Pipe Rupture Accident Research Articles

Design and optimization analysis of a new double-layer tube type heat exchanger for lead-bismuth reactors

Double-layer heat tubes have been designed to effectively reduce the occurrence of heat pipe rupture accidents. However, inter-tube thermal contact resistance can decrease heat transfer efficiency, thus hampering the heat dissipation in the primary loop system of lead-bismuth reactors. Therefore, optimizing the design of double-layer heat tubes is necessary. This work focuses on the double-layer heat exchanger of a lead-bismuth reactor and utilizes gallium-based graphene nanofluids as a thermal interface material to fill the gap between the heat tubes. Furthermore, the impact of the length, wall thickness, outer diameter, and spacing of heat tubes on the heat transfer performance of the double-layer heat exchanger with and without the nanofluids has been analyzed. The study aims to optimize the JF factor and cost-effectiveness ratio (CER). Genetic algorithms are employed to optimize and evaluate the heat transfer performance of the main heat exchanger based on the four aforementioned parameters. Consequently, a new design scheme is obtained for the double-layer heat exchanger, which increases the optimized overall heat transfer coefficient of the main heat exchanger by 5.79%, pressure drop in the primary loop by 2.32%, JF factor by 5%, and CER by 24.62%. These results demonstrate that the gallium-based graphene nanofluids can effectively enhance the heat transfer performance of the double-layer heat exchanger while reducing the likelihood of steam generator tube rupture accidents.

Jet impingement in high-energy piping systems, part I: Characteristics and model evaluation

Under postulated pipe rupture accidents in nuclear power plants (Loss-of-Coolant Accident or LOCA), jet impingement can occur where the high-energy fluid is discharged and cause damage to the surrounding structures, systems, and components (SSCs). As a part of addressing the potential non-conservatisms in the Standard model ANSI/ANS-58.2 (1988), the current work investigates characteristics of jet impingement in high energy piping systems as in nuclear power plants including pressure distribution within jets, pressure on target, and critical mass flux and performs model evaluation. The pressure prediction using ANSI/ANS-58.2 (1988) as well as critical mass flux models are evaluated using the experimental data available in literature. An organized experimental database with different inlet fluid conditions including subcooled water, saturated water/two-phase (0 ≤ x < 0.7), and saturated steam (x ≥ 0.7) is compiled for data analysis and model evaluation. It is found that a power law can be used to correlate the dimensionless static and stagnation pressures at the center of the free jets, while either a power law or an exponential function can be used to correlate the dimensionless stagnation pressure at the center of the targets in impinging jets. The difference in stagnation pressure between free and impinging jets is negligible for saturated steam and saturated water/two-phase jets, but it is significant for subcooled water jets. From model evaluations, the stagnation pressure within the saturated steam jets can generally be predicted well by the Standard model. Although stagnation pressures at far distances (z* > 3.3) can be underestimated, the dimensionless pressures at these axial locations are below 0.07. For saturated water/two-phase jets, the stagnation pressure can be predicted well using the Standard model for the large-scale test data (D > 0.28 m) but are underestimated at some axial locations for the medium- and small-scale test data (D < 0.15 m). Significant overestimation in stagnation pressure is observed for subcooled water jets with high degree of subcooling, which is because the critical mass flux is overestimated using Homogeneous Equilibrium Model (HEM). For subcooled water jets with low degree of subcooling, HEM underestimates the data. Different from HEM, Henry-Fauske model always overpredicts the experimental data, where the degree of overestimation increases as the degree of subcooling decreases. The critical mass flux for saturated steam jets and saturated water/two-phase jets can be predicted well by isentropic expansion model and HEM, respectively.

Jet impingement in high-energy piping systems, part II: Model improvement and guidance development

The Standard model in ANSI/ANS-58.2 (1988) has been widely employed in the design and safety analysis of nuclear power plants for jet impingement under postulated pipe rupture accidents. However, the potential non-conservatisms presented in Standard Review Plan (2015) need to be addressed when evaluating this model. The current work aims to improve the model on jet prediction including jet geometry and stagnation pressure, as well as to address the potential non-conservatisms. Based on the model evaluation results in Part I, new correlations for the jet core length and stagnation pressure in subcooled water and saturated water/two-phase jets are implemented to replace the existing correlations in the Standard model. It is shown that the revised code with the newly developed correlations predicts the stagnation pressure generally higher or conservatively for all conditions investigated.On the potential blast wave effect, the current work develops an approach to quantify the over-static and reflected pressures using the equivalent weight of TNT approach in literature. To employ this approach, the released energy in the equivalent of TNT needs to be determined from jet impingement. This energy is calculated using the vessel burst energy concept with conservative assumptions in the current work. For jet geometry, dynamic models to predict the jet spreading angle for saturated steam jets are compared with the Prandtl-Meyer angle for under-expanded supersonic jets. It is found that the jet spreading angle for saturated steam jets in the Standard model generally agrees with the dynamic model in literature for a wide range of initial vessel stagnation pressure conditions. Regarding the pressure distribution within jets, it is confirmed that edge-peaking pressure profiles can be observed for saturated steam jets at far distances from the break plane (z > 3.3 D); however, the dimensionless pressures at these locations are generally less than 0.07. Regarding the jet dynamic loading effect, the literature review suggests that dynamic amplification and resonance of impinged downstream structures and components might not occur for jet impingement in nuclear power plants.

Identification of hydrogen flammability in steam generator compartment of OPR1000 using MELCOR and CFX codes

Abstract The MELCOR code useful for a plant-specific hydrogen risk analysis has inevitable limitations in prediction of a turbulent flow of a hydrogen mixture. To investigate the accuracy of the hydrogen risk analysis by the MELCOR code, results for the turbulent gas behavior at pipe rupture accident were compared with CFX results which were verified by the American National Standard Institute (ANSI) model. The postulated accident scenario was selected to be surge line failure induced by station blackout of an Optimized Power Reactor 1000 MWe (OPR1000). When the surge line failure occurred, the flow out of the surgeline was strongly turbulent, from which the MELCOR code predicted that a substantial amount of hydrogen could be released. Nevertheless, the results indicated nonflammable mixtures owing to the high steam concentration released before the failure. On the other hand, the CFX code solving the three-dimensional fluid dynamics by incorporating the turbulence closure model predicted that the flammable area continuously existed at the jet interface even in the rising hydrogen mixtures. In conclusion, this study confirmed that the MELCOR code, which has limitations in turbulence analysis, could underestimate the existence of local combustible gas at pipe rupture accident. This clear comparison between two codes can contribute to establishing a guideline for computational hydrogen risk analysis.

Process of Air Ingress during a Depressurization Accident of GTHTR300

A depressurization accident is the design-basis accidents of a gas turbine high temperature reactor, GTHTR300, which is JAEA’s design and one of the Very-High-Temperature Reactors (VHTR). When a primary pipe rupture accident occurs, air is expected to enter the reactor core from the breach and oxidize in-core graphite structures. Therefore, it is important to know a mixing process of different kinds of gases in the stable and unstable density stratified fluid layer. In order to predict or analyze the air ingress phenomena during the depressurization accident, we have conducted an experiment to obtain the mixing process of two component gases and the characteristics of natural circulation. The experimental apparatus consists of a storage tank and a reverse U-shaped vertical rectangular passage. One side wall of the high temperature side vertical passage is heated and the other side wall is cooled. The other experimental apparatus consists of a cylindrical double coaxial vessel and a horizontal double coaxial pipe. The outside of the double coaxial vessel is cooled and the inside is heated. The results obtained in this study are as follows. When the primary pipe is connected at the bottom of the reactor pressure vessel, onset time of natural circulation of air is affected by not only molecular diffusion but also localized natural convection. When the wall temperature difference is large, onset time of natural circulation of air is strongly affected by natural convection rather than molecular diffusion. When the primary pipe is connected at the side of the reactor pressure vessel, air will enter the bottom space in the reactor pressure vessel by counter-current flow at the coaxial double pipe break part immediately. Afterward, air will enter the reactor core by localized natural convection and molecular diffusion.

Experiment and Dynamic Simulation of PIG Motion during Pigging Operation in a Slope Pipeline

Abstract Pigging techniques are widely used in the oil and gas industry. The unsteady motion of the PIG in an undulating pipe section during the pigging process after a water pressure test affects the stable operation of the pipeline and also causes a pipe rupture accident in serious cases. First, an experimental study was conducted to investigate the pigging process of air–water two phase pipe flows, and the PIG reverse movement and hydraulic pulse phenomenon were observed. Subsequently, a hydraulic transient model of the pigging process after a water pressure test was established in a dual-grid system. The model combined mass and motion equations of gas and liquid and PIG dynamic equations, considered three types of PIG motion states, namely positive movement, reverse movement and still, and used the method of characteristics to solve the equations. The model exhibits the ability for PIG tracing and hydraulic pulse prediction. It can be used to obtain the position and speed of the PIG. Finally, the field data and simulation results were compared, and the results indicated that they are essentially identical. This verified the accuracy of the model that is established in this study and the reliability of computed results and provided a reliable and effective theoretical basis for the development of field pigging plans.

Process of air ingress during double coaxial horizontal pipe rupture accident of VHTR

Process of air ingress during double coaxial horizontal pipe rupture accident of VHTR

Experiments and numerical analysis of a control method for natural circulation through helium gas injection

This study investigated a control method for natural circulation of air by helium gas injection. A depressurization accident is a design-basis accident of a very high temperature reactor. When a primary pipe rupture accident occurs, air is expected to enter the reactor pressure vessel from the breach. Thus, in-core graphite structures are oxidized. In order to predict and analyze the phenomena of air ingress during a depressurization accident, numerical analysis was carried out using a one-dimensional (1D) analysis code and three-dimensional computational fluid dynamics (3D CFD). An experiment was carried out regarding natural circulation using a circular pipe consisting of a reverse U-shaped channel. The channel consisted of two vertical heated and cooled pipes. The temperature difference between the vertical pipes was maintained at 40–80K, and a small amount of helium gas was injected into the channel. The injected volume of helium was about 3.1–12.5% of the total channel volume. After injecting helium gas, each component gas moved through molecular diffusion and very weak natural circulation. After approximately 1180s, ordinary natural circulation of air was suddenly produced. The numerical results of the 3D CFD code were in good agreement with the experimental results. The numerical results also showed that the natural circulation of air can be controlled by helium gas injection.

Analysis of the Consequences of a First-Loop Pipe Rupture Accident in IR-8 upon Conversion to Low-Enrichment Fuel

An analysis of design basis accidents has been completed and an analysis of the consequences of postulated beyond design basis accidents is now being performed on the IR-8 reactor. A computational analysis of the consequences of a beyond design basis accident with rupture of the delivery and intake piping in the first cooling loop of the reactor is being done. Preliminary assessments were initially made using the one-dimensional code BEREZA. In order to confirm the results and determine the time to onset of core drying, a three-dimensional thermohydraulic model of IR-8 was developed and constructed, and a computational analysis of this accident was performed using the systems code ATHLET.

ICONE23-1930 STUDY ON THE ONE-DIMENSIONAL NATURAL CIRCULATION PHENOMENON ON THE MIXING PROCESS OF TWO COMPONENT GASES IN A VERTICAL FLUID LAYER

A depressurization accident is one of the design-basis accidents of the VHTR. When the pipe rupture accident occurs, air is expected to enter the reactor core from the breach and oxidize in-core graphite structures. In order to predict or analyze the process of air ingress during the depressurization accident of the VHTR, it is very important to develop computer programs and to validate them by experiments. This study is to investigate the effect of one-dimensional natural circulation on the mixing process of two component gases by evaluating the onset time of natural circulation through the apparatus under the stable density stratified fluid layer. The experimental apparatus consists of a reverse U-shaped vertical slot and a storage tank. The left side vertical slot consists of the heated wall and the cooled wall. The right side vertical slot consists of the two cooled walls. These experimental results show that generation time of natural circulation was affected by molecular diffusion and localized natural convection. When the two components of gases have large density ratios and large Gr numbers, the mixing process of two components of gases was affected by more intensively molecular diffusion than localized natural convection when temperature difference was 50K. The mixing process of two component gas was affected by more intensively localized natural convection than molecular diffusion when temperature difference was 70 to 100K. However, two component gases were affected by more intensively molecular diffusion than localized natural convection at small density ratios and small Gr numbers.

Study on control method of natural circulation by injection of helium gas

This study is to investigate a control method of natural circulation of air by injection of helium gas. A depressurization accident is one of the design-basis accidents of a very high temperature reactor (VHTR). When the primary pipe rupture accident occurs in the VHTR, air is expected to enter into the reactor pressure vessel from the breach and oxidize in-core graphite structures. Finally, it seems to be probable that the natural circulation flow of air in the reactor pressure vessel produce continuously. In order to predict or analyze the air ingress phenomena during the depressurization accident of the VHTR, therefore, it is important to develop the method for prevention of air ingress during the accident. The experiment has been carried out regarding natural circulation using a circular tube consisting of a reverse U-shaped type. The vertical channel consists of the one side heated tube and the other side cooled tube. The experimental procedure is as follows. Firstly, the apparatus is filled with air and one vertical tube is heated. Then, natural circulation of air will be produced in the channel. After the steady state is established, a very small amount of helium gas injects from the top of the channel. The velocity, temperature of gas, and temperature of the tube wall are measured during the experiment. The analysis also has been carried out regarding natural circulation. The results were obtained as follows. The temperature difference between the vertical pipes was 50–130K, and a small amount of helium gas injected to the channel. The volume of injected helium gas is about 3.5–10% of the total volume of the channel. When the temperature difference between the vertical tubes was kept at 52K, the velocity of natural circulation flow became about 0.12m/s. During a steady state, a small amount of helium gas injected into the channel. Then, the flow velocity of natural circulation suddenly decreased. The volume of injected helium gas is about 3% of the total volume of the channel. The velocity became around zero. After 1500s elapsed, the natural circulation suddenly produced again. The experimental and analytical results show that the natural circulation flow of air can be controlled by injecting of helium gas.

Unanticipated Large-Diameter Pipe Rupture Accidents in the RBMK-1000 Circulation Loop with Reactor Emergency Cooling System Failure

The results of calculations of large-diameter pipe rupture accidents in the RBMK circulation loop (pressure header of the main circulation pump, down pipe of the steam drum, and distribution group header) with failure of the reactor emergency cooling system are presented. The main computational problem is to evaluate the onset time of fuel-element cladding depressurization in the damaged half of the core and the integrity of the channel tubes. It is shown that for rupture of the pressure header fuel-element cladding depressurization starts at the 10th minute after the initial event and depressurization of the down-pipe near the steam drum and distribution group header start at the 17th and 40th minutes, respectively. The possible rupture of the channel tubes can have the form of only individual small flaws, for which the structural integrity of the reactor space is preserved. The calculations were performed with the RELAP5/mod3.2 computer code adapted for RBMK.

D232 BWR配管減肉規格(2006年版)の改訂案の概要と技術的根拠(OS6 保全・設備診断技術(7))

The pipe rupture accident at Mihama Nuclear Power Station occurred in August 2004. Japan Society of Mechanical Engineers (JSME) established the rules on pipe wall thinning management for BWR and PWR in 2006 to prevent such pipe ruptures. The rules for each type of plant were based on field data obtained by wall thinning measurement at that time. In these rules, field data and technical knowledge regarding pipe wall thinning phenomena including FAC and LDI were adopted. Since the establishment of the rules, JSME has been working on the improvement of the methods for more proper prediction and updating the field data to enhance the pipe wall thinning management and to revise the present rules. In this paper, an overview for the revision proposals of the management for BWR is described according to the field data.

Radioactive Research on Main Steam Pipe Rupture Accident of Nuclear-Powered Equipment of Ship

Respectively considering both the condition of the main steam pipe rupture position at the reactor cabin and cabin behind the reactor cabin, the accident process of the main steam pipe rupture is analyzed, the mathematic and physical model about the transference, leakage and diffusion of the radioactive nuclide is established. Through the method of coupling the thermodynamic calculation result with the radioactive result analysis model, the calculation of the nuclide radioactive concentration in the cabin is realized. Further more the harm of the radioactivity has been analyzed. The research of the paper has provided a basis for the effective constitute of the radioprotection measure in the accident.

沸騰水型原子力発電所の配管減肉に関するフィールドデータの分析

Strongly motivated by every stakeholder not to repeat Mihama Nuclear Power Station pipe rupture accident in August 2004, JSME Main Committee on Codes & Standards on Power Generation Facilities immediately launched a special task force to develop Rules on Pipe Wall Thinning Management for BWR, PWR and fossil Power Plants respectively. The authors describes the process of the development of Rules for BWR Power Plants from the view point of collections and analysis of fields data of pipe wall thinning. Through its activities, the authors confirmed the existing findings, like the effect of Oxygen injection, turbulance and dependence on coolant temperature, derived from series of laboratory-scaled experiments in FAC and coolant velocities effects in LDI. Further based upon the said proven findings with field data, they explain the adequacy of major concept of the rule such as separate treatment of FAC (Flow Accelerated Corrosion) and LDI (Liquid Droplet Impingement).

Analysis of Data on Pipe Wall Thinning Phenomena by Fluid Flow in Pressurized Water Reactor Plant

In August 2004, a pipe rupture accident occurred at The Kansai Electric Power Company Mihama Unit 3. Japan Society of Mechanical Engineers (JSME) recognized that it was an urgent problem to settle the rules on pipe wall thinning management by fluid flow. JSME enacted rules on pipe wall thinning that stipulates the basic requirements of pipe wall thinning management, considering from these situations, and in 2006, the “Rules on Pipe Wall Thinning Management for PWR Power Plants” (JSME S NG 1-2006, JSME rules for PWR) that satisfies the requirements of rules on pipe wall thinning. This JSME rules for PWR stipulates the process of inspection plan, measurement, evaluation and maintenance. This paper focuses on the technical grounds that specify the management systems, pipe components and initial FAC rate.

Pressurizing behavior on ingress of coolant into pebble bed of blanket of Fusion DEMO reactor

In the safety design and technology of blankets for the Fusion DEMO reactor developed in Japan, coolant ingress in the blanket box structure is one of the most important events. Especially, the thermal hydraulics in the pebble bed in the case of the high-pressure coolant ingress is very important to evaluate the pressure propagation and coolant flow behavior. This paper presents the preliminary results of the pressure loss characteristics by the coolant ingress in the pebble bed. Experiments have been performed to simulate the helium coolant ingress into breeder and multiplier pebble beds and to evaluate the pressure loss in the pebble beds. The measured pressure loss is compared with the predicted values by Ergun's equation, which is the correlation equation on pressure loss of the flow through porous medium. Parameter study is carried out and dependencies of pressure loss through pebble bed on pressure and packing factor is evaluated. Method to reduce the maximum pressure in solid breeder blanket after pipe rupture accident is discussed. Enhancement of pressure loss in upstream and reduction of pressure loss in downstream is effective.

ICONE15-10127 Visualization Study of Helium-Air Counter Flow through a small Opening

Buoyancy-driven counter flows of helium-air were investigated through horizontal and inclined small openings. Counter flows may occur following a window opening as ventilation, fire in the room as well as a pipe rupture accident in a high temperature gas-cooled nuclear reactor. The experiment has carried out by a test chamber filled with helium and flow was visualized by the smoke wire method. The flow behavior has recorded by a high-speed camera with a computer system. The image of the flow was transferred to the digital data, thus the flow velocity was measured by PTV software. The mass fraction in the test chamber was measured by electronic balance. The detected data was arranged by the densimetric Floude number of the counter flow rate that derived from the dimensional analysis. The method of mass increment was developed and applied to measure the counter flow rate. By removing the cover plate placed on the top of the opening, the counter flow initiated. Air enters the test chamber and the mass of the gas mixture in the test chamber increased. The volumetric counter flow rate was evaluated from the mass increment data. In the case of inclination openings, the results of both methods were compared. The inclination angle for maximum densimetric Floude number decreased with increasing length-to-diameter ratio of the opening. For a horizontal opening, the results from the method of mass increment agreed with those obtained by other authors for a water-brine system.

Numerical Analysis on Air Ingress Behavior in GTHTR300-Cogeneration System

The objective of this study is to clarify safety characteristics of a High Temperature Gas-Cooled Reactor (HTGR) for the pipe rupture accident. Japan Atomic Energy Agency (JAEA) has been developing the analytical code for the safety characteristics of the HTGR and carrying out design study of the gas turbine high temperature reactor of 300MWe nominal-capacity for hydrogen production, the GTHTR300C (Gas Turbine High Temperature Reactor 300 for Cogeneration). A numerical analysis of heat and mass transfer fluid flow with multi-component gas mixture has been performed to obtain the variation of the density of the gas mixture, and the onset time of natural circulation of air. From the results obtained in this analysis, it was found that the duration time of the air ingress by molecular diffusion would increase due to the existence of the recuperator in the GTHTR300C system.

C204 高速度カメラ撮影による対向置換流挙動の可視化

Buoyancy-driven exchange flows of helium-air were investigated through inclined a small opening. Exchange flows may occur following a window opening as ventilation, fire in the room as well as a pipe rupture accident in a high temperature gas-cooled nuclear reactor. The experiment has carried out by a test chamber filled with helium and flow was visualized by the smoke wire method. By removing the cover plate placed on the top of the opening, the exchange flow initiated. Air enters the test chamber and the mass of the gas mixture in the test chamber increased. The flow behavior has recorded by a high-speed camera with a computer system. The image of the flow was transferred to the digital data, thus the flow velocity was measured by PTV software. The volumetric exchange flow rate was evaluated from the PTV method and mass increment method. In the case of inclination openings, the results of both methods were compared. The detected data was arranged by the densimetric Floude number of the exchange flow rate that derived from the dimensional analysis.