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

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.

scholarly journals Radioactive Material Transportation Considerations With Respect to DOE 3013 Storage Containers

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.

scholarly journals Definition of “Small Gram Quantity Contents” for Type B Radioactive Material Transportation Packages: Activity-Based Content Limitations

2011 ◽  
Author(s):  
Shiva Sitaraman ◽  
Soon Kim ◽  
Debdas Biswas ◽  
Ronald Hafner ◽  
Brian Anderson
Keyword(s):  
Dose Rate ◽  
Radioactive Material ◽  
Packaging Materials ◽  
Type B ◽  
Mass Limit ◽  
Sphere Model ◽  
External Dose ◽  
Calculated Mass ◽  
Material Transportation ◽  
Definition Of

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.

CASTOR® BR3: Design of a Transport and Storage Cask for Spent Fuel From a Belgian Nuclear Power Plant

2001 ◽  
Author(s):  
Ulrich Knopp
Keyword(s):  
Nuclear Power ◽  
Storage Period ◽  
Nuclear Research ◽  
Ductile Cast Iron ◽  
Radioactive Material ◽  
Research Centre ◽  
Spent Fuel ◽  
Interim Storage ◽  
And Storage

Abstract The CASTOR® BR3 cask has been designed and manufactured to accomodate irradiated fuel (U and MOX) from the BR3 test reactor at the nuclear research centre SCK/CEN in Dessel near Mol, Belgium, which is currently being dismantled. The CASTOR® BR3 is designed as a Type B(U)F package for transport and will be licensed in Belgium. In addition, the CASTOR® BR3 needs a license as a storage cask to be operated in an interim cask storage facility. To obtain these licenses, the cask design has to observe the international regulations for the safe transport of radioactive material as well as the special requirements for the cask storage. The CASTOR® BR3 is a member of the CASTOR® family of spent fuel casks, delivered by the German company GNB. In this way, the cask has such typical features as the following: • monolithic cask body made of ductile cast iron; • double-lid system consisting of primary and secondary lid for long-term interim storage of the fuel. This family of casks has been used for over 20 years for transport and storage of spent fuel. In this paper, the IAEA regulatory requirements for transport casks are summarized and it is shown by selected examples how these requirements have been converted into the cask design and the analyses performed for the cask. Finally, the cask features for an interim storage period of up to 50 years will be spotlighted. Main topics are the evaluation of the long term behaviour of selected cask components and the cask monitoring system for the surveillance of the leak tightness of the cask during the storage period.

Aging Performance of Viton® GLT O-Rings in Radioactive Material Packages (U)

2007 ◽  
Author(s):  
T. E. Skidmore ◽  
K. M. Counts ◽  
E. B. Fox ◽  
E. N. Hoffman ◽  
K. A. Dunn
Keyword(s):  
Storage Conditions ◽  
Radioactive Material ◽  
Test Results ◽  
Savannah River ◽  
Bearing Materials ◽  
Materials Used ◽  
Interim Storage ◽  
River Site ◽  
Term Storage

Radioactive material packages used for transportation of plutonium-bearing materials often contain multiple O-ring seals for containment. Packages such as the Model 9975 are also being used for interim storage of Pu-bearing materials at the Savannah River Site (SRS). One of the seal materials used in such packages is Viton® GLT fluoroelastomer. The aging behavior of containment vessel O-rings based on Viton® GLT at long-term containment term storage conditions is being characterized to assess its performance in such applications. This paper summarizes the program and test results to date.

Safety Aspects of Long-term Dry Interim Storage Of Type-b Spent Fuel and Hlw Transport Casks

2004 ◽  
Vol 15 (3-4) ◽  
pp. 207-214 ◽  
Author(s):  
D. Wolff ◽  
U. Probst ◽  
H. Völzke ◽  
B. Droste ◽  
R. Rödel
Keyword(s):  
Type B ◽  
Spent Fuel ◽  
Safety Aspects ◽  
Interim Storage

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

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.

Certification of SAFESHIELD 2999A

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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