Potential Radiological Consequences Research Articles

REMADA: An Emergency Decision-Making Software for Radiation Dose Estimation of Mitigation Tasks in a Reactor Building During Severe Nuclear Accidents

In the event of a postulated severe accident, fission products may leak into the reactor building through the containment wall, exerting a radiological impact on the emergency response team as they are tasked with performing mitigative missions. It is thus important to estimate the potential radiological consequences of the mission before taking action so that an optimized plan can be devised to avoid putting the team in harm’s way unintentionally. Some of the most well-known accident simulation codes were attempted to do the estimation, but were found to be too time consuming to get the results, making them not suitable for emergency use. The problem evidently arises from the fact that there are too many (about 200) compartments in a typical reactor building. In this study, the software REMADA is developed to simulate fission product dispersion in a reactor building (with about 200 compartments) within a reasonable timeframe, and to estimate the radiation doses to those who are carrying out mitigative missions in the reactor building. The results show that the software is not only fast, but also informative, to provide support for well-informed emergency decision making.

Интегральная оценка нуклидного состава и тепловыделения потенциальных ядерных отходов, образованных топливом в бассейне выдержки энергоблока 1 АЭС Фукусима Дайичи

Assuming an accident that may damage spent fuel assemblies at an NPP, nuclear fuel in the spent fuel pool can be considered as radioactive waste. Data on its nuclide composition and heat generation is seen important for the safety assessment of its storage and potential radiological consequences in case of an accident. The SOCRAT/V3 integral code designed to simulate severe accidents at nuclear power plants with VVER power units is able to estimate the nuclide inventory of irradiated fuel based on a one-through calculation and a built-in model. This code was used for a computational analysis of the spent fuel from the Fukushima Daiichi Unit 1, namely, evaluated was its nuclide composition and its decay heat release. The results were compared with the reference data from JAEA (Japan). The calculations demonstrate that in case of severe accident simulations, the use of an integral approach causes no large errors in the estimated nuclide inventory. For most dose-forming radionuclides in the spent fuel, relative deviation of activity from the reference data does not exceed 15%, which appears to be lower than the typical uncertainty under source term predictions. Despite certain overestimation tendency that can be noticed, the decay heat levels calculated for the spent fuel pool appear to be quite similar to those presented in the JAEA report differing by no more than 16% in the first years after the accident, and by 5% ten years after the accident.

Searching for Abandoned Radioactive Sources in Scrap Yards at Al-Tuwaitha Nuclear Site

In the present study, the scrap yards (located within the latitudes 33o 21’ 613” to 33o 21’ 290” North, and longitudes 44o 51’ 299” to 44o 50’ 904” East) at Al-Tuwaitha nuclear site were scanned for the period from January to February 2016 using hand-held radiation detectors with the intent of locating radioactive sources thrown within piles of scrap material. Only one 137Cs radioactive source has been found with an estimated activity of 6.3×108 Bq, which was found to be considerably less than the “dangerous quantity” D-value of 1×1011 Bq. Assessment of the potential radiological consequences arising from exposure to the source during normal conditions has been conducted using Rad Pro Calculator Version 3.26 (2009). The potential radiological dose arising from normal occupational exposure situation was estimated to be 139 mSv/y at 1 meter distance from the source, which was considerably higher than the occupational dose limit of 20 mSv/y recommended in the International Atomic Energy Agency (IAEA) GSR Part 3 (2014), indicating that this source must be removed and treated as radioactive waste.

Experience in radiological risk assessment of a surface waste disposal facility in Chişinău, Moldova

The long-term condition and potential radiological consequences of legacy radioactive waste stored in a RADON-type of near-surface disposal facility outside the city of Chişinău is of concern to the central government and health protection authorities of the Republic of Moldova. A ‘zero alternative scenario’ risk assessment has been undertaken in order to evaluate the potential radiological impact on humans and the environment of the facility, were it to be left in its current state with no remediation. The results have been used as a basis for regulatory decision making regarding remediation and decommissioning of the legacy radioactive waste facility. The aim of this study was two-fold: first to demonstrate a complete radiological risk assessment of a real site using a combination of methodologies developed by the IAEA (ISAM and BIOMASS), the second to illustrate the current state-of-the-art in respect of extracting site-specific information from site-descriptive material. We illustrate the practicality of employing geographic information systems techniques on site-specific topographic data to identify relevant biosphere dose objects, thereby allowing customisation of the generic ISAM model framework to site-specific conditions. As a result, a simple method is suggested to bound activity concentrations in well water based on an understanding of water balance in the local catchment area in which the biosphere dose object is embedded. With conservative assumptions, estimated doses from the calculation cases of the design scenario remain lower than the IAEA’s dose criteria and environmental screening values. However, the results also indicate that human intrusion activities after the institutional control period could lead to radiological exposures above the IAEA’s criteria for a period up to 100 000 years. The long-lived radionuclide 239Pu dominates doses for the on-site residence scenario. Remediation measures should be implemented were the waste to remain at its present place of disposal.

Interaction process between gaseous CH3I and NaCl particles: implication for iodine dispersion in the atmosphere.

Gaseous iodomethane (CH3I) is naturally emitted into the atmosphere by biological activity in oceans and during severe accidents (SAs) in nuclear power plants. In this latter case, a part of radioactive iodine such as 131I may be released. Improving the knowledge of CH3I transport and reactivity in the atmosphere is important since they are strongly linked to first the cycle of ozone and second to the dispersion of radioactive CH3I with potential radiological consequences on both the environment and human health. Here, the interaction process of CH3I with NaCl as a surrogate of atmospheric aerosols was investigated under ambient air conditions by using Diffuse Reflectance Fourier Transform Spectroscopy (DRIFTS). The DRIFTS spectra of NaCl clearly evidenced CH3I adsorption on the NaCl particle surface. A part of CH3I ((1.68 ± 0.85) × 1014 molecule per mgNaCl) was found to be strongly bonded to NaCl since no desorption was observed. The CH3I adsorption on the NaCl surface presented a 1st order kinetics relative to its gas phase concentration. The uptake coefficient was determined to be in the order of 10-11. These results show a low probability of CH3I to be taken up by halide-containing aerosols. These data are crucial for completing the iodine atmospheric chemical scheme.

Radionuclides emissions from sodium pool fires: Learning from the 1970s-era FANAL experiments

The emission of radionuclides from sodium fires is a very important concern when considering the potential radiological consequences of an accident at a sodium fast reactor. Radionuclides can be released when contaminated sodium from, for example, the primary coolant system, catches fire. Not all radionuclide species will behave the same way, however, and some will have a higher tendency to become aerosolized than others. This paper uses historical experimental results from the 1970s-era FANAL program in order to come up with estimates for the combustion partition coefficients for radioiodine, radiocesium, and several other radionuclides from pool fires, and uses information from the program to characterize the chemical interaction between radioiodine vapor and sodium combustion product aerosols.

ASTEC V2 severe accident integral code: Fission product modelling and validation

One main goal of the severe accident integral code ASTEC V2, jointly developed since almost more than 15 years by IRSN and GRS, is to simulate the overall behaviour of fission products (FP) in a damaged nuclear facility. ASTEC applications are source term determinations, level 2 Probabilistic Safety Assessment (PSA2) studies including the determination of uncertainties, accident management studies and physical analyses of FP experiments to improve the understanding of the phenomenology.ASTEC is a modular code and models of a part of the phenomenology are implemented in each module: the release of FPs and structural materials from degraded fuel in the ELSA module; the transport through the reactor coolant system approximated as a sequence of control volumes in the SOPHAEROS module; and the radiochemistry inside the containment nuclear building in the IODE module. Three other modules, CPA, ISODOP and DOSE, allow respectively computing the deposition rate of aerosols inside the containment, the activities of the isotopes as a function of time, and the gaseous dose rate which is needed to model radiochemistry in the gaseous phase.In ELSA, release models are semi-mechanistic and have been validated for a wide range of experimental data, and noticeably for VERCORS experiments. For SOPHAEROS, the models can be divided into two parts: vapour phase phenomena and aerosol phase phenomena. For IODE, iodine and ruthenium chemistry are modelled based on a semi-mechanistic approach, these FPs can form some volatile species and are particularly important in terms of potential radiological consequences. The models in these 3 modules are based on a wide experimental database, resulting for a large part from international programmes, and they are considered at the state of the art of the R&D knowledge.This paper illustrates some FPs modelling capabilities of ASTEC and computed values are compared to some experimental results, which are parts of the validation matrix.

Post-Closure Performance Assessment of a Deep Geological Repository for Advanced Heavy Water Reactor Fuels

Many countries worldwide are investigating the use of advanced fuels and fuel cycles for purposes such as increasing the sustainability of the nuclear fuel cycle, or decreasing the radiological impact of used fuel. One common metric used to assess the radiological impact to humans of fuels placed in a repository is the total radiotoxicity of the fuel, but this approach does not take into account how engineered and natural (i.e., rock) barriers can remove many radiotoxic nuclides from ground water before they reach the surface. In this study, we evaluate the potential radiological dose consequences of advanced fuels in the context of a full system model simulation for release and transport from a repository, transport through the surrounding geosphere, release to the biosphere and dose consequences for the target critical group. Heavy water moderated reactors, such as the CANDU® reactor, are well-suited to the use of advanced fuels, and the post-closure performance of a deep geological repository for spent natural uranium fuel from them has already been studied. For this study, two advanced fuels of current interest were chosen: a TRUMOX fuel designed to recycle plutonium and minor actinides and thereby reduce the amount of these materials going into disposal, and a plutonium thorium-based fuel whose main goal is to increase sustainability by reducing uranium consumption. The impact of filling a deep geological repository, of identical design to that for natural uranium, with used fuel from these fuel cycles was analyzed. It was found that the two advanced fuels analyzed had dose rates, to a hypothetical critical group of humans living above the repository, which remained a factor of 170 to 340 lower than the current acceptance limit for releases, while being 5.3 (for TRUMOX) and 2.6 (for thorium-plutonium) times higher than those of natural uranium. When the dose rates are normalized to total energy produced, the repository emissions are comparable. In this case, the maximum dose rates were found to be 6% lower for the TRUMOX fuel, and 16% higher for the plutonium thorium fuel, than for the used natural uranium fuel.

Radiological consequences of potential sabotage attack to storage casks on the ISFSI site

Abstract Interim storage is one major program of fuel cycle. Because there are some individuals staying or living very close to the ISFSI site that is being constructed, this paper examines the potential radiological consequences of hypothesized events for storage casks. This study uses ORIGEN, HOTSPOT and RESRAD software to calculate the source terms, acute dose and long-term dose, separately. The results show that explosive-induced buoyancy would disperse the source much farther than the site boundary in a sabotage accident. Nevertheless, the amount of radioactivity released to offsite would be very low, and the maximum dose received by any individual staying or living in the vicinity of ISFSI, even under the most conservative scenario, would be very much less than the local public relocation dose limit.

Ten years of Romanian experience on risk assessment owing to radioactive materials transport in Romania

AbstractThe transport of radioactive materials is a very important problem considering the potential risks and radiological consequences in carrying out the present activity. Based on the International Atomic Energy Agency (IAEA)'s Safety Standard TS-R-1 (1996 edition, as amended 2003), Romanian National Nuclear Regulatory Body – Romanian National Commission for Nuclear Activities Control (CNCAN) was adopted and implemented by act no. 374/October 2001, the safety regulations for the transport of radioactive materials in Romania under the title 'Fundamental regulations for a safe transport of radioactive materials, in Romania'. The present paper will present the main sources of radioactive materials in Romania, their transport routes with a particular interest paid to the radioactive wastes. Hypothetical scenarios for specific problems related to the identification and evaluation of the risks and potential radiological consequences associated with the transport of radioactive materials in Romania, for all ...

LAURISTON S. TAYLOR LECTURE: RADIATION PROTECTION IN THE AFTERMATH OF A TERRORIST ATTACK INVOLVING EXPOSURE TO IONIZING RADIATION

I would like to start this Twenty-Eighth Lauriston S. Taylor Lecture by expressing my gratitude to the National Council on Radiation Protection and Measurements (NCRP) for this unique occasion. I feel particularly honored for this opportunity to address a highly specialized and qualified audience of professionals who are specifically interested in what appears to be a forthcoming worldwide challenge, namely radiological terrorism and managing its potential radiological consequences.

Qualification Tests for a Type B(U) Package

AbstractThe primary objective for the safety of radioactive materials transport is to protect human health and the environment taking into consideration its potential risks and radiological consequences. Romania as a Member State of the International Atomic Energy Agency hasimplemented national regulations for the safe transport of radioactive materials in accordance with the Agency's recommendations as well as other international specialisedorganisations. The paper will describe the qualification tests performed for a Type B(U) package, intended to be used for the transport of the radioactive sources 241Am and 137Cs. For this kind of package the tests were performed for the first time inRomania and include: the water spray test, the 1.2 m free drop test, the stacking test, the penetration test, the 9 m free drop test, the thermal test and the submersion under a head of water of at least 15 m. The test facilities used for performing qualification tests for the Type B(U) package as well as experience and c...

Dose displays and record keeping.

In digital diagnostic and interventional radiology, patient dosimetry is essential. When considering the display requirements for interventionalists it is necessary to consider the basis for their information needs. An interventionalist must be made aware of the potential for deterministic and non-deterministic effects. Thus a display is required which provides an indication of whether skin effects are likely and this predicates a display of maximum estimated skin entrance dose. For non-deterministic effects, a display of dose-area product is useful. Unfortunately, neither display is generally available to the interventionalist, so inevitably they will remain in the dark about the potential radiological consequences of the procedure. Given the need to follow up patients, who may receive a high dose, it is important that the patient notes contain relevant dosimetry data. One way of achieving this in the future would be to link electronically the patient dosimetry device to the patient record in the Radiological Information System. This dictates a need to develop patient dosimetry record minimum dataset and equipment connectivity standards.

98/03378 Modelling the potential radiological consequences of radioactive waste dumping in the Kara Sea: Baxter, M. S. et al. J. Environ. Radioactivity, 1998, 39, (2), 161–181

98/03378 Modelling the potential radiological consequences of radioactive waste dumping in the Kara Sea: Baxter, M. S. et al. J. Environ. Radioactivity, 1998, 39, (2), 161–181

Modelling the potential radiological consequences of radioactive waste dumping in the Kara Sea

There has recently been growing concern over the dumping of high- and medium-level solid radioactive wastes in the Kara Sea by the former Soviet Union. The largest amounts of radioactive wastes were dumped primarily as nuclear reactors containing spent nuclear fuel. The present radionuclide inventory in dumped nuclear reactors is estimated at 4·7 PBq. Compartmental and hydrodynamic models have been developed and applied to describe the possible dispersal of radioactive contaminants and to predict the long-term radiological impact on global, regional and local scales. The collective committed effective dose to the world population based on the marine food ingestion pathway has been calculated as 2·2 man Sv. Modelling results suggest that only radiological effects on a local scale may be of importance. The global radiological impact of the disposals in the Kara Sea will be smaller than from other anthropogenic sources of radioactivity.

Risk Assessment of Radioactive and Hazardous Materials in DOE Defence Package Transport Accidents

AbstractA programme is being conducted by the US Department of Energy (DOE) to determine the risks to the public from domestic transport of radioactive and hazardous materials associated with nuclear defence materials. The programme will identify, quantify and manage potential risks to public health and safety including potential radiological and toxicological health consequences, particularly for those transport packages that may not meet the 10 CFR 71.73 (US Code of Federal Regulations) test criteria because of configuration or content. Lawrence Livermore National Laboratory and the University of Utah will provide the DOE with the bases and methodology for establishing rigorous, quantitative acceptance criteria for evaluating and certifying highway transport activities by DOE contractors. The Highway Transportation Risk Assessment (HITRA) model is described which is based upon probabilistic risk assessment (PRA) methodology and utilises transport route specific data. The HITRA model provides detailed, l...

Chapter 2: Description of Methods

AbstractThis investigation encompasses the determination of potential radiological consequences from normal (accident-free) waste transport and the transport accident risk. The investigation is based on the developmental stage as of autumn 1990 of the planning and licencing documents for the Konrad repository. This applies in particular to: (i) the intended waste and transport containers, (ii) the local dose rates and radioactivity inventories of the waste packages, and (iii) the planned sequence of the storage operations.

Chapter 5: Waste Transport; Transport Scenarios and Arrangements

The potential radiological consequences of waste transport to the Konrad waste disposal site depend on the nature and number of shipments and the associated waste transport routes.

Impact of Secondary Effects on the Reduction of Fission Product Source Terms in Class IX Reactor Accidents

This paper examines the potential radiological consequences of a hypothetical line-break, core melt type of accident (Class IX). This paper concludes that the potential for catastrophic environmental release of radioactive material in case of a major reactor accident seems to be greatly suppressed by inherent processes provided by the coolant, the primary system, and the containment, without significant reliance on engineered safety features. 18 refs.