The AGR-1 Advanced Gas Reactor (AGR) tristructural-isotropic-particle fuel experiment underwent 13 irradiation intervals from December 2006 until November 2009 within the Idaho National Laboratory Advanced Test Reactor in support of the Next Generation Nuclear Power Plant program. During this multi-year experiment, release-to-birth rate ratios were computed at the end of each operating interval to provide information about fuel performance. Fission products released during irradiation were tracked daily by the Fission Product Monitoring System using 8-h measurements. Birth rate calculated by MCNP with ORIGEN for as-run conditions were computed at the end of each irradiation interval. Each time step in MCNP provided neutron flux, reaction rates and AGR-1 compact composition, which were used to determine birth rate using ORIGEN. The initial birth-rate data, consisting of four values for each irradiation interval at the beginning, end, and two intermediate times, were interpolated to obtain values for each 8-h activity. The problem with this method is that any daily changes in heat rates or perturbations, such as shim control movement or core/lobe power fluctuations, would not be reflected in the interpolated data and a true picture of the system would not be presented. At the conclusion of the AGR-1 experiment, great efforts were put forth to compute daily birthrates, which were reprocessed with the 8-h release activity. The results of this study are presented in this paper.

This poster is one of two complementary posters. The Next Generation Safeguards Initiative (NGSI) of the U.S. DOE has initiated a multi-lab/university collaboration to quantify the plutonium (Pu) mass in, and detect the diversion of pins from, spent nuclear fuel assemblies with non-destructive assay (NDA). This research effort has the goal of quantifying the capability of 14 NDA techniques as well as training a future generation of safeguards practitioners. By November of 2010, we will be 1.5 years into the first phase (2.5 years) of work. This first phase involves primarily Monte Carlo modelling while the second phase (also 2.5 years) will focus on experimental work. The goal of phase one is to quantify the detection capability of the various techniques for the benefit of safeguard technology developers, regulators, and policy makers as well as to determine what integrated techniques merit experimental work, We are considering a wide range of possible technologies since our research horizon is longer term than the focus of most regulator bodies. The capability of all of the NDA techniques will be determined for a library of 64 17 x 17 PWR assemblies [burnups (15, 30, 45, 60 GWd/tU), initial enrichments (2, 3, 4, 5%) andmore » cooling times (1, 5, 20, 80 years)]. The burnup and cooling time were simulated with each fuel pin being comprised of four radial regions. In this paper an overview of the purpose will be given as well as a technical update on the following 6 neutron techniques: {sup 252}Cf Interrogation with Prompt Neutron Detection, Delayed Neutrons, Differential Die-Away, Differential Die-Away Self-Interrogation, Passive Neutron Albedo Reactivity, Self-Integration Neutron Resonance Densitometry. The technical update will quantify the anticipated performance of each technique for the 64 assemblies of the spent fuel library.« less

Energetic photon sources with energies greater than 6 MeV continue to be recognized as viable source for various types of inspection applications, especially those related to nuclear and/or explosive material detection. These energetic photons can be produced as a continuum of energies (i.e., bremsstrahlung) or as a set of one or more discrete photon energies (i.e., monoenergetic). This paper will provide a follow-on extension of the photon dose comparison presented at the 9(th) International Conference on Applications of Nuclear Techniques (June 2008). Our previous paper showed the comparative advantages and disadvantages of the photon doses provided by these two energetic interrogation sources and highlighted the higher energy advantage of the bremsstrahlung source, especially at long standoff distances (i.e., distance from source to the inspected object). This paper will pursue higher energy photon inspection advantage (up to 100 MeV) by providing dose and stimulated photonuclear interaction predictions in air and for an infinitely dilute interrogated material (used for comparative interaction rate assessments since it excludes material self-shielding) as the interrogation object positioned forward on the inspection beam axis at increasing standoff distances. In addition to the direct energetic photon-induced stimulation, the predictions will identify the importance of secondary downscattered/anenuated source-term effects arising from the photon transport in the intervening air environment.

Fast neutron dose recommendations for the graphite moderators of Magnox power plant are expressed in terms of Equivalent DIDO Nickel Dose (EDND). EDND is a calculated property that relates the damage rate in graphite at a given location to the fast neutron fluence experienced by graphite test samples irradiated in the DIDO reactor. This paper describes a study undertaken to determine the sensitivity of EDND recommendations to assumptions made in the modeling of neutron damage in graphite. These include different scattering cross sections and damage weighting functions. It is concluded that, although apparently significant differences exist between the available graphite damage response functions, their net effect is not significant in determining EDND. Therefore, confidence in past calculations has been improved and continued use of existing methods is supported.

An MCNPX-based calculational methodology has been developed to numerically simulate the complex electron–photon–neutron transport problem for the active interrogation system known as the pulsed photonuclear assessment (PPA) technique. The PPA technique uses a pulsed electron accelerator to generate bremsstrahlung photons in order to fission nuclear materials. Delayed neutron radiation is then detected with helium-3 neutron detectors as evidence of the nuclear material presence. Two experimental tests were designed, setup and run to generate experimental data for benchmarking purposes. The first test irradiated depleted uranium in air, and the second test, depleted uranium in a simulated cargo container (plywood pallet), using 10MeV electron pulses. Time-integrated, post-flash, delayed neutron counts were measured and compared to calculated count predictions in order to benchmark the calculational methodology and computer models. Comparisons between the experimental measurements and numerical predictions of the delayed neutron detector responses resulted in reasonable experiment/calculated ratios of 1.42 and 1.06 for the two tests. High-enriched uranium (HEU) predictions were also made with the benchmarked models.

Idaho National Laboratory, along with Idaho State University’s Idaho Accelerator Center and Los Alamos National Laboratory, is developing an electron accelerator-based, photonuclear inspection technology, called the Pulsed Photonuclear Assessment (PPA) system, for the detection of nuclear material concealed within air-, rail-, and, primarily, maritime-cargo transportation containers. This report summarizes the advances and progress of the system’s development in 2005. The contents of this report include an overview of the prototype inspection system, selected Receiver-Operator-Characteristic curves for system detection performance characterization, a description of the approach used to integrate the three major detection components of the PPA inspection system, highlights of the gray-scale density mapping technique being used for significant shield material detection, and higher electron beam energy detection results to support an evaluation for an optimal interrogating beam energy. This project is supported by the Department of Homeland Security Office of Research and Development and, more recently, the Domestic Nuclear Detection Office.

The Idaho National Laboratory has developed prototype shielded nuclear material detection systems based on pulsed photonuclear assessment (PPA) techniques for the inspection of cargo containers. During this work, increased nuclear material detection capabilities have been demonstrated at higher electron beam energies than those allowed by federal regulations for cargo inspection. This paper gives a general overview of a nuclear material detection system, the PPA technique and discusses the benefits of using these higher energies. This paper also includes a summary of the numerical and test results from LINAC operations up to 24MeV and discusses some of the federal energy limitations associated with cargo inspection.

The supercritical water-cooled nuclear reactor (SCWR) concept offers potential for superior economics due to its high thermal efficiency and plant simplification. However, design of a thermal-spectrum core for such a reactor is complicated by the relatively low density of the water coolant and therefore reduced moderation. This requires the SCWR design to include a dedicated moderator. One solution explored worldwide is based on the use of water rods. In this paper we assess the feasibility of a different approach based on solid moderators, which has some potential advantages including increased core thermal capacity, reduced coolant worth, and simplified vessel internals. The neutronic performance of several solid moderators was evaluated and compared to that of water rods. It was found that the only acceptable solid moderator is zirconium hydride. Axial and local peaking can be readily suppressed by modest variations of the enrichment in a manner similar to the boiling water reactor practice. The Doppler and coolant reactivity coefficients are both negative and in the range of light water reactor experience. The use of zirconium hydride as a stable structural core component was evaluated and found to be acceptable under steady-state and accident conditions. In addition to its chemical and mechanical stability, zirconium hydride can also be fabricated with existing technology. However, its impact on the SCWR cost of electricity generation is deemed significant.

An experimental. and calculational study has been performed to understand and prevent inadvertent activation of the criticality alarm system (CAS) from fuel-handling operations at the Irradiated Fuel Storage Facility. In conjunction with the study, the CAS neutron detectors were tested to verify the design specifications for gamma rejection capability and zero response limit. A minimum physical restrictive boundary around the CAS location was established based on a gamma ray dose rate limit of 10 rad/hr. The canister loaded with spent nuclear fuel must be moved in the area outside the exclusion zone so as not to trigger a false alarm from the CAS detectors.