The introduction of zirconium-hydride (ZrH_1.7) layer between the seed and blanket is very effective in reducing the coolant-void reactivity of steam-cooled FBR. The void reactivity reduction is attributed to the rapid increase of neutron absorption and to the decrease of neutron production in the blanket due to the moderation through the layer. The reason is that the neutron multiplication factor decreases, reflecting the neutron balance of the whole core. It is effective even the layer thickness is 1 or 2 cm. Compared with the conventional uniform introduction of this moderator into the seed, the fixed layer concept is more effective in reducing void reactivity and hardly deteriorate the breeding ratio. The negative void reactivity is proved for the non-flat large-sized radial heterogeneous core where the layers are placed between the seeds and blankets. The neutron absorption rate increases, the fast fission rate decreases in the central, inner and radial blankets.

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A Systematic Reliability Improvement (SRI) program has been developed to support the decision-making processes of preventive maintenance (PM) planning for enhancing plant/system reliability in nuclear power plants. The SRI program is based on improved information-handling methods in the Reliability-Centered Maintenance (RCM) methodology. This program enables us to select a proper PM plan using a computerized system that integrates the following three subsystems which have been developed and is implemented on an engineering work station.
(1) The maintenance management support subsystem can provide statistical parameters which indicate the characteristics of the failure mode for components based on the statistical analysis of field maintenance data.
(2) The FME/CA-database subsystem can manage the system/component failure modes and their characteristics which are estimated by experts at the design stage using 13 types of assessment rankings of the failure mode effects and criticality analysis (FME/CA).
(3) The PM planning support subsystem can support decision-making in determining the priority of PM improvement plans using a new method combining the above FME/CA assessment rankings and interactive logic tree analysis (1-LTA).
The effectiveness of the SRI program and its support systems has been validated through a feasibility study using simulated data on a primary loop recirculation system in BWR plants.

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The purpose of this study is to investigate the feasibility of visualization and void fraction measurement of air-water two-phase flow in a small diameter tube (I. D.: 4.08 mm) by using the real-time neutron radiography and image processing techniques. Video images of two-phase flow were taken by using the real-time neutron radiography system (thermal neutron radiography facility No. 2) installed at the Japan Research Reactor 3 M of the Japan Atomic Energy Research Institute. The shape of bubbles and its moving behavior were clearly observed from the video images. The image corrections for dark current, shading, field intensity fluctuation and electrical system drift were examined in order to measure the void fraction from the video images. Though, generally speaking, the effect of the scattered neutron could not be ignored for quantification of the images taken by the neutron radiography, the scattered neutron could not affect the final results of void fraction in the case of a small diameter tube. The void fraction calculated from the corrected images was correlated well with the drift flux equation, indicating that the existing drift flux equation could be applied to predict the void fraction of two-phase flow in a small diameter tube. It was demonstrated that the real-time neutron radiography technique could be useful for measuring the void fraction of two-phase flow in a small diameter tube.

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Possible mechanisms involved in the removal of oxide scale (chiefly Fe₃O₄) over carbon steel during the course of decontamination of mainly the internal surfaces of the primary system of a reactor have been proposed. During the initial stages, the externally added reducing agent was found to aid the reductive dissolution. But once the oxide layer became porous and the decontamination reagent found access to the base metal, electrons available from the base metal oxidation continued the reductive dissolution. Depending on the relative area of the base metal to the oxide layer exposed, base metal may or may not influence the oxide dissolution. The surface area of the specimen exposed and the rate of regeneration of the reagents also have an important role in the oxide scale removal.

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The NO_x composition and iodine species in the dissolution of spent fuels are discussed on the basis of thermochemical calculations and experimental results. The influence of NO_x sparging on the expulsion of iodine is also discussed. The dissolution of a spent PWR-fuel specimen (3 g) in 30 ml of 3.5 _M HNO₃ at 100°C is calculated to yield a concentration of 7×10^-2 atm of NO₂, which is 80% of the total NO_x in the dissolver. This NO₂ fraction is much higher than experimental values of 15% or less that were reported for dissolver off-gas cooled near to room temperature. The high NO₂ fraction suppresses the formation of iodate (IO₃^-) in the dissolution. The calculations predict that IO₃^- is not formed in 3.5 _M HNO₃ at 100°C at an NO₂ pressure ?? 3×10^-2 atm (3 kPa ). Attempts to expel iodine from the fuel solution indicated that the main iodine species in the solution was colloidal iodine and not iodate (IO₃^-) which earlier workers postulated. The obtained experimental results are consistent with the thermochemical predictions. For the decomposition of the colloidal iodine in the expulsion process, NO₂ sparging has a negative effect. This is because an increase in NO₂ pressure promotes the formation of colloidal Agl.

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The electrochemical reduction of uranyl ion has been studied by the cyclic voltammetry (CV) and the spectroelectrochemical method using optically transparent thin-layer electrode (OTTE) in basic carbonate and acidic aqueous solutions. The CV measurements showed that U( V ) was formed quasireversibly by the reduction of uranyl ion in both basic carbonate and acidic perchlorate solutions. Uranium( V) was found to be much more stable in the carbonate media and its absorption spectrum was measured by the OTTE cell under the applied potentials varying stepwise from -0.7 to -0.9 V vs. Ag/AgCl (3 _M NaCl). On the other hand, in the acidic solution the absorption spectra recorded at 20 s interval by the OTTE cell under the fixed applied potential at -0.4 V vs. Ag/AgCl (3 _M NaCl) exhibited only the spectral change from uranyl ion to the final reduction product U(IV). It has been suggested based on the present electrochemical study that uranyl is stable for the redox reaction in the deep groundwater, which is in reductive environment of -0.3 V (vs. NHE) and contains carbonate ion, but is reduced to U(IV) in the acidic groundwater in the same reductive environment.

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The dissolution behavior of Cu metal in hydrogen peroxide (H₂O₂)-ethylenediaminetetraacetic acid (EDTA) solution was studied through both electrochemical and chemical tests. In the electrochemical experiments, the open circuit potential (OCP) of Cu electrode at low H₂O₂ concentration below 0.5 wt% was determined by Cu anodic reaction and H₂O₂ cathodic reaction. On the other hand, the OCP in higher concentration above 2 wt% of H₂O₂ approached to 70 mV vs. SCE, since the redox reaction of H₂O₂ was so dominant reaction that the Cu anodic reaction was masked. The anodic polarization curve of Cu around the OCP of 70 mV in the range between -200 and 200 mV in the solution without H₂O₂ showed the maximum Cu dissolution reaction front, while those of both carbon steel and Inconel 600 in the same potential range were situated in the passivation region. The anodic polarization behaviors of Cu in the H₂O₂ concentrations above 1 wt% were examined by a potentiostatic method along with a weight loss measurement. The anodic Cu dissolution current estimated from the weight loss measurement in the solution containing more than 1 wt% of H₂O₂ was higher than that predicted by the polarization curve of Cu in the H₂O₂-free solution. The dissolution reaction rate of Cu in the H₂O₂ solution linearly increased with H₂O₂ concentration. This result indicated that the dissolution reaction rate of Cu in the presence of H₂O₂ was enhanced by a reaction between Cu and H₂O₂, resulting in the shift of the polarization curves depending on H₂O₂ concentration. The experimental correlation equation based on the above results described well the overall dissolution reaction rate of Cu in H₂O₂ concentration ranges below 1 wt% and above 2 wt% in the EDTA solution.

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The oxide film at the surface of a titanium electrode, which was observed by an Electron Probe Microanalyzer (EPMA) measurement, was found to hinder an electrolytic reduction of U(VI) to U(IV) in a nitric acid solution containing hydrazine. A method for the pretreatment of the titanium electrode to obtain a clear voltammogram for the reduction of U(VI) was investigated, and it was found that a controlled-potential electrolysis with a potential less than -0.8 V vs. SSE for about 30 min was adequate to remove the oxide film. The effect of the composition of the HNO₃+N₂H₅⁺ solution on the degree of pretreatment was examined. Voltammograms for the reduction of U(VI) at the oxide-free titanium electrode were recorded and analyzed.

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Sorption of Np(IV) on bentonite in the presence of carbonate was studied by a batch method. The sorption was considered mainly due to Np(OH)₂(CO₃)₂^2- and Np(OH)₄(CO₃)₂^4-. The distribution coefficient (K_d) of Np(OH)₂(CO₃)₂^2- and Np(OH)₄(CO₃)₂^4- was determined to be 10^4.01±0.24ml/g and 10^3.93±0.37 ml/g, respectively. The increasing Kd at high pH was attributed to the sorption of NpO₂(OH)₂^- with the K_d of 10^5.82±0.39 ml/g.
The formation of colloidal species was observed under a reducing condition. By assuming Np(OH)₄ as the colloidal species, the formation constant was evaluated to be β_c ?? 10^53.85. Under an oxidizing condition, the formation of the colloidal species was less observed possibly due to the transformation of Np(OH)₄ to NpO₂OH.

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The solubility of tri-n-butylphosphate (TBP) in aqueous solutions of plutonium nitrate (PuN) and in highly radioactive liquid waste (HRLW) of PUREX nuclear fuel reprocessing was investigated. By an empirical formula the solubility of TBP in PuN solutions was described in the range of 00.1 _M Pu and 18 _M HNO₃ concentrations. The following items were elucidated:
(1) The logarithm of TBP solubility (S) in the solution of interest varies inversely in proportion to the concentration of Pu(IV) in the range of 00.1 _M Pu(IV) at a constant concentration of HNO₃, indicating that Pu(IV) simply behaves as an electrolyte for the salting-out of TBP. Log S subsequently levels off with increasing Pu concentration, which would be due to a change in the principal dissolution form of TBP having an interaction with Pu(IV).
(2) The variation in S in PuN solutions (00.1 _M Pu) with nitric acid concentration shows almost the same tendency as that in HNO₃ solution.
(3) A dependency of S on fission product metal ions in HNO₃ for HRLW similar to that for PuN was observed.
(4) The logarithm of the ratio of TBP solubility in water to that in solution of interest was nearly proportional to 1/T for HRLW solution or for low concentration of PuN solution. That deviates from the linear relation at high temperature when the concentration of PuN is increased, which can be explained by the change in ionic form of Pu.

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Linear Discharge Type Analogue Charge Division is a method of determining the position of ionization in the Single Wire Position Sensitive Proportional Counter (SWPC). The division output is proportional to the position of ionization where the numerator input is the pulse (V₁) taken from the one end of the anode and the denominator input is the sum of the pulses (V₁+V₂) taken from both ends. There exists a difference between the timing of the peak of the numerator and denominator input pulses which varies with the position and shows a non-linear characteristic. As the peak-hold gate signal of the analogue charge division is obtained from the denominator input pulse, the effect is reflected on the output of the charge division.
The timing non-linearity of the input signal and the ultimate position output non-linearity are estimated analytically. The effect is evaluated for various parameters, such as, the intrinsic time constant of SWPC, the shaping time constant and gate signal's delay etc. The position output for a 1.8m long (15 μmφ) SWPC has been obtained experimentally and the observed output non-linearity is presented and compared with the estimation.

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To confirm the safety of the High Temperature Engineering Test Reactor (HTTR) facility which is being constructed as the first high temperature gas cooled reactor in Japan, the representative abnormal reactivity events assumed in the safety analysis of the HTTR were analyzed. The HTTR is a graphite moderated and He-gas-cooled reactor with thermal power of 30 MW, inlet coolant temperature of 395°C and outlet coolant temperature of 950°C.
This report presents the analytical results of two representative events, "Abnormal control rod withdrawal from a subcritical condition" and "Abnormal control rod withdrawal during the full power operation", showing that the safety of the HTTR is secured in conformity with the unique features of the HTTR with respect to the maximum fuel temperature, which is a key factor for the safety criteria.
The results of the safety analysis could demonstrate the safety of the HTTR facility with respect to abnormal reactivity events postulated in the HTTR, showing that the maximum fuel temperature is lower than the limit of the maximum fuel temperature of 1, 600°C.

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The precoat type condensate filter system, equipping most BWR plants in Japan, is sought to be improved in terms of filtering performance-i.e. to lower the concentration of crud in filtered condensate water and to extend the filter run length. Laboratory experiments were performed on various measures for improving the filter performance, among which those found to be promising were: (a) Use of more suitable precoat material (cation resin of suitably low cross-linkage and mixture with fibrous cation exchange resin), (b) Adoption of improved precoating procedure (body feed coating-i.e. feeding precoat material continuously during filter operation), (c) Improvement of flow pattern through filtering vessel (insertion of draft tube), and (d) Modification of filter system operating mode (swelling operation-i.e. interruptions in individual filter use in-between backwash operations).
Trial application on actual plants of some of the above measures has demonstrated that the envisaged improvement of filter system performance can duly be expected to be realized.

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The figure of merit (FOM) defined as 1/T have been used for comparing the statistical merits of different devices for radiation measurement by pulse counting, where T is the minimum combined time for the measurement of the radioactivity of a source and background radiation⁽¹⁾. In several types of FOM, the S²/B have been used generally as the FOM for a choice or an adjustment of a counting system, where S and B are the counting rate due to an unknown radioactive source and to background with dummy source, respectively. The criterion for the maximum ratio of S²/B is, however, usable only for a case of S<<B, but not applicable for S ?? B, and not determinable for an ideal system which has no background⁽²⁾⁽³⁾. So the FOM defined as S²/B is not necessarily the effective tool in comparing the statistical merit of a counting system directly.
In this article, for comparing the statistical merits of different measuring systems of radiation, we propose a counting factor (CF) as a FOM. This FOM was derived from the concept of noise factor which has been used in electronic devices to compare the noise performance of an ideal and an actual system.

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A three-dimensional analysis method for sloshing behavior of FBR is developed. The method treats the coolant in a reactor vessel as a potential flow with moving liquid surfaces. The Laplace equation of velocity potential is solved by a boundary element method with its boundary conditions described by a Bernoulli equation.
The method is applied to analysis of sloshing behavior of uni- and multi-vessel type FBRs and results are presented. The latter consists of vessels for the core, heat exchangers and pumps, all of which are connected by piping. In the uni-vessel type, heat exchangers and pumps are placed in the reactor vessel.

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