Probabilistic biosphere modeling for the long-term safety assessment of geological disposal facilities for radioactive waste using first- and second-order Monte Carlo simulation.

In the present paper, deterministic as well as first- and second-order probabilistic biosphere modeling approaches are compared. Furthermore, the sensitivity of the influence of the probability distribution function shape (empirical distribution functions and fitted lognormal probability functions) representing the aleatory uncertainty (also called variability) of a radioecological model parameter as well as the role of interacting parameters are studied. Differences in the shape of the output distributions for the biosphere dose conversion factor from first-order Monte Carlo uncertainty analysis using empirical and fitted lognormal distribution functions for input parameters suggest that a lognormal approximation is possibly not always an adequate representation of the aleatory uncertainty of a radioecological parameter. Concerning the comparison of the impact of aleatory and epistemic parameter uncertainty on the biosphere dose conversion factor, the latter here is described using uncertain moments (mean, variance) while the distribution itself represents the aleatory uncertainty of the parameter. From the results obtained, the solution space of second-order Monte Carlo simulation is much larger than that from first-order Monte Carlo simulation. Therefore, the influence of epistemic uncertainty of a radioecological parameter on the output result is much larger than that one caused by its aleatory uncertainty. Parameter interactions are only of significant influence in the upper percentiles of the distribution of results as well as only in the region of the upper percentiles of the model parameters.

Assessment of uncertainty in discrete fracture network modeling using probabilistic distribution method.

There have been widespread concerns about solute transport problems in fractured media, e.g. the disposal of high-level radioactive waste in geological fractured rocks. Numerical simulation of particle tracking is gradually being employed to address these issues. Traditional predictions of radioactive waste transport using discrete fracture network (DFN) models often consider one particular realization of the fracture distribution based on fracture statistic features. This significantly underestimates the uncertainty of the risk of radioactive waste deposit evaluation. To adequately assess the uncertainty during the DFN modeling in a potential site for the disposal of high-level radioactive waste, this paper utilized the probabilistic distribution method (PDM). The method was applied to evaluate the risk of nuclear waste deposit in Beishan, China. Moreover, the impact of the number of realizations on the simulation results was analyzed. In particular, the differences between the modeling results of one realization and multiple realizations were demonstrated. Probabilistic distributions of 20 realizations at different times were also obtained. The results showed that the employed PDM can be used to describe the ranges of the contaminant particle transport. The high-possibility contaminated areas near the release point were more concentrated than the farther areas after 5E6 days, which was 25,400 m2.

Radioactive Wastes.

Papers reviewed herein present a general overview of radioactive waste related activities around the world in 2016. The current reveiw include studies related to safety assessments, decommission and decontamination of nuclear facilities, fusion facilities, transportation. Further, the review highlights on management solutions for the final disposal of low and high level radioactive wastes (LLW and HLW), interim storage and final disposal options for spent fuel (SF), and tritiated wastes, with a focus on environmental impacts due to the mobility of radionuclides in ecosystem, water and soil alongwith other progress made in the management of radioactive wastes.

Uranium Mining and Norm in North America-Some Perspectives on Occupational Radiation Exposure.

All soils and rocks contain naturally occurring radioactive materials (NORM). Many ores and raw materials contain relatively elevated levels of natural radionuclides, and processing such materials can further increase the concentrations of naturally occurring radionuclides. In the U.S., these materials are sometimes referred to as technologically-enhanced naturally occurring radioactive materials (TENORM). Examples of NORM minerals include uranium ores, monazite (a source of rare earth minerals), and phosphate rock used to produce phosphate fertilizer. The processing of these materials has the potential to result in above-background radiation exposure to workers. Following a brief review of the sources and potential for worker exposure from NORM in these varied industries, this paper will then present an overview of uranium mining and recovery in North America, including discussion on the mining methods currently being used for both conventional (underground, open pit) and in situ leach (ISL), also referred to as In Situ Recovery (ISR), and the production of NORM materials and wastes associated with these uranium recovery methods. The radiological composition of the NORM products and wastes produced and recent data on radiological exposures received by workers in the North American uranium recovery industry are then described. The paper also identifies the responsible government agencies in the U.S. and Canada assigned the authority to regulate and control occupational exposure from these NORM materials.

Reconstruction of A Radiological Release Using Aerosol Sampling.

Reconstructing a radiological release using offsite air sampling is considered using measurements from the WIPP site event. Demonstration of the capabilities is shown using the HotSpot code as an example of the approach in general. The results show the empirical steps that could be folded into an iterative approach to back extrapolation of a radiological release from a nuclear facility. It is shown here that the limiting factor in detection is not counting statistics of the activity but rather the statistics for aerosols having small sample population numbers due to the lognormal distribution of particle radii.

Nuclear Waste Facing the Test of Time: The Case of the French Deep Geological Repository Project.

The purpose of this article is to consider the socio-anthropological issues raised by the deep geological repository project for high-level, long-lived nuclear waste. It is based on fieldwork at a candidate site for a deep storage project in eastern France, where an underground laboratory has been studying the feasibility of the project since 1999. A project of this nature, based on the possibility of very long containment (hundreds of thousands of years, if not longer), involves a singular form of time. By linking project performance to geology's very long timescale, the project attempts "jump" in time, focusing on a far distant future, without understanding it in terms of generations. But these future generations remain measurements of time on the surface, where the issue of remembering or forgetting the repository comes to the fore. The nuclear waste geological storage project raises questions that neither politicians nor scientists, nor civil society, have ever confronted before. This project attempts to address a problem that exists on a very long timescale, which involves our responsibility toward generations in the far future.

Decreasing Kd uncertainties through the application of thermodynamic sorption models.

Radionuclide retardation processes during transport are expected to play an important role in the safety assessment of subsurface disposal facilities for radioactive waste. The linear distribution coefficient (Kd) is often used to represent radionuclide retention, because analytical solutions to the classic advection-diffusion-retardation equation under simple boundary conditions are readily obtainable, and because numerical implementation of this approach is relatively straightforward. For these reasons, the Kd approach lends itself to probabilistic calculations required by Performance Assessment (PA) calculations. However, it is widely recognised that Kd values derived from laboratory experiments generally have a narrow field of validity, and that the uncertainty of the Kd outside this field increases significantly. Mechanistic multicomponent geochemical simulators can be used to calculate Kd values under a wide range of conditions. This approach is powerful and flexible, but requires expert knowledge on the part of the user. The work presented in this paper aims to develop a simplified approach of estimating Kd values whose level of accuracy would be comparable with those obtained by fully-fledged geochemical simulators. The proposed approach consists of deriving simplified algebraic expressions by combining relevant mass action equations. This approach was applied to three distinct geochemical systems involving surface complexation and ion-exchange processes. Within bounds imposed by model simplifications, the presented approach allows radionuclide Kd values to be estimated as a function of key system-controlling parameters, such as the pH and mineralogy. This approach could be used by PA professionals to assess the impact of key geochemical parameters on the variability of radionuclide Kd values. Moreover, the presented approach could be relatively easily implemented in existing codes to represent the influence of temporal and spatial changes in geochemistry on Kd values.

Remediating radium contaminated legacy sites: Advances made through machine learning in routine monitoring of "hot" particles.

The extensive use of radium during the 20th century for industrial, military and pharmaceutical purposes has led to a large number of contaminated legacy sites across Europe and North America. Sites that pose a high risk to the general public can present expensive and long-term remediation projects. Often the most pragmatic remediation approach is through routine monitoring operating gamma-ray detectors to identify, in real-time, the signal from the most hazardous heterogeneous contamination (hot particles); thus facilitating their removal and safe disposal. However, current detection systems do not fully utilise all spectral information resulting in low detection rates and ultimately an increased risk to the human health. The aim of this study was to establish an optimised detector-algorithm combination. To achieve this, field data was collected using two handheld detectors (sodium iodide and lanthanum bromide) and a number of Monte Carlo simulated hot particles were randomly injected into the field data. This allowed for the detection rate of conventional deterministic (gross counts) and machine learning (neural networks and support vector machines) algorithms to be assessed. The results demonstrated that a Neural Network operated on a sodium iodide detector provided the best detection capability. Compared to deterministic approaches, this optimised detection system could detect a hot particle on average 10cm deeper into the soil column or with half of the activity at the same depth. It was also found that noise presented by internal contamination restricted lanthanum bromide for this application.

Assessing the possible radiological impact of routine radiological discharges from proposed nuclear power stations in England and Wales.

The aim of this work is to assess the possible radiological impact on the population of the United Kingdom (UK) from new nuclear power stations proposed for up to eight sites in England and Wales. The radiological impact was measured in terms of collective dose to the UK, European and world populations from a single year's discharge integrated to 500 and 100 000 years and the annual dose to an average member of the UK population (known as the per-caput dose). The doses were calculated for two reactor types, UK EPR™ and AP1000™, using the annual expected discharges estimated by the designers of the reactors and assuming two reactors per site. In addition, typical individual doses to adults living close to the sites were calculated on the basis of continuous discharges for 60 years (the assumed lifetime of the reactors). The dose to a representative person (previously known as the critical group) was not calculated, as this has been done elsewhere. The assessments were carried out using the software program PC-CREAM 08(®) which implements the updated European Commission methodology for assessing the radiological impact of routine releases of radionuclides to the environment. The collective dose truncated to 500 years to the UK population was estimated to be 0.5 manSv assuming UK EPR reactors on all sites and 0.6 manSv assuming AP1000s on three sites with UK EPRs on the other sites. The most significant contribution to the collective dose to the UK population is due to the global circulation of carbon-14 released to the atmosphere. The annual dose to an average member of the UK population from all sites was calculated to be around 10 nSv y(-1) and would therefore contribute little to an individual's total radiation dose. All the calculated doses to a typical adult living near the sites assuming continuous discharges for 60 years were found to be below 1 µSv y(-1).

Plutonium in the WIPP environment: its detection, distribution and behavior.

The Waste Isolation Pilot Plant (WIPP) is the only operating deep underground geologic nuclear repository in the United States. It is located in southeastern New Mexico, approximately 655 m (2150 ft) below the surface of the Earth in a bedded Permian evaporite salt formation. This mined geologic repository is designed for the safe disposal of transuranic (TRU) wastes generated from the US defense program. Aerosol and soil samples have been collected near the WIPP site to investigate the sources of plutonium in the WIPP environment since the late 1990s, well before WIPP received its first shipment. Activities of (238)Pu, (239+240)Pu and (241)Am were determined by alpha spectrometry following a series of chemical separations. The concentrations of Al and U were determined in a separate set of samples by inductively coupled plasma mass spectrometry. The annual airborne concentrations of (239+240)Pu during the period from 1998 to 2010 show no systematic interannual variations. However, monthly (239+240)Pu particulate concentrations show a typical seasonal variation with a maximum in spring, the time when strong and gusty winds frequently give rise to blowing dust. Resuspension of soil particles containing weapons fallout is considered to be the predominant source of plutonium in the WIPP area. Further, this work characterizes the source, temporal variation and its distribution with depth in a soil profile to evaluate the importance of transport mechanisms affecting the fate of these radionuclides in the WIPP environment. The mean (137)Cs/(239+240)Pu, (241)Am/(239+240)Pu activity ratio and (240)Pu/(239)Pu atom ratio observed in the WIPP samples are consistent with the source being largely global fallout. There is no evidence of any release from the WIPP contributing to radionuclide concentrations in the environment.

Tritium release during nuclear power operation in China.

Overviews were evaluated of tritium releases and related doses to the public from airborne and liquid effluents from nuclear power plants on the mainland of China before 2009. The differences between tritium releases from various nuclear power plants were also evaluated. The tritium releases are mainly from liquid pathways for pressurised water reactors, but tritium releases between airborne and liquid effluents are comparable for heavy water reactors. The airborne release from a heavy water reactor is obviously higher than that from a pressurised water reactor.

An assessment of the radiological scenario around uranium mines in Singhbhum East district, Jharkhand, India.

The present work deals with the prevalent radiological scenario around uranium-mining sites in the Singhbhum East district of Jharkhand state, India. The concentration of naturally occurring radioactive materials (NORMs) was estimated from 27 soil samples collected around three regions in the study area, namely Bagjata, Turamdih and Jaduguda. The mean activity concentrations of (238)U in Bagjata, Turamdih and Jaduguda regions were found to be 128.6, 95.7 and 49.2 Bq kg(-1), respectively. Similarly for (232)Th and (40)K the activity concentrations were found to be 57.3, 78.4, 68.9 and 530, 425 and 615 Bq kg(-1) in the Bagjata, Turamdih and Jaduguda regions, respectively, which are comparable with other reported values worldwide, except for some high values. The calculated gamma dose rate, obtained from the concentrations of (238)U, (232)Th and (40)K in the samples, was compared with the observed dose rate in air. A good correlation (0.96) was observed between the calculated and the observed gamma dose rate. The annual outdoor effective dose rate was estimated and the values falls between 0.04-0.3, 0.07-0.3 and 0.07-.14 mSv y(-1) with mean values of 0.14, 0.12 and 0.11 mSv y(-1) for the Bagjata, Turamdih and Jaduguda regions, respectively. The terrestrial dose rates in all the three regions are comparable with other reported values worldwide, except for a few high values in Greece, Rio Grande Do Norte (Brazil) and Kalpakkam (India).

Composition and production rate of medical waste from a small producer in Greece.

The objective of this work was to determine the composition and production rate of medical waste from the health care facility of social insurance institute, a small waste producer in Xanthi, Greece. Specifically, produced medical waste from the clinical pathology (medical microbiology) laboratory, the X-ray laboratory and the surgery and injection therapy departments of the health facility was monitored for six working weeks. A total of 240 kg medical solid waste was manually separated and weighed and 330 L of liquid medical waste was measured and classified. The hazardous waste fraction (%w/w) of the medical solid waste was 91.6% for the clinical pathology laboratory, 12.9% for the X-ray laboratory, 24.2% for the surgery departments and 17.6% for the injection therapy department. The infectious waste fraction (%w/w) of the hazardous medical solid waste was 75.6% for the clinical pathology laboratory, 0% for the X-ray laboratory, 100% for the surgery departments and 75.6% for the injection therapy department. The total hazardous medical solid waste production rate was 64+/-15 g/patient/d for the clinical pathology laboratory, 7.2+/-1.6 g/patient/d for the X-ray laboratory, 8.3+/-5.1 g/patient/d for the surgery departments and 24+/-9 g/patient/d for the injection therapy department. Liquid waste was produced by the clinical pathology laboratory (infectious-and-toxic) and the X-ray laboratory (toxic). The production rate for the clinical pathology laboratory was 0.03+/-0.003 L/patient/d and for the X-ray laboratory was 0.06+/-0.006 L/patient/d. Due to the small amount produced, it was suggested that the most suitable management scheme would be to transport the hazardous medical waste, after source-separation, to the Prefectural Hospital of Xanthi to be treated with the hospital waste. Assuming this data is representative of other small medical facilities, medical waste production can be estimated for such facilities distributed around Greece.

Public exposure to hazards associated with natural radioactivity in open-pit mining in Ghana.

The results of studies carried out on public exposure contribution from naturally occurring radioactive materials (NORMS) in two open-pit mines in the Western and Ashanti regions of Ghana are reported. The studies were carried out under International Atomic Energy Agency-supported Technical Co-operation Project GHA/9/005. Measurements were made on samples of water, soil, ore, mine tailings and air using gamma spectrometry. Solid-state nuclear track detectors were used for radon concentration measurements. Survey was also carried out to determine the ambient gamma dose rate in the vicinity of the mines and surrounding areas. The effective doses due to external gamma irradiation, ingestion of water and inhalation of radon and ore dusts were calculated for the two mines. The average annual effective dose was found to be 0.30 +/- 0.06 mSv. The result was found to be within the levels published by other countries. The study provides a useful information and data for establishing a comprehensive framework to investigate other mines and develop guidelines for monitoring and control of NORMS in the mining industry and the environment as a whole in Ghana.

Lauriston S. Taylor Lecture: Yucca mountain radiation standards, dose/risk assessments, thinking outside the box, evaluations, and recommendations.

The Yucca Mountain high-level radioactive waste repository is designed to contain spent nuclear fuel and vitrified fission products. Due to the fact that it will be the first such facility constructed anywhere in the world, it has proved to be one in which multiple organizations, most prominently the U.S. Congress, are exercising a role. In addition to selecting a site for the facility, Congress specified that the U.S. Environmental Protection Agency (U.S. EPA) promulgate the associated Standards, the U.S. Nuclear Regulatory Commission establish applicable Regulations to implement the Standards, and the U.S. Department of Energy (U.S. DOE) design, construct, and operate the repository. Congress also specified that U.S. EPA request that the National Academy of Sciences (NAS) provide them guidance on the form and nature of the Standards. In so doing, Congress also stipulated that the Standards be expressed in terms of an "equivalent dose rate." As will be noted, this subsequently introduced serious complications. Due to the inputs of so many groups, and the fact that the NAS recommendations conflicted with the Congressional stipulation that the limits be expressed in terms of a dose rate, the outcome is a set of Standards that not only does not comply with the NAS recommendations, but also is neither integrated, nor consistent. The initial goals of this paper are to provide an independent risk/dose analysis for each of the eight radionuclides that are to be regulated, and to evaluate them in terms of the Standards. These efforts reveal that the Standards are neither workable nor capable of being implemented. The concluding portions of the paper provide guidance that, if successfully implemented, would enable U.S. DOE to complete the construction of the repository and operate it in accordance with the recommendations of NAS while, at the same time, provide a better, more accurate, understanding of its potential risks to the public. This facility is too important to the U.S. nuclear energy program to be impeded by inappropriate Standards and unnecessary regulatory restrictions. As will be noted, the sources of essentially all of the recommendations suggested in this paper were derived through applications of the principles of good science, and the benefits of "thinking outside the box."