Neutron Behavior in Cluster-Type Fuel Lattices, (II)

A multigroup method of calculation is presented for describing the neutron behavior in a cluster-type fuel lattice. It solves the integral transport equation by a semi-analytical method proposed in a previous paper for calculating collision probabilities in the lattice of a clustered fuel element. Using only fundamental nuclear data, it gives space and energy dependent neutron flux. The method has been programmed for HITAC-5020F (computer code named CLUSTER-III).

The accuracy of the method has been tested by comparing the calculation with the experiment described in Part (I) of this paper. The lattices are 28-pin clusters of UO₂ or PuO²+UO² fuel pins, with heavy-or light-water moderators and with light-water coolant containing varying void ratios. The quantities studied are micro-parameters, reaction distributions in energy and space, thermal disadvantage factors and the multiplication factors. It is found that the calculated results are generally in good agreement with experiment, typically within 10% for micro-parameters and thermal disadvantage factor, and within 1% for the multiplication factor.     

Lattice Parameter Measurements on Cluster-Type Fuel for Advanced Thermal Reactor

Abstract

The behavior of neutrons in a highly heterogeneous unit cell consisting of D₂O moderator, H₂O coolant and a 28-pin fuel cluster contained in a pressure tube has been studied through lattice parameter measurements covering three different ²³⁵U enrichments, four coolant void fractions and two lattice pitches. A single-region core configuration was adopted, with which measurements were made to determine—in relation to coolant void fraction—the critical D₂O level, as well as various lattice parameters

A strong dependence on coolant void was observed for the critical level and the lattice parameters, in the case of the smaller 22.5 cm pitch lattice, due to the positive effect on core reactivity exerted by the slowing-down faculty of H₂O in the epithermal energy region. With the larger 25.0 cm pitch, however, no meaningful dependence on void fraction was shown by any of the measured values, and this was ascribed to a compensating negative effect due to enhanced thermal neutron self-shielding in the fuel region produced by the H₂O coolant.

The results of cell calculations obtained by means of the METHUSELAH-II code showed generally good agreement with experimentally determined data, for both critical D₂O levels and lattice parameters, in the case of coolant-filled lattices (0, 30 and 70% void fractions). For lattices devoid of coolant (100% void fraction), however, discrepancies in lattice parameters—particularly in p ²⁸—produced corresponding deviations in core reactivity amounting to 1% in excess of those incurred with other void fractions.     

Neutron Behavior in Cluster-Type Fuel Lattices, (I)

Lattice parameters φ²⁸, φρ²⁵, ρ²⁸ and C^* were measured on cluster-type fuel lattices of the ATR (Advanced Thermal Reactor) by using a two region critical facility (D₂0-cluster test and H₂O-rod driver regions). Their dependence on lattice pitch, coolant-void ratio and fuel composition (whether UO₂ or PuO₂-UO₂) have been made clear by this experiment.

A foil handling technique has been developed for determining the lattice parameters of the Pu0₂- UO₂ fuel pins, and the resulting measurement errors are almost as small as those obtained on the U0₂ fuel pins.

The effects of the Cd-filter and of the presence of UO₂ buttons in the measurement of ρ²⁸ and χ²⁵ were studied experimentally and correction factors have been determined.

A method of observing the spatial distribution of the γ-ray source in an activated foil has been developed, and the relation between the spatial distribution and the coincidence counting efficiency of the foil has been examined.     

Fuel Temperature Coefficients of Reactivity for Heavy-Water-Moderated,Cluster-Type Fuel Lattices

Temperature coefficients of reactivity have been measured up to 600°C on cluster-type UO₂ fuel for three kinds of ²³⁵U enrichment and on a hollow cluster of sus-cladding tubes by using a hot He gas loop in a heavy-water-moderated, pressure-tube-type critical assembly. A new experimental method has been developed which accurately eliminates the reactivity disturbance caused by heat leakage in the measurement of an extremely small change in reactivity. The fuel (fuel pellet, cladding land pressure-tube) temperature coefficients of reactivity obtained for the temperature range below 300°C are +1.00±0.04, ?3.48±0.13 and - 6.36±0.25 in the unit of l0^-5% Δk/k.°C for 0.2%, 0.7% and 1.5%²³⁵U enrichment, respectively. In the higher temperature region above 300°C, each coefficient shifts to positive side by about 2x10^-5 Δk/k.°C. Temperature coefficient of reactivity for the hollow cluster of sus-cladding tubes (cladding and pressure-tube) has a large constant value with positive sign, + (6.42±0.26) x 10^-5 Δk/k.°C, all through the temperature range. A calculational model to analyze a hot-loop-type measurement of temperature coefficients with use of WIMS-D code was proposed and could be successfully applied to the present measurement.     

Measurement of Local Power Peaking Factors in Heavy-Water Moderated Plutonium Lattices

Local power peaking factors (LPF) in the heavy-water moderated plutonium lattices were measured by a new method of γ-scanning of fuel pins using the calculated power correction factor of which accuracy was evaluated with the aid of foil activation method. Accuracy in measurement of the LPF was evaluated to be within 1%.

By this measurement, behaviors of the LPF have been made clear concerning the differences in fuel materials, coolant materials and arrangement in fuel enrichments. Depression of the thermal power in the fuel cluster makes LPF in the plutonium fuel lattice larger than in the uranium lattice. This tendency is more remarkable in air coolant lattice than in H₂O coolant lattice. The value of LPF for the plutonium fuel cluster of different enrichments is smaller than that of a uniformly enriched fuel cluster. The reduction of LPF is smaller in H₂O coolant than in air coolant lattice.

The values of LPF by WIMS-D code based on the transport theory and by METHUSELAH-II code based on the diffusion theory are in agreement with the measured ones, within 1.5 and 2.4% respectively.     

熔盐实验堆中燃料核素的灵敏度系数计算分析

以熔盐实验堆为模型,采用MCNP5和SCALE5.1中的TSUNAMI-3D-K5对燃料核素的灵敏度系数进行计算与分析。结果表明,灵敏度系数与核素在MSRE中的含量、位置和核素的中子反应截面有关,得到灵敏度系数最大的核素²³⁵U的宏观裂变截面和宏观俘获截面的灵敏度系数分别为0.267和0.110。MCNP5和TSUNAMI-3D-K5计算不同能区下²³²Th宏观总截面和俘获截面的灵敏度系数曲线一致,曲线在0.1 eV附近有一小峰,振荡区域同截面共振区范围相同。     

Lattice Parameters of Mixed Oxides in Sm-U-O System

The oxidation-reduction properties of mixed oxides in the system Sm-U-O and the variation of their lattice parameters due to oxidation and reduction were examined by means of thermogravimetry and X-ray diffraction. For mixed oxides Sm_yU_1-yO_2±x prepared in air, the O/M ratio decreased linearly from 2.667 at y= 0 to nearly stoichiometric 2.02 at y=0.5. The 0/M ratios at y>0.5 were on the line connecting 2.667 at y= 0 with 1.50 at y =1. The single phase region of the fluorite structure was between y = 0.5 and 0.6. The lattice parameters of the fluorite phase at ≤0.6 were constant, 5.418×10^?10m. For mixed oxides prepared in hydrogen, the 0/M ratio decreased linearly from 2.00 at y = 0 to 1.84 at y = 0.5, and from 1.84 at y = 0.5 to 1.50 at y = 1. The quenched oxides were single phase compounds except those for y = 0.1, 0.6 and 0.9. The lattice parameter of the fluorite structure decreased from 5.4700 at y = 0 to 5.427 × 10^?10 m of Smo.₄U0.6O_1.854. and was kept constant between y = 0.4 and 0.5, then increased to 5.466 × m at y = 0.7. The Sm-U-O phase diagram at 1,000°C was constructed on the basis of the X-ray diffraction work.     

Experimental Determination of Spectral Indices in Low Enriched UO2-H2O Lattices

To get information about the neutron spectrum in low enriched UO₂-H₂O lattices, the spectral indices SI(U^8c/Dy) and SI(U^8c/U^5f) were measured on the basis of the parallel irradiation technique, which basically irradiates activation foils both in a neutron field to be investigated and in a reference field of thermal neutrons. In the present study, a fuel pellet of UO₂ was used for the measurement of activities caused by the neutron capture of ²³⁸U and the fission of ²³⁵U. Besides the technical details of the measurements, the origins of experimental errors are listed with the method how to eliminate them. The measurements were carried out in lattices of different fuel enrichment to demonstrate the capability of the present method, and the experimental results were compared with the calculated ones. It was found that the results of the present measurements are useful to assess the validity of the cell calculations.     

压水堆燃料棒尺寸及其富集度对k值影响的MCNP模拟

文章以大亚湾核电站压水堆核电机组为原型,根据程序要求建立起燃料棒的几何与核物理模型,进行MCNP程序模拟计算,并不断调整模型以减小计算方差,得到两组不同D与ε下的keff值并分别作图。实验结果表明:D与keff为类抛物线函数关系,ε与keff为对数函数关系。     

Accuracy of Cell Calculation Methods Used for Analysis of High Conversion Light Water Reactor Lattice

Validation tests were made for the accuracy of cell calculation methods used in analyses of tight lattices of a mixed-oxide (MOX) fuel core in a high conversion light water reactor (HCLWR). A series of cell calculations was carried out for the lattices referred from an international HCLWR benchmark comparison, with emphasis placed on the resonance calculation methods; the NR, IR approximations, the collision probability method with ultra-fine energy group. Verification was also performed for the geometrical modelling; a hexagonal/cylindrical cell, and the boundary condition; mirror/white reflection. In the calculations, important reactor physics parameters, such as the neutron multiplication factor, the conversion ratio and the void coefficient, were evaluated using the above methods for various HCLWR lattices with different moderator to fuel volume ratios, fuel materials and fissile plutonium enrichments.

The calculated results were compared with each other, and the accuracy and applicability of each method were clarified by comparison with continuous energy Monte Carlo calculations. It was verified that the accuracy of the IR approximation became worse when the neutron spectrum became harder. It was also concluded that the cylindrical cell model with the white boundary condition was not so suitable for MOX fuelled lattices, as for UO₂ fuelled lattices.     

Investigation on Criticality and Infinite Multiplication Factor of High Conversion Light Water Reactor Using Zone-Type FCA-HCLWR Core Fueled with Enriched Uranium

—A series of reactor physics experiments have been carried out at the FCA to examine the availability of the nuclear data and computational method currently employed to evaluate the nuclear characteristics of the High Conversion Light Water Reactor. Experimental results of the effective and infinite multiplication factors k_eff and k_∞ are compared with the calculated ones for three zone-type FCA-HCLWR cores fueled with enriched uranium. The calculated k_eff and k_∞ values with use of the SRAC system and the cross section set based on the JENDL-2 data file show a good agreement with the measured ones. The calculated-to-experimental (C/E) values for k_eff and k_∞ do not depend on the cell parameters such as the fuel enrichment, the moderator voidage state and the moderator-to-fuel volume ratio, and these values are similar with each other There is also no inconsistency between the C/E values for k_eff and k_∞ : The average C/E values are 989 and 0 988 for k_eff and k_∞ respectively     

Anisotropic Diffusion Effect on Criticality of Plate Lattice Fast Assembly

The neutron diffusion in plate lattice is generally somewhat anisotropic. In case of usual plate cells for the mockup of LMFBR composition, the diffusion coefficient for parallel direction to lattice plate, based on Benoist's theory, proves to be larger by 2～4% than that for perpendicular direction, which is considered to affect the criticality of plate lattice fast assembly.

A practical treatment of the anisotropic diffusion effect on criticality has been proposed, in which, like the transport correction, the anisotropic diffusion effect is treated as a correction term to be applied to the conventional isotropic diffusion calculation. The method is applied to actual plate lattice critical assemblies, already built in FCA, ZPR or ZEBRA. The anisotropy correction on criticality turns out to amount to the order of -0.2～-0.4%Δ k/k for the normal plate lattice-core. The amount of anisotropy correction is further enhanced in case of an assembly consisting of plate lattice-blanket or sodium-voided core. The anisotropic diffusion effect on criticality is, therefore, important for the analysis of criticality of plate lattice assembly, and should be corrected in addition to the conventional heterogeneity effect. The present method, based on the perturbation theory, is practical and useful.     

Proposal of Method to Estimate Criticality Correction for Anisotropic Diffusion in Plate Lattice Fast Assembly

Measurements on the neutron and γ-ray attenuation in multi-legged air filled ducts have been carried out using a Cockroft-Walton type neutron generator. The measured spectra were obtained with an NE-213 liquid scintillator using pulse shape discrimination to resolve neutron and γ-ray pulse height data and using a spectral unfolding code to convert these data to energy spectra. Results are given for two rectangular duct geometries. A series of integral measurements have also been carried out with a long counter for fast neutrons and a radiation survey meter to measure the γ-ray dose. The biological dose arising due to the streaming of neutrons was measured with a rem counter. The results show the relative decrement of the integral counts along the axis of the duct. The spectral measurements reveal the presence of resonant windows of oxygen present in the concrete medium. Also the 14 MeV component is found to be dominant in the first leg of the duct whereas degraded neutrons dominate in the second leg. The results are compared with MORSE calculations.     

Reactivity Control Method for Extended Burnup of Boiling Water Reactor Fuel Bundles

Abstract

A reactivity control method was proposed for a boiling water reactor (BWR) fuel bundle, which has a potential for higher burnup with an increase in fuel enrichment. The new method optimized the distribution and amount of nonboiling water area in a fuel bundle in order to enhance the reactivity control capacity.

Using the method, a 9×9 lattice fuel bundle with a small-sized channel box, large-sized water rods and a reduced fuel rod diameter was proposed for the discharged burnup of 70 GWd/t and the operational cycle length of 18 months. The core, which consists of the proposed fuel bundles with the bundle-averaged enrichment of 5.8% and includes other modifications concerning a neutron low leakage loading pattern, natural uranium axial blankets, and spectral shift with recirculation flow control, has a cold shutdown margin greater than the design limit (1%Δk) with minimum fuel bundle shuffling. Further, it has potentials for natural uranium savings of about 20% per unit power and reduction in the amount of reprocessing of about 60% per unit power, compared with current BWR designs.     

Computation on Fuel Particle Size Capable of Being Regarded as Homogeneous in Nuclear Criticality Safety Analysis

For powdered fuel processed in the nuclear fuel facilities, flooding is often thought to be the severest condition regarding the nuclear criticality safety evaluation. The reactivity of such a heterogeneous system as powdered fuel in water should be almost equal to that of the homogeneous one, when the fuel particle size is very small. The neutron multiplication factor was calculated for an infinite cubic array of slightly enriched UO₂ sphere particles immersed in water with various enrichments, water to fuel ratios and fuel particle sizes. The calculations were performed with a computer code module based on the collision probability method to solve the ultra-fine energy group equations of neutrons. The change in the neutron multiplication factor from the homogeneous system is dominated first by the change in the resonance escape probability and second by the change in the thermal utilization factor; these changes and therefore their sum, depend almost completely on the mean uranium concentration (or water to fuel volume ratio) and rarely on uranium enrichment up to 10 wt% for a fuel particle size of 1mm. The dependence determines the fuel particle size regarded as homogeneous in proportion to the negligible relative error of the neutron multiplication factors.