[en] Full text: Uranuim enrichment of homogenous and profiled fuel assemblies was verified by gamma spectrometry. The fuel rods in the assemblies contained uranium enriched to 1.6 to 4.4% among which those of 1.6, 2.4 and 3.6% were of homogenous composition. In one of the types of heterogenous (^profiled⁾ assemblies the enrichment in the central region is the highest (4%) and it is lower at the peripheral region (3.6 and 3.3%). In the other type it is 4.4% in the central part and 3.6 and 4% at the periphery. Three gamma detectors were used in the study two HPGe detectors of different resolution and efficiency and a CdZnTe (CZT) detector, which fits into the central tube of the assemblies In this way possible to obtain information from both the inner and outer parts of the assemblies. It proved to be possible to distinguish different types of assemblies within a reasonable measurement time(15-20 min). For the HPGe measurements the assemblies had to be lifted out from their storage rack, while when using the CZT detector, the assemblies could be left at their storage position. A routine project is going on for enrichment verification of -10% of the freshly arrived assemblies at a random basis in the fresh fuel storage under normal operational conditions, upon inspection, having minimum impact on the NPP work, it is also tested that 232U content of the examined assemblies is much smaller than what is expected for pellets made of reprocessed material.

[en] This paper reports on the results of the experiments performed on spent VVER-440 fuel assemblies at the Paks Nuclear Power Plant (NPP), Hungary. The fuel assemblies submerged in the service pit were examined by high-resolution gamma spectrometry (HRGS). The assemblies were moved to the front of a collimator tube built in the concrete wall of the pit in the reactor block at the NPP, and lifted down and up under water for scanning by the refueling machine. The HPGe detector was placed behind the collimator in an outside staircase. The measurements involved scanning of the assemblies along their length of all the 6 sides, at 5-12 measurement positions side by side. Axial and azimuthal burnup profiles were taken in this way. Assembly groups for measurements were selected according to their burnup (10–50 GWd/tU) and special positions (e. g. control assembly, neighbour of control assembly). Burnup differences were well observable between assembly sides looking towards the center of the core and opposite directions. Also, burnup profiles were different for control assemblies and normal (working) fuel assemblies. The ratio of the measured activities of Cs-134 and Cs-137 was evaluated by relative efficiency (intrinsic) calibration. Measurement uncertainty is around 3 %. Taking into account irradiation history and cooling time (i. e.the time elapsed since the discharge of the assembly out of the core), the activity ratio Cs-134/Cs-137 shows good correlation with the declared burnup.

[en] VVER-440 fuel assemblies sunk in the loading pit were measured by gamma spectrometry. The assemblies were moved to the front of a 1.6 m long collimator built in the concrete wall of the pit in the reactor block and lifted down and up under water for scanning by the refuelling machine. The HPGe detector was placed behind the collimator in an outside staircase. The measurements involved scanning of the assemblies along their length of all the 6 sides, at 5-12 measurement positions side by side. Axial an azimuthal burnup profiles were taken up in this way. Assembly groups for measurements were selected according to their burnup (10-45 GW.d.t^-1U) and special positions (e. g. control assembly, neighbor of control assembly). The ratio of the activities of Cs-134 to Cs-137 was found to be proportional to the burnup. Activity ratios were evaluated by intrinsic efficiency calibration. Burnup differences were well observable between assembly sides looking at center and opposite directions. Also, burnup profiles are different for control assemblies and normal fuel assemblies. Uncertainty is around 3%. Taking into account irradiation history and cooling time, the ratio Cs-134/Cs-137 shows good correlation with declared burnup. (author)

[en] Presently 19-element natural uranium fuel bundles are used in 220 MW(e) Indian PHWRs. The core average design discharge burnup for these bundles is 7000 MW.d.Te^-1U and maximum burnup for assembly goes upto of 15 000 MW.d.Te^-1U. Use of fuel materials like MOX, Thorium, slightly enriched uranium, etc., in place of natural uranium in 19-element fuel bundles, in 220 MW(e) PHWRs is being investigated to achieve higher burnups. The maximum burnup investigated with these bundles is 30 000 MW.d.Te^-1U. In PHWR fuel elements no plenum space is available and the cladding is of collapsible type. Studies have been carried out for different fuel element target burnups with different alternative concepts. Modification in pellet shape and pellet parameters are considered. These studies for the PHWR fuel elements/assemblies have been elaborated in this paper. (author)

[en] Enrichment of uniformly and non-uniformly enriched (“profiled”) fuel assemblies in a range of 1.6–4.4% was verified by gamma-ray spectrometry at a nuclear power plant (NPP). HPGe detectors and a CdZnTe (CZT) detector, the latter fitting into the central tube of the assemblies, were used for obtaining information from outer and inner fuel rods. A procedure which has minimal impact on the NPP work was developed for verifying freshly arrived assemblies under normal operational conditions, and is now in routine use. - Highlights: ► Homogeneous/profiled fuel assemblies' enrichment verified by HRGS and MRGS. ► Enrichment of fuel rods in a range of 1.6–4.4%. ► HPGe viewed assemblies from outside, CZT fitted into assemblies' central tube. ► Distinguishing different types of assemblies in 15–20 min measurement time. ► Routine fresh fuel enrichment verification under normal NPP operational conditions.