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Journal Articles

Development of fabrication technology for oxidation-resistant fuel elements for high-temperature gas-cooled reactors

Aihara, Jun; Honda, Masaki*; Ueta, Shohei; Ogawa, Hiroaki; Ohira, Koichi*; Tachibana, Yukio

Nippon Genshiryoku Gakkai Wabun Rombunshi, 18(1), p.29 - 36, 2019/03

Japan Atomic Energy Agency carried out development of fabrication technology of oxidation resistant fuel element for improvement of safety of high temperature gas-cooled reactors in serious oxidation accident, based on precursor research in former JAEA. Dummy coated fuel particles (alumina particles) were over-coated with mixed powder of Si, C and small amount of resin to form over-coated particles, and over-coated particles were molded and hot-pressed to sinter dummy oxidation resistant fuel elements with SiC/C mixed matrix. We fabricated dummy oxidation resistant fuel elements with matrix whose Si/C mole ratio (about 0.551) is three times as large as that in precursor research. Si peak was not detected by X-ray diffraction of matrix. Better oxidation resistant was confirmed with oxidation test in 20% O$$_{2}$$ at 1673 K than that of ordinal fuel compact with ordinal graphite/carbon matrix. All dummy coated fuel particles were held in specimen after 10 h oxidation.

Journal Articles

Development of security and safety fuel for Pu-burner HTGR; Test and characterization for ZrC coating

Ueta, Shohei; Aihara, Jun; Goto, Minoru; Tachibana, Yukio; Okamoto, Koji*

Mechanical Engineering Journal (Internet), 5(5), p.18-00084_1 - 18-00084_9, 2018/10

To develop the security and safety fuel (3S-TRISO fuel) for Pu-burner high temperature gas-cooled reactor (HTGR), R&D on zirconium carbide (ZrC) directly coated on yttria stabilized zirconia (YSZ) has been started in the Japanese fiscal year 2015. As results of the direct coating test of ZrC on the dummy YSZ particle, ZrC layers with 18 - 21 microns of thicknesses have been obtained with 0.1 kg of particle loading weight. No deterioration of YSZ exposed by source gases of ZrC bromide process was observed by Scanning Transmission Electron Microscope (STEM).

Journal Articles

Study on Pu-burner high temperature gas-cooled reactor in Japan; Design study of fuel and reactor core

Goto, Minoru; Aihara, Jun; Inaba, Yoshitomo; Ueta, Shohei; Fukaya, Yuji; Okamoto, Koji*

Proceedings of 9th International Topical Meeting on High Temperature Reactor Technology (HTR 2018) (USB Flash Drive), 6 Pages, 2018/10

JAEA has conducted design studies of a Pu-burner HTGR. The Pu-burner HTGR incinerates Pu by fission, and hence a high burn-up is required for the efficient incineration. In the fuel design, a thin ZrC layer, which acts as an oxygen getter and suppresses the internal pressure, was coated on the fuel kernel to prevent the CFP failure at the high burn-up. A stress analysis of the SiC layer, which acts as a pressure vessel for the CFP, was performed for with consideration of the depression effect due to the ZrC layer. As a result, the CFP failure fraction at high burn-up of 500 GWd/t satisfied the target value. In the reactor core design, an axial fuel shuffling was employed to attain the high burn-up, and the nuclear burn-up calculations with the whole core model and the fuel temperature calculations were performed. As a result, the nuclear characteristics, which are the shutdown margin and the temperature coefficient of reactivity, and the fuel temperature satisfied their target values.

Journal Articles

Study on Pu-burner high temperature gas-cooled reactor in Japan; Test and characterization for ZrC coating

Ueta, Shohei; Aihara, Jun; Mizuta, Naoki; Goto, Minoru; Fukaya, Yuji; Tachibana, Yukio; Okamoto, Koji*

Proceedings of 9th International Topical Meeting on High Temperature Reactor Technology (HTR 2018) (USB Flash Drive), 7 Pages, 2018/10

The security and safety fuel (3S-TRISO fuel) employs the coated fuel particle with a fuel kernel made of plutonium dioxide (PuO$$_{2}$$) and yttria stabilized zirconia (YSZ) as an inert matrix. Especially, a zirconium carbide (ZrC) coating is one of key technologies of the 3S-TRISO, which performs as an oxygen getter to reduce the fuel failure due to internal pressure during the irradiation. R&Ds on ZrC coating directly on the dummy CeO$$_{2}$$-YSZ kernel have been carried in the Japanese fiscal year 2017. As results of ZrC coating tests by the bromide chemical vapor deposition process, stoichiometric ZrC coatings with 3 - 18 microns of thicknesses were obtained with 0.1 kg of particle loading weight.

Journal Articles

Investigation of irradiated properties of extended burnup TRISO fuel

Shaimerdenov, A.*; Gizatulin, S.*; Kenzhin, Y.*; Dyussambayev, D.*; Ueta, Shohei; Aihara, Jun; Shibata, Taiju

Proceedings of 9th International Topical Meeting on High Temperature Reactor Technology (HTR 2018) (USB Flash Drive), 6 Pages, 2018/10

The Institute of Nuclear Physics of the Republic of Kazakhstan (INP) conducts an irradiation test and post-irradiation examinations (PIEs) of the high-temperature gas-cooled reactor (HTGR) fuel and materials to develop the extend burnup fuel up to 100 GWd/t-U collaboratively with the Japan Atomic Energy Agency (JAEA) under projects in a frame of the International Science and Technology Centre (ISTC). Cylindrical fuel compact specimens consisting of newly-designed TRISO (tri-structural isotropic) coated fuel particles and a matrix made of graphite material were manufactured in Japan. An irradiation test of the fuel specimens using a helium-gas swept capsule designed and constructed in the INP has been performed up to 100 GWd/t-U in the WWR-K research reactor by April 2015. In the next stage, PIEs with the irradiated fuel specimens have been started in February 2017 as a new ISTC project. Several PIE technologies by non-destructive and destructive techniques with irradiated fuel compacts were developed by the INP. This report presents the developed technologies and interim results of the PIE for high burning TRISO fuel.

JAEA Reports

Comparison between HTFP code and minory changed FORNAX-A code

Aihara, Jun; Ueta, Shohei; Goto, Minoru; Inaba, Yoshitomo; Shibata, Taiju; Ohashi, Hirofumi

JAEA-Technology 2018-002, 70 Pages, 2018/06

JAEA-Technology-2018-002.pdf:1.46MB

HTFP code is code for calculation of additional release amount of fission product (FP) from fuel rod in high temperature gas-cooled reactor (HTGR) after stop of fission. Minory changed Fornax-A code also can calculate that. Therefore, release behavior of Cs calculated with HTFP code was compared with that calculated with minory modified FORNAX-A code in this report. Release constants of Cs evaluated with minory modified FORNAX-A code are rather different from default values for HTFP code.

Journal Articles

Development of security and safety fuel for Pu-burner HTGR, 2; Design study of fuel and reactor core

Goto, Minoru; Ueta, Shohei; Aihara, Jun; Inaba, Yoshitomo; Fukaya, Yuji; Tachibana, Yukio; Okamoto, Koji*

Proceedings of 25th International Conference on Nuclear Engineering (ICONE-25) (CD-ROM), 6 Pages, 2017/07

A PuO$$_{2}$$-YSZ fuel kernel with a ZrC coating, which enhances safety, security and safeguard, namely: 3S-TRISO fuel, was proposed to introduce to the plutonium-burner HTGR. In this study, the efficiency of the ZrC coating as the free-oxygen getter was examined based on a thermochemical calculation. A preliminary study on the feasibility of the 3S-TRISO fuel was conducted focusing on the internal pressure. Additionally, a nuclear feasibility of the reactor core was studied. As a result, all the amount of the free-oxygen is captured by a thin ZrC coating under 1600$$^{circ}$$C and coating ZrC on the fuel kernel should be very effective method to suppress the internal pressure. The internal pressure of the 3S-TRISO fuel at 500 GWd/t is lower than that of UO$$_{2}$$ kernel TRISO fuel whose feasibility had been already confirmed and the 3S-TRISO fuel should be feasible. The fuel shuffling allows to achieve 500 GWd/t. The temperature coefficient of reactivity is negative during the operation period and thus the nuclear feasibility of the reactor core should be achievable.

Journal Articles

Development of security and safety fuel for Pu-burner HTGR, 5; Test and characterization for ZrC coating

Ueta, Shohei; Aihara, Jun; Goto, Minoru; Tachibana, Yukio; Okamoto, Koji*

Proceedings of 25th International Conference on Nuclear Engineering (ICONE-25) (CD-ROM), 4 Pages, 2017/07

To develop the security and safety fuel (3S-TRISO fuel) for Pu-burner high temperature gas-cooled reactor (HTGR), R&D on zirconium carbide (ZrC) directly coated on yttria stabilized zirconia (YSZ) has been started in the Japanese fiscal year 2015. As results of the direct coating test of ZrC on the dummy YSZ particle, ZrC layers with 18 - 21 microns of thicknesses have been obtained with 0.1 kg of particle loading weight. No deterioration of YSZ exposed by source gases of ZrC bromide process was observed by Scanning Transmission Electron Microscope (STEM).

JAEA Reports

Confirmation of feasibility of fabrication technology and characterization of high-packing fraction fuel compact for HTGR

Mizuta, Naoki; Ueta, Shohei; Aihara, Jun; Shibata, Taiju

JAEA-Technology 2017-004, 22 Pages, 2017/03

JAEA-Technology-2017-004.pdf:2.71MB

Confirmation of feasibility of fabrication technology and characterization of the high-packing fraction fuel compact of High Temperature Gas Reactor (HTGR) fuel were carried out. Fuel compacts were fabricated with CFP packing fraction targeted at 33 percent by the same manufacturing condition of HTTR fuel compact. SiC-defective fraction, compressive strength and internal CFP distribution of the compact, important parameters to guarantee its integrity, were evaluated. The high-packing fuel compacts showed as same level of SiC-defective fraction as that of HTTR first loading fuel, 8$$times$$10$$^{-5}$$, and larger compressive strength than the HTTR fuel criteria, 4,900N. The feasibility of fabrication technology and the performance for the high-packing fraction fuel compact was confirmed.

Journal Articles

Nuclear thermal design of high temperature gas-cooled reactor with SiC/C mixed matrix fuel compacts

Aihara, Jun; Goto, Minoru; Inaba, Yoshitomo; Ueta, Shohei; Sumita, Junya; Tachibana, Yukio

Proceedings of 8th International Topical Meeting on High Temperature Reactor Technology (HTR 2016) (CD-ROM), p.814 - 822, 2016/11

Japan Atomic Energy Agency (JAEA) has started R&D for apply SiC/C mixed matrix to fuel element of high temperature gas-cooled reactors (HTGRs) to improve oxidation resistance of fuel. Nuclear thermal design of HTGR with SiC/C mixed matrix fuel compacts was carried out as a part of above R&Ds. Nuclear thermal design was carried out based on a small sized HTGR for developing countries, HTR50S. Maximum enrichment of uranium is set to be 10 wt%, because coated fuel particles with 10 wt% uranium have been fabricated in Japan. Numbers of kinds of enrichment and burnable poisons (BPs) were set to be same as those of original HTR50S (3 and 2, respectively). We succeeded in nuclear thermal design of a small sized HTGR which performance was equivalent to original HTR50S, with SiC/C mixed matrix fuel compacts. Based on nuclear thermal design, intactness of coated fuel particles was evaluated to be kept on internal pressure during normal operation.

Journal Articles

Irradiation test and post irradiation examination of the high burnup HTGR fuel

Ueta, Shohei; Aihara, Jun; Shaimerdenov, A.*; Dyussambayev, D.*; Gizatulin, S.*; Chakrov, P.*; Sakaba, Nariaki

Proceedings of 8th International Topical Meeting on High Temperature Reactor Technology (HTR 2016) (CD-ROM), p.246 - 252, 2016/11

In order to examine irradiation performance of the new Tri-structural Isotropic (TRISO) fuel for the High Temperature Gas-cooled Reactor (HTGR) at the burnup around 100 GWd/t, a capsule irradiation test was conducted by WWR-K research reactor in the Institute of Nuclear Physics (INP) of Kazakhstan. The irradiated TRISO fuel was designed by Japan Atomic Energy Agency (JAEA) and fabricated in basis of the HTTR fuel technology in Japan. The fractional release of fission gas from the fuel during the irradiation shows good agreement with the predicted one released from as-fabricated failed TRISO fuel. It was suggested that unexpected additional fuel failure would not occur during the irradiation up to 100 GWd/t. In addition, the post-irradiation examination (PIE) with the irradiated fuel is planned to qualify TRISO fuel integrity and upgrade HTGR fuel design for further burnup extension.

JAEA Reports

Application of FORNAX-A

Aihara, Jun; Ueta, Shohei; Nishihara, Tetsuo

JAEA-Technology 2015-040, 32 Pages, 2016/02

JAEA-Technology-2015-040.pdf:0.83MB

Original FORNAX-A is a calculation code for amount of fission product (FP) released from fuel rods of pin-in-type high temperature gas-cooled reactors (HTGRs). This report is for explanation what calculations become possible with minor changed FORNAX-A.

Journal Articles

Conceptual study of a plutonium burner high temperature gas-cooled reactor with high nuclear proliferation resistance

Goto, Minoru; Demachi, Kazuyuki*; Ueta, Shohei; Nakano, Masaaki*; Honda, Masaki*; Tachibana, Yukio; Inaba, Yoshitomo; Aihara, Jun; Fukaya, Yuji; Tsuji, Nobumasa*; et al.

Proceedings of 21st International Conference & Exhibition; Nuclear Fuel Cycle for a Low-Carbon Future (GLOBAL 2015) (USB Flash Drive), p.507 - 513, 2015/09

A concept of a plutonium burner HTGR named as Clean Burn, which has a high nuclear proliferation resistance, had been proposed by Japan Atomic Energy Agency. In addition to the high nuclear proliferation resistance, in order to enhance the safety, we propose to introduce PuO$$_{2}$$-YSZ TRISO fuel with ZrC coating to the Clean Burn. In this study, we conduct fabrication tests aiming to establish the basic technologies for fabrication of PuO$$_{2}$$-YSZ TRISO fuel with ZrC coating. Additionally, we conduct a quantitative evaluation of the security for the safety, a design of the fuel and the reactor core, and a safety evaluation for the Clean Burn to confirm the feasibility. This study is conducted by The University of Tokyo, Japan Atomic Energy Agency, Fuji Electric Co., Ltd., and Nuclear Fuel Industries, Ltd. It was started in FY2014 and will be completed in FY2017, and the first year of the implementation was on schedule.

JAEA Reports

HTFP for calculation of amount of additionally released fission products from fuel rods of pin-in-block-type high temperature gas-cooled reactors during accident

Nomoto, Yasunobu; Aihara, Jun; Nakagawa, Shigeaki; Isaka, Kazuyoshi; Ohashi, Hirofumi

JAEA-Data/Code 2015-008, 39 Pages, 2015/06

JAEA-Data-Code-2015-008.pdf:10.32MB

HTFP is a calculation code for amount of additionally released fission product (FP) from fuel rods of pin-in-type according to transient of core temperature at the accident of high temperature gas-cooled reactors (HTGRs). This code analyzes FP release inventory from core according to the transient of core temperature at the accident as an input data and considering FP release rate from a fuel compact and a graphite sleeve and radioactive decay of FP. This report describes the outline of HTFP code and its input data. The computed solutions using the HTFP code were compared to those of HTCORE code, which was used for the design of the High Temperature Engineering Test Reactor (HTTR) to validate the analysis models of the HTFP code. The comparison of HTFP code results with HTCORE code results showed the good agreement.

Journal Articles

Development plan of high burnup fuel for high temperature gas-cooled reactors in future

Aihara, Jun; Ueta, Shohei; Honda, Masaki*; Blynskiy, P.*; Gizatulin, S.*; Sakaba, Nariaki; Tachibana, Yukio

Journal of Nuclear Science and Technology, 51(11-12), p.1355 - 1363, 2014/11

 Times Cited Count:6 Percentile:36.56(Nuclear Science & Technology)

Plan and status of research and development (R&D) were described on coated fuel particle (CFP) and fuel compacts for core of small sized high temperature gas-cooled reactor (HTGR) HTR50S at 2nd step of phase I (second core of HTR50S). Specifications of existing CFPs for high burnup (HTR50S2-type-CFPs) were adopted as specifications of CFPs, to reduce the R&D. HTR50S2-type-CFPs were fabricated based on technology developed in High Temperature Engineering Test Reactor (HTTR) project. First irradiation test of HTR50S2-type-CFPs is now being carried out. In addition, R&D for fuel compact with high packing fraction is needed, because volume fraction of fuel kernel to whole of HTR50S2-type-CFP is rather smaller than that of the HTTR-type-CFP. In addition, we describe outline of R&D plans for core of HTR50S in phase II and naturally safe HTGR.

Journal Articles

Fuel performance under continuous high-temperature operation of the HTTR

Ueta, Shohei; Aihara, Jun; Sakaba, Nariaki; Honda, Masaki*; Furihata, Noboru*; Sawa, Kazuhiro

Journal of Nuclear Science and Technology, 51(11-12), p.1345 - 1354, 2014/11

 Times Cited Count:3 Percentile:61.51(Nuclear Science & Technology)

Although the HTTR fuel was the first mass-produced HTGR fuel in Japan, it has been fabricated with the highest quality in the world on the basis of design principle and safety criteria to minimize both as-fabricated failure and on-operating-additional failure of the coated fuel particle. A precise technology for evaluating irradiation performance of the HTTR fuel has been developed by measuring fission gases released from fuel and constructing the fission gas release model. Through the HTTR operations including the continuous high temperature operation with 950 $$^{circ}$$C and 50 days, the measured fractional release of fission gas from fuel resulted less than 1.2 $$times$$ 10$$^{-8}$$, and a superior irradiation performance of Japanese HTGR fuel has been demonstrated. The HTTR fuel technologies could make a prospect for realization of the practical HTGR as well as the VHTR as a Generation IV nuclear power plant before the rest of the world.

JAEA Reports

Preliminary evaluation of integrity of coated fuel particles under normal operation in core of small-sized HTGR system HTR50S at 1st. step of Phase I

Aihara, Jun; Goto, Minoru; Inaba, Yoshitomo; Isaka, Kazuyoshi; Ohashi, Hirofumi; Tachibana, Yukio

JAEA-Technology 2014-009, 29 Pages, 2014/05

JAEA-Technology-2014-009.pdf:3.51MB

Japan Atomic Energy Agency (JAEA) is carrying out conceptual design of a 50 MWt small-sized high temperature gas cooled reactor (HTGR), HTR50S. In this report, integrity of coated fuel particles (CFPs) is evaluated for core of HTR50S of 1st. step of phase I (first core of HTR50S) under normal operation. CFPs are considered to be failed by fuel kernel migration by temperature gradient in CFPs or corrosion of SiC layer by fission product Pd (Pd corrosion) or increase in internal pressure under normal operation. In this report, integrity of CFPs is to be maintained for each phenomenon.

Journal Articles

R&D plan for development of oxidation-resistant graphite and investigation of oxidation behavior of SiC coated fuel particle to enhance safety of HTGR

Ueta, Shohei; Sumita, Junya; Shibata, Taiju; Aihara, Jun; Fujita, Ichiro*; Ohashi, Jun*; Nagaishi, Yoshihide*; Muto, Takenori*; Sawa, Kazuhiro; Sakaba, Nariaki

Nuclear Engineering and Design, 271, p.309 - 313, 2014/05

 Times Cited Count:4 Percentile:51.53(Nuclear Science & Technology)

A new concept of the high temperature gas-cooled reactor (HTGR) is proposed as a challenge to assure no event sequences to the harmful release of radioactive materials even when the design extension conditions (DECs) occur by deterministic approach based on the inherent safety features of the HTGR. The air/water ingress accident, one of the DECs for the HTGR, is prevented by additional measures (e.g. facility for suppression to air ingress). With regard to the core design, it is important to prevent recriticality accidents by keeping the geometry of the fuel rod which consists of the graphite sleeve, fuel compact and SiC-TRISO (TRIstructural-ISOtropic) coated fuel particle, and by improving the oxidation resistance of the graphite when air/water ingress accidents occur. Therefore, it is planned to develop the oxidation-resistant graphite, which is coated with gradient SiC layer. It is also planned that the experimental identification of the condition to form the stable oxide layer (SiO$$_{2}$$) for SiC layer on the oxidation-resistant graphite and on the SiC-TRISO fuel. This paper describes the R&D plan for un-irradiation and irradiation test under simulating air/water ingress accident condition to develop oxidation-resistant graphite and to investigate the oxidation behavior of SiC coated fuel particle.

JAEA Reports

FORNAX-A for calculation of fission product release amount from fuel rods of pin-in-block-type high temperature gas-cooled reactors

Aihara, Jun; Ueta, Shohei; Nakagawa, Shigeaki; Sawa, Kazuhiro; Ohashi, Hirofumi; Tachibana, Yukio

JAEA-Data/Code 2013-025, 64 Pages, 2014/03

JAEA-Data-Code-2013-025.pdf:2.54MB

FORNAX-A is a calculation code for amount of fission product (FP) released from fuel rods of pin-in-type high temperature gas-cooled reactors (HTGRs). This report is for explanation of outline and basic formulae of FORNAX-A code. FORNAX-A is based on Fick's laws of diffusion and can calculate FP release amount from fuel rod under normal operation and accidents without failure (including oxidation) of graphite sleeves and fuel compacts and without melting of fuel kernel, for example, stopping fission and increase in temperature.

JAEA Reports

Conceptual design of small-sized HTGR system, 5; Safety design and preliminary safety analysis

Ohashi, Hirofumi; Sato, Hiroyuki; Tazawa, Yujiro; Aihara, Jun; Nomoto, Yasunobu; Imai, Yoshiyuki; Goto, Minoru; Isaka, Kazuyoshi; Tachibana, Yukio; Kunitomi, Kazuhiko

JAEA-Technology 2013-017, 71 Pages, 2014/02

JAEA-Technology-2013-017.pdf:3.64MB

Japan Atomic Energy Agency (JAEA) has started a conceptual design of a 50 MWt small-sized high temperature gas cooled reactor (HTGR) for steam supply and electricity generation (HTR50S). Though the safety design of HTR50S was determined based on that of the High Temperature Engineering Test Reactor (HTTR) for the early deployment of HTR50S, the shutdown cooling system, which is the forced cooling heat removal system, was categorized as non-safety class to optimize the protection to provide the highest level of safety that can reasonably be achieved, and the vessel cooling system, which is categorized as the safety class system, was designed as a passive safety features. The preliminary safety analysis of HTR50S for the rupture of co-axial hot gas duct in primary cooling system and the tube rupture of steam generator was conducted to confirm the adequacy of the safety design. It was confirmed that the analysis results satisfied the acceptance criteria.

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