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  • ASTM
    C1347-08(2014)e1 Standard Practice for Preparation and Dissolution of Uranium Materials for Analysis
    Edition: 2014
    $93.60
    Unlimited Users per year

Description of ASTM-C1347 2014

ASTM C1347-08-e1-Reapproved2014

Historical Standard: Standard Practice for Preparation and Dissolution of Uranium Materials for Analysis




ASTM C1347

Scope

1.1 This practice covers dissolution treatments for uranium materials that are applicable to the test methods used for characterizing these materials for uranium elemental, isotopic, and impurities determinations. Dissolution treatments for the major uranium materials assayed for uranium or analyzed for other components are listed.

1.2 The treatments, in order of presentation, are as follows:

Procedure Title

Section

Dissolution of Uranium Metal and Oxide with Nitric Acid

8.1

Dissolution of Uranium Oxides with Nitric Acid and Residue
  Treatment

8.2

Dissolution of Uranium-Aluminum Alloys in Hydrochloric Acid
 with Residue Treatment

8.3

Dissolution of Uranium Scrap and Ash by Leaching with Nitric
 Acid and Treatment of Residue by Carbonate Fusion

8.4

Dissolution of Refractory Uranium-Containing Material by
 Carbonate Fusion

8.5

Dissolution of Uranium—Aluminum Alloys
Uranium Scrap and Ash, and Refractory
Uranium-Containing Materials by
Microwave Treatment

8.6


1.3 The values stated in SI units are to be regarded as standard. No other units of measurement are included in this standard.

1.4 This standard does not purport to address all of the safety concerns, if any, associated with its use. It is the responsibility of the user of this standard to establish appropriate safety and health practices and determine the applicability of regulatory limitations prior to use. Specific hazards statements are given in Section 7.


Keywords

dissolution; microwave dissolution; microwave oven; uranium-aluminum alloy; uranium ash; uranium dissolution; uranium metal; uranium oxide; uranium scrap                   


ICS Code

ICS Number Code 27.120.30 (Fissile materials and nuclear fuel technology)


DOI: 10.1520/C1347-08R14E01

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