The Experts below are selected from a list of 207 Experts worldwide ranked by ideXlab platform
Chenglin Zhu - One of the best experts on this subject based on the ideXlab platform.
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Conceptual Core design of an innovative small pwr utilizing fully ceramic microencapsulated fuel
Progress in Nuclear Energy, 2014Co-Authors: Xiang Dai, Xinrong Cao, Chenglin ZhuAbstract:The fully ceramic microencapsulated (FCM) fuel, which has the benefits of retaining fission products, high burnup, and proliferation resistance, is promising to fit in well with small reactors. This paper aims to combine FCM fuel concept with small PWR technology to design a 350 MWt PWR Core using FCM fuel, achieving soluble boron free (SBF) operation and at least five-year Core life without refueling. In the study, single batch refueling pattern is chosen to avoid frequent refueling. In addition, CASMO-4E/SIMULATE-3 code package is used for nuclear design calculation, and the double heterogeneity of FCM fuel is certified negligible. In FCM fuel design, the packing fraction of TRISO particles is chosen 46%, and with the TRISO particle design, the volume fraction of UO2 kernels in FCM fuel compact is achieved 17.87%. Calculation shows that under the single batch refueling pattern, directly using FCM fuel in conventional PWR design can hardly reach the long Core life purpose. In this Core design, fuel rod number, FCM fuel compact diameter, and axial active fuel height are all properly enlarged compared to output power level, leading to 13.6 t fuel inventory as well as about six-year Core life with 93 w/o UO2 enrichment. To achieve SBF operation in the Core, Pu-240 is added in some assemblies, and Gd2O3-UO2 rods are loaded in other assemblies, which together hold-down reactivity efficiently, and make the reactivity vary flatly with burnup changes. As many as 10 regulating banks are arranged since the integral worth of each bank is restricted to avoid adverse axial power distribution. With the combination of Pu-240 adding, Gd2O3-UO2 rods loading and control rods arrangement, normal operation and hot shut down can be achieved without soluble boron in the Core, and the cold shut down condition needs boric acid injection. It is estimated that the average center temperature of fuel kernel at fuel rod centerline is 845 K, which is much less than that of conventional PWR. (C) 2014 Elsevier Ltd. All rights reserved.
Lee Deokjung - One of the best experts on this subject based on the ideXlab platform.
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Core design of long-cycle small modular lead-cooled fast reactor
'Wiley', 2019Co-Authors: Tung Dong Cao Nguyen, Choe Jiwon, Ebiwonjumi Bamidele, Lemaire Matthieu, Lee DeokjungAbstract:A Core design of small modular liquid-metal fast reactor (SMLFR) cooled by lead-bismuth eutectic (LBE) was developed for power reactors. The main design constraint on this reactor is a size constraint: The Core needs to be small enough so that (1) it can be transported in a spent nuclear fuel (SNF) cask to meet the electricity demands in remote areas and off-grid locations or so that (2) it can be used as a power source on board of nuclear icebreaker ships. To satisfy this design requirement, the active Core of the reactor is 1 m in height and 1.45 m in diameter. The reactor is fueled with natural and 13.86% low-enriched uranium nitride (UN), as determined through an optimization study. The reactor was designed to achieve a thermal power of 37.5 MW with an assumption of 40% thermal efficiency by employing an advanced energy conversion system based on supercritical carbon dioxide (S-CO2) as working fluid, in which the Brayton cycle can achieve higher conversion efficiencies and lower costs compared to the Rankine cycle. The outer region of the Core with low-enriched uranium (LEU) performs the function of Core ignition. The center region plays the role of a breeding blanket to increase the Core lifetime for long cycle operation. The Core working fluid inlet and outlet temperatures are 300 degrees C and 422 degrees C, respectively. The primary coolant circulation is driven by an electromagnetic pump. Core performance characteristics were analyzed for isotopic inventory, criticality, radial and axial power profiles, shutdown margins (SDM), reactivity feedback coefficients, and integral reactivity parameters of the quasi-static reactivity balance. It is confirmed through depletion calculations with the fast reactor analysis code system Argonne Reactor Computation (ARC) that the designed reactor can be operated for 30 years without refueling. Preliminary thermal-hydraulic analysis at normal operation is also performed and confirms that the fuel and cladding temperatures are within normal operation range. The safety analysis performed with the ARC code system and the UNIST Monte Carlo code MCS shows that the Conceptual Core is favorable in terms of self-controllability, which is the first step towards inherent safety
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Conceptual Core Design of a Small Modular Fast Reactor Cooled by Lead-Bismuth Eutectic
Mexican Nuclear Society, 2018Co-Authors: Nguyen, Dong Cao Tung, Choe Jiwon, Ebiwonjumi Bamidele, Lemaire Matthieu, Lee DeokjungAbstract:A preliminary design of a Small Modular Fast Reactor (SMFR) cooled by Lead-Bismuth Eutectic (LBE) and based on the ALFRED Core (Advanced Lead-cooled Fast Reactor European Demonstrator) was achieved through three-dimensional simulation Core analysis with MC2-3/TWODANT/REBUS-3 - the deterministic code system from Argonne National Laboratory (ANL) for fast reactor analysis. The reactor was designed to achieve a thermal power of 37.5 MW with 40% thermal efficiency assumption. The two major design goals consisted in a cycle length of 30 Effective Full Power Years (EFPYs) and a small-size active Core which can be transported in a spent nuclear fuel (SNF) cask (1 m in height and less than 1.2 m in diameter). Several sensitivity tests on fuel material and geometry parameters were conducted to meet these two requirements. The selection of Uranium Nitride (UN) fuel allowed to improve significantly the fuel efficiency while the choice of a fuel pin radius of 0.52 cm, resulting in a fuel-to-coolant volume fraction of 1.413, enabled to increase the Core lifetime over 30 years without refueling. The reactivity control system was also evaluated by calculating the shutdown margin and it is checked that the control rods system can provide enough shutdown margins to cover the excess reactivity and the temperature defect. Further study is required to analyze the Core safety (void coefficient, fuel Doppler coefficients, expansion coefficients and radiation creep) and reduce the initial excess reactivity (by adding burnable poison, introducing enrichment zoning or increasing the number of control rods). Overall, it is confirmed that the Conceptual Core satisfies the target design ideas and this preliminary design work constitutes the very first step in a long Core design process
Alexander Agung - One of the best experts on this subject based on the ideXlab platform.
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Conceptual Core design study for indonesian space reactor isr
Progress in Nuclear Energy, 2020Co-Authors: Muhammad Farid Khandaq, Andang Widi Harto, Alexander AgungAbstract:Abstract Space exploration is very important for the future of the earth and human beings as it may eliminate earth overpopulation and overcome diminishing of earth resources. One of the obstacles of the space exploration mission is the energy source for the spacecraft. One alternative is using a nuclear reactor as an energy source in spacecraft. A Conceptual design of Indonesian Space Reactor (ISR) has been carried out to explore such a possibility. ISR is a liquid metal Na-78 K cooled space reactor with a fast neutron spectrum. It is designed to provide at least 500 kWth power for operating time more than 10 years at full power. The reactor uses 55% high-enriched uranium nitrate as fuel. The ISR hexagonal Core is comprised of 61 fuel pins and is designed in the form of a hollow cylinder with an individual cooling channel in each fuel pin. The reactor is also equipped with spectral shift absorbers (SSA) made of Re and Mo-30Re alloy to control the reactivity. Neutronic calculations have been performed to obtain optimum design parameters without compromising safety requirements. These design parameters include variation in uranium enrichment, reactor dimension, reflector thickness and control drum (absorber) design and dimension. The accepted reactor design has an excess reactivity of 4023 ± 9 pcm and shutdown margin of 4852 ± 9 pcm and the reactor is estimated to have a lifetime of 28 years. The temperature and void reactivity coefficients are all negative, implying inherent safety. Several accident scenarios were also considered in this work, both during launch failure and normal operation. It is found that to keep the reactor subcritical for a submerged reactor following a launch failure, the reflector segment should be discarded. Meanwhile, some portions of fuel pins should be removed from the Core during operational accidents.
Xiang Dai - One of the best experts on this subject based on the ideXlab platform.
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Conceptual Core design of an innovative small pwr utilizing fully ceramic microencapsulated fuel
Progress in Nuclear Energy, 2014Co-Authors: Xiang Dai, Xinrong Cao, Chenglin ZhuAbstract:The fully ceramic microencapsulated (FCM) fuel, which has the benefits of retaining fission products, high burnup, and proliferation resistance, is promising to fit in well with small reactors. This paper aims to combine FCM fuel concept with small PWR technology to design a 350 MWt PWR Core using FCM fuel, achieving soluble boron free (SBF) operation and at least five-year Core life without refueling. In the study, single batch refueling pattern is chosen to avoid frequent refueling. In addition, CASMO-4E/SIMULATE-3 code package is used for nuclear design calculation, and the double heterogeneity of FCM fuel is certified negligible. In FCM fuel design, the packing fraction of TRISO particles is chosen 46%, and with the TRISO particle design, the volume fraction of UO2 kernels in FCM fuel compact is achieved 17.87%. Calculation shows that under the single batch refueling pattern, directly using FCM fuel in conventional PWR design can hardly reach the long Core life purpose. In this Core design, fuel rod number, FCM fuel compact diameter, and axial active fuel height are all properly enlarged compared to output power level, leading to 13.6 t fuel inventory as well as about six-year Core life with 93 w/o UO2 enrichment. To achieve SBF operation in the Core, Pu-240 is added in some assemblies, and Gd2O3-UO2 rods are loaded in other assemblies, which together hold-down reactivity efficiently, and make the reactivity vary flatly with burnup changes. As many as 10 regulating banks are arranged since the integral worth of each bank is restricted to avoid adverse axial power distribution. With the combination of Pu-240 adding, Gd2O3-UO2 rods loading and control rods arrangement, normal operation and hot shut down can be achieved without soluble boron in the Core, and the cold shut down condition needs boric acid injection. It is estimated that the average center temperature of fuel kernel at fuel rod centerline is 845 K, which is much less than that of conventional PWR. (C) 2014 Elsevier Ltd. All rights reserved.
C Artioli - One of the best experts on this subject based on the ideXlab platform.
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Conceptual Core design study for a high flux lfr demonstrator
Progress in Nuclear Energy, 2012Co-Authors: Sara Bortot, Patrizio Console Camprini, Giacomo Grasso, C ArtioliAbstract:A preConceptual Core design study for a pool-type Lead-cooled Fast Reactor (LFR) demonstrator (DEMO) has been developed in the frame of an I-NERI between the Italian National Agency for the New Technologies, Energy and Sustainable Economic Development (ENEA) and Argonne National Laboratory (ANL), based on the European Lead-cooled System (ELSY) reference concept. A demonstration reactor is expected to prove the viability of technology to be implemented in the first-of-a-kind industrial power plant. DEMO specifications as a nuclear power facility demonstrating ELSY main features and performance, besides validating design methodology and tools, have been defined. Suitable design parameters have been set to meet the foremost objective of reaching a high fast neutron flux while respecting all technological constraints. Preliminary thermal-hydraulic analyses have been carried out to verify safety limits were not exceeded.