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Journal Articles

Kinetic study on eutectic melting process between boron carbide and stainless steel in sodium-cooled fast reactor

Kikuchi, Shin; Sakamoto, Kan*; Takai, Toshihide; Yamano, Hidemasa

Nippon Kikai Gakkai 2020-Nendo Nenji Taikai Koen Rombunshu (Internet), 4 Pages, 2020/09

In a postulated severe accidental condition of sodium-cooled fast reactor (SFR), eutectic melting between boron carbide (B$$_{4}$$C) as control rod element and stainless steel (SS) as control rod cladding or related structure may occur. Thus, behavior of B$$_{4}$$C-SS eutectic melting is one of the phenomena to evaluate the core disruptive accidents in SFR. In order to clarify the kinetic feature of B$$_{4}$$C-SS eutectic melting process in the interface, the thinning test for SS crucibles using the pellets of B$$_{4}$$C or SS with low B$$_{4}$$C concentration were performed to obtain the rate constant with dependence of B$$_{4}$$C concentration against SS. It was found that the rate constants of eutectic melting between SS and SS low B$$_{4}$$C concentration were smaller than that of B$$_{4}$$C-SS in the high temperature range. Besides, the rate constant of eutectic melting between SS and B$$_{4}$$C containing SS became small when decreasing the B$$_{4}$$C concentration against SS.

Journal Articles

Post-test material analysis of eutectic melting reaction of boron carbide and stainless steel

Yamano, Hidemasa; Takai, Toshihide; Furukawa, Tomohiro

Nippon Kikai Gakkai Rombunshu (Internet), 86(883), p.19-00360_1 - 19-00360_13, 2020/03

It is necessary to simulate a eutectic melting reaction and relocation behavior of boron carbide (B$$_{4}$$C) as a control rod material and stainless steel (SS) during a core disruptive accident in an advanced sodium-cooled fast reactor designed in Japan because the B$$_{4}$$C-SS eutectic relocation behavior has a large uncertainty in the reactivity history based on a simple calculation. A physical model simulating the eutectic melting reaction and relocation was developed and implemented into a severe accident simulation code. The developed model must be validated by using test data. To validate the physical model, therefore, the visualization tests of SS-B$$_{4}$$C eutectic melting reaction was carried out by contacting SS melts of several kg with a B$$_{4}$$C pellet heated up to about 1500 $$^{circ}$$C. The tests have shown the eutectic reaction visualization as well as freezing and relocation of the B$$_{4}$$C-SS eutectic in upper part of the solidified test piece due to the density separation. Post-test material analyses by using X-ray diffraction and transmission electron microscope techniques have indicated that FeB appeared at the B$$_{4}$$C-SS contact interface and (Fe,Cr)$$_{2}$$B at the top surface of the test piece. Glow discharge optical emission spectrometry has been applied to quantitative analysis of boron concentration distributions. The boron concentration was high at the upper surface and near the original position of the B$$_{4}$$C pellet.

Journal Articles

Study on eutectic melting behavior of control rod materials in core disruptive accidents of sodium-cooled fast reactors, 1; Project overview

Yamano, Hidemasa; Takai, Toshihide; Furukawa, Tomohiro; Kikuchi, Shin; Emura, Yuki; Kamiyama, Kenji; Fukuyama, Hiroyuki*; Higashi, Hideo*; Nishi, Tsuyoshi*; Ota, Hiromichi*; et al.

Proceedings of International Nuclear Fuel Cycle Conference / Light Water Reactor Fuel Performance Conference (Global/Top Fuel 2019) (USB Flash Drive), p.418 - 427, 2019/09

Eutectic reactions between boron carbide (B$$_{4}$$C) and stainless steel (SS) as well as its relocation are one of the key issues in a core disruptive accident (CDA) evaluation in sodium-cooled fast reactors. Since such behaviors have never been simulated in CDA numerical analyses, it is necessary to develop a physical model and incorporate the model into the CDA analysis code. This study is focusing on B$$_{4}$$C-SS eutectic melting experiments, thermophysical property measurement of the eutectic melt, and physical model development for the eutectic melting reaction. The eutectic experiments involve the visualization experiments, eutectic reaction rate experiments and material analyses. The thermophysical properties are measured in the range from solid to liquid state. The physical model is developed for a severe accident computer code based on the measured data of the eutectic reaction rate and the physical properties. This paper describes the project overview and progress of experimental and analytical studies by 2017. Specific results in this paper is boron concentration distributions of solidified B$$_{4}$$C-SS eutectic sample in the eutectic melting experiments, which would be used for the validation of the eutectic physical model implemented into the computer code.

Journal Articles

Study on eutectic melting behavior of control rod materials in core disruptive accidents of sodium-cooled fast reactors, 3; Effect of B$$_{4}$$C addition on thermophysical properties of austenitic stainless steel in a liquid state

Fukuyama, Hiroyuki*; Higashi, Hideo*; Yamano, Hidemasa

Proceedings of International Nuclear Fuel Cycle Conference / Light Water Reactor Fuel Performance Conference (Global/Top Fuel 2019) (USB Flash Drive), p.853 - 857, 2019/09

Thermophysical properties of molten mixture of 316L stainless steel (SS316L) and control-rod material (B$$_{4}$$C) are necessary for the development of computer simulation codes that describe core degradation mechanisms during severe accidents in nuclear power plants involving sodium-cooled fast reactors. The effect of B$$_{4}$$C addition to SS316L on the solidus and liquidus temperatures were first measured by differential scanning calorimetry. An electromagnetic levitation technique performed in a static magnetic field was used to measure the density, surface tension, normal spectral emissivity, specific heat capacity, and thermal conductivity of molten SS316L and SS316L containing B$$_{4}$$C. The effects of B$$_{4}$$C addition to SS316L on the thermophysical properties were studied up to 10 mass%.

Journal Articles

Study on eutectic melting behavior of control rod materials in core disruptive accidents of sodium-cooled fast reactors, 5; Validation of a multi-phase model for eutectic reaction between molten stainless steel and B$$_{4}$$C

Liu, X.*; Morita, Koji*; Yamano, Hidemasa

Proceedings of International Nuclear Fuel Cycle Conference / Light Water Reactor Fuel Performance Conference (Global/Top Fuel 2019) (USB Flash Drive), p.47 - 51, 2019/09

Investigation of the eutectic reaction in a core disruptive accident of sodium cooled reactor is of importance since reactor criticality will be affected by the change in reactivity after eutectic reaction. In this study, we performed 1st step of validation analysis using a fast reactor safety analysis code, SIMMER-III, with the developed model based on a new series of experiments, where a B$$_{4}$$C pellet was immersed into a molten stainless steel (SS) pool. The simulation results showed the general behavior of eutectic material formation measured in the experiments reasonably. The eutectic reaction consumes solid B$$_{4}$$C and liquid SS, and then the liquid eutectic composition is produced at the early stage of reaction due to the high temperature of molten SS. Movement of the eutectic material in the molten pool leads to the redistribution of boron element. Molten SS pool then freezes to solid SS and movement of eutectic material is stopped by surrounding solid SS. Boron concentration in the pool was measured after molten SS freezes into a solid. Simulation results indicate that boron tends to accumulate in the upper part of the molten pool. This is attributed to the buoyancy force acting on lighter boron in the molten SS pool. A parametric study was also conducted by changing the initial temperature of B$$_{4}$$C pellet and SS to investigate the temperature sensitivity on the eutectic reaction behavior.

Journal Articles

Numerical analysis of core disruptive accident in a metal-fueled sodium-cooled fast reactor

Yamano, Hidemasa; Tobita, Yoshiharu

Proceedings of 11th Korea-Japan Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-11) (Internet), 3 Pages, 2018/11

Based on the event tree analysis, the present numerical analyses investigated the capability of fuel discharge through the one-dimensional single fuel assembly geometry and the two-dimensional geometry of a CRGT channel with neighboring fuel assemblies. The single fuel assembly analyses showed that the fuel blockage formed in the lower shielding region because fuel solidified by contacting with cold sodium in case of no fission gas release. On the assumption that fission gas was released, the molten fuel successfully relocated below the core. The next analyses using the CRGT channel indicated a significant fuel discharge through the CRGT channel. This is because the fuel temperature was still high just after the CRGT wall failure and sodium in the CRGT channel was quickly voided just after the ingress of a small amount of molten fuel.

Journal Articles

Validation of three-dimensional finite-volume-particle method for simulation of liquid-liquid mixing flow behavior

Kato, Masatsugu*; Funakoshi, Kanji*; Liu, X.*; Matsumoto, Tatsuya*; Liu, W.*; Morita, Koji*; Kamiyama, Kenji

Proceedings of 11th Korea-Japan Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-11) (Internet), 7 Pages, 2018/11

Journal Articles

Sedimentation behavior of mixed solid particles

Sheikh, Md. A. R.*; Son, E.*; Kamiyama, Motoki*; Morioka, Toru*; Matsumoto, Tatsuya*; Morita, Koji*; Matsuba, Kenichi; Kamiyama, Kenji; Suzuki, Toru*

Journal of Nuclear Science and Technology, 55(6), p.623 - 633, 2018/06

 Times Cited Count:4 Percentile:24.89(Nuclear Science & Technology)

During core-disruptive accidents in sodium-cooled fast reactors, the sedimentation behavior of fragmented debris is crucial for in-vessel retention. The height of the beds formed may influence both the cooling of the bed and the neutronic characteristics. To develop an experimental database of bed formation behavior, a series of experiments using simulant materials, namely, Al$$_{2}$$O$$_{3}$$, ZrO$$_{2}$$, and stainless steel, were performed under gravity-driven discharge of solid particles from a nozzle into a quiescent cylindrical water pool. The bed height was measured for particles of different size, density, and sphericity, and an injection nozzle with varying diameter, injection velocity, and injection height. From these experiments, an empirical correlation was established to predict the bed height for both homogeneous and mixed particles for the different properties. This correlation reproduces reasonably well the experimental trend in bed height.

Journal Articles

Thermophysical properties of molten stainless steel containing 5mass%-B$$_{4}$$C

Fukuyama, Hiroyuki*; Higashi, Hideo*; Yamano, Hidemasa

Proceedings of 2018 International Congress on Advances in Nuclear Power Plants (ICAPP 2018) (CD-ROM), p.1014 - 1019, 2018/04

In this study, densities, surface tensions, normal spectral emissivities, heat capacities and thermal conductivities of molten SUS316L and SUS316L containing 5mass%-B$$_{4}$$C were measured by the electromagnetic levitation technique in a static magnetic field.

Journal Articles

Event sequence analysis of core disruptive accident in a metal-fueled sodium-cooled fast reactor

Yamano, Hidemasa; Tobita, Yoshiharu; Kubo, Shigenobu; Ueda, Nobuyuki*

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

In this study, the event sequence analysis of CDA in a large metal-fueled SFR has been performed in order to investigate reactivity progression and molten fuel relocation behavior in the metal-fueled SFR. The initiating phase analysis during the CDA initiated by unprotected loss-of-flow accidents has been conducted using the CANIS code, which showed a small power peak. Using the initial conditions based on the initiating phase analysis, the SIMMER-III code was applied to a whole-core scale analyses to clarify the event sequence including the reactivity progression and molten fuel relocation. As a result, recriticality in the whole core analysis resulted in a very mild energy release. The mild energy release in the metal-fueled core can be attributed to the small specific heat of metal fuel and the large prompt negative reactivity feedback mechanism.

Journal Articles

Improvements to the simmer code model for steel wall failure based on EAGLE-1 test results

Toyooka, Junichi; Kamiyama, Kenji; Tobita, Yoshiharu; Suzuki, Toru

Proceedings of 10th Japan-Korea Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-10) (USB Flash Drive), 7 Pages, 2016/11

Journal Articles

Experimental database for bed formation behaviors of solid particles

Sheikh, M. A. R.*; Son, E.*; Kamiyama, Motoki*; Morioka, Toru*; Matsumoto, Tatsuya*; Morita, Koji*; Matsuba, Kenichi; Kamiyama, Kenji; Suzuki, Toru

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

During the material relocation phase of core disruptive accidents in sodium-cooled fast reactors, the sedimentation behavior of fragmented debris leading to the formation of debris beds is crucial for in-vessel retention by debris bed cooling. In this study, a series of experiments using simulant materials was performed to develop an experimental database of bed formation behavior by gravity driven discharge of solid particles from a nozzle into a quiescent cylindrical water pool. The bed height as well as the bed shape was measured. Three types of spherical and non-spherical particles, namely Al$$_{2}$$O$$_{3}$$, ZrO$$_{2}$$ and stainless steel with different size were employed to study the effect of key experimental parameter on debris bed mound shape. Based on the experimental results, an empirical correlation as experimental database was proposed to predict the particle bed height. The proposed correlation reasonably reproduces the experimental trend of the bed height variation on the crucial factors. This result demonstrates a wide applicability of the proposed empirical model to predict the bed height in terms of all crucial factors with reasonable accuracy.

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

Development of the evaluation methodology for the material relocation behavior in the core disruptive accident of sodium-cooled fast reactors

Tobita, Yoshiharu; Kamiyama, Kenji; Tagami, Hirotaka; Matsuba, Kenichi; Suzuki, Toru; Isozaki, Mikio; Yamano, Hidemasa; Morita, Koji*; Guo, L.*; Zhang, B.*

Journal of Nuclear Science and Technology, 53(5), p.698 - 706, 2016/05

AA2015-0794.pdf:2.46MB

 Times Cited Count:6 Percentile:29.32(Nuclear Science & Technology)

The in-vessel retention (IVR) of core disruptive accident (CDA) is of prime importance in enhancing safety characteristics of sodium-cooled fast reactors (SFRs). In the CDA of SFRs, molten core material relocates to the lower plenum of reactor vessel and may impose significant thermal load on the structures, resulting in the melt through of the reactor vessel. In order to enable the assessment of this relocation process and prove that IVR of core material is the most probable consequence of the CDA in SFRs, a research program to develop the evaluation methodology for the material relocation behavior in the CDA of SFRs has been conducted. This program consists of three developmental studies, namely the development of the analysis method of molten material discharge from the core region, the development of evaluation methodology of molten material penetration into sodium pool, and the development of the simulation tool of debris bed behavior.

JAEA Reports

Phase Change Predictions for Liquid Fuel in Contact with Steel Structure using the Heat Conduction Equation

Brear, D. J.

PNC-TN9410 98-005, 53 Pages, 1998/01

PNC-TN9410-98-005.pdf:2.09MB

When liquid fuel makes contact with steel structure the liquid can freeze as a crust and the structure can melt at the surface. The melting and freezing processes that occur can influence the mode of fuel freezing and hence fuel relocation. Furthermore the temperature gradients established in the fuel and steel phases determine the rate at which heat is transferred from fuel to steel. In this memo the 1-D transient heat conduction equations are applied to the case of initially liquid UO$$_{2}$$ brought into contact with solid steel using up-to-date materials properties. The solutions predict criteria for fuel crust formation and steel melting and provide a simple algorithm to determine the interface temperature when one or both of the materials is undergoing phase change. The predicted steel melting criterion is compared with available experimental results.

JAEA Reports

Trending analysis of precursor events

Watanabe, Norio

JAERI-Tech 97-067, 55 Pages, 1998/01

JAERI-Tech-97-067.pdf:2.57MB

no abstracts in English

Journal Articles

Development of the EWS version of the severe accident analysis code THALES-2

Muramatsu, Ken; *; *; *; *

Dai-6-Kai Kakuritsuronteki Anzen Hyoka (PSA) Ni Kansuru Kokunai Shimpojiumu Rombunshu (IAE-9206), p.171 - 175, 1993/01

no abstracts in English

Journal Articles

Analysis of uncertainty factors on source terms for severe accidents

*; Muramatsu, Ken

Dai-6-Kai Kakuritsuronteki Anzen Hyoka (PSA) Ni Kansuru Kokunai Shimpojiumu Rombunshu (IAE-9206), p.25 - 33, 1993/01

no abstracts in English

Journal Articles

Source term evaluation for small break LOCA sequences at a PWR

Muramatsu, Ken; *

Dai-6-Kai Kakuritsuronteki Anzen Hyoka (PSA) Ni Kansuru Kokunai Shimpojiumu Rombunshu (IAE-9206), p.3 - 8, 1993/01

no abstracts in English

Journal Articles

Current status of probabilistic safety assessment

Muramatsu, Ken; Homma, Toshimitsu

Hoken Butsuri, 28, p.355 - 362, 1993/00

no abstracts in English

63 (Records 1-20 displayed on this page)