Use of Subsize Specimens for Determination of Radiation Embrittlement of Operating Reactor Pressure Vessels

2009 ◽  
pp. 424-424-16 ◽  
Author(s):  
AD Amayev ◽  
VI Badanin ◽  
AM Kryukov ◽  
VA Nikolayev ◽  
MF Rogov ◽  
...  
Keyword(s):  
Pressure Vessels ◽  
Radiation Embrittlement ◽  
Operating Reactor ◽  
Reactor Pressure

Determination of Copper in A533b Steel for the Assessment of Radiation Embrittlement Using Laser-Induced Breakdown Spectroscopy

1996 ◽  
Vol 50 (3) ◽  
pp. 306-309 ◽  
Author(s):  
Wolfgang E. Ernst ◽  
Dave F. Farson ◽  
D. Jason Sames
Keyword(s):  
Laser Beam ◽  
Fiber Optics ◽  
Nuclear Power ◽  
Power Plants ◽  
Pressure Vessels ◽  
Laser Induced Breakdown Spectroscopy ◽  
Radiation Embrittlement ◽  
Breakdown Spectroscopy ◽  
Laser Induced Breakdown

Determination of radiation embrittlement in nuclear reactor pressure vessels is crucial to assessing safe operative lifetimes for many aging nuclear power plants. Conservative nuclear fluence estimates and trace impurity diagnosis of the weldment material are the basis of radiation embrittlement analysis. Copper is thought to be a key impurity contributing to radiation embrittlement. In this paper, the application of laser-induced breakdown spectroscopy (LIBS) as a means to assess radiation embrittlement by the detection and quantification of copper in A553b steel was investigated. A LIBS configuration completely coupled by fiber optics was attempted, but because of low laser power and fiber losses, fiber-optic delivery of the laser beam was unsuccessful. Consequently, hard optics (lenses and mirrors) were employed for laser beam delivery. The plasma emission was delivered successfully via fiber optics to the detection apparatus. Copper measurements were made from custom-fabricated steel samples. Comparison of the LIBS results to an independent atomic absorption spectrophotometry (AAS) analysis showed LIBS to be of comparable accuracy, especially in low-level copper samples.

Precise Determination of Trace Hydrogen in SA508-3 Steel for Nuclear Reactor Pressure Vessels

2018 ◽  
Vol 10 (11) ◽  
pp. 1651-1657 ◽  
Author(s):  
Luhan Hao ◽  
Wei Zhao ◽  
Yan Peng ◽  
Mingyue Sun ◽  
Dianzhong Li ◽  
...  
Keyword(s):  
Nuclear Reactor ◽  
Precise Determination ◽  
Pressure Vessels ◽  
Reactor Pressure

Residual Stresses in Austenitic Cladding

2007 ◽  
Author(s):  
Milan Brumovsky ◽  
Vladislav Pistora ◽  
Ivan Kupka
Keyword(s):  
Residual Stresses ◽  
Water Environment ◽  
Strength Properties ◽  
Pressure Vessels ◽  
Primary Circuit ◽  
Adjacent Area ◽  
Pressurized Thermal Shock ◽  
Welding Direction ◽  
Reactor Pressure

Reactor pressure vessels (RPV) are usually manufactured with austenitic cladding on their inner surface as a protection against corrosion from the primary circuit water environment. Thus, they are not included into the strength calculations of pressure vessels due to their lower strength properties and much smaller thickness in comparison with those of vessels as they are taken only as a corrosion layer. In the same time, due to different thermal coefficients and Young moduli, welding of austenitic cladding results in a high residual stresses in the cladding and also in the adjacent area in the base ferritic metal. These residual stresses as well as stresses resulted from the temperature field in the vessels represent necessary inputs into pressurized thermal shock calculations. WWER (Water-Water Energy Reactor = PWR type) reactor pressure vessels have relatively thick cladding — nominally 8 mm — made from two layers: first layer of 25/10 type welded by one pass while the second layer of 18/10/Ti typed is usually welded by three passes. The main part of the vessels was performed by strip welding with strips of 60 mm wide. Results of residual stresses measurements are given in the paper. Method with incremental milling of beams was used for the measurements and determination of residual stresses. Tests were performed on specimens in as-welded state and also after final heat treatment of the vessels, i.e. after several stress relieves including first hydrotest in shop. As residual stresses depends strongly also on direction of welding, beams were oriented in both directions — parallel and perpendicular to the welding direction. Results of these measurements are shown and discussed in the paper.

Current Issues in Operated WWER Type RPVs

2008 ◽  
Author(s):  
Milan Brumovsky´ ◽  
Milos Kytka
Keyword(s):  
Fracture Toughness ◽  
Pressure Vessels ◽  
Point Of View ◽  
Radiation Embrittlement ◽  
Irradiation Effects ◽  
Static Fracture ◽  
High Level ◽  
Near Future ◽  
Surveillance Specimens ◽  
Reactor Pressure

RPVs of WWER type reactors are manufactured from other type of steels (15Kh2MFA of Cr-Mo-V type for WWER-440 and 15Kh2NMFA of Ni-Cr-Mo-V type for WWER-1000) and according to other Codes and standards than PWR ones, thus some specific problems are currently more important for WWER. The principal problem lies in relatively small number of manufactured and operated WWER type NPPs. Even though a high level of unification in RPVs exists — practically only two designs of RPVs exists (WWER-440 and WWER-1000) — total number is still small. All WWER-440 RPV are practically identical, either they were manufactured for V-230 or V-213 model: the only difference is in the purity of used materials and existence/non-existence of the surveillance programmes. (Fact that some V-230 type vessels were not covered by austenitic cladding is not important from irradiation effects point of view.) Regarding WWER-440/V-230 types, it is necessary to take into account, that even though most of them were successfully annealed, only some of them are still in operation but most of them will be closed in near future. Similar situation is with WWER-1000 RPVs, either they were manufactured for V-320 (most frequent), or V-338 or the newest V-428 — differences are practically only in the content of nickel in critical weldments and/or in design of surveillance specimens capsules. But, Large advantage of all WWER surveillance programmes is in loading static fracture toughness specimens in all programmes. The papers tries to summarize and analyze all current issues connected with radiation embrittlement of operated reactor pressure vessels of WWER type.

Applicability of Sub-Size Charpy Specimens to Transition Behaviour of RPV Materials

2007 ◽  
Vol 348-349 ◽  
pp. 977-980
Author(s):  
Michal Falcnik ◽  
Petr Novosad ◽  
Pavel Pesek ◽  
Mylos Kytka
Keyword(s):  
Fracture Toughness ◽  
Dynamic Fracture ◽  
Standard Specimen ◽  
Dynamic Fracture Toughness ◽  
Radiation Embrittlement ◽  
Toughness Testing ◽  
Transition Behaviour ◽  
Reactor Pressure ◽  
Selection Of

The advancement of proper methodology to determination of VVER reactor pressure vessel (RPV) materials transition behaviour has been followed. The project included selection of proper specimen geometries (standard 10x10x55 mm, sub-size 3x4x27 mm and 5x5x27.5 mm Charpy-V samples) and impact as well as dynamic fracture toughness testing. All the fundamental criterions have been applied to consider an applicability of small-sized specimens to Cr-Mo-V and Cr-Ni-Mo-V steel radiation embrittlement studies. Selected specimen geometries have been found to be fully valid for impact and dynamic fracture data and capable for prediction of standard specimen behaviour from the small specimens. Within testing of irradiated materials, it has been found not fully validity of all the compared transition temperatures.

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