scholarly journals Analysis for the Effect of Spatial Discretization Method on AP1000 Reactor Pressure Vessel Fluence Calculation

2016 ◽  
Vol 2016 ◽  
pp. 1-10 ◽  
Author(s):  
Junxiao Zheng ◽  
Bin Zhang ◽  
Shengchun Shi ◽  
Yixue Chen
Keyword(s):  
Fast Neutron ◽  
Pressure Vessel ◽  
Power Plants ◽  
Structural Integrity ◽  
Neutron Fluence ◽  
Mesh Size ◽  
Reactor Pressure Vessel ◽  
Life Extension ◽  
Discrete Ordinates ◽  
Reactor Pressure

Maintaining the structural integrity of the reactor pressure vessel (RPV) is a critical concern related to the safe operation of nuclear power plants. To estimate the structural integrity over the designed lifetime and to support analyses for a potential plant life extension, an accurate calculation of the fast neutron fluence (E>1.0 MeV orE>0.1 MeV) at the RPV is significant. The discrete ordinates method is one of the main methods to solve such problems. During the calculation process, many factors will affect the results. In this paper, the deviations introduced by different differencing schemes and mesh sizes on the AP1000 RPV fast neutron fluence have been studied, which are based on new discrete ordinates code ARES. The analysis shows that the differencing scheme (diamond difference with or without linear zero fix-up, theta weighted, directional theta weighted, and exponential directional weighted) introduces a deviation within 4%. The coarse mesh (4 × 4 cm meshes inXYplane) leads to approximately 23.7% calculation deviation compared to those of refined mesh (1 × 1 cm meshes inXYplane). Comprehensive study on the deviation introduced by differencing scheme and mesh size has great significance for reasoned evaluation of RPV fast neutron fluence calculation results.

Ex-Vessel Neutron Dosimetry Results in the Vicinity of RPV Supports

2007 ◽  
Author(s):  
Daniel N. Hopkins ◽  
Eugene T. Hayes ◽  
Arnold H. Ferro
Keyword(s):  
Fast Neutron ◽  
Cross Section ◽  
Pressure Vessel ◽  
Transport Theory ◽  
Neutron Fluence ◽  
Reactor Pressure Vessel ◽  
Life Extension ◽  
Neutron Dosimetry ◽  
Pressurized Water ◽  
Reactor Pressure

Neutron-induced embrittlement of the reactor pressure vessel has been a long standing concern for pressurized water reactors (PWR). To date, the beltline region of the pressure vessel, defined as the portion of the pressure vessel experiencing fast neutron fluence (E > 1.0 MeV) equal to or greater than 1017 n/cm2, has been the primary focus of evaluations assessing this embrittlement. These evaluations typically include a calculation of the neutron flux incident on the reactor pressure vessel beltline region, which is in part validated by direct comparison with dosimetry measurements. Two general types of measurements are commonly used, those being dosimetry sets that are included as part of the in-vessel surveillance capsules, and at some plants, those that are included in supplemental surveillance programs such as Ex-Vessel Neutron Dosimetry. In the context of life extension, the beltline region as defined above is getting larger. Present fluence calculations for a number of plants indicate that beltline region at the end of the 60 years of operation will extend to the bottom of the reactor pressure vessel nozzle welds. This extended beltline creates a new problem in terms of validating the neutron fluence calculations in this region well above the top of the active fuel, in that there are no measurements available to confirm calculated results in this new region of interest. Prior to the start-up of Cycle 11 at Comanche Peak Unit 1, an Ex-Vessel Neutron Dosimetry Program was initiated. This program included placement of neutron dosimetry sensor sets in the vicinity of the reactor pressure vessel supports. At the conclusion of Cycle 11, the first set of dosimetry was replaced and the irradiated set analyzed. The Ex-Vessel Neutron Dosimetry set from Cycle 11 was analyzed using a 2D/1D flux synthesis technique using the two dimensional discrete ordinates transport theory calculations (DORT) along with the BUGLE 96 cross-section library and the SNLRML neutron dosimetry cross-section library. The measurements in the vicinity of the vessel supports compare well with the transport calculations, thus confirming that the expected fast neutron fluence (E > 1.0 MeV) in the vicinity of the reactor vessel supports is below the 1018 n/cm2.

Structural Integrity Research for Reactor Pressure Vessel Under In-Vessel Retention of a Core Melt

2016 ◽  
Author(s):  
Yongjian Gao ◽  
Yinbiao He ◽  
Ming Cao ◽  
Yuebing Li ◽  
Shiyi Bao ◽  
...  
Keyword(s):  
Pressure Vessel ◽  
Material Properties ◽  
Power Plants ◽  
Structural Integrity ◽  
Plastic Material ◽  
Plastic Region ◽  
Reactor Pressure Vessel ◽  
Elastic Plastic ◽  
Elastic Plastic Material ◽  
Reactor Pressure

In-Vessel Retention (IVR) is one of the most important severe accident mitigation strategies of the third generation passive Nuclear Power Plants (NPP). It is intended to demonstrate that in the case of a core melt, the structural integrity of the Reactor Pressure Vessel (RPV) is assured such that there is no leakage of radioactive debris from the RPV. This paper studied the IVR issue using Finite Element Analyses (FEA). Firstly, the tension and creep testing for the SA-508 Gr.3 Cl.1 material in the temperature range of 25°C to 1000°C were performed. Secondly, a FEA model of the RPV lower head was built. Based on the assumption of ideally elastic-plastic material properties derived from the tension testing data, limit analyses were performed under both the thermal and the thermal plus pressure loading conditions where the load bearing capacity was investigated by tracking the propagation of plastic region as a function of pressure increment. Finally, the ideal elastic-plastic material properties incorporating the creep effect are developed from the 100hr isochronous stress-strain curves, limit analyses are carried out as the second step above. The allowable pressures at 0 hr and 100 hr are obtained. This research provides an alternative approach for the structural integrity evaluation for RPV under IVR condition.

Nuclear Reactor Pressure Vessel Integrity Assessment: Enhancement Provided by 3D Modeling Flaw Stability Calculations

2013 ◽  
Author(s):  
Adolfo Arrieta-Ruiz ◽  
Eric Meister ◽  
Henriette Churier
Keyword(s):  
Pressure Vessel ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Reactor ◽  
Structural Integrity ◽  
Reactor Pressure Vessel ◽  
Integrity Assessment ◽  
Potential Damage ◽  
Vessel Integrity ◽  
Reactor Pressure

Structural integrity of the Reactor Pressure Vessel (RPV) is one of the main considerations regarding safety and lifetime of Nuclear Power Plants (NPP) since this component is considered as not reasonably replaceable. Brittle fracture risk associated with the embrittlement of RPV steel in irradiated areas is the main potential damage. In France, deterministic integrity assessment for RPV is based on the crack initiation stage. The stability of an under-clad postulated flaw in the core area is currently evaluated under a Pressurized Thermal Shock (PTS) through a fracture mechanics simplified method. One of the axes of EDF’s implemented strategy for NPP lifetime extension is the improvement of the deterministic approach with regards to the input data and methods so as to reduce conservatisms. In this context, 3D finite element elastic-plastic calculations with flaw modelling have been carried out recently in order to quantify the enhancement provided by a more realistic approach in the most severe events. The aim of this paper is to present both simplified and 3D modelling flaw stability evaluation methods and the results obtained by running a small break LOCA event.

Verifying Calculation of Reactor Pressure Vessel Fast Neutron Fluence

2014 ◽  
Vol 986-987 ◽  
pp. 985-989
Author(s):  
Qiao Feng Liu ◽  
Jing Ru Han ◽  
Hai Ying Chen ◽  
Chun Ming Zhang
Keyword(s):  
Standard Deviation ◽  
Fast Neutron ◽  
Radiation Protection ◽  
Pressure Vessel ◽  
Neutron Fluence ◽  
Reactor Pressure Vessel ◽  
Light Water Reactor ◽  
Light Water ◽  
Water Reactor ◽  
Reactor Pressure

The reactor pressure vessel is an unchangeable component of the light water reactor. To some extent, the life of the pressure vessel depends on the fast neutron fluence. In addition, the fast neutron fluence is an important parameter for radiation protection. So, the fast neutron fluence is one of the main parameters which should be verifying calculated by the reviewers. The verifying calculation of the fast neutron fluence of one reactor pressure vessel is presented in this paper, and the standard deviation between the verifying and designing calculations is lower than 10%. The reasons for the deviation are discussed.

Boiling Water Reactor Pressure Vessel Integrity Evaluation by Probabilistic Fracture Mechanics

2010 ◽  
Author(s):  
Bo-Yi Chen ◽  
Chin-Cheng Huang ◽  
Hsoung-Wei Chou ◽  
Ru-Feng Liu ◽  
Hsien-Chou Lin
Keyword(s):  
Fracture Mechanics ◽  
Conditional Probability ◽  
Pressure Vessel ◽  
Structural Integrity ◽  
Neutron Fluence ◽  
Probability Of Failure ◽  
Life Extension ◽  
Water Reactor ◽  
License Renewal ◽  
Reactor Pressure

The Chinshan boiling water reactor (BWR) units 1 and 2, owned by Taiwan Power Company (TPC), started commercial operations in 1978 and 1979, respectively. The reactor pressure vessel (RPV) welds unavoidably degrade with the long time operation because of the fast-neutron fluence exposure. This effect should be considered in the life extension and license renewal application. Thus, the structural integrity of the axial and circumferential welds at the beltline region of reactor vessel must be evaluated carefully. The probabilistic fracture mechanics (PFM) analysis code: Fracture Analysis of Vessels – Oak Ridge (FAVOR), which has been verified by USNRC, is adopted in this work to calculate the conditional probability of initiation (CPI) and the conditional probability of failure (CPF) for the welds with 32 and 64 effective full power years (EFPY) operation, respectively. The Monte Carlo technique is involved in the simulation. This is the first time that the PFM technique is adopted for evaluating the risk of nuclear power plant components in Taiwan. Actual geometries, material properties, chemistry components, neutron fluence and operation conditions are used for the plant specific analyses. Moreover, the design basis transients/accidents described in the final safety analysis report are also taken into account. The computed results show that the failure probabilities of welds are less than 10−10 per year. Only the axial weld, W-1001-08, is found to have the probability of failure. 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 analysis results for the life extension and the license renewal applications.

On the effects of oxygen cross-sections in the fast neutron fluence analysis for a reactor pressure vessel scraping test

2008 ◽  
Vol 35 (4) ◽  
pp. 565-569 ◽  
Author(s):  
Alexander Vasiliev ◽  
Hakim Ferroukhi ◽  
Martin A. Zimmermann ◽  
Rakesh Chawla
Keyword(s):  
Fast Neutron ◽  
Cross Sections ◽  
Pressure Vessel ◽  
Neutron Fluence ◽  
Reactor Pressure Vessel ◽  
Reactor Pressure

Boiling Water Reactor Pressure Vessel Integrity Evaluation by Probabilistic Fracture Mechanics (PVP2010-25195)

2012 ◽  
Vol 135 (1) ◽  
Author(s):  
Bo-Yi Chen ◽  
Chin-Cheng Huang ◽  
Hsoung-Wei Chou ◽  
Ru-Feng Liu ◽  
Hsien-Chou Lin
Keyword(s):  
Fracture Mechanics ◽  
Conditional Probability ◽  
Pressure Vessel ◽  
Structural Integrity ◽  
Neutron Fluence ◽  
Probability Of Failure ◽  
Reactor Pressure Vessel ◽  
Circumferential Welds ◽  
Probabilistic Fracture Mechanics ◽  
Reactor Pressure

The reactor pressure vessel (RPV) welds unavoidably degrade with the long time operation because of the fast neutron fluence exposure. Thus, the structural integrity of the axial and circumferential welds at the beltline region of reactor vessel must be evaluated carefully. The probabilistic fracture mechanics (PFM) analysis code: Fracture analysis of vessels—Oak Ridge (FAVOR), which has been verified by USNRC, is adopted in this work to calculate the conditional probability of initiation (CPI) and the conditional probability of failure (CPF) for the welds with 32 and 64 effective full power years (EFPY) operation, respectively. The Monte Carlo technique is involved in the simulation. This is the first time that the PFM technique is adopted for evaluating the risk of nuclear power plant components in Taiwan. Actual geometries, material properties, alloying elements, neutron fluence, and operation conditions are used for the plant specific analyses. Moreover, the design basis transients/accidents described in the final safety analysis report are also taken into account. The computed results show that the failure probabilities of welds are less than 10−10 per year. Only the axial weld, W-1001-08 has the probability of failure. 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 analysis results for the life extension and the license renewal applications.

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