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JAEA Reports

Report of summer holiday practical training 2020; Feasibility study on nuclear battery using HTTR core; Feasibility study for nuclear design, 3

Ishitsuka, Etsuo; Mitsui, Wataru*; Yamamoto, Yudai*; Nakagawa, Kyoichi*; Ho, H. Q.; Ishii, Toshiaki; Hamamoto, Shimpei; Nagasumi, Satoru; Takamatsu, Kuniyoshi; Kenzhina, I.*; et al.

JAEA-Technology 2021-016, 16 Pages, 2021/09

JAEA-Technology-2021-016.pdf:1.8MB

As a summer holiday practical training 2020, the feasibility study for nuclear design of a nuclear battery using HTTR core was carried out, and the downsizing of reactor core were studied by the MVP-BURN. As a result, it is clear that a 1.6 m radius reactor core, containing 54 (18$$times$$3 layers) fuel blocks with 20% enrichment of $$^{235}$$U, and BeO neutron reflector, could operate continuously for 30 years with thermal power of 5 MW. Number of fuel blocks of this compact core is 36% of the HTTR core. As a next step, the further downsizing of core by changing materials of the fuel block will be studied.

JAEA Reports

Mesh effect around burnable poison rod of cell model for HTTR fuel block

Fujimoto, Nozomu*; Fukuda, Kodai*; Honda, Yuki*; Tochio, Daisuke; Ho, H. Q.; Nagasumi, Satoru; Ishii, Toshiaki; Hamamoto, Shimpei; Nakano, Yumi*; Ishitsuka, Etsuo

JAEA-Technology 2021-008, 23 Pages, 2021/06

JAEA-Technology-2021-008.pdf:2.62MB

The effect of mesh division around the burnable poison rod on the burnup calculation of the HTTR core was investigated using the SRAC code system. As a result, the mesh division inside the burnable poison rod does not have a large effect on the burnup calculation, and the effective multiplication factor is closer to the measured value than the conventional calculation by dividing the graphite region around the burnable poison rod into a mesh. It became clear that the mesh division of the graphite region around the burnable poison rod is important for more appropriately evaluating the burnup behavior of the HTTR core..

Journal Articles

High temperature gas-cooled reactors

Takeda, Tetsuaki*; Inagaki, Yoshiyuki; Aihara, Jun; Aoki, Takeshi; Fujiwara, Yusuke; Fukaya, Yuji; Goto, Minoru; Ho, H. Q.; Iigaki, Kazuhiko; Imai, Yoshiyuki; et al.

High Temperature Gas-Cooled Reactors; JSME Series in Thermal and Nuclear Power Generation, Vol.5, 464 Pages, 2021/02

As a general overview of the research and development of a High Temperature Gas-cooled Reactor (HTGR) in JAEA, this book describes the achievements by the High Temperature Engineering Test Reactor (HTTR) on the designs, key component technologies such as fuel, reactor internals, high temperature components, etc., and operational experience such as rise-to-power tests, high temperature operation at 950$$^{circ}$$C, safety demonstration tests, etc. In addition, based on the knowledge of the HTTR, the development of designs and component technologies such as high performance fuel, helium gas turbine and hydrogen production by IS process for commercial HTGRs are described. These results are very useful for the future development of HTGRs. This book is published as one of a series of technical books on fossil fuel and nuclear energy systems by the Power Energy Systems Division of the Japan Society of Mechanical Engineers.

JAEA Reports

Report of summer holiday practical training 2019; Feasibility study on nuclear battery using HTTR core; Feasibility study for nuclear design, 2

Ishitsuka, Etsuo; Nakashima, Koki*; Nakagawa, Naoki*; Ho, H. Q.; Ishii, Toshiaki; Hamamoto, Shimpei; Takamatsu, Kuniyoshi; Kenzhina, I.*; Chikhray, Y.*; Matsuura, Hideaki*; et al.

JAEA-Technology 2020-008, 16 Pages, 2020/08

JAEA-Technology-2020-008.pdf:2.98MB

As a summer holiday practical training 2019, the feasibility study for nuclear design of a nuclear battery using HTTR core was carried out, and the $$^{235}$$U enrichment and burnable poison of the fuel, which enables continuous operation for 30 years with thermal power of 5 MW, were studied by the MVP-BURN. As a result, it is clear that a fuel with $$^{235}$$U enrichment of 12%, radius of burnable poison and natural boron concentration of 1.5 cm and 2wt% are required. As a next step, the downsizing of core will be studied.

JAEA Reports

Study on control rod model in HTTR core analysis

Nagasumi, Satoru; Matsunaka, Kazuaki*; Fujimoto, Nozomu*; Ishii, Toshiaki; Ishitsuka, Etsuo

JAEA-Technology 2020-003, 13 Pages, 2020/05

JAEA-Technology-2020-003.pdf:1.5MB

The influence of the control rod model on the nuclear characteristics of the HTTR has been evaluated, by creating detailed control rod model, in which geometric shape was close to that of the actual control rod structure, in MVP code. According to refinement of the control rod model, the critical control rod position was 11 mm lower than that of the conventional model, and this was close to the measured value of 1775 mm. The reactivity absorbed by the shock absorber located at the tip of the control rod was 0.2%$$Delta$$k/k, and this was 14 mm difference at the critical control rod position. Considering the effect of refinement of the control rod and the effect of the shock absorber, the correction amount for the analysis value in SRAC code due to the shape effect of the control rod, is -0.05%$$Delta$$k/k in reactivity, and -3 mm in the critical control rod position at low temperature criticality.

JAEA Reports

Report of summer holiday practical training 2018; Feasibility study on nuclear battery using HTTR core; Feasibility study for nuclear design

Ishitsuka, Etsuo; Matsunaka, Kazuaki*; Ishida, Hiroki*; Ho, H. Q.; Ishii, Toshiaki; Hamamoto, Shimpei; Takamatsu, Kuniyoshi; Kenzhina, I.*; Chikhray, Y.*; Kondo, Atsushi*; et al.

JAEA-Technology 2019-008, 12 Pages, 2019/07

JAEA-Technology-2019-008.pdf:2.37MB

As a summer holiday practical training 2018, the feasibility study for nuclear design of a nuclear battery using HTTR core was carried out. As a result, it is become clear that the continuous operations for about 30 years at 2 MW, about 25 years at 3 MW, about 18 years at 4 MW, about 15 years at 5 MW are possible. As an image of thermal design, the image of the nuclear battery consisting a cooling system with natural convection and a power generation system with no moving equipment is proposed. Further feasibility study to confirm the feasibility of nuclear battery will be carried out in training of next fiscal year.

Journal Articles

Loss of core cooling test with one cooling line inactive in Vessel Cooling System of High-Temperature Engineering Test Reactor

Fujiwara, Yusuke; Nemoto, Takahiro; Tochio, Daisuke; Shinohara, Masanori; Ono, Masato; Takada, Shoji

Journal of Nuclear Engineering and Radiation Science, 3(4), p.041013_1 - 041013_8, 2017/10

In HTTR, the test was carried out at the reactor thermal power of 9 MW under the condition that one cooling line of VCS was stopped to simulate the partial loss of cooling function from the surface of RPV in addition to the loss of forced cooling flow in the core simulation. The test results showed that temperature change of the core internal structures and the biological shielding concrete was slow during the test. Temperature of RPV decreased several degrees during the test. The temperature decrease of biological shielding made of concrete was within 1$$^{circ}$$C. The numerical result simulating the detail configuration of the cooling tubes of VCS showed that the temperature rise of cooling tubes of VCS was about 15$$^{circ}$$C, which is sufficiently small, which did not significantly affect the temperature of biological shielding concrete. As the results, it was confirmed that the cooling ability of VCS can be kept in case that one cooling line of VCS is lost.

JAEA Reports

Development of fuel temperature calculation code "FTCC" for high temperature gas-cooled reactors

Inaba, Yoshitomo; Isaka, Kazuyoshi; Shibata, Taiju

JAEA-Data/Code 2017-002, 74 Pages, 2017/03

JAEA-Data-Code-2017-002.pdf:2.36MB

In order to ensure the thermal integrity of fuel in High Temperature Gas-cooled Reactors (HTGRs), it is necessary that the maximum fuel temperature in normal operation is to be lower than a thermal design target. In the core thermal-hydraulic design of block-type HTGRs, the maximum fuel temperature should be evaluated considering data such as core geometry and specifications, power density and neutron fluence distributions, and core coolant flow distribution. The fuel temperature calculation code used in the design stage of the High Temperature engineering Test Reactor (HTTR) presupposes to run on UNIX systems, and its operation and execution procedure are complicated and are not user-friendly. Therefore, a new fuel temperature calculation code, named FTCC, which has a user-friendly system such as a simple and easy operation and execution procedure, was developed. This report describes the calculation objects and models, the basic equations, the strong points (improvement points from the HTTR design code), the code structure, the using method of FTCC, and the result of a validation calculation with FTCC. The calculation result obtained by FTCC provides good agreement with that of the HTTR design code, and then FTCC will be used as one of the design codes for high temperature gas-cooled reactors. In addition, the effect of hot spot factors and fuel cooling forms on reducing the maximum fuel temperature is investigated with FTCC. As a result, it was found that the effect of center hole cooling for hollow fuel compacts and gapless cooling with monolithic type fuel rods on reducing the temperature is very high.

Journal Articles

Development of fuel temperature calculation code for HTGRs

Inaba, Yoshitomo; Nishihara, Tetsuo

Annals of Nuclear Energy, 101, p.383 - 389, 2017/03

 Times Cited Count:6 Percentile:65.17(Nuclear Science & Technology)

In order to ensure the thermal integrity of fuel in High Temperature Gas-cooled Reactors (HTGRs), it is necessary that the maximum fuel temperature in normal operation is to be lower than a thermal design target. In the core thermal-hydraulic design of block-type HTGRs, the maximum fuel temperature should be evaluated considering data such as thermal power, core geometry, power density and neutron fluence distributions, and core coolant flow distribution. The fuel temperature calculation code used in the design stage of the High Temperature engineering Test Reactor (HTTR) presupposes to run on UNIX systems, and its operation and execution procedure are complicated and are not user-friendly. Therefore, a new fuel temperature calculation code named FTCC which has a user-friendly system such as a simple and easy operation and execution procedure, was developed. This paper describes calculation objects and models, basic equations, improvement points from the HTTR design code in FTCC, and the result of a validation calculation with FTCC. The calculation result obtained by FTCC provides good agreement with that of the HTTR design code, and then FTCC will be used as one of the design codes for HTGRs. In addition, the effect of cooling forms on the maximum fuel temperature is investigated by using FTCC. As a result, it was found that the effect of center hole cooling for hollow fuel compacts and gapless cooling with monolithic type fuel rods on reducing the temperature is very high.

Journal Articles

New reactor cavity cooling system (RCCS) with passive safety features; A Comparative methodology between a real RCCS and a scaled-down heat-removal test facility

Takamatsu, Kuniyoshi; Matsumoto, Tatsuya*; Morita, Koji*

Annals of Nuclear Energy, 96, p.137 - 147, 2016/10

 Times Cited Count:5 Percentile:54.39(Nuclear Science & Technology)

After Fukushima Daiichi nuclear disaster by TEPCO, a cooling system to prevent core damage became more important from the perspective of defense in depth. Therefore, a new, highly efficient RCCS with passive safety features without a requirement for electricity and mechanical drive is proposed. Employing the air as the working fluid and the ambient air as the ultimate heat sink, the new RCCS design strongly reduces the possibility of losing the heat sink for decay heat removal. The RCCS can always stably and passively remove a part of the released heat at the rated operation and the decay heat after reactor shutdown. Specifically, emergency power generators are not necessary and the decay heat can be passively removed for a long time, even forever if the heat removal capacity of the RCCS is sufficient. We can also define the experimental conditions on radiation and natural convection for the scale-down heat removal test facility.

Journal Articles

Loss of core cooling test without one cooling line in Vessel Cooling System (VCS) of High Temperature engineering Test Reactor (HTTR)

Fujiwara, Yusuke; Nemoto, Takahiro; Tochio, Daisuke; Shinohara, Masanori; Ono, Masato; Hamamoto, Shimpei; Iigaki, Kazuhiko; Takada, Shoji

Proceedings of 24th International Conference on Nuclear Engineering (ICONE-24) (DVD-ROM), 7 Pages, 2016/06

In HTTR, the test was carried out at the reactor thermal power of 9 MW under the condition that one cooling line of VCS was stopped to simulate the partial loss of cooling function from the surface of RPV in addition to the loss of forced cooling flow in the core simulation. The test results showed that temperature change of the core internal structures and the biological shielding concrete was slow during the test. Temperature of RPV decreased several degrees during the test. The temperature decrease of biological shielding made of concrete was within 1$$^{circ}$$C. The numerical result simulating the detail configuration of the cooling tubes of VCS showed that the temperature rise of cooling tubes of VCS was about 15 degree C, which is sufficiently small, which did not significantly affect the temperature of biological shielding concrete. As the results, it was confirmed that the cooling ability of VCS can be kept in case that one cooling line of VCS is lost.

Journal Articles

New reactor cavity cooling system with a novel shape and passive safety features

Takamatsu, Kuniyoshi; Matsumoto, Tatsuya*; Morita, Koji*

Proceedings of 2016 International Congress on Advances in Nuclear Power Plants (ICAPP 2016) (CD-ROM), p.1250 - 1257, 2016/04

After Fukushima Daiichi nuclear disaster by TEPCO, a cooling system to prevent core damage became more important from the perspective of defense in depth. Therefore, a new, highly efficient RCCS with passive safety features without a requirement for electricity and mechanical drive is proposed. Employing the air as the working fluid and the ambient air as the ultimate heat sink, the new RCCS design strongly reduces the possibility of losing the heat sink for decay heat removal. The RCCS can always stably and passively remove a part of the released heat at the rated operation and the decay heat after reactor shutdown. Specifically, emergency power generators are not necessary and the decay heat can be passively removed for a long time, even forever if the heat removal capacity of the RCCS is sufficient. We can also define the experimental conditions on radiation and natural convection for the scale-down heat removal test facility.

Journal Articles

Visualization of internal structures of reactor core in the HTTR; Proposal of non-destructive inspection by cosmic-ray muon radiography

Takamatsu, Kuniyoshi

Hokeikyo Nyusu, (56), p.2 - 4, 2015/10

JP, 2010-166333   Licensable Patent Information Database   Patent publication (In Japanese)

In our study, we focused on a nondestructive inspection method by cosmic-ray muons which could be used to observe the internal reactor from outside the RPV and the CV. We conducted an observation test on the HTTR to evaluate the applicability of the method to the internal visualization of a reactor. We also analytically evaluated the resolution of existing muon telescopes to assess their suitability for the HTTR observation, and were able to detect the major structures of the HTTR based on the distribution of the surface densities calculated from the coincidences measured by the telescopes. Our findings suggested that existing muon telescopes could be used for muon observation of the internal reactor from outside the RPV and CV.

Journal Articles

Visualization of internal structures of reactor core in the HTTR by cosmic-ray muon radiography; Non-destructive inspection of internal structures of reactor core

Takamatsu, Kuniyoshi

Nihon Genshiryoku Gakkai-Shi ATOMO$$Sigma$$, 57(6), p.389 - 393, 2015/06

JP, 2010-166333   Licensable Patent Information Database   Patent publication (In Japanese)

In our study, we focused on a nondestructive inspection method by cosmic-ray muons which could be used to observe the internal reactor from outside the RPV and the CV. We conducted an observation test on the HTTR to evaluate the applicability of the method to the internal visualization of a reactor. We also analytically evaluated the resolution of existing muon telescopes to assess their suitability for the HTTR observation, and were able to detect the major structures of the HTTR based on the distribution of the surface densities calculated from the coincidences measured by the telescopes. Our findings suggested that existing muon telescopes could be used for muon observation of the internal reactor from outside the RPV and CV.

Journal Articles

Cosmic-ray muon radiography for reactor core observation

Takamatsu, Kuniyoshi; Takegami, Hiroaki; Ito, Chikara; Suzuki, Keiichi*; Onuma, Hiroshi*; Hino, Ryutaro; Okumura, Tadahiko*

Annals of Nuclear Energy, 78, p.166 - 175, 2015/04

 Times Cited Count:7 Percentile:60.95(Nuclear Science & Technology)

In our study, we focused on a nondestructive inspection method by which cosmic-ray muons could be used to observe the internal reactor from outside the RPV and the CV. We conducted an observation test on the HTTR to evaluate the applicability of the method to the internal visualization of a reactor. We also analytically evaluated the resolution of existing muon telescopes to assess their suitability for the HTTR observation, and were able to detect the major structures of the HTTR based on the distribution of the surface densities calculated from the coincidences measured by the telescopes. Our findings suggested that existing muon telescopes could be used for muon observation of the internal reactor from outside the RPV and CV.

Journal Articles

Assessment of calculation model for annular core on the HTTR

Nojiri, Naoki; Handa, Yuichi*; Shimakawa, Satoshi; Goto, Minoru; Kaneko, Yoshihiko*

Nihon Genshiryoku Gakkai Wabun Rombunshi, 5(3), p.241 - 250, 2006/09

It was shown from the annular core experiment of the HTTR that the discrepancy of excess reactivity between experiment and analysis reached about 3 % Dk/k at maximum. Sensitivity analysis for the annular core of the HTTR was performed to improve the discrepancy. The SRAC code system was used for the core analysis. As the results of the analysis, it was found clearly that the multiplication factor of the annular core is affected by (1) mesh interval in the core diffusion calculation, (2) mesh structure of graphite region in fuel lattice cell and (3) the Benoist's anisotropic diffusion coefficients. The significantly large discrepancy previously reported was reduced down to about 1 % Dk/k by the revised annular core model.

Journal Articles

Improvement of core dynamics analysis of control rod withdrawal test in HTGR

Takamatsu, Kuniyoshi; Nakagawa, Shigeaki

Nihon Genshiryoku Gakkai Wabun Rombunshi, 5(1), p.45 - 56, 2006/03

The HTTR (High Temperature Engineering Test Reactor), which has thermal output of 30MW, coolant inlet temperature of 395$$^{circ}$$C and coolant outlet temperature of 850$$^{circ}$$C/950$$^{circ}$$C, is a first high temperature gas-cooled reactor (HTGR) in Japan. The HTGR has a high inherent safety potential to accident condition. Safety demonstration tests using the HTTR are underway in order to demonstrate such excellent inherent safety features of the HTGR. A one-point core dynamics approximation with one fuel channel model had applied to this analysis. It was found that the analytical model for core dynamics couldn't simulate the reactor power behavior accurately. This report proposes an original method using temperature coefficients of some regions in the core. It is crucial to evaluate this method precisely to simulate a performance of HTGR during the test.

Journal Articles

Annular core experiments in HTTR's start-up core physics tests

Fujimoto, Nozomu; Yamashita, Kiyonobu*; Nojiri, Naoki; Takeuchi, Mitsuo; Fujisaki, Shingo; Nakano, Masaaki*

Nuclear Science and Engineering, 150(3), p.310 - 321, 2005/07

 Times Cited Count:5 Percentile:37.38(Nuclear Science & Technology)

Annular cores were formed in startup-core-physics tests of the High Temperature Engineering Test Reactor (HTTR) to obtain experimental data for verification of calculation codes. The first criticality, control rod positions at critical conditions, neutron flux distribution, excess reactivity etc. were measured as representative data. These data were evaluated with Monte Carlo code MVP that can consider the heterogeneity of coated fuel particles (CFP) distributed randomly in fuel compacts directly. It was made clear that the heterogeneity effect of CFP on reactivity for annular cores is smaller than that for fully-loaded cores. Measured and calculated effective multiplication factors (k) were agreed with differences less than 1%$$Delta$$k. Measured neutron flux distributions agreed with calculated results. The revising method was applied for evaluation of excess reactivity to exclude negative shadowing effect of control rods. The revised and calculated excess reactivity agreed with differences less than 1%$$Delta$$k/k.

Journal Articles

Core thermal-hydraulic design

Takada, Eiji*; Nakagawa, Shigeaki; Fujimoto, Nozomu; Tochio, Daisuke

Nuclear Engineering and Design, 233(1-3), p.37 - 43, 2004/10

 Times Cited Count:10 Percentile:58.07(Nuclear Science & Technology)

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 $$^{circ}$$C for the maximum through the burn-up cycle, which satisfies the design limit of 1495 $$^{circ}$$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 $$^{circ}$$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 $$^{circ}$$C. It is confirmed that the core thermal-hydraulic design gives conservative results.

Journal Articles

Experience of HTTR construction and operation; Unexpected incidents

Fujimoto, Nozomu; Tachibana, Yukio; Saikusa, Akio*; Shinozaki, Masayuki; Isozaki, Minoru; Iyoku, Tatsuo

Nuclear Engineering and Design, 233(1-3), p.273 - 281, 2004/10

 Times Cited Count:0 Percentile:0.01(Nuclear Science & Technology)

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.

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