[en] Highlights: • Core Damage Frequency and Large Early Release Frequency. • Multi –Unit Risk Metrics. • Aggregation of CDF of NPP through Mean Values. • Aggregation of CDF of NPP as Random Variable. - Abstract: The nuclear generating sites around the world are mostly twin unit and multi-unit sites. The PSA risk metrics Core Damage Frequency (CDF) and Large Early Release Frequency (LERF) currently are based on per reactor reference. The models for level 1 and level 2 PSA have been developed based on single unit. The Fukushima accident has spawned the need to address the issue of site base risk metrics, Site Core Damage Frequency (SCDF) and Site Large Early Release Frequency (SLERF), on the site years rather than reactor years. It is required to develop a holistic framework for risk assessment of a site. In the context of current study, the holistic framework refers to integration of risk from all units, dependencies due to external events and operation time of individual units. There is currently no general consensus on how to arrive at site-specific risk metrics. Some documents provide suggestions for site CDF and site LERF. This paper proposes a new method of aggregation of risk metric from the consideration of operating time of individual units under certain assumptions with a purpose to provide a new conceptual aspect for multi-unit PSA. The result of a case study on hypothetical data shows that site level CDF is not sum of CDF of all units but around 18% higher than unit level CDF. When the CDF is considered to be a random variable then, the new methodology produces site CDF as 50% higher than single unit CDF. These two approaches have been detailed in the paper. For a general data set of CDF for individual units, site CDF would more than individual unit CDF however, it would not be multiples of a single unit value.

[en] This article presents the INES scale and shows how the gravity of a reactor accident is assessed. The INES scale is made of 7 levels: 1 - anomaly, 2 - incident, 3 - severe incident, 4 - accident with local consequences, 5 - accident with extended consequences, 6 - severe accident, and 7 - major accident. Each level is illustrated by a real accident. Only 2 accidents ranked at the highest level (7): Tchernobyl (Ukraine - 1986) and Fukushima Daiichi (Japan - 2011), they were characterized by a complete meltdown of the reactor core. A partial meltdown occurred for the following reactor accidents: Three Mile Island (USA - 1979), Saint Laurent des Eaux (France - 1969 and 1980), Lucens (Switzerland - 1969), Chapelcross (UK - 1967), Windscale (UK - 1957), EBR-1 (US - 1955) and NRX (Canada - 1952). A quick methodology to assess the gravity of an accident is to estimate the quantity of equivalent radioactivity released in the atmosphere, for instance the equivalent of a few thousands Tera decay/second of iodine 131 corresponds to the level 5. The number of people irradiated is also an important parameter as well as the radiation exposure, for instance for an individual dose over 200 mSv, involving 100 people or more corresponds to a level 5 event, 10 people or more to a level 4 event and less than 10 people to a level 3 event. (A.C.)

[en] Goals: – Develop the fundamental scientific basis to understand, predict, and measure changes in materials and systems, structures and components (SSCs) as they age in environments; – Apply this knowledge to develop and demonstrate methods and technologies that support safe and economical long-term operation of existing reactors; – Research new technologies that enhance plant performance, economics, and safety.

[en] The multiple safety systems including high pressure safety injection (HPSI), low pressure safety injection (LPSI), safety injection tank (SIT), and etc. have been designed to protect the core under the accidents in NPPs. If the decay heat from reactor is not removed due to the failure of safety systems under the accidents, the core can be melted. Therefore, the monitoring of reactor internal phenomenon is very important to prevent core meltdown. The deep learning model can be simulated for reactor internal phenomena without knowledge of physical. Deep learning is one of the most active research fields in recent years because computer’s performance has been improved. It has been widely used not only in science but also in various industries such as medicine, advertising, and finance. Deep learning is a technology that applies information processing methods of human brain to machines. The basic structure of Deep learning has a multilayer perceptron (MLP) structure consisting of three or more hidden layers. The MLP is a neural network composed of several nodes and layers. The location of data is as follows: The thermal distribution of core cell, core baffle, bypass, support barrel, down comer, and vessel cylinder. The data was obtained by using MELCOR which is the severe accident analysis code. The operators can maintain the integrity of the reactor when an unexpected severe accident is occurred in the nuclear power plant because the developed model can predict the reactor internal phenomena by the thermal distribution of vessel cylinder.

[en] The phenomenon and mechanism of FCI (Fuel Coolant Interaction) has been widely studied around the world in the past few decades. A series of experiments were performed and several FCI models were developed on the basis of these experiments. However, there are still large uncertainties in the models of FCI and limitations to predict FCI process, especially the reactor scale process. To study the mechanism of FCI, a new FCI experimental facility was designed and further experiments were performed by Shanghai Jiao Tong University, China. The photography of FCI process were obtained by 2 high-speed cameras recording from 2 different directions vertical to each other. Water level changed can also be got from images of FCI process. Pressure peak produced by intense interaction and temperature of coolant are recorded. To discuss the influence of different factors for FCI, numbers of variables are considered in these experiments, including jet material, melt temperature, coolant type, coolant subcooled temperature, release heights, break size and interaction pool size. This paper focuses on the FCI responses for different melt temperatures. Tin, with the melt point of 231.9 C. degrees, was chosen as the melt material since it is possible to acquire a large temperature range of melt from 400 to 1300 C. degrees. The phenomenon responses of FCI process and the particle size responses for different melt temperature are discussed in the paper. (authors)

No abstract available

No abstract available

[en] The forecast of atmospheric diffusion described here allows statements to be made about the size and location of the area in which maximum radiation burdens have to be expected from emission after an accident. If the source strength can be estimated from the course of the accident, the value of the maximum burden to be expected can be predicted in addition. Based on these statements, first effective actions can be taken to protect the population in case of serious hypothetical accidents. (orig.)

[en] This report contains the safety analyses of the KALIMER-600 conceptual design which KAERI has been developing under the Long-term Nuclear R and D Program. The analyses have been performed reflecting the design developments during the second year of the 4th design phase in the program. The specific presentations are the key design features with the safety principles for achieving the safety objectives, the event categorization and safety criteria, and results on the safety analyses for the DBAs and ATWS events, the containment performance, and the channel blockages. The safety analyses for both the DBAs and ATWS events have been performed using SSC-K version 1.3., and the results have shown the fulfillment of the safety criteria for DBAs with conservative assumptions. The safety margins as well as the inherent safety also have been confirmed for the ATWS events. For the containment performance analysis, ORIGEN-2.1 and CONTAIN-LMR have been used. In results, the structural integrity has been acceptable and the evaluated exposure dose rate has been complied with 10 CFR 100 and PAG limits. The analysis results for flow blockages of 6-subchannels, 24-subchannels, and 54- subchannels with the MATRA-LMR-FB code, have assured the integrity of subassemblies

[en] In the context of the simulation of the Severe Accidents (SA) in Light Water Reactors (LWR), we are interested on the in-core corium pool propagation transient in order to evaluate the corium relocation in the vessel lower head. The goal is to characterize the corium and debris flows from the core to accurately evaluate the corium pool propagation transient in the lower head and so the associated risk of vessel failure. In the case of LWR with heavy reflector, to evaluate the corium relocation into the lower head, we have to study the risk associated with focusing effect and the possibility to stabilize laterally the corium in core with a flooded down-comer. It is necessary to characterize the core degradation and the stratification of the corium pool that is formed in core. We assume that the core degradation until the corium pool formation and the corium pool propagation could be modeled separately. In this document, we present a simplified geometrical model (0D model) for the in-core corium propagation transient. A degraded core with a formed corium pool is used as an initial state. This state can be obtained from a simulation computed with an integral code. This model does not use a grid for the core as integral codes do. Geometrical shapes and 0D models are associated with the corium pool and the other components of the degraded core (debris, heavy reflector, core plate... ). During the transient, these shapes evolve taking into account the thermal and stratification behavior of the corium pool and the melting of the core surrounding components. Some results corresponding to the corium pool propagation in core transients obtained with this model on a LWR with a heavy reflector are given and compared to the grid approach of the integral codes MAAP4. (authors)