Nuclear Materials

This paper provides an overview of the geochemical alteration of basaltic glass, considered for over 20 years as a suitable natural analogue for nuclear borosilicate glasses. The available data show that natural basaltic glasses may survive for million of years under subsurface conditions. Mineralogical studies show that an alteration layer called palagonite forms on the surface of the basaltic in response to the chemical attack by water. Under some environmental conditions, the alteration layer consists of an amorphous gel-like material, leading many authors to suspect hydration of the glass by water permeation and alkali interdiffusion. In other cases, the alteration layer is crystallized to some extent and contains clay minerals (smectite). Such layers are formed mainly on the younger natural glass samples (<1 My) by a process of coprecipitation of the elements dissolved from the glass. On samples older than 1 My, the alteration layers also contain zeolites. In the presence of these hydrated aluminosilicates, a hydrated residual glass is systematically observed, which thus forms as a consequence of interdiffusion processes. Leach tests conducted under controlled laboratory conditions at temperatures up to 200°C on both natural glass samples and synthetic basaltic glass provide identical kinetic results and alteration mechanisms. When compared with the data for SON68 nuclear borosilicate glass, the initial dissolution rates show the same activation energy (about 72 kJ mol À1) and consequently similar dissolution mechanisms. Moreover, when altered under static conditions at high reaction progress, both basaltic glass and nuclear glass have similar behavior characterized by a significant drop in the dissolution rate, up to three to five orders of magnitude lower than the initial dissolution rate. The time-dependence of the thickness of the altered layers measured on natural glass samples confirms this kinetic trend over time: the long-term dissolution rate is very low. This decrease may be related to diffusion mechanisms involving key chemical species and controlled by the mineralogy of the palagonite layer.

In modeling of release of a dust particle from a plasma-facing wall it is usually assumed that electron velocity distribution is Maxwellian. However, the absorption of fast electrons by the conducting wall can lead to truncation of fast component of reflecting electrons from the wall. In this work we study the effect of truncation of electron velocity distribution on the release condition of a conducting spherical dust particle from the plasma-facing wall. The truncation increases the electric field at the wall surface compared to that calculated in absence of the truncation. The stronger electric field makes the dust particle hard released when the gravitational force is directed from the wall and applied wall potential is shallower than the floating one.

Sodium n-propoxide and sodium iso-propoxides were synthesized and characterized. Thermal decomposition of these compounds was studied using thermogravimetric technique (TGA) coupled with mass spectrometry (MS) under non-isothermal and isothermal conditions. Various analytical techniques namely atomic emission spectroscopy (AES), infrared spectroscopy (IR), powder X-ray diffraction (XRD), elemental and volumetric analyses were employed to characterize these compounds and their decomposition residues. Kinetic parameters, namely, the activation energy and pre-exponential factor were deduced from the dynamic TGA and MS data. The activation energies derived from isothermal runs for the thermal decomposition of sodium n-propoxide and sodium iso-propoxide were 151.45 ± 2.16 and 128.07 ± 3.44 kJ mol À1 , respectively. Decomposition of sodium n-propoxide and sodium iso-propoxide results in the formation of gaseous products of saturated and unsaturated hydrocarbons leaving behind residue consisting of sodium carbonate, sodium hydroxide and free carbon.

The salient characteristics of martensite formation kinetics in 9Cr-1W-0.23V-0.063Ta-0.09C-0.02N (wt.%) reduced activation steel have been investigated using differential scanning calorimetry and supplemented by microscopy characterisation. Accurate determination of the martensite start (M S) and finish (M f) temperatures as a function of cooling rate (1-99 K min À1) and holding time (1-120 min) in the c-austenite phase field have been made. The critical cooling rate to form nearly 100% martensite is found to be of the order of 5-6 K min À1. The measured latent heat for the martensite transformation is found to be in the range 62-75 J g À1. In the case of austenitisation at 1323 K, it is found that the grain size of the parent austenite exhibited significant growth and the observed M S temperature too exhibited a systematic increase with austenite grain size. However, for those samples that are solution treated at a slightly lower temperature of 1253 K, the M S is found to be initially insensitive to the increase in holding time for up to about 30 min. This trend is followed later on by the usual increasing character of M S for further rise in the holding periods, up to about 120 min. This unusual initial behaviour is also reflected by the corresponding variation in the hardness values of the quenched in martensite. Moreover, the growth of parent austenite grains has also been restricted during the initial phase of solution annealing at 1253 K. This is attributed to the presence of undissolved carbide particles, which exert a pinning influence on the movement of austenite grain boundaries. The observed athermal martensite formation kinetics is described well by the Koistinen-Marburger relation.

Early interest in metallic plutonium fuels for fast reactors led to much research on plutonium alloy systems including binary solid solutions with the addition of aluminum, gallium, or zirconium and low melting eutectic alloys with iron and nickel or cobalt. There was also ...

The on-heating phase transformation temperatures up to the melting regime and the specific heat capacity of a reduced activation ferritic-martensitic steel (RAFM) with a nominal composition (wt%): 9Cr-0.09C-0.56Mn-0.23V-1W-0.063Ta-0.02N, have been measured using high temperature differential scanning calorimetry. The a-ferrite + carbides ? c-austenite transformation start and finish temperatures, namely Ac 1 , and Ac 3 , are found to be 1104 and 1144 K, respectively for a typical normalized and tempered microstructure. It is also observed that the martensite start (M S) and finish (M f) temperatures are sensitive to the austenitising conditions. Typical M S and M f values for the 1273 K normalized and 1033 K tempered samples are of the order 714 and 614 K, respectively. The heat capacity C P of the RAFM steel has been measured in the temperature range 473-1273 K, for different normalized and tempered samples. In essence, it is found that the C P of the fully martensitic microstructure is found to be lower than that of its tempered counterpart, and this difference begins to increase in an appreciable manner from about 800 K. The heat capacity of the normalized microstructure is found to vary from 480 to 500 J kg À1 K À1 at 500 K, where as that of the tempered steel is found to be higher by about, 150 J kg À1 K À1 .

Laser Raman spectroscopy and X-ray photoelectron spectroscopy were used to study the evolution of composition of oxide films in the presence of zirconia coatings on miniature HT-9 alloy specimens subjected to elevated temperature in air. The experiments expanded on previous efforts to develop a quickscreening technique for candidate alloys for cladding materials (HT-9) and actinide-based mixed oxide fuel mixtures (represented by the zirconia coating) by investigating the effect of both coating composition and annealing conditions on the high temperature reactions. In particular, the presence of the element Ga (a potential impurity in mixed oxide fuel) in the initial zirconia coating was found to accelerate the rate of oxide growth relative to that of yttria-stabilized zirconia studied previously. In addition, HT-9 samples that were subjected to different annealing conditions gave different results. The results suggest that the presence of Ga in a mixed oxide fuel will enhance the oxidation of HT-9 cladding under the conditions of this study, although the extent of enhancement is influenced by the type of annealing previously applied to the cladding material. Due to the multi-component layered structure of the oxidation products discovered in this fuel-cladding combination, it appears unlikely that Raman spectroscopy would be applicable as a stand-alone technique for the intended quick-screening approach to all but the simplest systems. Other techniques with better depth resolution, such as photoelectron spectroscopy, are required to follow the changes in composition of individual layers as they evolve during the oxidation process.

Preliminary investigations revealed that beryllides like Be 12 Ti may be much more suitable for neutron multipliers in future fusion power plants than pure beryllium. With suitable mechanical alloying and hot pressing methods, beryllium and beryllium-titanium tablets with grain sizes of 5 lm have been produced and analyzed by means of scanning electron microscopy (SEM), X-ray diffractometry, transmission electron microscopy (TEM) and plasma emission spectrometry. The fabricated tablets of Be-7.7 at.%Ti, corresponding to a chemical composition of the Be 12 Ti phase, have been arcmelted in an oxygen-free atmosphere. X-ray diffractometry, scanning and high resolution transmission electron microscopy measurements revealed that the vast majority of the ingots consist of nearly pure Be 12 Ti phase. Nineteen specimens have been fabricated from the ingots and shipped to HFR-Petten for the HIDOBE benchmark irradiation program that started in June 2005 with an aim to achieve 6000 appm He in Be and Be alloys at blanket relevant irradiation temperatures.

The formation and medium-term evolution of point defect and solute-rich clusters under neutron irradiation have been modelled in a complex Fe-CuMnNiSiP alloy representative of RPV steels, by means of first principle-based atomistic kinetic Monte Carlo simulations. The results obtained reproduce most features observed in available experimental studies, highlighting the very good agreement between both series. According to simulation, solute-rich clusters form and develop via an induced segregation mechanism on either the vacancy or interstitial clusters, and these point defect clusters are efficiently generated only in cascade debris and not Frenkel pair flux. The results have revealed the existence of two distinct populations of clusters with different characteristic features. Solute-rich clusters in the first group are bound essentially to interstitial clusters and they are enriched in Mn mostly, but also Ni to a lesser extent. Over the low dose regime, their density increases in the alloy as a result of the accumulation of highly stable interstitial clusters. In the second group, the solute-rich clusters are merged with vacancy clusters, and they contain mostly Cu and Si, but also substantial amount of Mn and Ni. The formation of a sub-population of pure solute clusters has been observed, which results from annihilation of the low stable vacancy clusters on sinks. The results indicate finally that the Mn content in clusters is up to 50%, Cu, Si, and Ni sharing the other half in more or less equivalent amounts. This composition has not demonstrated any noticeable modification with increasing dose over irradiation.

The irradiation response of PH13-8Mo stainless steel was measured up to 2.5 dpa at 200 and 300°C irradiation temperatures. The PH13-8Mo, a martensitic precipitation-hardened steel, was produced by Hot Isostatic Pressing at 1030°C. The fatigue tests (high cycle fatigue and fatigue crack propagation) showed a test temperature dependency but no irradiation effects. Tensile tests showed irradiation hardening (yield stress increase) of approximately 37% for 200°C irradiated material tested at 60°C and approximately 32% for 300°C irradiated material tested at 60°C. This contradicts the shift in reference temperature (T 0) measured in toughness tests (Master Curve approach), where the DT 0 for 300°C irradiated is approximately 170°C and the DT 0 for the 200°C irradiated is approximately 160°C. This means that the irradiation hardening of PH13-8Mo steel is not suitable to predict the shift in the reference temperature for the Master Curve approach.

The results of post-irradiation examinations of a pressure tube of fuel channel No. G-12 of KANUPP have been described. A detailed study was made in Canada by AECL. A parallel investigation on its seven rings of about 50 mm length each was also carried out at PINSTECH. Visual inspection showed normal oxidation effects. Gamma spectrometry showed the presence of 95Zr and 95Nb. Microstructural study revealed the characteristic alpha plus a transformed beta phase structure.

Long term tritium retention is one of the most critical issues for ITER and future fusion devices. While a global analysis of the T retention can be made by T accountancy in the activated phase of ITER, fuel retention and control must be already addressed in the non-activated phase, to identify the mechanism, location and amount of retention, its dependence on plasma and wall conditions and to qualify T retention mitigation and control techniques. For this purpose a new diagnostic, laser induced desorption spectroscopy of retained fuel has been developed and applied in TEXTOR. Hydrogen isotopes are desorbed from re-deposited layers on graphite plates by rapid heating with laser radiation. The released particles have been quantified in situ by spectroscopic measurements of hydrogen lines in a tokamak plasma. The results were validated by ex situ analysis of the hydrogen content of deposited a-C:H layers.

The behavior of argon and nitrogen, injected into single null diverted Ohmic discharges, has been compared. The screening of the divertor and scrape-off layer plasma is studied by comparing the number of impurity atoms in the core with the number injected by a calibrated gas puff. The effect of the poloidal position of injection on screening has been investigated. The core nitrogen density is dependent of an poloidal position of injection while the argon density is not. The spatial distribution of low charge states in the divertor has also been studied using a multichord visible spectrometer. Initial results of nitrogen modeling with the Monte Carlo code DIVIMP are presented.

An alteration phase that formed during the corrosion of commercial oxide spent nuclear fuel has been characterized with analytical transmission electron microscopy (AEM). The phase is a Cs-Ba uranyl molybdate oxide hydrate that has an orthorhombic structure related to the alkaline earth uranyl oxide hydrates of the protasite-group minerals. On the basis of the compositional analysis and a proposed model of the structure, the ideal structural formula is (Cs0.8Bao.6)(UOE)5(MOOE)O4(OH) 6 • nH20 (where n is around 6). Low levels of strontium are also present in the phase. The estimated unit cell parameters are a = 0.754 nm, b = 0.654 nm, and c = 3.008 nm. Although many of the phases formed during corrosion of spent oxide fuel are similar to those observed in natural uraninite deposits, such as Pefia Blanca in Mexico, there are important differences owing to the presence of fission products in the spent fuel. Thus, accurate determination of corrosion processes in actual radioactive waste forms is important. This study suggests that the natural U-Mo deposits at Shelby, WY, and Bates Mountain Tuff, NV, may be good analogues for the long-term behavior of U-Mo phases formed due to spent fuel corrosion.

Batch precipitation of ammonium diuranate (ADU) was carried out at various temperatures and at different flow rates of ammonia and their effects on the characteristics of ADU such as the nature of settling, moisture content, ammonia-to-uranium mole ratio and nitrate content have been studied. The effect of temperature of precipitation and the technological significance of savings in terms of heating requirements and quantity of precipitating agent ammonium hydroxide has been brought out. Also, the influence on other mentioned characteristics has been viewed in this angle. An attempt is made to explain the agglomeration of ADU semi-quantitatively, in terms of second order reaction kinetics by classifying the conditions of precipitation into two distinct phenomena, namely, the diffusion and activation controlled mechanisms. The activation energy involved in agglomeration of ADU under the experimental conditions was also calculated.

The influence of nitrogen co-deposition in ITER mixed layers on deuterium retention and release dynamics was studied in situ by Nuclear Reaction Analysis (NRA) and thermodesorption spectroscopy. W:Al and W:Be mixed layers deposited by TVA method were used and D uptake was compared in order to verify the possibility of Al being used as proxy material for Be in experiments. Samples were exposed to neutral deuterium atom beam with fluence of 3.24×10 19 D/cm 2 (flux 4.5×10 14 D/cm 2 s) at 370 K and NRA with 3 He ions was used for depth profile analysis of the deuterium uptake after the exposure. For the investigation of deuterium release dynamics the samples were linearly heated to around 1100 K and during this process NRA spectra at 2.5 MeV 3 He energy were collected every minute. Complementary to NRA a quadrupole mass spectrometer was used, following masses 2, 3 and 4. The simulation of the deuterium thermal desorption was performed and desorption energies of different desorption sites were calculated. Nitrogen co-deposition in the mixed layers was found to have an important influence on deuterium retention, since the concentration of deuterium in the sample increased by a factor of 4.8 in the presence of nitrogen in the W:Al compared to the nitrogen-free sample.

Uniaxial creep-to-rupture tests were performed on T91 in air and in flowing lead-bismuth eutectic melts. Compared to specimens tested in air, the specimens tested in liquid-metal show: (i) strain and strain rate increase up to a factor of about 50 (strain rate); (ii) time-to-rupture decrease; (iii) rapid transition into the third creep stage at high stress (above 180 MPa). The analysis of the test results revealed several important surface phenomena, which lead to different behavior of the specimens tested in lead-bismuth eutectic melts compared to those tested in air. Under high stress, and therefore high strain, the crack propagation process is mostly controlled by the reduction of the surface energy due to Pb and Bi adsorption on the steel surface. Under low stress (140 and 160 MPa) and low strain, this process is delayed due to the competing mechanism of healing the oxide scale cracks.

A new methodology is proposed to correlate the upper shelf energy (USE) of full-size and subsize Charpy specimens of a nuclear reactor pressure vessel plate material. The methodology appears to be more satisfactory than those methodlogies proposed earlier. The USE was normalized by a normalization factor involving the dimensions of the Charpy specimen, the elastic stress concentration factor, and the plastic constraint at the notch root. The normalized values of the USE were found to be invariant with speciment size. In addition, it was also found that the ratio of the USE of unirradiated to that of irradiated materials was approximately the same for full-, half-, and third-size specimens. The ductile-to-brittle transition temperture (DBTT) increased due to irradiation at 150°C to a nominal fluence of 1.0 × 1019 n/cm2 (E >MeV) by 78, 83 and 70°C for full-, half-, and third-size specimens, respectively. These shifts in DBTT appeared to be independent of specimen size and notch geometry.

Mg–6Zn–5Al–4RE (RE= Mischmetal, mass%) alloy was prepared by metal mould casting method. The microstructure and mechanical properties of the as-cast and heat-treated alloys were investigated. The results show that the phase compositions of the as-cast state alloy are supersaturated solid solution α-Mg, lamellar β-Al12Mg17, polygonal Al3RE and cluster Al2REZn2 phases. The mechanical properties, especially the ultimate tensile strength and elongation of the alloy were significantly improved by the heat treatment. Fracture surface ...

Recently, in situ Raman setups under irradiation become available, and uranium dioxide is one of the most studied compounds. This paper reports the temperature dependence of the only Raman-active mode in fluorine UO2. Besides the interest in terms of phonon anharmonicity knowledge, this can be used to estimate the heating induced by irradiation: The frequency shift and width of the T2g peak can be used as an internal indirect probe of the local temperature in the 20–590 °C range.

As well as acting as a moderator and reflector, graphite is used as a structural component in many gas-cooled fission nuclear reactors. Therefore the ability to predict the structural integrity of the many graphite components which make up a graphite reactor core is important in safety case assessments and reactor core life prediction. This involves the prediction of the service

Gadolinium oxide homogeneously mixed with uranium oxide nuclear fuel is used as burnable poison in modem LWR. Solid solutions of UOz with Gd20, additions of between 5 and 10 wt% have been prepared according to the following routes: (a) mechanical mixing of both ceramic powders; (b) AUC and ADU coprecipitation. Microstructural analyses have been carried out by optical ceramography and electron microprobe for observation of heterogeneities. Sintered densities as well as O/U ratios of the ceramic pellets have also been measured.

Transition metal ferrocyanides have important applications in the selective removal of radioactive caesium from low level and intermediate level radioactive liquid waste streams. The microcrystalline nature of these materials renders them useless for application in column mode operations. Special preparation procedures have been developed to prepare granular solids by in situ precipitation of metal ferrocyanides on organic anion exchangers, which is suitable for column mode operations. The elemental compositions of the metal ferrocyanides precipitated inside the pores of anion exchanger were determined by analysing the dissolved samples using ICP-AES system and flame photometer. From the XRD and EDX analyses and the elemental composition of the synthesized materials, the nature of the compound formed inside the anion exchanger was found to be cobalt ferrocyanide. From SEM analysis of the samples, the particle size of the cobalt ferrocyanide precipitated inside the anion exchanger was found to be much less than that of cobalt ferrocyanide precipitated outside. The efficiency of these materials for removal of Cs was evaluated by measuring the distribution coefficient (Kd), ion exchange capacity and kinetics of Cs uptake. The Kd of the materials loaded on anion exchanger was found to be of the order of 10 5 ml/g. The Cs uptake kinetics of the materials loaded on anion exchanger was slower than that of precipitated materials. The ion exchange capacity of the cobalt ferrocyanide loaded on anion exchanger was found to be much higher than that of the precipitated cobalt ferrocyanide.

A self-consistent description (performed by means of the B2-EIRENE code package) of a hydrogen plasma including electrons, ions and neutral background gas is used to investigate the processes and plasma behavior under conditions expected in MAGNUM-PSI. Several cases varying in gas-puffing, pumping rate, and plasma parameters are simulated. In all cases a detached plasma regime is achieved. The plasma density increases considerably for higher neutral pressures up to (1-5) · 10 14 cm À3 . The particle flux to the target is $10 24 m À2 s À1 and the plasma heat flux is $10 MW m À2 . The latter is significantly reduced in front of the target due to electron and ion cooling resulting from ionization and dissociation of H 2 molecules, and charge exchange/elastic collisions. Under the conditions of investigation, the losses due to molecule activated recombination are dominant compared with 3-body recombination of atomic ions.

The influence of composition variations on long-term glass behavior was investigated for three nuclear glass composition domains: the French SON 68 (R7T7-type) glass, the Na-Mg borosilicate AVM glass and the aluminosilicate VRZ glass defined as part of the investigation of new containment matrices based on zirconolite (CaZrTi 2 O 7 ). The initial alteration rates for glasses from different domains are comparable. Conversely, the alteration kinetics at advanced stages of reaction progress are very different, with decreases in the rates corresponding to different kinetic profiles, i.e. altered thickness versus time. The altered glass thickness can depend on the initial alteration rate and especially on the decrease in the rate, or it can be determined by the high residual alteration rate. The variation of the alteration rates over time appears to be related to the alteration film that forms on the surface of the material in particular the presence of any secondary crystalline phases. For AVM glass, the high residual rate is attributed to phyllosilicate phases rich in magnesium.

The effective atomic numbers, Z eff of some glass systems with and without Pb have been calculated in the energy region of 1 keV-100 GeV including the K absorption edges of high Z elements present in the glass. Also, these glass systems have been compared with some standard shielding concretes and commercial window glasses in terms of mean free paths and total mass attenuation coefficients in the continuous energy range. Comparisons with experiments were also provided wherever possible for glasses. It has been observed that the glass systems without Pb have higher values of Z eff than that of Pb based glasses at some high energy regions even if they have lower mean atomic numbers than Pb based glasses. When compared with some standard shielding concretes and commercial window glasses, generally it has been shown that the given glass systems have superior properties than concretes and window glasses with respect to the radiation-shielding properties, thus confirming the availability of using these glasses as substitutes for some shielding concretes and commercial window glasses to improve radiation-shielding properties in the continuous energy region.

In the near surface of plasma facing materials, high concentrations of hydrogen and helium isotopes can build up, which will interact with the point defects resulting from the bombardment of the surface as well as with the impurities of the materials. It is important to develop an understanding of the evolution of W microstructure in such conditions and to be able to model this evolution. The task is very complex, as many elements have to be included in the model which must be all parameterized correctly. Isochronal annealings experiments are simple experiments which can help in the making of more complicated models. In this work, an object Kinetic Monte Carlo technique parameterized on ab initio calculations as been used to model He desorption in W. The He atoms and the self interstitial atoms have been found to be very mobile but they can bind quite strongly with impurities such as carbon or molybdenum atoms. The evolution of the number of defects in the Kinetic Monte Carlo simulation was found to be in good agreement with the resistivity changes observed during an He desorption experiment of above threshold He implantation in a thin wire of tungsten.

High-temperature solid state syntheses of monazite powders containing plutonium (III), plutonium (IV) and americium (III) were performed. Resulting powders were characterized by X-ray diffraction. Pu 3+ PO 4 was readily obtained as a single phase by heating a Pu 4+ O 2 -NH 4 H 2 PO 4 mixture under argon atmosphere. Traces of tetravalent plutonium phosphate Pu 4+ P 2 O 7 were detected when synthesized under air atmosphere. The incorporation of (Pu 4+ , Ca 2+ ) in the monazite structure was investigated under air and argon atmosphere. We showed that Pu 4+ is fully reduced in Pu 3+ under argon atmosphere whereas, under air, the compound with the formula Pu 3þ 0:4 Pu 4þ 0:3 Ca 2þ 0:3 PO 4 was obtained. Pure Am 3+ PO 4 was also synthesized under argon atmosphere. X-ray patterns revealed a complete amorphisation of the monazite structure at a relatively low cumulative alpha dose for 241 AmPO 4 .

Hot compression testing of Mo-TZM (Mo-0.5Ti-0.1Zr-0.02C) alloy was carried out between 600 and 900°C employing strain rates from 0.001 s À1 to 1 s À1. Both the constant strain rate and strain rate change test results showed that Mo-TZM possesses low strain rate sensitivity in this temperature range. Activation energy calculated by using strain rate change data and plotting temperature compensated strain rate (Z) vs. shear modulus corrected flow stress (r/G) was found to be 290 kJ/mol. Electron back scattered diffraction (EBSD) and transmission electron microscopic (TEM) results obtained from the rapidly cooled deformed specimens revealed the formation of subgrains. Flow stress-plastic strain results and misorientation angles between subgrains showed an anomalous behavior at 800°C.

The oxidation behavior of ferritic-martensitic (F-M) alloys in supercritical water (SCW) was studied in order to evaluate the suitability of these alloys for use in the Gen IV supercritical water reactor (SCWR) concept. A series of exposure tests in SCW were performed with three F-M alloys: T91, HCM12A, and HT-9. The effect of temperature was evaluated over the range of 400-600°C and the dissolved oxygen concentration was controlled at <10 ppb (deaerated condition), 100 and 300 ppb. The oxidation behavior was determined from weight gain measurements along with oxide structure analysis. The results indicated that the oxidation rate was strongly dependent on temperature and followed an Arrhenius behavior. Activation energies for oxidation were 172, 177, and 189 kJ/mol for HT-9, HCM12A, and T91, respectively. The time dependence of the oxidation rate followed an exponential law with time exponents $0.3-0.42. Reduction in oxidation rate was observed at intermediate values (100-300 ppb) of dissolved oxygen concentration. The oxide formed on the alloy surface consisted of an outer layer of porous magnetite (Fe 3 O 4 ) and an inner layer of iron chromium oxide, (Fe, Cr) 3 O 4 with spinel structure. A transition region lies beneath the inner oxide in which the metal content increases to bulk values and the oxygen content decreases to nearly zero. Iron chromium oxide, (Fe, Cr)O, with the wustite structure was observed in the transition layer at 600°C. The relatively good agreement between the activation energies for oxidation and that for grain boundary diffusion of oxygen support an oxidation mechanism based on short circuit oxygen diffusion to the oxide-metal interface.

A modified version of the concentration-dependent model (CDM) potential (A. Caro et al., Phys. Rev. Lett. 95 (2005) [1] has been developed to study defects in Fe-Cr for different Cr concentrations. A comparison between this new potential and DFT results for a variety of point defect configurations is performed in order to test its reliability for radiation damage studies. The effect of Cr concentration on the vacancy formation energy in Fe-Cr alloys is analyzed in detail. This study shows a linear dependence of the vacancy formation energy on Cr concentration for values above 6% of Cr. However, the formation energy deviates from the linear interpolation in the region below 6% Cr concentration. In order to understand this behavior, the influence of the relative positions between Cr atoms and vacant sites on the vacancy formation energy has been studied.

Tensile testing has been performed at 25 and at 400 °C on two ferritic/martensitic steels (JFMS and HT-9) after irradiation in FFTF to up to 70 dpa at 373–433 °C. As observed in previous studies, this range of irradiation temperatures has a significant effect on hardening. The ...

An X-ray diffraction method was applied to measure the mean crystallite size (MCS) of the UO,,, powder, where x varied from 0.10 to 0.20. The MCS determination of UO z+x powder plays an important role in the process control of UOz pellets productions. The method is based on the Fourier analysis of X-ray diffraction profiles corrected for instrumental line broadening. The MCS was derived after performing appropriate correction of the diffraction profile to minimize the effect due to the presence of very low amount of the U,09 phase in UOzcx powders.

The degradation in mechanical properties of Inconel 625 ammonia cracker tubes occurs during the service for long duration in heavy water plants. The present study brings out the possibility of using Poisson's ratio (derived from measurement of time of flight of ultrasonic waves) in combination with hardness measurements, as an effective non-destructive tool for assessment of in-service degradation of Inconel 625 cracker tubes and qualification of re-solution annealing heat treatment for their rejuvenation. Further, the study also indicates the feasibility of extending the life of some of the tubes beyond the presently followed 120 000 h, before they are taken up for re-solution annealing, without affecting their serviceability. However, further studies are required to identify quantitative criterion for Poisson's ratio and hardness values, for deciding on the basis for removal of the tubes for rejuvenation.

In the Generation IV Materials Program cross-cutting task, irradiation and testing were carried out to address the issue of high temperature irradiation effects with selected current and potential candidate metallic alloys. The materials tested were (1) a high-nickel iron-base alloy (Alloy 800H); (2) a nickel-base alloy (Alloy 617); (3) two advanced nano-structured ferritic alloys (designated 14YWT and 14WT); and (4) a commercial ferritic-martensitic steel (annealed 9Cr-1MoV). Small tensile specimens were irradiated in rabbit capsules in the High-Flux Isotope Reactor at temperatures from about 550 to 700°C and to irradiation doses in the range 1.2-1.6 dpa. The Alloy 800H and Alloy 617 exhibited significant hardening after irradiation at 580°C; some hardening occurred at 660°C as well, but the 800H showed extremely low tensile elongations when tested at 700°C. Notably, the grain boundary engineered 800H exhibited even greater hardening at 580°C and retained a high amount of ductility. Irradiation effects on the two nano-structured ferritic alloys and the annealed 9Cr-1MoV were relatively slight at this low dose.

The consequences of irradiation damage in austenitic stainless steels on their mechanical properties, namely the yield stress, are investigated both experimentally and theoretically. The observed hardening is correlated with the quantitative characteristics of irradiation defects population. A simple model allowing for the defaulting of Frank loops under stress predicts the hardening and its saturation at large doses.

Zircaloy cladding oxidation is mostly represented by parabolic rate correlation. But the correlation approach is not suitable for long-term isothermal oxidation or oxidation occurs under steam starvation conditions and cannot obtain the detailed oxygen distribution which impacts the detailed mechanical behavior. To obtain the detailed oxygen distribution, a multi-phase diffusion problem with moving boundaries was introduced to simulate the cladding oxidation [9,10]. However, the hysteresis phenomenon related to the coexistence of monoclinic-tetragonal phases of zirconia which are very important to model the cladding oxidation during a LOCA, is not analyzed. In this study, a cladding oxidation model based on diffusion equations in the temperature range from 923 K to 2098 K which contains b-Zr, a-Zr, monoclinic-ZrO 2 , tetragonal-ZrO 2 , and cubic-ZrO 2 is developed and the detailed oxygen distribution in the cladding could be obtained. It showed that the simulations of short-term and long-term isothermal oxidation, transient oxidation, and oxidation under steam starvation conditions were reasonable through comparing with the experimental data. We found that our model can give a reasonable simulation of the hysteresis phenomenon of monoclinic-tetragonal phase transformation during transient oxidation as well as a much better simulation of the hypothetical LOCA transient oxidation experiments in ORNL than that by the code based on the parabolic rate correlation. This indicates that the developed model can accurately simulate the cladding oxidation during a LOCA transient.

CuCrZr is a high copper alloy widely used as electrical and thermal conducting material, especially in heat exchangers in nuclear reactors. In this respect, the physical and fatigue properties of CuCrZr have been extensively studied. The electrochemical behavior of CuCrZr, on the other hand, has not been adequately investigated. In the present study, the effect of pH on the corrosion behavior of CuCrZr in aqueous solutions without and with chloride (0.6 M NaCl) was studied. The pH of the solutions is found to exert significant influence on the corrosion behavior of CuCrZr. In acidic solutions without chloride, the corrosion of CuCrZr is ascribed to active dissolution with soluble products. In neutral and alkaline solutions without NaCl, the presence of oxides on the surface of CuCrZr leads to a noble shift in corrosion potential and passivation results in increased corrosion resistance. In chloride solutions at various pH values, the chloride ions influence the formation of the surface layers and the anodic dissolution process during polarization. At high pH, CuCrZr shows significant passivity and high corrosion

During normal operation of PWRs, routine fuel rods failures result in release of radioactive fission products (RFPs) in the primary coolant of PWRs. In this work, a stochastic model has been developed for simulation of failure time sequences and release rates for the estimation of fission product activity in primary coolant of a typical PWR under power perturbations. In the first part, a stochastic approach is developed, based on generation of fuel failure event sequences by sampling the time dependent intensity functions. Then a three-stage model based deterministic methodology of the FPCART code has been extended to include failure sequences and random release rates in a computer code FPCART-ST, which uses state-of-the-art LEOPARD and ODMUG codes as its subroutines. The value of the 131 I activity in primary coolant predicted by FPCART-ST code has been found in good agreement with the corresponding values measured at ANGRA-1 nuclear power plant. The predictions of FPCART-ST code with constant release option have also been found to have good agreement with corresponding experimental values for time dependent 135 I, 135 Xe and 89 Kr concentrations in primary coolant measured during EDITHMOX-1 experiments.

Journal of Nuclear Materials 170 (1990) 169-177 North-Holland 169 URANIUM NITRIDE FUEL SWELLING CORRELATION Steven B. ROSS * and Mohamed S. EL-GENK Institute for Space Nuclear Power Studies, Chemical and Nuclear Engineering Department, The University ...

At temperatures above the (01+ p)/p transformation temperature for zirconium alloys, steam reacts with P-ZI to form a superficial layer of zirconium oxide (Z102) and an intermediate layer of oxygen-stabilized a-ZI. Reaction kinetics and the rate of growth of the combined (Z102 + (Y-ZI) layer for Zircaloy-2 and Zircaloy-4 oxidation in steam were measured over the temperature range 1050-1850°C. The reaction rates for both alloys were similar, obeyed parabolic kinetics and were not limited by gas phase diffusion. The parabolic rate constants were consistently less than those given by the Baker and Just correlation for zirconium oxidation in steam. A discontinuity was found in the temperature dependence of both the reaction rate and the rate of growth of the combined (ZrOz + or-ZI) layer. The discontinuity is attributed to a change in the oxide microstructure at the discontinuity temperature, an observation which is consistent with the zirconium-oxygen phase diagram.

The development of an intergranular stress corrosion crack initiation site in thermally sensitized type 304 austenitic stainless steel has been observed in situ in high temperature oxygenated water using digital image correlation of time-resolved optical observations.  The grain boundary normal stresses were calculated using the Schmid-Modified Grain Boundary Stress (SMGBS) model of Was et al, applying three-dimensional data for the grain boundary planes and grain orientations.  The initiation site coincided with the most highly stressed sensitised boundary, demonstrating the importance of the combined contributions to crack initiation of grain boundary structure and plastic strain incompatibility.

The influence of ageing heat treatment on alloy A-286 microstructure and stress corrosion cracking behaviour in simulated Pressurized Water Reactor (PWR) primary water has been investigated. A-286 microstructure was characterized by transmission electron microscopy for ageing heat treatments at 670°C and 720°C for durations ranging from 5 h to 100 h. Spherical c 0 phase with mean diameters ranging from 4.6 to 9.6 nm and densities ranging from 8.5 · 10 22 m À3 to 2 · 10 23 m À3 were measured. Results suggest that both the c 0 phase mean diameter and density quickly saturate with time for ageing heat treatment at 720°C while the c 0 mean diameter increases significantly up to 100 h for ageing heat treatment at 670°C. Grain boundary g phase precipitates were systematically observed for ageing heat treatment at 720°C even for short ageing periods. In contrast, no grain boundary g phase precipitates were observed for ageing heat treatments at 670°C except after 100 h. Hardening by c 0 precipitation was well described by the dispersed barrier hardening model with a c 0 barrier strength of 0.23. Stress corrosion cracking behaviour of A-286 was investigated by means of constant elongation rate tensile tests at 1.5 · 10 À7 s À1 in simulated PWR primary water at 320°C and 360°C. In all cases, initiation was transgranular while propagation was intergranular. Grain boundary g phase precipitates were found to have no significant effect on stress corrosion cracking. In contrast, yield strength and to a lesser extent temperature were found to have significant influences on A-286 susceptibility to stress corrosion cracking.

Recent research results obtained in Europe, Japan, China and the USA on reduced activation ferritic/martensitic (RAFM) steels are reviewed. The present status of different RAFM steel products (plate, powder HIPped steel, many types of fusion and diffusion welds, unirradiated and irradiated states) is sufficient to present a strong case for the use of the steels in ITER test blanket modules. For application in DEMO, more research is needed, including the use of the International Fusion Materials Irradiation Facility (IFMIF) in order to quantify the effects of large amounts of transmutation products, such as helium and hydrogen.

India's nuclear programme envisages a large-scale utilisation of thorium, as it has limited deposits of uranium but vast deposits of thorium. The large-scale utilisation of thorium requires the adoption of closed fuel cycle. The stable nature of thoria and the radiological issues associated with thoria poses challenges in the adoption of a closed fuel cycle. A thorium fuel based Advanced Heavy Water Reactor (AHWR) is being planned to provide impetus to development of technologies for the closed thorium fuel cycle. Thoria fuel has been loaded in Indian reactors and test irradiations have been carried out with (Th-Pu) MOX fuel. Irradiated thorium assemblies have been reprocessed and the separated 233 U fuel has been used for test reactor KAMINI. The paper highlights the Indian experience with the use of thorium and brings out various issues associated with the thorium cycle.