Neutron production methods are an integral part of research and analysis for an array of applications. This paper examines methods of neutron production, and the advantages of constructing a radioisotopic neutron irradiator assembly using 252Cf. Characteristic neutron behavior and cost-benefit comparative analysis between alternative modes of neutron production are also examined. The irradiator is described from initial conception to the finished design. MCNP modeling shows a total neutron flux of 3 × 105 n/(cm2·s) in the irradiation chamber for a 25
Neutrons are useful in a variety of applications, spanning from laboratory investigations and field measurements, to national security and medical treatment. The manner by which neutrons are generated determines the particular energies and spectrum of the resulting emissions. Many studies exist on irradiators; however, most of the literature is rather dated. This paper seeks to modernize information on the design, construction, and modeling of a 252Cf neutron irradiator and provide a starting foundation to benefit others who might undertake such an endeavor.
Neutron irradiators are often employed in materials research, utilizing various techniques and methods such as neutron activation analysis (NAA), neutron radiography, and neutron diffraction for elemental analyses. Activation products can be measured using a high-purity germanium detector with gamma spectroscopy analysis software [
This paper is organized as follows. Section
In general, there are three main sources of neutrons: (
Comparison of neutron sources.
Neutron source | Production mode | Neutron yield (n/s) | Flux | Average neutron energies (MeV) | Spectrum | Cost ($) |
---|---|---|---|---|---|---|
241AmBe | ( | 7.0 × 10−5 | Low | 3.3, 4.3, 4.9, 6.8, 7.7, 9.5 | Continuous | 104 |
241AmF | 4.0 × 10−6 | 1.8, 2.1 | ||||
241AmB | 1.4 × 10−5 | 2.6 | ||||
239PuBe [ | 3.8 × 10−5 | ~3.3 | ||||
| ||||||
D-D generator [ | | 106–1011 | Moderate | 2.45 | Monoenergetic | 105 |
D-T generator [ | | 108–5 × 1013 | 14.1 | |||
| ||||||
252Cf [ | Spontaneous fission | 1.2 × 10−1 | Low | 2.0 | Fission | 104 |
Research reactor [ | Fission | >1016 | High | >108 [ |
The emitted neutrons may be focused into a beam for activation experiments, regardless of the neutron source. This neutron streaming is accomplished by surrounding the source location with reflecting or moderating material and leaving an empty pathway for the neutrons to propagate through freely (minimal interactions), with the volume along the path having a higher flux due to the scattering effects induced by the peripheral materials.
The purpose of a research reactor is to produce and sustain nuclear fission for experimentation with high neutron fluxes preferred, rather than electric power production. In the late 1950s, the Atomic Energy Commission (AEC) financially supported the construction of research reactors at universities in the United States. Many of these were Training Research Isotope-General Atomics (TRIGA) reactors. For almost 60 years, TRIGA reactor popularity among North American universities has helped to support the peaceful applications of nuclear technology [
Nuclear Regulatory Commission licensed research reactors at US universities (does not include those at industrial and other government sites such as Department of Energy national laboratories); data are compiled from multiple volumes of [
The TRIGA design serves as a comparative example for the other neutron sources examined in this paper. Several updates to the TRIGA reactor design have occurred and demonstrate a wide range of research capabilities and radiological characteristics. TRIGA reactors provide a regional compartment of high flux where material sample activation near the reactor core can be carried out. The characterized spectrum of neutrons in this irradiating region varies with the different TRIGA models owned by various organizations. For instance, the U.S. Geological Survey operates a 1 MW TRIGA capable of producing a
Commissioning a new nuclear research reactor requires considerable capital expenditure and extensive planning. This is a long-term commitment by an institution which assumes all responsibilities including financial requirements far into the future. A conservative estimate of over $200 million in capital is required for commissioning a basic research reactor, before factoring in many other significant costs such as the buildings and licensing [
Recently, accelerator-based systems seem to be receiving more interest than reactors for providing a high neutron flux. However, accelerators generate monoenergetic neutrons which limits the range of applicability for a multitude of research areas. Advantages of accelerators include selectivity and control of particular monoenergetic neutron energies and fluxes as well as the option to relocate the source and cost several orders of magnitude less than that of a research reactor. Accelerators require higher capital compared to radioisotopic neutron sources and are less portable than the isotopic-based irradiators, requiring extensive dismantling and reassembly in order to transport between locations; generally, accelerators are stationary structures.
Conventional accelerators utilize a D-D or D-T nuclear reaction to generate neutrons. Deuteron cations, required for both reactions, are produced in a radiofrequency ion source and are accelerated to high energy. A focused beam of accelerated deuterons collides with one of two titanium compound targets, TiD or TiT, determining the energy of the neutrons ejected. The nuclear reaction, D
Radioisotopic sources have many benefits that include easy transportability, low relative cost, zero to little maintenance, and no external power requirements. Radioisotopic-produced neutrons are generated by any of three primary modes of energy decay: (
The neutron energy spectrum, emitted by sources like 241AmBe or that of PuBe, is determined by the alpha emission energy contributed in the
The neutron yield of a 241AmBe source depends on the composition and volume of the AmO2 and 9Be used to make up a sealed neutron source. 241AmBe users benefit from an exceptionally stable flux over many years owed to its long half-life. Comparatively, 241AmBe incurs a smaller expenditure among the radioisotope sources. However, the neutron flux is also smaller than the other sources. This is disadvantageous when large samples need to be activated since the radiation intensity will be disproportionate closest to the source, as the sample periphery becomes activated at a slower rate than the parts of the sample closest to the source. An important property of 241AmBe is its high ratio of
Inexpensive neutron sources like PuBe have been utilized frequently in the petroleum industry for oil well logging and calibrating measuring instruments. PuBe13 is a single face centered cubic structure synthesized from pure 239Pu metal (oxide) and beryllium. This neutron source has certain advantages over similar
Like most of the radioisotopes discussed, 252Cf produces both neutrons and gamma rays among other radiations. With an overall half-life of 2.645 y, 252Cf decays primarily by alpha emission with only 3% of the nuclear transformations being spontaneous fission (SF). This Cf2O3 SF source is currently produced by only two facilities in the world, Oak Ridge National Laboratory in the US and the Research Institute of Atomic Reactors in Dimitrovgrad, Russia. Cermet wire is formed by the suspension of the metal-oxide compound in a palladium matrix. The wire is encased in a palladium tube and then doubly-encapsulated in an additional stainless steel tube. The typical composition of the oxide source is as follows: 2 weight percent (w/o) of 249Cf, 15 w/o of 250Cf, 4 w/o of 251Cf, and 79 w/o of 252Cf [
252Cf decay characteristics.
Parameter | Value |
---|---|
Half-life (effective: | 2.645 y |
Half-life (spontaneous fission) | 85.5 y |
Decay mode | |
Alpha energies | 6.076 and 6.118 MeV |
Accelerator sources were the most common neutron source available in the mid-1980s when the high initial cost to transmute plutonium into 252Cf prevented many from purchasing this SF source [
Having made these comparisons, our inclination was to use the spontaneous fission 252Cf source. Our primary intended use was to induce low-levels of radioactivity (neutron activation) in small material samples. In a similar application, the University of Minnesota built a neutron irradiation apparatus using a borrowed 11.5 mg 252Cf source to test the radiation hardness of photodiodes and other electronic components [
With the 252Cf source chosen, the potential irradiator characteristics were examined. First and foremost, the irradiator apparatus needed to incorporate passively safe design aspects. The irradiator was intended to be simple in operation and durable over time under heavy use. To that end, several existing irradiators were analyzed for design exemplars as well as individual strengths and weaknesses to improve upon. Although the purposes of the irradiators researched differ, they all utilize a radioisotopic-based neutron source within a structure of moderating and shielding materials. This is the common theme among various neutron irradiators and howitzers.
One example of an education-oriented neutron irradiator is the cylindrical Plexiglas, water moderated, “Visiflux” neutron howitzer, designed by ATOMIC Accessories, Inc. That howitzer was constructed for a university laboratory as an additional resource for teaching radiochemistry, nuclear engineering, and similar disciplines of study. The Visiflux weighs approximately 56.7 kg without water and has a 0.61 m diameter, standing roughly 0.71 m tall. The fully assembled howitzer hosts a neutron source in its center, surrounded by a shield of water. The water shield acts to moderate higher energy neutrons down to thermal equilibrium and helps protect students from the ejected neutrons. The Visiflux handles activities of up to 185 GBq (5 curies), using PoBe or PuBe neutron sources interchangeably [
In the past, commercialized neutron irradiators have been designed with parameters to make them appealing to a broad range of potential buyers, but at the cost of a high retail price point. Neutron irradiators can be constructed for a specialized purpose and for a lower cost than that of most commercially available howitzers. Faculty at Arizona State University constructed a neutron howitzer in 1966 for student use in the nuclear physics laboratory. That howitzer is described as a 55-gallon steel barrel, filled with paraffin up to 0.1 m from the top. Wheels affixed to the drum base allowed it to be transported through various laboratories. It did not require substantial shielding, employing a 1 Ci PuBe source, which produced a neutron flux of 3 × 103 n/(cm2·s). The small flux meant safer operation by students and minimal shielding cost. However, this necessitated increased sample irradiation times to achieve a specified activation level in a material sample. The design of that howitzer was limited to irradiating only basic shapes and small sample sizes. Although Rawls and Voss quoted an exceptionally low cost to make the howitzer, many of the materials used were already owned or donated to the project construction. The funds needed to reconstruct the same irradiator from nothing would actually be significantly higher than stated [
Another irradiator described in the literature varied significantly in shape and use compared to the above howitzer. This contrasting irradiator was a dual-hemisphere utilizing a 5 Ci 241AmBe source at its center and constructed specifically for qualitative-quantitative materials analysis. Paraffin wax with 5% being boric acid internally (the dual-hemisphere) and 20% boric acid in the outer components of containment encapsulated the 241AmBe source. Also, lead shielding was arranged around the periphery of the inner paraffin shielding. This helped to provide some attenuation of prompt gamma radiation being emitted from the neutron captures by the borated material [
A setup capable of irradiating small samples was needed in the absence of an onsite nuclear reactor. The 252Cf-based neutron irradiator is primarily fabricated from high density and purity polyethylene material, as well as interlocking lead bricks. The major components, shown in Figure
Exploded view irradiator components from computer-aided modeling software.
The irradiator design had five main goals: (
Based on the available funds, a 25
252Cf source data without moderation.
252Cf mass | 25 |
Neutron emission | 5.75 × 107 n/s |
Neutron dose at 1 m in air | 0.575 mSv/h |
Neutron energy | 2.3 MeV (average) |
Gamma exposure at 1 m in air | 0.03478 mGy/h |
Gamma energies | 0.2–1.8 MeV |
The shipping container houses the source internally within a borated polyethylene cylinder at the drum core. The drum itself weighs 230 kg, thus providing a stable base for the rest of the irradiator to be constructed on. The drum acquires its shielding properties and weight from being filled with a paraffin-lead mixture. A polyvinyl chloride (PVC) pipe with a 10.2 cm ID (10.8 cm OD) runs vertically through the center of the shipping container. The cylinder housing the source fits snug in the PVC pipe and rests at the bottom of the container core, approximately halfway down the drum. The PVC pipe extrudes from the top surface of the drum by almost the same height as the drum edge (~10 cm).
The source is repositioned between the shipping container core (i.e., the shielded, nonirradiating position) and the irradiating position in the sample chamber, via a control cable. The cable access, located below the rotation grips at the top of the assembly, allows the operator to shift the source vertically. The movement is constrained between the barrel center and the sample chamber ceiling. Two labeled markers affixed to the cable indicate the source location: (
Certainly, operator dose is kept as low as reasonably achievable (ALARA) by limiting their exposure time, following the irradiator operating procedure, and maximizing distance from the irradiating chamber when possible. Furthermore, the design reduces the possibility of radiation streaming. For example, the moderating tower was assembled with horizontally stacked polyethylene (PE) sheets which were then encased with borated PE oriented vertically. Gamma exposure was initially found to exceed 0.15 mSv/h (15 mrem/h) near the irradiator and was addressed by implementing a 5 cm (2 inches) thick lead wall structure surrounding the irradiator assembly. The completed irradiator is shown in Figure
Assembled neutron irradiator with lead shielding.
This section describes the assembly of the irradiator components in order to better understand its operation. The components themselves were fabricated at the university using machining equipment controlled by our computer-aided drawings. The construction of the overall irradiator is described in the sequence that the components were assembled.
The stationary donut-shaped base, as shown in Figure
(a) Stationary donut-shaped component that rests on the shipping drum; (b) bottom of the RIC that rests and slides on the component shown left.
The RIC assembly consists of the pure polyethylene irradiation chamber and attached borated cylinders extending below the RIC into the source container, and above to a point of access for rotating the apparatus 180°. The RIC has one bore hole in line with the source. The bore hole runs through the bottom cylindrical extension shown attached in Figure
To decrease sample irradiation times, the moderating tower component was utilized. The pure polyethylene used as moderator has a small coefficient of friction and can therefore cause problems with component alignment when assembled. Radial and lateral movement was constricted by the addition of a strong skeletal-like internal frame. The tower internal support is four aluminum rods that run through the corners of the stacked square PE layers. This prevents any wiggle in the horizontal plane and keeps the sides of the moderating tower flush.
The tower is engulfed with no less than 10 cm of borated PE on the periphery. The outer borated PE shield, with a thermal neutron one-tenth thickness of 0.45 inches, has primary responsibility for the overall neutron attenuation and exposure reduction by the irradiator. The periphery of the moderation tower is padded with no less than 10 cm (4 in.) of neutron capturing material. The RIC extensions are constructed of borated material so no gap is created in the outer shielding by the protrusion.
The last component, a lead wall, provides additional protection against high-energy photon exposure. The 252Cf neutron source produces gamma radiation by two mechanisms: (
The irradiator may be used by a single operator to perform sample irradiations. Irradiation times vary with sample material and desired activity levels. Scoping calculations are made to estimate a particular sample irradiation time and help assure that no samples become too radioactive, thereby risking unnecessary exposure to the sample handler when removed or exceeding radioactive material license possession limits. The calculations are performed prior to an irradiation, where the
The sample chamber is accessed only when the source is in the nonirradiating position within the drum, and the lines of sight of the bore holes are disconnected between the drum core and the RIC where the samples are located. This alignment reduces neutron streaming into the sample chamber from the neutron source stored directly below, as well as making the source more difficult to pull up accidentally. The RIC is configured by rotating the handles at the top of the irradiator, as shown in Figure
For measurement of exposure outside of the irradiator, a Ludlum model 12-4 Bonner sphere-type neutron dose rate meter employing a 3He proportional detector is used. Additionally, gamma radiation exposure from the irradiator and any activated samples are measured using a Ludlum Model 2 survey meter with a detachable Geiger pancake probe (model 44-9). Table
Radiation measurements at the chamber access door.
Source location | Neutron dose rate | Gamma exposure rate |
---|---|---|
Nonirradiating position (down) | 2 | 8 |
Irradiation position (up) | 42 | 0.15 mGy/h |
Monte Carlo radiation transport simulations investigated two aspects of the irradiator. The first was to evaluate how well the polyethylene moderates neutrons by simulating the 252Cf source in the irradiator and comparing those results to a baseline calculation assuming that the source was in air. Second, the effectiveness of the borated polyethylene layers to shield workers from neutron and gamma-ray radiation was characterized and compared to measurements at positions outside the irradiator where human contact might be made.
The Monte Carlo N-Particle (MCNP) transport code [
MCNP model of neutron irradiator.
Neglecting conventional substances such as water, air, and aluminum, Table
Properties of select irradiator materials.
Material | Density (g/cm3) | Composition (mass fraction) |
---|---|---|
Cf2O3 in Pd matrix | 12.02 | Pd (0.999); Cf2O3 (0.001) with the Cf comprised of 252Cf (0.79); 251Cf (0.04); 250Cf (0.15); 249Cf (0.02) |
Source encapsulation | 8.03 | Stainless steel 304L (1.0) |
Pure polyethylene | 0.92 | H (0.1437); C (0.8563) |
Borated polyethylene | 0.95 | H (0.116); C (0.612); O (0.222); B (0.050) |
An average of 3.7675 neutrons [
The corresponding 252Cf gamma emission spectrum is not provided in MCNP. Consequently, published data were consulted to obtain a photon source description. It was important to consider only those published gamma-ray spectra which provide the actual number of gamma rays being emitted [
To characterize the neutron irradiator performance, MCNP was used to obtain the dose rate to water phantoms located (a) within the inner irradiation chamber, (b) in front of the access door, and (c) near the handlebars where the turntable placement is controlled at the top of the irradiator. None of these locations are shielded by the lead structure. The front and top water phantoms were placed relatively close to the irradiator surface as measurements from a pancake probe are used for simulation verification in the case of gamma rays. Three separate simulations were run to account individually for the delayed and prompt gamma rays, as well as the secondary gamma rays which are emitted as a result of
Calculated gamma-ray dose rates to various locations of the irradiator. Note the use of different vertical axes.
The total gamma-ray dose rates obtained in the MCNP simulations were 159
Similar simulations were done in MCNP to characterize the neutron dose rates. The water phantoms were relocated according to the position of the center of the neutron dose rate meter. The MCNP dose values were converted to a unit of rem by applying the neutron quality factors [
Neutron dose rates to water phantoms at various irradiator locations.
It was found that the neutron dose rate at the top location is sensitive to the vertical position of the 252Cf source within the irradiation cavity since the source comes into closer proximity with the highly shielding borated polyethylene upper extension cylinder. Shifting the 252Cf source 1.27 cm (0.5 inches) upward, which is within the resolution of the marker on the source cable, changes the total dose rate by 13%. Between the fluctuation of the analog readout (due to the low dose rate) and the imprecision in placing the 252Cf source in the same exact vertical position from irradiation to irradiation, the MCNP results were shown to agree with 10% of the neutron survey meter measurements which falls within the manufacturer true value margin for the neutron dose rate [
The particle fluxes to a water phantom placed inside the irradiation chamber for neutrons and the three sources of gamma rays were computed. The total neutron flux was found to be 3.03 × 105 n/(cm2·s) while the total gamma-ray flux was 8.02 × 105
(a) Neutron and (b) gamma-ray interactions with the irradiator.
The MCNP computed neutron spectrum within the sample chamber is plotted in Figure
Calculated neutron spectrum within the sample chamber with and without the pure and borated polyethylene present.
This neutron source installation was researched, designed, and conducted to provide a modern radioisotopic neutron irradiator. 252Cf has become more affordable over the last few decades due to advances in nuclear technology and isotope production methods. This paper furnishes the scientific community with a guide to constructing an irradiation apparatus. The moderated 25
The authors declare that they have no competing interests.