Howard O. Menlove

A 252Cf based nondestructive assay system for fissile material

Abstract A modulated 252 Cf source assay system “Shuffler” based on fast-or-thermal-neutron interrogation combined with delayed-neutron counting has been developed for the assay of fissile material. The 252 Cf neutron source is repetitively transferred from the interrogation position to a shielded position while the delayed neutrons are counted in a high efficiency 3 He neutron well-counter. For samples containing plutonium, this well-counter is also used in the passive coincidence mode to assay the effective 240 Pu content. The design of an optimized neutron tailoring assembly for fast-neutron interrogation using a Monte Carlo Neutron Computer Code is described. The Shuffler system has been applied to the assay of fuel pellets, inventory samples, irradiated fuel and plutonium mixed-oxide fuel. The system can assay samples with fissile contents from a few milligrams up to several kilograms using thermal-neutron interrogation for the low mass samples and fast-neutron interrogation for the high mass samples. Samples containing 235 U− 238 U, or 233 U−Th, or UO 2 −PuO 2 fuel mixtures have been assayed with the Shuffler system.

Two specialized delayed-neutron detector designs for assays of fissionable elements in water and sediment samples

Abstract Two specialized neutron-sensitive detectors are described which are employed for rapid assays of fissionable elements by sensing for delayed neutrons emitted by samples after they have been irradiated in a nuclear reactor. The more sensitive of the two detectors, designed to assay for uranium in water samples, is 40% efficient; the other, designed for sediment sample assays, is 27% efficient. These detectors are also designed to operate under water as an inexpensive shielding against neutron leakage from the reactor and neutrons from cosmic rays.

Neutron proportional counter design for high gamma-ray environments

Abstract A parametric analysis of the performance of 3 He neutron proportional detector tubes in mixed gamma-ray/neutron environments has been performed. The objective of this study was to determine the optimum tube design configuration for minimized gamma-ray sensitivity based upon commercially available components. The parameters examined in the study were the tube wall material, the admix gas identity, the total fill pressure, and the tube lining material. The study resulted in the quantification of the limitations of typical 3 He tube designs and in the development of a new tube design which extends the usefulness of 3 He tubes in mixed gamma-ray/neutron environments.

Automated Nondestructive Assay Instrumentation for Nuclear Materials Safeguards

Four systems developed at the Los Alamos Scientific Laboratory for nondestructive analysis of nuclear fuel materials will be described. These systems utilize either minicomputers or a programmable calculator for measurement control and data analysis, and are typical of a variety of automated measurement systems developed for nuclear materials safeguards applications.

Neutron-burst detectors for cold-fusion experiments

Abstract We have designed several neutron detectors for use in measuring neutrons from cold-fusion experiments. The high-efficiency detectors are based on 3He gas tubes in a CH2 moderator. The total efficiency of our most advanced detector for 2.3 MeV (252Cf) neutrons is 44%. The detector consists of two independent segments making up inner and outer rings of 3He tubes. The inner detector has nine 3He tubes and the outer detector has 42 3He tubes. The low-background inner detector has a singles count background of 97 counts/h and a coincidence count background of only 0.7 counts/h. The corresponding singles efficiency is 19%. The inside ring of tubes is undermoderated and it is more sensitive to lower-energy source neutrons. The outside ring of tubes is overmoderated and thus more sensitive to high-energy neutrons. The ratio of the outside tubes to the inside tubes gives the approximate average neutron energy based on differential transmission as a function of energy. Monte Carlo transport calculations have been performed to establish the energy calibration of the outside/inside tube-ring ratio. The time-correlation (coincidence) information is required to investigate neutron pulsing, bursts and other nonrandom emissions caused by nonequilibrium conditions and instabilities in the samples. The time correlations in our detector are obtained by using the shift register-type electronic circuitry. Our detector system has been applied successfully to the measurement of neutron yield and multiplicity distribution from cold-fusion electrolysis samples and D2-gas-type samples.

Development of Self-Interrogation Neutron Resonance Densitometry to Improve Detection of Possible Diversions for PWR Spent Fuel Assemblies

A new nondestructive assay technique called self-interrogation neutron resonance densitometry (SINRD) is currently being developed at Los Alamos National Laboratory to improve existing nuclear safeguards and material accountability measurements for light water reactor fuel assemblies. The viability of using SINRD to improve the detection of possible diversion scenarios for pressurized water reactor 17 × 17 spent low-enriched uranium (LEU) and mixed oxide (MOX) fuel assemblies was investigated via Monte Carlo N-Particle eXtended transport code (MCNPX) simulations. The following capabilities were assessed: (a) verification of the burnup of a spent fuel assembly, (b) ability to distinguish fresh and one-cycle spent MOX fuel from three- and four-cycle spent LEU fuel, and (c) sensitivity and penetrability to the removal of fuel pins. SINRD utilizes 244Cm spontaneous-fission neutrons to self-interrogate the spent fuel pins. The amount of resonance absorption of these neutrons in the fuel can be quantified using a set of fission chambers (FCs) placed adjacent to the assembly. The sensitivity of SINRD is based on using the same fissile materials in the FCs as are present in the fuel because the effect of resonance absorption lines in the transmitted flux is amplified by the corresponding (n, f) reaction peaks in the FC. SINRD requires calibration with a reference assembly of similar geometry in a similar measurement configuration with the same surrounding moderator as the spent fuel assemblies. However, this densitometry method uses ratios of different detectors so that several systematic errors related to calibration and positioning cancel in the ratios.

Development of Self-Interrogation Neutron Resonance Densitometry to Quantify the Fissile Content in PWR Spent LEU and MOX Assemblies

Abstract A new nondestructive assay technique called self-interrogation neutron resonance densitometry (SINRD) is currently being developed at Los Alamos National Laboratory to improve existing nuclear safeguards and material accountability measurements for light water reactor fuel assemblies. The viability of using SINRD to quantify the fissile content (235U and 239Pu) in pressurized water reactor 17 × 17 spent low-enriched uranium and mixed-oxide fuel assemblies in water was investigated via Monte Carlo N-particle extended transport code simulations. SINRD utilizes 244Cm spontaneous fission neutrons to self-interrogate the fuel pins. The amount of resonance absorption of these neutrons in the fuel can be quantified using 235U and 239Pu fission chambers placed adjacent to the assembly. The sensitivity of this technique is based on using the same fissile materials in the fission chambers as are present in the fuel because the effect of resonance absorption lines in the transmitted flux is amplified by the corresponding (n,f) reaction peaks in the fission chamber. SINRD requires calibration with a reference assembly of similar geometry. However, this densitometry method uses ratios of different fission chambers so that most systematic errors related to calibration and positioning cancel in the ratios.

Distributed source term analysis, a new approach to nuclear material inventory verification

The Distributed Source-Term Analysis (DSTA) technique is a new approach to measuring in-process material holdup that is a significant departure from traditional hold-up measurement methodology. The DSTA method is a means of determining the mass of nuclear material within a large, diffuse, volume using passive neutron counting. The DSTA method is a more efficient approach than traditional methods of holdup measurement and inventory verification. The time spent in performing DSTA measurement and analysis is a fraction of that required by traditional techniques. The error ascribed to a DSTA survey result is generally less than that from traditional methods. Also, the negative bias ascribed to γ-ray methods is greatly diminished because the DSTA method uses neutrons which are more penetrating than γ-rays.

Radiation damage to 3He proportional counter tubes

Abstract Neutron damage to 3 He proportional-counter tubes is inhibited by coating the cathodes of the tubes with activated charcoal. This coating also improves pulse-height resolution.

252Cf-based hydrogen analyzer☆

Abstract A system is described for the nondestructive analysis of hydrogen. 3 He tube are used to detected 252 Cf fission neutrons which have been slowed by collisions with hydrogen atoms. The system is capable of measuring hydrogen in amounts as small ase 5 mg.