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.

Optimization study for thermal efficiency of supercritical water reactor nuclear power plant

2014 ◽  
Vol 63 ◽  
pp. 541-547 ◽  
Author(s):  
Yali Su ◽  
Khurrum Saleem Chaudri ◽  
Wenxi Tian ◽  
Guanghui Su ◽  
Suizheng Qiu
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Supercritical Water ◽  
Thermal Efficiency ◽  
Nuclear Power ◽  
Water Reactor ◽  
Optimization Study ◽  
Supercritical Water Reactor

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.

Flow and Temperature Fields Measurement Inside Rod Bundle by the Combined Use of PIV and LIF Technique

2018 ◽  
Author(s):  
Xing Li ◽  
Sichao Tan ◽  
Zhengpeng Mi ◽  
Peiyao Qi ◽  
Yunlong Huang
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Great Influence ◽  
Reactor Core ◽  
Transfer Characteristic ◽  
Index Of Refraction ◽  
Temperature Fields ◽  
Spacer Grid ◽  
Rod Bundle ◽  
Combined Use

Thermal hydraulic research of reactor core is important in nuclear energy applications, the flow and heat transfer characteristics of coolant in reactor fuel assembly has a great influence on the performance and safety of nuclear power plants. Particle image velocimetry (PIV) and Laser induced fluorescence (LIF) are the instantaneous, non-intrusive, whole-field fluid mechanics measuring method. In this study, the simultaneous measurement of flow field and temperature field for a rod bundle was conducted using PIV and LIF technique. A facility system, utilizing the matching index of refraction approach, has been designed and constructed for the measurement of velocity and temperature in the rod bundle. In order for further study on complex heat and mass transfer characteristic of rod bundle, the single-phase experiments on the heating conditions are performed. One of unique characteristics of the velocity and temperature distribution downstream the spacer grid was obtained. The experimental results show that the combined use of PIV and LIF technique is applied to the measurement of multi-physical field in a rod bundle is feasible, the measuring characteristics of non-intrusive ensured accuracy of whole field data. The whole field experimental data obtained in rod bundle benefits the design of spacer grid geometry.

Expectations for Inservice Testing Programs at New Nuclear Power Plants

2017 ◽  
Author(s):  
Thomas G. Scarbrough
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Reactor Core ◽  
Nuclear Regulatory Commission ◽  
Lessons Learned ◽  
Mechanical Equipment ◽  
American Society ◽  
Regulatory Commission ◽  
Safety Systems

In a series of Commission papers, the U.S. Nuclear Regulatory Commission (NRC) described its policy for inservice testing (IST) programs to be developed and implemented at nuclear power plants licensed under 10 CFR Part 52. This paper discusses the expectations for IST programs based on those Commission policy papers as applied in the NRC staff review of combined license (COL) applications for new reactors. For example, the design and qualification of pumps, valves, and dynamic restraints through implementation of American Society of Mechanical Engineers (ASME) Standard QME-1-2007, “Qualification of Active Mechanical Equipment Used in Nuclear Power Plants,” as accepted in NRC Regulatory Guide (RG) 1.100 (Revision 3), “Seismic Qualification of Electrical and Active Mechanical Equipment and Functional Qualification of Active Mechanical Equipment for Nuclear Power Plants,” will enable IST activities to assess the operational readiness of those components to perform their intended functions. ASME has updated the Operation and Maintenance of Nuclear Power Plants (OM Code) to improve the IST provisions for pumps, valves, and dynamic restraints that are incorporated by reference in the NRC regulations with applicable conditions. In addition, lessons learned from performance experience and testing of motor-operated valves (MOVs) will be implemented as part of the IST programs together with application of those lessons learned to other power-operated valves (POVs). Licensee programs for the Regulatory Treatment of Non-Safety Systems (RTNSS) will be implemented for components in active nonsafety-related systems that are the first line of defense in new reactors that rely on passive systems to provide reactor core and containment cooling in the event of a plant transient. This paper also discusses the overlapping testing provisions specified in ASME Standard QME-1-2007; plant-specific inspections, tests, analyses, and acceptance criteria; the applicable ASME OM Code as incorporated by reference in the NRC regulations; specific license conditions; and Initial Test Programs as described in the final safety analysis report and applicable RGs. Paper published with permission.

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