DOUBLE WALL BAYONET TUBE STEAM GENERATOR INVESTIGATION IN HERO EXPERIMENTAL CAMPAIGN

2019 ◽  
Vol 2019.27 (0) ◽  
pp. 2208
Author(s):  
Pierdomenico Lorusso ◽  
Alessio Pesetti ◽  
Mariano Tarantino
Keyword(s):  
Steam Generator ◽  
Experimental Campaign ◽  
Double Wall

ICONE19-43935 Development of a Helical-Coil Double Wall Tube Steam Generator for 4S Reactor

2011 ◽  
Vol 2011.19 (0) ◽  
pp. _ICONE1943-_ICONE1943
Author(s):  
Yuko Kitajima ◽  
Takehisa Hino ◽  
Katsuhiko Sato ◽  
Shigeki Maruyama ◽  
Noboru Jimbo
Keyword(s):  
Steam Generator ◽  
Helical Coil ◽  
Double Wall

E216 Development of advanced high reliable helical type double wall tube steam generator

2009 ◽  
Vol 2009.14 (0) ◽  
pp. 461-462
Author(s):  
Noboru JIMBO ◽  
Kunio SHIMANO ◽  
Shigeki MARUYAMA ◽  
Noriyasu KOBAYASHI ◽  
Takehisa HINO
Keyword(s):  
Steam Generator ◽  
Double Wall

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.

Experimental Investigation in LIFUS5/Mod2 Facility of Spiral-Tube Steam Generator Rupture Scenarios for ELFR

2017 ◽  
Author(s):  
Alessio Pesetti ◽  
Alessandro Del Nevo ◽  
Andrea Neri ◽  
Stefano Cati ◽  
Valerio Sermenghi ◽  
...  
Keyword(s):  
Steam Generator ◽  
Tube Bundle ◽  
Quality Data ◽  
Water Tank ◽  
Spiral Tube ◽  
Pressure Transducers ◽  
Experimental Campaign ◽  
Evolution Path ◽  
Pressure Peak ◽  
Lower Temperature

In the framework of the European Commission LEADER project, an experimental campaign of seven tests was performed in the LIFUS5/Mod2 facility, at ENEA CR Brasimone, for investigating the postulated Steam Generator Tube Rupture (SGTR) event in a relevant configuration for the Spiral-Tube Steam Generator (STSG) of the European Lead Fast Reactor (ELFR). The LIFUS5/Mod2 facility is composed by a water tank of 15 L injecting subcooled water up to 200 bar into the reaction tank of 100 L (420 mm of diameter), which is connected by a 3 inch pipe to the dump tank of 2 m3. A dedicated test section was designed, assembled and implemented in the reaction tank. It is composed by 188 tubes, vertically disposed with triangular pitch inside a cylindrical support. This tube bundle is representative of a portion of the STSG of ELFR. The cylindrical support is closed at the lower and upper end by two tube plates and has a perforated lateral shell (300 mm of diameter and 400 mm high). The reaction tank is filled by Lead-Bismuth Eutectic alloy (LBE) at 400°C up to the top tube plate, with an argon cover gas at about 2 bar. The water is injected at about 180 bar and 270°C through the central tube, at middle height of the bundle. The water-LBE interaction is characterised by high quality data acquisition system: 6 fast Pressure Transducers (PTs) working at 10 kHz for precisely characterize the first narrow injection peaks, 70 low constant time Thermocouples (TCs) to understand the vapour evolution path and 13 strain gages (SGGs) for measuring the strain of the bundle and main vessel. The overall LEADER experimental campaign is constituted by seven tests, divided in three series (B1, B2 and B3), characterized by different injection orifice diameters of 4, 8.9 and 12.6 mm, respectively. This paper presents the experimental results of the first two tests of series B2 (B2.1 and B2.2) having 8.9 mm of injection orifice. The first test analysed showed a first narrow pressure peak of about 32 bar, some milliseconds after the cap rupture instant. The following pressurization due to the evaporation of water entered into the reaction vessel was of an analogues magnitude for both the tests (about 50 bar) and lasted some tenths of second. The water/LBE interaction lower temperature was reached on the inner ranks of tubes, about 150°C. The outer rank was cooled down to about 300°C. The strain gage measurements showed a decreasing deformation on the tubes toward the outer positions. No ruptures were observed on tubes surrounding the injector. The amount of LBE transported into the dump tank was strongly dependent on the LBE level in the reaction tank at the start of the tests and about 200 kg.

Assessment of SIMMER-III Code Based on Steam Generator Tube Rupture Experiments in LIFUS5/Mod2 Facility

2016 ◽  
Author(s):  
Alessio Pesetti ◽  
Alessandro Del Nevo ◽  
Nicola Forgione
Keyword(s):  
Steam Generator ◽  
Volume Fraction ◽  
Water Pressure ◽  
Water Injection ◽  
Measurement Data ◽  
Pressure Profile ◽  
Reaction Vessel ◽  
Steam Generator Tube ◽  
Experimental Campaign ◽  
Injection Line

An experimental campaign investigating the postulated Steam Generator Tube Rupture (SGTR) event, in relevant configurations for Heavy Liquid Metal Reactors (HLMRs), was carried out in the separate-effect facility LIFU5/Mod2, at ENEA CR Brasimone. Ten tests were performed injecting pressurized subcooled water into the reaction tank partially filled by Lead-Bismuth Eutectic alloy (LBE) at 400°C with a cover gas of argon at about 2 bar. Fast pressure transducers, thermocouples and strain gages provided high-quality measurement data for improving the phenomena understanding and supporting the development and validation phase of computer codes for SGTR numerical simulation. The experimental campaign is composed by two series of tests, characterized by different water pressure: 40 and 16 bar. The first two tests belonging to the low pressure experiments are presented, highlighting the pressurization time trends of the water injection tank, injection line and reaction vessel. The injected water mass flow rate and temperature trends in the reaction vessel were measured. The former test is the reference one and the latter was carried out for investigating the injection of water with higher sub-cooling. A post-test analysis of the two mentioned tests was carried out by SIMMER-III code. The pressure profile in the water injection tank was set as boundary condition of the calculation. The numerical analysis provided injection line and reaction tank pressurization in agreement with the experimental data. The lower water temperature test provided a better accordance with the measured data, due to the lower evaporation along the injection line. The SIMMER-III analysis also studied the water-LBE interaction from the volume fraction point of view and the energy released in the total reaction tank and in its cover gas.

A High Reliability Straight Tube LMFBR Steam Generator Design

1980 ◽  
Vol 102 (3) ◽  
pp. 568-572 ◽  
Author(s):  
A. H. Spring ◽  
D. D. DeFur
Keyword(s):  
Thermal Expansion ◽  
Steam Generator ◽  
High Reliability ◽  
Straight Tube ◽  
Steam Generators ◽  
Geometric Size ◽  
Differential Thermal Expansion ◽  
Generator Design ◽  
Double Wall ◽  
Reliability And Availability

Straight-tube counterflow steam generators for LMFBR applications have been proposed, tested, fabricated and operated with varying degrees of success. A new embodiment of the straight-tube concept is described which incorporates a number of unique features which contribute to high reliability and availability. These features include a replaceable bellows for accommodation of differential thermal expansion between shell and tubes and a redundant, crevice-free tube-to-tubesheet joint design. The design can also easily incorporate single-wall or double-wall tubes. Single and double-wall tube versions are described whose thermal and geometric size are based on anticipated manufacturing limitations. The results of scoping tests of the tube-to-tubesheet welds are described which provide positive indications of the soundness of the weld design.

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