Test plan for thermogravimetric analyses of BWR spent fuel oxidation

Preliminary studies indicated the need for additional low-temperature spent fuel oxidation data to determine the behavior of spent fuel as a waste form for a tuffy repository. Short-term thermogravimetric analysis tests were recommended in a comprehensive technical approach as the method for providing scoping data that could be used to (1) evaluate the effects of variables such as moisture and burnup on the oxidation rate, (2) determine operative mechanisms, and (3) guide long-term, low-temperature oxidation testing. The initial test series studied the temperature and moisture effects on pressurized water reactor fuel as a function of particle and grain size. This document presents the test matrix for studying the oxidation behavior of boiling water reactor fuel in the temperature range of 140 to 225{degree}C. 17 refs., 7 figs., 3 tabs

Test plan for long-term, low-temperature oxidation of BWR spent fuel

Preliminary studies indicated the need for more spent fuel oxidation data in order to determine the probable behavior of spent fuel in a tuff repository. Long-term, low-temperature testing was recommended in a comprehensive technical approach to (1) confirm the findings of the short-term thermogravimetric analysis tests; (2) evaluate the effects of variables such as burnup, atmospheric moisture,and fuel type on the oxidation rate; and (3) extend the oxidation data base to representative repository temperatures and better define the temperature dependence of the operative oxidation mechanisms. This document presents the test plan to study the effects of atmospheric moisture and temperature on oxidation rate and phase formation using a large number of boiling-water reactor fuel samples. Tests will run for up to two years, use characterized fragmented and pulverized fuel samples, cover a temperature range of 110{degree}C to 175{degree}C, and be conducted with an atmospheric moisture content ranging from <{minus}55{degree}C to {approximately}80{degree}C dew point. After testing, the samples will be examined and made available for leaching testing. 15 refs., 2 figs., 2 tabs

Access Strategies for Network Caching

Having multiple data stores that can potentially serve content is common in modern networked applications. Data stores often publish approximate summaries of their content to enable effective utilization. Since these summaries are not entirely accurate, forming an efficient access strategy to multiple data stores becomes a complex risk management problem.This paper formally models this problem, and introduces practical algorithms with guaranteed approximation ratios, and in particular we show that our algorithms are optimal in a variety of settings. We also perform an extensive simulation study based on real data, and show that our algorithms are more robust than existing heuristics. That is, they exhibit near optimal performance in various settings whereas the efficiency of existing approaches depends upon system parameters that may change over time, or be otherwise unknown

Initial results from dissolution rate testing of N-Reactor spent fuel over a range of potential geologic repository aqueous conditions

Hanford N-Reactor spent nuclear fuel (HSNF) may ultimately be placed in a geologic repository for permanent disposal. To determine whether the engineered barrier system that will be designed for emplacement of light-water-reactor (LWR) spent fuel will also suffice for HSNF, aqueous dissolution rate measurements were conducted on the HSNF. The purpose of these tests was to determine whether HSNF dissolves faster or slower than LWR spent fuel under some limited repository-relevant water chemistry conditions. The tests were conducted using a flowthrough method that allows the dissolution rate of the uranium matrix to be measured without interference by secondary precipitation reactions that would confuse interpretation of the results. Similar tests had been conducted earlier with LWR spent fuel, thereby allowing direct comparisons. Two distinct corrosion modes were observed during the course of these 12 tests. The first, Stage 1, involved no visible corrosion of the test specimen and produced no undissolved corrosion products. The second, Stage 2, resulted in both visible corrosion of the test specimen and left behind undissolved corrosion products. During Stage 1, the rate of dissolution could be readily determined because the dissolved uranium and associated fission products remained in solution where they could be quantitatively analyzed. The measured rates were much faster than has been observed for LWR spent fuel under all conditions tested to date when normalized to the exposed test specimen surface areas. Application of these results to repository conditions, however, requires some comparison of the physical conditions of the different fuels. The surface area of LWR fuel that could potentially be exposed to repository groundwater is estimated to be approximately 100 times greater than HSNF. Therefore, when compared on the basis of mass, which is more relevant to repository conditions, the HSNF and LWR spent fuel dissolve at similar rates

Technical description of the NRC long-term whole-rod and crud performance test

Westinghouse Hanford Company (WHC) and EG and G-Idaho are jointly conducting a long-term, low-temperature, spent-fuel, whole rod and crud behavior test to provide the Nuclear Regulatory Commission (NRC) with information to assist in the licensing of light water reactor (LWR) spent-fuel, dry storage facilities. Readily available fuel rods from an H.B. Robinson Unit 2 (PWR) fuel assembly and a Peach Bottom-II (BWR) fuel assembly were selected for use in the 50-month test. Both intact and defected rods will be tested in inert and oxidizing atmospheres. A 230/sup 0/C test temperature was selected for the first 10-month run. Both nondestructive and destructive examinations are planned to characterize the fuel rod behavior during the 5-y test. Four interim examinations and a final examination will be conducted. Crud spallation behavior will be investigated by sampling the crud particulate from the test capsules at each of the four interim examinations and at the end of the test. The background to whole rod testing, description of rod breach mechanisms, and a detailed description of the test are presented in this document

Rationale for determining MCC spent fuel acquisitions

The Yucca Mountain Site Characterization Project of the US Department of Energy (DOE) is investigating the suitability of the Topopah Spring Tuff at Yucca Mountain, Nevada, for use as a disposal site for spent nuclear fuel and other high-level waste forms. The performance of the high-level waste forms and the engineered barrier system at the site must be shown to comply with the requirements in 10 CFR 60. Lawrence Livermore National Laboratory (LLNL) has the responsibility for determining the performance of the US commercial reactor spent nuclear fuels under potential repository conditions. Pacific Northwest Laboratory (PNL) performs testing of these highly radioactive materials in support of the LLNL program. This report summarizes the rationale for selecting additional spent fuels that should be acquired to support the LLNL and PNL testing programs. These programs have identified specific attributes that may affect spent fuel behavior in a repository

Path Integral Approach to Strongly Nonlinear Composite

We study strongly nonlinear disordered media using a functional method. We solve exactly the problem of a nonlinear impurity in a linear host and we obtain a Bruggeman-like formula for the effective nonlinear susceptibility. This formula reduces to the usual Bruggeman effective medium approximation in the linear case and has the following features: (i) It reproduces the weak contrast expansion to the second order and (ii) the effective medium exponent near the percolation threshold are $s=1$, $t=1+\kappa$, where $\kappa$ is the nonlinearity exponent. Finally, we give analytical expressions for previously numerically calculated quantities.Comment: 4 pages, 1 figure, to appear in Phys. Rev.

A fuel response model for the design of spent fuel shipping casks

The radiological source terms pertinent to spent fuel shipping cask safety assessments are of three distinct origins. One of these concerns residual contamination within the cask due to handling operations and previous shipments. A second is associated with debris (''crud'') that had been deposited on the fuel rods in the course of reactor operation, and a third involves the radioactive material contained within the rods. Although the lattermost source of radiotoxic material overwhelms the others in terms of inventory, its release into the shipping cask, and thence into the biosphere, requires the breach of an additional release barrier, viz., the fuel rod cladding. Hence, except for the special case involving the transport of fuel rods containing previously breached claddings, considerations of the source terms due to material contained in the fuel rods are complicated by the need to address the likelihood of fuel cladding failure during transport. The purpose of this report is to describe a methodology for estimating the shipping cask source terms contribution due to radioactive material contained within the spent fuel rods. Thus, the probability of fuel cladding failure as well as radioactivity release is addressed. 8 refs., 2 tabs

Performance of EBR-II MARK-II metallic driver fuel up to 675/sup 0/C

The performance of EBR-2 Mark-II fuel at 675/sup 0/C is significantly different from that at 590/sup 0/C. Despite the differences, the elements were all successfully irradiated to 8 at. % burnup before being removed from the reactor unbreached. Some of the more significant differences in performance at the higher temperatures are (1) The fuel-element cladding showed a double-peaking strain profile, with enhanced swelling and creep due to extensive carbide precipitation, an FCCI enhanced fission gas stress, or FCMI. (2) The FCCI zone in the fuel was cracked and represented 25% of the cladding thickness. This zone was substantially larger than observed at 590/sup 0/C. (3) The fuel pin lifted 12 mm to the restrainer dimple, so some type of restrainer may be necessary in future designs. A restrainer had previously been deemed unnecessary based on irradiations at 590/sup 0/C. (4) The fission-gas release is less at a fuel-centerline temperature of 720/sup 0/C compared at 650/sup 0/C. This difference is primarily due to the upper third of the element having very little open porosity. Owing to the lack of bnd sodium infiltration into the fuel at the higher temperature, the fission gas pressure is the same at both temperatures

Test plan for thermogravimetric analyses of BWR spent fuel oxidation

Preliminary studies indicated the need for additional low-temperature spent fuel oxidation data to determine the behavior of spent fuel as a waste form for a tuffy repository. Short-term thermogravimetric analysis tests were recommended in a comprehensive technical approach as the method for providing scoping data that could be used to (1) evaluate the effects of variables such as moisture and burnup on the oxidation rate, (2) determine operative mechanisms, and (3) guide long-term, low-temperature oxidation testing. The initial test series studied the temperature and moisture effects on pressurized water reactor fuel as a function of particle and grain size. This document presents the test matrix for studying the oxidation behavior of boiling water reactor fuel in the temperature range of 140 to 225{degree}C. 17 refs., 7 figs., 3 tabs