Compact Nuclear Power Source Benchmark Study. MARK PAULSON (University of Wisconsin Madison, WI 53715) TAEK K. KIM (Argonne National Laboratory, Argonne, IL, 60439)

The compact nuclear power source (CNPS) design was a small medium-enriched uranium-graphite-moderated reactor developed at Lose Alamos National Lab (LANL). Although the program was canceled before a complete demonstration CNPS was built, a series of critical experiments were performed at LANL given the unique CNPS core design. Today the most likely candidate for a Generation IV reactor is a Very High Temperature Reactor (VHTR) with similar design parameters as the CNPS. However, the published literature pertaining to the CNPS critical experiments does not consistently report the reactor's design parameters. This paper presents the results of a series of Monte-Carlo method benchmarking and sensitivity modeling of the CNPS design. The various CNPS configurations reported in the literature were evaluated and results of core eigenvalues and core physics data was compared. The modeling results were comparable but did not exactly reproduce the values found in the CNPS literature. The sensitivity results provide a more accurate understanding of the CNPS design and bias of benchmarking data. However, further modeling is needed to develop the CNPS benchmark problem for application to a VHTR Gen IV reactor design.

Comparison of Electron Beam Diffraction and Soft X-ray Attenuation Methods to Achieve Approximate Thickness of Lithium Charge State Stripper for RIA. ANTONIO JOHNSON (University of Missouri-Rolla Rolla, MO 65409) STEVE LOMPERSKI (Argonne National Laboratory, Argonne, IL, 60439)

The Rare Isotope Accelerator is a high priority new facility that will provide researchers with the opportunity to learn more about short-lived isotopes that are too unstable to study with conventional facilities. This type of accelerator is capable of delivering high-energy radioactive beams of isotopes from Hydrogen to Uranium for experiments relating to astrophysics and nuclear science. The high-energy radioactive beams are more effective at reaching their final speeds (~0.81 c) if they are stripped of their electrons by a charge-state stripper. Experiments have shown that lithium can withstand a 12 MeV/u ion beam in high vacuum. Research has also shown that the thickness of the liquid lithium charge state stripper is a key parameter to determine the behavior of the lithium as it interacts with the ion beam. Therefore the thickness of the lithium film must be measured and monitored. Many techniques have been investigated to monitor and/or measure the thickness of the liquid lithium film. The search for the most promising technique is primarily based on commercial availability and difficulty of implementation. The electron beam diffraction and soft x-ray attenuation techniques are the most promising methods among many possibilities for example, micrometer types, interferometry, laser triangulation, Terra Hertz wave, and Eddy currents)[4] because they meet the standards needed to drive the RIA project further toward completion. Many criteria are used to judge the suitability of a particular film thickness measurement technique. These include hardware and software, compatibility, development cost, accuracy, reliability, and safety. Inside the lithium loop, the equipment is exposed to high temperature (250oC), high vacuum (2x10-4 Torr), and the possibility of lithium splashing. Equipment that can perform in this environment without disrupting the very nature of RIA system is required. The electron beam diffraction method is the better choice for the reason that it is commercially available, less expensive, easier to administer, and more productive in a vacuum bakeable environment than soft x-ray attenuation.

Creating an Interactive Timeline of Events Pertaining to the Creation and Completion of a Nuclear Waste Repository at Yucca Mountain, Nevada. SCOTT SPYCHALA (University of Missouri-Rolla Rolla, MO 65401) STEVE LOMPERSKI (Argonne National Laboratory, Argonne, IL, 60439)

Realization of nuclear waste storage in the United States including delays to date in the Yucca Mountain Project is well known as a tremendous challenge from the perspective of Federal project management. As with any major governmental project (The Hoover Dam, ORNL's Spallation Neutron Source) the Yucca Mountain Project (YMP) requires several governmental entities to make individual but coordinated decisions. The Nuclear Waste Policy act requires that all responsible bodies submit reports containing performance metrics and objectives. While these reports are issued separately and progress stepwise to a common goal, there is very little examination of the dynamics of the project in its entirety. The Yucca Mountain Interactive Timeline (YaMIT) project is an attempt to statically and dynamically model the YMP as a whole. In order to create the static model, the dynamic modeling program Vensim was used. We created a static model that identifies stakeholders and their tasks along one axis, with time along the other. Task to task relationships are connected by functional arrows. While Vensim was able to provide initial logic for the dynamic model, limitations prevented full development, so the programming language Java was used to complete of the simulation. Constant, linear and non-linear utility functions in time were incorporated into tasks that downstream predicted an end-of -project timeline. The dynamic model may be, in scope and with respect to utility functions, a breakthrough in large-scale project modeling. This dynamic portion of the YaMIT project is especially useful in determining the influence of multitask coupling and sensitivity analysis. We are developing a first-of-a-kind user friendly interface to assess the impact of decisions and completion of tasks in multi-stakeholder Federal projects.

Millimeter wavelength RADAR detection of air ions and aerosols produced from ionizing radiation. MIKE HULL (University of Illinois at Urbana Champaign Urbana Champaign, IL 61820) NACHAPPA GOPALSAMI (Argonne National Laboratory, Argonne, IL, 60439)

Current methods of detecting ionizing radiation sources depend on the detector physically lying within the range of the radiation. For many sources, this is a matter of an inch or two in air. As a more versatile means of detecting radiation, a milimeter wave-length RADAR system is being developed and calibrated. To test the RADAR, it is first necessary to design an experiment where a known activity of radioactive substance is released into a controlled volume with variable temperatures and humidities, including the supersaturation humidities found in the atmosphere (above power plants, ie). For this purpose, the processes of droplet formation, vapor condensation, air chilling and adiabatic expansions were studied, and in the end, a 12" cube transparent polycarbonate chamber was carefully designed and then constructed. With consideration of the temperature and pressure ranges required for the experiment, appropriate and relatively inexpensive humidity detector, thermometer, and pressure gauge were procured. Testing is currently being done on this chamber to see whether the method of uniform cooling is sufficient to reach supersaturation, or whether condensation will occur before the center has cooled, and thus a volume expansion technique is required. The answer to this research is a necessary one for the further development of the millimeter RADAR radiation detector.

Obtaining Region Dependent Reaction Rates using MCNP4C for the International Criticality Safety Benchmark Evaluation Project Zero-Power Reactor 6/7 Benchmark. THOMAS GOTER (University of Missouri-Rolla Rolla, MO 65401) STEVE LOMPERSKI (Argonne National Laboratory, Argonne, IL, 60439)

Benchmark calculations for the International Criticality Safety Benchmark Evaluation Project (ICSBEP) prove to be useful in the validation of new codes and cross section data. Improved codes and updated cross section data in turn leads to better neutron path modeling and simulation. We use Monte Carlo N-Particle transport code, commonly known as MCNP, to calculate the regional reaction rates for the ICSBEP benchmark for Assembly 7 of Zero Power Reactor 6 (ZPR 6/7). A ZPR is a rudimentary experimental reactor operated at such a low power that a coolant is not required. In order to accurately simulate the reactor assembly, exact dimensions and nuclide compositions must be known. For simplification of the problem the as built geometry was transposed to an R-Z geometry. The transposition was automatically accomplished through the use of VIM, but an error in criticality still resulted from it. VIM is a separate nuclear criticality safety code which also is used in benchmark evaluations. The reaction rates desired are fission production, fission, radiative capture, (n, 2n) production and leakage. After the results are calculated they are then compared with results from VIM. The difference between VIM and MCNP is how they treat the resonance range of the cross section. Error calculations showed a very small amount of relative error between the two codes. There was a much larger error between the as-built criticality and the R-Z criticality. This means that the transposition through the use of VIM was the cause of that error since the as-built model obtained a criticality level much closer to the actual ZPR 6/7 experimental value. Future work will be to continue adding to the database of benchmarks for the ICSBEP and to encourage a greater level of international cooperation in the ICSBEP.

Optimum Fuel Design for the One-pass Deep Burn Concept. JEROME CASE (University of Wisconsin Madison, WI 53706) DR. TEAK K. KIM (Argonne National Laboratory, Argonne, IL, 60439)

The objective of the Deep Burn Modular Helium Reactor (DB-MHR) project is to maximize the transmutation of heavy elements, especially plutonium, in spent fuel from Commercial Light Water Reactors (LWR's) while providing a safe, clean, reliable and competitive source of electricity. This study provides a TRISO fuel configuration, which balances burnup of transuranic actinides, including overall burnup and plutonium transmutation, and the cycle length for a four-batch fuel cycle. This optimized configuration is intended for future design and core performance calculations for the once-through, single-phase fuel DB-MHR design. The most favorable fuel design found in this study has a 224µm diameter fuel kernel and a packing fraction of 15%. This combination results in a fuel-to-moderator ratio, cycle length and transmutation similar to previously recorded configurations. Specifically, a fuel-to-moderator ratio of 0.0003255 supplies a cycle length of 272 days and a transuranic transmutation rate of 57.9%. Plutonium transmutation is 61.3% and the total burnup of the fuel is 544 GWd/t.

Out-of-beam studies of beta decay products near ⁴⁸Ca. JARED NANCE (Beloit College Beloit, WI 53511) ROBERT JANSSENS (Argonne National Laboratory, Argonne, IL, 60439)

Traditional gamma-ray spectroscopy focuses on 'in-beam', or prompt gamma rays as a probe of nuclear structure. In this paper, I describe a data mining experiment that was performed on data collected at the Gammasphere facility at Argonne National Laboratories that focused on 'out-of-beam' events. These events are typically missed by in-beam studies, as they take longer to produce and are of much lower relative intensity than in-beam reaction products. Using this technique, the existing level schemes for ^46,47,49Ca were verified. Furthermore, I propose a new energy level in ⁴⁷Ca (E=3143keV).

Radiation Damage Studies at the Advanced Photon Source. JAMES YOUNG (Diablo Valley College Pleasant Hill, CA 94523) MARIA PETRA (Argonne National Laboratory, Argonne, IL, 60439)

The rare-earth permanent magnets in the insertion devices (IDs) of a storage ring are subjected to a harsh radiation environment that can lead to radiation-induced demagnetization significant enough to require the removal of the magnets for remagnetization. Integrated radiation dose measurements on the IDs give an indication of their radiation sensitivity and lifetime in the radiation environment of the accelerator facility and identification of regions where the probability for damage is highest. To date, at the Advanced Photon Source (APS) 7 GeV electron storage ring, the most severe radiation damage has been observed in those sectors that are equipped with small 5-mm vacuum chambers (i.e., sectors 3 and 4). The dose profile along the length of the ID in sector 4 was measured with alanine dosimeters, and magnetic measurements on that ID were also performed following a 3-month-long operation period (i.e., the 2005-1 run.) Helmholtz coil measurements of the magnetic field degradation of a sampling of magnets from the sector 4 ID were performed. The recorded doses along the length of the ID revealed significant dose nonuniformities with the highest doses being observed towards the downstream end of the ID where the greater magnetic degradation also occurs. In addition, transverse Hall probe x-scans of sample magnets from that ID showed that the sample magnets are primarily demagnetized directly under the beam. This study, as part of the continuing ID dosimetric and magnet damage studies on the APS IDs produced results consistent with previous studies. These results will contribute to the body of knowledge concerning the determination of the threshold for damage and identification of areas where the likelihood of damage is highest.

Transient CO[2] Line Pipe Break Analysis for the Liquid Metal Secure Transportable Autonomous Reactor (STAR-LM). ALEKSANDAR MILICEVIC (University of Wisconsin Madison, WI 53715) SIENICKI, JAMES J. (Argonne National Laboratory, Argonne, IL, 60439)

A simple transient model for a 400MWt Liquid Metal Secure Transportable Autonomous Reactor (STAR-LM) system has been developed and incorporated into a computer code for a pipe break analysis on the gas turbine Brayton cycle utilizing supercritical carbon dioxide as the working fluid. The transient model was used to determine if a break in the pipe at the turbine inlet would affect reactor power, and to determine what size break would affect reactor power most significantly. A break in the pipe (50 cm in diameter) at the turbine inlet does affect reactor power, and a maximum increase in power (4%) is calculated for a 12.5 cm in diameter sized break.