Naeem M. Abdurrahman

Spent-fuel assay performance and Monte Carlo analysis of the Rensselaer slowing-down-time spectrometer

The slowing-down-time method for the nondestructive assay of light water reactor (LWR) spent fuel is under development at Rensselaer Polytechnic Institute. A series of assay measurements of an LWR fuel assembly replica were carried out at the Rensselaer lead slowing-down-time spectrometer facility by using [sup 238]U and [sup 232]Th threshold fission detectors and [sup 235]U and [sup 239]Pu probe chambers. An assay model relating the assay signal and the signals of the probe chambers to the unknown masses of the fissile isotopes in the fuel assembly was developed. The probe chamber data were used to provide individual fission counting spectra of [sup 235]U and [sup 239]Pu inside the fuel assembly and to simulate spent-fuel assay signals. The fissile isotopic contents of the fuel were determined to better than 1%. Monte Carlo analyses were performed to simulate the experimental measurements, determine certain parameters of the assay system, and investigate the effect of the fuel assembly and hydrogen impurities on the performance of the system. The broadened resolution of the system caused by the presence of the fuel was still found to be sufficient for the accurate and separate assay of the uranium and plutonium fissiles in spent fuel.

Design of a spent-fuel assay device using a lead spectrometer

The neutron slowing-down-time method for nondestructive assay of light water reactor spent fuel has been under development for many years. Results for a newly optimized design of a lead slowing-down-time spectrometer for spent-nuclear-fuel assay are presented. Monte Carlo analyses were performed to optimize the design of the assay device, determine its main parameters, investigate the effects of the spent-fuel assembly and the detector impurities on its performance, determine the fission signatures of the fissile isotopes in spent-fuel elements, and simulate the assay signal as a function of the slowing-down time, assuming threshold fission chambers for the assay detectors. The assay signals from the threshold detectors were analyzed to predict the unknown masses of the fissile isotopes in a typical spent commercial light water reactor fuel element. The broadened resolution of the system caused by the presence of the spent fuel inside the spectrometer pile was found sufficient to separate the signatures of the U and Pu fissiles in spent fuel.

Neutron Spectrometry for the Assay of High Fissile Content Spent Fuel

Abstract A Monte Carlo study of the neutron slowing-down spectrometry technique for measuring fissile isotopic content in irradiated fuel has been completed. The neutron spectrometer system is characterized in terms of design, slowing-down time relation, isotopic response functions, and assay signals. The nonlinear effect of interrogating neutron self-shielding for a high fissile content fuel is compared to the same parameter for a low fissile content fuel. Simulated assays of 23 different fuel assemblies with a broad range of total fissile mass content (1.3 to 83 wt%) and fissile isotopic ratios are performed and analyzed using two different methods: a linear system model using a least-squares regression analysis and a radial basis neural network. Mean errors using the linear system model for the 23 different fuel types were approximately 20% for 235U and 43% for total plutonium. The radial basis neural network assay signal solutions showed promising results, considerably better than the linear model: 4.9% for 235U, 5.4% for total plutonium, and 0.5% for total fissile content.

Neutron tomographic fissile assay in spent fuel using the lead slowing down time spectrometer

Abstract A tomographical fissile assay technique for light water reactor spent fuels was developed based on the lead slowing down time spectrometer (LSDTS). The Monte Carlo method was used for system performance and for simulation of the sensitivity of detection of fissile materials using neutron emission tomography. A typical spent PWR fuel element was simulated and the spatial and mass sensitivity of the fissile components were investigated. From these studies, the LSDTS system is shown to be a very sensitive device for analyzing the spatial distribution of total fissile materials of a spent fuel assembly in a properly selected assay neutron energy range. This method is also applicable to nuclear waste assay.

Benchmark calculations of the saxton plutonium program critical experiments

The Saxton critical experiments, which used mixed-oxide (MOX) fuel of 6.6 wt% PuO 2 in natural UO 2 and UO 2 fuel of 5.74 wt% 235 U, are analyzed with MCNP-4B and continuous-energy cross-section libraries ENDF/ B-V and ENDF/B-VI. An excellent agreement of calculated and experimental effective multiplication factors for the entire set of 1.3208-cm MOX lattices and 1.4224-cm MOX and UO 2 lattices was obtained. The analysis of criticality calculations for the five different lattice pitches show a bias with lattice pitch, which led to an increase of ∼0.8% when doubling the lattice pitch. Good agreement between calculated and measured data was obtained for some of the relative power distribution experiments for MOX single-region cores and MOX/UO 2 multiregion cores; however, for others the agreement was less satisfactory. No significant difference in the results for relative power with the two libraries was observed.

Benchmark calculations of the ESADA single-region mixed-oxide critical experiments

In 1967, a series of critical experiments were conducted at the Westinghouse Reactor Evaluation Center under the joint sponsorship of the Empire State Atomic Development Associates (ESADA) and Westinghouse using mixed-oxide (MOX) fuel. During the experimental program, both single- and multiregion critical core configurations were constructed for different fuel types and lattice pitches. Two types of MOX fuels and a low-enriched UO 2 fuel were used. A description of selected single-region ESADA experiments with criticality benchmark calculation results for those experiments as well as sensitivity analysis results on some configurations is given. Criticality calculations were performed using MCNP-4A with both ENDF/B-V and ENDF-B/VI cross-section libraries. The calculational results show that the calculated eigenvalues with ENDF/B-V cross-section libraries are higher than calculated eigenvalues with ENDF/B-VI cross-section libraries for all types of fuel. Calculational results also indicate that there is an increase in k eff with increasing lattice pitch with both cross-section libraries. The uncertainties in k eff value due to some uncertainties in the measured data are also calculated.

Neutronics Benchmarks for the Utilization of Mixed Oxide Fuel in Water Reactors

This paper presents results of criticality calculations performed for three sets of benchmark problems conducted with four computer code systems. The problems presented in this paper include both experimental and computational benchmarks. These calculations were performed for the Amarillo Resource Center for Plutonium (ANRCP) in conjunction with the Oak Ridge National Laboratory (ORNL). Each of the following sections contains an abbreviated problem statement, a brief description of the calculational methodology, and primary results. MCNP-4A with ENDF/B-VI is used for the Monte Carlo calculations. WIMSD4-m/TWODANT, WIMSD4-m/DIF3D, and CASMO-3 are used for different parts of the deterministic calculations. Due to space constraints, the descriptions given below are quite brief. Please contact the authors if further information is desired.