OS8-4 Reactor core thermal characterization evaluation for the supercritical water reactor using subchannel analysis

2006 ◽  
Vol 2006.11 (0) ◽  
pp. 207-210
Author(s):  
Kazuaki Kitou ◽  
Yoshihiko Ishii ◽  
Masayoshi Matsuura ◽  
Shungo Sakurai
Keyword(s):  
Supercritical Water ◽  
Reactor Core ◽  
Thermal Characterization ◽  
Water Reactor ◽  
Subchannel Analysis ◽  
Supercritical Water Reactor

scholarly journals Analysis of numerical studies on the thermal-hydraulic and neutronic thermal-hydraulic stability of supercritical water reactors

2021 ◽  
Vol 7 (4) ◽  
pp. 311-318
Author(s):  
Artavazd M. Sujyan ◽  
Viktor I. Deev ◽  
Vladimir S. Kharitonov
Keyword(s):  
Supercritical Water ◽  
Nuclear Power ◽  
Power Plants ◽  
Negative Impact ◽  
Reactor Core ◽  
Resistance Coefficient ◽  
Coolant Flow ◽  
Water Reactor ◽  
Hydraulic Stability ◽  
Supercritical Water Reactor

The paper presents a review of modern studies on the potential types of coolant flow instabilities in the supercritical water reactor core. These instabilities have a negative impact on the operational safety of nuclear power plants. Despite the impressive number of computational works devoted to this topic, there still remain unresolved problems. The main disadvantages of the models are associated with the use of one simulated channel instead of a system of two or more parallel channels, the lack consideration for neutronic feedbacks, and the problem of choosing the design ratios for the heat transfer coefficient and hydraulic resistance coefficient under conditions of supercritical water flow. For this reason, it was decided to conduct an analysis that will make it possible to highlight the indicated problems and, on their basis, to formulate general requirements for a model of a nuclear reactor with a light-water supercritical pressure coolant. Consideration is also given to the features of the coolant flow stability in the supercritical water reactor core. In conclusion, the authors note the importance of further computational work using complex models of neutronic thermal-hydraulic stability built on the basis of modern achievements in the field of neutron physics and thermal physics.

General Corrosion Properties of Modified PNC1520 Austenitic Stainless Steel in Supercritical Water as a Fuel Cladding Candidate Material for Supercritical Water Reactor

2010 ◽  
Vol 73 ◽  
pp. 72-77
Author(s):  
Yoshihisa Nakazono ◽  
Takeo Iwai ◽  
Hiroaki Abe
Keyword(s):  
Stainless Steel ◽  
Austenitic Stainless Steel ◽  
Supercritical Water ◽  
Reactor Core ◽  
General Corrosion ◽  
Fuel Cladding ◽  
Corrosion Properties ◽  
Energy Agency ◽  
Water Reactor ◽  
Supercritical Water Reactor

The Super-Critical Water-cooled Reactor (SCWR) has been designed and investigated because of its high thermal efficiency and plant simplification. As the operating temperature of supercritical water reactor will be between 280°C and 620°C with a pressure of 25MPa, the selection of materials is difficult and important. The PNC1520 austenitic stainless steel developed by Japan Atomic Energy Agency (JAEA) as a nuclear fuel cladding material for a Na-cooled fast breeder reactor. The corrosion data of PNC1520 in supercritical water (SCW) is required but does not exist. The purpose of the present study is to research the corrosion properties for PNC1520 austenitic stainless steel in supercritical water. The supercritical water corrosion test was performed for the standard PNC1520 (1520S), the Ti-additional type of PNC1520 (1520Ti) and the Zr-additional type of PNC1520 (1520Zr) by using a supercritical water autoclave. In view of general corrosion, 1520Zr may have larger possibility than 1520S and 1520Ti to adopt a supercritical water reactor core fuel cladding.

scholarly journals Desain Inti Reaktor SCWR (Supercritical Water Reactor) Model Teras Silinder (r, z) dengan Bahan Bakar Thorium Hasil Daur Ulang

2019 ◽  
Vol 7 (2) ◽  
pp. 215-222
Author(s):  
Setiyaningsih Setiyaningsih ◽  
◽  
Yanti Yulianti ◽  
Simon Sembiring ◽  
◽  
...  
Keyword(s):  
Power Density ◽  
Supercritical Water ◽  
Power Factor ◽  
Critical Condition ◽  
Peak Power ◽  
Reactor Core ◽  
Multiplication Factor ◽  
Reactor Model ◽  
Water Reactor ◽  
Supercritical Water Reactor

The Research of the supercritical water reactor (SCWR) core design of the cylindrical core model (r, z) using the SRAC program has been done. The SRAC basic code was PIJ and CITATION. PIJ was used to calculate the fuel level and CITATION was used to calculate the reactor core level. The calculation of the reactor core has been done on the 1/4 cylinder core (r, z) and the geometry of the fuel cell was the cylindrical cell. Reactor fuel material was thorium burned 40 GWd/t and 30 GWd/t. The neutron parameters in this research were fuel enrichment, burn up, reactor core size, reactor core configurations, multiplication factor, and power density distribution. Multiplication factor (k-effective) in this research was 1.000004, which is reactor was in a critical condition. The reactor core in critical condition had the size of radius (r) was 130 cm, height (z) was 270 cm and fuel enrichment 2.8262%. The maximum power density was 130.0808 Watts /cm3 which was located at a radius of 25 cm and 135 cm high. The peak power factor in the radial direction was 1.6063 and the peak power factor in the axial direction was 1.3189.

Stability analysis of a square rod bundle sub-channel in supercritical water reactor

2013 ◽  
Author(s):  
Wang Hai-jun ◽  
You Ting ◽  
Zhang Lei ◽  
Gu Hong-fang ◽  
Luo Yu-shan ◽  
...  
Keyword(s):  
Stability Analysis ◽  
Supercritical Water ◽  
Water Reactor ◽  
Rod Bundle ◽  
Supercritical Water Reactor

Stability Analysis of a Supercritical Water Natural Circulation Loop

2012 ◽  
Author(s):  
ZhongChun Li ◽  
JiYang Yu ◽  
XiaoMing Song
Keyword(s):  
Supercritical Water ◽  
Natural Circulation ◽  
Basic Research ◽  
Circulation Loop ◽  
Two Phase ◽  
Water Reactor ◽  
Natural Circulation Loop ◽  
The Stability ◽  
Engineering Physics ◽  
Supercritical Water Reactor

As a part of “supercritical water reactor basic research”, the stability of the natural circulation research plays an important role on the feasibility of supercritical water reactor and experiment research. In this paper, the stability of a supercritical water natural circulation loop built by Department of Engineering Physics, Tsinghua University was studied by numerical method. It was confirmed that the static or Ledinegg instability doesn’t occur in HACA system, and there are no instabilities existing when the inlet enthalpy is larger than critical enthalpy. Instability was observed by numerical way, which is similar to DWOs and PDOs in two phase natural circulation loop. The system parameters’ influence on the instability of supercritical natural circulation loop was studied.

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