Analysis of Core Physics Experiments on Fresh and Irradiated PWR UO2 Fuels in the REBUS Program

Critical experiments of two cores each loaded with fresh 5 × 5 test PWR-type fuel rods of ²³⁵U enrichment of 3.8 wt% or irradiated 5 × 5 test rods of rod average burnup of 55 GWd/t in the REBUS program were analyzed using diffusion, transport, and continuous-energy Monte Carlo calculation codes coupled with nuclear data libraries based on JENDL-3.2 and JENDL-3.3. Biases in effective multiplication factors k _eff's of the critical cores were about ?1:2%Δk for the diffusion calculations (JENDL-3.2), ?0:5%Δk for the transport calculations (JENDL-3.3), and ?0:5 and 0.1%Δk for the Monte Carlo calculations (JENDL-3.3 and JENDL-3.2, respectively). The measured core fission rate and Sc- or Co-activation rate distributions were generally well reproduced using the three types of calculation. The burnup reactivity determined using the measured water level reactivity coefficients was ?2:35 ± 0:07Δk/kk′. The calculated result of the Monte Carlo calculations agreed with it; however, the diffusion and transport calculations overestimated the absolute value by about 7%, which would be mainly attributed to the errors in the calculation of the reactivity caused by changing the fuel compositions from fresh fuel to irradiated fuel.     

Analysis of Measured Isotopic Compositions of High-Burnup BWR MOX Fuel

Analysis of measured isotopic compositions of four high-burnup BWR MOX fuel samples was performed by using a general-purpose neutronic calculation code SRAC and a continuous-energy Monte Carlo burnup code MVP-BURN. The initial Pu fissile content of the samples was 5.52 wt%, and the burnups ranged from 50 to 80 GWd/t. It is confirmed that a geometrical model including the effect of UO₂ assemblies adjacent to the MOX assembly is necessary in the burnup calculations to obtain accurate calculated isotopic compositions. The calculated results of MVP-BURN with JENDL-3.3 taking such effect into account show more accurate results for major actinides (U, Pu, and Am isotopes) and most fission products than those of infinite assembly calculations. The paper also shows the results calculated using SRAC with JENDL-3.3, ENDF/B-VII, and JEFF-3.1.     

Development of drift-flux model based on 8 × 8 BWR rod bundle geometry experiments under prototypic temperature and pressure conditions

The drift-flux model is one of the imperative concepts used to consider the effects of phase coupling on two-phase flow dynamics. Several drift-flux models are available that apply to rod bundle geometries and some of these are implemented in several nuclear safety analysis codes. However, these models are not validated by well-designed prototypic full bundle test data, and therefore, the scalability of these models has not necessarily been verified. The Nuclear Power Engineering Corporation (NUPEC) conducted void fraction measurement tests in Japan with prototypic 8 × 8 BWR (boiling water reactor) rod bundles under prototypic temperature and pressure conditions. Based on these NUPEC data, a new drift-flux model applicable to predicting the void fraction in a rod bundle geometry has been developed. The newly developed drift-flux model is compared with the other existing data such as the two-phase flow test facility (TPTF) data taken at the Japan Atomic Energy Research Institute (JAERI) [currently, Japan Atomic Energy Agency (JAEA)] and low pressure adiabatic 8 × 8 bundle test data taken at Purdue University in the United States. The results of these comparisons show good agreement between the test data and the predictions. The effects of power distribution, spacer grids, and the bundle geometry on the newly developed drift-flux model have been discussed using the NUPEC data.     

Burnup characteristics and fuel cycle economics of mixed uranium–thorium fuel in a simplified small pebble bed reactor

The limitation of natural uranium resources and the improvement of economic values of nuclear reactors are important issues to be solved in the future development of these reactors. In our previous study, we presented an innovative design for simplifying a pebble bed reactor, and the optimization of this design showed that burnup values could be increased and natural uranium uses could be reduced. The purposes of the current study were to design a simplified pebble bed reactor by removing the unloading device from the reactor system and to further optimize the burnup characteristics of this reactor with a peu à peu fuel-loading scheme by introducing thorium in the fuel configuration as a fertile material. Another goal was to optimize the fuel composition so that the system could achieve even better burnup characteristics and use scarce uranium resources more efficiently. Using a specially developed computer code, we analyzed and optimized the performance of a 110-MWt simplified pebble bed reactor using a peu à peu fuel-loading scheme. An optimized design using 30% of fertile thorium mixed with uranium fuel with 15% ²³⁵U enrichment and a 7% packing fraction calculated to achieve a high burnup of 140 GWD/T for more than 21 years' operation time that could save 13 to 33% of natural uranium use compared with the savings noted in our previous study. Neutronic, burnup and fuel economic analysis for this optimized design are discussed in this study.     

Characterization of recrystallization of 12Cr and 15Cr ODS ferritic steels

The recrystallization behavior of 12Cr and 15Cr oxide dispersion-strengthened (ODS) ferritic steels, which are the promising candidate materials for long-life core materials of the advanced fast breeder reactors, was investigated in terms of an intermediate softening heat treatment. It was clarified that keeping recovery structure at the intermediate heat treatment is indispensable for producing recrystallized structure at the final heat treatment. Prevention of repeating recrystallization is owing to the stable {100} 〈110〉 texture formation with less stored strain energy by the cold-rolling of the recrystallized structure. The two-step softening process was proposed to suppress the recrystallization and obtain adequate hardness reduction at the intermediate heat treatment. This process is effective for producing a stable recrystallized structure at the final heat treatment of the manufacturing process of ODS ferritic steel cladding.     

Rewetting front propagation under anticipated operational occurrences for boiling water reactors – development of two-dimensional analytical model

An analytical model to predict a rewetting velocity applicable to high pressure and high flow rate condition during anticipated operational occurrences (AOOs) is developed by applying Wiener–Hopf technique coupled with appropriate kernel substitutions. The model considers the effects of enhanced cooling in the vicinity to liquid film front termed “precursory cooling” and heat input from fuel pellets on back side of wall as boundary conditions of a heat conduction equation. A simplified two-dimensional model neglecting an effect of axial heat conduction is also proposed. It is found through the comparison among the models and experimental data that the contribution of the heat conduction in the wall-depth direction is essential in the prediction of the rewetting velocity at the thermal-hydraulic condition simulating AOOs and the axial heat conduction has little influence when an enhanced heat transfer coefficient in the dried-out region is appropriately given as a function of distance from the liquid film front.     

Experimental study on fuel-discharge behaviour through in-core coolant channels

In core-disruptive accidents of sodium-cooled fast reactors, fuel discharge from the core region reduces the possibility of severe re-criticality events. In-core coolant channels with large hydraulic diameters, such as the control-rod guide tube and a concept of the Fuel Assembly with Inner Duct Structure have a potential to provide effective fuel-discharge paths if effects of sodium in these paths on molten fuel discharge are limited. Two series of experiments were conducted to investigate fuel-discharge behaviour through the sodium-filled channels. In the first series of experiments, an alloy with low melting temperature was ejected into a water channel to clarify dominant phenomena for melt discharge through the coolant-filled channel and to develop methodologies for evaluating the effects of coolant on melt discharge. In the second series of experiments, a molten alumina was discharged through the sodium-filled channel in order to verify the applicability of the knowledge and evaluation methodologies obtained in the first series of experiments to the sodium-filled channel. These series of experiments showed that the discharge path can be entirely voided by the vaporisation of a part of the coolant at the initial melt discharge phase that this is followed by coolant vapour expansion and that melt penetrates significantly into the voided channel. Preliminary extrapolation of the present results to the in-core coolant channel suggests that the effects of the sodium on fuel discharge are limited and, therefore, in-core coolant channels will provide effective fuel-discharge paths for reducing neutronic activity.     

Development of versatile nuclear heating calculation system in MARBLE

A versatile nuclear heating calculation system is developed in MARBLE, a neutronics calculation platform based on the object-oriented language. The system provides a variety of heating calculation methods, from the conventional simple methods based on energy-independent Q-values to a detailed method based on NJOY/HEATR. A particular feature is in hybrid methods that consider energy dependence employed in NJOY/HEATR in the system. The hybrid methods provide tools to assess errors in the simple methods and to evaluate the heat transport property by neutrons and photons precisely. The accuracy of the hybrid methods and the errors in the simple methods are investigated in a Monju design calculation. It is confirmed that the hybrid methods realize the nuclear heating by the detailed method within an accuracy of 0.1%. The neglect of incident energy dependence of the fission Q-value in the simple methods causes clear overestimation in the Q-value but its influence on the total heating is less than 1%. In the heat transport evaluation, it is observed that the heat transported from the fuel regions mainly appears in the blanket regions and the contribution by neutrons is larger than that by photons.     

Intergranular stress corrosion cracking initiation model of austenite stainless steels used in BWRs for optimized water chemistry control

A calculation model on intergranular stress corrosion cracking (IGSCC) initiation time of materials used in boiling water reactors (BWRs) has been developed to evaluate effectiveness of water chemistry control for mitigation of the IGSCC. The model was composed of four terms which determine passive film break time: (1) a chemical term based on electrochemical corrosion potential (ECP) and impurity concentration; (2) a mechanical term based on strain rate; (3) a material term based on sensitization; and (4) an irradiation term based on acceleration of corrosion by γ-rays and neutron irradiation. The contribution of the chemical term in the passive film break was calculated based on a deterministic local corrosion model. Then, the local corrosion model was modified by adding mechanical acceleration of the film rupture to treat the IGSCC phenomenon. The model could reproduce the behavioral tendency seen in the slow strain rate tensile test on high carbon contents with sensitization heat treatment (for example, 620°C × 24 h). Under BWR operating conditions, IGSCC initiation time could be extended by a factor of 5 by lowering the electric conductivity from 1.0 to 0.06 μS/cm. If the ECP was reduced below the critical potential by a mitigation method, the IGSCC initiation time was predicted to become sufficiently long for pipings and components.     

Determining factors for anodic polarization curves of typical structural materials of boiling water reactors in high temperature – high purity water

In order to examine the anodic polarization characteristics of typical structural materials of boiling water reactors (BWRs), the anodic polarization curves of type 316L stainless steel (316L SS) and Alloy 182 were measured in deaerated high purity water at 553 K using the previously reported measurement method which was confirmed suitable for high temperature – high purity water. In order to specify which constituent element determines the dissolution characteristics of these materials, the anodic polarization curves of pure iron, pure nickel, and pure chromium were also surveyed. The anodic polarization curve of 316L SS was determined to have active, passive, and transpassive states which were the same as type 304 SS (304 SS) showed. But, Alloy 182 had different polarization characteristics especially near the corrosion potential as it had no active state. From comparison results of the polarization characteristics of these materials and their constituent elements, the corrosion characteristics of these materials were concluded to be mainly determined by the corrosion characteristics of chromium.