Double-differential neutron emission cross sections (DDXs) of ^{6}Li, ^{7}Li and ^{9}Be were measured for 18.0MeV and 11.5MeV incident neutrons produced by the T(d, n) and ^{15}N(d, n) reactions respectively, using the Tohoku University Dynamitron time-of-flight (TOF) spectrometer. The data were obtained at 13 laboratory angles, and angular-differential cross sections (ADXs) of elastic and inelastic scattering neutrons were derived from the DDXs. For 11.5MeV neutrons, we obtained the neutron emission spectra over the secondary neutron energies by newly employing the double TOF method as well as the conventional one. In the measurements at 18.0MeV, we achieved better energy resolution than in our previous studies by using a neutron detector that has a larger solid angle and a thinner tritium target. The experimental results of DDXs and ADXs were compared with our previous results and the evaluated data given in JENDL-3.2, JENDL Fusion File and ENDF/B-VI. It is found that the JENDL data reproduce the experimental ones very well.
Double differential (n, α) reaction cross sections of ^{58}Ni and ^{nat}Ni were measured for 4.2-6.5MeV neutrons with energy resolution good enough to separate α-particles from the low-lying levels of residual nuclei by using a gridded ionization chamber. Angular distribution and excitation functions were derived for α_{0}, α_{1} and α_{i ?? 2} components (α-particles to the ground level, the lst level and levels higher than the 2nd level, respectively). The experimental results were compared with these obtained from calculation based on Hauser-Feshbach model employing the optical potential and the level density parameters derived to reproduce the experimental values of total, (n, p) and (n, α) cross sections. The calculation showed fair agreement with the experimental data while it underestimated the (n, α) cross section above 6MeV.
We have developed a radiation monitor by using plastic scintillation fibers (PSF), which we call the fiber optic radiation monitor. For wide-area radiation monitoring, the detection part needs to be longer, therefore we have produced a prototype monitor by combining the PSF and the silica fiber for optical propagation. In this paper we first explain the characteristics when silica fiber of 100m long is connected to the PSF. We found that in this case, the position resolution would drop. In order to determine the cause, we observed the aspects of optical pulse propagation. It was found that the light is transmitted through the silica fiber and is detected at fiber end as single photons, and the resolution deteriorated due to the difference in the propagation path of each of photons. We next explain the prototype fiber optic radiation monitor with a detection length of 20m and a total length of 60m, and its performance.
An approximating formula recently proposed by the authors for γ-ray buildup factors of multilayered shields was applied for very thick shields up to 40mfp. For this purpose, modifications were made to the model and the fitting method to improve the data reproducibility. The previous model was expanded so that it included both the plane-normal and point isotropic geometries. The verification test of the modified model was made for three materials; water, iron and lead. The separately published data of double-layered shields for point isotropic buildup factors calculated by EGS4 from 0.1MeV to 10MeV were used as well as newly calculated data at 1MeV for the plane-normal geometry, and data for the point isotropic geometry of triple-layered shields at 1 and 10MeV. The present formula generally shows a very good reproducibility of the multilayer buildup factors, even in case of very thick shielding problems. The observed error between the approximating description and the EGS4 data is 15% in the intermediate energy range, about 30% in the higher energy range, and 35% at 0.3MeV. However, the error in the approximation reaches a factor of 4 in the worst case at 0.1MeV.
A new transport theory code for two-dimensional calculations of both square and hexagonal fuel lattices by the method of characteristics has been developed. The ray tracing procedure is based on the macroband method, which permits more accurate spatial integration in comparison to the equidistant method of tracing. The neutron source within each region is approximated by a linear function and linearly anisotropic scattering can be optionally accounted for. Efficient new techniques for both azimuthal and polar integration are presented. The spatial discretization problem in case of P_{1}-scattering has been studied. Detailed analyses show that the P_{1}-scattering in case of regular infinite array of fuel cells is significant, especially for MOX fuel, while the transport correction is inadequate in case of real geometry multi-group calculations. Finally, the complicated nature of the angular flux in MOX and UO_{2} fuel cells is demonstrated.
A thin β-ray detector for surface contamination monitoring has been newly developed. This detector, which we call a "wavelength-shifting" β-ray detector, consists of a thin square plastic scintillator plus wavelength-shifting fibers (WLSFs) to collect scintillation photons. The WLSFs are attached to opposite edges of the scintillator, where they convert highly concentrated scintillation photons into fluorescence. The fluorescent photons can either be detected directly at the ends of the WLSFs or through transparent optical fibers by small photo-multiplier tubes (PMTs). This design provides a thin detector profile while offering good β-ray detection performance. We first carry out a theoretical evaluation of the light output of this design. Then we describe verification tests aimed at device evaluation, including some performance tests using trial-manufactured detectors. The results demonstrate that the new design offers good sensitivity and uniformity, and that the required detection limit for surface contamination monitoring is achieved. Finally, we confirm that the newly developed detectors are suitable for practical application.
Several experimental results show that bubbles can easily be captured in the wake formed by leading bubbles when multiple bubbles are rising in a liquid. It is suggested from this experimental result that the effect of bubble wake should be included in the constitutive relationships representing the interfacial drag force. In the present study, steam-water bubbly flow experiments were performed to develop a new interfacial drag force model including the effect of bubble wake. Since the validity of the existing constitutive equations have been tested mainly for two-phase flow in small-diameter pipes, our study focused on two-phase flow in a large-diameter pipe. Using a one-dimensional two-fluid model, the applicability of the new interfacial drag force model to our experimental conditions was investigated. As a result, it was shown that the present model markedly improves the accuracy of the predicted results. It was therefore demonstrated that the present bubble wake model is effective at least for the conditions which were used for model development. Its applicability to different conditions will be discussed in a subsequent study.
The effectiveness of an operator-initiated steam generator (SG) secondary-side depressurization on the core cooling performance during small-break loss of coolant accidents (SBLOCAs) in a pressurized water reactor (PWR) with total failure of the high pressure injection (HPI) systems is studied. The study is based on experiments conducted in the ROSA-V Large Scale Test Facility (LSTF) and analyses with the RELAP5/Mod3 code. The sensitivity of the core minimum liquid level and peak cladding temperature (PCT) to the secondary-side depressurization rate and the initiation time of the depressurization is evaluated analytically for various break sizes. It is shown that the PCT takes a maximum value for break areas between 1.0% and 1.5% of the cold leg cross-sectional area. The conditions which the depressurization rate and the initiation time should satisfy to limit the maximum PCT are derived.
Transient behaviors of plasma and in-vessel components have been investigated considering the divertor plasma state (detached/attached) transition. The SAFALY code consisting of a zero-dimensional plasma model and a one-dimensional heat transfer model of components has been modified to take account of the divertor plasma state transition on the basis of the updated divertor plasma physics. Several plasma events, i.e., over fueling, sudden auxiliary heating injection and confinement improvement events which would be expected to result in overpower, were selected for the International Thermonuclear Experimental Reactor (ITER) and the transient behaviors were calculated on the assumption of a combined failure of plasma control and machine interlock in addition with a postulated plasma transient. The results show that plasma burning passively terminates due to sublimated impurity penetration from the carbon target surface, but there are possibilities of dry out of the coolant for the high heat flux in sudden attached state transition under the multifailure of plasma control. However, effects by the aggravating failure of the divertor are expected to be safely terminated by the confinement boundary, the vacuum vessel and its pressure suppression system.
Conversion coefficients from neutron and proton fluences to effective dose are calculated in the range of incident neutron energy from 20MeV to 10GeV. Two different versions of effective dose are treated, respectively calculated using: (a) the radiation weighting factor w_{R}, and (b) the Q-L relationship given in ICRP 60. Monte Carlo calculations are performed applying the HETC-3STEP and the MORSE-CG/KFA in the HERMES code system. The calculations are based on a modified MIRD5 anthropomorphic phantom, which is irradiated in anterior-posterior and posterior-anterior directions of beam. For effective dose calculation using the Q-L relationship, a database was compiled and added to the HETC-3STEP, to derive the average quality factor. The effective dose derived using wR proved to overestimate that obtained with the Q-L relationship by about 80% at 10GeV incident neutron energy in the case of conversion from neutron fluence. For proton fluence, the corresponding overestimation reaches a maximum factor of 4.
In research reactors, plate-type fuel elements are generally adopted so as to produce high power densities and are cooled by a downward flow. A core flow reversal from a steady-state forced downward flow to an upward flow due to natural convection should occur during operational transients such as "Loss of the primary coolant flow". Therefore, in the thermal hydraulic design of research reactors, critical heat flux (CHF) under a counter-current flow limitation (CCFL) or a flooding condition are important to determine safety margins of fuel against CHF during a core flow reversal. The authors have proposed a CHF correlation scheme for the thermal hydraulic design of research reactors, based on CHF experiments for both upward and downward flows including CCFL condition. When the CHF correlation scheme was proposed, a subcooling effect for CHF correlation under CCFL condition had not been considered because of a conservative evaluation and a lack of enough CHF data to determine the subcooling effect on CHF. A too conservative evaluation is not appropriate for the design of research reactors because of construction costs etc. Also, conservativeness of the design must be determined precisely. In this study, therefore, the subcooling effect on CHF under the CCFL conditions in vertical rectangular channels heated from both sides were investigated quantitatively based on CHF experimental results obtained under uniform and non-uniform heat flux conditions. As a result, it was made clear that CHF in this region increase linearly with an increase of the channel inlet subcooling and a new CHF correlation including the effect of channel inlet subcooling was proposed. The new correlation could be adopted under the conditions of the atmospheric pressure, the inlet subcooling less than 78K, the channel gap size between 2.25 to 5.0mm, the axial peaking factor between 1.0 to 1.6 and L/De between 71 to 174 which were the ranges investigated in this study.
Inspection of pipe inner wall is the most important problem for nuclear power plant. At present there are no non-contact holding technique for the inspection system. For that purpose, we are developing a non-contact inspection system. This paper describes a "first" prototype model of the holder using air pressure, and results of preliminary experiment on air pressure distribution, holding attitude and static holding stiffness. In prototype mechanisms, when the supplied air pressure was 1.2MPa the angle of inclination was within 1', and the holding stiffness was 2N/μm(air film thickness 100μm).
Responses of a whole-body counter to distribution of ingested 137Cs within the body were evaluated using Monte Carlo simulation. The uncertainties in evaluation of 137Cs body burdens due to counting efficiencies of the whole-body counter were estimated. It was found from calculation that counting efficiencies of the whole-body counter are largely dependent on the 137Cs distribution within the body, and 137Cs body burdens would be underestimated by a factor of 3 in the worst case. To testify a calibration method for the whole-body counter using Monte Carlo simulation, counting efficiencies for simple-geometric-form models and phantoms were obtained by simulation and actual measurements. The calculations by Monte Carlo simulation are in good agreement with the measurements.