Liquid Li is a strong reducing agent, and thus a coating layer of many common oxides is not stable on the wall material. Therefore, intentional coating with insulator ceramics, which are stable in liquid Li, is necessary. Figure 2 shows free energy for oxidation and nitridation for various elements. From the

thermodynamic viewpoint, CaO, Y2O3 and Er2O3, and AlN are expected to be stable. (Note that the resistivity of TiN is not sufficient for the coating.) The instability, however, is a function of the impurity level of Li.10

Early development efforts focused on CaO and AlN by in situ formation during exposure to chemi­cally controlled lithium.

In the development of magnetic fusion power plants, the tritium breeding function is effectively integrated in the blanket, which also serves as the main thermal power conversion system and an effective shield for the adjacent reactor components from neutrons and g-rays. An integrated tritium breeding blanket acts as a shield, as a heat exchanger, and as a breeding zone of tritium fuel, as pictured in Figure 2. The blanket has a primary or first wall that faces the plasma, and this component is in direct contact with the edge of the fusion plasma. The latter is typically designed to remove the plasma radiative power and part of the nuclear heating, either with or without a cooling circuit separate from the blanket system.

In order to obtain a closed D-T fuel cycle for a fusion power plant, it is mandatory that the tritium production rate is, at least effectively, equal to its consumption rate and accounts for decay and losses at scheduled or unscheduled plant outages; this prin­ciple is usually called ‘tritium self-sufficiency.’ These conditions will not be achieved in near-term fusion devices, where tritium resources available from fis­sion plants can be used and where production from a so-called ‘driver’ blanket is an additional or alterna­tive source for fuel supply.

Effective tritium production requires that the lith­ium compounds are located in such a way that the

maximum capture of D—T neutrons is obtained in the so-called tritium breeding blanket. As most fusion devices require partial use of the plasma­facing area for plasma heating, plasma diagnostics, plasma control, and fuel exhaust, the effective cap­ture of D—T neutrons for breeding tritium requires the use of neutron multipliers in the blanket. Net tritium breeding ratios (TBRs) foreseen for power plants should be about 1.05—1.1.

The breeder material used in blanket designs that have attractive thermal efficiency for magnetic con­finement power plants should conform to certain requirements. It should

1. breed tritium in a relatively small volume with a high production rate

2. release tritium in a manner that allows fast proces­sing into plasma fueling

3. possess physical and chemical stability at high temperature

4. display compatibility with adjacent structures and other blanket components

5. exhibit adequate irradiation behavior

6. not pose specific safety risks under off-normal and accidental conditions

7. have activation characteristics allowing recycling or treatment as low-active waste.

Lithium-based ceramics are recognized as attrac­tive tritium breeding materials for the first generation of fusion power plants, due to their inherent thermal stability and chemical inertness.

This chapter describes the development ofceramic breeder (CB) blankets and material production routes applied or investigated and summarizes the properties and R&D results for a number of lithium-based ceramic materials.1 In this chapter the chemical for­mulas are used, though the actual composition are very often non-stoichiometric, which is more evident when a larger fraction of lithium has been burned. Most of the work presented in this chapter is subject to rapid evolutions in local, national or international programs. The authors like to stress that any of the activities, e. g. those concerning ITER can be quite different in their evolution.

Permeability of hydrogen and its isotopes is generally defined as the steady-state diffusional transport of atoms through a material that supports a differential pressure of the hydrogen isotope. Assuming steady state, semi-infinite plate, and Fick’s first law for dif­fusion J — —D(dc/dx), we can express the steady — state diffusional flux of tritium as

Ji = — D(Cxi — Cxi) [15]

x2 — x1

where cx is the concentration at position x within the thickness of the plate. Using chemical equilibrium (eqn [7]) and assuming that the tritium partial pres­sure is negligible on one side of the plate of thickness t, the steady-state diffusional flux can be expressed as

Ji — ^ pTT [16]

and the permeability, F, is defined as:

F = DK [17]

Substituting eqns [2] and [5] into eqn [17], the per­meability can be expressed as a function of tempera­ture in the usual manner:

F = K0D0 exp[-(AHs + Ed)/RT [18]

Permeability is a material property that charac­terizes diffusional transport through a bulk material, that is, it is a relative measure of the transport of tritium when diffusion-limited transport domi­nates; see LeClaire21 for an extensive discussion of permeation. By definition, the permeability (as well as diffusivity and solubility) of hydrogen isotopes through metals is independent of surface condition, since it is related to diffusion of hydrogen through the material lattice (diffusivity) and the thermody­namic equilibrium between the gas and the metal (solubility).

In practice, experimental measurements are strongly influenced by surface condition, such that the measured transport properties may not reflect diffusion-limited transport. Under some conditions (such as low pressure or due to the presence of resid­ual oxygen/moisture in the measurement system), the theoretical proportionality between the square root of pressure and hydrogen isotope flux does not describe the transport;21,22 thus, studies that do not verify diffusion-limited transport should be viewed critically. In particular, determination of the diffusivity of hydrogen and its isotopes is partic­ularly influenced by the surface condition of the specimen, since diffusivity is determined from tran­sient measurements. While permeation measurements (being steady-state measurements) are relatively less sensitive to experimental details, the quality of reported solubility relationships depends directly on the quality of diffusion, since solubility is typically determined from the measured permeability and dif — fusivity.1 In addition, trapping affects diffusivity and must, therefore, be mitigated in order to produce solubility relationships that reflect the lattice disso­lution of hydrogen and its isotopes in the metal. These characteristics of the actual measurements explain the fidelity of permeation measurements between studies in comparison with the much larger variation in the reported diffusivity and solubility.

The performance of bulk W for ITER has been investigated using water-cooled divertor designs, that is, flat tile, macrobrush, and, most relevant, monoblock options. One important factor in these design solutions is the maximum allowable distance between the front surface and coolant to accommodate the heat without melting96 and, if possible, to avoid recrystallization during normal operating conditions. The ability to estimate this parameter requires not only the thermal conductivity of the materials but also the amount of allowable damage at the interface. This requires knowing not only the damage produced during
operation but also understanding the manufacturing accuracy and reproducibility because tens of thousands of armor/heat sink joints will be produced. Studies on this issue have shown that the current W monoblock design with a defect extension up to 50° appears to be suitable for the upper part of the vertical target (P = 10MWm~), but is not well adapted to a heat flux of 20 MW m~ , which is neces­sary for application at the strike point of the vertical target, as systematic defect propagation was observed. A tungsten flat tile design with 6-mm long defects in the material interface was studied and proved to be compatible with fluxes of 5 MW m~ but was unable to sustain cyclic fluxes of 10 MW m~2.212

These results confirm that the monoblock geome­try generally proves to have superior behavior under high heat flux testing when compared with flat tile geometry. However, it is worthwhile to continue the investigation of the flat W tile design for low-flux regions despite the hazard of cascade failure of the flat tiles106 for two reasons: cost and weight.

Besides this characterization, a number of high heat flux tests have been carried out on mock-ups and prototypes without artificial defects representing the different design options to assess the ‘fitness for pur­pose’ of the developed technologies.33’90’161’213-218 The results obtained for small test mock-ups of the flat-tile and monoblock design can be transferred to large-scale prototypes for the divertor vertical target. Independent of the type of pure W or W-La2O3 armor material used in these prototypes’ the W parts survived in the nonneutron-irradiated condition up to 1000 cycles at 20MWm~2 in the monoblock design21 ‘ and up to 1000 cycles at 18 MW m~2 in the flat tile
design (see Figure 7).213 This is far beyond the design requirements for use in the upper part of the vertical target (P = 5 MW m~ ) and, in case of the monoblock design, even meets the design requirements for the strike point area of the vertical target.

Alternative concepts such as explosive bonding of tungsten to a heat sink material,2 PS on a

Cu-alloy216 or on EUROFER steel44 could probably be of use in the divertor but even more for first wall applications for fusion machines beyond ITER. How­ever, these concepts often suffer from high interfacial stresses as a result of the CTE difference between the W coating and the substrate.

4.19.1 Introduction

Beryllium, once called ‘the wonder metal of the future, 1 is a low-density metal that gained early prominence as a neutron reflector in weapons and fission research reactors. It then found a wide range of applications in the automotive, aerospace, defense, medical, and electronic industries. Also, because of its unique physical properties, and especially favorable plasma compatibility, it was considered and used in the past for protection of internal components in various magnetic fusion devices (e. g., UNITOR, ISX-B, JET). Most important future (near-term) applications in this field include (1) the installation of a completely new beryllium wall in the JET tokamak, which has been completed by mid of 2011 and consists of 1700 solid Be tiles machined from 41 of beryllium; and (2) ITER, the world’s largest experimental facility to demon­strate the scientific and technical feasibility of fusion power, which is being built in Cadarache in the South of France. ITER will use beryllium to clad the first wall (^700 m2 for a total weight of about 121 of Be).

Although beryllium has been considered for other applications in fusion (e. g., as neutron multiplier in the design of some types of thermonuclear breeding blankets of future fusion reactors and for hohlraums in inertial confinement fusion), this chapter will be limited to discussing the use of beryllium as a plasma-facing material in magnetic confinement devices, and in particular in the design, research, and development work currently underway for the JET and the ITER tokamaks. Considerations related to health and safety procedures for the use of beryllium relevant for construction and operation in tokamaks are not discussed here.

Designing the interface between a thermonuclear plasma and the surrounding solid material environ­ment has been arguably one of the greatest technical challenges of ITER and will continue to be a chal­lenge for the development of future fusion power reactors. The interaction between the edge plasma and the surrounding surfaces profoundly influences conditions in the core plasma and can damage the surrounding material structures and lead to long machine downtimes for repair. Robust solutions for issues of plasma power handling and plasma-wall interactions (PWIs) are required for the realization of a commercially attractive fusion reactor. A mix of several plasma-facing materials is currently pro­posed in ITER to optimize the requirements of areas with different power and particle flux characteristics (i. e., Be for the first wall, carbon-fiber composite (CFC) for the divertor strike point tiles, and W elsewhere in the divertor). Inevitably, this is expected to lead to cross-material contamination and the for­mation of material mixtures, whose behavior is still uncertain and requires further investigation.

The use of beryllium for plasma-facing — component (PFC) applications has been the subject of many reviews during the last two decades (see, e. g., Wilson et a/2 and Raffray et al.3 and references therein). Much of this fusion-related work has been summarized in a series oftopical workshops on beryl­lium that were held in the past, bringing together leading researchers in the field of beryllium tech­nology and disseminating information on recent progress in the field.4 Comprehensive reviews have also appeared recently in specialized journals5,6 con­taining state-of-the-art information on a number of topics such as manufacturing and development of coat­ing techniques, component design, erosion/deposition, tritium retention, material mixing and compatibility problems, safety of beryllium handling, etc.

This chapter reviews the properties of beryllium that are of primary relevance for plasma protection applications in near-term magnetic fusion devices (i. e., PWIs, thermal and mechanical properties, fab — ricability and ease of joining, chemical reactivity, etc.) together with the available knowledge on perfor­mance and operation in existing fusion machines. Special attention is given to beryllium’s erosion and deposition, the formation of mixed materials, and the hydrogen retention and release characteristics that play an important role in plasma performance, com­ponent lifetime, and operational safety. The status of the available techniques presently considered for joining the beryllium armor to the heat sink material of Cu alloys for the fabrication of beryllium-clad actively cooled components for the ITER first wall is briefly discussed together with the results of the performance and durability heat flux tests conducted in the framework of the ITER first-wall qualification programme. The effects of neutron irradiation on the degradation of the properties of beryllium itself and of the joints are also briefly analyzed.

This chapter is organized as follows. Section

4.19.2 provides some background information for the reader and briefly reviews (1) the problem of PWIs in tokamaks; (2) the history of plasma-facing materials in fusion devices and the rationale for choosing beryllium as the material for the first wall ofJET and ITER; and (3) the experience with the use of beryllium in tokamaks to date. Section 4.19.3 describes in detail the beryllium PWI-relevant prop­erties such as erosion/deposition, hydrogen retention and release, and chemical effects such as material mixing, all of which influence the selection of beryl­lium as armor material for PFCs. Section 4.19.4 briefly reviews a limited number of selected phy­sical and mechanical properties of relevance for the fabrication of heat exhaust components and the effects of neutron irradiation on material properties. Section 4.19.5 describes the fabrication issues and the progress of joining technology and high heat flux durability of beryllium-clad PFCs. Section 4.19.6 describes the main issues associated with the JET and ITER first-wall designs and discusses some con­straints foreseen during operation. The prospects of beryllium for applications in fusion reactors beyond ITER are briefly discussed. Finally, a summary is provided in Section 4.19.7.

4.19.2 Background

4.19.2.1 Synopsis of PWIs in Tokamaks

A detailed discussion on this subject is beyond the scope of this review. The relevant PWIs are compre­hensively reviewed by Federici et a/.7,8 More recent interpretations of the underlying phenomena and impact on the ITER device can be found in Roth eta/.9 Here we summarize some of the main points.

PWIs critically affect tokamak operation in many ways. Erosion by the plasma determines the lifetime ofPFCs, and creates a source ofimpurities, which cool and dilute the plasma. Deposition of material onto PFCs alters their surface composition and, depending on the material used, can lead to long-term accumula­tion of large in-vessel tritium inventories. This latter phenomenon is especially exacerbated for carbon — based materials but there are still some concerns with beryllium. Retention and recycling of hydrogen from PFCs affects fuelling efficiency, plasma density control, and the density of neutral hydrogen in the plasma boundary, which impacts particle and energy transport.

The primary driver for the interactions between the core plasma, edge plasma, and wall is the power generated in the plasma core that must be handled by the surrounding structures. Fusion power is obtained by the reaction of two hydrogen isotopes, deuterium (D) and tritium (T), producing an a-particle and a fast neutron. Although the kinetic energy carried by the 14.1 MeV neutron escapes the plasma and could be converted in future reactors beyond ITER to thermal energy in a surrounding blanket system, the kinetic energy of the a-particle is deposited in the plasma. The fraction of this power that is not radiated from the plasma core as bremsstrahlung or line radi­ation (and that on average is distributed uniformly on the surrounding structures) is transported across field lines to the edge plasma and intersects the material surfaces in specific areas leading to intense power loads. The edge plasma has a strong influence on the core plasma transport processes and thereby on the energy confinement time. A schematic represen­tation of the regions of the plasma and boundary walls in a divertor tokamak is portrayed in Figure 1 taken from Federici et a/.7

The outermost closed magnetic field surface forms an X-point of zero poloidal magnetic field within the vessel. This boundary is called the ‘last closed flux

Figure 1 Poloidal cross-section of a tokamak plasma with a single magnetic null divertor configuration, illustrating the regions of the plasma and the boundary walls where important PWIs and atomic physics processes take place. The characteristic regions are (1) the plasma core,

(2) the edge region just inside the separatrix, (3) the scrape-off-layer (SOL) plasma outside the separatrix, and (4) the divertor plasma region, which is an extension of the SOL plasma along field lines into the divertor chamber.

The baffle structure is designed to prevent neutrals from leaving the divertor. In the private flux region below the X-point, the magnetic field surfaces are isolated from the rest of the plasma. Reproduced with permission from Federici, G.; Skinner, C. H.; Brooks, J. N.; etal. Plasma-material interactions in current tokamaks and their implications for next-step fusion reactors. Nucl. Fusion 2001, 41, 1967-2137 (review special issue), with permission from IAEA.

surface’ (LCFS) or ‘separatrix.’ Magnetic field surfaces inside the LCFS are closed, confining the plasma ions. The edge region, just inside the LCFS, contains signif­icant levels of impurities not fully ionized, and perhaps neutral particles. Impurity line radiation and neutral particles transport some power from here to the wall. The remaining power enters the region outside the LCFS either by conduction or, depending on the degree to which neutrals penetrate the plasma, by convection. This region is known as the scrape-off — layer (SOL) as here power is rapidly ‘scraped off’ by electron heat conduction along open field lines, which are diverted to intersect with target regions that are known as ‘divertors.’ Poloidal divertors have been very successful at localizing the interactions of plasma ions with the target plate material in a part of the machine geometrically distant from the main plasma where any impurities released are well screened from the main plasma and return to the target plate.

The plasma density and temperature determine the flux density and energy of plasma ions striking the plasma-wetted surfaces. These, in turn, deter­mine the rate of physical sputtering, chemical sput­tering, ion implantation, and impurity generation. The interaction of the edge plasma with the sur­rounding solid material surfaces is most intense in the vicinity of the ‘strike point’ where the separatrix intersects the divertor target plate (see inset in Figure 1). In addition, the plasma conditions deter­mine where eroded material is redeposited, and to what degree codeposition of tritium occurs at the wall. The plasma power flow also determines the level of active structural cooling required.

Typical plasma loads and the effects expected during normal operation and off-normal operation in ITER are summarized in Table 1.

Because of the very demanding power handling requirements (predicted peak value of the heat flux in the divertor near the strike-points is >10MWm~ ) and the predicted short lifetime due to sputtering erosion arising from very intense particle fluxes (^1023-1024 particles m~2s_1) and damage during transient events, beryllium has been excluded from use in the ITER divertor and is instead the material selected for the main chamber wall of ITER.

Recent observations in present divertor tokamaks have shown that plasma fluxes to the main wall are dominated by intermittent events leading to fast plasma particle transport that reaches the PFCs along the magnetic field (see Loarte et a/.10 and references therein). The quasistationary heat fluxes to the main wall are thought to be dominated by convective

aW is also considered as an alternative.

bMultifaceted asymmetric radiation from the edge (MARFE).

cVertical displacement event (VDE).

transport,11 but still remain to be clarified. Although the steady-state parallel power fluxes associated with these particle fluxes will only be of the order of several MWm-2 in the ITER QdT = 10 reference scenario, local overheating of exposed edges of main wall PFCs can occur because of limitations in the achievable alignment tolerances. Similarly, transient events are expected to cause significant power fluxes to reach first-wall panels in ITER along the field lines. Edge localized modes (ELMs) deposit large amounts of energy in a short time, and in some cases in a toroidally localized fashion, which can lead to strong excursions in PFC surface temperatures. While the majority of ELM energy is deposited on divertor surfaces, a significant fraction is carried to surfaces outside the divertor. There are obvious concerns that ELMs will lead to damage of the divertor and the first wall.12 An additional concern is that even without erosion, thermal shock can lead to degradation of material thermomechanical properties, for example, loss of ductility leading to an enhanced probability of mechanical failure or spalling (erosion). Research efforts to characterize the ELMs in the SOL are described elsewhere.13-15 There are still large uncer­tainties in predicting the thermal loads of ELMs on the ITER beryllium first wall and the range of parallel energy fluxes varies from 1.0 MJ m~2 (controlled ELMs) to 20MJm~2 (uncontrolled ELMs).16,17 Even for controlled ELMs, such energy fluxes are likely to cause melting of up to several tens of micrometers of beryllium at the exposed edges,18 which could cause undesirable impurity influxes at every ELM.10,11

Compared to disruptions, the thermal effects from runaway electrons are confined to a much smaller area, but the localized damage is expected to be more severe and can cause severe melting/vaporization in virtually all materials and can lead to surface spallation. These events have been observed to cause severe damage to graphite tiles in present day tokamaks. While the beryllium in the strike region will probably be severely melted, the most critical issues for runaway electron damage and VDE are damage of coolant pipes with resulting risk of water spillage. Because of the deep

penetration and large spatial dispersion of the high — energy electrons, a thick armor may be required to avoid overheating of the coolant channels with subsequent coolant leakage. As thicker armor implies higher surface temperatures, the best solution may be local regions that are either uncooled or with thick armor that receives low heat flux during normal opera­tions. Typically a runaway electron energy deposition transient of 50MJm~2 over 0.3 s on the Be first-wall modules results in a maximum heat flux to the coolant of ~7.4MW m~ , a maximum Cu alloy temperature ^640 °C, and a Be melt layer thickness ~1.8mm.2 A reduction of the armor thickness will lead to an increase in the maximum Cu alloy temperature and could lead to the damage of joints.

Within the fusion community, there is an acute awareness of the necessity to construct a suitable irradiation testing facility for materials, which will enable both testing and development of materials for future fusion reactor devices with a fusion-like neutron spectrum. Within this context, both concep­tual and engineering design activities were under­taken during the 1990s within the IEA framework with the view of providing such a facility, the IFMIF (International Fusion Materials Irradiation Facility).46-50 This work has been recently renewed under the EU-Japan Broader Approach (BA) activ­ities with the EVEDA (Engineering Validation and Engineering Design Activities) tasks.51,52 However, at the present time no entirely suitable irradiation testing facility exists, and as a consequence experi­ments have been performed in nuclear fission reac­tors and particle accelerators, as well as g — and X-ray sources, in an attempt to simulate the real operating conditions of the insulating materials and compo­nents. The experiments required must simulate the neutron and g radiation field, that is, the displace­ment and ionization damage rates, the radiation envi­ronment, that is, vacuum and temperature, and also the operating conditions such as applied voltage, or mechanical stress. As will be seen, for the insulator physical properties, it is furthermore essential that in situ testing is carried out to determine whether or not the required physical properties of the material or component are maintained during irradiation. Examples of this include the electrical conductivity, which can increase many orders of magnitude due to the ionizing radiation, or optical windows, which may emit intense RL.

Experimental nuclear fission reactors clearly have the advantage of producing a radiation field consist­ing of both neutrons and g-rays, although in most cases the actual neutron energy spectrum and the dpa to ionization and He ratios are not those which will be experienced in a fusion reactor.50 However, it is worthwhile noting that to date experimental fission reactors have mainly been used for irradiations in the metals programs where the emphasis is on the neutron flux and little consideration is given to the g field. As a result, the irradiation channels have in general been designed and installed with this criterion. However, it should be possible to select positions within the reactors which, together with suitable neutron absorber materials and neutron to g converters, provide acceptable radiation fields. The main difficulties with in-reactor experiments come from the inaccessibility of the radiation volume and are concerned with the problem of carrying out in situ measurements and achieving the correct irra­diation environment. While considerable success has been attained in the in situ measurement require­ment, with parameters such as electrical conductivity, optical absorption and emission, and even radiofre­quency dielectric loss being determined, the problem of irradiating in vacuum still remains, with most experiments being performed in a controlled He environment. Irradiation in a controlled atmosphere such as He causes an immediate problem for in situ electrical and dielectric measurements because of the radiation-enhanced electrical conductivity of the gas,53 and even in the case of irradiation in vacuum at about 10~3mbar spurious leakage currents will occur.54 Furthermore, many in-reactor experiments rely on nuclear heating to reach the required temperature, and hence have difficulty maintaining a controlled temperature, in part because of the changes in the reactor power, and also because of the problem of calculating the final sample or component tempera­ture. These aspects will be further discussed later. One additional difficulty comes from the nuclear activation of the sample or component, which gener­ally means that postirradiation examination (PIE) has either to be carried out in a hot cell or postponed until the material can be safely handled.

Particle accelerators, on the other hand, are ideal for carrying out in situ experiments in high vacuum and at well-controlled temperatures because of the easy access and the very localized radiation field. High levels of displacement damage and ionization can be achieved with little or no nuclear activation. It is however in the nonnuclear aspect ofthe radiation field where their disadvantage is evident, and great care has to be taken to ensure that appropriate dis­placement rates are deduced to enable reliable com­parison with the expected fusion damage. A further serious disadvantage is due to the limited irradiation volume and particle penetration depth. This in gen­eral means that only small thin material samples or components can be tested.

The present-day situation of materials and com­ponent radiation testing for fusion applications takes full advantage not only of fission reactors and particle accelerators, but also 60Co g irradiation facilities and even X-ray sources. The use of such widely different radiation sources can be justified as long as the influ­ence of the type of radiation on the physical parame­ter of interest is known. This, in certain cases, is true for radiation-induced electrical conductivity and RL for example, where for low total fluences it is the ionizing component of the radiation field which is important. In situ measurements can now be made during irradiation of the important electrical, dielec­tric, and optical properties. In addition other aspects such as mechanical strength and tritium diffusion are being assessed during irradiation. Undoubtedly, suc­cessful modeling could be of help to address this diverse use of irradiation sources; however, general modeling for the insulators has hardly got off the ground because of the difficulties associated with describing radiation effects in polyatomic band — structured materials. As a result, in contrast to the extended activity for metallic structural materials, to date there has been no coordinated activity for the insulators, with only specific models for aspects such as electrical and thermal conductivity being developed.

If properly designed and constructed, the concrete structures in NPPs generally have substantial safety margins; however, additional information for quanti­fying the available margins of degraded structures is desired. In addition, how age-related degradation may affect dynamic properties (e. g., stiffness, frequency, and dampening), structural response, structural resistance/ capacity, failure mode, and location of failure initiation is not well understood. A better knowledge of the effects of aging degradation on structures and passive components is necessary to help ensure that the current licensing basis is maintained under all loading conditions.96

Decisions as to whether to invest in maintenance and rehabilitation of structures, systems, and compo­nents as a condition for continued service and risk mitigation, and the appropriate level of investment, should consider the nature and level of uncertainties in their current condition and in future demands.107,108 Recent advances in structural reliability analysis, uncertainty quantification, and probabilistic risk assessment make it possible to perform such evalua­tions and to devise uniform risk-based criteria by which existing facilities can be evaluated to achieve a desired performance level when subjected to uncertain demands.109,110 Consideration of in situ conditions, redundancy, and uncertainties in important engineer­ing parameters often can lead to significant economic benefits when assessing the condition of an existing structure in a (possibly) degraded condition, and the maintenance or rehabilitation strategies that might be required as a condition for future service. Reliability-based approaches have been applied to the NPP concrete structures11 , 2 and in evaluation of the prestress level in concrete containments with unbonded tendons.113

Degradation effects can be quantified with fragil­ity curves developed for both undegraded and degraded components.114 Fragility analysis is a tech­nique for assessing, in probabilistic terms in the pres­ence of uncertainties, the capability of an engineered system to withstand a specified event. Fragility mod­eling requires a focus on the behavior ofthe system as a whole and specifically, on things that can go wrong with the system. The fragility modeling process leads to a median-centered (or likely) estimate of system performance, coupled with an estimate of the varia­bility or uncertainty in performance. The fragility concept has found widespread usage in the nuclear industry, where it has been used in seismic prob­abilistic safety and/or margin assessments of safety- related plant systems.115 The fragility modeling procedures applied to degraded concrete members can be used to assess the effects of degradation on plant risk and can lead to the development of probability-based degradation acceptance limits. This approach has been applied to a limited extent to degraded flexural members and shear walls.96 Additional work is desired in this area for the purpose of refining and applying the time-dependent reliabil­ity methodology for optimizing in-service inspec — tion/maintenance strategies and for developing and evaluating improved quantitative models for predicting future performance (or failure probability) of a degraded concrete structure, either at present or at some future point in time.

As the ceramic breeder material is used in the form of pebble beds, the macroscopic properties of the gran­ular material are of particular interest. Most of these properties (mechanical, thermal, etc.) cannot be deduced directly or precisely from the properties of the base material or the single pebble; therefore, dedicated experiments are required. In these experi­ments, particular attention is paid to reproduce rep­resentative conditions in terms of, for example, pebble bed typical dimensions, packing factor, reduc­ing atmosphere, and so on, as envisaged for the blan­ket component.

This section mostly follows the approach devel­oped for the European blanket program and the work Reimann and coworkers94-106,110,115 have performed from the late 1990s onwards.

In addition to forming Al2O3, which is known to decrease hydrogen permeation, aluminization of steels

forms aluminide intermetallics that are believed to also lower permeability. Most studies of aluminized samples either have intentionally grown an oxidized layer in order to achieve greater PRFs or at least have not attempted to suppress the formation of surface Al2O3 prior to permeation testing. To our knowledge, no permeation measurements on oxide-free alumi — nides have been performed. However, different pro­cesses lead to oxide scales of differing composition, thickness, and defect density, and the PRF may not be attributable to oxides alone. Steels have been alumi­nized by the hot dip process (described earlier), as well as by various chemical vapor deposition (CVD), spray, packed cementation bed, and hot isostatic press­ing (HIP) techniques. For those techniques that lay down a substantial amount of materials that does not react with the matrix (such as HIP), an aluminum- containing iron alloy can be used in preference to aluminum to offer a higher temperature barrier per — formance.189 Oftentimes, the aluminized layer will be made up of mixed FeAl, FeAl2, Fe2Al3, Fe2Al5, FeAl3, Fe3Al, and even Fe4Al13 intermetallics.190 Nickel, chro­mium, and mixed-aluminides are also formed.191,192 Due to the aluminum-rich intermetallics on the sur­face, aluminized material will often have a mixed oxide scale that is rich in Al2O3.193,194 PRFs are generally larger than for a pure aluminum layer, varying between 10 and 10 0 00,175,195 while barriers containing a clean Al2O3 surface often have the greatest PRF.196 It should be noted that aluminum additions can also stabilize ferrite in some austenitic stainless steels, producing a duplex microstructure and an increased permeation.197

4.16.3.3.2 Nitrides

As with oxides, either the native nitrides of the base metal or depositions of other nitrides can be made to serve as barriers. One of the most common native nitrides is Fe2N, which forms when the upstream face of steel samples are nitrided and reduces permeabil­ity by one to three orders of magnitude.198-200

After oxides and aluminide, TiN coatings are one of the most researched barriers because of their good adhesion and the ease of deposition.175 Reported PRFs for nitride barriers vary widely, from less than an order of magnitude to six orders of magnitude. TiN barriers reduce permeability the most when they are placed on the high-pressure side of samples, and much less change in permeation is observed when they are placed downstream.201,202

Boron nitride has been shown to reduce the permeability of hydrogen in 304SS by one to two orders of magnitude.175,203 It forms both cubic and hexagonal structures. Checchetto et a/204 noted that the hexagonal structures absorb a greater amount of hydrogen isotopes (because of a larger number of trapping sites; particularly dangling B and nitrogen bonds), but also display a greater diffusivity of hydro­gen. The latter effect may be due to preferential diffusion along the a direction of the hexagonal lattice.

Bazzanella et a/.205 found that a 1.7 pm (Al, Ti)N coating reduces the deuterium permeability of 0.1-mm thick 316L by two to three orders of magnitude. From permeation transients, they speculated that this reduction was primarily due to the very low diffusiv — ity of D in (Al, Ti)N.