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Interest in high-quality integral benchmark data is increasing as efforts to quantify and reduce calculational uncertainties accelerate to meet the demands of next generation reactor and advanced fuel cycle concepts. The International Reactor Physics Experiment Evaluation Project (IRPhEP) and the International Criticality Safety Benchmark Evaluation Project (ICSBEP) continue to expand their efforts and broaden their scope to identify, evaluate, and provide integral benchmark data for method and data validation. Benchmark model specifications provided by these two projects are used heavily by the international reactor physics, nuclear data, and criticality safety communities. Thus far, 14 countries have contributed to the IRPhEP, and 20 have contributed to the ICSBEP. The status of the IRPhEP and ICSBEP is discussed in this paper, and the future of the two projects is outlined and discussed. Selected benchmarks that have been added to the IRPhEP and ICSBEP handbooks since PHYSOR’06 are highlighted, and the future of the two projects is discussed.

The authors report a measurement of resonance parameters of the orbitally excited (L = 1) narrow B{sup 0} mesons in decays to B{sup (*)+}{pi}{sup -} using 1.7 fb{sup -1} of data collected by the CDF II detector at the Fermilab Tevatron. The mass and width of the B*{sub 2}{sup 0} state are measured to be m(B*{sub 2}{sup 0}) = 5740.2{sub -1.8}{sup +1.7}(stat.){sub -0.8}{sup +0.9}(syst.) MeV/c{sup 2} and {Lambda}(B*{sub 2}{sup 0}) = 22.7{sub -3.2}{sup +3.8}(stat.){sub -10.2}{sup +3.2}(syst.) MeV/c{sub 2}. The mass difference between the B*{sub 2}{sup 0} and B{sub 1}{sup 0} states is measured to be 14.9{sub -2.5}{sup +2.2}(stat.){sub -1.4}{sup +1.2}(syst.) MeV/c{sup 2}, resulting in a B{sub 1}{sup 0} mass of 5725.3{sub -2.2}{sup +1.6}(stat.){sub -1.5}{sup +1.4}(syst.) MeV/c{sup 2}. This is currently the most precise measurement of the masses of these states and the first measurement of the B*{sub 2}{sup 0} width.

The Next Generation Nuclear Plant (NGNP) and other High-Temperature Gas-cooled Reactor (HTGR) Projects require research, development, design, construction, and operation of a nuclear plant intended for both high-efficiency electricity production and high-temperature industrial applications, including hydrogen production. During the life cycle stages of an HTGR, plant systems, structures and components (SSCs) will be developed to support this reactor technology. To mitigate technical, schedule, and project risk associated with development of these SSCs, a large-scale test facility is required to support design verification and qualification prior to operational implementation. As a full-scale helium test facility, the Component Test facility (CTF) will provide prototype testing and qualification of heat transfer system components (e.g., Intermediate Heat Exchanger, valves, hot gas ducts), reactor internals, and hydrogen generation processing. It will perform confirmation tests for large-scale effects, validate component performance requirements, perform transient effects tests, and provide production demonstration of hydrogen and other high-temperature applications. Sponsored wholly or in part by the U.S. Department of Energy, the CTF will support NGNP and will also act as a National User Facility to support worldwide development of High-Temperature Gas-cooled Reactor technologies.

The spatial arrangements of hematite particles within aqueous soil and clay samples are investigated with soft X-ray microscopy, taking advantage of the elemental contrast at the Fe-L edge around E = 707 eV. In combination with stereo microscopy, information about spatial arrangements are revealed and correlated to electrostatic interactions of the different mixtures. Manipulation of a sample mounted to the microscope is possible and particles added while imaging can be detected.

One key long-standing issue that must be overcome to fully realize the successful growth of nuclear power is to determine other benefits of nuclear energy apart from meeting the electricity demands. The Next Generation Nuclear Plant (NGNP) will most likely be producing electricity and heat for the production of hydrogen and/or oil retrieval from oil sands and oil shale to help in our national pursuit of energy independence. For nuclear process heat to be utilized, intermediate heat exchange is required to transfer heat from the NGNP to the hydrogen plant or oil recovery field in the most efficient way possible. Development of nuclear reactor - process heat technology has intensified the interest in liquid metals as heat transfer media because of their ideal transport properties. Liquid metal heat exchangers are not new in practical applications. An important rational for considering liquid metals is the potential convective heat transfer is among the highest known. Thus explains the interest in liquid metals as coolant for intermediate heat exchange from NGNP. For process heat it is desired that, intermediate heat exchangers (IHX) transfer heat from the NGNP in the most efficient way possible. The production of electric power at higher efficiency via the …

The Advanced Test Reactor (ATR) is one of the world’s premiere test reactors for performing long term, high flux, and/or large volume irradiation test programs. The ATR is a very versatile facility with a wide variety of experimental test capabilities for providing the environment needed in an irradiation experiment. These capabilities include simple capsule experiments, instrumented and/or temperature-controlled experiments, and pressurized water loop experiment facilities. Monitoring systems have also been utilized to monitor different parameters such as fission gases for fuel experiments, to measure specimen performance during irradiation. ATR’s control system provides a stable axial flux profile throughout each reactor operating cycle, and allows the thermal and fast neutron fluxes to be controlled separately in different sections of the core. The ATR irradiation positions vary in diameter from 16 mm to 127 mm over an active core height of 1.2 m. This paper discusses the different irradiation capabilities with examples of different experiments and the cost/benefit issues related to each capability. The recent designation of ATR as a national scientific user facility will make the ATR much more accessible at very low to no cost for research by universities and possibly commercial entities.

We present the latest measurements on production, spectroscopy, lifetimes and branching fractions for b-mesons, b-baryons and quarkonia. We also discuss recent results on B{sub s}{sup 0} mixing as well as on CP violation for the B{sub s}{sup 0} meson and for b-baryons. These results were obtained by analyzing data collected by the CDF II detector at Fermilab.

We report a measurement of the single top quark production cross section in 2.2 fb{sup -1} of p{bar p} collision data collected by the Collider Detector at Fermilab at {radical}s = 1.96 TeV. Candidate events are classified as signal-like by three parallel analyses which use likelihood, matrix element, and neural network discriminants. These results are combined in order to improve the sensitivity. We observe a signal consistent with the standard model prediction, but inconsistent with the background only model by 3.7 standard deviations with a median expected sensitivity of 4.9 standard deviations. We measure a cross section of 2.2{sub -0.6}{sup +0.7}(stat+sys) pb, extract the CKM matrix element value |V{sub tb}| = 0.88{sub -0.12}{sup +0.13}(stat + sys) {+-} 0.07(theory), and set the limit |V{sub tb}| > 0.66 at the 95% C.L.

The U.S. Nuclear Regulatory Commission maintains a set of risk models covering the U.S. commercial nuclear power plants. These standardized plant analysis risk (SPAR) models include several loss-of-coolant accident (LOCA) initiating events such as small (SLOCA), medium (MLOCA), and large (LLOCA). All of these events involve a loss of coolant inventory from the reactor coolant system. In order to maintain a level of consistency across these models, initiating event frequencies generally are based on plant-type average performance, where the plant types are boiling water reactors and pressurized water reactors. For certain risk analyses, these plant-type initiating event frequencies may be replaced by plant-specific estimates. Frequencies for SPAR LOCA initiating events previously were based on results presented in NUREG/CR-5750, but the newest models use results documented in NUREG/CR-6928. The estimates in NUREG/CR-6928 are based on historical data from the initiating events database for pressurized water reactor SLOCA or an interpretation of results presented in the draft version of NUREG-1829. The information in NUREG-1829 can be used several ways, resulting in different estimates for the various LOCA frequencies. Various ways NUREG-1829 information can be used to estimate LOCA frequencies were investigated and this paper presents two methods for the SPAR model standard …

A 'multiple beamlet' experiment aimed at investigating the transverse space charge effect was recently conducted at the Argonne Wakefield Accelerator. The experiment generated a symmetric pattern of 5 beamlets on the photocathode of the RF gun with the drive laser. We explored the evolution of the thereby produced 5 MeV, space-charge dominated electron beamlets in the 2m drift following the RF photocathode gun for various external focusing. Two important effects were observed and benchmarked using the particle-in-cell beam dynamics code IMPACT-T. In this paper, we present our experimental observation and their benchmarking with Impact-T.

A system analysis has been performed to assess the efficiency and carbon utilization of a nuclear-assisted coal gasification process. The nuclear reactor is a high-temperature helium-cooled reactor that is used primarily to provide power for hydrogen production via high-temperature electrolysis. The supplemental hydrogen is mixed with the outlet stream from an oxygen-blown coal gasifier to produce a hydrogen-rich gas mixture, allowing most of the carbon dioxide to be converted into carbon monoxide, with enough excess hydrogen to produce a syngas product stream with a hydrogen/carbon monoxide molar ratio of about 2:1. Oxygen for the gasifier is also provided by the high-temperature electrolysis process. Results of the analysis predict 90.5% carbon utilization with a syngas production efficiency (defined as the ratio of the heating value of the produced syngas to the sum of the heating value of the coal plus the high-temperature reactor heat input) of 66.1% at a gasifier temperature of 1866 K for the high-moisture-content lignite coal considered. Usage of lower moisture coals such as bituminous can yield carbon utilization approaching 100% and 70% syngas production efficiency.

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This session will examine a method developed by Federal and Contractor personnel at the U.S. Department of Energy, National Nuclear Security Administration Nevada Site Office (NNSA/NSO) to examine long-term maintenance of DOE Integrated Safety Management System (ISMS) criteria, including safety culture attributes, as well as identification of process improvement opportunities. This process was initially developed in the summer of 2000 and has since been expanded to recognize the importance of safety culture attributes, and associated safety culture elements, as defined in DOE M 450.4-1, “Integrated Safety Management System Manual.” This process has proven to significantly enhance collective awareness of the importance of long-term ISMS implementation as well as support commitments by NNSA/NSO personnel to examine the continued effectiveness of ISMS processes.

Hydraulic and Clean-in-Place Evaluations for a 12.5 cm Annular Centrifugal Contactor at the INL Troy G. Garn, Dave H. Meikrantz, Nick R. Mann, Jack D. Law, Terry A. Todd Idaho National Laboratory Commercially available, Annular Centrifugal Contactors (ACC) are currently being evaluated for processing dissolved nuclear fuel solutions to selectively partition integrated elements using solvent extraction technologies. These evaluations include hydraulic and clean-in-place (CIP) testing of a commercially available 12.5 cm unit. Data from these evaluations is used to support design of future nuclear fuel reprocessing facilities. Hydraulic testing provides contactor throughput performance data on two-phase systems for a wide range of operating conditions. Hydraulic testing results on a simple two-phase oil and water system followed by a 30 % Tributyl phosphate in N-dodecane / nitric acid pair are reported. Maximum total throughputs for this size contactor ranged from 20 to 32 liters per minute without significant other phase carryover. A relatively new contactor design enhancement providing Clean-in-Place capability for ACCs was also investigated. Spray nozzles installed into the central rotor shaft allow the rotor internals to be cleaned, offline. Testing of the solids capture of a diatomaceous earth/water slurry feed followed by CIP testing was performed. Solids capture efficiencies …

Developing new nuclear grades of graphite used in the core of a High Temperature Gas-cooled Reactor (HTGR) is one of the critical development activities being pursued within the Next Generation Nuclear Plant (NGNP) program. Graphite’s thermal stability (in an inert gas environment), high compressive strength, fabricability, and cost effective price make it an ideal core structural material for the HTGR reactor design. While the general characteristics necessary for producing nuclear grade graphite are understood, historical “nuclear” grades no longer exist. New grades must be fabricated, characterized, and irradiated to demonstrate that current grades of graphite exhibit acceptable non-irradiated and irradiated properties upon which the thermo-mechanical design of the structural graphite in NGNP is based. The NGNP graphite R&D program has selected a handful of commercially available types for research and development activities necessary to qualify this nuclear grade graphite for use within the NGNP reactor. These activities fall within five primary areas; 1) material property characterization, 2) irradiated material property characterization, 3) modeling, and 4) ASTM test development, and 5) ASME code development efforts. Individual research and development activities within each area are being pursued with the ultimate goal of obtaining a commercial operating license for the nuclear graphite from …

Various tools exist to analyze enterprise network systems and to produce attack graphs detailing how attackers might penetrate into the system. These attack graphs, however, are often complex and difficult to comprehend fully, and a human user may find it problematic to reach appropriate configuration decisions. This paper presents methodologies that can 1) automatically identify portions of an attack graph that do not help a user to understand the core security problems and so can be trimmed, and 2) automatically group similar attack steps as virtual nodes in a model of the network topology, to immediately increase the understandability of the data. We believe both methods are important steps toward improving visualization of attack graphs to make them more useful in configuration management for large enterprise networks. We implemented our methods using one of the existing attack-graph toolkits. Initial experimentation shows that the proposed approaches can 1) significantly reduce the complexity of attack graphs by trimming a large portion of the graph that is not needed for a user to understand the security problem, and 2) significantly increase the accessibility and understandability of the data presented in the attack graph by clearly showing, within a generated visualization of the network …

In the mid 80’s, a four-group/two-region, entirely analytical 1D nodal benchmark appeared. It was readily acknowledged that this special case was as far as one could go in terms of group number and still achieve an analytical solution. In this work, we show that by decomposing the solution to the multigroup diffusion equation into homogeneous and particular solutions, extension to any number of groups is a relatively straightforward exercise using the mathematics of linear algebra.

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A variable input coupler has been designed for the Fermilab vertical cavity test facility (VCTF), a facility for CW RF vertical testing of bare ILC 1.3 GHz 9-cell SRF cavities at 2K, to provide some flexibility in the test stand RF measurements. The variable coupler allows the cavity to be critically coupled for all RF tests, including all TM010 passband modes, which will simplify or make possible the measurement of those modes with very low end-cell fields, e.g., {pi}/9 mode. The variable coupler assembly mounts to the standard input coupler port on the cavity, and uses a cryogenic motor submerged in superfluid helium to control the antenna position. The RF and mechanical design and RF test results are described.

Data collected in Run II of the Fermilab Tevatron are searched for indications of new electroweak-scale physics. Rather than focusing on particular new physics scenarios, CDF data are analyzed for discrepancies with the standard model prediction. A model-independent approach (VISTA) considers gross features of the data, and is sensitive to new large cross-section physics. Further sensitivity to new physics is provided by two additional algorithms: a Bump Hunter searches invariant mass distributions for 'bumps' that could indicate resonant production of new particles; and the Sleuth procedure scans for data excesses at large summed transverse momentum. This combined global search for new physics in 2.0 fb{sup -1} of p{bar p} collisions at {radical}s = 1.96 TeV reveals no indication of physics beyond the standard model.

TRUEX mass transfer tests were performed using a single stage commercially available 12.5 cm centrifugal contactor and stable cerium (Ce) and europium (Eu). Test conditions included throughputs ranging from 2.5 to 15 Lpm and rotor speeds of 1750 and 2250 rpm. Ce and Eu extraction forward distribution coefficients ranged from 13 to 19. The first and second stage strip back distributions were 0.5 to 1.4 and .002 to .004, respectively, throughout the dynamic test conditions studied. Visual carryover of aqueous entrainment in all organic phase samples was estimated at < 0.1 % and organic carryover into all aqueous phase samples was about ten times less. Mass transfer efficiencies of = 98 % for both Ce and Eu in the extraction section were obtained over the entire range of test conditions. The first strip stage mass transfer efficiencies ranged from 75 to 93% trending higher with increasing throughput. Second stage mass transfer was greater than 99% in all cases. Increasing the rotor speed from 1750 to 2250 rpm had no significant effect on efficiency for all throughputs tested.

A superconducting RF cavity has to be shielded from magnetic fields present during cool down below the critical temperature to avoid freezing in the magnetic flux at localized impurities, thereby degrading the cavity intrinsic quality factor Q{sub 0}. The magnetic shielding designed for the Fermilab vertical cavity test facility (VCTF), a facility for CW RF vertical testing of bare ILC 1.3 GHz 9-cell SRF cavities, was recently completed. For the magnetic shielding design, we used two cylindrical layers: a room temperature 'outer' shield of Amumetal (80% Ni alloy), and a 2K 'inner' shield of Cryoperm 10. The magnetic and mechanical design of the magnetic shielding and measurement of the remanent magnetic field inside the shielding are described.