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Ono, Masato; Goto, Minoru; Shinohara, Masanori; Nojiri, Naoki; Tochio, Daisuke; Shimazaki, Yosuke; Yanagi, Shunki
JAEA-Technology 2013-001, 35 Pages, 2013/03
JAEA-Technology-2013-001.pdf:6.04MB
The temperature coefficient measurements of the HTTR have been carried out. In the beginning of the operation, temperature coefficients at the reactor power of 30 kW and 9 MW were obtained through 1999 to 2000. The operation days of the HTTR fuel reached 375 Effective Full Power Days (EFPD), which is over a half of design operation days (660 EFPD). The temperature coefficient measurements were conducted at the same power levels of 30 kW and 9 MW to evaluate burnup effect. Also, to measure temperature coefficient in high accuracy, technique of core temperature control and technique of core temperature homogenization were established.
Rai, D.*; Yui, Mikazu
JAEA-Technology 2013-002, 35 Pages, 2013/05
JAEA-Technology-2013-002.pdf:1.18MB
The solubility method is one of the most powerful tools to obtain reliable thermodynamic data for (1) solubility products of discrete solids and double salts, (2) complexation constants for various ligands, (3) development of data in a wide range of pH values, (4) evaluation of data for metals that form very insoluble solids (e.g. tetravalent actinides), (5) determining solubility-controlling solids in defferent types of wastes and (6) elevated temperature for redox sensitive systems. This document is focused on describing various aspects of obtaining thermodynamic data using the solubility method. This manuscript deals with various aspects of conducting solubility studies, including selecting the study topic, modeling to define important variables, selecting the range of variables and experimental parameters, anticipating results, general equipment requirements, conducting experiments, and interpreting experimental data.
Hirayama, Takashi; Kannari, Masaaki
JAEA-Technology 2013-003, 33 Pages, 2013/06
JAEA-Technology-2013-003.pdf:1.95MB
The optimization of the JAEA network system has been promoted in accordance with the optimization plan which has the fundamental principles of ensuring its dependability, information security and usability. In respect to ensuring the dependability, we addressed to (a) the reduction of both trouble probability and recovery time, and (b) an execution of the business continuity plan in time of large-scale earthquake. For the latter, we installed an e-mail backup server and an alternate connection to the internet in Kansai Photon Science Institute (Kizu-area) based on lesson learned from the experience of the Tohoku earthquake on March 11th, 2011. In addition, we introduced a backup system for data and servers of other main IT infrastructure services. This report documents the configuration and operation of the backup system.
Mizuno, Tomoyasu; Koyama, Shinichi; Kaito, Takeji; Uwaba, Tomoyuki; Tanaka, Kenya
JAEA-Technology 2013-004, 16 Pages, 2013/05
JAEA-Technology-2013-004.pdf:1.52MB
In order to be an alternative concept to the conventional concept consisting of mixed oxide (MOX) fuel, Metallic fuel, U-Pu(TRU)-Zr metallic fuel slug and ODS cladding were considered for Sodium-cooled fast reactor (SFR) cycle system. The capability of the U-Pu(TRU)-Zr metallic fuel with ODS cladding under a high burnup condition was calculated and conducted by a simplified calculation grogram developed in JAEA. The fuel temperature profiles, gap width profiles, and radial temperature distribution at EOL were evaluated. Those results show that the fuel pin had enough safety margin to fuel melting under the irradiation. Evaluation of the profiles of plenum gas pressure and the cladding deformation after irradiation shows that the fuel pin had enough plenum volume not to cause considerable cladding deformations by plenum gas pressure. In case of 0.4% Am bearing fuel, calculation result shows that fuel centerline temperature becomes high, but increase from U-Pu-Zr fuel is insignificant.
Mizuno, Tomoyasu; Koyama, Shinichi; Kaito, Takeji; Uwaba, Tomoyuki; Tanaka, Kenya
JAEA-Technology 2013-005, 17 Pages, 2013/05
JAEA-Technology-2013-005.pdf:1.7MB
A mixed oxide fuel pin concept with annular pellets and an oxide dispersion strengthened martensitic steel (ODS) cladding is a possible driver fuel for commercialized Sodium-cooled fast reactor (SFR) core. The capability of annular MOX fuel pins with (U,Pu) oxide fuel and Am bearing oxide fuel under a high burnup condition was evaluated by a fuel performance analysis code CEDAR developed in JAEA. The fuel temperature profiles, fuel and cladding deformation profiles, and radial temperature distribution at EOL were evaluated. Those results show that the fuel pin had enough safety margin to fuel melting under the irradiation. Also, the profiles of pressure on the cladding inner surface and the cladding deformation after irradiation were evaluated. Those results show that the gap of the fuel pin at fabrication had enough width not to occur the considerable fuel-cladding mechanical interaction (FCMI).
Mizuno, Tomoyasu; Koyama, Shinichi; Kaito, Takeji; Uwaba, Tomoyuki; Tanaka, Kenya
JAEA-Technology 2013-006, 17 Pages, 2013/05
JAEA-Technology-2013-006.pdf:1.72MB
As a fuel concept for commercialized Sodium-cooled fast reactor (SFR) system, minor actinides (MA) bearing oxide fuel with oxide dispersion strengthened martensitic steel (ODS) cladding was considered under homogeneous TRU recycling strategy. The MA content is calculated to be around 5% of heavy metal in case of trans-uranium (TRU) feed from light water reactor (LWR) spent fuel during the transition phase from LWR to fast reactor era. The fuel temperature profiles, fuel and cladding deformation profiles, and radial temperature distribution at end of life (EOL) were evaluated by fuel performance analytical code CEDAR developed in JAEA to investigate the irradiation behavior of annular MOX fuel pins with (U,Pu) oxide fuel and Am bearing oxide fuel under a high burnup condition. Also, the profiles of pressure on the cladding inner surface and the cladding deformation after irradiation were evaluated.
Mizuno, Tomoyasu; Koyama, Shinichi; Kaito, Takeji; Uwaba, Tomoyuki; Tanaka, Kenya
JAEA-Technology 2013-007, 17 Pages, 2013/05
JAEA-Technology-2013-007.pdf:1.69MB
As a swelling resistant austenitic steel, PNC316, is candidate cladding tube material of the first core of demonstration Sodium-cooled fast reactor (SFR). The irradiation behavior of an annular MOX fuel pin with (U,Pu) oxide fuel contained in PNC 316 cladding was evaluated by a fuel performance analysis code CEDAR developed in JAEA. The fuel temperature profiles, fuel and cladding deformation profiles, and radial temperature distribution at EOL were evaluated. Those results show that the fuel pin keeps its integrity at least up to 100 GWd/t of peak burnup.
Nagata, Hiroshi; Inoue, Shuichi; Yamaura, Takayuki; Tsuchiya, Kunihiko; Nagao, Yoshiharu
JAEA-Technology 2013-008, 30 Pages, 2013/06
JAEA-Technology-2013-008.pdf:2.24MB
Refurbishment of JMTR was completed in FY2010. For damage caused by the 2011 off the Pacific coast of Tohoku Earthquake, the repair of facilities was completed in October 2012. Currently, the JMTR is in preparation for restart. Irradiation tests for LWRs safety research, science and technologies, etc. are expected after the JMTR restart. On the other hand, aiming at the attractive irradiation testing reactor, the usability improvement has been discussed. As a part of the usability improvement, shortening of turnaround time was discussed focusing on the fabrication process of irradiation capsules, where the fabrication process was analyzed and reviewed by referring a trial fabrication of the mockup capsule. As a result, it was found that the turnaround time can be shortened 2 months from fabrication period of 6 months with communize of irradiation capsule parts, application of ready-made instrumentation including the sheath heater, reconsideration of inspection process, etc.
Mizuno, Tomoyasu; Koyama, Shinichi; Kaito, Takeji; Uwaba, Tomoyuki; Tanaka, Kenya
JAEA-Technology 2013-009, 12 Pages, 2013/06
JAEA-Technology-2013-009.pdf:1.3MB
Metallic fuel, U-Pu(TRU)-Zr is a fuel candidate for Sodium-cooled fast reactor (SFR) selected as a possible promising future nuclear reactor system in Generation-IV international forum (GIF). Design studies were performed in the Japanese feasibility study on commercialized fast reactor cycle system, and the irradiation behavior of metallic fuel is under investigation through analytical fuel performance code calculations with preliminary analytical models. As fuel temperature analyses at overpower events are also major interest, some calculations of U-Pu(TRU)-Zr fuel irradiation performance were conducted by a simplified calculation program developed in JAEA. The calculated fuel temperature at the maximum power of overpower events, 110-120% of steady state power, was around 1100K in maxim. It is clear that this temperature was low enough to avoid fuel melting in the event.
Mizuno, Tomoyasu; Koyama, Shinichi; Kaito, Takeji; Uwaba, Tomoyuki; Tanaka, Kenya
JAEA-Technology 2013-010, 17 Pages, 2013/06
JAEA-Technology-2013-010.pdf:2.46MB
Metallic fuel, U-Pu(TRU)-Zr is a fuel candidate for Sodium-cooled fast reactor (SFR) selected as a possible promising future nuclear reactor system in Generation-IV international forum (GIF). Design studies were performed in the Japanese feasibility study on commercialized fast reactor cycle system, and the irradiation behavior of metallic fuel is under investigation through analytical fuel performance code calculations with preliminary analytical models. Some calculations of U-Pu(TRU)-Zr fuel irradiation performance were conducted by a simplified calculation grogram developed in JAEA. Axial profile of fuel pin centerline temperature calculated by using effective fuel thermal conductivity where sodium ingress into fuel was considered fits well with actual fuel micro structures after the irradiation. The effective fuel thermal conductivity with sodium ingress is suitable for the irradiation behavior investigation.
Mizuno, Tomoyasu; Koyama, Shinichi; Kaito, Takeji; Uwaba, Tomoyuki; Tanaka, Kenya
JAEA-Technology 2013-011, 10 Pages, 2013/06
JAEA-Technology-2013-011.pdf:2.02MB
In order to evaluate integrity limiting parameters of fuel pins during fast reactor core transient events, such as fuel center line temperature and cladding maximum temperature, the fast reactor fuel pin performance code CEDAR was used for calculation. The temperature histories of fuel pins during a loss of flow (LOF) type transient events was calculated based on Ross & Stoute type gap conductance model and constant gap conductance model used in a core transient calculation code like HIPRAC. The calculated maximum temperatures of cladding and adjacent coolant channel were lower in the case with Ross & Stoute type model than in the case of constant gap conductance model due to the dynamic change of gap conductance of the former case. It is indicated that core transient calculations with constant gap conductance give conservative cladding and coolant temperatures than that with Ross & Stoute type gap conductance model which is thought to be realistic.
Mizuno, Tomoyasu; Koyama, Shinichi; Kaito, Takeji; Uwaba, Tomoyuki; Tanaka, Kenya
JAEA-Technology 2013-012, 13 Pages, 2013/06
JAEA-Technology-2013-012.pdf:2.34MB
A mixed oxide fuel pin concept with annular pellets and an ODS cladding is a possible driver fuel for commercialized Sodium-cooled fast reactor (SFR) core. This fuel concept was considered with low breeding ratio as a standard, break-even breeding cores and cores with high breeding ratio (high breeding cores). Some calculations of fuel pin irradiation performance of (U,Pu) oxide fuel and minor actinides bearing oxide fuel were conducted by a fuel performance analysis code CEDAR developed in JAEA to understand the steady state irradiation behavior of fuel pins for the cores with high breeding ratio. The fuel temperature profiles, fuel and cladding deformation profiles, and radial temperature distribution at end of life (EOL) were evaluated. Those results show that the MOX fuel pin having the specifications and irradiation conditions used in this investigation would be irradiated moderately up to approximately 250 GWd/t with well integrity.
Takemoto, Noriyuki; Sugaya, Naoto; Otsuka, Kaoru; Hanakawa, Hiroki; Onuma, Yuichi; Hosokawa, Jinsaku; Hori, Naohiko; Kaminaga, Masanori; Tamura, Kazuo*; Hotta, Koji*; et al.
JAEA-Technology 2013-013, 44 Pages, 2013/06
JAEA-Technology-2013-013.pdf:4.42MB
A real-time simulator for operating both a reactor and irradiation facilities of a materials testing reactor, Simulator of Materials Testing Reactors, was developed for understanding reactor behavior and upskilling in order to utilize for a nuclear human resource development and to promote partnership with developing countries which have a plan to introduce nuclear power plant. The simulator is designed based on the JMTR (Japan Materials Testing Reactor) and it simulates operation, irradiation tests and various kinds of anticipated operational transients and accident conditions caused by the reactor and irradiation facilities. The development of the simulator was sponsored by the Japanese government as one of the specialized projects of advanced research infrastructure in order to promote basic as well as applied researches. This report summarizes the simulation model, hardware specification and operation procedure of the simulator.
Harigae, Hitoshi; Takagi, Tsuyohiko; Hamano, Tomoharu; Nakamura, Shoichi; Oba, Toshio; Ebashi, Masaaki; Okuda, Eiichi; Kinoshita, Tomonobu
JAEA-Technology 2013-014, 150 Pages, 2013/07
JAEA-Technology-2013-014.pdf:24.38MB
In-Vessel Transfer Machine (IVTM) came off from the gripper claw in the Auxiliary Handling Machine (AHM) and fell at a height of approximately two meters during a withdrawal work of the IVTM in the Fast Breeder Reactor (FBR) Monju. The withdrawal work of IVTM from the reactor vessel by AHM was performed. The work, however, was suspended due to the excessive load alarm. To grasp the situation of the IVTM fall, observation of the machine was necessary. An interior observation and an exterior observation of the dropped IVTM were performed. As a result of these observations, the radially deformed lower end of the upper guide tube was observed at the connection part, and it was jammed in the fuel throat sleeve when the dropped IVTM was withdrawn. Based on this information, the IVTM could be safely withdrawn from the reactor vessel with the fuel throat sleeve.
Sato, Hiroyuki; Ohashi, Hirofumi; Tazawa, Yujiro; Imai, Yoshiyuki; Nakagawa, Shigeaki; Tachibana, Yukio; Kunitomi, Kazuhiko
JAEA-Technology 2013-015, 68 Pages, 2013/06
JAEA-Technology-2013-015.pdf:2.22MB
In present study, requirements in order to design, construct and operate hydrogen production plants coupled to HTGRs under conventional chemical plant standards are identified. In addition, design considerations for safety design of nuclear facility are suggested. Furthermore, feasibility of proposed safety design and design considerations are clarified.
Ohashi, Hirofumi; Sato, Hiroyuki; Yan, X.; Sumita, Junya; Nomoto, Yasunobu; Tazawa, Yujiro; Noguchi, Hiroki; Imai, Yoshiyuki; Tachibana, Yukio
JAEA-Technology 2013-016, 176 Pages, 2013/09
JAEA-Technology-2013-016.pdf:8.62MB
JAEA has started a conceptual design of a 50MWt small-sized high temperature gas cooled reactor for steam supply and electricity generation (HTR50S), which is a first-of-kind of the commercial plant or a demonstration plant of a small-sized HTGR system for steam supply to the industries and district heating and electricity generation by a steam turbine. The plant design of HTR50S for the steam supply and electricity generation was performed based on the plant specification and the requirements for each system taking into account for the increase of the reactor outlet coolant temperature from 750
C to 900C and the installation of IHX. The technical feasibility of HTR50S was confirmed because the designed systems satisfies the design requirements. The conceptual plant layout was also determined. This paper provides the summary of the plan design and technical feasibility of HTR50S.
Ohashi, Hirofumi; Sato, Hiroyuki; Tazawa, Yujiro; Aihara, Jun; Nomoto, Yasunobu; Imai, Yoshiyuki; Goto, Minoru; Isaka, Kazuyoshi; Tachibana, Yukio; Kunitomi, Kazuhiko
JAEA-Technology 2013-017, 71 Pages, 2014/02
JAEA-Technology-2013-017.pdf:3.64MB
Japan Atomic Energy Agency (JAEA) has started a conceptual design of a 50 MWt small-sized high temperature gas cooled reactor (HTGR) for steam supply and electricity generation (HTR50S). Though the safety design of HTR50S was determined based on that of the High Temperature Engineering Test Reactor (HTTR) for the early deployment of HTR50S, the shutdown cooling system, which is the forced cooling heat removal system, was categorized as non-safety class to optimize the protection to provide the highest level of safety that can reasonably be achieved, and the vessel cooling system, which is categorized as the safety class system, was designed as a passive safety features. The preliminary safety analysis of HTR50S for the rupture of co-axial hot gas duct in primary cooling system and the tube rupture of steam generator was conducted to confirm the adequacy of the safety design. It was confirmed that the analysis results satisfied the acceptance criteria.
Kato, Yuko; Yabuki, Kentaro*; Okawachi, Yasushi
JAEA-Technology 2013-018, 118 Pages, 2013/07
JAEA-Technology-2013-018.pdf:3.73MB
The prototype fast breeder reactor MONJU resumed the system startup test (SST) on May 6th 2010 after fourteen years and five months shutdown since the sodium leakage of the secondary heat transport system in December 1995 and reached criticality on May 8th. Core confirmation test (CCT) is the first step of SST which consists of three steps, and finished on July 22nd after 78 days test. Control rod reactivity worth measurements were carried out in order to calibrate the reactivity worth of control rods and back-up rods. In addition, we also aimed at a basic data acquisition for the control rod reactivity worth calibration.
Okada, Yuji; Magome, Hirokatsu; Hanawa, Hiroshi; Omi, Masao; Kanno, Masaru; Iida, Kazuhiro; Ando, Hitoshi; Shibata, Mitsunobu; Yonekawa, Akihisa; Ueda, Haruyasu
JAEA-Technology 2013-019, 236 Pages, 2013/10
JAEA-Technology-2013-019.pdf:45.07MB
In Japan Atomic Energy Agency, in order to solve the problem in the long-term operation of a light water reactor, preparation which does the irradiation experiment of light-water reactor fuel and material is advanced. JMTR stopped after the 165th operation cycle in August 2006, and is advancing renewal of the irradiation facility towards re-operation. This material irradiation test facility and power ramping test facility for doing the neutron irradiation test of the fuel and material for light water reactors is scheduled to be manufactured and installed between the 2008 fiscal year and the 2012 fiscal year. This report summarizes manufacture and installation of the material irradiation test facility for IASCC research carried out from the 2008 fiscal year to the 2010 fiscal year.
Noguchi, Hiroki; Kubo, Shinji; Iwatsuki, Jin; Onuki, Kaoru
JAEA-Technology 2013-020, 38 Pages, 2013/07
JAEA-Technology-2013-020.pdf:3.24MB
Hazardous substances such as sulfuric acid, sulfur dioxide and hydrogen iodide acid are employed in thermochemical Iodine-sulfur (IS) process. It is necessary to take safety measure against workers and external environments to study experimentally on IS process. Presently we have been conducting to verify the soundness of main components made of engineering material in actual corrosive condition. An integrity test apparatus for the components of sulfuric acid decomposition was set up. We will use the hazardous substances such as sulfuric acid and sulfur dioxide and perform the experiment in pressurized condition in this integrity test. Safety measures for test apparatus, operation and abnormal situation were considered prior to starting the test. This report summarized the consideration results for the safety measures on the integrity test apparatus for the components of sulfuric acid decomposition.
Watahiki, Shunsuke; Hanakawa, Hiroki; Imaizumi, Tomomi; Nagata, Hiroshi; Ide, Hiroshi; Komukai, Bunsaku; Kimura, Nobuaki; Miyauchi, Masaru; Ito, Masayasu; Nishikata, Kaori; et al.
JAEA-Technology 2013-021, 43 Pages, 2013/07
JAEA-Technology-2013-021.pdf:5.12MB
The number of research reactors in the world is decreasing because of their aging. On the other hand, the necessity of research reactor, which is used for human resources development, progress of the science and technology, industrial use and safety research is increasing for the countries which are planning to introduce the nuclear power plants. From above background, the Neutron Irradiation and Testing Reactor Center began to discuss a basic concept of Multipurpose Compact Research Reactor (MCRR) for education and training, etc., on 2010 to 2012. This activity is also expected to contribute to design tool improvement and human resource development in the center. In 2011, design study of reactor core, irradiation facilities with high versatility and practicality, and hot laboratory equipment for the production of Mo-99 was carried out. As the result of design study of reactor core, subcriticality and operation time of the reactor in consideration of an irradiation capsule, and about the transient response of the reactor to the reactivity disturbance during automatic control operation, it was possible to do automatic operation of MCRR, was confirmed. As the result of design study of irradiation facilities, it was confirmed that the implementation of an efficient mass production radioisotope Mo-99 can be expected. As the result of design study with hot laboratory facilities, Mo-99 production, RI export devised considered cell and facilities for exporting the specimens quickly was designed.
Kuji, Masayoshi*; Asai, Hideaki*; Hashizume, Shigeru; Horiuchi, Yasuharu; Sato, Toshinori; Matsui, Hiroya
JAEA-Technology 2013-022, 72 Pages, 2013/10
JAEA-Technology-2013-022.pdf:5.3MB
Rock mass classifications are used for design and construction of underground structures. However, the classification methods commonly used in Japan are qualitative and inadequate for estimating the actual mechanical properties of a rock mass based on site specific geological features. Considering the design, construction and safe operation of large underground facilities, an important requirement is to utilize a rock mass classification method that can estimate site specific rock mechanical properties based on surface-based investigations and geological observations during excavation. For this study, a new quantitative rock mass classification method based on JGS standard was proposed and applied to the sedimentary formations and the granite at MIU. The results were compared with the rock mass classification system developed by CRIEPI and commonly used in JAPAN. Then the applicability of the new rock mass classification could be evaluated.
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.
Shibata, Akira; Miura, Kuniaki*; Takeuchi, Tomoaki; Otsuka, Noriaki; Nakamura, Jinichi; Tsuchiya, Kunihiko
JAEA-Technology 2013-024, 21 Pages, 2013/10
JAEA-Technology-2013-024.pdf:12.62MB
In the Fukushima-Daiichi Nuclear Power Plant Accident, the measurements of water level in pressure vessel and spent fuel pool were impossible due to station blackout, and it resulted in difficulty for countermeasures against the accidents and for understanding of the situations of reactor core after accidents. Therefore, we started to develop a new water level sensor for l with high reliability, which works with small electric power. This report describes reviews of conventional water level sensor and design and production of new water level sensor. After production of the sensor, performance tests were performed between room temperature and about 95 C, and the it was confirmed that the sensor is able to measure water level with the accuracy of 20 mm. As the results, a perspective to use the new water level sensor as water level indicator for spent fuel pools and reactor vessels after severe accident.
Mo manufacturing process, 1; Reusability evaluation of Mo absorbent (Joint research)Kimura, Akihiro; Niizeki, Tomotake*; Kakei, Sadanori*; Chakrova, Y.*; Nishikata, Kaori; Hasegawa, Yoshio*; Yoshinaga, Hideo*; Chakrov, P.*; Tsuchiya, Kunihiko
JAEA-Technology 2013-025, 40 Pages, 2013/10
JAEA-Technology-2013-025.pdf:2.62MB
Neutron Irradiation and Testing Reactor Center has developed the production of a medical isotope of Mo, the parent nuclide of 
Tc by the (n,
) method using JMTR. The (n,) method has an advantage of easy manufacturing process and low radioactive wastes generation. However, the low radioactivity concentration of Tc is remaining as an issue. Therefore, PZC and PTC have been developed as adsorbent of molybdenum. Meanwhile, it is necessary to recycle the absorbent and Mo for the reduction of the radioactive waste of used-adsorbent and the effective use of limited resources, respectively. This report summarizes results of the synthesis of Mo adsorbents such as PZC and PTC, and the performance tests.
Takato, Kiyoto; Murakami, Tatsutoshi; Suzuki, Kiichi; Shibanuma, Kimikazu; Hatanaka, Nobuhiro; Yamaguchi, Bungo; Tobita, Yoshimasa; Shinozaki, Masaru; Iimura, Naoto; Okita, Takatoshi; et al.
JAEA-Technology 2013-026, 42 Pages, 2013/10
JAEA-Technology-2013-026.pdf:3.17MB
In order to cope with making a commercial fast reactor fuel burn-up higher, oxygen-to-metal (O/M) ratio in the fuel specification is designed to 1.95. As the test for the fabrication of such low O/M ratio pellets, two kinds of O/M ratio preparation tests of different reduction mechanism were done. In the first test, we evaluated the technology to prepare the O/M ratio low by annealing the sintered pellets in production scale. In addition, we know from past experience that O/M ratio of the sintered pellets can be reduced by residual carbon when the de-waxed pellets with high carbon content are sintered. Thus, in another test, the green pellets containing a large amount of organic additives were sintered and we evaluated the technology to produce the low O/M ratio sintered pellets by the reduction due to residual carbon. From the first test results, we found a tendency that the higher annealing temperature or the longer annealing time resulted in the lower O/M ratio. However, the amount of O/M ratio reduction was small and it is estimated that a substantial annealing time is necessary to prepare the O/M ratio to 1.95. It is considered that reducing O/M ratio by annealing was difficult because atmosphere gas containing oxygen released from pellets remained and the O/M ratio was changed to the value equilibrated with the gas having high oxygen potential. From another test results, it was confirmed that O/M ratio was reduced by the reduction due to residual carbon. We found that it was important to manage an oxygen potential of atmosphere gas in a sintering furnace low to reduce the O/M ratio effectively.
Ishigami, Tsutomu; Sukegawa, Takenori*; Mukai, Masayuki
JAEA-Technology 2013-027, 124 Pages, 2013/10
JAEA-Technology-2013-027.pdf:5.9MB
In order to safely and efficiently implement decommissioning of nuclear installations, it is important to beforehand predict decommissioning project management data (PMD) and to develop a decommissioning plan based on the predicted results. The PMD prediction is made with PMD evaluation equations including model parameters such as unit work activity coefficients. Although model parameter values developed so far include uncertainties, little evaluation of the uncertainties and resulted uncertainties in predicted PMD has been made. However information on the uncertainties is valuable in flexibly studying and developing a decommissioning plan. We therefore studied and evaluated uncertainties in model parameters by analyzing the JPDR decommissioning experience data. This report describes an evaluation method of the model parameter uncertainties and their evaluated results.
Watanabe, Akihiko; Sakai, Kenji; Oi, Motoki; Meigo, Shinichiro; Takada, Hiroshi
JAEA-Technology 2013-028, 21 Pages, 2013/11
JAEA-Technology-2013-028.pdf:2.46MB
The General Control System (GCS) of the MLF of J-PARC has a problem that it costs very much in the maintenance because of its poor flexibility on OS, etc. For resolving the problem, we have re-examined framework and application softwares for the MLF-GCS, in considering functions that PLCs in many local control panels are controlled by the plural exclusive PCs, and operating data over 7000 are acquired, stored and distributed with suitable data format by shared servers. Furthermore, we have made a prototype of an upgraded GCS and evaluated its concrete performances with true data such as communication speed between the PLCs and PCs, control functions from operating windows, storage capability of data server, and long-term stability of the system. In conclusion, we decided to adopt following softwares for the upgraded GCS: EPICS as framework software, Takebishi OPC server as data input/output module, CSS as user interface window and PostgreSQL for the data storage server.
Sagawa, Naoki; Izaki, Kenji; Mizuniwa, Harumi*
JAEA-Technology 2013-029, 28 Pages, 2013/11
JAEA-Technology-2013-029.pdf:3.35MB
Developed in the Mixed Oxide fuel fabrication facility, Basic handling of IP, analysis method, detection of Pu and radioactivity quantification. These are the conditions that the exposure conditions and the analysis condition are constant. However, in the case of Contamination has occurred in the workplace, contaminated samples are not only Pu. It may contain a Pb or RnTn. Then, if other work is being carried out in a room that is the operation of IP, it's difficult to darken the room. PSL is reduced when light hits the IP. In this study, we have investigated in order to upgrade radiation protection, in the case of Containing the Pb or PSL reduction by light. Additionally, regarding change of analysis condition, the analysis method was examined when the resolution was changed to 50 micro, 100 micro and 200 micro.
Nakanishi, Chika; Sato, Takeshi; Sato, Sohei; Nagai, Haruyasu; Kakefuda, Toyokazu; Katata, Genki; Tsuzuki, Katsunori; Ikeda, Takeshi; Okuno, Hiroshi; Yamamoto, Kazuya; et al.
JAEA-Technology 2013-030, 105 Pages, 2013/10
JAEA-Technology-2013-030.pdf:29.0MB
North Korea carried out the third nuclear test in February 2013. Due to the request of the Ministry of Education, Culture, Sports, Science and Technology (MEXT), Nuclear Emergency Assistance and Training Center (NEAT) and Nuclear Science and Engineering Directorate (NSED) of JAEA predicted the atmospheric dispersion of radionuclide by WSPEEDI-II for the purpose of contributing to the environmental monitoring plan. From February 12 to 22, they provided daily reports on the prediction to the MEXT and the Ministry of Defense. MEXT has published these reports on the website. Since April 2012, NEAT and NSED had prepared to predict by the framework for the prediction around the clock during 10months until February 2013. This report described this experience and pointed issues out on this system.
Okano, Fuminori; Ikeda, Yoshitaka; Sakasai, Akira; Hanada, Masaya; Ichige, Hisashi; Miyo, Yasuhiko; Kaminaga, Atsushi; Sasajima, Tadayuki; Nishiyama, Tomokazu; Yagyu, Junichi; et al.
JAEA-Technology 2013-031, 42 Pages, 2013/11
JAEA-Technology-2013-031.pdf:18.1MB
The disassembly of JT-60 tokamak device and its peripheral equipments, where the total weight was about 6200 tons, started in 2009 and accomplished in October 2012. This disassembly was required process for JT-60SA project, which is the Satellite Tokamak project under Japan-EU international corroboration to modify the JT-60 to the superconducting tokamak. This work was the first experience of disassembling a large radioactive fusion device based on Radiation Hazard Prevention Act in Japan. The cutting was one of the main problems in this disassembly, such as to cut the wielded parts together with toroidal field coils, and to cut the vacuum vessel into two. After solving these problems, the disassembly completed without disaster and accident. This report presents the outline of the JT-60 disassembly, especially tokamak device.
Okano, Fuminori; Masaki, Kei; Yagyu, Junichi; Shibama, Yusuke; Sakasai, Akira; Miyo, Yasuhiko; Kaminaga, Atsushi; Nishiyama, Tomokazu; Suzuki, Sadaaki; Nakamura, Shigetoshi; et al.
JAEA-Technology 2013-032, 32 Pages, 2013/11
JAEA-Technology-2013-032.pdf:8.86MB
Japan Atomic Energy Agency started to construct a fully superconducting tokamak experiment device, JT-60SA, to support the ITER since January, 2013 at the Fusion Research and Development Directorate in Naka, Japan. The JT-60SA will be constructed with enhancing the previous JT-60 infrastructures, in the JT-60 torus hall, where the ex-JT-60 machine was disassembled. The JT-60SA Cryostat Base, for base of the entire tokamak structure, were assembly as first step of this construction. The Cryostat Base (CB, 250 tons) is consists of 7 main made of stainless steel, 12m diameter and 3m height. It was built in the Spain and transported to the Naka site with the seven major parts split, via Hitachi port. The assembly work of these steps, preliminary measurements, sole plate adjustments of its height and flatness, and assembly of the CB. Introduces the concrete result of assembly work and transport of JT-60SA cryostat base.
Takebe, Shinichi; Sasaki, Toshihisa; Saito, Tatsuo; Yamaguchi, Naoko
JAEA-Technology 2013-033, 87 Pages, 2013/11
JAEA-Technology-2013-033.pdf:2.49MB
Materials, etc. in a non-controlled area is, if the criteria (10 microsieverts per year) listed below that in "A guideline regarding treatment of materials in nuclear facilities considering the influence of fallout released from the accident of TEPCO's Fukushima Dai-ichi Nuclear Power Station" (March 30, 2012), that in accordance with relevant laws and regulations, such as "Waste Disposal and Cleaning Act" (Act No. 137 of 1970), effective use as a resource or be properly disposed of is required. In this paper, in order to effectively use as resources or properly disposed of the materials, etc. in a non-controlled area, radioactivity concentration of materials, etc. in you see " For clearance level specified in Radiation Hazards Prevention Law" and "Radionuclide Concentrations for Materials not Requiring Treatment as Radioactive Wastes Generated from Dismantling etc. of Reactor Facilities and Nuclear Fuel Use Facilities (in Japanese), NSC Japan, 2005." corresponds to the dose of the above criteria the presented results have been estimated as an example.
Homma, Fumitaka; Inoi, Hiroyuki; Watanabe, Shuji; Fukutani, Koji*
JAEA-Technology 2013-034, 57 Pages, 2013/12
JAEA-Technology-2013-034.pdf:11.01MB
Emergency generator of HTTR started in the blackout occurred just after an Tohoku Pacific Ocean Earthquake on March 11, 2011 with an intensity of 5 upper on the Japanese seven stage seismic scale and its duration time was long. In addition, we suffer from multiple severe aftershocks just after the start of emergency generators. Emergency generator of HTTR was able to supply output electric power sufficiently and stably to required loads. We carried on integrity check of the emergency generator for the HTTR after the earthquake. In particular, we put emphasis on finding faults caused by thee earthquake shaking. As a result, we found that the erosion in a combustion liner, and the condition of erosion was very strange and rare. Therefore, we carried out investigations of causes of erosion, and change of specifications for combustion liner to prevent erosion. This measure improve the reliability for the further Large-Scale earthquake.
Yoshida, Kazuo; Abe, Hitoshi
JAEA-Technology 2013-035, 14 Pages, 2013/12
JAEA-Technology-2013-035.pdf:2.18MB
Boiling accidents of reprocessed liquid wastes are postulated to be occurred caused by the loss of cooling function for waste storage tanks due to a total loss of AC power persisting over a long period of time at fuel reprocessing facilities. Some amounts of radioactive materials could be released from facilities by vapor flow from a boiling liquid waste storage tank. Thermal-hydraulic behavior of water and nitric acid vapor and aerosol behaviors in compartments of facility building are needed to be analyzed for assessing amount of released radioactive materials to outside of facilities. The amount of water and nitric acid vapor, which is one of key parameters to estimate the duration of evaporation to dryness and the amounts of transferring radioactive materials, is calculated based on the liquid/vapor equilibrium data of liquid wastes. This report described the experiment details and its results to obtain the equilibrium data of multicomponent nitrate solution.
Nakata, Hisakazu; Amazawa, Hiroya; Sakai, Akihiro; Kurosawa, Ryohei; Kanno, Naohiro*; Kashima, Takahiro*; Sakamoto, Yoshiaki
JAEA-Technology 2013-036, 47 Pages, 2014/02
JAEA-Technology-2013-036.pdf:5.27MB
The Low-level Radioactive Waste Disposal Project Center will construct near surface disposal facilities. The disposal facilities consist of concrete pit type for low-level radioactive wastes and trench type for very low level radioactive wastes. As for the trench type disposal facility, two kinds of facility designs are on projects -one for normal trench type disposal facilities and the other for trench type disposal facilities with geomembrane liners that could prevent from causing environmental effects of non radioactive toxic materials. This study examined mechanical strength and permeability properties to assess the durability on the basis of an indoor accelerated exposure experiment targeting the liner materials presumed to avail the conceptual design so far. Its results will be used for the basic and detailed design henceforth by confirming the empirical degradation characteristic with the progress of the exposure time.
Tsuda, Shuichi; Yoshida, Tadayoshi; Nakahara, Yukio; Sato, Tetsuro; Seki, Akiyuki; Matsuda, Norihiro; Ando, Masaki; Takemiya, Hiroshi; Tanigaki, Minoru*; Takamiya, Koichi*; et al.
JAEA-Technology 2013-037, 54 Pages, 2013/10
JAEA-Technology-2013-037.pdf:4.94MB
JAEA has been performing dose rate mapping in air using a car-borne survey system KURAMA-II. The KURAMA system is a GPS-aided mobile radiation monitoring system that has been newly developed by Kyoto University Research Reactor Institute in response to the nuclear disaster. The KURAMA system is composed of an energy-compensated scintillation survey meter for measuring dose rate, electric device for controlling both the dose rates and the position data from a GPS module, a computer server for processing and analyzing data from KURAMA, and client PCs for providing for end users. The KURAMA-II has been improved in small-packaging, durability, and automated data transmission. In consequence, dose rate mapping in wide area has become possible in shorter period of time. This report describes the construction of KURAMA-II, its application and a suggestion of how to manage a large number of KURAMA-II.
Okuda, Eiji; Suzuki, Toshiaki; Fujinaka, Hideaki
JAEA-Technology 2013-038, 42 Pages, 2014/01
JAEA-Technology-2013-038.pdf:82.62MB
With the incident as an opportunity, repair techniques were developed in Joyo. The cover gas boundary components on rotating plug were removed as a preparation work for Joyo restoration work. The removed components are the door valve for fuel handling hole, hold-down shaft driving mechanism and door valve for inspection hole (A). Because these components were designed as the maintenance free. These components haven't been overhauled for more than 30 years since the construction of Joyo. Since the cover gas radioactivity was quite low due to the long time reactor shutdown, the new method was applied in this removal work. In order to prevent the contaminant of impurities in cover gas, the boundary was secured by temporary green house with slight negative pressure. Achievement of this work and accumulated experience will be able to provide valuable insights for further improving and verifying repair techniques in sodium cooled fast reactors.
Sakai, Akihiro; Kurosawa, Ryohei; Hara, Hironori*; Nakata, Hisakazu; Amazawa, Hiroya; Arikawa, Masanobu*; Sakamoto, Yoshiaki
JAEA-Technology 2013-039, 228 Pages, 2014/02
JAEA-Technology-2013-039.pdf:24.05MB
The sensitivity analysis of doses in terms of the environmental conditions was performed by statistical method in order to make the technical basis for the siting criteria of near surface disposal facility for low level radioactive waste generated from research, industrial and medical facilities. Doses calculated at all assumed pathways in more than 97.5% of calculation cases were able to be reduced below the target dose after control period (0.01 mSv/y) by means of equipping the disposal facility with additional engineered barriers. As a result, we concluded it was possible to safely and rationally design disposal facilities in most of the environmental parameters related to safety assessment. Another sensitivity analysis was done in order to discuss the area of disposal site. Dose at the site boundary were able to be reduce below the target dose during operation (0.05 mSv/y) whenever the distances from these facilities to the site boundary were more than 120 m, respectively.
Nakamura, Yasuyuki; Tezuka, Masashi; Iwai, Hiroki; Sano, Kazuya
JAEA-Technology 2013-040, 80 Pages, 2014/02
JAEA-Technology-2013-040.pdf:4.29MB
It was reported that Fukushima Daiichi Nuclear Power Plant (1F) had been lost the function of cooling the reactor by the Tohoku Earthquake. It is assumed that the original shapes of the internal core are not be kept and the inside of the reactor make so narrow in the space, however the fuel debris and the molten internal core will have to be removed for the decommissioning of 1F. The cutting methods for those removal works will have to be selected depending on the situation of the inside of the reactor. In consideration of above situations, the plasma-arc cutting method, Fugen has much data of underwater cutting for the reactor dismantling and there are experiences of the reactor dismantling in both domestic and international, will be being developed for the fuel debris removal works and so on.
Iwai, Hiroki; Nakamura, Yasuyuki; Tezuka, Masashi; Sano, Kazuya
JAEA-Technology 2013-041, 57 Pages, 2014/02
JAEA-Technology-2013-041.pdf:7.01MB
It was reported that Fukushima Daiichi Nuclear Power Plant (1F) had been lost the function of cooling the reactor by the Tohoku Earthquake. It is assumed that the original shapes of the internal core are not kept and the inside of the reactor makes so narrow in the space, however the fuel debris and the molten internal core will have to be removed for the decommissioning of 1F. The cutting methods for those removal works will have to be selected depending on the situation of the inside of the reactor. In consideration of above situations, the abrasive water jet cutting method, Fugen has much data of underwater cutting for the reactor dismantling and there are experiences of the reactor maintenance and dismantling in both domestic and international, will be being developed for the fuel debris removal works and so on. In the fiscal year 2012, in order to confirm the cutting performance of the cutting machine, the cutting tests were carried out to acquire the fundamental data.
Ono, Masato; Shinohara, Masanori; Iigaki, Kazuhiko; Tochio, Daisuke; Nakagawa, Shigeaki; Shimazaki, Yosuke
JAEA-Technology 2013-042, 45 Pages, 2014/01
JAEA-Technology-2013-042.pdf:10.51MB
In HTTR, it has passed about two years since the last performance confirmation test. During two years, the integrity of active equipment, leakage efficiency of coolant pressure boundary of piping and vessel and control system performance due to influence of damage and deterioration by earthquake and aging were not confirmed. To confirm them, the cold test by using HTTR was conducted and the system performances such as above mentioned items were evaluated by comparing with the plant data obtained by the past cold test. In the result, no abnormity was found in all the data in the cooling system of HTTR, and it was confirmed that the integrity of facilities and instruments of HTTR was maintained in good condition.
Shibata, Akira; Takeuchi, Tomoaki; Otsuka, Noriaki; Saito, Takashi; Onoda, Shinobu; Oshima, Takeshi; Tsuchiya, Kunihiko
JAEA-Technology 2013-043, 24 Pages, 2014/01
JAEA-Technology-2013-043.pdf:2.76MB
The big tsunami wave caused by the Great East Japan Earthquake triggered station black out of the Fukushima Daiichi Nuclear Power Plant. The schedule until ending of the decommission is shown in the guidance from Japanese government. But the dose rate in the reactor vessel is quite high and it is not possible to specify the position of melted fuel debris by visual inspection. That is one of the most important issues in this process. This report describes the development of Self Powered Gamma Detector (SPGD) for the purpose to specify the position of melted fuel debris and situation in the reactor by measuring rate in the reactor of the Fukushima Daiichi Nuclear Power Plant. The irradiation examinations by changing the parameter of emitters figure were performed and the dependency of SPGD output on emitter shape was summarized.
Tsuyuguchi, Koji; Kuroiwa, Hiroshi; Kawamoto, Koji; Yamada, Nobuto; Onuki, Kenji; Iwatsuki, Teruki; Takeuchi, Ryuji; Ogata, Nobuhisa; Suto, Masahiro; Mikake, Shinichiro
JAEA-Technology 2013-044, 89 Pages, 2014/02
JAEA-Technology-2013-044.pdf:11.09MB
JAEA-Technology-2013-044-appendix(CD-ROM).zip:316.8MB
This document summarizes the data of pilot boreholes (12MI27, 12MI33) in the -500m Access/Research Gallery-North. The geological, hydraulic and geochemical data were obtained. In addition, groundwater monitoring system was installed in closure test gallery for the flooding test in phase III research. The results of investigation, biotite granite with medium to coarse-grained equigranular texture are characterized. Rock mass classification is B from CH class. Minor fault with fault gouge that was not presumed by an original model are observed in 12MI33. Density of fracture in 12MI27 near the Main-shaft fault tends to be compared to 12MI33. Water inflow in both boreholes is less. Permeability ranges from 4.8E-10 to 6.1E-09m/sec at the zone without alteration and with low inflow, from 1.1E-07 to 2.7E-07m/sec at the zone without alteration and with high inflow, respectively. Groundwater chemistry is rich in Na and Cl ion.
Hiyama, Kazuhisa; Hanawa, Nobuhiro; Kurosawa, Akihiko; Eguchi, Shohei; Hori, Naohiko; Kusunoki, Tsuyoshi; Ueda, Hisao; Shimada, Hiroshi; Kanda, Hiroaki*; Saito, Isamu*
JAEA-Technology 2013-045, 32 Pages, 2014/02
JAEA-Technology-2013-045.pdf:5.83MB
This report summarizes regarding to develop of real-time multifunctional access control system which is able to manage worker's access control and exposure dose at real-time in the reactor building, besides worker's location and worker might be fall down by accident.
Kasai, Noboru; Iwanade, Akio; Ueki, Yuji; Saiki, Seiichi; Hoshina, Hiroyuki; Seko, Noriaki
JAEA-Technology 2013-046, 25 Pages, 2014/02
JAEA-Technology-2013-046.pdf:9.63MB
To remove their radioactive species which have long radioactive half-life from the circumstances as rapidly as possible, we developed novel radioactive cesium adsorbents containing ammonium 12-molybdophosphate, which had adsorption selectivity for cesium ion, by radiation grafting method. The bench-scale equipment 150 times as large volume as laboratory scale was established for graft polymerization. The radioactive cesium adsorbents 1,000 times as large as laboratory scale were successfully synthesized with the bench-scale equipment. Moreover, the adsorption performance with radioactive cesium in environmental water was evaluated at field tests in Fukushima Prefecture. As a result, the adsorbents could successfully remove radioactive cesium dissolved in environmental water below the detection limit of radioactivity concentration.
Ijiri, Yuji*; Noda, Masaru*; Nobuto, Jun*; Matsui, Hiroya; Mikake, Shinichiro; Hashizume, Shigeru
JAEA-Technology 2013-047, 819 Pages, 2014/03
JAEA-Technology-2013-047-01.pdf:41.49MBJAEA-Technology-2013-047-02.pdf:25.26MB
The researches on engineering technology in the Mizunami Underground Research Laboratory plan consists of (1) research on engineering technology at a deep underground, and (2) research on engineering technology as abasis of geological disposal. The former research mainly aimed in this study are categorized in (a) development of design and construction planning technologies, (b) development of construction technology, (c) development of countermeasure technology, (d) development of technology for security. In this study, the researches on engineering technology are proceeded in these four categories by using data measured down to GL-460m during construction as a part of the second phase of the MIU plan.
pellets (Joint research)Kimura, Akihiro; Nishikata, Kaori; Nikolayevich, A.*; Vladimirovna, T.*; Chakrova, Y.*; Tsuchiya, Kunihiko
JAEA-Technology 2013-048, 30 Pages, 2014/03
JAEA-Technology-2013-048.pdf:3.0MB
In this study, the irradiation tests of the high-density MoO pellets and PIEs were carried out with WWR-K for the realization of Mo/Tc production procedure by the (n,) method. High-density MoO pellets were irradiated. After neutron irradiation, the irradiated pellets were carried out PIEs, and the pellets were sound from the results. The irradiated pellets were also dissolved with NaOH solution at 100C. The solution speed of the pellets at 100C was faster than that at 50C and the it was clear that dissolved temperature of pellet was important factor for the solution speed. Mo adsorption/Tc elution tests were carried out with PZC and PTC. It was obtained that the properties of Mo adsorption/Tc elution of these Mo adsorbents was equivalent in previous results. As the these results, the prospects are bright for the realization of Mo production procedure by the (n,) method.
-eyeIIKanayama, Fumihiko; Okada, Takashi; Fukushima, Mineo; Yoshimoto, Katsunobu*; Hanyu, Toshinori; Kawanobe, Takayuki
JAEA-Technology 2013-049, 60 Pages, 2014/03
JAEA-Technology-2013-049.pdf:25.84MB
For planning of removing fuels and debris from the Unit 2 reactor building, TEPCO has already started to measure dose rate over the floor by remotely operated vehicle. Because the measured data were widely distributed in the range of several decades to one thousand mSv/h, it is necessary for TEPCO to survey of contamination distribution on operation floor 2 for more detail planning. JAEA estimated sensitivity of developed gamma camera system named "-eye II" in consistency with actual radiation condition, and carried a demonstration experiment at Fukushima Daiichi N.P.P. to confirm a strength of jamming by back ground dose. Then, JAEA surveyed contamination distribution of operating floor using -eye II. At the result of survey, it was found that, - main radiation source in survey area was located on upper reactor well, - western floor in survey area was lower the margin of capacity of -eye II, -there was a highly contaminated spot on the floor near the opened BOP.
Zaima, Naoki; Nakashima, Shinichi; Nakatsuka, Yoshiaki; Fujiki, Naoki*; Kado, Kazumi
JAEA-Technology 2013-050, 39 Pages, 2014/03
JAEA-Technology-2013-050.pdf:3.56MB
A uranium mass assay system NWAS, for 200-litter wastes drums applied by NDA method was developed and accumulated the data of the actual uranium bearing wastes drums. The system consists of the 16 pieces of Helium-3 proportional counters for neutron detection generated from U-234(
,n) reaction or U-238 spontaneous fissions with polyethylene moderation and a Germanium solid state detector for ray detection as to determine uranium enrichment. The satisfactory works had been continued and the uranium determination data of 850 drums had been accumulated approximately. On the other hand several considerable problems on the system or methodology had been revealed technically or analytically through the measurements experiences. Furthermore as the next improvement plans, the active neutrons assay for uranium bearing wastes drums are now progressing. The results of complications will lead us to the progressive next steps.