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In part I of this helium-effects study on ferritic/martensitic steels, results were presented on tensile and Charpy impact properties of 9Cr–1MoVNb (modified 9Cr–1Mo) and 12Cr–1MoVW (Sandvik HT9) steels and these steels containing 2% Ni after irradiation in the High Flux Isotope Reactor (HFIR) to 10–12 dpa at 300 and 400 °C and in the Fast Flux Test Facility (FFTF) to 15 dpa at 393 °C. The results indicated that helium caused an increment of hardening above irradiation hardening produced in the absence of helium. In addition to helium-effects studies on ferritic/martensitic steels using nickel doping, studies have also been conducted over the years using boron doping, ion implantation, and spallation neutron sources. In these previous investigations, observations of hardening and embrittlement were made that were attributed to helium. In this paper, the new results and those from previous helium-effects studies are reviewed and analyzed.

Tensile and Charpy specimens of 9Cr–1MoVNb (modified 9Cr–1Mo) and 12Cr–1MoVW (Sandvik HT9) steels and these steels doped with 2% Ni were irradiated at 300 and 400 °C in the High Flux Isotope Reactor (HFIR) up to ≈12 dpa and at 393 °C in the Fast Flux Test Facility (FFTF) to ≈15 dpa. In HFIR, a mixed-spectrum reactor, ( n , α ) reactions of thermal neutrons with 58 Ni produce helium in the steels. Little helium is produced during irradiation in FFTF. After HFIR irradiation, the yield stress of all steels increased, with the largest increases occurring for nickel-doped steels. The ductile–brittle transition temperature (DBTT) increased up to two times and 1.7 times more in steels with 2% Ni than in those without the nickel addition after HFIR irradiation at 300 and 400 °C, respectively. Much smaller differences occurred between these steels after irradiation in FFTF. The DBTT increases for steels with 2% Ni after HFIR irradiation were 2–4 times greater than after FFTF irradiation. Results indicated there was hardening due to helium in addition to hardening by displacement damage and irradiation-induced precipitation.

Reduced activation ferritic/martensitic steel, like F82H has been developed as a structural material for in vessel components because of its superior resistance to irradiation damage. As a blanket fabrication process, hot isostatic pressing (HIP) bonding has the great merit of near-net-shaping processing. The degassing conditions and surface roughness were investigated as parameters of HIP conditions. Although the surface roughness and degassing conditions had slight effects on tensile properties, the lack of degassing caused significant degradation of impact properties. A dissimilar metal joint between sintered tungsten and F82H was fabricated by a spark plasma sintering (SPS) method. The joint had no defects in spite of the large difference in thermal expansion coefficient between tungsten and F82H. It is considered that formation of a compliant layer of the ferritic phase can lead to successful bonding for the tungsten and F82H joint even without an artificial interlayer.

The dependence of helium production on radiation hardening and -embrittlement has been examined in a reduced-activation martensitic F82H steel (8Cr–2W–0.2V–0.04Ta–0.1C) irradiated at 250 °C to 2.3 dpa. In this study, 10 B and 11 B-doped specimens were irradiated to minimize the errors from the effect of B on mechanical properties by comparing the results. The specimens used were 10 B-doped, 10 B + 11 B-doped and 11 B-doped F82H steels. The total amounts of doping boron were about 60 mass ppm. The range of helium concentration produced in the specimens was from about 5 to about 330 appm. Tensile and fracture toughness tests were performed after neutron irradiation. 50 MeV-He 2+ irradiation was also performed to implant about 85 appm He atoms at 120 °C by AVF cyclotron to 0.03 dpa, and small punch testing was performed to obtain ductile-to-brittle transition temperatures (DBTT). Radiation hardening of the neutron-irradiated specimens increased slightly with increasing helium production. The 100 MPa m 1/2 DBTT for the F82H + 11 B, F82H + 10 B + 11 B, and F82H + 10 B specimens were 40, 110, and 155 °C, respectively. The shifts of DBTT due to helium production were evaluated as about 70 °C by 190 appm He and 115 °C by 330 appm He. In cyclotron experiment using standard F82H, a similar DBTT shift due to He was measured. These results suggest that helium production can increase the DBTT.

F82H (Fe–8Cr–2W) and its variant doped with 2%Ni were irradiated up to 20 dpa at 300 °C in the High Flux Isotope Reactor (HFIR) at Oak Ridge National Laboratory. Post irradiation tensile testing was performed at room temperature. During testing, the images of the specimens including the necked region were continuously recorded. Tests on cold worked material were also carried out for comparison. From the load–displacement curves and the strain distributions obtains from the images, flow stress levels and strain hardening behavior was evaluated. A preliminary constitutive equation for the plastic deformation of irradiated F82H is presented. The results suggest that the irradiation mainly causes defect-induced hardening while it did not strongly affect strain hardening at the same flow stress level for F82H irradiated at 300 °C. The strain hardening of Ni doped specimens was, however, strongly affected by irradiation. Results provide basics to determine allowable stress levels at temperatures below 400 °C.

The objectives of this work are to evaluate the effects of thermal history corresponding to a blanket fabrication process on Reduced Activation Ferritic/Martensitic steel (RAF/Ms) microstructure, and to establish appropriate Hot Isostatic Pressing (HIP) conditions without degradation in the microstructures. One of RAF/Ms F82H and its modified versions were investigated by metallurgical methods after isochronal heat treatments up to 1473 K simulating HIP thermal history. Although conventional F82H showed significant grain growth after conventional solid HIP conditions, F82H with 0.1 wt% tantalum maintained a fine grain structure after the same heat treatment. It is considered that the grain coarsening was caused by dissolution of tantalum-carbide which immobilizes grain boundaries. On the other hands, conventional RAF/Ms with coarse grains were recovered by post HIP normalizing at temperatures below the TaC solvus temperature. This process can refine the grain size of F82H to more than ASTM grain size number 7.

The effects of irradiation on the Charpy impact properties of reduced-activation ferritic/martensitic steels were investigated on a microstructural basis. It was previously reported that the ductile–brittle transition temperature (DBTT) of F82H-IEA and its heat treatment variant increased by about 130 K after irradiation at 573 K up to 5 dpa. Moreover, the shifts in ORNL9Cr–2WVTa and JLF-1 steels were much smaller, and the differences could not be interpreted as an effect of irradiation hardening. The precipitation behavior of the irradiated steels was examined by weight analysis and X-ray diffraction analysis on extraction residues, and SEM/EDS analysis was performed on extraction replica samples and fracture surfaces. These analyses suggested that the difference in the extent of DBTT shift could be explained by (1) smaller irradiation hardening at low test temperatures caused by irradiation-induced lath structure recovery (in JLF-1), and (2) the fracture stress increase caused by the irradiation-induced over-solution of Ta (in ORNL9Cr–2WVTa).

A Japanese ITER test blanket module (TBM) is planed to use reduced-activation martensitic steel F82H. Feasibility of F82H for ITER test blanket module is discussed in this paper. Several kinds of property data, including physical properties, magnetic properties, mechanical properties and neutron-irradiation data on F82H have been obtained, and these data are complied into a database to be used for the designing of the ITER TBM. Currently obtained data suggests F82H will not have serious problems for ITER TBM. Optimization of F82H improves the induced activity, toughness and HIP resistance. Furthermore, modified F82H is resistant to temperature instability during material production.

The Japanese reduced activation ferritic steels (RAFSs) R&D road map towards DEMO is shown. The important steps include high-dose irradiation in fission reactors such as the high flux isotope reactor at Oak Ridge National Laboratory, irradiation tests with 14 MeV neutrons in the International Fusion Materials Irradiation Facility and application to ITER test blanket modules to provide an adequate database of RAFSs for the design of DEMO. The current status of RAFS development is also introduced. The major properties of concern are well-known, and process technologies are mostly ready for fusion application. RAFSs are now certainly ready to proceed to the next stage. A materials database is already in hand, and further progress is anticipated with the design of the ITER test blanket. Oxide dispersion strengthening steels are quite promising for high temperature operation of the blanket system, with potential improvements in radiation resistance and in corrosion resistance.

In order to determine the contributions of different microstructural features to strength and to deformation mode, microstructure of deformed flat tensile specimens of irradiated reduced activation F82H (IEA heat) base metal (BM) and its tungsten inert-gas (TIG) weldments (weld metal and weld joint) were investigated by transmission electron microscopy (TEM), following fracture surface examination by scanning electron microscopy (SEM). After irradiation, the fracture surfaces of F82H BM and TIG weldment showed a martensitic mixed quasi-cleavage and ductile-dimple fracture. The microstructure of the deformed region of irradiated F82H BM contained dislocation channels. This suggests that dislocation channeling could be the dominant deformation mechanism in this steel, resulting in the loss of strain-hardening capacity. While, the necked region of the irradiated F82H TIG, where showed less hardening than F82H BM, showed deformation bands only. From these results, it is suggested that the pre-irradiation microstructure, especially the dislocation density, could affect the post-irradiation deformation mode.

Previous work reported that a TIG weld joint of F82H exhibited low irradiation hardening in a tensile test, compared to the base metal. Microhardness tests and microstructure observation on the neutron-irradiated TIG weld joint of F82H revealed that the over-tempered zone in the heat-affected zone (HAZ) exhibited this good performance. The region in the HAZ where the prior austenite grain size became very fine during welding also exhibited lower irradiation hardening. Hypotheses for these low-hardening mechanisms were proposed based on the phase diagram and grain size.

The effects of irradiation up to 20 dpa on the Charpy impact properties of reduced-activation ferritic/martensitic steels (RAFs) were investigated. The ductile-brittle transition temperature (DBTT)...

The synergistic effect of displacement damage and hydrogen or helium atoms on microstructures in F82H steel irradiated at 250–400 °C to 2.8–51 dpa in HFIR has been examined using isotopes of 54 Fe or 10 B . Hydrogen atoms increased slightly the formation of dislocation loops and changed the Burgers vector for some parts of dislocation loops, and they also affected on the formation of cavity at 250 °C to 2.8 dpa. Helium atoms also influenced them at around 300 °C, and the effect of helium atoms was enhanced at 400 °C. Furthermore, the relations between microstructures and radiation-hardening or ductile to brittle transition temperature (DBTT) shift in F82H steel were discussed. The cause of the shift increase of DBTT is thought to be due to the hardening of dislocation loops and the formation of α′-precipitates on dislocation loops.

An alloy of F82H prepared using the isotope 54Fe in order to encourage H and He production in a fission reactor has been irradiated in the HFIR JP20 experiment at three temperatures to 7 dpa as TEM disks. Irradiated disks were shear punch tested, examined by TEM, analyzed for He and H content, and compared with previous results in order to quantify irradiation hardening due to transmutation-induced H and He. Hardening due to irradiation is found following irradiation at 300 and 400 °C, that is intermediate between that at lower and higher dose, but hardening is negligible following irradiation at 500 °C. Microstructural examinations show typical behavior of irradiation as a function of irradiation temperature, with moderate swelling after 400 °C irradiation but few bubbles after irradiation at 300 °C. Correlations of change in hardening with He and H content show little indication of transmutation-induced hardening, but measured H levels do not agree with predictions and therefore H production and analysis requires further study.

In the absence of a 14 MeV neutron source, the effect of helium on structural materials for fusion must be simulated using fission reactors. Helium effects in ferritic/martensitic steels have been studied by adding nickel and boron and irradiating in a mixed-spectrum reactor. Although the nickel- and boron-doping techniques have limitations and difficulties to estimate helium effects on the ferritic/martensitic steels, past irradiation experiments using these techniques have demonstrated similar effects on the swelling and Charpy impact properties that are indicative of a helium effect. Although both techniques have disadvantages, it should be possible to plan experiments using the nickel- and boron-doping techniques to develop an understanding of the effects of helium on mechanical properties.

This paper presents materials data for design of ITER test blanket modules with the reduced activation ferritic martensitic steel type F82H as structural material. From the physical properties databases, variations of modulus of elasticity, density, thermal conductivity, thermal diffusivity, specific heat, mean and instantaneous linear coefficients of thermal expansion versus temperature are derived. Also reported are Poisson's ratio and magnetic properties. From the tension test results, nominal and minimum stress values of S y and S u are derived and used for calculation of allowable primary membrane stress intensity S m . Likewise, uniform and total elongations, as well as reduction of area data, are used for calculation of minimum and true ductility at rupture values. From the instrumented Charpy impact and fracture toughness test data, ductile to brittle transition temperature, toughness and behavior of material in different fracture modes are evaluated. The effect of specimen size and geometry are discussed but preference is given to standard size specimens. From the fatigue data, total strain range versus number of cycles to failure curves are plotted and used to derive fatigue design curves, using a reduction factor of 2 on strain and a reduction factor of 20 on number of cycles to failure. Cyclic hardening curves are also derived and compared with monotonic hardening curves. From the creep data, time dependent allowable stresses S r and S t are calculated. Combination of tension and creep results are used to deduce S mt and isochronus curves. Finally, irradiated and aged materials data are compared to insure that the safety margins incorporated in unirradiated design limits are not exceeded.

Miniature tensile and Charpy specimens of four ferritic/martensitic steels were irradiated at 300°C and 400°C in the high flux isotope reactor (HFIR) to a maximum dose of ≈12 dpa. The steels were standard F82H (F82H-Std), a modified F82H (F82H-Mod), ORNL 9Cr–2WVTa, and 9Cr–2WVTa–2Ni, the 9Cr–2WVTa containing 2% Ni to produce helium by (n,α) reactions with thermal neutrons. More helium was produced in the F82H-Std than the F82H-Mod because of the presence of boron. Irradiation embrittlement in the form of an increase in the ductile–brittle transition temperature (ΔDBTT) and a decrease in the upper-shelf energy (USE) occurred for all the steels. The two F82H steels had similar ΔDBTTs after irradiation at 300°C, but after irradiation at 400°C, the ΔDBTT for F82H-Std was less than for F82H-Mod. Under these irradiation conditions, little effect of the extra helium in the F82H-Std could be discerned. Less embrittlement was observed for 9Cr–2WVTa steel irradiated at 400°C than for the two F82H steels. The 9Cr–2WVTa–2Ni steel with ≈115 appm He had a larger ΔDBTT than the 9Cr–2WVTa with ≈5 appm He, indicating a possible helium effect.

F82H(8Cr–2WVTa steel) IEA heat was used to prepare tungsten-inert-gas (TIG) and electron-beam (EB) weld joints, followed by heat treatment at 720°C for 1 h. Hardening in the weld metal and softening in the heat-affected zone (HAZ) were detected in TIG weld joints. In EB weld joints, hardening in the weld metal was more clearly observed but HAZ softening was hardly observed. Hardness of TIG weld metal was reduced after 550°C thermal-aging, but softening of the base metal was only observed after 650°C thermal-aging. M 23 C 6 phase was the major precipitate in aged base metal and weld joints. The amount of precipitates in aged weld metal was lower than that of normalized and tempered base metal. W-rich Laves phase was also detected in aged weld metal, HAZ and base metal.

Reduced-activation ferritic/martensitic steel, F82H (8Cr–2W–0.2V–0.04Ta–0.1C), and variants doped with isotopically tailored boron were irradiated at 673 K up to 51 dpa in the high flux isotope reactor (HFIR). The concentrations of 10B in these alloys were 4, 62, and 325 appm during HFIR irradiation which resulted in the production of 4, 62 and 325 appm He, respectively. After irradiation, transmission electron microscopy (TEM) was carried out. The number density of cavities increased and the average diameter of cavities decreased with increasing amounts of 10B. The number density decreased and the average diameter increased with increasing displacement damage. Swelling increased as a function of displacement damage and He concentration.

F-82H alloys isotopically enriched in 54 Fe up to 86% were irradiated in the high flux isotope reactor (HFIR) to determine the accelerated production of helium and hydrogen due to isotopic effects. Results are compared to calculations using isotopic helium production cross-sections from ENDF/B-VI or GNASH and measured neutron spectra. Helium measurements demonstrated an accelerated helium (appm)/dpa ratio of 2.3 after a 1.25-year irradiation, an increase of a factor of 4.3 over natural iron. The accelerated helium production is due to higher helium production cross-sections for 54 Fe and 55 Fe. Alloys doped with 55 Fe could achieve helium/dpa ratios up to about 20, well above the fusion reactor ratio of 10. Hydrogen measurements were performed using a newly developed quadrupole mass spectrometer system at PNNL. Calculations predict that hydrogen production will be accelerated by about a factor of 13 over natural iron. However, measurements show that most of this hydrogen is not retained in the samples.

The effects of helium production and heat treatment on the swelling of F82H steel irradiated in the HFIR to 51 dpa have been investigated using 10B, 58Ni and 60Ni-doped specimens. The swelling of tempered F82H-std and F82H doped with 10B irradiated at 400°C ranged from 0.52% to 1.2%, while the swelling of the non-tempered F82H doped with 58Ni or 60Ni was less than 0.02%. At 300°C the swelling in all steels was insignificant. In the F82H + Ni, a high number of density carbides formed in the matrix at these temperatures. The production of helium atoms enhanced the swelling of the F82H steel. However, the non-tempered treatment for the F82H + Ni suppressed remarkably the swelling. The cause of low swelling in the F82H + Ni may be due to the occurrence of the high density of carbides acting as sinks or the decrease of mobility of vacancies interacting with carbon atoms in matrix.

The effects of neutron spectrum on tensile properties of the low-activation martensitic steel F82H (8Cr–2WVTa) was examined using a thermal neutron shield to tailor the neutron spectrum for steels irradiated in the high flux isotope reactor (HFIR). The yield stresses of spectrally tailored specimens irradiated in HFIR to 5 dpa at 300°C and 500°C are on trend lines obtained from unshielded irradiation in HFIR. No significant effect of the neutron spectrum on tensile properties could be detected.

Microstructural evolution of welded austenitic stainless steel irradiated with mixed-spectrum neutrons was examined by transmission electron microscopy (TEM). TEM disks were obtained from electron-beam (EB) welded plates of JPCA, which is a Ti-midified austenitic stainless steel. Specimens were irradiated in HFIR up to 17 dpa at 670 and 770 K, and the estimated helium concentration was around 1100 appm. Cavities formed at 670 K irradiation were very small (

The austenitic stainless steel (Fe–0.06C–0.5Si–1.8Mn–14Cr–16Ni–2Mo–0.24Ti) was irradiated in a triple ion facility and the High Flux Isotope Reactor. The materials used were in the solution annealed (SA) and 15% cold-worked (CW) condition. TEM and tensile specimens were irradiated to a dose level of 30 and 10 dpa at 200°C. Some of the specimens were annealed after the irradiation at 500°C for 8 h in a vacuum. Microhardness tests were carried out on the surface of the TEM disks at room temperature. Tensile tests were carried out at 200°C in a vacuum with strain rate of about 1×10−3 s−1. The microhardness of both SA and CW increased by ion irradiation and then decreased by annealing. The yield strengths of the neutron irradiated SA and CW decreased to 610 and 650 MPa by annealing, respectively. The strain to necking of the irradiated CW recovered from 0.7% to 7.6%. The fracture mode remained ductile in each case.

Disk compact tension and sheet tensile specimens of the ferritic-martensitic steels F82H and Sandvik HT-9 were irradiated in the High Flux Isotope Reactor (HFIR) at 90°C and 250°C to neutron doses of 1.5–2.5 dpa. For both steels, radiation hardening was accompanied by a reduction in strain hardening capacity (SHC). When combined with other literature data it is apparent that severe loss of SHC occurs in F82H for irradiation temperatures below ∼400°C and in HT-9 for irradiation temperatures below ∼250°C. For both alloys, severe loss of SHC does not correlate with brittle behavior during fracture toughness testing.

Vanadium, V1Ni, V10Ti and V10Ti1Ni (at%) were irradiated in HFIR to doses ranging from 18 to 30 dpa and temperatures between 300 and 600°C. Since the irradiation was conducted in a highly thermalized neutron spectrum without shielding against thermal neutrons, significant levels of chromium (15–22%) were formed by transmutation. The addition of such large chromium levels caused strong embrittlement. At higher irradiation temperatures radiation-induced segregation of transmutant Cr and solute Ti at specimen surfaces caused strong increases in the density of the alloy. The resultant shrinkage, possibly compounded by thermal cycling, led to cracks developing at all intersections of grain boundaries with the specimen surface. This caused specimens irradiated at 500°C or below to often fail during retrieval from the reactor, as well as during electropolishing and other handling operations. At 600°C, the cracking and embrittlement processes are so severe that only a fine dust, composed mostly of individual grains or chunks of grains, was found in the irradiation capsule.

Tensile and Charpy impact properties of neutron irradiated F82H (Fe8Cr2WVTa) with and without boron have been investigated to obtain the basic irradiation response on mechanical properties in low damage regime less than 1 dpa at the temperature ranging from 300° to 590°C. Boron-doped steel was used for the helium effect due to (n, α) reaction. Typical irradiation hardening was observed at 300°C. The irradiation above 520°C did not reveal increase in yield stress, but the specimen irradiated at 590°C showed some reduction in elongation in room temperature tensile testing. Slight difference in the tensile properties between boron-doped and boron-free were observed at 590°C. No changes in ductile brittle transition temperature (DBTT) occurred at a temperature between 335° and 460°C by Charpy impact testing.

Two experiments have been conducted to quantify the effects of neutron irradiation on the deformation and fracture behavior of solution annealed austenitic stainless steels irradiated to doses ranging from 3 to 19 dpa at temperatures from 60 to 400°C. For all alloys, yield strength increases rapidly with dose in the 60–300°C regime. Radiation hardening is accompanied by changes in the flow properties with the appearance of an initial yield drop and a significant reduction in strain hardening capacity. The magnitude of the changes is dependent upon both neutron dose and irradiation temperature, with reductions in strain hardening capacity occurring most rapidly in the range 250–350°C. It is shown that for neutron doses up to about 3 dpa, although the changes in deformation mode reduce the fracture thoughness, the toughness remains satisfactorily high.

Addition of 1 at% nickel to vanadium and V-10Ti, followed by irradiation along with the nickel-free metals in HFIR to 2.3x1026 n m−2, E > 0.1 MeV (corresponding to 17.7 doa) at 400°C, has been used to study the influence of helium on microstructural evolution and embrittlement. Approximately 15.3% of the vanadium transmuted to chromium in these alloys. The ∼50 appm helium generated from the 58Ni(n, γ)59Ni(n, α56 Fesequence was found to exert much less influence than either the nickel directly or the chromium formed by transmutation. The V-10Ti and V-10Ti-1Ni alloys developed an extreme fragility and broke into smaller pieces in response to minor physical insults during density measurements. A similar behavior was not observed in pure V or V-1Ni. Helium's role in determination of mechanical properties and embrittlement of vanadium alloys in HFIR is overshadowed by the influence of alloying elements such as titanium and chromium. Both elements have been shown to increase the DBTT rather rapidly in the region of 10% (Cr + Ti). Since Ci is produced by transmutation of V, this is a possible mechanism for the embrittlement. Large effects on the DBTT may have also resulted from uncontrolled accumulation of interstitial elements such as C, N, and O during irradiation.

Type 316 stainless steel from the core of the experimental fast breeder reactor (FBR) JOYO was examined by the slow strain rate tensile (SSRT) test in pure, oxygenated-water and air and by the electrochemical potentiokinetic reactivation (EPR) test to evaluate a susceptibility to the irradiation assisted stress corrosion cracking (IASCC) and the radiation-in-duced segregation (RIS). The solution annealed and 20% cold-worked materials had been irradiated at 425°C to a neutron fluence of 8.3 × 1026n/m2 (〉 0.1 MeV) which is equivalent to 40 displacement per atom (dpa). Intergranular cracking was induced by the SSRT in water at 200 and 300°C, but was not observed on specimen tested in water at 60°C and in air at 300°C. This indicates that irradiation increased a susceptibility to stress corrosion cracking (SCO in water. After the EPR test, grain boundary etching was observed in addition to grain face etching. This suggests Cr depletion may have occurred both at grain boundary and at defect clusters during the irradiation. The results are compared with the behavior of similar materials irradiated with different neutron spectrum.

Since 1975, extensive studies on transient fuel behavior under reactivity initiated accident (RIA) conditions have been continued in the Nuclear Safety Research Reactor (NSRR) of Japan Atomic Energy Research Institute. A new experimental program with preirradiated LWR fuel rods as test samples has recently been started. In this program, transient behavior and failure initiation have been studied with 14 × 14 type PWR fuel rods preirradiated to a burnup of 20 to 42 MWd/kgU. The test fuel rods contained in a capsule filled with the coolant water were subjected to a pulse irradiation in the NSRR to simulate a prompt power surge in an RIA. The effects of preirradiation on the transient fission gas release, pellet-cladding mechanical interaction and fuel failure were clearly observed through the transient in-core measurements and postirradiation examination.

Uranium isotope separation by redox chromatography is analytically studied. The periodic withdrawal of products and tails and the introduction of natural feed are simulated on the assumption of a square cascade for a uranium adsorption band. The influences on the separative power and the lead time until product withdrawal are investigated by varying the magnitude of the isotope separation factor, uranium band length, tails concentration, and so on. Simulating calculations indicate that using ion-exchange resins to achieve uranium isotope separation requires a very long lead time for the production of highly enriched uranium.

The releases of xenon from three (Th, U)O2 specimens with different U contents were measured over a wide range of fission dose from 2.9 × 1019 to 2.2 × 1022 fissions m−3 by using a post-irradiation technique. The releases were found to decrease with dose and to level off at higher doses. Measurements of the changes in lattice parameter and specific surface area of the same specimens enabled one to conclude that the decrease in release originates in the trapping of xenon by the vacancies and vacancy clusters induced by fission fragments. And the release mechanisms of fission gas were proposed based on the proper evaluation of the observation on radiation damage and recovery in oxide fuel.

Reactivity enhancement was studied using a reaction between a laser excited atomic beam and a molecular beam. This method was applied to lithium isotope separation, in which the excited Li isotopic atom reacts with CHCIF 2 . It is found that the specified lithium isotope is enriched in the reaction product LiF, whereas LiCl has no selectivity.

Oxygen-potential (ΔGO2) measurements employing a thermogravimetric method have been performed for Th0.80U0,20O2 + x. A complete set of data is presented at 1273–1473 K in the ranges 2.000 ≲ OM ≲ 2.024 and −95 ≲ ΔGO2$−32 kcal/mol. Partial molar entropies and enthalpies of solution of oxygen in the mixed oxide were derived from the temperature variation of ΔGO2. Vapor pressures over Th0.80U0,20O2 + x at 2000 at 2300 K were calculated from our experimental ΔGO2 data and the known free energies of formation for gaseous and condensed oxides. It is predicted that with an increase in O/M ratio the vapor pressure of UO3(g) increases rapidly while maintaining an extremely lower pressure of ThO2(g).

The release of 133Xe from (Th,U)O2 was studied at a low fission density by using a post-irradiation technique. The uranium concentrations of the specimens ranged from 0.15 to 20 mol% U02. Heating curves of the release gave almost the same pattern in shape, while the total release, which was the combination of an in-pile release and a post-irradiation release up to 1000°C, increased with uranium concentration except for the nominally pure ThO2. Effects of preparation conditions of specimens such as atmosphere, temperature and stoichiometry were also studied and found to be minor. Possible release mechanisms were discussed.

The lattice parameter change of Th 0.937 U 0.063 O 2 fuel kernels for high temperature gas cooled reactors (HTGR) was studied after irradiation from 1.1 × 10 15 to 4.4 × 10 17 f/cm 3 using a micro-beam X-ray diffractometer. It was found that the lattice parameter increases initially in a linear relationship with the increasing fission dose and then reaches a saturation value ( Δ a a ⋍9.4 × 10 −4 ) at about 1 × 10 17 f / cm 3 . The damage volume and the fission-induced defect number per fission were estimated using a damage saturation function for the fission dose dependence of the lattice parameter changes. The isochronal annealing results indicated that the recovery was roughly characterized by two stages. The isothermal annealing results showed that the recovery process seems to obey a second order kinetics with the activation energy depending on the fraction of recovery.

The oxygen-potential dependence of the stability of fission-product phases has been studied for (Th0.81U0.19)O2 simulating 21.5% FIMA. In reducing environments Mo-Ru-Pd alloy and (Ba, Sr)(Zr, Ce)O3 exist stably in the fuel matrix, whereas in oxidizing environments (Ba, Sr)MoO4 and Nd2(Zr, Ce)2O7 become more stable than the perovskite-type zirconate, with oxidative loss of Mo from the alloy. The oxygen-potential threshold for such a change in the chemical state of the fuel is in the range ΔG(O2) = −62.0 to −69.6 kcal/mol at 1500°C. The threshold ΔG(O2)-value will coincide with that for the onset of Mo oxidation in the alloy; i.e., − −65 kcal/mol at 1500°C. The possible role of molybdenum in an oxide fuel pin is briefly considered on the basis of oxidation behavior of the molybdenum and the fuel matrix.

The evaporation mechanism and the rate-controlling step were clarified for polycrystalline Th 0.90 U 0.10 O 2.05 and Th 0.75 U 0.25 O 2.13 solid solutions. The evaporation annealing was performed under a flowing air atmosphere at 1650°C for 32 h. The solid solutions showed incongruent evaporation in which uranium oxide preferentially vaporized. The rate of the vaporization was found to be controlled by the cation diffusion in the solid phase and enhanced by the presence of grain boundaries. In the case of the UO 2 -rich solid solution many cracks and pores were produced beneath the surface due to the enhancement.

The polished hemispheres of ThO2 kernels were irradiated by fission fragments, which recoiled from the covered Al-U alloy foils. After irradiation, the lattice parameters of the kernels were measured by microbeam X-ray diffractometer. It was found that the lattice parameter increased initially in a linear relationship with increasing fission fragment dose and reached a saturation value at doses higher than about 1×1016 fission fragments⋅cm-3. The saturation value was smaller for the fine-grained specimen than for the coarsegrained. The recovery began at about 400°C for both the specimens and lasted to higher temperature for the coarse-grained.

The chemical form of the solid fission products has been studied for (Th0.81U0.19) O2 simulating 21.5% FIMA in an HTGR environment. Experiments have been performed with X-ray diffraction, electron-probe microanalysis, ceramography and hardness measurement. The results showed that fission-product phases of two types, Mo-Ru-Pd and (Ba, Sr)(Zr, Ce) O3, are present in the simulated fuel pellet. The fuel matrix comprises (Th1-xUx, Zr, Ce, RE) O2 with an x value of 0.067(RE = Nd, La, Pr, Y, Sm). Dissolution of the rare earths (RE) and the residual Zr plus Ce in (Th0.933U0.067) O2 was accompanied by contraction of the unit cell of the oxide matrix. Reaction behavior in the selected fission product system BaO-ZrO2-Nd2O3 was also investigated. The results showed that in the presence of BaO, Nd2Zr2O7 is converted to barium zirconate: at 1630°C, Nd 2 Zr 2 O 7 + 2 BaO → 2 BaZrO 3 + Nd 2 O 3 . This fact, combined with thermochemical assessment, confirms the relative stability of (Ba, Sr)(Zr, Ce) O3 against Nd2Zr2O7 in the simulated (Th0.81U0.19) O2. From these results and fission product inventories, it is inferred that the chemical state of high-burnup ThO2 is very similar to that of (Th0.81U0.19) O2.