• 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.

• 설비공학논문집
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• 제8권2호
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• pp.279-287
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• 1996

A cycle analysis is performed to investigate how the non-ideal gas behavior of helium reduces the heating capacity of VM heat pumps. Since the operating pressures of VM heat pumps are as high as 1 to 20 MPa, the compressibility factor of helium becomes clearly greater than 1 and the non-ideal behavior always represents a thermal loss in heating. To calculate the amount of the losses, an adiabatic cycle analysis is performed with the real properties of helium and the net enthaply flows through the two regenerators are numerically obtained. It is shown that the non-ideal gas losses could be as much as 8% in the heating capacity when the operating pressures are greater than 10MPa. The effects of the operating temperatures and the dead volumes on the loss are presented.

• Nuclear Engineering and Technology
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• 제52권4호
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• pp.689-699
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• 2020

In this work, new multi-recycling options of TRU nuclides using PWR fuel assemblies comprised of MOX and FCM (Fully Ceramic Micro Encapsulated) fuels are suggested and neutronically analyzed. These options do not use a fully recycling of TRU but a partial recycling where TRUs from MOX fuels are recycled while the ones from FCM fuels are not recycled due to their high consumption rate resulted from high burnup. In particular, additional external TRU feed in MOX fuels for each cycle was considered to significantly increase the TRU consumption rate and the finally selected option is to use external TRU and enriched uranium feed as a makeup for the heavy metal consumption in MOX fuels. This hybrid external feeding of TRU and enriched uranium in MOX fuel was shown to be very effective in significantly increasing TRU consumption rate, maintaining long cycle length, and achieving negative void reactivity worth during recycling.

• Nuclear Engineering and Technology
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• 제45권6호
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• pp.803-810
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• 2013

The U.S. Department of Energy's Fuel Cycle Technologies R&D program under the Office of Nuclear Energy is working to advance technologies to enhance both the existing and future fuel cycles. One thrust area is in developing enabling technologies for next generation nuclear materials management under the Materials Protection, Accounting and Control Technologies (MPACT) Campaign where advanced instrumentation, analysis and assessment methods, and security approaches are being developed under a framework of Safeguards and Security by Design. An overview of the MPACT campaign's activities and recent accomplishments is presented along with future plans.

• 한국경영과학회지
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• 제31권1호
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• pp.25-36
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• 2006

Consider a supply chain in which a vendor supplies a product to a buyer. We assume that the buyer's and vendor's inventory cost structures are different, resulting in differences in inventory order/delivery cycle times. Here, if one party insists on its individually optimal order/delivery quantity, the other party will suffer from mismatches in cycle times. Under this scenario, coordination contracts that make use of either a Net Term/Two parts Term Trade Credit or a Quantity Discount are designed to align individually optimal order Quantities. We compare and analyze the perform ances of these contracts. The focus of the comparison is the ability of contracts to generate a lower cost for the supply chain. We show that a Trade Credit policy can be effectively used to coordinate a supply chain. In many cases it will result in a lower supply chain cost compared to that achieved by using a Quantitative Discount policy.

• Nuclear Engineering and Technology
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• 제49권1호
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• pp.216-223
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• 2017

In this work, linear and mass attenuation coefficients, effective atomic number and electron density, mean free paths, and half value layer and $10^{th}$ value layer values of barium-bismuth-borosilicate glasses were obtained for 662 keV, 1,173 keV, and 1,332 keV gamma ray energies using MCNP-4C code and XCOM program. Then obtained data were compared with available experimental data. The MCNP-4C code and XCOM program results were in good agreement with the experimental data. Barium-bismuth-borosilicate glasses have good gamma ray shielding properties from the shielding point of view.

• Nuclear Engineering and Technology
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• 제48권2호
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• pp.434-447
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• 2016

The aim of this paper is the study of instability state of boiling water reactors with a method based in largest Lyapunov exponents (LLEs). Detecting the presence of chaos in a dynamical system is an important problem that is solved by measuring the LLE. Lyapunov exponents quantify the exponential divergence of initially close state-space trajectories and estimate the amount of chaos in a system. This method was applied to a set of signals from several nuclear power plant (NPP) reactors under commercial operating conditions that experienced instabilities events, apparently each of a different nature. Laguna Verde and Forsmark NPPs with in-phase instabilities, and Cofrentes NPP with out-of-phases instability. This study presents the results of intrinsic instability in the boiling water reactors of three NPPs. In the analyzed cases the limit cycle was not reached, which implies that the point of equilibrium exerts influence and attraction on system evolution.

• Nuclear Engineering and Technology
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• 제46권1호
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• pp.81-92
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• 2014

This study suggests 5 evaluation criteria (safety and technology, environmental impact, economic feasibility, social factors, and institutional factors) and 24 evaluation indicators for a NFC (nuclear fuel cycle) derived using factor analysis. To do so, a survey using 1 on 1 interview was given to nuclear energy experts and local residents who live near nuclear power plants. In addition, by conducting a factor analysis, homogeneous evaluation indicators were grouped with the same evaluation criteria, and unnecessary evaluation criteria and evaluation indicators were dropped out. As a result of analyzing the weight of evaluation criteria with the sample of nuclear power experts and the general public, both sides recognized safety as the most important evaluation criterion, and the social factors such as public acceptance appeared to be ranked as more important evaluation criteria by the nuclear energy experts than the general public.

• Nuclear Engineering and Technology
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• 제51권2호
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• pp.479-489
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• 2019

Low-cycle fatigue (LCF) tests were performed for Alloy 690 and 316 SS in a simulated pressurized water reactor (PWR) environment. Alloy 690 showed about twice longer LCF life than 316 SS at the test condition of 0.4% amplitude at strain rate of 0.004%/s. Observation of the oxide layers formed on the fatigue crack surface showed that Cr and Ni rich oxide was formed for Alloy 690, while Fe and Cr rich oxide for 316 SS as an inner layer. Electrochemical analysis revealed that the oxide layers formed on the LCF crack surface of Alloy 690 had higher impedance and less defect density than those of 316 SS, which resulted in longer LCF life of Alloy 690 than 316 SS in a simulated PWR environment.

• Nuclear Engineering and Technology
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• 제53권11호
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• pp.3517-3527
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• 2021

New 80-MW (electric) ultra-long-life sodium cooled fast reactor core having inherent safety characteristics is designed with heterogeneous fuel assemblies comprised of driver and blanket fuel rods. Several options using upper sodium plenum and SSFZ (Special Sodium Flowing Zone) for reducing sodium void reactivity are neutronically analyzed in this core concept in order to improve the inherent safety of the core. The SSFZ allowing the coolant flow from the peripheral fuel assemblies increases the neutron leakage under coolant expansion or voiding. The Monte Carlo calculations were used to design the cores and analyze their physics characteristics with heterogeneous models. The results of the design and analyses show that the final core design option has a small burnup reactivity swing of 618 pcm over ~54 EFPYs cycle length and a very small sodium void worth of ~35pcm at EOC (End of Cycle), which leads to the satisfaction of all the conditions for inherent safety with large margin based on the quasi-static reactivity balance analysis under ATWS (Anticipated Transient Without Scram).