Effect of Initial and Boundary Conditions on the Formation of Reactor Vessel Pressurized Thermal Shock

2013 ◽  
Author(s):  
Maksym Vyshemirskyi ◽  
Aleksander Mazurok ◽  
Anatoliy Nosovskyy
Keyword(s):  
Boundary Conditions ◽  
Thermal Shock ◽  
Nuclear Power ◽  
Reactor Vessel ◽  
Energy Agency ◽  
Detailed Model ◽  
Initial Event ◽  
Pressurized Thermal Shock ◽  
Initial And Boundary Conditions ◽  
Main Steam

The effect of initial and boundary conditions on the reactor vessel Pressurized Thermal Shock (PTS) was analyzed. For example unit 1 of South-Ukrainian Nuclear Power Plant (NPP) with WWER-1000/V-302 reactor was selected. Thermal-hydraulic analysis was made in accordance with methodology of Reactor Pressure Vessel (RPV) brittle fracture analysis by using RELAP5/mod3.2 code. Also, a detailed model of downcomer, which provides a realistic behavior of the fluid, including the mutual mixing of flows with different temperatures was used for these calculations. As an Initial Event (IE) the large secondary leak, namely Steam Generator (SG) Main Steam Line (MSL) break, was considered. According to the International Atomic Energy Agency (IAEA) recommendations all calculations were performed under conservative approach. Also, the asymmetric overcooling was considered. As a result of a series of thermal-hydraulic calculations the most conservative scenario for a group of IE associated with the secondary leaks was obtained.

Fast Neutron Reduction Techniques for Plant Lifetime Extension

2010 ◽  
Author(s):  
Choon Sung Yoo ◽  
Byoung Chul Kim ◽  
Tae Je Kwon
Keyword(s):  
Fast Neutron ◽  
Thermal Shock ◽  
Neutron Irradiation ◽  
Neutron Transport ◽  
Reactor Vessel ◽  
Pressurized Water ◽  
Reduction Techniques ◽  
Pressurized Thermal Shock ◽  
Water Reactors ◽  
Inner Vessel

A Pressurized Thermal Shock (PTS) Event is an event or transient in pressurized water reactors (PWRs) causing severe overcooling (thermal shock) concurrent with or followed by significant pressure in the reactor vessel. A PTS concern arises if one of these transients acts in beltline region of a reactor vessel where a reduced fracture resistance exists due to neutron irradiation. Such an event may produce a flaw or cause the propagation of a flaw postulated to exist near the inner vessel wall surface, thereby potentially affecting the integrity of the vessel. In this paper fast neutron flux reduction techniques were implemented to reduce the potential risk of PTS due to the neutron irradiation on the pressure vessel beltline region. And the RTPTS value for the end of life of the plant was projected using the fast neutron fluence obtained by neutron transport calculations according to the various core loading pattern and reduction program possible for the future cycles.

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.

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.

Stable and Unstable Crack Growth of A508 Class 3 Plates Subjected to Combined Force of Thermal Shock and Tension

1989 ◽  
Vol 111 (3) ◽  
pp. 234-240 ◽  
Author(s):  
G. Yagawa ◽  
Y. Ando ◽  
K. Ishihara ◽  
T. Iwadate ◽  
Y. Tanaka
Keyword(s):  
Thermal Shock ◽  
Nuclear Power ◽  
Power Plants ◽  
Fracture Behavior ◽  
Structural Integrity ◽  
Three Dimensional ◽  
Test Results ◽  
Unstable Crack ◽  
Pressurized Thermal Shock ◽  
Series Of Experiments

An urgent problem for nuclear power plants is to assess the structural integrity of the reactor pressure vessel under pressurized thermal shock. In order to estimate crack behavior under combined force of thermal shock and tension simulating pressurized thermal shock, two series of experiments are demonstrated: one to study the effect of material deterioration due to neutron irradiation on the fracture behavior, and the other to study the effect of system compliance on fracture behavior. The test results are discussed with the three-dimensional elastic-plastic fracture parameters, J and Jˆ integrals.

Stochastic Simulation of PWR Vessel Integrity for Pressurized Thermal Shock Conditions

1984 ◽  
Vol 106 (3) ◽  
pp. 223-229 ◽  
Author(s):  
P. S. Jackson ◽  
D. S. Moelling
Keyword(s):  
Thermal Shock ◽  
Stochastic Simulation ◽  
Bayesian Method ◽  
Reactor Vessel ◽  
Pressurized Water Reactor ◽  
Sampling Errors ◽  
Pressurized Water ◽  
Simulation Methodology ◽  
Pressurized Thermal Shock ◽  
Vessel Integrity

A stochastic simulation methodology is presented for performing probabilistic analyses of Pressurized Water Reactor vessel integrity. Application of the methodology to vessel-specific integrity analyses is described in the context of Pressurized Thermal Shock (PTS) conditions. A Bayesian method is described for developing vessel-specific models of the density of undetected volumetric flaws from ultrasonic inservice inspection results. Uncertainty limits on the probabilistic results due to sampling errors are determined from the results of the stochastic simulation. An example is provided to illustrate the methodology.

A quantitative methodology for reactor vessel pressurized thermal shock decision making

1983 ◽  
Vol 75 (3) ◽  
pp. 405-414 ◽  
Author(s):  
D.S. Ackerson ◽  
K.R. Balkey ◽  
T.A. Meyer ◽  
R.P. Ofstun ◽  
S.D. Rupprecht ◽  
...  
Keyword(s):  
Decision Making ◽  
Thermal Shock ◽  
Reactor Vessel ◽  
Quantitative Methodology ◽  
Pressurized Thermal Shock

Emergency Response Guidance for Reactor Vessel Pressurized Thermal Shock Events

1986 ◽  
Vol 108 (3) ◽  
pp. 346-351
Author(s):  
W. T. Kaiser ◽  
B. S. Monty
Keyword(s):  
Thermal Shock ◽  
Temperature Limit ◽  
Reactor Vessel ◽  
Pressurized Water ◽  
Plant Safety ◽  
Pressurized Thermal Shock ◽  
Water Reactors ◽  
Accident Conditions ◽  
Core Cooling ◽  
Emergency Operating

The operational concern of pressurized thermal shock (PTS) can be minimized by proper operator guidance. This paper presents a method for calculating a pressure temperature limit curve for reactor vessel integrity which can be used to identify an ongoing potential PTS event. This method has been developed for use and is applicable to all pressurized water reactors. The curve is used in emergency operating procedures developed to prioritize various plant safety concerns including PTS and core cooling to ensure proper operator action during accident conditions. This paper emphasizes the development of the pressure-temperature limit and how it is used within the emergency operating procedures.

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