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

https://doi.org/10.1007/s10973-019-08045-7

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)/CaCO 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 compounds.

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.

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 (BC) 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 BC-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.

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

https://doi.org/10.1299/transjsme.17-00382

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.

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.

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 NaO). Based on the corrosive data, authors quantitatively evaluated the influence factors of wastage and formulated the average corrosive equations.

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.

Progress of thermal hydraulic evaluation methods and experimental studies on a sodium-cooled fast reactor and its safety in Japan

Kamide, Hideki; Ohshima, Hiroyuki; Sakai, Takaaki; Tanaka, Masaaki

Nuclear Engineering and Design, 312, p.30 - 41, 2017/02

https://doi.org/10.1016/j.nucengdes.2016.09.026

In the framework of the Generation-IV International Forum, the safety design criteria (SDC) incorporating safety-related Research and Development results on innovative technologies and lessons learned from Fukushima Dai-ichi Nuclear Power Plants accident has been established to provide the set of general criteria for the safety designs of structures, systems and components of Generation-IV Sodium-cooled Fast Reactors (Gen-IV SFRs). A number of thermal-hydraulic evaluations are necessary to meet the concept of the criteria in the design studies of Gen-IV SFRs. This paper focuses on four kinds of thermal-hydraulic issues associated with the SDC, i.e., fuel subassembly thermal-hydraulics, natural circulation decay heat removal, core disruptive accidents, and thermal striping. Progress of evaluation methods on these issues is shown with activities on verification and validation (V and V) and experimental studies towards commercialization of SFR in Japan. These evaluation methods are planned to be eventually integrated into a comprehensive numerical simulation system that can be applied to all possible phenomena in SFR systems and that can be expected to become an effective tool for the development of human resource and the handing our knowledge and technologies down.

An Empirical correlation to predict the distance for fragmentation of simulated Molten-Core materials discharged into a sodium pool

Matsuba, Kenichi; Isozaki, Mikio; Kamiyama, Kenji; Suzuki, Toru; Tobita, Yoshiharu

Proceedings of 11th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, Operation and Safety (NUTHOS-11) (USB Flash Drive), 8 Pages, 2016/10

In order to evaluate the distance for fragmentation of molten core material discharged into the lower sodium plenum during core disruptive accidents in sodium-cooled fast reactors, experiments with simulated molten materials and coolants (water, sodium) was carried out, where an empirical correlation of the distance for fragmentation was developed. The empirical correlation developed by this study showed a good agreement with the measurement results obtained by the present experiments. It was found that in order to well-predict the distance for fragmentation in sodium, thermal phenomena, such as sodium boiling and resultant vapor expansion, needed to be considered.

Distance for fragmentation of a simulated molten-core material discharged into a sodium pool

Matsuba, Kenichi; Isozaki, Mikio; Kamiyama, Kenji; Tobita, Yoshiharu

Journal of Nuclear Science and Technology, 53(5), p.707 - 712, 2016/05

https://doi.org/10.1080/00223131.2015.1113897

In order to develop an evaluation method of the distance for fragmentation of molten core material discharged into the sodium plenum, a sodium experiment with visual observation was conducted using an X-ray imaging system. In the current experiments, 0.9 kg of molten aluminum (initial temperature: around 1473 K) was discharged into a sodium pool (initial temperature: 673 K) through a nozzle (inner diameter: 20 mm). Based on the experimental results, the distance for fragmentation of the liquid column was estimated to be 100 mm in the experiments. Through the sodium experiment, useful knowledge was obtained for the future development of an evaluation method of the distance for fragmentation of molten core material. As a next step, sodium experiments using higher-density molten materials will be conducted to enrich the experimental knowledge. Besides, a new semi-empirical correlation will be developed to evaluate more appropriately the distance for fragmentation under CDA conditions.