Study on In-Service Inspection Program and Inspection Technologies for Commercialized Sodium-Cooled Fast Reactor

2006 ◽  
Author(s):  
Masato Ando ◽  
Shigenobu Kubo ◽  
Yoshio Kamishima ◽  
Toru Iitsuka
Keyword(s):  
Steam Generator ◽  
Fast Reactor ◽  
Nuclear Power ◽  
Boundary Function ◽  
Sodium Coolant ◽  
Cycle Systems ◽  
And Performance ◽  
Double Wall ◽  
Wall System ◽  
Service Inspection

The objective of in-service inspection of a nuclear power plant is to confirm integrity of function of components necessary to safety, and satisfy the needs to protect plant investment and to achieve high plant ability. The sodium-cooled fast reactor, which is designed in the feasibility study on commercialized fast reactor cycle systems in Japan, has two characteristics related to in-service inspection. The first is that all sodium coolant boundary structures have double-wall system. Continuous monitoring of the sodium coolant boundary structures are adopted for inspection. The second characteristic is the steam generator with double-wall-tubes. Volumetric testing is adopted to make sure that one of the tubes can maintain the boundary function in case of the other tube failure. A rational in-service inspection concept was developed taking these features into account. The inspection technologies were developed to implement in-service inspection plan. The under-sodium viewing system consisted of multi ultrasonic scanning transducers, which was used for imaging under-sodium structures. The under-sodium viewing system was mounted on the under-sodium vehicle and delivered to core internals. The prototype of under-sodium viewing system and vehicle were fabricated and performance tests were carried out under water. The laboratory experiments of volumetric testing for double-wall-tubes of steam generator, such as ultrasonic testing and remote-field eddy current testing, were performed and technical feasibility was assessed.

Simulation Study of Dynamic Characteristic of Once-Through Steam Generator for CEFR

2016 ◽  
Author(s):  
Ye Shangshang ◽  
Yang Hongyi ◽  
Wang Xiaokun
Keyword(s):  
Dynamic Characteristic ◽  
Steam Generator ◽  
Fast Reactor ◽  
Nuclear Power ◽  
Simulation Method ◽  
Nucleate Boiling ◽  
Simulation System ◽  
Stable Operation ◽  
Steam Quality ◽  
Water Steam

The reliability of steam generator is extremely important for the sodium-cooled fast reactor nuclear power plant safety and stable operation. The convective heat transfer mechanism of the once-through steam generator (OTSG) of China Experimental Fast Reactor (CEFR) was researched. The water/steam side was divided into four areas according to the heat flux and steam quality, named subcooled, nucleate-boiling, film-boiling, and superheater. In order to accurate determine the DNB, the CHF table was used in this paper. Based on the homogeneous flow model and fixed boundary method, a thermal-hydraulic simulation system, which named OTAC, was established in this paper. To evaluate its performance, the predictions of this method were compared with PSM-W code. The maximum difference between the temperatures predicted by this model and PSM-W was ∼5K. The calculated results are consistent with the actual experiment data, which indicates the correctness of the mathematical model and simulation method. Static and dynamic characteristic researches of CEFR OTSG have done in the simulation system. And the system can be used to simulate the OTSG dynamic in real-time.

Inversion of Eddy Current Test Signals Obtained from Steam Generator Tubes

2005 ◽  
Vol 297-300 ◽  
pp. 2219-2224 ◽  
Author(s):  
Young H. Kim ◽  
Sung Jin Song ◽  
Jin Soo Hur ◽  
Eui Lae Kim ◽  
Chang Jae Yim ◽  
...  
Keyword(s):  
Finite Element ◽  
Steam Generator ◽  
Eddy Current ◽  
Nuclear Power ◽  
Principal Component ◽  
Automated System ◽  
Pressurized Water ◽  
Current Test ◽  
Flaw Characterization ◽  
Service Inspection

Eddy current testing (ECT) is widely used in in-service inspection as well as pre-service inspection of the steam generator (SG) tubes in nuclear power plant of pressurized water reactor type. The interpretation of ECT signals, however, is truly a difficult task so that the reliability enhancement of signal interpretation is strongly desired. An enhanced interpretation tools for ECT signals have been developed by the novel combination of neural networks and finite element modeling for quantitative flaw characterization SG tubes. A database was constructed using synthetic ECT signals generated by the finite element models and principal component analysis (PCA) was adopted in order to optimize the feature set of ECT signals. The improvement in the performances by the features with PCA and the excellent performance for the experimental ECT signals demonstrate the high potential of the developed inversion tools for reliable interpretation of eddy current signals. To explore the possibility of applying the developed approach in practical inspection, we developed an automated system (laboratory prototype) that can acquire experimental ECT signals from SG tubes and carry out the quantitative flaw characterization in a real time fashion by applying the approach developed in the present work.

scholarly journals A Conceptual Approach to Eliminate Bypass Release of Fission Products by In-Containment Relief Valve under SGTR Accident

2018 ◽  
Vol 2018 ◽  
pp. 1-12
Author(s):  
Taeseok Kim ◽  
Wonjun Choi ◽  
Joongoo Jeon ◽  
Nam Kyung Kim ◽  
Hoichul Jung ◽  
...  
Keyword(s):  
Steam Generator ◽  
Nuclear Power ◽  
Power Plants ◽  
Fission Products ◽  
Mitigation Strategies ◽  
Severe Accident ◽  
Base Case ◽  
Conceptual Approach ◽  
Relief Valve ◽  
Radioactive Nuclides

During a hypothesized severe accident, a containment building is designed to act as a final barrier to prevent release of fission products to the environment in nuclear power plants. However, in a bypass scenario of steam generator tube rupture (SGTR), radioactive nuclides can be released to environment even if the containment is not ruptured. Thus, thorough mitigation strategies are needed to prevent such unfiltered release of the radioactive nuclides during SGTR accidents. To mitigate the consequence of the SGTR accident, this study was conducted to devise a conceptual approach of installing In-Containment Relief Valve (ICRV) from steam generator (SG) to the free space in the containment building and it was simulated by MELCOR code for numerical analysis. Simulation results show that the radioactive nuclides were not released to the environment in the ICRV case. However, the containment pressure increased more than the base case, which is a disadvantage of the ICRV. To minimize the negative effects of the ICRV, the ICRV linked to Reactor Drain Tank (RDT) and cavity flooding was performed. Because the overpressurization of containment is due to heat of ex-vessel corium, only cavity flooding was effective for depressurization. The conceptual design of the ICRV is effective in mitigating the SGTR accident.

Structural Integrity Assessment of Steam Generator Tubes Using a New EPRI Statistical Approach

2002 ◽  
Author(s):  
Carlos Alexandre de Jesus Miranda ◽  
Miguel Mattar Neto
Keyword(s):  
Steam Generator ◽  
Wall Thickness ◽  
Nuclear Power ◽  
Structural Integrity ◽  
Tube Wall ◽  
New Approach ◽  
Inconel 600 ◽  
Integrity Assessment ◽  
Tube Wall Thickness ◽  
Uniform Tube

A fundamental step in tube plugging management of a Steam Generator (SG), in a Nuclear Power Plant (NPP), is the tube structural integrity evaluation. The degradation of SG tubes may be considered one of the most serious problems found in PWRs operation, mainly when the tube material is the Inconel 600. The first repair criterion was based on the degradation mode where a uniform tube wall thickness corrosion thinning occurred. Thus, a requirement of a maximum depth of 40% of the tube wall thickness was imposed for any type of tube damage. A new approach considers different defects arising from different degradation modes, which comes from the in-service inspections (NDE) and how to consider the involved uncertainties. It is based on experimental results, using statistics to consider the involved uncertainties, to assess structural limits of PWR SG tubes. In any case, the obtained results, critical defect dimensions, are within the regulatory limits. In this paper this new approach will be discussed and it will be applied to two cases (two defects) using typical data of SG tubes of one Westinghouse NPP. The obtained results are compared with ‘historical’ approaches and some comments are addressed from the results and their comparison.

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