Effect of Pressurized Water Reactor Environment on Material Parameters of 316 Stainless Steel: A Cyclic Plasticity Based Evolutionary Material Modeling Approach

2015 ◽  
Author(s):  
Subhasish Mohanty ◽  
William K. Soppet ◽  
Saurindranath Majumdar ◽  
Krishnamurti Natesan
Keyword(s):  
Stainless Steel ◽  
Fatigue Life ◽  
Nuclear Reactor ◽  
Axial Stress ◽  
Cyclic Plasticity ◽  
Model Parameters ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Test Conditions ◽  
Over Time

At present, the fatigue life of nuclear reactor components is estimated based on empirical approaches, such as stress/strain versus life (S∼N) curves and Coffin-Manson type empirical relations. In most cases, the S∼N curves are generated from uniaxial fatigue test data, which may not truly represent the multi-axial stress state at the component level. Also, the S∼N curves are based on the final life of the specimen, which may not accurately represent the mechanistic time-dependent evolution of material behavior. These discrepancies lead to large uncertainties in fatigue life estimations. We propose a modeling approach based on evolutionary cyclic plasticity that can be used for developing finite element models of nuclear reactor components subjected to multi-axial stress states. These models can be used for more accurately predicting the stress-strain evolution over time in reactor components and, in turn, fatigue life. The model parameters were estimated for 316 stainless steel material, which are widely used in U.S. nuclear reactors. The model parameters were estimated for different test conditions to understand their evolution over time and their sensitivity to particular test conditions, such as the pressurized water reactor coolant condition.

scholarly journals Finite Element Based Full-Life Cyclic Stress Analysis of 316 Grade Nuclear Reactor Stainless Steel Under Constant, Variable, and Random Fatigue Loading

2018 ◽  
Vol 140 (5) ◽  
Author(s):  
Bipul Barua ◽  
Subhasish Mohanty ◽  
Joseph T. Listwan ◽  
Saurindranath Majumdar ◽  
Krishnamurti Natesan
Keyword(s):  
Finite Element ◽  
Stainless Steel ◽  
Stress Analysis ◽  
Nuclear Reactor ◽  
Three Dimensional ◽  
Fatigue Loading ◽  
Cyclic Stress ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Fatigue Evaluation

Although S∼N curve-based approaches are widely followed for fatigue evaluation of nuclear reactor components and other safety critical structural systems, there is a chance of large uncertainty in estimated fatigue lives. This uncertainty may be reduced by using a more mechanistic approach such as physics based three-dimensional (3D) finite element (FE) methods. In a recent paper (Barua et al., 2018, ASME J. Pressure Vessel Technol., 140(1), p. 011403), a fully mechanistic fatigue modeling approach which is based on time-dependent stress–strain evolution of material over the entire fatigue life was presented. Based on this approach, in this work, FE-based cyclic stress analysis was performed on 316 nuclear grade reactor stainless steel (SS) fatigue specimens, subjected to constant, variable, and random amplitude loading, for their entire fatigue lives. The simulated results are found to be in good agreement with experimental observation. An elastic-plastic analysis of a pressurized water reactor (PWR) surge line (SL) pipe under idealistic fatigue loading condition was performed and compared with experimental results.

Rupture Hardware Minimization in Pressurized Water Reactor Piping

1989 ◽  
Vol 111 (1) ◽  
pp. 64-71 ◽  
Author(s):  
S. K. Mukherjee ◽  
J. J. Szy Slow Ski ◽  
V. Chexal ◽  
D. M. Norris ◽  
N. A. Goldstein ◽  
...  
Keyword(s):  
Stainless Steel ◽  
Ferritic Steel ◽  
High Energy ◽  
Pressurized Water Reactor ◽  
Pipe Failure ◽  
Pressurized Water ◽  
Screening Criteria ◽  
Water Reactor ◽  
Leak Before Break ◽  
Generic Sense

For much of the high-energy piping in light water reactor systems, fracture mechanics calculations can be used to assure pipe failure resistance, thus allowing the elimination of excessive rupture restraint hardware both inside and outside containment. These calculations use the concept of leak-before-break (LBB) and include part-through-wall flaw fatigue crack propagation, through-wall flaw detectable leakage, and through-wall flaw stability analyses. Performing these analyses not only reduces initial construction, future maintenance, and radiation exposure costs, but also improves the overall safety and integrity of the plant since much more is known about the piping and its capabilities than would be the case had the analyses not been performed. This paper presents the LBB methodology applied at Beaver Valley Power Station—Unit 2 (BVPS-2); the application for two specific lines, one inside containment (stainless steel) and the other outside containment (ferritic steel), is shown in a generic sense using a simple parametric matrix. The overall results for BVPS-2 indicate that pipe rupture hardware is not necessary for stainless steel lines inside containment greater than or equal to 6-in. (152-mm) nominal pipe size that have passed a screening criteria designed to eliminate potential problem systems (such as the feedwater system). Similarly, some ferritic steel line as small as 3-in. (76-mm) diameter (outside containment) can qualify for pipe rupture hardware elimination.

Depletion Calculation for a Nodal Reactor Physics Code

2010 ◽  
Author(s):  
Antonio Carlos Marques Alvim ◽  
Fernando Carvalho da Silva ◽  
Aquilino Senra Martinez
Keyword(s):  
Diffusion Equation ◽  
Nuclear Reactor ◽  
Reactor Core ◽  
Expansion Method ◽  
Design System ◽  
Pressurized Water Reactor ◽  
Reactor Physics ◽  
Pressurized Water ◽  
Nodal Method ◽  
Orthogonal Polynomial Expansion

This paper deals with an alternative numerical method for calculating depletion and production chains of the main isotopes found in a pressurized water reactor. It is based on the use of the exponentiation procedure coupled to orthogonal polynomial expansion to compute the transition matrix associated with the solution of the differential equations describing isotope concentrations in the nuclear reactor. Actually, the method was implemented in an automated nuclear reactor core design system that uses a quick and accurate 3D nodal method, the Nodal Expansion Method (NEM), aiming at solving the diffusion equation describing the spatial neutron distribution in the reactor. This computational system, besides solving the diffusion equation, also solves the depletion equations governing the gradual changes in material compositions of the core due to fuel depletion. The depletion calculation is the most time-consuming aspect of the nuclear reactor design code, and has to be done in a very precise way in order to obtain a correct evaluation of the economic performance of the nuclear reactor. In this sense, the proposed method was applied to estimate the critical boron concentration at the end of the cycle. Results were compared to measured values and confirm the effectiveness of the method for practical purposes.

Cyclic Ramberg-Osgood Parameters of 316SS Weld Metal Under In-Air and PWR-Water Environments

2021 ◽  
Author(s):  
Jae Phil Park ◽  
Subhasish Mohanty ◽  
Chi Bum Bahn
Keyword(s):  
Room Temperature ◽  
Cyclic Plasticity ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Shakedown Analysis ◽  
Life Estimation ◽  
Fatigue Life Estimation ◽  
Tensile Curve ◽  
Asme Code ◽  
Primary Water

Abstract At present the available Ramberg-Osgood (R-O) parameters for different metals (e.g. in ASME code and other literature) are static (generally based on a tensile curve). These static R-O parameters cannot accurately model the cyclic plasticity behavior. This work presents the cyclic R-O material hardening parameters for 316 stainless steel similar metal welds. The parameters were estimated under various conditions (in-air at room temperature, 300°C in-air, and in-air at primary water conditions for a pressurized water reactor (PWR)). It is anticipated that the reported results would be useful for computational mechanics based shakedown analysis and fatigue life estimation of PWR components.

Tensile Tests for Cast Stainless Steel: Evolution of the RCC-M Code

2018 ◽  
Author(s):  
Arnaud Blouin ◽  
Mathieu Couvrat ◽  
Félix Latourte ◽  
Julian Soulacroix
Keyword(s):  
Tensile Strength ◽  
Stainless Steel ◽  
Ultimate Tensile Strength ◽  
Tensile Tests ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Different Diameters ◽  
Cast Stainless Steel ◽  
Acceptance Tests ◽  
Do So

In the framework of a pressurized water reactor primary loop replacement, elbows of different types were produced in cast austenitic stainless steel grade Z3CN 20-09 M. For that type of component, acceptance tests to check the sufficient mechanical properties include room and hot temperature tensile tests, following the RCC-M CMS – 1040 and EN 10002 specifications. A large test campaign on standard 10mm diameter specimens was performed and exhibited a high scattering in yield stress and ultimate tensile strength values. As a consequence, some acceptance tensile tests failed to meet the required minimal values, especially the ultimate tensile strength. Optical and electronic microscopy revealed that the low values were due to the presence of very large grain compared to the specimen gage diameter. However, tensile tests strongly rely on the hypothesis that the specimen gage part can be considered as a representative volume element containing a number of grains large enough so that their variation in size and orientation gives a homogeneous response. To confirm the origin of the scattering, a huge experimental tensile test campaign with specimens of different diameters was conducted. In parallel, FE calculations were also performed. From all those results, it was concluded that it was necessary to improve the RCC-M code for that type of test for cast stainless steel: to do so, a modification sheet was sent and is being investigated by AFCEN.

Study on the Viability of the Recycling by Electric Arc Melting of Zirconium Alloys Scraps Aiming the Scalability of the Process

2018 ◽  
Vol 930 ◽  
pp. 495-500
Author(s):  
Cristiano Stefano Mucsi ◽  
L.A.M. dos Reis ◽  
Maurilio Pereira Gomes ◽  
L.A.T. Pereira ◽  
Jesualdo Luiz Rossi
Keyword(s):  
Nuclear Reactor ◽  
Zirconium Alloys ◽  
Vacuum Arc ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Vacuum Arc Remelting ◽  
Arc Melting ◽  
Vacuum Arc Furnace ◽  
Fuel Elements

Turning chips of zirconium alloys are produced in large quantities during the machining of alloy rods for the fabrication of the end plugs for the Pressurized Water Reactor (PWR) fuel elements parts of Angra II nuclear reactor (Brazil – Rio de Janeiro). This paper presents a study on the search for an efficient way for the cleaning, quality control and Vacuum Arc Remelting (VAR) of pressed zirconium alloys chips to produce a material viable to be used in the production of the fuel rod end plugs. The process starts with cutting oil clean out. The first step in this process consists in soaking a bunch of chips in clean water, to remove soluble cutting oils, followed by an alkaline degreasing bath and a wash with a high-pressure flow of water. Drying is performed by a flux of warm air. The oil free chips are then subjected to a magnet in order to detect and collect any magnetic material, essentially ferrous, that may be present in the original chips. Samples of the material are collected and then melted in a small non consumable electrode vacuum arc furnace for evaluation by Energy Dispersive X-ray Fluorescence Spectrometry (EDXRFS) in order to define the quality of the chips. The next step consists in the 15 ton hydraulic pressing the chips in a die with 40 mm square section and 500 mm long, producing an electrode with 20% of the Zircaloy bulk density. The electrode was finally melted in a laboratory scale modified VAR furnace located at the CCTM–IPEN, producing 0.8 kg ingots. The authors conclude that the samples obtained from the fuel element industry can be melting in a VAR furnace, modified to accommodate low density electrodes, allowing a reduction up to 40 times the original storage volume, however, it is necessary to remelt the ingots to correct their composition in order to recycle the original zirconium alloys chips. in a process to reduce volume and allow the reutilization of valuable Zircaloy scraps.

Simulation of Triaxial Residual Stress Mapping for a Hollow Cylinder

2012 ◽  
Author(s):  
Mitchell D. Olson ◽  
Wilson Wong ◽  
Michael R. Hill
Keyword(s):  
Residual Stress ◽  
Stainless Steel ◽  
Nuclear Power ◽  
Complex Geometry ◽  
Cylindrical Body ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Weld Residual Stress ◽  
Novel Method ◽  
Stress Mapping

This paper describes a novel method to determine a two-dimensional map of the triaxial residual stress on a radial-axial plane of interest in a hollow cylindrical body. With the description in hand, we present a simulation to validate the steps of the method. The simulation subject is a welded cylindrical nozzle typical of a nuclear power pressurized water reactor pressurizer; in the weld region, the nozzle inner diameter is roughly 132 mm (5.2 inch) and the wall thickness is roughly 35 mm (1.4 inch). The pressure vessel side of the nozzle is carbon steel (with a thin stainless steel lining), the piping side is austenitic stainless steel, and between the two are weld and buttering deposits of nickel alloy. Weld residual stresses in such nozzles have important effects on crack growth rates in fatigue and stress corrosion cracking, therefore measurements of weld residual stress can help provide inputs for managing aging reactor fleets. Nuclear power plant welds often have large and complex geometry, which has made residual stress measurements difficult, and this work provides a proof of concept for a new experimental technique for measurements on welded nozzles.

scholarly journals Fatigue Damage Management According to Performance Based Maintenance (PBM) Concept

2021 ◽  
Author(s):  
Masayuki Kamaya
Keyword(s):  
Fatigue Life ◽  
Crack Growth ◽  
Thermal Loading ◽  
Water Environment ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Term Operation ◽  
Primary Water ◽  
Crack Growth Curve

Abstract A maintenance concept of performance based maintenance (PBM) has been proposed by the current author. According to the PBM concept, inspection results are considered in determining the next inspection schedule. In this study, this concept was applied to fatigue degradation for stainless steel components in the pressurized water reactor (PWR) primary water environment. It is possible to estimate the fatigue life for the PWR water environment from that obtained in an air environment and the parameter Fen, which represents the ratio of the fatigue life in the air and PWR water environments. It was shown that the fatigue life prediction using Fen can be replaced by the crack growth analysis using the growth rate for the PWR water environment. Then, the crack growth was predicted for a thermal loading assuming the growth occurred in the PWR water environment. It was shown that the duration until the next inspection could be optimized based on the inspection results together with the crack growth curve. A long term operation before the inspection resulted in a longer duration until the next inspection.

A Study of Cavitation Erosion

2002 ◽  
Author(s):  
Hiromu Isaka ◽  
Masatsugu Tsutsumi ◽  
Hiroyuki Kobayashi ◽  
Tadashi Shiraishi
Keyword(s):  
Stainless Steel ◽  
Erosion Rate ◽  
Cavitation Erosion ◽  
Design Criteria ◽  
Hard Material ◽  
Test Results ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Type Stainless Steel ◽  
Critical Cavitation

The authors performed experimental study for the purpose of the following two items from a viewpoint of cavitation erosion of a cylindrical orifice in view of a problem at the letdown orifice in PWR (Pressurized Water Reactor). 1. To get the critical cavitation parameter of the cylindrical orifice to establish the design criteria for prevention of cavitation erosion, and 2. to ascertain the erosion rate in such an eventuality that the cavitation erosion occurs with the orifice made of stainless steel with precipitation hardening (17-4-Cu hardening type stainless steel), so that we confirm the appropriateness of the design criteria. Regarding the 1st item, we carried out the cavitation tests to get the critical cavitation parameters inside and downstream of the orifice. The test results showed that the cavitation parameter at inception is independent of the length or the diameter of the orifice. Moreover, the design criteria of cavitation erosion of cylindrical orifices have been established. Regarding the 2nd item, we tested the erosion rate under high-pressure conditions. The cavitation erosion actually occurred in the cylindrical orifice at the tests that was strongly resemble to the erosion occurred at the plant. It will be seldom to reproduce resemble cavitation erosion in a cylindrical orifice with the hard material used at plants. We could establish the criteria for preventing the cavitation erosion from the test results.

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