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The reactor assembly
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Ebook "Nuclear science and safety in Europe (NATO security through science series)" contains a selection of invited talks at the ARW. They are grouped in two chapters, one on results of theoretical and experimental studies of nuclear forces (strong interaction physics), and the other on their possible applications. Here, of particular interest is the coverage of the recent developments in the construction of safe sub-critical (assemblies) reactors.
289p
manmanthanhla0201
26-02-2024
2
1
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For the future of nuclear power, the design and development of an economical, accident tolerant fuel (ATF) for use in the current pressurized water reactors (PWRs) are highly desirable and essential. It is reported that the composite fuels are advantageous over the conventional UO2 fuel due to their higher thermal conductivities and a higher uranium densities.
12p
vicedric
08-02-2023
7
4
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The comparative results generally show a good agreement between TRITON and Serpent with the benchmark data, indicating that the TRITON and Serpent models developed herein for the PWR MOX assemblies can be applied to group constant generation to be further used in transient analyses of PWR MOX fueled cores.
5p
visnape
30-01-2023
7
4
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The VVER-1000 reactors are nowadays widespread used in Russia and numerous countries. With almost 3000 MW of thermal, VVER-1000 is a candidate to transmute minor actinides (MAs) which are the main contributors to the radiotoxicity and decay heat of nuclear spent fuel (SNF).
11p
visnape
30-01-2023
8
4
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In this paper, the Da Lat Nuclear Research Reactor (DNRR) core has been modeled by SCALE/TRITON code to generate two-group homogenized cross-sections for 3D kinetics calculations. In the calculation, plate-type model has been applied in selfshielding structure while the fuel assemblies have been grouped for the cross-section generation.
10p
vidumbledore
19-01-2023
13
4
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The Virtual Environment for Reactor Applications (VERA) core physics benchmark problem #6, 3D Hot Full Power (HFP) assembly, from the Consortium for Advanced Simulation of Light Water Reactors (CASL) was simulated using the MC21 continuous energy Monte Carlo code coupled with the COBRA-IE subchannel thermal-hydraulics code using the R5EXEC coupling framework.
14p
vironald
15-12-2022
7
4
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A simulation-based study concerning three non-destructive approaches by which water in spent nuclear fuel assemblies might be quantified with neutrons is described. Three fuel types have been considered: thirty-six spent Advanced Gas-cooled Reactor (AGR) fuel pins contained in a stainless-steel can; a prototype fast reactor (PFR) spent fuel assembly and a light water reactor (LWR) spent fuel assembly – with the PFR and LWR assemblies containing mixed oxide fuels.
25p
vironald
15-12-2022
13
4
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The moderator system in a CANDU reactor provides the unique ability to provide emergency cooling to the fuel in postulated severe accidents during the early phases of the transient. A key criteria which dictates the effectiveness of heat removal during these events is the integrity of the fuel channel assembly.
13p
vironald
15-12-2022
10
5
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Validating boiling water reactor (BWR) spent nuclear fuel inventory calculations is challenging due to the complexity of BWR assembly designs, the lack of publicly available radiochemical assay measurements, and limited access to documentation on fuel design and operating conditions.
15p
vironald
15-12-2022
9
3
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Gas-Cooled Fast reactor (GFR) is one of the favourable Generation IV type concepts under development. ALLEGRO reactor will be a demonstrator reactor of this gas-cooled fast reactor technology. Compared to conventional gas-cooled reactors in the case of a GFR reactor the power density is one order of magnitude higher, and the high temperature fuel assemblies are cooled by a low-density coolant (He), therefore special attention is paid to the heat transfer and heat removal performance of the reactor.
15p
vironald
15-12-2022
14
3
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This paper presents the analysis of sensitivity and uncertainty for the infinite multiplication factor (kinf) for the VVR-M2 typed HEU and LEU fuel assemblies of the Dalat nuclear research reactor (DNRR) using the MCNP6.1-Whisper1.1 code.
14p
vimelindagates
18-07-2022
7
2
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In this study, the SCALE/TRITON code (based on deterministic method) and the Serpent 2 code (based on Monte Carlo method) were utilized to prepare the group constants of the pressurized water reactor (PWR) mixed-oxide (MOX) fuel assemblies for transient analyses of PWR MOX fueled cores in normal operation and control rod ejection accident condition with 3D reactor kinetics codes.
10p
meyerowitz
25-12-2021
7
0
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This paper presents the application of an evolutionary simulated annealing (ESA) method to design a small 200 MWt reactor core. The core design is based on a reference ACPR50 reactor deployed in a floating nuclear power plant. The core consists of 37 typical 17x17 PWR fuel assemblies with three different U-235 enrichments of 4.45, 3.40 and 2.35 wt%.
8p
princessmononoke
29-11-2021
10
1
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"The journal of Nuclear science and technology - Volume 8/Number 4, 2018" present reactivity induced transient analysis when the occurrence of leakage in the dry irradiation channels of the Dalat Nuclear Research Reactor; burnup calculation of the OECD VVER-1000 LEU benchmark assembly using MCNP6 and SRAC2006; particle identification for neutron rich nuclei 63,65Cr from knockout reactions...
48p
trinhthamhodang1218
14-03-2021
13
1
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This paper presents assumptions, computer models and the results of analysis for such event in the DNRR by using MCNP5 code (code for neutronics analysis) and EUREKA-2/RR code (code for transient analysis). The calculation results include value of reactivity insertion, change in power of reactor, as well as surface temperature of the hottest fuel assembly. This research contributes to updating the reactor operating procedure.
9p
trinhthamhodang1218
14-03-2021
13
1
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This paper presents the conceptual design of a 300 MWt small modular reactor (SMR) using fuel assemblies of the AP1000 reactor. Numerical calculations have been performed to evaluate a proper active core size and core loading pattern using the SRAC code system and the JENDL-4.0 data library.
6p
trinhthamhodang1218
14-03-2021
13
2
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An investigation on the nuclear transmutation of elemental long-lived fission product (LLFP) in a fast reactor is being conducted focusing on the I-129 LLFP (half-life 15.7 million years) to reduce the environmental burden. The LLFP assembly is loaded into the radial blanket region of a Japanese MONJU class sodium-cooled fast reactor (710 MWth, 148 days/cycle). The iodine element containing I-129 LLFP (without isotope separation) is mixed with YD2 and/or YH2 moderator material to enhance the nuclear transmutation rate.
8p
trinhthamhodang1218
14-03-2021
10
2
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The main aim of this work was to design, fabricate and evaluate the performance of a small-scale solid waste pyrolysis waste converter reactor. The system’s main components were the furnace housing assembly, the reactor assembly, the piping system, the heat exchanger assembly (cooling condenser) and the collection system.
16p
cleopatrahuynh
01-06-2020
11
1
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The paper covers the results of VVER core reflooding studies in fuel assembly (FA) mockup of 126 fuel rod simulators with axial power peaking. The experiments were performed for two types of flooding. The first type is top flooding of the empty (steamed) FA mockup.
11p
christabelhuynh
30-05-2020
34
1
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This paper will present the progress in the project with respect to liquid metal cooled reactor thermal-hydraulics (liquid metal heat transfer, fuel assembly thermal-hydraulics, pool thermal-hydraulics, and system thermal-hydraulics).
8p
christabelhuynh
29-05-2020
9
2
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