In internal combustion engines, a significant portion of the total fuel energy is consumed to overcome the mechanical friction between the cylinder liner and the piston rings. The engine work loss through friction gradually reduces during the engine break-in period, as the result of liner surface topography changes caused by wear. This work is the first step toward the development of a physics-based liner wear model to predict the evolution of liner roughness and ring pack lubrication during the break-in period. Two major mechanisms are involved in the wear model: plastic deformation and asperity fatigue. The two mechanisms are simulated through a set of submodels, including elastoplastic asperity contact, crack initiation, and crack propagation within the contact stress field. Compared to experimental measurements, the calculated friction evolution of different liner surface finishes during break-in exhibits the same trend and a comparable magnitude. Moreover, the simulation results indicate that the liner wear rate or duration of break-in depends greatly on the roughness, which may provide guidance for surface roughness design and manufacturing processes.

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Since the nuclear energy industry in China has developed rapidly, the demand for spent fuel reprocessing is becoming more and more distinct. In a reprocessing facility, power supply system (PSS) is important for operation-stability and safety-guarantee of the whole facility. Therefore, the reliability of the PSS deserves in-depth investigation and analysis. Fault tree (FT) methodology is the most accustomed methodology for system-reliability analysis and it has been standardized and applied widely. GO methodology is another effective methodology for system-reliability analysis but it has not been applied widely. In this work, GO methodology was applied to analyze the reliability of the PSS in a typical reprocessing facility. First, for modeling expediently, based on the fact that tie breakers are set in the system, tie breaker operator was defined by myself. Then, GO methodology modeling and quantitative analysis were performed, minimal cut sets (MCSs) and average unavailability of the system were obtained. Finally, parallel analysis between GO methodology and fault tree methodology was also performed. The results from the two methodologies are completely coincident. The results of this work show the following two points: 1. Setup of tie breakers in the PSS of the reprocessing facility is rational and necessary; 2. For reliability analysis of the PSS of the reprocessing facility, parallel with fault tree methodology, GO methodology has two distinct advantages: Its modeling is much easier and its chart is much more succinct.

Fukushima accident shows again that the probability of reactor accident exists even though it is extremely small. In case of emergency in nuclear power plant, emergency condition of the reactor plays an important role in decision making. During emergency response, especially early stage of severe accident with large release of radioactive nuclides, decision making for protection actions should be based on emergency condition in NPPs. If emergency condition could be prognosed, more time could be bought for decision making and emergency response. In this paper, method for prognosis of large break loss of coolant (LBLOCA) initiated severe accident progression was established based on transient analysis for M310 reactor. Mass and energy conservation equations are the basis of the method. Separated flow model is used for prognosis of emergency condition for large break loss of coolant accident initiated severe accident. These conservation equations are solved approximately in order to significantly increase calculation speed. The active core is divided into 4 radial rings and 10 axial levels, which means there are 40 cells. Heat transfer calculation in the core is done using four experimental correlations. Based on the method established in this paper, a code using for prognosis of LBLOCA initiated severe accident emergency condition was developed. Research on method for prognosis of other severe accidents are being conducted.

The safety characteristics and potential hazards of reprocessing facilities are different from those of nuclear power plants (NPPs). Emergency action level (EAL) development is an important aspect of emergency preparedness, and EAL is an important basis for emergency response of reprocessing facilities. EAL quantitative research can enhance its operability. At present, the domestic and foreign literature, generally only gave the principle method for EAL development. There are no operational guidance documents for specific EAL quantification in reprocessing facilities. According to the features of the functions and configurations of the reprocessing facilities, two additional categories EAL, E category for the spent fuel pool accident and W category for the high level liquid waste tank accident, have been added. Meanwhile, four categories-S, F, A, H are retained with similar implication as NPPs. On the basis of existing principles and practical experience, specially the reference from DOE G 151.1-1A, EAL quantification was developed according to the characteristics and symptoms related to the safety of reprocessing facilities. EAL quantification for several accidents was developed, and it was with good maneuverability. The study for EAL quantification in reprocessing facilities shows that, quantitative, indicative, practical EALs can make emergency response more accurate and efficient, and ongoing research is strongly requisite.