2. Nuclear Physics and Nuclear Data

概要

INTRODUCTION: The electron linear accelerator at the Research Reactor Institute, Kyoto University (KUR- RI-LINAC) had been originally established in 1965 by the High Voltage Engineering Co., USA and started as a 23 MeV machine. In 1971, the machine power had been increased to 46 MeV. The KURRI-linac has two different operation pulse modes. One is a long mode with a maxi- mum repetition rate of 120 Hz, a pulse width of 0.1–4.0 μs and a peak current of about 0.5 A for the measurement at low energies below 10 eV. Another is a short mode with a maximum repetition rate of 300 Hz, a pulse width of 2–100 ns and a peak current of about 5 A for the measurement at high energies above 1 eV. It is worth noting that the peak current of short mode is ten times as large as that of long mode. In measurements of nuclear data, a water-cooled tantalum (Ta) target as a pho- to-neutron target and a light water moderator are used. There are two kinds of the moderator. One is a water tank type and another is an octagonal shape moderator called “pac-man type”. In order to measure accurate nuclear data, it is very important to evaluate the energy resolution (ΔE/E) of a moderator. For example, the energy resolu- tion for pac-man type moderator had been calculated about 0.7 % between energy range of 0.1eV from 10keV [1]. However, measurement and detail evaluation of en- ergy resolutions for the tank type moderator has not car- ried out in KURRI-LINAC. Furthermore, the relationship between energy resolution and beam pulse width is not discussed in Reference [1].
In this study, the energy resolutions of neutron flux from a tank type moderator were obtained using the TOF tech- nic in 4 operational mode with different pulse width.

INTRODUCTION: In terms of nuclear transmutation stud- ies with neutrons of minor actinides contained in nuclear wastes, accurate neutron capture cross sections are required to obtain transmutation rates. The present work selected 237Np nuclide among minor actinides because 237Np causes long-term radio- toxicity in nuclear waste management due to a long half-life of
2.14×106 years.[1] When examining the situation of ther- mal-neutron capture cross-section (σ0) of 237Np, there have been discrepancies in reported data even in recent year as shown in Figure 1. Previous study [2] has demonstrated that the σ0 can be measured using the graphite thermal column equipment (TC-Pn) installed in KUR. Consequently, the present work
aimed to measure the σ0 of 237Np by an activation method using a well-thermalized neutron field in the TC-Pn.

INTRODUCTION: The decay data of the fission products are important for the decay heat evaluation and also structure of neutron-rich nuclei. Many neutron-rich nuclei with mass numbers near 150 do not have detailed decay schemes because of their short half-lives and low fission yields. The nucleus of 155Pr was proposed to have a half-life of 1.49 s by Wu et al. [1] and a level scheme of the daughter nucleus 155Nd was reported by Hwang et al. using spontaneous fission of 252Cf [2], but, no decay γ-rays have been reported. Based on the experimental results of 153,154Pr in the last year and the expected yields of an on-line mass separator KUR-ISOL, the observation of the γ-rays in the β decay of 155Pr are expected. In this year, to identify γ rays in the β decay of 155Pr, β-γ coincidence measurements were carried out using a high-efficiency clover detector coupled with β-ray detectors. The clover detector has four large Ge crystals with a diameter of 80 mm and a length of 90 mm arranged in the shape of a four-leaf clover around a through hole with a diameter of 15 mm. The β-ray detectors to set in the through hole of the clover detector developed by Ishikawa [3] were modified.

INTRODUCTION: In the present study, we have de- veloped a current-mode neutron detector [1] that can be used in intense neutron fields such as boron neutron cap- ture therapy (BNCT) fields. It is necessary to measure the neutron flux for a large dynamic range because of connec- tion between a BNCT field in a hospital and a neutron cal- ibration field in the National Institute of Advanced Indus- trial Science and Technology (AIST). The difference of neutron fluxes between the BNCT field and the calibration field is more than 5 orders of magnitude. We have devel- oped a new 3He gas detector (proportional counter or ion- ization chamber) with the current mode. The current mode gas detector is expected to be high radiation re- sistance in comparison with a photo-multiplier tube in the scintillation detector. However, subtraction of gamma-ray component is a problem that need to be overcome in the current-mode neutron detector. In the present study, we ex- perimentally verify the gamma-ray subtraction method in the current-mode 3He gas detector using 3He proportional counters with 4 different gas pressures.

NTRODUCTION: In a reactor core, γ rays are radiated by some kinds of neutron induced reactions such as the fission, capture, inelastic scattering, etc. Accordingly, γ ray spectroscopy is potentially useful to identify and quantify those reac- tions. Focusing on that, one of the authors has studied γ ray spectroscopy for Kyoto University Critical Assembly (KUCA) [1]. In KUCA, the uranium fuel of 93 wt% -235U enrichment is scheduled to be alternated to the fuel with lower enrichment (LEU). By the reduction of the enrich- ment, the reactions of 238U would be more significant in the core. In order to quantify the capture reaction rates of
238U, we need neutron induced γ ray emission data. For the reason, prompt γ ray spectrum measurements have
been conducted for uranium metallic samples irradiated by neutrons of white spectrum at the LINAC facility [1,2].

INTRODUCTION: The Japan Atomic Energy Agency (JAEA) is conducting research and development of active neutron NDA techniques [1] for nuclear nonproliferation and security. As a part of these projects, technology de- velopments for a compact neutron resonance transmis- sion analysis (NRTA) [2] system using a laser-driven neutron source (LDNS) was performed [3]. Using an ex- tremely short pulse laser (~ps), an LDNS can provide short-pulsed fast neutrons (~ns) [4]. This is potentially useful to perform high resolution NRTA measurements (neutron time-of-flight (TOF) measurements) with a short flight path.
In experiments of NRTA, one of the origins of back- ground is the emission of 2.2-MeV gamma rays of 1H(n, γ)2H reaction in a moderator for pulse neutron generation. The 2.2-MeV gamma-ray emission decreases exponen- tially just after a neutron pulse generation with a decay time of few hundreds μs. In a measurement using short flight path, therefore, the background overlaps neutron resonance dips of a spectrum, resulting in decline of sen- sitivity of the NRTA system. To overcome the problem, we developed a multi-layer 6Li glass scintillation (MLS) detector [5] that has low sensitivity to high energy gam- ma rays.
In the previous experiment [5], we compared the per- formance of an MLS detector (total 6Li thickness: 0.5 cm) and a conventional 6Li glass scintillation detector (6Li thickness: 1.0 cm). Influence of the 2.2-MeV gamma ray in TOF spectra (<100 μs) were examined. From the study, we concluded that the MLS detector could reduce the sensitivity to 2.2 MeV gamma rays.
Based on the previous results [5], we manufactured a new and upgraded MLS detector. The total thickness of 6Li glass was increased to 1.5 cm to achieve better neu- tron detection efficiency. Another photomultiplier tube (PMT) was introduced to eliminate noise signals by using the coincidence technique. The performance of the new MLS detector was tested at the Kyoto University Institute for Integrated Radiation and Nuclear Science – Linear Accelerator (KURNS-LINAC). Neutron transmission spectra were measured varying the thicknesses of reso- nance samples.

INTRODUCTION: The thorium based nuclear fuel cycle has many ad- vantages with respect to the reduction of the risk of pro- liferation of fissile material and the reduction of buildup of long-lived higher actinides. In the previous works, we have already finished the neutron capture cross section measurement of 232Th [1, 2]. Uranium-233 is the main fuel in the thorium fuel cycle, and its fission reaction plays an important role in the reactor system. Therefore, it is necessary to improve the accuracy of nuclear data associated with the neutron induced fission cross sections of 233U for the feasibility study of the thorium nuclear system. We carried out the neutron induced fission cross-section measurement of 233U.

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