vgbe energy journal 5 (2022) - International Journal for Generation and Storage of Electricity and Heat

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Assessment of loss of shutdown cooling system accident during mid-loop operation in LSTF experiment using SPACE Code Minhee Kim, Junkyu Song and Kyungho Nam Abstract Bewertung des Ausfalls des Nachkühlsystems während des Mitte- Loop-Betriebs im LSTF-Experiment unter Verwendung des SPACE-Codes Während eines Anlagenstillstands wird die Wärmeabfuhr aus dem Reaktorkern, bei Verbleib der Brennelemente im Kern, durch das Nachkühlsystem sicher gestellt. Der Ausfall dieses Systems kann zu einem Verlust der Wärmeabfuhr aus dem Kern führen und stellt daher ein sicherheitstechnisch relevantes Ereignis dar. Während bestimmter Instandhaltungsarbeiten z. B. am Dampferzeuger, wird das Kühlmittelniveau bis zur Mitte der Rohre Autors Minhee Kim Junkyu Song Kyungho Nam Central Research Institute Korea Hydro & Nuclear Power Co. Daejeon, Korea The SPACE code, which is Safety and Performance Analysis Code for Nuclear Power Plants, has been developed in recent years by the Korea Hydro & Nuclear Power Co. through collaborative works with other Korean nuclear industries and research institutes, and is approved by Korea Institute of Nuclear Safety (KINS) in March 2017. The SPACE is a best-estimate two-phase threefield thermal-hydraulic analysis code in order to analyze the performance and evaluate the safety of pressurized water reactors. Each field equation is discretized based on finite volume approach on a structured mesh and an unstructured mesh together with an one-dimensional pipe meshes [7]. For time integration method, the semi-imdes heißen und des kalten Stangs abgesenkt. Dies wird als Mitte-Loop-Betrieb bezeichnet, und damit ist der Kühlmittelstand am niedrigsten mit Brennstoff im Kern. Daher stellt der Verlust der Nachwärmeabfuhr während des Mitte-Loop-Betriebs den relevanten Betriebszustand dar. Die vorliegende Arbeit konzentriert sich auf die Bewertung von SPACE 3.0 bei der Vorhersage des primären und sekundären Systemverhaltens nach einem Nachwäremeabfuhr-Verluststörfall während des Mitte-Loop- Betriebs des LSTF-Experiments unter Bezugnahme auf den Bericht NUREG/IA-0143. Die berechneten Ergebnisse werden mit den Ergebnissen von RELAP5 und den experimentellen Daten in Bezug auf das stationäre und instationäre Verhalten verglichen. l I Introduction During a plant outage, while the fuel remains in the core, the core is cooled by the Residual Heat Removal (RHR) system. The loss of the RHR can lead to loss of heat removal from the core and is a safety concern. During certain stage of maintenance, such as installation of steam generator nozzle dams, the RCS coolant level is lower to centerline of hot leg and cold leg pipes. This is called mid-loop operation and the coolant level is lowest while the fuel remains in the core. Therefore, the loss of RHR during midloop operation represents the most limiting condition for loss of RHR incidents. The accident can be occurred by an isolation valve closure or a loss of vital ac power in the RHR suction line, or a loss of RHR pump flow due to air ingestion. If the loss of RHR flow should continue for a certain period of time, the reactor vessel coolant has possibility on core boiling and uncover. In order to analyze the thermal hydraulic phenomena following the loss of RHR accident, the numerical and experimental studies have been performed. Nakamura et al. conducted the experiments of loss of RHR accident during mid-loop operation in the ROSA-IV/LSTF facility [1]. In the experiments, the primary pressurization behavior after the coolant boiling in the core was observed. Also, the system integral responses were investigated through analyzing the steam and non-condensable gas behavior in the RCS. The opening location and size in a pressurizer or a horizontal leg was analyzed as major experimental parameters. In numerical approach, the major thermal hydraulic phenomena and process were evaluated using RELAP5 system code [1, 2]. The calculation results were compared with ROSA-IV/LSTF experimental data. The present paper is focused on the assessment of SPACE 3.0 in predicting the system primary and secondary behavior following the loss-of-RHR accident during the mid-loop operation of LSTF experiment in reference to NUREG/IA-0143 report [2]. The calculated results are compared with RELAP5 results and experimental data in terms of steady-state and transient behavior. II Code descriptions 46 | vgbe energy journal
Assessment of loss of shutdown cooling system accident using SPACE Code plicit scheme is used. The SPACE code is package of input and output, hydrodynamic model, heat structure model, and reactor kinetics model. The input package performs a reading the input and restart files, a parsing the data files, an allocating the memory, and checking the unit conversion. Hydrodynamic model package is composed of constitutive models, special process models, and component models, and hydraulic solver. The hydraulic solver is based on two-fluid, threefield governing equations, which are composed of gas, continuous liquid, and droplet fields. Therefore, the SPACE code has the advantage in solving a dispersed liquid field as well as vapor and continuous liquid fields in comparison with existing nuclear reactor system analysis codes, such as RELAP5 [ISL, 2001], TRACE [NRC, 2000], CATHARE [Robert et al., 2003], and MARS-KS [KAERI, 2006]. Constitutive models are composed of correlations by the flow regime map to simulate the mass, momentum, and energy distributions, such as surface area and surface heat transfer, surface-wall friction, droplet separation and adhesion, and wall-fluid heat transfer. In order to analyzed the physical phenomena of the NPP, special process and system components are modeled. Major special process and component models are critical flow model, counter current flow limit model, off-take model, abrupt area change model, 2-phase level tracking model, pump model, safety injection tank model, valve model, pressurizer model, and separator model, etc. The package of heat structure model calculates the heat addition transfer and removal. The heat structure model includes transient heat conduction of rectangular or cylindrical geometry, and has various boundary conditions of convection, radiation, user defined variables such as temperature, heat flux, and heat transfer coefficient. In order to calculate the nuclear fission heat of a fuel rod, the point kinetics methodology is used in the heat conduction equation. Reactivity feedbacks are considered in terms of moderator temperature, boron concentration fuel moderator density, reactor scram, and power defect. Decay heat of ANS- 73, -79, and -2005 models are also implemented. The 3.0 version of the code was released through various validation and verification using the separated or integral loop test data and the plant operating data. The approved code version will be used in the safety analysis of operating PWR and the design of an advanced reactor. III Modeling information F i g u r e 1 shows the nodalization to simulate the LSTF facility with the SPACE code. The modeling is based on 179 hydrodynamic cell and 202 heat structures. The reactor pressure vessel includes the lower plenum, upper plenum, downcomer, and core, upper head and guide thimble channel (cell 100 to 156). The core is modeled as two channel with 12 cells per each channel connected by crossflow. The two channel arrangement is adopted in order to assess the multi-dimensional effect such as the natural circulation behavior in the core. The power distribution of the two channel core is 60 % for high power channel and 40 % for low power channel. The LSTF system are described by an intactloop (cell 400 to 499) and a broken-loop (cell 200 to 299) in an almost symmetrical way. Each loop consists of a SG inlet and outlet, loop seal, SG U-tube, reactor coolant pump, hot leg, and cold leg. The pressurizer is connected to the hot leg of intact-loop through the surge line elements. The secondary sides of two SGs (cell 300 to 399 and 500 to 599) are composed using an identical nodalization. In order to analyze the cold leg opening with loss of RHR accident, the openings are modeled by a trip valve. The opening sizes are Fig. 1. Nodalization diagram for LSTF experiment. vgbe energy journal 5 · 2022 | 47
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