Kyushu University Academic Staff Educational and Research Activities Database
List of Papers
Nozomu Fujimoto Last modified date:2024.05.07

Professor / Nuclear Energy System / Department of Applied Quantum Physics and Nuclear Engineering / Faculty of Engineering


Papers
1. Hai Quan Ho, Toshiaki Ishii, Satoru Nagasumi, Masato Ono, Yosuke Shimazaki, Etsuo Ishitsuka, Hiroaki Sawahata, Minoru Goto, Irwan Liapto Simanullang, Nozomu Fujimoto, Kazuhiko Iigaki, Feasibility of using BeO rods as secondary neutron sources in the long-life fuel cycle high-temperature gas-cooled reactor, Nuclear Engineering and Design 417 (2024) 112795, 2024.03.
2. Irwan Liapto Simanullang, Shohei Kawaguchi, Nozomu Fujimoto, Toshiaki Ishii, Satoru Nagasumi, Hai Quan Ho, Kunihiro Nakajima, Etsuo Ishitsuka, Kazuhiko Iigaki, Preliminary study of burnup measurement and relative power distribution in the HTTR using gamma-ray measurement, ICNC 2023 - The 12th International Conference on Nuclear Criticality Safety, 2023.10.
3. Hai Quan Ho, Minoru Goto, Satoru Nagasumi, Toshiaki Ishii, Yosuke Shimazaki, Irwan Liapto Simanullang, Nozomu Fujimoto, Etsuo Ishitsuka, Kazuhiko Iigaki, High-temperature operation mode of HTTR for hydrogen production facility, 30th International Conference on Nuclear Engineering(ICONE30)ICONE30-1543, 2023.05.
4. Seiji Yamasaki, Soichiro Moriya, Irwan Liapto Simanullang, Nozomu Fujimoto, Atsushi Sakon, Tadafumi Sano, Takahashi Yoshiyuki, Evaluation of burnable poison reactivity worth at the KUCA graphite moderated system, 30th International Conference on Nuclear Engineering(ICONE30)ICONE30-1271, 2023.05.
5. Irwan Liapto. Simanullang, Katuki. Fukuhara, Keisuke. Morita, Yuji. Fukaya, Ho Hai Quan, Satoru. Nagasumi, Kazuhiko. Iigaki, Etsuo. Ishitsuka, Nozomu Fujimoto, Preparation Method of ORIGEN2 Library for High-Temperature Gas-Cooled Reactors (HTGRs), 29th International Conference on Nuclear Engineering ICONE29, ICONE29-90802, 2022.08.
6. Hai Quan Ho, Satoru Nagasumi, Youske Shimazaki, Toshiaki Ishii, Kazuhiko Iigaki, Minoru Goto, Irwan Liapto Simanullang, Nozomu Fujimoto, Etsuo Ishituka, Prediction of the Operating Control Rod Position of the HTTR with supervised Machine Learning, 29th International Conference on Nuclear Engineering ICONE29, ICONE29- 90826, 2022.08.
7. Hai Quan Ho, Toshiaki Ishii, Satoru Nagasumi, Masato Ono, Yosuke Shimazaki, Etsuo Ishitsuka, Minoru Goto, Irwan Liapto Simanullang, Nozomu Fujimoto, Kazuhiko Iigaki, Calculation of shutdown gamma distribution in the high temperature engineering test reactor, Nuclear Engineering and Design, 10.1016/j.nucengdes.2022.111913, 396 (2022) 111913, 2022.08.
8. Irwan Liapto Simanullang, Naoki Nakagawa, Hai Quan Ho, Satoru Nagasumi, Etsuo Ishitsuka, Kazuhiko Iigaki, Nozomu Fujimoto, Evaluation of power distribution calculation of the very high temperature reactor critical assembly (VHTRC) with Monte Carlo MVP3 code, Annals of Nuclear Energy, 10.1016/j.anucene.2022.109314, Annals of Nuclear Energy 177(2022)109314, 2022.07.
9. Nozomu Fujimoto, Kenichi Tada, Hai Quan Ho, Shimpei Hamamoto, Satoru Nagasumi, Etsuo Ishitsuka, Nuclear data processing code FRENDY: A verification with HTTR criticality benchmark experiments, Annals of Nuclear Energy, 10.1016/j.anucene.2021.108270, 158, 108270, 2021.04.
10. Hai Quan Ho, Nozomu Fujimoto, Shimpei Hamamoto, Satoru Nagasumi, Minoru Goto, Etsuo Ishitsuka, Preparation for restarting the high temperature engineering test reactor: Development of utility tool for auto seeking critical control rod position, Nuclear Engineering and Design, 10.1016/j.nucengdes.2021.111161, 377, 2021.03.
11. Hai Quan Ho, Yuki Honda, Shimpei Hamamoto, Toshiaki Ishii, Shoji Takada, Nozomu Fujimoto, Etsuo Ishitsuka, Promising Neutron Irradiation Applications at the High Temperature Engineering Test Reactor, Journal of Nuclear Engineering and Radiation Science., 10.1115/1.4044529, Vol.6, 2020.06.
12. Hai Quan Ho, Hiroki Ishida, Shimpei Hamamoto, Toshiaki Ishii, Nozomu Fujimoto, Naoyuki Takaki, Etsuo Ishitsuka, Conceptual design of direct production of 99mTc at the high temperature engineering test reactor, Nuclear Engineering and Design, 352, 110174, 2019.07.
13. Hai Quan Ho, Yuki Honda, Shimpei Hamamoto, Toshiaki Ishii, Nozomu Fujimoto, Etsuo Ishitsuka, Feasibility study of large-scale production of iodine-125 at the high temperature engineering test reactor, Applied Radiation and Isotopes, 10.1016/j.apradiso.2018.07.024, 140, 209-214, 2018.10, The feasibility of a large-scale iodine-125 production from natural xenon gas at high-temperature gas-cooled reactors (HTGRs) was investigated. A high-temperature engineering test reactor (HTTR), which is located in Japan at Oarai-machi Research and Development Center, was used as a reference HTGR reactor in this study. First, a computer code based on a Runge-Kutta method was developed to calculate the quantities of isotopes arising from the neutron irradiation of natural xenon gas target. This code was verified with a good agreement with a reference result. Next, optimization of irradiation planning was carried out. As results, with 4 days of irradiation and 8 days of decay, the
125
I production could be maximized and the
126
I contamination was within an acceptable level. The preliminary design of irradiation channels at the HTTR was also optimized. The case with 3 irradiation channels and 20-cm diameter was determined as the optimal design, which could produce approximately 1.8 × 10
5
GBq/y of
125
I production..
14. Yuki Honda, Nozomu Fujimoto, Hiroaki Sawahata, Shoji Takada, Kazuhiro Sawa, Burn-up dependency of control rod position at zero power criticality in the High Temperature Engineering Test Reactor, American Society of Mechanical Engineering, 10.1115/1.4033812, 2017, Vol.3, 011013-1, 2017.01, The High Temperature Engineering Test Reactor (HTTR) is a block type fuel High Temperature Gas-cooled Reactor (HTGR) constructed in Japan, firstly. The operating data of the HTTR with burn-up is very important for developments of HTGRs. Many test data have been collected in the HTTR. Many tests are carried out in low power operation. On the other hand, the full power operation is not enough. There is a temperature distribution in a core in full power operation. The temperature distribution in a core makes it difficult to validate the calculation code. Additionally, it is difficult to measure core temperature in HTTR. On the other hands, the data of the control rod position at criticality at zero power have been measured at the beginning of each operation cycle and the temperature distribution in a core at zero power is uniform. Therefore, the data at zero power are suitable for confirm the characteristics of burn-up and validation of calculation code. In this study, the calculated control rod positions at zero power criticality with burn-up are compared with the experimental data with correlation of core temperature. The calculated results of criticality control rod position at zero power show good agreement to the experimental data. It means that calculated result shows appropriate decrease in uranium and accumulation in plutonium decrease in burnable absorber with burn-up..
15. Yuki Honda, Nozomu Fujimoto, Hiroaki Sawahata, Shoji Takada, Kazuhiro Sawa, Study on Sensitivity of Control Rod Cell Model in Reflector Region of High-Temperature Engineering Test Reactor, American Society of Mechanical Engineering, 10.1115/1.4033813, January 2017, Vol. 3, 011005-1, 2017.01, The high-temperature engineering test reactor (HTTR) is a block-type high-temperature gas-cooled reactor (HTGR). There are 32 control rods (16 pairs) in the HTTR. Six of the pairs of control rods are located in a core region and the remainder are located in a reflector region surrounding the core. Inserting all control rods simultaneously at the reactor scram in a full-power operation presents difficulty in maintaining the integrity of the metallic sleeve of the control rod because the core temperature of the HTTR is too high. Therefore, a two-step control rod insertion method is adopted for the reactor scram. The calculated control rod worth at the first step showed a larger underestimation
than the measured value in the second step, although the calculated results of the excess reactivity tests showed good agreement with the measured result in the criticality tests of the HTTR. It is concluded that a cell model for the control rod guide block with the control rod in the reflector region is not suitable. In addition, in the core calculation, the macroscopic cross section of a homogenized region of the control rod guide block with the control rod is used. Therefore, it would be one of the reasons that the neutron flux distribution around the control rod in control rod guide block in the reflector region cannot be simulated accurately by the conventional cell model. In the conventional cell model, the control rod guide block is surrounded by the fuel blocks only, although the control rods in the reflector region are surrounded by both the fuel blocks and the reflector blocks. The difference of the neutron flux distribution causes the large difference of a homogenized macroscopic cross section set of the control rod guide block with the control rod. Therefore, in this paper, the cell model is revised for the control rod guide block with the control rod in the reflector region to account for the actual configuration around the control rod guide block in the reflector region. The calculated control rod worth at the first step using the improved cell model shows better results than the previous one..
16. H. H. Quan, Koji Morita, Y. Honda, Nozomu Fujimoto, S. Takada, Benchmark study on realized random packing model for coated fuel particles of HTTR using MCNP6, 2017 International Congress on Advances in Nuclear Power Plants: A New Paradigm in Nuclear Power Safety, ICAPP 2017 2017 International Congress on Advances in Nuclear Power Plants, ICAPP 2017 - A New Paradigm in Nuclear Power Safety, Proceedings, 2017.01, The Coated Fuel Particle plays an important role in the excellent safety feature of the High Temperature Gascooled Reactor. However, the random distribution of CFPs also makes the simulation of HTGR fuel become more complicated. The Monte Carlo N-particle (MCNP) code is one of the most well-known codes for validation of nuclear systems; unfortunately, it does not provide an appropriate function to model a statistical geometry explicitly. In order to deal with the stochastic media, a utility program for the random model, namely Realized Random Packing (RRP), has been developed particularly for High Temperature engineering Test Reactor (HTTR). This utility program creates a number of random points in an annular geometry. Then, these random points will be used as the center coordinate of CFPs in the MCNP6 input file and therefore the actual random arrangement of CFPs can be simulated explicitly. First, a pin-cell calculation was carried out to validate the RRP by comparing with Statistical Geometry (STG) model of MVP code. After that, the comparison between the RRP model (MCNP) and STG model (MVP) was shown in whole core criticality calculation, not only for the annular core but also for the fully-loaded core. The comparison of numerical results showed that the RRP model and STG model differed insignificantly in the multiplication factor as expected, regardless of the pin-cell or whole core calculations. In addition, the RRP model did not make the calculation time increase a lot in comparison with the conventional regular model (uniform arrangement)..
17. Daisuke Tochio, Nozomu Fujimoto, Thermal mixing characteristics of helium gas in high-temperature gas-cooled reactor, (I) thermal mixing behavior of helium gas in HTTR, Journal of Nuclear Science and Technology, 10.1080/00223131.2015.1054910, 53, 3, 425-431, 2016.03, The future high-temperature gas-cooled reactor (HTGR) is now designed in Japan Atomic Energy Agency. The reactor has many merging points of helium gas with different temperatures. It is needed to clear the thermal mixing characteristics of helium gas at the pipe in the HTGR from the viewpoint of structure integrity and temperature control. Previously, the reactor inlet coolant temperature was controlled lower than specific one in the high-temperature engineering test reactor (HTTR) due to lack of mixing of helium gas in the primary cooling system. Now, the control system is improved to use the calculated bulk temperature of reactor inlet helium gas. In this paper, thermal–hydraulic analysis on the primary cooling system of the HTTR was conducted to clarify the thermal mixing behavior of helium gas. As a result, it was confirmed that the thermal mixing behavior is mainly affected by the aspect ratio of annular flow path, and it is needed to consider the mixing characteristics of helium gas at the piping design of the HTGR..
18. Yuki Honda, Nozomu Fujimoto, Sawahata Hiroaki, Sawa Kazuhiro, Improvement of cell model for control rod in reflector region of high temperature test engineering reactor, 23rd International Conference on Nuclear Engineering: Nuclear Power - Reliable Global Energy, ICONE 2015 ICONE 2015 - 23rd International Conference on Nuclear Engineering Nuclear Power - Reliable Global Energy, 2015-January, 2015.01, The High Temperature Engineering Test Reactor (HTTR) [1] is a block type fuel High Temperature Gas-cooled Reactor. There are 32 control rods (16 pairs) in the HTTR. The 6 pairs of control rods are inserted into a core region and the others are inserted in a reflector region surrounding the core. The core temperature of the HTTR is too high to insert all control rods simultaneously at reactor scram near full power operation for keeping integrity of control rods metallic sleeve. Therefore, a two-step control rods insertion method for reactor scram is adopted. The reactivity inserted at the two-step control rod insertion method was measured at HTTR criticality tests. The calculated reactivity at the firststep showed larger underestimation than that of the second-step. On the other hand, calculated results of excess reactivity at the HTTR criticality tests showed good agree with tests. It is considered that a cell model for reflector region control rod is not suitable. Therefore, this paper focuses on a new cell model for control rods in a reflector region. In a previous control rod cell model, control rod is surrounded by fuel blocks only. The surrounding condition of the new cell model corresponds to the configuration around the reflector region control rod. The calculated reactivity at the first-step using the new cell model shows better results than previous calculation. It is considered that the new cell model brings appropriate neutron flux distribution around control rods in reflector region..
19. John D. Bess, Nozomu Fujimoto, Benchmark evaluation of start-up and zero-power measurements at the high-temperature engineering test reactor, Nuclear Science and Engineering, 10.13182/NSE14-14, 178, 3, 414-427, 2014.11, Benchmark models were developed to evaluate six cold-critical and two warm-critical, zero-power measurements of the high-temperature engineering test reactor (HTTR). Additional measurements of the subcritical configuration of the fully loaded core, core excess reactivity, shutdown margins, six isothermal temperature coefficients, and axial reaction-rate distributions were also evaluated as acceptable benchmark experiments. Insufficient information is publicly available to develop finely detailed models of the HTTR as much of the design information is still proprietary. The uncertainties in the benchmark models are judged to be of sufficient magnitude to encompass any biases and bias uncertainties incurred through the simplification process used to develop the benchmark models. However, use of the benchmark critical configurations of the HTTR for nuclear data adjustment is not recommended as the impact of these biases has not been addressed with rigorous detail. The impact of any simplification biases, if any, is not expected to significantly impact evaluation of the other reactor physics measurement calculations. Dominant uncertainties in the experimental keff for all core configurations come from uncertainties in the impurity content of the various graphite blocks that compose the HTTR. Monte Carlo calculations of kff are between ∼0.9% and ∼2.7% greater than the benchmark values. Reevaluation of the HTTR models as additional information becomes available could improve the quality of this benchmark and possibly reduce the computational biases. High-quality characterization of graphite impurities would significantly improve the quality of the HTTR benchmark assessment. Simulations of the other reactor physics measurements are in good agreement with the benchmark experiment values. The complete benchmark evaluation details are available in the 2014 edition of the International Handbook of Evaluated Reactor Physics Benchmark Experiments..
20. Atsushi Shimizu, Takayuki Furusawa, Fumitaka Homma, Hiroyuki Inoi, Masayuki Umeda, Masaaki Kondo, Minoru Isozaki, Nozomu Fujimoto, Tatsuo Iyoku, Operation and maintenance experience from the HTTR database, Atomic Energy Society of Japan, 10.1080/00223131.2014.946568, 51, 11-12, 1444-1451, 2014.08, The Japan Atomic Energy Agency has been establishing a database of operation and maintenance experience for the High Temperature Engineering Test Reactor. The objective of this database is to share information from operation and maintenance experience and make use of the knowledge gained in the design, construction, and operation of future High Temperature Gas-cooled Reactors (HTGRs). Between 1997 and 2012, more than 1000 events have been registered in this database system.
This paper describes trends in operation and maintenance events recorded in this database, including experience gained from the Great East Japan Earthquake. The paper also identifies the following significant items that are expected to be useful in the design of future HTGRs: (1) performance degradation of helium gas compressors, (2) malfunction of the reserved shutdown system in the reactivity control system, (3) problems with emergency gas turbine generators, and (4) consequences of the Great East Japan Earthquake..
21. John D. Bess, Nozomu Fujimoto, Benchmark Evaluation of Start-Up and Zero-Power Measurements at the High-Temperature Engineering Test Reactor, Nuclear Science and Engineering, 178, 414-427, 2014.06, Benchmark models were developed to evaluate six cold-critical and two warm-critical, zeropower
measurements of the high-temperature engineering test reactor (HTTR). Additional measurements
of the subcritical configuration of the fully loaded core, core excess reactivity, shutdown margins, six
isothermal temperature coefficients, and axial reaction-rate distributions were also evaluated as
acceptable benchmark experiments. Insufficient information is publicly available to develop finely
detailed models of the HTTR as much of the design information is still proprietary. The uncertainties in
the benchmark models are judged to be of sufficient magnitude to encompass any biases and bias
uncertainties incurred through the simplification process used to develop the benchmark models.
However, use of the benchmark critical configurations of the HTTR for nuclear data adjustment is not
recommended as the impact of these biases has not been addressed with rigorous detail. The impact of
any simplification biases, if any, is not expected to significantly impact evaluation of the other reactor
physics measurement calculations. Dominant uncertainties in the experimental keff for all core
configurations come from uncertainties in the impurity content of the various graphite blocks that
compose the HTTR. Monte Carlo calculations of keff are between *0.9% and *2.7% greater than the
benchmark values. Reevaluation of the HTTR models as additional information becomes available could improve the quality of this benchmark and possibly reduce the computational biases. High-quality characterization of graphite impurities would significantly improve the quality of the HTTR benchmark assessment. Simulations of the other reactor physics measurements are in good agreement with the benchmark experiment values. The complete benchmark evaluation details are available in the 2014 edition of the International Handbook of Evaluated Reactor Physics Benchmark Experiments..
22. Masanori Shinohara, Yoshitomo Inaba, Shimpei Hamamoto, Nozomu Fujimoto, Establishment of floating support technology applied to high-temperature components and piping in HTTR, journal of nuclear science and technology, 10.1080/00223131.2014.967734, 51, 1398-1406, 2014.01, In the primary cooling system of the High Temperature Engineering Test Reactor (HTTR) with an outlet coolant temperature of 950°C, high-temperature components and piping such as an intermediate heat exchanger and coaxial double piping reach very high temperature, and large and complex thermal displacements arise in them. In order not only to absorb the thermal displacements but also to withstand earthquakes, the HTTR has adopted a new three-dimensional floating support system. In the limited space of the containment vessel, the support system can support the components and piping's own weights and follow the thermal displacements and have seismic capacity. On the other hand, the adoption of the support system was unprecedented in nuclear plants. Thus, the effectiveness of the support system was demonstrated through the HTTR operation. In this paper, by using the HTTR operation data, the thermal displacement behavior of the high-temperature components and piping is investigated, and the behavior and characteristics are simulated numerically. In addition, the aftermath of the Great East Japan Earthquake on the HTTR is confirmed. As a result, the effectiveness of the three-dimensional floating support system adopted by the HTTR is verified..
23. Kuniyoshi Takamatsu, Kazuhiro Sawa, Kazuhiko Kunitomi, Ryutaro Hino, Masuro Ogawa, Yoshihiro Komori, Toshio Nakazawa, Tatsuo Iyoku, Nozomu Fujimoto, Tetsuo Nishihara, Masayuki Shinozaki, High-temperature continuous operation of the HTTR, Transactions of the Atomic Energy Society of Japan, 10.3327/taesj.J11.020, 10, 4, 290-300, 2011.12, A high-temperature gas-cooled reactor (HTGR) is a graphite-moderated and helium gas-cooled reactor. It is particularly attractive due to its capability of producing high-temperature helium gas, and its passive and inherent safety features. To enable nuclear energy application to a wide range of heat process industries, Japan Atomic Energy Agency (JAEA) has continued extensive effort for the development of the HTGR using the high-temperature engineering test reactor (HTTR), which is the first HTGR in Japan with a thermal power of 30 MW, and operates it at the site of the JAEA's Oarai Research and Development Center. The HTTR has successfully completed a full-power high-temperature (950°C) continuous operation for 50 days from January to March in 2010. Through this operation, the potential of a stable high-temperature heat supply to heat application systems, such as a hydrogen production system, was demonstrated. This paper presents the operation results including reactor characteristics..
24. Minoru Goto, Shusaku Shiozawa, Nozomu Fujimoto, Shigeaki Nakagawa, Yasuyuki Nakao, Experimental validation of effectiveness of rod-type burnable poisons on reactivity control in HTTR, Nuclear Engineering and Design, 240, 2994-2998, 2010.10, In block-type high temperature gas-cooled reactors (HTGRs), insertion depth of control rods (CRs) into a core should be retained shallow to keep fuel temperature below 1495 ◦C through a burnup period, and hence excess reactivity should be reduced through a different method. Loading burnable poisons (BPs) into the core is considered as a method to resolve this problem as in case of light water reactors (LWRs).
Effectiveness of BPs on reactivity control in LWRs has been validated by experimental data, however, this has not been done yet for HTGRs, because there was not enough burnup characteristics data for HTGRs required for the validation. The High Temperature Engineering Test Reactor (HTTR) is a block-type HTGRs and it adopts rod-type BPs to control reactivity. The HTTR has been operated up to middle burnup, and thereby the experimental data was expected to show effect of the BPs on the reactivity control. Hence, in order to validate effectiveness of rod-type BPs on reactivity control in the HTTR, we investigated on the HTTR results whether the BPs have functioned as designed. As a result, the CRs insertion depth has been retained shallow within allowable range, and then effectiveness of rod-type BPs on reactivity control in the HTTR was validated..
25. Minoru Goto, Nozomu Fujimoto, Shigeaki Nakagawa, Analysis for HTTR burnup characteristics, International Congress on Advances in Nuclear Power Plants 2009, ICAPP 2009, 1, 309-314, 2009.05, Comparison of burnup characteristics of the High Temperature Engineering Test Reactor (HTTR) were conducted between experimental data and analytical results, and then applicability of a three dimensional whole core burnup calculation method to burned block-type HTGRs was studied. The calculations were performed by the SRAC/COREBN based on diffusion theory. We focused on changes in control rods (CRs) position with burnup as one of the HTTR burnup characteristics, and analyzed that in rated power with 30MW and zero power, respectively. In the HTTR experiments, different tendencies were observed in the changes in CRs position with burnup between rated and zero power. CRs position in rated power decreased with burnup, namely, insertion depth of CRs into a core increased with burnup. Meanwhile that in zero power was almost constant through burnup. In the analysis, similar results to the above were obtained. Thus, applicability of the three dimensional whole core burnup calculation method to burned block-type HTGRs was confirmed by experimental data..
26. Daisuke Tochio, Junya Sumita, Eiji Takada, Nozomu Fujimoto, Shigeaki Nakagawa, Evaluation of fuel temperature on high temperature test operation at high temperature gas-cooled reactor 'HTTR', Transactions of the Atomic Energy Society of Japan, 10.3327/taesj2002.5.57, 5, 1, 57-67, 2006.01, High Temperature Engineering Test Reactor (HTTR) of high temperature gas-cooled reactor at Japan Atomic Energy Agency (JAEA) achieved the reactor outlet coolant temperature of 950°C for the first time in the world at Apr. 19, 2004. To ensure the thermal integrity of fuel in high temperature test operation, it is necessary that fuel temperature is designed appropriately by fuel temperature designing method, and that estimated maximum fuel temperature is lower than the thermal limit temperature. In this report, by constructing newly a realistic core-shape representing model, the current fuel temperature estimation model is improved. Moreover fuel temperature in high-temperature test operation is estimated with the newly-constructed model, and it is confirmed that estimated maximum fuel temperature in high temperature test operation is lower than the thermal limit temperature..
27. Nozomu Fujimoto, Kiyonobu Yamashita, Naoki Nojiri, Mituo Takeuchi, Shingo Fujisaki, Masaaki Nakano, Annular core experiments in HTTR's start-up core physics tests, Nuclear Science and Engineering, 10.13182/NSE03-79, 150, 3, 310-321, 2005.01, Annular cores were formed in start-up core physics tests of the High Temperature Engineering Test Reactor (HTTR) to obtain experimental data for verification of design codes. The first criticality, control rod (CR) positions at critical conditions, neutron flux distribution, excess reactivity, etc., were measured as representative data. These data were evaluated with the MVP Monte Carlo code, which can consider directly the heterogeneity of coated fuel particles (CFPs) distributed randomly in fuel compacts. It was made clear that the heterogeneity effect of CFPs on keff's for annular cores is smaller than that for fully loaded cores. The measured and the calculated k eff's agreed with each other with differences
28. Naoki Nojiri, Satoshi Shimakawa, Nozomu Fujimoto, Minoru Goto, Characteristic test of initial HTTR core, Nuclear Engineering and Design, 10.1016/j.nucengdes.2004.08.015, 233, 1-3, 283-290, 2004.10, This paper describes the results of core physics test in start-up and power-up of the HTTR. The tests were conducted in order to ensure performance and safety of the high temperature gas cooled reactor, and was carried out to measure the critical approach, the excess reactivity, the shutdown margin, the control rod worth, the reactivity coefficient, the neutron flux distribution and the power distribution. The expected core performance and the required reactor safety characteristics were verified from the results of measurements and calculations..
29. Eiji Takada, Shigeaki Nakagawa, Nozomu Fujimoto, Daisuke Tochio, Core thermal-hydraulic design, Nuclear Engineering and Design, 10.1016/j.nucengdes.2004.07.009, 233, 1-3, 37-43, 2004.10, The core thermal-hydraulic design for the HTTR is carried out to evaluate the maximum fuel temperature at normal operation and anticipated operation occurrences. To evaluate coolant flow distribution and maximum fuel temperature, we use the experimental results such as heat transfer coefficient, pressure loss coefficient obtained by mock-up test facilities. Furthermore, we evaluated hot spot factors of fuel temperatures conservatively. As the results of the core thermal-hydraulic design, an effective coolant flow through the core of 88% of the total flow, is achieved at minimum. The maximum fuel temperature appears during the high-temperature test operation, and reaches 1492 °C for the maximum through the burn-up cycle, which satisfies the design limit of 1495 °C at normal operation. It is also confirmed that the maximum fuel temperature at any anticipated operation occurrences does not exceed the fuel design limit of 1600 °C in the safety analysis. On the other hand, result of re-evaluation of analysis condition and hot spot factors based on operation data of the HTTR, the maximum fuel temperature for 160 effective full power operation days is estimated to be 1463 °C. It is confirmed that the core thermal-hydraulic design gives conservative results..
30. Nozomu Fujimoto, Yukio Tachibana, Akio Saikusa, Masayuki Shinozaki, Minoru Isozaki, Tatuo Iyoku, Experience of HTTR construction and operation - Unexpected incidents, Nuclear Engineering and Design, 10.1016/j.nucengdes.2004.08.014, 233, 1-3, 273-281, 2004.10, From a viewpoint of heat leakage, there were two incidents during HTTR power-rise-tests. One was a temperature rise of the primary upper shielding, and the other was a temperature rise of the core support plate. Causes of the both incidents were small amount of helium flow in structures. For the temperature rise of the primary upper shielding, countermeasures to reduce the small amount of helium flow, enhancement of heat release and installation of thermal insulator were taken. For the temperature rise of the core support plate, temperature evaluations were carried out again considering the small amount of helium flow and design temperature of the core support plate was revised. By these countermeasures, the both temperatures were kept below their limits..
31. Nozomu Fujimoto, Naoki Nojiri, Hiroei Ando, Kiyonobu Yamashita, Nuclear design, Nuclear Engineering and Design, 10.1016/j.nucengdes.2004.07.008, 233, 1-3, 23-36, 2004.10, The high-temperature engineering test reactor (HTTR) has been designed for an outlet temperature of 950 °C. That is the highest temperature in the world for a block-type high-temperature gas-cooled reactor (HTGR). The functions of the reactivity control system are determined considering the operational conditions, and the reactivity balance is planned so that the design requirements are fully satisfied. Moreover, the reactivity coefficients are evaluated to confirm the safety characteristics of the reactor. The power distribution in the core was optimized by changing the uranium enrichment to maintain the fuel temperature at less than the limit (1600 °C). Deviation from the optimized distribution due to the burnup of fissile materials was avoided by flattening time-dependent changes in local reactivities. Flattening was achieved by optimizing the specifications of the burnable poisons. The original nuclear design model had to be modified based on the first critical experiments. The Monte Carlo code MVP was also used to predict criticality of the initial core. The predicted excess reactivities are now in good agreement with the experimental results..
32. Nozomu Fujimoto, Naoki Nojiri, Kiyonobu Yamashita, Validation of the nuclear design code system for the HTTR using the criticality assembly VHTRC, Nuclear Engineering and Design, 10.1016/j.nucengdes.2004.08.005, 233, 1-3, 155-162, 2004.10, The high temperature engineering test reactor is the first block-type HTGR designed for a 950 °C outlet gas temperature which uses low-enriched uranium fuel with burnable poison rod. For validation of the nuclear design code system for the HTTR, a critical assembly of VHTRC had been constructed. The calculation uncertainties of effective multiplication factor, neutron flux distribution, burnable poison reactivity worth, and control rod worth, temperature coefficients were evaluated. Calculation accuracy of a Monte Carlo code is also evaluated..
33. Takeshi Takeda, Shigeaki Nakagawa, Fumitaka Honma, Eiji Takada, Nozomu Fujimoto, Safety shutdown of the high temperature engineering test reactor during loss of off-site electric power simulation test, Journal of Nuclear Science and Technology, 10.1080/18811248.2002.9715285, 39, 9, 986-995, 2002.01, The high temperature engineering test reactor (HTTR) is a graphite-moderated and helium-gas-cooled reactor, which is the first high temperature gas-cooled reactor in Japan. The HTTR achieved its first full power of 30 MW at rated operation on December 7 in 2001. In the rise-to-power test of the HTTR, simulation test of anticipated operational occurrence with scram was carried out by manual shutdown of off-site electric power from 30 MW operation. Because helium circulators and water pumps coasted down immediately after the loss of off-site electric power, mass flow rates of helium and water decreased to the scram points. Sixteen pairs of control rods were inserted at two-steps into the core by gravity within the design criterion of 12 s. In 51 s after the loss of off-site electric power, the auxiliary cooling system started up by supplying electricity from emergency power feeders. In 40 min after the startup of the auxiliary cooling system, one of two auxiliary helium circulators stopped for reducing thermal stresses of core graphite components such as fuel blocks. Temperature of hot plenum block among core graphite structures decreased continuously after the startup of the auxiliary cooling system. Blackout sequences of the HTTR dynamic components were in accordance with the design. As a result of the loss of off-site electric power simulation test, it was confirmed that the HTTR shuts down safely after the scram..
34. Nozomu Fujimoto, Masaaki Nakano, Mitsuo Takeuchi, Shingo Fujisaki, Kiyonobu Yamashita, Start-up core physics tests of high temperature engineering test reactor (HTTR), (II). First criticality by an annular form fuel loading and its criticality prediction method, Nippon Genshiryoku Gakkaishi/Journal of the Atomic Energy Society of Japan, 10.3327/jaesj.42.458, 42, 5, 458-464, 2000.01, The HTTP has achieved the first criticality on 11/10/98. The fuels were loaded into the core from outer region to inner region to obtain characteristic data of annular cores. The annular core is expected to be a core type of future HTGR. Fuel loading schedule was planned based on preliminary calculations by Monte Carlo method. These calculations predicted the first criticality at 16 ± 1 columns. However, the reactor achieved the first criticality at 19 columns. The first criticality was re-predicted by comparing measured and calculated 1/M curves. The calculated 1/M curves were obtained for different critical mass adjusting some parameters such as the amount of impurities, etc. The method is called "1/M sandwich method". This method well predicted the number of fuel columns at the first criticality. The combination of this method with Monte Carlo calculation was a rational method for predicting the first criticality. It was confirmed that Monte Carlo calculation could be used for the evaluation of HTTR with
35. Kiyonobu Yamashita, Nozomu Fujimoto, Mituo Takeuchi, Shingo Fujisaki, Masaaki Nakano, Masayuki Umeda, Takeshi Takeda, Haruyoshi Mogi, Toshiyuki Tanaka, Startup core physics tests of High Temperature Engineering Test Reactor(HTTR), (I) test plan, fuel loading and nuclear characteristics tests, Nippon Genshiryoku Gakkaishi/Journal of the Atomic Energy Society of Japan, 10.3327/jaesj.42.30, 42, 1, 30-42, 2000.01, High Temperature Engineering Test Reactor (HTTR) is a graphite-moderated, helium-cooled reactor which has 30 MW of thermal power and 950°C of outlet coolant-gas temperature. The fuel loading of the HTTR was started on July 1, 1998, from the core periphery. The first criticality was attained in annular type core of 19 columns on Nov. 10, 1998. The startup core physics tests consisted mainly of tests for licensing and tests for establishing the technology bases necessary for HTGRs. It was confirmed in the former tests that all fuel blocks are loaded in certain positions and the excess reactivity is less than the limit. Experimental data for the annular core are obtained in the latter tests. Also, it was confirmed that the inverse kinetics method and delayed integral counting method are useful for the measurement of scram reactivity even if it takes about 10s for rod insertions. Furthermore, control rod worth curve, axial neutron flux distribution, etc. were measured to grasp the core performance. All tests planned in the startup core physics tests had been successfully performed and were completed on Jan. 21, 1999. It was confirmed from the tests that the HTTR was capable to step up to the power ascension tests..
36. Kiyonobu Yamashita, Mituo Takeuchi, Nozomu Fujimoto, Shingo Fujisaki, Masaaki Nakano, Naoki Nojiri, Seiji Tamura, Measuring method of reactivity worth of control rod with long falling time by IKRD technique., Nippon Genshiryoku Gakkaishi/Journal of the Atomic Energy Society of Japan, 10.3327/jaesj.41.35, 41, 1, 35-38, 1999.01.
37. Kiyonobu Yamashita, Kazumi Tokuhara, Nozomu Fujimoto, Weapons-grade plutonium burning with high-temperature gas-cooled reactors using plutonium burner balls and thorium breeder balls, Nuclear Science and Engineering, 10.13182/NSE97-A24460, 126, 1, 94-100, 1997.01, A concept for a new reactor system is developed where weapons-grade plutonium can be made worthless for weapons use. It is a pebble bed-type high-temperature gas-cooled reactor that uses plutonium burner ball and thorium breeder ball fuels. The residual amount of 239Pu in spent plutonium balls becomes 135Xe..
38. Kiyonobu Yamashita, Ryuichi Shindo, Isao Murata, So Maruyama, Nozomu Fujimoto, Takeshi Takeda, Nuclear design of the high-temperature engineering test reactor (HTTR), Nuclear Science and Engineering, 10.13182/NSE96-A24156, 122, 2, 212-228, 1996.02, The high-temperature engineering test reactor has been designed whose outlet gas temperature is 950°C. That is the highest temperature in the world for a block-type high-temperature gas-cooled reactor. The power distribution in the core was optimized by changing the uranium enrichment to maintain the fuel temperature at less than the limit (1600°C). Deviation from the optimized distribution due to the burnup of fissile materials was avoided by flattening time-dependent changes in local reactivities. Flattening was achieved by optimizing the specifications of the burnable poisons. Control rod destruction of the optimized power distribution was avoided by limiting the depth of insertion. The insertion depth of the control rods is limited by reducing the excess reactivity of the whole core by the burnable poisons to the minimum value necessary for operations..
39. Nozomu Fujimoto, Kiyonobu Yamashita, Kazumi Tokuhara, HTGR type minor actinide transmutation reactor, Nippon Genshiryoku Gakkaishi/Journal of the Atomic Energy Society of Japan, 10.3327/jaesj.38.304, 38, 4, 304-306, 1996.01.
40. So Maruyama, Nozomu Fujimoto, Yukio Sudo, Tomoyuki Murakami, Sadao Fujii, Evaluation of core thermal and hydraulic characteristics of HTTR, Nuclear Engineering and Design, 10.1016/0029-5493(94)90084-1, 152, 1-3, 183-196, 1994.11, Japan Atomic Energy Research Institute has started the development of the high temperature engineering test reactor (HTTR), a graphite-moderated, helium gas-cooled reactor with 30 MW thermal power and maximum outlet coolant temperature of 950 °C. This paper describes the core thermal and hydraulic (T/H) design procedure, including the validation of the computer code system, design criteria pertaining to the fuel design limit and the evaluated core T/H charateristics. The core T/H design of the HTTR has been carried out considering the specific characteristics of the core structure and the fuel based on R&D results. The coolant flow rate and temperature distribution are evaluated by the flow network analysis code flownet. The fuel temperature distribution is evaluated by the fuel temperature analysis code temdim with multi-cylindrical model using hot spot factors. Fuel design limit for anticipated operational occurrences and fuel temperature limit for normal operation are specified at 1600°C and 1495°C, respectively based on experimental results. Several design considerations are also adopted to realize a high reactor outlet coolant temperature of 950°C. As a result of core T/H design, the effective core flow rate and maximum fuel temperature during the high temperature test operation are 88% and 1492°C, respectively..
41. Shinzo SAITO, Toshiyuki TANAKA, Yukio SUDO, Osamu BABA, Masami SHINDO, Shusaku SHIOZAWA, Haruyoshi MOGI, Minoru OKUBO, Noboru ITO, Ryuichi SHINDO, Noriaki KOBAYASHI, Ryoichi KURIHARA, Kimio HAYASHI, Kazuhiko HADA, Yuji KURATA, Kiyonobu YAMASHITA, Kozo KAWASAKI, Tatsuo IYOKU, Kazuhiko KUNITOMI, So MARUYAMA, Masahiro ISHIHARA, Kazuhiro SAWA, Nozomu FUJIMOTO, Isao MURATA, Shigeaki NAKAGAWA, Yukio TACHIBANA, Tetsuo NISHIHARA, Shinichi OSHITA, Masayuki SHINOZAKI, Takeshi TAKEDA, Shigeaki SAKABA, Akio SAIKUSA, Yujiro TAZAWA, Yoshio FUKAYA, Hiroshi NAGAHORI, Takayuki KIKUCHI, Satoshi KAWAJI, Minoru ISOZAKI, Shinjiro MATSUZAKI, Iwao SAKAMA, Kunio HARA, Noriaki UEDA, Shigeru KOKUSEN, Design of High Temperature Engineering Test Reactor (HTTR), JAERI-1332, 1994.09, Construction of High Temperature Engineering Test Reactor (HTTR) is now underway to establish and upgrade basic technologies for HTGRs and to conduct innovative basic research at high temperatures. The HTTR is a graphite-moderated and helium gas-cooled reactor with 30MW in thermal output and outlet coolant temperature of 850 °C for rated operation and 950 °C for high temperature test operation. It is planned to conduct various irradiation tests for fuels and materials, safety demonstration tests and nuclear heat application tests.
JAERI received construction permit of HTTR reactor facility in February 1990 after 22 months of safety review. This report summarizes evaluation of nuclear and thermal-hydraulic characteristics, design outline of major systems and components, and also includes relating R&D result and safety evaluation. Criteria for judgment, selection of postulated events, major analytical conditions for anticipated operational occurrences and accidents, computer codes used in safety analysis and evaluation of each event are presented in the safety evaluation..
42. Isao Murata, Kiyonobu Yamashita, So Maruyama, Ryuichi Shindo, Nozomu Fujimoto, Yukio Sudo, Evaluation Of Local Power Distribution With Fine-Mesh Core Model For High Temperature Engineering Test Reactor (HTTR), journal of nuclear science and technology, 10.1080/18811248.1994.9735115, 31, 1, 62-72, 1994.01, In the high temperature gas-cooled reactors (HTGRs), the radial and axial heterogeneity resulted from a combination of fuel rods, burnable poison rods, block end graphite and so on causes local power peakings which increase the fuel temperature locally. An method was developed for calculating the local power and the fuel temperature distributions. This method deals with all heterogeneity effects of a whole core in the radial and axial directions with a design code system including a vectorized 3-dimensional diffusion code. The uncertainty of the method had been evaluated through the analyses of the power distribution obtained by critical experiments with the Very High Temperature Reactor Critical Assembly (VHTRC). The difference was less than 3% between the calculated and measured power distributions. From the results, it was confirmed that this method could predict the local power distribution of the HTGR with high accuracy. This method was applied to the evaluation of the fuel temperature of the HTTR. It was shown that the maximum fuel temperature would be lower than the design limit of 1,495°C for the normal operation and that of 1,600°C for the anticipated operational transients..
43. So Maruyama, Kiyonobu Yamashita, Nozomu Fujimoto, Isao Murata, Yukio Sudo, Tomoyuki Murakami, Sadao Fujii, Evaluation of hot spot factors for thermal and hydraulic design of HTTR, journal of nuclear science and technology, 10.1080/18811248.1993.9734606, 30, 11, 1186-1194, 1993.01, High Temperature Engineering Test Reactor (HTTR) is a graphite-moderated and helium gas-cooled reactor with 30 MW in thermal power and 950°C in reactor outlet coolant temperature. One of the major items in thermal and hydraulic design of the HTTR is to evaluate the maximum fuel temperature with a sufficient margin from a viewpoint of integrity of coated fuel particles. Hot spot factors are considered in the thermal and hydraulic design to evaluate the fuel temperature not only under the normal operation condition but also under any transient condition conservatively. This report summarizes the items of hot spot factors selected in the thermal and hydraulic design and their estimated values, and also presents evaluation results of the thermal and hydraulic characteristics of the HTTR briefly..
44. Kiyonobu Yamashita, So Maruyama, Isao Murata, Ryuichi Shindo, Nozomu Fujimoto, Yukio Sudo, Tetsuo Nakata, Kazumi Tokuhara, Optimization of power distribution to achieve outlet gas-coolant temperature of 950°c for httr, journal of nuclear science and technology, 10.1080/18811248.1992.9731553, 29, 5, 472-481, 1992.05, This report presents the optimization result with respect to the spatial power distribution of the High Temperature engineering Test Reactor (HTTR) core to achieve a high outlet coolant gas temperature of 950° C. At first, the power distribution optimization procedure was developed to achieve a high outlet coolant gas temperature while maintaining the fuel temperature as low as possible. Secondarily, the optimization procedure thus developed was applied for the power distribution design of the HTTR core. The maximum nominal fuel temperature was reduced about 300°C through the optimization and was 1,321°C. By the power distribution optimization, the maximum fuel temperature was maintained less than the fuel temperature design limit of 1,600°C, even accounting for the temperature increase at the hot spot and the anticipated operational occurrences..
45. Kazuhiko Kudo, Nozomu Fujimoto, Masao Ohta, Takaaki Ohsawa, Yasuyuki Nakao, Yoshikuni Shinohara, CONTROL SYSTEM DESIGN OF VERY HIGH TEMPERATURE GAS COOLED REACTOR FOR START, STOP AND LOAD FOLLOW OPERATIONS., Memoirs of the Kyushu University, Faculty of Engineering, 47, 1, 75-84, 1987.03, Very High Temperature Gas Cooled Reactor (VHTR) is one of future reactors for multi-purpose use. The reactor core has large amount of graphite moderator with much heat capacity. So the response of the core outlet gas temperature has very long time constant for disturbances. In this report, the control system design has been studied for the start up, the shut down and the load follow operations using PID controllers..