In 2001 the Nuclear Safety Division (NSD) of the UK Health and Safety Executive (HSE) decided to ... more In 2001 the Nuclear Safety Division (NSD) of the UK Health and Safety Executive (HSE) decided to underwrite the Nuclear Graphite Research Group (NGRG) at the University of Manchester, UK with the aim of providing a source of independent research and advice to the HSE(NSD). Since then the group has rapidly expanded to sixteen members and attracted considerable funding from the nuclear power industry and the regulator for a wide range of research and consultancy work. It is now also part of the Material Performance Centre within the BNFL Universities Research Alliance. Extensive collaboration exists between the group and other nuclear research institutes, both in the UK and overseas.
This paper briefly describes some of the research programmes being carried out by the NGRG at Manchester.
Processed XRD patterns of carbon materials used in the production of near-isotropic Gilsocarbon f... more Processed XRD patterns of carbon materials used in the production of near-isotropic Gilsocarbon for nuclear application in AGRs (Advanced Gas-cooled Reactors). Samples of anisotropic Pile Grade A (PGA) nuclear graphite from Magnox reactors are also shared. Type of processing: Baseline removal by Savitzky-Golay smoothing method (with BSpline interpolation method) Studied carbon materials: Coal tar pitch, Calcined (Gilsonite) coke, Green article (mixture of calcined coke and coal tar pitch), Baked article (carbonised green article), Gilsocarbon (graphitised baked article), irradiated gilsocarbon samples (AGR1997, AGR 2000, AGR2006), pristine PGA, irradiated PGA (iPGA)
Primitive (unprocessed) XRD patterns of carbon materials used in the production of near-isotropic... more Primitive (unprocessed) XRD patterns of carbon materials used in the production of near-isotropic Gilsocarbon for nuclear application in AGRs (Advanced Gas-cooled Reactors). Samples of anisotropic Pile Grade A (PGA) nuclear graphite from Magnox reactors are also shared. Studied carbon materials: Coal tar pitch, Calcined (Gilsonite) coke, Green article (mixture of calcined coke and coal tar pitch), Baked article (carbonised green article), Gilsocarbon (graphitised baked article), irradiated gilsocarbon samples (AGR1997, AGR 2000, AGR2006), pristine PGA, irradiated PGA (iPGA)
Abstract There is a significant amount of historic graphite data used in the safety cases for ope... more Abstract There is a significant amount of historic graphite data used in the safety cases for operating reactors and in research aimed at informing the designs of Generation IV High Temperature Gas-cooled Reactors (HTRs) and Molten Salt Reactors (MSR). Over time graphite irradiation ageing has been expressed by a number of units, including Calder Hall Equivalent Dose (CED MWd ATE −1 ), Equivalent DIDO Nickel Dose (EDND n cm −2 ), E n > 0.18 MeV (n cm −2 ) and the now preferred, displacements per atom (dpa), although the former units are still often used. Fortunately, each of these units has a standard value and there are conversion factors between units. This paper examines the origin of each of these units and conversion factors using current nuclear data and codes, to test their validity and margin of error. This work gives the reader an understanding of the derivation and origin of the various graphite dose units and quantifies the values and tolerance of conversion factor quoted in the literature. The background to these conversion factors and the methods used to derive the units are reassessed using the latest nuclear data and modern codes. Most of the calculated and standard values were found to be in reasonable agreement supporting their continued use. However, in the case of the unit based on the energy range of E n > 1.0 MeV, significant differences are seen between the calculated and accepted conversion factors as well as between reactor spectrums. This latter finding was not unexpected as a significant amount of carbon atoms are displaced at lower energies above ∼0.1 MeV.
A continuum damage mechanics (CDM) failure model, which can predict both the initiation and propa... more A continuum damage mechanics (CDM) failure model, which can predict both the initiation and propagation of cracks in brittle materials such as ceramics and nuclear graphite, has been developed and implemented into the commercial finite element code ABAQUS. In this model, both the stress-based failure criterion and fracture-mechanics-based failure criterion are used to define the damage surface, and a softening parameter 'n' is adopted to control the rate of damage development. Specifically, a change in the softening parameter changes the shrinkage rate of the damage surface, thus giving rise to different traction-displacement behaviors at the interface where crack initiation and propagation are taking place. This paper studies the effect of the degree of softening in the CDM model on the predicted strength of brittle materials subjected to loadings with different stress gradients. 0. FOREWORD The failure of brittle materials, such as nuclear graphite and ceramics, is not only a function of component geometry and size but also of loading mode. Therefore, full-size component testing is often required to determine the failure loads. Although various computational methods have been developed to predict the lifetime and failure load of brittle components, the absence of a satisfactory failure model is still an impediment to accurate failure predictions. Several commonly used failure models of graphite were reviewed in Ref. (1). The authors showed that simple models based on critical stress, critical strain and critical strain energy density criteria were remarkably unsuccessful in describing the experimental results. A continuum damage mechanics (CDM) model (2) has been developed for nuclear graphite at the Manchester School of Engineering (MSE), University of Manchester. The CDM model can be viewed as an extension or generalization of the cohesive zone model (3), in that it couples together the effects of mode-I and mode-II loadings. An interface with no thickness is introduced into the continuum solid at positions where potential crack surfaces may form. Within the interfacial constitutive law, which relates tractions and displacements across the interface, a damage parameter is introduced to take account of the effects of damage such as mcro-cracking. A damage surface is then constructed using both the conventional stress-based and fracture-mechanics-based failure criteria. Therefore, the CDM model can predict both the initiation and propagation of cracks in components with high stress concentration or even stress singularity.
Graphite, an attractive moderating material due to its unique characteristics, has been used in d... more Graphite, an attractive moderating material due to its unique characteristics, has been used in different reactor generations and is currently a candidate for future Generation IV high-temperature reactors and molten salt-cooled reactors. The graphite material is complex and its properties are dependent on the length scale. Furthermore, the bulk graphite properties are dependent on many factors, such as manufacturing, grain size, and operational environment. These properties also change over time due to irradiation. Based on experience gained from different current graphite moderated reactors, referring to both past experience and data more recently obtained in various EU Framework projects, this chapter introduces the basics of graphite as a nuclear material for application to Generation IV reactors. Current procedures and understanding of graphite manufacturing, material properties models, component structural integrity assessment, and waste management are presented to help future reactor designers. Specific issues related to high-temperature nuclear graphite are also discussed.
Abstract Nuclear graphite components are produced from polycrystalline artificial graphite manufa... more Abstract Nuclear graphite components are produced from polycrystalline artificial graphite manufacture from a binder and filler coke with approximately 20% porosity. During the operational lifetime, nuclear graphite moderator components are subjected to fast neutron irradiation which contributes to the change of material and physical properties such as thermal expansion co-efficient, young's modulus and dimensional change. These changes are directly driven by irradiation-induced changes to the crystal structure as reflected through the bulk microstructure. It is therefore of critical importance that these irradiation changes and there implication on component property changes are fully understood. This work examines a range of irradiated graphite samples removed from the British Experimental Pile Zero (BEPO) reactor; a low temperature, low fluence, air-cooled Materials Test Reactor which operated in the UK. Raman spectroscopy and high-resolution transmission electron microscopy (HRTEM) have been employed to characterise the effect of increased irradiation fluence on graphite microstructure and understand low temperature irradiation damage processes. HRTEM confirms the structural damage of the crystal lattice caused by irradiation attributed to a high number of defects generation with the accumulation of dislocation interactions at nano-scale range. Irradiation-induced crystal defects, lattice parameters and crystallite size compared to virgin nuclear graphite are characterised using selected area diffraction (SAD) patterns in TEM and Raman Spectroscopy. The consolidated ‘D’peak in the Raman spectra confirms the formation of in-plane point defects and reflected as disordered regions in the lattice. The reduced intensity and broadened peaks of ‘G’ and ‘D’ in the Raman and HRTEM results confirm the appearance of turbulence and disordering of the basal planes whilst maintaining their coherent layered graphite structure.
As an outstanding moderator, graphite is broadly applied in nuclear reactors. Irradiation will ch... more As an outstanding moderator, graphite is broadly applied in nuclear reactors. Irradiation will change the pore structure and content, which is intimately related to the properties of graphite. However, the irradiation effect of nuclear graphite on pores requires a more profound understanding, especially the evolution on a threedimensional(3D) view. In this work, FIB-SEM tomography intuitively reveal the pore structure and porosity evolution of ion irradiated graphite grades IG110 in a 3D view, which was irradiated with 30 MeV Ni 5+ ion at 400 ℃. Besides, the quantitative statistics of porosity indicates a negative correlation between porosity with the ion irradiation damage and confirms the consistency of ion deposition sites and irradiation damage peak. Moreover, individual microcracks of the different ion irradiated depths and virgin graphite are extracted and displayed in various 3D perspectives,which exhabit a much more complex morphology than lenticular structure and demonstrates the closed trend of the microcracks under irradiation.Through the display of 3D pore structure and quantitative statistics of porosity, aims to provide a further understanding of ion irradiation behavior of nuclear graphite.
The graphite core of an Advanced Gas-cooled Reactor (AGR) consists of a complex assembly of a lar... more The graphite core of an Advanced Gas-cooled Reactor (AGR) consists of a complex assembly of a large number of graphite components, such as fuel bricks, interstitial bricks and keys. Fast neutron damage to the graphite causes significant dimensional and material properties changes. Furthermore, in a component, such as a fuel brick, there are significant thermal gradients. The dimensional and property changes which vary across a typical fuel brick as a result of the fluence profiles coupled with the thermal gradients lead to the generation of internal stresses. As the reactor ages, the stresses particularly at keyway corners may lead to the development of cracks in the fuel bricks. In this paper, the behaviour of a cracked brick is investigated and compared to the behaviour of an intact brick. Moreover, the effects of interactions of a cracked or an intact fuel brick with its surrounding components are evaluated. Various numerical models are presented to provide useful information for...
In 2001 the Nuclear Safety Division (NSD) of the UK Health and Safety Executive (HSE) decided to ... more In 2001 the Nuclear Safety Division (NSD) of the UK Health and Safety Executive (HSE) decided to underwrite the Nuclear Graphite Research Group (NGRG) at the University of Manchester, UK with the aim of providing a source of independent research and advice to the HSE(NSD). Since then the group has rapidly expanded to sixteen members and attracted considerable funding from the nuclear power industry and the regulator for a wide range of research and consultancy work. It is now also part of the Material Performance Centre within the BNFL Universities Research Alliance. Extensive collaboration exists between the group and other nuclear research institutes, both in the UK and overseas.
This paper briefly describes some of the research programmes being carried out by the NGRG at Manchester.
Processed XRD patterns of carbon materials used in the production of near-isotropic Gilsocarbon f... more Processed XRD patterns of carbon materials used in the production of near-isotropic Gilsocarbon for nuclear application in AGRs (Advanced Gas-cooled Reactors). Samples of anisotropic Pile Grade A (PGA) nuclear graphite from Magnox reactors are also shared. Type of processing: Baseline removal by Savitzky-Golay smoothing method (with BSpline interpolation method) Studied carbon materials: Coal tar pitch, Calcined (Gilsonite) coke, Green article (mixture of calcined coke and coal tar pitch), Baked article (carbonised green article), Gilsocarbon (graphitised baked article), irradiated gilsocarbon samples (AGR1997, AGR 2000, AGR2006), pristine PGA, irradiated PGA (iPGA)
Primitive (unprocessed) XRD patterns of carbon materials used in the production of near-isotropic... more Primitive (unprocessed) XRD patterns of carbon materials used in the production of near-isotropic Gilsocarbon for nuclear application in AGRs (Advanced Gas-cooled Reactors). Samples of anisotropic Pile Grade A (PGA) nuclear graphite from Magnox reactors are also shared. Studied carbon materials: Coal tar pitch, Calcined (Gilsonite) coke, Green article (mixture of calcined coke and coal tar pitch), Baked article (carbonised green article), Gilsocarbon (graphitised baked article), irradiated gilsocarbon samples (AGR1997, AGR 2000, AGR2006), pristine PGA, irradiated PGA (iPGA)
Abstract There is a significant amount of historic graphite data used in the safety cases for ope... more Abstract There is a significant amount of historic graphite data used in the safety cases for operating reactors and in research aimed at informing the designs of Generation IV High Temperature Gas-cooled Reactors (HTRs) and Molten Salt Reactors (MSR). Over time graphite irradiation ageing has been expressed by a number of units, including Calder Hall Equivalent Dose (CED MWd ATE −1 ), Equivalent DIDO Nickel Dose (EDND n cm −2 ), E n > 0.18 MeV (n cm −2 ) and the now preferred, displacements per atom (dpa), although the former units are still often used. Fortunately, each of these units has a standard value and there are conversion factors between units. This paper examines the origin of each of these units and conversion factors using current nuclear data and codes, to test their validity and margin of error. This work gives the reader an understanding of the derivation and origin of the various graphite dose units and quantifies the values and tolerance of conversion factor quoted in the literature. The background to these conversion factors and the methods used to derive the units are reassessed using the latest nuclear data and modern codes. Most of the calculated and standard values were found to be in reasonable agreement supporting their continued use. However, in the case of the unit based on the energy range of E n > 1.0 MeV, significant differences are seen between the calculated and accepted conversion factors as well as between reactor spectrums. This latter finding was not unexpected as a significant amount of carbon atoms are displaced at lower energies above ∼0.1 MeV.
A continuum damage mechanics (CDM) failure model, which can predict both the initiation and propa... more A continuum damage mechanics (CDM) failure model, which can predict both the initiation and propagation of cracks in brittle materials such as ceramics and nuclear graphite, has been developed and implemented into the commercial finite element code ABAQUS. In this model, both the stress-based failure criterion and fracture-mechanics-based failure criterion are used to define the damage surface, and a softening parameter 'n' is adopted to control the rate of damage development. Specifically, a change in the softening parameter changes the shrinkage rate of the damage surface, thus giving rise to different traction-displacement behaviors at the interface where crack initiation and propagation are taking place. This paper studies the effect of the degree of softening in the CDM model on the predicted strength of brittle materials subjected to loadings with different stress gradients. 0. FOREWORD The failure of brittle materials, such as nuclear graphite and ceramics, is not only a function of component geometry and size but also of loading mode. Therefore, full-size component testing is often required to determine the failure loads. Although various computational methods have been developed to predict the lifetime and failure load of brittle components, the absence of a satisfactory failure model is still an impediment to accurate failure predictions. Several commonly used failure models of graphite were reviewed in Ref. (1). The authors showed that simple models based on critical stress, critical strain and critical strain energy density criteria were remarkably unsuccessful in describing the experimental results. A continuum damage mechanics (CDM) model (2) has been developed for nuclear graphite at the Manchester School of Engineering (MSE), University of Manchester. The CDM model can be viewed as an extension or generalization of the cohesive zone model (3), in that it couples together the effects of mode-I and mode-II loadings. An interface with no thickness is introduced into the continuum solid at positions where potential crack surfaces may form. Within the interfacial constitutive law, which relates tractions and displacements across the interface, a damage parameter is introduced to take account of the effects of damage such as mcro-cracking. A damage surface is then constructed using both the conventional stress-based and fracture-mechanics-based failure criteria. Therefore, the CDM model can predict both the initiation and propagation of cracks in components with high stress concentration or even stress singularity.
Graphite, an attractive moderating material due to its unique characteristics, has been used in d... more Graphite, an attractive moderating material due to its unique characteristics, has been used in different reactor generations and is currently a candidate for future Generation IV high-temperature reactors and molten salt-cooled reactors. The graphite material is complex and its properties are dependent on the length scale. Furthermore, the bulk graphite properties are dependent on many factors, such as manufacturing, grain size, and operational environment. These properties also change over time due to irradiation. Based on experience gained from different current graphite moderated reactors, referring to both past experience and data more recently obtained in various EU Framework projects, this chapter introduces the basics of graphite as a nuclear material for application to Generation IV reactors. Current procedures and understanding of graphite manufacturing, material properties models, component structural integrity assessment, and waste management are presented to help future reactor designers. Specific issues related to high-temperature nuclear graphite are also discussed.
Abstract Nuclear graphite components are produced from polycrystalline artificial graphite manufa... more Abstract Nuclear graphite components are produced from polycrystalline artificial graphite manufacture from a binder and filler coke with approximately 20% porosity. During the operational lifetime, nuclear graphite moderator components are subjected to fast neutron irradiation which contributes to the change of material and physical properties such as thermal expansion co-efficient, young's modulus and dimensional change. These changes are directly driven by irradiation-induced changes to the crystal structure as reflected through the bulk microstructure. It is therefore of critical importance that these irradiation changes and there implication on component property changes are fully understood. This work examines a range of irradiated graphite samples removed from the British Experimental Pile Zero (BEPO) reactor; a low temperature, low fluence, air-cooled Materials Test Reactor which operated in the UK. Raman spectroscopy and high-resolution transmission electron microscopy (HRTEM) have been employed to characterise the effect of increased irradiation fluence on graphite microstructure and understand low temperature irradiation damage processes. HRTEM confirms the structural damage of the crystal lattice caused by irradiation attributed to a high number of defects generation with the accumulation of dislocation interactions at nano-scale range. Irradiation-induced crystal defects, lattice parameters and crystallite size compared to virgin nuclear graphite are characterised using selected area diffraction (SAD) patterns in TEM and Raman Spectroscopy. The consolidated ‘D’peak in the Raman spectra confirms the formation of in-plane point defects and reflected as disordered regions in the lattice. The reduced intensity and broadened peaks of ‘G’ and ‘D’ in the Raman and HRTEM results confirm the appearance of turbulence and disordering of the basal planes whilst maintaining their coherent layered graphite structure.
As an outstanding moderator, graphite is broadly applied in nuclear reactors. Irradiation will ch... more As an outstanding moderator, graphite is broadly applied in nuclear reactors. Irradiation will change the pore structure and content, which is intimately related to the properties of graphite. However, the irradiation effect of nuclear graphite on pores requires a more profound understanding, especially the evolution on a threedimensional(3D) view. In this work, FIB-SEM tomography intuitively reveal the pore structure and porosity evolution of ion irradiated graphite grades IG110 in a 3D view, which was irradiated with 30 MeV Ni 5+ ion at 400 ℃. Besides, the quantitative statistics of porosity indicates a negative correlation between porosity with the ion irradiation damage and confirms the consistency of ion deposition sites and irradiation damage peak. Moreover, individual microcracks of the different ion irradiated depths and virgin graphite are extracted and displayed in various 3D perspectives,which exhabit a much more complex morphology than lenticular structure and demonstrates the closed trend of the microcracks under irradiation.Through the display of 3D pore structure and quantitative statistics of porosity, aims to provide a further understanding of ion irradiation behavior of nuclear graphite.
The graphite core of an Advanced Gas-cooled Reactor (AGR) consists of a complex assembly of a lar... more The graphite core of an Advanced Gas-cooled Reactor (AGR) consists of a complex assembly of a large number of graphite components, such as fuel bricks, interstitial bricks and keys. Fast neutron damage to the graphite causes significant dimensional and material properties changes. Furthermore, in a component, such as a fuel brick, there are significant thermal gradients. The dimensional and property changes which vary across a typical fuel brick as a result of the fluence profiles coupled with the thermal gradients lead to the generation of internal stresses. As the reactor ages, the stresses particularly at keyway corners may lead to the development of cracks in the fuel bricks. In this paper, the behaviour of a cracked brick is investigated and compared to the behaviour of an intact brick. Moreover, the effects of interactions of a cracked or an intact fuel brick with its surrounding components are evaluated. Various numerical models are presented to provide useful information for...
The structural integrity of nuclear graphite bricks is important in Advanced Gas-cooled Reactors ... more The structural integrity of nuclear graphite bricks is important in Advanced Gas-cooled Reactors (AGRs) as they not only provide moderation but channels for fuel, cooling and control rods. AGR graphite moderator bricks are subjected to fast neutron irradiation and radiolytic oxidation during reactor operation, leading to component dimensional and material properties changes. With irradiation ageing the graphite components deform in the axial and radial directions causing changes to bore diameter and brick length. Furthermore, deformation at the brick end faces (called dishing) can lead to the formation of axial gaps between the bricks which can then lead to fuel channel bowing. However, it is believed that irradiation creep reduces the potential size of these axial gaps and retains bricks in contact to some extent. Therefore, as the reactors age it is important to understand the nature of this contact behaviour between bricks in the fuel channels. This paper focuses primarily on the finite element modelling of contact behaviour between bricks and the effects of irradiation-induced creep on contact conditions and hence, ovality of the brick bore. Multilayer fuel brick models and a brick with a rigid body on the top surface have been modelled. The results show that multilayer models are required to understand the contact conditions between the bricks throughout the life of a reactor.
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Articles by Barry Marsden
This paper briefly describes some of the research programmes being carried out by the NGRG at Manchester.
Papers by Barry Marsden
This paper briefly describes some of the research programmes being carried out by the NGRG at Manchester.