University Research Reactor Research Articles

Slow-positron beamline temperature rise reduction at Kyoto University Research Reactor

Temperature rises of a reactor-based slow-positron beamline at Kyoto University Research Reactor (KUR) were measured during reactor operation and solenoid-coil excitation. During KUR operation at 5 MW, the temperature of the top of the beamline reached 300 °C. This temperature which is close to the melting point (321 °C) of Cd was used to enhance positron generation. On the other hand, the temperature of the flange supporting the beamline was approximately 50 °C and it was sufficiently low in terms of the strength of the beamline. The temperature of the top of the beamline was successfully reduced to 240 °C by introducing a He gas flow around the vacuum duct of the beamline. Beamline temperatures calculated using a finite element method were in agreement with measured temperatures. Such calculation is useful for future experiments with longer or irregular KUR operation.

Release kinetics of tritium generation in neutron irradiated biphasic Li2TiO3–Li4SiO4 ceramic breeder

Biphasic Li2TiO3–Li4SiO4 tritium breeders with different phase ratios were prepared by the solution combustion synthesis method. The as-prepared biphasic Li2TiO3–Li4SiO4 and single-phase materials were irradiated by the thermal neutron with the flux of 2.75 × 1013 n cm−2 s−1 at Kyoto University Research Reactor (KUR). Tritium is generated based on the following reaction: 6Li+1n→3T+4He+4.8 MeV. The results of tritium release experiment show that tritium release starts at the lower temperature range for the biphasic materials. The tritium release temperature and the shape of tritium-TDS (Thermal desorption spectroscopy) spectra are dependent on the phase ratio of Li2TiO3 to Li4SiO4. In the case of Li2TiO3/Li4SiO4 = 2.0, tritium release starts at 400 K, and the tritium migration is mainly controlled by the diffusion process. With the increase of Li4SiO4 content in the biphasic materials, tritium release is influenced by the de-trapping process.

Measurements of thermal-neutron capture cross-section and resonance integral of neptunium-237

ABSTRACT The thermal-neutron capture cross-section (σ 0) and resonance integral (I 0) were measured for the 237Np(n,γ) 238Np reaction by an activation method. A method with a Gadolinium filter, which is similar to the Cadmium difference method, was used to measure the σ 0 with paying attention to the first resonance at 0.489 eV of 237Np, and a value of 0.133 eV was taken as a cut-off energy. Neptunium-237 samples were irradiated at the pneumatic tube of the Kyoto University Research Reactor in Institute for Integral Radiation and Nuclear Science, Kyoto University. Wires of Co/Al and Au/Al alloys were used as monitors to determine thermal-neutron fluxes and epi-thermal Westcott’s indices at an irradiation position. A γ-ray spectroscopy was used to measure activities of 237Np, 238Np and neutron monitors. On the basis of Westcott’s convention, the σ 0 and I 0 values were derived as 186.9 ± 6.2 barn, and 1009 ± 90 barn, respectively.

Flow visualization of heavy oil in a packed bed using real-time neutron radiography

Neutron radiography can visualize heavy oil flows in metallic reactors used to upgrade heavy oils. Thus, we visualized heavy oil flows in a packed bed using real-time neutron radiography to support its capability and to offer basic experimental data necessary for verification and validation of numerical simulations. Atmospheric residue was used as the heavy oil sample and heavy oil and N2 gas were concurrently supplied into a packed bed from above a reactor simulating a trickle bed reactor. Reactor temperatures were set at either 100°C or 250°C to change heavy oil viscosity. The heavy oil flow rate was kept constant at 2.5 mL/min and the N2 gas flow rate was set at either 1 L/min or 3 L/min at 25°C. A series of neutron radiography experiments was conducted at the B-4 neutron imaging facility in the Kyoto University Research Reactor (KUR) with a thermal neutron flux of 5 × 107 n/cm2⋅s. We performed image processing for the neutron radiographs to calculate neutron beam attenuation and clarify flow behavior. Visualization results show differences in flow behavior depending on operating factors. Temperature had a particularly substantial effect on flow behavior because heavy oil viscosity depends strongly on temperature. The flows also showed different behaviors for 1 mm and 3 mm packed particle sizes because the void ratios in the packed bed, which were preliminarily observed by X-ray computational tomography (CT), change with particle size. Channeling flow was observed with 3 mm particles and a temperature of 250°C. Furthermore, this work suggests that neutron radiography can be used to investigate heavy oil flows in metallic reactors.

Development of a real-time in-core neutron monitoring system for neutron flux variations during neutron activation analysis

The pneumatic NAA facility at The University of Texas at Austin is widely used for the measurement of trace elemental concentrations. It has been shown for extremely short duration NAA irradiations, sample activation measurements can vary by up to 12% from normalized values. The typical university research reactor’s external neutron monitoring instruments are unable to detect small-localized variations in neutron population likely resulting in the observed error. The primary method to reduce this uncertainty is to irradiate traceable flux monitor samples in addition to the experimental samples to be analyzed. This method doubles the required sample prep, measurement, and analysis effort. An in-core neutron monitoring system was designed and installed adjacent to the pneumatic NAA sample system terminus to track these localized fluctuations. This design consists of a miniature in-core fission chamber fitted through the upper grid plate of the reactor core. The system is controlled by a NI LabVIEW data logging application. This system provides the capability of monitoring sample irradiance in real-time. In preliminary testing, the system was able to track short irradiation NAA sample variance to within 3% of normalized values. Thus, a significant reduction in the uncertainty of NAA measurements for trace elemental concentrations is achievable.

Boron Neutron Capture Therapy for High-Grade Skull-Base Meningioma.

Objectives Boron neutron capture therapy (BNCT) is a nuclear reaction-based tumor cell-selective particle irradiation that occurs when nonradioactive Boron-10 is irradiated with low-energy neutrons to produce high-energy α particles (10B [ n , α] 7Li). Possible complications associated with extended surgical resection render high-grade meningioma (HGM) a challenging pathology and skull-base meningiomas (SBMs) even more challenging. Lately, we have been trying to control HGMs using BNCT. This study aims to elucidate whether the recurrence and outcome of HGMs and SBMs differ based on their location. Design Retrospective review. Setting Osaka Medical College Hospital and Kyoto University Research Reactor Institute. Participants Between 2005 and 2014, 31 patients with recurrent HGM (7 SBMs) were treated with BNCT. Main Outcome Measures Overall survival and the subgroup analysis by the anatomical tumor location. Results Positron emission tomography revealed that HGMs exhibited 3.8 times higher boron accumulation than the normal brain. Although tumors displayed transient increases in size in several cases, all lesions were found to decrease during observation. Furthermore, the median survival time of patients with SBMs post-BNCT and after being diagnosed as high-grade were 24.6 and 67.5 months, respectively (vs non-SBMs: 40.4 and 47.5 months). Conclusions BNCT could be a robust and beneficial therapeutic modality for patients with high-grade SBMs.

Boron neutron capture therapy for vulvar melanoma and genital extramammary Paget\u2019s disease with curative responses

BackgroundAlthough the most commonly recommended treatment for melanoma and extramammary Paget’s disease (EMPD) of the genital region is wide surgical excision of the lesion, the procedure is highly invasive and can lead to functional and sexual problems. Alternative treatments have been used for local control when wide local excision was not feasible. Here, we describe four patients with genital malignancies who were treated with boron neutron capture therapy (BNCT).MethodsThe four patients included one patient with vulvar melanoma (VM) and three with genital EMPD. They underwent BNCT at the Kyoto University Research Reactor between 2005 and 2014 using para-boronophenylalanine as the boron delivery agent. They were irradiated with an epithermal neutron beam between the curative tumor dose and the tolerable skin/mucosal doses.ResultsAll patients showed similar tumor and normal tissue responses following BNCT and achieved complete responses within 6 months. The most severe normal tissue response was moderate skin erosion during the first 2 months, which diminished gradually thereafter. Dysuria or contact pain persisted for 2 months and resolved completely by 4 months.ConclusionsTreating VM and EMPD with BNCT resulted in complete local tumor control. Based on our clinical experience, we conclude that BNCT is a promising treatment for primary VM and EMPD of the genital region.Trial registration numbers UMIN000005124

Development of a neutron source for imaging at the electron linac facility in Kyoto University Research Reactor Institute

The Kyoto University electron linac, KURRI-LINAC, has been used for executing a project to develop methodology for the integrity evaluation of nuclear fuels, named the N-DeMAIN (Development of Non-Destructive Methods Adapted for Integrity test of Next generation nuclear fuels) project in Japan. As a non-destructive method, the neutron resonance transmission analysis was adopted for the identification and the quantification of nuclide in the fuels. Furthermore, determination of temperature distribution in the fuels based on Doppler broadening was also planned as well as traditional transmission imaging.A target-moderator-reflector system (TMRS) and a beam collimator system has been redesigned to obtain the highest intensity with the use of simulation calculation. The neutron beam line is placed at an angle of 135° with respect to the electron beam line. Optimal arrangement of the target and the moderator was investigated. After then, collimator design was performed, and a collimator system to shape the neutron and γ ray beams was designed. Test experiments were performed by using this system.Here, we explain the results of design study of the TMRS and the collimator, and the test experiments.

The effects of a university research reactor’s outreach program on students’ attitudes and knowledge about nuclear radiation

ABSTRACTBackground Many students carry negative beliefs about nuclear radiation and have concerns regarding the safety of nuclear power. However, most of these students have not had any personal or educational experiences with nuclear radiation or reactors on which to form these opinions.Purpose To examine the influence that a visit to a nuclear reactor could have upon students’, the impact of a reactor’s outreach program on students’ attitudes toward nuclear power and their scientific knowledge about nuclear radiation and reactors was explored.Program description The students in this study visited a research and teaching reactor on a university campus. Instructed by reactor staff, they participated in a series of hands-on activities that explored and explained ionizing radiation, and they toured the reactor facilities.Sample The subjects were 32 students, 13–14 years of age, from a rural area of the United States.Design and methods The students were surveyed before and after the program using open-ended and Likert scale items, and were asked to complete a nuclear science version of the Draw-A-Scientist-Test.Results The findings indicate that the students’ attitudes toward nuclear radiation and nuclear power became more positive and judicious after visiting the reactor, and there was a corresponding increase in student knowledge about nuclear reactors and nuclear radiation. Before the program, some of the students carried common scientific misconceptions about nuclear power and nuclear radiation, and the outreach was successful in challenging those misconceptions if they were directly addressed.Conclusions Increased instruction on the topics of nuclear radiation and reactors led students to have views on these topics that were more knowledgeable and discriminating. The results imply that outreach programs at reactors can be valuable in terms of educating students and developing their attitudes toward nuclear power and nuclear radiation.

Resonance analysis of 151,153Eu from neutron capture cross section measurements in the energy range from 1 to 20 eV

ABSTRACT The neutron capture cross sections of Europium-151 and Europium-153 have been measured by the time-of-flight method in the energy range from 0.005 to 100 eV using the Kyoto University Research Reactor Institute-Linear Accelerator (KURRI-LINAC). An assembly of Bismuth Germanate (BGO) scintillators was used to detect the prompt capture of γ rays. The absolute values of the neutron capture cross sections of 151Eu and 153Eu were deduced by normalizing the thermal capture cross sections in JENDL-4.0 and ENDF/B-VII.1, respectively. Then, we have obtained the resonance parameters of 20 resonances in 151Eu and 17 resonances in 153Eu using the code SAMMY. For the 3.36-eV resonance of 151Eu, the evaluated resonance peak area in JENDL-4.0 is about 95% smaller than the present result. For the 7.00-, 7.22-, and 7.42-eV resonance; we confirmed that there are significant differences between the measured peaks and evaluated peaks in JENDL-4.0, ENDF/B-VII.1, and JEFF-3.2. For the 153Eu, the evaluated resonance peak areas in JENDL-4.0, ENDF/B-VII.1, and JEFF-3.2 are about 15% larger than the measured resonance peak areas at the 2.46-, 3.29-, and 3.94-eV resonances.

Resonance Analysis for Neutron Capture Cross Sections of Th-232

In recent years, a great interest has been taken in the thorium fuel cycle. The neutron capture reaction for 232Th is very important to evaluate the 233U production in the thorium fuel cycle.In the present study, we have carried out the resonance analysis of neutron capture cross sections of 232Th obtained by using the Kyoto University Research Reactor Institute – Linear Accelerator (KURRI-LINAC). We performed the resonance analysis using the code SAMMY for the 10 resolved resonances in the energy range from 20 to 210 eV. The obtained parameters for the 69.2-, 113.1-, 120.9- and 170.4-eV resonances are in disagreement with the evaluated values, whereas the parameters for other six resonances were in general agreement.In order to verify the validity for the resonance analysis, we also investigated the neutron self-shielding and multiple-scattering effects on the neutron capture cross sections obtained for 232Th samples with three different thicknesses. The correction factors were estimated by the Monte Carlo simulation using the evaluated cross sections in JENDL-4.0 and the cross section data revised by the present resonance analysis. The comparison between the capture cross sections corrected with the JENDL-4.0 is consistent with the resonance analysis results. Moreover, it was confirmed that the revised cross section data enable the correction factors to reproduce the experimental results for the 69.2- and 120.9-eV resonances. In the case of 170.4-eV resonance, the disagreement between the experimental results was not improved in this study. Therefore, further study is necessary to improve the resonance data.

Creation of an Internal Cladding in Sapphire Optical Fiber Using the $^{6}$ Li(n, $\alpha)^{3}$ H Reaction

The focus of this work is to extend optical frequency domain reflectometry sensing to sapphire optical fiber, by creating a cladding in the fiber. A cladding was created by irradiating a sapphire optical fiber, which was surrounded by an annulus of Li-6 enriched lithium carbonate (Li2CO3) powder, in The Ohio State University Research Reactor. The 6Li(n, $\alpha)^{3}\text{H}$ reaction created high energy alpha particles and tritons that irradiated the fiber simultaneously to a depth of 24 microns along the entire periphery of the sapphire fiber, thereby slightly reducing the index of refraction in the fiber’s periphery and creating a cladding within the fiber. Transmitted light intensity profiles show that the ion implanted cladding made the fiber’s intensity profile nearly single mode. The cladding survived to the highest temperature that was tested (1500 °C). The cladding made the fiber sufficiently single mode that the cladded sapphire fiber, when read out with an optical backscatter reflectometer, produced distributed temperature measurements along the entire length of the fiber.

Neutron capture cross section measurements of 151,153Eu using a pair of C6D6 detectors

ABSTRACTWe have measured the neutron capture cross sections of 151Eu and 153Eu by the time-of-flight (TOF) method in the range from 0.005 eV to keV region using the Kyoto University Research Reactor Institute - Linear Accelerator (KURRI-LINAC). We employed a pair of C6D6 liquid scintillators for the prompt capture γ-ray measurement. The pulse-height weighting technique was employed to obtain the capture yields from the γ-ray spectra of 151,153Eu. The obtained thermal cross sections at 0.0253 eV are 9051 ± 683 b for 151Eu and 364 ± 44 b for 153Eu, respectively. The resonance integrals have been derived as 3490 ± 162 b for 151Eu and 1538 ± 106 b for 153Eu.The obtained capture cross sections were compared with the previously reported experimental data and the evaluated data. The evaluated data in JENDL-4.0 and JEFF-3.2 show good agreement with the present experiment results of 151Eu, however, the evaluated data in ENDF/B-VII.1 are larger than the present experiment results of 151Eu about 10% to 20% in the energy region from 0.03 to 0.2 eV. For the neutron capture cross sections of 153Eu, the evaluated data in ENDF/B-VII.1 and Widder's data are in good agreement with the present results in the energy region below 0.35 eV.

Selective boron delivery by intra-arterial injection of BSH-WOW emulsion in hepatic cancer model for neutron capture therapy

Boron neutron-capture therapy (BNCT) has been used to inhibit the growth of various types of cancers. In this study, we developed a 10BSH-entrapped water-in-oil-in-water (WOW) emulsion, evaluated it as a selective boron carrier for the possible application of BNCT in hepatocellular carcinoma treatment. We prepared the 10BSH-entrapped WOW emulsion using double emulsification technique and then evaluated the delivery efficacy by performing biodistribution experiment on VX-2 rabbit hepatic tumour model with comparison to iodized poppy-seed oil mix conventional emulsion. Neutron irradiation was carried out at Kyoto University Research Reactor with an average thermal neutron fluence of 5 × 1012 n cm-2. Morphological and pathological analyses were performed on Day 14 after neutron irradiation. Biodistribution results have revealed that 10B atoms delivery with WOW emulsion was superior compared with those using iodized poppy-seed oil conventional emulsion. There was no dissemination in abdomen or lung metastasis observed after neutron irradiation in the groups treated with 10BSH-entrapped WOW emulsion, whereas many tumour nodules were recognized in the liver, abdominal cavity, peritoneum and bilateral lobes of the lung in the non-injected group. Tumour growth suppression and cancer-cell-killing effect was observed from the morphological and pathological analyses of the 10BSH-entrapped WOW emulsion-injected group, indicating its feasibility to be applied as a novel intra-arterial boron carrier for BNCT. Advances in knowledge: The results of the current study have shown that entrapped 10BSH has the potential to increase the range of therapies available for hepatocellular carcinoma which is considered to be one of the most difficult tumours to cure.

DNA Double-strand Breaks Induced byFractionated Neutron Beam Irradiation for Boron Neutron Capture Therapy.

To use the 53BP1 foci assay to detect DNA double-strand breaks induced by fractionated neutron beam irradiation of normal cells. The Kyoto University Research Reactor heavy-water facility and gamma-ray irradiation system were used as experimental radiation sources. After fixation of Chinese Hamster Ovary cells with 3.6% formalin, immunofluorescence staining was performed. Number and size of foci were analyzed using ImageJ software. Fractionated neutron irradiation induced 25% fewer 53BP1 foci than single irradiation at the same dose. By contrast, gamma irradiation induced 30% fewer 53BP1 foci than single irradiation at the same dose. Fractionated neutron irradiation induced larger foci than gamma irradiation, raising the possibility that persistent unrepaired DNA damage was amplified due to the high linear energy transfer component in the neutron beam. Unrepaired cluster DNA damage was more prevalent after fractionated neutron irradiation than after gamma irradiation.

OBSIDIAN SUB-SOURCES AT THE ZARAGOZA-OYAMELES QUARRY IN PUEBLA, MEXICO: SIMILARITIES WITH ALTOTONGA AND THEIR DISTRIBUTION THROUGHOUT MESOAMERICA

We present data produced through archaeological and geological survey, as well as geochemical analysis of the Zaragoza-Oyameles obsidian source area located on the northern and western flanks of the Los Humeros Caldera in eastern Puebla, Mexico. One result of the intensive archaeological surface survey of this obsidian source area was the identification of 117 obsidian flow-band exposures. Geologic samples from 40 of these were submitted for instrumental neutron activation analysis. Eighty-five projectile points collected from the surface were characterized using portable X-ray fluorescence. These analyses identified three sub-sources: Z-O1, Potreros Caldera, and Gomez Sur. The Gomez Sur sub-source appears chemically similar to the previously identified Altotonga source, located 25 km to the northeast. Results of the geological survey help elucidate the relationship of Altotonga obsidian with the Zaragoza-Oyameles source area. The data from the three sub-sources are compared to all consumer site data attributed to the Zaragoza-Oyameles source in the Missouri University Research Reactor database. Results indicate that the majority of consumer samples throughout Mesoamerica match the Z-O1 sub-source, while 4 percent match the Potreros Caldera sub-source. This information, combined with the Gomez Sur data, is discussed in terms of economic relations with the regional center of Cantona. Obsidian procurement and distribution may have been more nuanced than previously modeled. We suggest that a number of potentially independent communities in addition to Cantona may have been involved in distributing this obsidian throughout Mesoamerica.

Influence of the neutron transport tube on neutron resonance densitometry

Neutron Resonance Densitometry (NRD) is a non-destructive assay technique of nuclear materials in particle-like debris that contains various materials. An aim of NRD is to quantify nuclear materials in a melting fuel of Fukusima Daiichi plant, spent nuclear fuel and annihilation disposal fuel etc. NRD consists of two techniques of Neutron Resonance Transmission Analysis (NRTA) and Neutron Resonance Capture Analysis (NRCA) or Prompt Gamma-ray Analysis (PGA). A density of nuclear material isotopes is decided with NRTA. The materials absorbing a neutron in a wide energy range such as boron in a sample are identified by NRCA/PGA. The information of NRCA/PGA is used in NRTA analysis to quantify nuclear material isotopes. A neutron time of flight (TOF) method is used in NRD measurements. A facility, consisting of a neutron source, a neutron flight path, and a detector is required. A short flight path and a strong neutron source are needed to downsize such a facility and put NRD into practical use. A neutron transport tube covers a flight path to prevent noises. In order to investigate the effect of neutron transport tube and pulse width of a neutron source, we carried out NRTA experiments with a 2-m short neutron transport tube constructed at Kyoto University Research Reactor Institute - Linear Accelerator (KURRI-LINAC), and impacts of shield of neutron transport tube and influence of pulse width of a neutron source were examined. A shield of the neutron transport tube reduced a background and had a good influence on the measurement. The resonance dips of 183 W at 27 eV was successfully observed with a pulse width of a neutron source less than 2 μs.

Evaluation of the optical performance of a brightness enhancement system developed for the KUR slow positron beamline

A slow positron beamline using a nuclear reactor is under development at the Kyoto University Research Reactor (KUR). A brightness enhancement system was designed to reduce the initial positron beam size (~30 mm) below typical specimen sizes (< 10 mm) and keep the intensity of the beam as high as possible. After installation of the brightness enhancement system at the beamline, we tested its optical performance using an electron beam. The experimental result indicated that the system would have enough capability to obtain a small-sized beam suitable for positron measurements.

Evaluation of a positron-beam-pulsing system in KUR reactor-based positron beam facility

The pulsing performance of the positron-beam-pulsing system of the slow positron beamline at Kyoto University research Reactor was tested using electron beams trimmed by 5-mm- or 15-mm-diameter apertures. Electron pulses with full width at half maximum of 143 ps were obtained with a 5-mm-diameter beam which corresponds to the diameter of the brightness-enhanced positron beam. This indicates the pulsing system has a sufficient pulsing performance for positron annihilation lifetime measurements.

Early clinical experience utilizing scintillator with optical fiber (SOF) detector in clinical boron neutron capture therapy: its issues and solutions.

BackgroundReal-time measurement of thermal neutrons in the tumor region is essential for proper evaluation of the absorbed dose in boron neutron capture therapy (BNCT) treatment. The gold wire activation method has been routinely used to measure the neutron flux distribution in BNCT irradiation, but a real-time measurement using gold wire is not possible. To overcome this issue, the scintillator with optical fiber (SOF) detector has been developed. The purpose of this study is to demonstrate the feasibility of the SOF detector as a real-time thermal neutron monitor in clinical BNCT treatment and also to report issues in the use of SOF detectors in clinical practice and their solutions.Material and methodsClinical measurements using the SOF detector were carried out in 16 BNCT clinical trial patients from December 2002 until end of 2006 at the Japanese Atomic Energy Agency (JAEA) and Kyoto University Research Reactor Institute (KURRI).ResultsThe SOF detector worked effectively as a real-time thermal neutron monitor. The neutron fluence obtained by the gold wire activation method was found to differ from that obtained by the SOF detector. The neutron fluence obtained by the SOF detector was in better agreement with the expected fluence than with gold wire activation. The estimation error for the SOF detector was small in comparison to the gold wire measurement. In addition, real-time monitoring suggested that the neutron flux distribution and intensity at the region of interest (ROI) may vary due to the reactor condition, patient motion and dislocation of the SOF detector.ConclusionClinical measurements using the SOF detector to measure thermal neutron flux during BNCT confirmed that SOF detectors are effective as a real-time thermal neutron monitor. To minimize the estimation error due to the displacement of the SOF probe during treatment, a loop-type SOF probe was developed.