原子炉内熱流動現象の更なる高精度な解析を目指して,燃料集合体内の液膜・液滴挙動,原子炉圧力容器内など複雑な体系下での3次元二相流挙動,シビアアクシデント時に見られる複雑な相変化や化学反応が共存した多相・多成分熱流動等について,現象の微視的機構に着目したシミュレーションの研究が進められている。また,連続体力学的手法に基づく超微細メッシュ法,粒子法,格子ボルツマン法,オートマトン法,分子動力学的視点からモンテカルロ法などの数値解析技術を駆使する微視的な熱流動現象のシミュレーションも多く試みられるようになってきている。これどの数値解析技術を駆使する微視的な熱流動現象のシミュレーションも多く試みられるようになってきている。これでの運動に着目した巨視的シミュレーションなど既存の利用可能な技術と微視的シミュレーション技術との間の関係を明らかにし,今後の研究の方向性と課題を展望する。

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Recent topics of nuclear criticality safety in the United States are presented. A guidance and two standards issued by U.S.DOE, Nuclear Criticality Safety Guide LA-12808, a work group for ANSI/ANS-8.15, a continuous energy version of KENO V.a, and the Zeus benchmark experiment are descried below. Source data of these topics are from journals, reports, transactions of American Nuclear Society and so on issued from 1995 until 1999, and internet sites.

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Mock-ups of acid recovery evaporators which are made of Ti-5%Ta alloy and Zr were tested under non-radioactive condition for 40, 000h to solve the corrosion problem in the reprocessing field. A long-term durability of both evaporators was demonstrated by results of operation data, evaporation performance, and corrosion resistance.
With regards to corrosion resistance, cracks and local corrosion such as pitting and intergranular attack were not observed on both evaporators after the test, and a corrosion of weld was not selective. The average corrosion rates at measuring points were less than 0.1mm/yr respectively, however, thickness of the Ti-5%Ta alloy evaporator was slightly reduced in vapor phase region. In addition, the destructive inspection showed that the mechanical properties of both materials were not degraded during the test.
From an engineering perspective, an excellent durability of Ti-5%Ta alloy and Zr evaporators under high oxidative environment was successfully demonstrated through this mock-up test.

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Using probabilistic fracture mechanics (PFM), we performed qualitative evaluations on effectiveness of various maintenance strategies for steam generator (SG) tubes used in pressurized water reactors (PWRs). Here inspection accuracy, inspection interval, sampling inspection and SCC allowance operation were selected as parameters for sensitivity analyses. Although the analyses were mainly conducted for SG tubes made of Inconel 600mill anneal (MA) material, additional analyses were performed for assumed materials whose resistance to PWSCC was improved.
The results demonstrated that failure probabilities were influenced significantly by the crack propagation law in I600 material. In both I600 and the assumed materials, the improvement of inspection accuracy would dramatically reduce the failure probabilities. With respect to SCC allowance operation crack, the results showed that it is preferable to improve the inspection accuracy.
The probabilistic approach conducted in this study enables to make effective maintenance strategies with being taken account of various maintenance methods.

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In order to evaluate neutron irradiation effect on divertor target plate component, brazed specimens of C/C composite armor tiie on alumina dispersed copper and HIPed specimen of beryllium on alumina dispersed copper were irradiated up to a fluence of 2.8×10²³n/m² (>0.18MeV) at 746-788K, and their strength tests were performed.
From these tests, the following results were obtained.
(1) MFC-1/alumina dispersed copper brazed-specimen
Tensile and 4 point bending strength of irradiated brazed-specimen was 30-55% of that of unirradiated one. Tensile and 4point bending strength of unirradiated and irradiated brazed-specimen increased as the test temperature was increased, however these strength decreased above 600K. Remarkable degradation of tensile strength was observed for the irradiated brazed-specimen above 600K. Fracture position of irradiated specimen shifts from MFC-1 side brazed zone to alumina dispersed copper as the testing temperature increases.
(2) Beryllium/alumina dispersed copper HIPed specimen
Under tensile and bending loading, irradiated HIPed-specimen exhibits similar deformation behavior as copper alloy, but fracture strength of the specimen was dominated by fracture strain limitation of beryllium material. Tensile and 4 point bending strength of irradiated HIPed-specimen showed 10-20% lower values than that of unirradiated one. The room-temperature strength of irradiated joint was about 3 times higher than that of the irradiated MFC-1/alumina dispersed copper brazed-specimen. Tensile and 4 point bending fracture strains of irradiated joint showed 0-30% lower values than that of unirradiated one. Fracture of irradiated HIPed-specimens was observed at joined area.

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