Integral Effect Test Facility Research Articles

A review of natural circulation flow in integral effect test facilities

The proposed NCFM map was reviewed using natural circulation data identified in the PWR simulated integral effect test (IET) facility with a height (length) scale ratio of 1/1. Based on this, the characteristics of natural circulation flow in the ATLAS and FESTA facilities, which are major IET facilities in Korea, are compared. It was confirmed that the natural circulation flow characteristics were well maintained within the NCFM area even at the half-height test facility of ATLAS. It was also confirmed that the natural circulation flow characteristics were well maintained within the NCFM area at the full-height test facility of FESTA, and the natural circulation flow characteristics were also maintained outside the NCFM region for a relatively small amount of coolant inventories.

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.

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.

Development of fault diagnosis for nuclear power plant using deep learning and infrared sensor equipped UAV

Fault component detection is necessary for safety and maintenance in large-scale industrial fields including nuclear power plants. Therefore, this study proposes a method for diagnosing a power plant composed of numerous components based on deep learning using a UAV with an IR sensor and a camera. The proposed method could diagnose the components and recognize the fault component in real time. In this study, a thermal–hydraulic integral effect test facility, which is a scaled-down nuclear power plant, is utilized considering the nuclear power plant. The database for the application of deep learning was performed by combining an IR intensity map and general image to enhance the performance of component classification and fault detection. Deep learning was applied using object detection and classification methods based on convolutional neural networks (CNNs) that are effective for image processing. As a result, this technology can diagnose the multi-component by a single measurement instrument. The optimal performance of component classification and fault detection was 55.9 ms per 16 batches, demonstrating a mean average precision (mAP) of 0.9913. This technology could be applied to various industries as a comprehensive component condition monitoring method for operating efficiency and safety.

Deep-learning-based system-scale diagnosis of a nuclear power plant with multiple infrared cameras

Comprehensive condition monitoring of large industry systems such as nuclear power plants (NPPs) is essential for safety and maintenance. In this study, we developed novel system-scale diagnostic technology based on deep-learning and IR thermography that can efficiently and cost-effectively classify system conditions using compact Raspberry Pi and IR sensors. This diagnostic technology can identify the presence of an abnormality or accident in whole system, and when an accident occurs, the type of accident and the location of the abnormality can be identified in real-time. For technology development, the experiment for the thermal image measurement and performance validation of major components at each accident condition of NPPs was conducted using a thermal-hydraulic integral effect test facility with compact infrared sensor modules. These thermal images were used for training of deep-learning model, convolutional neural networks (CNN), which is effective for image processing. As a result, a proposed novel diagnostic was developed that can perform diagnosis of components, whole system and accident classification using thermal images. The optimal model was derived based on the modern CNN model and performed prompt and accurate condition monitoring of component and whole system diagnosis, and accident classification. This diagnostic technology is expected to be applied to comprehensive condition monitoring of nuclear power plants for safety.

Similarity evaluation of the pump simulation loop in STELLA-2 for conservation of mechanical sodium pump characteristics

The STELLA-2 is a large-scale sodium thermal-hydraulic integral effect test facility and supports the development of PGSFR. The facility adopted Pump Simulation Loop System (PSLS) concept for the mechanical sodium pump in the reference reactor to control and to measure the primary sodium flow. Since the component (mechanical pump) is replaced by the loop, it is very important to evaluate the similarity between the pump and the loop. In this paper, to simulate the characteristic of the mechanical sodium pump, the pressure loss along the various options of the loop was evaluated and the comprehensive validity of each design options was analyzed. Using the similarity criteria based on the Richardson number and Euler number conservation, the PSLS design was finalized and the result was within the acceptable error range. Finally, the result of this study was used for construction of the overall facility, STELLA-2.

Design of large-scale sodium thermal-hydraulic integral effect test facility, STELLA-2

The STELLA program was launched to support the PGSFR development in 2012 and for the 2nd stage, the STELLA-2 facility was designed to investigate the integral effect of safety systems including the comprehensive interaction among PHTS, IHTS and DHRS. In STELLA-2, the long-term transient behavior after accidents can be observed and the overall safety aspect can also be evaluated. In this paper, the basic design concept from engineering basis to specific design is described. The design was aimed to meet similarity criteria and requirements based on various non-dimensional numbers and the result satisfied the key features to explain the reasoning of safety evaluation. The result of this study was used to construct the facility and the experiment is on-going. In general, the final design meets the similarity criteria of the multi-dimensional physics inside the reactor pool. And also, for the conservation of natural circulation phenomena, the design meets the similarity requirements of geometry and thermo-dynamic behavior.

Safety evaluation of nuclear power plant based on multiscale analyses

A loss-of-coolant accident of nuclear power plant is simulated using the system analysis code. The analytical condition is based on the international standard problem 26 conducted with the integral effect test facility. The experimental analysis is also performed, and the result of plant analysis is compared with the experiment and the experimental analysis. The discharge coefficient of critical flow model is determined so as to obtain the agreement of reactor pressure between the experiment and the experimental analysis, and is used for the plant analysis. The thermal-hydraulic phenomena in the experiment such as core heat up are simulated well by the two analyses, while some problems in the experiment for simulating plant accidents are made clear. Parametric sensitivity analyses of plant accidents are performed as an example of reliable safety evaluation using the validated plant model.

Physical Phenomena in Nuclear Thermal Hydraulics and Current Status

116 nuclear Thermal-Hydraulic Phenomena T-HP are identified in the present paper, following documents issued during the last three decades by the Committee on the Safety of Nuclear Installations of Nuclear Energy Agency of the Organization for Economic Cooperation and Development (OECD/NEA/CSNI) and by the International Atomic Energy Agency (IAEA). The derived T-HP list includes consideration of experiments performed in Separate Effect Test (SET) and Integral Effect Test (IET) facilities relevant to reactor coolant system and containment of Water Cooled Nuclear Reactors (WCNR). We consider a dozen WCNR types: Pressurized Water Reactors (PWR), Boiling Water Reactors (BWR), Russian reactors (VVER-440, VVER-1000 and RBMK), pressure tube heavy water reactors by Canada (CANDU) and India (PHWR) and so-called ‘advanced’ reactors (e.g. AP-1000 and APR-1400 designed in US and Korea, respectively). We envisage a variety of applications for the T-HP list. Four of the phenomena are helpful to characterize the current state of art in nuclear thermal-hydraulics: Counter Current Flow Limitation (CCFL), Critical Heat Flux (CHF), reflood and Two-Phase Critical Flow (TPCF). Furthermore, the T-HP identification contributes to addressing the scaling issue, performing uncertainty evaluations, developing constitutive equations and ‘special models’ in codes and prioritizing the research.

Integral effect test for steam line break with coupling reactor coolant system and containment using ATLAS-CUBE facility

To improve safety analysis technology for a nuclear reactor containment considering an interaction between a reactor coolant system (RCS) and containment, this study aims at an experimental investigation on the integrated simulation of the RCS and containment, with an integral effect test facility, ATLAS-CUBE. For a realistic simulation of a pressure and temperature (P/T) transient, the containment simulation vessel was designed to preserve a volumetric scale equivalently to the RCS volume scale of ATLAS. Three test cases for a steam line break (SLB) transient were conducted with variation of the initial condition of the passive heat sink or the steam flow direction. The test results indicated a stratified behavior of the steam-gas mixture in the containment following a high-temperature steam injection in prior to the spray injection. The test case with a reduced heat transfer on the passive heat sink showed a faster increase of the P/T inside the containment. The effect of the steam flow direction was also investigated with respect to a multi-dimensional distribution of the local heat transfer on the passive heat sink. The integral effect test data obtained in this study will contribute to validating the evaluation methodology for mass and energy (M/E) and P/T transient of the containment.

Counterpart test for LSTF 1% cold-leg break LOCA (SB-CL-32) utilizing ATLAS test facility

The counterpart test of the OECD-ATLAS project named A5.1 was performed using a thermal–hydraulic integral effect test facility, ATLAS. The target scenario for the A5.1 test was 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. The test objective is to investigate the thermal hydraulic phenomena during a cold leg SBLOCA such as core heat-up, loop seal clearing (LSC), and the effect of AM actions. To achieve the test objectives, scaling analysis between LSTF and ATLAS was conducted first. The detailed test conditions such as initial and boundary conditions and sequence of events were determined based on the scaling analysis result. The experimental result showed that overall thermal hydraulic behavior which can be expected in this kind of SBLOCA scenario was observed. However, significant differences were found in the A5.1 test result compared with the LSTF SB-CL-32 test result in the view point of the loop seal clearing phenomenon and the maximum heater rod surface temperature behavior. These differences result from the different design between two test facilities which are refer to the different prototype nuclear power plant. Thus, several test condition parameters were analyzed as the sensitivity study utilizing the thermal hydraulic system analysis code to check their effect on the system response. From the sensitivity analysis result, it was confirmed that the A5.1 test was well modeled as a counterpart test of LSTF and realized the thermal hydraulic behavior during SBLOCA transient. This test result can be utilized to evaluate the prediction capability of thermal hydraulic system analysis codes and to validate the scale-up capability of integrated effect test (IET) test data to the nuclear power plant condition.

Design and operation of the transparent integral effect test facility, URI-LO for nuclear innovation platform

Conventional integral effect test facilities were constructed to enable the precise observation of thermal–hydraulic phenomena and reactor behaviors under postulated accident conditions to prove reactor safety. Although these facilities improved the understanding of thermal–hydraulic phenomena and reactor safety, applications of new technologies and their performance tests have been limited owing to the cost and large scale of the facilities. Various nuclear technologies converging 4th industrial revolution technologies such as artificial intelligence, drone, and 3D printing, are being developed to improve plant management strategies. Additionally, new conceptual passive safety systems are being developed to enhance reactor safety. A new integral effect test facility having a noticeable scaling ratio, i.e., the (UNIST reactor innovation loop (URI-LO), is designed and constructed to improve the technical quality of these technologies by performance and feasibility tests. In particular, the URI-LO, which is constructed using a transparent material, enables better visualization and provides physical insights on multidimensional phenomena inside the reactor system. The facility design based on three-level approach is qualitatively validated with preliminary analyses, and its functionality as a test facility is confirmed through a series of experiments. The design feature, design validation, functionality test, and future utilization of the URI-LO are introduced.

Core makeup tank behavior investigation during ACME integral effect tests

Passive safety systems are widely used in the advance reactors, such as AP1000 and CAP1400. As an important component in the passive safety system, the core makeup tank (CMT) plays a key role for the safety injection and the core decay heat removal. Since the CMT not only provides the full-pressure injection but also triggers the auto depressurization system by its level signal during small-break loss of coolant accident (SBLOCA), the CMT thermal hydraulic behaviors affect the SBLOCA transient process. The thermal hydraulic behaviors of the CMT were investigated by conducting different SBLOCA tests on the ACME integral effect test facility, which includes the 2.5 cm, 5 cm and 20 cm breaks, and double ended direct vessel injection (DVI) line break. Based on the test observations, the CMT injection response to SBLOCA was divided into four phases: the circulation, the drain transition, the injection inhabitation and the final drain. The CMT injection characteristics and the dominate factors were scrutinized based on the phase division. It was found that the break condition has significant impacts on the CMT operating characteristics in each phase. The CMT circulation flow rate can be predicted, and the break size determines the circulation phase length. The CMT drain transition process is controlled by the PBL voiding process that has a close relation to the liquid level change in the cold leg. The ACC-CMT interaction was modeled, and the prediction result agreed with the test result. The comparison tests showed that the ACC nitrogen gas injection still affected the CMT drain behavior. A similar thermal stratification pattern formed during the different SBLCOA tests, and the thermal stratification remained not only in the CMT internal fluid but also in the CMT wall. The estimation of the flashing amount showed it did not play a significant role for the CMT liquid level decline, but it could induce the increases in the pressure difference between the CMT and the system and thus accelerate the CMT drain rate. The CMT internal fluid temperature determined the wall temperature, and a reverse heat transfer process occurred due to the flashing induced wall cooling during the ADS depressurization.

Integral Effect Test on Top-Slot Break Scenario With 4 Inches Cold Leg Break LOCA in ATLAS Facility

An experimental simulation for a loss of coolant accident with a 4 inch cold leg top-slot break was performed using ATLAS, an integral effect test facility at KAERI, Korea. The aims of this test are to resolve a safety issue and to validate the thermal hydraulic safety analysis codes for the activity of the 4th ATLAS-Domestic Standard Problem. The related safety issue for this test is that a loop seal clearing and loop seal reformation of APR1400 can lead to a core temperature excursion because of the characteristics of the loop seal geometry. In the experimental results, the core heater temperature slowly increased after loop seal reformation. This increase is attributed to the accumulated steam at the upper head. The accumulated steam increased the pressure and saturated the temperature in the core. The core heater temperature then increased due to the increased saturated temperature. This temperature increase therefore was not a core temperature excursion. Sensitivity analysis results using MARS-KS, which is a best-estimate thermal hydraulic system analysis code, are also presented in this study. In the 4th ATLAS-Domestic Standard Problem, 15 participants performed calculations and sensitivity analyses using various thermal hydraulic safety analysis codes (MARS-KS, SPACE, RELAP5, and TRACE). From this exercise, several parameters that affect the calculation results are brought out. The sensitivity analysis results using MARS-KS with some of these parameters are also presented. The MARS-KS calculation results are similar to the experimental data. The effects of critical flow model, break line modeling, and fine node modeling are also discussed.

Numerical assessment of LOCA experiments of RD-14M using MARS-KS code

The code assessment on the design basis events has been of fundamental importance to the safety analyses of the nuclear power plants. When a system analysis code is released to the new version by improving the existing models and correcting the found errors, it is essential to make sure that the updated code can predict and simulate reasonably the important phenomena possibly observed during the transient. This work has focused on strengthening the predictability and reliability of the system analysis code by benchmarking the invaluable loss-of-coolant-accident (LOCA) data obtained from an integral effect test facility (RD-14M). In this study, the verification tests of the inlet-header break simulating the LOCA of the CANDU-6 plants were selected as a benchmark case for MARS-KS code evaluation. First, a proper analysis model to benchmark the RD-14M experiment was envisioned into the MARS-KS input model and its steady-state conditions were achieved successfully. Subsequently, through the transient analysis experimental data of void fraction, header pressure, differential pressure between headers, and sheath temperature were compared with the code results. The code input models predicted the experimental results of the transient conditions with a good agreement. Through this study, it is expected that the MARS-KS code validation for the CANDU safety analysis could be improved significantly. In addition, this study is expected to contribute to developing practical assessment problems and providing a meaningful example for direction to code improvement.

Design of electric core simulator system (ECSS) for large-scale integral effect test facility

The second phase of STELLA program (STELLA-2) has been started to verify and validate the performance of DHRS of PGSFR, and the construction of the facility is expected to be completed in 2019. The STELLA-2 is not only aimed for investigation of decay heat removal characteristics but also demonstrates its performance while interacting with PHTS of pool-type design. For comprehensive understanding of the phenomena, all systems and components of significance in the reactor need to be modeled as closely as possible and the core is one of such significant areas to be taken into account. To simulate the thermal-fluidic behavior and to observe the Peak Cladding Temperature (PCT) phenomena in core, the Electric Core Simulator System (ECSS) has been designed through similarity consideration by the scaling analysis with conservation of dimensionless number groups under the global/local scaling law. In this paper, the design consideration and process of the electric heater rod and the ECSS is described. The equivalent property concept was introduced and the thermal-fluidic behavior in PHTS pool was also conserved. Finally, the ECSS design was evaluated by MARS-LMR code and the comparison result between reactor core and that of ECSS in STELLA-2 is also described.

Scaling and integral effect test on transition from depressurization to long-term gravity injection

The passive safety commercial pressurized water reactors (PWRs), AP1000 and CAP1400 are equipped with the automatic depressurization system (ADS). The fourth stage of ADS (ADS-4) has the largest discharging capacity among the all ADS stages. During a small break loss-of-coolant accident (SBLOCA), only when the system is fully depressurized by the ADS-4 to near the containment atmospheric pressure, the in-containment refueling water storage tank (IRWST) can deliver the large coolant inventory for the long-term core cooling by its water head. Before the IRWST injection, the core reaches its minimum inventory. So the transition from the ADS-4 depressurization to the IRWST injection is a very challenging period in a SBLOCA. Therefore, to evaluate the passive system performance by the integral effect test (IET), the transition must be simulated properly. This work conducted the scaling analysis for the ADS-4 depressurization and the IRWST injection for designing the related components of the ACME IET facility. A uniform scaling law based on the existing T-branch liquid entrainment models was proposed. For scaling the onset of the IRWST injection, the requirement for reproducing the synergetic effect in an IET was introduced in order to preserve the important event/process occurrence time with the proper conditions, and the scaling laws for the full pressure IRWST injection were obtained. For investigating the transition thermal hydraulic behaviors and the ADS-4 configuration influences on the core safety, one cold leg break test and three double-ended direct-vessel-injection (DEDVI) line break tests with the different ADS-4 failure modes were selected from the ACME SBLOCA test matrix, and the main test data related to the core safety were analyzed and compared. It was found that the ADS-4 actuation significantly accelerated the system depressurization but worsened the core inventory depletion at the same time. In addition, the break condition and the ADS-4 failure mode could differently affect the system depressurization and the core level. The ACC injection during the ADS-4 depressurization in DEDVI was critical for the core level makeup. The insufficient ADS-4 venting capacity delayed IRWST injection, but also mitigated the core inventory depletion. It suggests the ADS-4 design can be potentially optimized in the passive safety system design.

Integral effect test on cooling performance of hybrid safety injection tank

This paper reports an experimental research on the performance of the passive emergency core cooling system (PECCS) using an integral effect test facility in Korea Atomic Energy Research Institute (KAERI). The PECCS consists of two hybrid safety injection tanks (H-SIT), two medium pressure safety injection tanks (MP-SITs) and an automatic depressurization system consisting of four stages. In this study, an integral effect test (IET) was conducted on a loss of coolant accident (LOCA), which requires the operation of the primary makeup system. The H-SIT (SITs #1 and #3) injection starts successfully when the primary side pressure falls to 10.0 MPa, and the MP-SIT (SITs #2 and #4) injection starts when the pressure falls further to 4.21 MPa. The coolant injection of H-SIT occurs intermittently at certain points of time, rather than continuously. It can be seen that the flow generation of the H-SIT near the pressure plateau occurs exactly when the secondary side is depressurized.