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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

Temperature coefficient measurement test of HTTR; Burn-up characteristic of temperature coefficients at reactor power 30 kW and 9 MW

Ono, Masato; Goto, Minoru; Shinohara, Masanori; Nojiri, Naoki; Tochio, Daisuke; Shimazaki, Yosuke; Yanagi, Shunki

JAEA-Technology 2013-001, 35 Pages, 2013/03

JAEA-Technology-2013-001.pdf:6.04MB

The temperature coefficient measurements of the HTTR have been carried out. In the beginning of the operation, temperature coefficients at the reactor power of 30 kW and 9 MW were obtained through 1999 to 2000. The operation days of the HTTR fuel reached 375 Effective Full Power Days (EFPD), which is over a half of design operation days (660 EFPD). The temperature coefficient measurements were conducted at the same power levels of 30 kW and 9 MW to evaluate burnup effect. Also, to measure temperature coefficient in high accuracy, technique of core temperature control and technique of core temperature homogenization were established.

JAEA Reports

Safety demonstration test using the High Temperature Engineering Test Reactor (HTTR); Cold test of the loss of forced cooling

Shinohara, Masanori; Yanagi, Shunki; Tochio, Daisuke; Shimazaki, Yosuke; Nojiri, Naoki; Owada, Hiroyuki; Sato, Nao; Sagawa, Hiroshi; Umeda, Masayuki

JAEA-Technology 2011-029, 39 Pages, 2011/12

JAEA-Technology-2011-029.pdf:3.03MB

JAEA plans and performs the safety demonstration test using the HTTR to develop High Temperature Gas Reactor technologies. Cold test of the loss of forced cooling was conducted prior to the safety demonstration test, to check test procedure and plant behavior. Cold test consists of two phases, Phase1, 1 or 2 Vessel Cooling System (VCS) terminates, in the Phase2, all 3 Gas circulators and 1 VCS terminates. Cold test could confirm test process, and obtain data necessary to analysis and 2-dimensional horizontal sectional model analysis was verified to simulate actual measurement value.

JAEA Reports

Handling of HTTR second driver fuel elements in assembling and storage working

Tomimoto, Hiroshi; Kato, Yasushi; Owada, Hiroyuki; Sato, Nao; Shimazaki, Yosuke; Kozawa, Takayuki; Shinohara, Masanori; Hamamoto, Shimpei; Tochio, Daisuke; Nojiri, Naoki; et al.

JAEA-Technology 2009-025, 29 Pages, 2009/06

JAEA-Technology-2009-025.pdf:21.78MB

The first driver fuel of the HTTR (High Temperature Engineering test Reactor) was loaded in 1998 and the HTTR reached first criticality state in the same year. The HTTR has been operated using the first driver fuel for a decade. In Fuel elements assembling, 4770 of fuel rods which consist of 12 kinds of enrichment uranium are loaded into 150 fuel graphite blocks for HTTR second driver fuel elements. Measures of prevention of fuel rod miss loading, are employed in fuel design. Additionally, precaution of fuel handling on assembling are considered. Reception of fuel rods, assembling of fuel elements and storage of second driver fuels in the fresh fuel storage rack in the HTTR were started since June, 2008. Assembling, storage and pre-service inspection were divided into three parts. The second driver fuel assembling was completed in September, 2008. This report describes concerns of fuel handling on assembling and storage work for the HTTR fuel elements.

JAEA Reports

Result of long-term operation of HTTR; Rated/parallel-loaded 30-days operation

Tochio, Daisuke; Nojiri, Naoki; Hamamoto, Shimpei; Inoi, Hiroyuki; Sekita, Kenji; Kondo, Masaaki; Saikusa, Akio; Kameyama, Yasuhiko; Saito, Kenji; Fujimoto, Nozomu

JAEA-Technology 2009-005, 47 Pages, 2009/05

JAEA-Technology-2009-005.pdf:4.01MB

HTTR is now conducted in-service operation through the rise-to power operation with rated operation or high-temperature test operation from achievement of first criticality at 1998. In order to demonstrate to supply stable heat to heat utilization system for long-term, HTTR was conducted rated/parallel-loaded 30-days operation. This paper reports the characteristics of long-term operation for HTTR.

Journal Articles

Contribution to improvement of HTGR technology by using HTTR operation data

Nakagawa, Shigeaki; Tochio, Daisuke; Shinohara, Masanori; Nojiri, Naoki; Nishihara, Tetsuo; Goto, Minoru; Takamatsu, Kuniyoshi

Proceedings of 2009 International Congress on Advances in Nuclear Power Plants (ICAPP '09) (CD-ROM), p.9476_1 - 9476_6, 2009/05

Journal Articles

Present status of HTTR and its test experience

Iyoku, Tatsuo; Nojiri, Naoki; Fujimoto, Nozomu; Shinohara, Masanori; Ota, Yukimaru; Tachibana, Yukio

Proceedings of 16th Pacific Basin Nuclear Conference (PBNC-16) (CD-ROM), 6 Pages, 2008/10

The High Temperature Gas-cooled Reactor (HTGR) is expected to be one of the most promising energy sources not only for electricity generation and but also for process heat applications such as hydrogen production, desalination, etc. In Japan, since 1960s Japan Atomic Energy Agency (JAEA) has been developing HTGR technologies such fuel, high temperature metal, graphite, core physics, thermal hydraulics control and so forth. These technologies were well developed and used to design and construct the Japan's first HTGR, High Temperature Engineering Test Reactor (HTTR). It is a graphite-moderated and helium-cooled HTGR with the rated thermal power of 30 MW and the maximum outlet coolant temperature of 950$$^{circ}$$C. The HTTR achieved the reactor outlet coolant temperature of 950$$^{circ}$$C on April 19, 2004. It is the highest coolant temperature outside reactor pressure vessel in the world. So far, basic performance data of the HTTR during the power-up and long-term high temperature operation tests are accumulated. Except that, various unique tests concerning the HTGR safety are conducted to confirm inherent safety characteristics of the HTGR.

JAEA Reports

Ultrasonic test results for the reactor pressure vessel of the HTTR; Longitudinal welding line of bottom dome

Nojiri, Naoki; Owada, Hiroyuki; Kato, Yasushi

JAEA-Technology 2008-045, 38 Pages, 2008/06

JAEA-Technology-2008-045.pdf:6.95MB

This paper describes the inspection method, the measured area, etc. of the ultrasonic test of the in-service inspection (ISI) for welding lines of the reactor pressure vessel of the HTTR and the inspection results of the longitudinal welding line of the bottom dome. The pre-service inspection (PSI) results for estimation of occurrence and progression of defects to compare the ISI results is described also.

JAEA Reports

HTTR operation data base, 2; Examples of the HTTR core characteristics data base, etc

Nojiri, Naoki; Owada, Hiroyuki; Fujimoto, Nozomu

JAEA-Data/Code 2007-013, 93 Pages, 2007/06

JAEA-Data-Code-2007-013.pdf:5.34MB

For the future HTGR development and the management of the High Temperature engineering Test Reactor(HTTR), the HTTR operation data base is constructed. The data base consists of the sorted or evaluated data based on the measured values from the HTTR's operation, such as excess reactivity of the core, temperature at facilities of the core and the plant, impurities in coolant and so on. The data base also consists of some sub-databases which have objects related to the future HTGR development or the HTTR's operational management in order to manage the important operation data systematically on a long term. This paper describes examples of the HTTR common data base, the HTTR nuclear characteristics data base, the helium purity control data base and the other data base.

Journal Articles

Present status of HTTR and its operational experience

Iyoku, Tatsuo; Nojiri, Naoki; Tochio, Daisuke; Mizushima, Toshihiko; Tachibana, Yukio; Fujimoto, Nozomu

Proceedings of 15th International Conference on Nuclear Engineering (ICONE-15) (CD-ROM), 8 Pages, 2007/04

A HTGR is particularly attractive because of its capability of producing high temperature helium gas and its inherent safety characteristics. Hence, the HTTR wasconstructed at the Oarai Research Establishment of the Japan Atomic Energy Agency. The HTTR achieved the full power of 30MW and reactor outlet coolant temperature of about 850$$^{circ}$$C on December 7, 2001. After several operation cycles, the HTTR achieved the reactor outlet coolant temperature of 950$$^{circ}$$C on April 19, 2004. It is the highest coolant temperature outside reactor pressure vessel in the world. Extensive tests are planned in the HTTR and a process heat application system will be coupled to the HTTR, where hydrogen will be produced directly from the nuclear energy.

JAEA Reports

HTTR operation data base, 1; Outline and structure

Nojiri, Naoki; Tochio, Daisuke; Hamamoto, Shimpei; Umeda, Masayuki; Fujimoto, Nozomu; Iyoku, Tatsuo; Takeda, Tetsuaki

JAEA-Data/Code 2006-022, 61 Pages, 2006/10

JAEA-Data-Code-2006-022.pdf:5.68MB

For the future HTGR development and the management of the High Temperature engineering Test Reactor (HTTR), the HTTR operation data base is constructed. The data base consists of the sorted or evaluated data based on the measured values from the HTTR's operation such as excess reactivity of the core, temperature at facilities of the core and the plant, impurity in coolant and so on. The data base also consists of some sub-databases which have objects related to the future HTGR development or the HTTR's operational management in order to manage the important operation data systematically on a long term. This paper describes the outline and structure of the HTTR operation data base. Also, as an example, some part of the HTTR common data-base, the HTTR nuclear characteristics data-base and the Helium purity control data-base are described.

Journal Articles

Neutronics calculations of HTTR with several nuclear data libraries

Goto, Minoru; Nojiri, Naoki; Shimakawa, Satoshi

Journal of Nuclear Science and Technology, 43(10), p.1237 - 1244, 2006/10

 Times Cited Count:4 Percentile:32.55(Nuclear Science & Technology)

Benchmark calculations for several HTTR core conditions were performed with four cross-section sets which were generated from JENDL-3.3, JENDL-3.2, ENDF/B-6.8 and JEFF-3.0 using a continuous energy Monte Carlo code MVP. The core conditions were an annular form core at room temperature and a full core with cylindrical form at room temperature and at full power operation. Study of k$$_{eff}$$ discrepancies caused by difference of the nuclear data libraries and identification of nuclide which gives large effects on the k$$_{eff}$$ discrepancies were carried out. Comparison of the k$$_{eff}$$ between calculations and experiments was also carried out. As the results, for each HTTR core conditions, JENDL-3.3 gives the k$$_{eff}$$ agreement with the experiments within 1.5%$$Delta$$k, JENDL-3.2 gives the k$$_{eff}$$ agreement within 1.7%$$Delta$$k, and ENDF/B-6.8 and JEFF-3.0 give k$$_{eff}$$ agreement within 1.8%$$Delta$$k. There is little k$$_{eff}$$ discrepancy between ENDF/B-6.8 and JEFF-3.0. The $$k_{eff}$$ discrepancy between JENDL-3.3 and JENDL-3.2 is caused by difference of U-235 data and has temperature dependency. The k$$_{eff}$$ discrepancy between JENDL-3.3 and ENDF/B-6.8 or JEFF-3.0 is caused by difference of graphite data.

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.

JAEA Reports

Burnup characteristics of burnable poison and core characteristics of HTTR

Fujimoto, Nozomu; Nojiri, Naoki

JAEA-Technology 2005-008, 45 Pages, 2006/01

JAEA-Technology-2005-008.pdf:1.9MB

The HTTR uses burnable poison to compensate the change in reactivity with burnup. The burnable poison is a rod type and the change in effective absorption cross section become large due to its shape. Therefore, change in effective cross section of burnable poison due to burnup, temperature, fuel enrichment, etc. are evaluated and the core calculation model for burnup calculation. Using the model, burnup calculations are carried out and compared with results of experiments and other code. It become clear that the model shows reasonable results.

Journal Articles

Nuclear characteristics of High Temperature Engineering Test Reactor (HTTR)

Goto, Minoru; Nojiri, Naoki; Nakagawa, Shigeaki; Fujimoto, Nozomu

Koon Gakkai-Shi, 32(1), p.11 - 15, 2006/01

no abstracts in English

Journal Articles

Application of miniature pulsed magnets to synchrotron X-ray spectroscopy and neutron diffraction

Matsuda, Yasuhiro*; Murata, Yuto*; Inami, Toshiya; Owada, Kenji; Nojiri, Hiroyuki*; Oyama, Kenji*; Kato, Naoki*; Murakami, Yoichi*; Iga, Fumitoshi*; Takabatake, Toshiro*; et al.

Journal of Physics; Conference Series, 51, p.490 - 493, 2006/00

 Times Cited Count:8 Percentile:92.95

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.

JAEA Reports

Rise-to-power test in high temperature engineering test reactor in the high temperature test operation mode; Test progress and summary of test results up to 30MW of reactor thermal power

Takamatsu, Kuniyoshi; Nakagawa, Shigeaki; Sakaba, Nariaki; Takada, Eiji*; Tochio, Daisuke; Shimakawa, Satoshi; Nojiri, Naoki; Goto, Minoru; Shibata, Taiju; Ueta, Shohei; et al.

JAERI-Tech 2004-063, 61 Pages, 2004/10

JAERI-Tech-2004-063.pdf:3.14MB

The High Temperature engineering Test Reactor (HTTR) is a graphite moderated and gas cooled reactor with the thermal power of 30MW and the reactor outlet coolant temperature of 850$$^{circ}$$C/950$$^{circ}$$C. Rise-to-power test in the HTTR was performed from March 31th to May 1st in 2004 as phase 5 test up to 30MW in the high temperature test operation mode. It was confirmed that the thermal reactor power and the reactor outlet coolant temperature reached to 30MW and 950$$^{circ}$$C respectively on April 19th in the single operation mode using only the primary pressurized water cooler. The parallel loaded operation mode using the intermediate heat exchanger and the primary pressurized water cooler was performed from June 2nd and JAERI (Japan Atomic Energy Research Institute) obtained the certificate of the pre-operation test on June 24th from MEXT (Ministry of Education Culture Sports Science and Technology) after all the pre-operation tests were passed successfully in the high temperature test operation mode. Achievement of the reactor-outlet coolant temperature of 950$$^{circ}$$C is the first time in the world. It is possible to extend highly effective power generation with a high-temperature gas turbine and produce hydrogen from water with a high-temperature. This report describes the results of the high-temperature test operation of the HTTR.

JAEA Reports

DELIGHT-8; One dimensional fuel cell burnup analysis code for High Temperature Gas-cooled Reactor (HTGR) (Joint research)

Nojiri, Naoki; Fujimoto, Nozomu; Mori, Tomoaki; Obata, Hiroyuki*

JAERI-Data/Code 2004-012, 65 Pages, 2004/10

JAERI-Data-Code-2004-012.pdf:7.77MB

DELIGHT code is a fuel cell burnup analysis code which can produce the group constants necessary for High Temperature Gas-cooled Reactors (HTGR) core analyses. Collision probability method is used to the lattice calculation. The lattice calculation model is a cylinder type fuel or a ball type fuel of the HTGR. This code characterizes the burnup calculation considering the double heterogeneity caused by coated fuel particles of the HTGR fuel. DELIGHT code has updated its nuclear data library to the latest JENDL-3.3 data, and included new burnup chain models in order to calculate high burnup HTGR cores. The material regions of the periphery burnable poisons (BPs) were divided into details in order to improve calculation accuracy of the BP lattice calculation. This report presents the revised points of the DELIGHT-8 and can be used as user's manual.

Journal Articles

Nuclear design

Fujimoto, Nozomu; Nojiri, Naoki; Ando, Hiroei*; Yamashita, Kiyonobu*

Nuclear Engineering and Design, 233(1-3), p.23 - 36, 2004/10

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

In the nuclear design of the HTTR, 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$$^{circ}$$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.

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