In-reactor Experiments Research Articles

Wireless coupling for LVDT sensor for reactor applications

Wireless coupling can offer significant advantages for in-reactor applications. For instance, wireless signal couplings can avoid electrical feedthroughs penetrating fuel cladding or other high-temperature pressure barriers in fuel performance testing. This paper presents a feasibility study of a wireless coupling technology for the linear variable differential transducer (LVDT), as LVDTs have been successfully demonstrated in reactor applications and can be used to measure a variety of physical parameters. The coupling technology utilizes mutual inductance at a low frequency to transfer power into and signal out of a stainless steel pressure barrier. Using Kirchhoff’s law and magnetic circuit analysis, a theoretical model is developed, including the LVDT, wireless power coupling parts, and wireless signal coupling parts. The LVDT is assumed to have a constant inductance of the primary coil and the sum of two secondary coils. An experimental system is built to test and verify the performance of the technology with a commercial LVDT. The experimental setup has an input voltage limitation to prevent the excessive heating. The non-linearity of the commercial LVDT is ±0.77% as measured in the lab. After coupling the LVDT with the wireless system, the measurement non-linearity is increased to ±1.26%, which shows the wireless power and signal transfer can be applied with LVDTs with limited additional measurement uncertainty. The theoretical model prediction agrees with the experiment results, with the maximum discrepancy of 1.5% between prediction and experiment. The model and experimental results of this study can be directly used to guide the design and optimization of wireless LVDT measurement systems for future in-reactor experiments.

The WWR-K Reactor Experimental Base for Studies of the Tritium Release from Materials Under Irradiation

The WWR-K is 6-MW(thermal) light-water, tank-type reactor with thermal neutron spectrum. It is the exclusive multipurpose research reactor in the Republic of Kazakhstan. The WWR-K is owned by the Institute of Nuclear Physics of the Ministry of Energy of the Republic of Kazakhstan. The coolant is desalted water. The moderator and the reflector are desalted water and beryllium. The reactor operates on uranium dioxide that is enriched to 19.7% by 235U. The reactor is equipped with irradiation channels with the following characteristics: the thermal neutron flux density in the core center comprises 2 · 1014 cm−2s−1, whereas the fast neutron flux density (En > 0.1 MeV) comprises ~8 · 1013 cm−2s−1; in the core periphery, fluxes of the thermal and fast neutrons comprise, respectively, ~8 · 1013 and ~6 · 1012 cm−2s−1. The regular irradiation cycle length is 21 days. The annual number of cycles is ten.Since WWR-K reactor startup, the studies of various prospective reactor materials and fuels have been carried out in its core. Since 2000, activities on in-reactor tests of fusion reactor materials have been performed at the WWR-K reactor, such as experiments on tritium release out of lithium ceramics. Tests forced development and fabrication of an installation for in-reactor studies of tritium release from various candidate materials of fusion reactor blankets in the inert gas environment of an ampoule with specimens under study. Also, a technique has been developed for assessment of the time of tritium retention in materials under irradiation.In 2018, the WWR-K reactor facility was upgraded for studies of fusion reactor materials under irradiation, which makes it possible to carry out experiments on irradiation of specimens at vacuum conditions.This work presents the experimental facility description and block circuit along with its general technical capacities as applied to the expected studies of tritium release of fusion reactor materials at the WWR-K reactor. The developed irradiation ampoule device is presented schematically as well. Also, the obtained results of the neutron-physical, thermophysical, and vacuum calculations for the in-reactor experiment on irradiation of fusion reactor blanket materials are given.

Fracture resistance of imperfect filler particles within nuclear graphite during irradiation

Although there have been many in-reactor experiments of nuclear graphite in history, its microstructure evolution during irradiation have not been well studied. One filler particle within nuclear graphite was investigated during ion irradiation. Remarkable dimensional changes and distortion of graphite filler were observed. Different from the well-known Mrozowski crack, cracks formed by graphite sheets fracture were found in filler and showed continuous opening during irradiation, resulting in porosity increase and decline in integrity of nuclear graphite. But evidences show that these cracks did not grow along the c-axis, indicating a strong resistance to irradiation-induced fracture of un-fractured graphite sheets. Although the dimensional evolution of graphite crystallite continues at the highest dose of 10 dpa, the Raman spectra showed that the structure of damaged graphite lattice became very stable at doses above 0.84 dpa during room-temperature irradiation.

Analysis, Design, and Development of a Compact LVDT for In-Reactor Experiments

ABSTRACTA compact linear variable differential transformer (LVDT) is newly designed, fabricated, and tested for its possible use in the in-reactor experiments, due to the restricted space available in reactors for experiments. This paper discusses the approach adopted in obtaining analytical equations for designing the LVDT and verifying of the result using experiment and by computation using COMSOL software. In the analytical work, an integral equation is derived using equations for current carrying conductor between the primary and secondary coils. It is then solved by applying three-dimensional coordinate geometry using MATLAB. Results obtained from solving analytical equations, COSMOL computations and experiment are seen to be matching with each other.

Modeling for Late-Phase Hydrogen Generation in a Degraded Core

Evaluations of severe accident conditions for water-cooled reactors with metallic fuel pin cladding must consider the oxidation of this material for accident sequences that could lead to high metal temperatures in a steam environment. Such representations are included in integral accident analysis computer codes. If the oxidation causes sufficiently high temperatures to melt, or liquefies the core materials, the core geometry changes as the melt drains downward and freezes on cooler structures promoting blockages and redirection of steam flowing through the fuel assemblies. Once this configuration forms, the accident condition is characterized as the late phase of core oxidation. The Phebus in-reactor experiments investigated hydrogen generation in this compacted core state and measured the generation rates over several thousand seconds. This paper investigates the role of countercurrent steam-hydrogen flow to the debris upper surface as a limit for the generation rate and finds that this provides a close description of the behavior for the Phebus experiments. Applying this mechanism to reactor accident conditions shows how this should be considered in the Severe Accident Management Guidelines.

Study of the Effect of Structural-Technological Parameters on the Thermo-Radiation Stability of VVER and PWR Fuel

The IFA-503.1 and −503.2 in-reactor experiments in the HBWR (Norway) are comparable to the microstructural studies of four types of VVER and PWR uranium dioxide fuel fabricated to standard specifications. On the basis of analysis, the radiation stability of VVER fuel as compared with PWR fuel isevaluated and a quantitative dependence of the effect of the initial submicron porosity on the thermoradiation densification of all experimental fuels is obtained. Data confirming a correlation between the radiation densification and a thermal test for dimensional stability are obtained for the fuels studied. Complex experimental studies made it possible to expand the database for verifying thermo-mechanical models and codes intended for predicting the behavior of high-burnup VVER.

The behaviour of weldable strain gauges under nuclear reactor core conditions

Electrical resistance strain gauges are a very useful tool to measure mechanical parameters such as deformation, stress, dynamic strain, vibration, etc. This paper presents our experience with strain gauges in nuclear reactor environments.The nature of nuclear applications and the desirable characteristics of nuclear strain gauges are discussed. Several in-reactor experiments featuring the use of strain gauges are described. The behaviour of weldable strain gauges under intense nuclear radiation is discussed. Experimental results and techniques for the successful applications of strain gauges in nuclear environments are presented.It is concluded that weldable stain gauges can be used successfully under the very hostile conditions of nuclear reactor cores if appropriate procedures are followed.

Ion bombardment induced surface electrical degradation monitoring by means of luminescence in aluminas

Oxide ceramics for use as electrical insulators in future fusion devices, will be exposed to ionization and displacement damage (neutrons, gammas, ion bombardment). Enhanced oxygen loss due to ion bombardment increases surface electrical conductivity, and at the same time the surface emits light due to ion beam induced luminescence (IBIL). Results for 3 types of α-alumina and sapphire measuring electrical surface conductivity and IBIL as a function of dose at different temperatures between 20 and 200°C, show a clear correlation between luminescence and surface electrical degradation. This indicates the potential to remotely monitor insulating material degradation not only in ITER and beyond, but also in the more immediate in-reactor experiments required for materials testing. Partial reduction of degradation by heating in air suggests the possibility for in situ recovery of the insulating properties.

A general model for predicting coolant activity behaviour for fuel-failure monitoring analysis

A mathematical treatment has been developed to predict the release of volatile fission products from operating defective nuclear fuel elements. The fission product activity in both the fuel-to-sheath gap and primary heat transport system as a function of time can be predicted during all reactor operating conditions, including: startup, steady-state, shutdown, and bundle-shifting manoeuvres. In addition, an improved ability to predict the coolant activity of the 135Xe isotope in commercial reactors is discussed. A method is also proposed to estimate both the burnup and the amount of tramp uranium deposits in-core. The model has been validated against in-reactor experiments conducted with defective fuel elements containing natural and artificial failures at the Chalk River Laboratories. Lastly, the model has been benchmarked against a defective fuel occurrence in a commercial reactor.

Modeling dislocation structure development and creep–swelling coupling in neutron irradiated stainless steel

Anisotropic nucleation and growth of multi-classes of dislocation loops under the combined actions of fast-neutrons and an external applied stress are considered in modeling dislocation structure development in metals and alloys. The stochastic nature of the nucleation kinetics is formulated via the Fokker–Planck equation. The strain derived from the climb of the anisotropic dislocation structure is separable into volumetric and deviatoric components, corresponding respectively to swelling and creep. The creep contribution resulting from the development of the stress-induced dislocation anisotropy is found to be very significant and exhibits a strong correlation with swelling. For stainless steel, our model explains very well the complex deformation behavior observed in a wide variety of in-reactor experiments.

A model for predicting coolant activity behaviour for fuel-failure monitoring analysis

A mathematical treatment has been developed to predict the release of volatile fission products from operating defective nuclear fuel elements. The activity in both the fuel-to-clad gap and coolant as a function of time can be predicted during all reactor operations including steady operation as well as reactor shutdown, startup and bundle-shifting maneuvers. The model has been implemented as the STAR (Steady-state and Transient Activity Release) code based on a finite-element solution of the mass transport equations. The model parameters are derived from in-reactor experiments conducted with defective fuel elements containing natural and artificial failures at the Chalk River Laboratories. The STAR code has also been successfully validated against an analytical solution and benchmarked against several defect occurrences in a commercial reactor.

Mechanism of the Excitation of Vibrations in RBMK Fuel Assemblies

It is shown that the high intensity of the vibrations of RBMK fuel assemblies is due to the parametric excitation of oscillations when an axial flow of the coolant interacts with the fuel elements. The periodicity of this coolant action is due to plug-flow regimes of the two-phase coolant, pumped-induced oscillations of the coolant flow rate, and turbulization of the flow by the spacing lattices and fittings. It is shown that an in-reactor experiment must be performed in order to obtain a relation between the steam-content distribution and the vibrations in the RBMK fuel assemblies. For this, it is suggested that the steam-content distribution and the vibrations of the fuel elements be measured by using a standard system of measurements of the thermohydraulic and neutron-physical parameters of RBMK reactors and by analysing the noise in the indications of the sensors monitoring the height distribution of the energy release.

A continuum model for cracked UO 2 fuel

This paper presents a constitutive model for uranium dioxide fuel pellets in light water reactor fuel rods. The proposed model accounts for the fuel mechanical behaviour under pellet cracking, fragment relocation and pellet-clad mechanical interaction. Moreover, the detrimental effect of cracking on the fuel thermal conductivity is considered in the model. An essential part of the model is the representation of pellet cracks, which significantly affect both the mechanical and thermal behaviour of nuclear fuel under operation. Cracking is modelled in a continuum context, where cracks are represented by nonelastic strains in the material. The continuum representation is particularly suitable for finite element computer codes, since cracking can be treated in the same manner as plasticity and creep. The model is derived in the form of a nonlinear constitutive relation for the fuel material, that may be implemented in either two- or three-dimensional finite element fuel performance computer codes. The fundamentals of the model are presented, and issues concerning its numerical implementation are discussed. The model's ability to capture important aspects of the cracked fuel behaviour is also illustrated by comparisons with in-reactor experiments.

Chemical processes in defective LWR fuel rods

The results of several experimental studies aimed at improving understanding of the chemical processes that cause severe degradation of defective light-water reactor fuel cladding are reported. The competition between oxidation and hydriding of zirconium and zircaloy exposed to steam-hydrogen mixtures at 70 bar and 350–400°C was studied by thermogravimetry. A critical H 2/H 2O ratio of the gas separates regimes of these two types of reaction. For sponge-Zr, the critical ratios at 350 and 400°C are ≈ 5000 and ≈ 200, respectively. Radiolysis of steam by alpha particles was studied mass spectrometrically. The yield of the hydrogen radiolysis product in saturated steam at 290°C was found to be ≈ 8 molecules per 100 eV of deposited energy. An in-reactor experiment demonstrated that fission-fragment-irradiated steam can oxidize UO 2 to UO 2+ x .

A generalized model for radiation-induced amorphization and crystallization of U 3Si and U 3Si 2 and recrystallization of UO 2

A rate-theory model of radiation-induced amorphization and crystallization of U 3Si during ion irradiation has been generalized to include U 3Si 2 and UO 2. The generalized model has been applied to ion-irradiation and in-reactor experiments on U 3Si and U 3Si 2 and provides an interpretation for the amorphization curve (dose required to amorphize the material as a function of temperature), for the ion-radiation-induced nanoscale polycrystallization of these materials at temperatures above the critical temperature for amorphization, as well as for the role of the small crystallites in retarding amorphization. An alternative mechanism for the evolution of recrystallization nuclei is described for a model of irradiation-induced recrystallization of UO 2 wherein the stored energy in the UO 2 is concentrated in a network of sinklike nuclei that diminish with dose due to interaction with radiation-produced defects. The sinklike nuclei are identified as cellular dislocation structures that evolve relatively early in the irradiation period. The complicated kinetics involved in the formation of a cellular dislocation network are approximated by the formation and growth of subgrains due to the interaction of shock waves produced by fission-induced damage to the UO 2.

Tritium/hydrogen barrier development

A review of the hydrogen permeation barriers which can be applied to the structural metals used in fusion power plants is presented. Both implanted and chemically available hydrogen isotopes must be controlled in fusion plants. The need for permeation barriers appears strongest in Pb-17Li blanket designs, although barriers are also necessary for other blanket and coolant systems. Barriers which provide greater than a 1000- fold reduction in the permeation of structural metals are desired. In laboratory experiments, aluminide and titanium ceramic coatings provide permeation reduction factors (PRFs) of 1000 to over 100 000 with a wide range of scatter. The rate-controlling mechanism for hydrogen permeation through these barriers may be related to the number and type of defects in the barriers. Although these barriers appear robust and resistant to liquid metal corrosion, irradiation tests which simulate blanket environments result in very low PRFs in comparison with laboratory experiments, i.e. less than 150. It is anticipated from fundamental research activities that the radiation- and electric-field-induced enhancement of hydrogen diffusion in oxides may contribute to the lower PRFs during in-reactor experiments.

Fuel Performance and Fission Product Release Studies for Defected Fuel Elements

Results from numerous in-reactor experiments with uranium dioxide fuel elements that contain defects in the Zircaloy cladding are reviewed. The various factors that influence the rate of physical deterioration of a defected element are examined. Experimental and theoretical investigations into the release behavior of radioactive nobel gases and iodine are considered for both the steady-state and transient situation, focusing on the relationship between the release behavior and the state of deterioration of a fuel element. Application of this work to power reactor operation is discussed. 95 ref.

A computer-simulation assessment of the reliability of thermal-gradient samples for predicting compositional profiles produced by thermotransport in metallic nuclear fuel elements

Redistribution of Zr in metallic nuclear fuels is potentially important to fuel performance. However, it is difficult to determine why Zr redistributes because the reactor environment provides several driving forces. This environment cannot be reproduced outside the reactor, but reactor testing is time-consuming and expensive. Therefore, in-reactor experiments must be as definitive as possible. Computer simulation can guide reactor testing in two ways. Firstly, simpler situations, with fewer variables, can be modeled mathematically, and the predictions compared to in-reactor results to assess importance of modeled variables. The second way in which computer simulation can be useful is by helping researchers assess differences between in-reactor results and corresponding out-of-reactor experiments. Methods presented in this paper can help researchers to interpret such differences. This paper also addresses the more general question of whether TGSs are reliable indicators of in-reactor thermotransport. To do so, this study makes conservative estimates of potential differences between compositional profiles in TGSs and fuel elements. If these overestimated differences are not large enough to make TGSs misrepresent in-reactor thermotransport effects, it will be concluded that TGSs are reliable indicators. This paper ignores irradiation effects and assumes conservation of lattice sites. Consistent with the desire to produce conservative estimates,more » materials properties and other parameters used in this work overestimate differences between fuel elements and TGSs unless stated otherwise.« less

Computer simulation of elemental redistribution in nuclear fuel elements as a thermotransport phenomenon

Redistribution of Zr in metallic nuclear fuels is potentially important to fuel performance. Unfortunately, Zr redistribution is difficult to understand because the irradiation environment provides several driving forces that could induce it. Among these driving forces are fluxes of point defects, diffusion of fission products, and high thermal gradients. This environment is difficult to reproduce outside of the reactor; however, reactor testing is time-consuming and expensive. Therefore, in-reactor experiments must be as definitive as possible. Computer simulation can serve as a guide to reactor testing in such circumstances. Simpler situations, with fewer variables, can be modeled mathematically, and the predictions compared to in-reactor results to assess the importance of the modeled variables. This paper investigates the ability of thermotransport to cause extensive redistribution. An interesting aspect of this approach is that because Zr is non-fissile, its migration changes the temperature profile. Hence, at steady state, the Zr distribution must satisfy the thermotransport equation while producing a temperature profile that allows it to do so. Because the purpose of this study is to investigate general features of compositional profiles produced by thermotransport, a binary alloy was modeled. This study ignores radiation-induced vacancy effects and assumes conservation of lattice sites.

Fission product release from nuclear fuel by recoil and knockout

An analytical model has been developed to describe the fission product release from nuclear fuel arising from the surface-fission release mechanisms of recoil and knockout. Release expressions are evaluated and compared to the short-lived activity measurements from in-reactor experiments with intact operating fuel. Recoil is shown to be an important process for releasing fission products from free UO 2 surfaces into the fuel-to-sheath gap. The model is also applied to tramp uranium in a power reactor primary heat transport circuit where it is demonstrated that recoil is the dominant release mechanism for small particles of fuel which are deposited on in-core surfaces. A methodology is established whereby release from surface contamination can be distinguished from that of fuel pin failure.