[en] The results of investigation into the thermal hydraulic conditions in a ring channel when interrupting the cooling water feed are presented. Experiments have been carried out in a channel with an electrical heated inner tube of 14K18N9T steel in the range of regime parameters:Δtsub(n)=25; 100 deg C; P=3-11.9 MPa; rhow=700-2100 kg/m²s; q=250-850 kW/m²; the temperature of the beginning of cooling tsub(p)=300-700 deg C. Analyzed are in detail Main stages of development of emergency conditions: channel dewatering and heating up, supply of a channel with water and heat transfer in the regime of film boiling, damping of heated-up wall are analyzed in detail. Formulae for the calculation of these processes and experimental data are given

[en] MCC rupture on outlet, inlet the reactor (with coolant double end leakage); MCC rupture in the bottom part of a loop seal (with coolant double end leakage); Rupture of the connecting pipeline of HA (JNG10-40) of the ECCS (D = 300 mm); Loss of an alternating current sources, failure of safety active systems for more than 24 hours, failure of all diesel-generators; emergency supply from accumulators; Water injection from a fuel storage pool begins after emptying of tanks GE-2 (JNG50-80) with the total mass flow rate of 3.2 kg/s. The RELAP5/ANGAR code was used. The sequence of events and the work of the systems at a guillotine rupture of the MCC is described. The computing analysis of the beyond design accident which includes MCC rupture with loss of all sources of an alternating current, including diesel engines-generators, demonstrated the following: (i) the intervention of the safety systems of the NVNPP-2 project meets all Russian and international requirements to localizing functions by NVNPP-2 containment at beyond design accidents with leaks from the reactor facility; (ii) water injection from ST after the termination of work; (iii) GE-2 provides reliable core cooling for the beyond design basis accident scenarios during 74-88 hours in dependence on the leak size and location. (P.A.)

[en] Full text of publication follows: Paper represents some results of computational research executed within the framework of verification of the KORSAR thermal hydraulic code. This code was designed in the NITI by A.P. Aleksandrov (Russia). The general purpose of the work was development of a nodding scheme of the PSB-VVER integral facility, scheme testing and computational modelling of the experiment 'The PSB-VVER Natural Circulation Test With Stepwise Reduction of the Primary Inventory'. The NC test has been performed within the framework of the OECD PSB-VVER Project (task no. 3). This Project is focused upon the provision of experimental data for codes assessment with regard to VVER analysis. Paper presents a nodding scheme of the PSB-VVER facility and results of pre- and post-test calculations of the specified experiment, obtained with the KORSAR code. The experiment data and the KORSAR pre-test calculation results are in good agreement. A post-test calculation of the experiment with KORSAR code has been performed in order to assess the code capability to simulate the phenomena relevant to the test. The code showed a reasonable prediction of the phenomena measured in the experiment. (authors)

[en] For the first time on the basis of experiments as were performed at the two-loop integral thermal testing unit ISB-WWER modulating the first circuit of the reactor facility with WWER-1000 five standard problems were studied in 1993 - 2001 for the investigation into accident regimes of loss of coolant in the first circuit. Verification of native and frontier calculating thermal hydraulic codes: TECH', KORSAR, ATHLET, CATHARE, RELAP was checked. Results of verification calculations were sufficiently consistent with experimental data. Most of processes and events that were a possibility in emergency conditions with WWER-1000 at low and middle flow of coolant were reproduced in the experiments. Analysis of the results demonstrates that with use of these calculating codes accident processes are simulated

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[en] The computerized simulations of the accident and abnormal conditions at nuclear power plants are very important tools for its safety assessment. This presentation is a survey of computer codes used in the USSR for safety analysis of WWER type reactors. It also presents a review of the experimental facilities for safety experiments in the field. The plans in safety analysis programs for today and the future are presented

No abstract available

[en] The possibility of the increasing of the steam equalization ability of the submerged perforated sheet with the help of the submerged perforated sheet with non-uniform perforation was studied. The PGV test facility was constructed in JSC “EREC” for the experimental investigation of this problem. Several series of experiments were performed in PGV test facility in 2010-2013. The constructive shortcomings of the test facility have been identified and corrected in the course of this work. After removing of all constructive shortcomings in the test facility experimental researches were performed with using submerged perforated sheets with uniform and non-uniform perforations. Experimental data obtained in the experiments are reported. The hydraulic resistance of the submerged perforated sheet with uniform perforation for different flow rates of the steam and two-phase correction factor were determined on the basis of the experimental data analysis. The steam equalization ability of the SPS was evaluated on the basis of the experimental data analysis by different methods

[en] The report outlines thermophysical calculation models and programs for computers used for safety analysis of VVER reactor plants in the Soviet Union and describes laboratory equipment used and experimental studies made for handling the thermophysical processes. It also summarizes the present status and future prospects of theoretical and experimental investigations on safety issues in nuclear power plants equipped with a VVER reactor. The accidental instances that are taken into consideration in safety studies for nuclear power plants include coolant loss in the primary system, disturbance of normal operation conditions, reactivity-related accident, transient process in failure of emergency shutdown required, and melting of fuel. Two types of calculation codes are used: those for detailed description of each element and physical process in a plant and those for complete description of the entire plant. The second part of the report lists various subjects of thermophysical experiments and studies related to safety issues involving the design and construction of VVER reactor plants and summarizes tools used for comprehensive experimental investigations of accident instances. The third part discusses some practical safety problems and future directions of calculation programs and experimental investigations. Major studies currently under way or under plan are also listed. (Nogami, K.)