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Journal Articles

Measurements of the neutron capture cross section of $$^{243}$$Am around 23.5 keV

Kodama, Yu*; Katabuchi, Tatsuya*; Rovira Leveroni, G.; Kimura, Atsushi; Nakamura, Shoji; Endo, Shunsuke; Iwamoto, Nobuyuki; Iwamoto, Osamu; Hori, Junichi*; Shibahara, Yuji*; et al.

Journal of Nuclear Science and Technology, 58(11), p.1159 - 1164, 2021/11

Journal Articles

Neutron beam filter system for fast neutron cross-section measurement at the ANNRI beamline of MLF/J-PARC

Rovira Leveroni, G.; Kimura, Atsushi; Nakamura, Shoji; Endo, Shunsuke; Iwamoto, Osamu; Iwamoto, Nobuyuki; Katabuchi, Tatsuya*; Terada, Kazushi*; Kodama, Yu*; Nakano, Hideto*; et al.

Nuclear Instruments and Methods in Physics Research A, 1003, p.165318_1 - 165318_10, 2021/07

 Times Cited Count:0 Percentile:0.03(Instruments & Instrumentation)

Journal Articles

Spin-orbit-induced Ising ferromagnetism at a van der Waals interface

Matsuoka, Hideki*; Barnes, S. E.*; Ieda, Junichi; Maekawa, Sadamichi; Bahramy, M. S.*; Saika, B. K.*; Takeda, Yukiharu; Wadachi, Hiroki*; Wang, Y.*; Yoshida, Satoshi*; et al.

Nano Letters, 21(4), p.1807 - 1814, 2021/02

 Times Cited Count:1 Percentile:72.02(Chemistry, Multidisciplinary)

Journal Articles

The Development of method for extinguishing / treating metallic sodium using calcium chloride

Abe, Yuta; Nagai, Keiichi; Maie, Mitsuyoshi*; Nakano, Natsuko*; Kawashima, Yuichi*; Takesue, Naohisa*; Saito, Junichi

Dai-23-Kai Doryoku, Enerugi Gijutsu Shimpojiumu Koen Rombunshu (USB Flash Drive), 5 Pages, 2018/06

no abstracts in English

JAEA Reports

The Second periodic safety review report of Tokai Reprocessing Plant

Shirai, Nobutoshi; Miura, Yasushi; Tachibana, Ikuya; Omori, Satoru; Wake, Junichi; Fukuda, Kazuhito; Nakano, Takafumi; Nagasato, Yoshihiko

JAEA-Technology 2016-007, 951 Pages, 2016/07

JAEA-Technology-2016-007-01.pdf:11.93MB
JAEA-Technology-2016-007-02.pdf:4.7MB

The periodic safety review of TRP is to confirm the safety activities and get effective additional measures the facility safety and its reliability. We implemented 4 items; for (1) evaluation of safety activity implementation, we confirmed we are adequately expanding its safety activities by the necessary documents and schemes. For (2) evaluation of status of safety activities reflecting the latest technical knowledges, we confirmed we reflect latest knowledges for improvement of safety and reliability. For (3) technical evaluation about aging degradation, we can keep the safety of the facilities important to safety and the sea discharge line, under assumption of the present maintenance, because of "focuses for aging degradation". For (4) planning measures about a 10-years-plan that the operator shall implement to keep the facility condition, by the technical evaluation, we found no additional safety plans into maintenance strategies.

Journal Articles

Estimation of corrosion amounts for carbon steel in $$gamma$$-ray irradiated neutral water condition

Yamamoto, Masahiro; Komatsu, Atsushi; Sato, Tomonori; Nakano, Junichi; Ueno, Fumiyoshi

Proceedings of 17th Asian Pacific Corrosion Control Conference (APCCC-17) (USB Flash Drive), 8 Pages, 2016/01

In Fukushima-Daiichi Nuclear Power Station, decommissioning procedures are continuing more than 30 years until fuel debris removal. It is important to keep soundness of primary container vessel (PCV), made of carbon steel, during these procedures. Corrosion of carbon steel is well-known to be controlled by cathodic reaction, in usual, oxygen reduction reaction. Corrosion of carbon steel could be mitigated by nitrogen injection procedure. However, a lot of radioactive materials exist in cooling water, an effect of radiolysis product on corrosion is an important problem. Clarifying an irradiation effect for corrosion of carbon steel, corrosion test was conducted in $$^{60}$$Co $$gamma$$-ray irradiated condition. Electrochemical measurements were conducted to determine cathodic current density of samples. Corrosion rates of carbon steel decrease with time in both $$gamma$$-ray irradiated and non-irradiated conditions. Measured values of cathodic current density gradually decreased with time and then stayed at constant value.

Journal Articles

Localized corrosion behavior of stainless steel in the diluted artificial sea-water contacted with zeolite under $$gamma$$-ray irradiation

Kato, Chiaki; Sato, Tomonori; Nakano, Junichi; Ueno, Fumiyoshi; Yamagishi, Isao; Yamamoto, Masahiro

Nihon Genshiryoku Gakkai Wabun Rombunshi, 14(3), p.181 - 188, 2015/09

In relation to the consideration for long-term storage of spent Cs adsorption vessels containing zeolites in the Fukushima Daiichi Nuclear Power Station, corrosion of the vessel material in the spent Cs adsorption vessel is one of important issues. We performed electrochemical tests of stainless steel (SUS 316L) in the zeolites containing artificial seawater under $$gamma$$-ray irradiation. The spontaneous potential ($$E_{rm SP}$$) and critical pitting potential ($$V_{rm c}$$), of SUS 316L were measured to understand the corrosion resistance of the stainless steel in this study. The rest potential of the stainless steel increased with increasing time after $$gamma$$-ray irradiation. The $$E_{rm SP}$$, defined as the steady rest potential, increased with increasing dose rate, while increasing $$E_{rm SP}$$ was suppressed by contact with the zeolites. Concentration of H$$_{2}$$O$$_{2}$$ in bulk water increased with increasing dose rate. The concentration increasing was suppressed by contact with the zeolites due to decomposition of H$$_{2}$$O$$_{2}$$. There was good relationship between $$E_{rm SP}$$ and the concentration of H$$_{2}$$O$$_{2}$$. The $$V_{rm c}$$ of SUS 316L contacted with the zeolites decreased with increasing Cl$$^{-}$$ ion concentration and is slightly smaller than the $$V_{rm c}$$ in the bulk water. The contact with the zeolites causes the suppressant of increasing $$E_{rm SP}$$ under the irradiation. The contact with the zeolite can reduce probability in the localized corrosion for SUS 316L.

Journal Articles

Estimation method for corrosion rate of carbon steel in water with $$gamma$$-ray irradiated condition

Yamamoto, Masahiro; Sato, Tomonori; Komatsu, Atsushi; Nakano, Junichi; Ueno, Fumiyoshi

Proceedings of European Corrosion Congress 2015 (EUROCORR 2015) (USB Flash Drive), 7 Pages, 2015/09

In Fukushima-Daiichi Nuclear Power Station, decommissioning procedures are continuing and it will take more than 30 years. As some structures are made of carbon steel, degradation by corrosion is large problem for structural reliability. To clarify an irradiation effect for corrosion of carbon steel, corrosion test was con-ducted in $$^{60}$$Co $$gamma$$-ray irradiated condition. Corrosion test results showed that corrosion rates of $$gamma$$-ray irradiated condition increased with $$gamma$$-ray dose rates. The oxidant concentrations were also increased with $$gamma$$-ray dose rate. From these results, a new estimation method for corrosion rate of carbon steel in water with $$gamma$$-ray irradiated condition using radiolysis calculation is introduced and discussed.

Journal Articles

Effect of $$gamma$$-ray irradiation on corrosion of low alloy steel in neutral water

Yamamoto, Masahiro; Nakano, Junichi; Komatsu, Atsushi; Sato, Tomonori; Tsukada, Takashi

Proceedings of 19th International Corrosion Congress (19th ICC) (CD-ROM), 6 Pages, 2014/11

Corrosion protection of RPV and PCV is an important issue for the long term maintenance until the end of the decommissioning procedures. One of the uncertain factors for the issue is an effect of radioactivity on corrosion of LAS and CS. Corrosion tests using LAS and CS were conducted in $$gamma$$-rays irradiated condition. Oxygen and hydrogen peroxide concentrations in the water were measured after the tests. Corrosion test results indicated that the amounts of corrosion increased by $$gamma$$-rays irradiation both air and nitrogen atmosphere. And also corrosion amounts increased with $$gamma$$-ray dose rates. Electrochemical analyses indicated that cathodic reaction of Hydrogen peroxide was controlled by diffusion process. The measured diffusion constant of H$$_{2}$$O$$_{2}$$ was about 0.75 times to that of oxygen. From these results, it is estimated that corrosion of LAS and CS in $$gamma$$-ray irradiated condition was evaluated by the cathodic reduction reaction of oxidant.

Journal Articles

Corrosion of the stainless steel in the zeolite containing diluted artificial seawater under $$gamma$$-ray irradiation

Kato, Chiaki; Sato, Tomonori; Nakano, Junichi; Ueno, Fumiyoshi; Yamagishi, Isao

Proceedings of 2014 Nuclear Plant Chemistry Conference (NPC 2014) (USB Flash Drive), 9 Pages, 2014/10

As a part of consideration for long-term storage of spent zeolite adsorption vessels in the Fukushima Daiichi Nuclear Power Station, corrosion of vessel material in the spent zeolite adsorption vessel is one of important issue. We performed electrochemical tests of stainless steel (type 316L) in the zeolite containing artificial seawater under $$gamma$$-ray irradiation. Steady spontaneous potential (Esp) and pitting potential (VC), of type 316L was measurement. $$^{60}$$Co $$gamma$$-rays source was used under irradiation. Dose rate of $$gamma$$-ray irradiation was controlled for 5 kGy/h and 400 Gy/h. In anode polarization curves, there was no clear difference under irradiation and non-irradiation. The corrosion potential of type 316L increased with increasing time after $$gamma$$-ray irradiation. The Esp was shifted to nobler by $$gamma$$-rays irradiation, while increasing Esp was suppressed by contacted with zeolite.

Journal Articles

Effects of hydrazine addition and N$$_{2}$$ atmosphere on the corrosion of reactor vessel steels in diluted seawater under $$gamma$$-rays irradiation

Nakano, Junichi; Kaji, Yoshiyuki; Yamamoto, Masahiro; Tsukada, Takashi

Journal of Nuclear Science and Technology, 51(7-8), p.977 - 986, 2014/07

 Times Cited Count:3 Percentile:28.59(Nuclear Science & Technology)

Seawater was injected into the reactor cores in the Fukushima Daiichi Nuclear Power Station. Corrosion of reactor vessel steels is considered to progress. To evaluate durability of the reactor vessel steels, corrosion tests were conducted in diluted seawater at 50 $$^{circ}$$C under $$gamma$$-rays irradiation. 10 mg/L and 100 mg/L N$$_{2}$$H$$_{4}$$ were added to diluted seawater. Without addition of N$$_{2}$$H$$_{4}$$, weight loss in the vessel steels irradiated with the 0.2 kGy/h dose rate was comparable with those without irradiation and weight loss in the vessel steels irradiated with the 4.4 kGy/h dose rate was higher than those without irradiation. Under irradiation, weight loss in the vessel steels in diluted seawater containing N$$_{2}$$H$$_{4}$$ was comparable with that in diluted seawater without N$$_{2}$$H$$_{4}$$. When gas phase in the flask was replaced with N$$_{2}$$, weight loss in the vessel steels, and O$$_{2}$$ and H$$_{2}$$O$$_{2}$$ concentrations in the diluted seawater decreased.

Journal Articles

Characterization and storage of radioactive zeolite waste

Yamagishi, Isao; Nagaishi, Ryuji; Kato, Chiaki; Morita, Keisuke; Terada, Atsuhiko; Kamiji, Yu; Hino, Ryutaro; Sato, Hiroyuki; Nishihara, Kenji; Tsubata, Yasuhiro; et al.

Journal of Nuclear Science and Technology, 51(7-8), p.1044 - 1053, 2014/07

 Times Cited Count:9 Percentile:65.28(Nuclear Science & Technology)

For safe storage of zeolite wastes generated by treatment of radioactive saline water at the Fukushima Daiichi Nuclear Power Station, properties of the Herschelite adsorbent were studied and its adsorption vessel was evaluated for hydrogen production and corrosion. Hydrogen production depends on its water level and dissolved species because hydrogen is oxidized by radicals in water. It is possible to evaluate hydrogen production rate in Herschelite submerged in seawater or pure water by taking into account of the depth effect of the water. The reference vessel of decay heat 504 W with or without residual pure water was evaluated for the hydrogen concentration by thermal hydraulic analysis using obtained fundamental properties. Maximum hydrogen concentration was below the lower explosive limit (4 %). The steady-state corrosion potential of a stainless steel 316L increased with absorbed dose rate but its increase was repressed by the presence of Herschelite. At 750 Gy/h and $$<$$60$$^{circ}$$C which were values evaluated at the bottom of the vessel of 504 W, the localized corrosion of SUS316L contacted with Herschelite would not immediately occur under 20,000 ppm of Cl$$^{-}$$ concentration.

Journal Articles

Corrosion of carbon steel and low-alloy steel in diluted seawater containing hydrazine under $$gamma$$-rays irradiation

Nakano, Junichi; Yamamoto, Masahiro; Tsukada, Takashi

Nihon Genshiryoku Gakkai Wabun Rombunshi, 13(1), p.1 - 6, 2014/03

The seawater was injected into reactor cores of Units 1, 2 and 3 in the Fukushima Daiichi Nuclear Power Plant. To investigate effects of $$gamma$$ rays irradiation on corrosion of carbon steel and low alloy steel, corrosion tests were performed in the diluted seawater at 50 $$^{circ}$$C under $$gamma$$ rays irradiation with dose rates of 4.4 kGy/h and 0.2 kGy/h. Hydrazine (N$$_{2}$$H$$_{4}$$) was added to the diluted seawater. In the diluted seawater without N$$_{2}$$H$$_{4}$$, the weight losses of the steels irradiated with 0.2 kGy/h were similar to those of the unirradiated steels, and the weight losses of the steels irradiated with 4.4 kGy/h increased to approximate 1.7 times of those of the unirradiated steels. The weight losses of the steels irradiated in the diluted seawater containing N$$_{2}$$H$$_{4}$$ were similar to those in the diluted seawater without N$$_{2}$$H$$_{4}$$. When N$$_{2}$$ was introduced to gas phase in the flasks during $$gamma$$ rays irradiation, the weight losses of the steels decreased.

Journal Articles

Effects of temperature on stress corrosion cracking behavior of stainless steel and outer oxide distribution in cracks due to exposure to high-temperature water containing hydrogen peroxide

Nakano, Junichi; Sato, Tomonori; Kato, Chiaki; Yamamoto, Masahiro; Tsukada, Takashi; Kaji, Yoshiyuki

Journal of Nuclear Materials, 444(1-3), p.454 - 461, 2014/01

 Times Cited Count:1 Percentile:10.7(Materials Science, Multidisciplinary)

Cracking growth tests were conducted in high-temperature water containing hydrogen peroxide (H$$_{2}$$O$$_{2}$$) at 561 to 423 K to evaluate the effects of H$$_{2}$$O$$_{2}$$ on stress corrosion cracking (SCC) of stainless steel (SS) at temperature lower than the boiling water reactor (BWR) operating temperature. Small compact tension (CT) specimens were prepared from thermally sensitized type 304 SS. Despite the observation of only a small portion intergranular SCC (IGSCC) near the side groove of the CT specimen at 561 K in high-temperature water containing 100 ppb of H$$_{2}$$O$$_{2}$$, the IGSCC area expanded to the central region of the CT specimens at 423 and 453 K. Effects of H$$_{2}$$O$$_{2}$$ on SCC appeared intensely at temperature lower than the BWR operating temperature. To estimate the environment in the cracks, outer oxide distribution on the fracture surface and fatigue pre-crack were examined by laser Raman spectroscopy and thermal equilibrium calculation was performed.

Journal Articles

Morphology of stress corrosion cracking due to exposure to high-temperature water containing hydrogen peroxide in stainless steel specimens with different crevice lengths

Nakano, Junichi; Sato, Tomonori; Kato, Chiaki; Yamamoto, Masahiro; Tsukada, Takashi; Kaji, Yoshiyuki

Journal of Nuclear Materials, 441(1-3), p.348 - 356, 2013/10

 Times Cited Count:2 Percentile:21.27(Materials Science, Multidisciplinary)

Crack growth tests were performed in high-temperature water containing hydrogen peroxide (H$$_{2}$$O$$_{2}$$) to evaluate the relationships between the crevice structure and H$$_{2}$$O$$_{2}$$ on stress corrosion cracking (SCC) growth morphology of stainless steel (SS). Small compact tension (CT) specimens with different fatigue pre-crack lengths were prepared. 20$$sim$$300 ppb H$$_{2}$$O$$_{2}$$ was injected into the high-temperature water at 561 K. Intergranular SCC (IGSCC) was observed only near the side grooves of the CT specimens. Owing to pre-crack shortening, the IGSCC area expanded to the central region of the CT specimens. The effects of H$$_{2}$$O$$_{2}$$ on SS appeared intensely near the surfaces exposed to high levels of H$$_{2}$$O$$_{2}$$. The calculations for the percentage of H$$_{2}$$O$$_{2}$$ remaining showed that the effects of H$$_{2}$$O$$_{2}$$ flowed from both sides of the crack were more obvious than those flowed from the crack mouth.

Journal Articles

Carbon transport and fuel retention in JT-60U with higher temperature operation based on postmortem analysis

Yoshida, Masafumi; Tanabe, Tetsuo*; Adachi, Ayumu*; Hayashi, Takao; Nakano, Tomohide; Fukumoto, Masakatsu; Yagyu, Junichi; Miyo, Yasuhiko; Masaki, Kei; Itami, Kiyoshi

Journal of Nuclear Materials, 438, p.S1261 - S1265, 2013/07

 Times Cited Count:6 Percentile:49.64(Materials Science, Multidisciplinary)

Fuel retention rates and carbon re-deposition rates in the plasma shadowed areas in JT-60U were measured. Distributions of the fuel retention as well as the carbon re-deposition in the whole in-vessel of a large tokamak were clarified for the first time in the world. The fuel retention in the plasma shadowed areas was about two times larger than that in the carbon re-deposited layers on the plasma facing surface, although the amount of the carbon re-deposited on the plasma shadowed areas were about a half of that on the plasma facing surface, because of relatively lower temperature in the shadow areas causing higher hydrogen saturation concentration in the carbon re-deposited layers. The total fuel retention rate in JT-60U, including previously measured for all plasma facing areas, was evaluated to be 1.3$$times$$10$$^{20}$$ H+Ds$$^{-1}$$, which was lower than that in other devices, due to probably to higher temperature operation in JT-60U.

Journal Articles

Hydrogen isotopes retention in gaps at the JT-60U first wall tiles

Yoshida, Masafumi; Tanabe, Tetsuo*; Hayashi, Takao; Nakano, Tomohide; Fukumoto, Masakatsu; Yagyu, Junichi; Miyo, Yasuhiko; Masaki, Kei; Itami, Kiyoshi

Fusion Science and Technology, 63(1T), p.367 - 370, 2013/05

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

In this study, the retentions of hydrogen isotopes (H and D) in the gaps in JT-60U are clarified. Carbon tiles used in 1992-2004 were poloidally and toroidally taken out from outboard first wall in JT-60U to measure the retentions. The H and D retentions in the samples were measured by thermal desorption spectrometry (TDS). The H+D retention in the top side was higher than that of the bottom side, which might be due to thicker re-deposited carbon layers on the surface of the top side. The retentions in the surface of the side surfaces were slightly lower than that in the plasma facing surface where the retention was saturated to be 3-4e22 atoms/m$$^{2}$$. The retention rate was evaluated to be 3e17 H+D atoms/m$$^{2}$$/s from the measured retentions in two different discharge times by assuming the retention to increase linearly with the discharge time.

Journal Articles

Influence of microstructure on IASCC growth behavior of neutron irradiated type 304 austenitic stainless steels in simulated BWR condition

Kaji, Yoshiyuki; Miwa, Yukio*; Shibata, Akira; Nakano, Junichi; Tsukada, Takashi; Takakura, Kenichi*; Nakata, Kiyotomo*

International Journal of Nuclear Energy Science and Engineering, 2(3), p.65 - 71, 2012/09

Crack growth rate (CGR) tests have been conducted with neutron irradiated compact tension (CT) specimens. The specimens were irradiated in the core region of the Japan Materials Testing Reactor (JMTR) in simulated BWR water environments at 288 $$^{circ}$$C from 0.37 to 5.55$$times$$10$$^{25}$$ n/m$$^{2}$$ (E$$>$$ 1 MeV) (0.62 to 9.2 dpa). The CGRs of base metals in high electrochemical corrosion potential (ECP) condition with 10 $$<$$ stress intensity factor, K $$<$$ 30 MPam$$^{1/2}$$, increased with increasing neutron fluence until 2 dpa and the CGRs were almost the same from 2 to 10 dpa. We investigated the influence of microstructure on CGR by microstructure observation and local strain measurement around the precipitate. This paper will discuss the relationship between CGR and microstructure, radiation hardening, radiation induced segregation.

Journal Articles

Effects of temperature on SCC propagation in high temperature water injected with hydrogen peroxide

Nakano, Junichi; Sato, Tomonori; Kato, Chiaki; Kaji, Yoshiyuki; Yamamoto, Masahiro; Tsukada, Takashi

Proceedings of 2012 Nuclear Plant Chemistry Conference (NPC 2012) (CD-ROM), 9 Pages, 2012/09

It has been reported that the corrosion behavior of stainless steels in high temperature water with hydrogen peroxide (H$$_{2}$$O$$_{2}$$) was different from those with O$$_{2}$$. To evaluate the effect of H$$_{2}$$O$$_{2}$$ on stress corrosion cracking (SCC), SCC growth tests were conducted in high temperature water injected with H$$_{2}$$O$$_{2}$$. In 100 ppb H$$_{2}$$O$$_{2}$$ at 561 K, an intergranular SCC (IGSCC) was observed only small portion of area near the side grooves of the CT specimen. In 100 ppb H$$_{2}$$O$$_{2}$$ at 453 K, however, IGSCC extended to the central region of the CT specimen. Effects of H$$_{2}$$O$$_{2}$$ on SCC growth behavior appeared stronger at lower temperature due to a reduction of the thermal decomposition of H$$_{2}$$O$$_{2}$$. Moreover, outer oxide layer of oxide film formed on the crack of the CT specimen was examined to estimate environmental situations in the cracks and a thermal equilibrium calculation was performed.

JAEA Reports

Evaluating techniques and phenomena of Stress Corrosion Cracking (SCC) in Light Water Reactors (LWRs); SCC evaluating techniques for predicting core internal and pipe aging of LWRs, technical data collection (Contract research)

Yamamoto, Masahiro; Kato, Chiaki; Sato, Tomonori; Nakano, Junichi; Ugachi, Hirokazu; Tsukada, Takashi; Kaji, Yoshiyuki; Tsujikawa, Shigeo*; Hattori, Shigeo*; Yoshii, Tsuguyasu*; et al.

JAEA-Review 2012-007, 404 Pages, 2012/03

JAEA-Review-2012-007.pdf:36.72MB

There are many LWRs which have been operated for more than 20 years in Japan and it is expected that technique corresponding to aging plants are necessary established for safety operation in LWRs. A lot of troubles related to SCC are reported and many investigations are concerned with SCC mechanism and technical evaluation. In this paper, those research data were collected as possible widely and reviewed systematically. Current circumstances concerned with SCC in LWRs were reviewed specifically as follows: SCC incidents, SCC evaluation methods for crack initiation and propagation, the investigations concerned with SCC mechanism and monitoring technique for corrosive environment. Influences with reactor types (BWR, PWR), materials (stainless steels, Ni alloys) and SCC evaluating methods (laboratories and actual plants) were summarized as graphs and tables easy to understand in common/difference points concerned with SCC. From these arranged results, future themes were considered and remarked SCC phenomenon was summarized in actual plants. As for SCC evaluations under the accelerate conditions in the laboratory test, it was suggested that a computational prediction and modeling including statistical technique and microscopic analysis in crack initiation were important. Furthermore it was suggested that monitoring techniques predicting SCC initiation and grasping plant circumstance in operation and feasibility in actual plants were important.

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