iTeh StandardsiTeh Standards
EURUSD
Your shopping cart is empty.
ASTM

ASTM C1233-15

(Practice)

Standard Practice for Determining Equivalent Boron Contents of Nuclear Materials

Standard Practice for Determining Equivalent Boron Contents of Nuclear Materials

ABSTRACT
This practice details the recommended method for calculating the equivalent boron content (EBC) values of nuclear elements and materials that are of potential significance as thermal neutron poisons. EBC factors are determined from the atomic weight of elements and the thermal neutron absorption cross section in barns. These may be used depending upon the actual neutron energy characteristics of the applicable reactor system. The elements aluminum, fluorine, rubidium, barium, lead, silicon, beryllium, neon, tin, bismuth, oxygen, zirconium, carbon, magnesium, cerium, and phosphorus are not required to be included in the EBC calculations as their contribution to the total poison effect is not considerably significant.
SIGNIFICANCE AND USE
4.1 Use of this standard practice yields an equivalent boron content (EBC) that can be used to characterize the neutron-absorbing properties of a nuclear material. The elements included in the calculation are typically chosen so that the EBC represents either the entire material (for example, for a moderator) or the impurities in the material (for example, for a nuclear fuel). This practice is typically used for materials in which thermal neutron absorption is undesirable. The EBC is not intended for use as an input to any neutronic calculation. The EBC factors in Table 1 were selected to represent neutron absorption in water reactors under normal operating conditions. It is the responsibility of the user to evaluate their suitability for other purposes. (A) Neutron Cross Sections , Vol 1, Parts A and B, Academic Press, New York, 1981 and 1984, respectively.(B) Holden, N. E., and Martin, R. L., Pure and Applied Chemistry, Vol 56, p. 653, 1984.(C) When present in small concentrations, this element should be excluded from determinations of the total EBC.(D) In the absence of other data, the neutron capture cross section for a Maxwellian flux is used.(E) Cross section is primarily due to a single isotope, whose isotopic abundance is variable in nature. The value can vary between 733 and 779 barns depending upon the source. See Holden, N. E., Neutron Capture Cross Section Standards for BNL-325, Fourth Ed., BNL-NCS-51388, January 1981.(F) Cross section is primarily due to a single isotope, whose isotopic abundance is variable in nature. The value can vary between 69 and 72 barns depending upon the source. See Holden, N. E., Neutron Capture Cross Section Standards for BNL-325, Fourth Ed., BNL-NCS-51388, January 1981.
SCOPE
1.1 This standard details a recommended practice for the calculation of the Equivalent Boron Content (EBC) for nuclear materials. The EBC is used to provide a measure of the macroscopic neutron absorption cross section of a nuclear material. EBC factors for the natural elements are determined from their atomic masses and thermal neutron absorption cross sections. This practice is illustrated by using EBC factors that are based on thermal neutron (2200 m/s) absorption cross sections. Other EBC factors may be used depending upon the actual neutron energy spectrum.  
1.2 The EBC is a characteristic of a homogeneous material. Characterization of inhomogeneous materials and calculation of neutron multiplication factors require techniques that are beyond the scope of this practice.  
1.3 The values stated in SI units are to be regarded as standard. No other units of measurement are included in this standard.

General Information

Status
Historical
Publication Date
31-May-2015
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

Buy Standard

ASTM C1233-15 - Standard Practice for Determining Equivalent Boron Contents of Nuclear MaterialsASTM C1233-15 - Standard Practice for Determining Equivalent Boron Contents of Nuclear Materials
PreviousNext
Standard
ASTM C1233-15 - Standard Practice for Determining Equivalent Boron Contents of Nuclear Materials
English language
3 pages
sale 15% off
Preview
sale 15% off
Preview
REDLINE ASTM C1233-15 - Standard Practice for Determining Equivalent Boron Contents of Nuclear MaterialsREDLINE ASTM C1233-15 - Standard Practice for Determining Equivalent Boron Contents of Nuclear Materials
PreviousNext
Standard
REDLINE ASTM C1233-15 - Standard Practice for Determining Equivalent Boron Contents of Nuclear Materials
English language
3 pages
sale 15% off
Preview
sale 15% off
Preview

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: C1233 − 15
Standard Practice for
Determining Equivalent Boron Contents of Nuclear
1
Materials
This standard is issued under the fixed designation C1233; 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 (´) indicates an editorial change since the last revision or reapproval.
1. Scope Nuclear-Grade Uranyl Nitrate Solutions
C859 Terminology Relating to Nuclear Materials
1.1 This standard details a recommended practice for the
calculation of the Equivalent Boron Content (EBC) for nuclear
3. Terminology
materials. The EBC is used to provide a measure of the
3.1 Terminology used in this practice is in accordance with
macroscopic neutron absorption cross section of a nuclear
material. EBC factors for the natural elements are determined Terminology C859.
from their atomic masses and thermal neutron absorption cross
4. Significance and Use
sections. This practice is illustrated by using EBC factors that
are based on thermal neutron (2200 m/s) absorption cross
4.1 Use of this standard practice yields an equivalent boron
sections. Other EBC factors may be used depending upon the
content (EBC) that can be used to characterize the neutron-
actual neutron energy spectrum.
absorbing properties of a nuclear material. The elements
included in the calculation are typically chosen so that the EBC
1.2 The EBC is a characteristic of a homogeneous material.
represents either the entire material (for example, for a mod-
Characterization of inhomogeneous materials and calculation
erator) or the impurities in the material (for example, for a
of neutron multiplication factors require techniques that are
nuclear fuel). This practice is typically used for materials in
beyond the scope of this practice.
which thermal neutron absorption is undesirable. The EBC is
1.3 The values stated in SI units are to be regarded as
not intended for use as an input to any neutronic calculation.
standard. No other units of measurement are included in this
The EBC factors in Table 1 were selected to represent neutron
standard.
absorption in water reactors under normal operating condi-
tions. It is the responsibility of the user to evaluate their
2. Referenced Documents
suitability for other purposes.
2
2.1 ASTM Standards:
C696 Test Methods for Chemical, Mass Spectrometric, and
5. Procedures for EBC Determination
Spectrochemical Analysis of Nuclear-Grade Uranium Di-
5.1 Agreement shall be reached between the buyer and
oxide Powders and Pellets
seller as to which elements shall be analyzed for calculation of
C698 Test Methods for Chemical, Mass Spectrometric, and
their EBC. It is recommended that B, Cd, Dy, Eu, Sm, and Gd
Spectrochemical Analysis of Nuclear-Grade Mixed Ox-
be included in this calculation. Analytical methods for such
ides ((U, Pu)O )
2
elements shall be those given in Test Methods C696, C698,
C761 Test Methods for Chemical, Mass Spectrometric,
C761, and C799 as applicable or as otherwise agreed upon
Spectrochemical, Nuclear, and RadiochemicalAnalysis of
between buyer and seller.
Uranium Hexafluoride
C799 Test Methods for Chemical, Mass Spectrometric, 5.2 TheindividualEBCvaluesarecalculatedusingtheEBC
Spectrochemical, Nuclear, and RadiochemicalAnalysis of
factors from Table 1 as follows:
EBC of constituent =
1
This practice is under the jurisdiction of ASTM Committee C26 on Nuclear
(EBC factor of constituent)(µg of constituent/gof material)
Fuel Cycle and is the direct responsibility of Subcommittee C26.02 on Fuel and
where:
Fertile Material Specifications.
EBC factor of constituent = (atomic mass of boron)(σaof
Current edition approved June 1, 2015. Published June 2015. Originally
constituent)/[(atomic mass of constituent)(σa of boron)], and
approved in 1993. Last previous approved in 2009 as C1233 – 09. DOI: 10.1520/
C1233-15.
σa = atomic neutron absorption cross section in barns.
2
For referenced ASTM standards, visit the ASTM website, www.astm.org, or
The EBC factors have been calculated using a value of 764
contact ASTM Customer Service at service@astm.org. For Annual Book of ASTM
barns for the neutron absorption cross section (σa) of boron.
Standards volume information, refer to the standard’s Document Summary page on
the ASTM website. This value may vary in nature according to the isotopic
Copyright © ASTM International, 100 Barr Harbor Drive, PO Box C700, West Conshohocken, PA 19428-2959. United States
1

---------------------- Page: 1 ----------------------
C1233 − 15
TABLE 1 Equivalent Boron Content Factors
A
Neutron Absorption Cross Section (Barns) at B
Element Atomic Mass EBC Factor
2200 m/s
C
Aluminium 0.231 26.98 1.21E-04
D
Antimony 5.1 121.75 5.93E-04
Argon 0.68 39.95 2.41E-04
...

This document is not an ASTM standard and is intended only to provide the user of an ASTM standard an indication of what changes have been made to the previous version. Because
it may not be technically possible to adequately depict all changes accurately, ASTM recommends that users consult prior editions as appropriate. In all cases only the current version
of the standard as published by ASTM is to be considered the official document.
Designation: C1233 − 09 C1233 − 15
Standard Practice for
Determining Equivalent Boron Contents of Nuclear
1
Materials
This standard is issued under the fixed designation C1233; 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 (´) indicates an editorial change since the last revision or reapproval.
1. Scope
1.1 This standard details a recommended practice for the calculation of the Equivalent Boron Content (EBC) for nuclear
materials. The EBC is used to provide a measure of the macroscopic neutron absorption cross section of a nuclear material. EBC
factors for the natural elements are determined from their atomic masses and thermal neutron absorption cross sections. This
practice is illustrated by using EBC factors that are based on thermal neutron (2200 m/s) absorption cross sections. Other EBC
factors may be used depending upon the actual neutron energy spectrum.
1.2 The EBC is a characteristic of a homogeneous material. Characterization of inhomogeneous materials and calculation of
neutron multiplication factors require techniques that are beyond the scope of this practice.
1.3 The values stated in SI units are to be regarded as standard. No other units of measurement are included in this standard.
2. Referenced Documents
2
2.1 ASTM Standards:
C696 Test Methods for Chemical, Mass Spectrometric, and Spectrochemical Analysis of Nuclear-Grade Uranium Dioxide
Powders and Pellets
C698 Test Methods for Chemical, Mass Spectrometric, and Spectrochemical Analysis of Nuclear-Grade Mixed Oxides ((U,
Pu)O )
2
C761 Test Methods for Chemical, Mass Spectrometric, Spectrochemical, Nuclear, and Radiochemical Analysis of Uranium
Hexafluoride
C799 Test Methods for Chemical, Mass Spectrometric, Spectrochemical, Nuclear, and Radiochemical Analysis of Nuclear-
Grade Uranyl Nitrate Solutions
C859 Terminology Relating to Nuclear Materials
3. Terminology
3.1 Terminology used in this practice is in accordance with Terminology C859.
4. Significance and Use
4.1 Use of this standard practice yields an equivalent boron content (EBC) that can be used to characterize the
neutron-absorbing properties of a nuclear material. The elements included in the calculation are typically chosen so that the EBC
represents either the entire material (for example, for a moderator) or the impurities in the material (for example, for a nuclear fuel).
This practice is typically used for materials in which thermal neutron absorption is undesirable. The EBC is not intended for use
as an input to any neutronic calculation. The EBC factors in Table 1 were selected to represent neutron absorption in water reactors
under normal operating conditions. It is the responsibility of the user to evaluate their suitability for other purposes.
5. Methods ForProcedures for EBC Determination
5.1 Agreement shall be reached between the buyer and seller as to which elements shall be analyzed for calculation of their
EBC. It is recommended that B, Cd, Dy, Eu, Sm, and Gd be included in this calculation. Analytical methods for such elements
shall be those given in Test Methods C696, C698, C761, and C799 as applicable or as otherwise agreed upon between buyer and
seller.
1
This practice 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.
Current edition approved June 1, 2009June 1, 2015. Published July 2009June 2015. Originally approved in 1993. Last previous approved in 20032009 as
C1233 – 03.C1233 – 09. DOI: 10.1520/C1233-09.10.1520/C1233-15.
2
For referenced ASTM standards, visit the ASTM website, www.astm.org, or contact ASTM Customer Service at service@astm.org. For Annual Book of ASTM Standards
volume information, refer to the standard’s Document Summary page on the ASTM website.
Copyright © ASTM International, 100 Barr Harbor Drive, PO Box C700, West Conshohocken, PA 19428-2959. United States
1

---------------------- Page: 1 ----------------------
C1233 − 15
TABLE 1 Equivalent Boron Content Factors
A
Neutron Absorption Cross Section (Barns) at
B
Element Atomic Mass EBC Factor
2200 m/s
C
Aluminium 0.231 26.98 1.21E-04
D
Antimony 5.1 121.75 5.93E-04
Argon 0.68 39.95 2.41E-04
Arsenic 4.5 74.92 8.50E-04
C D
Barium 1.2 137.33 1.24E-04
C D
Beryllium 0.0076 9.01 1.19E-05
C
Bismuth 0.0338 208.98 2.29E-06
E
Boron 764 10.81 1.00E+00
Bromine 6.9 79.91 1.
...

Questions, Comments and Discussion

Ask us and Technical Secretary will try to provide an answer. You can facilitate discussion about the standard in here.

This May Also Interest You

ASTM
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.

  • Standard
    6 pages
    English language
    sale 15% off
  • Standard
    6 pages
    English language
    sale 15% off
ASTM
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.
SCOPE
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.

  • Standard
    10 pages
    English language
    sale 15% off
  • Standard
    10 pages
    English language
    sale 15% off
ASTM
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.

  • Standard
    4 pages
    English language
    sale 15% off
ASTM
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.  
1.6 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 Technic...

  • Technical specification
    5 pages
    English language
    sale 15% off
  • Technical specification
    5 pages
    English language
    sale 15% off
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...

  • Guide
    17 pages
    English language
    sale 15% off
  • Guide
    17 pages
    English language
    sale 15% off
ASTM
ASTM C859-23(Terminology)
Standard Terminology Relating to Nuclear Materials

SCOPE
1.1 This terminology standard contains 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.  
1.2 While subcommittees within Committee C26 are free to only provide terms and definitions within individual standards, each subcommittee may request the addition of utilized terms and definitions to this terminology standard if it believes that such serves the broader interest of Committee C26 and the nuclear fuel cycle profession. Therefore, terms and definitions proposed for inclusion in Terminology C859 need not be used in more than one committee standard before being considered.  
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.  
1.4 Subcommittee C26.10 (Nondestructive Assay) also has a terminology standard applicable to its standards: Terminology C1673.  
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.

  • Standard
    6 pages
    English language
    sale 15% off
  • Standard
    6 pages
    English language
    sale 15% off
ASTM
ASTM C1554-18(2023)(Guide)
Standard Guide for Materials Handling Equipment for Hot Cells

SIGNIFICANCE AND USE
4.1 Materials handling equipment operability and long-term integrity are concerns that originate during the design and fabrication sequences. Such concerns are most efficiently addressed during one or the other of these stages. Equipment operability and integrity can be compromised during handling and installation sequences. For this reason, the subject equipment should be handled and installed under closely controlled and supervised conditions.  
4.2 This guide is intended as a supplement to other standards (Section 2, Referenced Documents), and to federal and state regulations, codes, and criteria applicable to the design of equipment intended for this use.  
4.3 This guide is intended to be generic and to apply to a wide range of types and configurations of materials handling equipment.  
4.4 The term materials handling equipment is used herein in a generic sense. It includes manipulators, cranes, carts or bogies, and special equipment for handling tools and material in hot cells.  
4.5 This service imposes stringent requirements on the quality and the integrity of the equipment, as follows:  
4.5.1 Boots and similar protective covers should not restrict movement of the equipment, should be properly sealed to the equipment and should withstand the radiation, cell atmosphere, dust, cell temperatures, chemical exposures, and cleaning and decontamination reagents, and also resist snags and tearing.  
4.5.2 Materials handling equipment should be capable of withstanding rigorous chemical cleaning and decontamination procedures.  
4.5.3 Materials handling equipment should be designed and fabricated to remain dimensionally stable throughout its life cycle.  
4.5.4 Attention to fabrication tolerances is necessary to allow the proper fit-up between components for the proper installation and mounting of materials handling equipment in hot cells, for example, when parts or components are being replaced. Fabrication tolerances should be controlled to provide sufficie...
SCOPE
1.1 Intent:  
1.1.1 This guide covers materials handling equipment used in hot cells (shielded cells) for the processing and handling of nuclear and radioactive materials. The intent of this guide is to aid in the selection and design of materials handling equipment for hot cells in order to minimize equipment failures and maximize the equipment utility.  
1.1.2 It is intended that this guide record the principles and caveats that experience has shown to be essential to the design, fabrication, installation, maintenance, repair, replacement, and decontamination and decommissioning of materials handling equipment capable of meeting the stringent demands of operating, dependably and safely, in a hot cell environment where operator visibility is limited due to the radiation exposure hazards.  
1.1.3 This guide may apply to materials handling equipment in other radioactive remotely operated facilities such as suited entry repair areas and canyons, but does not apply to materials handling equipment used in commercial power reactors.  
1.1.4 This guide covers mechanical master-slave manipulators and electro-mechanical manipulators, but does not cover electro-hydraulic manipulators.  
1.2 Applicability:  
1.2.1 This guide is intended to be applicable to equipment used under one or more of the following conditions:
1.2.1.1 The materials handled or processed constitute a significant radiation hazard to man or to the environment.
1.2.1.2 The equipment will generally be used over a long-term life cycle (for example, in excess of two years), but equipment intended for use over a shorter life cycle is not excluded.
1.2.1.3 The equipment can neither be accessed directly for purposes of operation or maintenance, nor can the equipment be viewed directly, for example, without shielded viewing windows, periscopes, or a video monitoring system.  
1.3 User Caveats:  
1.3.1 This standard is not a substitute for applied engin...

  • Guide
    10 pages
    English language
    sale 15% off
ASTM
ASTM C1703-18(2023)(Practice)
Standard Practice for Sampling of Gaseous Uranium Hexafluoride for Enrichment

SIGNIFICANCE AND USE
5.1 Uranium hexafluoride is normally produced and handled in large (typically 1 to 14-ton) quantities and must, therefore, be characterized by reference to representative samples (see ISO 7195). The samples are used to determine compliance with the applicable commercial specification C787. The quantities involved, physical properties, chemical reactivity, and hazardous nature of UF6 are such that for representative sampling, specially designed equipment must be used and operated in accordance with the most carefully controlled and stringent procedures. This practice can be used by UF6 converters to review the effectiveness of existing procedures or as a guide to the design of equipment and procedures for future use.  
5.2 The intention of this practice is to avoid liquid UF6 sampling once the cylinder has been filled. For safety reasons, manipulation of large quantities of liquid UF6 should be avoided when possible.  
5.3 It is emphasized that this practice is not meant to address conventional or nuclear criticality safety issues.
SCOPE
1.1 This practice covers methods for withdrawing representative sample(s) of uranium hexafluoride (UF6) during a transfer occurring in the gas phase. Such transfer in the gas phase can take place during the filling of a cylinder during a continuous production process, for example the distillation column in a conversion facility. Such sample(s) may be used for determining compliance with the applicable commercial specification, for example Specification C787.  
1.2 Since UF6 sampling is taken during the filling process, this practice does not address any special additional arrangements that may be agreed upon between the buyer and the seller when the sampled bulk material is being added to residues already present in a container (“heels recycle”). Such arrangements will be based on QA procedures such as traceability of cylinder origin (to prevent for example contamination with irradiated material).  
1.3 If the receiving cylinder is purged after filling and sampling, special verifications must be performed by the user to verify the representativity of the sample(s). It is then expected that the results found on volatile impurities with gas phase sampling may be conservative.  
1.4 This practice is only applicable when the transfer occurs in the gas phase. When the transfer is performed in the liquid phase, Practice C1052 should apply. This practice does not apply to gas sampling after the cylinder has been filled since the sample taken will not be representative of the cylinder.  
1.5 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.  
1.6 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.

  • Standard
    3 pages
    English language
    sale 15% off
ASTM
ASTM C1347-08(2023)(Practice)
Standard Practice for Preparation and Dissolution of Uranium Materials for Analysis

SIGNIFICANCE AND USE
4.1 The materials covered that must meet ASTM specifications are uranium metal and uranium oxide.  
4.2 Uranium materials are used as nuclear reactor fuel. For this use, these materials must meet certain criteria for uranium content, uranium-235 enrichment, and impurity content, as described in Specifications C753 and C776. The material is assayed for uranium to determine whether the content is as specified.  
4.3 Uranium alloys, refractory uranium materials, and uranium containing scrap and ash are unique uranium materials for which the user must determine the applicability of this practice. In general, these unique uranium materials are dissolved with various acid mixtures or by fusion with various fluxes.
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, health, and environmental practices and determine the applicability of regulatory limitations prior to use. Specific hazards statements are given in Section 7.  
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.

  • Standard
    9 pages
    English language
    sale 15% off
ASTM
ASTM C1165-23(Test Method)
Standard Test Method for Determining Plutonium by Controlled-Potential Coulometry in H2SO4 at a Platinum Working Electrode

SIGNIFICANCE AND USE
5.1 This test method is used to ascertain whether or not materials meet specifications for plutonium concentration or plutonium mass fraction.  
5.1.1 The materials (nuclear grade plutonium nitrate solutions, plutonium metal, plutonium oxide powder, and mixed oxide and carbide powders and pellets) to which this test method applies are subject to nuclear safeguards regulations governing their possession and use. However, adherence to this test method does not automatically guarantee regulatory acceptance of the resulting safeguards measurements. It remains the sole responsibility of the user of this test method to ensure that its application to safeguards has the approval of the proper regulatory authorities.  
5.1.2 When used in conjunction with appropriate certified reference materials (CRMs), this test method can demonstrate traceability to the international measurements system (SI).  
5.2 Fitness for Purpose of Safeguards and Nuclear Safety Application—Methods intended for use in safeguards and nuclear safety applications shall meet the requirements specified by Guide C1068 for use in such applications.  
5.3 A chemical calibration of the coulometer is necessary for accurate results.
FIG. 1 Example of a Cell Design Used at Los Alamos National Laboratory (LANL)
SCOPE
1.1 This test method covers the determination of milligram quantities of plutonium in unirradiated uranium-plutonium mixed oxide having a U/Pu ratio range of 0.1 to 10. This test method is also applicable to plutonium metal, plutonium oxide, uranium-plutonium mixed carbide, various plutonium compounds including fluoride and chloride salts, and plutonium solutions.  
1.2 The recommended amount of plutonium for each aliquant in the coulometric analysis is 5 mg to 10 mg. Precision worsens for lower amounts of plutonium, and elapsed time of electrolysis becomes impractical for higher amounts of plutonium.  
1.3 The quantity values stated in SI units are to be regarded as standard. The quantity values with non-SI units are given in parentheses for information only.  
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. Specific precautionary statements are given in Section 9.  
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.

  • Standard
    6 pages
    English language
    sale 15% off
  • Standard
    6 pages
    English language
    sale 15% off