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Journal Articles

Experiments of self-wastage phenomena elucidation in steam generator tube of sodium-cooled fast reactor

Umeda, Ryota; Shimoyama, Kazuhito; Kurihara, Akikazu

Nippon Genshiryoku Gakkai Wabun Rombunshi, 19(4), p.234 - 244, 2020/12

Sodium-water reaction caused by failure of the steam generator tube of sodium-cooled fast reactor induce the wastage phenomenon, which has erosive and corrosive feature. In this report, the authors have performed the self-wastage experiments under high sodium temperature condition to evaluate the effect of wastage form/geometry by using two types of initial defect such as the micro fine pinhole and fatigue crack, and water leak rate on self-wastage rate. Based on the consideration of crack type influence, it was confirmed that self-wastage rate did not strongly depend on the initial defect geometry. As a mechanism of the self-plug phenomenon, it is speculated that sodium oxide intervenes and inhibits the progress of self-wastage. The dependence of initial sodium temperature on self-wastage rate was clearly observed, and new self-wastage correlation was derived considering the initial sodium temperature.

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

Now is the time of fast reactor

Negishi, Hitoshi; Kamide, Hideki; Maeda, Seiichiro; Nakamura, Hirofumi; Abe, Tomoyuki

Nippon Genshiryoku Gakkai-Shi, 62(8), p.438 - 441, 2020/08

Prototype Fast Breeder Reactor, Monju, was under decommission since April, 2018. It is the first time for Japan to make a sodium cooled reactor into decommission. It is significant work and will take 30 years. The Monju has provided wide spectrum and huge amount of findings and knowledge, e.g., design, R&D, manufacturing, construction, and operation up to 40% of full power over 50 years of development history. It is significant to utilize such findings and knowledge for the development and commercialization of a fast rector in Japan.

Journal Articles

Study on eutectic melting behavior of control rod materials in core disruptive accidents of sodium-cooled fast reactors, 3; Kinetic study of boron carbide-stainless steel eutectic melting by differential thermal analysis

Kikuchi, Shin; Yamano, Hidemasa; Nakamura, Kinya*

Proceedings of 2020 International Conference on Nuclear Engineering (ICONE 2020) (Internet), 9 Pages, 2020/08

In a postulated severe accidental condition of sodium-cooled fast reactor (SFR), eutectic melting between boron carbide (B$$_{4}$$C) and stainless steel (SS) may take place. Thus, kinetic behavior of B$$_{4}$$C-SS eutectic melting is one of the important phenomena to be considered when evaluating the core disruptive accidents in SFR. In this study, for the first step to obtain the fundamental information on kinetic feature of B$$_{4}$$C-SS eutectic melting, the thermal analysis using the pellet type samples of B$$_{4}$$C and Type 316L SS as different experimental technique was performed. The differential thermal analysis endothermic peaks for the B$$_{4}$$C-SS eutectic melting appeared from 1483K to 1534K and systematically shifted to higher temperatures when increasing heating rate. Based on this kinetic feature, apparent activation energy and pre-exponential factor for the B$$_{4}$$C-SS eutectic melting were determined by Kissinger method. It was found that the kinetic parameters obtained by thermal analysis were comparable to the literature values.

JAEA Reports

Phase 1 code assessment of SIMMER-III; A Computer program for LMFR core disruptive accident analysis

Kondo, Satoru; Tobita, Yoshiharu

JAEA-Research 2019-009, 382 Pages, 2020/03

JAEA-Research-2019-009.pdf:22.82MB

The SIMMER-III computer code, developed at the Japan Atomic Energy Agency (JAEA, the former Power Reactor and Nuclear Fuel Development Corporation), is a two-dimensional, multi-velocity-field, multi-component fluid-dynamics code, coupled with a space- and time-dependent neutron kinetics model. The code is being used widely for simulating complex phenomena during core-disruptive accidents (CDAs) in liquid-metal fast reactors (LMFRs). In parallel to the code development, a comprehensive assessment program was performed in two phases: Phase 1 for verifying individual fluid-dynamics models; and Phase 2 for validating its applicability to integral phenomena important to evaluating LMFR CDAs. The SIMMERIII assessment program was participated by European research and development organizations, and the achievement of Phase 1 was compiled and synthesized in 1996. This report has been edited by revising and reproducing the original 1996 informal report, which compiled the achievement of Phase 1 assessment. A total of 34 test problems were studied in the areas: fluid convection, interfacial area and momentum exchange, heat transfer, melting and freezing, and vaporization and condensation. The problems identified have been reflected to the Phase 2 assessment and later model development and improvement. Although the revisions were made in the light of knowledge base obtained later, the original individual contributions by the participants, both positive and negative, are retained except for editorial changes.

Journal Articles

Comparative study on the thermal behavior of structural concretes of sodium-cooled fast reactor

Kikuchi, Shin; Koga, Nobuyoshi*; Yamazaki, Atsushi*

Journal of Thermal Analysis and Calorimetry, 137(4), p.1211 - 1224, 2019/08

 Times Cited Count:2 Percentile:58.41(Thermodynamics)

In this study, two siliceous concretes with similar specification as structural concretes of SFR were selected for the comparative study of the thermal behavior. The thermal behavior of the structural concretes was investigated in a temperature range from room temperature to 1900 K using TG-differential thermal analysis (DTA) and other supplementary techniques. The softening and melting of the concretes initiated in the thermal decomposition product of the cement portion in the temperature range 1400-1600 K. Because the compositional difference between the cement portion of two different siliceous concretes was characterized by different Ca(OH)$$_{2}$$/CaCO$$_{3}$$ ratios, the melting temperature ranges of those thermal decomposition products are not so significantly different. On the other hand, the melting of the aggregate is directly influenced by the initial composition of SiO$$_{2}$$ compounds.

Journal Articles

Development of a fast reactor and related thermal hydraulics studies in Japan

Ohshima, Hiroyuki; Kamide, Hideki

Proceedings of 18th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-18) (USB Flash Drive), p.2095 - 2107, 2019/08

Development of a sodium-cooled fast reactor has been implemented in Japan from the viewpoint of severe accident countermeasures. This paper describes the progress of research and development related to safety enhancement and the severe accident countermeasures. A volcanic PRA methodology was developed for the proper consideration of external hazards. Water and sodium experiments were carried out for the decay heat removal in a core disruptive accident (CDA), and also thermal hydraulic interactions between the core and upper plenum where dipped heat exchanger was operated. In order to elucidate the behavior of molten fuel during CDA, basic experiments of core melt fragmentation in deep and shallow sodium pools were carried out. X-ray visualization showed the liquid column of molten steel was intensively fragmented nearly simultaneously with a rapid expansion of sodium vapor.

JAEA Reports

Prototype fast breeder reactor Monju; Its history and achievements

Tsuruga Comprehensive Research and Development Center

JAEA-Technology 2019-007, 159 Pages, 2019/07

JAEA-Technology-2019-007.pdf:19.09MB
JAEA-Technology-2019-007-high-resolution1.pdf:42.36MB
JAEA-Technology-2019-007-high-resolution2.pdf:33.56MB
JAEA-Technology-2019-007-high-resolution3.pdf:38.14MB
JAEA-Technology-2019-007-high-resolution4.pdf:48.82MB
JAEA-Technology-2019-007-high-resolution5.pdf:37.61MB

This report summarizes the history and achievements of the prototype fast breeder reactor Monju. The development of Monju started in 1968 as a prototype reactor following the experimental fast reactor Joyo. The development covers all the activity related to the fast reactor; plant design, mockup tests, construction, operation, and plant management. This report summarizes the history and achievements for 11 technical areas: history and principal achievements, design and construction, operation test, plant safety, core physics, fuel, plant system, sodium technology, materials and mechanical design, plant management, and trouble management.

Journal Articles

A Proposal of inelastic constitutive equations of lead alloys used for structural tests simulating severe accident conditions

Hashidate, Ryuta; Onizawa, Takashi; Wakai, Takashi; Kasahara, Naoto*

Proceedings of 2019 ASME Pressure Vessels and Piping Conference (PVP 2019) (Internet), 10 Pages, 2019/07

Under the severe accident conditions, structural materials of nuclear power plants are subjected to excessive high temperature. Although it is very essential to clarify how the structure collapses under the severe accident conditions, the failure mechanisms in such high temperatures are not clarified. However, it is very difficult and expensive to perform structural tests using actual structural materials. Therefore, we propose to use lead alloys instead of actual structural materials. Because the strength of lead alloys is much poorer than that of the actual structural materials, failure can be observed at low temperature and by small load. For demonstration of analogy between the failure mechanisms of lead alloys structure at low temperature and those of the actual structures at high temperature, numerical analyses are required. So, we confirm the material characteristics of lead alloys and develop inelastic constitutive equations of lead alloy required for finite element analyses.

Journal Articles

Three-dimensional numerical study on pool stratification behavior in molten corium-concrete interaction (MCCI) with MPS method

Li, X.; Sato, Ikken; Yamaji, Akifumi*; Duan, G.*

Proceedings of 26th International Conference on Nuclear Engineering (ICONE-26) (Internet), 8 Pages, 2018/07

Molten corium-concrete interaction (MCCI) is an important ex-vessel phenomenon that could happen during the late phase of a hypothetical severe accident in a light water reactor. In the present study, a three-dimensional (3-D) numerical study has been performed to simulate COMET-L3 test carried out by KIT with a stratified molten pool configuration of simulant materials with improved MPS method. The heat transfer between corium/crust/concrete was modeled with heat conduction between particles. Moreover, the potential influence of the siliceous aggregates was also investigated by setting up two different case studies since there was previous study indicating that siliceous aggregates in siliceous concrete might contribute to different axial and radial concrete ablation rates. The simulation results have indicated that metal melt as corium in MCCI can have completely different characteristics regarding concrete ablation pattern from that of oxidic corium, which needs to be taken into consideration when assessing the containment melt-through time in severe accident management.

Journal Articles

The Development of method for extinguishing / treating metallic sodium using calcium chloride

Abe, Yuta; Nagai, Keiichi; Maie, Mitsuyoshi*; Nakano, Natsuko*; Kawashima, Yuichi*; Takesue, Naohisa*; Saito, Junichi

Dai-23-Kai Doryoku, Enerugi Gijutsu Shimpojiumu Koen Rombunshu (USB Flash Drive), 5 Pages, 2018/06

no abstracts in English

Journal Articles

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

Kikuchi, Shin; Yamano, Hidemasa

Dai-23-Kai Doryoku, Enerugi Gijutsu Shimpojiumu Koen Rombunshu (USB Flash Drive), 4 Pages, 2018/06

In a postulated severe accidental condition of sodium-cooled fast reactor (SFR), eutectic reaction between boron carbide (B$$_{4}$$C) and stainless steel (SS) may probably occur. Elucidation on the behavior of cited eutectic reaction is very important in terms of evaluation of core disruptive accidents in SFRs. For the first step to clarify the kinetic feature of B$$_{4}$$C-SS eutectic reaction, the preliminary thermogavimetry-differential thermal analysis (TG-DTA) measurements using individual reagent were performed to obtain the fundamental information and to confirm the applicability of sample crucibles. It was found that alumina crucible was applicable in terms of eutectic behavior. Based on the DTA curves at different heating rates, the kinetic parameters were roughly estimated by using Kissinger method.

Journal Articles

Application of nontransfer type plasma heating technology for core-material-relocation tests in boiling water reactor severe accident conditions

Abe, Yuta; Sato, Ikken; Nakagiri, Toshio; Ishimi, Akihiro; Nagae, Yuji

Journal of Nuclear Engineering and Radiation Science, 4(2), p.020901_1 - 020901_8, 2018/04

A new experimental program using non-transfer type plasma heating is under consideration in JAEA to clarify the uncertainty on core-material relocation (CMR) behavior of BWR. In order to confirm the applicability of this new technology, authors performed preparatory plasma heating tests using small-scale test pieces (107 mm $$times$$ 107 mm $$times$$ 222 mm (height)). An excellent perspective in terms of applicability of the non-transfer plasma heating to melting high melting-temperature materials such as ZrO$$_{2}$$ has been obtained. In addition, molten pool was formed at the middle height of the test piece indicating its capability to simulate the initial phase of core degradation behavior consistent with the real UO$$_{2}$$ fuel Phebus-FPT tests. Furthermore, application of EPMA, SEM/EDX and X-ray CT led to a conclusion that the pool formed consisted mainly of Zr with some concentration of oxygen which tended to be enhanced at the upper surface region of the pool. Based on these results, an excellent perspective in terms of applicability of the non-transfer plasma heating technology to the Severe Accident (SA) experimental study was obtained.

Journal Articles

Evaluation of target-wastage in consideration of sodium-water reaction environment formed on the periphery of an adjacent tube in steam generator of sodium-cooled fast reactor

Kurihara, Akikazu; Umeda, Ryota; Shimoyama, Kazuhito; Kikuchi, Shin

Nippon Kikai Gakkai Rombunshu (Internet), 84(859), p.17-00382_1 - 17-00382_11, 2018/03

Wastage on adjacent tubes (target-wastage) arise from water/steam leak in steam generators of sodium-cooled fast reactors (sodium-water reaction). Target-wastage is likely to be caused by liquid droplet impingement erosion (LDI) and Na-Fe composite oxidation type corrosion with flow (COCF) in an environment marked by high temperature and high-alkali (reaction jet) due to sodium-water reaction. In the previous study, the authors quantitatively evaluated the effect of material temperature and fluid velocity on COCF rate, and revealed that COCF was sodium-iron composite oxidation type corrosion from metallographic observation and element assay. In this study, the applicability of new wastage correlations was confirmed for each tube in sodium-water reaction test with straight vertical tube bundle under practical steam generator operation condition. The authors established that the new wastage correlations were applicable to each tube of tube bundle in the above test, and the time progress of wastage was qualitatively investigated for the two penetrated tubes in the period including the water and/or steam blowdown.

Journal Articles

Thermal behavior of sodium hydroxide-structural concrete composition of sodium-cooled fast reactor

Kikuchi, Shin; Koga, Nobuyoshi*

Journal of Thermal Analysis and Calorimetry, 131(1), p.301 - 308, 2018/01

 Times Cited Count:2 Percentile:76.27(Thermodynamics)

Under postulated accidental condition of sodium-cooled fast reactor (SFR), liquid sodium spill into the floor may lead to fail the steel liner resulting in sodium-concrete reaction (SCR). In this study, the sodium hydroxide (NaOH)-silica (SiO$$_{2}$$) reaction as one of possible secondary stage reactions was investigated for safety assessment of SFR. Thermal behavior of NaOH-SiO$$_{2}$$ reaction such as reaction onset was determined using a differential scanning calorimetry (DSC). As a result of DSC measurement, it was revealed that NaOH-SiO$$_{2}$$ reaction occurs as rapid reaction right after NaOH melting at 583 K. Therefore, it was expected that NaOH-SiO$$_{2}$$ reaction is dominant in the time frame of secondary stage of SCR if significant amount of NaOH has been generated during the initial stage reaction.

JAEA Reports

Cutting operation of simulated fuel assembly heating examination by AWJ

Abe, Yuta; Nakagiri, Toshio; Watatani, Satoshi*; Maruyama, Shinichiro*

JAEA-Technology 2017-023, 46 Pages, 2017/10

JAEA-Technology-2017-023.pdf:8.01MB

This is a report on Abrasive Water Jet (AWJ) cutting work carried out on specimen, which was used for Simulated Fuel Assembly Heating Examination by Collaborative Laboratories for Advanced Decommissioning Science (CLADS) molten core behavior analysis group in February 2016. The simulated fuel assembly is composed of Zirconia for the outer crucible/simulated fuel, stainless steel for the control blade and Zircaloy (Zr) for the cladding tube/channel box. Therefore, it is necessary to cut at once substances having a wide range of fracture toughness and hardness. Moreover, it is a large specimen with an approximate size of 300 mm. In addition, epoxy resin has high stickiness, making it more difficult to cut. Considering these effects, AWJ cutting was selected. The following two points were devised, and this specimen could be cut with AWJ. If it was not possible to cut at one time like a molten portion of boride, it was repeatedly cut. By using Abrasive Suspension Jet (ASJ) system with higher cutting ability than Abrasive Injection Jet (AIJ, conventional method) system, cutting time was shortened. As a result of this work, the cutting method in Simulated Fuel Assembly Heating Examination was established. Incidentally, in the cutting operation, when the cutting ability was lost at the tip of the AWJ, a curved cut surface, which occurs when the jet flowed away from the feeding direction, could be confirmed at the center of the test body. From the next work, to improve the cutting efficiency, we propose adding a mechanism such as turning the cutting member itself for re-cutting from the exit side of the jet and appropriate traverse speed to protect cut surface.

Journal Articles

Evaluation and demonstration of cutting the fuel assembly heating examination by AWJ

Maruyama, Shinichiro*; Watatani, Satoshi*

Mitsui Sumitomo Kensetsu Gijutsu Kenkyu Kaihatsu Hokoku, (15), p.107 - 112, 2017/10

It is essential to estimate characteristics and forms of fuel debris for safe and reliable removing at the decommissioning of the Fukushima Daiichi Nuclear Power Plant (1F). For the estimation, melting behavior of fuel assembly in the accident is being researched. To proceed the research, the fuel debris were need to cut, and the abrasive water jet (AWJ) which had enough results for cutting ceramic material or mixed material of zirconium alloy and stainless. The test results demonstrated that AWJ could cut the fuel assembly and accumulated the cutting data which will be subservient when removing the fuel debris in future.

JAEA Reports

Phenomenon elucidation experiment for target wastage caused in steam generator of sodium-cooled fast reactor; Corrosion experiment in flowing high-temperature sodium hydroxide environment

Umeda, Ryota; Shimoyama, Kazuhito; Kurihara, Akikazu

JAEA-Technology 2017-018, 70 Pages, 2017/08

JAEA-Technology-2017-018.pdf:9.67MB

In case of the water leak into sodium in a SG of SFRs due to tube failure, reaction jet is formed by sodium-water reaction with exothermic heat. The reaction jet forms highly alkaline environment with high temperature and high pressure, which cause local thinning of adjacent heat transfer tubes (target wastage). In this report, for the purpose of elucidation of target wastage, the authors developed the experimental apparatus and experimental technique which enable the separate evaluation of wastage influence factors, including temperature, impingement velocity, reagent ratio and so on by using high temperature sodium hydroxide as major reaction product and sodium monoxide as secondary reaction product. In addition, the impingement corrosion experiments have been conducted by using high temperature reagents (NaOH and Na$$_{2}$$O). Based on the corrosive data, authors quantitatively evaluated the influence factors of wastage and formulated the average corrosive equations.

Journal Articles

Progress of design and related researches of sodium-cooled fast reactor in Japan

Kamide, Hideki; Sakamoto, Yoshihiko; Kubo, Shigenobu; Oki, Shigeo; Ohshima, Hiroyuki; Kamiyama, Kenji

Proceedings of International Conference on Fast Reactors and Related Fuel Cycles; Next Generation Nuclear Systems for Sustainable Development (FR-17) (USB Flash Drive), 10 Pages, 2017/06

Development of a sodium-cooled fast reactor has been implemented in Japan from the viewpoint of severe accident countermeasures in order to strengthen safety of a fast reactor since the Great East Japan Earthquake. This paper describes the progress of design study and research and development related to safety enhancement and the severe accident countermeasures. For the purpose of strengthening of decay heat removal function, several researches have been carried out on the decay heat removal in a core disruptive accident (CDA), diversity and applicability of decay heat removal systems, and thermal hydraulic evaluation methods. In order to elucidate the behavior of molten fuel during CDA, some in-pile and out-of-pile tests has been performed by international collaboration including basic experiments. Core design was also improved from the viewpoint of preventing the occurrence of severe accident.

Journal Articles

SAS4A analysis study on the initiating phase of ATWS events for generation-IV loop-type SFR

Kubota, Ryuzaburo; Koyama, Kazuya*; Moriwaki, Hiroyuki*; Yamada, Yumi*; Shimakawa, Yoshio*; Suzuki, Toru; Kawada, Kenichi; Kubo, Shigenobu; Yamano, Hidemasa

Proceedings of 2017 International Congress on Advances in Nuclear Power Plants (ICAPP 2017) (CD-ROM), 10 Pages, 2017/04

This paper describes an analysis study on the initiating phase of the ATWS events with SAS4A in order to confirm the appropriateness of the core design for the medium-scale SFR (750MWe-1765MWt). Not using a conventional lumping method that multiple fuel sub-assemblies having a similar characteristic were assigned to one channel (representing fuel assembly in SAS4A), each channel represents only the sub-assemblies of identical operating condition. In addition, the detailed power and reactivity distribution were set reflecting the change of insertion position of control rods. Applying these detailed analysis conditions, the SAS4A analyses were performed for unprotected loss-of-flow (ULOF) and unprotected transient overpower (UTOP) during both of the nominal power and the partial power operation. As a result, more proper event progression including incoherency of events especially fuel dispersion after fuel failure was successfully evaluated and then this analysis study suggested that the power excursion with prompt criticality leading to large mechanical energy release can be prevented in the initiating phase of the current design.

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