Report of Tokai Works  PNCT-AR-67

[目次]

  • CONTENTS
  • NUCLEAR FUEL RESEARCH LABORATORY
  • Ceramics Group
  • 1. The Effects of Chloride Ions and U(VI)Ions in the Sol-Gel Process of UO2 / 1
  • 2. Densification Mechanism of Sol-Gel UO2 in Sintering Process / 1
  • 3. The Sol-Gel Process Starting from Uranyl Nitrate Solutions / 3
  • 4. A Modified Sol-Gel Process Using Fine UO2 Powder as a Starting Material / 4
  • 5. Low Temperature Sintering of UO2 Pellets from the Active Powders Prepared by Hydrolysis of Uranous Solution / 5
  • 6. Preparation of UC Aggregates by the Application of the Sol-Gel Process / 6
  • 7. Preparation of Uranium Monocarbide Pellet / 6
  • 8. Some Improvements of Sol-Gel UO2 Preparation Technology / 7
  • 9. Vibratory Compaction of Sol-Gel UO2 in a Thermal Reactor Fuel Rod / 8
  • 10. Fabrication of a Thermal Reactor Fuel Rod by Vib-Swaging Process / 9
  • 11. Preparation of Non-Stoichiometric Uranium Dioxide / 10
  • 12. Ultrasonic Attenuation in Sintered UO2 Pellets with Excess Oxygen / 11
  • 13. Oxidation Behaviours of Uranium Carbides / 13
  • 14. Observation of Etch Pits in Uranium Oxides / 13
  • 15. Thermal Conductivity of UO2±x by Using a Transient Technique / 16
  • 16. Lattice and Grain-Boundary Diffusion of Protactinium in ThO2 and ThO2-UO2 / 18
  • 17. The Release of Xe-133 from Irradiated UO2 by the Step-Way Heating at Intervals of 50℃ / 18
  • 18. Irradiation Testing of Vibratory Compacted Natural UO2 Prepared by the Sol-Gel Process-Post-Irradiation Examination- / 19
  • Reprocessing Group
  • 19. The Fundamental Distribution Data of Plutonium between Tri-n-butyl Phosphate and Nitric Acid / 21
  • 20. Process Study for Aqueous Reprocessing(I)-Experimental Installation and Operation Test- / 22
  • 21. Process Study for Aqueous Reprocessing(II)-Extraction Test of U-HNO3 and Pu-HNO3 Systems- / 23
  • 22. Process Study for Aqueous Reprocessing(III)-Computer Evaluation of Plutonium Purification Process with TBP- / 24
  • 23. Process Study for Aqueous Reprocessing(IV)-Test of U/Pu Partition Process with U(IV)as a Reductant- / 25
  • 24. Recovery of Krypton and Xenon in the Off-Gases from Reprocessing of Spent Fuels.
  • (I) The Solubility of Various Gases in Carbon Tetrachloride / 26
  • 25. Recovery of Krypton and Xenon in the Off-Gases from Reprocessing of Spent Fuels.
  • (II) Dynamic Adsorption of Krypton by Activated Carbon / 27
  • 26. Use of Zeolites on the Treatment of Low-Level Radioactive Liquid Waste / 28
  • Other Groups
  • 27. The Uranium(IV)-(VI)Electron Exchange Reactions in the Anion Exchange Resin,Tri-n-Octyl Amine and Tri-Butyl Phosphate / 30
  • 28. Behavior of Uranium Isotope in Breakthrough Operation / 32
  • 29. Geochemical Characteristics of Tōno Uranium Deposits / 33
  • 30. Radioactivity in Shakanai Black Ore Mine / 33
  • PLUTONIUM FUEL DIVISION
  • Fuel Design Section
  • 31. Design and Specification of IFA-159 PuO2-UO2 Fuel Assembly for the Irradiation in HBWR / 34
  • 32. Design and Specifications of IFA-159 PuO2-UO2 Fuel Assembly for the Irradiation in the Halden Boiling Water Reactor / 34
  • 33. Experimental Studies for Basic Reactor Physics of PuO2-UO2 Fuel
  • 1) Reactivity Effect of the Single Fuel Rod / 35
  • 34. Experimental Studies for Basic Reactor Physics of PuO2-UO2 Fuel
  • 2) Measurement of Fine Thermal Neutron Flux Distribution in the Single Fuel Rod / 35
  • 35. Experimental Studies for Basic Reactor Physics of PuO2-UO2 Fuel
  • 3) Theoretical Analysis of Fine Thermal Neutron Flux Distribution in the Single Fuel Rod / 36
  • 36. Experimental Studies for Basic Reactor Physics of PuO2-UO2 Fuel
  • 4) Measurement of Time Decay of Fission Product Activity / 36
  • 37. Fuel Design for Plutonium Recycling in Boiling Water Power Reactor(II)Vipac Type Fuel / 37
  • 38. Development of Transient Temperature Calculation Code / 37
  • 39. In-Core Measurement of Thermal Conductivity of Vipac UO2 Fuel / 38
  • 40. Experimental Study on the Consequences of Fuel Cladding Failure(II) / 38
  • 41. Theoretical Analysis of Thermal Conductivity for Porous Ceramic Fuel / 39
  • Fuel Development Section
  • 42. The Irradiation Testing Program in Enrico-Fermi Fast Breeder Reactor(III)Production of Fuel Materials / 39
  • 43. PuO2-UO2 Pellet Fabrication(II) / 40
  • 44. Preparation of UO2-PuO2 Pellets for the Irradiation Testing in Halden HBWR / 41
  • 45. PuO2-UO2 Hollow Pellet Fabrication / 41
  • 46. Fabrication of PuO2-UO2 Fuel Specimens for the Screening Irradiation Test in GETR / 43
  • 47. Fabrication of Enrico-Fermi Reactor Irradiation Test Specimen by Vibratory Compaction / 44
  • 48. Preparation of Plutonium Hydroxide Gel by the Sol-Gel Process / 44
  • 49. Preparation of High Density and Coarse Particles of UO2-PuO2 by the Sol-Gel Process / 44
  • 50. Preparation of PuO2-UO2 Dense Oxide by the Sol-Gel Process(IV)-Preparation of High Plutonium Enriched PuO2-UO2- / 45
  • 51. Thermal Conductivity of UO2±x / 45
  • 52. Phase Study on U-Pu-O System(I) / 45
  • 53. X-Ray Diffraction Work for Plutonium Oxides(II) / 46
  • 54. Electron Microscopic Replica Technique for Plutonium Bearing Oxides / 46
  • 55. Determination of Plutonium Distribution in PuO2-UO2 Fuel by Autoradiography-Photometry / 46
  • 56. Analysis of Released Gas from UO2-PuO2 / 47
  • 57. Radiochemical Determination of Isotopic Ratio in Plutonium / 47
  • 58. Determination of Plutonium by Potentiometric Titration / 48
  • 59. Determination of Plutonium in Waste Solution from Production Process / 48
  • 60. Analysis of Impurities in PuO2 and PuO2-UO2 by Emission Spectrography / 48
  • TECHNICAL DIVISION
  • Analytical Section
  • 61. Spectrochemical Determination of Calcium in Plutonium and Uranium Mixed Oxide / 49
  • 62. Spectrophotometric Determination of Trace Amounts of Uranium by Divenzoylmethane / 50
  • 63. Analytical Methods for Routine Analysis in Chemical Reprocessing Study(I) / 51
  • 64. Measurement of U-235 by Microwave Discharge Method / 52
  • 65. Spectrophotometric Determination of Vanadium in Uranium Oxides with N-Benzoylphenylhydroxylamin / 54
  • Inspection Section
  • 66. Ultrasonic Testing of Zircaloy Sheath Tubing / 54
  • 67. Ultrasonic Testing of Zircaloy Sheath Tubing after Autoclave Treatment / 57
  • Production Section
  • 68. Fabrication Experience of Natural Uranium Thin Plate-Type Fuels for Fast Critical Assembly(I)Casting of Slabs to be Rolled / 58
  • HEALTH AND SAFETY SECTION
  • 69. A Use of Glass Dosimeters for Personnel Monitoring in Plutonium Facilities / 59
  • 70. Local Ventilation Device Used in Plutonium Safety Handling / 60
  • 71. Problems of Plutonium Contamination Control / 61
  • 72. Control of Airborne Contamination in the Plutonium Fuel Development Laboratory at P.N.C. / 61

「国立国会図書館デジタルコレクション」より

この本の情報

書名 Report of Tokai Works
著作者等 動力炉核燃料開発事業団
シリーズ名 PNCT-AR ; 67-68
巻冊次 PNCT-AR-67
出版元 Tokai Works, Power Reactor & Nuclear Fuel Development Corporarion
刊行年月 1968
ページ数 2冊
大きさ 30cm
全国書誌番号
22056718
※クリックで国立国会図書館サーチを表示
言語 英語
出版国 日本
この本を: 
このエントリーをはてなブックマークに追加

Yahoo!ブックマークに登録
この記事をクリップ!
Clip to Evernote
このページを印刷

外部サイトで検索

この本と繋がる本を検索

ウィキペディアから連想