Monte Carlo N-Particle Version Research Articles

Specific Absorbed Fractions for Spontaneous Fission Neutron Emitters in the ICRP Reference Pediatric Voxel Phantom Series.

Specific absorbed fractions (SAFs) are key components in the workflow of internal exposure assessment following the intake of a radionuclide, allowing quick conversion of particle energy released in a source region to the expected absorbed dose in target regions throughout the body. For data completeness, SAFs for spontaneous fission neutron emitters are currently needed for the recently adopted ICRP reference pediatric voxel phantom series. With 77 source regions within each reference individual and 28 radionuclides decaying via spontaneous fission, full Monte Carlo simulation requires significant computation time. In order to reduce this burden, a novel method for neutron SAF estimation was undertaken. The Monte Carlo N-Particle version 6.1 (MCNP6) simulation package was chosen to simulate the 252 Cf Watt fission neutron spectrum originating from 15 source regions in each phantom; dose estimation within 41 target tissues allowed for assessment of the SAF value for each source-target pair. For the remaining source regions, chord length distributions were computed using MATLAB code to determine the separation between the source-target pairs within the pediatric phantom series. These distance distributions were used in conjunction with a 252 Cf neutron dose point kernel calculated in soft tissue, which was modified to account for the source region's depth from the surface of the body. Lastly, the 252 Cf SAF dataset was extended to the other 27 spontaneous fission neutron emitters based on differences in the Watt fission spectrum parameters of each radionuclide. This methodology has been shown to accurately estimate spontaneous fission neutron SAFs to within 20% of the Monte Carlo estimated value for most source-target pairs in the ICRP reference pediatric series.

Neutronic Performance of a Compact 10 MWe Nuclear Reactor with Low Enrichment (ThxU1\u2212x)N Fuel

Recent interest in compact nuclear reactors for applications in space or in remote locations drives innovation in nuclear fuel design, especially non-oxide ceramic nuclear fuels. This work details neutronic modeling designed to support the development of a new nuclear fuel concept based on a mixture of thorium and uranium nitride. A Monte Carlo N-Particle Version 6.2 (MCNP-6) model of a compact 10 MWe reactor design which incorporates (ThxU1−x)N fuel is presented. In this context, a “compact” reactor is a completely assembled reactor which may be emptied of coolant and transported by specialized commercial vehicle, deployed by a C130J aircraft, or launched into space. Core geometry, reflector barrels, and the heat exchange zones are designed to support reduction of overall reactor volume of core components while maintaining criticality with a fixed total fuel mass of 4500 kg. Dense mixed nitrides of thorium nitride (ThN) additions in uranium nitride (UN) in 5 wt.% increments between 0.05 le x le 0.5 have been considered for calculation of k_{infty } and k_{{{text{effective}}}}. ThN additions in UN results in a slight increase in the magnitude of the temperature coefficient of reactivity, which is negative by design. The isotopic distribution of the principal actinide inventory as a function of burnup, time, and initial fuel composition is presented and discussed within the context of the proliferation risk of this core design.

Safety Analysis of Indonesia's Multipurpose Reactor During Irradiation of Natural UO2 Pin Target at Power Ramp Test Facility

Abstract To support the utilization of the power ramp test facility (PRTF) at the reactor Serba Guna GA Siwabessy (RSG-GAS) reactor, the investigation of safety parameters both on the target and the reactor during the irradiation of natural UO2 has been conducted. The program for the analysis of reactor transient (PARET) and Monte Carlo N-particle version 6 (MCNP6) codes were integrated for thermal-hydraulic and neutronic analysis under the reactivity insertion phenomenon. The reactivity insertion and linear heat rate (LHR) were calculated by MCNP6 to confirm that the maximum values are well below the safety limits. Also, the code computed the reactor kinetic parameters to provide inputs for the PARET code to evaluate the reactor response due to the target movement at the PRTF. The target moves from the farthest position to the nearest one traveling 440 mm and then moves back to the original position. The safety analysis indicated that both maximum reactivity insertion and LHR values did not exceed the safety limits. The target induces maximum reactivity insertion of 0.023%dk/k and produces the maximum LHR of 137.53 W/cm and 275.06 W/cm; when the reactor operates at 15 MW and 30 MW, respectively. The PARET simulation results showed that the reactor power oscillated due to the target movement. Therefore, the recommended velocities of the target are below 0.17 m/s and 0.08 m/s for reactor power of 15 MW and 30 MW, respectively, to avoid the reactor scram, and no safety limits are exceeded.

Startup Test Plan and Predictions for Highly Enriched Uranium to Low-Enriched Uranium Fuel Conversion at the University of Missouri Research Reactor

Nonpower reactors licensed by the U.S. Nuclear Regulatory Commission require a startup test plan as part of any facility modification to verify operability prior to resumption of operations. In order to support conversion of the University of Missouri Research Reactor from the use of highly enriched uranium to low-enriched uranium (LEU) fuel, a startup test plan has been devised to measure certain reactor physics parameters for the initial all-fresh LEU core licensing documentation that will be submitted. These parameters include the approach to critical, primary coolant void coefficient of reactivity, flux trap void coefficient of reactivity, determination of flux trap sample reactivity worth, radial and axial thermal neutron flux mapping, control blade worth calibration, primary and pool coolant temperature coefficient of reactivity, and flux mapping of experimental positions. In this paper, predictions for these parameters made using the Monte Carlo N-Particle Version 5 (MCNP5) radiation transport code are reported. These predictions will support the startup tests by providing a baseline set of expectations and additional insight into the performance of the LEU core.

Determination of Dose Distributions by High-energy Electrons in Alumina Pellets Using Monte Carlo Simulations

Electron beam accelerators are broadly used to study the effects of ionizing radiation on the properties of materials. These accelerators are also used to study radiation effects in some important ceramics such as alumina. The produced high-energy electrons at MeV energies will deposit doses that are typically non-uniformly distributed within the entire material. Such distributions provide valuable information regarding the degree of internal and surface changes owing to the irradiation. In the present study, the Monte Carlo (MC) method was used to obtain the dose distributions of high-energy electrons to alumina ceramic in pellet form. The distributions obtained via MCNP5 (Monte Carlo N-Particle version 5) were normalized by experimental values. For the energy of the simulated electrons, both monoenergetic and polyenergetic source electron energies were evaluated. The distributions of the depth-dose, depth-energy, and the top and bottom surface 2D dose mesh were obtained. Results showed that either assumption of monoenergetic or polyenergetic source electron energies resulted in similar depth-dose curves. The depth-energy results showed a larger spread in energy distributions for polyenergetic as compared to monoenergetic source electrons; nonetheless, the average doses for both were similar. The 2D dose maps showed a uniform dose distribution at the top pellet surface and a significantly non-uniform distribution at the bottom of the pellet. These results emphasize the effectiveness of the MC method in determining non-uniform doses within materials, which can be used in quantifying the changes in the material properties at each separate area of the electron irradiated object.

Comparison of Lead, Copper and Aluminium as Gamma Radiation Shielding Material through Experimental Measurements and Simulation Using MCNP Version 4c

The paper compares Lead, Copper and Aluminium as gamma radiation shielding material using both experimental and simulation techniques. Cs- 137 (662KeV), Na-22 (511KeV) and Na- 22(1274KeV) were used as gamma radiation sources and a sodium iodide (NaI) detector was used to detect the radiation. Variations were noted for detected gamma count rates by changing shielding material thickness. In the experimental approach, thickness was varied by placing sheets of a particular test material one by one. For simulation, Monte Carlo n- Particle (MCNP) code version 4c was used and the geometry of the whole experimental setup was plotted in it. The results were then compared for each test material and it was found that lead is the best shielding material for gamma radiation followed by copper and aluminium.

Dosimetric and microdosimetric analyses for blood exposed to reactor-derived thermal neutrons

Thermal neutrons are found in reactor, radiotherapy, aircraft, and space environments. The purpose of this study was to characterise the dosimetry and microdosimetry of thermal neutron exposures, using three simulation codes, as a precursor to quantitative radiobiological studies using blood samples. An irradiation line was designed employing a pyrolytic graphite crystal or—alternatively—a super mirror to expose blood samples to thermal neutrons from the National Research Universal reactor to determine radiobiological parameters. The crystal was used when assessing the relative biological effectiveness for dicentric chromosome aberrations, and other biomarkers, in lymphocytes over a low absorbed dose range of 1.2–14 mGy. Higher exposures using a super mirror will allow the additional quantification of mitochondrial responses. The physical size of the thermal neutron fields and their respective wavelength distribution was determined using the McStas Monte Carlo code. Spinning the blood samples produced a spatially uniform absorbed dose as determined from Monte Carlo N-Particle version 6 simulations. The major part (71%) of the total absorbed dose to blood was determined to be from the 14N(n,p)14C reaction and the remainder from the 1H(n,γ)2H reaction. Previous radiobiological experiments at Canadian Nuclear Laboratories involving thermal neutron irradiation of blood yielded a relative biological effectiveness of 26 ± 7. Using the Particle and Heavy Ion Transport Code System, a similar value of ∼19 for the quality factor of thermal neutrons initiating the 14N(n,p)14C reaction in soft tissue was determined by microdosimetric simulations. This calculated quality factor is of similar high value to the experimentally-derived relative biological effectiveness, and indicates the potential of thermal neutrons to induce deleterious health effects in superficial organs such as cataracts of the eye lens.

Assessment of gamma radiation shielding properties of concrete containers containing recycled coarse aggregates

This paper presents an assessment of gamma radiation shielding performance, specifically in terms of surface dose rate, of concrete containing virgin and recycled coarse aggregates (RCA) to be used for a transportable concrete container for radioactive waste. In order to evaluate radiation shielding performance of the transportable concrete container a numerical simulation method is performed using Monte Carlo N-particle version 6.1 (MCNP6.1). Prior to evaluating radiation shielding performance of the transportable concrete containers, radiation shielding properties of four different concrete mixtures which have two different compressive strengths, 40 and 70 MPa, containing natural coarse aggregate (NCA) and recycled coarse aggregate (RCA), respectively, are assessed using experimental and numerical simulation methods in terms of linear attenuation coefficient (μ) to verify the reliability of the numerical simulation method. Density, compressive strength, and static modulus of elasticity tests are conducted to determine the parameters for transportable concrete container design. Based on the physical properties of concrete a transportable concrete container is designed. In order to assess the radiation shielding with respect to the maximum surface dose rate of the concrete containers, three different radioactive wastes are assumed to be loaded in the containers with metal drums and radiation shielding analysis is carried out using MCNP6 simulation code for the four different concrete containers, comparing with the existing carbon steel container. It was found that the results from numerical simulation are in good agreement with the experimentally determined results. In terms of the maximum surface dose rate, all concrete containers showed considerably lower surface dose rate than the existing carbon steel container.

Effect of age-dependent bone electron density on the calculated dose distribution from kilovoltage and megavoltage photon and electron radiotherapy in paediatric MRI-only treatment planning.

MRI-only treatment planning (TP) can be advantageous in paediatric radiotherapy. However, electron density extraction is necessary for dose calculation. Normally, after bone segmentation, a bulk density is assigned. However, the variation of bone bulk density in patients makes the creation of pseudo CTs challenging. This study aims to assess the effects of bone density variations in children on radiation attenuation and dose calculation for MRI-only TP. Bone contents of <15-year-old children were calculated, and substituted in the Oak Ridge National Laboratory paediatric phantoms. The percentage depth dose and beam profile of 150 kVp and 6 MV photon and 6 MeV electron beams were then calculated using Xcom, MCNPX (Monte Carlo N-particle version X) and ORLN phantoms. Using 150 kVp X-rays, the difference in attenuation coefficient was almost 5% between an 11-year-old child and a newborn, and ~8% between an adult and a newborn. With megavoltage radiation, the differences were smaller but still important. For an 18 MV photon beam, the difference of radiation attenuation between an 11-year-old child and a newborn was 4% and ~7.4% between an adult and a newborn. For 6 MeV electrons, dose differences were observed up to the 2 cm depth. The percentage depth dose difference between 1 and 10-year-olds was 18.5%, and between 10 and 15-year-olds was 24%. The results suggest that for MRI-only TP of photon- or electron-beam radiotherapy, the bone densities of each age group should be defined separately for accurate dose calculation. Advances in knowledge: This study highlights the need for more age-specific determination of bone electron density for accurate dose calculations in paediatric MRI-only radiotherapy TP.

Evaluation of tomographic image quality of extended and conventional parallel hole collimators using maximum likelihood expectation maximization algorithm by Monte Carlo simulations.

One of the major problems associated with parallel hole collimators (PCs) is the trade-off between their resolution and sensitivity. To solve this problem, a novel PC - namely, extended parallel hole collimator (EPC) - was proposed, in which particular trapezoidal denticles were increased upon septa on the side of the detector. In this study, an EPC was designed and its performance was compared with that of two PCs, PC35 and PC41, with a hole size of 1.5 mm and hole lengths of 35 and 41 mm, respectively. The Monte Carlo method was used to calculate the important parameters such as resolution, sensitivity, scattering, and penetration ratio. A Jaszczak phantom was also simulated to evaluate the resolution and contrast of tomographic images, which were produced by the EPC6, PC35, and PC41 using the Monte Carlo N-particle version 5 code, and tomographic images were reconstructed by using maximum likelihood expectation maximization algorithm. Sensitivity of the EPC6 was increased by 20.3% in comparison with that of the PC41 at the identical spatial resolution and full-width at tenth of maximum here. Moreover, the penetration and scattering ratio of the EPC6 was 1.2% less than that of the PC41. The simulated phantom images show that the EPC6 increases contrast-resolution and contrast-to-noise ratio compared with those of PC41 and PC35. When compared with PC41 and PC35, EPC6 improved trade-off between resolution and sensitivity, reduced penetrating and scattering ratios, and produced images with higher quality. EPC6 can be used to increase detectability of more details in nuclear medicine images.

Development of a method for on-line determination of chlorine impurity in crude oil by using fast neutrons

An important challenge in the petroleum industry is to reduce the salt content to acceptable levels in crude oil. Continuous monitoring of crude oil salinity is thus an important aspect. However, development of safe and reliable on-line and non-intrusive monitoring systems is still a challenging task. In this work, a method for non-intrusive, on-line monitoring of the chlorine impurity in crude oil based on prompt gamma-ray neutron activation analysis (PGNAA) in conjunction with a portable pulsed neutron generator producing 14MeV neutrons in deuterium-tritium (DT) nuclear reactions is proposed and tested. Simulations with Monte Carlo N-Particle version 6 (MCNP6) transport code combined with Center for Engineering Applications of Radioisotopes Detector Response Function (CEARDRF) and a specific purpose Monte Carlo code called Center for Engineering Applications of Radioisotopes Coincidence Prompt Gamma-Ray (CEARCPG) were used for the design optimization of the experiments. Preliminary results indicate that a minimum detection limit for chlorine impurity of about 71mg/L is achievable for an integration time of 30min.

Technical Note: Monte Carlo calculations of the AAPM TG-43 brachytherapy dosimetry parameters for a new titanium-encapsulated Yb-169 source.

Due to a number of distinct advantages resulting from the relatively low energy gamma ray spectrum of Yb‐169, various designs of Yb‐169 sources have been developed over the years for brachytherapy applications. Lately, Yb‐169 has also been suggested as an effective and practical radioisotope option for a novel radiation treatment approach often known as gold nanoparticle‐aided radiation therapy (GNRT). In a recently published study, the current investigators used the Monte Carlo N‐Particle Version 5 (MCNP5) code to design a novel titanium‐encapsulated Yb‐169 source optimized for GNRT applications. In this study, the original MC source model was modified to accurately match the specifications of the manufactured Yb‐169 source. The modified MC model was then used to obtain a complete set of the AAPM TG‐43 parameters for the new titanium‐encapsulated Yb‐169 source. The MC‐calculated dose rate constant for this titanium‐encapsulated Yb‐169 source was 1.19 ± 0.03 cGy·h−1·U−1, indicating no significant change from the values reported for stainless steel‐encapsulated Yb‐169 sources. The source anisotropy and radial dose function for the new source were also found similar to those reported for the stainless steel‐encapsulated Yb‐169 sources. The current results suggest that the use of titanium, instead of stainless steel, to encapsulate the Yb‐169 core would not lead to any major change in the dosimetric characteristics of the Yb‐169 source. The results also show that the titanium encapsulation of the Yb‐169 source could be accomplished while meeting the design goals as described in the current investigators’ published MC optimization study for GNRT applications.

Development of remaining wall thickness measurement system for boiler wall tube using gamma scattering technique

In this study, the General Monte Carlo N-Particle version 5 (MCNP5) simulation of the measuring system for the remaining thickness of the wall tube using gamma scattering technique is performed to investigate the applicability of this technique and to find the optimum geometrical setup for the experimental setup. The numerical results show that the optimal geometry condition to provide the highest ratio between the gamma flux changes per thickness variation, namely, the measurement sensitivity, is the alignment of the source incidence angle of 30 degree and detector scattering angle of 30 degrees. Sequentially, the preliminary experiment of thickness measurement system for the tube using the gamma scattering technique is conducted based on the selected geometrical test setup by the simulation. It is found that the experimental results have generally a good agreement with the calculated results. Conclusively, it is suggested that the gamma scattering technique has a potential method to measure the remaining wall thickness in the boiler wall tube.

Realistic dosimetry for studies on biological responses to X-rays and γ-rays

ABSTRACTA calibration coefficient R (= DA/DE) for photons was employed to characterize the photon dose in radiobiological experiments, where DA was the actual dose delivered to cells and DE was the dose recorded by an ionization chamber. R was determined using the Monte Carlo N-Particle version 5 (MCNP-5) code. Photons with (i) discrete energies, and (ii) continuous-energy distributions under different beam conditioning were considered. The four studied monoenergetic photons had energies E = 0.01, 0.1, 1 and 2 MeV. Photons with E = 0.01 MeV gave R values significantly different from unity, while those with E > 0.1 MeV gave R ≈ 1. Moreover, R decreased monotonically with increasing thickness of water medium above the cells for E = 0.01, 1 or 2 MeV due to energy loss of photons in the medium. For E = 0.1 MeV, the monotonic pattern no longer existed due to the dose delivered to the cells by electrons created through the photoelectric effect close to the medium–cell boundary. The continuous-energy distributions from the X-Rad 320 Biological Irradiator (voltage = 150 kV) were also studied under three different beam conditions: (a) F0: no filter used, (b) F1: using a 2 mm-thick Al filter, and (c) F2: using a filter made of (1.5 mm Al + 0.25 mm Cu + 0.75 mm Sn), giving mean output photon energies of 47.4, 57.3 and 102 keV, respectively. R varied from ~1.04 to ~1.28 for F0, from ~1.13 to ~1.21 for F1, and was very close to unity for F2.

Use of novel thermobrachytherapy seeds for realistic prostate seed implant treatments.

A practical means of delivering both therapeutic radiation and hyperthermia to a deep-seated target has been identified in the literature as highly desirable, provided it is capable of generating sufficient temperatures over the defined target volume. The authors present continued development of a dual-modality thermobrachytherapy (TB) seed, investigating its capabilities in delivering prescribed hyperthermia to realistic deep-seated targets. The TB seed is based on the ubiquitous low dose-rate (LDR) brachytherapy permanent implant. Heat is generated by incorporating a ferromagnetic core within the seed and placing the patient in an oscillating external magnetic field, producing eddy currents within the core and hence Joule heating. A strategically selected Curie temperature results in thermal self-regulation. The magnetic and thermal properties of the TB seed were studied experimentally by means of seed prototypes placed in a tissue-mimicking phantom and heated with an industrial induction heater, as well as computationally in the finite element analysis solver COMSOL Multiphysics. Patient-specific seed distributions derived from LDR permanent prostate implants previously conducted at their institution were modeled in COMSOL to evaluate their ability to adequately cover a defined target volume and to overcome the loss of heat due to blood perfusion within tissue. The calculated temperature distributions were analyzed by generating temperature-volume histograms, which were used to quantify coverage and temperature homogeneity for varied blood perfusion rates, seed Curie temperatures, and thermal power production rates. Use of additional hyperthermia-only (HT-only) seeds in unused spots within the implantation needles was investigated, as was an increase in these seeds' core size to increase their power. The impact of the interseed attenuation and scatter (ISA) effect on radiation dose distributions of this seed was also quantified by Monte Carlo studies in the software package Monte Carlo N-Particle Version 5. Increasing the power production of the seeds, as well as increasing their Curie point, would increase the maximum blood perfusion rate that a given seed distribution could overcome to obtain an acceptable temperature distribution. However, this would also increase the maximum temperatures generated at the seed surfaces. Auxiliary HT-only seeds serve to improve the temperature uniformity within the target, as well as decrease the seed power generation requirements. Both an increase in their core size and an increase in both seed types' Curie temperatures enhance the resulting temperature coverage. The interseed and scatter effect caused by both the TB and HT-only seeds was found to reduce the dose to 90% of the target volume (D90) by a factor of 1.10 ± 0.02. A systematic approach of combining LDR prostate brachytherapy with hyperthermia is described, and its ability to provide sufficient and uniform temperature distributions in realistic patient-specific implants evaluated. A combination of TB and HT-only seeds may be used to produce a uniform temperature distribution in a defined target. Various modeled changes to their design, such as optimization of their Curie temperature, improve their ability to overcome the thermal effects of blood perfusion. The enhanced ISA of the TB and HT-only seeds must be taken into account for dose calculations, but is manageable.

Comparison of filtered x-ray spectra and depth doses derived from a hybrid Monte Carlo model of an orthovoltage x-ray unit with experimental measurements

The Monte Carlo N-Particle Version 5 (MCNP5) code system was used to create an MC model of the head of a Philips RT-250 unit and a 30 × 30 × 30 cm3 water phantom. For the x-ray machine head, this MC model includes the vacuum region, beryllium window, collimators, inherent filters, and exterior steel housing. For increased computational efficiency of MC simulations, the primary x-ray spectrum from the target was pre-calculated using a well-validated analytical software package. The current hybrid MC model was tested for the three common kilovoltage x-ray beams available from the machine: 250 kVp (0.35 mm Cu), 125 kVp (0.1 mm Cu), and 75 kVp (0.1 mm Cu). Specifically, calculated percentage-depth-dose (PDD) values and x-ray spectra were compared with experimental data from film/ionization chamber and Compton-scatter spectrum measurements, respectively. A comparison between the calculated PDD and measured data for all three x-ray beams showed overall agreement within 3%. The measured x-ray spectra were also compared with the MC results showing reasonable agreement between the two, in terms of the spectral shapes (or effects of filtration on the final spectra). In conclusion, despite its simplicity, the current hybrid MC model was capable of predicting the dosimetric and spectral characteristics of each beam from the RT-250 orthovoltage unit, demonstrating its applicability as a convenient and efficient computational tool for various preclinical and clinical applications.

Intratumorally Injected 177Lu-Labeled Gold Nanoparticles: Gold Nanoseed Brachytherapy with Application for Neoadjuvant Treatment of Locally Advanced Breast Cancer.

Improvements in the treatment of locally advanced breast cancer (LABC) are needed. Our objective was to study a radiation nanomedicine (gold nanoseeds) composed of 30-nm gold nanoparticles (AuNP) modified with polyethyleneglycol (PEG) chains linked to DOTA for complexing the β-particle emitter (177)Lu and to panitumumab for targeting epidermal growth factor receptors (EGFR) ((177)Lu-T-AuNP) as a novel neoadjuvant brachytherapy for LABC. Nontargeted gold nanoseeds ((177)Lu-NT-AuNP) were constructed without panitumumab for comparison. (177)Lu-T-AuNP or (177)Lu-NT-AuNP was injected intratumorally in CD-1 athymic mice bearing subcutaneous EGFR-positive MDA-MB-468 human breast cancer tumors. Biodistribution and small-animal SPECT/CT imaging studies were performed to evaluate tumor and normal organ localization. A short-term (15 d) study was conducted to select the most effective amount of (177)Lu-T-AuNP or (177)Lu-NT-AuNP for treatment with long-term observation (90-120 d). Normal organ toxicities were assessed by monitoring body weight, blood cell counts, and serum alanine aminotransferase and creatinine. Radiation-absorbed doses in the tumor and normal organs were estimated by Monte Carlo N-Particle version 5.0 modeling. Tumor radioactivity concentrations were high at 1 h after injection (>300-400 percentage injected dose per gram [%ID/g]) but decreased by 2-3-fold at 48 h after injection. Normal organ uptake was low (<0.5 %ID/g) except for the liver and spleen (<3 %ID/g), increasing by 2-5-fold at 48 h after injection. Treatment with 4.5 MBq (6 × 10(11) AuNP) of (177)Lu-T-AuNP or (177)Lu-NT-AuNP arrested tumor growth over 90 d without normal organ toxicity, whereas tumors continued to grow in mice treated with unlabeled T-AuNP or (177)Lu-labeled PEG polymer not linked to AuNP. Survival was prolonged up to 120 d in mice treated with (177)Lu-T-AuNP or (177)Lu-NT-AuNP. Radiation-absorbed doses to the tumor were 30 and 22 Gy for (177)Lu-T-AuNP and (177)Lu-NT-AuNP, respectively. Some tumor regions received high radiation doses (250-1,300 Gy). Normal organ doses were low (0.04-0.6 Gy). Gold nanoseeds injected intratumorally were highly effective for inhibiting the growth of breast cancer tumors in CD-1 athymic mice and caused no normal organ toxicity. These results are promising for their application for neoadjuvant brachytherapy of LABC. Because EGFR targeting was not required, the approach is broadly applicable to LABC with different phenotypes.

Evaluation of Neutron Radiography Reactor LEU-Core Start-Up Measurements

Benchmark models were developed to evaluate the cold-critical start-up measurements performed during the fresh core reload of the neutron radiography (NRAD) reactor with low-enriched uranium fuel. Experiments include criticality, control rod worth measurements, shutdown margin, and excess reactivity for four core loadings with 56, 60, 62, and 64 fuel elements. The worths of four graphite reflector block assemblies and an empty dry tube used for experiment irradiations were also measured and evaluated for the 60-fuel-element core configuration. Dominant uncertainties in the experimental keff come from uncertainties in the manganese content and impurities in the stainless steel fuel cladding as well as the 236U and erbium poison content in the fuel matrix. Calculations with MCNP5 (Monte Carlo N-Particle version 5-1.60) and ENDF/B-VII.0 neutron nuclear data are ˜1.4% (9σ) greater than the benchmark model eigenvalues, which is commonly seen in Monte Carlo simulations of other TRIGA (Training, Research, Isotopes, General Atomics) reactors. Simulations of the worth measurements are within the 2σ uncertainty for most of the benchmark experiment worth values. The complete benchmark evaluation details are available in the 2014 edition of the International Handbook of Evaluated Reactor Physics Benchmark Experiments.

Low energy electron generator design and depth dose prediction for micro-superficies tumors treatment purposes

We investigate deposited energy and linear energy transfer (LET) of low energy ejection electrons in air and water layers of a generator design via a plasma source. A structured model of a concave cold cathode electron generator was designed and simulated by using Monte Carlo n-particle version X 2.7.0 (MCNPX) code. A negative dc high voltage was applied to a concave cathode up to −12 kV to determine electron energy activity. Results determined that the geometric dimensions of field size toward the anode increased in relation to the angle of the conic beam, widening the accumulated bulks. The increased field size increased the anode current, which also resulted in an increase of electron energy, a reduction in LET, a stretched build-up area and a dose curve that shifted to a higher depth. The biological effect of low energy electron radiation can be increased with an increase of LET; as the depth dose decreased, the electron energy increased at the same time. The study of electron irradiation as a conic beam from an electron generator may provide an accurate investigation of the indirect effect of low energy electrons on bystander cells.

Neutronic analysis of the Geological Survey TRIGA Reactor

The United States Geological Survey TRIGA Reactor (GSTR) is a 1MW reactor located in Lakewood, Colorado. In support of the GSTR’s relicensing efforts, this project developed and validated a Monte Carlo N-Particle Version 5 (MCNP5) model of the GSTR reactor. The model provided estimates of the excess reactivity, power distribution and the fuel temperature, water temperature, void, and power reactivity coefficients for the current and limiting core. The MCNP5 model predicts a limiting core excess reactivity of $6.48 with a peak rod power of 22.2kW. The fuel and void reactivity coefficients for the limiting core are strongly negative, and the core water reactivity coefficient is slightly positive, consistent with other TRIGA analyses. The average fuel temperature reactivity coefficient of the full power limiting core is −0.0135$/K while the average core void coefficient is −0.069$/K from 0% to 20% void. The core water temperature reactivity coefficient is+0.012$/K.