TRACE code modeling of the horizontal steam generator of the PACTEL facility and calculation of a loss-of-feedwater experiment

Vihavainen, Juhani; Riikonen, Vesa; Kyrki-Rajamäki, Riitta

doi:10.1016/j.anucene.2010.06.013

Cited by 3 publications

(2 citation statements)

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“…PACTEL is included also in the OECD NEA's validation matrixes available in references [9,10]. The PACTEL experiments have provided valuable data for the validation processes of several codes, such as APROS, RELAP, CATHARE, and TRACE, for example, presented in reference [11].…”

Section: Pactel Experimentsmentioning

confidence: 99%

PACTEL and PWR PACTEL Test Facilities for Versatile LWR Applications

Kouhia

Purhonen

Riikonen

et al. 2012

Science and Technology of Nuclear Installations

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This paper describes construction and experimental research activities with two test facilities, PACTEL and PWR PACTEL. The PACTEL facility, comprising of reactor pressure vessel parts, three loops with horizontal steam generators, a pressurizer, and emergency core cooling systems, was designed to model the thermal-hydraulic behaviour of VVER-440-type reactors. The facility has been utilized in miscellaneous applications and experiments, for example, in the OECD International Standard Problem ISP-33. PACTEL has been upgraded and modified on a case-by-case basis. The latest facility configuration, the PWR PACTEL facility, was constructed for research activities associated with the EPR-type reactor. A significant design basis is to utilize certain parts of PACTEL, and at the same time, to focus on a proper construction of two new loops and vertical steam generators with an extensive instrumentation. The PWR PACTEL benchmark exercise was launched in 2010 with a small break loss-of-coolant accident test as the chosen transient. Both facilities, PACTEL and PWR PACTEL, are maintained fully operational side by side.

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Section: Pactel Experimentsmentioning

confidence: 99%

PACTEL and PWR PACTEL Test Facilities for Versatile LWR Applications

Kouhia

Purhonen

Riikonen

et al. 2012

Science and Technology of Nuclear Installations

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“…Nowadays, TRACE code (TRAC/RELAP Advanced Computational Engine) is being developed to make use of the most favorable characteristics of RELAP-5 and TRAC codes to simulate both, PWR and BWR, technologies. This code is nowadays being studied and it has been used in some interesting analyses as (Wang et al , 2009) and (Vihavainen et al , 2010).…”

Section: In Order To Improve the Spent Fuel Pools Safety Analysis Itmentioning

confidence: 99%

Use of TRACE best estimate code to analyze spent fuel storage pools safety

Alberola

Sánchez-Saez

Martorell

2014

Progress in Nuclear Energy

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Nuclear policies have experienced an important change since Fukushima Daiichi nuclear plant accident and the safety of spent fuels has been in the spot issue among all the safety concerns. At many countries, the spent fuel of nuclear power plants is not reprocessed so it has to be stored inside the facilities, normally in pools. As nuclear power plants use best estimate codes to perform safety analysis, it is interesting to assess the capacity of such codes to simulate spent fuel pools behavior. This paper uses TRACE thermal-hydraulic code to simulate steady state and transient conditions of spent fuel pools. The steady state results are compared with plant measurements of Maine Yankee with a good agreement between the calculations and the measurements. The transient simulated is a loss of cooling together with a loss of coolant through the transfer channel. TRACE heat transfer radiation model has been activated using the parameters obtained from COBRA thermal-hydraulic code.

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