DEVELOPMENT AND APPLICATION OF THE NEUTRONICS/THERMAL-HYDRAULICS COUPLING CODE FOR SAFETY ANALYSIS OF EBR-II LOSS OF HEAT SINK TESTS WITHOUT SCRAM

被引：0

作者：

Lu, Daogang ^{[1
]}

Guo, Chao ^{[1
]}

Sui, Danting ^{[1
]}

机构：

[1] North China Elect Power Univ, Nucl Sci & Engn Sch, Beijing 102206, Peoples R China

来源：

PROCEEDINGS OF THE 24TH INTERNATIONAL CONFERENCE ON NUCLEAR ENGINEERING, 2016, VOL 4 | 2016年

关键词：

Neutronic-thermalhydraulic coupling; safety analysis; EBR-II; LOHSWS; LOSS-OF-FLOW; THERMAL-HYDRAULICS;

D O I：

暂无

中图分类号：

TH [机械、仪表工业];

学科分类号：

0802 ;

摘要：

In the GEN IV technology evaluations, the LMFBR (Liquid Metal Fast Breeder Reactor) system which includes SFR (Sodium-cooled Fast Reactor) and LFR (Lead-cooled Fast Reactor) was top-ranked in sustainability due to its closed fuel cycle and it is top-ranked in proliferation resistance and physical protection because it employs a long-life core. It is necessary to conduct the coupled neutronics and thermal-hydraulics simulation when the feedback effects are significant in the safety analysis of Anticipated Transients Without Scram (ATWS) in LMFBR. Thus, a neutronics-thermalhydraulics coupling code for safety analysis of LMFBR was developed and used to analyze whole-plant transient behavior of the Experimental Breeder Reactor II (EBR-II) under Loss of Heat Sink Without Scram (LOHSWS) tests in this paper. The two mixing zone method for cold pool coupled with SAC-CFR was used and the predicted results agree well with measurements which are taken from EBR-II LOHSWS test data.

引用

页数：9

共 30 条

[21] Analysis of loss of flow without scram test in the FFTF reactor - Part II: System thermal hydraulics with point neutron kinetics
Ponomarev, Alexander
Nikitin, Evgeny
Fridman, Emil
PROGRESS IN NUCLEAR ENERGY, 2024, 170
[22] Development of safety analysis methodology for moderator system failure of CANDU-6 reactor by thermal-hydraulics/physics coupling
Kim, Jong Hyun
Jin, Dong Sik
Chang, Soon Heung
NUCLEAR ENGINEERING AND DESIGN, 2013, 263 : 241 - 254
[23] Coupled 3D neutronics/thermal-hydraulics analysis of Superph′enix start-up tests with DYN3D/ATHLET code system
Ponomarev, Alexander
Nikitin, Evgeny
Fridman, Emil
NUCLEAR ENGINEERING AND DESIGN, 2023, 412
[24] Thermal-hydraulic analysis of EBR-II Shutdown Heat Removal Tests SHRT-17 and SHRT-45R
Yue, Nina
Ma, Zaiyong
Cal, Rong
Hu, Benxue
Su, G. H.
Qiu, Suizheng
PROGRESS IN NUCLEAR ENERGY, 2015, 85 : 682 - 693
[25] Development and application of a thermal-hydraulics analysis code for small modular natural circulation lead or lead-alloy cooled fast reactors
Chen, Zhao
Zhou, Guangming
Zhao, Pengcheng
Chen, Hongli
ANNALS OF NUCLEAR ENERGY, 2014, 73 : 68 - 74
[26] Development of neutronics and thermal hydraulics coupled code - SAC-RIT for plate type fuel and its application to reactivity initiated transient analysis
Singh, Tej
Kumar, Jainendra
Mazumdar, Tanay
Raina, V. K.
ANNALS OF NUCLEAR ENERGY, 2013, 62 : 61 - 80
[27] First-Order Comprehensive Adjoint Sensitivity Analysis Methodology for Critical Points in Coupled Nonlinear Systems. II: Application to a Nuclear Reactor Thermal-Hydraulics Safety Benchmark
Cacuci, Dan Gabriel
FLUIDS, 2021, 6 (01)
[28] Nanofluid application for heat transfer, safety, and natural circulation enhancement in the NuScale nuclear reactor as a small modular reactor using computational fluid dynamic (CFD) modeling via neutronic and thermal-hydraulics coupling
Rahnama, Z.
Ansarifar, G. R.
PROGRESS IN NUCLEAR ENERGY, 2021, 138
[29] Code development and assessment for reactor outage thermal-hydraulic and safety analysis-midloop operation with loss of residual heat removal
Liang, TKS
Ko, FK
NUCLEAR TECHNOLOGY, 2000, 129 (01) : 13 - 25
[30] Thermal-hydraulic modeling for deterministic safety analysis of portable equipment application in the VVER-1000 nuclear reactor during loss of ultimate heat sink accident using RELAP5/MOD3.2 code
Tabadar, Z.
Ansarifar, G. R.
Pirouzmand, A.
ANNALS OF NUCLEAR ENERGY, 2019, 127 : 53 - 67

← 1 2 3 →