Fission Product Gamma Rays Research Articles

Method for calculating delayed gamma-ray response in the ACRR Central Cavity using MCNP

This document presents the process for a new method developed for the characterization of the delayed gamma-ray radiation fields in pulse reactors like the Annular Core Research Reactor (ACRR). The environments used to test this method in the ACRR Central Cavity were various standard irradiation configurations that used various amounts of moderators and shielding. All environment configurations used the same fission product gamma-ray source energy spectrum. This method required the use of fission sites recorded in the Monte Carlo N-Particle (MCNP) KCODE source tape. A FORTRAN script was written to translate and extract the coordinates for the fission sites. The fission sites were then input into an MCNP source mode input file. Using a MATLAB script, a parametric analysis was performed and determined that 10K fission sites are an appropriate number of coordinates to converge to the correct answer. The method gave excellent results as compared to previous methods. This method can be applied to other pulse research reactors as well.

Signal Processing and Data Acquisition for the UNM Fission Spectrometer to Measure Binary Fission Product Mass, Energy, Velocity, Atomic Number, and Gamma Rays, Correlated Particle-By-Particle

Signal Processing and Data Acquisition for the UNM Fission Spectrometer to Measure Binary Fission Product Mass, Energy, Velocity, Atomic Number, and Gamma Rays, Correlated Particle-By-Particle

Measurement of the gamma-ray energy spectrum of the educational Kinki University Reactor (UTR-KINKI)

The gamma-ray energy spectrum of the Kinki University Reactor (UTR-KINKI) was estimated from Ge detector measurements combined with Monte Carlo N-particle transport criticality calculations. The gamma rays mainly originated from prompt fission components, although small amounts of gamma rays from (n,γ) reactions, fission product gamma rays, and activation gamma rays were detected. The averaged gamma-ray tissue kerma rate in the irradiation port during UTR-KINKI operation at 1W was calculated as 10.5cGy/h based on the estimated gamma-ray energy spectrum. This value is consistent with a previous measurement with paired ionization chambers and a tissue equivalent gas proportional counter. This result demonstrates the reliability of the estimated gamma-ray energy spectrum.

A generalized method for characterization of 235U and 239Pu content using short-lived fission product gamma spectroscopy

Rapid nondestructive assay methods for trace fissile material analysis are needed in both nuclear forensics and safeguards communities. To address these needs, research at the Oak Ridge National Laboratory High Flux Isotope Reactor Neutron Activation Analysis facility has developed a generalized nondestructive assay method to characterize materials containing fissile isotopes. This method relies on gamma-ray emissions from short-lived fission products and makes use of differences in fission product yields to identify fissile compositions of trace material samples. Although prior work has explored the use of short-lived fission product gamma-ray measurements, the proposed method is the first to provide a complete characterization of isotopic identification, mass ratios, and absolute mass determination. Successful single fissile isotope mass recoveries of less than 6% recovery bias have been conducted on standards of 235U and 239Pu as low as 12ng in less than 10 minutes. Additionally, mixtures of fissile isotope standards containing 235U and 239Pu have been characterized as low as 198ng of fissile mass with less than 7% recovery bias. The generalizability of this method is illustrated by evaluating different fissile isotopes, mixtures of fissile isotopes, and two different irradiation positions in the reactor. It is anticipated that this method will be expanded to characterize additional fissile nuclides, utilize various irradiation facilities, and account for increasingly complex sample matrices.

Differentiating special nuclear materials through computationally generated gamma-ray spectra

This work focuses on providing a first step interpretation of fission product gamma-ray spectra from 0.1 h to 10 days after a fission event. The activity of fission products from mass chains 89 through 92 and 131 through 141 was evaluated using independent fission yields. An MCNP model was used to simulate the gamma-ray spectra from a single HPGe crystal for the radionuclides of interest for a sample of U-235 and Pu-239 assuming a fission neutron spectrum. The ratios of the activities were evaluated for differences in the spectrum. The analysis was optimized for radionuclides with prominent gamma-ray branches and a large difference in intensities between U-235 fission versus Pu-239 fission.

Spectroscopy of high rate events during active interrogation

Pulsed accelerators are being used for radiographic inspections in homeland security applications. Adding spectroscopic detection capability enables additional information to be extracted but challenges signal processing hardware. This work compares two approaches to cost-effective digital spectroscopy systems designed to handle these high rate scenarios. In the first system, the signal from a PMT is directly digitized and then deconvolved on a PC using standard techniques. Bypassing the preamplifier, this system can handle high count rates (>10 6 cps) from an LaCl 3 detector with no dead time and an energy resolution of 5% at 1.33 MeV. Spectra acquired from fission product gamma rays between interrogation pulses are shown. At higher rates, this approach is limited by the DAQ transfer rate to the PC. Instead, we tested a fast ADC (125 MSPS) and an FPGA that performed onboard, real-time digital signal processing (DSP). The energy spectrum is generated on the FPGA board during the measurement and transferred later. This system was tested in nuclear resonance fluorescence (NRF) experiments. Resonance peaks from NRF reactions are shown.

Observation of /sup 238/U photofission products

Sophisticated solutions are being developed by researchers to improve detection of clandestine fissile material. In this work, a simpler approach is investigated as an interim solution. Using the existing 9-MeV X-ray radiography units deployed for cargo inspection, delayed fission product gamma rays from induced photofission are seen minutes after probing a small sample of /sup 238/U. By acquiring both time and energy information, a relatively large number of usable fission products can be identified. Using the radiographic images along with an efficient portable scintillator capable of detecting gamma rays above 3 MeV may provide a cost-effective improvement to existing inspection systems.

Measurements of Delayed Fission Product Gamma-Ray Transmission Through Low-Enriched UO2 Fuel Pin Lattices in Air

Measurements of delayed fission product gamma-ray transmission through low-enriched UO2 fuel pin lattices in an air medium were conducted at the Rensselaer Polytechnic Institute Reactor Critical Facility (RCF). The RCF core consists of excess Special Power Excursion Reactor Test fuel pins enriched to 4.81 wt% 235U and clad in stainless steel. An experimental apparatus was constructed to hold various arrangements of fuel pin lattices. The arrangements consisted of a single activated source pin taken from the reactor core surrounded by inactive fuel pins in an air medium. A sodium-iodide detector and gamma-ray spectroscopy system was used to generate a pulse-height spectrum of the gamma-ray radiation for detector positions outside the lattice. The change in radiation intensity as the detector is rotated about the vertical axis of the lattice, the “channeling effect,” was measured. Measurements of the channeling effect were performed for six experimental arrangements: 3 × 3, 5 × 5, and 7 × 7 lattices, with both the corner and the center positions containing the activated pin. The results of the measurements demonstrate that the gamma-ray radiation intensity can vary widely as a function of angle relative to the angle of rotation about the vertical axis of the lattice.

Analysis of fissile materials by high-energy neutron-induced fission decay gamma-rays

Author(s): Firestone, R.B.; English, G.A.; Reijonen, J.; Gicquel, F.; Leung, K.-N.; Perry, D.L.; Garabedian, G.; Bandong, B.; Revay, Zs.; Molnar, G.L. | Abstract: Thermal neutrons from the Budapest Research Reactor and fast neutrons from the Berkeley Neutron Generator Facility have been used to analyze uranium. It has been shown that both prompt and delayed gamma rays from neutron capture and fission product decay can be used to analyze uranium concentrations and 235U enrichment. Detection of neutrons from the spontaneous fission of 238U has also been demonstrated for uranium analysis. The observation of high-energy gamma rays following the decay of short-lived fission products is a sensitive indication of fissile material, and the ratio of fission product gamma ray intensities can uniquely determine the concentrations of fission isotopes.

Using the Monte Carlo Coupling Technique to Evaluate the Shielding Ability of a Modular Shielding House to Accommodate Spent-Fuel Transportable Storage Casks

The Monte Carlo coupling technique with the coordinate transformation is used to evalulate the shielding ability of a modular shielding house that accommodates four spent-fuel transportable storage casks for two units. The effective dose rate distributions can be obtained as far as 300 m from the center of the shielding house. The coupling technique is created with the Surface Source Write (SSW) card and the Surface Source Read/Coordinate Transformation (SSR/CRT) card in the MCNP 4C continuous energy Monte Carlo code as the “SSW-SSR/CRT calculation system.”In the present Monte Carlo coupling calculation, the total effective dose rates 100, 200, and 300 m from the center of the shielding house are estimated to be 1.69, 0.285, and 0.0826 (μSv/yr per four casks), respectively. Accordingly, if the distance between the center of the shielding house and the site boundary of the storage facility is kept at >300 m, approximately 2400 casks are able to be accommodated in the modular shielding houses, under the Japanese severe criterion of 50 μSv/yr at the site boundary. The shielding house alone satisfies not only the technical conditions but also the economic requirements.It became evident that secondary gamma rays account for >60% of the effective total dose rate at all the calculated points around the shielding house, most of which are produced from the water in the steel-water-steel shielding system of the shielding house. The remainder of the dose rate comes mostly from neutrons; the fission product and 60Co activation gamma rays account for small percentages. Accordingly, reducing the secondary gamma rays is critical to improving not only the shielding ability but also the radiation safety of the shielding house.

Identification of fissile materials from fission product gamma-ray spectra

A new active–passive nondestructive assay technique has been developed to identify fissile materials by observing the gamma-ray emissions from induced fission products. This technique entails using a neutron source to induce fissions in unknown containerized fissile materials and to subsequently observe high energy peaks (>800 keV) in the fission product gamma-ray spectrum. Ratios of the observed peak intensities are then formed. The values of these ratios can be compared to calculated theoretical values to identify the fissile material. The present work focuses on the differentiation between 235 U and 239 Pu .

Investigation of the recombination losses in a three-electrode cylindrical ionization chamber developed for gamma ray dosimetry of fission product activity

A three-electrode ionization chamber has been designed and developed for the gamma ray dosimetry of fission product activity and reported elsewhere. In this paper, the ( I, V) characteristics of the chamber filled, with argon gas at 1.24 MPa (180 psi) pressure, for fission product gamma rays from spent fuel have been studied. To do so, the chamber was irradiated with gamma rays using different numbers of (i.e. up to 4) spent fuel elements. The plateau region is reached above 1200 V and the detector operating voltage is found to be 2 kV. It is observed that in the plateau region the slope increases with an increase in the exposure rate. The ( 1 I , 1 V ) and ( I, 1 V 2 ) characteristic curves reveal the presence of the initial and volume recombination losses. The volume recombination losses are found to be smaller than the initial recombination losses. Both these losses increase with the increasing exposure rate but the increase in the volume recombination losses is slightly greater than that of the initial recombination losses.

Design and study of the characteristics of a three electrode experimental ionization chamber for gamma ray dosimetry of spent fuel

The (I, V) characteristics of two and three electrode ionization chamber filled with argon gas have been studied. To determine the sensitivity and the response with increase in exposure rate, the chamber was tested with a 60Co commercial irradiator. The response is linear up to more than 1.5 krad/h. The experimentally measured sensitivity of the chamber is 1.849×10−13 A/cm3 per rad/h when the argon gas pressure in the chamber is 1.24 GPa (180 psi). The effect of transparency of the intermediate electrod on the saturation current due to 137Cs gamma-rays has also been studied. The experimental results show that the electrode with holes of small diameter acts as a better intermediate electrode as compared to the electrodes without holes or with holes of a larger diameter. The chamber has also been teste with fission product gamma-rays from spent fuel elements of a typical pool type research reactor. The results indicate that the presence of an intermediate electrode lowers the operating voltage by 50% and reduces the slope in the plateau region.

A well logging technique for the in situ determination of 90Sr

A technique for determining the 90Sr concentration in soils and sediments by measurement of bremsstrahlung radiation in the energy range 60–236 keV with a large intrinsic germanium diode capable of logging wells as small as 3 in diameter is described. Corrections are made for changing background levels as a function of depth and for interferences from any other radionuclides present. A 90Sr concentration as low as 9 nCi cm −3 can be measured in only 10 s in typical fission product gamma-ray activities of 100 nCi cm −3. If no interfering radionuclides are present, 90Sr concentrations of <10 pCi cm −3 cm can be determined. relationship between detector size and sensitivity is discussed. Detection limits are given for different counting times as a function of interfering activity levels.

Fission Yields for Thermal-Neutron Fission of Plutonium-241

Cumulative fission yields for 17 fission products (16 mass chains) created by thermal-neutron fission of 241Pu were determined from analysis of gross fission product gamma-ray spectra obtained using a large-volume Ge(Li) detector. Uncertainties assigned to nine of the measured yields are smaller than existing evaluated uncertainties.

Quick Separation of Fission Product Molybdenum and Gamma-Rays of Mo-102

Quick Separation of Fission Product Molybdenum and Gamma-Rays of Mo-102

Structure Shielding Against Initial Radiations from Nuclear Explosions. I. Attenuation of Air Secondary and Fission Product Gamma Rays

Calculations and resulting data are described which are intended for use in estimating the protection afforded by buildings against nuclear radiations emitted from a nuclear burst in the first half...

Measurement and Analysis of Parameters in Tight 232Th-235U and 232Th-233U Lattices Moderated with Heavy Water

Parameter measurements and calculations in two D2O-moderated thorium-uranium critical assemblies are described in detail. The first, designated ETA-I, contained 6.7 wt% 235UO2-ThO2 fuel rods 0.66 cm in diameter, clad in aluminum. The second assembly, ETA-II, contained 3.0 wt% 233UO2-ThO2 fuel rods 1.09 cm in diameter, clad in Zircaloy-2. A relatively hard spectrum was obtained in both lattices. Because leakage was small and the central flux spectra were closely asymptotic, these assemblies provided a cleanly analyzable test of important reaction cross sections. Parameters measured in ETA-I were 232Th fission/235U fission (δ02), 232Th capture/235U fission (CR*), and epithermal/thermal-neutron ratios for 235U fission (δ25) and 232Th capture (p02). Corresponding measurements with 233U were made in ETA-II; δ02 was measured directly in fuel and ThO2 pellets by observing specific fission product gamma rays with a high resolution Ge(Li) detector. With this system, 232Th capture gamma-ray activity was also measured directly in a fuel pellet. The epithermal/thermal-neutron ratios were measured by the thermal-neutron subtraction technique relative to 164Dy, and CR* was measured relative to its thermal-neutron value in a very soft spectrum. A three-dimensional adjoint Monte Carlo program was used to calculate foil flux perturbations and cadmium cutoff energies. Supplemental fast/epithermal-neutron spectrum comparisons were made between the ETA assemblies and the 4/1 TRX critical assembly by means of the following activation detectors: 197Au(n,γ), 55Mn(n,γ), 235U(n,f), 238U(n,f), 27Al(n,α), and 96Zr(n, γ). The lattices were analyzed explicitly with a full energy range Monte Carlo program having a detailed cross-section description. ENDF/B Version 2 and light-water breeder reactor (LWBR) cross-section sets were used. On the whole, agreement with experiment was very good.

Prompt Critical Assembly Gamma Ray Sources

The method of obtaining the gamma dose rate during a prompt critical burst from the “Godiva” reactor is presented. In this measurement, the fission product gamma ray contribution to the total gamma ray dose must be determined. The determination of this quantity is described along with the results. Conclusions are drawn as to the relative contributions to the total dose from prompt, capture, and inelastic collision gamma ray sources.

Photoneutrons and the Control of a Pool Type Reactor

In a pool type reactor installation, the fission chambers or ionization chambers controlling the reactor detect two types of neutrons, e.g., thermalized fission neutrons and photoneutrons produced around the detector in a D(γ, η) H reaction.If the photoneutrons are produced by fission product gamma rays, there will be a superimposed neutron flux that may lead to unsafe operating conditions.This effect has been analytically and experimentally studied, and it is shown here that the unsafe conditions can be suppressed either by placing the detector closer to the reactor or by limiting the rate of change of reactor flux to a safe value.