Dose Rate Measurements Around Consignments of Radioactive Material

2000 ◽  
Vol 11 (1-2) ◽  
pp. 141-146
Author(s):  
R. Gelder ◽  
K. B. Shaw
Keyword(s):  
Dose Rate ◽  
Radioactive Material

Unit Dose Factors for Transportation of Radioactive Materials

2015 ◽  
Author(s):  
Kevin J. Connolly ◽  
Elena Kalinina
Keyword(s):  
Dose Rate ◽  
Nuclear Power ◽  
Large Scale ◽  
Spent Nuclear Fuel ◽  
Radioactive Material ◽  
Unit Distance ◽  
The Public ◽  
Environmental Impact Statements ◽  
Unit Dose ◽  
Interim Storage

It will be necessary in the future to transport spent nuclear fuel on a large-scale basis from nuclear power plant sites to interim storage and/or a repository. Shipments of radioactive material are required to comply with regulations limiting the dose rate to no more than 0.1 mSv (10 mrem) per hour at 2 meters from the sides of the vehicle transporting the package. Determining the resulting dose to the public will be necessary for a number of reasons (e.g., stakeholder concerns, environmental impact statements). In order to understand the dose consequence of such a transportation system, this paper describes a method for determining unit dose factors. These are defined as the dose to the public per unit distance traveled along a road, rail, or waterway from one shipment assuming unit values for the other route specific parameters. The actual dose to the public is calculated using unit dose factors, the dose rate due to the radiation field emanating from the package, and characteristics of the route itself. Route specific parameters include the speed of the conveyance, the population density, and characteristics of the environment surrounding the route; these are provided by a routing tool. Using these unit dose factors, in conjunction with a routing tool, it will be possible to quantify the collective dose to the public and understand the ramifications of choosing specific routes.

ACHIEVING A LINEAR DOSE RATE RESPONSE IN PULSE-MODE SILICON PHOTODIODE SCINTILLATION DETECTORS OVER A WIDE RANGE OF EXCITATIONS

2014 ◽  
Vol 27 ◽  
pp. 1460136
Author(s):  
LEWIS CARROLL
Keyword(s):  
Signal Processing ◽  
Dose Rate ◽  
Pulse Generator ◽  
Current Mode ◽  
Pulse Mode ◽  
Radioactive Material ◽  
Activity Levels ◽  
Silicon Photodiode ◽  
Processing Scheme ◽  
Wide Range

We are developing a new dose calibrator for nuclear pharmacies that can measure radioactivity in a vial or syringe without handling it directly or removing it from its transport shield “pig”. The calibrator's detector comprises twin opposing scintillating crystals coupled to Si photodiodes and current-amplifying trans-resistance amplifiers. Such a scheme is inherently linear with respect to dose rate over a wide range of radiation intensities, but accuracy at low activity levels may be impaired, beyond the effects of meager photon statistics, by baseline fluctuation and drift inevitably present in high-gain, current-mode photodiode amplifiers. The work described here is motivated by our desire to enhance accuracy at low excitations while maintaining linearity at high excitations. Thus, we are also evaluating a novel “pulse-mode” analog signal processing scheme that employs a linear threshold discriminator to virtually eliminate baseline fluctuation and drift. We will show the results of a side-by-side comparison of current-mode versus pulse-mode signal processing schemes, including perturbing factors affecting linearity and accuracy at very low and very high excitations. Bench testing over a wide range of excitations is done using a Poisson random pulse generator plus an LED light source to simulate excitations up to ∼106 detected counts per second without the need to handle and store large amounts of radioactive material.

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).

DOSE RATE IN A DEACTIVATED URANIUM MINE

2019 ◽  
Vol 7 (2B) ◽  
Author(s):  
Wagner De Souza Pereira ◽  
Alphonse Kelecom ◽  
Ademir Xavier Da Silva ◽  
José Marquez Lopes ◽  
Alessander Sá Do Carmo ◽  
...  
Keyword(s):  
Standard Deviation ◽  
Dose Rate ◽  
Electronic Monitoring ◽  
Limit Of Detection ◽  
Monitoring Program ◽  
Radioactive Material ◽  
Uranium Mine ◽  
Treatment Unit ◽  
Thermoluminescence Dosimetry ◽  
Area Monitoring

The Ore Treatment Unit is a deactivated uranium mine and milling situated in Caldas, MG, BR. Although disabled, there are still areas considered controlled and supervised from the radiological point of view. In these areas, it is necessary to keep an occupational monitoring program to ensure the workers' safety and to prevent the dispersion of radioactive material. For area monitoring, the dose rate, in µSv∙h-1, was measured with Geiger Müller (GM) area monitors or personal electronic monitors type GM and thermoluminescence dosimetry (TLD), in mSv∙month-1, along the years 2013 to 2016. For area monitoring, 577 samples were recorded; for personal dosimeters monitoring, 2,656; and for TLD monitoring type, 5,657. The area monitoring showed a mean dose rate of 6.42 µSv∙h-1 associated to a standard deviation of 48 µSv∙h-1 with a maximum recorded value of 685 µSv∙h-1. 96 % of the samples were below the derived limit per hour for workers (10 µSv∙h-1). For the personal electronic monitoring, the average of the data sampled was 15.86 µSv∙h-1, associated to a standard deviation of 61.74 µSv∙h-1. 80 % of the samples were below the derived limit and the maximum recorded was 1,220 µSv∙h-1. Finally, the TLD showed a mean of 0.01 mSv∙h-1 (TLD detection limit is 0.2 mSv∙month-1 equivalent to 0.28 µSv∙h-1), associated to a standard deviation of 0.08 mSv∙h-1. 98% of the registered values were below 0.2 mSv∙month-1 and less than 2 % of the measurements had values above the limit of detection. The samples show areas with low risk of external exposure, as can be seen by the TLD evaluation. Specific areas with greater risk of contamination have already been identified, as well as operations at higher risks. In these cases, the use of the individual electronic dosimeter is justified for a more effective monitoring. Radioprotection identified all risks and was able to extend individual electronic monitoring to all risk operations, even with the use of the TLD.

scholarly journals Definition of “Small Gram Quantity Contents” for Type B Radioactive Material Transportation Packages: Activity-Based Content Limitations

2011 ◽  
Author(s):  
Shiva Sitaraman ◽  
Soon Kim ◽  
Debdas Biswas ◽  
Ronald Hafner ◽  
Brian Anderson
Keyword(s):  
Dose Rate ◽  
Radioactive Material ◽  
Packaging Materials ◽  
Type B ◽  
Mass Limit ◽  
Sphere Model ◽  
External Dose ◽  
Calculated Mass ◽  
Material Transportation ◽  
Definition Of

This paper presents a compendium of allowable masses for a variety of gamma and neutron emitting isotopes (with varying impurity levels of beryllium in some of the actinide isotopes) that, when loaded in an unshielded radioactive material transportation packaging, do not result in an external dose rate on the surface of the package that exceeds 190 mrem/hr (190 mrem/hr was chosen to provide 5% conservatism relative to the regulatory limit). These mass limits define the term “Small Gram Quantity” (SGQ) contents in the context of radioactive material transportation packages. The term SGQ is isotope-specific and pertains to contents in radioactive material transportation packages that do not require shielding and still satisfy the external dose rate requirements. Since these calculated mass limits are for contents without shielding, they are conservative for packaging materials that provide some limited shielding or if the contents are placed into a shielded package. Two sets of mass limit results are presented: (1) mass limits calculated with a “voided sphere” model, and (2) mass limits calculated with the unshielded radioactive material transportation packaging Model 9977-96.

scholarly journals Assessment of exposure to ionizing radiation in Chernobyl tree frogs (Hyla orientalis)

2021 ◽  
Author(s):  
Pablo Burraco ◽  
Clement Car ◽  
Jean-Marc Bonzom ◽  
German Orizaola
Keyword(s):  
Ionizing Radiation ◽  
Total Dose ◽  
Dose Rate ◽  
Animal Health ◽  
Radioactive Material ◽  
Internal Exposure ◽  
Accurate Estimation ◽  
Exposure Levels ◽  
Tree Frogs ◽  
Current Exposure

Ionizing radiation can damage organic molecules, causing detrimental effects on human and wildlife health. The accident at the Chernobyl nuclear power plant (1986) represents the largest release of radioactive material to the environment. An accurate estimation of the current exposure to radiation in wildlife, often reduced to ambient dose rate assessments, is crucial to understand the long-term impact of radiation on living organisms. Here, we present an evaluation of the sources and variation of current exposure to radiation in breeding Eastern tree frogs (Hyla orientalis) males living in the Chernobyl Exclusion Zone. Total dose rates in H. orientalis were highly variable, although generally below widely used thresholds considered harmful for animal health. Internal exposure was the main source of absorbed dose rate (81% on average), with 90Sr being the main contributor (78% of total dose rate, on average). These results highlight the importance of assessing both internal and external exposure levels in order to perform a robust evaluation of the exposure to radiation in wildlife. Further studies incorporating life-history, ecological, and evolutionary traits are needed to fully evaluate the effects that these exposure levels can have in amphibians and other taxa inhabiting radio-contaminated environments.

Data Assimilation for Fukushima Nuclear Accident Assessments

2017 ◽  
Author(s):  
Mei Xu ◽  
Biao Yuan ◽  
Liangyu Wang ◽  
Lijun Zhang
Keyword(s):  
Data Assimilation ◽  
Dose Rate ◽  
Radiation Field ◽  
Nuclear Power ◽  
Nuclear Accident ◽  
Radioactive Material ◽  
Fukushima Nuclear Accident ◽  
The Real ◽  
Kalman Filter Algorithm ◽  
Good Agreement

In order to investigate the feasibility of data assimilation in a real nuclear accident environment, measurements of Fukushima nuclear accident were considered. The data assimilation system was constructed by using the Lagrangian puff model as the radioactive material diffusion model, and 86 group real dose rate data from the accident as the observations, and the Ensemble Kalman Filter algorithm as the assimilation algorithm. The experimental results show that the assimilated nuclear accident radiation field is in good agreement with the actual measurements, the land contaminated areas are concentrated in the northwest of the nuclear power plant. With the increase of the real measurements, the error of the radiation field decreases with time. Compared with the results with no assimilation, the uncertainty of assimilated dose rate was reduced more than 80%. Through the data assimilation, the whole error of the radiation field is about 30%. The utilization of the real measurements can reduce the uncertainty of the model prediction.

The Nuclear Decommissioner and the Regulation of Low Dose Radiation

2003 ◽  
Author(s):  
Shankar Menon ◽  
Luis Valencia ◽  
Lucien Teunckens
Keyword(s):  
Dose Rate ◽  
Nuclear Industry ◽  
Radioactive Material ◽  
High Dose ◽  
Activity Levels ◽  
Actual Measurement ◽  
Economic Implications ◽  
High Background ◽  
Naturally Occurring ◽  
Biological Studies

The management of the large quantities of very low level radioactive material that arise during the decommissioning of the increasing numbers of nuclear power stations reaching the end of their commercially useful lives, has become a major subject of discussion. This has very significant economic implications for the nuclear decommissioner. Much larger quantities — 2–3 orders of magnitude larger — of material, radiologically similar to the candidate material for recycling from the nuclear industry, arise in non-nuclear industries like coal, fertiliser, oil and gas, mining, etc. In such industries, naturally occurring radioactivity is artificially concentrated in products, by-products or waste to form TENORM (Technologically Enhanced Naturally Occurring Radioactive Material). It is only in the last decade that the international community has become aware of the prevalence of TENORM, specially the activity levels and quantities arising in so many non-nuclear industries. The first reaction of international organisations seems to have been to propose different standards for the nuclear and non-nuclear industries, with very stringent release criteria for radioactive material from the regulated nuclear industry and up to thirty to a hundred times more liberal criteria for the release/exemption of TENORM from the as yet unregulated non-nuclear industries. The radiological effects of these TENORM releases have recently been dramatically highlighted by the Marina II study, which showed that over 90% of the total exposures of the European population from discharges into the North European marine waters are from radioactive discharges from non-nuclear industries. The results of an international project to validate, by actual measurement, dose calculation codes RESRAD-RECYCLE (USA) and CERISE (France) for recycling, have indicated an overestimation of doses by the codes by an order of magnitude. For the nuclear decommissioner and other producers of large volumes of slightly radioactively contaminated material, clearance levels determined on the basis of such a degree of conservatism in calculations can lead to huge volumes of material unnecessarily being condemned to burial as radioactive waste. Earlier estimates of the quantitative risk levels of exposure to ionising radiation have almost exclusively been based on doses taken by exposed populations of Hiroshima and Nagasaki (ICRP 60). The populations studied have been exposed to over 200 mSv at a dose rate of 6 Sv/s. The effects of such high dose/dose-rate exposure are being used as the basis for risk judgment at doses/dose-rates lower by a factor 1012–1015. The validity of such an extrapolation in risk judgement is an area of prime interest for discussion. In this connection, an interesting development, for both the nuclear and non-nuclear industries, is the increased scientific scrutiny that the populations of naturally high background dose level areas of the world are being subject to. Preliminary biological studies have indicated that the inhabitants of such areas, exposed to many times the permitted occupational doses for nuclear workers, have not shown any differences in cancer mortality, life expectancy, chromosome aberrations or immune function, in comparison with those living in normal background areas. The paper discusses these and other strategic issues regarding the management of redundant low radiation material from both the nuclear and non-nuclear industries, underlining the need for consistency in regulatory treatment.

Measurement of ambient dose rates in tantalite mining sites in Oke ̶ Ogun, southwest, Nigeria

2021 ◽  
Vol 20 (2) ◽  
pp. 1-12
Author(s):  
A.E. Ajetunmobi ◽  
S.K. Alausa ◽  
J.O. Coker ◽  
T.W. David ◽  
A.T. Talabi
Keyword(s):  
Cancer Risk ◽  
Ionizing Radiation ◽  
Dose Rate ◽  
Health Effect ◽  
Radioactive Material ◽  
Annual Dose ◽  
Mining Sites ◽  
Dose Rates ◽  
Safe Limit ◽  
Ambient Dose

The work scenarios involved in the mining of tantalite a radioactive material expose the miners to ionizing radiation from the ore and the surrounding environment. The dose level in the mine air may be higher than the safe limit due to various contributory sources of ionizing radiation such as radionuclides from rocks, effluents, sand, and radon gas that emanates from caves and this can be of health detriment to the miners. Measurements of ambient dose rates in four selected mining sites have been investigated. Gamma absorbed dose rates were measured in air onsite at Komu, Sepenteri, Gbedu, and Eluku mining sites in Oke-Ogun areas of Oyo State, Nigeria using GammaRAE II dosimeter. Radiation dose to risk software was used to estimate the cancer risk for the period the miners spent onsite. The measured mean dose rate at the sites falls within the range of (19-240) nSv/y and the estimated annual dose rate, cumulative dose, and cancer risk fall within the range of (37-314) μSv/y, (4.0 ̶ 11.1) mSv and (0.5 ̶ 4.5) E-04 respectively. The upper limits of the range for the radiological parameters are all above the safe limit. The health implication of that is that increased work activities at these mining sites may over the years have a negative health effect on the miners. The exposure time of workers can be reduced through proper planning of working shifts for the miners.

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