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JAEA Reports

Proposal of safety demonstration test plan of HTTR by cold test of loss of forced cooling with vessel cooling system inactive

Takada, Shoji; Shinohara, Masanori; Seki, Tomokazu; Shimazaki, Yosuke; Ono, Masato; Tochio, Daisuke; Iigaki, Kazuhiko; Sawa, Kazuhiro

JAEA-Technology 2014-001, 34 Pages, 2014/03

JAEA-Technology-2014-001.pdf:4.46MB

The loss of forced cooling with vessel cooling system inactive has been planned by using HTTR at the reactor power 9 MW. In this test, the forced cooling of reactor core is lost and the vessel cooling system which removes decay heat from core is tripped. In the test, the technical items such that the temperature of water cooling tubes is expected to be higher are considered. The methods to solve such technical items were proposed. The proposed methods were verified based on the test data of the cold test toward the proposal of test plan of safety demonstration test. In the cold test, the two water trains of vessel cooling system was tripped under the condition that the reactor was heated up without nuclear heating. The reactor inlet temperature was set at 120 and 150$$^{circ}$$C.

JAEA Reports

Development of instrument to measure transmission power density distribution using dielectric disk in millimeter waveguide

Yokokura, Kenji; Moriyama, Shinichi; Kobayashi, Takayuki; Hiranai, Shinichi; Sawahata, Masayuki; Terakado, Masayuki; Hinata, Jun; Wada, Kenji; Sato, Yoshikatsu; Hoshino, Katsumichi; et al.

JAEA-Technology 2014-002, 64 Pages, 2014/03

JAEA-Technology-2014-002.pdf:6.83MB

A new instrument has been developed to measure spatial distribution of power density and total power of the millimeter wave, by measuring temperature rise of dielectric material inserted in the waveguide. For a measurement, a dielectric disk with thermally insulated support is inserted into the few millimeters gap in the waveguide. The disk is heated by the millimeter wave pulse for 0.1$$sim$$0.2 s, and immediately after the pulse, it is pulled up and its temperature distribution is measured by an infrared camera to estimate the spatial power density distribution of the millimeter wave. In the other hand, total transmission power is estimated by the disk temperature reached equilibrium. In the measurement test, deformation of the power density distribution was successfully detected when the mirror angle was intentionally changed in the matching optics unit (MOU) at the waveguide input from the gyrotron. The test result shows that the instrument is effective for both adjustment of MOU without opening the vacuum boundary and to detect any abnormal transmission during operation of the ECH system. The test also shows high reliability of the instrument which stands with 1 MW high power transmission without any arcing or pressure rise in vacuum region. The instrument will be contributed to keep good condition of high power long pulse ECH system by detecting abnormal transmission in the waveguide in operation without open vacuum boundary.

JAEA Reports

Disassembly of JT-60 tokamak device and ancillary facilities for JT-60 tokamak

Okano, Fuminori; Ichige, Hisashi; Miyo, Yasuhiko; Kaminaga, Atsushi; Sasajima, Tadayuki; Nishiyama, Tomokazu; Yagyu, Junichi; Ishige, Yoichi; Suzuki, Hiroaki; Komuro, Kenichi; et al.

JAEA-Technology 2014-003, 125 Pages, 2014/03

JAEA-Technology-2014-003.pdf:13.32MB

The disassembly of JT-60 tokamak device and its peripheral equipments, where the total weight was about 5400 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 and ancillary facilities for tokamak device.

JAEA Reports

Development of specimen preparation techniques for pitting potential measurement of irradiated fuel cladding tubes

Suzuki, Kazuhiro; Motooka, Takafumi; Tsukada, Takashi; Terakawa, Yuto; Ichise, Kenichi; Numata, Masami; Kikuchi, Hiroyuki

JAEA-Technology 2014-004, 29 Pages, 2014/03

JAEA-Technology-2014-004.pdf:3.66MB

By the effect of the Great East Japan Earthquake, seawater was injected into spent fuel pools in unit 2, 3 and 4 at Fukushima Daiichi Nuclear Plant in order to cool spent fuels. It is known that chloride ion contained in seawater could cause pitting corrosion for metallic materials. It was concerned that radioactive products inside of fuel cladding tubes might be escaped through the pits. Therefore we have investigated the pit initiation condition of fuel cladding tubes by measuring pitting potential in order to evaluate stability of the enclosure function of fuel cladding tubes in spent fuel pools containing sea salt. In this report, we describe the development of specimen preparation techniques for pitting measurement of spent fuel cladding tubes having high radioactivity. By accomplishing of the development of the specimen preparation techniques, we could evaluate pit initiation condition of spent fuel cladding tubes in water containing sea salt.

JAEA Reports

Development of vapor pressure measurement method with static vapor pressure apparatus for high-corrosion solutions under high-pressure

Takai, Toshihide; Kubo, Shinji

JAEA-Technology 2014-005, 29 Pages, 2014/03

JAEA-Technology-2014-005.pdf:38.08MB

Concerning the iodine-sulfur thermochemical water-splitting process, expanding range of the properties of the HI-I$$_{2}$$-H$$_{2}$$O system (HI$$_{x}$$) is essential for designing distillation columns and for making good choices of operating conditions. A measurement method with a static vapor pressure apparatus was developed for determining vapor pressure of high-temperature and high-pressure HI$$_{x}$$ (up to 3 MPa and 160 $$^{circ}$$C). Preliminary tests employing the pressure gauge for the sample chamber were carried out for comparisons of the direct pressure values and the indirect values. The results of the test using sample solutions of water, HI-H$$_{2}$$O system, and HI$$_{x}$$ showed the two sets of data accorded well, so that the practicability of this vapor pressure measurement method is validated.

JAEA Reports

Storage management of disassembled and radioactive components of JT-60 tokamak device; Storage of radioactive components by containers

Nishiyama, Tomokazu; Miyo, Yasuhiko; Okano, Fuminori; Sasajima, Tadayuki; Ichige, Hisashi; Kaminaga, Atsushi; Miya, Naoyuki; Sukegawa, Atsuhiko; Ikeda, Yoshitaka; Sakasai, Akira

JAEA-Technology 2014-006, 30 Pages, 2014/03

JAEA-Technology-2014-006.pdf:4.87MB

JT-60 tokamak device and the peripheral equipment were disassembled so as to be upgraded to the superconducting tokamak JT-60SA. The disassembled components were stored into storage and airtight containers at the radioactive control area. The total weight and the total number of those components are about 1,100 tons and about 11,500 except for large components. Radiation measurements and records of the radioactive components were required one by one under the law of Act on Prevention of Radiation Disease Due to Radioisotopes, etc. for the control of transport and storage from the radioactive control area to the other area. The storage management of the radioactive components was implemented by establishing the work procedure and the component management system by barcode tags. The radioactive components as many as 11,500 were surely and effectively stored under the law. The report gives the outline of the storage of JT-60 radioactive components by the storage containers.

JAEA Reports

Stabilization of uranium hexafluoride by hydrolysis method for decommissioning of safeguard laboratory facility

Inagawa, Jun; Hotoku, Shinobu; Oda, Tetsuzo; Aoyagi, Noboru; Magara, Masaaki

JAEA-Technology 2014-007, 48 Pages, 2014/03

JAEA-Technology-2014-007.pdf:5.76MB

In safeguard laboratory (SGL) facility of Nuclear Science Research Institute of JAEA, uranium hexafluoride (UF$$_{6}$$) of enriched uranium of various enrichment was used for research and development of a spectrometric method for the determination of the enrichment of uranium in April 1983 through March 1993. After completion of this R&D, the UF$$_{6}$$ has been stored in SGL facility. It was decided that the UF$$_{6}$$ is carried to out of the facility, because the SGL facility will be decommissioning until March 2015. To transport and store in safety after transportation, it is necessary that the UF$$_{6}$$ should be converted to stable chemical form. Hydrolysis of UF$$_{6}$$ to uranyl fluoride (UO$$_{2}$$F$$_{2}$$) and evaporation to solid state were selected for the stabilization method. The equipment for hydrolysis and evaporation was installed in the SGL facility. Stabilization was operated in this equipment, and all of the UF$$_{6}$$ in the SGL facility was converted to UO$$_{2}$$F$$_{2}$$ solid state in October 2012 through August 2013. In this report, results of examination and operation for stabilization of UF$$_{6}$$ were reported.

JAEA Reports

Monju system start-up test report evaluation of the feedback reactivity

Miyagawa, Takayuki*; Kitano, Akihiro; Okawachi, Yasushi

JAEA-Technology 2014-008, 60 Pages, 2014/05

JAEA-Technology-2014-008.pdf:29.75MB

The prototype fast breeder reactor Monju resumed the system startup test (SST) on May 6, 2010 after fourteen year and five month shutdown since the sodium leakage of the secondary heat transport system in December 1995 and reached criticality on May 8, 2010. Core Confirmation Test (CCT) is the first step of SST which consists of three steps, and finished on July 22 after 78 days test. In the evaluation of the feedback reactivity at the part of the CCT, the "self-stability" of Monju was observed when the positive reactivity was inserted with the control rod withdrawal, due to the negative feedback property of the reactor, and due to the control properties of the auxiliary cooling system. Parameters represented with reactor power, sodium temperature of the primary loops became to be stable after transient without any operations. Additionally, the quantitative feedback reactivity was evaluated using the results of this test tentatively.

JAEA Reports

Preliminary evaluation of integrity of coated fuel particles under normal operation in core of small-sized HTGR system HTR50S at 1st. step of Phase I

Aihara, Jun; Goto, Minoru; Inaba, Yoshitomo; Isaka, Kazuyoshi; Ohashi, Hirofumi; Tachibana, Yukio

JAEA-Technology 2014-009, 29 Pages, 2014/05

JAEA-Technology-2014-009.pdf:3.51MB

Japan Atomic Energy Agency (JAEA) is carrying out conceptual design of a 50 MWt small-sized high temperature gas cooled reactor (HTGR), HTR50S. In this report, integrity of coated fuel particles (CFPs) is evaluated for core of HTR50S of 1st. step of phase I (first core of HTR50S) under normal operation. CFPs are considered to be failed by fuel kernel migration by temperature gradient in CFPs or corrosion of SiC layer by fission product Pd (Pd corrosion) or increase in internal pressure under normal operation. In this report, integrity of CFPs is to be maintained for each phenomenon.

JAEA Reports

Outline of device to confirm the responsiveness of control systems for HTTR

Homma, Fumitaka; Hirato, Yoji; Saito, Kenji

JAEA-Technology 2014-010, 64 Pages, 2014/05

JAEA-Technology-2014-010.pdf:35.55MB

After an accident at Fukushima Daiichi Nuclear Plant, the High Temperature Engineering Test Reactor (HTTR) should be kept shut-down till adaptability to the new nuclear safety regulation established by the Nuclear Regulation Authority (NRA) will be confirmed. However, we have to keep the operational ability for the HTTR is indispensable. The operation training for the HTTR using the simulator of light water reactors is not effective. Because the HTTR is the only test reactor in the world and one has many characteristics different from light water reactors. It is an urgent problem to utilize the device which confirms the responsiveness of the control systems for the HTTR (simulator) as a simulator for operation training. This report shows specifications of device, results of simulation and the matter which we should carry out in future.

JAEA Reports

Results of pilot borehole investigation in -500m access/research gallery-south (12MI32 borehole)

Kawamoto, Koji; Kuroiwa, Hiroshi; Yamada, Nobuto; Onuki, Kenji; Omori, Kazuaki; Takeuchi, Ryuji; Ogata, Nobuhisa; Omori, Masaki; Watanabe, Kazuhiko

JAEA-Technology 2014-011, 92 Pages, 2014/07

JAEA-Technology-2014-011.pdf:24.65MB
JAEA-Technology-2014-011-appendix(DVD).zip:331.54MB

This document summarizes the data of pilot boreholes (12MI32) in the -500m Access/Research Gallery-South. The geological, hydraulic and geochemical data were obtained. In addition, groundwater monitoring system was installed to observe the groundwater pressure in initial condition and change during the excavation of gallery. The results of investigation, biotite granite with medium to coarse-grained equigranular texture are characterized. Rock mass classification is B from CM class. Minor fault with fault breccia are observed around 48.90mabh. However, S200_13 fault and IF_SB3_13_3 fault (that were presumed by an original model) were not observed. Density of fracture is large in the section of 40.00 to 80.00mabh. Water inflow was a maximum of 600 L/min in 78.83mabh. Permeability ranges from 2.0E-9 to 1.5E-08m/sec at the zone with low inflow, from 1.1E-05 to 1.6E-05m/sec at the zone with high inflow, respectively. Groundwater chemistry is rich in Na and Cl ion.

JAEA Reports

Further study of measurement of performance of the NDA using Q2 system for uranium waste drum

Naganuma, Masaki; Ohara, Yoshiyuki; Miyamoto, Yasunori*; Murashita, Tatsuya*; Makita, Akinori*; Nohiro, Tetsuya*

JAEA-Technology 2014-012, 11 Pages, 2014/06

JAEA-Technology-2014-012.pdf:1.06MB

In Japan Atomic Energy Agency Ningyo-toge environmental engineering center, exploration for uranium and technical development of uranium refining, conversion and enrichment which are the front end of a nuclear fuel cycle have been performed since 1955. In 2002, we introduced Q2 low-level-waste drum measuring system which is a bulk measuring method of the passive $$gamma$$ ray. In 2007, OS2 analyzing operation system which was used in Q2, was replaced with windows system. This replacement improved the performance of the analysis of Q2. But quantified values of uranium obtained from win system did not correspond exactly to OS2 system. We considered whether the drum which was measured by OS2 system, was measured again by windows system. But it was difficult to measure these drum by win system. So in this study, we studied a calculation method for adjusting quantified values of uranium obtained by each system.

JAEA Reports

Evaluation of infiltration water through the upper cover soil in trench type disposal facility for low level radioactive wastes generated from research, industrial and medical facilities

Kurosawa, Ryohei; Sakai, Akihiro; Nakata, Hisakazu; Amazawa, Hiroya; Sakamoto, Yoshiaki

JAEA-Technology 2014-013, 89 Pages, 2014/06

JAEA-Technology-2014-013.pdf:23.93MB

In the safety assessment for the trench disposal facilities, outflow of radioactive material to the environment is assumed to be due to the percolating rain water into the waste layer, because the waste layer is established above the groundwater level. Therefore, in dose assessment of trench type disposal facilities, it is important to evaluate how the structure of the upper cover soil layers affects the suppressed amount of water infiltration to the waste layer due to rainfall.

JAEA Reports

Evaluation of fission product and actinide release behavior during BWR severe accident focusing on the chemical forms; 2013 annual report

Fukushima Project Team, Oarai Research and Development Center; Fukushima Fuels and Materials Department, Oarai Research and Development Center; Oarai Research and Development Center, Technology Development Department

JAEA-Technology 2014-014, 60 Pages, 2014/07

JAEA-Technology-2014-014.pdf:57.06MB

We have launched a new research program since 2011 for the evaluation of fission product and actinide release behaviour under the severe accident. Chemical forms of fission products were focused on for more accurate evaluation of source term issues. This report describes the progresses and achievements of the research program in 2013. In order to clarify the Cs and I chemistry in the BWR system during the severe accident which include moderator material B$$_{4}$$C, the three research items were configurated: the fission product release kinetics evaluation, the fission product chemical form evaluation and fundamental data acquisition. Basic knowledge on the chemistry of B$$_{4}$$C and CsI have been obtained by using non-radioactive samples.

JAEA Reports

Design of auxiliary shield for remote controlled metallographic microscope

Matsui, Hiroki; Okamoto, Hisato

JAEA-Technology 2014-015, 43 Pages, 2014/06

JAEA-Technology-2014-015.pdf:7.71MB

The remote controlled optical microscope installed in the lead cell at the Reactor Fuel Examination Facility in Japan Atomic Energy Agency has been upgraded to a higher performance unit to study the effect of the microstructural evolution in clad material on the high burn-up fuel behavior under the accident condition. The optical pass of the new microscope requires a new through hole in the shielding lead wall of the cell. To meet safety regulations, auxiliary lead shieldings were designed to cover the lost shielding function of the cell wall. Particle and Heavy Ion Transport Code System (PHITS) was used to calculate and determine the shape and setting positions of the shielding unit. Seismic assessments of the unit were also performed.

JAEA Reports

Development of breast cancer irradiation technique for BNCT at JRR-4

Nakamura, Takemi; Horiguchi, Hironori; Yanagie, Hironobu*; Arai, Masaji

JAEA-Technology 2014-016, 61 Pages, 2014/06

JAEA-Technology-2014-016.pdf:30.48MB

In the department of research reactor and tandem accelerator, developments of irradiation technique with application enlargement for breast cancer on BNCT have been performed in the second medium term plans. We compiled this report about the technological development to solve several problems with the irradiation of breast cancer in the medical irradiation facility of JRR-4. In the present study, design fabrication of a collimator for breast cancer, dose evaluation analysis by clinical model, investigation of dose enhancement at deeper region and investigation of fixing method for breast cancer irradiation were studied. By these evaluation results, we verified that the developed breast cancer irradiation technique can be applied to BNCT medical irradiation of JRR-4. These results are expected to be able to contribute to breast cancer irradiation techniques of other reactor-based BNCT and future accelerator-based BNCT.

JAEA Reports

Evaluation of the performance of the shields in the EPMAs used for radioactive samples

Matsui, Hiroki; Suzuki, Miho; Obata, Hiroki; Kanazawa, Hiroyuki

JAEA-Technology 2014-017, 57 Pages, 2014/06

JAEA-Technology-2014-017.pdf:20.43MB

The Reactor Fuel Examination Facility in JAEA has been used for Post Irradiation Examinations to verify the reliability and safety of the nuclear fuels irradiated in commercial reactors. EPMA (Electron Probe Micro Analyzer) has been utilized for the qualitative analysis of the fission product in the fuel pellet and the detailed observation of the oxide layers formed at the inner and outer surfaces of fuel cladding. Commercial EPMAs were remodeled so that the EPMAs can be applied for radioactive samples. Several shields was set in the EPMA to avoid the $$gamma$$-rays which radiate from a radioactive sample to the proportional counter in the EPMA. It is important to calculate this shielding performance adequately to maintain the precision of analysis. This report describes the results of re-evaluation of the performance of the shields in the EPMAs in the RFEF by using the Particle and Heavy Ion Transport Code System and the examination results of $$gamma$$-ray effect to the X-ray spectrum data by using a radioactive sample.

JAEA Reports

Development of microbeam formation and single-ion hit technologies at the TIARA cyclotron

Yokota, Wataru; Sato, Takahiro; Kamiya, Tomihiro; Okumura, Susumu; Kurashima, Satoshi; Miyawaki, Nobumasa; Kashiwagi, Hirotsugu; Yoshida, Kenichi; Funayama, Tomoo; Sakashita, Tetsuya; et al.

JAEA-Technology 2014-018, 103 Pages, 2014/09

JAEA-Technology-2014-018.pdf:123.66MB

The world's first microbeam focusing technology for heavy ions of hundreds MeV accelerated by a cyclotron has been developed at the TIARA facility in the Takasaki Advanced Radiation Research Institute of the Japan Atomic Energy Agency. The technology enables us to form a microbeam of less than 1 $$mu$$m in diameter and to shoot a specified point on a target by one ion (single-ion hit) with spatial accuracy of microbeam size. In the course of the development, a cyclotron technology to accelerate a small energy-spread beam of hundres MeV, which is necessary for focusing to 1 $$mu$$m, has been developed as well as a beam focusing apparatus, beam size measurement and so forth based on the several-MeV microbeam/single-ion hit system of the TIARA electrostatic accelerators. Applicability of the technologies was examined by actual use in irradiation experiment and the result were fed back to them. This paper reports the process and the results of the development over ten years.

JAEA Reports

Study on engineering technologies in the Mizunami Underground Research Laboratory; FY2012 (Contract research)

Fukaya, Masaaki*; Noda, Masaru*; Hata, Koji*; Takeda, Yoshinori*; Akiyoshi, Kenji*; Ishizeki, Yoshikazu*; Kaneda, Tsutomu*; Sato, Shin*; Shibata, Chihoko*; Ueda, Tadashi*; et al.

JAEA-Technology 2014-019, 495 Pages, 2014/08

JAEA-Technology-2014-019.pdf:82.23MB

The researches on engineering technology in the Mizunami Underground Research Laboratory (MIU) plan consists of (1) research on engineering technology deep underground, and (2) research on engineering technology as a basis of geological disposal. The former research is mainly aimed in this study, which is categorized in (a) development of design and construction planning technologies, (b) development of construction technologies, (c) development of countermeasure technologies, and (d) development of technologies for security. In this study, the researches on engineering technology are being conducted in these four categories by using data measured during construction as a part of the second phase of the MIU plan.

JAEA Reports

Influence evaluation of corrosion on long-term integrity of spent fuel assembly component materials exposed to unusual corrosive environment at unit 1-4 spent fuel pools of Fukushima Dai-ichi Nuclear Power Stations

Fukushima Project Team, Oarai Research and Development Center; Fukushima Fuels and Materials Department, Oarai Research and Development Center

JAEA-Technology 2014-020, 52 Pages, 2014/07

JAEA-Technology-2014-020.pdf:21.21MB

In this study, a screening study on corrosion phenomena and a preliminary investigation for an evaluation method on long-term integrity of FAs experienced unusual corrosive environment were carried out in views of fundamental features of corrosion. The screening study have led to the following two features of FAs from the viewpoint of the integrity as important phenomena to be further investigated; "fission product confinement of cladding tube" and "structural integrity of FA". In terms of "fission product confinement of cladding tube", it was shown experimentally that influence of the exposure to an unusual corrosive environment was low. On the other hand, in terms of "structural integrity of FA", a concept of experimental methodology for predicting long-term corrosion behavior was preliminary studied for preferentially selected FA local parts composed of different metals.

JAEA Reports

Evaluation formulas of manpower needs for dismantling of equipments in uranium refining and conversion plant

Izumo, Sari; Usui, Hideo; Kubota, Shintaro; Tachibana, Mitsuo; Kawagoshi, Hiroshi; Takahashi, Nobuo; Morimoto, Yasuyuki; Tokuyasu, Takashi; Tanaka, Yoshio; Sugitsue, Noritake

JAEA-Technology 2014-021, 79 Pages, 2014/07

JAEA-Technology-2014-021.pdf:22.8MB

Japan Atomic Energy Agency has developed PROject management data evaluation code for DIsmantling Activities (PRODIA) to make an efficient decommissioning for nuclear facilities. PRODIA is a source code which provides estimated value such as manpower needs, costs, etc., for dismantling by evaluation formulas according to the type of nuclear facility. Evaluation formulas of manpower needs for dismantling of equipments about reprocessed uranium conversion in Uranium Refining and Conversion Plant are developed in this report. In the result, 7 formulas for prepare process, 24 formulas for dismantling process and 8 formulas for clean-up process are derived. It is confirmed that an unified evaluation formula can be used instead of 8 formulas about dismantling process of steel equipment for uranium conversion process, and 3 types of simplified formula can be used for preparation process and clean-up process respectively.

JAEA Reports

Evaluation formulas of manpower needs for dismantling of equipment in FUGEN, 3; Dismantling process of the condenser removal

Kubota, Shintaro; Izumo, Sari; Usui, Hideo; Kawagoshi, Hiroshi; Koda, Yuya; Nanko, Takashi

JAEA-Technology 2014-022, 22 Pages, 2014/07

JAEA-Technology-2014-022.pdf:3.5MB

Japan Atomic Energy Agency (JAEA) has been developing the PRODIA code which supports to make decommissioning plan and has been preparing evaluation functions. Manpower needs for the dismantling the condenser that had conducted from 2010 to 2012 was analyzed and compared with existing evaluation functions. Applicability of evaluation function for a large scale reactor facility was confirmed in dismantling of the heat insulating materials and feed water heaters and reliability of unit productivity factor was improved. Evaluation function of work for clearance was made in dismantling of pipes and supports. Statistically meaningful data was provided from the dismantling of the condenser. Manpower needs for dismantling of a condenser has positive correlation to the weight of equipment and can be described in linear expression. Reliability of each unit productivity factor will be improved with accumulating actual dismantling data in future.

JAEA Reports

Installation of the water environment irradiation facility for the IASCC research under the BWR/PWR irradiation environment, 2

Magome, Hirokatsu; Okada, Yuji; Hanawa, Hiroshi; Sakuta, Yoshiyuki; Kanno, Masaru; Iida, Kazuhiro; Ando, Hitoshi; Yonekawa, Akihisa; Ueda, Haruyasu; Shibata, Mitsunobu

JAEA-Technology 2014-023, 267 Pages, 2014/07

JAEA-Technology-2014-023-01.pdf:103.68MB
JAEA-Technology-2014-023-02.pdf:71.92MB

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 was advanced. JMTR stopped after the 165th operation cycle in August 2006, and is advancing renewal of the irradiation facility towards re-operation. The material irradiation test facility was installed from 2008 fiscal year to 2012 fiscal year in JMTR. This report summarizes manufacture and installation of the material irradiation test facility for IASCC research carried out from 2012 to 2014 in the follow-up report reported before (JAEA-Technology 2013-019).

JAEA Reports

Inspection and repair techniques in the reactor vessel of the experimental fast reactor Joyo; Development of repair techniques for UCS replacement of Joyo, 2

Ito, Hiromichi; Takamatsu, Misao; Kawahara, Hirotaka; Nagai, Akinori

JAEA-Technology 2014-024, 28 Pages, 2014/07

JAEA-Technology-2014-024.pdf:17.45MB

Because the gap between the UCS and rotation plug's guide sleeve is 5 mm in minimum, there is a risk of deformation of the UCS and guide sleeve with interference between UCS and guide sleeve in the UCS replacement work. In order to reduce the risk, R&D for below subjects is required.(1) UCS jack-up equipment with strict control of inclination, (2) Detection and escape method for interference between UCS and guide sleeve. In order to solve above (1), the jack-up equipment with applying three-point suspension is developed. Then, in the aspect of above (2), load-measuring devices are installed on three jacks of jack-up equipment. By means of detection eccentric load with interference, deformation of UCS and guide sleeve are prevented. And also, the location of interference can be investigated based on eccentric loads of three jacks. The performance is verified in the ex-vessel mock-up test using full-scale dummy of UCS.

JAEA Reports

Study of origin on tritium release into primary coolant for research and testing reactors; Tritium release rate evaluated from JMTR, JRR-3M and JRR-4 operation data

Ishitsuka, Etsuo; Motohashi, Jun; Hanawa, Yoshio; Komeda, Masao; Watahiki, Shunsuke; Mukanova, A.*; Kenzhina, I. E.*; Chikhray, Y.*

JAEA-Technology 2014-025, 77 Pages, 2014/08

JAEA-Technology-2014-025.pdf:43.46MB

It has been shown that tritium concentration in the primary coolant of the JMTR and JRR-3M increases during its operation. In this report, to clarify the tritium sources, the tritium release rate into the primary coolant in each operation cycle for the JMTR, JRR-3M and JRR-4 was evaluated. As a result, the tritium release rate is $$<$$ 8 Bq/Wd in the JRR-4, which has not the beryllium core components installed, and no increase in the tritium concentration during reactor operation is observed. In contrast, the tritium release rate is about 10$$sim$$95 and 60$$sim$$140 Bq/Wd in the JRR-3M and JMTR respectively, which cores contain beryllium components, and where the tritium content increases while reactor operates. It is also observed that the amount of released tritium is lower in the case of new beryllium components installation, and increases with the reactor operating cycle.

JAEA Reports

Development of experimental method for self-wastage behavior in sodium-water reaction; Development of test rig (SWAT-2R) and study for experimental procedure

Abe, Yuta; Shimoyama, Kazuhito; Kurihara, Akikazu

JAEA-Technology 2014-026, 40 Pages, 2014/07

JAEA-Technology-2014-026.pdf:33.12MB

In case of water leak from a penetrated crack on a tube of steam generator in the sodium cooled fast reactor (SFR), self-wastage, that increases the size of leak, may take place by corrosion related to chemical reaction between sodium and water. For the safety evaluation of the accident, JAEA has been developing the analytical method of self-wastage using the multi-dimensional sodium-water reaction code. This report describes the development of new experimental rig (SWAT-2R). SEAT-2R enables to examine corrosion effecting factors that were ambiguous in the previous studies. The report includes description of development of micro-leak test piece, examination of experimental procedure. The results will provide fundamental data for validation of the self-wastage analytical method.

JAEA Reports

Development of a "scroll pump operation status monitoring system(SCP-MS)" for use at a synchrotron radiation beamline

Yamaoka, Shingo; Shimizu, Yuka*; Fukuda, Yoshihiro*; Shobu, Takahisa; Konishi, Hiroyuki

JAEA-Technology 2014-027, 21 Pages, 2014/08

JAEA-Technology-2014-027.pdf:28.95MB

At SPring-8 synchrotron radiation beamlines, it is essential to maintain a vacuum between the radiation source and the experimental station. This is achieved by using scroll pumps and turbo molecular pumps. However, scroll pump malfunctions have been reported at BL22XU. Since many of the pumps are located inside radiation-shielding hutches, malfunctions often go un-noticed. As a result, operations can continue despite the malfunction. To facilitate the early detection of scroll pump malfunctions, we have developed a "scroll pump operating status monitoring system (SCP-MS)". The system simultaneously measures motor current and vacuum pressure at the scroll pump. It is possible to monitor pumps from outside of the shielding hutch, something which was not possible until now. The (SCP-MS) has been installed to monitor scroll pumps in actual operation, to monitor the change of the motor current value and vacuum pressure. We report on the detail of the system.

JAEA Reports

Application of Cherenkov light observation to reactor measurements, 1; Estimation of reactor power from Cherenkov light intensity

Yamamoto, Keiichi; Takeuchi, Tomoaki; Sano, Tadafumi*; Homma, Ryohei*; Kimura, Nobuaki; Otsuka, Noriaki; Kosuge, Fumiaki*; Nakajima, Ken*; Tsuchiya, Kunihiko

JAEA-Technology 2014-028, 56 Pages, 2015/01

JAEA-Technology-2014-028.pdf:9.23MB

Development of the reactor measurement system was started to obtain real-time in-core nuclear and thermal information, where the quantity measurement of brightness of Cherenkov light was applied. The system would be applied as monitoring system in severe accidents and for advanced operation management technology in existing LWRs. In this report, the calculation and the observation results were summarized about the quantity of the Cherenkov light caused by the $$gamma$$ and $$beta$$ ray emitted from the fuels in the core of Kyoto University Research Reactor.

JAEA Reports

Improvement of the environment for the diffusion experiment using granite samples and results of pore physicality measurement and mineralogical test

Yamashita, Riyo; Hama, Katsuhiro; Takeuchi, Ryuji; Morikawa, Keita*; Hosoya, Shinichi*; Nakamura, Toshiaki*; Tanaka, Yumiko*

JAEA-Technology 2014-029, 118 Pages, 2014/09

JAEA-Technology-2014-029.pdf:25.16MB

This study is to gain a better understanding of mass transfer phenomena in the geological environment as well as to develop technologies for: measurement of the solute transport parameters, model construction, numerical analysis and validation of all those technologies based on the existing information. As part of solute transport study, laboratory experiments were planned to understand the influence of the geological characteristics of fracture on the solute transport parameters, also understand the differences in test results by the different sizes of the samples used for an experiment, and moreover to validate the parameters obtained by numerical analysis.

JAEA Reports

Development on radiation damage calculation method for HTGR in-core structural material

Fukaya, Yuji; Goto, Minoru; Shibata, Taiju

JAEA-Technology 2014-030, 29 Pages, 2014/08

JAEA-Technology-2014-030.pdf:17.7MB

A study on radiation damage calculation method for in-core structural material of HTGR had been performed. Firstly, a theory and a calculation method for radiation damage were investigated. Secondly, a DPA cross-section calculation method using NJOY, which is the typical reactor constants generation code, was established. Moreover, DPA calculation method was established. To evaluate these evaluations simply, calculation method was developed including the function of NPRIM which includes NJOY code as a solver and was developed in previous study. In addition, necessary items were identified to improve the method for accuracy.

JAEA Reports

HTTR demonstration test plan for industrial utilization of nuclear heat

Sato, Hiroyuki; Ohashi, Hirofumi; Yan, X.; Kubo, Shinji; Nishihara, Tetsuo; Tachibana, Yukio; Inagaki, Yoshiyuki

JAEA-Technology 2014-031, 30 Pages, 2014/09

JAEA-Technology-2014-031.pdf:17.95MB

In the present study, identification of test items to be validated in the HTTR demonstration test to accomplish a formulation of safety requirement and design consideration for coupling a hydrogen production plant to a nuclear facility as well as confirmation of overall performance of helium gas turbine system. In addition, a plant concept for the heat utilization system to be connected with the HTTR is clarified.

JAEA Reports

The Periodic safety review report of Tokai Reprocessing Plant

Fukuda, Kazuhito; Tomioka, Kenichiro*; Omori, Satoru; Nakano, Takafumi; Nagasato, Yoshihiko

JAEA-Technology 2014-032, 566 Pages, 2014/11

JAEA-Technology-2014-032.pdf:32.45MB
JAEA-Technology-2014-032(errata).pdf:9.54MB

The Periodic Safety Review of TRP is assessment of the validity of safety activities in order to get assurance for continuous operation by adding effective items to extract and to execute for TRP safety. We performed 4 items; for (1) evaluation of safety activity at TRP, as we confirmed organization was ordered and managed. For (2) evaluation of status of safety activities reflecting the latest knowledge, we confirmed improvement of safety was continued adequately reflecting from the experience for safety. For (3) technical review on aging for the safety related structures, systems and components, we evaluated the guaranty of safety under assumption of continuous maintenance till the next Periodic Safety Review. For (4) establishment of long term maintenance program, we found no additional activities into maintenance programs, however, for several installations we established a plan and utilized them for reliability.

JAEA Reports

Development of annular fuel design code CEPTAR-D; Study on applicability for PCMI stress evaluation

Kamei, Miho; Ozawa, Takayuki

JAEA-Technology 2014-033, 36 Pages, 2014/11

JAEA-Technology-2014-033.pdf:3.93MB

Annular fuel pellet would be available to improve fast reactor fuel performance, and we have developed the "CEPTAR" to apply the annular fuel design taking into account the irradiation behaviors. CEPTAR computes the stress and strain in fuel pellet and cladding by using the generalized plane strain analysis method and the void migration model is applied to compute the fuel restructuring. On the other hand, taking into account the licensability, the fuel restructuring three-region model is applied to the fast reactor fuel design code. In this study, we developed "CEPTAR-D", in which fuel restructuring model of CEPTAR was exchanged into the "fuel restructuring three-region" model, to apply to the fuel design, and verified thermal and mechanical computations by using the results of short-term and long-term irradiation tests. Consequently, the computation accuracy of CEPTAR-D was as well as that of CEPTAR, and it was confirmed that CEPTAR-D could reasonably evaluate the stress due to PCMI.

JAEA Reports

Fabrication technology development and characterization of irradiation targets for $$^{99}$$Mo/$$^{rm 99m}$$Tc production by (n,$$gamma$$) method

Nishikata, Kaori; Kimura, Akihiro; Ishida, Takuya; Shiina, Takayuki*; Ota, Akio*; Tanase, Masakazu*; Tsuchiya, Kunihiko

JAEA-Technology 2014-034, 34 Pages, 2014/10

JAEA-Technology-2014-034.pdf:3.26MB

As a part of utilization expansion after the Japan Material Testing Reactor (JMTR) re-start, research and development (R&D) on the production of medical radioisotope $$^{99}$$Mo/$$^{99m}$$Tc by (n, $$gamma$$) method using JMTR has been carried out in the Neutron Irradiation and Testing Reactor Center of the Japan Atomic Energy Agency. $$^{99}$$Mo is usually produced by fission method. On the other hand, $$^{99}$$Mo/$$^{99m}$$Tc production by the (n, $$gamma$$) method has advantages for radioactive waste, cost reduction and non-proliferation. However, the specific radioactivity per unit volume by the (n, $$gamma$$) method is low compared with the fission method, and that is the weak point of the (n, $$gamma$$) method. This report summarizes the investigation of raw materials, the fabrication tests of high-density MoO$$_{3}$$ pellets by the plasma sintering method for increasing of $$^{98}$$Mo contents and the characterization of sintered high-density MoO$$_{3}$$ pellets.

JAEA Reports

Study on engineering technologies in the Mizunami Underground Research Laboratory (FY 2013); Development of design and construction planning and countermeasure technologies (Contract research)

Kobayashi, Shinji*; Niimi, Katsuyuki*; Okihara, Mitsunobu*; Tsuji, Masakuni*; Yamada, Toshiko*; Sato, Toshinori; Mikake, Shinichiro; Horiuchi, Yasuharu*; Aoyagi, Yoshiaki

JAEA-Technology 2014-035, 172 Pages, 2015/01

JAEA-Technology-2014-035.pdf:91.27MB

The researches on engineering technology in the Mizunami Underground Research Laboratory (MIU) plan consists of (1) development of design and construction planning technologies, (2) development of construction technology, (3) development of countermeasure technology, (4) development of technology for security, and (5) development of technologies regarding restoration or reversal and mitigating of the excavation effect. To develop design and construction planning technologies, and countermeasure technology, the analysis of measured data during earthquake and seismic movement characteristics at deep underground, and the examination of grouting method were carried out. The knowledge of the seismic movements at deep underground was obtained by which observation records of seismometers at Mizunami underground research laboratory were analyzed to verify the earthquake-resistant design of the shafts and tunnels. As for" Study on grouting method at deep underground", Existing post-grouting methods for crystalline rock were reviewed, the applicability of pre-grouting technology was evaluated and study on experiment plan in MIU was carried out following the previous year.

JAEA Reports

Seismically-induced reactor coolant leakage as an allegedly-possible cause of accident at unit 1 of Fukushima Dai-ichi Nuclear Power Station

Kukita, Yutaka; Watanabe, Norio

JAEA-Technology 2014-036, 38 Pages, 2014/11

JAEA-Technology-2014-036.pdf:26.08MB

NAIIC emphasized the possibility of seismically-induced reactor coolant leakage and implied its causal connection to the accident, in particular at the Fukushima Daiichi Unit 1. This view of NAIIC has been addressed by the Accident Investigation Committee established by the Cabinet decision, NISA, and the Secretariat of NRA. Based on seismic response analyses, plant records and simulations, their reports uniformly note that seismically-induced leakage is unlikely to be a causal factor for the core damage though the possibility of insignificantly small leakage cannot be ruled out completely. Also refuted are some of the arguments made by NAIIC as grounds for suspecting safety-significant leakage. The present report re-examines the leak detection capability through the review of plant instruments and post-accident simulations, and adds some arguments in order to resolve the issue raised by NAIIC without technical ambiguity as far as possible. As well, the plant design uniqueness of Unit 1, the history of facility changes, the operating procedures and the actual operations are looked into to raise issues for further investigation.

JAEA Reports

Measurement of pure water resistivity in a high temperature region

Yamanaka, Haruhiko; Maejima, Tetsuya; Terunuma, Yuto; Watanabe, Kazuhiro; Kashiwagi, Mieko; Hanada, Masaya

JAEA-Technology 2014-037, 12 Pages, 2014/12

JAEA-Technology-2014-037.pdf:4.27MB

Resistivity of a high temperature pure water has been measured up to 180$$^{circ}$$C which is the maximum water temperature in the ITER Neutral Beam Injector. The resistivity of the pure water is decreased by increasing the water temperature. It was found that even different resistivity water of 9 M$$Omega$$cm and 5 M$$Omega$$cm showed almost the same resistivity at the higher temperature region of 100$$^{circ}$$C. The resistivity of 0.36 M$$Omega$$cm was measured at the temperature of 180$$^{circ}$$C. This resistivity agreed well to the calculated value for the theoretical pure water.

JAEA Reports

Verification and validation of the THYTAN code for the graphite oxidation analysis in the HTGR systems

Shimazaki, Yosuke; Isaka, Kazuyoshi; Nomoto, Yasunobu; Seki, Tomokazu; Ohashi, Hirofumi

JAEA-Technology 2014-038, 51 Pages, 2014/12

JAEA-Technology-2014-038.pdf:3.84MB

The analytical models for the evaluation of graphite oxidation were implemented into the THYTAN code, which employs the mass balance and a node-link computational scheme to evaluate tritium behavior in the High Temperature Gas-cooled Reactor (HTGR) systems for hydrogen production, to analyze the graphite oxidation during the air or water ingress accidents in the HTGR systems. This report describes the analytical models of the THYTAN code in terms of the graphite oxidation analysis and its verification and validation (V&V) results. Mass transfer from the gas mixture in the coolant channel to the graphite surface, diffusion in the graphite, graphite oxidation by air or water, chemical reaction and release from the primary circuit to the containment vessel by a safety valve were modeled to calculate the mass balance in the graphite and the gas mixture in the coolant channel. The computed solutions using the THYTAN code for simple questions were compared to the analytical results by a hand calculation to verify the algorithms for each implemented analytical model. A representation of the graphite oxidation experimental was analyzed using the THYTAN code, and the results were compared to the experimental data and the computed solutions using the GRACE code, which was used for the safety analysis of the High Temperature Engineering Test Reactor (HTTR), in regard to corrosion depth of graphite and oxygen concentration at the outlet of the test section to validate the analytical models of the THYTAN code. The comparison of THYTAN code results with the analytical solutions, experimental data and the GRACE code results showed the good agreement.

JAEA Reports

Development of refilling techniques of LA-type bituminized waste products

Irisawa, Keita; Komatsuzaki, Toshio; Kawato, Yoshimi; Sakakibara, Tetsuro; Nakazawa, Osamu; Meguro, Yoshihiro

JAEA-Technology 2014-039, 28 Pages, 2014/12

JAEA-Technology-2014-039.pdf:6.13MB

In JAEA, 13,296 drums of low-radioactivity bituminized waste products (BWPs) have been stored in asphalt solidification storages. In order to effectively utilize the space of the BWP in a repository site, we studied refilling techniques of the BWP from the drum to a box-shaped container. Tentative processes, which we devised, consisted of (1) take-off of BWP from the drum, (2) separation of a post filling part from BWP and (3) filling of BWP to a box-shaped container. Two methods for each process were selected, and work efficiencies of the methods were investigated by using a synthetic BWP.

JAEA Reports

Study on engineering technologies in the Mizunami Underground Research Laboratory (FY 2013); Development of recovery and mitigation technology on excavation damage (Contract research)

Fukaya, Masaaki*; Hata, Koji*; Akiyoshi, Kenji*; Sato, Shin*; Takeda, Yoshinori*; Miura, Norihiko*; Uyama, Masao*; Kaneda, Tsutomu*; Ueda, Tadashi*; Toda, Akiko*; et al.

JAEA-Technology 2014-040, 199 Pages, 2015/03

JAEA-Technology-2014-040.pdf:37.2MB

The researches on engineering technology in the Mizunami Underground Research Laboratory (MIU) project consists of (1) development of design and construction planning technologies, (2) development of construction technology, (3) development of countermeasure technology, (4) development of technology for security, and (5) development of technologies for restoration and/or reduction of the excavation damage. The researches on engineering technology such as verification of the initial design were being conducted by using data measured during construction as a part of the second phase of the MIU plan. Examination about the plug for reflood test in the GL-500m Access/Research Gallery-North as part of the development of technologies for restoration and/or reduction of excavation damage were carried out. Specifically, Literature survey was carried out about the plug, based on the result of literature survey, examination of the design condition, design of the plug and rock stability using numerical simulation, selection of materials for major parts, and grouting for water inflow from between rock and plug, were carried out in this study.

JAEA Reports

Development of the control system with versatile PLCs for the long-pulse negative ion source

Komata, Masao; Shimizu, Tatsuo; Ozeki, Masahiro; Kojima, Atsushi; Hanada, Masaya

JAEA-Technology 2014-041, 50 Pages, 2015/01

JAEA-Technology-2014-041.pdf:22.68MB

In JT-60 Super Advanced, the machine for nuclear fusion research with superconducting magnets for long pulse operation, the negative-ion-base neutral beam injector is required to extend the pulse duration time 10 s to 100 s. In order to realize the long-pulse N-NB injector, the control system of the power supplies for the negative ion source has been newly developed. The control system with use of the versatile devices such as PLC was designed for an ease extension of the functions. Since the control system should have the many different functions which require the wide range of the sampling time of 1 milli-second to 10, all of the functions are performed by distributing PLCs for each of the function. The developed control system has been applied in the tests of the JT-60 negative ion source, where a 100 s negative ion beam has been successfully produced. Through this test, the controllability of this system has been confirmed to be feasible for JT-60SA operation.

JAEA Reports

Disassembly of the NBI system on JT-60U for JT-60 SA

Akino, Noboru; Endo, Yasuei; Hanada, Masaya; Kawai, Mikito*; Kazawa, Minoru; Kikuchi, Katsumi*; Kojima, Atsushi; Komata, Masao; Mogaki, Kazuhiko; Nemoto, Shuji; et al.

JAEA-Technology 2014-042, 73 Pages, 2015/02

JAEA-Technology-2014-042.pdf:15.1MB

According to the project plan of JT-60 Super Advanced that is implemented as an international project between Japan and Europe, the neutral beam (NB) injectors have been disassembled. The disassembly of the NB injectors started in November, 2009 and finished in January, 2012 without any serious problems as scheduled. This reports the disassembly activities of the NB injectors.

JAEA Reports

Development of three-dimensional diffusion and burn-up code HIZER for Monju core management

Kato, Shinya; Shimomoto, Yoshihiko; Kato, Yuko; Kitano, Akihiro

JAEA-Technology 2014-043, 36 Pages, 2015/02

JAEA-Technology-2014-043.pdf:8.94MB

The core management and operation code system aims to perform core management task efficiently by systematic management of data, analyses and edits, which are needed in the reactor core management and operation. The system consists of the five calculation modules: the reactor constant generation module, the neutronic-thermal calculation module, the radiation analysis module, the core structural integrity estimation module, and the core operation analysis module. In these modules, the neutronic-thermal calculation module is based on the dedicated three-dimensional diffusion and burn-up code HIZER. HIZER can execute core calculations easily for specific design specification and operation patterns of Monju, enabling efficient and accurate evaluation of the Monju core characteristics. This report describes its calculation method and validation results.

JAEA Reports

Conceptual study of transmutation experimental facility, 5; Investigation of MA fuel handling

Sugawara, Takanori; Nishihara, Kenji; Sasa, Toshinobu; Tsujimoto, Kazufumi; Tazawa, Yujiro; Oigawa, Hiroyuki

JAEA-Technology 2014-044, 59 Pages, 2015/03

JAEA-Technology-2014-044.pdf:14.46MB

Transmutation Physics Experimental Facility (TEF-P) planned in the J-PARC is a critical assembly with low thermal output and it will treat large amount of highly-radioactive minor actinide (MA) fuels in the experiments. Handling of the MA fuels in each stage of storage, transport and loading/unloading to the core was conceptually investigated, then, criticality, dose and cooling performance were assessed. For the criticality, it was shown that the effective multiplication factors in each step, storage, transport and loading, were sufficiently lower than 1.0. For the dose, the dose for workers will be reduced by installing remote handling devices to treat the MA fuels. For the cooling performance, it was found that the temperature of the core would be kept low in the normal operation. On the other hand, in the case which the air conditioning or the blower for the core stopped for long period, it was shown that there would be a possibility of the MA fuel failure.

JAEA Reports

Performance confirmation of MONJU failed fuel detection and location system, 1

Morohashi, Yuko; Suzuki, Satoshi

JAEA-Technology 2014-045, 116 Pages, 2015/03

JAEA-Technology-2014-045.pdf:33.37MB

The failed fuel detection and location (FFDL) system collects the tagging gas that migrates into the reactor cover gas from a failed pin. The tagging gas is made of stable isotopes of Kr and Xe. The isotopic composition of the tagging gas can be made specific to each assembly. The assembly containing a failed fuel pin can be identified by analyzing the isotopic composition. The FFDL system is comprised of two tagging gas concentration devices. The concentration rate is designed to be higher than 200. Past examinations demonstrated that the concentration rate meets the requirement with a noble gas concentration of 1ppm. However, the actual noble gas concentration emitted from a failed fuel is assumed to be much lower. In the present study, the performance of FFDL system was investigated by measuring low concentration gas of the actual fuel failure level. As a result, the concentration rate was confirmed to be more than tens of thousands, which sufficiently satisfies the design demand.

JAEA Reports

User's guide of cement solidification test for incinerated ash

Nakayama, Takuya; Kawato, Yoshimi; Osugi, Takeshi; Shimazaki, Takejiro; Hanada, Keiji; Suzuki, Shinji; Sakakibara, Tetsuro; Nakazawa, Osamu; Meguro, Yoshihiro

JAEA-Technology 2014-046, 56 Pages, 2015/03

JAEA-Technology-2014-046.pdf:7.61MB

The combustible and flame-retardant radioactive wastes generated as a result of the research activities in Japan Atomic Energy Agency (JAEA) are incinerating to reduce their volume. The incinerated ash is planned to be solidified using cement for disposal. Since the properties of ashes generated in each institute of JAEA are varied with the type of incinerator and the wastes to be incinerated, it is necessary to do fundamental solidification tests in each institute to decide operating conditions of the planning cement solidification facility. It is important to standardize evaluating methods of cement and solidified waste because some characters depend on measuring method. This user's guide have been prepared how to decide the cement solidifying conditions of ash to design the cement solidification facility in JAEA. Requirements on the regulations of solidified radioactive waste have been examined and seven technical criteria, e.g. compressive strength, fluidity, have been selected as characters to be evaluated. Some empirical notes about selection of cement, admixtures, procedure on making a test piece, evaluation of expanding, compressive strength, solubility have been described. The strategy of tests and tips for finding optimized solidification condition has been summarized. Finally the example of optimized conditions satisfied the requirements and some problems to be solved have been described.

JAEA Reports

Mock-up test of the modified STACY (Accuracy verification of water-feed- stop switch detector)

Seki, Masakazu; Izawa, Kazuhiko; Sono, Hiroki

JAEA-Technology 2014-047, 22 Pages, 2015/03

JAEA-Technology-2014-047.pdf:2.33MB

The Japan Atomic Energy Agency is conducting a reactor modification project of the Static Experiment Critical Facility (STACY). In the modification, STACY is to be converted from a thermal reactor using solution fuel into that using fuel rods and light water moderator. Reactivity of the modified STACY is controlled by the water level fed in the core tank as well as the present STACY. Regarding water level detection, however, a float-type water-feed-stop switch is adopted in the modified STACY because the electro-conductivity-type switch of the present STACY for uranyl nitrate solution cannot detect demineralized water used in the modified STACY. For safety operation of the modified STACY, the float-type switch needs accurate and reliable detection of water level at any temperature. This report describes a mock-up test on accuracy verification of the float-type water-feed-stop switch in whole range of water temperature (room temperature $$sim$$70$$^{circ}$$C) in the modified STACY operation.

JAEA Reports

Environmental impact of nitrate nitrogen in sub-surface disposal system on surface water and sensitivity analysis of distribution coefficient of nitrate ion and porosity of waste layer on the environmental safety assessments

Nakamura, Yasuo; Nakatani, Takayoshi

JAEA-Technology 2014-048, 18 Pages, 2015/03

JAEA-Technology-2014-048.pdf:7.75MB

Sodium nitrate is included bituminized waste generating from the reprocessing plant of spent fuel which is disposed of in sub-surface disposal facility. Because the sodium nitrate is soluble material in surface water, it is a concern impact on surface water. Such as non-radioactive materials are not strictly regulated by "the Law for the Regulations of Nuclear Source Material, Nuclear Fuel Material and Reactors", but should be considered by related laws and regulations according to former basic policy. Because it is regulated as nitrate nitrogen by "The Basic Environment Law", the valuation of the environmental impact on general sub-surface disposal system was carried out. As the results, the concentration of nitrate nitrogen in river water whose annual quantity of water is rather than 1$$times$$10$$^{5}$$m$$^{3}$$/y is below the regulated value at the small scale surface waters as evaluation point.

JAEA Reports

Characteristics of OSL dosimeter as individual monitoring for external radiation

Suzuki, Akifumi; Suzuki, Takehiko; Takahashi, Masa; Nakata, Toru; Murayama, Takashi; Tsunoda, Masahiko

JAEA-Technology 2014-049, 19 Pages, 2015/03

JAEA-Technology-2014-049.pdf:9.12MB

Optically Stimulated Luminescence, OSL, dosimeters have been used as individual dosimeters for external radiation in Nuclear Science Research Institute and so on since October, 2014 as successor of the RPL glass dosimeters. Characteristics of the OSL dosimeters such as dose linearity, energy response, angular dependence, fading characteristics and responses at mixed irradiation fields were examined prior to the start of use. As a result, it was found that the OSL dosimeters met the performances that the national standard (JIS Z 4339) determined. The characteristics of OSL dosimeters were comparable with those of the RPL glass dosimeters. In conclusion, it was confirmed the OSL dosimeters had sufficient performances for the practical use on individual monitoring. This report shows the testing methods and the results for the characteristics of OSL dosimeters.

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