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Journal Articles

Experimental study on the coolability of molten core materials discharged into a depth- and volume-limited sodium plenum

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

Proceedings of 31st International Conference on Nuclear Engineering (ICONE31) (Internet), 7 Pages, 2024/11

During a severe accident in sodium-cooled fast reactors, molten core materials could be discharged from the core region toward the lower sodium region of the reactor vessel through coolant channels, such as control rod guide tubes. Typical SFRs have a sodium plenum with limited depth and volume, such as the core inlet plenum located under the core region. Therefore, it is important to evaluate the coolability of molten core materials discharged into a depth- and volume-limited sodium plenum. In the present study, to deepen the understanding on the coolability of molten core materials discharged into such a sodium plenum, conditions under which molten core materials form solidified fragments were discussed based on an experiment discharging a molten fuel simulant (molten Al2O3) into a test vessel filled with liquid sodium.

Journal Articles

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.

Journal Articles

A Status of experimental program to achieve in-vessel retention during core disruptive accidents of sodium-cooled fast reactors

Kamiyama, Kenji; Matsuba, Kenichi; Kato, Shinya; Imaizumi, Yuya; Mukhamedov, N.*; Akayev, A.*; Pakhnits, A.*; Vurim, A.*; Baklanov, V.*

Proceedings of International Conference on Fast Reactors and Related Fuel Cycles; Sustainable Clean Energy for the Future (FR22) (Internet), 9 Pages, 2022/04

Journal Articles

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

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.

Journal Articles

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.

Journal Articles

Results of an out-of-pile experiment for fragmentation of a simulated molten core material discharged into a shallow sodium pool

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

Proceedings of 11th Korea-Japan Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-11) (Internet), 4 Pages, 2018/11

In Core Disruptive Accidents of Sodium-cooled Fast Reactors, molten core material would be discharged through control rod guide tubes into the inlet coolant plenums beneath the rector cores. The inlet coolant plenums have quite limited heights and sodium inventories. Therefore, in the inlet plenums, molten core material with a jet-like shape would impinge on the bottom of the plenum before it breaks up into fragments. In this study, to clarify fragmentation behavior in a shallow sodium pool whose height and volume are so limited that jet impingement on the bottom is expected, an out-of-pile experiment discharging molten alumina into a sodium pool was conducted. Although a small amount of alumina agglomeration was found on the bottom plate (steel disk) installed in the sodium pool, most of the molten alumina was fragmented into debris particles. Results obtained in the present experiment suggest that molten core material is fragmented and quenched even in a shallow sodium pool.

Journal Articles

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 (Al$$_{2}$$O$$_{3}$$) were analyzed. The results clarified that while Al$$_{2}$$O$$_{3}$$ 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.

Journal Articles

A Recent experimental program to evidence in-vessel retention by controlled material relocation during core disruptive accidents of sodium-cooled fast reactors

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

Proceedings of 10th Japan-Korea Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-10) (USB Flash Drive), 5 Pages, 2016/11

Molten fuel discharge through control rod guide tubes (CRGTs) is a key process that dominates the termination of core disruptive accidents of sodium-cooled fast reactors, since fuel dispersion from the core contributes to the achievement of both deeper subcriticality in the degraded core and formation of coolable debris bed. Within a framework of a collaborative research program between Japan Atomic Energy Agency and National Nuclear Center of the Republic of Kazakhstan, called EAGLE program, a new experimental program has been started with out-of-pile experiments to clarify the fuel discharge through CRGTs. This paper presents the status of the new program, including experimental results obtained so far.

Journal Articles

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

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.

Journal Articles

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

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

Proceedings of 23rd International Conference on Nuclear Engineering (ICONE-23) (DVD-ROM), 6 Pages, 2015/05

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 0.4 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 the tendency of measured mass median diameters (i.e., non-dependence on Weber number) suggests that before hydrodynamic instabilities sufficiently grow to induce fragmentation, thermal phenomena such as local coolant vaporization and resultant vapor expansion accelerate fragmentation.

Journal Articles

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

Journal Articles

Experimental study on material relocation during core disruptive accident in sodium-cooled fast reactors; Results of a series of fragmentation tests for molten oxide discharged into a sodium pool

Matsuba, Kenichi; Kamiyama, Kenji; Konishi, Kensuke; Toyooka, Junichi; Sato, Ikken; Zuev, V. A.*; Kolodeshnikov, A. A.*; Vasilyev, Y. S.*

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

A series of fragmentation tests (FR tests) for molten oxide was conducted to obtain experimental knowledge on the distance for fragmentation of molten core material discharged into the lower sodium plenum. Approx. 7$$sim$$14 kg of molten alumina was discharged into a sodium pool (depth: 1.3 m, diameter: 0.4 m, temperature: approx. 673 K) through a duct (inner diameter: 40$$sim$$63 mm). The test results showed that the molten alumina was fragmented into particles within 1 m from the surface of the sodium pool. The estimated distances for fragmentation in the FR tests were roughly 60$$sim$$70% lower than the predictions by the existing representative correlation. The experimental knowledge confirms the possibility that the distance for fragmentation of the molten core material can be significantly reduced due to thermal interactions in the lower sodium plenum.

JAEA Reports

Retesting of $$^{99}$$Mo adsorption and $$^{rm 99m}$$Tc elution characteristics of molybdenum adsorbents for (n,$$gamma$$) method; Joint experiment report on irradiation technology of RI production (STC No.2$$-$$II) (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 $$^{99}$$Mo, which is the parent nuclide of radiopharmaceutical $$^{rm rm 99m}$$Tc, by (n,$$gamma$$) method. The cooperation experiments for $$^{99}$$Mo adsorption and $$^{rm 99m}$$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 $$^{99}$$Mo adsorption capability was the same level as reference data, however the $$^{rm 99m}$$Tc elution capability with PZC was lower than reference data in this test. Therefore, re-experiments of $$^{99}$$Mo adsorption and $$^{rm 99m}$$Tc elution with both methods were carried out at NNC. As a result, the $$^{99}$$Mo adsorption and $$^{rm 99m}$$Tc elution capabilities were obtained as the same levels as reference data. Additionally, $$^{rm 99m}$$Tc solution was high purity by the elution method connected with alumina column.

JAEA Reports

$$^{99}$$Mo adsorption and $$^{rm 99m}$$Tc elution characteristics of molybdenum adsorbents for (n,$$gamma$$); Method-joint experiment report on irradiation technology of RI production (STC No. 2-II) (Joint research)

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

JAEA-Technology 2009-075, 23 Pages, 2010/02

JAEA-Technology-2009-075.pdf:7.41MB

Japan Materials Testing Reactor (JMTR) of the Japan Atomic Energy Agency (JAEA) has a plan to produce $$^{99}$$Mo, which is the parent nuclide of radiopharmaceutical $$^{rm 99m}$$Tc, by (n,$$gamma$$) method. The $$^{99}$$Mo adsorption and $$^{rm 99m}$$Tc elution characteristics of molybdenum adsorbents should be evaluated since the specific activity of $$^{99}$$Mo obtained by (n,$$gamma$$) method is low. Therefore, $$^{99}$$Mo adsorption and $$^{rm 99m}$$Tc elution tests with molybdenum adsorbents for the (n,$$gamma$$) method such as poly-zirconium compound (PZC) and molybdate zirconium gel were carried out under cooperation with the Kazakhstan National Nuclear Energy Center (NNC). As a result, the $$^{99}$$Mo adsorption performance of the adsorbents was the same level as conventional data, whereas the $$^{rm 99m}$$Tc elution performance of the adsorbents was lower than conventional data. The $$^{99}$$Mo adsorption and $$^{rm 99m}$$Tc elution performance will be investigated again in future.

Journal Articles

Tritium generation in lithium ceramics Li$$_{2}$$TiO$$_{3}$$ 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 (Li$$_{2}$$TiO$$_{3}$$) 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 Li$$_{2}$$TiO$$_{3}$$ in the WWR-K of NNC-RK are described in this paper. 96at% $$^{6}$$Li-enriched Li$$_{2}$$TiO$$_{3}$$ 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 Li$$_{2}$$TiO$$_{3}$$ 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.

Oral presentation

Experimental studies on discharge of molten-core materials during core disruptive accidents for sodium-cooled fast reactors; Consideration on discharge behavior of molten-core materials based on results of post-test investigations

Kamiyama, Kenji; Matsuba, Kenichi; Tobita, Yoshiharu; Toyooka, Junichi; Pakhnits, A. V.*; Vityuk, V. A.*; Kukushkin, I.*; Vurim, A. D.*; Baklanov, V. V.*; Kolodeshnikov, A. A.*

no journal, , 

no abstracts in English

Oral presentation

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

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

no journal, , 

To clarify the molten material outflow behavior from the reactor core region through the control rod guide tube(CRGT) in the core disruptive accident of SFR, the out-of-pile tests simulating the molten material dropping into the CRGT, they are imitated by the molten alumina and the stainless cup respectively, were conducted. In this presentation, the evaluation result of the tests will be reported.

Oral presentation

Study on discharge behavior of molten core materials in core on core disruptive accidents of sodium cooled fast reactors; Consideration on discharge behavior through a sodium-filled channel with an internal structure

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

no journal, , 

In order to enrich experimental knowledge of discharge behavior of molten core materials through a sodium-filled channel in core disruptive accidents of sodium cooled fast reactors, an out-of-pile experiment was conducted, in which molten alumina was used as a molten-fuel simulant and it penetrated into a sodium-filled channel with an internal structure reducing a flow area. In this presentation, consideration on effects of the internal structure on melt discharge-behavior will be presented based on experimental results.

Oral presentation

Experimental study on fragmentation behavior of molten core material during core disruptive accident for sodium-cooled fast reactors; Discussion on the fragmentation mechanism

Matsuba, Kenichi; Kamiyama, Kenji; Tobita, Yoshiharu; Toyooka, Junichi; Zuev, V.*; Kolodeshnikov, A.*; Vasilyev, Y.*

no journal, , 

In order to obtain experimental knowledge on fragmentation behavior of molten core material discharged into the lower sodium plenum in the reactor vessel during core disruptive accidents of sodium cooled fast reactors, a series of fragmentation experiments have been carried out with a molten oxide. Based on the experimental results, dominant mechanisms for the fragmentation behavior was discussed.

Oral presentation

Experimental studies on discharge of molten-core materials during core disruptive accidents for sodium-cooled fast reactors; Results of post-test investigations on the in-pile test devices

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

no journal, , 

no abstracts in English

29 (Records 1-20 displayed on this page)