Small-break Loss-of-coolant Accident Tests Research Articles

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.

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.

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.

Top-down scaling methodology from the LSTF facility to a three loop PWR plant applied to a SBLOCA event – The ROSA 1.2 test

The scaling methodology describes quantitatively the possible differences in the behavior between a reduced size experimental facility and a full scale commercial plant for a determined transient. Then scaling is an essential tool in the design and operation processes of a reduced size facility in order to be able to simulate the behavior of large commercial plants. Since, in this way, it is possible to advance in the development of safety systems, prediction of situations likely to produce accidental sequences, etcetera. Therefore the scaling calculations should be carried out from the earlier stages of the experimental facility conceptual design, so that its experimental data results could be directly extrapolated to a commercial plant. Evidently, the scaling analysis has to be carried out for the experimental test plans, in order to check the capacity to transpose the experimental data of the reduced size facilities to the commercial plants. Consequently, the major objective of the scaling methodology is to evaluate quantitatively the applicability of small size test facility data to predict the behavior of full size commercial plants.The top-down scaling step of the H2TS scaling methodology, between the LSTF experimental facility and a Pressurized Water Reactor of Siemens-KWU type for the ROSA 1.2 test, has been carried out along this document to evaluate the global system behavior. In order to achieve this main purpose, the general description of both facilities, the transient scenario, and the results of the scaling analysis for a small break loss of coolant accident (SBLOCA) in the hot leg are shown. The scaling analysis methodology used is the top-down global system analysis, from which, it is intended to establish thermal–hydraulic similarity between a scaled facility and a full scale industrial plant. With this aim, the accidental sequence of the ROSA 1.2 test has been divided into its five main time phases, analyzing separately each of them. In each phase the similarity groups (π-monomials and π-monomial groups) have been defined and applied. In such a way that only a reduced number π-monomials are needed to develop the scaling methodology, in addition, many of these similarity groups appear in several phases.The key aspect of this document is to show the applicability of the top-down scaling analysis methodology to a commercial plant of a different type to that of the reference NPP at which the experimental facility replicates, i.e., test carried out in a Westinghouse type experimental facility, scaling analysis applied to a Siemens-KWU type. In such a way that, it has been proved the possibility to use the measurements made in an experimental facility of different type of the reference plant for an accidental scenario, in particular, applicability of a SBLOCA test data from an experimental facility of type Westinghouse to a commercial plant of type Siemens-KWU.

Investigation of thermal hydraulic behavior of SBLOCA tests in SMART-ITL facility

An experimental investigation of the thermal hydraulic behavior of small break loss of coolant accident (SBLOCA) tests in an integral test loop (ITL) of a system-integrated modular advanced reactor (SMART) was performed, especially on the pressurizer safety valve (PSV) line and safety injection (SI) line breaks. Compared to typical phases of SBLOCAs in conventional and advanced active pressurized water reactors (PWRs), four sequential phases were identified: a blowdown to the upper downcomer (UDC) under saturation conditions, a pressure plateau under forced circulation, boil-off after the reactor coolant pump (RCP) trip, and core level restoration after the safety injection tank (SIT) injection or long-term cooling. The pressure plateau was deeply dependent under UDC saturation conditions and there were mass transfers among the reactor vessel (RV) regions during the pressure plateau. In the boil-off phase, a short reverse flow occurred from the lower dowoncomer (LDC) to the primary side of steam generator (SGP) owing to the occurrence of vaporization in the LDC region. The core level restoration is mainly dependent on the SIT injections. In the secondary system, the fluid conditions on the secondary side of steam generator (SGS) outlets were changed during the tests from superheat to saturation, from saturation to subcooling, and from subcooling to superheat. In the passive safety injection system (PSIS), there was a short reverse flow in the pressure balance lines (PBLs) of the core make-up tanks (CMTs) just after the break, and sufficient injection flowrates of the CMTs were achieved after the partial clearing/blocking of the PBL. In the case of a SIT actuation signal, duration times for the hydraulic equilibrium in pressure among the CMTs, SITs, and RV were needed.

Validation of computer code ‘ATMIKA’ against RD-14M Small Break LOCA experiments

The simulation of Small Break Loss-of-Coolant Accident (SBLOCA) experiments in the RD-14M integral test facility is performed under the auspices of International Atomic Energy Agency (IAEA) as an International Collaborative Standard Problem (ICSP) with the objective to benchmark and validate the in-house developed system thermal–hydraulic neutronic computer code ‘ATMIKA’, extensively used to analyze postulated events in Indian PHWRs. RD-14M is an 11MW, full-elevation-scaled extensively instrumented thermal hydraulic Canadian test facility, possessing most of the key components of a CANDU (CANada Deuterium Uranium) primary heat transport system (PHTS). The loop configuration is similar to figure-of-eight geometry of a typical CANDU circuit and it is intended to reproduce the important geometric features of a reactor PHTS and the appropriate operating conditions. ‘ATMIKA’ prediction and its comparison against SBLOCA experimental results are compared in this paper. A specific SBLOCA experiment ‘B9006’ is selected for the Computer code ‘ATMIKA’ predictions. Test B9006 is a 7-mm inlet header break experiment with pressurized accumulator emergency coolant injection (ECI) and represents most complete SBLOCA test conducted in RD-14M that includes all the phases of the transient (blow-down, high-pressure ECI, secondary pressure ramp (crash cool), refill, low pressure ECI, exponential pump ramp, and natural circulation). This simulation demonstrates that ‘ATMIKA’ is adequately capable of predicting the break discharge, PHTS depressurization, channel flow rate, channel voiding, fuel sheath temperatures and high pressure core injection flow through ECCS accumulator and initiation of low pressure ECI for test B9006. ‘ATMIKA’ predicted results are compared with experimental results and it is seen that predicted results for all phases of transient are in good agreement with experimental results.

Reflux condensation behavior in SBLOCA tests of ATLAS facility

The behavior of the reflux condensation heat transfer in a hot side steam generator (SG) U-tubes during a cold leg (CL) pipe and a direct vessel injection (DVI) line break in small break loss-of-coolant accident (SBLOCA) tests of the ATLAS facility was investigated including MARS code calculations. Among the SBLOCA tests, a 6″-CL pipe and 50%-DVI line break SBLOCA test were selected to investigate the behavior of the reflux condensation. A reflux condensation heat transfer seemed to occur from the time the SG U-tubes were half-empty to near the loop seal clearing (LSC). It was found that a transition regime existed between the reflux condensation heat transfer and reverse heat transfer. The remaining reflux condensate in SG U-tubes owing to the counter-current flow limit (CCFL) phenomenon and a separating effect of liquid carry-over and/or entrainment with steam moisturizing seemed to affect the thermal-hydraulic behavior of the transition regime. It was also found that the steam flowrate of the loop pipings and SG U-tubes seemed to have a strong effect on the duration time of the transition regime, e.g., a larger steam flowrate results in a longer duration. From a comparison of the reflux condensation behavior between the ATLAS tests and MARS code calculations, overall qualitative agreements were found between the two cases. The largest discrepancies were found in the SG inlet plenum water level between the two cases, and the authors suggest that the combination effects of the remaining reflux condensate in SG U-tubes and a separating effect of liquid carry-over and/or entrainment with steam moisturizing are the reasons for the discrepancies in the water levels of the SG inlet plenum, which seems to be difficult to simulate in the current version of the MARS-KS code.

Analyses of ACME integral test results on CAP1400 small-break loss-of-coolant-accident transient

CAP1400 has been developed as a new passive safety and generation III + reactor based on AP600/1000 technology. Like previous reactors, the small-break loss-of-coolant-accident (SBLOCA) is still one of the most important safety problems that have to be considered before it's widely applications. Although extensive theoretical and experimental results of SBLOCA have been reported in previous studies, the dominant SBLOCA transient mechanisms and the important thermal hydraulic phenomena have not been clearly explained, especially in the passive safety system, due to the complexity of the passive safety nuclear reactor geometry coupled with the transient two-phase fluid interactions. In this study, a new constructed thermal hydraulic integral effect test facility, named as advanced core-cooling mechanism experiment (ACME), which is designed and constructed for supporting CAP1400 safety review, was used to investigate the SBLOCA transient in passive safety plants. 1-inch, 2-inch and 8-inch cold leg breaks and a double-ended direct-vessel-injection (DEDVI) line break were selected as the critical and essential SBLOCA tests. The system pressure, injection flow rate, core level and fluid temperature were considered as the safety related parameters to investigate the system responses to the SBLOCA. In addition, the conditions that affect the SBLOCA chronology were investigated as well. Also, a few important thermal hydraulic phenomena observed in the ACME SBLOCA test were reported, and their effects on the transient were analyzed. The dominant SBLOCA transient mechanisms in the passive safety system were presented. This study not only helps to improve the understanding of SBLOCA mechanisms and important thermal hydraulic phenomena in passive safety systems, but also sheds some lights on SBLOCA analysis, test facility scaling, safety performance evaluation, code simulation and even the design optimization for the passive safety PWRs.

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 loop seal clearings for the SBLOCA tests

An analytical investigation of the loop seal clearing (LSC) in the small-break loss-of-coolant accident (SBLOCA) for direct vessel injection (DVI) line and cold leg (CL) breaks was performed using the MARS-KS code. LSC behaviors of the ATLAS test and analyses showed quite similar trends with each other, but those of the APR1400 analyses, different from the ATLAS test and analysis. With respect to the effect of modeling, the LSC behaviors showed a strong dependence of the system and component modelings. Critical thermal hydraulic parameters important to reactor safety, i.e., core and downcomer water level, cladding temperature, and crossover leg (COL) water level, were also evaluated for the two typical scenarios of an SBLOCA. Most of the thermal hydraulic behaviors were greatly dependent on the system and component modelings of the MARS analyses including the multi-dimensional effect of the reactor vessel (RV) and the application of the counter-current flow limit (CCFL) model.

An experimental investigation of loop seal clearings in SBLOCA tests

An investigation of the loop seal clearing (LSC) in a small-break loss-of-coolant accident (SBLOCA) for direct vessel injection (DVI) line and cold leg (CL) breaks was performed. The behavior of an LSC appears to be closely related to the break location and break size. In the tests of SBLOCAs, a loop seal or cross-over leg (COL) in the broken loop was cleared first, and the number of loop seals cleared was dependent on the break size. The larger the break size was, the more the loop seals or COLs that were cleared. The location of the LSCs appeared to have a consistent behavior under each scenario. In the SBLOCA tests, the downcomer water level just before an LSC was a very important parameter to the peak cladding temperature (PCT). The initiation of an LSC might not be related to the existing flooding condition, but to the magnitude of the pressure difference between the reactor upper head and downcomer, which is sufficient to push the upflow leg of a COL. The sustaining of an LSC without refilling was evaluated using the test data and existing flooding conditions.

Reliability of Core Exit Thermocouple for Accident Management Action during SBLOCA and Abnormal Transient Tests at ROSA/LSTF

Presented are experimental results on the general performances of core exit thermocouples (CETs) to detect core overheat for accident management (AM) action by using the Large-Scale Test Facility (LSTF) of the ROSA Program of the Japan Atomic Energy Agency. The LSTF is a full-height, full-pressure, and 1/48-volumetric-scaled model of a 4-loop pressurized water reactor (PWR). This study was motivated by a significant delay in the time and temperature rise of the CETs from core heat-up during a vessel top head small break loss-of-coolant accident (SBLOCA) test. A certain delay in time and temperature rise of the CETs was also observed in various SBLOCA and abnormal transient tests. Such CET performances are derived from thirteen LSTF tests as follows: (1) general CET performances are obtained in the form of equations including cases under limited influences of water fall-back from hot legs, (2) the major reason for the delay is the interaction of three-dimensional steam flows with low-temperature structures in and around the core and core exit, (3) break location was insignificant except for the PV top and bottom break cases, and (4) CET superheat is suitable for AM instead of the temperature value for significantly high or low pressure transients. The applicability of the results to PWR is discussed further.

A blind simulation of RD-14M small-break LOCA experiments using CATHENA code

The results of two Small Break Loss of Coolant Accident (SBLOCA) experiments in RD-14M test facility and their predictions by CATHENA code, which is used to analyze postulated events in CANDU reactors, are compared in this paper. Two specific SBLOCA experiments selected for the CATHENA code predictions are B9006 and B9802. Test B9006 is a 7-mm inlet header break experiment with pressurized accumulator emergency coolant injection and represents most complete SBLOCA test conducted in RD-14M. Test B9802 is a 3-mm inlet header break experiment with full channel power to study boiling in channels and condensation in steam generators in a slowly depressurizing loop rather than a blow-down. These blind simulations demonstrate that CATHENA code is capable of adequately predicting the primary pressure depressurization, channel flow rate, channel voiding for tests B9006 and B9802, and the high pressure core injection flow by CATHENA accumulator model and switching time from high pressure to low pressure injections for test B9006. The CATHENA predictions of fuel sheath temperatures for test B9006 are in much better agreement with the test measurements than those for test B9802, because CATHENA code could not capture the oscillatory behavior of channel flow and consequently sheath temperature when local fuel sheath surface was being intermittently dried and rewet under near stagnant flow and full power conditions of test B9802. Therefore, it is concluded that further works are required to appropriately predict the sheath temperature spike and its fluctuation during the transient of test B9802 where the critical heat flux and post-dryout heat transfer are important governing phenomena.

Reliability of Core Exit Thermocouple for Accident Management Action during SBLOCA and Abnormal Transient Tests at ROSA/LSTF

Presented are experimental results on the general performances of core exit thermocouples (CETs) to detect core overheat for accident management (AM) action by using the Large-Scale Test Facility (LSTF) of the ROSA Program of the Japan Atomic Energy Agency. The LSTF is a full-height, full-pressure, and 1/48-volumetric-scaled model of a 4-loop pressurized water reactor (PWR). This study was motivated by a significant delay in the time and temperature rise of the CETs from core heat-up during a vessel top head small break loss-of-coolant accident (SBLOCA) test. A certain delay in time and temperature rise of the CETs was also observed in various SBLOCA and abnormal transient tests. Such CET performances are derived from thirteen LSTF tests as follows: (1) general CET performances are obtained in the form of equations including cases under limited influences of water fall-back from hot legs, (2) the major reason for the delay is the interaction of three-dimensional steam flows with low-temperature structures in and around the core and core exit, (3) break location was insignificant except for the PV top and bottom break cases, and (4) CET superheat is suitable for AM instead of the temperature value for significantly high or low pressure transients. The applicability of the results to PWR is discussed further.

INTEGRAL BEHAVIOR OF THE ATLAS FACILITY FOR A 3-INCH SMALL BREAK LOSS OF COOLANT ACCIDENT

A small-break loss of coolant accident (SB-LOCA) test with a break size equivalent to a 3-inch cold leg break of the APR1400 was carried out as the first transient integral effect test using the ATLAS (Advanced Thermal-hydraulic Test Loop for Accident Simulation). This was the first integral effect test to investigate the integral performance of the test facility and to verify its simulation capability for one of the design-basis accidents. Reasonably good thermal hydraulic data was obtained so that an integral performance of the fluid sub-systems was identified and control performance of the ATLAS was confirmed under real thermal hydraulic conditions. Based on the measured data, a post-test calculation was carried out using the best-estimate thermal hydraulic safety analysis code, MARS 3.1, and the similarity between the expected and actual data was investigated. On the whole, the post-test calculation reasonably predicts the major thermal hydraulic parameters measured during the SB-LOCA test. The obtained data will be used to enhance the simulation capability of the ATLAS and to improve an input model of the ATLAS for simulation of other target scenarios.

The results of 5% small-break LOCA tests and natural circulation tests at the ROSA-IV LSTF

Five 5% small-break loss-of-coolant accident (SBLOCA) experiments and two natural circulation experiments were conducted at the ROSA-IV Large Scale Test Facility (LSTF). The liquid holdup in the upflow side of steam generator (SG) U-tubes temporarily depressed the core collapsed liquid level below the bottom of core during the loop seal clearing in the cold-leg break SBLOCA tests. This phenomena was affected by the core power and core bypass but was affected little by the actuation of the high pressure injection system. Overall thermal-hydraulic phenomena in a loop seal line break test was similar to that of cold-leg break tests, however, the liquid holdup phenomena played a little role. In a hot-leg break test a temporary but rapid depression of the core liquid level was observed immediately after the initiation of accumulator injection which caused condensation and depressurization in the cold leg. The change of natural circulation flow rate with the decrease of primary system mass inventory was qualitatively the same as observed in Semiscale, LOBI and PKL. The SG effective overall heat transfer coefficient below the secondary-side collapsed liquid level was weakly dependent on the secondary side liquid level and the core power. The measured minimum heat transfer coefficient was 1.7 kW/m 2K for the full secondary side mass inventory.