Design and corrosion study of a closed spallation target module of an accelerator-driven system (ADS)

At the Forschungszentrum Karlsruhe (FZK) the characteristics of an accelerator-driven subcritical reactor system (ADS) are critically evaluated, mainly with respect to the potential of transmutation of minor actinides and long-lived fission products, to the feasibility and to safety aspects. The work is concentrating on system design, neutronics, thermalhydraulics, safety, materials and corrosion. This article describes the FZK approach to design a closed 4 MW(th) spallation target module with a solid beam window and eutectic lead–bismuth (Pb–Bi) as spallation material and cooling fluid, which is going to be implemented in the FZK three-beam concept of an ADS. This multi-beam concept shows significant improvements towards single-beam concepts from the literature with respect to power distribution in the subcritical blanket and thermal loads of heat removal from the beam window and the spallation region. For some selected martensitic and austenitic steels, corrosion tests in static lead are performed to examine their suitability as structural or window materials. Alloying aluminum into the surface layer by high-power electron beam treatment, corrosion can be reduced to nearly zero. One prerequisite to minimize corrosion is a proper oxygen control system (OCS) via the gas-phase to set the oxygen concentration in the liquid Pb–Bi. The dynamic behaviour of this oxygen control system is described. Finally, the KArlsruhe Lead LAboratory (KALLA) is introduced, the objectives of which are technological, thermal-hydraulic and corrosion investigations into the beam window, the spallation target module and the primary system of an ADS.     

The impact of the composition of structural steels on their corrosion stability in liquid Pb-Bi at 500 and 400 °C with different oxygen concentrations

In recent years, heavy liquid metals have found exercise as possible coolants and targets in the conversion of radioactive elements in accelerator driven systems (ADS). Liquid lead-bismuth eutectic alloy is one of candidates for this using tanks to its suitable nuclear and physical properties. Performed examination was aimed at research of compatibility choice materials for parts of ADS with liquid Pb-Bi eutectic alloy, influence of composition choice materials on their corrosion resistance, influence of temperature and oxygen content. We performed corrosion tests of 1000 h each on approximately 20 types of structural steels (austenitic, ferritic and martensitic) in convection loops with flowing Pb-Bi at 500 and 400 °C and using different oxygen concentrations. The impact of Fe, Cr, Ni, Mn, Si, Al and Mo content on the corrosion stability of these steels was measured without and after preliminary passivation through creating thin spinel or oxide layers on their surface.     

Behavior of steels in flowing liquid PbBi eutectic alloy at 420-600 °C after 4000-7200 h

This paper presents the results of steel exposure up to 7200 h in flowing LBE at elevated temperatures and is a follow-up paper of that with results of an exposure of up to 2000 h. The examined AISI 316 L, 1.4970 austenitic and MANET 10Cr martensitic steels are suitable as a structural material in LBE (liquid eutectic Pb₄₅Bi₅₅) up to 550 °C, if 10⁻⁶ wt% of oxygen is dissolved in the LBE. The martensitic steel develops a thick magnetite and spinel layer while the austenites have thin spinel surface layers at 420 °C and thick oxide scales like the martensitic steel at 550 °C. The oxide scales protect the steels from dissolution attack by LBE during the whole test period of 7200 h. Oxide scales that spall off are replaced by new protective ones. At 600 °C severe attack occurs already after 2000 and 4000 h of exposure. Steels with 8-15 wt% Al alloyed into the surface suffer no corrosion attack at all experimental temperatures and exposure times.     

Evaluation method of corrosion lifetime of conventional stainless steel canister under oceanic air environment

Natural exposure and accelerated corrosion tests of conventional stainless steels for canisters of Types 304, 304L, and 316(LN) for concrete casks were conducted using several test specimens and 1/5 scale canister models. The welding residual stress of a full-scale model canister was also measured and the lifetime of sealability of canisters against corrosion evaluated. The maximum pitting rate and crevice corrosion rate of Type 304 were approximately 20 and 30 μm/year. Many SCC in the 4 Point Bending (4PB) test specimens were found to initiate from the bottom of the corrosion area by pitting or crevice corrosion. The SCC propagation rates in Types 304 and 304L under natural conditions were around 1.2E−12 to 1.8E−11 m/s in the K (Stress Intensity Factor) range of 0.6–9.0 MPa m^1/2, and that of the accelerated test (60 °C, 95% RHS, filled with NaCl mist) around 1.0E−10 to 3.5E−9 m/s in the K range of 0.5–30 MPa m^1/2. The SCC propagation rates under both natural and accelerated conditions were independent of K. The lifetime of sealability estimated from 1/5 scale models was longer than that from the small bending test specimens and has a safety margin as a structure.     

Review of liquid metal corrosion issues for potential containment materials for liquid lead and lead–bismuth eutectic spallation targets as a neutron source

Lead (Pb) and lead–bismuth eutectic (44Pb–56Bi) have been the two primary candidate liquid metal target materials for the production of spallation neutrons. Selection of a container material for the liquid metal target will greatly affect the lifetime and safety of the target subsystem. For the liquid lead target, niobium–1 wt% zirconium (Nb–1Zr) is a candidate containment material for liquid lead, but its poor oxidation resistance has been a major concern. In this paper, the oxidation rate of Nb–1Zr was studied based on the calculations of thickness loss resulting from oxidation. According to these calculations, it appeared that uncoated Nb–1Zr may be used for a 1-year operation at 900°C at P_O2=1×10^–6 Torr, but the same material may not be used in argon with 5-ppm oxygen. Coating technologies to reduce the oxidation of Nb–1Zr are reviewed, as are other candidate refractory metals such as molybdenum, tantalum, and tungsten. For the liquid lead–bismuth eutectic target, three candidate containment materials are suggested, based on a literature survey of the materials’ compatibility and proton irradiation tests: Croloy 2-1/4, modified 9Cr–1Mo, and 12Cr–1Mo (HT-9) steel. These materials seem to be used only if the lead–bismuth is thoroughly deoxidized and treated with zirconium and magnesium.     

Corrosion characteristics of reduced activation ferritic steel, JLF-1 (8.92Cr–2W) in molten salts Flibe and Flinak

Static corrosion tests were performed in molten salts, LiF–BeF₂ (Flibe) and LiF–NaF–KF (Flinak), at 500 °C and 600 °C for 1000 h. The purpose is to investigate the corrosion characteristics of reduced activation ferritic steels, JLF-1 (8.92Cr–2W) in the fluids. The concentration of hydrogen fluoride (HF) in the fluids was measured by slurry pH titration method before and after the exposure. The HF concentration determined the fluoridation potential. The corrosion was mainly caused by dissolution of Fe and Cr into the fluids due to fluoridation and/or electrochemical corrosion. Carbon on the surface might be dissolved into the fluids due to the corrosion, and this resulted to the decrease of carbide on the surface. The corrosion depth of the JLF-1 specimen, which was obtained from the weight losses, was 0.637 μm in Flibe at 600 °C and 6.73 μm in Flinak at 600 °C.     

Irradiation behavior of atomized U–10wt.% Mo alloy aluminum matrix dispersion fuel meat at low temperature

In order to examine the in-reactor behavior of very-high-density dispersion fuels for high flux performance research reactors, U–10wt.% Mo alloy dispersions in an aluminum matrix have been irradiated at low temperature in the Advanced Test Reactor (ATR). The alloy fuel dispersant was produced by a centrifugal atomization process. The fuel shows stable in-reactor irradiation behavior to a fission density of 5×10²⁷ m⁻³ at an irradiation temperature of 65 °C. The fuel–matrix interaction layer growth rate is similar to that observed in uranium-silicide fuels. The fuel particles have a fine and a relatively narrow fission gas bubble size distribution. There appears to be features in the microstrucure that are the result of segregation of the microstructure in to molybdenum rich and depleted regions on solidification.     

Behavior of mechanical properties of nickel-alloyed reactor pressure vessel steel under neutron irradiation and post-irradiation annealing

The effect of neutron irradiation and post-irradiation thermal annealing on tensile and impact properties of Cr–Ni–Mo steel used for WWER-1000 reactor pressure vessel (RPV) manufacturing was studied. A gap in yield stress and ultimate tensile stress fluence dependence at the fluence range of 0–3×10²³ neutrons m⁻² was observed while ductile-to-brittle transition temperature (DBTT) was continuously increasing with damage dose. The post-irradiation annealing recovery of tensile properties was found to be higher than the one of impact properties. Over-recovery of tensile properties due to 460 and 490°C post-irradiation annealings were observed. The annealing effectiveness of WWER-440 and WWER-1000 grades was compared. Nickel was supposed to affect both the radiation sensitivity and the post-irradiation residual DBTT shift of WWER-1000 type steel.     

Power up-grading study for the first Egyptian research reactor

In the present work, power up-grading study is performed, for the first Egyptian Research Reactor (ET-RR-1), using the present fuel basket with 4×4 fuel rods, (17.5 mm pitch), and a proposed fuel basket with 5×5 fuel rods, (14.0 mm pitch), without violating the thermal hydraulic safety criteria. These safety criteria are; fuel centerline temperature (fuel melting), clad surface temperature (surface boiling), outlet coolant temperature, and maximum heat flux (critical heat flux ratio). Different thermal reactor powers (2–10 MW) and different core coolant flow rates (450, 900, 1350 m³ h⁻¹) are considered. The thermal hydraulic analysis was performed using the subchannel code COBRA-IIIC for the estimation of temperatures, coolant velocities and critical heat flux. The neutronic calculations were performed using WIMS-D4 code with 5 — group neutron cross section library. These cross sections were adapted to use in the two-dimensional (2-D) diffusion code DIXY for core calculations. The study concluded that ET-RR-1 power can be upgraded safely up to 4 MW with the present 4×4-fuel basket and with the proposed 5×5-fuel basket up to 5 MW with the present coolant flow rate (900 m³ h⁻¹). With the two fuel arrays, the reactor power can be upgraded to 6 MW with coolant flow rate of 1350 m³ h⁻¹ without violating the safety criterion. It is also concluded that, loading the ET-RR-1 core with the proposed fuel basket (5×5) increases the excess reactivity of the reactor core than the present 4×4 fuel matrix with equal U-235 mass load and gave better fuel economy of fuel utilization.     

Fluence dependence of the surface roughness of InP after bombardment

InP(1 0 0) surfaces were sputtered under ultrahigh vacuum conditions by 5 keV ions at an angle of incidence of 41° to the sample normal. The fluence, , used in this study, varied from 1 × 10¹⁴ to 5 × 10¹⁸ cm⁻². The surface topography was investigated using field emission scanning electron microscopy (FE-SEM) and atomic force microscopy (AFM). At the lower fluences ( 5 × 10¹⁶ cm⁻²) only conelike features appeared, similar in shape as was found for noble gas ion bombardment of InP. At the higher fluences, ripples also appeared on the surface. The bombardment-induced topography was quantified using the rms roughness. This parameter showed a linear relationship with the logarithm of the fluence. A model is presented to explain this relationship. The ripple wavelength was also determined using a Fourier transform method. These measurements as a function of fluence do not agree with the predictions of the Bradley–Harper theory.     

Corrosion of stainless steels in lead-bismuth eutectic up to 600 °C

An experimental program has been carried out to understand the differences in the corrosion behaviour between different stainless steels: the austenitic steels 304L and 316L, the martensitic steels F82Hmod, T91 and EM10, and the low alloy steel P22. The influence of oxygen level in Pb-Bi, temperature and exposure time is studied. At 600 °C, the martensitic steels and the P22 steel exhibit thick oxide scales that grow with time, following a linear law for the wet environment and a parabolic law for the dry one. The austenitic stainless steels show a better corrosion behaviour, especially AISI 304L. Under reducing conditions, the steels exhibit dissolution, more severe for the austenitic stainless steels. At 450 °C, all the materials show an acceptable behaviour provided a sufficient oxygen level in the Pb-Bi. At reducing conditions, the martensitic steels and the P22 steel have a good corrosion resistance, while the austenitic steels exhibit already dissolution at the longer exposures.     

Corrosion and deposition of ferrous alloys in molten lead-bismuth

This paper reports corrosion and deposition data from tests carried out with liquid eutectic lead-bismuth (Pb-55 at.% Bi) filled steel tubes (austenitic and martensitic) under a thermal gradient (500-280 °C) for 3000 h. For the austenitic steel, the surface exposed to Pb-55Bi exhibited a ferritic corrosion layer depleted in nickel and chromium at temperature above 450 °C. In the temperature range 450-360 °C, deposits composed of iron and chromium were found. There is a temperature effect on composition with a change from iron-rich to chromium-rich with decreasing temperature. For the martensitic steel, a corrosion without corrosion layer was observed above 480 °C. Only one type of deposit consisting of 98Fe-2Cr was found in the 400-480 °C temperature range.     

The effect of radiation on the anaerobic corrosion of steel

To ensure the safe encapsulation of spent nuclear fuel elements for geological disposal, SKB of Sweden are considering using a canister, which consists of an outer copper canister and a cast iron insert. Previous work has investigated the rate of gas generation due to the anaerobic corrosion of ferrous materials over a range of conditions. This paper examines the effect of radiation on the corrosion of steel in repository environments. Tests were carried out at two temperatures (30 °C and 50 °C), two dose rates (11 Gray h⁻¹ and 300 Gray h⁻¹) and in two different artificial groundwaters, for exposure periods of several months. Radiation was found to enhance the corrosion rate at both dose rates but the greatest enhancement occurred at the higher dose rate. The corrosion products were predominantly magnetite, with some indications of unidentified higher oxidation state corrosion products being formed at the higher dose rates.     

Vapour void fraction in subcooled flow boiling

Experimental data on steam void fraction and axial temperature distribution in an annular boiling channel for low mass-flux forced and natural circulation flow of water with inlet subcooling have been obtained. The ranges of variables covered are: mass flux = 1.4 × 10⁴−1.0 × 10⁵ kg/hr m²; heat flux = 4.5 × 10³−7.5 × 10⁴ kcal/hr m²; and inlet subcooling = 10–70°C. The present and literature data match well with the theoretical void predictions using a four-step method similar to that suggested by Zuber and co-workers.     

Irradiation damage induced on polyethylene terephtalate by 1.6 MeV deuteron ions

The irradiation damage caused on polyethylene terephtalate (Mylar, PET) samples by 1.6 MeV deuteron ions has been measured using simultaneously the nuclear reaction analysis (NRA) and the transmission energy loss (TEL) techniques. The irradiation was carried out at normal incidence relative to the target surface with the irradiation beam being used as the analysis beam. The evolution of the overall damage during irradiation was evaluated by measuring the variation of the energy loss of the deuteron beam passing through the target. For this purpose, a solid state Si detector placed at a forward angle of 30° relative to the incident beam direction was used. The NRA spectra recorded by a second Si detector located backward at 150° allowed the evaluation of the carbon and the oxygen depletion. The beam spot size was circular in shape and 1 mm in diameter and the beam current was set at 5 nA. The ion fluence was increased up to the value of 2.5 × 10¹⁶ deuterons/cm². It was observed that the target energy loss decreased steadily as the fluence increased and levelled off at high fluence. The ¹⁶O(d,p₀)¹⁷O, ¹⁶O(d,p₁)¹⁷O^* and ¹²C(d,p₀)¹³C reactions were used for monitoring the evolution of the oxygen and carbon content as a function of the deuteron fluence. A monotonic decrease of the oxygen content with the increase of ion fluence was observed. At the highest fluence the oxygen depletion reached a value of about 75%. For carbon, a weak depletion was observed at fluence ranging from 2.5 × 10¹⁵ d/cm² to 1.0 × 10¹⁶ d/cm² followed by a levelling-off with a total loss around 20%.     

Results of steel corrosion tests in flowing liquid Pb/Bi at 420-600 °C after 2000 h

Corrosion tests were carried out on austenitic AISI 316L and 1.4970 steels and on MANET steel up to 2000 h of exposure to flowing (up to 2 m/s) Pb/Bi. The concentration of oxygen in the liquid alloy was controlled at 10⁻⁶ wt%. Specimens consisted of tube and rod sections in original state and after alloying of Al into the surface. After 2000 h of exposure at 420 and 550 °C the specimen surfaces were covered with an intact oxide layer which provided a good protection against corrosion attack of the liquid Pb/Bi alloy. After the same time corrosion attack at 600 °C was severe at the original AISI 316L steel specimens. The alloyed specimens containing FeAl on the surface of the alloyed layer still maintained an intact oxide layer with good corrosion protection up to 600 °C.     

Effect of irradiation by 140 Mev Ag ions on the optical and electrical properties of polypropylene/TiO2 composite

Changes in the optical, structural, dielectric properties and surface morphology of a polypropylene/TiO₂ composite due to swift heavy ion irradiation were studied by means of UV–visible spectroscopy, X-ray diffraction, impedance gain phase analyzer and atomic force microscopy. Samples were irradiated with 140 MeV Ag¹¹⁺ ions at fluences of 1 × 10¹¹ and 5 × 10¹² ions/cm². UV–visible absorption analysis reveals a decrease in optical direct band gap from 2.62 to 2.42 eV after a fluence of 5 × 10¹² ions/cm². X-ray diffractograms show an increase in crystallinity of the composite due to irradiation. The dielectric constants obey the Universal law given by ε α f ⁿ⁻¹, where n varies from 0.38 to 0.91. The dielectric constant and loss are observed to change significantly due to irradiation. Cole–cole diagrams have shown the frequency dependence of the complex impedance at different fluences. The average surface roughness of the composite decreases upon irradiation.     

Compatibility of FBR structural materials with supercritical carbon dioxide

A key problem in the application of a supercritical carbon dioxide (CO₂) turbine cycle to a fast breeder reactor is the corrosion of structural materials brought about by supercritical CO₂ at high temperatures. In this study, long-term (8000 h) compatibility tests on candidate materials, two high-chromium martensitic steels (12Cr- and 9Cr-steels) and an austenitic stainless steel (316FR), were performed at 400-600 °C in supercritical CO₂ pressurized at 20 MPa, and corrosion allowances for the steels were proposed for application to preliminary reactor design.Although high temperature oxidation was measured in all steels, the behavior differed greatly. For martensitic steels, weight gain exhibited parabolic growth as exposure time increased at each temperature. Neither exfoliation of the oxide nor the breakage was observed during the 8000 h of exposure. The corrosion behavior was equivalent to that seen in supercritical CO₂ at 10 MPa, and it was confirmed that no effects of CO₂ pressure were present under the CO₂ turbine cycle operation conditions. Based on the results, corrosion allowances for temperature-dependant parabolic growth were proposed. For 316FR steel, weight gain was significantly lower than that of martensitic steels, with a maximum value of 6.2 g/m² at 600 °C for 8000 h. Since no dependency of temperature and immersion time on weight gain such as the martensitic steels was noted, corrosion allowances proportional to time was proposed. Estimated corrosion allowances for the martensitic and austenitic steels were 380 μm and 220 μm, respectively, for reactors, whose design life is rated at 60 years.     

Corrosion experiments and materials developed for the Japanese HLM systems

The static corrosion tests in lead-bismuth eutectic (LBE) were conducted from 450 °C to 600 °C to understand corrosion behavior and develop corrosion resistant materials for heavy liquid metal systems. While increase of Cr content in steels enhances corrosion resistance in LBE, the effect approaches a constant value above 12 wt% of Cr. Corrosion depth in LBE increases with increasing temperature and corrosion attack becomes severe above 550 °C even under the condition of high oxygen concentration. Nickel dissolution and Pb-Bi penetration occur in 316SS and JPCA above 550 °C under the condition of high oxygen concentration. When oxygen concentration decreases below the level of Fe oxide formation, corrosion attack on these steels also becomes violent due to dissolution of various elements and grain boundary corrosion. Whereas additions of 1.5 wt% Si to T91 and 2.5 wt% Si to 316SS improve corrosion resistance, the effect is insufficient taking fluctuation of oxygen concentration in LBE into consideration. Furthermore, addition of 1.5 wt% Si to T91 causes rise in DBTT. A new coating method using Al, Ti and Fe powders produces corrosion resistant coating layers on 316SS. The coating layers containing 6-8 wt% Al exhibit good corrosion resistance at 550 °C for 3000 h in LBE containing 10⁻⁶-10⁻⁴ wt% of oxygen.     

Dissolution of uranium metal without hydride formation or hydrogen gas generation

This study shows that metallic uranium will cleanly dissolve in carbonate-peroxide solution without generation of hydrogen gas or uranium hydride. Metallic uranium shot, 0.5–1 mm diameter, was reacted with ammonium carbonate–hydrogen peroxide solutions ranging in concentration from 0.13 M to 1.0 M carbonate and 0.50 M to 2.0 M peroxide. The dissolution rate was calculated from the reduction in bead mass, and independently by uranium analysis of the solution. The calculated dissolution rate ranged from about 4 × 10⁻³ to 8 × 10⁻³ mm/h, dependent primarily on the peroxide concentration. Hydrogen analysis of the etched beads showed that no detectable hydrogen was introduced into the uranium metal by the etching process.