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Steven C. Marschman

Bio: Steven C. Marschman is an academic researcher. The author has contributed to research in topics: Hanford Site & Spent nuclear fuel. The author has an hindex of 2, co-authored 2 publications receiving 5 citations.

Papers
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ReportDOI
10 Aug 1999
TL;DR: In this paper, the authors proposed a dry storage concept for storing spent nuclear fuel (SNF) discharged from the N-Reactor at Hanford Site in order to protect the Columbia River.
Abstract: Approximately 2100 metric tons of spent nuclear fuel (SNF) discharged from the N-Reactor have been stored underwater at the K-Basins in the 100 Area of the Hanford Site. The spent fuel has been stored in the K-East Basin since 1975 and in the K-West Basin since 1981. Some of the SNF elements in these basins have corroded because of various breaches in the Zircaloy cladding that occurred during fuel discharge operations and/or subsequent handling and storage in the basins. Consequently, radioactive material in the fuel has been released into the basin water, and water has leaked from the K-East Basin into the soil below. To protect the Columbia River, which is only 380 m from the basins, the SNF is scheduled to be removed and transported for interim dry storage in the 200 East Area, in the central portion of the Site. However, before being shipped, the corroded fuel elements will be loaded into Multi-Canister OverPacks and conditioned. The conditioning process will be selected based on the Integrated Process Strategy (IPS) (WHC 1995), which was prepared on the basis of the dry storage concept developed by the Independent Technical Assessment (ITA) team (ITA 1994).

3 citations

ReportDOI
24 Sep 1999
TL;DR: A series of whole element drying tests (reported in separate documents, see Section 8.0) have been conducted by Pacific Northwest National Laboratory (PNNL) on several intact and damaged fuel elements recovered from both the K-East and K-West basins as discussed by the authors.
Abstract: The water-filled K-Basins in the Hanford 100-Area have been used to store N-Reactor spent nuclear fuel (SNF) since the 1970s. Because some leaks in the basins have been detected and some of the fuel is breached due to handling damage and corrosion, efforts are underway to remove the fuel elements from wet storage. An Integrated Process Strategy (IPS) has been developed to package, dry, transport, and store these metallic uranium fuels in an interim storage facility on the Hanford Site (WHC 1995). Information required to support the development of the drying processes, and the required safety analyses, is being obtained from characterization tests conducted on fuel elements removed from the K-Basins. A series of whole element drying tests (reported in separate documents, see Section 8.0) have been conducted by Pacific Northwest National Laboratory (PNNL) on several intact and damaged fuel elements recovered from both the K-East and K-West Basins. This report documents the results of the first of those tests (Run 1), which was conducted on an N-Reactor inner fuel element (1990) that had been stored underwater in the K-West Basin (see Section 2.0). This fuel element was subjected to a combination of low- and high-temperature vacuum drying treatments that were intended to mimic, wherever possible, the fuel treatment strategies of the IPS. The testing was conducted in the Whole Element Furnace Testing System, described in Section 3.0, located in the Postirradiation Testing Laboratory (PTL, 327 Building). The test conditions and methodology are given in Section 4.0, and the experimental results provided in Section 5.0. These results are further discussed in Section 6.0.

2 citations


Cited by
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Journal ArticleDOI
TL;DR: In this article, the interaction of water vapor with UO 2 (001) by using low energy electron diffraction (LEED), X-ray photoelectron spectroscopy (XPS), low energy ion scattering (LEIS) and electron stimulated desorption (ESD) was studied.

52 citations

ReportDOI
06 Dec 2004
TL;DR: In this paper, a survey of the technical literature was performed to summarize the mechanical properties of inorganic components in K Basin sludge, including irradiated uranium metal and uranium dioxide.
Abstract: A survey of the technical literature was performed to summarize the mechanical properties of inorganic components in K Basins sludge. The components included gibbsite, ferrihydrite, lepidocrocite and goethite, hematite, quartz, anorthite, calcite, basalt, Zircaloy, aluminum, and, in particular, irradiated uranium metal and uranium dioxide. Review of the technical literature showed that information on the hardness of uranium metal at irradiation exposures similar to those experienced by the N Reactor fuel present in the K Basins (typically up to 3000 MWd/t) were not available. Measurements therefore were performed to determine the hardness of coupons taken from three irradiated N Reactor uranium metal fuel elements taken from K Basins. Hardness values averaged 30 {+-} 8 Rockwell C units, similar to values previously reported for uranium irradiated to {approx}1200 MWd/t. The physical properties of candidate uranium metal and uranium dioxide surrogates were gathered and compared. Surrogates having properties closest to those of irradiated uranium metal appear to be alloys of tungsten. The surrogate for uranium dioxide, present both as particles and agglomerates in actual K Basin sludge, likely requires two materials. Cerium oxide, CeO2, was identified as a surrogate of the smaller UO2 particles while steel grit was identified for the UO2 agglomerates.

13 citations

Journal ArticleDOI
TL;DR: He transport in UO(2), which is an important material with respect to both civilian and military applications, is explored and it is found that He transport proceeds much faster through the grain-boundary and grain-junction network compared with intracrystalline UO (2) regions, in accordance with experiment.
Abstract: We present a general strategy for generating full atomistic models of nanopolycrystalline materials including bulk and thin film. In particular, models for oxide nanoparticles were constructed using simulated amorphisation and crystallisation and used to populate a library of oxide nanoparticles (amorphous and crystalline) with different radii. Nanoparticles were then taken from this library and positioned, within a specific volume, using Monte Carlo techniques, to facilitate a tight-packed structure. The grain-size distribution of the polycrystalline material was controlled by selecting particular sized nanoparticles from the library. The (randomly oriented) grains facilitated a polycrystalline oxide, which comprised a network of general grain-boundaries. To help validate the model, gas diffusion through the (polycrystalline) oxide material was then simulated and the activation energy calculated directly. Specifically, we explored He transport in UO(2), which is an important material with respect to both civilian and military applications. We found that He transport proceeds much faster through the grain-boundary and grain-junction network compared with intracrystalline UO(2) regions, in accordance with experiment.

4 citations

ReportDOI
30 Sep 2002
TL;DR: The Hanford Spent Nuclear Fuel Project focuses its efforts on determining how to safely move the degraded N-Reactor spent fuel from water-stored basins to a dry storage facility.
Abstract: The Hanford Spent Nuclear Fuel Project focuses its efforts on determining how to safely move the degraded N-Reactor spent fuel from water-stored basins to a dry storage facility Based on the laboratory data, the project chose to use a conservative enhancement factor in analyzing the oxidation behavior of the spent metallic fuel However, there is a need for the project to increase the fuel throughput for the drying treatment process by implementing certain design optimization steps The study discussed in this paper re-evaluated the previous laboratory data in conjunction with the cold vacuum drying (CVD) process experience and determined whether the built-in level of conservatism could accommodate the potential changes in the process without compromising public and worker safety An established oxidation reaction-rate constant was used to accurately determine the reactive surface areas of corroded N-Reactor fuel elements The surface areas calculated for 6 different N-Reactor elements that were stored in the K-West Basin and shipped to Pacific Northwest National Laboratory for drying studies ranges from as low as 00018 m2 for a broken element to 81 m2 for a highly corroded SNF element 5744U The SNF element 0309M that was a clean broken piece was used to calibrate themore » calculation method The result using the SNF reaction rate constant (ie, kSNF) gave a very good (ie, 00018 m2) agreement with the geometrical value of 00015 m2 Having established that the hydrogen generation can be used to determine the exposed surface area of these irregular corroded SNF elements, the calculations was extended to provide a good estimate of the exposed uranium surface area of SNF elements loaded into the multi-canister overpacks (MCOs)« less

3 citations

20 Apr 2000
TL;DR: Plys et al. as discussed by the authors compiled a topical reference on the phenomena, experiences, experiments, and prediction of uranium pyrophoricity for the Hanford Spent Nuclear Fuel Project (SNFP) with specific applications to SNFP process and situations.
Abstract: We have compiled a topical reference on the phenomena, experiences, experiments, and prediction of uranium pyrophoricity for the Hanford Spent Nuclear Fuel Project (SNFP) with specific applications to SNFP process and situations. The purpose of the compilation is to create a reference to integrate and preserve this knowledge. Decades ago, uranium and zirconium fires were commonplace at Atomic Energy Commission facilities, and good documentation of experiences is surprisingly sparse. Today, these phenomena are important to site remediation and analysis of packaging, transportation, and processing of unirradiated metal scrap and spent nuclear fuel. Our document, bearing the same title as this paper, will soon be available in the Hanford document system [Plys, et al., 2000]. This paper explains general content of our topical reference and provides examples useful throughout the DOE complex. Moreover, the methods described here can be applied to analysis of potentially pyrophoric plutonium, metal, or metal hydride compounds provided that kinetic data are available. A key feature of this paper is a set of straightforward equations and values that are immediately applicable to safety analysis.

2 citations