High Temperature Engineering Test Reactor Research Articles

Limited linear source approximation with edge detection for convergence stability of method of characteristics

ABSTRACT A new implementation of the limited linear source approximation (LLSA) is proposed. The LLSA was previously proposed to eliminate local negative source in flux regions to mitigate numerical instability of the method of characteristics (MOC) with linear source approximation (LSA). In the present LLSA implementation, the convex edges of flux regions are used to check the local negative source to decrease the computational load. The present method is implemented in the transport code GENESIS, and its effectiveness is verified through the two dimensional C5G7 benchmark problem and the simplified two-dimensional high-temperature engineering test reactor core. The calculation results indicate that the present LLSA implementation efficiently mitigates the numerical instability of MOC with LSA. Additional computational time is less than 1% of total computation time.

Development of a Surrogate System of a Plant Dynamics Simulation Model and an Abnormal Situation Identification System for Nuclear Power Plants Using Deep Neural Networks

In the case of a new nuclear reactor, existing evaluation experience is limited; thus, accidents and troubles may occur as a result of such lack of experience. To deal with such situations, it is desirable to use a virtual nuclear plant to reproduce behaviors under various conditions and identify unknown anomalies from the behaviors. Then, when an abnormal situation occurs, one can quickly determine the cause of the abnormality to operate plant equipment and return the plant to a stable condition as quickly as possible. Two types of deep neural network (DNN) systems have been constructed to support the identification of unknown anomalies and the determination of their causes. One is a surrogate system that can estimate physical quantities of a nuclear power plant in a computational time of several orders less than a physical simulation model. The other is an abnormal situation identification system that can estimate the state of the disturbance causing an anomaly from physical quantities of a nuclear power plant. Both systems are trained and tested using data obtained from the analytical code for incore and plant dynamics (ACCORD), which reproduces the steady and dynamic behavior of the actual High Temperature Engineering Test Reactor (HTTR) under various scenarios. The DNN models are built by adjusting the main hyperparameters. Through these procedures, these systems are shown to be able to perform with a high degree of accuracy.

Feasibility of using BeO rods as secondary neutron sources in the long-life fuel cycle high-temperature gas-cooled reactor

External sources of neutron provide stable and sufficient neutron for initial startup of a nuclear reactor. They also provide signals for neutron detectors to monitor the safety of reactor during shutdown. In the high temperature engineering test reactor (HTTR), 252Cf is used as the external neutron source. However, the 252Cf sources must be renewed every approximately 7 years because of its relatively short half-life of 2.6 years. The renewal of 252Cf sources requires a high cost and a very complicated procedure. This study investigated the feasibility of using BeO rods as the secondary neutron sources to avoid renewing the 252Cf neutron sources periodically. The BeO rods could exist in the reactor for a long time so that if the reactor operates long enough, the neutron flux at the wide-range monitoring detectors remains more than 10n.s−1.cm−2 even if the reactor is shutdown for as long as 5 years. The results of this study indicated that using BeO rods as the secondary neutron sources would be an attractive option for the future HTGR design with a long-life fuel cycle.

Numerical study of the flow and heat transfer characteristics in a scale model of the vessel cooling system for the HTTR

The reactor cavity cooling system (RCCS) is a passive reactor safety system commonly present in the designs of High-Temperature Gas-cooled Reactors (HTGR) that removes heat from the reactor pressure vessel by means of natural convection and radiation. It is one of the factors responsible for ensuring that the reactor does not melt down under any plausible accident scenario. For the simulation of accident scenarios, which are transient phenomena unfolding over a span of up to several days, intermediate fidelity methods and system codes must be employed to limit the models’ execution time. These models can quantify radiation heat transfer well, but heat transfer caused by natural convection must be quantified with the use of correlations for the heat transfer coefficient. It is difficult to obtain reliable correlations for HTGR RCCS heat transfer coefficients experimentally due to such a system’s size. They could, however, be obtained from high-fidelity steady-state simulations of RCCSs. The Rayleigh number in RCCSs is too high for using a Direct Numerical Simulation (DNS) technique; thus, a Reynolds-Averaged Navier–Stokes (RANS) approach must be employed. There are many RANS models, each performing best under different geometry and fluid flow conditions. To find the most suitable one for simulating an RCCS, the RANS models need to be validated. This work benchmarks various RANS models against three experiments performed on the HTTR RCCS Mockup by the Japanese Atomic Energy Agency (JAEA) in 1993. This facility is a 1/6 scale model of a vessel cooling system (VCS) for the High Temperature Engineering Test Reactor (HTTR), which is operated by JAEA. Multiple RANS models were evaluated on a simplified 2d-axisymmetric geometry. They were found to reproduce the experimental temperature profiles with errors of up to 22% for the lowest temperature benchmark and 15% for the higher temperature benchmarks. The results highlight that the pragmatic turbulence models need to be validated for high Rayleigh natural convection-driven flows and improved accordingly, more publicly available experimental data of RCCS resembling experiments is needed and indicate that a 2d-axisymmetric geometry approximation is likely insufficient to capture all the relevant phenomena in RCCS simulations.

Reliability and availability simulation for vessel cooling system of the high temperature engineering test reactor

The article deals with the problem of risk analysis for new and unique nuclear facilities where safety-related systems are continuously operated during normal exploitation. A non-standard approach to system modeling was applied which goes beyond traditional method suitable for on-demand safety systems with a fixed mission time. As an example, the Vessel Cooling System (VCS) of the Japanese High Temperature Engineering Test Reactor (HTTR) has been analyzed in terms of life-cycle reliability and availability. Normal operation and emergency conditions have been considered, differing in the requirements for the system performance. In the former, the VCS reliability for cooling the biological shield around the Reactor Pressure Vessel (RPV) has been examined. In the latter, the VCS availability to remove the residual heat from the reactor core and RPV has been simulated. The results are of high importance for the safety and profitability of the HTTR-based electricity-hydrogen cogeneration plant (HTTR-GT/H2). The high reliability of the VCS under emergency conditions has been confirmed, thus contributing to the probabilistic safety assessment (PSA). Moreover, the fraction of the Forced Outage Rate (FOR) caused by VCS failures has been determined, thus contributing to the plant profitability studies.

Benchmarking exercise of indigenous Monte Carlo neutron transport code PATMOC for hexagonal full core reactor geometry

An indigenous Monte Carlo neutron transport code, named PATMOC, has been designed and developed at Bhabha Atomic Research Centre (BARC) for reactor physics design applications. The code has been extensively tested for its algorithmic accuracy against numerous international numerical benchmarks, has also been validated on practical nuclear reactor systems like Advanced Heavy Water Reactor Critical Facility (AHWR-CF) and Apsara-U pool type research reactor located at BARC, and is now in regular use for day-to-day reactor physics studies. These reactor systems and most of the benchmark problems used for the verifications and validation of the code have square lattice arrangement in the core. In this paper, we present the results of verification of PATMOC code on whole-core reactor system with hexagonal lattice arrangement of fuel elements. In this exercise, we have analyzed the High Temperature engineering Test Reactor (HTTR) whole-core benchmark for integral and differential parameters at both lattice and core level. The results demonstrate the algorithmic accuracy of the code on hexagonal reactor systems also. This exercise will help make use of this code for reactor physics design studies also for the reactor systems with hexagonal arrangement.

Improved multiphysics model of the High Temperature Engineering Test Reactor for the simulation of loss-of-forced-cooling experiments

We present a multiphysics model of the High Temperature Engineering Test Reactor for comparison with past and predict future loss-of-forced-cooling (LOFC) experiments. The approach selected combines (1) 3-D full-core superhomogenization-corrected neutronics, (2) 3-D full-core homogenized or semi-heterogeneous heat transfer (macroscale), (3) 2-D axisymmetric fuel rod heat transfer (pin-scale), and (4) 1-D thermal-hydraulics channels. Although large uncertainties remain, the time and magnitude of the first power peak after re-criticality is predicted within 1.5 h and 175 kW, respectively. The novelty of our work includes (1) a new macroscale/pin-scale heat transfer coupling approach relying on gap conductance to drastically speed up numerical convergence by two orders of magnitude, (2) determination of a radial effective thermal conductivity, reproducing the semi-heterogeneous re-criticality time within one hour using a homogenized macroscale model, and (3) a preliminary study of the reactor’s early behavior following a LOFC event, enabling further assessment of numerical models against fission power measurements.

Study on evaluation method of kernel migration of TRISO fuel for High Temperature Gas-cooled Reactor

Kernel migration of TRi-structual ISOtopic (TRISO) fuel for High Temperature Gas-cooled Reactor (HTGR) has been analyzed to investigate the potential dominating effects. Kernel migration is a major fuel failure mode and dominant to determine the lifetime of the fuel for High Temperature engineering Test Reactor (HTTR). However, the result and reliability depend on the evaluation method. Therefore, we propose the recommendation method by taking into account of actual distribution of Coated Fuel Particles (CFPs) and the resulting heterogeneous fuel temperature calculation with such distribution. As a result, the evident and conservative method can be proposed, and the method shows that the Kernel Migration Rate (KMR) is predicted to be about 10% less compared with the most conservative evaluation.

Present status of JAEA’s R&D toward HTGR deployment

Japan Atomic Energy Agency (JAEA) has been promoting the research and development (R&D) of High Temperature Gas-cooled Reactor (HTGR). HTGR having superior safety features can supply high temperature heat above 900 °C and it makes possible to use nuclear energy to heat source for various industries as well as to power generation. Promotion of R&D on HTGR and hydrogen production system is stated in Japanese policies, e,g, The 6th Strategic Energy Plan and The Green Growth Strategy Through Achieving Carbon Neutrality in 2050. R&D on reactor technologies is carried out by using High Temperature engineering Test Reactor (HTTR) which is the first HTGR in Japan. On July 30th 2021, the HTTR was resumed without significant reinforcements. On January 28th 2022, a safety demonstration test under the OECD/NEA LOFC project was carried out. JAEA is promoting R&D on a carbon-free hydrogen production by thermochemical water splitting Iodine-Sulfur process (IS process). Hydrogen production test for 150 h with 0.03 m3/h-H2 was successfully performed in January 2019. JAEA conducts design study for various HTGR systems toward commercialization. The major one is GTHTR300C for hydrogen and electric power cogeneration. JAEA has launched a new test program about demonstration of hydrogen production by the HTTR. Steam methane reforming hydrogen production system was selected for the first demonstration by 2030. The other carbon-free hydrogen production method will be demonstrated, later.

MCNP6 calculation of neutron flux map in the httr during normal operation

Detailed neutron flux distribution is important to understand the neutronic behavior during operation as well as to precise the core optimization and safety analysis of a reactor. In the literature, no calculations have been performed to show the detailed neutron flux map for the high temperature engineering test reactor (HTTR) because of the limitation of the old neutronic codes and the low performance of the computing system. HTTR is a prismatic type reactor, helium gas-cooled, and graphite-moderated, providing 30 MWth power and up to 950 oC outlet temperature. The present work deals with MCNP6 Monte-Carlo calculation to determine the detailed neutron flux map in the HTTR during normal operation. At first, the calculation of neutron flux at several positions in the reactor was validated by comparing the corresponding reaction rate between the calculation and measurement. After that detailed neutron flux with the small cells of 1cm×1cm×10cm was obtained for the entire reactor core using the fmesh tally of MCNP6 code.

Calculation of shutdown gamma distribution in the high temperature engineering test reactor

Estimation of decay gamma distribution in a reactor core is essential for safely conducting various works after reactor shutdown such as periodic maintenance, shuffling fuel, removing spent fuel at the end of cycle, etc. Because of the dependency on the complex operating history of the reactor, attempting to calculate the decay gamma rays distribution in the core remains a challenge. This study shows a method to calculate the shutdown gamma distribution in the HTTR core by coupling a Monte-Carlo transport calculation code MCNP6 and an activation code ORIGEN2 to take advantage of spatial dependence and transport abilities of MCNP6 and the detailed fission products tracking during burnup and cooling of ORIGEN2. As result, the three-dimensional shutdown gamma distribution in the HTTR core for different cooling times and spatial locations could be obtained accurately.

Lifetime reliability and availability simulation for the electrical system of HTTR coupled to the electricity-hydrogen cogeneration plant

The lifetime reliability and availability for the Electrical Facility of the High Temperature Engineering Test Reactor (HTTR) have been evaluated. The impact of the electrical malfunctions on the Forced Outage Rate (FOR) of the HTTR-based electricity-hydrogen cogeneration has been determined. Normal operation and emergency conditions have been considered for the standard system design and its modified configuration of higher redundancy. The Monte Carlo method has been applied for the uncertainty assessment followed by the Fussell-Vesely importance evaluation. A novel methodological concept has been proposed for the integration of reliability and availability simulations of alternative models corresponding to several options of the system's design considered under various analytical assumptions. It provides a computational framework suitable for the incorporation of the alternative models into one simulation case driven by the common set of reliability data at each iteration step. This enables efficient comparison of the models in terms of reliability, availability, and uncertainty. The results of this work are of high practical importance for the safety and reliability of the HTTR-based cogeneration plant. The former was confirmed by the high availability of safety-related electrical loads, while the latter needs optimization to meet the industry standards concerning FOR.

Experimental study on natural convection in the vertical enclosure of a double coaxial cylinder

Abstract This paper presents results of an experimental study on natural convection in the vertical enclosure of a double coaxial cylinder where the inner wall is at constant heat flux. It is the objective of this experiment to understand the basic characteristics of heat transfer under the condition of natural convection coupled with thermal radiation in a vertical enclosure. The range of Rayleigh numbers based on the width of the double coaxial cylinder is set to be 5.6 × 105 < Ra < 1.04 × 108, the radius ratio is 2.73, the aspect ratio is 3.43 (cylinder length divided by annular gap). The heat transfer coefficient of natural convection coupled with thermal radiation was obtained as function of the Rayleigh number, aspect ratio of the enclosure, and temperature of the hot and cold surfaces. The experimental results show that thermal radiation can’t be neglected in analyzing heat transfer of vertical annulus with natural convection. The results provide the basic data for the design and the performance assessment of the future passive cooling system used in a high-temperature engineering test reactor.

RELAP5-3D modelling of the HTTR-GT/H2 secondary system and turbomachinery

High Temperature Gas-cooled Reactors (HTGRs) can be used for the cogeneration of electricity and hydrogen. The JAEA-operated High-Temperature engineering Test Reactor (HTTR-GT/H2) is exploring using the reactor for electricity and hydrogen production. A RELAP5-3D model of the HTTR-GT/H2 secondary system has been developed using design information. The various components and heat exchangers in the secondary system were investigated, modified, and improved where applicable. With these changes made, the model was run, and the results were compared to the design conditions. The results for the sole-power generation mode were shown to fit the design conditions very well. The largest temperature difference was on the order of 7 K, and the largest pressure difference was on the order of 0.05 MPa. The results for the hydrogen cogeneration mode did not fit the design conditions as well as the other case. The largest temperature difference was about 39 K and the largest pressure difference was about 0.27 MPa at the compressor outlet. The larger differences for the hydrogen cogeneration mode are attributed to the various complex components and the flow being split in the secondary loop.A transient reduction in heat removal capability of the secondary system was investigated. Reactor temperatures were predicted to rise as a result. The core reactivity response due to this increase in temperature is investigated and is expected to add negative reactivity to the reactor.

Improving the safety of the high temperature gas-cooled reactor ‘HTTR’ based on Japan’s new regulatory requirements

Following the Fukushima Daiichi Nuclear Power Plant accident in 2011, the Japan Atomic Energy Agency adapted High-Temperature engineering Test Reactor (HTTR) to meet the new regulatory requirements that began in December 2013. The safety and seismic classifications of the existing structures, systems, and components were discussed to reflect insights regarding High Temperature Gas-cooled Reactors (HTGRs) that were acquired through various HTTR safety tests. Structures, systems, and components that are subject to protection have been defined, and countermeasures to manage internal and external hazards that affect safety functions have been strengthened. Additionally, measures are in place to control accidents that may cause large amounts of radioactive material to be released, as a beyond design basis accident. The Nuclear Regulatory Commission rigorously and appropriately reviewed this approach for compliance with the new regulatory requirements. After nine amendments, the application to modify the HTTR’s installation license that was submitted in November 2014 was approved in June 2020. This response shows that facilities can reasonably be designed to meet the enhanced regulatory requirements, if they reflect the characteristics of HTGRs. We believe that we have established a reference for future development of HTGR.

Calculation of Shutdown Gamma Distribution in the High Temperature Engineering Test Reactor

Calculation of Shutdown Gamma Distribution in the High Temperature Engineering Test Reactor

Improved Multiphysics Model of the High Temperature Engineering Test Reactor for the Simulation of Loss-of-Forced-Cooling Experiments

Improved Multiphysics Model of the High Temperature Engineering Test Reactor for the Simulation of Loss-of-Forced-Cooling Experiments

Economic Dispatch Model of Nuclear High-Temperature Reactor with Hydrogen Cogeneration in Electricity Market

Hydrogen produced without carbon emissions could be a useful fuel as nations look to decarbonize their electricity, transport, and industry sectors. Using the iodine–sulfur (IS) cycle coupled with a nuclear heat source is one method for producing hydrogen without the use of fossil fuels. An economic dispatch model was developed for a nuclear-driven IS system to determine hydrogen sale prices that would make such a system profitable. The system studied is the HTTR-GT/H2, a design for power and hydrogen cogeneration at the Japan Atomic Energy Agency’s High Temperature Engineering Test Reactor. This study focuses on the development of the economic model and the role that input data plays in the final calculated values. Using a historical price duration curve shows that the levelized cost of hydrogen (LCOH) or breakeven sale price of hydrogen would need to be 98.1 JPY/m3 or greater. Synthetic time histories were also used and found the LCOH to be 67.5 JPY/m3. The price duration input was found to have a significant effect on the LCOH. As such, great care should be used in these economic dispatch analyses to select reasonable input assumptions.

Seismic classification of high temperature engineering test reactor

In the late 1980s, the seismic classification of the high temperature engineering test reactor (HTTR) was formulated during its design stage. Because of the HTTR’s lack of operation experiences to sufficiently understand the safety characteristics of high temperature gas cooled reactors (HTGRs) at that time, the seismic classification of commercial light water reactors (LWRs) was applied to HTTR. However, subsequent operation experiences and test results using HTTR made it clear that the seismic classification of commercial LWRs was somewhat too conservative for the HTGRs. The seismic classification of HTTR was reevaluated using the following approach. Class S facilities with high safety requirements ensured that they maintained their function even when the safety functions of Class B and Class C facilities were lost during earthquake-triggered accidents, thereby safely controlling the accident without exposing the members of the public in the vicinity of the reactor to excessive radiation exposure. Consequently, Class S facilities were downgraded compared with the commercial LWRs. Moreover, an evaluation was conducted to verify the validity of the new seismic classification under the conditions where the functions of Class B and Class C facilities are lost. It was confirmed that the reactor was stable and that the evaluated public dose was below the authorized limit. In June 2020, the Nuclear Regulatory Authority in Japan confirmed that the results of the seismic classification conformed to the standard rules of the reactor installation change.

Effect of nuclear heat caused by the 6Li(n,α)T reaction on tritium containment performance of tritium production module in High-Temperature Gas-Cooled reactor for fusion reactors

Tritium is required for research and development activities for the deuterium–tritium (DT) fusion reactor and fueling the DEMOnstration Power Station (DEMO). However, tritium is a very rare nuclide and must be produced artificially. Tritium production by loading Li compounds (Li rods) into burnable poison holes of a high-temperature gas-cooled reactor (HTGR) has been proposed (H. Matsuura, et al., Nucl. Eng. Des. 243 (2012) 95–101.). Al2O3 and Zr are used to prevent tritium leaks. Nuclear reaction heat caused by the nuclear reaction (e.g., 6Li(n,α)T reaction) can cause a spatial temperature profile in the Li rods and may change its tritium containment performance, because Al2O3 and Zr performance strongly depend on these temperatures. The effect of nuclear reaction heat by the 6Li(n,α)T reaction on the tritium containment performance of the Li rods was evaluated by simulation. The temperatures of the Li rods for the high-temperature engineering test reactor (HTTR) and gas turbine high-temperature reactor 300 (GTHTR300) increased by 36 K and 46 K, and the leaked tritium decreased by 32% and 37% via nuclear reaction heat, respectively.