Effect of Coolant Water Temperature of Emergency Core Cooling System on Failure Probability of Reactor Pressure Vessel

2020 ◽  
Vol 143 (3) ◽  
Author(s):  
Kai Lu ◽  
Jinya Katsuyama ◽  
Koichi Masaki ◽  
Tadashi Watanabe ◽  
Yinsheng Li
Keyword(s):  
Water Temperature ◽  
Failure Probability ◽  
Pressure Vessel ◽  
Cooling System ◽  
Reactor Pressure Vessel ◽  
Pressurized Water ◽  
Model Plant ◽  
Coolant Water ◽  
Core Cooling ◽  
Reactor Pressure

Abstract Structural integrity assessment of reactor pressure vessel (RPV) is important for the safe operation of nuclear power plant. For an RPV in a pressurized water reactor (PWR), pressurized thermal shock (PTS) resulted from rapid coolant water injection due to a loss-of-coolant accident is an issue of particular concern. The coolant water temperature in the emergency core cooling system (ECCS) can influence the integrity of RPV subjected to PTS events; thus, this paper is focused on investigating the effect of coolant water temperature of ECCS on failure probability of an RPV. First, thermal-hydraulic (TH) analyses were conducted for a Japanese PWR model plant by using RELAP5, and different coolant water temperatures in ECCS were considered to investigate the effect of coolant water temperature on TH behaviors during a PTS event. Using the TH analysis results, probabilistic fracture mechanics (PFM) analyses were performed for the RPV of the Japanese model plant. Based on the PFM analysis results, the effect of coolant water temperature on failure probability of the RPV was quantified.

Effect of Coolant Water Temperature of ECCS on Failure Probability of RPV

2019 ◽  
Author(s):  
Jinya Katsuyama ◽  
Koichi Masaki ◽  
Kai Lu ◽  
Tadashi Watanabe ◽  
Yinsheng Li
Keyword(s):  
Water Temperature ◽  
Thermal Shock ◽  
Failure Probability ◽  
Structural Integrity ◽  
Cooling System ◽  
Pressure Vessels ◽  
Coolant Temperature ◽  
Pressurized Water ◽  
Coolant Water ◽  
Core Cooling

Abstract For reactor pressure vessels (RPVs) of pressurized water reactors, temperature of the coolant water in the emergency core cooling system (ECCS) may influence the structural integrity of the RPV during pressurized thermal shock (PTS) events. By focusing on a mitigation measure to raise the coolant water temperature of ECCS for aged RPVs to reduce the effect of thermal shock due to PTS events, we performed thermal hydraulic analyses and probabilistic fracture mechanics analyses by using RELAP5 and PASCAL4, respectively. The analysis results show that the failure probability of RPV decreased dramatically when the coolant temperature in accumulator as well as in the high- and low-pressure injection systems (HPI/LPI) was increased, although the increase in coolant temperature in the HPI/LPI only did not lead to a decrease in the failure probability.

Optimizing Maintenance Strategy of a Reactor Pressure Vessel Using 3D-CFD and FEM Based Probabilistic Pressurized Thermal Shock Analysis

2017 ◽  
Author(s):  
Xiaoyong Ruan ◽  
Toshiki Nakasuji ◽  
Kazunori Morishita
Keyword(s):  
Spatial Distribution ◽  
Thermal Shock ◽  
Pressure Vessel ◽  
Nuclear Power ◽  
Structural Integrity ◽  
Cooling System ◽  
Reactor Pressure Vessel ◽  
Pressurized Thermal Shock ◽  
Core Cooling ◽  
Reactor Pressure

The structural integrity of a reactor pressure vessel (RPV) is important for the safety of a nuclear power plant. When the emergency core cooling system (ECCS) is operated and the coolant water is injected into the RPV due to a loss-of-coolant accident (LOCA), the pressurized thermal shock (PTS) loading takes place. With the neutron irradiation, PTS loading may lead a RPV to fracture. Therefore, it is necessary to evaluate the performance of RPV during PTS loading to keep the reactor safety. In the present study, optimization of RPV maintenance is considered, where two different attempts are made to investigate the RPV integrity during PTS loading by employing the deterministic and probabilistic methodologies. For the deterministic integrity evaluation, 3D-CFD and finite element method (FEM) simulations are performed, and stress intensity factors (SIFs) are obtained as a function of crack position inside the RPV. As to the probabilistic integrity evaluation, on the other hand, a more accurate spatial distribution of SIF on the RPV is calculated. By comparing the distribution thus obtained with the fracture toughness included as a part of the master curve, the dependence of fracture probabilities on the position inside the RPV is obtained. Using the spatial distribution of fracture probabilities in RPV, the priority of the inspection and maintenance is finally discussed.

Study on Structural Integrity Assessment of Reactor Pressure Vessel Based on Three-Dimensional Thermal-Hydraulics and Structural Analyses

2014 ◽  
Author(s):  
Jinya Katsuyama ◽  
Genshichiro Katsumata ◽  
Kunio Onizawa ◽  
Tadashi Watanabe ◽  
Yutaka Nishiyama
Keyword(s):  
Pressure Vessel ◽  
Structural Integrity ◽  
Three Dimensional ◽  
Reactor Core ◽  
Reactor Pressure Vessel ◽  
Thermal Hydraulics ◽  
Pressurized Water ◽  
Integrity Assessment ◽  
Coolant Water ◽  
Reactor Pressure

In the structural integrity assessment of a pressurized water reactor pressure vessel (RPV) during pressurized thermal shock (PTS) events, the thermal history of the coolant water and the heat transfer coefficient between the coolant water and RPV are dominant factors. These values can be determined on the basis of thermal-hydraulics (TH) analysis simulating PTS events and Jackson-Fewster correlation. Subsequently, using these values, structural integrity assessments of RPV are performed by structural analysis; e.g., loading that affects crack propagation is evaluated. Three-dimensional TH and structural analyses are recommended for precise assessments of the structural integrity of RPV. In this study, we performed TH and structural analyses simulating typical PTS events using three-dimensional models of cold-leg, downcomer and RPV to more accurately assess the structural integrity of RPV. From these analyses, we obtained loading histories from the reactor core region of RPV in which a crack is postulated in the structural integrity assessment. We discuss the conservativeness of current analysis methods on the structural integrity assessment of RPV through the comparison of loading conditions due to PTS events.

Loss of Core Cooling Test Without One Cooling Line in Vessel Cooling System (VCS) of High Temperature Engineering Test Reactor (HTTR)

2016 ◽  
Author(s):  
Yusuke Fujiwara ◽  
Takahiro Nemoto ◽  
Daisuke Tochio ◽  
Masanori Shinohara ◽  
Masato Ono ◽  
...  
Keyword(s):  
Control System ◽  
Pressure Vessel ◽  
Cooling System ◽  
Reactor Power ◽  
Reactor Pressure Vessel ◽  
Measured Temperature ◽  
Helium Gas ◽  
Test Reactor ◽  
Core Cooling ◽  
Reactor Pressure

In the High Temperature engineering Test Reactor (HTTR), the Vessel Cooling System (VCS) which is arranged around the reactor pressure vessel, removes residual heat and decay heat from the reactor core passively when the forced core cooling is lost. The test was carried out at the reactor thermal power of 9 MW, under the condition that the reactor power control system and the reactor inlet coolant temperature control system was isolated, and three helium gas circulators in the primary cooling system were stopped to simulate the loss of forced cooling flow in the core. Moreover, one cooling line of the VCS was stopped to simulate the partial loss of cooling function from the surface of the reactor pressure vessel. The test results showed that the reactor power immediately decreased to almost zero and was stable as soon as the helium gas circulators were stopped. The power decrease is caused by negative feedback effect of reactivity. On the other hand, temperature changes of core internal structures, the reactor pressure vessel and the biological shielding concrete were slowly during the test. The measured temperature of the reactor pressure vessel decreased for several degrees during the test. The measured temperature increase of biological shielding made of concrete was small within 1 °C. The numerical analysis showed that the temperature increase of VCS cooling tube was about 15°C which is sufficiently small, which did not significantly affect the temperature of biological shielding. As the results, it was confirmed that the cooling ability of VCS can be kept sufficiently even in case that one of two water cooling lines of VCS is lost.

scholarly journals Analysis of the Technological Evolution of Materials Requirements Included in Reactor Pressure Vessel Manufacturing Codes

2021 ◽  
Vol 13 (10) ◽  
pp. 5498
Author(s):  
Alvaro Rodríguez-Prieto ◽  
Mariaenrica Frigione ◽  
John Kickhofel ◽  
Ana M. Camacho
Keyword(s):  
Sustainable Development ◽  
Pressure Vessel ◽  
Nuclear Regulatory Commission ◽  
Green Energy ◽  
Reactor Pressure Vessel ◽  
Technical Literature ◽  
Pressurized Water ◽  
Water Reactors ◽  
Regulatory Commission ◽  
Reactor Pressure

The growth of green energy technologies within the frame of the 7th Sustainable Development Goal (SDG) along with the concern about climatic changes make nuclear energy an attractive choice for many countries to ensure energy security and sustainable development as well as to actively address environmental issues. Unlike nuclear equipment (immovable goods), which are often well-catalogued and analyzed, the design and manufacturing codes and their standardized materials specifications can be considered movable and intangible goods that have not been thoroughly studied based on a detailed evaluation of the scientific and technical literature on the reactor pressure vessel (RPV) materials behavior. The aim of this work is the analysis of historical advances in materials properties research and associated standardized design codes requirements. The analysis, based on the consolidated U.S. Nuclear Regulatory Commission (NRC) Regulatory Guide (RG) 1.99 Rev.2 model, enables determination of the best materials options, corresponding to some of the most widely used material specifications such as WWER 15Kh2MFAA (used from the 1970s and 1980s; already in operation), ASME SA-533 Grade B Cl.1 (used in pressurized water reactor-PWR 2nd–4th; already in operation), DIN 20MnMoNi55 and DIN 22NiMoCr37 (used in PWR 2nd–4th) as well as ASTM A-336 Grade F22V (current designs). Consequently, in view of the results obtained, it can be concluded that the best options correspond to recently developed or well-established specifications used in the design of pressurized water reactors. These assessments endorse the fact that nuclear technology is continually improving, with safety being its fundamental pillar. In the future, further research related to the technical heritage from the evolution of materials requirements for other clean and sustainable power generation technologies will be performed.

Failure Probability Assessment for a Boiling Water Reactor Pressure Vessel Under Low Temperature Over-Pressure Event

2012 ◽  
Author(s):  
Hsoung-Wei Chou ◽  
Chin-Cheng Huang ◽  
Bo-Yi Chen ◽  
Hsien-Chou Lin ◽  
Ru-Feng Liu
Keyword(s):  
Low Temperature ◽  
Failure Probability ◽  
Pressure Vessel ◽  
Fracture Probability ◽  
Pressure Vessels ◽  
Reactor Pressure Vessel ◽  
Boiling Water ◽  
Boiling Water Reactor ◽  
Water Reactor ◽  
Reactor Pressure

The fracture probability of a boiling water reactor pressure vessel for a domestic nuclear power plant in Taiwan has been numerically analyzed using an advanced version of ORNL’s FAVOR code. First, a model of the vessel beltline region, which includes all shell welds and plates, is built for the FAVOR code based on the plant specific parameters of the reactor pressure vessel. Then, a novel flaw model which describes the flaw types of surface breaking flaws, embedded weld flaws and embedded plate flaws are simulated along both inner and outer vessel walls. When conducting the fracture probability analyses, a transient low temperature over-pressure event, which has previously been shown to be the most severe challenge to the integrity of boiling water reactor pressure vessels, is considered as the loading condition. It is found that the fracture occurs in the fusion-line area of axial welds, but with only an insignificant failure probability. The low through-wall cracking frequency indicates that the analyzed reactor pressure vessel maintains sufficient stability until either the end-of-license or for doubling of the present license of operation.

Dosimetry Assessments for the Reactor Pressure Vessel and Core Barrel in UK Pressurized Water Reactor (PWR) Plant

2012 ◽  
Vol 9 (4) ◽  
pp. 104016 ◽  
Author(s):  
D. A. Thornton ◽  
D. A. Allen ◽  
A. P. Huggon ◽  
D. J. Picton ◽  
A. T. Robinson ◽  
...  
Keyword(s):  
Pressure Vessel ◽  
Reactor Pressure Vessel ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Water Reactor ◽  
Reactor Pressure

Reactor Pressure Vessel Integrity in Light of the Evolution of Materials Science and Engineering

2005 ◽  
Vol 473-474 ◽  
pp. 287-292
Author(s):  
Péter Trampus
Keyword(s):  
Pressure Vessel ◽  
Structural Integrity ◽  
Materials Science ◽  
Reactor Pressure Vessel ◽  
Special Focus ◽  
Science And Engineering ◽  
Pressurized Water ◽  
Integrity Assessment ◽  
Materials Science And Engineering ◽  
Reactor Pressure

Structural integrity of the reactor pressure vessel of pressurized water reactors is one of the key safety issues in nuclear power operation. Integrity may be jeopardized during operational transients. The problem is compounded by radiation damage of the vessel structural materials. Structural integrity assessment as an interdisciplinary field is primarily based on materials science and fracture mechanics. The paper gives an overview on the service induced damage processes and associated changes of mechanical properties, the prediction of degradation and the assessment of the entire component against brittle fracture with a special focus on how the evolution of materials science and engineering has contributed to reactor vessel structural integrity assessment.

Structural Reliability Evaluation on the Pressurized Water Reactor Pressure Vessel Under Pressurized Thermal Shock Events

2014 ◽  
Author(s):  
Hsoung-Wei Chou ◽  
Chin-Cheng Huang
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Thermal Shock ◽  
Pressure Vessel ◽  
Nuclear Power ◽  
Reactor Pressure Vessel ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Pressurized Thermal Shock ◽  
Reactor Pressure

The failure probability of the pressurized water reactor pressure vessel for a domestic nuclear power plant in Taiwan has been evaluated according to the technical basis of the USNRC’s new pressurized thermal shock (PTS) screening criteria. The ORNL’s FAVOR code and the PNNL’s flaw models are employed to perform the probabilistic fracture mechanics analysis based on the plant specific parameters of the domestic reactor pressure vessel. Meanwhile, the PTS thermal hydraulic and the probabilistic risk assessment data analyzed from a similar nuclear power plant in the United States for establishing the new PTS rule are applied as the loading condition. Besides, an RT-based regression formula derived by the USNRC is also utilized to verify the through-wall cracking frequencies. It is found that the through-wall cracking of the analyzed reactor pressure vessel only occurs during the PTS events resulted from the stuck-open primary safety relief valves that later reclose, but with only an insignificant failure risk. The results indicate that the Taiwan domestic PWR reactor pressure vessel has sufficient structural margin for the PTS attack until either the end-of-license or for the proposed extended operation.

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