An Assisted Flow Circulation Option for Integral Pressure Water Reactors

2006 ◽  
Author(s):  
Dario F. Delmastro ◽  
Luciano Patruno ◽  
Viviana Masson
Keyword(s):  
Pressure Vessel ◽  
Natural Circulation ◽  
Thermal Power ◽  
Primary Circuit ◽  
Forced Circulation ◽  
Water Reactor ◽  
Jet Pumps ◽  
Pressure Water ◽  
Flow Circulation ◽  
Water Reactors

In this paper the use of jet pumps for assissting the coolant circulation in a integral pressure water reactor is considered. The integral pressure water reactors are characterized by the presence of all high pressure and temperature components inside a single pressure vessel. In this way the core, steam generators, pressurizer and pumps are located inside the same pressure vessel. This kind of systems usually consider natural circulation for the low power modules and forced circulation for the high power ones. CAREM-25 is the prototype reactor of CAREM design. It is a 100 MW thermal power integral pressure water reactor that use natural circulation in the primary circuit. The possibility, advantages and constraints, of using jet pumps for assissting the primary flow circulation and increase the reactor power are analyzed.

Development of Jet Pumps for Boiling Water Reactor Recirculation Systems

1974 ◽  
Vol 96 (1) ◽  
pp. 7-12 ◽  
Author(s):  
A. A. Kudirka ◽  
D. M. Gluntz
Keyword(s):  
Pressure Vessel ◽  
First Generation ◽  
Boiling Water ◽  
Boiling Water Reactors ◽  
Water Reactor ◽  
Recirculation System ◽  
Jet Pumps ◽  
Water Reactors ◽  
And Performance ◽  
Reactor Pressure

The efficiency of jet pumps used in Boiling Water Reactors has been significantly improved through extensive development efforts. These jet pumps are used inside the reactor pressure vessel as part of the reactor recirculation system to circulate subcooled water at about 530 deg F. The development of the first-generation and improved, second-generation jet pumps is surveyed, and the economic advantages of using them in a reactor are discussed. The development, geometry, and performance of these two jet pumps are described, their performance is compared, and the significant increase in efficiency obtained, resulting in the highest value reported in literature, is illustrated.

Analysis and Research on Natural Circulation Capacity of HFETR

2016 ◽  
Author(s):  
Taozhong Xu ◽  
Caiyu Deng ◽  
Yuxin Xiang
Keyword(s):  
Pressure Vessel ◽  
Cold Water ◽  
Flow Direction ◽  
Natural Circulation ◽  
Reactor Power ◽  
Hot Water ◽  
Forced Circulation ◽  
The Core ◽  
Establishment Process ◽  
Residual Heat

Natural circulation is being used as an important circulation to remove reactor residual heat. In the core of High Flux Engineering Trial Reactor of China (HFETR), the coolant is driven by pumps normally and flows from upside to downside in the core. When HFETR is shut down or runs in low power, the natural circulation between the hot water in the core and the cold water in the reflector inside the pressure vessel is established to cool down the core. Since the natural circulation processed only in the pressure vessel, the accident pumps need to be turned on periodically to remove reactor residual heat. The inversion of flow direction in HFETR and internal natural circulation lead to a different natural circulation establishment process from traditional reactor in which coolant flows form down to top normally. In this paper the transition between the natural circulation and forced circulation is analyzed by RELAP5/MOD3 code. The results showed that the accident pump could be turned off in the power of 850kW; The time, at which the accident pump needs to be turned on to transit the natural circulation to forced circulation, is decided by the temperature of the water in top of pressure vessel, and a formula between temperature of the water in the top of pressure vessel and the reactor power was obtained. The research results have theoretical and practical value to the full use of the natural circulation ability, as well as the safety of the engineering reactors or similar test facilities.

The Development of Radiation Embrittlement Models for U.S. Power Reactor Pressure Vessel Steels

2006 ◽  
Author(s):  
J. A. Wang ◽  
N. S. V. Rao ◽  
S. Konduri
Keyword(s):  
Pressure Vessel ◽  
Prediction Models ◽  
Irradiation Temperature ◽  
Reactor Pressure Vessel ◽  
Boiling Water ◽  
Radiation Embrittlement ◽  
Pressurized Water ◽  
Water Reactor ◽  
Water Reactors ◽  
Reactor Pressure

The information fusion technique is used to develop radiation embrittlement prediction models for reactor pressure vessel (RPV) steels from U.S. power reactors, including boiling water reactors and pressurized water reactors. The Charpy transition temperature-shift data is used as the primary index of RPV radiation embrittlement in this study. Six parameters—Cu, Ni, P, neutron fluence, irradiation time, and irradiation temperature—are used in the embrittlement prediction models. The results indicate that this new embrittlement predictor achieved reductions of about 49.5% and 52% in the uncertainties for plate and weld data, respectively, for pressurized water reactor and boiling water reactor data, compared with the Nuclear Regulatory Commission Regulatory Guide 1.99, Rev. 2. The implications of dose-rate effect and irradiation temperature effects for the development of radiation embrittlement models are also discussed.

scholarly journals Generation-IV Multi-Application Small Light Water Reactor (MASLWR)

2002 ◽  
Author(s):  
S. Michael Modro ◽  
James Fisher ◽  
Kevan Weaver ◽  
Pierre Babka ◽  
Jose Reyes ◽  
...  
Keyword(s):  
Natural Circulation ◽  
Thermal Power ◽  
Heat Removal ◽  
Steam Pressure ◽  
Light Water Reactor ◽  
Primary System ◽  
Light Water ◽  
Water Reactor ◽  
System Pressure ◽  
Environmental Laboratory

The Idaho National Engineering and Environmental Laboratory (INEEL), Nexant Inc. and the Oregon State University (OSU) have developed a Multi-Application Small Light Water Reactor (MASLWR) concept. The MASLWR is a small, safe and economic natural circulation pressurized light water reactor. MASLWR reactor module consists of an integral reactor/steam generator located in a steel cylindrical containment. The entire module is to be entirely shop fabricated and transported to site on most railways or roads. Two or more modules are located in a reactor building, each being submersed in a common, below grade cavity filled with water. For the most severe postulated accident, the volume of water in the cavity provides a passive ultimate heat sink for 3 or more days allowing the restoration of lost normal active heat removal systems. MASLWR thermal power of a single module is 150 MWt, primary system pressure 10.5 MPa, steam pressure 1.52 MPa and the net electrical output is 35–50 MWe.

BWR Steam Dryer Issues and Lessons Learned Related to Flow Induced Vibration

2013 ◽  
Author(s):  
Chakrapani Basavaraju ◽  
Kamal A. Manoly ◽  
Martin C. Murphy ◽  
William T. Jessup
Keyword(s):  
Pressure Vessel ◽  
Nuclear Power ◽  
Structural Integrity ◽  
Thermal Power ◽  
Lessons Learned ◽  
Flow Induced Vibration ◽  
Safety Function ◽  
Water Reactors ◽  
American Society ◽  
Loose Parts

Steam dryers in Boiling Water Reactors, located in the upper steam dome of the reactor pressure vessel, are not pressure retaining components and are not designed and constructed to the provisions of Section III of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code. As such, these components do not correspond to any specific safety class referenced in the Code. Although the steam dryers in BWRs perform no safety function, they must maintain the structural integrity in order to avoid the generation of loose parts that may adversely impact the capability of other plant equipment to perform their safety functions. Therefore guidance from Section III of the ASME Code is utilized in the design and fabrication of replacement dryers as well as for design modifications of the existing dryers for extended power uprates. The majority of licensees of operating nuclear plants are applying for EPU, which generally increases the thermal power output to 20% above the original licensed thermal power. Nuclear power plant components such as steam dryers can be subjected to strong fluctuating loads and can experience unexpected high cycle fatigue due to adverse flow effects while operating at EPU conditions. However, there are some unique challenges related to steam dryer operation at EPU conditions requiring special considerations to prevent fatigue damage from the effects of flow induced vibration. This paper examines the issues and lessons learned related to FIV considerations during EPU reviews of BWR steam dryers.

FIV Considerations for BWR Steam Dryers for EPU Operation

2011 ◽  
Author(s):  
Chakrapani Basavaraju ◽  
Kamal A. Manoly ◽  
Meena Khanna
Keyword(s):  
Pressure Vessel ◽  
Nuclear Power ◽  
Structural Integrity ◽  
Thermal Power ◽  
Operating Experience ◽  
Flow Induced Vibration ◽  
Safety Function ◽  
Water Reactors ◽  
American Society ◽  
Loose Parts

Steam dryers in Boiling Water Reactors (BWRs), located in the upper steam dome of the reactor pressure vessel, are not pressure retaining components and are not designed and constructed to American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code Section III safety class. Although the steam dryers in BWRs perform no safety function, they must maintain the structural integrity in order to avoid the generation of loose parts that may adversely impact the capability of other plant equipment to perform their safety functions. The majority of licensees of many operating nuclear plants are applying for extended power uprate (EPU), which increases the thermal power output up to 20% above the original licensed thermal power (OLTP). Nuclear power plant components such as steam dryers can be subjected to strong fluctuating loads and can experience unexpected high cycle fatigue due to adverse flow effects while operating at EPU conditions. However, there are some unique challenges for operation of dryers at EPU conditions requiring special considerations to prevent fatigue damage from the effects of flow induced vibration (FIV). This paper examines the FIV considerations and margin recommendations for fatigue stresses due to many uncertainties in the prediction of the fluctuating pressure loading acting on the steam dryer and the structural analysis of the steam dryer, very limited bench marking of the methods, and operating experience with some dryer failures.

Paper 7: Typical Heat Transfer Phenomena in a Natural Circulation Supercritical Pressure Water System

1967 ◽  
Vol 182 (9) ◽  
pp. 42-51 ◽  
Author(s):  
C. Merlini
Keyword(s):  
Heat Transfer ◽  
Natural Circulation ◽  
Pressure Fluctuations ◽  
Water System ◽  
Liquid Region ◽  
Forced Circulation ◽  
Compressed Liquid ◽  
Pressure Water ◽  
Flow Oscillations ◽  
Prandtl Numbers

This paper presents a study of the natural circulation heat transfer to water flowing under supercritical pressures. Flow oscillations occurred when crossing the critical state, both from the compressed liquid region and from the supercritical vapour region. The threshold was found to correspond to the outlet film temperature, being close to the transposed critical value. Differential pressure fluctuations with frequencies of 25–60 Hz were encountered. These vibrations are generated in the heated section and propagate in the system when resonance occurs. The heat transfer data, when compared with available forced circulation correlations, showed the need for a new correlation having smaller exponents of Reynolds and Prandtl numbers. This was attributed to the power-dependence of flow rate.

Study on Design of Scaled Down Test Facilities for Investigation of Instabilities in Supercritical Water Reactor

2013 ◽  
Author(s):  
Daya Shankar ◽  
Dipankar N. Basu ◽  
Manmohan Pandey
Keyword(s):  
Supercritical Water ◽  
Natural Circulation ◽  
Plant Design ◽  
Power Requirement ◽  
Water Reactor ◽  
Reference Design ◽  
Water Reactors ◽  
R 134A ◽  
The Cost ◽  
Supercritical Water Reactor

Supercritical Water Reactor (SCWR) proposes higher thermal efficiency and simpler plant design compared to modern Boiling Water Reactors. High pressure, temperature and power requirement in SCWR, however, escalates the cost of an experimental facility significantly. Present work, therefore, focuses on designing downscaled test facilities for stability analysis of SCWR. The facilities are conceptualized to model the European reference design of SCWR under both forced and natural circulation condition. R-134a is identified as the scaling fluids through fluid-to-fluid modeling, along with two others from literature. Similarity variables are obtained following two different approaches, starting from fundamental conservation equations. Dimensional and non-dimensional representations of important geometric, kinematic and dynamic parameters are evaluated and compared. Comparisons between two different approaches, as well as between forced and natural circulation have been presented for each scaling fluid.

Natural Circulation Characteristics of an Integral Pressurized Water Reactor

2006 ◽  
Author(s):  
Junli Gou ◽  
Suizheng Qiu ◽  
Guanghui Su ◽  
Dounan Jia
Keyword(s):  
Steady State ◽  
Steam Generator ◽  
Natural Circulation ◽  
Heat Removal ◽  
Primary Circuit ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Water Reactor ◽  
Secondary Side ◽  
Circulation Characteristics

Natural circulation potential is of great importance to the inherent safety of a nuclear reactor. This paper presents a theoretical investigation on the natural circulation characteristics of an integrated pressurized water reactor. Through numerically solved the one-dimensional model, the steady-state single phase conservative equations for the primary circuit and the steady-state two-phase drift-flux conservative equations for the secondary side of the once-through steam generator, the natural circulation characteristics are studied. Based on the preliminary calculation analysis, it is found that natural circulation mass flow rate is proportional to the exponential function of the power, and the value of the exponent is related to working conditions of the steam generator secondary side. The higher height difference between the core center and the steam generator center is favorable to the heat removal capacity of the natural circulation.

Sign in / Sign up

Export Citation Format

Share Document