Wwer-1000 Reactor Simulator Download

Abstract:The safety performance of nuclear power plants (NPPs) is a very important factor in evaluating nuclear energy sustainability. Safety analysis of passive and active safety systems have a positive influence on reactor transient mitigation. One of the common transients is primary coolant leg rupture. This study focused on guillotine large break loss of coolant (LB-LOCA) in one of the reactor vessels, in which cold leg rupture occurred, after establishment of a steady-state condition for the VVER-1000. The reactor responses and performance of emergence core cooling systems (ECCSs) were investigated. The main safety margin considered during this simulation was to check the maximum value of the clad surface temperature, and it was then compared with the design licensing limit of 1474 K. The calculations of event progression used the engineering-level RELAP5/SCDAPSIM/MOD3.5 thermal-hydraulic program, which also provide a more detailed treatment of coolant system thermal hydraulics and core behavior. The obtained results show that actuation of ECCSs at their actuation set points provided core cooling by injecting water into the reactor pressure vessel, as expected. The peak cladding temperature did not overpass the licensing limit during this LB-LOCA transient. The primary pressure above the core decreased rapidly from 15.7 MPa to 1 MPa in less than 10 s, then stabilizes up to the end of transient. The fuel temperature decreased from 847 K to 378 K during the first 30 s of the transient time. The coolant leakage reduced from 9945 kg/s to approximately 461 kg/s during the first 190 s in the transient. Overall, the study shows that, within the design of the VVER-1000, safety systems of the have inherent robustness of containing guillotine LB-LOCA.Keywords: nuclear safety; severe accident; LOCA; RELAP5/SCDAPSIM/MOD3.5; emergency core cooling system

The space - energy distribution of the mixed neutron - photon radiation field has been measured over the Reactor Pressure Vessel (RPV) simulator thickness in the WWER-1000 engineering benchmark assembly in the LR-0 experimental reactor (in Nuclear Research Institute e plc) with a scintillation spectrometer. The spectra have been measured before the RPV in one quarter, one half, and three quarters of its thickness and behind the RPV in the energy range 0.5 - 10 MeV. The evaluated integral fluxes above 1 MeV and their ratio are compared with the MCNP and DORT calculation, the measured spectra are presented graphically. The measurements are being performed in the frame of the project REDOS [1], 5th Framwork Programme of the European Community 1998 - 2002.

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The sub-channel thermal-hydraulic analysis of a nuclear reactor is essential for assessing of its safety aspects. In this paper, the VVER-1000 has been selected as an example of the third generation reactors since it meets most of the international safety standards and because it has been taken as a base for designing the VVER-1200 which is belonging to the III+ generation. A steady state mathematical model has been proposed and solved to validate and assure that the hottest channel temperature limits are satisfied. The various temperature distributions, the critical heat flux and the departure from nucleate boiling ratio (DNBR) for the hottest channel were evaluated. Also, a transient state model has also been presented and solved using the finite difference method with the aid of MATLAB algorithm. An exponential loss of flow rate of the reactor core coolant was triggered from the steady state conditions. We assumed that the neutron flux and the generated power were unchanged during the postulated event. The average core coolant flow time constant was treated as a single parameter expressing the rapidity of the event. A value of 250 seconds time constant was assumed for slow transient, whereas 10 seconds was assumed for fast one. The reactor core was assumed to be protected through the reactor control system and mitigated according to the regular emergency operating procedures. The time dependent temperature distributions were calculated for the cladding of the hottest coolant channel. For each value of the temperature, the response time required for reaching unsafe conditions was evaluated, discussed and presented. ff782bc1db