The paper presents new, transient simulation results of condensing steam in the event of emergency core cooling where water is injected into the cold leg during a postulated loss-of-coolant-accident (LOCA) of a pressurized water reactor (PWR), a phenomenon known as Pressurized Thermal Shock (PTS). The model is first validated for the Lim et al. (1984) experiment involving a smooth to wavy turbu...

The flashing phenomenon is relevant to nuclear safety analysis, for example by a loss of coolant accident and safety release scenarios. It has been studied intensively by means of experiments and simulations with system codes, but computational fluid dynamics (CFD) simulation is still at the embryonic stage. Rapid increasing computer speed makes it possible to apply the CFD technology in such c...

Operators of nuclear power plants may not be equipped with sufficient information during a loss-of-coolant accident (LOCA), which can be fatal, or they may not have sufficient time to analyze the information they do have, even if this information is adequate. It is not easy to predict theprogressionof LOCAs innuclear power plants. Therefore, accurate informationon theLOCAbreakpositionandsize sh...

The “best-estimate plus uncertainty” (BEPU) methodology is the term used in nuclear engineering community when dealing with uncertainty quantification issues realistic numerical simulation models. One of most critical hypothesis these studies choice probability distributions uncertain input variables which are propagated through model. Bringing stringent justifications to BEPU approach, especia...

The Helium Cooled Pebble Bed (HCPB) breeding blanket, being developed by the Karlsruhe Institute of Technology (KIT) and its partners is one two driver blanket candidates to be selected for European demonstration fusion power plant (EU DEMO). in-box Loss Coolant Accident (LOCA) a postulated initiating event (BB) that must accounted within design basis. In this paper, BB cap region analyzed abil...

Significance: Sep 30, 2002 Identified By: NRC Item Type: FIN Finding UNEXPECTED CHANGES IN UNIT 2 OPERATING PARAMETERS AND DRYER FAILURE DUE TO FLOW INDUCED VIBRATION. The failure to consider the impact of new flow induced vibration failure mechanisms on the Unit 2 steam dryer as part of the extended power uprate analysis resulted in unexpected and unpredictable changes in reactor power, reacto...

The safety verification of nuclear systems can be done by analyzing the outputs of Best-Estimate Thermal-Hydraulic (BE-TH) codes, which allow predicting the system response under safe and accidental conditions with greater realism as compared to conservative TH codes. In this case, it is necessary to quantify and control the uncertainties in the analysis, which affect the estimated safety margi...

The EU-DEMO must demonstrate the possibility of generating electricity through nuclear fusion reactions. Moreover, it denote necessary technologies to control a powerful plasma with adequate availability and meet safety requirements for plant licensing. However, extensive radioactive materials inventory, complexity plant, presence massive energy sources require rigorous approach fully realize p...

Zirconium-based claddings with an outer chromium coating resistant to corrosion are studied and developed as evolutionary Enhanced Accident Tolerant Fuel (E-ATF) concept for light water reactors. However, in hypothetical LOss-of-Coolant-Accident (LOCA) conditions, following clad ballooning burst, the does not allow protect inner surface of cladding from High Temperature (HT) steam oxidation ass...

During a Severe Accident (SA) occurring in nuclear power plant, many Fission Products (FP) are released from the degraded fuel and transported Reactor Coolant System (RCS). Depending on their volatility, FP can be either deposited surface of (RCS) or into containment building where they may environment case early failure. This was for Fukushima Daiichi (FD) accident with important releases whic...