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Development of a thermal-hydraulic analysis code for thermal sizing of once-through steam generators for SMR

PROGRESS IN NUCLEAR ENERGY 2021年 139卷

作者： Park, Youngjae Kim, Iljin Kim, Hyungdae Kang, Kyungjoon Kyung Hee Univ Dept Nucl Engn Yongin South Korea Korea Atom Energy Res Inst Daejeon South Korea

A one-dimensional thermal-hydraulic analysis code was developed for the conceptual design and sizing of oncethrough steam generators (OTSGs) with application to small modular nuclear reactors. The OTSG in the model is represented by one characteristic double-tube heat exchanger with countercurrent flow. The secondary side of the OTSG is divided into subcooled liquid region, saturated two-phase mixture region, and superheated vapor region. Heat transfer and pressure drop correlations for each flow region were selected and implemented. To benchmark predictability of the developed computer code, the B&W OTSG whose design data is available in open literature was simulated. The results were compared both to the design data of the OTSG and to the analysis results obtained using a best-estimate thermal-hydraulic analysis code. The overall characteristics of the heat transfer area, pressure, and temperature distributions calculated with the developed code were in good agreement with the design data as well as the results from the best-estimate code. Lastly, a new concept of OTSG that incorporates feedwater circulation to reduce its volume for SMR application was simulated using the developed code. This demonstrates that the code can be used for design innovation and performance analysis of OTSGs.

关键词： computer code Once-through steam generator SMR Thermal-hydraulics Thermal sizing

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CSPACE for a simulation of core damage progression during severe accidents

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NUCLEAR ENGINEERING AND TECHNOLOGY 2021年第12期53卷 3990-4002页

作者： Song, JinHo Son, Dong-Gun Bae, JunHo Bae, Sung Won Ha, KwangSoon Chung, Bub-Dong Choi, YuJung Korea Atom Energy Res Inst 989-111 Daeduk Daero Daejon 34057 South Korea FNC Technol Co Ltd Heungdeok IT Valley Bldg 32F13Heungdeok 1 Ro Yongin 16954 Gyeonggi Do South Korea Korea Hydro & Nucl Power Co Ltd Cent Res Inst 70Yuseong Daero 1312 Gil Daejon 34101 South Korea

CSPACE (Core meltdown, Safety and Performance Analysis code for nuclear power plants) for a simulation of severe accident progression in a Pressurized Water Reactor (PWR) is developed by coupling of verified system thermal hydraulic code of SPACE (Safety and Performance Analysis code for nuclear power plants) and core damage progression code of COMPASS (Core Meltdown Progression Accident Simulation Software). SPACE is responsible for the description of fluid state in nuclear system nodes, while COMPASS is responsible for the prediction of thermal and mechanical responses of core fuels and reactor vessel heat structures. New heat transfer models to each phase of the fluid, flow blockage, corium behavior in the lower head are added to COMPASS. Then, an interface module for the data transfer between two codes was developed to enable coupling. An implicit coupling scheme of wall heat transfer was applied to prevent fluid temperature oscillation. To validate the performance of newly developed code CSPACE, we analyzed typical severe accident scenarios for OPR1000 (Optimized Power Reactor 1000), which were initiated from large break loss of coolant accident, small break loss of coolant accident, and station black out accident. The results including thermal hydraulic behavior of RCS, core damage progression, hydrogen generation, corium behavior in the lower head, reactor vessel failure were reasonable and consistent. We demonstrate that CSPACE provides a good platform for the prediction of severe accident progression by detailed review of analysis results and a qualitative comparison with the results of previous MELCOR analysis. (c) 2021 Korean Nuclear Society, Published by Elsevier Korea LLC. This is an open access article under the CC BY-NC-ND license (http://***/licenses/by-nc-nd/4.0/).

关键词： Severe accident computer code CSPACE SBO Core damage progression Coupling

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SPECTRA: a synchrotron radiation calculation code

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JOURNAL OF SYNCHROTRON RADIATION 2001年第6期8卷 1221-1228页

作者： Tanaka, T Kitamura, H Inst Phys & Chem Res Coherent Synchrotron Light Source Phys Lab Mikazuki Hyogo 6795148 Japan

An application program to calculate various characteristics of synchrotron radiation, called SPECTRA, is described. The program does not need any other commercial software and is equipped with a full graphical user interface which makes data input quite easy. Equations on synchrotron radiation from arbitrary-field sources in a near-field region are derived, as are simplified expressions for ideal devices using a far-field approximation. Effective numerical methods implemented in SPECTRA to reduce computation time are explained, and several examples are presented.

关键词： computer code synchrotron radiation sources numerical methods

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Development and validation of the MAVR-TA code for analyzing the release and transport of fission products during a severe accident at a nuclear power plant with VVER. Part 1-Release of fission products from a fuel

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NUCLEAR ENGINEERING AND DESIGN 2021年 385卷

作者： Shmelkov, Yu. B. Zvonarev, Yu. A. Shutov, N. V. Petrov, L. V. NRC Kurchatov Inst 1 Akad Kurchatova Pl Moscow 123182 Russia

This paper provides information on the development and validation of the MAVR-TA code. The MAVR-TA code is designed to simulate the fission products release from fuel and melt, and the transfer of fission products in the primary circuit and in the containment during a severe accident at NPP with VVER. The first part of the work provides information on the models of the MAVR-TA code, associated with modeling of the fission products release from the fuel. The paper provides information about description of the models and their approaches. In addition, it provides a description of the experimental tests, which were used to validate the models of the fission products release in the MAVR-TA code. As a result, the applicability of the models for calculating the fission products release from UO2 and MOX fuel has been demonstrated.

关键词： Validation Severe accident Fission products computer code Fuel MOX fuel

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How the code analyst can influence the calculation and use good practices to minimize pitfalls

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NUCLEAR TECHNOLOGY 1997年第1期117卷 40-48页

作者： Shotkin, LM U.S. Nuclear Regulatory Commission Mail Stop T-10E37 Washington DC 20555 United States Harvard University United States U.S. Nucl. Regulatory Commn. Off. R. United States

Almost 20 yr ago, the U.S. Nuclear Regulatory Commission (NRC) thermal-hydraulic code development effort made the transition from a homogeneous equilibrium formulation to a two-fluid formulation. The objective was to introduce a more physically based model so that the code analyst would have to make fewer choices in the input deck to simulate expected phenomena. There were still several options left open to the user, especially the noding for the simulation. Recent experience with NRC analyses, as well as with International Standard Problems, has shown that there can still be a considerable ''user effect'' in the use of even the two-fluid codes. Two specific examples are given. Using the RELAP5 code as a specific prototype to focus the discussion, examples are given of the choices currently available to the analyst. Similar choices are available in almost all thermal-hydraulic system codes. These example choices serve to show that even though thermal-hydraulic system codes are reaching a certain state of maturity, the user must still make many choices in setting up an input deck or in running a calculation. There are several pitfalls that the user can encounter, and there are good practices that can avoid many of these pitfalls. Specific examples of current practices for minimizing pitfalls and increasing good practices are discussed. They apply to any thermal-hydraulic system code. These include the following: training and mentoring for the code analyst;user guidelines documentation;internal review and quality assurance of the input deck by knowledgeable individuals;use of standard noding;and fixed noding for each test facility and reactor system. Guidance for the average code analyst is provided in terms of common pitfalls typically encountered and suggestions for good practices in choosing input deck options.

关键词： code analyst code user computer code

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SAGE MD: molecular-dynamic software package to study properties of materials with different models for interatomic interactions

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COMPUTATIONAL MATERIALS SCIENCE 2003年第2期28卷 107-124页

作者： Selezenev, AA Aleynikov, AY Gantchuk, NS Yermakov, PV Labanowski, JK Korkin, AA Sarov Open Comp Ctr Sarov 607189 Russia Ohio Supercomp Ctr Columbus OH 43212 USA

The paper describes molecular-dynamic modules and the graphical user interface of the computer code-SAGE MD. Algorithms for numerical solution of equations of motion and approaches used in representing interatomic interaction potentials are described and discussed. A few representative cases were chosen to illustrate application of the code to the derivation of properties of materials with different types of chemical bonding. (C) 2003 Elsevier B.V. All rights reserved.

关键词： computer code molecular dynamics method interatomic potentials molecular simulation

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Development of a Genetic Algorithm for the search of Optical Model parameters

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NUCLEAR INSTRUMENTS & METHODS IN PHYSICS RESEARCH SECTION B-BEAM INTERACTIONS WITH MATERIALS AND ATOMS 2011年第24期269卷 2984-2989页

作者： Abriola, D. IAEA Nucl Data Sect A-1400 Vienna Austria

The analysis of elastic scattering cross sections in terms of the Optical Model is subject to a series of well known ambiguities. Diverse assumptions about the initial values or shape of the potentials frequently produce different parameters, leading to different physical interpretations of the observed data. It would be important to have a starting set of "user independent" optical potentials that fit the experimental data to allow the evaluator to consider a large array of possibilities before committing to a particular optical potential. This work presents a Genetic Algorithm (GA) code that simulates natural selection and evolution, allowing a "blind search" of the multiparametric chi(2) surface. In this GA, the genes subject to evolution are the parameters of the optical potential. The GA variables, operators and procedures are described, and the GA is applied to two cases in which the elastic scattering cross section is adjusted: one for the Li-7 + Al-27 system at energies close to the Coulomb barrier where the interaction occurs near the nuclear surface, and another for the O-16 + O-16 system where the two nuclei deeply interpenetrate each other. Further developments are described. (C) 2011 Elsevier B.V. All rights reserved.

关键词： Optical Model computer code Differential cross section Evaluation Genetic Algorithm Elastic scattering

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Bruce Schneier on AI Security (Interview)

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computer 2024年第11期57卷 99-101页

作者： Berghe, Hal Univ Nevada Comp Sci Las Vegas NV 89154 USA

In this interview, Bruce Schneier reflects on the security challenges of artificial intelligence.

关键词： Social Media Rule Based Political System European Union Business Model Modern Medicine computer code Social Media Sites Fake News Market System Disinformation Security Vulnerabilities Artificial Intelligence Systems Absence Of Values Trust In Systems Pufferfish Tax Avoidance Operant Chambers Bullshit Thing Of The Past

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Development of time dependent safety analysis code for plasma anomaly events in fusion reactors

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JOURNAL OF NUCLEAR SCIENCE AND TECHNOLOGY 1997年第3期34卷 229-239页

作者： Honda, T Bartels, HW Uckan, NA Seki, Y Okazaki, T ITER JOINT CENT TEAM SAN DIEGO CA USA OAK RIDGE NATL LAB OAK RIDGE TN 37831 USA JAPAN ATOM ENERGY RES INST NAKA FUS RES ESTAB NAKA IBARAKI 31101 JAPAN

A safety analysis code SAFALY has been developed to analyze plasma anomaly events in fusion reactors, e.g., a loss of plasma control. The code is a hybrid code comprising a zero-dimensional plasma dynamics and a one-dimensional thermal analysis of in-vessel components. The code evaluates the time evolution of plasma parameters and temperature distributions of in-vessel components. As the plasma-safety interface model, we proposed a robust plasma physics model taking into account updated data for safety assessment. For example, physics safety guidelines for beta limit, density limit and H-L mode confinement transition threshold power, etc. are provided in the model. The model of the in-vessel components are divided into twenty temperature regions in the poloidal direction taking account of radiative heat transfer between each surface of each region. This code can also describe the coolant behavior under hydraulic accidents with the results by hydraulics code and treat vaporization (sublimation) from plasma facing components (PFCs). Furthermore, the code includes the model of impurity transport form PFCs by using a transport probability and a time delay. Quantitative analysis based on the model is possible for a scenario of plasma passive shutdown. We examined the possibility of the code as a safety analysis code for plasma anomaly events in fusion reactors and had a prospect that it would contribute to the safety analysis of the International Thermonuclear Experimental Reactor (ITER).

关键词： ITER plasma anomaly events safety analysis computer code plasma dynamics thermal analysis beta limit density limit confinement transition in-vessel components

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Analysis of dust and fission products in a pebble bed NGNP

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NUCLEAR ENGINEERING AND DESIGN 2012年 251卷 433-442页

作者： Stempniewicz, M. M. Winters, L. Caspersson, S. A. NRG Arnhem NL-6800 ES Arnhem Netherlands Westinghouse Elect Co Windsor CT USA

In the HTR pebble bed reactor graphite dust is generated during normal reactor operation due to pebble-to-pebble interactions. This dust will be deposited throughout the primary system. Furthermore, the dust will become radioactive due to sorption of fission products released during normal operation. This paper presents an analysis of dust transport and deposition and fission product transport and plate-out during normal operation of a NGNP (next generation nuclear plant) Westinghouse pebble bed design. The main objective is to determine the amount and location of deposited graphite dust in the system and the amount of radioactive isotopes adsorbed on the structures and the dust during normal operation. The results will be used in planning of maintenance activities. Moreover the present results may be used in a next step to perform a depressurized loss of forced cooling (D-LOFC) analysis and to determine the amount of radioactivity released to the atmosphere during the accident. The analysis was performed using the SPECTRA code. Based on the assumed dust source term, during the 60 years lifetime of the pebble bed reactor concept which was analyzed approximately 1630 kg dust enters the primary helium flow. It was determined: (1) That 86% of the dust settles on the graphite structures inside the reactor vessel. Of the remaining graphite dust 2/3 collects in the low temperature intermediate heat exchanger (IHX) and the remaining dust in the high temperature IHX and the connecting pipes of the primary system. (2) Agglomeration of dust from smaller into larger particles is observed at all locations. The resuspension of agglomerated dust particles is an important phenomenon that limits the build-up of a dust layer on the surface of the IHXs. A large deposited dust layer is observed on the control rod drive (CRD) walls in the graphite structures. The dust accumulates on the CRD walls because, due to very low gas velocities, resuspension of the agglomerated particles does n

关键词： CIP core inlet pipe CRD control rod drives CV control volume FHSS fuel handling and storage system Flownex computer code HC helium circulator HTR high temperature reactor IHX intermediate heat exchanger JN junction D-LOFC depressurized loss of forced circulation’ LPC low pressure compressor MELCOR computer code developed at Sandia NGNP next generation nuclear plant NHSS nuclear heat supply system PATRAS computer code created in Jülich PBMR pebble bed modular reactor PCHX process coupling heat exchanger PHTS primary heat transport system RADAX computer code created in Jülich, developed further RT radioactive particle transport package in SPECTRA RUS reactor unit system SC 1-D solid heat conductor SG steam generator SHTS secondary heat transport system SPATRA computer code created in Jülich SPECTRA computer code developed at NRG WEC Westinghouse Electric Company, LLC

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