• Nuclear Engineering and Technology
• /
• v.40 no.1
• /
• pp.87-92
• /
• 2008

Tokamak reactor system analysis code was developed at KAERI (Korea Atomic Energy Research Institute) and is used here for the conceptual development of a DEMO reactor. In the system analysis code, prospects of the development of plasma physics and the relevant technology are included in a simple mathematical model, i.e., the overall plant power balance equation and the plasma power balance equation. This system analysis code provides satisfactory results for developing the concept of a DEMO reactor and for identifying the necessary R&D areas, both in the physics and technology areas for the realization of the concept. With this system analysis code, the performance of a DEMO reactor with a limited extension of the plasma physics and technology adopted in the ITER design. The main requirements for the DEMO reactor were selected as: 1) demonstrate tritium self-sufficiency, 2) generate net electricity, and 3) achieve a steady-state operation. It was shown that to access an operational region for higher performance, the main restrictions are presented by the divertor heat load and the steady-state operation requirements.

• Nuclear Engineering and Technology
• /
• v.53 no.9
• /
• pp.2816-2829
• /
• 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
• /
• v.54 no.1
• /
• pp.226-233
• /
• 2022

In the present work, a phase-field model was developed to investigate the influence of recrystallization on bubble evolution during irradiation. Considering the interaction between bubbles and grain boundary (GB), a set of modified Cahn-Hilliard and Allen-Cahn equations, with field variables and order parameters evolving in space and time, was used in this model. Both the kinetics of recrystallization characterized in experiments and point defects generated during cascade were incorporated in the model. The bubble evolution in recrystallized polycrystalline of U-Mo alloy was also investigated. The simulation results showed that GB with a large area fraction generated by recrystallization accelerates the formation and growth of bubbles. With the formation of new grains, gas atoms are swept and collected by GBs. The simulation results of bubble size and distribution are consistent with the experimental results.

• Nuclear Engineering and Technology
• /
• v.56 no.8
• /
• pp.3255-3267
• /
• 2024

This paper presents a high-fidelity simulation of the Organization for Economic Co-operation and Development (OECD) Nuclear Energy Agency (NEA) 3D whole-core Watts Bar benchmark using the UNIST in-house STREAM3D (Steady State and Transient Reactor Analysis code with Method of Characteristics) neutronic code. The benchmark encompasses various whole-core exercises, including single physics problems, multiphysics simulations, and depletion problems. When comparing parameters during the zero-power physics tests, including ITC, DBW, CRW, and criticality tests, STREAM3D results indicate a strong agreement with the measured data and KENO-VI. The comparison with the MC21/CTF code in 3D HFP BOC condition demonstrated strong agreement, with only a 0.42% difference in the normalized radial power distribution, a 0.38 K difference in the RMS of the assembly coolant exit temperature, and a mere 4 ppm difference in CBC.

• Nuclear Engineering and Technology
• /
• v.53 no.10
• /
• pp.3398-3405
• /
• 2021

This study observed the effect of neutron irradiation and ageing on the microstructure, hardness, and corrosion resistance of SAV-1 (Al-Mg-Si) alloy. The investigated material was irradiated with neutrons to fluences of 10²¹-10²⁶ n/m² in the WWR-K research reactor and kept in dry storage. Long-term irradiation led to an increase in hardness of the alloy and a deterioration of pitting corrosion resistance. Post-irradiation ageing for 1 h at 100-300 ℃ resulted in a decrease in microhardness of the irradiated SAV-1. The effect of post-irradiation ageing on pitting corrosion was made clear through the formation of Guinier-Preston zones and secondary precipitates in the Al matrix. Ageing at 250 ℃ corresponded to the development of stable microstructure and the highest corrosion resistance for the irradiated samples. Mg₂Si, Si, and needle-shaped β" precipitates were formed in SAV-1 alloy that was irradiated with low fluences. β" and clusters of rod-shaped B-type precipitates were observed in highly irradiated samples. The precipitates were similar to those seen in non-irradiated pseudo-binary Al-Mg₂Si alloys with Si excess.

• Nuclear Engineering and Technology
• /
• v.55 no.5
• /
• pp.1791-1801
• /
• 2023

The multi-physics coupled methodologies that have been widely used to analyze the complex process occurring in nuclear reactors have also been used to the R&D of numerical reactors. The advancement in the field of computer technology has helped in the development of these methodologies. Herein, we report the integration of ADPRES code and RELAP5 code into the SALOME-ICoCo framework to form a multi-physics coupling platform. The platform exploits the supervisor architecture, serial mode, mesh one-to-one correspondence and explicit coupling methods during analysis, and the uncertainty analysis tool URANIE was used. The correctness of the platform was verified through the NEACRP-L-335 benchmark. The results obtained were in accordance with the reference values. The platform could be used to accurately determine the power peak. In addition, design margins could be gained post uncertainty analysis. The initial power, inlet coolant temperature and the mass flow of assembly property significantly influence reactor safety during the rod ejections accident (REA).

• Nuclear Engineering and Technology
• /
• v.52 no.6
• /
• pp.1099-1109
• /
• 2020

The performance of gadolinium burnable absorber (GdBA) for reactivity control in UO₂ and (U, Th)O₂ fuels and its impact on spent fuel characteristics was performed. Five fuel assemblies: one without GdBA fuel rod and four each containing 16, 24, 34 and 44 GdBA fuel rods in both fuels were investigated. Reactivity swing in all the FAs with GdBA rods in UO₂ fuel was higher than their counterparts with similar GdBA fuel rods in (U, Th)O₂ fuel. The excess reactivity in all FAs with (U, Th)O₂ fuel was higher than UO₂ fuel. At the end of single discharge burn-up (~ 49.64 GWd/tHM), the excess reactivity of (U, Th) O₂ fuel remained positive (16,000 pcm) while UO₂ fuel shows a negative value (-6,000 pcm), which suggest a longer discharge burn-up in (U, Th)O₂ fuel. The concentration of plutonium isotopes and minor actinides were significantly higher in UO₂ fuel than in (U, Th)O₂ fuel except for ²³⁶Np. However, the concentration of non-actinides (gadolinium and iodine isotopes) except for ¹³⁵Xe were respectively smaller in (U, Th)O₂ fuel than in UO₂ fuel but may be two times higher in (U, Th)O₂ fuel due to its potential longer discharge burn-up.

• Nuclear Engineering and Technology
• /
• v.56 no.5
• /
• pp.1574-1583
• /
• 2024

Small Modular Reactors (SMRs) are compact nuclear reactors designed to generate electric power up to 300 MWe. They could be assembled in factory, and then transported to be directly installed on-stie. CAREM (Central Argentina de Elementos Modulares) is a national SMR development project, based on light water reactor technology supervised by Argentina's National Atomic Energy Commission (CNEA). It is a natural circulation-based SMR with an indirect-cycle, including specific items and parts that simplify the design and improve safety performance. In this paper, the thermal-hydraulic study of CAREM-25 advanced small modular reactor is conducted by using COBRA-EN modified code and the Porous Media Approach (PMA) for the first time. According to PMA approach, each fuel assembly is modeled and divided into a network of lumped regions. While complex geometries are defined, the thermal-hydraulic parameters such as temperature and density are calculated for coolant and fuel rods. The obtained results show that the temperature in the fuel center may reach a peak around 1280 K in the hottest fuel assembly. Finally, the comparison of results from both methods (modified COBRA-EN and PMA) presented an appropriate consistency.

• Nuclear Engineering and Technology
• /
• v.53 no.10
• /
• pp.3133-3150
• /
• 2021

This work deals with the validation of a high-fidelity multiphysics system coupling the Serpent 2 Monte Carlo neutron transport code with SUBCHANFLOW, a subchannel thermalhydraulics code, and TRANSURANUS, a fuel-performance analysis code. The results for a full-core pin-by-pin burnup calculation for the ninth operating cycle of the Temelín II VVER-1000 plant, which starts from a fresh core, are presented and assessed using experimental data. A good agreement is found comparing the critical boron concentration and a set of pin-level neutron flux profiles against measurements. In addition, the calculated axial and radial power distributions match closely the values reported by the core monitoring system. To demonstrate the modeling capabilities of the three-code coupling, pin-level neutronic, thermalhydraulic and thermomechanic results are shown as well. These studies are encompassed in the final phase of the EU Horizon 2020 McSAFE project, during which the Serpent-SUBCHANFLOW-TRANSURANUS system was developed.

• Nuclear Engineering and Technology
• /
• v.56 no.9
• /
• pp.3969-3980
• /
• 2024

Over the years, numerous evaluations of material attractiveness have been performed for conventional light water reactors to better understand the nature of the spent fuel material and its desirability for misuse at different points in the nuclear fuel cycle. However, availability of such assessments for newer, Generation IV reactors such as Molten Salt Reactors is rather limited. In the present study we address the gap in knowledge of material attractiveness for molten salt reactor systems and describe the nature of irradiated fuel salts which the nuclear safeguards community might be faced with in the near future as more and more such reactors enter commission and operation. Within the scope of the paper, we use a large database of simulated irradiated fuel salt isotopics (and other derived quantities such as gamma activity, decay heat, and neutron emission rates) developed specifically for a molten salt reactor concept in order to shed some light on possible weapons usability of uranium and plutonium present in the irradiated fuel salts. This has been achieved by proposing a new attractiveness metric that is better suited for quantifying attractiveness of irradiated salts from a model molten salt concept. The said metric has been computed using a database that has been created by simulating the irradiation of molten fuel salt in a concept core over a wide range of operational parameters (burnup, initial enrichment, and cooling time) using the Monte-Carlo particle transport code, Serpent. With the help of this attractiveness metric, the findings from this study have shown that in relative terms, molten salt spent fuel is more attractive than spent fuel produced by a conventional light water reactor. The findings also underscore the need for strengthened safeguards measures for such spent fuel. These results are expected to be useful in the future for regulatory authorities as well as for nuclear safeguards inspectors for designing a functional safeguards verification routine for irradiated fuel of such unique nature.