Theodore H. Bauer

Separations and Transmutation Criteria to Improve Utilization of a Geologic Repository

Abstract This paper describes the results of a study that uses the thermal performance of the repository to establish chemical separations and transmutation criteria for commercial spent nuclear fuel of benefit to a geologic repository, as measured by the allowable increase in utilization of repository space. The method for determining the chemical elements to be separated is based on the thermal performance of the repository. The important chemical elements are identified, the order of importance of the separated elements is established, and the relationship between the efficiency of the chemical separation and the resulting increase in utilization is determined. The proposed repository at Yucca Mountain is used as an example of a geologic repository for the purposes of illustrating the magnitude of the benefits that are possible and the implications for repository size and operation. This work is being done in support of the U.S. Department of Energy Advanced Fuel Cycle Initiative, where numerous reactor, processing, and recycling strategies are being examined to determine the impact on issues important to the viability of nuclear electricity generation, including the disposal of spent nuclear fuel and nuclear waste.

BEHAVIOR OF MODERN METALLIC FUEL IN TREAT TRANSIENT OVERPOWER TESTS

AbstractResults and analyses of margin to cladding failure and prefailure axial expansion of metallic fuel are reported for Transient Reactor Test Facility in-pile transient overpower tests M2 through M7. These include the first such tests on binary and ternary alloy fuel of the Integral Fast Reactor concept and fuel burnups to 10 at. %. The fuel was tested at full coolant flow and subjected to an exponential power rise on an 8-s period until either incipient or actual cladding failure was achieved. Objectives, designs, and methods are described with emphasis on developments unique to metal fuel safety testing. Test results include the following: (a) temperature, flow, and pressure data; (b) fuel motion diagnostic data from the fast neutron hodoscope; and (c) test remains described by both destructive and nondestructive posttest examination. The resulting M-series data base for cladding failure threshold and prefailure fuel expansion is presented. The nature of the observed cladding failure and resultant ...

Impact on geologic repository usage from limited actinide recycle in pressurized light water reactors.

A project has been conducted as part of the U.S. Department of Energy Advanced Fuel Cycle Initiative to evaluate the impact of limited actinide recycling in light water reactors on the utilization of a geologic repository where loading of the repository is constrained by the decay heat of the emplaced materials. In this study, it was assumed that spent PWR fuel was processed, removing the uranium, plutonium, americium, and neptunium, along with the fission products cesium and strontium. Previous work had demonstrated that these elements were responsible for limiting loading in the repository based on thermal constraints. The plutonium, americium, and neptunium were recycled in a PWR, with process waste and spent recycled fuel being sent to the repository. The cesium and strontium were placed in separate storage for 100–300 years to allow for decay prior to disposal. The study examined the effect of single and mutliple recycles of the recovered plutonium, americium, and neptunium, as well as different processing delay times. The potential benefit to the repository was measured by the increase in utilization of repository space as indicated by the allowable linear loading in the repository drifts (tunnels). The results showed that limited recycling would provide only a small fraction of the benefit that could be achieved with repeated processing and recycling, as is possible in fast neutron reactors.

Fuel motion in overpower tests of metallic integral fast reactor fuel

In this paper results from hodoscope data analyses are presented for transient overpower (TOP) tests M5, M6, and M7 at the Transient Reactor Test Facility, with emphasis on transient feedback mechanisms, including prefailure expansion at the tops of the fuel pins, subsequent dispersive axial fuel motion, and losses in relative worth of the fuel pins during the tests. Tests M5 and M6 were the first TOP tests of margin to cladding breach and prefailure elongation of D9-clad ternary (U-Pu-Zr) integral fast reactor-type fuel. Test M7 extended these results to high-burnup fuel and also initiated transient testing of HT-9-clad binary (U-Zr) Fast Flux Test Facility driver fuel. Results show significant prefailure negative reactivity feedback and strongly negative feedback from fuel driven to failure.

Quantitative studies of severe fuel damage using delayed neutron data

In this paper, a technique is presented to quantify fuel damage in a liquid-metal reactor through fast-running computer analysis of delayed neutron data, suitable for analysis of both small-scale in-pile experiments and full-scale plants. Validating analyses are described for five in-pile severe accident simulations performed within the Sodium Loop Safety Facility and Mol-7C test programs. Comparison is made of measured and calculated amounts of fuel damage.

Features of postfailure fuel behavior in transient overpower and transient undercooled/overpower tests in the transient reactor test facility

Prototypic oxide fuel was subjected to simulated, fast reactor severe accident conditions in a series of in-pile tests in the Transient Reactor Test Facility reactor. Seven experiments were performed on fresh and previously irradiated oxide fuel pins under transient overpower and transient undercooled. overpower accident conditions. For each of the tests, fuel motions were observed by the hodoscope. Hodoscope data are correlated with coolant flow, pressure, and temperature data recorded by the loop instrumentation. Data were analyzed from the onset of initial failure to a final mass distribution at the end of the test. In this paper results of these analyses are compared to pre- and posttest accident calculations and to posttest metallographic accident calculations and to posttest metallographic examinations and computed tomographic reconstructions from neutron radiographs.