Characteristics of Neutron Guide Tubes in Upgraded JRR-3

Abstract

The JRR-3 has been upgraded to be a new high performance research reactor JRR-3M with neutron guide tubes on a large scale and a cold neutron source. The neutron fluxes and spectra were measured at the end of the two thermal and three cold neutron guide tubes. The gain of the cold neutron source is also found from these spectra. The neutron fluxes of thermal neutron guide tubes with characteristic wavelength 2 Å are 1.2x10⁸ n/cm².s at a reactor power of 20 MW. The neutron fluxes of cold guide tubes are 2.0x 10⁸ n/cm².s with characteristic wavelength 4 Å and 1.4x10⁸ n/cm².s with 6 A when the cold neutron source is operated. The neutron spectra measured by the time-of-flight method agree well with their designed ones. The gains of the cold neutron source are 8 for 4 Å and 20 for 6 Å at a reactor power of 20 MW.     

Chemical Dosimeters for Mixed Radiations in a Nuclear Reactor

The dose rates due to mixed reactor radiations were measured by five gaseous chemical dosimeters – nitrous oxide (natural), ¹⁵N-enriched nitrous oxide, ethylene, ethane and carbon dioxide. The observed dose rates for these gases at the same irradiation position in a nuclear reactor were, 1.8×10⁸, 1.5×l0⁸, 1.9×10⁸, 2.5×l0⁸ and 1.0×10⁸ rad/hr, respectively. These values were compared with those calculated from the mass stopping power of the gases for secondary electrons produced by γ-rays and those from thermal and fast neutron fluxes. No contradiction was found among them. A method of analysis of the reactor radiation dose rates into γ, thermal and fast neutron components is proposed, which is based solely on chemical dosimetry.     

Positron annihilation lifetime study of JRQ and JPA irradiated and annealed reactor pressure vessel steels

The model reactor pressure vessel steels known as JRQ and JPA were manufactured in Japan for the IAEA neutron embrittlement research studies. These model alloys belong to the commercially used steel A533B-1 type and show relatively large changes in mechanical properties after relevant neutron irradiation. The neutron irradiation was performed by different neutron fluxes as well as different neutron fluences (up to about 150 × 10¹⁸ cm⁻² (E > 0.5 MeV)). For a better understanding of the neutron embrittlement, the Positron Annihilation Lifetime Spectroscopy (PALS) technique was applied in 2014. PALS measurement of irradiated specimens was performed using three detectors set-up due to induced ⁶⁰Co radioactivity of the studied specimens. We confirmed that the JPA steel, considered to be high-copper steel, is much more sensitive to defect creation due to neutron irradiation than the low-copper JRQ steel.     

Quantitative Analysis of Boron with Solid State Track Detector

In a study aimed at the application of solid state track detectors to the quantitative analysis of boron, track detector plates of cellulose nitrate were set in polypropylene test tubes, filled with various concentrations of boron-agar solution, and irradiated with thermal neutrons. After irradiation, the cellulose track detectors were separated from the solution, and etched with suitable chemical reagents. The number of etch-pits observed on the plate surface was counted by means of an optical microscope. The experiments proved that the number of etch-pits per unit area P (cm^?2) is proportional to the product of the ¹⁰B-concentration τ (atom·cm^?3) and the thermal neutron fluence φ (cm^?2): P=bτφ, where b is the ¹⁰B-detection efficiency. If p is the number of ¹⁰B(n, a)⁷Li reactions per cubic centimeter, P=pρ, where p is the etch-pit formation efficiency. The maximum values of b and p obtained in these experiments were 1.3 × 10^?24 cm³, and 3.3 × 10^?4 cm, respectively. The mean effective range of α and ⁷Li particles in the solution was found to equal 2p.     

Construction of Neutron Guide Tubes in Upgraded JRR-3

Five neutron guide tubes have been installed in the upgraded JRR-3 (Japan Research Reactor No. 3). Two of them are for thermal neutrons and the other three are for cold ones. The characteristic wavelength of the thermal neutron guide tubes is 2 Å, and those of the cold neutron guide tubes are 4 and 6 Å. The longest guide tube is 59.9 m long and the total length of guide tubes is 232.1 m.

The beam sizes are 2 cm × 20 cm for the thermal neutron beams and 2 cm × 12 cm for the cold neutron beams. A curved part of the neutron guide is assembled by a polygonal approximation with use of 85 cm long straight units. The neutron mirrors of these units are made of natural Ni deposited borosilicate glasses. The Ni layer is about 2,000 Å in thickness.

The mean fabrication error of guide tube units is 4 μm. The mean installation errors are 8 μm for the positional abutment error and 5 × 10^?6 rad for the angular error. The neutron losses by these errors will be about 5%, and the neutron fluxes at the exits of the neutron guides are estimated to be about 2 × 10⁸ n/cm²·s.     

Thick Lithium Metal Target for Fast Neutron Production

A high flux fast neutron was produced by the (d, n) reaction of a lithium metal target. A thick lithium layer for the target was prepared by a simple method of melt-coating on a copper plate. The fast neutron (>9 MeV) flux at a distance of 6mm from the target was (9.0±1.6)×l0⁶ n/cm²·sec·μA with use of a 2.0 MeV deuteron beam. A flux of 2.7×10⁹ n/cm²·sec was obtained by bombarding the target with the deuteron beam of 300 μA.     

Studies of Ceramic Fuels with the Use of Fission Gas Release Loop, (I)

Fission gas release from a UO²-graphite mixture was studied during irradiation with the use of the Fission Gas Release Loop in the JRR-3 reactor. The release rates of fission krypton and xenon increased proportionally with neutron flux (6×10¹⁰–6×10¹² n/cm²·sec) and exponentially with temperature (400°–1,000°C). A burst of fission gas was observed when the specimen was abruptly heated to a higher temperature. These results can be explained by a mechanism whereby fission gas is trapped in defects created in graphite by fission fragments and released through annealing of the defects.     

Thermal Neutron Monitoring Using Cellulose Nitrate Track Detector by Means of Spark Counter

Cellulose nitrate films were examined for thermal neutron monitoring in contact with several kinds of (n, α) converter materials such as a plastic sheet doped with 1% boron, a solidified boron oxide plate, a single crystal plate of boron oxide and that of lithium fluoride. After thermal neutron irradiation in a research reactor, the films were etched in alkaline solution. The etch-pits on the films were counted with a spark counter. The single crystal converter of lithium fluoride was found to be the most sensitive for thermal neutron. A method for producing the single crystal converter was described. Linear relation between thermal neutron fluence and spark counts was found to be in the fluence range of 2 × 10^?6–3 × 10⁷n/cm² in case of using a lithium fluoride single crystal converter.     

Optimization of moderator assembly for neutron flux measurement: experimental and theoretical approaches

A moderator of paraffin wax assembly has been demonstrated where its thickness can be optimized to thermalize fast neutrons. The assembly is used for measuring fast neutron flux of a neutron probe at different neutron energies, using BF03(U10and 200) and3He(U0.500)neutron detectors. The paraffin wax thickness was optimized at 6 cm for the neutron probe which contains an Am–Be neutron source. The experimental data are compared with Monte Carlo simulation results using MCNP5 version 1.4. Neutron flux comparison and neutron activation techniques are used for measuring neutron flux of the neutron probe to validate the optimum paraffin moderator thickness in the assembly. The neutron fluxes are measured at(1.17 ± 0.09) 9 105 and(1.19 ± 0.1) 9 105n/s, being in agreement with the simulated values. The moderator assembly can easily be utilized for essential requirements of neutron flux measurements.     

Resonance Parameters of Tantalum-181 in Neutron Energy Range from 100 to 4,300eV

Neutron transmission measurements were performed on natural tantalum (abundance ratio 99.988% for ¹⁸¹Ta) in the energy range of 100–4,300 eV using the Japan Atomic Energy Research Institute linac. The transmissions were measured using 55 and 190 m time-of-flight spectrometers for two and three samples of different thicknesses, respectively. These transmission data were simultaneously analyzed with a least squares fitting program based on a multl-level Breit-Wigner formula, and resonance energies and neutron width were obtained for 696 resonances of ¹⁸¹Ta.

The statistical analysis of these parameters gave the s-wave average level spacing of D=4.10±0.14 eV and s-wave neutron strength functions of (1.67±0.13) × 10^?4, (1.09 ± 0.09) × 10^?4 and (1.42 ± 0.20) × 10^?4 for the energy intervals from 100 ? 1,700 eV, 1,700–3,400 eV and 3,400–4,300 eV, respectively. This significant difference among the neutron strength function for each energy interval is a prominent result of the present experiments and is of great interest.     

Measurement and Calculation of Neutron Diffusion Parameters in Water

Neutron diffusion parameters in water at a room temperature of 10°C have been measured by the pulsed neutron method for the range of geometrical buckling from 0.093 to 1.36 cm^-2. The results are 205±4 μ60 for the neutron mean life time due to absorption, 34,120±610 cm²·sec^-1 for the diffusion coefficient and 3,350±560 cm⁴·sec^-1 for the diffusion cooling coefficient.

The decay constant has been calculated as a function of buckling for the Nelkin and the Rad-kowsky scattering models of water on the assumption of linear anisotropic scattering. The calculated diffusion coefficients, 36,290 cm²·sec^-1 for the Nelkin model and 37.610 cm²·sec^-1 for the Radkowsky model, are somewhat higher than the experimental result.

It is shown that the calculated diffusion coefficient approaches the experimental value if we use μ-(E), the mean value of cosine of scattering angle, obtained from the Beyster's experiment instead of that for the Nelkin model.     

Thermal conductivity of SiC after heavy ions irradiation

In this study, we performed irradiation experiments on α-SiC samples, with heavy ions at room temperature (74 MeV Kr, fluence of 5 × 10¹⁴ ions cm^?2). This energy results in an irradiated layer of about 9.6 μm for SiC. TEM and Raman analyses reveal a graded damaged material. In the electronic interactions domain SiC is weakly damaged whereas it becomes fully amorphous in the nuclear interactions domain. According to the structural examinations, the irradiated SiC is considered as a multilayered material. Thermal conductivity in both electronic and nuclear interactions domains is measured as a function of temperature and annealing temperature. It appears that such an approach is reliable to estimate thermal conductivity of ceramics under neutron irradiation.     

Neutron Resonance Parameters of Silver-107 and Silver-109

Neutron transmission measurements were carried out on the separated isotopes of silver using the time-of-flight facility at the Japan Atomic Energy Research Institute electron linear accelerator. Neutrons were detected with the ⁶Li-glass detectors at 56 and 191 m. The samples used were metallic powder enriched to 98.2% for ¹⁰⁷Ag and 99.3% for ¹⁰⁹Ag. Transmission data were analyzed with the multi-level Breit-Wigner formula incorporated in a least squares fitting program. Resonance energies and neutron widths were determined for the large number of resolved resonances in the neutron energy region of 400 eV~7 keV. The s-wave strength functions and average level spacings were obtained to be; S₀= (0.43±0.05) × 10^?4, D₀ = 20±2 eV for ¹⁰⁷Ag and S₀= (0.45 ± 0.05) × 10^?4, D₀ = 20 ± 2eV for ¹⁰⁹Ag.     

Effect of metal electrode on characteristics of gamma-irradiated silicon carbide detector

Silicon carbide (SiC) is a highly promising semiconductor neutron-detector material for harsh environments such as nuclear reactor cores and spent-fuel storage pools. In the present study, three 4H–SiC p–i–n diode detectors were fabricated as variations of those metal-electrode structures. The I–V characteristics and alpha-particle responses of the detectors were measured before and after gamma-ray exposure. The detector with a Ti/Au electrode showed the lowest change of leakage current after irradiation; none of the detectors showed any change in the charge-collection efficiency when a sufficient electric field was applied after gamma irradiation of up to 8.1 MGy.     

Use of concretes for high-temperature shielding of nuclear reactors

The authors have studied the possibility of using chromite and chomotte heat-resistant concretes for the thermal shields of reactors. They observe neutron fluxes of various intensities (up to 10¹³ neutrons/cm²·sec, with spectrum similar to fission spectrum), absorbed by shields of these materials. They compute the transmission of neutrons and of fluxes of gamma quanta and the heat emission in the shielding. They calculate the temperatures in the shielding for various neutron fluxes, concrete thicknesses and cooling conditions. They perform a statistical calculation of the temperature stresses for shielding constructed of heat-resistant ferroconcrete.It was established that nuclear reactor shields can be made from heat-resistant ferroconcrete when the neutron fluxes on the concrete are up to 10¹³ neutrons/cm²·sec, for temperatures up to 1000–1100° C and temperature differences of up to 900° C.Translated from Atomnaya Énergiya, Vol. 19, No. 6, pp. 524–529, December, 1965Report read by G. I. Budker at the International Conference on High-Energy Accelerators (Frascati, Italy).     

Production of 11C and 7 Be by TRIGA-II Research Reactor

¹¹C has been produced by irradiating for 30 min in the pneumatic tube of a TRIGA-III research reactor about 700 mg of metaboric acid (HBO₂), using the ^11B (p,n) ¹¹C reaction initiated by neutron-hydrogen recoil. The fast neutron flux (>lMeV) was 9×10¹¹n/cm²·sec. The ¹¹C produced was separated from the irradiated target by distillation of CO₂. The yield of ¹¹C at the end of irradiation was 0.27 μ.Ci.

⁷Be also was produced by irradiating for 15 hr 1 g of lithium hydroxide in the central experimental tube of the reactor, through the ⁷Li(p,n) ⁷Be reaction initiated by neutron-hydrogen recoil. The fast neutron flux was 2.2×10¹² n/cm²·sec. Separation of carrier-free ⁷Be from the irradiated target was obtained by co-precipitation with ferric hydroxide. The yield of 7Be was 0.2 μCi.

For determining the influence on the ⁷Be yield brought by differences in the chemical composition of target, irradiations were carried out on lithium compounds containing hydrogen, such as lithium hydroxide, lithium hydroxide hydrate, lithium acetate and lithium citrate. Among these four lithium compounds, lithium hydroxide provided the highest yield of ⁷Be per gram of target.     

Supermirror neutron guide system for neutron resonance spin echo spectrometers at a pulsed neutron source

A neutron guide system for neutron resonance spin echo spectrometers has been constructed at BL06 of the Japan Proton Accelerator Research Complex, Materials and Life Science Experimental Facility. The spectrometers consist of two types of neutron spin echo instruments, a modulated intensity by zero effort instrument (MIEZE) and a neutron resonance spin echo instrument (NRSE), to cover a wide energy range for various sample environments. A neutron beam from the moderator is deflected by supermirror neutron guides, split, and separately guided into the MIEZE and NRSE. The characteristic wavelengths of the neutron guide tube for the MIEZE and NRSE are 2.9 and 4.9 Å, respectively. The cross sections of the exit of the MIEZE and NRSE guides are 15 mm × 50 mm and 30 mm × 50 mm, respectively. The neutronics and shielding design were optimized by using the heavy ion transport code system (PHITS), and the absolute average neutron fluxes at the exits of the MIEZE and NRSE guides are estimated to be 2.7 × 10⁸ and 6.9 × 10⁸ n/cm²/s/MW, respectively. The measured fluxes of the MIEZE and NRSE neutron guides are 0.56 and 0.95 times the calculated values, respectively.     

实验测量CFBR-Ⅱ堆启动Am-Be中子源的初始增殖倍数

由于CFBR-Ⅱ堆原启动中子源²⁵²Cf源半衰期较短,中子发射率已降低至无法满足使用要求,因此采用半衰期较长的Am-Be中子源替代。通过分别测量²⁵²Cf源与Am-Be源在裸源情形下的计数率以及处于活性区中心时引起的泄漏中子计数率,建立比例关系,借助于²⁵²Cf源对应的初始增殖倍数间接给出了Am-Be源对应的初始增殖倍数,为反应堆运行提供参数。     

Application of a 6LiF Small Neutron Detector with an Optical Fiber to Tritium Production Rate Measurement in D-T Neutron Fields

⁶LiF small neutron detectors with an optical fiber have been used to measure ⁶Li(n,α)T reaction rate distributions at thermal research reactors and accelerator facilities. In the present study, we developed an experimental method for the measurement of tritium production rate (TPR) of ⁶Li using this small detector in deuterium-tritium (D-T) neutron fields. Reaction rate measurements with the detector were conducted in the D-T neutron fields at the Fusion Neutronics Source (FNS) facility. From the results, we determined that this detector can be used to measure the TPR distribution in soft neutron spectrum fields such as in a Be assembly. It is difficult to obtain ⁶Li(n,α)T reaction rate separately in hard neutron spectrum fields such as in a Li₂O assembly, because many kinds of charged particle production reactions need to be taken into consideration. However, a time-dependent reaction rate measurement method combined with the ⁶LiF detector and the ZnS detector is effective to separate the ⁶Li(n,α)T reaction from other reactions even in a hard spectrum field, and it can be applied to the measurement of the TPR distribution accurately.