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This project has received funding from the European Union's Horizon 2020 research and innovation programme under agreement No 945275

2023 Frequency and amplitude of flashing-induced instability in an open natural circulation loop

Frequency and amplitude of flashing-induced instability in an open natural circulation loop

Abstract

Several nuclear reactor designs rely on passive containment cooling systems. The so-called containment wall condenser relies on natural circulation loops to extract heat from high-temperature steam in the containment to a water tank at ambient pressure. In such passive systems, phase changes can happen and cause flow instabilities in the cooling loop. The flashing-induced instability occurs when the heated fluid in the riser suddenly vaporizes due to a hydrostatic pressure decrease. This instability causes periodic flow peaks, which are of major concern but whose characteristics have not been studied quantitatively.

This paper presents two analytical models that predict the flashing frequency and a maximum flow amplitude from geometry and basic operating parameters such as power level and reservoir temperature. The expressions are derived from a physical analysis and do not involve any calibration constants. The flashing frequency appears to be driven by the power level, the inlet temperature and the riser pipe geometry. For the amplitude, the maximum flow rate can be expressed in a Froude number that depends only on the total pressure losses. These models are validated against PASI experiments and system-scale simulations with the CATHARE 3 code, both performed as part of the European Commission funded PASTELS project. Additional data from numerous experimental studies in the literature are used to extend the validity range of the frequency model.

Successfully validated against experimental data and additional simulations, these models provide an explicit relationship between oscillations characteristics and design parameters, making them valuable tools for nuclear engineers.

2023 Assessment of TRACE code for modeling of passive safety system during long transient SBO via PKL/SACO facility

Passive system SACO PKL Safety condenser

Abstract

Passive safety systems are integrated into the latest generation of Light Water Reactors (LWRs), including small modular reactors. This paper employs the US-NRC TRACE thermal hydraulic code to examine the performance of a passive safety condenser known as SACO, designed to serve as the ultimate heat sink for dissipating decay heat during accident scenarios. The TRACE model is constructed with reference to the PKL/SACO test facility. The safety condenser (SACO) is interconnected with the PKL facility via the secondary side of steam generator 1, effectively serving as a third natural circulation cooling loop during accident scenarios. In the present research, the thermal-hydraulic behavior of the PKL facility is investigated in the presence of the SACO passive safety system during an extended SBO with Loss of AC Power accident scenario. This SBO can be categorized into three distinct phases depending on the activation of the SACO system and the refilling process of the SACO pool. The first phase is depressurizing using primary and secondary relief valves, the second phase is cooling down using SACO system, and the third phase is the refilling of SACO pool. The findings indicate that the SACO system effectively manages to dissipate all decay heat, even though there is temporary evaporation of the SACO water pool. Furthermore, this study provides sensitivity analysis for the assessments of system codes on the selection of maximum time step.

2023 PASTELS project - overall progress of the project on experimental and numerical activities on passive safety systems

PASTELS project overall progress of the project on experimental and numerical activities on passive safety systems

Abstract

Nuclear accidents such as Fukushima Daiichi have highlighted the potential of passive safety systems to replace or complement active safety systems as part of the overall prevention and/or mitigation strategies. In addition, passive systems are key features of Small Modular Reactors (SMRs), for which they are becoming almost unavoidable and are part of the basic design of many reactors available in today’s nuclear market. Nevertheless, their potential to significantly increase the safety of nuclear power plants still needs to be strengthened, in particular the ability of computer codes to determine their performance and reliability in industrial applications and support the safety demonstration.

The PASTELS project (September 2020–February 2024), funded by the European Commission “Euratom H2020” programme, is devoted to the study of passive systems relying on natural circulation. The project focuses on two types, namely the SAfety COndenser (SACO) for the evacuation of the core residual power and the Containment Wall Condenser (CWC) for the reduction of heat and pressure in the containment vessel in case of accident.

A specific design for each of these systems is being investigated in the project. Firstly, a straight vertical pool type of SACO has been implemented on the Framatome’s PKL loop at Erlangen. It represents a tube bundle type heat exchanger that transfers heat from the secondary circuit to the water pool in which it is immersed by condensing the vapour generated in the steam generator. Secondly, the project relies on the CWC installed on the PASI test loop at LUT University in Finland. This facility reproduces the thermal-hydraulic behaviour of a Passive Containment Cooling System (PCCS) mainly composed of a CWC, a heat exchanger in the containment vessel connected to a water tank at atmospheric pressure outside the vessel which represents the ultimate heat sink. Several activities are carried out within the framework of the project. Different tests are conducted on these integral test facilities to produce new and relevant experimental data allowing to better characterize the physical behaviours and the performances of these systems for various thermo-hydraulic conditions. These test programmes are simulated by different codes acting at different scales, mainly system and CFD codes. New “system/CFD” coupling approaches are also considered to evaluate their potential to benefit both from the accuracy of CFD in regions where local 3D effects are dominant and system codes whose computational speed, robustness and general level of physical validation are particularly appreciated in industrial studies. In parallel, the project includes the study of single and two-phase natural circulation loops through a bibliographical study and the simulations of the PERSEO and HERO-2 experimental facilities. After a synthetic presentation of the project and its objectives, this article provides the reader with findings related to the physical analysis of the test results obtained on the PKL and PASI installations as well an overall evaluation of the capability of the different numerical tools to simulate passive systems.

2023 TRACE code simulation of the interaction between reactor coolant system and containment building with passive heat removal system

References Aksan, et al., 2009 N. Aksan et al., “Passive safety systems and natural circulation in water cooled nuclear power plants.” 2009. Google Scholar Al-Yahia et al., 2022 O.S. Al-Yahia, I. Clifford, K. Nikitin, H. Ferroukhi “Parametric Study on the

Abstract :

Several integral and separate effect test facilities are constructed around the world on scaled-down bases to simulate the complex thermal-hydraulic phenomena of the commercial nuclear power plant. PKL is one of the integral test facilities of four loops of western-type KWU pressurized water reactor (PWR) with 1:1 in height scale and 1:145 in volume and power scale. Using the PKL facility, a large database has been created through numerous experimental tests investigating transients such as Loss of Coolant Accidents (LOCAs) and station blackouts. On the other hand, the PASI facility is a separate effect test facility that has been built at LUT University to simulate the Containment Wall Cooling (CWC) passive safety system for the AES-2006 NPP type at a 1:2 height scale, 1:281 in volume, and 1:141 power scale. This paper utilizes the newly released US-NRC TRACE (version 5 – patch 7) system thermal-hydraulics code to investigate the thermal hydraulic interaction phenomena between the reactor coolant system (RCS) and containment building, referring to the PKL facility as the RCS and the PASI facility as the containment heat removal system. During LOCAs, the pressure in the containment is a key parameter to evaluate the core behavior and the passive containment cooling system (PCCS) performance. This study emphasizes the ability of CWC to reduce the pressure in the containment and remove the core residual heat during LOCA accidents. Additionally, it contributes to evaluating the capability of system analysis codes to model the complex interaction between the RCS and containment building.

The TRACE model for the PASI facility has been validated against the respective experimental results. Then, the PKL G7.1 test has been utilized to validate the PKL TRACE model during the hot leg SBLOCA test. The results of TRACE simulations show good agreement with experimental data, including natural circulation behavior, core exit temperature (CET), and peak cladding temperature (PCT). After the validation stage, PKL and PASI TRACE models are coupled to create a virtual experimental test. The coupled model uses the CWC scaled up from the PASI facility. During SBLOCA, steam released from the primary loop to the containment is condensed on the CWC heat exchanger tubes. The results illustrate the complex feedback interaction between the RCS and containment building. The CWC can reduce the pressure in the containment during LOCAs and extract a large amount of the core residual heat. This research can provide a better understanding of the natural circulation and flow instability of the containment passive cooling system, and its influence on the RCS performance. As well as, it provides additional reference data to system code users for the further development of scaling and coupling methodologies. The coupling approach evaluates the capabilities of TRACE code for modeling the containment building.

Publications

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Assessment of TRACE code for modeling of passive safety system during long transient SBO via PKL/SACO facility
Frequency and amplitude of flashing-induced instability in an open natural circulation loop
PASTELS project - overall progress of the project on experimental and numerical activities on passive safety systems
TRACE code simulation of the interaction between reactor coolant system and containment building with passive heat removal system