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Journal articles on the topic 'Best-Estimate code'

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Author: Grafiati

Published: 14 March 2023

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1

Prošek, Andrej, and Borut Mavko. "RELAP5/MOD3.3 Best Estimate Analyses for Human Reliability Analysis." Science and Technology of Nuclear Installations 2010 (2010): 1–12. http://dx.doi.org/10.1155/2010/797193.

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To estimate the success criteria time windows of operator actions the conservative approach was used in the conventional probabilistic safety assessment (PSA). The current PSA standard recommends the use of best-estimate codes. The purpose of the study was to estimate the operator action success criteria time windows in scenarios in which the human actions are supplement to safety systems actuations, needed for updated human reliability analysis (HRA). For calculations the RELAP5/MOD3.3 best estimate thermal-hydraulic computer code and the qualified RELAP5 input model representing a two-loop pressurized water reactor, Westinghouse type, were used. The results of deterministic safety analysis were examined what is the latest time to perform the operator action and still satisfy the safety criteria. The results showed that uncertainty analysis of realistic calculation in general is not needed for human reliability analysis when additional time is available and/or the event is not significant contributor to the risk.

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2

Gonfiotti, Bruno, Michela Angelucci, Bradut-Eugen Ghidersa, Xue Zhou Jin, Mihaela Ionescu-Bujor, Sandro Paci, and Robert Stieglitz. "Best-Estimate for System Codes (BeSYC): A New Software to Perform Best-Estimate Plus Uncertainty Analyses with Thermal-Hydraulic and Safety System Codes for Both Fusion and Fission Scenarios." Applied Sciences 12, no. 1 (December 29, 2021): 311. http://dx.doi.org/10.3390/app12010311.

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The development and the validation of old and new software in relevant DEMO reactor conditions have been exploited in the latest years within the EUROfusion Consortium. The aim was to use—if possible—the software already validated for fission reactors and to fill the gaps with new ad-hoc software. As contribution to this effort, the Karlsruhe Institute of Technology (KIT) developed and tested a novel software to apply the Best-Estimate Model Calibration and Prediction through Experimental Data Assimilation methodology to the system codes RELAP5-3D, MELCOR 1.8.6, and MELCOR 2.2. This software is called Best-estimate for SYstem Codes (BeSYC), and it is developed as a MATLAB App. The application is in charge of applying the mathematical framework of the methodology, writing and executing the code runs required by the methodology, and printing the obtained results. The main goal of BeSYC is to wrap up the methodology in a software suitable to be used by any user through a simple graphical user interface. Albeit developed in the fusion research context, BeSYC can be applied to any reactor/scenario type supported by the specific system code. The goals of BeSYC, the mathematical framework, the main characteristics, and the performed verification and validation activities are described in this paper.

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3

Takeuchi, K., and M. Y. Young. "Assessment of flooding in a best estimate thermal hydraulic code (W̱COBRA/TRAC)." Nuclear Engineering and Design 186, no. 1-2 (November 1998): 225–55. http://dx.doi.org/10.1016/s0029-5493(98)00224-6.

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4

Cross, A. W., D. P. DiVincenzo, and B. M. Terhal. "A comparative code study for quantum fault tolerance." Quantum Information and Computation 9, no. 7&8 (July 2009): 541–72. http://dx.doi.org/10.26421/qic9.7-8-1.

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We study a comprehensive list of quantum codes as candidates for codes used at the physical level in a fault-tolerant code architecture. Using the Aliferis-Gottesman-Preskill (AGP) ex-Rec method we calculate the pseudo-threshold for these codes against depolarizing noise at various levels of overhead. We estimate the logical noise rate as a function of overhead at a physical error rate of $p_0=1 \times 10^{-4}$. The Bacon-Shor codes and the Golay code are the best performers in our study.

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5

FURUYA, Masahiro, Yoshihisa Nishi, and Nobuyuki Ueda. "S083034 Predictability of Best-Estimate Code, TRACE for Flashing-Induced Density Wave Oscillations." Proceedings of Mechanical Engineering Congress, Japan 2012 (2012): _S083034–1—_S083034–5. http://dx.doi.org/10.1299/jsmemecj.2012._s083034-1.

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6

Carlos, S., F. Sanchez-Saez, and S. Martorell. "Use of TRACE best estimate code to analyze spent fuel storage pools safety." Progress in Nuclear Energy 77 (November 2014): 224–38. http://dx.doi.org/10.1016/j.pnucene.2014.07.008.

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7

Vinai, Paolo, Rafael Macian-Juan, and Rakesh Chawla. "A statistical methodology for quantification of uncertainty in best estimate code physical models." Annals of Nuclear Energy 34, no. 8 (August 2007): 628–40. http://dx.doi.org/10.1016/j.anucene.2007.03.003.

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8

Petruzzi, Alessandro, Francesco D'Auria, Tomislav Bajs, Francesc Reventos, and Yassin Hassan. "International Course to Support Nuclear Licensing by User Training in the Areas of Scaling, Uncertainty, and 3D Thermal-Hydraulics/Neutron-Kinetics Coupled Codes: 3D S.UN.COP Seminars." Science and Technology of Nuclear Installations 2008 (2008): 1–16. http://dx.doi.org/10.1155/2008/874023.

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Thermal-hydraulic system computer codes are extensively used worldwide for analysis of nuclear facilities by utilities, regulatory bodies, nuclear power plant designers, vendors, and research organizations. The computer code user represents a source of uncertainty that can influence the results of system code calculations. This influence is commonly known as the “user effect” and stems from the limitations embedded in the codes as well as from the limited capability of the analysts to use the codes. Code user training and qualification represent an effective means for reducing the variation of results caused by the application of the codes by different users. This paper describes a systematic approach to training code users who, upon completion of the training, should be able to perform calculations making the best possible use of the capabilities of best estimate codes. In other words, the program aims at contributing towards solving the problem of user effect. In addition, this paper presents the organization and the main features of the 3D S.UN.COP (scaling, uncertainty, and 3D coupled code calculations) seminars during which particular emphasis is given to the areas of the scaling, uncertainty, and 3D coupled code analysis.

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9

Vileiniskis, V., and A. Kaliatka. "Best estimate analysis of PHEBUS FPT1 experiment bundle phase using ASTEC code ICARE module." Kerntechnik 76, no. 4 (August 2011): 254–60. http://dx.doi.org/10.3139/124.110158.

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10

Bousbia-Salah, Anis, and Francesco D'Auria. "Use of coupled code technique for Best Estimate safety analysis of nuclear power plants." Progress in Nuclear Energy 49, no. 1 (January 2007): 1–13. http://dx.doi.org/10.1016/j.pnucene.2006.10.002.

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11

Vadi, Roozbeh, and Kamran Sepanloo. "Investigation of a LOCA in a typical MTR by a novel best-estimate code." Progress in Nuclear Energy 86 (January 2016): 141–61. http://dx.doi.org/10.1016/j.pnucene.2015.10.013.

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12

Agnello, G., P. A. Di Maio, A. Bersano, and F. Mascari. "Cold Leg LBLOCA uncertainty analysis using TRACE/DAKOTA coupling." Journal of Physics: Conference Series 2177, no. 1 (April 1, 2022): 012023. http://dx.doi.org/10.1088/1742-6596/2177/1/012023.

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Abstract Safety analyses for nuclear power plants were carried out in the past using a conservative approach. With the increase of the phenomenological knowledge, through experimental data, and computational power, it became possible to adopt best estimate thermal- hydraulic system codes to perform deterministic safety analyses. However, some uncertainties are still present in the models, correlations, initial and boundary conditions, etc. Therefore, it is fundamental to quantify the uncertainty of calculation. This approach is called “Best Estimate Plus Uncertainty” (BEPU). Among the available uncertainty analysis methodologies, the probabilistic method to propagate input uncertainty is widely adopted. In the present study, an uncertainty analysis of a cold leg large break loss of coolant accident in a generic PWR-900 MWe has been developed and it has been carried out coupling the best estimate thermal- hydraulic system code TRACE and the uncertainty quantification tool DAKOTA in the SNAP environment/architecture.

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13

Bae, Sung-Won, and Bub-Dong Chung. "DEVELOPMENT OF THE MULTI-DIMENSIONAL HYDRAULIC COMPONENT FOR THE BEST ESTIMATE SYSTEM ANALYSIS CODE MARS." Nuclear Engineering and Technology 41, no. 10 (December 31, 2009): 1347–60. http://dx.doi.org/10.5516/net.2009.41.10.1347.

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14

Kang, Seok-Ju, Hae-Yong Jeong, Sung-Won Bae, Chi-Woong Choi, Kwi-Seok Ha, and Jae-Seung Suh. "Best estimate calculation and uncertainty quantification of sodium-cooled fast reactor using MARS-LMR code." Annals of Nuclear Energy 115 (May 2018): 138–53. http://dx.doi.org/10.1016/j.anucene.2018.01.033.

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15

Sengupta, Samiran, Vijay K. Veluri, R. Patel, S. Mammen, and S. Bhattacharya. "Application of best estimate thermal hydraulic code for safety analysis of high flux research reactor." Life Cycle Reliability and Safety Engineering 6, no. 2 (June 2017): 109–17. http://dx.doi.org/10.1007/s41872-017-0011-y.

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16

Glaeser, Horst. "GRS Method for Uncertainty and Sensitivity Evaluation of Code Results and Applications." Science and Technology of Nuclear Installations 2008 (2008): 1–7. http://dx.doi.org/10.1155/2008/798901.

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During the recent years, an increasing interest in computational reactor safety analysis is to replace the conservative evaluation model calculations by best estimate calculations supplemented by uncertainty analysis of the code results. The evaluation of the margin to acceptance criteria, for example, the maximum fuel rod clad temperature, should be based on the upper limit of the calculated uncertainty range. Uncertainty analysis is needed if useful conclusions are to be obtained from “best estimate” thermal-hydraulic code calculations, otherwise single values of unknown accuracy would be presented for comparison with regulatory acceptance limits. Methods have been developed and presented to quantify the uncertainty of computer code results. The basic techniques proposed by GRS are presented together with applications to a large break loss of coolant accident on a reference reactor as well as on an experiment simulating containment behaviour.

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17

Jeong, Hae-Yong. "Analysis of Loss of Condenser Vacuum Accident using a Conservative Approach with a Best-Estimate Code." Journal of Energy Engineering 24, no. 4 (December 31, 2015): 175–82. http://dx.doi.org/10.5855/energy.2015.24.4.175.

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18

Lim, Chul-Kyu, Sang-Koo Han, Sook-Kwan Kim, Dong-Sik Jin, Bong-Jin Ko, Yong-Ho Hong, Kwang-Il Ahn, et al. "Best-estimate analysis for a MSGTR accident of CANDU-6 plants using the MAAP-ISAAC code." Nuclear Engineering and Design 359 (April 2020): 110452. http://dx.doi.org/10.1016/j.nucengdes.2019.110452.

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19

Ceuca, Cristian Sabin, H. Austregesilo, and R. Macian-Juan. "BEST-ESTIMATE SIMULATIONS OF CONDENSATION-INDUCED WATER HAMMER IN HORIZONTAL PIPES WITH THE SYSTEM ANALYSIS CODE ATHLET." Multiphase Science and Technology 26, no. 4 (2014): 305–27. http://dx.doi.org/10.1615/multscientechn.v26.i4.30.

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20

Marao, A., T. Kaliatka, A. Kaliatka, and E. Ušpuras. "Adaptation of the FEMAXI-6 code and RBMK fuel rods model testing employing the best estimate approach." Kerntechnik 75, no. 3 (April 2010): 72–80. http://dx.doi.org/10.3139/124.110069.

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21

Kim, Kyung-Doo, Jae-Jun Jeong, Seung-Wook Lee, Myeong-Soo Lee, Jae-Seung Suh, Jin-Hyuk Hong, and Yong-Kwan Lee. "Development of NSSS Thermal-Hydraulic Model for KNPEC-2 Simulator Using the Best-Estimate Code RETRAN-3D." Nuclear Technology 148, no. 1 (October 2004): 3–11. http://dx.doi.org/10.13182/nt04-a3543.

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22

TAKIWAKI, Kenya, Yutaka TAKEUCHI, Norio SAKAI, and Taku KITAMURA. "A205 Development of BOP models for best estimate transient analysis code and application to plant cycle evaluation." Proceedings of the National Symposium on Power and Energy Systems 2011.16 (2011): 225–26. http://dx.doi.org/10.1299/jsmepes.2011.16.225.

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23

Hadjam, Ahmed, Ferhat Souidi, Ahcene Loubar, and Marcel Weber. "Simulation of a LBLOCA in the CALLISTO test facility using the best estimate computer code RELAP5/SCDAP3.2." Nuclear Engineering and Design 262 (September 2013): 153–67. http://dx.doi.org/10.1016/j.nucengdes.2013.03.052.

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24

Taleyarkhan, R. P., A. F. McFarlane, R. T. Lahey, and M. Z. Podowski. "Benchmarking and qualification of the nufreq-npw code for best estimate prediction of multichannel core stability margins." Nuclear Engineering and Design 151, no. 1 (November 1994): 157–71. http://dx.doi.org/10.1016/0029-5493(94)90040-x.

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25

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-house advanced thermal-hydraulic subchannel code CTF and the US NRC system code TRACE. CTF is a version of COBRA-TF whose models have been continuously improved and validated by the RDFMG group at PSU. TRACE is a reactor systems code developed by US NRC to analyze transient and steady-state thermal-hydraulic behavior in LWRs and it has been designed to perform best-estimate analyses of LOCA, operational transients, and other accident scenarios in PWRs and BWRs. The paper presents CTF and TRACE models for the PSBT void distribution exercises. Code-to-code and code-to-data comparisons are provided along with a discussion of the void generation and void distribution models available in the two codes.

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26

Adorni, Martina, Alessandro Del Nevo, Francesco D'Auria, and Oscar Mazzantini. "A Procedure to Address the Fuel Rod Failures during LB-LOCA Transient in Atucha-2 NPP." Science and Technology of Nuclear Installations 2011 (2011): 1–11. http://dx.doi.org/10.1155/2011/929358.

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Depending on the specific event scenario and on the purpose of the analysis, the availability of calculation methods that are not implemented in the standard system thermal hydraulic codes might be required. This may imply the use of a dedicated fuel rod thermomechanical computer code. This paper provides an outline of the methodology for the analysis of the 2A LB-LOCA accident in Atucha-2 NPP and describes the procedure adopted for the use of the fuel rod thermomechanical code. The methodology implies the application of best estimate thermalhydraulics, neutron physics, and fuel pin performance computer codes, with the objective to verify the compliance with the specific acceptance criteria. The fuel pin performance code is applied with the main objective to evaluate the extent of cladding failures during the transient. The procedure consists of a deterministic calculation by the fuel performance code of each individual fuel rod during its lifetime and in the subsequent LB-LOCA transient calculations. The boundary and initial conditions are provided by core physics and three-dimensional neutron kinetic coupled thermal-hydraulic system codes calculations. The procedure is completed by the sensitivity calculations and the application of the probabilistic method, which are outside the scope of the current paper.

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27

Mathis, Alexander, Andreas V. M. Herz, and Martin Stemmler. "Optimal Population Codes for Space: Grid Cells Outperform Place Cells." Neural Computation 24, no. 9 (September 2012): 2280–317. http://dx.doi.org/10.1162/neco_a_00319.

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Rodents use two distinct neuronal coordinate systems to estimate their position: place fields in the hippocampus and grid fields in the entorhinal cortex. Whereas place cells spike at only one particular spatial location, grid cells fire at multiple sites that correspond to the points of an imaginary hexagonal lattice. We study how to best construct place and grid codes, taking the probabilistic nature of neural spiking into account. Which spatial encoding properties of individual neurons confer the highest resolution when decoding the animal's position from the neuronal population response? A priori, estimating a spatial position from a grid code could be ambiguous, as regular periodic lattices possess translational symmetry. The solution to this problem requires lattices for grid cells with different spacings; the spatial resolution crucially depends on choosing the right ratios of these spacings across the population. We compute the expected error in estimating the position in both the asymptotic limit, using Fisher information, and for low spike counts, using maximum likelihood estimation. Achieving high spatial resolution and covering a large range of space in a grid code leads to a trade-off: the best grid code for spatial resolution is built of nested modules with different spatial periods, one inside the other, whereas maximizing the spatial range requires distinct spatial periods that are pairwisely incommensurate. Optimizing the spatial resolution predicts two grid cell properties that have been experimentally observed. First, short lattice spacings should outnumber long lattice spacings. Second, the grid code should be self-similar across different lattice spacings, so that the grid field always covers a fixed fraction of the lattice period. If these conditions are satisfied and the spatial “tuning curves” for each neuron span the same range of firing rates, then the resolution of the grid code easily exceeds that of the best possible place code with the same number of neurons.

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28

DUONG, Thanh Tung, and Yuichi OTSUKA. "Development of Statistical Safety Analysis for Best-Estimate Code for Loss of Coolant Accident Using Relap5/Mod 3.3." Proceedings of Conference of Hokuriku-Shinetsu Branch 2020.57 (2020): D013. http://dx.doi.org/10.1299/jsmehs.2020.57.d013.

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29

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 understanding of the plant response and further corrective measures. The abnormal event with reactor trip and safety injection signal actuation was simulated with the latest RELAP5/MOD3.3 Patch 05 best-estimate thermal hydraulic computer code. The measured and simulated data agree well considering the major plant system responses and operator actions. This suggests that the RELAP5 code simulation is good representative of the plant response and can complement not available information from plant measured data. In such a way, an event can be better understood.

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30

Skrzypek, M., E. Skrzypek, M. Stempniewicz, and J. Malesa. "Study on the DLOFC accident of the GEMINI+ conceptual design of HTGR reactor with MELCOR and SPECTRA." Journal of Physics: Conference Series 2048, no. 1 (October 1, 2021): 012043. http://dx.doi.org/10.1088/1742-6596/2048/1/012043.

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Abstract The work presented in this paper was performed within the Euratom Horizon 2020 GEMINI Plus project. Behavior of the HTGR reactor under severe accident conditions was investigated and the maximum fuel temperature was observed. Due to application of the TRISO particles and SiC layers in the fuel element, no damage of the fuel is expected up to 1600°C. Under the cooperation in the project between Nuclear Research Group (NRG) and National Centre for Nuclear Research (NCBJ) a code-to-code calculations were carried out between the SPECTRA and MELCOR codes. SPECTRA code, developed by the NRG is a thermal hydraulic analysis code and MELCOR 2.1.6342 used by NCBJ developed by SANDIA National Laboratory is fast running severe accident code. Both codes have already HTGR specific models build in. The following accident was analyzed and will be presented: Depressurized Loss of Forced Circulation (DLOFC) with 65 mm break at the top of reactor vessel. The scenario was calculated applying following sets of assumptions: best estimate and conservative. Plant behavior was analyzed including primary and secondary side of the reactor. As the results of applying conservative assumptions, it was found that fuel temperature excides the acceptable limit of 1620°C. Therefore, changes in the core design were proposed by project participants. Analyses of the new core showed acceptable temperatures. In the paper the results of code-to-code comparison are presented. Both codes have shown a good agreement of presented following characteristics on maximum fuel temperature, relative power and Reactor Cavity Cooling System power, primary pressure and break flow.

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31

Frepoli, Cesare. "An Overview of Westinghouse Realistic Large Break LOCA Evaluation Model." Science and Technology of Nuclear Installations 2008 (2008): 1–15. http://dx.doi.org/10.1155/2008/498737.

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Since the 1988 amendment of the 10 CFR 50.46 rule in 1988, Westinghouse has been developing and applying realistic or best-estimate methods to perform LOCA safety analyses. A realistic analysis requires the execution of various realistic LOCA transient simulations where the effect of both model and input uncertainties are ranged and propagated throughout the transients. The outcome is typically a range of results with associated probabilities. The thermal/hydraulic code is the engine of the methodology but a procedure is developed to assess the code and determine its biases and uncertainties. In addition, inputs to the simulation are also affected by uncertainty and these uncertainties are incorporated into the process. Several approaches have been proposed and applied in the industry in the framework of best-estimate methods. Most of the implementations, including Westinghouse, follow the Code Scaling, Applicability and Uncertainty (CSAU) methodology. Westinghouse methodology is based on the use of the WCOBRA/TRAC thermal-hydraulic code. The paper starts with an overview of the regulations and its interpretation in the context of realistic analysis. The CSAU roadmap is reviewed in the context of its implementation in the Westinghouse evaluation model. An overview of the code (WCOBRA/TRAC) and methodology is provided. Finally, the recent evolution to nonparametric statistics in the current edition of the W methodology is discussed. Sample results of a typical large break LOCA analysis for a PWR are provided.

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32

Perin, Yann, and Javier Jimenez Escalante. "Application of the best-estimate plus uncertainty approach on a BWR ATWS transient using the NURESIM European code platform." Nuclear Engineering and Design 321 (September 2017): 48–56. http://dx.doi.org/10.1016/j.nucengdes.2017.05.018.

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Shamir, Maoz. "The Scaling of Winner-Takes-All Accuracy with Population Size." Neural Computation 18, no. 11 (November 2006): 2719–29. http://dx.doi.org/10.1162/neco.2006.18.11.2719.

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Empirical studies seem to support conflicting hypotheses with regard to the nature of the neural code. While some studies highlight the role of a distributed population code, others emphasize the possibility of a “single-best-cell” readout. One particularly interesting example of single-best-cell readout is provided by the winner-takes-all (WTA) approach. According to the WTA, every cell is characterized by one particular preferred stimulus, to which it responds maximally. The WTA estimate for the stimulus is defined as the preferred stimulus of the cell with the strongest response. From a theoretical point of view, not much is known about the efficiency of single-best-cell readout mechanisms, in contrast to the considerable existing theoretical knowledge on the efficiency of distributed population codes. In this work, we provide a basic theoretical framework for investigating single-best-cell readout mechanisms. We study the accuracy of the WTA readout. In particular, we are interested in how the WTA accuracy scales with the number of cells in the population. Using this framework, we show that for large neuronal populations, the WTA accuracy is dominated by the tail of the single-cell-response distribution. Furthermore, we find that although the WTA accuracy does improve when larger populations are considered, this improvement is extremely weak compared to other types of population codes. More precisely, we show that while the accuracy of a linear readout scales linearly with the population size, the accuracy of the WTA readout scales logarithmically with the number of cells in the population.

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Petruzzi, Alessandro, and Francesco D'Auria. "Approaches, Relevant Topics, and Internal Method for Uncertainty Evaluation in Predictions of Thermal-Hydraulic System Codes." Science and Technology of Nuclear Installations 2008 (2008): 1–17. http://dx.doi.org/10.1155/2008/325071.

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The evaluation of uncertainty constitutes the necessary supplement of best-estimate calculations performed to understand accident scenarios in water-cooled nuclear reactors. The needs come from the imperfection of computational tools, on the one side, and the interest in using such a tool to get more precise evaluation of safety margins. The paper reviews the salient features of three independent approaches for estimating uncertainties associated with predictions of complex system codes. Namely, the propagations of code input error and calculation output error constitute the keywords for identifying the methods of current interest for industrial applications, while the adjoint sensitivity-analysis procedure and the global adjoint sensitivity-analysis procedure, extended to performing uncertainty evaluation in conjunction with concepts from data adjustment and assimilation, constitute the innovative approach. Throughout the developed methods, uncertainty bands can be derived (both upper and lower) for any desired quantity of the transient of interest. For one case, the uncertainty method is coupled with the thermal-hydraulic code to get the code with capability of internal assessment of uncertainty, whose features are discussed in more detail.

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35

Khedr, Ahmed, Martina Adorni, and Francesco d’Auria. "The effect of code user and boundary conditions on RELAP calculations of MTR research reactor transient scenarios." Nuclear Technology and Radiation Protection 20, no. 1 (2005): 16–22. http://dx.doi.org/10.2298/ntrp0501016k.

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The safety evaluation of nuclear power and re search reactors is a very important step before their construction and during their operation. This evaluation based on the best estimate calculations requires qualified codes qualified users, and qualified nodalizations. The effect of code users on the RELAP5 results during the analysis of loss of flow transient in MTR research reactors is presented in this pa per. To clarify this effect, two nodalizations for research reactor different in the simulation of the open water surface boundary conditions of the reactor pool have been used. Very different results are obtained with few choices for code users. The core natural circulation flow with the be ginning of core boiling doesn't stop but in creases. The in creasing in the natural circulation flow shifts out the boiling from the core and the clad temperature decreases be low the local saturation temperature.

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36

Angelucci, Michela, Bruno Gonfiotti, Bradut-Eugen Ghidersa, Xue Zhou Jin, Mihaela Ionescu-Bujor, Sandro Paci, and Robert Stieglitz. "Post-Test Numerical Analysis of a Helium-Cooled Breeding Blanket First Wall under LOFA Conditions with the MELCOR Fusion Code." Applied Sciences 12, no. 1 (December 24, 2021): 187. http://dx.doi.org/10.3390/app12010187.

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The validation of numerical tools employed in the analysis of incidental transients in a fusion reactor is a topic of main concern. KIT is taking part in this task providing both experimental data and by performing numerical analysis in support of the main codes used for the safety analyses of the Helium Cooled Pebble Bed (HCPB) blanket concept. In recent years, an experimental campaign has been performed in the KIT-HELOKA facility to investigate the behavior of a First Wall Mock-Up (FWMU) under Loss Of Flow Accident (LOFA) conditions. The aim of the experimental campaign was twofold: to check the expected DEMO thermal-hydraulics conditions during normal and off-normal conditions and to provide robust data for code validation. The present work is part of these validation efforts, and it deals with the analysis of the LOFA experimental campaign with the system code MELCOR 1.8.6 for fusion. A best-estimate methodology has been used in support of this analysis to ease the distinction between user’s assumptions and code limitations. The numerical analyses are here described together with their goals, achievements, and lesson learnt.

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37

Hamad, Ahmed A., Mohammed Taih Gatte, and Laith Ali Abdul-Rahaim. "Efficient systematic turbo polar decoding based on optimized scaling factor and early termination mechanism." International Journal of Electrical and Computer Engineering (IJECE) 13, no. 1 (February 1, 2023): 629. http://dx.doi.org/10.11591/ijece.v13i1.pp629-637.

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In this paper, an efficient early termination (ET) mechanism for systematic turbo-polar code (STPC) based on optimal estimation of scaling factor (SF) is proposed. The gradient of the regression line which best fits the distance between a priori and extrinsic information is used to estimate the SF. The multiplication of the extrinsic information by the proposed SF presents effectiveness in resolving the correlation issue between intrinsic and extrinsic reliability information traded between the two typical parallel concatenated soft-cancellation (SCAN) decoders. It is shown that the SF has improved the conventional STPC by about 0.3 dB with an interleaver length of 64 bits, and about 1 dB over the systematic polar code (SPC) at a bit error rate (BER) of . A new scheme is proposed as a stopping criterion, which is mainly based on the estimated value of SF at the second component decoder and the decoded frozen bits for each decoding iteration. It is shown that the proposed ET results in halving the average number of iterations (ANI) without adding considerable complexity. Moreover, the modified codes present comparable results in terms of BER to the codes that utilize fix number of iterations.

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38

Włostowski, Mateusz, Paweł Domitr, and Piotr Darnowski. "A Sensitivity Study of Critical Flow Modeling with MELCOR 2.2 Code Based on the Marviken CFT-21 Experiment." Energies 14, no. 16 (August 13, 2021): 4985. http://dx.doi.org/10.3390/en14164985.

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The paper presents a study of the critical flow phenomena modeling with MELCOR 2.2.15254 severe accident computer code. The Marviken Critical Flow Test number 21 (CFT-21) experiment was selected as a representative critical flow-focused Separate Effect Test (SET). Various modeling aspects were investigated, including the nodalization, model setup, parameters, and sensitivity coefficients. A local-type sensitivity study was performed to analytically identify the significant parameters and assess their impact on the modeling. A dedicated regression-based approach, using standard deviation, was developed to find the best-fit MELCOR modeling parameters. The primary purpose of this work was to determine the appropriate approach to model critical flow with MELCOR 2.2, investigate the model performance, assess the influence of nodalization choices, identify significant sensitivity parameters, and prepare recommendations with an emphasis on best-estimate modeling. An additional outcome was the benchmark of the recent code version with the Marviken test.

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39

Prošek, Andrej. "RELAP5 Calculations of Bethsy 9.1b Test." Science and Technology of Nuclear Installations 2012 (2012): 1–11. http://dx.doi.org/10.1155/2012/238090.

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Recently, several advanced computational tools for simulating reactor system behavior during real and hypothetical transient scenarios were developed. The TRAC/RELAP Advanced Computational Engine (TRACE) is the latest in a series of advanced, best-estimate reactor system codes developed by the United States Nuclear Regulatory Commission (US NRC). Nevertheless, the RELAP5/MOD3.3 computer code will be maintained in the next years. The purpose of the present study was to assess how the accuracy of Bethsy 9.1b test calculation depends on the US NRC RELAP5 code version used. Bethsy 9.1b test (International Standard Problem no. 27) was 5.08 cm equivalent diameter cold leg break without high-pressure safety injection and with delayed ultimate procedure. Seven different RELAP5 code versions were used and as much as possible the same input model. The obtained results indicate that the results obtained by the oldest and latest RELAP5 versions are in general comparable for Bethsy 9.1b test. This is very important for the validity of the results, obtained in the past with older RELAP5 versions. Due to the fact that observation was restricted to Bethsy 9.1b posttest, with its own physical phenomena, this conclusion could be generalized only for scenarios having similar range of the considered Bethsy transient conditions.

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40

Jones, D. P., and J. E. Holliday. "Elastic-Plastic Analysis of the PVRC Burst Disk Tests With Comparison to the ASME Code Primary Stress Limits." Journal of Pressure Vessel Technology 122, no. 2 (March 7, 2000): 146–51. http://dx.doi.org/10.1115/1.556164.

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This paper provides a comparison between finite element analysis results and test data from the Pressure Vessel Research Council (PVRC) burst disk program. Testing sponsored by the PVRC over 20 yr ago was done by pressurizing circular flat disks made from three different materials until failure by bursting. The purpose of this reanalysis is to investigate the use of finite element analysis (FEA) to assess the primary stress limits of the ASME Boiler and Pressure Vessel Code (hereafter the Code), and to qualify the use of elastic-plastic (EP-FEA) for limit-load calculations. The three materials tested represent the range of strength and ductility found in modern pressure vessel construction and include a low-strength, high-ductility material, a medium-strength, medium-ductility material, and a high-strength, low-ductility, low-alloy material. Results of elastic and EP-FEA are compared to test data. Stresses from the elastic analyses are linearized for comparison of Code primary stress limits to test results. Elastic-plastic analyses are done using both best-estimate and elastic-perfectly plastic (EPP) stress-strain curves. Both large strain-large displacement (LSLD) and small strain-small displacement (SSSD) assumptions are used with the EP-FEA. Analysis results are compared to test results to evaluate the various analysis methods, models, and assumptions as applied to the bursting of thin disks. The test results show that low-strength, high-ductility materials have a higher burst capacity than do high-strength, low-ductility materials. Linearized elastic FEA stresses and ASME Code primary stress limits provide excessive margins to failure for the burst disks for all three materials. The results of these studies show that LSLD EP-FEA can provide a best-estimate analysis of the disks, but the accuracy depends on the material stress-strain curve. This work concludes that SSSD EPP analysis methods provide a robust and viable alternative to the current elastic linearization method of satisfying the primary stress limits of the Code. [S0094-9930(00)01602-4]

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Wong, C. M., and W. K. Tso. "Seismic loading for buildings with setbacks." Canadian Journal of Civil Engineering 21, no. 5 (October 1, 1994): 863–71. http://dx.doi.org/10.1139/l94-092.

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Dynamic analysis is in general accepted as the best method to obtain the seismic load distribution for buildings with a setback. However, most building codes require the base shear obtained by dynamic analysis to be calibrated by the static base shear obtained using the code's equivalent static load procedure. In obtaining the code static base shear, two issues often arise among the design professionals. First, it is unclear whether the code static base shear is applicable for buildings with setbacks because the period prescribed by the code to be used in the base shear formula is in general not pertinent to buildings with setbacks. Second, it is uncertain whether the higher mode period should be used in computing the base shear when the modal weight of a higher mode is larger than that of the fundamental mode — a case often encountered in designing buildings with setbacks. This paper is an attempt to resolve the above issues. For the first issue, modification factors were derived for adjusting the code period formula so that it can provide a more reasonable estimate for the period of a building with a setback. For the second issue, it was demonstrated in this paper that for cases where the modal weight of a higher mode is larger than that of the fundamental mode, using the higher mode period for base shear calculation will result in unnecessarily conservative design. Key words: earthquake, seismic, irregular buildings, setback, dynamic analysis.

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Antariksawan, Anhar Riza, Surip Widodo, and Hendro Tjahjono. "PARAMETRIC STUDY OF LOCA IN TRIGA-2000 USING RELAP5/SCDAP CODE." JURNAL TEKNOLOGI REAKTOR NUKLIR TRI DASA MEGA 19, no. 2 (May 26, 2017): 59. http://dx.doi.org/10.17146/tdm.2017.19.2.3279.

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A postulated loss of coolant accident (LOCA) shall be analyzed to assure the safety of a research reactor. The analysis of such accident could be performed using best estimate thermal-hydraulic codes, such as RELAP5. This study focuses on analysis of LOCA in TRIGA-2000 due to pipe and beam tube break. The objective is to understand the effect of break size and the actuating time of the emergency core cooling system (ECCS) on the accident consequences and to assess the safety of the reactor. The analysis is performed using RELAP/SCDAPSIM codes. Three different break size and actuating time were studied. The results confirmed that the larger break size, the faster coolant blow down. But, the siphon break holes could prevent the core from risk of dry out due to siphoning effect in case of pipe break. In case of beam tube rupture, the ECCS is able to delay the fuel temperature increased where the late actuation of the ECCS could delay longer. It could be concluded that the safety of the reactor is kept during LOCA throughout the duration time studied. However, to assure the integrity of the fuel for the long term, the cooling system after ECCS last should be considered. Keywords: safety analysis, LOCA, TRIGA, RELAP5 STUDI PARAMETRIK LOCA DI TRIGA-2000 MENGGUNAKAN RELAP5/SCDAP. Kecelakaan kehilangan air pendingin (LOCA) harus dianalisis untuk menjamin keselamatan suatu reaktor riset. Analisis LOCA dapat dilakukan menggunakan perhitungan best-estimate seperti RELAP5. Penelitian ini menekankan pada analisis LOCA di TRIGA-2000 akibat pecahnya pipa dan tabung berkas. Tujuan penelitian adalah memahami efek ukuran kebocoran dan waktu aktuasi sistem pendingin teras darurat (ECCS) pada sekuensi kejadian dan mengkaji keselamatan reaktor. Analisis dilakukan menggunakan program perhitungan RELAP/SCDAPSIM. Tiga ukuran kebocoran dan waktu aktuasi ECCS berbeda dipilih sebagai parameter dalam studi ini. Hasil perhitungan mengonfirmasi bahwa semakin besar ukuran kebocoran, semakin cepat pengosongan tangki reaktor. Lubang siphon breaker dapat mencegah air terkuras dalam hal kebocoran pada pipa. Sedang dalam hal kebocoran pada beam tube, ECCS mampu memperlambat kenaikan temperatur bahan bakar. Dari studi ini dapat disimpulkan bahwa keselamatan reaktor dapat terjaga pada kejadian LOCA, namun pendinginan jangka panjang perlu dipertimbangkan untuk menjaga integritas bahan bakar.Kata kunci: analisis keselamatan, LOCA, TRIGA, RELAP5

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Ui, Atsushi, and Takamasa Miyaji. "Peach Bottom 2 Turbine Trip Simulation Using TRAC-BF1/COS3D, a Best-Estimate Coupled 3-D Core and Thermal-Hydraulic Code System." Nuclear Science and Engineering 148, no. 2 (October 2004): 281–90. http://dx.doi.org/10.13182/nse04-a2458.

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44

Jin, Dong-Sik, Sook-Kwan Kim, Sang-Koo Han, Chul-Kyu Lim, Bong-Jin Ko, Yong-Ho Hong, Kwang-Il Ahn, et al. "Best-estimate severe accident and source term analysis for an ISLOCA scenario of a CANDU-6 plant using the MAAP-ISAAC code." Nuclear Engineering and Design 358 (March 2020): 110443. http://dx.doi.org/10.1016/j.nucengdes.2019.110443.

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45

Pošta, Borna, and Siniša Šadek. "Mathematical model of the NPP Krško PCFV system for the RELAP5 computer code." Journal of Energy - Energija 66, no. 1-4 (June 23, 2022): 226–40. http://dx.doi.org/10.37798/2017661-4106.

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Containment building is the final barrier for radioactive releases from a nuclear power plant (NPP). Preserving its integrity will minimize these releases even in a case of a severe accident with core degradation and melt discharge in the containment accompanied with the pressure and temperature increase. Installation of a venting system with ability to filter radioactive fission products is a preferred way to deal with the issue in present and future NPPs, especially after the Fukushima accident. Such system, called passive containment filtered venting system (PCFV), was installed in 2013 in the NPP Krško. Thermal hydraulic model of the PCFV system which included aerosol and iodine filters, associated pipings and valves was developed for the best-estimate computer code RELAP5. Main results and discussion are presented and compared with relevant plant documentation.

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46

Racca, Stefano, and Tomasz Kozlowski. "Trace Code Validation for BWR Spray Cooling Injection and CCFL Condition Based on GÖTA Facility Experiments." Science and Technology of Nuclear Installations 2012 (2012): 1–17. http://dx.doi.org/10.1155/2012/282987.

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Best estimate codes have been used in the past thirty years for the design, licensing, and safety of NPP. Nevertheless, large efforts are necessary for the qualification and the assessment of such codes. The aim of this work is to study the main phenomena involved in the emergency spray cooling injection in a Swedish-designed BWR. For this purpose, data from the Swedish separate effect test facility GÖTA have been simulated using TRACE version 5.0 Patch 2. Furthermore, uncertainty calculations have been performed with the propagation of input errors method, and the identification of the input parameters that mostly influence the peak cladding temperature has been performed.

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Tong, Michael T., Ian Halliwell, and Louis J. Ghosn. "A Computer Code for Gas Turbine Engine Weight and Disk Life Estimation." Journal of Engineering for Gas Turbines and Power 126, no. 2 (April 1, 2004): 265–70. http://dx.doi.org/10.1115/1.1691980.

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Reliable engine-weight estimation at the conceptual design stage is critical to the development of new aircraft engines. It helps to identify the best engine concept amongst several candidates. In this paper, the major enhancements to NASA’s engine-weight estimate computer code (WATE) are described. These enhancements include the incorporation of improved weight-calculation routines for the compressor and turbine disks using the finite difference technique. Furthermore, the stress distribution for various disk geometries was also incorporated, for a life-prediction module to calculate disk life. A material database, consisting of the material data of most of the commonly used aerospace materials, has also been incorporated into WATE. Collectively, these enhancements provide a more realistic and systematic way to calculate the engine weight. They also provide additional insight into the design tradeoff between engine life and engine weight. To demonstrate the new capabilities, the enhanced WATE code is used to perform an engine weight/life tradeoff assessment on a production aircraft engine.

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48

Martin, Robert P., and Larry D. O'Dell. "Development Considerations of AREVA NP Inc.'s Realistic LBLOCA Analysis Methodology." Science and Technology of Nuclear Installations 2008 (2008): 1–13. http://dx.doi.org/10.1155/2008/239718.

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The AREVA NP Inc. realistic large-break loss-of-coolant-accident (LOCA) analysis methodology references the 1988 amended 10 CFR 50.46 allowing best-estimate calculations of emergency core cooling system performance. This methodology conforms to the code scaling, applicability, and uncertainty (CSAU) methodology developed by the Technical Program Group for the United States Nuclear Regulatory Commission in the late 1980s. In addition, several practical considerations were revealed with the move to a production application. This paper describes the methodology development within the CSAU framework and utility objectives, lessons learned, and insight about current LOCA issues.

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Ebiwonjumi, Bamidele, Peng Zhang, and Deokjung Lee. "SENSITIVITY ANALYSIS OF PWR SPENT FUEL DUE TO MODELLING PARAMETER UNCERTAINTIES USING SURROGATE MODELS." EPJ Web of Conferences 247 (2021): 15009. http://dx.doi.org/10.1051/epjconf/202124715009.

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In the BEPU (Best Estimate Plus Uncertainty) framework, uncertainty quantification (UQ) is a requirement to improve confidence and reliability of code predictions. Over the years, a lot of works have been done to quantify uncertainties in code predictions of spent nuclear fuel (SNF) characteristics due to nuclear data uncertainties. The purpose of this study is to quantify the uncertainty in pressurized water reactor (PWR) fuel assembly radiation source terms (isotopic inventory, activity, decay heat, neutron and gamma source) due to uncertainties in modeling parameters. The deterministic code STREAM is used to predict the source terms of a typical PWR fuel assembly following realistic and detailed irradiation history. For the sensitivity analysis (SA) and UQ, surrogate models are developed based on polynomial chaos expansion (PCE) and variance-based global sensitivity indices (i.e., Sobol’ indices) are employed. The global SA identifies the less important uncertain parameters, showing that the number of uncertain input parameters can be reduced. The surrogate model offers a significantly reduced computational burden even with large number of samples required for the SA/UQ of the model response.

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Zhou, Lin, Qingsheng Zhao, Shukai Chi, Yanlong Li, Lanjun Liu, and Qianxiang Yu. "A fractional Fourier transform–based channel estimation algorithm in single-carrier direct sequence code division multiple access underwater acoustic communication system." International Journal of Distributed Sensor Networks 15, no. 1 (January 2019): 155014771982600. http://dx.doi.org/10.1177/1550147719826001.

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Due to the complexity and variability of the underwater acoustic channel, the communication signal is affected by multi-path, time delay, and Doppler frequency shift. Based on the advantageous characteristics of fractional Fourier transform on chirp signal processing, a fractional Fourier transform–based algorithm using combined linear frequency–modulated signal is proposed, which can estimate parameters of underwater acoustic channel and has a better performance than the existing methods. To distinguish multi-user in underwater acoustic communication system, a single-carrier direct sequence code division multiple access communication system combined with the fractional Fourier transform–based algorithm is proposed. Thus, a preliminary study on underwater multi-target identification is carried out. The simulation and experimental results show that the fractional Fourier transform–based algorithm is simple and effective, and the energy can be focused at the “best” fractional order, which can directly determine the multi-path number and complete the channel estimation. The proposed single-carrier direct sequence code division multiple access communication system has good performance on bit error rate when we use corresponding spreading code to distinguish multi-user.

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