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ASTM C1462-00(2005)

(Specification)

Standard Specification for Uranium Metal Enriched to More than 15% and Less Than 20%235 U

Standard Specification for Uranium Metal Enriched to More than 15% and Less Than 20%<sup>235</sup> U

SCOPE
1.1 This specification covers nuclear grade uranium metal that has either been processed through an enrichment plant, or has been produced by the blending of highly enriched uranium with other uranium, to obtain uranium of any  235U concentration below 20 % (and greater than 15 %) and that is intended for research reactor fuel fabrication. The scope of this specification includes specifications for enriched uranium metal derived from commercial natural uranium, recovered uranium, or highly enriched uranium. Commercial natural uranium, recovered uranium and highly enriched uranium are defined in Section . The objectives of this specification are to define the impurity and uranium isotope limits for commercial grade enriched uranium metal.
1.2 This specification is intended to provide the nuclear industry with a standard for enriched uranium metal which is to be used in the production of research reactor fuel. In addition to this specification, the parties concerned may agree to other appropriate conditions.
1.3 The scope of this specification does not comprehensively cover all provisions for preventing criticality accidents or requirements for health and safety or for shipping. Observance of this standard does not relieve the user of the obligation to conform to all applicable international, federal, state, and local regulations for processing, shipping, or any other way of using uranium metal (see, for example, C 996 regarding references).

General Information

Status
Historical
Publication Date
31-May-2005
ICS
27.120.30 - Fissile materials and nuclear fuel technology
Technical Committee
C26 - Nuclear Fuel Cycle
Drafting Committee
C26.02 - Fuel and Fertile Material Specifications
Current Stage
Ref Project

Relations

Replaces
ASTM C1462-00 - Standard Specification for Uranium Metal Enriched to More than 15% and Less Than 20%<sup>235</sup> U
Effective Date
01-Jun-2005
Replaced By
ASTM C1462-00(2008) - Standard Specification for Uranium Metal Enriched to More than 15 % and Less Than 20 % <sup> 235</sup>U
Effective Date
01-Jan-2008

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ASTM C1462-00(2005) - Standard Specification for Uranium Metal Enriched to More than 15% and Less Than 20%<sup>235</sup> UASTM C1462-00(2005) - Standard Specification for Uranium Metal Enriched to More than 15% and Less Than 20%<sup>235</sup> U
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ASTM C1462-00(2005) - Standard Specification for Uranium Metal Enriched to More than 15% and Less Than 20%<sup>235</sup> U
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Standards Content (Sample)


NOTICE: This standard has either been superseded and replaced by a new version or withdrawn.
Contact ASTM International (www.astm.org) for the latest information
Designation: C 1462 – 00 (Reapproved 2005)
Standard Specification for
Uranium Metal Enriched to More than 15 % and Less Than
20 % U
This standard is issued under the fixed designation C 1462; the number immediately following the designation indicates the year of
original adoption or, in the case of revision, the year of last revision. A number in parentheses indicates the year of last reapproval. A
superscript epsilon (e) indicates an editorial change since the last revision or reapproval.
1. Scope C 859 Terminology Relating to Nuclear Materials
C 996 Specification for Uranium Hexafluoride Enriched to
1.1 This specification covers nuclear grade uranium metal
less than 5 % U
that has either been processed through an enrichment plant, or
C 1233 Practice for Determining the Equivalent Boron
has been produced by the blending of highly enriched uranium
Content of Nuclear Material
with other uranium, to obtain uranium of any U concentra-
C 1295 Test Method for Gamma Energy Emission from
tion below 20 % (and greater than 15 %) and that is intended
Fission Products in Uranium Hexafluoride and Uranium
for research reactor fuel fabrication. The scope of this speci-
Solutions
fication includes specifications for enriched uranium metal
C 1347 PracticeforPreparationandDissolutionofUranium
derived from commercial natural uranium, recovered uranium,
Materials for Analysis
or highly enriched uranium. Commercial natural uranium,
2.2 ANSI Standard
recovered uranium and highly enriched uranium are defined in
ANSI-ASME NQA-1 Quality Assurance Program Require-
Section 3. The objectives of this specification are to define the
ments for Nuclear Facility Applications
impurity and uranium isotope limits for commercial grade
2.3 U.S. Government Documents
enriched uranium metal.
Code of Federal Regulations, Title 10, Part 50, (Appendix
1.2 This specification is intended to provide the nuclear
B)
industrywithastandardforenricheduraniummetalwhichisto
be used in the production of research reactor fuel. In addition
3. Terminology
to this specification, the parties concerned may agree to other
3.1 Definitions of Terms Specific to This Standard—Terms
appropriate conditions.
shall be defined in accordance withTerminology C 859, except
1.3 The scope of this specification does not comprehen-
for the following:
sively cover all provisions for preventing criticality accidents
3.1.1 commercial grade enriched uranium metal—uranium
or requirements for health and safety or for shipping. Obser-
metal derived from commercial natural uranium, recovered
vance of this standard does not relieve the user of the
uranium, or uranium obtained from the blending of highly
obligation to conform to all applicable international, federal,
enriched uranium with commercial natural uranium, recovered
state, and local regulations for processing, shipping, or any
uranium, or depleted uranium.
other way of using uranium metal (see, for example, C 996
3.1.2 commercial natural uranium—any form of natural
regarding references).
unirradiated uranium (containing 0.711 + 0.004 g U per 100
2. Referenced Documents g U).
2 3.1.3 depleted uranium—any form of unirradiated uranium
2.1 ASTM Standards:
with a U content less than commercial natural uranium.
C 696 Test Methods for Chemical, Mass Spectrometric, and
3.1.4 highly enriched uranium—any form of uranium hav-
Spectrochemical Analysis of Nuclear-Grade Uranium Di-
ing a U content equal to or in excess of 20 %.
oxide Powers and Pellets
3.1.5 recovered uranium—any form of uranium that has
C 799 Test Methods for Chemical, Mass Spectrometric,
been exposed in a neutron irradiation facility and either has
Spectrochemical, Nuclear, and Radiochemical Analysis of
been subsequently chemically separated from the fission prod-
Nuclear-Grade Uranyl Nitrate Solutions
ucts and transuranic isotopes so generated, or may be used as
is, due to low irradiation levels.
This specification is under the jurisdiction of ASTM Committee C26 on
Nuclear fuel Cycle and is the direct responsibility of Subcommittee C26.02 on Fuel
and Fertile Material Specifications. Available from theAmerican National Standards Institute, 11 W. 42nd St, 13th
Current edition approved June 1, 2005. Published December 2005. Originally Floor, New York, NY 10036.
approved in 2000. Last previous edition approved in 2000 as C 1462–00. Available from the Superintendent of Documents, U.S. Government Printing
Annual Book of ASTM Standards, Vol 12.01. Office, Washington, DC 20402.
Copyright © ASTM International, 100 Barr Harbor Drive, PO Box C700, West Conshohocken, PA 19428-2959, United States.
C 1462 – 00 (2005)
3.1.5.1 Discussion—The requirements for recovered ura- 5.1.5 Equivalent Boron Content—For research reactor use,
nium are intended to be typical of reprocessed spent fuel from the total equivalent boron content (EBC) shall not exceed 4.0
research reactors that have achieved burn-up levels of up to µg/gU. The list of elements to be considered for the EBC
50 % of the originally contained fissile material or may have calculation shall be agreed upon between the buyer and the
been utilized in critical facilities. It is recognized that different sell
...

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ASTM C1295-24(Test Method)
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1.2 The lower limit of detection is estimated at 5000 MeV Bq/kg (MeV kg-1/s-1) of uranium and is the square root of the sum of the squares of the individual reporting limits of the nuclides to be measured. The limit of detection was determined on a pure, aged natural uranium (ANU) solution. The value is dependent upon the detector efficiency and background that can be achieved.  
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1.1 This terminology standard covers terms, definitions, descriptions of terms, nomenclature, and explanations of acronyms and symbols specifically associated with standards under the jurisdiction of Committee C26 on Nuclear Fuel Cycle. The content of this terminology standard may also be applicable to documents not under the jurisdiction of Committee C26, in which case this terminology standard may be referenced in those documents.  
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1.3 In general, technical terms that are defined in common dictionaries would not also be defined in this terminology standard unless there is a need to emphasize a specific definition in making appropriate use of a Committee C26 standard.  
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ASTM C1672-23(Test Method)
Standard Test Method for Determination of the Uranium, Plutonium or Americium Isotopic Composition or Concentration by the Total Evaporation Method Using a Thermal Ionization Mass Spectrometer

SIGNIFICANCE AND USE
5.1 The total evaporation method is used to measure the isotopic composition of uranium, plutonium, and americium materials, and may be used to measure the elemental concentrations of these elements when employing the IDMS technique.  
5.2 Uranium and plutonium compounds are used as nuclear reactor fuels. In order to be suitable for use as a nuclear fuel the starting material must meet certain criteria, such as found in Specifications C757, C833, C753, C776, C787, C967, C996, or as specified by the purchaser. The uranium concentration, plutonium concentration, or both, and isotope abundances are measured by TIMS following this method.  
5.3 Americium-241 is the decay product of 241Pu isotope. The abundance of the 241Am isotope together with the abundance of the 241Pu parent isotope can be used to estimate radio-chronometric age of the Pu material for nuclear forensic applications Ref (6). The americium concentration and isotope abundances are measured by TIMS following this method.  
5.4 The total evaporation method allows for a wide range of sample loading with no significant change in precision or accuracy. The method is also suitable for trace-level loadings with some loss of precision and accuracy. The total evaporation method and modern instrumentation allow for the measurement of minor isotopes using ion counting detectors, while the major isotope(s) is(are) simultaneously measured using Faraday cup detectors.  
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1.1 This method describes the determination of the isotopic composition, or the concentration, or both, of uranium, plutonium, and americium as nitrate solutions by the total evaporation method using a thermal ionization mass spectrometer (TIMS) instrument. Purified uranium, plutonium, or americium nitrate solutions are deposited onto a metal filament and placed in the mass spectrometer. Under computer control, ion currents are generated by heating of the filament(s). The ion currents are continually measured until the whole deposited solution sample is exhausted. The measured ion currents are integrated over the course of the measurement and normalized to a reference isotope ion current to yield isotope ratios.  
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ASTM C1432-23(Test Method)
Standard Test Method for Determination of Impurities in Plutonium: Acid Dissolution, Ion Exchange Matrix Separation, and Inductively Coupled Plasma-Atomic Emission Spectroscopic (ICP/AES) Analysis

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5.1 This test method can be used on plutonium matrices in nitrate solutions.  
5.2 This test method has been validated for all elements listed in Test Methods C757 except sulfur (S) and tantalum (Ta).  
5.3 This test method has been validated for all of the cation elements measured in Table 1. Phosphorus (P) requires a vacuum or an inert gas purged optical path instrument.
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1.1 This test method covers the determination of 25 elements in plutonium (Pu) materials. The Pu is dissolved in acid, the Pu matrix is separated from the target impurities by an ion exchange separation, and the concentrations of the impurities are determined by inductively coupled plasma-atomic emission spectroscopy (ICP-AES).  
1.2 This test method is specific for the determination of impurities in 8 M HNO3 solutions. Impurities in other plutonium materials, including plutonium oxide samples, may be determined if they are appropriately dissolved (see Practice C1168) and converted to 8 M HNO3  solutions.  
1.3 The values stated in SI units are to be regarded as standard. The values given in parentheses are mathematical conversions that are provided for information only and are not considered standard. Additionally, the non-SI units of molarity and centimeters of mercury are to be regarded as standard.  
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ASTM C1807-15(2023)(Guide)
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5.1 This guide assists in satisfying requirements in such areas as safeguards, SNM inventory control, nuclear criticality safety, waste disposal, and decontamination and decommissioning (D&D). This guide can apply to the measurement of holdup in process equipment or discrete items whose neutron production properties may be measured or estimated. These methods may meet target accuracy for items with complex distributions of SNM in the presence of moderators, absorbers, and neutron poisons; however, the results are subject to larger measurement uncertainties than measurements of less complex items.  
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1.1 This guide describes passive neutron measurement methods used to nondestructively estimate the amount of neutron-emitting special nuclear material compounds remaining as holdup in nuclear facilities. Holdup occurs in all facilities in which nuclear material is processed. Material may exist, for example, in process equipment, in exhaust ventilation systems, and in building walls and floors.  
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1.2 This guide includes information useful for management, planning, selection of equipment, consideration of interferences, measurement program definition, and the utilization of resources.  
1.3 Counting modes include both singles (totals) or gross counting and neutron coincidence techniques.  
1.3.1 Neutron holdup measurements of uranium are typically performed on neutrons emitted during (α, n) reactions and spontaneous fission using singles (totals) or gross counting. While the method does not preclude measurement using coincidence or multiplicity counting for uranium, measurement efficiency is generally not sufficient to permit assays in reasonable counting times.  
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ASTM C1168-23(Practice)
Standard Practice for Preparation and Dissolution of Plutonium Materials for Analysis

SIGNIFICANCE AND USE
5.1 Plutonium and uranium mixtures are used as nuclear reactor fuels. For use as a nuclear reactor fuel, the material must meet certain criteria for combined uranium and plutonium content, effective fissile content, and impurity content as described in Specifications C757 and C833. After dissolution using one of the procedures described in this practice, the material is assayed for plutonium and uranium to determine if the content is correct as specified by the purchaser.  
5.2 Unique plutonium materials, such as alloys, compounds, and scrap metals, are typically dissolved with various acid mixtures or by fusion with various fluxes. Many plutonium salts are soluble in hydrochloric acid. One or more of the procedures included in this practice may be effective for some of these materials; however their applicability to a particular plutonium material shall be verified by the user.
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1.1 This practice is a compilation of dissolution techniques for plutonium materials that are applicable to the test methods used for characterizing these materials. Dissolution treatments for the major plutonium materials assayed for plutonium or analyzed for other components are listed. Aliquots of the dissolved materials are dispensed on a mass basis when one of the analyses must be of high precision, such as plutonium assay; otherwise they are dispensed on a volume basis.  
1.2 Procedures in this practice are intended for the dissolution of plutonium metal, plutonium oxide, and uranium-plutonium mixed oxides. Aliquots of dissolved materials are analyzed using test methods, such as those developed by Subcommittee C26.05 on Methods of Test, to demonstrate compliance with applicable requirements. These may include product specifications such as Specifications C757 and C833.  
1.3 One or more of the procedures in this practice may be applicable to unique plutonium materials, such as alloys, compounds, and scrap materials. The user must determine the applicability of this practice to such materials.  
1.4 The treatments, in order of presentation, are as follows:    
Procedure Number  
Procedure Title  
Section  
1  
Dissolution of Plutonium Metal with Hydrochloric Acid at Room Temperature  
9  
2  
Dissolution of Plutonium Metal with Hydrochloric Acid and Heating  
10  
3  
Dissolution of Plutonium Metal with Sulfuric Acid  
11  
4  
Dissolution of Plutonium Oxide and Uranium-Plutonium Mixed
Oxide by the Sealed-Reflux Technique  
12  
5  
Dissolution of Plutonium Oxide and Uranium-Plutonium Mixed Oxides by Sodium Bisulfate Fusion  
13  
6  
Dissolution of Uranium-Plutonium Mixed Oxides and Low-Fired Plutonium Oxide in Beakers  
14  
7  
Open-Vessel (with Reflux Condenser) Dissolution of Plutonium Oxide Powder  
15  
8  
Open-Vessel (with Reflux Condenser) Dissolution of Mixed Oxide Powder  
16  
9  
Closed-Vessel Hot Block Dissolution of Plutonium Oxide Powder  
17  
10  
Open-Vessel (with Reflux Condenser) Dissolution of Mixed Oxide Pellets  
18  
1.5 The values stated in SI units are to be regarded as standard. The non-SI unit of molarity (M) is also to be regarded as standard. Values in parentheses (non-SI units), where provided, are for information only and are not considered standard.  
1.6 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, health, and environmental practices and determine the applicability of regulatory limitations prior to use.  
1.7 This international standard was developed in accordance with internationally recognized principles on standardization established in the Decision on Principles for the Development of International Standards, Guides and Recommendations issued by the World Trade Organization Technical Barriers to Trade (TBT) Committee.

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ASTM C1268-23(Test Method)
Standard Test Method for Quantitative Determination of 241Am in Plutonium by Gamma-Ray Spectrometry

SIGNIFICANCE AND USE
5.1 This test method allows the determination of 241Am in a plutonium solution without separation of the americium from the plutonium. It is generally applicable to any solution containing 241Am.  
5.2 The 241Am in solid plutonium materials may be determined when these materials are dissolved (see Practice C1168).  
5.3 When the plutonium solution contains unacceptable levels of fission products or other materials, this method may be used following a tri-n-octylphosphine oxide (TOPO) extraction, ion exchange or other similar separation techniques (see Test Methods C758 and C759).  
5.4 This test method is less subject to interferences from plutonium than alpha counting since the energy of the gamma ray used for the analysis is better resolved from other gamma rays than the alpha particle energies used for alpha counting.  
5.5 The minimal sample preparation reduces the amount of sample handling and exposure to the analyst.  
5.6 This test method is applicable only to homogeneous solutions. This test method is not suitable for solutions containing solids.  
5.7 Solutions containing 241Am at concentrations as little as 1 × 10−5 g/L may be analyzed using this method. The lower limit depends on the detector used and the counting geometry. Solutions containing high concentrations may be analyzed following an appropriate dilution.
SCOPE
1.1 This test method covers the quantitative determination of 241Am by gamma-ray spectrometry in plutonium nitrate solution samples that do not contain significant amounts of radioactive fission products or other high specific activity gamma-ray emitters.  
1.2 This test method can be used to determine the 241Am in samples of plutonium metal, oxide and other solid forms, when the solid is appropriately sampled and dissolved.  
1.3 The values stated in SI units are to be regarded as standard. Additionally, the non-SI units of electron volts, kiloelectron volts, and liters 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, health, and environmental practices and determine the applicability of regulatory limitations prior to use.  
1.5 This international standard was developed in accordance with internationally recognized principles on standardization established in the Decision on Principles for the Development of International Standards, Guides and Recommendations issued by the World Trade Organization Technical Barriers to Trade (TBT) Committee.

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ASTM C833-23(Specification)
Standard Specification for Sintered (Uranium-Plutonium) Dioxide Pellets for Light Water Reactors

ABSTRACT
This specification covers finished sintered and ground (uranium-plutonium) dioxide pellets for use in thermal reactors. It applies to uranium-plutonium dioxide pellets containing plutonium additions up to 15 % weight. The diversity of manufacturing methods shall be recognized by which uranium-plutonium dioxide pellets are produced and the many special requirements for chemical and physical characterization that may be imposed by the operating conditions to which the pellets will be subjected in specific reactor systems. The following are different chemical requirements that shall be determined: uranium content, plutonium content, impurity content, stoichiometry, moisture content, gas content, and americium-241 content. Nuclear requirements such as isotopic content, plutonium equivalent at a given date, equivalent boron content, and reactivity shall also be determined. Physical properties of the pellets like dimensions, density, grain size, pore morphology, plutonium-oxide homogeneity, plutonium-oxide particle size, plutonium-oxide particle distribution, integrity, and surface cracks shall be determined as well. The surfaces of finished pellets shall be visually free of loose chips, oil, macroscopic inclusions, and foreign materials. An estimate of the fuel pellet irradiation stability shall be obtained unless adequate allowance for such effects are factored into the fuel rod design. The estimate of the stability shall consist of either conformance to the thermal stability test as specified in the or by adequate correlation of manufacturing process or microstructure to in-reactor behavior, or both.
SCOPE
1.1 This specification covers finished sintered and ground (U, Pu)O2 pellets for use in light water reactors. It applies to (U, Pu)O2 pellets containing a plutonium mass fraction up to 15 % (that is, mass of Pu divided by the sum of masses U, Pu, and Am yielding 0.15 or less).  
1.2 Pellets produced under this specification are available in four grades.  
1.2.1 Grade R—240Pu / (Pu + Am) isotope mass fraction is at least 19 %.  
1.2.2 Grade F—240Pu / (Pu + Am) isotope mass fraction is at least 7 % and less than 19 %.  
1.2.3 Grade N1—240Pu / (Pu + Am) isotope mass fraction is less than 7 %.  
1.2.4 Grade N2—240Pu /239Pu isotope mass fraction does not exceed 0.10 (10 %).  
1.3 There is no discussion of or provision for preventing criticality incidents, nor are health and safety requirements, the avoidance of hazards, or shipping precautions and controls discussed. Observance of this specification does not relieve the user of the obligation to be aware of and conform to all applicable international, federal, state, and local regulations pertaining to possessing, processing, shipping, or using source or special nuclear material. Examples of U.S. government documents are Code of Federal Regulations Title 10, Part 50—Domestic Licensing of Production and Utilization Facilities; Code of Federal Regulations Title 10, Part 71—Packaging and Transportation of Radioactive Material; and Code of Federal Regulations Title 49, Part 173—General Requirements for Shipments and Packaging.  
1.4 The values stated in SI units are to be regarded as standard. No other units of measurement are included in this standard.  
1.5 The following safety hazards caveat pertains only to the technical requirements portion, Section 4, of this specification: 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, health, and environmental practices and determine the applicability of regulatory limitations prior to use.  
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ASTM C1428-18(2023)(Test Method)
Standard Test Method for Isotopic Analysis of Uranium Hexafluoride by Single–Standard Gas Source Multiple Collector Mass Spectrometer Method

SIGNIFICANCE AND USE
5.1 Uranium hexafluoride is a basic material used to produce nuclear reactor fuel. To be suitable for this purpose, the material must meet criteria for isotopic composition. This test method is designed to determine whether the material meets the requirements described in Specifications C787 and C996.
SCOPE
1.1 This test method is applicable to the isotopic analysis of uranium hexafluoride (UF6) with  235U concentrations less than or equal to 5 % and 234U,  236U concentrations of 0.0002 to 0.1 %.  
1.2 This test method may be applicable to the analysis of the entire range of 235U isotopic compositions providing that adequate Certified Reference Materials (CRMs or traceable standards) are available.  
1.3 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, health, and environmental practices and determine the applicability of regulatory limitations prior to use.  
1.4 This international standard was developed in accordance with internationally recognized principles on standardization established in the Decision on Principles for the Development of International Standards, Guides and Recommendations issued by the World Trade Organization Technical Barriers to Trade (TBT) Committee.

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ASTM
ASTM C1062-23(Guide)
Standard Guide for Design, Fabrication, and Installation of Nuclear Fuel Dissolution Facilities

SIGNIFICANCE AND USE
4.1 The purpose of this guide is to provide information that will help to ensure that nuclear fuel dissolution facilities are conceived, designed, fabricated, constructed, and installed in an economic and efficient manner. This guide will help facilities meet the intended performance functions, eliminate or minimize the possibility of nuclear criticality and provide for the protection of both the operator personnel and the public at large under normal and abnormal (emergency) operating conditions as well as under credible failure or accident conditions.
SCOPE
1.1 It is the intent of this guide to set forth criteria and procedures for the design, fabrication and installation of nuclear fuel dissolution facilities. This guide applies to and encompasses all processing steps or operations beyond the fuel shearing operation (not covered), up to and including the dissolving accountability vessel.  
1.2 Applicability and Exclusions:  
1.2.1 Operations—This guide does not cover the operation of nuclear fuel dissolution facilities. Some operating considerations are noted to the extent that these impact upon or influence design.
1.2.1.1 Dissolution Procedures—Fuel compositions, fuel element geometry, and fuel manufacturing methods are subject to continuous change in response to the demands of new reactor designs and requirements. These changes preclude the inclusion of design considerations for dissolvers suitable for the processing of all possible fuel types. This guide will only address equipment associated with dissolution cycles for those fuels that have been used most extensively in reactors as of the time of issue (or revision) of this guide. (See Appendix X1.)  
1.2.2 Processes—This guide covers the design, fabrication and installation of nuclear fuel dissolution facilities for fuels of the type currently used in Pressurized Water Reactors (PWR). Boiling Water Reactors (BWR), Pressurized Heavy Water Reactors (PHWR) and Heavy Water Reactors (HWR) and the fuel dissolution processing technologies discussed herein. However, much of the information and criteria presented may be applicable to the equipment for other dissolution processes such as for enriched uranium-aluminum fuels from typical research reactors, as well as for dissolution processes for some thorium and plutonium-containing fuels and others. The guide does not address equipment design for the dissolution of high burn-up or mixed oxide fuels.
1.2.2.1 This guide does not address special dissolution processes that may require substantially different equipment or pose different hazards than those associated with the fuel types noted above. Examples of precluded cases are electrolytic dissolution and sodium-bonded fuels processing. The guide does not address the design and fabrication of continuous dissolvers.  
1.2.3 Ancillary or auxiliary facilities (for example, steam, cooling water, electrical services) are not covered. Cold chemical feed considerations are addressed briefly.  
1.2.4 Dissolution Pretreatment—Fuel pretreatment steps incidental to the preparation of spent fuel assemblies for dissolution reprocessing are not covered by this guide. This exclusion applies to thermal treatment steps such as “Voloxidation” to drive off gases prior to dissolution, to mechanical decladding operations or process steps associated with fuel elements disassembly and removal of end fittings, to chopping and shearing operations, and to any other pretreatment operations judged essential to an efficient nuclear fuels dissolution step.  
1.2.5 Fundamentals—This guide does not address specific chemical, physical or mechanical technology, fluid mechanics, stress analysis or other engineering fundamentals that are also applied in the creation of a safe design for nuclear fuel dissolution facilities.  
1.3 The values stated in inch-pound units are to be regarded as standard. The values given in parentheses are mathematical conversions to SI unit...

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  • Guide
    17 pages
    English language
    sale 15% off