Development of a thermal-hydraulic analysis code for primary system of an integrated small modular pressurized water reactor

被引：0

作者：

Wang, Dongqing ^{[1
]}

Chen, Huandong ^{[1
]}

Chen, Jiayue ^{[1
]}

Zhan, Dekui ^{[2
]}

Xia, Shaoxiong ^{[2
]}

机构：

[1] Sun Yat Sen Univ, Sino French Inst Nucl Engn & Technol, Zhuhai 519082, Guangdong, Peoples R China

[2] China Nucl Power Technol Res Inst Co Ltd, Reactor Engn & Safety Res Ctr, Shenzhen 518031, Peoples R China

来源：

ANNALS OF NUCLEAR ENERGY | 2021年 / 152卷

基金：

中国博士后科学基金; 国家重点研发计划;

关键词：

Thermal-hydraulic analysis code; Primary system; IRIS; Sub-channel model; Helically coiled steam generator model; STEAM-GENERATOR; TRANSIENT ANALYSIS; HEAT-TRANSFER; FLOW; MODEL; SIMULATION; DESIGN; TUBES; CORE; DROP;

D O I：

10.1016/j.anucene.2020.108019

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：

In this study, a thermal-hydraulic analysis code has been developed for primary system of the integrated small modular pressurized water reactor IRIS (International Reactor Innovative and Secure). Numerical model of the primary system has been established, and characteristics of equipment and layout are considered. The sub-channel model is used for performance analysis of hot channel in the core. In modeling of the helically coiled steam generator, effects of secondary flow inside coiled tubes are taken into account. Based on the model established, the simulation code THAS (Thermal-Hydraulic Analysis code of SMRs) has been developed. To benchmark this code, steady state performance of the system has been calculated and compared with design parameters as well as results obtained with RELAP5. Furthermore, transient performance of the system is also simulated and compared with RELAP5. The comparison presents a good agreement, which confirms simulation capacity and accuracy of the THAS code. (C) 2020 Elsevier Ltd. All rights reserved.

引用

页数：13

共 50 条

[1] PRESSURIZED WATER-REACTOR THERMAL-HYDRAULIC CORE ANALYSIS WITH THE FLICA COMPUTER CODE
RAYMOND, P
SPINDLER, B
LENAIN, R
[J]. NUCLEAR ENGINEERING AND DESIGN, 1990, 124 (03) : 299 - 313
[2] Development of a thermal-hydraulic analysis software for the Chinese advanced pressurized water reactor
Wu, Y. W.
Su, G. H.
Qiu, S. Z.
Zhuang, C. J.
[J]. NUCLEAR ENGINEERING AND DESIGN, 2010, 240 (01) : 112 - 122
[3] Development of a thermal-hydraulic analysis code for the Pebble Bed Water-cooled Reactor
Cai, Xiaoyu
Qiu, Suizheng
Tian, Wenxi
Su, Guanghui
[J]. NUCLEAR ENGINEERING AND DESIGN, 2011, 241 (12) : 4978 - 4988
[4] A PLANT-COMPUTER-BASED PRESSURIZED WATER-REACTOR PRIMARY SYSTEM THERMAL-HYDRAULIC MODEL
KAO, SP
MEYER, JE
[J]. TRANSACTIONS OF THE AMERICAN NUCLEAR SOCIETY, 1985, 50 (NOV): : 637 - 638
[5] SIMULATION RESEARCH ON THERMAL-HYDRAULIC PERFORMANCE OF A NATURAL CIRCULATION INTEGRATED PRESSURIZED WATER REACTOR
Zhao, Yanan
Peng, Minjun
Xia, Genglei
Lv, Lianxin
[J]. PROCEEDINGS OF THE 26TH INTERNATIONAL CONFERENCE ON NUCLEAR ENGINEERING, 2018, VOL 9, 2018,
[6] Development of Thermal-Hydraulic Analysis Code for Plate Type Fuel Reactor
Li Lei
Zhang Zhijian
[J]. 2010 ASIA-PACIFIC POWER AND ENERGY ENGINEERING CONFERENCE (APPEEC), 2010,
[7] Neutronics and thermal hydraulic coupling analysis of integrated pressurized water reactor
Khan, Salah Ud-Din
Peng, Minjun
Khan, Shahab Ud-Din
[J]. INTERNATIONAL JOURNAL OF ENERGY RESEARCH, 2013, 37 (13) : 1709 - 1717
[8] UKAP - A CODE FOR THERMAL-HYDRAULIC ANALYSIS OF A REACTOR CORE
HUHN, J
[J]. KERNENERGIE, 1989, 32 (05): : 193 - 198
[9] A RELAP5 MODEL FOR THE THERMAL-HYDRAULIC ANALYSIS OF A TYPICAL PRESSURIZED WATER REACTOR
Agamy, Said
Metwally, Adul Mohsen
Al-Ramady, Amir Mohammad
Elaraby, Sayed Mohamed
[J]. THERMAL SCIENCE, 2010, 14 (01): : 79 - 88
[10] Uncertainty analysis of neutronic/thermal-hydraulic coupling in pressurized water reactor fuel assemblies
Kong, Deyan
Gao, Zhibo
Wu, Di
Cheng, Jie
Wang, Jianjun
[J]. ANNALS OF NUCLEAR ENERGY, 2024, 207

← 1 2 3 4 5 →