Conference Agenda

In 2024, a novel separate-effect-test (SET) facility for two-phase critical (TPCF) flow studies was commissioned at LUT University’s Nuclear Engineering Laboratory. CRiticAl Flow Test facility (CRAFTY) utilizes a straight long tube for discharging subcooled water from an upstream pressure vessel to atmospheric pressure. Prior, a plethora of two-phase critical flow experiments have been carried out in the world. A preliminary literary survey found out that there is a lack of two-phase critical flow experiments utilizing very long length-to-diameter (L/D) ratio tubes (>200). In a postulated primary-to-secondary leak in a PWR, the L/D ratio of the tube can be upwards from 1000 depending on the steam generator design.

In CRAFTY, the L/D ratio and tube diameter can be conveniently changed with interchangeable discharge tubes. A discharge tube with an inner diameter of 13 mm and closely resembling the VVER-440 steam generator tube (inner diameter of 13.2 mm) was utilized in the tests conducted in 2024. The length-to-diameter ratio of the tube was 350 which is close to the half of an average length of the VVER-440 steam generator tube.

Altogether 12 discharge experiments with subcooling varying from 5 °C to 60 °C and upstream pressure from 5 MPa to 8 MPa were conducted. The nominal pressure difference between the primary and secondary circuit in VVER-440 is around 7 MPa.

This paper discusses the experiment results, introduces a simplified critical mass flux model utilizing a modified Jakob number, and presents some simulation results obtained with a system thermal hydraulic code.

In order to realistically simulate the thermal-hydraulic behavior and accident progression during a multiple failure accident, an integral effect test was performed to simulate an SBLOCA (Small break loss-of-coolant accident) with failure of safety injection in the ATLAS-CUBE test facility, which can simulate the thermal hydraulic interaction between the RCS (Reactor coolant system) and the containment. With the break at the cold leg, failure of the safety injection was assumed, whereas an accident management (AM) action was implemented to initiate the safety injection pumps (SIP). The test result confirmed the sufficient grace time during the multiple failure scenarios, including safety injection failure and loop seal clearing phenomena. The compartments acted as passive thermal sinks, effectively maintaining containment pressure below the set-point of spray injection, and ensuring long-term cooling without spray system operation. The test data in the ATLAS-CUBE facility was utilized to assess SPACE and CAP codes. The linked calculation of both codes was performed with considering the M/E (Mass and energy) transport and the P/T (Pressure and temperature) build-up in the containment. From comparing the test and calculation result, it was found that a higher pressure and temperature of the containment was predicted in the multi-volume of the containment in the CAP code calculation. The uniform temperature inside the containment in the single-volume case could overestimate the heat transfer at the passive heat sink and it affected a slower increase of the pressure and temperature of the containment.

Natural circulation imbalance or interruption (NCI) phenomena observed in C3.1 test of OECD-ATLAS3 project will be described in the present paper. When the primary forced flow is lost, the reactor core decay heat is generally removed through natural circulation (NC) convection: the flow is driven by the coolant density differences in the steam generators (SGs) as heat sink and in the reactor pressure vessel (RPV) as heat source.
During an asymmetric cooldown conditions between two steam generators, especially for the case of APR1400, Natural circulation imbalance between the two loops might occur. Moreover, natural circulation imbalance trend can be resulted in NCI phenomenon that means a complete loss of natural circulation flow in a certain loop (or loops) of the primary system. This NCI scenario may prevent the RCS to reach the entry conditions for the residual heat removal system (RHRS). Therefore, an understanding on the NCU is important.
C3.1 test consists of two separated tests named by Run1 and Run2 to compare an effect of primary coolant inventory. These two runs were performed on December 4th and 6th, 2023, respectively. Main objectives of the test were to investigate natural circulation imbalance and NCI phenomenon induced from an asymmetric cooldown transient operation, and to investigate an effect of related parameters such as the decay core power rate, injection flow rate of auxiliary feed water to unaffected SG, and discharge flow rate of ADV. In the paper, test results were compared with post-test calculation results.

Ensuring the safety of nuclear power plants is paramount, Integral Effect Test (IET) facilities have been utilized to verify the performance of reference reactors and assess the application of newly proposed technologies. Before the construction of IET facilities, System behavior analysis should be conducted to ensure that IET facilities can adequately represent reference reactors. In this study, Platform for Advanced TRaining and Integrated OPR1000 Thermal-hydraulic Test (PATRIOT), using refrigerant as the working fluid, was demonstrated to exhibit behavior similar to a reference reactor under station blackout (SBO) conditions through the utilization of system analysis codes. The PATRIOT, developed at UNIST based on the OPR-1000 design, operates with R134a refrigerant at 26.5 bar on the primary side and 13.5 bar on the secondary side. The MARS-KS code was used to analyze SBO behavior, and the R134a properties were generated within compatible pressure ranges for system analysis. The results were compared to the ATLAS, an IET facility developed by KAERI for APR-1400, which has similar design characteristics to OPR-1000. Compared to ATLAS, PATRIOT exhibited less pressure reduction and faster onset of dry-out phenomena, attributed to the lower latent heat and heat transfer of R134a. Despite these differences, the behavior of PATRIOT was similar to ATLAS, which demonstrated the feasibility of utilizing R134a in IET facilities. Therefore, It is confirmed that PATRIOT can simulate the reference reactor. Furthermore, considering the necessity of refrigerants for IET facilities to scale down, this study could contribute to the development and validation of refrigerant-based IET facilities.

In September 2024, SMART100 with a Passive Safety Injection System and Containment Pressure and Radioactivity Suppression System obtained standard design approval from the Korean regulatory agency. SMART-ITL built to evaluate the operating performance and safety of SMART100 was equipped with all passive safety systems except CPRSS. It is designed to simulate most accidents that can occur in SMART100, including transient accidents such as CLOF, SGTR, and FLB as well as SBLOCA. The role of the PSIS during SBLOCA is to supply coolant to the reactor for 72 hours without operator intervention, and its injection performance by gravity head was verified using SMART-ITL. The Passive Residual Heat Removal System operated in almost all accidents occurring in SMART100 is a natural circulation cooling system in which the condensation heat exchanger connected to the secondary side of the steam generator is contained in the Emergency Cooling Tank, and removes the core residual heat absorbed in the steam generator to the ECT. The heat removal performance of the PRHRS was verified through various types of accident simulation tests. This paper deals with the performance of the PSIS and the PRHRS confirmed from the thermal-hydraulic test results using SMART-ITL. In all individual accidents where the passive safety systems were activated, they performed sufficiently to bring the reactor coolant system to a safe shutdown.