Design of a Gas Test Loop Facility for the Advanced Test Reactor

The Office of Nuclear Energy within the U.S. Department of Energy (DOE-NE) has identified the need for irradiation testing of nuclear fuels and materials, primarily in support of the Generation IV (Gen-IV) and Advanced Fuel Cycle Initiative (AFCI) programs. These fuel development programs require a unique environment to test and qualify potential reactor fuel forms. This environment should combine a high fast neutron flux with a hard neutron spectrum and high irradiation temperature. An effort is presently underway at the Idaho National Laboratory (INL) to modify a large flux trap in the Advanced Test Reactor (ATR) to accommodate such a test facility [1,2]. The Gas Test Loop (GTL) Project Conceptual Design was initiated to determine basic feasibility of designing, constructing, and installing in a host irradiation facility, an experimental vehicle that can replicate with reasonable fidelity the fast-flux test environment needed for fuels and materials irradiation testing for advanced reactor concepts. Such a capability will be needed if programs such as the AFCI, Gen-IV, the Next Generation Nuclear Plant (NGNP), and space nuclear propulsion are to meet development objectives and schedules. These programs are beginning some irradiations now, but many call for fast flux testing within this decade

Runoff production on forest roads in a steep, mountain catchment

This study investigated how roads interact with hillslope flow in a steep, forested landscape dominated by subsurface flow and how road interactions with hillslope flow paths influence hydrologic response during storms in a second-order catchment. Runoff was measured continuously from 12 subcatchments draining to road segments and covering 14% of a 101-ha, second-order catchment (WS3) in the Andrews Forest, Oregon. Observed runoff over the 1996 water year was compared to predictions for runoff timing and interception of a hillslope water table based on a simple model of kinematic subsurface storm flow. Observed runoff behavior was consistent with model estimates, a finding that underscores the utility of this simple approach for predicting and explaining runoff dynamics on forest roads constructed on steep hillslopes. Road segments in the study area interacted in at least four distinct ways with complex landforms, potentially producing very different effects depending on landform characteristics. Hillslope length, soil depth, and cutbank depth explained much of the variation in road runoff production among subcatchments and among storm events. Especially during large storm events, a majority of instrumented road segments intercepted subsurface flow and routed it to ditches and thence directly to streams with a timing that contributed to the rising limb of the catchment-wide hydrograph. The approach used in this study may be useful for model development and for targeting road segments for removal or restoration

Displacement Kerma Cross Sections for Neutron Interactions in Molybdenum

Modifications to the displacement kerma cross section methods employed in the NJOY99 nuclear data processing code are described. Calculations were performed with the modified code for molybdenum using ENDF-6 neutron interaction data. Results are presented for a range of plausible Ed values

A method for neutron dosimetry in ultrahigh flux environments

A method for neutron dosimetry in ultrahigh flux environments is developed, and devices embodying it are proposed and simulated using a Monte Carlo code. The new approach no longer assumes a linear relationship between the fluence and the activity of the nuclides formed by irradiation. It accounts for depletion of the original ``foil`` material and for decay and depletion of the formed nuclides. In facilities where very high fluences are possible, the fluences inferred by activity measurements may be ambiguous. A method for resolving these ambiguities is also proposed and simulated. The new method and proposed devices should make possible the use of materials not traditionally considered desirable for neutron activation dosimetry

Neutronics methods, models, and applications at the Idaho National Engineering Laboratory for the advanced neutron source reactor three-element core design

A summary of the methods and models used to perform neutronics analyses on the Advanced Neutron Source reactor three-element core design is presented. The applications of the neutral particle Monte Carlo code MCNP are detailed, as well as the expansion of the static role of MCNP to analysis of fuel cycle depletion calculations. Results to date of these applications are presented also. A summary of the calculations not yet performed is also given to provide a {open_quotes}to-do{close_quotes} list if the project is resurrected

Minerva User Manual Version 1.0

MINERVA (Modality-Inclusive Environment for Radiotherapeutic Variable Analysis) is a Java-based patient-centric radiation treatment planning system (RTPS) for computational dosimetry and treatment planning in emerging areas of radiotherapy for cancer and other diseases. MINERVA was primarily developed at the Idaho National Laboratory (INL) and Montana State University (MSU). MINERVA allows the radiotherapist to make side-by-side comparison of plans for multiple treatment modalities with a common anatomical basis for the computational geometry, calculate doses for combinations of different radiotherapy modalities, and perform dose analysis and reporting functions. This provides the therapist with a consistent basis for selecting the modality or combination of modalities to use for treatment of the patient. MINERVA employs an integrated, lightweight plug-in architecture to accommodate multi-modal treatment planning using standard interface components. The MINERVA design facilitates integration of improved or emerging treatment planning technologies. MINERVA consists of the basic radiation treatment planning software modules managed by a consistent patient interface for developing multi-modal radiotherapy patient treatment plans. One of MINERVA's main functions is to provide a graphical environment for constructing and displaying uniform volume-element-based solid models derived from medical images. These solid models form the geometric basis of the target areas for the radiation transport model

Non-linear optical susceptibilities, Raman efficiencies and electrooptic tensors from first-principles density functional perturbation theory

The non-linear response of infinite periodic solids to homogenous electric fields and collective atomic displacements is discussed in the framework of density functional perturbation theory. The approach is based on the 2n + 1 theorem applied to an electric-field-dependent energy functional. We report the expressions for the calculation of the non-linear optical susceptibilities, Raman scattering efficiencies and electrooptic coefficients. Different formulations of third-order energy derivatives are examined and their convergence with respect to the k-point sampling is discussed. We apply our method to a few simple cases and compare our results to those obtained with distinct techniques. Finally, we discuss the effect of a scissors correction on the EO coefficients and non-linear optical susceptibilities

Reactor physics methods, models, and applications used to support the conceptual design of the Advanced Neutron Source

This report summarizes the neutronics analysis performed during 1991 and 1992 in support of characterization of the conceptual design of the Advanced Neutron Source (ANS). The methods used in the analysis, parametric studies, and key results supporting the design and safety evaluations of the conceptual design are presented. The analysis approach used during the conceptual design phase followed the same approach used in early ANS evaluations: (1) a strong reliance on Monte Carlo theory for beginning-of-cycle reactor performance calculations and (2) a reliance on few-group diffusion theory for reactor fuel cycle analysis and for evaluation of reactor performance at specific time steps over the fuel cycle. The Monte Carlo analysis was carried out using the MCNP continuous-energy code, and the few- group diffusion theory calculations were performed using the VENTURE and PDQ code systems. The MCNP code was used primarily for its capability to model the reflector components in realistic geometries as well as the inherent circumvention of cross-section processing requirements and use of energy-collapsed cross sections. The MCNP code was used for evaluations of reflector component reactivity effects and of heat loads in these components. The code was also used as a benchmark comparison against the diffusion-theory estimates of key reactor parameters such as region fluxes, control rod worths, reactivity coefficients, and material worths. The VENTURE and PDQ codes were used to provide independent evaluations of burnup effects, power distributions, and small perturbation worths. The performance and safety calculations performed over the subject time period are summarized, and key results are provided. The key results include flux and power distributions over the fuel cycle, silicon production rates, fuel burnup rates, component reactivities, control rod worths, component heat loads, shutdown reactivity margins, reactivity coefficients, and isotope production rates

Some recent developments in treatment planning software and methodology for BNCT

Over the past several years/the Idaho National Engineering Laboratory (INEL) has led the development of a unique, internationally-recognized set of software modules (BNCT rtpe) for computational dosimetry and treatment planning for Boron Neutron Capture Therapy (BNCT). The computational capability represented by this software is essential to the proper administration of all forms of radiotherapy for cancer. Such software addresses the need to perform pretreatment computation and optimization of the radiation dose distribution in the target volume. This permits the achievement of the optimal therapeutic ratio (tumor dose relative to critical normal tissue dose) for each individual patient via a systematic procedure for specifying the appropriate irradiation parameters to be employed for a given treatment. These parameters include angle of therapy beam incidence, beam aperture and shape,and beam intensity as a function of position across the beam front. The INEL software is used for treatment planning in the current series of human glioma trials at Brookhaven National Laboratory (BNL) and has also been licensed for research and developmental purposes to several other BNCT research centers in the US and in Europe

MINERVA - A Multi-Modal Radiation Treatment Planning System

Recently, research efforts have begun to examine the combination of BNCT with external beam photon radiotherapy (Barth et al. 2004). In order to properly prepare treatment plans for patients being treated with combinations of radiation modalities, appropriate planning tools must be available. To facilitiate this, researchers at the Idaho National Engineering and Environmental Laboratory (INEEL)and Montana State University (MSU) have undertaken development of a fully multi-modal radiation treatment planning system