Theodore A. Parish

Reduction in the Toxicity of Fission Product Wastes Through Transmutation with Deuterium-Tritium Fusion Neutrons

A waste management concept that employs irradiation of fission products in fusion reactor blankets is evaluated. The purpose of the irradiation is to reduce the toxicity of the material to be commi...

Operator declaration verification technique for spent fuel at reprocessing facilities

Abstract A verification technique for use at reprocessing facilities, which integrates existing technologies to strengthen safeguards through the use of environmental monitoring, has been developed at Los Alamos National Laboratory. This technique involves the measurement of isotopic ratios of stable noble fission gases from on-stack emissions during reprocessing of spent fuel using high-precision mass spectrometry. These results are then compared to a database of calculated isotopic ratios using a data analysis method to determine specific fuel parameters (e.g., burnup, fuel type, reactor type, etc.). These inferred parameters can be used to verify operator declarations. The integrated system (mass spectrometry, reactor modeling, and data analysis) has been validated using on-stack measurements during reprocessing of fuel from a US production reactor. These measurements led to an inferred burnup that matched the declared burnup to within 3.9%, suggesting that the current system is sufficient for most safeguards applications. Partial system validation using gas samples from literature measurements of power reactor fuel has been reported elsewhere. This has shown that the technique developed here may have some difficulty distinguishing pressurized water reactor (PWR) from boiling water reactor (BWR) fuel; however, it consistently can distinguish light water reactor (either PWR or BWR) fuels from other reactor fuel types. Future validations will include advanced power reactor fuels (such as breeder reactor fuels) and research reactor fuels as samples become available.

Measurements of delayed neutron decay constants and fission yields from 235U, 237Np, 241Am, and 243Am

Delayed neutron yields and decay constants for 235 U, 237 Np, 241 Am, and 243 Am were measured at the Texas AM Waldo, Karam, and Meyer; and Tuttle. Very good agreement was obtained, especially for 235 U.

AN EXTENDED STEP CHARACTERISTIC METHOD FOR SOLVING THE TRANSPORT EQUATION IN GENERAL GEOMETRIES

A method for applying the discrete ordinates method to solve the Boltzmann transport equation on arbitrary two-dimensional meshes has been developed. The finite difference approach normally used to approximate spatial derivatives in extrapolating angular fluxes across a cell is replaced by direct solution of the characteristic form of the transport equation for each discrete direction. Thus, computational cells are not restricted to the geometrical shape of a mesh element characteristic of a given coordinate system. However, in terms of the treatment of energy and angular dependencies, this method resembles traditional discrete ordinates techniques. By using the method developed here, a general two-dimensional space can be approximated by an irregular mesh comprised of arbitrary polygons. Results for a number of test problems have been compared with solutions obtained from traditional methods, with good agreement. Comparisons include benchmarks against analytical results for problems with simple geometry, as well as numerical results obtained from traditional discrete ordinates methods by applying the ANISN and TWOTRAN-II computer programs.

Status of six-group delayed neutron data and relationships between delayed neutron parameters from the macroscopic and microscopic approaches

Work performed in part for an American Nuclear Society Standards Committee Subgroup (ANS 19.9) to assess the status of delayed neutron data is summarized. Recent measurements of delayed neutron emission conducted at Texas A and M University are also described. During the last 10 yr, there have been advances in nuclear data libraries (e.g., improved fission product yields) that make it possible to quantitatively predict delayed neutron emission from basic data. The six-group delayed neutron data available in the literature from both macroscopic level experiments and microscopic level calculations for several actinide isotopes are compared. Results are also presented from recent experimental measurements of delayed neutron emission and delineates some of the relationships between these measurements and microscopic level predictions. For example, from the experimental measurements, Keepin`s delayed neutron group 1 is shown to correspond mainly to a single isotope. {sup 87}Br, as expected from microscopic level theory, and Keepin`s group 2 is shown to correspond to primarily two separate isotopes. {sup 137}I and {sup 88}Br. In the future, it may be useful to use properties of specific isotopes to replace Keepin`s delayed neutron groups 1, 2, 3, and 4 for prescribing delayed neutron data for actinides.

Calculated Actinide and Fission Product Concentration Ratios for Gaseous Effluent Monitoring Using Monteburns 3.01

Techniques have been developed at Los Alamos National Laboratory for accurately calculating certain spent-fuel isotope concentration ratios for pressurized water reactor assemblies using a linked MCNP/ORIGEN2 code named Monteburns 3.01, without resorting to an assembly or full-core calculation. The effects of various fuel parameters such as the number of radial fuel regions per pin, burnup step size, reactor power, reactivity control mechanisms, and axial profiles have been studied. The significance of each factor was determined. A method was also proposed for calculating spent-fuel inventories as a function of burnup for a wide range of reactors and fuel types. It was determined that accurate calculations can be obtained using a three-dimensional, modified pin cell with seven radial fuel regions and two (flat-flux) axial fuel regions calculated with 2000 MWd/tonne U burnup steps for burnups ranging from 0 to 50 000 MWd/tonne U. The calculational technique was benchmarked to measured values from the Calvert Cliffs Unit 1 reactor, and good agreement from the point of view of calibrating a monitoring instrument was found for most cases.

Derivation and Implementation of the Three-Region Problem in the SOURCES Code System

The SOURCES code system was designed to calculate neutron production rates and spectra in materials due to the decay of radionuclides [specifically from (α,n) reactions, spontaneous fission, and delayed neutron emission]. The current version (SOURCES-3 A) is capable of calculating (α,n) source rates and spectra for three types of problems: homogeneous materials, interface problems, and beam problems. Recent interest in (α,n) sources has prompted the development of a fourth scenario: the three-region problem. To allow SOURCES to confront this problem, the relevant equations defining the α-particle source rates and spectra at each interface and the neutron source rates and spectra per unit area of interface were derived. These equations (in discretized form) were added as a new subroutine to the SOURCES code system (dubbed SOURCES-4A). The new code system was tested by analyzing the results for a simple three-region problem in two limits: with an optically thin ‘intermediate region’ and with an optically thick ‘intermediate region.’ To further validate the code system, SOURCES-4A will be experimentally benchmarked as measured data becomes available.

Measurements of nuclear data of minor actinides for transmutation of high-level waste

For nuclear transmutation of minor actinides, delayed neutron emission measurement for241Am was carried out in thermal neutron irradiation location. The neutron capture cross sections of241Am were also measured radiochemically. The transmutation process of241Am in reactor is discussed by calculating the yields of minor actinides with the nuclear data measured in this study and the evaluated values. The accelerator driven transmutation of minor actinides by high-flux neutrons from spallation reactions is also presented.

Measured and calculated radionuclide production from copper, gold, and lead spallation targets

Abstract In order to verify the production rates of radionuclides calculated by the Los Alamos High Energy Transport code (LAHET), a series of experiments were conducted using the Texas A&M University Cyclotron to measure radionuclide yields from copper, gold, and lead targets bombarded by 120 MeV deuterons. The measured yields were compared to a set of calculated yields generated from LAHET simulations. Several irradiation times and decay times were used to improve sensitivity to both short-lived and long-lived nuclides. Agreement between the measured and calculated radionuclide yields was found to be generally within a factor of two.

Neutron production measurements from an LiD fusion plate using an NE-213 detector

Abstract To characterize the high energy neutron production from an LiD plate constructed at Texas A&M University, measurements and calculations were performed. Two types of measurements were used to substantiate the high energy neutron flux. One used an NE-213 detector system to measure the high energy neutron production from the LiD plate and the other employed foil activation to measure the neutron flux entering the irradiation cell. A new code, FERDNA, was produced by not only converting the original unfolding FERD code to Microsoft FORTRAN but also embellishing the code with several subroutines to perform integration over energy ranges specified by the user. Raw pulse-height distributions from an NE-213 detector were directly transferred from an MCA to a MacIntosh where FERDNA was executed. The LiD plate was designed and constructed by considering (1) neutron production due to 6 Li(n, α) 3 T reactions and subsequent T-D or T-Li fusion as the tritons slow down and (2) the escape probability of tritons from the LiD material. This allowed the needed 6 Li enrichment and thickness of the plate to be selected. The number of high energy neutrons produced per thermal neutron incident upon the LiD plate was measured as 1.03×10 −4 ±8.5% n/n th .