Chemical aspects of iodine-induced stress corrosion cracking failure of zircaloy-4 tubing above 50°C

The chemical environment associated with iodine-induced SCC failure of Zircaloy-4 tubing above 500°C has been characterized. At the critical iodine concentrations which result in SCC initiation and propagation, most of the iodine is present as condensed zirconium subiodides (I/Zr ? 0.4). Only a small part of the iodine remains in the gas phase as ZrI₄. The gaseous ZrI₄ is probably responsible for crack initiation and propagation. The critical ZrI₄ pressures for SCC failure have been estimated in zircaloy/iodine reaction experiments performed with unstressed zircaloy tube specimens. These pressures were confirmed in additional creep rupture tests conducted under controlled ZrI₄ partial pressure conditions. The estimated critical ZrI₄ pressure above which low-ductility SCC failure of the zircaloy tubing always occurs, independent of time-to-failure, varies between 0.005 bar at 550°C and 0.043 bar at 800°C. Below the critical values, however, a rather wide range of ZrI₄ pressures is associated with the onset of the SCC, especially at temperatures below 800°C. A comparison of the experimental results with available thermochemical data in the Zr-I system indicates that the main reaction involved during crack propagation is chemisorption of iodine-containing species on the fresh zircaloy surfaces created by metal straining at the crack tip.     

The iodine-induced strain rate sensitivity of Zircaloy fuel rod cladding

The strain rate sensitivities and failure times characteristic of iodine-induced stress corrosion cracking of Zircaloy fuel rod cladding are important because they can be related to certain power ramping rates which may increase fuel rod failure probabilities. The CCSCC model developed in this paper approximates these failure characteristics by simulating the transition from the slower (less observable) non-corrosive creep cracking (CC) regime to the faster (more observable) stress corrosion cracking (SCC) regime. Components of the CCSCC model include: ZrI₄ production by chemical reaction between Zircaloy and iodine; diffusion of the ZrI₄ to the crack tip; chemisorption, embrittlement, and damage accumulation at the crack tip; and crack initiation times and growth rates.Results indicate that the Zr-I₂ SCC process is dominated by competition between chemical reaction and material creep rate phenomena, rather than by stress thresholds or by the requirement that a complete monolayer of ZrI₄ form on the exposed Zircaloy surface at the crack tip. Failure times are dominated by the time required to initiate active crack growth. The SCC process is apparently not limited by diffusion kinetics on the time scale of laboratory experiments. Conflicting results were found concerning the chemical reaction rate at the crack tip.     

The effect of a single tensile overload on stress corrosion cracking growth of stainless steel in a light water reactor environment

It has been found that a single tensile overload applied during constant load amplitude might cause crack growth rate retardation in various crack propagating experiments which include fatigue test and stress corrosion cracking (SCC) test. To understand the affecting mechanism of a single tensile overload on SCC growth rate of stainless steel or nickel base alloy in light water reactor environment, based on elastic-plastic finite element method (EPFEM), the residual plastic strain in both tips of stationary and growing crack of contoured double cantilever beam (CDCB) specimen was simulated and analyzed in this study. The results of this investigation demonstrate that a residual plastic strain in the region immediately ahead of the crack tips will be produced when a single tensile overload is applied, and the residual plastic strain will decrease the plastic strain rate level in the growing crack tip, which will causes crack growth rate retardation in the tip of SCC.     

Evaluation of SCC crack behavior in zirconium and zircaloy-2 using nonlinear fracture mechanics parameters

Applicability of nonlinear fracture mechanics parameters, i.e. J-integral, crack tip opening displacement (CTOD), and crack tip opening angle (CTOA), to evaluation of stress corrosion crack (SCC) propagation rate was investigated using fully annealed zirconium plates and Zircaloy-2 tubing, both of which produce SCC with comparatively large plastic strain in an iodine environment at high temperatures.Tensile SCC tests were carried out at 300°C for center-notched zirconium plates and internal gas pressurization SCC tests at 350°C, for Zircaloy-2 tubing, to measure the SCC crack propagation rate. The J-integral around semi-elliptical SCC cracks produced in Zircaloy-2 tubing was calculated by a three-dimensional finite element method (FEM) code.The test results revealed that the SCC crack propagation rate dc/dt could be expressed as a function of the J-integral, which is the most frequently used parameter in nonlinear fracture mechanics, by the equation dc/dt = C · Jⁿ, where C and n were experimental constants.Among the other parameters, CTOD and CTOA, the latter appeared to be useful for assessing the crack propagation rate, because it had a tendency to hold a constant value at various crack depths.     

An investigation of the degradation characteristics for casting stainless steel, CF8M, under high temperatures

Stainless steel castings are used in pipes and valves subjected to high pressure and temperatures. The primary coolant system of a nuclear power plant is made of a stainless steel casting and the operating temperatures are in the range of 290–330°C. If the coolant system is exposed to these temperature ranges for a long period, it may be possible to experience degradation of the material. The present investigation is concerned with the degradation characteristics of CF8M (cast duplex stainless steel), exposed to the thermal and σ-phase degradation temperatures, 430 and 700°C, respectively. After the CF8M specimens are held 100–3600 h at 430°C for the thermally degraded specimens and maintained 20 min to 150 h at 700°C for the σ-phase degraded specimens, respectively, all specimens are water quenched. Each specimen of the thermally and σ-phase degraded materials is classified into five classes depending on the holding time at the given temperatures. In order to investigate the characteristics of degradation, microstructure, micro Vickers hardness, tensile, impact tests, and fatigue crack growth tests are performed for each class of the specimens. From the present investigation the following results were obtained: (1) the difference between the thermally and σ-phase degraded specimens can be distinguished through their microstructures, (2) hardness and tensile strength are increased with degradation, while elongation, reduction area, and impact energy are decreased by increasing the degradation, (3) the fatigue crack growth rate (FCG) of the σ-phase degradation at 700°C is larger than that of the thermally degraded specimens, and (4) the FCG for both thermally and σ-phase degraded specimens are larger than those of the virgin (nondegraded) specimens.     

Stress corrosion cracking of Zircaloys in unirradiated and irradiated CsI

Unirradiated split-ring specimens of Zircaloy fuel cladding, coated with CsI, cracked when stressed at elevated temperatures. The specimens have been reexamined fractographically and metallographically in order to confirm that the cause of cracking was stress corrosion (SCC) and not delayed hydride cracking (DHC). Further specimens have been cracked at 350°C by a solution of CsI in a fused mixture of nitrates of rubidium, cesium, strontium and barium, by a similar mechanism. CsI dissolved in a fused molybdate melt was not stable at 400°C, and rapidly evolved iodine, leaving a melt that was incapable of causing SCC. Irradiation of stressed split-ring specimens of Zircaloy fuel cladding in a γ-irradiator of 10⁶ R/h and in the U-5 loop in the NRU reactor at an estimated 10⁹ R/h caused SCC when the specimens were packed in dry CsI powder. Care had to be taken to dry the CsI, otherwise cracking occurred by a DHC mechanism from hydrogen absorbed from residual moisture in the CsI. Fractography showed that the crack surfaces obtained with dry CsI were typical of iodine-induced SCC rather than cesium-induced metal vapour embrittlement. Thus, if a transport process is provided for the iodide to obtain access to the zirconium surface, CsI is capable of causing SCC of Zircaloy. This transport process might be ionic diffusion in a fission product oxide melt in the fuel-clad gap, however, radiolysis of CsI to form a volatile iodine species in a radiation field is the more probable explanation of PCI failures.     

Crack growth in the through-thickness direction of hydrided thin-wall Zircaloy sheet

In a reactivity-initiated accident, cladding failure may occur by crack initiation within a defect such as a hydride rim or blister and subsequent crack propagation through the thickness of the thin-wall cladding. In such a circumstance, determining the cladding resistance to crack propagation in the through-thickness direction is crucial to predicting cladding failure. To address this issue, through-thickness crack propagation in hydrided Zircaloy-4 sheet was analyzed at 25 °C, 300 °C, and 375 °C. At 25 °C, the fracture toughness decreased with increasing hydrogen content and with an increasing fraction of radial hydrides. Hydride particles fractured ahead of the crack tip, creating a path for crack growth. At both 300 °C and 375 °C, the resistance to crack-growth initiation was sufficiently high that crack extension was often caused by crack-tip blunting. There was no evidence of hydride particles fracturing near the crack tip, and no significant effect of hydrogen content on fracture toughness was observed at these elevated temperatures.     

Delayed hydride cracking of spent fuel rods in dry storage

Failures of zirconium alloy cladding tubes during a long-term storage at room temperature were first reported by Simpson and Ells in 1974, which remains unresolved by the old delayed hydride cracking (DHC) models. Using our new DHC model, we examined failures of cladding tubes after their storage at room temperature. Stress-induced hydride phase transformation from γ to δ at a crack tip creates a difference in hydrogen concentration between the bulk region and the crack tip due to a higher hydrogen solubility of the γ-hydride, which is a driving force for DHC at low temperatures. Accounting for our new DHC model and the failures of zirconium alloy cladding tubes during long-term storage at room temperature, we suggest that the spent fuel rods to be stored either in an isothermal condition or in a slow cooling condition would fail by DHC during their dry storage upon cooling to below 180 °C. Further works are recommended to establish DHC failure criterion for the spent fuel rods that are being stored in dry storage.     

PWSCC of thermally treated alloy 600 pulled from a Korean plant

Three tubes of alloy 600 were pulled out from a Korean nuclear power plant. The microstructure was analyzed using an optical microscope and TEM. Information on the crack length and depth was obtained by metallography, and crack detection and evolution were evaluated by analyzing the eddy current data obtained from each in-service-inspection (ISI). The carbon content in the pulled tubes was higher (around 0.03 wt.%) than that (around 0.015 wt.%) of Alloy 600 tubing used in other operating nuclear power plants. Most carbides in the pulled tubing were distributed in the grains rather than along the grain boundaries. The poor microstructure might come from high carbon contents, low temperature annealing, or high residual stresses during tube straightening. Mill annealing temperature should be high enough to dissolve all carbon in order to decorate the grain boundaries with semi-continuous carbide precipitation during 700 °C thermal treatment. Shot peening seemed to suppress the growth of the axial cracks, while it was analyzed to play a role in increasing crack growth in the wall thickness direction.     

Elastic anisotropy of zirconium alloy fuel cladding

Some methods to determine the anisotropic elasticity coefficients of zirconium alloy fuel cladding are discussed together with the conventional elastic constants.A simplified method, which uses the f parameters, was proposed, and the validity and applicability of the method were also investigated. The integration method, which was originally proposed by Rosenbaum et al., was found to be in excellent agreement with the experimental values of our own twisting test from room temperature to 800°C. The proposed f parameter method was also found to agree well with the values obtained by the integration method or experiment, especially at high temperatures near 700°C. It became evident that the elastic property of the typical fuel cladding was roughly isotropic at room temperature, and that the elastic anisotropy monotonically increased with temperature. Some stress or strain distributions of the fuel cladding were also obtained using anisotropic elasticity constants. The stress induced in the fuel cladding with simulated ridge deformation was very little affected by the difference in texture, but was more influenced by the elastic constants employed.     

Thermo-mechanical analysis of supercritical pressure light water-cooled fast reactor fuel rod by FEMAXI-6 code

Thermo-mechanical behaviors of supercritical pressure light water cooled fast reactor (SWFR) fuel rod and cladding have been investigated by FEMAXI-6 (Ver.1) code with high enriched MOX fuel at elevated operating condition of high coolant system pressure (25 MPa) and high temperature (500 °C in core average outlet temperature). Fuel rod failure modes and associated fuel rod design criteria that is expected to be limiting in SWFR operating condition have been investigated in this fuel rod design study. Fuel centerline temperature is evaluated to be 1853 °C and fission gas release fraction is about 45% including helium production. Cumulative damage fraction is evaluated by linear life fraction rule with time-to-rupture correlation of advanced austenitic stainless steel. In a viewpoint of mechanical strength of fuel cladding against creep rupture and cladding collapse at high operation temperature, currently available stainless steels or being developed has a potential for application to SWFR. Admissible design range in terms of initial gas plenum pressure and its volume ratio are suggested for fuel rod design The stress ranges suggested by this study could be used as a preliminary target value of cladding material development for SWFR application.     

Subchannel analysis of supercritical light water-cooled fast reactor assembly

Subchannel analyses have been carried out for supercritical water-cooled fast reactor fuel assembly. Peak cladding surface temperature difference arising from subchannel heterogeneities have been calculated by using the improved subchannel analysis code STARS and was evaluated to be about 18.5 °C. Several suggestions have been also made for reducing the PCST difference arising from channel heterogeneity. Influences of local power peaking on deflection of cladding surface temperature are explained with pin power distribution taken from core depletion calculation in this paper. Maximum cladding surface temperature at nominal condition is evaluated to be 645.3 °C over the cycle. Statistical thermal design uncertainty associated with PCST calculation is evaluated by Monte-Carlo sampling technique combined with subchannel analysis code. Maximum statistical design uncertainty of PCST is calculated to be 31 °C and is in a good agreement with that from RTDP method. Influence of downward flow in seed region on system sensitivity is investigated by improved Monte-Carlo thermal design procedure. Limiting thermal condition of MCST is 681 °C (650 °C of nominal + 31 °C) within 95/95 limit for SWFR.     

Reactor vessel cladding separate effects studies

The existence of a layer of tough weld overlay cladding on the interior of a light-water reactor pressure vessel could mitigate damage caused during certain overcooling transients. The potential benefit of the cladding is that it could keep a short surface flaw, which would otherwise become long, from growing either by impeding crack initiation or by arresting a running crack. Two aspects critical to cladding behavior will be reported: irradiation effects on cladding toughness and the response of mechanically loaded, flawed structures in the presence of cladding.A two-phase irradiation experiment is being conducted. In the first phase, Charpy impact and tensile specimens from a single wire, submerged-arc stainless steel weld overlay were irradiated to 2 × 10²³ neutrons/m² (>1 MeV) at 288°C. Typical, good quality pressure vessel cladding exhibited very little irradiation-induced degradation. However, ductile-to-brittle transition behavior, caused by temperature-dependent failure of the residual δ-ferrite, was observed. In contrast, specimens from a highly diluted, poor quality weldment were markedly embrittled. In the second phase of irradiations, now in progress, a commercially produced three-wire series arc weldment will be evaluated under identical irradiation and testing conditions as the first series. In addition, 0.5T compact specimens of both weldments and higher fluences will be examined.A two-phase program is also being conducted utilizing relatively large bend specimens that have been clad and flawed on the tension surface. The testing rationale is that if a surface flaw is pinned by the cladding and cannot grow longer, it will also not grow beyond a certain depth, thereby arresting the entire flaw in a stress field in which it would otherwise propagate through the specimen. The results of phase one showed that single wire cladding with low-to-moderate toughness appeared to have a limited ability to mitigate crack propagation. For the second phase, three-wire cladding has been deposited on a base plate with a very high ductile-to-brittle transition temperature allowing testing to ascertain the crack inhibiting capability of tough upper shelf cladding.     

Influence of iodine on the burst strain of Zircaloy-4 cladding tubes under simulated reactor accident conditions

The Stress Corrosion Cracking (SCC) behavior of short tubular Zircaloy-4 (Zry) specimens by the action of iodine was investigated out-of-pile between 600 and 1100°C under inert gas conditions. The Zry cladding tubes were used as-received, with an inner preliminary oxidation and pre-damaged, respectively. Moreover, the influence of the UO₂ oxygen potential on the SCC behavior was studied. The burst and creep-rupture tests (time-to-rupture ?15 min) clearly show that the deformation behavior of Zry cladding tubes below 850°C is influenced by the presence of iodine. A low ductility failure of Zry tubes takes place that is characterized by little deformation as compared to reference specimens without iodine. Moreover, in the isothermal, isobaric experiments, the time-to-failure of the iodine containing specimens is markedly shorter as compared to the iodine-free reference specimens. Internal preoxidation of the cladding tube or UO₂ in the specimens exert an additional influence on the mechanical properties of Zry. SEM examinations of the rupture surface of the cladding tubes show that in the presence of iodine and at burst temperatures below 850°C the cracks in the cladding material are mainly intergranular followed by ductile residual rupture.     

Thermo-physical properties and transient heat transfer of concrete at elevated temperatures

The objective of this study is to produce our own experimental data of physical properties of domestic concrete used in Korean NPPs, and to study on the thermal behavior of concrete exposed to high temperature conditions. The compressive strength and chemical composition of the concrete used in the Yonggwang NPP units 3 and 4 were analyzed. The chemical composition of Korean concrete is similar to that of US basaltic concrete. The thermal properties of the concrete, such as density, conductivity, diffusivity, and specific heat were also measured with a wide temperature range of 20–1100 °C. Most thermo-physical properties of concrete decrease with an increase in temperature except for the specific heat, and particularly the conductivity and the diffusivity are a 50% lower at 900 °C as compared with the values at room temperature. The specific heat increases until 500 °C, decreases from 700 to 900 °C, and then increases again when temperature is above 900 °C. In this work, we also have performed CORCON analysis and MCCI experiments to simulate a transient thermal behavior of concrete exposed to high temperature conditions. The measured maximum downward heat flux to the concrete specimen was estimated to be about 2.1 MW m⁻² and the maximum erosion rate of the concrete to be 175 cm h⁻¹ with maximum erosion depth of about 2 cm. In the CORCON analysis, it is found that the concrete compositions have an important effect upon concrete erosion.     

European experiences of the proposed ASTM test method for crack arrest toughness of ferritic materials

The proposed ASTM test method for measuring the crack arrest toughness of ferritic materials using wedge-loaded, side-grooved, compact specimens was applied to three steels: A514 bridge steel tested at −30°C (CV30–50°C), A588 bridge steel tested at −30°C (CV30–65°C), and A533B pressure vessel steel tested at +10°C (CV30-12°C) and +24°C (CV30+2°C). Five sets of results from different laboratories are discussed here; in four cases FOX DUR 500 electrodes were used for notch preparation, in the remaining case HARDEX-N electrodes were used. In all cases, notches were prepared by spark erosion, although root radii varied from 0.1–1.5 mm. Although fast fractures were successfully initiated, arrest did not occur in a significant number of cases.The results showed no obvious dependence of crack arrest toughness, K_a, (determined by a static analysis) on crack initiation toughness, K₀. It was found that K_a decreases markedly with increasing crack jump distance, Δα/W. A limited amount of further work on smaller specimens of the A533B steel showed that lower K_a values tended to be recorded.It is concluded that a number of points relating to the proposed test method and notch preparation are worthy of further consideration. It is pointed out that the proposed validity criteria may screen out lower bound data. Nevertheless, for present practical purposes, K_a values may be regarded as useful in providing an estimate of arrest toughness — although not necessarily a conservative estimate.     

Stress corrosion cracking of low-alloy reactor pressure vessel steels under boiling water reactor conditions

The stress corrosion cracking (SCC) behaviour of different reactor pressure vessel (RPV) steels and weld filler/heat-affected zone materials was characterized under simulated boiling water reactor (BWR) normal water (NWC) and hydrogen water chemistry (HWC) conditions by periodical partial unloading, constant and ripple load tests with pre-cracked fracture mechanics specimens. The experiments were performed in oxygenated or hydrogenated high-purity or sulphate/chloride containing water at temperatures from 150 to 288 °C. In good agreement with field experience, these investigations revealed a very low susceptibility to SCC crack growth and small crack growth rates (<0.6 mm/year) under most BWR/NWC and material conditions. Critical water chemistry, loading and material conditions, which can result in sustained and fast SCC well above the ‘BWRVIP-60 SCC disposition lines’ were identified, but many of them generally appeared atypical for current optimized BWR power operation practice or modern RPVs. Application of HWC always resulted in a significant reduction of SCC crack growth rates by more than one order of magnitude under these critical system conditions and growth rates dropped well below the ‘BWRVIP-60 SCC disposition lines’.     

Finite element fracture mechanics analysis of creep rupture of fuel element cladding

A finite element fracture mechanics technique is applied for simulating the elevated temperature creep rupture behavior of initially defected austenitic stainless steel fuel element cladding. The basic analytical approach consists of determining total instantaneous strain energy release rates G_T, and the corresponding values of the stress intensity factor K_T from sequential linear elastic finite element solutions and relating these to either an effective creep fracture toughness parameter G_ec (or K_ec) or to creep crack growth rates , obtained from test results.An initial application of this approach has been made to simulate the creep rupture behavior of initially defected type 316 austenitic stainless steel fuel element cladding in the 20% cold worked condition, tested at 650°C. This application has provided a relationship in the simple familiar form: , where σ is the nominal loop stress, a is the initial depth of a longitudinal crack, h is the cladding thickness, t_r is the time to rupture, and q is a structure sensitive parameter which accounts for the influence of the environment. is a function, obtained from finite element solutions, which accounts for the geometric differences between the present structure and the classical Griffith plate. The function ) is obtained from creep rupture tests of cladding with varying initial flaw depths and times to rupture under corrosive as well as inert environments.Performing time-dependent analyses, a preliminary relationship is obtained between the instantaneous values G_T and K_T, and crack growth rates under corrosive and non-corrosive environments. The analytical predictions of critical combinations of cladding flaw configurations, stresses, times to rupture and crack growth rates are in good agreement with the limited test data available for comparison. Current applications are aimed at the long-term cyclic creep fracture behavior of fast reactor fuel elements, using a nonlinear finite element code. In addition, multiple intergranular fracture configurations are being investigated.     

An analysis of iodine stress corrosion cracking of zircaloy-4 tubing

A fracture mechanics approach to interpreting iodine-vapor stress-corrosion cracking in unirradiated Zircaloy-4 tubing is presented in which crack velocities are related to the fourth power on the stress intensity factor, $\text{K}{}_{\mathrm{I}}$. The crack growth power law on $\text{K}{}_{\mathrm{I}}$ is shown to predict well the time-to-failure in internally pressurized Zircaloy-4 tubing at 360 and 400°C reported by Busby, Tucker and McCauley. The temperature dependency on iodine stress corrosion cracking in Zircaloy can be described by an Arrhenius-type equation in which the activation energy $\text{Q}$ for recrystallized and cold-reduced Zircaloy was determined to be 42.9 and 35.9 kcal/mole, respectively. It is concluded that the geometry of the initial surface flaw, through its attendant elastic stress field, is directly responsible in controlling the SCC time-to-failure, cold working having a relatively small effect on increasing the susceptibility to SCC. The effects of neutron flux on iodine stress corrosion cracking of Zircaloy-4 tubing in-reactor are still unknown.     

Experiments on air ingress during severe accidents in LWRS

Various kinds of experiments on the oxidation of Zircaloy-4 cladding material in different scales and under different conditions at temperatures 800–1300 °C (small scale) and up to 2000 °C (large scale) are presented. The focus of this work was on prototypic mixed air–steam atmospheres and sequential reaction in steam and air, where no data were available before. The separate-effects tests were performed to support the large scale bundle test QUENCH-10 and to deliver first data for model development.