检索结果-内蒙古大学图书馆

Verification of kinetic capabilities of the GNOMER neutron diffusion code

ANNALS OF NUCLEAR ENERGY 2017年 109卷 431-439页

作者： Merljak, Vid Kromar, Marjan Trkov, Andrej Jozef Stefan Inst Jamova Cesta 39 SI-1000 Ljubljana Slovenia Vienna Int Ctr Int Atom Energy Agcy POB 100 A-1400 Vienna Austria

Recently, the GNOMER neutron diffusion code was updated to support kinetic simulations. This direct approach to reactor kinetics is further branched into two operational modes: the so-called Lambda mode and the (normal) implicit mode. Both approaches are described in detail. Also, calculation of point kinetic parameters is implemented and discussed. To verify GNOMER's new kinetic capabilities, two numerical tests are considered;namely, the point kinetic test case and a rod insertion simulation on a model of a typical pressurised water reactor (PWR). The reference solutions were produced by a point kinetic script and the SKETCH-N diffusion code, respectively. Very good agreement is found for all cases considered. Unusual behaviour and discrepancies - where observed - are explained satisfactorily. With the evidence provided in the paper, GNOMER is deemed to be a reliable tool for (both static and) dynamic simulation. (C) 2017 Elsevier Ltd. All rights reserved.

关键词： Reactor kinetics 3D simulation computer code Verification Diffusion theory

来源：评论

学校读者我要写书评

暂无评论

Improvements in the particle and heavy-ion transport code system (PHITS) for simulating neutron-response functions and detection efficiencies of a liquid organic scintillator

引用

JOURNAL OF NUCLEAR SCIENCE AND TECHNOLOGY 2022年第8期59卷 1047-1060页

作者： Satoh, Daiki Sato, Tatsuhiko Japan Atom Energy Agcy Nucl Sci & Engn Ctr Ibaraki Japan

In this study, we simulated the neutron-response functions and detection efficiencies of a liquid organic scintillator using the particle and heavy-ion transport code system (PHITS). We incorporated the algorithm and database of the neutron-response simulation code SCINFUL-QMD into PHITS. Then, we updated the total, elastic, and inelastic cross-section data of the hydrogen and carbon nuclei for neutrons and developed a new scorer to analyze the light outputs from a scintillator. The calculation results of the neutron-response functions and the detection efficiencies were compared with results of SCINFUL-QMD, the previous PHITS with the new scorer, and the reported measurements. It was found that the improved PHITS successfully reproduced the results calculated by SCINFUL-QMD, except for around 150 MeV where a discontinuity in the detection-efficiency curve was observed in the SCINFUL-QMD values. Our results showed better agreement with the measured data than the results of the previous PHITS. The uncertainties of the detection efficiencies calculated by PHITS using the present extensions were estimated to be approximately 15% for neutrons in the energy region below 100 MeV.

关键词： Carbon computer code cross section detection efficiency light output Monte Carlo calculation neutron organic scintillator PHITS response function

来源：评论

学校读者我要写书评

暂无评论

Thermal-hydraulic analysis code development for sodium heated once-through steam generator

引用

ANNALS OF NUCLEAR ENERGY 2019年 127卷 385-394页

作者： Xu, Rongshuan Zhang, Dalin Tian, Wenxi Qiu, Suizheng Su, G. H. Xi An Jiao Tong Univ Sch Nucl Sci & Technol State Key Lab Multiphase Flow Power Engn Shaanxi Key Lab Adv Nucl Energy & Technol Xian 710049 Shaanxi Peoples R China

Sodium heated once-through steam generator (OTSG) is an essential component of Sodium cooled Fast Reactor (SFR). It transfers the heat from hot sodium to water and converts water to superheat steam. Steam can roll the turbine and generate electricity. Therefore, the safety of the steam generator is vital to the operation of the plant. China is developing the China Demonstration SFR (CDSFR) after successful construction of China Experiment Fast Reactor (CEFR). In order to make sure the safety of the steam generator under various transient conditions and support the design of the steam generator for CDSFR, a Transient analysis code of Once-through Steam generator for Sodium cooled fast reactor (TCOSS) has been developed. To benchmark the developed physical model, the calculated results of the code have been compared with some design data of OTSGs and they are found in a good agreement. In addition, the transient results of the code have been compared with the experimental results of a shutdown experiment for the validation of the code. The predictions of the code agree well with the experimental data. Hence, the code can be utilized to predict the variations of thermal-hydraulic parameters in steam generator under different transient conditions. In addition, predictions can support the design of the steam generator for CDSFR. (C) 2018 Published by Elsevier Ltd.

关键词： Once-through steam generator Sodium cooled fast reactor computer code Thermal-hydraulic model Transient analysis

来源：评论

学校读者我要写书评

暂无评论

SPECTRA: a synchrotron radiation calculation code

引用

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

来源：评论

学校读者我要写书评

暂无评论

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

引用

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

来源：评论

学校读者我要写书评

暂无评论

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

引用

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

来源：评论

学校读者我要写书评

暂无评论

Development of dynamics code and proposal of start-up procedure for accelerator driven system

引用

JOURNAL OF NUCLEAR SCIENCE AND TECHNOLOGY 2006年第1期43卷 20-31页

作者： Sugawara, T Iwasaki, T Chiba, T Nishihara, K Japan Atom Energy Agcy Tokai Ibaraki 3191195 Japan Tohoku Univ Aoba Ku Sendai Miyagi 9808579 Japan

A dynamics code for Accelerator Driven System (ADS) has been developed to analyze behaviors of a subcritical core coupled with an accelerator. This computer code named DSE (Dynamics calculation code system for Sub-critical system with External neutron source) is able to treat changes of the neutron source distribution produced by the spallation reactions, which have not been treatable by existing dynamics codes. The DSE code consists of the high energy part and the reactor part. For the high energy part, a database for neutron source distributions which were results of transport calculations in the high energy region was developed to treat the changes of beam parameters and target parameters. For the reactor part, the IBIS code which is a dynamics code for fast reactors was adapted for the core dynamics calculation. Some modifications were carried out to treat the dynamics calculation in a subcritical state with an external neutron source. The DSE was used to calculate various events from which new procedures for ADS operation control were investigated. From these calculation results, it was found that the change of the surface location of a heavy liquid metal target (target level) was useful to control the power from 0 to 100% with a gradual variation. A new procedure for the ADS start-up was proposed by using the change of the target level and beam intensity instead of the present one which is simply to start the accelerator only. The DSE and the proposed procedure will contribute to the research and development of the ADS.

关键词： accelerator driven system (ADS) dynamics code operation control DSE computer code

来源：评论

学校读者我要写书评

暂无评论

Assessment of TAPINS code for application to thermal-hydraulic analysis of an integral reactor, REX-10

引用

JOURNAL OF NUCLEAR SCIENCE AND TECHNOLOGY 2013年第9期50卷 924-941页

作者： Lee, Yeon-Gun Park, Goon-Cherl Jeju Natl Univ Inst Nucl Sci & Technol Jeju Si 690756 Jeju South Korea Seoul Natl Univ Dept Nucl Engn Seoul 151744 South Korea

The thermal-hydraulic analysis program for integral reactor system (TAPINS) is a thermal-hydraulic system code developed by Seoul National University for transient analysis of an integral reactor, REX-10. Specialized for a fully passive integral pressurized water reactor, TAPINS adopts a one-dimensional four-equation drift-flux model for two-phase flows. It also consists of component models for the core, the helical-coil steam generator, and the steam-gas pressurizer. This paper presents the developmental assessment of TAPINS to validate its applicability to the thermal-hydraulic analysis of REX-10. Assessment problems are determined by taking into account thermal-hydraulic phenomena expected during design basis accidents of REX-10, including the loss-of-feedwater accident and the small-break loss-of-coolant accident. To confirm the predictive capability of TAPINS for these phenomena, the TAPINS model is validated against four sets of separate effects problems, including the pressurizer insurge test, the subcooled boiling experiment, the critical flow test, and the Edwards pipe problem. In addition, the calculation results of TAPINS are compared with the experimental data obtained from a series of integral effects tests using a scaled apparatus of REX-10. From the validation results, it is demonstrated that TAPINS can provide the reasonable prediction on the thermal-hydraulic responses of REX-10 during the transient and accident conditions.

关键词： PWR-type reactor REX-10 thermal hydraulics computer code TAPINS validation integral effects tests

来源：评论

学校读者我要写书评

暂无评论

Development of 3D Finite Element code of Incompatible Displacement Mode for Flexural Analysis

引用

ARABIAN JOURNAL FOR SCIENCE AND ENGINEERING 2014年第8期39卷 6009-6016页

作者： Alwathaf, Ahmed Hasan Sanaa Univ Fac Engn Dept Civil Engn Sanaa Yemen

For three-dimensional finite element modeling, it is necessary to use a solid element taking into account the flexural response and also should be simple for computer storage. One of the elements that can attain the previous requirements is the incompatible (non-conforming) element. In this paper, formulation and algorithm of the developed 3D finite element code of incompatible displacement mode for flexural analysis are presented using 3D 8-noded isoparametric incompatible brick element. Several comparisons and a parametric study have been carried out to verify the incompatible brick element and the developed code. Accurate results have been obtained by the 3D incompatible element in flexural analysis of beams in comparison with the standard element.

关键词： 3D finite element Incompatible displacement mode computer code Algorithm Flexural analysis Stiffness

来源：评论

学校读者我要写书评

暂无评论

An Implicit Numerical Method for Integration of the Conservation Equations Incorporated into the KORSAR code Two-Fluid Model

引用

THERMAL ENGINEERING 2024年第4期71卷 301-308页

作者： Yudov, Yu. V. Danilov, I. G. Aleksandrov Res Technol Inst Sosnovyi Bor 188450 Russia

A noniterative implicit method for solving the discrete conservation equations of the KORSAR/GP computer code (hereinafter referred to as KORSAR) two-fluid model is presented. The KORSAR/GP code has been developed jointly by specialists of the Federal State Unitary Enterprise Aleksandrov Research Institute of Technology (NITI) and the special design bureau OKB Gidropress. In 2009, the code was certified at the Federal Service for Environmental, Technological, and Nuclear Supervision (Rostekhnadzor) as applied to numerical safety assessment of VVER-type power reactor plants. The code uses the semi-implicit numerical scheme, which limits the integration time step by the Courant condition with respect to the velocity of a two-phase flow. To cut down the time it takes to calculate prolonged transients in reactor plants, an implicit numerical method, which does not limit the time step by the Courant condition, has been developed on the basis of the SETS (stability-enhancing two-step) method. It is based on the semi-implicit scheme. Prior to its application, discrete phase momentum conservation equations with the convective terms written in implicit form are solved at each time step. After the semi-implicit step, the specific (per unit volume) mass and energy of the phases, which are donor quantities in the convective terms of the transport equations, are calculated at the new time layer. Unlike the SETS method, the implicit method developed for the KORSAR code employs a semi-implicit scheme with linearization of unsteady terms describing the change in the specific mass and energy of a two-phase flow. This approach enables us to solve discrete equations in a noniterative manner. However, the implementation of this procedure requires that the unknown scalar variables, such as the phase specific enthalpies, the vapor volume fraction, and the pressure, be determined in the computational cells. Therefore, the semi-implicit scheme with linearization of unsteady terms with reca

关键词： interfacial heat and mass transfer two-fluid model donor quantities semi-implicit scheme implicit method computer code Courant condition unsteady terms time layer

来源：评论

学校读者我要写书评

暂无评论

建议与咨询留下您的常用邮箱和电话号码，以便我们向您反馈解决方案和替代方法

时间限定

文献类型

馆藏选择

核心期刊

语言

文献类型

帮助

文字说明：

检索规则说明：

检索范例：

分类表

所选分类

限定检索结果

文献类型

馆藏范围

日期分布

学科分类号

主题

机构

作者

语言

请选择保存的检索档案：

请选择收藏分类：

通借通还

建议与咨询 留下您的常用邮箱和电话号码，以便我们向您反馈解决方案和替代方法

时间限定

文献类型

馆藏选择

核心期刊

语言

文献类型

帮助

文字说明：

检索规则说明：

检索范例：

分类表

所选分类

限定检索结果

文献类型

馆藏范围

日期分布

学科分类号

主题

机构

作者

语言

请选择保存的检索档案： 新增检索档案 确定 取消

请选择收藏分类： 新增自定义分类 确定 取消

通借通还

建议与咨询留下您的常用邮箱和电话号码，以便我们向您反馈解决方案和替代方法

请选择保存的检索档案：

请选择收藏分类：