4.08.8.3.1 9Cr — and 12Cr-ODS steel cladding in BOR-60

In order to weld 9Cr — and 12Cr-ODS steel claddings with end-plugs for the manufacture of fuel pins, the PRW method was developed in JAEA, which makes joining possible in the solid state condition.70 This method is based on the electrical resistance heating of the components, while maintaining a continuous force sufficient to forge-weld without melting. The appropriate conditions, for example, electric current, voltage, and contact force, were selected. For the PRW — welded specimens, tensile, internal burst, and creep rupture tests, were conducted and their integrity was confirmed. In addition, a nondestructive ultrasonic inspection method was developed to assure the integ­rity of the weld between the cladding and end-plug.

Using this PRW method, upper end-plugs were welded for two types of 9Cr-ODS steel cladding (Mm13) and 12Cr-ODS steel cladding (F13) at JAEA. Figure 4171 shows a cross-section of the welded part between the 9Cr-ODS steel cladding and end-plug. The ODS steel cladding welded to the upper end — plug was shipped to the fuel production facility of the Institute ofAtomic Reactor (RIAR) in Russia where the MOX and UO2 granulated fuels, as well as uranium metal getter particles, were vibro-packed into the ODS steel cladding, and the lower end-plug was welded by the TIG end-face method. The TIG-welded part at

the lower end-plug ensured that its integrity would be maintained at a lower temperature of 400 °C. The inspection and quality control of the fabricated ODS fuel pins were done through X-ray analysis, gamma scanning, and leak testing, etc., which confirmed that the fuel pins satisfied BOR-60 requirements. The fuel pins were loaded into two dismountable experimental assemblies to satisfy the cladding middle wall tempera­ture within 700 °C and 650 °C, and irradiation was conducted in the BOR-60 up to 5 at.% burnup and 25 dpa as the collaborative work between JAEA in Japan and RIAR of Research.71

The results of the postirradiation examination are shown in Figure 42 in the optical micrographs of the upper part of the fuel column of 9Cr-ODS steel fuel; no obvious corrosion inside the cladding was observed.72 The maximum depth of corrosion of 25 pm is partially confirmed in the upper part of the fuel column. The inner corrosion of the ODS cladding can be reduced by using a lower O/M

100 nm

Figure 43 Precipitation occurring during the in-pile service (a) a’-phases at 400 °C, ~0dpa, (b) w-phases at 523 °C to 78.8dpa, and (c) Laves phases at 580°C to 30.5dpa. Reproduced from Dubuisson, P.; Schill, R.; Higon, M.-P.; Grislin, I.; Seran, J.-L. In Effects of Radiation on Materials: 18th International Symposium; Nanstad, R. K., Hamilton, M. L., Garner, F. A., Kumar, A. S., Eds.; American Society for Testing and Materials: Philadelphia, PA; p 882, ASTM STP 1325.

ratio fuel, even in lower Cr content cladding such as 9Cr-ODS steel.

The reactor designer requires a high-density, very pure graphite, with a high scattering cross-section, a low absorption cross-section, and good thermal and mechanical properties, both in the unirradiated and irradiated condition. The purity is important to ensure not only a low absorption cross-section but also that during operation the radioactivity of the graphite remains as low as possible for waste disposal purposes.

Artificial graphite is manufactured from coke obtained either from the petroleum or coal industry, or in some special cases (such as Gilsocarbon, a UK grade of graphite) from a ‘graphitizable’ coke derived

from naturally occurring pitch deposits.9 The raw coke is first calcined to remove volatiles and then ground or crushed for uniformity, before being blended and mixed with a pitch binder. (Crushed ‘scrap’ artifi­cial graphite may be added to help with heat removal during the subsequent baking. For nuclear graphite, this should be of the same grade as the final product.) This mixture is then formed into blocks using one of various techniques such as extrusion, pressing, hydro­static molding, or vibration molding, to produce the so-called ‘green article.’ The ‘green’ blocks are then put into large ‘pit’ or ‘intermittent’ gas or oil-fired furnaces. The blocks are usually arranged in staggers, covered by a metallurgic coke, and baked at around 800 °C in a cycle lasting about 1 month to produce carbon blocks. These carbon blocks may be used for various industrial purposes such as blast furnace liners; it has even been used for neutron shielding in some nuclear reactors. (Care must be taken as the carbon blocks are not as pure as graphite and may lead to waste disposal issues at the end of the reactor life.)

To improve the properties of the graphite pro­duced from the carbon block, the carbon block is often impregnated with a low-density pitch under vacuum in an autoclave. To facilitate the entry of the pitch into the body of the block, the block surface may be broken by grinding. After impregnation the blocks are then rebaked. This process of impregna­tion and rebaking may be repeated 2, 3, or 4 times. However, the improvement in the properties by this process is subject to diminishing rewards.

The next process is graphitization at about 2800­3000 °C by passing a large electrical current at low voltage through the blocks either in an ‘Acheson furnace’ or using an ‘in-line furnace.’ In both cases, the blocks are covered by a metallurgical coke to prevent oxidation. This graphitization cycle may take about 1 month. If necessary, there may be a final purification step. This involves heating the graphite blocks to around 2400 °C in a halogen gas atmosphere to remove impurities. The final product can then be machined into the many intricate com­ponents required in a nuclear reactor.

For quality assurance purposes, during manufac­ture the blocks are numbered at an early stage and this number follows the block through the manu­facturing process. This is clearly an expensive manufacturing process and therefore, at each stage, quality control is very important. Many samples will be taken from the blocks to ensure that the final batch (or heat) is of appropriate quality compared to previ­ous heats. It is important that the reactor operators retain this data in electronic form as it may be required to investigate any anomalous behavior as the reactor ages. Samples of ‘virgin’ unirradiated graphite blocks should also be retained for future reference. Records should include information on the batch or heat, property measurements, nonde­structive testing (NDT) results, and measurements of impurities. It is not enough just to have the ‘ash’ content after incineration and the ‘boron equivalent’ as some impurities, such as nitrogen, chlorine, and cobalt, will cause significant issues related to reactor operation and final waste disposal. It is important that the reactor operator takes responsibility for these measurements as in the past it has been found that reactor designers and graphite manufacturers close down or merge, and records are lost.

Final inspection will uncover issues related to damage, imperfection, quality, etc. Therefore, a ‘con­cessions’ policy is required to determine what is acceptable and where such components can be used in reactor. Again, the reactor operator will require an electronic record of these concessions.

 

1 Krr(T)

 

[52] ox

In recent years, it has been realized that the use of the product rule is simplistic, and most probably, only applicable for low irradiation data, up to a fluence not far beyond dimensional change turnaround and only for relatively low weight loss. Therefore, there has been a recent trend to use empirical fits to reactor or MTR data where available.

By the late 1940s, it was known that graphite compo­nents, when subjected to fast neutron irradiation, suffered significant dimensional change. It was thought that, because of the flux gradient across the brick section, these dimensional changes would gen­erate significant stresses in hollow graphite modera­tor blocks and that this would lead to significant

г
—— ——
№ я. • ■
350
Gilsocarbon (300-400 °C) □ Petroleum coke (560°C) О Unidentified coke (560 °C) О Gilsocarbon (560 °C) a Pitch coke (300-440 °C) E/E0 ■ Petroleum coke (300-440 °C) • Unidentified coke (300-440 °C)———————————————————————————— V(E/E0)

Figure 52 Change in strength of irradiated Gilsocarbon graphite compared to change in modulus. Modified from Brocklehurst, J. E. Chem. Phys. Carbon 1977, 13, 145-272.

Porosity

component failures within a few years of reactor operation. Therefore, the fuel channels in the early reactors, such as the Windscale Piles, were designed to avoid the buildup of stress.

By the 1950s, it was realized that there was an irradiation-induced mechanism that was relieving

stresses generated by dimensional change and the term ‘irradiation induced plasticity’85,86 was coined to describe this mechanism. Later, around 1960, the term ‘irradiation creep’87 started to be used for the difference between dimensional change in loaded and unloaded graphite specimens irradiated to the same dose.

Data on the long-term performance of the reinforced concrete materials are of importance for demonstrat­ing the durability of the NPP concrete structures and in predicting their performance under the influence of pertinent aging factors and environmental
stressors. This information also has application to establishing limits on hostile environmental exposure for these structures and to developing inspection and maintenance programs that will prolong com­ponent service life and improve the probability of the component surviving an extreme event such as a loss-of-coolant accident.

Prior reviews of research conducted on concrete materials and structures indicate that only limited data are available on the long-term (40-80 years) properties of reinforced concrete materials.26 Where concrete properties have been reported for condi­tions that have been well documented, the results were generally for concretes having ages <5 years or for specimens that had been subjected to extreme, nonrepresentative environmental conditions such as accelerated corrosion or aging. Relatively few inves­tigations have been reported providing results on examinations of structures that had been in service for the time period of interest, 20-100 years, and they did not generally provide the ‘high quality’ informa­tion (e. g., baseline material characteristics and changes in material properties with time) that is desired for meaningful assessments to indicate how the structures have changed under the influence of aging factors and environmental stressors.

Limited data on the long-term performance of reinforced concrete materials reported in the litera­ture, results from concrete cores removed from NPPs, and specimens cast in conjunction with NPP facilities have been reported.78 As shown in Figure 6, these results generally show an increase in compres­sive strength (relative to the 28-day reference

strength) at a decreasing rate with age, but the data obtained from the literature were for concrete ages <50years and the nuclear plant data for ages <27 years. (Using results obtained from concrete cores removed from residential buildings and bridges, one reference indicates that although the concrete strength and modulus exhibit an increase with age, the ability of concrete to resist shear and torsion may decrease.79) Long-term laboratory results in the fig­ure for the Portland Cement Association (PCA) and University of Wisconsin (WIS) studies that attained the largest increases in strength were generally for concrete mixes having high water-cement ratios (lower reference compressive strength) and that had access to moisture for continued cement hydration. NPP concretes had higher reference compressive strengths and were essentially maintained under sealed conditions. With the availability of decommis­sioned NPPs and plant modifications requiring removal of materials, opportunities exist to obtain samples for use in providing an improved under­standing of the effects of extended exposure under the unique conditions found in NPPs. In addition to aging, areas of interest would be the effects of long­term thermal loadings at moderate temperature levels and effects of irradiation on load-bearing con­crete structures operating for >40 years. Additional applications of a concrete material sampling activity would be for assessment of construction quality, development of improved damage models, assess­ment and validation of nondestructive testing meth­ods, and evaluation of the performance of repair activities.

With respect to post-tensioning systems, current examination programs such as ASME Section XI Subsection IWL80 are adequate for determining the condition of the post-tensioning system materials and evaluating the effects of conventional degradation. Isolated incidences of wire failure due to corrosion have occurred. Leakage of tendon sheathing filler (ungrouted tendons) has occurred at a few plants but, except for the potential loss of corrosion protec­tion, the problem appears to be primarily aesthetic.81 Tendon forces at most plants are acceptable by a significant margin, but larger than anticipated loss of force has occurred at a few older plants. The hypothetical effect of reduced prestressing force and degradation of prestressing tendons (e. g., broken wires) has been investigated for a typical PWR post — tensioned concrete containment during a loss-of- coolant accident using finite-element analysis.56 (Results for the scenario investigated indicated that loss of prestressing force leads to increased concrete cracking at lower pressurization levels, complete fail­ure of selected hoop tendons can have a significant impact on the containment ultimate capacity, and failure of selected vertical tendons does not have a significant impact on ultimate capacity.) With the potential use of grouted tendon systems in some of the new reactor designs proposed for construction in the United States, improved guidance on in-service inspection of grouted tendons is desired. Other potential research topics related to post-tensioning systems include development of an improved relation­ship between the end-anchorage force measured by the lift-off test and change in mean force along the tendon length for unbonded tendons, as well as an assessment of the validity of using estimates of time-dependent loss of prestressing force based on limited-duration relaxation tests (e. g., 1000 h) and concrete creep results (e. g., 6 months): at a plant 60-year old, this involves application of time factors of 500 and 120, respectively.

The origin for the stability of the (a) loops in zirco­nium is attributed to the relative packing density of the prismatic plane compared to the basal plane, which depends on the c/a ratio of the hcp lattice. Foll and Wilkens64 have proposed that when the c/a ratio is higher than 3, loops are formed in the basal plane with Burgers vector 1/6(2023), whereas if c/a is lower than 3, then loops are formed in the prismatic plane with Burgers vector (a) = 1/3(1120). For all hcp metals, this means that loops are formed in the prismatic plane except for Zn and Cd. This is not the case for Zr, Ti, and Mg where loops are also formed in the basal planes, depending on the irradiation dose, irradiation temperature, and purity of the metal.56,57

MD computations for a-zirconium have also shown that most of the small interstitial clusters produced in the cascade have the form of a dislocation loop with Burgers vector (a) = 1/3(1120). The small vacancy clusters are also found in the prismatic plane.8,28,65 For larger point-defect clusters,66 it is shown that the point-defect clusters in the prismatic plane always relax to perfect dislocation loops with Burgers vector (a) = 1/3(1120). On the other hand, vacancy clusters in the basal plane form a hexagonal loop enclosing a stacking fault with 1 /2(0001) Burgers vector.

The simultaneous observation of vacancy and interstitial (a) loops in zirconium alloys45,48,50,54,61 is a rather surprising feature.53,57 Indeed, as discussed for usual cubic metals, interstitial loops tend to grow under irradiation and the vacancy loops tend to shrink since the edge dislocations are biased toward SIAs due to the EID.

According to Griffiths,57 the coexistence of these two types of loops in zirconium can be explained by a modified SIA bias in zirconium due to (i) a rela­tively small relaxation volume of SIA relative to vacancy (low bias), (ii) interaction with impurities, and (iii) spatial partitioning of vacancy loops and interstitial loops as a result of elastic interactions or
anisotropic diffusion. Other authors53’68 think that this phenomenon is due to a subtle balance of the bias factors of the neighboring point-defect sinks that lead to an increasing bias as the loop size increases if the loop density is high. Woo44 considers that the coexistence of both types of (a) loops can be explained in the frame of the DAD model’ which induces a strong DAD-induced bias. Indeed’ in this model, the (a) type loops are shown to be relatively neutral and may therefore receive a net flow of either interstitials or vacancies, depending on the sink situ­ation in their neighborhood.

Finally, recent computations,69 using the Monte Carlo method, that take into account the large vacancy and interstitial point-defect clusters created inside the cascade as an input microstructure show that both vacancy and interstitial loops are able to grow simultaneously, the proportion of vacancy loops increasing with increasing irradiation temperature. This last phenomenon can be related to the so-called production bias discussed previously.1

4.02.9.5.1 Dependence of irradiation creep on dpa rate

As mentioned earlier, once swelling begins, irradia­tion creep quickly assumes all the parametric depen­dencies of void swelling. However, for many years it
was assumed that the B0 component of creep was also strongly dependent on dpa rate, increasing as the dpa rate fell, as shown in Figure 81.

The original research that established this percep­tion was performed by Lewthwaite and Mosedale on various cold-worked steels in the Dounreay Fast Reactor at temperatures in the 270-350 °C range.178

The explanation advanced for such a dependence was the decreasing amount of annihilation of point defects by recombination at lower dpa rates, where such an effect is expected to be more pronounced at the lower irradiation temperatures characteristic of this experiment.

An earlier review article was published where this and other data sets were assessed to determine the appropriate rate dependence.1 Some data sets avail­able at that time supported a flux dependence and other data sets supported an independence of dpa rate. On balance it appeared that a strong depen­dence of irradiation creep rate on dpa rate was the more defendable conclusion.

With hindsight and additional published data sup­porting the opposite conclusion, it was later realized that apparent dependence of creep rate on dpa rate was an artifact of the analysis procedure used by

Mosedale and Lewthwaite. The authors had not prop­erly separated the transient and post-transient strains, and all of the lower flux data were in the higher-rate transient regime. When the DFR creep data were reanalyzed by Garner and Toloczko, the creep com­pliance B0 was found to be independent of dpa rate.179

Evidence from various independent studies using techniques such as FEGSTEM or 3DAP has con­firmed the small solute clusters to be multi-alloyed with minor constituents of the steel. An example of the composition of a cluster formed in an irradiated low Ni A533B weld metal is shown in Figure 7. The results of the 3DAP study clearly show that the CECs are alloyed

Table 6 Techniques for the characterization of irradiation-induced solute clusters

Existing Comment

techniques

Nickel

Copper

Figure 7 Example of single irradiation-induced cluster approximately 3nm in diameter.

with Mn, Ni, and Si; there is also evidence of an association with P near the cluster-matrix interface (not shown). Similar cluster compositions have been found by other workers using other atom probes and FEGSTEM (which is not sensitive to Fe in clusters), and there is good agreement between the techniques.

The atom probe data indicate that the clusters are dilute in that the majority element is Fe.57-59 In contrast, analysis of SANS data is often undertaken assuming that irradiation-induced clusters are non­magnetic (implicitly Fe-free). The assumption of low levels of Fe in the Cu clusters was supported by thermodynamic calculations that predicted low levels of Fe in such precipitates.47 Furthermore, PAS-OEMS on as-irradiated (290 °C) Fe-0.9Cu or Fe-0.9Cu-1Mn indicate that the clusters in this con­dition are not magnetic and therefore very unlikely to contain Fe,60 although it should be noted that clusters in model alloys may be different from those found in RPV steels. Carter demonstrated that the scattering observed in SANS experiments is not inconsistent with the presence of magnetic clusters.57 More recently, Morley et al61 attempted to characterize the extent of trajectory aberrations in the atom probe that might give rise to an incorrect estimate of the true Fe content in small clusters in thermally aged (330-405 °C) RPV steels. He concluded that the clusters do contain Fe, but the levels are lower than those measured. Furthermore, he found that the con­centration of Fe in the precipitate phase is a function of aging temperature with less Fe at higher aging temperatures. Consensus has not yet been reached on the precise Fe content in irradiation-induced clus­ters in ferritic steels and is the subject of ongoing research.

Abbreviations

CTT Continuous cooling transformation

CEN-SCK Centre d’Etude de l’energie Nucleaire — Studiecentrum voor Kernenergie

4.08.1 Introduction

Recent progress in oxide dispersion strengthened (ODS) steels produced by mechanical alloying (MA) techniques allows them to be used as fuel cladding in sodium-cooled fast reactors (SFR). The thermally stable oxide particles dispersed in the ferritic matrix improve the radiation resistance and creep resistance at high temperature. As a result, ODS steels have a strong potential for high burnup (long-life) and high — temperature applications typical for SFR fuels. The attractiveness of ODS steels is due not only to the nanosize oxide particles composed of Y—Ti—O atoms but also to their controlled micron-size grain mor­phology. We review existing knowledge on the crys­talline structure and lattice coherency of these nanosize particles with their surrounding matrix, since these factors dominate the dispersion and strength-determining mechanism through dislocation interaction. The development of manufacturing pro­cesses is a principal issue for hardened ODS steels to realize long, thin-walled ODS steel cladding on pro­duction scales. There was the long-standing problem in low hoop strength due to the extremely elongated fine grains parallel to the rolling direction. To soften hard­ened cold-rolled products and modify their grain morphology, martensitic 9Cr-ODS steels and ferritic 12Cr-ODS steels have been developed. Current prog­ress in the development of these ODS steel claddings, including their relevant mechanical properties, for example, tensile and creep rupture strengths in the hoop directions, and irradiation performance, is reviewed. The development of Al-added high Cr — ODS steel cladding is also addressed, with a focus on superior resistance to oxidation and corrosion in a lead — bismuth eutectic (LBE), and supercritical pressurized water (SCPW) in the international Generation IV advanced nuclear power system. Nanocluster ODS steels,1 for example, 14YWT, etc., for fusion blanket structure materials, are not addressed in this chapter.

4.09.2.1.1 Elastic stress

Stresses in welds can be determined via several computational and experimental techniques. Compu­tational methods are generally based on finite element methods’ while experimental techniques include X-ray and neutron diffraction, hole drilling, and sur­face deformation mapping (e. g., slitting). Details of the application of these techniques can be found in several research proceedings55—57 and recent books.58’59

4.09.2.1.2 Plastic strain

The evolution of automated electron backscatter dif­fraction analysis has made the mapping and quantifi­cation of plastic strains in welds accessible via the scanning electron microscope.43’60-64 Strains can be visualized qualitatively via the intragrain misorienta — tion (Figures 11 and 19(c)) of the diffraction pattern or quantitatively (Figure 13) via the average intra­grain misorientation (i. e., the ‘AMIS’ parameter) of many grains and an appropriate calibration curve. Calibration curves from uniaxial tensile samples for several nickel-based alloys are given in Figure 16.

For reference, the measured plastic strain in sev­eral different welds and a heat-affected zone are compared in Table 2. Appreciable plastic strains (2-4%) occur even in unconstrained bead-on-plate welds and a wide range of strains (^2% to almost 30%) may be found, depending on the precise weld geometry, constraint, and welding practice. An example of the effect of welding practice is

given in Figure 17 for a 2 in. thick, Alloy 690 narrow groove weld made with EN82H filler metal via automatic gas tungsten arc welding (A-GTAW) 42,43 If welded with no ‘repairs’ (i. e.,
autogenous remelting of beads to improve bead — to-bead tie-in), it shows ^5.5% plastic strain near the weld root. This plastic strain increases if the beads above the weld are remelted as shown in the

graph with strains of ~11.5%, 15.0%, and 16.5% with 1,2, and 3 simulated ‘repairs’ above the weld. As expected, high levels of plastic strain lead to increased yield strength, decreased ductility, and increased susceptibility to stress corrosion cracking.

As supplied, graphite components contain a significant amount of both open and closed porosity in a variety of shapes and sizes, from the nm scale to the mm scale, as illustrated in Section 4.11.3. The open pore volume (OPV) is defined as the volume of pores accessible to helium, closed pore volume (CPV) is the volume of pores not accessible to helium, and the total pore volume (TPV) is the volume of open and closed pore.

The effect of pore size on the radiolytic oxidation rate was investigated by Labaton et al11 who found

the maximum range to be 1.5-5 pm. Taking this into account and referring to eqns [I]—[III] above, the oxidation process will be expected to be more effi­cient in the smaller pores than in the larger pores.