• primary circuit feed and bleed system;
• gas removal system;
• system of fine-tuning injection into the pressurizer;
• systems for simulation of failures to reseat of the pressurizer relief valve and of the steam relief valve to atmosphere (BRU-A);
• system for simulation of main circulation pumps (MCP) overshoot and seizure.

       In order to investigate the accidents, similar to the ultimate design basis accident, a system is installed for organizing double-ended leak from the “hot” and “cold” pipelines, as well as a system for measuring the rate of coolant efflux.
       The primary and the secondary circuits of the test facility are operated under rated pressures of the prototype reactor. The maximal design pressure of the primary circuit of the test facility is 18 MPa, whereas the maximal design pressure of the secondary circuit is 13 Mpa.
       The test facility is equipped with the state of the art research automation system and I&C. The research automation system includes 1000 measurement channels with sampling frequencies 20 Hz and 16 channels with sampling frequencies 100 Hz. The test facility is equipped with the devices for measuring coolant temperature and temperature of simulated fuel elements, pressure and pressure differential, void fraction, coolant flow rate, surface heat loss of the components, voltage, current and power.
       

PSB-VVER testbed	Control room of PSB-VVER

Purpose of the PSB–VVER test facility and areas of experimental studies

       The PSB-VVER test facility is designed for integrated investigation of the phenomenological thermo-hydraulics of the reactor unit during DBA and BDBA. It is used for modeling an initiator and subsequent response of the safety and protection systems, as well as for modeling additionally imposed failures and accident management procedures. The data obtained during the experiment reflects the specifics of the accident under investigation and anti-accident procedures is used for verification of thermo-hydraulic codes. The codes are then used for more accurate analysis of the corresponding steady-state condition. Thus, the PSB VVER test facility plays a key role in the experimental testing and analytical justification of the accident modes and accident prevention and management measures.

Results

       In 1998, the «characterization» experiments were made, with the following purposes:

• adjustment I&C and research automation system;
• determination of the thermo-hydraulic characteristics of the primary circuit and its components;
• determination of the heat loss magnitudes for the primary and secondary circuit components;
• determination of the natural circulation parameters of the one-phase primary coolant flow.

       In 1999, 13 experiments were made to model coolant leak from different locations in the primary circuit. The effects of the size and location of the leak for the accident progression was investigated, along with the effects of failures of different emergency cooling systems.
       In 2000, the PSB VVER test facility was under reconstruction, which included:

1. Installation of special systems allowing maximally realistically modeling the processes occurring in the VVER reactor unit during accidents and transients, as well as investigating the accident management options. Specifically, this included:
   - primary circuit feed and bleed system
   - gas removal system
   - system of fine-tuning injection into the pressurizer
   - systems for simulation of large LOCA
   - hydroaccumulators of second stage, HA-2 of with regard for influence of the passive heat removal system (PHRS) and containment.
   
2. Modernization of the major systems of the test facility: feed­water system, pressurizers, leak imitation modules, active ECCS, pressurizer relief valve, main circulation pumps.
3. Modernization of the research automation system, resulting in the increase of the measurement channel number from 320 to 1000 (sampling frequency to 20 Hz), and addition of 16 channels with the sampling frequency of up to 1 Hz for the primary circuit pressure measurements in case of large LOCA.
4. Modernization of instrumentation and control system (I&C).
   

In 2001, the following experimental studies were conducted at the PSB VVER test facility:

1. Primary-to-secondary leakage (1,4%);
   1. 2,4 % leak from the cold leg;
   2. 11% leak from the cold leg;
   3. 16% leak from the cold leg;
   4. 11% leak from the exit chamber of the reactor model;
   5. SG steam pipe break.

   In 2002, five types of experiments were conducted at the PSB VVER test facility:

   1. Coolant leak from the exit chamber of the reactor model;
      1. Coolant leak from the cold pipeline;
      2. Large double-sided LOCA 2x25% from the hot pipe;
      3. Loss of four MCP;
      4. Loss of one MCP.

      In 2003, two types of experiments were conducted at the PSB VVER test facility:

      • tow-phase natural circulation (step-by-step draining of the primary circuit in different scenarios);
      • loss of feedwater (simulation of an accident scenario for Kozloduy NPP).
        

      In the first part of 2004, the experiment «4,1% leak from the cold pipeline» which is a follow-up experiment performed at the integral test facility LOBI (Italy) was conducted at the PSB-VVER test facility.

             In the second part of 2004 and in 2005 15 experiments were carried out at the PSB-VVER test facility against the TACIS Contract R2.03/97 dedicated to the development and justification of accident management procedures for NPP with VVER-1000. During these experiments, there were simulated main accident conditions and accident management procedures. The main objective of this group of tests was to obtain appropriate experimental data required for validation of evaluating best estimated codes to be used for the development of accident management procedures for VVER-1000 NPP.
             In 2006, an experimental investigation of the influence of new passive safety systems (HA-2 and PHRS [the project “NPP-2006”]) on the thermal state of fuel elements during double-ended break (2?100%) of the “hot” pipeline was carried out at the PSB-VVER test facility. Two experiments (the second is to the recurrence) were carried out. Duration of each experiment was  24 hours. Using the experimental data code RELAP5 was validated as applied to the analyses of the emergency regime at the NPP. The evaluation of this regime was carried out. Analyses of the calculated and experimental data showed the effectiveness of the passive safety systems (HA-2 and PHRS) in accidents during the depressurization of the primary circuit of the NPP-2006.
             The creation of databases of experimental data of the design accidents and beyond design accidents, including procedures of management accidents and operation of new passive safety systems is the main result of the experimental investigations, which were carried out at the PSB-VVER test facility in 1998-2006. Evaluating codes TRAP, CORSAR, BAGIRA, RELAP5, CATHARE, ATHLET, used for safety analyses of the NPP with the VVER, were validated and certified on the basis of this database.
             The experimental investigation of the accident with double break of the main circulation pipeline at the entrance to the reactor was carried out at the PSB-VVER test facility in 2008. Unique experimental data for the validation of codes used for analyses of accidents at the NPP with VVER, including NPP-2006, were obtained.
             Execution of the experimental investigations are planned at PSB-VVER test facility in 2009–2011 at the following areas:

      • regimes with large leak of coolant;
      • regimes with noncondensible gases;
      • standing regimes.
        

      Experiment with 11% leak of coolant through “cold” pipeline

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       The ISB-VVER test facility is a two-loop model of primary NPP circuit with VVER-1000 reactor (V-320 model). The volume and power scaling factor of the testbend is 1:3000, the elevations of the major test facility equipment are consistent with those for the prototype reactor. The test facility is designed to perform experimental study of thermo-hydraulic processes during transients and accidents, including «small» and «medium» LOCA’s break, SG tube breaks, reactor power surges.
       One loop with one steam generator simulates emergency loop of the reactor’s plant. Another test facility loop (triple) with three steam generators simulates three operable loops of RP. Circulation pumps are installed in both loops. The test facility includes also a pressurizer model with electrical heaters, models of three independent emergency core cooling systems (ECCS): high pressure injection system, accumulators and low pressure injection system.
        Reactor model includes the external downcomer, core model, upper plenum and core by-pass. The core model is a bundle of fuel simulators (19 items) with indirect heating.
       To collect, process, display and achieve data obtained from measurement transducers, the test facility is equipped with DAS system possessing 328 measurement channels, 128 of them have the sampling frequencies ~18 Hz, the rest - 1 Hz. The test facility is equipped with the tools for measuring coolant and bundle simulators temperature, pressure, pressure difference, coolant flow rate and electrical power.

Results

       A number of experiments to research of thermo-hydraulic process in VVER-1000 during transients and accidents ordered by Atomenergoproect, Belgatom-Siemens-Framatom consortium and Rossendorf Research Center (FZR), Germany, have been performed. Three standard problems have been realized for validation of national and foreign thermo-hydraulic codes.