Experimental validation of the modified linear scaling methodology for scaling ECC bypass phenomena in DVI downcomer

In the direct vessel injection (DVI) system downcomer, the direct emergency core coolant (ECC) bypass is activated during the reflood phase of a large-break loss-of-coolant accident (LBLOCA) by the interaction between the downward-flowing liquid-film and the transverse gas flow. Direct ECC bypass is reportedly the major bypass mechanism of ECC, and various experiments have been performed to obtain detailed information about the ECC bypass in a DVI downcomer. These lead to a proposed new scaling methodology, named ‘modified linear scaling’, which is expected to preserve the phase distribution in the downcomer and the ECC bypass phenomena. In the present study, modified linear scaling was experimentally validated in air–water tests comprising Test 21-D of the upper plenum test facility (UPTF). The counterpart tests of UPTF Test 21-D were performed in 1/7.3 and 1/4.0 scale models of a UPTF downcomer, and the test results were compared with the experimental data of UPTF Test 21-D. Two important parameters of direct ECC bypass – the normalized liquid-spreading width on the downcomer wall and the direct ECC bypass fraction, which is the fraction of input water bypassed out the broken cold-leg – were considered in the validation. The comparison revealed that the scaling parameters of direct ECC bypass are well preserved in the prototype and reduced models, from which we conclude that the modified linear scaling methodology is appropriate for designing a reduced test facility and for a scaling analysis of direct ECC bypass in the reflood phase of an LBLOCA.     

Scaling for the ECC bypass phenomena during the LBLOCA reflood phase

As one of the advanced design features of the APR1400 (Advanced Power Reactor), a direct vessel injection (DVI) system is adopted instead of the conventional cold leg injection (CLI) system. It is known that the DVI system greatly enhances the reliability of the emergency core cooling (ECC) system. However, there is still a dispute on its performance in terms of water delivery to the reactor core during the reflood period of a large-break loss-of-coolant accident (LOCA). Thus, experimental validation is underway. In this paper, a new scaling method, using the time and velocity reduced “modified linear scaling law”, is suggested for the design of a scaled-down experimental facility to investigate the direct ECC bypass phenomena in the PWR downcomer.     

An Experimental Study of Thermal-hydraulic Phenomena in the Downcomer with a Direct Vessel Injection System of APR1400 during the LBLOCA Reflood Phase

The direct vessel injection (DVI) mode is adopted as a safety injection system in the place of a conventional cold leg injection (CLI) mode in the Advanced Power Reactor 1400MW (APR1400). It is expected that “sweep-out” and “direct ECC (Emergency Core Cooling) water bypass” are two most important bypass mechanisms of ECC water injected through the DVI lines during the LBLOCA reflood phase. Using the test facility of plane-channel type scaled down to 1/7 ratio of the prototype reactor (APR1400), we carry out the following tests for the investigation of the two mechanisms: water film spreading test, sweep-out test, and direct ECC water bypass test. From the water film spreading test, it was found that the curvature effect is negligible and the present modified linear scaling law is more appropriate than the linear scaling law. In the sweep-out test, the continuous onset is used to analyze the water height in the downcomer and the amount of ECC water bypass by sweep-out is compared with the previous correlations. The direct ECC water bypass test is performed in order to understand the ECC water film behavior in the downcomer.     

UPTF experiment: a synopsis of full scale test results

A series of 14 tests has been run at UPTF – a 1:1 scale test facility – to investigate the thermohydraulic phenomena in a PWR primary system during blowdown, refill and reflood phases. A synopsis of the most significant test results is given to characterize the controlling phenomena in a full scale primary system under LOCA conditions.     

PSB-VVER simulation of Kozloduy NPP “loss of feed water transient”

This paper provides a comparison between the PSB test facility experimental results obtained during the simulation of loss of feed water transient (LOFW) and the calculation results received by INRNE computer model of the same test facility. Integral thermal-hydraulic PSB-VVER test facility located at Electrogorsk Research and Engineering Center on NPPs Safety (EREC) was put in operation in 1998. The structure of the test facility allows experimental studies under steady state, transient and accident conditions.RELAP5/MOD3.2 computer code has been used to simulate the loss of feed water transient in a PSB-VVER model. This model was developed at the Institute for Nuclear Research and Nuclear Energy for simulation of loss of feed water transient.The objective of the experiment “loss of feed water”, which has been performed at PSB-VVER test facility is simulation of Kozloduy NPP LOFW transient. One of the main requirements to the experiment scenario has been to reproduce all main events and phenomena that occurred in Kozloduy NPP during the LOFW transient. Analyzing the PSB-VVER test with a RELAP5/MOD3.2 computer code as a standard problem allows investigating the phenomena included in the VVER code validation matrix as “integral system effects” and ”natural circulation“. For assessment of the RELAP5 capability to predict the “Integral system effect” phenomenon the following RELAP5 quantities are compared with external trends: the primary pressure and the hot and cold leg temperatures. In order to assess the RELAP5 capability to predict the “Natural circulation” phenomenon the hot and cold leg temperatures behavior have been investigated.This report was possible through the participation of leading specialists from Kozloduy NPP and with the support of Argonne National Laboratory (ANL), under the International Nuclear Safety Program (INSP) of the United States Department of Energy.     

Major findings from LBLOCA reflood tests using the ATLAS facility

A set of LBLOCA (large-break loss of coolant accident) reflood tests was performed in the first phase of the ATLAS (advanced thermal-hydraulic test loop for accident simulation) program. Their main objectives were to identify the major thermal-hydraulic characteristics during the reflood phase of a LBLOCA for APR1400 and to provide qualified data for APR1400 licensing. The ATLAS reflood test program could be divided into two phases (Phase-1 and Phase-2) according to the target period to be simulated. The Phase-1 tests were parametric effect tests for downcomer boiling in the late reflood phase of LBLOCA and the Phase-2 tests were integral effect tests for the entire reflood phase of LBLOCA. The experimental results from both Phase-1 and Phase-2 tests reproduced typical thermal-hydraulic trends expected to occur during the APR1400 LBLOCA scenario. A separate effect test was also performed under a low reflooding rate condition to provide data to validate the RELAP5 reflood models, and its experimental results showed a gradual reflooding in the core, a subsequent quenching of the core heater rods and the cooling of the reactor pressure vessel downcomer.     

IIST and LSTF counterpart test on PWR station blackout transient

A reduced-height, reduced-pressure (RHRP) integral system test facility at the Institute of Nuclear Energy Research (INER) has been established for simulating the thermal-hydraulics of a Westinghouse three-loop pressurized water reactor (PWR). To understand whether or not the physical phenomena observed in this RHRP integral system test facility during a station blackout (SB) transient can be reliably extrapolated to those for an actual plant, a counterpart test based on the same scenarios as those of the full-height, full-pressure (FHFP) large-scale test facility (LSTF) test was performed. To see the result of differences in the design, scaling approach and facility operational conditions in the systems, the present study examines their effects on the SB transient, particularly for the tests performed at full and reduced pressures. We also identify the occurrence of key thermal-hydraulic phenomena, as well as their possible distortions. Results of the INER integral system test (ISST) facility and LSTF tests showed the common thermal-hydraulic phenomena, such as the secondary coolant boil-off and the subsequent primary coolant saturation, pressurization, coolant inventory depletion and redistribution, and core uncovery caused by coolant boil-off. The sequence and timing of the significant events during the SB transient studied in the RHRP IIST facility are also consistent (in most cases) with those for the SB transient studied in the FHFP LSTF.     

A separate effect test on the cooling behavior in a pressurized water reactor core under a low reflooding rate condition by using ATLAS and RELAP5/MOD3.3 code assessment

A separate effect test was performed on the cooling behavior in a PWR core under a low reflooding rate condition by using the ATLAS (Advanced Thermal–Hydraulic Test Loop for Accident Simulation) which is a thermal–hydraulic integral effect test facility for the pressurized water reactors APR1400 and OPR1000. Although several integral tests for the reflood phase of a large break loss of coolant accident (LBLOCA) of APR1400 have been performed with the ATLAS, the previous integral effect tests for the reflood phase of a LBLOCA are not easily simulated by existing codes, such as the RELAP5/MOD3, due to a unique phenomena in ATLAS, that resulted from an injection of large amount of subcooled water onto the heated wall of which temperature was higher than the target value.     

Commissioning of the ATLAS thermal-hydraulic integral test facility

KAERI recently constructed a new thermal-hydraulic integral test facility for advanced pressurized water reactors (PWRs) – ATLAS. The ATLAS facility has the following characteristics: (a) 1/2-height&length, 1/288-volume, and full pressure simulation of APR1400, (b) maintaining a geometrical similarity with APR1400 including 2(hot legs) × 4(cold legs) reactor coolant loops, direct vessel injection (DVI) of emergency core cooling water, integrated annular downcomer, etc., (c) incorporation of specific design characteristics of OPR1000 such as cold leg injection and low-pressure safety injection pumps, (d) maximum 10% of the scaled nominal core power. The ATLAS will mainly be used to simulate various accident and transient scenarios for evolutionary PWRs, OPR1000 and APR1400: the simulation capability of broad scenarios including the reflood phase of a large-break loss-of-coolant accident (LOCA), small-break LOCA scenarios including DVI line breaks, a steam generator tube rupture, a main steam line break, a feed line break, a mid-loop operation, etc. The ATLAS is now in operation after an extensive series of commissioning tests in 2006.     

Direct vessel inclined injection system for reduction of emergency core coolant direct bypass in advanced reactors

Multidimensional thermal hydraulics in the APR1400 (Advanced Power Reactor 1400 MWe) downcomer during a large-break loss-of-coolant accident (LBLOCA) plays a pivotal role in determining the capability of the safety injection system. APR1400 adopts the direct vessel injection (DVI) method for more effective core penetration of the emergency core cooling (ECC) water than the cold leg injection (CLI) method in the OPR1000 (Optimized Power Reactor 1000 MWe). The DVI method turned out to be prone to occasionally lack in efficacious delivery of ECC to the reactor core during the reflood phase of a LBLOCA, however. This study intends to demonstrate a direct vessel inclined injection (DVII) method, one of various ideas with which to maximize the ECC core penetration and to minimize the direct bypass through the break during the reflood phase of a LBLOCA. The 1/7 scaled down THETA (Transient Hydrodynamics Engineering Test Apparatus) tests show that a vertical inclined nozzle angle of the DVII system increases the downward momentum of the injected ECC water by reducing the degree of impingement on the reactor downcomer, whereby lessening the extent of the direct bypass through the break. The proposed method may be combined with other innovative measures with which to ensure an enough thermal margin in the core during the course of a LBLOCA in APR1400.     

Effect of the yaw injection angle on the ECC bypass in comparison with the horizontal DVI

The comparison tests for the direct emergency core cooling (ECC) bypass fraction were experimentally performed with a typical direct vessel injection (DVI) nozzle and an ECC column nozzle having a yaw injection angle to the gravity axis. The ECC yaw injection nozzle is newly introduced to make an ECC water column in the downcomer region. The yaw injection angle of the ECC water relative to the gravity axis is varied from 0 to (±)90° stepped by 45°. The tests are performed in the air–water separate effect test facility (direct injection visualization and analysis (DIVA)), which is a 1/7.07 linearly scaled-down model of the APR1400 nuclear reactor. The test results show that (1) if the ECC water column is injected into the wake region which is induced by the hot leg blunt body in the downcomer annulus, the ECC bypass fraction is greatly reduced compared with the typical horizontal ECC injection which makes ECC film on the downcomer wall. At the same time, the ECC penetration toward the lower downcomer region becomes larger than those of a typical horizontal type of direct vessel injection on the downcomer wall vertically. (2) If the ECC water column is injected near the broken cold leg, the ECC water is directly bypassed. Thus, the ECC penetration fraction is greatly reduced compared with a typical film type of the horizontal ECC injection. (3) In order to minimize the ECC bypass fraction, the ECC water should be injected toward the wake region of the hot leg blunt bodies.     

TRAC-PF1/MOD1 calculations and data comparisons for mist small-break LOCA, feed-and-bleed and steam-generator tube-rupture experiments

Los Alamos National Laboratory is a participant in the Integral System Test (IST) program initiated in June 1983 for the purpose of providing integral system test data on specific issues/phenomena relevant to post-small-break loss-of-coolant accidents, loss of feedwater and other transients in Babcock and Wilcox (B&W) nuclear plant designs. The Multi-Loop Integral System Test (MIST) facility is the largest single component in the IST program. MIST is a 2 × 4 [two hot legs and steam generators (SGs), four cold legs and reactor coolant pumps] representation of B&W lowered-loop reactor systems. It is a full-height, full-pressure facility with 1/817 power and volume scaling. Efforts are under way at Los Alamos to assess TRAC-PF1/MOD1 against data from the MIST facility.Calculations and data comparisons for TRAC-PF1/MOD1 assessment are presented for three transients run in the MIST facility. The energy removal and depressurization mechanisms in these tests are identified and the phenomena occurring in these tests compared. The tests analyzed are MIST Test 3109AA, the nominal small-break LOCA, Test 330302, a feed and bleed test with delayed high-pressure injection; and Test 3404AA, an SG tube-rupture test with the affected SG isolated. TRAC was able to predict these phenomena although the timing and magnitude of events were not always in good agreement.The MIST test have demonstrated the thermal-hydraulic phenomena expected to occur during transients in B&W nuclear plants. Because of scaling atypicalities, test results cannot be extrapolated directly to plant conditions. Although the phenomena were demonstrated in the MIST tests, there may be differences in the timing, magnitude and sequences of events in plant transients. Assessment calculations, three of which are presented here, have shown that the TRAC computer code can predict the major trends and phenomena occurring during the MIST tests with reasonable qualitative agreement. This includes complex sequences of events. Reasonable qualitative agreement is defined as meaning that major trends are predicted correctly, although TRAC values are frequently outside the range of data uncertainty. These assessment results, taken with assessment results from other facilities at a wide range of scales, provide us with confidence that the TRAC code can adequately simulate the transient phenomena possible in B&W nuclear plants.     

UPTF experiment: Flow phenomena during full-scale loop seal clearing of a PWR

To investigate the flow phenomena in the primary system of a pressurized water reactor (PWR) during a loss-of-coolant accident (LOCA) occurring with a small or intermediate break, experiments were performed at the full-scale Upper Plenum Test Facility (UPTF). Within the Transient and Accident Management (TRAM) program integral and separate effect tests were carried out to study loop seal clearing and to provide data for the further improvement of computer codes concerning the reactor safety analysis. This paper describes the UPTF tests that focus on the sequence of loop seal clearance in a four-loop operation for two different cold leg break sizes and the residual water levels, the flow patterns in, and the pressure drops across a single loop seal during the clearing. The UPTF results obtained from a single-loop seal operation are compared with experimental data and correlations available in the literature. Two correlations are proposed which allow the quantification of residual water levels in the loop seal under PWR conditions. It is shown that the steam–water test results gained from the full-scale UPTF with realistic PWR loop seal geometry differ from those obtained from the full or small-scale test facilities under air–water conditions. The UPTF experiments indicate the substantial need for steam-water test data from a full-scale facility with realistic PWR geometries in order to validate PWR LOCA thermal-hydraulic system codes to predict loop seal clearing correctly.     

Scaling methodology for a reduced-height reduced-pressure integral test facility to investigate direct vessel injection line break SBLOCA

A scaling methodology for a small-scale integral test facility was investigated in order to analyze thermal-hydraulic phenomena during a DVI (direct vessel injection) line SBLOCA (small break loss-of-coolant accident) in an APR1400 (advanced power reactor 1400 MWe) pressurized water reactor. The test facility SNUF (Seoul National University Facility) was utilized as a reduced-height and reduced-pressure integral test loop. To determine suitable test conditions for simulating the prototype in the SNUF experiment, the energy scaling methodology was propose to scale the coolant mass inventory and the thermal power for a reduced-pressure condition. The energy scaling methodology was validated with a system code (MARS) analysis for an ideally scaled-down SNUF model and that predicted a reasonable transient of pressure and coolant inventory when compared to the prototype model. For the actually constructed SNUF, the effect of scaling distortions in the test facility's thermal power and the loop geometry was analytically investigated. To overcome the limitation of the thermal power supply in the facility, the convective heat transfer between primary and secondary systems at the steam generator U-tubes was excluded and a modified power curve was applied for simulating the core decay heat. From the code analysis results for the actual SNUF model, the application of the modified power curve did not affect the major events occurring during the transient condition. The results revealed that the scaling distortion in the actual SNUF geometry also did not strongly disturb significant thermal-hydraulic phenomena such as the downcomer seal clearing. Thus, with an adoption of the energy scaling methodology, the thermal-hydraulic phenomena observed in the SNUF experiment can be properly utilized in a safety analysis for a DVI line break SBLOCA in the APR1400.     

Advanced DVI for ECC direct bypass mitigation

An ECC direct bypass fraction during a late reflood phase of a LBLOCA is strongly dependent on the characteristics of the cross flow and the geometrical configuration of a DVI in the downcomer of a pressurized light water reactor. The important design parameters of a DVI are the elevation, the azimuthal angle, and the separator to prevent a steam-water interaction. An ECC sub-channel to separate or to isolate an ECC water from a high-speed cross flow is one of the important design features to mitigate the ECC bypass phenomena. A dual core barrel cylinder as an ECC flow separator is located between a reactor vessel and a core barrel outer wall in the downcomer annulus. A new narrow gap between the core barrel and the additional dual core barrel plays the role of a downward ECC flow channel or an ECC flow separator in a high-speed cross flow field of the downcomer annulus. The flow zone around a broken cold leg in the downcomer annulus has the role of a high ECC direct bypass due to a strong suction force while the wake zone of a hot leg has the role of an ECC penetration. Thus, the relative azimuthal angle of the DVI nozzle from the broken cold leg is an important design parameter. A large azimuthal angle from a cold leg to a hot leg needs to avoid a high suction flow zone when an ECC water is being injected. The other enhancing mechanism of an ECC penetration is a grooved core barrel which has small rectangular-shaped grooves vertically arranged on the core barrel wall of the reactor vessel downcomer annulus. These grooves have the role for a generation of a vortex induced by a high-speed cross flow. Since the stagnant flow in a lateral direction and rotational vortex provides the pulling force of an ECC drop or film to flow down into the lower downcomer annulus by gravity, the ECC direct bypass fraction is reduced when compared to the current design of a smoothed wall. An open channel of grooves generates a stagnant vortex, while a closed channel of grooves creates an isolated ECC downward flow channel from a high-speed lateral flow. In this study, new design concepts for a dual core barrel cylinder, grooved core barrel, and a reallocation of the DVI azimuthal angle are proposed and tested by using an air-water 1/5 scaled air-water test facility. The ECC direct bypass reduction performances of the new design concepts have been compared with that of the standard type of a DVI injection. The azimuthal angle of the DVI nozzle from a broken cold leg varies from −15° to +52° toward a hot leg. The test results show that the azimuthal injection angle is an effective parameter to reduce the ECC direct bypass fraction. The elevation of the DVI nozzle is also an important parameter to reduce the ECC direct bypass fraction. The most effective design for reducing the ECC direct bypass fraction is a dual core barrel. The reduction fraction when compared to the standard DVI is about −30% for the dual core barrel while it is −15% for the grooved core barrel.     

Similarity study of ROSA-III and fist large break cuonterpart tests to BWR large break LOCA

A large break test in a recirculation pump suction line with the assumption of LPCI-diesel generator failure was conducted at the ROSA-III test facility of Japan Atomic Energy Research Institute. A counterpart test was also performed at the FIST test facility of General Electric Company. The objective of the tests was to develop common understanding and interpretation of the controlling thermal-hydraulic phenomena during a large break LOCA of a BWR. The fundamental thermal-hydraulic phenomena in the ROSA-III and FIST tests such as the system pressure, mixture level and fuel rod surface temperatures agreed well. The FIST test had more bundle uncovery than that in ROSA-III since lower plenum steam in the FIST test flowed out of the jet pumps when they uncovered allowing more liquid to drain from the bundle. The ROSA-III and FIST tests and a BWR counterpart were analyzed with the RELAP5/MODI (cycle 018) code. The similarity of the ROSA-III and FIST large break tests to a BWR large break LOCA has been confirmed through comparison of calculated results though they are slightly different in details. It is perhaps desirable to reexamine the DNB and interphase drag correlations and the jet pump models usedin the code.     

Algorithm for Cause Analysis Using Cause Consequence Model

In an integral system test on the reflood phenomena of a PWR LOCA with the Cylindrical Core Test Facility (CCTF), a high pressure drop was observed through the broken cold leg of the pressure vessel side. In order to understand the pressure drop and to assess the applicability of the CCTF result to the LOCA analyses for PWRs, the pressure drop characteristics through the broken cold leg is analyzed with CCTF and FLECHT SET data. The high pressure drop is explained quantitatively with the homogeneous flow model of the two-phase flow. The difference of the pressure drop between the FLECHT SET and the CCTF is attributed to the differences of the flow area scaling of the broken cold leg and the ECC water injection method. It is confirmed analytically that the high pressure drop as in the CCTF tests is expectable in a PWR system with a cold leg break due to the pressure losses at the broken cold leg nozzle and the break.     

基于ECC安注热混合试验的LOCUST 1.2分析与验证

在失水事故（LOCA）工况下安注系统投入使用时,蒸汽与安注冷却剂会发生流体热力学混合,热混合过程中冷腿段的冷却是直接影响堆芯再淹没与否的重要因素。中国广核集团有限公司自主研发了一款两相流热工水力系统分析软件LOCUST,可用于压水堆核电厂事故工况的分析计算。基于西安交通大学堆芯应急冷却系统（ECCS-XJTU）试验台架进行的堆芯应急冷却（ECC）安注热混合试验,本文使用LOCUST软件对ECC热混合试验进行了几何建模及计算分析。ECC热混合试验工况主要为不同流量下主管纯蒸汽与安注管过冷水的混合,蒸汽流量为25~125 kg/h,过冷水流量为100~500 kg/h。模拟计算结果和试验结果的对比分析表明：试验段出口质量流量计算值的最大相对误差在13.8%以内,混合后温度计算值的最大相对误差在8%以内,LOCUST在计算高温蒸汽和过冷水混合时的计算结果相对保守,总体上验证了LOCUST在LOCA下两相热混合安注计算的可靠性和准确性。     

Three-dimensional analysis of flow characteristics on the reactor vessel downcomer during the late reflood phase of a postulated LBLOCA

A three-dimensional CFD analysis has been performed on the flow characteristics in the reactor vessel downcomer during the late reflood phase of a postulated large-break loss-of-coolant accident (LBLOCA), in order to validate the modified linear scaling methodology that was applied in the MIDAS test facility of Korea Atomic Energy Research Institute. The vertical and circumferential velocity similarities are numerically tested for the 1/1 and 1/5 linear scale models for the APR1400 reactor vessel downcomer. The effects of scale on flow patterns, pressure and velocity distributions, and the impinging jet behavior are analyzed with the FLUENT code. In addition, a simplified half cylinder model with a single emergency core cooling (ECC) nozzle is numerically tested to investigate the scale effect on the spreading width and break-up of ECC water film. The qualitative and quantitative results indicate that the 1/5 modified linear scale model of the reactor vessel downcomer would reasonably preserve the hydrodynamic similarity with APR1400.     

Testing of the multi-application small light water reactor (MASLWR) passive safety systems

Experimental thermal hydraulic research has been conducted at Oregon State University for the purpose of assessing the performance of a new reactor design concept, the multi-application small light water reactor (MASLWR). The MASLWR is a pressurized light water reactor design with a net output of 35 MWe that uses natural circulation in both normal and transient operation. Due to its small size, portability and modularity, the MASLWR design is well suited to help fill the potential need for grid appropriate reactor designs for smaller electricity grids as may be found in developing or remote regions. The purpose of the OSU MASLWR test facility is to assess the operation of the MASLWR under normal full operating pressure and full temperature conditions and to assess the passive safety systems under transient conditions. The data generated by the testing program will be used to assess computer code calculations and to provide a better understanding of the thermal-hydraulic phenomena in the design of the MASLWR NSSS. During this testing program, four tests were conducted at the OSU MASLWR test facility. These tests included one design basis accident and one beyond design basis accident. During the performance of these tests, plant operations to include start up, normal operation and shut down evolutions were demonstrated successfully.