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基于铝箔封装碳酸锂探测片产氚率测量的液闪样品制备方法

Preparation method of liquid scintillation sample for tritium production rate measurement based on lithium carbonate pellet with aluminum foil coat
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摘要 氚增殖包层中产氚率的测量是聚变核能系统中需要研究的重要问题之一,本文开展了用于产氚率测量的Al箔封装碳酸锂探测片液闪样品制备化学处理方法的研究。结果表明,首先采用氢氧化钠溶液来溶解Al箔,然后再用盐酸溶解碳酸锂探测片的溶解方式,能制成透明度高且无分层的液闪样品。为了提高测氚计数效率和保证样品兼容性,对20 m L的标准液闪样品,闪烁液体积应至少取12 m L,同时还应将液闪样品保持在10-20°C范围进行储存和测量。 Background: The measurement of tritium production rate for tritium breeding blanket is one key issue of fusion system. Purpose: A preparation method of liquid scintillation sample for tritium production rate meastlrement based on lithium carbonate pellet with aluminum foil coat was studied. Methods: According to the basic principle of liquid scintillation technique counting, it is necessary to bring lithium carbonate pellet with aluminum foil coat into the scintillation solution by chemistry process. Results & Conclusion: From the results, the liquid scintillation sample prepared by sodium hydroxide and hydrochloric acid can keep clear fluid without the liquid-liquid separation. To improve tritium counting efficiency and with a good compatibility, liquid scintillation sample for 20 mL should keep least 12-mL liquid scintillation and the temperature during the process of preparation and measurement must be 10-20 ℃ centigrade.
出处 《核技术》 CAS CSCD 北大核心 2017年第6期27-31,共5页 Nuclear Techniques
基金 国家磁约束核聚变能发展专项(No.2015GB108000) 高增益包层实验模块中子学性能实验研究项目(No.2015GB108006)资助~~
关键词 Al箔 碳酸锂 液闪样品 产氚率测量 Aluminum foil, Li2CO3 pellet, Liquid scintillation sample, Tritium production rate measurement
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  • 1Sawan M E , et al. Physics and technology conditions for attaining tritium self-sufficiency for the DT fuel cycle [J].Fusion Engineering and Design, 2006, 81:1131- 1144.
  • 2Abdou M, Void E, Gung C, Youssef M, Shin K. Deuteriumtritium fuel self-sufficiency in fusion reactors[J]. Fusion Technology, 1986,9 : 250-285.
  • 3Billone Michael C, et al. Status of beryllium development for fusion applications[R], presented at the Third International Symposium on Fusion Nuclear Technology. University of California at Los Angeles, 1994.
  • 4Pint B A, et al. Liquid metal compatibility issues for test blanket modules[J].Fusion Engineering and Design, 2006, 81: 901-908.
  • 5Norajitra P, et al. The EU advanced lead lithium blanket concept using SiCf/SiC flow channel inserts as electrical and thermal insulators[J]. Fusion Engineering and Design , 2001,58-59:629-634.
  • 6Moriyama H, et al. Molten salts in fusion nuclear technology[J]. Fusion Engineering and Design, 1998, 39- 40: 627-637.
  • 7Kulsartov T V, et al. Investigation of hydrogen isotope permeation through F82H steel with and without a ceramic coating of Cr2O3-SiO2 including CrPO4 (out-of- pile tests)[J].Fusion Engineering and Design, 2006, 81: 701-705.
  • 8Reimann J, Hermsmeyer S. Thermal conductivity of compressed ceramic breeder pebble beds[J].Fusion Engineering and Design, 2002, 61-62:345-351.
  • 9Muroga T, et al. Blanket neutronics of Li/vanadium-alloy and Flibe/vanadium-alloy systems for FFHR[J]. Fusion Engineering and Design, 2006, 81:1203-1209.
  • 10Maisonnier D, et al. Power plant conceptual studies in Europe[J]. Nuclear Fusion, 2007,47:1524-1532.

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