Nuclear Technology

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ISSN / EISSN : 0029-5450 / 1943-7471
Published by: Informa UK Limited (10.1080)
Total articles ≅ 8,461
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Chun-Yen Li, , Marco Pellegrini, Nejdet Erkan, Koji Okamoto
Published: 22 November 2021
Nuclear Technology pp 1-17; https://doi.org/10.1080/00295450.2021.1973181

Abstract:
For the Japan Sodium-cooled Fast Reactor (JSFR), should the hypothesized core disruptive accident (CDA) happened, the in-vessel retention (IVR) will be the main target to achieve. In the heat-removal phase of the CDA, the debris bed will be piled up on the debris catcher. The capability of stable cooling and avoiding recriticality on the debris bed will be the main issues for achieving IVR. Previous studies have shown that the homogeneous debris bed can attain stable cooling and eliminate the probability of recriticality. Besides, self-leveling, which is a mechanism redistributing and flattening the debris bed by the natural circulation or vaporization from surrounding coolant, can further suppress the debris bed’s thickness to below the coolable thickness. However, in the real situation, the debris bed is composed of mixed-density debris particles. Hence, when these mixed-density debris particles start to redistribute due to self-leveling, the debris bed will form a heterogeneous density distribution. Under this scenario, the capability of coolability and the probability of recriticality could deviate from the previous study. Therefore, it is necessary to obtain a verified coupled model between the computational fluid dynamics (CFD) and the discrete element method (DEM) to track the mixed-density debris particles’ movement under the phenomenon of self-leveling. In this paper, first, the experiments simulating self-leveling on the mixed-density particle bed are performed. Afterward, the random heavy particle movement’s experimental data are extracted and transformed into the statistics form as the benchmark materials. Finally, the CFD-DEM model is validated via a series of sensitivity studies. The verified CFD-DEM can be expected to simulate the self-leveling behavior on the mixed-density debris bed and the real reactor case.
Published: 15 November 2021
Nuclear Technology pp 1-15; https://doi.org/10.1080/00295450.2021.1985912

Abstract:
A previous study concluded that the robust, multimodule design of the NuScale small modular reactor plant can provide power at an unprecedented level of availability to mission critical facilities. This study extends the analysis to include a microgrid power distribution and delivery system to demonstrate the increased availability of power delivered to a customer. A hypothetical 12-module NuScale plant located on the Clinch River site in Tennessee is assumed to supply power from three modules to Oak Ridge National Laboratory (ORNL) through the Tennessee Valley Authority (TVA) transmission system. Combinations of transmission and power generation equipment failures that might interrupt power, and the associated frequency and duration of these failures, are identified and the potential for power interruption to ORNL is evaluated. The analysis first evaluates the existing transmission infrastructure and availability of power to ORNL to establish a baseline availability. Then, a connection from the NuScale plant through the local TVA transmission system (option 1) and a direct connection from the NuScale plant to the ORNL distribution system (option 2) are evaluated, as well as three sensitivity cases. The existing power distribution and delivery system at ORNL is already highly reliable resulting from multiple diverse power generators feeding a robust power delivery system. The primary driver of macrogrid power unavailability is the existing power generation sources, which includes two coal plants and two hydroelectric generators, rather than transmission equipment. Adding a 12-module NuScale plant to the system further reduces the unavailability of power to ORNL by over two orders of magnitude in both cases of considering only local power sources and the macrogrid as a whole. When considering only local generators, the inclusion of a NuScale plant improves the average availability of power to ORNL from three-nines to over five-nines. If the large-scale macrogrid is also included, average availability is increased to nine-nines.
, Joseph Nielsen, Joshua Cogliati, Charles Wemple
Published: 15 November 2021
Nuclear Technology pp 1-11; https://doi.org/10.1080/00295450.2021.1980320

Abstract:
The neutronics software, HELIOS, was validated in 2015 for performing core reload design and safety analysis of the Advanced Test Reactor. However, when HELIOS was benchmarked against historic fission-wire measurements (i.e., zero-power full-core measurements), a statistically resolved calculation-to-measurement bias was discovered. The azimuthal power along each fuel plate computed by HELIOS has consistently shown to underpredict measurements made by fission wires in historic zero-power tests near the fuel element side plates. It was hypothesized during the HELIOS software validation work that this bias is attributable to local moderation in coolant vents in the side plates axially just above and below the fission wires on the fuel plate edges. This work used detailed MCNP and MC21 models of the side plate vents to test this hypothesis. By comparing the average azimuthal biases between HELIOS and two-dimensional and three-dimensional (3-D) MCNP models and a 3-D MC21 model, it was found that the HELIOS azimuthal bias is not due to the measurement.
, L. Bures, K. Mikityuk
Published: 13 November 2021
Nuclear Technology pp 1-13; https://doi.org/10.1080/00295450.2021.1971025

Abstract:
Helium gases are utilized to remove fission products from the molten salt fast reactor (MSFR) core during operation. Helium gases and other volatile fission products may be introduced into the intermediate heat exchanger channels. The effect of these gases on heat transfer is essential for the MSFR to operate properly, especially in laminar flow regimes. The computational fluid dynamics code PSI-BOIL was selected to examine this problem because of its interface tracking capability. A periodic square duct simulation created the flow regime, resulting in a sliding bubble regime. Following that, we examined the impact of heat transfer using an extended nonperiodic channel simulation with a succession of corner bubble arrays. Due to the combined effects of low thermal diffusivity and laminar flow characteristics, it is shown that the length of heat transfer augmentation may extend to at least five bubble diameters downstream of the bubble placement. Finally, we examined the impact of interphasic heat transfer between an inert gas and a liquid. The bulk of the heat transfer amplification effect was due to the motion of the bubbles rather than interphasic heat transfer.
, Todd S. Palmer, Samuel Bays
Published: 13 November 2021
Nuclear Technology pp 1-21; https://doi.org/10.1080/00295450.2021.1960783

Abstract:
The field of reactor design is rich with opportunities for applications of computational optimization algorithms; these applications can range from preliminary core design to reactor shuffling patterns. Many of these schemes rely on sets of previously generated solutions (sometimes referred to as “generations”) to inform future decisions. While it is important to build upon prior knowledge, this process requires a full generation of solutions to be formed before future solutions can be examined. Rather than relying on a generational scheme to perform an optimization, we propose using an agent-based approach in conjunction with a blackboard framework for performing reactor design optimizations. Utilizing an agent-based approach allows agents to perform tasks independently, while retaining the ability to build off of previous solutions. We develop an agent-based blackboard system (ABBS) for determining the Pareto front (PF) in sodium fast reactor design optimization problems and compared this with the Non-Dominated Sorting Genetic Algorithm II (NSGA-II). Our goal is to evaluate the viability of the ABBS in producing a PF that is comparable with the NSGA-II algorithm. The design space consists of the fuel height, fuel smear, and plutonium fraction in the core, and we seek to minimize the reactivity swing and plutonium mass, while maximizing the burnup. The diversity, coverage, and spread of the PFs generated by the two methods are examined, and the ABBS is able to converge to the same PF as the NSGA-II algorithm. These results show that the ABBS is able to find optimal designs that are similar to those found by the NSGA-II algorithm. We conclude our study by applying the ABBS to the design of a sodium-cooled fast reactor to dispose of weapons-grade plutonium. The ABBS finds a core design that can burn upwards of 17.5 kg of weapons-grade plutonium per year and degrade an additional 195 kg of weapons-grade plutonium per year into non-weapons-grade material.
, Clayton G. Turner
Published: 13 November 2021
Nuclear Technology pp 1-10; https://doi.org/10.1080/00295450.2021.1977085

Abstract:
New nuclear reactor designs that incorporate heat pipes are being investigated for possible near-term deployment in terrestrial applications. This study explores the use of screen-covered axially grooved sodium heat pipes and their applicability for providing heat removal for microreactors. A sodium working fluid is appropriate for microreactors operating in the 5 to 20 MW(thermal) range at approximately 650°C. HTPIPE, a legacy software code, was validated for the case of screen-covered grooves and used to perform steady-state analyses to determine the performance limits of a proposed heat pipe design. The performance limits of a sodium heat pipe with a screen-covered square grooved wick structure is compared to that of an equivalent heat pipe with an annular wick. In a horizontal orientation at an operating temperature of 650°C,the performance limits for the heat pipe with an annular wick configuration are 15% higher than for the screen-covered grooved wick. At operating temperatures below 777°C, the annular wick outperforms the screen-covered grooved wick, and at temperatures above 777°C, the screen-covered grooved wick outperforms the annular wick. However, the marginal performance gain at higher temperatures may not justify the use of heat pipes with a screen-covered grooved wick structure due to increased manufacturing costs.
L. C. Olson, , H. M Ajo
Published: 11 November 2021
Nuclear Technology pp 1-10; https://doi.org/10.1080/00295450.2021.1988821

Abstract:
The Savannah River National Laboratory evaluated several options for disposition of stainless steel (SS)–clad plutonium metal, particularly Pu-10.6 at. % Al (Pu- 1.3 wt% Al) alloy fuel. One technology considered was alloying fuel with SS. The goal of the alloying would be to make a SS-Pu alloy that was a nonproliferable waste form with secondary Pu-rich microencapsulated regions distributed throughout the refractory SS. The microencapsulation of the Pu regions should therefore allow the waste form to meet the requirements for a low attractiveness waste as defined by the U.S. Department of Energy. Plutonium-bearing alloys at these levels could potentially be suitable for disposal at a waste isolation pilot plant. Four metal ingots were successfully fabricated using U and Al as a surrogate for Pu-Al. The U was distributed and microencapsulated by the alloy matrix, thereby setting the stage for subsequent tests using SS-clad fuel elements containing Pu-10.6Al.
Tate Shorthill, , , Heng Ban
Published: 5 November 2021
Nuclear Technology pp 1-20; https://doi.org/10.1080/00295450.2021.1957659

Abstract:
Digital instrumentation and control (I&C) upgrades are a vital research area for the nuclear industry. Despite their performance benefits, deployment of digital I&C in nuclear power plants (NPPs) has been limited. Digital I&C systems exhibit complex failure modes including common cause failures (CCFs), which can be difficult to identify. This paper describes the development of a redundancy-guided application of the Systems-Theoretic Process Analysis and fault tree analysis for the hazard analysis of digital I&C in advanced NPPs. The resulting Redundancy-Guided Systems-Theoretic Hazard Analysis (RESHA) is applied for the case study of a representative state-of-the-art digital reactor trip system. The analysis qualitatively and systematically identifies the most critical CCFs and other hazards of digital I&C systems. Ultimately, the RESHA can help researchers make informed decisions for how, and to what degree, defensive measures such as redundancy, diversity, and defense in depth can be used to mitigate or eliminate the potential hazards of digital I&C systems.
, Andrew T. Godfrey, , , Shane C. Henderson,
Published: 1 November 2021
Nuclear Technology pp 1-17; https://doi.org/10.1080/00295450.2021.1957660

Abstract:
The Consortium for Advanced Simulation of Light Water Reactors (CASL) Virtual Environment for Reactor Applications (VERA) is a reactor simulation software. It offers unique capabilities by combining high-fidelity in-core radiation transport with temperature feedback by using MPACT (a deterministic neutron transport code) and COBRA-TF (a thermal-hydraulic code) with follow-on, fixed-source transport calculations using the Shift Monte Carlo code to calculate ex-core quantities of interest. In these coupled calculations, MPACT provides Shift with the fission source for follow-on ex-core calculations. These ex-core simulations can be set up to calculate detector responses, as well as the flux and fluence in ex-core regions of interest, such as the reactor pressure vessel, nozzle, and irradiated capsules. A Watts Bar Nuclear Plant Unit 1 (WBN1) ex-core model was developed, as described in this paper, and this model was used to perform coupon calculations. The results for the coupon flux calculations show close agreement with the reference values for cycle 1 produced by the two-dimensional Discrete Ordinates Transport (DORT) code and presented in a BWXT Services Inc. report. However, differences in the results (10%) seen in cycles 2 and 3 and the reasons for these differences are discussed in this paper. The VERA WBN1 model was also used to perform a vessel fluence calculation for cycle 1. Additionally, a collaboration between CASL and Duke Energy led to the first code-to-code validation of VERA for reactor ex-core applications that used a model for the Shearon Harris reactor. Results from this collaboration show excellent agreement between VERA and the Monte Carlo N-Particle Transport Code for the detector response calculations. The work performed under this collaboration is also detailed in this paper.
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