Use of Small Specimens for Evaluation of Irradiated Materials

2013 ◽  
Author(s):  
Mikhail A. Sokolov ◽  
Randy K. Nanstad
Keyword(s):  
Fracture Toughness ◽  
Small Volume ◽  
Radioactive Material ◽  
Surveillance Programs ◽  
Series Of Experiments ◽  
Point Bend ◽  
Impact Data ◽  
Surveillance Specimens ◽  
Reactor Pressure ◽  
Fracture Toughness Specimen

Small specimens are playing the key role in evaluating properties of irradiated materials. The use of small specimens provides several advantages. Typically, only small volume of material can be irradiated in a reactor at desirable conditions in terms of temperature, neutron flux, and neutron dose. Small volume of irradiated material may also allow for easier handling of specimens. Smaller specimens reduce the amount of radioactive material, minimizing personnel exposures and waste disposal. However, use of small specimens imposes variety of challenges as well. These challenges are associated with proper accounting for size effects and transferability of small specimen data to the real structures of interest. The PCVN specimen as well as any fracture toughness specimen that can be made out of the broken halves of standard Charpy specimens may have exceptional utility for evaluation of RPVs. The Charpy V-notch specimen is the most commonly used specimen geometry in surveillance programs. Precracking and testing of Charpy surveillance specimens would allow one to determine and monitor directly actual fracture toughness instead of requiring indirect predictions using correlations established with impact data. However, there is a growing number of indications that there might be a bias in the reference fracture toughness transition temperature, To values derived from PCVN and compact specimens. The present paper summarizes data from the series of experiments that use subsize specimens for evaluation of the transition fracture toughness of reactor pressure vessel (RPV) steels. Two types of compact specimens and three types of three-point bend specimens from five RPV materials were used in these subsize experiments. The current results showed that To determined from PCVN specimens with width (W) to thickness (B) ratio W/B = 1, on average, are lower than To determined from compact specimens with W/B = 2. At the same time, three-point bend specimens with W/B = 2 exhibited To values that were very similar to To values derived from compact specimens. Constraint corrections developed by Dodds et al. are applied to assess the bias.

On Bias in To Values Derived With Compact and PCVN Specimens

2004 ◽  
Author(s):  
Mikhail A. Sokolov ◽  
Randy K. Nanstad
Keyword(s):  
Fracture Toughness ◽  
National Laboratory ◽  
Oak Ridge ◽  
Surveillance Programs ◽  
Series Of Experiments ◽  
Point Bend ◽  
Impact Data ◽  
Surveillance Specimens ◽  
Reactor Pressure ◽  
Fracture Toughness Specimen

The Heavy-Section Steel Irradiation (HSSI) Program at Oak Ridge National Laboratory (ORNL) includes a task to investigate the bias in the reference fracture toughness transition temperature values, To, derived with the pre-cracked Charpy (PCVN) and compact specimens. The PCVN specimen, as well as any other fracture toughness specimen that can be made out of the broken Charpy specimens, may have exceptional utility for the evaluation of RPV steels. The Charpy V-notch specimen is the most commonly used specimen geometry in surveillance programs. Precracking and testing of Charpy surveillance specimens would allow one to determine and monitor directly actual fracture toughness instead of requiring indirect predictions using correlations established with impact data. However, there are a growing number of indications that there might be a bias in To values derived from PCVN and compact specimens. The present paper summarizes data from the series of experiments that use subsize specimens for evaluation of the transition fracture toughness of reactor pressure vessel (RPV) steels conducted within the HSSI Program. Two types of compact specimens and three types of three-point bend specimens from five RPV materials were used in these subsize experiments. The current results showed that To determined from PCVN specimens with width (W) to thickness (B) ratio W/B=1, on average, are lower than To determined from compact specimens with W/B=2. At the same time, three-point bend specimens with W/B=2 exhibited To values that were very similar to To values derived from compact specimens.

Use of Mini-CT Specimens for Fracture Toughness Characterization of Low Upper-Shelf Linde 80 Weld

2017 ◽  
Author(s):  
Mikhail A. Sokolov
Keyword(s):  
Fracture Toughness ◽  
Pressure Vessels ◽  
Charpy Specimen ◽  
Good Correspondence ◽  
Asme Code ◽  
Surveillance Programs ◽  
Impact Data ◽  
Reactor Pressure ◽  
Fracture Toughness Specimen

Any fracture toughness specimen that can be made out of the broken halves of standard Charpy specimens may have exceptional utility for evaluation of reactor pressure vessels since it would allow one to determine and monitor directly actual fracture toughness instead of requiring indirect predictions using correlations established with impact data. The Charpy V-notch specimen is the most commonly used specimen geometry in surveillance programs and most likely to be used in advanced reactors as per ASME code. The advantage of the Mini-CT specimen technique is that multiple specimens can be machined from one half of a broken Charpy specimen, used in a standard surveillance capsule of a reactor pressure vessel. Up to now, most of the work on validation of this type of the specimens has been performed on base metal. In this study, Mini-CT specimens were used to perform fracture toughness characterization of low upper-shelf Linde 80 weld, designated WF-70. This weld was utilized in the Midland beltline weld and has been previously well characterized at ORNL with various types and sizes of fracture toughness specimens. The Mini-CT specimens were machined from broken previously tested Charpy V-notch specimens. Despite very small size and relatively small number of Mini-CT specimen tested, the transition fracture toughness temperature, To, derived from these Mini-CT specimens is in very good correspondence with To reported from analysis of a large number of larger fracture toughness specimens.

Fracture Toughness Characterization of Low Upper-Shelf Linde 80 Weld Using Mini-C(T) Specimens

2018 ◽  
Author(s):  
Michael R. Ickes ◽  
J. Brian Hall ◽  
Robert G. Carter
Keyword(s):  
Fracture Toughness ◽  
Direct Measurement ◽  
Pressure Vessel ◽  
Pressure Vessels ◽  
Reactor Pressure Vessel ◽  
Weld Metals ◽  
Surveillance Programs ◽  
Test Reactor ◽  
Reactor Pressure

The Charpy V-notch specimen is the most commonly used specimen geometry in reactor pressure vessel irradiation surveillance programs and there is an extensive stored inventory of irradiated broken Charpy specimens. The advantage of the mini-C(T) (4mm thick C(T)) specimen technique is that multiple specimens can be machined from each half of broken irradiated Charpy specimens. Fracture toughness specimens that can be machined from broken halves of standard Charpy specimens enable the direct measurement of fracture toughness which can be used for engineering evaluation of reactor pressure vessels. Work to validate the mini-C(T) specimens has been performed mostly on unirradiated reactor pressure vessel base and weld metals . In this study, mini-C(T) specimens were tested providing fracture toughness characterization of an irradiated low upper-shelf Linde 80 weld (WF-70). This weld was utilized in the Midland beltline and has been previously well characterized at ORNL with various types and sizes of fracture toughness specimens. The mini-C(T) specimens were machined from broken previously tested Charpy V-notch size specimens which were irradiated in a material test reactor. The effect of different methods of measuring the displacement on the results is assessed. The ASTM E1921 results are compared to previous test data produced from larger fracture toughness specimens. In addition, the sensitivity of T0 to the ASTM E1921 censoring value is discussed.

Fracture Toughness in Postulated Crack Area of PTS Evaluation in Highly-Neutron Irradiated RPV Steel

2021 ◽  
Author(s):  
Yoosung Ha ◽  
Masaki Shimodaira ◽  
Hisashi Takamizawa ◽  
Tohru Tobita ◽  
Jinya Katsuyama ◽  
...  
Keyword(s):  
Fracture Toughness ◽  
Vickers Hardness ◽  
Heat Affected Zone ◽  
Pressure Vessel ◽  
Structural Integrity ◽  
Rpv Steel ◽  
Pressurized Thermal Shock ◽  
Inner Surface ◽  
Surveillance Specimens ◽  
Reactor Pressure

Abstract The Japanese Electric Association Code 4206-2016 requires that the semi-elliptical crack sized 10 mm in depth × 60 mm in length shall be postulated near the inner surface of a reactor pressure vessel (RPV) in pressurized thermal shock events. The fracture toughness distribution was investigated in the postulated crack area under the PTS events of unirradiated and highly-neutron irradiated RPV steels. Vickers hardness in heat-affected zone (HAZ) due to stainless overlay cladding and 10 mm from the cladding were higher than that of a quarter thickness position, where the surveillance specimens are machined, for both unirradiated (E1) and irradiated (up to 1 × 1020 n/cm2, WIM) materials. Fracture toughness of HAZ and 10 mm from the cladding was higher for the above highly-neutron irradiated material. The same result was obtained in the unirradiated material. Therefore, it was confirmed that fracture toughness obtained from surveillance specimens can provide conservative assessment of structural integrity of RPV.

Embrittlement Trend Curves for Notch and Fracture Toughness of 15Kh2MFAA Type Steel

2017 ◽  
Author(s):  
Milan Brumovsky ◽  
Radim Kopřiva ◽  
Miloš Kytka
Keyword(s):  
Fracture Toughness ◽  
Temperature Dependence ◽  
Pressure Vessel ◽  
Master Curve ◽  
Charpy Impact ◽  
Type Steel ◽  
Impact Tests ◽  
Surveillance Programs ◽  
Vessel Integrity ◽  
Reactor Pressure

Reactor pressure vessel integrity and lifetime evaluation is based on the use of fracture mechanics apparatus but most of the material vessel material data and their degradation during operation are based on results from Charpy V-notch impact tests. Then, empirical correlations between transition shift of temperature dependence of notch toughness and fracture toughness are applied. Elaboration of „Master Curve“ approach for fracture toughness experimental data analysis allows to use fracture toughness data directly to the reactor pressure vessel integrity evaluation. Wider use of this approach is limited by the lack of appropriate database from surveillance specimen test data, as mostly only Charpy impact specimen are included into the Surveillance specimen programs. Fortunately, all WWER Surveillance programs contain also fracture toughness specimens, either pre-cracked Charpy size or CT-0.5. Thus, database of fracture toughness data from Surveillance programs of WWER-440/V-213C type reactor pressure vessels, operated in the Czech Republic, Slovakia and Hungary and manufactured only by one manufacturer - SKODA JS, was collated and analyzed. These vessels were manufactured from 15Kh2MFAA type steel and appropriate weld metal, both of Cr-Mo-V type with low content of detrimental impurities — P and Cu. Analysis of the data in fluence interval up to 6×1024 m−2 (with neutron energies En larger than 0.5 MeV) show that transition temperature shifts in fracture toughness temperature dependence are higher than for Charpy impact tests. Several formulae have been applied for fitting these shift dependencies with chemical composition of materials and finally new Embrittlement Trend Curves for Charpy shifts have been corrected. Additionally, new Embrittlement Trend Curves for fracture toughness shifts based on “Master Curve” approach have been also proposed. Both trends are using simple power law on fluence with exponents around 0.6 and depend on phosphorus and copper contents even though effect of other elements has been also checked.

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.

scholarly journals Uncertainty determination of fast neutron fluence onto the WWER pressure vessel metal surveillance specimens

2021 ◽  
Vol 22 (1) ◽  
pp. 42-47
Author(s):  
O.M. Pugach ◽  
◽  
S.M. Pugach ◽  
V.L. Diemokhin ◽  
V.N. Bukanov ◽  
...  
Keyword(s):  
Fast Neutron ◽  
Pressure Vessel ◽  
Neutron Fluence ◽  
Neutron Transport ◽  
Monte Carlo Code ◽  
Neutron Fluences ◽  
Surveillance Programs ◽  
Surveillance Specimens ◽  
Reactor Pressure

The standard surveillance programs of WWER reactors do not allow to measure the surveillance specimens irradiation conditions with the required accuracy. Therefore, the special methodology for the determination of the surveillance specimens irradiation conditions of the reactor pressure vessel metal has been developed by the specialists of the INR of NASU and is successfully applied. The developed methodology bases on the use of the Monte-Carlo code for neutron transport calculations to the surveillance specimens locations. The methodology improvement is described. The fundamentals of the calculation-experimental determination of the fast neutron fluences onto surveillance specimens and their uncertainties are presented.

Comparative analysis and verification of engineering methods of shallow cracks effect for fracture toughness prediction for reactor pressure vessels

2020 ◽  
pp. 140-165
Author(s):  
V. I. Kostylev ◽  
B. Z. Margolin
Keyword(s):  
Experimental Data ◽  
Fracture Toughness ◽  
Comparative Analysis ◽  
Nuclear Reactor ◽  
Probabilistic Approach ◽  
Pressure Vessels ◽  
Local Methods ◽  
European Method ◽  
Shallow Crack ◽  
Reactor Pressure

The main features of shallow cracks fracture are considered, and a brief analysis of methods allowing to predict the temperature dependence of the fracture toughness KJC (T) for specimens with shallow cracks is given. These methods include DA-method, (JQ)-method, (J-T)-method, “local methods” with its multiparameter probabilistic approach, GP method uses power approach, and also two engineering methods – RMSC (Russian Method for Shallow Crack) and EMSC (European Method for Shallow Crack). On the basis of 13 sets of experimental data for national and foreign steels, a detailed verification and comparative analysis of these two engineering methods were carried out on the materials of the VVER and PWR nuclear reactor vessels considering the effect of shallow cracks.

Sign in / Sign up

Export Citation Format

Share Document