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Radiochemie pp 712-774 | Cite as

Die Wiederaufarbeitung von Uran-Plutonium-Kernbrennstoffen

  • F. Baumgärtner
  • H. Philipp
Conference paper
Part of the Fortschritte der Chemischen Forschung book series (TOPCURRCHEM, volume 12/4)

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Literatur

  1. 1).
    Jelinek-Fink, P., u. B. Skorpil: Angereichertes UF6 — Preise bei Bezug von der USAEC. Atomwirtschaft 13, 365 (1968).Google Scholar
  2. 2).
    Rometsch, R.: Wiederaufarbeitung abgebrannter Kernbrennstoffe in Europa — Aufgaben der Eurochemic. Atomwirtschaft 11, 423 (1966).Google Scholar
  3. 3).
    Zusammengestellt nach Angaben von H. Grümm et al. Ergänzendes Material zum Bericht: Kernbrennstoffbedarf und Kosten verschiedener Reaktortypen in Deutschland, KFK-466 (1966).Google Scholar
  4. 4).
    Kennedy, J. W., G. T. Seaborg, E. Segrè, and A. C. Wahl: Fissionable Isotope of a New Element: 94239. Phys. Rev. 70, 555 (1946).CrossRefGoogle Scholar
  5. 5).
    Symposium on Chemical Processing of Irradiated Fuels, Hanford 1959. US-Report TID-7583 (1959).Google Scholar
  6. 6).
    Stoller, S. M., and R. B. Richards: Reactor Handbook, Vol. II, S. 124. Fuel Processing (1961).Google Scholar
  7. 7).
    Seaborg, G. T.: The Plutonium Story in the Transuranium Elements. Edison Wesley 1958.Google Scholar
  8. 8).
    Thompson, S. G., and G. T. Seaborg: The First Use of Bismuth Phosphate for Separating Plutonium from Uranium and Fission Products, Process Chemistry. London: Pergamon Press 1956.Google Scholar
  9. 9).
    Lawrowski, S., L. R. Dawson, and I. E. Tepe: Solvent Extraction Process for Concentration and Isolation of Product. US-Report CN-2511 (1945).Google Scholar
  10. 10).
    Perlman, J.: Separation Processes. US-Report CN-3627 (1946).Google Scholar
  11. 11).
    Culler Jr., F. L., and F. R. Bruce: The Processing of Uranium-Aluminium Fuel Elements. Proc. Intern. Conf. Peaceful Uses At. Energy, Geneva 9, 484 (1955).Google Scholar
  12. 12).
    Flagg, J. F.: Chemical Processing of Reactor Fuels. New York: Academic Press 1961.Google Scholar
  13. 13).
    Lawrowski, S., and M. Levenson: Redox Process — A Solvent Extraction Processing Method for Irradiated Uranium, Process Chemistry, Vol. II. US-Report TID-7534 (1957).Google Scholar
  14. 14).
    Warf, J. C.: Extraction of Ce(IV) Nitrate by Butylphosphate. J. Am. Chem. Soc. 71, 3257 (1949); ursprünglich US-Reports CC-2402 (1945) und ISC-8 (1947).CrossRefGoogle Scholar
  15. 15).
    Ellison, C. V., D. E. Ferguson, and T. C. Runion: Solvent Extraction Recovery from Metal Waste. US-Report ORNL-258 (1949).Google Scholar
  16. 16).
    Ferguson, D. E., and T. C. Runion: Tributylphosphate Solvent Extraction of U from Metal Waste. US-Report ORNL-260 (1949).Google Scholar
  17. 17).
    Culler, F. L.: Reprocessing of Reactor Fuel and Blanket Materials by Solvent Extraction. Proc. Intern. Conf. Peaceful Uses At. Energy, Geneva 9, 467 (1955).Google Scholar
  18. 18).
    Gresky, A. T.: Solvent Extraction Separation of U-233 and Thorium from Fission Products by Means of Tributyl-Phosphate. Proc. Intern. Conf. Peaceful Uses At. Energy, Geneva 9, 505 (1955).Google Scholar
  19. 19).
    Lanham, W. B., and T. C. Runion: Purex Process for Pu and U Recovery. US-Report ORNL-479 (1949).Google Scholar
  20. 20).
    Goldschmidt, B., P. Regnaut, and J. Prevot: Solvent Extraction of Plutonium from Uranium Irradiated in Atomic Piles. Proc. Intern. Conf. Peaceful Uses At. Energy, Geneva 9, 492 (1955).Google Scholar
  21. 21).
    Flanary, I. R.: Solvent Extraction Separation of Uranium and Plutonium from Fission Products by Means of Tributyl-Phosphate. Proc. Intern. Conf. Peaceful Uses At. Energy, Geneva 9, 528 (1955).Google Scholar
  22. 22).
    Irish, E. R., and W. H. Reas: The Purex Process. US-Report HW-49483A.Google Scholar
  23. 23).
    Campbell, W. M.: Canadian Fuel Processing. Nucleonics 14, 92 (1956).Google Scholar
  24. 24).
    Gowing, M.: Britain and Atomic Energy 1939–1945. London: Macmillan 1964.Google Scholar
  25. 25).
    Howells, G. R., T. G. Hughes, D. R. Mackey, and K. Saddington: The Chemical Processing of Irradiated Fuels from Thermal Reactors. Proc. 2. Intern. Conf. Peaceful Uses At. Energy, Geneva 17, 3 (1958).Google Scholar
  26. 26).
    Fletcher, J. M.: Ethers as Extractants. In: Aqueous Processing Chemistry for Irradiated Fuels, Brüssel 1963, ENEA-Eurochemic.Google Scholar
  27. 27).
    Warner, B. F., W. W. Marshall, A. Naylor, and G. D. C. Short: The Development of the New Separation Plant, Windscale. Proc. 3. Intern. Conf. Peaceful Uses At. Energy, Geneva 10, 224 (1964).Google Scholar
  28. 28).
    Baroncelli, F., G. Galleri, A. Moccia, G. Scibona, and M. Zifferero: The Eurex Flowsheet: Processing of Irradiated U-Al Alloys by Amine Solvent Extraction. Comitato Nazionale Energia Nucleare, Rom 1963.Google Scholar
  29. 29).
    Zifferero, M.: Proposed Flowsheets for Amine Systems, Aqueous Processing Chemistry for Irradiated Fuels, Brüssel 1963, ENEA-Eurochemic.Google Scholar
  30. 30).
    Brown, K. B., C. F. Coleman, D. J. Crouse, J. O. Denis, and J. G. Moore: The Use of Amines Extractants for Uranium from Acidic Sulfate Liquors. US-Report AECD-4142 (1954).Google Scholar
  31. 31).
    Moore, J. G., K. B. Brown, and C. F. Coleman: Further Studies of Amines as Extractants for Uranium from Acid Sulfate Solutions. US-Report AECD-4145 (1955).Google Scholar
  32. 32).
    Arnold, W. D., and D. J. Crouse: Further Evaluation of Amines as Extractants for Uranium from Sulfate Liquors. US-Report ORNL-3030 (1961).Google Scholar
  33. 33).
    Coleman, C. F., K. B. Brown, J. G. Moore, and D. J. Crouse: Solvent Extraction with Alkyl Amines. Ind. Eng. Chem. 50, 1756 (1958).CrossRefGoogle Scholar
  34. 34).
    Brown, K. B., C. F. Coleman, D. J. Crouse, C. A. Blake, and A. D. Ryon: Solvent Extraction Processing of Uranium and Thorium Ores. Proc. 2. Intern. Conf. Peaceful Uses At. Energy, Geneva 3, 472 (1958).Google Scholar
  35. 35).
    Diamond, R. M., and D. G. Tuck: Extraction of Inorganic Compounds into Organic Solvents. Progr. Inorg. Chem. 2, 157 (1960).Google Scholar
  36. 36).
    Naylor, A.: Fission Product Chemistry in Relation to TBP Processes, in Reprocessing of Fuel from Present and Future Power Reactors. Kjeller-Report KR-126, S. 101 (1967).Google Scholar
  37. 37).
    Brown, P. G. M., J. M. Fletcher, C. J. Hardy, J. Kennedy, D. Scargill, A. G. Wain, and J. L. Woodhead: The Significance of Certain Complexes of Ruthenium, Niobium, Zirconium, and Uranium in Plant Processes. Proc. 2. Intern. Conf. Peaceful Uses At. Energy, Geneva 17, 118 (1958).Google Scholar
  38. 38).
    Scargill, D., C. E. Lyon, N. R. Large, and J. M. Fletcher: Nitratoaquo Complexes of Nitrosylruthenium-III, J. Inorg. Nucl. Chem. 27, 161 (1965).CrossRefGoogle Scholar
  39. 39).
    Klaas, J.: RuNO-Nitro Complexes. Conversion Rates and Extraction Coefficients, Proceedings XI. Intern. Conf. on Coordination Chemistry 1968. Amsterdam: Elsevier Publ. Co. 1968.Google Scholar
  40. 40).
    Wallace, R. M.: The Composition of Some Nitrato Nitrosylruthenium Complexes. J. Inorg. Nucl. Chem. 20, 283 (1961).CrossRefGoogle Scholar
  41. 41).
    Fletcher, J. M., C. E. Lyon, and A. G. Wain: Partion Coefficients of Nitratonitrosylruthenium Complexes between Nitric Acid and TBP-Phases. J. Inorg. Nucl. Chem. 27, 1841 (1965).CrossRefGoogle Scholar
  42. 42).
    Brown, P. G. M.: Nitrato Complexes of Nitrosylruthenium. J. Inorg. Nucl. Chem. 13, 73 (1960).CrossRefGoogle Scholar
  43. 43).
    Finsterwalder, L.: über die Extraktionskinetik von Plutonium IV und Uran VI im Purex-Prozeß. Dissertation TH München 1968.Google Scholar
  44. 44).
    Ortega, J., L. R. Salvador, and B. Lopez-Perez: The Use of Tartrates for the Separation of Uranium(VI) and Plutonium(IV) by Extraction with TBP, Solvent Extraction Chemistry, S. 335. New York: John Wiley 1967.Google Scholar
  45. 45).
    Naylor, A.: Plutonium-Uranium Separation Techniques in TBP Systems, in Reprocessing of Fuel from Present and Future Power Reactors. Norwegischer Report KR-126, S. 172 (1967).Google Scholar
  46. 46).
    Edwall, B.: Some Experience in the Use of Uranium(IV) Nitrate, in Aqueous Reprocessing Chemistry for Irradiated Fuels, Brüssel 1963, ENEA-Eurochemic.Google Scholar
  47. 47).
    Lopez-Menchero, E., L. Gehem, H. Eschrich, P. Hansen, J. Centeno, and R. Aerts: Study of Uranium(IV) Nitrate as Reductant for Plutonium. I. The Preparation of Uranium(IV) Nitrate Solutions. Eurochemic-Report ETR-180.Google Scholar
  48. 48).
    Biddle, P., H. A.C. McKay, and J. H. Miles: The Role of Nitrous Acid in the Reduction of Plutonium(IV) and Uranium(VI) by Uranium(IV) in TBP Systems. In: Solvent Extraction Chemistry of Metals, S. 133. London: Macmillan 1965.Google Scholar
  49. 49).
    McKay, H. A. C.: Uranium(IV) Nitrate as a Reducing Agent for Uranium-Plutonium Separation, in Aqueous Reprocessing Chemistry for Irradiated Fuels, S. 281. Brüssel 1963, ENEA-Eurochemic.Google Scholar
  50. 50).
    Talmont, L. R.: Utilization of Uranium (IV) Nitrate in a Second Cycle of Purification of Plutonium by TBP. In: Solvent Extraction Chemistry of Metals, S. 103. London: Macmillan 1965.Google Scholar
  51. 51).
    Regnault, P., D. Faugeras, A. Brut, R. Helon, and A. Redon: The Processing of Irradiated Uranium in the Fontenay aux Roses Pilot Plant. Proc. 2. Intern. Conf. Peaceful Uses At. Energy, Geneva 17, 73 (1958).Google Scholar
  52. 52).
    Karraker, D. G.: Temperature Effects on TBP Solvent Extraction Processes. Proc. 2. Intern. Conf. Peaceful Uses At. Energy, Geneva 17, 333 (1958).Google Scholar
  53. 53).
    Eurochemic First Activity Report 1959-1961. OECD-ENEA Paris 1961.Google Scholar
  54. 54).
    Stoller, S. M., and R. B. Richards: Reactor Handbook, Vol. II, S. 154/155. Fuel Processing (1961).Google Scholar
  55. 55).
    Hoffart, A. J., and R. D. Thomson: Effects of ZrO (NO3)2 and HNO3 upon the Chemical Stability of TBP. US-Report IDO-14643 (1967).Google Scholar
  56. 56).
    Hardy, C. J., and D. Scargill: Extraction of Zr from Nitrate Solution by DBP. J. Inorg. Nucl. Chem. 17, 337 (1961).CrossRefGoogle Scholar
  57. 57).
    Faugeras, P., and X. Talmont: Radiolysis and Hydrolysis of TBP and their Effects, Proc. XI. Intern. Conf. Coordination Chemistry 1968. Amsterdam: Elsevier Publ. Co. 1968.Google Scholar
  58. 58).
    Blake Jr., C. A.: Solvent Stability in Nuclear Fuel Processing, Evaluation of the Literature, Calculation of Radiation Dose and Effects of Iodine and Plutonium. US-Report ORNL-4212.Google Scholar
  59. 59).
    McKay, H. A. C.: TBP-Meeting-Point of Science and Technology, Solvent Extraction Chemistry, S. 185. Amsterdam: North-Holland Publ. Co. 1967.Google Scholar
  60. 60).
    Vogel, R. C, A. A. Jonke, and R. K. Steunenberg: The Non Aqueous Processing of Spent Fast Reactor Fuels, Symposium on Dry Reprocessing, Mol. 1968.Google Scholar
  61. 61).
    Jonke, A. A.: Reprocessing of Nuclear Reactor Fuels by Processes Based on Volatilization, Fractional Distillation, and Selective Adsorption. At. Energy Rev. 3, (1), 3–60 (1965).Google Scholar
  62. 62).
    Vogel, R. C., W. H. Carr, G. I. Gathers, J. Fischer, L. P. Hatch, R. W. Horton, A. A. Jonke, R. P. Milford, J. J. Reilly, and G. Strickland: Fluoride Volatility Processes for Recovery of Fissionable Materials from Irradiated Reactor Fuels. Proc. 3. Intern. Conf. Peaceful Uses At. Energy, Geneva 10, 491–498 (1964).Google Scholar
  63. 63).
    Schmets, J., G. Camozzo, A. Francesconi, P. Godrie, R. Heremans, G. Pierini, and P. Speeckaert: Reprocessing of Nuclear Fuels by Volatilization. Proc. 3. Intern. Conf. Peaceful Uses At. Energy, Geneva 10, 520–529 (1964).Google Scholar
  64. 64).
    Bourgeois, M., and P. Faugeras: The Processing of Irradiated Fuels by the Halogens and their Compounds. Proc. 3. Intern. Conf. Peaceful Uses At. Energy, Geneva 10, 483–490 (1964).Google Scholar
  65. 65).
    pierce, R. D., and L. Burris Jr.: Pyroprocessing of Reactor Fuels. In: Reactor Technology, Selected Reviews. US-Report TID-8540, S. 411–476 (July 1964).Google Scholar
  66. 66).
    Lawroski, S., and L. Burris Jr.: Processing of Reactor Fuel Materials by Pyrometallurgical Methods, At. Energy Rev. 2(3), 3–69 (1964).Google Scholar
  67. 67).
    Faugeras, P.: Chemical Treatment of Irradiated Nuclear Fuels. Energie Nucl. 7, 214–227 (1965).Google Scholar
  68. 68).
    Burris, L., K. M. Harmon, G. E. Brand, E. W. Murback, and R. K. Steunenberg: Pyrometallurgical and Pyrochemical Fuel Processing. Proc. 3. Intern. Conf. Peaceful Uses At. Energy, Geneva 10, 501–510 (1964).Google Scholar
  69. 69).
    Benedict, G. E., W. R. Bond, G. Jansen Jr., L. G. Morgan, and J. R. Lundquist: Status of the Salt Cycle Process for Processing Oxide Fuels. US-Report BNWL-SA-205 (Sept. 1965).Google Scholar
  70. 70).
    Burris, L., and G. A. Bennett: Dry Processes. Reactor Fuel Process. 9(1), 36–38 (1965–1966).Google Scholar
  71. 71).
    Reas, W. H.: The Aqua-Fluor Process. In: Reprocessing of Fuel from Present and Future Power Reactors. Kjeller-Report KR-126, S. 361.Google Scholar
  72. 72).
    -Commercial Nuclear Fuel Recovery. In: Reprocessing of Fuel from Present and Future Power Reactors. Kjeller-Report KR-126, S. 342.Google Scholar
  73. 73).
    Uriarte, A. L., and R. H. Rainey: Dissolution of High Density UO2, PuO2 and UO2-PuO2 Pellets in Inorganic Acids. US-Report ORNL-3695 (April 1965).Google Scholar
  74. 74).
    Bähr, W., u. T. Dippel: über die Auflösung von PuO2-haltigen Brüterbrennstoffen in Salpetersäure für die wäßrige Wiederaufarbeitung nach dem Purex-Verfahren. KFK-673 (1967).Google Scholar
  75. 75).
    Goode, J. H.: Hot CeU Dissolution of Highly Irradiated 20% PuO2-80 UO2 Fast Reactor Specimens. US-Report ORNL-3754 (October 1965).Google Scholar
  76. 76).
    Baumgärtel, G., W. Ochsenfeld u. H. Schmieder: Die Verteilung der Metallnitrate im System Pu(NO3)4-UO2(NO3)2-HNO3/TBP-Dodecan, KFK-680 (1967).Google Scholar
  77. 77).
    Ochsenfeld, W., H. Schmieder u. S. TheiΒ: Wäßrige Wiederaufarbeitung der Brennelemente Schneller Brüter. I. Die gemeinsame Extraktion und Trennung von Makromengen Plutonium und Uran im Purex-Prozeß. KFK-911 (1969).Google Scholar
  78. 78).
    Baumgärtner, F., W. Ochsenfeld, B. Roth u. L. Finsterwalder: Das Verhalten hoher Plutoniumkonzentrationen im Purex-Prozeß und die Entwicklung schneller Extraktoren zur Wiederaufarbeitung von Kernbrennstoffen. KFK-652 (1967).Google Scholar
  79. 79).
    Rainey, R. H.: Hydrogen Reduction of Pu(IV) to Pu(III). Nucl. Appl. 1, 310 (1965).Google Scholar
  80. 81).
    Stoller, S. M., and R. B. Richards: Reactor Handbook, Vol. II, S. 445. Fuel Processing (1961).Google Scholar
  81. 82).
    Jouannaud, C.: Discussion Fuel Reprocessing (I) Expérience de six années de fonctionnement de l'usine de retraitement de Marcoule. Proc. 3. Intern. Peaceful Uses At. Energy, Geneva 10, 215 (1965).Google Scholar
  82. 83).
    Nicholson, E. L.: Preliminary Investigation of Processing Fast-Reactor Fuel in an Existing Plant. US-Report ORNL-TM-1784 (May 1967).Google Scholar
  83. 84).
    Nuclear Fuel Services, Inc., Safety Analysis — Spent Fuel Reprocessing Plant, Vol. II (July 1962).Google Scholar
  84. 85).
    Lewis, W. H., M. E. Weech, and B. E. Knight: Criticality Control in the Nuclear Fuels Services Processing Plant. US-Report SC-DC-67-1305. Nuclear Criticality Safety Conference (ANS), Las Vegas, Nevada (December 1966).Google Scholar
  85. 86).
    U. S. Atomic Energy Commission, Division of Operational Safety, Operational Accidents and Radiation Exposure Experience within the USAEC, 1943–1964. U.S. Government Printing Office, Washington (April 1965).Google Scholar
  86. 87).
    Johnson, W. A.: Nuclear Safety of Fissile Material Outside Reactors, Nucl. Safety 8(1), 16–19 (1967).Google Scholar
  87. 88).
    Paxton, H. C.: The Nature and Consequences of Nuclear Accidents. US-Report SC-DC-67-1305. Nuclear Criticality Safety Conference (ANS), Las Vegas, Nevada (December 1966).Google Scholar
  88. 89).
    Flanary, J. R., F. H. Goode, M. J. Bradley. J. W. Ulmann, L. M. Ferris, and G. C. Wall: Hot Cell Studies of Aqueous Dissolution Processes for Irradiated Carbide Reactor Fuels. US-Report ORNL-3660 (Sept. 1964).Google Scholar
  89. 90).
    Ferris, L. M., and M. J. Bradley: Off Gases from the Reactions of Uranium Carbides with Nitric Acid at 90‡ C. Us-Report ORNL-3719 (December 1964).Google Scholar
  90. 91).
    Züst, H. E., H. R. v. Gunten u. P. Baertschi: Verfahren und Einrichtung zur Aufbereitung von in Kernreaktoren verbrauchten Spaltstoffelementen auf Carbidbasis. DBP-Auslegeschrift 1464647 (1963).Google Scholar
  91. 82).
    Stade u. Würgassen: Nucl. Eng. 12, 756 (1967).Google Scholar
  92. 93).
    Ochsenfeld, W.: Wiederaufarbeitung der Brennelemente Schneller Brutreaktoren. Atomwirtschaft 13, 422 (1968).Google Scholar
  93. 94).
    Shaw, M.: Current AEC Programms. In: Nuclear Power Fuel Processing — Technology and Economics, Conf-670542.Google Scholar
  94. 95).
    Europäisches Institut für Transuranelemente Karlsruhe, Progress Report No. 6, Juli–Dezember 1968, Kap. 6.1.2.Google Scholar
  95. 96).
    Blomeke, I. O., and M. F. Todd: Uranium-235 Fission-Product Production as a Function of Thermal Neutron Flux, Irradiation Time, and Decay Time. 1. Atomic Concentrations and Gross Totals. US-Report ORNL-2127, Part I, Vol. 1 (1957).Google Scholar
  96. 97).
    Stoller, S. M., and R. B. Richards: Reactor Handbook, Vol. II, S. 147. Fuel Processing (1961).Google Scholar
  97. 98).
    Flagg, J. F.: Chemical Processing of Reactor Fuels, S. 209. New York: Academic Press 1961.Google Scholar
  98. 99).
    Burris Jr., L., and G. Dillon: Estimation of Fission Product Spectra in discharged Fuel from Fast Reactors. US-Report ANL-5742 (1957).Google Scholar
  99. 100).
    Salomon, L., and E. Lopez-Menchero: Stability of HNO3-TBP Diluent Systems: Bibliography of Data Up to June 1966. Eurochemic-Report ETR-203.Google Scholar
  100. 101).
    Rigg, T., and W. Wild: Radiation Effects in Solvent Extraction Processes. Progress in Nuclear Energy, Series III, Vol. 2, S. 320. London: Pergamon Press 1958.Google Scholar
  101. 102).
    Scadden, E. M., and N. E. Ballon: Solvent Extraction Separations of Zirconium and Niobium. Anal. Chem. 11, 1602 (1953).CrossRefGoogle Scholar
  102. 103).
    Blake, C. A., A. T. Gresky, J. M. Schmitt, and R. G. Mansfield: Comparison of Dialkyl Phenylphosphonates with Tri-n-Butyl Phosphate in Nitrate Systerns: Extraction Properties, Stability, and Effect of Diluent on the Recovery of Uranium and Thorium from Spent Fuels. US-Report ORNL-3374 (1963).Google Scholar
  103. 104).
    Marston, A. L., D. L. West, and R. N. Wilhite: Selection, Cost and Performance of n-Paraffin Diluents. In: Solvent Extraction Chemistry of Metals, S. 213. London: Macmillan 1965.Google Scholar
  104. 105).
    Stieglitz, L., W. Ochsenfeld u. H. Schmieder: Der Einfluß der Radiolyse von TBP auf die Pu-Ausbeute im Purex Prozeß bei hohem Plutoniumgehalt. KFK-691 (1969).Google Scholar
  105. 106).
    Burger, L. L.: The Decomposition Reactions of Tributyl Phosphate and its Diluents and their Effects on Uranium Recovery Processes. In: Progress in Nuclear Energy, Series III, Vol. 2, S. 307. London: Pergamon Press 1958.Google Scholar
  106. 107).
    Moore, R. H.: Investigation of Solvent Degradation Products in Recycled Uranium Recovery Plant Solvent. US-Report HW-34502 (1955).Google Scholar
  107. 108).
    Wallace, R. M., and T. H. Siddell: Effect of Solvent Degradation on the Purex Process. US-Report DP-286 (1958).Google Scholar
  108. 109).
    Davies, W.: Solubilities of Uranyl and Iron(III) Dibutyl and Monobutyl Phosphates in TBP Solvent Extraction Solutions. US-Report ORNL-3084. (1961).Google Scholar
  109. 110).
    Naylor, A.: TBP Extraction Systems. 2. TBP and Diluent Regradation, in Reprocessing of Fuel from Present and Future Power Reactors. Kjeller-Report KR-126, S. 120.Google Scholar
  110. 111).
    Schlea, C. S., and A. S. Jennings: Behaviour of Actinides and Fission Products in Tri-n-butyl Phosphate and in Di-2-amyl 2 butylphosphonate Solvent Extraction Processes Using Short-residence Contactors. In: Solvent Extraction Chemistry of Metals, S. 81. London: Macmillan 1965.Google Scholar
  111. 112).
    Ochsenfeld, W., u. H. Schmieder: Verfahren zur vollständigen Rückextraktion von wertvollen vierwertigen Metallen, insb. Plutonium aus sauren Organophosphorverbindungen, die mit diesen vierwertigen Metallen besonders stabile Komplexe bilden. DBP-Anmeldung 1968.Google Scholar
  112. 113).
    Huggard, A. J., and B. F. Warner: Investigations to Determine the Extent of Degradation of TBP/Odorless Kerosene Solvent in the New Separation Plant Windscale. Nucl. Sci. Eng. 17, 638 (1963).Google Scholar
  113. 114).
    Lane, E. S.: Performance and Degradation of Diluents for TBP and the Cleanup of Degraded Solvents. Nucl. Sci. Eng. 17, 620 (1963).Google Scholar
  114. 115).
    Blake, C. A., W. Davis, and /. M. Schmitt: Properties of Degraded TBP-Amsco Solutions and Alternative Extractant-Diluent Systems. Nucl. Sci. Eng. 17, 626 (1963).Google Scholar
  115. 116).
    -and J. M. Schmitt: Stability of Aromatic Diluents and Solvent Extraction Reagents in Radiochemical Processing. In: Solvent Extraction Chemistry of Metals, S. 161. London: Macmillan 1965.Google Scholar
  116. 117).
    Orth, D. A., and T. W. Olcott: Purex Process Performance Versus Solvent Exposure and Treatment. Nucl. Sci. Eng. 17, 593 (1963).Google Scholar
  117. 118).
    -: Performance of 3% and 30% TBP Processes. In: Solvent Extraction Chemistry of Metals, S. 47. London: Macmillan 1965.Google Scholar
  118. 119).
    Barconcelli, F., and G. Grossi: Chemical Degradation of Aromatic Diluents Exposed to Nitric Acid Attack. In: Solvent Extraction Chemistry of Metals, S. 197. London: Macmillan 1965.Google Scholar
  119. 120).
    Lane, E. S.: Degraded TBP-Kerosene Clean-up. UKAEA-Report AERE-M. 809 (1961).Google Scholar
  120. 121).
    Cooper, V. R., and M. T. Wallung Jr.: Aqueous Processes for Separation and Decontamination of Irradiated Fuels. Proc. 2. Intern. Conf. Peaceful Uses At. Energy, Geneva 17, 291 (1958).Google Scholar
  121. 122).
    Rigg, T.: The Breakdown of Tri-n-Butyl Phosphate Solvents During the Processing of Extremely Radioactive Nuclear Fuels, Part I + II. UKAEA-Report IGR-R/W 203 und IGR-R/W 233 (1957).Google Scholar
  122. 123).
    Kennedy, J., J. W. A. Peckett, and J. M. Fletcher: The Influence of High Radiation Level on the Retention of Ruthenium in Processing Solvents. In: Solvent Extraction Chemistry of Metals, S. 187. London: Macmillan 1965.Google Scholar
  123. 124).
    Zebroski, E. L., H. W. Alter, and G. D. Collins: Plutonium Fuel Fabrication and Reprocessing for Fast Ceramic Reactors. US-Report GEAP-3876 (1962).Google Scholar
  124. 125).
    Centeno, J., and R. de Witte: Experiments in Eurochemic Extraction Evaporation Pilot Plant on First Cycle of HEU Flowsheet. Eurochemic-Report ETR-191 (1967).Google Scholar
  125. 126).
    Salomon, L., E. Verbecken, and E. Lopez-Menchero: Predictions on the Behaviour of First Cycle Solvent During the Processing of Highly Irradiated Fuel. Eurochemic-Report ETR-213 (1967).Google Scholar
  126. 127).
    Eargle, J. E., C. W. Swindell, and R. I. Martens: Large Scale Processing of Highly Irradiated Plutonium by Solvent Extraction and Ion Exchange, Ind. Eng. Chem. Process Design Develop. 6, 348 (1967).CrossRefGoogle Scholar
  127. 128).
    Allerdice, R. H.: Reprocessing of Fuel from Dounreay Fast Reactor. In: Reprocessing of Fuel from Present and Future Power Reactors. Kjeller-Report KR-126, S. 394.Google Scholar
  128. 129).
    Clark Jr., A. T.: Performances of a 10 inch Centrifugal Contactor. US-Report DP-752 (1962).Google Scholar
  129. 130).
    Kishbanyh, A. A.: Performances for a Multi-stage Centrifugal Contactor. US-Report DP-841 (1963).Google Scholar
  130. 131).
    Webster, D. S.: Hydraulic Performance of a 5 inch Centrifugal Contactor. US-Report DP-370 (1962).Google Scholar
  131. 132).
    Schlea, C. S., H. E. Henry, M. R. Caverly, and W. J. Jenkins: Purex Process Performance with Short Residence Time Contactors. US-Report DP-809 (1963).Google Scholar
  132. 133).
    Whatley, M. E., and W. M. Woods: The performance of an Advanced Experimental Stacked-Clone Contactor: A High-Performance Solvent Extraction Machine with Potential for Application to Very Highly Radioactive Solutions. US-Report ORNL-3533 (April 1964).Google Scholar
  133. 134).
    Whatley, M. W., W. S. Groenier, and W. M. Woods: Stacked-Clone Contactor: A High-Performance Hydroclone Solvent Extraction Device. US-Report ORNL-TM-1936 (August 7, 1967).Google Scholar
  134. 135).
    Groenier, W. S., and M. E. Whatley: Stacked-Clone Contactor. In: M. E. Whatley, P. A. Haas, R. W. Horson, A, D. Ryon, J. C. Suddath, and C. D. Watson: Unit Operations Section Quarterly Progress Report, April–June 1966, US-Report ORNL-4074 (April 1967).Google Scholar
  135. 136).
    Roth, B.: Zentrifugalextraktoren für die Wiederaufarbeitung von Kernbrennstoffen mit hohem Abbrand und Plutonium-Gehalt. KFK-862 (1969).Google Scholar
  136. 137).
    König, L. A., u. S. Zehme: Die maximal zulässige Aktivitätsabgabe über die Abluft einer kerntechnischen Anlage. KFK-340 (1965).Google Scholar
  137. 138).
    Merryman, J. R., and J. H. Pashley: Engineering Development of an Adsorption Process for the Concentration and Collection of Krypton and Xenon. US-Report K-1745 (1968).Google Scholar
  138. 139).
    Albenesius, E. L.: Tritium as a Product of Fission. Phys. Rev. Letters 3, 274 (1959).CrossRefGoogle Scholar
  139. 140).
    -and R. S. Ondrejoin: Nuclear Fission Produces Tritium. Nucleonics 18, No. 9, 100 (1960).Google Scholar
  140. 141).
    Sloth, E. N., D. L. Horrocks, E. J. Boyce, and M. H. Studier: Tritium in the Thermal Neutron Fission of Uranium-235. J. Inorg. Nucl. Chem. 24, 337 (1962).CrossRefGoogle Scholar
  141. 142).
    Mecham, W. J.: Studies and Evaluations: Problems of Tritium in Power Reactor Fuel Cycles. US-Report ANL-7375 (1967).Google Scholar
  142. 143).
    Dudey, N. D.: Review of Low Mass Atom Production in Fast Reactors. US-Report ANL-7434 (1968).Google Scholar
  143. 144).
    Nuclear Fuel Services, Inc., Spent Fuel Processing Plant-Safety Analysis. AEC-Docket No. 50-201 (October 1962).Google Scholar
  144. 145).
    Schmidt, W. C.: Treatment of Gaseous Effluents, Symposium on the Reprocessing of Irradiated Fuels, Brussels, May 20, 1957. US-Report TID-7534, S. 371.Google Scholar
  145. 146).
    Marter, W. L.: Radioiodine Release Incident at the Savannah River Plant. US-Report DPSPU-63-30-26B (May 1963).Google Scholar
  146. 147).
    Bruce, F. R.: The Behavior of Fission Products in Solvent Extraction Processes, Progress in Nuclear Energy. III. Process Chemistry, S. 134, 144; F. R. Bruce, J. M. Fletcher, and H. H. Hyman, editors. New York: McGraw-Hill 1956.Google Scholar
  147. 148).
    Culler, F. L.: Reprocessing of Reactor Fuel and Blanket Materials by Solvent Extraction, Progress in Nuclear Energy. III. Process Chemistry, S. 183; F. R. Bruce, J. M. Fletcher, and H. H. Hyman, editors. New York: McGraw-Hill 1956.Google Scholar
  148. 149).
    Browning Jr., W. E.: Removal of Fission Product Activity from Gases, Nucl. Safety 1 (3), 42 (1960).Google Scholar
  149. 150).
    Cowser, K. E.: Current Practices in the Release and Monitoring of I131 at NRTS, Hanford, Savannah River and ORNL. US-Report ORNL-NSIC-3 (August 1964).Google Scholar
  150. 151).
    Leger, B. M., B. M. Legler, S. F. Fairbourne, P. N. Kelly, and R. A. Robinson: Startup Operation of a Production Facility for Separating Ba-140 from MTR Fuel. US-Report IDO-14414, S. 70 (September 1957).Google Scholar
  151. 152).
    Paige, D. M., P. N. Kelly, and E. S. Grimmett: Two Gas Cleaning Problems at the Idaho Chemical Processing Plant Site. In; Fifth AEC Air Cleaning Conference, June 24–27, 1957, TID-7551, S. 17 (April 1958); Nucl. Safety 1(3), 41 (March 1960).Google Scholar
  152. 153).
    Cederberg, G. K., and D. K. MacQueen: Containment of Iodine-131 Released by the Rala Process. US-Report IDO-14566 (October 1961).Google Scholar
  153. 154).
    Bower, J. R., and G. K. Cederberg: Removal of Iodine from Ba40 Process Off-Gas. In: ICPP Technical Progress Report for July–Sept. 1958. US-Report IDO-14457, S. 71. (W. E. Browning, Nucl. Safety 1 (3), 43 (March 1960).Google Scholar
  154. 155).
    Schmidt, W. C.: Treatment of Gaseous Effluents, Symposium on the Reprocessing of Irradiated Fuels, Brussels, May 20, 1957. US-Report TID-7534, S. 371.Google Scholar
  155. 156).
    Blasewitz, A. G., and W. C. Schmidt: Treatment of Radioactive Waste Gases, P/1397. Proc. 2. Intern. Conf. Peaceful Uses At. Energy, Geneva 18, 184 (1958).Google Scholar
  156. 157).
    Michels, L. R.: Design and Operating Considerations for Off-Gas Systems in Nuclear Processing Plants. Chem. Eng. Syrap. Ser., No. 28, 56, 12 (1910).Google Scholar
  157. 158).
    Keilholtz, G. W., and C. J. Barton: Behavior of Iodine in Reactor Containment System. US-Report ORNL-NSIC-4, S. 64 (February 1965).Google Scholar
  158. 159).
    Browning Jr., W. E.: Removal of Radioiodine from Reactor Gases. Nucl. Safety 2 (3), 35 (1960/61).Google Scholar
  159. 160).
    -: Removal of Radioiodine from Gases. Nucl. Safety 4 (2), 83 (1963).Google Scholar
  160. 161).
    -: Removal of Radioiodine from Gases. Nucl. Safety 6 (3), 272 (1964/65).Google Scholar
  161. 162).
    McCormack, J. D.: Some Observations on Iodine Removal from Plant Streams with Charcoal. Presented at Eighth AEC Air Cleaning Conference October 22–25, 1963, Oak Ridge, Tennessee. US-Report TID-7677, S. 38.Google Scholar
  162. 163).
    Jacobsen, W. R., and L. Jolly: Measurement of Radioiodine in Purex Stack Gases. US-Report DPSPU-63-30-4 B (May 1963).Google Scholar
  163. 164).
    McHenry, R. E.: Removal of Radioiodine from Hot-Off-Gas System by Charcoal Absorption. Isotopes Development Center Progress Report December 1962–January 1963. US-Report ORNL-TM-532, S. 27.Google Scholar
  164. 165).
    Chamberlain, A. C., A. E. J. Eggleton, W. J. Megaw, and J. B. Morris: Physical Chemistry of Iodine and Removal of Iodine from Gas Streams. Nucl. Energy, Parts A/B 17, 519 (1963).CrossRefGoogle Scholar
  165. 166).
    Kovach, J. L.: Evaluation of the Ignition Temperature of Activated Charcoals in Dry Air. Nucl. Safety 8 (1), 41 (1966).Google Scholar
  166. 167).
    Adams, R. E., and W. E. Browning: Estimate of the Probability and Consequences of Ignition of the HRT Charcoal Beds. US-Report CF-58-6-6, S. 8 (June 1958).Google Scholar
  167. 168).
    -, W. E. Browning Jr., Wm. B. Cottrell, and G. W. Parker: The Release and Absorption of Methyl Iodide in the HFIR Maximum Credible Accident. US-Report ORNL-TM-1291, S. 24 (October 1, 1965).Google Scholar
  168. 169).
    Ackley, R. C, R. E. Adams, W. E. Browning Jr., G. E. Creek, and G. W. Parker: Retention of Methyl Iodide by Charcoal under Accident Conditions. In: Nuclear Safety Program Semi-Annual Progress Report for Period ending Dec. 31, 1965. US-Report ORNL-3915, S. 61.Google Scholar
  169. 170).
    Adams, R. E., R. C. Ackley, and W. E. Browning Jr.: Removal of Radioactive Methyl Iodide from Steam Air Systems. US-Report ORNL-4040 (January 1967).Google Scholar
  170. 171).
    Colline, D. A., L. R. Taylor, and R. Taylor: The Development of Impregnated Charcoals for Trapping Methyl Iodide at High Humidity, presented at Ninth AEC Air Cleaning Conference, Boston, Mass., Sept. 13–16, 1966. US-Report CONF-660904, S. 159.Google Scholar
  171. 172).
    Scharmann, A.: Radionuklidbatterien in Energiedirektumwandlung. München: K. J. Euler 1967.Google Scholar
  172. 173).
    Seaborg, G. T.: Mass Production and Practical Applications of Actinide Elements. Isotopes Radiation Technol. 6, 1 (1968).Google Scholar
  173. 174).
    Industrial Applications for Isotopic Power Generators. Intern. Symposium AERE, Harwell 1967 (ENEA).Google Scholar
  174. 175).
    Born, H. J.: Aktiniden für Radionuklidbatterien. Atompraxis 15, 32 (1969).Google Scholar
  175. 176).
    Roberts, F. P., and H. H. Van Tuyl: Promethium-146, Fission Product, and Transuranium Isotope Content of Power Reactor Fuels. US-Report BNWL-45 (1965).Google Scholar
  176. 177).
    Vondy, D. R., J. A. Lane, and A. T. Gresky: Production of Np237 and pu238 in Thermal Power Reactors. Ind. Eng. Chem. Process Design Develop. 3, 293 (1964).CrossRefGoogle Scholar
  177. 178).
    Deonigi, D. E., and E. A. Eschbach: Production and Indifference Pricing of Transuranium Isotopes. US-Report BNWL-223 (1966).Google Scholar
  178. 179).
    Bähr, W.: Extraktionsverhalten von Neptunium bei der Wiederaufarbeitung von bestrahlten Kernbrennstoffen nach dem Purex-Prozeß. KFK-797 (1969).Google Scholar
  179. 180).
    Isaacson, R. E., and B. F. Judson: Ind. Eng. Chem. Process Design Develop. 3, 296 (1964).CrossRefGoogle Scholar
  180. 181).
    Poe, W. L., A. W. Joyce, and J. R. Martens: Np237 and Pu238 Separation of Savannah River Plant. Ind. Eng. Chem. Process Design Develop. 3, 314 (1964).CrossRefGoogle Scholar
  181. 182).
    Savannah River Laboratory, Large Scale Production and Applications of Radioisotopes. US-Report DP-1066, Vol. II, 1966.Google Scholar
  182. 183).
    Ferguson, D. E.: ORNL Transuranium Program. The production of Transuranium Elements. Nucl. Sci. Eng. 17, 435 (1963).Google Scholar
  183. 184).
    -: Transuranium-Element Processing. Isotopes Radiation Technol. 4 (4), 321 (1967).Google Scholar
  184. 185).
    Ryan, V. A., and J. W. Pringle: Preparation of Pure Americium. US-Report RFP-130 (1960).Google Scholar
  185. 186).
    Kingsley, R. S.: A Multi-Column Ion Exchange Purification-Concentration Process for Americium. US-Report RL-SEP-729 REV (1965).Google Scholar
  186. 187).
    Koch, G., u. J. Schön: Isolierung von Americium-241 aus Plutonium-Schrott durch Extraktion mit Tricaprylmethylammoniumnitrat. KFK-783 (1968).Google Scholar
  187. 188).
    Wheelwright, E. J., F. B. Roberts, L. A. Bray, G. L. Ritter, and L. A. Bolt: Simultaneous Recovery and Purification of Pm, Am, and Cm by the Use of Alternating DTPA and NTA Cation-Exchange Flowsheets. US-Report BNWL-SA-1492 (1968).Google Scholar
  188. 189).
    Höhlein, B. G., H. J. Born u. W. Weinländer: Die Isolierung von Cm-242 aus neutronenbestrahltem Am-241. Radiochim. Acta 10, 85 (1969).Google Scholar
  189. 190).
    Müller, W.: Die Gewinnung von Transplutoniumelementen aus bestrahltem Americium-241. Atompraxis 15, 35 (1969).Google Scholar
  190. 191).
    Eubanks, J. D., and G. A. Burney: Curium Process Development, I. General Purpose Description. US-Report DP-1009 (1966).Google Scholar
  191. 192).
    Henry, H. E.: Isolation of Americium and Curium from Al(NO3)3-NaNO3-HNO3 Solutions by Batch Extraction with Tributyl Phosphate. US-Report DP-972 (1965).Google Scholar
  192. 193).
    Moore, F. L.: Improved Extraction Method for Isolation of Trivalent Actinide-Lanthanide Elements from Nitrate Solutions. Anal. Chem. 38, 510 (1966).CrossRefGoogle Scholar
  193. 194).
    Ooyen, J. van: Quarternary Ammonium Nitrates as Extractants for Trivalent Actinides. In: Solvent Extraction Chemistry, S. 485. Amsterdam: North-Holland Publ. Co. 1967.Google Scholar
  194. 195).
    Horwitz, E. P., C. A. A. Bloomquist, L. J. Sauro, and D. J. Henderson: The Liquid-Liquid-Extraction of Certain Tripositive Transplutonium Ions from Salted Nitrate Solutions with a Tertiary and Quaternary Amine, J. Inorg. Nucl. Chem. 28, 2312 (1966).Google Scholar
  195. 196).
    Peppard, D. F., G. W. Mason, W. J. Driscoll, and R. J. Sironen: Acidic Esters of Orthophosphoric Acid as Selective Extractants for Metallic Cations — Tracer Studies. J. Inorg. Nucl. Chem. 7, 276 (1958).CrossRefGoogle Scholar
  196. 197).
    ---, and S. McCarty: Application of Phosphoric Acid Esters to the Isolation of Certain Trans-Plutonides by Liquid-Liquid Extraction. J. Inorg. Nucl. Chem. 12, 141 (1959).CrossRefGoogle Scholar
  197. 198).
    Baes Jr., C. F.: The Extraction of Metallic Species by Dialkylphosphoric Acids. J. Inorg. Nucl. Chem. 24, 707 (1962).CrossRefGoogle Scholar
  198. 199).
    Leuze, R. E., R. D. Baybarz, and B. Weaver: Application of Amine and Phosphonate Extractants to Transplutonium Element Production. Nucl. Sci. Eng. 17, 252 (1963).Google Scholar
  199. 200).
    Baybarz, R. D., B. S. Weaver, and H. B. Kinser: Isolation of Transplutonium Elements by Tertiary Amine Extraction. Nucl. Sci. Eng. 17, 457 (1963).Google Scholar
  200. 201).
    Leuze, R. D., R. D. Baybarz, F. A. Kappelmann, and B. Weaver: Behaviour of the Transplutonium Elements in Solvent Extraction Systems. In: Solvent-Extraction Chemistry of Metals, S. 423. London: Macmillan 1965.Google Scholar
  201. 202).
    Moore, F. L.: New Approach to Separation of Trivalent Actinide Elements from Lanthanide Elements — Selective Liquid-Liquid Extraction with Tricaprylmethylammonium Thiocyanate. Anal. Chem. 36, 2158 (1964).CrossRefGoogle Scholar
  202. 203).
    Gerontopulos, P. Th., L. Rigali, and P. G. Barbano: Separation of Americium-(III) from Lanthanides by Quaternary Ammonium Salt Extraction. Radiochim. Acta 4, 75 (1965).Google Scholar
  203. 204).
    Lloyd, M. H., and R. E. Leuze: Anion. Exchange Separation of Trivalent Actinides and Lanthanides. Nucl. Sci. Eng. 11, 274 (1961).Google Scholar
  204. 205).
    -: An Anion Exchange Process for Americium-Curium Recovery from Plutonium Process Waste. Nucl. Sci. Eng. 17, 452 (1963).Google Scholar
  205. 206).
    Coleman, J. S., R. A. Penneman, T. K. Keenan, L. E. Lamar, D. E. Armstrong, and L. B. Asprey: An Anion-Exchanger Process for Gram-Scale Separation of Americium from Rare Earth. J. Inorg. Nucl. Chem. 3, 327 (1957).CrossRefGoogle Scholar
  206. 207).
    Keenan, T. J.: Rapid and Efficient Purification of Americium. J. Inorg. Nucl. Chem. 20, 185 (1961).CrossRefGoogle Scholar
  207. 208).
    Higgins, G. H., and W. W. T. Crane: Proc. 2. Intern. Conf. Peaceful Uses At. Energy, Geneva 17, 245 (1958).Google Scholar
  208. 209).
    Weaver, B., and F. A. Kappelmann: TALSPEAK: A New Method of Separat ing Americium and Curium from the Lanthanides by Extraction from an Aqueous Solution of an Aminopolyacetic Acid Complex with a Monoacidic Organophosphate or Phosphonate. US-Report ORNL-3559 (1964).Google Scholar
  209. 210).
    --: Preferential Extraction of Lanthanides over Trivalent Actinides by Monoacidic Organophosphates from Carboxylic Acids and from Mixtures of Carboxylic and Aminopolyacetic Acids. J. Inorg. Nucl. Chem. 30, 263 (1968).CrossRefGoogle Scholar
  210. 211).
    Koch, G.: A Study of Americium-Curium Recovery from Fuel Processing High Level Waste Solutions. In: Solvent Extraction Chemistry, im Druck Proc. V. Intern. Conf. Solvent Extraction Chemistry, 1968. New York: John Wiley 1969.Google Scholar
  211. 212).
    Adar, S., R. K. Sjomblom, R. F. Barnes, P. R. Fields, E. K. Hulet, and H. D. Wilson: Ion-Exchange Behaviour of the Transuranium Elements in LiNO3 Solutions. J. Inorg. Nucl. Chem. 25, 447 (1963).Google Scholar
  212. 213).
    Marcus, Y., M. Givon, and G. R. Choppin: Anion Exchange of Metal Complexes XII — The Actinide(III)-Nitrate System. J. Inorg. Nucl. Chem. 25, 1457 (1963).CrossRefGoogle Scholar
  213. 214).
    Horwitz, E. P., K. A. Orlandini, and C. A. A. Bloomquist: The Separation of Americium and Curium by Extraction Chromatography Using a High Molecular Weight Quaternary Ammonium Nitrate. Inorg. Nucl. Chem. Letters 2, 87 (1966).CrossRefGoogle Scholar
  214. 215).
    Bockkarev, V. A., et E. N. Voevodin: Séparation de 1'Americium et du Curium par la Méthode d'échange Anionique avec Utilisation des Solutions concentant une Mélange d'Alcohol Methylique et d'Acide Azotique comme Eluant. Radiokhimija 7, 461 (1965); Frz. übersetzung: Radiochimie 7, 485 (1965).Google Scholar
  215. 216).
    Burney, G. A.: Nucl. Appl. 4, 217 (1968).Google Scholar

Copyright information

© Springer-Verlag 1969

Authors and Affiliations

  • F. Baumgärtner
    • 1
  • H. Philipp
    • 2
  1. 1.Kernforschungszentrum KarlsruheLehrstuhl für Radiochemie der Universität HeidelbergDeutschland
  2. 2.Literaturabteilung des Kernforschungszentrums KarlsruheDeutschland

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