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JAEA Reports

Technical design of the pressure-resistant chamber for open inspections of the storage containers of nuclear fuel materials

Marufuji, Takato; Sato, Takumi; Ito, Hideaki; Suzuki, Hisashi; Fujishima, Tadatsune; Nakano, Tomoyuki

JAEA-Technology 2019-006, 22 Pages, 2019/05

JAEA-Technology-2019-006.pdf:2.84MB

Radioactive contamination incident occurred at Plutonium Fuel Research Facility (PFRF) in Oarai Research and Development Institute, Japan Atomic Energy Agency on June 6, 2017. During inspection work of storage container containing nuclear fuel materials, the PVC bag packaging in the storage container ruptured when a worker opened the lid in the hood, and a part of contents was spattered over the room. The cause of the increase of internal pressure of the storage container was gas generation by alpha radiolysis of the epoxy resin mixed with nuclear fuel materials. Opening inspection of about 70 similar containers stored in PFRF has been planned to confirm the condition of the contents and to stabilize the stored materials containing organic compounds. For safe and reliable open inspection of the storage containers with high internal pressure in the glove box, it is necessary to develop a pressure-resistant chamber in which the storage containers are opened and the contents are inspected under gastight condition. This report summarizes the concerns and countermeasures of the chamber design and the design results of the chamber.

Journal Articles

Effect of re-oxidation rate of additive cations on corrosion rate of stainless steel in boiling nitric acid solution

Irisawa, Eriko; Yamamoto, Masahiro; Kato, Chiaki; Motooka, Takafumi; Ban, Yasutoshi

Journal of Nuclear Science and Technology, 56(4), p.337 - 344, 2019/04

 Percentile:100(Nuclear Science & Technology)

JAEA Reports

Activity median aerodynamic diameter relating to contamination at Plutonium Fuel Research Facility in Oarai Research and Development Center; Particle size analysis for plutonium particles using imaging plate

Takasaki, Koji; Yasumune, Takashi; Hashimoto, Makoto; Maeda, Koji; Kato, Masato; Yoshizawa, Michio; Momose, Takumaro

JAEA-Review 2019-003, 48 Pages, 2019/03

JAEA-Review-2019-003.pdf:3.81MB

June 6, 2017, at Plutonium Fuel Research Facility in Oarai Research and Development Center of JAEA, when five workers were inspecting storage containers containing plutonium and uranium, resin bags in a storage container ruptured, and radioactive dust spread. Though they were wearing a half face mask respirator, they inhaled radioactive materials. In the evaluation of the internal exposure dose, the aerodynamic radioactive median diameter (AMAD) is an important parameter. We measured 14 smear samples and a dust filter paper with imaging plates, and estimated the AMAD by image analysis. As a result of estimating the AMAD, from the 14 smear samples, the AMADs are 4.3 to 11 $$mu$$m or more in the case of nitrate plutonium, and the AMADs are 5.6 to 14 $$mu$$m or more in the case of the oxidized plutonium. Also, from the dust filter paper, the AMAD is 3.0 $$mu$$m or more in the case of nitrate plutonium, and the AMAD is 3.9 $$mu$$m or more in the case of the oxidized plutonium.

JAEA Reports

Restoration activity of the contamination accident at plutonium fuel research facility

Restoration Activity Team for the PFRF Contamination Incident

JAEA-Review 2019-001, 58 Pages, 2019/03

JAEA-Review-2019-001.pdf:10.74MB

The contamination accident occurred in a laboratory room (Room No.108) of Plutonium Fuel Research Facility (PFRF) in Japan Atomic Energy Agency (JAEA), Oarai Research and Development Institute on June 6, 2017. The polyvinyl chloride (PVC) bags burst just after the lid of one storage container was opened during the inspection of storage containers for U and Pu in the ventilation hood. At that time, part of nuclear materials in the storage container were scattered all over the room. Five workers in the room were subjected to plutonium contamination, which resulted in internal exposure. In order to restore the Room No.108 of PFRF, the Restoration Activity Team organized in JAEA carried out the decontamination work after the investigation of the contamination level in the room. The team decontaminated the surface of walls, ceiling, gloveboxes and other experimental instruments. Depending on the contamination distribution and installation state of the instruments, suitable decontamination methods were selected. In addition to the manual wiping using wet clothes, the exfoliation method using a strippable paint was applied for constricted areas. As a result, the loose alpha-contamination level fell below the detection limit throughout the room. On the other hand, the fixed contamination was covered with plastic sheets after the decontamination by a strippable paint. We hope that the restoration activity described in this report will provide useful information for the management of decommissioning facilities, especially for facilities treating alpha-radioactive materials such as plutonium.

Journal Articles

Study on Pu-burner high temperature gas-cooled reactor in Japan; Design study of fuel and reactor core

Goto, Minoru; Aihara, Jun; Inaba, Yoshitomo; Ueta, Shohei; Fukaya, Yuji; Okamoto, Koji*

Proceedings of 9th International Topical Meeting on High Temperature Reactor Technology (HTR 2018) (USB Flash Drive), 6 Pages, 2018/10

JAEA has conducted design studies of a Pu-burner HTGR. The Pu-burner HTGR incinerates Pu by fission, and hence a high burn-up is required for the efficient incineration. In the fuel design, a thin ZrC layer, which acts as an oxygen getter and suppresses the internal pressure, was coated on the fuel kernel to prevent the CFP failure at the high burn-up. A stress analysis of the SiC layer, which acts as a pressure vessel for the CFP, was performed for with consideration of the depression effect due to the ZrC layer. As a result, the CFP failure fraction at high burn-up of 500 GWd/t satisfied the target value. In the reactor core design, an axial fuel shuffling was employed to attain the high burn-up, and the nuclear burn-up calculations with the whole core model and the fuel temperature calculations were performed. As a result, the nuclear characteristics, which are the shutdown margin and the temperature coefficient of reactivity, and the fuel temperature satisfied their target values.

Journal Articles

Demonstration of $$gamma$$-ray pipe-monitoring capabilities for real-time process monitoring safeguards applications in reprocessing facilities

Rodriguez, D.; Tanigawa, Masafumi; Nishimura, Kazuaki; Mukai, Yasunobu; Nakamura, Hironobu; Kurita, Tsutomu; Takamine, Jun; Suzuki, Satoshi*; Sekine, Megumi; Rossi, F.; et al.

Journal of Nuclear Science and Technology, 55(7), p.792 - 804, 2018/07

 Percentile:100(Nuclear Science & Technology)

Nuclear material in reprocessing facilities is safeguarded by random sample verification with additional continuous monitoring applied to solution masses and volume in important tanks to maintain continuity-of-knowledge of process operation. Measuring the unique $$gamma$$ rays of each solution as the material flows through pipes connecting all tanks and process apparatuses could potentially improve process monitoring by verifying the compositions in real time. We tested this $$gamma$$ ray pipe-monitoring method using plutonium-nitrate solution transferred between tanks at the PCDF-TRP. The $$gamma$$ rays were measured using a lanthanum-bromide detector with a list-mode data acquisition system to obtain both time and energy of $$gamma$$-ray. The analysis and results of this measurement demonstrate an ability to determine isotopic composition, process timing, flow rate, and volume of solution flowing through pipes, introducing a viable capability for process monitoring safeguards verification.

Journal Articles

Development of security and safety fuel for Pu-burner HTGR, 2; Design study of fuel and reactor core

Goto, Minoru; Ueta, Shohei; Aihara, Jun; Inaba, Yoshitomo; Fukaya, Yuji; Tachibana, Yukio; Okamoto, Koji*

Proceedings of 25th International Conference on Nuclear Engineering (ICONE-25) (CD-ROM), 6 Pages, 2017/07

A PuO$$_{2}$$-YSZ fuel kernel with a ZrC coating, which enhances safety, security and safeguard, namely: 3S-TRISO fuel, was proposed to introduce to the plutonium-burner HTGR. In this study, the efficiency of the ZrC coating as the free-oxygen getter was examined based on a thermochemical calculation. A preliminary study on the feasibility of the 3S-TRISO fuel was conducted focusing on the internal pressure. Additionally, a nuclear feasibility of the reactor core was studied. As a result, all the amount of the free-oxygen is captured by a thin ZrC coating under 1600$$^{circ}$$C and coating ZrC on the fuel kernel should be very effective method to suppress the internal pressure. The internal pressure of the 3S-TRISO fuel at 500 GWd/t is lower than that of UO$$_{2}$$ kernel TRISO fuel whose feasibility had been already confirmed and the 3S-TRISO fuel should be feasible. The fuel shuffling allows to achieve 500 GWd/t. The temperature coefficient of reactivity is negative during the operation period and thus the nuclear feasibility of the reactor core should be achievable.

Journal Articles

Completion of solidification and stabilization for Pu nitrate solution to reduce potential risks at Tokai Reprocessing Plant

Mukai, Yasunobu; Nakamichi, Hideo; Kobayashi, Daisuke; Nishimura, Kazuaki; Fujisaku, Sakae; Tanaka, Hideki; Isomae, Hidemi; Nakamura, Hironobu; Kurita, Tsutomu; Iida, Masayoshi*; et al.

Proceedings of 2017 International Congress on Advances in Nuclear Power Plants (ICAPP 2017) (CD-ROM), 8 Pages, 2017/04

TRP has stored the plutonium in solution state for long-term since the last PCDF operation in 2007 was finished. After the great east Japan earthquake in 2011, JAEA had investigated the risk against potential hazard of these solutions which might lead to make hydrogen explosion and/or boiling of the solution accidents with the release of radioactive materials to the public when blackout. To reduce the risk for storing Pu solution (about 640 kg Pu), JAEA planned to perform the process operation for the solidification and stabilization of the solution by converted into MOX powder at PCDF in 2013. In order to perform PCDF operation without adaption of new safety regulation, JAEA conducted several safety measures such as emergency safety countermeasures, necessary security and safeguards (3S) measures with understanding of NRA. As a result, the PCDF operation had stared on 28th April, 2014, and successfully completed to convert MOX powder on 3rd August, 2016 for about 2 years as planned.

JAEA Reports

Stabilization of MOX dissolving solution at STACY

Kobayashi, Fuyumi; Sumiya, Masato; Kida, Takashi; Kokusen, Junya; Uchida, Shoji; Kaminaga, Jota; Oki, Keiichi; Fukaya, Hiroyuki; Sono, Hiroki

JAEA-Technology 2016-025, 42 Pages, 2016/11

JAEA-Technology-2016-025.pdf:17.88MB

A preliminary test on MOX fuel dissolution for the STACY critical experiments had been conducted in 2000 through 2003 at Nuclear Science Research Institute of JAEA. Accordingly, the uranyl / plutonium nitrate solution should be reconverted into oxide powder to store the fuel for a long period. For this storage, the moisture content in the oxide powder should be controlled from the viewpoint of criticality safety. The stabilization of uranium / plutonium solution was carried out under a precipitation process using ammonia or oxalic acid solution, and a calcination process using a sintering furnace. As a result of the stabilization operation, recovery rate was 95.6% for uranium and 95.0% for plutonium. Further, the recovered oxide powder was calcined again in nitrogen atmosphere and sealed immediately with a plastic bag to keep its moisture content low and to prevent from reabsorbing atmospheric moisture.

Journal Articles

Thermophysical properties of americium-containing barium plutonate

Tanaka, Kosuke; Sato, Isamu; Hirosawa, Takashi; Kurosaki, Ken*; Muta, Hiroaki*; Yamanaka, Shinsuke*

Journal of Nuclear Science and Technology, 52(10), p.1285 - 1289, 2015/10

 Times Cited Count:2 Percentile:66.76(Nuclear Science & Technology)

Polycrystalline specimens of americium-containing barium plutonate have been prepared by mixing the appropriate amounts of (Pu$$_{0.91}$$Am$$_{0.09}$$)O$$_{2}$$ and BaCO$$_{3}$$ powders followed by reacting and sintering at 1600 K under the flowing gas atmosphere of dry-air. The sintered specimens had a single phase of orthorhombic perovskite structure and were crack-free. The elastic moduli were determined from the longitudinal and shear sound velocities. The Debye temperature was also determined from the sound velocities and lattice parameter measurements. The thermal conductivity was calculated from the measured density at room temperature, literature values of heat capacity, and thermal diffusivity measured by laser flash method in vacuum. The thermal conductivity of americium-containing barium plutonate was roughly independent of the temperature and was almost the same magnitude as that of BaPuO$$_{3}$$ and BaUO$$_{3}$$.

Journal Articles

Conceptual study of a plutonium burner high temperature gas-cooled reactor with high nuclear proliferation resistance

Goto, Minoru; Demachi, Kazuyuki*; Ueta, Shohei; Nakano, Masaaki*; Honda, Masaki*; Tachibana, Yukio; Inaba, Yoshitomo; Aihara, Jun; Fukaya, Yuji; Tsuji, Nobumasa*; et al.

Proceedings of 21st International Conference & Exhibition; Nuclear Fuel Cycle for a Low-Carbon Future (GLOBAL 2015) (USB Flash Drive), p.507 - 513, 2015/09

A concept of a plutonium burner HTGR named as Clean Burn, which has a high nuclear proliferation resistance, had been proposed by Japan Atomic Energy Agency. In addition to the high nuclear proliferation resistance, in order to enhance the safety, we propose to introduce PuO$$_{2}$$-YSZ TRISO fuel with ZrC coating to the Clean Burn. In this study, we conduct fabrication tests aiming to establish the basic technologies for fabrication of PuO$$_{2}$$-YSZ TRISO fuel with ZrC coating. Additionally, we conduct a quantitative evaluation of the security for the safety, a design of the fuel and the reactor core, and a safety evaluation for the Clean Burn to confirm the feasibility. This study is conducted by The University of Tokyo, Japan Atomic Energy Agency, Fuji Electric Co., Ltd., and Nuclear Fuel Industries, Ltd. It was started in FY2014 and will be completed in FY2017, and the first year of the implementation was on schedule.

JAEA Reports

Essentials of neutron multiplicity counting mathematics; An Example of U-Pu mixed dioxide

Hosoma, Takashi

JAEA-Research 2015-009, 162 Pages, 2015/08

JAEA-Research-2015-009.pdf:22.3MB

Neutron coincidence counting assay systems have been developed in the last two decades. Objects would extend to high-mass uranium-plutonium dioxide containing other spontaneous fission nuclei, so essentials of neutron multiplicity counting were reconsidered and expanded: (a) Formulae of multiplicity distribution were algebraically derived up to septuplet using a probability generating function; (b) Leakage multiplication was evaluated not by Monte Carlo method but by an average length from an arbitrary point inside a sample to an arbitrary point on its surface and a probability of induced fission within the length; (c) Mechanism of coincidence counting was associated with a couple of different time axes in Poisson process, and consequently a pair of close-to-coincident neutrons from the process was derived. For the formulae, new expressions using combination were wrote down. For spectrum and mean free path, actually treated uranium-plutonium dioxide was selected as an example.

Journal Articles

Introduction to development of advanced safeguards and security NDA technologies by JAEA-ISCN

Seya, Michio; Kureta, Masatoshi; Soyama, Kazuhiko; Nakamura, Hironobu; Harada, Hideo; Hajima, Ryoichi

Proceedings of INMM 55th Annual Meeting (Internet), 10 Pages, 2014/07

JAEA has been implementing development programs of basic technologies of the following advanced NDA (non-destructive assay) of nuclear material (NM) for nuclear safeguards and security. (1) Alternative to $$^{3}$$He neutron detection using ZnS/B$$_{2}$$O$$_{3}$$ ceramic scintillator, (2) NRD (neutron resonance densitometry) using NRTA (neutron resonance transmission analysis) and NRCA (neutron resonance capture analysis), (3) NRF (nuclear resonance fluorescence)-NDA using laser Compton scattered (LCS) $$gamma$$-rays (intense mono-energetic $$gamma$$-rays). The development program (1) is for NDA systems that use ZnS/B$$_{2}$$O$$_{3}$$ ceramic scintillator as alternative neutron detector to $$^{3}$$He for coming shortage of its supply. The program (2) is for a NDA system of isotopic composition measurement (non-destructive mass spectroscopy) in targets such as particle-like melted fuel debris using NRTA and NRCA. The program (3) is for NDA systems using a specific NRF reaction of certain Pu/U isotope caused by mono-energetic LCS $$gamma$$-ray with energy tuned to the specific excited state of the isotope. This paper introduces above three programs.

Journal Articles

Neutron-sensitive ZnS/$$^{10}$$B$$_{2}$$O$$_{3}$$ ceramic scintillator detector as an alternative to a $$^{3}$$He-gas-based detector for a plutonium canister assay system

Nakamura, Tatsuya; Ozu, Akira; To, Kentaro; Sakasai, Kaoru; Suzuki, Hiroyuki; Honda, Katsunori; Birumachi, Atsushi; Ebine, Masumi; Yamagishi, Hideshi*; Takase, Misao; et al.

Nuclear Instruments and Methods in Physics Research A, 763, p.340 - 346, 2014/05

 Times Cited Count:2 Percentile:72.96(Instruments & Instrumentation)

A neutron-sensitive ZnS/$$^{10}$$B$$_{2}$$O$$_{3}$$ ceramic scintillator detector was developed as an alternative to a $$^{3}$$He-gas-based detector for use in a plutonium canister assay system. The detector has a modular structure, with a flat ZnS/$$^{10}$$B$$_{2}$$O$$_{3}$$ceramic scintillator strip that is installed diagonally inside a light-reflecting aluminium case with a square cross section. The prototype detectors, which have a neutron-sensitive area of 30 mm $$times$$ 250 mm, exhibited a sensitivity of 21.7-23.4 $$pm$$ 0.1 cps$$/$$nv for thermal neutrons, a $$^{137}$$Cs $$gamma$$-ray sensitivity of 1.1-1.9 $$pm $$0.2 $$times$$ 10$$^{-7}$$ and a count variation of less than 6% over the detector length. A trial experiment revealed a temperature coefficient of less than -0.24$$pm$$ 0.05% / $$^{circ}$$C over the temperature range of 20-50$$^{circ}$$C.

Journal Articles

Vaporization behaviour of Pu-Cd intermetallic compounds

Nakajima, Kunihisa; Nakazono, Yoshihisa; Arai, Yasuo

Recent Advances in Actinide Science, p.448 - 450, 2006/06

A pyrochemical process for the metallurgical treatment of the spent nuclear fuels contains the Cd-distillation step, where Pu is recovered as a residue. For understanding this Cd-ditillation behavior the samples of PuCd$$_{2}$$+PuCd$$_{4}$$ and PuCd$$_{4}$$+PuCd$$_{6}$$ were prepared and Knudsen-cell mass-spectrometric measurements were carried out to determine the vapour pressures of Cd(g) over these samples. Further,the thermodynamic quantities of PuCd$$_{2}$$ and PuCd$$_{4}$$ were evaluated from these vapour pressures.

Journal Articles

VEGA; An Experimental study of radionuclides release from fuel under severe accident conditions

Kudo, Tamotsu; Hidaka, Akihide*; Fuketa, Toyoshi

Proceedings of 2005 Water Reactor Fuel Performance Meeting (CD-ROM), p.883 - 889, 2005/10

The VEGA program have been performed at Japan Atomic Energy Research Institute (JAERI). The program was comprised of series of experiments on radionuclides release from fuel under severe accident conditions and post-test evaluation with numerical calculations. Effects on the release of ambient pressure, fuel temperature, inert or steam environment and MOX-effect were studied in the program. These effects had been hardly investigated in previous studies due to difficulties in experiments with high temperature and pressure conditions. Release of cesium was mitigated at elevated pressure in comparison with atmospheric pressure. Cesium release was enhanced in the temperature region where fuel foaming occurred below the melting point of UO$$_{2}$$. Release of cesium and ruthenium under steam condition was greater than that under the inert helium condition. Released mass of plutonium above 2800 K was higher by nearly three orders of magnitude than that in lower temperature than 2800 K.

Journal Articles

Development of a simple reprocessing process using selective precipitant for uranyl ions; Precipitation behaviors of plutonium and other transuranium elements

Morita, Yasuji; Kawata, Yoshihisa*; Mineo, Hideaki; Koshino, Nobuyoshi*; Asanuma, Noriko*; Ikeda, Yasuhisa*; Yamasaki, Kazuhiko*; Chikazawa, Takahiro*; Tamaki, Yoshihisa*; Kikuchi, Toshiaki*

Proceedings of International Conference on Nuclear Energy System for Future Generation and Global Sustainability (GLOBAL 2005) (CD-ROM), 6 Pages, 2005/10

N-cyclohexyl-2-pyrrolidone (NCP) can selectively precipitate U(VI) ions in aqueous nitric acid solutions. Utilizing this property, we have been developing a simple reprocessing process of spent nuclear fuel based only on precipitation method. In the first precipitation step, only U is separated by precipitation in a yield of about 70%, and in the second precipitation step both U and Pu are recovered and separated from fission products (FP) and other transuranium elements (TRU). In JAERI, precipitation behaviors of Pu and other TRU were examined experimentally, and the results showed the feasibility of the process establishement.

Journal Articles

Separation of Np from U and Pu using a salt-free reductant for Np(VI) by continuous counter-current back-extraction

Ban, Yasutoshi; Asakura, Toshihide; Morita, Yasuji

Proceedings of International Conference on Nuclear Energy System for Future Generation and Global Sustainability (GLOBAL 2005) (CD-ROM), 5 Pages, 2005/10

An idea for controlling Np behavior in the Purex process is that Np(VI) extracted by TBP is selectively reduced to Np(V) by salt-free reagents and separated from U and Pu. Allylhydrazine is expected as a selective Np(VI) reductant from a view point of reduction rates for Np(VI) and Pu(IV). To confirm the applicability of allylhydrazine, a continuous counter-current back-extraction test of Np(VI) has been carried out using a miniature mixer-settler that consists of two steps: U-Pu recovery (3 stages) and Np separation (4 stages). Experimental results show that at least 90% of Np in feed are back-extracted and separated from U and Pu, therefore, it is confirmed that allylhydrazine is expected to be a selective salt-free reductant of Np(VI).

Journal Articles

R&D on safeguards environmental sample analysis at JAERI

Sakurai, Satoshi; Magara, Masaaki; Usuda, Shigekazu; Watanabe, Kazuo; Esaka, Fumitaka; Hirayama, Fumio; Lee, C. G.; Yasuda, Kenichiro; Kono, Nobuaki; Inagawa, Jun; et al.

Proceedings of International Conference on Nuclear Energy System for Future Generation and Global Sustainability (GLOBAL 2005) (CD-ROM), 6 Pages, 2005/10

no abstracts in English

Journal Articles

Investigation on Innovative Water Reactor for Flexible Fuel Cycle (FLWR), 2; Recycle characteristics

Okubo, Tsutomu; Uchikawa, Sadao; Kugo, Teruhiko; Akie, Hiroshi; Takeda, Renzo*

Proceedings of International Conference on Nuclear Energy System for Future Generation and Global Sustainability (GLOBAL 2005) (CD-ROM), 6 Pages, 2005/10

In order to ensure sustainable energy supply in the future based on the commercialized LWR technologies, a concept of Innovative Water Reactor for Flexible Fuel Cycle (FLWR) has been investigated in JAERI. Results on the FLWR recycling characteristics under possible various reprocessing schemes are presented in the present paper. The results show the recycling is possible a few times at most as long as the fissile Pu content stays over 60%, even in the high conversion type core with the conversion ratio around 0.9, under the simplified PUREX reprocessing, with relatively high average decontamination factor. For breeding core, the results have indicated that even under the reprocessing with relatively low DFs and with whole MA, the recycling is also feasible, suggesting all MAs from the core can be possibly recycled itself, although the core performances are a little degraded depending on MA and FP contents.

152 (Records 1-20 displayed on this page)