As part of the Department of Energy's Light Water Reactor Sustainability (DOE/LWRS) program, we are developing time-independent material models based on tensile tests and time-dependent material models based on cyclic tests for different reactor materials, such as 316 stainless steel (SS), 508 lowalloy steel (LAS) base metal, 316 SS-316 SS similar metal welds, and 316 SS-508 LAS dissimilar metal welds. Also, materials models are being developed under different environmental conditions, such as in air (at room temperature and 300 o C) and PWR primary loop water (at 300 o C). In our previous work, we presented time-dependent material models for 316 SS base metals [15][16][17]. In this report, we present tensile and fatigue test results and associated material models under different test and environmental conditions for 508 LAS base metal and 316 SS-316 SS similar metal welds.

The effect of thermal aging on tensile properties of cast stainless steels during service in light water reactors has been evaluated. Tensile data for several experimental and commercial heats of cast stainless steels are presented. Thermal aging increases the tensile strength of these steels. The high-C Mo-bearing CF-8M steels are more susceptible to thermal aging than the Mo-free CF-3 or CF-8 steels. A procedure and correlations are presented for predicting the change in tensile flow and yield stresses and engineering stress-vs.-strain curve of cast stainless steel as a function of time and temperature of service. The tensile properties of aged cast stainless steel are estimated from known material information, i.e., chemical composition and the initial tensile strength of the steel. The correlations described in this report may be used for assessing thermal embrittlement of cast stainless steel components. Executive SummaryCast duplex stainless steels used in light water reactor (LWR) systems for primary pressure-boundary components are susceptible to thermal embrittlement at reactor operating temperatures. Thermal aging of cast stainless steels at these temperatures causes an increase in hardness and tensile strength and a decrease in ductility, impact strength, and fracture toughness of the material and the Charpy transition curve shifts to higher temperatures. Investigations at Argonne National Laboratory have shown that thermal embrittlement of cast stainless steel components may occur within the reactor design lifetime of 40 yr. Various grades and heats of cast stainless steel exhibit varying degrees of thermal embrittlement. In general, the low-C CF-3 steels are the most resistant t o thermal embrittlement, and the Mobearing, high-C CF-8M steels are the least resistant. An assessment of mechanical-property degradation due to thermal embrittlement is therefore required to evaluate the performance of cast stainless steel components during prolonged exposure to service temperatures, because rupture of the primary pressure boundary could lead to a loss-of-coolant accident and possible exposure of the public to radiation.This report presents tensile-property data on several heats of cast stainless steels aged up to 58,000 h at temperatures between 290 and 450°C (554 and 752°F). The tensile data are analyzed to establish the effects of thermal aging on tensile strength and engineering stress-strain behavior (represented by the Ramberg-Osgood equation) of cast stainless steels. A procedure and correlations are presented for predicting the change in tensile flow and yield stress, and in the engineering stress-vs.-strain curve of cast stainless steel components due to thermal aging during service in LWRs. The tensile properties of aged cast stainless steel are estimated from information that is readily available from certified material test records for the component, i.e., chemical composition and the initial tensile strength of the unaged material. The correlations described in this report may be used ...

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.

Reactor vessel internal components made of nickel-base alloys are susceptible to environmentally assisted cracking (EAC). A better understanding of the causes and mechanisms of this cracking may permit less conservative estimates of damage accumulation and requirements on inspection intervals. The objective of this work is to evaluate and compare the resistance of Alloys 600 and 690 and their welds, such as Alloys 82, 182, 52, and 152, to EAC in simulated light water reactor environments. The existing crack growth rate (CGR) data for these alloys under cyclic and constant loads have been evaluated to establish the effects of alloy chemistry, cold work, and water chemistry. The experimental fatigue CGRs are compared with CGRs that would be expected in air under the same mechanical loading conditions to obtain a qualitative understanding of the degree and range of conditions for significant environmental enhancement in growth rates. The existing stress corrosion cracking (SCC) data on Alloys 600 and 690 and Alloy 82, 182, and 52 welds have been compiled and analyzed to determine the influence of key parameters on growth rates in simulated PWR and BWR environments. The SCC data for these alloys have been evaluated with correlations developed by Scott and by Ford and Andresen.

We are undertaking a systematic study at Argonne National Laboratory to evaluate the uniaxial creep behavior of V-Cr-Ti alloys in a vacuum environment as a function of temperature in the range of 650-800°C and at applied stress levels of 75-380 MPa. Creep strain in the specimens is measured by a linear-variable-differential transducer, which is attached between the fixed and movable pull rods of the creep assembly. Strain is measured at sufficiently frequent intervals during testing to define the creep strain/time curve. A linear least-squares analysis function is used to ensure consistent extraction of minimum creep rate, onset of tertiary creep, and creep strain at the onset of tertiary creep. Creep test data, obtained at 650, 700, 725, and 800°C, showed power-law creep behavior. Extensive analysis of the tested specimens is conducted to establish hardness profiles, oxygen content, and microstructural characteristics. The data are also quantified by the Larson-Miller approach, and correlations are developed to relate time to rupture, onset of tertiary creep, times for 1 and 2% strain, exposure temperature, and applied stress.