• Energy Research
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• 12. Responsible consumption close
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• Annals of Nuclear Energy close

Abstract This paper is concerned with the neutronics analysis of extremely simplified recycling technologies of spent fuels in a small breed-and-burn fast reactor (B&BR). The discharged fuels of the first generation B&BR, which achieved an average burnup of 160 GWd/MTHM, were used to construct a second generation B&BR core. Two types of high proliferation resistant recycling technologies, melt refining and the newly suggested super-simplified melt and treatment, were applied to process and treat the discharged fuels. Because the burnup profile of discharged fuels varies largely depending on its position in the core, the recycling of the discharged fuels was also carried out by grouping them into recycling regions including 1, 3, and 6 recycling regions. In this study, the core performance of the 2nd generation B&BR loaded with the recycled fuel, which was produced by different recycling technologies and recycling regions, was analyzed and compared. An optimum fuel loading scheme was also adopted to maximize the performance of the 2nd generation B&BR in terms of the burnup reactivity change, core lifetime, and power profiles.

Abstract The radiological environment around Wolsong NPP site has been monitored since Wolsong Unit 1 started commercial operation in 1983, as required by regulations in Korea. As a result of the periodic safety review on Wolsong Unit 3 & 4 in 2011, the regulator recommended that an assessment on the long-term accumulation trend of environmental radioactivity around Wolsong Unit 3 & 4 should be performed. In response to this, the assessment was performed based on environmental monitoring data and non-parametric statistical tool (i.e., Mann–Kendall Test). The environmental monitoring data included indicator organisms, soil and marine deposit which were usually monitored for verification of the typical radioactivity level. Most results of the assessment showed no trend and only a few results showed a slight increase. It was concluded that the radioactivity accumulation due to the operation of the plant was insignificant and thus no concern has been raised regarding any effect on the safety and health of the residents around Wolsong NPP site. It is expected that the result of this study could be utilized to establish the plan and policy for environmental monitoring at other NPP sites.

Abstract The oldest commercial reactor in South Korea, Kori Unit 1 Nuclear Power Plant (NPP), will be permanently shut down starting in 2017. Since decommissioning of NPPs will generate a huge amount of radioactive metallic waste, proper treatment for decommissioning wastes is one of the key factors to decommission a plant successfully. Many studies have been conducted and experience accumulated for waste minimization and the reduction of disposal volume. Melting technology is one of the proven technologies for effectively managing the metallic waste and has been adopted by many countries. However, a licensed melting facility to dispose of metallic wastes from the NPP decommissioning has not been implemented in Korea. The purpose of this study is to establish a management plan for metallic waste and to evaluate the preliminary doses according to the acceptance criteria and operations at the facility. Based on the available research papers and empirical data from domestic and international experience, the operational concept and the melting technology principle were established and its dose and volume reduction effects were evaluated. The concept of operating a melting plant is to finally free release the ingots after melting or free release after a planned storage period. Based on this concept, the radioactive concentration of the scrap metal which can be treated was derived, demonstrating that melting can be applied to low level and very low level wastes above the clearance level. The RESRAD-RECYCLE computer code was used to assess worker doses and was rated within the regulatory limits in most scenarios, except for slag workers. Factors affecting the dose, such as methods of radiation protection, operating method and working time, will be important factors in establishing the melting facility and the radiological limits should also be determined accordingly.

Abstract The Macro Level Grid scheme for the efficient application of the subgroup method is presented, that employs the number density consideration factor and the temperature consideration factor for the treatment of non-uniform number densities and temperature distributions in a core. This scheme provides the efficient resonance treatment in direct whole core calculations of power reactors that involve thermal feedback and isotopic depletion. The new method solves the subgroup fixed source problem only 8 times per energy group with 8 macroscopic subgroup levels, regardless of the number of resonance isotopes in the problem of interest. The escape cross section of each isotope is obtained by interpolation using the pre-calculated ones at the specified macroscopic subgroup levels. This scheme turns out to be superior to the conventional scheme in terms of computing time and accuracy. More than 30% of the computing time for fixed source problems is saved compared to the conventional one with negligible reactivity errors of about a few pcm, whereas the conventional one has consistent reactivity errors of about +60 ∼ +150 pcm for typical pin-cell problems in a light water reactor. Moreover, it turns out that the new method provides high accuracy not only for very heterogeneous uranium dioxide and mixed oxide pin-cell checkerboard problems, but also for the depletion calculation of a multi-assembly problem involving hot full power thermal feedback with significantly shortened times.

Abstract This work investigates the depletion capability implemented in lattice physics code STREAM for the prediction of pressurized water reactor (PWR) uranium dioxide (UO2) spent nuclear fuel (SNF) isotopic inventory. The validation of this capability is performed by comparison of STREAM calculation results to measured SNF assay data obtained from PWRs Takahama-3, Calvert Cliffs and GKN II. The depletion analysis is conducted with the ENDF/B-VII.0 library and uses a pin cell model of the fuel rods from which the fuel samples were taken. The Chebyshev Rational Approximation Method (CRAM) is used to solve the depletion equation with about 1300–1600 isotopes in the depletion chain. 16 actinides and 23 fission products are analyzed in 14 spent UO2 fuel samples. The actinides are isotopes of uranium, neptunium, plutonium, americium and curium. The fission products nuclides include isotopes of cesium, neodymium, europium, samarium as well as 106Ru, 144Ce, 155Gd, 99Tc, 90Sr, 109Ag, and 103Rh. The sensitivity of some of the nuclides to the details of the power history and the adjustment of the fuel sample burnup is discussed. The impact of using ENDF/B-VII.0 library instead of ENDF/B-VI.8 is also discussed. Most of the nuclides analyzed are well predicted within ±7% of the experiment for actinides and fission products. STREAM depletion results are also compared to the codes SWAT, HELIOS and SCALE results based on publicly available information in literature, to check the performance of STREAM relative to other codes for the prediction of SNF isotopic inventory. The comparison to other code systems shows that the implementation in STREAM is of comparable accuracy. Overall, this paper demonstrates that the depletion capability in STREAM can be reliably applied to predict the isotopic inventory of PWR UO2 SNF for burnup ranging from 14 to 54 GWd/t and initial enrichment ranging from 3.0 to 4.1 wt% 235U.

Abstract This study presents the first application of the advanced Rossi-alpha method (theoretically introduced by Kong et al., 2014) on the reactivity measurements in a research reactor: detector count signals at the Kyoto University Critical Assembly (KUCA) facility. The detector signals in the KUCA A-type core are analyzed by three subcriticality measurement methods: (1) Feynman-alpha (F-α) method, (2) Rossi-alpha (R-α) method, and (3) advanced Rossi-alpha (advanced R-α) method. Four cases are analyzed for two different subcritical states of the core and two different neutron source locations. Two different negative reactivity ρ values are obtained by the measurements of control rod worth and regarded as the reference reactivity values, comparing the results by the four methods. The F-α shows reactivity errors ranging between 7.1 and 7.3% due to its use of variance-to-mean ratios of detector count signals, which are not very sensitive to neutron background noise. However, the fitting uncertainties associated to the F-α results are large, ranging between 5.4 and 12.8% at one standard deviation. The R-α shows small fitting uncertainties ranging between 2.8 and 3.8%, although reactivity errors are in the range of 3.5–26.5% due to the neutron background noise. Finally, the advanced R-α that explicitly models the neutron background noise contrary to the previous methods shows the reactivity errors in the range of 1.0–11.8%, and provides the lowest uncertainties of the measured ρ in the range of 0.4–0.9%. In conclusion, among the four methods applied to the reactivity measurements at KUCA, the advanced R-α reveals the best accuracy with the lowest uncertainties.

Abstract This study is intended to suggest an ergonomic evaluation on the working postural comfort. This study issued for the first time a methodology in view of combination between visual field and comfort. Especially, the ergonomic evaluation using the virtual decommissioning environments is user-friendly because setup of physical mock-up environments is difficult. This study verified the front and standing postures are best working postures during movement under radiation environments of nuclear facilities. It is expected that this methodology will make it possible to establish the ergonomic plan for decommissioning of nuclear facilities and safety of decommissioning will be improved and also decommissioning costs also can be reduced.

Abstract In many industrial settings such as chemical plants and electrical power generating plants, operator performance measures have been used for multi-objectives such as control room design, human system interface evaluation, training and procedures. Because of multiple uses of performance measures, many methods and criteria for measuring human performance have been developed. Among these methods, functional performance measures and task performance measures are well known and widely used to measure human performance. Although these two methods are used independently for many industries, and most studies related to operators’ performance use one of these measures, there has been no research to directly identify the relationships between the two measures. In this paper, task performance and functional performance measures are compared and investigated in terms of whether they have any relationships under the same Interfacing System Loss of Coolant Accident (ISLOCA) scenario considered in the experiment. Furthermore, to gain realistic results, these two performance measures are applied to Nuclear Power Plant Main Control Room (NPP MCR) operators to measure their performances in a full-scope simulator. Finally, a regression line plotting the relation between task and functional performance score and the correlation of the two performance measures are investigated.

Abstract It is necessary to accurately predict the minimum point of pressure drop to ensure the safety of nuclear reactors. However, the non-uniform heat flux distribution along the transverse direction is encountered when the plate-type nuclear fuels are used. This study shows the effect of a transversely non-uniform heat flux on the minimum point of the pressure drop. The pressure drop-flow rate curve under the non-uniform heat flux was obtained by the experiment, and the trend of curve was different with the one of uniform heat flux case. Under the non-uniform heat flux, even when the inlet mass flow rate decreased, the value of the pressure drop was constant for a while with the development of a two-phase flow. With further reduction of inlet mass flow rate, the pressure drop started to decrease until the minimum point of the pressure drop was reached. Moreover, the inlet mass flow rate at the minimum point of pressure drop is much lower than that in the uniform heat flux case. For a detail analysis, the numerical approach is proposed along with the application of multi-channel concept. A single narrow rectangular channel is divided along the transverse direction, and the heat flux is given non-uniformly to the divided channels. Although the pressure drop is separately calculated for each divided channel, the mass is transferred between the channels. In the calculation, the mass flow rate is non-uniformly distributed in the transverse direction. If the mass flow rate is uniformly distributed, the non-uniform heat flux causes an unbalanced pressure drop because of the non-uniform distribution of void fraction. As a result, at the edges where the void fraction is high, the mass flow rate is transferred to the middle of channel to balance the pressure drop in transverse direction. When the void fraction in the middle becomes significantly large, the minimum point of the pressure drop can be obtained.