Dose Rate Profile Surrounding a Repository

2013 ◽  
Author(s):  
J. Parson ◽  
A. Brandl ◽  
N. Zoeger ◽  
R. Koppitsch
Keyword(s):  
Monte Carlo ◽  
Dose Rate ◽  
Simulation Data ◽  
Dose Rates ◽  
Monte Carlo Techniques ◽  
Analytical Functions ◽  
Detector Distance ◽  
Waste Repository ◽  
Rate Profile

The focus of this research was to analyze the dose rate profile around a waste repository using Monte Carlo techniques. Dose rates at various heights and distances were analyzed outside of the waste repository using MCNPX [1]. The heights measured extended the height of the building and the distances varied between 0 and 22 m away from the waste repository. The simulation data were fitted by smooth analytical functions for different height levels and distances, such that vertical and horizontal dose rates as functions of source-detector distance were achieved.

Measurements of the High Dose Rate Profiles Inside a Shutdown CANDU® Reactor

2012 ◽  
Vol 1 (1) ◽  
pp. 27-34
Author(s):  
C. Jewett ◽  
J. Chow ◽  
D. Comeau ◽  
G. Jonkmans ◽  
B. Smith ◽  
...  
Keyword(s):  
Monte Carlo ◽  
Monte Carlo Simulations ◽  
Dose Rate ◽  
High Dose Rate ◽  
Measured Data ◽  
High Dose ◽  
Dose Rates ◽  
Diode Detector ◽  
Γ Ray ◽  
Rate Profile

The positions of the components of a reactor can change over time, due to radiation damage, sagging, etc. Thus, it is important to determine their positions. To satisfy this requirement of the staff at the Point Lepreau Generating Station, a method to determine the positions of reactor components has been developed and demonstrated. This method combines the use of dose rate measurements and Monte Carlo simulations. It first involves measuring the high γ-ray dose rates as a function of position within a reactor. Then it entails comparing these measurements with Monte Carlo simulations. In order to perform such measurements, a silicon diode detector and a scan drive system have been developed. In 2009, measurements of the γ-ray dose rate profile of the shut down Point Lepreau Generating Station reactor were conducted. By comparing the locations of the local peaks in the dose rate data, it was possible to determine the distances between the steel reactor components. The measured data were then compared with Monte Carlo simulations to determine how precisely one could locate the positions of the adjuster rods. Using this technique, it was found that the retracted adjuster rods were 440 ± 60 mm below their designed positions.

scholarly journals Field gamma spectrometry, Monte Carlo simulations and potential of non-invasive measurements

2012 ◽  
Vol 39 (1) ◽  
pp. 40-47 ◽  
Author(s):  
Guillaume Guérin ◽  
Norbert Mercier
Keyword(s):  
Monte Carlo ◽  
Monte Carlo Simulations ◽  
Dose Rate ◽  
Gamma Dose ◽  
Luminescence Dating ◽  
Gamma Dose Rate ◽  
Invasive Technique ◽  
Simple Method ◽  
Dose Rates ◽  
Non Invasive

Abstract The determination of gamma dose rates is of prior importance in the field of luminescence dating methods. In situ measurements are usually performed by the insertion of dosimeters or a portable gamma spectrometer cell in sediments. In this paper, Monte-Carlo simulations using the Geant4 toolkit allow the development of a new technique of insitu gamma dose rate evaluations: a spectrometer cell is placed on the surface of sediments under excavation to acquire successive spectra as sediments are removed by excavations. The principle of this non-invasive technique is outlined and its potential is discussed, especially in the case of environments in which radioelements are heterogeneously distributed. For such cases, a simple method to reconstruct gamma dose rate values with surface measurements using an attenuator is discussed, and an estimation of errors is given for two simple cases. This technique appears to be applicable, but still needs experimental validation.

Electron and Beta Dose Rates of UO2 Pellet and Fuel Rod

2013 ◽  
Author(s):  
Guoqing Zhang ◽  
Xuexin Wang ◽  
Jiangang Zhang ◽  
Dajie Zhuang ◽  
Chaoduan Li ◽  
...  
Keyword(s):  
Monte Carlo ◽  
Dose Rate ◽  
Radiation Protection ◽  
High Dose Rate ◽  
High Dose ◽  
Electron Dose ◽  
Dose Rates ◽  
Fuel Rod ◽  
Calculation Results ◽  
Beta Dose

The isotopes of uranium and their daughter nuclides inside the UO2 pellet emit mono-energetic electrons and beta rays, which generate rather high dose rate near the UO2 pellet and could cause exposure to workers. In this work calculations of electron dose rates have been carried out with Monte Carlo codes, MCNPX and Geant4, for a UO2 pellet and a fuel rod. Comparisons between calculations and measurements have been carried out to verify the calculation results. The results could be used to estimate the dose produced by electrons and beta rays, which could be used to make optimization for radiation protection purpose.

MCNP Simulation of In-Core Dose Rates for an Offline CANDU® Reactor

2017 ◽  
Vol 3 (3) ◽  
Author(s):  
Jordan G. Gilbert ◽  
Scott Nokleby ◽  
Ed Waller
Keyword(s):  
Monte Carlo ◽  
Dose Rate ◽  
Integrated Circuit ◽  
Multidisciplinary Team ◽  
Operating Conditions ◽  
Inspection System ◽  
Dose Rates ◽  
Pressure Tubes ◽  
Circuit Components ◽  
Safe Operating

Inspections of pressure tubes in CANDU® reactors are a key part of maintaining safe operating conditions. The current inspection system, the channel inspection and gauging apparatus for reactors (CIGAR), performs the job well but is limited by the fact that it can only inspect one channel at a time. A multidisciplinary team is currently developing a novel robotic inspection system. As part of this work, a Monte Carlo N-particle (MCNP) model has been developed in order to predict the dose rates that the improved inspection system will be exposed to and, from this, predict the component lifetime. This MCNP model will be capable of predicting in-core dose rates at any location within the reactor, and as such could be used for other situations where the in-core dose rate needs to be known. Based on estimates from this model, it is expected that at 7 days after shutdown, the improved inspection system could survive in core for approximately 7 h, providing it uses a tungsten shield 2.5 cm in thickness around the integrated circuit components. This is expected to be sufficient to perform a single inspection.

Cemented Containers Radiological Data From a Disused Uranium Mine Low-Level Waste Repository: A Calculated-Experiment Cross-Check for Data Verification and Validation

2003 ◽  
Author(s):  
D. G. Cepraga ◽  
G. Cambi ◽  
M. Frisoni ◽  
D. Ene
Keyword(s):  
Experimental Data ◽  
Dose Rate ◽  
Measured Data ◽  
Radioactive Material ◽  
Uranium Mine ◽  
Low Level ◽  
Dose Rates ◽  
Waste Repository ◽  
Central Plan ◽  
Decay Gamma

Code validation problems involve calculation of experiments and a comparison experiment-calculation. Experimental data and physical properties of these systems are used to determine the range of applicability of the validation. Once a sequence-code of calculations has been validated, it has to be underlined that the comparison experimental-calculated results involving “complex systems” or “complex experimental measures” permits also a bi-lateral cross-check between the calculation scheme and the experimental procedures. The results of the testing and the validation effort related to the collection of information and measured data and the comparison between code results with experimental data coming from a “low-level waste” repository are presented in this paper. The Baita-Bihor repository, sited into former disused uranium mine in Transylvania, has been considered as the source of experimental data. The study was developed through the following steps: a) collection and processing of measured data (radioactivity content and dose rate), from the cemented containers of the Baita-Bihor repository; b) decay gamma source calculation by the ANITA-2000 code package (the input data for the calculations are the measured isotope activities for each container); c) decay gamma transport calculation by the SCALENEA-1 shielding Sn sequence approach (Nitawl-Xsdrnpm-Xsdose modules of the Scale 4.4a code system, using the Vitenea-J library, based on FENDL/E-2 data) to obtain dose rates on the surfaces and at various points outside the containers; d) comparison experimental-calculated dose rates, taking into account also the measurement uncertainties. The new version of the ANITA-2000 activation code package used makes possible to assess the behaviour of irradiated materials independently from the knowledge of the irradiation scenario but using only data on the isotope radioactive material composition. Radioactive waste disposed of at Baita Bihor repository consists of worn reactor parts, resins and filters, packing materials, mop heads, protective clothing, temporary floor coverings and tools, the sources normally generated during the day-to-day operation of research reactors, the remediation-treatment stations and the medicine and biological activities. The low and intermediate wastes are prepared for shipping and disposal in the treatment stations by confining them in a cement matrix inside 220 litre metallic drums. Each container consists of an iron cladding filled by concrete Portland. Radioisotope composition and radioactivity distributions inside the drum are measured. The gamma spectroscopy has been used for. The calibration technique was based on the assumption of a uniform distribution of the source activity in the drum and also of a uniform sample matrix. Dose rate measurements are done continuously, circularly, in the central plan on the surface of the drum and 1 m from the surface, in the air. A “stuffing factor” model has been adopted to simulate, for the calculation, the spatial distribution of the gamma sources in the concrete region. In order to guarantee a complete Quality Assurance for codes and procedures, a simulation of the radioactive containers to evaluate the dose rates was done also by using the Monte Carlo MCNP-4C code. Its calculation results are in a very good agreement with those obtained by the Sn approach (discrepancies are around 2%, using the spherical approximation).

scholarly journals Environmental dose rate assessment of ITER using the Monte Carlo method

2014 ◽  
Vol 29 (1) ◽  
pp. 34-39
Author(s):  
Alireza Karimian ◽  
Amir Beheshti ◽  
Mohammadreza Abdi ◽  
Iraj Jabbari
Keyword(s):  
Monte Carlo ◽  
Monte Carlo Method ◽  
Dose Rate ◽  
Equivalent Dose ◽  
Photon Radiation ◽  
Normal Operation ◽  
Whole Body ◽  
Dose Rates ◽  
The Monte Carlo Method ◽  
Radiosensitive Organs

Exposure to radiation is one of the main sources of risk to staff employed in reactor facilities. The staff of a tokamak is exposed to a wide range of neutrons and photons around the tokamak hall. The International Thermonuclear Experimental Reactor (ITER) is a nuclear fusion engineering project and the most advanced experimental tokamak in the world. From the radiobiological point of view, ITER dose rates assessment is particularly important. The aim of this study is the assessment of the amount of radiation in ITER during its normal operation in a radial direction from the plasma chamber to the tokamak hall. To achieve this goal, the ITER system and its components were simulated by the Monte Carlo method using the MCNPX 2.6.0 code. Furthermore, the equivalent dose rates of some radiosensitive organs of the human body were calculated by using the medical internal radiation dose phantom. Our study is based on the deuterium-tritium plasma burning by 14.1 MeV neutron production and also photon radiation due to neutron activation. As our results show, the total equivalent dose rate on the outside of the bioshield wall of the tokamak hall is about 1 mSv per year, which is less than the annual occupational dose rate limit during the normal operation of ITER. Also, equivalent dose rates of radiosensitive organs have shown that the maximum dose rate belongs to the kidney. The data may help calculate how long the staff can stay in such an environment, before the equivalent dose rates reach the whole-body dose limits.

Bayesian Estimation of DSGE Models

2015 ◽  
Author(s):  
Edward P. Herbst ◽  
Frank Schorfheide
Keyword(s):  
Monte Carlo ◽  
Sequential Monte Carlo ◽  
Likelihood Function ◽  
Academic Research ◽  
Dynamic Stochastic General Equilibrium ◽  
Computational Techniques ◽  
Dsge Models ◽  
Monte Carlo Techniques ◽  
Dynamic Stochastic ◽  
Theoretical Foundations

Dynamic stochastic general equilibrium (DSGE) models have become one of the workhorses of modern macroeconomics and are extensively used for academic research as well as forecasting and policy analysis at central banks. This book introduces readers to state-of-the-art computational techniques used in the Bayesian analysis of DSGE models. The book covers Markov chain Monte Carlo techniques for linearized DSGE models, novel sequential Monte Carlo methods that can be used for parameter inference, and the estimation of nonlinear DSGE models based on particle filter approximations of the likelihood function. The theoretical foundations of the algorithms are discussed in depth, and detailed empirical applications and numerical illustrations are provided. The book also gives invaluable advice on how to tailor these algorithms to specific applications and assess the accuracy and reliability of the computations. The book is essential reading for graduate students, academic researchers, and practitioners at policy institutions.

Neutron dose evaluation of Elekta Linac at two energies (10 & 18 MV) by MCNP code and comparison with experimental measurements

2014 ◽  
Vol 6 (1) ◽  
pp. 1006-1015
Author(s):  
Negin Shagholi ◽  
Hassan Ali ◽  
Mahdi Sadeghi ◽  
Arjang Shahvar ◽  
Hoda Darestani ◽  
...  
Keyword(s):  
Monte Carlo ◽  
Measurement Data ◽  
Calculated Data ◽  
High Energy ◽  
Neutron Dose ◽  
Dose Assessment ◽  
Photon Flux ◽  
Dose Equivalent ◽  
Monte Carlo Techniques ◽  
Neutron Dose Equivalent

Medical linear accelerators, besides the clinically high energy electron and photon beams, produce other secondary particles such as neutrons which escalate the delivered dose. In this study the neutron dose at 10 and 18MV Elekta linac was obtained by using TLD600 and TLD700 as well as Monte Carlo simulation. For neutron dose assessment in 2020 cm2 field, TLDs were calibrated at first. Gamma calibration was performed with 10 and 18 MV linac and neutron calibration was done with 241Am-Be neutron source. For simulation, MCNPX code was used then calculated neutron dose equivalent was compared with measurement data. Neutron dose equivalent at 18 MV was measured by using TLDs on the phantom surface and depths of 1, 2, 3.3, 4, 5 and 6 cm. Neutron dose at depths of less than 3.3cm was zero and maximized at the depth of 4 cm (44.39 mSvGy-1), whereas calculation resulted  in the maximum of 2.32 mSvGy-1 at the same depth. Neutron dose at 10 MV was measured by using TLDs on the phantom surface and depths of 1, 2, 2.5, 3.3, 4 and 5 cm. No photoneutron dose was observed at depths of less than 3.3cm and the maximum was at 4cm equal to 5.44mSvGy-1, however, the calculated data showed the maximum of 0.077mSvGy-1 at the same depth. The comparison between measured photo neutron dose and calculated data along the beam axis in different depths, shows that the measurement data were much more than the calculated data, so it seems that TLD600 and TLD700 pairs are not suitable dosimeters for neutron dosimetry in linac central axis due to high photon flux, whereas MCNPX Monte Carlo techniques still remain a valuable tool for photonuclear dose studies.

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