• Nuclear Engineering and Technology
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• v.46 no.3
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• pp.407-412
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• 2014

In this research, we investigated the burnup characteristics and the conversion of fertile $^{232}Th$ into fissile $^{233}U$ in the core of a Sodium-Cooled Fast Reactor (SFR). The SFR fuel assemblies were designed for burning $^{232}Th$ fuel (fuel pin 1) and $^{233}U$ fuel (fuel pin 2) and include mixed minor actinide compositions. Monte Carlo simulations were performed using Serpent Code1.1.19 to compare with CRAM (Chebyshev Rational Approximation Method) and TTA (Transmutation Trajectory Analysis) method in the burnup calculation mode. The total heating power generated in the system was assumed to be 2000 MWth. During the reactor operation period of 600 days, the effective multiplication factor (keff) was between 0.964 and 0.954 and peaking factor is 1.88867.

• Nuclear Engineering and Technology
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• v.54 no.3
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• pp.803-810
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• 2022

The small modular liquid-metal fast reactor (SMFR) is an important component of advanced nuclear systems. SMFRs exhibit relatively low breeding capability and constraint space for control rod installation. Consequently, control rods are deeply inserted at beginning and are withdrawn gradually to compensate for large burnup reactivity loss in a long lifetime. This paper is committed to investigating the impact of control rod compensation operation on core neutronics characteristics. This paper presents a whole core fine depletion model of long lifetime SMFR using OpenMC and the influence of depletion chains is verified. Three control rod position schemes to simulate the compensation process are compared. The results show that the fine simulation of the control rod compensation process impacts significantly the fuel burnup distribution and absorber consumption. A control rod equivalent position scheme proposed in this work is an optimal option in the trade-off between computation time and accuracy. The control position is crucial for accurate power distribution and void feedback coefficients in SMFRs. The results in this paper also show that the pin level power distribution is important due to the heterogeneous distribution in SMFRs. The fuel burnup distribution at the end of core life impacts the worth of control rods.

• Nuclear Engineering and Technology
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• v.56 no.5
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• pp.1762-1776
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• 2024

A spent nuclear fuel (SNF) analysis module for the Vodo-Vodyanoi Energetichesky Reactor (VVER) was developed and validated in this study. This advancement expands the application area of the existing nodal diffusion code, RAST-V, and reduces the need for additional code during 3D core simulations for SNF analysis, leading to increased efficiency in simulation time. RAST-V uses Lagrange interpolation and a power correction factor derived from the Bateman equation to bypass the re-depletion calculations, which are used to solve the microdepletion chain. This approach improved the efficiency of analysis. To mirror the conditions during the 3D core simulations, the module used history indices related to the moderator temperature, fuel temperature, and boron concentration. The module can predict 1620 isotopes. This paper presents the validation of this isotope inventory prediction and the application of burnup credit. The VVER analysis module was validated using 28 samples discharged from the Novovoronezh-4. Most isotopes were within 10 % of the boundaries of the measurements. This study successfully offers verification results using VVER benchmarks and discusses the application of burnup credit using a VVER-440 cask.

• Nuclear Engineering and Technology
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• v.43 no.5
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• pp.405-412
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• 2011

Used nuclear fuel will likely be stored dry for extended periods of time in the USA. Until a final disposition pathway is chosen, the storage periods will almost definitely be longer than were originally intended. The ability of the important-tosafety structures, systems, and components (SSCs) to continue to meet storage and transport safety functions over extended times must be determined. It must be assured that there is no significant degradation of the fuel or dry cask storage systems. Also, it is projected that the maximum discharge burnups of the used nuclear fuel will increase. Thus, it is necessary to obtain data on high burnup fuel to demonstrate that the used nuclear fuel remains intact after extended storage. An evaluation was performed to determine the conditions that may lead to failure of dry storage SSCs. This paper documents the initial technical gap analysis performed to identify data and modeling needs to develop the desired technical bases to ensure the safety functions of dry stored fuel.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.9 no.3
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• pp.191-198
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• 2011

In general, conventional criticality analyses for spent fuel transport/dry storage systems have been performed based on assumption of fresh fuel concerning the potential uncertainties from number density calculation of Transuranic and Fission Products in spent fuel. However, because of economic loss due to the excessive criticality margin, recently the design of transport/dry storage systems with Burnup Credit(BUC) application has been actively developed. The uncertainties in criticality analyses on transport/storage systems with BUC technique show strong dependance upon initial enrichment and burnup rate, whereas those in the conventional criticality evaluation based on fresh fuel assumption do not show such a dependance. In this study, regulatory-required uncertainties of the criticality analyses for BK 26 Cask, which is conceptually designed spent fuel transport cask with BUC corresponding to the limiting circumstances on nuclear power plants in Korea, are evaluated as a function of initial enrichment and burnup rate. Results of this study will be used as basic data for spent fuel loading curve of BK 26 Cask.

• Nuclear Engineering and Technology
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• v.55 no.6
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• pp.2088-2095
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• 2023

The safety of nuclear installations is largely determined by the tightness of fuel elements cladding. As the Fukushima nuclear accident showed, the main task in case of loss of power supply is to ensure reliable removal of residual heat release from spent fuel pool (SFP) with irradiated fuel assemblies (IFAs). The paper presents the results of calculated-experimental studies and thermal-hydraulic modeling of temperature storage modes of IFAs in SFP. Experimental studies of SFP's temperature regime and calculated evaluation of residual heat removal due to the thermal conductivity of building structures surrounding the SFP were performed. To ensure the safe operation of research reactors, it's necessary to know the IFA's residual heat power (RHP) in the reactor and SFP, which is determined depending on the operating time of fuel assemblies (FAs) and the IFAs calculated holding time. The FAs operating time depends on the reactor energy output. The IFAs calculated holding time is determined by the fuel burnup, U-235 mass in the fuel, and reactor utilization factor. The IFAs fuel burnup was calculated using the MCU-PTR program. Also presented are the RHP's calculation results using some of the empirical dependencies. The concept of a passive heat removal system (PHRS) based on thermosyphon's operating principle was proposed.

• Nuclear Engineering and Technology
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• v.35 no.5
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• pp.471-481
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• 2003

In order to evaluate the microstructural characteristics of irradiated fuel claddings (Zircaloy-4), two irradiated specimens having different burnups (18 GWD/MTU and 42 GWD/MTU) were prepared from the G23-M4 fuel rods, which were loaded in Kori Unit 1 for 4 fuel cycles. The oxide thickness, hydride morphology and hardness change were characterized by an optical microscope and a micro-hardness tester after preparing the irradiated specimens in the PIE (Post Irradiation Examination) facilities. The dislocation loops and the amorphous transformation of precipitates induced by the neutron irradiation in the nuclear plant were also examined by a TEM. As the burnup increased from 18 GWD/MTU to 42 GWD/MTU, the oxide thickness increased from $6.0{\mu}m\;to\;25.3{\mu}m$ and the contents of hydrogen pick-up in the Zr matrix also greatly increased. In the comparison of the hardness of the unirradiated fuel cladding, the amount of hardness increase was nearly $9.3\%\;and\;24.2\%$ for the irradiated samples of 18 and 42 GWD/MTU, respectively. Both -type dislocation loops and -type dislocation components were observed simultaneously in the irradiated specimens and the densities of the dislocation was increased by increasing the burnup. The precipitates in both the irradiated specimens were amorphously transformed by the neutron irradiation and the trend of the amorphous transformation of the precipitates was enhanced at a higher burnup.

• Nuclear Engineering and Technology
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• v.53 no.9
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• pp.2816-2829
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• 2021

Decay heat residuals of spent nuclear fuel (SNF), i.e., the differences between calculations and measurements, were obtained previously for various spent fuel assemblies (SFA) using the Polaris module of the SCALE code system. In this paper, we compare decay heat residuals to their uncertainties, focusing on four PWRs and four BWRs. Uncertainties in nuclear data and model inputs are propagated stochastically through calculations using the SCALE/Sampler super-sequence. Total uncertainties could not explain the residuals of two SFAs measured at GE-Morris. The combined z-scores for all SFAs measured at the Clab facility could explain the resulting deviations. Nuclear-data-related uncertainties contribute more in the high burnup SFAs. Design and operational uncertainties tend to contribute more to the total uncertainties. Assembly burnup is a relevant variable as it correlates significantly with the SNF decay heat. Additionally, burnup uncertainty is a major contributor to decay heat uncertainty, and assumptions relating to these uncertainties are crucial. Propagation of nuclear data and design and operational uncertainties shows that the analyzed assemblies respond similarly with high correlation. The calculated decay heats are highly correlated in the PWRs and BWRs, whereas lower correlations were observed between decay heats of SFAs that differ in their burnups.

• Nuclear Engineering and Technology
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• v.56 no.7
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• pp.2704-2710
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• 2024

Nuclear archaeology research provides scientific methods to reconstruct the operating histories of fissile material production facilities to account for past fissile material production. While it has typically focused on analyzing material in permanent reactor structures, spent fuel or high-level waste also hold information about the reactor operation. In this computational study, we explore a Bayesian inference framework for reconstructing the operational history from measurements of isotope ratios from a sample of nuclear waste. We investigate two different inference models. The first model discriminates between three potential reactors of origin (Magnox, PWR, and PHWR) while simultaneously reconstructing the fuel burnup, time since irradiation, initial enrichment, and average power density. The second model reconstructs the fuel burnup and time since irradiation of two batches of waste in a mixed sample. Each of the models is applied to a set of simulated test data, and the performance is evaluated by comparing the highest posterior density regions to the corresponding parameter values of the test dataset. Both models perform well on the simulated test cases, which highlights the potential of the Bayesian inference framework and opens up avenues for further investigation.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.6 no.3
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• pp.225-232
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• 2008

An estimation has been made for inventories and characteristics of spent nuclear fuel(SNF) to be generated from existing and planned nuclear power plants based on the 3rd Basic Plan for Electric Power Demand and Supply. The characteristics under consideration in this study are dimensions, a fuel rod array, a weight, $^{235}U$ enrichment, and the discharge burnup in terms of fuel assembly. These are essentially needed for designing a pyroprocessing facility. It is appeared that the anticipated quantity by the end of 2077 is about 23,000 tU for PWR spent nuclear fuel. It is revealed that the proportion of SNF with the initial $^{235}U$ enrichment below 4.5 weight percent(wt.%) is approximately 95 % in total. For SNF with 16$\times$16 fuel rod array the proportion is expected approximately 74% in total. It appears that the average burnup of SNF will be 55 GWd/tU after the medium and/or latter part of 2010s while the average burnup is 45 GWd/tU at present. Finally, a requirement in terms of reference SNF for designing the pyroprocessing facility has been derived from the above-mentioned results. The anticipated SNF seems to be 16$\times$16 Korean Standard Fuel Assembly with a cross section of 21.4 cm$\times$21.4 cm, a length of 453 cm, a mass of 672 kg, the initial $^{235}U$ enrichment of 4.5 wt.%, and the discharge burnup of 55 GWd/tU.