Practical Thermal Evaluation Methods for HAC Fire Analysis in Type B Radioactive Material (RAM) Packages

Author(s):  
Narendra K. Gupta ◽  
Stephen J. Hensel ◽  
Glenn Abramczyk

Title 10 of the United States Code of Federal Regulations Part 71 for the Nuclear Regulatory Commission (10 CFR Part 71.73[1]) requires that Type B radioactive material (RAM) packages satisfy certain Hypothetical Accident Conditions (HAC) thermal design requirements to ensure package safety during accidental fire conditions. Compliance with thermal design requirements can be met by prototype tests, analyses only or a combination of tests and analyses. Normally, it is impractical to meet all the HAC using tests only and the analytical methods are too complex due to the multi-physics non-linear nature of the fire event. Therefore, a combination of tests and thermal analyses methods using commercial heat transfer software are used to meet the necessary design requirements. The authors, along with his other colleagues at Savannah River National Laboratory in Aiken, SC, USA, have successfully used this ‘tests and analyses’ approach in the design and certification of several United States’ DOE/NNSA certified packages, e.g. 9975, 9977, 9978, 9979, H1700, and Bulk Tritium Shipping Package (BTSP). This paper will describe these methods and it is hoped that the RAM Type B package designers and analysts can use them for their applications.

Author(s):  
Ronald S. Hafner ◽  
Gerald C. Mok ◽  
Lisle G. Hagler

The U.S. Nuclear Regulatory Commission (USNRC) contracted with the Packaging Review Group (PRG) at Lawrence Livermore National Laboratory (LLNL) to conduct a single, 30-ft shallow-angle drop test on the Combustion Engineering ABB-2901 drum-type shipping package. The purpose of the test was to determine if bolted-ring drum closures could fail during shallow-angle drops. The PRG at LLNL planned the test, and Defense Technologies Engineering Division (DTED) personnel from LLNL’s Site-300 Test Group executed the plan. The test was conducted in November 2001 using the drop-tower facility at LLNL’s Site 300. Two representatives from Westinghouse Electric Company in Columbia, South Carolina (WEC-SC); two USNRC staff members; and three PRG members from LLNL witnessed the preliminary test runs and the final test. The single test clearly demonstrated the vulnerability of the bolted-ring drum closure to shallow-angle drops—the test package’s drum closure was easily and totally separated from the drum package. The results of the preliminary test runs and the 30-ft shallow-angle drop test offer valuable qualitative understandings of the shallow-angle impact. • A drum package with a bolted-ring closure may be vulnerable to closure failure by the shallow-angle drop, even if results of the steep-angle drop demonstrate that the package is resistant to similar damage. • Although there exist other mechanisms, the shallow-angle drop produces closure failure mainly by buckling the drum lid and separating the drum lid and body, which the bolted ring cannot prevent. • Since the closure failure by the shallow-angle drop is generated mainly by structural instabilities of a highly discontinuous joint, the phenomenon can be rather unpredictable. Thus, a larger-than-normal margin of safety is recommended for the design of such packages. • The structural integrity of the bolted-ring drum closure design depends on a number of factors. To ensure that the drum closure survives the shallow-angle drop, the following general qualitative rules should be observed: – The drum closure components should be quality products made of ductile materials, and the torque value for tightening the bolted ring should be included in the SAR and operating procedures to ensure quality. – The package should not be too heavy. – The package internal structure should be impact-absorbent and resistant to disintegration and collapse under high compressive load. However, a strong internal structure may defeat the purpose of protecting the containment vessel from damage during a free drop. • If not previously tested, drum packages with bolted-ring drum closures should be drop-tested at shallow angles. Due to the unpredictable nature of the behavior, the demonstration should be completed by test and on a case-by-case basis. The test plan should take into account the behavior’s sensitivity to the details of the package design and the impact condition. • Because the shallow-angle drop can open the drum closure, organizations using these types of drum packages should assess the consequences of exposing the radioactive contents in the containment vessel to unconsidered external elements or conditions. This work was supported by the United States Nuclear Regulatory Commission under a Memorandum of Understanding with the United States Department of Energy, and performed under the auspices of the U.S. Department of Energy by Lawrence Livermore National Laboratory under Contract W-7405-Eng-48.


Author(s):  
Andrew Celovsky ◽  
Randy Lesco ◽  
Brian Gale ◽  
Jeffrey Sypes

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.


PEDIATRICS ◽  
1990 ◽  
Vol 85 (4) ◽  
pp. 698-704
Author(s):  
Sunil K. Sood ◽  
Robert S. Daum

Several Haemophilus influenzae type b vaccines have been licensed and recommended for administration to children in the United States. These vaccines have consisted of purified polyribosylribitol-phosphate (PRP), the capsular polysaccharide of H influenzae type b,1 alone or covalently bound to one of several carrier proteins. Two of these saccharide-protein conjugate vaccines are now licensed, a polysaccharide-diphtheria toxoid conjugate (PRP-D)2 and an oligosaccharide-mutant diphtheria toxin conjugate (HbOC).3 Two others, a polysaccharide- Neisseria meningitidis outer membrane protein conjugate (PRP-OMPC)4 and a polysaccharide-tetanus toxoid conjugate (PRP-T),5 are currently in clinical trials. One concern with the use of PRP vaccine was the suggestion that the incidence of invasive disease caused by H influenzae type b in the immediate period after immunization might be increased; this idea was supported by evidence from several sources. In a case-control study of the efficacy of PRP vaccine, Black et al6 found that 4 children were hospitalized for invasive disease within 1 week of immunization, a rate of invasive disease 6.4 times greater (95% confidence interval [CI], 2.1 to 19.2) than the background rate in unvaccinated children. In Minnesota, the relative risk for invasive disease in the first week after immunization was 6.2 (95% CI, 0.6 to 45.9),7 and the results of a study conducted by the Centers for Disease Control in six areas of the United States revealed a 1.8-fold (95% CI, 0.3 to 10.2) increase in the occurrence of invasive disease caused by H influenzae type b in the first week after immunization.8 Moreover, among 16 cases of disease caused by H influenzae type b occurring within 14 days of immunization that were passively reported to the FDA,9 10 were clustered within the first 72 hours.


Author(s):  
Terry L. Dickson ◽  
Shah N. Malik ◽  
Mark T. Kirk ◽  
Deborah A. Jackson

The current federal regulations to ensure that nuclear reactor pressure vessels (RPVs) maintain their structural integrity when subjected to transients such as pressurized thermal shock (PTS) events were derived from computational models that were developed in the early to mid 1980s. Since that time, there have been advancements in relevant technologies associated with the physics of PTS events that impact RPV integrity assessment. Preliminary studies performed in 1999 suggested that application of the improved technology could reduce the conservatism in the current regulations while continuing to provide reasonable assurance of adequate protection to public health and safety. A relaxation of PTS regulations could have profound implications for plant license extension considerations. Based on the above, in 1999, the United States Nuclear Regulatory Commission (USNRC) initiated a comprehensive project, with the nuclear power industry as a participant, to re-evaluate the current PTS regulations within the framework established by modern probabilistic risk assessment (PRA) techniques. During the last three years, improved computational models have evolved through interactions between experts in the relevant disciplines of thermal hydraulics, PRA, human reliability analysis (HRA), materials embrittlement effects on fracture toughness (crack initiation and arrest), fracture mechanics methodology, and fabrication-induced flaw characterization. These experts were from the NRC staff, their contractors, and representatives from the nuclear industry. These improved models have now been implemented into the FAVOR (Fracture Analysis of Vessels: Oak Ridge) computer code, which is an applications tool for performing risk-informed structural integrity evaluations of embrittled RPVs subjected to transient thermal-hydraulic loading conditions. The baseline version of FAVOR (version 1.0) was released in October 2001. The updated risk-informed computational methodology in the FAVOR code is currently being applied to selected domestic commercial pressurized water reactors to evaluate the adequacy of the current regulations and to determine whether a technical basis can be established to support a relaxation of the current regulations. This paper provides a status report on the application of the updated computational methodology to a commercial pressurized water reactor (PWR) and discusses the results and interpretation of those results. It is anticipated that this re-evaluation effort will be completed in 2002.


Author(s):  
Michael Needham

Why is the detection of radioactive sources important to the solid waste industry?: Radioactive material is used extensively in the United States in research, medicine, education, and industry for the benefit of society (e.g. smoke detectors, industrial process gauges, medical diagnosis/treatment). Generally speaking, the Nuclear Regulatory Commission and state governments regulate the use and disposal of radioactive materials. Licensed radioactive waste disposal facilities receive the bulk of the waste generated in the United States with exceptions for low-level waste (e.g. medical patient waste) that may be disposed of as municipal waste. According to the Conference of Radiation Control Program Directors, Inc (CRCPD)., there has been an increasing number of incidence involving the detection of prohibited radioactive wastes at solid waste management facilities. While the CRCPD acknowledges that the increased incidence may be partially attributed to the growing number of solid waste facilities that have detection systems, undetected sources of ionizing radiation can harm the environment, have a negative impact on employee health and safety, and result in significant remedial actions. Implementing an effective detection/response plan can aid in the proper management of radioactive waste and serve to minimize the potential for negative outcomes.


Author(s):  
Russell Wagner

The U.S. Nuclear Regulatory Commission (NRC) has provided set guidance that hydrogen concentrations in radioactive material packages be limited to 5 vol% unless the package is designed to withstand a bounding hydrogen deflagration or detonation. The NRC guidance further specifies that the expected shipping time for a package be limited to one-half the time to reach 5 vol% hydrogen. This guidance has presented logistical problems for transport of retrieved legacy waste packages on the Department of Energy (DOE) Hanford Site that frequently contain greater than 5 vol% hydrogen due to their age and the lack of venting requirements at the time they were generated. Such packages do not meet the performance-based criteria for Type B packaging, and are considered risk-based packages. Duratek Technical Services (Duratek) has researched the true risk of hydrogen deflagration and detonation with closed packages, and has developed technical justification for elevated concentration limits of up to 15 vol% hydrogen in risk-based packages when transport is limited to the confines of the Hanford Site. Duratek has presented elevated hydrogen limit justification to the DOE Richland Operations Office and is awaiting approval for incorporation into the Hanford Site Transportation Safety Document. This paper details the technical justification methodology for the elevated hydrogen limits.


Author(s):  
John Minichiello ◽  
Ernest B. Branch ◽  
Timothy M. Adams ◽  
Yasuhide Asada ◽  
Richard W. Barnes

The new rules for seismic piping design in Section III that were developed and included in the requirements in 1994 Addenda of the ASME Boiler and Pressure Vessel Code (B&PV Code) generated considerable discussion within the industry and from the United States Nuclear Regulatory Commission, (USNRC). The USNRC initiated a review of the results of the previous EPRI/NRC experimental program and the Japanese industry started its own experimental program. To accommodate and address developments resulting from these efforts, the ASME, B&PV Code established a Special Working Group (SWG) to continue the review and study of the questions and information generated. This paper reports on the efforts of this SWG which resulted in refinements of the revised rules. These refinements have been accepted for inclusion in Section III of the ASME, B&PV Code.


Author(s):  
Terry Dickson ◽  
Shengjun Yin ◽  
Mark Kirk ◽  
Hsuing-Wei Chou

As a result of a multi-year, multi-disciplinary effort on the part of the United States Nuclear Regulatory Commission (USNRC), its contractors, and the nuclear industry, a technical basis has been established to support a risk-informed revision to pressurized thermal shock (PTS) regulations originally promulgated in the mid-1980s. The revised regulations provide alternative (optional) reference-temperature (RT)-based screening criteria, which is codified in 10 CFR 50.61(a). How the revised screening criteria were determined from the results of the probabilistic fracture mechanics (PFM) analyses will be discussed in this paper.


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