Emergency Core Cooling System Research Articles

VIKTORIA experiments in the frame of R&D project on sump filtration during a loss of coolant accident

During a Loss of Coolant Accident (LOCA), in PWR’s, water is injected by the Emergency Core Cooling System (ECCS) to ensure the long-term core coolability and by the Containment Spray System (CSS) to remove residual heat and to maintain containment integrity. After the drainage of the Refueling Water Storage Tank (RWST), water is taken from sumps in the lower part of the reactor building. A filtering system is implemented to collect debris produced by the pipe break as well as other latent materials, such as fiberglass, paint and concrete particles, and to minimize the amount of debris entering in the ECCS and CSS systems. IRSN has launched an experimental R&D project investigating the clogging of sump filters by integral tests performed in the VIKTORIA loop. The main objectives are to investigate the head loss of the filter (physical and chemical clogging) for prototypic upstream debris source term in relevant thermal hydraulic conditions (water temperature and flow velocity on the filter surface) in compliance with the temperature profile and the chemistry of the water in sumps during a LOCA transient. For that, the VIKTORIA loop was equipped successively with two types of 2 m2 filters used in 900MWe NPP’s. The tests highlight the settling of the largest particles (concrete, painted chips) and part of the fibers; the transport of debris (roughly 55 to 65 % of the injected debris source term) leads to the physical clogging of the filter. The debris carried to the filter generate at 80 °C (with chemistry) a very quick increase of the pressure drop across the filter (≈7 kPa) that could be due to rapid chemical effects further to fibers corrosion. At the end of the 80 °C plateau, a pressure drop increase was observed due to the temperature decrease in agreement with the water viscosity evolution. The two types of filters (rectangular pockets or planar types) behave very differently with rather low head losses for the second type; ≈1.5 kPa. Nevertheless, downstream debris source term appears to be more significant with the planar filter (compared to the filter with pockets) during the first hour of injection before the fibrous debris bed formation. We have also observed a more stabilized “cake”, during the (7 days) medium term evolution compared to that for rectangular pockets filter due to motion of debris inside pockets. The recent experiments performed with less amount of fibers (by replacement of part of fibrous materials by RMI metallic insulation) led to significantly reduce the head loss without any consequences on the downstream behavior.

Validation and parametric study of FFRD model in DRACCAR code

The recent revision of Emergency Core Cooling System (ECCS) regulations in Korea has necessitated the consideration of Fuel Fragmentation, Relocation, and Dispersal (FFRD) phenomena in nuclear reactor safety analyses. Consequently, the Korea Atomic Energy Research Institute (KAERI) has developed and integrated the FFRD model into the domestically licensed safety analysis code, SPACE. Globally, US NRC’s FRAPTRAN and IRSN’s DRACCAR are available for evaluating FFRD phenomena. The FRAPTRAN code incorporates the FFR model, developed by Quantum Technology, while the fuel dispersal model is currently not included. In contrast, the DRACCAR code functions as an integrated analysis platform capable of modeling multi-dimensional thermal, hydraulic, mechanical, and chemical phenomena during a Loss of Coolant Accident (LOCA). This study conducts a thorough examination of the FFRD model in the DRACCAR code and validates its applicability through analysis using the Halden IFA-650 tests. The results demonstrate satisfactory predictive capabilities. Furthermore, a parametric study of key FFRD model parameters enhances the understanding of the FFRD model in the DRACCAR code. The development of detailed physical models in the FFRD model could significantly enhance the performance of the DRACCAR code, warranting the establishment of a comprehensive framework for these advancements. In the future, code-to-code comparisons between the DRACCAR code and other domestically developed integrated analysis platforms will be conducted to investigate various phenomena in depth.

Analysis of the hydrogen safety of the NPP with a VVER-1200/V-491 reactor in a severe beyond design based accident

The results of a computational analysis of possible modes of combustion of a hydrogen-air mixture in containment during the evolution of a severe beyond design basis accident (BDBA) with the COCOSYS code are presented. The containment of the NPP with a VVER-1200/V-491 reactor was selected as the object of study. BDBA with loss of coolant occurs due to a break of the injection pipeline of the pressurizer system (LOCA DN179) and the simultaneous failure of all active channels of the emergency core cooling system (ECCS). The calculated parameters of the thermodynamic state (pressure and temperature) of the gas mixture are given, and the values of the concentration distribution of hydrogen inside the containment are presented. After calculating the hydrogen distribution and mixing in all compartments of the containment, an evaluation was made of the flammability of the mixture and the potential for self-ignition in the containment by using a three-component Shapiro-Moffette diagram. It was concluded that in BDBA (LOCA DN179 with failure of active ECCS) detonation of the hydrogen-containing mixture is excluded, and deflagration is possible only in the emergency compartment of steam generators with a pipeline break. Thus, the hydrogen risk mitigation has been achieved in accordance with the standards established by the Belarusian regulator, provided that the localization safety systems are operational in the event of hydrogen deflagration. And the efficiency of the hydrogen removal system from the containment using catalytic recombination is considered sufficient for the considered BDBA.

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.

Dispersal of high-burnup fuel fragment surrogate particles during and after loss-of-coolant accident tests

The issue of fuel fragmentation, relocation, and dispersal is critical in the licensing and use of high-burnup (>62GWd/MTU) nuclear fuel in light water reactors (LWRs). In this work, two test series are reported that examine the fragment dispersal during a burst event and an additional dispersal following the burst due to vibrations in the rod, such as those induced by accident recovery systems. To examine dispersal during the balloon and burst portion of a Loss-of-Coolant Accident event, HfO2 fragments and yttria-stabilized zirconia pellets were filled into an as-fabricated cladding tube, which was then pressurized and subjected to loss-of-coolant accident testing in steam. Results from this test were found to be highly non-prototypic, and dispersal was both significantly more violent and significantly greater in magnitude than identified for actual fuel tests. These findings were attributed to the conservative (more dispersive) nature of the particles chosen for dispersal and to the details of the test conditions used that led to particularly wide bursts. The second set of tests examined dispersal following the burst when vibrations were induced in the rod, primarily via recovery activities such as Emergency Core Cooling System actuation leading to rapid water addition. Post-burst dispersal testing was performed by inducing sinusoidal oscillations with 2–25 nm peak-to-peak amplitude and 2–5 Hz frequencies in pre-burst rods that had been refilled with HfO2 fragments or high-burnup fragment surrogate mixture of HfO2 fragments and yttria-stabilized zirconia sands. Testing revealed that rods with large burst openings (7 mm wide in this work) led to unmitigated dispersal from above the burst zone but that smaller bursts (5 mm wide), although still much larger than the mean fragment size of 3 mm, led to effectively no dispersal because of interparticle locking. Additionally, mixture and moisture were found to impact the amount of dispersal: mixtures increased dispersal, and moisture drastically reduced it. The implications of these findings on likely dispersal from actual fuel are discussed.

CFD Applications to Pressurized Thermal Shock-Related Phenomena

In pressurized water reactor accident scenarios, the injection of water from the emergency core cooling system (ECCS) (ECC injection) might induce a pressurized thermal shock (PTS), affecting the reactor pressure vessel (RPV) integrity. Therefore, PTS is a vital research issue in reactor safety, and its analysis is essential for evaluating the integrity of RPVs, which determines the reactor life. The PTS analysis comprises a coupled analysis between thermal–hydraulic and structural analyses. The thermal–hydraulic approach is particularly crucial, and reliable computational fluid dynamic (CFD) simulations should play a vital role in the future because predicting the temperature gradient of the RPV wall requires data on the transient temperature distribution of the downcomer (DC). Since one-dimensional codes cannot predict the complex three-dimensional flow features during ECC injection, PTS is one reactor safety issue where CFD simulation can benefit from complement evaluations with thermal–hydraulic system analysis codes. This study reviewed from the viewpoint of the turbulence models most affecting PTS analysis based on papers published since 2010 on single- and two-phase flow CFD simulation for the experiment on PTS performed in the Rossendorf coolant mixing model (ROCOM), transient two-phase flow (TOPFLOW), upper plenum test facility (UPTF), and large-scale test facility (LSTF). The results revealed that in single-phase flow CFD simulation, where knowledge and experience are sufficient, various turbulence models have been considered, and many analyses using large eddy simulation (LES) have been reported. For two-phase flow analysis of air–water conditions, interface capturing/tracking methods were used in addition to two-fluid models. The standard k–ε and shear stress transport (SST) k–ω models were still in the validated phase, and various turbulence models have yet to be fully validated. In the two-phase flow analysis of steam–water conditions, many studies have used two-fluid models and Reynolds-averaged Navier-Stoke (RANS), and NEPTUNE_CFD, in particular, has been reported to show excellent prediction performance based on years of accumulated validation.

Analyses of the unavailability dynamics of emergency core cooling system

Abstract The Probabilistic Safety Assessment (PSA) studies can be used to evaluate the time-dependence of the safety systems availability. The purpose of this work is to analyze the unavailability dynamics of the Emergency Core Cooling System (ECCS). Using the PSA methodology, the ECCS was modeled. In order to show how the unavailability of the system varies as function of time (mission time, test interval and component ageing), sensitivity analyses are carried out. Following these analyses, polynomial functions of ECCS unavailability as function of time have been obtained. These functions allow to estimate the unavailability without modelling other ECCS fault trees being a simplified estimation of the system unavailability. In the same time, taking into account that the ECCS unavailability must be less than 10−2, the best time variations can be established to have the unavailability complied with this criterion. Also, the best time variations could be similarly established for any safety systems having the unavailability less than 10−3.

Evaluating direct vessel injection accident-event progression of AP1000 and key figures of merit to support the design and development of water-cooled small modular reactors

The passive safety systems (PSSs) within water-cooled reactors are meticulously engineered to function autonomously, requiring no external power source or manual intervention. They depend exclusively on inherent natural forces and the fundamental principles of reactor physics, such as gravity, natural convection, and phase changes, to manage, alleviate, and avert the release of radioactive materials into the environment during accident scenarios like a loss-of-coolant accident (LOCA). PSSs are already integrated into such operating commercial reactors as the Advanced Pressurized Reactor-1000 MWe (AP1000) and the Water-Water Energetic Reactor-1200 MWe (WWER-1200) are adopted in most of the upcoming small modular reactor (SMR) designs. Examples of water-cooled SMR PSSs are the passive emergency core-cooling system (ECCS), passive containment cooling system (PCCS), and passive decay-heat removal system, the designs of which vary based on reactor system-design requirements. However, understanding the accident-event progression and phases of a LOCA is pivotal for adopting a specific PSS for a new SMR design. This study covers the accident-event progression for direct vessel injection (DVI) small-break loss-of-coolant accident (SB-LOCA), associated physics phenomena, knowledge gaps, and important figures of merit (FOMs) that may need to be evaluated and assessed to validate thermal-hydraulics models with an available experimental dataset to support new SMR design and development.

Post-quench ductility limits of coated ATF with various zirconium-based alloys and coating designs

Post Quench Ductility (PQD) of various Cr-coated cladding designs with respect to base cladding materials, Cr coating methods, and its thickness were investigated to explore Emergency Core Cooling System (ECCS) limits. Both inner wall and outer wall of coated (Cr and Cr/CrN) claddings were steam oxidized at ∼1204 °C and then water quenched. The post-oxidized specimens were ring compressed to attain offset strains based on which ductility was assessed in compliance with the U.S. Nuclear Regulatory Commission (U.S. NRC)’s protocols. Weight gain-based Equivalent Cladding Reacted (ECR) limits of coated specimens (Cr and Cr/CrN) are consistently lower than those of the base cladding materials due to the early load drop under Ring Compression Test (RCT). Yet, time needed to reach the ECR limit is still increased for coated specimens because it effectively undergoes single side oxidation with protective coating. Cracks imitated from ZrCr2 are primarily responsible for the early major load drop observed for tested coating thicknesses (8.9 µm and 18.9 µm), and hence reduced ECR limits compared to bare Zircaloy. The cracks from ZrCr2 was promoted by an increased oxygen concentration of the interfacing Zr matrix owing to the diffusion of oxygen through the coating. The thicker coating reduces this oxygen level, thereby delaying the load drop from ZrCr2 during RCT and increases the ECR limit. The maximum attainable ECR limit of coated ECR limits is affected by the base Zircaloy as it gives as the ECR ceiling from which ECR reduction for coated cladding occurs. For the tested base cladding materials (Opt. ZIRLO™ and HANA-6), ECR 19 %, which is the limit for Cr-coated (8.9 µm) HANA-6, may serve as the lower envelope limit. Hence, ECR 19 % (single-side ECR) can conservatively serve as the conservative, yet non-design specific, Cr-coated ECR limit for the most of the modern base Zircaloy materials (Opt. ZIRLO™ and HANA-6).

Fuel integrity evaluation for the shutdown BWR-4 plant during LOCA event

Two BWR units in the Chinshan stopped operation since 2017. However, the spent fuel assemblies still reside in the reactor core, and the plant staffs defined this configuration as the Mode 5 condition. If a Loss-of-Coolant Accident (LOCA) occurs in this condition, the water level will drop and then maintain at approximately 2/3 of the height of the reactor core. The upper 1/3 of the fuel will be exposed to air, and then the fuel cladding temperature will increase significantly. This study use RELAP5 to investigate the effect of different Emergency Core Cooling System (ECCS) to mitigate the LOCA. It is found that the core spray (CS) system can sufficiently suppress the fuel temperature. However, the temperature increase of the uncovered fuel becomes larger if the low pressure coolant injection (LPCI) system which does not provide direct core injection is operable. Based on the results, the peak cladding temperature of both cases are still below the acceptance criteria of 600 °C.

Application of REPAS to analyze the sump clogging issue following a LOCA and its impact on the reliability of the ECCS long-term core cooling function

Sump clogging has been identified as a relevant issue after an accident occurred in the Barsebäck-2 nuclear power plant in Sweden (1992). Following a steam line Loss of Coolant Accident (LOCA) due to the inadvertent opening of a safety relief valve, the jet stripped some insulation material from nearby pipes. The insulation debris were transported to the inlet of the strainers for the drywell spray system, thus clogging the intake. The accident was not serious but showed that the Emergency Core Cooling System (ECCS) could have failed. As a consequence, several actions have been undertaken by international organizations, regulatory bodies and nuclear power plant owners to characterize this issue, and propose solutions and improvements. In the present paper, the Reliability Evaluation of Passive Safety Systems (REPAS) methodology is applied to analyze the sump clogging issue and its effect on the long-term core cooling function. Originally developed to evaluate the reliability of passive systems, REPAS is here applied for the first time to an active system. The application is performed using the TRAC/RELAP Advanced Computational Engine (TRACE) code, developed by the USNRC, to simulate a generic three-loop PWR, with an active decay heat removal system.

Two-phase CFD simulations of COSI tests and evaluation of condensation distribution

This study presents the two-phase CFD simulation of experiments carried out on COSI facility in the frame of Pressurized Thermal Shock (PTS). PTS may occur during a cold-water injection, by the Emergency Core Cooling System (ECCS) during a Loss-Of-Coolant Accident (LOCA). The simulations were carried out using the Neptune_CFD code. A sensitivity study on Direct Contact Condensation (DCC) models was performed with a comparison with experimental data. At the end of this study, it appears that the turbulence of the free surface is rather low. Consequently, a low-turbulence heat transfer model is more appropriate. Then, two sensitivity studies on global and local mesh refinement are carried out, with the creation of an optimized mesh. This optimized mesh is used with the low-turbulence model to simulate the 40 test cases with no water level and co-current steam flow. Deviations in temperature and total condensation rate are less than 5% and 25% respectively. Finally, the condensation rate between stratified and non-stratified regions is evaluated. Condensation tends to occur in non-stratified zones, with imbalance that can be very high, reaching almost 80% of total condensation.

Analysis of thermal hydraulic system code LOCUST V1.2 with ECC thermal mixing test facility

The safety injection (SI) system might be put into use, and the coolant at high temperature will mix with the supcooled water under Loss Of Coolant Accident (LOCA). The thermo-hydraulic phenomena associated with the thermal mixing of coolant and supcooled water will directly affect the judgment on core reflooding. China General Nuclear Power Group (CGN) independently developed and designed a thermal hydraulic system code named LOCUST, and the verification of thermal mixing models in LOCUST is necessary. This paper will introduce the thermal mixing tests at the T-junction based on Emergency Core Cooling System in Xi'an Jiaotong University (ECCS-XJTU) experimental facility, which were mainly conducted for the mixing between subcooled water injected from the SI pipe with a range of 25–125 kg/h in mass flow and the pure steam in the primary pipe with a range of 100–500 kg/h in mass flow. Besides, the fluid temperature and the dynamic vapor quality after mixing in tests are analyzed, and the simulations of 25 thermal mixing tests using the LOCUST 1.2 code are performed. The results show that the maximum relative error of LOCUST in mass flow of liquid is within 13.8 %, and the maximum relative error of LOCUST in temperature is within 8 %, which validates the reliability and accuracy of simulations of LOCUST for two-phase thermal mixing in LOCA.

Effect evaluation on performance issues of passive safety system– Part Ⅱ. Passive emergency core cooling system

In this study, the effect of performance issues on the Passive Emergency Core Cooling System (PECCS), a type of Passive Safety System (PSS), was evaluated. The PECCS relies on natural phenomena such as gravity, density difference, and pressure difference to operate. Its driving force is lower compared to that of the active safety system, making it necessary to evaluate its performance in various transient situations. Major performance issues were identified through effect evaluation using a conceptual problem for the PECCS, and their effects on the system were analyzed using the system analysis code MARS-KS. Therefore, there are concerns about whether the PSS can sufficiently perform safety functions in various situations. In Korea, performance evaluation methodology for PSSs is being developed to dispel these concerns. As part of the methodology development, this study conducted the effect evaluation on performance issues. The study identified several major performance issues for the PECCS, including those related to pressure and heat loss of piping within the system, uncertainties in tank wall condensation modeling, and checks valve modeling. The evaluation of performance issues has led to the identification of the main factors to consider in PECCS modeling, which include: 1) the geometry and pressure drop of PECCS, 2) the occurrence of wall condensation in pipes and tanks, and 3) the behavior of check valves. Proper modeling of these factors is important in achieving accurate analysis results and preventing potential issues, such as premature check valve closure or incomplete coolant injection. In conclusion, the results of this study are expected to contribute to deriving safety analysis modeling guidelines for PSSs and design considerations for NPPs that utilize PSSs by improving the completeness of the developing performance evaluation methodology.

The DISNY facility for sub-cooled flow boiling performance analysis of CRUD deposited zirconium alloy cladding under pressurized water reactor condition: Design, construction, and operation

The CRUD on the fuel cladding under the pressurized water reactor (PWR) operating condition causes several issues. The CRUD can act as thermal resistance and increases the local cladding temperature which accelerate the corrosion process. The hideout of boron inside the CRUD results in axial offset anomaly and reduces the plant's shutdown margin. Recently, there are efforts to revise the acceptance criteria of emergency core cooling systems (ECCS), and additionally require the modeling of the thermal resistance effect of the CRUD during the performance analysis. There is an urgent need for the evaluation of the effect of the CRUD deposition on the cladding heat transfer under PWR operating conditions, but the experimental database is very limited. The experimental facility called DISNY was designed and constructed to analyze the CRUD-related multi-physical phenomena, and the performance analysis of the constructed DISNY facility was conducted. The thermal-hydraulic and water chemistry conditions to simulate the CRUD growth under PWR operating conditions were established. The design characteristics and feasibility of the DISNY facility were validated by the MARS-KS code analysis and separate performance tests. In the current study, detailed design features, design validation results, and future utilization plans of the proposed DISNY facility are presented.

Investigation of the source term and radiological consequences of small break LOCA with failure of the emergency core cooling system in Bushehr nuclear power plant

In the present study, the in-containment source term derivation, environmental dispersion and deposition as well as radiological consequences of Bushehr Nuclear Power Plant (BNPP) due to Small Break Loss of Coolant Accident (SB-LOCA) have been studied. First, the calculation of in-containment release and distribution of the source terms due to the severe accident of SB-LOCA with failure of the active part of the Emergency Core Cooling System (ECCS) has been carried out. In this calculation, two scenarios, including with and without operator control actions, have been considered. Then, the environmental leakage of radionuclides as well as their dispersion and deposition considering two meteorological scenarios have been investigated. Consequently, public exposure has been calculated for each meteorological scenario and the obtained results have been reported and discussed. For two accidents with the same meteorological conditions, at a distance of 400 m from the release point, the received dose in the first 24 h is 18 and 342 mSv without and with operator action, respectively. The results show that due to the intervention of the operator and the use of emergency cooling supply systems, more water enters the reactor. Therefore, if the operator's actions do not prevent the core from melting, more source terms have entered the containment after the core melts and worse conditions than the baseline scenario without operator action occurred.

Analysis of iron-chromium-aluminum samples exposed to accident conditions followed by quench in the QUENCH-19 experiment

The QUENCH-19 experiment was a first-of-its-kind full-bundle test simulating accident conditions followed by water quench on accident-tolerant fuel (ATF) cladding. A type of FeCrAl(Y) alloy, B136Y3, was developed at Oak Ridge National Laboratory and tested at the Karlsruhe Institute of Technology using Kanthal APM corner rods, a shroud, and Kanthal AF spacer grids. Testing conditions were similar to those in QUENCH-15—which tested ZIRLO cladding behavior—so that B136Y3 and ZIRLO cladding could be compared. QUENCH-19 consisted of an initial pre-oxidation heating followed by a transient. Then, a maximum power hold, which was not present in QUENCH-15, was executed to extend the heating period for the FeCrAl(Y) rods. Finally, a rapid water quench was executed that was similar to emergency core coolant system (ECCS) actuation. Compared with the ZIRLO rods in QUENCH-15, the bundle in QUENCH-19 released significantly less H2 (9.2 g vs. 47.6 g) and achieved a much lower maximum temperature (1455°C vs. 1880°C). Furthermore, no breakaway oxidation was observed in QUENCH-19. Metallographic mounts revealed that despite the symmetry of the setup, at elevations near the maximum temperature, cladding and thermocouples were heavily damaged, substantial melting and oxidation occurred, and the cladding underwent chemical interaction with the thermocouple sheaths. Additionally, the ZrO2 spacers detrimentally interacted with the cladding, leading to mixed oxide debris and the full destruction of some rods. Additional failure was found in certain cooler rods that may have risen due to the high thermal expansion coefficient of FeCrAl alloys. This paper presents an analysis of this work, which suggests that FeCrAl cladding can chemically survive anticipated loss-of-coolant accident events followed by rapid ECCS quench if the correct geometry and core design are present.

Assessment of spray system capabilities in reduction of containment pressure and deposition of fission products during a severe accident in VVER-1000

In this study, we have investigated the spray system's capabilities to reduce fission products in the containment atmosphere and control the containment pressure in a VVER-1000 nuclear power plant. Large Break Loss of Coolant Accident (LB-LOCA) without access to the active part of the Emergency Core Cooling System (ECCS), as a worst-case severe accident, has been investigated as a case study. Only containment is simulated via the MELCOR code, while leakage of various materials from the nuclear power plant's core and the primary circuit is considered as the boundary conditions. Following the Fukushima accident in 2011, many nuclear safety regulatory organizations focused on using portable equipment that is essential for decay heat removal in nuclear power plants during severe accidents. Based on the literature studies, the accident management scenarios have changed recently since severe accidents inside nuclear power plants have worried many scientists and the public. Evaluating the current spray system in a severe condition based on which this system was not designed is vital and novel. We considered that the spray system would be fully restored and perform its safety function using the existing portable equipment to depressurize the containment and deposition of the fission products. To calculate the maximum allowed delay time of the spray system recovery, the accident was modeled without the operation of the spray system while a safety criterion was considered to avoid reaching the ultimate pressure of containment. For a more detailed evaluation of the spray system, two different operating modes including the activation of one spray train and the activation of both spray trains have been selected. According to the obtained results, if the spray system is revived before 4,000s from the beginning of the accident, this system would be able to reasonably mitigate the accident consequences. Moreover, without the need for any external water supply source for the spray system, this system can be permanently fed from the containment sump for at least 800,000s after the initiation of the accident, requiring the relevant heat exchanger of the spray system to be recovered for this period.

Proposal of quantification method of dynamic system reliability model of digital RPS using Markov state-transition model

ABSTRACT In safety systems of the latest nuclear power plants such as the Advanced Boiling Water Reactor (ABWR) and the latest Pressurized Water Reactor (PWR), digital circuits are used for start-up signals of a reactor protection system (RPS) and emergency core cooling systems (ECCS). And several studies have been conducted to develop the method to build a system reliability model of RPS to be combined with probabilistic risk assessment (PRA) so far. However, these studies have not been sufficiently progressed in terms of the quantitative evaluation and coupling with the conventional PRA method. Despite of the efforts of Task group of Digital system reliability failure mode taxonomy (DIGREL, later becoming Working Group on Digital Instrumentation and Control [WGDIC]) of Organization for Economic Co-operation and Development (OECD)/Nuclear Energy Agency (NEA) in this field, the situation has not changed sufficiently. From these backgrounds, this paper presents a rational quantitative evaluation method of the dynamic system reliability model of the digital RPS which can represent time-dependency and is consistent with the conventional PRA method.

Large-eddy simulation on two-liquid mixing in the horizontal leg and downcomer (the TAMU-CFD Benchmark), with respect to fluctuation behavior of liquid concentration

Pressurized Thermal Shock (PTS) is induced potentially by the rapid cooling of the cold-leg and downcomer wall in the primary system of a Pressurized Water Reactor (PWR) due to the initiation of Emergency Core Cooling System (ECCS). Thus, fluids mixing in a horizontal cold-leg and downcomer should be predicted accurately. However, turbulence production and damping often hinders this prediction due to the presence of the density gradients. Hence, the Fifth International Benchmark Exercise, “the cold-leg mixing Computational Fluid Dynamics (CFD) Benchmark,” was conducted under the support of OECD/NEA. The experiment was designed for visualization of the mixing phenomena of two liquids with different densities. The heavy liquid was a simulant of cold water from ECCS, in a horizontal leg and downcomer. We used the Large-eddy Simulation (LES) to investigate the time fluctuation behaviors of velocity and liquid concentration. The CFD simulation was performed with two turbulence models and three different numerical meshes. We investigated the characteristics of the appearance frequency of the heavy liquid concentration with the statistical method. Based on our findings, we propose further experiments and numerical investigations to understand the fluid mixing phenomena related to PTS.