Thermal-hydraulic Test Loop Research Articles

Scaling analysis of nuclear power plant (APR1400) utilizing ATLAS integral effect test result of A5.1 (Counterpart test for LSTF 1% cold-leg break LOCA)

ATLAS is the thermal hydraulic integral effect test loop that can simulate the same pressure and temperature condition with nuclear power plant. It was designed with a height ratio of 1/2 and a volume ratio of 1/288 with reference to APR1400. The counterpart test against LSTF which was named as A5.1 test was performed within the OECD/NEA ATLAS international joint research project to compare major thermal hydraulic phenomena between two test facilities. The main scenario for the A5.1 test is the LSTF SB-CL-32 test. The SB-CL-32 test is a 1% horizontal small break loss of coolant accident (SBLOCA) at the cold leg with cold leg injection of the emergency core cooling system (ECCS) and accident management (AM) action.From the counterpart test results, the significant differences were found in the viewpoint of the thermal–hydraulic phenomena. It was concluded that the different result were caused by the different design characteristic of each test facility because they referred the different nuclear power plant. Thus, the applicability of the ATLAS test results to its reference nuclear power plant was evaluated by utilizing ATLAS A5.1 test results which was the counterpart test with the LSTF.The transient scenario and initial and boundary conditions of the A5.1 test were adopted to APR1400 after applying scaling ratios of ATLAS. The MARS-KS code was utilized to analyzed thermal hydraulic behaviors of APR1400 and the analysis results were compared with the scaled A5.1 test result. From the result of comparison, it was concluded that the ATLAS test results predicted the thermal hydraulic behavior of the nuclear power plant well. Thus, it was confirmed that ATLAS is a well-designed integral effect test facility which is designed with an appropriate scaling method.

Validation of TRACE 5.0 for steam line break with passive auxiliary feedwater system via the ATLAS facility

The TRAC/RELAP Advanced Computational Engine (TRACE) reactor systems code was validated for the Passive Auxiliary Feedwater System (PAFS) using the experimental data from the Advanced Thermal-Hydraulic Test Loop for Accident Simulation (ATLAS) facility. The accident scenario which emulates a Double-Ended Guillotine Break (DEGB) in the main steam line of Steam Generator 1, was meticulously analyzed, with a focus on transient events and steady-state conditions with an error margin of less than 2%. Comprehensive comparative analyses of key thermal hydraulic parameters, including system pressures, temperatures, flow rates, and water levels, were conducted. Discrepancies were systematically identified, investigated, and effectively resolved. The experimental results underscore the successful role of passive safety systems in mitigating potential accident impacts, emphasizing their pivotal contribution to the overall safety and reliability of nuclear reactor systems. Furthermore, the simulation results effectively mirror the experimental data, demonstrating TRACE's capability to adequately simulate the behavior of a passive feedwater system during a main steam line break.

Analysis of control rod driving mechanism nozzle rupture with loss of safety injection at the ATLAS experimental facility using MARS-KS and TRACE

Korea Atomic Energy Research Institute (KAERI) has operated an integral effect test facility, the Advanced Thermal-Hydraulic Test Loop for Accident Simulation (ATLAS), with reference to the APR1400 (Advanced Power Reactor 1400) for tests for transient and design basis accidents simulation. A test for a loss of coolant accident (LOCA) at the top of the reactor pressure vessel (RPV) had been conducted at ATLAS to address the impact of the loss of safety injections (LSI) and to evaluate accident management (AM) actions during the postulated accident. The experimental data has been utilized to validate system analysis codes within a framework of the domestic standard problem program organized by KAERI in collaboration with Korea Institute of Nuclear Safety. In this study, the test has been analyzed by using thermal-hydraulic system analysis codes, MARS-KS 1.5 and TRACE 5.0 Patch 6, and a comparative analysis with experimental and calculation results has been performed. The main objective of this study is the investigation of the thermal-hydraulic phenomena during a small break LOCA at the RPV upper head with the LSI as well as the predictability of the system analysis codes after the AM actions during the test. The results from both codes reveal that overall physical behaviors during the accident are predicted by the codes, appropriately, including the excursion of the peak cladding temperature because of the LSI. It is also confirmed that the core integrity is maintained with the proposed AM action. Considering the break location, a sensitivity analysis for the nodalization of the upper head has been conducted. The sensitivity analysis indicates that the nodalization gave a significant impact on the analysis result. The result emphasizes the importance of the nodalization which should be performed with a consideration of the physical phenomena occurs during the transient.

Analysis of 4-inch cold leg top-slot break LOCA in ATLAS experimental facility using MARS-KS

Abstract Korea Atomic Energy Research Institute (KAERI) has operated an integral effect test facility, the Advanced Thermal-Hydraulic Test Loop for Accident Simulation (ATLAS) with reference to the Advanced Power Reactor 1400 MW (APR1400) for transient and design basis accidents (DBAs) simulation. An experiment for a 4-inch cold leg top-slot break was performed at ATLAS to resolve a safety issue that the loop seal reformation (LSR) of APR1400 could lead to the increase of peak cladding temperature. In addition, the experimental data has been utilized to validate system codes within a framework of domestic standard problem (DSP) program organized by KAERI in collaboration with Korea Institute of Nuclear Safety (KINS). In this study, the experiment has been analyzed by thermal-hydraulic system analysis code, MARS-KS 1.5 and a comparison with experimental and calculation results has been performed. Since the top-slot break is not a typical break geometry for safety analyses, this study aims at examining the applicability of MARS-KS to the top-slot break accident where the LSR occurs repeatedly. The results revealed that overall physical behavior during the accident was predicted by the code, appropriately. MARS-KS showed the excursion of the peak cladding temperature because of the LSR as in experiment. It has been confirmed that the core integrity was maintained because the temperature excursion by the LSR was not large enough to alter the acceptance criteria. In addition, it is presented the results of the sensitivity analysis of parameter that affect the figure of merits.

Integral-Effect Test for Loss of Residual Heat Removal System During Mid-Loop Operation in ATLAS Facility

A test called B4.2 in the OECD-ATLAS2 project was performed to simulate loss of the residual heat removal system (RHRS) during mid-loop operation (MLO) using a thermal-hydraulic (T-H) integral-effect test facility: the Advanced Thermal-Hydraulic Test Loop for Accident Simulation (ATLAS). The main purpose of this test was to investigate a T-H transient in the reactor coolant system (RCS) during loss of the RHRS and to evaluate the effectiveness of reflux condensation and the capability of a safety injection tank (SIT) on shutdown coolability. The initial and boundary conditions for the B4.2 test were appropriately determined according to a state of MLO corresponding to 65 h after reactor trip in the Advanced Power Reactor 1400 MW(electric) (APR1400). During the loss of RHRS accident transient simulation, major T-H parameters such as system pressures, temperatures, and collapsed water levels in the RCS were measured, and unique T-H phenomena such as reflux/cocurrent condensations, off-take, countercurrent flow, and countercurrent flow limitation were investigated. In this paper, the overall T-H behavior in the RCS during a simulated loss of the RHRS with SITs is highlighted to provide a better understanding of T-H phenomena regarding coolability with reflux condensation.

Comparison of Blind and Open Calculation Results for Top-Slot Break LOCA in Fourth ATLAS Domestic Standard Problem

The advanced thermal-hydraulic test loop for accident simulation (ATLAS) was developed and operated at The Korea Atomic Energy Research Institute. The ATLAS is operated to simulate accidents in pressurized water reactors (PWRs). A domestic standard problem (DSP) using the ATLAS was proposed for transferring the database of the integral effect test to Korean nuclear researchers and developing the safety analysis methodology of PWRs. The fourth DSP (DSP-04) exercise was performed during 2015–2017 with 15 participants (13 organizations), that are universities, government, and nuclear industries. In DSP-04, a top-slot break at the cold leg was chosen as the target scenario to resolve an issue about the effect of loop seal clearing and the reformation on the peak cladding temperature during the cold leg top-slot break LOCA for APR1400. The participants performed a code calculation for the experimental simulation and sensitivity studies for the enhancement of the code. This paper includes brief information about the experimental and major code assessment results.

Simulation of subcooled flow instability for high flux research reactors with ATHLET

Abstract The ATHLET code has been improved and validated for the safety analysis of high flux research reactors against experimental data of the Thermalhydraulic Test Loop of Oak Ridge National Laboratory (ORNL). Therefore, an extensive series of “stiff mode” experiments were recalculated. The comparison between experiments and ATHLET posttest calculations shows that the extended code simulates accurately the thermalhydraulic conditions and the flow instabilities in a wide range of inlet velocity, heat and mass flux.

Integral Effect Test and MARS-KS Calculation with Uncertainty Propagation Analysis for Direct Vessel Injection Line Break Intermediate-Break Loss-of-Coolant Accident

To investigate thermal-hydraulic phenomena during an intermediate-break loss-of-coolant accident (IBLOCA) and evaluate the effect of a direct vessel injection (DVI) line break, an integral effect test using the Advanced Thermal-hydraulic Test Loop for Accident Simulation (ATLAS) test facility was conducted as the B3.2 test item of the international cooperation project Organisation for Economic Co-operation and Development (OECD)–ATLAS Project Phase 2 (ATLAS-2) (OECD-ATLAS2). The initial and boundary conditions for the test were determined referring to the Advanced Power Reactor 1400 MW(electric) (APR1400) as a prototype with three-level scaling methodology. A single-failure criterion was applied to the operation of the safety injection pump (SIP), and four safety injection tanks (SITs) were available to cool down the reactor coolant system. In the test result, as the break nozzle was located at the DVI line, the clearance of the upper downcomer could make an effective flow path of the steam toward the break and quench the reactor core. Maximum cladding temperature was measured before clearance of the upper downcomer. Coolant inventory in the reactor pressure vessel was maintained due to the safety injection without any further core heatup. So, it was proved that the current design of the safety systems in APR1400 had a sufficient long-term cooling capability with a single SIP during a DVI line break IBLOCA. The ATLAS test data were utilized to evaluate the prediction capability of the thermal-hydraulic system code Multi-dimensional Analysis of Reactor Safety KINS Standard (MARS-KS) for a DVI line break IBLOCA scenario. The calculation result with the uncertainty propagation analysis using the PArallel computing Platform IntegRated for Uncertainty and Sensitivity analysis (PAPIRUS) toolkit proved that major phenomena such as uncovery of the core or intermittent injection of the SIT flow could be reasonably predicted by the MARS-KS code.

Validation of interfacial area concentration model for simulating bubbly and cap/slug flow behaviors in large diameter pipes using LSTF, ATLAS, and horizontal pipe experiment data

This study discusses Interfacial Area Concentration (IAC) models developed for bubbly and cap/slug flow regimes in a large diameter pipe incorporating recently developed small bubble fraction correlations in various flow conditions. The main focus of this study was the robust simulation of the behaviors of bubbly and cap/slug flows in large diameter pipes observed in Large Scale Test Facility (LSTF), Advanced Thermal-hydraulic test Loop for Accident Simulation (ATLAS), and horizontal flow test facility. Since the IAC was one of the key parameters to achieve this, an IAC model for large diameter pipes was first derived based on the existing Hibiki-Ishii IAC model for bubbly flow, and the Shen-Hibiki IAC model for cap bubbly-to-churn flow. The volume fraction of group-1 bubbles, i.e., small spherical bubbles, in the IAC model, which have been developed based on the assumption that the group-1 bubbles exponentially decay in the slug flow regime, was refined. As a result, a more physical model of the group-1 bubble number density that remains constant in the cap/slug flow regime was imposed for large diameter pipes. The validity of the IAC model for large diameter pipes in predicting the interfacial drag force between gas and liquid phases in large pipes of multiple test facilities was then examined. The validation of the proposed IAC model was conducted through simulations of 1) the loss of Residual Heat Removal (RHR) during the mid-loop operation of the LSTF, 2) the 6-inch Small Break Loss of Coolant Accident (SBLOCA) of the ATLAS, and 3) the two-phase flow behaviors under a wide range of fluid conditions in the horizontal pipe test facility. Improved results were obtained with the proposed model as it indicated that there was good agreement between the calculated results and the experimental data throughout the whole transient.

MARS best-estimate evaluation using ATLAS test data on 8.5″ cold leg break LOCA

A MARS best-estimate evaluation was performed for 8.5″ cold leg break loss-of-coolant accident (LOCA) in the advanced thermal–hydraulic test loop for accident simulation (ATLAS) facility. The ATLAS system was modeled in one-dimensional manner and the break nozzle was modeled by a break valve using the Henry-Fauske critical flow. First, a best-fitted critical flow discharge coefficient was obtained by comparing the test data and the calculations especially on the pressurizer (PZR) pressure and integrated break flow. Then, for a trial calculation related to the counter-current flow limit (CCFL) phenomena for the three locations, all the same options adopted in the LSTF (large scale test facility) evaluation were used. No good agreements were obtained between the test data and the calculations on the core collapsed water level and core cladding temperature. Another approach for the CCFL options was selected: Wallis model for inlets of the SG U-tube and the SG inlet plenum bottom; Bankoff model for the fuel alignment plate (FAP). From trial calculations, best-fitted CCFL options were obtained by comparing the test data and the calculations on the core collapsed water level and core cladding temperature.

Comparison of loop seal clearing behavior between SBLOCA and SBO-related tests in the ATLAS facility

Loop seal clearing (LSC) behavior was investigated for small-break loss-of-coolant accident (SBLOCA) and station blackout (SBO)-related tests in the Advanced Thermal-Hydraulic Test Loop for Accident Simulation (ATLAS) facility. Major characteristics of LSC behavior between two scenarios were identified and summarized. In SBLOCA tests, the first LSC provided an instant event connecting the pressure plateau phase to the boil-off phase, and tended to induce a resultant peak cladding temperature (PCT) for larger break sizes. Every LSC induced an intermittent inventory redistribution between the downcomer (DC) and the core. In SBO-related tests, most LSCs occurred under a nearly stagnant or even reverse loop flow condition, especially in the hot legs, and showed little effect on PCT occurrence. After LSC, water level changes in the core and loop seals showed synchronizing trends but opposite directions. LSC also induced mass transfer between the DC and core. And a visual observation on the sequence of an LSC was discussed.

Overall thermal-hydraulic behavior in an SBO test using HSIT in the ATLAS facility

A station blackout test called SBO-HSIT (Station Blackout-Hybrid Safety Injection Tank) was performed at the ATLAS (Advanced Thermal-Hydraulic Test Loop for Accident Simulation) facility. An overall review of the SBO-HSIT test was performed, including comparisons with a previous SBO test and small break loss-of-coolant accident (SBLOCA) tests relating to similar thermal-hydraulic behavior. The decay heat release in the SBO-HSIT test was carried out through the main steam safety valves (MSSVs) until the steam generator (SG) dryout, and after that by blowdown through the power operated safety relief valve (POSRV). Most of the major thermal-hydraulic behavior happened during this blowdown period. It is noteworthy that the reactor pressure vessel (RPV) took charge of the pressurizer (PZR) function after it reached a saturated condition and then the PZR became a simple buffer tank along the POSRV flow path. This seemed to hinder the PZR inventory from being drained to the primary circuits. Loop flows showed unstable conditions under a slightly subcooled condition of the upper plenum of RPV (RPV-UP) region, and the natural circulation was broken after the SG U-tube became empty. Loop seal clearings (LSCs) occurred twice during the test and the downcomer (DC) water levels maintained significantly higher values compared to the SBLOCA cases. Pseudo differential pressure (DP) values appeared in some pressure balance lines (PBLs) due to the existence of void in the PBLs, and the injection flowrates from the HSITs tended to increase after the intrusion of voids into the corresponding PBL. Quite a large amount of emergency core cooling water (ECCW) inventory remained in the closed HSITs in the test. Finally, a core heatup occurred at low core water level despite a rather large amount of PZR inventory, whose drainage was hindered by the pressurizer function of the RPV, and of HSITs inventories, whose drainage was interrupted by the closure logic for HSITs, respectively.

Laplace transform finite volume modeling of water hammer along fluid–structure interaction

Water hammer or propagation of pressure waves generates profound forces through pipelines of industrial high pressure processes which causes structural vibration of the pipe in both radial and axial directions. To model the sudden rupture of a pipeline system the fluid–structure interaction, FSI, is taken into account by coupling the structural vibration equations to the fluid dynamic equation. In this paper, Laplace transform finite volume, LTFV, which is a new technique along the finite element method is developed to treat fluid transient and the structural vibration equations respectively.To evaluate the numerical results, a Thermal Hydraulic Test Loop (THTL facility) which has been designed and constructed for experimental research on the physical phenomena, characteristics and performance of the safety systems involved in plants is used. To conduct tests for representing a sudden break condition in the loop, the THTL facility has been equipped by devices and sensors to record pressure and vibration signals during simulated accidents. Under steady condition, by an electrical signal an electric valve, Break valve, is opened and simultaneously pressure along pipe vibration signals close to valve is recorded. Comparing the experimental data to results from numerical modeling validated the implemented method.

Simulation of pressure waves propagation following LOCA in piping systems using Laplace Transform Finite Volume

The detailed dynamic modeling and the simulation of the rapid depressurization of PWR following leak or loss of coolant accident, LOCA, is a key element of the safety analysis in the nuclear power plants. Early in a LOCA, the blowdown at the break point causes the propagation of an acoustic wave through the primary circuit. The local pressure gaps due to the depressurization wave propagation may lead to the component recoils and the internal structure movements. In this paper to obtain the pressure depressurization rate behavior in hypothetical LOCA, a thermal hydraulic test loop, THTL facility, has been designed and constructed. The pressure change data are provided to model the trend of the pressure decreasing along the water hammer oscillation peaks that are produced as soon as the accident is initiated. For determination of the intensity of the water hammer pressure waves due to LOCA, the governing equation is solved using Laplace Transform Finite Volume, LTFV, method and the boundary conditions are set on using the obtained experimental data from the THTL facility.To evaluate the method, the analysis of the water hammer data which is obtained experimentally from THTL facility are compared with simulation results of LTFV technique which reveals the reliability of the method.

Analysis of LSC phenomena of ATLAS cold leg SBLOCA tests using MARS-KS code

ABSTRACT Loop seal clearing (LSC) is an important phenomenon for the safety of a pressurized water reactor (PWR) during a small-break loss-of-coolant accident (SBLOCA). The investigation on an LSC phenomenon of 4″, 6″, and 8.5″ break cold leg SBLOCAs simulated by Advanced thermal–hydraulic Test Loop for Accident Simulation (ATLAS) is performed using a Multi-dimensional Analysis of Reactor Safety-KINS Standard (MARS-KS) code. The LSC triggers earlier for larger break sizes during tests and calculations. LSCs occur during the simultaneous sudden decrease of steam condensation rate and the sudden increase of the break volumetric flow rate while the core volumetric flow rate increases slowly in calculation. The five phases of an SBLOCA transient are blowdown, pressure plateau, LSC, boil-off, and core-recovery phase, which can be identified by observing the volumetric flow rate and the time-dependent pressure variation. Loop seal refilling (LSR) occurs due to the trivial steam flow rate to the crossover leg inlet in calculation. The sensitivity analysis shows that the combination of countercurrent flow limitation (CCFL) model option for hot leg and steam generator (SG) inlet (Kutateladze, c = 1.36, m = 1), crossover legs (Kutateladze, c = 1, m = 1), and SG U-tubes (Wallis, c = 1, m = 1) provide good prediction of the LSC phenomenon and thermal-hydraulics behaviors in an SBLOCA transient by MARS-KS code calculation.

A pre-test analysis of ATLAS SBO with RCP seal leakage using MARS code

The accident in Fukushima Daiichi nuclear power plants shows the important of developing coping strategies for extended station blackout (SBO) scenarios of the nuclear power plants (NPPs). Many NPPs in United State of America are applying FLEX approach as main coping strategies for extended station blackout (SBO) scenarios. In FLEX strategies, outside water injection to reactor cooling system (RCS) and steam generators (SGs) is considered as an effective method to remove residual heat and maintain the inventory of the systems during the accident. This study presents a pretest calculation using MARS code for the Advanced Thermal-hydraulic Test Loop for Accident Simulation (ATLAS) SBO experiment with RCP seal leakage scenario. In the calculation, the turbinedriven auxiliary feed water pumps (TDAFPs) are firstly used after SBO initiation. Then, the outside cooling water injection method is used for long term cooling. In order to minimize operator actions and satisfy requirements of APR1400 emergency operation procedure (EOP), the SGs Atmospheric Dump Valve (ADV) opening ratio, auxiliary feed water (AFW) and outside cooling water injection flow rates were investigated to have suitable values. The analysis results would be useful for performing the experiment to verify the APR 1400 extended SBO optimum mitigation strategy using outside cooling water injection.

Code assessment of ATLAS integral effect test simulating main steam-line break accident of an advanced pressurized water reactor

ABSTRACT KAERI has been operating an integral effect test facility, Advanced Thermal–Hydraulic Test Loop for Accident Simulation (ATLAS), for accident simulations of advanced pressurized water reactors. As an integral effect test database for major design basis accidents has been accumulated, a domestic standard problem (DSP) exercise using ATLAS was proposed in order to transfer the database to domestic nuclear industries and to contribute to improving the safety analysis technology for pressurized water reactors (PWRs). As the third DSP exercise, a double-ended guillotine break of the main steam-line at an 8% power without loss of off-site power was decided as a target scenario. Seventeen domestic organizations joined this DSP exercise. They include universities, government, and nuclear industries. The participants of DSP-03 were classified into three groups and each group has focused on the specific subject related to the enhancement of the code assessment; (1) scaling capability of the ATLAS test data by comparing with the code analysis for a prototype, (2) multi-dimensional thermal–hydraulic phenomena anticipated during the steam-line break transient, (3) effect of various models in the one-dimensional safety analysis codes.

WITHDRAWN: Thermal-hydraulic test loop identification using nonlinear method; an approach to mathematical modeling of the energy system

In this study, two mathematical identification methods, used to identify the thermodynamic parameters of Columbia University 1.5 MW test loop as a low energy system. This test loop is a scaled loop of Savanah River Site (SRS) reactors and the experiment was the onset of flow instability experiment. Recursive least squares (RLS) algorithm and a non-linear (NL) algorithm are the two algorithms used for identification. Test section flow rate and pressure drop considered as the single input and the single output of the system respectively. 600 data set entries are the input used to calculate the parameter vectors and 190 data set entries used for evaluation of the models. For conservative calculations, the test data selected over the range of the training data. The system has been identified in the best manner using both methods with a calculation error less than 1%. By comparing calculated parameters with the measured data it can be concluded that non-linear algorithm has more speed to determine the coefficients of the proposed equations and can be used as a reactor transient monitoring tool like a sensor. However, the RLS algorithm properly identified the system parameters via the linear method, too.

Comparison of critical flow models for SBLOCA tests of the ATLAS facility

A comparison of critical flow models between the Trapp–Ransom and Henry–Fauske models for all SBLOCA (small break loss of coolant accident) scenarios of the ATLAS (advanced thermal–hydraulic test loop for accident simulation) facility was performed using the MARS-KS code. For the comparison of the two critical models, the accumulated break mass was selected as the main parameter for the comparison between the calculations and tests. Four cases showed the same respective discharge coefficients between the two critical models, e.g., 6″ CL (cold leg) break and 25%, 50%, and 100% DVI (direct vessel injection) breaks. In the case of the 4″ CL break, no reasonable results were obtained with any possible Cd values. In addition, typical system behaviors, e.g., PZR (pressurizer) pressure and collapsed core water level, were also compared between the two critical models. From the comparison between the two critical models for CL breaks, the Trapp–Ransom model predicted quite well with respect to the other model for the smallest and larger breaks, e.g., 2″, 6″, and 8.5″ CL breaks. In addition, from the comparison between the two critical models for DVI breaks, the Trapp–Ransom model predicted quite well with respect to the other model for the smallest and larger breaks, e.g., 5%, 50%, and 100% DVI breaks. In the case of 50% and 100% breaks, the two critical models predicted the test data quite well.

The characteristics of break flows of SBLOCA tests in ATLAS facility

The characteristics of break flows of small break loss of coolant accident (SBLOCA) tests in the ATLAS (Advanced thermal–hydraulic Test Loop for Accident Simulation) facility were investigated. For an investigation of the characteristics of break flows for SBLOCA tests, cold leg (CL) pipe break and direct vessel injection (DVI) line break tests were compared with respect to several kinds of viewpoints, e.g., pressurizer (PZR) pressure vs. accumulated break mass; comparison of accumulated break masses; occurrence of choking at the break nozzle; comparison of upstream conditions; and comparison of the discharge coefficient. In the comparison of PZR pressure vs. accumulated break mass, the overall trends were similar between CL pipe and DVI line breaks. In the comparison of accumulated break masses, the accumulated break mass of the CL pipe break was larger than that of the DVI line break for four different break sizes. In the comparison of the occurrence of choking at the break nozzle, it was found that there should be choking at the break nozzle in the CL pipe and DVI line SBLOCA tests. For a comparison of the upstream conditions, upstream fluid temperatures showed effects of downcomer wall temperature in a DVI line break. For a comparison of the Cd values, the geometry effect, e.g., the L/D ratio, was discussed, especially regarding the two smallest break tests.