• Nuclear Engineering and Technology
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• v.29 no.6
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• pp.488-499
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• 1997

The measurements of the moderator temperature coefficient (MTC) are performed to demonstrate that the calculational model produces results that are consistent with the measurements. Since negative MTC is also a technical specification value that may limit the cycle length, it is important to measure it as accurately as possible. In this report, preferred choice of test method depending on the time in cycle, best power indication and temperature definition in MTC calculation were determined based on the MTC test results taken during initial startup testing and at 2/3 cycle burnup in the Yonggwang nuclear power plant. The results show that the ratio and rodded methods provided good agreement with the predictions during initial startup testing. However, near end-of-cycle the depletion method gives better results, and so is suggested to be used in the MTC measurements at 2/3 cycle burnup. The use of primary Delta T power as a power indicator in the MTC calculations is highly advisable since it responds with good consistent results very quickly to changes unlike secondary calorimetric power. For the appropriate temperature definitions used in the MTC calculations, it is considered that the arithmetic average temperature measured simply by inlet and outlet thermocouples is preferred. Although volumetric average temperature provides better results, the improvement is not sufficient to compensate for the simplicity of calculations by arithmetic average temperature.

• Nuclear Engineering and Technology
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• v.51 no.5
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• pp.1355-1364
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• 2019

A recently published nuclear forensics methodology for source discrimination of separated weapons-grade plutonium utilizes intra-element isotope ratios and a maximum likelihood formulation to identify the most likely source reactor-type, fuel burnup and time since irradiation of unknown material. Sensitivity studies performed here on the effects of random measurement error and the uncertainty in intra-element isotope ratio values show that different intra-element isotope ratios have disproportionate contributions to the determination of the reactor parameters. The methodology is robust to individual errors in measured intra-element isotope ratio values and even more so for uniform systematic errors due to competing effects on the predictions from the selected intra-element isotope ratios suite. For a unique sample-model pair, simulation uncertainties of up to 28% are acceptable without impeding successful source-reactor discrimination. However, for a generic sample with multiple plausible sources within the reactor library, uncertainties of 7% or less may be required. The results confirm the critical role of accurate reactor core physics, fuel burnup simulations and experimental measurements in the proposed methodology where increased simulation uncertainty is found to significantly affect the capability to discriminate between the reactors in the library.

• Nuclear Engineering and Technology
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• v.52 no.3
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• pp.538-543
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• 2020

Boric acid is used as a coolant for pressurized-water reactors, and the degree of burnup is controlled by the concentration of boric acid. Therefore, accurate measurement of the concentration of boric acid is an important factor in reactor safety. An improved system was proposed for the accurate determination of boron concentration. A new boron-concentration measurement technique, called multisensitive detection, was developed to improve the measurement accuracy of boron meters. In previous studies, laboratory-scale experiments were performed based on different sensitivity detectors, confirming a 65% better accuracy than conventional single-detector boron meters. Based on these experimental results, an experimental system simulating the coolant-circulation environment in the reactor was constructed; accuracy analysis of the boron meter with a multisensitivity detector was performed at the actual coolant pressure and temperature. In this study, the boron concentration conversion equation was derived from the calibration test, and the accuracy of the boron concentration conversion equation was examined through a repeatability test. Through the experiment, it was confirmed that the accuracy was up to 87.5% higher than the conventional single-detector boron meter.

• Journal of Powder Materials
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• v.11 no.4
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• pp.314-321
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• 2004

The elastic moduli of simulated dry process fuels with varying composition and density were measured in order to analyze the mechanical properties of a dry process fuel pellet. Resonant ultrasound spectroscopy(RUS) which can determine all elastic moduli with one set of measurements for a rectangular parallelepiped sample was used to measure the elastic moduli of UO$_{2}$ and simulated dry process fuel. The simulated dry process fuel showed a higher value of Young's modulus than UO$_2$ due to the presence of metallic precipitates and solid solution elements in the UO$_{2}$ matrix. The correlation between Young's modulus and porosity(P) of simulated dry process fuel was found to be 231.4-651.8 P (GPa) at room temperature. Dry process fuel with a higher burnup showed higher Young's modulus because total content of fission product element was increased.

• Nuclear Engineering and Technology
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• v.30 no.2
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• pp.173-179
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• 1998

The corrosion behavior of KOFA cladding which is a standard Zircaloy-4 manufactured by Westinghouse Specialty Metal Plant according to the Siemens/KWU's HCW (Highly Cold Worked) standard Zircaloy-4 specification was analyzed using the oxide measurement data of KOFA fuel irradiated in Kori-2 nuclear power plant. Analysis of the measured KOFA cladding oxidation showed that oxidation of KOFA cladding was lower than the design prediction based upon Siemens/KWU's HCW standard Zircaloy-4 cladding. Although the measured fuel rods have relatively low burnup and oxidation and the amount of the measured data are small, analysis of manufacturing and in-reactor operation conditions of KOFA cladding indicates that the differences in the manufacturing processes and chemical composition of the Siemens/KWU's HCW (Highly Cold Worked) standard Zircaloy-4 and KOFA cladding may have somewhat contributed to lower corrosion of KOFA cladding than the expected.

• Nuclear Engineering and Technology
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• v.53 no.11
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• pp.3563-3579
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• 2021

Measured decay heat data of light water reactor (LWR) spent nuclear fuel (SNF) assemblies are adopted to train machine learning (ML) models. The measured data is available for fuel assemblies irradiated in commercial reactors operated in the United States and Sweden. The data comes from calorimetric measurements of discharged pressurized water reactor (PWR) and boiling water reactor (BWR) fuel assemblies. 91 and 171 measurements of PWR and BWR assembly decay heat data are used, respectively. Due to the small size of the measurement dataset, we propose: (i) to use the method of multiple runs (ii) to generate and use synthetic data, as large dataset which has similar statistical characteristics as the original dataset. Three ML models are developed based on Gaussian process (GP), support vector machines (SVM) and neural networks (NN), with four inputs including the fuel assembly averaged enrichment, assembly averaged burnup, initial heavy metal mass, and cooling time after discharge. The outcomes of this work are (i) development of ML models which predict LWR fuel assembly decay heat from the four inputs (ii) generation and application of synthetic data which improves the performance of the ML models (iii) uncertainty analysis of the ML models and their predictions.

• Nuclear Engineering and Technology
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• v.55 no.10
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• pp.3525-3534
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• 2023

Non-uniformity of Pu and Cm composition in used nuclear fuel was analyzed to determine its effect on Pu accountancy in pyroprocessing, while employing the Pu-to-²⁴⁴Cm ratio method. Burnup simulation of a typical pressurized water reactor fuel assembly, required for the analysis, was carried out using MCNP code. Used fuel nuclide composition, as a function of nine axial and two radial meshes, were evaluated. The axial variation of neutron flux and self-shielding effects were found to affect the uniformity of Pu and Cm compositions and in turn the Pu-to-²⁴⁴Cm ratio. However, the results of the study showed that these non-uniformities do not affect the use of Pu-to-²⁴⁴Cm ratio method for Pu accountancy, if the measurement samples are drawn from the voloxidized powder at the feed step of pyroprocessing. 'Material Unaccounted For' and its uncertainty estimates are also presented for a pyrprocessing facility to verify safeguards monitoring requirements of the IAEA.

• Nuclear Engineering and Technology
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• v.21 no.4
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• pp.277-286
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• 1989

Destructive methods are used for the turnup determination of an irradiated PWR fuel. One of the methods includes U, Pu, Nd-148 and Nd-(145+146) determination by an isotope dilution mass spectrometry using triple spikes (U-233, Pu-242 and Nd-150). The method involves two sequential ion exchange resin separation procedures. Pu is eluted from the first anion exchange resin column (Dowex AG 1$\times$8) with 12 M HCl-0.1 M HI mixed solution, followed by U elution with 0.1 M HCl. Nd is isolated from other fission products on the second anion exchange resin column (Dowex AG 1$\times$4) with a nitric acid-methanol eluent. Each fraction is analysed by thermal ionization mass spectrometry. The difference between Nd-148 and Nd-(145+146) method is found with an average 2.07%. The results are compared with those by the heavy element method using U and Pu isotopes and by the destructive y-spectrometric measurement of Cs-137. The dependences of isotope composition of U and Pu on burn-up, and correlation between those isotopes are illustrated graphically.

• Nuclear Engineering and Technology
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• v.42 no.5
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• pp.576-581
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• 2010

KAERI (Korea Atomic Energy Research Institute) fabricated MOX (Mixed Oxide) fuel pellets as a cooperation project with PSI (Paul Scherrer Institut) for an irradiation test in the Halden reactor. The MOX pellets were fitted into fuel rods that included instrumentation for measurement in IFE (Institutt for Energiteknikk). The fuel rods were assembled into the test rig and irradiated in the Halden reactor up to 50 MWd/kgHM. The irradiated fuel rods were transported to the IFE, where ceramography was carried out. The fuel rods were cut transversely at the relatively higher burn-up locations and then the radial cross sections were observed. Micrographs were analyzed using an image analysis program and grain sizes along the radial direction were measured by the linear intercept method. Radial cracks in the irradiated MOX were observed that were generally circumferentially closed at the pellet periphery and open in the hot central region. A circumferential crack was formed along the boundary between the dark central and the outer regions. The inner surface of the cladding was covered with an oxide layer. Pu-rich spots were observed in the outer region of the fuel pellets. The spots were surrounded by many small pores and contained some big pores inside. Metallic fission product precipitates were observed mainly in the central region and in the inside of the Pu spots. The average areal fractions of the metallic precipitates at the radial cross section were 0.41% for rod 6 and 0.32% for rod 3. In the periphery, pore density smaller than 2 ${\mu}m$ was higher than that of the other regions. The grain growth occurred from 10 ${\mu}m$ to 12 ${\mu}m$ in the central region of rod 6 during irradiation.

• Nuclear Engineering and Technology
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• v.46 no.6
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• pp.837-846
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• 2014

A lead slowing down spectrometer (LSDS) is under development for analysis of isotopic fissile material contents in pyro-processed material, or spent fuel. Many current commercial fissile assay technologies have a limitation in accurate and direct assay of fissile content. However, LSDS is very sensitive in distinguishing fissile fission signals from each isotope. A neutron spectrum analysis was conducted in the spectrometer and the energy resolution was investigated from 0.1eV to 100keV. The spectrum was well shaped in the slowing down energy. The resolution was enough to obtain each fissile from 0.2eV to 1keV. The detector existence in the lead will disturb the source neutron spectrum. It causes a change in resolution and peak amplitude. The intense source neutron production was designed for ~E12 n's/sec to overcome spent fuel background. The detection sensitivity of U238 and Th232 fission chamber was investigated. The first and second layer detectors increase detection efficiency. Thorium also has a threshold property to detect the fast fission neutrons from fissile fission. However, the detection of Th232 is about 76% of that of U238. A linear detection model was set up over the slowing down neutron energy to obtain each fissile material content. The isotopic fissile assay using LSDS is applicable for the optimum design of spent fuel storage to maximize burnup credit and quality assurance of the recycled nuclear material for safety and economics. LSDS technology will contribute to the transparency and credibility of pyro-process using spent fuel, as internationally demanded.