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論文

Modeling of the P2M past fuel melting experiments with the FEMAXI-8 code

Mohamad, A. B.; 宇田川 豊

Nuclear Technology, 210(2), p.245 - 260, 2024/02

 被引用回数:2 パーセンタイル:57.39(Nuclear Science & Technology)

In the Power to Melt and Maneuverability (P2M) project, a simulation exercise on two past power ramp experiments xM3 on medium burn-up rod and HBC4 on high burn-up rod were performed with the fuel performance code FEMAXI-8 to investigate the fuel behavior under high power and high-temperature conditions toward centerline fuel melting. In order to treat fuel melting, empirical melting temperature models have been incorporated into the FEMAXI-8 code. The present analysis gave reasonable predictions not only on cladding deformation but also on the fuel melting behavior of the HBC4 rod, in which the UO$$_{2}$$ liquidus temperature was reached during the transient. On the other hand, model improvement appeared to be needed for a more accurate treatment of fuel melting behavior of the xM3 rod, in which fuel center temperature reached solidus line, whereas may not reached liquidus line. A reasonable agreement of estimated FGR with the measurement suggested that the high temperature FGR at the given conditions are essentially temperature dependent phenomenon: rate-limited primarily by thermally activated elementary processes such as fission gas diffusion.

論文

Microstructural evolution of intermetallic phase precipitates in Cr-coated zirconium alloy cladding in high-temperature steam oxidation up to 1400$$^{circ}$$C

Mohamad, A. B.; 根本 義之; 古本 健一郎*; 岡田 裕史*; 佐藤 大樹*

Corrosion Science, 224, p.111540_1 - 111540_15, 2023/11

 被引用回数:2 パーセンタイル:33.88(Materials Science, Multidisciplinary)

The steam oxidation test on the Cr-coated Zry cladding was studied up to 1400$$^{circ}$$C to understand the oxidation behavior under the accidental conditions. The double-sided oxidation test study showed that Cr coating can protect Zry cladding at 1200$$^{circ}$$C within 5 min. Cr coating has a protective effect on the Zry cladding up to 1200$$^{circ}$$C in a steam environment. However, in the oxidation test up to 1200$$^{circ}$$C/30 min and 1300$$^{circ}$$C/5 min, Cr coating can no longer protect Zry cladding. Furthermore, at 1300$$^{circ}$$C, the intermetallic phase of the Zr(Cr, Fe)$$_{2}$$ phase that precipitated within the Zry substrate formed as globule microstructures with Fe enrichment. In addition, the transition of the intermetallic phase within the Zry substrate from the solid to the pre-liquid and liquid phases was observed, where it was determined at 1350$$^{circ}$$C/60 min and 1400$$^{circ}$$C/30 min within the ZrO$$_{2}$$ phase (outer side region). The oxidation of the Zr(Cr, Fe)$$_{2}$$ interlayer was also determined in this study, where it resulted in the formation of the oxide phase of Cr, Zr, and Fe. It is worth mentioning that further experiments, such as mechanical testing and modeling, should be considered to support the degradation of the Cr-coated Zry cladding mainly when the liquid phase of the intermetallic phase is obtained for beyond design-basis accident environment.

論文

Chemical interaction between Sr vapor species and nuclear reactor core structure

Mohamad, A. B.; 中島 邦久; 三輪 周平; 逢坂 正彦

Journal of Nuclear Science and Technology, 60(3), p.215 - 222, 2023/03

 被引用回数:0 パーセンタイル:0.00(Nuclear Science & Technology)

The distribution of Sr in the reactor would be influenced by a chemical reaction of Sr vapor species with a structural material of internal reactor and fuel cladding materials; stainless steel (SS) or Zircaloy (Zry) cladding during 1F-NPS accident. The chemical interaction between Sr-Zry and Sr-SS has been described. The reaction tests have been performed to investigate the chemical interaction behavior under possible severe accident conditions. The tests have been conducted up to 1523 K under steam atmosphere. It was confirmed that Sr-Zr-O and Sr-Si-O compounds were formed through 2 kinds chemical interactions; gas-solid reaction and liquid-solid reaction. The gas and liquid species of Sr in a good contact with the solid Zry and SS to form Sr-Zr-O and Sr-Si-O compounds, respectively. Sr was deposited onto the Zry and SS surfaces and lead to the formation of reaction product. Thus, this study highlights the possibility that Sr was deposited and retained in the core structure where the temperature was elevated during the accident in the 1F-NPS.

口頭

CrコーティングZr合金製被覆管を用いたATFのSAMPSONによる解析手法の開発

手塚 健一*; 木野 千晶*; 山下 晋; Mohamad, A. B.; 根本 義之

no journal, , 

原子力発電所の過酷事故発生時に水素発生・炉心溶融の進展を抑制することを目的とした事故耐性燃料の開発が進んでおり、PWR用には、CrコーティングしたZr被覆管(Crコーティング被覆管)が検討されている。本研究では、Crコーティング被覆管を用いたATFの事故時挙動を評価するための解析手法を開発した。解析ツールとして、国産のSA解析コードであり、ソースコードに容易にアクセス可能なSAMPSONを用いた。解析の結果、現行被覆管に比べて、Crコーティング被覆管を用いることで、事故模擬条件において有意な水素発生抑制効果を確認することができた。

口頭

The Behavior of Cr-coated Zry cladding under high-temperature steam oxidation

Mohamad, A. B.; 古本 健一郎; 根本 義之; 井岡 郁夫; 佐藤 大樹*; 岡田 裕史*; 山下 真一郎; 逢坂 正彦

no journal, , 

Chromium (Cr) coated zirconium (Zr) based alloy cladding is the promising material for a near term accident tolerant fuel (ATF). Cr-coated Zr based cladding was fabricated by sputtering technique and HT oxidation tests were conducted up to different high temperatures (1100$$^{circ}$$C to 1400$$^{circ}$$C). The present work aims to investigate the metallography of Cr-coated Zr cladding after HT steam test. The result showed Cr$$_{2}$$O$$_{3}$$ layers were formed as a protective oxide layer at the outmost layer for all samples. However, Cr-Zr-Fe phases were observed In particular, the main phases observed in the cross-sectional area were Cr$$_{2}$$O$$_{3}$$, Cr, ZrO$$_{2}$$, Cr-Zr, and Zr situated from outer to inner of the sample after HT test. The details of the microstructure and mechanism of these samples will be discussed in the presentation.

口頭

Overview of ATF R&D program in Japan

山下 真一郎; Mohamad, A. B.; 井岡 郁夫; 根本 義之; 川西 智弘; 逢坂 正彦; 加治 芳行

no journal, , 

福島第一原子力発電所事故を教訓に、冷却材喪失等の過酷条件においても損傷しにくく、高い信頼性を有する新型燃料の開発への関心が高まり、世界中の多くの国々において事故耐性を高めた新型燃料(ATF: Accident Tolerant Fuel)の研究開発が進められている。国内におけるATF開発は、経済産業省資源エネルギー庁からの委託を受けて共通基盤技術開発を担う原子力機構と、各ATF候補材料の要素技術開発を担うプラントメーカ、燃料メーカのコンソーシアムが、密接に連携協力しながら進めてきている。本講演では、各ATF要素技術の進捗状況報告に先立ち、本邦で進められているATF開発の状況について、国内でのATF開発における原子力機構の役割等の説明を基盤研究の成果を交えながら概要を説明する。

口頭

Accident-Tolerant Fuel R&D Program in Japan

山下 真一郎; Mohamad, A. B.; 井岡 郁夫; 根本 義之; 川西 智弘; 加治 芳行; 逢坂 正彦; 村上 望*; 大脇 理夫*; 佐々木 政名*; et al.

no journal, , 

日本の事故耐性燃料(ATF)研究開発プログラムは、軽水商用炉の炉心、燃料の評価及び実際の設計、研究開発の経験等を最大限に利用するために国内のプラントメーカ、燃料製造メーカ、大学などと協力し2015年より進められてきている。現在国内で検討されているATF候補材料は、潜在的にPWR及びBWRへの適用が期待できる炭化ケイ素/炭化ケイ素(SiC/SiC)複合材料、BWR向けに開発が進められている酸化物分散によって強化されたFeCrAl鋼、PWR向けCr-コーティングジルカロイ被覆管である。また、被覆管材料に加えて、SiC製のBWR用チャンネルボックスや事故耐性制御棒に関する研究開発も行われている。本講演では、ATFプログラムにおけるJAEAの役割を含めて、現在までの研究開発の進捗状況を概説する。

口頭

Transition of the Zr(Cr, Fe)$$_{2}$$ intermetallic phase up to the eutectic temperature

Mohamad, A. B.; 根本 義之; 古本 健一郎*; 岡田 裕史*; 佐藤 大樹*

no journal, , 

The development of Accident Tolerant Fuel (ATF) had been started by conducting the investigation on new concepts to improve the safety of Light Water Reactors (LWRs). It is well known that the Cr coating on Zry cladding has shown an improvement in behavior under accident conditions and normal operation. In the Cr-Zr system, the eutectic phase of ZrCr$$_{2}$$ is present at 1332$$^{circ}$$C and forms as intermetallic compounds. There is still lack of data on the evolution of the intermetallic phase when the oxidation temperature reaches the eutectic temperature of Cr-Zr. Therefore, the purpose of this study will be to understand the solid-to-liquid phase transition of Zr(Cr, Fe)$$_{2}$$. High temperature oxidation tests were performed in a steam atmosphere to the target temperature (i.e., 1100$$^{circ}$$C, 1200$$^{circ}$$C, 1300$$^{circ}$$C, 1350$$^{circ}$$C, and 1400$$^{circ}$$C) for different exposure times of 5, 30, and 60 min. From the tests, the transition of Zr(Cr, Fe)$$_{2}$$ that formed at the Cr-Zr interface and also that precipitated in the Zry cladding were studied with varied oxidation time and temperatures. The microstructural evolution of the intermetallic phase was observed in the Zr substrate within the progress of the oxidation of Cr-coated Zry. A dendritic structure was observed at 1400$$^{circ}$$C, indicating the formation of the Zr(Cr, Fe)$$_{2}$$ liquid phase when the oxidation temperature is above the eutectic temperature.

口頭

The Transition of protective coating to no-longer protective coating of Cr-coated Zry cladding in high temperature steam oxidation

Mohamad, A. B.; 根本 義之; 古本 健一郎*; 岡田 裕史*; 佐藤 大樹*

no journal, , 

The development of Accident Tolerant Fuel (ATF) started with the investigation of new concepts to improve the safety of Light Water Reactors (LWR). It is well known that the Cr coating on Zry cladding has shown improved behaviour under accident conditions and in normal operation. However, many questions remain about the oxidation behaviour of Cr-coated Zry cladding as it approaches the Cr-Zr eutectic temperature. In the present study, the steam oxidation tests were carried out under different oxidation conditions in order to understand the oxidation behaviour of the Cr-coated material mainly above the eutectic temperature. The results obtained showed that the Cr coating can protect the Zry substrate at 1100$$^{circ}$$C to 1200$$^{circ}$$C/5min. However, at 1200$$^{circ}$$C/30min, the Cr coating no longer protected the Zry substrate. This is due to the formation of Zr at the Cr grain boundary where it becomes a short path for O diffusion and reacts with the Zry substrate.

口頭

JAEAにおけるATF基礎基盤研究

Mohamad, A. B.; 根本 義之; 相馬 康孝; 石島 暖大; 佐藤 智徳; 井岡 郁夫; Pham, V. H.; 三輪 周平; 中島 邦久; 加治 芳行; et al.

no journal, , 

ATF等の新型燃料実用化においては、関連技術開発やそれらの基となる科学的知見の取得及び拡充が不可欠である。原子力機構は、照射試験実施による燃料ふるまい解析技術基盤の構築のための研究開発を行い、長期を要する開発において、開発内容やスケジュールの予見性向上に貢献していくべきと認識している。このため、実装化が最も早いCrコーティング被覆管に関して、燃料ふるまいのメカニズムに立ち返り、「長期照射時の影響」「事故時影響」に関する科学的知見を拡充することを目的とした基礎基盤研究計画を立案し、研究をすすめている。本発表では各研究項目の内容や期待される成果、これまでに得られた結果等を紹介する。

口頭

Improving chemical thermodynamics knowledge of severe accidents within the OECD-TCOFF2 Project

Journeau, C.*; Bechta, S.*; Komlev, A.*; 倉田 正輝; 多木 寛; 松本 俊慶; Mohamad, A. B.; Barrachin, M.*; Quaini, A.*; Bottomley, D.*; et al.

no journal, , 

The OECD project TCOFF-2 (Thermodynamic Chemistry of Fission Products - Part 2) is an extension of the original project that was part of the near-term projects intended to support the Fukushima Dai-ichi decommissioning efforts. This second part continues to be mainly financed by Japanese Ministry (MEXT) with JAEA-CLADS support. TCOFF2 started in August 2022 and includes certain new material requirements compared to the first TCOFF project. These are increased emphasis on accident tolerant fuels (ATFs), certain actinides and fission products related to volatility/leachability but also certain combinations of the ceramic oxide systems important for MCCI behaviour. Following a PIRT review of phenomena in TCOFF 1, there was in TCOFF2 a Task 1 to prioritise current needs, particularly for relevance to severe accident phenomena and alternative materials (ATFs). This included a re-evaluation and ranking of the thermodynamic systems to make a Systems Identification and Ranking Table (SIRT) for the improvement of the thermodynamic database foreseen in TCOFF2. The further systems for evaluation were then proposed by the partners according to their particular requirements; this resulted in over 150 thermodynamic systems. In the mid-year meeting (June 2023) discussions were made to rationalise these into the most important 20 systems. The rationale for the reduction of systems to this limit will be explained in the talk, both those systems that were omitted as well as those finally included. These systems will then be used as a priority list of work for the experimental call to members that will be launched by the TCOFF-2 project towards the end of the year 2023 with the intention to initiate the first projects soon afterwards.

口頭

Study on coating technic to enhance accident tolerance of fuel cladding, 3; Irradiation behavior of the Cr coated MDA cladding

Mohamad, A. B.; Chien, J.; 井岡 郁夫; 鈴木 恵理子; 近藤 啓悦; 根本 義之; 大久保 成彰; 山下 真一郎; 岡田 裕史*; 佐藤 大樹*

no journal, , 

As a candidate for accident tolerent fuel (ATF) cladding tubes, chromium (Cr) coated Zry cladding tubes are being developed. To realize the Cr-coated Zry cladding for future cladding application, the integrity of this material needs to be confirmed with the reactor environment conditions. In order to understand an effect of irradiation on the Cr-coated Zry cladding, ion irradiation test is carried out on the cross-sectional specimens. In this study, the following content mainly focuses on the microstructural evolution and mechanical behavior induced by ion irradiation in Cr-coated Zry cladding. The 10 MeV-Fe$$^{3+}$$ irradiation was chosen to induce the damage on the cross section of the Cr-coated Zry cladding. The sample was irradiated at 350 $$^{o}$$C and the peak irradiation damage was approximately 30 dpa. TEM-EDS shows that the Fe-enrichment peaks are observed around 15 nm at the interface regions between the coating and the Zry substrate for the sample irradiated up to 30 dpa. In addition, the hardness of irradiated sample is higher compared to that un-irradiated sample as result of irradiation-induced hardening. The details of the irradiation effect on the un- and irradiated Cr-coated Zry cladding will be discussed in detail during the presentation.

口頭

事故耐性向上を目指した燃料被覆管のコーティング技術に関する研究,1; JAEAにおける事故耐性コーティング技術研究と装置開発

山下 真一郎; Mohamad, A. B.; 根本 義之; 相馬 康孝; 石島 暖大; 佐藤 智徳; 井岡 郁夫; Pham, V. H.; 三輪 周平; 中島 邦久; et al.

no journal, , 

原子力機構(JAEA)では事故耐性の向上を目指した燃料被覆管の各種コーティング技術の研究を行っている。講演では全体概要の他、それらの研究に用いることを目的としたJAEAでの新規の装置開発について紹介する。

口頭

Oxidation behavior of Cr-coated Zry cladding in steam environments

Mohamad, A. B.; 根本 義之; 古本 健一郎*; 岡田 裕史*; 佐藤 大樹*

no journal, , 

It is widely recognized that the Cr coating on Zry cladding has shown an improvement in the behavior under accident conditions and normal operation. Many research groups around the world have conducted the high-temperature oxidation and LOCA tests on Cr-coated Zry under accident conditions and explained the degradation phenomena from these tests. Although many literatures have revealed the mechanism and phenomena of the degradation of the Cr-coated, there is still a lack of data on the Zr-Cr-Fe phase or intermetallic phase behavior when the temperature reaches and exceeds the eutectic temperature of Zr-Cr (1332$$^{circ}$$C). In the present study, a high temperature steam oxidation test is carried out from 1100 to 1400$$^{circ}$$C in order to understand the behavior of Cr-coated Zry as it approaches the eutectic temperature. Fromelectron probe microanalysis, the Fe enrichment of the Zr(Cr,Fe)$$_{2}$$ phase is identified for the sample tested at 1300$$^{circ}$$C. In addition, the liquid formation of the Zr(Cr,Fe)$$_{2}$$ phase is observed at 1300$$^{circ}$$C.

口頭

Fundamental research program on zircalloy with accident tolerance

Mohamad, A. B.; 相馬 康孝; 根本 義之; 阿部 陽介; 井岡 郁夫; 佐藤 智徳; 石島 暖大; 三輪 周平; 中島 邦久; 加治 芳行; et al.

no journal, , 

日本原子力研究開発機構(以下、JAEA)では、2019年に事故耐性を兼ね備えたジルカロイに関する基礎研究を立ち上げ取り組んできている。基礎研究を実施する主目的は、長期の通常運転時、冷却水喪失事故(以下、LOCA)時、設計基準外事象(以下、B-DBA)時、過酷事故(以下、SA)時におけるジルカロイ挙動の理解を深化させること、そして国内メーカで開発されているクロムコーティングジルカロイの実装を支援すること、である。JAEAはまた、通常運転時、LOCA時、B-DBA時、SA時における事故耐性コーティングジルカロイの挙動理解に必要な基礎技術開発も行っている。例えば、通常運転条件を模擬するために軽水炉の冷却条件を組合わせたイオン照射試験技術を開発している。また、被覆管の破断やバルーニングを詳細に理解するために、LOCA試験で得られた結果を機械学習に取り込んだ解析等もしている。さらには、高温酸化試験のような分離効果試験なども実施している。加えて、B-DBAやSA時の核分裂生成ガスの放出についても研究プログラムに含まれている。将来的には、これらの基礎技術を用いて得られた研究結果は、統合されて燃料ふるまい解析コードに導入されることによって原子炉の運転条件下での燃料ふるまいの予測に用いられる。

口頭

Overview of ATF R&D program in Japan

山下 真一郎; Mohamad, A. B.; 相馬 康孝; 根本 義之; 井岡 郁夫; 加治 芳行; 逢坂 正彦

no journal, , 

2015年以降、日本の事故耐性燃料(ATF)研究開発プログラムは、国内商用炉の燃料や炉心の設計・評価、研究開発における経験を最大限に活用するため、発電プラントメーカ、燃料メーカや大学と連携し進められてきている。本講演では、同プログラムにおける原子力機構の役割を含めて、現在の研究開発について進捗状況の概要を紹介する。

口頭

Updating fission product chemistry database based on recent investigation in Fukushima-Daiichi Nuclear Power Station, 1; Overview of fundamental study related to fission product chemistry

三輪 周平; 中島 邦久; 唐澤 英年; Rizaal, M.; Luu, V. N.; Mohamad, A. B.

no journal, , 

東京電力福島第一原子力発電所内のセシウム等のFPの分布や性状を把握することが廃炉に向けた重要な課題であり、日本原子力研究開発機構ではそれらに大きな影響を与えるFP化学に着目した基礎研究を実施し、事故時の原子力発電所内のFP化学を評価するためのデータベースECUMEを開発している。ECUMEは福島第一原子力発電所の事故により明らかになった重要な現象、例えば、セシウムの制御材ホウ素との反応や、構造材との反応に関するデータやモデルを収納している。近年の内部調査により明らかとなったシールドプラグでの高線量等の原因を明らかにするため、コンクリートや他の炉内物質との化学反応を調べ、モデル化を行い、ECUMEの更新を進めている。

口頭

System Identification and Ranking Table (SIRT) for chemical thermodynamics of severe accidents

Journeau, C.*; Bechta, S.*; Komlev, A.*; 倉田 正輝; 多木 寛; 松本 俊慶; Mohamad, A. B.; Barrachin, M.*; Quaini, A.*; Gu$'e$neau, C.*; et al.

no journal, , 

Within the second phase of the OECD/NEA project on Thermodynamic Characterisation of Fuel Debris and Fission Products Based on Scenario Analysis of Severe Accident Progression (TCOFF-2), a Systems Identification and Ranking Table (SIRT), has been derived from the usual PIRT and adapted to chemical thermodynamics and material science. Firstly, a draft list of systems of interests has been distributed to TCOFF-2 partners. After review, a final list of 154 systems has been considered. Then system ranking has been carried out. Two series of figures of merit have been considered: importance for safety including the contribution to severe accident phenomena and to fission product behaviour and source term as well as needs for further R&D, based on the lack of existing data and the needs to improve relevant existing thermodynamic databases (mostly NUCLEA and TAF-ID). Each participating organization provided ranking of the systems for these Figure of Merits. Relevance to safety has been organized in 10 columns: Classical LWR / High BU, High enrichment fuel/ Fukushima Daiichi, TMI2, Chornobyl/Near-term ATF cladding (Cr-coated Zry)/ Near-term ATF cladding (FeCrAl) /Long-term ATF cladding (SiC) /Advanced fuels (UN, U-Si) /Advanced Modular Reactors / Post-Accident Leaching). Received rankings have then been agglomerated and averaged. A dedicated finalization meeting has been held in June 2023 in which the systems having the highest relevance and the largest needs for improvement have been considered. It must be noted that some systems for which there is a rather good knowledge, like (U,O, Zr) have not been selected, nor systems in which thermodynamic results are very hard to attain due to kinetics effect such as (U,H,O). After some system grouping, consensus have been reached on a list of 20 systems (13 linked to current or near-term fuels and 7 linked to longer-term fuels) having the highest priority.

口頭

事故耐性向上を目指した燃料被覆管のコーティング技術に関する研究,2-1; JAEAにおける事故耐性コーティング技術研究と装置開発

Mohamad, A. B.; 根本 義之; 相馬 康孝; 石島 暖大; 佐藤 智徳; 井岡 郁夫; Pham, V. H.; 三輪 周平; 中島 邦久; 加治 芳行; et al.

no journal, , 

原子力機構(JAEA)では事故耐性の向上を目指した燃料被覆管の実用化に資する基礎研究の一環として、各種コーティング被覆管の特性評価を進めている。講演では全体概要の他、特性評価試験に用いることを目的としたJAEAでの新規装置開発の状況、高温水蒸気酸化特性評価試験の結果等について紹介する。

口頭

連続濃度傾斜を有するコンビナトリアル試料を用いたFe-Cr-Al合金中のCrリッチ粒子形成に関するハイスループット実験

阿部 陽介; Mohamad, A. B.; 佐々木 泰祐*; 山下 真一郎; 大久保 成彰; 鵜飼 重治

no journal, , 

軽水炉の事故耐性燃料被覆管として開発中のFe-Cr-Al(ODS)合金では、Crリッチ脆化相($$alpha$$'粒子)形成による脆化が危惧されるものの、脆化が発現するCrとAlの組成範囲は未確定で、照射の影響も十分理解されていない。本研究では、Fe-Cr-Al合金を拡散対とした相互拡散熱処理により、連続濃度傾斜を有するコンビナトリアル(コンビネーション(組合せ)とマテリアル(材料)の合成語)試料を作製し、熱時効前後での組成と硬さのハイスループット測定を行うことで、固溶強化や$$alpha$$'粒子生成に対する組成の影響を検討した。SEM/EDSによるCrとAlの2次元濃度測定の結果、得られた傾斜濃度がカバーする組成範囲が、$$alpha$$'粒子の析出境界の決定に必要な組成範囲を概ね満たすことを確認した。拡散熱処理後の硬さデータから固溶強化を検討した結果、冷却速度が遅い場合にスピノーダル分解やAl析出物の形成が示唆された。また、熱時効後の硬さデータから$$alpha$$'粒子の析出境界を精緻に評価した。

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