Overview and details of the sessions of this conference. Please select a date or location to show only sessions at that day or location. Please select a single session for detailed view (with abstracts and downloads if available).
Session Chair: Alexandre Ezzidi Nakata, Nuclear Power Engineeing Consulting, Japan
Thermal-Hydraulic Analysis of ATLAS facility during Loss of Residual Heat Removal System (RHRS) using RELAP5 code
Omar S. Al-Yahia, Ho Joon Yoon
The residual heat removal system (RHRS) provides enough cooling capabilities during a mid-loop operation and shutdown condition. In the process, the water inventory is reduced and the PZR manway is opened, whereas the RHRS operates under the atmospheric pressure. However, any failure occurs to the RHRS will lead to a sequence of unfavorable events during the transient conditions. As a result, the possibility of core failure will raise. Recently, a series of the experimental test has been performed at KAERI using ATLAS (Advanced Thermal-hydraulic Test Loop for Accident Simulation) facility to enlarge the database for loss of RHRS during a mid-loop operation, in which the shutdown coolability with loss of RHRS has been investigated with respect to the reflux condensation phenomena. After loss of RHRS, it is essential to ensure that the natural circulation and reflux condensation supply the required heat removal capability for core coolability to avoid the core damage. In the present study, transient analysis of thermal-hydraulic parameters during loss of RHRS is performed using RELAP5 code. Reflux condensation occurs inside the SG U-tube for a short period owing to the countercurrent flow limit (CCFL). However, a reverse heat transfer initiates when the temperature of the secondary side of SG becomes higher than the temperature of the primary side. During the core coolant evaporation process, the water level inside RV will reduce with time, apportion of steam will leak through the PZR manway.
Transition boiling heat transfer correlation with the consideration of hysteresis effect for the best-estimation of reflood phenomena using SPACE code
Woonho Jeong, Yong Hoon Jeong
Korea Advanced Institute of Science and Technology
Heat removal rate from the surface of fuel cladding has been one of the biggest issues on nuclear power plant safety analysis. The heat transfer rate at the cladding should be accurately estimated especially during the reflood phase. Therefore, various thermal-hydraulic safety analysis codes provide correlations to predict the heat flux during the reflood. Currently, thermal-hydraulic safety analysis codes adopted instantaneous local conditions hypothesis to predict the heat flux during the boiling process. However, the heat flux, especially in the transition boiling region, is affected by the boiling process it has passed through. As a result, the heat transfer rate during the surface-cooling process and the surface-heating process are different although the local condition is identical. Transition boiling heat transfer correlations for the surface-cooling process was suggested based on Bjornard and Griffith's correlation. FLECHT-SEASET data was used to validate the correlation which adopted to the SPACE code. Predicted peak cladding temperature with the new correlation was almost identical for various FLECHT-SEASET test conditions. Quenching time was delayed for the cases with the flow velocity over 7.5 m/sec. But with the lower flow velocity, quenching time was quicken. The prediction performance of quenching time was enhanced by 10% for the cases validated. And the quenching time was more accurately predicted for the experiments with the lower flow velocity.
Experimental study of the thermophysical properties of boric acid solutions at the parameters typical of the WWER emergency mode taking into account the features of water chemistry
Azamat Sakhipgareev, Aleksander Shlepkin, Andrei Morozov, Aleksandra Soshkina
Leypunsky Institute for Physics and Power Engineering (IPPE)
WWER-1200 reactor plant is the result of the evolutionary development and improvement of existing reactors, which have proven their reliability over many years of trouble-free operation. Power units constructed under the WWER-1200 project are already in operation at the Novovoronezh and Leningrad NPPs. Passive cooling systems for the WWER-1200 reactor core include passive core flooding systems from the hydraulic accumulators of the first and second stages (HA-1 and HA-2), as well as a passive heat removal system (PHRS). In the event of LOCA, the passive safety systems operation provides at least 24 hours of cooling of the reactor core by feeding a solution of boric acid with a concentration of 16 g/kg from hydro accumulators, as well as the condensate flow from the steam generators operating in the condensation mode. Currently, the task of extending the duration of passive core cooling systems operation for new WWER reactor projects has been formulated. In the case of a prolonged supply of a solution of boric acid into the core, its concentration may increase. In this case, conditions of the core may change due to the transition to the cooling mode with a highly concentrated solution of boric acid. To assess the maximum possible duration of operation of passive safety systems, it is necessary to calculate the process of cooling the WWER reactor during an accident. For the correct calculation of the processes in the core, it is necessary to know the thermophysical properties of highly concentrated boric acid. Experimental studies of the thermophysical properties (density and kinematic viscosity) of boric acid solutions in the concentration range of 2.5-230 g/kg, taking into account the WWER primary circuit water chemistry features were carried out at the IPPE.
Analysis of the natural circulation onset transients on the bayonet tube HERO-2 facility with RELAP5
Within the framework of a National Research Program funded by the Italian Minister of Economic Development, ENEA in collaboration with SIET leaded some studies on heat exchangers for SMR and Gen-IV reactors application.
At SIET laboratories in Piacenza, an important series of experimental campaigns have been carried out to support basic studies on the bayonet tubes with the test section HERO-2. The test section consists in a couple of bayonet tubes externally heated by electric resistors with a total power of 50 kW. The facility is able to operate at a pressure of 70 bars and a flow rate of 0.1 kg/s per tube.
The thermal-hydraulic tests conducted have allowed the creation of a valuable database to characterize the bayonets heat exchange and thermal-hydraulic instabilities and, in natural circulation, to evaluate the performances in a typical DHR operations. The experimental database is useful for the qualification of computer codes used to support the design and safety analysis of innovative nuclear reactors, in particular the comprehensive test campaign carried-out in 2018, where three tests at different pressure conditions have been performed in order to verify the onset of natural circulation.
This paper presents a review of the RELAP5 post-test analysis of the natural circulation steady-states tests used to calibrate the model, in order to get a unique tool able to simulate the whole closed configuration campaign. Then, the model has been assessed with the simulation of the transient tests.
The results have shown a good capability of the RELAP5 model in retracing the transient tests with some limits linked to the in-pool condensation phenomena. Moreover, the bayonet tubes have proved to be a viable way to design DHR systems for future reactor generation.