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論文

Features of a control blade degradation observed ${it in situ}$ during severe accidents in boiling water reactors

Pshenichnikov, A.; 山崎 宰春; Bottomley, D.; 永江 勇二; 倉田 正輝

Journal of Nuclear Science and Technology, 56(5), p.440 - 453, 2019/05

In the present paper new results using ${it in situ}$ video, are presented regarding BWR control blade degradation up to 1750 K at the beginning of a nuclear severe accident. Energy-dispersive X-ray spectrometry (EDS) mapping indicated stratification of the absorber blade melt with formation of a chromium and boride-enriched layer. High content-B- and C-containing material with increased melting temperature acted like a shielding and was found to prevent further relocation of control blade claddings. The interacted layers around the B$$_{4}$$C granules prevented direct steam attack of residual B$$_{4}$$C. The results provide new insights for understanding of the absorber blade degradation mechanism under reducing conditions specific to Fukushima Dai-Ichi Unit 2 resulting from prolonged steam starvation.

論文

High temperature oxidation test of simulated BWR fuel bundle in steam-starved conditions

山崎 宰春; Pshenichnikov, A.; Pham, V. H.; 永江 勇二; 倉田 正輝; 徳島 二之*; 青見 雅樹*; 坂本 寛*

Proceedings of Annual Topical Meeting on Reactor Fuel Performance (TopFuel 2018) (Internet), 8 Pages, 2018/10

燃料集合体の酸化及び水素吸収はその後の事故進展挙動に影響を与えることから、PWR燃料集合体では、実効的な水蒸気流量としてg-H$$_{2}$$O/sec/rodという単位が導入されており、事故進展評価の重要なパラメータといて用いられている。一方BWRにおいては、燃料集合体の構成がPWRとは異なることにより、PWRで用いられている規格化された水蒸気流量ではチャンネルボックスの内外での酸化及び水素吸収の差が正確に評価できない。そのため、PWRで用いられているg-H$$_{2}$$O/sec/rodという規格化された水蒸気流量に代わる、適切な評価パラメータがBWRでも必要である。そこで、ジルカロイの水蒸気枯渇条件での酸化及び水素吸収データを取得するため、実機を模擬したBWRバンドル試験体を用いて高温酸化試験を行なった。BWRにおける水蒸気流量を規格化するため、水蒸気流路断面積を考慮したパラメータを検討した。

口頭

Specifications of CLADS test facilities for fuel degradation simulation

Pshenichnikov, A.; 山崎 宰春; 永江 勇二; 倉田 正輝

no journal, , 

Japan Atomic Energy Agency (JAEA) has to establish the scientific basis for the decommissioning of the Boiled Water Reactors (BWR) at Fukushima Daiichi Nuclear Power Station and provide the information on the nuclear fuel behavior during the nuclear accidents. To address these problems JAEA has developed several test facilities which combine unique features for the R&D related to Severe Accident (SA) propagation behavior for BWRs. The technical data and main features of the facilities will be introduced to the International audience to give an understanding of JAEA research abilities and possible collaboration in the severe accidents research.

口頭

CLADS test facilities on BWR core material degradation

永江 勇二; 柴田 裕樹; Pshenichnikov, A.; 山崎 宰春; 倉田 正輝

no journal, , 

Two control blade degradation test facilities which were designed for the investigation of high temperature degradation of the nuclear reactor elements, have been set up in Tomioka-site. One was a test facility for small-scale samples and another one was test facility for large-scale samples. In situ video of the experiment and mass-spectrometric analysis of the output gases can be carried out during tests, and heating rate can be controlled up to 1 K/s. The working area of the large-scale facility consists of the isolated heating area 1.2 m height with the water cooled lower part to achieve the natural temperature gradient conditions as during severe accident. Steam rate in large-scale facilities can be controlled in the range of 15-100 g/min, and a test piece for the control blade degradation tests can be used to simulate a part of the control blade and fuel assembly. Chemical reaction between stainless steel/B$$_{4}$$C and Zry is affected by the oxide layer on Zry. The oxidation behavior depends on steel flow rate and tendency of control blade degradation possibly has threshold condition. According to pretest, critical condition for reaction between molten control rod and channel box can be recognized. Further analysis and tests in various steam rate will be conducted to evaluate the critical condition.

口頭

Focusing the test strategy of CLADS-MADE-01 and first results of the BWR control blade degradation test

Pshenichnikov, A.; 山崎 宰春; 永江 勇二; 倉田 正輝

no journal, , 

Current presentation gives a deep analysis of the progression of the 1F accident at Unit 2. Based on that analysis the explanation of the strategy for a severe accident test program of JAEA-CLADS to contribute to understanding of accident progression at 1F Unit 2 is given. First result of CLADS-MADE-01 (BWR control blade degradation test) is presented.

口頭

Consideration for modeling of Zry oxidation and hydrogen uptake in steam-starved conditions

山崎 宰春; Pshenichnikov, A.; Pham, V. H.; 永江 勇二; 倉田 正輝

no journal, , 

燃料集合体の酸化及び水素吸収はその後の事故進展挙動に影響を与えることから、PWR燃料集合体では、実効的な水蒸気流量としてg-H$$_{2}$$O/sec/rodという1rodあたりの単位が導入されており、事故進展評価の重要なパラメータとして用いられている。一方BWRにおいては、燃料集合体の構成がPWRとは異なることにより、PWRで用いられている1rodあたりの水蒸気流量ではチャンネルボックスの内外での酸化及び水素吸収の差が正確に評価できない。そのため、PWRで用いられている規格化された水蒸気流量に代わる、適切な評価パラメータがBWRでも必要である。そこで、ジルカロイの水蒸気枯渇条件での酸化及び水素吸収データを取得するため、4本の燃料棒をチャンネルボックスで囲んだ試験体を用いて水蒸気枯渇条件にて高温酸化試験を行なった。BWRにおける水蒸気流量を規格化するため、水蒸気流路断面積を考慮したパラメータを検討した。

口頭

シビアアクシデント時の燃料破損・溶融過程解析手法の高度化,1-4; ジルカロイ酸化/水素化モデルの整備

山崎 宰春; Pshenichnikov, A.; 永江 勇二; 倉田 正輝; 坂本 寛*; 徳島 二之*; 青見 雅樹*

no journal, , 

本研究では、被覆管の水蒸気が枯渇した条件での被覆管の酸化、及び水素吸蔵現象をモデル化することを目的とし、異なる2つの水蒸気供給速度による酸化/水素吸蔵挙動の違いに係るデータを取得した。これまで被覆管1本での酸化/水素吸蔵挙動を把握し、モデル化を進めているが、BWRではチャンネルボックスの酸化をモデルに考慮する必要があると考えられる。そこで、被覆管4本の周囲にチャンネルボックスを配置した試験体を用いて、1300$$^{circ}$$Cで酸化/水素吸蔵挙動試験を実施した。水蒸気供給量が少ない場合、試験体チャンネルボックス内上部にて酸化膜が減少するとともに水素吸蔵量が増加し、被覆管1本での試験で認められた酸化/水素吸蔵挙動と同じ傾向が見られた。

口頭

Status and first results of the BWR control blade degradation test under steam-starved conditions

Pshenichnikov, A.; 山崎 宰春; 永江 勇二; 倉田 正輝

no journal, , 

The present study is focused on the issue of a control blade degradation, which is the first step during BWR severe accidents. In particular problem of heterogeneous core degradation, boron behaviour, control blade liquefaction and melt relocation were analysed. Melt relocation features: onset, progress, candling, blockage formation and its subsequent melting was directly observed in situ using video cameras. Carefully tailored experimental conditions of the CLADS-MADE-01 test, which include preliminary oxidation, transient heating and long steam starvation phase, enabled successful reproducing of an accident scenario close to one that assumed to happen at Fukushima-1 Unit 2.

口頭

The Behaviour of materials in case of solidified absorber melt - oxidized BWR channel box interaction revealed after CLADS-MADE-01 test

Pshenichnikov, A.; 山崎 宰春; 永江 勇二; 倉田 正輝

no journal, , 

The paper summarizes the first results of a thorough SEM investigation uncovering the process of channel box wall penetration by Fe-Cr-Ni-B containing melt. The preliminary oxidation of channel box is shown to play an important role on severe accident progression resulted in the suppression of channel box massive destruction. Only one small droplet came out to the other side of channel box. The mechanism of local beginning of oxide layer destruction with subsequent Zircaloy-4 channel box penetration is under discussion.

口頭

Characterization of the Fukushima Unit-2 sediments / debris based on the on-site video investigations in comparison to the debris obtained after integral CLADS-MADE-01 test

Pshenichnikov, A.

no journal, , 

The published results on the debris characterization obtained after CLADS-MADE-01 control blade degradation test will be compared to debris inside the PCV of 1F unit 2 recently investigated by TEPCO using new developed robot.

口頭

Analysis of the video data from Fukushima Dai-ichi Unit 2 pedestal debris inspection in comparison to the CLADS-MADE-01 debris

Pshenichnikov, A.; 倉田 正輝; 永江 勇二; 山崎 宰春

no journal, , 

The new data from video investigation of the 1F Unit 2 pedestal debris performed by TEPCO was analysed. The debris features as derived from visual appearance on the video compared with the debris obtained after the CLADS-MADE-01 test. Some speculative conclusions concerning the properties and possible nature of the debris can be made.

口頭

Boron behaviour issue during control blade degradation at the beginning phase of postulated scenario for 1F Unit 2 accident reproduced in the CLADS MADE 01 test

Pshenichnikov, A.; 倉田 正輝; 永江 勇二; 山崎 宰春

no journal, , 

The problem of control blade degradation which involves boron carbide in the eutectic reaction is one of the key issues for understanding of the accident progression at the beginning phase of the severe accident. In the 1f Unit 2 postulated integral test conditions the interaction of B with the hot steam was found to be rather difficult because of melting and encapsulation of the B materials in the debris. Thus the amount of B release for BWRs in Japan can be significantly lower. The presentation discusses the possibility of unpredictable B accumulation in some unpredictable places in the reactor core.

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