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At least 757 records · Page 42Linked to original sources
PubMed2005

Benchmarking of MCNP for calculating dose rates at an interim storage facility for nuclear waste.

During the operation of research facilities at Research Centre Jülich, Germany, nuclear waste is stored in drums and other vessels in an interim storage building on-site, which has a concrete shielding at the side walls. Owing to the lack of a well-defined source, measured gamma spectra were unfolded to determine the photon flux on the surface of the containers. The dose rate simulation, including the effects of skyshine, using the Monte Carlo transport code MCNP is compared with the measured dosimetric data at some locations in the vicinity of the interim storage building. The MCNP data for direct radiation confirm the data calculated using a point-kernel method. However, a comparison of the modelled dose rates for direct radiation and skyshine with the measured data demonstrate the need for a more precise definition of the source. Both the measured and the modelled dose rates verified the fact that the legal limits (<1 mSv a(-1)) are met in the area outside the perimeter fence of the storage building to which members of the public have access. Using container surface data (gamma spectra) to define the source may be a useful tool for practical calculations and additionally for benchmarking of computer codes if the discussed critical aspects with respect to the source can be addressed adequately.

Benchmarking↗
PubMed2005

Shielding for the upgraded Duke Free Electron Laser Laboratory.

The Duke FEL Laboratory is undergoing a series of upgrades staggered over time that will greatly increase the capabilities of the machines and by the same token the importance of radiation safety issues. In this paper, we present the scope of the planned upgrades and provide several specific examples of shielding calculations. We also present our effort to correlate calculations with experimental measurements.

Computer Simulation↗
PubMed2005

Three-dimensional shielding calculations for the IFMIF neutron source using a coupled Monte Carlo/Deterministic computational scheme.

Shielding calculations for the International Fusion Materials Irradiation Facility (IFMIF) are complicated due to the geometrical complexity of the target system and the large-scale bulk shields around the source target. Three-dimensional shielding calculations were performed by using a newly developed Monte Carlo/Deterministic computational scheme. The neutron-photon fluxes and dose rate distributions in the back wall of the Test Cell and the access/maintenance room are presented and compared with previous shielding calculations. The results demonstrate that this coupled scheme is an useful computational tool for three-dimensional shielding analyses of complex and large nuclear facilities.

Computer Simulation↗
PubMed2005

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

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.

Computer Simulation↗
PubMed2005

Calculation of eigenfunction fluxes in nuclear systems.

A new Monte Carlo method is being developed to calculate eigenfunction fluxes in critical or near-critical nuclear systems. The correct estimation of fluxes is essential for radiation protection and shielding near these systems, in addition to isotope production, isotope depletion, nuclear criticality and other applications. The proposed method applies to Monte Carlo criticality eigenvalue calculations in which the fission sites in one generation are used as fission sources in subsequent generations. The usual Monte Carlo power iteration method for such problems often calculates fluxes (eigenfunctions) that are inaccurate and very different in symmetric parts of a problem geometry. The proposed method calculates flux distributions by estimating an approximate fission matrix. The way the fission matrix is estimated and used differs from other recent works. Preliminary results are promising.

Algorithms↗
PubMed2005

The refined shielding design for the cyclotron room of the Buddhist Tzu Chi General Hospital.

Full-scale Monte Carlo simulations of the cyclotron room of the Buddhist Tzu Chi General Hospital were carried out to improve the inadequate maze design. The double differential neutron source from the 18O(p,n)18F reaction was adopted for the calculation. The weight window variance reduction technique, where the weight window was set by applying the adjoint flux, has been implemented in the MCNP run to facilitate the calculation of the dose rates outside the cyclotron room. Dose rates including neutron and gamma-ray components were calculated for some maze shielding modifications.

Computer Simulation↗
PubMed2005

Measurement of neutron inelastic scattering cross sections for 52Cr from threshold up to 18 MeV.

The powerful white neutron spectrum from Gelina is used for the investigation of gamma ray production cross sections in the reaction 52Cr(n,n'gamma)52Cr. The gamma rays from the inelastic process are detected with two large volume HPGe detectors placed at 110 degrees and 150 degrees with respect to the beam direction. The neutron flux has been measured with a 235U fission chamber. For the main transitions the inelastic cross section is given with an incident neutron energy resolution ranging from 1.12 keV at 1 MeV to 35.7 keV at 10 MeV, the overall statistical error being <5%. Based on the known 52Cr level scheme, the level and total inelastic cross sections have been obtained up to 3.77 MeV excitation energy. Progress on the implementation of a new acquisition system based on a fast digitiser is presented in this paper.

Chromium Isotopes↗
PubMed2005

Shielding for a cyclotron used for medical isotope production in China.

Monte Carlo and discrete ordinate calculations have been performed to determine the doses at several locations in a positron emission tomography (PET) facility in China, where the radiation source is a cyclotron that is used for the production of the isotopes necessary for PET scans. The energy-dependent neutron source term is obtained by calculations using the ALICE code, and is interpolated for input to Monte Carlo and discrete ordinate calculations. The building that houses the cyclotron has a labyrinth of walls to minimise dose to operators and to other occupants of the building. Unbiased Monte Carlo calculations did not converge after more than one week of CPU time, whereas direction biasing alone resulted in convergence in several days. A study of several biasing techniques indicated that about a factor of 3 in computational efficiency is obtained using evaluated biasing methods. The use of adjoint fluxes for biasing Monte Carlo calculations can improve computational efficiencies by one or two orders of magnitude for some problems.

China↗
PubMed2005

Radiation safety assessment of a system of small reactors for distributed energy.

A passively safe small reactor for a distributed energy system, PSRD, is an integral type of light-water reactor with a thermal output of 100 or 300 MW aimed to be used for supplying district heat, electricity to small grids, and so on. Candidate locations for the PSRD as a distributed energy source are on-ground, deep underground, and in a seaside pit in the vicinity of the energy consumption area. Assessments of the radiation safety of a PSRD were carried out for three cases corresponding to normal operation, shutdown and a hypothetical postulated accident for several siting candidates. Results of the radiation safety assessment indicate that the PSRD design has sufficient shielding performance and capability and that the exposure to the general public is very low in the case of a hypothetical accident.

Algorithms↗
PubMed2005

On the high-resolution gamma-ray spectrometric measurement of 40K in natural and synthetic materials.

Many regulatory agencies require that all building materials and industrial waste be tested for their naturally occurring radioactive material (NORM) concentrations before they can be used or thrown away. Usually the NORM concentrations of 40K, 232Th and 238U are measured by gamma-ray spectrometry using high-purity germanium or NaI(Tl) detectors. 40K is measured through its 1460.8 keV gamma line, which is mixed with 1459.2 keV line of 228Ac from the chain of 232Th. This fact ignored till now, requires a correction in the computation of 40K concentration. Although in many cases the error is <1%, there are cases where there have been higher errors. It should be emphasised that even if the correction in 40K concentration is large, the correction of the external dose index is negligible owing to the weighing factor being higher for 232Th than for 40K (by at least a factor of 10).

Gamma Rays↗
PubMed2005

Radiotoxicity and decay heat power of spent nuclear fuel of VVER type reactors at long-term storage.

Radiotoxicity and decay heat power of the spent nuclear fuel of VVER-1000 type reactors are calculated during storage time up to 300,000 y. Decay heat power of radioactive waste (radwaste) determines parameters of the heat removal system for the safe storage of spent nuclear fuel. Radiotoxicity determines the radiological hazard of radwaste after its leakage and penetration into the environment.

Half-Life↗
PubMed2005

Modelling of the shielding capabilities of the existing solid radioactive waste storages at Ignalina NPP.

There is only one nuclear power plant in Lithuania--Ignalina NPP (INPP). The INPP operates two similar units with design electrical power of 1500 MW. The units were commissioned in 1983 and 1987 respectively. From the beginning of the INPP operation all generated solid radioactive waste was collected and stored at the Soviet type solid radwaste facility located at INPP site. The INPP solid radwaste storage facility consists of four buildings, namely building No. 155, No. 155/1, No. 157 and No. 157/1. The buildings of the INPP solid radwaste storage facility are reinforced concrete structures above ground. State Nuclear Safety Inspectorate (VATESI) has specified that particular safety analysis must be performed for existing radioactive waste storage facilities of the INPP. As part of the safety analysis, shielding capabilities of the walls and roofs of these buildings were analysed. This paper presents radiation shielding analysis of the buildings No. 157 and No. 157/1 that are still in operation. The buildings No. 155 and No. 155/1 are already filled up with the waste and no additional waste loading is expected.

Computer Simulation↗
PubMed2005

Study of the neutron beam line shield design for JSNS.

The JSNS, a spallation neutron source of J-PARC (JAERI-KEK Joint Project of the High Intensity Proton Accelerator) has 23 neutron beam lines. In the present study, a database was formulated for an optimum shielding design using the MCNP-X code. The calculations involved two steps. In the first step, the neutron distributions were created in the typical neutron beam line with a model that included the spallation neutron source target. The neutron currents evaluated flowed from the duct into the duct wall which was the boundary source for the bulk shield surrounding the beam line. In the second step, bulk-shield calculations were performed for the various shielding materials (iron, concrete, heavy concrete and so on) used and their composites up to thicknesses of 3 m. The results were compared with each other. Composite material shields of iron and such hydrogeneous materials as polyethylene or concrete were more effective. A typical design was prepared for a beam line within 25 m distance from a moderator, as a sample.

Computer Simulation↗
PubMed2005

Test case of the long term preliminary performance assessment for the L&IL radioactive waste repository Baita Bihor, Romania.

The paper presents the assessment methodology developed in the framework of the long-term preliminary performance assessment of Baita Bihor Romanian national low-level and intermediate-level radioactive waste repository. Conservative assumptions, input parameters and tools used in this approach reflect limitations imposed by the current assessment capability, as well as limited knowledge and understanding of the disposal system. The resultant dose rates are presented and analysed. Conclusions are drawn relating to the safety of the repository.

Computer Simulation↗
PubMed2005

MCNPX vs. DORT for SNS shielding design studies.

Radiation transport occurs through the 18 m long access way adjacent to the Spallation Neutron Source accelerator tunnel and the 2.2 m thick massive shielding door which closes the access way. A variety of typical materials for accelerator shielding, such as concrete and steel, were used for construction of the door to study radiation penetration. A comparison was carried out using both Monte Carlo (code MCNPX) and discrete ordinates (code DORT) methods. The beam losses during the accelerator operation are the sources for the radiation calculations. Analyses show that the results from the two methods are in good agreement.

Computer Simulation↗
PubMed2005

The TRADE experiment: shielding calculations for the building hosting the subcritical system.

The TRADE project (TRiga Accelerator Driven Experiment), to be performed at the existing TRIGA reactor at ENEA Casaccia, has been proposed as a validation of the accelerator-driven system (ADS) concept. TRADE will be the first experiment in which the three main components of an ADS--the accelerator, spallation target and sub-critical blanket--are coupled at a power level sufficient to encounter reactivity feedback effects. As such, TRADE represents the necessary intermediate step in the development of hybrid transmutation systems, its expected outcomes being considered crucial--in terms of proof of stability of operation, dynamic behaviour and licensing issues--for the subsequent realisation of an ADS Transmutation Demonstrator. An essential role in the feasibility study of the experiment is played by radioprotection calculations. Such a system exhibits new characteristics with respect to a traditional reactor, owing to the presence of the proton accelerator. As beam losses always occur under normal operating conditions of an accelerator, shielding studies need to be performed not only around the reactor but also along the beam line from the accelerator to the spallation target. This paper illustrates a preliminary evaluation, using Monte Carlo methods, of the additional shielding to be located around the reactor structures, the beam transport line and the existing reactor building to allow access into the reactor hall and to restrict the doses outside to their legal limits.

Computer Simulation↗
PubMed2005

Calculation of absorbed dose around a facility for disposing of low activity natural radioactive waste (C3-dump).

A C3-dump is a facility for disposing of low activity natural radioactive waste containing the uranium series 238U, the thorium series 232Th and 40K. Only the external radiation owing to gamma rays, X-rays and annihilation photons is considered in this study. For two situations--the semi-infinite slab and the tourist geometry--the conversion coefficients from specific activity to air kerma rate at 1 m above the relevant level are calculated. In the first situation the waste material is in contact with the air but in the tourist geometry it is covered with a 1.35 m thick layer. For the calculations, the Monte Carlo radiation transport code MCNP is used. The yield and photon energy for each radionuclide are according to the database of Oak Ridge National Laboratory. For the tourist situation, the depth-dose distribution through the covering layer is calculated and extrapolated to determine the exit dose.

Computer Simulation↗
PubMed2005

Dose rate calculations for a reconnaissance vehicle.

A Chemical Nuclear Reconnaissance System (CNRS) has been developed by the British Ministry of Defence to make chemical and radiation measurements on contaminated terrain using appropriate sensors and recording equipment installed in a land rover. A research programme is under way to develop and validate a predictive capability to calculate the build-up of contamination on the vehicle, radiation detector performance and dose rates to the occupants of the vehicle. This paper describes the geometric model of the vehicle and the methodology used for calculations of detector response. Calculated dose rates obtained using the MCBEND Monte Carlo radiation transport computer code in adjoint mode are presented. These address the transient response of the detectors as the vehicle passes through a contaminated area. Calculated dose rates were found to agree with the measured data to be within the experimental uncertainties, thus giving confidence in the shielding model of the vehicle and its application to other scenarios.

Automobiles↗