Reactor Project Research Articles

Terrapower’s Molten Chloride Fast Reactor (MCFR): Advancements in Electrochemical Monitoring and Control

TerraPower’s Molten Chloride Fast Reactor (MCFR) project is pioneering a nuclear reactor technology that leverages fast spectrum chloride as a fuel salt and chloride salts as coolant salts, offering vast potential for decarbonizing various economic sectors. The fast spectrum enhances the neutron economy, mitigating fission product poisoning. Utilizing a NaCl-UCl3 binary eutectic as the fuel salt, the reactor undergoes significant chemical transformations during burn-up. The resulting species from the fission process, including Kr(0), Xe(0), Rb(I), Cs(I), Sr(II), Ba(II), Zr(II), Nb(0), Mo(0), Te(0), I(0), Pd(II), Re(I), Rh(II), Ag(I), Cd(II), rare earths(III), are in oxidation states equal to or lower than the initial U(III). This transformation leads to an oxidizing melt, potentially converting U(III) to U(IV), which can trigger oxidative corrosion of the structural materials.To address these challenges, TerraPower is employing diverse electrochemical and electroanalytical approaches for process monitoring and control, helping bridge existing knowledge gaps in fuel and coolant salts. These methods provide key insights into species thermodynamics and kinetics, enabling direct measurement of redox information crucial for the operation of a molten salt nuclear reactor. This presentation will discuss TerraPower’s efforts in applying diverse electrochemical and electroanalytical approaches towards process monitoring and process control, addressing the knowledge gaps that exist for both fuel and coolant salts.Basic electrochemical experiments on initial fuel salt provide data for models that predict and prevent corrosion, identify species affecting reactor physics, and offer real-time monitoring of the need to remove fission product poisons. Spectroelectrochemical methods are adapted to examine how redox transformations impact the coordination and speciation of uranium complexes in the fuel, crucial for predicting precipitation and volatilization of U(IV) species. This presentation will discuss TeraPower’s progress in electrochemical and spectroelectrochemical experiments.TerraPower has designed two novel electrochemical and spectroelectrochemical cells that serve as testbeds for understanding fundamental redox chemistry and developing probes for process monitoring. For instance, measuring salt potential relative to the ratio of uranium oxidation states enables monitoring the U(III):U(IV) ratio, crucial for maintaining an optimal reactor environment. The temperature dependency of this value will also be characterized, providing a reference for system perturbations.The presentation will also touch upon the influence of fission and corrosion products on the equilibrium behavior of uranium throughout the fuel's operational life. Fission products may reach substantial concentrations in the salt, imposing varied redox conditions that drive different behaviors in uranium. The dependence of uranium oxidation state ratio on burnup will be evaluated, and method development will explore the relationship between chlorine potential and species solubility.Through this comprehensive approach, TerraPower aims to advance the MCFR technology, ensuring safe and efficient nuclear energy generation with significant benefits for decarbonizing the economy.

Selection of EU-DEMO divertor operating conditions: water cooling thermal-hydraulic parameters and power exhaust capabilities

In the context of EUROfusion activities for the development of the DEMO reactor project, the divertor design is a major challenge. It must sustain very high heat, ion particle and neutron fluxes allowing, at the same time, the shielding of the vacuum vessel and the vacuum pumping for reducing the plasma pollution. The conceptual divertor design is based on the use of EUROFER97 for the divertor cassette body, while tungsten monoblocks bonded to CuCrZr pipes are used for plasma-facing targets. EUROFER97 was selected considering its reduced long-term activation and superior creep and swelling resistance under neutron irradiation. However, depending on the operating temperature under neutron irradiation, a pronounced shift of the Ductile to Brittle Transition Temperature (DBTT) is expected. At the same time, for the plasma-facing targets, the coolant temperature has to be identified such to allow sufficient heat removal capacity at the strike point. This study explores alternative cooling conditions for the divertor system that are able to ensure the fulfillment of functional and system requirements and to allow for divertor cassette body re-use during plant lifetime. The main aim is to identify the best water cooling thermal-hydraulic conditions avoiding material embrittlement (for EUROFER 97) and softening/hardening (for copper alloy pipes). At the same time, the goal is to reduce the inventories (enthalpy of the cooling circuit) and the radwaste at the end of divertor lifetime.

Analysis of hydrogen control in a Small Modular Reactor during TLOFW severe accident

During the Fukushima Daiichi nuclear accident in 2011, hydrogen explosions occurred in all units from Unit 1 to Unit 3. Consequently, one of the lessons learned from the Fukushima Daiichi accident is the necessity of implementing hydrogen control and mitigation strategies for most Nuclear Power Plants (NPPs). This paper focuses on the incorporation of Passive Autocatalytic Recombiners (PARs) during the design phase of a small modular Pressurized Water Reactor (SMR-PWR) project. The numerical analyses are conducted using the MELCOR v. 2.2 code. Two scenarios are compared: the Total Loss of Feed Water (TLOFW) severe accident with and without PARs. Saphiro’s diagram is utilized to investigate whether the mixture's composition (hydrogen, air, steam) is flammable for both scenarios. It has been observed that the inclusion of PARs leads to a reduction in hydrogen risk (detonative or deflagrative) as the final hydrogen concentration values fall below the flammability limit. This study is preliminary, and further research is required.

PyroGenesis announces accelerated construction of fumed silica reactor project with arrival of all major equipment

PyroGenesis announces accelerated construction of fumed silica reactor project with arrival of all major equipment

Optimized modular nuclear reactor project utilizing artificial intelligence: Seed-blanket concept

This article proposes a Light Water Small Modular Reactor (LW-SMR) based on the seed-blanket concept with thorium fuel, utilizing Artificial Intelligence (AI) for optimization. The objectives include defining SMR parameters and developing parallel multi-objective AI models using Genetic Algorithm (GA) and Particle Swarm Optimization PSO. A GA has been developed and applied to a simplified fuel assembly, while a PSO algorithm has been used together with the SERPENT code to perform precise calculations for SMR simulation. The results obtained demonstrated a great improvement when compared to classical parametric studies, demonstrating the possibility of designing a fuel assembly using artificial intelligence.

Investigation of hydrogen generation from the reaction of beryllium pebbles with water vapor for WCCB-TBM

Be pebble, as a neutron multiplier material, is being explored for the development of a Water-Cooled Ceramic Breeder Test Blanket Module (WCCB-TBM) in Japan for the International Thermonuclear Experiment Reactor (ITER) project. The Be pebbles were fabricated by the rotating electrode process (REP). However, as the safety of the test blanket module (TBM) could be compromised by hydrogen generation via Be–H2O reaction owing to the leakage of pressurized water from the cooling pipes inside TBM, the safety assessment of TBM must be conducted using the Be–H2O reactivity data for the Be pebbles whose fabrication method is the same as that for TBM (i.e., REP). Thus, in this study, a Be–H2O reactivity test was conducted using Be pebbles that were fabricated by REP to characterize the generated hydrogen and evaluate the hydrogen generation (HG) rates toward WCCB-TBM safety assessment. First, the Be–H2O reactivity data were obtained at ≥ 300 °C (the coolant temperature for the TBM). The experimental results and oxidation rate of the Be pebbles revealed that only the Be surface may have been oxidized at 300 °C–600 °C, although its inside was oxidized at 700 °C–1000 °C. Further, the experimental data–derived HG rates at 300 °C–1000 °C increased with the temperature and the tendency changed in the three temperature regions. The HG rates did not contradict those reported by previous studies. Additionally, these regions corresponded to each expected Be–H2O reaction behavior at 300 °C–1000 °C. At a Be temperature of < 600 °C, the generated hydrogen from TBM will not exceed 2.5 kg of a limitation by the water leakage on the estimation based on the HG rates and the qualitative effects of the safety functions.

The economic logic of open science in fusion energy research: A policy perspective

This policy perspective explains the economic logic of open science in fusion energy research (FER) via application of the Mertonian norms of communalism, universalism, disinterestedness, originality, and skepticism. FER is transitioning between science and technology and so provides a fitting example of the productive balance between the community of scientists, who generally favor full disclosure of their results, and that of technologists, who typically favor secrecy or exclusive possession, as the New Economics of Science (NES) explains. For scientists, the publicly funded International Thermonuclear Experimental Reactor (ITER) project is an example of the benefits of open science because all its intellectual property is equally shared by its members. For technologists, the next Demonstration (DEMO) phase, is being characterized by an influx of private funding and by a shift from open science to proprietary technologies, which can negatively impact future research due to exclusive possession. The evidence provides support to continuing with a systematic approach to international environmental technology policy that favors managed co-opetition via open science because this can provide greater economic and social utility to both communities of scientists and of technologists.

EFFECT OF TRANSIENT LAYERS ON PLASMA ENERGY TRANSFER TO DIFFERENT SURFACES UNDER QSPA EXPOSURES

The plasma energy transfer to plasma-facing materials, as well as the energy and particles exhaust, needs to be extensively studied for the implementation of the next-step fusion reactor project. Analysis of plasma-surface interaction features has been performed using QSPA exposures of reference plasma-facing materials. The parameters of the plasma streams imitated conditions of transient events in a fusion reactor. The influence of an external magnetic field on the energy balance during the plasma-surface interaction is also discussed.

Irradiation performance of a U-7Mo in Al-Si matrix dispersion full-size fuel plate assembly

The Korea Atomic Energy Research Institute (KAERI) is leading the Ki-Jang Research Reactor (KJRR) project with the intent to develop a new reactor for medical isotope production and other nuclear research purposes. The KJRR core is designed to use high density fuel system where uranium alloyed with 7 wt% molybdenum (U-7Mo) particles are dispersed in a matrix of aluminum alloyed with 5 wt% silicon (Al-5Si) and clad in aluminum alloy 6061 (Al-6061) to form fuel plates. KAERI developed a fabrication facility to construct KJRR fuel assemblies and partnered with the Idaho National Laboratory (INL) to irradiate a full-size fuel assembly, with 21 total fuel plates, in the Advanced Test Reactor (ATR). Irradiation testing and subsequent Post Irradiation Exam (PIE) campaigns were performed successfully over a multi-year project. Monte Carlo neutronic calculations, coupling with a depletion code, were performed based on ATR’s as-run power history which showed that the highest power plate (plate 20) reached 83.1 % end-of-life (EOL) local burnup based on initial 235U content. Finite element thermal modeling was performed based as-run power history which showed a beginning-of-life (BOL) peak local heat flux of 184 W/cm2. No anomalous fuel performance was observed during the irradiation and target test conditions were achieved. PIE showed favorable performance of the fuel assembly regarding all important phenomena. This paper describes the KJRR fuel assembly irradiation conditions and PIE data to support the conclusion that it performed well, without evidence of unexpected or problematic fuel performance, within an irradiation test designed to bound the KJRR design environment.

Challenges in high-fidelity thermal–hydraulic simulation of SFR cores: Insights and PACA-S4FR solutions

Numerical simulations in studying thermal–hydraulic behavior in sodium-cooled fast reactors (SFR) have gained attention for their safety and cost advantages. While low-fidelity numerical simulations commonly used in industry have the advantage of lower computational costs, they face challenges in accurately representing intricate flow phenomena. On the other hand, high-fidelity numerical simulations offer excellent characterization of turbulence but demand significant computational resources. The emergence of exascale supercomputers has provided opportunities and challenges for large-scale high-fidelity numerical simulations. In this paper, firstly, we review the research progress of thermal–hydraulic numerical simulations of SFR cores using both low-fidelity and high-fidelity methods. Secondly, we estimate the computational and storage requirements for high-fidelity numerical simulations of SFR core based on calculations performed on China Experimental Fast Reactor (CEFR). Subsequently, we analyze the challenges of conducting high-fidelity thermal–hydraulic simulations of the entire SFR core, including large-scale mesh generation, appropriate numerical discretization methods, efficient simulation acceleration techniques, coupled simulations, and the complexities of validation and verification processes. Finally, we present the research progress of PACA-S4FR, a high-fidelity thermal–hydraulic simulation software in the China Virtual Reactor (CVR1.0) project, along with the ongoing efforts to address challenges in high-fidelity simulations. This paper contributes to advancing the comprehension of SFR core thermal hydraulics, potentially aiding in safety and performance enhancements in SFR.

Development and validation of Loss of Coolant Accident (LOCA) simulation capability in the ENIGMA fuel performance code for zirconium-based cladding materials

Loss-of-coolant accident (LOCA) modelling forms a pivotal element in the licensing of light water reactor technologies. Within this context, the National Nuclear Laboratory has further developed its ENIGMA fuel performance code to accurately simulate fuel behaviour during limiting design-basis accidents. The enhanced ENIGMA code now incorporates models for high-temperature cladding creep, cladding oxidation and hydriding at elevated temperatures, cladding phase change, and cladding failure. A detailed examination of existing LOCA-capable models relevant to industrial-standard fuel performance codes was conducted, which led to the introduction of an industrial state-of-the-art capability within ENIGMA for modelling uranium dioxide fuel in various zirconium alloy cladding materials. The efficacy of the enhanced ENIGMA code was validated using data from the IFA-650.10 LOCA experiment carried out as part of the OECD Halden Reactor Project. ENIGMA’s predictions were compared with both the experimental data and predictions of other fuel performance codes from the International Atomic Energy Agency’s Fuel Modelling in Accident Conditions (FUMAC) coordinated research project. The comparison generally showed good agreement, highlighting the effectiveness of the improved ENIGMA code.

Performance evaluation of nuclear fuel in a reactor based on the CAREM 25

Nuclear energy has been proven as a viable option for strengthening a power grid, and the expansion of the Brazilian nuclear programme is becoming a reality in light of new reactor projects currently in the licensing and/or construction stage in the country. The object of study in this paper is a small modular reactor (SMR) based on the CAREM 25 Reactor designed in Argentina that represents an innovation in nuclear reactor research in Latin America and belongs to the category of advanced reactors. It is a reduced-size Pressurized Water Reactor (PWR) with integrated primary and secondary circuits, passive and redundant safety systems, light water cooling, and a power output of 25 MW. In this study, the behaviour of certain safety parameters related to a nuclear fuel rod operating at high burnup is evaluated, considering different types of coatings commonly used in the country. The aspects addressed are related to plant operational safety and fuel integrity, such as fuel centreline temperature, cladding surface temperature, oxide layer thickness on the rod surface, and hydrogen concentration in the cladding. The results obtained show that the tool used for simulating the performance of fuel rods in conventional PWR reactors, such as the FRAPCON line, can be applied to smaller modular reactors as well, along with the feasibility of using well-established coatings in the country.

Technological readiness of alternative reactor concepts

Abstract. Some alternative reactor concepts are frequently discussed in the literature as relevant for the disposal of high-level radioactive waste. We present results on the technological readiness of and research and development on those reactor concepts from a project commissioned by the Federal Office for the Safety of Nuclear Waste Management on novel reactors (Pistner et al., 2023). The focus is on 7 technology lines and 10 specific reactor concepts within those lines. These are sodium-cooled fast reactors (SFRs), lead-cooled fast reactors (LFRs), gas-cooled fast reactors (GFRs), very high-temperature reactors (VHTRs), super-critical water-cooled reactors (SCWRs), molten-salt reactors (MSRs) and accelerator-driven systems (ADSs). For all technology lines considered, extensive research and development work has been taking place for several decades and in some cases since the middle of the last century. Depending on the technology line, technical test stands for individual phenomena have been built and operated and so have smaller experimental reactors and larger demonstration reactors. Nevertheless, until today no commercially competitive reactor concept exists in the field of alternative reactor concepts. The most extensive technical experience is available for the SFR and VHTR technology lines. However, proof of reliable operation under economic boundary conditions is still required for SFRs and VHTRs. For other technology lines, neither has technical feasibility in the form of a demonstration reactor been demonstrated so far (LFR, MSR) nor are more extensive findings from smaller experimental reactors available (GFR, SCWR, ADS). To plan, license, construct and operate such experimental and demonstration reactors, a period of at least 1 to 2 decades must be assumed for each reactor project and probably substantially more based on historical experience. It has been shown that developers' schedules are often characterized by overly optimistic assumptions, that developments are delayed by years or even decades, and that in many cases specific approaches are discontinued completely because the underlying technological difficulties could not be overcome. Thus, the time still required for the development of alternative reactor concepts is probably in the range of several decades. Against this background, it cannot be assumed that such reactor concepts will be used on a relevant scale by the middle of this century. Additionally we will present general results of this study regarding criteria such as safety, fuel supply and waste disposal, proliferation, and costs and point out advantages and disadvantages of those systems over light-water reactors.

He atoms diffusion and aggregation in Li2TiO3: A molecular dynamics study

Advanced reactor projects will consider the utilization of Li2TiO3 as a potential material for neutron moderation and certain radiation shielding applications in fission reactors, as well as for tritium breeder blanket material in fusion reactors. He is one of the particles produced by the reaction of Li with neutrons during the service of Li2TiO3. The diffusion and aggregation of He in Li2TiO3 can cause embrittlement failure of Li2TiO3. Understanding the diffusion and aggregation behavior of helium in Li2TiO3 is crucial for evaluating service life, as well as for the selection and design of advanced nuclear reactor components. In this paper, the diffusion of He in Li2TiO3 and the nucleation behavior of a He cluster were studied by molecular dynamics simulations. The following results were mainly obtained: (1) The long-range diffusion activation energy of He atoms in Li2TiO3 is about 0.5 eV, and it varies significantly with the change of Li vacancy content. (2) He atoms tend to be distributed in the Li layers after sufficient diffusion in Li2TiO3, so the best nucleation site for He clusters in Li2TiO3 may be located in the Li layers. (3) After a He cluster reaches a certain size, it can trigger defects by extruding an O-LiTi2-O layer in its immediate vicinity on one side and migrating with the extruded O-LiTi2-O layer. (4) A He cluster can promote their own growth by releasing pressure.

Development and Verification of a Computational Fluid Dynamics Model for a PWR-type Small Modular Reactor Subchannel

With regard to nuclear technology, innovation is critical not only in reducing the current and future costs of deploying Gen III reactors in the short to medium term, but also in enabling the advancement of Gen IV reactors and Small Modular Reactors (SMRs). Considering this aspect, Computational Fluid Dynamics (CFD) can be viewed as an innovative technology for nuclear thermal-hydraulics, increasing the range of possible analyzes with detailed results that are progressively enhanced by advances in processing power, seen in the last decades. This study presents the construction of a CFD model of a subchannel in a pressurized water reactor (PWR) using ANSYS Fluent. The data used for the analysis were obtained from a conceptual design of a Small Modular Reactor, and the development of the model was based on best practice guidelines from existing literature. Turbulence was modeled using the Standard k-ε Model, while a cosine-shaped heat generation profile was used for the fuel. Convergence and mesh studies were conducted to ensure the stability and accuracy of the numerical solution. The results indicated the development of a turbulent profile in the coolant flow, consistent with the inlet conditions. The temperature distribution and pressure drop along the subchannel were also examined. Verification performed by the comparison between the numerical and analytical outcomes revealed low errors. Maximum temperatures in each component remained below the safety limits prescribed for nuclear reactor projects. Based on these results, it can be concluded that the CFD model developed in this study is suitable for thermal-hydraulics analysis of PWR reactors, particularly when examining preliminary assessments of conceptual SMR configurations.

Evaluation of the Basic Neutronics and Thermal-Hydraulics for the Safety Evaluation of the Advanced Micro Reactor (AMR)

South Africa requires safe affordable distributed base load energy, one way to achieve this is to use nuclear power integrated with renewable energy sources on a decentralized basis. This suggests the development of its own micro modular nuclear reactor, to supply energy to towns, small communities, mines and processing plants. Large Light Water Reactors (LWRs) are expensive and require a large infrastructure development. A High Temperature Reactor (HTR) called the Advanced Micro Reactor (AMR) is in the process of being developed and the design philosophy is to design for inherent safety, maximally using technology that has been developed and validated in previous HTR programs albeit in a completely different and unique configuration. The concept is based on existing knowhow and experience/expertise in South Africa during the time of the Pebble Bed Modular reactor (PBMR) project. These AMR reactors are to be factory built to obtain good quality control and rolled out to various sites. Once the reactor has reached its end of life, it would be returned to a licensed organisation for refuelling. The AMR produces 10MW of thermal power. The reactor configuration uses hexagonal graphite blocks for structural and moderator material, which are arranged to form a cylindrical core layout. The fuel assemblies are silicon carbide tubes that house coated particle fuel, immersed in a lead-bismuth eutectic alloy (LBE). Each fuel assembly is contained in a boring within the graphite moderator that allows an annulus for cooling. There are 420 fuel assemblies in the core. Low enriched fuel in the form of UO₂ or UCO is used. Helium gas is used as coolant. The coolant enters the core at 450°C and exits at 750°C. The mechanical, neutronic and thermal-hydraulic design of the AMR, is being evaluated with assistance from STL Nuclear (Pty) Ltd., the University of Pretoria (UP), the North-West University and the South African Nuclear Energy Corporation (NECSA). The OSCAR-5 code package, together with the Serpent neutronic code were used to perform the basic neutronic studies while the Flownex package was used to determine the thermal-hydraulic and safety evaluation for the Design Base Accident (DBA) specifically the Depressurized Loss of Forced Cooling (DLOFC) event.

Urenco Leaves U-Battery Modular Reactor Project

Urenco is leaving the U-Battery project having ‘exhausted attempts to secure the commitment of new commercial investors.’

THE EFFECT OF A SMALL HELIUM ADDITION ON THE PLASMA-SURFACE INTERACTION IN QSPA

The synergistic effects of tungsten exposure to combined hydrogen and helium particle fluxes as well as transient thermal loads need to be extensively studied for implementation of fusion reactor project. The mixture of hydrogen and helium was used as the working gas for plasma stream generation within the QSPA-M accelerator. The parameters of the mixed hydrogen and helium plasma were similar to those of pure hydrogen plasma generated in QSPA-M. It was shown that the small addition of helium (5 %) to hydrogen does not strongly influence plasma surface interaction. The influence of the external magnetic field on plasma surface interaction is also discussed.

Three-dimensional modeling and simulation of RF cavity for a transmitter with bandwidth of 40–100 MHz at 100-kilowatt level

A 1.5 MW final power amplifier is to be developed for ion cyclotron resonance heating (ICRH) system for the China fusion engineering testing reactor (CFETR) project. First, the feasibility of driver power amplifier with input matching circuit and output resonator cavity and a newly developed tetrode was cross-verified through three-dimensional (3D) simulations and experiments. A single coaxial cavity in parallel with the tetrode covered the frequency range of 40 to 100 MHz at a 100-kilowatt level. In contrast to the previous method of deducing matching parameters using RF circuit and transmission line theory, the use of 3D transmitter simulations provided more accurate matching parameters of tuning components at different operating frequencies. A new phenomenon of high frequency oscillation between the short strip inductor and the shield cavity was found in the input simulation results, which was subsequently verified by experiments. Specific reasons for dividing the output into two parts of high and low frequency were found through output circuit simulations. One is that the length of short stub decreases with the increase of operating frequency, thus decreasing the frequency resolution if all frequencies were using low frequency port. The other is that the coupling capacitor increases sharply with the increase of frequency due to parasitic effect in the low output. Further, the simulation facilitated the study of parasitic effects on matching parameters caused by strip-inductance, inter electrode capacitances of tetrode, and RF cavity. Thus, facilitating the design of final power amplifier in the future. Finally, the amplifier installed with the new tetrode DB968 achieved the design objectives in the test experiment with 150 and 110 kW power outputs at 48 and 100 MHz, respectively.

Development of adjusted nuclear data library for fast reactor application

In Japan, development of adjusted nuclear data library for fast rector application based on the cross-section adjustment method has been conducted since the early 1990s. The adjusted library is called the unified cross-section set, which is an ABBN-type group constant set with 70-group energy structure. The first version was developed in 1991 and is called ADJ91. After that, ADJ98, ADJ2000, ADJ2000R, and ADJ2010 were constantly developed. For instance, ADJ2010 was developed based on JENDL-4.0, which provides covariance data needed to apply the cross-section adjustment method, by using 488 integral experimental data acquired in typical fast reactor systems. ADJ2010 has been used as the standard cross-section set for nuclear design in the fast reactor cycle technology development project (FaCT) and the succeeding fast reactor projects. In parallel, the integral experimental data were further expanded to improve the design prediction accuracy of the core loaded with MA and/or degraded Pu. Using the additional integral experimental data, development of the next version of ADJ2017 was started in 2017. In 2022, the latest unified cross-section set AJD2017R was developed based on JENDL-4.0 by using 619 integral experimental data. An overview of the latest version with a review of previous ones will be shown. On the other hand, the latest Japanese evaluated nuclear data library JENDL-5 was released at the end of 2021. In the development of JENDL-5, some of the integral experimental data used in ADJ2017R were explicitly utilized in the nuclear data evaluation. However, this is not reflected in the covariance data. This situation needs to be considered when developing a next version of the unified cross-section set based on JENDL-5. Preliminary adjustment calculation based on JENDL-5 is performed using C/E (calculation/experiment) values simply evaluated by a sensitivity analysis. The preliminary result of the JENDL-5-based adjustment will be also discussed.