Failure behavior of Zircaloy-4 cladding after oxidation and water quench

Simulated LOCA (loss of coolant accident) tests and subsequent mechanical tests on Zircaloy-4 cladding were carried out to evaluate the failure behavior of the cladding. Zircaloy-4 claddings were oxidized in a steam environment from 900 to 1250 °C for a given time period followed by a flooding of cool water to simulate LOCA tests. After the simulated LOCA test, the ductility of the oxidized cladding was evaluated by mechanical tests such as ring compression test and 3-point bend test. Evaluation of the absorbed contents such as hydrogen and oxygen were also carried out. The results showed that Zircaloy-4 cladding failed during thermal shock when the ECR (equivalent cladding reacted) value exceeded 20%. Lower boundary of brittle failure at thermal shock corresponds to 20% of ECR line calculated by the Baker-Just equation regardless of test temperature. On the other hand, boundary of ductile failure by the mechanical test did not followed after the ECR line. It rapidly decreased above 1000 °C to show that all Zircaloy-4 claddings behaved brittle fracture above 1150 °C when it oxidized at 300 s. Microstructural analysis revealed that boundary of ductile failure by the mechanical test fitted well when the absorbed oxygen content inside the prior-β layer was below 0.5 wt%.     

Oxidation kinetics of Zircaloy-4 and Zr-1Nb-1Sn-0.1Fe at temperatures of 700-1200 °C

The oxidation characteristics for the Zircaloy-4 and Zr-1.0Nb-1.0Sn-0.1Fe alloys were investigated in the temperature ranges of 700-1200 °C for 3600 s under steam supply conditions, using a modified thermo-gravimetric analyzer. The oxidation at these temperatures generally complied with the parabolic rate law for the examined duration up to 3600 s. However, the parabolic rate was not obeyed in the temperature ranges of 800-1050 °C. The oxidation kinetics were changed depending on the oxidation temperatures due to the phase transformations of the base metal and its oxide. The oxidation rate exponents of the Zr-1.0Nb-1.0Sn-0.1Fe alloy at all the temperatures were higher than those of Zircaloy-4. Considering the data controlled by the parabolic rates at 700, 1100, 1150, and 1200 °C, the oxidation rate constants were the same slopes as the Baker-Just relationship. The rate transition at 800 °C could have resulted from the phase transformation of the base metal and those at 1000 and 1050 °C could have resulted from the lateral cracks in the oxide due to the ZrO₂ phase transformation from a monoclinic structure to a tetragonal structure.     

High-temperature steam-oxidation behavior of Zr-1Nb cladding alloy E110

Oxidation experiments were conducted at 1000-1200 °C in flowing steam with samples of as-received Zr-1Nb alloy E110 tubing and/or polished E110 tubing. The purpose was to determine the oxidation behavior of this alloy under postulated loss-of-coolant accident conditions in light water reactors. The as-received E110 tubing exhibited a high degree of susceptibility to nodular oxidation and breakaway oxidation at relatively low test times, as compared to other cladding alloys. The nodules grew much more rapidly at 1000 °C than 1100 °C, as did the associated hydrogen uptake. The oxidation behavior was strongly affected by the surface condition of the materials. Polishing to ≈0.1 μm roughness (the roughness of the as-received tubing was ≈0.4 μm) delayed breakaway oxidation. Polishing also removed surface impurities. For polished samples oxidized at 1100 °C, no significant nodular oxidation appeared up to 1000 s. For polished samples oxidized at 1000 °C, hydrogen uptake >100 wppm was delayed from ≈300 s to >900 s. Weight-gain coefficients were determined for pre-breakaway oxidation of polished-only and machined-and-polished E110 tubing samples: 0.162 (mg/cm²)/s^0.5 at 1000 °C and 0.613 (mg/cm²)/s^0.5 at 1100 °C.     

The effect of high pressure steam on the oxidation of low-Sn Zircaloy-4 at temperatures between 700 and 900 °C

To see the effect of high pressure steam on the oxidation of low-Sn Zircaloy-4, we measured the weight gain of low-Sn Zircaloy-4 claddings at temperatures between 700 and 900 °C under steam pressure up to 10 MPa. High pressure steam enhanced the oxidation, and the measured weight gain by oxidation was exponentially proportional to the applied steam pressure. The cubic rate law was still applicable to the oxidation under high pressure steam. High pressure steam made poor-quality monoclinic oxide containing many cracks. Rapid transformation of tetragonal oxide to monoclinic oxide due to high pressure steam seems to be the reason for the formation of poor-quality monoclinic oxide.     

Prototypical experiments relating to air oxidation of Zircaloy-4 at high temperatures

The mechanism of the reaction between Zircaloy-4 and air at temperatures from 800 to 1500 °C was studied. Air attack under prototypical conditions with air ingress during a hypothetic severe nuclear reactor accident was investigated. Oxidation in air and in air and nitrogen-containing atmospheres leads to a major degradation of the cladding material. The main mechanism is the formation of zirconium nitride and its re-oxidation. Pre-oxidation in steam prevents air attack as long as the oxide scale is intact. Under steam/oxygen starvation conditions, the oxide scale is reduced and significant external nitride formation takes place. When modeling air ingress in severe accident computer codes, parabolic correlations for oxidation in air may be applied only for high temperatures (>1400 °C) and for pre-oxidized cladding (?1100 °C). Under all other conditions, faster, rather linear reaction kinetics should be applied.     

Comparison of the air oxidation behaviors of Zircaloy-4 implanted with yttrium and cerium ions at 500 °C

As a valuable process for surface modification of materials, ion implantation is eminent to improve mechanical properties, electrochemical corrosion resistance and oxidation behaviors of varieties of materials. To investigate and compare the oxidation behaviors of Zircaloy-4, implantation of yttrium ion and cerium ion were respectively employed by using an MEVVA source at the energy of 40 keV with a dose ranging from 1×10¹⁶ to 1×10¹⁷ ions/cm². Subsequently, weight gain curves of the different specimens including as-received Zircaloy-4 and Zircaloy-4 specimens implanted with the different ions were measured after oxidation in air at 500 °C for 100 min. It was obviously found that a significant improvement was achieved in the oxidation behaviors of implanted Zircaloy-4 compared with that of the as-received Zircaloy-4, and the oxidation behavior of cerium-implanted Zircaloy-4 was somewhat better than that of yttrium-implanted specimen. To obtain the valence and the composition of the oxides in the scale, X-ray photoemission spectroscopy was used in the present study. Glancing angle X-ray diffraction, employed to analyze the phase transformation in the oxide films, showed that the addition of yttrium transformed the phase from monoclinic zirconia to tetragonal zirconia, yet the addition of cerium transformed the phase from monoclinic zirconia to hexagonal zirconia. In the end, the mechanism of the improvement of the oxidation behavior was discussed.     

A creep rupture criterion for Zircaloy-4 fuel cladding under internal pressure

The usual criterion which limits the cladding strain to 0.01 to prevent the creep rupture under internal pressure seems too conservative for application to transport and interim storage. So we have analysed CEA’s data on this subject for CWSR Zircaloy-4 in order to find a less conservative criterion. Temperatures between 350 and 470 °C were studied for stresses between 100 and 550 MPa, according to the irradiation level from 0 to 9.5 × 10²⁵ n m⁻². Except for high stressed irradiated material (because of low ductility), the plastic instability appears as the major mechanism of rupture. For the unirradiated material, it is essentially due to the stress increase with strain. This instability is accelerated by annealing for the irradiated one at moderate or low stress. From these considerations, we propose a new rupture criterion for CWSR Zircaloy-4 cladding submitted to internal pressure, for both unirradiated and irradiated materials.     

Degradation and oxidation of B4C control rod segments at high temperatures

Extensive series of test were performed of the degradation of boron carbide absorber rods and the oxidation of the resultant absorber melts. Various types of control rod segments made of commercial materials used in French 1300 MW PWRs were investigated in the temperature range between 800 °C and 1700 °C in a steam atmosphere. The gaseous reaction products were analyzed quantitatively by mass spectroscopy for evaluation of the oxidation rates. Extensive post-test examinations were performed by light microscopy, scanning electron microscopy as well as EDX and Auger spectroscopy. Rapid melt formation due to eutectic interactions of stainless steel (cladding tube) and B₄C, on the one hand, and steel and Zircaloy-4 (guide tube), on the other hand, was observed at temperatures above 1250 °C. Complex multi-component, multi-phase melts were produced. ZrO₂ oxide scale on the outside kept the melt within the guide tube, thus preventing its early relocation and oxidation. Rapid oxidation of the absorber melts and remaining boron carbide pellets took place after failure of the protective oxide shell above 1450 °C. Only very little methane was produced in these tests which is of interest in fission product gas chemistry because of the production of organic iodine.     

Air oxidation of Zircaloy-4, M5® and ZIRLO™ cladding alloys at high temperatures

The paper presents the results of isothermal and transient oxidation experiments of the advanced cladding alloys M5® and ZIRLO™ in comparison to Zircaloy-4 in air at temperatures from 973 to 1853 K.Generally, oxidation in air leads to a strong degradation of the cladding material. The main mechanism of this process is the formation of zirconium nitride and its re-oxidation. From the point of view of safety, the barrier effect of the fuel cladding is lost much earlier than during accident transients with a steam atmosphere only.Comparison of the three alloys investigated reveals a qualitatively similar, but quantitatively varying oxidation behavior in air. The mainly parabolic oxidation kinetics, where applicable, is comparable for the three alloys. Strong differences of up to 500% in oxidation rates were observed after transition to linear kinetics at temperatures below 1300 K.The paper presents kinetic rate constants as well as critical times and oxide scale thicknesses at the point of transition from parabolic to linear kinetics.     

High-pressure hydriding of Zircaloy cladding by the thermogravimetry and tube-burst techniques

Hydriding kinetics of modified Zircaloy claddings was studied by the thermogravimetric method at 400 °C and the tube-burst technique at 315 °C. Some specimens were prefilmed with a thin oxide layer by air oxidation on both the inner and outer surfaces which were either pickled or blasted. In the thermogravimetric test, the hydriding rates of bare cladding specimens (no oxide prefilm) were in the range 0.9-1.6 mg/cm² h with little effect of the surface treatment. Incubation times were less than 1 h or even zero. In the tube-burst test, immediate and extensive hydrogen uptake was observed for these non-coated specimens. On the other hand, the cladding specimens with oxide prefilm exhibited lower hydriding rates ranging from 0.01 to 0.05 mg/cm² h and incubation times increased to 42 h. In addition, no hydrogen uptake was observed for all oxide-coated specimens for 100-750 h.     

Ring test of 304 stainless steel tube oxidized in high temperature steam

The residual mechanical strength and ductility of 304 stainless steel claddings were studied in tests of ring compression, ring tension and hardness. The tests were made after the specimens had been heated in steam and in argon gas to temperatures between 900° C and 1350° C. The effects of crystal grain growth and oxygen absorption in the metal on the residual properties were also studied. The experiments showed that the claddings retained their ductility well at temperatures below 1300° C and when the oxide scale thickness was below 40% of the initial wall thickness. The claddings maintained a constant level of residual strength when heated at temperatures over 1000° C or when the oxide thickness was over five percent. These results were used to support our proposals for limits on stainless steel cladding damage in a LOCA.     

Synopsis and outcome of the QUENCH experimental program

The paper gives an overview of the main outcome of the QUENCH program launched in 1997 at the Karlsruhe Institute of Technology (KIT), formerly Karlsruhe Research Center (FZK). The research program comprises bundle experiments as well as complementary separate-effects tests. The focus of the experiments performed from 1997 to 2009 was on scenarios of severe accidents whereas that of the future test program will be on large-break loss-of-coolant accidents (LOCA) in the frame of design-basis accidents, and debris coolability, in the frame of severe accidents. The major objective of the program is to deliver experimental and analytical data to support the development and validation of quench and quench-related models as used in code systems that model severe accident progression in light water reactors.So far, 15 integral bundle QUENCH experiments with 21-31 electrically heated fuel rod simulators of 2.5 m length have been conducted. The following parameters and their influence on bundle degradation and reflood have been investigated: degree of pre-oxidation, temperature at initiation of reflood, flooding rate, influence of neutron absorber materials (B₄C, AgInCd), air ingress, and influence of the type of cladding alloy.In six tests, reflooding of the bundle led to a temporary temperature excursion driven by runaway oxidation of zirconium alloy components and resulting in release of a significant amount of hydrogen, typically two orders of magnitude greater than in those tests with “successful” quenching in which cool-down was rapidly achieved. Considerable formation, relocation, and oxidation of melt were observed in all tests with escalation. The temperature boundary between rapid cool-down and temperature escalation was typically in the range of 2100-2200 K in the “normal” quench tests, i.e. in tests without absorber and/or steam starvation. Tests with absorber and/or steam starvation were found to lead to temperature escalations at lower temperatures.All phenomena occurring in the bundle tests have been investigated additionally in parametric and more systematic separate-effects tests. Oxidation kinetics of various cladding alloys, including advanced ones, have been determined over a wide temperature range (873-1773 K) in different atmospheres (steam, oxygen, air, and their mixtures). Hydrogen absorption by different zirconium alloys was investigated in detail, recently also using neutron radiography as non-destructive method for determination of hydrogen distribution in claddings. Furthermore, degradation mechanisms of absorber rods including B₄C and AgInCd as well as the oxidation of the resulting low-temperature melts have been studied. Steam starvation was found to cause deterioration of the protective oxide scale by thinning and chemical reduction.The most recent topic of the QUENCH program has been investigation of the behavior of advanced cladding materials (ACM) in comparison with the classical Zircaloy-4. Although separate-effects tests have shown some differences in oxidation kinetics, the influence of the various cladding alloys on the integral bundle behavior during oxidation and reflooding was only limited.     

The effect of oxidation-and-quenching process during a LOCA on the behavior of the oxidation and embrittlement of Zircaloy-4 cladding under reheating transients

Two-sided oxidation tests, ring compression tests and semi-integral quench tests on Zircaloy-4 cladding specimens were conducted under temperature transient conditions simulating a post-quench reheat transient in order to evaluate the effect of high-temperature oxidation and quenching during a loss-of-coolant accident (LOCA) on the behavior of the oxidation and embrittlement of the cladding under a loss of long-term core-cooling condition. Test specimens prepared from non-irradiated Zircaloy-4 cladding tube were oxidized at a temperature between 1173 and 1473 K in steam flow and quenched by soaking the specimen in room temperature water. Re-heating tests were performed on the specimens in steam flow at a temperature between 1173 and 1473 K. The suppression of oxide layer growth and weight gain was observed under certain reheating-after-quenching conditions. Nevertheless, it seemed that the temperature transients including quenching-and-reheating process did not significantly affect the embrittlement of cladding. It was found that the embrittlement behavior of cladding during the temperature transients including quenching-and-reheating process could be dealt with on the basis of the Equivalent Cladding Reacted (ECR) based on the Baker–Just correlation.     

Hoop strength and ductility evaluation of irradiated fuel cladding

The objective of this study is to evaluate the hoop-directional mechanical properties comprising strength such as yield strength and ultimate tensile strength as well as mechanical ductility such as uniform elongation and total elongation. Therefore, in this paper, the ring tensile tests were performed in order to evaluate the mechanical properties of high burn-up fuel cladding under a hoop loading condition in a hot cell. The tests were performed with Zircaloy-4 nuclear fuel cladding whose burn-up is approximately 65,000 MWd/tU in the temperature range of room temperature to 800 °C. All the experiments were carried out at a constant strain rate of 0.01/s.On the basis of the ring tensile tests for a high burn-up Zircalay-4 cladding, the following conclusions were drawn. Firstly, the mechanical properties are abruptly degraded beyond 600 °C, which corresponds to a design-basis accident condition such as a RIA. Secondly, the un-irradiated fuel cladding showed ductile fracture behaviors such as 45° shear type fracture, cup and cone type fracture, cup and cup type fracture and chisel edge type fracture. While the high burn-up Zircalay-4 cladding showed a brittle fracture behavior even at the high temperatures (e.g. over 600 °C) which are achievable during a RIA. Thirdly, in the case of the high burn-up Zircalay-4 cladding, the strength, ductility and the energy to break are strongly dependent on the material property itself which are degraded by oxidation and hydriding during an operation rather than the temperature. Fourthly, hydride rim formation in the vicinity of metal-oxide interface can play an important role in the degradation of the mechanical properties for high burn-up fuel cladding.     

Investigation on boron carbide oxidation for nuclear reactor safety: Experiments in highly oxidising conditions

The oxidation kinetics of boron carbide pellets were investigated in steam/argon mixtures in the temperature range 1200-1800 °C for steam partial pressures between 0.2 and 0.8 bar and total flows (steam + argon) between 2.5 and 10 g/min resulting in gas velocities from 1.01 to 5.34 m/s. A kinetic model for boron carbide pellet oxidation depending on temperature, steam partial pressure and flow velocity is obtained. The activation energy of the oxidation process was determined to be 163 ± 8 kJ/mol. The strong influence of temperature and steam partial pressure on the boron carbide oxidation kinetics is confirmed. The obtained data suggest the coexistence of two kinetic regimes, one at 1200 °C and the other at 1400-1800 °C, with different dependence on steam partial pressure.     

Separate-effect tests on zirconium cladding degradation in air ingress situations

In the event of air ingress during a reactor or spent fuel pond low probability accident, the fuel rods will be exposed to air-containing atmospheres at high temperatures. In comparison with steam, the presence of air is expected to result in a more rapid escalation of the accident.A state-of-the-art review performed before SARNET started showed that the existing data on zirconium alloy oxidation in air were scarce. Moreover, the exact role of zirconium nitride on the cladding degradation process was poorly understood. Regarding the cladding behaviour in air + steam or nitrogen-enriched atmospheres (encountered in oxygen-starved conditions), almost no data were available.New experimental programmes comprising small-scale tests have therefore been launched at FZK, IRSN (MOZART programme in the frame of the International Source Term Program—ISTP) and INR. Zircaloy-4 cladding in PWR (FZK, IRSN) and in CANDU (INR) geometry are investigated. On-line kinetic data are obtained on centimetre size tube segments, by thermogravimetry (FZK, IRSN and INR) or by mass spectrometry (FZK). Plugged tubes 15 cm long (FZK) are also investigated. The samples are air-oxidised either in the “as-received” state, or after pre-oxidation in steam. “Analytical” tests at constant temperature and gas composition provide basic kinetic data, while more prototypical temperature transients and sequential gas compositions are also investigated. The temperature domains extend from 600 °C up to 1500 °C. Systematic post-test metallographic inspections are performed.The paper gives a synthesis of the results obtained, comparing them in terms of kinetics and oxide scale structure and composition. A comparative analysis is performed with results of the QUENCH-10 (Q-10) bundle test, which included an air ingress phase. It is shown how the data contribute to a better understanding of the cladding degradation process, especially regarding the role of nitrogen. For modelling of the oxide scale degradation under air exposure, important features that have to be taken into account are highlighted.     

Out-of-pile and In-pile Perfomance of Advanded Zirconium Alloys (HANA) for High Burn-up Fuel

The performance of the advanced Zr alloys (HANA) for a high burn-up fuel has been evaluated in the out-of-pile and in-pile conditions. The corrosion resistance of the HANA claddings was superior to Zicaloy-4 in a PWR-simulating loop condition. The improved corrosion resistance of the HANA claddings was attributed to the fine distribution of the precipitate. HANA claddings showed a higher creep resistance as compared to Zircaloy-4 from the thermal creep test. The deformation behavior of HANA in a LOCA condition was similar to Zircaloy-4. Threshold ECR value of HANA was higher than the conventional value of 17% in Zircaloy-4, which is mainly due to the fact that the Nb decreases the oxidation rate as well as the hydrogen pickup. Fretting wear test revealed that HANA claddings have a similar wear resistance to Zircaloy-4. From the irradiation test up to burn-up of about 12 GWd/MtU, HANA claddings showed a better corrosion resistance as well as a better creep resistance than Zircaloy-4. The in-pile corrosion resistance of the HANA claddings was improved by 40–50% as compared to Zircaloy-4 on the basis of the oxide thickness measurements.     

Failure-Bearing Capability of Oxidized Zircaloy-4 Cladding under Simulated Loss-of-Coolant Condition

With a view to examining the failure-bearing capability of Zircaloy-4 cladding under postulated Loss-of-Coolant Accident condition in LWRs, integral tests of rod-burst, oxidation and thermal-shock were performed using simulated fuel containing A1₂0₃ pellets sheathed in Zircaloy-4 specimen cladding, filled with He gas, and sealed. This simulated fuel rod was oxidized in steam flowing at the isothermal oxidation temperatures between 920 and 1,330°C for duration ranging of 3~180 min after the cladding burst. After isothermal oxidation, the rod was quenched with bottom-flooding water under the condition of constraint or no constraint.

The failure boundary oxidation condition of the cladding on quenching under no constraint condition lay in the region of 35~38% ECR for the isothermal oxidation temperatures between 1,050 and 1,330°C. For the temperatures ranging 970~1,050°C, the boundary value of ECR was somewhat lower than that obtained for higher temperatures.

The failure boundary oxidation condition of the cladding on quenching under constraint condition lay in the region of 19~24% ECR for the isothermal oxidation temperatures between 930 and 1,310°C. It is sufficiently large compared with the criterion of 15% ECR in Japanese acceptance criteria for ECCS. Hydrogen absorbed by the Zircaloy-4 cladding as well as oxygen played a dominant role in the fracture behavior of the rod during flooding under constraint condition.     

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.     

Developments in PWR cladding corrosion: materials and models

In pressurized water reactors Zircaloy-4 is a standard fuel cladding material. The aim of this paper is to present and evaluate corrosion data generated both in-reactor, and out-of-reactor on PWR claddings made of both Zircaloy-2 and Zircaloy-4 materials. The oxide thickness measurements of cladding tubes irradiated in the Ringhals 3 reactor, and oxide weight gain measurements carried out in Sandvik autoclaves at 400°C, 10.3 MPa clearly show that the stress relief annealed Zircaloy-2 is more corrosion resistant than Zircaloy-4 produced with an identical fabrication route. Furthermore, autoclave tests indicate that the hydrogen pickup fraction of the two alloys is very similar. The obtained data have been evaluated in regard to chemical composition and heat treatment. In addition, computer models, which simulate thermal and hydraulic reactor conditions and corrosion kinetic processes simultaneously, have been used to predict the in-reactor corrosion behaviour of the claddings.