li 7 , Be 9 (called flibe). Flibe allows the final-focus magnets to be closer to the target, which helps to reduce the focus spot size and in turn the size of the driver, with a large reduction of the cost of HIF electricity. Consequently the superconducting coils of the magnets closer to the D-T neutron source will potentially suffer higher damage though they can stand only a certain amount of energy deposited before quenching. This work has been primarily focusing on verifying that total energy deposited by fusion neutrons and induced γ rays remain under such limit values and the final purpose is the optimization of the shielding of the magnetic lens system from the points of view of the geometrical configuration and of the physical nature of the materials adopted. The system is analyzed in terms of six geometrical models going from simplified up to much more realistic representations of a system of 192 beam lines, each focused by six magnets. A 3-D transport calculation of the radiation penetrating through ducts, that takes into account the complexity of the system, requires Monte Carlo methods. The technical nature of the design problem and the methodology followed were presented in a previous paper [Nucl. Instr. and Meth. A 464 (2001) 410] by summarizing briefly the results for the deposited energy distribution on the six focal magnets of a beam line. Now a comparison of the performances of the two codes TART98 [TART98: A Coupled Neutron-Photon 3-D Combinational Geometry Monte Carlo Transport Code, Lawrence

Shielding performance of the NET vacuum vessel

International Nuclear Information System (INIS)

Arkuszewski, J.J.; Jaeger, J.F.

1988-01-01

To corroborate 1-D deterministic shielding calculations on the Next European Torus (NET) vacuum vessel/shield and shielding blanket, 3-D Monte Carlo calculations have been done with the MCNP code. This should provide information on the poloidal and the toroidal variations. Plasma source simulation and the geometrical model are described, as are other assumptions. The calculations are based on the extended plasma power of 714 MW. The results reported here are the heat deposition in various parts of the device, on the one hand, and the neutron and photon currents at the outer boundary of the vacuum vessel, on the other hand. The latter are needed for the detailed design of the super-conducting magnetic coils. A reasonable statistics has been obtained on the outboard side of the torus, though this cannot be said for the inboard side. The inboard is, however, much more toroidally symmetric than the outboard, so that other methods could be applied such as 2-D deterministic calculations, for instance. (author) 4 refs., 44 figs., 42 tabs

Development of three-dimensional program based on Monte Carlo and discrete ordinates bidirectional coupling method

International Nuclear Information System (INIS)

Han Jingru; Chen Yixue; Yuan Longjun

2013-01-01

The Monte Carlo (MC) and discrete ordinates (SN) are the commonly used methods in the design of radiation shielding. Monte Carlo method is able to treat the geometry exactly, but time-consuming in dealing with the deep penetration problem. The discrete ordinate method has great computational efficiency, but it is quite costly in computer memory and it suffers from ray effect. Single discrete ordinates method or single Monte Carlo method has limitation in shielding calculation for large complex nuclear facilities. In order to solve the problem, the Monte Carlo and discrete ordinates bidirectional coupling method is developed. The bidirectional coupling method is implemented in the interface program to transfer the particle probability distribution of MC and angular flux of discrete ordinates. The coupling method combines the advantages of MC and SN. The test problems of cartesian and cylindrical coordinate have been calculated by the coupling methods. The calculation results are performed with comparison to MCNP and TORT and satisfactory agreements are obtained. The correctness of the program is proved. (authors)

MARS14 deep-penetration calculation for the ISIS target station shielding

International Nuclear Information System (INIS)

Nakao, Noriaki; Nunomiya, Tomoya; Iwase, Hiroshi; Nakamura, Takashi

2004-01-01

The calculation of neutron penetration through a thick shield was performed with a three-dimensional multi-layer technique using the MARS14(02) Monte Carlo code to compare with the experimental shielding data in 1998 at the ISIS spallation neutron source facility of Rutherford Appleton Laboratory. In this calculation, secondary particles from a tantalum target bombarded by 800-MeV protons were transmitted through a bulk shield of approximately 3-m-thick iron and 1-m-thick concrete. To accomplish this deep-penetration calculation, a three-dimensional multi-layer technique and energy cut-off method were used considering a spatial statistical balance. Finally, the energy spectra of neutrons behind the very thick shield could be calculated down to the thermal energy with good statistics, and the calculated results typically agree well within a factor of two with the experimental data over a broad energy range. The 12 C(n,2n) 11 C reaction rates behind the bulk shield were also calculated, which agree with the experimental data typically within 60%. These results are quite impressive in calculation accuracy for deep-penetration problem

Gamma dose from activation of internal shields in IRIS reactor.

Science.gov (United States)

Agosteo, Stefano; Cammi, Antonio; Garlati, Luisella; Lombardi, Carlo; Padovani, Enrico

2005-01-01

The International Reactor Innovative and Secure is a modular pressurised water reactor with an integral design. This means that all the primary system components, such as the steam generators, pumps, pressuriser and control rod drive mechanisms, are located inside the reactor vessel, which requires a large diameter. For the sake of better reliability and safety, it is desirable to achieve the reduction of vessel embrittlement as well as the lowering of the dose beyond the vessel. The former can be easily accomplished by the presence of a wide downcomer, filled with water, which surrounds the core region, while the latter needs the presence of additional internal shields. An optimal shielding configuration is under investigation, for reducing the ex-vessel dose due to activated internals and for limiting the amount of the biological shielding. MCNP 4C calculations were performed to evaluate the neutron and the gamma dose during operation and the 60Co activation of various shields configurations. The gamma dose beyond the vessel from activation of its structural components was estimated in a shutdown condition, with the Monte Carlo code FLUKA 2002 and the MicroShield software. The results of the two codes are in agreement and show that the dose is sufficiently low, even without an additional shield.

Gamma dose from activation of internal shields in IRIS reactor

International Nuclear Information System (INIS)

Agosteo, S.; Cammi, A.; Garlati, L.; Lombardi, C.; Padovani, E.

2005-01-01

The International Reactor Innovative and Secure is a modular pressurised water reactor with an integral design. This means that all the primary system components, such as the steam generators, pumps, pressurizer and control rod drive mechanisms, are located inside the reactor vessel, which requires a large diameter. For the sake of better reliability and safety, it is desirable to achieve the reduction of vessel embrittlement as well as the lowering of the dose beyond the vessel. The former can be easily accomplished by the presence of a wide downcomer, filled with water, which surrounds the core region, while the latter needs the presence of additional internal shields. An optimal shielding configuration is under investigation, for reducing the ex-vessel dose due to activated internals and for limiting the amount of the biological shielding. MCNP 4C calculations were performed to evaluate the neutron and the gamma dose during operation and the 60 Co activation of various shields configurations. The gamma dose beyond the vessel from activation of its structural components was estimated in a shutdown condition, with the Monte Carlo code FLUKA 2002 and the MicroShield software. The results of the two codes are in agreement and show that the dose is sufficiently low, even without an additional shield. (authors)

Shielding design for positron emission tomography facility

International Nuclear Information System (INIS)

Abdallah, I.I.

2007-01-01

With the recent advent of readily available tracer isotopes, there has been marked increase in the number of hospital-based and free-standing positron emission tomography (PET) clinics. PET facilities employ relatively large activities of high-energy photon emitting isotopes, which can be dangerous to the health of humans and animals. This coupled with the current dose limits for radiation worker and members of the public can result in shielding requirements. This research contributes to the calculation of the appropriate shielding to keep the level of radiation within an acceptable recommended limit. Two different methods were used including measurements made at selected points of an operating PET facility and computer simulations by using Monte Carlo Transport Code. The measurements mainly concerned the radiation exposure at different points around facility using the survey meter detectors and Thermoluminescent Dosimeters (TLD). Then the set of manual calculation procedures were used to estimate the shielding requirements for a newly built PEF facility. The results from the measurement and the computer simulation were compared to the results obtained from the set manual calculation procedure. In general, the estimated weekly dose at the points of interest is lower than the regulatory limits for the little company of Mary Hospital. Furthermore, the density and the HVL for normal strength concrete and clay bricks are almost similar. In conclusion, PET facilities present somewhat different design requirements and are more likely to require additional radiation shielding. Therefore, existing shields at the little Company of Mary Hospital are in general found to be adequate and satisfactory and additional shielding was found necessary at the new PET facility in the department of Nuclear Medicine of the Dr. George Mukhari Hospital. By use of appropriate design, by implying specific shielding requirements and by maintaining good operating practices, radiation doses to

Two-dimensional radiation shielding optimization analysis of spent fuel transport container

International Nuclear Information System (INIS)

Tian Yingnan; Chen Yixue; Yang Shouhai

2013-01-01

The intelligent radiation shielding optimization design software platform is a one-dimensional multi-target radiation shielding optimization program which is developed on the basis of the genetic algorithm program and one-dimensional discrete ordinate program-ANISN. This program was applied in the optimization design analysis of the spent fuel transport container radiation shielding. The multi-objective optimization calculation model of the spent fuel transport container radiation shielding was established, and the optimization calculation of the spent fuel transport container weight and radiation dose rate was carried by this program. The calculation results were checked by Monte-Carlo program-MCNP/4C. The results show that the weight of the optimized spent fuel transport container decreases to 81.1% of the origin and the radiation dose rate decreases to below 65.4% of the origin. The maximum deviation between the calculated values from the program and the MCNP is below 5%. The results show that the optimization design scheme is feasible and the calculation result is correct. (authors)

Monte-Carlo validation of secondary sodium activation in a pool-type LMFBR

International Nuclear Information System (INIS)

Plamiotti, G.; Rado, V.; Salvatores, M.

1980-09-01

The secondary sodium activation in a pool-type LMFBR is the main parameter to be accurately evaluated in the shield design. In the present work a complete two dimensional description of the system, including core, shielding and sodium up to Heat Exchangers, is coupled to local Heat Exchanger Monte-Carlo calculations. This refined calculation is used to deduce a simplified method to take into account the coupling of radial propagation in the Heat Exchanger and its finite cylindrical structure

Markov chain Monte Carlo techniques applied to parton distribution functions determination: Proof of concept

Science.gov (United States)

Gbedo, Yémalin Gabin; Mangin-Brinet, Mariane

2017-07-01

We present a new procedure to determine parton distribution functions (PDFs), based on Markov chain Monte Carlo (MCMC) methods. The aim of this paper is to show that we can replace the standard χ2 minimization by procedures grounded on statistical methods, and on Bayesian inference in particular, thus offering additional insight into the rich field of PDFs determination. After a basic introduction to these techniques, we introduce the algorithm we have chosen to implement—namely Hybrid (or Hamiltonian) Monte Carlo. This algorithm, initially developed for Lattice QCD, turns out to be very interesting when applied to PDFs determination by global analyses; we show that it allows us to circumvent the difficulties due to the high dimensionality of the problem, in particular concerning the acceptance. A first feasibility study is performed and presented, which indicates that Markov chain Monte Carlo can successfully be applied to the extraction of PDFs and of their uncertainties.

Advanced methodologies of evaluating the radiation sources and ionising radiation shieldings for reducing the irradiation in nuclear field personnel

International Nuclear Information System (INIS)

Pantazi, D.; Mateescu, S.; Stanciu, M.

2003-01-01

One of the technical measures of protection against ionizing radiations is the radiation shielding. The process of implementing modern and efficient methods of evaluating the radiation shielding implies advanced calculation methods. That means using from simpler 1-D or 2-D computing codes such as MicroShield or QAD up to systems of codes such as SCALE (containing several independent modules) or the Monte Carlo multipurpose and many particles, MCNP, transport code. The main objective of this work is to present the Monte Carlo based evaluation of the dose rates from the CANDU type spent fuel all along the path of its handling up to intermediate storage. These values will be then compared with the values obtained from calculations with different computing programs. To obtain this objective two problems were approached: - establishing geometrical models according to the definition used by MCNP code so that the characteristics of CANDU type nuclear fuel are taking into account; - checking the validity of the proposed models by comparing the MCNP results with those obtained with other computing codes specific for shielding evaluation and radiation dose calculation

Problems of the power plant shield optimization

International Nuclear Information System (INIS)

Abagyan, A.A.; Dubinin, A.A.; Zhuravlev, V.I.; Kurachenko, Yu.A.; Petrov, Eh.E.

1981-01-01

General approaches to the solution of problems on the nuclear power plant radiation shield optimization are considered. The requirements to the shield parameters are formulated in a form of restrictions on a number of functionals, determined by the solution of γ quantum and neutron transport equations or dimensional and weight characteristics of shield components. Functional determined by weight-dimensional parameters (shield cost, mass and thickness) and functionals, determined by radiation fields (equivalent dose rate, produced by neutrons and γ quanta, activation functional, radiation functional, heat flux, integral heat flux in a particular part of the shield volume, total energy flux through a particular shield surface are considered. The following methods of numerical solution of simplified optimization problems are discussed: semiempirical methods using radiation transport physical leaks, numerical solution of approximate transport equations, numerical solution of transport equations for the simplest configurations making possible to decrease essentially a number of variables in the problem. The conclusion is drawn that the attained level of investigations on the problem of nuclear power plant shield optimization gives the possibility to pass on at present to the solution of problems with a more detailed account of the real shield operating conditions (shield temperature field account, its strength and other characteristics) [ru

Proposal of a Technical Design of the ATLAS Forward Region Shielding

CERN Document Server

Leroy, C; Palla, J; Pospísil, S; Puchmajer, P; Sodomka, J; Stekl, I

2002-01-01

The aim of this note is to present a technical design of the ATLAS forward region. The concept is based on segmented shielding supported by the results of an experiment performed at CERN - PS and Monte Carlo simulations extending these results to ATLAS situation. This concept is translated into a practical engineering design.

Monte Carlo approach to define the refrigerator capacities for JT-60SA

International Nuclear Information System (INIS)

Wanner, Manfred; Barabaschi, Pietro; Lamaison, Valerie; Michel, Frederic; Reynaud, Pascal; Roussel, Pascal

2011-01-01

The JT-60SA cryogenic system shall provide refrigeration to keep the superconducting magnets and their structures at 4.4 K, cryo-pumps at 3.7 K, thermal shields at 80-100 K, and deliver a flow of 50 K helium to the current leads. A Monte Carlo method is proposed to determine the capacity contingencies for the refrigeration system. Attributing individual contingencies and distribution probability functions to the design variables allows the different load contributions to be statistically averaged. The total refrigeration contingency is derived for each temperature level from the 95% confidence level of the integrated distribution function.

Burnup Estimation of Rhodium Self-Powered Neutron Detector Emitter in VVER Reactor Core Using Monte Carlo Simulations

OpenAIRE

Khrutchinsky, А. А.; Kuten, S. A.; Babichev, L. F.

2011-01-01

Estimation of burn-up in a rhodium-103 emitter of self-powered neutron detector in VVER-1000 reactor core has been performed using Monte Carlo simulations within approximation of a constant neutron flux.

Considerations on scattering and leak radiation for effective determination of secondary shielding in X-rays rooms of megavoltage

International Nuclear Information System (INIS)

Borges, Diogo da S.; Lava, Deise D.; Affonso, Renato R.W.; Moreira, Maria de L.; Guimaraes, Antonio C.F.

2014-01-01

This paper addresses the development of a algorithm capable of analyzing the thickness of the secondary shielding due to the production of secondary beams. The production of this beam requires consideration of scattering angle, as well as factors normally used for screening of medical facilities using radiographic techniques. Besides the beam emanated from scattering radiation, is is necessary to evaluate the contribution of leakage radiation, originating from equipment used for the production of the primary beam. A view of the mutual contribution of these radiation to the formation of the secondary beam has shown the need of using shieldings in adjacent walls of the room. The code was validated by comparison with an example case provided by NCRP-151 Report. In this report calculations for determining the secondary barrier for small angles are presented, that deserves greater attention for shielding and statements related to radiotherapy procedures of Modulated intensity. The results are consistent with those provided in the report, which makes the code can be used as a practical tool for the determination of effective shielding beams of megavoltage X-rays

Shielding Benchmark Computational Analysis

International Nuclear Information System (INIS)

Hunter, H.T.; Slater, C.O.; Holland, L.B.; Tracz, G.; Marshall, W.J.; Parsons, J.L.

2000-01-01

Over the past several decades, nuclear science has relied on experimental research to verify and validate information about shielding nuclear radiation for a variety of applications. These benchmarks are compared with results from computer code models and are useful for the development of more accurate cross-section libraries, computer code development of radiation transport modeling, and building accurate tests for miniature shielding mockups of new nuclear facilities. When documenting measurements, one must describe many parts of the experimental results to allow a complete computational analysis. Both old and new benchmark experiments, by any definition, must provide a sound basis for modeling more complex geometries required for quality assurance and cost savings in nuclear project development. Benchmarks may involve one or many materials and thicknesses, types of sources, and measurement techniques. In this paper the benchmark experiments of varying complexity are chosen to study the transport properties of some popular materials and thicknesses. These were analyzed using three-dimensional (3-D) models and continuous energy libraries of MCNP4B2, a Monte Carlo code developed at Los Alamos National Laboratory, New Mexico. A shielding benchmark library provided the experimental data and allowed a wide range of choices for source, geometry, and measurement data. The experimental data had often been used in previous analyses by reputable groups such as the Cross Section Evaluation Working Group (CSEWG) and the Organization for Economic Cooperation and Development/Nuclear Energy Agency Nuclear Science Committee (OECD/NEANSC)

Shielding calculations by using the analytic methods : Application to the radio-isotopes production in the CENM reactor

International Nuclear Information System (INIS)

Elmorabit, A.; Labrim, H.

2010-01-01

Full text: this work is part of developing an analytical method for solving the neutrons transport equation in improving the treatment of the anisotropy of neutron scattering through heterogeneous shielding. We also develop the tools necessary for the formation of multigroup libraries (cross section) with the best choice of the weighting function. Among the radioprotection problems of radioisotopes production experiments in the research reactor core is mainly the photons gamma generation produced by radiative capture: activation of samples and their capsules. So, in order to review the safety of operating personnel and the public is essential to quantify the neutrons flux and gamma photons produced. In this study a numerical methods is used in two different Fortran program to solve the neutron transport problem and to determine the neutron and photon flux. This program based on the Monte Carlo method: the neutron is born with a unit statistical weight, this corrected after each imposed scattering event during its whole history within the shield. The final neutron statistical weight is used in an appropriate estimator to determine the searched response. The generated gamma rays by neutron capture are calculated of different isotopes, and then the equivalent dose rate is evaluated in biological tissue for different neutron source energies. We have identified and studied the choice of the best weighting function to calculate a library of multigroup cross sections self protected by using the energy weighting function. A Fortran program is used as a mathematical tool to solve the neutron slowing down equation in infinite homogeneous medium for different dilutions. We determined the energetic flux distribution and the effective integrals. The results of both calculations are in a good agreement; the relative error is less than 0.5%.

Monte Carlo benchmarking: Validation and progress

International Nuclear Information System (INIS)

Sala, P.

2010-01-01

Document available in abstract form only. Full text of publication follows: Calculational tools for radiation shielding at accelerators are faced with new challenges from the present and next generations of particle accelerators. All the details of particle production and transport play a role when dealing with huge power facilities, therapeutic ion beams, radioactive beams and so on. Besides the traditional calculations required for shielding, activation predictions have become an increasingly critical component. Comparison and benchmarking with experimental data is obviously mandatory in order to build up confidence in the computing tools, and to assess their reliability and limitations. Thin target particle production data are often the best tools for understanding the predictive power of individual interaction models and improving their performances. Complex benchmarks (e.g. thick target data, deep penetration, etc.) are invaluable in assessing the overall performances of calculational tools when all ingredients are put at work together. A review of the validation procedures of Monte Carlo tools will be presented with practical and real life examples. The interconnections among benchmarks, model development and impact on shielding calculations will be highlighted. (authors)

Use of Existing CAD Models for Radiation Shielding Analysis

Science.gov (United States)

Lee, K. T.; Barzilla, J. E.; Wilson, P.; Davis, A.; Zachman, J.

2015-01-01

The utility of a radiation exposure analysis depends not only on the accuracy of the underlying particle transport code, but also on the accuracy of the geometric representations of both the vehicle used as radiation shielding mass and the phantom representation of the human form. The current NASA/Space Radiation Analysis Group (SRAG) process to determine crew radiation exposure in a vehicle design incorporates both output from an analytic High Z and Energy Particle Transport (HZETRN) code and the properties (i.e., material thicknesses) of a previously processed drawing. This geometry pre-process can be time-consuming, and the results are less accurate than those determined using a Monte Carlo-based particle transport code. The current work aims to improve this process. Although several Monte Carlo programs (FLUKA, Geant4) are readily available, most use an internal geometry engine. The lack of an interface with the standard CAD formats used by the vehicle designers limits the ability of the user to communicate complex geometries. Translation of native CAD drawings into a format readable by these transport programs is time consuming and prone to error. The Direct Accelerated Geometry -United (DAGU) project is intended to provide an interface between the native vehicle or phantom CAD geometry and multiple particle transport codes to minimize problem setup, computing time and analysis error.

Radiation attenuation by lead and nonlead materials used in radiation shielding garments

International Nuclear Information System (INIS)

McCaffrey, J. P.; Shen, H.; Downton, B.; Mainegra-Hing, E.

2007-01-01

The attenuating properties of several types of lead (Pb)-based and non-Pb radiation shielding materials were studied and a correlation was made of radiation attenuation, materials properties, calculated spectra and ambient dose equivalent. Utilizing the well-characterized x-ray and gamma ray beams at the National Research Council of Canada, air kerma measurements were used to compare a variety of commercial and pre-commercial radiation shielding materials over mean energy ranges from 39 to 205 keV. The EGSnrc Monte Carlo user code cavity.cpp was extended to provide computed spectra for a variety of elements that have been used as a replacement for Pb in radiation shielding garments. Computed air kerma values were compared with experimental values and with the SRS-30 catalogue of diagnostic spectra available through the Institute of Physics and Engineering in Medicine Report 78. In addition to garment materials, measurements also included pure Pb sheets, allowing direct comparisons to the common industry standards of 0.25 and 0.5 mm 'lead equivalent'. The parameter 'lead equivalent' is misleading, since photon attenuation properties for all materials (including Pb) vary significantly over the energy spectrum, with the largest variations occurring in the diagnostic imaging range. Furthermore, air kerma measurements are typically made to determine attenuation properties without reference to the measures of biological damage such as ambient dose equivalent, which also vary significantly with air kerma over the diagnostic imaging energy range. A single material or combination cannot provide optimum shielding for all energy ranges. However, appropriate choice of materials for a particular energy range can offer significantly improved shielding per unit mass over traditional Pb-based materials

Monte Carlo calculation of ''skyshine'' neutron dose from ALS [Advanced Light Source

International Nuclear Information System (INIS)

Moin-Vasiri, M.

1990-06-01

This report discusses the following topics on ''skyshine'' neutron dose from ALS: Sources of radiation; ALS modeling for skyshine calculations; MORSE Monte-Carlo; Implementation of MORSE; Results of skyshine calculations from storage ring; and Comparison of MORSE shielding calculations

Determination of the neutron activation profile of core drill samples by gamma-ray spectrometry.

Science.gov (United States)

Gurau, D; Boden, S; Sima, O; Stanga, D

2018-04-01

This paper provides guidance for determining the neutron activation profile of core drill samples taken from the biological shield of nuclear reactors using gamma spectrometry measurements. Thus, it provides guidance for selecting a model of the right form to fit data and using least squares methods for model fitting. The activity profiles of two core samples taken from the biological shield of a nuclear reactor were determined. The effective activation depth and the total activity of core samples along with their uncertainties were computed by Monte Carlo simulation. Copyright © 2017 Elsevier Ltd. All rights reserved.

A contribution Monte Carlo method

International Nuclear Information System (INIS)

Aboughantous, C.H.

1994-01-01

A Contribution Monte Carlo method is developed and successfully applied to a sample deep-penetration shielding problem. The random walk is simulated in most of its parts as in conventional Monte Carlo methods. The probability density functions (pdf's) are expressed in terms of spherical harmonics and are continuous functions in direction cosine and azimuthal angle variables as well as in position coordinates; the energy is discretized in the multigroup approximation. The transport pdf is an unusual exponential kernel strongly dependent on the incident and emergent directions and energies and on the position of the collision site. The method produces the same results obtained with the deterministic method with a very small standard deviation, with as little as 1,000 Contribution particles in both analog and nonabsorption biasing modes and with only a few minutes CPU time

Malakit: an innovative pilot project to self-diagnose and self-treat malaria among illegal gold miners in the Guiana Shield.

Science.gov (United States)

Douine, Maylis; Sanna, Alice; Galindo, Muriel; Musset, Lise; Pommier de Santi, Vincent; Marchesini, Paola; Magalhaes, Edgard Dias; Suarez-Mutis, Martha; Hiwat, Helene; Nacher, Mathieu; Vreden, Stephen; Garancher, Laure

2018-04-10

Illegal gold miners in French Guiana, a French overseas territory ('département') located in Amazonia, often carry malaria parasites (up to 46.8%). While the Guiana Shield Region aims at malaria elimination, the high prevalence of Plasmodium in this hard-to-reach population in conjunction with frequent incorrect use of artemisinin-based anti-malarials could favour the emergence of resistant parasites. Due to geographical and regulatory issues in French Guiana, usual malaria control strategies cannot be implemented in this particular context. Therefore, new strategies targeting this specific population in the forest are required. Numerous discussions among health institutions and scientific partners from French Guiana, Brazil and Suriname have led to an innovative project based on the distribution of kits for self-diagnosis and self-treatment of Plasmodium infections. The kit-distribution will be implemented at "resting sites", which are areas across the border of French Guiana regularly frequented by gold miners. The main objective is to increase the appropriate use and complete malaria treatment after a positive malaria diagnosis with a rapid test, which will be evaluated with before-and-after cross-sectional studies. Monitoring indicators will be collected from health mediators at the time of kit distribution and during subsequent visits, and from illegal gold miners themselves, through a smartphone application. The project funding is multisource, including Ministries of Health of the three countries, WHO/PAHO, and the European Union. This project will start in April 2018 as a 18 month pilot study led by the Clinical Investigation Centre of Cayenne. Results should be available at the end of 2019. This innovative approach may have several limitations which should be taken into account, as potential side effects, kit misuse or resale, declarative main criteria, or no Plasmodium vivax curative treatment. Close monitoring is thus needed. This project may be the

Resonance self-shielding effect in uncertainty quantification of fission reactor neutronics parameters

International Nuclear Information System (INIS)

Chiba, Go; Tsuji, Masashi; Narabayashi, Tadashi

2014-01-01

In order to properly quantify fission reactor neutronics parameter uncertainties, we have to use covariance data and sensitivity profiles consistently. In the present paper, we establish two consistent methodologies for uncertainty quantification: a self-shielded cross section-based consistent methodology and an infinitely-diluted cross section-based consistent methodology. With these methodologies and the covariance data of uranium-238 nuclear data given in JENDL-3.3, we quantify uncertainties of infinite neutron multiplication factors of light water reactor and fast reactor fuel cells. While an inconsistent methodology gives results which depend on the energy group structure of neutron flux and neutron-nuclide reaction cross section representation, both the consistent methodologies give fair results with no such dependences.

Self-learning kinetic Monte Carlo simulations of self-diffusion of small Ag islands on the Ag(111) surface

International Nuclear Information System (INIS)

Shah, Syed Islamuddin; Nandipati, Giridhar; Rahman, Talat S; Karim, Altaf

2016-01-01

We studied self-diffusion of small two-dimensional Ag islands, containing up to ten atoms, on the Ag(111) surface using self-learning kinetic Monte Carlo (SLKMC) simulations. Activation barriers are calculated using the semi-empirical embedded atom method (EAM) potential. We find that two- to seven-atom islands primarily diffuse via concerted translation processes with small contributions from multi-atom and single-atom processes, while eight- to ten-atom islands diffuse via single-atom processes, especially edge diffusion, corner rounding and kink detachment, along with a minimal contribution from concerted processes. For each island size, we give a detailed description of the important processes, and their activation barriers, responsible for its diffusion. (paper)

Self-shielding phenomenon modelling in multigroup transport code Apollo-2; Modelisation du phenomene d'autoprotection dans le code de transport multigroupe Apollo 2

Energy Technology Data Exchange (ETDEWEB)

Coste-Delclaux, M

2006-03-15

This document describes the improvements carried out for modelling the self-shielding phenomenon in the multigroup transport code APOLLO2. They concern the space and energy treatment of the slowing-down equation, the setting up of quadrature formulas to calculate reaction rates, the setting-up of a method that treats directly a resonant mixture and the development of a sub-group method. We validate these improvements either in an elementary or in a global way. Now, we obtain, more accurate multigroup reaction rates and we are able to carry out a reference self-shielding calculation on a very fine multigroup mesh. To end, we draw a conclusion and give some prospects on the remaining work. (author)

Research on Primary Shielding Calculation Source Generation Codes

Science.gov (United States)

Zheng, Zheng; Mei, Qiliang; Li, Hui; Shangguan, Danhua; Zhang, Guangchun

2017-09-01

Primary Shielding Calculation (PSC) plays an important role in reactor shielding design and analysis. In order to facilitate PSC, a source generation code is developed to generate cumulative distribution functions (CDF) for the source particle sample code of the J Monte Carlo Transport (JMCT) code, and a source particle sample code is deveoped to sample source particle directions, types, coordinates, energy and weights from the CDFs. A source generation code is developed to transform three dimensional (3D) power distributions in xyz geometry to source distributions in r θ z geometry for the J Discrete Ordinate Transport (JSNT) code. Validation on PSC model of Qinshan No.1 nuclear power plant (NPP), CAP1400 and CAP1700 reactors are performed. Numerical results show that the theoretical model and the codes are both correct.

Extended pattern recognition scheme for self-learning kinetic Monte Carlo simulations

International Nuclear Information System (INIS)

Shah, Syed Islamuddin; Nandipati, Giridhar; Kara, Abdelkader; Rahman, Talat S

2012-01-01

We report the development of a pattern recognition scheme that takes into account both fcc and hcp adsorption sites in performing self-learning kinetic Monte Carlo (SLKMC-II) simulations on the fcc(111) surface. In this scheme, the local environment of every under-coordinated atom in an island is uniquely identified by grouping fcc sites, hcp sites and top-layer substrate atoms around it into hexagonal rings. As the simulation progresses, all possible processes, including those such as shearing, reptation and concerted gliding, which may involve fcc-fcc, hcp-hcp and fcc-hcp moves are automatically found, and their energetics calculated on the fly. In this article we present the results of applying this new pattern recognition scheme to the self-diffusion of 9-atom islands (M 9 ) on M(111), where M = Cu, Ag or Ni.

Analysis of coupled neutron-gamma radiations, applied to shieldings in multigroup albedo method

International Nuclear Information System (INIS)

Dunley, Leonardo Souza

2002-01-01

The principal mathematical tools frequently available for calculations in Nuclear Engineering, including coupled neutron-gamma radiations shielding problems, involve the full Transport Theory or the Monte Carlo techniques. The Multigroup Albedo Method applied to shieldings is characterized by following the radiations through distinct layers of materials, allowing the determination of the neutron and gamma fractions reflected from, transmitted through and absorbed in the irradiated media when a neutronic stream hits the first layer of material, independently of flux calculations. Then, the method is a complementary tool of great didactic value due to its clarity and simplicity in solving neutron and/or gamma shielding problems. The outstanding results achieved in previous works motivated the elaboration and the development of this study that is presented in this dissertation. The radiation balance resulting from the incidence of a neutronic stream into a shielding composed by 'm' non-multiplying slab layers for neutrons was determined by the Albedo method, considering 'n' energy groups for neutrons and 'g' energy groups for gammas. It was taken into account there is no upscattering of neutrons and gammas. However, it was considered that neutrons from any energy groups are able to produce gammas of all energy groups. The ANISN code, for an angular quadrature order S 2 , was used as a standard for comparison of the results obtained by the Albedo method. So, it was necessary to choose an identical system configuration, both for ANISN and Albedo methods. This configuration was six neutron energy groups and eight gamma energy groups, using three slab layers (iron aluminum - manganese). The excellent results expressed in comparative tables show great agreement between the values determined by the deterministic code adopted as standard and, the values determined by the computational program created using the Albedo method and the algorithm developed for coupled neutron

Monte Carlo simulation: tool for the calibration in analytical determination of radionuclides

International Nuclear Information System (INIS)

Gonzalez, Jorge A. Carrazana; Ferrera, Eduardo A. Capote; Gomez, Isis M. Fernandez; Castro, Gloria V. Rodriguez; Ricardo, Niury Martinez

2013-01-01

This work shows how is established the traceability of the analytical determinations using this calibration method. Highlights the advantages offered by Monte Carlo simulation for the application of corrections by differences in chemical composition, density and height of the samples analyzed. Likewise, the results obtained by the LVRA in two exercises organized by the International Agency for Atomic Energy (IAEA) are presented. In these exercises (an intercomparison and a proficiency test) all reported analytical results were obtained based on calibrations in efficiency by Monte Carlo simulation using the DETEFF program

Neutron shielding material based on colemanite and epoxy resin

International Nuclear Information System (INIS)

Okuno, K.

2005-01-01

In recent years, there has been a need for compact shielding design such as self-shielding of a PET cyclotron or up-gradation of radiation machinery in existing facilities. In these cases, high performance shielding materials are needed. Concrete or polyethylene have been used for a neutron shield. However, for compact shielding, they fall short in terms of performance or durability. Therefore, a new type of neutron shielding material based on epoxy resin and colemanite has been developed. Slab attenuation experiments up to 40 cm for the new shielding material were carried out using a 252 Cf neutron source. Measurement was carried out using a REM-counter, and compared with calculation. The results show that the shielding performance is better than concrete and polyethylene mixed with 10 wt% boron oxide. From the result, we confirmed that the performance of the new material is suitable for practical use. (authors)

A Cross-Section Adjustment Method for Double Heterogeneity Problem in VHTGR Analysis

International Nuclear Information System (INIS)

Yun, Sung Hwan; Cho, Nam Zin

2011-01-01

Very High Temperature Gas-Cooled Reactors (VHTGRs) draw strong interest as candidates for a Gen-IV reactor concept, in which TRISO (tristructuralisotropic) fuel is employed to enhance the fuel performance. However, randomly dispersed TRISO fuel particles in a graphite matrix induce the so-called double heterogeneity problem. For design and analysis of such reactors with the double heterogeneity problem, the Monte Carlo method is widely used due to its complex geometry and continuous-energy capabilities. However, its huge computational burden, even in the modern high computing power, is still problematic to perform wholecore analysis in reactor design procedure. To address the double heterogeneity problem using conventional lattice codes, the RPT (Reactivityequivalent Physical Transformation) method considers a homogenized fuel region that is geometrically transformed to provide equivalent self-shielding effect. Another method is the coupled Monte Carlo/Collision Probability method, in which the absorption and nu-fission resonance cross-section libraries in the deterministic CPM3 lattice code are modified group-wise by the double heterogeneity factors determined by Monte Carlo results. In this paper, a new two-step Monte Carlo homogenization method is described as an alternative to those methods above. In the new method, a single cross-section adjustment factor is introduced to provide self-shielding effect equivalent to the self-shielding in heterogeneous geometry for a unit cell of compact fuel. Then, the homogenized fuel compact material with the equivalent cross-section adjustment factor is used in continuous-energy Monte Carlo calculation for various types of fuel blocks (or assemblies). The procedure of cross-section adjustment is implemented in the MCNP5 code

Neutron shielding material

International Nuclear Information System (INIS)

Suzuki, Shigenori; Iimori, Hiroshi; Kobori, Junzo.

1980-01-01

Purpose: To provide a neutron shielding material which incorporates preferable shielding capacity, heat resistance, fire resistance and workability by employing a mixture of thermosetting resin, polyethylene and aluminium hydroxide in special range ratio and curing it. Constitution: A mixture containing 20 to 60% by weight of thermosetting resin having preferable heat resistance, 10 to 40% by weight of polyethylene powder having high hydrogen atom density and 1000 to 60000 of molecular weight, and 15 to 55% by weight of Al(OH) 3 for imparting fire resistance and self-fire extinguishing property thereto is cured. At this time approx. 0.5 to 5% of curing catalyst of the thermosetting resin is contained in 100 parts by weight of the mixture. (Sekiya, K.)

Comparison of MCNP4C and experimental results on neutron and gamma ray shielding effects for materials

Energy Technology Data Exchange (ETDEWEB)

Cha, Kyoon Ho; Lee, Eun Ki [KEPRI, Taejon (Korea, Republic of)

2004-07-01

MCNP code is a general-purpose Monte Carlo radiation transport code that can numerically simulate neutron, photon, and electron transport. Increasing the speed of computing machine is making numerical transport simulation more attractive and has led to the widespread use of such code. This code can be used for general radiation shielding and criticality accident alarm system related dose calculations, so that the version 4C2 of this code was used to evaluate the shielding effect against neutron and gamma ray experiments. The Ueki experiments were used for neutron shielding effects for materials, and the Kansas State University (KSU) photon skyshine experiments of 1977 were tested for gamma ray shielding effects.

Preliminary neutron shielding calculations of the electronics in the EAST BES systems focusing on neutron induced displacement damage

Energy Technology Data Exchange (ETDEWEB)

Náfrádi, Gábor, E-mail: nafradi@reak.bme.hu [Institute of Nuclear Techniques (NTI), Budapest University of Technology and Economics (BME), H-1111 Budapest (Hungary); Kovácsik, Ákos, E-mail: kovacsik.akos@reak.bme.hu [Institute of Nuclear Techniques (NTI), Budapest University of Technology and Economics (BME), H-1111 Budapest (Hungary); Németh, József, E-mail: nemeth.jozsef@wigner.mta.hu [Institute for Particle and Nuclear Physics, Wigner Research Centre for Physics (Wigner RCP), Hungarian Academy of Sciences (HAS), POB 49, 1525 Budapest (Hungary); Pór, Gábor, E-mail: por@reak.bme.hu [Institute of Nuclear Techniques (NTI), Budapest University of Technology and Economics (BME), H-1111 Budapest (Hungary); Zoletnik, Sándor, E-mail: zoletnik.sandor@wigner.mta.hu [Institute for Particle and Nuclear Physics, Wigner Research Centre for Physics (Wigner RCP), Hungarian Academy of Sciences (HAS), POB 49, 1525 Budapest (Hungary)

2016-11-15

Monte Carlo N-Particle (MCNP) calculations were carried out to compare neutron shielding capabilities of three frequently used neutron shielding materials: polyethylene without neutron absorbers, polyethylene with boron absorbers and polyethylene with lithium absorbers, according to Non Ionizing Energy Loss (NIEL). The results of 1D shielding calculations showed that simple neutron moderating materials can provide sufficient and cheap shielding against 2.45 MeV and 14.1 MeV fusion neutrons, in terms of 1 MeV neutron equivalent flux, in silicon targets, which is the most commonly used material of electronic components. Based on these results a new shielding concept is proposed which can be taken into consideration where the reduction of displacement damage is the main goal and the free space available for shielding is limited. Based on this shielding concept detailed 3D calculations were carried out to describe the properties of the neutron shielding of the Beam Emission Spectroscopy (BES) system installed at the EAST tokamak.

An Analysis on the Characteristic of Multi-response CADIS Method for the Monte Carlo Radiation Shielding Calculation

International Nuclear Information System (INIS)

Kim, Do Hyun; Shin, Chang Ho; Kim, Song Hyun

2014-01-01

It uses the deterministic method to calculate adjoint fluxes for the decision of the parameters used in the variance reductions. This is called as hybrid Monte Carlo method. The CADIS method, however, has a limitation to reduce the stochastic errors of all responses. The Forward Weighted CADIS (FW-CADIS) was introduced to solve this problem. To reduce the overall stochastic errors of the responses, the forward flux is used. In the previous study, the Multi-Response CADIS (MR-CAIDS) method was derived for minimizing sum of each squared relative error. In this study, the characteristic of the MR-CADIS method was evaluated and compared with the FW-CADIS method. In this study, how the CADIS, FW-CADIS, and MR-CADIS methods are applied to optimize and decide the parameters used in the variance reduction techniques was analyzed. The MR-CADIS Method uses a technique that the sum of squared relative error in each tally region was minimized to achieve uniform uncertainty. To compare the simulation efficiency of the methods, a simple shielding problem was evaluated. Using FW-CADIS method, it was evaluated that the average of the relative errors was minimized; however, MR-CADIS method gives a lowest variance of the relative errors. Analysis shows that, MR-CADIS method can efficiently and uniformly reduce the relative error of the plural response problem than FW-CADIS method

Graphs of neutron cross sections in JSD1000 for radiation shielding safety analysis

International Nuclear Information System (INIS)

Yamano, Naoki

1984-03-01

Graphs of neutron cross sections and self-shielding factors in the JSD1000 library are presented for radiation shielding safety analysis. The compilation contains various reaction cross sections for 42 nuclides from 1 H to 241 Am in the energy range from 3.51 x 10 -4 eV to 16.5 MeV. The Bondarenko-type self-shielding factors of each reaction are given by the background cross sections from σ 0 = 0 to σ 0 = 10000. (author)

Shielding performances analysis for the IFMIF test facility based on high-fidelity Monte Carlo neutronic calculations

Energy Technology Data Exchange (ETDEWEB)

Kondo, Keitaro, E-mail: kondo.keitaro@jaea.go.jp; Arbeiter, Frederik; Fischer, Ulrich; Lu, Lei; Qiu, Yuefeng; Tian, Kuo

2015-10-15

Highlights: • A detailed geometry model with pipe penetrations and gaps was prepared for the IFMIF test cell. • The neutron streaming effect due to gaps and pipes with shielding plugs was investigated. • The present analysis revealed that the streaming effect can be mitigated by some counter measures. • Occupational workers can access to the room above the test cell during operation. - Abstract: The IFMIF Test Cell (TC) design was developed and optimized in the EVEDA phase, and finally the reference TC design was proposed. The present study is devoted to further investigations of open issues on the reference TC design. In order to examine the neutron streaming effect caused by pipe penetrations and gaps around removable shielding plugs, a new geometry model for neutronic analyses has been prepared directly from engineering CAD data by utilizing the McCad conversion software. All removable shielding plugs are separately described in the model and a detailed description of pipes was incorporated into the model. The calculation result suggests that the streaming effect is mitigated if the pipe penetration is designed appropriately, while the gaps around the shielding plugs above the TC have large impact on the radiation dose in the access cell. The concept of the reference TC design has been basically validated from the neutronics point of view, although the streaming effect should be compensated by the shielding capability of the test cell cover plate so that occupational workers can access to the access cell during operation.

Shielding performances analysis for the IFMIF test facility based on high-fidelity Monte Carlo neutronic calculations

International Nuclear Information System (INIS)

Kondo, Keitaro; Arbeiter, Frederik; Fischer, Ulrich; Lu, Lei; Qiu, Yuefeng; Tian, Kuo

2015-01-01

Highlights: • A detailed geometry model with pipe penetrations and gaps was prepared for the IFMIF test cell. • The neutron streaming effect due to gaps and pipes with shielding plugs was investigated. • The present analysis revealed that the streaming effect can be mitigated by some counter measures. • Occupational workers can access to the room above the test cell during operation. - Abstract: The IFMIF Test Cell (TC) design was developed and optimized in the EVEDA phase, and finally the reference TC design was proposed. The present study is devoted to further investigations of open issues on the reference TC design. In order to examine the neutron streaming effect caused by pipe penetrations and gaps around removable shielding plugs, a new geometry model for neutronic analyses has been prepared directly from engineering CAD data by utilizing the McCad conversion software. All removable shielding plugs are separately described in the model and a detailed description of pipes was incorporated into the model. The calculation result suggests that the streaming effect is mitigated if the pipe penetration is designed appropriately, while the gaps around the shielding plugs above the TC have large impact on the radiation dose in the access cell. The concept of the reference TC design has been basically validated from the neutronics point of view, although the streaming effect should be compensated by the shielding capability of the test cell cover plate so that occupational workers can access to the access cell during operation.

Neutron and gamma sensitivities of self-powered detectors: Monte Carlo modelling

Energy Technology Data Exchange (ETDEWEB)

Vermeeren, Ludo [SCK-CEN, Nuclear Research Centre, Boeretang 200, B-2400 Mol, (Belgium)

2015-07-01

This paper deals with the development of a detailed Monte Carlo approach for the calculation of the absolute neutron sensitivity of SPNDs, which makes use of the MCNP code. We will explain the calculation approach, including the activation and beta emission steps, the gamma-electron interactions, the charge deposition in the various detector parts and the effect of the space charge field in the insulator. The model can also be applied for the calculation of the gamma sensitivity of self-powered detectors and for the radiation-induced currents in signal cables. The model yields detailed information on the various contributions to the sensor currents, with distinct response times. Results for the neutron sensitivity of various types of SPNDs are in excellent agreement with experimental data obtained at the BR2 research reactor. For typical neutron to gamma flux ratios, the calculated gamma induced SPND currents are significantly lower than the neutron induced currents. The gamma sensitivity depends very strongly upon the immediate detector surroundings and on the gamma spectrum. Our calculation method opens the way to a reliable on-line determination of the absolute in-pile thermal neutron flux. (authors)

Performance of quantum Monte Carlo for calculating molecular bond lengths

Energy Technology Data Exchange (ETDEWEB)

Cleland, Deidre M., E-mail: deidre.cleland@csiro.au; Per, Manolo C., E-mail: manolo.per@csiro.au [CSIRO Virtual Nanoscience Laboratory, 343 Royal Parade, Parkville, Victoria 3052 (Australia)

2016-03-28

This work investigates the accuracy of real-space quantum Monte Carlo (QMC) methods for calculating molecular geometries. We present the equilibrium bond lengths of a test set of 30 diatomic molecules calculated using variational Monte Carlo (VMC) and diffusion Monte Carlo (DMC) methods. The effect of different trial wavefunctions is investigated using single determinants constructed from Hartree-Fock (HF) and Density Functional Theory (DFT) orbitals with LDA, PBE, and B3LYP functionals, as well as small multi-configurational self-consistent field (MCSCF) multi-determinant expansions. When compared to experimental geometries, all DMC methods exhibit smaller mean-absolute deviations (MADs) than those given by HF, DFT, and MCSCF. The most accurate MAD of 3 ± 2 × 10{sup −3} Å is achieved using DMC with a small multi-determinant expansion. However, the more computationally efficient multi-determinant VMC method has a similar MAD of only 4.0 ± 0.9 × 10{sup −3} Å, suggesting that QMC forces calculated from the relatively simple VMC algorithm may often be sufficient for accurate molecular geometries.

Biases in Monte Carlo eigenvalue calculations

Energy Technology Data Exchange (ETDEWEB)

Gelbard, E.M.

1992-12-01

The Monte Carlo method has been used for many years to analyze the neutronics of nuclear reactors. In fact, as the power of computers has increased the importance of Monte Carlo in neutronics has also increased, until today this method plays a central role in reactor analysis and design. Monte Carlo is used in neutronics for two somewhat different purposes, i.e., (a) to compute the distribution of neutrons in a given medium when the neutron source-density is specified, and (b) to compute the neutron distribution in a self-sustaining chain reaction, in which case the source is determined as the eigenvector of a certain linear operator. In (b), then, the source is not given, but must be computed. In the first case (the ``fixed-source`` case) the Monte Carlo calculation is unbiased. That is to say that, if the calculation is repeated (``replicated``) over and over, with independent random number sequences for each replica, then averages over all replicas will approach the correct neutron distribution as the number of replicas goes to infinity. Unfortunately, the computation is not unbiased in the second case, which we discuss here.

Biases in Monte Carlo eigenvalue calculations

Energy Technology Data Exchange (ETDEWEB)

Gelbard, E.M.

1992-01-01

The Monte Carlo method has been used for many years to analyze the neutronics of nuclear reactors. In fact, as the power of computers has increased the importance of Monte Carlo in neutronics has also increased, until today this method plays a central role in reactor analysis and design. Monte Carlo is used in neutronics for two somewhat different purposes, i.e., (a) to compute the distribution of neutrons in a given medium when the neutron source-density is specified, and (b) to compute the neutron distribution in a self-sustaining chain reaction, in which case the source is determined as the eigenvector of a certain linear operator. In (b), then, the source is not given, but must be computed. In the first case (the fixed-source'' case) the Monte Carlo calculation is unbiased. That is to say that, if the calculation is repeated ( replicated'') over and over, with independent random number sequences for each replica, then averages over all replicas will approach the correct neutron distribution as the number of replicas goes to infinity. Unfortunately, the computation is not unbiased in the second case, which we discuss here.

Biases in Monte Carlo eigenvalue calculations

International Nuclear Information System (INIS)

Gelbard, E.M.

1992-01-01

The Monte Carlo method has been used for many years to analyze the neutronics of nuclear reactors. In fact, as the power of computers has increased the importance of Monte Carlo in neutronics has also increased, until today this method plays a central role in reactor analysis and design. Monte Carlo is used in neutronics for two somewhat different purposes, i.e., (a) to compute the distribution of neutrons in a given medium when the neutron source-density is specified, and (b) to compute the neutron distribution in a self-sustaining chain reaction, in which case the source is determined as the eigenvector of a certain linear operator. In (b), then, the source is not given, but must be computed. In the first case (the ''fixed-source'' case) the Monte Carlo calculation is unbiased. That is to say that, if the calculation is repeated (''replicated'') over and over, with independent random number sequences for each replica, then averages over all replicas will approach the correct neutron distribution as the number of replicas goes to infinity. Unfortunately, the computation is not unbiased in the second case, which we discuss here

Validation of SCALE code package on high performance neutron shields

International Nuclear Information System (INIS)

Bace, M.; Jecmenica, R.; Smuc, T.

1999-01-01

The shielding ability and other properties of new high performance neutron shielding materials from the KRAFTON series have been recently published. A comparison of the published experimental and MCNP results for the two materials of the KRAFTON series, with our own calculations has been done. Two control modules of the SCALE-4.4 code system have been used, one of them based on one dimensional radiation transport analysis (SAS1) and other based on the three dimensional Monte Carlo method (SAS3). The comparison of the calculated neutron dose equivalent rates shows a good agreement between experimental and calculated results for the KRAFTON-N2 material.. Our results indicate that the N2-M-N2 sandwich type is approximately 10% inferior as neutron shield to the KRAFTON-N2 material. All values of neutron dose equivalent obtained by SAS1 are approximately 25% lower in comparison with the SAS3 results, which indicates proportions of discrepancies introduced by one-dimensional geometry approximation.(author)

The evaluation of the radiation shielding ability of lead glass

International Nuclear Information System (INIS)

Tsuda, Keisuke; Fukushi, Masahiro; Myojoyama, Atsushi; Kitamura, Hideaki; Nakaya, Giichiro; Hassan, Nabil; Inoue, Kazumasa; Kimura, Junichi; Sawaguchi, Masato; Kinase, Sakae; Saito, Kimiaki

2008-01-01

Positron emission tomography (PET) scanning with the tracer 2-[F-18] Fluoro-2deoxy-D-glucose (FDG) is widely used in the clinical PET. However, the photon energy used in the PET scans is considerably higher than that of the X-rays traditionally used in the diagnoses. The radiation protection in the PET institution, therefore, is the remaining problem. Meanwhile, lead glass has attracted considerable attention as a radiation-shielding material for the PET institution. The aim of the present study was to evaluate the radiation-shielding ability of the lead glass against the positron emitters. The shielding ability evaluations were done both in the actual experiments and in the Monte Carlo simulation. The lead glass, the object of evaluation in this study, proved to have sufficient protective effect. The development and the spread of a thinner and lighter lead glass with the same effective dose transmission factor should be expected in the near future. (author)

Monte Carlo Simulations of New 2D Ripple Filters for Particle Therapy Facilities

DEFF Research Database (Denmark)

Ringbæk, Toke Printz; Weber, Uli; Petersen, Jørgen B.B.

2014-01-01

). At the Universitätsklinikum Gießen und Marburg, Germany, a new second generation RiFi has been developed with two-dimensional groove structures. In this work we evaluate this new RiFi design. Methods: The Monte Carlo (MC) code SHIELD-HIT12A is used to determine the RiFi- induced inhomogeneities in the dose distribution...... for various ion types, initial particle energies and distances from the RiFi to the phantom surface as well as in the depth of the phantom. The beam delivery and monitor system (BAMS) used at Marburg, the Heidelberg Ionentherapiezentrum (HIT), Universit ̈tsklinikum Heidelberg, Germany and the GSI...... Helmholtzzentrum für Schwerionenforschung, Darmstadt, Germany is modeled and simulated. To evaluate the PTV dose coverage performance of the new RiFi design, the heavy ion treatment planning system TRiP98 is used for dose optimization. SHIELD-HIT12A is used to prepare the facility-specific physical dose kernels...

RESONANCE SELF-SHIELDING EFFECT IN UNCERTAINTY QUANTIFICATION OF FISSION REACTOR NEUTRONICS PARAMETERS

Directory of Open Access Journals (Sweden)

GO CHIBA

2014-06-01

Full Text Available In order to properly quantify fission reactor neutronics parameter uncertainties, we have to use covariance data and sensitivity profiles consistently. In the present paper, we establish two consistent methodologies for uncertainty quantification: a self-shielded cross section-based consistent methodology and an infinitely-diluted cross section-based consistent methodology. With these methodologies and the covariance data of uranium-238 nuclear data given in JENDL-3.3, we quantify uncertainties of infinite neutron multiplication factors of light water reactor and fast reactor fuel cells. While an inconsistent methodology gives results which depend on the energy group structure of neutron flux and neutron-nuclide reaction cross section representation, both the consistent methodologies give fair results with no such dependences.

Practical Application of Monte Carlo Code in RTP

International Nuclear Information System (INIS)

Mohamad Hairie Rabir; Julia Abdul Karim; Muhammad Rawi Mohamed Zin; Na'im Syauqi Hamzah; Mark Dennis Anak Usang; Abi Muttaqin Jalal Bayar; Muhammad Khairul Ariff Mustafa

2015-01-01

Monte Carlo neutron transport codes are widely used in various reactor physics applications in RTP and other related nuclear and radiation research in Nuklear Malaysia. The main advantage of the method is the capability to model geometry and interaction physics without major approximations. The disadvantage is that the modelling of complicated systems is very computing-intensive, which restricts the applications to some extent. The importance of Monte Carlo calculation is likely to increase in the future, along with the development in computer capacities and parallel calculation. This paper presents several calculation activities, its achievements and challenges in using MCNP code for neutronics analysis, nuclide inventory and source term calculation, shielding and dose evaluation. (author)

Self-shielding phenomenon modelling in multigroup transport code Apollo-2; Modelisation du phenomene d'autoprotection dans le code de transport multigroupe Apollo 2

Energy Technology Data Exchange (ETDEWEB)

Coste-Delclaux, M

2006-03-15

This document describes the improvements carried out for modelling the self-shielding phenomenon in the multigroup transport code APOLLO2. They concern the space and energy treatment of the slowing-down equation, the setting up of quadrature formulas to calculate reaction rates, the setting-up of a method that treats directly a resonant mixture and the development of a sub-group method. We validate these improvements either in an elementary or in a global way. Now, we obtain, more accurate multigroup reaction rates and we are able to carry out a reference self-shielding calculation on a very fine multigroup mesh. To end, we draw a conclusion and give some prospects on the remaining work. (author)

Neutron shielding and activation of the MASTU device and surrounds

Energy Technology Data Exchange (ETDEWEB)

Taylor, David, E-mail: david.taylor@ccfe.ac.uk [EURATOM/CCFE Fusion Association, Culham Science Centre, Abingdon OX14 3DB (United Kingdom); Lilley, Steven; Turner, Andrew [EURATOM/CCFE Fusion Association, Culham Science Centre, Abingdon OX14 3DB (United Kingdom); Davis, Andrew [Now at College of Engineering, University of Wisconsin, Madison, WI 53706 (United States)

2014-10-15

Highlights: We model neutron shielding for the planned MASTU device; nadequacies in the existing shielding design are remedied; Levels of public exposure are considered; We model activated gamma emission for the device under a worst case scenario. Abstract: A significant functional upgrade is planned for the Mega Ampere Spherical Tokamak (MAST) device, located at Culham in the UK, including the implementation of a notably greater neutral beam injection power. This upgrade will cause the emission of a substantially increased intensity of neutron radiation for a substantially increased amount of time upon operation of the device. Existing shielding and activation precautions are shown to prove insufficient in some regards, and recommendations for improvements are made, including the following areas: shielding doors to MAST shielded facility enclosure (known as “the blockhouse”); north access tunnel; blockhouse roof; west cabling duct. In addition, some specific neutronic dose rate questions are addressed and answered; those discussed here relate to shielding penetrations and dose rate reflected from the air above the device (“skyshine”). It is shown that the alterations to shielding and area access reduce the dose rate in unrestricted areas from greater than 100 μSv/h to less than 2 μSv/h averaged over the working day. The tools used for this analysis are the MCNP (Monte Carlo N-particle) code, used to calculate the three-dimensional spatial distribution of neutron and photon dose rates in and around the device and its shields, and the nuclear inventory code FISPACT, run under the umbrella code MCR2S, used to calculate the time-dependent shutdown dose rate in the region of the device at several decay times.

Shielded transient self-interaction of a bunch entering a circle from a straight path

International Nuclear Information System (INIS)

Li, R.; Bohn, C.L.; Bisognano, J.J.

1997-01-01

When a short (mm-length) bunch with high (nC-regime) charge is transported through a magnetic bending system, self-interaction via coherent synchrotron radiation (CSR) and space charge may alter the bunch dynamics significantly. The authors consider a Gaussian rigid-line-charge bunch following a straight-path trajectory into a circle, with the trajectory centered between two infinite, parallel, perfectly conducting plates. Transients associated with CSR and space charge generated from source particles both on the straight path and the circle are calculated, and their net effect on the radiated power is contrasted with that of shielded steady-state CSR

Secondary gamma-ray data for shielding calculation

International Nuclear Information System (INIS)

Miyasaka, Sunichi

1979-01-01

In deep penetration transport calculations, the integral design parameters is determined mainly by secondary particles which are produced by interactions of the primary radiation with materials. The shield thickness and the biological dose rate at a given point of a bulk shield are determined from the contribution from secondary gamma rays. The heat generation and the radiation damage in the structural and shield materials depend strongly on the secondary gamma rays. In this paper, the status of the secondary gamma ray data and its further problems are described from the viewpoint of shield design. The secondary gamma-ray data in ENDF/B-IV and POPOP4 are also discussed based on the test calculations made for several shield assemblies. (author)

A parallelization study of the general purpose Monte Carlo code MCNP4 on a distributed memory highly parallel computer

International Nuclear Information System (INIS)

Yamazaki, Takao; Fujisaki, Masahide; Okuda, Motoi; Takano, Makoto; Masukawa, Fumihiro; Naito, Yoshitaka

1993-01-01

The general purpose Monte Carlo code MCNP4 has been implemented on the Fujitsu AP1000 distributed memory highly parallel computer. Parallelization techniques developed and studied are reported. A shielding analysis function of the MCNP4 code is parallelized in this study. A technique to map a history to each processor dynamically and to map control process to a certain processor was applied. The efficiency of parallelized code is up to 80% for a typical practical problem with 512 processors. These results demonstrate the advantages of a highly parallel computer to the conventional computers in the field of shielding analysis by Monte Carlo method. (orig.)

Parameters calculation of shielding experiment

International Nuclear Information System (INIS)

Gavazza, S.

1986-02-01

The radiation transport methodology comparing the calculated reactions and dose rates for neutrons and gama-rays, with experimental measurements obtained on iron shield, irradiated in the YAYOI reactor is evaluated. The ENDF/B-IV and VITAMIN-C libraries and the AMPX-II modular system, for cross sections generation collapsed by the ANISN code were used. The transport calculations were made using the DOT 3.5 code, adjusting the boundary iron shield source spectrum to the reactions and dose rates, measured at the beginning of shield. The neutron and gamma ray distributions calculated on the iron shield presented reasonable agreement with experimental measurements. An experimental arrangement using the IEA-R1 reactor to determine a shielding benchmark is proposed. (Author) [pt

Perturbative expansions from Monte Carlo simulations at weak coupling: Wilson loops and the static-quark self-energy

Science.gov (United States)

Trottier, H. D.; Shakespeare, N. H.; Lepage, G. P.; MacKenzie, P. B.

2002-05-01

Perturbative coefficients for Wilson loops and the static-quark self-energy are extracted from Monte Carlo simulations at weak coupling. The lattice volumes and couplings are chosen to ensure that the lattice momenta are all perturbative. Twisted boundary conditions are used to eliminate the effects of lattice zero modes and to suppress nonperturbative finite-volume effects due to Z(3) phases. Simulations of the Wilson gluon action are done with both periodic and twisted boundary conditions, and over a wide range of lattice volumes (from 34 to 164) and couplings (from β~9 to β~60). A high precision comparison is made between the simulation data and results from finite-volume lattice perturbation theory. The Monte Carlo results are shown to be in excellent agreement with perturbation theory through second order. New results for third-order coefficients for a number of Wilson loops and the static-quark self-energy are reported.

Modeling of the 3RS tau protein with self-consistent field method and Monte Carlo simulation

NARCIS (Netherlands)

Leermakers, F.A.M.; Jho, Y.S.; Zhulina, E.B.

2010-01-01

Using a model with amino acid resolution of the 196 aa N-terminus of the 3RS tau protein, we performed both a Monte Carlo study and a complementary self-consistent field (SCF) analysis to obtain detailed information on conformational properties of these moieties near a charged plane (mimicking the

A general transform for variance reduction in Monte Carlo simulations

International Nuclear Information System (INIS)

Becker, T.L.; Larsen, E.W.

2011-01-01

This paper describes a general transform to reduce the variance of the Monte Carlo estimate of some desired solution, such as flux or biological dose. This transform implicitly includes many standard variance reduction techniques, including source biasing, collision biasing, the exponential transform for path-length stretching, and weight windows. Rather than optimizing each of these techniques separately or choosing semi-empirical biasing parameters based on the experience of a seasoned Monte Carlo practitioner, this General Transform unites all these variance techniques to achieve one objective: a distribution of Monte Carlo particles that attempts to optimize the desired solution. Specifically, this transform allows Monte Carlo particles to be distributed according to the user's specification by using information obtained from a computationally inexpensive deterministic simulation of the problem. For this reason, we consider the General Transform to be a hybrid Monte Carlo/Deterministic method. The numerical results con rm that the General Transform distributes particles according to the user-specified distribution and generally provide reasonable results for shielding applications. (author)

The determination of self-powered neutron detector sensitivity on thermal and epithermal neutron flux densities

International Nuclear Information System (INIS)

Erben, O.

1980-01-01

The coefficients of thermal and epithermal neutron flux density depression and self-shielding for the SPN detectors with vanadium, rhodium, silver and cobalt emitters are presented, (for cobalt SPN detectors the functions describing the absorbtion of neutrons along the emitter cross-section are also shown). Using these coefficients and previously published beta particle escape efficiencies, sensitivities are determined for the principal types of detectors produced by Les Cables de Lyon and SODERN companies. The experiments and their results verifying the validity of the theoretical work are described. (author)

The determination of beam quality correction factors: Monte Carlo simulations and measurements.

Science.gov (United States)

González-Castaño, D M; Hartmann, G H; Sánchez-Doblado, F; Gómez, F; Kapsch, R-P; Pena, J; Capote, R

2009-08-07

Modern dosimetry protocols are based on the use of ionization chambers provided with a calibration factor in terms of absorbed dose to water. The basic formula to determine the absorbed dose at a user's beam contains the well-known beam quality correction factor that is required whenever the quality of radiation used at calibration differs from that of the user's radiation. The dosimetry protocols describe the whole ionization chamber calibration procedure and include tabulated beam quality correction factors which refer to 60Co gamma radiation used as calibration quality. They have been calculated for a series of ionization chambers and radiation qualities based on formulae, which are also described in the protocols. In the case of high-energy photon beams, the relative standard uncertainty of the beam quality correction factor is estimated to amount to 1%. In the present work, two alternative methods to determine beam quality correction factors are prescribed-Monte Carlo simulation using the EGSnrc system and an experimental method based on a comparison with a reference chamber. Both Monte Carlo calculations and ratio measurements were carried out for nine chambers at several radiation beams. Four chamber types are not included in the current dosimetry protocols. Beam quality corrections for the reference chamber at two beam qualities were also measured using a calorimeter at a PTB Primary Standards Dosimetry Laboratory. Good agreement between the Monte Carlo calculated (1% uncertainty) and measured (0.5% uncertainty) beam quality correction factors was obtained. Based on these results we propose that beam quality correction factors can be generated both by measurements and by the Monte Carlo simulations with an uncertainty at least comparable to that given in current dosimetry protocols.

Neutron Skyshine in shielding projects of radiotherapy: comparison between theoretical approach and simulation by Monte Carlo method; 'Skyshine' de neutrons em projetos de blindagens de radioterapia: comparacao entre abordagem teorica e simulacao por metodo de Monte Carlo

Energy Technology Data Exchange (ETDEWEB)

Falcao, R.C.; Facure, A. [Comissao Nacional de Energia Nuclear (CNEN), Rio de Janeiro, RJ (Brazil); Santini, E.S. [Comissao Nacional de Energia Nuclear (CNEN), Rio de Janeiro, RJ (Brazil); Centro Brasileiro de Pesquisas Fisicas (CBPF), Rio de Janeiro, RJ (Brazil); Silva, A.X. [Coordenacao dos Programas de Pos-Graduacao de Engenharia (PEN/COPPE/UFRJ), RJ (Brazil). Programa de Engenharia Nuclear

2005-07-01

In this work, the MCNP code is used to simulate the transport of neutrons in a room of radiotherapy, whose shieldings are designed according to the method of skyshine (scattering in the atmosphere). The simulations are compared with the results obtained from empirically established expressions, which are normally used for designing the ceilings of the rooms facilities, ensuring that dose rates (neutrons + photons) around them do not exceed the maximum limits allowed by the standards of the CNEN. Good agreement is observed between the doses calculated according to these expressions and those obtained through simulation by Monte Carlo in the case of rooms without ceiling, and an overestimate of the calculations by a factor 2 or 3 in relation to the simulations, in the case of rooms with ceiling.

Electron correlation effects on geometries and 19F shieldings of fluorobenzenes

International Nuclear Information System (INIS)

Webb, G.A.; Karadakov, P.B.; England, J.A.

2000-01-01

In order to include the effects of electron correlation in ab initio molecular orbital calculations it is necessary to go beyond the single determinant Hartree-Fock (HF) level of theory. In the present investigation the influences of both dynamic and non-dynamic correlation effects on the optimised geometries and 19 F nuclear shielding calculations of the twelve fluorobenzenes are reported.The non-dynamic electron correlation effects are represented by complete-active space self-consistent field (CASSCF) calculations. Second- and fourth-order Moller-Plesset (MP2 and MP4) calculations are used to describe the dynamic electron correlation effects. Some density-functional (DFT) results are also reported which do not distinguish between dynamic and non-dynamic electron correlation. Following the correlated geometry optimisations 19 F nuclear shielding calculations were performed using the gauge-included atomic orbitals (GIAO) procedure, these were undertaken with wave functions which include various levels of electron correlation including HF, CASSCF and MP2. For the calculations of the optimised geometries, and some of the nuclear shieldings the 6-13G** basis set s used whereas the locally-dense [6-13G** on C and H and 6-311++G(2d,2p) on F] set is used for some of the shielding calculations. A comparison of the results of HF shielding calculations using other basis sets is included. Comparison of the calculated geometry and shielding results with relevant, reported, experimental data is made. (author)

Successful vectorization - reactor physics Monte Carlo code

International Nuclear Information System (INIS)

Martin, W.R.

1989-01-01

Most particle transport Monte Carlo codes in use today are based on the ''history-based'' algorithm, wherein one particle history at a time is simulated. Unfortunately, the ''history-based'' approach (present in all Monte Carlo codes until recent years) is inherently scalar and cannot be vectorized. In particular, the history-based algorithm cannot take advantage of vector architectures, which characterize the largest and fastest computers at the current time, vector supercomputers such as the Cray X/MP or IBM 3090/600. However, substantial progress has been made in recent years in developing and implementing a vectorized Monte Carlo algorithm. This algorithm follows portions of many particle histories at the same time and forms the basis for all successful vectorized Monte Carlo codes that are in use today. This paper describes the basic vectorized algorithm along with descriptions of several variations that have been developed by different researchers for specific applications. These applications have been mainly in the areas of neutron transport in nuclear reactor and shielding analysis and photon transport in fusion plasmas. The relative merits of the various approach schemes will be discussed and the present status of known vectorization efforts will be summarized along with available timing results, including results from the successful vectorization of 3-D general geometry, continuous energy Monte Carlo. (orig.)

EBT-P gamma-ray-shielding analysis

International Nuclear Information System (INIS)

Gohar, Y.

1983-01-01

First, a one-dimensional scoping study was performed for the gamma-ray shield of the ELMO Bumpy Torus proof-of-principle device to define appropriate shielding material and determine the required shielding thickness. The dose-equivalent results are analyzed as a function of the radiation-shield thickness for different shielding options. A sensitivity analysis for the pessimistic case is given. The recommended shielding option based on the performance and cost is discussed. Next, a three-dimensional scoping study for the coil shield was performed for four different shielding options to define the heat load for each component and check the compliance with the design criterion of 10 watts maximum heat load per coil from the gamma-ray sources. Also, a detailed biological-dose survey was performed which included: (a) the dose equivalent inside and outside the building, (b) the dose equivalent from the two mazes of the building, and (c) the skyshine contribution to the dose equivalent

Automatic modeling for the monte carlo transport TRIPOLI code

International Nuclear Information System (INIS)

Zhang Junjun; Zeng Qin; Wu Yican; Wang Guozhong; FDS Team

2010-01-01

TRIPOLI, developed by CEA, France, is Monte Carlo particle transport simulation code. It has been widely applied to nuclear physics, shielding design, evaluation of nuclear safety. However, it is time-consuming and error-prone to manually describe the TRIPOLI input file. This paper implemented bi-directional conversion between CAD model and TRIPOLI model. Its feasibility and efficiency have been demonstrated by several benchmarking examples. (authors)

Monte Carlo simulation of nuclear energy study (II). Annual report on Nuclear Code Evaluation Committee

Energy Technology Data Exchange (ETDEWEB)

NONE

2001-01-01

In the report, research results discussed in 1999 fiscal year at Nuclear Code Evaluation Committee of Nuclear Code Research Committee were summarized. Present status of Monte Carlo simulation on nuclear energy study was described. Especially, besides of criticality, shielding and core analyses, present status of applications to risk and radiation damage analyses, high energy transport and nuclear theory calculations of Monte Carlo Method was described. The 18 papers are indexed individually. (J.P.N.)

Perturbative expansions from Monte Carlo simulations at weak coupling: Wilson loops and the static-quark self-energy

International Nuclear Information System (INIS)

Trottier, H.D.; Shakespeare, N.H.; Lepage, G.P.; Mackenzie, P.B.

2002-01-01

Perturbative coefficients for Wilson loops and the static-quark self-energy are extracted from Monte Carlo simulations at weak coupling. The lattice volumes and couplings are chosen to ensure that the lattice momenta are all perturbative. Twisted boundary conditions are used to eliminate the effects of lattice zero modes and to suppress nonperturbative finite-volume effects due to Z(3) phases. Simulations of the Wilson gluon action are done with both periodic and twisted boundary conditions, and over a wide range of lattice volumes (from 3 4 to 16 4 ) and couplings (from β≅9 to β≅60). A high precision comparison is made between the simulation data and results from finite-volume lattice perturbation theory. The Monte Carlo results are shown to be in excellent agreement with perturbation theory through second order. New results for third-order coefficients for a number of Wilson loops and the static-quark self-energy are reported

Shielding calculations for industrial 5/7.5MeV electron accelerators using the MCNP Monte Carlo Code

Science.gov (United States)

Peri, Eyal; Orion, Itzhak

2017-09-01

High energy X-rays from accelerators are used to irradiate food ingredients to prevent growth and development of unwanted biological organisms in food, and by that extend the shelf life of the products. The production of X-rays is done by accelerating 5 MeV electrons and bombarding them into a heavy target (high Z). Since 2004, the FDA has approved using 7.5 MeV energy, providing higher production rates with lower treatments costs. In this study we calculated all the essential data needed for a straightforward concrete shielding design of typical food accelerator rooms. The following evaluation is done using the MCNP Monte Carlo code system: (1) Angular dependence (0-180°) of photon dose rate for 5 MeV and 7.5 MeV electron beams bombarding iron, aluminum, gold, tantalum, and tungsten targets. (2) Angular dependence (0-180°) spectral distribution simulations of bremsstrahlung for gold, tantalum, and tungsten bombarded by 5 MeV and 7.5 MeV electron beams. (3) Concrete attenuation calculations in several photon emission angles for the 5 MeV and 7.5 MeV electron beams bombarding a tantalum target. Based on the simulation, we calculated the expected increase in dose rate for facilities intending to increase the energy from 5 MeV to 7.5 MeV, and the concrete width needed to be added in order to keep the existing dose rate unchanged.

Monte Carlo simulations of low background detectors

International Nuclear Information System (INIS)

Miley, H.S.; Brodzinski, R.L.; Hensley, W.K.; Reeves, J.H.

1995-01-01

An implementation of the Electron Gamma Shower 4 code (EGS4) has been developed to allow convenient simulation of typical gamma ray measurement systems. Coincidence gamma rays, beta spectra, and angular correlations have been added to adequately simulate a complete nuclear decay and provide corrections to experimentally determined detector efficiencies. This code has been used to strip certain low-background spectra for the purpose of extremely low-level assay. Monte Carlo calculations of this sort can be extremely successful since low background detectors are usually free of significant contributions from poorly localized radiation sources, such as cosmic muons, secondary cosmic neutrons, and radioactive construction or shielding materials. Previously, validation of this code has been obtained from a series of comparisons between measurements and blind calculations. An example of the application of this code to an exceedingly low background spectrum stripping will be presented. (author) 5 refs.; 3 figs.; 1 tab

Measurements and FLUKA Simulations of Bismuth, Aluminium and Indium Activation at the upgraded CERN Shielding Benchmark Facility (CSBF)

Science.gov (United States)

Iliopoulou, E.; Bamidis, P.; Brugger, M.; Froeschl, R.; Infantino, A.; Kajimoto, T.; Nakao, N.; Roesler, S.; Sanami, T.; Siountas, A.; Yashima, H.

2018-06-01

The CERN High energy AcceleRator Mixed field (CHARM) facility is situated in the CERN Proton Synchrotron (PS) East Experimental Area. The facility receives a pulsed proton beam from the CERN PS with a beam momentum of 24 GeV/c with 5·1011 protons per pulse with a pulse length of 350 ms and with a maximum average beam intensity of 6.7·1010 protons per second. The extracted proton beam impacts on a cylindrical copper target. The shielding of the CHARM facility includes the CERN Shielding Benchmark Facility (CSBF) situated laterally above the target that allows deep shielding penetration benchmark studies of various shielding materials. This facility has been significantly upgraded during the extended technical stop at the beginning of 2016. It consists now of 40 cm of cast iron shielding, a 200 cm long removable sample holder concrete block with 3 inserts for activation samples, a material test location that is used for the measurement of the attenuation length for different shielding materials as well as for sample activation at different thicknesses of the shielding materials. Activation samples of bismuth, aluminium and indium were placed in the CSBF in September 2016 to characterize the upgraded version of the CSBF. Monte Carlo simulations with the FLUKA code have been performed to estimate the specific production yields of bismuth isotopes (206 Bi, 205 Bi, 204 Bi, 203 Bi, 202 Bi, 201 Bi) from 209 Bi, 24 Na from 27 Al and 115 m I from 115 I for these samples. The production yields estimated by FLUKA Monte Carlo simulations are compared to the production yields obtained from γ-spectroscopy measurements of the samples taking the beam intensity profile into account. The agreement between FLUKA predictions and γ-spectroscopy measurements for the production yields is at a level of a factor of 2.

The Benchmark experiment on stainless steel bulk shielding at the Frascati neutron generator

International Nuclear Information System (INIS)

Batistoni, P.; Angelone, M.; Martone, M.; Pillon, M.; Rado, V.

1994-11-01

In the framework of the European Technology Program for NET/ITER, ENEA (Italian Agency for New Technologies, Energy and Environment) - Frascati and CEA (Commissariat a L'Energie Atomique) - Cadarache collaborated on a Bulk Shield Benchmark Experiment using the 14-MeV Frascati Neutron Generator (FNG). The aim of the experiment was to obtain accurate experimental data for improving the nuclear database and methods used in shielding designs, through a rigorous analysis of the results. The experiment consisted of the irradiation of a stainless steel block by 14-MeV neutrons. The neutron reaction rates at different depths inside the block were measured by fission chambers and activation foils characterized by different energy response ranges. The experimental results have been compared with numerical results calculated using both S N and Monte Carlo transport codes and as transport cross section library the European Fusion File (EFF). In particular, the present report describes the experimental and numerical activity, including neutron measurements and Monte Carlo calculations, carried out by the ENEA Italian Agency for New Technologies, Energy and Environment) team

Scattered dose to thyroid from prophylactic cranial irradiation during childhood: a Monte Carlo study

International Nuclear Information System (INIS)

Mazonakis, Michalis; Tzedakis, Antonis; Damilakis, John; Varveris, Haris; Kachris, Stefanos; Gourtsoyiannis, Nicholas

2006-01-01

The purpose of this study was to estimate the scattered dose to thyroid from prophylactic cranial irradiation during childhood. The MCNP transport code and mathematical phantoms representing the average individual at ages 3, 5, 10, 15 and 18 years old were employed to simulate cranial radiotherapy using two lateral opposed fields. The mean radiation dose received by the thyroid gland was calculated. A 10 cm thick lead block placed on the patient's couch to shield the thyroid was simulated by MCNP code. The Monte Carlo model was validated by measuring the scattered dose to the unshielded and shielded thyroid using three different humanoid phantoms and thermoluminescense dosimetry. For a cranial dose of 18 Gy, the thyroid dose obtained by Monte Carlo calculations varied from 47 to 79 cGy depending upon the age of the child. Appropriate placement of the couch block resulted in a thyroid dose reduction by 39 to 54%. Thyroid dose values at all possible positions of the radiosensitive gland with respect to the inferior field edge at five different patient ages were found. The mean difference between Monte Carlo results and thyroid dose measurements was 9.6%. (note)

Testing of the PELSHIE shielding code using Benchmark problems and other special shielding models

International Nuclear Information System (INIS)

Language, A.E.; Sartori, D.E.; De Beer, G.P.

1981-08-01

The PELSHIE shielding code for gamma rays from point and extended sources was written in 1971 and a revised version was published in October 1979. At Pelindaba the program is used extensively due to its flexibility and ease of use for a wide range of problems. The testing of PELSHIE results with the results of a range of models and so-called Benchmark problems is desirable to determine possible weaknesses in PELSHIE. Benchmark problems, experimental data, and shielding models, some of which were resolved by the discrete-ordinates method with the ANISN and DOT 3.5 codes, were used for the efficiency test. The description of the models followed the pattern of a classical shielding problem. After the intercomparison with six different models, the usefulness of the PELSHIE code was quantitatively determined [af

Neutron shielding studies on an advanced molten salt fast reactor design

International Nuclear Information System (INIS)

Merk, Bruno; Konheiser, Jörg

2014-01-01

Highlights: • Material damage due to irradiation has already been discovered at the MSRE. • Neutronic analysis of MSFR with curved blanket wall geometry. • Neutron fluence limit at the wall of the outer vessel can be kept for 80 years. • Shielded MSFR core will be of same dimension than a SFR core. - Abstract: The molten salt reactor technology has gained some new interest. In contrast to the historic molten salt reactors, the current projects are based on designing a molten salt fast reactor. Thus the shielding becomes significantly more challenging than in historic concepts. One very interesting and innovative result of the most recent EURATOM project on molten salt reactors – EVOL – is the fluid flow optimized design of the inner reactor vessel using curved blanket walls. The developed structure leads to a very uniform flow distribution. The design avoids all internal structures. Based on this new geometry a model for neutron physics calculation is presented. The major steps are: the modeling of the curved geometry in the unstructured mesh neutron transport code HELIOS and the determination of the real neutron flux and power distribution for this new geometry. The developed model is then used for the determination of the neutron fluence distribution in the inner and outer wall of the system. Based on these results an optimized shielding strategy is developed for the molten salt fast reactor to keep the fluence in the safety related outer vessel below expected limit values. A lifetime of 80 years can be assured, but the size of the core/blanket system will be comparable to a sodium cooled fast reactor. The HELIOS results are verified against Monte-Carlo calculations with very satisfactory agreement for a deep penetration problem

Analysis of the ITER computational shielding benchmark with the Monte Carlo TRIPOLI-4® neutron gamma coupled calculations

International Nuclear Information System (INIS)

Lee, Yi-Kang

2016-01-01

Highlights: • Verification and validation of TRIPOLI-4 radiation transport calculations for ITER shielding benchmark. • Evaluation of CEA-V5.1.1 and FENDL-3.0 nuclear data libraries on D–T fusion neutron continuous energy transport calculations. • Advances in nuclear analyses for nuclear heating and radiation damage in iron. • This work also demonstrates that the “safety factors” concept is necessary in the nuclear analyses of ITER. - Abstract: With the growing interest in using the continuous-energy TRIPOLI-4 ® Monte Carlo radiation transport code for ITER applications, a key issue that arises is whether or not the released TRIPOLI-4 code and its associated nuclear data libraries are verified and validated for the D–T fusion neutronics calculations. Previous published benchmark results of TRIPOLI-4 code on the ITER related activities have concentrated on the first wall loading, the reactor dosimetry, the nuclear heating, and the tritium breeding ratio. To enhance the TRIPOLI-4 verification and validation on neutron-gamma coupled calculations for fusion device application, the computational ITER shielding benchmark of M. E. Sawan was performed in this work by using the 2013 released TRIPOLI-4.9S code and the associated CEA-V5.1.1 data library. First wall, blanket, vacuum vessel and toroidal field magnet of the inboard and outboard components were fully modelled in this 1-D toroidal cylindrical benchmark. The 14.1 MeV source neutrons were sampled from a uniform isotropic distribution in the plasma zone. Nuclear responses including neutron and gamma fluxes, nuclear heating, and material damage indicator were benchmarked against previous published results. The capabilities of the TRIPOLI-4 code on the evaluation of above physics parameters were presented. The nuclear data library from the new FENDL-3.0 evaluation was also benchmarked against the CEA-V5.1.1 results for the neutron transport calculations. The results show that both data libraries can be

Shielded transient self-interaction of a bunch entering a circle from a straight path

International Nuclear Information System (INIS)

Li, R.; Bohn, C.L.; Bisognano, J.J.

1997-01-01

Recent developments in electron-gun and injector technologies enable production of short (mm-length), high-charge (nC-regime) bunches. In this parameter regime, the curvature effect on the bunch self-interaction, by way of coherent synchrotron radiation (CSR) and space-charge forces as the beam traverses magnet bends, may cause serious emittance degradation. In this paper, the authors study an electron bunch orbiting between two infinite, parallel conducting plates. The bunch moves on a trajectory from a straight path to a circular orbit and begins radiating. Transient effects, arising from CSR and space-charge forces generated from source particles both on the bend and on the straight path prior to the bend, are analyzed using Lienard-Wiechert fields, and their overall net effect is obtained. The influence of the plates on the transients is contrasted to their shielding of the steady-state radiated power. Results for emittance degradation induced by this self-interaction are also presented

SCALE6.1 Hybrid Shielding Methodology For The Spent Fuel Dry Storage

International Nuclear Information System (INIS)

Matijevic, M.; Pevec, D.; Trontl, K.

2015-01-01

The SCALE6.1/MAVRIC hybrid deterministic-stochastic shielding methodology was used for dose rates calculation of the generic spent fuel dry storage installation. The neutron-gamma dose rates around the cask array were calculated over a large problem domain in order to determine the boundary of the controlled area. The FW-CADIS methodology, based on the deterministic forward and adjoint solution over the phase - space, was used for optimized, global Monte Carlo results over the mesh tally. The cask inventory was modeled as homogenized material corresponding to 20 fuel assemblies from a standard mid - sized PWR reactor. The global simulation model was an array of 32 casks in 2 rows with concrete foundations and external air, which makes a large spatial domain for shielding calculations. The dose rates around the casks were determined using FW-CADIS method with weighted adjoint source and mesh tally covering a portion of spatial domain of interest. The conservatively obtained dose rates give the upper boundary, since the activation reduction of sources was not taken into account when sequential filling of the dry storage will start. The effective area of the dry storage installation can be additionally reduced with lowering concrete foundation under the ground, embankment raising, and with extra concrete walls, that would additionally lower the dominant gamma dose rates. (author).

Resonance self-shielding methodology of new neutron transport code STREAM

International Nuclear Information System (INIS)

Choi, Sooyoung; Lee, Hyunsuk; Lee, Deokjung; Hong, Ser Gi

2015-01-01

This paper reports on the development and verification of three new resonance self-shielding methods. The verifications were performed using the new neutron transport code, STREAM. The new methodologies encompass the extension of energy range for resonance treatment, the development of optimum rational approximation, and the application of resonance treatment to isotopes in the cladding region. (1) The extended resonance energy range treatment has been developed to treat the resonances below 4 eV of three resonance isotopes and shows significant improvements in the accuracy of effective cross sections (XSs) in that energy range. (2) The optimum rational approximation can eliminate the geometric limitations of the conventional approach of equivalence theory and can also improve the accuracy of fuel escape probability. (3) The cladding resonance treatment method makes it possible to treat resonances in cladding material which have not been treated explicitly in the conventional methods. These three new methods have been implemented in the new lattice physics code STREAM and the improvement in the accuracy of effective XSs is demonstrated through detailed verification calculations. (author)

Determination of the optical properties of turbid media from a single Monte Carlo simulation

International Nuclear Information System (INIS)

Kienle, A.; Patterson, M.S.

1996-01-01

We describe a fast, accurate method for determination of the optical coefficients of 'semi-infinite' and 'infinite' turbid media. For the particular case of time-resolved reflectance from a biological medium, we show that a single Monte Carlo simulation can be used to fit the data and to derive the absorption and reduced scattering coefficients. Tests with independent Monte Carlo simulations showed that the errors in the deduced absorption and reduced scattering coefficients are smaller than 1% and 2%, respectively. (author)

Evaluating secondary neutron doses of a refined shielded design for a medical cyclotron using the TLD approach

International Nuclear Information System (INIS)

Lin, Jye-Bin; Tseng, Hsien-Chun; Liu, Wen-Shan; Lin, Ding-Bang; Hsieh, Teng-San; Chen, Chien-Yi

2013-01-01

An increasing number of cyclotrons at medical centers in Taiwan have been installed to generate radiopharmaceutical products. An operating cyclotron generates immense amounts of secondary neutrons from reactions such the 18 O(p, n) 18 F, used in the production of FDG. This intense radiation can be hazardous to public health, particularly to medical personnel. To increase the yield of 18 F-FDG from 4200 GBq in 2005 to 48,600 GBq in 2011, Chung Shan Medical University Hospital (CSMUH) has prolonged irradiation time without changing the target or target current to meet requirements regarding the production 18 F. The CSMUH has redesigned the CTI Radioisotope Delivery System shield. The lack of data for a possible secondary neutron doses has increased due to newly designed cyclotron rooms. This work aims to evaluate secondary neutron doses at a CTI cyclotron center using a thermoluminescent dosimeter (TLD-600). Two-dimensional neutron doses were mapped and indicated that neutron doses were high as neutrons leaked through self-shielded blocks and through the L-shaped concrete shield in vault rooms. These neutron doses varied markedly among locations close to the H 2 18 O target. The Monte Carlo simulation and minimum detectable dose are also discussed and demonstrated the reliability of using the TLD-600 approach. Findings can be adopted by medical centers to identify radioactive hot spots and develop radiation protection. - Highlights: • Neutron doses were verified using TLD approach. • Neutron doses were increased at cyclotron centers. • Revised L-shaped shield suppresses effectively the neutrons. • Neutron dose can be attenuated to 1.13×10 6 %

Determination of the neutron energy and spatial distributions of the neutron beam from the TSR-II in the large beam shield

International Nuclear Information System (INIS)

Clifford, C.E.; Muckenthaler, F.J.

1976-01-01

The TSR-II reactor of the ORNL Tower Shielding Facility has recently been relocated within a new, fixed shield. A principal feature of the new shield is a beam port of considerably larger area than that of its predecessor. The usable neutron flux has thereby been increased by a factor of approximately 200. The bare beam neutron spectrum behind the new shield has been experimentally determined over the energy range from 0.8 to 16 MeV. A high level of fission product gamma ray background prevented measurement of bare beam spectra below 0.8 MeV, however neutron spectra in the energy range from 8 keV to 1.4 MeV were obtained for two simple, calculable shielding configurations. Also measured in the present work were weighted integral flux distributions and fast neutron dose rates

NOTE: Monte Carlo evaluation of kerma in an HDR brachytherapy bunker

Science.gov (United States)

Pérez-Calatayud, J.; Granero, D.; Ballester, F.; Casal, E.; Crispin, V.; Puchades, V.; León, A.; Verdú, G.

2004-12-01

In recent years, the use of high dose rate (HDR) after-loader machines has greatly increased due to the shift from traditional Cs-137/Ir-192 low dose rate (LDR) to HDR brachytherapy. The method used to calculate the required concrete and, where appropriate, lead shielding in the door is based on analytical methods provided by documents published by the ICRP, the IAEA and the NCRP. The purpose of this study is to perform a more realistic kerma evaluation at the entrance maze door of an HDR bunker using the Monte Carlo code GEANT4. The Monte Carlo results were validated experimentally. The spectrum at the maze entrance door, obtained with Monte Carlo, has an average energy of about 110 keV, maintaining a similar value along the length of the maze. The comparison of results from the aforementioned values with the Monte Carlo ones shows that results obtained using the albedo coefficient from the ICRP document more closely match those given by the Monte Carlo method, although the maximum value given by MC calculations is 30% greater.

Determining optical and radiation characteristics of cathode ray tubes' glass to be reused as radiation shielding glass

Science.gov (United States)

Zughbi, A.; Kharita, M. H.; Shehada, A. M.

2017-07-01

A new method of recycling glass of Cathode Ray Tubes (CRTs) has been presented in this paper. The glass from CRTs suggested being used as raw materials for the production of radiation shielding glass. Cathode ray tubes glass contains considerable amounts of environmentally hazardous toxic wastes, namely heavy metal oxides such as lead oxide (PbO). This method makes CRTs glass a favorable choice to be used as raw material for Radiation Shielding Glass and concrete. The heavy metal oxides increase its density, which make this type of glass nearly equivalent to commercially available shielding glass. CRTs glass have been characterized to determine heavy oxides content, density, refractive index, and radiation shielding properties for different Gamma-Ray energies. Empirical methods have been used by using the Gamma-Ray source cobalt-60 and computational method by using the code XCOM. Measured and calculated values were in a good compatibility. The effects of irradiation by gamma rays of cobalt-60 on the optical transparency for each part of the CRTs glass have been studied. The Results had shown that some parts of CRTs glass have more resistant to Gamma radiation than others. The study had shown that the glass of cathode ray tubes could be recycled to be used as radiation shielding glass. This proposed use of CRT glass is only limited to the available quantity of CRT world-wide.

Shielding structure analysis for LSDS facility

Energy Technology Data Exchange (ETDEWEB)

Choi, Hong Yeop; Kim, Jeong Dong; Lee, Yong Deok; Kim, Ho Dong [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2014-05-15

The nuclear material (Pyro, Spent nuclear fuel) itself and the target material to generate neutrons is the LSDS system for isotopic fissile assay release of high intensity neutron and gamma rays. This research was performed to shield from various strong radiation. A shielding evaluation was carried out with a facilities model of LSDS system. The MCNPX 2.5 code was used and a shielding evaluation was performed for the shielding structure and location. The radiation dose based on the hole structure and location of the wall was evaluated. The shielding evaluation was performed to satisfy the safety standard for a normal person (1 μSv/h) and to use enough interior space. The MCNPX2.5 code was used and a dose evaluation was performed for the location of the shielding material, shielding structure, and hole structure. The evaluation result differs according to the shielding material location. The dose rate was small when the shielding material was positioned at the center. The dose evaluation result regarding the location of the shielding material was applied to the facility and the shielding thickness was determined (In 50 cm + Borax 5 cm + Out 45cm). In the existing hole structure, the radiation leak is higher than the standard. A hole structure model to prevent leakage of radiation was proposed. The general public dose limit was satisfied when using the concrete reinforcement and a zigzag structure. The shielding result will be of help to the facility shielding optimization.

Shielding structure analysis for LSDS facility

International Nuclear Information System (INIS)

Choi, Hong Yeop; Kim, Jeong Dong; Lee, Yong Deok; Kim, Ho Dong

2014-01-01

The nuclear material (Pyro, Spent nuclear fuel) itself and the target material to generate neutrons is the LSDS system for isotopic fissile assay release of high intensity neutron and gamma rays. This research was performed to shield from various strong radiation. A shielding evaluation was carried out with a facilities model of LSDS system. The MCNPX 2.5 code was used and a shielding evaluation was performed for the shielding structure and location. The radiation dose based on the hole structure and location of the wall was evaluated. The shielding evaluation was performed to satisfy the safety standard for a normal person (1 μSv/h) and to use enough interior space. The MCNPX2.5 code was used and a dose evaluation was performed for the location of the shielding material, shielding structure, and hole structure. The evaluation result differs according to the shielding material location. The dose rate was small when the shielding material was positioned at the center. The dose evaluation result regarding the location of the shielding material was applied to the facility and the shielding thickness was determined (In 50 cm + Borax 5 cm + Out 45cm). In the existing hole structure, the radiation leak is higher than the standard. A hole structure model to prevent leakage of radiation was proposed. The general public dose limit was satisfied when using the concrete reinforcement and a zigzag structure. The shielding result will be of help to the facility shielding optimization

Shielding design for the front end of the CERN SPL.

Science.gov (United States)

Magistris, Matteo; Silari, Marco; Vincke, Helmut

2005-01-01

CERN is designing a 2.2-GeV Superconducting Proton Linac (SPL) with a beam power of 4 MW, to be used for the production of a neutrino superbeam. The SPL front end will initially accelerate 2 x 10(14) negative hydrogen ions per second up to an energy of 120 MeV. The FLUKA Monte Carlo code was employed for shielding design. The proposed shielding is a combined iron-concrete structure, which also takes into consideration the required RF wave-guide ducts and access labyrinths to the machine. Two beam-loss scenarios were investigated: (1) constant beam loss of 1 Wm(-1) over the whole accelerator length and (2) full beam loss occurring at various locations. A comparison with results based on simplified approaches is also presented.

Prospect on general software of Monte Carlo method

International Nuclear Information System (INIS)

Pei Lucheng

1992-01-01

This is a short paper on the prospect of Monte Carlo general software. The content consists of cluster sampling method, zero variance technique, self-improved method, and vectorized Monte Carlo method

Optimization of multi-layered metallic shield

International Nuclear Information System (INIS)

Ben-Dor, G.; Dubinsky, A.; Elperin, T.

2011-01-01

Research highlights: → We investigated the problem of optimization of a multi-layered metallic shield. → The maximum ballistic limit velocity is a criterion of optimization. → The sequence of materials and the thicknesses of layers in the shield are varied. → The general problem is reduced to the problem of Geometric Programming. → Analytical solutions are obtained for two- and three-layered shields. - Abstract: We investigate the problem of optimization of multi-layered metallic shield whereby the goal is to determine the sequence of materials and the thicknesses of the layers that provide the maximum ballistic limit velocity of the shield. Optimization is performed under the following constraints: fixed areal density of the shield, the upper bound on the total thickness of the shield and the bounds on the thicknesses of the plates manufactured from every material. The problem is reduced to the problem of Geometric Programming which can be solved numerically using known methods. For the most interesting in practice cases of two-layered and three-layered shields the solution is obtained in the explicit analytical form.

Nuclear characteristics of epoxy resin as a space environment neutron shielding

Energy Technology Data Exchange (ETDEWEB)

Adeli, Ruhollah [Nuclear Science and Technology Research Institute, Yazd (Iran, Islamic Republic of). Central Iran Research Complex; Shirmardi, Seyed Pezhman [Nuclear Science and Technology Research Institute, Tehran (Iran, Islamic Republic of). Radiation Application Research School; Mazinani, Saideh [Amirkabir Nanotechnology Research Institute, Tehran (Iran, Islamic Republic of); Ahmadi, Seyed Javad [Nuclear Science and Technology Research Institute, Tehran (Iran, Islamic Republic of). Nuclear Fuel Cycle Research School

2017-03-15

In recent years many investigations have been done for choosing applicable light neutron shielding in space environmental applications. In this study, we have considered the neutron radiation-protective characteristics of neat epoxy resin, a thermoplastic polymer material and have compared it with various candidate materials in neutron radiation protection such as Al 6061 alloy and Polyethylene. The aim of this investigation is the effect of type of moderator for fast neutron, notwithstanding neutron absorbers fillers. The nuclear interactions and the effective dose at shields have been studied with the Monte Carlo N-Particle transport code (MCNP), using variance reductions to reduce the relative error. Among the candidates, polymer matrix showed a better performance in attenuating fast neutrons and caused a lower neutron and secondary photon effective dose.

Influence of preheating on API 5L-X80 pipeline joint welding with self shielded flux-cored wire

International Nuclear Information System (INIS)

Cooper, R.; Silva, J. H. F.; Trevisan, R. E.

2004-01-01

The present work refers to the characterization of API 5L-X80 pipeline joints welded with self-shielded flux cored wire. This process was evaluated under preheating conditions, with an uniform and steady heat input. All joints were welded in flat position (1G), with the pipe turning and the torch still. Tube dimensions were 762 mm in external diameter and 16 mm in thickness. Welds were applied on single V-groove, with six weld beads, along with three levels of preheating temperatures (room temperature, 100 degree centigree, 160 degree centigree). These temperatures were maintained as inter pass temperature. The filler metal E71T8-K6 with mechanical properties different from parent metal was used in under matched conditions. The weld characterization is presented according to the mechanical test results of tensile strength, hardness and impact test. The mechanical tests were conducted according to API 1104, AWS and ASTM standards. API 1104 and API 51 were used as screening criteria. According to the results obtained, it was possible to remark that it is appropriate to weld API 5L-X80 steel ducts with Self-shielded Flux Cored wires, in conformance to the API standards and no preheat temperature is necessary. (Author) 22 refs

Reclaiming Self-Determination from the Indian Self-Determination and Education Assistance Act of 1975

Science.gov (United States)

Wilson, Michael D.

2012-01-01

This paper examines the way the term "self-determination" is used in the Indian Self-Determination and Education Assistance Act of 1975. Its main thesis is that the Act does not in fact offer tribal governments self-determination, but instead reaffirms old power configurations that go back to the Indian Reorganization Act of 1934.…

SU-C-BRB-06: Utilizing 3D Scanner and Printer for Dummy Eye-Shield: Artifact-Free CT Images of Tungsten Eye-Shield for Accurate Dose Calculation

International Nuclear Information System (INIS)

Park, J; Lee, J; Kim, H; Kim, I; Ye, S

2015-01-01

Purpose: To evaluate the effect of a tungsten eye-shield on the dose distribution of a patient. Methods: A 3D scanner was used to extract the dimension and shape of a tungsten eye-shield in the STL format. Scanned data was transferred into a 3D printer. A dummy eye shield was then produced using bio-resin (3D systems, VisiJet M3 Proplast). For a patient with mucinous carcinoma, the planning CT was obtained with the dummy eye-shield placed on the patient’s right eye. Field shaping of 6 MeV was performed using a patient-specific cerrobend block on the 15 x 15 cm 2 applicator. The gantry angle was 330° to cover the planning target volume near by the lens. EGS4/BEAMnrc was commissioned from our measurement data from a Varian 21EX. For the CT-based dose calculation using EGS4/DOSXYZnrc, the CT images were converted to a phantom file through the ctcreate program. The phantom file had the same resolution as the planning CT images. By assigning the CT numbers of the dummy eye-shield region to 17000, the real dose distributions below the tungsten eye-shield were calculated in EGS4/DOSXYZnrc. In the TPS, the CT number of the dummy eye-shield region was assigned to the maximum allowable CT number (3000). Results: As compared to the maximum dose, the MC dose on the right lens or below the eye shield area was less than 2%, while the corresponding RTP calculated dose was an unrealistic value of approximately 50%. Conclusion: Utilizing a 3D scanner and a 3D printer, a dummy eye-shield for electron treatment can be easily produced. The artifact-free CT images were successfully incorporated into the CT-based Monte Carlo simulations. The developed method was useful in predicting the realistic dose distributions around the lens blocked with the tungsten shield

SU-C-BRB-06: Utilizing 3D Scanner and Printer for Dummy Eye-Shield: Artifact-Free CT Images of Tungsten Eye-Shield for Accurate Dose Calculation

Energy Technology Data Exchange (ETDEWEB)

Park, J; Lee, J [Program in Biomedical Radiation Sciences, Department of Transdisciplinary Studies, Graduate School of Convergence Science and Technology, Seoul National University, Seoul (Korea, Republic of); Institute of Radiation Medicine, Seoul National University Medical Research Center, Seoul (Korea, Republic of); Kim, H [Institute of Radiation Medicine, Seoul National University Medical Research Center, Seoul (Korea, Republic of); Interdisciplinary Program in Radiation Applied Life Science, Seoul National University College of Medicine, Seoul (Korea, Republic of); Kim, I [Department of Radiation Oncology, Seoul National University Hospital, Seoul (Korea, Republic of); Interdisciplinary Program in Radiation Applied Life Science, Seoul National University College of Medicine, Seoul (Korea, Republic of); Ye, S [Program in Biomedical Radiation Sciences, Department of Transdisciplinary Studies, Graduate School of Convergence Science and Technology, Seoul National University, Seoul (Korea, Republic of); Institute of Radiation Medicine, Seoul National University Medical Research Center, Seoul (Korea, Republic of); Department of Radiation Oncology, Seoul National University Hospital, Seoul (Korea, Republic of); Interdisciplinary Program in Radiation Applied Life Science, Seoul National University College of Medicine, Seoul (Korea, Republic of); Advanced Institutes of Convergence Technology, Seoul National University, Suwon (Korea, Republic of)

2015-06-15

Purpose: To evaluate the effect of a tungsten eye-shield on the dose distribution of a patient. Methods: A 3D scanner was used to extract the dimension and shape of a tungsten eye-shield in the STL format. Scanned data was transferred into a 3D printer. A dummy eye shield was then produced using bio-resin (3D systems, VisiJet M3 Proplast). For a patient with mucinous carcinoma, the planning CT was obtained with the dummy eye-shield placed on the patient’s right eye. Field shaping of 6 MeV was performed using a patient-specific cerrobend block on the 15 x 15 cm{sup 2} applicator. The gantry angle was 330° to cover the planning target volume near by the lens. EGS4/BEAMnrc was commissioned from our measurement data from a Varian 21EX. For the CT-based dose calculation using EGS4/DOSXYZnrc, the CT images were converted to a phantom file through the ctcreate program. The phantom file had the same resolution as the planning CT images. By assigning the CT numbers of the dummy eye-shield region to 17000, the real dose distributions below the tungsten eye-shield were calculated in EGS4/DOSXYZnrc. In the TPS, the CT number of the dummy eye-shield region was assigned to the maximum allowable CT number (3000). Results: As compared to the maximum dose, the MC dose on the right lens or below the eye shield area was less than 2%, while the corresponding RTP calculated dose was an unrealistic value of approximately 50%. Conclusion: Utilizing a 3D scanner and a 3D printer, a dummy eye-shield for electron treatment can be easily produced. The artifact-free CT images were successfully incorporated into the CT-based Monte Carlo simulations. The developed method was useful in predicting the realistic dose distributions around the lens blocked with the tungsten shield.

Shielding Calculations for Industrial 5/7.5MeV Electron Accelerators Using the MCNP Monte Carlo Code

International Nuclear Information System (INIS)

Peri, E.; Orion, I.

2014-01-01

High energy X-rays from accelerators are used to irradiate food ingredients to prevent growth and development of unwanted biological organisms in food, in order to extend the shelf life of products. High energy photons can cause food activation due to (D 3 ,n) reactions. Until 2004, to eliminate the possibility of food activation, the electron energy was limited to 5 MeV X-rays for food irradiation. In 2004, the FDA approved the usage of up to 7.5 MeV, but only with tantalum and gold targets (1). Higher X-ray energy results an increased flux of X-rays in the forward direction, increased penetration, and higher photon dose rate due to better electron-to-photon conversion. These improvements could decrease the irradiation time and allow irradiation of larger packages, thereby providing higher production rates with lower treatment cost. Medical accelerators usually work with 6-18 MV electron energy with tungsten target to convert the electron beam to X-rays. In order to protect the patients, the accelerator head is protected with a heavy lead shielding; therefore, the bremsstrahlung is emitted only in the forward direction. There are many publications and standards that guide how to design optimal shielding for medical accelerator rooms. The shielding data for medical accelerators is not applicable for industrial accelerators, since the data is for different conversion targets, different X-Ray energies, and only for the forward direction. Collimators are not always in use in industrial accelerators, and therefore bremsstrahlung photons can be emitted in all directions. The bremsstrahlung spectrum and dose rate change as a function of the emission angle. The dose rate decreases from maximum in the forward direction (0°) to minimum at 180° by 1-2 orders of magnitude. In order to design and calculate optimal shielding for food accelerator rooms, there is a need to have the bremsstrahlung spectrum data, dose rates and concrete attenuation data in all emission directions

Monte Carlo Study of Four-Dimensional Self-avoiding Walks of up to One Billion Steps

Science.gov (United States)

Clisby, Nathan

2018-04-01

We study self-avoiding walks on the four-dimensional hypercubic lattice via Monte Carlo simulations of walks with up to one billion steps. We study the expected logarithmic corrections to scaling, and find convincing evidence in support the scaling form predicted by the renormalization group, with an estimate for the power of the logarithmic factor of 0.2516(14), which is consistent with the predicted value of 1/4. We also characterize the behaviour of the pivot algorithm for sampling four dimensional self-avoiding walks, and conjecture that the probability of a pivot move being successful for an N-step walk is O([ log N ]^{-1/4}).

CHARGE-2/C, Flux and Dose Behind Shield from Electron, Proton, Heavy Particle Irradiation

International Nuclear Information System (INIS)

Ucker, W.R.; Lilley, J.R.

1994-01-01

1 - Description of problem or function: The CHARGE code computes flux spectra, dose and other response rates behind a multilayered spherical or infinite planar shield exposed to isotopic fluxes of electrons, protons and heavy charged particles. The doses, or other responses, to electron, primary proton, heavy particle, electron Bremsstrahlung, secondary proton, and secondary neutron radiations are calculated as a function of penetration into the shield; the materials of each layer may be mixtures of elements contained in the accompanying data library, or supplied by the user. The calculation may optionally be halted before the entire shield is traversed by specifying a minimum total dose rate; the computation stops when the dose drops below this value. The ambient electron, proton and heavy particle spectra may be specified in tabular or functional form. These incident charged particle spectra are divided into energy bands or groups, the number or spacing of which are controlled by input data. The variation of the group boundary energies and group spectra as a function of shield penetration uniquely determines charged particle dose rates and secondary particle production rates. The charged particle shielding calculation is essentially the integration of the range- energy equation which expresses the variation of particle energy wit distance travelled. 2 - Method of solution: The 'straight-ahead' approximation is used throughout, that is the changes in particle direction of motion due to elastic scattering are ignored. This approximation is corrected, in the case of electrons, by applying transmission factors obtained from Monte Carlo calculations. Inelastic scattering between protons and the shielding material is assumed to produce two classes of secondaries 1) Cascade protons and neutrons, emitted in the same direction as the primaries 2) Evaporation neutrons, emitted isotropically. The transmission of secondary protons is analyzed in exactly the same way as the

Design of the magnetized muon shield for the prompt-neutrino facility

International Nuclear Information System (INIS)

Baltay, C.; Bosek, N.; Couch, J.

1982-01-01

The main technical challenge in the design of the prompt neutrino beam is the magnetized muon shield. Two satisfactory alternate designs have been developed for such a shield during this past year and the background muon fluxes have been calculated by three independent programs at Columbia, Fermilab, and MIT. The background muon fluxes have been calculated to be satisfactory in all of the detectors that might use the beam. In Section III of this report we describe in detail the three Monte Carlo programs used in these calculations. In Section IV we give the details of the flux calculations for the E-613 shield and the comparisons with the observed fluxes with various configurations of that shield. In Section V we describe the designs that have been developed for the neutrino area shield. In Section VI we discuss the problem of proton beam transport losses and the associated muon fluxes. Finally, in Section VII a comparison of the two solutions is made which covers cost, effectiveness, schedule and responsiveness to future unknowns. We conclude that there are not overwhelming reasons for the choice of one design over the other. However, for a variety of secondary reasons the superconducting design offers advantages. We therefore propose the construction of the prompt neutrino facility with the superconducting magnet design

Determination of half-value layers and tenth-value layer to barite as shielding against X radiation in radiological protection

International Nuclear Information System (INIS)

Lopes, G.A.; Aragao Filho, G.L.; Almeida Junior, A.T.; Santos, M.A.P.; Araujo, F.G.S.; Nogueira, M.S.

2013-01-01

The barium mortar has been widely used as radiation shielding material for X and gamma radiations in Brazil, by presenting some advantages as the high rate of efficiency in radiation shielding, the easy handling and application, the facility to be found in the national market and low cost. The determination of the half-value layers (HVL) and tenth-value layer (TVL) of different types of barite becomes the major factor to characterize the attenuation of these materials, in order to ensure the efficiency and quality of projects shielding, by ensuring the safety of workers occupationally exposed to radiation and of individuals to the public. Thus, plates of different thickness of mortar of barite were made for determination of their HVL and TVL. The plates were irradiated with X-ray qualities for radiological protection according to standard ISO 4037. A system of CdTe spectrometry was used to acquire spectra transmitted, in the presence of each plate, and their combinations. The areas of the spectra obtained, depending on the total thickness of the plates used in the arrangement were used to determine the attenuation curves. From these curves obtained in this work was to establish the HVL and TVL

75 FR 57519 - Weather Shield Manufacturing, Medford, WI; Notice of Negative Determination Regarding Application...

Science.gov (United States)

2010-09-21

... DEPARTMENT OF LABOR Employment and Training Administration [TA-W-72,673] Weather Shield...), applicable to workers and former workers of Weather Shield Manufacturing, Inc., Medford, Wisconsin (subject... administrative support services related to the production of doors and windows at various Weather Shield...

Determination of true coincidence correction factors using Monte-Carlo simulation techniques

Directory of Open Access Journals (Sweden)

Chionis Dionysios A.

2014-01-01

Full Text Available Aim of this work is the numerical calculation of the true coincidence correction factors by means of Monte-Carlo simulation techniques. For this purpose, the Monte Carlo computer code PENELOPE was used and the main program PENMAIN was properly modified in order to include the effect of the true coincidence phenomenon. The modified main program that takes into consideration the true coincidence phenomenon was used for the full energy peak efficiency determination of an XtRa Ge detector with relative efficiency 104% and the results obtained for the 1173 keV and 1332 keV photons of 60Co were found consistent with respective experimental ones. The true coincidence correction factors were calculated as the ratio of the full energy peak efficiencies was determined from the original main program PENMAIN and the modified main program PENMAIN. The developed technique was applied for 57Co, 88Y, and 134Cs and for two source-to-detector geometries. The results obtained were compared with true coincidence correction factors calculated from the "TrueCoinc" program and the relative bias was found to be less than 2%, 4%, and 8% for 57Co, 88Y, and 134Cs, respectively.

ICRF antenna Faraday shield plasma sheath model

International Nuclear Information System (INIS)

Whealton, J.H.; Ryan, P.M.; Raridon, R.J.

1990-01-01

A two-dimensional nonlinear formulation that explicitly considers the plasma edge near a Faraday shield in a self-consistent manner is used in the modeling of the ion motion for a Faraday shield concept and model suggested by Perkins. Two models are considered that may provide significant insight into the generation of impurities for ion cyclotron resonance heating (ICRH) antennas. In one of these models a significant sheath periodically forms next to the Faraday screen, with ion acoustic waves heating the ions in the plasma. (orig.)

Hierarchy Formation and Self-Determination

Directory of Open Access Journals (Sweden)

Stefano I. Di Domenico

2014-12-01

Full Text Available We examined how self-determination, the subjective experience of one’s behavior as internally initiated and personally endorsed, depends on one’s standing in real-world social hierarchies. We predicted that those with the traits most relevant to status attainment would be those afforded the most opportunities to be self-determining. We examined the trait of physical attractiveness, given its documented association with social status and no known association with self-determination. First-year undergraduates living in same-sex residences rated their housemates’ social status, while an independent set of observers rated the participants’ physical attractiveness. Consistent with prediction, physically attractive individuals attained the highest levels of social status; in turn, those who attained the highest levels of social status experienced the highest levels of self-determination. These findings provide new insights into self-determination as an inherently relational phenomenon and specifically highlight the formative influence of social status on people’s capacities for self-determination.

Evaluation of the room shielding thickness of Hi-Art tomotherapy system

International Nuclear Information System (INIS)

Liu Haikuan; Wu Jinhai; Gu Naigu; Gao Yiming; Wang Li; Huang Weiqin; Wang Fengxian

2010-01-01

In this paper, we calculate and evaluate the room shielding thickness of a Hi-Art tomotherapy system, which is a new type of radiotherapy facility. Due to the self-shielding of the accelerator,only scattered beam and beam leakage were considered in calculating the room shielding thickness. The radiation field of the tomotherapy system was used as the basic data to calculate the shielding thickness of every 15 degree solid angle. The maximum shielding thickness required of each shielding wall was at the position with the angle of 15 degree, and the calculated shielding thickness were 1023, 975, 917, 1460, 1147 and 1189 mm for the east wall,south wall,west wall, north wall, the roof and the floor,respectively. According to the calculation results, all shielding walls, ceiling and floor could meet the requirement of the radiation protection, but the north wall thickness of 1200 mm was a little thinner. (authors)

A comparative study for different shielding material composition and beam geometry applied to PET facilities: simulated transmission curves

OpenAIRE

Hoff, Gabriela; Costa, Paulo Roberto

2013-01-01

The aim of this work is to simulate transmission data for different beam geometry and material composition in order to evaluate the effect of these parameters on transmission curves. The simulations are focused on outgoing spectra for shielding barriers used in PET facilities. The behavior of the transmission was evaluated as a function of the shielding material composition and thickness using Geant4 Monte Carlo code, version 9.2 p 03.The application was benchmarked for barited mortar and com...

CAD-Based Shielding Analysis for ITER Port Diagnostics

Directory of Open Access Journals (Sweden)

Serikov Arkady

2017-01-01

Full Text Available Radiation shielding analysis conducted in support of design development of the contemporary diagnostic systems integrated inside the ITER ports is relied on the use of CAD models. This paper presents the CAD-based MCNP Monte Carlo radiation transport and activation analyses for the Diagnostic Upper and Equatorial Port Plugs (UPP #3 and EPP #8, #17. The creation process of the complicated 3D MCNP models of the diagnostics systems was substantially accelerated by application of the CAD-to-MCNP converter programs MCAM and McCad. High performance computing resources of the Helios supercomputer allowed to speed-up the MCNP parallel transport calculations with the MPI/OpenMP interface. The found shielding solutions could be universal, reducing ports R&D costs. The shield block behind the Tritium and Deposit Monitor (TDM optical box was added to study its influence on Shut-Down Dose Rate (SDDR in Port Interspace (PI of EPP#17. Influence of neutron streaming along the Lost Alpha Monitor (LAM on the neutron energy spectra calculated in the Tangential Neutron Spectrometer (TNS of EPP#8. For the UPP#3 with Charge eXchange Recombination Spectroscopy (CXRS-core, an excessive neutron streaming along the CXRS shutter, which should be prevented in further design iteration.

CAD-Based Shielding Analysis for ITER Port Diagnostics

Science.gov (United States)

Serikov, Arkady; Fischer, Ulrich; Anthoine, David; Bertalot, Luciano; De Bock, Maartin; O'Connor, Richard; Juarez, Rafael; Krasilnikov, Vitaly

2017-09-01

Radiation shielding analysis conducted in support of design development of the contemporary diagnostic systems integrated inside the ITER ports is relied on the use of CAD models. This paper presents the CAD-based MCNP Monte Carlo radiation transport and activation analyses for the Diagnostic Upper and Equatorial Port Plugs (UPP #3 and EPP #8, #17). The creation process of the complicated 3D MCNP models of the diagnostics systems was substantially accelerated by application of the CAD-to-MCNP converter programs MCAM and McCad. High performance computing resources of the Helios supercomputer allowed to speed-up the MCNP parallel transport calculations with the MPI/OpenMP interface. The found shielding solutions could be universal, reducing ports R&D costs. The shield block behind the Tritium and Deposit Monitor (TDM) optical box was added to study its influence on Shut-Down Dose Rate (SDDR) in Port Interspace (PI) of EPP#17. Influence of neutron streaming along the Lost Alpha Monitor (LAM) on the neutron energy spectra calculated in the Tangential Neutron Spectrometer (TNS) of EPP#8. For the UPP#3 with Charge eXchange Recombination Spectroscopy (CXRS-core), an excessive neutron streaming along the CXRS shutter, which should be prevented in further design iteration.

Accuracy evaluation of the current data and method applied to shielding design of the Fusion Experimental Reactor (FER)

International Nuclear Information System (INIS)

Mori, Seiji; Kobayashi, Takeshi; Seki, Yasushi

1988-06-01

Shielding benchmarking study of the current data and method applied to the Fusion Experimental Reactor (FER) was performed. First, neutron and gamma ray fluxes were calculated by the one-dimensional S N code using various cross section libraries and the continuous energy Monte Carlo code. The results were compared in terms of the S N /MC ratio. The worst ratios are about 0.5 and 0.25 for neutron flux and gamma ray flux, respectively. Next, the analytical calculations of the iron sphere transmission experiment of 14 MeV neutrons were performed to examine the accuracy of cross section data of iron, which is the most important material of shield. The E/C ratio is larger than 2 even if the continuous energy Monte Carlo code was used. Thirdly, the influence of geometrical representation of the shield was investigated by comparing the homogeneous model and the heterogeneous model (alternating layers of SS316 and water). As a result, it was made clear that the homogeneous model underestimates neutron flux by a factor of 2. Finally, the necessity of benchmark experiment and improvement of cross section library was pointed out as the further R and D issues. (author)

Monte Carlo simulations and experimental results on neutron production in the spallation target QUINTA irradiated with 660 MeV protons