Study on the discharge behavior of the molten-core materials through the control rod guide tube; Investigations of the effect of an internal structure in the control rod guide tube on the discharge behavior

Kato, Shinya; Matsuba, Kenichi; Kamiyama, Kenji; Akaev, A.*; Vurim, A.*; Baklanov, V.*

Proceedings of 13th International Topical Meeting on Nuclear Reactor Thermal-Hydraulics, Operation and Safety (NUTHOS-13) (Internet), 12 Pages, 2022/09

The In-Vessel Retention (IVR) of molten-core in Core Disruptive Accidents (CDAs) is of prime importance in enhancing the safety of sodium-cooled fast reactors. One of the main subjects in ensuring IVR is to design the Control Rod Guide Tube (CRGT) which allows effective discharge of molten core materials from the core region. The effectiveness of the CRGT design is assessed through CDA analyses, and it is reasonable for these analyses to develop a computer code collaborated with experimental researches. Thus, experiments addressing the discharge behavior of the molten-core materials through the CRGT have proceeded as one of the subjects in the collaboration research named the EAGLE-3 project, and the obtained experimental results are reflected in the development of the SIMMER code. In this project, a series of out-of-pile tests using molten-alumina as the fuel simulant was conducted to understand the discharge behavior of molten-core materials through the CRGT. In this study, in order to investigate the effect of an internal structure in the CRGT on the discharge behavior of the molten-core materials, the data of an out-of-pile test in which the molten-alumina penetrated to a duct with the internal structure were analyzed. In addition, the post-test analysis using the SIMMER code was conducted and the results were compared with the test results.

Fragmentation and cooling behavior of a simulated molten core material discharged into a sodium pool with limited depth and volume

Matsuba, Kenichi; Kato, Shinya; Kamiyama, Kenji; Akayev, A. S.*; Baklanov, V. V.*

Proceedings of 28th International Conference on Nuclear Engineering (ICONE 28) (Internet), 4 Pages, 2021/08

https://doi.org/10.1115/ICONE28-64500

In order to obtain experimental knowledge on fragmentation and cooling behavior of molten core material discharged into regions where the depth and volume of sodium are limited, a series of out-of-pile experiments using molten alumina as a simulant for molten core material was conducted. It was found that following mechanisms might be involved in the fragmentation and cooling behavior in a shallow sodium pool: (1) FCI which occurs at location of impingement of the molten jet on the bottom plate promotes fragmentation. (2) If there is a sufficient amount of sodium as a heat sink outside the region, heat exchange by sodium flow in and out due to vapor expansion and condensation suppresses the sodium temperature rise. (3) This temperature suppression contributes to effective cooling of molten core material. In the future study, in order to confirm the mechanisms which was clarified in this study, analytical evaluation of the experimental result will be carried out using a simulation tool.

Study on the discharge behavior of molten-core through the control rod guide tube in the core disruptive accident of SFR

Kato, Shinya; Matsuba, Kenichi; Kamiyama, Kenji; Ganovichev, D. A.*; Baklanov, V. V.*

Proceedings of 2019 International Congress on Advances in Nuclear Power Plants (ICAPP 2019) (Internet), 9 Pages, 2019/05

In order to ensure In-Vessel Retention (IVR) of molten-core in Core Disruptive Accident (CDA), we are investigating the possibility of the molten-core discharge through the control rod guide tube (CRGT) to prevent energetics due to exceeding the prompt criticality. Internal structures of the CRGT, such as a sodium-flow regulator when the CRGT is connected to the high-pressure plenum, may disturb the discharge of molten-core from the core region. Based on above background, an experimental program to clarify characteristics of molten-core discharge through the CRGT has been commenced as one of subjects under a joint study with National Nuclear Center of the Republic of Kazakhstan (NNC-RK) named EAGLE-3 project. An experiment using molten-alumina as fuel simulant and sodium was conducted at the out-of-pile test facility owned by NNC-RK to investigate sodium cooling effect around the sodium flow regulator on its destruction. The experimental result represented that void development at the initiation of molten-alumina discharge eliminated liquid-phase sodium from the discharge path and this also eliminated sodium cooling effect around the sodium flow regulator. As a result, early destruction of the sodium flow regulator and massive discharge of molten alumina occurred in turn.

An Experimental study on the fragmentation and accompanying cooling behaviors of a simulated molten oxide fuel penetrating into a sodium pool

Matsuba, Kenichi; Kamiyama, Kenji; Toyooka, Junichi; Zuev, V. A.*; Kolodeshnikov, A. A.*

Proceedings of 17th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-17) (USB Flash Drive), 11 Pages, 2017/09

To clarify jet fragmentation and accompanying cooling behaviors of molten core materials discharged into sodium, results of the out-of-pile experiments with a simulant material (AlO) were analyzed. The results clarified that while AlO jets were entirely fragmented into smaller particles during their penetration to several tenths of a meter in depth of sodium, an additional significant distance was needed to cool down the particles to the degree that thermal loading on the steel structures could be neglected. Based on the results, it is concluded that in terms of the reduction of thermal load on the lower structures in the reactor vessels, the cooling distance after fragmentation should be treated.

Experimental discussion on fragmentation mechanism of molten oxide discharged into a sodium pool

Matsuba, Kenichi; Kamiyama, Kenji; Toyooka, Junichi; Tobita, Yoshiharu; Zuyev, V. A.*; Kolodeshnikov, A. A.*; Vassiliev, Y. S.*

Mechanical Engineering Journal (Internet), 3(3), p.15-00595_1 - 15-00595_8, 2016/06

http://doi.org/10.1299/mej.15-00595

To develop a method for evaluating the distance for fragmentation of molten core material discharged into sodium, the particle size distribution of alumina debris obtained in the FR tests was analyzed. The mass median diameters of solidified alumina particles were around 0.3 mm, which are comparable to particle sizes predicted by hydrodynamic instability theories such as Kelvin-Helmholtz instability. However, even though hydrodynamic instability theories predict that particle size decreases with an increase of Weber number, such the dependence of particle size on We was not observed in the FR tests. It can be interpreted that this tendency of measured mass median suggests that before hydrodynamic instabilities sufficiently grow to induce fragmentation, thermal phenomena such as local coolant vaporization and resultant vapor expansion accelerate fragmentation.

An Experimental study on heat transfer from a mixture of solid-fuel and liquid-steel during core disruptive accidents in Sodium-Cooled Fast Reactors

Kamiyama, Kenji; Konishi, Kensuke; Sato, Ikken; Toyooka, Junichi; Matsuba, Kenichi; Suzuki, Toru; Tobita, Yoshiharu; Pakhnits, A. V.*; Vityuk, V. A.*; Vurim, A. D.*; et al.

Proceedings of 10th International Topical Meeting on Nuclear Thermal Hydraulics, Operation and Safety (NUTHOS-10) (USB Flash Drive), 8 Pages, 2014/12

Retesting of Mo adsorption and Tc elution characteristics of molybdenum adsorbents for (n,) method; Joint experiment report on irradiation technology of RI production (STC No.2II) (Joint research)

Kimura, Akihiro; Izumo, Hironobu; Tsuchiya, Kunihiko; Hori, Naohiko; Ishihara, Masahiro; Bannykh, V.*; Gluschenko, N.*; Chakrova, Y.*; Chakrov, P.*

JAEA-Testing 2010-002, 20 Pages, 2010/08

JAEA-Testing-2010-002.pdf:2.92MB

JMTR has a plan to produce Mo, which is the parent nuclide of radiopharmaceutical Tc, by (n,) method. The cooperation experiments for Mo adsorption and Tc elution with the Poly-Zirconium Compound (PZC) and the Molybdate Zirconium Gel (Zr-gel) methods were carried out at Kazakhstan National Nuclear Energy Center (NNC) in October, 2009. The Mo adsorption capability was the same level as reference data, however the Tc elution capability with PZC was lower than reference data in this test. Therefore, re-experiments of Mo adsorption and Tc elution with both methods were carried out at NNC. As a result, the Mo adsorption and Tc elution capabilities were obtained as the same levels as reference data. Additionally, Tc solution was high purity by the elution method connected with alumina column.

Tritium generation in lithium ceramics LiTiO for fusion reactor blanket

Tazhibayeva, I. L.*; Kenzhin, E. A.*; Kulsartov, T. V.*; Kuykabayeva, A. A.*; Shestakov, V.*; Chikhray, E.*; Gizatulin, S.*; Maksimkin, O. P.*; Beckman, I. N.*; Kawamura, Hiroshi; et al.

Questions of Atomic Science and Technology, 2, p.3 - 11, 2008/00

Lithium titanate (LiTiO) was chosen as a tentative reference material from viewpoints of good tritium recovery at low temperatures and of low tritium inventory and chemical stability for the breeding blanket in fusion reactors. The results of the irradiation tests of LiTiO in the WWR-K of NNC-RK are described in this paper. 96at% Li-enriched LiTiO pebbles and disks were prepared as the irradiation specimens and these specimens were irradiated during 220 days (5350 hours) at the reactor power of 6 MWt. Tritium release was measured continuously during irradiation tests and tritium release properties were evaluated. The mechanics describing generation and release of tritium from the irradiated LiTiO were analyzed. There was estimated tritium loss due to recoil energy and binding of tritium in HTO, and there was calculated stationary tritium release due to diffusion under constant temperature and under thermal cycling.

The Result of a wall failure in-pile experiment under the EAGLE project

Konishi, Kensuke; Toyooka, Junichi; Kamiyama, Kenji; Sato, Ikken; Kubo, Shigenobu*; Kotake, Shoji*; Koyama, Kazuya*; Vurim, A. D.*; Gaidaichuk, V. A.*; Pakhnits, A. V.*; et al.

Nuclear Engineering and Design, 237(22), p.2165 - 2174, 2007/11

https://doi.org/10.1016/j.nucengdes.2007.03.012

The WF (Wall Failure) test of the EAGLE program, in which 2kg of uranium dioxide fuel-pins were melted by nuclear heating, was successfully conducted in the IGR of NNC/Kazakhstan. In this test, a 3mm-thick stainless steel (SS) wall structure was placed between fuel pins and a 10mm-thick sodium-filled channel (sodium gap). During the transient, fuel pins were heated, which led to the formation of a fuel-steel mixture pool. Under the transient nuclear heating condition, the SS wall was strongly heated by the molten pool, leading to wall failure. The time needed for fuel penetration into the sodium-filled gap was very short (less than 1 second after the pool formation). The result suggests that molten core materials formed in hypothetical LMFBR core disruptive accidents have a certain potential to destroy SS-wall boundaries early in the accident phase, thereby providing fuel escape paths from the core region. The early establishment of such fuel escape paths is regarded as a favorable characteristic in eliminating the possibility of severe re-criticality events.

The Result of medium scale in-pile experiment conducted under the EAGLE-project

Konishi, Kensuke; Toyooka, Junichi; Kamiyama, Kenji; Sato, Ikken; Kubo, Shigenobu*; Kotake, Shoji*; Koyama, Kazuya*; Vurim, A. D.*; Gaidaichuk, V. A.*; Pakhnits, A. V.*; et al.

Proceedings of Technical Meeting on Severe Accident and Accident Management (CD-ROM), 16 Pages, 2006/03

no abstracts in English