• Nuclear Engineering and Technology
• /
• v.15 no.4
• /
• pp.280-285
• /
• 1983

Effects of fission neutron spectra in the reactor calculations have been analysed through applications of several cases of spectra in the criticality calculations of fast critical assemblies. They were the application of Maxwellian or Watt-Cranberg type formulae, of region dependent spectrum, of composition dependent spectrum, of fission transfer matrix, and the effects due to the selection of nuclear temperature in Maxwellian formula.

• Nuclear Engineering and Technology
• /
• v.45 no.7
• /
• pp.827-838
• /
• 2013

Interaction layer growth between U-Mo alloy fuel particles and Al in a dispersion fuel is a concern due to the volume expansion and other unfavorable irradiation behavior of the interaction product. To reduce interaction layer (IL) growth, a small amount of Si is added to the Al. As a result, IL growth is affected by the Si content in the Al matrix. In order to predict IL growth during fabrication and irradiation, empirical models were developed. For IL growth prediction during fabrication and any follow-on heating process before irradiation, out-of-pile heating test data were used to develop kinetic correlations. Two out-of-pile correlations, one for the pure Al matrix and the other for the Al matrix with Si addition, respectively, were developed, which are Arrhenius equations that include temperature and time. For IL growth predictions during irradiation, the out-of-pile correlations were modified to include a fission-rate term to consider fission enhanced diffusion, and multiplication factors to incorporate the Si addition effect and the effect of the Mo content. The in-pile correlation is applicable for a pure Al matrix and an Al matrix with the Si content up to 8 wt%, for fuel temperatures up to $200^{\circ}C$, and for Mo content in the range of 6 - 10wt%. In order to cover these ranges, in-pile data were included in modeling from various tests, such as the US RERTR-4, -5, -6, -7 and -9 tests and Korea's KOMO-4 test, that were designed to systematically examine the effects of the fission rate, temperature, Si content in Al matrix, and Mo content in U-Mo particles. A model converting the IL thickness to the IL volume fraction in the meat was also developed.

• Nuclear Engineering and Technology
• /
• v.46 no.2
• /
• pp.159-168
• /
• 2014

In order to advance understanding of the breakaway swelling behavior of U-Mo/Al dispersion fuel under a high-power irradiation condition, the effects of fuel-matrix interaction on the fuel performance of U-Mo/Al dispersion fuel were investigated. Fission gas release into large interfacial pores between interaction layers and the Al matrix was analyzed using both mechanistic models and observations of the post-irradiation examination results of U-Mo dispersion fuels. Using the model predictions, advantageous fuel design parameters are recommended to prevent breakaway swelling.

• Nuclear Engineering and Technology
• /
• v.55 no.10
• /
• pp.3732-3753
• /
• 2023

The Real-time Analysis for Particle-transport and In-situ Detection (RAPID) Code System, developed based on the Multi-stage Response-function Transport (MRT) methodology, enables real-time simulation of nuclear systems such as reactor cores, spent nuclear fuel pools and casks, and sub-critical facilities. This paper presents the application of a novel fission matrix-based burnup methodology to the well-characterized JSI TRIGA Mark II research reactor. This methodology allows for calculation of nuclear fuel depletion by combination and interpolation of RAPID's burnup dependent fission matrix (FM) coefficients to take into account core changes due to burnup. The methodology is compared to experimentally validated Serpent-2 Monte Carlo depletion calculations. The results show that the burnup methodology for RAPID (bRAPID) implemented into RAPID is capable of accurately calculating the k_eff burnup changes of the reactor core as the average discrepancies throughout the whole burnup interval are 37 pcm. Furthermore, capability of accurately describing 3D fission source distribution changes with burnup is demonstrated by having less than 1% relative discrepancies compared to Serpent-2. Good agreement is observed for axially and pin-wise dependent fuel burnup and nuclear fuel nuclide composition as a function of burnup. It is demonstrated that bRAPID accurately describes burnup in areas with high gradients of neutron flux (e.g. vicinity of control rods). Observed discrepancies for some isotopes are explained by analyzing the neutron spectrum. This paper presents a powerful depletion calculation tool that is capable of characterization of spent nuclear fuel on the fly while the reactor is in operation.

• Nuclear Engineering and Technology
• /
• v.46 no.2
• /
• pp.147-158
• /
• 2014

In order to investigate how the microstructure of fuel/matrix-interaction (FMI) layers change during irradiation, different U-7Mo dispersion fuel plates have been irradiated to high fission density and then characterized using scanning electron microscopy (SEM). Specifially, samples from irradiated U-7Mo dispersion fuel elements with pure Al, Al-2Si and AA4043 (~4.5 wt.%Si) matrices were SEM characterized using polished samples and samples that were prepared with a focused ion beam (FIB). Features not observable for the polished samples could be captured in SEM images taken of the FIB samples. For the Al matrix sample, a relatively large FMI layer develops, with enrichment of Xe at the FMI layer/Al matrix interface and evidence of debonding. Overall, a significant penetration of Si from the FMI layer into the U-7Mo fuel was observed for samples with Si in the Al matrix, which resulted in a change of the size (larger) and shape (round) of the fission gas bubbles. Additionally, solid fission product phases were observed to nucleate and grow within these bubbles. These changes in the localized regions of the microstructure of the U-7Mo may contribute to changes observed in the macroscopic swelling of fuel plates with Al-Si matrices.

• Nuclear Engineering and Technology
• /
• v.54 no.8
• /
• pp.2792-2800
• /
• 2022

In the core of the WWR-K reactor, a long-term irradiation of tristructural isotopic (TRISO)-coated fuel particles (CFPs) with a UO₂ kernel was carried out under high-temperature gas-cooled reactor (HTGR)-like operating conditions. The temperature of this TRISO fuel during irradiation varied in the range of 950-1100 ℃. A fission per initial metal atom (FIMA) of uranium burnup of 9.9% was reached. The release of gaseous fission products was measured in-pile. The release-to-birth ratio (R/B) for the fission product isotopes was calculated. Aspects of fuel safety while achieving deep fuel burnup are important and relevant, including maintaining the integrity of the fuel coatings. The main mechanisms of fuel failure are kernel migration, silicon carbide corrosion by palladium, and gas pressure increase inside the CFP. The formation of gaseous fission products and carbon monoxide leads to an increase in the internal pressure in the CFP, which is a dominant failure mechanism of the coatings under this level of burnup. Irradiated fuel compacts were subjected to electric dissociation to isolate the CFPs from the fuel compacts. In addition, nondestructive methods, such as X-ray radiography and gamma spectrometry, were used. The predicted R/B ratio was evaluated using the fission gas release model developed in the high-temperature test reactor (HTTR) project. In the model, both the through-coatings of failed CFPs and as-fabricated uranium contamination were assumed to be sources of the fission gas. The obtained R/B ratio for gaseous fission products allows the finalization and validation of the model for the release of fission products from the CFPs and fuel compacts. The success of the integrity of TRISO fuel irradiated at approximately 9.9% FIMA was demonstrated. A low fuel failure fraction and R/B ratios indicated good performance and reliability of the studied TRISO fuel.

• Nuclear Engineering and Technology
• /
• v.53 no.8
• /
• pp.2616-2628
• /
• 2021

The results of effective irradiation swelling in a wide range of burnup levels are numerically obtained for an inert matrix fuel, which are verified with DART model. The fission gas swelling of fuel particles is calculated with a mechanistic model, which depends on the external hydrostatic pressure. Additionally, irradiation and thermal creep effects are included in the inert matrix. The effects of matrix creep strains, external hydrostatic pressure and temperature on the effective irradiation swelling are investigated. The research results indicate that (1) the above effects are coupled with each other; (2) the matrix creep effects at high temperatures should be involved; and (3) ranged from 0 to 300 MPa, a remarkable dependence of external hydrostatic pressure can be found. Furthermore, an explicit multi-variable mathematic model is established for the effective irradiation swelling, as a function of particle volume fraction, temperature, external hydrostatic pressure and fuel particle fission density, which can well reproduce the finite element results. The mathematic model for the current volume fraction of fuel particles can help establish other effective performance models.

• Nuclear Engineering and Technology
• /
• v.30 no.5
• /
• pp.452-461
• /
• 1998

The pin power distribution is determined by analyzing the rotational gamma scanning data for 36 element fuel bundle of HANARO. A fission monitor of Nb$^{95}$ is chosen by considering the criteria of the half-life, fission yield, emitting ${\gamma}$-ray energy and probability. The ${\gamma}$-ray spectra were measured in Korea Atomic Energy Research Institute(KAERI) by using a HPGe detector and by rotating the fuel bundle at steps of 10$^{\circ}$. The counting rates of Nb$^{95}$ 766 keV ${\gamma}$-rays are determined by analyzing the full absorption peak in the spectra. A 36$\times$36 response matrix is obtained from calculating the contribution of each rod at every scanning angle by assuming 2-dimensional and parallel beam approximations for the measuring geometry. In terms of the measured counting rates and the calculated response matrix, an inverse problem is set up for the unknown distribution of activity concentrations of pins. To select a suitable solving method, the performances of three direct methods and the iterative least-square method are tested by solving simulation examples. The final solution is obtained by using the iterative least-square method that shows a good stability. The influences of detection error, step size of rotation and the collimator width are discussed on the accuracy of the numerical solution. Hence an improvement in the accuracy of the solution is proposed by reducing the collimator width of the scanning arrangement.

• Nuclear Engineering and Technology
• /
• v.43 no.3
• /
• pp.309-316
• /
• 2011

Molybdenum is the most abundant fission product since its fission yield is equivalent to that of xenon, and it has a very special role in the chemistry of nuclear fuel because it influences the oxygen potential of $UO_2$ fuel. In this study, the distribution of molybdenum in spent $UO_2$ fuel specimens with 33.3, 41.0 and 57.6 GWd/tU burnup was measured by a LA-ICP-MS system and the reproducibility of the measured data was obtained. The Mo distribution was almost constant along the radius of a fuel except an increase at the periphery of the fuel. It showed a drop in reproducibility with relatively high deviation of measured values for the highest burnup fuel. To explain this, the state of molybdenum in a $UO_2$ matrix and its effect on the oxidation behavior of $UO_2$ were investigated. The low reproducibility was explained by the segregation of molybdenum, and the inhibition of oxidation by the molybdenum was also observed.

• Nuclear Engineering and Technology
• /
• v.28 no.3
• /
• pp.229-246
• /
• 1996

A mechanistic model has been developed to predict the release behavior of fission gas during steady-state irradiation of LWR UO$_2$ fuel. Under the assumption that UO$_2$ grain surface is composed of fourteen identical circular faces and grain edge bubble can be represented by a triangulated tube around the circumference of three circular grain faces, it introduces the concept of continuous formation of open grain edges tunnels that is proportional to grain edge swelling. In addition, it takes into account the interaction between the gas release from matrix to grain boundary and the reintroduction of gas atoms into the matrix by the irradiation-induced re-solution of grain face bubbles. It also treats analytically the behavior of intragranular, intergranular, and grain edge bubbles under the assumption that both intragranular and intergranular bubbles are uniform in both radius and number density. Comparison of the present model with experimental data shows that the model's prediction produces reasonable agreement for fuel with centerline temperatures of 1000 to 140$0^{\circ}C$, wide scatter band for fuel with centerline temperatures lower than 100$0^{\circ}C$, and underprediction for fuel with centerline temperatures higher than 140$0^{\circ}C$.