Achievements and further plans for the OECD Halden reactor project materials programme

A review of irradiation effects on LWR core internal materials – IASCC susceptibility and crack growth rates of austenitic stainless steels

“…Tensile properties data have been obtained on various grades of austenitic SSs, including weld heat affected zone (HAZ) material, weld metal, and CF-8 cast austenitic SSs, irradiated at temperatures of 300-400°C (572-752°F) in fast reactors and LWRs [38][39][40][41][42]. The 0.2% yield strength and total elongation of solutionannealed and CW SSs at elevated temperatures are plotted in Figs.…”

“…For solution-annealed Type 304 SS, they predict a value of 800 MPa for both saturated yield strength and ultimate strength, and for CW Type 316 SS, the predicted saturated yield and ultimate strengths are 970 and 980 MPa, respectively. However, most of the data for yield and ultimate strength of LWR-irradiated materials are above the MRP curve [38][39][40][41][42]. Similarly, most of the data for uniform and total elongation of LWR-irradiated materials are below the MRP curve.…”

“…For some irradiated austenitic SS materials, small increases in the stress intensity factor K due to either load perturbations or the reloading portion of the partial unload/reload cycle during a SCC growth rate test result in a rapid or step-like crack extension [40]. The growth rates during such periods are as high as 10 À7 m/s.…”

In-situ Creep Testing Capability Development for Advanced Test Reactor

“…IFE/HRP in-core sensors for fuel and material performance can detect fuel temperature, fission gas release, fuel swelling/densification, cladding creep, corrosion/crud buildup, and crack-growth rates. [4][5][6][7][8] Many of HRP's in-core instrumentation relies upon LVDTs. The LVDT is a versatile instrument used to transform a mechanical movement into an electrical signal.…”

In-Core SCC Growth Behavior of Type 304 Stainless Steel in BWR Simulated High-Temperature Water at JMTR

“…Under the IASCC test program of the OECD Halden Reactor Project, 3) the IASCC susceptibility of stainless steels (and nickel base alloys), the effects of radiation and water chemistry on IASCC, the quantification of crack growth rate as a function of stress intensity factor and the benefits of mitigation measures, such as hydrogen water chemistry (HWC), have been addressed. [4][5][6] In this program, successful qualification of instrumentation and loading techniques required for in-core monitoring of cracking behavior during exposure to representative LWR (typically boiling water reactor (BWR)) environments have been performed, and the use of ''1/4T'' compact tension (CT) specimens prepared from irradiated stainless steels in crack growth studies has been demonstrated. 4,5) Furthermore, the water loops can be operated under different thermal-hydraulic and water chemistry conditions, covering a range of BWR and pressurized water reactor (PWR) requirements at the Halden reactor.…”