Fuel behavior in simulated reactivity initiated accident (RIA) was studied under high pressure and temperature coolant condition of LWR operating environment. Test fuel rods were pulse-irradiated in the Nuclear Safety Research Reactor (NSRR) to simulate transient power generations at RIAs. From the results, it was clarified that the cladding collapsed under high external pressure but that the basic mechanisms and threshold enthalpies of incipient fuel failure were not different from those observed in the tests conducted under atmospheric pressure, room temperature and stagnant condition.

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Specimen holders with multiple holes are often used in capsules for material irradiation experiments. In reactor, all parts of capsule are heat sources due to gamma-heating. Then, in thermal analysis, such cylinders must be considered as heat generators with multiple holes.
To estimate, for instance, the maximum temperature of test pieces inserted in the holes, temperature distribution in the cylinder must be analysed. Then, it was tried to obtain an analytical expression of the temperature distribution in the cylinder.
An equation was obtained by solving the steady-state heat conduction equation with practical assumptions. The equation will be useful in design and safety evaluation of capsules for material irradiation experiments.

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Inconel alloys have been used extensively as heat exchange tubing materials in the power generation industry because of their excellent mechanical properties at elevated temperatures. Evaluation of the alloys for High Temperature Gas-Cooled Reactor (HTGR) applications has proved that Ni-Cr-Co alloys have good high temperature strength, creep properties and weld-ability.
In this work, the cyclic fatigue behavior and creep crack growth rate of Inconel 617 alloy were tested for analysis of time-dependent crack propagation characteristics at 650°C. Micro-structural examination was performed for observation of grain boundary carbide precipitation growth and the coalescence of intergranular microcracks that would enhance the crack propagation rate. It was observed that at higher temperatures, low fatigue frequency, and higher R-ratio (mean-stress), the effects of dynamic recovery tended to become stronger because the mobility of dislocation increased with increasing temperature. As a result, the subgrain grew to a larger size. Eventually, the influence of time-dependent processes on crack growth rate can effectively represent the overall effects on a creep ductile material at elevated temperatures.

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The sensitivity of uranium solubility to variation of ligand concentrations in the groundwater was systematically evaluated. Predominant dissolution reactions between uranium solubility limiting solids and predominant aqueous species were obtained by means of calculating uranium solubility and speciation using the NEA Thermochemical Data Base of uranium and the geochemical code SOLGASWATER. Logarithm of the mass law equations for the predominant dissolution reactions allowed that logarithmic concentration of total dissolved uranium species, log[UO₂²+]_Τ, was a linear function of pH and logarithm of ligand concentration. The equation for the difference of log[UO₂²+]_Τ between the U-H₂O system and U-ligand-H₂O system was derived from the linear functions in the common pe-pH area of the two systems. Differential coefficients obtained from the difference of log[UO₂²+]_Τ allowed to judge which predominant dissolution reactions are sensitive to the variation of ligand concentrations as follows:
(1) The uranium solubility increases most sensitively with ligand concentrations, according to the following equations;
at pe 10,
UO₃•2H₂O(cr)+3HCO^-₃, =UO₂(CO₃)₃^4-+3H₂O(1)+H⁺ (the CO₂-closed system),
UO₃•2H₂O(cr)+3H₂CO₃(aq)=UO₂(CO₃)₃^4-+3H₂O(1)+4H⁺ (the CO₂-open system).
and at pe -4,
UO₂(cr)+4H⁺+4F^-=UF₄(aq)+2H₂O(1).
(2) The uranium solubility decreases with ligand concentrations, according to the following equations;
at pe 10,
UO₂CO₃(cr)+2H⁺=UO₂²⁺+H₂CO₃(aq)
(UO₂)₃(PO₄)₂•6H₂O(cr)+4H⁺=-3UO₂²⁺+2H₂PO^-₄+6H₂O(1),
(UO₂)₃(PO₄)₂•6H₂O(cr)=3UO₂(OH)₂(aq)+2H₂PO^-₄+2H⁺,
and at pe -4,
USiO₄(cr) +4H₂O(1) =U (OH)₄(aq)+H₄SiO₄(aq).

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A new type of β-ray calibration source was designed for estimating tissue doses from skin contamination. Thin polyethylene ion-exchange membranes were applied for the source preparation. The method of determining tissue absorbed dose rates with an extrapolation chamber was discussed for the calibration sources prepared with ¹⁴⁷Pm, ²⁰⁴Tl and ⁹⁰Sr-⁹⁰Y. The influence of backscattered β-rays by source mount materials was estimated as a function of membrane source thickness. The membrane sources are quite suitable for the present purpose, because they have large uniform radioactive areas and can be easily handled without radioactive contamination.

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The iron, carbon and beryllium cross sections in JENDL-3 have been tested by the continuous energy Monte Carlo analysis of the neutron shielding benchmark experiments. The iron cross sections have been tested with analysis of the ORNL and the Winfrith experiments using the fission neutron sources, and also the LLNL iron experiment using the D-T neutron source. The carbon and beryllium cross sections have been tested with the JAERI-FNS TOF experiments using the D-T neutron source.
Revision of the subroutine TALLYD and an appropriate weight-window-parameter assignment have been accomplished in the MCNP code. In consequence, the FSD for each energy bin is reduced so small that the Monte Carlo results for neutron energy spectra could be recognized to be reliable.
The Monte Carlo calculations with JENDL-3 indicate a good agreement with the benchmark experiments in a wide energy range, as a whole. Particularly, for the Winfrith iron experiment, the results with JENDL-3 give better agreement, just below the iron 24 keV window, than that with ENDF/B-IV. For the JAERI-FNS TOF graphite experiment, the calculated angular fluxes with JENDL-3 give closer agreement than that with ENDF/B-IV at several peaks and dips caused by the inelastic scattering.
However, distinct underestimation is observed in the calculated energy spectrum with JENDL-3 between 0.8 and 3.0 MeV for the two iron experiments using fission neutron sources.

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