T.E. Valentine

Evaluation of prompt fission gamma rays for use in simulating nuclear safeguard measurements

Abstract Nondestructive assay methods that rely on measurement of correlated gamma rays from fission have been proposed as a means to determine the mass of fissile materials. Sensitivity studies for such measurements will require knowledge of the multiplicity of prompt gamma rays from fission; however, a very limited number of multiplicity distributions have been measured. A method is proposed to estimate the average number of gamma rays from any fission process by using the correlation of neutron and gamma emission in fission. Using this method, models for the total prompt gamma ray energy from fission adequately reproduce the measured value for thermal neutron induced fission of 233 U. Likewise, the average energy of prompt gamma rays from fission has been adequately estimated using a simple linear model. Additionally, a method to estimate the multiplicity distribution of prompt gamma rays from fission is proposed based on a measured distribution for 252 Cf. These methods are only approximate at best and should only be used for sensitivity studies. Measurements of the multiplicity distribution of prompt gamma rays from fission should be performed to determine the adequacy of the models proposed in this article.

Subcritical reactivity monitoring in accelerator driven systems

Abstract In this paper, an absolute measurements technique for the subcriticality determination is presented. The development of accelerator driven systems (ADS) requires the development of methods to monitor and control the subcriticality of this kind of system, without interfering with its normal operation mode. This method is based on the stochastic neutron and photon transport theory that can be implemented by presently available neutron transport codes. As a by-product of the methodology a monitoring measurement technique has been developed and verified using two coupled Monte Carlo programs. The first one, LAHET, simulates the spallation collisions and the high energy transport and the other, MCNPDSP, is used to estimate the counting statistics from neutron ray counter in fissile system, and the transport for neutrons with energies less than 20 Mev. Through the analysis of the counter detectors it is possible to determine the kinetics parameters and the k eff value. We present two different ways to obtain these parameters using the accelerator or using a Cf-252 source. A good agreement between theory and simulations has been obtained with both sources.

MCNP-DSP: A neutron and gamma ray Monte carlo calculation of source-driven noise-measured parameters

Abstract The 252Cf-source-driven noise analysis measurement method was developed to determine the subcriticality and other properties of configurations of fissile material. The method provides measured parameters that can also be used for nuclear weapons identification, nuclear materials control and accountability, quality assurance, process monitoring, and verification of calculation models and cross section data used for criticality safety analyses. MCNP-DSP was developed to calculate the measured frequency analysis parameters, time analysis quantities such as autocorrelation and cross-correlation functions, and the time distribution of counts after 252Cf fission for both neutrons and/or gamma rays using continuous energy cross sections. MCNP-DSP can be used to validate calculational methods and cross section data sets with measured data from subcritical experiments. In most cases the frequency analysis parameters are more sensitive to cross section changes by as much as 1 or 2 orders of magnitude and thus may be more useful than comparisons of neutron multiplication factors for calculational validation. The use of MCNP-DSP model in place of point kinetics model to interpret subcritical experiments extends the usefulness of this measurement method to systems with much lower neutron multiplication factors. MCNP-DSP can also be used to determine the calculational bias in the neutron multiplication factor (a quantity which is essential to the criticality safety specialist) from in-plant subcritical experiments. This paper describes how MCNP-DSP calculates the measured parameters from the 252Cf-source-driven time and frequency analysis measurement.

Modal influence of the detector location for the noise calculation of the ADS

Abstract The purpose of this paper is to investigate the space dependence of neutron noise in a source-driven subcritical system. The noise produced by the fluctuations of the source are measured from the cross spectrum between the source and a detector located inside the system using the methodology of Munoz-Cobo et al. (Munoz-Cobo, J.L., Rugama, Y., Valentine, T., Mihalczo, J., Perez, R., 2002. Annals of Nuclear Energy in press). The prompt neutron decay constant obtained from the source-detector cross spectrum is dependent on the detector location because of the influence of higher modes of the neutron flux. One group diffusion theory is used to determine the eigenvalues and eigenfunctions of the fundamental and higher modes. This analytical approximation will be used to explain the detector location effect through quantitative evaluation of a specific model. The analytical system subcriticality will be obtained from the eigenvalue equation in the static case at two different subcritical situations.

A stochastic transport theory of neutron and photon coupled fields: neutron and photon counting statistics in nuclear assemblies

Abstract The behavior of neutrons and gamma rays in a nuclear reactor or configuration of fissile material can be represented as a stochastic process. The observation of this stochastic process is usually achieved by measuring the fluctuations of the neutron and gamma ray population on the system. The general theory of the stochastic neutron field has been developed to a high degree. However, the theory of the stochastic nature of the gamma rays and neutrons couples the two processes. The generalized probability balances are developed from which the first and higher moments of the neutron and gamma rays fields are obtained. The paper also provides a description of the probability generating functions for both photon and neutron detectors that are the foundations for measurements of the fluctuations. The formalism developed in this paper for the representation of the statistical descriptors of the neutron-photon coupled field is applicable for many neutron noise analysis measurements.

252Cf-source-correlated transmission measurements for uranyl fluoride deposit in a 24-in-OD process pipe

Characterization of a hydrated uranyl fluoride (UO{sub 2}F{sub 2}{center_dot}nH{sub 2}O) deposit in a 17-ft-long, 24-in.-OD process pipe at the former Oak Ridge Gaseous Diffusion Plant was successfully performed by using {sup 252}Cf-source-correlated time-of-flight (TOF) transmission measurements. These measurements of neutrons and gamma rays through the pipe from an external {sup 2521}Cf fission source were used to measure the deposit profile and its distribution along the pipe, the hydration (or H/U), and the total uranium mass. The measurements were performed with a source in an ionization chamber on one side of the pipe and detectors on the other. Scanning the pipe vertically and horizontally produced a spatial and time-dependent radiograph of the deposit in which transmitted gamma rays and neutrons were separated in time. The cross-correlation function between the source and the detector was measured with the Nuclear Weapons Identification System. After correcting for pipe effects, the deposit thickness was determined from the transmitted neutrons and H/U from the gamma rays. Results were consistent with a later intrusive observation of the shape and the color of the deposit; i.e., the deposit was annular and was on the top of the pipe at some locations, demonstrating the usefulness of this method for deposit characterization.

New Neutron Cross‐Section Measurements at ORELA for Improved Nuclear Data Calculations

The Oak Ridge Electron Linear Accelerator (ORELA) was used to measure neutron total and capture cross sections of aluminum, silicon, chlorine, fluorine, and potassium in the energy range from 100 eV to ~600 keV. These measurements were carried out to support the Nuclear Criticality Safety Program (NCSP). Concerns about the use of existing cross section data in nuclear criticality calculations have been a prime motivator for the new cross section measurements. Our results are substantially different from the evaluated nuclear data files of ENDF/B-VI and JENDL-3.2.

High-Resolution Transmission Measurements of 233U Using a Cooled Sample at the Temperature T=11 K

Abstract For the first time, high-resolution transmission data of 233U have been obtained using a cooled sample. The samples were cooled to T = 11 K using a cryogenic device, which reduced the Doppler broadening of resonances by 50% compared to room-temperature measurements. The measurements were carried out at the Oak Ridge Electron Linear Accelerator over the energy range from 0.6 eV to 300 keV at the 80-m flight path station. Corrections were made for experimental effects, and the average total cross section in this energy range was determined. Results are compared to previous measurements.

Neutron Cross Sections Measurements for Light Elements at ORELA and their Application in Nuclear Criticality

The Oak Ridge Electron Linear Accelerator (ORELA) was used to measure neutron total and capture cross sections of aluminium, natural chlorine and silicon in the energy range from 100 eV to ~600 keV. ORELA is the only high power white neutron source with excellent time resolution and ideally suited for these experiments still operating in the USA. These measurements were carried out to support the Nuclear Criticality Predictability Program. Concerns about the use of existing cross section data in the nuclear criticality calculations using Monte Carlo codes and benchmarks have been a prime motivator for the new cross section measurements. More accurate nuclear data are not only needed for these calculations but also serve as input parameters for s-process stellar models.

Subcritical Measurements of a Plutonium Sphere Reflected by Polyethylene and Acrylic

Abstract Subcritical measurements were conducted with an alpha-phase plutonium sphere using the 252Cf source-driven noise analysis method. Measurements were performed with both polyethylene and acrylic reflectors. For each reflector type, five different reflector thicknesses were investigated: 0 (bare), 1.27, 2.54, 3.81, and 7.62 cm. A certain ratio of spectral quantities that depends on the fluctuations in the fission chain multiplication process was measured for each configuration. In addition, two types of Monte Carlo calculations were employed to estimate the keff and spectral ratio values of each configuration. From the measured and computed quantities, the multiplication and uncertainty of the system can be inferred. The polyethylene measurements compared well to previous measurements conducted with the same plutonium sphere and polyethylene reflector thicknesses. The acrylic measurements provide benchmark data of an alpha-phase plutonium sphere reflected by acrylic.