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Miyakawa, Kazuya; Hayano, Akira; Sato, Naomi; Nakata, Kotaro*; Hasegawa, Takuma*
JAEA-Data/Code 2023-009, 103 Pages, 2023/09
JAEA-Data-Code-2023-009.pdf:9.29MB
JAEA-Data-Code-2023-009-appendix1(DVD-ROM).zip:271.51MBJAEA-Data-Code-2023-009-appendix2(DVD-ROM).zip:883.78MBJAEA-Data-Code-2023-009-appendix3(DVD-ROM).zip:10.29MB
This borehole investigation was carried out to confirm the validity of the distribution of low flow areas deep underground estimated based on the geophysical survey in FY 2020, as a part of an R&D supporting program titled "Research and development on Groundwater Flow Evaluation Technology in Bedrock" under contract to the Ministry of Economy, Trade and Industry (2021, 2022 FY, Grant Number: JPJ007597). The borehole name is Horonobe Fossil seawater Boring-1 and is referred to as HFB-1 borehole. HFB-1 is a vertical borehole drilled adjacent to the Horonobe Underground Research Laboratory (URL), which was drilled from the surface to a depth of 200 m in FY2021 and from a depth of 200 m to 500 m in FY2022. This report summarizes information related to the drilling of HFB-1 and various data (rock core description, geophysical logging, chemical analysis, etc.) obtained from the borehole investigation.
Kirishima, Akira*; Terasaki, Mariko*; Miyakawa, Kazuya; Okamoto, Yoshihiro; Akiyama, Daisuke*
Chemosphere, 289, p.133181_1 - 133181_12, 2022/04
Times Cited Count：1 Percentile：6.09(Environmental Sciences)no abstracts in English
Nakayama, Masashi; Saiga, Atsushi; Kimura, Shun; Mochizuki, Akihito; Aoyagi, Kazuhei; Ono, Hirokazu; Miyakawa, Kazuya; Takeda, Masaki; Hayano, Akira; Matsuoka, Toshiyuki; et al.
JAEA-Research 2019-013, 276 Pages, 2020/03
JAEA-Research-2019-013.pdf:18.72MB
The Horonobe Underground Research Laboratory (URL) Project is being pursued by the Japan Atomic Energy Agency (JAEA) to enhance the reliability of relevant disposal technologies for geological disposal of High-level Radioactive Waste through investigations of the deep geological environment within the host sedimentary rock at Horonobe Town in Hokkaido, north Japan. The investigations will be conducted in three phases, namely "Phase 1: Surface based investigations", "Phase 2: Construction phase" (investigations during construction of the underground facilities) and "Phase 3: Operation phase" (research in the underground facilities). According to the research plan described in the 3rd Mid- and Long- term Plan of JAEA, "Near-field performance study", "Demonstration of repository design option", and "Verification of crustal-movement buffering capacity of sedimentary rocks" are important issues of the Horonobe URL Project, and schedule of future research and backfill plans of the project will be decided by the end of 2019 Fiscal Year. The present report summarizes the research and development activities of these 3 important issues carried out during 3rd Medium to Long-term Research Phase.
Kimuro, Shingo; Kirishima, Akira*; Kitatsuji, Yoshihiro; Miyakawa, Kazuya; Akiyama, Daisuke*; Sato, Nobuaki*
Journal of Chemical Thermodynamics, 132, p.352 - 362, 2019/05
Times Cited Count：10 Percentile：44.91(Thermodynamics)A combination of potentiometry and calorimetry was used for the determination of the thermodynamic quantities of complexation of generic and groundwater humic acid (HA), which was isolated from deep groundwater at Horonobe, Hokkaido, Japan, with copper (II) ions and uranyl (VI) ions. The apparent complexation constant of Horonobe HA was independent of the pH, whereas that of generic HA was dependent on the pH. This observation indicates that the polyelectrolyte effect of Horonobe HA is negligible because of its small molecular size. In addition, the effect of the heterogeneity of Horonobe HA was not significant. Moreover, the complexation enthalpy of Horonobe HA was consistent with that of homogeneous poly(acrylic acid), which means the complexation of Horonobe HA was not affected by the functional group heterogeneity. Consequently, the characteristic complexation mechanism of Horonobe HA was revealed based on the determined thermodynamic quantities.
Kimuro, Shingo*; Kirishima, Akira*; Nagao, Seiya*; Saito, Takumi*; Amano, Yuki; Miyakawa, Kazuya; Akiyama, Daisuke*; Sato, Nobuaki*
Journal of Nuclear Science and Technology, 55(5), p.503 - 515, 2018/05
Times Cited Count：7 Percentile：58.07(Nuclear Science & Technology)no abstracts in English
Sato, Toshinori; Sasamoto, Hiroshi; Ishii, Eiichi; Matsuoka, Toshiyuki; Hayano, Akira; Miyakawa, Kazuya; Fujita, Tomoo*; Tanai, Kenji; Nakayama, Masashi; Takeda, Masaki; et al.
JAEA-Research 2016-025, 313 Pages, 2017/03
JAEA-Research-2016-025.pdf:45.1MB
The Horonobe Underground Research Laboratory (URL) Project is being pursued by the Japan Atomic Energy Agency (JAEA) to enhance the reliability of relevant disposal technologies through investigations of the deep geological environment within the host sedimentary formations at Horonobe, northern Hokkaido. This report summarizes the results of the Phase II investigations carried out from April 2005 to June 2014 to a depth of 350m. Integration of work from different disciplines into a "geosynthesis" ensures that the Phase II goals have been successfully achieved and identifies key issues that need to made to be addressed in the Phase II investigations Efforts are made to summarize as many lessons learnt from the Phase II investigations and other technical achievements as possible to form a "knowledge base" that will reinforce the technical basis for both implementation and the formulation of safety regulations.
Sasaki, Takayuki*; Kokami, Takayuki*; Kobayashi, Taishi*; Kirishima, Akira*; Murakami, Hiroaki; Amano, Yuki; Mizuno, Takashi; Iwatsuki, Teruki; Sasamoto, Hiroshi; Miyakawa, Kazuya
Journal of Nuclear Science and Technology, 54(3), p.373 - 381, 2017/03
Trace amounts of natural thorium and uranium in deep groundwater were investigated at two underground research laboratories situated at Horonobe and Mizunami, Japan. The groundwater was sampled from underground boreholes, and the colloid contribution was checked by in situ two size-fractionated ultrafiltration systems. A decrease in the concentration after in situ filtration suggested the presence of natural colloids and suspended matter that were carriers of a portion of the elements. The result of the Th and U concentrations in groundwater after 10 kDa filtration was analyzed thermodynamically using existing hydrogeological and geochemical data such as the mineral components in the groundwater at a given pH, ionic strength, concentration of co-existing ions, redox potential, and solid phase assumed. A crystalline solid phase made the solubility very low compared with that of the amorphous phase, and the solubility agreed well with the concentrations measured.
Kirishima, Akira*; Kuno, Atsushi*; Amamiya, Hiroki; Kubota, Takumi*; Kimuro, Shingo*; Amano, Yuki; Miyakawa, Kazuya; Iwatsuki, Teruki; Mizuno, Takashi; Sasaki, Takayuki*; et al.
Chemosphere, 168, p.798 - 806, 2017/02
Times Cited Count：3 Percentile：10.21(Environmental Sciences)For better understanding of the migration behavior of minor actinides (MA) in deep groundwater, the interaction of doped rare earth elements (REEs) and components in Horonobe deep groundwater was studied. Appx. 10 ppb of rare earth elements, i.e., Y, La, Ce, Pr, Nd, Sm, Eu, Gd, Tb, Dy, Er, Tm and Yb were doped to the sample groundwater collected from a packed sections in borehole drilled from 140 m depth experiment drift of Horonobe underground research laboratory (URL), Hokkaido, Japan. Then, that groundwater was sequentially filtrated by 0.2 micron pore filter, 10 kDa, 3 kDa and 1 kDa of nominal molecular weight limit (NMWL) ultrafilters by keeping inert condition. After that, the filtrate solutions were analyzed by ICP-MS to determine the concentrations of retained REEs at each filtration steps, while the used filters were analyzed by the neutron activation analysis (NAA) and TOF-SIMS element mapping to know the amount and chemical speciation of trapped fraction of the REEs on each filter. A remarkable relation between the retention ratios of REEs in the filtrate solutions and the ionic radius was observed, i.e., smaller rare earth element solves more in liquid phase under the Horonobe groundwater condition. NAA and TOF-SIMS analyses revealed that certain portions of REEs were trapped by 0.2 micron pore filters as rare earth phosphates which corresponded with the predicted predominant species by a chemical equilibrium calculation for the Horonobe groundwater condition, while small portions of colloidal REEs were trapped by 10 kDa and 3 kDa NMWL ultrafilters. The result suggested that phosphate anion plays an important role in the chemical behavior of REEs in saline (seawater based) groundwater, which could be referred for the prediction of migration behavior of trivalent actinide released from the repository of radioactive waste in far future.
Nishi, Yoshihisa*; Ueda, Nobuyuki*; Kinoshita, Izumi*; Miyakawa, Akira; Kato, Mitsuya*
Proceedings of 14th International Conference on Nuclear Engineering (ICONE-14) (CD-ROM), 10 Pages, 2006/07
CERES is plant system analysis code for LMRs developed by the Central Research Institute of Electric Power Industry (CRIEPI). CERES has a function of calculating multidimensional flow in the plena of a coolant in addition to that in one-dimensional plant network calculation. To verify the CERES code, analyses were performed by using the result of the plant trip test from the partial power operation of the prototype FBR "MONJU" that had been executed in December, 1995. The verification work was performed as a joint research of CRIEPI and JAEA. (1)Analysis concerning the primary/secondary/auxiliary cooling system (the plenum in the reactor vessel (R/V) is modeled in R-Z 2-dimension). (2)Analysis concerning the flow pattern in the plenum of R/V (the plenum is modeled in 3-dimension). (3)Analysis concerning the flow pattern inside the IHX plenum (the plenum in the IHX is modeled in 3-dimension). Analytical results by the CERES code showed good agreement with the results of the test of the "MONJU". Fundamental abilities of the CERES as a plant dynamics calculation code had been verified through these analyses. Additionally, some characteristic flows in plena of "MONJU" became clear by these analyses.
Morizono, Koji; Rekimoto, Masafumi; Jo, Takahisa; Miyakawa, Akira
Proceedings of International Conference on Nuclear Energy System for Future Generation and Global Sustainability (GLOBAL 2005) (CD-ROM), 4 Pages, 2005/10
The construction program of the prototype FBR MONJU which has suspended since the sodium leak accident in 1995 has recently restarted and is under construction to implement various improvements. The test run program following the reconstruction work could roughly divided into two phases of Function test and System start-up test (SST). The requirements, test items and the schedule of the function test program of MONJU have been discussed.
Power up test(40% power)Miyakawa, Akira; Hattori, Kazuhiro; Jo, Takahisa
JNC TN2410 2005-002, 278 Pages, 2005/07
JNC-TN2410-2005-002.pdf:30.04MB
In December, 1992, the performance tests (System Startup Tests; SST) of Monju had been started after the integrated function tests. However, coolant leakage in secondary heat transport system occurred at 8th December, 1995. And so, the performance test was interrupted at the stage of the plant output about 40%. Then, it is also in a plant stop state now.In SST, the plant data was measured in all each stage of reactor power from fuel loading to rated power. And, based on the measurement data, the check of the system performance, the evaluation of appropriate design, and the consolidation of the measurement data for future etc. were carried out. There were about 130 test items in SST, 2/3 items of total were indispensable to starting of a plant (these are equivalent to starting test items of LWR), 1/3 items were R&D. Here, summary of past performance test result are organized.
Nishi, Yoshihisa*; Ueda, Nobuyuki*; Kinoshita, Izumi*; Miyakawa, Akira; Kato, Mitsuya*
JNC TY2400 2005-001, 66 Pages, 2005/06
JNC-TY2400-2005-001.pdf:7.59MB
Multi-dimensional thermal-hydraulic characteristic of the coolant in the reactor vessel (R/V) influences the temperature at the plant transitional condition of fast breeder reactor (FBR). CRIEPI is developing plant dynamics calculation code CERES for FBR that adds multi-dimensional thermal-hydraulic analysis function to one-dimensional system calculation code to evaluate the temperature distribution in high accuracy. The temperature distribution affects the integrity of equipments of FBR. To verify the CERES code, analyses were performed by using the result of the plant trip test from the partial power operation of the prototype fast breeder reactor
Miyakawa, Akira
IAEA/TWGFR Gijutsu Kaigi, 0 Pages, 2004/12
Monju system start-up tests
Miyake, Osamu; Hamada, Hirotsugu; Tanabe, Hiromi; Wada, Yusaku; Miyakawa, Akira; Okabe, Ayao; Nakai, Ryodai; Hiroi, Hiroshi
JNC TN2400 2003-003, 225 Pages, 2004/02
JNC-TN2400-2003-003.pdf:40.45MB
The model has been developed for the assessment of the overheating tube failure in an event of sodium-water reaction accident of fast breeder reactor's steam generators (SGs). The model has been applied to the Monju SG studies.
Miyake, Osamu; Hamada, Hirotsugu; Tanabe, Hiromi; ; Miyakawa, Akira; Okabe, Ayao;
JNC TN9400 2001-130, 235 Pages, 2002/03
JNC-TN9400-2001-130.pdf:7.05MB
The model has been developed for the assessment of the overheating tube failure in an event of sodium-water reaction accident of fast breeder reactor's steam generators (SGs). The model has been applied to the Monju SG studies. Major results obtained in the studies are as follows: (1)To evaluate the structural integrity of tube material, the strength standard for 2.25Cr-1Mo steel was established taking account of time dependent effect based on the high temperature (700-1200
C) creep data. This standard has been validated with the tube rupture simulation test data. (2)The conditions for overheating by the high temperature reaction were determined by use of the SWAT-3 experimental data. The realistic local heating conditions (reaction zone temperature and related heat transfer conditions) for the sodium-water reaction were proposed as the cosine-shaped temperature profile. (3)For the cooling effects inside of target tubes, LWR's studies of critical heat flux (CHF) and post-CHF heat transfer correlations have been examined and considered in the model. (4)The model has been validated with experimental data obtained by SWAT-3 and LLTR. The results were satisfactory with conservatism. The PFR superheater leak event in 1987 was studied, and the cause of event and the effectiveness of the improvement after the leak event could be identified by the analysis. (5)The model has been applied to the Monju SG studies. It is revealed consequently that no tube failure occurs in 100%, 40%, and 10% water flow operating conditions when an initial leak is detected by the cover gas pressure detection system.
Miyake, Osamu; Hamada, Hirotsugu; Tanabe, Hiromi; Okabe, Ayao; Miyakawa, Akira
JNC TN9400 2001-099, 76 Pages, 2001/11
JNC-TN9400-2001-099.pdf:2.13MB
The JNC technical report "The Development and Application of overheating Failure Model of FBR Steam Generator Tubes" summarized the assessment method and its application for the overheating tube failure in an event of sodium-water reaction accident of fast breeder reactor's steam generators (SGs). This report describes the following items studied after the publication of the above technical report. (1)0n the basis of the SWAT-3 experimental data, realistic local heating conditions (reaction zone temperature and related heat transfer conditions) for the sodium-water reaction were proposed. New correlations are cosine-shaped temperature profiles with 1,170 C maximum for the 100% and 40% Monju operating conditions, and those with 1,110 C maximum for the 10% condition. (2)For the cooling effects inside of target tubes, LWR's studies of critical heat flux (CHF) and post-CHF heat transfer correlations have been examined and considered in the assessment. The revised assessment adopts the Katto's correlation for CHF, and the Condie-Bengston IV correlation for post-CHF. (3)Other additional examination for the assessment includes treatments of the whole heating effect (other than the local reaction zone) due to the sodium-water reaction, and the temperature-dependent thermal properties of the heat transfer tube material (2.25Cr-1Mo steel). The revised overheating tube failure assessment method has been applied to the Monju SG studies. It is revealed consequently that no tube failure occurs in 100%, 40%, and 10% operating conditions when an initial leak is detected by the cover gas pressure detection system. The assessment for the SG system improved for the detection and blowdown systems shows even better safety margins against the overheating tube failure.
maintenance of the steam generator model (2)Miyakawa, Akira; *
JNC TN2400 99-003, 315 Pages, 1999/06
The action of the pressure descent inside the pipe which transmits the inside sodium temperature of the steam occurrencevessel, heat in MONJU, and the water pipe and seething and others, in the examination, to evaluate the character which can't be measured in detail and it aimed at untying an again rapid change like the time of the rapid blow going down of the steam occurrencevessel in the stable target, and the maintenance of the steam occurrencevessel model had been advanced. In this calligraphy, as a (PLANT TRIP) examination result that it was confirmed that a trouble with the turbine was presumed on December 1, 1995 and a plant stopped safely as the design and the examination of the difference from the analytic result which did a survey before the improvement, and the actual opportunity examination is done by using the steam occurrence vessel model that development maintenance has been advanced to this and the analytic code which improvement is added to is used, and a re-analysis is done, and the result that a comparative examination with the actual opportunity examination data was carried out is reported. And, the maintenance of the plant movement characteristic analytic code is advanced even at present, as for PLANT TRIP examination reappearance analytic result, the result is reported one after another, then, it is reported one after another when there is a change in the model structure, the input and output environment.
Miyakawa, Akira; *
JNC TN2400 99-002, 50 Pages, 1999/02
In Monju, in order to evaluate a plant thermal behavior with sufficient accuracy, the analysis code has been improved according to the data acquired by SKS and SST which were carried out before. Here, the main purpose is improvement of the air cooler model which is the important component which remove the decay hot at the time of a trip. In order to improve, input data of air cooler model was arranged using data measured by the plant trip test. The result is as follows. First, data of the air cooler the outlet Na temperature obtained by the plant trip test was arranged, and it compared with the result of the analysis carried out conventionally. After plant trip, the vane was completely open about 10 minutes. During this period it was found that the outlet Na temperature of an analysis exceeds that of the test. In order to examine the cause of this difference, we investigated using examination results, such as the inlet Na temperature and Na flow rate of an air cooler, and compared the examination result and the calculation result. As a result, the amount of heat exchange found to be calculate little than the actual value. Performance data of an air cooler blower was adjusted according to this result. Consequently, the analysis result has been improved. However, when a vane is controlled and the half-opening state, the outlet Na temperature of an analysis result came to be less than the examination result. This reason is that the flow rate of air in a half-opening state is estimated more than the actual value. By the cause, it is considered to have calculated greatly the amount of heat exchange of an air cooler. Reflecting this result, to reduce the flow rate of air, the pressure loss coefficient of the vane was change to be high so that the amount of heat exchange might become small. Input data was corrected as the above-mentioned. The air cooler model has been improved compared with the result before adjustment. As a result, close agreement between the analysis ...
Ohtaki, Akira; Miyakawa, Akira; Nakai, Satoru
PNC TN9410 95-060, 204 Pages, 1995/02
The modular integrated plant dynamics analysis code "Super-COPD" is being developed. The validation of the code by using the MONJU functional test data was carried out to evaluate the accuracy of the calculation models and the validation results that had been carried out by using scale model test data. The procedure of the validation was as follows: (1)Setting up of the MONJU system data. The plant system data were set up by the tests data. (2)Refinement of the I/O. An input/output program for the MONJU test data was added to the code. (3)Evaluation of the test data. The test data to be used for the validation and program module to be validated were identified. The intermediate heat exchanger (IHX), steam generator (SG) and auxiliary cooling system (ACS) were selected as validation components. (4)Validation. The performance of some components was evaluated. The calculation and modification of the code were carried out. The following results were obtained. (1)IHX. The recommended calculation mesh number and plenum heat capacity and the items to be validated by the start up test were obtained. (2)SG. The recommended heat loss, validity of plenum model and the effectiveness of the flow rate to the thermal calculation were confirmed. A new model for the change of effective flow area in the SG may be needed at very low flow rate. (3)ACS. The ACS thermal/hydraulics were evaluated. The accuracy of air cooler thermal/hydraulics models, interlock/control system models, vane/blower models were validated. The accuracy of the calculation models and the validation results by using scale models was confirmed. Validation of the code using the start up test results under power operation is scheduled.
Nishi, Yoshihisa*; Miyakawa, Akira
no journal, ,
CERES is plant system analysis code for LMRs developed by CRIEPI. CERES has a function of calculating multidimensional flow in the plena of a coolant in addition to that in one-dimensional plant network calculation. To verify the CERES code, analyses were performed by using the result of the plant trip test from the partial power operation of the prototype FBR "MONJU" that had been executed in December, 1995. The verification work was performed as a joint research of CRIEPI and JAEA. The analysis intended for primary/secondary/auxiliary cooling system and the analysis that was focused on three dimension plenum thermal-hydraulic characteristics were executed. Analytical results by the CERES code showed good agreement with the results of the test of the "MONJU". Fundamental abilities of the CERES as a plant dynamics calculation code had been verified through these analyses.
