Irradiation devices for the study of creep and swelling in ceramic fuels

In fuel element design for advanced nuclear reactors perfect knowledge of fuel behaviour under irradiation plays a decisive role, above all for long service lives and high burnups. Therefore, the development of fast breeder fuel elements within the framework of the Karlsruhe Fast Breeder Project included various irradiation rigs which allow continuous measurement during irradiation of fuel specimen creep and swelling. A survey is presented of some of these irradiation rigs.     

Transient induced creep for irradiated fuels

Creep of nuclear fuels under irradiation has generally been considered as an athermal process. A transient state of irradiation induced creep, however, was treated with kinetic equations based on thermodynamics. For this purpose, the kinetic equations for enhanced diffusion, annihilation of excess vacancies and migration probability under irradiation were derived. Effective works were introduced to the above processes as a result of the fission damage, and reduced activation energies for each process were defined. Based on the knowledge obtained in the transient state, the effective activation energies in the steady state were discussed. The above concepts were examined using the experiments of Clough.     

Creep behavior of ceramic nuclear fuels under neutron irradiation

A theoretical estimation of the irradiation-induced creep rate of UO₂ resulted in a creep rate range between about 6 × 10⁻⁶/h and 8 × 10⁻⁵/h for a fission rate of 1 × 10¹⁴ f/cm³·s and a stress of 2 kgf/mm². It is essentially due to the “thermal rods” along the fission fragment tracks. Therefore, creep rates should only weakly depend on temperature (below 1000–1200°C) and must be markedly lower for carbide and nitride fuel.     

The plastic deformation behaviour of ceramic nuclear fuels at stress changes

The irradiation creep experiments of the BR2 Mol-12 series in capsules allowing an in-pile variation of the stress applied on the fuel specimen showed a typical accelerated deformation of mixed oxide as well as mixed carbide fuel. Quantitative relations for the accelerated deformation depending upon the irradiation parameters are derived from the measured creep data. Magnitude and time dependence of the accelerated deformation effects are discussed and compared with the steady-state creep and swelling. Accelerated fuel deformations due to stress changes can be a beneficial effect lowering the fuel/ cladding mechanical interaction during load-follow-on operation.     

Irradiation creep of oxide dispersion strengthened (ODS) steels for advanced nuclear applications

Ferritic oxide dispersion strengthened steels with different microstructure were in-beam creep tested in a temperature range from 300 to 500 °C. Irradiation was by He-ions. Elongation was determined as a function of stress and irradiation damage rate. Damage was investigated by transmission electron microscopy. A thorough analysis of the loops developing during irradiation creep did not show any dependence of orientation or size on the direction of the applied stress. At 400 °C radiation induced segregation was found (most probably an iron aluminide) which had no effect on irradiation creep. No pronounced influence of microstructure or dispersoid size on the irradiation creep behavior was detected. Irradiation creep compliance of PM2000 with dispersoids of about 30 nm diameter were found to differ little from material with dispersoids of only 2-3 nm diameter. This is in contrast to thermal creep where dislocation-obstacle interactions are extremely important. An assessment of the technical relevance of irradiation creep in advanced nuclear systems is presented.     

Irradiation creep in Zr single crystals

A single crystal of crystal bar Zr was irradiated, unstressed, at 570 K in a fast (> 1 MeV) neutron flux of $\text{5.5 × 10}{}^{16}\text{n/m}{}^2\text{-s}$. After a dose of $\text{6 × 10}{}^{23}\text{n/m}{}^2$ a tensile stress of 25 MPa was applied during a period of steady reactor power. The loading strain was an order of magnitude smaller than that observed when an identical, unirradiated, crystal was loaded to the same stress. There followed a period of primary creep during which the creep rate decreased to a value of $\text{5 × 10}{}^{\mathrm{?6}}\text{h}{}^{\mathrm{?1}}$ in the first 24 hours of the test. For the final 2000 hours of the test the specimen was observed to creep at a rate of $\text{1 × 10}{}^{\mathrm{?6}}\text{h}{}^{\mathrm{?1}}$ when the reactor was at full power. During shutdowns, the creep rate decreased with time. The results will be discussed and compared with predictions from current theories for the mechanism of irradiation enhanced creep in light of the micro-structures observed.     

Irradiation induced creep experiments in the BR2 reactor using the resonant cavity method

A 16Cr16NiNb stainless steel (steel 1.4981) will be used as wrapper material for SNR 300. Therefore, some in-pile creep tests have been performed with this material in the temperature range 420–700°C. The main objective of this programme was to see, whether the creep rates of steel 1.4981 followed at low fluences ($\text{? 2.5 × 10}{}^{22}\text{n/cm}{}^2$, $\text{E > 0.1}\text{MeV}$) the same rules as for other austenitic stainless steels. The experiments were performed in the BR2 reactor at Mol/Belgium, using creep rigs which were developed and manufactured by CEN Grenoble. The creep strains were measured by the resonant cavity method. The paper describes the main characteristics of the creep capsules, and reports on the performance of those types of rigs. Finally, the experimental results are presented and discussed.     

Irradiation creep relaxation of void swelling-driven stresses

Swelling-driven-creep test specimens are used to measure the compressive stresses that develop due to constraint of irradiation void swelling. These specimens use a previously non-irradiated 20% CW Type 316 stainless steel holder to axially restrain two Type 304 stainless steel tubular specimens that were previously irradiated in the US Experimental Breeder Reactor (EBR-II) at 490 °C. One specimen was previously irradiated to fluence levels in the void nucleation regime (9 dpa) and the other in the quasi-steady void growth regime (28 dpa). A lift-off compliance measurement technique was used post-irradiation to determine compressive stresses developed during reirradiation of the two specimen assemblies in Row 7 of EBR-II at temperatures of 547 °C and 504 °C, respectively, to additional damage levels each of about 5 dpa. Results obtained on the higher fluence swelling-driven-creep specimen show that compressive stress due to constraint of swelling retards void swelling to a degree that is consistent with active load uniaxial compression specimens that were irradiated as part of a previously reported multiaxial in-reactor creep experiment. Swelling results obtained on the lower fluence swelling-driven creep specimen show a much larger effect of compressive stress in reducing swelling, demonstrating that the larger effect of stress on swelling is on void nucleation as compared to void growth. Test results are analyzed using a recently proposed multiaxial creep-swelling model.     

Irradiation creep of stainless steel in bending

The development is described of a test to measure irradiation enhanced creep in bending of 20% cold-worked Type-316 stainless steel. The test will be irradiated in the experimental fast reactor EBR-II. The rationale used in design selection is described. The selected beam designs, the supportive tests in other stress states and the measurement techniques are described in detail.     

Irradiation creep in non-fissile metals and alloys

The principal methods used in measuring irradiation creep in non-fissile metals and alloys are described and the limitations of the techniques emphasised. The theoretical models of irradiation creep are surveyed and the experimental data on thermal and fast reactor core component materials, such as zirconium alloys and austenitic steels, are reviewed. In particular, the effects of compositional and metallurgical variables and irradiation parameters (temperature, dose and dose rate) on the magnitudes of the irradiation creep are assessed. Finally, the additional theoretical studies required to further the understanding of the phenomenon and the experimental work necessary for optimising the design and operation of thermal and fast reactors are summarised.     

Irradiation creep in deuteron bombarded stainless steel

A commercial and a high purity version of cold worked type 316 stainless steel was irradiated with 9 MeV deuterons at 300°C under tensile stresses between 100 and 350 MPa and the irradiation creep rate was measured. The results are qualitatively discussed in the light of present theoretical models.     

Irradiation creep under 60 MEV alpha irradiation

Accelerator-produced charged-particle beams have advantages over neutron irradiation for studying radiation effects in materials, the primary advantage being the ability to control precisely the experimental conditions and improve the accuracy in measuring effects of the irradiation. An apparatus has recently been built at ORNL to exploit this advantage in studying irradiation creep. These experiments employ a beam of 60 MeV alpha particles from the Oak Ridge Isochronous Cyclotron (ORIC). The experimental approach and capabilities of the apparatus are described. The damage cross section, including events associated with inelastic scattering and nuclear reactions, is estimated. The amount of helium that is introduced during the experiments through inelastic processes and through backscattering is reported. Based on the damage rate, the damage processes and the helium-to-dpa ratio, the degree to which fast reactor and fusion reactor conditions may be simulated is discussed. Recent experimental results on the irradiation creep of type 316 stainless steel are presented, and are compared to light ion results obtained elsewhere. These results include the stress and temperature dependence of the formation rate under irradiation. The results are discussed in relation to various irradiation creep mechanisms and to damage microstructure as it evolves during these experiments.     

Recent activities in the field of nuclear materials and nuclear fuels

On the research and development of nuclear materials and fuels, many of outstanding papers have been presented in scientific journals, including the Journal of Nuclear Science and Technology. Some topics have been covered in this summary from the latest activities of nuclear materials and fuels.     

Hydrogen for synthetic fuels via nuclear energy

Fluctuations in availability and recent increases in price of petroleum have had profound effects on the national economy. As synthetic fuels, in particular, hydrogen, become increasingly attractive, nuclear energy has a role in developing such fuels. It is postulated that the nuclear radiation of the fission process itself can be utilized directly in fluid fueled devices or radiation and heat can be used in special purpose solid-fuel reactors. Both fusion and fission are considered in this light.     

The porosity correction factor for the thermal conductivity of ceramic fuels

The variation of the porosity correction factor, β, with temperature and pore shape was studied by using the Fricke equation for the thermal conductivity of a two-phase medium containing the second phase as randomly distributed ellipsoids. The temperature variation occurs via a parameter, γ, related to the conductivity of the pores. The effect of pore shape was determined via the axial ratio, ε, of oblate or prolate ellipsoidal pores. The results are presented graphically in curves showing the variation of β with γ and ε.