An inverse source estimation method is proposed to reconstruct emission rates of multi-radionuclides using local gamma dose rate measurements under the data assimilation framework. It involves the Proper Orthogonal Decomposition (POD)-based ensemble four-dimensional variational data assimilation (PODEn4DVar) algorithm and a transfer coefficient matrix (TCM) created using FLEXPART, a Lagrangian atmospheric dispersion model. PODEn4DVar is a hybrid data assimilation method that exploits the strengths of both the ensemble Kalman filter (EnKF) and the 4DVar assimilation method. With an explicit expression of control (state) variables in the cost functional, the data assimilation process is substantially simplified than traditional 4D variational method. By setting a unit emission rate and running the ATDM model (FLEXPART in this article) driven by meteorological fields forecasted with WRF, we get the transfer coefficient matrix with the progression of nuclear accident. TCM not only acts as observation operator in PODEn4DVar, but also eliminates the control run in traditional data assimilation framework. The method is tested by twin experiments with ratios of nuclides assumed to be known. With pseudo observations based on Fukushima Daiichi nuclear power plant (FDNPP) accident, most of the emission rates were estimated accurately, except under conditions when wind blew off land toward the sea and at extremely slow wind speeds near the FDNPP. Because of the long duration of accident and variability of meteorological fields, measurements from land only in local area is unable to offer enough information to support emergency response. With abundant measurements of gamma dose rate, emission rates can be reconstructed sequentially with the progression of nuclear accident. Therefore, the proposed method has the potential to be applied to nuclear emergency response after improvement.

The control room radiological habitability (CRRH) is important for staff safety in a nuclear power plant, which is also a licensing requirement of the High-temperature Reactor Pebble-bed Module (HTR-PM) in China. Meanwhile, the complexity of the dose assessment increases for the multi-reactor site, which put forward higher requirements for building layout. The CRRH is investigated comprehensively for the multi-reactor site at Shidao Bay in this study. For a large-break loss of coolant accident of HTR-PM and CAP1000 in Shidao Bay nuclear power site, this study estimates doses of body, thyroid and skin due to three exposure pathways using NRC-recommended ARCON96 and dose calculation method in RG 1.195. To perform a realistic evaluation, the latest design and site-specific information are utilized as the input parameters, including the unique accidental source term of HTR-PM and the RG1.183-recommended source term of CAP1000, the release and ventilation parameters, the final layout and the meteorological data in a whole year. The evaluation results demonstrate that the individual dose level of staff in the control room is far below the requirement of the regulatory guide, which guarantees the CRRH of HTR-PM. The contribution of primary radionuclides suggests that tellurium and iodine are primary contributors of the inhalation dose of body and thyroid, which is worthy of paying particular attention to the CRRH design in HTR-PM.

Determination of the SMR emergency planning zone (EPZ) is one of the important external constraint factor of its marketing and application, which means that it is very important to formulate appropriate classification criteria and establish proper size range. In China, due to the requirement of “Criteria for emergency planning and preparedness for nuclear power plants: Part 1, The dividing of emergency planning zone.” (GB/T 17680.1-2008), for PWR nuclear power plant, its external plume EPZ should be within 7km–10km, and its internal plume EPZ should be within 3km∼5km. However, the scope of the standard for the emergency planning area is currently limited to conventional nuclear power plants, and for the current SMR, its emergency planning size is not included. In this paper, we will analyze the classification method of SMR EPZ based on the traditional Nuclear Power Plants feedback experience, including selection of source term, accident cutoff probability, determination method of the plume EPZ and the ingestion EPZ. Three typical nuclear power plant sites in China are chosen as CAP200 case study sites, including two inland nuclear power plant sites and one coastal site. The three sites can represent most of the meteorological and terrain characters of China nuclear power plants. According to the CAP200 source term and meteorological data of the sites, MACCS2 computer program is used to calculate the severe accidents consequence. Conclusions show that for the CAP200 SMR, the accident cutoff probability can be 1.0E−08 to 1.0E−07 per reactor per year, and its project dose exceeding probability in the three sites boundary is far below 30%, which directs that for CAP200 SMR, its plume and ingestion emergence planning zone is limited to the on-site area, and its off-site emergency response can be simplified.

Food contamination has aroused public concern since Fukushima accident. As emergency preparedness is often viewed as an important approach to protect staff working on site and public around the site, ingestion emergency planning zone (EPZ) is applied to protect public from the exposure of contaminated food. Ingestion EPZ is one of the technical foundations for nuclear emergency preparedness, which will be influenced by design features of plant and characteristics of the site. This paper is devoted to the research on the optimization of ingestion EPZ sizing from the view of the atmospheric dispersion model and the food chain model, which are crucial points for the sizing of ingestion EPZ. Compared to the traditional straight-line Gaussian plume model with a quite conservative assumption that plume segments always transport in the downwind direction, the Lagrangian Gaussian puff model considers the swing of wind direction over time, which makes the simulation more realistic. With the results of radionuclide concentrations evaluated by the dispersion model, the transportation of the radionuclides in food is simulated by the food chain model. The traditional food chain model is essentially a static model with no consideration that food contamination level has a strong dependence on the accident date, which may overstate the risk from nuclear plant accidents and result in unfounded fear of public. The dynamic food chain model, which takes daily changes of plant biomass, or livestock feeding periods in consideration, has been developed to estimate radionuclide concentrations in different foodstuffs. On basis of the study of the dispersion models and food chain models above, we evaluate the ingestion EPZ size of Tianwan NPP by choosing the comparatively realistic ones from them. In the scenario considered in this paper, the simulation domain of Tianwan NPP within 80km-range and hourly time-step is applied, and meteorological conditions are carefully set according to observation data in recent years. Results show that there is significant margin and conservatism in the traditional ingestion EPZ sizing. Radionuclide concentrations predicted by the Lagrangian Gaussian puff model is almost an order of magnitude lower than the Gaussian plume model. Moreover, the dynamic food chain model considers the seasonal effect that simulation results of radionuclide concentrations in foodstuffs are significantly higher in summer than in winter, which helps to make a more realistic consideration of ingestion pathway. This research gives an example of the application of new models for the optimization of ingestion EPZ sizing, which may contribute to strengthen public confidence in nuclear safety and emergency preparedness.

The main control room (MCR) ventilation system has been designed to maintain habitability of the control room envelope both under normal condition and accident condition. The system adopting dual air intakes adds one more air intake for accidents at suitable position. During accidents, the air intake with lower contamination will be selected and the other with higher contamination will be isolated, to reduce the amount of radioactive substances entering MCR extremely and enhance the habitability of MCR envelop. This paper is devoted to research on the impact of switching time interval (STI) for dual intakes on workers in main control room during accidents. As the contamination condition varies, the switching action will be happened. Switching time interval (STI) referred in this paper means the time between two switching actions. When accidents occur, the air intake will operate and switch between two intakes automatically. The action of switching will be influenced by several parameters: the meteorological conditions of the site, the response features of the monitoring instruments and the source term released to the environment after accidents. Analysis of these parameters and their sensitivity analysis are performed, which show that the ventilation system cannot afford too frequent switching actions resulted from instantaneous sudden changes of intake’s activity. That’s the reason why it is necessary to set a minimum STI which means the contamination of one intake have to be lower than the other intake and this dominant position should be kept longer than the minimum STI, if not, the switching action will not be happened. As it is essential to set a minimum STI to prevent frequent switching of system, the analysis of its impact on the atmospheric relative concentrations and the doses of the workers in main control room are performed on basis of specific site meteorological condition and the response characteristic of dose monitoring instruments. Three kinds of accident release conditions are considered, which are relief valve release, containment leakage and elevated funnel release. The atmospheric relative concentrations and the doses of the workers in MCR are evaluated for every case and compared with the dose limits. Finally an acceptable minimum STI of dual air intakes is recommended.

In this paper, air-immersion, ground deposition, ingestion and inhalation of airborne radioactive effluent released from nuclear power plant under normal operating conditions is studied according to the atmospheric diffusion and ground deposition patterns and parameters that are suitable for the environmental characteristics of the nuclear power plant site, and the public living habits and food chain parameters around the site. Based on the Gaussian plume model, with a radius of 80 kilometers we divide 1, 2, 3, 5, 10, 20, 30, 40,50,60,70,80 km concentric circles around the nuclear power plant site. The 16 compass azimuth axial are the sector center-line, forming a total of 192 sub-regions, atmospheric diffusion of radionuclides is simulated in the assessment area of the region. The annual average atmospheric dispersion factor is calculate by using hourly observation data of wind direction, rainfall and atmospheric stability of the meteorological tower and the ground station, taking into account the ground reflection during transmission, the the decay of the radionuclide, and the loss brought by the wet and dry settling that caused by gravity and rain washing. The airborne radioactive effluent is deposited on the ground or plant surface by dry settling and wet settling in the process of atmospheric environment changing and diffusion. Radioactivity of per unit area brought about by dry settling and rain fall settling is described by the deposition coefficient and deposition speed. The long-term ground deposition factor and ground annual concentration in the evaluation area were calculated under the situation of airborne radioactive effluents in the nuclear power station mixing emission, and the calculated result of radionuclide concentration in the air and soil was compared with the natural background value and the actual monitoring value. Based on the radionuclide deposited on the ground and air through the terrestrial food radioactive transfer mode, together with a large number of environmental surveys data on the population distribution, agriculture, farming, animal husbandry and people’s living and eating habits in the 80km around nuclear station, combing with the actual situation of nuclear power station, the calculation model is amended accordingly. Using reasonable dose mode to calculate the maximum individual and entire public effective dose of the residents in the assessment area, and the results will be compared with other human activities. By comparing the calculated results of radionuclide concentration and radiation dose, it provide quantitative reference information for us understanding the influence of nuclear power station on the surrounding radiation environment, and to meet the requirements of nuclear power plant influence on surrounding environment and people under normal operating conditions.

Operation intervention level (OIL) is a type of action level that can be used immediately and directly (without further assessment) to determine the appropriate protective actions and other response actions on the basis of an environmental measurement. OIL is used to implementation actions to protect the public due to the severe conditions at a nuclear power plant. This paper presents an integrated solution that integrates remote access and remote support over the Internet. The system real-time acquire the data of source term, meteorological, off-site monitoring points, monitoring vehicles, execute operation intervention level calculation, protective action decision-making to public based on data from database or input by user. The system simultaneous input multiple monitor data, and protection of the public action based on the results of multiple OIL decisions. The system display off-site monitoring points, protection of the public area of operations and administrative village using WebGIS.

PAVAN is an atmospheric dispersion program for evaluating design basis accidental releases of radioactive materials from nuclear power plant. It was developed by Pacific Northwest Laboratory on the basic of the atmospheric dispersion models described in RG 1.145 by NRC (U.S. Nuclear Regulatory Commission). Using the joint frequency of wind direction, wind speed and atmospheric stability, the atmospheric relative concentration values for the exclusion area boundary and outer LPZ boundary of nuclear power plant are calculated and given by the program. Once the program was introduced, it has been widely used in the radioactive accident consequence assessment, especially in the FSAR (Final Safety Analysis Report) and Report of EIA (Environmental Impact Assessment) of NPPs in China. The theory basis and general method of PAVAN is introduced in this paper. And specialty of the X/Q points based on joint frequency data is discussed. The envelope algorithm of PAVAN is also introduced and discussed. The paper presents an improved algorithm based on PAVAN which uses the hourly meteorological data as input instead of joint frequency data. In this algorithm, the size of X/Q points is related to the quantity of the hourly meteorological data. When the quantity is large enough, e.g. 17520 sets of hourly meteorological data in two years, the envelope curve for X/Q points fit more exactly than PAVAN. Using the observed meteorological data, the improved algorithm is compared with PAVAN. The result proves that the former is more accurate. In general, the improved algorithm is relatively conservative. In some situation, the conservativeness is not certain. The factors which result in the uncertainty are deeply discussed. Further optimized are performed by the algorithm. The number of points to seek in envelope curve fitting is set to be dynamic and be a quarter of total number of X/Q points to be fitted. The result shows that increasing the number of points to seek in the iteration process of envelope curve fitting will lead to more conservative X/Q values. Additionally, the optimized algorithm provides X/Q value of 50% probability level for overall site. The value is not relatively conservative. From the standpoint of statistical probability, it is more realistic and is acceptable for potential accident consequence assessment. Especially, when X/Q value of 95% probability level for overall site is too conservative to accept, the value of 50% probability level can be used to replace the conservative value.

A Probabilistic Risk Assessment (PRA) should be performed not only for earthquake and tsunami which are major natural events in Japan, but also for other natural external hazards. However, PRA methodologies for other external hazards and their combination have not been sufficiently developed. This study is intended to develop PRA methodology for a combination of low temperature and snow for a Sodium-cooled Fast Reactor (SFR) that uses the ambient air as its ultimate heat sink for decay heat removal under accident conditions. Annual excess probabilities of low temperature and of snow are statistically estimated based on the meteorological records of low temperature, snow depth and daily snowfall depth. To identify core damage sequence, an event tree was developed by considering the impact of low temperature and snow on decay heat removal systems (DHRSs), e.g., plugged intake and/or outtake for the DHRS and for the emergency diesel generator (EDG), unopenable door on the access routes due to accumulated snow, failure of the intake filters due to accumulated snow, possibility of freezing of the water in cooling circuits. Recovery actions (i.e., snow removal and filter replacement) to prevent loss of DHRS function were also considered in developing the event tree. Furthermore, considering that a dominant contributor to snow risk can be failure of snow removal around the intake and outtake induced by loss of the access routes, this study has investigated effects of electric heaters installed around the intake and outtake as an additional countermeasure. By using the annual excess probabilities and failure probabilities, the event tree was quantified. The result showed that a dominant core damage sequence is failure of the electric heaters and loss of the access routes for snow removal against the combination hazard at daily snowfall depth of 2 m/day, duration time (snow and low temperature) of 1 day.

EDF operates 28 natural draft cooling towers in nuclear power plants. The ageing of their atmospheric cooling tower shells is periodically monitored. Proactive maintenance strategies require ranking the towers according to the risk of failure and the observed damage. The ranking includes all sorts of monitoring data acquired at the plant: foundation settlements, material properties, quantified crack patterns, shell deformation and meteorological data... This combined ranking rely on two pillars: a ranking of the towers based on their shell surface faults observed on-site and another ranking based on their safety margin in terms of structural behavior. The aim of this paper is to present the second pillar of the ranking. The objective of the methodology is to calculate for each tower a failure quantitative risk index based on failure analysis of reinforced concrete. It includes three modules: a mechanical module which is the core module, an ageing module considering the carbonation and corrosion and a decisional module allowing ranking towers. Putting all towers on a ranking scale based on their global risk index allowing decision-maker to optimize their cooling tower maintenance program.

The Reactor Cavity Cooling System (RCCS) is a passive safety concept under consideration for the overall safety strategy of advanced reactors such as the High Temperature Gas-Cooled Reactor (HTGR). One such variant, air-cooled RCCS, uses natural convection to drive the flow of air from outside the reactor building to remove decay heat during normal operation and accident scenarios. The Natural convection Shutdown heat removal Test Facility (NSTF) at Argonne National Laboratory (“Argonne”) is a half-scale model of the primary features of one conceptual air-cooled RCCS design. The facility was constructed to carry out highly instrumented experiments to study the performance of the RCCS concept for reactor decay heat removal that relies on natural convection cooling. Parallel modeling and simulation efforts were performed to support the design, operation, and analysis of the natural convection system. Throughout the testing program, strong influences of ambient conditions were observed in the experimental data when baseline tests were repeated under the same test procedures. Thus, significant analysis efforts were devoted to gaining a better understanding of these influences and the subsequent response of the NSTF to ambient conditions. It was determined that air humidity had negligible impacts on NSTF system performance and therefore did not warrant consideration in the models. However, temperature differences between the building exterior and interior air, along with the outside wind speed, were shown to be dominant factors. Combining the stack and wind effects together, an empirical model was developed based on theoretical considerations and using experimental data to correlate zero-power system flow rates with ambient meteorological conditions. Some coefficients in the model were obtained based on best fitting the experimental data. The predictive capability of the empirical model was demonstrated by applying it to the new set of experimental data. The empirical model was also implemented in the computational models of the NSTF using both RELAP5-3D and STAR-CCM+ codes. Accounting for the effects of ambient conditions, simulations from both codes predicted the natural circulation flow rates very well.

The authors have developed a probabilistic risk assessment method on a forest fire as one of external hazards. A hazard curve by heat effect of a forest fire had been obtained by using a logic tree in our previous study. The main application target of the forest fire probabilistic risk assessment is for sodium-cooled fast reactor systems. Databases for a hazard curve evaluation were based on forest fire records, meteorological and vegetation data of a studied area which is near a typical sodium-cooled fast reactor in Japan. There are two intensity parameters of heat effect of a forest fire, namely, reaction intensity and frontal fireline intensity. The hazard curves of these two intensities obtained in our previous study were referred to as “reference case” where constant breakout frequency throughout a day, equal probability distribution for potential breakout points, and firefighting effect on a forest fire were assumed as a priori . The reference reaction intensity and the fireline intensity became 935 kW/m 2 and 107 kW/m for the annual exceedance frequency of 10 −4 /year, respectively. This paper describes a sensitivity study of the hazard curves on condition parameters where frequency/probability variables in the logic tree were varied within respective fluctuation ranges in order to evaluate quantitative effects on the frequency and/or intensity of the hazard curves. As for the forest fire breakout frequency and propagation probability, important variables are “fluctuation of breakout time”, “probability distribution fluctuation of breakout point”, and “firefighting effect on a probability of forest fire arrival at a nuclear power plant (NPP)”. The intensities increase in daytime due to sunshine, and the breakout probability in daytime is statistically 2.8 times higher than a daily average, and that in nighttime is 1/9 of the average. As a result, the hazard curves of the reaction intensity and the fireline intensity increased around 4% and 14% respectively in intensity direction in comparison with those of the reference case. The “fluctuation of breakout time” only affects the intensities of the hazard curves, but not the frequency. As for the “probability distribution fluctuation of breakout point”, one selected point is given higher probability than the other points. The hazard curves vary around +70% to −40% in frequency direction; each breakout point has different distance to the NPP and the forest fire arrival probability varies with a propagation duration. Namely, the longer duration, the higher probability of the extinguishment by firefighting, accordingly the lower probability of the arrival at the NPP. The “probability distribution fluctuation of breakout point” affects only the frequency of the hazard curves, but not the intensities. “Firefighting effect on a probability of forest fire arrival at an NPP” was conservatively assumed for the sensitivity study in which there is no firefighting action outside the NPP, hence all potential forest fires arrive at the NPP. The hazard curves remarkably increase around 40 to 80 times in frequency direction in comparison with those of the reference case. This is because most of forest fires in Japan are extinguished within one to two hours by fire departments, and the conditional probability of a forest fire arrival at an NPP from a potential breakout point with kilometer range distance was evaluated to be very low (i.e. less than a few percent). The “firefighting effect on a probability of forest fire arrival at an NPP” only affects the frequency of the hazard curves, but not the intensity. This study indicated that the most significant factor in the forest fire hazard curve is whether the firefighting action outside an NPP is expected before the arrival at an NPP.

The Risø Mesoscale PUFF model (RIMPUFF) predicts dispersion of released hazardous material and plays an important role in the emergency response system of many Chinese nuclear power plants (NPPs). The wind field over the calculation domain is a critical input of RIMPUFF, which dominates the performance of RIMPUFF. Due to the complex topography of most Chinese NPP sites, it remains challenging to provide a refined wind field for RIMPUFF prediction. To solve the problem, California Meteorological Model (CALMET) is coupled with RIMPUFF for wind field calculation in this study. Moreover, the computing scope and capability of RIMPUFF are enhanced in order to obtain more accurate prediction for the emergency response and source term inversion in nuclear accidents. Except for that the unlimited amount of grid with higher resolution is supported, a new sampling module is added to RIMPUFF for predicting the concentration of radioactive materials and dose at number-unbound arbitrary location. To verify the CALMET-RIMPUFF method, a wind tunnel experiment that replicates the topography of one Chinese NPP site within 10 km-range, is conducted. In the experiment scenario, the speed and vertical profile of the incoming flow is carefully set according to the annual mean wind speed and wind profile data measured in recent years on the meteorological tower of this NPP. The results demonstrate that the wind field calculated by CALMET is consistent with the topography. With this wind field, the RIMPUFF-predicted concentration distribution matches the measurements well both qualitatively and quantitatively. Moreover, the calculation of U.S. Environmental Protection Agency (EPA) statistical evaluation metrics indicate that the random scatter is within a factor of 1.8 and the FAC2 is nearly 80%. It proves the acceptability of CALMET-RIMPUFF over the complex topography of Chinese NPP sites.

The present study discusses the source term derivation and dose result calculation for a hypothetical accident sequence in the Greek Research Reactor – 1 (GRR-1). A loss-of-coolant accident (LOCA) has been selected as a credible accident sequence. The source term derivation has been based on the GRR-1 confinement performance where the inventory has been computed assuming continuous reactor operation. A core damage fraction of 30% has been considered for the calculations while conservative core release fractions have been employed. The radionuclides released from the reactor core to the confinement atmosphere have been subjected to natural decay, deposition on and resuspension from various internal surfaces before being led to the release pathway. It has been assumed that an emergency shutdown is initiated immediately after the beginning of the accident sequence and the emergency ventilation system is also activated. Subsequently, the source term has been derived comprising of noble gases, iodine and aerosol. The JRODOS computational software for off-site nuclear emergency management has been utilized to estimate the dose results from the LOCA-initiated source term that is released in its entirety from the reactor stack at ambient temperature. The Local Scale Model Chain in conjunction with the DIPCOT atmospheric dispersion model that is embedded in JRODOS have been used with proper parameterization of the calculation settings. Five weather scenarios have been selected as representative of typical meteorological conditions at the reactor site. The scenarios have been assessed with the use of the Weather Research and Forecast model. Total effective, skin, thyroid, lung and inhalation doses downwind of the reactor building and up to a distance of 10 km have been calculated for each weather scenario and are presented. The total effective gamma dose rate at a fixed distance from the reactor building has been assessed. The radiological consequences of the dose results are discussed.

A PWR reactor coolant system is a highly complex physical process: heterogeneous power, flow and temperature distributions are difficult to be accurately measured, because instrumentations are limited in number, thus leading to the relevant safety and protection margins. This situation is in many ways similar to climate and weather models: a complex process that is not possible to sample and measure as finely as wanted. Meteorology and climate sciences have adapted and improved the Data Assimilation techniques in order to improve the accuracy of description and prediction in their fields. EDF R&D is seeking to assess the potential benefits of applying Data Assimilation to a PWR’s RCS (Reactor Coolant System) measurements: is it possible to improve the estimates for parameters of a reactor’s operating set-point, i.e. improving accuracy and reducing uncertainties of measured RCS parameters? In this paper we study the feasibility of enhanced estimation of PWR primary parameters, by using twin experiments for assessing Data Assimilation benefits. We simulated test samples with a 0D-Model, and used these samples in a Monte-Carlo approach to get background terms for Data Assimilation. This successful preliminary study will lead to further assessments with real plant data.

Assuring the protection of people and the environment from unnecessary exposure to radiation is of great concern to nuclear electric power generators and regulators. In order to secure and maintain a license for operation of a nuclear power plant in the United States, the applicant is required to assess postulated scenarios to determine if an accidental chemical release either onsite or offsite will result in a design-basis event. When determining design-basis events, accident categories such as fire, explosion and/or toxic vapor cloud formation are considered. Evaluations must consider whether the postulated accidental chemical release will result in operator impairment or damage to safety related structures, systems or components (SSCs), either of which may prevent safe shutdown of the nuclear plant. A critical step in performing such safety evaluations is selection of an appropriate model which will most closely reflect the behavior of a chemical release in the scenario under consideration. Many chemical dispersion models are available for use in safety evaluations for accidental chemical releases; however, it is imperative that the model selected be appropriate for the postulated release conditions. Model selection should be based on careful evaluation of factors such as release location, meteorological conditions, terrain, chemical inventory and storage conditions, as well as the physical and chemical properties of the chemical under consideration in the postulated release. Failure to select a model suitable for the conditions under which the chemical is released or to appropriately evaluate the physical properties of the chemical and the corresponding limitations of the model may result in underpredicting or overpredicting the impact of a fire, explosion or toxic chemical release. Underpredicting could leave the facility susceptible to damage of safety related SSCs or lead to operator impairment both of which may affect the ability of the plant to safely operate following an accident. While overpredicting the impacts could lead to unnecessary and costly overdesign. This paper will address the considerations that must be evaluated when selecting a chemical dispersion model and will illustrate the importance of model selection in performing nuclear safety evaluations through examples and case studies.

The Chernobyl accident and Fukushima 1 Nuclear Power Plant accident are the most serious accidents in the history of the nuclear technology and industry. A large amount of radioactive materials from nuclear power plant were released, leading to huge damage and long-term effect on the environment as well as the human health neighbor to the plant. Therefore, simulating the transport and transformation of radionuclides in the atmosphere is significant for decision makers to take steps at all level. Now, many different dispersion models are widely applied and used to simulate the transport and transformation of radionuclide such as Gaussian model, Lagrangian model and Eulerian model. Though the Eulerian or Lagrangian models have several advantages, such as high spatial resolution, fully 3D descriptions of the meteorological, the simple Gaussian plume model is still widely chosen because of its higher accuracy and faster calculation. In this study, the atmospheric dispersion of leaked radioactive material during nuclear accident is simulated by using Gaussian plume model. The relative concentration distribution of the radionuclides and the trajectory of the distribution centrode are obtained in taking account of different geographical environments, wind direction, wind velocity, and stability category. These results can provide a favorable evidence for the management of nuclear accident emergency.

The Ultimate Heat Sink (UHS) is a large body of water supply that can be used to cool vital nuclear power plant systems during normal operation and for accident conditions. Due to more stringent environmental and water permit requirements, many new nuclear design proposals have selected the relatively smaller sized mechanical-draft cooling tower with a basin for their UHS. UHS sizing analysis is a critical licensing task for some new generation nuclear power plants Combined Operating License Applications (COLA). In this paper, a potential UHS is sized for a representative new generation nuclear power plant considering worst case design inputs and modeling assumptions. Over 30 years of historical site meteorological data are processed using an automated technique to identify limiting conditions based on resulting worst UHS design parameters, such as the maximum basin evaporative water loss and the highest basin temperature. The impacts of the cooling tower entrance recirculation effect to these design parameters are also investigated. This paper models the transient plant heat loads in detail for various design basis accident conditions. The large-break LOCA heat load is determined to be bounding for the basin evaporative water loss, while a small-break LOCA heat load may result in the highest basin water temperature. This paper also illustrates that the bounding basin water temperature can result when the peak wet bulb temperature is coincident with the peak UHS heat load. The results of this paper are of interest for new generation nuclear power plants as the paper determines impacts of limiting conditions in assessing the design margins for UHS sizing.

The Darlington New Nuclear Project (DNNP) is a proposal by Ontario Power Generation for the site preparation, construction, and operation of up to four new nuclear reactor units for the production of up to 4800 MW of electrical generating capacity. The site was selected by the Ontario Government and is located at the existing Darlington Nuclear site, which is located on the north shore of Lake Ontario and about 70 km east of Toronto, Canada. The regulatory process for the project began in September 2006 and has included the completion of a comprehensive Environmental Assessment (EA), the submission of a Licence to Prepare Site Application, and a three-week public hearing from March 21 to April 8, 2011. This paper provides an overview of various site evaluation and safety studies that were performed in order to demonstrate that the DNNP site meets the Canadian regulatory requirements. The site evaluation studies are also consistent with the principles in the IAEA document NS-R-3, “ Site Evaluation for Nuclear Installations ” and its associated guides. Accordingly, the site evaluation studies considered the following hazards: extreme meteorological events, flooding hazards, seismic hazards, geotechnical hazards, external human-induced events, and potential dispersion of radioactive material with off-site dose consequences. These hazards were assessed in terms of risk to the new nuclear units and ultimately to the public and the environment. Since a reactor technology has not yet been selected for the DNNP, a multi-technology approach was followed for both the EA and site evaluation studies. This involved the use of a bounding Plant Parameters Envelope (PPE), similar to the US-based PPE approach, encompassing the following reactor designs: the US EPR (1580 MWe), the AP1000 (1037 MWe), the ACR-1000 (1085 MWe), and the Enhanced CANDU 6 (740 MWe). Additionally, to assess the impact of protective measures on the local population (e.g., in terms of temporary evacuation), bounding source terms were derived based on the regulatory safety goals for both Small Release Frequency and Large Release Frequency. These generic source terms are expected to bound the releases from any credible accidents, for any reactor designs considered licensable in Canada. In each of the hazard areas, the risk was determined to be acceptably low or could be reduced to an acceptable level through design mitigation. The overall conclusion is that the DNNP site is suitable for the new nuclear units.

The accidents of Chernobyl and Fukushima have shown the necessity to better understand all the mechanisms implied in the scavenging of aerosol particles released to the atmosphere during a nuclear accident. Among all the phenomena involved in the deposition of aerosol particles, we focus here on the aerosol particles scavenging by the raindrops below the clouds, also called washout (as opposed to the rainout, which concerns scavenging inside the clouds). The strategy of IRSN to enhance the knowledge and the modelling of any mechanism involved in the washout of aerosol particles by rain spans from environmental studies, to analytical ones. The semi-analytical approach chosen here is halfway between these two modes of reasoning. A companion paper is also submitted to the conference to present the microphysical approach chosen at IRSN. In order to perform this study, aerosol particles were dispersed in the TOSQAN chamber, which is a large cylindrical enclosure (4.8 m height with 1.5 m internal diameter). The aerosol particles once dispersed, synthetic rains of different kinds (from stratiform to convective rains) can be activated. Finally, the instantaneous spectral scavenging coefficients are determined from the spectral decrease of aerosol particles concentration in the chamber as a function of time. In order to be able to produce synthetic rains representative of any tropospheric events, a special generator has been designed; it is based on a vibro-rotative disk. This generator is able to produce monodispersed rains at the top of the TOSQAN chamber with rainfall rates from 7 to 15 mm/h and drops diameters from 0.5 to 2.5 mm injected at velocities close to their terminal one. During these tests, the spectral aerosol concentration is measured in line with the help of a Welas granulometer. This instrument is based on white light scattering. The results of these experiments highlight the influence of “meteorological” conditions inside the chamber on the washout of the chamber atmosphere, especially when the relative humidity is reaching saturation.