Advanced Thermal-hydraulic Test Loop For Accident Simulation Research Articles

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.

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.

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.

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.

An analytical investigation of cold leg small break accidents of the ATLAS facility

A previous parametric study of the direct vessel injection (DVI) line breaks was re-evaluated to see its applicability to that of the cold leg (CL) pipe breaks in advanced thermal-hydraulic test loop for accident simulation (ATLAS). Evaluation results of the tests and analyses for the major parameters, e.g., the pressurizer (PZR) pressure, downcomer water level, collapsed core water level, and clad temperature, were compared for four different CL pipe break scenarios. The overall trends of the major parameters showed reasonable behaviors between the tests and analyses. The clad temperature showed conservative behaviors in the analyses using the suggested options. The suggested counter-current flow limit (CCFL) options for the fuel alignment plate (FAP) and cross-over legs (COLs) can be applicable to any small-break loss-of-coolant accident (SBLOCA) scenario for the CL pipe and DVI line breaks in the ATLAS tests.

Thermal–hydraulic analysis and code assessment for ATLAS 6-inch cold leg break (SBLOCA) test using MARS-KS

The thermal–hydraulic analysis using MARS-KS code was performed for 6-inch cold leg break test of ATLAS (Advanced Thermal–Hydraulic Test Loop for Accident Simulation), which was the second domestic standard problem. The calculation results were compared with experimental data to assess the code capability to simulate the transient thermal–hydraulic behavior for small break loss of coolant accident (SBLOCA). The sequence of events, except for the location of loop seal clearing (LSC) and safety injection tank (SIT) injection time was predicted well. The loop seals of 1A and 2B intermediate legs were cleared at 398s in the experiment, while that of 1A was only cleared in the calculation at the same time. The prediction showed good agreement with the experimental data for pressurizer pressure and break mass flow rate. The sudden decrease and increase of water level at the LSC time were predicted qualitatively. After LSC, there was significant water level dip at SIT injection time which was not seen in the experiment. In addition, sensitivity study to investigate the cause of core level dip at SIT injection time was performed and discussions were made for it. In conclusion, MARS-KS code has good capabilities to simulate cold leg break SBLOCA, however, including interfacial heat and mass transfer, especially condensation model needs to be improved to predict more accurate results.

Development of a phenomena identification ranking table for simulating a station blackout transient of a pressurized water reactor with a thermal-hydraulic integral effect test facility

KAERI (Korea Atomic Energy Research Institute) will launch an OECD/NEA project by utilizing a thermal-hydraulic integral effect test facility, ATLAS (Advanced Thermal-hydraulic Test Loop for Accident Simulation) from April, 2014. In the OECD-ATLAS project, design extension conditions (DECs) such as a station blackout (SBO) and a total loss of feed water (TLOFW) will be experimentally investigated to identify the major thermal-hydraulic phenomena of high risk multiple failure accidents considering magnified safety concerns after the Fukushima accident. A SBO is one of the most important DECs in that without any proper operator actions, a total loss of heat sink leads to core uncover, to core damage, and ultimately a core melt-down scenario under high pressure. Due to this safety importance, a SBO is considered to be a base test item of the OECD-ATLAS project.In this study, a phenomena identification ranking table (PIRT) has been developed for identifying the major parameters affecting the thermal-hydraulic phenomena in SBO transients. Development of a PIRT for the SBO transients is expected to contribute to making strategy for performing a thermal-hydraulic integral effect test with ATLAS which includes determination of proper test conditions and improvement of measurements. The primary safety criterion (PSC) was determined to be the core mixture level from the expert panel discussion. As for the major components of the nuclear power plant, the rank of the importance and the knowledge level were summarized for each of the three temporal phases. Taking into account the important phenomena from the PIRT result, it can be confirmed that ATLAS has a capability of simulating the major thermal-hydraulic phenomena during a SBO transient as realistically as possible. The present PIRT will be directly used to make strategy for performing a thermal-hydraulic integral effect test with ATLAS in the framework of the OECD-ATLAS project.

SECOND ATLAS DOMESTIC STANDARD PROBLEM (DSP-02) FOR A CODE ASSESSMENT

KAERI (Korea Atomic Energy Research Institute) has been operating an integral effect test facility, the Advanced Thermal-Hydraulic Test Loop for Accident Simulation (ATLAS), for transient and accident simulations of advanced pressurized water reactors (PWRs). Using ATLAS, a high-quality integral effect test database has been established for major design basis accidents of the APR1400 plant. A Domestic Standard Problem (DSP) exercise using the ATLAS database was promoted to transfer the database to domestic nuclear industries and contribute to improving a safety analysis methodology for PWRs. This 2nd ATLAS DSP (DSP-02) exercise aims at an effective utilization of an integral effect database obtained from ATLAS, the establishment of a cooperation framework among the domestic nuclear industry, a better understanding of the thermal hydraulic phenomena, and an investigation into the possible limitation of the existing best-estimate safety analysis codes. A small break loss of coolant accident with a 6-inch break at the cold leg was determined as a target scenario by considering its technical importance and by incorporating interests from participants. This DSP exercise was performed in an open calculation environment where the integral effect test data was open to participants prior to the code calculations. This paper includes major information of the DSP-02 exercise as well as comparison results between the calculations and the experimental data.

Analysis of an ATLAS 6-in. cold-leg break simulation with MARS code

A transient analysis of an ATLAS (Advanced Thermal–Hydraulic Test Loop for Accident Simulation) 6-in. cold-leg break was carried out with the MARS (Multi-dimensional Analysis for Reactor Safety) safety analysis code in the frame of the Domestic Standard Problem exercise in Korea. The calculated major sequence of events of the 6-in. cold-leg break simulation showed good agreement with the measured data. The calculated break mass flowrate was predicted well, whereas accumulated mass of the break outflow was underestimated due to the underestimation of the break mass flowrate in a later phase. The general trends of the collapsed water level were well predicted in the core and the downcomer region. The loop seal clearing phenomena were observed at about 400s in the 1-A and 2-B intermediate legs in the calculation results which is identical to the experiment. To investigate the emergency core coolant (ECC) bypass phenomena, variation of the boron concentration in safety injection water was adopted. In the loop seal clearing phase of the 6-in. cold-leg break, about 40–50% of the ECC bypass fraction was calculated and after the loop seal clearance, 20–30% of ECC water bypassed, that is, not participating in core cooling directly. The boron tracking approach is considered to be a feasible methodology with which quantify the ECC bypass flow in the early phase of a small-break loss-of-coolant accident (SBLOCA).