Definition of "Small Gram Quantity Contents" for Type B Radioactive Material Transportation Packages: Activity-Based Content Limitations, Rev. 1

This paper presents a compendium of allowable masses for a variety of gamma and neutron emitting isotopes (with varying impurity levels of beryllium in some of the actinide isotopes) that, when loaded in an unshielded radioactive material transportation packaging, do not result in an external dose rate on the surface of the package that exceeds 190 mrem/hr (190 mrem/hr was chosen to provide 5% conservatism relative to the regulatory limit). These mass limits define the term “Small Gram Quantity” (SGQ) contents in the context of radioactive material transportation packages. The term SGQ is isotope-specific and pertains to contents in radioactive material transportation packages that do not require shielding and still satisfy the external dose rate requirements. Since these calculated mass limits are for contents without shielding, they are conservative for packaging materials that provide some limited shielding or if the contents are placed into a shielded package. Two sets of mass limit results are presented: (1) mass limits calculated with a “voided sphere” model, and (2) mass limits calculated with the unshielded radioactive material transportation packaging Model 9977-96.

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Radioactive Material Transportation Considerations With Respect to DOE 3013 Storage Containers

Transportation, Storage, and Disposal of Radioactive Materials ◽

10.1115/pvp2004-2783 ◽

2004 ◽

Author(s):

S. J. Hensel ◽

T. T. Wu ◽

B. R. Seward

Keyword(s):

Nuclear Regulatory Commission ◽

Storage Conditions ◽

Radioactive Material ◽

Department Of Transportation ◽

Type B ◽

Long Term Storage ◽

Leak Tightness ◽

Material Transportation ◽

Regulatory Commission

This paper evaluates sealed hardware that meets the requirements of DOE-STD-3013, “Criteria for Preparing and packaging Plutonium Metals and Oxides for Long-Term Storage” [1] with respect to radioactive material (Type B quantity) transportation requirements. The Standard provides criteria for packaging of the plutonium materials for storage periods of at least 50 years. The standard requires the hardware to maintain integrity under both normal storage conditions and under anticipated handling conditions. To accomplish this, the standard requires that the plutonium be loaded in a minimum of two nested stainless steel sealed containers that are both tested for leak-tightness per ANSI N14.5. As such the 3013 hardware is robust. While the 3013 STD may provide appropriate storage criteria, it is not intended to provide criteria for transporting the material under the requirements of the Department of Transportation (DOT). In this evaluation, it is assumed that the activity of plutonium exceeds A1 and/or A2 curies as defined in DOT 49 CFR 173.431 and therefore must be shipped as a Type B package meeting the Nuclear Regulatory Commission (NRC) requirements of 10 CFR 71. The evaluation considers Type B shipment of plutonium in the 3013 hardware within a certified package for such contents.

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Extending Periodic Leakage Rate Testing of 9977 Packages With Elastomeric O-Rings

Volume 7: Operations, Applications and Components ◽

10.1115/pvp2014-28950 ◽

2014 ◽

Author(s):

Zenghu Han ◽

Vikram N. Shah ◽

Yung Y. Liu

Keyword(s):

Leakage Rate ◽

Radioactive Material ◽

Ring Test ◽

Type B ◽

Testing Interval ◽

Rate Testing ◽

Arrhenius Function ◽

Material Transportation ◽

Interim Storage

According to ANSI N14.5, the periodic leakage rate testing of Type B radioactive material transportation packages is performed within 12 months prior to each shipment. The purpose of performing periodic leakage rate testing is to confirm that packages built to an approved design can perform their containment function as required after a period of use. However, certain transportation packages, e.g., Model 9975 and 9977 Type B packages, have been used for interim storage for a period > 12 months, and it is desirable to extend the periodic leakage rate testing interval to reduce personnel radiation exposure and cost. Long-term leak performance tests on O-ring test fixtures have been conducted at 200°F (366K) and higher temperatures since 2004 for the purpose of interim storage of 9975 packages. The test data are adopted and evaluated in this paper by using the Arrhenius function and the Weibull statistics to establish the basis for extending the periodic leakage rate testing interval. The results show that the testing interval can be extended to 5 and 2 years for Model 9977 packages with Viton® GLT and GLT-S elastomeric O-rings (Parker Seals V0835-75 and VM835-75), respectively, if the O-ring service temperature is kept below 200°F (366K) and verified with continuous temperature monitoring.

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A Comparison of 10 CFR 71 and the IAEA TS-R-1 Radioactive Material Transportation Regulations

Volume 7: Operations, Applications and Components ◽

10.1115/pvp2008-61593 ◽

2008 ◽

Author(s):

Christopher S. Bajwa ◽

Earl P. Easton ◽

Nancy L. Osgood ◽

Ronald B. Pope

Keyword(s):

International Atomic Energy Agency ◽

Atomic Energy ◽

Radioactive Material ◽

Type B ◽

Radioactive Materials ◽

Energy Agency ◽

National Legislation ◽

The Us ◽

Material Transportation ◽

Safe Transport

The US regulations for certification of Type B packages are based in large part on those of the International Atomic Energy Agency (IAEA); however, the US has chosen to differ (or deviate) in some respects, from the IAEA regulations based on its national legislation, its technical experience, and efforts to minimize burden on shippers of radioactive materials in the US. This paper will provide a brief overview of some of the differences between 10 CFR Part 71 “Packaging and Transportation of Radioactive Material”, as implemented January 2008, and IAEA TS-R-1 “Regulations for the Safe Transport of Radioactive Material”, 2005 edition, discuss some of the differences between the two sets of regulations, and the reasons for those differences.

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Certification of SAFESHIELD 2999A

Volume 7: Operations, Applications, and Components ◽

10.1115/pvp2006-icpvt-11-93117 ◽

2006 ◽

Author(s):

Yung Liu ◽

Steve Bellamy ◽

James Shuler

Keyword(s):

Lawrence Livermore National Laboratory ◽

National Laboratory ◽

Argonne National Laboratory ◽

Department Of Energy ◽

Radioactive Material ◽

Fissile Material ◽

Type B ◽

Technical Issues ◽

Material Transportation ◽

The U.S

Based on the U.S. Department of Transportation regulations in 49 CFR 173.7(d), the U.S. Department of Energy (DOE) Order 460.1B codifies the authority of certification of Type-B and fissile material transportation packaging to the Office of Environmental Management (EM), except for materials of interest to national security, naval propulsion systems, and civilian radioactive waste management. DOE Order 460.1B also stipulates that the EM certification of Type B and fissile materials transportation packaging shall be in accordance with the U.S. Nuclear Regulatory Commission safety standards in 10 CFR Part 71. The Office of Licensing (EM-24) is supported by technical review teams at Argonne National Laboratory, Lawrence Livermore National Laboratory, and Savannah River National Laboratory. SAFESHIELD 2999A is a Type-B radioactive material transportation packaging designed for use by the DOE’s Isotope Program. The contents of the packaging consist of source capsules of Co-60, Cs-137, or Ir-192; solid and liquid-metal accelerator targets; ion exchange resins; and target processing wastes. No fissile materials are included. Protection against radiation and confinement of radioactivity are, therefore, the two major requirements for the safety performance of the SAFESHIELD 2999A packaging under both normal conditions of transport and hypothetical accidents. The Safety Analysis Report for Packaging (SARP) of SAFESHIELD 2999A underwent four revisions by the applicant during the entire certification review. This paper will highlight some of the technical issues in areas such as contents, shielding, and quality assurance, and will discuss how these issues interact and affect other areas such as structural, thermal, containment, operating procedures, and acceptance tests and maintenance. Also discussed in the paper is the use of an independent third party to facilitate resolution of the technical issues and move the process forward for certification of SAFESHIELD 2999A.

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A Review of a Radioactive Material Shipping Container Including Design, Testing, Upgrading Compliance Program and Shipping Logistics

Transportation, Storage, and Disposal of Radioactive Materials ◽

10.1115/pvp2003-2134 ◽

2003 ◽

Author(s):

Andrew Celovsky ◽

Randy Lesco ◽

Brian Gale ◽

Jeffrey Sypes

Keyword(s):

Radioactive Material ◽

Type B ◽

Design Elements ◽

Shipping Container ◽

Global Transport ◽

Leak Tightness ◽

Design Requirements ◽

Compliance Program ◽

Integrity Tests ◽

Key Aspects

Ten years ago Atomic Energy of Canada developed a Type B(U)-85 shipping container for the global transport of highly radioactive materials. This paper reviews the development of the container, including a summary of the design requirements, a review of the selected materials and key design elements, and the results of the major qualification tests (drop testing, fire test, leak tightness testing, and shielding integrity tests). As a result of the testing, improvements to the structural, thermal and containment design were made. Such improvements, and reasons thereof, are noted. Also provided is a summary of the additional analysis work required to upgrade the package from a Type B(U) to a Type B(F), i.e. essentially upgrading the container to include fissile radioisotopes to the authorized radioactive contents list. Having a certified shipping container is only one aspect governing the global shipments of radioactive material. By necessity the shipment of radioactive material is a highly regulated environment. This paper also explores the experiences with other key aspects of radioactive shipments, including the service procedures used to maintain the container certification, the associated compliance program for radioactive material shipments, and the shipping logistics involved in the transport.

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Design and Analysis of Shipping Container for Big Rock Decommissioned Reactor Vessel

Volume 7: Operations, Applications and Components ◽

10.1115/pvp2014-28798 ◽

2014 ◽

Author(s):

Bruce (Bart) Slimp ◽

Mick Papp ◽

Phuong H. Hoang

Keyword(s):

Structural Analysis ◽

Base Material ◽

Reactor Vessel ◽

Radioactive Material ◽

Type B ◽

Shipping Container ◽

Analysis Methodology ◽

Maximum Wall Thickness ◽

Hypothetical Accident ◽

Accident Conditions

A major milestone in 2003 on the Big Rock Point (BRP) decommissioning project involved shipping the Reactor Vessel (RV) in a steel cask for burial. The Reactor Vessel Transport System (RVTS) cask was a sealed integral container, which provided necessary radiological shielding and containment of radioactive waste for shipping and disposal. The RVTS, using the provisions of the ASME BPVC Section III, Subsection NB, was designed as a Type B package in accordance with the requirements of 10 CFR Part 71. This included meeting Normal Condition of Transport (NCT) and the Hypothetical Accident Conditions (HAC) loading per 10 CFR 71, Regulatory Guide 7.6, “Design Criteria for the Structural Analysis of Shipping Cask Containment Vessels,” Regulatory Guide 7.8, “Load Combinations for the Structural Analysis of Shipping Casks for Radioactive Material” and 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.” The RVTS was designed to withstand accelerations and shocks postulated during highway and rail transit using guidelines from the Association of American Railroads (AAR) and ANSI N14.2. The design analysis methodology, fabrication process and transportation planning for the Big Rock RVTS Cask are presented in this paper.

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Implementation of Decommissioning Materials Conditional Clearance Process to the OMEGA Calculation Code

ASME 2010 13th International Conference on Environmental Remediation and Radioactive Waste Management, Volume 1 ◽

10.1115/icem2010-40120 ◽

2010 ◽

Author(s):

Matej Zachar ◽

Vladimi´r Danisˇka ◽

Vladimi´r Necˇas

Keyword(s):

Radioactive Material ◽

Model Assessment ◽

Exposure Limits ◽

Conditional Release ◽

Release Process ◽

Model Calculations ◽

Nuclear Installation ◽

Clearance Level ◽

Definition Of ◽

New Procedures

The activities performed during nuclear installation decommissioning process inevitably lead to the production of large amount of radioactive material to be managed. Significant part of materials has such low radioactivity level that allows them to be released to the environment without any restriction for further use. On the other hand, for materials with radioactivity slightly above the defined unconditional clearance level, there is a possibility to release them conditionally for a specific purpose in accordance with developed scenario assuring that radiation exposure limits for population not to be exceeded. The procedure of managing such decommissioning materials, mentioned above, could lead to recycling and reuse of more solid materials and to save the radioactive waste repository volume. In the paper an implementation of the process of conditional release to the OMEGA Code is analyzed in details; the Code is used for calculation of decommissioning parameters. The analytical approach in the material parameters assessment, firstly, assumes a definition of radiological limit conditions, based on the evaluation of possible scenarios for conditionally released materials, and their application to appropriate sorter type in existing material and radioactivity flow system. Other calculation procedures with relevant technological or economical parameters, mathematically describing e.g. final radiation monitoring or transport outside the locality, are applied to the OMEGA Code in the next step. Together with limits, new procedures creating independent material stream allow evaluation of conditional material release process during decommissioning. Model calculations evaluating various scenarios with different input parameters and considering conditional release of materials to the environment are performed to verify the implemented methodology. Output parameters and results of the model assessment are presented, discussed and concluded in the final part of the paper.

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Licensing of the ES-3100 Type B Shipping Container: Replacement of the DOT 6M Container

Volume 7: Operations, Applications, and Components ◽

10.1115/pvp2005-71601 ◽

2005 ◽

Author(s):

Jeffrey G. Arbital ◽

Dean R. Tousley ◽

James C. Anderson

Keyword(s):

State Of The Art ◽

Nuclear Regulatory Commission ◽

Department Of Energy ◽

Radioactive Material ◽

Shipping Container ◽

Nuclear Security ◽

Material Transportation ◽

Enriched Uranium ◽

Regulatory Commission ◽

The U.S

The National Nuclear Security Administration (NNSA) is shipping bulk quantities of fissile materials for disposition purposes, primarily highly enriched uranium (HEU), over the next 15 to 20 years. The U.S. Department of Transportation (DOT) specification 6M container has been the workhorse for NNSA and many other shippers of radioactive material. However, the 6M does not conform to the safety requirements in the Code of Federal Regulations (10 CFR 71[1]) and, for that reason, is being phased out for use in the secure transportation system of the U.S. Department of Energy (DOE) in early 2006. BWXT Y-12 is currently developing the replacement for the DOT 6M container for NNSA and other users. The new package is based on state-of-the-art, proven, and patented technologies that have been successfully applied in the design of other packages. The new package will have a 50% greater capacity for HEU than the 6M, and it will be easier to use with a state-of-the-art closure system on the containment vessel. This new package is extremely important to the future of fissile, radioactive material transportation. An application for license was submitted to the U.S. Nuclear Regulatory Commission (NRC) in February 2005. This paper reviews the license submittal, the licensing process, and the proposed contents of this new state-of-the-art shipping container.

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