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.     

Simulation of irradiation-induced creep in nickel

An experimental technique has been developed to simulate neutron irradiation-induced creep by charged particle bombardment. The experimental apparatus permits on-line computer monitoring of experimental parameters while temperature, stress, and flux are maintained at the desired levels. A typical result obtained with a 0.38 mm (0.015 in.) thick, high-purity nickel specimen bombarded with 22 MeV deuterons at 224°C (435°F) and at a stress of 345 MPa (50.12 ksi) is presented. The result demonstrates that charged particle irradiation can successfully be used to simulate irradiation-induced creep reproducibly in materials whose thickness is typical of nuclear fuel cladding.     

High temperature deformation characteristics of Zirlo™ tubing via ring-creep and burst tests

Fuel cladding tubing acting as a barrier between coolant and radioactive fuel pellets in light water reactors undergo a combination of mechanical and thermal effects along with corrosive conditions during normal operations as well as accident situations, such as LOCA, etc. Therefore, the mechanical integrity of the cladding tubing is of critical importance. In this study, high temperature deformation characteristics of niobium-containing zirconium alloy cladding materials (Zirlo™) have been evaluated via both ring-creep and burst tests. Creep-rupture data are presented in terms of Larson-Miller parameters (LMP). Data (creep rate vs. stress) from ring-creep and burst tests are analyzed, and operating deformation mechanisms are elucidated. This study demonstrates that the hoop creep data obtained from ring-creep and burst tests are equivalent, and one can be replaced with the other, if needed, in order to evaluate creep life.     

A model to predict the failure of zircaloy-4 fuel sheathing during postulated loca conditions

A failure model has been developed and used to predict the burst temperature and burst strain for Zircaloy-4 fuel sheathing in inert or steam atmospheres in the 900 to 1600 K range. The model assumes that the deformation of the thin-walled tube is controlled by steady-state creep and that there is a relationship between the tangential stress and the temperature at the instant of failure.For any temperature and pressure sequence, the steady-state creep rate is numerically integrated to obtain the tangential strain and stress as a function of time. The steady-state creep rate of oxidized Zircaloy is calculated using a homologous temperature concept in which it is assumed that fuel sheaths with the same homologous temperature have the same properties. Failure is predicted when the tangential stress reaches the value given by the burst-stress/burst-temperature correlation. The model also provides for a circumferential variation of temperature by approximating the thin-walled tube as a membrane. Both local and average stresses and strains can be calculated.The model has been used to assess the effect of some important variables such as anisotropy, heating rate, oxidation and circumferential temperature variation on sheath failure. Predictions using the model have been compared with data obtained in inert and steam atmospheres, and with burst strain data for a known circumferential temperature variation. The model accurately predicts the results of the experiments.     

Investigation of in-reactor creep and irradiation growth of zirconium-2.5 wt% niobium channel tubes

We summarize the diametral creep results obtained in the MR reactor of the Kurchatov Institute of Atomic Energy on zirconium-2.5 wt% niobium pressure tubes of the type used in RBMK-1000 power reactors. The experiments that lasted up to 30 000 h cover a temperature range of 270 to 350°C, neutron fluxes between 0.6 and 4.0 ×10¹³ n/cm² · s (E > 1 MeV) and stresses of up to 16 kgf/mm². Diametral strains of up to 4.8% have been measured. In-reactor creep results have been analyzed in terms of thermal and irradiation creep components assuming them to be additive. The thermal creep rate is given by a relationship of the type ε_th = A₁ exp [(A₂ + A _3σt) T] and the irradiation component by ε_rad = A_4σtø(T − A₅), where T = temperature, σ_t = hoop stress, ø = neutron flux and a₁ to A₅ are constants. Irradiation growth experiments carried out at 280° C on specimens machined from pressure tubes showed a non-linear dependence of growth strain on neutron fluence up to neutron fluences of 5 × 10²⁰ n/cm². The significance of these results to the elongation of RBMK reactor pressure tubes is discussed.     

A measurement of the irradiation-induced creep of mixed carbide nuclear fuel

In the BR 2 reacior at Mol, Belgium, a measurement of the irradiation induced creep of mixed carbide nuclear fuel up to high burnup was carried out The dependence upon applied stress and burnup of 95% dense (U, Pu) C was measured within a temperature range between 500 and 720°C and at fission rates between 1.0−1.5 × 10¹⁴ f/cm³ s. The used irradiation device was a Confluent-type capsule that allowed a variation of stress as well as temperature during irradiation. The length changes of the fuel specimen were determined by means of the microwave cavity resonance method. The obtained creep rates are proportional to stress and burnup-independent. The irradiation creep rates are about one order of magnitude below those of mixed oxide fuel. The fission product swelling rate increased with burnup form initially 1.2 to 3.0 vol% per % burnup. At stress changes the fuel showed a transient swelling up to 0.2 vol%. The theoretical background of carbide irradiation creep is briefly discussed.     

Contribution of uranium diffusion on creep behavior of uranium dicarbide

Compressive creep tests of uranium dicarbide (UC₂) have been conducted. The general equation best describing the creep rate over the temperature range 1200–1400°C and over the stress range 2000–15000 psi is represented by the sum of two exponential terms ge =A(σ/E)^0.9 exp(−39.6 ± 1.0/RT) + B(σ/E)^4.5 exp(−120.6 ± 1.7/RT), where pre-exponential factors are A(σ/E)^0.9 = 12.3/h at low stress region (3000 psi) and B(σ/E)^4.5 = 3.17 × 10¹³/h at high stress region (9000 psi), and the activation energy is given in kcal/mol. Each term of this experimental equation indicates that important processes occurring during the steady state creep are grain-boundary diffusion of the Coble model at low stress region and the Weertman dislocation climb model at high stress region. Both mechanisms are related to migration of uranium vacancies.     

Effect of microstructure on light ion irradiation creep in nickel

The concept of inhomogeneous slip or localized deformation is introduced to account for a weak dependence of irradiation creep on initial microstructure. Specimens of pure nickel (Ni) with three different microstructures were irradiated at 473 K with 15–17 MeV douterons in the Pacific Northwest Laboratory (PNL) light ion irradiation creep apparatus. A dispersed barrier model for Climb-Glide (CG) creep was unable to account for the observed creep rates and creep strains. The weak dependence on microstructure was consistent with the Stress Induced Preferential Absorption (SIPA) creep mechanism but a high stress enhanced bias had to be assumed to account for the creep rates. Also, SIPA was unable to account for the observed creep strains. The CG and SIPA modeling utilized rate theory calculations of point defect fluxes and transmission electron microscopy for sink sizes and densities.     

Material Properties of a Natural UF6 Transport Cylinder Vessel at High Temperature

Abstract

The IAEA Regulations for the Safe Transport of Radioactive Material are to be revised in 1996 and the fire test (800°C for 30 min) could become a requirement for the natural UF₆ transport cylinder. ASME SA 516 carbon steel is used as the structural material for this type of cylinder. It is very important to obtain high temperature data for SA 516 steel to be able to evaluate the integrity of the UF₆ transport cylinder vessel in the fire test. CRIEPI has therefore conducted material tests on SA 516 at high temperatures. The AC₁ and AC₃ transformation points of actual SA 516 steels have been measured. Tensile tests up to 900°C were conducted using USA, French and Japanese manufactured materials and the influence of phase transformation assessed. Preliminary creep tests show that assessment by creep strength can give a more conservative estimation than using the tensile strength. Creep deformation equations have been obtained using uniaxial creep tests and internal pressure creep tests. In addition, by the use of internal pressure creep rupture tests, the relation between the circumferential stress, the test temperature and the rupture time has been obtained.     

Effect of ion irradiation and indentation depth on the kinetics of deformation during micro-indentation of Zr-2.5%Nb pressure tube material at 25 °C

Micro-indentation creep tests were performed at 25 °C on radial-normal samples cut from Zr-2.5Nb CANDU pressure tube material in both the as-fabricated condition and after irradiation with 8.5 MeV Zr⁺ ions. The average indentation stress, and hence the yield stress, was found to increase with decreasing indentation depth and with increasing levels of ion irradiation. The activation energy of the indentation creep rate and hence the, activation energy of the obstacles that limit the rate of dislocation glide, was independent of indentation depth but increased from ΔG₀ = 0.185 to 0.215 μb³ with increasing ion irradiation damage. The magnitude of the activation energy indicates that ion irradiation introduces a new type of obstacle into the microstructure which reduces the low temperature indentation creep rate of Zr-2.5Nb pressure tubes. This is supported by TEM images showing that Zr⁺ ion irradiation produces small, nanometer size, dislocation loops which act as obstacles to dislocation glide and thus influence both the yield stress and the activation energy of the low-temperature thermal creep of Zr-2.5Nb pressure tube material. These findings suggest that neutron irradiation will have similar effect upon yield stress and low-temperature thermal creep as the Zr⁺ ion irradiation since both create similar crystallographic defects in Zr-2.5Nb pressure tubes.     

On the anomalous stress-dependence of creep rate in precipitation strengthened alloys

The stress and temperature dependence of secondary creep rate have been analysed in the temperature range of 773–823 K for a 22% Cr-34% Ni austenitic steel (alloy 800) strengthened by a small volume fraction of γ′(Ni₃Ti, Al). In this respect two regimes have been distinguished at low and high stresses with the activation energies corresponding to those of the grain boundary and bulk diffusions, respectively. The very high stress-dependence (20) observed at elevated stresses, in comparison with those at low stresses (3–5), is shown to be enhanced by the metastability of the matrix and the resultant marked deformation inequalities during secondary creep. The effect of prior cold working and ageing are discussed. A correlation between low stress and high stress data through internal friction stress estimations is sought, and the possibility of the Coble creep mechanism becoming operative at low stresses is foreseen.     

Final-stage densification of mixed oxide fuels

The final-stage densification of hypostoichiometric (U, Pu)O_2−x has been studied using a double-action punch and die, in the temperature range of 1325 to 1550°C between stresses of 7.6 and 76 MPa. At low stresses the densification rate is nearly proportional to stress and has an activation energy comparable to that measured for steady-state creep. The stress exponent increases as stress is increased. The data compare tolerably well with the predictions of theoretical pressures-intering maps once grain growth (which was found to occur at the higher temperatures) is accounted for. The controlling mechanism throughout the experiments is shown to be lattice diffusion.     

The irradiation creep of nickel and AISI 321 stainless steel during 4 MeV proton bombardment

An apparatus has been developed to study the creep of thin metal specimens under tensile stress during bombardment by 4 MeV protons from the Harwell Van de Graaff Accelerator. The specimen is held in a helium atmosphere and the proton beam reaches it through a thin metal window at the end of the accelerator beam line. The proton beam passes through the thin (25 μm) specimen, losing ~1.5 MeV in the process (most of which contributes to heating the specimen) and creating almost uniform radiation damage at the rate of $\text{(1–10) × 10}{}^{\mathrm{?7}}$ displacements per atom per second (dpa s^?1). The specimen temperature is monitored by infra-red pyrometry and controlled to ± 0.2°C by additional DC heating via the infra-red pyrometer output to compensate for ion beam fluctuations. The irradiation creep strain of the specimen is continuously measured with a sensitivity of $\text{5 × 10}{}^{\mathrm{?6}}$ by a linear variable differential transformer. Irradiation times up to about 100h with reasonable beam stability are possible. Results are presented of the irradiation creep behaviour of pure Ni and both solution treated and cold-worked AISI 321 stainless steel bombarded in the temperature range 400–600°C under tensile stresses in the range 20–250 MPa.     

Fast neutron induced creep in pyrocarbon under constant stress

Pyrocarbon is used as a coating material in the fuel of high-temperature nuclear reactors, and a thorough understanding of its irradiation behaviour includes a knowledge of its ability to creep under fast neutron irradiation. An experiment is described which demonstrates fast neutron-induced creep of a pyrolytic carbon under constant applied stress. This differs from previous work which has obtained creep ductility data from restrained shrinkage tests. The specimens were centre-loaded discs freely supported at the rim, thus subjected to a constant biaxial bend stress. On each specimen, elastic and plastic strains were produced and measured using the same geometry and loading arrangement, to allow the creep strain to be expressed simply in terms of initial elastic strain units. Results were obtained on specimens of initial density 1.95 g/cm and 1.64 g/cm³ up to a fast neutron dose of 4 × 10²⁰ n/cm² (DNE) at a temperature of 1000°C. The low-density specimens showed both the greater shrinkage and the greater creep strain, and average creep rates were 0.5 and 1.0 elastic units per 10²⁰ n/cm² (DNE) for the high and low-density specimens respectively. These constant-stress creep results are shown to be consistent with other data on pyrocarbon. They differ from graphite creep data in that the two pyrocarbons give creep strains per unit initial elastic strain which depend on their initial densities.     

Simple Method for Evaluating Mesh Size Effects in Control Rod Worth Analysis

It has been pointed out that the reactor coolant system piping could fail prior to the meltthrough of the reactor pressure vessel in a high pressure sequence of pressurized water reactor severe accidents. In order to apply to the evaluation of the piping failure which influences the subsequent accident progression, models for the strength of piping materials at high temperatures were examined. It was found that 0.2% proof stress and ultimate tensile strength above 1,073 K obtained from tensile tests was reproduced by a quadratic equation of the reciprocal absolute temperature. Short-term creep rupture time and minimum creep rate at high temperatures were well correlated by the modified Norton's Law as a function of stress and temperature, which implicitly expressed the effect of the precipitation and the resolution of precipitates on the creep strength. The modified Norton's Law gave better results than the conventional Larson-Miller method. Relating applied stress vs. minimum creep rate and tensile properties vs. applied strain rate obtained from the creep and tensile tests, a temperature range where the dynamic recrystallization significantly occurred was evaluated.     

Coupled thermal structural analysis of LWR vessel creep failure experiments

Considering the hypothetical core melt down scenario for a light water reactor (LWR) the failure mode of the reactor pressure vessel (RPV) has to be investigated to determine the loadings on the containment. The failure of reactor vessel retention (FOREVER)-experiments, currently underway, are simulating the thermal and pressure loadings on the lower head for a melt pool with internal heat sources. Due to the multi-axial creep deformation of the vessel with a non-uniform temperature field these experiments are an excellent source of data for validation of numerical creep models. Therefore, a finite element (FE) model has been developed based on a commercial multi-purpose code. Using the computational fluid dynamics (CFD) module the temperature field within the vessel wall is evaluated. The transient structural mechanical calculations are performed using a new numerical approach, which avoids the use of a single creep law employing constants derived from the data for a limited stress and temperature range. Instead of this a three-dimensional array is developed where the creep strain rate is evaluated according to the values of the actual total strain, temperature and equivalent stress. Care has to be exercised performing post-test calculations particularly in the comparisons of the measured data and the numerical results. Considering the experiment FOREVER-C2, for example, the recorded creep process appears to be tertiary, if a constant temperature field is assumed. But, small temperature increase during the creep deformation stage could also explain the observed creep behavior. Such considerations provide insight and better predictive capability for the vessel creep behavior during prototypic severe accident scenarios.     

Stress state dependence of in-reactor creep and swelling. Part 2: Experimental results

Irradiation creep constitutive equations, which were developed in Part I, are used here to analyze in-reactor creep and swelling data obtained ca. 1977-1979 as part of the US breeder reactor program. The equations were developed according to the principles of incremental continuum plasticity for the purpose of analyzing data obtained from a novel irradiation experiment that was conducted, in part, using Type 304 stainless steel that had been previously irradiated to significant levels of void swelling. Analyses of these data support an earlier observation that all stress states, whether tensile, compressive, shear or mixed, can affect both void swelling and interactions between irradiation creep and swelling. The data were obtained using a set of five unique multiaxial creep-test specimens that were designed and used for the first time in this study. The data analyses demonstrate that the constitutive equations derived in Part I provide an excellent phenomenological representation of the interactive creep and swelling phenomena. These equations provide nuclear power reactor designers and analysts with a first-of-its-kind structural analysis tool for evaluating irradiation damage-dependent distortion of complex structural components having gradients in neutron damage rate, temperature and stress state.     

Recommendations for calculating the thermal creep properties of uranium dioxide when analyzing fuel-element serviceability

Recommendations for calculating the thermal creep properties of uranium dioxide for fuel element serviceability analysis programs are developed on the basis of a physical model. The deformation processes in the model include diffusion and diffusion-controlled motion of dislocations. It is shown on the basis of the analysis of the thermodynamics of point defects in ionic crystals that the diffusion of ions is controlled by a vacancy mechanism and that the diffusion coefficient depends on the temperature and oxygen coefficient. The model includes the influence of temperature, stress, fuel density, grain size, and oxygen ratio on the creep rate. The relations obtained in the this work have made it possible to improve by approximately a factor of 10 the agreement between the calculations and experimental data as compared with the empirical relation used previously to describe the characteristics of creep.     

High temperature high dose C ion implantation in epitaxial SiGe

Si_{1 − x}Ge_x epitaxial layers fully strained (x = 0.27) and relaxed (x = 0.55) have been implanted with C ions at 500°C. Implantation energy and doses were selected to obtain the C peak in the central region of the SiGe layer, with a concentration similar to the Ge content. The implanted layers have been analyzed by Raman scattering, X-ray diffraction, transmission electron microscopy and secondary ion mass spectroscopy. The data obtained show the direct synthesis of β-SiC precipitates aligned in relation to the SiGe lattice after the implantation, as well as a Ge enrichment and stress relaxation of the SiGe lattice. For the relaxed layer a significant Ge redistribution from the implanted region is observed.