Yuji Kurata

Corrosion studies in liquid Pb–Bi alloy at JAERI: R & D program and first experimental results

Abstract The program of corrosion study in liquid Pb–Bi at Japan Atomic Energy Research Institute (JAERI) is described. It was planned to clarify effects of parameters such as temperature, temperature difference between high-temperature and low-temperature parts, oxygen concentration in liquid Pb–Bi, flow rate, irradiation, stress and chemical composition of materials. The program contains the basic corrosion study in static Pb–Bi for elucidation of corrosion mechanism and effects of oxygen concentration in liquid Pb–Bi and ion irradiation on corrosion behavior. It also contains the corrosion study in flowing Pb–Bi using a corrosion testing loop. From the initial result of the static corrosion tests in oxygen saturated Pb–Bi at 550 ° C for 500 h, it was shown that the thickness of the corrosion film decreases with increasing Cr content in steels.

Corrosion–erosion test of SS316 in flowing Pb–Bi

Abstract Corrosion tests of austenitic stainless tube were done under flowing Pb–Bi conditions for 3000 h at 450 °C. Specimens were 316SS produced as a tubing form with 13.8 mm outer diameter, 2 mm thickness and 40 cm length. During operation, maximum temperature, temperature difference and flow velocity of Pb–Bi at the specimen were kept at 450, 50 °C, and 1 m/s, respectively. After the test, specimen and components of the loop were cut and examined by optical microscope, SEM, EDX, WDX and X-ray diffraction. Pb–Bi adhered on the surface of the specimen even after Pb–Bi was drained out to the storage tank from the circulating loop. Results differed from a stagnant corrosion test in that the specimen surface became rough and the corrosion rate was maximally 0.1 mm/3000 h. Mass transfer from the high temperature to the lower temperature area was observed: crystals of Fe–Cr were found on the tube surface in the low-temperature region. The sizes of crystals varied from 0.1 to 0.2 mm. The depositing crystals were ferrite grains and the chemical composition ratio (mass%) of Fe to Cr was 9:1.

Experimental investigation of lead-bismuth evaporation behavior

An experimental study on the equilibrium evaporation of lead-bismuth eutectic (LBE) was conducted to acquire essential and fundamental knowledge about the vaporization of LBE. The experiments were conducted using the transpiration method in which saturated vapor in an isothermal evaporation pot was transported by inert carrier gas and collected outside of the pot. The weight of the LBE pool in the pot was about 500 g. The investigated temperature range was from 450 to 750° C. We obtained data on the saturated vapor pressure of LBE, the vapor concentration of Pb, Bi and Bi2 in the LBE saturated gas phase, the activity coefficient of Pb in the LBE, and the LBE evaporation rate. The LBE vapor pressure equation is represented by logP LBE(Pa)=10.2–10, 100/T(K) in the temperature range from 550 to 750°C.

Creep rupture properties of a NiCrW superalloy in air environment

The creep rupture properties in air environment were studied at 900, 1000 and 1050°C for bar, plate and seamless tube materials of a NiCrW superalloy developed for use at service temperatures around 1000°C. The long-term creep rupture strength is estimated by applying the time-temperature parameter method to the creep rupture data. Data of anomalous behaviour due to oxidation strengthening, which is observed in creep curves with rupture times above ∼ 10 000 h at 1000°C, are corrected for the application of the time-temperature parameter method. The Larson-Miller parameter method is better than the Orr-Sherby-Dorn parameter method in respect of curve fitting to the present creep rupture data. The creep rupture strength of the bar material with grain sizes above ASTM No. 2 and of the plate material is above 9.8 MPa for a 1 × 105 h life at 1000°C, which is the final target of this program. The creep rupture strength increases with increasing grain size and heat treatment temperature.

Fission product release from Triso-coated UO2 particles at 1940 to 2320°C

Abstract The fission product release from TRISO-coated UO 2 particles was measured by post-activation heating at 1940 to 2320°C for use in a safety analysis. The results are analyzed mathematically with effective diffusion coefficients in each medium. 103 Ru, 99 Mo and 95 Nb are released at 1940 to 2320°C and have high effective diffusion coefficients. Although 140 Ba and 137 Cs are retained in TRISO-coated particles at 2050°C, they are released rapidly at 2320°C. This is attributed to the transition of beta to alpha SiC at 2320°C. 141 Ce, 140 La and 95 Zr are released little if any at 2320°C. Rare gas nuclides, iodine and tellurium seem to be retained in coated particles at this high temperature.

FEASIBILITY OF LEAD-BISMUTH-COOLED ACCELERATOR-DRIVEN SYSTEM FOR MINOR-ACTINIDE TRANSMUTATION

The feasibility for the lead-bismuth-cooled accelerator-driven system (ADS) to transmute minor actinides partitioned from high-level radioactive waste is discussed. Since lead-bismuth will cause considerable corrosion and erosion effects at high temperature, the fuel-clad temperature must be kept as low as possible. Moreover, the most critical issue of the ADS design is the engineering viability of the high-power spallation target and the beam window. The thermal-hydraulic and structural analysis was carried out for both the fuel assembly and the beam window. In addition to the analysis in steady state, the transient behaviors were also studied during typical transient and unprotected accidents. The results showed that engineering viability is reasonably achievable in the nominal operation. For the beam trip, which will be the most frequent transient, the number of events to cause the failure of the beam window is estimated as more than 105. For safety aspects of the ADS during unprotected accidents, the estimated results showed that unprotected loss of flow would cause the most significant problem, if the beam operation was kept. Therefore, high reliability of the beam shutdown is required for the ADS safety.

CREEP PROPERTIES OF 20% COLD-WORKED HASTELLOY XR

Abstract The creep properties of Hastelloy XR, in solution-treated and in 20% cold-worked conditions, were studied at 800, 900 and 1000°C. At 800°C, the steady-state creep rate and rupture ductility decrease, while rupture life increases after cold work to 20%. Although the steady-state creep rate and ductility also decrease at 900°C, the beneficial effect on rupture life disappears. Cold work to 20% enhances creep resistance of this alloy at 800 and 900°C due to a high density of dislocations introduced by the cold work. Rupture life of the 20% cold-worked alloy becomes shorter and the steady-state creep rate larger at 1000°C during creep of the 20% cold-worked alloy. It is emphasized that these cold work effects should be taken into consideration in design and operation of high-temperature structural components of high-temperature gas-cooled reactors.

In-pile and post-irradiation creep of type 304 stainless steel under different neutron spectra

In addition to post-irradiation creep tests, in-pile creep tests were performed using newly developed technology with in situ measurement under different neutron spectra. The in-pile creep properties of type 304 stainless steel at 550°C appear to depend on neutron spectrum, but a spectral effect on post-irradiation creep properties is not clearly seen. The rupture time of in-pile creep under a high thermal neutron flux condition is the shortest. The order of the rupture time following the high thermal flux condition is post-irradiation creep, in-pile creep with a thermal neutron shield condition and finally creep of unirradiated material, all in increasing order. It is suggested that the acceleration of creep deformation and fracture observed in irradiation creep tests may be related to enhancement of thermal creep in terms of FMD increased under a high thermal neutron flux in addition to increased helium embrittlement.

Creep and Rupture Behavior of a Special Grade Hastelloy-X in Simulated HTGR Helium

Creep and rupture tests were conducted for Hastelloy-XR (a modified version of the conventional Hastelloy alloy X) at 800, 900, and 1000/sup 0/C in simulated high-temperature gas-cooled reactor helium. Creep testing machines with special control of helium chemistry were used. As a result, the scatter of creep-rupture data could be reduced, and the variability of creep-rupture behavior due to manufacturing history could be resolved. Results of metallography and carbon analysis of ruptured specimens showed that the material improved resistance to corrosion in the helium environment, and carbon intrusion during the steady-state creep stage was suppressed to a negligible level. Under refined test conditions combined with the quality controlled material, it was demonstrated that there was little significant difference between helium and air in the creep-rupture results obtained at 800 to 1000/sup 0/C up to about 10/sup 4/h. The importance of maintaining the protective function of the surface oxide film of alloys was stressed in securing reproducibility and predictability of long-time creep performance.

Creep Properties of Base Metal and Welded Joint of Hastelloy XR Produced for High-Temperature Engineering Test Reactor in Simulated Primary-Coolant Helium

Creep tests of base metal, weld metal and welded joint of Hastelloy XR, which had the same chemical composition as Hastelloy XR produced for an intermediate heat exchanger of the High-Temperature Engineering Test Reactor, were conducted in simulated primary coolant helium. The weld metal and welded joint showed almost equal to or longer rupture time than the base metal of Hastelloy XR at 850 and 900°C, although they gave shorter rupture time at 950° C under low stress and at 1,000°C. The welded joint of Hastelloy XR ruptured at the base metal region at 850 and 900°C. On the other hand, it ruptured at the weld metal region at 950 and 1,000°C. The steady-state creep rate of weld metal of Hastelloy XR was lower than that of base metal at 850, 900 and 950°C. The creep rupture strengths of base metal, weld metal and welded joint of Hastelloy XR obtained in this study were confirmed to be much higher than the design allowable creep-rupture stress (SR ) of the Design Allowable Limits below 950°C.