In-reactor creep of zirconium alloys by thermal spikes

The size and duration of thermal spikes from fast neutrons have been calculated for zirconium alloys, showing that spikes up to 1.8 nm radius may exist for $\text{2 × 10}{}^{\mathrm{?11}}$s at greater than melting point, at 570 K ambient temperature. Creep rates have been calculated assuming that the elastic strain from the applied stress relaxes in the volume of the spikes (by preferential loop alignment or modification of an existing dislocation network). The calculated rates are consistent with strain rates observed in long term tests in-reactor, if spike lifetimes are 2 to 2.5 × 10^?11s.     

In-reactor study of the friction characteristics of reactor materials

Conclusions An apparatus was developed for measuring the frictional characteristics and was tested during in-reactor experiments. Apparently, the significant increase in the coefficient of static friction observed during the in-reactor irradiation process must be considered when evaluating the stress (deformed) state of the fuel elements during power fluctuations (jumps), i. e., in case of power increase after relatively long-term operation at a low power.Translated from Atomnaya Énergiya, Vol. 61, No. 3, pp. 175–178, September, 1986.     

In-reactor tubular creep of Zircaloy-2 at 260 to 300 °C

Tubular specimens of Zircaloy-2, 23 mm diameter, have been creep tested in-reactor at 260 to 300°C (530 to 570 K). The specimens were biaxially stressed by internal pressure, with transverse stresses from 100 to 300 MN/m². Zircaloy-2 was tested in three conditions; 20% cold drawn, 70% tube reduced then stress-relieved and annealed.All creep curves, both in and out of neutron fluxes, can be represented by straight lines on log strain-log time curves. Fast neutron flux increased the slopes of the log-log creep curves of the cold-worked materials. These slopes increased from 0.24–0.27 for unirradiated specimens (and specimens in the thermal neutron flux) to 0.42–0.47 for specimens in a fast neutron flux. This means that creep rate does not diminish with time as rapidly in-reactor as out-reactor. The creep behaviour of the annealed Zircaloy-2 was little affected by fast neutron flux.     

Crack growth initiation in concrete-like materials in the presence of creep

A numerical procedure that employs the finite element method and the state variables approach is proposed to analyse the critical condition of a crack in an aging viscoelastic body under sustained load. A far field solution proposed by Schapery is used. An example shows how a crack can become critical after some finite time that depends on the characteristics of the concrete.     

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.     

In-reactor creep of zircaloy-2, zircaloy-4 and Zr-1.15 wt % Cr-0.1 wt % Fe at 568 K derived from their stress-relaxation behaviour

The in-reactor stress relaxation of zirconium alloys can be represented by the equation $\text{σ/σ}{}_0\text{= D exp (?Kt)}$ where the unrelaxed stress ratio is equal to a factor times the exponential of the relaxation rate. Creep rates derived from the bent-beam stress-relaxation data are compared with uniaxial creep results and a good correlation is shown for Zircaloy-2 through a continuous trend curve from low-stress stress relaxation to the high-stress creep behaviour. Specimens quenched from the β-phase had the lowest creep rates compared with those which were annealed or cold-worked, while increasing cold-work increased the creep rates for most specimens. The effects of cold-work and results from post-irradiation annealing of stress-relaxation test specimens suggest that for the low stress régime ($\text{σ <}\text{1}\text{3}\text{σ}$ yield) the operation of a dislocation climb/glide mechanism in addition to the stress-directed loop formation mechanism is required for in-reactor creep.     

Development of high temperature structural materials for the HTGR

In accordance with the HTGR program in Japan, a series of R&D for high temperature structural materials in particular with respect to the HTTR design code has been performed in JAERI for more than 20 years. This paper introduces R&D results of the pressure retaining low alloy steel 2 1/4Cr-1Mo and the high temperature structural alloys Hastelloy XR and Ni-Cr-W superalloy for the design code together with some fruits of recent studies.     

Development of the Climb Induced Glide concept to describe in-reactor creep of fcc materials

The Climb Induced Glide model (CIG) for irradiation creep is developed using a plastic flow law which has been successfully applied in the correlation of Type 316 stainless steel rupture data. This model is used to predict the stress and temperature dependence of irradiation creep and the transition from irradiation to thermal creep. The predictions of this model are compared and found to be qualitatively consistent with experimental data and microstructural information. This model allows prediction of deformation behavior covering strain rates from 1 × 10⁻¹³ s⁻¹ to 1 s⁻¹.     

The first materials science investigations of plutonium in the laboratories of NII-9

The formation of the institute NII-9 (now the State Science Center of the Russian Federation—A. A. Bochvar All-Russia Scientific-Research Institute of Standardization in Machine Engineering), by a decree of the State Committee on Defense, in 1945 in Moscow and the production at this Institute in 1948 of substantial (milligram) quantities of metallic plutonium from uranium blocks irradiated in the F-1 reactor are recounted. Data on the organization of the work and the creation of special equipment and methods for performing investigations on microsamples of a radioactive, toxic, chemically active material are presented. These investigations led to the discovery of six allotropic modifications of plutonium the temperature ranges of their stability, the melting temperature, the characteristic features of the volume and temperature expansion at phase transitions, and the mechanical properties of the phases. The participation of the Institute staff and of A. A. Bochvar himself in the fabrication of the plutonium warhead in 1949 for the first Soviet atomic bomb is reflected. The directions of the investigations of plutonium performed in subsequent years are indicated. State Science Center of the Russian Federation—A. A. Bochvar All-Russia Scientific-Research Institute of Standardization in Machine Engineering. Translated from Atomnaya énergiya, Vol. 87, No. 3, pp. 170–174, September, 1999.     

In-reactor stress relaxation of selected metals and alloys at low temperatures

Sttess relaxation of bent beam specimens under fast neutron irradiation at 340 and 570 K has been studied for a range of materials, as follows: several stainless steels, a maraged steel, AISI4140, Ni, Inconel X-750, Ti, Zircaloy-2, Zr-2.5% Nb and Zr₃ Al. All specimens were in the annealed or solution-treated condition. Where comparisons were possible, the creep coefficients derived from the stress relaxation tests were found to be consistent with other studies of irradiation-induced creep. The steels showed the lowest rates of stress relaxation; the largest rates were observed with Zr-Nb, Ti and Ni. For most materials, the creep coefficient at 340 K was equal to or greater than that at 570 K. Such weak temperature dependence is not easily reconciled with existing models of irradiation creep based on dislocation climb, such as SIPA or climb-induced glide. Rate theory calculations indicate that because the vacancy mobility becomes very low at the lower temperature, recombination should dominate point defect annealing, resulting in a very low creep rate compared to that at the higher temperature. It is shown that the weak temperature dependence observed experimentally cannot be accounted for by the inclusion of more mobile divacancies in the calculation.     

Research and development at the structural materials information center

The U.S. Nuclear Regulatory Commission has initiated a Structural Aging Program at the Oak Ridge National Laboratory to identify potential structural safety issues related to continued service of nuclear power plants and to establish criteria for evaluating and resolving these issues. One of the tasks in this program focuses on the establishment of a Structural Materials Information Center where long-term and environment-dependent properties of concretes and other structural materials are being collected and assembled into a data base. These properties will be used to evaluate the current condition of critical structural components in nuclear power plants and to estimate their future performance during the continued service period.     

The role of dispersed barriers in the pulsed irradiation creep of magnetic fusion reactor materials

The effects of irradiation pulsing on the climb-glide creep, and the role of the glide barrier height are investigated. Only tokamak-type pulsing is considered. We develop a new formulation for the creep strain increment per pulse for large barriers, for which more than one pulse is needed for glide to occur. This formulation is applied to typical tokamak-type conditions, including the UWMAK-I and INTOR designs. It is concluded that no significant enhancement over steady irradiation occurs for τ_v ? burn-time. However, in long burn-time Tokamaks with τ_v?on-time and off-time, it is found that the pulsed creep enhancement can be significant. For example, for a duty factor of 0.9 the enhancement is about 3 for small barriers using a dose-equivalent average damage rate when comparing pulsed and steady irradiation. The maximum enhancements are diminished to about 2 when equal instantaneous damage rates are used.