DEVELOPMENT OF THE SPACE CODE FOR NUCLEAR POWER PLANTS

被引：91

作者：

Ha, Sang Jun ^{[1
]}

Park, Chan Eok ^{[2
]}

Kim, Kyung Doo ^{[3
]}

Ban, Chang Hwan ^{[4
]}

机构：

[1] Korea Elect Power Corp Res Inst, Taejon 305760, South Korea

[2] KEPCO Engn & Construct Co Inc, Taejon 305353, South Korea

[3] Korea Atom Energy Res Inst Daejon, Yusong, Korea Chang Hwa, South Korea

[4] Korea Nucl Fuel, Taejon 305353, South Korea

来源：

NUCLEAR ENGINEERING AND TECHNOLOGY | 2011年 / 43卷 / 01期

关键词：

System Code; SPACE; Nuclear Safety Analysis; Thermal-Hydraulics; Multi-Fields Model; Code Validation; FLOW;

D O I：

10.5516/NET.2011.43.1.045

中图分类号：

TL [原子能技术]; O571 [原子核物理学];

学科分类号：

0827 ; 082701 ;

摘要：

The Korean nuclear industry is developing a thermal-hydraulic analysis code for safety analysis of pressurized water reactors (PWRs). The new code is called the Safety and Performance Analysis Code for Nuclear Power Plants (SPACE). The SPACE code adopts advanced physical modeling of two-phase flows, mainly two-fluid three-field models which comprise gas, continuous liquid, and droplet fields and has the capability to simulate 3D effects by the use of structured and/or non-structured meshes. The programming language for the SPACE code is C++ for object-oriented code architecture. The SPACE code will replace outdated vendor supplied codes and will be used for the safety analysis of operating PWRs and the design of advanced reactors. This paper describes the overall features of the SPACE code and shows the code assessment results for several conceptual and separate effect test problems.

引用

页码：45 / 62

页数：18

共 27 条

[1]

[Anonymous], 1981, NUMERICAL HEAT TRANS

[2]

BANKOFF SG, 1985, NUCL SAFETY, V26, P139

[3]

Bennett A. W., 1976, AERER5373

[4]

Cathcart J., 1977, ORNLNUREG17

[5] A THEORETICAL BASIS FOR LOCKHART-MARTINELLI CORRELATION FOR 2-PHASE FLOW [J].

CHISHOLM, D .

INTERNATIONAL JOURNAL OF HEAT AND MASS TRANSFER, 1967, 10 (12) :1767-&

[6]

CHURCHILL SW, 1977, CHEM ENG-NEW YORK, V84, P91

[7]

CLAXTON KT, 1972, HTFSDR28AERER7162

[8]

EDWARDS AR, 1970, J BRIT NUCL ENER SOC, V9, P125

[9] 2-PHASE CRITICAL FLOW OF ONE-COMPONENT MIXTURES IN NOZZLES, ORIFICES, AND SHORT TUBES [J].

HENRY, RE ;

FAUSKE, HK .

JOURNAL OF HEAT TRANSFER, 1971, 93 (02) :179-&

[10]

Hirsch C., 1991, NUMERICAL COMPUTATIO

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