An Experimental Investigation of High Level Vibration on the Residual Heat Removal Line of a Pressurized Water Reactor

2009 ◽  
Author(s):  
Jeffrey A. Brown ◽  
Robert D. Blevins ◽  
H. Joseph Fernando
Keyword(s):  
Heat Removal ◽  
Coupled System ◽  
Acoustic Resonance ◽  
Acoustic Mode ◽  
Side Branch ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Water Reactor ◽  
High Level ◽  
Residual Heat

This paper presents the results of a scaled aero acoustic test that modeled a side branch resonance observed in the residual heat removal suction line of a large pressurized water reactor. Resolution of the acoustic resonance was sought by detuning the eddy shedding frequency from the fundamental side branch acoustic mode. The specific physical modifications and their ability to detune the coupled system are presented.

Analysis of PRHRS for an Integral Pressurized Water Reactor Using RELAP5/MOD3.4

2010 ◽  
Author(s):  
Xuhua Ye ◽  
Minjun Peng ◽  
Jiange Liu
Keyword(s):  
Natural Circulation ◽  
Heat Removal ◽  
Parameter Study ◽  
Pressurized Water Reactor ◽  
Steam Generators ◽  
Primary Coolant ◽  
Pressurized Water ◽  
Water Reactor ◽  
Pressure Peak ◽  
Residual Heat

An investigation on the thermal hydraulic characteristics of the passive residual heat removal system (PRHRS) which is used in an integral pressurized water reactor (INSURE-100) is presented in this paper. The main components of primary coolant system are enclosed in reactor vessel. Primary fluid flow circle is natural circulation. The PRHRS can remove the energy from the primary side as long as the residual heat exchanger (RHE) is submerged in the emergency cooldown tank (ECT). The parameter study is performed by considering the effects of an effective height between the steam generators and the RHE and a valve actuation time, which are useful for the design of the PRHRS. The mass flow in the PRHRS has been affected by the height difference between the steam generators and the RHE. The pressure peak of the primary side and PRHRS has been affected by the valve action time.

Loss of Feed Water Accident in an Integral Pressurized Water Reactor (IPWR)

2010 ◽  
Vol 171-172 ◽  
pp. 379-384
Author(s):  
Khan Salah Ud Din ◽  
Min Jun Peng ◽  
Muhammad Zubair
Keyword(s):  
Heat Removal ◽  
Feed Water ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Water Reactor ◽  
Heat Removal System ◽  
Water Accident ◽  
Relap5 Code ◽  
Removal System ◽  
Residual Heat

In this paper research has been carried out on Loss of Feed Water Accident (LOFW) scenario of the Integral Pressurized Water Reactor ( IPWR) under two circumstances by the use of thermal hydraulic system code i.e Relap5/Mod3.4. In the first one, Passive Residual Heat Removal System (PRHRS) which is designed to absorb core residual heat in case of transient conditions is included which has the function of operating under the accident vulnerabilities. Concerning with the second case i.e without the use of PRHRS rather a tank of water which has the capacity of about 8% of the total feed water supply and is operated under accident scenario is considered. Taken into account these conditions,first the nodalization diagram of the two cases have been figured out then according to the LOFW accident time event scenario use the Relap5 code to simulate the accident. Finally the graphical explanation (separately) of the two cases with graphical approach as well as the conclusion is given at the end.

Evaluation of a SGTR Accident in the Multi-Application Integrated Pressurized Water Reactor

2010 ◽  
Author(s):  
Liguo Jiang ◽  
Minjun Peng ◽  
Jiange Liu
Keyword(s):  
Steam Generator ◽  
Flow Model ◽  
Heat Removal ◽  
Injection System ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Water Reactor ◽  
Heat Removal System ◽  
Removal System ◽  
Residual Heat

One of more frequent events in the Pressurized Water Reactor (PWR) is Steam Generator Tube Rupture (SGTR) accident, which is among the main accidents in the field of nuclear safety. This paper studies the SGTR event in the Multi-application Integrated Pressurized Water Reactor (IPWR) using the best-estimate thermal-hydraulic code RELAP5/MOD3.4. In the reactor of IPWR, several Once-Through Steam Generator (OTSG) cassettes are used and located between the core support and the pressure vessel. The tube rupture location is on the top of the tube sheet of a steam generator. Three different tube rupture modeling methods and several different subcooled discharge coefficients in the critical flow model are considered and compared. In the safety analysis, high pressure safety injection system, core makeup system and Passive Residual Heat Removal System (PRHRS) that would affect the accident consequences are considered.

scholarly journals Thermal Hydraulic Analysis of a Passive Residual Heat Removal System for an Integral Pressurized Water Reactor

2009 ◽  
Vol 2009 ◽  
pp. 1-12 ◽  
Author(s):  
Junli Gou ◽  
Suizheng Qiu ◽  
Guanghui Su ◽  
Douna Jia
Keyword(s):  
Heat Exchanger ◽  
Steam Generator ◽  
Heat Removal ◽  
Heat Transfer Area ◽  
Pressurized Water Reactor ◽  
Height Difference ◽  
Pressurized Water ◽  
Heat Removal System ◽  
Removal System ◽  
Residual Heat

A theoretical investigation on the thermal hydraulic characteristics of a new type of passive residual heat removal system (PRHRS), which is connected to the reactor coolant system via the secondary side of the steam generator, for an integral pressurized water reactor is presented in this paper. Three-interknited natural circulation loops are adopted by this PRHRS to remove the residual heat of the reactor core after a reactor trip. Based on the one-dimensional model and a simulation code (SCPRHRS), the transient behaviors of the PRHRS as well as the effects of the height difference between the steam generator and the heat exchanger and the heat transfer area of the heat exchanger are studied in detail. Through the calculation analysis, it is found that the calculated parameter variation trends are reasonable. The higher height difference between the steam generator and the residual heat exchanger and the larger heat transfer area of the residual heat exchanger are favorable to the passive residual heat removal system.

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.

Heat Removal Tests for Pressurized Water Reactor Containment Spray by Large-Scale Facility

1983 ◽  
Vol 63 (2) ◽  
pp. 316-329 ◽  
Author(s):  
Yasuo Motoki ◽  
Mitsuo Naritomi ◽  
Mitsugu Tanaka ◽  
Gunji Nishio ◽  
Kazuichiro Hashimoto ◽  
...  
Keyword(s):  
Large Scale ◽  
Heat Removal ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Water Reactor

Questions Relative to CHF Testing for New Pressurized Water Reactor Advanced Fuel Assemblies

2010 ◽  
Author(s):  
Christophe Herer
Keyword(s):  
Heat Transfer ◽  
Heat Flux ◽  
Heat Removal ◽  
Nucleate Boiling ◽  
Operating Conditions ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Water Reactor ◽  
Fuel Assemblies ◽  
Advanced Fuel

One of the limiting conditions during operation of a Pressurized Water Reactor is cladding integrity in case of occurrence of any conditions I or II events. The decoupling criterion is the absence of Departure from Nucleate Boiling (DNB) during the full sequence of any of these transients. Heat transfer between the clad and the water is limited by the DNB phenomenon when local surface heat flux is greater than the so-called Critical Heat Flux (CHF). Heat production at the surface is higher than heat removal capacity by the coolant therefore a vapor blanket is formed around the clad; consequently the heat transfer will drastically drop resulting in a sudden significant increase of the local wall temperature and clad damage may appear if no corrective action is initiated. DNB can not be estimated with physical principles only. Experimental support is needed for evaluation. Occurrence of DNB is evaluated using the Departure from Nucleate Boiling Ratio (DNBR) which is a function of both core thermal hydraulic (T/H) parameters and design of the fuel assembly. Advanced fuel assemblies claim higher CHF values compared to previous designs. Along with increased DNB performances for advanced fuel assemblies, CHF correlation development and advanced methodologies enable to extend normal operating conditions of a nuclear plant. On the one hand, CHF performances really increased allow additional margin related to the loss of fuel cladding integrity whereas on the other hand optimized correlations and advanced methodologies reduce this margin. An accurate assessment of the CHF performance of the advanced fuel assemblies is therefore required. This paper will raise issues regarding the assessment of the CHF performance of new advanced fuel assemblies design. The issues will be focused on the reliability of the experimental assessment of the CHF values and the accuracy of the transposition of mock up geometries to plant core configuration (representativity of the experiments). The verification that the tests conditions (pressure, flowrate, quality, heat flux …) ensure a proper coverage of all core conditions encountered during any of the conditions I & II transients is closely linked to DNBR methods and will not be extensively covered in this paper. This paper suggests some thoughts about relevance of the demonstration carried out by vendors on these matters.

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