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APPLICATIONS OF CALIFORNIUM-252 NEUTRON SOURCES IN MEDICINE, RESEARCH, AND INDUSTRY
Abstract
The 252 Cf radioisotope is an intense neutron emitter that is readily encapsulated in compact, portable, sealed sources. Californium-252 is used commercially as a reliable, cost-effective neutron source for prompt gamma neutron activation analysis (PGNAA) of coal, cement, and minerals, and for detection and identification of explosives, land mines, and unexploded military ordnance. Other uses include neutron radiography, materials characterization and nuclear assay, reactor start-up sources, and calibration standards. Also highlighted are the treatment of cancer using 252 Cf, experiments at the Californium User Facility for Neutron Science (CUF), and the 252 Cf distribution programs of the U.S. Department of Energy (DOE).
... The photon spectrum of the 252 Cf source has photon energies in the range of 0.01-10 MeV [25]. One microgram of 252 Cf emits 2.314 9 10 6 fast neutrons/s [26] and has photon yield of about 1.3 9 10 7 photons/s [27]. Maxwellian fission distribution was defined as the distribution for the neutrons emitted by the 252 Cf. ...
... Maxwellian fission distribution was defined as the distribution for the neutrons emitted by the 252 Cf. The Maxwellian neutron spectrum emitted by 252 Cf has the average energy of 2.1 MeV and the most probable energy of *0.7 MeV [26]. The F6 tally was used to score the neutron and photon doses in this step. ...
The aim of this study is to compare dose enhancement of various agents, nanoparticles and chemotherapy drugs for neutron capture therapy. A (252)Cf source was simulated to obtain its dosimetric parameters, including air kerma strength, dose rate constant, radial dose function and total dose rates. These results were compared with previously published data. Using (252)Cf as a neutron source, the in-tumour dose enhancements in the presence of atomic (10)B, (157)Gd and (33)S agents; (10)B, (157)Gd, (33)S nanoparticles; and Bortezomib and Amifostine chemotherapy drugs were calculated and compared in neutron capture therapy. Monte Carlo code MCNPX was used for simulation of the (252)Cf source, a soft tissue phantom, and a tumour containing each capture agent. Dose enhancement for 100, 200 and 500 ppm of the mentioned media was calculated. Calculated dosimetric parameters of the (252)Cf source were in agreement with previously published values. In comparison to other agents, maximum dose enhancement factor was obtained for 500 ppm of atomic (10)B agent and (10)B nanoparticles, equal to 1.06 and 1.08, respectively. Additionally, Bortezomib showed a considerable dose enhancement level. From a dose enhancement point of view, media containing (10)B are the best agents in neutron capture therapy. Bortezomib is a chemotherapy drug containing boron and can be proposed as an agent in boron neutron capture therapy. However, it should be noted that other physical, chemical and medical criteria should be considered in comparing the mentioned agents before their clinical use in neutron capture therapy.
... The cap- ture product dose (absorbed dose by capture materials) resulted from the capture of thermal neutrons by 10 B, 157 Gd and 33 S was calculated using the fluence-to-kerma conversion factors [13]. The neutron dose is the sum of source fast neutron dose resulted from elastic scatter- ing of fast neutrons in water and the capture product dose which is resulted from thermal neutron capture by 10 B, 157 Gd and 33 S. The neutron energy spectrum for 252 Cf source was assumed to be Maxwellian spectrum with an average energy of 2.1 MeV and the most prob- able energy of ~0.7 MeV [14]. Photon spec- trum of the 252 Cf source was taken from the study by Fortune, and has photon energies in the range of 0.01-9.79 ...
Background:
In neutron interaction with matter and reduction of neutron energy due to multiple scatterings to the thermal energy range, increasing the probability of thermal neutron capture by neutron captures makes dose enhancement in the tumors loaded with these materials.
Objective:
The purpose of this study is to evaluate dose distribution in the presence of (10)B, (157)Gd and (33)S neutron capturers and to determine the effect of these materials on dose enhancement rate for (252)Cf brachytherapy source.
Methods:
Neutron-ray flux and energy spectra, neutron and gamma dose rates and dose enhancement factor (DEF) are determined in the absence and presence of (10)B, (157)Gd and (33)S using Monte Carlo simulation.
Results:
The difference in the thermal neutron flux rate in the presence of (10)B and (157)Gd is significant, while the flux changes in the fast and epithermal energy ranges are insensible. The dose enhancement factor has increased with increasing distance from the source and reached its maximum amount equal to 258.3 and 476.1 cGy/h/µg for (157)Gd and (10)B, respectively at about 8 cm distance from the source center. DEF for (33)S is equal to one.
Conclusion:
Results show that the magnitude of dose augmentation in tumors containing (10)B and (157)Gd in brachytherapy with (252)Cf source will depend not only on the capture product dose level, but also on the tumor distance from the source. (33)S makes dose enhancement under specific conditions that these conditions depend on the neutron energy spectra of source, the (33)S concentration in tumor and tumor distance from the source.
... The capture product dose (absorbed dose by capture materials) resulted from the capture of thermal neutrons by 10 B, 157 Gd and 33 S was calculated using the fluence-to-kerma conversion factors [13]. The neutron dose is the sum of source fast neutron dose resulted from elastic scattering of fast neutrons in water and the capture product dose which is resulted from thermal neutron capture by 10 B, 157 Gd and 33 S. The neutron energy spectrum for 252 Cf source was assumed to be Maxwellian spectrum with an average energy of 2.1 MeV and the most probable energy of ~0.7 MeV [14] . Photon spectrum of the 252 Cf source was taken from the study by Fortune, and has photon energies in the range of 0.01–9.79 ...
Abstract
Background: In neutron interaction with matter and reduction of neutron energy due to multiple scatterings to the thermal energy range, increasing the probability of thermal neutron capture by neutron captures makes dose enhancement in the tumors loaded with these materials.
Objective: The purpose of this study is to evaluate dose distribution in the presence of 10B, 157Gd and 33S neutron capturers and to determine the effect of these materials on dose enhancement rate for 252Cf brachytherapy source.
Methods: Neutron-ray flux and energy spectra, neutron and gamma dose rates and dose enhancement factor (DEF) are determined in the absence and presence of 10B, 157Gd and 33S using Monte Carlo simulation.
Results: The difference in the thermal neutron flux rate in the presence of 10B and 157Gd is significant, while the flux changes in the fast and epithermal energy ranges are insensible. The dose enhancement factor has increased with increasing distance from the source and reached its maximum amount equal to 258.3 and 476.1 cGy/h/µg for 157Gd and 10B, respectively at about 8 cm distance from the source center. DEF for 33S is equal to one.
Conclusion: Results show that the magnitude of dose augmentation in tumors containing 10B and 157Gd in brachytherapy with 252Cf source will depend not only on the capture product dose level, but also on the tumor distance from the source. 33S makes dose enhancement under specific conditions that these conditions depend on the neutron energy spectra of source, the 33S concentration in tumor and tumor distance from the source.
... One microgram of 252 Cf emits 1.3 × 10 7 photons/s 16 and has a neutron yield of about 2.314 × 10 6 neutrons/s. 17 These values were utilized in the process of conversion of the Monte Carlo output (MeV/g) to the dose rate in terms of (cGy/(h g)). ...
Aim:
The aim of this study is to assess the effect of the compositions of various soft tissues and tissue-equivalent materials on dose distribution in neutron brachytherapy/neutron capture therapy.
Background:
Neutron brachytherapy and neutron capture therapy are two common radiotherapy modalities.
Materials and methods:
Dose distributions were calculated around a low dose rate (252)Cf source located in a spherical phantom with radius of 20.0 cm using the MCNPX code for seven soft tissues and three tissue-equivalent materials. Relative total dose rate, relative neutron dose rate, total dose rate, and neutron dose rate were calculated for each material. These values were determined at various radial distances ranging from 0.3 to 15.0 cm from the source.
Results:
Among the soft tissues and tissue-equivalent materials studied, adipose tissue and plexiglass demonstrated the greatest differences for total dose rate compared to 9-component soft tissue. The difference in dose rate with respect to 9-component soft tissue varied with compositions of the materials and the radial distance from the source. Furthermore, the total dose rate in water was different from that in 9-component soft tissue.
Conclusion:
Taking the same composition for various soft tissues and tissue-equivalent media can lead to error in treatment planning in neutron brachytherapy/neutron capture therapy. Since the International Commission on Radiation Units and Measurements (ICRU) recommends that the total dosimetric uncertainty in dose delivery in radiotherapy should be within ±5%, the compositions of various soft tissues and tissue-equivalent materials should be considered in dose calculation and treatment planning in neutron brachytherapy/neutron capture therapy.
... The photon spectrum of the 252 Cf source has photon energies in the range of 10 keV-10 MeV [29]. One microgram of 252 Cf source emits 2.314 × 10 6 fast neutrons/s [30] and also has a photon yield of about 1.3 × 10 7 photons/s [31]. Maxwellian fission distribution has been utilized as the distribution for the neutrons emitted by the 252 Cf source. ...
The aim of this study is evaluation of the effect of diameter of (10)B nanoparticles and various neutron capture cross-section libraries on macroscopic dose enhancement in boron neutron capture therapy (BNCT).
MCNPX Monte Carlo code was used for simulation of a (252)Cf source, a soft tissue phantom and a tumor containing (10)B nanoparticles. Using (252)Cf as a neutron source, macroscopic dose enhancement factor (MDEF) and total dose rate in tumor in the presence of 100, 200, and 500 ppm of (10)B nanoparticles with 25 nm, 50 nm, and 100 nm diameters were calculated. Additionally, the effect of ENDF, JEFF, JENDL, and CENDL neutron capture cross-section libraries on MDEF was evaluated.
There is not a linear relationship between the average MDEF value and nanoparticles' diameter but the average MDEF grows with increased concentration of (10)B nanoparticles. There is an increasing trend for average MDEF with the tumor distance. The average MDEF values were obtained the same for various neutron capture cross-section libraries. The maximum and minimum doses that effect on the total dose in tumor were neutron and secondary photon doses, respectively. Furthermore, the boron capture related dose component reduced in some extent with increase of diameter of (10)B nanoparticles.
Based on the results of this study, it can be concluded that from physical point of view, various nanoparticle diameters have no dominant effect on average MDEF value in tumor. Furthermore, it is concluded that various neutron capture cross-section libraries are resulted to the same macroscopic dose enhancements. However, it is predicted that taking into account the biological effects for various nanoparticle diameters will result in different dose enhancements.
... Californium-252 sources of up to 25Ci can be leased by universities where security arrangements may be much weaker than in industry. 25 Polonium-210, made notorious through the Litvinenko affair, is generally found in smaller quantities. An individual static eliminator unit may contain only 0.11Ci, but hundreds of such units may be used or stored in the same facility. ...
When applying photon radiotherapy at the cervical carcinoma it has been stated that, in spite of an important progress in the radiotherapy technique and quality assurance, and consiquently of different radiosensitivity in various tumorous populations, a pronounced progress of curative results has not been reached. The application of 252Cf as a source of gamma-neutron radiation in brachytherapy creates certain presuppositions to overcome this resistency. Design and Subjects: Since Janauary 1985 and till December 1992, 294 patients with a cervical carcinoma were, in a randomized study, cured out of which 184 (81 IIb. and 103 IIIb. grades) were treated intracavitarily 252Cf and 110 (50 IIb. and 60 IIIb. grades) patients were treated by gamma radiation only. As criteria of randomized choice the patients, age, progress of illness, histology and histological grading were applied. Methods and Results: In all patients an equal dosis of 56 Gy-equivalent was intracavitarily applied, supported by external radiation of 40 Gy. The total applied dosis in point A was 85 Gy, or 59 Gy in point B, resp. In patients treated by 252Cf these sources were applied in the first treatment week while using the following schemes: in 23 patients only a dosis of 9 Gy (56 Gy-eq) of the neutron component was applied, whereas in 117 and 44 patients a dosis of 6 Gy (40 Gy-eq) or 2 Gy (16 Gy-eq) of the neutron component were used, added by gamma radiation of 16 or 40 Gy, resp. RBE of the neutron component had the value 6. In patients treated by gamma radiation was the intracavitary dosis of 56 Gy applied in the two fraction. The resuslts of 5-year survival scoring in 252Cf patients are 14,3% (70,6% vs. 56,3%; p < 0,05) better when compared with a conventional treatment, and they are even better than that in more progressive grades of illness where in IIb. grade this difference makes 11,9% (83,9% vs. 72.0%), and IIIb grade making 16,9% (60,2% vs. 43,3%; p < 0,05). A better survival scoring in 252Cf patients is the result of a significant decrease of relapse cases in the pelvis of 16,1% (20,7% vs. 31,8%), which in IIIb. gr. cases amoounts to 24,7% (20,3% vs. 45,0%). The 252Cf application in the brachyterapy of cervical carcinoma significantly reduces the early postirradiation proctitides and minimizes treatment risks. Conclusions: The 252Cf application in the brachytherapy of cervical carcinoma creates totally new quality prerequisites for significantly better treatment results and a lowering of side effect and treatment complications.
The Californium User Facility for Neutron Science has been established at Oak Ridge National Laboratory (ORNL). The Californium User Facility (CUF) is a part of the larger Californium Facility, which fabricates and stores compact ²âµÂ²Cf neutron sources for worldwide distribution. The CUF can provide a cost-effective option for research with ²âµÂ²Cf sources. Three projects at the CUF that demonstrate the versatility of ²âµÂ²Cf for biological and biomedical neutron-based research are described: future establishment of a ²âµÂ²Cf-based neutron activation analysis system, ongoing work to produce miniature high-intensity, remotely afterloaded ²âµÂ²Cf sources for tumor therapy, and a recent experiment that irradiated living human lung cancer cells impregnated with experimental boron compounds to test their effectiveness for boron neutron capture therapy.
A product of the nuclear age, Californium-252 (²âµÂ²Cf) has found many applications in medicine, scientific research, industry, and nuclear science education. Californium-252 is unique as a neutron source in that it provides a highly concentrated flux and extremely reliable neutron spectrum from a very small assembly. During the past 40 years, ²âµÂ²Cf has been applied with great success to cancer therapy, neutron radiography of objects ranging from flowers to entire aircraft, startup sources for nuclear reactors, fission activation for quality analysis of all commercial nuclear fuel, and many other beneficial uses, some of which are now ready for further growth. Californium-252 is produced in the High Flux Isotope Reactor (HFIR) and processed in the Radiochemical Engineering Development Center (REDC), both of which are located at the Oak Ridge National Laboratory (ORNL) in Oak Ridge, Tennessee. The REDC/HFIR facility is virtually the sole supplier of ²âµÂ²Cf in the western world and is the major supplier worldwide. Extensive exploitation of this product was made possible through the ²âµÂ²Cf Market Evaluation Program, sponsored by the United States Department of Energy (DOE) [then the Atomic Energy Commission (AEC) and later the Energy Research and Development Administration (ERDA)]. This program included training series, demonstration centers, seminars, and a liberal loan policy for fabricated sources. The Market Evaluation Program was instituted, in part, to determine if large-quantity production capability was required at the Savannah River Laboratory (SRL). Because of the nature of the product and the means by which it is produced, ²âµÂ²Cf can be produced only in government-owned facilities. It is evident at this time that the Oak Ridge research facility can meet present and projected near-term requirements. The production, shipment, and sales history of ²âµÂ²Cf from ORNL is summarized herein.
The Nuclear Materials Identification System (NMIS), developed by the Oak Ridge National Laboratory and Oak Ridge Y-12 Plant (Y-12), has been successfully used at Y-12 for nuclear material control and accountability (NMC&A). It is particularly useful in the high gamma-ray background of storage arrays and for shielded HEU. With three systems in use at Y-12, NMIS has enhanced the NMC&A capability for verification and for confirmation of materials in storage and for HEU receipts by providing capability not available or practical by other NDA methods for safeguards. It has recently cost-effectively quantified the HEU mass and enrichment of hundreds of HEU metal items to within a total spread of {+-} 5% (3 sigma) with and mean deviations for all HEU verified of + 0.2% for mass and -0.2% for enrichment. Three cart portable systems are easily moved around with minimal impact on facility operations since no permanent dedicated floor space is required. The positive impact of NMIS at the Oak Ridge Y-12 Plant is improved and more cost effective NMC&A as well as the resolution of NMC&A findings. Its operation at the Y-12 Plant is essential for compliance with the NMC&A requirements of the US Department of Energy. NMIS portability has allowed one system to be moved temporarily to the former K-25 Gaseous Diffusion Plant for characterization of a large deposit of hydrated uranyl fluoride. The impact of this NMIS application was enhanced and verified nuclear criticality safety that led to the safe removal of a large deposit originally estimated by gamma-ray spectrometry and neutron counting to contain 1300 kg of 3.3 wt% ²³âµU material. NMIS has also been operational at Los Alamos National Laboratory and Pantex.
A thermal neutron analysis (TNA) detector was developed as a confirmatory sensor for the Canadian ILDP multisensor teleoperated vehicle mounted land mine detector system. ILDP is the only multisensor mine detection system with a confirmation sensor which can reduce false alarms to acceptable levels. The TNA has been experimentally proven to be capable of detection of anti-tank and large anti- personnel mines in acceptable short time periods in its intended role. It has performed well, in extreme climates, in Canadian and U.S. stand-alone tests and in U.S. tests of the complete ILDP system. Current research is aimed at developing a version which is ready for production and field.
Neutron activation analysis using low-flux isotopic neutron sources is put to use in addressing areas of applied interest in managing the Savannah River Site. Some of the applications are unique due the site's operating history and its chemical processing facilities. Because sensitivity needs for many of the applications are not severe, they can be accomplished using a similar to 6 mg Cf-252 neutron activation analysis facility. Overviews of the facility and several example applications are presented.
Results on the radiation hardness of photodiodes to fast neutrons are presented. Four photodiodes (three avalanche photodiodes from two manufacturers, and one PIN photodiode) were exposed to neutrons from a source at Oak Ridge National Laboratory. The effects of this radiation on many parameters such as gain, intrinsic dark current, quantum efficiency, noise, capacitance, and voltage and temperature coefficients of the gain for these devices for fluences up to ∼2×1013 neutrons/cm2 are shown and discussed. While degradation of APDs occurred during neutron irradiation, they remained photosensitive devices with gain.
Nd–Fe–B permanent magnets are highly desirable for use in the insertion devices of synchrotron radiation sources due to their high remanence, or residual magnetic induction, and intrinsic coercivity. However, the radiation environment within high-energy storage rings makes essential the determination of the degree of radiation sensitivity as well as the mechanisms of radiation-induced demagnetization. Sample Nd–Fe–B permanent magnets were irradiated at the Advanced Photon Source with bending magnet X-rays up to an absorbed dose of approximately 280 Mrad (1 Mrad=10kGy). Sample magnets were also irradiated with 60Co γ-rays up to an absorbed dose of 700 Mrad at the National Institute of Standards and Technology's standard gamma irradiation facility. Changes in the residual induction were found to be within the experimental uncertainties for both the X-ray and γ-ray irradiations. Sample Nd–Fe–B permanent magnets were then irradiated at Oak Ridge National Laboratory's Californium User Facility for Neutron Science with fast neutrons up to a total fast fluence of 1.61×1014 n/cm2, and with thermal neutrons up to a total thermal fluence of 2.94×1012 n/cm2. The fast-neutron irradiation revealed changes between residual induction measurements of the sample magnets before and after irradiation of 0.6% and greater for fast-neutron fluence levels of 2×1013 n/cm2 and above. Thermal-neutron irradiation revealed changes in the residual induction measurements of the sample magnets before and after irradiation that were within the experimental uncertainties for thermal-neutron fluence levels up to 3×1012 n/cm2.
Experiments were carried out to investigate the possible use of neutron backscattering for the detection of landmines buried in the soil. Several landmines, buried in a sand-pit, were positively identified. A series of Monte Carlo simulations were performed to study the complexity of the neutron backscattering process and to optimize the geometry of a future prototype. The results of these simulations indicate that this method shows great potential for the detection of non-metallic landmines (with a plastic casing), for which so far no reliable method has been found.
The radioisotope 252Cf is routinely encapsulated into compact, portable, intense neutron sources with a 2.6-yr half-life. A source the size of a person's little finger can emit up to 10(11) neutrons s(-1). Californium-252 is used commercially as a reliable, cost-effective neutron source for prompt gamma neutron activation analysis (PGNAA) of coal, cement and minerals, as well as for detection and identification of explosives, land mines and unexploded military ordinance. Other uses are neutron radiography, nuclear waste assays, reactor start-up sources, calibration standards and cancer therapy. The inherent safety of source encapsulations is demonstrated by 30 yr of experience and by US Bureau of Mines tests of source survivability during explosions. The production and distribution center for the US Department of Energy (DOE) Californium Program is the Radiochemical Engineering Development Center (REDC) at Oak Ridge National Laboratory (ORNL). DOE sells 252Cf to commercial reencapsulators domestically and internationally. Sealed 252Cf sources are also available for loan to agencies and subcontractors of the US government and to universities for educational, research and medical applications. The REDC has established the Californium User Facility (CUF) for Neutron Science to make its large inventory of 252Cf sources available to researchers for irradiations inside uncontaminated hot cells. Experiments at the CUF include a land mine detection system, neutron damage testing of solid-state detectors, irradiation of human cancer cells for boron neutron capture therapy experiments and irradiation of rice to induce genetic mutations.
We describe a new and novel means for detecting nonmetallic land mines. Timed neutron detection (TND) utilizes the unique neutron-moderating properties common to most explosives and all polymers and incorporates timing aspects to improve the measurement process. TND offers the possibility of a low-cost detector for non-metallic land mines with small radiation exposure. Experimental results with actual land mines and simulants are described. Computer simulations provide an understanding of how the performance of detectors based on this approach will be affected by environmental and operator parameter variations.
A small-diameter, double-contained <sup>252</sup>Cf fission
chamber design that can be inserted into a 0.95-cm (0.375-in) diameter
hole or tube (such as the control rod guide tubes in pressurized water
reactor (PWR) fuel elements) is described. This fission chamber can be
used as a neutron driver for subcriticality measurements with irradiated
PWR fuel elements. This prototype contains ~0.1 μg of
<sup>252</sup>Cf, but the design has the flexibility to accommodate
larger amounts of fissionable material required for PWR tests. Details
of the design are presented, as well as some of the instrument's
operational parameters and performance characteristics
A nondestructive assay method to identify chemical warfare (CW)
agents and high explosive (HE) munitions was tested with actual chemical
agents and explosives. The assay method exploits the gamma radiation
produced by neutron interactions inside a container or munition to
identify the elemental composition of its contents. The characteristic
gamma-ray signature of the chemical elements chlorine, phosphorus, and
sulfur were observed from the CW agent containers and munitions, in
sufficient detail to allow identification of agents GC [sarin], HD
[mustard gas] and VX from one another, and from HE-filled munitions. By
detecting of the presence of nitrogen, the key indicator of explosive
compounds, and the absence of elements Cl, P, and S, HE shells were also
clearly identified
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Neutron Activation Analysis of Environmental Samples at the Californium User Facility
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MCNP™—A General Monte Carlo N-Particle Transport Code, Version 4B
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