著者

爾見 豊

出版者

一般社団法人 日本原子力学会

雑誌

日本原子力学会和文論文誌 (ISSN:13472879)

巻号頁・発行日

pp.J19.029, (Released:2021-05-12)

参考文献数

15

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Before the Fukushima-Daiichi Accident, Japanese nuclear power plants (NPPs) were utilized with a capacity factor of around 70%, 20% lower than the US capacity factor of around 90%, which is a consequence of Japanese NPPs being operated with a shorter fuel cycle and longer outage period. One reason for this situation is that Japanese decision making is strongly focused on equipment reliability. In a typical pressurized-water reactor, however, core damage frequency (CDF) during refueling outage is higher than that during operation, that is, a short fuel cycle could possibly increase the total CDF of NPPs. In this paper, a decision-making rule using an index representing CDF par power generation is firstly proposed. Secondly, using this rule, a decision process is simulated to optimize the fuel cycle and refueling outage period while showing the effects on CDF reduction in each plant. Thirdly, by applying this decision process to all Japanese NPPs, the total CDF reduction in Japan is indicated. This simulation shows that the change of decision-making rule will bring about an 18% CDF reduction or 16% increase in power generation in total in Japan. At the same time, each NPP gains strong incentive to improve its own safety because this new rule permits a higher capacity factor operation only for the NPPs that are safer than the average.

著者

中林 亮 杉山 大輔

出版者

一般社団法人 日本原子力学会

雑誌

日本原子力学会和文論文誌 (ISSN:13472879)

巻号頁・発行日

vol.19, no.2, pp.47-64, 2020

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<p>Expert elicitation has traditionally been accepted in some countries as a way to quantify the uncertainty of radionuclide migration parameters in the safety assessment of radioactive waste disposal. However, expert elicitation has not yet been explicitly performed in the field of radioactive waste disposal in Japan. To discuss the applicability of expert elicitation in Japan, here we broadly review the histories and methodologies of expert elicitation in some papers and review in more detail case studies on the utilization of expert elicitation in the safety assessment of radioactive waste disposal in the US, UK, and Sweden. From the literature review, we suggest that it is valuable to adopt expert elicitation to quantify the uncertainty of parameters in Japan. In particular, the documentation of each elicitation step is critical to ensuring the traceability and transparency of expert elicitation. The documentation enables the regulator to evaluate whether the expert judgment including the elicitation process is adequate. Furthermore, we recommend providing not only an aggregated expert judgment for safety assessment but also the distribution of individual expert judgements. Individual expert judgments will be used for related analyses (<i>e.g.</i>, sensitivity or uncertainty analyses), leading to increased confidence in the safety assessment.</p>

著者

福村 卓也 福谷 耕司 有岡 孝司

出版者

Atomic Energy Society of Japan

雑誌

日本原子力学会和文論文誌 (ISSN:13472879)

巻号頁・発行日

vol.9, no.3, pp.318-329, 2010 (Released:2012-02-08)

参考文献数

12

被引用文献数

2 2

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Some pressurized water reactor (PWR) plants have switched secondary system feed water treatment to ethanol-amine (ETA) injection from all-volatile treatment (AVT) to reduce iron transfer in the steam generator (SG). However, the effect of ETA injection on FAC rate has not been studied systematically. To assess the effect of ETA injection on FAC rate, the water chemistries in secondary systems were calculated by considering the thermal decomposition of hydrazine in SG and the vapor/liquid partition of ammonia, ETA, and hydrazine in SG and in a moisture-separator-and-reheater (MSR). Then, we measured the FAC rate experimentally by rotation tests to examine the effect of ETA injection. The high pH condition of ETA injection reduced the FAC rate more than the low pH condition of AVT. No chemical effect on the FAC rate was observed between ETA injection and AVT at 180°C. We also evaluated the FAC rate using magnetite solubility with and without ETA injection. The evaluation showed that ETA injection reduces the FAC rate of the secondary system.

著者

坂本 浩幸 赤木 洋介 山田 一夫 舘 幸男 福田 大祐 石松 宏一 松田 樹也 齋藤 希 上村 実也 浪平 隆男 重石 光弘

出版者

一般社団法人 日本原子力学会

雑誌

日本原子力学会和文論文誌 (ISSN:13472879)

巻号頁・発行日

vol.17, no.2, pp.57-66, 2018 (Released:2018-05-15)

参考文献数

22

被引用文献数

1

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Concrete debris contaminated with radioactive cesium and other nuclides has been generated by the accident in the Fukushima Dai-ichi nuclear power plant. Moreover, there is concern that a large amount of radioactive concrete waste will be generated by the decommissioning of nuclear power plants in the future. Although conventional techniques are effective in decontaminating concrete with flat surfaces such as floors and walls, it is not clear what techniques to apply for decontaminating radioactive concrete debris. In this study, focusing on a pulsed power discharge technique, fundamental experimental work was carried out and the applicability of the technique to decontaminating radioactive concrete debris associated with the accident in the Fukushima Dai-ichi nuclear power plant was evaluated. The decontamination of concrete by applying the aggregate recycling technique using the pulsed power discharge technique was evaluated by measuring the radioactivity concentration of the divided aggregate and sludge from the contaminated concrete using a Ge-semiconductor detector. It indicated a reduction of the radioactivity concentration in the recovered aggregate and an increase in the radioactivity concentration in the sludge. These findings suggest that the division of the contaminated concrete debris into aggregate and sludge could result in the decontamination and reuse of the aggregate, which would reduce the amount of contaminated concrete debris.

著者

高松 操 川原 啓孝 伊藤 裕道 宇敷 洋 鈴木 信弘 佐々木 純 大田 克 奥田 英二 小林 哲彦 長井 秋則 坂尾 龍太 村田 長太郎 田中 淳也 松坂 康智 立野 高寛 原 正秀 岡﨑 弘祥

出版者

一般社団法人 日本原子力学会

雑誌

日本原子力学会和文論文誌 (ISSN:13472879)

巻号頁・発行日

vol.15, no.1, pp.32-42, 2016 (Released:2016-02-15)

参考文献数

10

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In the experimental fast reactor “Joyo”, it was confirmed that the top of the irradiation test subassembly of the material testing rig named “MARICO-2” was broken and bent onto the in-vessel storage rack as an obstacle, damaging the upper core structure (UCS). In this paper, we describe the in-vessel repair techniques for UCS replacement, which are developed in Joyo. The UCS replacement was conducted in the following four stages: (1) jack-up of the existing damaged UCS, (2) retrieval of the existing damaged UCS, (3) installation of the O-ring, and (4) insertion of the new UCS. Since the UCS replacement was not anticipated in the original design, the work conditions at Joyo were carefully investigated, and the obtained results were applied to the design of special handling equipment. The UCS replacement was successfully completed in 2014. In-vessel repair techniques for sodium-cooled fast reactors (SFRs) are important in confirming the safety and integrity of SFRs. However, the techniques demonstrated in the actual reactor environment with high temperature, high radiation dose, and remaining sodium are insufficient to secure the reliability of these techniques. The experience and knowledge accumulated in the UCS replacement provide valuable insights into further improvements of in-vessel repair techniques for SFRs.

著者

菅原 慎悦 稲村 智昌 木村 浩 班目 春樹

出版者

Atomic Energy Society of Japan

雑誌

日本原子力学会和文論文誌 (ISSN:13472879)

巻号頁・発行日

vol.8, no.2, pp.154-164, 2009 (Released:2012-02-22)

参考文献数

20

被引用文献数

6 8

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In Japan, safety of nuclear facilities is regulated by the central government and local governments are responsible for protecting the local public. To operate nuclear facilities in local communities, local governments would conclude safety agreements with power companies. In recent years, local governments have used the safety agreements as excuse for delaying the operations of nuclear facilities. The legal basis of the safety agreements was questioned by some who considered that this was the cause of the stranded relationship between local governments and power companies, and in some cases, the interrupted nature of electricity supply. To understand the sources of this difficult relationship, safety agreements must be analyzed, although these documents may have undergone revisions, and various regulations may have changed. By analyzing the safety agreements and revisions, we found that the relationship between local governments and power companies gradually changed over time, which can be divided into the following 3 stages: (1) in the early 70s, the dawn stage when local governments groped with the situation of nuclear facilities built in their communities; (2) from late 70s to 90s, the stage when local governments demanded information, and (3) from late 90s to present, the stage when local governments demand information and trustworthiness. This paper shows the results of analyzing the relationship changes between local governments and power companies. We conclude that viewpoints of local governments on nuclear power evolve, as social responsibilities of power companies stipulated in safety agreements also evolve over time.

著者

大橋 弘史 Steven R. SHERMAN

出版者

Atomic Energy Society of Japan

雑誌

日本原子力学会和文論文誌 (ISSN:13472879)

巻号頁・発行日

vol.7, no.4, pp.439-451, 2008 (Released:2012-03-02)

参考文献数

21

被引用文献数

2

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Tritium migration behavior in the next-generation nuclear plant (NGNP) employing a high-temperature electrolysis (HTE) process to produce hydrogen is estimated by numerical analysis. Estimated tritium concentrations in the hydrogen product and tertiary heat transport fluid in heat exchangers in the HTE process are higher than the limit in drinking water defined by the U.S. Environmental Protection Agency (EPA) and in the effluent at the boundary of an unrestricted area defined by the U.S. Nuclear Regulatory Commission (NRC), respectively. The effects of some countermeasures (i.e., reducing tritium release rate, increasing purification system capacity, removing tritium at high-temperature positions in the heat transport fluids, reducing the permeability of heat exchangers, and hydrogen or water injection in the heat transport fluids) to reduce tritium concentrations in the hydrogen product and tertiary heat transport fluid are proposed and evaluated. The alternative countermeasure proposed in this study to decrease the tritium permeation rate by water injection, which produces HTO from HT according to an isotope exchange reaction (HT+H2O=H2+HTO) in the heat transport fluids, is effective for decreasing the tritium concentrations.

著者

木村 謙仁 木村 浩

出版者

一般社団法人 日本原子力学会

雑誌

日本原子力学会和文論文誌 (ISSN:13472879)

巻号頁・発行日

vol.17, no.2, pp.42-56, 2018 (Released:2018-05-15)

参考文献数

63

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OPECST (Office parlementaire d’evaluation des choix scientifiques et technologiques) helps decision making of the French Parliament by assessing options of scientific and technological policy. In this process, OPECST holds public hearings and gathers opinions from various participants: experts, industry, citizens, and so on. It held many public hearings and assessments on nuclear energy policy in the 1990s, when French people started to demand more transparency and independence of the nuclear safety regime than before. So we can assume that OPECST helped the reform in the 1990s and in the 2000s that finally established the Law on Transparency and Security in the Nuclear Field. This research aims to precisely clarify its function through a survey of all the reports of OPECST on nuclear safety policy published in this period and of political decisions related to them. As a result, it is shown that the function of OPECST consists of three elements: it defines problems, elaborates policy recommendations from various opinions, and accumulates its survey results in the form of reports. Nowadays, the form of the discussion carried out by OPECST is changing, so we have to learn both its history and recent activities to make a policy-making system that would be suitable in Japan.

著者

鈴木 章悟 永山 明彦 大井 昇

出版者

一般社団法人 日本原子力学会

雑誌

日本原子力学会和文論文誌 (ISSN:13472879)

巻号頁・発行日

vol.4, no.1, pp.1-6, 2005-03-25 (Released:2010-01-21)

参考文献数

5

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Uranium glass is yellowish green glass which is produced by adding a small amount of uranium as a coloring agent into ordinary glasses. The glass has a characteristic of emitting strong fluorescent light of around 550nm under ultraviolet light. Most of uranium glass was fabricated before World War 2 but still small amount of uranium glass is being fabricated nowadays. In this study, various kinds of uranium glass were subjected to the measurement of the γ-ray by Ge semiconductor detector. It was found that in older glasses (fabricated from the late 19th century to the early 20th century), γ-ray peaks from 214Pb and 214Bi were clearly identified which are in equilibrium with 226Ra that was apparently not completely separated from uranium due to the level of purification technology at that period. From the comparison of gamma peaks of 235U and 234Pa, it was found that many uranium glass samples fabricated after World War 2 use depleted uranium. The content of 40K in uranium glass was found to vary with where specimens were produced.

著者

篠田 佳彦 山野 直樹

出版者

一般社団法人 日本原子力学会

雑誌

日本原子力学会和文論文誌 (ISSN:13472879)

巻号頁・発行日

vol.14, no.2, pp.95-112, 2015 (Released:2015-05-15)

参考文献数

32

被引用文献数

1 3

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The Fukushima Daiichi nuclear accident has led to changes in the acceptance of nuclear power in many people. The authors conducted an opinion survey of 300 adult inhabitants of Tsuruga city in Fukui prefecture, Japan. The aim of this survey is to obtain people’s opinions concerning radiation and its risks. Authors classified Tsuruga inhabitants on the basis of responses to questions on the concept and knowledge of risk and the cognition of radiation by factor and cluster analyses of multivariable analysis. Using the results of these analyses, Tsuruga inhabitants have been assigned to five categories: “acceptance group,” “anxiety group,” and three intermediate groups.

著者

綿引 政俊 赤井 昌紀 中井 宏二 家村 圭輔 吉野 正則 平野 宏志 北村 哲浩 鈴木 一敬

出版者

Atomic Energy Society of Japan

雑誌

日本原子力学会和文論文誌 (ISSN:13472879)

巻号頁・発行日

pp.1111290028, (Released:2011-11-30)

参考文献数

9

被引用文献数

1

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Gloveboxes used for plutonium fuel development and fabrication are eventually dismantled for replacement. Since equipment interior and the inner surface of gloveboxes are contaminated with radioactive materials, glovebox dismantling is performed by workers wearing an air fed suit with mechanical tools in a plastic enclosure system to control the spread of contamination. Various improvements of the enclosure system are implemented including the modification of the rooms to decontaminate and undress the air fed suit and the introduction of an inflammable filter and a safety film near the size reduction workspace against fire. We describe the countermeasures deployed in the enclosure system against potential hazards and how these devices work in the real dismantling activities.

著者

酒谷 圭一 中谷 隆良 船橋 英之

出版者

一般社団法人 日本原子力学会

雑誌

日本原子力学会和文論文誌 (ISSN:13472879)

巻号頁・発行日

vol.14, no.4, pp.261-267, 2015 (Released:2015-11-15)

参考文献数

14

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Corrosion rate data for activated metal wastes are necessary for the prediction of the radiological impact of radioactive waste disposal. However, there are no such data available in the consideration of Zr-2.5 wt%Nb alloy, which is used in pressure tubes of the Fugen Nuclear Power Plant. Since the pressure tubes are destined for sub-surface disposal, it is necessary to obtain the corrosion rate of Zr-2.5 wt%Nb alloy under the disposal conditions. In this study, corrosion tests were conducted in high alkalinity and deoxidized water at 30℃ by a gas-accumulating-type corrosion test. The corrosion rate decreased to 3.3-3.9 nm/y and the corrosion thickness increased in proportion to the cubic root of corrosion time until 24 months after the commencement of the test. The results indicate that the corrosion rate would decrease in proportion to the minus cubic root of corrosion time squared if the empirical corrosion characteristic that had been obtained in material research for light-water reactors is applicable.

著者

佃 由晃 林 洋 上村 勝一郎 服部 年逸 金子 浩久 師岡 慎一 光武 徹 秋葉 美幸 安部 信明 藁科 正彦 増原 康博 木村 次郎 田辺 朗 西野 祐治 井坂 浩順 鈴木 理一郎

出版者

Atomic Energy Society of Japan

雑誌

日本原子力学会和文論文誌 (ISSN:13472879)

巻号頁・発行日

vol.1, no.4, pp.384-403, 2002

被引用文献数

5

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Nuclear Power Engineering Corporation (NUPEC) has conducted a proving test for thermal-hydraulic performance of BWR fuel (high-burnup 8×8, 9×9) assemblies entrusted by the Ministry of Economy, Trade and Industry (NUPECTH-B Project). The high-burnup 8×8 fuel (average fuel assembly discharge burnup: about 39.5GWd/t), has been utilized from 1991. And the 9×9 fuel (average fuel assembly discharge burnup: about 45GWd/t), has started to be used since 1999. There are two types (A-type and B-type) of fuel design in 9×9 fuel assembly.<BR>Using an electrically heated test assembly which simulated a BWR fuel bundle on full scale, flow induced vibration, pressure drop, critical power under steady state condition and post-boiling transition (post-BT) tests were carried out in an out-of pile test facility that can simulate the high pressure and high temperature conditions of BWRs. This paper completed the results of 9×9 fuel combined with the previously reported results of high-burnup 8×8 fuel.<BR>As a result of NUPEC-TH-B Project, the validity of the current BWR thermal-hydraulic design method was confirmed and the reliability of BWR thermo-hydraulic fuel performance was demonstrated. Based on the test data, a new correlation of the estimation of fuel rod vibration amplitude, new post-BT heat transfer and rewet correlations for the estimation of fuel rod surface temperature were developed.

著者

西原 健司 山岸 功 安田 健一郎 石森 健一郎 田中 究 久野 剛彦 稲田 聡 後藤 雄一

出版者

Atomic Energy Society of Japan

雑誌

日本原子力学会和文論文誌 (ISSN:13472879)

巻号頁・発行日

vol.11, no.3, pp.247, 2012 (Released:2012-08-15)

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日本原子力学会和文論文誌 Vol. 11, No. 1 (2012), pp.13-19   著者の申し出により,14–16頁の Table 4, 7(1/2), 7(2/2)に誤りがありましたので,PDFの通り訂正いたします。

著者

三田地 紘史

出版者

一般社団法人 日本原子力学会

雑誌

日本原子力学会和文論文誌 (ISSN:13472879)

巻号頁・発行日

vol.16, no.4, pp.173-179, 2017 (Released:2017-11-15)

参考文献数

20

被引用文献数

2

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A study is performed on a molten salt fast reactor (MSFR) of 1.5 GWe output. The reactor is started up by using transuranium elements reprocessed from spent fuel of a BWR. The fuel salt of the reactor is the mixed fluoride salt NaF–KF–UF4–TRUF3, which is reprocessed almost continuously by an oxide-precipitation process during the reactor operation. By performing calculations using the nuclear analysis code PIJ–BURN in SRAC2006 and the nuclear data file of JENDL–3.3, the following results are obtained. (1) The burn-up characteristics of the reactor are mainly determined by the uranium inventory (Uinv) in the reactor and the reprocessing cycle (Lrep), which is the time interval necessary to reprocess all the fuel salt in the primary loop. (2) A large Uinv and short Lrep enhance the breeding performance of the reactor. (3) The period necessary to keep the radioactive waste under control will be about 400 years in the case of Lrep longer than 400 efpd. (4) Power stations consisting of 20 MSFRs (total output of 30 GWe) can operate for 600 years by utilizing 14,000 t of uranium obtained from the spent fuel of LWRs in Japan.

著者

鵜原 壽 結城 英二 月山 和樹 神澤 真人 妹尾 幸一 大田 幸平 嶋本 文夫 大田 成幸

出版者

一般社団法人 日本原子力学会

雑誌

日本原子力学会和文論文誌 (ISSN:13472879)

巻号頁・発行日

vol.13, no.4, pp.127-135, 2014 (Released:2014-11-15)

参考文献数

20

被引用文献数

1 2

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A variety of technologies are applied to the decontamination of radiocesium from water using inorganic adsorbents such as zeolites, Prussian blue (PB) and its analogues. However, these adsorbents are difficult to apply. Although zeolites work as good adsorbents for cesium (Cs) in freshwater, their adsorption ability is extremely low in seawater and fly ash extracts with a high salt concentration. In contrast, PB and its analogues maintain their selective adsorption ability for Cs even in water containing salts, but a high level of cyan remains in the treated water. In this study, we introduce a new technology that utilizes complexes between PB and hydroxides of transition metals (PB-X) for the decontamination of Cs from water and report results of demonstration tests on simulated seawater and fly ash extract. Furthermore, the excellent results of the PB-X method applied to the extracts from fly ash contaminated with radiocesium (more than 8000 Bq/kg) are also shown. It has been proved that radiocesium activities are not only below the detection limit (<10 Bq/kg) and the content of cyan can be controlled under the regulation value of tap water in the water treated with PB-X.

著者

伊藤 健一 宮原 英隆 氏家 亨 武島 俊達 横山 信吾 中田 弘太郎 永野 哲志 佐藤 努 八田 珠郎 山田 裕久

出版者

一般社団法人 日本原子力学会

雑誌

日本原子力学会和文論文誌 (ISSN:13472879)

巻号頁・発行日

vol.11, no.4, pp.255-271, 2012 (Released:2012-11-15)

参考文献数

27

被引用文献数

8 11

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In the radiation dosimetry of radiocesium in Iitate, Fukushima, the level of radiocesium around the environment did not exceed the criteria in liquid phases such as puddle water, but was distributed in solid phases such as some soil types and organic matter. On the other hand, retting of the cut bamboo grass and hemlock fir in water allowed the release of radiocesium, about 230 Bq/kg exceeding the criteria for a bathing area. The flow-thru test using zeolite showed the removal of radiocesium from the liquid phase. The wet classification test was performed for 3 types of radiocesium-contaminated soil. According to the results of wet classification, radiocesium was detected and its level exceeded the cropping restriction level in almost all classified particle fractions. The decontamination effect of wet classification on radiocesium contamination was smaller than that on heavy metal contamination. Specifically, the wet classification could not induce volume reduction. Accordingly, preprocessing and intermediate treatments such as dispersion or attrition by vibration or mixing in the wet classification process were devised and examined as improved processing techniques. As a result, the effectual volume reduction of the radiocesium-contaminated soil was confirmed by adding an intermediate process such as the surface attrition in the vibrator.

著者

安藤 真樹 松田 規宏 斎藤 公明

出版者

一般社団法人 日本原子力学会

雑誌

日本原子力学会和文論文誌 (ISSN:13472879)

巻号頁・発行日

pp.J16.023, (Released:2017-02-28)

参考文献数

29

被引用文献数

21

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In order to determine the contribution of radioactive cesium due to the Fukushima Daiichi nuclear power plant accident to the ambient dose equivalent rates measured by car-borne surveys, natural background radiation was evaluated for eastern Japan as municipality-averaged values. The window count method for the distinction between natural and artificial radioactive nuclides was applied to car-borne surveys using KURAMA–II. The distribution of the evaluated natural background radiation reflected geological features, and it was found that the radiation measured along paved roads reflected the distribution of terrestrial gamma rays. The contribution of the radioactive cesium as of 2014 for the municipalities designated as the Intensive Contamination Survey Area was beyond the uncertainty of the natural background radiation. That for the other municipalities, however, was found to be almost negligible.

著者

平山 英夫 松村 宏 波戸 芳仁 佐波 俊哉

出版者

一般社団法人 日本原子力学会

雑誌

日本原子力学会和文論文誌 (ISSN:13472879)

巻号頁・発行日

vol.12, no.4, pp.304-310, 2013 (Released:2013-11-15)

参考文献数

4

被引用文献数

1 3

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A method was presented to estimate radionuclide concentration in plume using the pulse height distribution measured by a LaBr3 scintillation detector and its calculated response to radionuclides in plume with egs5. Radionuclide concentration was estimated from the ratio between the peak count rates corresponding to each radionuclide in the measured pulse height distribution on an expressway on March 15 and in the calculated one from each radionuclide in plume using the egs5 Monte Carlo code. The pulse height distribution reconstructed based on the estimated concentrations agrees well with the measured one at the time that the contribution from radionuclides deposited on a ground surface is negligible.

著者

石田 倫彦 林 芳昭 上田 吉徳 吉田 一雄

出版者

一般社団法人 日本原子力学会

雑誌

日本原子力学会和文論文誌 (ISSN:13472879)

巻号頁・発行日

vol.9, no.1, pp.71-81, 2010 (Released:2012-02-08)

参考文献数

17

被引用文献数

1 2

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A special committee on “Research on the analysis methods for accident consequence of nuclear fuel facilities (NFFs)” was organized by the Atomic Energy Society of Japan (AESJ) under the entrustment of Japan Atomic Energy Agency (JAEA). The committee aims to research on the state-of-the-art consequence analysis method for Probabilistic Safety Assessment (PSA) of NFFs, such as fuel reprocessing and fuel fabrication facilities. The objective of this research is to obtain the useful information related to the establishment of quantitative performance objectives and to risk-informed regulation through qualifying issues needed to be resolved for applying PSA to NFFs. The research activities of the committee were mainly focused on the analysis method of consequences for postulated accidents with potentially large consequences in NFFs, e.g., events of criticality, spill of molten glass, hydrogen explosion, boiling of radioactive solution, and fire (including rapid decomposition of TBP complexes), resulting in the release of radioactive materials into the environment. The results of the research were summarized in a series of six reports, which consist of a review report and five technical ones. In this technical report, the research results about basic experimental data related to the consequence of the radiolytically generated hydrogen gas explosion postulated in the radioactive solution reserve tank caused by the loss of dilution air supply were summarized.