An Sn-albedo Monte Carlo-albedo Monte Carlo (Sn-AMC-AMC) coupling technique has been developed and implemented in the AMC code system MORSE-ALB in order to make an effective and accurate shielding calculation for large and complex geometry. This paper describes the feature of the present coupling method which includes new source geometry in the Sn-AMC coupling calculation. Applicability of the present method was studied by analyzing the experiments of neutron streaming in "JOYO". The Sn-AMC-AMC coupling calculation reproduced the measured data within the accuracy of one order of magnitude. Considering that neutron fluxes attenuate by 812 orders of magnitude from the core to the measuring positions and the streaming paths are complicated, the agreement can be regarded good.

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The 10, 5 and 2.5% cold leg break loss-of-coolant accident experiments were conducted by using the Large Scale Test Facility (LSTF) of the Rig of Safety Assessment (ROSA)-IV program. In the early stage of the 5% break experiment, the core collapsed liquid level was depressed nearly to the core bottom and the dryout of the core was observed. However, the core liquid level depression without the core dryout was observed in the 10 and 2.5% break experiments. In the three break experiments, the core liquid levels were recovered just after the loop seal clearing.
The manometric effect due to the liquid seal formation in the loop seal and the liquid holdup in the steam generator (SG) U-tubes upflow-side caused a depression of the core collapsed liquid level. The liquid holdup in the U-tubes upflow-side was observed after the termination of the two-phase circulation due to the phase separation at the U-tubes top. The counter current flow limiting (CCFL) and the condensation of steam was considered to be the main reason for the liquid holdup. In the 10, 5 and 2.5% break experiments, the termination of the two-phase circulation and the loop seal clearing were observed approximately at 4060% and 30% mass inventory in the primary system, respectively.

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Creep characteristics of Zircaloy-4 stress-releaved cladding under internal pressure were studied. Creep tests were conducted under 21 conditions chosen from the temperature range of 603693 K and the hoop stress of 49314 MPa. The maximum accumulated test period was 3, 000 h. Diametral creep data were analyzed by separating the primary (transient) and the secondary (steady-state) creep, based on Dorn's quasitheretical model, and the following equations were derived:
Total creep strain:
Saturated primary creep strain:
Steady-state creep rate:
The apparent activation energy of the steady-state creep, which is 2.72 t× 10⁵ J/mol, is in good agreement with these of self-diffusion of Zr in Zr-Sn alloys and suggests the self-diffusion is the control mechanism.

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The equilibrium distributions of thorium and radium between molten LiCl-LiF salt and liquid bismuth were determined over a wide range of salt composition at 913 and 1, 023 K. The distribution behavior of thorium was apparently affected by salt composition and temperature, but that of radium was little affected. The effect of salt composition on the distribution behavior was well explained by considering the formation of complex compounds in the salt phase. The thermodynamical properties of the complex compounds were also evaluated.

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Pyrolysis of spent ion exchange resins is used to reduce radioactive waste volume and to make the final waste form more stable. The weight loss of cation exchange resin after pyrolysis is only 50w/0 while that of anion exchange resin is 90 w/0. Fundamental experiments were performed to investigate the reason for the small weight loss of the former,
The cation resin consists of base polymer and functional sulfonic acid groups. Chemical analyses of the pyrolysis products showed that 65% of the functional groups decomposed at about 300°C and generated SO₂ gas. However, only a small amount of the base polymer was pyrolyzed even at 600°C and the weight loss was only 50 w/0. The IR and XPS studies on the residue showed that 35% of the functional sulfonic acid groups was converted to sulfonyl and sulfur bridges between the base polymers during pyrolysis. These bridges made the base polymers thermally stable. Therefore, the small weight loss of the cation resin was attributed to formation of bridges, which originated from the functional groups.

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A CR-39 plastic track detector has been applied to thermal neutron dosimetry by combining with a converter using (n, α) reactions. A ceramic boron nitride (BN) was utilized as the promising converter because of high boron-concentration and its smooth surface. The contribution of background tracks was evaluated in the region of CR-39 plastic unattached to BN converter. The detector was exposed to a reference neutron field generated by ²⁵²Cf source in a water tank in order to investigate the detection characteristics. It was found that the efficiency was (1.0 ± 0.1) × 10^-3 pits/n, corresponding to the sensitivity of (9.6 ± 0.9) × 10² pits/mm2/mSv. The linear response was experimentally confirmed between 0.035 and 0.7 mSv. The minimum detectable dose equivalent was also estimated to be about 0.005 mSv.

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The performance of PWR power plants is evaluated through analysis of their power generation records, and the directions are given in which measures might be effectively sought for further improvement of plant performance and productivity. The boundary between what has already been achieved in performance of such plants and what remains to be done with further efforts in development and demonstration is clearly identified.
To supplement the traditionally adopted "capacity factor" (average power/nominal capacity) for assessing the improvements gained in the performance of uprated fuels and reactor cores, additional new yardsticks are proposed to represent the productivity of nuclear fuel and reactor core. For evaluating the performance of fuel, the proposed variable is based on the correlation between specific power and annual core average burnup, i.e. thermal power and annual heat generation per unit mass of fuel. Similarly for the reactor core, the variable is based on the correlation between thermal power and heat generation per unit core volume, and the advantages of adopting the proposed variables are discussed.
Operating experience with PWR plants indicates that, with nuclear fuel, their relatively short service life compared with the core structure and other reactor components has permitted reliable, effective service of uprated fuel in high-performance plants to be demonstrated over periods extending beyond the service life of individual fuels. This is not the case, however, with the core structure and other reactor components: most existing plants have hardly been in service for more than half of the expected service life of these components, therefore data available today are insufficient for evaluation of their long-term performance. An analysis is presented on possible repercussions to be expected from the current trend in development, which tends towards higher core power rating, and it is pointed out that certain plant components will possibly come to be exposed to increasingly severe thermal conditions, to call for further efforts in development and demonstration to ensure their continued reliability in service.

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Waterlogged fuel rod experiments performed at the NSRR are analyzed using the computer code WTRLGD, which was devised for the analyses of thermo-dynamical behavior of a waterlogged fuel rod. The numerical results are compared with the data from the experiments in order to assess the validity of the computer code. Parameters in the analyses are volumetric fraction of water, reactor period, gap width, a pin hole and the end peaks. Thus the analyses cover almost all the waterlogged fuel rod experiments at the NSRR.
The comparison shows good agreement between the experimental results and numerical ones on the transient thermo-dynamical behaviors of fuel, such as, rod internal pressure, cladding surface temperature and cladding strain. The numerical results also quantitatively agree with the experimental data concerning the effects of the above parameters on failure threshold energy. From the above findings, the computer code is assessed to be valid enough for the analyses of the failure behavior of the waterlogged fuel rod under a reactivity initiated accident condition.

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