Publisher：Nuclear Engineering and Design  

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.

DOI： 10.1016/j.nucengdes.2023.112795

Scopus

Language：English   Publishing type：Research paper (international conference proceedings)  

Publisher：International Conference on Nuclear Engineering, Proceedings, ICONE  

The conceptual design of a demonstration hydrogen production facility using heat supply from the high-temperature engineering test reactor (HTTR) is being researched and developed at Japan Atomic Energy Agency (JAEA). This facility produces hydrogen with a thermochemical water-splitting Iodine-Sulphur (IS) process that requires high-temperature heat to extract hydrogen efficiently. The HTTR could supply the heat to the IS demonstration plant through the secondary helium loop, coupling the IS plant to the HTTR. The rated operation mode of the HTTR gives helium outlet temperature of 850οC with 660 effective full power days (EFPD). However, in order to achieve hydrogen with high efficiency, the helium outlet temperature should be as high as 950οC. Increasing the outlet temperature increases the reactor core temperature, and as a result the operation time decreases. If the operation time is reduced too much, it is not feasible to use the HTTR as a heat supply for the IS plant. Therefore, the purpose of this study is to estimate the operation time of the HTTR at high operation mode of 950οC to confirm whether it could supply long enough high-temperature heat for the demonstration hydrogen IS plant. The thermal-hydraulic model is also revised using the latest calculation method to improve the accuracy of the temperature distribution in the HTTR. As results, the core temperature increase by about 50 to 100οC when the outlet temperature increases from 850 to 950οC. Although the increase of core temperature makes keff decrease by about 0.3 %Δk/k, the HTTR can still operate approximately 660 EFPD. Therefore, it is possible to use the HTTR for long-term high-temperature heat supply to the demonstration hydrogen production IS plant.

Scopus

Publisher：International Conference on Nuclear Engineering, Proceedings, ICONE  

In a block-type High-Temperature Gas-Cooled Reactor (HTGR), a fixed number of burnable poisons are loaded at the beginning of operation time to control a large amount of excess reactivity. Therefore, it is necessary to evaluate the burnable poison (BP) reactivity worth to achieve the optimum design of the HTGR. In this study, an experiment to measure the burnable poison worth reactivity was conducted at the B-Core of Kyoto University Critical Assembly (KUCA). B-core is solid moderator materials such as polyethylene and graphite combined with fuel plates to form the fuel element. The experiment was performed to measure the reactivity worth of a small cadmium plate (15 × 15 × 0.5 mm) at the B-Core of KUCA. In the experiment, there are 8-unit cells in a fuel element. In this study, the unit cell position of cadmium is called the cadmium unit cell. The experiments were carried out by varying the cadmium plate position inside the cadmium unit cell. This study evaluated the cadmium reactivity worth using the Monte Carlo MVP3. The objective of this study was to evaluate the appropriate results between the measured and the calculation values. The simulation using MVP3 code was conducted by varying the number of batches in the calculation. The results showed that the maximum discrepancy between experimental and calculated results was 24% for 5,000 batches. However, the discrepancy decreased when the number of batches increased to 50,000. The cadmium reactivity worth difference between the experiment and simulation was approximately 18 % depending on the cadmium plate position in the cadmium unit cell.

Scopus

Language：English   Publishing type：Research paper (international conference proceedings)  

DOI： https://doi.org/10.1115/ICONE29-90802

Language：English   Publishing type：Research paper (international conference proceedings)  

DOI： https://doi.org/10.1115/ICONE29-90826

Publisher：Annals of Nuclear Energy  

Power distribution plays a significant role in preventing the fuel temperature from exceeding the safety limit of 1600 °C in high-temperature gas-cooled reactors. Experiments for measuring the power distribution in the graphite-moderated system were performed at the Very High-Temperature Reactor Critical Assembly (VHTRC) facility. The power distribution was determined from the measured Cu activation rate for both the radial and axial distributions. In this study, the pin-wise power distribution of the VHTRC was evaluated with the Monte Carlo MVP3 code. The difference between the calculated and measured results was less than 1 % for the axial and radial distributions. The significant results were concerned with the area around the fuel and reflector regions in the axial direction, where the average discrepancy between the calculated and measured values was 0.8 %. This result showed improved agreement compared to the diffusion calculation that was conducted in the previous study.

DOI： 10.1016/j.anucene.2022.109314

Scopus

Publisher：Annals of Nuclear Energy  

The distribution of the number of fissions has been studied as the consequence of a postulated criticality accident using a randomized model for the material distribution in fuel debris with the power law spectrum (1/fβ). The f is the frequency domain variable while the β value is the parameter related to randomness varying in a certain range. This study investigated the effect of β on the temperature coefficient of reactivity and number of fissions. The calculation results for the temperature coefficient of reactivity show that the shape of the histogram changes from symmetrical to positive skewness by increasing the β value. Moreover, large values of β increase the uncertainty of the number of fissions. The results of this study are worthwhile to realize the behavior of the temperature coefficient of reactivity and number of fissions in the randomized model of fuel debris using the 1/fβ spectrum model.

DOI： 10.1016/j.anucene.2022.109167

Scopus

Publisher：Nuclear Engineering and Design  

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.

DOI： 10.1016/j.nucengdes.2022.111913

Scopus

Language：English   Publishing type：Research paper (scientific journal)  

DOI： https://doi.org/10.1016/j.nucengdes.2022.111913

Language：English   Publishing type：Research paper (scientific journal)  

DOI： https://doi.org/10.1016/j.anucene.2022.109314

Language：English   Publishing type：Research paper (scientific journal)  

DOI： https://doi.org/10.1016/j.anucene.2022.109167

Other Link： https://www.sciencedirect.com/science/article/pii/S030645492200202X

Publisher：International Conference on Nuclear Engineering, Proceedings, ICONE  

The ORIGEN2 code has been used for fuel depletion calculations of many kinds of reactor fuels but there is no library for high temperature gas cooled reactors (HTGRs). A set of the ORIGEN2 library for the HTGR has been established to evaluate the fuel burnup characteristics. In this study, the ORIGEN2 library was prepared for the high temperature engineering test reactor (HTTR). The HTTR is the first Japanese prismatic type HTGR. The burnup dependent neutron spectrum is necessary for generating the ORIGEN2 library. A pin-cell burnup calculation was conducted to obtain the burnup dependent neutron spectrum in the fuel compact of HTTR. Then, the ORIGEN2 library was generated based on the neutron spectrum of the pin cell model. The calculation results that were calculated by the ORIGEN2 code was validated by comparison with a detailed calculation with use of the MVP-BURN code. This code-To-code method was used to validate the ORIGEN2 code calculation because of no assay data of HTTR spent fuels. One of the isotopes that evaluated was 239Pu. The calculation results showed that the amount of 239Pu calculated by ORIGEN2 code was higher about 35 % than that of calculated by the MVP-BURN code. It showed that the ORIGEN2 library using the neutron spectrum of a pincell burnup model was not enough for evaluating burnup characteristics of the HTTR. Therefore, an improvement was performed to evaluate the ORIGEN2 library. In this study, the ORIGEN2 library was generated based on the neutron spectrum of a core burnup calculation. The calculation results showed that the ORIGEN2 code and the MVP-BURN code was in a good agreement. The maximum difference of 239Pu amount between the ORIGEN2 and MVP-BURN became 2.4 %.

DOI： 10.1115/ICONE29-90755

Scopus

Publisher：International Conference on Nuclear Engineering, Proceedings, ICONE  

During operation of the HTTR, hundreds of technical signals and operating conditions must be observed and evaluated to ensure safe operation of the reactor. The accumulated experiment data of the HTTR is not only important for the HTTR operation, but also for the basic development of the HTGR in general. Artificial intelligence (AI) and particularly machine learning (ML) could give the ability to make predictions as well as allow the extraction of key information about physical process from large datasets. Hence, there is a lot of potential to apply AI and ML to predict the operating and safety parameters of the HTTR. In this study, the control rod position of the HTTR is predicted based on ML without using the conventional neutronic codes. The ML with a linear regression algorithm finds a functional relationship between the input dataset and a reference dataset, constructing a function that predicts control rod position from the other operation conditions. As result, the ML gives a good prediction of the HTTR control rod position with less than 5% difference compared to that in the experiment. With increasingly complicated experiments that create a large amount of data, ML is also expected to improve the design and safety analysis of the HTTR in the future.

DOI： 10.1115/ICONE29-90818

Scopus

Language：English   Publishing type：Research paper (scientific journal)  

DOI： https://doi.org/10.1016/j.anucene.2020.107675

Language：English   Publishing type：Research paper (scientific journal)  

DOI： https://doi.org/10.1016/j.anucene.2018.06.014

Language：English   Publishing type：Research paper (scientific journal)  

DOI： https://doi.org/10.1016/j.nima.2018.06.066

Language：English   Publishing type：Research paper (scientific journal)  

DOI： https://doi.org/10.1016/j.egypro.2017.09.476

Language：English   Publishing type：Research paper (scientific journal)  

DOI： https://doi.org/10.1016/j.anucene.2017.04.019

Language：English   Publishing type：Research paper (scientific journal)  

DOI： https://doi.org/10.1016/j.anucene.2016.02.027

Language：English   Publishing type：Research paper (international conference proceedings)  

Language：English   Publishing type：Research paper (international conference proceedings)  

Event date： 2024.3

Language：English   Presentation type：Oral presentation (general)  

Country：Thailand  

Event date： 2024.3

Language：English   Presentation type：Symposium, workshop panel (public)  

Country：Indonesia  

Event date： 2023.9

Language：English   Presentation type：Oral presentation (general)  

Country：Japan  

Event date： 2023.9

Language：English   Presentation type：Oral presentation (general)  

Country：Japan  

Event date： 2023.8

Language：Others  

Country：Indonesia  

Event date： 2023.3

Language：Japanese   Presentation type：Oral presentation (general)  

Venue：Tokyo University   Country：Japan  

Event date： 2022.9

Language：Japanese   Presentation type：Oral presentation (general)  

Country：Japan  

Event date： 2022.9

Language：Japanese   Presentation type：Oral presentation (general)  

Country：Japan  

Event date： 2022.9

Language：Japanese   Presentation type：Oral presentation (general)  

Country：Japan  

Event date： 2022.8

Language：English  

Country：China  

Event date： 2021.12

Language：Japanese  

Venue：Online Presentation   Country：Japan  

Event date： 2021.12

Language：Japanese  

Venue：Online Presentation   Country：Japan  

Event date： 2021.10

Language：English   Presentation type：Oral presentation (general)  

Country：United States  

Event date： 2020.3

Language：English   Presentation type：Oral presentation (general)  

Country：Japan  

Event date： 2019.3

Language：English   Presentation type：Oral presentation (general)  

Country：Japan  

Event date： 2017.9

Language：English   Presentation type：Oral presentation (general)  

Country：Japan  

Event date： 2016.9

Language：English   Presentation type：Oral presentation (general)  

Country：Japan  

Event date： 2016.3

Language：English   Presentation type：Oral presentation (general)  

Country：Japan  

Event date： 2015.9

Language：English   Presentation type：Oral presentation (general)  

Country：Japan  

Language：Indonesian   Presentation type：Oral presentation (general)  

Venue：Consulate General of The Republic of Indonesia in Osaka   Country：Japan  

Language：English   Presentation type：Oral presentation (general)  

Venue：Beijing   Country：China