MONJU Reactor Research Articles

Friction factor and Reynolds number correlation for finned tube bundle of the air cooler of Monju reactor

Applicability of the analysis model of the cooling ability under a high temperature and low flowrate condition should be improved to evaluate the plant safety in case of the severe accident due to long-term station blackout.The present study reviews experiments on pressure drop behavior for complicated tube bundle geometry to the flow path and then develops a new correlation equation based on a computational fluid dynamics analysis of the Monju air cooler reflecting the actual geometry and plant data.After a basic simulation model being developed for a typical pressure drop experiment, the simulation is applied to the Monju air cooler that has a finned tube bundle. The obtained relationship between friction factors and Reynolds numbers ranging from 10 to 10,000 are fitted to a power function to derive a correlation equation of the fin tube bundle friction factor. The derived correlation equation can estimate pressure loss in the finned tube bundle more precisely than that used in the Monju design. It is applicable to the future reactor design of the air cooler, especially when the cooling ability in low Reynolds number is requested.

Uncertainty Evaluation of Anticipated Transient without Scram Plant Response in the Monju Reactor Considering Reactivity Coefficients within the Design Range

Uncertainty Evaluation of Anticipated Transient without Scram Plant Response in the Monju Reactor Considering Reactivity Coefficients within the Design Range

Numerical study on the thermal stratification characteristics in the upper plenum of sodium-cooled fast reactor (SFR)

In this paper, the IAEA benchmark, a trip transient scenario from 40% rated power of Monju reactor in Japan, was studied using CFD method. The reactor steady state and shutdown transient scenarios were simulated and the 3-D distributions of important thermal hydraulic parameters in the SFR upper plenum was achieved. The steady state calculated results agreed well with experimental data and acceptable deviations were obtained, which indicate the application and prediction accuracy of CFD method. The complex development of thermal stratification in the upper plenum of SFR under accident conditions were obtained. The detailed flow fields in the bottom of center measuring column and near with the through flow holes were revealed. The influence of different local structures on the thermal stratification in the SFR upper plenum were understood. This work could provide a certain basis for the SFR upper plenum structure design and operation strategy making.

Application of a 3D model in the transient system analysis of sodium-cooled fast reactor

In this study, the application of a 3D model for the sodium pool of the sodium-cooled fast reactor was realized. The basic equations and numerical solution methods of the model were given in detail. The model was verified by calculating the 3D temperature field of the region above the MONJU reactor core after the reactor was shut down and the pump stopped. Subsequently, the 3D sodium pool model was used to simulate and analyse the Chinese Experimental Fast Reactor (CEFR). The 3D temperature fields of the inner hot pool under the accidents of the control rod uncontrolled withdrawal and one primary pump shaft seizure at a 102.5% rated power were simulated to analyse the axial position where the thermal fatigue of the casing material was most likely to occur. The results showed that in the height range of 0.3–0.5 m of the enclosure, the thermal stratification phenomenon was the greatest, and the material was most prone to thermal fatigue.

Minor Actinide Transmutation in Supercritical-CO2-Cooled and Sodium-Cooled Fast Reactors with Low Burnup Reactivity Swings

This paper presents the investigation of minor actinide (MA) transmutation in supercritical CO2-cooled and sodium-cooled fast reactors (S-CO2-FR and SFR) with the thermal output of 600 MW(thermal) for simultaneously attaining low burnup reactivity swings and reducing long-life radioactive waste. Minor actinides are loaded uniformly in the fuel of the cores, and the MA contents are determined to minimize the burnup reactivity swings. In the S-CO2-FR, the burnup reactivity swing is minimized to 0.11% ∆k/kk’ when the MA content is 6.0 wt%. In the SFR, the MA content was determined to reduce the burnup reactivity swing while maintaining sodium void reactivity under a design limitation of 5 $. The burnup reactivity swing of the SFR is reduced to 1.94% ∆k/kk’, whereas sodium void reactivity is about 4.7 $ when 10.0 wt% MAs are loaded. The low burnup reactivity swing enables minimization of control rod operation during fuel burnup. The number of control rods in the two reactors is reduced to ten, which is half of a typical sodium-cooled mixed-oxide fuel MONJU reactor without MA loading. The MA transmutation rates in the S-CO2-FR and SFR are 42.2 and 52.2 kg/year, respectively, which are equivalent to the production rates in seven and nine light water reactors of the same electrical output.

Numerical modeling of sodium fire – Part II: Pool combustion and combined spray and pool combustion

The risk of sodium-air reaction has received considerable attention after the sodium-fire accident in Monju reactor. The fires resulting from the sodium-air reaction can be detrimental to the safety of a sodium fast reactor. Therefore, predicting the consequences of a sodium fire is important from a safety point of view. A computational method based on CFD is proposed here to simulate sodium pool fire and understand its characteristics. The method solves the Favre-averaged Navier-Stokes equation and uses a non-premixed mixture fraction based combustion model. The mass transfer of sodium vapor from the pool surface to the flame is obtained using a sodium evaporation model. The proposed method is then validated against well-known sodium pool experiments of Newman and Payne. The flame temperature and location predicted by the model are in good agreement with experiments. Furthermore, the trends of the mean burning rate with initial pool temperature and oxygen concentration are captured well. Additionally, parametric studies have been performed to understand the effects of pool diameter and initial air temperature on the mean burning rate.Furthermore, the sodium spray and sodium pool combustion models are combined to simulate simultaneous spray and pool combustion. Simulations were performed to demonstrate that the combined code could be applied to simulate this. Once sufficiently validated, the present code can be used for safety evaluation of a sodium fast reactor.

Generating Unstructured Nuclear Reactor Core Meshes in Parallel

Recent advances in supercomputers and parallel solver techniques have enabled users to run large simulations problems using millions of processors. Techniques for multiphysics nuclear reactor core simulations are under active development in several countries. Most of these techniques require large unstructured meshes that can be hard to generate in a standalone desktop computers because of high memory requirements, limited processing power, and other complexities. We have previously reported on a hierarchical lattice-based approach for generating reactor core meshes. Here, we describe efforts to exploit coarse-grained parallelism during reactor assembly and reactor core mesh generation processes. We highlight several reactor core examples including a very high temperature reactor, a full-core model of the Korean MONJU reactor, a ¼ pressurized water reactor core, the fast reactor Experimental Breeder Reactor-II core with a XX09 assembly, and an advanced breeder test reactor core. The times required to generate large mesh models, along with speedups obtained from running these problems in parallel, are reported. A graphical user interface to the tools described here has also been developed.

NUMERICAL ANALYSIS OF THERMAL STRATIFICATION IN THE UPPER PLENUM OF THE MONJU FAST REACTOR

A numerical analysis of thermal stratification in the upper plenum of the MONJU fast breeder reactor was performed. Calculations were performed for a 1/6 simplified model of the MONJU reactor using the commercial code, CFX-13. To better resolve the geometrically complex upper core structure of the MONJU reactor, the porous media approach was adopted for the simulation. First, a steady state solution was obtained and the transient solutions were then obtained for the turbine trip test conducted in December 1995. The time dependent inlet conditions for the mass flow rate and temperature were provided by JAEA. Good agreement with the experimental data was observed for steady state solution. The numerical solution of the transient analysis shows the formation of thermal stratification within the upper plenum of the reactor vessel during the turbine trip test. The temporal variations of temperature were predicted accurately by the present method in the initial rapid coastdown period (~300 seconds). However, transient numerical solutions show a faster thermal mixing than that observed in the experiment after the initial coastdown period. A nearly homogenization of the temperature field in the upper plenum is predicted after about 900 seconds, which is a much shorter-term thermal stratification than the experimental data indicates. This discrepancy may be due to the shortcoming of the turbulence models available in the CFX-13 code for a natural convection flow with thermal stratification.

Analysis of the Natural Convection Flow in the Upper Plenum of the MONJU Reactor with Trio_U

The IAEA has coordinated a benchmark project on natural convection phenomena in the upper plenum of the MONJU reactor. JAEA has provided both detailed geometrical data of the plant and complete thermalhydraulic boundary conditions describing a pump trip transient, accomplished during the start-up experiments of the reactor. For the initial conditions of the pump trip transient, extensive sensitivity analyses have been made with the CFD code Trio_U. These calculations show a high sensitivity of the global flow pattern in the MONJU upper plenum depending on the initial order of the numerical scheme and the modelling of the geometrically complex upper core structure. During the pump trip, the formation of a thermal stratification within the plenum has been observed which persists for almost two hours. All calculations have shown a homogenization of the temperature in the plenum after about 15 minutes. A slight reduction of the mixing in the upper plenum could have been achieved by modifying the form of the flow holes in the inner barrel (fillets instead of sharp edges) in order to reduce their axial pressure loss.

몬주 고속증식로 상부플레넘에서의 열성층에 관한 전산유체역학 해석

A numerical analysis of thermal stratification in the upper plenum of the MONJU fast breeder reactor was performed. Calculations were performed for a 1/6 simplified model of the MONJU reactor using the commercial code, CFX-13. To better resolve the geometrically complex upper core structure of the MONJU reactor, the porous media approach was adopted for the simulation. First, a steady state solution was obtained and the transient solutions were then obtained for the turbine trip test conducted in December 1995. The time dependent inlet conditions for the mass flow rate and temperature were provided by JAEA. Good agreement with the experimental data was observed for steady state solution. The numerical solution of the transient analysis shows the formation of thermal stratification within the upper plenum of the reactor vessel during the turbine trip test. The temporal variations of temperature were predicted accurately by the present method in the initial rapid coastdown period (~300 seconds). However, transient numerical solutions show a faster thermal mixing than that observed in the experiment after the initial coastdown period. A nearly homogenization of the temperature field in the upper plenum is predicted after about 900 seconds, which is a much shorter-term thermal stratification than the experimental data indicates. This discrepancy is due to the shortcoming of the turbulence models available in the CFX-13 code for a natural convection flow with thermal stratification.

Uncertainty analysis of the prototype FBR Monju with the JENDL-4.0 nuclear data set

This paper discusses the uncertainty analysis of the criticality of the prototype Fast Breeder Reactor Monju with JENDL-4.0 nuclear data. The JENDL-4.0 nuclear data are more physically correct and complete than previous JENDL releases. The well-established ERANOS code was used. Since no cross section libraries based on JENDL-4.0 exist for ERANOS, these libraries were generated as part of the work. To validate the newly made libraries, benchmark calculations were performed. A model of the Monju reactor was prepared, based exclusively on publicly available information. The criticality of May 2010 of Monju was adequately calculated. A perturbation analysis identified the most important differences between JENDL-3.3 and JENDL-4.0. A strong contribution of 241Am was found. In the subsequent uncertainty analysis, it was found that the overall uncertainty is roughly the same between JENDL-3.3 and JENDL-4.0. This result was not as expected. Two main causes have been identified: firstly, in JENDL-4.0 covariance data has been added for isotopes and energy ranges for which there previously was no covariance data. Secondly, for some isotopes the covariances in JENDL-4.0 are larger than in JENDL-3.3. For several main reactions and isotopes, the uncertainty is largely reduced between JENDL-3.3 and JENDL-4.0. But for many isotopes with a non-negligible contribution the uncertainty is increased in JENDL-4.0. The balance is a slight increase in the overall uncertainty when JENDL-4.0 is used.

Heat Transfer in Intermediate Heat Exchanger under Low Flow Rate Conditions

This paper describes the heat transfer in intermediate heat exchangers (IHXs) of liquid-metal-cooled fast reactors when the flow rate is low, such as under natural-circulation conditions. Although empirical correlations of heat transfer coefficients for IHXs were derived using test data of the fast reactors Monju and Joyo and of a 50-MW steam generator facility, the measured heat transfer coefficient was very low compared to the well-known correlation for liquid metals proposed by Seban and Shimazaki. The heat conduction (HC) in IHX is discussed as a possible cause of the low Nusselt number. However, the present results show that HC is not significant under natural-circulation conditions, and the HC term in the energy equation can be neglected in the one-dimensional plant dynamics calculation. Simulations relating natural-circulation transients tested at the Monju reactor were conducted using the NETFLOW++ code with the proposed empirical correlations. Good agreement was obtained for long-term behavior.

Breeding of 242mAm in a Fast Reactor

There is growing interest in the use of 242mAm as a nuclear fuel. Since the thermal absorption cross section of 242mAm is very high (σa = 8950 b), the best way to obtain 242mAm is by the capture of fast or epithermal neutrons in 241Am. As a result, we have considered replacing the radial blanket of a fast reactor, which is usually depleted uranium, with 241AmO2.We chose a 714-MW(thermal) MONJU reactor, and we replaced some of the radial blanket and the outer core assemblies with 10 676 kg of 241AmO2 fuel. We calculated the reactor core by using the MCNP Monte Carlo code.The total amount of 242mAm becomes stabilized after 16 yr, but the enrichment does not. In our calculation, ~7.2% enrichment is obtained after 18 yr. Obtaining higher enrichments might indicate that 242mAm nuclear fuel can be used without further enrichment in many cases.The results presented in this paper are considered an upper limit scenario. In particular the target 241Am loading is not likely to be available soon, but 242mAm production from lesser amounts is easily scaled down proportional to the actual mass irradiated.

Development of Manufacturing Process of PNC-FMS Wrapper Tube with SUS316 Short Joint

When a ferritic-martensitic stainless steel (PNC-FMS) wrapper tube having far greater swelling resistance against neutron irradiation is applied in the JOYO or MONJU reactor, it becomes necessary to weld it with SUS316 austenitic stainless steel (entrance nozzle and handling head). Such welding between PNC-FMS and SUS316 causes the delta (δ) ferrite formation at heat-affected zone, which leads to significant toughness degradation. In addition, bending of wrapper tube caused by their differential thermal expansion should be straightened. For preventing those problems, manufacturing process of the complex wrapper tube was developed. This process involves TIG-welding with SUS316 short pipe joints in 50mm length to both ends of a PNC-FMS round tube, and then performing the drawing and normalizing and tempering. Normalizing induces complete disappearance of the δ ferrite in the course of wrapper tube manufacturing. The mechanical properties of PNC-FMS/SUS316 welded zone were confirmed to be equivalent to those of the base metal even after thermal aging.