Advanced Gas-cooled Reactor Research Articles

Measurements of radioxenon activities during periods of gaseous release from an advanced gas-cooled reactor

Measurements of radioxenon activities during periods of gaseous release from an advanced gas-cooled reactor

3D analysis of crack propagation in nuclear graphite using DVC coupled with finite element analysis

The initiation and propagation of cracks within nuclear graphite components compromise the lifespan of advanced gas-cooled reactors and even lead to safety incidents. A thorough understanding of fracture properties in nuclear graphite is crucial for the effective management and design of graphite structural components. In this work, an in-situ test of a double cleavage drilled compression (DCDC) specimen was conducted using X-ray tomographic imaging to fully analyze the 3D crack propagation behavior in IG-11 nuclear graphite. “Subvolume Splitting” digital volume correlation (SS-DVC) was employed to accurately quantify the displacement fields surrounding the crack. The entire 3D crack propagation and crack surface around the crack front were precisely captured through 3D imaging processing of the gray level residual (GLR) field. Additionally, crack opening displacements (CODs) and stress intensity factors (SIFs) were determined using finite element (FE) analysis with extended finite element method (XFEM). Experimental results indicate that although mode I crack remains predominant throughout the whole process, complex crack propagation exists around the crack front. The utilization of irregular crack surfaces in FE analysis helps to obtain refined displacement fields and accurate SIFs of the crack front. This work provides comprehensive insights into crack propagation and reveals the influence of complex crack morphology on SIF calculations, facilitating accurate prediction of structural integrity and rational design of graphite components.

Microstructural analysis of ex-service neutron irradiated stainless steel nuclear fuel cladding by high-speed AFM

Fuel cladding in advanced gas-cooled nuclear reactors is made of an austenitic stainless steel referred to as 20Cr/25NiNb. This material can undergo microstructural changes in these extreme environments through a combination of thermal effects and radiation damage. In this work, the microstructures present in an ex-service irradiated 20Cr/25Ni-Nb sample are studied using correlated high-speed atomic force microscopy (HS-AFM) and electron microscopy techniques for the first time. Correlated topographic, crystallographic, and chemical information from the sample surface enabled identification of secondary phase precipitates (SPPs) including M23C6, sigma phase, NbC, and G phase. These SPPs can have adverse effects on the material properties. HS-AFM analysis showed surface textures unique to each SPP. Voids formed due to irradiation were also identified across the surface by HS-AFM. These voids were found to be larger in size and depth along grain boundaries. Further analysis identified a relationship between void size and frequency and grain boundary misorientation and SPP presence.

Gamma-spectrometry measurements of filtration media from an Advanced Gas-cooled Reactor

Filtration media used to quantify particulate and gaseous releases have been collected from Hartlepool Power Station in the United Kingdom and measured using high-sensitivity gamma-spectrometry systems. Radionuclides that are relevant to the monitoring regime of the Comprehensive Nuclear-Test-Ban Treaty (CTBT) have been detected. Results are reported and compared to detections recorded on the International Monitoring System (IMS). Time series activity plots have been produced and results interpreted with respect to known plant activities. The reported results improve the understanding of trace-level radionuclide emissions from Advanced Gas-cooled Reactors (AGRs) and aid interpretation of IMS measurements. This work is being performed as part of the Xenon Environmental Nuclide Analysis at Hartlepool (XENAH) collaboration between the Atomic Weapons Establishment (AWE, UK), EDF Energy (UK), Pacific Northwest National Laboratory (PNNL, US) and the Swedish Defence Agency (FOI, Sweden).

On the reduced damage tolerance of fine-grained nuclear graphite at elevated temperatures using in situ 4D tomographic imaging

Fine-grained nuclear graphite is one of the key structural materials for high temperature gas-cooled reactors as well as several Generation IV nuclear fission reactor designs. However, its deformation and fracture behaviours at elevated temperatures are not well understood. In light of this, the current study focused on investigating the flexural strength and fracture toughness of two fine-grained graphite (SNG623 and T220) using real-time X-ray computed micro-tomography imaging at room temperature, 750 °C and 1100 °C. Specifically, nonlinear-elastic fracture mechanics-based JR(Δa) R-curves at these temperatures were presented with evolution of damage and failure micro-mechanisms, local strain distributions and J-integral fracture analysis, purveying notable findings. Compared to the coarser-grained Gilsocarbon nuclear graphites used in the current UK Advanced Gas-cooled Reactors (AGRs), these modern fine-grained graphites display deficient fracture resistance in the form of far less stable crack growth prior to catastrophic fracture and reduced failure strain at 1100 °C. Moreover, their elevation in strength and toughness at high temperatures is remarkably lower than that of Gilsocarbon graphite. Based on in situ high-temperature Raman spectroscopy mapping, we believe that one of the major causes of this behaviour can be attributed to the smaller magnitude of ‘frozen-in’ residual stress relaxed at elevated temperatures compared with Gilsocarbon graphite.

Seismic response prediction using intensity measures: Graphite nuclear reactor core model case study

Seismic response analyses of structures have conventionally used the peak ground acceleration or spectral acceleration as an intensity measure to estimate the engineering demand parameters. An extensive shaking table test program was carried out on a quarter-sized advanced gas-cooled reactor (AGR) core model to investigate the global dynamic behavior of the system with degraded graphite components while subjected to seismic excitation. Evaluation of the most widely considered intensity measures, with respect to their capability for predicting the seismic response of an AGR core–like structure, is performed. Twenty intensity measures of 16 distinct seismic input motions are formulated and correlated, with experimental measurements describing the dynamic response of the reactor core model. Linear correlations are constructed for each intensity measure to statistically determine the best metric for predicting the seismic response of the AGR core model, and statistical analysis indicates that the acceleration spectrum intensity (ASI) is best suited to characterize and describe the structural demand of an AGR core-like structure when subjected to seismic loading. A response prediction tool is developed, based on empirically derived linear correlations, to estimate column distortions and determine the critical input motion for further experimental and numerical studies. Statistical analysis indicates that predicted column distortions, compared against direct experimental displacements, are significant, repeatable, and accurate.

Study of the Perturbation in Temperature Profile of an AGR Fuel Pin for Surface Roughness of Cladding by CFD Simulation in Ansys Fluent

The surface roughness of nuclear fuel cladding plays a crucial role in the thermal-hydraulic response of the Advanced Gas Cooled reactor (AGR). In the present work, the change in the temperature distribution from an isolated AGR fuel rod to primary coolant due to cladding roughness was studied by computational fluid dynamics (CFD) simulation in Ansys Fluent software. Square transverse ribs of the various pitch to height ratios _p / k_ were considered as the surface roughness. Radial temperature profiles from fuel to coolant were generated. Lower fuel temperature was found for the fuel rod with a rough cladding surface as compared to the smooth cladding surface. The peak fuel temperature was determined and found to decrease with decreasing values of _p / k_ . Temperature drop across the fuel and from fuel to coolant was also studied. DUJASE Vol. 7(1) 9-15, 2022 (January)

The Influence of High-Temperature Helium and the Amount of Revert Material on the Material Properties of Inconel 738

Nickel-based alloys are considered promising materials for primary circuits of high-temperature gas reactors (HTGRs), specifically for gas turbines. The primary helium (He) coolant in the gas-turbine-based HTGRs is expected to reach temperatures of up to 900 °C; therefore, the selected materials should adequately perform over a long service life at such an environment. A promising manufacturing method in the production of reactor components is precision casting, where the content of revert (recyclate) material in the alloy differs and can influence the material behavior. In our study, Inconel alloy 738 was manufactured by casting 50% and 100% of revert material and tested in HTGR conditions to examine the influence of helium coolant on the material’s properties. Tensile specimens were exposed at 900 °C for 1000 h in helium containing a specified amount of gaseous impurities. Scanning electron microscopy (SEM) with energy dispersive spectroscopy (EDS), in combination with X-ray diffraction (XRD) and nano-, microhardness methods, were used for material characterization after performing the tensile tests at room temperature. The presence of three types of layers was observed: a thin layer formed by aluminum and chromium oxides on the surface; non-uniform surface oxides Ti3O5 with inner (Al,Cr)2O3; and the inner fine-grained Inconel Cr-enriched phase (approx. 10–20 µm below the surface), which can act as a protective surface layer. Mechanical properties of both revert materials decreased after exposure to HTGR conditions but did not show a significant difference as a result of the content of the revert material. The increase of nano-hardness in line profiles throughout the specimen’s cross-section was observed locally at the surface oxides and in the precipitates and grain boundaries. After exposure, Rp0.2 values decreased by 20% and 17.7%, and Rm values by 12.3% and 20.8% in samples with 50 and 100% revert content, respectively. Furthermore, a decrease in microhardness values (HV0.1) was detected by 4.98% in longitude and 5.80% in cross-section for samples with 50% revert material and by 3.85% in longitude and 7.86% in cross-section for samples with 100% revert material. It can be concluded that both revert materials have similar corrosion resistance in HTGR conditions. The presented results complement the knowledge about the degradation of alloys in the coolant environment of advanced gas-cooled reactors.

Unified viscoplasticity model for type 316 stainless steel subjected to isothermal and anisothermal-mechanical loading

316FR-type low carbon stainless steel is usually used for many advanced gas cooled reactor (AGR) power plants which are subjected to operating temperatures in the range of 500–650 °C under thermo-mechanical loading, which may lead to isothermal low cycle fatigue (LCF) and thermal mechanical fatigue (TMF) damage. The strain controlled thermo-mechanical fatigue test with and without holding time was carried out for 316FR-type stainless steel under the temperature cycling between 500 °C and 650 °C, and the strain ranges are Δε = ± 0.4%, ±0.8%, ±1.0% and ±1.2%, respectively. Within the framework of Abdel Karim and Ohno model, an improved unified viscoplastic constitutive model (UVM) is proposed to simulate the cyclic plastic behavior of 316FR stainless steel. The improved UVM includes strain rate sensitivity, temperature rate dependence, static recovery, average stress evolution and strain range dependence of cyclic hardening. And then based on the radial return mapping and backward Euler integration methods, the implicit stress integration algorithm and the new expressions of the consistent tangent modulus are derived respectively, the improved UVM is implemented by a User Material Subroutine into ABAQUS software again. The good correlation between simulation results and experimental data is obtained.

Data‐driven prediction of the probability of creep–fatigue crack initiation in 316H stainless steel

AbstractStainless steel components in advanced gas‐cooled reactors (AGRs) are susceptible to creep–fatigue cracking at high temperatures. Quantifying the probability of creep–fatigue crack initiation requires probabilistic numerical simulations; these are complex and computationally intensive. Here, we present a data‐driven approach to develop fast probabilistic surrogate models of creep–fatigue crack initiation in 316H stainless steel. We perform a set of Monte Carlo simulations based on the R5V2/3 high temperature assessment procedure and determine the sensitivity of the probability of crack initiation to loads and operating conditions. The data are used to train different supervised machine learning models considering Bayesian hyperparameter optimization. We discuss the relative performance of such models and show that a gradient tree boosting algorithm results in surrogate models with the highest accuracy.

Observations of high burnup structure in AGR fuel

Nuclear power plants (NPPs) within the UK are largely of the commercial advanced gas-cooled reactor (AGR) design. The key elements of the reactor and fuel design differ somewhat to those seen in Light Water Reactors (LWRs), most markedly the cladding, an austenitic stainless steel, and annular ceramic UO2 fuel pellets. The graphite moderated AGR cores also experience much higher temperatures than PWR (Pressurised water reactor) cores.In order to fully understand the requirements for enhanced fuel performance and interim/long term spent fuel storage, some efforts must be made to compare and contrast the two fuel types, to understand if LWR data are fully applicable to AGR data in various operational and storage scenarios. One such microstructural effect requiring assessment is a fuel microstructure commonly referred to as high burnup structure (HBS).The current study describes a series of post-irradiation examinations (PIE) undertaken on AGR fuel. A total of circa 30 AGR samples are examined in which HBS is present, and a further similar size of samples in which no HBS was present. Observations include SEM (scanning electron microscopy) and LOM (light optical microscopy) providing a characterisation of the appearance of HBS in AGR. The results are supported by burnup profile modelling, and assessment of fuel temperature from metallographic markers.The results are consistent with AGR outer ring fuel exhibiting an onset of HBS at a rod average burnup of ∼30 MWd/kgU and a 100 μm HBS thickness at a rod average burnup of 40 MWd/kgU. These values are lower than the comparable (pellet average) burnup for PWR fuel. The difference in burnup is shown to be due to the neutron spectrum in AGR fuel elements, and which also leads to a marked variation in HBS thickness around the circumference of AGR fuel. There is no suggestion that the local onset burnup for HBS formation differs between the two fuel types.

Simulated advanced gas-cooled reactor spent nuclear fuels: Determination of the O/U ratio - an XRD, XPS and Raman study

This paper reports a detailed chemical and materials characterization study of novel Simulated Spent Nuclear Fuels (SIMFuel) which replicate the chemical and microstructural state of Spent Nuclear Fuel (SNF) discharged from a UK Advanced Gas-cooled Reactor (AGR). Recent advances in the analysis of pure UO2 samples by XPS, micro-Raman spectroscopy and XRD, have been assessed for deployment in the characterisation of SIMFuels generally and these AGR SIMFuels in particular. All three analytical methods reveal the extent that the UO2 bulk matrix is defected in the SIMFuels by the presence of lanthanide dopants. XPS, by inspection of the U4f peaks and their satellites, indicates the presence of U(V) in the UO2 matrix as a means to charge compensate for the incorporation of Ln(III) states and presence of a slight hyperstoichiometry in the UO2 solid solution. This was corroborated by detailed Raman analysis of the UO2 matrix of the SIMFuels – wherein the (Ba,Sr)ZrO3 and metallic particle phases, that simulate the precipitated grey phases and ε-particles formed in real spent fuel, were avoided – which indicated the presence of both interstitial oxygens and oxygen vacancies, the latter as a parallel means of charge compensation for the presence of Ln(III). Further confirmation was provided by XRD measurements through observation of a lattice parameter contraction arising from the presence of the smaller U(V) ion in the U(IV)O2 matrix. XPS and Raman allow for the reporting of O/U ratios in the SIMFuels. These XPS and Raman-derived O/U ratios are in good agreement, indicating: (i) that the SIMFuels are near stoichiometric/slightly hyperstoichiometric, concurring with previous analogous studies on real and simulated light water reactor fuels; and (ii) that either technique may find application in the analysis of real fuel analysis, both pre- and post-irradiation.

A surrogate machine learning model for advanced gas-cooled reactor graphite core safety analysis

A surrogate machine learning model was developed with the aim of predicting seismic graphite core displacements from crack configurations for the advanced gas-cooled reactor. The model was trained on a dataset generated by a software package which simulates the behaviour of the graphite core during a severe earthquake. Several machine learning techniques, such as the use of convolutional neural networks, were identified as highly applicable to this particular problem. Through the development of the model, several observations and insights were garnered which may be of interest from a graphite core analysis and safety perspective. The best performing model was capable of making 95% of test set predictions within a 20 percentage point margin of the ground truth.

Design space exploration for salt-cooled reactor system: Part I – Thermal hydraulic design

Fluoride-salt-cooled High-temperature Reactors (FHRs) are a variant of nuclear reactor offering high power densities and operating temperatures. This paper proposes a new design stream of FHR, leveraging developments and experience from the British Advanced Gas-cooled Reactor (AGR) fleet, the only commercial fleet of high-temperature reactors to-date. The design space for the proposed AGR-like FHR is explored using a thermal-hydraulic simulation and optimisation package, seeking to maximise the reactor’s economic performance. Starting from a salt-cooled AGR assembly and applying reasonable thermal-hydraulic constraints, this exploration includes varying the assembly fuel type and geometry, coolant salt, as well as design of the pumps and heat exchangers. Some of the resulting designs promise substantially higher power densities than AGRs and improved economics.

Creep deformation phenomena in near-surface carburised layers of 316H stainless steels

The local creep-induced phenomena occurring within the near-surface carburised layer in two pre-carburised type 316H austenitic stainless steels, containing 0.98 or 1.52 wt% Mn, was assessed at 550 °C. The pre-carburisation exposed both steel grades at either 550 °C (8000 h) or 600 °C (3000 h) to simulated CO2-rich gas environment of UK Advanced Gas-cooled Reactors. The local properties within the carburised layer causes a reduction in minimum creep rate as compared to the non-carburised condition. In both steels carburised at 600 °C, ferrite is observed to form in the most heavily carburised regions, including slip bands, during the initial plastic deformation to the target applied stress. The ferrite area fraction further increases with creep strain. Within the ferritic areas, new fine grains are observed in the creep tested specimens and the fraction of high angle grain boundaries is >60%. Those fine grains have a lower degree of lattice strain compared to the surrounding ferrite area and especially to the austenite matrix. The lattice strain localises within the carburised layer, and the formation of ferrite from the strained austenite, and potentially a dynamic recrystallization within the ferrite regions, pinpoints to a dominant strain relief-driven mechanism. Lattice strain localises primarily in the pre-carburised austenite grains and its carbide-decorated boundaries, and induces the continuous formation and refinement of ferrite grains, and the development of intergranular cracks that evolve further into the bulk structure. These results emphasise the necessity to account for environmental-creep interactions at elevated temperatures in structural assessment procedures, in the current gas-cooled fission reactor fleet and beyond into (very) high-temperature Gen IV and advanced modular reactor designs.

A mathematical framework for evaluation of column shape profiles for an advanced gas-cooled reactor core model subject to seismic excitation

The safe shutdown of Advanced Gas-cooled Reactor (AGR) nuclear power stations in response to a seismic event is vital to their safety case. The tubular graphite bricks used to moderate the neutrons within an AGR core are arranged in columns whose bores provide channels for either fuel or control rods. Earthquake-induced distortion of the channels could impede the insertion of the control rods and compromise the safe shut-down, maintenance and servicing of the reactor. This paper presents a mathematical framework, utilising Euler mechanics, to evaluate the column shape displacement profiles of fuel and control rod channels within a state-of-the-art quarter-sized physical model of an AGR core when subjected to seismic loading. The data obtained from sensors installed within the model bricks, and configured to monitor interface displacements, are used to infer the global behaviour of a multi-stacked brick column subject to seismic excitation. Directly measured displacements of the top of the brick columns, obtained using a motion capture vision system, are compared with the displacements calculated using the framework presented, verifying the validity of the procedure. Statistical analysis is employed to quantify and characterise the performance of the Euler mechanics method. For multiple model build configurations, which represent different brick-cracking scenarios in an aged core, the Pearson correlation factor between the direct and indirect measurements is evaluated for the top of the column displacements giving an average value of 0.96 in the direction of the input motion. This shows that good agreement is achieved for the column shape displacement time-histories. The seismic responses are shown to be significantly larger in amplitude in the presence of large numbers of cracked bricks.

Large-Scale Modeling of Damage and Failure of Nuclear Graphite Moderated Reactor

Abstract The UK Advanced Gas-Cooled reactors (AGRs) have cores made of graphite bricks with dual functions: as structural elements of the core, providing space for and separating fuel and control rods; and as moderator of the nuclear reaction. Nuclear graphite is a quasi-brittle material, where the dominant mechanism for failure is cracking. While cracking of isolated bricks is expected due to operation-induced changes in graphite microstructure and stress fields, these could be tolerated as far as the overall structural function of the core is maintained. Assessment of the whole core behavior has been previously done with whole scale models where bricks have been considered as rigid body elements connected by elastic-brittle springs. This approach does not allow for the realistic assessment of the stresses in the bricks and associated brick cracking. Reported here are results from an ongoing project, which addresses this shortcoming. The proposed model uses deformable bricks with appropriate interactions, allowing for physically realistic whole core analysis. The results are focused on the damage that a graphite moderated reactor develops during a life cycle, how this affects the behavior of the whole core, and how changes in bricks' behavior impacts the core integrity. The proposed methodology is a major step toward high-fidelity assessment of AGRs' fitness for service, required for supporting continuous safe operation and life-extension decisions.

Development of a Low-Cost 6 DOF Brick Tracking System for Use in Advanced Gas-Cooled Reactor Model Tests.

This paper presents the design of a low-cost, compact instrumentation system to enable six degree of freedom motion tracking of acetal bricks within an experimental model of a cracked Advanced Gas-Cooled Reactor (AGR) core. The system comprises optical and inertial sensors and capitalises on the advantages offered by data fusion techniques. The optical system tracks LED indicators, allowing a brick to be accurately located even in cluttered images. The LED positions are identified using a geometrical correspondence algorithm, which was optimised to be computationally efficient for shallow movements, and complex camera distortions are corrected using a versatile Incident Ray-Tracking calibration. Then, a Perspective-Ray-based Scaled Orthographic projection with Iteration (PRSOI) algorithm is applied to each LED position to determine the six degree of freedom pose. Results from experiments show that the system achieves a low Root Mean Squared (RMS) error of 0.2296 mm in x, 0.3943 mm in y, and 0.0703 mm in z. Although providing an accurate measurement solution, the optical tracking system has a low sample rate and requires the line of sight to be maintained throughout each test. To increase the robustness, accuracy, and sampling frequency of the system, the optical system can be augmented with an Inertial Measurement Unit (IMU). This paper presents a method to integrate the optical system and IMU data by accurately timestamping data from each set of sensors and aligning the two coordinate axes. Once miniaturised, the developed system will be used to track smaller components within the AGR models that cannot be tracked with current instrumentation, expanding reactor core modelling capabilities.

Characterization of radiation damage in 3D printed SiC

• High-purity SiC composites fabricated by additive manufacturing were irradiated. • Radiation-induced defects are easily observed in α-SiC particles but not CVI matrix. • Defect swelling measured by STEM-EELS is similar in α-SiC particles and CVI matrix. • Defect types and swelling are consistent with other forms of nuclear-grade SiC. • Interface denuded zone reveals sink strength bias and relative defect-type mobility. The SiC fuel matrix for advanced gas-cooled high temperature reactors as part of the Transformational Challenge Reactor program serves as the fuel particle container structure, a barrier to fission gas release, and a heat transfer medium. Its performance is particularly important because the fuel matrix must demonstrate good structural stability and thermal behaviors. An additive manufacturing methodology combining a binder jet 3D printing process with chemical vapor infiltration (CVI) for the production of SiC was recently developed. In this study, post irradiation examination by transmission electron microscopy shows that defect accumulation within the printed particles is very similar to other forms of high-purity SiC. However, damage accumulation was not directly observed in the CVI matrix because black spot damage and dislocation loops are difficult to image within the nanoscale highly faulted CVI matrix and because interstitial defects may rapidly annihilate at the stacking faults. Therefore, electron energy loss spectroscopy (EELS) analysis was used to analyze defect swelling in both the printed particles and the CVI matrix. The EELS analysis helped reveal that the radiation-induced swelling in the CVI matrix is similar to that of the printed SiC particles. This work shows that 3D printed SiC has behavior that is comparable to SiC processed by other means and that 3D printing could serve as a suitable processing technique for high-purity SiC for nuclear applications.

Quantifying water in spent fuel assemblies with neutrons: A simulation-based approach

A simulation-based study concerning three non-destructive approaches by which water in spent nuclear fuel assemblies might be quantified with neutrons is described. Three fuel types have been considered: thirty-six spent Advanced Gas-cooled Reactor (AGR) fuel pins contained in a stainless-steel can; a prototype fast reactor (PFR) spent fuel assembly and a light water reactor (LWR) spent fuel assembly – with the PFR and LWR assemblies containing mixed oxide fuels. The three approaches are investigated with MCNP5 based on neutron interrogation: fast neutron detection of the perturbed fast neutron flux and two approaches based on the 2.223 MeV γ-ray photon emission resulting from the 1H(n,γ)2H capture reaction, stimulated with moderated and unmoderated neutrons. For each of these approaches the following perspectives have been considered: neutron transmission, the influence of radiation emitted by the associated spent fuel inventory, and the dependencies on water quantity and location. The moderated γ-ray assay technique is observed to return the most significant distinction between scenarios with and without water in the fuel assemblies, notwithstanding the need for efficiency response functions to be overlaid on the neutron models as part of the interpretation of the MCNP model results. Water present in all of the spent fuels considered is observed to perturb the incident neutron flux to a degree significant within the simulation uncertainties. The AGR and PFR cases are not undermined by the intrinsic neutron field from the fuel; whereas for the LWR-MOX case, the simulations suggest that this will limit its potential. Successful detection of water quantities down to 10 g is anticipated for AGR, whereas reduced mass sensitivity is forecast for PFR and LWR-MOX, where it is assumed that the water will be more dispersed. Water located in fewer locations is observed to be harder to discern, suggesting that small, localised quantities of water could pose a challenge for these techniques.