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.

scholarly journals Effect of Nitrogen Ion Implantation on the Cavitation Erosion Resistance and Cobalt-Based Solid Solution Phase Transformations of HIPed Stellite 6

Materials ◽  
2021 ◽  
Vol 14 (9) ◽  
pp. 2324
Author(s):  
Mirosław Szala ◽  
Dariusz Chocyk ◽  
Anna Skic ◽  
Mariusz Kamiński ◽  
Wojciech Macek ◽  
...  
Keyword(s):  
Plastic Deformation ◽  
Stainless Steel ◽  
Ion Implantation ◽  
Phase Transformations ◽  
Erosion Rate ◽  
Cavitation Erosion ◽  
Matrix Phase ◽  
Nitrogen Ion Implantation ◽  
Stellite 6 ◽  
Nitrogen Ion

From the wide range of engineering materials traditional Stellite 6 (cobalt alloy) exhibits excellent resistance to cavitation erosion (CE). Nonetheless, the influence of ion implantation of cobalt alloys on the CE behaviour has not been completely clarified by the literature. Thus, this work investigates the effect of nitrogen ion implantation (NII) of HIPed Stellite 6 on the improvement of resistance to CE. Finally, the cobalt-rich matrix phase transformations due to both NII and cavitation load were studied. The CE resistance of stellites ion-implanted by 120 keV N+ ions two fluences: 5 × 1016 cm−2 and 1 × 1017 cm−2 were comparatively analysed with the unimplanted stellite and AISI 304 stainless steel. CE tests were conducted according to ASTM G32 with stationary specimen method. Erosion rate curves and mean depth of erosion confirm that the nitrogen-implanted HIPed Stellite 6 two times exceeds the resistance to CE than unimplanted stellite, and has almost ten times higher CE reference than stainless steel. The X-ray diffraction (XRD) confirms that NII of HIPed Stellite 6 favours transformation of the ε(hcp) to γ(fcc) structure. Unimplanted stellite ε-rich matrix is less prone to plastic deformation than γ and consequently, increase of γ phase effectively holds carbides in cobalt matrix and prevents Cr7C3 debonding. This phenomenon elongates three times the CE incubation stage, slows erosion rate and mitigates the material loss. Metastable γ structure formed by ion implantation consumes the cavitation load for work-hardening and γ → ε martensitic transformation. In further CE stages, phases transform as for unimplanted alloy namely, the cavitation-inducted recovery process, removal of strain, dislocations resulting in increase of γ phase. The CE mechanism was investigated using a surface profilometer, atomic force microscopy, SEM-EDS and XRD. HIPed Stellite 6 wear behaviour relies on the plastic deformation of cobalt matrix, starting at Cr7C3/matrix interfaces. Once the Cr7C3 particles lose from the matrix restrain, they debond from matrix and are removed from the material. Carbides detachment creates cavitation pits which initiate cracks propagation through cobalt matrix, that leads to loss of matrix phase and as a result the CE proceeds with a detachment of massive chunk of materials.

The Analytical and Experimental Research of In-Vessel Corium Retention in CPR1000

2010 ◽  
Author(s):  
Xing Chen ◽  
Shishun Zhang ◽  
Jiming Lin ◽  
Huiyong Zhang
Keyword(s):  
Experimental Research ◽  
Success Probability ◽  
Outer Wall ◽  
Test Results ◽  
Pressurized Water Reactor ◽  
Structural Failure ◽  
Pressurized Water ◽  
Thermal Failure ◽  
Near Future ◽  
Preliminary Phase

The analytical and experiment research of In-Vessel Corium Retention (IVR) in the Chinese Pressurized-water Reactor 1000 MWe (CPR1000) are introduced. The IVR research consists of preliminary phase and detailed phase. The analysis of thermal failure, structural failure and penetration failure of Reactor Pressure Vessel (RPV) and the experimental research of External Reactor Vessel Cooling (ERVC) are performed at preliminary phase. Analysis results show that the RPV failure is the dominated by thermal failure mode and the probability of the thermal failure is very low. Test results show that the IVR success probability for CPR1000 is about 99% if the Critical Heat Flux (CHF) of CPR1000 is the same as that of AP600. Further works, including the ERVC enhancement design, the CHF test of the RPV outer wall and the recalculation of the IVR success probability for CPR1000, will be performed at detailed phase in the near future.

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.

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.

PKL Tests on Heterogeneous Inherent Boron Dilution Following SB-LOCA: Applicability to Reactor Scale

2010 ◽  
Author(s):  
Klaus Umminger ◽  
Simon Philipp Schollenberger ◽  
Se´bastien Cornille ◽  
Claire Agnoux ◽  
Delphine Quintin ◽  
...  
Keyword(s):  
Cooling System ◽  
Natural Circulation ◽  
Limited Range ◽  
Test Facility ◽  
Test Results ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
The Core ◽  
Physical Phenomena ◽  
Boron Dilution

In the course of a small break LOCA in a Pressurized Water Reactor (PWR) the flow regime in the Reactor Cooling System (RCS) passes through a number of different phases and the filling level may decrease down to the point where the decay heat is transferred to the secondary side under Reflux-Condenser (RC) conditions. During RC, the steam formed in the core condensates in the Steam Generator (SG) U-tubes. For a limited range of break size and configuration, a continuous accumulation of condensate may cause the formation of boron-depleted slugs. If natural circulation reestablishes, as the RCS is refilled, boron-depleted slugs might be transported to the Reactor Pressure Vessel (RPV) and to the core. To draw conclusions on the risk of boron dilution processes in SB-LOCA transients, two important issues, the limitation of slug size and the onset of Natural Circulation (NC) have to be assessed on the basis of experimental data, as system Thermal-Hydraulic codes are limited in their capability to replicate the complex physical phenomena involved. The OECD PKL III tests were performed at AREVA’s PKL test facility in Erlangen, Germany, to evaluate important phases of the boron dilution transient in PWRs. Several integral and separate effect tests were conducted, addressing the inherent boron dilution issue. The PKL III integral transient test runs provide sufficient data to state major conclusions on the formation and maximum possible size of the boron-depleted slugs, their boron concentration and their transport into the RPV with the restart of NC. Some of these conclusions can be applied to reactor scale. It has to be mentioned, that even though this paper is based on PKL test results obtained within the OECD PKL project, the conclusions of this paper reflect the views of the authors and not necessarily of all the members of the OECD PKL project.

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.

Oxidation resistance and stress corrosion cracking susceptibility of 308L and 309L stainless steel cladding layers in simulated PWR primary water

CORROSION ◽  
10.5006/3699 ◽  
2021 ◽  
Author(s):  
Tongming Cui ◽  
Qi Xiong ◽  
Jiarong Ma ◽  
Kun Zhang ◽  
Zhanpeng LU ◽  
...  
Keyword(s):  
Stainless Steel ◽  
Oxidation Resistance ◽  
Layer Structure ◽  
Initiation Site ◽  
Pressurized Water Reactor ◽  
Internal Oxide ◽  
Pressurized Water ◽  
The Matrix ◽  
Primary Water ◽  
Cladding Layers

Exposure and slow strain rate tensile (SSRT) tests were conducted in a simulated pressurized water reactor (PWR) primary water to investigate the oxidation resistance and SCC susceptibility of 308L and 309L stainless steel (SS) cladding layers. A double-layer structure oxide layer grown on 308L SS and 309L SS contained the Cr-enriched nanocrystalline internal layer and the Fe-enriched spinel oxide in the external layer. Ni-enrichment at the matrix/oxide (M/O) boundary was observed. The internal oxide film on 309L SS was thicker and had a lower Cr content than that on 308L SS. Preferential dissolution of inclusions led to pits on 308L SS and 309L SS surfaces during the exposure tests. More inclusions in 309L would decrease its SCC resistance due to the pits can act as the SCC initiation site. 308L SS had a lower susceptibility of SCC than 309L SS in PWR primary water. Lower ferrite content, higher strength/hardness reduced the oxidation and SCC resistance of 309L SS cladding. The effect of ferrite on oxidation and SCC of the SS claddings was discussed.

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.

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