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JAEA Reports

Assessment of the probability of aircraft crashing for HTTR

Ono, Masato; Hanawa, Yoshio; Sonobe, Hiroshi; Nishimura, Arashi; Sugaya, Naoto; Iigaki, Kazuhiko

JAEA-Technology 2020-010, 14 Pages, 2020/09

JAEA-Technology-2020-010.pdf:1.74MB

In response to new standard for regulating research and test reactor which is enforced December 18, 2013, it was carried out assessment of the probability of aircraft crashing for HTTR. According to assessment method provided in the Assessment Criteria of the Probability of Aircraft Crashing on Commercial Power Reactor Facilities, assessment was conducted targeting reactor building, spent fuel storage building and cooling tower. As a result, it was confirmed that the probability was 5.98$$times$$10$$^{-8}$$, which is lower than the assessment criteria 10$$^{-7}$$.

Journal Articles

Evaluation of impact with high temperature engineering test reactor using realistic model of stack and reactor building

Ono, Masato; Fujiwara, Yusuke; Matsumoto, Tetsuro*; Iigaki, Kazuhiko

Nippon Genshiryoku Gakkai Wabun Rombunshi, 19(2), p.110 - 120, 2020/06

Integrity confirmation for buildings against collisions of projectiles has been conducted to evaluate collisions between a projectile with simple shape and a wall using empirical formulas. It is a matter of fact, there is a possibility that structures with complex shape such as stack may collide with a reactor building. However, there were not so many studies of collisions between structures with complex shape and buildings in the literature. Impact evaluation was carried out using reactor building and stack with real shape and adequate physical property. It was found that ceiling of reactor building was not damaged by the collision, confirming that there was no effect inside of reactor building.

Journal Articles

Maintenance management of HTTR (Characteristics and achievements of maintenance management)

Shimazaki, Yosuke; Yamazaki, Kazunori; Iigaki, Kazuhiko

Hozengaku, 18(1), p.16 - 20, 2019/04

no abstracts in English

JAEA Reports

Error estimation in observed acceleration data toward V&V of a seismic simulation

Suzuki, Yoshio; Iigaki, Kazuhiko

JAEA-Data/Code 2018-009, 41 Pages, 2018/09

JAEA-Data-Code-2018-009.pdf:3.14MB

Toward Verification & Validation (V&V) of a seismic simulation of entire nuclear plant, an approach to estimate errors included in observed acceleration data is proposed. On the comparison between simulation results and experimental/observational results in the process of V&V, errors which might be included in experimental/observational data should be estimated. It is considered that there exist following two causes for errors in observed acceleration data; measurement accuracy of an accelerometer measurement system and disturbance included in measured data. Techniques based on the specification of an accelerometer measurement system and the time series analysis are respectively adopted to estimate those errors. To clarify the actual procedure, those techniques are applied to acceleration data observed at High Temperature engineering Test Reactor (HTTR) at the Oarai Research and Development Institute of Japan Atomic Energy Agency.

JAEA Reports

Excellent feature of Japanese HTGR technologies

Nishihara, Tetsuo; Yan, X.; Tachibana, Yukio; Shibata, Taiju; Ohashi, Hirofumi; Kubo, Shinji; Inaba, Yoshitomo; Nakagawa, Shigeaki; Goto, Minoru; Ueta, Shohei; et al.

JAEA-Technology 2018-004, 182 Pages, 2018/07

JAEA-Technology-2018-004.pdf:18.14MB

Research and development on High Temperature Gas-cooled Reactor (HTGR) in Japan started since late 1960s. Japan Atomic Energy Agency (JAEA) in cooperation with Japanese industries has researched and developed system design, fuel, graphite, metallic material, reactor engineering, high temperature components, high temperature irradiation and post irradiation test of fuel and graphite, high temperature heat application and so on. Construction of the first Japanese HTGR, High Temperature engineering Test Reactor (HTTR), started in 1990. HTTR achieved first criticality in 1998. After that, various test operations have been carried out to establish the Japanese HTGR technologies and to verify the inherent safety features of HTGR. This report presents several system design of HTGR, the world-highest-level Japanese HTGR technologies, JAEA's knowledge obtained from construction, operation and management of HTTR and heat application technologies for HTGR.

Journal Articles

Comprehensive seismic evaluation of HTTR against the 2011 off the Pacific coast of Tohoku Earthquake

Ono, Masato; Iigaki, Kazuhiko; Sawahata, Hiroaki; Shimazaki, Yosuke; Shimizu, Atsushi; Inoi, Hiroyuki; Kondo, Toshinari; Kojima, Keidai; Takada, Shoji; Sawa, Kazuhiro

Journal of Nuclear Engineering and Radiation Science, 4(2), p.020906_1 - 020906_8, 2018/04

On March 11th, 2011, the 2011 off the Pacific coast of Tohoku Earthquake of magnitude 9.0 occurred. When the great earthquake occurred, the High Temperature Engineering Test Reactor (HTTR) had been stopped under the periodic inspection and maintenance of equipment and instruments. A comprehensive integrity evaluation was carried out for the HTTR facility because the maximum seismic acceleration observed at the HTTR exceeded the maximum value of design basis earthquake. The concept of comprehensive integrity evaluation is divided into two parts. One is the "visual inspection of equipment and instruments". The other is the "seismic response analysis" for the building structure, equipment and instruments using the observed earthquake. All equipment and instruments related to operation were inspected in the basic inspection. The integrity of the facilities was confirmed by comparing the inspection results or the numerical results with their evaluation criteria. As the results of inspection of equipment and instruments associated with the seismic response analysis, it was judged that there was no problem for operation of the reactor, because there was no damage and performance deterioration. The integrity of HTTR was also supported by the several operations without reactor power in cold conditions of HTTR in 2011, 2013 and 2015. Additionally, the integrity of control rod guide blocks was also confirmed visually when three control rod guide blocks and six replaceable reflector blocks were taken out from reactor core in order to change neutron startup sources in 2015.

Journal Articles

3D crash calculation of stack and reactor building of HTTR

Ono, Masato; Fujiwara, Yusuke; Iigaki, Kazuhiko; Matsumoto, Tetsuro*; Taki, Nobuhiro*

Proceedings of European Research Reactor Conference 2018 (RRFM 2018) (Internet), 7 Pages, 2018/03

Integrity confirmation of building against the crash of flying object due to the tornado was carried out by formulas which calculate by simple shape. However, there was no study on crash calculation using complex shape such as the stack and reactor building. In this study, the crash calculation was carried out by a real shape model of High Temperature Engineering Test Reactor (HTTR) stack and reactor building using three-dimensional analysis code (VPS: Virtual Performance Solution). In the calculation, parameters of VPS were conservatively set in accordance with the formulas, which are formulated based on results of crash experiments and approved by the Nuclear Regulatory Authority. The crash calculation of stack and reactor building of HTTR was carried out using VPS. As a result, the integrity of building against the crash by stack was confirmed.

Journal Articles

Measurement of temperature response of intermediate heat exchanger in heat application system abnormal simulating test using HTTR

Ono, Masato; Fujiwara, Yusuke; Honda, Yuki; Sato, Hiroyuki; Shimazaki, Yosuke; Tochio, Daisuke; Homma, Fumitaka; Sawahata, Hiroaki; Iigaki, Kazuhiko; Takada, Shoji

Proceedings of 2017 International Congress on Advances in Nuclear Power Plants (ICAPP 2017) (CD-ROM), 5 Pages, 2017/04

Japan Atomic Energy Agency (JAEA) has carried out research and developments towards nuclear heat utilization of High Temperature Gas-cooled Reactor (HTGR) using High Temperature Engineering Test Reactor (HTTR). The nuclear heat utilization systems connected to HTGR will be designed on the basis of non-nuclear-grade standards in terms of easier entry for the chemical plant companies and the construction economics of the systems. Therefore, it is necessary that the reactor operations continue even if abnormal events occur in the systems. Heat application system abnormal simulating test with HTTR was carried out in non-nuclear heating operation to focus on the thermal effect in order to obtain data of the transient temperature behavior of the metallic components in the Intermediate Heat Exchanger (IHX). The IHX is the key components to connect the HTTR with the heat application system. In the test, the coolant helium gas temperature was heated up to 120$$^{circ}$$C by the compression heat of the gas circulators in the HTTR under the ideal condition to focus on the heat transfer. The tests were conducted by decreasing the helium gas temperature stepwise by increasing the mass flow rate to the air cooler. The temperature responses of the IHX were investigated. For the components such as the heat transfer tubes and heat transfer enhancement plates of IHX, the temperature response was slower in the lower position in comparison with the higher position. The reason is considered that thermal load fluctuation is imposed in the secondary helium gas which flows from the top to the bottom in the heat transfer tubes of the IHX. The test data are useful to verify the numerical model of the safety evaluation code.

Journal Articles

Investigation of countermeasure against local temperature rise in vessel cooling system in loss of core cooling test without nuclear heating

Ono, Masato; Shimizu, Atsushi; Kondo, Makoto; Shimazaki, Yosuke; Shinohara, Masanori; Tochio, Daisuke; Iigaki, Kazuhiko; Nakagawa, Shigeaki; Takada, Shoji; Sawa, Kazuhiro

Journal of Nuclear Engineering and Radiation Science, 2(4), p.044502_1 - 044502_4, 2016/10

In the loss of forced core cooling test using High Temperature engineering Test Reactor (HTTR), the forced cooling of reactor core is stopped without inserting control rods into the core and cooling by Vessel Cooling System (VCS) to verify safety evaluation codes to investigate the inherent safety of HTGR be secured by natural phenomena to make it possible to design a severe accident free reactor. The VCS passively removes the retained residual heat and the decay heat from the core via the reactor pressure vessel by natural convection and thermal radiation. In the test, the local temperature was supposed to exceed the limit from the viewpoint of long-term use at the uncovered water cooling tube by thermal reflectors in the VCS, although the safety of reactor is kept. Through a cold test, which was carried out by non-nuclear heat input from gas circulators with stopping water flow in the VCS, the local higher temperature position was specified although the temperature was sufficiently lower than the maximum allowable working temperature, and natural circulation of water had insufficient cooling effect on the temperature of water cooling tube below 1$$^{circ}$$C. Then, a new safe and secured procedure for the loss of forced core cooling test was established, which will be carried out soon after the restart of HTTR.

Journal Articles

Visualization in response analyses for a nuclear power plant

Nakajima, Norihiro; Nishida, Akemi; Miyamura, Hiroko; Iigaki, Kazuhiko; Sawa, Kazuhiro

Kashika Joho Gakkai-Shi (USB Flash Drive), 36(Suppl.2), 4 Pages, 2016/10

Since nuclear power plants have dimensions approximately 100m$$^{3}$$ and their structures are an assembly made up of over 10 million components, it is not convenient to experimentally analyze its behavior under strong loads of earthquakes, due to the complexity and hugeness of plants. The proposed system performs numerical simulations to evaluate the behaviors of an assembly like a nuclear facility. The paper discusses how to carry out visual analysis for assembly such as nuclear power plants. In a result discussion, a numerical experiment was carried out with a numerical model of High Temperature engineering Test Reactor of Japan Atomic Energy Agency and its result was compared with observed data. A good corresponding among them was obtained as a structural analysis of an assembly by using visualization. As a conclusion, a visual analytics methodology for assembly is discussed.

Journal Articles

Confirmation of seismic integrity of HTTR against 2011 Great East Japan Earthquake

Ono, Masato; Iigaki, Kazuhiko; Shimazaki, Yosuke; Shimizu, Atsushi; Inoi, Hiroyuki; Tochio, Daisuke; Hamamoto, Shimpei; Nishihara, Tetsuo; Takada, Shoji; Sawa, Kazuhiro; et al.

Proceedings of 24th International Conference on Nuclear Engineering (ICONE-24) (DVD-ROM), 12 Pages, 2016/06

On March 11th, 2011, the Great East Japan Earthquake of magnitude 9.0 occurred. When the great earthquake occurred, the HTTR had been stopped under the periodic inspection and maintenance of equipment and instrument. In the great earthquake, the maximum seismic acceleration observed at the HTTR exceeded the maximum value in seismic design. The visual inspection of HTTR facility was carried out for the seismic integrity conformation of HTTR. The seismic analysis was also carried out using the observed earthquake motion at HTTR site to confirm the integrity of HTTR. The concept of comprehensive integrity evaluation for the HTTR facility is divided into two parts. One is the inspection of equipment and instrument. The other is the seismic response analysis using the observed earthquake. For the basic inspections of equipment and instrument were performed for all them related to the operation of reactor. The integrity of the facilities is confirmed by comparing the inspection results or the numerical results with their evaluation criteria. As the result of inspection of equipment and instrument and seismic response analysis, it was judged that there was no problem to operate the reactor, because there was no damage and performance deterioration, which affects the reactor operation. The integrity of HTTR was also supported by the several operations without reactor power in cold conditions of HTTR in 2011, 2013 and 2015.

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 degree 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

Thermal mixing behavior in the annulus of co-axial double-walled piping in HTGR

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

Proceedings of 24th International Conference on Nuclear Engineering (ICONE-24) (DVD-ROM), 9 Pages, 2016/06

From the HTTR operational experience, it is needed to clear the thermal mixing characteristics of the helium gas at the annulus of the co-axial double-walled piping in HTGR. In this paper, thermal-hydraulic analysis on the helium gas at the annular flow path of the co-axial double pipe with T-junction was conducted. The analysis was performed under the condition of the different annular flow path height and with the different flow rate of the higher- and the lower-temperature helium gas. It is shown that the thermal mixing behavior is not so much affected by the flow rate of higher- and lower-temperature helium gas, and it is difficult to mix the helium gas with the smaller height of the annular flow path. It is confirmed that it is difficult to mix the helium gas in the annular flow path of the co-axial double-walled piping by using the hydraulic behavior, and it is necessary to arrange the mixing promotor in the annular flow path.

Journal Articles

Characteristic confirmation test by using HTTR and investigation of absorbing thermal load fluctuation

Honda, Yuki; Tochio, Daisuke; Sato, Hiroyuki; Nakagawa, Shigeaki; Ono, Masato; Fujiwara, Yusuke; Hamamoto, Shimpei; Iigaki, Kazuhiko; Takada, Shoji

Proceedings of 24th International Conference on Nuclear Engineering (ICONE-24) (DVD-ROM), 5 Pages, 2016/06

The characteristic confirmation test has been demonstrating by using the High Temperature engineering Test Reactor (HTTR). The thermal load fluctuation test, which is one of marginal performance test is planned to be carried out after restarting of the HTTR. The preliminary analysis for the thermal load fluctuation test has been investigated. In the analysis, the reactor outlet temperature can continue to be stable against the reactor inlet temperature changing by thermal fluctuation. It means that HTGR have the capability of absorbing thermal fluctuation. This paper focuses on the investigation of mechanism of absorbing thermal fluctuation. With additional analysis, it is cleared that the large negative graphite moderator reactivity enhances the capability of absorbing thermal fluctuation. In addition, in the middle of the core, graphite moderator reactivity insertion trend are inverted. This trend is unique to HTGR because of large temperature difference between core inlet and outlet.

Journal Articles

Evaluation on seismic integrity of HTTR core components

Ono, Masato; Iigaki, Kazuhiko; Shimazaki, Yosuke; Tochio, Daisuke; Shimizu, Atsushi; Inoi, Hiroyuki; Hamamoto, Shimpei; Takada, Shoji; Sawa, Kazuhiro

Proceedings of International Topical Meeting on Research Reactor Fuel Management and Meeting of the International Group on Reactor Research (RRFM/IGORR 2016) (Internet), p.363 - 371, 2016/03

HTTR is graphite moderated and helium gas-cooled reactor with prismatic fuel elements and hexagonal blocks. Here, the graphite block is brittle materials and might be damaged by collision of neighboring blocks by the large earthquake. A seismic observation system is installed in the HTTR site to confirm a behavior of a seismic event. On March 11th, 2011, off the Pacific coast of Tohoku Earthquake of magnitude 9.0 occurred. After the accident at the TEPCO Fukushima Daiichi Nuclear Power Station, the safety of nuclear reactors is the highest importance. To confirm the seismic integrity of HTTR core components, the seismic analysis was carried out using the evaluation waves based on the relationship between the observed earthquake motion at HTTR site and frequency transfer function. In parallel, confirmation tests of primary cooling system on cold state and integrity confirmation of reactor buildings and component support structures were also carried out. As a result, it was found that a stress value of the graphite blocks satisfied an allowable value, and the integrity of the HTTR core components was ensured. The integrity of HTTR core components was also supported by the operation without reactor power in cold conditions of HTTR. The obtained data was compared with the normal plant data before the earthquake. As the result, the integrity of the HTTR facilities was confirmed.

Journal Articles

Influence of differences between seismic safety evaluation methods for equipment and piping of a nuclear facility

Nishida, Akemi; Iigaki, Kazuhiko; Sawa, Kazuhiro; Li, Y.

Proceedings of 2015 ASME Pressure Vessels and Piping Conference (PVP 2015) (Internet), 7 Pages, 2015/07

The objective of this research was to investigate the influence of differences between methods for evaluating the seismic safety of the equipment and piping of a nuclear facility. For the input ground motion, one wave was chosen from among 200 waves of input ground motions of maximum acceleration of 700-1100 cm/s$$^{2}$$ created for the Oarai District of the Ibaraki Prefecture. Seismic safety evaluations were performed using the conventional method, which relies on floor response spectrum data, and using the multi-input method. The differences between the two methods were summarized. The target equipment and piping system were cooling systems in a model plant. It was found that the response predicted by the multi-input method was approximately half of the response predicted by the conventional method. The third trial evaluation method using the floor response of a three-dimensional building model as input was also reported.

Journal Articles

Seismic response simulation of High-Temperature Engineering Test Reactor building against 2011 Tohoku earthquake

Nishida, Akemi; Nakajima, Norihiro; Kawakami, Yoshiaki; Iigaki, Kazuhiko; Sawa, Kazuhiro

Proceedings of 23rd International Conference on Nuclear Engineering (ICONE-23) (DVD-ROM), 7 Pages, 2015/05

The R&D on the three dimensional vibration simulation technologies for a nuclear facility is one of missions of Center for Computational Science and e-Systems, Japan Atomic Energy Agency. Until now, three dimensional building and equipment models of HTTR (High Temperature Engineering Test Reactor) have been constructed and been performed validation of the models by comparison with seismic observed records. In this report, the results obtained by seismic observation simulation on the Tohoku earthquake occurred in the 3/11/2011 using three dimensional models of the HTTR building are shown. The simulation results show good agreement with the real observation data.

Journal Articles

Numerical modeling assistance system in finite element analysis for the structure of an assembly

Nakajima, Norihiro; Nishida, Akemi; Kawakami, Yoshiaki; Suzuki, Yoshio; Sawa, Kazuhiro; Iigaki, Kazuhiko

Proceedings of 23rd International Conference on Nuclear Engineering (ICONE-23) (DVD-ROM), 5 Pages, 2015/05

A numerical analysis controlling and managing system is implemented on K, which controls the modelling process and data treating, although the manager only controls a structural analysis by finite element method. The modeling process is described by the list of function ID and its procedures in a data base. The manager executes the process by order in the list for simulation procedures. The manager controls the intention of an analysis by changing the analytical process one to another. Experiments were carried out with static and dynamic analyses.

Journal Articles

Investigation of characteristics of natural circulation of water in vessel cooling system in loss of core cooling test without nuclear heating

Takada, Shoji; Shimizu, Atsushi; Kondo, Makoto; Shimazaki, Yosuke; Shinohara, Masanori; Seki, Tomokazu; Tochio, Daisuke; Iigaki, Kazuhiko; Nakagawa, Shigeaki; Sawa, Kazuhiro

Proceedings of 23rd International Conference on Nuclear Engineering (ICONE-23) (DVD-ROM), 5 Pages, 2015/05

In the loss of forced core cooling test using High Temperature engineering Test Reactor (HTTR), the forced cooling of reactor core is stopped without inserting control rods into the core and cooling by Vessel Cooling System (VCS) to demonstrate the inherent safety of HTGR be secured by natural phenomena to make it possible to design a severe accident free reactor. In the test, the local temperature was supposed to exceed the limit from the viewpoint of long-term use at the uncovered water cooling tube by thermal reflectors in the VCS, although the safety of reactor is kept. The local higher temperature position was specified although the temperature was sufficiently lower than the maximum allowable working temperature, and natural circulation of water had insufficient cooling effect on the temperature of water cooling tube below 1$$^{circ}$$C. Then, a new safe and secured procedure for the loss of forced core cooling test was established, which will be carried out soon after the restart of HTTR.

Journal Articles

A Safety evaluation of HTTR core graphite structures against the Great East Japan Earthquake

Iigaki, Kazuhiko; Ono, Masato; Shimazaki, Yosuke; Tochio, Daisuke; Shimizu, Atsushi; Inoi, Hiroyuki; Takada, Shoji; Sawa, Kazuhiro

Mechanical Engineering Journal (Internet), 1(4), p.TEP0021_1 - TEP0021_13, 2014/08

On March 11th, 2011, the Great East Japan Earthquake which is one of the largest earthquakes in japan occurred and the maximum acceleration in observed seismic wave in the HTTR exceeded the design value in a part of input seismic motions. Therefore, a visual inspection, a seismic analysis and a performance confirmation test of facilities were carried out in order to confirm the integrity of facility after the earthquake. The seismic analysis was carried out for the reactor core structures by using the response magnification factor method. As the results of the evaluation, the generated stress in the graphite blocks in the reactor core at the earthquake were well below the allowable values of safety criteria, and thus the structural integrity of the reactor core was confirmed. The integrity of reactor core was also supported by the visual inspections of facilities and the operation without reactor power in cold conditions of HTTR.

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