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

https://doi.org/10.1016/C2018-0-05383-9

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 950C, 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.

Data on migration analysis for TRU waste disposal; Results of plutonium solubility experiments in porewater of cement hydrates

Suguro, Toshiyasu; Nishikawa, Yoshiaki*; Watahiki, Takashi*; Kagawa, Akio

JAEA-Technology 2013-023, 22 Pages, 2013/10

JAEA-Technology-2013-023.pdf:2.41MB

For safety assessment of TRU waste disposal, solubility of plutonium was investigated under hardened cement paste porewater condition. Polycarboxylic acid compound, which have the possibility to be used for the TRU waste disposal, was selected as the cement admixture for the experiment. Initial concentration of Pu was 10 M in the experiment. The porewater of hardened cement paste was obtained by squeezing out the kneading of ordinary portland cement and deionized water with the cement admixture. The porewater of hardened cement paste without cement admixture is also used for the experiment. The maximum experimental period was 154 days. The experiment was carried out at room temperature (298 5 K) under argon atmosphere, in which oxygen concentration was lower than 1 ppm. Pu concentration in the porewater of hardened cement paste with or without the cement admixture were in the order of 10 mol/dm after 154 days. This value is comparable to the solubility of Pu(IV) under high pH condition, suggesting that the solubility of Pu was not affected by the cement admixture in hardened cement paste.

Validation of IHX temperature calculation code for future HTGR

Nakagawa, Shigeaki; Saikusa, Akio; Tochio, Daisuke; Takeda, Tetsuaki*

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

The intermediate heat exchanger (IHX) is one of key components in the very high temperature reactor (VHTR) system. The IHX is a helium-helium heat exchanger and the secondary hot helium gas heated up to about 900 C in the IHX is provided to the hydrogen production facility such as IS system which produced the hydrogen by the thermo-chemical water-splitting iodine-sulfur process. The calculation to obtain a precise temperature distribution inside the IHX is required to the reliable design in the VHTR system with the design lifetime of 60 years. The 30 days operation in the HTTR with the reactor outlet coolant temperature of 850 C has been performed and the temperature data for the IHX was obtained. The temperature calculation was performed to simulate the temperature distribution inside the IHX during the rated operation of the HTTR. The calculation result shows a good agreement with the experimental data and this calculation code was validated. It was confirmed that the IHX temperature calculation code was able to simulate precisely the temperature distribution inside the heat exchanger.

Hydration effect on low-frequency protein dynamics observed in simulated neutron scattering spectra

Jochi, Yasumasa*; Nakagawa, Hiroshi; Kataoka, Mikio; Kitao, Akio*

Biophysical Journal, 94(11), p.4435 - 4443, 2008/06

https://doi.org/10.1529/biophysj.107.118042

Hydration effects on protein dynamics were investigated by comparing the frequency dependence of the calculated neutron scattering spectra between full and minimal hydration states at temperatures between 100 and 300 K. The protein boson peak is observed in the frequency range 1-4 meV at 100 K in both states. The peak frequency in the minimal hydration state shifts to lower than that in the full hydration state. Protein motions with frequency higher than 4 meV were shown to undergo almost harmonic motion in both states at all temperatures simulated, whereas those with frequency lower than 1 meV dominate the total fluctuations above 220 K and contribute to the origin of the glass-like transition. At 300 K, the boson peak becomes buried in the quasi-elastic contributions in the full hydration state, but is still observed in the minimal hydration state. The boson peak is observed when protein dynamics are trapped within a local minimum of its energy surface. Protein motions, which contribute to the boson peak, are distributed throughout the whole protein. Fine structure of the dynamics structure factor is expected to be detected by the experiment if a high resolution instrument is developed in the near future.

Data on plutonium sorption onto rock; Results of the experiment for data on plutonium sorption onto tuff under conditions of reducing and of presence of nitrate

Suguro, Toshiyasu; Nishikawa, Yoshiaki*; Komuro, Takashi*; Kagawa, Akio; Kashiwazaki, Hiroshi; Yamada, Kazuo

JAEA-Technology 2007-058, 20 Pages, 2007/11

JAEA-Technology-2007-058.pdf:3.26MB

For safety assessment of TRU waste disposal, data on sorption data of plutonium on Tuff have been obtained by a static batch-type experiment. Because the repository condition will be reducing and be affected by considerable amount of nitrate in waste, the authors carried out the experiments using Tuff under the reducing (NaSO as added as reductant) and anoxic condition (O1 ppm) and solution of 0 to 0.5 M NaNO. The experimental results suggest that distribution coefficient (Kd) ranges 0.2 to 0.7 m kg in case of L/S=0.1 m kg. Similarly the Kd ranges, 1 to 7 m kg at L/S=1 m kg. However, almost samples of the solution after experiments were plutonium solubility less than detection limit(10mol/dm) of alpha spectrometer. The reason, it is guessed plutonium coprecipitation with calcium hydroxide, because experiments using saturated calcium hydroxide in the liquid.

Dynamical heterogeneity of protein dynamics studied by elastic incoherent neutron scattering and molecular simulations

Nakagawa, Hiroshi; Tokuhisa, Atsushi*; Kamikubo, Hironari*; Jochi, Yasumasa*; Kitao, Akio*; Kataoka, Mikio*

Materials Science & Engineering A, 442(1-2), p.356 - 360, 2006/12

https://doi.org/10.1016/j.msea.2005.12.082

The dynamical heterogeneity of a globular soluble protein was studied by elastic incoherent neutron scattering and molecular simulations. The q-dependence of the elastic incoherent neutron scattering shows a non-Gaussianity, a deviation from Gaussian approximation. We determined that the dynamical heterogeneity explains the non-Gaussianity, although the anharmonicity is also plausible origin. Molecular dynamics simulations confirmed that the non-Gaussianity is mainly due to the dynamical heterogeneity at a lower energy resolution, =1meV. On the other hand, the contribution from the anharmonicities to the non-Gaussianity became substantial at a higher resolution, =10eV. Regardless, the dynamical heterogeneity is the dominant factor for the non-Gaussianity.

Evaluation of heat exchange performance for secondary pressurized water cooler in HTTR

Tochio, Daisuke; Watanabe, Shuji; Saikusa, Akio; Oyama, Sunao; Nemoto, Takahiro; Hamamoto, Shimpei; Shinohara, Masanori; Isozaki, Minoru; Nakagawa, Shigeaki

JAEA-Technology 2006-005, 83 Pages, 2006/02

JAEA-Technology-2006-005.pdf:6.09MB

In High Temperature Engineering Test Reactor (HTTR), the rated thermal power of 30MW, the generated heat at reactor core is finally dissipated at the air-cooler by way of the heat exchangers of the primary cooling system, such as the intermediate heat exchanger (IHX) and the secondary pressurized water cooler (SPWC). The heat exchangers in the main cooling system are required the heat exchange performance to remove the reactor-generated-heat of 30MW under the condition of reactor coolant outlet temperature of 850 C/ 950 C. Therefore, the heat exchanges are required to satisfy the design criteria of heat exchange performance. In this report, heat exchange performance of the SPWC in the main cooling system was evaluated with the rise-to-power-up test and the in-service operation data. Moreover, evaluated value is compared with designed one, it is confirmed that the SPWC has required heat exchange performance.

Data on test results of vessel cooling system of High Temperature Engineering Test Reactor

Saikusa, Akio*; Nakagawa, Shigeaki; Fujimoto, Nozomu; Tachibana, Yukio; Iyoku, Tatsuo

JAERI-Data/Code 2002-027, 34 Pages, 2003/02

JAERI-Data-Code-2002-027.pdf:1.22MB

High Temperature Engineering Test Reactor (HTTR) is the first graphite-moderated helium gas cooled reactor in Japan. The rise-to-power test of the HTTR started on September 28,1999 and thermal power of the HTTR reached its full power of 30 MW on December 7, 2001. Vessel Cooling System (VCS) of the HTTR is a first Reactor Cavity Cooling System applied for High Temperature Gas Cooled Reactors. The VCS cools the core indirectly through the reactor pressure vessel to keep core integrity during the loss of core flow accidents such as depressurization accident. Minimum heat removal of the VCS to satisfy its safety requirement is 0.3MW at 30 MW power operation. Through the performance test of the VCS in the rise-to-power test of the HTTR, it is confirmed that the VCS heat removal at 30 MW power operation is higher than 0.3MW. This paper shows outline of the VCS and test results on the VCS performance.

Rise-to-power test in High Temperature Engineering Test Reactor; Test progress and summary of test results up to 30MW of reactor thermal power

Nakagawa, Shigeaki; Fujimoto, Nozomu; Shimakawa, Satoshi; Nojiri, Naoki; Takeda, Takeshi; Saikusa, Akio; Ueta, Shohei; Kojima, Takao; Takada, Eiji*; Saito, Kenji; et al.

JAERI-Tech 2002-069, 87 Pages, 2002/08

JAERI-Tech-2002-069.pdf:10.12MB

Rise-to-power test in the HTTR has been performed from April 23rd to June 6th in 2000 as phase 1 test up to 10MW, from January 29th to March 1st in 2001 as phase 2 test up to 20MW in the rated operation mode and from April 14th to June 8th in 2001 as phase 3 test up to 20MW in the high temperature test operation mode. Phase 4 test to achieve the thermal reactor power of 30MW started from October 23rd in 2001. On December 7th it was confirmed that the thermal reactor power reached to 30MW and the reactor outlet coolant temperature reached to 850C. JAERI obtained the certificate of pre-operation test from MEXT because all the pre-operation tests by MEXT were passed successfully. From the test results of rise-up-power test up to 30MW, the performance of reactor and cooling system were confirmed, and it was confirmed that an operation of reactor facility could be performed safely. Some problems to be solved were found through tests. By means of solving them, the reactor operation with the reactor outlet coolant temperature of 950C will be achievable.

Cause and countermeasure for heat up of HTTR core support plate at power rise tests

Fujimoto, Nozomu; Takada, Eiji*; Nakagawa, Shigeaki; Tachibana, Yukio; Kawasaki, Kozo; Saikusa, Akio; Kojima, Takao; Iyoku, Tatsuo

JAERI-Tech 2001-090, 69 Pages, 2002/01

JAERI-Tech-2001-090.pdf:7.88MB

HTTR has carried out many kinds of tests as power rise tests in which reactor power rises step by step after attained the first criticality. In the tests, temperature of a core support plate showed higher results than expected value at each power level, the temperature was expected to be higher than the maximum working temperature at 100% power level. Therefore, tests under the high temperature test operation mode, in which the core flow rate was different, were carried out to predict the temperature at 100% power precisely, and investigate the cause of the temperature rise. From the investigation, it was clear that the cause was gap flow in a core support structure. Furthermore, it was estimated that the temperature of the core support plate rose locally due to change in gap width between the core support plate and a seal plate due to change in core pressure drop. The maximum working temperature of the core support plate was revised. The integrity of core support plate under the revised maximum working temperature condition was confirmed by stress analyses.