Sodium experiments on natural circulation decay heat removal and 3D simulation of plenum thermal hydraulics

Kamide, Hideki; Ono, Ayako; Kimura, Nobuyuki; Endo, Junji*; Watanabe, Osamu*

Proceedings of International Conference on Fast Reactors and Related Fuel Cycles; Safe Technologies and Sustainable Scenarios (FR-13), Companion CD (CD-ROM), 11 Pages, 2015/04

http://www-pub.iaea.org/MTCD/Publications/PDF/SupplementaryMaterials/P1665CD/Track3_Safety.pdf#page=21

http://www.iaea.org/nuclearenergy/nuclearpower/Downloadable/Meetings/2013/2013-03-04-03-07-CF-NPTD/T3.3/T3.3.kamide.pdf

Natural circulation decay heat removal is one of the significant issues for fast reactor safety, especially in long term station blackout events. Several sodium experiments were carried out using a 7-subassmbly core model for core thermal hydraulics under natural circulation conditions and for onset transients of natural circulation in a decay heat removal system (DHRS) including natural draft. Significant heat removal via inter-wrapper flow was confirmed in the experiments. Solidification of sodium in an air cooler is one of key issues in loss of heat sink events. Natural circulation characteristics under long-term decay heat removal were also obtained. Multi-dimensional phenomena, e.g., thermal stratification and bypass flow in plenums and/or heat exchangers, may influence the natural circulation. Thus, 3D simulation method was developed for entire region in the primary loop. Comparison of temperature distributions in a DHRS heat exchanger between experiment and analysis was done.

Study on high cycle thermal fatigue in mixing tee; Evaluation of transfer characteristics of temperature fluctuation from fluid to structure

Kimura, Nobuyuki; Kobayashi, Jun; Kameyama, Yuri*; Nagasawa, Kazuyoshi*; Ezure, Toshiki; Ono, Ayako; Kamide, Hideki

JAEA-Research 2014-009, 104 Pages, 2014/07

JAEA-Research-2014-009.pdf:15.23MB
JAEA-Research-2014-009-appendix(CD-ROM).pdf:17.88MB

In this study, water experiments (WATLON) were carried out to clarify the unsteady behavior of heat transfer under wall jet condition in the mixing tee. In experiments, heat transfer coefficients between fluid and wall in the mixing region were obtained from temperature measurements using thermocouples (movable tree type in fluid and embedded type in wall). To clarify the relation between the local velocity and the wall temperature, those were measured simultaneously by the Particle Image Velocimetry (PIV) and the thermocouple measurement, respectively. Sampling time of the velocity by the PIV and the temperature by the thermocouple were synchronized in the measurement. The experimental results showed that the heat transfer coefficient was from 2 - 6 time larger than the reference value predicted by the Dittus-Boelter correlation in straight pipes and was increased as the local velocity near the wall.

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.

Experimental study on prevention of high cycle thermal fatigue in a compact reactor vessel of advanced sodium cooled reactor; Thermal striping phenomena at bottom of upper internal structure

Kobayashi, Jun; Kimura, Nobuyuki; Tobita, Akira; Kamide, Hideki

JAEA-Research 2012-014, 40 Pages, 2012/07

JAEA-Research-2012-014.pdf:7.8MB

As the temperatures difference between the control rod channels and the core fuel subassemblies is around 100 C centigrade, temperature fluctuation due to the fluid mixing at the core outlet may cause high cycle thermal fatigue at the bottom of upper internal structure (UIS) in JSFR. Then, a water experiment was conducted using an 1/3 scale 60 sector model. As a result, thermal striping phenomena in the region between the fuel subassembly outlet and the bottom of the UIS were grasped. The modified geometry of the UIS bottom and the handling head of the primary CR channel was created so as to suppress the cold jets from the CR channels. The comparison of measured temperature fluctuations around the CR channels revealed that the modified geometry was effective to decrease the temperature fluctuation intensity and amplitude in the sensitive frequency band to the stress conversion.

Development of flow-induced vibration evaluation methodology for large-diameter piping with elbow in Japan sodium-cooled fast reactor

Yamano, Hidemasa; Tanaka, Masaaki; Kimura, Nobuyuki; Ohshima, Hiroyuki; Kamide, Hideki; Watanabe, Osamu*

Nuclear Engineering and Design, 241(11), p.4464 - 4475, 2011/11

https://doi.org/10.1016/j.nucengdes.2011.06.005

This paper describes the current status of flow-induced vibration evaluation methodology development for the primary piping in JSFR, in particular emphasizing on the development approach and research activities that investigate unsteady hydraulic characteristics in a short-elbow piping. Experimental efforts have been made using 1/3-scale and 1/10-sca1e single elbow test sections for the hot-leg piping and 1/4-scale and 1/7-scale triple-elbow test sections for the cold-leg piping. Recent experiments using the 1/3-scale test section revealed that a swirl flow at the inlet of the hot-leg piping hardly influenced the pressure fluctuations onto the pipe. Simulation activities include Unsteady Reynolds Averaged Navier Stokes equation (U-RANS) and Large Eddy Simulation (LES) approaches. Numerical results using the U-RANS approach appear in this paper, indicating its applicability to the hot-leg piping experiments.

Experimental investigation on bubble characteristics entrained by surface vortex

Ezure, Toshiki; Kimura, Nobuyuki; Miyakoshi, Hiroyuki; Kamide, Hideki

Nuclear Engineering and Design, 241(11), p.4575 - 4584, 2011/11

https://doi.org/10.1016/j.nucengdes.2010.09.021

The cover gas entrainment at the free surface of sodium coolant becomes one of the significant issues according to the compact sizing of reactor vessel in the latest design of sodium cooled fast reactor. In the present study, some basic water experiments for the gas entrainment due to the surface vortex were performed in order to obtain the fundamental knowledge about the entrained bubble size. Distributions of entrained bubble diameters in several experimental conditions were obtained from bubble images using an image processing technique. Velocity fields around vortices and surface dimple shapes (gas cores) due to surface vortices were measured to grasp those influences on bubble shapes. The result showed that mean equivalent diameters of bubbles were varied from 1.3 to 2.1 mm in the range of present experimental conditions. The bubble sizes were influenced by the thickness of gas core.

Sodium experiment on fully natural circulation systems for decay heat removal in Japan sodium-cooled fast reactor

Kamide, Hideki; Kobayashi, Jun; Ono, Ayako; Kimura, Nobuyuki; Watanabe, Osamu*

Proceedings of 14th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-14) (CD-ROM), 16 Pages, 2011/09

Fully natural circulation system is adopted in a decay heat removal system (DHRS) of Japan sodium cooled fast reactor (JSFR). Sodium experiments were carried out for heat transfer characteristics of a sodium-sodium heat exchanger of PRACS and start-up transient of the DHRS loop with parameters of pressure loss coefficients in the loops. Influences of the pressure loss coefficient in the primary loop and the DHRS loop were limited on the core temperature and also heat removal of PRACS due to recovery of natural circulation head via the increase of temperature difference in each loop.

Experimental study on influences of kinematic viscosity on occurrences of cavitation due to sub-surface vortex

Ezure, Toshiki; Kimura, Nobuyuki; Kobayashi, Jun; Kamide, Hideki

Proceedings of 14th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-14) (CD-ROM), 13 Pages, 2011/09

In order to clarify the influence of kinematic viscosity () on the occurrence of vortex cavitation, a water experiment was carried out in a cylindrical tank with a suction pipe. The occurrences of vortex cavitation were measured under several fluid temperature conditions between 10C and 80C ( : 1.310 to 3.710m/s). The velocity fields around vortex were also measured by Particle Image Velocimetry. The influence of was observed under relatively high conditions. However, that influence diminished with the decrease of or the increase of suction velocity. And also, normalized circulation was found as an index to estimate such influences of or suction velocity on the vortex cavitation.

Sodium experiments on decay heat removal system of Japan sodium cooled fast reactor; Start-up transient of decay heat removal system

Kamide, Hideki; Kobayashi, Jun; Ono, Ayako; Kimura, Nobuyuki; Miyakoshi, Hiroyuki; Watanabe, Osamu*

Proceedings of 7th Korea-Japan Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-7) (CD-ROM), 8 Pages, 2010/11

Fully natural circulation system is adopted in a decay heat removal system (DHRS) of Japan Sodium Cooled Fast Reactor (JSFR). The JSFR has two loops of the main heat transport system in order to reduce number of components and the construction cost. The DHRS of JSFR consists of one DRACS and two PRACS, which have a heat exchanger in a primary-side inlet plenum of IHX in each loop. Sodium experiments were carried out for heat transfer characteristics of a sodium-sodium heat exchanger of PRACS and also start-up transient of the DHRS loop. Heat transfer coefficient on the tube outer surface was in good agreement with a conventional correlation under operation condition in the reactor. The transient experiments for the start-up of DHRS loop showed that smooth increase of natural draft in the air duct followed by the sodium flow rate in the DHRS loop. Some delay of the flow rate increase was recognized in the DHRS loop as compared with that of the natural draft in the air cooler.