Breeder Reactor Research Articles

Robust controller design strategies with improved performance for MSBR core

Designing controllers for power regulation of nuclear reactors is a challenging task because of its non-linear behavior, and fluctuation of system parameters over time in response to changes in power levels. Motivated by the aforementioned fact, proportional-integral (PI) controllers are designed in this manuscript to regulate the core power of the molten salt breeder reactor (MSBR) by employing the all stabilizing region and quantitative feedback theory (QFT) technique. In the first approach, the entire stability region is plotted in the (Kp,Ki)-plane. In order to ensure safe operation, the reactivity of control rods must lie in the specified range. Hence, the control effort constrained stability region (CECSR) is obtained within the all stability region to design the PI controller. A novel PI controller and a pre-filter based on gain-phase shaping on Nichols chart are also proposed. Robustness, disturbance rejection and relative stability of the system are addressed by obtaining the performance bounds on Nichols chart. Simulation results illustrate that the closed-loop system is exhibiting a noteworthy improvement in the closed loop performance over the latest reported control strategies.

Development of Whole System Digital Twins for Advanced Reactors: Leveraging Graph Neural Networks and SAM Simulations

In this work, we introduce a novel method to develop whole system digital twins (DTs) for advanced nuclear reactors. This method treats a complex reactor system as a heterogeneous graph: with the system components as different types of graph nodes and their physical interconnections as edges. Based on the heterogeneous graph, a graph neural network combining graph convolution and temporal node attention is developed as the DT, facilitating a comprehensive understanding of the system’s dynamic behavior. By utilizing the System Analysis Module (SAM) code for simulating various operational transients, we develop a graph-based database that trains the DT. This DT is characterized by two primary functions: It can infer the entire system’s status using sparse node information, and it can predict the progress of transients based on current and historical system information. Our approach is validated through case studies on the Experimental Breeder Reactor II (EBR-II) system and a generic Fluoride-salt-cooled High-temperature Reactor (gFHR), demonstrating the DT’s accuracy in forecasting operational transients. The DT’s rapid computation capabilities enhance its potential for supporting advanced reactor operations, offering benefits in intelligent simulation, autonomous control, and anomaly detection, paving the way for improved safety analysis and intelligent component health management for advanced reactor systems and reducing their operations and maintenance cost.

Conceptual design study of the tritium breeding unit for verifying the long-term performance of the breeding blanket

Fusion reactors produce high-energy neutrons through the Deuterium-Tritium (DT) reaction. The long-term performance verification and structural integrity assessment of the breeding blanket under continuous plasma operation are essential for the development of Demonstration fusion power reactor (DEMO) breeding blankets. The Korea Institute of Fusion Energy (KFE) has been planning the construction of a test facility called the Integrated Breeding Test Facility (IBTF) that uses a linear accelerator as the neutron source. A solid beryllium target has been considered to generate fusion-like neutrons. A conceptual design of the Tritium Breeding Unit (TBU), which is a key component of the breeding blanket, has been carried out. This study evaluated tritium production under fusion-like conditions using factors such as the position and thickness of the graphite block, the thickness of the first wall, and the radial length of the breeding zone. The results showed that increasing graphite block thickness and optimizing the dimensions of the TBU enhanced tritium production. However, these changes affected the structural integrity of the TBU. The side wall of the TBU experienced membrane plus bending stress, which exceeded the Level C and D acceptance criteria under the assumption of In-Box LOCA. Measures were investigated to ensure structural integrity while improving tritium production of the TBU.

Molten Uranium Breeder Reactor (MUBR) and Its Development Steps

Background The Molten Uranium Breeder Reactor (MUBR) is a radical new mixed-energy spectrum breed and burn reactor concept. The MUBR is fueled with molten uranium metal fuel in large fuel tubes instead of thin fuel rods, and is cooled by circulating the molten fuel through a heat exchanger. The purpose of this research is to find MUBR configurations with a SCALE (RSICC request/license 203869) burnup at least 10 times greater than the initial fissile content. Methods A proprietary computer program uses parameters to generate MCNP (RSICC request/license 176034) or SCALE input files and initiate MCNP or SCALE burn simulations and other analysis. This allows relatively fast comparison of different parameters to optimize the configuration. Results MUBR configurations have been found which have SCALE burn simulations through 120 years of fuel life with a burnup of 35% of the initial fuel mass when the initial fuel was Low Enriched Uranium (LEU) (3% U-235). Conclusions These results suggest that if it can be constructed and operated reliably for a long time, the MUBR would be a true breed and burn reactor. Compared to Light Water Reactors (LWR) this would provide many advantages for the nuclear industry and the world such as: reactor operation without shutdowns for refueling or fuel manipulation, no need for special fuel, ten times as much energy per ton of fuel, one tenth as much Used Nuclear Fuel (UNF) per megawatt hour of electricity, no UNF storage required every two years, increased energy security, and reduced nuclear proliferation risk. While development of the MUBR concept from a concept to a prototype reactor will be costly, the advantages suggest that the concept merits further study.

Measurement of fast neutron induced (n,γ) reaction cross-section of 68Zn, 96Zr, 121Sb and 123Sb in the energy range of 1 to 2 MeV

The (n,γ) reaction cross-section for the elements 68Zn, 96Zr, 121Sb and 123Sb, present in the reactor structural/shielding materials, was measured by neutron activation technique in the neutron energy region of 1–2 MeV as very limited data is available in this energy range. Further, the neutron spectrum peaks in this energy region for the fast breeder reactors and proposed accelerator driven sub-critical systems. The natural strontium (natSr) element was used as a neutron flux monitor by considering effective combined reaction cross-section for 86Sr(n,γ)87Srm and 87Sr(n,n′)87Srm reactions. The pellets of mixture of sample and monitor were irradiated by a quasi-mono energetic fast neutron beam, generated by 7Li(p,n)7Be reaction at FOTIA, Bhabha Atomic Research Centre, Mumbai, India. The activity of activation products was measured by off-line gamma-ray spectrometry using High Purity Germanium Detector (HPGe). The present data with improved uncertainty and covariance analysis enhance the cross-section data base for better constraining the evaluated data and theoretical models. The theoretical (n,γ) reaction cross-sections were calculated using TALYS 1.96, which could reasonably explain the present data with the Fermi gas level density prescription.

Microstructural constituents in compositionally graded NiCrFeSiBC hardfacing alloy/316L SS additive manufactured deposits

Ni-Cr-Fe-Si-B-C hardfacing alloys (D50) provide excellent resistance to abrasive and adhesive wear, high temperature oxidation and corrosion. These alloys are candidate materials for hardface coatings on in-reactor components of fast breeder reactors. Conventionally these deposits are made on steels through the welding route. Due to high fluidity of the molten alloy, generation of residual stress and significant dilution of the deposit, achieving crack free deposits is often a very difficult and time consuming task. To address such issues arising out of conventional fabrication routes, compositionally graded coatings have been tried through laser additive manufacturing. A 3 layered compositionally graded deposit (D50/50D50-50SS/SS) was made on 316 L SS build plate and the phase constituents and possible correlation with micro hardness was analyzed. Uppermost D50 layer consisted of γ-Ni, γ-Ni + Ni3B and γ-Ni + Cr5B3 eutectics, CrB, Cr7C3, Cr23C6, Cr3C2 and Ni3Si precipitates. The middle layer consisted of γ-(Ni,Fe), γ-(Ni,Fe)+Cr5B3, Cr5B3, Cr3C2, Cr7C3,Cr23C6 and Ni3Si. The SS layer closest to the build plate had γ-Fe, Cr3C2 and Cr7C3 precipitates. Corresponding to the variation in microstructure a gradual change in hardness also can be seen starting from the top layer to SS build plate: 580 ± 35 HV0.1, 532 ± 17 HV0.1, 348 ± 21 HV0.1 and 230 ± 10 HV0.1, which is promising.

Multiphysics analysis of mechanical responses in micro-reactors under varied operating conditions

Micro-reactors are a promising alternative for power production applications where traditional methods may not be feasible (e.g., space missions). Their design requires critical selection of materials, especially for moderator and fuel, to ensure long-term operation and optimal performance. This research focuses on developing computational analysis models for micro-reactors, aiming to explore their structural behavior under thermal and mechanical loadings. The micro-reactor model is based on the Zero Energy Breeder Reactor Assembly (Zebra) Kilowatt (kW) reactor design conceptualized at Los Alamos National Laboratory. This study investigates the thermal and mechanical responses of three distinct reactor models considering factors, such as boundary conditions, material properties, and thermal-mechanical loading. In this study, Beryllium Oxide (BeO), Beryllium (Be), Aluminum Oxide (Al2O3), and Magnesium Oxide (MgO) are chosen as materials for the reflector. Yttrium Hydride (YH) and Zirconium Hydride (ZH) are chosen as moderators. Uranium Nitride (UN) and Uranium Molybdenum (UMo) are chosen as reactor fuel materials. The thermal analysis predicts the deformation due to the thermal expansion within the reactor under steady operating conditions at 25 kWth, mimicking the maximum rated power of the Zebra kW design. The mechanical study simulates the system's response to forced vibration loads applied along the cooling channels, matching the natural frequency of the system. The results from both analyses show that Beryllium Oxide, Yttrium Hydride, and Uranium Nitride demonstrate the most optimal stress distribution in the system compared to other configurations investigated here. The results of this study can be developed into datasets to facilitate future machine learning studies, enabling the prediction of mechanical and thermal responses for micro-reactors.

Enhancing fuel breed-burn performance in a sodium-cooled fast reactor using a novel reactivity control method

This study aims to enhance the fuel cycle and fissile breeding performance of a sodium-cooled fast breeder reactor (FBR) by utilizing minor actinides (MAs) as a means of reactivity control alongside partially-inserted control rods. Choosing the PFBR-500 as the reference design, four core models, designated as Cases A, B, C, and D, utilizing various proportions of minor actinides (MAs) were built and simulated using OpenMC. The MA concentrations were optimized to compensate for the withdrawal of control rods and ensure the same initial reactivity for all cores. Burnup analysis over 365 EFPDs revealed a significant increase in cycle length and burnup for the modified cores along with a modest rise in breeding ratio. Notably, Case C, employing 3.45 wt.% MAs in the 88 inner-core fuel subassemblies achieved an extra 62.25 EFPDs cycle length, a 33.74% rise in single-cycle burnup, and a 3.86% increase in breeding gain compared to the reference. Loading MAs into the inner-core region proved to be more effective in enhancing both fertile-to-fissile conversion (thus extending the cycle length and fuel burnup) and transmutation than utilizing MAs throughout the core due to greater neutron flux at the core center. While Case D utilizing 2.2 wt.% MAs both in the inner and outer core fuel subassemblies had the highest overall MA loading, it demonstrated lower increments in cycle length, burnup, and breeding gain compared to Case C. Case C also exhibited the highest overall destruction rate (approximately 24%/y) for Np and Am isotopes, and successfully transmuted around 24.08 kg 237Np, 9.33 kg 241Am and 3.82 kg 243Am over the course of a single year. The addition of MAs also achieved a slight flattening of the axial and radial flux profile and a decrease in the flux peaking factor. However, it slightly lowered the beta-effective, Doppler constant, and control rod assembly worth, and shifted the coolant void reactivity worth to the positive side.

Investigation on numerical simulation models for heat transfer characteristics of lead-bismuth eutectic

The liquid lead-bismuth eutectic (LBE) has distinct thermal and physical properties that challenge the effectiveness of traditional RANS turbulence models and turbulent Prandtl number (Prt) models for accurately simulating its flow and heat transfer characteristics. This study aims to investigate numerical simulation models of LBE based on existing experimental data. Simulations are conducted using different combinations of three turbulence models and four Prt models, and comparisons are made with experimental data. The local and overall heat transfer characteristics of LBE are analyzed, and the applicability of each model combination is evaluated. Results indicate that for simulating local heat transfer characteristics, the SST k-ω turbulence model combined with the Prt model proposed by Cheng et al. yields the highest accuracy. Additionally, the empirical correlation for heat transfer proposed by Kutateladze provides the best prediction of the local Nusselt number. For overall heat transfer characteristics, the combination of the SST k-ω turbulence model and the Prt model introduced by Reynolds et al. demonstrates the highest accuracy and applicability. This investigation might offer a pivotal benchmark for the discernment of appropriate computational fluid dynamics (CFD) models for liquid metal breeder reactor (LMBR) applications utilizing lead-bismuth eutectic (LBE) as coolant.

Influence of thermo-mechanical and solution annealing treatments on the flow behaviour of 14Cr-15Ni stainless steel

ABSTRACT Thermo-mechanical treatment (TMT) involving cold work and annealing is prevalent in enhancing the mechanical properties of precipitation-strengthened stainless steels. This investigation evaluates the effect of TMT and solution annealing treatment (SAT) on the flow properties of a Ti-added 14Cr-15Ni austenitic stainless steel, which is used as cladding material for fast breeder reactors (FBRs). Comparisons drawn with the solution-annealed (SAT) condition showed that TMT enhances the yield stress more significantly compared to the ultimate tensile strength. Analysis of the work-hardening behaviour suggested that though TMT enhanced the initial work-hardening rate, its microstructure was more susceptible to dynamic recovery. Despite exhibiting lower strength, the SAT process enhanced the ductility at all the investigated test temperatures. Adoption of this TMT process on this material has the potential to enhance the performance of the FBR's cladding.

A comprehensive study on the phase composition, mechanical properties and stability of Li4SiO4-Li2ZrO3 biphasic ceramics

Lithium orthosilicate (Li4SiO4) is regarded as a candidate for tritium breeding in fusion reactors. In this study, the Li4SiO4-Li2ZrO3 biphasic was developed to improve the sinterability, density, and mechanical properties of Li4SiO4. The Li4SiO4-xLi2ZrO3 (x = 0.5, 1) composite powders were prepared using solid-state reaction via in-situ method. Li6Zr2O7 existed in the ceramic powders at low calcination temperatures. When the calcination temperature was increased to 900 °C, Li6Zr2O7 was transformed into Li2ZrO3 due to the decomposition of Li6Zr2O7 at high temperatures. The TEM observation confirmed that the powders consisted of Li4SiO4 and Li2ZrO3. The Li4SiO4 and Li4SiO4-xLi2ZrO3 (x = 0.5, 1) pebbles were fabricated by the sol-gel method. The measurement results showed that the pebbles had a narrow size distribution and fine sphericity. The density of Li4SiO4-Li2ZrO3 pebbles reached 96.01% of the theoretical density when it was sintered at 1000 °C for 4 h. Compared with the Li4SiO4, the grain size of ceramic pebbles was significantly reduced. Owing to decreased grain size, 139.0 N crush load for the pebbles and 92.4 MPa bending strength for the sintered bodies were achieved. Besides, the ceramics with the Li4SiO4 to Li2ZrO3 ratio of 2: 1 exhibited preferable mechanical properties. Further, to investigate the chemical stability of biphasic ceramics, the structure and mechanical properties were examined under high temperatures and continuous inert gas purging, simulating the working condition of fusion reactors. It was shown that there was almost no change in phase composition and grain size after purging for 60 h at 650 °C. For the mechanical properties, the crush load was decreased initially due to the cracking in the surface region of the ceramic and then increased because the cracking was recovered.

Evaluating hybrid nuclear-thermal power systems for reduced carbon footprint: A comparative analysis

The article provides an estimated substantiation of the concept of expanding the fuel and energy base and reducing the volume of greenhouse gas emissions with the joint use of organic and nuclear fuel in the structure of hybrid nuclear-thermal power plants. It is shown that the technological scheme (the production of saturated steam in a nuclear steam-producing plant - NPPU, and then overheating to the maximum achieved parameters in the thermal power industry due to organic fuel) leads to a decrease in the specific consumption of gas fuel for the production of electrical energy by 25–30 %, emissions of combustion products by 2.5 times (compared to thermal power plants), consumption of industrial water (compared to nuclear power plants) - by 1.5 times. It is indicated that it is possible to expand the fuel base of modern nuclear energy with thermal neutron reactors through the use of technologies mastered (breeder reactors) and promising (hybrid thermonuclear reactors). The fundamental advantage of the considered complex - JPPP-TPP - its feasibility in a short period of time with a significant reduction in the carbon footprint per unit of energy produced, which is an important condition for the sustainable growth of energy production while maintaining the ecological balance in the geosphere.

Hydrodynamics coupled circulating-fuel reactor equations in comoving frame and their analytical solutions for molten salt reactors

Effects of relative motion between moving fuel and neutrons are usually not considered in the analysis of Circulating Fuel Reactors (CFRs). In this study, we formulate neutron transport equation for CFRs in a hydrodynamic representation in terms of velocities relative to moving fuel. Using the P1 approximation in the comoving transport equation, the diffusion equation for CFRs is obtained. Mass transport of precursors is considered in the formulations. Comoving frame CFRs equations are analytically solved for critical slab problem and a closed form criticality condition is obtained. Comoving representation has introduced corrections to Eulerian cross sections arising from the acceleration term. Hydrodynamics coupling has posed density modifications to cross sections. A parametric study of corrective terms is carried out to calculate effects of these corrections on a generic MSR and on Molten Salt Breeder Reactor (MSBR).

Lessons Learned from Failures and Design Improvements for Large Diameter Seals

Large diameter seals are used in Prototype Fast Breeder Reactor (PFBR) to prevent the escape of radioactive argon cover gas through wide annular gaps. The design of these seals shall address different challenges than normal seals. The large tolerances needed on the width of annulus, achievable tolerances on rubber, clearances in the seal holder, etc., influence the seal performance. The paper discusses analytical and experimental work done to develop robust large diameter seals. The contours of the present study are based on the feedback from the testing and qualification of small size model seals. Various permutations of geometry for seal and its holder are numerically investigated. The study mainly highlighted the influence of seal holder on the seal performance. Various sensitivity studies done on selected seal and seal holder demonstrated robustness of the design solutions.Graphical

Fabrication, sintering behavior, and strength of tritium breeding ceramic pebbles with Pb addition

Lithium ceramics are regarded as potential candidates for tritium breeding in fusion reactors. Lead (Pb) was added to the ceramic to improve the strength and tritium release properties. The Li2TiO3-0.5Li4SiO4–Pb ceramic powders with the Pb contents of 5 wt% and 20 wt% were prepared by microwave-induced solution combustion synthesis. It was confirmed that the Pb was in the form of the Li2PbO3 phase in the powders. The Li2TiO3, Li4SiO4, and Li2PbO3 crystallites were composited at nanoscale level, which was confirmed by the transmission electron microscopy (TEM). During high-temperature sintering, the Li2PbO3 was transformed into PbO, which facilitated the densification process at low temperatures. The chemical state in the sintered body was characterized by X-ray photoelectron spectroscopy (XPS) with 3 keV Ar+ sputtering. The binding energies for Pb were 137.2 and 142.2 eV. There was almost no change in peak position during Ar+ sputtering, showing that Pb was firmly adherent with oxygen atoms. Besides, the as-synthesized powders were used for the pebble fabrication. The average crush load of Li2TiO3-0.5Li4SiO4–20%Pb pebbles sintered at 1223 K was 124 N, which was much higher than that without Pb addition. It was proved that the crush load increased almost linearly with the increase in the size of the pebbles. Thus, the bending strength of the sintered ceramic body was used to evaluate the mechanical properties. Further, the grain growth kinetics was studied. It was confirmed that the Pb addition reduced the activation energy of grain growth, which realized the densification of the ceramic pebbles at low temperatures.

Qualification and Quantification of Porosity at the Top of the Fuel Pins in Metallic Fuels Using Image Processing

Approximately 130,000 metal fuel pins were irradiated in the Experimental Breeder Reactor II (EBR-II) during its 30 years of operation to develop and characterize existing and prospective fuels. For many of the metal fuel irradiation experiments, neutron radiography imaging was performed to characterize fuel behavior, such as fuel axial expansion. While several fuel expansion results obtained from neutron radiography imaging have been published, the analysis of neutron radiography for the purpose of describing statistical properties of porous matter formed on top of the fuel pins, also referred to as fluff in previous publications, is significantly less represented in the literature with just a single paper so far. This study aims to validate and augment results reported in previous publications using automated image processing. The paper describes the statistical properties of the porous matter in terms of nine parameters derived from radiography images and correlates those parameters with such fuel properties as composition, expansion, temperature, and burnup. The reported results are based on 1097 fuel pins of eight different fuel compositions. For three major fuel types, U-10Zr, U-8Pu-10Zr, and U-19Pu-10Zr, a clear negative correlation is found between the Pu content and five parameters describing the amount of porous matter generated. The parameters describing granularity properties, however, showed either negative correlation or nonlinear dependency from fuel composition. The parameters describing the amount showed a positive correlation with fuel axial expansion, while granularity parameters showed a negative correlation with axial expansion. The dependency on cladding temperature was found to be weak. A positive correlation is demonstrated for volume parameters and fuel burnup. In general, reported results confirm and validate findings published in previous studies using a much larger number of pins and automated processing techniques, which easily lend themselves to reproducibility, thus avoiding subjective bias.

Development of a simplified one-dimensional CDA bubble model

Sodium-cooled fast reactors (SFRs) have high safety with an extremely low probability of a core disruptive accident (CDA). However, from a defense-in-depth perspective, the CDA sequence is still worth studying. Severe accidents in SFRs, such as unprotected loss of flow, might lead to a CDA during which fuel and some fission products could be instantly released from a large CDA bubble through potential leak paths in the top shield structure. Therefore, a reasonable prediction of dynamic behavior of a large-scale bubble within the sodium pool is vital for accurately evaluating the migration of source terms. In this study, we propose a simplified one-dimensional CDA bubble model that can handle heat and mass transfer in two-phase multicomponent materials in different computational domains during the rising of bubbles through the sodium pool toward the cover-gas region. In this model, an entrainment model based on the Rayleigh–Taylor instability and Kelvin–Helmholtz instability was used to explain the coolant entrainment through the gas/liquid boundary and jet fragmentation. The model also addresses the mitigation effect of non-condensable gases on the condensation of fuel, steel and sodium vapor at bubble interface. We evaluated the model using a past experiment on the expansion of a two-phase, large bubble with high pressure in a stagnant liquid pool conducted by Purdue University in the late 1970 s using a 1/7-scale model of Clinch River Breeder Reactor. Good agreement with the experimental data demonstrates that the developed model can reasonably represent the essential characteristics of dynamic behavior of a large, high-pressure bubble with heat and mass transfer in two-phase multicomponent materials. This is valuable for evaluating the migration of the source terms, which will be carried out in future studies.

The effect of Li2TiO3 pebbles on the mechanical properties of ARAA under breeding blanket conditions

The breeding blanket serves as a critical element for achieving a self-sustaining tritium cycle and generating sufficient energy in fusion reactors. This study focuses on candidate materials of the helium-cooled solid breeder blanket designed for a demonstration fusion reactor breeding blanket in Korea. We conducted compatibility tests between advanced reduced-activation alloy (ARAA) and Li2TiO3 pebbles at 550 °C for 28 days in He-0.1 vol% H2 purge gas conditions. Surface oxidation on ARAA was observed regardless of pebble presence. Double oxide films comprised of Fe-rich and Cr-rich layers formed on the surface of ARAA with pebbles. Lithium deposition was verified in both contact and non-contact areas of ARAA. Despite surface oxidation, the impact on tensile strength and elongation was found to be negligible.

Neutronic Analysis of the Conceptual Molten Uranium Breeder Reactor Using MCNP and SCALE Tools

The Molten Uranium Breeder Reactor (MUBR) is a radically new reactor concept with a mixed-energy spectrum. MUBR is fueled with molten uranium metal in large-diameter fuel tubes and is cooled by circulating molten uranium fuel through a heat exchanger. The reactor has heavy water as moderator, and the reactivity of the reactor is primarily controlled by the voiding effect of the moderator through an innovative control cavity structure design. Because the MUBR design is vastly different from most existing fission reactors, neutronics analysis must be performed for many different combinations of design parameters to identify viable and optimum design configurations. To facilitate the neutronics analysis, a proprietary program called MUBR6gen is being developed to provide a pipeline tool to expedite the process. MUBR6gen employs two well-established neutronics codes, i.e., MCNP and SCALE, to perform standard neutronics calculations for MUBR by automating input preparation and output processing. In addition, MUBR6gen ensures consistency of the MCNP and SCALE inputs and compares the outputs of the two codes to warrant the simulation results. Augmented with MUBR6gen, standard neutronics analysis was carried out on a small-scale MUBR design, which serves as a model problem in the paper. The neutronics performance characteristics of the model reactor were obtained and discussed in a code-to-code pattern. An overall very good agreement between the results of the two neutronics codes was established. Based on the success of the model problem analysis, further neutronics analysis using MUBR6gen was extended for a set of MUBR variant designs. Meaningful and promising fuel cycle analysis results for the 10 different designs were achieved and discussed. These results are used to identify the best MUBR candidates in terms of fuel lifetime and utilization efficiency for future applications.

Evaluation of Reaction Rate Distribution for Shielding Region in the Prototype Fast Reactor Monju

The reliability and usefulness of the reaction rate distribution data measured in the prototype fast breeder reactor Monju were examined through a comparison with a calculation using JENDL-4.0, mainly focusing on shielding regions around the reactor core. The reaction rates of 238U (n,f) and 58Ni (n,p) sensitive to high-energy neutrons were all judged reliable. The calculation-to-experiment values were slightly worse in the shielding regions; however, those for the 58Ni (n,p) reaction rates were improved by employing JEFF-3.3 instead of JENDL-4.0. A different tendency was observed between the two reactions, probably due to the 238U (n,f) cross section in the energy range of around 700 eV. The reaction rates of 235U (n,f), 239Pu (n,f), 238U (n,γ), and 197Au (n,γ) sensitive to the lower-energy neutrons were mostly judged reliable. The data in the lower axial shielding region are less reliable but may be acceptable for the shielding calculation.