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Journal Articles

Development of fast reactor containment safety analysis code, CONTAIN-LMR, 1; Outline of development project

Miyahara, Shinya; Seino, Hiroshi; Ohno, Shuji; Konishi, Kensuke

Proceedings of 23rd International Conference on Nuclear Engineering (ICONE-23) (DVD-ROM), 5 Pages, 2015/05

A CONTAIN-LMR code has been developed in JAEA for application to PRA of LMFRs since the original CONTAIN code had been introduced from SNL of U.S. in 1982. The code is a best-estimate, integrated analysis tool for predicting the physical, chemical and radiological conditions inside a containment building of LMFRs following a severe accident with reactor vessel melt-through. The code is also able to predict the source term to the environment in the accident. This code can treat many important phenomena consistently such as sodium fire, radioactive aerosol behavior, hydrogen burn, sodium-concrete reaction and core debris-concrete interaction occurred in the accident with inter-cell heat and mass flow under the multiple cell geometry. This paper describes the chronology of the code development in JAEA briefly as an introduction, and after that, the outline of computational models in the code, the examples of the code validation, and the future plan of the code application are described.

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 studies on the upward fuel discharge for elimination of severe recriticality during core-disruptive accidents in sodium-cooled fast reactors

Kamiyama, Kenji; Konishi, Kensuke; Sato, Ikken; Toyooka, Junichi; Matsuba, Kenichi; Zuyev, V. A.*; Pakhnits, A. V.*; Vityuk, V. A.*; Vurim, A. D.*; Gaidaichuk, V. A.*; et al.

Journal of Nuclear Science and Technology, 51(9), p.1114 - 1124, 2014/09

AA2013-0469.pdf:1.18MB

 Times Cited Count:5 Percentile:43.1(Nuclear Science & Technology)

Journal Articles

Experimental study on fuel-discharge behavior through in-core coolant channels

Kamiyama, Kenji; Saito, Masaki*; Matsuba, Kenichi; Isozaki, Mikio; Sato, Ikken; Konishi, Kensuke; Zuyev, V. A.*; Kolodeshnikov, A. A.*; Vassiliev, Y. S.*

Journal of Nuclear Science and Technology, 50(6), p.629 - 644, 2013/06

 Times Cited Count:11 Percentile:19.82(Nuclear Science & Technology)

In core disruptive accidents of sodium cooled fast reactors, fuel discharge from the core region reduces the possibility of severe re-criticality events. In-core coolant channels such as the control-rod guide tube and a concept of the FAIDUS (Fuel Assembly with Inner Duct Structure) provide effective fuel discharge paths if effects of sodium in these paths on molten fuel discharge are limited. Two series of experiments conducted in the present study showed that the discharge path can be entirely voided by the vaporization of a part of the coolant at the initial melt discharge phase, that this is followed by coolant vapor expansion, and that melt penetrates significantly into the voided channel. In conclusion, the effects of the sodium on fuel discharge are limited and therefore in-core coolant channels provide effective fuel discharge paths for reducing neutronic activity.

Journal Articles

Experimental studies on upward fuel discharge during core disruptive accident in sodium-cooled fast reactors

Kamiyama, Kenji; Konishi, Kensuke; Sato, Ikken; Toyooka, Junichi; Matsuba, Kenichi; Zuyev, V. A.*; Pakhnits, A. V.*; Vurim, A. D.*; Gaidaichuk, V. A.*; Kolodeshnikov, A. A.*; et al.

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

In order to eliminate energetics potential in the case of postulated core disruptive accidents (CDAs) of sodium-cooled fast reactors, introduction of a fuel subassembly with an inner duct structure has been considered. Recently, a design option which leads molten fuel to discharge upward is considered to minimize developmental efforts for the fuel subassembly fabrication. In this paper, a series of out-of-pile tests and one in-pile test were presented. The out-of-pile tests were conducted to investigate the effects of driving pressures on upward discharge, and the in-pile test was conducted to demonstrate a sequence of upward discharge behavior of molten-fuel. Based on these experimental results, it is concluded that the most of molten-fuel is expected to complete discharging upward before core melting progression, and thereby, introduction of the fuel subassembly with the upward discharge duct has the sufficient potential to eliminate energetics events.

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.

Journal Articles

Safety strategy of JSFR eliminating severe recriticality events and establishing in-vessel retention in the core disruptive accident

Sato, Ikken; Tobita, Yoshiharu; Konishi, Kensuke; Kamiyama, Kenji; Toyooka, Junichi; Nakai, Ryodai; Kubo, Shigenobu*; Kotake, Shoji*; Koyama, Kazuya*; Vassiliev, Y. S.*; et al.

Journal of Nuclear Science and Technology, 48(4), p.556 - 566, 2011/03

In the JSFR design, elimination of severe recriticality events in the Core Disruptive Accident (CDA) is intended as an effective measure to assure retention of the core materials within the reactor vessel. The design strategy is to control the potential of excessive void reactivity insertion in the Initiating Phase selecting appropriate design parameters such as maximum void reactivity on one hand, and to exclude core-wide molten-fuel-pool formation, which has been the main issue of CDA, with introduction of Inner Duct on the other hand. The effectiveness of these measures are reviewed based on existing experimental data and evaluations performed with validated analysis tools. It is judged that the present JSFR design can exlude severe power burst events.

Journal Articles

SIMMER-III analysis of eagle-1 in-pile tests focusing on heat transfer from molten core material to steel-wall structure

Toyooka, Junichi; Kamiyama, Kenji; Konishi, Kensuke*; Tobita, Yoshiharu; Sato, Ikken

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

In this study, the heat flux from the molten core materials to the outer surface of the inner duct (the pool-to-duct heat flux) was evaluated based on all the EAGLE-1 in-pile experiments available. Through the evaluation, it was understood that the pool-to-duct heat flux was so high in all the in-pile experiments that the duct wall failed without coolant boiling in its behind. It was also indicated that the presence of steel in the molten core mixture played a key role in this high heat flux. Application of the SIMMER-III code for these tests suggested that some model improvements were necessary to simulate pool-to-duct heat transfer behavior in a mechanistic way.

JAEA Reports

Fundamental study on discharging of molten core material through the in-core coolant channel

Kamiyama, Kenji; Isozaki, Mikio; Imahori, Shinji; Konishi, Kensuke; Matsuba, Kenichi; Sato, Ikken

JAEA-Research 2008-059, 33 Pages, 2008/07

JAEA-Research-2008-059.pdf:10.82MB

In CDA of LMFBR, molten core materials would discharge from the core region through the coolant paths. Rapid vaporization of the coolant by mixing of the molten core materials provides effective evacuation of the liquid coolant from the path and reduces significantly possibility of core-material freezing and blockage formation inside the paths. This characteristic enhances early discharge of molten-core materials and reduces possibility of severe re-criticality events. In this study, melt discharge experiments were conducted with a coolant channel simulating the discharge path with an enhanced length of the path compared with that of the realistic design structure. An alloy and water were used as simulant of the molten fuel and sodium respectively. This series of experiments showed that the discharge path can be entirely voided by vaporization of a part of the coolant at the initial melt discharge phase, followed by vapor expansion toward the end of the coolant channel. Furthermore, it was revealed that the condition where coolant void expansion started can be defined by melt-coolant sensible heats ratio and the heated height of the coolant. The heat balance evaluation during the coolant void expansion phase shows that the film condensation heat transfer should be considered. The coolant-void-expansion behavior in the discharge path of the realistic design condition was estimated based on an application of this knowledge to existing experiments with molten oxide and sodium.

Journal Articles

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

 Times Cited Count:33 Percentile:7.15(Nuclear Science & Technology)

The WF (Wall Failure) test of the EAGLE program, in which $$sim$$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.

Journal Articles

The Eagle project to eliminate the recriticality issue of fast reactors; Progress and results of in-pile tests

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

Proceedings of 5th Korea-Japan Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-5), p.465 - 471, 2006/11

no abstracts in English

Journal Articles

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

JAEA Reports

Study on Numerical Simulation for Thermal-Hydraulic Phenomena of Multiphase, Multicomponent Flows; Transient Vaporization/Condensation Phenomena in Multicomponent System,1

Morita, Koji*; Matsumoto, Tatsuya*; Fukuda, Kenji*; Tobita, Yoshiharu; Yamano, Hidemasa; Konishi, Kensuke; Sato, Ikkenn

JNC-TY9400 2003-011, 56 Pages, 2003/04

JNC-TY9400-2003-011.pdf:2.31MB

It is one of the important problems for more reliable reactor safety evaluation to improve numerical simulation techniques for involved thermal-hydraulic phenomena of multiphase, multicomponent flows in core disruptive accidents.In the present joint research, physical model development and experimental investigation were conducted for transient condensation phenomena of a vapor bubble with noncondensable gases to improve applicability of a fast-reactor safety analysis code for the phase-transition phenomena in multicomponent systems.In this fiscal year, preliminary experiments using noncondensable gas were performed for the transient bubble condensation phenomena, and then basic data were obtained for large-scale bubble behavior without condensation.In addition, a multiple-scale flow-regime model treating large-scale bubbles was newly proposed for the fast-reactor safety analysis code and applied to the analysis of the preliminary experiments successfully.

Oral presentation

EAGLE project; Experimental study on elimination of the Re-criticality issue during CDAs, 12; The Prompt result of in-pile large scale dry test

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

no journal, , 

no abstracts in English

Oral presentation

EAGLE project; Experimental study on elimination of the Re-criticality issue during CDAs, 13; Results of the sodium test in the out-of-pile program

Kamiyama, Kenji; Konishi, Kensuke; Sato, Ikken; Kubo, Shigenobu*; Kotake, Shoji*; Shimakawa, Yoshio*; Koyama, Kazuya*; Zuyev, V.*; Vassiliev, Y. S.*; Kolodeshnikov, A.*

no journal, , 

no abstracts in English

Oral presentation

EAGLE project; Experimental study on elimination of the Re-criticality issue during CDAs, 14; The Result and interpretation of in-pile middle scale test

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

no journal, , 

no abstracts in English

Oral presentation

EAGLE project: Experimental study on elimination of re-criticality issue during CDAs, 15; The Result of the first in-pile integral test

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.

no journal, , 

no abstracts in English

Oral presentation

EAGLE-project: Experimental study on elimination of re-criticality issue during CDAs, 18; The Result of the second in-pile integral test

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

no journal, , 

The EAGLE experimental program is dedicated to show experimental evidences supporting a safety logic eliminating the recriticality issue in the core disruptive accidents (CDAs) of sodium-cooled fast breeder reactors. In order to confirm an inherent nature of early fuel escape from the core region, both in-pile (using IGR) and out-of-pile experiments have been performed in the program. This presentation shows the preliminary interpretation of the second integral experiment, in which fuel discharge through a duct-type escape path (initially filled with sodium) was investigated using about 8kg of molten fuel. Energy insertion in this second experiment was smaller than that in the first experiment. The duct-wall failure timing was a little delayed compared with that in the first experiment, and the fuel discharged through the duct intermittently.

Oral presentation

EAGLE-project: Experimental study on elimination of re-criticality issue during CDAs, 19; Identification of fuel movement using neutron detectors

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

no journal, , 

The EAGLE experimental program is dedicated to show experimental evidences supporting a safety logic eliminating the recriticality issue in the core disruptive accidents (CDAs) of sodium-cooled fast breeder reactors. Two kinds of neutron detectors (one was placed in the test section and another was placed around IGR driver core for power monitoring use) were analyzed to get prospect that the data include information of molten-fuel motion in the in-pile test.

Oral presentation

Overview on the EAGLE experimental program aiming at resolution of the re-criticality issue for the fast reactors

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

no journal, , 

In the EAGLE program, several in-pile and out-of-pile tests have been conducted by August 2006, under a co-operation between JAEA and NNC/RK. The main objectives of these tests are; (1) to demonstrate effectiveness of special design concepts to eliminate the re-criticality issue in the course of CDAs of SFRs, and (2) to acquire basic information on early-phase relocation of molten-core materials toward cold regions surrounding the core, which would be applicable to various core design concepts. As the final step of this program, integral in-pile tests simulating realistic accident conditions were conducted. Geometry of the test apparatus adopted in these tests is corresponding to a typical special design concept equipped with a "discharge duct" within each fuel sub-assembly. In these tests, fuel-steel mixture pool was successfully realized and discharge of the pool materials through the duct was observed.

33 (Records 1-20 displayed on this page)