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At least 775 records · Page 43Linked to original sources
PubMed2005

Radiation impact caused by activation of air from the future GSI accelerator facility fair.

The Gesellschaft für Schwerionenforschung in Darmstadt is planning a new accelerator Facility for Antiproton and Ion Research (FAIR). Two future experimental areas are regarded to be the most decisive points concerning the activation of air. One is the area for the production of antiprotons. A second crucial experimental area is the so-called Super Fragment Separator. The production of radioactive isotopes in air is calculated using the residual nuclei option of the Monte Carlo program FLUKA. The results are compared with the data for the activation of air given by Sullivan and in IAEA report 283. The resulting effective dose is calculated using a program package from the German Federal Office for Radiation Protection, the Bundesamt für Stranlenschutz. The results demonstrate that a direct emission of the total radioactivity produced into the air will probably conflict with the limits of the German Radiation Protection Ordinance. Special measures have to be planned in order to reduce the amount of radioactivity released into the air.

Air↗
PubMed2005

Radiation safety design for the J-PARC Project.

The High-Intensity Proton Accelerator Project, named J-PARC, is in progress, with the aim of enabling studies on the latest basic science and the advancement of nuclear technology. In the project, a high-energy proton accelerator complex with the world's highest instantaneous intensity is under construction. In order to establish a reasonable shielding design, both simplified and detailed design methods were used in the shielding design of J-PARC. This paper reviews the present status of the radiation safety design study for J-PARC.

Computer Simulation↗
PubMed2005

Interim storage of spent and disused sealed sources: optimisation of external dose distribution in waste grids using the MCNPX code.

Radioactive sealed sources are in use worldwide in different fields of application. When no further use is foreseen for these sources, they become spent or disused sealed sources and are subject to a specific waste management scheme. Portugal does have a Radioactive Waste Interim Storage Facility where spent or disused sealed sources are conditioned in a cement matrix inside concrete drums and following the geometrical disposition of a grid. The gamma dose values around each grid depend on the drum's enclosed activity and radionuclides considered, as well as on the drums distribution in the various layers of the grid. This work proposes a method based on the Monte Carlo simulation using the MCNPX code to estimate the best drum arrangement through the optimisation of dose distribution in a grid. Measured dose rate values at 1 m distance from the surface of the chosen optimised grid were used to validate the corresponding computational grid model.

Computer Simulation↗
PubMed2005

An additional performance of HTRS: the waste radiotoxicity minimisation.

The management of radioactive waste is a key issue for the present and future use of nuclear energy. In this frame, high temperature reactors (HTRs) have, among others, the capability to burn actinides. After a short introduction on HTRs, the performances of two MC-based burnup codes (Monte Carlo continuous energy burnup and MONTEBURNS) in assessing the ability of these reactors to burn actinides are compared. These codes are necessary for performing ultra-high burnup calculations on HTRs. The best one, in this specific case, results to be MONTEBURNS. It was analysed using HTRs loaded with the following: (1) 1st generation Pu, 600 equivalent full power days; (2) 2nd generation Pu, 645 equivalent full power days; and (iii) 33% 1st generation Pu and 67% Th, 705 equivalent full power days. Finally, it is possible to conclude that HTRs can reduce time when the waste is considered dangerous. Even if the amount of reduction does not solve the whole problem, it represents an important step in the management of radioactive waste.

Computer Simulation↗
PubMed2005

Calculations of neutron and photon source terms and attenuation profiles for the generic design of the SPEAR3 storage ring shield.

The FLUKA Monte Carlo particle generation and transport code was used to calculate shielding requirements for the 3 GeV, 500 mA SPEAR3 storage ring at the Stanford Synchrotron Radiation Laboratory. The photon and neutron dose equivalent source term data were simulated for a 3 GeV electron beam interacting with two typical target/shielding geometries in the ring. The targets simulated are a rectangular block of 0.7 cm thick copper and a 5 cm thick iron block, both tilted at 1 degree relative to the beam direction. Attenuation profiles for neutrons and photons in concrete and lead as a function of angle at different shield thicknesses were calculated. The first, second and equilibrium attenuation lengths of photons and neutrons in the shield materials are derived from the attenuation profiles. The source term data and the attenuation lengths were then used to evaluate the shielding requirements for the ratchet walls of all front-ends of the SPEAR3 storage ring.

Computer Simulation↗
PubMed2005

SOURCES: a code for calculating (alpha,n), spontaneous fission, and delayed neutron sources and spectra.

SOURCES is a computer code that determines neutron production rates and spectra from (alpha,n) reactions, spontaneous fission and delayed neutron emission owing to the decay of radionuclides in homogeneous media, interface problems and three-region interface problems. The code is also capable of calculating the neutron production rates due to (alpha,n) reactions induced by a monoenergetic beam of alpha particles incident on a slab of target material. The (alpha,n) spectra are calculated using an assumed isotropic angular distribution in the centre-of-mass system with a library of 107 nuclide decay alpha-particle spectra, 24 sets of measured and/or evaluated (alpha,n) cross sections and product nuclide level branching fractions, and functional alpha particle stopping cross sections for Z < 106. Spontaneous fission sources and spectra are calculated with evaluated half-life, spontaneous fission branching and Watt spectrum parameters for 44 actinides. The delayed neutron spectra are taken from an evaluated library of 105 precursors. The code outputs the magnitude and spectra of the resultant neutron sources. It also provides an analysis of the contributions to that source by each nuclide in the problem.

Algorithms↗
PubMed2005

Application of MCBEND to PBMR shielding analysis.

Shielding analysis of an early design of Pebble Bed Modular Reactor (PBMR) has been carried out by using the Monte Carlo code MCBEND. The issues of concern were damage to the core barrel and the reactor pressure vessel (RPV), activation of the core barrel, RPV, top plate and bottom plate, and also burn-up of boron in the control layer underneath the core. The analysis below the core was complicated due to the presence of the de-fuelling chute, which meant that multiplication had to be taken into account. The analysis of boron burn-up was particularly challenging and was tackled using a combination of MCBEND and the criticality code MONK in the depletion mode. The application of MCBEND to the shielding analysis of the PBMR is described, with particular attention being paid to the regions below the core.

Algorithms↗
PubMed2005

Space radiation measurements on-board ISS--the DOSMAP experiment.

The experiment 'Dosimetric Mapping' conducted as part of the science program of NASA's Human Research Facility (HRF) between March and August 2001 was designed to measure integrated total absorbed doses (ionising radiation and neutrons), heavy ion fluxes and its energy, mass and linear energy transfer (LET) spectra, time-dependent count rates of charged particles and their corresponding dose rates at different locations inside the US Lab at the International Space Station. Owing to the variety of particles and energies, a dosimetry package consisting of thermoluminescence dosemeter (TLD) chips and nuclear track detectors with and without converters (NTDPs), a silicon dosimetry telescope (DOSTEL), four mobile silicon detector units (MDUs) and a TLD reader unit (PILLE) with 12 TLD bulbs as dosemeters was used. Dose rates of the ionising part of the radiation field measured with TLD bulbs applying the PILLE readout system at different locations varied between 153 and 231 microGy d(-1). The dose rate received by the active devices fits excellent to the TLD measurements and is significantly lower compared with measurements for the Shuttle (STS) to MIR missions. The comparison of the absorbed doses from passive and active devices showed an agreement within +/- 10%. The DOSTEL measurements in the HRF location yielded a mean dose equivalent rate of 535 microSv d(-1). DOSTEL measurements were also obtained during the Solar Particle Event on 15 April 2001.

Aerospace Medicine↗
PubMed2005

Patient neutron dose equivalent exposures outside of the proton therapy treatment field.

A large fraction of dose to healthy tissue located outside of the treatment field during proton therapy is attributable to neutrons produced in the beam-delivery apparatus. In this work, the neutron dose equivalent (H) per therapeutic proton absorbed dose (D) was estimated for typical treatment conditions as a function of range modulation width, angle with respect to the incident proton beam, and the distance from the isocentre at the Harvard Cyclotron Laboratory's (Cambridge, MA) passively spread treatment field using Monte Carlo simulations. For a beam with 16 cm penetration (depth) and a 5 x 5 cm2 lateral field size at the patient location along the incident beam direction at 100 cm from the isocentre, the predicted H/D values are 0.35 and 0.60 mSv Gy(-1) from the simulations and measurements, respectively. At all locations, the predicted H/D values are within a factor of 2 and 3 of the measured result for no modulation and 8.2 cm of modulation, respectively.

Body Burden↗
PubMed2005

Monte Carlo evaluations and the source normalisation problem.

It is well known that Transport Theory calculations give results which are immediately intelligible, as they take into account the source explicitly. Monte Carlo evaluations on the contrary are given per source particle, and they have to be normalized according to the actual source to be completely intelligible. Unfortunately in MC evaluations sometimes one can be led, to reach good results from a statistic point of view (that is low variance values), to manipulate somehow the source instead of sharply define it, in order either to fully exploit the features of his own code or to bypass and simplify actual physical difficulties, but on the other hand this procedure strongly increases the complexity of the source normalization process, which becomes (or can become) the true problem indeed. Two examples will be analyzed: a shielding problem of a transportation cask which will be investigated by means of MCNP, and a shielding problem of a proton beam which will be investigated by means of FLUKA. In both the problems a correct, sharply defined description of the source leads to hardly manageable situations, when "tricky" adjustments on the source instead lead smoothly to statistically good results while the source normalizations are anyway solved without too much complexity.

Computer Simulation↗
PubMed2005

Evaluation of the neutron spectrum and dose assessment around the venus reactor.

An assessment of the neutron field near the VENUS reactor is made in order to evaluate the neutron dose to the operators, particularly in an area near the reactor shielding and in the control room. Therefore, a full MCNPX model of the shielding geometry was developed. The source term used in the simulation is derived from a criticality calculation done beforehand. Calculations are compared to routine neutron dose rate measurements and show good agreement. The MCNPX model developed easily allows core adaptations in order to evaluate the effect of future core configuration on the neutron dose to the operators.

Algorithms↗
PubMed2005

SPE dose prediction using locally weighted regression.

When astronauts are outside earth's protective magnetosphere, they are subject to large radiation doses resulting from solar particle events (SPEs). The total dose received from a major SPE in deep space could cause severe radiation poisoning. The dose is usually received over a 20-40 h time interval but the event's effects may be mitigated with an early warning system. This paper presents a method to predict the total dose early in the event. It uses a locally weighted regression model, which is easier to train and provides predictions as accurate as neural network models previously used.

Algorithms↗
PubMed2005

ISFSI site boundary radiation dose rate analyses.

Across the globe nuclear utilities are in the process of designing and analysing Independent Spent Fuel Storage Installations (ISFSI) for the purpose of above ground spent-fuel storage primarily to mitigate the filling of spent-fuel pools. Using a conjoining of discrete ordinates transport theory (DORT) and Monte Carlo (MCNP) techniques, an ISFSI was analysed to determine neutron and photon dose rates for a generic overpack, and ISFSI pad configuration and design at distances ranging from 1 to -1700 m from the ISFSI array. The calculated dose rates are used to address the requirements of 10CFR72.104, which provides limits to be enforced for the protection of the public by the NRC in regard to ISFSI facilities. For this overpack, dose rates decrease by three orders of magnitude through the first 200 m moving away from the ISFSI. In addition, the contributions from different source terms changes over distance. It can be observed that although side photons provide the majority of dose rate in this calculation, scattered photons and side neutrons take on more importance as the distance from the ISFSI is increased.

Computer Simulation↗
PubMed2005

MYRRHA, a Pb-Bi experimental ADS: specific approach to radiation protection aspects.

Since 1998, SCK*CEN, in partnership with IBA s.a. and many European research laboratories, is designing a multipurpose accelerator driven system (ADS) for Research and Development (R&D) applications-MYRRHA-and is conducting an associated R&D support programme. MYRRHA is an ADS under development at Mol in Belgium and is aiming to serve as a basis for the European experimental ADS to provide protons and neutrons for various R&D applications. It consists of a proton accelerator delivering a 350 MeV x 5 mA proton beam to a liquid Pb-Bi spallation target that in turn couples to a Pb-Bi cooled, subcritical fast core. In the first stage, the project focuses mainly on demonstration of the ADS concept, safety research on sub-critical systems and nuclear waste transmutation studies. In a later stage, the device will also be dedicated to research on structural materials, nuclear fuel, liquid metal technology and associated aspects, and on sub-critical reactor physics. Subsequently, it will be used for research on applications such as radioisotope production. A first preliminary conceptual design file of MYRRHA was completed by the end of 2001 and has been reviewed by an International Technical Guidance Committee, which concluded that there are no show stoppers in the project and even though some topics such as the safety studies and the fuel qualification need to be addressed more deeply before concluding it. In this paper, we are reporting on the state-of-the art of the MYRRHA project at the beginning of 2004 and in particular on the radiation shielding assessment and the radiation protection particular aspects through a remote handling operation approach in order to minimise the personnel exposure to radiation.

Bismuth↗
PubMed2005

Occupational radiation exposure to norms in a gold mine.

Preliminary studies have been conducted into the occupational radiation exposure to NORMS from surface and underground mining operations in a gold mine in the Ashanti Region of Ghana. A brief description of the methods and instrumentation is presented. The annual effective dose has been estimated to be 0.26 +/- 0.11 mSv for surface mining and 1.83 +/- 0.56 mSv for the underground mines using the ICRP dose calculation method. The results obtained are found to be within the allowable limit of 20 mSv per annum for occupational exposure control recommended by the ICRP.

Background Radiation↗
PubMed2005

Assessment of plutonium exposures for an epidemiological study of US nuclear workers.

An ongoing case-control study evaluating the association between workplace external radiation exposures and leukaemia mortality required an assessment of internal plutonium exposures as a potential confounder. Of the study participants, 1,092 were employed at four Department of Energy sites where plutonium-bearing materials were processed or stored. Exposures were assessed by first categorising exposure potentials based on available bioassay data, then estimating doses for workers in the highest categories using recent recommendations of the International Commission on Radiological Protection. Given the aetiology of leukaemia, equivalent dose to active bone marrow was chosen as the exposure variable. There were 556 workers each with at least one plutonium bioassay result, assigned to one of three evaluation categories. Dose estimates were made for 115 workers resulting in a collective equivalent dose of 2.1 person-Sv for 2,822 exposure-years, compared with 29.8 person-Sv estimated from photon exposures. Modelling uncertainties were examined by comparison of results from independent analyses and by Monte Carlo simulation.

Case-Control Studies↗
PubMed2005

Radiation quality of photons in small and large receptors--a microdosimetric analysis.

The quality of different photon radiations in small and large receptors is assessed in terms of microdosimetry. Monte Carlo simulations are performed for the electrons released by photons. To represent the case of a small receptor, only electrons released by the non-degraded incident photons are tracked. For a large receptor, all electrons released in the complete degradation process are followed. Dose averaged values, y(D), of the lineal energy are derived from the simulated tracks for monoenergetic photons from 10 keV to 2 MeV. In microdosimetry, the dose mean lineal energy is the major parameter of radiation quality. The results demonstrate distinct differences in radiation quality between high-energy gamma rays and conventional X rays that are present not just in small cell samples but also in large receptors, such as the human body. The values of y(D) suggest that, for both small and large receptors, conventional X rays should be about twice as effective as gamma rays from 60Co.

Algorithms↗
PubMed2005

PRDC--a software package for personnel radiation dose calculation.

To determine effective dose, we usually need to use a very complicated human body model and a sophisticated computer code to transport radiations in the body model and surrounding medium, which is not very easy to practicing health physicists in the field. This study develops and tests a software package, called PRDC (Personnel Radiation Dose Calculation), which calculates effective dose and radiation doses to various organs/tissues and personal dosemeters based on a series of interpolations.

Algorithms↗