• Nuclear Engineering and Technology
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• 제56권9호
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• pp.3730-3739
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• 2024

The high-temperature molten salt pump is the core equipment in the small modular molten salt reactor with media temperatures up to 700 ℃. The cantilever ratio of the molten salt pump is usually large. Excessively large cantilever ratios cause increased deformations and rotational amplitudes at the impeller, thus affecting the operational stability of the main pump; small cantilever ratios cause heavy temperature gradients, thus affecting the structural integrity evaluation. This paper used numerical simulation methods to calculate and analyze the temperature field, stress, and structural integrity, optimized the pump shaft cantilever length of the original scheme based on structural integrity using the dichotomy method, and analyzed the rotor dynamics of the optimization results. The results of this study show that the thermal expansion load caused by the temperature difference has a significant mechanical effect on the structure; the first-order critical speed of the rotor system of the optimized schemes has been improved, and the amplitude of the unbalanced response has been significantly reduced, which not only improves the operational stability of the rotor, also contributes to the compact design of the main pump of a small modular molten salt reactor.

• Nuclear Engineering and Technology
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• 제48권3호
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• pp.711-718
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• 2016

Steady-state operation of a fusion power plant requires external current drive to minimize the power requirements, and a high fraction of bootstrap current is required. One of the external sources for current drive is lower hybrid current drive, which has been widely applied in many tokamaks. Here, using lower hybrid simulation code, we calculate electron distribution function, electron currents and phase velocity changes for two options of demonstration reactor at the launched lower hybrid wave frequency 5 GHz. Two plasma scenarios pertaining to two different demonstration reactor options, known as pulsed (Option 1) and steady-state (Option 2) models, have been analyzed. We perceive that electron currents have major peaks near the edge of plasma for both options but with higher efficiency for Option 1, although we have access to wider, more peripheral regions for Option 2. Regarding the electron distribution function, major perturbations are at positive velocities for both options for flux surface 16 and at negative velocities for both options for flux surface 64.

• Nuclear Engineering and Technology
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• 제56권9호
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• pp.3626-3643
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• 2024

One of the issues in the development of the sodium-cooled fast reactor (SFR) using transuranic (TRU) metallic fuel is the absence of criticality benchmark experiment that faithfully mocks up the nuclear characteristics of the target design for validation of the reactor core design code and its uncertainty quantification (UQ). This study aims to quantify the criticality uncertainty of a typical TRU burner with metallic fuel by using the standard upper safety limit (USL) estimation framework based on the similarity analysis of existing benchmark experiments but elaborated in two aspects:1) application of two-sided rather than one-sided tolerance interval and 2) inclusion of additional uncertainty to account for fission products and minor actinides not included in the benchmark experiments. To conduct the similarity analysis and evaluate the nuclear-data induced uncertainty, existing, well-verified computing codes were integrated, including the nuclear data sampling code SANDY, the nuclear data processing code NJOY, and the continuous-energy Monte Carlo code McCARD. Finally, using the SFR benchmark database comprising both publicly available and proprietary benchmark experiments, the criticality uncertainty of the TRU core model with metallic fuel was evaluated.

• Nuclear Engineering and Technology
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• 제54권8호
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• pp.2792-2800
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• 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
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• 제55권4호
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• pp.1439-1447
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• 2023

The fuel cycle characteristics of the IVG.1M reactor were studied within the framework of the research reactor conversion program to modernize the IVG.1M reactor. Optimum use of the nuclear fuel and reactor was achieved through routine methods which included partial fuel reloading combined with scheduled maintenance operations. Since, the additional problem in planning the fuel cycle of the IVG.1M reactor was the poisoning of the beryllium parts of the core, reflector, and control system. An assessment of the residual power and composition of spent fuel is necessary for the selection and justification of the technology for its subsequent management. Computational studies were performed using the MCNP6.1 program and the neutronics model of the IVG.1M reactor. The proposed scheme of annual partial fuel reloading allows for maintaining a high reactor reactivity margin, stabilizing it within 2-4 β_eff for 20 years, and achieving a burnup of 9.9-10.8 MW × day/kg U in the steady state mode of fuel reloading. Spent fuel immediately after unloading from the reactor can be placed in a transport packaging cask for shipping or safely stored in dry storage at the research reactor site.

• 한국의학물리학회:학술대회논문집
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• 한국의학물리학회 1999년도 Japanese Journal of Medical Physics
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• pp.54-57
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• 1999

• Nuclear Engineering and Technology
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• 제48권3호
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• pp.635-641
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• 2016

Because of prohibitive data storage requirements in large-scale simulations, the memory problem is an obstacle for Monte Carlo (MC) codes in accomplishing pin-wise three-dimensional (3D) full-core calculations, particularly for whole-core depletion analyses. Various kinds of data are evaluated and quantificational total memory requirements are analyzed based on the Reactor Monte Carlo (RMC) code, showing that tally data, material data, and isotope densities in depletion are three major parts of memory storage. The domain decomposition method is investigated as a means of saving memory, by dividing spatial geometry into domains that are simulated separately by parallel processors. For the validity of particle tracking during transport simulations, particles need to be communicated between domains. In consideration of efficiency, an asynchronous particle communication algorithm is designed and implemented. Furthermore, we couple the domain decomposition method with MC burnup process, under a strategy of utilizing consistent domain partition in both transport and depletion modules. A numerical test of 3D full-core burnup calculations is carried out, indicating that the RMC code, with the domain decomposition method, is capable of pin-wise full-core burnup calculations with millions of depletion regions.

• Nuclear Engineering and Technology
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• 제39권2호
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• pp.95-102
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• 2007

The conceptual design of the MIT modular pebble bed reactor is described. This reactor plant is a 250 Mwth, 120 Mwe indirect cycle plant that is designed to be deployed in the near term using demonstrated helium system components. The primary system is a conventional pebble bed reactor with a dynamic central column with an outlet temperature of 900 C providing helium to an intermediate helium to helium heat exchanger (IHX). The outlet of the IHX is input to a three shaft horizontal Brayton Cycle power conversion system. The design constraint used in sizing the plant is based on a factory modularity principle which allows the plant to be assembled 'Lego' style instead of constructed piece by piece. This principle employs space frames which contain the power conversion system that permits the Lego-like modules to be shipped by truck or train to sites. This paper also describes the research that has been conducted at MIT since 1998 on fuel modeling, silver leakage from coated fuel particles, dynamic simulation, MCNP reactor physics modeling and air ingress analysis.

• Nuclear Engineering and Technology
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• 제48권2호
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• pp.304-317
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• 2016

An accelerator-driven system consists of a subcritical reactor and a controllable external neutron source. The reactor in an accelerator-driven system can sustain fission reactions in a subcritical state using an external neutron source, which is an intrinsic safety feature of the system. The system can provide efficient transmutations of nuclear wastes such as minor actinides and long-lived fission products and generate electricity. Recently at Kyoto University Research Reactor Institute (KURRI; Kyoto, Japan), a series of reactor physics experiments was conducted with the Kyoto University Critical Assembly and a Cockcrofte-Walton type accelerator, which generates the external neutron source by deuteriu-metritium reactions. In this paper, neutronic analyses of a series of experiments have been re-estimated by using the latest Monte Carlo code and nuclear data libraries. This feasibility study is presented through the comparison of Monte Carlo simulation results with measurements.

• Nuclear Engineering and Technology
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• 제51권4호
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• pp.949-953
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• 2019

The burnup equation of nuclides is one of the most equations in nuclear reactor physics, which is generally coupled with transport calculations. The burnup equation describes the variation of the nuclides with time. Because of its very stiffness and the need for large time step, this equation is solved by special methods, for example transmutation trajectory analysis (TTA) or the matrix exponential methods where the matrix exponential is approximated by CRAM. However, TTA or CRAM functions well when the flux is constant. In this work, a new method is proposed when the flux changes. It's an improved method compared to TTA or CRAM. Furtherly, this new method is based on TTA or CRAM, and it is more accurate than them. The accuracy and efficiency of this method are investigated. Several cases are used and the results show the accuracy and efficiency of this method are great.