Neutron transmission measurements were performed on natural tantalum (abundance ratio 99.988% for ¹⁸¹Ta) in the energy range of 1004, 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 multi-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, 7003, 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.

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To develop an analytical method for a DC electromagnetic flow coupler, an electric circuit analog has been considered. The adopted analog consists of an infinite series of the lumped parameter circuit of the flow coupler cross section and allows an analysis of end effects which appear outside the inlet and outlet of the flow ducts. The ordinary differential equation of the second order was derived for the current distributions from the circuit equations of the present analog. The present method predicted very well the current distributions of the analytical solutions obtained by Birtzvalk et al. for the coupler with infinite width and length.
Further, the overall efficiency of the typical flow coupler evaluated by the present method gave a similar dependence on the flow ratio to that of Hughes & Alexion, although differences appeared in the results of the pressure rise and drop in the coupler's two ducts. From the current distributions obtained, these differences were attributed to the electrical coupling of the pump and the generator ducts in the fringing magnetic field which was considered by the present method, but not by Hughes & Alexion. Finally, the current flow patterns are given with an improved distribution of the external magnetic field including the wall effect.

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A unified definition for cell-averaged diffusion constants has been established on the basis of transport and/or diffusion perturbation theories. The diffusion constants are derived by equating each component of reactivities obtained by transport theory for a heterogeneous cell to a corresponding component of reactivities by diffusion theory for a homogenized cell. The derived diffusion constants are applied to infinite uniform lattices and heterogeneous lattices composed of different cells. In infinite lattices the present diffusion constants are compared to the conventional flux-weighted cross sections and Benoist's anisotropic diffusion coefficients. In heterogeneous lattices they are compared with Rowlands-Eaton cross sections and the extended Benoist's diffusion coefficient. The present formula is applied to a heterogeneous slab lattice. It reveals that the present method gives a more accurate result for Κen compared to the case with the conventional diffusion coefficient defined by D = 1/3Σ_ιγ or the extended Benoist's diffusion coefficient together with the conventional flux-weighted cross sections.

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Under appropriate conditions, stochastic processes are described by the ARMA model, however, the AR model is popularly used in reactor noise analysis. Hence, the properties of AR model as an approximate representation of the ARMA model should be made clear. Here, convergence of AR-parameters and PSD of AR model were studied through numerical analysis on specific examples such as the neutron noise in subcritical reactors, and it was found that :
(1) The convergence of AR-parameters and AR model PSD is governed by the "zero nearest to the unit circle in the complex plane" (μ^-1, |μ|<l) of the ARMA model transfer function.
(2) The AR-parameters of AR(M) model have biases from those for the infinite model order, and these biases decrease approximately in proportion to |μ|^M.
(3) The AR model of the neutron noise of subcritical reactors needs a large model order because of an ARMA-zero very close to unity corresponding to the decay constant of the 6-th group of delayed neutron precursors.
(4) In applying AR model for system identification, much attention has to be paid to a priori unknown error as an approximate representation of the ARMA model in addition to the statistical errors.

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We attempted to develop an effective control system that can successfully manage the nuclear steam supply (NSS) system of a PWR power plant in an operational mode requiring relatively small variations of power. A procedure is proposed for synthesizing control system that is a simple, yet practiced, suboptimal control system. The suboptimal control system is designed in two steps; application of the optimal control theory, based on the linear state-feedback control and the use of an approximate model-following method. This procedure can appreciably reduce the complexity of the structure of the controller by accepting a slight deviation from the optimality and by the use of the output-feedback control. This eliminates the engineering difficulty caused by an incompletely state-feedback that is sometimes encountered in practical applications of the optimal state-feedback control theory to complex large-scale dynamical systems. Digital simulations and graphical studies based on the Bode-diagram demonstrate the effectiveness of the suboptimal control, and the applicability of the proposed design method as well.

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In order to improve performance of sodium vapor traps, the effects of the operating conditions on the sodium trapping efficiency has been analyzed using a simplified model of the mesh packed region. Then experimental work was performed to evaluate the model validity.
The calculated and the measured values agreed with the maximum deviation of 15%, which led to the following conclusions :
(1) When mist containing sodium vapor is supplied to the vapor trap, the sodium trapping efficiency is greatly lowered because of condensation of vapor to mist.
(2) Even if the sodium vapor contained no mist initially, mist is formed when the cover gas is cooled so rapidly that the nuclei for condensation begin to grow. This mist formation causes a decreased sodium trapping efficiency.
(3) The sodium trapping efficiency can be maximized by selecting restrictive conditions for the vapor trap operation, where the inlet sodium vapor concentration in cover gas is kept in a state of unsaturated and the relative heat capacity rate of the cooling air to the flowing cover gas is maintained between 2 and 5.

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This study is based on the fuel centerline temperature measurement conducted in the Japan Materials Testing Reactor (JMTR). The specimen used in this experiment is the assembly composed of three fuel rods, two high performance fuel rods (one has the Cubarrier cladding and the other has the Zr-liner cladding) and one reference fuel rod (i.e. it has the normal cladding). As to the fuel centerline temperature, it was investigated whether there are (1) differences between the reference fuel rod and the high performance fuel rods and (2) differences between two high performance fuel rods. Consequently, the differences of the fuel centerline temperatures were obvious among three fuel rods. And it was concluded that the inside surface roughness of the cladding of the fuel rods have the important effect on the differences of their fuel centerline temperatures because the fuel centerline temperature of the fuel rod with the Zr-liner cladding of which the inside surface is rougher than that of the normal cladding is higher than that of the fuel rod with the normal cladding etc.

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A series of cell calculations for the Chernobyl reactor was performed using the SRAC code system to provide its fundamental neutronic characteristics for the accident analysis at JAERI. The calculations are based on a two-step cell modelling. The primary cell is supposed on a unit square graphite block of 25 cm×25 cm which contains a fuel assembly or a control channel. The secondary cell is supposed on a unit of 16 channels where 14 fuel and 2 control channels are located so as to simulate the whole core.
Detailed investigation on fractional change of reaction rates for each nuclide along with increase of void fraction was carried out. The analysis clarified the mechanism to induce the positive void coefficient, together with its burn-up dependence and the increase due to withdrawal of control rod.
A comparison of the effect of void fraction on the reaction rates in the primary cell between a Monte Carlo code VIM and SRAC shows consistent results within the statistical error.
The calculated results for the composition of discharged fuel, void and other reactivity coefficients, kinetic parameters and their burn-up dependence show satisfactorily good agreement with those reported by the Soviet Union and some institutions.

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In most of the calculations using analytical methods a reactor core is approximated as cylinder and the reactor parameters are calculated using two-dimensional computer codes. While such calculations are useful in scoping studies in view of azimuthal asymmetry in the actual reactor core these calculations could entail errors of unknown magnitude. The present study reports our estimate of such errors in Κ_eff with the instance of fast reactor having 22 and 23 fuel subassemblies. The Κ_eff are calculated using Monte Carlo code KENO and Hansen-Roach cross section set, modelling the core in two different ways, (1) by approximating the core to a cylinder (2-D calculation), (2) by near exact representation of the core (3-D calculation). The difference in Κ_eff is appreciable between 2-D and 3-D calculations.
Experimental values are adduced in support of these calculations.

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A small loss-of-coolant accident (LOCA) test was conducted simulating a PWR instrument tube break as one of the break location parameter tests in the ROSA-IV large scale test facility (LSTF). Similarities and differences are studied experimentally between the test and a cold leg break (CLB) LOCA test.

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The determination of neutron energy is important as well as the measurement of neutron fluence in D-T neutron source characterization works for the experimental research studies on the fusion reactor developments, such as neutron cross section measurements, blanket/shielding neutronics and material irradiation experiments. The actual neutron energy values of the ordinary D-T neutron source using a thick Ti-T target is determined by the two effects ; one is the kinematic relation of T (d, n)⁴ He reaction process, depending on the deuteron accelerating energy and neutron emission angle, and another is the thickness of the tritium layer in the target, which gives the effective deuteron energy reacting on tritium atoms through the energy moderation of incident deuterons in the finite thick target material.
As one of the practical techniques to determine this source neutron energy, an activationrate ratio method based on ⁹³Nb(n, 2n)^92mNb and ⁹⁰Zr(n, 2n)⁸⁹Zr reactions was proposed and has been applied through the international intercomparison work of D-T neutron source characteristics among each national standard laboratory, such as NBS, PTB, ETL etc.⁽¹⁾ In this intercomparison, several laboratories' values have been found to be different from the neutron energy values determined by this Zr/Nb activation-rate ratio method, which means systematic discrepancy over experimental uncertainties in the conversion curve from the activation-rate ratio data to the mean neutron energy.
In order to improve the applicability of this Zr/Nb activation-rate ratio method, the present paper gives the basic data on their reaction cross sections and activation-rate ratios in the D-T neutron energy region between 13 MeV and 16 MeV, which has been precisely calibrated at the Japanese 14 MeV standard neutron field, and also discussions are made on the accuracy of the mean neutron energy conversion curve recommended here.

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