Fission Rates Measured Using High-Energy Gamma-Rays from Short Half-Life Fission Products in Fresh and Spent Nuclear Fuel

Hanna Kröhnert
EPFL, 2011
EPFL thesis

Abstract

In recent years, higher discharge burn-ups and initial fuel enrichments have led to more and more heterogeneous core configurations in light water reactors (LWRs), especially at the beginning of cycle when fresh fuel assemblies are loaded next to highly burnt ones. As this trend is expected to continue in the future, the Paul Scherrer Institute has, in collaboration with the Swiss Association of Nuclear Utilities, swissnuclear, launched the experimental programme LIFE@PROTEUS. The LIFE@PROTEUS programme aims to better characterise interfaces between burnt and fresh UO2 fuel assemblies in modern LWRs. Thereby, a novel experimental database is to be made available for enabling the validation of neutronics calculations of strongly heterogeneous LWR core configurations. During the programme, mixed fresh and highly burnt UO2 fuel lattices will be investigated in the zero-power research reactor PROTEUS. One of the main types of investigations will be to irradiate the fuel in PROTEUS and measure the resulting fission rate distributions across the interface between fresh and burnt fuel zones. The measurement of fission rates in burnt fuel re-irradiated in a zero-power reactor requires, however, the development of new experimental techniques which are able to discriminate against the high intrinsic activity of the fuel. The principal goal of the present research work has been to develop such a new measurement technique. The selected approach is based on the detection of high-energy gamma-ray lines above the intrinsic background (i.e. above 2200 keV), which are emitted by short-lived fission products freshly created in the fuel. The fission products 88Kr, 142La, 138Cs, 84Br, 89Rb, 95Y, 90mRb and 90Rb, with half-lives between 2.6 min and 2.8 h, have been identified as potential candidates. During the present research work, the gamma-ray activity of short-lived fission products has, for the first time, been measured and quantitatively evaluated for re-irradiated burnt UO2 fuel samples with burn-ups of about 36 and 46GWd/t. Based on experiments carried out with these fuel samples in a reference test lattice of the PROTEUS reactor, fresh-to-burnt-fuel fission rate ratios could be determined. The 1σ uncertainties on the derived fission rate ratios are 1.7 to 3.4% and are mainly due to the statistical uncertainties. Calculated values of the fission rate ratios, as obtained using the Monte Carlo code MCNPX, have been shown to agree with the experimental results within these uncertainties. In deriving fresh-to-burnt-fuel fission rate ratios,142La and 138Cs have emerged as the preferred fission products. Their fission yields for the main fissile isotope (235U, 239Pu and 241Pu) are similar, which makes them relatively insensitive to the exact composition of the burnt fuel. Finally, a measurement station for the future LIFE@PROTEUS experiments has been proposed and evaluated, along with a detailed formulation of recommendations for optimised irradiation and measurement strategies. The estimated accuracy for the foreseen measurements of fission rate ratios between fresh and highly burnt fuel pins is 1 to 2%. The contribution of nuclear-data related uncertainties have been pointed out as possibly representing the main constraint on the achievable accuracy in future experiments. In brief, the present research work has established a novel experimental technique for measuring and comparing fission rates in fresh and highly burnt fuels in a zero-power research reactor such as PROTEUS. Moreover, possibilities have been presented for the further optimisation needed for a future, routine application of the technique.

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Hanna Kröhnert
EPFL, 2011
EPFL thesis

Abstract

Official source

https://infoscience.epfl.ch/record/154773?ln=en

About this result

Related concepts (28)

Related concepts

Related publications (40)

Related publications

Current fuel management strategies for light water reactors (LWRs), in countries with high back-end costs, progressively extend the discharge burnup at the expense of increasing the 235U enrichment of the fresh UO2 fuel loaded. In this perspective, standard non-destructive assay (NDA) techniques, which are very attractive because they are fast, cheap, and preserve the fuel integrity, in contrast to destructive approaches, require further validation when burnup values become higher than 50 GWd/t. This doctoral work has been devoted to the development and optimisation of non-destructive assay (NDA) techniques based on gamma-ray emissions from irradiated fuel. It represents an important extension of the unique, high-burnup related database, generated in the framework of the LWR-PROTEUS Phase II experiments. A novel tomographic measurement station has been designed and developed for the investigation of irradiated fuel rod segments. A unique feature of the station is that it allows both gamma-ray transmission and emission computerised tomography to be performed on single fuel rods. Four burnt UO2 fuel rod segments of 400 mm length have been investigated, two with very high (52 GWd/t and 71 GWd/t) and two with ultra-high (91 GWd/t and 126 GWd/t) burnup. Several research areas have been addressed, as described below. The application of transmission tomography to spent fuel rods has been a major task, because of difficulties of implementation and the uniqueness of the experiments. The main achievements, in this context, have been the determination of fuel rod average material density (a linear relationship between density and burnup was established), fuel rod linear attenuation coefficient distribution (for use in emission tomography), and fuel rod material density distribution. The non-destructive technique of emission computerised tomography (CT) has been applied to the very high and ultra-high burnup fuel rod samples for determining their within-rod distributions of caesium and europium fission-product radionuclides. As indicated above, results provided by the transmission tomography measurements were employed in the emission tomography reconstruction phase, together with a calculated global efficiency matrix and input sinograms derived from the processing of measured projections. Different tomographic algorithms were tested and "tuned", on the basis of known test distributions, before being applied to the actual fuel rod measurements. Amongst the various possibilities, the Paraboloidal Surrogates Coordinate Ascent (PSCA) penalised likelihood method has been chosen for presentation of the final results, because it ensures high precision, especially in resolving the most difficult peripheral regions of the rods. The results of the emission tomography have indicated large central depressions in the caesium distributions, but of varying extent from sample to sample. Particularly interesting is the case of the 126 GWd/t sample, showing a very deep central depression (a factor of ∼2.5 for 137Cs, a factor of ∼3 for 134Cs). Differences in the relative activity distributions of 137Cs and 134Cs have, in fact, been observed for all the samples. The depression of 134Cs is more marked than that of 137Cs, probably due to the different origins of the two isotopes. In contrast, the europium shows an almost flat distribution. In order to support the tomographically measured caesium distributions, the results of destructive chemical techniques applied on samples from the same fuel rod (126 GWd/t sample) were examined and found to show reasonably good agreement with the tomography, thus confirming the depressed distributions at the centre of the rod. In addition to the tomographic reconstructions, the present research has also investigated the possibility to use single isotope activities, and/or isotopic concentration ratios from 134Cs, 137Cs, and 154Eu, as burnup indicators at very high and ultra-high burnups. The corresponding non-destructive measurements, performed using three different approaches, have been compared with chemical assays, as well as with reactor physics calculations (CASMO). Whereas the chemical results confirm the current gamma-spectroscopic measurements within uncertainties, the agreement between measurements and calculations is not satisfactory. It has been shown that certain indicators, well established for application at low and medium burnups, suffer a serious loss of reliability at high burnup. Nevertheless, the possibility of successfully employing other burnup monitors has been clearly highlighted. Considering that the overall effort required for the destructive chemical analysis of fuel samples is much greater, the current investigations have clearly demonstrated that non-destructive gamma spectrometry, in conjunction with corresponding transmission and emission tomography measurements, can indeed be considered a valid approach for characterising fuel burnup in the high/ultra-high range.

EPFL2007

Appraisal of Core Analysis Methods against Zero-Power Experiments on Full-Scale BWR Fuel Assemblies

Flavio Dante Giust

The present doctoral research aims at the appraisal of nodal core simulators used for the calculation of commercial boiling water reactor (BWR) cores, against measurements carried out at the Paul Scherrer Institute under the LWR-PROTEUS experimental programme. The research focuses mainly on the prediction of radial and axial total-fission rate (i.e. power) distributions in the vicinity of core heterogeneities, caused by features such as the presence of control blades, enrichment boundaries or partial length rods. As such, this thesis complements previous validation work performed for LWR calculational methods on the basis of integral experiments in zero-power research reactors, the latter corresponding largely to the validation of two-dimensional, reflected fuel-assembly calculations. Two commercial code systems have been investigated currently: HELIOS/PRESTO-2, presently used for core monitoring at the Leibstadt Nuclear Power Plant (Switzerland), and CASMO-5/SIMULATE-5, which represents the most recent generation of the widely applied CASMO/SIMULATE system. Six different LWR-PROTEUS configurations have been modelled, for which the nodal reconstructed total-fission rate distributions have been compared against experimental results, as well as against reference, three-dimensional whole-reactor Monte Carlo calculations using the MCNPX code. To start with, a methodology has been developed and tested for appropriate representation of the multi-zone experimental configurations, as employed in the LWR-PROTEUS programme, by means of reduced-geometry models set up using the investigated nodal code systems. The approach adopted has been to apply, in each case, case-dependent three-dimensional boundary conditions to the nodal code's modelling of the 3x3 array of BWR assemblies constituting the LWR-PROTEUS test zone. This has been done in terms of so-called Partial Current Ratios (PCRs), which describe the relation between the incoming and outgoing neutron currents across the test-zone boundary. These PCRs have been derived from the fore-mentioned MCNPX calculation of the multi-zone PROTEUS reactor, thus allowing for the adequate description of the three-dimensional effects associated with the interaction between the test zone and its surroundings. The results obtained for the reference LWR-PROTEUS configuration, with a test zone consisting of an unperturbed regular array of identical, axially uniform BWR fuel assemblies (of type SVEA-96+), showed that both the investigated nodal methodologies reproduce the experimental total-fission rate distribution with very good accuracy, equivalent to that achievable with high-order transport calculations. This has provided adequate indication that the developed methodology of using three-dimensional PCRs, calculated by means of an appropriate whole-reactor MCNPX model, offers a reliable platform for the desired validation. The full insertion of a L-shaped hafnium control blade, especially fabricated for the LWR-PROTEUS programme, has permitted the study of the behaviour of calculated total-fission rate distributions in the presence of strong radial and azimuthal flux gradients. In this case, the use of reflective boundary conditions in the lattice calculations, combined with the azimuthal non-uniformity of the currents that couple the test zone with the outer zones of the reactor, produce systematic deviations which, together, lead to some loss of accuracy in comparison with the regular, uncontrolled case. The three-dimensional effects of the changes in enrichment and gadolinium content between two different axial sections of the fuel assembly have also been studied. In this case, the calculations of the axial total-fission rate distributions are seen to generally reproduce the experimental data with good accuracy, although certain systematic effects are observed. In particular, at local level, and very near the strong axial heterogeneity caused by the enrichment boundary, relatively large deviations have been found to occur. However, due to their very local character, these deviations are not considered to represent a relevant issue in connection with power reactor monitoring and design. Also for the demanding case of the LWR-PROTEUS configuration with a partially inserted control blade, the results obtained in this thesis show a very good behaviour. As in the boundary enrichment case, however, some very localized deviations occur in the vicinity of the strong axial flux gradients. In a separate phase of the LWR-PROTEUS programme, BWR assemblies with partial length rods (PLRs), viz. of type SVEA-96 Optima2, were studied in the central test zone. The results of the comparisons made currently for the corresponding LWR-PROTEUS configurations have shown that, also for fuel assemblies with axial heterogeneities of this type, the performance of both the investigated nodal methodologies is very satisfactory for predicting the global distribution of the total-fission rate. The local three-dimensional effects caused by the PLRs are also predicted with good accuracy, similar, in fact, to that obtained with high-order three-dimensional transport calculations. The present research shows that the two nodal code systems investigated – HELIOS/PRESTO-2 and CASMO-5/SIMULATE-5 – reproduce the LWR-PROTEUS experimental results with high accuracy, both for the radial and the three-dimensional (i.e. at pellet level) comparisons. Significant deviations occur almost exclusively within a short distance from the nodal interface, in cases featuring strong gradients across the core midplane. The axial and radial shapes of the total-fission rate distributions are well predicted in most cases, which represents an important observation concerning the monitoring of operational limits in power reactor cores. Thus, overall – within the range of applicability of the experimental conditions studied in the LWR-PROTEUS programme – the results of the comparisons performed in this thesis confirm that nodal methodologies with pin-power reconstruction have a very high level of performance. In fact, the accuracy that they can achieve for the assembly-internal total-fission rate distribution is, in general, higher than that practically achievable in an operating nuclear power plant. This is due to unavoidable, additional uncertainties in the characterisation of a BWR core under power reactor conditions. Finally, the present research has provided quantitative insights into the applicability of integral data, produced in critical facilities such as PROTEUS, for validating the calculation of power-reactor core heterogeneities. One has thus been able to demonstrate that the experimental evidence from the LWR-PROTEUS programme indeed constitutes a very valuable basis for the validation and appraisal of three-dimensional nodal methodologies with pin-power reconstruction.

EPFL2012

Inert matrix fuel for the utilisation of plutonium

Manuel Alexandre Pouchon

Utilising fuel resources responsibly, reducing waste volume and emissions as well as conflict potentials within the international community (non-proliferation, energy demand) are among the principles for the judgment of sustainable development. Utilising and burning plutonium in a light water reactor has been shown to be feasible for the disposition of the large amount of excess plutonium produced in today's power reactors and resulting from the disarmament efforts of the super powers. With regard to material technology aspects, efforts have concentrated on the evaluation of fabrication feasibility and on the determination of the physicochemical properties of a single phase zirconium/erbium/plutonium oxide material stabilised as a cubic solution by yttrium for plutonium and minor actinide incineration and transmutation. Due to the absence of uranium as origin for plutonium build-up, such a nuclear fuel is called Inert Matrix Fuel. Irradiation testing with a dedicated experiment in the material test reactor in Halden, Norway, is underway within the framework of the OECD-Halden Reactor Project.

Elsevier2001

EPFL2007

Appraisal of Core Analysis Methods against Zero-Power Experiments on Full-Scale BWR Fuel Assemblies

Flavio Dante Giust

EPFL2012