Safety-related LWR fuel studies in Japan during the past several years are overviewed. Fuel Specialists Group (NEN-ANSEN), JAERI and NUPEC have been the main members in this field. Participation in International Projects at Halden, Studsvik, Battelle etc. has been useful and fruitful in obtaining the wide range of data for LWR fuel behaviour in normal and transient operating conditions. So-called "reliability tests" were carried out at NUPEC supported by the Government to verify the integrity of then-current design LWR fuels. Demonstration of the higher-duty fuel program is also overviewed. Burn-up Extension Program has been going on also as an activities of NUPEC.
For fuel behaviour under accident conditions, JAERI made important contributions through the experiments in NSRR. Fuel behaviour during severe-core damage has been considered based on the data obtained from the International Corporation Programs.

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Analytical prediction model of critical heat flux (CHF) has been developed on the basis of film dryout criterion due to droplets deposition and entrainment in annular mist flow. Critical heat flux in round tubes were analyzed by the Film Dryout Analysis Code in Subchannels (FIDAS) which is based on the three-fluid, three-field and newly developed film dryout model. Predictions by FIDAS were compared with the world-wide experimental data on CHF obtained in water and Freon for uniformly and non-uniformly heated tubes under vertical upward flow condition. Furthermore, CHF prediction capability of FIDAS was compared with those of other film dryout models for annular flow and Katto's CHF correlation. The predictions of FIDAS are in sufficient agreement with the experimental CHF data, and indicate better agreement than the other film dryout models and empirical correlation of Katto.

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In order to clarify differences in reflooding behavior between a PWR in which a part of the ECC water is directly injected into the downcomer and that in which it is injected into the cold legs, an experimental study has been performed with Cylindrical Core Test Facility. Two experiments were conducted under the same test conditions only except for injection location of the Low Pressure Coolant Injection (LPCI) water. In one test LPCI water was injected into the downcomer, whereas into the cold legs in the other.
Through investigation of the experimental data, the following has been clarified : Periodic oscillation of pressures in primary system, fluid temperatures around downcomer and so forth appear under the downcomer injection, whereas does not appear under the cold leg injection. The cause of the oscillation is that water at the top of the downcomer is subcooled under the downcomer injection due to incomplete mixing between steam flowing into the downcomer and LPCI water. Downcomer water is saturated under the cold leg injection. However, average values of the oscillatory data are almost the same as those of the cold leg injection, and hence, overall reflooding behavior is almost the same in the two cases.

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Nuclear spin relaxation time T₁ of 19F in molten lithium trifluoroberyllate (LiBeF₃) was measured at frequency of 20 MHz by means of a pulse n. m. r. technique. In the plot of T₁ against temperature T shows a minimum at about 813 K and T₁^-1can be expressed as [r/(1+ω²τ²)+4τ/(1+4ω²τ²)] in terms of correlation time τ. We obtained the relation τ(s) = 1.9 x10^-16 exp[1.04(eV) /kT]. The behavior of T₁ is attributed to the dissociation of F^- ions from [BeF₄]^2- complex ions and long distance diffusion of the F^- ions in the melt. At lower temperatures than 690 K the other mechanism prevails where the main contribution to the relaxation is the rotation of the complex ions.

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Although one-dimensional analysis developed by Jones & Furry is useful for predicting separative performances of hot-wire thermal diffusion column for isotope separation, it is not so simple and requires inevitably numerical calculations with computers, because it involves the column constants which cannot be evaluated analytically. Approximate formulae for the column constants were derived here by replacing the logarithmic function for the radial temperature distribution, which makes analytical integration impossible when exists at a denominator, with the radially constant and column-design-parameter dependent reference mean temperature. Laborious repeated integrals, with semi-explicit axial convection velocity profile being input, were performed manually through ignoring S or ratio of radius between hot wire and cold wall, to the second and higher powers. The results were verified by the symbolic manipulating language REDUCE. Although the formulae obtained in terms of δ and ΔT/T_c, or the ratio of temperature difference between hot and cold surfaces to cold wall temperature, were rather rough in approximation, calculations for dependence of column constants on (ΔT/T_c) showed fairly good agreements with those by original expressions.

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This report describes the results of reactivity initiated accident (RIA) experiments using UO₂ fuel pellets containing additives which enhance the pellet-cladding interaction (PCI) resistance of BWR fuels.
The results obtained were :
(1) The failure thresholds of the PCI resistant fuels with additives were not less than 260 cal/g•UO₂ (220 cal/g•UO₂ in enthalpy) and no measurable differences in failure thresholds were observed between the fuels with additives and the conventional BWR (and NSRR standard) fuels.
(2) No measurable difference in the threshold for mechanical energy release existed between the PCI resistant fuel with additives and the conventional BWR (and NSRR standard) fuels. The threshold was estimated to be approximately 335 cal/g•UO₂ (285 cal/g•UO₂ in enthalpy) for the fuels with additive. This is also the value previously established for the NSRR standard fuel rods.

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Dryout experiments of water have been conducted in an annulus with inside heating (heat flux from inner wall only) under high-pressure, low-flow and mixed inlet conditions which are of importance in the core thermal-hydraulic behavior during a loss-of-coolant accident (LOCA) and also partially during an anticipated transient without scram (ATWS) of a nuclear reactor. The experimental conditions have covered ranges of pressure of 3 MPa, mass flux from 105 to 320 kg/m²-s and inlet quality from 0.15 to 0.90. The dryout data have been compared with several existing empirical critical heat flux (CHF) correlations and a new correlation. The Katto correlation predicts best the CHF among the existing correlations examined. However, even the Katto correlation overpredicts the CHF by factors up to 2 at about 1/6 data points of the present dryout data. The present dryout data are divided into two groups (regions) according to the value of a non-dimensional number l_bo/d_he, where l_bo is the assumed boiling length and d_he the heated equivalent diameter. A new correlation covering both the regions has been developed by correlating the present dryout data in terms of two non-dimensional numbers. The new correlation performs best among the correlations examined in predicting the present dryout data.

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Neutron-induced reaction cross sections of ³He and ⁴He have been evaluated in the energy range of 10^-5eV20 MeV. Evaluated quantities are the total and elastic scattering cross sections of ⁴He and the total, elastic scattering, (n, p) and (n, d) reaction cross sections of ³He as well as the angular distributions of elastically scattered neutrons for both isotopes. The total and elastic scattering cross sections of ⁴He were analyzed with the R-matrix theory in the overall energy region. As for ³He, the R-matrix calculation was performed to evaluate the total, elastic scattering and (n, p) reaction cross sections below 1 MeV, and the evaluation of the cross sections in the energy region above 1 MeV was based on available experimental data. The evaluated data have been compiled into JENDL-3 in the ENDF-5 format.

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