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A series of critical experiments were performed using heterogeneous cores at the Static Experiment Critical Facility (STACY) of the Japan Atomic Energy Agency (JAEA) in order to obtain systematic benchmark criticality data concerning the dissolving process in reprocessing plants. Focusing on the use of gadolinium as a soluble poison, critical mass was measured for a combination of uranium dioxide fuel rods (5 wt% 235U) and uranyl nitrate solution (6 wt% 235U) poisoned with gadolinium (Gd). Fuel rods were arrayed with a 1.5 cm lattice interval in the poisoned fuel solution in a 60-cm-diameter cylindrical tank. The uranium concentration of the solution was roughly kept at about 330 gU/L, and the Gd concentrations were varied up to 0.1 gGd/L. The other series of experiments were also conducted, as reference cases, varying uranium concentration in the fuel solution without Gd. The results provided benchmark criticality data for validation of neutron multiplication factor calculation on heterogeneous systems s...

A series of critical experiments were performed using heterogeneous cores at the Static Experiment Critical Facility (STACY) of Japan Atomic Energy Agency (JAEA) in order to obtain systematic benchmark data concerning the dissolving process in a reprocessing plant. Focusing on the introduction of the burn-up credit, critical mass measurement was conducted for a combination of uranium dioxide fuel rods (5 wt% 235U) and uranyl nitrate solution (6 wt% 235U) poisoned with pseudo-fission-product (FP) elements—samarium, cesium, rhodium, and europium. Fuel rods were arrayed at a 1.5-cm lattice interval in the poisoned fuel solution in a 60-cm-diameter cylindrical tank. The uranium concentration of the solution was roughly kept at about 320 gU/L, and the FP element concentrations were adjusted to be equivalent to that in a burn-up of about 30GWd/t. The result provides basic experimental data for validation of computational methods to evaluate the reactivity effect of each FP element, as well as benchmark critical...

In the preparation process of a MOX powder for a MOX fuel fabrication facility, zinc stearate will be added into the MOX powder as an additional material to adjust the density. If too much of the material is added by mistake, the criticality characteristics of the MOX fuel will be influenced because of its neutron moderation effect. If a criticality condition should be induced by the excess zinc stearate, melting and pyrolysis of the material could be caused by the fission energy, and, therefore, the dynamic characteristics of the MOX fuel would be affected in a feedback mechanism. To provide a quantitative evaluation of such dynamics, thermal properties such as exothermic and endothermic calorific values, reaction rates, etc. were measured and correlated with the respective physical variations and release behavior of pyrolysis gas with the pyrolysis reaction. It was found that the exo/endothermic behavior of the material with rising temperature could be divided into six regions and that a rapid rise in pressure occurred over about 400°C caused by the pyrolysis reaction. Furthermore, on the basis of these results, an evaluation model for the thermal properties under criticality conditions was also investigated.

Single-unit criticality condition data were calculated for homogeneous uranium materials in six chemical forms for revision of the Data Collection appendix to the Nuclear Criticality Safety Handbook. The calculated criticality condition data were the estimated critical and the estimated lower-limit critical masses and volumes of spheres, diameters of infinitely-long cylinders and thicknesses of infinite-area slabs of uranium materials in six chemical forms encountered in criticality safety evaluation of nuclear fuel cycle facilities. The chemical forms were U-H 2 O, UO 2 -H 2 O, UO 2 F 2 , UO 2 (NO 3 ) 2 , ADU-H 2 O and UF 6 -HF, the first and last ones of which were not dealt with the Data Collection. The calculations were made with the continuous-energy Monte Carlo criticality calculation code MVP and the Japanese Evaluated Nuclear Data Library JENDL-3.2. The values and precision of the present calculations are discussed in comparison with previous results.

A method for calculating effective microscopic and macroscopic cross sections for media containing randomly dispersed particles, which was originally derived by Shmakov et al., is improved to extend its applicability to a wider range of criticality calculations. The newly modified method can be applied to media containing more than one particle type. This technique is incorporated into a continuous energy Monte Carlo code MCNP and can be applied to a wide variety of criticality problems. The cause of an inadequacy of the original method for larger particles is investigated, and the accuracy is found to be improved by using an adjustment parameter. This method originally underestimated fission neutrons’ tracks in particles, which degrades calculation results for larger particles. A simplified method to implement the effect of the fission neutron tracks into MCNP is developed. We demonstrate that the new technique is successfully applied to MOX fuel rods with plutonium spots, fuel solution containing absorber particles, etc. The newly improved method can treat media containing particles with unequal diameters. However, a sample calculation shows the accuracy of criticality calculations becomes worse with increasing variation in particle diameters. The new method can also successfully perform a criticality calculation for media containing different particle compositions with an equal diameter.

In conventional criticality evaluations of nuclear powder systems, effects of particulate behavior were not considered. In other words, it is difficult to take into account the particle motion in the criticality evaluations. We have developed a novel criticality evaluation code to resolve this problem. The criticality evaluation code, coupling a discrete element method simulation code with a continuous-energy Monte Carlo transport code, makes it possible to study the effects of the particulate dynamics on criticality. This criticality evaluation code is applied to the mixed-oxide (MOX) fuel powder agitation process. The criticality evaluations are performed while mixing the MOX fuel powder and an additive powder in a stirred vessel to investigate the effects of the powder free surface deformation and the particulate mixture state on the effective multiplication factor. The evaluation results reveal that the effective multiplication factor decreases due to the powder boundary deformation while it increases as the mixture condition of MOX powder and Zn-St powder is close to homogeneous.

Simplified evaluation models are developed at the Japan Atomic Energy Research Institute (JAERI) to predict the first peak power, energy, and total fission numbers during a criticality accident for...

A new algorithm of Monte Carlo criticality calculations for implementing Wielandt's method, which is one of acceleration techniques for deterministic source iteration methods, is developed, and the algorithm can be successfully implemented into MCNP code. In this algorithm, part of fission neutrons emitted during random walk processes are tracked within the current cycle, and thus a fission source distribution used in the next cycle spread more widely. Applying this method intensifies a neutron interaction effect even in a loosely-coupled array where conventional Monte Carlo criticality calculation methods have difficulties, and a converged fission source distribution can be obtained with fewer cycles. Computing time spent for one cycle, however, increases because of tracking fission neutrons within the current cycle, which eventually results in an increase of total computing time up to convergence. In addition, statistical fluctuations of a fission source distribution in a cycle are worsened by applying ...

As a part of the development of a subcriticality monitoring system, a system which has a time series data acquisition function for detector signals and a real time evaluation function of the α value with the Feynman-α method was established, with which the kinetic parameter (α value) was measured at the STACY heterogeneous core. Hashimoto's difference filter was implemented in the system, enabling the measurement in a critical condition. The measurement result of the new system agreed with the pulsed neutron method. The α value was also evaluated by the newly developed algorithm of α-mode calculations for the Monte Carlo method, which was compared with the experimental value. Those computational and experimental α values show agreement within their uncertainties.

The gamma-ray monitor installed at Processing Facility 1 in the JCO Tokai-works recorded the gamma dose rate change over time in proportion to power (nuclear fission rate) in the precipitation vessel from the beginning of the criticality accident to the end of the critical condition. The shape of the gamma dose rate record from 25 min after occurrence to the point of termination ('plateau' part) was normalized using the fission number 2.2 x 10{sup 18}, and the absolute value of the power was evaluated.

Evaluated criticality benchmark data obtained at the Static Criticality Experiment Facility (STACY) account for a large percentage of low-enriched uranium (LEU) solution systems documented in the 'International Handbook of Evaluated Criticality Safety Benchmark Experiments'. These data are available for validation of computer codes and nuclear data used for criticality safety analyses of LEU solution systems. The calculated k{sub eff}'s for the water-reflected STACY criticality experiments have been overestimated with JENDL-3.2 by {approx}0.7%. These overestimations were kept in mind while making modifications of the fission spectrum and the fission cross section of {sup 235}U, and the (n,p) cross section of {sup 14}N in JENDL-3.3. Because of these modifications, the k{sub eff}'s calculated with JENDL-3.3 were largely improved. The contributions of these modifications in JENDL-3.3 with respect to JENDL-3.2 and ENDF/B-VI.5 were investigated by performing perturbation calculations. The overestimation of the elastic-scattering cross section of {sup 56}Fe in the mega-electron-volt range was one of the reasons for the k{sub eff} overestimations for the STACY experiments with JENDL-3.2. The modification of {sup 56}Fe cross sections in JENDL-3.3 reduces k{sub eff}'s in the STACY experiments by 0.2%. The dependence of calculated k{sub eff}'s on uranium concentration still exists in JENDL-3.3. The overestimation of calculatedmore » k{sub eff}'s for the STACY experiments with JENDL-3.3 is not insignificant and is as much as 0.6%. These problems are to be resolved in a future evaluation of the cross-section library.« less

The reactivity effect of neutron interaction between two identical units containing low enriched (10% 235U enrichment) uranyl nitrate solution without neutron isolator was measured in the STACY. The unit has 350 mm of thickness and 690 mm of width and distance between those two units was adjustable from 0 to 1,450 mm. Condition of the solution was about 290gU/l in uranium concentration, about 0.8 N in free nitric acidity, 24–27°C in temperature and about 1.4g/cm3 in solution density. The reactivity effect was estimated from variation of critical solution level from 495 to 763 mm depending on the core distance. The reactivity effect was also evaluated by the solid angle method and a computational method using the continuous energy Monte Carlo code MCNP-4C and the nuclear data library JENDL 3.2. Those estimations were compared and the distance to isolate neutron interaction was overestimated by the solid angle method. The computational method considering neutron reflection from surrounding structures reprod...

The third series of the critical experiments on 10% enriched uranyl nitrate solution has been performed at the Static Experiment Critical Facility (STACY) of the Japan Atomic Energy Research Institute. Water-reflected and unreflected 80-cm-diam cylindrical cores were used to obtain the systematic data of critical solution height and differential reactivity for various uranium concentrations from 190 to 240 g/l. The numerically evaluated extrapolation length of neutron flux distribution was in good agreement with the experimental result. The effective neutron multiplication factor k{sub eff} for each core configuration and the effect of uncertainties on k{sub eff} were also numerically evaluated with both the detailed experimental configuration of critical cores and a benchmark model provided for the validation of nuclear calculation codes. The MCNP 4B was used for the evaluation calculations with JENDL-3.2 cross-section library, and the value of the k{sub eff} of the benchmark model was reproduced within the difference of 0.05% {delta}k{sub eff} for the water-reflector cores and 0.17% {delta}k{sub eff} for the unreflected cores.

The effect of higher-harmonic fluxes on a subcriticality measurement by the exponential experiment was investigated. An analysis method for obtaining higher-harmonic eigenvalues and eigenfunctions of the neutron diffusion equation of which the eigenvalues are the decay constants in the exponential experiment was developed. We attempted to reproduce three-dimensional neutron flux distributions formed in the exponential experiment with superposition of higher-harmonic fluxes. Exponential experiments were performed for a subcritical rectangular fuel rod array of the TCA (Tank-type Critical Assembly) and the vertical neutron flux distributions at several horizontal locations were measured for various neutron source positions. Higher-harmonic eigenvalues and eigenfunctions in the exponential experiment were calculated to reproduce the vertical neutron fluxes. It was found that they were well reproduced within the fuel rod array or in the water-reflector region close to the array by taking account of the fewer-...

The kinetic parameter βeff/l of low enriched (10% EU) uranyl nitrate solution was measured by the pulsed neutron source (PNS) method in STACY This measurement was repeated systematically over several uranium concentrations from about 195 gU/l to about 430 gU/l. Used core tanks were two cylindrical tanks whose diameters are 600 mm and 800 mm and one slab tank which has 280 mm thickness and 700 mm width. In this report, experimental data such as solution conditions, critical solution level for each solution condition, subcritical solution levels where measurements were conducted, measured decay time constants of prompt neutron and extrapolated βeff/l values are described as well as basic principle of the PNS method. The kinetic parameter βeff/l values were evaluated also by computation with the diffusion code CITATION in SRAC and the nuclear data library JENDL 3.2. Strong linear correlation has been found between kinetic parameter βeff/l and uranium concentration regardless of differences of reflecting cond...

Mechanisms of a positive temperature reactivity coefficient that occurs in a dilute plutonium solution are investigated based on the perturbation theory and the four-factor formula. The temperature coefficient of a solution fuel is positive if the adjoint flux increases with neutron energy between 0.05 and 0.2 eV. As compared to {sup 239}Pu, {sup 241}Pu has a tendency to make the temperature coefficient of a plutonium solution positive because of the energy dependence of the capture cross section of {sup 241}Pu. As {sup 241}Pu in a plutonium solution decays into {sup 241}Am with time, the temperature coefficient of the solution becomes more positive. Since the capture cross sections of most neutron absorbers such as boron and gadolinium decrease with increasing neutron energy between 0.05 and 0.2 eV, soluble absorbers in a plutonium solution make the temperature coefficient positive for higher-concentration plutonium solutions. Cadmium and samarium dissolved in a dilute plutonium solution can exceptionally keep the temperature coefficient negative because of the energy dependence of the capture cross sections. A fixed neutron absorber generally makes the temperature coefficient of a plutonium solution negative regardless of the property of absorber materials.

(2002). Evaluation of Photon Shielding Capability for Epithermal Neutron Detection during Reactivity-Initiated Power Burst Experiments. Journal of Nuclear Science and Technology: Vol. 39, No. 7, pp. 800-803.

The second series of critical experiments with 10% enriched uranyl nitrate solution using a 28-cm-thick slab core have been performed with the Static Experiment Critical Facility of the Japan Atomic Energy Research Institute. Systematic critical data were obtained by changing the uranium concentration of the fuel solution from 464 to 300 gU/l under various reflector conditions. In this paper, the thirteen critical configurations for water-reflected cores and un reflected cores are identified and evaluated. The effects of uncertainties in the experimental data on k eff are quantified by sensitivity studies. Benchmark model specifications that are necessary to construct a calculational model are given. The uncertainties of k eff's included in the benchmark model specifications are approximately 0.1%δk eff. The thirteen critical configurations are judged to be acceptable benchmark data. Using the benchmark model specifications, sample calculation results are provided with several sets of standard codes and c...

A modified quasi-steady-state method has been developed in order to evaluate the mean power during a nuclear excursion in fissile solution. The conventional method used the critical equation based on the one-group theory in order to calculate the reactivity. However, the one-group approximation reduces the calculation accuracy, and the geometrical buckling used in the critical equation is not applicable to complex geometries. Thus, we have modified the method to use the reactivity feedback coefficients, which are widely used in the calculation of one-point reactor kinetics. Although the modified method requires an external calculation to obtain the feedback coefficients, it is applicable to complex geometries and provides more accurate results than does the one-group approximation when the proper coefficients are given. Moreover, a new method to calculate the boiling power has been developed. In this method, the power corresponding to the void fraction that compensated for the inserted reactivity along wi...

For the decommissioning of the Fukushima Daiichi Nuclear Power Stations, fuel debris involving molten structural materials should be retrieved from each reactor unit. The fuel debris, which is of uncertain chemical composition and physical state, needs to be treated with great care from the standpoint of criticality safety. For developing criticality control for the fuel debris, the Japan Atomic Energy Agency (JAEA) has been planning to modify the Static Experiment Critical Facility (STACY) and to pursue critical experiments on fuel debris. STACY, a facility using solution fuel, is to be converted into a thermal critical assembly using fuel rods and a light water moderator. A series of critical experiments will be conducted at the modified STACY using simulated fuel debris samples. The simulated fuel debris samples are to be manufactured by mixing uranium oxide and reactor structural materials with various chemical compositions. This report summarizes a facility development project for an experimental study on criticality control for fuel debris using the modified STACY and simulated fuel debris samples.

One of the ways of improving fuel cycle cost is to reduce neutron leakage from a core using a reflector. For this purpose, experiments were carried out to investigate the reactivity effect of an iron reflector in a light-water-moderated core using the critical assembly, TCA. The experiment showed that iron reflectors of 15 cm thickness made the core more reactive than water and that the increasing the thickness from 2.2 cm to 15 cm produced 1.8%Δk/k core reactivity gain. The experiment was analyzed with the two-dimensional transport code PHOENIX-P and the continuous energy Monte Carlo code MVP. From the analyses, it has been found that, if the ENDF/B-VI data for iron isotopes are used, the calculated reactivity effect gives good agreement with the experiment and that the epi-thermal and thermal capture reaction rate distributions measured with gold wires and the resultant spectral index distribution are also well reproduced. The experiment and calculations have revealed that increasing a baffle plate thic...

β 2 -glycoprotein I (β 2 -GPI=apolipoprotein H) is an important autoantigen in patients with the antiphospholipid syndrome. It also plays a role in lipoprotein metabolism, such as anti-atherogenic property, triglyceride removal, and enhancement of lipoprotein lipase. Serum β 2 -GPI concentration of 812 apparently healthy Japanese individuals was measured by sandwich EIA. Two families with complete β 2 -GPI deficiency were identified. In one family, all affected had increased serum LDL-cholesterol levels or smaller particle sizes of LDL, while the other had no apparent abnormality in lipid metabolism. Individuals investigated had no history of thrombosis or overt abnormalities in hemostatic tests. A thymine corresponding to position 379 of the β 2 -GPI cDNA was deleted in every β 2 -GPI deficient individual. The incidence of this heterozygous deficiency determined by RFLP was 6.3% in Japanese and none in Caucasians. Heterozygotes had significantly lower concentrations of serum β 2 -GPI than did those without the mutation, yet no significantly different lipid profiles, such as total cholesterol, triglyceride, HDL-cholesterol, LDL-cholesterol, apoA-I, apoB and Lp(a), were observed. A low concentration of β 2 -GPI seemed not to be associated with apparent abnormality in lipoprotein metabolism.

Anomalous fission source convergence in a Monte Carlo criticality calculation for a weakly coupled array of two fissile material units are demonstrated. Introducing coupling coefficients among array units, it is quantitatively explained that this anomaly is caused by an insufficient restoring force to the true distribution and its large statistical uncertainty, especially, in a symmetric system. A new approach for estimating the fission source intensity ratio in an array is proposed by obtaining the eigenvector of a coupling coefficient matrix. This method also gives the uncertainty of the ratio as well as the ratio, which is available for evaluating the accuracy of the obtained ratio. The correlation between a calculated k eff and the fission source intensity ratio is formulated. It is illustrated theoretically and empirically that there is no significant correlation in a symmetric two-unit array system. In general, care should be taken that a calculated k eg may be biased by an incorrect fission source ...

The neutron source introduction method was applied to absolute measurements of low reactor power at the Static Experiment Critical Facility STACY. To obtain the effective neutron source intensity more accurately, which is a key parameter for the source introduction method, the neutron source is newly defined as fission neutrons from the first fission reaction caused by neutrons emitted from the external neutron source. To obtain the newly defined effective neutron source intensity, the probability that a neutron from the external neutron source causes a fission reaction is calculated using the Monte Carlo code MCNP. This calculation took into consideration the three-dimensional complicated core structures. Furthermore, the fission reaction distribution, fundamental mode forward and adjoint flux distribution in a critical state were calculated using the three-dimensional transport code THREEDANT. Following the principle of the neutron source introduction method, an external neutron source was inserted near...

In order to operate accelerator driven subcritical system safely, it is necessary to monitor the subcriticality in real-time. In this paper, we propose a highly accurate real-time subcriticality-monitoring system by combining the existing multiple measurement techniques, a number of neutron detectors, and computations. Combining the advantages of each technique and averaging among the multiple measurements, the system provides high accurate subcriticality in real-time. This system was examined by subcritical experiments using STAtic experiment Critical facilitY. The experimental results showed that the subcriticalities of the results of this system agreed with the reference values, and it showed the system is useful as a real-time monitor.

Critical experiments on 10% enriched uranyl nitrate solution have been conducted for obtaining benchmark data using a Static Experiment Critical Facility, STACY, in the Nuclear Fuel Cycle Safety Engineering Research Facility, NUCEF. Two cylindrical and one slab tank were used to construct a single core system. Critical solution levels were measured for three core tanks with changing the uranium concentration of uranyl nitrate solution. Fundamental data were obtained for unreflected or water reflected cores. The sensitivities of the fuel composition and tank geometry on the criticality were calculated, and total uncertainty of the experiments are evaluated less than 0.15% Δ k. Calculations of the neutron multiplication factor keff for the critical conditions were performed using a continuous-energy Monte Carlo code MCNP-4B employing JENDL-3.2. This paper presents the core configurations of STACY, their experimental uncertainties and calculation results with above calculation system.

Unexpected deviations of ex-core neutron detector signals were observed during a voyage of the Japanese nuclear ship, Mutsu. From detailed three-dimensional analyses, this phenomenon was determined to be caused by an asymmetrical neutron source distribution in the core due to a small misalignment between the two control rods of a control rod group. A systematic ex-core detector response experiment was performed during the Mutsu's third experimental voyage to gain some understanding of the relationship between the control rod pattern and the detector response characteristics. Results obtained from analyses of the experiment indicate that the Crump-Lee technique, using calculated three-dimensional source distributions for various control rod patterns, provides good agreement between the calculated and measured detector responses. Xenon transient analyses were carried out to generate accurate three-dimensional source distributions for predicting the time-dependent detector response characteristics. Two types of ex-core detector responses are caused by changes in the control rod pattern in the Mutsu reactor: the detector response ratio tends to decrease with the withdrawal of a group of control rods as a pair, and a difference in the positions of the control rods in a group causes signal deviations among the four ex-core detectors. Control rod misalignment does not greatlymore » affect the mean value of the four detector signals, and the deviation can be minimized if the two rods within a group are set at the same elevation at the time of detector calibration.« less

Described in this report are evaluation tests of neutron-induced malfunction of real-time personal dosimeters and electronic modules using the Transient Experiments Criticality Facility, TRACY, at the Japan Atomic Energy Agency. In the event of radiological or nuclear terrorism, radiation levels are expected to be high, especially if nuclear-critical devices are used. In such devices, fission chain reactions develop that may lead to nuclear criticality with strong lethal emissions of both neutrons and gamma-rays. A wide area of radius 400m at least should be cordoned off. In such a situation real-time dosimeters with wireless data transmission capability would provide essential support for first-response teams. The reliability of these apparatuses, however, has not been evaluated sufficiently under such high-dose conditions. We have evaluated these at TRACY, which is a reactor that can realize controlled nuclear excursions using 10% 235U-enriched uranyl nitrate solutions as fuel.

Experimental and computational studies have been performed on the temperature coefficients of reactivity in light-water moderated and reflected UO2 cores with soluble poisons such as boron and gadolinium. Experiments were carried out using the Tank-type Critical Assembly (TCA) in Japan Atomic Energy Research Institute (JAERI). Temperature coefficients of the cores with soluble poisons were measured by changing the temperature of the moderator and reflector from the room temperature to about 60°C. The dependence of temperature coefficients on the core configuration and the concentration of soluble poison was studied with the water level worth method. Temperature coefficients were calculated with a diffusion code CITATION included in the SRAC code system and a perturbation code CIPER for comparison with the experimental data. It was found that the temperature coefficients are always negative in the experimental cores (the water to fuel volume ratio (Vm/Vf) of 1.83) containing boron as soluble poison. On the...

症例は49歳,男性.眼科にてBehcet病の治療中,高熱をきたし近医入院.その後,意識障害が出現し,脳MRIにて右大脳基底核に腫瘤性病変を認め,髄液所見とあわせ神経Behcet病と診断された.ステロイドパルス療法を開始後当科転科.転科時に腸管穿孔がみられ緊急手術となった.全身管理のもとにステロイドパルス療法を追加し,良好な経過をとった.神経Behcet病にはステロイド大量投与を含む迅速な治療が必要であるが,腸管穿孔の合併にも注意すべきと考えられた.

A 42-year-old Japanese woman with systemic lupus erythematosus (SLE) developed Parkinsonian-like movements. Steroid pulse therapy was most effective and additional anti-Parkinsonian drugs were not required. Although psychosis, seizures and meningitis are common central nervous system (CNS) manifestations in SLE patients, Parkinsonian-like symptoms are extremely rare. The putative genesis and treatment of CNS lupus are discussed.

In-situ measurement of photoelectron spectra of polypyrrole during electrochemical undoping/doping cycles has been carried out by using an open-type electrochemical cell. It has been observed that the ionization potential decreases with decreased electrochemical potential. This result seems to be reasonable because the decreased electrochemical potential corresponds to the undoping or recovery of electrons into vacant state of valence band.

Build up of Pu in a co-decontamination process of fuel reprocessing due to shift of process parameters from normal conditions and its effect on nuclear criticality of the system were investigated with computer codes. Being based on the flow sheet of a Purex reprocessing plant, the calculation shows that Pu accumulates mainly in the aqueous phase of extraction stages under the conditions in which concentration of U increases in the organic phase, i.e. a decrease of the extractant flow rate or the TBP concentration, an increase of feed flow rate or U concentration, or a decrease of scrub acidity. Relating the maximum Pu concentration and the time required for maximum build up to the change of the flow sheet conditions, an effective neutron multiplication factor keff of the system involving Pu build up was calculated for the first mixer-settler bank of a model plant. The analysis revealed that the extractor would become critical in approximately 45 h at earliest provided that shift of the parameters stayed within a specified range during operation. Required performance of the process control instruments, safety margin in concentration limit, and other operation safety considerations necessary for the prevention of Pu build up in the extraction bank are discussed.

Build-up of Pu in a partitioning process of Purex fuel reprocessing at the abnormal conditions and its effect on nuclear criticality of the system were investigated with computer codes. Being based on the flow sheet of a model mixer settler process, the calculation shows that a considerable increase of Pu concentration occures solely in the aqueous phase of the 1BX bank under the condition that flow rate of Pu strip solution decreases. The decrease of Pu strip flow causes not only a decrease of the total aqueous flow which consequently concentrates Pu(III) in the aqueous phase, but also an increase of acidity of the aqueous and the organic phases which changes the kinetics of the redox reactions of Pu with U(IV) and nitrous acid. Kinetic data of the reoxidation of Pu(III) with nitrous acid were critically evaluated to be used in calculation.The analysis revealed that the model 1BX bank would become critical in approximately 80h if Pu strip flow would cease. Requirements to assure criticality safety of the partitioning process, e.g. prevention and detection of the cessation of Pu strip flow, detection methods of abnormal distribution profile of Pu or U, were discussed.