UO2 Core Research Articles

Re-evaluation of the Effective Delayed Neutron Fraction Measured by the Substitution Technique for a Light Water Moderated Low-enriched Uranium Core

The effective delayed neutron fraction βeff for a light water moderated low-enriched UO2 core has been re-evaluated to obtain benchmark data for the validation of calculation codes and nuclear data. Originally, the βeff value was measured by the substitution method. In that method, the βeff value was obtained from measured reactivity change by substituting a Sb-Cd-Pb absorber rod for a 2.6 wt% UO2 rod for all core regions. In the present evaluation, we have employed the latest value for the buckling coefficient of reactivity to re-evaluate the substitution reactivity with high accuracy. In addition, the correction factor, which was ignored in the previous measurement, has been calculated to compensate the difference in the absorption cross sections of fuel and absorber rods. Consequently, the obtained βeff value in the present evaluation was 0.00771±0.00017, and it is more credible than the previous one. The present result is available as benchmark data for the verification of delayed neutron data for light water reactors. For comparison, we have calculated the βeff value using a transport code TWODANT with the JENDL-3.2 nuclear data library. The calculated βeff value overestimated the experiments; the difference slightly exceeded the experimental error.

High energy X-ray microscopy for characterisation of fuel particles

For the first time different high energy microanalysis techniques were combined to characterise individual micrometer sized radioactive particles. It was shown that particle characteristics including weathering rates and mobilisation of associated radionuclides are source specific and release-scenario dependent. Fuel particles released during the explosion are characterised by UO 2-cores with surrounding layer of reduced U with low weathering rates. In contrast, fuel particles released during the subsequent fire show UO 2-core with surrounding layers of oxidised U 2O 5/U 3O 8 with high weathering rates

Re-evaluation of the Effective Delayed Neutron Fraction Measured by the Substitution Technique for a Light Water Moderated Low-enriched Uranium Core.

The effective delayed neutron fraction βeff for a light water moderated low-enriched UO2 core has been re-evaluated to obtain benchmark data for the validation of calculation codes and nuclear data. Originally, the βeff value was measured by the substitution method. In that method, the βeff value was obtained from measured reactivity change by substituting a Sb-Cd-Pb absorber rod for a 2.6 wt% UO2 rod for all core regions. In the present evaluation, we have employed the latest value for the buckling coefficient of reactivity to re-evaluate the substitution reactivity with high accuracy. In addition, the correction factor, which was ignored in the previous measurement, has been calculated to compensate the difference in the absorption cross sections of fuel and absorber rods. Consequently, the obtained βeff value in the present evaluation was 0.00771±0.00017, and it is more credible than the previous one. The present result is available as benchmark data for the verification of delayed neutron data for light water reactors. For comparison, we have calculated the βeff value using a transport code TWODANT with the JENDL-3.2 nuclear data library. The calculated βeff value overestimated the experiments; the difference slightly exceeded the experimental error.

Analysis of Mixed Oxide Fuel Critical Experiments with Neutronics ANalysis Codes for Boiling Water Reactors

Critical experiments of UO2 and full mixed oxide (MOX) fuel cores conducted at the Tank-type Critical Assembly (TCA) were aNalyzed using BWR design-purpose codes HINES and CERES with ENDF/B files and Monte Carlo fine analysis codes VMONT and MVP with the JENDL-3.2 library. The averaged values of the multiplication factors calculated with HINES/CERES, VMONT and MVP agreed with those of experiments within 0.3%ΔAk. The values by the design-purpose codes showed a small difference of 0.1%Δk between UO2 and MOX cores. Monte Carlo code results showed that the JENDL-3.2 library had a tendency to overestimate the multiplication factors of UO2 cores by about 0.3%Δk compared with those values of MOX cores. The root mean square errors of calculated power distributions were less than 1% for HINES/CERES and VMONT. These results showed that (1) the accuracy of these codes when applied to full MOX cores was almost the same as their accuracy for UO2 cores, which confirmed the accuracy of present core design codes for full MOX cores; and (2) the accuracy of the 190-energy-group Monte Carlo calculation code VMONT was almost the same as that of the continuous-energy Monte Carlo calculation code MVP.

Analysis of Mixed Oxide Fuel Critical Experiments with Neutronics Analysis Codes for Boiling Water Reactors.

Critical experiments of UO2 and full mixed oxide (MOX) fuel cores conducted at the Tank-type Critical Assembly (TCA) were aNalyzed using BWR design-purpose codes HINES and CERES with ENDF/B files and Monte Carlo fine analysis codes VMONT and MVP with the JENDL-3.2 library. The averaged values of the multiplication factors calculated with HINES/CERES, VMONT and MVP agreed with those of experiments within 0.3%ΔAk. The values by the design-purpose codes showed a small difference of 0.1%Δk between UO2 and MOX cores. Monte Carlo code results showed that the JENDL-3.2 library had a tendency to overestimate the multiplication factors of UO2 cores by about 0.3%Δk compared with those values of MOX cores. The root mean square errors of calculated power distributions were less than 1% for HINES/CERES and VMONT. These results showed that (1) the accuracy of these codes when applied to full MOX cores was almost the same as their accuracy for UO2 cores, which confirmed the accuracy of present core design codes for full MOX cores; and (2) the accuracy of the 190-energy-group Monte Carlo calculation code VMONT was almost the same as that of the continuous-energy Monte Carlo calculation code MVP.

Nuclear Design Study on a Super High-burnup Fuel Assembly for Pressurized Water Reactors Using Transuranium

A conceptual design study was carried out on a super high-burnup mixed-oxide (MOX) fuel assembly (SHB FA) for pressurized water reactors (PWRs) using transuranium (TRU). This study aims to avoid the surplus plutonium (Pu) accumulation and to reduce the accumulation of long-lived radioactive minor actinides (MAS) by utilizing the currently existing PWRs under the condition that the Japanese program to develop fast breeder reactors (FBRs) is tend to delay. For this purpose, an SHB FA with discharged burnup of −80 GWd/t was investigated by utilizing MAS positively as both burnable absorbers and fissile suppliers and loading high-content Pu. It is possible to load the SHB FAs in a current PWR together with UO2 FAs and to use 2.5 times as much amount of Pu as that in a standard 1/3 MOX core. Moreover, it is found to be possible to reduce the total number of fresh FAs further from that of a high-burnup (55 GWd/t in maximum) UO2 (4.9 wt%) core and also to reduce the accumulation of MAS in the nuclear fuel cycle significantly.

Nuclear Design Study on a Super High-burnup Fuel Assembly for Pressurized Water Reactors Using Transuranium.

A conceptual design study was carried out on a super high-burnup mixed-oxide (MOX) fuel assembly (SHB FA) for pressurized water reactors (PWRs) using transuranium (TRU). This study aims to avoid the surplus plutonium (Pu) accumulation and to reduce the accumulation of long-lived radioactive minor actinides (MAS) by utilizing the currently existing PWRs under the condition that the Japanese program to develop fast breeder reactors (FBRs) is tend to delay. For this purpose, an SHB FA with discharged burnup of −80 GWd/t was investigated by utilizing MAS positively as both burnable absorbers and fissile suppliers and loading high-content Pu. It is possible to load the SHB FAs in a current PWR together with UO2 FAs and to use 2.5 times as much amount of Pu as that in a standard 1/3 MOX core. Moreover, it is found to be possible to reduce the total number of fresh FAs further from that of a high-burnup (55 GWd/t in maximum) UO2 (4.9 wt%) core and also to reduce the accumulation of MAS in the nuclear fuel cycle ...

96/03844 Viscosity of coal-derived liquids. 1. A group contribution method for pure model compound liquids

96/03844 Viscosity of coal-derived liquids. 1. A group contribution method for pure model compound liquids

Evaluation of Critical Bucklings of Light-Water Moderated Low Enriched UO2Cores by the Variable Loading Method

Abstract Horizontal extrapolated distances Xh and critical bucklings B2 C for light-water moderated low enriched UO2 cores were evaluated using the variable loading method. A set of critical height (water levels) of the cores having various horizontal dimensions was used in this method. Errors of the present method were smaller than the previous ones by the flux shape method in which the horizontal extrapolated distance was determined from the horizontal power distribution. The presently obtained Xh and B2 C reproduced well the change in the critical water level for the whole range of the horizontal core size, while the previous ones could be used only for a limited range. Moreover, the effective multiplication factors obtained through the cell calculation of the SRAC code by using the presently evaluated critical bucklings agreed much better with the rigorous ones by the continuous energy Monte Carlo calculation using a full core model, than those by using the previous values. Therefore, it was concluded that the variable loading method can give more accurate values for Xh and B2 C than the flux shape method for the cores investigated.

Evaluation of Critical Bucklings of Light-Water Moderated Low Enriched UO2 Cores by the Variable Loading Method.

Abstract Horizontal extrapolated distances Xh and critical bucklings B2 C for light-water moderated low enriched UO2 cores were evaluated using the variable loading method. A set of critical height (water levels) of the cores having various horizontal dimensions was used in this method. Errors of the present method were smaller than the previous ones by the flux shape method in which the horizontal extrapolated distance was determined from the horizontal power distribution. The presently obtained Xh and B2 C reproduced well the change in the critical water level for the whole range of the horizontal core size, while the previous ones could be used only for a limited range. Moreover, the effective multiplication factors obtained through the cell calculation of the SRAC code by using the presently evaluated critical bucklings agreed much better with the rigorous ones by the continuous energy Monte Carlo calculation using a full core model, than those by using the previous values. Therefore, it was concluded that the variable loading method can give more accurate values for Xh and B2 C than the flux shape method for the cores investigated.

Temperature Effects on Reactivity in Light Water Moderated U02Cores with Soluble Poisons

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...

Temperature Effects on Reactivity in Light Water Moderated UO2 Cores with Soluble Poisons.

Temperature Effects on Reactivity in Light Water Moderated UO2 Cores with Soluble Poisons.

Contact conductivity between a UO2 core and cladding

Contact conductivity between a UO2 core and cladding

Effect of temperature distribution on the swelling of UO2 and UO2-PuO2 cores

Effect of temperature distribution on the swelling of UO2 and UO2-PuO2 cores

Heat conductivity between UO2 core and casing in fast reactor fuel elements

1. The coefficient of heat conductivity between the fuel and OKh16N15M3B steel casing in fuel elements are determined from the changes of the structure of UO2 core in conditions of operation of fuel elements in fast reactors. 2. For the temperature of the casing equal to 500°C and linear thermal power equal to 550 W/cm and higher (medium-helium) the heat conductivity coefficient is ∼0.5 W/cm2·°C for relative diametral gap δ/D=0.008; with the increase of the diametral gap to 0.052 α decreases by a factor of 2. For δ/D≥0.05 the local value of α may be 1.5–2 times smaller than the average value because of the eccentric location of the fuel pellet in the casing. 3. In fuel elements filled with argon during assembly α is 1.2–2 times lower than in fuel elements with helium filling for identical δ/D and other conditions remaining unchanged. This difference is more significant for larger values of δ/D, which indicates a mixed mechanism of heat transfer from the fuel to the casing, i. e. through the solid contact and the gaseous layer. 4. It is found from the size of the melting zone that after attaining its maximum value the temperature at the center of the core decreases in the process of single emergence into power and during irradiation for 3–30 hr at constant power. The decrease is apparently due to the increase of heat conductivity of the fuel as a result of condensation and due to an increase in α. 5. It is found that the diametral gap, which had a relative value of 0.052–0.087 before irradiation, was filled by melted UO2 in some segments of the fuel elements. The casing was not damaged during its contact with UO2.

Kinetics of the Graphite‐Uranium Dioxide Reaction from 1400° to 1756°C

The reaction of microspheres of UOz with graphite was studied from 1400° to 1756°C. When a spherically symmetrical layer of carbide was produced around the UO2 core, only UC2 was formed, and the diffusion of oxygen through this layer was rate‐controlling. The Arrhenius relation for this system is kD= 21.0 exp(−90, 000/RT) cm2/sThe reaction of a geometrically nonsymmetrical configuration of UO2 and UC2 was also studied. Comparison of the reactions in the symmetrical and nonsymmetrical systems demonstrated that the kinetic behavior of the two systems is quite different and that the conversion in the nonsymmetrical system was 2 to 5 times faster. The importance of these observations to kinetic results reported in the literature for analogous systems is discussed.

Dynamics of an Adiabatic Point Reactor

The problem of adiabatic excursions in a reactor is studied in general. We let the prompt temperature reactivity feedback be an unspecified function of temperature, ρ = ρ0 = ρ0 + f(T), where ρ is total reactivity, ρ0 initial step reactivity and f(T) the feedback function. The similarity of the behavior of the reactor for different f(T) is established by means of a topological (qualitative) analysis. A quantitative asymptotic solution of the non-linear system of DE describing the reactor is presented. In delayed critical excursions, the delayed neutrons play a determining role. In the first part of a prompt excursion, the delayed-neutron source is nil; however this is not so in the second part, where it contributes appreciably to the excursion. These conclusions are shown to be valid in general, and allow us to write down almost directly the (approximate) quantitative solution of the non-linear system for any f(T). These results are correlated with the experimental data for the adiabatic excursions of a UO2 core in SPERT I; in this case the (prompt) dependence of the reactivity on energy is of the form ρ = ρ0 - 4.588 × 10-4E0.74.

Self-Limiting Power Excursion Tests of a Water Moderated Low-Enrichment UO2 Core

"Self-Limiting Power Excursion Tests of a Water Moderated Low-Enrichment UO2 Core." Nuclear Science and Engineering, 15(4), p. 482