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Journal articles on the topic 'Reactor system code'

Author: Grafiati

Published: 4 June 2021

Last updated: 8 February 2022

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1

Osusky, Filip, Stefan Cerba, Jakub Luley, Branislav Vrban, and Jan Hascik. "COUPLED SIMULATION OF GAS COOLED FAST REACTOR FUEL ASSEMBLY WITH NESTLE CODE SYSTEM." Acta Polytechnica CTU Proceedings 14 (May 17, 2018): 34. http://dx.doi.org/10.14311/app.2018.14.0034.

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The paper is focused on coupled calculation of the Gas Cooled Fast Reactor. The proper modelling of coupled neutronics and thermal-hydraulics is the corner stone for future safety assessment of the control and emergency systems. Nowadays, the system and channel thermal-hydraulic codes are accepted by the national regulatory authorities in European Union for license purposes, therefore the code NESTLE was used for the simulation. The NESTLE code is a coupled multigroup neutron diffusion code with thermal-hydraulic sub-channel code. In the paper, the validation of NESTLE code 5.2.1 installation

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Fable, E., C. Angioni, M. Siccinio, and H. Zohm. "Plasma physics for fusion reactor system codes: Framework and model code." Fusion Engineering and Design 130 (May 2018): 131–36. http://dx.doi.org/10.1016/j.fusengdes.2018.03.061.

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3

Wang, Yan. "The Prospect of the Modeling Simulation for the Passive Containment Cooling System of the Advanced PWR in China." Applied Mechanics and Materials 88-89 (August 2011): 82–87. http://dx.doi.org/10.4028/www.scientific.net/amm.88-89.82.

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The passive containment cooling system (PCCS) is one main component of the various passive safety systems in the advanced 3rd generation pressurized water reactor. Several containment analysis codes are modified for the simulation and analysis on the PCCS by researchers. These codes for the PCCS were validated by comparison with transient test data from some separate effect tests and integral tests, and used to evaluate the heat-removed capability of the PCCS under the postulated events, such as a loss-of-coolant accident (LOCA) and a main steam line break (MSLB). The advanced 3rd generation p

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4

Petruzzi, Alessandro, and Francesco D'Auria. "Thermal-Hydraulic System Codes in Nulcear Reactor Safety and Qualification Procedures." Science and Technology of Nuclear Installations 2008 (2008): 1–16. http://dx.doi.org/10.1155/2008/460795.

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In the last four decades, large efforts have been undertaken to provide reliable thermal-hydraulic system codes for the analyses of transients and accidents in nuclear power plants. Whereas the first system codes, developed at the beginning of the 1970s, utilized the homogenous equilibrium model with three balance equations to describe the two-phase flow, nowadays the more advanced system codes are based on the so-called “two-fluid model” with separation of the water and vapor phases, resulting in systems with at least six balance equations. The wide experimental campaign, constituted by the i

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Pérez Mañes, Jorge, Victor Hugo Sánchez Espinoza, Sergio Chiva, and Robert Stieglitz. "A New Coupled CFD/Neutron Kinetics System for High Fidelity Simulations of LWR Core Phenomena: Proof of Concept." Science and Technology of Nuclear Installations 2014 (2014): 1–13. http://dx.doi.org/10.1155/2014/294648.

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The Institute for Neutron Physics and Reactor Technology (INR) at the Karlsruhe Institute of Technology (KIT) is investigating the application of the meso- and microscale analysis for the prediction of local safety parameters for light water reactors (LWR). By applying codes like CFD (computational fluid dynamics) and SP3 (simplified transport) reactor dynamics it is possible to describe the underlying phenomena in a more accurate manner than by the nodal/coarse 1D thermal hydraulic coupled codes. By coupling the transport (SP3) based neutron kinetics (NK) code DYN3D with NEPTUNE-CFD, within a

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Hong, Bong-Guen, Dong-Won Lee, and Sang-Ryul In. "TOKAMAK REACTOR SYSTEM ANALYSIS CODE FOR THE CONCEPTUAL DEVELOPMENT OF DEMO REACTOR." Nuclear Engineering and Technology 40, no. 1 (2008): 87–92. http://dx.doi.org/10.5516/net.2008.40.1.087.

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Seungsu, Yuk, Tak Nam-il, and Chang Jo Keun. "DEVELOPMENT OF PIN-LEVEL NEUTRONICS/THERMAL-FLUID ANALYSIS COUPLED CODE SYSTEM FOR A BLOCK-TYPE HTGR CORE." EPJ Web of Conferences 247 (2021): 02041. http://dx.doi.org/10.1051/epjconf/202124702041.

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Recently, the coupling between computer codes that simulate different physical phenomena has attracted for more accurate analysis. In the case of high-temperature gas-cooled reactor (HTGR), the coupling between neutronics and thermal-fluid analysis is necessary because of large change of temperature in the reactor core. Korea Atomic Energy Research Institute (KAERI) has developed the coupled code system between a reactor physics analysis code CAPP and a thermal-fluid system safety analysis code GAMMA+ for a block-type HTGR. The CAPP/GAMMA+ coupled code system provides more accurate block-wise

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Oka, H., M. Nishikawa, T. Takeishi, et al. "Calculation Code of System Effect Using Serial Reactor Model." Fusion Science and Technology 41, no. 3P2 (2002): 658–62. http://dx.doi.org/10.13182/fst02-a22669.

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Nakata, Alexandre Ezzidi, Masanori Naitoh, and Chris Allison. "NEED OF A NEXT GENERATION SEVERE ACCIDENT CODE." JURNAL TEKNOLOGI REAKTOR NUKLIR TRI DASA MEGA 21, no. 3 (2019): 119. http://dx.doi.org/10.17146/tdm.2019.21.3.5630.

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Two international severe accident benchmark problems have been performed recently by using several existing parametric severe accident codes: The Benchmark Study of the Accident at the Fukushima Daiichi Nuclear Power Plant (BSAF) and the Benchmark of the In-Vessel Melt Retention (IVMR) Analysis of a VVER-1000 Nuclear Power Plant (NPP). The BSAF project was organized by the Nuclear Power Engineering Center (NUPEC) of the Institute of Applied Energy (IAE) in Japan for the three Boiling Water Reactors (BWRs) of the Fukushima NPP. The IVMR Project was organized by the Joint Research Center (JRC) o

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Luthfi, Wahid, and Surian Pinem. "VALIDATION OF SRAC CODE SYSTEM FOR NEUTRONIC PARAMETERS CALCULATION OF THE PWR MOX/UO2 CORE BENCHMARK." Urania : Jurnal Ilmiah Daur Bahan Bakar Nuklir 27, no. 1 (2021): 47. http://dx.doi.org/10.17146/urania.2021.27.1.6238.

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VALIDATION OF SRAC CODE SYSTEM FOR NEUTRONIC PARAMETERS CALCULATION OF THE PWR MOX/UO2 CORE BENCHMARK. Determination of neutronic parameter value is an important part in determining reactor safety, so accurate calculation results can be obtained. This study is focused on the validation of SRAC code system in the calculation of neutronic parameters value of a PWR (Pressurized Water Reactor) reactor core. MOX/UO2 Core Benchmark was choosed because it is used by several researchers as a reference core for code validation in the determination of neutronic parameters of a reactor core. The neutroni

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Bousbia-Salah, Anis, Areeya Jirapongmed, Tewfik Hamidouche, John White, Francesco D'Auria, and Martina Adorni. "Assessment of RELAP5 model for the University of Massachusetts Lowell Research Reactor." Nuclear Technology and Radiation Protection 21, no. 1 (2006): 3–12. http://dx.doi.org/10.2298/ntrp0601003b.

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RELAP5 is a system code developed at the Idaho National Environmental and Engineering Laboratory for thermal hydraulic analysis of nuclear reactors. The code RELAP5 is widely used for safety analysis studies of commercial nuclear power plants. However, recent released version of RELAP5/3.2 and over present significant capabilities for analysis of nuclear reactor research systems. As a contribution to the assessment of RELAP5/3.3 for research reactor safety analysis, experimental data from the University of Massachusetts Lowell Research Reactor - UMLRR are used. The UMLRR is a 1 MW light water

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Reux, C., L. Di Gallo, F. Imbeaux, et al. "DEMO reactor design using the new modular system code SYCOMORE." Nuclear Fusion 55, no. 7 (2015): 073011. http://dx.doi.org/10.1088/0029-5515/55/7/073011.

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13

Mikityuk, Konstantin, Sandro Pelloni, Paul Coddington, Evaldas Bubelis, and Rakesh Chawla. "FAST: An advanced code system for fast reactor transient analysis." Annals of Nuclear Energy 32, no. 15 (2005): 1613–31. http://dx.doi.org/10.1016/j.anucene.2005.06.002.

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14

Feghhi, S. A. H., S. Jafarikia, and F. Abtin. "Miniature neutron source reactor burnup calculations using IRBURN code system." Annals of Nuclear Energy 47 (September 2012): 242–48. http://dx.doi.org/10.1016/j.anucene.2012.04.016.

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15

Sun, Jun, Ximing Sun, and Yanhua Zheng. "Transient Study on the HTR-PM with TINTE-vPower Coupling Code Package." Science and Technology of Nuclear Installations 2020 (August 25, 2020): 1–14. http://dx.doi.org/10.1155/2020/5090597.

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The high-temperature gas-cooled reactor pebble-bed module (HTR-PM) nuclear power plant consists of two nuclear steam supply system modules, each of which drives the steam turbine by the superheated steam flow and is fed by the heated-up water flow. The shared steam/water system induces mutual effects on normal operation conditions and transients of the nuclear power plant, which is worthy of safety concerns and intensive study. In this paper, a coupling code package was developed with the TINTE and vPower codes to understand how the HTR-PM operated. The TINTE code was used to analyze the react

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16

JAEGER, Wadim, Wolfgang LISCHKE, and Victor Hugo SANCHEZ ESPINOZA. "ICONE15-10438 SAFETY RELATED INVESTIGATIONS OF THE VVER-1000 REACTOR TYPE BY THE COUPLED CODE SYSTEM TRACE/PARCS." Proceedings of the International Conference on Nuclear Engineering (ICONE) 2007.15 (2007): _ICONE1510. http://dx.doi.org/10.1299/jsmeicone.2007.15._icone1510_231.

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17

Asada, Yasuhide. "Japanese Activities Concerning Nuclear Codes and Standards—Part II." Journal of Pressure Vessel Technology 128, no. 1 (2005): 64–70. http://dx.doi.org/10.1115/1.2138063.

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This series of papers has been issued to give general views on recent Japanese activities related to nuclear codes and standards. Part II of the series describes component aging aspects and future trends. The component aging aspects include evaluation methods for vessels based on elastic-plastic fracture mechanics, environmental fatigue evaluation guidelines and inspection and evaluation guidelines for reactor internals. With respect to future trends, the development of the International Thermonuclear Experimental Reactor code and the System-Based Code are introduced.

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18

Pesic, Milan. "MCNP5 study on kinetics parameters of coupled fast-thermal system HERBE." Nuclear Technology and Radiation Protection 26, no. 2 (2011): 134–39. http://dx.doi.org/10.2298/ntrp1102134p.

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New validation of the well-known Monte Carlo code MCNP5 against measured criticality and kinetics data for the coupled fast-thermal HERBE System at the Reactor B critical assembly is shown in this paper. Results of earlier calculations of these criticality and kinetics parameters, done by combination of transport and diffusion codes using two-dimension geometry model are compared to results of new calculations carried out by the MCNP5 code in three-dimension geometry. Satisfactory agreements in comparison of new results with experimental data, in spite complex heterogeneous composition of the

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19

Li, Dong, Cheng Peng, and Rao Hao. "Simulation investigation of ocean motion on small reactor based on modified system code." E3S Web of Conferences 194 (2020): 01012. http://dx.doi.org/10.1051/e3sconf/202019401012.

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Offshore nuclear small reactor is a promising research direction during recent years. In order to carry out system simulation and safety analysis under accident conditions for marine reactors, the classic thermal-hydraulic system analysis code RELAP5 is modified by taking heaving, inclining and rolling motion into consideration in this paper. Additional forces are added to the momentum equation for heaving and rolling, and control volume coordination are transformed for inclining and rolling. Then, a simplified small reactor with two loops is simulated by the modified code and the effect of oc

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20

Kim, Hyoung Tae, Se-Myong Chang, Jong-Hyeon Shin, and Yong Gwon Kim. "The Feasibility of Multidimensional CFD Applied to Calandria System in the Moderator of CANDU-6 PHWR Using Commercial and Open-Source Codes." Science and Technology of Nuclear Installations 2016 (2016): 1–10. http://dx.doi.org/10.1155/2016/3194839.

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The moderator system of CANDU, a prototype of PHWR (pressurized heavy-water reactor), has been modeled in multidimension for the computation based on CFD (computational fluid dynamics) technique. Three CFD codes are tested in modeled hydrothermal systems of heavy-water reactors. Commercial codes, COMSOL Multiphysics and ANSYS-CFX with OpenFOAM, an open-source code, are introduced for the various simplified and practical problems. All the implemented computational codes are tested for a benchmark problem of STERN laboratory experiment with a precise modeling of tubes, compared with each other a

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21

Patrulescu, I., M. Constantin, E. Rades, and V. Balaceanu. "Development of CANDU reactor physics calculation system based on WIMS code." Annals of Nuclear Energy 24, no. 14 (1997): 1105–25. http://dx.doi.org/10.1016/s0306-4549(96)00101-6.

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22

Aboanber, Ahmed Ebrahim, Naohiro Hirakawa, and Mohammed Omar Shaker. "Efficient code system of core characteristics for small cylindrical thermal reactor." Annals of Nuclear Energy 26, no. 6 (1999): 489–508. http://dx.doi.org/10.1016/s0306-4549(98)00065-6.

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23

Hermawan, Dedy P., Rionaldy, and Syarip. "Neutronic Analysis of SAMOP Reactor Experimental Facility Using SCALE Code System." Journal of Physics: Conference Series 1090 (September 2018): 012032. http://dx.doi.org/10.1088/1742-6596/1090/1/012032.

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24

Wang, Guanbo, Dazhi Qian, Jimin Ma, et al. "Fuel management code system against experimental measurements of the CMRR reactor." Annals of Nuclear Energy 130 (August 2019): 124–30. http://dx.doi.org/10.1016/j.anucene.2019.02.038.

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25

Prošek, Andrej, and Marko Matkovič. "RELAP5/MOD3.3 Analysis of the Loss of External Power Event with Safety Injection Actuation." Science and Technology of Nuclear Installations 2018 (2018): 1–14. http://dx.doi.org/10.1155/2018/6964946.

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The code assessment typically comprises basic tests cases, separate effects test, and integral effects tests. On the other hand, the thermal hydraulic system codes like RELAP5/MOD3.3 are primarily intended for simulation of transients and accidents in light water reactors. The plant measured data come mostly from startup tests and operational events. Also, for operational events the measured plant data may not be sufficient to explain all details of the event. The purpose of this study was therefore besides code assessment to demonstrate that simulations can be very beneficial for deep underst

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26

Zizin, M. N., V. F. Boyarinov, V. A. Nevinitsa, P. A. Fomichenko, Yu N. Volkov, and A. E. Kruglikov. "Verification of the stationary module of the ShIPR software system for modelling experiments of the ASTRA HTGR type critical facility." Kerntechnik 85, no. 1 (2020): 4–8. http://dx.doi.org/10.1515/kern-2020-850103.

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Abstract Coupled neutronic and thermal hydraulic calculation codes are verified for calculating the design of modern and prospective types of nuclear reactors. This verification is done by comparing experimental and calculated results for stationary and transient conditions. This paper presents ShIPR (Shell of Intelligent Package for Reactor) Integrated Development Environment with automatic generation of head programs based on the chain of computational modules. The aim of this study is to find the reason of a discrepancy in the modelling of sub-critical states that was found in previous work

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27

Milosevic, Miodrag, Ehud Greenspan, and Jasmina Vujic. "Monte Carlo with fuel burnup method for the ENHS benchmark calculations." Nuclear Technology and Radiation Protection 18, no. 2 (2003): 3–11. http://dx.doi.org/10.2298/ntrp0302003m.

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Estimates of the uncertainties arising from approximations in the methods used in different nuclear data processing and neutron transport codes are usually obtained by inter-comparing calculations made using different code systems. This paper gives details of an investigation of differences between results obtained by using different codes for a single zone model of the Encapsulated Nuclear Heat Source (ENHS) benchmark core fuelled with metallic alloy of Pu, U, and Zr. The ENHS is a new lead-bismuth or lead cooled novel reactor concept for 20 effective full power years without refuelling and w

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Oggioni, Carlo, Chris Keckler, and Massimiliano Fratoni. "ANALYSIS OF MODIFIED PASSIVE SAFETY SYSTEM IN FAST RECTORS TRANSIENTS." EPJ Web of Conferences 247 (2021): 07016. http://dx.doi.org/10.1051/epjconf/202124707016.

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The Autonomous Reactivity Control (ARC) system is a passive safety system aiming to provide an additional negative reactivity feedback during reactor transient scenarios. This paper shows how the performance of the ARC system can be enhanced by introducing a hydraulic diode that allows for different engagement and disengagement speeds of system. The benefits of the proposed system is assessed in a reference soidum-cooled fast reactor (SFR) during multiple postulated transient scenarios. The reactor and plant dynamic response are evaluated using SAS4A/SASSYS-1, whereas for the internal ARC syst

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Aboanber, A. E., and Y. M. Hamada. "PWS: an efficient code system for solving space-independent nuclear reactor dynamics." Annals of Nuclear Energy 29, no. 18 (2002): 2159–72. http://dx.doi.org/10.1016/s0306-4549(02)00034-8.

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30

Mandelli, Diego, Steven Prescott, Curtis Smith, et al. "A Flooding Induced Station Blackout Analysis for a Pressurized Water Reactor Using the RISMC Toolkit." Science and Technology of Nuclear Installations 2015 (2015): 1–14. http://dx.doi.org/10.1155/2015/308163.

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In this paper we evaluate the impact of a power uprate on a pressurized water reactor (PWR) for a tsunami-induced flooding test case. This analysis is performed using the RISMC toolkit: the RELAP-7 and RAVEN codes. RELAP-7 is the new generation of system analysis codes that is responsible for simulating the thermal-hydraulic dynamics of PWR and boiling water reactor systems. RAVEN has two capabilities: to act as a controller of the RELAP-7 simulation (e.g., component/system activation) and to perform statistical analyses. In our case, the simulation of the flooding is performed by using an adv

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31

Ma, Yugao, Minyun Liu, Erhui Chen, et al. "RMC/ANSYS MULTI-PHYSICS COUPLING SOLUTIONS FOR HEAT PIPE COOLED REACTORS ANALYSES." EPJ Web of Conferences 247 (2021): 06007. http://dx.doi.org/10.1051/epjconf/202124706007.

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The heat pipe cooled reactor is a solid-state reactor using heat pipes to passively transfer heat generated from the reactor, which is a potential and near-term space nuclear power system. This paper introduces the coupling scheme between the continuous energy Reactor Monte Carlo (RMC) code and the finite element method commercial software ANSYS. Monte Carlo method has the advantages of flexible geometry modeling and continuous-energy nuclear cross sections. ANSYS Parametric Design Language (APDL) is used to determine the detailed temperature distributions and geometric deformation. The on-the

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Hong, Bong Guen, Dong Won Lee, Suk-Kwon Kim, and Yonghee Kim. "Development of a tokamak reactor system code and its application for concept development of a demo reactor." Fusion Engineering and Design 83, no. 10-12 (2008): 1615–18. http://dx.doi.org/10.1016/j.fusengdes.2008.07.037.

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Acharya, Govatsa, Fredrik Dehlin, Sara Bortot, and Ignas Mickus. "INVESTIGATION OF A SELF-ACTUATED, GRAVITY-DRIVEN SHUTDOWN SYSTEM IN A SMALL LEAD-COOLED REACTOR." EPJ Web of Conferences 247 (2021): 07007. http://dx.doi.org/10.1051/epjconf/202124707007.

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Passive safety systems in a nuclear reactor allow to simplify the overall plant design, beside improving economics and reliability, which are considered to be among the salient goals of advanced Generation IV reactors. This work focuses on investigating the application of a self-actuated, gravity-driven shutdown system in a small lead-cooled fast reactor and its dynamic response to an initiating event. The reactor thermal-hydraulics and neutronics assessment were performed in advance. According to a first-order approximation approach, the passive insertion of shutdown assembly was assumed to b

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Du, Xue, Junsheng Hu, and Genglei Xia. "Operation characteristic of supercritical carbon dioxide-cooled reactor system under coordination control scheme." International Journal of Advanced Robotic Systems 17, no. 3 (2020): 172988142093383. http://dx.doi.org/10.1177/1729881420933833.

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Nuclear power plant is completely independent of air and has high power density, which improves unmanned marine vehicle maneuvering performance, maximum speed, and concealment. The supercritical carbon dioxide-cooled reactor system that has the advantages of compact system and high cycle efficiency is a promising energy conversion system candidate for generation IV reactors. However, due to the dramatic variations of supercritical fluid thermophysical properties, the operation of supercritical carbon dioxide-cooled reactor system is complex and it is difficult to control it. In this article, a

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35

Morreale, A. C., M. J. Brown, and S. M. Petoukhov. "PRELIMINARY METHODOLOGY FOR THE ANALYSIS OF THE NATIONAL RESEARCH UNIVERSAL REACTOR USING INTEGRATED SEVERE ACCIDENT MODELLING CODES." AECL Nuclear Review 4, no. 1 (2015): 53–65. http://dx.doi.org/10.12943/anr.2014.00035.

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The National Research Universal (NRU) Reactor is a multi-purpose research reactor located at Atomic Energy of Canada Limited (AECL) Chalk River Laboratories. The severe accident case for the NRU has been explored through deterministic and probabilistic safety analysis (PSA) including multi-level PSAs that detail the progression and consequences of a severe accident in the NRU. These previous calculations lack the interconnected and comprehensive features of a full severe accident modelling code that is now the standard for severe accident analysis of power reactors. It was of interest within A

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Luthfi, Wahid, and Surian Pinem. "CALCULATION OF 2-DIMENSIONAL PWR MOX/UO2 CORE BENCHMARK OECD NEA 6048 WITH SRAC CODE." JURNAL TEKNOLOGI REAKTOR NUKLIR TRI DASA MEGA 22, no. 3 (2020): 89. http://dx.doi.org/10.17146/tdm.2020.22.3.5955.

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The mixed uranium-plutonium oxide fuel (MOX/UO2) is an interesting fuel for future power reactors. This is due to the large amount of plutonium that can be processed from spent fuel of nuclear plants or from plutonium weapons. MOX/UO2 fuel is very flexible to be applied in thermal reactors such as PWR and it is more economical than UO2 fuel. However, due to the different nature of neutron interactions of MOX in PWR, it will change the reactor core design parameters and also its safety characteristic. The purpose of this study is to determine the accuracy of SRAC2006 code system in generation o

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37

Avramova, M., A. Velazquez-Lozada, and A. Rubin. "Comparative Analysis of CTF and Trace Thermal-Hydraulic Codes Using OECD/NRC PSBT Benchmark Void Distribution Database." Science and Technology of Nuclear Installations 2013 (2013): 1–12. http://dx.doi.org/10.1155/2013/725687.

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The international OECD/NRC PSBT benchmark has been established to provide a test bed for assessing the capabilities of thermal-hydraulic codes and to encourage advancement in the analysis of fluid flow in rod bundles. The benchmark was based on one of the most valuable databases identified for the thermal-hydraulics modeling developed by NUPEC, Japan. The database includes void fraction and departure from nucleate boiling measurements in a representative PWR fuel assembly. On behalf of the benchmark team, PSU in collaboration with US NRC has performed supporting calculations using the PSU in-h

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Cho, Yun-Je, Hyoung-Kyu Cho, and Goon-Cherl Park. "Assessment of MARS-GCR Code for a New Reactor Cavity Cooling System Design." Nuclear Technology 162, no. 1 (2008): 92–106. http://dx.doi.org/10.13182/nt08-a3935.

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Chung, Y. J., I. S. Jun, S. H. Kim, S. H. Yang, H. R. Kim, and W. J. Lee. "Development and assessment of system analysis code, TASS/SMR for integral reactor, SMART." Nuclear Engineering and Design 244 (March 2012): 52–60. http://dx.doi.org/10.1016/j.nucengdes.2011.12.013.

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40

Lee, Yeon-Gun, Jong-Won Kim, and Goon-Cherl Park. "Development of a thermal–hydraulic system code, TAPINS, for 10MW regional energy reactor." Nuclear Engineering and Design 249 (August 2012): 364–78. http://dx.doi.org/10.1016/j.nucengdes.2012.04.020.

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Warmer, F., S. B. Torrisi, C. D. Beidler, et al. "System Code Analysis of HELIAS-Type Fusion Reactor and Economic Comparison With Tokamaks." IEEE Transactions on Plasma Science 44, no. 9 (2016): 1576–85. http://dx.doi.org/10.1109/tps.2016.2545868.

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42

Moloko, Lesego E., and Mohamed G. A. H. Belal. "Improved SAFARI-1 Research Reactor Irradiation Position Modeling in OSCAR-3 Code System." IEEE Transactions on Nuclear Science 58, no. 4 (2011): 1907–12. http://dx.doi.org/10.1109/tns.2011.2156428.

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TSUCHIHASHI, Keichiro. "Development, Validation and Applications of SRAC: JAERI Thermal Reactor Standard Neutronics Code System." Journal of Nuclear Science and Technology 26, no. 1 (1989): 23–27. http://dx.doi.org/10.1080/18811248.1989.9734265.

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KUGO, Teruhiko, Masayuki NAKAGAWA, and Keichiro TSUCHIHASHI. "Development of Intelligent Code System to Support Conceptual Design of Nuclear Reactor Core." Journal of Nuclear Science and Technology 34, no. 8 (1997): 760–70. http://dx.doi.org/10.1080/18811248.1997.9733740.

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FUKUMURA, Nobuo, Kazuhiro TANJI, Atsushi TAKAMURA, and Takeshi SUGITA. "Development of Radioactivity Estimation Code System after Final Reactor Shut Down of FBR." Proceedings of the National Symposium on Power and Energy Systems 2004.9 (2004): 465–72. http://dx.doi.org/10.1299/jsmepes.2004.9.465.

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Meftah, B., A. Halilou, F. Letaim, S. Mazidi, M. Y. Mokeddem, and F. Zeggar. "BURNUR.SYS: A 2-D code system for NUR research reactor burn up analysis." Annals of Nuclear Energy 35, no. 4 (2008): 591–600. http://dx.doi.org/10.1016/j.anucene.2007.08.020.

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Cui, Manman, Yun Guo, and Zhijian Zhang. "Transient simulation code development of primary coolant system of Chinese Experimental Fast Reactor." Annals of Nuclear Energy 53 (March 2013): 158–69. http://dx.doi.org/10.1016/j.anucene.2012.09.022.

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Aksan, N. "International Standard Problems and Small Break Loss-of-Coolant Accident (SBLOCA)." Science and Technology of Nuclear Installations 2008 (2008): 1–22. http://dx.doi.org/10.1155/2008/814572.

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Best-estimate thermal-hydraulic system codes are widely used to perform safety and licensing analyses of nuclear power plants and also used in the design of advance reactors. Evaluation of the capabilities and the performance of these codes can be accomplished by comparing the code predictions with measured experimental data obtained on different test facilities. OECD/NEA Committee on the Safety of Nuclear Installations (CSNI) has promoted, over the last twenty-nine years, some forty-eight international standard problems (ISPs). These ISPs were performed in different fields as in-vessel therma

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Park, Hyun-Sik, Byung-Yeon Min, Youn-Gyu Jung, Yong-Cheol Shin, Yung-Joo Ko, and Sung-Jae Yi. "Design of the VISTA-ITL Test Facility for an Integral Type Reactor of SMART and a Post-Test Simulation of a SBLOCA Test." Science and Technology of Nuclear Installations 2014 (2014): 1–14. http://dx.doi.org/10.1155/2014/840109.

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To validate the performance and safety of an integral type reactor of SMART, a thermal-hydraulic integral effect test facility, VISTA-ITL, is introduced with a discussion of its scientific design characteristics. The VISTA-ITL was used extensively to assess the safety and performance of the SMART design, especially for its passive safety system such as a passive residual heat removal system, and to validate various thermal-hydraulic analysis codes. The VISTA-ITL program includes several tests on the SBLOCA, CLOF, and PRHRS performances to support a verification of the SMART design and contribu

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Zhang, D. L., P. Song, S. Wang, et al. "Analysis code development for the direct reactor auxiliary cooling system of the pool-type sodium-cooled fast reactor." Kerntechnik 83, no. 3 (2018): 232–36. http://dx.doi.org/10.3139/124.110878.

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