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ASTM C1334-96(2000)

(Specification)

Standard Specification for Uranium Oxides with a 235U Content of Less Than 5% for Dissolution Prior to Conversion to Nuclear-Grade Uranium Dioxide

Standard Specification for Uranium Oxides with a <sup>235</sup>U Content of Less Than 5% for Dissolution Prior to Conversion to Nuclear-Grade Uranium Dioxide

SCOPE
1.1 This specification covers uranium trioxide (UO3 ) or U3 O8  powders, or mixtures of the two, that are intended for dissolution into uranyl nitrate solution meeting the requirements of Specification C788 prior to conversion into nuclear grade UO2  powder with a  235 U content of less than 5%. This specification defines the impurity and uranium isotope limits for such urania powders that are to be dissolved prior to processing to nuclear grade UO2  as defined in Specification C753.  
1.2 This specification provides the nuclear industry with a general standard for such uranium oxide powders. It recognizes the diversity of conversion processes and the processes to which such powders are subsequently to be subjected. It is therefore anticipated that it may be necessary to include supplementary specification limits by agreement between the buyer and seller.  
1.3 The scope of this specification does not comprehensively cover all provisions for preventing criticality accidents, for health and safety, or for shipping. Observance of this specification does not relieve the user of the obligation to conform to all international, national, state and local regulations for processing, shipping, or any other way of using urania powders (see 2.2 and 2.3).

General Information

Status
Historical
Publication Date
09-Jun-2000
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

Replaced By
ASTM C1334-05 - Standard Specification for Uranium Oxides with a <sup>235</sup>U Content of Less Than 5 % for Dissolution Prior to Conversion to Nuclear-Grade Uranium Dioxide
Effective Date
01-Jan-2005
Referred By
ASTM C1413-18 - Standard Test Method for Isotopic Analysis of Hydrolyzed Uranium Hexafluoride and Uranyl Nitrate Solutions by Thermal Ionization Mass Spectrometry
Effective Date
10-Jun-2000

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ASTM C1334-96(2000) - Standard Specification for Uranium Oxides with a <sup>235</sup>U Content of Less Than 5% for Dissolution Prior to Conversion to Nuclear-Grade Uranium DioxideASTM C1334-96(2000) - Standard Specification for Uranium Oxides with a <sup>235</sup>U Content of Less Than 5% for Dissolution Prior to Conversion to Nuclear-Grade Uranium Dioxide
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ASTM C1334-96(2000) - Standard Specification for Uranium Oxides with a <sup>235</sup>U Content of Less Than 5% for Dissolution Prior to Conversion to Nuclear-Grade Uranium Dioxide
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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 1334 – 96 (Reapproved 2000)
Standard Specification for
Uranium Oxides with a U Content of Less Than 5 % for
Dissolution Prior to Conversion to Nuclear-Grade Uranium
Dioxide
This standard is issued under the fixed designation C 1334; 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 799 Test Methods for Chemical, Mass Spectrometric,
Spectrochemical, Nuclear, and Radiochemical Analysis of
1.1 This specification covers uranium trioxide (UO)or
Nuclear-Grade Uranyl Nitrate Solutions
U O powders, or mixtures of the two, that are intended for
3 8
C 859 Terminology Relating to Nuclear Materials
dissolution into uranyl nitrate solution meeting the require-
C 996 Specification for Uranium Hexafluoride Enriched to
ments of Specification C 788 prior to conversion into nuclear
Less Than 5 % U
grade UO powder with a U content of less than 5 %. This
C 1233 Practice for Determining Equivalent Boron Con-
specification defines the impurity and uranium isotope limits
tents of Nuclear Materials
for such urania powders that are to be dissolved prior to
E 11 Specification for Wire-Cloth Sieves for Testing Pur-
processing to nuclear grade UO as defined in Specification
poses
C 753.
E 105 Practice for Probability Sampling of Materials
1.2 This specification provides the nuclear industry with a
2.2 ANSI Standard:
generalstandardforsuchuraniumoxidepowders.Itrecognizes
ASME NQA-1 Quality Assurance Program, Requirements
the diversity of conversion processes and the processes to
for Nuclear Facilities
which such powders are subsequently to be subjected. It is
2.3 U.S. Government Document:
therefore anticipated that it may be necessary to include
Federal Regulations Title 10, (Energy) Part 50, Domestic
supplementary specification limits by agreement between the
Licensing of Production and Utilization Facilities
buyer and seller.
1.3 The scope of this specification does not comprehen-
3. Terminology
sively cover all provisions for preventing criticality accidents,
3.1 Definition of Term Specific to This Standard:
for health and safety, or for shipping. Observance of this
3.1.1 Terms shall be defined in accordance with Terminol-
specification does not relieve the user of the obligation to
ogy C 859, except for the following:
conform to all international, national, state and local regula-
3.1.2 Commercial Grade Uranium Oxide, n—uranium tri-
tionsforprocessing,shipping,oranyotherwayofusingurania
oxide (UO ), U O , or a mixture of the two, made from
3 3 8
powders (see 2.2 and 2.3).
unirradiated uranium. It is recognized some contamination
2. Referenced Documents with reprocessed uranium may occur during routine process-
ing; this is acceptable provided that the specification for
2.1 ASTM Standards:
Commercial Grade Uranium Oxide as set forth in 4.1 is met.
C 696 Test Methods for Chemical, Mass Spectrometric, and
Spectrochemical Analysis of Nuclear-Grade Uranium Di-
4. Isotopic Content
oxide Powders and Pellets
4.1 For Commercial Grade Uranium Oxide with an isotopic
C 753 Specification for Nuclear-Grade, Sinterable Uranium
content of U between that of natural uranium and 5 %, the
Dioxide Powder
isotopic limits of Specification C 996 shall apply. The specific
C 788 Specification for Nuclear-Grade Uranyl Nitrate So-
isotopic measurements required by Specification C 996 may be
lution
1 3
This specification is under the jurisdiction of ASTM Committee C26 on Annual Book of ASTM Standards, Vol 14.02.
Nuclear Fuel Cycle and is the direct responsibility of Subcommittee C26.02 on Fuel Available from American National Standards Institute, 11 West 42nd St., 13th
and Fertile Material Specifications. Floor, New York, NY 10036.
Current edition approved April 10, 1996. Published May 1996. 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 1334
waived, provided that the seller can demonstrate compliance buyer and seller. Packing or agglomeration during shipping, or
through, for instance, the seller’s quality assurance records. both, may be a concern.
4.2 For commercial uranium oxides not having an assay in 5.8 Dissolvability—At the buyer’s request, a dissolvability
the range set forth in 4.1, the isotopic requirements shall be as test shall be performed by a procedure and to a specification as
agreed upon between the buyer and seller. mutually agreed upon between buyer and seller. The test
parameters(suchastime,temperature,nitricacidmolarity)and
5. Physical and Chemical Requirements
characteristics to be measured (such as dissolution rate, in-
solubles, foam generation) shall be defined by agreement
5.1 Uranium Content—The uranium content shall be deter-
between the buyer and seller.
mined using methods described in Test Methods C 696 and
C 799, or as agreed upon between the buyer and seller. Based
6. Lot Requirements
on the oxygen-to-uranium ratio, the minimum total uranium
6.1 A lot is defined as a quantity of Commercial Grade
content shall also be agreed upon between the buyer and seller.
Uranium Oxide powder that is uniform in isotopic, chemical,
5.2 Impurity Content—The impurity content shall not ex-
and flowability characteristics.
ceed the individual element limit specified in Table 1 on a
6.2 The identity of a lot shall be retained throughout.
uranium weight basis. The summation of the contribution of
6.3 Apowder lot shall form the basis for defining sampling
each of the impurity elements listed inTable 1 shall not exceed
plans used to establish conformance to this specification.
1000 µg/gU. The impurity content shall be determined using
6.4 Sampling plans shall be mutually agreed upon by the
methods described in Test Methods C 696 or as agreed upon
buyer and seller. A suggested sampling procedure is given in
between the buyer and seller.
Annex A1.
5.3 Equivalent Boron Content—The total equivalent
...

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1.1 This test method covers the measurement of gamma energy emitted from fission products in uranium hexafluoride (UF6) and uranyl nitrate solution. This test method may also be used to measure the concentration of some uranium decay products. It is intended to provide a method for demonstrating compliance with UF6 Specifications C787 and C996, uranyl nitrate Specification C788, and uranium ore concentrate Specification C967.  
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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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.  
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1.2 In principle, the total evaporation method should yield isotope ratios that do not require mass bias correction. In practice, samples may require this bias correction. Compared to the conventional TIMS method described in Test Method C1625, the total evaporation method is approximately two times faster, improves precision of the isotope ratio measurements by a factor of two to four, and utilizes smaller sample sizes. Compared to the C1625 method, the total evaporation method provides “major” isotope ratios 235U/238U, 240Pu/239Pu, and 241Am/243Am with improved accuracy.  
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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.  
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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).  
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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.  
1.1.1 The most frequent uses of passive neutron holdup techniques are for the measurement of uranium or plutonium deposits in processing facilities.  
1.2 This guide includes information useful for management, planning, selection of equipment, consideration of interferences, measurement program definition, and the utilization of resources.  
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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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Standard Test Method for Quantitative Determination of 241Am in Plutonium by Gamma-Ray Spectrometry

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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).  
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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.
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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.  
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ASTM C1168-23(Practice)
Standard Practice for Preparation and Dissolution of Plutonium Materials for Analysis

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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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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 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...

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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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