Fracture Probability Assessment for Embrittled Reactor Pressure Vessels Under Ultimate Response Guideline Operation

2016 ◽  
Author(s):  
Hsoung-Wei Chou ◽  
Chin-Cheng Huang
Keyword(s):  
Nuclear Power ◽  
Structural Integrity ◽  
Water Injection ◽  
Fracture Probability ◽  
Pressure Vessels ◽  
Radioactive Material ◽  
Severe Accident ◽  
Core Damage ◽  
Compound Disaster ◽  
Reactor Pressure

After the Fukushima nuclear accident occurred in Japan on March 11, 2011, the compound disaster beyond design basis which may severely impact the nuclear safety has been noticed and paid much more attention. In addition to the original emergency operating procedures (EOPs) and severe accident management guidance (SAMG) of nuclear power plant, the licensee in Taiwan developed the ultimate response guideline (URG) when EOPs and SAMG cannot be performed effectively due to loss of power and water supply by the Fukushima-like compound disaster. Once the URG procedures are initiated, the operators will conduct reactor depressurization, low pressure water injection and containment venting to strictly prevent the core damage and the release of radioactive material. In the paper, the fracture probabilities of boiling water reactor (BWR) pressure vessels with incremental levels of radiation embrittlement under URG operation are evaluated by probabilistic fracture mechanics (PFM) analysis. First, the models of PFM FAVOR code concerning the beltline shell welds of reactor pressure vessels (RPVs) associated with a very conservative flaw distribution are built. Then, the hypothetical transients of URG operation obtained from the thermal hydraulic analyses for Taiwan domestic BWRs are applied as the loading condition. The analysis results demonstrate that performing URG operation will not cause significant fracture probability for RPV, even at an extremely embrittled condition. The URG procedures can ensure the prevention of core damage as well as maintenance of structural integrity of RPV in the situation of long-term loss of electric power when suffering from the Fukushima-like accidents.

Probabilistic Structural Integrity Analysis of Reactor Pressure Vessels During Pressurized Thermal Shock Events

2013 ◽  
Vol 136 (1) ◽  
Author(s):  
Koichi Masaki ◽  
Jinya Katsuyama ◽  
Kunio Onizawa
Keyword(s):  
Crack Initiation ◽  
Thermal Shock ◽  
Structural Integrity ◽  
Fracture Probability ◽  
Pressure Vessels ◽  
Sensitivity Analyses ◽  
Conditional Probabilities ◽  
Integrity Assessment ◽  
Pressurized Thermal Shock ◽  
Reactor Pressure

To apply a probabilistic fracture mechanics (PFM) analysis to the structural integrity assessment of a reactor pressure vessel (RPV), a PFM analysis code has been developed at JAEA. Using this PFM analysis code, pascal version 3, the conditional probabilities of crack initiation (CPIs) and fracture for an RPV during pressurized thermal shock (PTS) events have been analyzed. Sensitivity analyses on certain input parameters were performed to clarify their effect on the conditional fracture probability. Comparisons between the conditional probabilities and the temperature margin (ΔTm) based on the current deterministic analysis method were made for various model plant conditions for typical domestic older types of RPVs. From the analyses, a good correlation between ΔTm and the conditional probability of crack initiation was obtained.

Application of Flaw Updating Process on Probabilistic Structural Evaluation for a Reactor Pressure Vessel Under Pressurized Thermal Shocks

2018 ◽  
Author(s):  
Pin-Chiun Huang ◽  
Hsoung-Wei Chou ◽  
Yuh-Ming Ferng
Keyword(s):  
Pressure Vessel ◽  
Nuclear Power ◽  
Structural Integrity ◽  
Fracture Probability ◽  
Reactor Pressure Vessel ◽  
Structural Evaluation ◽  
User Facility ◽  
Research User ◽  
Reactor Pressure ◽  
Thermal Shocks

The structural integrity of a reactor pressure vessel (RPV) is one of the most important issues for the operation of nuclear power plant. Nowadays, the probabilistic fracture mechanics (PFM) technique is widely used in evaluating the structural integrity of RPVs. However, the flaw characteristics used for PFM analysis are mainly derived from the Pressure Vessel Research User Facility (PVRUF) and Shoreham vessel inspection database, which may not be able to truly represent the vessel-specific condition of the analyzed RPV. In this work, the NUREG-2163 procedure which modifies the flaw characteristic parameters is employed. The Bayesian updating process which combines the prior flaw data with non-destructive examination (NDE) results as well as uncertainties is used to develop the posterior vessel-specific flaw distributions. Subsequently, the updated flaw files are used for PFM analysis to investigate the effects of NDE updated flaw characteristics on the fracture probability of RPV subjected to pressurized thermal shocks. Considering the updated flaws based on the NDE data, the analyzed results could be more plant-specific to predict the fracture risks of RPVs during operation.

Probabilistic Structural Integrity Analysis of Reactor Pressure Vessels During PTS Events

2012 ◽  
Author(s):  
Kunio Onizawa ◽  
Koichi Masaki ◽  
Jinya Katsuyama
Keyword(s):  
Crack Initiation ◽  
Structural Integrity ◽  
Fracture Probability ◽  
Pressure Vessels ◽  
Sensitivity Analyses ◽  
Conditional Probabilities ◽  
Integrity Assessment ◽  
Pressurized Thermal Shock ◽  
Integrity Analysis ◽  
Reactor Pressure

In order to apply a probabilistic fracture mechanics (PFM) analysis to the structural integrity assessment of a reactor pressure vessel (RPV), PFM analysis code has been developed at JAEA. Using the PFM analysis code, PASCAL version 3, the conditional probabilities of crack initiation and fracture for an RPV during pressurized thermal shock events have been analyzed. Sensitivity analyses on some input parameters were performed to clarify the effect on the conditional fracture probability. Comparison between the conditional probabilities and temperature margin (ΔTm) from current deterministic analysis method were made for some model plant conditions of domestic typical old-type RPVs. From the analyses, a good correlation between ΔTm and the conditional probability of crack initiation was obtained.

Structural Integrity Assessment of Doel 3 and Tihange 2 RPVs Affected by Hydrogen Flakes: Refined X-FEM Analyses

2016 ◽  
Author(s):  
Pierre Dulieu ◽  
Valéry Lacroix
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Extended Finite Element Method ◽  
Structural Integrity ◽  
Pressure Vessels ◽  
Extended Finite Element ◽  
Integrity Assessment ◽  
Safety Case ◽  
Component Manufacturing ◽  
Reactor Pressure

During the 2012 outage at Doel 3 and Tihange 2 Nuclear Power Plants, specific ultrasonic in-service inspections revealed a large number of quasi-laminar indications in the base metal of the reactor pressure vessels, mainly in the lower and upper core shells. The observed indications could subsequently be attributed to hydrogen flaking induced during the component manufacturing process. As a consequence, a Flaw Acceptability Assessment had to be performed as a part of the Safety Case demonstrating the fitness-for-service of these units. In that framework, detailed analyses using eXtended Finite Element Method were conducted to model the specific character of hydrogen flakes. Their quasi-laminar orientation as well as their high density required setting up 3D multi-flaws model accounting for flaw interaction. These calculations highlighted that even the most penalizing flaw configurations are harmless in terms of structural integrity despite the consideration of higher degradation of irradiated material toughness.

Development of Probabilistic Evaluation Models of Fracture Toughness KIc and KIa for Japanese RPV Steels

2015 ◽  
Author(s):  
Jinya Katsuyama ◽  
Genshichiro Katsumata ◽  
Kunio Onizawa ◽  
Kazuya Osakabe ◽  
Kentaro Yoshimoto
Keyword(s):  
Fracture Toughness ◽  
Structural Integrity ◽  
Fracture Probability ◽  
Pressure Vessels ◽  
Confidence Limits ◽  
Lognormal Distributions ◽  
Integrity Assessment ◽  
Probabilistic Evaluation ◽  
Evaluation Models ◽  
Reactor Pressure

Probabilistic fracture mechanics (PFM) analysis code PASCAL3 has been developed to apply the PFM analysis to the structural integrity assessment of domestic reactor pressure vessels (RPVs). In this paper, probabilistic evaluation models of fracture toughness KIc and KIa which have the largest scatter among the associated factors based on the database of Japanese RPV steels are presented. We developed probabilistic evaluation models for KIc and KIa based on the Weibull and lognormal distributions, respectively. The models are compared with the existing lower bound of fracture toughness in the Japanese code and probabilistic model in USA. As the results, the 5% confidence limits of the models established in present work corresponded to lower bounds of fracture toughness in the Japanese code. The comparison in the models between present work and USA showed significant differences that may have an influence on fracture probability of RPV.

What Could Have Saved Fukushima From Its Severe Accident

2016 ◽  
Author(s):  
Kenji Iino ◽  
Ritsuo Yoshioka ◽  
Masao Fuchigami ◽  
Masayuki Nakao
Keyword(s):  
Electric Power ◽  
Nuclear Power ◽  
Water Injection ◽  
Radioactive Material ◽  
Severe Accident ◽  
Energy Agency ◽  
Event Scale ◽  
Tsunami Waves ◽  
Ac Power ◽  
Utility Companies

The Great East Japan Earthquake on March 11, 2011 triggered huge tsunami waves that devastated the northeast region of Japan along the Pacific coastline. The Tokyo Electric Power Company (TEPCO) owned Fukushima Daiichi Nuclear Power Plant (Fukushima-1) survived the earthquake, however, not the tsunami that followed. Four of the 6 reactor units underwent Station Blackout. Unit 5 lost all its own AC power, however, it shared AC power with Unit 6. Units 1, 3, and 4 had hydrogen explosions that destroyed their reactor buildings, and even worse, 1, 2, and 3 had core meltdowns to release a large amount of radioactive material to their surroundings. The accident was rated Level 7 on the International Nuclear Event Scale, the worst level defined by International Atomic Energy Agency (IAEA). Reports and papers have been published by a number of entities including the Japanese Diet, Government, TEPCO, IAEA, and more. They give detail explanation of how the accident developed into a nuclear disaster explaining the direct and background causes and faults made after the accident broke out. Finding the accident process, i.e., how it happened, and its causes of why it happened, are the most important first steps in accident analysis. Figuring out how to prevent similar events in the future, or even if it is possible to do so, however, is equally important for our future. We started our study in 2014 to find what actions TEPCO could have taken before the accident for preventing it from growing into a catastrophe. Then in February 2015, we set the goal of our study group to find answers to the following two questions: A. Was the huge tsunami, induced by a huge earthquake, predictable at Fukushima-1? B. If it was predictable, what preparations at Fukushima-1 could have reduced the severity of the accident? In response to our invitation to experts in the nuclear field, active and retired people gathered from academia, manufacturers, utility companies, and even regulators. After a series of tense discussions, we reached the conclusions that: Aa. Tsunami of the level that hit Fukushima-1 in 2011 was well predictable, and, Ba. The accident would have been much less severe if the plant had prepared a set of equipment, and most of all, had exercised actions against such tsunami. Preparation at the plant to prevent the severe accident consisted of the following items 1 through 7, and drills in 8: 1. A number of 125Vdc and 250Vdc batteries, 2. Portable underwater pumps, 3. Portable AC generators with sufficient gasoline supply to run the pumps, and 4. High voltage AC power truck This set applied only to this specific accident. For preparing against many other situations that could have taken place at Fukushima-1, we recommend having, in addition, the following equipment and modifications. 5. Portable compressor to drive air-operated valves for venting, 6. Watertight modification to RCIC and HPCI control and instrumentation, 7. Fire engines for alternate low pressure water injection after vent (Fukushima-1 had three). Just making these preparations would not have been sufficient. Activating valves with DC batteries, for example, takes disengaging the regular power supply lines and hooking up the batteries. 8. Drills against extended loss of all electric power and seawater pump This item 8, on and off-site drills was the most important preparation that should had been made. All other necessary preparations to save the plant in such cases would have followed logically.

Creep Evaluation for a PWR Reactor Pressure Vessel Lower Head Under Severe Accident Conditions Considering Sustained Internal Pressure

2018 ◽  
Author(s):  
YongJian Gao ◽  
Ming Cao ◽  
YinBiao He
Keyword(s):  
Internal Pressure ◽  
Failure Modes ◽  
Structural Integrity ◽  
Creep Damage ◽  
Pressure Vessels ◽  
Mitigation Strategies ◽  
Field Analysis ◽  
Severe Accident ◽  
Lower Head ◽  
Reactor Pressure

In-Vessel Retention (IVR) is one of appropriate severe accident mitigation strategies for AP1000 Nuclear Power Plant (NPP), and assurance of prevention against to thermal failure and structural failure of Reactor Pressure Vessels (RPV) is the prerequisite of IVR. A Finite Element Model fora RPV considering lower head melting was established, the creep calculation was carried out after the temperature field analysis, and the stress-strain responses for different times were obtained. By means of choosing representative evaluation sections and applying the Accumulative Damage Theory based on Larson-Miller Parameter, the Creep Damage calculations and evaluations were conducted. The results showed that the failure modes associated with creep rupture would not happen under IVR condition when a certain amount of internal pressure sustained. The approaches employed in this paper could be utilized in structural integrity evaluation of RPV under IVR for other new type NPPs.

The Influence of Chemistry Concentration Uncertainty on the Probabilistic Fracture Mechanics Analysis

2011 ◽  
Author(s):  
Bo-Yi Chen ◽  
Chin-Cheng Huang ◽  
Hsuing-Wei Chou ◽  
Ru-Feng Liu ◽  
Hsien-Chou Lin
Keyword(s):  
Fracture Mechanics ◽  
Pressure Vessel ◽  
Nuclear Power ◽  
Structural Integrity ◽  
Fracture Probability ◽  
Reactor Pressure Vessel ◽  
Oak Ridge ◽  
Probabilistic Fracture Mechanics ◽  
Full Power ◽  
Reactor Pressure

The chemistry concentration uncertainty of cooper and nickel significantly affects the shift in reference nil-ductility transition temperature (ΔRTNDT). The uncertainty comes from the methods and equipments applied in measurements, the lack of specimen in surveillance capsule, and the non-homogeneous of material. The variations of ΔRTNDT result in the differences of failure probability of reactor pressure vessel. In this study, the structural integrity of Chinshan boiling water reactor RPV shell welds was evaluated by probabilistic fracture mechanics code-Fracture Analysis of Vessel – Oak Ridge (FAVOR). The influence of chemistry concentration uncertainty on the fracture probability of Chinshan nuclear power plant RPV with 32 and 64 effective full power years (EFPY) operation was discussed. The results of this work can be used to evaluate the structural integrity of the welds located at the RPV beltline region, and provide the aging assessment of reactor pressure vessel.

scholarly journals Severe Accident Simulation of the Laguna Verde Nuclear Power Plant

2012 ◽  
Vol 2012 ◽  
pp. 1-11 ◽  
Author(s):  
Gilberto Espinosa-Paredes ◽  
Raúl Camargo-Camargo ◽  
Alejandro Nuñez-Carrera
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Water Injection ◽  
Reactor Core ◽  
Cooling Water ◽  
Fission Products ◽  
Severe Accident ◽  
Core Damage ◽  
Laguna Verde

The loss-of-coolant accident (LOCA) simulation in the boiling water reactor (BWR) of Laguna Verde Nuclear Power Plant (LVNPP) at 105% of rated power is analyzed in this work. The LVNPP model was developed using RELAP/SCDAPSIM code. The lack of cooling water after the LOCA gets to the LVNPP to melting of the core that exceeds the design basis of the nuclear power plant (NPP) sufficiently to cause failure of structures, materials, and systems that are needed to ensure proper cooling of the reactor core by normal means. Faced with a severe accident, the first response is to maintain the reactor core cooling by any means available, but in order to carry out such an attempt is necessary to understand fully the progression of core damage, since such action has effects that may be decisive in accident progression. The simulation considers a LOCA in the recirculation loop of the reactor with and without cooling water injection. During the progression of core damage, we analyze the cooling water injection at different times and the results show that there are significant differences in the level of core damage and hydrogen production, among other variables analyzed such as maximum surface temperature, fission products released, and debris bed height.

Systematic Source Term Analyses for Level 3 PSA of a BWR With Mark-II Type Containment With THALES-2 Code

2002 ◽  
Author(s):  
Jun Ishikawa ◽  
Ken Muramatsu ◽  
Toru Sakamoto
Keyword(s):  
Nuclear Power ◽  
Failure Modes ◽  
Radioactive Material ◽  
Severe Accident ◽  
Source Terms ◽  
Energy Research ◽  
Core Damage ◽  
Reactor Building ◽  
Level 3 ◽  
Accident Scenarios

The THALES-2 code is an integrated severe accident analysis code developed at the Japan Atomic Energy Research Institute in order to simulate the accident progression and transport of radioactive material for probabilistic safety assessment (PSA) of a nuclear power plant. As part of a level 3 PSA being performed at JAERI for a 1,100MWe BWR-5 with a Mark-II containment, a series of calculations were performed by THALES-2 to evaluate the source terms for extensive accident scenarios. For some of the containment failure modes not modeled in THALES-2, such as steam explosion, simple models were coupled with the analysis results of THALES-2 to estimate the source terms. This paper presents the methods and insights from the analyses. An insight from the analyses was that the source terms depend more strongly on the differences in the containment function failure scenarios, such as overpressure failure, controlled containment venting, and small leakage to the reactor building, than those core damage sequences.

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