Loss Of Accident Research Articles

Cold Leg LBLOCA uncertainty analysis using TRACE/DAKOTA coupling

Safety analyses for nuclear power plants were carried out in the past using a conservative approach. With the increase of the phenomenological knowledge, through experimental data, and computational power, it became possible to adopt best estimate thermal- hydraulic system codes to perform deterministic safety analyses. However, some uncertainties are still present in the models, correlations, initial and boundary conditions, etc. Therefore, it is fundamental to quantify the uncertainty of calculation. This approach is called “Best Estimate Plus Uncertainty” (BEPU). Among the available uncertainty analysis methodologies, the probabilistic method to propagate input uncertainty is widely adopted. In the present study, an uncertainty analysis of a cold leg large break loss of coolant accident in a generic PWR-900 MWe has been developed and it has been carried out coupling the best estimate thermal- hydraulic system code TRACE and the uncertainty quantification tool DAKOTA in the SNAP environment/architecture.

Improvement and Validation of the System Analysis Model and Code for Heat-Pipe-Cooled Microreactor

Heat-pipe-cooled microreactors (HPMR) use a passive high-temperature alkali metal heat pipe to directly transfer the heat of solid core to the hot end of the intermediate heat exchanger or thermoelectric conversion device, thus avoiding a single point failure. To analyze and evaluate the transient safety characteristics of an HPMR system under accident conditions, such as heat pipe failure in the core or a loss of system heat sink and other accidents, a previously developed model for transient analysis of a heat-pipe-cooled space nuclear reactor power system (HPSR) was improved and validated in this study. The models improved mainly comprise: (1) An entire 2-D solid-core heat transfer model is established to analyze the accident conditions of core heat pipe failure and system heat sink loss. In this model, radial and axial Fourier heat conduction equations are used to divide the core into r-θ direction control volumes. The physical parameters of the material in the control volume are calculated according to the volume-weighted average. (2) By coupling the heat transfer limit model and the two-dimensional thermal resistance network model, the transient model of a heat pipe for HPMR system analysis is improved. (3) Conversion system models are established to simulate the system characteristics of the advanced HPMR concept, such as thermoelectric conversion, Stirling conversion, and the open Brayton conversion analysis model. Based on the improved models, the HPMR system analysis program TAPIRSD was developed, which was verified by experimental data of the separated conversion components and the ground nuclear test device KRUSTY. The maximum deviation of the power output predicted by the energy conversion model is less than 8%. The accident conditions of the KRUSTY tests, such as load change, core heat pipe failure, and heat sink loss accident, were studied by using TAPIRSD. The results show that the simulation results of the TAPIRSD code agree well with the experimental data of the KRUSTY prototype reactor. The maximum error between the TAPIRSD code prediction and the measured value of the core temperature under accident conditions is less than 10 K, and the maximum deviation is less than 2%. The results show that the developed code can predict the transient response process of the HPMR system well. At the same time, the accuracy and reliability of the improved model are proved. The TAPIRSD is suitable for system transient analysis of different types of HPMRs and provides an optional tool for the system safety characteristics analysis of HPMR.

Research on Diagnosis of Loss of Flow Accidents in Nuclear Reactor Coolant System

The reactor coolant system is one of the most important systems in a nuclear power plant. The safe and stable operation of the reactor coolant system is related to the safety of the nuclear power plant. With the rapid development of nuclear power technology, it is particularly important to ensure the safety of nuclear power plants. This paper takes the third-generation reactor AP1000 as the research object, firstly analyzes the typical accidents of the coolant system, and establishes a three-layer fault model, and then introduces the counterfactual reasoning method to take the Loss of Flow Accident as an example to identify the accident, and compares it with the traditional data-driven The Bayesian network diagnosis method was compared with the Bayesian network diagnosis method, and finally the accident was diagnosed by the simulation software Pctran. The results show that the method can accurately identify the fault and is better than the Bayesian method, which can provide decision support for accident handling and ensure the safety of the reactor.

Numerical analysis on in-core ignition and subsequent flame propagation to containment in OPR1000 under loss of coolant accident

Since Fukushima nuclear power plant (NPP) accident in 2011, the importance of research on various severe accident phenomena has been emphasized. Particularly, detailed analysis of combustion risk is necessary following the containment damage caused by combustion in the Fukushima accident. Many studies have been conducted to evaluate the risk of local hydrogen concentration increases and flame propagation using computational code. In particular, the potential for combustion by local hydrogen concentration in specific areas within the containment has been emphasized. In this study, the process of flame propagation generated inside a reactor core to containment during a loss of coolant accident (LOCA) was analyzed using MELCOR 2.1 code. Later in the LOCA scenario, it was expected that hydrogen combustion occurred inside the reactor core owing to oxygen inflow through the cold leg break area. The main driving force of the oxygen intrusion is the elevated containment pressure due to the molten corium–concrete interaction. The thermal and mechanical loads caused by the flame threaten the integrity of the containment. Additionally, the containment spray system effectiveness in this situation was evaluated because changes in pressure gradient and concentrations of flammable gases greatly affect the overall behavior of ignition and subsequent containment integrity.

Is It Possible to Improve Urinary Incontinence and Quality of Life in Female Patients? A Clinical Evaluation of the Efficacy of Top Flat Magnetic Stimulation Technology.

Background and Objectives: Urinary incontinence is the accidental loss of urine that can occur at any age, especially among women over 50; however, its prevalence is increasing. This study aimed to assess the efficacy and safety of a device that uses Top Flat Magnetic Stimulation for the management of women with mixed urinary incontinence. Materials and Methods: A total of 50 female patients underwent a total of six treatment sessions performed twice a week for three consecutive weeks, for 28 min. Three questionnaires were used for the evaluation of the urinary improvements: Urinary Incontinence Short Form (ICIQ-UI-SF), Incontinence Questionnaire Overactive Bladder Module (ICIQ-OAB), and Incontinence Impact Questionnaire-Short Form (IIQ-7). The questionnaires were compiled immediately before each treatment, within the treatment period (until the sixth treatment session), and up to 3 months of follow-up. Results: Improvement in quality of life is observed according to the results of the questionnaire, already after six sessions and at three months follow-up. ICIQ-UI-SF’s average score at baseline was 20.2 (±1.1), and it significantly declined (p < 0.01) to 1.8 (±2.4) after six sessions and to 4.8 (±2.8) at 3 months follow-up; ICIQ-OAB’s average score significantly decreased (p < 0.01) from 10.4 (±3.2) at baseline to 1.4 (±0.8) after six sessions and 2.3 (±1.6) at 3 months follow-up. IIQ-7’s average score significantly decreased (p < 0.01) from 20.2 (±1.7) at baseline to 0.4 (±0.5) after six sessions and to 4.7 (±2.8) at 3 months follow-up. Conclusions: Our results suggest that Top Flat Magnetic Stimulation technology can reduce mixed urinary incontinence symptoms for all women examined, leading to an improvement in the patient quality of life and patient awareness of their pelvic floor area with good results.

Simulation of the gamma dose rate in a loss of pool water accident of the second Egyptian research reactor ET-RR-2

Abstract The second Egyptian research reactor ET-RR-2, is a pool type reactor. A sudden loss of pool water would leave the core region uncovered. The reactor core is surrounded by chimney chambers with water isolated from the pool water. This accident would lead to significant external doses. A model is developed and used to calculate the dose rates for key access-areas and traffic plans from indirect line of sight of the core which have a maximum dose rate. The model developed uses the discrete ordinate method as implemented in the code DOT3.5.

Studies on the behaviour of a passive containment cooling system for the Indian Advanced Heavy Water Reactor

Abstract A passive containment cooling system has been proposed for the advanced heavy water reactor being designed in India. This is to provide long term cooling for the reactor containment following a loss of coolant accident. The system removes energy released into the containment through immersed condensers kept in a pool of water. An important aspect of immersed condenser’s working is the potential degradation of immersed condenser’s performance due to the presence of noncondensable gases. An experimental programme to investigate the passive containment cooling system behaviour and performance has been undertaken in a phased manner. In the first phase, system response tests were conducted on a small scale model to under­stand the phenomena involved. Tests were conducted with constant energy input rate and with varying energy input rate simulating decay heat. With constant energy input rate, pressures in volume V1 and V2 reached almost steady value. With varying energy input rate V1 pressure dropped below the pressure in V2. The system could efficiently purge air from V1 to V2. The paper deals with the details of the tests conducted and the results obtained.

Experimental studies on condensation of steam mixed with noncondensable gas inside the vertical tube in a pool filled with subcooled water

Abstract A passive containment cooling system with immersed condensers has been proposed as one of the alternatives for the Advanced Heavy Water Reactor (AHWR) being designed in India. The system removes residual/decay heat released into the containment through the immersed condensers kept in a pool of water following loss of coolant accident. An important aspect of the immersed condensers is the potential degradation of its performance due to the presence of noncondensable gases. Experiments are performed to obtain reliable data on condensation phenomena in presence of air. These experiments are conducted on full-scale tubes of condensers immersed in a pool of water maintaining similar conditions as in the prototype of AHWR. A method has been proposed for the determination of the local heat transfer rate using correlations given in literature. The parametric effects of air mass fraction, pressure, steam flow, etc. on condensation heat transfer in presence of noncondensable gas have been studied. The experimental results are compared with the correlations given in literature.

High-fidelity FE models for assessing progressive collapse robustness of RC ordinary moment frame (OMF) buildings

Most of the buildings in Saudi Arabia were constructed using reinforced concrete (RC) ordinary moment frames (OMFs), which, in most cases, had discontinuous bottom beam rebars across the beam-column joints. Consequently, they may be susceptible to the potential of progressive collapse owing to accidental loss of one (or more) exterior column(s) in an extreme event. Therefore, it is imperative to examine the progressive collapse robustness of RC OMF buildings. In this paper, 16 half-scale RC OMF assemblies (skeletal frames comprising of multi-span beams and two-story columns) were investigated using validated FE modeling for progressive collapse potential in the event of middle column removal. The analysis was carried out with the help of LS-DYNA software. It incorporated nonlinear constitutive modeling of concrete and steel rebars considering the effect of strain rate, and modeling of bond-slip behavior at the interaction of steel rebars with concrete. Out of the 16 assemblies, 12 specimens represented the actual detailing of an existing 8-story building. The studied variables included axial load on columns, percentage of continuous bottom beam rebars across the beam-column joint (with respect to the total number of bottom rebars), and number of beam spans. The performance of the 12 assemblies was evaluated in accordance with mode of failure and characteristics of load–displacement envelope. A simple approach was suggested to assess the progressive collapse risk level of RC OMFs. Five risk levels were proposed, ranging from very low to very high. Very high to high risk was predicted for the 12 studied assemblies. The last four assemblies of the analysis matrix were newly designed to reduce the progressive collapse risk to a moderate (or low) level.

Independent qualification of the CIAU tool based on the uncertainty estimate in the prediction of the LOBI test A1-93

Abstract The Code with – the capability of – Internal Assessment of Uncertainty (CIAU) is a tool proposed by the Dipartimento di Ingegneria Meccanica, Nucleare e della Produzione (DIMNP) of the University of Pisa. Other Institutions including the nuclear regulatory body from Brazil, “Comissäo Nacional de Energia Nuclear”, contributed to the development of the tool. The CIAU aims at providing the currently available Relap5/Mod3.2 system code with the integrated capability of performing not only relevant transient calculations but also the related estimates of uncertainty bands. The Uncertainty Methodology based on Accuracy Extrapolation (UMAE) is used to characterize the uncertainty in the prediction of system code calculations for light water reactors and is ‘internally’ coupled with the above system code. Following an overview of the CIAU development, the present paper deals with the independent qualification of the tool. The qualification test is performed by estimating the uncertainty bands that should envelope the prediction of the LOBI test Al-93 (small break loss of coolant accident). The current limitation of the ‘error’ database, implemented into the CIAU prevented a final demonstration of the qualification. However, all the steps for the qualification process are demonstrated.

Study on parameter importance ranking of the OECD/NEA PKL4 test i2.2 with efficient moment-independent sensitivity analysis

The intermedium break loss of coolant accident (IBLOCA) benchmark study was carried out on the OECD/NEA PKL-4 facility test i2.2 to explore the influence of break size on the peak cladding temperature (PCT). Results of the blind calculations performed by different participants showed that the transient behaviors of the scenario can be well predicted, but further improvements were needed for some local phenomena. In this paper, the global sensitivity analysis was carried out to investigate the influences of 36 input parameters on the PCT and minimum core water level (MCWL) with an efficient moment-independent method, and the sensitivity measures as well as corresponding importance rankings were obtained. The obtained results can deepen the understandings of the key phenomena in the IBLOCA test, and provide guidance to the model optimization and code verification. In addition, the adopted sensitivity analysis method can be extended to other tests of the PKL facility.

Inherent Safety Characteristics of Lead Bismuth Eutectic-Cooled Accelerator Driven Subcritical Systems

An accelerator driven subcritical system (ADS) is a new nuclear energy system which could not only produce clean energy but also incinerate nuclear waste. In this paper, inherent safety analysis of an ADS is performed with neutronics and thermal-hydraulics coupled code named ARTAP. Five typical accidents are carried out, including the cases of proton beam interruption, transient overpower, reactivity insertion, loss of flow, and loss of heat sink. The transient simulations are performed in the average channel and the hottest channel of the fuel pin in the ADS core. The simulation results for beam interruption show that the highest temperature of the pellet is in the middle of the fuel element in the average channel, while the peak temperature of the cladding is in the top of the fuel element. After the beam is interrupted for 20s, the maximum temperature drops at the fuel center, the cladding inner surface, and the outlet coolant in the hottest channel are 644.46K, 162.27K, and 136.42K respectively. For transient overpower accidents with the increase of beam intensity, the maximum temperature of the fuel and the cladding are below the safety limit. Concerning the reactivity insertion accident, it is found that the ADS has good inherent safety and its margin of criticality safety is large. The calculation results for loss of flow show that the power drop is small due to low sensitivity of the subcritical core to negative reactivity feedback, and the maximum temperature of the cladding reaches 1726K, which means the fuel element would rupture. However, the power and the temperatures of fuel, cladding, and coolant could decrease quickly to the safety level after the proton accelerator is cut off under a loss of flow accident. The results also show that the peak temperature of the cladding is lower than the safety limit under a loss of heat sink accident. The present simulation results reveal that the ADS has a remarkable advantage against severe accidents. It also implies that its inherent safety characteristics could ensure reactor shutdown by cutting off the proton beam after accidents occur.

Uncertainty Study of the In-Vessel Phase of a Severe Accident in a Pressurized Water Reactor

A comprehensive uncertainty analysis in the event of a severe accident in a two-loop pressurized water reactor is conducted using an uncertainty package integrated in the ASYST code. The plant model is based on the nuclear power plant (NPP) Krško, a Westinghouse-type power plant. The station blackout scenario with a small break loss of coolant accident is analyzed, and all processes of the in-vessel phase are covered. A best estimate plus uncertainty (BEPU) methodology with probabilistic propagation of input uncertainty is used. The uncertain parameters are selected based on their impact on the safety criteria, the operation of the NPP safety systems and to describe uncertainties in the initial and boundary conditions. The number of required calculations is determined by the Wilks formula from the desired percentile and confidence level, and the values of the uncertain parameters are randomly sampled according to appropriate distribution functions. Results showing the thermal hydraulic behaviour of the primary system and the propagation of core degradation are presented for 124 successful calculations, which is the minimum number of required calculations to estimate a 95/95 tolerance limit at the 3rd order of the Wilks formula application. A statistical analysis of the dispersion of results is performed afterwards. Calculation of the influence measures shows a strong correlation between the decay heat and the representative output quantities, which are the mass of hydrogen produced during the oxidation and the height of molten material in the lower head. As the decay heat increases, an increase in the production of hydrogen and the amount of molten material is clearly observed. The correlation is weak for other input uncertain parameters representing physical phenomena, initial and boundary conditions. The influence of the order of the Wilks formula is investigated and it is found that increasing the number of calculations does not significantly change the bounding values or the distribution of results for this particular application.

중대재해 예방을 위한 안전보건관리체계의 공시 항목 연구

The Serious Accidents Punishment Act was enacted in a surprise move to raise corporate managers' interest and awareness of fatal accidents, effective on January 2022. The government policy to reduce serious disasters by punishments has significant limitations. Thus, it is important to conduct a study on estimating industrial accident costs in the safety and health management system so that companies can voluntarily check the workplace realities. In order to strengthen the occupational safety and health (OSH) management system, this study focuses on the development of safety accident cost disclosure items, estimating industrial accident loss costs, both direct and indirect. The OSH disclosure items can be divided into general information of companies, OSH management system, OSH budget and expenses, certification status, and corporate industrial accident information. And also the OSH disclosure survey form was developed.

Deriving formula to simulate the radiological behavior during LOCA in open pool type reactor

This study aims to model the radiological consequences that result from loss of coolant accident in open pool research reactor of 22 MW power. The loss of coolant accident results from rupture in the cooling systems of the reactor and consequently results in decreasing the water level in the reactor pool. Decreasing the water level, that represents the shielding material above the reactor core, results in increasing the dose rate in the top region in the reactor pool as well as the reactor building. LOCA normally results in shutting down the reactor due to the radioprotection system provided in the upper region of the reactor pool. A failure in shutdown systems of the reactor during LOCA is rare but expected and it represents the worst case. So, evaluating LOCA from the radiological point of view must be investigated for the safe shutdown mode as well as the failure of shutdown case. The study also includes occurring LOCA for shutdown reactor taking the decay time into consideration. ORIGEN-S module from scale system was used to estimate the delayed neutrons and gamma rays from the core for the shutdown mode. KENO-VI module from Scale system was used to estimate the prompt neutrons and gamma rays generated from the core for the failure case. Then, MAVRIC/Scale module was used to estimate the dose rate at the top point of the reactor pool for the two cases during LOCA.Studying the dose rate during LOCA aims to determine the water level that verifies the radiological safety limit at the top point of the reactor pool for the two cases. The calculations introduced a relationship between the water level that verifies the permissible radiological limit (WL) and the decay time (DT) in the range between 0 to 30 d.

Heat transfer analysis of heat pipe cooled device with thermoelectric generator for nuclear power application

Heat pipe cooled reactor featured with compact structure, high reliability, and inherent safety could be applied for the energy system for multipurpose. Numerical simulation methods are widely used to simulate the normal operating conditions of heat pipe reactors and even for accident conditions. In this paper, the ANSYS Fluent is used to perform thermal calculations on the heat pipe thermoelectric generator (HPTEG) prototype for both normal and accident conditions which include heat pipe failure and loss of heat sink accident conditions. The results show that the numerical simulation results are in good agreement with the experimental results, which proves that the fidelity of heat pipe model. Further analyses show that the failure of heat pipe would cause an increase in the power and operating temperature of the adjacent heat pipes, which may increase the possibility of failure of adjacent heat pipes. The maximum core temperature of the HPTEG prototype under the loss of heat sink accident is 1600 K, which exceeds the copper’s melting point temperature 1356 K. It is obvious that the loss of heat sink accident would cause the core temperature increase sharply, which is likely to cause damage to the core. This work provides reference to safety analysis and evaluation of the heat pipe cooled reactor.

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).

Evaluation of advanced accumulator in a VVER-1000 reactor in loss of coolant accident

The advanced accumulator is a passive system which gradually decreases the high flow into a low flow. The role of the fluidic device is to obtain an additional safety margin by increasing golden time to cope in the LOCA. In this paper, the effect of using the safety injection tank and the fluidic device, called advanced accumulator, in a wide spectrum of LOCA break sizes without emergency core cooling systems (ECCS) of a VVER-1000 is evaluated. Firstly, the main component of primary and secondary loops of Bushehr nuclear power plant (BNPP) is modeled by RELAP5. Secondly, the performance of the model is assessed in steady-state and LOCA conditions by comparing the obtained results with the plant’s FSAR. Thirdly, a model is developed in RELAP5 to simulate the behavior of advanced accumulator and validated by blowdown test in a test facility and MARS-KS code. Finally, the performance of advanced accumulator in the spectrum of 200–300 mm break sizes of LOCA without ECCS. In this range of breaks, the core damage times (CDT) developed from peak cladding temperature (PCT) of 1204 °C have been calculated. In accordance with the results, the advanced accumulators can increase the CDT in the whole range of breaks.

Numerical study of blowdown period of double-end break loss of coolant accident in PWR based on RELAP5

Loss of coolant accident (LOCA) is one of the most severe accidents of nuclear power plants (NPP). During the early blowdown period, the system is in a high-pressure gradient and high-velocity gradient state. Therefore, the hydrodynamic forces and safety analysis are highly related to calculating thermal-hydraulic parameters in the flow field. This paper studied the factors and modeling methods that affect the calculation of thermal-hydraulic characteristics under large break LOCA conditions. Firstly, the reactor core and primary system of a typical two-loop pressurized water reactor (PWR) NPP is modeled using the system analysis code RELAP5 Mod4.0. The normal operation state is obtained after the steady calculation. Then, transient calculation of double-end break LOCA is implemented considering the effects of break opening time and the pressurizer. The influence of the results on the calculation of hydrodynamic force has also been analyzed. Finally, the modeling method of reactor pressurized vessel (RPV) downcomer is presented. It is found that the flow field can be better simulated when the downcomer is modeled with parallel channels and several transverse junctions. Thus, this paper provides meaningful suggestions for modeling and simulation in the early blowdown period under LOCA conditions and contributes to further hydrodynamic load calculation and reactor safety analysis.

Development of a Compact Multivariable Sensor Probe for Two-Phase Detection in High Temperature PbLi–argon Vertical Columns

Liquid gas two phase flow is a common occurrence in various industrial applications. For nuclear fusion applications with a Li based breeder, existence of two phase flow may lead to critical issues including decreased tritium breeding ratio, generation of hot spots and improper nuclear shielding. Additionally, a very large density ratio of liquid metal to gas mandates relevant experiments towards development and validation of software tools. PbLi has gained immense focus for its various advantages and is utilized in several breeding blanket concepts. In view of above mentioned requirements, a two phase detection tool is imperative for liquid PbLi environment. For liquid metal applications, electrical conductivity based probes are most suitable in terms of ruggedness, fabrication ease and operational simplicity. However, corrosive nature and high operational temperature for PbLi put severe demands on electrical insulation, a foremost requirement for electrical conductivity based schemes. In this study, a multivariable probe based on electrical conductivity and temperature measurement schemes is developed using Al2O3 coating as electrical insulation. Developed probe is validated in PbLi/Ar vertical column with bulk PbLi temperature upto 400 degC and time averaged void fraction upto 0.95 covering flow regimes from dispersed bubbly flow upto in box loss of coolant accident. Developed probe provides high reliability and excellent temporal resolution towards individual bubble detection through electrical conductivity based principle alongwith simultaneous two phase temperature trends. Present paper provides details about sensor probe fabrication, calibration, two phase test facility, void fraction estimations, bubble frequency and bubble residence time estimations alongwith critical observations from preliminary experimental investigations.