Small Break Loss Of Coolant Accident Research Articles

CINEMA-SMART development and its application to SBLOCA analysis for Korean small integral reactor

The Korean small integral reactor of SMART (System-integrated Modular Advanced ReacTor) was developed, which contained core, reactor coolant pumps, steam generators and pressurizer within a single reactor pressure vessel. In the code development of severe accident sequence analysis for the SMART, a CINEMA (Code for INtegrated severe accidEnt Management Analysis)-SMART has been developing from an initiation event to a containment failure including fission product behavior in the vessel and the containment. The CINEMA-SMART is composed of CSPACE, SACAP (Severe Accident Containment Analysis Package), and SIRIUS (SImulation of Radioactive nuclide Interaction Under Severe accident), which are capable of core melt progression with thermal hydraulic analysis of the RCS (Reactor Coolant System), severe accident analysis of the containment, and fission product analysis, respectively. The main sequence of a SBLOCA (Small Break Loss Of Coolant Accident) for the SMART has been analyzed using CINEMA-SMART computer code. This analysis is focused on an evaluation of core melt progression and containment thermal hydraulic behavior during a severe accident. The CINEMA results have shown that the integrity of the containment can be maintained during the SBLOCA accident in the SMART.

Numerical study on the thermal-hydraulic behavior in the ultimate heat sink of passive residual heat removal system in the SMART

Computational Fluid Dynamics (CFD) analysis was performed to evaluate the thermal–hydraulic characteristics of emergency cooldown tank (ECT) in a Passive Residual Heat Removal System (PRHRS) of SMART. ECT is the ultimate heat sink of the PRHRS. Experimental results about the small-break loss-of-coolant accident (SBLOCA) scenario for a 0.4 in. line break in the safety-injection system (SIS) were used to compare and validate the CFD results. The geometrical model for CFD considered only the ECT and internal heat exchanger. The boundary and initial conditions of CFD analysis were based on the experimental data. The transient analysis was performed for 6000 s. The temperature distribution of the CFD results is in good agreement with the experiment results. The water level in ECT increases due to the volume expansion according to the temperature, which was well simulated in CFD analysis.

Assessment of the core-catcher in the VVER-1000 reactor containment under various severe accidents

The core catcher is used as a passive safety system in new generation nuclear power plants to create a space in the containment for the placing and cooling of the molten corium under various severe accidents. This research investigates the role of the core catcher in the VVER-1000 reactor containment system in mitigating the effects of core meltdown under various severe accidents within the context of the Ex-vessel Melt Retention (EVMR) strategy. Hence, a comparison study of three severe accidents is conducted, including Station Black-Out (SBO), SBO combined with the Large Break Loss of Coolant Accident (LB-LOCA), and SBO combined with the Small Break Loss of Coolant Accident (SB-LOCA). Numerical comparative simulations are performed for the aforementioned scenario with and without the EX-vessel core-catcher. The results showed that considering the EX-Vessel core catcher reduces the amount of hydrogen by about 18.2 percent in the case of SBO + LB-LOCA, and hydrogen production decreases by 12.4 percent in the case of SBO + SB-LOCA. Furthermore, in the presence of an EX-Vessel core-catcher, the production of gases such as CO and CO2 for the SBO accident is negligible. It was revealed that the greatest decrease in pressure and temperature of the containment is related to the SBO accident.

RELAP5 validation for core makeup tank analysis based on separate effect tests

The RELAP5 code has been developed for best-estimate transient simulation of light water reactor coolant systems during accident and has been widely used in the field of nuclear thermal-hydraulics and safety analysis. The core make-up tank (CMT) is a crucial device of the passive core cooling system in the AP series pressurized water reactor nuclear power plants. Since the CMT not only provides full-pressure injection but also triggers auto depressurization system by its level signal during small-break loss of coolant accident. Therefore, it is necessary to carry out CMT separate effect tests to verify the feasibility and accuracy of RELAP5/MOD 3.4 to simulate CMT. The previous studies found that the lower version of the RELAP5 program has problems in simulating CMT, and the applicability of the latest RELAP5/MOD3.4 has not been evaluated. In this paper, a CMT separate effect test facility were presented and then RELAP5/MOD 3.4 was used to establish a CMT test bench model to evaluate the applicability, and obtain the general rules and methods of RELAP5/MOD 3.4 to simulate CMT. It’s found that RELAP5/MOD 3.4 can accurately simulate most of the CMT operating conditions. However, large uncertainty is expected in the simulation of delayed time and the flowrate of delayed drainage stage is not affected by the heat transfer coefficient of steam direct contact condensation. The simulation of delay time depends on experimental data and the CMT delayed drainage experiment cannot verify the accuracy of the RELAP5/MOD 3.4 direct contact condensation model.

The scaling effect analysis of counterpart test between LSTF and ATLAS IET facilities on 1% top head break LOCA

ATLAS B5.1 test is a 1% small break loss of coolant accident (SBLOCA) at the top of the reactor pressure vessel with a total failure of a high-pressure injection (HPI) system. It was conducted as a counterpart test of LSTF SB-PV-07. The LSTF is scaled from a three-loop pressurized water reactor (PWR) Westinghouse design, while the ATLAS facility is scaled from APR1400. Thus, to determine B5.1 test conditions, a scaling analysis has been conducted between the two test facilities of LSTF and ATLAS. We reviewed the scaling method used for the ATLAS facility and found the initial and boundary condition for the counterpart test. Then, we compared the experimental data of the ATLAS B5.1 test against that of LSTF SB-PV-07. To reduce the scaling distortion between two IET facilities, we modified the RELAP5 input of the ATLAS facility toward the LSTF facility considering the elevation difference and design difference. The internal design difference between the ALTAS and LSTF facilities generated a significant difference in the plant transient beyond the scaling distortion.

Separate-effect drainage tests in a core make-up tank

Passive safety systems are widely used in the advanced nuclear reactors, such as AP1000 and CAP1400. As an important component in the passive safety system, the core make-up tank (CMT) plays a key role for the safety injection and the core decay heat removal. Although the structure of the CMT system is simple, it involves many thermal-hydraulic phenomena, most of which significantly impact the peak temperature of the fuel cladding and the liquid level of the core. Since the CMT not only provides full-pressure injection but also triggers the auto depressurization system by its liquid level signal during small-break loss of coolant accident (SBLOCA), its thermal hydraulic behaviors significantly affect the performance of the passive core cooling system (PXS). Therefore, separate effect tests are inevitable to verify the feasibility of CMT applied to passive safety and injection systems and to study its operating characteristics. In this work, a separate-effect CMT test facility was designed and constructed. Experiments were performed to investigate the CMT's thermal hydraulic characteristics under the effects of system pressure, steam condensation, and flow friction. The study found that the delay drainage phenomenon is the result of a pressure balance between the pipe line and the gravity head. Whether delayed drainage occurs depends on the degree of subcooling of the fluid at the top of the CMT. The flowrate is not affected by the heat transfer coefficient of steam direct contact condensation. The flow friction of the drainage pipeline has no effect on the delayed drainage process, but has a great influence on the gravity drainage process.

Experimental study of the transient characteristics of a natural circulation loop under SBLOCA

Abstract In order to understand the influences of various factors on the depressurization rate under small break LOCA occurring at the upper plenum in a natural circulation loop, a basic analysis has been carried out. It is shown that the depressurization rate is influenced by system pressure, back pressure, break area, water mass in the system, heating power, operational state of the second loop, and so on. Experimental results of flow and heat transfer are also presented in this paper. The experiment was carried out at the 5 MW nuclear heating reactor test loop HRTL-5. The transient processes are described and heat transfer coefficients are estimated. The results show that flashing improves the natural circulation and the heat transfer, so the wall temperature of the heated rod can be kept at a lower level and the reactor is safe. The flow oscillation, which leads to cyclical variations of important parameters, is described and its development mechanisms are studied. It is pointed out that the interaction of pressure decreasing and bubbles producing caused by flashing is the main reason of the flow oscillation.

CFD investigation of flow reversal in inverted U-tube steam generator under two-phase natural circulation

In this paper, a critical mass flow rate (CMFR) has been obtained for three different steam generators (i.e. two marine-type and one commercial) at different inlet void fractions using the CFD method to study the sustainability of natural circulation (NC) due to depressurization in two-phase NC mode. Because of depressurization, the transients in the inlet void fraction have been considered as multiple steady-state inputs, and the range of safe operating mass flow rate has been obtained. The results show that the characteristics curves shift with inlet void fraction thus limiting the operating mass flow rate range between 0.046 kg/s to 0.050 kg/s for M_SG1 and 0.051 kg/s to 0.055 kg/s for M_SG2. The effect of pipe roughness on CMFR has also been studied. The results can be used for the optimized design of the U-tube steam generator (UTSG) that will safely perform the heat removal operation during a small break loss of coolant accident (SBLOCA).

The Huge Missing Factor in Leak-Before-Break Analysis: How a Circumferential Through-Wall-Crack in a Pipe System Changes the Flexibility and Reduces the Applied Moments

Abstract In typical leak-before-break (LBB) analyses in the nuclear industry, the uncracked piping normal operating forces and moments are applied in a cracked-pipe analytical procedure to determine normal leakage, and the combined forces and moments under normal operating condition and safe shutdown earthquake seismic loading are used in a fracture analysis to predict margins on “failure.” The International Piping Integrity Research Program (IPIRG) performed in 1990–1998 provided some insights to typical LBB behaviors where pipe system tests were conducted with simulated seismic loadings. The test results showed a large margin on LBB, which was also recognized in 2011 when the Argentinian Atucha II plant was analyzed using a robust full FE model. It was found that when circumferential through-wall cracks were put in the highest stressed locations, the applied moment dropped for both normal operating and N + safe shutdown earthquake (SSE) loading as the crack length increased. The through-wall crack size for causing a double-ended guillotine break (DEGB) was greater than 90% of the circumference. Similar results were also found for a petrochemical pipe system where thermal expansion stresses are much higher than the primary stresses. Even with very low toughness materials, the critical crack size leading to DEGB was greater than 80% of the circumference. The implication of this work is that pragmatically there is much higher margin for DEGB failure in nuclear plant operation, and efforts would be better focused on the potential for a small-break loss-of-coolant accident (SB-LOCA).

Numerical Analysis for Hydrogen Flame Acceleration during a Severe Accident Initiated by SBLOCA in the APR1400 Containment

We performed a hydrogen combustion analysis in the Advanced Power Reactor 1400 MWe (APR1400) containment during a severe accident initiated by a small break loss of coolant accident (SBLOCA) which occurred at a lower part of the cold leg using a multi-dimensional hydrogen analysis system (MHAS) to confirm the integrity of the APR1400 containment. The MHAS was developed by combining MAAP, GASFLOW, and COM3D to simulate hydrogen release, distribution and combustion in the containment of a nuclear power plant during the severe accidents in the containment of a nuclear power reactor. The calculated peak pressure due to the flame acceleration by the COM3D, using the GASFLOW results as an initial condition of the hydrogen distribution, was approximately 555 kPa, which is lower than the fracture pressure 1223 kPa of the APR1400 containment. To induce a higher peak pressure resulted from a strong flame acceleration in the containment, we intentionally assumed several things in developing an accident scenario of the SBLOCA. Therefore, we may judge that the integrity of the APR1400 containment can be maintained even though the hydrogen combustion occurs during the severe accident initiated by the SBLOCA.

Sensitivity analysis of the IAEA SPE-2 small-break LOCA experiment with RELAP5, TRACE and APROS system codes

The Hungarian PMK-2 test facility is a scaled-down model of the VVER-440/213 type pressurized water reactors of Paks NPP, on which several measurement series have been carried out since the 80s. One of these series include four Standard Problem Exercises (SPE), primarily small-break loss-of coolant accident (SBLOCA) transients, performed in cooperation with IAEA. New RELAP5, TRACE and APROS models have already been developed and a comparative analysis of the results has been published recently for the PMK SPE-4 experiment. In that SBLOCA scenario, no high-pressure injection system (HPIS) was available, but a secondary side bleed and feed operation has been performed during the transient. This paper is a continuation of the previous one and presents the required model modifications and the analysis of the second SPE. The initiating event, similarly to the SPE-4 experiment, is a 7.4% break located at the top of the downcomer, while the main differences are that one line of the HPIS is available and no secondary side feed and bleed operation is performed in this case. One of the main objectives of the experiment was to verify that these safety measures are enough to prevent core dry-out. In addition to the investigation of this question, a sensitivity study was also carried out with respect to selected parameters that might have a significant impact on the characteristic of the simulated processes. The assessment of the results has also been performed with two quantitative methods, namely the Fast Fourier Transform Based Method with Signal Mirroring (FFTBM-SM) and Stochastic Approximation Ratio Based Method (SARBM).

Assessment of Trace for Candu Applications Against Rd-14m Small-Break Loca Experiment

Assessment of Trace for Candu Applications Against Rd-14m Small-Break Loca Experiment

Steam condensation separate effect tests in core make-up tank

The core make-up tank (CMT) is a crucial device of the passive core cooling system in the Westinghouse AP-series pressurized water reactor nuclear power plants. In the prototype plant, the CMT is a vertical cylindrical vessel with hemispherical top and bottom end sockets. Although the structure of the CMT system is simple, it involves many thermal-hydraulic phenomena, most of which significantly impact the peak temperature of the fuel cladding and the liquid level of the core. Since the CMT not only provides full-pressure injection but also triggers the auto depressurization system by its liquid level signal during small-break loss of coolant accident (SBLOCA), its thermal hydraulic behaviors affect the performance of the passive core cooling system (PXS). Therefore, it is necessary to carry out separate effect tests to verify the feasibility of CMT applied to passive safety and injection systems and to study the operating characteristics of CMT. This article first presents the design of a CMT separate effect test facility and analysis of the operating characteristics of the CMT with the experimental data. The influence of pressure, non-condensable gas, wall condensation, and thermal resistance on the CMT behavior were investigated. The study found that the non-condensable gas would stratify in the CMT, which had negligible effect on the wall condensation, but could greatly reduce the gas-liquid interface condensation.

Verification of Uncertainty Distributions for Thermal-Hydraulic Mixing Parameters Used in KWU-MIX for Pressurized Thermal Shock

The ability to extend the operating life of a pressurized water reactor depends in part on the ability of the reactor pressure vessel to withstand thermal shock concurrent with significant pressure. If during the course of a small-break loss-of-coolant accident (SBLOCA), the primary-side pressure is reduced sufficiently, cold make-up water is supplied to the cold leg by the emergency core cooling system. If incomplete mixing occurs between the cold injected water and the hot water in the primary circuit, a stream of cool water flows along the bottom of the cold leg into the downcomer. There, the cool water forms a downward-flowing buoyant plume surrounded by the hot water in the downcomer. The time-dependent spatial distributions of the temperatures and heat transfer coefficients on the inside surface of the reactor pressure vessel are important in determining compliance with regulatory requirements. The simulation of the mixing in the cold leg and downcomer is typically performed with flow-mixing computer codes, most of which use either computational fluid dynamics techniques or mechanistic models. The computer code used for this work, called KWU-MIX, makes use of mechanistic models. In previous works, the uncertainties in parameters associated with the most important phenomena that contribute to the temperature distributions were quantified by comparing experimentally derived values of the parameters with values from the mechanistic models. In this work, those uncertainties are propagated through the flow-mixing code in order to quantify the uncertainty in the calculated temperature distributions. An example of the propagation of uncertainties is given for conditions typical of a SBLOCA. Random values from each of the uncertainty distributions for the parameters of all of the most important phenomena were selected for each of 100 simulations of the typical accident conditions. The results of the 100 simulations were analyzed statistically in order to quantify the best-estimate temperature distribution and its uncertainty. The resulting best-estimate temperature distribution and its uncertainty were compared with experimental data obtained in the Upper Plenum Test Facility at the same typical accident conditions. The results of the comparison show that the uncertainty in the calculated temperature distribution bounds the experimental values.

Study on NC in primary loop and reverse flow in SG during SBO combining with SBLOCA

Marine nuclear power plants can rely on Natural Circulation (NC) to remove the heat generated by the reactor under normal operation or accident condition, which is the key to keeping the reactor safe. Based on the steady-state operation of single-phase NC and Forced Circulation (FC), the characteristics analysis of NC in primary loop and the reverse flow in the Steam Generator (SG) are analyzed employed by RELAP5/MOD3.3 code under the conditions of the Station Blackout (SBO) accident overlapped with 5-mm Small Break Loss of Coolant Accident (SBLOCA) in the cold leg happened and the Passive Residual Heat Removal System (PRHRS) of the secondary side operated. The results show that, with the development of the accidents, the loops A and B of the SG would go through the flow modes of the single-phase NC, the two-phase NC, and the reflux condensation. The occurrence of the three flow modes and their durations are dramatically different in the two loops for the influence of the pressurizer. The residual heat removal efficiency during the single- or two-phase NC is greater than that during the reflux condensation. The mass flow rate of the primary loop increases modestly, as the single-phase NC converting to the two-phase NC. During the NCs, the mass flow rate of the obverse flow in the SG is greater than that of the reverse flow, however, during the reflux condensation, the two mass flow rate are expected to be very close and the dynamic flow instabilities may appear in the primary loop.

Experimental Study on Sensitivity of PRHR Pipeline Break Location on ACME Test Facility

To study the safety performance of the advanced passive (AP) nuclear power plant (NPP) under passive system failure condition, the experimental study on the loss of coolant from the passive residual heat removal (PRHR) pipeline break is performed by the advanced core-cooling mechanism experiment (ACME) facility, during which the effect of main test sequences and break location on the key parameters in different phases of the accident is analyzed. As demonstrated by the study results, the ACME PRHR pipeline break test sequences, basically same as those for the small-break loss of coolant accident (SBLOCA) of the cold leg, reproduce the safety characteristics in the natural circulation phase of the passive NPP, blowout phase of the automatic depressurization system (ADS) and safety injection (SI) phase of the in-containment refueling water storage tank (IRWST); in the test at different break locations, the passive core cooling system (PXS) can always ensure the water makeup for core and the core active region remains below the mixed liquid level; and the break locations have notable effect on the ACME LOCA accident sequence, initial depressurization rate of reactor coolant system (RCS), PRHR heat exchanger (HX) flow, blowout flow, core level, IRWST SI flow and other parameters, and have little impact on the SI flow of the core makeup tank (CMT) and accumulator (ACC).

Small break loss of coolant accident of 200 cm² in cold leg of primary loop of Angra2 nuclear power reactor evaluation

The aim of this paper is evaluated the consequences to ANGRA 2 nuclear power reactor and to identify the flow regimes, the heat transfer modes, and the correlations used by RELAP5/MOD3.2.gama code in ANGRA 2 during the Small-Break Loss-of-Coolant Accident (SBLOCA) with a 200cm2 of rupture area in the cold leg of primary loop. The Chapter 15 of the Final Safety Analysis Report of ANGRA 2 (FSAR-A2) reports this specific kind of accident. The results from this work demonstrated the several flow regimes and heat transfer modes that can be present in the core of ANGRA 2 during the postulated accident. The results obtained for ANGRA 2 nuclear reactor core during the postulated accident were satisfactory when compared with the FSAR-A2. Additionally, the results showed the correct actuation of the ECCS guaranteeing the integrity of the reactor core.

Study on estimation of break diameter in the small break loss of coolant accident based on NSGA-II

The break diameter estimation in the small break loss of coolant accident (SBLOCA) is an important work in the field of nuclear safety. It can help the operators to judge the severity of the SBLOCA in time and play an important role in reducing the damage caused by the accident. In this study, a break diameter estimation model for the blowdown process of the SBLOCA is built based on the NSGA-II and the blowdown model of the pressurizer in the SBLOCA. The model uses monitoring data such as liquid level, temperature and pressure of the pressurizer during the blowdown process to calculate the break diameter in the SBLOCA. It can be analyzed from the results that the break diameter estimation model can estimate the break diameters accurately and quickly in the SBLOCA, the errors between the calculating results of the break diameter estimation model and the actual results are no more than 18.9%, and the calculation time of the model is less than 310 s. When the calculation process reaches the maximum generation of the model, the values of fitness functions in the model approaches the minimum values.

An innovation idea for SBLOCA mitigation strategy: Self-coagulation system in comparison with human hemostatic mechanism

The loss of coolant accident (LOCA), as one of the design basis accidents (DBAs), is a hypothetical accident that is usually considered in the design of nuclear power plant. LOCA is caused by small/large breaks in the reactor primary coolant and pressure boundary and may result in a loss of reactor coolant at a rate in excess of the reactor makeup system capability. Currently, the typical mitigation measures for LOCA are through the vessel high/low pressure coolant injection (HPCI/LPCI), core spray system, or reactor core isolation cooling (RCIC) system, to protect reactor from fuel melting and core degradation. However, is it possible to directly repair the small breaks passively? Inspired by the hemostasis and coagulation mechanism of human blood vessels, the authors are considering to develop a technology which would allow the reactor self-coagulation during the small break loss of coolant accident (SBLOCA). With the assistance of this innovative technology, the speed of losing coolant through reactor primary loop breaks could be significantly reduced or even the loss is stopped. With more and longer existing coolant in the core, the possibility of core degradation will also be significantly reduced. This innovative technology needs to be designed feasible and simple, which can be directly used into the existing nuclear power plants to improve the inherent safety of the cooling system, benefit the economy and safety of the nuclear power plant. In this paper, the authors compare the reactor primary loop of nuclear power plant with the blood circulation system of human beings from comprehensive and multi-angle perspectives, discuss the similarity, feasibility, and economy of nuclear power plant self-coagulation system based on the existing hemostasis and coagulation mechanism of human blood vessels. If possible, this technology will play a revolutionary positive role in alleviating DBAs related to SBLOCA.

A nodal sensitivity study of MELCOR simulation for severe accidents in a pressurized water reactor

This paper presents a sensitivity study of nodal scheme in MELCOR simulation of severe accidents in a pressurized water reactor, with the objective to estimate the nodal effects on some in-vessel and ex-vessel processes and phenomena, including thermal–hydraulic response, core degradation and relocation, hydrogen generation, vessel failure, containment pressurization and venting, source term. For this purpose, three nodal schemes (i.e., coarse, medium and fine meshes) of the COR package of the MELCOR code are chosen to analyze two severe accident scenarios: small break loss of coolant accident (SBLOCA) and large break loss-of-coolant accident (LBLOCA), both combined with station blackout. The results show that the nodal schemes mainly affect the calculations of heat transfers from the core to coolant and heat structures, relatively affecting the core degradation and relocation to the lower head of the reactor pressure vessel. As for the consequences, the coarse mesh tends to predict slower core relocation progressions and a later failure of RPV lower head. Moreover, more hydrogen generation by cladding oxidation can be observed in the coarse mesh case. The nodal schemes have little impact on the estimation of in-containment source term. Meanwhile, the simulations with fine mesh may also provide more detailed distributions of corium masses and temperatures, as well as heat fluxes, affecting thermal and mechanic behavior of RPV lower head.