Jet Pump Noise Analysis for BWRs

2004 ◽  
Author(s):  
Rogelio Castillo-Dura´n ◽  
Javier Ortiz-Villafuerte ◽  
He´ctor Herna´ndez-Lo´pez ◽  
Gustavo Alonso-Vargas ◽  
Gabriel Calleros-Micheland
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Pressure Drop ◽  
Nuclear Power ◽  
Noise Analysis ◽  
Jet Pump ◽  
Operation Conditions ◽  
Jet Pumps ◽  
Process Computer ◽  
Abnormal Operation

The use of noise analysis for detection of BWR component malfunction is a powerful tool in determining abnormal operation conditions, during the life of a nuclear power plant. Since the 80s, several nuclear reactors have reported problems related with jet pumps and recirculation loops. The NRC, in the IE Bulletin 80-07, recommended performing periodic monitoring to individual pressure drop jet pumps, to prevent structural failure. In this work, noise analysis methods are used for detection of jet pumps abnormal operation conditions in a BWR. Power signals obtained from the backup process computer of a BWR are analyzed with a home-developed software, called NOISE, for noise diagnostic of power signals. The computer program takes individual signals from the tabular report of the process computer. The normalized power spectral density (NPSD) is then obtained, using a Prime Factor Algorithm to calculate the Fast Fourier Transform. The NPSD of the jet pumps pressure drop, of Unit 2 of the Laguna Verde Nuclear Power Plant, showed a noticeable change in jet pump 6 during 2003, considering the period from the startup test to operation during 2003. This abnormal condition was due to that the jet pump throat was partially blocked. The noise analysis methodology is shown to be a useful tool for malfunction detection, and could be applied to create a data bank for monitoring the dynamic behavior of BWR jet pumps.

scholarly journals Sensitivity Evaluation of AP1000 Nuclear Power Plant Best Estimation Model

2017 ◽  
Vol 2017 ◽  
pp. 1-13 ◽  
Author(s):  
Hao Shi ◽  
Qi Cai ◽  
Yuqing Chen
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Pressure Drop ◽  
Temperature Variation ◽  
Steam Generator ◽  
Nuclear Power ◽  
Channel Number ◽  
Design Parameters ◽  
Coolant Temperature ◽  
Node Number

The best estimation process of AP1000 Nuclear Power Plant (NPP) requires proper selections of parameters and models so as to obtain the most accurate results compared with the actual design parameters. Therefore, it is necessary to identify and evaluate the influences of these parameters and modeling approaches quantitatively and qualitatively. Based on the best estimate thermal-hydraulic system code RELAP5/MOD3.2, sensitivity analysis has been performed on core partition methods, parameters, and model selections in AP1000 Nuclear Power Plant, like the core channel number, pressurizer node number, feedwater temperature, and so forth. The results show that core channel number, core channel node number, and the pressurizer node number have apparent influences on the coolant temperature variation and pressure drop through the reactor. The feedwater temperature is a sensitive factor to the Steam Generator (SG) outlet temperature and the Steam Generator outlet pressure. In addition, the cross-flow model nearly has no effects on the coolant temperature variation and pressure drop in the reactor, in both the steady state and the loss of power transient. Furthermore, some fittest parameters with which the most accurate results could be obtained have been put forward for the nuclear system simulation.

Hydraulic Characteristics Research on SG Under Tube Plugging Operations Using FLUENT

2018 ◽  
Author(s):  
Xiaohan Zhao ◽  
Mingjun Wang ◽  
Wenxi Tian ◽  
G. H. Su ◽  
Suizheng Qiu
Keyword(s):  
Heat Transfer ◽  
Power Plant ◽  
Nuclear Power Plant ◽  
Pressure Drop ◽  
Nuclear Power ◽  
Reactor System ◽  
Economic Losses ◽  
Maximum Pressure ◽  
Hydraulic Characteristics ◽  
Tube Plugging

Steam Generator (SG) is a critical equipment in the nuclear power plant, it is the huge heat exchanger in reactor system which can achieve removing fission energy from the reactor system effectively to ensure safety of the whole nuclear system. It is located between the primary and the secondary loop in reactor system act as the intermediate hub of energy and the security barrier in nuclear power plant. Generally, there are numerous of U-shaped heat transfer tubes in SG it is one of the weakest structures throughout the primary loop system. So the integrity of the SG especially its heat transfer tubes is important to the safety of reactor operation. The degradation problem of heat transfer tubes together with ruptures accidents often occur under suffer environments in reactors, which include thermal stress, mechanical stress and so on, it is noteworthy that this kind of accidents is inevitable due to the limited properties of existing materials. The performance of the SG is seriously affected by the number of failure tubes. Plugging operations through various mechanical means is the most common method to solve the tubes ruptures problems which can reduce the economic losses to the utmost extent. However, plugging operations will make huge impact on the thermal hydraulic performances of both sides of SG. It’s meaningful to research the characteristics of the plugging affects under different operations. In this paper the hydraulic characteristics of primary side in AP1000 SG under a certain fraction of heat transfer tube plugging conditions is researched. Three dimensional hydraulic characteristics of primary side coolant in SG under different plugging conditions are obtained by using the thermal hydraulic software FLUENT. The typical plugging fraction in this simulation model is 10 percent, and the effect of plugging locations also be considered through changing the plugging positions using the zone marking method. The results shows that the pressure drop under the structure integrated SG is 358.01MPa which is accordance with the results from Westinghouse 343KPa. The pressure drop values varies when changing positions of the plugging tubes under the same plugging fraction condition. The flow fields in bottom head also change meanwhile and the maximum pressure drop can reach up to 388.05KPa when the plugging fraction is 10%. The growth rate become significant when tube plugging fraction larger than 5%, and differences between maximum and minimum values of total pressure drop under different plugging positions become larger gradually. Finally the local resistance coefficients and flow field distributions of primary side in SG under various plugging conditions are obtained which is meaningful for the reactor safety and it can be a good reference for the maintenance of SG.

scholarly journals Transient Study on the HTR-PM with TINTE-vPower Coupling Code Package

2020 ◽  
Vol 2020 ◽  
pp. 1-14
Author(s):  
Jun Sun ◽  
Ximing Sun ◽  
Yanhua Zheng
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Water Flow ◽  
Nuclear Power ◽  
Data Exchange ◽  
Reactor Power ◽  
Normal Operation ◽  
Primary Circuit ◽  
Fuel Temperature ◽  
Operation Conditions

The high-temperature gas-cooled reactor pebble-bed module (HTR-PM) nuclear power plant consists of two nuclear steam supply system modules, each of which drives the steam turbine by the superheated steam flow and is fed by the heated-up water flow. The shared steam/water system induces mutual effects on normal operation conditions and transients of the nuclear power plant, which is worthy of safety concerns and intensive study. In this paper, a coupling code package was developed with the TINTE and vPower codes to understand how the HTR-PM operated. The TINTE code was used to analyze the reactor core and primary circuit, while the vPower code simulated the steam/water flow in the conventional island. Two TINTE models were built and coupled to one vPower model through the data exchange in the steam generator models. Using this code package, two typical transients were simulated by decreasing the primary flow rate or introducing the negative reactivity of one module. Important parameters, including the reactor power, the fuel temperature, and the reactor inlet and outlet helium temperatures of two modules, had been studied. The calculation results preliminarily proved that this code package can be further used to evaluate working performance of the HTR-PM.

Implementation of New Reactivity Measurement System and New Reactor Noise Analysis Equipment in a VVER-440 Nuclear Power Plant

2010 ◽  
Vol 57 (5) ◽  
pp. 2689-2696 ◽  
Author(s):  
János Vegh ◽  
Sándor Kiss ◽  
Sándor Lipcsei ◽  
Csaba Horvath ◽  
István Pos ◽  
...  
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Measurement System ◽  
Nuclear Power ◽  
Noise Analysis ◽  
Reactivity Measurement ◽  
Reactor Noise

Research and Solution on Vibration and Noise Caused by Orifice Cavitations of PTR System

2013 ◽  
Author(s):  
Ting Yang ◽  
Yao Dong Wang ◽  
Xu Lun Jiang ◽  
Chun Ming Wu
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Pressure Drop ◽  
Nuclear Power ◽  
System Function ◽  
Spent Fuel ◽  
Design Code ◽  
Field Investigations ◽  
Nuclear Power Plant Unit ◽  
Cooling Ability

There are serious vibration and noise in the pipes of cooling circle in Reactor Cavity and Spent Fuel Pit Cooling and Treatment System (PTR) system during the commissioning test of a Nuclear Power Plant Unit 1, after the improvement of the cooling ability for PTR was actualized. It has been confirmed that the orifice cavitations is the cause of vibration and noise. Engineering modification plans have been proposed: add the series number of throttle orifices and decrease pressure drop of each throttle orifices. Proper engineering redundancy shall be considered according to the test phenomenon and data to keep the orifices from the zones of cavitations. The field investigations after the modification shows that the vibration and noise have been controlled in the acceptable limit while both system function and structure integrity meet the requirement of design code. Thus it can be said that the system modification is successful.

"Chemical Effect" Head Loss Pressure Drop for the Nuclear Power Plant Safety

2019 ◽  
Vol 28 (24) ◽  
pp. 37-59 ◽  
Author(s):  
Robert Choromokos ◽  
Pete Mast ◽  
Jong-Hee M. Park
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Pressure Drop ◽  
Nuclear Power ◽  
Head Loss ◽  
Chemical Effect ◽  
Plant Safety ◽  
Nuclear Power Plant Safety

Integrity Evaluation of Korean Nuclear Reactor Pressure Vessel Under Pressurized Thermal Shock Conditions According to JEAC

2013 ◽  
Author(s):  
Se-Chang Kim ◽  
Jae-Boong Choi ◽  
Doo-Ho Cho ◽  
Sang-Min Lee ◽  
Yong-Beum Kim ◽  
...  
Keyword(s):  
Finite Element ◽  
Power Plant ◽  
Nuclear Power Plant ◽  
Thermal Shock ◽  
Pressure Vessel ◽  
Nuclear Power ◽  
Reactor Pressure Vessel ◽  
Operation Conditions ◽  
Pressurized Thermal Shock ◽  
Reactor Pressure

In nuclear power plant, reactor pressure vessel (RPV) is the primary equipment that contains reactor cores and coolant. The RPV integrity should be evaluated in consideration with transient operation conditions and material deterioration. Especially, the pressurized thermal shock (PTS) has been considered as one of the most important issues regarding the RPV integrity since Rancho Seco nuclear power plant accident in1978. In this paper, integrity evaluation of Korean RPV was performed by using finite element analysis. PTS conditions like small break loss of coolant accident (SBLOCA) and Turkey Point steam line break (TP-SLB) were applied as loading conditions. Neutron fluence data of actual RPV operated over 30 years was used to determine fracture toughness of RPV material. The 3-dimensional finite element model including circumferential surface crack was generated for fracture mechanics analysis. The RPV integrity was evaluated according to Japan Electric Association Code (JEAC).

Safety Approach and Ageing Management in Czech Rules and Standards

2009 ◽  
Author(s):  
Jaroslav Bartonicek ◽  
Lubomir Junek ◽  
Milan Vrana ◽  
Stanislav Vejvoda
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
High Energy ◽  
Proactive Approach ◽  
Plant Safety ◽  
Operation Conditions ◽  
Concept Usage ◽  
Time Required

There are basic technical (protection) objectives determined for assurance of nuclear power plant safety and the following generally belong among them: - Reactor pressure vessel shut down, - Long term maintenance of sub-critical state, - Long term cooling, - Prevention of radioactivity leakage. To ensure these objectives multi-step concept of deep protection is used for the design of a nuclear power plant and it includes: - Prevention of failures during normal and abnormal operation, - Control of failures and their consequences, - Minimizing of risks during accidents. Failure of operating systems is conservatively postulated for determination of systems requirements using for failure ensure as piping breaks. Ensure of these postulated failures come under multilevel safety approach. Failure consequences should be mainly ensured by design measures as separation of high energy piping, whip piping restrains etc. Efficiency of design measures have to be demonstrated. This passive safety procedure during design of new NPP can be applied. Application of this passive procedure for operating NPP can lead to technical and economical problems. It can be done by non precise and non sufficient requirements, current standards and documents. Leak before break concept (LBB) is very often out due to break operation conditions for successful concept usage. Beak preclusion concept was defined in Germany thirty years ago. The concept is developed from this time. Required quality of SSC is basic of this concept. The quality has to be received during manufacturing and assembly of new components to system or the quality passport has to be documented for SSC in operation before enlistment to the concept. During next operation they are sufficient and redundant measures necessary to control and to manage ageing phenomena (conceptual, technological, and physical) for exclusion of premature ageing. This proactive approach is also basic of documents from the last year’s required ageing and lifetime management. In Czech NPPs postulated failures and their consequences in accord with producer knowledge state at that time were insured. Postulated failures and their consequences were insured partly design measures and partly design supposed quality too. It is very difficult to realize new requirements on needed design provision on NPP in operation. It is impracticable in any cases. Needed national law for approach application exists in Czech from 1997. Regulation on lifetime management and national nuclear standards with specific requirements exist in Czech too. There are backgrounds for application proactive approach as it is used in Germany NPPs. New safety approach was provided in Czech NPPs. SSC are separated into three groups on the base safety approach: - SSC must not fail (guarantee of quality), - SSC may fail in rare case (preventative maintenance), - SSC may fail (failure orientated maintenance). The contribution deals about new Czech safety concept aspects, boundary conditions, needed document and proactive measures.

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