Irradiation-induced Dimensional Changes Research Articles

Relationship between nuclear graphite moderator brick bore profile measurement and irradiation-induced dimensional change

A stress analysis for a hypothetical nuclear graphite moderator brick is presented, considering dimensional and other property changes due to fast neutron irradiation, to illustrate the relationship between the change in moderator brick bore profile and dimensional change of the material. The results give the stresses and deformations of the brick during operation and at shutdown, with the effect of irradiation creep on the deformation of the brick also considered. The analyses provide information useful to reactor designers and operators for planning graphite monitoring campaigns.

Irradiation effects on thermal expansion of SiC/SiC composite materials

Irradiation-induced dimensional change and thermal expansion of two kinds of composites, self-particle reinforced SiC p/SiC composites and a Hi-Nicalon™ SiC fiber reinforced SiC f/SiC composite, and monolithic α-SiC were measured after irradiation at 0.2 dpa with irradiation temperatures of 573, 673 and 843 K using the JMTR. From the measurement, swelling was observed for the SiC p/SiC composites and the monolithic α-SiC, on the contrary, the SiC f/SiC composites showed a shrinkage. The measured thermal expansion increased with increasing the specimen temperature below the irradiation temperature, and then rapidly decreased over the irradiation temperature. The so-called ‘temperature monitor effect’ of the silicon carbide was clearly observed for all specimens, the monolithic α-SiC and both composites.

Neutron Irradiation-Induced Dimensional Changes in MEMS Glass Substrates

AbstractA study is made of radiation-induced expansion/compaction in Pyrex and Hoya SD-2 glasses, which are used as substrates for MEMS devices. Glass samples were irradiated with a neutron fluence composed primarily of thermal neutrons, and a flotation technique was employed to measure the resulting density changes in the glass. Transport of Ions in Matter (TRIM) calculations were performed to relate fast (∼1 MeV) neutron atomic displacement damage to that of boron thermal neutron capture events, and measured density changes in the glass samples were thus proportionally attributed to thermal and fast neutron fluences. The trend for strain with thermal neutron fluence (n/cm2) was found to be a linear compaction of -2.8×10−20 for Pyrex and -1.0×10−21 for Hoya SD-2. For fast neutron fluence, the trend for strain was also linear: -6.1×10−21for Pyrex and -7.9×10−22 for Hoya SD-2. The measured radiation-induced compaction of Pyrex is found to agree with that of previous studies. To our knowledge, this work represents the first study of compaction in Hoya SD-2 with neutron fluence. Hoya SD-2 is of considerable importance to MEMS, owing to its close thermal expansivity match to silicon from 25-500 C.

Phenomenological Inelastic Constitutive Equations for SiC and SiC Fibers Under Irradiation

Experimental data on irradiation-induced dimensional changes and creep in beta-silicon carbide (SiC) and SiC fibers are analyzed with the objective of studying the constitutive behavior of these materials under high-temperature irradiation. The data analysis includes the empirical representation of irradiation-induced dimensional changes in an SiC matrix and SiC fibers as functions of time and irradiation temperature. The analysis also includes the formulation of simple scaling laws to extrapolate the existing data to fusion conditions on the basis of the physical mechanisms of radiation effects on crystalline solids. Inelastic constitutive equations are then developed for SCS-6 SiC fibers, Nicalon fibers, and chemical vapor deposition SiC. The effects of applied stress, temperature, and irradiation fields on the deformation behavior of this class of materials are simultaneously represented. Numerical results are presented for the relevant creep functions under the conditions of the fusion reactor (ARIES IV) first wall. The developed equations can be used in estimating the macromechanical properties of SiC-SiC composite systems as well as in performing a time-dependent micromechanical analysis that is relevant to slow crack growth and fiber pullout under fusion conditions.

Production bias due to clustering of point defects in irradiation-induced cascades

Abstract In the conventional rate-theory treatment of irradiation-induced dimensional changes based on the concept of dislocation bias, it is implicitly assumed that interstitials and vacancies are produced continuously and uniformly in the form of Frenkel pairs. This implies that all produced point defects are free and are available for annihilation at sinks. This approach may not be appropriate, however, for cases where the damage is produced in the form of cascades. In the cascades both vacancies and interstitials are produced in a highly segregated fashion in that the distributions of vacancies and interstitials are separated from each other in space. This spatial segregation is maintained even after the end of the thermal spike phase during which vacancies condense in the central region of the cascades whereas the high concentration of highly mobile interstitials form clusters at a certain distance from the cascade centre. The consideration of interstitial clustering within the cascade region yields ...

Neutron irradiation-induced dimensional changes in a cesium-highly oriented pyrolytic graphite intercalation compound

Direct observations of Cs-intercalated, highly oriented pyrolytic graphite (HOPG) exposed to neutron irradiation show that the presence of Cs dramatically enhances the radiation-induced dimensional changes. A qualitative explanation of this behavior employs an island domain or “pleated layer” model of the graphite intercalation compound (GIC) structure. The inferences from this model are consistent with earlier data for neutron-induced dimensional changes in oriented pure graphite as a function of crystallite size.

A model to describe the irradiation-induced dimensional changes in polycrystalline graphites and carbons

A model is presented that describes the irradiation-induced dimensional changes of polycrystalline graphites and carbons in terms of the preferred orientation and shape change of the individual crystallites. It is assumed that the shape change of each crystallite arises from the additive effects of uninhibited irradiation growth and plastic deformation induced by internal stresses. Plastic deformation is assumed to occur by steady state creep varying as the nth power of the internally generated stresses. The predictions of the model agree well with experimental observations.

Irradiation-induced dimensional changes of poorly crystalline carbons

Data are presented on irradiation-induced dimensional changes of poorly crystalline carbons at high temperatures (>900°C). The materials surveyed include: 1. (1) carbon fibers, 2. (2) glassy carbons, 3. (3) carbonaceous matrix materials for HTGR fuel rods and 4. (4) pyrocarbons. The materials above are listed in order of increasing stability, with maximum strains ranging from more than 50% for fibers to less than 10% for pyrocarbons. Dimensional changes of highly anisotropic carbon fibers appear to be sensitive to irradiation temperature, as slightly anisotropic pyrocarbons are, whereas temperature seems to have little influence on the behavior of isotropic glassy carbons over the range from 600 to 1350°C. Dimensional changes for graphite-filled matrix materials were roughly isotropic on the average and did not seem to be strongly temperature dependent for the lower fluences investigated. Increased graphite filler lowered volumetric dimensional changes of the matrix in agreement with a rule-of-mixtures relationship between change components for the filler and the less-stable binder phases. Instabilities of all of the poorly crystalline materials were generally greater than those for more crystalline carbons under the same conditions, including highly orientated graphites that approximate single-crystal behavior.

The Mechanical Behavior of Biso-Coated Fuel Particles during Irradiation. Part II: Prediction of Biso Particle Behavior during Irradiation with a Stress-Analysis Model

The initial crystalline anisotropy and density of the outer pyrocarbon coating of a two-layer Biso fuel particle have a very large effect on calculated diametral changes and coating failure probabilities during irradiation. Of the two, the anisotropy is by far the most important parameter affecting the coating failure probability. These effects arise because of variations in the irradiation-induced dimensional changes, irradiation-induced changes in crystalline anisotropy, and mechanical properties of pyrolytic carbons with these two structural parameters. In support of the stress-analysis model used for these calculations, predicted diametral changes of Biso-coated particles agree well with those observed in irradiation experiments.

Irradiation Performance of Fort St. Vrain High-Temperature Gas-Cooled Reactor Fuel in Capsule F-30

Representative production fuel fabricated for the Fort St. Vrain (FSV) high-temperature gas-cooled reactor (HTGR) was tested in capsule F-30. The irradiation conditions experienced by the fuel encompassed the FSV service conditions designed for a 6-yr fuel cycle. Fuel specimens were irradiated at temperatures ranging from 825 to 1250/sup 0/C (1098 to 1523 K) and to a peak fast-neutron exposure of 9.4 x 10/sup 25/ n/m/sup 2/ (E greater than 29 fJ)/sub HTGR/, which is 18 percent beyond the design FSV peak fast-neutron exposure. In-pile fission gas release measurements and postirradiation examination indicated good irradiation stability of the fuel specimens. The 13 bonded fuel rods were intact, and their irradiation-induced dimensional changes were in good agreement with dimensional change curves used in the FSV core design. Total fuel particle failure fractions determined by visual examination, metallography, and fission gas release measurements were consistent with the criterion of less than 1 percent failure at peak exposure conditions assumed in FSV design and licensing evaluations. Fuel performance in the FSV reactor was evaluated using the capsule F-30 irradiation results. The good irradiation behavior of production fuel in this test gives a high degree of confidence in the performance of the FSV core throughoutmore » its lifetime and demonstrates the conservative nature of the FSV fuel particle design.« less

Interactions between fuel pins and assembly components

The structure of a fuel element has the primary function of holding fuel pins in a regular array throughout their irradiation life, and secondary functions of permitting loading and unloading operations and transportation to be carried out without damage. Differential thermal expansions, irradiation-induced dimensional changes and flow-induced phenomena have to be accommodated. These factors can lead to fuel pin damage and ultimate loss of cladding integrity, or to distortions which may affect the thermal performance either in normal operation or in a loss-of-coolant accident. This paper discusses the various types of interaction that have been experienced and their consequences, as well as the design principles that should be followed to avoid them.

Behavior of pyrolytic carbons irradiated under mechanical restraint

The behavior of isotropic pyrolytic carbons irradiated free from stress is compared with that of the same carbons when the irradiation-induced shrinkage is prevented by an unyielding substrate. Failure of the restrained specimens cannot be predicted from the irradiation-induced dimensional changes of the unrestrained specimens, and fracture strengths do not appear to decrease during irradiation in either a restrained or an unrestrained state. However, restraint and the resulting irradiation-induced creep produce an increase in the preferred orientation of the carbon cystallites. This increase in preferred orientation results in increased irradiation-induced dimensional change rates which can explain the discrepancies in the irradiation behavior of the restrained and unrestrained carbons.

163. Influence of microstructure of irradiation-induced dimensional changes in annealed oriented polycarbons

163. Influence of microstructure of irradiation-induced dimensional changes in annealed oriented polycarbons

Irradiation-Induced Dimensional Change in Boronated Graphite

Boronated graphites containing 22 to 43 wt% boron as B4C and representing current warm-pressing or extrusion technology were irradiated at 650 ± 100°C to fast-neutron fluences up to 7 × 1021 n/cm2 ...

High-fluence irradiation-induced dimensional changes of co-deposited carbon-silicon alloys

The high-fluence irradiation-induced dimensional changes of co-deposited carbon-silicon alloys have been investigated at several temperatures. The silicon, which is generally incorporated as silicon carbide, an irradiation-stable phase, leads to reduced density and dimensional changes if sufficient silicon carbide is present. The density of the carbon phase tends to decrease with increasing silicon carbide content and is also an important variable. For silicon carbide fractions above about 0·3 and carbon densities below about 1·9g/cm 3, the dimensional changes do not depend strongly on the irradiation temperature and are considerably smaller than those of unalloyed carbons.

On mathematical models for calculating the mechanical behaviour of coated fuel particles

Coated fuel particles, as used in high-temperature gas-cooled reactors, consist of a fuel kernel and a spherically symmetric multilayer coating. During their lifetime they undergo varying stress states, caused by the simultaneous occurrence of a number of basic physical effects. These effects are discussed on the basis of the presently known data and classified in “actions” and “reactions”: The actions, tending to cause deformations, are fuel swelling, fission gas development, irradiation-induced dimensional changes and thermal expansion; the reactions, tending to restrain the deformations, are elastic and plastic effects, characterized by the build up of stresses. Model calculation of the stress and deformation state in the system is complicated by delayed elastic- and plastic effects, called transient- and steady-state creep, and by the anisotropy and the differential behaviour in and between the materials: UO 2, pyrocarbon, SiC. Model development, which is reviewed here, however allowed one to describe the total “stress history”: it begins with differential thermal expansion during and after fabrication and continues with fast and thermal neutron induced dimensional and density changes, fuel swelling and gas pressure development. It still proceeds when mechanical effects such as decoupling, recoupling and failure of single layers occur as a consequence of critical states of stress and deformation. It does not end before all layers are broken. The concepts of such a highly developed mathematical model are discussed. They are based on a division of the stress history into single consecutive intervals, each one with small enough deformation- and stress variations to allow treatment by the methods of the “classical” linear theory of elasticity. Calculated examples are given, which predict the mentioned physical and mechanical effects. Present practical experience and knowledge of data show, that mathematical model considerations cannot replace test experiments as much as in classical fields. They are important however in designing the experiments and necessary in interpreting them; they provide the only reasonable possibility of extrapolating the results of test irradiations to power reactor conditions.

Irradiation-induced dimensional changes in co-deposited carbon-silicon coatings

Irradiations of pure and silicon-alloyed isotropic carbons (deposited below 1500°C) at 910°–1300°C to fast-neutron fluences in the range 1.4–4.8 × 10 21 n/ cm 2 ( E > 0.18 MeV) show that silicon concentrations above about 18 wt-% are effective in reducing the irradiation-induced dimensional changes. For example, at 910°C the lineal shrinkages caused by irradiation are reduced about a factor of 3 by the addition of 34 wt-% silicon. At 1250° to 1300°C the corresponding reduction is about a factor of 2. At 910°C, irradiation creep strains as high as 5.5% per 10 21 n/cm 2 were measured for the unalloyed carbons. At 1250°–1300°C, total creep strains of 10 per cent were observed for unalloyed carbons, and 8 per cent were observed for the alloyed coatings. Since none of the creep specimens fractured, these are not limiting values. Comparison data for isotropic carbons deposited below 1500°C (LTI carbons) with those for isotropic carbons deposited above 1500°C (HTI carbons) shows that at high fluences the LTI carbons expand at a rate that is substantially less than the corresponding rate for HTI carbons.

Irradiation-induced dimensional changes in imperfect graphite crystals

Irradiation-induced dimensional changes in imperfect graphite crystals

Irradiation-induced dimensional changes and creep of isotropic carbon

This paper describes an investigation of the dimensional behavior of unrestrained and spherically restrained isotropic carbons deposited at both high temperatures (> 1500°C; HTI carbons) and low temperatures (< 1500°C; LTI carbons) during irradiation at 600°, 1000°, and 1250°C to fast-neutron fluences near 2 × 10 21 n/ cm 2( E > 0.18 MeV). The investigation covered, in particular, the effect of the initial density on the unrestrained dimensional changes and on the total strain accommodated without fracture in the spherically restrained specimens. For isotropic carbons (Bacon anisotropy factor < 1.1), the unrestrained dimensional changes were primarily those due to irradiation-induced densification, and the strain accommodated consisted mostly of irradiation-induced creep strain. The largest dimensional changes and highest total strains accommodated were in low-density carbons. At 1000°C the accommodated strain rates were 0.04 per 10 21 n/cm 2 and 0.05 per 10 21 n/cm 2 for LTI and HTI carbons, respectively. At 600°C the respective accommodated strain rates were 0.028 per 10 21 n/cm 2 and 0.025 per 10 21 n/cm 2. At 1250°C both an upper and a lower limit on the density of the restrained LTI carbons that survived irradiation were observed. The lower density limit apparently revealed the maximum densification strain rate that can be tolerated. The existence of the upper density limit is discussed.

Irradiation-induced dimensional change in boron carbide

Irradiation-induced dimensional change in boron carbide