Mechanical properties of ferrous alloys after fast neutron-irradiation at high temperature

The results are given of an international “round-robin” experiment to study the nature of the damage structure in neutron irradiated zirconium and zircaloy-2 using transmission electron microscopy. The damage structure consists entirely of $\text{1}\text{3}\text{α<11}\text{2}\text{?}\text{0>}$ dislocation loops and no evidence has been found for c-component loops. Both vacancy and interstitial loops were found in specimens irradiated at 400 °C, with an excess of vacancy loops. Quantitative measurements of loop size distributions and loop concentrations are reported. All specimens exhibited “corduroy” contrast to varying degrees. The importance of choice of imaging conditions to minimize the contrast from thin foil artefacts such as oxide films and surface hydrides is stressed. The significance of the results is briefly discussed with reference to current theories of irradiation growth.     

Damage in neutron-irradiated molybdenum: (I). Characterization of as-irradiated microstructure

The as-irradiated microstructure of molybdenum, irradiated in the EBR-II reactor at six different temperatures in the range 430–1000°C (0.24–0.44 $\text{T}{}_{\mathrm{m}}$) to a fast neutron fluence of ≈ 1 × 10²² n · cm^?2 ($\text{E > 1 MeV}$), has been characterized as black spot clusters, loops, rafts, voids (random and ordered) and dislocations. Present results show that both the void density, $\text{N}$_v and the void size, $\text{d}{}_{\mathrm{v}}$, are independent of irradiation temperature in the range 430–700° C. Above the 700° C irradiation temperature the void density decreases and the void size increases exponentially with increasing irradiation temperature and they have been expressed empirically as $\text{N}{}_{\mathrm{<mtext>v</mtext>}}\text{= 3.6 × 10}{}^{20}\text{exp}\text{(?26.9 T/T}{}_{\mathrm{<mtext>m</mtext>}}\text{)}$, $\text{d}{}_{\mathrm{<mtext>v</mtext>}}\text{= 1.5}\text{exp}\text{(9.44 T/T}{}_{\mathrm{<mtext>m</mtext>}}\text{)}$, where $\text{T/T}{}_{\mathrm{<mtext>m</mtext>}}$ is the irradiation temperature presented as a fraction of the melting point. The void density of all available published data has been used to show that the void density is (a) a strong function of irradiation temperature for a constant number of displacements per atom (dpa) and (b) a function of reactor power and spectrum when normalized to dpa.     

Magnetic properties of a highly neutron-irradiated nuclear reactor pressure vessel steel

We report results of minor B–H loop measurements on a highly neutron-irradiated A533B-type reactor pressure vessel steel. A minor-loop coefficient, which is a sensitive indicator of internal stress, changes with neutron fluence, but depends on relative orientation to the rolling direction in the low fluence regime. At a higher fluence of ～10 × 10²³ m^?2, on the other hand, an anomalous increase of the coefficient was detected irrespective of the orientation. The results were interpreted as due to competing irradiation mechanisms of the formation of Cu-rich precipitates, recovery process, and the formation of late-blooming Mn–Ni–Si-rich clusters.     

Annealing study of gold,zinc and germanium alloyed aluminum neutron-irradiated at liquid nitrogen temperature

A study of the recovery of the electrical resistivity of the dilute aluminum alloys Al-Au (0.02–0.25 at % Au), Al-Zn (0.05–1.0 at % Zn) and Al-Ge (0.02–0.5 at % Ge) after neutron-irradiation at a total dose of 9 × 10¹⁸n/cm²at 78 K was undertaken. The annealing spectra of Al-Au and Al-Zn alloys showed two distinct substages in stage II, while in Al-Ge alloys a five-peak structure was apparent. At the low-temperature side of the main peak of stage III a shoulder is definitely apparent in the three alloy systems. After stage III, Al-Ge alloys appear to have two doublets and a reverse recovery resistivity, which can be attributed to the formation of Guinier-Preston zones. The break-up of interstitials from impurity atoms or their clusters and the trapping of vacancy-type defects during their migration in stage III, are the mechanisms which can explain stages II and III respectively. The activation energy for the main peak of stage III was in all cases found to be 0.60 eV, and the peak obeys second-order kinetics.     

Mechanical properties and microstructure in neutron-irradiated nickel-based alloys and stainless steels for supercritical water-cooled-reactor fuel cladding

ABSTRACT

To investigate the irradiation behavior of mechanical properties and microstructural changes of commercial Ni-based alloys and improved stainless steels, a neutron-irradiation experiment was performed at the Joyo reactor, and post-irradiation examinations with tensile tests and TEM observations were carried out. The room-temperature tensile tests showed that all specimens that were irradiated at 485°C exhibited significant hardening and ductile behavior, especially in alloy 625. The irradiation hardening of all specimens irradiated at 668°C was less than that of specimens irradiated at 485°C. The fine-grained stainless steel, T3 and the Zr-added stainless steels, H1 and H2 showed good mechanical-property performance with keeping ductility after neutron irradiation. Most alloys and steels showed ductile behavior on the fracture surface except for alloy 625 specimen. The TEM observations showed that a high density of tangled dislocations and irradiation-induced defect clusters formed in the stainless steels and Ni-based alloys irradiated at 485°C. At 668°C, the material microstructures coarsened and their dislocation density decreased significantly. Long rod-like precipitates of Zr(Cr, Fe) compounds formed in the H1 and H2 steels that were modified with Zr. The yield stress drop of T3 steel in tensile stress was observed and is caused by grain-size coarsening at an irradiation of 668°C.     

Tensile and fracture toughness properties of neutron-irradiated CuCrZr

Tensile and fracture toughness properties of a precipitation-hardened CuCrZr alloy were investigated in two heat treatment conditions: solutionized, water quenched and aged (CuCrZr SAA), and hot isostatic pressed, solutionized, slow-cooled and aged (CuCrZr SCA). The second heat treatment simulated the manufacturing cycle for large components, and is directly relevant for the ITER divertor components. Specimens were neutron irradiated at ∼80 °C to two fluences, 2 × 10²⁴ and 2 × 10²⁵ n/m² (E > 0.1 MeV), corresponding to displacement doses of 0.15 and 1.5 displacements per atom (dpa). Tensile and fracture toughness tests were carried out at room temperature. Significant irradiation hardening and plastic instability at yield occurred in both heat treatment conditions with a saturation dose of ∼0.1 dpa. Neutron irradiation slightly reduced fracture toughness in CuCrZr SAA and CuCrZr SCA. The fracture toughness of CuCrZr remained high up to 1.5 dpa (J_Q > 200 kJ/m²) for both heat treatment conditions.     

Mechanical design methods for high temperature reactor components

The work concerned with design codes for high temperature reactor (HTR) components operating at temperatures above 800°C is summarized. Using the experimental results from the German HTR materials development programmes, in particular the time dependent properties, the structural design analysis for an intermediate heat exchanger is discussed, with reference to creep, fatigue, creep buckling and creep ratchetting. The analysis provides the basis for a critical consideration of ASME Code, Case N 47, and the applicability of the code case rules for service temperatures above 800°C.     

Low temperature neutron irradiation effects on microstructure and tensile properties of molybdenum

Polycrystalline molybdenum was irradiated in the hydraulic tube facility at the High Flux Isotope Reactor to doses ranging from 7.2 × 10⁻⁵ to 0.28 dpa at 80 °C. As-irradiated microstructure was characterized by room-temperature electrical resistivity measurements, transmission electron microscopy (TEM) and positron annihilation spectroscopy (PAS). Tensile tests were carried out between −50 and 100 °C over the strain rate range 1 × 10⁻⁵ to 1 × 10⁻² s⁻¹. Fractography was performed by scanning electron microscopy (SEM), and the deformation microstructure was examined by TEM after tensile testing. Irradiation-induced defects became visible by TEM at 0.001 dpa. Both their density and mean size increased with increasing dose. Submicroscopic three-dimensional cavities were detected by PAS even at 0.0001 dpa. The cavity density increased with increasing dose, while their mean size and size distribution was relatively insensitive to neutron dose. It is suggested that the formation of visible dislocation loops was predominantly a nucleation and growth process, while in-cascade vacancy clustering may be significant in Mo. Neutron irradiation reduced the temperature and strain rate dependence of the yield stress, leading to radiation softening in Mo at lower doses. Irradiation had practically no influence on the magnitude and the temperature and strain rate dependence of the plastic instability stress.     

Effects of exposure to high-temperature helium containing oxygen on the mechanical properties of molybdenum and TZM-Mo alloy at room temperature

The effects of exposure to helium containing oxygen of 0.1–115 vpm at 1000°C on the mechanical properties of molybdenum and TZM-Mo alloy at room temperature were studied. The stress-relieved molybdenum specimen which was not recrystallized at test temperature showed the ductility after exposure to helium containing oxygen. The recrystallized molybdenum and TZM lost ductility after exposure to helium containing oxygen of 0.1–13 vpm in a few hours. The embrittlement of molybdenum was considered to be due to the grain boundary weakening. Molybdenum to which carbon was added seemed to hinder the grain boundary weakening by the oxygen contamination. Both stress-relieved and recrystallized TZM specimens picked up oxygen linearly with time of exposure to helium. The increase in oxygen content of TZM, which was considered to be caused by the internal oxidation of titanium and zirconium, results in the embrittlement of TZM.     

Defect assessment procedures at high temperature

Within Nuclear Electric PLC, a comprehensive assessment procedure for the high-temperature response of structures is being produced. The procedure is referred to as R5 and is written as a series of step-by-step instructions in a number of volumes. This paper considers in detail those parts of R5 which address the behaviour of defects. The defect assessment procedures may be applied to defects found in service, postulated defects, or defects formed during operation as a result of creep-fatigue loading. In the last case, a method is described for deducing from endurance data the number of cycles to initiate a crack of a specified size. Under steady loading, the creep crack tip parameter C* is used to assess crack growth. Under cyclic loading, the creep crack growth during dwell periods is still governed by C* but crack growth due to cyclic excursions must also be included. This cyclic crack growth is described by an effective stress intensity factor range. A feature of the R5 defect assessment procedures is that they are based on simplified methods and approximate reference stress methods are described which enable C* in a component to be evaluated. It is shown by comparison with theoretical calculations and experimental data that reliable estimates of C* and the associated crack growth are obtained provided realistic creep strain rate data are used in the reference stress approximation.     

中子福照硅单晶正电子捕获的温度效应

用正电子湮没寿命测量方法研究了在中子辐照硅单晶缺陷中正电子捕获的温度效应。寿命谱被成功地分成两个组分,长寿命组分在109—300K的温度范围有325±9 ps的常数寿命值,它被认为是在负荷电双空位捕获的正电子寿命。正电子捕获率显示一个强的负温度效应.这个捕获特征能用正电子在负荷电双空位的级联捕获来描述。在低温区正电子捕获截面以T~(-1.7)随温度变化。     

Fast flux irradiation tests performed at high temperature

The first gas-cooled fast breeder reactor (GCFR) fast flux irradiation experiment [F-1(X094)] consists of seven fuel rods clad in 20% cold-worked 316 stainless steel. The rods are individually encapsuled, with sodium filling the gaps within the capsule walls. The rods are fueled with (15% Pu, 85% U)O₂ and have depleted UO₂ lower and upper axial blankets and charcoal to trap volatile fission products. The cladding i.d. temperature range covered by these rods is 570–760°C (1055–1400°F).The in-reactor performance of the fuel rods in the F-1 high-temperature experiment, which achieved a burnup of 121 MWd/kg (13.0 at.%) on the lead rod, is described. All rods in the experiment have remained intact. The results of interim examinations [at 25 and 50 MWd/kg (2.7 and 5.4 at.%)] of fuel and fission product behavior and transport and comparisons of observed results with LIFE-III code predictions are described.The F-3 experiment, which consists of ten encapsulated GCFR fuel rods with surface-roughened (ribbed) cladding, shares a nineteen capsule subassembly with Argonne National Laboratory. Temperatures are controlled over the range 675°C (1250°F) to 750°C (1380°F). Irradiation is in the core region of the EBR-II and thus permits achievement of a higher fluence-to-burnup ratio than that obtained in the F-1 experiment.Preliminary results of a planned interim examination at an exposure of 46 MWd/kg (4.9 at.%) burnup and a fluence of 5.2 × 10²² n/cm² show that cladding failures occurred in nine of the ten rods. Preliminary indications are that the failures are due to defects in the sodium bond between the fuel rod and the capsule.The tests completed and currently under way have been scoping in nature, and irradiation in EBR-II of GCFR prototypical fuel (pressure equalized) rods with ribbed cladding is required to provide the information needed for reactor design on effects of exposure to high fluence and burnup and on design reliability for a statistically significant number of rods. The design and the operating conditions for the F-5 experiment being prepared for this purpose are described.     

Oxide layer stability in lead-bismuth at high temperature

Materials protection by ‘in situ’ oxidation has been studied in stagnant lead-bismuth, with different oxygen levels (H₂/H₂O ratios of 0.3 and 0.03), at temperatures from 535 °C to 600 °C and times from 100 to 3000 h. The materials tested were the martensitic steels F82Hmod, EM10 and T91 and the austenitic stainless steels, AISI 316L and AISI 304L. The results obtained point to the existence of an apparent threshold temperature above which corrosion occurs and the formation of a protective and stable oxide layer is not possible. This threshold temperature depends on material composition, oxygen concentration in the liquid lead-bismuth and time. The threshold temperature is higher for the austenitic steels, especially for the AISI 304L, and it increases with the oxygen concentration in the lead-bismuth. The oxide layer formed disappear with time and, after 3000 h all the materials, except AISI 304L, suffer corrosion, more severe for the martensitic steels and at the highest temperature tested.     

Gamma-rays irradiation effects on polysulfone at high temperature

The temperature dependence of the irradiation effects on polysulfone was studies by measuring the molecular weight, glass transition temperature, gel fraction and evolved gas. Polysulfone was irradiated with gamma-rays at room temperature, 100, 150, 180 and 210 °C. The change of molecular weight distribution and glass transition temperature showed occurrences of a main chain scission at room temperature and cross-linking at high temperature. The decrease of gel dose, the increases of gel fraction and total gas evolution with increasing temperature was observed. The evolution of CO, CO₂ and SO₂ gases increased at high temperature, while yield of evolved H₂ was independent of irradiation temperature. The probability of the cross-linking was clearly increased by irradiation at high temperature above 180 °C, though the chain scission was not changed very much.     

The high temperature gas-cooled reactor as a source of high temperature process heat

The nuclear reactor has established itself as a future major supplier of electrical energy. The industrial market for other forms of energy, however, is almost as large and represents a new potential for the use of nuclear reactors. The high temperature gas-cooled reactor (HTGR) has been developed for commercial application in the electric power generation field. Since the HTGR is capable of delivering process heat in the temperature range of 1000–1500°F, it has many applications that would not be possible at the lower operating temperatures of water-cooled reactors. This paper briefly summarizes the development of the HTGR and outlines its salient technical features. Modifications to the reactor that enable it to be used as a process heat source are discussed. Specific applications are developed for coal gasification, steelmaking, and hydrogen production.     

Low temperature fatigue crack propagation in neutron-irradiated type 316 steel and weld metal

The fast cycling fatigue crack propagation characteristics of type 316 steel and weld metal have been investigated at 380°C after irradiation to 1.72–1.92 × 10²⁰n/cm² (E>1 MeV) and 2.03×10²¹n/cm² (E>1 MeV) at the same temperature. With mill-annealed type 316 steel, modest decreases in the rates of crack propagation were observed for both dose levels considered, whereas for cold-worked type 316 steel irradiation to 2.03 ×10²¹ n/cm² (E>1 MeV) caused increases in the rate of crack propagation. For type 316 weld metal, increases in the rate of crack propagation were observed for both dose levels considered.The diverse influences of irradiation upon fatigue crack propagation in these materials are explained by considering a simple continuum mechanics model of crack propagation, together with the results of control tensile experiments made on similarly irradiated materials.     

Low temperature fatigue crack propagation in neutron-irradiated type 316 steel and weld metal

The fast cycling fatigue crack propagation characteristics of type 316 steel and weld metal have been investigated at 380°C after irradiation to 1.72?1.92 × 10²⁰ n/cm²($\text{E > 1}$ MeV) and 2.03 × 10²¹ n/cm² ($\text{E > 1}$ MeV)at the same temperature. With mill-annealed type 316 steel, modest decreases in the rates of crack propagation were observed for both dose levels considered, whereas for cold-worked type 316 steel irradiation to 2.03 × 10²¹ n/cm² ($\text{E > 1}$ MeV) caused increases in the rate of crack propagation. For type 316 weld metal, increases in the rate of crack propagation were observed for both dose levels considered.The diverse influences of irradiation upon fatigue crack propagation in these materials are explained by considering a simple continuum mechanics model of crack propagation, together with the results of control tensile experiments made on similarly irradiated materials.