The mechanical property performance of pure Mo is strongly controlled by the grain size, oxygen, nitrogen, and carbon concentrations as well as alloy additions. This is true for unirradiated as well as irradiated properties.71,76,83,84 The sensitivity to embrittlement at irradiation temperatures <873 K can be mitigated through a reduction in oxygen and nitrogen while keeping the carbon-to-oxygen ratio high to reduce the segregation of oxygen and nitrogen to the grain boundaries. A reduction in the grain size can further increase the number of sinks and reduce the mean distance that irradiation-induced defects must travel at temperatures at which mobility is limited.

Irradiated mechanical properties of wrought LCAC-Mo in both the recrystallized and stress — relieved conditions have been examined over several decades.81,82,84-86,99-106 In general, LCAC-Mo un­dergoes significant increases in tensile strength through radiation hardening <873 K, which produces reductions in ductility and high DBTT values. A summary of tensile properties as a function of irra­diation temperature and dose is shown in Figure 14. Irradiated stress-relieved LCAC-Mo shows less radia­tion embrittlement than as-crystallized materials at temperatures <1208 K.1 0 However, at higher irradia­tion temperatures, recrystallization of the stress — relieved material occurs, leading to large changes in the microstructure and less desirable properties.

Increases in hardness of 56% and 112% for stress — relieved and recrystallized LCAC-Mo, respectively, are reported following irradiation to 1.2 dpa at 543 K.100 The increase in hardness decreases slightly following 878 K irradiations to 2.4 dpa, but returns to values near those for the unirradiated material for irradiation at temperatures >1208 K. Materials irra­diated in the stress-relieved condition at tempera­tures >1173 K can result in softening compared to unirradiated materials because ofrecrystallization.107 A comparison of the changes in irradiated material hardness as a function of dose and temperature is plotted in Figure 15(a) for LCAC-Mo, TZM, and ODS-Mo. The two latter alloys are discussed in detail later. In the three materials investigated by Cockeram et a/.,107 the largest increase in hardness was measured for irradiations at 873 K, counter to what is observed in tensile properties for irradiation between 573 and 873 K.81,85 However, tensile failure in the lower temperature irradiated samples generally occurs before the samples yield because of the eleva­tion ofthe flow stress above the fracture stress for these test conditions. Recovery of the hardness, measured through 1 h anneals at increasing temperatures, is plot­ted in Figure 15(b) for the ^573 K irradiated material. The start of recovery is near 800 K with full recovery of the LCAC-Mo hardness by 1253 K. For LCAC-Mo irradiated at a higher temperature of 873 K, recovery of hardness begins near 1163 K and is completed near 1463 K.

Substantial increases in the DBTT to values >773 K are observed for irradiated LCAC-Mo over a range of fluences for irradiation temperatures <873 k.26,82,85,103,105,108 A summary of DBTT values for LCAC-Mo is presented in Figure 16, along with data from high-purity grain-refined LCAC-Mo, TZM, and ODS-Mo, which is discussed next. In general, recovery in the DBTT of LCAC-Mo is not observed until irradiation temperatures above 873­973 K, depending on material conditions. A reduced sensitivity to low-temperature irradiation embrittle­ment of LCAC-Mo is observed in materials with reduced levels of impurities. A high-purity form of LCAC-Mo (HP-LCAC-Mo) was developed through the use of 1873 K hydrogen atmosphere annealing of LCAC-Mo plates prior to further arc casting, extru­sion, and rolling into sheet stock.82 Levels of oxygen and nitrogen were < 4 wppm each, with carbon at 20wppm. Average grain diameters of 1.3 and 452 mm lengths were produced, and represent a considerable change in aspect ratio compared to LCAC-Mo values of 4-5 mm diameter and 78-172 mm lengths.100 The DBTT of 573 K irradiated HP-LCAC-Mo showed no increase over the unirradiated value (123 K) for irradiations up to 0.11 dpa, and increased to 723 K by 1.29 dpa. For 873 K irradiations, the DBTT remained below 223 K up to 1.29 dpa.

The majority of the irradiated mechanical prop­erty database for TZM is limited to displacement damage <5 dpa and irradiation temperatures <1000K, with much of the testing conducted at tem­peratures below that used in the irradiation. The data­base is also sparse, with little connectivity between

different experimental examinations. Furthermore, significant differences are observed in the nonirradi — ated tensile values of TZM because of variations in material processing leading to differences in grain size, impurity level, distribution of the strengthening phase, and differences in testing procedure. In general,

I

1100

900

Figure 15 (a) Change in hardness as a function of neutron dose for LCAC-Mo, TZM, and ODS-Mo irradiated between 573 and 1173 K up to 13.1 dpa. (b) Recovery of hardness as a function of isochronal annealing temperature for material irradiated ~573K. Adapted from Cockeram, B. V.; Smith, R. W.; Byun, T. S.; Snead, L. L. J. Nucl. Mater. 2009, 393, 12-21.

Irradiation temperature (°C) Figure 16 Summary of DBTT values as a function of neutron irradiation temperature for LCAC-Mo and TZM. Adapted from Cockeram, B. V.; Smith, R. W.; Leonard, K. J.; Byun, T. S.; Snead, L. L. J. Nucl. Mater. 2008, 382, 1-23; Cockeram, B. V.; Smith, R. W.; Snead, L. L. J. Nucl. Mater. 2005, 346, 165.

LCAC-Mo • Wiffen26 (13.3-23.4 dpa) A Chakin and Kazakov108 (2.1-3dpa) О Hasegawa eta/.106 (10.5-50dpa) Abe et a/.104,105 (0.6-1.4dpa) ▼ Webster et a/.148 (7.4-10 dpa) ф Smith and Michel149 (5.3 dpa) Scibetta et a/.109 (0.15-0.19 dpa) Cockeram et a/.85 (12.3-13dpa) HP-LCAC-Mo Q Cockeram et a/.82 (1.29 dpa) TZM A Cockeram et a/.85 (12.3-13dpa) ■ Wiffen26 (16 dpa) ODS-Mo ^ Cockeram et a/.85 (12.3-13.1 dpa) <§> Cockeram et a/.82 (1.29 dpa)

displacement damage to the TZM alloy produces a large increase in the yield strength of the material with a corresponding drop in total elongation, with the level of change increasing with dose and/or lower irradiation temperatures. Some recovery of properties begins to be observed at test conditions above 973 K for materials irradiated at lower temperatures. A compila­tion of earlier and more recent tensile data for irra­diated TZM is provided in Figure 17. Irradiation <1.2 dpa and temperatures >800 K showed little effec­tive strengthening above the unirradiated values, for tensile tests below the irradiation temperature, 85,109,110 though higher displacement doses resulted in a signifi­cant increase.81 Total elongation in material irradiated < 1000 K was limited, with uniform elongation values <1%.m Plastic instability following yielding was also observed in irradiated TZM as well as LCAC-Mo. The plastic instability stress, defined as the maximum true stress in which plastic necking occurs, is strongly dependent on test temperature but nearly independent of fluence and irradiation temperature.81

Increases in the DBTT for irradiated TZM are significant and do not diminish until irradiation tem­peratures slightly higher than that of LCAC-Mo,81 but comparisons may be difficult because of differ­ences in materials and test methods. The DBTT value for unirradiated TZM is ^200 K and increases with increasing irradiation damage. An increase of 230 K in the DBTT over unirradiated values was reported for TZM irradiated to 2-4.8 dpa at 644-661 K.111 DBTTs
of 750 K for Mo-0.5Ti irradiated to ^16 dpa at 727 K26 and 1073 K for TZM irradiated to 12.3 dpa at 573 K81,85 were also reported. A summary of DBTT as a function ofirradiation temperature for LCAC-Mo and TZM is shown in Figure 16 for irradiation doses <50 dpa.8 Excessive embrittlement is observed for irradiations at <773 K. A reduction in the DBTT for LCAC-Mo appears near 873 K, while the DBTT of TZM remains high. Recovery of DBTT to near unir­radiated values occurs for irradiations at >1073 K.81, The stage V recovery temperature for vacancy diffu­sion in Mo is ~-873 K; however, the kinetics for micro­structural changes to occur are still relatively slow at this temperature. Therefore, embrittlement issues can be present until 1073 K.

Very little fracture toughness data exist for irradiated TZM. For precracked compact tension specimens irradiated to 0.29-0.35 dpa at 313-695 K, a 4MPaVm decrease in fracture toughness (15- 20MPaVm at Trm, unirradiated84,109) was observed up to the irradiation temperature.109 Kitsunai etal.112 examined the impact toughness of irradiated TZM and alloys, incorporating 0.1-1 wt% TiC additions to Mo. The Mo-TiC alloys showed dramatically improved toughness levels over TZM with increasing TiC concentration in samples irradiated to 0.08 dpa between temperatures of 573 and 773 K. Shown in Figure 18 is the shift in DBTT to lower tempera­tures for the irradiated TiC-containing Mo over the TZM alloy. For the 1 wt% containing sample,

the DBTT remained unchanged with irradiation, despite a ^50% increase in Vickers microhardness. More surprisingly, the Mo-1%TiC sample increased in toughness following 0.8 dpa irradiation attributed to grain boundary strengthening by (Ti, Mo)C radiation-enhanced precipitates.

Developed to improve the low-temperature duc­tility and weld characteristics of unalloyed Mo, the Mo-Re alloys have gained considerable attention over the past decade for use in nuclear applications. Single-phase solid solution a-Mo phase field extends
up to ^42 wt% Re, above which the a-MoRe2 phase precipitates. At higher Re concentrations, the W-MoRe3 phase is present. However, the exact phase boundaries are not well delineated at temperatures below 1773 K113,114 mainly because of the slow kinet­ics in phase development.115

The Mo-Re alloys show a hardening response to irradiation stronger than that of the pure metal and TZM following irradiation.116-118 The hardening response of Mo-Re alloys ranging in composition from 2 to 13 as well as 41 wt% Re following

Figure 18 Total absorbed energy versus test temperature for TZM and 0.1-1.0 wt% TiC additions to Mo irradiated to 0.08dpa at 573-773 K. Reproduced from Kitsunai, Y.; Kurishita, H.; Narui, M.; Kayano, H.; Hiraoka, Y. J. Nucl. Mater. 1996, 239, 253-260.

irradiation up to 20 dpa at temperatures between 681 and 1072 Kwas examined by Nemoto etal.116 A linear increase in hardness with Re concentration was ob­served for the unirradiated controls as well as samples irradiated at temperatures <874 K. For samples irra­diated at 1072 K, little variation was observed with increasing Re content, though hardness values remained nearly double that of the unirradiated mate­rial. The dependence of hardness on irradiation tem­perature and fluence for Mo-5Re and Mo-41Re in comparison with LCAC-Mo is presented in Figure 19.

The high degree of radiation hardening at tem­peratures <1000 K exhibited by the Mo-Re alloys is further reflected in low ductility and reported embrittlement. Tensile elongation values of <0.3% were reported for Mo-5Re irradiated to 0.16 dpa and tested at the irradiation temperature of 320 K,11 though higher total elongations of 8% were reported for Mo-5Re fast reactor irradiated to 0.29 dpa and tested near the irradiation temperature of ^723 K.1 Unlike LCAC-Mo and TZM, the little deformation occurring in Mo-Re is mostly uniform at tempera­tures below 1000 K,11 ,119 while a small degree of work softening has been observed at higher tempera — tures.120 No evidence of dislocation channeling was found during microstructural examination of tensile tested Mo-5Re irradiated to 0.16 dpa at 373 K.118

For comparable irradiation fluences, dislocation loop concentrations are approximately two times higher than TZM and four to six times higher than Mo, while dislocation loop diameters are smaller in the Mo-5Re alloy.118 Void development begins to appear in Mo-Re alloys above 623 K118 and shows a slight increase in size with Re concentration (with corresponding decrease in number density) up to 10wt% with no further increase for the 41% Re alloy.116 Irradiation-induced void swelling in Mo with Re concentrations <13wt% is 0.5—1.5% for ^21 dpa irradiated material at 681-1072 K, while swelling for 41 wt% Re was near 0.1%.116

Radiation-induced precipitation has been reported by Nemoto eta/.116 in Mo—(2—41)Re alloys irradiated to 21 dpa between 681 and 1072 K, and by Edwards et a/.121 in Mo—41Re irradiated 28—96 dpa at 743— 1003 K. An initial formation of hcp-structured precipi­tates with a thin plate-like morphology consisting of solid solution Re and Os was observed, appearing with the {110}Mo//{0001}Re, <111>Mo//<2110>Re orien­tation relationship.121 The precipitation of these plates on dislocation loops resulted in the high density ofplates observed, which dominates the microstructure. On fur­ther irradiation to higher doses or higher temperatures, these plates develop and coarsen into the w-phase. This nonequilibrium phase development in Re-lean alloys was originally observed by Erck and Rehn12 in Mo—(27—30)Re irradiated by 1.8 MeV He ions at 1023— 1348 K. The a-MoRe2 phase was also reported appear­ing in all the Mo—Re samples examined by Nemoto and coworkers,116 but was suppressed in the stress-relieved specimens compared to the recrystallized materials.

While a limited (<1%) amount of ductility was reported in Mo—5Re alloys irradiated at tempera­tures <1000K to 34dpa,109,118,119 embrittlement of Mo—(1—20)Re irradiated 723—1073 K in a fast reactor up to 5 dpa and Mo—(13 and 47)Re irradiated 373— 673 K in a mixed spectrum reactor to 2 dpa has been reported by Fabritsiev and Pokrovsky.62 The reduc­tion in tensile strength, in many cases below the unir­radiated values with no plasticity occurring in the samples, was attributed to the hardening of the mate­rial by radiation-created defects along with RIS of oxygen, nitrogen, and transmuted impurities to the grain boundaries. The oxygen and nitrogen content in the embrittled alloys was reported to be near 70 appm.

Irradiation hardening above 900 MPa was also observed in Mo—41Re and Mo—47.5Re samples irra­diated to 1.46 dpa at temperatures >1073 K.120 Failure of Mo—41Re samples irradiated to 1.46 either prior to yielding or after ^5% elongation upon reaching 1600 MPa was observed at 1073 K (Tirr = Ttest). Exam­ples of the tensile curves for the two alloys in the irradiated, 1100 h aged and as-annealed condition tested at 1073 K is shown in Figure 20. Radiation

Irradiation temperature (K)

Mo (Nemoto et a/.116 18-21 dpa) ■ Mo-5Re (Nemoto et a/.116 18-21 dpa)

A Mo-10Re (Nemoto et a/.116 18-21 dpa) x Mo-41Re (Nemoto et a/.116 18-21 dpa)

♦ Mo-5Re (Hasegawa et a/.117 6.8-34 dpa) —O— Mo-5Re (Hasegawa et a/.117 6.8-34 dpa)

—□— Mo-41Re (Hasegawa et a/.117 6.8-34 dpa)

Figure 19 Vickers hardness as a function of neutron irradiation temperature and dose for LCAC-Mo, Mo-5Re, and Mo-41 Re alloys. Displacement damage levels are provided in the key. Reproduced from Nemoto, Y.; Hasegawa, A.; Satou, M.; Abe, K.; Hiraoka, Y. J. Nucl. Mater. 2004, 324, 62-70; Hasegawa, A.; Ueda, K.; Satou, M.; Abe, K. J. Nucl. Mater. 1998, 258-263, 902-906.

Figure 20 Comparison of stress-strain curves for neutron irradiated, 1100 h and as-annealed Mo-41Re and Mo-47.5 Re samples at 1073 K (Tirr = Ttest). Adapted from Busby, J. T.; Leonard, K. J.; Zinkle, S. J. Effects of neutron irradiation on refractory metal alloys, ORNL/LTR/NR-PROM1/05-38; Oak Ridge National Laboratory: Oak Ridge, TN, Dec 2005; Busby, J. T.; Leonard, K. J.; Zinkle, S. J. J. Nucl. Mater. 2007, 366, 388-406.

hardening to levels over twice the as-annealed condi­tion was observed for the alloys irradiated at 1223 and 1373 K; however, total elongation was between 4 and 12%. Analysis of the fractured surfaces of these samples revealed intergranular failure, with the sever­ity increasing with irradiation temperature. A compari­son of mechanical property data of Mo-Re samples from the sources discussed is shown in Figure 21.

The degree of RIS influencing the properties of Mo-Re alloys varies with temperature, dose rate, and total fluence. At temperatures <0.3 Tm, the recombi­nation of vacancies and interstitials generated by displacement damage dominates because of the lim­ited defect mobility, and therefore RIS is not a factor. At temperatures >0.5 Tm, a reduced driving force for segregation occurs because of the high thermal defect concentrations. At intermediate temperatures (^850-1430 K for Mo-Re), the radiation generated point defects diffuse to defect sinks such as grain
boundaries or dislocations. Any preferential coupling of vacancies or interstitial defects fluxes with solute atoms, including transmuted species, will create enrichment at the defect sinks. This is observed in the nucleation of Re-rich phases in the microstruc­tures of neutron-irradiated samples116,121 and the degradation in mechanical properties and transition to intergranular fracture in higher Re concentration alloys.62,120 Further information on RIS can be found in Chapter 1.18, Radiation-Induced Segregation.

Through modeling and experimental work, Erck and Rehn123 showed that the degree of segregation per dpa reaches a maximum for Mo-30 at.% Re (^45 wt% Re) near 1223 K and that for Mo-7 at.% Re (~13wt% Re) near 1473 K. While the Mo-5Re alloys irradiated up to 20 dpa show some limited ductility,109,118,119 the maximum irradiation tempera­tures were <1073 K and are therefore at or below the lower temperature limit expected for RIS.

The Mo—(41 and 47.5)Re alloys irradiated at 1073— 1373 K120 showed indications of RIS even at relatively low damage levels, part of which may have been a contribution of a thermal aging component, which in the unirradiated as-aged Mo-41Re and Mo—47.5Re showed increases in Re at the grain boundaries, lead­ing to precipitation of s — and w-phases at the grain boundaries in the 47.5Re containing alloy.115 Utilizing Mo-Re alloys with a more moderate Re content may improve the irradiation performance of these alloys, especially when considering higher doses and/or lon­ger irradiation times at temperatures at which thermal precipitation effects may further compound RIS influ­ences on mechanical properties.

Additional information on the fracture toughness data for Mo-Re alloys is also needed. Preliminary data by Scibetta and coworkers109 on precracked com­pact tension specimens of Mo-5Re showed reduc­tions in fracture toughness from unirradiated values of 17-23 MPa Vm at room temperature and 623 K, to ^11MPaVm for 0.35 dpa irradiated at 313 K, and 15MPaVm for 0.29 dpa irradiated at 643 K. These low irradiated values, at which no ductile crack growth was observed in the specimens, are a concern.

Recent work examining the irradiated properties of wrought, commercially available, ODS-Mo contain­ing lanthanum oxide particles has shown promising results.81,82,124 The fabrication methods produce a microstructure consisting of elongated grains with appreciable texturing and alignment of the oxide par­ticles. The high degree of working associated with fabrication produces a <2 pm grain size, which is stabilized from growth by the ODS particles. Irradia­tion of the ODS-Mo up to 13.1 dpa at 567 K and 883-882 K produced an increase in yield strength of 57-173%,124 while irradiation at 1143-1273 K produced a 10-34% increase. The irradiated tensile properties of ODS-Mo as a function of irradiation temperature and dose from the work of Cockeram and coworkers124 are shown in Figure 22. The increases in radiation strength are comparable to the higher limits for LCAC-Mo. The most striking result of the ODS work is the improvement in the DBTT for the irradiated samples.82,124 For 567 K irradiation to 12.3 dpa, the DBTT is 1073 K and is comparable to that of LCAC-Mo and TZM (Figure 16). How­ever, the DBTT for ODS-Mo irradiated to 13.1 dpa at 833-882 K is ^298 K, while that of LCAC-Mo is 573 and 973 K for TZM. For irradiation to 13.1 dpa at 1143-1209 K, the DBTT is 173 K, unchanged from the nonirradiated material, while those of LCAC-Mo and TZM are between 223-273 K.

Figure 22 Yield stress and total elongation as a function of test temperature for lanthanum oxide ODS-Mo neutron irradiated up to 13 dpa at temperatures between 567 and 1209 K. Adapted from Cockeram, B. V.; Smith, R. W.; Snead, L. L. J. Nucl. Mater. 2005, 346, 165-184.

The reduced susceptibility to irradiation embrit­tlement of ODS-Mo is in part due to the grain size reduction and presence ofthe oxide particles. Reduc­ing the distance of possible defect sinks such as grain boundaries and offering additional sites such as the oxide/matrix interface are particularly critical at lower irradiation temperatures at which defect mobility is limited. In addition, Cockeram and coworkers81,12 describe the fine but elongated grain structure as enhancing the plain strain condition acting on each plane of the lamina-shaped grains formed during the fracture process, inducing larger plastic deformation in irradiation-hardened material. This is also true for the HP-LCAC-Mo containing the high aspect ratio grains compared to other forms of LCAC-Mo produced. Currently, no fracture data on the irradiated

ODS-Mo are available. Unirradiated fracture tough­ness values are between 23 and 38 MPaVm, depending on the grain orientation tested.83