• Nuclear Engineering and Technology
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• v.43 no.6
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• pp.477-488
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• 2011

A formulation is proposed for modelling the process of intra-granular diffusion of fission gas during irradiation of $UO_2$ under both normal operating conditions and power transients. The concept represents a simple extension of the formulation of Speight, including an estimation of the contribution of bubble motion to fission gas diffusion. The resulting equation is formally identical to the diffusion equation adopted in most models that are based on the formulation of Speight, therefore retaining the advantages in terms of simplicity of the mathematical-numerical treatment and allowing application in integral fuel performance codes. The development of the new model proposed here relies on results obtained by means of molecular dynamics simulations as well as finite element computations. The formulation is proposed for incorporation in the TRANSURANUS fuel performance code.

• Nuclear Engineering and Technology
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• v.50 no.2
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• pp.259-267
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• 2018

Focused ion beam-scanning electron microscope and electron backscattered diffraction examinations were conducted in the center of a $73\;GWd/t_U\;UO_2$ fuel. They showed the formation of subdomains within the initial grains. The local crystal orientations in these domains were close to that of the original grain. Most of the fission gas bubbles were located on the boundaries. Their shapes were far from spherical and far from lenticular. No interlinked bubble network was found. These observations shed light on previous unexplained observations. They plead for a revision of the classical description of fission gas release mechanisms for the center of high burn-up $UO_2$. Yet, complementary detailed observations are needed to better understand the mechanisms involved.

• Nuclear Engineering and Technology
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• v.7 no.2
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• pp.119-126
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• 1975

On the assumption that the spectral distribution of prompt fission neutrons released from supercritical accidents can be expressed by the generalized Cranberg form with two spectral parameters, which is then transformed into the single parameter form, a variation of the fission spectrum-averaged cross-sections for some threshold reactions with varying the spectral parameter has teen calculated using an electronic computer. It appears that the average cross-sections are very sensitive to the spectral deformation, especially those for the detectors having the threshold at high neutron energy are high compared to those for the detectors of which the threshold energies are comparatively low.

• Nuclear Engineering and Technology
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• v.49 no.5
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• pp.1095-1099
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• 2017

A fission source can act as a stabilization element in coupled Monte Carlo simulations. We have observed this while studying numerical instabilities in nonlinear steady-state simulations performed by a Monte Carlo criticality solver that is coupled to a xenon feedback solver via fixed-point iteration. While fixed-point iteration is known to be numerically unstable for some problems, resulting in large spatial oscillations of the neutron flux distribution, we show that it is possible to stabilize it by reducing the number of Monte Carlo criticality cycles simulated within each iteration step. While global convergence is ensured, development of any possible numerical instability is prevented by not allowing the fission source to converge fully within a single iteration step, which is achieved by setting a small number of criticality cycles per iteration step. Moreover, under these conditions, the fission source may converge even faster than in criticality calculations with no feedback, as we demonstrate in our numerical test simulations.

• Nuclear Engineering and Technology
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• v.54 no.4
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• pp.1195-1205
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• 2022

Recent developments on spectral diffusion algorithms, i.e., algorithms which exploit the projection of the solution on the eigenfunctions of the Laplacian operator, demonstrated their effective applicability in fast transient conditions. Nevertheless, the numerical error introduced by these algorithms, together with the uncertainties associated with model parameters, may impact the reliability of the predictions on short-lived volatile fission product release from nuclear fuel. In this work, we provide an upper bound on the numerical error introduced by the presented spectral diffusion algorithm, in both constant and time-varying conditions, depending on the number of modes and on the time discretization. The definition of this upper bound allows introducing a methodology to a priori bound the numerical error on short-lived volatile fission product retention.

• Nuclear Engineering and Technology
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• v.56 no.2
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• pp.444-450
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• 2024

Based on multiple measurements of ionization loss, the Time Projection Chamber (TPC) combines strong tracking ability with particle identification ability in a large momentum range, which is an important advantage of TPC detection technology over traditional ionization measurement technology. According to these two characteristics of TPC, applying it to the measurement of fission cross-section can greatly improve the measurement accuracy. During the measurement of the fission cross-section, the number of target nuclei is required to be accurately measured. So this paper introduces a method for measuring the number of ²³⁵U target nuclei using a fission TPC system. The measurement result agrees with the reference value, and relative error is around 1 %.

• Nuclear Engineering and Technology
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• v.56 no.10
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• pp.4158-4168
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• 2024

The Advanced Fuels Campaign's Fission Accelerated Steady State Testing (FAST) approach at Idaho National Laboratory creates a benchmark for evaluating accelerated irradiation via control rodlets and advanced metal fuel alloys for sodium-cooled fast reactors (SFRs). FAST experiments have been developed to generate prototypic temperature conditions during steady state irradiations of scaled geometric fuel pins. This approach helps to attain higher burn ups at a much faster rate than previous irradiation tests. For this study, the results from profilometry, fission gas release, and metallography of a FAST experiment are presented. Profilometry determined 0 % effective strain in the rodlets. The fission gas release fraction was measured from puncture/collection analysis. Constituent redistribution was observed in two specimens despite the peak fuel temperatures being below the normal ranges in which redistribution is expected. Metallography of the two higher temperature specimens showed typical swelling with the solid pin closing the fuel-cladding gap and the annular specimen having a fully closed annulus. Additionally, metallography indicated no swelling, no redistribution, and a homogenous microstructure for specimens with lower irradiation temperature. Post irradiation examination of FAST rodlets generally showed the expected representative behavior of metallic fuels within SFRs.

• Proceedings of the Korean Nuclear Society Conference
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• 2003.10a
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• pp.113-113
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• 2003

• Nuclear Engineering and Technology
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• v.52 no.9
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• pp.2130-2137
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• 2020

The standard method for calculating anti-neutrino emissions from a reactor involves knowing the fractional fission rates for the most important fissioning nuclides in the reactor. To calculate these rates requires detailed reactor physics calculations based upon the reactor design, fuel design, burnup dependent fuel composition, location of specific fuel assemblies in the core and detailed operational data from the reactor. This has only been published for a few reactors during specific time periods, whereas to be of practical use for anti-neutrino reactor monitoring it is necessary to be able to predict these on the publicly available information from any reactor, especially if using these data to subtract the anti-neutrino signal from other reactors to identify an undeclared reactor and monitor its operation. This paper proposes a method to estimate the fission fractions for a specific reactor based upon publicly available information and provides a database based upon a series of spent fuel inventory calculations using the FISPIN10 code and its associated data libraries.

• Nuclear Engineering and Technology
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• v.52 no.9
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• pp.1974-1982
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• 2020

Numerical prediction of fission product plateout and circulating coolant activities under normal operating conditions is crucial in the design of a high temperature gas-cooled reactor (HTGR). The results are used for the maintenance and repair of the components as well as the safety analysis regarding early source terms under loss of coolant accident scenarios. In this work, a new computer code named POSCA (Plate-Out Surface and Circulating Activities) was developed based on a one-dimensional model to evaluate fission product plateout and circulating coolant activities within the primary circuit of a HTGR. The verification and validation of study for the POSCA code was done using available analytical results and two in-pile experiments (i.e., OGL-1 and VAMPYR-1). The results of the POSCA calculations show that POSCA is able to simulate plateout and circulating coolant activities in a HTGR with fast computation and reasonable accuracy.