[en] Nuclear safety analysis is always the fundamental item for nuclear power development. Within the safety analysis, the most important aspect is the thermal hydraulic analysis of operation conditions and accidents (including design-basis accident (DBA) and beyond-design-basis accident (BDBA)). The loss-of-coolant accident (LOCA) is the most concerned DBA for nuclear thermal hydraulic safety analysis. After the development of around half a century, the concerned issue in thermal hydraulic safety analysis has switched in the recent decade from large-break LOCA (LBLOCA) and small-break LOCA (SBLOCA) to intermediate-break LOCA (IBLOCA) scenario. IBLOCA scenario has been the focus of experimental studies in Rig-of-Safety assessment/large scale test facility (ROSA/LSTF), advanced thermal-hydraulic test loop for accident simulation (ATLAS), and most recently the PKL (German abbreviation for #Quotation Mark#Primärkreislauf#Quotation Mark#, #Quotation Mark#primary loop#Quotation Mark# in English) facility. Based on the IBLOCA test scenario of the LSTF facility, it was found that the phenomena of core heat up and peak cladding temperatures (PCTs) are very sensitive to the break size and the operation of safety injections. Unfortunately, most of the system thermal-hydraulic (STH) codes could not reproduce these processes in different IBLOCA scenarios. In order to confirm and to solve this problem, the PKL I2.2 IBLOCA benchmark was resorted to, as a counterpart test similar to the IBLOCA scenario in LSTF. Some typical STH codes should be evaluated by the test data of PKL I2.2 IBLOCA benchmark. ATHLET (short for Analyses of THermal-hydraulics for LEaks and Transients) was used for the simulation of PKL I2.2 IBLOCA benchmark and the model assessment/modification of ATHLET was focused on. The main steps for ATHLET input deck preparation, the simulation results - including steady state and transient - were described in details in this work. In order to make the results convincing, the related state-of-art methodology was used: a nodalization qualification method was described and applied before the benchmark scenario simulation; a well-known method – Fast Fourier Transform Based Method (FFTBM) was introduced for the evaluation of the effectiveness of ATHLET on PKL I2.2 IBLOCA simulation. Based on the analysis of the transient results and the FFTBM results, one may come to the conclusion that most of the variables in PKL I2.2 IBLOCA benchmark were predicted very well by ATHLET, which confirmed its effectiveness on IBLOCA simulation. But unfortunately the PCT was not reproduced in the simulation. According to the average amplitude (AA) values from the FFTBM method and the results of the sensitivity study which - based on new developed methodology (two-layer FFTBM – MSM coupling sensitivity study method, MSM here means Morris screening method), this failure is most likely related to the break mass flow modelling. Consequently, a new two-phase model (non-equilibrium and non-homogeneous two phase critical flow model (NNTPCM)) for the analysis of two-phase critical discharge was developed as a potential critical flow model (CFM) in ATHLET. The model allows thermodynamic non-equilibrium and hydrodynamic non-homogeneity between the liquid and vapor phases. It comes out as the solution of the six conservation equations of mass, momentum and energy for separated phases (the present ATHLET CFM - Critical Discharge Rate 1 Dimension model (CDR1D) - is a 4-equation model). The model is able to simulate several flow regimes, from subcooled to annular flows. Closure was achieved by a set of constitutive relations chosen from an extensive literature review. Two kinds of choking criteria (determinant and pressure gradient) are discussed. For the determinant criterion, a compatibility condition should be considered for the system of ordinary differential equations (ODEs) describing the two-phase flow to have a solution at choking point. In order to confirm the two criteria, they were numerically investigated for long pipe, short pipe and orifice discharge tests. The results obtained by using the two different criteria are consistent as long as the pressure gradient threshold value remains large enough. Simultaneously, according to the results, this value is larger for the case of orifice and short pipe discharges (compared with a long pipe discharge). The model was validated by the experimental data from Al-Sahan tests (long pipe discharge), Celata test (nozzle discharge), Dobran test (long pipe discharge), Sozzi–Wutherland tests (short pipe discharge) and Henry tests (which comprises 9 subcooled and 10 saturated upstream conditions). The comparison of results showed excellent agreement with measured critical mass fluxes (but also with pressure profiles in Al-Sahan and Henry tests). The calculation results were the best ones, compared with other models from literature. A special attention was paid to the understanding of the choking process by analyzing the evolution of the main constitutive parameters, aspect seldom considered in previous studies. According to this analysis of the constitutive parameters, some interesting conclusions are extracted: the interfacial area becomes maximum at the transition point from bubble to slug/churn flow; the virtual mass force becomes important and sometimes decisive for choked flow; for long pipe, the thermodynamic non-equilibrium plays negligible role because of the good heat transfer between the two phases but the hydrodynamic non-homogeneity has to be taken into account since the velocity difference becomes very large at the choked point; on the contrary, the hydrodynamic non-homogeneity may be neglected but the thermodynamic non-equilibrium considered for short pipe or orifice because of the superheated liquid and small velocity difference. As a potential substitute for CDR1D model in ATHLET, the methodology for the plugin in ATHLET has been described in details. To validate the effectiveness of the model and to verify its ability in replacing the CDR1D model, several Marviken full scale critical flow tests and PKL I2.2 IBLOCA benchmark were chosen for the model validation; the model was compared with both test data and also with the results obtained by the ATHLET built-in CDR1D model. The results showed that NNTPCM could get better or at least comparable results than CDR1D model for the simulation of thermal-hydraulic scenarios in PWRs.

[en] The modern technology of construction of the spent fuel assembly storage system with high storage density in nuclear power plants is introduced. With no change in the existing fuel storage pool, the use of the high density storage rack system will double the storage capacity for the spent fuel. High economic benefit will be achieved by spreading the use of this modern technology in nuclear power plants. The technology background, the distinguishing structural features, and the requirements and contents for design and analysis of the high density storage rack system are discussed

[en] The dynamic response time-history analysis technology for nonlinear mechanical systems with multi-degrees of freedom is introduced. Mechanical systems with almost any non-linear (and linear, of course) properties can be treated by this advanced technology. Through comparison with conventional methods for dynamic analysis, such as response spectrum analysis, etc, the significant advantages and developing status of this technology are shown. The results from this technology for a typical equipment seismic response analysis in nuclear power plant is also presented to show its suitability for engineering application

[en] In seismic dynamic response analysis of structures and equipment, time-history analysis is now widely used. The 3-D seismic acceleration time-histories or 3-D seismic displacement time-histories are required in the 3-D seismic dynamic response analysis as the seismic excitation input data. Because of the lack of actual acceleration time-histories for the field where the structures or equipment are installed, the general practice is to use the synthetic seismic acceleration time-histories, which are derived from the design seismic response spectra of the field, as the seismic excitation input data. However, from one specified design response spectrum indefinite solutions of acceleration time-histories can be derived depending on the values of the input parameters. Not all the derived synthetic time-histories can be used as seismic excitation input data. Only those which meet the acceptance criteria can be used. The factors (input parameters), which will affect the time-history solution from a specified seismic response spectrum, and the acceptance criteria are discussed

[en] A recent simulation method called reverse Monte Carlo (RMC) applicable without interaction potential is used to study the aqueous electrolyte system LiCl-6H₂O. Artifacts are appeared in some pair distribution functions particularly a small pick near the first coordination of g_OO(r) and also near the g_OCl(r) one. One try to remedy for that artifact with introducing a specified potential for the oxygen atoms and a Coulomb potential for the rest of the atomic species. An improvement in the first coordination of this function is noticed suggesting a useful test of an interaction potential model for classical methods as Monte Carlo (MC) and molecular dynamic (MD)

[en] Scientific and reasonable operator management is the basis of nuclear security. It was paid more attention after the three-mile island accident. The prediction of operators' basic behavior parameters is the premise and foundation of scientific and reasonable operator management. Grey theory happens to solve the dilemma encountered in prediction and decision-making of operator behavior in operator management of NPP. The procedure is divided into two steps: according to the history record of operators' behavior parameter, a differential equation model using grey theory is set up to predict the future behavior of operators and use grey theory to make decision for operator management. The calculation result is helpful for operator management and also useful for operators to find their shortcoming. Grey theory using in the study provides a new idea and method for future operator management in NPP

[en] After Fukushima nuclear accident, much attention has been paid to the safety of the spent fuel pool (SFP). AP1000 spent fuel pool cooling system (SPS) is not a safety related system. It is not designed for design basis accident. But the safety of SFP in exterior disaster and severe accident is very important. The extensive damage mitigation guidelines (EDMG) of NEI to AP1000 SFP accident mitigation was studied, including internal strategy and external strategy. The results are useful for design AP1000 nuclear power plant and the mitigation when SFP exterior disaster happens. (author)

[en] Scientific and reasonable operator management is the basis of nuclear safety. It is paid more attention after the three-mile island accident. The prediction of operators' basic behavior parameters is the premise and foundation of scientific and reasonable operator management. Grey theory happened to solve the dilemma encountered in prediction and decision-making of operator behavior in operator management of nuclear power plant. The procedure was divided into two steps: 1) According to the historical record of operators' behavior parameters, a differential equation model using grey theory was set up to predict the future behavior of operators; 2) operator management decision-making was made based on grey theory. The calculation result is not only helpful for operator management but also useful for operators to find their shortcomings. Grey theory used in the study provides a new idea and method for future operator management in nuclear power plant. (author)

[en] A new measurement of the photodissociation of CH₃I at 193 nm is reported in which we use a combination of vacuum ultraviolet photoionization and velocity map ion imaging. The iodine photofragments are probed by single-photon ionization at photon energies above and below the photoionization threshold of I(²P_3/2). The relative I(²P_3/2) and I^*(²P_1/2) photoionization cross sections are determined at these wavelengths by using the known branching fractions for the photodissociation at 266 nm. Velocity map ion images indicate that the branching fraction for I(²P_3/2) atoms is non-zero, and yield a value of 0.07 ± 0.01. Interestingly, the translational energy distribution extracted from the image shows that the translational energy of the I(²P_3/2) fragments is significantly smaller than that of the I^*(²P_1/2) atoms. This observation indicates the internal rotational/vibrational energy of the CH₃ co-fragment is very high in the I(²P_3/2) channel. The results can be interpreted in a manner consistent with the previous measurements, and provide a more complete picture of the dissociation dynamics of this prototypical molecule

[en] The Failure Assessment Diagram (FAD) procedure is introduced. The method, using FAD to distinguish different failure models and to analyse the stable growth of crack, is explained. The technical background of flaw evaluation procedures for nuclear piping in ASME Code is analysed. The special treatment in these procedures are discussed