Study on uncertainty evaluation methodology for decay heat removal experiment in sodium experimental facility

Akimoto, Yuta; Ezure, Toshiki; Onojima, Takamitsu; Kurihara, Akikazu

Proceedings of 30th International Conference on Nuclear Engineering (ICONE30) (Internet), 9 Pages, 2023/05

A numerical analysis method has been developed to evaluate thermal-hydraulic behaviors in a reactor vessel under the operation of a NC-DHRS at the Japan Atomic Energy Agency. During the validation of the evaluation method, in addition to uncertainties due to the numerical solution and input parameters in simulations, it is important to quantify uncertainties due to the experimental data. From this perspective, JAEA has been developing an experimental database and uncertainty evaluation methods for sodium experiments during operation of the NC-DHRS. In this study, the authors have proposed an uncertainty evaluation approach during relative calibrations of thermocouples in sodium experiments. The proposed approach was applied to experimental data obtained in a sodium NC-DHRS experiment conducted at PLANDTL-2. As a result, uncertainties of the experimental data were successfully evaluated and the applicability of the method to temperature measurement in sodium experiments was confirmed.

Study on multi-dimensional core cooling behavior of sodium-cooled fast reactors under DRACS operating conditions

Ezure, Toshiki; Onojima, Takamitsu; Tanaka, Masaaki; Kobayashi, Jun; Kurihara, Akikazu; Kameyama, Yuri*

Proceedings of 18th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-18) (USB Flash Drive), p.3355 - 3363, 2019/08

Steady-state sodium experiments under the operating conditions of a decay heat removal system (DHRS) were carried out as part of the safety enhancement of sodium-cooled fast reactors using the PLANDTL 2 facility, which has 30 heated channels with electric heaters and 25 no-heated channels as the simulated core. In the experiments, a direct reactor auxiliary cooling system (DRACS) with a dipped type direct heat exchanger (DHX) in the upper plenum was used as the DHRS. This paper reports on the preliminary experimental results of the PLANDTL 2 experiments under the DRACS operating conditions without flow in the primary circuit, including the thermal hydraulic interactions between the upper plenum and the core under the DHX operating conditions and the resulting core cooling behavior from the outside of the multiple rows of the fuel assemblies

A Study on the thermal-hydraulics in the damaged subassemblies under the operation of decay heat removal system

Ono, Ayako; Onojima, Takamitsu; Doda, Norihiro; Miyake, Yasuhiro*; Kamide, Hideki

Proceedings of 2016 International Congress on Advances in Nuclear Power Plants (ICAPP 2016) (CD-ROM), p.2183 - 2192, 2016/04

Some auxiliary cooling systems to remove the decay heat of the core are under consideration for a sodium-cooled fast reactor, and two of the typical systems are primary reactor auxiliary cooling system (PRACS) and direct reactor auxiliary cooling system (DRACS). In this study, sodium experiments were conducted in order to confirm the applicability of the PRACS and DRACS under a situation assuming the severe accidents with core melting. The plant dynamics test loop was used for these experiments, which contains a simulated core, the PRACS and DRACS. The core melt situation is simulated by shutting off the inlet of subassemblies (S/A). The experimental results revealed the cooling process of the partially/completely inactive S/A and confirmed the long-term heat removal by the PRACS/DRACS.

Experimental study on thermal stratification phenomena in compact reactor vessel of sodium cooled fast reactor; Evaluation on stratification interface behavior under natural circulation condition

Hagiwara, Hiroyuki; Kimura, Nobuyuki*; Onojima, Takamitsu; Nagasawa, Kazuyoshi*; Kamide, Hideki; Tanaka, Masaaki

JAEA-Research 2014-014, 178 Pages, 2014/09

JAEA-Research-2014-014.pdf:53.12MB

Thermal stratification in the upper plenum is one of the most important issues of a reactor vessel in sodium cooled fast reactor. The steep temperature gradient across the stratification interface may cause the thermal load against the reactor vessel wall. In this study, the water experiment was carried out using the 1/11 scale upper plenum model of the Japan sodium-cooled fast reactor (JSFR) in order to evaluate the thermal stratification under the natural circulation condition and a direct heat exchanger (DHX) operation condition. The temperature gradient under the natural circulation condition was approximately 1/3 times smaller than that under the forced circulation condition. In the DHX operation case, the steep temperature gradient occurred in the lower region of upper plenum due to the cold fluid from the outlet of DHX.

Fundamental behavior of vortex cavitation in a 1/22 scaled upper plenum model of sodium-cooled fast reactor; Influences of kinematic viscosity and system pressure

Ezure, Toshiki; Ito, Kei; Onojima, Takamitsu; Kimura, Nobuyuki; Kamide, Hideki

Proceedings of 8th Japan-Korea Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-8) (USB Flash Drive), 7 Pages, 2012/12

In this study, water experiments were performed in the 1/22 scaled upper plenum model of JSFR. Occurrence behavior of vortex cavitation was grasped quantitatively by means of the visualization and image analyses under several conditions of kinematic viscosity ) and pressure (). The experimental results showed that the vortex cavitation has dependence on the variation of and P. The increase of at least in the present small model, leaded to the restriction of cavitation as assumed by Burgers model. And also, the restriction level of vortex cavitation according to the increase was smaller than the evaluation using cavitation factor.

Experimental study on thermal stratification in a reactor vessel of innovative sodium cooled fast reactor; Characteristics of stratification interface under natural circulation operation

Kimura, Nobuyuki; Onojima, Takamitsu; Kamide, Hideki

Proceedings of 9th International Topical Meeting on Nuclear Thermal-Hydraulics, Operation and Safety (NUTHOS-9) (CD-ROM), 12 Pages, 2012/09

In the Japan Sodium-cooled Fast Reactor, thermal stratification phenomena occur in the reactor vessel during scram transient. In the study, the characteristics of stratification interface were investigated under the natural circulation operation during the scram transient using the 1/11th scale upper plenum model. The experimental results showed that the temperature gradient under the natural circulation operation was reduced to 1/2.6-1/6.2 in comparison with that under the forced circulation operation.

Development of sodium small leak detection technique Preliminary experimental study on sodium aerosol detection sensitivity using laser induced breakdown spectroscopy

Nagai, Keiichi; Nagai, Keiichi; Otaka, Masahiko; Miyakoshi, Hiroyuki; Onojima, Takamitsu

JNC TN9400 2003-058, 35 Pages, 2003/05

JNC-TN9400-2003-058.pdf:1.15MB

A preliminary examination was carried out for evaluation of the detection sensitivity of Laser Sodium Leak Detector (LLD) based on a principle of Laser Induced Breakdown Spectroscopy (LIBS). Evaluation criteria and examination conditions were planned based on the results of preliminary experiments.The main results are as follows:(1) Signal intensity of LLD was obtained with parameter of sodium concentration in combustion aerosols. The signal intensity in the combustion aerosols was nearly equivalent to that in case of sodium mist using carrier gas of nitrogen. It was shown that LLD was effective to detect sodium in the combustion aerosols.(2) Diameter or chemical component of sodium aerosols are one of significant factors for the detection sensitivity of LLD. Preliminary experiments were carried out with parameters of humidity, oxygen concentration, and pressure of carrier gas. The obtained experimental data of a few cases showed that influence of these parameters was limited on the detection sensitivity of LLD.3) Based on the preliminary experimental results, main conditions of a sensitivity evaluation test plan were decided for LLD.

Experimental study on thermal stratification phenomena in Sodium-cooled Fast Reactor, 2; Influence evaluation which a temperature difference and a flow of DHX outlet give to a thermal stratification interface

Hagiwara, Hiroyuki; Kimura, Nobuyuki*; Nagasawa, Kazuyoshi*; Onojima, Takamitsu; Kamide, Hideki; Tanaka, Masaaki

no journal, , 

Influence of DHX operating conditions in natural circulation on thermal stratification surface occurring under DHX was studied by using 1/11 scale upper plenum water test for Japan Sodium-cooled Fast Reactor. As the result, it was found that temperature difference between fuel subassembly outlet after scrum and DHX outlet became smaller or DHX outlet velocity became lower, temperature gradient decreased.

Study on chemical reactivity control of liquid sodium, 15; Oxidation property

Nagai, Keiichi; Nishimura, Masahiro; Onojima, Takamitsu; Saito, Junichi; Ara, Kuniaki

no journal, , 

no abstracts in English

PLANDTL-2 experiment for evaluation of decay heat removal in sodium-cooled fast reactors; Multi dimensional cooling behavior through inter wrapper gap region under steady state conditions

Ezure, Toshiki; Onojima, Takamitsu; Kobayashi, Jun; Kurihara, Akikazu; Tanaka, Masaaki

no journal, , 

As a part of the safety enhancement of sodium-cooled fast reactors, steady state sodium experiments were carried out using PLANDTL-2 facility, which has 30 heated channels using electric heater and 25 no-heated channels as the simulated core, to evaluate the core cooling behavior under decay heat removal systems (DHRSs) operating conditions. Temperature distributions in inter-wrapper gap region (IWG) were grasped quantitatively under the several experimental conditions using DHRSs. From the results, multi-dimensional core cooling behavior in the IWG was grasped.