High Temperature Transients Research Articles

Commissioning of a Preheat Strategy of a Molten Salt Test Receiver

Solar power tower plants that are operated with molten salt offer the advantage that the molten salt can be used as a heat transfer medium as well as storage medium. Consequently, solar energy can be stored easily and efficiently in only one loop without transformation losses. Furthermore, by using large-scale storage facilities, almost a 24/7 base load operation is possible. Since the molten salt begin to crystallize at temperature of 238 °C, the absorber tube must be preheated before the plants’ operation can begin. During the preheating a subset of heliostats will be aimed on the receiver surface to ensure a solar flux density that leads to tube temperatures above the solar salt crystallization point. Since the absorber tubes are empty during the preheating, there is a risk of high temperature gradients and transients and thus high thermal stresses that may lead to fatigue damage. To prevent this, a preheating strategy for a molten salt test receiver was developed and tested, taking into account ambient conditions and time. The present work shows the results of the commissioning of the developed preheating strategy and discusses its potential for improvement.

Oxidation behavior and microstructural evolution of Zr-1.0Nb-0.01Cu during high-temperature transients

To study the oxidation behavior and microstructural evolution of Zr-1.0 Nb-0.01Cu claddings during high-temperature transients simulating LOCA scenarios, samples were oxidized to 850, 900, 1000, 1100 and 1200 ℃ at heating rates of 25 and 50 ℃/min. The oxidation data show a similar trend from a parabolic to a cubic rate law. Then, the cross-sectional morphology, phase distribution, orientation, crystal structure and stress distribution were characterized. Monoclinic ZrO2 oxide shows a fiber texture. The suboxide ZrO at the ZrO2/Zr3O interface from presence to absence is a continuous process with increasing temperature. The presence of ZrO could be associated with local stresses.

A novel approach on designing ultrahigh burnup metallic TWR fuels: Upsetting the current technological limits

The grand challenge of “net-zero carbon” emission calls for technological breakthroughs in energy production. The traveling wave reactor (TWR) is designed to provide economical and safe nuclear power and solve imminent problems, including limited uranium resources and radiotoxicity burdens from back-end fuel reprocessing/disposal. However, qualification of fuels and materials for TWR remains challenging and it sets an “end of the road” mark on the route of R&D of this technology. In this article, a novel approach is proposed to maneuver reactor operations and utilize high-temperature transients to mitigate the challenges raised by envisioned TWR service environment. Annular U-50Zr fuel and oxidation dispersion strengthened (ODS) steels are proposed to be used instead of the current U-10Zr and HT-9 ferritic/martensitic steels. In addition, irradiation-accelerated transport of Mn and Cr to the cladding surface to form a protective oxide layer as a self-repairing mechanism was discovered and is believed capable of mitigating long-term corrosion. This work represents an attempt to disruptively overcome current technological limits in the TWR fuels.Impact statementAfter the Fukushima accident in 2011, the entire nuclear industry calls for a major technological breakthrough that addresses the following three fundamental issues: (1) Reducing spent nuclear fuel reprocessing demands, (2) reducing the probability of a severe accident, and (3) reducing the energy production cost per kilowatt-hour. An inherently safe and ultralong life fast neutron reactor fuel form can be such one stone that kills the three birds. In light of the recent development findings on U-50Zr fuels, we hereby propose a disruptive, conceptual metallic fuel design that can serve the following purposes at the same time: (1) Reaching ultrahigh burnup of above 40% FIMA, (2) possessing strong inherent safety features, and (3) extending current limits on fast neutron irradiation dose to be far beyond 200 dpa. We believe that this technology will be able to bring about revolutionary changes to the nuclear industry by significantly lowering the operational costs as well as improving the reactor system safety to a large extent.Graphical abstract

Novel method of ammonium nitrate quantification in SCR catalysts

NH4NO3 or Ammonium nitrate (AN) can be formed onto NH3-SCR catalysts in the low-T range typical of urban or low-load operation, resulting in self-inhibition of the catalysts. Also, subsequent decomposition of AN during high-temperature transients may lead to the undesired formation of N2O, a potent greenhouse gas. Quantification of AN formation and decomposition has been not adequately addressed so far due to experimental difficulties.We propose an original method for measuring the amount of AN formed on SCR catalysts, and validate it in bench-flow runs over a commercial Cu-CHA formulation. In order to quantify the AN on the catalyst, we rely on its reaction with NO according to NO+NH4NO3→NO2+N2+2H2O, further investigated via DRIFTs in this work. Based on this stoichiometry, the amount of deposited AN can be estimated from the integral molecular balances of reliably and precisely measurable species, such as NO and NO2. This approach is initially applied over catalyst powder in a rig where N-balances could be evaluated directly, and subsequently validated over more realistic coated monolith catalyst samples in a bench-flow reactor in which N2 was used as carrier gas. In the validation tests, known amounts of AN were initially deposited on the catalyst by controlled pre-dosing of an AN aqueous solution: such amounts were found in satisfactory agreement with the estimates obtained by our method. We expect the proposed methodology for AN quantification to be generally applicable to other types of SCR catalysts as well, including V-based systems and Fe-zeolites.

CHARACTERIZING HIGH-TEMPERATURE DEFORMATION OF INTERNALLY HEATED NUCLEAR FUEL ELEMENT SIMULATORS

The sag behaviour of a simulated nuclear fuel element during high-temperature transients has been investigated in an experiment utilizing an internal indirect heating method. The major motivation of the experiment was to improve understanding of the dominant mechanisms underlying the element thermo-mechanical response under loss-of-coolant accident conditions and to obtain accurate experimental data to support development of 3-D computational fuel element models. The experiment was conducted using an electrically heated CANDU fuel element simulator. Three consecutive thermal cycles with peak temperatures up to ≈1000 °C were applied to the element. The element sag deflections and sheath temperatures were measured. On heating up to 600 °C, only minor lateral deflections of the element were observed. Further heating to above 700 °C resulted in an element multi-rate creep and significant permanent bow. Post-test visual and X-ray examinations revealed a pronounced necking of the sheath at the pellet-to-pellet interface locations. A wall thickness reduction was detected in the necked region that is interpreted as a sheath longitudinal strain localization effect. The sheath cross-sectioning showed signs of a “hard” pellet–cladding interaction due to the applied cycles. A 3-D model of the experiment was generated using the ANSYS finite element code. As a fully coupled thermal mechanical simulation is computationally expensive, it was deemed sufficient to use the measured sheath temperatures as a boundary condition, and thus an uncoupled mechanical simulation only was conducted. The ANSYS simulation results match the experiment sag observations well up to the point at which the fuel element started cooling down.

Low temperature high density plasma nitriding of stainless steel molds for stamping of oxide glasses

Various kinds of stainless steels have been widely utilized as a die for mold- and direct-stamping processes of optical oxide glasses. Since they suffered from high temperature transients and thermal cycles in practice, they must be surface-treated by dry and wet coatings, or, by plasma nitriding. Martensitic stainless steel mold was first wet plated by the nickel phosphate (NiP), which was unstable at the high temperature stamping condition; and, was easy to crystalize or to fracture by itself. This issue of nuisance significantly lowered the productivity in fabrication of optical oxide-glass elements. In the present paper, the stainless steel mold was surface-treated by the low-temperature plasma nitriding. The nitrided layer by this surface modification had higher nitrogen solute content than 4 mass%; the maximum solid-solubility of nitrogen is usually 0.1 mass% in the equilibrium phase diagram. Owing to this solid-solution with high nitrogen concentration, the nitrided layer had high hardness over 1400 HV within its thickness of 50 μm without any formation of nitrides after plasma nitriding at 693 K for 14.4 ks. This plasma-nitrided mold was utilized for mold-stamping of two colored oxide glass plates at 833 K; these plates were successfully deformed and joined into a single glass plate by this stamping without adhesion or galling of oxide glasses onto the nitrided mold surface.

Characterization of aerosols from RDD surrogate compounds produced by fast thermal transients

Experimental tests have been performed to characterize the aerosols representative of radiological dispersion devices (RDDs, a.k.a. “dirty bombs”) by applying to chosen surrogate compound rapid high temperature transients, vaporizing the sample and forming aerosols mainly by rapid cooling of the vapour. The materials, which were tested in their non-radioactive form, had been chosen from the radioactive sources widely used in industries and nuclear medicine applications, Co, CsCl, Ir and SrTiO3. Our analyses permitted the characterization of the inhalable fraction of the aerosols released, and the study of the influence of cladding materials on the aerosol release and on its characteristics.

Structural materialization of stainless steel molds and dies by the low temperature high density plasma nitriding

Various kinds of stainless steels have been widely utilized as a mold substrate material for injection molding and as a die for mold-stamping and direct stamping processes. Since they suffered from high temperature transients and thermal cycles in practice, they must be surface-treated by dry and wet coatings, or, by plasma nitriding. Martensitic stainless steel mold was first wet plated by the nickel phosphate (NiP), which was unstable at the high temperature stamping condition; and, was easy to crystalize or to fracture by itself. This issue of nuisance significantly lowered the productivity in fabrication of optical elements at present. In the present paper, the stainless steel mold was surface-treated by the low-temperature plasma nitriding. The nitrided layer by this surface modification had higher nitrogen solute content than 4 mass%; the maximum solid-solubility of nitrogen is usually 0.1 mass% in the equilibrium phase diagram. Owing to this solid-solution with high nitrogen concentration, the nitrided layer had high hardness of 1400 Hv within its thickness of 40 μm without any formation of nitrides after 14.4 ks plasma nitriding at 693 K. This nitrogen solid-solution treated stainless steel had thermal resistivity even at the mold-stamping conditions up to 900 K.

High-temperature stability of laser-joined silicon carbide components

Silicon carbide is recommended for applications in energy technology due to its good high-temperature corrosion resistance, mechanical durability, and abrasion resistance. The prerequisite for use is often the availability of suitable technologies for joining or sealing the components. A laser-induced process using fillers and local heating of the components represents a possible low-cost option.Investigations in which yttrium aluminosilicate glass was used for laser-induced brazing of SiC components of varying geometry are presented. A four-point bending strength of 112MPa was found for these joints. In burst tests, laser-joined components were found to withstand internal pressures of up to 54MPa.Helium leak tests yielded leak rates of less than 10–8mbarls−1, even after 300h at 900°C. In contrast, the assemblies showed an increased leak rate after annealing at 1050°C.The short process time of the laser technique – in the range of a few seconds to a few minutes – results in high temperature gradients and transients. SEM analysis showed that the filler in the seam predominantly solidifies in a glassy state. Crystallization occurred during later thermal loading of the joined components, with chemical equilibrium being established. Differences in seam structures yielded from different cooling rates in the laser process could not be equalized by annealing.The results demonstrated the long-term stability of laser-brazed SiC assemblies to temperatures in the range of glass transformation (900°C) of the yttrium aluminosilicate filler. In technological investigations, the suitability of the laser joining technique for sealing of SiC components with a geometry approximating that of a fuel element sleeve pin (pin) in a gas-cooled fast reactor was proven.

Investigation of the Ru-43LV fuel behaviour under LOCA conditions in a CANDU reactor

Abstract Presently, INR Pitesti is developing an advanced fuel design RU-43LV (recovered uranium fuel bundle with 43 elements and low void reactivity feature) based on recovered uranium from LWR. Compared with the current design of 37 natural uranium element (NU-37) fuel bundle, RU-43LV will have higher power capability and higher burn-up potential in CANDU reactors of Cernavodâ-Romania Nuclear Power Plant (NPP). Fuel burn-up of RU-43LV fuel will be about two times the burn-up usually achieved in CANDU reactors fuelled with natural uranium fuel. The effect of the design changes of RU-43LV bundle on the reactor safety has been analyzed and the results are presented in this paper. As part of the conceptual design study, the performance of the RU-43LV fuelled core during a large loss-of-coolant accident (LLOCA) was assessed with the use of several computer codes. The most relevant calculations performed regarding RU-43 RV fuel safety are presented in this paper. Also, the stages of an experimental program aiming to study RU-43LV fuel behaviour in high temperature transients are briefly described.

Deformation and ultrafine dynamic recrystallization of quartz in pseudotachylyte-bearing brittle faults: A matter of a few seconds

Tectonic pseudotachylytes, i.e. quenched friction-induced silicate melts, record coseismic slip along faults and are mainly reported from the brittle crust in association with cataclasites. In this study, we document the occurrence of recrystallization of quartz to ultrafine-grained (grain size 1–2 μm) aggregates along microshear zones (50–150 μm thick) in the host rock adjacent to pseudotachylytes from two different faults within quartzite (Schneeberg Normal Fault Zone, Eastern Alps), and tonalite (Adamello fault, Southern Alps) in the brittle crust. The transition from the host quartz to microshear zone interior includes: (i) formation of high dislocation densities; (ii) fine (0.3–0.5 μm) polygonization to subgrains defined by disordered to well-ordered dislocation walls; (iii) development of a mosaic aggregate of dislocation-free new grains. The crystallographic preferred orientation (CPO) of quartz towards the microshear zone shows a progressive misorientation from the host grain, by subgrain rotation recrystallization, to a nearly random CPO possibly related to grain boundary sliding. These ultrafine aggregates appear to be typically associated with pseudotachylytes in nature. We refer the crystal plastic deformation of quartz accompanied by dramatic grain size refinement to the coseismic stages of fault slip due to high differential stress and temperature transients induced by frictional heating. Microshear zones localized on precursory fractures developed during the stages of earthquake rupture propagation and the very initial stages of fault slip. Thermal models indicate that the process of recrystallization, including recovery processes, occurred in a time lapse of a few tens of seconds.

Oxygen, Hydrogen and Main Alloying Chemical Elements Partitioning Upon Alpha←→Beta Phase Transformation in Zirconium Alloys

Due to their adequate properties, zirconium alloys are the reference materials for the nuclear fuel cladding tubes of Light Water Reactors (LWR). During some hypothetical accidental High Temperature (HT) transients, the materials should experience heavy steam oxidation and deep metallurgical evolutions. This promotes Alpha-Beta phase transformations and an associated strong partitioning of oxygen/hydrogen and of the main chemical alloying elements (Nb, Sn, Fe and Cr). Moreover, it has been shown quite recently that such chemical elements partitioning during on-cooling Beta-to-Alpha transformation can strongly impact the residual mechanical properties of HT oxidized materials. Thus, it appeared that it was important to better quantify and, if possible, to compute the quite complex phase equilibrium that occurs in multi-alloyed zirconium materials in the presence of both oxygen and hydrogen. For that, systematic studies have been performed on industrial alloys, charged with oxygen and/or hydrogen. After applying different heating/cooling scenarii, both Electron Microprobe using Wave Dispersive Spectrometry (WDS) and Nuclear Microprobe using Elastic Recoil Detection Analysis (ERDA) have been applied. Finally, to support the observed chemical elements partitioning between the Alpha and Beta allotropic phases, some thermodynamic calculations have been performed thanks to the development and the use of a specific thermodynamic database for zirconium alloys called “Zircobase".

Hydride behavior in Zircaloy cladding tube during high-temperature transients

In order to study the hydride behavior in high burnup fuel cladding during temperature transients expected in anticipated operational occurrences and accidents, unirradiated hydrided Zircaloy-4 cladding tubes were rapidly heated to temperatures ranging from 673 to 1173 K and annealed for holding time ranging from 0 to 3600 s. Hydrides were localized in the peripheral region of the cladding tubes prior to the annealing, as observed in high burnup fuel cladding. The localized hydride layer (hydride rim) was annealed out, and the radial hydride distribution became uniform after the annealing at 873 K for 600 s, 973 K for 60 s, or 1173 K for 0 s. The annealing out of the hydride rim is caused by the phase transformation from the α + δ phase to the α + β or β phase in the hydride rim and the subsequent drastic increase in the solid solubility and diffusion of hydrogen in Zircaloy. On the other hand, the radial distribution and morphology of hydrides did not change at lower temperatures: Thus, the hydride remains almost intact below the phase transformation temperature for the short time ranges.

Comparison between CAREB code calculations and LOCA test results in the FUMEX III project

Abstract The IAEA initiated a Coordinated Research Project (CRP) on improvement of computer codes used for fuel behaviour simulation under the name: FUMEX III. The Institute for Nuclear Research (INR) Pitesti participated at this CRP with ROFEM and CAREB computer codes. Recently, both codes have been improved with new models in order to extend their capabilities. The behaviour of fuel elements during high-temperature transients like LOCA is of importance to safety and licensing of power reactors. CAREB was developed for fuel transients analyses, such as LOCA and RIA. In this paper a comparison between CAREB code calculations and measured data from FIO-131 LOCA tests is presented. Several parameters were considered, including fuel sheath strains, internal element gas pressure, fuel centerline and sheath temperature, thicknesses of ZrO2 on the sheath. Fuel behavior during high-temperature transient was reasonably well modeled by CAREB code. New LOCA tests are planed to be performed in the C2-LOCA facility of the TRIGA research reactor at INR Pitesti in order to extend the experimental data base used for transient code validation.

Simulation of pressure tube deformation during high temperature transients

In the case of a loss of coolant accident (LOCA) with/without loss of emergency core cooling system (ECCS), in a PHWR, the pressure tube temperature could rise significantly depending on the scenario due to fuel heat up, with the system internal pressure also varying from 9 to 0.1 MPa during the event. This transient can cause metallurgical and geometrical changes in the pressure tube. Depending on the temperature reached, its structural integrity needs to be ensured during such events with the possibility of pressure tube–Calandria Tube contact. At high temperature the pressure tube deforms plastically due to internal pressure and fuel weight. Both ballooning and sagging behaviour are observed in the pressure tube. However, the relative contribution of each deformation is dictated by the pressure and temperature transient. Pressure tube sagging would always lead to an initial bottom contact. In such cases, studies have shown that the temperature difference between the top and bottom of the pressure tube could be as large as 500 °C. Along with the internal pressure, the thermal stress developed may threaten the integrity of the pressure tube. Modelling has been carried out to simulate the simultaneous sagging and ballooning deformation behaviour expected during the channel heat up condition. This paper describes the prediction of creep deformation in the context of the pressure tube configuration in Indian PHWR's and the study of the simultaneous ballooning and sagging behaviour of a pressure tube. The paper also presents the application of this model for a specific postulated LOCA in Indian PHWRs.

Development of a new cost-effective ceramic composite for re-entry heat shield applications

Within the Manned Space Transportation Programme ESA have launched study projects to assess an advanced heat shield concept for re-entry capsules relying on a C/SiC ceramic matrix composite for the hot structure. MAN Technologie has developed a C/SiC composite with a polymer derived SiC matrix and carbon fibre woven fabrics as reinforcement which is being tested in this technology study. It is manufactured by a liquid polymer impregnation (LPI) process. Repeated impregnation and pyrolysis cycles are employed to reach the proper density of 1.75 g/cm 3. The main mechanical properties of the fabric reinforced LPI-C/SiC, determined with test pieces taken from complex parts, are a tensile strength of 270 MPa along with a strain to failure of 1.2%, a compressive strength of 440 MPa, and an interlaminar shear strength of 30 MPa. A strength retention of 80% was found after 30 min exposure to air at 1870K and in LCF tests with a peak stress of 200 MPa a million cycles were achieved without rupture. Integrally laminated panel segments have successfully been tested up to 1800 K in air of 20 hPa with high temperature gradients (>1000 K) and transients (30 K/sec) without damage. these results demonstrate that the LPI-C/SiC with fabric reinforcement provides a proper performance and reliability to open up the option for a CMC heat shield application in advanced crew transfer vehicles.

Method for describing the Zircaloy-4 oxidation using the analytical solution of the oxygen diffusion equation

The oxidation of the Zircaloy-4 fuel element sheath at high temperature transients is treated as a moving boundary one-dimensional diffusion problem. It is usually solved by numerical methods. A method is provided in this work to analytically solve the oxygen diffusion equations. The temperature transient has been approximated as a stepwise function of time. { T i ,( t 0 i , t i )} i = 1, n . The main assumption was that in the beginning of the ith step of the temperature transi the oxide grown during the previous steps, x 0 i , could be considered as being formed at a constant temperature, T i , but during another time interval, named equivalent time interval Δt eq. Calculations have been made with a routine, OXCON, developed using this model. The results approach well the predictions of validated corrosion codes, FROM and PRECIP-II, which solve the diffusion equations using numerical methods.

Materials interactions during high temperature transients: Discussion about the use of the kinetic rate constants in Zircaloy oxidation

The codes that simulate materials interactions in nuclear core degradation or core meltdown generally use the kinetic rate constants obtained from isothermal diffusion experiments. They calculate the interphase boundaries (IB) movements during temperature transients as a succession of isothermal steps at different temperatures. An alternative procedure based on a diffusion model is presented here, which we consider more appropriate for studying materials interactions during high temperature transients. As a system to be simulated we have chosen the transient oxidation of Zry at high temperatures (1000–1300°C). It is shown that the evolution of the oxidized Zry when described by both methods throws different results. The same could happen in other materials interacting at high temperature transients. In these cases diffusion codes should be used; otherwise kinetic parameters obtained from isothermal measurements have to be validated in each case.

Kinetics of liquid phase formation due to solid/solid chemical interactions and its modeling; application to the Zircaloy/Inconel system

Abstract According to recent results of LWR bundle meltdown experiments the formation of liquid phases was observed as a consequence of chemical interactions between the various bundle components below their melting points. The reaction kinetics of the different material combinations in a LWR fuel rod bundle was studied. In this paper the chemical interactions between Inconel spacer grids and as-received and preoxidized Zircaloy (Zry) fuel rod claddings, respectively, are described. The experimental results between 1000 and 1200°C show that the interactions obey parabolic rate laws. The fundamental difference of the experiments with preoxidized Zry (oxide layer thickness ≦ 100 μm) compared to as-received Zry is the delay time for the start of the interactions between Inconel and Zircaloy. A model is presented to describe the experimental results of the solid Inconel/solid Zircaloy interaction kinetics. The simple relations obtained can be used as modules in SFD (severe fuel damage) code systems to describe the behavior of a LWR core during high temperature transients.

Simulation of oxidation phenomena during high temperature transients: Application to zirconium alloys in steam

A model is proposed for describing the phenomena occurring during high temperature transient oxidation in zirconium alloys. Several phases are present and phase transformations take place, particularly during cooling, in a two-phase region where precipitation occurs. This zone is very important because of its influence on the mechanical behaviour of Zircaloy at high temperatures. Experimental measurements give evidence of the existence of this two-phase zone and that it is highly dependent on the oxidation distribution in the metal beneath the oxide layer. A numerical code (TRANOX) was developed to take account not only of the broadening of the single-phase layers but also of the evolution of the two-phase region at high temperatures. In this work, good agreement between the model and the experimental results was found.