DOE-STD-1026-2016 Reference Guide

NNSA Package Certification Engineer Qualification Standard DOE-STD-1026-2016 June 2016 Reference Guide The Functional Area Qualification Standard References Guides are developed to assist operators, maintenance personnel, and the technical staff in the acquisition of technical competence and qualification within the Technical Qualification Program (TQP). Please direct your questions or comments related to this document to Learning and Career Management, TQP Manager, NNSA Albuquerque Complex. This page is intentionally blank. i TABLE OF CONTENTS FIGURES ....................................................................................................................................... ii TABLES ......................................................................................................................................... ii VIDEOS ......................................................................................................................................... ii ACRONYMS ................................................................................................................................ iv PURPOSE ...................................................................................................................................... 1 SCOPE ........................................................................................................................................... 1 PREFACE ...................................................................................................................................... 1 TECHNICAL COMPETENCIES ............................................................................................... 3 1. NNSA package certification engineers must demonstrate a working level knowledge of DOE O 461.1B or latest revision, Packaging and Transfer or Transportation of Materials of National Security Interest..................................................................................... 3 2. NNSA package certification engineers must demonstrate a working level knowledge of DOE O 460.1C, Packaging and Transportation Safety. ........................................................... 5 3. NNSA package certification engineers must demonstrate a working level knowledge of DOE O 461.2, Onsite Packaging and Transfer of Materials of National Security Interest and DOE M 441.1-1, Nuclear Material Packaging Manual. ................................................... 8 4. NNSA package certification engineers must demonstrate a working level knowledge of the information required in chapter 1 of a SARP. .................................................................. 15 5. NNSA package certification engineers must demonstrate a working level knowledge of the structural evaluation information required in chapter 2 of a SARP. ................................. 28 6. NNSA package certification engineers must demonstrate a working level knowledge of the thermal evaluation information required in chapter 3 of a SARP. .................................... 47 7. NNSA package certification engineers must demonstrate a working level knowledge of the containment information required in chapter 4 of a SARP. .............................................. 55 8. NNSA package certification engineers must demonstrate a working level knowledge of the shielding evaluation information required in chapter 5 of a SARP. ................................. 70

9. NNSA package certification engineers must demonstrate a working level knowledge of the criticality evaluation information required in chapter 6 of a SARP. ................................ 75 10. NNSA package certification engineers must demonstrate a working level knowledge of the package operations information required in chapter 7 of a SARP. ................................... 89 11. NNSA package certification engineers must demonstrate a working level knowledge of the acceptance tests and maintenance program information required in chapter 8 of a SARP....................................................................................................................................... 94 12. NNSA package certification engineers must demonstrate a working level knowledge of the required QAP information required in chapter 9 of a SARP. ......................................... 107 13. NNSA package certification engineers must demonstrate a working level knowledge of the required elements and their evaluation within the probabilistic risk assessment section of a TSRA. ................................................................................................................ 113 14. NNSA package certification engineers must demonstrate a working level knowledge of the required format and content of a HAR............................................................................ 117 ii 15. NNSA package certification engineers must demonstrate a working level knowledge of the following guides and regulations: ................................................................................... 121 16. NNSA package certification engineers must demonstrate a working level knowledge of the process for certifying United Kingdom (UK) packages within the United States (U.S.). .................................................................................................................................... 122 17. NNSA package certification engineers must demonstrate a working level knowledge of the transportation requirements and the interface with the Office of Secure Transportation (OST). ........................................................................................................... 125 18. NNSA package certification engineers must demonstrate a familiarity level knowledge of the Price-Anderson Amendment Act of 1988 (PAAA) and its relationship to subparts A and B of 10 CFR 830. ....................................................................................................... 126 19. NNSA package certification engineers must demonstrate the technical writing and assessment/performance skills necessary to execute the responsibilities of a package certification engineer. ........................................................................................................... 133 20. NNSA package certification engineers must demonstrate a working level knowledge of assessment techniques and DOE O 226,1B, Implementation of Department of Energy Oversight Policy.................................................................................................................... 134 Selected Bibliography and Suggested Reading ...................................................................... 142

FIGURES Figure 1. Typical containment system components .......................................................................16 Figure 2. Sample package ..............................................................................................................32 Figure 3. Uranium enrichment .......................................................................................................69 Figure 4. Effect of nonhydrogenous diluent ..................................................................................83 Figure 5. Operational considerations in packaging design ............................................................93 Figure 6. Relationship of acceptance/maintenance programs to other disciplines ........................95 Figure 7. Typical drum-type package ............................................................................................96 TABLES Table 1. Types of labels .................................................................................................................22 Table 2. Separation of class 7 packages by rail .............................................................................23 Table 3. Separation between class 7 packages by highway ...........................................................24 Table 4. Allowable quantities of tritium ........................................................................................25 Table 5. Typical NNSA package content materials .......................................................................31 Table 6. Packaging materials and their functions ..........................................................................34 Table 7. Design categories .............................................................................................................36 Table 8. Summary of the fabrication criteria based on the ASME codea ......................................37 Table 9. Insolation data ..................................................................................................................43 Table 10. Criteria for free drop test (weight distance) ...................................................................43 Table 11. Accepted leakage test and sensitivities ..........................................................................59 VIDEOS Video 1. Decay heat .......................................................................................................................28 Video 2. Austenitic stainless steel .................................................................................................39 file://doe.local/dfs/NA/abq/qtd/WG/a_TQP%20Recast/Reference%20Guides/NNSA%20Package%20Certification%20Engineer/NNSA%20Package%20Certification%20Engineer%20%20RAM%205-6-15_AB_BD-T.docx#_Toc453576110 iii Video 3. Thermal insulation material ............................................................................................55 Video 4. Uranium enrichment........................................................................................................69 Video 5. Plutonium ........................................................................................................................70 Video 6. Risk assessment and job hazard analysis ......................................................................114 Video 7. Safe transport of radioactive materials ..........................................................................123 iv

ACRONYMS AC administrative control ADAST Assistant Deputy Administrator for Secure Transportation AEA Atomic Energy Act AEGL acute exposure guideline level AI anodic index ALARA as low as reasonably achievable ANSI American National Standards Institute ASME American Society of Mechanical Engineers ASTM American Society for Testing and Materials AWS American Welding Society BPVC boiler and pressure vessel code Bq becquerel CAS contractor assurance system CEDE committed effective dose equivalent Ci curie cm centimeter CO certifying official CoC certificate of compliance CRAD criteria review and approach document CSI criticality safety index CV containment vessel DID defense-in-depth DOE U.S. Department of Energy DOT U.S. Department of Transportation DSA documented safety analysis ERPG emergency response planning guideline FAE first article evaluation FAQS functional area qualification standard FN ferrite number ft feet gs inertial load factors h hour HA hazard analysis HAC hypothetical accident condition HAR hazards analysis report HAZ Heat-affected zone HCO headquarters certifying official HMR hazardous material regulation HS-40 HSS Office of Enforcement and Oversight HSS Office of Health, Safety, and Security IAEA International Atomic Energy Agency in. inch ISM integrated safety management KSA knowledge, skill, and ability v lb pound lbf pound force LDM low dispersible material MAWP maximum allowable working pressure MC&A material control and accountability MNOP maximum normal operating pressure MNSI materials of national security interest MOD (United Kingdom) Ministry of Defense mrem millirem MSLD mass spectrometer leak detector NCT normal condition of transport NNSA National Nuclear Security Administration Np neptunium NRC Nuclear Regulatory Commission NTS noncompliance tracking system OEM Office of Emergency Management OST Office of Secure Transportation OTA offsite transportation authorization OTC offsite transportation certificate OTD offsite transportation direction PAAA Price Anderson Amendment Act PCD Packaging Certification Division PRA probabilistic risk assessment psig pounds per square inch gauge PSO program secretarial officer Pu plutonium QA quality assurance QAP quality assurance program ref-cm3/s reference cubic centimeter per second SAC specific administrative control SARP safety analysis report for packaging SCAPA Subcommittee on Consequence Assessment and Protective Actions SER safety evaluation report SFRP shipment forecast and request procedure SG safety guide SNM special nuclear material SS safety-significant SSC structures, systems, and components STAAB Secure Transportation Asset Advisory Board STD standard STPSC Secure Transportation and Packaging Steering Committee TBq terabecquerel TEEL temporary emergency exposure limit TI transport index TQP Technical Qualification Program TSD transportation safety document vi TSR technical safety requirement TSRA transportation system risk assessment TSRP Transportation Safety Review Panel TSS Transportation Safeguards System U uranium UK United Kingdom U.S. United States USL upper subcritical limit USQ unreviewed safety question USQD unreviewed safety question determination 1

PURPOSE The purpose of this reference guide is to provide a document that contains the information required for a Department of Energy (DOE)/National Nuclear Security Administration (NNSA) technical employee to successfully complete the NNSA Package Certification Engineer Functional Area Qualification Standard (FAQS). Information essential to meeting the qualification requirements is provided; however, some competency statements require extensive knowledge or skill development. Reproducing all the required information for those statements in this document is not practical. In those instances, references are included to guide the candidate to additional resources. SCOPE This reference guide has been developed to address the competency statements in the April 2016 edition of DOE-Standard (STD)-1126-2016, NNSA Package Certification Engineer Functional Area Qualification Standard. The qualification standard for NNSA Package Certification Engineer contains 20 competency statements. PREFACE Competency statements and supporting knowledge and/or skill statements from the qualification standard are shown in contrasting bold type, while the corresponding information associated with each statement is provided below it. A comprehensive list of acronyms, abbreviations, and symbols is provided at the beginning of this document. It is recommended that the candidate review the list prior to proceeding with the competencies, as the acronyms, abbreviations, and symbols may not be further defined within the text unless special emphasis is required. The competencies and supporting knowledge, skill, and ability (KSA) statements are taken directly from the FAQS. Most corrections to spelling, punctuation, and grammar have been made without remark. Only significant corrections to errors in the technical content of the discussion text source material are identified. Editorial changes that do not affect the technical content (e.g., grammatical or spelling corrections, and changes to style) appear without remark. When they are needed for clarification, explanations are enclosed in brackets. Every effort has been made to provide the most current information and references available as of June 2016. However, the candidate is advised to verify the applicability of the information provided. It is recognized that some personnel may oversee facilities that utilize predecessor documents to those identified. In those cases, such documents should be included in local qualification standards via the TQP. In the cases where information about an FAQS topic in a competency or KSA statement is not available in the newest edition of a standard (consensus or industry), an older version is referenced. These references are noted in the text and in the bibliography. 2 This reference guide includes streaming videos to help bring the learning experience alive. To activate the video, click on any hyperlink under the video title. Note: Hyperlinks to videos are shown in entirety, due to current limitations of eReaders and must be accessed through Google Chrome. 3 TECHNICAL COMPETENCIES 1. NNSA package certification engineers must demonstrate a working level knowledge of DOE O 461.1B or latest revision, Packaging and Transfer or Transportation of Materials of National Security Interest. The information for the KSAs in this competency statement is taken from DOE O 461.1B, unless stated otherwise. a. Describe the purpose of the SARP, the HAR, and the TSRA.

Safety Analysis Report for Packaging (SARP) A SARP is a document that conforms to the Nuclear Regulatory Commission (NRC) Regulatory Guide 7.9, Standard Format and Content of Part 71 Applications for Approval of Packages for Radioactive Material, and provides a comprehensive technical evaluation of a package. The SARP consists of sections containing general information; structural, thermal, containment, shielding, and criticality evaluations; operating procedures; acceptance tests; and maintenance and quality assurance programs (QAPs). The purpose of the SARP is to demonstrate conformity with the applicable sections of 10 CFR 71, “Packaging and Transportation of Radioactive Material,” and 49 CFR 171–180. Hazards Analysis Report (HAR) The HAR is a document that is submitted to the NNSA certifying official (CO) to support an applicant’s request for an authorization for offsite transport of noncompliant packages and/or shipping configurations that contain less than a type B quantity of radioactive material or no radioactive materials, but which contain regulated hazardous materials as defined in 49 CFR 171–180. The HAR identifies the type and quantity of hazardous material, proposed packaging/handling gear, mode of transportation, shipment destinations, tie-down procedures, tests that will be performed on the unit, the post-test status of hazardous components, and procedures to verify that the shipment can be conducted safely (if applicable). Transportation System Risk Assessment (TSRA) The TSRA is a document that is submitted to the Assistant Deputy Administrator for Nuclear Safety and Operations to support an applicant’s request for an authorization for offsite transport of noncompliant packages and/or shipping configurations that contain a type A(F) or type B quantity of radioactive material and may contain regulated hazardous materials as defined in 49 CFR 171-180. b. Discuss the significance of the information required in a HAR. Contractor requests for nonfissile less than type B offsite transportation authorizations (OTAs) for noncompliant packages and/or shipping configurations containing hazardous materials and/or no more than type A quantities of radioactive material must be supported by a HAR. c. Discuss the information required in a TSRA. The TSRA records the hazards, the assessment of the hazards, the analysis methods, and analysis and results used to determine the frequency and consequences of events that could pose risks to 4 the workers, public, and/or the environment during the proposed offsite transportation of an uncertified package or a special assembly that contains a type A(F) or type B quantity of radioactive material in a prescribed transportation system and routes. d. Discuss what function the TSRP serves, how it is organized and staffed, and the basic steps in the review process. The Transportation Safety Review Panel (TSRP) is a committee chaired by a Federal employee and composed of persons with appropriate expertise that performs technical reviews to verify compliance with DOE O 461.1B and makes recommendations for offsite transportation certification or authorization. e. Discuss the purposes of the OTA, OTC, OTD and how they differ. Describe the information required in an OTA, OTC and OTD.

Offsite Transportation Authorization (OTA) The OTA is an NNSA CO approval that details the transportation configuration, authorized contents, regulatory and emergency response hazards, and transportation restrictions. The OTA is issued for nonfissile less than type B quantities of material. An OTC authorizes packages for shipment of radioactive materials within the TSS and for commercial carriers. An OTA may detail required positive measures, ACs, a declared maximum number of shipments per calendar year and a maximum number of units per trailer. An OTA may be issued for a one-time shipment or for a transportation campaign over a period of time not to exceed five years, at which time it must be re-authorized. Offsite Transportation Certificate (OTC) The OTC is an NNSA CO prepared document, analogous to an NRC or DOE headquarters CO certificate of compliance (CoC) that describes the compliant package configuration, authorized contents, and transportation restrictions. An OTC authorizes packages for shipment of radioactive materials within the transportation safeguards system (TSS) and for commercial carriers. An OTC can declare essential positive measures, administrative controls (ACs), and a maximum number of specified packages per transporter. Issuance of an OTC is demonstration of compliance with 10 CFR 71. It is issued for either a one-time use or multiple uses up to five years, at which point it must be renewed. Offsite Transportation Direction (OTD) The OTD is an NNSA approval that details the transportation configuration, authorized contents, regulatory and emergency response hazards, and transportation restrictions. The OTD is issued for type A(F) or type B quantities of material. An OTD always stipulates specific conditions of operations, including required transporter and shipment configuration. An OTD may detail required positive measures, ACs, a declared maximum number of shipments per calendar year and a maximum number of units per trailer. The OTD must state the national security purpose, the reason(s) why a compliant shipment cannot be made, and the reasons for determining that adequate safety has been achieved. An OTD may be issued for a one-time shipment or for a transportation campaign over a period of time not to exceed five years, at which time it must be re-authorized. 5 f. Explain the purpose of the SER and discuss the information required in the SER for HARs, SARPs, and TSRAs. The safety evaluation report (SER) is a document that provides the results of the TSPA’s safety evaluation, including its independent review of the HAR, SARP, and/or TSRA. The SER highlights whether or not the shipment of the packages and/or shipping configurations proposed in the HAR presents an undue risk to the health and safety of the public, workers, and/or the environment. The NNSA CO must document the TSRA review in a SER. The SER documents the analysis of a TSRA for consideration in granting an OTD. g. Discuss what is required for a contractor to be an authorized user of an NNSA type B package. Contractors must conduct type B and other fissile radioactive materials packaging and transportation activities in accordance with a quality assurance (QA) plan that follows the requirements of 10 CFR 71, Subpart H. The QA plan must be submitted through the contractor’s responsible DOE field organization to the NNSA CO for review and approval. The contractor is not authorized to begin packaging or transportation operations until it has been designated an authorized user of that package by the NNSA CO, and the NNSA CO will not designate the contractor an authorized user of the package until the contractor’s packaging procedures and QA plan have been approved.

2. NNSA package certification engineers must demonstrate a working level knowledge of DOE O 460.1C, Packaging and Transportation Safety. The following is taken from DOE O 460.1C. a. Discuss the significance of the offsite safety requirements. Each entity subject to DOE O 460.1C must perform packaging and transportation activities in accordance with the Department of Transportation (DOT) requirements of the Hazardous Materials Regulations (49 CFR 171–180). b. Describe the application process for NRC or DOT certified packagings. For a new NRC or DOT packaging certificate, each entity must file a request for a new certificate with the headquarters certifying official (HCO) or NNSA CO, as appropriate. When DOE or NNSA is the holder of a packaging certificate issued by the NRC or DOT, each entity must file a request for revisions to or renewal of the existing NRC or DOT certificate with the HCO or NNSA CO, as appropriate. In all cases the HCO or NNSA CO will review and forward, if appropriate, the request to the NRC or DOT. 6 c. Describe the application process for other type B or fissile materials certified packagings. For a new DOE or NNSA type B or fissile material packaging, each entity must submit an application to the HCO or NNSA CO, as appropriate, that includes a SARP and any other supporting documentation to demonstrate that the packaging meets the requirements of 10 CFR 71, Subparts E, F, G, and H, and any other applicable standards for certification prior to use. d. Discuss the QA requirements of this Order. Each entity that participates in the design, fabrication, procurement, use, or maintenance of a hazardous materials packaging must  have a QAP approved and audited by o the HCO or NNSA CO, as appropriate, for certified type B and fissile radioactive materials packagings satisfying the requirements of 10 CFR 71, Subpart H, “Quality Assurance,” or o the head of operations office or field office/site office manager, as appropriate, for all other radioactive and hazardous materials packagings, satisfying the requirements of DOE O 414.1D, Quality Assurance;  report deviations from the applicable requirements in compliance with DOE O 231.1B, Environment, Safety, and Health Reporting; and  additionally, report deviations to the HCO or the NNSA CO within 30 days in which there is o any instance in which there is significant reduction in the effectiveness of any approved type B or fissile packaging during use; o any discovery of any defects with safety significance in type B or fissile packaging after first use, with details of the means employed to repair the defects and prevent their recurrence; or o any instances in which the conditions of approval in the CoC were not observed in making a shipment. e. Discuss the significance of the onsite safety requirements. Onsite transfer of hazardous materials, substances, and wastes must be conducted in accordance with one of the following:  49 CFR 171–180 and the Federal Motor Carrier Safety Regulations (49 CFR 350–399)  A transportation safety document (TSD) approved by the head of operations office or field office/site office manager, as appropriate: o The TSD must describe the methodology and compliance process to meet equivalent safety for any deviation from 49 CFR 171–180 and 49 CFR 350–399. o For onsite transfers subject to 10 CFR 830, “Nuclear Safety Management,” the TSD

must comply with the safety basis requirements of 10 CFR 830, Appendix A to Subpart B, to identify the conditions, safe boundaries, and hazard controls necessary to protect workers, the public, and the environment from adverse consequences. o For multiple-tenant DOE/NNSA sites, safety documents for several contractor organizations may be combined into a single document. 7 o For onsite transfers not subject to 10 CFR 830, the TSDs must be approved and in effect no later than one year from incorporation of the contract requirements document of DOE O 460.1C into contracts. f. Discuss the DOT HMR in 49 CFR Parts 171–180. The contractor must perform offsite packaging and transportation activities in accordance with the DOT requirements of the hazardous materials regulations (HMR) (49 CFR 171–180). For specific radioactive material packagings for offsite shipments, the following apply:  Each contractor that offers for transportation or transports radioactive material in a type B or fissile material packaging, as appropriate, certified by the HCO, NNSA CO or the NRC, must meet the conditions specified in the CoC or OTC, as appropriate, for the package issued by the HCO, NNSA CO or NRC and register in writing with the HCO or the NNSA CO prior to use.  For an import or an export shipment pursuant to 49 CFR 173.471, 173.472 or 173.473, each contractor must use a packaging certified by the U.S. competent authority (DOT) where the DOE or NNSA and the contractor have been registered with the DOT as a user, and the contractor has the required documentation for the use and maintenance of the packaging and makes the shipments in accordance with the terms of the certificate issued by the DOT.  For a new NRC or DOT packaging certificate, the contractor must file a request for a new certificate with the responsible head of the operations office or the field office/site office manager for processing through the HCO or NNSA CO, as appropriate. When DOE or NNSA is the holder of a packaging certificate issued by the NRC or DOT, the contractor must file a request for revisions to or renewal of existing NRC or DOT certificate with the responsible head of the operations office or the field office/site office manager for processing through the HCO or NNSA CO.  For a new DOE or NNSA type B or fissile material packaging, each contractor must submit an application to the responsible head of the operations office or the field office/site office manager for processing through the HCO or NNSA CO. This application must include a SARP and any other supporting documentation to demonstrate the packaging meets the requirements of 10 CFR 71, Subparts E, F, G, and H, and any other applicable standards for certification prior to use. g. Describe the DOT special permit application process in 49 CFR 107 Subpart B. Any offsite hazardous materials packaging or shipment that is regulated by DOT and is not prepared in accordance with the HMR must be prepared in accordance with a valid DOT special permit. DOE applications for a DOT special permit must be submitted to the HCO to review, process, and forward to DOT. NNSA applications for a DOT special permit must be submitted to the NNSA CO to review, process, and forward to DOT. Applications must be prepared in accordance with the procedures in 49 CFR 107.105, “Application for Special Permit.” 8 3. NNSA package certification engineers must demonstrate a working level knowledge of DOE O 461.2, Onsite Packaging and Transfer of Materials of National Security Interest and DOE M 441.1-1, Nuclear Material Packaging Manual.

a. Discuss the purpose and scope of DOE O 461.2. Describe the following:  Packaging and transfer procedures  Transfer authorization  Review and approval process  Quality assurance  Transfer operations  Scheduling TSS transfers The following is taken from DOE O 461.2. Packaging and Transfer Procedures Each site must maintain a set of packaging and transfer procedures, approved by the appropriate authority. The responsible field organization managers must provide the oversight to ensure the implementation of written procedures and compliance with the packaging and transfer authorization basis. NNSA type A(F) or type B packages issued an OTC require packaging procedures to be approved by the NNSA Service Center. Non-NNSA approved type A(F) or type B packaging procedures shall be approved by the appropriate authority. Transfer procedures must be approved by the contractor appropriate authority. Non-type A or B packaging (e.g., IP-1, IP-II) procedures must be approved by the contractor appropriate authority. Transfer Authorization Compliant transfers of materials of national security interest (MNSI) do not require additional authorization. Noncompliant transfers must be authorized through a DOE-issued TSD SER. Compliant transfers of radioactive material—Compliant transfers of radioactive MNSI do not require special authorization:  Compliant packages must be used. Type A(F) or type B quantities of material must be packaged in a DOE, NNSA, or NRC certified package in accordance with the CoC or OTC.  Transfer motor vehicles must follow DOT regulations.  Personnel performing transfer operations must follow DOT regulations. Noncompliant transfers of radioactive material—Noncompliant transfers for all MNSI must be authorized by an approved safety basis:  The TSD establishes the approved safety envelope for packaging and transfer operations for MNSI. The TSD, when approved by a SER, satisfies the requirements of Attachment 3 of DOE O 461.2 and/or 10 CFR 830, Appendix A to Subpart B, Section F, “Documented Safety Analysis.” The SER presents the results of the DOE review team 9 and provides the framework for approval of the TSD giving the contractor authority to transfer these materials.  Noncompliant transfers that were evaluated as part of DOE-approved safety basis per 10 CFR 830, Appendix A to Subpart B, Section F, do not require a TSD. Review and Approval Process A DOE team reviews the TSD and makes recommendations to the responsible field organization manager for approving the application/documentation for noncompliant transfers. Quality Assurance Packaging and transfer activities for MNSI must comply with DOE O 414.1D. Each site must maintain a compliant QAP approved by the appropriate field organization manager, for the packaging and transfer of MNSI. Compliant transfers of type A(F) or type B quantities of radioactive and fissile material must be conducted in accordance with a QA plan that meets the requirements of 10 CFR 71, Subpart H. Transfer Operations Each compliant transfer must be prepared and transported in accordance with the applicable hazardous materials regulations 49 CFR 171–180. All transfer activities performed under the TSS must be conducted according to 10 CFR 830. Government or contractor vehicles must be operated in compliance with the applicable Federal motor carrier safety regulations 49 CFR 350–399. All noncompliant transfers must be performed under the purview of an approved TSD.

Transfer of nuclear explosives must meet the requirements of DOE O 452.1E, Nuclear Explosive and Weapon Surety Program, and DOE O 452.2E, Nuclear Explosive Safety. Scheduling TSS Transfers If TSS resources are required for onsite transfers, secure transportation shipping requirement forecasts must be developed for the Assistant Deputy Administrator for Secure Transportation (ADAST). Refer to DOE O 461.1C for the requirements and process to request and schedule TSS resources. b. Discuss the purpose and scope of DOE M 441.1-1. Describe the following:  Overview and responsibilities  Scope of materials  Nuclear material packaging requirements The following is taken from DOE M 441.1-1. Overview and Responsibilities DOE M 441.1-1, Nuclear Material Packaging Manual, provides detailed packaging requirements for protecting workers from internal exposure to nuclear materials stored outside of an approved engineered contamination barrier. The nuclear materials of concern in DOE M 441.1-1 are those whose composition and quantity create the potential for an airborne 10 contamination hazard that could result in a facility worker receiving an internal radiation dose in excess of five rem committed effective dose equivalent (CEDE). DOE M 441.1-1 does not address the shielding of nuclear materials that generate significant external direct radiation fields because those situations are addressed in other directives and rules. FIELD ELEMENT MANAGERS’ RESPONSIBILITIES Review and approve the following contractor products:  Technical basis for nuclear material packaging systems including the evaluation of the chemical, radiological, and physical characteristics of the stored nuclear material and the nuclear material storage packaging designs  Nuclear materials packaging surveillance programs  Process for documenting its nuclear materials storage program Ensure oversight of contractor’s implementation of the requirements of DOE M 441.1-1. Ensure a process exists for proposing, reviewing, approving, and implementing site corrective actions when necessary to ensure the requirements of DOE M 441.1-1 are met and to address conditions that are not protective of the public, workers, or the environment. Establish a risk-based prioritized schedule for implementing the nuclear material packaging requirements for material stored prior to the issuance of DOE M 441.1-1. Maintain records according to National Archives and Records Administration-approved DOE record schedules. Scope of Materials Field element managers must ensure that nuclear materials that are stored outside engineered contamination barriers are packaged in accordance with the requirements described in Chapter III of DOE M 441.1-1. The nuclear materials of concern are those whose composition and quantity create the potential for an airborne contamination hazard that could result in a facility worker receiving an internal radiation dose in excess of five rem CEDE. Nuclear materials of this type with a total quantity in a storage package exceeding the A2 thresholds established in 49 CFR 173.435, “Table of A1 and A2 Values for Radionuclides,” are subject to the specific container requirements of DOE M 441.1-1. A detailed description of the methodology for determining the type and quantities of materials is contained in Attachment 3 of DOE M 441.1-1.

Nuclear Material Packaging Requirements STORED MATERIAL CHARACTERISTICS Contractors must ensure the chemical, radiological, and physical characteristics of the stored material are evaluated for the lifetime of the storage of the material and are appropriate for the material package including the following:  Explosion sensitive and/or flammable materials must be evaluated to determine if safe storage can be achieved or stabilization is necessary.  Gas generation rates and gas composition must be evaluated and measures must be taken to minimize the formation or accumulation of gases inside the storage package with particular attention paid to limiting flammable gases. Minimization measures may 11 include placing limitations on contents of containers, stabilizing materials, or venting and filtering containers.  Incompatible materials whose interaction could lead to failure of a containment barrier must not be packaged together.  Physical and chemical form of material must be evaluated, including its corrosivity and potential for oxidative expansion to ensure proper packaging.  Moisture content must be evaluated to determine if safe storage can be achieved or if stabilization is necessary. Stabilization must be considered for materials that can absorb/adsorb significant quantities of moisture when in contact with air. Moisture content amount must not violate the requirements by interacting with materials being stored to induce corrosion or to increase the potential for flammable mixtures or overpressure.  Pyrophoricity must be evaluated to determine if safe storage can be achieved or if stabilization is necessary.  Radioactive decay heat must be evaluated to ensure no unacceptable thermal degradation occurs. Package designers must include a maximum heat load in the design specifications.  Radiation fields must be evaluated to ensure they will not cause failure of a containment barrier or other containers included in the storage package. Plastic should not be in direct contact with alpha-emitting materials. Package designers must include radiation fields resulting from radioactive decay of the stored material including daughter products that might have worse radiation fields than the originally stored material. Consideration should also be given to the radiation fields of contiguous containers in a storage array.  Solution composition and its effect on the selection of packaging type, storage duration, and surveillance must be evaluated. Special consideration must be given to gas generation and venting during storage life. PACKAGING DESIGN CRITERIA AND PERFORMANCE OBJECTIVES Contactors must ensure concerns with corrosion, radiolytic and thermal degradation, oxidative expansion, pressurization, incompatible materials, and usage (handling) that may result in container failure are addressed as part of the package design. Contactors must ensure nuclear material storage packages meet the following design criteria and performance objectives:  Corrosion resistance—The container, including filter vent and seals, must be constructed entirely of materials that are resistant to corrosion and chemical degradation from the materials being stored and generated gases as well as the ambient storage environment.

 Thermal degradation resistance—The container, including filter vent and seals, must be designed so that heat generation of the contents does not challenge the integrity of the container for the design life of the container.  Radiation resistance—The container, including filter vent and seals, must be designed so that radiation from stored material and the storage environment does not challenge the integrity of the container for the container’s design life.  Oxidative expansion accommodation—The container must be designed to accommodate the maximum volume increase and potential effects on the container from the oxidative expansion of stored material that may occur for the container’s design life.  Pressurization—Containers must be designed to withstand pressure due to gas generation, material oxidation, volume expansion, or release of volatiles during the design storage period. Measures should be taken to minimize the formation or accumulation of gases 12 inside the storage package. Such measures include placing limitations on the contents of containers and permitted length of storage time, or providing mechanisms in the storage package design, such as venting.  Incompatible and pyrophoric materials—Incompatible materials must not be stored together. Furthermore, pyrophoric materials must be containerized to prevent interaction with oxidative environments using techniques such as inert environments, double hermetic seals, encapsulation, etc. to prevent a pyrophoric reaction. Sites are encouraged to stabilize pyrophoric materials prior to storage.  Filter performance—Filters must be capable of venting the maximum credible gas flow rate. Filters must be compatible with the material being stored, including any product of reaction between the environment and the material.  Design life—The design life of the materials used in the container assembly as fabricated must be calculated or tested and the container must be examined under a regular documented maintenance and surveillance program. The impact of organic or other materials, such as silicone gaskets, that are subject to degradation must be accounted for in these calculations and in maintenance and surveillance plans. The container must meet design objectives such as the design release rate and design qualification release rate for the design life of the container. Container designs incorporating filters require useful-life determinations and performance verification measurements at the rated filtration efficiency.  Container closure—Storage containers must be securely closed in a way that precludes accidental opening or breaching during normal operational conditions and/or postulated drops.  Design release rate performance objective—The package must have a design release rate that will maintain the potential internal exposure to workers as low as reasonably achievable (ALARA) during normal storage or handling of the package.  Design qualification release rate performance objective—The package must have a post- drop design qualification release rate that will prevent the exposure of workers to greater than five rem CEDE. The drop test must be from the maximum working or storage height but not less than four feet.  Nondestructive examination—The package configuration should allow for nondestructive contents verification, inspection, and surveillance such as by radiography and weighing.  Quality assurance—Packages must be designed, tested, and procured in accordance with the QA requirements of 10 CFR 830 and DOE O 414.1D.

 Labeling—Each storage container or package of radioactive material must have a unique permanent identification marking or bear a durable, clearly visible label for identification and documentation purposes. Labeling must meet the requirements of 10 CFR 835.605, “Labeling Items and Containers.” Where practical and appropriate, combine radiological, criticality safety, and material control and accountability (MC&A) data into a single label. PACKAGING SURVEILLANCE PROGRAM Field elements must ensure a surveillance program is established and implemented to ensure the nuclear material storage package continues to meet its design criteria:  General—The surveillance program must evaluate appropriate attributes of stored packages to determine whether the packages can continue to be stored safely. Surveillance must also validate the package design life. 13  ALARA—The surveillance program must be structured to minimize the overall risk to the facility workers and maintain radiation doses ALARA during the performance of surveillance activities. At a minimum, the following concepts should be considered: o Coordinating surveillance activities with MC&A measurements and inspections o Coordinating surveillance activities with package movements o Using MC&A and routine radiological survey data in the surveillance program o Performing surveillance on packages as they are being opened for material use o Conducting remote surveillance (e.g., camera, load cell, etc.) for items generating high radiation fields  Objectives and techniques—The objective of the surveillance program is to identify indications of package degradation early to remediate the degraded package and to identify similar packages and materials that may need remediation. Surveillance techniques must be specified to provide early indications of package degradation which includes seal failure or loss of venting capability (if present). Any signs of degradation should be used for comparison with similar packages and containers. At a minimum, the following techniques must be considered: o Visual inspection of the container for indications of corrosion or pressurization, including examining the container for signs of degradation prior to routine handling of the container or opening for material use o Weight measurement for indications of mass change of the storage package that could indicate loss of package seal resulting in  possible oxidation and expansion of metal contents  absorption of moisture which could increase corrosion of the package o Contamination surveys which could indicate loss of container seal o Testing of vent filters for plugging o Radiography that could indicate pressurization or degradation of inner containers o Opening package for examination of interior contents and seals  Frequency—Surveillance frequencies must be specified for each package design and content combination. At a minimum, the frequency should be based on the potential failure mechanism, failure consequence, and package design life. Statistical sampling of packages for surveillance may be used for large, uniform populations of packages and/or contents.  Procedures—Surveillance procedures must specify the surveillance techniques to be used and must have defined acceptance criteria. Procedures must be established to address initial response to packages that fail the acceptance criteria.

 Evaluation—Surveillance data must be evaluated at least annually. Results of the surveillance program must be used to improve packaging design and modify the surveillance program, when appropriate. Sites should consider sharing data with other sites. DOCUMENTATION Field elements must ensure documentation of the nuclear material storage package design and surveillance is maintained:  Records retention—Packaging and surveillance records must be maintained in accordance with DOE O 243.1B, Records Management Program. 14  Content elements: o Material Information—Records must include, at a minimum, available information on the following material characteristics:  Description of the chemical and physical form  Best available isotopic content  Weight o Package information—At a minimum, the records must include identification of the following package characteristics:  Date of packaging for each container.  Baseline package gross weight.  Unique identification number associated with each package.  Configuration—quantity and type of containers in a package. This is not required for packages that existed before the manual was issued; however, it is a recommended good practice. o Surveillance information—At a minimum, the records must include the following surveillance and inspection data:  Unique identification number associated with each package  Surveillance and radioactive survey results and dates  Dates, location, and results of inspections  Name and site qualifications of individuals performing inspections o Technical basis information—At a minimum, records must include information pertaining to the technical basis for packaging and storing applicable nuclear materials, including a description of how the applicable stored material characteristics and packaging design criteria are satisfied, supported by test results and other evidence, as appropriate; threshold calculations for the packaged material; and surveillance and monitoring information.  Database: o A database of relevant information about stored materials and packages including material information, packaging information, and surveillance information for each storage package must be maintained. An electronic database should be used to simplify scheduling of surveillance and repackaging of packages and evaluation of surveillance data. The database may consist of several files, which in themselves may be databases, some of which may be classified. For completeness, the database should be coordinated and generally compatible with the MC&A databases. 15 4. NNSA package certification engineers must demonstrate a working level knowledge of the information required in chapter 1 of a SARP. a. Define and discuss the following items related to a packaging:  Containment features/boundaries  Neutron and gamma shielding features  Lifting and tie-down devices, impact limiters, and other structural features  Heat transfer features  Packaging tags, markings, and labels Containment Features/Boundaries The following is taken from NNSA Safety Guide (SG)-100, Chapter 4. The design of an NNSA package normally starts with the containment system. The containment system is defined as an assemblage of all the components required to retain the contents. In general, this includes the containment vessel (CV), seals, leak test and vent/fill port components, and closure bolts. The containment boundary is an assemblage of all the components required to retain the contents and is in direct communication with the internal cavity of the CV. Another way to interpret this is that any component whose failure could directly result in leakage of the contents is part of the containment boundary. Examples of containment boundary components are the CV, CV primary seal, and vent/fill port seal and plug.

The exact containment boundary should be defined, including the CV, welds, drain or fill ports, valves, seals, test ports, pressure relief devices, lids, cover plates, and other closure devices. Penetrations such as fill ports, valves, and test ports may be needed for operating purposes, such as backfilling for leak testing or for heat transfer enhancement. A CV usually has a bolted or other type of threaded closure to accommodate the loading and unloading of contents. The closure contains a seal or seals that minimize leakage from the CV to the environment. If there are multiple seals used for a single closure, the seal that is identified as the containment boundary seal should be clearly defined. CONTAINMENT VESSEL The function of all the CV and closure components is to maintain the containment boundary so that all normal condition of transport (NCT) and hypothetical accident conditions (HAC) containment requirements are met. The general standards for all packages listed in 10 CFR 71.43, “General Standard for all Packages,” also contain requirements that influence the structural design of the package containment. For most NNSA packages, the CV is designed as an American Society of Mechanical Engineers (ASME) boiler and pressure vessel code (BPVC) pressure vessel. In other words, the CV is designed to be capable of holding pressure. The CVs of NNSA packages are usually configured as a hollow, thin-walled cylindrical shell with one welded end and a bolted flange closure at the other end for loading and unloading the contents. (See figure 1.) The welded end closure of the CV is usually ellipsoidal, torispherical, hemispherical, or a flat head and welded to the CV shell by an ASME BPVC permissible weld joint. The bolted flange closure is usually a hub flange welded to the CV shell by an ASME BPVC permissible weld 16 joint and bolted to a blind flange, which acts as the closure lid. The material of construction of an NNSA package CV component is usually stainless steel manufactured to ASME BPVC standards for pressure-retaining components or American Society for Testing and Materials (ASTM) standards that are essentially identical to ASME BPVC standards. Source: NNSA SG-100 Figure 1. Typical containment system components The closures of a containment system of NNSA packages are usually bolted closures and are the weakest component of the package. The structural integrity and containment performance of the bolted closure depend upon the number, strength, and tightness of the closure bolts and are vital in maintaining containment and adequate seal compression. CLOSURE SYSTEM SEALS The closure system of an NNSA package CV usually consists of a closure lid bolted to a hub flange. Closure system sealing is normally provided by two seals installed in dovetail or 17 modified dovetail grooves on the underside of the closure lid. At assembly, the two seals are compressed between the polished machined surfaces on the hub flange and the closure lid grooves by installing and tightening the closure bolts to form the seal. In some designs, the closure lid is equipped with a vent/fill port which leads directly to the vessel cavity and a leak test port which leads to a small annulus between the two seals. The double-seal arrangement, coupled with a vent/fill and leak test ports, makes it possible to easily perform leakage tests of the inner seal. This is accomplished using various methods via a leak test port located between the two seals.

CLOSURE BOLTS The weakest link in the CV closure system is typically the closure bolt joint. The integrity of the containment system depends on the number, strength, and tightness of the closure bolts. However, the structural behavior of bolted joints is very complex due to the interactions with the closure lid, vessel wall, and, at times, the seals. In addition, the behavior of bolted joints varies depending upon the design and application. Existing national standards, such as the ASME BPVC for vessels, have significantly different design parameters and loads from those of an NNSA CV. The closure bolts of an NNSA package CV must be able to experience significant fire and impact loads, as specified in normal conditions of transport and hypothetical transport condition requirements that are not experienced by a pressure vessel, and still perform their intended function. Neutron and Gamma Shielding Features The following is taken from NNSA SG-100. For the majority of NNSA packages, structural parts of the CV are sufficient to provide shielding of short-range radiation such as alpha and beta, and usually are sufficient to reduce gamma and neutron radiation to acceptable levels. In cases where this is not true, shielding against gamma radiation can be achieved by using relatively dense materials such as lead, steel, or depleted uranium. At times, materials such as polyethylene or boron carbide are added to provide neutron shielding. In all cases, the design should ensure the mechanical configuration of any shielding materials is maintained. In addition, when lead is used as a shield material, lead slump and changes in the shielding configuration should be considered. This can happen at relatively low temperatures. To prevent decomposition, it is important to protect any plastic or other hydrogenous neutron shielding from high temperatures or from intense gamma radiation. Loads that can result in rupture or severe distortion of the CV also pose a threat to the shielding system. HAC stresses and concentrated loads may reduce the effectiveness of the shielding, even if the CV maintains containment. Vibration and differential thermal expansion may also reduce the effectiveness of the shielding if the mechanisms holding the shielding components in place are subject to fatigue. The package design should be evaluated to ensure the materials selected for the shielding are compatible under all conditions of transport. The materials selected should be such that there will be no significant chemical, galvanic, or other reaction among the package components and 18 contents. The primary concern during NCT and HAC is thermal degradation or decomposition, particularly when plastics or other low-temperature materials are used. While not normally encountered in NNSA packages, radiolysis of hydrocarbons is a concern at higher radiation levels. Decomposition can affect the shielding properties of the material and can also cause pressurization of the CV if gases are generated. The most significant concern is a case in which decomposition of materials in the shielding results in a loss of mechanical properties or a physical configuration change that may lead to a loss of shielding effectiveness. Lifting and Tie-Down Devices, Impact Limiters, and Other Structural Features The following is taken from NNSA SG-100.

TIE-DOWN DEVICES This section discusses the requirements for securing the packaging to a transport vehicle, including those for devices, inspection, and certification. Instructions may be included in the generic operating procedure or may be contained within a separate attachment or appendix. The following items may be included under this category:  Tie-down devices (including tie-down straps, chains, etc.) that secure the package to its transport vehicle or shipping skid or pallet  Inspection and certification requirements for the tie-down devices or shipping skids  Alternatively acceptable means to provide proper package tie-down Tie-down procedures used in the TSS conveyances are approved by the Office of Secure Transportation (OST). IMPACT LIMITERS AND OTHER STRUCTURAL FEATURES A number of empirical and classical calculational methods have been developed for scoping or preliminary determination of satisfactory or optimum performance of impact limiter or energy absorbing devices. In the past, the most common method used was a combination of testing, empirical, and classical methods. More recently, computer programs capable of modeling dynamic loadings on packages are being used to determine satisfactory or optimum performance of impact limiting materials and energy absorbing devices. In any method used, the most important parameter is the mechanical properties of the impact energy absorbing material under dynamic loading. The main mechanical property of interest is the force deflection or stress-strain behavior of the material under dynamic load conditions. In the past, such materials as Celotex and/or plywood, polyurethane foam, wood, and Kaolite have been used for NNSA package impact limiters. In all cases, impact energy absorption is based on the crushing of these materials. The mechanical property information developed for these materials ranges from static force deflection curves to dynamic stress strain data. In this section, some classical and empirical methods are discussed that have been used in design and material selection of package impact limiting devices. They are based on selecting the materials and design of impact limiters by calculating the loads transmitted to critical components. Also discussed are the key parameters for assessing impact limiter performance using computer programs, which numerically simulate impacts of the package design. 19 In the past, several classical and empirical methods have been used for designing crushable impact limiting devices. Most are based on idealizing the containment system as a rigid body surrounded by a crushable material that deforms on impact. These methods are based on the assumption that the natural frequency of the containment system is substantially higher than that of the impact absorbing system and dampening can be ignored. The deceleration inertial factor on the containment system is then conservatively determined by balancing the impact energy to work done or by using the impulse momentum theory combined with empirical test data on the behavior of the energy absorbing material. One empirical method, based on the impulse momentum theory for estimating maximum deceleration inertial load factors (gs) for free drops of packages protected by impact absorbing materials onto d surface, is provided in the Shock and Vibration Handbook. Based on field observations, knowing the drop height and assuming a sinusoidal deceleration pulse, the rise time (time to reach peak deceleration) and the peak deceleration can be determined and are dependent on the impact energy absorbing (or cushioning) material. The peak deceleration (G) in gs is determined based on the following equation:

where: h = the drop height in inches tr = the rise time in seconds The rise times given in table 41.1 of the Shock and Vibration Handbook are the approximated rise times from an assumed sinusoidal pulse for various packaging impact energy absorbing materials being dropped onto a rigid floor. The rise times provided are for metal containers, wooden boxes, cartons, and packages cushioned by 1-inch and 3-inch latex hairs. The reference also provides information and data on the material properties for impact absorbing materials. Unfortunately, the material data provided in this reference is limited by both type and the drop heights from which this data was obtained. For design purposes, this method provides the means for quickly and conservatively estimating impact limiting material performance if the rise time or deceleration time history data from a similar package impact limiter design using similar materials is known. The most common classical method is the energy balance method. This method essentially equates the kinetic energy of the falling object to the work done in crushing the impact limiting material. The work done on the impact limiter can be represented as the stress applied multiplied by the volume crushed by the material. Assuming a uniform crush strength of the material, the volume of deformation is determined. Based on the geometry and from the volume, the area of crush is calculated and multiplied by the assumed uniform crush strength of the material. The product is then divided by the gross weight of the package to determine the average deceleration inertial factor. In using this method, the strain on the material crush is usually limited to 60 percent to account for material lock-up. The limitation of this method is that the crush strength of the impact limiting material must be relatively constant and known, the lock-up strain must be either known or estimated, and the determined deceleration initial load factors is only the average deceleration of the system. 20 A variation on this energy balance method is used in specifying the impact limiting properties of polyurethane foam. This method is presented in the General Plastics product information for LAST-A-FOAM® FR-3700 impact limiting and insulating materials. In this iterative method, a theoretical deceleration distance must be determined to reduce the forces on the containment system components to some desired level. This is simply done by dividing the free drop height by the desired inertial deceleration factor to determine the theoretical deceleration distance. A theoretically perfect impact limiting material would decelerate a payload uniformly through 100 percent of its thickness. It should be noted that impact limiting materials typically have efficiencies of 25 to 50 percent of theoretically perfect foams. As a result, the actual deceleration levels would be 2 to 4 times the theoretically determined values. If this deceleration distance is adequate, the next step is to determine if the energy levels are within the energy absorbing range of the foam. This is done by determining the strain energy density required by the foam, which is determined by dividing the kinetic energy of the object free dropped by the volume of the foam. From this strain energy density, the required density range of the foam is determined by integrating the stress-strain curve for various foam densities and equating this to the required strain energy density that needs to be absorbed. In the case of the foam manufactured by General Plastics, the foam dynamic properties have been well characterized by testing. The data and methods for determining desired foam properties are available in General Plastics publication General Plastics LAST-A-FOAM® FR-3700 for Crash and Fire Protection of Nuclear Material Shipping Containers. This method can also be applied to other impact limiting materials as long as the dynamic stress-strain or force-deflection properties and the impact energy absorbing efficiencies of the material are known. As with the previous methods, the limitation of this method is that the crush strength of the impact limiting material must be relatively constant and known, lock-up strain must be either known or estimated, and the determined deceleration initial load factors are only the average deceleration of the system.

There are also several empirical methods for design of impact limiters using steel structures. These are usually in the form of fins, tubular members, and steel cage type construction. Also in some cases, the spent nuclear fuel casks have been constructed of concentric cylinders with lead shielding poured into the annulus. In this case, the deformation of the outer shell and lead can be used as impact limiters as long as regulatory shielding limits are maintained. Although impact limiting devices made from steel structures are used in many commercial and DOE environmental management applications, they are not typically used in NNSA packages. Heat Transfer Features The following is taken from NNSA SG-100. Specific heat is a measure of the amount of energy it takes to heat a specified quantity of a material a specified number of degrees. The units for this property can be expressed in kJ/(kg-K). The heat capacity of most materials increases with temperature. Density is the mass per volume of a material and units are kg/m3. For most materials, density decreases as temperature rises, although some materials may contract. Thermal conductivity, heat capacity, and density are needed to calculate non-steady state (transient) conduction heat transfer rates. 21 The only material property needed to calculate heat transfer rates due to radiation is emissivity (or absorbtivity). Emissivity is a measure of how efficiently a surface of a material emits radiant energy compared with an ideal radiator (blackbody). The emissivity of a material surface can vary greatly depending on the finish of the surface. For instance, the emissivity of stainless steel can vary from 0.17 (highly polished) to 0.90 (oxidized) depending on the finish of the surface. While selecting a material with either a low or high emissivity due to its surface finish may help protect a package, attaining, keeping, certifying, and documenting the specified surface finish on a package part may be difficult. Most applications of radiative heat transfer dealing with a package are for the package surroundings to the package exterior (or vice versa). However, if gaps exist within the package, heat transfer across these gaps will be by radiation and convection (unless the space is a vacuum, in which case heat transfer is only by radiation). Several properties must be known for the calculation of convection heat transfer coefficients, and they will always be fluid properties (i.e., either gas or liquid). The fluid properties needed to calculate the natural convection heat transfer coefficient are dependent on the correlations used, but generally the properties are fluid viscosity, density, volumetric thermal expansion coefficient, thermal conductivity, and heat capacity. For the vast majority of cases, the fluid will be air, although some other liquids, such as nitrogen and helium, are used in package design. Packaging Tags, Markings, and Labels The following is taken from the United States (U.S.) Nuclear Regulatory Commission, Labeling. Labeling Labels are used to visually indicate the type of hazard and the level of hazard contained in a package. Labels rely principally on symbols to indicate the hazard.

Although the package required for transporting radioactive material is based on the activity inside the package, the label required on the package is based on the radiation hazard outside the package. Radioactive material is the only hazardous material which has three possible labels, depending on the relative radiation levels external to the package. Also, labels for radioactive material are the only ones that require the shipper to write some information on the label. The information is a number called the transport index (TI), which, in reality, is the highest radiation level at 1 meter from the surface of the package. 22 The three labels are commonly called white I, yellow II, and yellow III, referring to the color of the label and the Roman numeral prominently displayed. A specific label is required if the surface radiation limit and the limit at 1 meter satisfy the requirements in table 1. Table 1. Types of labels Label Surface Radiation Level Radiation Level at 1 Meter White I Does not exceed 0.5 millirem (mrem) 1 hour (h) Not applicable Yellow II Does not exceed 50 mrem/h and Does not exceed 1 mrem/h Yellow III Exceeds 50 mrem/h or Exceeds 1 mrem/h Source: U.S. Nuclear Regulatory Commission, Labeling Because the TI is the radiation level at 1 meter, it is clear that a white I label has no TI. A yellow II must have a TI no greater than 1, and a yellow III may have a TI greater than 1. b. Define and discuss the following items related to radioactive material contents:  Maximum activity  Fissile material spatial separation and quantity limits  Neutron absorbers or moderators  Materials subject to chemical or galvanic reactions  Decay heat Maximum Activity According to The American Nuclear Society Tool Kit Glossary, maximum activity is the maximum rate of disintegration (transformation) or decay of radioactive material. The units of activity are the curie (Ci) and the becquerel (Bq). Fissile Material Spatial Separation and Quantity Limits The following is taken from 49 CFR 174.700. The number of packages of class 7 (radioactive) materials that may be transported by rail car or stored at any single location is limited to a total TI and a total criticality safety index (CSI) of not more than 50 each. Each package of class 7 (radioactive) material bearing radioactive yellow II or radioactive yellow III labels may not be placed closer than 3 feet to an area that may be continuously occupied by any passenger, rail employee, or shipment of one or more animals, nor closer than 15 feet to any package containing undeveloped film (if so marked). If more than one package of class 7 materials is present, the distance must be computed from table 2 on the basis of the total TI number (determined by adding together the TI numbers on the labels of the individual packages) of packages in the rail car or storage area. 23 Table 2. Separation of class 7 packages by rail Total Transport Index Minimum separation distance to nearest undeveloped film Minimum distance to area of persons or minimum distance from dividing partition of a combination car Meters Feet Meters Feet None 0 0 0 0 0.1 to 10.0 4.5 15 0.9 3 10.1 to 20.0 6.7 22 1.2 4 20.1 to 30.0 7.7 29 1.5 5 30.1 to 40.0 10 33 1.8 6 40.1 to 50.0 10.9 36 2.1 7 Note: The distance in this table must be measured from the nearest point on the nearest packages of class 7 materials. Source: 49 CFR 174.700 The following is taken from 49 CFR 177.842.

The number of packages of class 7 (radioactive) materials in any transport vehicle or in any single group in any storage location must be limited so that the total TI number does not exceed 50. The total TI of a group of packages and overpacks is determined by adding together the TI number on the labels on the individual packages and overpacks in the group. Packages of class 7 (radioactive) material bearing radioactive yellow II or radioactive yellow III labels may not be placed in a transport vehicle, storage location, or in any other place closer than the distances shown in table 3 to any area which may be continuously occupied by any passenger, employee, or animal, nor closer than the distances shown in the table to any package containing undeveloped film (if so marked), and must conform to the following conditions:  If more than one of these packages is present, the distance must be computed from table 3 on the basis of the total TI number determined by adding together the TI number on the labels on the individual packages and overpacks in the vehicle or storeroom.  Where more than one group of packages is present in any single storage location, a single group may not have a total TI greater than 50. Each group of packages must be handled and stowed not closer than 20 feet (measured edge to edge) to any other group. 24 Table 3. Separation between class 7 packages by highway Total Transport Index Minimum separation distance in meters (feet) to nearest undeveloped film in various times of transit Minimum distance in meters (feet) to area of persons, or minimum distance in meters (feet) from dividing partition of cargo compartments Up to 2 hours 2–4 hours 4–8 hours 8–12 hours Over 12 hours None 0.0 (0) 0.0 (0) 0.0 (0) 0.0 (0) 0.0 (0) 0.0 (0) 0.1 to 1.0 0.3 (1) 0.6 (2) 0.9 (3) 1.2 (4) 1.5 (5) 0.3 (1) 1.1 to 5.0 0.9 (3) 1.2 (4) 1.8 (6) 2.4 (8) 3.4 (11) 0.6 (2) 5.1 to 10.0 1.2 (4) 1.8 (6) 2.7 (9) 3.4 (11) 4.6 (15) 0.9 (3) 10.1 to 20.0 1.5 (5) 2.4 (8) 3.7 (12) 4.9 (16) 6.7 (22) 1.2 (4) 20.1 to 30 2.1 (7) 3.0 (10) 4.6 (15) 6.1 (20) 8.8 (29) 1.5 (5) 30.1 to 40 2.4 (8) 3.4 (11) 5.2 (17) 6.7 (22) 10.1 (33) 1.8 (6) 40.1 to 50.0 2.7 (9) 3.7 (12) 5.8 (19) 7.3 (24) 11.0 (36) 2.1 (7) Source: 49 CFR 177.842 QUANTITY LIMITS The following is taken from DOE-HDBK-1129-2008. Once an item has met the DOE and DOT requirements for shipment (e.g., properly packaged, radioactively surveyed, properly marked, properly completed shipping papers), the item inside the approved shipping package can be shipped to a new location. During shipment, the item inside the approved shipping package is expected to be subject to the normal activities associated with its movement from one location to another; for example, loading and unloading the package from vehicles, transport to a shipping area, storage in the shipping area prior to transport, loading and unloading the package onto trucks/trains/airplanes, and storage in the receiving area after arrival at the new destination. The packaging required by DOT and NRC regulations is designed to protect the workers, the public, and the environment from the radioactive material during normal package handling, transport, and shipping/receiving storage. The applicable requirements for various quantities of tritium for transportation and storage are roughly as follows:  Limited quantity—Limited quantities of tritium can be packaged and shipped in strong,

tight containers (paper boxes, paint cans) with proper markings.  Type A quantity (21.6 to < 1,100 Ci)—Type A quantities of tritium use DOT specification 7A containers, properly marked and surveyed prior to shipment. A number of different packages are available from small cans, 55- and 85-gallon drums, 4 x 4 x 7- foot steel boxes, up to and including oversized, specially designed containers. These containers are relatively inexpensive. 25  Type B quantity (> 1,100 Ci)—Type B quantities of tritium must be shipped in a certified DOT type B package. There is only a limited number of these expensive type B packages available for tritium shipment, and special routing (i.e., prescribed routes employing highway route control such as using major highways or bypassing cities) is required.  Type B quantity, low-level radioactive waste—For the purposes of storage at the waste site, type B quantity solid waste can be stored in type A containers. At the waste generation location, items containing greater than 1,100 Ci of tritium are normally stored in type A containers. These type A containers containing over 1,100 Ci must then be placed into type B containers for shipping to DOE waste sites. At the DOE waste site, the type A containers can be removed from the type B package and stored in the type A package. This allows the expensive type B package to be returned to the shipper and reused for another type B shipment. Table 4. Allowable quantities of tritium Form Shipping Package Type Maximum Quantity Specific Activity of Tritium per Package Comments All physical forms None • Activity concentration for exempt material of 10-6 TBq • Activity limit for exempt consignment of 10-3 TBq (0.027 Ci) Both conditions must be satisfied. Solid Limited quantity LSA-I 0.04 TBq (1.1 Ci) LSA-II 4 × 10-5 TBq/g (0.001 Ci/g) The conveyance limit for combustible solids is 4,000 TBq (1.1 × 105 Ci). LSA-III 0.08 TBq/g (2.16 Ci/g) SCO-I Limit based on surface contamination The maximum nonfixed contamination on accessible surfaces is 4 Bq/cm2. The maximum fixed contamination on accessible surfaces is 4 × 104 Bq/cm2. The total surface contamination on the inaccessible surfaces is limited to 104 Bq/cm2. The conveyance limit is 4,000 TBq. SCO-II Limit based on surface contamination The maximum nonfixed contamination on accessible surfaces is 400 Bq/cm2. The maximum fixed contamination on accessible surfaces is 8 × 108 Bq/cm2. The total surface contamination on the inaccessible surfaces is limited to 8 × 105 Bq/cm2. The conveyance limit is 4,000 TBq. Type A 40 TBq (1,100 Ci) Type B Limited by CoC for the package The type B UC-609 package is limited to 150 g of tritium. 26 Form Shipping Package Type Maximum Quantity Specific Activity of Tritium per Package Comments Liquid Limited quantity (tritiated water) 37 TBq (1,000 Ci) 3.7 TBq (100 Ci) 0.037 TBq (1 Ci) <0.0037 TBq (0.1Ci/L) 0.0037 to 0.037 TBq (0.1 to 1.0 Ci/L) >0.037 TBQ/L (1.0 Ci/L). Limited quantity (other liquids) 0.004 TBq (0.108 Ci) LSA-II 4 × 10-4 TBq/g (0.001 Ci/g) The conveyance limit is 4,000 TBq (1.1 × 105 Ci). Type A 40 TBq (1,100 Ci) Type B Limited by CoC for the package The type B UC-609 package is limited to 150 g of tritium. Gas Limited quantity 0.8 TBq (21.6 Ci/g) LSA-II 0.004 TBq/g (0.1 Ci/g) The conveyance limit is 4,000 TBq (1.1 × 105 Ci). Type A 40 TBq (1,100 Ci) Type B Limited by CoC for the package The type B UC-609 package is limited to 150 g of tritium.

Note: LSA (low specific activity) SCO (surface contaminated object) Source: DOE-HDBK-1129-2008 Neutron Absorbers or Moderators The following is taken from Wikipedia, Neutron Poisons. A neutron poison (also called a neutron absorber or a nuclear poison) is a substance with a large neutron absorption cross section, in applications such as nuclear reactors. In such applications, absorbing neutrons is normally an undesirable effect. However, neutron absorbing materials, also called poisons, are intentionally inserted into some types of reactors to lower the high reactivity of their initial fresh fuel load. Some of these poisons deplete as they absorb neutrons during reactor operation, while others remain relatively constant. The following is taken from NNSA SG-100. The presence of a neutron absorber or neutron absorbing material in the packaging will be revealed by the fabrication records. A test should be performed on the package to ensure the neutron absorbing material is functional. This may be accomplished by loading the packaging with a measured neutron emitting source and taking direct measurements. If no neutrons are detected, the absorber is still functional and the packaging may be used for shipment. A moderator is a material, typically light nuclei, used to reduce the kinetic energy of neutrons by scattering collisions without appreciable neutron capture. Materials Subject to Chemical or Galvanic Reactions CHEMICAL REACTIONS The following is taken from eHow, Types of Chemical Reactions of Metals. http://en.wikipedia.org/wiki/Neutron_absorber http://en.wikipedia.org/wiki/Cross_section_(physics) http://en.wikipedia.org/wiki/Nuclear_reactor http://en.wikipedia.org/wiki/Neutron http://en.wikipedia.org/wiki/Neutron_absorber 27 A chemical reaction is a process by which one type of chemical substance is converted to another type of chemical substance. All chemical reactions are characterized by a change in energy. Reactions either release or absorb energy from their surroundings as they proceed. In metals, there are four types of chemical reactions: synthesis reactions, decomposition reactions, single displacement reactions, and combustion reactions When two elements combine to form a compound, it is called a synthesis reaction. Basically, two reactants combine to form a product. Types of synthesis reaction include: combining a metal and a nonmetal to form a compound; reaction between a metallic oxide and water to form a hydroxide; and a metal reacting with oxygen to form metal oxide. Examples include sodium reacting with chlorine to form sodium chloride and aluminum reacting with fluorine to form aluminum fluoride. When a compound breaks into its elements or into simpler compounds, it is called a decomposition reaction. In this reaction, a reactant breaks into two or more products. Types of decomposition reaction include: metallic carbonates, when heated, produce metallic oxide and carbon dioxide; metallic chlorides break into a metallic element and chlorine; and metallic chlorates break into metallic chlorides and oxygen. GALVANIC REACTIONS The following is taken from eHow, Aluminum Galvanic Reaction. A galvanic reaction is the movement of ions from one type of metal to another, which results in the corrosion and consumption of the metal losing ions and the fortification against corrosion of the metal gaining ions.

When two metals come into contact, a galvanic reaction begins if three conditions are in place: the metals are electrochemically dissimilar; there is an electrically conductive path between the two metals; and there is a path by which metal ions can move. Water that contains salt, such as seawater, provides electrical conductivity as well as a path for metal ion migration. Metals with high anodic index (AI) values are more prone to galvanic corrosion than those with low AI values. Aluminum is considered a relatively reactive metal, with an AI value between 0.75 V and 0.95 V, depending on the composition of the aluminum alloy. To prevent galvanic corrosion of aluminum when in proximity with a lower AI metal, such as corrosion-resistant steel (AI 0.50 V), either the metals must be treated with a nonreactive coating, such as paint, or a sacrificial metal such as magnesium (AI 1.75 V) must be attached to the aluminum. Decay Heat The following is taken from Massachusetts Institute of Technology, Nuclear Science and Engineering, Explanation of Nuclear Reactor Decay Heat. Nuclear reactors produce electricity in a similar way to conventional coal plants in that they heat steam to drive a turbine that spins an electric generator. However, they differ on how that heat is produced. Coal plants burn coal to heat a boiler that produces the steam while nuclear reactors use nuclear fission to create the heat. 28 The heat in an operating reactor is produced mainly by the fission of fissile isotopes such as uranium (U)-235 and plutonium (Pu)-239. When a neutron causes one of these isotopes to split, a large amount of energy is released, which is then deposited in the fuel, cladding, coolant, and structures. On average, approximately eighty percent of the energy released in a fission reaction is imparted to the two or more fission products and these deposit their energy in the fuel, since they have a very short range. The rest of the energy is released in the form of neutrons, and other forms of radiation. When there is a SCRAM (an emergency shutdown of a nuclear reactor), where all the control rods are inserted and the reactor is shutdown, the fission reactions essentially stop and the power drops drastically to about seven percent of full power in one second. The power does not drop to zero because of the radioactive isotopes that remain from the prior fissioning of the fuel. These radioactive isotopes, also called fission products, continue to produce various types of radiation as they decay, such as gamma rays, beta particles, and alpha particles. The decay radiation then deposits most of its energy in the fuel, and this is what is referred to as decay heat. As these radioactive isotopes continue to decay, more and more of them reach a stable state and stop emitting radiation, and thus no longer contribute to the decay heat. The decay heat must be removed at the same rate it is produced or the reactor core will begin to heat up. The removal of this heat is the function of the various reactor core cooling systems that provide water flow through the reactor core and then reject the heat elsewhere. Video 1. Decay heat https://www.youtube.com/watch?v=Wb1JFzBgSl0 5. NNSA package certification engineers must demonstrate a working level knowledge of the structural evaluation information required in chapter 2 of a SARP. a. Discuss the following:  NNSA SG-100 as it pertains to structural design and development: 1) mechanical

design, 2) material selection, 3) fabrication (including welding), 4) examination, 5) testing, 6) SARP preparation, and 7) certification.  Determination of the design criteria for package design per NRC Regulatory Guide 7.11, Fracture Toughness Criteria of Base Material for Ferritic Steel Shipping Cask Containment Vessels with a Maximum Wall Thickness of 4 Inches (0.1 m)  NUREG/CR-3854, Fabrication Criteria for Shipping Containers  The purpose and scope of NRC Regulatory Guide 7.6, Design Criteria for the Structural Analysis of Shipping Cask Containment Vessels  The purpose and scope of NRC Regulatory Guide 7.8, Load Combinations for the Structural Analysis of Shipping Casks for Radioactive Material  Material properties and specifications  The application of austenitic stainless steel as opposed to ferritic steels in the design and construction of small type B packages  Chemical, galvanic, and other reactions, and the application of corrosion- inhibiting coatings  General requirements for all packages per 10 CFR 71.43  10 CFR 71 requirements for lifting and tie-down devices https://www.youtube.com/watch?v=Wb1JFzBgSl0 29 NNSA Safety Guide 100 STRUCTURAL DESIGN NRC Regulatory Guide 7.6 was developed in the late 1970s specifically for spent nuclear fuel casks. It is based on the 1977 version of the ASME BPVC, Section III, Division 1, Subsection NB for Class 1 components. Therefore, much of the guidance in NRC Regulatory Guide 7.6 is based on the “design by analysis” and stress categorization methods outlined in Article NB-3000 for Class 1 components. At that time, the predominate method of analysis was classical linear elastic methods. NRC Regulatory Guide 7.6 has not changed since its issuance in 1978, while the ASME BPVC and analytical methods have evolved over time. As a result, the NNSA guidance in Section 2.5 suggests that the current version of the ASME BPVC, Section III, Division 1, Subsection NB be used, along with the appropriate requirements from NRC Regulatory Guide 7.6 that are specific to packaging. The basis for this is that the packaging performance tests and the failure modes addressed by NRC Regulatory Guide 7.6 have not changed substantially. The appropriate criteria that should be considered are those outlined in NRC Regulatory Guide 7.6 regulatory positions 3, 4, 6, and 7. It should be recognized that these criteria are based on linear elastic methods and should be applied only to the containment system. Other appropriate nonlinear methods can be applied to other components, such as the impact limiters. Also, as the ASME BPVC design by analysis method is based on the maximum shear stress theory of failure, the stress intensity, which is the difference between the maximum and minimum principal stresses on the section being evaluated, must be evaluated. In using the guidance provided in Section 2.5 and NRC Regulatory Guide 7.6, it should be recognized that HAC performance test evaluations may use the criteria specified in ASME BPVC, Section III, Division 1, Appendices, Appendix F for Category I content package containment systems. Appendix F provides acceptance criteria for elastic system analysis and plastic system analysis in Section F-1330 and Section F-1340, respectively. Both sections define acceptance criteria for level D service limits for components designed in accordance with Subsection NB. Section F-1341 allows the use of any one of the following structural acceptance criteria for demonstrating the acceptability of components:  Elastic analysis (F-1341.1)  Plastic analysis (F-1341.2)  Collapse load analysis (F-1341.3)  Plastic instability analysis (F-1341.4)  Interaction analysis (F-1341.5)

MECHANICAL DESIGN NRC Regulatory Guide 7.6 describes design criteria acceptable to the NRC staff for use in the structural analysis of spent nuclear fuel cask CVs. These criteria are used to demonstrate the CV meets the performance tests of 10 CFR 71.71, “Normal Conditions of Transport,” and 10 CFR 71.73, “Hypothetical Accident Conditions.” NRC Regulatory Guide 7.6 was developed because there were no standards for demonstrating satisfactory performance of shipping cask CVs at the time. Although originally written for spent nuclear fuel cask, NRC Regulatory Guide 7.6 has traditionally been applied as the acceptable performance criteria for structural analysis of all type B package CVs with category I contents. The design criteria established in NRC Regulatory Guide 7.6 are only applicable to the structural evaluation of the CV of a package. 30 NRC Regulatory Guide 7.6 has adopted portions of the ASME BPVC that use the design by analysis approach for class 1 reactor components to form acceptable design criteria for shipping cask CVs. As with the 1977 ASME BPVC, NRC Regulatory Guide 7.6 is based on linear elastic methods, stress categorization, and the principle of superposition for determining the effect of combined loads on CVs. The authors of NRC Regulatory Guide 7.6 recognized that ASME BPVC, Section III contains requirements for the design of nuclear power plant components, so many of its failure modes are not considered applicable to shipping cask CV design. However, NRC Regulatory Guide 7.6 provides design criteria extracted from the 1977 ASME BPVC that are considered acceptable to the NRC for assessing the adequacy of CV designs. These requirements are stated as regulatory positions in NRC Regulatory Guide 7.6. Because NRC Regulatory Guide 7.6 is based on ASME BPVC design by analysis methods, the definitions and classification of stresses are essentially the same as the 1977 ASME BPVC, Section III, Division 1, Subsection NB. In assessing the performance of a package under the performance tests specified in 10 CFR 71.71 for NCT and 10 CFR 71.73 for HAC, NRC Regulatory Guide 7.6 applies the ASME BPVC design by analysis approach. NRC Regulatory Guide 7.6 has adapted the service loading limits for reactor components to packaging performance test conditions. In NRC Regulatory Guide 7.6, NCT service loading limits are defined as level A service limits and HAC service loading limits are defined as level D service limits, as amended by NRC Regulatory Guide 7.6 for shipping cask CVs. In applying these limits, NRC Regulatory Guide 7.6 applies the ASME BPVC methods for determination of satisfactory performance of the CV. The ASME BPVC method applies the Tresca-Guest or maximum shear stress theory of failure by defining the term “stress intensity” as twice the maximum shear stress and equal to the difference between the algebraically largest and smallest principal stresses at any given point. These stress intensities are then categorized into three stress categories: primary, secondary, and peak. In addition, the primary stress intensity has three sub-categories: general membrane, local membrane, and bending. The stress is divided into various categories because the limit analysis method used in the ASME BPVC indicates that some stresses may be permitted to go to higher values. Packaging materials are selected based on the package content material and the package safety functions required to transport this material. Materials typically packaged in NNSA packages are given in table 5.

31 Table 5. Typical NNSA package content materials Material Form Source of Nuclear and Physical Properties 233U, 234U, 235U, 236U, 238U Solid (parts, powder, chips, etc.): pure metal, oxides, nitrides, other compounds SCALE: A Modular Code for Performing Standardized Computer Analyses for Licensing Evaluation, NUREG/CR-0200, Rev.7, 2004 238Pu, 239Pu, 240Pu Solid (parts, powder, chips, etc.): pure metal, oxides, nitrides, other compounds SCALE: A Modular Code for Performing Standardized computer Analyses for Licensing Evaluation, NUREG/CR-0200,Rev.7, 2004 Tritium Solid, liquid, or gas (mixtures and compounds) Health Physics Manual of Good Practices for Tritium Facilities, MLM-3719, Draft EG&G Mound Applied Technologies, December 1991 Source: NNSA SG-100 The materials used in NNSA packages must enable the package to satisfy the following safety functions: contain the radioactive material; provide shielding from radiation; and maintain subcriticality. The packaging must also keep the contents in place (confinement). These functions, in part, are accomplished by designing structural integrity into the packaging as well as designing the package to accommodate the thermal loads which may be experienced by the packaging. The following paragraphs discuss the material characteristics needed to accomplish each of these functions. A sample packaging showing the function(s) of each of the components and possible materials for those components is shown in figure 2. Containment The containment boundary must prevent the release of radioactive materials during NCT and HACs as stipulated in the regulations. The walls, heads, welds, fasteners, and seals that make up the containment boundary must be capable of withstanding the mechanical, thermal, and physical conditions to which they may be subjected. Shielding The radiation dose rate must not exceed regulatory limits during use. For NCT, the regulations stipulate dose rates at the surface of the package and at 1m from the package surface; whereas for HAC only dose rates at 1 meter from the package surface are specified. If the packaging is deformed during HAC, consideration must be given to the possibility that the radiation source may be nearer to the surface of the packaging and, consequently, the radiation dose rate at 1 meter may be higher after deformation. In some cases, packaging design features that satisfy functions other than shielding also provide some shielding. In other cases, the materials used or the thicknesses of the materials used will have to be changed to provide adequate shielding. In 32 addition to transportation, the packaging may be required to meet more stringent facility dose rates. These lower criteria must be considered during the material and fabrication selection process. Source: NNSA SG-100 Figure 2. Sample package Subcriticality Subcriticality must be maintained during all NCT and HAC. In many cases, subcriticality is dependent on keeping moderating material out of the containment boundary, and on maintaining adequate spacing between multiple containers. With the advent of the crush test, possible reduced spacing of post-HAC units must also be considered. Structural Integrity Structural integrity is often supplied by impact limiters. The components of the packaging that act as an impact limiter are usually those components at or near the outer surface of the 33 packaging. The impact limiters absorb energy from the HAC drops and crush tests such that the other functions of the packaging are preserved.

Thermal The characteristics of the materials used must accommodate the thermal loads that may be applied to the package. Thermal loads include radioactive decay heat from the contents, heating from the sun (insolation), and heat from a HAC thermal event. The environment can also affect the thermal load on the package. Various combinations of thermal loading can occur during the shipment of a package. The thermal properties of the materials used and the configuration of those materials must be such that containment, shielding, and subcriticality are maintained. Confinement The components of the packaging that provide confinement must hold the package together during HAC to the extent that the other functions of the packaging are maintained. Usually, the confinement components are those that make up the outer surface of the packaging. In the sample packaging (figure 2), the confinement boundary consists of the stainless steel drum and drum lid. After the drop test and crush test portions of the HAC, the extent of the deformation and opening of the drum must be such that allowable radiation doses are not exceeded and containment is maintained during the fire test. Packing Material Packing material may be used inside the CV and holds the contents in place. It may also enhance the structural integrity and thermal capabilities of the package, as well as provide confinement. No single material can accomplish all of these functions; however, some materials contribute to more than one function. Materials should also be durable, easy to work with, and cost-effective. Materials that have been successfully used for various functions and components are listed in table 6. Materials other than these may be used in NNSA packagings; however, justification for their use must be documented. 34 Table 6. Packaging materials and their functions Material Function Containment Shielding Subcriticality Structural Integrity Thermal Confinement Packing Material Austenitic stainless steel X x x x x X Carbon steel X X X X High strength steel X X X Aluminum alloy 6061 X X X Cellulosic insulating board X X Inorganic refractory X X Ceramic fiber materials X X Fir plywood X X Redwood X X Urethane foam X X Silicone foam X Ethylene propylene X Source: NNSA SG-100 FABRICATION Fabrication generally addresses forming, fitting, and aligning; welding and brazing; heat treatment; and mechanical joints. If the fabrication specifications are prescribed by an acceptable code or standard, the code or standard should be identified on the engineering drawings. Unless otherwise indicated, specifications of the same code or standard used for design should also be used for fabrication. For components that have no fabrication codes or standards that are applicable, the specifications on which the evaluation depends should be described, and the method of control used to assure that the specifications are met should be identified. This information should be indicated on the engineering drawings. EXAMINATION Examination addresses the methods and criteria by which fabrication is determined to be acceptable. Unless it is justified otherwise, specifications of the same code or standard used for fabrication should also be used for examination. For components that have no fabrication codes or standards that are applicable, the examination method and acceptance criteria used should be described. 35

TESTING In general, the requirements for pressure testing of components in Article 6000 are the same in Subsection NB and ND. Essentially when the construction is complete the component is closed and tested hydrostatically at a pressure of 1.25 times the design pressure. In the past this test was considered a structural integrity test even though it did not subject the vessel to conditions as severe as it might be expected to experience in service. The test is normally at atmospheric temperature and the only loads applied are the pressure loading and the weight of the test fluid. In cases where hydrostatic testing is not appropriate, pneumatic testing may be used. However, pneumatic pressure tests are specified to be 1.1 times the design pressure. No pressure testing is required for components constructed to Subsections NF and NG. In the case of vessels constructed to the requirements, the requirements for pressure testing of components is a hydrostatic test at 1.3 times the maximum allowable working pressure (MAWP). For packaging purposes, MAWP is the same as the design pressure. As with Section III, Division 1, pneumatic testing is permitted where hydrostatic testing is not appropriate. However, pneumatic pressure tests are specified to be 1.1 times the MAWP. SARP PREPARATION The culmination of the design and verification phase will lead to the preparation of the SARP. The SARP should clearly demonstrate that all the requirements have been met with a clear margin of safety to protect the public, the workers, and the environment. The information in the SARP is unique to the packaging and content. A suggested format for a SARP has been provided by the NRC in Regulatory Guide 7.9. Revision 2 of NRC Regulatory Guide 7.9, was issued in March of 2005. NNSA SG-500 suggests the use of NRC Regulatory Guide 7.9 for SARP development. Appendix B of NNSA SG-500 provides additional guidance on the SARP development process. NRC Regulatory Guide 7.9 has been developed from years of experience to standardize the SARP preparation and review process. While adherence to NRC Regulatory Guide 7.9 is not a requirement, it is highly recommended that it be followed closely. By using NRC Regulatory Guide 7.9, the SARP will be more easily and quickly reviewed. If applicants deviate from the suggested format and/or content, it is recommended that such deviations be well-documented in the SARP to aid the reviewer. SARPs submitted to the DOE and NNSA, unlike those submitted to the NRC, must contain a section on QA. This is a deviation from the format outlined in NRC Regulatory Guide 7.9. CERTIFICATION Packages are usually certified for five years, and applications must be made to recertify packages when the certificate is due to expire. The application will include a revised SARP, supplemental report, and completed NNSA SG-200 checklist, and any other supporting documents that should be submitted no later than 26 work weeks prior to the expiration of the certificate. Additional time may be required for the recertification process if significant comments are generated during the review cycle or there are delays in transmitting sensitive or classified documents. 36

Numerous issues may impact the amount of time and effort needed for the recertification effort. Packages requiring design changes will require more effort and time than packages that simply need recertification. A listing of all packaging serial numbers that were used under the existing or previous certificate and their refurbishment history should be available for review by the CO. If the refurbishment history indicates an excessive amount of repair/refurbishment of packaging components, replacement of these components with more robust components may be warranted. In addition, the level of the change to the safety class will also affect the level of change to the SARP. If new or revised rules and regulations become effective, between the time the package was certified/recertified and this recertification, the changes to the rules and/or regulations must be properly considered, as the packaging may not meet the new requirements. The extent of the SARP revision will depend on how the new regulations are to be implemented, the extent of the rule changes, and how they will impact the packaging to be recertified. Determination of the Design Criteria The following is taken from U.S. Nuclear Regulatory Commission, NRC Regulatory Guide 7.11. Table 7. Design categories Levels of safety Category I Category II Category III Low specific activity Greater than 30,000 Ci or greater than 3,000 A1 or greater than 3,000 A2 Less than 30,000 Ci and less than 3,000A1 and less than 3,000 A2 Special form Greater than 3,000 A1 or greater than 30,000 Ci Between 3,000 A1 and 30 A1 and not greater than 3,000 Ci Less than 30 A1 and less than 30,000 Ci Normal form Greater than 3,000 A2 or greater than 30,000 Ci Between 3,000 A2 and 30 A2 and not greater than 30,000 Ci Less than 30 A2 and less than 30,000 Ci Source: U.S. Nuclear Regulatory Commission, Regulatory Guide 7.11 NUREG/CR-1815, Recommendations for Protecting Against Failure by Brittle Fracture in Ferritic Steel Shipping Containers up to Four Inches Thick, contains background and other information pertinent to the development of the criteria in this guide. These criteria are divided into three categories that are associated with the levels of safety appropriate for the radioactive contents being transported. Table 7 identifies the radioactivity limits for each of the three categories. The criteria contained in Section 5 of NUREG/CR-1815, other than for full-scale destructive testing and qualifying procedures for reduced stress levels, are acceptable to the NRC staff for assessing the fracture toughness of thin-wall base material ferritic steel CVs for the categories identified in table 7. A category I container qualified in accordance with NRC Regulatory Guide 7.11 is acceptable for transporting either category II or category III radioactive materials. Similarly, a category II container qualified in accordance with the guide is acceptable for transporting category III materials. 37 Except in those cases in which an applicant or licensee proposes an acceptable alternative method for complying with specified portions of the NRC’s regulations, the methods described in NRC Regulatory Guide 7.11 (which reflects public comments) will be used by the NRC staff in evaluating base material for all applications for new package designs and all requests that existing package designs be designated as type B(U) or type B(M) packages. NUREG/CR-3854, Fabrication Criteria for Shipping Containers The following is taken from U.S. Nuclear Regulatory Commission, NUREG/CR-3854.

NUREG/CR-3854 provides fabrication criteria for the metal components of shipping containers used for transporting radioactive materials. The criteria are divided into three categories that are associated with the levels of safety for the types and quantities of radioactive materials being transported. For each category, the fabrication criteria are subdivided into three component safety groups that are formed according to their safety function. The categories and component safety group designations are the same as those used in developing the welding criteria. The fabrication criteria are based on the ASME BPVC as summarized in table 8 for each of the categories and component safety groups. Section 4.0 of NUREG/CR-3854 provides the detailed fabrication criteria, including any exceptions or modifications to the ASME code. The selected ASME code criteria provide levels of confidence in controlling fabrication processes consistent with the categories and component safety groups. The criteria should be used with the welding criteria when fabricating shipping containers for transporting radioactive materials. Table 8. Summary of the fabrication criteria based on the ASME codea Component safety groupb Container contents Category I Category II Category III Containment Section III Subsection NB Section III Subsection ND Section VIII Division 1 Criticality Section III, Subsection NG Other safety Section VIII, Division 1 or Section III, Subsection NF a See Section 4.0 of NUREG/CR-3854 for detailed criteria for each category, component safety group, and fabrication process. b DOT specifications 17C or 17H are acceptable for the fabrication of drums and pails used as shipping containers. Source: NUREG/CR-3854 The criteria for fabricating metal component of shipping containers used for transporting radioactive materials are based on the ASME code as summarized in table 8. An acceptable method of assuring compliance with the criteria is to have the construction of a shipping container carried out by a fabricator having a valid certificate of authorization for the use of the ASME code stamp for the appropriate section. It is not intended that the ASME code stamp be applied to the shipping container. A fabricator having a Section III subsection NB, ND, or NF certificate of authorization is considered to be qualified for Section VIII fabrication and would not require a Section VIII certificate of authorization. 38 Purpose and Scope of NRC Regulatory Guide 7.6 10 CFR 71.35, “Package Evaluation,” requires that packages used to transport radioactive materials meet the normal and HACs of Appendices A and B, respectively, to 10 CFR 71. This guide describes design criteria acceptable to the NRC staff for use in the structural analysis of the CVs of type B packages used to transport irradiated nuclear fuel. Alternative design criteria may be used if judged acceptable by the NRC staff in meeting the structural requirements of 10 CFR 71.35. Purpose and Scope of NRC Regulatory Guide 7.8 10 CFR 71.71 and 10 CFR 71.73 describe normal conditions of transport and HACs that produce thermal and mechanical loads that serve as the structural design bases for the packaging of radioactive material for transport. Initial conditions must be assumed before analyses can be performed to evaluate the response of structural systems to prescribed loads. This regulatory guide presents the initial conditions that are considered acceptable by the NRC staff for use in the structural analysis of type B packages used to transport radioactive material in the contiguous United States.

Any information collection activities mentioned in NRC Regulatory Guide 7.8 are contained as requirements in 10 CFR 71, which provides the regulatory basis for this guide. Material Properties and Specifications The following is taken from NNSA SG-100. The materials selected for the fabrication of a packaging designed to transport nuclear materials are often chosen for their ability to shield and contain the radioactive contents. However, if fissile material is being transported, the materials of construction can also be used to maintain the spatial distribution and subcriticality within the package. The materials of construction for the packaging containment system should ensure structural containment of the hazardous material. These materials of construction, supported by the material certifications and structural analysis, must be shown by design and testing to adequately contain the radioactive material. 10 CFR 71.33, “Package Description,” requires applications to include a description of the proposed package in sufficient detail to identify the package accurately and provide a sufficient basis for evaluation of the package. The description must include specific information regarding the materials of construction, weights, dimensions, and fabrication methods. 10 CFR 71.31, “Contents of Application,” requires the applicant to include the identification of any established codes and standards proposed for use in the package design, fabrication, assembly, testing, maintenance, and use. In the absence of any codes and standards, it requires the applicant to describe and justify the basis and rationale used to formulate the package’s QAP. Although materials of construction are not specifically mentioned, their selection is an integral part of the design process. Therefore, the codes and standards that are used to determine the properties for the materials of construction need to be identified. 39 10 CFR 71.31(c) provides the applicant the option of describing and justifying the basis and rationale used to formulate the package’s QAP in the absence of any material codes or standards. However, the current regulatory practice has been to limit the choice of materials for vital package safety components to those that have existing codes and/or standards established by nationally recognized organizations or long-standing commercial manufacturers of products used in similar applications. The regulatory requirement in combination with the regulatory practice compel the applicant to use materials whose characteristics and properties are well-known and documented by testing, and verified by operational experience in similar applications. This ensures that the materials selected have material properties, which serve as a basis for the various safety analyses that have been extensively tested and proven. Any materials proposed for vital package safety components whose material properties are not described by such codes or standards must be adequately characterized and well-documented. The application should include a characterization of the materials with respect to the values of their mechanical, thermal, and physical properties; the means by which the quality of the materials is ensured; and any effects that the fabrication processes may have on the materials.

The material properties are external measures of how the materials behave in the presence of applied loads and environmental conditions such as internal or external pressure, dynamic and vibration loads, interactions with fluids, hot and cold temperatures, and so on. The objective is to use these properties to fully describe how the packaging materials will react to any combination of the applied loads and environmental conditions. The Application of Austenitic Stainless Steel The following is taken from the American Welding Society (AWS), Classification of Stainless Steel. AUSTENITIC Austenitic stainless steels are the most weldable of the stainless steels and can be divided rather loosely into three groups: common chromium-nickel; manganese-chromium-nickel-nitrogen; and specialty alloys. Austenitic is the most popular stainless steel group and is used for numerous industrial and consumer applications, such as in chemical plants, power plants, food processing and dairy equipment. Austenitic stainless steels have a face-centered cubic structure. Though generally very weldable, some grades can be prone to sensitization of the weld heat-affected zone (HAZ) and weld metal hot cracking. Video 2. Austenitic stainless steel https://www.youtube.com/watch?v=frwLBFoRj-0 FERRITIC Ferritic stainless steel consists of iron-chromium alloys with body-centered cubic crystal structures. They can have good ductility and formability, but high-temperature strengths are relatively poor when compared to austenitic grades. Some ferritic stainless steels used, for example, in mufflers, exhaust systems, kitchen counters and sinks, cost less than other stainless https://www.youtube.com/watch?v=frwLBFoRj-0 40 steels. Other more highly alloyed steels low in chromium and nickel are more costly, but are highly resistant to chlorides. SELECTING STAINLESS STEEL The selection of a particular type of stainless steel will depend on what requirements a particular application poses. Environment, expected part life and extent of acceptable corrosion all help determine what type of stainless to use. In most cases, the primary factor is corrosion resistance, followed by tarnish and oxidation resistance. Other factors include the ability to withstand pitting, crevice corrosion, and intergranular attack. The austenitic/higher chromium stainless steels, usually required in very high or very low temperatures, are generally more corrosion- resistant than the lower chromium ferritic or martensitic stainless steels. Most stainless steels are considered to have good weldability. It is important to make sure joint surfaces and any filler metal be kept free from oxide, organic material, or other contamination. A principal concern in selecting welding filler metals for stainless steels is to match the important properties of the base metal. In addition, for nominally austenitic and duplex stainless steels, one should have some control over the weld metal’s ferrite content. Specification of ferrite in nominally austenitic and duplex stainless steel welds are based upon ferrite numbers (FN) defined in AWS A4.2M/A4.2:2006, Standard Procedures for Calibrating Magnetic Instruments to Measure the Delta Ferrite Content of Austenitic and Duplex Ferritic-Austenitic Stainless Steel Weld Metal. Recommended by the ASME code, the magnetically determined FN is much simpler to obtain and is more reproducible than metallographically determined percent ferrite.

When selecting stainless steels, a welder must also consider something called sensitization. Ferritic stainless steels and some austenitic stainless steels, which contain appreciable free carbon can be rendered sensitive to intergranular corrosion in the HAZ of a weld. This sensitization occurs where a peak temperature of about 900 to 1,600°F is reached in the HAZ. Chromium carbides precipitate on grain boundaries, and in the process of doing so, chromium as an alloy element is depleted in the metal adjacent to the grain boundaries. Then, in corrosive service, this chromium-depleted metal is selectively attacked. Low welding heat input can limit, but not eliminate, sensitization. The best methods of preventing sensitization are selection of very low carbon base metal or selection of a grade stabilized with titanium or niobium, such as types 321 or 347. Note also that sensitization is almost never a weld metal problem—it is largely a HAZ problem. Chemical, Galvanic, and Other Reactions The following is taken from NNSA SG-100. The materials selected should be such that there will be no significant chemical, galvanic, or other reaction among the package components and contents. The primary concern during NCT and HAC is thermal degradation or decomposition, particularly when plastics or other low temperature materials are used. While not normally encountered in NNSA packages, radiolysis of hydrocarbons is a concern at higher radiation levels. Decomposition can affect the shielding properties of the material and can also cause pressurization of the CV if gases are generated. The most significant concern is a case in which decomposition of materials in the shielding results in 41 a loss of mechanical properties or a physical configuration change that may lead to a loss of shielding effectiveness. General Requirements for all Packages per 10 CFR 71.43 The following is taken from 10 CFR 71.43. The smallest overall dimension of a package may not be less than four inches. The outside of a package must incorporate a feature, such as a seal, that is not readily breakable and that, while intact, would be evidence that the package has not been opened by unauthorized persons. Each package must include a containment system securely closed by a positive fastening device that cannot be opened unintentionally or by a pressure that may arise within the package. A package must be made of materials and construction that ensure there will be no significant chemical, galvanic, or other reaction among the packaging components, among package contents, or between the packaging components and the package contents, including possible reaction resulting from in-leakage of water, to the maximum credible extent. Account must be taken of the behavior of materials under irradiation. A package valve or other device, the failure of which would allow radioactive contents to escape, must be protected against unauthorized operation and, except for a pressure relief device, must be provided with an enclosure to retain any leakage. A package must be designed, constructed, and prepared for shipment so that under the tests specified in 10 CFR 71.71, there would be no loss or dispersal of radioactive contents, no significant increase in external surface radiation levels, and no substantial reduction in the effectiveness of the packaging. A package must be designed, constructed, and prepared for transport so that in still air at 100°F and in the shade, no accessible surface of a package would have a temperature exceeding 122°F in a nonexclusive use shipment, or 185°F in an exclusive use shipment.

A package may not incorporate a feature intended to allow continuous venting during transport. 10 CFR 71 Requirements for Lifting and Tie-Down Devices The following is taken from 10 CFR 71.45. LIFTING DEVICES Any lifting attachment that is a structural part of a package must be designed with a minimum safety factor of three against yielding when used to lift the package in the intended manner, and it must be designed so that failure of any lifting device under excessive load would not impair the ability of the package to meet other requirements of 10 CFR 71.45, “Lifting and Tie-Down Standards for All Packages.” Any other structural part of the package that could be used to lift the package must be capable of being rendered inoperable for lifting the package during transport, or must be designed with strength equivalent to that required for lifting attachments. 42 TIE-DOWN DEVICES: If there is a system of tie-down devices that is a structural part of the package, the system must be capable of withstanding, without generating stress in any material of the package in excess of its yield strength, a static force applied to the center of gravity of the package having a vertical component of two times the weight of the package with its contents, a horizontal component along the direction in which the vehicle travels of ten times the weight of the package with its contents, and a horizontal component in the transverse direction of five times the weight of the package with its contents. Any other structural part of the package that could be used to tie down the package must be capable of being rendered inoperable for tying down the package during transport, or must be designed with strength equivalent to that required for tie-down devices. Each tie-down device that is a structural part of a package must be designed so that failure of the device under excessive load would not impair the ability of the package to meet other requirements of 10 CFR 71.45. b. Discuss the following related to normal conditions of transport:  Evaluation methods  Most limiting initial test conditions  Most damaging orientations Evaluation Methods The following is taken from 10 CFR 71.71. EVALUATION Evaluation of each package design under normal conditions of transport must include a determination of the effect on that design of the conditions and tests specified in 10 CFR 71.71. Separate specimens may be used for the compression test, and the penetration test, if each specimen is subjected to the water spray test before being subjected to any of the other tests. INITIAL CONDITIONS With respect to the initial conditions for the tests in 10 CFR 71.71, the demonstration of compliance with the requirements of 10 CFR 71.71 must be based on the ambient temperature preceding and following the tests remaining constant at that value between -20ºF and +100ºF which is most unfavorable for the feature under consideration. The initial internal pressure within the containment system must be considered to be the maximum normal operating pressure (MNOP), unless a lower internal pressure consistent with the ambient temperature considered to precede and follow the tests is more unfavorable. CONDITIONS AND TESTS Heat—An ambient temperature of 100ºF in still air, and insolation according to table 9. 43 Table 9. Insolation data Form and location of surface Total insolation for a 12-hour period (g cal/cm2) Flat surfaces transported horizontally: Base Other surfaces

Flat surfaces not transported horizontally Curved surfaces None 800 200 400 Source: 10 CFR 71.71 Cold—An ambient temperature of -40ºF in still air and shade. Reduced external pressure—An external pressure of 3.5 pound force (lbf)/in.2 absolute. Increased external pressure—An external pressure of 20 lbf/in.2 absolute. Vibration—Vibration normally incident to transport. Water spray—A water spray that simulates exposure to rainfall of approximately 5 cm/h (2 in./h) for at least 1 hour. Free drop—Between 1.5 and 2.5 hours after the conclusion of the water spray test, a free drop through the distance specified in table 10 onto a flat, essentially unyielding, horizontal surface, striking the surface in a position for which maximum damage is expected. Table 10. Criteria for free drop test (weight distance) Package weight Free drop distance Kilograms Pounds Meters Feet Less than 5,000 Less than 11,000 1.2 4 5,000 to 10,000 11,000 to 22,000 0.9 3 10,000 to 15,000 22,000 to 33,100 0.6 2 More than 15,000 More than 33,100 0.3 1 Source: 10 CFR 71.71 CORNER DROP A free drop onto each corner of the package in succession, or in the case of a cylindrical package onto each quarter of each rim, from a height of 0.1 foot (ft) onto a flat, essentially unyielding, horizontal surface. This test applies only to fiberboard, wood, or fissile material rectangular packages not exceeding 110 pounds (lbs) and fiberboard, wood, or fissile material cylindrical packages not exceeding 220 lbs. 44 COMPRESSION For packages weighing up to 11,000 lbs, the package must be subjected, for a period of 24 hours, to a compressive load applied uniformly to the top and bottom of the package in the position in which the package would normally be transported. The compressive load must be the greater of the following:  The equivalent of 5 times the weight of the package  The equivalent of 2 lbf/in.2 multiplied by the vertically projected area of the package PENETRATION Impact of the hemispherical end of a vertical steel cylinder of 1.25 inch diameter and 13 lbs mass, dropped from a height of 40 inches onto the exposed surface of the package that is expected to be most vulnerable to puncture. The long axis of the cylinder must be perpendicular to the package surface. Most Limiting Initial Test Conditions The following is taken from NNSA SG-100. NRC Regulatory Guide 7.9 identifies the tests that must be performed prior to the first use of the packaging: visual inspection for physical defects, structural and pressure tests, leak tests, and individual component tests. NRC Regulatory Guide 7.9 states that “components, such as gaskets, should be tested under conditions that simulate the most severe service conditions under which they are to perform.” Such testing should be performed on all components whose failure would impair the effectiveness of the packaging, not just gaskets. Clearly it is not required that all specimens of a specific component be tested, just that enough units must be tested to provide a high level of confidence that any specimen of a component specified for such service can be expected to perform properly. This is especially true for components that can only be tested in a destructive manner, such as rupture disks. It is also acceptable to test the component to criteria that exceed the most severe condition, assuming that exceeding the most severe condition will not have a detrimental effect on the performance of the package. These components include, but are not limited to, valves, rupture disks, fluid transport devices, as well as gaskets. All components must be procured under an approved QAP.

If any of the packaging components are required for shielding integrity or thermal protection, tests must be performed to ensure the shielding requirements for the packaging are met and the thermal protection maintains a temperature that is consistent with the thermal analyses and/or tests conducted for NCT. Acceptance criteria for each test and inspection must be specified, and the packaging must not be used until all acceptance criteria are met. The acceptance and/or verification testing, inspection, and maintenance procedures must describe actions to be taken if acceptance criteria are not met. Tests and inspections performed and their acceptance criteria must be determined on the basis of the quality category of the component, or of the packaging as a whole. Most Damaging Orientations According to International Atomic Energy Agency (IAEA) TS-G.1-1 (ST-2), Advisory Material for the IAEA Regulations for the Safe Transport of Radioactive Material, packages must be tested in whichever orientation results in the most damage. 45 According to 10 CFR 71.71, the impact of the hemispherical end of a vertical steel cylinder of 1.25 inches diameter and 13 lbs mass, dropped from a height of 40 inches onto the exposed surface of the package that is expected to be most vulnerable to puncture. The long axis of the cylinder must be perpendicular to the package surface. According to NNSA SG-100, Rev 2, Safety Guide: Design and Development Guide for NNSA Type B Packages, if applicable, the effects of the dynamic crush test on a package should be described. The most unfavorable orientation should be justified. c. Discuss the criteria that determine whether the dynamic crush test applies and describe how this test affects DOE’s ability to add new contents to packages certified prior to October 2004. The following is taken from NNSA SG-100. 10 CFR 71.73, Subsection c2 was modified by the addition of the following sentence: “For packages containing fissile material, the radioactive contents greater than 1000 A2 criterion does not apply.” (The pre-October 1, 2004, 10 CFR 71.73(c)(2) did not require this test for packages with radioactive material content less than or equal to 1,000 A2.) 10 CFR 71.73(c)(2) now reads as follows: Subjection of the specimen to a dynamic crush test by positioning the specimen on a flat, essentially unyielding horizontal surface so as to suffer maximum damage by the drop of a 1,100-lb mass from 30 ft onto the specimen. The mass must consist of a solid mild steel plate 40 inches by 40 inches and must fall in a horizontal attitude. The crush test is required only when the specimen has a mass not greater than 1,100 lbs, an overall density not greater than 62.4 lb/ft3 based on external dimension, and radioactive contents greater than 1,000 A2 not as special form radioactive material. For packages containing fissile material, the radioactive contents greater than 1,000 A2 criterion does not apply. d. Describe the 10 CFR 71 tests required for hypothetical accident conditions and discuss the requirements that pertain to the sequence of HAC tests and the allowable and typical methods of demonstrating compliance. The following is taken from 10 CFR 71.73. Test Procedures Evaluation for HAC is to be based on sequential application of the tests specified in 10 CFR 71.73, in the order indicated, to determine their cumulative effect on a package or array of packages.

Test Conditions With respect to the initial conditions for the tests, except for the water immersion tests, to demonstrate compliance with the requirements of 10 CFR 71.73 during testing, the ambient air temperature before and after the tests must remain constant at that value between -29oC (20oF) and +38oC (+100º) which is most unfavorable for the feature under consideration. The initial internal pressure within the containment system must be the MNOP, unless a lower internal 46 pressure, consistent with the ambient temperature assumed to precede and follow the tests, is more unfavorable. Tests Tests for HACs must be conducted as follows:  Free drop—A free drop of the specimen through a distance of 30 ft onto a flat, essentially unyielding, horizontal surface, striking the surface in a position for which maximum damage is expected.  Crush—Subjection of the specimen to a dynamic crush test by positioning the specimen on a flat, essentially unyielding horizontal surface so as to suffer maximum damage by the drop of a 1,100-lb mass from 30 ft onto the specimen. The mass must consist of a solid mild steel plate 40 inches by 40 inches and must fall in a horizontal attitude. The crush test is required only when the specimen has a mass not greater than 1,100 lbs, an overall density not greater than 62.4 lb/ft3 based on external dimension, and radioactive contents greater than 1,000 A2 not as special form radioactive material. For packages containing fissile material, the radioactive contents greater than 1,000 A2 criterion does not apply.  Puncture—A free drop of the specimen through a distance of 40 inches in a position for which maximum damage is expected, onto the upper end of a solid, vertical, cylindrical, mild steel bar mounted on an essentially unyielding, horizontal surface. The bar must be 6 inches in diameter, with the top horizontal and its edge rounded to a radius of not more than 0.25 inch, and of a length as to cause maximum damage to the package, but not less than 8 inches long. The long axis of the bar must be vertical.  Thermal—Exposure of the specimen fully engulfed, except for a simple support system, in a hydrocarbon fuel/air fire of sufficient extent, and in sufficiently quiescent ambient conditions, to provide an average emissivity coefficient of at least 0.9, with an average flame temperature of at least 1,475ºF for a period of 30 minutes, or any other thermal test that provides the equivalent total heat input to the package and which provides a time averaged environmental temperature of 800ºC. The fuel source must extend horizontally at least 40 inches, but may not extend more than 10 ft, beyond any external surface of the specimen, and the specimen must be positioned 40 inches above the surface of the fuel source. For purposes of calculation, the surface absorptivity coefficient must be either that value which the package may be expected to possess if exposed to the fire specified or 0.8, whichever is greater; and the convective coefficient must be that value which may be demonstrated to exist if the package were exposed to the fire specified. Artificial cooling may not be applied after cessation of external heat input, and any combustion of materials of construction must be allowed to proceed until it terminates naturally.  Immersion, fissile material—For fissile material subject to 10 CFR 71.55, “General Requirements for Fissile Material Packages,” in those cases where water in-leakage has not been assumed for criticality analysis, immersion under a head of water of at least 3 ft in the attitude for which maximum leakage is expected.

 Immersion, all packages—A separate, undamaged specimen must be subjected to water pressure equivalent to immersion under a head of water of at least 50 ft. For test purposes, an external pressure of water of 21.7 lbf/in.2 gauge is considered to meet these conditions. 47 e. Discuss the NNSA SG-600, Regulatory Compliance Testing of NNSA Type B Packages. NNSA Service Center SG-600 is a comprehensive collection of information covering the many different subject areas that must be addressed to complete a successful package testing sequence. The purpose of NNSA SG-600 is to provide the necessary information to be used in conjunction with all applicable regulations to test packages for the transport of SNM components. The guide can also be used to test packages for other unirradiated nuclear materials. These specific radioactive materials can be shipped only in certified type B(F) containers that meet all applicable 10 CFR 71 and 49 CFR 100–178 requirements. Although much of this document has universal application, the guide is specifically focused on drum-type, type B, fissile packages used by the DOE/NNSA for transporting SNM components. The primary goal of NNSA SG-600 is to document successful testing practices that have resulted in certified type B packages and to make the information on these practices readily available to NNSA organizations and/or individuals who can benefit from its use. 6. NNSA package certification engineers must demonstrate a working level knowledge of the thermal evaluation information required in chapter 3 of a SARP. a. Discuss the NNSA SG-100 Section 3.2.3.2 (Physical Systems) through Section 3.2.5.2 (Successful Designs) as they pertain to thermal evaluation. Packages used in the shipment of radioactive material must be capable of withstanding intense thermal environments while preventing the release of contents and maintaining shielding and nuclear subcriticality. Achieving this capability requires the use of construction materials that enable the package to withstand serious thermal insult. Insulating materials are used to slow the transfer of heat toward the package content during a thermal accident condition. The material used to provide this thermal protection may also be used to increase structural stability. For packages designed to carry contents with significantly large heat sources, the issue of thermal protection becomes much more complicated. In these cases, the insulating material must work effectively to reduce the heat added to the package during an upset condition while also allowing internally generated heat to escape under regular operating conditions. These are conflicting requirements that must be carefully balanced during the package design process. Other forms of thermal protection can be employed. Passive systems, such as heat pipes, can be used, but their viability in the event of an accident must be demonstrated before package certification. Fins can also be used to increase the rate of heat transfer from the package content to the package’s surroundings. However, fin cooling systems can also increase the rate of heating to the interior portions of the package during an accident scenario. One possibility is to make the fins of a material that is stable at lower temperatures but experiences a phase change at elevated temperatures. The problem with this design theory is that, while the fins may melt when exposed to the accident scenario outlined in 10 CFR 71, they may not melt if they were near but not fully engulfed in such an accident. In this case, it is possible for the fins to stay intact and actually increase the quantity of heat absorbed by the package.

48 Conduction pathways can be used to help remove heat from the CV on packages with a significant internal heat generation. Any type of high thermal conductivity material can be used as long as it does not interfere with any of the package’s other functions. Special care must be taken to ensure such pathways do not allow heat an easy access to sensitive portions of the packaging during a thermal accident situation. In general, the use of pressure vents and valves is not encouraged on packages in the NNSA. Pressure vents and valves should be used only if they are absolutely necessary and must comply with 10 CFR 71.43(e). If they are used, the concept of a fail-safe mode must be applied, and the applicant must be able to convince the regulators that a fail-safe situation does exist. Thermophysical Properties Several properties affect how a material or group of materials will respond to a thermal input. The discussion of these properties provided in this guide is brief, and more information can be found in most standard heat transfer texts. Thermal conductivity is a measure of a material’s ability to conduct heat. Units for this quantity are expressed in W/m-K, or more explicitly W-m/m2-K. Metals tend to be the most conductive class of materials, but conductivity within this group varies greatly. Of the common materials used for construction of packaging parts, aluminum is the most highly conductive with a thermal conductivity of about 240W/m-K. Carbon steel has a much lower conductivity of about 60 W/m-K, and stainless steels have an even lower conductivity of about 15 W/m-K. Insulating materials, as their name suggests, have a much lower thermal conductivity, typically on the order of 0.05 W/m-K, but this number is somewhat misleading. Most insulating materials employ air pockets or gaps. Through this type of material, heat transfer occurs by natural convection and radiation, not conduction. In most cases, these other heat transfer mechanisms are less efficient than conduction; thus, the overall or apparent thermal conductivity of the material is very low. In general, solids are more conductive than liquids, which are more conductive than gases. However, insulating materials (which are usually considered solids) often display thermal conductivities equal to or less than that of some liquids and gases as these materials are actually a mixture of solids and gases. Thermal conductivities of common materials can be found in most heat transfer texts. For less common materials only the thermal conductivity of a material is needed to calculate steady-state conduction heat transfer rates. Specific heat is a measure of the amount of energy it takes to heat a specified quantity of a material a specified number of degrees. The units for this property can be expressed in kJ/(kg-K). The heat capacity of most materials increases with temperature. Density is the mass per volume of a material and units are kg/m3. For most materials, density decreases as temperature rises, although some materials may contract. All three of the material properties that have been discussed so far (thermal conductivity, heat capacity, and density) are needed to calculate non-steady state (transient) conduction heat transfer rates. The only material property needed to calculate heat transfer rates due to radiation is emissivity (or absorbtivity). Emissivity is a measure of how efficiently a surface of a material emits radiant energy compared with an ideal radiator (blackbody). The emissivity of a

49 material surface can vary greatly depending on the finish of the surface. While selecting a material with either a low or high emissivity due to its surface finish may help protect a package, attaining, keeping, certifying, and documenting the specified surface finish on a package part may be difficult. Most applications of radiative heat transfer dealing with a package are for the package surroundings to the package exterior (or vice versa). However, if gaps exist within the package, heat transfer across these gaps will be by radiation and convection. Several properties must be known for the calculation of convection heat transfer coefficients, and they will always be fluid properties. The fluid properties needed to calculate the natural convection heat transfer coefficient are dependent on the correlations used, but generally the properties are fluid viscosity, density, volumetric thermal expansion coefficient, thermal conductivity, and heat capacity. For the vast majority of cases, the fluid will be air, although some other liquids, such as nitrogen and helium, are used in package design. Properties for these fluids can be found in most heat transfer texts. One of the most important physical attributes involving thermal aspects of package design is the decomposition of materials of construction at elevated temperatures. Many forms of decomposition are possible and include dehydration, melting, pyrolysis, and combustion. Not all of these processes need to be avoided; some can serve as energy absorption mechanisms, while others tend to have degrading effects on package performance. Dehydration is the act of vaporizing free water. The phase change of the H2O from liquid to gas (steam) requires a considerable quantity of energy. Thus, during a thermal excursion, this process can absorb a considerable quantity of energy thereby allowing interior package temperatures to remain lower then would otherwise occur. Generally, the bulk of the steam produced is driven from the package, due to a buoyancy gradient, and escapes through vent holes or other openings possibly caused by previous HAC structural insults. Materials used for structural impact absorption and thermal insulation, such as Kaolite 1600TM and Celotex exhibit this trait. In fact, for the inorganic Kaolite 1600TM material, the only form of decomposition during thermal excursions up to about 2,300°F is the vaporization of free water from within the material matrix. Although the vaporization is usually good for a package’s thermal characteristics, there can be negative effects on the shielding and subcriticality properties of a shipping package due to the dehydration of materials of construction. This should be carefully considered when both shielding and criticality calculations are made. It is important to understand the materials of construction regarding the possibility of dehydration and to take these processes into account in subsequent calculations. Many organic materials, such as Celotex and polyurethane foams, undergo chemical decomposition at the elevated temperatures associated with the HAC thermal event. If the decomposition occurs with oxygen present, the result is typically combustion; if oxygen is not present, the result is typically pyrolysis. Each of these processes can be either exothermic (heat releasing) or endothermic (heat absorbing), and should be examined for each material that is considered for use. Typically, the decomposition process—whether it be combustion, pyrolysis or a combination of both—is very complicated, especially for organic materials.

50 For most cases, the decomposition processes are not fully predictable or understood; thus, conservatism is important when dealing with this aspect of package design. Finally, melting is typically associated with elastomeric O-rings that form the CV seal. Clearly, this type of behavior is to be avoided in package designs as melting of a containment boundary O-ring leads immediately to loss of containment. Typically, the manufacturers of the elastomeric O-rings used in drum-type packages specify a maximum continuous use temperature as well as specified durations at higher temperatures that the O-ring will survive. It may be necessary to determine, by test or analysis, the duration that a seal may be above the continuous use temperature, during the design or certification process. It should be noted that some plastic behavior has been seen in O-rings that continue to provide a leak-tight seal. Squared-off O-rings have been found in drum-type package design test units, even though prior to removal of the CV lid the O-rings provided a leak-tight containment seal. Successful Designs Many package designs ranging from very simple to quite complex have been successful. Packages with very small heat sources tend to have the more simplistic designs, whereas those with sizable heat sources tend to be more extravagant. The simplest and most common package design now used in the NNSA is a stainless steel CV surrounded by a thick layer of material that acts as a thermal insulation and an impact limiter. This layer is then surrounded by the confinement drum, which is generally constructed of either mild or stainless steel. This very basic design is usually used for packages with contents that have a reasonable heat output. If some very basic design strategies are followed, this package design can be effective and relatively inexpensive. The main concerns with this type of design are the use of the organic materials and their propensity to offgas when exposed to high temperatures. This property makes analytical modeling of the hypothetical thermal accident almost impossible unless an extremely conservative approach is taken. Offgassing notwithstanding, this design has repeatedly been able to withstand the rigors of the hypothetical thermal accident environment without loss of function. A more sophisticated version of this design uses two layers of thermal insulation/impact limiting material. These two layers are sealed off from one another by creating compartments for the two layers. The outermost compartment contains the insulating material. In the event of a accident involving a fire, the outside wall of the insulation compartment would be exposed to very high temperatures. Heat from this exposure would be transferred into the insulating material adjacent to this wall. As this insulation is heated, decomposition processes will occur and hot gasses will be formed. As these gasses are formed, the insulation compartment will pressurize. Most of the gases will escape through vent holes provided for such an event; however, this internal pressurization will force hot gasses to the inside wall of the compartment. Due to compartmentalization, offgas from the outer most layer cannot reach the CV, but rather can reach only the outside of the inner wall of the insulation compartment. The insulation/impact limiting material used for the inner layer is different from that used on the outer layer. During physical thermal accident testing, some slight degradation of the inner thermal insulation does occur. But, in general, CVs withstand this

51 type of test with a much smaller accumulation of offgas condensate on their outer surfaces then would the simpler design described earlier. For some packages with significant heat sources, it is very important to spread out the heat quickly such that the content does not overheat. One method that has been employed is to use an aluminum packing material inside the CV to help carry off the heat coming from the source. Aluminum is an excellent conductor of heat and thus works very well for this purpose. Initial designs for this package called for the use of aluminum shot to be used. The package is quite small, and the weight of the shot was significant when compared with the rest of the package. As an alternative, an aluminum packing very similar to aluminum straws was tried. This packing was just as effective at removing heat at only about one-third the weight of the aluminum shot. The ability of aluminum to transfer heat so rapidly simulates a heat source the volume of the CV interior rather than just the size of the content. That is, heat generated by the content is transferred to the aluminum surrounding it. The aluminum is so efficient at conducting heat there is little temperature gradient throughout this material. Because the entire CV is filled with either the content or the aluminum, the temperature profile throughout the entire CV cavity is nearly constant. In this way, the temperature of the content is minimized, and the heat transfer from the CV is maximized, to the extent possible for the package design being used. When employing this concept, the content is not constrained within the CV, and some shifting of the content may occur. Another strategy used to keep heat down inside the CV is the use of aluminum conduction bars that attach directly to the content and to the lid of the CV. For this container, it was necessary to hold the content in place, and a polypropylene material was selected for this duty. Placement of the content is critical to ensure there is good physical contact between the content and the conduction bars; therefore, the conduction bars are built into the CV lid such that good contact there is assured. This design is effective, and there is no reason why other similar designs could not be constructed. One possibility is to build an aluminum holder into the CV with the holder attached to the CV sidewall. Hence, both objectives of helping to remove heat while securing the content in a specific location are met. b. Discuss NNSA SG-140.1, Combination Test/Analysis Method Used to Demonstrate Compliance to DOE Type B Packaging Thermal Test Requirements. The following is taken from NNSA SG-140.1. Applicants requesting certification of packaging for transport of defense programs radioactive material must, in general, show that the packaging conforms to the DOE performance requirements adopted from 10 CFR 71. The subject of DOE-SG-140.1 is the DOE interpretation of the hypothetical accident scenario described in 10 CFR 71.73, as follows: Thermal. Exposure of the specimen fully engulfed, except for a simple support system, in a hydrocarbon fuel/air fire of sufficient extent, and in sufficiently quiescent ambient conditions, to provide an average emissivity coefficient of at 52

least 0.9, with an average flame temperature of at least 800°C (1,475°F) for a period of 30 minutes, or any other thermal test that provides the equivalent total heat input to the package and which provides a time averaged environmental temperature of 800°C. The fuel source must extend horizontally at least 1 m (40 in.), but may not extend more than 3 m (10 ft), beyond any external surface of the specimen, and the specimen must be positioned 1 m (40 in.) above the surface of the fuel source. For purposes of calculation, the surface absorptivity coefficient must be either that value which the package may be expected to possess if exposed to the fire specified or 0.8, whichever is greater; and the convective coefficient must be that value which may be demonstrated to exist if the package were exposed to the fire specified. Artificial cooling may not be applied after cessation of external heat input, and any combustion of materials of construction, must be allowed to proceed until it terminates naturally. NNSA SG-140.1 has been prepared as a guideline to ensure applicants are aware of DOE’s interpretation of the requirement and to provide consistent review policy. It should be recognized by DOE regulatory certifying personnel and applicants using this guideline that other viable means of demonstrating conformance may exist and that it is not the intent of NNSA SG-140.1 to imply that any such methods may or may not be acceptable. Specifically, NNSA SG-140.1 is applicable to a combination test/analysis methodology that involves heating a package in a furnace or other enclosed radiative environment where the package and the furnace have similar geometric shapes or where dimensional considerations allow the assumption of similar geometric shape. c. Discuss the purpose and scope of ASTM E 2230, Standard Practice for Thermal Qualification of Type B Packages for Radioactive Material. The major objective of ASTM E 2230 is to provide a common reference document for applicants and certification authorities on the accepted practices for accomplishing package thermal qualification. Details and methods for accomplishing qualification are described in this document in more specific detail than available in the regulations. Methods that have been shown by experience to lead to successful qualification are emphasized. Possible problems and pitfalls that lead to unsatisfactory results are also described. ASTM E 2230 defines detailed methods for thermal qualification of type B radioactive materials packages under 10 CFR 71 in the United States or, under IAEA regulation TS-R-1, Regulations for the Safe Transport of Radioactive Material. Under these regulations, packages transporting what are designated to be type B quantities of radioactive material shall be demonstrated to be capable of withstanding a sequence of hypothetical accidents without significant release of contents. ASTM E 2230 is used to measure and describe the response of materials, products, or assemblies to heat and flame under controlled conditions, but does not by itself incorporate all factors required for fire hazard or fire risk assessment of the materials, products, or assemblies under actual fire conditions. 53 d. Discuss the NCT thermal analyses/tests that are evaluated in the SARP. The following is taken from NNSA SG-100.

The following example demonstrates an actual thermal acceptance test used to evaluate new drum-type packagings. The test was developed to ensure that the heat load capabilities of the package are acceptable for NCT. A representative sample packaging is tested as follows:  Assemble a packaging. In the CV, include a device that generates the maximum permissible heat load of the package being emulated.  During assembly, attach thermocouples to the outside top and side of the CV and to the outside top and side of the drum surfaces.  Place the packaging in a normal condition environment of 100°F until it reaches steady- state temperatures. Record the temperatures.  The temperatures should not exceed 250°F on the CV and 136°F on the drum surface. If the temperatures are exceeded, notify the design agency for the packaging. e. Discuss the HAC thermal analyses/tests that are evaluated in the SARP. This topic is covered in KSA 3d. f. Discuss the thermal test methods used to demonstrate compliance with 10 CFR 71 thermal requirements. The following is taken from NNSA SG-600. The thermal testing of a package represents a challenge to the packaging designer and test engineer. Along with this testing challenge may come some details that cannot be foreseen. The thermal test procedure is covered in detail by ASTM Standard Practice E 2230. The ASTM E 2230 discusses the use of pool fires, furnaces, radiant heat facilities, and analysis for performing the thermal test described in 10 CFR 71.73. For pool fires and furnaces, the material covered in ASTM E 2230 is relatively conclusive. However, for radiant testing and analysis, some additional discussion is warranted. It is paramount that applicants and testing authorities fully discuss their planned thermal testing methodology with NNSA Packaging Certification Division (PCD) early in the test planning process to allow for changes to their plans based on PCD input, if needed. Thermal Test by Analysis Because the vast majority of packages certified by NNSA are relatively small, NNSA expects the applicants to perform actual thermal testing, rather than a thermal analysis, for package certification. Large spent fuel casks typically use metallic seals that do not fail even when exposed to the severe environment of the thermal test for very long periods of time. In contrast, most packages certified by NNSA use elastomeric O-rings that could fail at relatively low temperatures compared with temperatures reached in the thermal test environment. Furthermore, full-scale test units for spent fuel casks can cost millions of dollars. 54 Because the cost of fabricating the test units for the package designs that NNSA certifies is typically much less, NNSA expects its applicants to build prototype units and test them to all of the requirements. Therefore, it is strongly recommended that some form of physical testing, instead of computer analysis, be used to demonstrate compliance with the 10 CFR 71.73 thermal test requirements. However, computer analysis or other types of analysis may be used to make adjustments to results obtained from actual physical thermal tests performed.

Thermal Test by Radiant Heat A radiant heat testing methodology has been developed at Sandia National Laboratories. The methodology uses a large array of very bright radiant heat lamps in proximity to the test unit to create an environment similar to that specified in 10 CFR 71.73. For a typical setup, a cylindrical package is oriented upright usually about two feet off the ground. The radiant heat lamp arrays are then placed so as to fully encircle the sides of the test unit. The heat lamps usually extend about one foot beyond the top and bottom of the test unit. The test unit surface temperature can be monitored and the radiant heat system can be controlled based on these surface temperature measurements. However, in most cases, the open space between the test unit and the heat lamps, and the ambient air above and below the test unit, allow considerable airflow from the area external to the test rig. That is, as the air inside is heated, it rises toward the top of the test rig and out the top. This lost heated air is replaced with ambient-temperature air entering at the bottom of the test rig. This creates a convection current that carries off much of the heat imparted from the heat lamps to the package. For any test method using the radiant heat technology, consideration of the convective cooling must be taken into account, and test methods that effectively deal with the issue of airflow while still meeting all the requirements of 10 CFR 71.73 must be used and demonstrated to the satisfaction of the competent authority. The materials of construction, especially any that are combustible, must be considered in determining the orientation in which the test unit will be placed for thermal testing, regardless of the type of thermal test method used. The possibility of chimney effects, especially in packages that have a high height-to-width ratio, should be considered. If previous testing created any breaches into the test unit’s insulation, and the insulation is known to be combustible or will degrade significantly during a thermal event, the test unit should be oriented so the atmospheric gases within the thermal test can flow into the breach and out through any other available area. The orienting of the test unit is subjective; however, someone with considerable knowledge of heat transfer should be consulted to help determine the test unit’s “worst case” orientation. This use of heat transfer expertise should be documented and provided as part of either the test report or the discussion in the associated SARP. Applicants who are thinking of conducting the thermal test by the use of radiant heating should discuss their plan with PCD and get PCD’s concurrence before finalizing their plan. 55 g. Discuss the application of fire resistant and consumable thermal insulation materials, and describe the benefits and disadvantages of each. The following is taken from the American Society of Mechanical Engineers, Operational Readiness of the ES-3100 Type B Shipping Container for Fissile Materials.

The DOT specification 6M containers had been the workhorse for bulk highly enriched uranium shipping containers for DOE and many other shippers for over twenty years. This DOT- specification container was terminated for shipment of radioactive material on September 30, 2008. The anticipation of this action prompted DOE to develop and implement the ES-3100 shipping container as a replacement for the 6M. The ES-3100 was first licensed in April 2006. Since then, the license has been revised nine times. The ES-3100 was operationally ready for use at several sites by September 2007, and is now being used on a regular basis for materials that had been shipped in the DOT 6M. The ES-3100 has also been certified for air transport, in support of foreign research reactor fuel supply and international nonproliferation efforts. This container has a CoC from the NRC and a competent authority certificate from the DOT. The utility of the ES-3100 continues to grow. The ES-3100 CoC allows many forms of fissile material to be shipped, and continues to be amended to authorize additional contents for a variety of shippers. Video 3. Thermal insulation material https://www.youtube.com/watch?v=T24o3ZXmJDA 7. NNSA package certification engineers must demonstrate a working level knowledge of the containment information required in chapter 4 of a SARP. a. Discuss the NNSA SG-100 Sections 4.2 through 4.8 as they pertain to containment. Chapter 4 addresses containment for type B, radioactive material packaging systems. Requirements for type B packaging systems are contained in the applicable sections of 10 CFR 71. Containment for packaging is demonstrated through leakage rate testing. Standards for leakage rate testing of type B packages are presented in American National Standards Institute (ANSI) N14.5,Radioactive Materials—Leakage Tests on Packages for Shipment. This chapter provides guidance on the requirements for development and design of containment systems and on preparation of the containment section of the SARP. This chapter identifies the regulatory requirements for containment, and describes, with examples, how those requirements may be satisfied in a packaging system. b. Discuss the containment criteria for normal form radioactive material under NCT and HAC. The following is taken from NNSA SG-100. Regulatory requirements for packaging and transportation of radioactive materials, components, and special assemblies are governed by DOE Orders and various Federal regulations. This section describes the primary regulatory documents that contain the requirements for containment, including DOE O 461.1B, 49 CFR 100–185, and 10 CFR 71. https://www.youtube.com/watch?v=T24o3ZXmJDA 56 DOE O 461.1B DOE O 461.1B, Packaging and Transfer or Transportation of Materials of National Security Interest, establishes requirements and responsibilities for offsite shipments and onsite transfers of naval nuclear fuel elements, category I and category II special nuclear material (SNM), nuclear components, special assemblies and other MNSI, and the certification of packages for category I and II SNM, nuclear components and other MNSI. DOE O 461.1B is applicable to all DOE elements, including the NNSA. DOE O 461.1B requires that shipments of radioactive materials be prepared and transported in accordance with the applicable DOT hazardous materials regulations in 49 CFR 100–185.

49 CFR 100–185 49 CFR 173, “Shippers—General Requirements for Shipments and Packagings,” lists general requirements for all hazardous materials and specific requirements for various classes of hazardous materials. General requirements are located in 49 CFR 173, Subpart A. 49 CFR 173.7(d) grants the DOE the authority to evaluate, approve, and certify packagings made by or under the direction of the DOE for the transportation of class 7 materials provided the DOE uses standards that are equivalent to those specified in 10 CFR 71. Specific regulatory requirements for transportation of radioactive materials are located in 49 CFR 173, Subpart I, “Class 7 (Radioactive) Materials.” In 49 CFR 173.410, “General Design Requirements,” DOT specifies general design requirements for radioactive material packages. DOT also requires that type B packages be designed and constructed to meet the applicable requirements specified in 10 CFR 71. A type B package is one designed to contain a type B quantity of radioactive material. A type B quantity is defined as any quantity of special form material having an activity of greater than A1 or any quantity of normal form material having an activity of greater than A2. 10 CFR 71 10 CFR 71, “Packaging and Transportation of Radioactive Material,” specifies the design, construction, and testing requirements for type B packages. Containment requirements are specified in 10 CFR 71.43, which states: (f) A package must be designed, constructed, and prepared for shipment so that under the tests specified in 10 CFR 71 there would be no loss or dispersal of radioactive contents… (h) A package may not incorporate a feature intended to allow continuous venting during transport. The phrase “no loss or dispersal of radioactive contents” is clarified in 10 CFR 71.51, “Additional Requirements for Type B Packages,” which states: (a) A Type B package,…, must be designed, constructed, and prepared for shipment so that under the tests specified in: (1) Section 71.71, “Normal Conditions of Transport,” there would be no loss or dispersal of radioactive contents—as demonstrated to a sensitivity of 10-6 A2 per hour, no significant 57 increase in external surface radiation levels, and no substantial reduction in the effectiveness of the packaging; and (2) Section 71.73, “Hypothetical Accident Conditions,” there would be no escape of krypton-85 exceeding 10 A2 in 1 week, no escape of other radioactive material exceeding a total amount A2 in 1 week… (b) Where mixtures of different radionuclides are present, the provisions of Appendix A, paragraph IV of this part shall apply, except that for krypton-85, an effective A2 value equal to 10 A2 may be used. (c) Compliance with the permitted activity release limits of paragraph (a) of this section may not depend on filters or on a mechanical cooling system. (d) For packages which contain radioactive contents with activity greater than 105 A2, the requirements of 10 CFR 71.61, “Special Requirements for Type B Packages Containing More than 105A2,” must be met. It must be demonstrated that containment, defined as a release rate of less than 10-6 A2 per hour, is maintained for packages subjected to the normal conditions of transport tests specified in 10 CFR 71.71. The tests specified in 10 CFR 71.71 consist of vibration, water spray, free drop, corner drop, compression, and penetration. The 10 CFR 71.71 tests are to be conducted at the most unfavorable conditions of external temperature and pressure.

Packages must also demonstrate that containment, defined as a release of less than 1 A2/wk (except krypton-85 may be released at up to 10 A2/wk), is maintained for packages subjected to the HAC tests specified in 10 CFR 71.73. The tests defined in 10 CFR 71.73 must all be applied to a single package in the specified sequence. The test sequence is a 30-ft free drop, followed by crush, puncture, and fire. The tests are to be performed at the most unfavorable temperature, and the initial pressure within the containment system must be the MNOP unless a lower internal pressure is more unfavorable. An immersion test is also required, but may be performed on an undamaged test specimen. There are important changes to the containment requirements included in the 10 CFR 71 version effective October 1, 2004. Each of the primary changes affecting containment requirements is discussed below. 10 CFR 71.61 was renamed and revised to specify requirements for the packages identified in its title. Earlier versions of 10 CFR 71 applied this requirement only to irradiated nuclear fuel shipments. 10 CFR 71.61 now reads as follows: A type B package containing more than 105 A2 must be designed so that its undamaged containment system can withstand an external water pressure of 290 psi for a period of not less than 1 hour without collapse, buckling, or in-leakage of water. 10 CFR 71.63, “Special Requirements for Plutonium Shipments,” was modified to eliminate 58 the double containment requirement for plutonium shipments. The section now reads as follows: Shipments containing plutonium must be made with the contents in solid form, if the contents contain greater than 0.74 TBq of plutonium. 10 CFR 71.73(c)(2), “Crush,” was modified by the addition of the following sentence: “For packages containing fissile material, the radioactive contents greater than 1,000 A2 criterion does not apply.” (The pre-October 1, 2004, 10 CFR 71.73 did not require this test for packages with radioactive material content less than or equal to 1,000 A2.) 10 CFR 71.73(c)(2) now reads as follows: Subjection of the specimen to a dynamic crush test by positioning the specimen on a flat, essentially unyielding horizontal surface so as to suffer maximum damage by the drop of a 1,100-lb mass from 30 ft onto the specimen. The mass must consist of a solid mild steel plate 40 inches by 40 inches and must fall in a horizontal attitude. The crush test is required only when the specimen has a mass not greater than 1,100 lbs, an overall density not greater than 62.4 lb/ft3 based on external dimension, and radioactive contents greater than 1,000 A2 not as special form radioactive material. For packages containing fissile material, the radioactive contents greater than 1,000 A2 criterion does not apply. c. Discuss the following related to ANSI N14.5:  Different methods for leak testing a small drum-type package  Standard(s) utilized by DOE for leak tests of type B radioactive materials  Packages and the following containment boundary test requirements for: 1) design verification, 2) fabrication, 3) pre-shipment, 4) periodic, and 5) maintenance Different Methods for Leak Testing The following is taken from NNSA SG-100. ANSI N14.5-1997, Annex A lists eleven accepted leakage test methods with corresponding test sensitivities. Table A.2 of Annex A lists the accepted tests and nominal sensitivities. The table is repeated here as table 11. Summaries of these descriptions are provided after the table.

It is important to design the containment system closures, valves, seals, etc. with the required leakage testing in mind. It should be understood that the reference leakage rate, LR, will likely limit the available test method. For example, if the reference leakage rate is determined to be leak-tight (10-7 reference cubic centimeter per second [ref-cm3/s]), the only leakage rate tests with acceptable sensitivities are gas-filled envelope and evacuated envelope (both techniques use a tracer gas [generally helium] placed on one side of a seal and a vacuum on the other side). In the gas-filled envelope and evacuated envelope techniques, mass spectrometry is used to detect the leakage of the tracer gas. It is also important to note that the current sensitivity of helium mass spectrometry, depending on the method and test conditions, can be orders of magnitude greater than listed in table 11. The other tests described here may be used for packages where less sensitivity is acceptable and for pre-shipment leakage testing, which only requires detection of a leakage rate of 10-3 ref- cm3/s. 59 Table 11. Accepted leakage test and sensitivities Test method Nominal test sensitivitya ref-cm3/s Pa-m3/s Gas pressure dropb,c 10-1 – 10-5 10-2 – 10-6 Gas pressure riseb,c 10-1 – 10-5 10-2 – 10-6 Gas filled envelope (gas detector) 10-3 – 10-8 10-4 – 10-9 Evacuated envelope (gas detector) 10-3 – 10-8 10-4 – 10-9 Evacuated envelope (with back pressurization) 10-3 – 10-8 10-4 – 10-9 Gas bubble techniques 10-3 10-4 Hot water bubbled,e Vacuum bubbled,e 10-3 10-4 Pressurized cavity bubbled,e 10-3 10-4 Soap bubbled,e 10-3 10-4 Tracer gas (sniffer technique) 10-3 – 10-6 10-4 – 10-7 Tracer gas (spray method) 10-3 – 10-6 10-4 – 10-7 aFor comparison purposes only. The listed values are referred to a common standard of dry air at 25oC and a pressure of 1 atm. Bubble test sensitivities are for 1 atm differential pressure except for the hot water bubble, which is for 0.25 atm differential. bSensitivity depends upon volume tested, test time, and instrument sensitivity to pressure differentials. cThe listed sensitivity is typical. The actual sensitivity when applied to a containment system may be calculated using ANSI N14.5-1997, Annex B. dThe listed sensitivity applies to test methods under normal field conditions. Under favorable, well-controlled conditions, this sensitivity could be increased by a factor of 10. eFor tests with a liquid-gas interface, account should be taken of surface tension and the hydrostatic head of the liquid bath (see ANSI N14.5-1997, Annex B). Source: NNSA SG-100 GAS PRESSURE DROP This method involves pressurizing the test item cavity or interspace and measuring the pressure drop. This method is particularly useful for testing double O-ring seals, where the small interspace volume between the seals makes the method most sensitive and the primary seal of the cavity does not have to be broken. The sensitivity of this method is inversely proportional to the test volume. To have a sensitive test using this method, it is critical to accurately measure the volume of the test cavity. Measured leakage rates in pressure drop and pressure rise tests using double O-rings may be actually higher than the actual leakage rate from the containment boundary because the measured leakage rate is across both of the O-ring seals as opposed to only across the containment boundary seal. 60

GAS PRESSURE RISE This method involves evacuating the test cavity and measuring a pressure rise during a specified time period. This method applies to test items with pressure tap connections, but can also be used for testing double O-ring seals. Again, test sensitivity is inversely proportional to test volume, and it is critical for a sensitive test to accurately measure the volume of the test cavity. Measured leakage rates in pressure drop and pressure rise tests using double O-rings may be actually higher than the actual leakage rate from the containment boundary because the measured leakage rate is across both of the O-ring seals as opposed to only across the containment boundary seal. GAS-FILLED ENVELOPE This method involves evacuating the test item connected to a gas detector and surrounding the item in an envelope filled with a tracer gas. This method is suitable for large test items that have a replaceable seal. The detector typically used for this test is a mass spectrometer that has the ability to detect the tracer gas at low concentrations. This test method can have sufficient sensitivity to verify a leak-tight seal. Because the containment void volume is typically evacuated in this test method, the connection point for the vacuum must also be tested. EVACUATED ENVELOPE (GAS DETECTOR) This method involves pressurizing the test item with a tracer gas while the item is placed in a vacuum chamber connected to a gas detector. This method is ideal for small test items that have a replaceable seal. This test method can have sufficient sensitivity to verify a leak-tight seal. This test method is often applied to packages with two concentric seals; the inner seal forms the containment boundary, and the outer seal establishes an interspace for testing. It is important to ensure both seals are not evaluated simultaneously. Additionally, the designer should ensure the two seal grooves communicate in this type configuration. EVACUATED ENVELOPE (WITH BACK PRESSURIZATION) This method involves externally pressurizing the test item in an envelope of the tracer gas for a period of time and subsequently transferring the test item to an evacuated envelope connected to a gas detector. This method is ideal for welded capsules from very small sizes up to the sizes limited by the pressurizing chamber. This method can have sufficient sensitivity to verify a leak- tight seal. HOT WATER BUBBLE This method involves submerging a room-temperature test item in hot water, which raises the internal pressure. A leak is indicated by a stream of bubbles. This method applies to welded capsules and small test items, usually without pressure tap connections. It can be used in the field without sophisticated equipment. This method provides sufficient sensitivity for pre-shipment leakage rate testing where a sensitivity of 1 x 10-3 ref-cm3/s is acceptable. VACUUM BUBBLE This method involves immersing the test item in a liquid and then producing a vacuum above the liquid in which the test item is submerged. A leak is indicated by a stream of bubbles. The method is suitable for welded capsules and small resealable items. This method provides sufficient sensitivity for pre-shipment leakage rate testing where a sensitivity of 1 x 10-3 ref- cm3/s is acceptable. 61

PRESSURIZED CAVITY BUBBLE This method involves pressurizing the test item and immersing it in a liquid. A leak is indicated by a stream of bubbles. This method applies to welded capsules, containers with pressure tap connections, or where the pressure required within the cavity can be obtained by the vaporization of solid carbon dioxide. This method provides sufficient sensitivity for pre-shipment leakage rate testing where a sensitivity of 1 x 10-3 ref cm3/s is acceptable. SOAP BUBBLE This method involves pressurizing the test item and coating the surface with a soap film or other viscous material. A leak is indicated by a soap bubble on the surface. This method applies to containers with pressure tap connections. The required pressure within the cavity can be obtained by the vaporization of solid carbon dioxide. Care must be taken to not exceed the test item design pressure. This method provides sufficient sensitivity for pre-shipment leakage rate testing where a sensitivity of 1 x 10-3 ref cm3/s is acceptable. TRACER GAS (SNIFFER TECHNIQUE) This method involves pressurizing the test item with the tracer gas. A leak is detected by moving a probe across the areas that are likely to leak. This method is best used on large items where the area of potential leak is clearly visible. There must be some facility for pressurizing the inside of the weld or seal. TRACER GAS (SPRAY METHOD) This method involves evacuating a test item connected to a gas and spraying the tracer gas over the surface. This method can be used for testing partially finished vessels provided that one side of a potential leak can be evacuated and the other side is easily accessible with a supply of tracer gas. Standard(s) Utilized by DOE for Leak Tests of Type B Radioactive Material Packages 10 CFR 34.27, “Leak Testing and Replacement of Sealed Sources.” NUREG 1609, Standard Review Plan for Transportation Packages for Radioactive Material, provides guidance for the review and approval of applications for packages used to transport radioactive material under 10 CFR 71. NRC Regulatory Guide 7.4, Leakage Tests on Packages for Shipment of Radioactive Material, endorses the methods and procedures developed by the ANSI Standards Committee on Packaging and Transportation of Radioactive and Nonnuclear Hazardous Materials in ANSI N14.5-1997 issued in 1997 and reaffirmed in 2008, as a process that the NRC staff considers acceptable for meeting the regulatory requirements. The IAEA SSR-6, Regulations for the Safe Transport of Radioactive Material specify permitted release of radioactivity under normal and accidental conditions of transport, in terms of activity per unit of time, for type B packaging used to transport radioactive materials. Generally, it is not practical to measure activity release directly. The usual method used is to relate activity release to nonradioactive fluid leakage, for which several leakage test procedures are available. The appropriate procedure will depend on its sensitivity and its application to a specific package. 62 ISO 12807, Safe Transport of Radioactive Materials—Leakage Testing on Packages, specifies gas leakage test criteria and test methods for demonstrating that packages used to transport radioactive materials comply with the package containment requirements defined in reference of annex F for  design verification  fabrication verification  pre-shipment verification  periodic verification

The regulations specify permissible activity release for normal and accidental conditions of transport. These activity release limits can be expressed in maximum permissible activity release rates for the radioactive material carried within a containment system. In general, it is not feasible to demonstrate that the activity release limits are not exceeded by direct measurement of activity release. In practice, the most common method to prove that a containment system provides adequate containment is to carry out an equivalent gas leakage rate test. Packages, and the Following Containment Boundary Test Requirements for 1) Design Verification. 2) Fabrication, 3) Pre-shipment, 4) Periodic, and 5) Maintenance DESIGN VERIFICATION The following is taken from NNSA SG-100. The selected method for verification may include testing, analysis, or a combination of the two. A detailed verification plan should be developed that addresses the methods of achieving the packaging requirements. The verification plan should include specific tests or analysis plan requirements. The test or analysis plan should then be followed by detailed test or analysis plan procedures. The test or analysis acceptance criteria must be clearly stated in the plans and procedures. The development of the verification plan, and the test or analysis plan and the associated procedures fulfill the requirements of 10 CFR 71.123, “Test Control,” by establishing measures to ensure applicable test programs are accomplished in accordance with written procedures. The verification plan and test or analysis plan identify the appropriate test prerequisites that are in turn properly translated into test procedures. Test results may be presented on calculation data sheets, test results data forms, photographs, videos, computer-generated graphics, plots, and charts. To meet the requirements of 10 CFR 71.123, measures must be established to ensure test results are documented, evaluated, and maintained as QA records. These records must be readily available should questions arise concerning operational aspects of the packages. The acceptability of the records should be determined by a qualified individual or group. Test results should be documented in a formal test report that will be used in the SARP either directly or by reference. 63 FABRICATION The following is taken from NNSA SG-100. Methods for fabricating the packaging will usually depend on the quantity of packages made, schedule for completion, and funding available. Complicated designs may require intricate fabrication techniques. The applicant’s QAP should detail the approach to the control of purchased items and services to fulfill the requirements of 10 CFR 71.115, “Control of Purchased Material, Equipment, and Services,” and 10 CFR 71.109, “Procurement Document Control.” Vendors should be carefully selected based on their capability to comply with applicable sections of 10 CFR 71, Subpart H, their facility and QAP, and their previous records and performance. Vendor evaluations are to be performed before the vendor is released for production. The evaluation of the QAP should be performed by qualified QA personnel from the applicant’s organization prior to initiation of activities affected by the program. Depending on the critical nature of the packaging and the vendor’s performance, the applicant should perform the following activities for each vendor to address the requirements of 10 CFR 71.115(a):  Evaluate and select packaging component vendors  Establish controls to be imposed on these vendors  Perform audits at vendors’ facilities (as required)  Establish conditions for the receiving inspection.

Procurement documents provided to the vendor will supply the design basis technical requirements, including the applicable regulatory requirements; material and component identification requirements; drawings; specifications; applicable codes and standards; special process instructions; and test and inspection requirements for the components to be fabricated. Procurement documents should also identify the records to be retained, controlled, and maintained by the vendor and the records to be delivered to the applicant prior to installation of hardware. These records should include the pertinent documentation to be furnished with the procured materials or services. If the pertinent documentation is in an electronic format, the software system the documentation is to be delivered in should be specified. If the item requires inspection and/or certification by the applicant, then documentation of such inspections and/or certifications should also be entered into the QA records. If the packaging components are manufactured in-house, then similar steps should be conducted for those manufacturing elements. In some cases, to ensure adequate quality, it may be necessary to require that, prior to fabrication, the vendor or in-house shop supply a sample of the component to be fabricated for a first article evaluation (FAE). This requirement should be specified in the procurement documents. The component will be subjected to verification tests and inspections that are specified in the procurement documents to ensure the article was fabricated in accordance with its design and that the manufacturing process was properly performed. If the FAE is acceptable, the vendor will be released to begin production. 64 PRE-SHIPMENT The following is taken from NNSA SG-100. The purpose of packaging pre-shipment leakage rate testing is to confirm the containment system is properly assembled for shipment. Pre-shipment leakage rate testing must be performed before each shipment, after the contents are loaded and the containment system is fully assembled. Pre-shipment leakage rate testing is required only for the containment boundary areas that have been opened during the unload-loading cycle. (Note: NNSA has determined that CV seal (O- ring), CV washer and/or CV fastener replacement is not necessarily a maintenance activity for NNSA certified type B packages.) The acceptance criterion for pre-shipment leakage rate testing shall be the greater of either 1) a leakage rate of not more than the reference air leakage rate, LR, or 2) no detected leakage when tested to a sensitivity of at least 10-3 ref-cm3/s. Pre-shipment leakage rate tests for packages that have non-reusable seals, or have reusable seals that have been replaced, shall demonstrate a leakage rate according to the guidance for maintenance leakage rate testing. PERIODIC The following is taken from NNSA SG-100. The purpose of periodic leakage rate testing is to confirm that the containment capabilities of packagings built to an approved design have not deteriorated during a period of use. Periodic testing must be performed within 12 months prior to the initiation of each shipment. Periodic leakage rate testing need not be performed for out-of-service packagings, but must be performed prior to placing the packaging back in service. Periodic leakage rate testing must be performed for all containment boundary seals, closures, valves, rupture disks, etc. Periodic leakage rate testing does not need to include inaccessible surfaces. If the contribution of the individual component leakage rate to the packaging leakage rate is unknown, the periodic leakage rate testing must be performed for the entire containment boundary.

The acceptance criterion for packaging fabrication leakage rate testing is the reference air leakage rate. MAINTENANCE The following is taken from NNSA SG-100. The purpose of the maintenance leakage rate test is to confirm that any maintenance, repair, or replacement of components has not degraded the containment system. Maintenance leakage rate testing must be performed prior to returning a package to service following maintenance, repair, or replacement of components of a containment system. Maintenance leakage rate testing need only be performed on the affected area of the package. Reuse of components like removable lids, O-rings, and closure bolts is not considered replacement. The containment boundaries of some packaging systems have multiple port openings. The total leakage rate from the package is the sum of the leakage rates from the multiple openings. If the contribution of the individual component leakage rate to the packaging 65 leakage rate is unknown, the maintenance leakage rate testing must be performed for the entire containment boundary. The acceptance criterion for packaging fabrication leakage rate testing is the reference air leakage rate. d. Discuss the following:  The definition of A1 and A2 and the units commonly used  How to calculate A2 for a mixture of radionuclides, and how to calculate the number of A2s in the package contents using the guidance in 10 CFR 71  The definition of a type A and type B package relative to the quantity of the content  The definition of exclusive use and highway route control, and how these declarations alter package transport, whether by the DOE or for the DOE by a commercial carrier  The definition of normal, depleted, and enriched uranium based on isotopic content  The definition of weapons grade, reactor grade, and heat source grade plutonium The Definition of A1 and A2 and the Units Commonly Used The following is taken from NNSA SG-100. A1 is the maximum activity of special form radioactive material permitted in a type A package. This value is either listed in 10 CFR 71, table A-1, or can be derived in accordance with the procedures prescribed in 10 CFR 71, Appendix A. A2 is the maximum activity of radioactive material, other than special form radioactive material, permitted in a type A package. This value is either listed in 10 CFR 71, table A-1, or can be derived in accordance with the procedures prescribed in 10 CFR 71, Appendix A. The following is taken from 10 CFR 71, Appendix A. The Ci values specified are obtained by converting from the TBq value. The TBq values are the regulatory standard. The Ci values are for information only and are not intended to be the regulatory standard. Where values of A1 and A2 are unlimited, it is for radiation control purposes only. For nuclear criticality safety, some materials are subject to controls placed on fissile material. For mixtures of radionuclides whose identities and respective activities are known, the following conditions apply: a. For special form radioactive material, the maximum quantity transported in a type A package is as follows: ∑ l 1 )i(A )i(B ≤1 66 b. For normal form radioactive material, the maximum quantity transported in a type A package is as follows: ∑B(i)/A2 (i) ≤ 1 where B(i) is the activity of radionuclide i, and A2(i) is the A2 value for radionuclide i. c. Alternatively, the A1 value for mixtures of special form material may be determined as follows: A1 for mixture = ∑ l 1 )i(A

)i(f 1 where f(i) is the fraction of activity for radionuclide I in the mixture, and A2 (i) is the appropriate A2 value for radionuclide I. d. Alternatively, the A2 value for mixtures of normal form material may be determined as follows: A2 for mixture = ∑ l 2 )i(A )i(f 1 where f(i) is the fraction of activity for radionuclide I in the mixture, and A2(i) is the appropriate A2 value for radionuclide I. e. The exempt activity concentration for mixtures of nuclides may be determined as follows: Exempt activity concentration for mixture = [ ]∑ l )i(A )i(f 1 where f(i) is the fraction of activity concentration of radionuclide I in the mixture, and [A] is the activity concentration for exempt material containing radionuclide I. f. The activity limit for an exempt consignment for mixtures of radionuclides may be determined as follows: Exempt consignment activity limit for mixture = ∑ l )i(A )i(f 1 where f(i) is the fraction of activity of radionuclide I in the mixture, and A is the activity limit for exempt consignments for radionuclide I. When the identity of each radionuclide is known, but the individual activities of some of the radionuclides are not known, the radionuclides may be grou