Physics-Infused AI/ML Based Digital-Twin Framework for Flow-Induced-Vibration Damage Prediction in a Nuclear Reactor Heat Exchanger

The PIPO-FE and VIBIC finite-element computer codes, developed and updated over the past 30 years, are used to calculate heat exchanger tube flow-induced vibration (FIV) response. PIPO-FE includes a linear forced-vibration analysis of heat exchanger tubes subjected to all major flow-induced excitation mechanisms, namely fluidelastic instability, random turbulence-induced excitation and periodic wake shedding. VIBIC is for both linear and non-linear transient dynamic simulations of heat exchanger tubes. When used to simulate a tube with clearance supports (non-linear case), VIBIC calculates tube wear work-rates to aid in the prediction of tube fretting-wear damage. All the excitation mechanisms included in PIPO-FE analyses can be simulated in VIBIC. In addition, VIBIC can model friction forces between a tube and its supports, squeeze film forces produced by the resistance of the fluid opposing the relative motion of the tube and supports, and constant loads. An important application of these codes is the analysis of the susceptibility of a heat exchanger tube to vibration damage. These codes may be used at the design stage to assess a new heat exchanger, or during the operational stage to investigate a tube failure and determine if the damage was caused by vibration. If a vibration problem exists, then the codes can be used to assess the effectiveness of any proposed design modifications. To properly assess tube vibration damage, the codes must predict vibration response accurately. This paper documents the validation process of code predictions against measurements from three flow-induced vibration experiments conducted at Chalk River Laboratories: 1. A single-span cantilever tube bundle subjected to two-phase air-water cross flow; 2. A single-span cantilever tube bundle subjected to single- and two-phase Freon cross flow; and 3. A single-span U-bend tube bundle subjected to single-phase water and two-phase air-water partial cross flow. PIPO-FE and VIBIC code predictions for fluidelastic instability ratio and the response to random turbulence-induced excitation are compared to each other for each of these three experiments. The predictions from the two codes are in good agreement. In addition, the predictions for frequency, damping ratio, fluidelastic instability ratio and the response to random turbulence-induced excitation from both codes are in reasonable agreement with the experimental results.

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Temperature Transients in a Triple Heat Exchanger

Proceedings of the Institution of Mechanical Engineers Part C Journal of Mechanical Engineering Science ◽

10.1243/pime_proc_1984_198_102_02 ◽

1984 ◽

Vol 198 (3) ◽

pp. 145-154

Author(s):

P C Chiu ◽

E H K Fung

Keyword(s):

Heat Exchanger ◽

Nuclear Reactor ◽

Reactor Power ◽

Solar Heating ◽

Inlet Temperature ◽

Exchange Processes ◽

Flow Experiments ◽

Heating Plant ◽

Reactor Power Plant ◽

Secondary Fluid

A triple heat exchanger, so called because there are three heat exchange processes taking place in it, was built to simulate the system behaviour of a nuclear reactor power plant or a solar heating plant which is characterized by the two circulating loops of the fluid flow. Experiments were carried out to study the temperature transients under disturbances in secondary fluid inlet temperature and power output from immersion heaters. Numerical results were obtained from the weighted residual formulation of the proposed dynamic model and they were shown to be in general agreement with the two sets of experimental responses.

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Step Response of a Single-Solid, Single-Fluid Heat Exchanger Having Sinusoidally Space-Dependent Internal Heat Generation : Specially the Response of a Heterogeneous Nuclear Reactor with Sinusoidally Space-Dependent Power Generation

Transactions of the Japan Society of Mechanical Engineers ◽

10.1299/kikai1938.28.898 ◽

1962 ◽

Vol 28 (192) ◽

pp. 898-911

Author(s):

Wen-Jei YANG

Keyword(s):

Power Generation ◽

Heat Exchanger ◽

Heat Generation ◽

Nuclear Reactor ◽

Internal Heat ◽

Internal Heat Generation ◽

Step Response ◽

Heterogeneous Nuclear Reactor

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Research on Characteristics of Fluid Exciting Force of Reactor Internals

10.1115/pvp2021-62128 ◽

2021 ◽

Author(s):

Zhipeng Feng ◽

Liwen Deng ◽

Xuan Huang ◽

Pingchuan Shen ◽

Shuai Liu ◽

...

Keyword(s):

Pressure Fluctuation ◽

Nuclear Reactor ◽

Wide Band ◽

Exciting Force ◽

Design Stage ◽

Response Analysis ◽

Flow Induced Vibration ◽

Analysis And Evaluation ◽

Safety And Reliability ◽

Force Characteristics

Abstract Flow-induced vibration is an important issue related to the safety and reliability of nuclear reactor, which need to be analyzed and evaluated in the design stage. In order to obtain the input loads and key parameters used in the calculation of flow-induced vibration of reactor vessel internals (RVIs) that need to satisfy the engineering requirements. The typical RVIs are selected as the research object, and the fluid exciting force characteristics are studied based on the computational fluid dynamics methods. The results show that the fluid exciting force acting on the RVIs is a wide-band stochastic process. For upper internal, the largest pressure fluctuation occurs at the guide tubes and support columns located near the outlet. Therefore, it is necessary to pay more attention to these guide tubes and support columns in response analysis. As for core barrel, the root mean square value of the pressure fluctuation changes drastically at the inlet and outlet location. For lower internal, the lower flow field of RVIs is relatively disordered, and its pressure fluctuation possesses irregular characteristics. Each component of lower internal need to be considered in analysis and evaluation.

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