Effect of temperature on graphite oxidation behavior

The temperature dependence of oxidation behavior for the graphite IG-11, used in the HTR-10, was investigated by thermogravimetric analysis in the temperature range of 400–1200 °C. The oxidant was dry air (water content <2 ppm) with a flow rate of 20 ml/min. The oxidation time was 4 h. The oxidation results exhibited three regimes: in the 400–600 °C range, the activation energy was 158.56 kJ/mol and oxidation was controlled by chemical reaction; in the 600–800 °C range, the activation energy was 72.01 kJ/mol and oxidation kinetics were controlled by in-pore diffusion; when the temperature was over 800 °C, the activation energy was very low and oxidation was controlled by the boundary layer. Due to CO production, the oxidation rate increased at high temperatures. The effect of burn-off on activation energy was also investigated. In the 600–800 °C range, the activation energy decreased with burn-off. Results of low temperature tests were very dispersible because the oxidation behavior at low temperatures is sensitive to inhomogeneous distribution of any impurity, and some impurities can catalyse graphite oxidation.     

The effects of heat treatments on the microstructure and mechanical properties of the Ti–2.19Al–2.35Zr alloy

Ti–2.19Al–2.35Zr alloy is one of the candidate materials for the steam generator tubing of an integrated reactor, System Modular Integrated Advanced ReacTor (SMART) being developed in Korea. In this study, the effects of heat treatments on the mechanical properties of Ti–2.19Al–2.35Zr alloy were evaluated. Mechanical tests were implemented to examine the effects of an annealing, cooling rate and re-annealing temperature/time on the mechanical properties of the alloy. The annealing temperatures ranged from 600 to 1050 °C and the cooling rates were controlled by introducing a water-quenching (WQ), air-cooling (AC), and furnace-cooling (FC). As for the re-annealing heat treatment, after a β water quenching, the re-annealing temperature was selected as 800 °C for the α-phase heat treatment and 940 °C for the α + β-phase heat treatment with various time intervals (1, 10 and 24 h). The results showed that an increase of the annealing temperature to above the β-region temperature induced an increase of the tensile strength and a decrease of the elongation in the 25 and 300 °C tests. A decrease of the cooling rates from water-quenching to a furnace-cooling revealed a decrease of the tensile strength and an increase of the elongation. Also an increase of the re-annealing time with different phase regimes exhibited a decrease of the strength and an increase of the elongation. These tendencies were more dominant in the 300 °C test rather than the room temperature test from the characteristics of the microstructures which were affected by the heat treatments.     

Fretting-wear of zirconium alloys

Fretting tests of Zircaloy fuel sheath bearing pads in contact with zirconium alloy (Zr–2.5Nb) pressure tube specimens were conducted at temperatures varying from 25 to 315 °C. The effects of motion type and amplitude, water chemistry, fuel sheath manufacturer and pressure tube surface finish were also investigated. The effect of temperature is the most significant. The pressure tube wear coefficient in the 225–286 °C for all four motions studied is considerably greater than that above 300 °C. The fretting rate for small amplitude motion representative of flow turbulence excitation is about equal at temperatures below 150 °C and above 300 °C, but is five to ten times greater in the 250–286 °C range.     

Thermal–hydraulic feasibility analysis on uprating the HTR–PM

In order to meet energy demand in China, the high temperature gas-cooled reactor–pebble-bed module (HTR–PM) is being developed. It adopts a two-zone core, in which graphite balls are loaded in the central zone and the outer part is fuel ball zone, and couple with a steam cycle. Outer diameter of the reactor core is 4.0 m and height of the core is 9.43 m. The helium inlet and outlet temperature are 250 and 750 °C, respectively. The reactor thermal power is 380 MW. Preliminary studies show that the HTR–PM is feasible technologically and economically. In order to increase the reactor thermal power of the HTR–PM, some efforts have been made. These include increasing the height of reactor core, optimizing the thickness of fuel zone and better selection of the scheme of central graphite zone, etc. Basic design concepts and thermal–hydraulic parameters of the HTR–PM are given. Measures to increase the thermal power are introduced. Thermal–hydraulic analysis results are presented. The results show that, from the viewpoint of thermal–hydraulics, it is possible to increase the reactor power.     

Wear behavior of graphite studies in an air-conditioned environment

The wear performance of graphite used in the high-temperature gas-cooled reactor (HTR-10) was researched. The wear mechanism, worn surface and wear debris were analyzed under SEM. Under test conditions, the wear rate was 2.27×10⁻⁷ g/m for surface contact, and 1×10⁻⁶ g/m level for line contact. The main wear mechanisms of graphite were groove and fatigue. The projected area of wear debris followed the logarithm normal school, giving most wear debris as a small sphere and large flake debris as only a small part.     

Moisture flow in concrete under steady state non-uniform temperature states: experimental observations and theoretical modelling

The migration of water in concrete at temperatures up to 400 °C is controlled dominantly by pore vapour pressures. The free pore volume is as-cast concrete serves as a reservoir during the migration process and plays an important role in the mitigation of high pore vapour pressures in the hottest regions.Experimental results are presented for two concretes containing limestone and basalt aggregates. They illustrate two sets of overheat conditions in reactor containment walls: (i) long-term service conditions at steady state liner temperatures in the range 105–200 °C, with and without pressure venting close to the liner; (ii) short-term transient behaviour for an accident with temperatures to 400 °C. The results show the distributions of free and bound water in walls of two thicknesses (1.55 and 3.1 m) after approximately 1.5 years from the imposition of a temperature crossfall. The vented experiments confirm significantly higher rates of drying and the ability of water to migrate towards higher temperature locations when driven by pore pressure gradients which are in opposition to the local temperature gradients.A theoretical model, based on pore pressure gradients as the driving potentials, is introduced and used to predict water migration in a concrete wall of 5 m thickness, heated at the inner face to 200 °C. It is suggested that thick walls will take many years to dry significantly, eventhough they dry simultaneously near to the liner and at the exposed cold face. Finally it is demonstrated that the theoretical model is capable of predicting this special behaviour and therefore has an advantage over diffusion-based analyses which cannot model this feature.     

Structural design of high temperature metallic components

In the high temperature engineering test reactor (HTTR), even at normal operation the service temperatures of class 1 metallic components reach temperatures above 900 °C when exposed to primary helium coolant of 950 °C. For these components, Hastelloy XR, which is the improved version of Hastelloy X, was developed and used for high temperature application.Some of the high temperature materials and their service temperatures, including Hastelloy XR, used for the class 1 and reactor internal metallic components of the HTTR are very well beyond the well-established Japanese elevated temperature structural design guideline. Moreover, at very high temperatures, where creep deformation is significant, the component design based on elastic analysis is impossible. Therefore, many research works on structural mechanics behavior were carried out to establish a high temperature structural design guideline and creep analysis methods. This paper reviews structural design of the high temperature components for the HTTR made of Hastelloy XR, 2 1/4Cr–1Mo steel, austenitic stainless steels SUS321TB and SUS316, and 1Cr–0.5Mo–V steel.     

Results of examinations of pressure vessel samples and instrument nozzles from the TMI-2 lower head

15 prism-shaped steel samples were removed from the lower head of the damaged Three Mile Island Unit 2 (TMI-2) nuclear reactor pressure vessel to assess the effects of approximately 19 tonne of molten core debris that had relocated there during the 1979 loss-of-coolant accident. Metallographic examinations of the samples revealed that inside-surface temperatures of 800–1100°C were attained during the accident, in an elliptical ‘hot spot’ with dimensions of about 1 m × 0.8 m. Tensile, creep and Charpy V-notch specimens were cut from the samples to assess the mechanical properties of the lower head material at temperatures up to the peak accident temperature. These properties were used in a margin-to-failure analysis of the lower head. Examinations of instrument nozzles removed from the lower head region assisted in defining the relocation scenario of the molten core debris and showed that the lower head was largely protected from catastrophic failure by a solidified layer below the molten core debris that acted as a partial thermal insulator.     

The effect of temperature and oxygen content on the flowing liquid metal corrosion of structural steels in the Pb–Bi eutectic

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, 450 and 400 °C and a temperature gradient of 100 °C. These experiments were performed in liquid Pb–Bi with different oxygen concentrations (from approximately 1 × 10⁻⁶ to 2 × 10⁻⁵ wt.%) to ascertain at what oxygen concentration and up to what temperature the oxygen technology can create protective oxide or spinel layers to reduce or prevent corrosion. The results showed that the structural materials contemplated for building an ADS system, including 9% Cr–1% Mo (W) martensitic steels and similar steels with a higher Si content (2–3%), can be used with their surface unpassivated at up to 450 °C and suffer only minimal corrosion (up to 5 μm/year). At higher temperatures, their surface must be passivated prior to and regularly during the operation; however, no technology to perform such passivation in the presence of Pb–Bi is known that this time. In addition, we measured the impact of various alloying elements, such as Fe, Cr, Ni, Mn, Si, Al and Mo, on the corrosion of such steels and searched for potential ways to passivate their surface or create protective oxide or spinel layers during operation by varying the amount of oxygen in liquid Pb–Bi.     

Development of constitutive equations for computer analysis of stainless steel components

Engineering application makes conflicting demands of constitutive equations which are difficult to satisfy simultaneously, so forcing considerable approximation. The difficulty is compounded by the frequent need, in anything but room temperature application, to be able to describe the behaviour of the structural material over a range of temperatures. This is illustrated by considering the spectrum of behaviour of Type 316 stainless steel from room temperature to operation at 600°C. It is found that a simple plasticity model describes the behaviour well at 400°C but is less adequate at 20°C in the presence of “cold creep”. There is a discussion of the way plasticity and creep can both be described, with a systematic interaction but without the restrictions of a single “unified relation” for all inelastic deformation.     

Assessment of irradiation temperature stability of the first irradiation test rig in the HTTR

The High Temperature Engineering Test Reactor (HTTR) can provide very large irradiation spaces at high temperatures for various irradiation tests. The first irradiation test rig for the HTTR, the I–I type irradiation equipment, was developed for an in-pile creep test on a stainless steel with large standard size specimens. The equipment uses the ambient high temperature of the core for the irradiation temperature control. The target irradiation temperatures are 550 and 600 °C with the target temperature deviation of ±3 °C. In this study, the specimen temperature stability at the irradiation test was assessed by both analytical and experimental approaches. The irradiation temperature changes at transient conditions were analyzed by a finite element method (FEM) code and the temperature controllability of the equipment was examined by a mockup test. The controllability was evaluated with the measured temperature transient data at the core graphite components in the Rise-to-Power tests of the HTTR. The result indicates that the temperature control method of the I–I type irradiation equipment is effective to keep the irradiation temperature stable in the irradiation test.     

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.     

Fracture toughness of narrow-gap welded joints in the nuclear pressure vessel steel 22 NiMoCr 37

The mechanical testing of narrow-gap welded joints in 100 and 200 mm thick sections of the steel 22 NiMoCr 37 has revealed that the weld metal, and not the heat affected zone (HAZ) or the weld metal-parent metal boundary. is the critical region. This modified gas-shielded welding process operates with a very low heat input of the order of 6.500 J cm⁻¹ pass⁻¹ and the combination of small diameter welding wires and high welding speeds contributes to the excellent joint properties in the as-welded condition.To investigate the effect of preheating and post-welding heat treatment on the mechanical properties of narrow-gap welds, tensile, notch impact, flat bend and fracture toughness test specimens were extracted from joints welded with the following conditions: (1) no preheating: no post-weld heat treatment; (2) no preheating: soaking at 300°C: (3) no preheating: stress-relief heat treatment at 600°C; (4) preheating 200–250°C; no post-weld heat treatment; (5) preheating 200–250°C; soaking at 300°C; (6) preheating 200–250°C; stress relief heat treatment at 600°C. Tensile testing at room temperature and at 250°C of round specimens oriented across the seam revealed the ultimate fracture to be always located in the base material remote from the welded zone. Although pores or slag inclusions had an influence on bend-test results of specimens in the as-welded condition, the results generally show failure free bends to 180°C with no evidence of cracking in the HAZ or at the fusion boundary.Using sharp-notched impact bend specimens with the notch located in the centre of the seam as well as in and across the HAZ, absorbed energy-test temperature curves have been determined for each welding condition. In comparison with the base material impact toughness, the weld exhibits superior toughness in the temperature range − 60 – 0°C, but yielded lower values at room temperature. After stress relieving at 600°C, the impact toughness of the weld reduced significantly, apparently due to precipitations occurring in the weld-metal microstructure. Test results from welded specimens with the no notch in the HAZ show this region to have superior notch impact toughness to the base material.Crack opening displacement (COD) specimens 45 × 90 × 380 mm with the fatigue crack located in the weld metal and in the HAZ were tested at 0 and 20°C using both the recommendation in BS DD 19: 1972 as well as acoustic emission measurements for the determination of COD values. For this method of fracture toughness testing it has been shown that the occurrence of a critical event must be clearly defined as corresponding to stable crack growth or alternatively to unstable crack propagation.     

Creep-fatigue behaviour of an AISI stainless steel at 550 °C

In the past, a lot of experimental studies have been devoted to creep-fatigue interactions in austenitic stainless steels. Tests have been carried mainly at temperatures of at least 600 °C and at high applied strains, which are supposed to be the most damaging. The present work is dedicated to mechanical tests, TEM observations and lifetime predictions at 550 °C which corresponds to the real industrial temperature in Liquid Metal Fast Breeder Reactors. It is shown that if pure fatigue test results are close to those performed at 600 °C, some of the creep-fatigue results are different, particularly for small applied strains which correspond once more to the industrial conditions. In the 0.25–0.3% strain amplitude range, the stress is larger with hold time than without whatever is the hold time up to 5 h. The numbers of cycles to failure are greatly reduced and no saturation with the hold time is observed, contrary to higher temperature results. The stress–strain behaviour is discussed considering several high temperature mechanisms such as ageing, recovery and viscoplasticity and using TEM observations and stress partitioning into kinematic, isotropic and thermal stresses. Finally, a simple linear damage accumulation model is applied to the 550 °C and 600 °C tests, using the measured stresses. The stress dependence on hold time can partly explain the observed failure results on fatigue life.     

Estimation of graphite density and mechanical strength variation of VHTR during air-ingress accident

This study focuses on predicting the changes in graphite density and mechanical strength in VHTR during the air-ingress accident via thermal hydraulic system analysis code. A simple graphite burn-off model was developed based on the similarities between a parallel electrical circuit and graphite oxidation. The developed model along with other comprehensive graphite oxidation models were integrated into the VHTR system analysis code, GAMMA. GT-MHR 600 MW t reactor was selected as a reference reactor. Based on the calculation, the main oxidation process was observed 5.5 days after the accident when followed by natural convection. The core maximum temperature reached 1430 °C, but never exceeded the maximum temperature criteria, 1600 °C. However, the oxidation process did significantly decrease the density of bottom reflector, making it vulnerable to mechanical stress. The stress on the bottom reflector is greatly increased because of the reduction of loaded surface area with graphite oxidation. The calculation proceeded until 11 days after the accident, resulting in an observed 4.5% decrease in density and a 25% reduction of mechanical strength.     

Radiation-Induced Change in the Properties of Isotropic Structural Graphite

Three types of samples of isotropic graphite with different grain density and size were irradiated in a BOR-60 reactor up to neutron fluence (1.7–2.8)·10²⁶ m^–2 (E > 0.18 MeV) at 360–400°C. After irradiation, the change in the dimensions, resistivity, linear thermal expansion coefficient and dynamic elastic modulus were investigated. It was determined that the density in the range 1.67–1.76 g/cm³ results in an increase of the maximum weight and depth of volume shrinkage of isotropic fine-grain graphite. An equation was proposed for fitting the temperature dependence of the critical neutron fluence in the range 380–780°C for the experimental graphite samples.     

An evaluation of reactor cooling and coupled hydrogen production processes using the modular helium reactor

The high-temperature characteristics of the modular helium reactor (MHR) make it a strong candidate for producing hydrogen using either thermochemical or high-temperature electrolysis (HTE) processes. Using heat from the MHR to drive a sulfur-iodine (SI) thermochemical hydrogen production process has been the subject of a U.S. Department of Energy sponsored Nuclear Engineering Research Initiative (NERI) project led by General Atomics, with participation from the Idaho National Laboratory (INL) and Texas A&M University. While the focus of much of the initial work was on the SI thermochemical production of hydrogen, recent activities included development of a preconceptual design for an integral HTE hydrogen production plant driven by the process heat and electricity produced by a 600 MW MHR.This paper describes ATHENA analyses performed to evaluate alternative primary system cooling configurations for the MHR to minimize peak reactor vessel and core temperatures while achieving core helium outlet temperatures in the range of 900–1000 °C that are needed for the efficient production of hydrogen using either the SI or HTE process. The cooling schemes investigated are intended to ensure peak fuel temperatures do not exceed specified limits under normal or transient upset conditions, and that reactor vessel temperatures do not exceed American Society of Mechanical Engineers (ASME) code limits for steady-state or transient conditions using standard light water reactor vessel materials. Preconceptual designs for SI and HTE hydrogen production plants driven by one or more 600 MW MHRs at helium outlet temperatures in the range of 900–1000 °C are described and compared. An initial SAPHIRE model to evaluate the reliability, maintainability, and availability of the SI hydrogen production plant is also described. Finally, a preliminary flowsheet for a conceptual design of an HTE hydrogen production plant coupled to a 600 MW modular helium reactor is presented and discussed.     

Hydrogen influence on mechanical and fracture mechanics characteristics of zirconium Zr–2.5Nb alloy at ambient and elevated temperatures

This article deals with the investigation of the hydrogen concentration and temperature influence onto mechanical and fracture mechanics characteristics of RBMK-1500 Ignalina NPP unit 2 reactor fuel channel material—Zr–2.5Nb zirconium alloy (TMT-2) at temperatures from ambient up to 300 °C. The investigation of mechanical characteristics was performed on tensile specimens, fracture mechanics characteristics K_Q, , J_IC—on compact specimens (B = 4 mm) of hydrogen-free and saturated by hydrogen (52, 100 and 140 ppm) at 20, 170, 200 and 300 °C. The investigation showed that temperature increasing calls mechanical strength decreasing, whereas the reductions of area increase. Stronger influence of hydrogen concentration onto mechanical characteristics is noticed only at 20–170 °C temperature, however this influence diminishes as the temperature increases and weakest hydrogen influence is given at 300 °C. Fracture toughness characteristics K_Q, more depends on temperature than on hydrogen concentration. Critical J_IC integral values for the specimens containing hydrogen were given lowest at 20 °C, increases when temperature were raised up to 140 °C and were given highest when it reaches 300 °C.The analysis of and J_IC dependence due to the mechanical characteristics of zirconium alloy has showed that the modified plasticity Z_mod = (R_p0.2/R_m)Z satisfactorily approximates the influence of temperature and hydrogen concentration on variation of these characteristics.     

Strain components of nuclear-reactor-type concretes during first heat cycle

Strains of three advanced-gas-cooled-reactor-type nuclear reactor concretes were measured during the first heat cycle and their relative thermal stability determined. It was possible to isolate for the first time the shrinkage component for the period during heating. Predictions of the residual strains for the loaded specimens can be made by simple superposition of creep and shrinkage components up to a certain critical temperature, which for basalt concrete is about 500 °C and for limestone concrete is about 200–300 °C. Above the critical temperature, an expansive “cracking” strain component is present. It is shown that the strain behaviour of concrete provides a sensitive indication of its thermal stability during heating and subsequent cooling.     

Experimental study on uniaxial and nonproportionally multiaxial ratcheting of SS304 stainless steel at room and high temperatures

An experimental study was achieved for the cyclic properties of SS304 stainless steel subjected to uniaxial strain-controlled, uniaxial and nonproportionally multiaxial stress-controlled cyclic loading at room and high temperatures. The effects of cyclic strain amplitude, mean strain, temperature and their histories on the cyclic deformation behavior of the material were investigated under the uniaxial strain-controlled cyclic loading. The uniaxial and nonproportionally multiaxial ratcheting was researched under the asymmetrical stress-controlled cyclic loading with variable stress amplitudes, mean stresses, loading paths and their histories at room and high temperatures. It is shown that the uniaxial cyclic properties under strain-controlled cyclic loading and the ratcheting under asymmetric uniaxial and nonproportionally multiaxial stress-controlled cyclic loading depend not only on the current temperature and loading state, but also greatly on the previous loading history and the shape of loading path. The material presents much greater cyclic hardening and less ratcheting in the range of 400–600 °C than at room temperature, due to the strong dynamic strain aging taken place in this temperature range. Some significant results were obtained for the constitutive modeling of cyclic plastic deformation such as ratcheting.