Long-Range Neutron Detection

A neutron detector designed for detecting neutron sources at distances of 50 to 100 m has been constructed and tested. This detector has a large surface area (1 m{sup 2}) to enhance detection efficiency, and it contains a collimator and shielding to achieve direction sensitivity and reduce background. An unusual feature of the detector is that it contains no added moderator, such as polyethylene, to moderate fast neutrons before they reach the {sup 3}He detector. As a result, the detector is sensitive mainly to thermal neutrons. The moderator-free design reduces the weight of the detector, making it more portable, and it also aids in achieving directional sensitivity and background reduction. Test results show that moderated fission-neutron sources of strength about 3 x 10{sup 5} n/s can be detected at a distance out to 70 m in a counting time of 1000 s. The best angular resolution of the detector is obtained at distances of 30 m or less. As the separation .distance between the source and detector increases, the contribution of scattered neutrons to the measured signal increases with a resultant decrease in the ability to detect the direction to a distant source. Applications for which the long-range detector appears to be suitable include detecting remote neutron sources (including sources in moving vehicles) and monitoring neutron storage vaults for the intrusion of humans and the effects they make on the detected neutron signal. Also, the detector can be used to measure waste for the presence of transuranic material in the presence of high gamma-ray background. A test with a neutron source (3 x 10{sup 5} n/s) in a vehicle showed that the detector could readily measure an increase in count rate at a distance of 10 m for vehicle speeds up to 35 mph (the highest speed tested). These results. indicate that the source should be detectable at this distance at speeds up to 55 mph

Direct Fast-Neutron Detection: A Progress Report

It is widely acknowledged that Mure neutron-detection technologies will need to offer increased performance at lower cost. One clear route toward these goals is rapid and direct detection of fast neutrons prior to moderation. This report describes progress to date in an effort to achieve such neutron detection via proton recoil within plastic scintillator. Since recording proton-recoil events is of little practical use without a means to discriminate effectively against gamma-ray interactions, the present effort is concentrated on demonstrating a method that distinguishes between pulse types. The proposed method exploits the substantial difference in the speed of fission neutrons and gamma-ray photons. Should this effort ultimately prove successful, the resulting. technology would make a valuable contribution toward meeting the neutron-detection needs of the next century. This report describes the detailed investigations that have been part of Pacific Northwest National Laborato@s efforts to demonstrate direct fast-neutron detection in the laboratory. Our initial approach used a single, solid piece of scintillator along with the electronics needed for pulse-type differentiation. Work to date has led to the conclusion that faster scintillator and/or faster electronics will be necessary before satisfactory gamma-ray discrimination is achieved with this approach. Acquisition and testing of both faster scintillator and faster electronics are currently in progress. The "advanced" approach to direct fast-neutron detection uses a scintillating assembly with an overall density that is lower than that of ordinary plastic scintillator. The lower average density leads to longer interaction times for both neutrons and gamma rays, allowing easier discrimination. The modeling, optimization, and design of detection systems using this approach are described in detail

Predicting 232U Content in Uranium

The minor isotope 232U may ultimately be used for detection or confirmation of uranium in a variety of applications. The primary advantage of 232 U as an indicator of the presence of enriched uranium is the plentiful and penetrating nature of the radiation emitted by its daughter radionuclide 208Tl. A possible drawback to measuring uranium via 232U is the relatively high uncertainty in 232U abundance both within and between material populations. An important step in assessing this problem is to ascertain what determines the 232U concentration within any particular sample of uranium. To this end, we here analyze the production and eventual enrichment of 232 U during fuel-cycle operations. The goal of this analysis is to allow approximate prediction of 232 U quantities, or at least some interpretation of the results of 232U measurements. We have found that 232U is produced via a number of pathways during reactor irradiation of uranium and is subsequently concentrated during the later enrichment of the uranium' s 235U Content. While exact calculations are nearly impossible for both the reactor-production and cascade-enrichment parts of the prediction problem, estimates and physical bounds can be provided as listed below and detailed within the body of the report. Even if precise calculations for the irradiation and enrichment were possible, the ultimate 212U concentration would still depend upon the detailed fuel-cycle history. Assuming that a thennal-diffusion cascade is used to produce highly enriched uranium (HEU), dilution of reactor-processed fuel at the cascade input and the long-term holdup of 232U within the cascade both affect the 232U concentration in the product. Similar issues could be expected to apply for the other isotope-separation technologies that are used in other countries. Results of this analysis are listed below: 0 The 232U concentration depends strongly on the uranium enrichment, with depleted uranium (DU) containing between 1600 and 8000 times less 232U than HEU does. * The 236U/232U concentration ratio in HEU is likely to be between 10{sup 6} and 2 x 10{sup 7}. 0 Plutonium-production reactors yield uranium with between I and 10 ppt of 232u. 0 Much higher 132U concentrations can be obtained in some situations. * Significant variation in the 232U concentration is inevitable. * Cascade enrichment increases the 232U concentration by a factor of at least 200, and possibly as much as 1000. 0 The actual 232U concentration depends upon the dilution at the cascade input

Long Range Neutron Detection: A Progress Report

The detection of neutron sources horn a considerable distance constitutes a problem that must be treated separately from the bulk of other neutron-detection applications. This report analyzes this problem, describes a number of possible approaches, and describes the design and construction of a square-meter detection system using the approach of moderator-free directional neutron detection. Although experimental results are not the focus of this report a few preliminary results are offered in the last section. Both theoretical and preliminary experimental results confirm that usefi.d detection of neutron sources for national-security applications is relatively easy at a distance of 50 meters, yet becomes somewhat challenging from a distance of 100 meters. The square-meter detection system designed for this effort was intended to be, in decreasing order of priority, optimally capable of neutron-source detection at 100 meters, lightweight and easy to use, and low in cost. Thus, the majority of design decisions were driven by the need to maximize sensitivity for remote source detection. Several surprises resulted from this design effort. First, we discovered that%, rather than cadmium or gadolinium, must be used as a shielding material. Second, we discovered that a relatively open collimator is best for remote detection. These and other design decisions are described in detail in the third section of this report. The final detector weighs roughly 45 kg and inco~orates hardware with a cost of roughly $1OOK. Of course, lighter or cheaper detection systems could be designed with some reduction in sensitivity. As designed, our l-square-meter moderator-free detection system is expected to be superior to conventional moderate-and-capture detection for some applications

Bubble Radiation Detection: Current and Future Capability

Despite a number of noteworthy achievements in other fields, superheated droplet detectors (SDDs) and bubble chambers (BCs) have not been used for nuclear nonproliferation and arms control. This report examines these two radiation-detection technologies in detail and answers the question of how they can be or should be ''adapted'' for use in national security applications. These technologies involve closely related approaches to radiation detection in which an energetic charged particle deposits sufficient energy to initiate the process of bubble nucleation in a superheated fluid. These detectors offer complete gamma-ray insensitivity when used to detect neutrons. They also provide controllable neutron-energy thresholds and excellent position resolution. SDDs are extraordinarily simple and inexpensive. BCs offer the promise of very high efficiency ({approximately}75%). A notable drawback for both technologies is temperature sensitivity. As a result of this problem, the temperature must be controlled whenever high accuracy is required, or harsh environmental conditions are encountered. The primary findings of this work are listed and briefly summarized below: (1) SDDs are ready to function as electronics-free neutron detectors on demand for arms-control applications. The elimination of electronics at the weapon's location greatly eases the negotiability of radiation-detection technologies in general. (2) As a result of their high efficiency and sharp energy threshold, current BCs are almost ready for use in the development of a next-generation active assay system. Development of an instrument based on appropriately safe materials is warranted. (3) Both kinds of bubble detectors are ready for use whenever very high gamma-ray fields must be confronted. Spent fuel MPC and A is a good example where this need presents itself. (4) Both kinds of bubble detectors have the potential to function as low-cost replacements for conventional neutron detectors such as {sup 3}He tubes. For SDDs, this requires finding some way to get boron into the detector. For BCs, this requires finding operating conditions permitting a high duty cycle

Direct Fast-Neutron Detection: A Progress Report

It is widely acknowledged that Mure neutron-detection technologies will need to offer increased performance at lower cost. One clear route toward these goals is rapid and direct detection of fast neutrons prior to moderation. This report describes progress to date in an effort to achieve such neutron detection via proton recoil within plastic scintillator. Since recording proton-recoil events is of little practical use without a means to discriminate effectively against gamma-ray interactions, the present effort is concentrated on demonstrating a method that distinguishes between pulse types. The proposed method exploits the substantial difference in the speed of fission neutrons and gamma-ray photons. Should this effort ultimately prove successful, the resulting. technology would make a valuable contribution toward meeting the neutron-detection needs of the next century. This report describes the detailed investigations that have been part of Pacific Northwest National Laborato@s efforts to demonstrate direct fast-neutron detection in the laboratory. Our initial approach used a single, solid piece of scintillator along with the electronics needed for pulse-type differentiation. Work to date has led to the conclusion that faster scintillator and/or faster electronics will be necessary before satisfactory gamma-ray discrimination is achieved with this approach. Acquisition and testing of both faster scintillator and faster electronics are currently in progress. The "advanced" approach to direct fast-neutron detection uses a scintillating assembly with an overall density that is lower than that of ordinary plastic scintillator. The lower average density leads to longer interaction times for both neutrons and gamma rays, allowing easier discrimination. The modeling, optimization, and design of detection systems using this approach are described in detail

Origin of the 871-keV gamma ray and the ``oxide'' attribute

This work concludes the investigation of the oxide attribute of current interest for the characterization of stored plutonium. Originally it was believed that the presence of oxide could be ascertained by measurement of the 871-keV line in a high-resolution gamma-ray spectrum. However, recent work has suggested that the 871-keV gamma ray in plutonium oxide arises from the reaction {sup 14}N({alpha},p){sup 17}O rather than the inelastic scattering reaction {sup 17}O({alpha},{alpha}{prime}){sup 17}O*. This conclusion, though initially surprising, was obtained during efforts to determine the relative importance of americium and plutonium alpha-particle decay for the production of the 871-keV gamma ray. Several questions were raised by previous experiments: What role, if any does {sup 17}O have in the generation of the 871-keV gamma ray? How does sufficient nitrogen come to be present in plutonium oxide? Under what conditions is the 871-keV gamma ray measurable in plutonium oxide? This paper describes the answers to these questions