Friction and wear of Inconel 690 and Inconel 600 for steam generator tube in room temperature water

In the steam generators of nuclear power plants, the flow of cooling water can cause the tubes to vibrate, resulting in fretting wear damage due to contacts between these tubes and their supports. The tubes are made of Inconel 690 and Inconel 600 and the supports are made of STS 304. In this paper, fretting wear tests in water were performed using the materials Inconel 690 and Inconel 600 in contact with STS 304. Fretting tests using a cross-cylinder type set up were conducted under various vibrating amplitudes and applied normal loads in order to measure friction forces and wear volumes. Also, conventional sliding tests using a pin-on-disk type set up were carried out to compare these test results.In the fretting tests, friction force was found to be strongly dependent on normal load and vibrating amplitude. Coefficients of friction decreased with an increase in the normal load and a decrease in the vibrating amplitude applied. Also, the wear of Inconel 600 and Inconel 690 was predicted using a work rate model. Depending on the normal load and vibrating amplitude applied, distinctively different wear mechanisms and often drastically different wear rates occurred. It was found that the fretting wear coefficients for Inconel 600 and Inconel 690 were 9.3×10⁻¹⁵ and 16.2×10⁻¹⁵ Pa⁻¹, respectively. This study shows that Inconel 690 can result in lesser friction forces and exhibits less wear resistance than Inconel 600 in room temperature water.     

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.     

Fission product release and its environment impact for normal reactor operations and for relevant accidents

The radioactive concentration in the primary loop and the radioactive release for both normal operations and accidents for the HTR-10 are calculated and presented in the paper. The coated-particle fuel is used in the HTR-10, which has good performance of retaining fission products. Therefore the radioactive concentration in the primary loop of the HTR-10 is very low, and the amount of radioactive release to the environment is also very small for both normal operation and accident conditions. The radiation doses to the public caused by radioactive release for both normal operations and accidents are given in the paper. The results show that the maximum individual effective dose to the public due to the release of airborne radioactivity during normal operations is only 1.4×10⁻⁴ mSv a⁻¹, which is much lower than the dose limit (1 mSv a⁻¹) stipulated by Chinese National Standard GB8703-86. For depressurization accident and water ingress accident, the maximum individual whole-body doses to man are only 7.7×10⁻² and 2.0×10⁻¹ mSv, thyroid doses only 1.7×10⁻¹ and 1.1 mSv, respectively. They are much lower than the prescribed minimum of emergency intervention level (whole-body dose: 5 mSv, thyroid dose: 50 mSv) for sheltering measures stipulated by the Chinese Nuclear Safety Criterion HAD002/03. The conclusion is that the environmental impact is very small for normal operations and accidents for the HTR-10, and the requirements stipulated in the Chinese Nuclear Safety Criterions are satisfied perfectly.     

HTGR core auxiliary cooling system availability and reliability

Using fault tree techniques, a quantitative estimate is made to predict both the start-up availability and operational reliability of the core auxiliary cooling system (CACS) of an HTGR following the postulated, simultaneous occurrence of a design basis depressurization accident (DBDA) and the complete loss of main loop cooling (LOMLC). The effects of a postulated, concurrent loss of offsite power are considered as well. Several potential common mode failures are identified. The limited availability of data presents a problem to numerical evaluation and estimates of uncertainty are at best crude. To provide a basis for measure of this uncertainty, the fault trees were solved using, on a consistent basis, either ‘optimistic’ failure rates, ‘pessimistic’ failure rates, or mean values (the geometric mean).Generally, about 80% of the failure rate data was larger than the ‘optimistic’ value, while only 20% was larger than the ‘pessimistic’ value. Predicated on a variety of assumptions, many of which may be unduly pessimistic, the CACS unavailability following a postulated DBDA and LOMLC has been estimated to be between 4 × 10⁻⁷ and 3 × 10⁻⁵ for the 2000 MW (th) HTGR and between 5 × 15⁻⁷ and 5 × 10⁻⁵ for the 3000 MW (th) HTGR. At the end of 300 hr, the estimated probability that the CACS will not leave sufficient core cooling capacity varies between 9 × 10⁻⁵ and 4 × 10⁻² for the smaller plant and 3 × 10⁻⁴ and 6 × 10⁻² for the larger plant. If it is further postulated that offsite power is concurrently lost, then the estimated mean unavailability at start-up is 3 × 10⁻³ for the 2000 MW (th) plant. The estimated mean probability that the CACS of the smaller plant will not be available at start-up and not be operational after 300 hr is 8 × 10⁻².     

Manufacture and characteristics of spherical fuel elements for the HTR-10

The R&D of spherical fuel elements for the 10 MW high temperature gas-cooled reactor (HTR-10) started in 1986 in China. A process known as cold quasi-isostatic molding was used for manufacturing spherical fuel elements, and about 20,540 spherical fuel elements were produced in 2000 and 2001. Fabrication technology and graphite matrix materials were investigated and optimized. Cold properties of the spherical fuel elements met the design specifications. The mean free uranium fraction of 44 batches was 4.57 × 10⁻⁵. In-pile irradiation test results showed that irradiation did not lead to apparent change in linear dimensional, geometrical density, porosity and strength of matrix graphite samples. No cracks and blisters were observed in spherical fuel elements. This indicated that matrix graphite and spherical fuel elements of HTR-10 met the requirement of design specifications.     

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.     

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.     

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.     

Temperature effect on IG-11 graphite wear performance

IG-11 graphite, used in the 10 MW high temperature gas-cooled test reactor (HTR-10), was tested under different temperatures on an SRV standard wear performance tester. The experiment temperatures were room temperature, 100, 200, 300 and 400 °C. According to the reactor structure, the experiments were designed to test graphite–graphite and graphite–stainless steel wear. The wear debris was collected, and the worn surfaces and debris were observed under scanning electronic microscope (SEM). It was found that there were different wear mechanisms at different temperatures. The main wear mechanism at room temperature was abrasive wear; at 200 °C, it was fatigue wear; at 400 °C, adhesive wear was observed. This difference was mainly due to the change of stress distribution at the contact area. The distribution of wear debris was also analyzed by EDX particle analysis software.     

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.     

An evaluation of containment inerting and air dilution systems as methods for post-accident hydrogen control in BWRS

Hydrogen control is important in post-accident situations because of possibilities for containment rupture due to hydrogen deflagration or detonation. Post-accident hydrogen generation in BWR containments is analyzed as a function of engineered hydrogen control system, assumed either nitrogen inerting or air dilution. Fault tree analysis was applied to assess the failure probability per demand of each system. These failure rates were then combined with the probability of accidents producing various hydrogen generation rates to calculate the overall system hydrogen control probability. Results indicate that both systems render approximately the same overall hydrogen control failure rate (air dilution: 8.3 × 10⁻²−1.1 × 10⁻²; nitrogen inerting: 1.3 × 10⁻²−2 × 10⁻³). Drywell entries and unscheduled shutdowns were also analyzed to determine the impact on the total BWR accident risk as it relates to the decay heat removal system. Results indicate that inerting may increase the overall risk due to a possible increase in the number of unscheduled shutdowns due to a lessened operator ability to correct and identify ‘unidentified’ leakage from the primary coolant system. Further, possible benefits of inerting due to reduced torus corrosion and fire risk in containment appear to be dominated by the possible operations-related disadvantages.     

SCALE calculation for evaluation of spent fuel condition during long-term storage

Experiences with an advanced spent nuclear fuel management in Slovakia are presented in this paper. The evaluation and monitoring procedures are based on practices at the Slovak wet interim spent fuel storage facility in NPP Jaslovské Bohunice. Since 1999, leak testing of WWER-440 fuel assemblies were completed using a special leak tightness detection system developed by Framatome-anp, “Sipping in Pool”. This system utilized external heating for the precise defects determination.Optimal methods for spent fuel disposal and monitoring were designed. A new conservative factor for specifying of spent fuel leak tightness is introduced in the paper. Limit values of leak tightness were established from the combination of SCALE4.4a (ORIGEN-ARP) calculations and measurements from the “Sipping in Pool” system. These limit values are: limiting fuel cladding leak tightness coefficient for tight fuel assembly – k_FCT(T) = 3 × 10⁻¹⁰, limiting fuel cladding leak tightness coefficient for fuel assembly with leakage – k_FCT(L) = 8 × 10⁻⁷.     

A design-phase PSA of a nuclear-powered hydrogen plant

A probabilistic safety assessment (PSA) is being developed for a steam-methane reforming hydrogen production plant linked to a high-temperature gas-cooled nuclear reactor (HTGR). This work is based on the Japan Atomic Energy Research Institute's (JAERI) High Temperature Engineering Test Reactor (HTTR) prototype in Japan. The objective of this paper is to show how the PSA can be used for improving the design of the coupled plants. A simplified HAZOP study was performed to identify initiating events, based on existing studies. The results of the PSA show that the average frequency of an accident at this complex that could affect the population is 7 × 10⁻⁸ year⁻¹ which is divided into the various end states. The dominant sequences are those that result in a methane explosion and occur with a frequency of 6.5 × 10⁻⁸ year⁻¹, while the other sequences are much less frequent. The health risk presents itself if there are people in the vicinity who could be affected by the explosion. This analysis also demonstrates that an accident in one of the plants has little effect on the other. This is true given the design base distance between the plants, the fact that the reactor is underground, as well as other safety characteristics of the HTGR.     

Wear and corrosion resistance of AZ31 magnesium alloy irradiated by high-intensity pulsed ion beam

The wear and corrosion resistance of AZ31 magnesium alloy irradiated by high-intensity pulsed ion beam (HIPIB) at an ion current density of 100–300 A/cm² with shot number of 1–10 are investigated by sliding wear test and potentiodynamic polarization measurement. The surface and cross-sectional morphologies, phase structure and surface microhardness of the irradiated AZ31 magnesium alloy samples are characterized by scanning electron microscopy (SEM), optical microscopy, X-ray diffraction (XRD) and Vickers tester, respectively. The HIPIB irradiation produces the hardened surface layers and improves abrasive wear resistance of all the samples. The wear volume of the irradiated samples at 200 A/cm² and 300 A/cm² with 10 shots as well as 100 A/cm² with 5 shots is about four times less than that of the original sample. The apparent increase in corrosion resistance is achieved for all the irradiated samples in 0.01 mol/l NaCl solution with a pH value of 12. The corrosion potential and pitting breakdown potential for the samples irradiated at 100 A/cm² with 5 shots are 560 and 630 mV higher than those of the original sample, −1560 mV and −1300 mV (SCE), respectively. It is found that the combined improvement in wear and corrosion resistance of AZ31 magnesium alloy is achieved by HIPIB irradiation, which is ascribed to the microstructural refinement and the chemical homogeneity of the irradiated magnesium alloy.     

Reactivity feedback coefficients of a material test research reactor fueled with high-density U3Si2 dispersion fuels

The reactivity feedback coefficients of a material test research reactor fueled with high-density U₃Si₂ dispersion fuels were calculated. For this purpose, the low-density LEU fuel of an MTR was replaced with high-density U₃Si₂ LEU fuels currently being developed under the RERTR program. Calculations were carried out to find the fuel temperature reactivity coefficient, moderator temperature reactivity coefficient and moderator density reactivity coefficient. Nuclear reactor analysis codes including WIMS-D4 and CITATION were employed to carry out these calculations. It is observed that the average values of fuel temperature reactivity feedback coefficient, moderator temperature reactivity coefficient and moderator density reactivity coefficient from 20 °C to 100 °C, at the beginning of life, followed the relationships (in units of Δk/k × 10⁻⁵ K⁻¹) −2.116 − 0.118 ρ_U, 0.713 − 37.309/ρ_U and −12.765 − 34.309/ρ_U, respectively for 4.0 ≤ ρ_U (g/cm³) ≤ 6.0.     

Wall temperature measurement and heat transfer correlation of turbulent supercritical carbon dioxide flow in vertical circular/non-circular tubes

Experimental data are presented which describe heat transfer characteristics of turbulent supercritical carbon dioxide flow in vertical tubes with circular, triangular, and square cross-sections. Experiments are conducted at a constant pressure of 8 MPa under various conditions such as inlet bulk temperatures ranging from 15 to 32 °C, imposed heat fluxes from 3 to 180 kW/m², and mass velocities from 209 to 1230 kg/m² s. The corresponding Reynolds and Grashof numbers are in the range of 3 × 10⁴ to 1.4 × 10⁵, and 5 × 10⁹ to 4 × 10¹¹, respectively. The test section is composed of an entrance region of 0.6 m long and a heating region of 1.2 m long. Wall temperatures are measured by thermocouples installed at the outer surface of the heating region. In order to identify the effect of the cross-sectional shape on the supercritical heat transfer, wall temperature distributions in the streamwise direction are compared at the same heat flux and mass velocity conditions. Based on the wall temperature data, an improved heat transfer correlation, which can be applicable to both forced convection and mixed convection regimes, is proposed, and compared with previous ones.     

JR curves of the low alloy steel 20 MnMoNi 5 5 with two different sulphur contents in oxygen-containing high temperature water at 240°C

J_R curves of the low alloy steel 20 MnMoNi 5 5 with two different sulphur contents (0.003 and 0.011 wt.%) were determined at 240°C in oxygen-containing high temperature water as well as in air. The tests were performed by the single-specimen unloading compliance technique at load line displacement rates from 1 × 10⁻⁴ down to 1 × 10⁻⁶ mm s⁻¹ on 20% side-grooved 2T CT specimens in an autoclave testing facility at an oxygen content of 8 ppm and a pressure of 7 MPa under quasi-stagnant flow conditions.In the case of testing in high temperature water, remarkably lower J_R curves than in air at the same load line displacement rate (1 × 10⁻⁴ mm s⁻¹) were obtained. A decrease in the load line displacement rate as well as an increase in the sulphur content of the steel caused a reduction of the J_R curves. At the fastest load line displacement rate a stretch zone could be detected fractographically on the specimens tested in air and in high temperature water and consequently J_i could be determined. When testing in high temperature water, the J_i value of the higher sulphur material type decreases from 45 N mm⁻¹ in air to 3 N mm⁻¹, much more than that of the optimized material type from 51 N mm⁻¹ in air to 20 N mm⁻¹ at 1 × 10⁻⁴ mm s⁻¹.     

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.     

Ductile–brittle transition of polycrystalline iron and iron–chromium alloys

Fracture toughness of polycrystalline Fe, Fe–3%Cr and Fe–9%Cr was measured by four-point bending of pre-cracked specimens at temperatures between 77 K and 150 K and strain rates between 4.46 × 10⁻⁴ and 2.23 × 10⁻² s⁻¹. For all materials, fracture behaviour changed with increasing temperature from brittle to ductile at a distinct brittle–ductile transition temperature (T_c), which increased with increasing strain rate. At low strain rates, an Arrhenius relation was found between T_c and strain rate in each material. At high strain rates, T_c was at slightly higher values than those expected from extrapolation of the Arrhenius relation from lower strain rates. This shift of T_c was associated with twinning near the crack tip. For each material, use of an Arrhenius relation for tests at strain rates at which specimens showed twinning gave the same activation energy as for the low strain rate tests. The values of activation energy for the brittle–ductile transition of polycrystalline Fe, Fe–3%Cr and Fe–9%Cr were found to be 0.21, 0.15 and 0.10 eV, respectively, indicating that the activation energy for dislocation glide decreases with increasing chromium concentration in iron.     

Consequence evaluation for pressurized thermal shock

In order to estimate the risk associated with Pressurized Thermal Shock (PTS), a sample calculation of the core melt frequency and offsite consequences has been performed for Oconee Unit 1, a Babcock and Wilcox pressurized water reactor located in the United States. Core melt frequency was derived from through-wall-crack frequency estimates based on thermal-hydraulic and fracture mechanics analyses performed by Oak Ridge National Laboratory and Pacific Northwest Laboratory. The mode and timing of containment response was estimated from previous risk studies for Oconee Unit 3 and other plants with large dry containments.The core melt frequency was calculated to be 6 × 10⁻⁶ per reactor year for operation at the PTS screening criterion. Operation of redundant and independent containment heat removal systems results in low probability of containment failure. The risk dominant scenario involves overpressure failure of containment due to failure of containment heat removal. Prompt containment failure was assigned a very low probability (10⁻⁴), and hydrogen burn failure was not considered.The central estimate of annual risk was 5 × 10⁻⁷ early fatalities, 2 × 10⁻⁴ latent cancer fatalities and 0.7 person-rem. These values are minimal compared with other severe accident scenarios.Uncertainties and sensitivies to important parameters are discussed. The response of other types of plants is briefly described.