(JAERI) The Third International Symposium on Advanced Nuclear Energy Research “Global Environment and Nuclear Energy”

Nodular corrosion on the surface of Zircaloy-2 fuel cladding tubes was investigated and the effects of irradiation time, linear heat rate, and fast neutron flux on the nucleation and growth of nodules were statistically studied. The nucleation and growth rate of nodules were found to decrease with irradiation time. Then, the nodular corrosion resistance of neutron irradiated Zircaloy-2 was examined in an out-of-pile corrosion test, and neutron irradiation was seemed to improve the Zircaloy-2 nodular corrosion resistance. Finally, microstructural evolution of Zircaloy-2 during neutron irradiation was investigated and radiation-induced dissolution of inter-metallic precipitates was confirmed. It was proposed that an increase of the dissolved alloying element concentration should improve the nodular corrosion resistance of Zircaloy-2 and re- strain the nucleation and growth of nodules on Zircaloy-2 fuel cladding tubes during irradiation in BWRs.     

Effect of Neutron Irradiation on Iodine Stress Corrosion Cracking Susceptibility of Zircaloy-2 Tubing

The effect of neutron irradiation on the iodine stress corrosion cracking (SCC) of Zircaloy-2 tubing of 8×8 type design was studied by means of ring tension test, using specimens unirradiated and irradiated to 3.2×l0¹⁹ and 3.0×10²⁰ n/cm² (E>lMeV). The SCC threshold stresses were determined from constant load tests and the SCC initiation stresses were defined from constant cross-head rate tests. Both stresses increased with increasing neutron fluence, reaching a maximum at a neutron fluence between 10¹⁹ and 10²⁰ n/cm² and then decreased. The tendency is qualitatively in good agreement with the hypothetical conclusion derived by Lunde & Videm, for SCC failure stresses from internal gas pressurization tests on various Zircaloy cladding tubes irradiated at different reactor conditions. The cause of the increase in the SCC susceptibility at neutron fluences above 10²⁰ n/cm² would be ascribed to radiation anneal hardening phenomenon and resultant inhomogeneous incipient deformation characterized by dislocation channels.     

The Effect of Annealing Parameter on Corrosion Resistance of Zircaloy-2

The effects of equal ΣAi for different combinations of the annealing temperature and annealing time on corrosion resistance and evolution of precipitates of Zircaloy-2 were investigated.

Nodular corrosion resistance in the out-of-pile corrosion test was degraded with increasing ΣAi only when it was increased by extending the annealing time at 894 K but did not depend on ΣAi which was increased by raising the annealing temperature for a constant annealing time of 2.5h. Extensive observation and micro-analysis of precipitates by analytical electron microscope (AEM) suggested the cause of degradation of nodular corrosion resistance to be the remarkable increase in volume fraction of Si-containing precipitates such as Zr₃Si and Zr₂Si, which were observed more frequently in large ΣAi only when it was increased by extending the annealing time at 894K.

On the other hand, uniform corrosion resistance was improved with increasing ΣAi irrespective of the annealing conditions, but was slightly better when ΣAi was increased by extending the annealing time under constant annealing temperature at 894K. This behavior can be under-stood from the reduction in content of alloying elements such as Fe, Cr and Ni which remained as solutes in the matrix at large ΣAi.     

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.     

Irradiated Behavior at High Burnup for HiFi Alloy

Irradiation tests of a BWR advanced Zr alloy (HiFi alloy) and Zircaloy-2 (Zry-2) were carried out in a Japanese commercial reactor and the irradiation performances of the materials were investigated. HiFi alloy and Zry-2 showed excellent resistance to corrosion up to 70 GWd/t, and furthermore, HiFi kept lower hydrogen pickup compared with Zry-2. TEM observation showed that the Fe/(Fe+Cr) ratio of Zr(Fe,Cr)2 type second phase particles (SPPs) for HiFi alloy and Zry-2 tended to decrease as fast neutron fluence increased and to saturate at high fluence. Zr-Fe-Cr SPPs did not completely disappear even for 6 cycles for the irradiated HiFi alloy and Zry-2. In order to clarify the mechanism of hydrogen absorption, an electrochemical technique was used for the oxide film of both materials as part of the out-of-pile test. The relation between the oxide surface potential and the hydrogen pickup fraction was estimated suggesting that the potential difference over the oxide film suppressed hydrogen (proton) diffusion in the oxide film.     

Zr-4合金a-Zr固溶体中的Fe、Cr含量分析

将电化学分离Zr-4合金中Zr(Fe,Cr)2第二相粒子的技术和原子吸收光谱分析技术相结合,建立了一种分析Zr-4合金a-Zr固溶体中Fe、Cr含量的新方法,并用这种新方法分析了不同热处理状态下Zr-4合金a-Zr固溶体中的Fe、Cr含量。分析结果表明,随着淬火温度的增加,a-Zr固溶体中的Fe、Cr含量和Fe/Cr比值均增加,而Zr(Fe,Cr)2第二相粒子中的Fe/Cr比值则相应降低。结合以前的工作,可得结论:过饱和固溶在a-Zr固溶体中的Fe、Cr含量对Zr-4合金的耐水侧腐蚀性能有重要影响。     

加工工艺对Zr—4管抗疖状腐蚀的影响

应用高压釜，金相及电子显微镜等研究了不同加工工艺的Ｚｒ－４包壳的疖状腐蚀行为。结果表明，改进工艺加工的管材比常规工艺加工的管材有更优良的抗疖状腐蚀性能，去应力试样比再结晶试样有较强的抗疖状腐蚀能力，影响Ｚｒ－４合金抗腐蚀性能的主要因素是Ｆｅ和Ｃｒ合金元素在α－Ｚｒ中的固溶含量，而不是第二相粒子的大小。     

In-Pile Eddy Current Test on PWR Fuel Rod Failed by Pellet-Cladding Interaction

Zirconium oxide nodules formed on BWR fuel rods were characterized quantitatively and correlated statistically with the rod operational parameters. Cladding specimens were obtained from fuel rods irradiated in a commercial BWR. Their burnup and fast neutron fluence ranged 17~38 GWd/t and 4×10²⁵~8×10²⁵ n/m², respectively. Characteristic variables of the nodules such as maximum thickness T ^max (μm) were measured on metallographs of the cladding cross sections. These variables were correlated by multiple regression analyses with the operational parameters, such as irradiation time t (d), linear heat rate p (kW/m) and fast neutron flux ø (n/m²-s). For example, the maximum thickness depended on linear heat rate and showed a saturating tendency with burnup B (GWd/t) (T^max ∝ t0.8+0.5 p2.3±0.9 or T ^max ∝B0.8+0.4p1.5±0.5). This decrease of growth rate with irradiation time was interpreted in terms of a microstructure change of Zircaloy-2 during neutron irradiation. Results of transmission microscopy and energy dispersive X-ray spectroscopy indicated that the alloying elements such as Fe, Cr and Ni dissolved from intermetallic precipitates into the base metal during neutron irradiation. Dissolution of the alloying elements might be effective in decreasing the growth rate of nodules.     

The effect of alloying modifications on hydrogen uptake of zirconium-alloy welding specimens during corrosion tests

The hydrogen uptake behavior during corrosion tests for electron beam welding specimens made out of Zircaloy-4 and zirconium alloys with different compositions was investigated. Results showed that the hydrogen uptake in the specimens after corrosion tests increased with increasing Cr content in the molten zone. This indicated that Cr element significantly affected the hydrogen uptake behavior. Fe and Cr have a low solubility in α-Zr and exist mainly in the form of Zr(Fe,Cr)₂ precipitates, which is extremely reactive with hydrogen in its metallic state. It is concluded that the presence of Zr(Fe,Cr)₂ second phase particles (SPPs) is responsible for the increase in the amount of hydrogen uptake in the molten zone of the welding samples after corrosion, as Zr(Fe,Cr)₂ SPPs embedded in α-Zr matrix and exposed at the metal/oxide interface could act as a preferred path for hydrogen uptake.     

Effect of additives on the susceptibility of zirconium alloys to nodular corrosion

A new approach based on a stability analysis of a uniformly growing oxide film was applied to estimate the effect of alloying additives on the susceptibility of zirconium alloys to nodular corrosion. The analytical results agree with available experimental data on effect of Fe and Ni on resistance of Zircaloy-2 and Zircaloy-4 to the growth of nodular oxide.     

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.     

Development of the high-energy neutron fluence rate standard field in Japan with a peak energy of 45 MeV using the 7Li(p,n)7Be reaction at TIARA

In this study, we developed a 45 MeV neutron fluence rate standard of Japan. Quasi-monoenergetic neutrons with a peak energy of 45 MeV in the neutron standard field were produced by the ⁷Li(p,n)⁷Be reaction using a 50-MeV proton beam from an azimuthally varying field (AVF) cyclotron of the Takasaki Ion Accelerators for Advanced Radiation Application (TIARA). The neutron energy spectrum was measured using an organic liquid scintillation detector and a ⁶Li-glass scintillation detector by the time-of-flight method, and using a Bonner sphere spectrometer by the unfolding method. The absolute neutron fluence was determined using a proton recoil telescope (PRT) composed of the liquid scintillation detector and a Si(Li) detector that was newly developed in the present study. The detection efficiency of the PRT was obtained using the MCNPX code. The peak neutron production cross section for the ⁷Li(p,n)⁷Be reaction was also derived from the neutron fluence in order to confirm the neutron fluence of the TIARA high-energy neutron field. The peak neutron production cross section obtained in the present study was in good agreement with those of previous studies. The characteristics of the 45-MeV neutron field in TIARA were successfully evaluated in order to calibrate high-energy neutron detectors and high-energy neutron dosimeters.     

The effect of heat treatment on the microstructure and the corrosion resistance of Zircaloy-4 in 450°c steam

The effect of beta treatment and the subsequent cold work and alpha annealing on the microstructure and corrosion resistance of Zircaloy-4 in 450°C steam was investigated. The distribution of second-phase particles in test specimens were found to be not uniform and the effect of cold work and alpha annealing was the growth and redistribution of second-phase particles. By using SEM-EDX, the particles were identified as Zr-Fe-Cr and Zr-Fe. One cycle of the cold work and alpha annealing process improved the uniform corrosion resistance but enhanced the nodular corrosion. The difference was explained by a nodular growth mechanism which was related to the distribution of second-phase particles, the build-up of hydrogen pressure, the weakening of oxide after rate transition, and crack propagation along grain boundaries.     

活化法测量散裂靶中子能谱的实验验证

散裂靶中子的能谱对加速器驱动次临界系统的倍增因数和嬗变率等影响很大,计算表明散裂靶中子谱在MeV能区与裂变中子谱相近。本文利用活化法测量临界装置的泄漏中子谱和中子注量率,提出了用In、Al、Mg、Ti、Au、Zn、Ni、Rh、Fe和Co等活化箔测量散裂靶中子能谱和中子注量率的方案。结果表明,将活化箔在散裂靶中子场中辐照5h,中子注量最高达5×1014 cm-2量级,辐照后1h内取出活化箔,根据半衰期的长短安排测量顺序,可测量散裂靶的中子能谱和中子注量率。     

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.     

Influence of second phase particles on corrosion resistance of N36 alloy in superheated steam

The nodular corrosion resistance of N36 (Zr–0.8Sn–1Nb–0.3Fe–0.12O, wt%) alloy with different temperatures of interstage annealing and optimized Zircaloy-4 (lower content of tin) cladding tubes were tested in 500 °C and 10.3 MPa superheated steam. The result showed that no nodular spots appeared on surface of N36 alloy cladding tubes after 500 h corrosion, while nodular spots on optimized Zircaloy-4 tubes after 8 h. The differences of weight gain between two kinds of N36 tubes suggested that the percentage (volume fraction) of tetragonal Zr dioxide in oxide film can be effectively influenced by characteristics, such as size, of the second phase particles on the oxide film-matrix boundary. The analysis by SEM and LRS clearly indicated that smaller and fine distributed second phase particles were beneficial to stabilize the oxide via endure the uniform stress so it could slow down the altering process from tetragonal Zr dioxide to monoclinic Zr dioxide.     

Crack Initiation Mechanism in Non-ductile Cracking of Irradiated 304L Stainless Steels under BWR Water Environment

The deformation behavior and initiation mechanisms of intergranular (IG) and transgranular (TG) cracks in irradiated 304L stainless steel were studied by slow-strain-rate tensile tests in inert gas and simulated BWR water environments, followed by fractographic and microstructural examinations. Neutron irradiation was made in test reactors to fluences of up to 6.2x10²⁰ n/cm² (E>1 MeV). Intergranular cracking occurred in water above a critical neutron fluence of around 1 × 10²⁰ n/cm², based on the results of the SSRT tests and SEM fractography. That critical fluence is mechanistically supported by irradiated, deformed microstructures exhibiting dislocation channeling at that fluence, while radiation-induced Cr depletion at the grain boundaries was minor. Transgranular cracking of the irradiated material occurred in water below the critical fluence, initiating in the non-uniformly strained surface region of the test bar in the later stages of plastic deformation. The initiation of TG cracking is hypothesized to be related to a high density of deformation twins. Intergranular cracking is proposed to have initiated where localized slip bands terminated at grain boundaries, while TG cracking is inferred to have initiated at deformation twin boundaries. High stress and strain concentrations at grain/twin boundaries would be the common cause of non-ductile crack initiation.     

纳米晶Al的快中子辐照效应研究

采用真空热压技术将自悬浮定向流法制得的纳米Al粉压制成平均晶粒尺寸约为120nm的块体,并对其进行了注量为1.9×1012~7.2×1014 cm-2的快中子(E1 MeV)辐照。通过X射线衍射(XRD)分析、扫描电子显微镜与能谱(SEM-EDS)分析和显微硬度测试研究了快中子辐照对纳米晶Al的微观结构和显微硬度的影响。研究结果表明:快中子辐照同时造成了纳米晶Al的平均晶粒尺寸增大和显微硬度提高。随快中子辐照注量的增大,纳米晶Al的平均晶粒尺寸和显微硬度分别增大了2.09%~9.09%和3.54%~4.37%。纳米晶Al的平均晶粒尺寸的增长率随快中子注量的增加而增大。     

Irradiation-Induced Dissolution of Precipitates in Zircaloy-2

Calibration curves of extremely low concentrations of the alloying elements Sn, Fe, Cr and Ni in Zircaloy were obtained, using standard samples, by energy dispersive X-ray spectroscopy to measure concentration distributions of alloying elements dissolved in the Zircaloy matrix. Their detectable limits were 0.21 at% for Sn, 0.06 at% for Fe. 0.04 at% for Cr and 0.03 at% for Ni. Then concentration distributions of alloying elements in unirradiated and neutron irradiated Zircaloy-2 were measured using these calibration curves. It was confirmed that neutron irradiation increased the dissolved concentrations of Fe. Cr and Ni. Furthermore, Cr diffused slower than Fe and Ni. It was suggested that the rate limiting process of irradiation-induced dissolution from Fe, Cr-type precipitates into the matrix was the diffusion of alloying atoms in the precipitates and that the dissolution process proceeded due to displacement of alloying atoms from the precipitates into the matrix and diffusion in the matrix.     

Residence Time of Crud on Surfaces of Channel Box in JPDR

Characterization of crud on surfaces of the channel box in JPDR has been carried out by means of chemical, radiochemical, X-ray diffraction and infrared spectrum analyses. The main cations in the crud are Fe and Ni: The sum of their weights amounts to more than 90% of the total weight of the cations found. The results of X-ray diffraction and infrared spectrum analyses revealed that the crud consisted of Ni_0.65Fe_2.35O₄, NiO and γ-FeOOH.

Based on the neutron flux calculated from the burn-up of ²³⁵U and ²³⁸U in the spent fuel, the apparent residence time of elements on the surface of the channel box was calculated to be 230 d for Co, 260 d for Ni and 70 d for Fe. The value for Fe should be taken as a minimum value, because of the presence of γ-FeOOH in the crud, which has been formed during the storage in a pond.

The present data are discussed in correlation with the one in the reactor water.