Randy K. Nanstad

Fracture Mechanics Data Deduced From Thermal-Shock and Related Experiments With LWR Pressure Vessel Material

Pressurized water reactors (PWRs) are susceptible to certain types of hypothetical accidents that can subject the reactor pressure vessel to severe thermal shock, that is, a rapid cooling of the inner surface of the vessel wall. The thermal-shock loading, coupled with the radiation-induced reduction in the material fracture toughness, introduces the possibility of propagation of preexistent flaws and what at one time were regarded as somewhat unique fracture-oriented conditions. Several postulated reactor accidents have been analyzed to discover flaw behavior trends; seven intermediate-scale thermal-shock experiments with steel cylinders have been conducted; and corresponding materials characterization studies have been performed. Flaw behavior trends and related fracture-mechanics data deduced from these studies are discussed.

Accelerated neutron embrittlement of ferritic steels at low fluence: Flux and spectrum effects

Abstract The results of Charpy V-notch surveillance testing of ferritic steels from the High Flux Isotope Reactor pressure vessel revealed significant radiation-induced embrittlement in A212 grade B, A350 grade LF3, and A105 grade II steels. The steels were irradiated at about 50°C for about 17.5 effective full-power years at a neutron flux (E > 1MeV) of 1012 to 1013 n m−2 s−1 to fluences of 1021 to 1022 n m−2. These fluences are only about one-tenth those required to cause the same embrittlement in the higher flux (~1017 n m−2 s−1) environments of test reactors. The findings suggest that the degree of embrittlement per unit fast fluence is increased at low neutron flux. Changes in neutron energy spectrum may be involved, too. Potential mechanisms for effects of neutron flux and neutron spectrum on embrittlement are discussed.

Irradiation effects on thermal conductivity of a light-water reactor pressure vessel steel

A Mn-1/2 Mo-1/2 Ni steel (ASTM A533 grade B class 1) similar to the material used for construction of early pressurized water reactors was examined for irradiation-induced changes in thermal conductivity. The mechanical properties of this steel have been shown to be sensitive to irradiation. Testing of a standard sample in the range 30 to 90°C showed that the thermal conductivity measurements had an absolute uncertainty of about ±3%. Results on the control and irradiated samples showed a small (1.7%) increase in thermal conductivity after irradiation at about 290°C to fluences up to 2.4 × 1023n/m2 (E > 1 MeV). Thus, for the neutron irradiation conditions investigated, the thermal conductivity of this pressure vessel steel is not degraded.

Effects of irradiation on the fracture properties of stainless steel weld overlay cladding

Abstract Stainless steel weld overlay cladding was fabricated using the submerged arc, single-wire, oscillating-electrode, and the three-wire, series-arc methods. Three layers of cladding were applied to a pressure vessel plate to provide adequate thickness for fabrication of test specimens, and irradiations were conducted at temperatures and to fluences relevant to power reactor operation. Post-irradiation test results of all cladding specimens show that, in the test temperature range from – 125 to 288°C, the yield strength increased by 40 to 5%, ductility increased insignificantly, and there was almost no change in ultimate tensile strength. All cladding exhibited ductile-to-brittle transition behavior during Charpy impact testing due to the dominance of delta-ferrite failures at low temperatures. On the upper shelf, energy was reduced up to 50% due to irradiation exposure. In addition, radiation damage resulted in 13 to 100 C shifts of the Charpy impact transition temperature at the 41 J level. Furthermore, irradiation exposure of 12.5 mm-thick compact specimens (0.5TCS), from the three-wire cladding to an average fluence of 2.41 × 10 19 neutrons/cm 2 (> 1 MeV ), resulted in decreases in the initiation ductile fracture toughness, J Ic , and the tearing modulus in the test temperature range from – 125 to 288°C. This is in agreement with the reduction in both the CVN upper-shelf energy and the CVN lateral expansion.

Fracture behavior of a neutron-irradiated stainless steel submerged arc weld cladding overlay☆

Abstract Stainless steel weld overlay cladding was irradiated at temperatures and fluences relevant to power reactor operation. The cladding was applied to a pressure vessel steel plate by the submerged arc, single-wire, oscillating-electrode method. Three layers of cladding were applied. The first layer was type 309, and the upper two layers were type 308 stainless steel. The type 309 was diluted considerably by excessive melting of the base plate. Charpy V-notch and tensile specimens were irradiated at 288°C to a fluence of 2 × 10 23 neutrons/m 2 (> 1 MeV). When irradiated, both types 308 and 309 cladding increased 5 to 40% in yield strength and slightly increased in ductility in the temperature range from 25 to 288°C. All cladding exhibited ductile-to-brittle transition behavior during impact testing caused by temperature dependent failure of the δ-ferrite phase. The type 308 cladding, microstructurally typical of that in reactor pressure vessels, showed very little degradation in either upper-shelf energy or transition temperature due to irradiation. Conversely, the impact properties of the specimens containing the highly diluted type 309 cladding, microstructurally similar to that produced during some off-normal welding conditions in existing reactors, experienced significant increases in transition temperature and drops of up to 50% in upper-shelf energy.

Generation IV Reactors Integrated Materials Technology Program Plan: Focus on Very High Temperature Reactor Materials

Since 2002, the Department of Energys (DOEs) Generation IV Nuclear Energy Systems (Gen IV) Program has addressed the research and development (R&D) necessary to support next-generation nuclear energy systems. The six most promising systems identified for next-generation nuclear energy are described within this roadmap. Two employ a thermal neutron spectrum with coolants and temperatures that enable hydrogen or electricity production with high efficiency (the Supercritical Water Reactor-SCWR and the Very High Temperature Reactor-VHTR). Three employ a fast neutron spectrum to enable more effective management of actinides through recycling of most components in the discharged fuel (the Gas-cooled Fast Reactor-GFR, the Lead-cooled Fast Reactor-LFR, and the Sodium-cooled Fast Reactor-SFR). The Molten Salt Reactor (MSR) employs a circulating liquid fuel mixture that offers considerable flexibility for recycling actinides and may provide an alternative to accelerator-driven systems. At the inception of DOEs Gen IV program, it was decided to significantly pursue five of the six concepts identified in the Gen IV roadmap to determine which of them was most appropriate to meet the needs of future U.S. nuclear power generation. In particular, evaluation of the highly efficient thermal SCWR and VHTR reactors was initiated primarily for energy production, and evaluation of the three fast reactor concepts, SFR, LFR, and GFR, was begun to assess viability for both energy production and their potential contribution to closing the fuel cycle. Within the Gen IV Program itself, only the VHTR class of reactors was selected for continued development. Hence, this document will address the multiple activities under the Gen IV program that contribute to the development of the VHTR. A few major technologies have been recognized by DOE as necessary to enable the deployment of the next generation of advanced nuclear reactors, including the development and qualification of the structural materials needed to ensure their safe and reliable operation. The focus of this document will be the overall range of DOEs structural materials research activities being conducted to support VHTR development. By far, the largest portion of materials R&D supporting VHTR development is that being performed directly as part of the Next-Generation Nuclear Plant (NGNP) Project. Supplementary VHTR materials R&D being performed in the DOE program, including university and international research programs and that being performed under direct contracts with the American Society for Mechanical Engineers (ASME) Boiler and Pressure Vessel Code, will also be described. Specific areas of high-priority materials research that will be needed to deploy the NGNP and provide a basis for subsequent VHTRs are described, including the following: (1) Graphite: (a) Extensive unirradiated materials characterization and assessment of irradiation effects on properties must be performed to qualify new grades of graphite for nuclear service, including thermo-physical and mechanical properties and their changes, statistical variations from billot-to-billot and lot-to-lot, creep, and especially, irradiation creep. (b) Predictive models, as well as codification of the requirements and design methods for graphite core supports, must be developed to provide a basis for licensing. (2) Ceramics: Both fibrous and load-bearing ceramics must be qualified for environmental and radiation service as insulating materials. (3) Ceramic Composites: Carbon-carbon and SiC-SiC composites must be qualified for specialized usage in selected high-temperature components, such as core stabilizers, control rods, and insulating covers and ducting. This will require development of component-specific designs and fabrication processes, materials characterization, assessment of environmental and irradiation effects, and establishment of codes and standards for materials testing and design requirements. (4) Pressure Vessel Steels: (a) Qualification of short-term, high-temperature properties of light water reactor steels for anticipated VHTR off-normal conditions must be determined, as well as the effects of aging on tensile, creep, and toughness properties, and on thermal emissivity. (b) Large-scale fabrication process for higher temperature alloys, such as 9Cr-1MoV, including ensuring thick-section and weldment integrity must be developed, as well as improved definitions of creep-fatigue and negligible creep behavior. (5) High-Temperature Alloys: (a) Qualification and codification of materials for the intermediate heat exchanger, such as Alloys 617 or 230, for long-term very high-temperature creep, creep-fatigue, and environmental aging degradation must be done, especially in thin sections for compact designs, for both base metal and weldments. (b) Constitutive models and an improved methodology for high-temperature design must be developed.

Atom probe tomography of radiation-sensitive KS-01 weld

An atom probe tomography characterization has been performed on a neutron-irradiated (fluence = 0.8 × 1019 n cm−2 (E > 1 MeV)) high copper (0.37%), high manganese (1.64%), high nickel (1.23%) and high chromium (0.47%) KS-01 test weld. This weld exhibited a high sensitivity to neutron irradiation. Atom probe tomography revealed that there was an unusually high supersaturation of copper in the matrix after the stress relief treatment, which resulted in a high number density (3 × 1024 m−3) of Cu–Mn–Ni-enriched precipitates after neutron irradiation. Their average size and composition were estimated to be ⟨lg⟩ = 2.6 ± 0.5 nm and Fe-17.0 ± 9.7 at% Cu, 31.9 ± 13.8% Ni, 31.7 ± 11.8% Mn. Phosphorus clusters and a Cr-, Mn-, Ni-, Cu-, C-, N-, Si- and Mo-enriched atmosphere, possibly associated with a dislocation, were also observed in the neutron irradiated material. Nickel, manganese, silicon, phosphorus and carbon segregation to a grain boundary were observed in the unirradiated condition. The microstructural and mechanical response to irradiation was consistent with other lower solute level steels.

Statistical analyses of fracture toughness results for two irradiated high-copper welds

The objectives of the Heavy-Section Steel Irradiation Program Fifth Irradiation Series were to determine the effects of neutron irradiation on the transition temperature shift and the shape of the K{sub Ic} curve described in Sect. 6 of the ASME Boiler and Pressure Vessel Code. Two submerged-arc welds with copper contents of 0.23 and 0.31% were commercially fabricated in 215-mm-thick plates. Charpy V-notch (CVN) impact, tensile, drop-weight, and compact specimens up to 203.2 mm thick (1T, 2T, 4T, 6T, and 8T C(T)) were tested to provide a large data base for unirradiated material. Similar specimens with compacts up to 4T were irradiated at about 288{degrees}C to a mean fluence of about 1.5 {times} 10{sup 19} neutrons/cm{sup 2} (>1 MeV) in the Oak Ridge Research Reactor. Both linear-elastic and elastic-plastic fracture mechanics methods were used to analyze all cleavage fracture results and local cleavage instabilities (pop-ins). Evaluation of the results showed that the cleavage fracture toughness values determined at initial pop-ins fall within the same scatter band as the values from failed specimens; thus, they were included in the data base for analysis (all data are designated K{sub Jc}).

Wide-plate crack-arrest tests utilizing a prototypical pressure vessel steel

Wide-plate crack-arrest tests are being performed at the National Bureau of Standards (Gaithersburg, MD) under the Heavy-Section Steel Technology (HSST) Program and are designed to provide fracture-toughness measurements at temperatures approaching or above the onset of the upper-shelf regime, in a rising toughness region and with increasing driving force. The test specimens are 1 × 1 × 0·1 m and possess a single-edge notch (crack) that initiates in cleavage propagation at low temperature and arrests in a region of increased fracture toughness. The toughness is achieved through a linear transverse temperature profile across the plate. Results obtained using a prototypical reactor pressure vessel steel (A533 grade B class 1 material) exhibit a significant increase in toughness at temperatures near and above the onset of Charpy upper shelf. Additionally, cleavage crack propagation and arrest at temperatures above the onset of Charpy upper shelf have been demonstrated.

Applicability of the Fracture Toughness Master Curve to Irradiated Highly Embrittled Steel and Intergranular Fracture

The Heavy-Section Steel Irradiation Program at Oak Ridge National Laboratory has evaluated a submerged-arc (SA) weld irradiated to a high level of embrittlement and a temper embrittled base metal that exhibits significant intergranular fracture relative to representation by the Master Curve. The temper embrittled steel revealed that the intergranular mechanism significantly extended the transition temperature range up to 150°C above To. For the irradiated highly embrittled SA weld study, a total of 21 1T compact specimens were tested at five different temperatures and showed the Master Curve to be nonconservative relative to the results, although that observation is uncertain due to evidence of intergranular fracture.