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Effects of HFIR neutron irradiation on fracture toughness properties of standard and Ni-doped F82H
F82H is the Japanese reference reduced-activation ferritic-martensitic (RAFM) steel for fusion blanket applications. The harsh environment of a fusion reactor, such as neutron irradiation and He/H damage, can result in significant degradation of F82H fracture toughness. Therefore, understanding the fracture toughness behavior of F82H in the fusion environment is critical to ensure the long-term safe operation of the fusion reactor. Here, we summarize seven irradiation campaigns of the High Flux Isotope Reactor (HFIR) at Oak Ridge National Laboratory (ORNL) covering five variants of F82H steels, including F82H IEA, F82H Mod3, F82H doped with 1.4% natural Ni, F82H doped with 1.4% 58 Ni, and F82H doped with 1.4% 60 Ni. The irradiation temperatures covered the range from 220 °C to 530 °C and the neutron irradiation dose spanned 4 dpa to 70 dpa. The effects of neutron irradiation temperature, dose, materials composition, Ni doping, and He production on F82H fracture toughness are discussed. Our results showed that irradiation embrittlement monotonically decreased with increasing irradiation temperature until 400 °C for F82H IEA and F82H Mod3. F82H Mod3 showed better fracture toughness than F82H IEA both before and after neutron irradiation. We determined that 1.4% Ni alloying can be applied to F82H for simulating He effect in a fission reactor without jeopardizing the fracture toughness of the material. However, more studies are needed to understand the effect of high dose (>20 dpa) and He production on F82H fracture toughness.
Fracture toughness evaluations for spent nuclear fuel dry storage canister welds and spent nuclear fuel clad-pellet structures
Integrity of spent nuclear fuel (SNF) interim storage canisters is very important to the safety of the back-end nuclear fuel cycle. Stress corrosion cracking (SCC) potential of interim storage canister has been considered as a high priority. Because no post-weld heat treatment was required for forming these canisters, the high tensile residual stress existed within these canister welds. This can change the fracture resistance capacity significantly as well as increase SCC potential. Due to relative thin shell thickness of a canister weldment, the spiral notch torsion test (SNTT) method was used to estimate the canister weldment fracture toughness. SNTT was developed to measure the intrinsic fracture toughness of structural materials using small specimens. The SNTT method has been applied to a wide variety of structural materials, such as low-alloy steels, stainless steel, aluminum alloy, ceramics, concrete, and composites. The SNTT system operates by applying pure torsion to cylindrical specimens with a notch line that spirals around the specimen at a 45° pitch. In order to carry out pure torsion load mode, biaxial tension/torsion tester was developed accordingly to perform SNTT protocol. Moreover, applying fracture mechanics approach to SNF system reliability investigation is warranted due to the inherent flaws and hydride structures existed in a SNF system after nuclear reactor operation. However, none of the existing fracture toughness data deal with fuel cladding specific geometry or spent fuel material conditions, such as cladding structure with the pellet-inserts and the associated pellet clad mechanical interactions induced mixed-mode damage mechanisms. Thus, the development of an intrinsic fracture mechanics approach that is suitable for SNF materials is needed. Furthermore, due to thin wall and small dimension of clad tubing structure, the SNTT method was used to estimate the clad tubing structure fracture toughness. Fracture testing were performed on the received stainless steel canister weldment, most SNTT weld samples fracture initiation sites are at heat-affected zone (HAZ) regions. The estimated fracture toughness J Q ’ for the baseline SS304 steel is at 283 kJ/m². The estimated JQ’ for the SS304/308 weld and baseline metals are 148 kJ/m 2 and 459 kJ/m 2 , respectively. Out of cell fracture testing for spent fuel structure were carried out on the surrogate rods made of Zr-4 clad and alumina inserts, the estimated fracture toughness values for baseline Zr-4 cladding with alumina-pellet inserts are: (1) For SNTT samples with a short or medium crack length, between 5.4-mm and 8-mm, the estimate J IQ upon fracture initiation for the baseline Zr-4 cladding is at 50 kJ/m 2 with 2-sigma uncertainty of 3.26 kJ/m 2 , and the associated K IQ is at 67.46 MPa$\sqrt{m}$; and (2) For SNTT samples with a long crack length, around 13-mm, the crack initialization is deviated from that of the Mode-I tensile fracture and appears to be a mixed-mode fracture of Mode I - tensile stress and Mode III - out of plane shear stress; the estimated J MQ is at 18.9 kJ/m 2 , the associated K MQ is at 41.4 MPa$\sqrt{m}$.
Fracture toughness of zirconia with a nanometer size notch fabricated using focused ion beam milling
Zirconia with 3 mol% yttria (3Y-TZP) has been used for dental crowns and bridges due to its excellent mechanical behavior. Performing fracture toughness testing on this nanograin material, however, can be a challenge. For reliable results, fracture toughness testing requires an extremely sharp notch in the test specimen that closely approximates a very sharp crack. This study was to investigate an alternative method to produce nanometer-sized notches, which are less than the average grain size of 3Y-TZP, during the preparation of single-edge V-notched beam specimens and report the resulting fracture toughness value. We present a method using focused ion beam (FIB) milling to fabricate nanometer-sized notches in 3Y-TZP. The notch tip is <100 nm wide, which is smaller than the grain size, and is consistent throughout the thickness of the specimen. The FIB-notched specimens show a much reduced average fracture toughness of 5.64 ± 1.14 MPa√m compared to 8.90 ± 0.23 MPa√m for the specimens without FIB-notches. The FIB-milling did not appear to create any monoclinic phase prior to fracture toughness testing. Fractures originated at the FIB-notches, and the notch size can be readily identified post-mortem using a microscope. A considerable amount of tetragonal-to-monoclinic phase transformation was observed throughout the fracture surfaces. FIB milling provides an alternative method to fabricate nanometer-sized notches that are smaller than the grain size of tetragonal zirconia polycrystal. The fracture toughness determined using FIB-notches was ~5.64 MPa√m, smaller than the specimens with V-notches fabricated using saw blades.
Fracture toughness evaluation for Zr-4 clad tubing structure with pellet inserts
Applying fracture mechanics approach to spent nuclear fuel (SNF) system reliability investigation is warranted due to the inherent flaws and hydride structures existed in a SNF system after nuclear reactor operation. However, none of the existing fracture toughness data deal with fuel cladding specific geometry or spent fuel material conditions, such as cladding structure with the pellet-inserts and the associated pellet clad mechanical interactions induced mixed-mode damage mechanisms. Thus, the development of an intrinsic fracture mechanics approach that is suitable for SNF materials is needed. Due to thin wall and small dimension of clad tubing structure, the spiral notch torsion test (SNTT) method of small specimen approach was used to estimate the clad tubing structure fracture toughness. The estimated fracture toughness for Zr-4 cladding with alumina-pellet inserts are presented in this report. For SNTT samples with a short or medium crack length, between 5.4-mm and 8-mm, the estimate JIQ upon fracture initiation for the baseline Zr-4 cladding is at 50 kJ/m2 with 2-sigma uncertainty of 3.26 kJ/m2, and the associated KIQ is at 67.46 MPa√m. For SNTT samples with a long crack length, around 13-mm, the crack initialization is deviated from that of the Mode-I tensile fracture and appears to be a mixed-mode fracture of Mode I – tensile stress and Mode III – out of plane shear stress; the estimated JMQ is at 18.9 kJ/m2, the associated KMQ is at 41.4 MPa√m.
Use of Mini-CT Specimens for Fracture Toughness Characterization of Irradiated Highly Embrittled Weld
In the 1990’s, the Heavy Section Steel Irradiation Program at the ORNL performed investigation of the shape of the fracture toughness master curve for reactor pressure vessel steel highly embrittled because of irradiation exposure. A radiation-sensitive reactor pressure vessel (RPV) weld with intentionally enhanced copper content, designated KS-01, has been characterized in terms of static initiation (KJc) and Charpy impact toughness in the unirradiated and irradiated conditions. The objective of this original project was to investigate the ability of highly embrittled material to maintain the shape of the unirradiated transition fracture toughness curve, as well as to examine the ability of the Charpy 41-J shift to predict the fracture toughness shift at such high level of embrittlement. Irradiation of this weld was performed at the University of Michigan Ford Reactor in the new HSSI irradiation-anneal-reirradiation (IAR) facility. Broken specimens from that project have been saved in ORNL storage. To verify applicability of Mini-CT specimens for fracture toughness characterization of RPV materials as part of the DOE Light Water Reactor Sustainability Program, Mini-CT specimens were machined from broken Charpy specimens and tested according to ASTM E1921 standard. As result of this study, the fracture toughness of this weld derived by testing Mini-CT specimens in the unirradiated and irradiated conditions is compared to previously reported fracture toughness derived by large number of conventional specimens.
Microstructure and fracture toughness characterization of three 9Cr ODS EUROFER steels with different thermo-mechanical treatments
Ferritic martensitic ODS steels are one of the candidate structural materials for future Gen-IV nuclear fission and fusion reactors. Here, the dependence of fracture toughness on microstructure was investigated by comparing three 9Cr ODS EUROFER steels manufactured through different thermo-mechanical processing routes. Quasi-static fracture toughness testing was performed with sub-sized C(T) specimens and microstructural characterization was carried out using scanning electron microscopy, electron backscatter diffraction, and transmission electron microscopy. It was found that at lower test temperatures (-100 – 22 °C), the fracture toughness was primarily controlled by crack initiation at sub-micron particles and by production of secondary cracks during fracture. At higher temperatures (above 100 °C), fracture toughness was predominantly controlled by the matrix ductility and the grain boundary strength with a relatively ductile coarse-grained alloy demonstrating higher fracture toughness compared to high-strength fine-grained alloys. These results and discussion show that variations in thermomechanical treatments can produce significant differences in microstructure and fracture toughness behavior of ferritic martensitic ODS steels.
Fracture toughness and crack propagation in LX-17 and PBX 9502 insensitive high explosives
High explosive materials, and in particular insensitive high explosive (IHE) materials, are of significant interest. TATB-based LX-17 and PBX 9502 are two IHE’s and mechanical properties are an important metric of performance. In this work we present, for the first time, a study of fracture as a function of temperature and displacement rate and starter crack length on billet pressed IHE materials using a standard three point bend technique. We demonstrated that the technique can yield consistent results over time by measuring samples from the same lots 4 years later and observed that the fracture toughness of LX-17 and PBX 9502 are similar. We also determined the temperature dependence for different starter crack lengths and found that the fracture toughness generally decreases with larger starter crack sizes and temperatures. For the LX-17, where a wide temperature range of -60 C to 50 C was investigated, the impact of temperature on fracture toughness between -20 C and 23 C was less than expected. For the PBX 9502 we observed the difference in fracture toughness values between -20 C and 23 C did depend on the particular lot, and also there was a tendency of the fracture toughness to scale with the tensile strength. We also investigated the effect of a lower displacement rate and observed that while the impact on the fracture toughness was relatively small, the samples surprisingly did not break (fail) after exceeding the peak load. We took advantage of this to perform x-ray computed tomography (CT) of the samples to image crack propagation in the intact samples. We observed that cracks were generally straight and unaffected by the microstructure in the LX-17, in contrast to the PBX 9502 where the direction of the cracks would frequently change and were likely influenced by the underlying microstructure.
Intrinsic and extrinsic effects on the fracture toughness of ductile metallic glasses
The potential application of metallic glasses (MGs) as structural materials demands accurate measurements of their toughness and the understanding of the underlying factors affecting it. Currently, it is challenging to precisely measure the toughness of MGs, especially ductile MGs. Measured toughness values are widely scattered, even for MGs with identical compositions. That is attributed to the combined effect of intrinsic and extrinsic factors including processing, sample geometries, and loading conditions. A fundamental understanding of the influences of these elements is thus of great significance. In the present study, molecular dynamics simulations are performed to investigate the influence of intrinsic and extrinsic effects on the fracture toughness of CuZr MGs. In particular, focused is placed on the effects of cooling rate, notch shape and size, strain rate, and temperature. The results indicate that the fracture toughness of a MG scales with the cooling rate used to prepare it. This is attributed to increased free volume content generated at high cooling rates, which enhances plastic deformation and amplifies the associated energy dissipation during plastic deformation events. Fracture toughness also increases with the strain rate, arguably due to strain rate-induced crack extension delay. Overall, the results demonstrate that the largest fracture toughness are achieved when MG samples are fabricated at high cooling rates and subjected to high strain rate deformation. In addition, results suggest that the fracture toughness decreases with increasing temperature, due to the significant decrease in strength. Here, the correlations revealed between these crucial intrinsic and extrinsic parameters and the calculated MG fracture toughness support the development of a framework to understand the root of the discrepancies in the measurement of the toughness of MGs and provide insights into the design of tough MGs for structural applications.
Influence of fatigue precracking and specimen size on Master Curve fracture toughness measurements of EUROFER97 and F82H steels
EUROFER97 and F82H steels are two leading reduced-activation ferritic-martensitic (RAFM) steels for fusion first wall and blanket applications. Exposure to the harsh environment of fusion reactors can result in severe degradation of fracture toughness. Thus, the post-irradiation evaluation of fracture toughness is critical for understanding the material behavior. Due to the space constraints of irradiation facilities and challenges in controlling a uniform irradiation condition for large size specimens, the development of small specimen test techniques (SSTT) is indispensable to evaluate the performance of irradiated materials. In this study, we evaluated specimen size effects on the Master Curve fracture toughness of EUROFER97 and F82H steels. A wide variety of specimens, including 0.5 T compact tension (C(T)) specimens, 0.16 T mini-compact tension (miniC(T)) specimens, and 1.65 mm miniature bend bar specimens, were tested. The testing methodology was based on the Master Curve method in the ASTM E1921 standard. No specimen size effect was observed in 0.5 T C(T) and 0.16 T miniC(T) specimens on the Master Curve reference temperature T0, while 1.65 mm miniature bend bar specimens yielded a higher T 0Q . A strong effect of fatigue precracking on T 0 for 0.5 T C(T) and 0.16 T miniC(T) specimens was observed, such that testing on specimens with skewed fatigue precrack fronts resulted in lower T 0 than for specimens with ASTM standard qualified straight fatigue precrack fronts. The results highlight the importance of experimental quality control in developing SSTT for Master Curve fracture toughness testing. Lastly, we also evaluated and provided recommendations on the minimum number of specimens needed for each specimen type for yielding reliable T 0Q values.
Fracture toughness of schist, amphibolite, and rhyolite from the Sanford Underground Research Facility (SURF), Lead, South Dakota
The Cracked Chevron Notched Brazilian Disc (CCNBD) method was selected for Mode I fracture toughness tests on Poorman schist, Yates amphibolite, and rhyolite dikes from the EGS Collab site at the SURF in Lead, South Dakota. The effects of lithology, anisotropy, and loading rate were investigated. Fracture toughness was greatest in amphibolite, with schist and rhyolite having similar toughness values ($K$ amphibolite > $K$ rhyolite ≈ $K$ schist ). The effects of anisotropy on fracture toughness were investigated in the foliated schist samples. Schist samples were prepared in three geometries (divider, arrester, and short transverse) which controlled how the fracture would propagate relative to foliations. The divider geometry was strongest and short transverse geometry was the weakest ($K$ divider > $K$ arrester > $K$ shorttransverse ). Fracture toughness was observed to decrease with decreasing loading rate. Optical and SEM microscopy revealed that for the short transverse geometry, fractures tended to propagate along grain boundaries, whereas in arrester and divider geometries fractures tended to propagate through grains. In foliated samples, the tortuosity of the fracture observed in thin section was greater in arrester and divider geometries than in short transverse geometries.
Tritium Aging Effects on Fracture Toughness of Stainless Steel Weldments
The long-term embrittlement effects of tritium and decay helium on the structural properties of stainless steels have been studied for years at Savannah River to provide required data for establishing safe operating conditions and lifetimes of pressure vessels used to contain tritium gas. In this study, the fracture toughness properties of the longest-aged tritium-precharged stainless-steel base metals and weldments tested at Savannah River were measured and compared to earlier results. The fracture toughness values were the lowest recorded here for tritium-exposed stainless steel. As-forged and as-welded specimens were thermally precharging with tritium gas at 34.5 MPa and 623 K, then aged for up to 17 years to buildin decay helium prior to testing. ASTM J-Integral fracture mechanics analyses, transmission electron microscopy (TEM), and small angle neutron scattering (SANS) examinations were conducted to characterize the effects of tritium and its radioactive decay product, helium-3. Results show that the fracture toughness values were reduced to less than 2-4% of the as-forged values for specimens with more than 1300 appm helium from tritium decay. The trend of decreasing fracture toughness values with increasing helium content was consistent with earlier observations, and the data show that Type 304L stainless steel is more resistant to tritium-induced cracking than Type 21-6-9 stainless at similar decay helium levels. The fracture toughness properties of long-aged weldments were also affected, but the reductions were not as severe over time because weldments did not retain as much tritium as did the base metals. TEM observations were used to characterize the effects of decay helium bubbles on the deformation substructures, but nanometer-sized helium bubbles were not easily resolved because of high dislocation densities within the forged microstructures. SANS results are presented that suggest the technique can provide information on decay helium bubble size, spacing, and distribution in these steels.
High temperature fracture toughness of single-layer CrAlN and CrAlSiN hard coatings
Hardness and toughness are among of the most important physical properties of hard coatings and it is important to analyze them at temperatures present during coating applications. CrAlN and CrAlSiN are hard coating materials frequently used in high temperature applications. Here, while both possess high oxidation resistance and thermal stability, they differ in their microstructure. In this work, the fracture toughness of magnetron-sputtered single-layer CrAlN and CrAlSiN hard coatings was determined up to 700 °C by in-situ scanning electron microscopy micro-cantilever deflection tests. The results show a decreasing trend of the fracture toughness with increasing temperature. While both coatings' fracture toughness trends are similar up to 400 °C, they diverge at higher temperatures. In general, the evolution of the fracture toughness with increasing temperature shows similarities to the observed changes in hardness in the same temperature range.
Grain Boundary and Lattice Fracture Toughness of UO 2 Measured Using Small-Scale Mechanics
Small-scale mechanical testing has been used to isolate the properties of individual grain boundaries and regions of the lattice in undoped UO 2 polycrystals. Notched microcantilever specimens were prepared by focused ion beam milling and tested in bending. Fracture toughness values were calculated from finite element analysis of the sample geometry at fracture load using the J-integral method. Lattice and grain boundary fracture toughness values of 1.17 ± 0.16 MPa m 1/2 and 1.08 ± 0.08 MPa m 1/2 , respectively, were obtained. The similarity between the lattice and grain boundary fracture toughness values is consistent with the preference for transgranular room-temperature fracture in UO 2 . The grain boundary and lattice fracture toughness values also agree well with the toughness of dense polycrystals, which suggests that microstructural-scale toughening mechanisms are limited in such samples.
Assessment of irradiation embrittlement effect on fatigue life of a pressurized-water reactor pressure vessel using the fracture toughness master curve approach
The reactor pressure vessel (RPV) is a critical structural component in pressurized-water reactors, and it is designed to withstand extreme conditions, such as high pressures, elevated temperatures, and prolonged radiation exposure. Ensuring RPV integrity is essential for the safe and reliable long-term operation of nuclear power plants, especially as aging mechanisms such as fatigue and irradiation embrittlement pose increased risks. Fatigue, caused by cyclic thermal and mechanical loading, can lead to crack initiation in localized high-stress regions. Simultaneously, neutron irradiation, particularly in the beltline region, progressively reduces fracture toughness, increasing susceptibility to brittle fracture. These combined effects of fatigue and irradiation embrittlement potentially impact the RPV structural integrity, necessitating fitness-for-service assessments. This study applies the fracture toughness master curve approach to evaluate the impact of irradiation embrittlement on RPV fatigue life. A coupled thermo-mechanical stress analysis identifies critical stress locations under normal service transients, pinpointing regions most vulnerable to fatigue crack initiation and growth. Stress intensity factors for postulated flaws at these locations are calculated, enabling an assessment of fatigue life under irradiated and unirradiated conditions. The results indicate that neutron irradiation embrittlement accelerates the conditions in which a critical crack can form and lead to failure, particularly at lower temperatures. The failure occurs where reduced fracture toughness limits the material’s resistance to crack growth. Axial cracks at nozzle corners are the most life limiting without irradiation, while circumferential cracks demonstrate longer fatigue lives. The findings highlight the importance of incorporating irradiation effects into fatigue life predictions to ensure the long-term structural integrity of RPVs.
Fracture toughness measurement of ceramics with fuel pellet geometry
The crack initiation of nuclear fuel pellets, driven by extreme temperature gradients, is a critical factor impacting reactor performance. Although this fact is widely acknowledged, there is limited fracture toughness data for specimens with representative geometry and fabricated using relevant processing methods. Here, to address this, a new chevron-notched short-rod specimen configuration with diameter and length both equal to 10 mm, representative of a conventional fuel pellet and tested in tension, was developed in an attempt to validly measure Mode I fracture toughness, K Ic . Test calibration was carried out by testing polycrystalline ZrO 2 , SiC, and Si 3 N 4 ceramics fabricated with conventional fuel pellet geometry and known fracture toughnesses. The results suggest that the method described in this paper produces valid K Ic measurements using conventional ceramic fuel pellets.
The comparative strength and fracture toughness properties of commercial 95W-3.5Ni1.5Fe and 95W-3.5Ni1.5Cu tungsten heavy alloys
Due to the sharp surface cracks resulting from high and cyclic thermal loads, fracture toughness is a crucial property for fusion divertor materials. Tungsten heavy alloys (WHA) have emerged as a leading candidate for plasma facing component applications. Both the WNiFe and WNiCu WHAs have been considered. However, there is no fracture toughness data in the literature on the WNiCu WHA. Therefore, maximum load elastic–plastic fracture toughness (K Jm ) of a commercial 95W-3.5Ni1.5Cu (NiCu) WHA plate is reported here for the first time. The NiCu WHA K Jm is compared to a corresponding K Jm for commercial 95W-3.5Ni1.5Fe (NiFe) WHA, along with comparisons of their respective microstructures, microhardness, and tensile properties. Although their room temperature yield strengths are similar (≈ 600 MPa), the NiFe WHA has higher ultimate stress (818 MPa vs. 642 MPa) and total elongation (8% vs. 1.2%) compared to the NiCu WHA. Most notably, the room temperature elastic–plastic K Jm for the NiFe WHA of ≈ 80 ± 8 MPa$\sqrt{m}$, is nearly twice that of the NiCu WHA at ≈ 47 ± 4 MPa$\sqrt{m}$. Further, while the NiFe WHA experienced stable ductile tearing, the crack propagated unstably (approximately elastically) after reaching the maximum load in the NiCu WHA. The higher K Jm of the NiFe WHA is attributed to its higher flow stress and ductility, along with the energy dissipation and microcrack dilatational stress shielding, provided by a wide distribution of arrested process-zone microcracks, which are largely absent in the NiCu WHA. Since, the 95W-NiFe WHA has distinctly better mechanical properties than 95W-NiCu WHA, it is a more suitable candidate for the fusion reactor application.
Fracture toughness evaluation for thin-shell stainless steel weldment
Integrity of spent nuclear fuel (SNF) interim storage canisters are very important to the safety of the back-end nuclear fuel cycle. Stress corrosion cracking (SCC) potential of interim storage canister has been considered as a high priority. Because no post-weld heat treatment was required for forming these canisters, the high tensile residual stress existed within these canister welds. This can change the fracture resistance capacity significantly as well as increase SCC potential. Due to relative thin shell thickness of a canister weldment, the spiral notch torsion test (SNTT) method of small specimen approach was used to estimate the canister weldment fracture toughness. Most SNTT weldment samples fracture initiation sites are at heat-affected zone (HAZ) regions. The estimated fracture toughness $J_Q'$ for the baseline SS304 steel is at 283 kJ/m 2 . The estimated $J_Q'$ for the SS304/308 weld from HAZ and weld specimens are 126 kJ/m 2 and 148.9 kJ/m 2 , respectively.