Reactivity and neutron emission measurements of highly burnt PWR fuel rod samples

Murphy, Michael; Jatuff, F.; Grimm, P.; Seiler, R.; Brogli, R.; Meier, G.; Berger, Hans-Dieter; Chawla, R.

doi:10.1016/j.anucene.2006.04.003

Cited by 18 publications

(6 citation statements)

References 3 publications

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“…The PIE data that are the integral parameters in this study come from the LWR Proteus Phase II (LWR-PII) experimental campaign [23,24]. The integral data were concentrations of FPs in spent fuel normalized to the total mass of uranium (all isotopes) in the fuel.…”

Section: Lwr Proteus Phase Ii: Pie Datamentioning

confidence: 99%

“…It is important to consider FYs in DA with PIE data because the FPs are highly sensitive to FYs and FYs can have large uncertainties. The DA framework was implemented around GEF and tested on CASMO-5M [22] simulations of the LWR-Phase II (LWR-PII) experimental campaign of Proteus [23,24]. The experiment, and simulations of it, was used to do DA for GEF's model parameters.…”

Section: Introductionmentioning

confidence: 99%

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Data assimilation of post-irradiation examination data for fission yields from GEF

Siefman

Hursin

Sjöstrand

et al. 2020

EPJ Nuclear Sci. Technol.

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Nuclear data, especially fission yields, create uncertainties in the predicted concentrations of fission products in spent fuel which can exceed engineering target accuracies. Herein, we present a new framework that extends data assimilation methods to burnup simulations by using post-irradiation examination experiments. The adjusted fission yields lowered the bias and reduced the uncertainty of the simulations. Our approach adjusts the model parameters of the code GEF. We compare the BFMC and MOCABA approaches to data assimilation, focusing especially on the effects of the non-normality of GEF’s fission yields. In the application that we present, the best data assimilation framework decreased the average bias of the simulations from 26% to 14%. The average relative standard deviation decreased from 21% to 14%. The GEF fission yields after data assimilation agreed better with those in JEFF3.3. For Pu-239 thermal fission, the average relative difference from JEFF3.3 was 16% before data assimilation and after it was 12%. For the standard deviations of the fission yields, GEF’s were 100% larger than JEFF3.3’s before data assimilation and after were only 4% larger. The inconsistency of the integral data had an important effect on MOCABA, as shown with the Marginal Likelihood Optimization method. When the method was not applied, MOCABA’s adjusted fission yields worsened the bias of the simulations by 30%. BFMC showed that it inherently accounted for this inconsistency. Applying Marginal Likelihood Optimization with BFMC gave a 2% lower bias compared to not applying it, but the results were more poorly converged.

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Section: Lwr Proteus Phase Ii: Pie Datamentioning

confidence: 99%

Section: Introductionmentioning

confidence: 99%

Data assimilation of post-irradiation examination data for fission yields from GEF

Siefman

Hursin

Sjöstrand

et al. 2020

EPJ Nuclear Sci. Technol.

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“…8,9) Phase II was related to the analysis of the reactivity effects associated with burnt samples introduced into a pressurised water reactor lattice. 10,11) Phase III concerned essentially the investigation of modern BWR-type SVEA-96 Optima2 assemblies, and, in particular, the effects due to the presence of part-length pins and the lattice response in a mixture of $34% D 2 O and $66% H 2 O to create reduced moderation conditions. 12) In the current study, an intensive interpretation of the Phase I Core 2A (I-2A) experiments is presented.…”

Section: Lwr-proteus Experimental Programmementioning

confidence: 99%

Experimental Validation of Reaction Rate Distributions for a SVEA-96+ BWR Assembly with Hafnium Control Blades

Plaschy

Murphy

Jatuff

et al. 2009

Journal of Nuclear Science and Technology

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The accurate estimation of reactor physics parameters related to the presence of cruciform absorber blades in boiling water reactors (BWRs) is important for safety assessment and for achieving flexible operation during the cycle. Characteristics that are affected significantly include distributions of the total fission (F tot ) and 238 U capture (C 8 ) rates for controlled regions. Representative experimental investigations have been performed in the framework of the LWR-PROTEUS programme. In particular, the LWR-PROTEUS I-2A experiments concerned the neutronics characterisation of a SVEA-96+ BWR assembly controlled with a hafnium (Hf) blade under full-density water moderation conditions. The current paper presents the comparisons of the measured F tot and C 8 pin-wise distributions with a variety of stochastic and deterministic calculations: (a) MCNPX2.5 using recent nuclear data libraries (JEFF-3.1, ENDF/B-VI.8, and JENDL-3.3), (b) PHOENIX4 using ENDF/B-VI.3, (c) BOXER using JEF-1, (d) CASMO4 using JEF-2.2, and (e) HELIOS1.6 using ENDF/B-VI.1. The calculation/experiment comparisons show standard deviations from 1.2% (MCNPX2.5) up to 1.9% (BOXER) for the prediction of the F tot distribution, the highest individual discrepancy (7.6% with BOXER) being seen close to the ''Hf-vertex.'' The C 8 comparisons show systematically better agreement than those of F tot , the lowest standard deviations being 1.0% (BOXER) and the highest only 1.4% (HELIOS). In addition, sensitivity studies highlight the greater importance of modelling aspects, compared with that of nuclear data libraries, for the achievement of satisfactory and validated F tot and C 8 predictions.

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“…The station is a structure consisting of a lead shielding enclosure, a tungsten collimator, a remote handling system with the associated control system, a high-purity germanium (HPGe) detector with a support for the cryostat, a system for the injection of an external gamma source, and a set of neutron detectors. It is operated in combination with the PROTEUS transport flask/sample changer, 10) which loads an individual fuel segment into the station from the top, after which a remotely controlled clamping system ensures that the sample is secured in an accurately known position by means of a 'V' channel (see Fig. 2).…”

Section: Tomographic Cell Stationmentioning

confidence: 99%

“…In this work, the applied procedures and results obtained for the gamma-transmission tomography conducted on burnt and fresh fuel samples, previously deployed in the LWR-PROTEUS Phase II experiments, 10) are presented. A fresh, 400-mm-long UO 2 segment containing 3.5% by weight 235 U and four burnt PWR UO 2 fuel rod segments also of 400 mm length, two with very high (52 GWd/t and 71 GWd/t) and two with ultrahigh (91 GWd/t and 126 GWd/t) pellet-averaged burnup values, have been investigated.…”

Section: Introductionmentioning

confidence: 99%

Nondestructive Determination of Fresh and Spent Nuclear Fuel Rod Density Distributions through Computerised Gamma-Ray Transmission Tomography

Caruso¹,

Murphy²,

Jatuff³

et al. 2008

Journal of Nuclear Science and Technology

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The morphology of single PWR UO 2 fuel rods has been investigated by employing computerised gamma-ray transmission tomography. Four highly burnt fuel rods of different burnups (from 52 GWd/t to 126 GWd/t) and a fresh rod have been investigated. This paper describes the tomographic station built to acquire the projections together with the related experimental procedure, followed by the description of the image reconstruction techniques implemented. The principal results obtained are presented and analysed at three levels. First, the derived linear attenuation coefficient matrix, which is very useful for the implementation of emission tomography, is discussed. Second, the variation of density as a function of rod radius is presented, showing an interesting morphological change of the fuel with burnup. Particularly for the highest burnup sample, the periphery is characterised by a lower density (a steep decrease of almost 10% from the adjacent region) characteristic of an ultrahigh-burnup high-porosity structure. Finally, the nondestructive derivation of the volume-averaged fuel density was performed, compared with other experimental data, and plotted as a function of burnup, so that an interesting linear relationship between density and burnup was established, providing a potential indicator for the noninvasive determination of nuclear fuel rod burnup.

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Reactivity and neutron emission measurements of highly burnt PWR fuel rod samples

Cited by 18 publications

References 3 publications

Data assimilation of post-irradiation examination data for fission yields from GEF

Data assimilation of post-irradiation examination data for fission yields from GEF

Experimental Validation of Reaction Rate Distributions for a SVEA-96+ BWR Assembly with Hafnium Control Blades

Nondestructive Determination of Fresh and Spent Nuclear Fuel Rod Density Distributions through Computerised Gamma-Ray Transmission Tomography

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