Validation of the RELAP5 code for the simulation of the Siphon Break effect in pool type research reactors (original) (raw)
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RELAP5 Modeling of a Siphon Break Effect on the Brazilian Multipurpose Reactor
Brazilian Journal of Radiation Sciences
This work presents the thermal-hydraulic simulation of the Brazilian Multipurpose Reactor (RMB) using the RELAP5/Mod3 code. The RMB will provide Brazil with a fundamental infrastructure for the national development on activities of the nuclear sector in the areas of social, strategic, industrial applications and scientific and technological development. A RELAP5/Mod3 code model was developed for thermal-hydraulic simulation of the RMB to analyze the phenomenology of the Siphon Breakers device (four flap valves in the cold leg and one open tube for the atmosphere in the hot leg) during a Loss of Coolant Accident (LOCA) at different points in the primary circuit. The Siphon Breaker device is an important passive safety system for research reactors in order to guarantee the water level in the core under accidental conditions. Different simulations were carried out at different location in the Core Cooling System (CCS) of the RMB, for example: LOCA before the CCS pumps with and without pump trip and LOCA after the CCS pumps and the heat exchanger. In all RELAP5/Mod3 code simulations, the Siphon Breaker device's performance after a LOCA was effective to allow enough air to enter the outlet pipe of the CCS in order to break the siphon effect and preventing the pool level from reaching the riser (chimney) and the RMB core discovering. In all cases, the reactor pool level stabilized at about 5.5 m after the end of the LOCA simulation and the fuel elements were kept underwater and cooled.
Numerical Simulation of Siphon Breaker of an Open-Pool Type Research Reactor
2017
A numerical investigation of the siphon breaker of an open-pool type nuclear research reactor was performed by Computational Fluid Dynamics. The computational model was assessed by solving a siphon break line design, for which experimental and numerical data are available. The multiphase problem was solved with Volume of Fluid Method and k-epsilon for turbulence modeling. Numerical results were in very good agreement with experimental data. The siphon breaker occurrence was verified and the undershooting height, measured from the pool level to the siphon break line end as well as the liquid mass flow rate inside the main pipe were well captured. The implemented model showed to be reliable for assessing this kind of passive safety systems.
Analysis of Loss of Flow Events on Brazilian Multipurpose Reactor Using the Relap5 Code
International Journal of Nuclear Energy, 2014
This work presents the thermal hydraulic simulation of the Brazilian multipurpose reactor (RMB) using a RELAP5/MOD3.3 model. Beyond steady state calculations, three transient cases of loss of flow accident (LOFA) in the primary cooling system have been simulated. The RELAP5 simulations demonstrate that after all initiating events, the reactor reaches a safe new steady state keeping the integrity and safety of the core. Moreover, a sensitivity study was performed to verify the nodalization behavior due to the variation of the thermal hydraulic channels in the reactor core. Transient calculations demonstrate that both nodalizations follow approximately the same behavior.
ICAPP 2015, 2015
Kakrapar Atomic Power Project units-3&4 (KAPP-3&4) are 700 MWe Pressurized Heavy Water Reactors (PHWR) are presently under construction. This paper presents the results of double ended large break LOCA (Loss of Coolant Accident) in Reactor Inlet Header (RIH) performed, for KAPP-3&4 as a part of safety studies to investigate the plant behavior. System thermal hydraulics code RELAP-5/MOD3.2 has been used for the analysis. Here the overall thermal hydraulics of the plant along with various control systems, trip and actuation logics have been simulated. High pressure accumulators and low pressure recirculation system of emergency core cooling system are modeled. The modeling of secondary system includes modeling of Atmospheric Steam Discharge Valves (ASDVs), Safety Relief Valves (SRVs), Condensate Steam Discharge Valves (CSDVs), Turbine (steam control) Governor valve, the U-tubes of the steam generator, the riser, the separator and the steam drum. Using this model, double ended guillotine break size in the Reactor Inlet Header(100% break) was analysed, for evaluating maximum peak clad temperature. Following postulated accidents, the event progression and the variation of different parameters like different Header pressures, mass flow rate in the core, fuel clad temperature, rate of discharge from break, injection rate of ECCS in core etc have been studied. The objective of this study is to evaluate the thermal hydraulic behavior of the core and to check the capability of ECCS to mitigate the event.
A RELAP5 model for the thermal-hydraulic analysis of a typical pressurized water reactor
Thermal Science, 2010
This study de scribes a RELAP5 com puter code for ther mal-hy drau lic anal y sis of a typ i cal pres sur ized wa ter re ac tor. RELAP5 is used to cal cu late the ther mal hy draulic char ac ter is tics of the re ac tor core and the pri mary loop un der steady-state and hy po thet i cal ac ci dents con di tions. New de signs of nu clear power plants are di rected to in crease safety by many methods like re duc ing the de pend ence on ac tive parts (such as safety pumps, fans, and die sel gen er a tors) and re plac ing them with pas sive fea tures (such as grav ity draining of cool ing wa ter from tanks, and nat u ral cir cu la tion of wa ter and air). In this work, high and me dium pres sure in jec tion pumps are re placed by pas sive in jec tion com po nents. Dif fer ent break sizes in cold leg pipe are sim u lated to an a lyze to what de gree the plant is safe (with out any op er a tor ac tion) by us ing only these pas sive com po nents. Also sta tion black out ac ci dent is sim u lated and the time re sponse of op er a tor ac tion has been dis cussed.
International Journal of Energy Production and Management, 2019
Thermal-hydraulic response of a reacTor core following large break loss-of-coolanT accidenT under flow blockage condiTion young seok bang & Joosuk lee korea institute of nuclear safety. absTracT since the revision of the requirements to consider the effect of fuel burnup on emergency core cooling system performance was proposed, flow blockage in reactor core has been one of the important issues in the thermal-hydraulic analysis of loss-of-coolant accident (loca). The present paper describes how much flow blockage would be expected following a large break loca based on the actual nuclear design data including the power and burnup of the fuel rods. a system thermal-hydraulic code, mars-ks, is used for calculation where the burnup specific data of the fuel rods is supported by a fuel performance code, fracon3. To recover the weakness of the system code in which the flow blockage under multiple rods configuration cannot be automatically simulated in hydraulic calculation, a special modelling scheme is developed and applied to the calculation. The effect of flow blockage on the thermal-hydraulic response of the reactor core is also discussed. To compensate for the uncertainty of the present flow blockage model, additional calculations are attempted for a wide range of the level of blockage.
Acta mechanica Malaysia, 2022
In this paper, a comparative analysis of loss of coolant accident (LOCA) in hot leg and cold leg of primary circuit in a VVER 1200 nuclear power plant is investigated. The effect of break size on the severity of the accident is observed. The break size was varied in the range 200-11350 cm 2. For all the accident scenarios, station blackout (SBO) condition is set up. Additionally, it is assumed that no ECCS (Emergency Core Cooling System) is available due to system malfunction. The whole scenario is simulated in PCTRAN (Personal Computer Transient Analyzer) software. Results reveal that with the increase in the size of the break area, the core uncovering time decreases sharply. However, for a break size of 2800 cm 2 or smaller, the water level in the core doesn't drop to zero, indicating that the core is partially uncovered throughout the accident scenario. In case of hot leg LOCA, the draining of the reactor vessel is observed to be more rapid compared to cold leg LOCA, while the core melting started earlier in case of cold leg.
Numerical Simulation Study on the Air/Water Counter- current Flow Limitation in Nuclear Reactors
Brazilian Journal of Radiation Sciences
After a loss-of-coolant accident (LOCA) in a Pressurized Water Reactor (PWR), the temperature of the fuel elements cladding increases dramatically due to the heat produced by the fission products decay, which is not adequately removed by the vapor contained in the core. In order to avoid this sharp rise in temperature and consequent melting of the core, the Emergency Core Cooling System is activated. This system initially injects borated water from accumulator tanks of the reactor through the inlet pipe (cold leg) and the outlet pipe (hot leg), or through the cold leg only, depending on the plant manufacturer. Some manufacturers add to this, direct injection into the upper plenum of the reactor. The penetration of water into the reactor core is a complex thermofluidodynamic process because it involves the mixing of water with the vapor contained in the reactor, added to that generated in the contact of the water with the still hot surfaces in various geometries. In some critical loc...