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[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] 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 Nuclear Engineering Department at Israel Electric Company has been engaged for a number of years in a joint research agreement with the Technion Nuclear Reactor Research Group on various thermal-hydraulic aspects of reactor design and safety. Besides developing their own analytical models, the researchers rely heavily on the RELAPS computer code in their analyses. The RELAPS series are general purpose, thermal-hydraulic system codes, used to simulate system response (such as the RCS) to transients and accidents. They are based on solving the equations of conservation of mass, energy and momentum within the system being modeled, where the model is a series of control volumes connected by junctions. The equations are solved simultaneously in each volume and junction using a finite difference numerical scheme. As an example, a recent report refers to containment response to a large LOCA in an AP600-like advanced rector. This work has been performed using RELAPS/Mod2. Accidents like LOCA represent design base events necessary to verify the adequacy of the emergency core cooling system, the passive containment system and other safety systems. The validation of simulation results is therefore important to the IEC staff responsible for monitoring the research. (author); 3 refs
[en] Full text: One of the major design basis accident is loss of coolant accident (LOCA). This accident is characterized by high sheath temperature and rapid depressurization of coolant. One of the ECCS acceptance criteria for this type of accident is for all LOCA events the release of radioactive material from the fuel in the reactor shall be limited such that the calculated radiological consequences at exclusion zone boundary are within acceptable limits for accidents as specified by AERB. The assessment of above criteria requires the estimation of fuel failure. Two phenomena which cause fuel failure during such condition are oxygen embrittlement and clad ballooning. Oxygen embrittlement is assessed by the amount of oxidation. Failure due to ballooning is analysed by calculating stress and strain on the clad and comparing with ultimate burst stress. Failure data generated by analysing these two phenomena are used in assessing damage on fuel. This paper discusses the criteria of fuel failure followed by various regulating agencies or countries along with Indian criteria. This paper also discusses the damage phenomena along with detailed method of assessment of fuel failure during LOCA
[en] Fuel cans of defined failure size and type were operated in the HSD loop of the FR2 reactor. The experiments were carried out with the purpose 1) of measuring the emission of fission products and fuel particles from failed fuel rods, 2) of determining the behaviour of failed fuel elements in constant-load and load-following operation, and 3) of determining the type of failure from the measuring signals. (orig./AK)
[de]Im Huellenschaden-Dampfkreislauf (HSD-Loop) des FR2 wurden Brennstaebe definierter Defektgroesse und -art betrieben. Ziel der Versuche war: 1) Messung der Freisetzung von Spaltprodukten und Brennstoffpartikeln aus beschaedigten Brennstaeben, 2) Bestimmung des Verhaltens defekter Brennelemente bei Konstantlast- und Lastfolgebetrieb, 3) Erkennen verschiedener Defektarten aus den Messsignalen. (orig./RW)
[en] Vibrations of reactor internals are studied with respect to noise diagnostics and the building of a diagnostic expert system. The basic theory is presented along with the first numerical results on the behaviour of the support cylinder represented by a cylindrical shell. The results show that level changes in the system pressure vessel-support cylinder have marked effect on the frequency spectrum of the support cylinder vibrations, which might be used to indicate the water level in the core during small LOCA and ATWS accidents and might form part of a diagnostic expert system. (author)
[en] The objective of this paper is to assess proposed transuranic waste accident analysis guidance and recent software improvements in a Windows-OS version of MACCS2 that allows the inputting of parameter uncertainty. With this guidance and code capability, there is the potential to perform a quantitative uncertainty assessment of unmitigated accident releases with respect to the 25 rem Evaluation Guideline (EG) of DOE-STD-3009-94 CN3 (STD-3009). Historically, the classification of safety systems in a U.S. Department of Energy (DOE) nuclear facility's safety basis has involved how subject matter experts qualitatively view uncertainty in the STD-3009 Appendix A accident analysis methodology. Specifically, whether consequence uncertainty could be larger than previously evaluated so the site-specific accident consequences may challenge the EG. This paper assesses whether a potential uncertainty capability for MACCS2 could provide a stronger technical basis as to when the consequences from a design basis accident (DBA) truly challenges the 25 rem EG