[en] The test rig IFA-657 contained four CT specimens prepared from high fluence, high yield strength 304 SS. Three of the CTs were prepared from Chooz A 304 SS with an irradiated YS of 890 MPa. The fluence for two of the CTs (CT3 and CT4) was 2.5 x 10²² n/cm² and the fluence of the third (CT2) was 1.2 x 10²² n/cm². The fourth CT (CT1) was prepared from Oskarshamn 2 304 SS (YS 745 MPa and fluence 0.9 x 10²² n/cm².). During in-pile testing over three 100-day irradiation cycles, the specimens were exposed to a coolant temperature of 335 C and PWR primary water chemistry conditions (with 2-3 ppm Li / 1000-1200ppm B and 2-3 ppm H₂). Some of the specimen cracking response was ambiguous and in order to screen the results, criteria were set for either selecting or rejecting the data. Crack growth data that were accepted were computed over time intervals > 100 hours, with rates > 10-8 mm/s and crack length increments > 0.01mm. Despite the screening, the crack growth responses of the three Chooz A CTs were not easily interpretable: CTs 2 and 3 were initially loaded a stress intensity level of ∼6.5-7.5 MPaνm and the dcpd measurements made during the first irradiation cycle indicated similar apparent growth rates (in the range 3x10^-8 - 1x10^-7 mm/s) for both specimens. Later post irradiation examination inspection of the fracture surfaces, however, did not show any evidence of environmentally assisted cracking having occurred in CT3. As the test progressed, the K level on CT2 was increased (to 8, 15, 18 and finally 37 MPaνm), with no apparent effect in producing enhanced growth rates. A limited data set was produced for CT4, with crack growth rates ranging from 2x10^-8 - 1x10^-7 mm/s being recorded at a K level of ∼14-15 MPaνm. Overall, little effect of K was observed for the Chooz A specimens and a possible explanation for the absence of a CGR-K dependency, yet to be confirmed, could be the presence of high Si segregation at the grain boundaries of this material. For both CT2 and CT4, load cycling and other load perturbations had a marked effect on cracking response. Small (of the order 025-0.5 MPaνm) increases in load resulted in intervals of rapid crack extension while cycling produced step-like increments in crack length. Similar observations have been reported in other studies on high yield strength materials and it remains unclear as to whether the steps are due to real, rapid crack advance or due to the breaking of uncracked ligaments of material. The most reasonable cracking response was obtained for CT1, prepared from the Oskarshamn 2 304L SS. At applied stress intensity levels of 13.6-17 MPaνm, crack growth rates of ∼2.4x10-8 to ∼1.2x10-7 mm/s were calculated for the specimen and the K dependency of K 2.6 is comparable to the K dependencies reported in the literature. The IFA 657 CTs exhibited growth rates that were similar to those reported for unirradiated cold worked materials of similar yield strength, tested at low potential. The crack growth rates measured on the specimens in this study were also compared with crack growth rate results from irradiated specimens that were tested in BWR conditions at high and at low potential. At low potential, the crack growth rates measured in IFA-657 were higher than those measured in BWR conditions, where temperatures are also lower; i.e. the higher rates measured for the PWR test are attributed primarily to the difference in test temperature. (Author)

[en] IFA-650.5, a test with pre-irradiated fuel in the Halden Project LOCA test series, was conducted on October 23rd, 2006. The fuel rod had been used in a commercial PWR and had a high burnup, 83 MWd/kgU. Experimental arrangements of the fifth test were similar to the preceding LOCA tests. The peak cladding temperature (PCT) level was higher than in the third and fourth tests, 1050 C. A peak temperature close to the target was achieved and cladding burst occurred at approx. 750 C. Within the joint programme framework of the Halden Project PIE was done, consisting of gamma scanning, visual inspection, neutron-radiography, hydrogen analysis and metallography / ceramography. An additional extensive PIE including metallography, hydrogen analysis, and hardness measurements of cross-sections at seven axial elevations was done. It was completed to study the high burnup and LOCA induced effects on the Zr-4 cladding, namely the migration of oxygen into the cladding from the inside surface, the cladding distension, and the burst (author)(tk)

[en] The LOCA experiment on the second rod (IFA-650.2) a fresh, low-tin Zr-4, pressurised PWR rod was carried out in May 2004. The main objective was to produce ballooning, to determine the time to burst and to assess the material oxidation and hydriding kinetics. The rod pressure at hot conditions and peak PCT were 70 bar and 1050 C, respectively. To document the effect of the LOCA testing on the cladding, rod 2 was subjected in PIE to visual inspection, profilometry and metallography. On visual inspection the clad shows a typical balloon. In the region of max ballooning the clad shows a 35 mm long, < 20 mm burst opening. In the balloon region the outer clad diameter increased by <35% and locally the wall thickness reduction is >55%. The entire rod is covered with a black oxide layer. Below and above the split opening the continuous oxide layer was 40 to 50μm both on water and fuel side of the clad. At the locations of the upper and lower cladding thermocouples the oxide thickness was in the range 24-27 μm. Widmanstaetten structure is seen in the bulk of the clad and confirms the high temperature experienced. A some 40μm wide, hard and brittle zone with oxygen rich globular α-grains is found both at the outer and the inner edge of the clad in the balloon region. The zone is widest near the axial split (crack). In the clad few, arbitrary oriented hydride platelets are observed in the balloon area. (Author)

[en] To study the thermomechanical behaviour, fission gas release, and fuel structural characteristics of high burnup MIMAS-MOX fuel two experimental rods, rods 5 and 6, were refabricated from fuel rods irradiated for four cycles in Gravelines-4 PWR, instrumented, pressurised with helium and in 1999 loaded into the burnup extension rig IFA 648.1 in the HBWR. In 2000, the rig was unloaded, and rod 6 was punctured, reinstrumented and repressurised with helium. The rods were then re-irradiated in the ramp rig IFA 629.3 to a final discharge burn up of .60 MWd/kg MOX. The rods were subjected to post irradiation examination (PIE). PIE showed that both high burnup rods remained intact after the stepwise power ramp up to 25 kW/m. Fission gas release and internal rod pressure, found from puncturing test and fission gas analysis, were consistent with in-pile measurement (rod 6). Fission gas release in rod 5 (not re-instrumented between IFA-648 and 629.3) was significantly higher than in rod 6. During ramping in IFA-629.3 the measured fuel temperature was higher in rod 5 than in rod 6. The burnup of rod 5 was also somewhat larger. Fission gas analysis showed an unusual high Xe/Kr ratio of .19.5 for both rods. Both fuel cross-sections showed an inhomogeneous microstructure with large Pu-islands (up to 300 μm) in the UO₂ matrix. The Puislands showed signs of a higher burn-up than the matrix (indicated by larger MFPs and larger porosity). Both fuel rods showed a rim structure and pellet clad mechanical interaction (PCMI) with chemical bonding between cladding and fuel. The waterside oxide layer of rod 5 (35-40μ) is somewhat thinner than that found on the rod 6. Permanent diameter increase was approximately 50 μ on rod 5 and 25 μ on rod 6. (Author)

[en] After more than 30 years of operation the lead shielded metallography hot cells needed a basic renewal and modernisation not least of the specimen preparation equipment. Preparation in hot cells of radioactive samples for metallography and ceramography is challenging and time consuming. It demands a special design and quality of all in-cell equipment and skill and patience from the operator. Essentials in the preparation process are: simplicity and reliability of the machines, and a good quality, reproducibility and efficiency in performance. Desirable is process automation, flexibility and an alara amounto of radioactive waste produced per sample prepared. State of the art preparation equipment for materialography seems to meet most of the demands, however, it cannot be used in hot cells without modifications. Therefore. IFE and Struers in Copenhagen modified a standard model of a Strues precision cutting machine and a microprocessor controlled grinding and polishing machine for Hot Cell application. Hot cell utilisation of the microcomputer controlled grinding and polishing machine and the existing automatic dosing equipment made the task of preparing radioactive samples more attractive. The new grinding and polishing system for hot cells provides good sample preparation quality and reproductibility at reduced preparation time and reduced amount of contaminated waste produced per sample prepared. the sample materials examined were irradiated cladding materials and fuels

[en] Thermal behaviour studies of gadolinia doped UO₂ fuel, as compared to UO₂ fuel, were performed in the HBWR in the test rig IFA-515. Two small diameter, hollow pellet fuel rods A1, with 11.5 wt% enriched UO₂, and A2, with 8 wt% Gadolina doped and 13 wt% enriched UO₂, irradiated from July 1994 to October 2000 to a final discharge burnup of 76 (A1) and 84 MWd/kgU (A2), were sent to Kjeller for post irradiation examination (PIE). Both rods were instrumented with Expansion Thermometers (ET). PIE was performed to obtain more information on fission gas release, fuel burnup, and structural characteristics of fuel and cladding. PIE included: axial gamma scanning - dimensional measurement - rod puncturing - fission gas analysis - burn up analysis - ceramography. The gamma activity and Cs isotope distributions measured were similar for both rods. The value measured for diametrical cladding creep-out was approximately 0.46% for rod A1 and 0.6% for rod A2. FGR was similar (-10%) for both rods, while the inventory of helium was somewhat larger in the UO₂ rod (A1). The Xe/Kr ratio was slightly higher in rod A2 with the higher burnup. Ceramography showed almost no cracks in A2, while A1 had circumferential and radial cracks. Both fuel rods revealed a closed cold gap and PCMI. Somewhat larger fuel swelling was seen in A2 as indicated by the diameter reduction of the pellet centre hole and the larger diameter. The two types of high burnup fuel showed different microstructures. A higher quantity of small roundish pores (evidence for retained fission gas) could be observed in A2. The pellet microstructure in A1 showed two radial characteristic zones and grain-growth, while A2, Gd-doped UO₂ fuel, exhibited only one zone across the radiuswith a characteristic 'rim structure'. (Author)

[en] The present leading option for plutonium disposition, either civilian or weapons Pu, is to burn it in LWRs after having converted it to MOX fuel. However, among the possible types of fuel which can be envisaged to burn plutonium in LWRs, innovative U-free fuels such as inert matrix and thoria fuel are novel concept in view of a more effective and ultimate solution from both security and safety standpoint. Inert matrix fuel is an non-fertile oxide fuel consisting of PuO₂, either weapon-grade or reactor-grade, diluted in inert oxides such as for ex. stabilized ZrO₂ or MgAl₂O₄, its primary advantage consisting in no-production of new plutonium during irradiation, because it does not contain uranium (U-free fuel) whose U-238 isotope is the departure nuclide for breeding Pu-239. Some thoria addition in the matrix (thoria-doped fuel) may be required for coping with reactivity feedback needs. The full thoria-plutonia fuel though still a U-free variant cannot be defined non-fertile any more because the U-233 generation. The advantage of such a fuel option consisting basically on a remarkable already existing technological background and a potential acceleration in getting rid of the Pu stocks. All U-free fuels are envisaged to be operated under a once-through cycle scheme being the spent fuel outlooked to be sent directly to the final disposal in deep geological formations without requiring any further reprocessing treatment, thanks to the quality-poor residual Pu and a very high chemical stability under the current fuel reprocessing techniques. Besides, inert matrix-thoria fuel technology is suitable for in-reactor MAs transmutation. An additional interest in Th containing fuel refers to applicability in ADS, the innovative accelerated driven subcritical systems, specifically aimed at plutonium, minor actnides and long lived fission products transmutation in a Th-fuel cycle scheme which enables to avoid generations of new TRUs. A first common irradiation experiment in Halden, aimed at assessing the basic under-irradiation behaviour of the new proposed fuels, is under preparation, in view of starting the irradiation beginning 2000 (author) (ml)

[en] Light water reactor cores may be subjected to thermal-hydraulic transients resulting in inadequate core cooling for short periods of time. In PWRs this would result in departure from nucleate boiling (DNB) and in BWRs the result would be short term dry-out at the fuel rod surface. Both situations lead to transitory temperature increases of the cladding and it is a safety requirement that after such an event reasonable fuel performance must be maintained up to the subsequent shutdown. In order to be able to assess post dry-out fuel performance, it is necessary to know what effect the transient heating has on the mechanical properties of irradiated cladding. To this end some out-of-pile experiments with pre-irradiated BWR fuel rod cladding have been earned out, which have yielded valuable information. In order to expand this database, and investigate any changes to the fuel as well as the clad, a series of dry-out experiments have been carried out at the Halden Project. The aim was to expose fresh and pre-irradiated fuel rods to short-term in-pile dry-outs of the type anticipated to occur after a pump trip in a BWR. This in-pile exposure was then to be followed by post irradiation examination (PIE) directed at discovering the microstructural changes induced in the fuel rod by an actual dry-out event and the consequence of these changes to fuel rod mechanical integrity. Further details are presented in the paper (author) (ml)