12 research outputs found
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Technical Review Report for the Safety Analysis Report for Packaging Model 9977 S-SARP-G-00001 Revision 2
This Technical Review Report (TRR) summarizes the review findings for the Safety Analysis Report for Packaging (SARP) for the Model 9977 B(M)F-96 shipping container. The content analyzed for this submittal is Content Envelope C.1, Heat Sources, in assemblies of Radioisotope Thermoelectric Generators or food-pack cans. The SARP under review, i.e., S-SARP-G-00001, Revision 2 (August 2007), was originally referred to as the General Purpose Fissile Material Package. The review presented in this TRR was performed using the methods outlined in Revision 3 of the Department of Energy's (DOE's) Packaging Review Guide (PRG) for Reviewing Safety Analysis Reports for Packages. The format of the SARP follows that specified in Revision 2 of the Nuclear Regulatory Commission's, Regulatory Guide 7.9, i.e., Standard Format and Content of Part 71 Applications for Approval of Packages for Radioactive Material. Although the two documents are similar in their content, they are not identical. Formatting differences have been noted in this TRR, where appropriate. The Model 9977 Package is a 35-gallon drum package design that has evolved from a family of packages designed by DOE contractors at the Savannah River Site. The Model 9977 Package design includes a single, 6-inch diameter, stainless steel pressure vessel containment system (i.e., the 6CV) that was designed and fabricated in accordance with Section III, Subsection NB, of the American Society of Mechanical Engineers Boiler & Pressure Vessel Code. The earlier package designs, i.e., the Model 9965, 9966, 9967 and 9968 Packages, were originally designed and certified in the 1980s. In the 1990s, updated package designs that incorporated design features consistent with new safety requirements, based on International Atomic Energy Agency guidelines, were proposed. The updated package designs were the Model 9972, 9973, 9974 and 9975 Packages, respectively. The Model 9975 Package was certified by the Packaging Certification Program, under the Office of Safety Management and Operations. Differences between the Model 9975 Package and the Model 9977 Package include: (1) The lead shield present in the Model 9975 Package is absent in the Model 9977 Package; (2) The Model 9975 Package has eight allowable contents, while the Model 9977 Package has a single allowable content. (3) The 6CV of the Model 9977 Package is similar in design to the outer Containment Vessel of the Model 9975 Package that also incorporates a 5-inch Containment Vessel as the inner Containment Vessel. (4) The Model 9975 Package uses a Celotex{reg_sign}-based impact limiter while the Model 9977 Package uses Last-A-Foam{reg_sign}, a polyurethane foam, for the impact limiter. (5) The Model 9975 Package has two Containment Vessels, while the Model 9977 Package has a single Containment Vessel
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Technical Review Report for the Mound 1KW Package Safety Analysis Report for Packaging Addendum No. 1, through Revision b
This Technical Review Report (TRR) documents the review, performed by the Lawrence Livermore National Laboratory (LLNL) staff, at the request of the U.S. Department of Energy (DOE), on the 'Mound 1KW Package Safety Analysis Report for Packaging, Addendum No. 1, Revision b', dated May 2007 (Addendum 1). The Mound 1KW Package is certified by DOE Certificate of Compliance (CoC) number USA/9516/B(U)F-85 for the transportation of Type B quantities of plutonium heat source material. The safety analysis of the package is documented in the 'Safety Analysis Report for Packaging (SARP) for the Mound 1KW Package' (i.e., the Mound 1KW SARP, or the SARP). Addendum 1 incorporates a new fueled capsule assembly payload. The following changes have been made to add this payload: (1) The primary containment vessel (PCV) will be of the same design, but will increase in height to 11.16 in.; (2) A new graphite support block will be added to support up to three fueled capsule assemblies per package; (3) The cutting groove height on the secondary containment vessel (SCV) will be heightened to accommodate the taller PCV; and (4) A 3.38 in. high graphite filler block will be placed on top of the PCV. All other packaging features, as described in the Mound 1KW SARP [3], remain unchanged. This report documents the LLNL review of Addendum 1[1]. The specific review for each SARP Chapter is documented herein
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Plutonium Disposition by Immobilization
The ultimate goal of the Department of Energy (DOE) Immobilization Project is to develop, construct, and operate facilities that will immobilize between 17 to 50 tonnes (MT) of U.S. surplus weapons-usable plutonium materials in waste forms that meet the ''spent fuel'' standard and are acceptable for disposal in a geologic repository. Using the ceramic can-in-canister technology selected for immobilization, surplus plutonium materials will be chemically combined into ceramic forms which will be encapsulated within large canisters of high level waste (HLW) glass. Deployment of the immobilization capability should occur by 2008 and be completed within 10 years. In support of this goal, the DOE Office of Fissile Materials Disposition (MD) is conducting development and testing (D&T) activities at four DOE laboratories under the technical leadership of Lawrence Livermore National Laboratory (LLNL). The Savannah River Site has been selected as the site for the planned Plutonium Immobilization Plant (PIP). The D&T effort, now in its third year, will establish the technical bases for the design, construction, and operation of the U. S. capability to immobilize surplus plutonium in a suitable and cost-effective manner. Based on the D&T effort and on the development of a conceptual design of the PIP, automation is expected to play a key role in the design and operation of the Immobilization Plant. Automation and remote handling are needed to achieve required dose reduction and to enhance operational efficiency
The influence of building height variability on pollutant dispersion and pedestrian ventilation in idealized high-rise urban areas
Studies are still required to understand how rural/marine wind remove ground-level pollutants released uniformly in street networks of high-rise urban areas. The link between building height variability and pollutant removal process still remains unclear. Several idealized urban-like neighbourhoods made of 9-row and 18-row small-scale high-rise square arrays (building width B = street width W, building packing density λ p = 0.25) were first numerically studied with a parallel approaching wind and neglecting thermal effects. Normalized pollutant transport rates and pedestrian purging flow rate were applied to quantify the contribution of pollutant removal by mean flow and turbulent diffusion and their net purging capacity. Results show that the prediction of isothermal turbulent flows agreed generally well with wind tunnel data. For 9-row arrays with building height variations (standard deviation of 0-57.1%) and the same average canopy height (H 0 = 2.33W), pollutant removal mainly depends on mean flows. Larger standard deviations tend to induce better pedestrian ventilation. In comparison to small and large standard deviations, medium values of 14.3-42.9% may experience smaller purging capacity by horizontal mean flows but significantly enhance that by vertical mean flows. For arrays with uniform heights, lowering aspect ratios (H/W = 2.33 and 2.67-1.5) or increasing street lengths (9-row to 18-row) may enhance the contribution of removing pollutants by turbulent diffusions across canopy roofs which may be similarly important as that by mean flows. Although further investigations are still required, this paper clarifies the relationship between building layouts, height variability and removal potential of ground-level pollutants in high-rise urban-like geometries. © 2012 Elsevier Ltd.link_to_subscribed_fulltex