ASTM C1062-23

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...

General Information

• Status: Published
• Publication Date: 31-Mar-2023
• ICS: 27.120.30 - Fissile materials and nuclear fuel technology
• Technical Committee: C26 - Nuclear Fuel Cycle
• Drafting Committee: C26.09 - Nuclear Processing

Relations

• Refers: ASTM C859-24 - Standard Terminology Relating to Nuclear Materials
• Effective Date: 01-Jan-2024
• Refers: ASTM C1217-00(2020) - Standard Guide for Design of Equipment for Processing Nuclear and Radioactive Materials
• Effective Date: 01-Jul-2020
• Refers: ASTM C859-14a - Standard Terminology Relating to Nuclear Materials
• Effective Date: 15-Jun-2014
• Refers: ASTM C859-14 - Standard Terminology Relating to Nuclear Materials
• Effective Date: 15-Jan-2014
• Refers: ASTM C859-13a - Standard Terminology Relating to Nuclear Materials
• Effective Date: 01-Jun-2013
• Refers: ASTM C859-13 - Standard Terminology Relating to Nuclear Materials
• Effective Date: 01-May-2013
• Refers: ASTM C1217-00(2012) - Standard Guide for Design of Equipment for Processing Nuclear and Radioactive Materials
• Effective Date: 01-Jun-2012
• Refers: ASTM C859-10b - Standard Terminology Relating to Nuclear Materials
• Effective Date: 01-Nov-2010
• Refers: ASTM C859-10a - Standard Terminology Relating to Nuclear Materials
• Effective Date: 01-Aug-2010
• Refers: ASTM C859-10 - Standard Terminology Relating to Nuclear Materials
• Effective Date: 01-Feb-2010
• Refers: ASTM C859-09 - Standard Terminology Relating to Nuclear Materials
• Effective Date: 15-Feb-2009
• Refers: ASTM C859-08 - Standard Terminology Relating to Nuclear Materials
• Effective Date: 15-Sep-2008
• Refers: ASTM C1217-00(2006) - Standard Guide for Design of Equipment for Processing Nuclear and Radioactive Materials
• Effective Date: 01-Jan-2006
• Refers: ASTM C1217-00 - Standard Guide for Design of Equipment for Processing Nuclear and Radioactive Materials
• Effective Date: 10-Jun-2000
• Refers: ASTM C859-92b - Standard Terminology Relating to Nuclear Materials (Withdrawn 2005)
• Effective Date: 01-Jan-1992

Overview

ASTM C1062-23: Standard Guide for Design, Fabrication, and Installation of Nuclear Fuel Dissolution Facilities provides comprehensive recommendations to ensure the safe and efficient development of facilities used in the nuclear fuel dissolution process. Published by ASTM International, this guide is intended to help the nuclear industry implement best practices for the conception, design, fabrication, construction, and installation of these specialized facilities, focusing on minimizing nuclear criticality risks and protecting personnel and the public under normal and emergency conditions.

This guide applies specifically to processing steps beyond the fuel shearing operation and up to the dissolving accountability vessel. It encompasses the most commonly used fuel types, particularly those deployed in Pressurized Water Reactors (PWR), Boiling Water Reactors (BWR), Pressurized Heavy Water Reactors (PHWR), and Heavy Water Reactors (HWR).

Key Topics

• Scope and Applicability

  • The guide defines criteria and procedures for the design, fabrication, and installation of nuclear fuel dissolution facilities.
  • It covers equipment and processes for widely used reactor fuels but excludes high burn-up and mixed oxide (MOX) fuels, as well as specialized dissolution processes such as electrolytic or sodium-bonded fuel processes.

• Criticality Control

  • Emphasizes minimizing the risk of nuclear criticality through geometrically favorable vessel design, use of neutron poisons, and concentration control.
  • Stresses rigorous criticality assessment and periodic quality assurance verification throughout equipment life cycles.

• Quality Assurance

  • Recommends implementation of formalized quality assurance programs in accordance with regulatory criteria (e.g., ASME NQA-1, 10 CFR 50 Appendix B).
  • Stipulates thorough documentation and retention of all relevant records pertaining to design, calculations, quality, and testing.

• Personnel and Responsibility

  • Design and construction personnel are expected to possess the requisite expertise to manage complexities and ensure safety.
  • Clear assignment of responsibilities and robust audit trails are encouraged for transparency and regulatory compliance.

• Design Considerations

  • Guidance provided for safe vessel design, heat transfer management, off-gas handling, use of corrosion-resistant materials, and mitigation of safety hazards.
  • Emphasizes provisions for chemical and physical containment under normal, abnormal, and accident scenarios.

Applications

ASTM C1062-23 is a crucial resource for organizations involved in the reprocessing and management of spent nuclear fuel. Applications include:

• Design of Nuclear Fuel Dissolution Facilities
  • Ensuring consistent and compliant infrastructure for dissolution processes in nuclear fuel reprocessing plants.
• Safety Assessments
  • Implementing advanced hazard assessments and risk mitigation strategies to prevent nuclear criticality incidents.
• Regulatory Compliance
  • Meeting industry, national, and international regulatory requirements for nuclear facility design and operation.
• Quality Management
  • Establishing traceable quality systems and long-term records management for facility lifecycle oversight.
• Operator and Public Protection
  • Integrating robust engineering controls to safeguard facility personnel and the general public from radiological and chemical hazards.

Related Standards

Several standards are referenced in ASTM C1062-23 to facilitate comprehensive design, fabrication, and quality assurance:

• ASTM Standards

  • ASTM C859: Terminology Relating to Nuclear Materials
  • ASTM C1217: Guide for Design of Equipment for Processing Nuclear and Radioactive Materials

• ASME Standards

  • ASME Boiler and Pressure Vessel Code (Sections II, V, VIII, IX)
  • ASME NQA-1: Quality Assurance Requirements for Nuclear Facility Applications

• American Nuclear Society (ANS) Standards

  • ANS 8.1: Nuclear Criticality Safety in Operations with Fissionable Materials Outside Reactors
  • ANS 8.3: Criticality Accident Alarm System
  • ANS 8.9: Nuclear Criticality Safety Criteria for Steel-Pipe Intersections Containing Aqueous Solutions of Fissile Materials
  • ANS 57.8: Fuel Assembly Identification

• Federal Regulations

  • 10 CFR 50: Licensing of Production and Utilization Facilities
  • 10 CFR 50, Appendix B: Quality Assurance Criteria for Nuclear Power Plants and Fuel Reprocessing Plants

ASTM C1062-23 provides essential guidance for the secure and efficient design, fabrication, and installation of nuclear fuel dissolution facilities, making it a reference standard for nuclear fuel reprocessing industry stakeholders and supporting safe management of nuclear materials.

Standards Content (Sample)

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.
Designation: C1062 − 23
Standard Guide for
Design, Fabrication, and Installation of Nuclear Fuel
Dissolution Facilities
This standard is issued under the fixed designation C1062; 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 pose different hazards than those associated with the fuel types
noted above. Examples of precluded cases are electrolytic
1.1 It is the intent of this guide to set forth criteria and
dissolution and sodium-bonded fuels processing. The guide
procedures for the design, fabrication and installation of
does not address the design and fabrication of continuous
nuclear fuel dissolution facilities. This guide applies to and
dissolvers.
encompasses all processing steps or operations beyond the fuel
1.2.3 Ancillary or auxiliary facilities (for example, steam,
shearing operation (not covered), up to and including the
cooling water, electrical services) are not covered. Cold chemi-
dissolving accountability vessel.
cal feed considerations are addressed briefly.
1.2 Applicability and Exclusions:
1.2.4 Dissolution Pretreatment—Fuel pretreatment steps in-
1.2.1 Operations—This guide does not cover the operation
cidental to the preparation of spent fuel assemblies for disso-
of nuclear fuel dissolution facilities. Some operating consider-
lution reprocessing are not covered by this guide. This exclu-
ations are noted to the extent that these impact upon or
sion applies to thermal treatment steps such as “Voloxidation”
influence design.
to drive off gases prior to dissolution, to mechanical decladding
1.2.1.1 Dissolution Procedures—Fuel compositions, fuel el-
operations or process steps associated with fuel elements
ement geometry, and fuel manufacturing methods are subject
disassembly and removal of end fittings, to chopping and
to continuous change in response to the demands of new
shearing operations, and to any other pretreatment operations
reactor designs and requirements. These changes preclude the
judged essential to an efficient nuclear fuels dissolution step.
inclusion of design considerations for dissolvers suitable for
1.2.5 Fundamentals—This guide does not address specific
the processing of all possible fuel types. This guide will only
chemical, physical or mechanical technology, fluid mechanics,
address equipment associated with dissolution cycles for those
stress analysis or other engineering fundamentals that are also
fuels that have been used most extensively in reactors as of the
applied in the creation of a safe design for nuclear fuel
time of issue (or revision) of this guide. (See Appendix X1.)
dissolution facilities.
1.2.2 Processes—This guide covers the design, fabrication
1.3 The values stated in inch-pound units are to be regarded
and installation of nuclear fuel dissolution facilities for fuels of
as standard. The values given in parentheses are mathematical
the type currently used in Pressurized Water Reactors (PWR).
conversions to SI units that are provided for information only
Boiling Water Reactors (BWR), Pressurized Heavy Water
and are not considered standard.
Reactors (PHWR) and Heavy Water Reactors (HWR) and the
1.4 This standard does not purport to address all of the
fuel dissolution processing technologies discussed herein.
safety concerns, if any, associated with its use. It is the
However, much of the information and criteria presented may
responsibility of the user of this standard to establish appro-
be applicable to the equipment for other dissolution processes
priate safety, health, and environmental practices and deter-
such as for enriched uranium-aluminum fuels from typical
mine the applicability of regulatory limitations prior to use.
research reactors, as well as for dissolution processes for some
1.5 This international standard was developed in accor-
thorium and plutonium-containing fuels and others. The guide
dance with internationally recognized principles on standard-
does not address equipment design for the dissolution of high
ization established in the Decision on Principles for the
burn-up or mixed oxide fuels.
Development of International Standards, Guides and Recom-
1.2.2.1 This guide does not address special dissolution
mendations issued by the World Trade Organization Technical
processes that may require substantially different equipment or
Barriers to Trade (TBT) Committee.
This guide is under the jurisdiction of ASTM Committee C26 on Nuclear Fuel
2. Referenced Documents
Cycle and is the direct responsibility of Subcommittee C26.09 on Nuclear
Processing.
2.1 Industry and National Consensus Standards—Industry
Current edition approved April 1, 2023. Published July 2023. Originally
and national consensus standards applicable in whole or in part
approved in 1986. Last previous edition approved in 2014 as C1062 – 00 (2014).
DOI: 10.1520/C1062-23. to the design, fabrication, and installation of nuclear fuel
Copyright © ASTM International, 100 Barr Harbor Drive, PO Box C700, West Conshohocken, PA 19428-2959. United States
C1062 − 23
dissolution facilities are referenced throughout this guide and 3.1.3 shall, should, and may—The word “shall” denotes a
include the following: requirement, the word “should” denotes a recommendation and
2 the word “may” indicates permission, neither a requirement
2.2 ASTM Standards:
nor a recommendation. In order to conform with this guide, all
C859 Terminology Relating to Nuclear Materials
actions or conditions shall be in accordance with its require-
C1010 Guide for Acceptance, Checkout, and Pre-
ments but they need not conform with its recommendations.
Operational Testing of a Nuclear Fuels Reprocessing
Facility (Withdrawn 2001) 3.2 Definitions of Terms Specific to This Standard:
C1217 Guide for Design of Equipment for Processing 3.2.1 accident—an unplanned event that could result in
Nuclear and Radioactive Materials unacceptable levels of any of the following:
2.3 ASME Standards: 3.2.1.1 equipment damage,
ASME Boiler and Pressure Vessel Code, Sections II, V, VIII, 3.2.1.2 injury to personnel,
and IX 3.2.1.3 downtime or outage,
ASME NQA-1 Quality Assurance Requirements for Nuclear 3.2.1.4 release of hazardous materials (radioactive or non-
Facility Applications radioactive).
2.4 ANS Standard: 3.2.1.5 radiation exposure to personnel, and
ANS Glossary of Terms in Nuclear Science and Technology 3.2.1.6 criticality.
(ANS Glossary) 3.2.2 accountability—the keeping of records on and the
ANS 8.1 Nuclear Criticality Safety in Operations with Fis- responsibility associated with being accountable for the
sionable Materials Outside Reactors amount of fissile materials entering and leaving a plant, a
ANS 8.3 Criticality Accident Alarm System location, or a processing step.
ANS 8.9 Nuclear Criticality Safety Criteria for Steel-Pipe
3.2.3 basic data—the fundamental chemical, physical, and
Intersections Containing Aqueous Solutions of Fissile
mathematical values, formulas, and principles, and the defini-
Materials
tive criteria that have been documented and accepted as the
ANS 57.8 Fuel Assembly Identification
basis for facilities design.
2.5 Federal Regulations —Federal Regulations that are
3.2.4 double contingency principle—the use of methods,
specifically applicable in whole or in part to the design,
measures, or factors of safety in the design of nuclear facilities
fabrication, and installation of nuclear fuel dissolution facilities
such that at least two unlikely, independent, and concurrent
include the following:
changes in process or operating conditions are required before
10 CFR 50 Licensing of Production and Utilization Facilities
a criticality accident is possible.
10 CFR 50, App B Quality Assurance Criteria for Nuclear
3.2.5 eructation—a surface eruption in a tank, vessel, or
Power Plants and Fuel Reprocessing Plants
liquefied pool caused by the spontaneous release of gas or
2.6 This guide does not purport to list all standards, codes,
vapor, or both, from within the liquid. An eructation may bear
or federal regulations, or combinations thereof that may apply
some resemblance to the flashing of superheated water; but it
to nuclear fuel dissolution facilities design.
best resembles a burping action that may or may not be
accompanied by dispersion of liquid droplets or particulates, or
3. Terminology
both, and by a variable degree of liquid splashing. The
potential for eructation is most often caused by an excessive
3.1 General:
heating rate combined with an inadequate agitation condition.
3.1.1 The terminology used in this guide is intended to
conform with industry practice insofar as is practicable, but the 3.2.6 geometrically favorable—a term applied to a vessel or
following terms are of a restricted nature, specifically appli- system having dimensions and a shape or configuration that
cable to this guide. Other terms and their definitions are provides assurance that a criticality incident cannot occur in the
contained in the ANS Glossary. vessel or system under a given set of conditions. The given
3.1.2 For definitions of general terms used to describe the conditions require that the isotopic composition, form,
design, fabrication, and installation of nuclear fuel dissolution concentration, and density of fissile materials in the system will
facilities refer to terminology in Terminology C859. duplicate those used in preparation of the criticality analysis.
These variables will remain within conservatively chosen
limits, and moderator and reflector conditions will be within
For referenced ASTM standards, visit the ASTM website, www.astm.org, or
contact ASTM Customer Service at [e-mail]. For Annual Book of ASTM some permitted range.
Standards volume information, refer to the standard’s Document Summary page on
3.2.7 poison or poisoned—any material used to minimize
the ASTM website.
the potential for criticality, usually containing quantities of one
The last approved version of this historical standard is referenced on
www.astm.org.
of the chemical elements having a high neutron absorption
Available from American Society of Mechanical Engineers (ASME), ASME
cross-section, for example, boron, cadmium, gadolinium, etc.
International Headquarters, Two Park Ave., New York, NY 10016-5990, http://
www.asme.org.
5 4. Significance and Use
Available from American Nuclear Society, 555f N. Kensington Ave., La Grange
Park, IL 60526.
4.1 The purpose of this guide is to provide information that
Available from U.S. Government Printing Office Superintendent of Documents,
will help to ensure that nuclear fuel dissolution facilities are
732 N. Capitol St., NW, Mail Stop: SDE, Washington, DC 20401, http://
www.access.gpo.gov. conceived, designed, fabricated, constructed, and installed in
C1062 − 23
an economic and efficient manner. This guide will help acceptance criteria by which attainment or conformity is to be
facilities meet the intended performance functions, eliminate or judged. Attainment or conformity verification requirements
minimize the possibility of nuclear criticality and provide for should be written into the Quality Assurance Inspection pro-
the protection of both the operator personnel and the public at cedures.
large under normal and abnormal (emergency) operating con-
5.6 Records Retention—All records pertaining to the basic
ditions as well as under credible failure or accident conditions.
data, design calculations, computer analysis, quality, quality
assurance, chemical or physical test results, inspections, and
5. General Requirements
other records that bear on the condition, safety, or integrity of
5.1 Basic Data and Design Criteria—The fundamental data
the dissolution system facilities shall be available for audit
and design criteria that form the basis for facilities design shall
purposes at any time subsequent to their creation.
be documented in an early stage such that evolving plant
concepts and engineering calculations have a solid and trace-
6. Equipment
able origin or foundation. Design criteria can be included in an
6.1 Design Considerations—The general principles used to
owner/client prepared data document or, when the owner/client
design dissolvers for nuclear fuels are essentially the same as
so instructs, they may be selected or developed by the
those widely employed in the design of processing equipment
responsible design, organization. Values, formulas, equations,
in the chemical industry. Design of nuclear processing facilities
and other data should derive from proven and scientifically and
presents three additional considerations: the possibility of
technically sound sources. Any and all changes to the basic
nuclear criticality, the dissipation of heat created by radioactive
data shall be documented and dated. Procedural requirements
decay, and the provision for the adequate containment of
associated with the authentication, documentation, and reten-
radioactive contaminants under both normal and abnormal
tion of the data base should be essentially equivalent to, and
conditions. The latter consideration demands a degree of
meet the intent of, ASME NQA-1.
quality and the application of quality assurance procedures that
5.2 Responsibility for Basic Data—The production,
are in excess of those that are normally required in the
authentication, and issue of the basic data document should be
chemical industry.
the responsibility of the owner/client. However, this responsi-
6.1.1 General considerations and accepted good practice in
bility may be delegated.
regard to the design of dissolvers and other processing vessels
5.2.1 The Architect-Engineering (AE) organization charged
for nuclear and radioactive materials is contained in guide
with design and engineering responsibility for the nuclear fuel
C1217.
dissolution facilities is generally held responsible for the
6.1.2 Design of dissolution equipment and facilities shall
adequacy, appropriateness, and completeness of the basic data.
include provisions to minimize the release of radioactive
The AE shall indicate the acceptance of this responsibility by
material from process vessels and equipment (including pipes
a signed client/AE acceptance document in testimony thereof.
or lines connecting to vessels or areas that are not normally
Such an acceptance document should be executed within 90
contaminated with radioactive material, such as cold reagent
days after receipt of the basic data document.
and instrument air) or confinement (for example, shielding cell
5.3 Quality Assurance—A formalized quality assurance pro-
walls) during normal and foreseeable abnormal conditions of
gram shall be conducted as required by 10 CFR 50, App B.
operation, maintenance, and decontamination.
This program shall be in general accordance with ASME
6.1.3 Offgas, vapor, droplet, and foaming disengagement
NQA-1.
space, equivalent to approximately 100 % freeboard should be
5.4 Personnel—Personnel associated with facility design included in sizing the dissolver. The dissolver fuel baskets
should be sized so that the fuel charge occupies no more than
and construction should collectively have the training,
experience, and competence to understand, analyze, engineer, 75 % of the basket depth. This will help to ensure confinement
of hulls and metal fragments during the dissolution cycle. Fuel
and resolve questions or problems associated with their as-
signed tasks. basket perforations (openings) should be limited in size to
retain metal fragments and yet allow free flow of dissolvent
5.4.1 Records shall be kept showing names and responsi-
bilities of personnel involved with and responsible for the solutions.
design, fabrication, inspection, and installation of nuclear fuel
6.1.4 Design should specify the controls and checks that are
dissolving facilities for purposes of auditing quality assurance
required to ensure that vessel design dimensions are achieved
(QA) records.
and maintained during fabrication and construction sequences.
This is a requirement for vessels designed to provide geometri-
5.5 Degree of Quality—The quality and integrity of materi-
cally favorable handling conditions for fissile materials.
als and workmanship associated with the design, fabrication,
6.1.5 Criticality assessment calculations (see 8.1) shall in-
and installation of nuclear fuels dissolution facilities shall be
clude an allowance to compensate for vessel fabrication
commensurate with calculated, demonstrable needs. Such
inaccuracies and corrosion. This compensatory calculation
needs arise from known and perceived risks, given physical
allowance is not to be construed as establishing or altering
and chemical principles, and applicable codes and regulations.
given dimensions or tolerances on design drawings.
5.5.1 In setting forth the need for any given level of quality
or integrity, the organization or individual responsible for 6.1.6 The layout and installation of equipment and piping
making any such determination shall document the tests and for the processing and transfer of aqueous solutions of enriched
C1062 − 23
uranyl nitrate should be in accordance with the requirements 8.1.1.2 Adding soluble neutron absorbers (poisons) with the
and constraints set forth in ANSI/ANS 8.9. dissolver solvent and other influent streams.
6.1.7 A gas sparge connection should be included in the
8.1.1.3 Controlling fissile material concentrations below
dissolver. Gas sparging serves as an aid to dissolution,
safe concentration limits.
agitation, and the removal of fission product gases such as
8.2 Design Considerations:
iodine, krypton, and xenon.
8.2.1 Geometry—In the development of the design for a
6.1.8 The layout of dissolver internals, vessel shape and
geometrically favorable nuclear fuel dissolving system, many
profiles, and the placement of sparger nozzles should accom-
precautions must be taken. Some of these special design
modate thorough hydraulic flushing of the bottom of the
considerations are as follows:
dissolver in order to facilitate the removal of sludges and
8.2.1.1 The system shall be designed for the most reactive
metallic fines.
fuel configuration likely to be encountered during the operating
6.1.9 The dissolution cycle vessels should contain provi-
life of the dissolver. Both expected variations in operating
sions for sampling liquid contents.
conditions and credible off-standard and accident conditions
7. Fuel Types
should be considered.
8.2.1.2 Suitable allowances shall be made in selecting
7.1 Cladding and Core Combinations—Nuclear fuels are
geometrically favorable slab thicknesses and cylinder diam-
invariably fabricated with a corrosion resistant metal cladding
eters to allow for fabrication tolerances and for expected
material covering the nuclear material in the core. The core
corrosion over the design lifetime of the vessels (see 6.1.5). It
material is exposed for dissolution by either chemical removal
may also be necessary to provide an allowance for slab
of the cladding or by mechanical chopping to expose the core.
distortion under maximum fill level and design pressure load
7.1.1 Some of the methods that have been used for cladding
conditions, or to provide stays or reinforcement such as to
removal or core exposure treatment, or both, are listed in Table
prevent distortion or variations in slab thickness under design
1.
and operational load conditions.
7.1.2 Core dissolution has been achieved almost exclusively
with hot nitric acid except for some very special fuels (see 8.2.1.3 Fissile material fines or precipitates may be inten-
Appendix X1). tionally or accidentally generated during the dissolution pro-
cess. The dissolver design must include provisions for safely
8. Criticality
accommodating them to a noncritical array. They can either be
removed from the system as generated, or provisions must be
8.1 General Considerations—Candidate dissolver (and dis-
included in the design of the dissolver for their safe accumu-
solver solutions hold/transfer vessel) concepts shall undergo a
lation and later removal (for example, in slabs or cylinders of
criticality assessment analysis prepared by a qualified engineer
geometrically favorable dimensions for these more nuclear-
or physicist, and the analysis shall be subject to a QA
reactive materials). Special precautions and design provisions
verification audit to ensure procedural and computational
are necessary in order to ensure that during removal operations,
accuracy. The calculational method and audit should satisfy the
the solids are not redisbursed into an unsafe geometry at
conditions of ANS 8.1. The analysis and audit should be
another location.
repeated at intervals during the design and operating sequences
as changes occur and as necessary to ensure that safe condi- 8.2.1.4 If heating or cooling jackets, or both, are included on
tions will prevail throughout the equipment’s life cycle. geometrically favorable cylinders or slabs, the geometrically
8.1.1 The need for and the extent of criticality control in the favorable dimension should include the thickness of the jacket,
processing of irradiated nuclear fuel is governed by the isotopic or special provisions should be included to prevent leakage of
composition of the fuel and by many other factors. In the dissolver solution into the jacket. (See 8.1.)
dissolution of nuclear fuels that are more enriched than natural
8.2.1.5 Dissolver dimensions should be fixed in such a
uranium (for example, that have a U content in excess of
manner as to prevent the introduction or charging of fuel in
approximately 0.72 %), precautions must be taken to prevent
amounts in excess of those provided for in the criticality
formation of a critical configuration. In designing a safe
analysis. This assumes that administrative controls will prevent
dissolver system capable of holding more than one critical
the charging of fuels having a higher fissile element content
mass, the following three methods, either alone or in
than that for which the dissolver was designed.
combination, are generally used and recommended for ensur-
8.2.1.6 Dissolver instrumentation shall be capable of pro-
ing nuclear safety:
viding an accurate assessment of vessel contents to the extent
8.1.1.1 Using subcritical geometry (for example, geometri-
that this is practicable and possible. Consideration may be
cally favorable vessel dimensions).
given to the installation of duplicate instruments when such
instrumentation is critical to safe operation and control of the
dissolver.
TABLE 1 Core Exposure Methods Cladding Material
8.2.1.7 The dissolution system shall be designed consistent
Zirconium Stainless
Core Aluminum
with the double contingency principle.
Alloy Steel
Oxide . . . Chop/Chemical Chop/Chemical
8.2.1.8 Nuclear interaction between the dissolver contents
Metal Chemical Chemical . . .
and the immediate environment at the installation location of
Alloy . . . Chop . . .
the dissolver shall be evaluated in developing its design.
C1062 − 23
8.2.1.9 Nuclear interaction between the contents of nearby 8.2.3 Nuclear Fuel Services, Inc. (NFS) Design—For the
or adjacent vessels in the vicinity of the dissolver shall be dissolution of power reactor fuels, dissolver designs have been
developed that use thin slabs (straight slabs or annular cylin-
evaluated when either of the volumes under consideration
ders) or long cylinders of subcritical dimensions. A typical
contains fissile materials. Neutron reflection from cell walls,
example of subcritical geometry, used in combination with
floors and ceilings, and from other nearby objects (for example,
concentration control, was the batch dissolver designed for use
equipment, piping, personnel) for a specific installation loca-
in the West Valley plant of Nuclear Fuel Services, Inc. (NFS).
tion shall also be considered. The geometrically favorable
The design employed six fuel baskets that were 8 ft (244 cm)
dimension(s) shall be reduced appropriately to take into
high, and were 8 in. (20 cm) or less in diameter. One basket
account any interaction between vessels’ contents and to
(with the enclosed fuel charge) was loaded into each of the six
account for the presence of interconnecting piping and appur-
cylinder ports with diameters of 10 in. (25 cm). The basket
tenances. In some instances, such as that in the NFS dissolver
diameter and the fuel loading selected for a particular fuel was
design discussed in 8.2.3, interaction between geometrically
one that limited the fissile materials concentration in the
favorable component shapes can be minimized or essentially
peripheral annulus to a width of 3 in. (8 cm) and the 10 in.
eliminated by interposing moderating materials (for example,
(25 cm) cylindrical areas to 60 % of the calculated critical
concrete) and neutron capture materials (for example,
concentration value when the fuel was dissolved. Nuclear
gadolinium, cadmium, boron) between the geometrically fa-
interaction between the six cylindrical sections was minimized
vorable compartments of a vessel.
by addition of natural boron with a mass fraction of 0.5 % to
8.2.1.10 For dissolver systems designed for less than full
the concrete core section of the dissolver that was positioned
neutron reflection (for example, dissolvers designed as geo-
and sized so as to provide for a minimum separation of 30 in.
metrically favorable configurations for mounting or placement
(76 cm) between the 10 in. (25 cm) diameter cylindrical areas.
in air cells), special precautions must be taken and operational
8.2.4 Allied-General Nuclear Services (AGNS) Design—
constraints invoked to ensure that excessive cell flooding is
Although the plant was not operated using irradiated fuels, the
precluded and that significant amounts of neutron reflecting
Allied-General Nuclear Services (AGNS) Barnwell plant dis-
and moderating materials are not brought into the immediate
solver illustrated a design using a soluble neutron poison in the
vicinity of the dissolver. This would include prohibitions
dissolver. It was intended that sufficient natural gadolinium (as
against the placement of another vessel in near proximity to the
gadolinium nitrate) be added to the nitric acid dissolvent such
dissolver in the cell, unless the criticality analysis is recalcu-
that no criticality would occur based on the fissile concentra-
lated and appropriate design changes are made.
tion of the unirradiated fuel (initial enrichment) to be dis-
8.2.1.11 Sumps designed to collect solutions that leak out
solved.
of, or overflow from, dissolvers shall also be of safe design;
8.2.5 Mention of specific dissolver designs does not consti-
that is, they shall have geometrically favorable dimensions or
tute an endorsement of one concept versus another. Other
other provisions such as poisoned raschig rings. Sumps should
critically safe dissolver designs are equally acceptable.
be designed to collect safely the maximum amount of liquid
8.3 Operating Considerations:
likely to come out of any one process vessel in a“ worst case”
8.3.1 Soluble Poisons—If soluble poisons are used to pro-
design basis accident (DBA) scenario. The sumps shall be
vide nuclear safety, the nuclear poison concentration selected
equipped with instrumentation and alarms that notify operating
shall be capable of ensuring dissolver nuclear safety for the
personnel of abnormal sump accumulations. Pumps, eductors,
most reactive fuel mixture to be processed.
or jets should be installed for moving solutions containing
8.3.1.1 The dissolver and associated dissolution system
fissile materials out of the sumps into a vessel having a
equipment shall be operated under conditions that ensure that
geometrically favorable shape and which is positioned in a
the poison concentration in the systems remains within the
manner such that the addition of sump contents will not initiate
prescribed range and that the nuclear poison remains in
a criticality incident due to interaction with adjacent vessels or
solution during normal operating conditions under predictable
masses.
abnormal operating conditions and under credible accident
8.2.1.12 When fuel reprocessing operations involve han-
conditions. Cold feed solutions that have the capability for
dling of fissile materials in amounts sufficient to create a
precipitation of either the soluble poison or the fissile materials
potential criticality hazard, the load conditions established for
should not be directly connected to (piped into) the dissolver.
vessel design shall include the potential shock loads and lateral
If such piping connections are employed, the lines shall contain
forces that may result from a design basis seismic event. The
lockable valving under supervisory control or other flow
forces developed by the design basis earthquake (DBE) shall
blockage provisions.
be accommodated by the vessel design without vessel collapse
8.3.1.2 When cooling jackets or heating jackets, or both, are
or distortion that would render a geometrically favorable shape
provided on a poisoned dissolver, the effects of coil or jacket
or dimension to be altered in such a manner as to allow a
heat transfer media leakage into the dissolver shall be consid-
criticality incident to occur in the vessel.
ered since dilution of the poison could produce a more reactive
8.2.2 Soluble Poisons—The use of soluble poisons, for
condition. Inclusion of poison in the cooling or heating media
example, chemical elements having high neutron absorption should be considered. Design must also consider the potential
cross-sections, in an alternative or supplementary method of
for leakage of fissile material solutions into heating and cooling
reducing the potential for a criticality incident. circuits and provide protection against conveyance of such
C1062 − 23
materials into areas occupied by operator personnel or into sludge tank. Extended leaching and rinse operations are carried
auxiliary systems equipment where criticality may potentially out in order to reduce the fissile material content to specifica-
occur. tion levels prior to removal and disposal of the sludge as waste.
8.3.2 Administrative Control of Charge Mass—Operational 9.1.3.1 Zirconium alloy fines and small pieces constitute a
spontaneous fire hazard. Zirconium alloy hulls that have been
control over the accumulation of a critical mass in the dissolver
vessel is an active means of preventing a criticality incident but fully stripped of heavy metal values are rinsed and passivated
with a caustic solution. It is recommended that the passivation
one which provides an added measure of protection. As
inferred, this is primarily an operational procedure, but facili- step be carried out in an inert (argon) atmosphere to prevent
fires.
ties design shall provide the informational feedback, through
instrumentation to enhance operational control. 9.1.4 Decay and Reaction Heat Control—It is recom-
mended that the dissolver incorporate separate heating and
cooling provisions (for example, coils) to allow close control
9. Dissolution
over dissolution solution temperatures and reaction rates dur-
9.1 Design Considerations—Dissolution processes are out-
ing both the cladding and the fuel dissolution steps, and to
lined in Appendix X1. Operating considerations incidental to
provide for temperature control in instances where exothermal
the use of each of the processes are discussed therein. Design
reactions occur. Heat removal capacity (coils or jacket heat
shall anticipate operation over a wide range of temperature,
transfer area or temperature differences) shall be sufficient to
pressure, and reaction rate conditions and use adequate mar-
remove the radioactive decay heat load as well as the reaction
gins of safety in the design. Some of the safety considerations,
heat.
and the sources of hazards and their mitigation or control, are
9.1.4.1 For those dissolvers employing heating jackets or
discussed in Appendix X3.
coils, and where control of the final concentration of the
9.1.1 Chemical Reactivity—The dissolver and the dissolver
dissolver solution is important, the heat transfer area should be
offgas handling and treatment equipment shall be designed as
positioned somewhat above the bottom of the dissolver at a
a complete entity, sized to handle the offgas load from the most
level that prevents over-concentration (by boil-up) of fissile
reactive dissolution chemistry that can be predicted for the
material solutions. Concentration, except for that which might
dissolver design and potential fuel charges being considered.
occur as a result of self-heating, would cease when heat
Typically, the offgas system capacity should be capable of
transfer surfaces are no longer submerged.
accommodating offgas surge rates or burps in the range of five
9.1.4.2 Cooling coils or jackets should be positioned in
to eight times the normal (production) processing rate over a
processing vessels in such a way as to be fully submerged
one to three minute time period. However, if the chemical
when vessels are filled to their normal operating levels. The
reactivity is controlled through solvent (acid) availability, the
heat transfer surface for cooling shall extend near to the bottom
offgas system should be capable of accommodating offgas
of the vessels in order to provide the means for removal of
surge rates of 1.3 to 1.5 times the normal processing rate.
decay heat from residual amounts of process solutions left in
9.1.1.1 Nuclear fuel dissolution sequences have many
the vessels.
similarities, but the sequential steps for any one process may
9.1.4.3 Dissolver steam and cooling water supplies should
not be fully applicable to other nuclear fuel dissolution cycles.
have temperature-activated interlocks. Settings of the inter-
9.1.1.2 The metal charge, in the form of chopped/sheared
locks should be fixed at points that will prevent overheating
fuel pins one to three inches long, is frequently added in
and excessive boil-up of process solutions and at points that
perforated metal baskets. For these cases, the dissolver design
will automatically introduce cooling water flow to cooling coils
may incorporate remotely operable provisions to raise the
in the event that set points for the vessel temperature are
charge basket above the solution level. This provides an
exceeded.
alternative means of reaction rate control for emergency use in
9.2 Operating Considerations—Fuels in particulate form
the event that the reaction rate becomes excessive, to the extent
are highly reactive in acid solutions. It is recommended that
that the offgas evolution rate threatens to overtax the capacity
dissolution cycles anticipate the presence of significant quan-
of the offgas treatment system.
tities of fines. Assuming that such a condition exists, operators
9.1.2 Corrosion—A variety of chemicals can be used to
should start each dissolution cycle with the use of dilute acid
dissolve particular fuels and residue sludges that may remain in
and chemical inhibitors that modify, and have a controlling
the dissolver at the conclusion of the dissolution cycle. The
effect on, the dissolution reaction chemistry.
designer must anticipate these, select appropriate materials of
9.2.1 The administrative and technical practices for critical-
construction, and provide a corrosion allowance that tends to
ity safety and control should conform with or meet the intent of
ensure contents confinement integrity over the design life of
those practices set forth in ANS 8.1.
the vessel. Organic acids and other chemicals used in decon-
tamination sequences need consideration, and the corrosive
10. Dissolver Vapors and Offgas
effects of ions released during the chemical dissolution cycle
should also be considered. Accelerated corrosion tests on
10.1 Design Considerations for Offgas Treatment—
candidate materials of construction are recommended.
Dissolver offgases generally pass through several sequential
9.1.3 Residues—The accumulation of metal fines and undis- treatment steps. The offgas treatment requirements depend on
solved fission products as a sludge in the dissolver will require the dissolution chemistry, the composition of the spent fuel
the capability for flush-out and removal of this material to a being dissolved, the gaseous and volatile radionuclides, other
C1062 − 23
contaminants in the offgas stream, and other factors. Treatment accumulation of these solids in the offgas equipment train. The
of dissolver vapors and offgas ensures that valuable process design of the scrubber shall accommodate recovery and recy-
materials are recovered, and both radioactive materials and any cling of the solids and fines and shall prevent a criticality
noxious or undesirable gas/vapor stream constituents are re- incident that might potentially occur through inadvertent accu-
moved to the extent practicable or required. Treatment methods mulation of fissile material fines.
for the removal of any particular offgas constituent may vary. 10.2.1 The condenser section of the dissolver should be
Typical offgas treatment steps are briefly described in the designed as a total reflux condenser, to return condensed
following paragraphs. Mention of a particular offgas treatment liquids to the dissolver, and to promote acid economy.
process is for purpose of illustration and does not constitute an Typically, gases are passed downwards through the condenser.
endorsement of the procedure as the best or only method for The condenser capacity should be sufficient to cope with peak
removal of contaminants from the offgas stream. boil-up and offgas loads without excessive pressure drop and
10.1.1 The treated offgas stream shall meet release criteria consequent pressurization of the dissolver assembly. The
condenser should be equipped with an acid spray connection to
for toxic and radioactive contaminants as established by law
and by basic data specifications. permit wash-down and decontamination of the coil assembly.
The design of the condenser and scrubber should provide for
10.1.2 The offgas systems for dissolvers are generally
reducing the temperature of the offgas stream to the ambient
designed to handle vapors or condensates, or both, that contain
cell or canyon temperature, or lower if practicable.
very low concentrations of fissile materials and are generally
10.2.2 The removal of NO gases may require the inclusion
not designed as a geometrically favorable system configura-
x
of a multi-tray absorption column, or other NO removal
tion. If foaming were to be encountered or excessive entrain-
x
methods such as the use of synthetic mordents in a packed
ment were experienced, dissolver solution or fissile fines could
column to catalyze selectively the ammonia reduction of NO
be carried into the offgas handling system. Special design
x
gases.
provisions to prevent or to mitigate dissolver foaming condi-
10.2.3 An atomized steam-driven or pumped solution jet
tions shall be considered. As a minimum, dissolver system
scrubber provides a means of solids removal, as well as means
design should include provisions and operating procedures, or
of cooling the offgas stream and assisting in the removal of
both, to return such carry-over materials to the dissolver and to
NO gases. The scrubber jet(s) may also serve as part of the
prevent their accumulation in the offgas system (for example,
x
vacuum system. When such a treatment step is included in the
vapor and offgas decontamination devices). Design provisions
offgas system, the motive system for maintaining a vacuum
(for example, overflows, instrumentation, and alarms) and
condition in the dissolver should be backed by an installed
operating precautions shall prevent flooding of the offgas
spare (alternative) vacuum-producing component or system
handling system with dissolver solution.
that will prevent over-pressurizing the dissolver in the event of
10.1.3 Specific design features shall be considered to ensure
steam or pump failure.
an adequate offgas flow control capability during all phases of
the dissolver operation (for example, charging, dissolution,
10.3 Ruthenium (Ru) Removal—The use of a silica gel bed
solutions transfer, reaction surges, standby, etc). A means of
is one of a number of accepted and effective processes for the
vacuum regulation (such as a vacuum breaker) shall be
removal of particulate or volatile Ru from the scrubbed offgas
included in the dissolver system design to avoid an excessive
stream.
vacuum on the dissolver, or one that could breach liquid seals
10.4 Iodine Removal—Silver-exchanged mordenite beds in
or upset weight factor instrumentation.
series is one accepted and effective process for the removal of
10.1.4 Designs based on low air in-leakage rates to the
iodine. The beds operate at a temperature of 150 °C. Silver-
dissolver offgas system should ensure that the low design basis
exchanged mordenite beds loaded with iodine are regenerated
rates can be maintained during the entire life cycle for the
with hydrogen. Iodine produced in the regeneration cycle is
facilities. The integrity and characteristics of closures design
collected on lead-based absorption beds.
would be a prime consideration here.
10.5 Krypton-85 Removal—One suggested process for the
10.1.5 Design of the offgas system shall include a pressure
removal of Kr from an offgas stream features a selective
relief system or component to limit the maximum dissolver
absorption step using refrigerant R-12 (dichlorodifluorometh-
system pressure to 3 to 5 psig, or to the design pressure limits
ane) as the absorption medium.
for the vent system. The relief system shall reset automatically.
10.1.6 Provisions should be included to permit periodic
10.6 Tritium Removal—Tritium may be recovered by oxi-
flushing of all offgas lines and equipment. Provisions to collect
dation and sorption techniques. One process is based on the
the flush water, together with any accumulated solids or
addition of excess hydrogen to the offgas stream that then
deposited fission products, or both, that are flushed out, are
passes through a Ni-Cr-Pd ribbon catalyst bed to oxidize the
necessary as part of the flush system.
hydrogen isotopes to HTO. The unit operates at 400 °C. The
HTO is then preferentially sorbed on molecular sieves (zeo-
10.2 Moisture and NO Removal—The removal of dusts,
x
lite).
excess moisture and NO (oxides of nitrogen) gases may be
x
affected by scrubbing, condensation, and adsorption tech- 10.7 Carbon-14 Removal—Carbon-14 may be removed as
niques. Oxygen addition may be employed to enhance NO CO gas by adsorption on zeolite molecular sieve beds. The
x 2
recovery. The offgas scrubber step is intended to remove solid CO gas is driven off the sorbent bed during periodic regen-
particulates carried off in the offgas stream and prevent the eration cycles and is adsorbed on a BaOH bed.
C1062 − 23
11. Dissolver Product Handling 12.1.3 The design of fuel transfer tubes or chutes with
valves or diversion gates to permit the intermittent addition of
11.1 Design Considerations—Metal solution from the fuel
sheared fuel to the dissolver shall be accomplished in a manner
dissolution step undergoes several treatments and measure-
that ensures mechanical failures will not create a hazardous
ments before being chemically adjusted so as to be suitable for
condition. The design shall also ensure that maintenance (or
feed to solvent extraction. The dissolver solution may be
removal and replacement) can be accomplished by remote
processed through a feed clarification step such as centrifuga-
handling techniques or other safe (low exposure) maintenance
tion. The accountability datum is established in a dedicated
procedures. The charge tube design shall effect control over the
accountability tank. Nuclear fuel dissolution systems design
escape of vapors or gases from closed systems.
shall consider the potential for criticality in dissolver solutions
12.1.4 The charging transfer operations equipment shall be
hold and transfer tankage as well as in the dissolver vessel (see
designed and installed in a manner such as to permit avoidance
Section 8). The use of a dilution eductor for solutions transfer
of or recovery from known and predictabl
...

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: C1062 − 00 (Reapproved 2014) C1062 − 23
Standard Guide for
Design, Fabrication, and Installation of Nuclear Fuel
Dissolution Facilities
This standard is issued under the fixed designation C1062; 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 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
This guide is under the jurisdiction of ASTM Committee C26 on Nuclear Fuel Cycle and is the direct responsibility of Subcommittee C26.09 on Nuclear Processing.
Current edition approved June 1, 2014April 1, 2023. Published June 2014July 2023. Originally approved in 1986. Last previous edition approved in 20082014 as
C1062 – 00 (2008).(2014). DOI: 10.1520/C1062-00R14.10.1520/C1062-23.
Copyright © ASTM International, 100 Barr Harbor Drive, PO Box C700, West Conshohocken, PA 19428-2959. United States
C1062 − 23
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 units that are provided for information only and are not considered standard.
1.4 This standard does not purport to address all of the safety concerns, if any, associated with its use. It is the responsibility
of the user of this standard to establish appropriate safety and healthsafety, 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.
2. Referenced Documents
2.1 Industry and National Consensus Standards—Industry and national consensus standards applicable in whole or in part to the
design, fabrication, and installation of nuclear fuel dissolution facilities are referenced throughout this guide and include the
following:
2.2 ASTM Standards:
C859 Terminology Relating to Nuclear Materials
C1010 Guide for Acceptance, Checkout, and Pre-Operational Testing of a Nuclear Fuels Reprocessing Facility (Withdrawn
2001)
C1217 Guide for Design of Equipment for Processing Nuclear and Radioactive Materials
2.3 ASME Standards:
ASME Boiler and Pressure Vessel Code, Sections II, V, VIII, and IX
ASME NQA-1 Quality Assurance Requirements for Nuclear Facility Applications
2.4 ANS Standard:
ANS Glossary of Terms in Nuclear Science and Technology (ANS Glossary)
ANS 8.1 Nuclear Criticality Safety in Operations with Fissionable Materials Outside Reactors
ANS 8.3 Criticality Accident Alarm System
ANS 8.9 Nuclear Criticality Safety Criteria for Steel-Pipe Intersections Containing Aqueous Solutions of Fissile Materials
ANS 57.8 Fuel Assembly Identification
2.5 Federal Regulations —Federal Regulations that are specifically applicable in whole or in part to the design, fabrication, and
installation of nuclear fuel dissolution facilities include the following:
10 CFR 50 Licensing of Production and Utilization Facilities
10 CFR 50, App B Quality Assurance Criteria for Nuclear Power Plants and Fuel Reprocessing Plants
2.6 This guide does not purport to list all standards, codes, or federal regulations, or combinations thereof that may apply to
nuclear fuel dissolution facilities design.
3. Terminology
3.1 General:
For referenced ASTM standards, visit the ASTM website, www.astm.org, or contact ASTM Customer Service at [e-mail]. For Annual Book of ASTM Standards
volume information, refer to the standard’s Document Summary page on the ASTM website.
The last approved version of this historical standard is referenced on www.astm.org.
Available from American Society of Mechanical Engineers (ASME), ASME International Headquarters, Two Park Ave., New York, NY 10016-5990, http://
www.asme.org.
Available from American Nuclear Society, 555f N. Kensington Ave., La Grange Park, IL 60526.
Available from U.S. Government Printing Office Superintendent of Documents, 732 N. Capitol St., NW, Mail Stop: SDE, Washington, DC 20401, http://
www.access.gpo.gov.
C1062 − 23
3.1.1 The terminology used in this guide is intended to conform with industry practice insofar as is practicable, but the following
terms are of a restricted nature, specifically applicable to this guide. Other terms and their definitions are contained in the ANS
Glossary.
3.1.2 For definitions of general terms used to describe the design, fabrication, and installation of nuclear fuel dissolution facilities
refer to terminology in Terminology C859.
3.1.3 shall, should, and may—The word “shall” denotes a requirement, the word “should” denotes a recommendation and the word
“may” indicates permission, neither a requirement nor a recommendation. In order to conform with this guide, all actions or
conditions shall be in accordance with its requirements but they need not conform with its recommendations.
3.2 Definitions of Terms Specific to This Standard:
3.2.1 accident—an unplanned event that could result in unacceptable levels of any of the following:
3.2.1.1 equipment damage,
3.2.1.2 injury to personnel,
3.2.1.3 downtime or outage,
3.2.1.4 release of hazardous materials (radioactive or nonradioactive).
3.2.1.5 radiation exposure to personnel, and
3.2.1.6 criticality.
3.2.2 accountability—the keeping of records on and the responsibility associated with being accountable for the amount of fissile
materials entering and leaving a plant, a location, or a processing step.
3.2.3 basic data—the fundamental chemical, physical, and mathematical values, formulas, and principles, and the definitive
criteria that have been documented and accepted as the basis for facilities design.
3.2.4 double contingency principle—the use of methods, measures, or factors of safety in the design of nuclear facilities such that
at least two unlikely, independent, and concurrent changes in process or operating conditions are required before a criticality
accident is possible.
3.2.5 eructation—a surface eruption in a tank, vessel, or liquefied pool caused by the spontaneous release of gas or vapor, or both,
from within the liquid. An eructation may bear some resemblance to the flashing of superheated water; but it best resembles a
burping action that may or may not be accompanied by dispersion of liquid droplets or particulates, or both, and by a variable
degree of liquid splashing. The potential for eructation is most often caused by an excessive heating rate combined with an
inadequate agitation condition.
3.2.6 geometrically favorable—a term applied to a vessel or system having dimensions and a shape or configuration that provides
assurance that a criticality incident cannot occur in the vessel or system under a given set of conditions. The given conditions
require that the isotopic composition, form, concentration, and density of fissile materials in the system will duplicate those used
in preparation of the criticality analysis. These variables will remain within conservatively chosen limits, and moderator and
reflector conditions will be within some permitted range.
3.2.7 poison or poisoned—any material used to minimize the potential for criticality, usually containing quantities of one of the
chemical elements having a high neutron absorption cross-section, for example, boron, cadmium, gadolinium, etc.
4. 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
C1062 − 23
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.
5. General Requirements
5.1 Basic Data and Design Criteria—The fundamental data and design criteria that form the basis for facilities design shall be
documented in an early stage such that evolving plant concepts and engineering calculations have a solid and traceable origin or
foundation. Design criteria can be included in an owner/client prepared data document or, when the owner/client so instructs, they
may be selected or developed by the responsible design, organization. Values, formulas, equations, and other data should derive
from proven and scientifically and technically sound sources. Any and all changes to the basic data shall be documented and dated.
Procedural requirements associated with the authentication, documentation, and retention of the data base should be essentially
equivalent to, and meet the intent of, ASME NQA-1.
5.2 Responsibility for Basic Data—The production, authentication, and issue of the basic data document should be the
responsibility of the owner/client. However, this responsibility may be delegated.
5.2.1 The Architect-Engineering (AE) organization charged with design and engineering responsibility for the nuclear fuel
dissolution facilities is generally held responsible for the adequacy, appropriateness, and completeness of the basic data. The AE
shall indicate the acceptance of this responsibility by a signed client/AE acceptance document in testimony thereof. Such an
acceptance document should be executed within 90 days after receipt of the basic data document.
5.3 Quality Assurance—A formalized quality assurance program shall be conducted as required by 10 CFR 50, App B. This
program shall be in general accordance with ASME NQA-1.
5.4 Personnel—Personnel associated with facility design and construction should collectively have the training, experience, and
competence to understand, analyze, engineer, and resolve questions or problems associated with their assigned tasks.
5.4.1 Records shall be kept showing names and responsibilities of personnel involved with and responsible for the design,
fabrication, inspection, and installation of nuclear fuel dissolving facilities for purposes of auditing quality assurance (QA) records.
5.5 Degree of Quality—The quality and integrity of materials and workmanship associated with the design, fabrication, and
installation of nuclear fuels dissolution facilities shall be commensurate with calculated, demonstrable needs. Such needs arise
from known and perceived risks, given physical and chemical principles, and applicable codes and regulations.
5.5.1 In setting forth the need for any given level of quality or integrity, the organization or individual responsible for making any
such determination shall document the tests and acceptance criteria by which attainment or conformity is to be judged. Attainment
or conformity verification requirements should be written into the Quality Assurance Inspection procedures.
5.6 Records Retention—All records pertaining to the basic data, design calculations, computer analysis, quality, quality assurance,
chemical or physical test results, inspections, and other records that bear on the condition, safety, or integrity of the dissolution
system facilities shall be available for audit purposes at any time subsequent to their creation.
6. Equipment
6.1 Design Considerations—The general principles used to design dissolvers for nuclear fuels are essentially the same as those
widely employed in the design of processing equipment in the chemical industry. Design of nuclear processing facilities presents
three additional considerations: the possibility of nuclear criticality, the dissipation of heat created by radioactive decay, and the
provision for the adequate containment of radioactive contaminants under both normal and abnormal conditions. The latter
consideration demands a degree of quality and the application of quality assurance procedures that are in excess of those that are
normally required in the chemical industry.
6.1.1 General considerations and accepted good practice in regard to the design of dissolvers and other processing vessels for
nuclear and radioactive materials is contained in guide C1217.
6.1.2 Design of dissolution equipment and facilities shall include provisions to minimize the release of radioactive material from
process vessels and equipment (including pipes or lines connecting to vessels or areas that are not normally contaminated with
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radioactive material, such as cold reagent and instrument air) or confinement (for example, shielding cell walls) during normal and
foreseeable abnormal conditions of operation, maintenance, and decontamination.
6.1.3 Offgas, vapor, droplet, and foaming disengagement space, equivalent to approximately 100 % freeboard should be included
in sizing the dissolver. The dissolver fuel baskets should be sized so that the fuel charge occupies no more than 75 % of the basket
depth. This will help to ensure confinement of hulls and metal fragments during the dissolution cycle. Fuel basket perforations
(openings) should be limited in size to retain metal fragments and yet allow free flow of dissolvent solutions.
6.1.4 Design should specify the controls and checks that are required to ensure that vessel design dimensions are achieved and
maintained during fabrication and construction sequences. This is a requirement for vessels designed to provide geometrically
favorable handling conditions for fissile materials.
6.1.5 Criticality assessment calculations (see 8.1) shall include an allowance to compensate for vessel fabrication inaccuracies and
corrosion. This compensatory calculation allowance is not to be construed as establishing or altering given dimensions or
tolerances on design drawings.
6.1.6 The layout and installation of equipment and piping for the processing and transfer of aqueous solutions of enriched uranyl
nitrate should be in accordance with the requirements and constraints set forth in ANSI/ANS 8.9.
6.1.7 A gas sparge connection should be included in the dissolver. Gas sparging serves as an aid to dissolution, agitation, and the
removal of fission product gases such as iodine, krypton, and xenon.
6.1.8 The layout of dissolver internals, vessel shape and profiles, and the placement of sparger nozzles should accommodate
thorough hydraulic flushing of the bottom of the dissolver in order to facilitate the removal of sludges and metallic fines.
6.1.9 The dissolution cycle vessels should contain provisions for sampling liquid contents.
7. Fuel Types
7.1 Cladding and Core Combinations—Nuclear fuels are invariably fabricated with a corrosion resistant metal cladding material
covering the nuclear material in the core. The core material is exposed for dissolution by either chemical removal of the cladding
or by mechanical chopping to expose the core.
7.1.1 Some of the methods that have been used for cladding removal or core exposure treatment, or both, are listed in Table 1.
7.1.2 Core dissolution has been achieved almost exclusively with hot nitric acid except for some very special fuels (see Appendix
X1).
8. Criticality
8.1 General Considerations—Candidate dissolver (and dissolver solutions hold/transfer vessel) concepts shall undergo a criticality
assessment analysis prepared by a qualified engineer or physicist, and the analysis shall be subject to a QA verification audit to
ensure procedural and computational accuracy. The calculational method and audit should satisfy the conditions of ANS 8.1. The
analysis and audit should be repeated at intervals during the design and operating sequences as changes occur and as necessary
to ensure that safe conditions will prevail throughout the equipment’s life cycle.
8.1.1 The need for and the extent of criticality control in the processing of irradiated nuclear fuel is governed by the isotopic
composition of the fuel and by many other factors. In the dissolution of nuclear fuels that are more enriched than natural uranium
TABLE 1 Core Exposure Methods Cladding Material
Stainless
A
Core Aluminum Zircaloy
Steel
Zirconium Stainless
Core Aluminum
Alloy Steel
Oxide . . . Chop/Chemical Chop/Chemical
Metal Chemical Chemical . . .
Alloy . . . Chop . . .
A
Zircaloy is a registered trademark of Westinghouse Electric Corporation,
Blairsville, PA.
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(for example, that have a U content in excess of approximately 0.72 %), precautions must be taken to prevent formation of a
critical configuration. In designing a safe dissolver system capable of holding more than one critical mass, the following three
methods, either alone or in combination, are generally used and recommended for ensuring nuclear safety:
8.1.1.1 Using subcritical geometry (for example, geometrically favorable vessel dimensions).
8.1.1.2 Adding soluble neutron absorbers (poisons) with the dissolver solvent and other influent streams.
8.1.1.3 Controlling fissile material concentrations below safe concentration limits.
8.2 Design Considerations:
8.2.1 Geometry—In the development of the design for a geometrically favorable nuclear fuel dissolving system, many precautions
must be taken. Some of these special design considerations are as follows:
8.2.1.1 The system shall be designed for the most reactive fuel configuration likely to be encountered during the operating life of
the dissolver. Both expected variations in operating conditions and credible off-standard and accident conditions should be
considered.
8.2.1.2 Suitable allowances shall be made in selecting geometrically favorable slab thicknesses and cylinder diameters to allow
for fabrication tolerances and for expected corrosion over the design lifetime of the vessels (see 6.1.5). It may also be necessary
to provide an allowance for slab distortion under maximum fill level and design pressure load conditions, or to provide stays or
reinforcement such as to prevent distortion or variations in slab thickness under design and operational load conditions.
8.2.1.3 Fissile material fines or precipitates may be intentionally or accidentally generated during the dissolution process. The
dissolver design must include provisions for safely accommodating them to a noncritical array. They can either be removed from
the system as generated, or provisions must be included in the design of the dissolver for their safe accumulation and later removal
(for example, in slabs or cylinders of geometrically favorable dimensions for these more nuclear-reactive materials). Special
precautions and design provisions are necessary in order to ensure that during removal operations, the solids are not redisbursed
into an unsafe geometry at another location.
8.2.1.4 If heating or cooling jackets, or both, are included on geometrically favorable cylinders or slabs, the geometrically
favorable dimension should include the thickness of the jacket, or special provisions should be included to prevent leakage of
dissolver solution into the jacket. (See 8.1.)
8.2.1.5 Dissolver dimensions should be fixed in such a manner as to prevent the introduction or charging of fuel in amounts in
excess of those provided for in the criticality analysis. This assumes that administrative controls will prevent the charging of fuels
having a higher fissile element content than that for which the dissolver was designed.
8.2.1.6 Dissolver instrumentation shall be capable of providing an accurate assessment of vessel contents to the extent that this
is practicable and possible. Consideration may be given to the installation of duplicate instruments when such instrumentation is
critical to safe operation and control of the dissolver.
8.2.1.7 The dissolution system shall be designed consistent with the double contingency principle.
8.2.1.8 Nuclear interaction between the dissolver contents and the immediate environment at the installation location of the
dissolver shall be evaluated in developing its design.
8.2.1.9 Nuclear interaction between the contents of nearby or adjacent vessels in the vicinity of the dissolver shall be evaluated
when either of the volumes under consideration contains fissile materials. Neutron reflection from cell walls, floors and ceilings,
and from other nearby objects (for example, equipment, piping, personnel) for a specific installation location shall also be
considered. The geometrically favorable dimension(s) shall be reduced appropriately to take into account any interaction between
vessels’ contents and to account for the presence of interconnecting piping and appurtenances. In some instances, such as that in
the NFS dissolver design discussed in 8.2.3, interaction between geometrically favorable component shapes can be minimized or
essentially eliminated by interposing moderating materials (for example, concrete) and neutron capture materials (for example,
gadolinium, cadmium, boron) between the geometrically favorable compartments of a vessel.
8.2.1.10 For dissolver systems designed for less than full neutron reflection (for example, dissolvers designed as geometrically
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favorable configurations for mounting or placement in air cells), special precautions must be taken and operational constraints
invoked to ensure that excessive cell flooding is precluded and that significant amounts of neutron reflecting and moderating
materials are not brought into the immediate vicinity of the dissolver. This would include prohibitions against the placement of
another vessel in near proximity to the dissolver in the cell, unless the criticality analysis is recalculated and appropriate design
changes are made.
8.2.1.11 Sumps designed to collect solutions that leak out of, or overflow from, dissolvers shall also be of safe design; that is, they
shall have geometrically favorable dimensions or other provisions such as poisoned raschig rings. Sumps should be designed to
collect safely the maximum amount of liquid likely to come out of any one process vessel in a“ worst case” design basis accident
(DBA) scenario. The sumps shall be equipped with instrumentation and alarms that notify operating personnel of abnormal sump
accumulations. Pumps, eductors, or jets should be installed for moving solutions containing fissile materials out of the sumps into
a vessel having a geometrically favorable shape and which is positioned in a manner such that the addition of sump contents will
not initiate a criticality incident due to interaction with adjacent vessels or masses.
8.2.1.12 When fuel reprocessing operations involve handling of fissile materials in amounts sufficient to create a potential
criticality hazard, the load conditions established for vessel design shall include the potential shock loads and lateral forces that
may result from a design basis seismic event. The forces developed by the design basis earthquake (DBE) shall be accommodated
by the vessel design without vessel collapse or distortion that would render a geometrically favorable shape or dimension to be
altered in such a manner as to allow a criticality incident to occur in the vessel.
8.2.2 Soluble Poisons—The use of soluble poisons, for example, chemical elements having high neutron absorption cross-sections,
in an alternative or supplementary method of reducing the potential for a criticality incident.
8.2.3 Nuclear Fuel Services, Inc. (NFS) Design—For the dissolution of power reactor fuels, dissolver designs have been developed
that use thin slabs (straight slabs or annular cylinders) or long cylinders of subcritical dimensions. A typical example of subcritical
geometry, used in combination with concentration control, was the batch dissolver designed for use in the West Valley plant of
Nuclear Fuel Services, Inc. (NFS). The design employed six fuel baskets that were 8 ft (244 cm) high, and were 8 in. (20 cm) or
less in diameter. One basket (with the enclosed fuel charge) was loaded into each of the six 10-in. (25 cm) diameter cylindrical
ports. cylinder ports with diameters of 10 in. (25 cm). The basket diameter and the fuel loading selected for a particular fuel was
one that limited the fissile materials concentration in the 3-in. (8 cm) wide peripheral annulus and the 10-in. peripheral annulus
to a width of 3 in. (8 cm) and the 10 in. (25 cm) cylindrical areas to 60 % of the calculated critical concentration value when the
fuel was dissolved. Nuclear interaction between the six cylindrical sections was minimized by addition of 0.5 wt % natural boron
natural boron with a mass fraction of 0.5 % to the concrete core section of the dissolver that was positioned and sized so as to
provide for a minimum of 30-in. (76 cm) separation between the 10-in. (25 cm) separation of 30 in. (76 cm) between the 10 in.
(25 cm) diameter cylindrical areas.
8.2.4 Allied-General Nuclear Services (AGNS) Design—Although the plant was not operated using irradiated fuels, the
Allied-General Nuclear Services (AGNS) Barnwell plant dissolver illustrated a design using a soluble neutron poison in the
dissolver. It was intended that sufficient natural gadolinium (as gadolinium nitrate) be added to the nitric acid dissolvent such that
no criticality would occur based on the fissile concentration of the unirradiated fuel (initial enrichment) to be dissolved.
8.2.5 Mention of specific dissolver designs does not constitute an endorsement of one concept versus another. Other critically safe
dissolver designs are equally acceptable.
8.3 Operating Considerations:
8.3.1 Soluble Poisons—If soluble poisons are used to provide nuclear safety, the nuclear poison concentration selected shall be
capable of ensuring dissolver nuclear safety for the most reactive fuel mixture to be processed.
8.3.1.1 The dissolver and associated dissolution system equipment shall be operated under conditions that ensure that the poison
concentration in the systems remains within the prescribed range and that the nuclear poison remains in solution during normal
operating conditions under predictable abnormal operating conditions and under credible accident conditions. Cold feed solutions
that have the capability for precipitation of either the soluble poison or the fissile materials should not be directly connected to
(piped into) the dissolver. If such piping connections are employed, the lines shall contain lockable valving under supervisory
control or other flow blockage provisions.
8.3.1.2 When cooling jackets or heating jackets, or both, are provided on a poisoned dissolver, the effects of coil or jacket heat
transfer media leakage into the dissolver shall be considered since dilution of the poison could produce a more reactive condition.
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Inclusion of poison in the cooling or heating media should be considered. Design must also consider the potential for leakage of
fissile material solutions into heating and cooling circuits and provide protection against conveyance of such materials into areas
occupied by operator personnel or into auxiliary systems equipment where criticality may potentially occur.
8.3.2 Administrative Control of Charge Mass—Operational control over the accumulation of a critical mass in the dissolver vessel
is an active means of preventing a criticality incident but one which provides an added measure of protection. As inferred, this is
primarily an operational procedure, but facilities design shall provide the informational feedback, through instrumentation to
enhance operational control.
9. Dissolution
9.1 Design Considerations—Dissolution processes are outlined in Appendix X1. Operating considerations incidental to the use of
each of the processes are discussed therein. Design shall anticipate operation over a wide range of temperature, pressure, and
reaction rate conditions and use adequate margins of safety in the design. Some of the safety considerations, and the sources of
hazards and their mitigation or control, are discussed in Appendix X3.
9.1.1 Chemical Reactivity—The dissolver and the dissolver offgas handling and treatment equipment shall be designed as a
complete entity, sized to handle the offgas load from the most reactive dissolution chemistry that can be predicted for the dissolver
design and potential fuel charges being considered. Typically, the offgas system capacity should be capable of accommodating
offgas surge rates or burps in the range of five to eight times the normal (production) processing rate over a one to three minute
time period. However, if the chemical reactivity is controlled through solvent (acid) availability, the offgas system should be
capable of accommodating offgas surge rates of 1.3 to 1.5 times the normal processing rate.
9.1.1.1 Nuclear fuel dissolution sequences have many similarities, but the sequential steps for any one process may not be fully
applicable to other nuclear fuel dissolution cycles.
9.1.1.2 The metal charge, in the form of chopped/sheared fuel pins one to three inches long, is frequently added in perforated metal
baskets. For these cases, the dissolver design may incorporate remotely operable provisions to raise the charge basket above the
solution level. This provides an alternative means of reaction rate control for emergency use in the event that the reaction rate
becomes excessive, to the extent that the offgas evolution rate threatens to overtax the capacity of the offgas treatment system.
9.1.2 Corrosion—A variety of chemicals can be used to dissolve particular fuels and residue sludges that may remain in the
dissolver at the conclusion of the dissolution cycle. The designer must anticipate these, select appropriate materials of construction,
and provide a corrosion allowance that tends to ensure contents confinement integrity over the design life of the vessel. Organic
acids and other chemicals used in decontamination sequences need consideration, and the corrosive effects of ions released during
the chemical dissolution cycle should also be considered. Accelerated corrosion tests on candidate materials of construction are
recommended.
9.1.3 Residues—The accumulation of metal fines and undissolved fission products as a sludge in the dissolver will require the
capability for flush-out and removal of this material to a sludge tank. Extended leaching and rinse operations are carried out in
order to reduce the fissile material content to specification levels prior to removal and disposal of the sludge as waste.
9.1.3.1 ZircaloyZirconium alloy fines and small pieces constitute a spontaneous fire hazard. ZircaloyZirconium alloy hulls that
have been fully stripped of heavy metal values are rinsed and passivated with a caustic solution. It is recommended that the
passivation step be carried out in an inert (argon) atmosphere to prevent fires.
9.1.4 Decay and Reaction Heat Control—It is recommended that the dissolver incorporate separate heating and cooling provisions
(for example, coils) to allow close control over dissolution solution temperatures and reaction rates during both the cladding and
the fuel dissolution steps, and to provide for temperature control in instances where exothermal reactions occur. Heat removal
capacity (coils or jacket heat transfer area or temperature differences) shall be sufficient to remove the radioactive decay heat load
as well as the reaction heat.
9.1.4.1 For those dissolvers employing heating jackets or coils, and where control of the final concentration of the dissolver
solution is important, the heat transfer area should be positioned somewhat above the bottom of the dissolver at a level that prevents
over-concentration (by boil-up) of fissile material solutions. Concentration, except for that which might occur as a result of
self-heating, would cease when heat transfer surfaces are no longer submerged.
9.1.4.2 Cooling coils or jackets should be positioned in processing vessels in such a way as to be fully submerged when vessels
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are filled to their normal operating levels. The heat transfer surface for cooling shall extend near to the bottom of the vessels in
order to provide the means for removal of decay heat from residual amounts of process solutions left in the vessels.
9.1.4.3 Dissolver steam and cooling water supplies should have temperature-activated interlocks. Settings of the interlocks should
be fixed at points that will prevent overheating and excessive boil-up of process solutions and at points that will automatically
introduce cooling water flow to cooling coils in the event that set points for the vessel temperature are exceeded.
9.2 Operating Considerations—Fuels in particulate form are highly reactive in acid solutions. It is recommended that dissolution
cycles anticipate the presence of significant quantities of fines. Assuming that such a condition exists, operators should start each
dissolution cycle with the use of dilute acid and chemical inhibitors that modify, and have a controlling effect on, the dissolution
reaction chemistry.
9.2.1 The administrative and technical practices for criticality safety and control should conform with or meet the intent of those
practices set forth in ANS 8.1.
10. Dissolver Vapors and Offgas
10.1 Design Considerations for Offgas Treatment—Dissolver offgases generally pass through several sequential treatment steps.
The offgas treatment requirements depend on the dissolution chemistry, the composition of the spent fuel being dissolved, the
gaseous and volatile radionuclides, other contaminants in the offgas stream, and other factors. Treatment of dissolver vapors and
offgas ensures that valuable process materials are recovered, and both radioactive materials and any noxious or undesirable
gas/vapor stream constituents are removed to the extent practicable or required. Treatment methods for the removal of any
particular offgas constituent may vary. Typical offgas treatment steps are briefly described in the following paragraphs. Mention
of a particular offgas treatment process is for purpose of illustration and does not constitute an endorsement of the procedure as
the best or only method for removal of contaminants from the offgas stream.
10.1.1 The treated offgas stream shall meet release criteria for toxic and radioactive contaminants as established by law and by
basic data specifications.
10.1.2 The offgas systems for dissolvers are generally designed to handle vapors or condensates, or both, that contain very low
concentrations of fissile materials and are generally not designed as a geometrically favorable system configuration. If foaming
were to be encountered or excessive entrainment were experienced, dissolver solution or fissile fines could be carried into the
offgas handling system. Special design provisions to prevent or to mitigate dissolver foaming conditions shall be considered. As
a minimum, dissolver system design should include provisions and operating procedures, or both, to return such carry-over
materials to the dissolver and to prevent their accumulation in the offgas system (for example, vapor and offgas decontamination
devices). Design provisions (for example, overflows, instrumentation, and alarms) and operating precautions shall prevent flooding
of the offgas handling system with dissolver solution.
10.1.3 Specific design features shall be considered to ensure an adequate offgas flow control capability during all phases of the
dissolver operation (for example, charging, dissolution, solutions transfer, reaction surges, standby, etc). A means of vacuum
regulation (such as a vacuum breaker) shall be included in the dissolver system design to avoid an excessive vacuum on the
dissolver, or one that could breach liquid seals or upset weight factor instrumentation.
10.1.4 Designs based on low air in-leakage rates to the dissolver offgas system should ensure that the low design basis rates can
be maintained during the entire life cycle for the facilities. The integrity and characteristics of closures design would be a prime
consideration here.
10.1.5 Design of the offgas system shall include a pressure relief system or component to limit the maximum dissolver system
pressure to 3 to 5 psig, or to the design pressure limits for the vent system. The relief system shall reset automatically.
10.1.6 Provisions should be included to permit periodic flushing of all offgas lines and equipment. Provisions to collect the flush
water, together with any accumulated solids or deposited fission products, or both, that are flushed out, are necessary as part of
the flush system.
10.2 Moisture and NO Removal—The removal of dusts, excess moisture and NO (oxides of nitrogen) gases may be affected by
x x
scrubbing, condensation, and adsorption techniques. Oxygen addition may be employed to enhance NO recovery. The offgas
x
scrubber step is intended to remove solid particulates carried off in the offgas stream and prevent the accumulation of these solids
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in the offgas equipment train. The design of the scrubber shall accommodate recovery and recycling of the solids and fines and
shall prevent a criticality incident that might potentially occur through inadvertent accumulation of fissile material fines.
10.2.1 The condenser section of the dissolver should be designed as a total reflux condenser, to return condensed liquids to the
dissolver, and to promote acid economy. Typically, gases are passed downwards through the condenser. The condenser capacity
should be sufficient to cope with peak boil-up and offgas loads without excessive pressure drop and consequent pressurization of
the dissolver assembly. The condenser should be equipped with an acid spray connection to permit wash-down and
decontamination of the coil assembly. The design of the condenser and scrubber should provide for reducing the temperature of
the offgas stream to the ambient cell or canyon temperature, or lower if practicable.
10.2.2 The removal of NO gases may require the inclusion of a multi-tray absorption column, or other NO removal methods
x x
such as the use of synthetic mordents in a packed column to catalyze selectively the ammonia reduction of NO gases.
x
10.2.3 An atomized steam-driven or pumped solution jet scrubber provides a means of solids removal, as well as means of cooling
the offgas stream and assisting in the removal of NO gases. The scrubber jet(s) may also serve as part of the vacuum system. When
x
such a treatment step is included in the offgas system, the motive system for maintaining a vacuum condition in the dissolver
should be backed by an installed spare (alternative) vacuum-producing component or system that will prevent over-pressurizing
the dissolver in the event of steam or pump failure.
10.3 Ruthenium (Ru) Removal—The use of a silica gel bed is one of a number of accepted and effective processes for the removal
of particulate or volatile Ru from the scrubbed offgas stream.
10.4 Iodine Removal—Silver-exchanged mordenite beds in series is one accepted and effective process for the removal of iodine.
The beds operate at a temperature of 150°C.150 °C. Silver-exchanged mordenite beds loaded with iodine are regenerated with
hydrogen. Iodine produced in the regeneration cycle is collected on lead-based absorption beds.
10.5 Krypton-85 Removal—One suggested process for the removal of Kr from an offgas stream features a selective absorption
step using refrigerant R-12 (dichlorodifluoromethane) as the absorption medium.
10.6 Tritium Removal—Tritium may be recovered by oxidation and sorption techniques. One process is based on the addition of
excess hydrogen to the offgas stream that then passes through a Ni-Cr-Pd ribbon catalyst bed to oxidize the hydrogen isotopes to
HTO. The unit operates at 400°C.400 °C. The HTO is then preferentially sorbed on molecular sieves (zeolite).
10.7 Carbon-14 Removal—Carbon-14 may be removed as CO gas by adsorption on zeolite molecular sieve beds. The CO gas
2 2
is driven off the sorbent bed during periodic regeneration cycles and is adsorbed on a BaOH bed.
11. Dissolver Product Handling
11.1 Design Considerations—Metal solution from the fuel dissolution step undergoes several treatments and measurements before
being chemically adjusted so as to be suitable for feed to solvent extraction. The dissolver solution may be processed through a
feed clarification step such as centrifugation. The accountability datum is established in a dedicated accountability tank. Nuclear
fuel dissolution systems design shall consider the potential for criticality in dissolver solutions hold and transfer tankage as well
as in the dissolver vessel (see Section 8). The use of a dilution eductor for solutions transfer from the dissolver may be included
by design as one means of minimizing chances of a criticality incident resulting from fissile materials solutions transf
...