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JAEA Reports

Fuel unloading work in decommissioning of the prototype fast breeder reactor Monju; First stage of Monju decommissioning project

Shiota, Yuki; Kudo, Junki; Tsuno, Hiromi; Takeuchi, Ryotaro; Ariyoshi, Hideo; Shiohama, Yasutaka; Hamano, Tomoharu; Takagi, Tsuyohiko; Nagaoki, Yoshihiro

JAEA-Technology 2023-002, 87 Pages, 2023/06

JAEA-Technology-2023-002.pdf:8.53MB

In the first stage of Monju decommissioning project, fuel unload work began to be carried out. There are two tasks in this work. One is Fuel Treatment and Storage work that gets rid of sodium on the fuel assemblies unloaded from Ex-Vessel fuel Storage Tank (EVST) and carries it in the storage pool, and the other is Fuel Unloading that the fuel assemblies in the reactor core is replaced with dummy fuels and stored in EVST. Fuel Treatment and Storage work and Fuel Unloading work are performed alternately, and 370 fuel assemblies in the core and 160 fuel assemblies in EVST are all carried to the storage pool. Monju had a large amount of sodium in the reactor vessel and EVST, and there was a residual risk of fuel failure due to the superposition of a large scale sodium fire. Therefore, in the first stage of the Monju decommissioning project, it was decided to take about 5.5 years to remove the residual risk by storing all the fuel rods in the fuel storage pool. There are few Fuel handling system of Sodium Fast Reactor in the world, so the driving record and experience are not enough. So, events that occur even if taken measure are assumed. The following three events apply to this; first, events that are difficult to prevent, events. Second, that are due to lack of experience, and final, events optimization of system is not enough. Plans were taken to suppress these events. This report summarizes the "Monju decommissioning project" work conducted so far in all four campaigns.

Journal Articles

Verification of fuel assembly bowing analysis model for core deformation reactivity evaluation

Doda, Norihiro; Uwaba, Tomoyuki; Ohgama, Kazuya; Yoshimura, Kazuo; Nemoto, Toshiyuki*; Tanaka, Masaaki; Yamano, Hidemasa

Nihon Kikai Gakkai Kanto Shibu Dai-29-Ki Sokai, Koenkai Koen Rombunshu (Internet), 5 Pages, 2023/03

An evaluation method for reactivity feedback due to core deformation during reactor power increase in sodium-cooled fast reactors is being developed for realistic core design evaluation. In this evaluation method, fuel assembly bowing was modeled with a beam element of the finite element method, and the assembly's pad contact between adjacent assemblies was modeled with a dedicated element which could consider the wrapper tube cross-sectional distortion and the pad stiffness depending on pad contact conditions. This fuel assembly bowing analysis model was verified for thermal bowing of a single assembly and assembly pad contact between adjacent assemblies in a core as past benchmark problems. The calculation results by this model showed good agreement with those of reference solutions of theoretical solutions or results by participating institutions in the benchmark. This study confirmed that the analysis model was able to calculate thermal assembly bowing appropriately.

Journal Articles

Anaerobic methane-oxidizing activity in a deep underground borehole dominantly colonized by $$Ca.$$ Methanoperedenaceae

Nishimura, Hiroki*; Kozuka, Mariko*; Fukuda, Akari*; Ishimura, Toyoho*; Amano, Yuki; Beppu, Hikari*; Miyakawa, Kazuya; Suzuki, Yohei*

Environmental Microbiology Reports (Internet), 9 Pages, 2023/00

 Times Cited Count:0 Percentile:0.01(Environmental Sciences)

The family $$Ca.$$ Methanoperedenaceae archaea mediate anaerobic oxidation of methane (AOM). We newly developed a high-pressure laboratory incubation system and investigated groundwater from 214- and 249-m deep boreholes at Horonobe Underground Research Laboratory, Japan, where the high and low abundances of $$Ca.$$ Methanoperedenaceae archaea have been revealed, respectively. We incubated the samples amended with or without amorphous Fe(III) and $$^{13}$$C-labelled methane at an in-situ pressure of 1.6 MPa. After three to seven-day incubation, AOM activities were not detected from the 249-m sample but from the 214-m sample. The AOM rates were 93.7$$pm$$40.6 and 27.7$$pm$$37.5 nM/day with and without Fe(III) amendment. Suspended particulates were not visible in the 249-m sample on the filter, while they were abundant and contained amorphous Fe(III) and Fe(III)-bearing phyllosilicates in the 214-m sample. This supports the in-situ activity of Fe(III)-dependent AOM in the deep subsurface borehole.

Journal Articles

JENDL-5 benchmark test for shielding applications

Konno, Chikara; Ota, Masayuki*; Kwon, Saerom*; Onishi, Seiki*; Yamano, Naoki*; Sato, Satoshi*

Journal of Nuclear Science and Technology, 24 Pages, 2023/00

 Times Cited Count:0 Percentile:0.04(Nuclear Science & Technology)

JENDL-5 was validated from a viewpoint of shielding applications under the Shielding Integral Test Working Group of the JENDL Committee. The following benchmark experiments were selected: JAEA/FNS in-situ experiments, Osaka Univ./OKTAVIAN TOF experiments, ORNL/JASPER sodium experiments, NIST iron experiment and QST/TIARA experiments. These experiments were analyzed with MCNP and nuclear data libraries (JENDL-5, JENDL-4.0 or JENDL-4.0/HE, ENDF/B-VIII.0 and JEFF-3.3). The analysis results demonstrate that JENDL-5 is comparable to or better than JENDL-4.0 or JENDL-4.0/HE, ENDF/B-VIII.0 and JEFF-3.3.

Journal Articles

Core and safety design for France-Japan common concept on sodium-cooled fast reactor

Takano, Kazuya; Oki, Shigeo; Ozawa, Takayuki; Yamano, Hidemasa; Kubo, Shigenobu; Ogura, Masashi*; Yamada, Yumi*; Koyama, Kazuya*; Kurita, Koichi*; Costes, L.*; et al.

EPJ Nuclear Sciences & Technologies (Internet), 8, p.35_1 - 35_9, 2022/12

The France and Japan teams have carried out collaborative works to have common technical views regarding a sodium-cooled fast reactor concept. Japan has studied the feasibility of an enhanced high burnup low-void effect (CFV) core and fuel using oxide dispersion-strengthened steel cladding in ASTRID 600. Regarding passive shutdown capabilities, Japan team has performed a preliminary numerical analysis for ASTRID 600 using a complementary safety device, called a self-actuated shutdown system (SASS), one of the safety approaches of Japan. The mitigation measures of ASTRID 600 against a severe accident, such as a core catcher, molten corium discharge assembly, and the sodium void reactivity features of the CFV core, are promising to achieve in-vessel retention for both countries. The common design concept based on ASTRID 600 is feasible to demonstrate the SFR core and safety technologies for both countries.

Journal Articles

Legal compliance activities and nuclear security culture development activities in JAEA

Amano, Tsukasa; Sato, Yoshiharu; Shibata, Ryodai; Yamazaki, Katsuyuki; Shiromo, Hideo; Nakamura, Hironobu

Dai-43-Kai Nihon Kaku Busshitsu Kanri Gakkai Nenji Taikai Kaigi Rombunshu (Internet), 4 Pages, 2022/11

About ten years have passed since the introduction of nuclear security compliance and security culture development activities. While there have been changes in domestic and international situations, it is essential for nuclear security that the entire organization maintains an attitude that emphasizes compliance with laws and regulations and nuclear security. JAEA has been effectively implementing various activities with evaluation and improvement. Especially, e-learning which combines education and awareness, case studies, and internal audits are considered effective in maintaining nuclear security compliance and security culture development activities.

Journal Articles

Clogging properties of HEPA filter induced by loading of soot from burned glove-box panel materials

Tashiro, Shinsuke; Ono, Takuya; Amano, Yuki; Yoshida, Ryoichiro; Watanabe, Koji*; Abe, Hitoshi

Nuclear Technology, 208(10), p.1553 - 1561, 2022/10

 Times Cited Count:0 Percentile:0.01(Nuclear Science & Technology)

To contribute to the confinement safety evaluation of the radioactive materials in the Glove box (GB) fire accident, combustion tests with the Polymethyl methacrylate (PMMA) and the Polycarbonate (PC) as typical panel materials for the GB have been conducted with a relatively large scale apparatus. As the important data for evaluating confinement safety, the release ratio and the particle size distribution of the soot generated from the burned materials were obtained. Furthermore, the rise of the differential pressure ($$Delta$$P) of the high efficiency particle air (HEPA) filter by the soot loading was also investigated. As results, the release ratio of the soot from the PC was about seven times as large as the PMMA. In addition, it was found that the behavior of the rise of the $$Delta$$P with soot loading could be represented uniformly regardless of kinds of combustion materials by considering effect of the loading volume of the soot particle in the relatively low loading region.

JAEA Reports

Fuel unloading operations -2020- in the decommissioning of the prototype fast breeder reactor "Monju"

Shiota, Yuki; Ariyoshi, Hideo; Shiohama, Yasutaka; Isobe, Yuta; Takeuchi, Ryotaro; Kudo, Junki; Hanaki, Shotaro; Hamano, Tomoharu; Takagi, Tsuyohiko

JAEA-Technology 2022-019, 95 Pages, 2022/09

JAEA-Technology-2022-019.pdf:7.59MB

In the first stage of "Monju" decommissioning project, "Fuel Unloading Operations" have been carrying out. The operations consists of two processes. The first process is "Fuel Treatment and Storage" is that the fuel assemblies unloaded from the Ex-Vessel fuel Storage Tank (EVST) are canned after sodium cleaning, and then transferred to the storage pool. The second process is "Fuel Unloading" that the fuel assemblies in the reactor core are replaced with dummy fuel assemblies and stored in the EVST. "Fuel Treatment and Storage" and "Fuel Unloading" are performed alternately until 370 fuel assemblies in the core and 160 fuel assemblies in the EVST are all transferred to the storage pool. This is a summary of "Fuel Unloading" in the third quarter of "Fuel Unloading Operation". In fiscal 2020, as "Fuel Unloading", 72 fuel assemblies and 74 blanket fuel assemblies were unloaded from the core, and stored in the EVST. From the EVST, 145 dummy fuel assemblies and 1 fixed absorber were loaded in the core instead. During these operations, a total of 36 cases alarming or equipment malfunctions classified into 4 types occurred. However, these events were estimated in advance, there were no significant events that menaces to safety of fuel assemblies and equipment. Therefore, there were no serious problem like fall of fuel assemblies and events that may affect schedule of the project like stick of gripper of ex-vessel fuel transfer machine. When equipment's work or performance fail, the operation continued with safety by elimination of causes of problem. Fuel handling system of Monju has function that is endemic to sodium cooling fast breeding reactor. Because continuous operations of fuel handling system with actual fuel assemblies start recently, we don't have as much experience as PWR and BWR. With estimation of various troubles, reduction of frequency of trouble occurrence and minimization of impacts on schedule performed.

Journal Articles

Normal spectral emissivity, specific heat capacity, and thermal conductivity of type 316 austenitic stainless steel containing up to 10 mass% B$$_{4}$$C in a liquid state

Fukuyama, Hiroyuki*; Higashi, Hideo*; Yamano, Hidemasa

Journal of Nuclear Materials, 568, p.153865_1 - 153865_12, 2022/09

 Times Cited Count:2 Percentile:80.5(Materials Science, Multidisciplinary)

The normal spectral emissivity, specific heat capacity and thermal conductivity of type 316 austenitic stainless steel (SS) containing boron carbide (B$$_{4}$$C) in a liquid state were experimentally measured over the composition range of SS-$$x$$ mass% B$$_{4}$$C (up to 10%) and wide temperature ranges using an electromagnetic levitator in a static magnetic field. The normal spectral emissivity and specific heat capacity were almost constant against temperature for all SS-B$$_{4}$$C melts, and the thermal conductivities of the melts had a negligible or small positive temperature dependence. The B$$_{4}$$C-content dependence of each property at 1800 K had a different tendency across the eutectic composition (around 3 mass% B$$_{4}$$C) of the SS-B$$_{4}$$C pseudo-binary system.

Journal Articles

Study on eutectic melting behavior of control rod materials in core disruptive accidents of sodium-cooled fast reactors, 1; Project overview and progress until 2020

Yamano, Hidemasa; Takai, Toshihide; Emura, Yuki; Fukuyama, Hiroyuki*; Higashi, Hideo*; Nishi, Tsuyoshi*; Ota, Hiromichi*; Morita, Koji*; Nakamura, Kinya*; Fukai, Hirofumi*; et al.

Proceedings of 13th International Topical Meeting on Nuclear Reactor Thermal-Hydraulics, Operation and Safety (NUTHOS-13) (Internet), 12 Pages, 2022/09

This paper describes the project overview and progress of experimental and analytical studies conducted until 2020. Specific results in this paper are the measurement of the eutectic reaction rates and the validation of physical model describing the eutectic reaction in the analysis code through the numerical analysis of the B$$_{4}$$C-SS eutectic reaction rate experiments in which a B$$_{4}$$C pellet was placed in a SS crucible.

JAEA Reports

Fuel Unloading Operations -2019- in the decommissioning of prototype fast breeder reactor "Monju"

Yabe, Takanori; Murakami, Makio; Shiota, Yuki; Isobe, Yuta; Shiohama, Yasutaka; Hamano, Tomoharu; Takagi, Tsuyohiko; Nagaoki, Yoshihiro

JAEA-Technology 2022-002, 66 Pages, 2022/07

JAEA-Technology-2022-002.pdf:10.45MB

In the first stage of "Monju" decommissioning project, "Fuel Unloading Operations" have been carrying out. The operations consists of two processes. The first process is "Fuel Treatment and Storage" is that the fuel assemblies unloaded from the Ex-Vessel fuel Storage Tank are canned after sodium cleaning, and then transferred to the storage pool. The second process is "Fuel Unloading" that the fuel assemblies in the reactor core are replaced with dummy fuel assemblies and stored in the Ex-Vessel fuel Storage Tank. "Fuel Treatment and Storage" and "Fuel Unloading" are performed alternately until 370 fuel assemblies in the core and 160 fuel assemblies in the Ex-Vessel fuel Storage Tank are all transferred to the storage pool. In fiscal 2018, as "Fuel Treatment and Storage", 86 fuel assemblies were transferred to the storage pool. As "Fuel Unloading", 76 dummy fuel assemblies were stored in the Ex-Vessel fuel Storage Tank. In fiscal 2019, as "Fuel Unloading", 60 fuel assemblies and 40 blanket fuel assemblies were unloaded from the core. These assemblies were stored in the Ex-Vessel fuel Storage Tank, and dummy fuel assemblies were loaded into the core instead. During these operations, a total of 38 cases of alarming or equipment malfunctions classified into 24 types occurred. However, no significant events that menaces to safety have occurred. The operations were continued safely by removing the direct factors for the malfunctions in the equipment operation and performance.

JAEA Reports

Fuel unloading work in decommissioning of the prototype fast breeder reactor Monju; Fuel treatment and storage work in 2018 and 2019

Shiota, Yuki; Yabe, Takanori; Murakami, Makio; Isobe, Yuta; Sato, Masami; Hamano, Tomoharu; Takagi, Tsuyohiko; Nagaoki, Yoshihiro

JAEA-Technology 2022-001, 117 Pages, 2022/07

JAEA-Technology-2022-001.pdf:25.55MB

In the first stage of Monju decommissioning project, fuel unload work began to be carried out. There are two tasks in this work. One is fuel treatment and storage work that gets rid of sodium on the fuel assemblies unloaded from Ex-Vessel fuel Storage Tank (EVST) and carries it in the storage pool, and the other is fuel unloading that the fuel assemblies in the reactor core is replaced with dummy fuels and stored in EVST. Fuel treatment and storage work and fuel unloading work are performed alternately, and 370 bodies in the core and 160 pieces in EVST are all carried to the storage pool. 86 fuel assemblies was carried to the storage pool in fuel treatment and storage work in 2018 and 76 dummy fuels were stored in EVST for fuel unloading work. During the work, 86 types and 232 alarms / malfunctions occurred, but there was no impact on safety. There was one equipment's failure at gripper's claw open / close clutch of ex-vessel fuel transfer machine B, but it was repaired and restarted. Also it was eliminated the cause of problem or concession that the equipment failure due to the sticking of the sodium compound and continuous use of the equipment. Some problems related to system control occurred, but the work was done after checking the safety. With estimation of various troubles, reduction of frequency of trouble occurrence and minimization of impacts on schedule performed.

Journal Articles

Experimental study on reaction behavior between control rod material and molten stainless steel for core disruptive accidents of sodium-cooled fast reactors

Emura, Yuki; Kamiyama, Kenji; Yamano, Hidemasa

Dai-26-Kai Doryoku, Enerugi Gijutsu Shimpojiumu Koen Rombunshu (Internet), 4 Pages, 2022/07

no abstracts in English

Journal Articles

Evolution of the reaction and alteration of granite with Ordinary Portland cement leachates; Sequential flow experiments and reactive transport modelling

Bateman, K.*; Murayama, Shota*; Hanamachi, Yuji*; Wilson, J.*; Seta, Takamasa*; Amano, Yuki; Kubota, Mitsuru*; Ouchi, Yuji*; Tachi, Yukio

Minerals (Internet), 12(7), p.883_1 - 883_20, 2022/07

 Times Cited Count:0 Percentile:0.02(Geochemistry & Geophysics)

Journal Articles

Differential pressure changes of a high airflow-type HEPA filter during solvent fire in reprocessing facilities

Tashiro, Shinsuke; Uchiyama, Gunzo; Ono, Takuya; Amano, Yuki; Yoshida, Ryoichiro; Abe, Hitoshi

Nuclear Technology, 208(7), p.1205 - 1213, 2022/07

 Times Cited Count:0 Percentile:0.01(Nuclear Science & Technology)

A clogging behavior of a high-efficiency particulate air (HEPA) filter at solvent fire accidents for reprocessing facilities has been studied. In this study, the burning rates of 30% tri-butyl phosphate (TBP)/dodecane (DD) mixed solvent and DD solvent and the differential pressure ($$Delta$$P) of a high airflow typed HEPA filter applied in the actual facilities in japan were measured. It was confirmed that the mainly burned was DD at the early stage of the mixed solvent burning and the TBP at the late stage. Furthermore, it was found that the $$Delta$$P rapidly rose at the late stage of the mixed solvent burning. The increase of the release ratio of the unburned particulate composition (TBP, its degraded solvent and inorganic phosphorus (P$$_{2}$$O$$_{5}$$)) was considered to contribute to the rapid rise. The correlating formulas with the $$Delta$$P and the mass of the loading particulates, except for the region of the rapid rise of $$Delta$$P, could be induced.

Journal Articles

Temporal variability of $$^{137}$$Cs concentrations in coastal sediments off Fukushima

Suzuki, Shotaro*; Amano, Yosuke*; Enomoto, Masahiro*; Matsumoto, Akira*; Morioka, Yoshiaki*; Sakuma, Kazuyuki; Tsuruta, Tadahiko; Kaeriyama, Hideki*; Miura, Hikaru*; Tsumune, Daisuke*; et al.

Science of the Total Environment, 831, p.154670_1 - 154670_15, 2022/07

 Times Cited Count:1 Percentile:41.02(Environmental Sciences)

Journal Articles

Development plan of failure mitigation technologies for improving resilience of nuclear structures

Kasahara, Naoto*; Yamano, Hidemasa; Nakamura, Izumi*; Demachi, Kazuyuki*; Sato, Takuya*; Ichimiya, Masakazu*

Transactions of 26th International Conference on Structural Mechanics in Reactor Technology (SMiRT-26) (Internet), 8 Pages, 2022/07

Utilizing fracture control, we are developing a technology to suppress the expansion of damage caused by an event that exceeds the design assumption. We made a plan to develop three issues; (1) Technology for mitigating failure consequence at extremely high temperatures, (2) Technology for mitigating failure consequence against excessive earthquakes, and (3) Methodology for improving reactor structure resilience.

Journal Articles

Vibration test and fatigue test for failure probability evaluation method with integrated energy

Kinoshita, Takahiro*; Okamura, Shigeki*; Nishino, Hiroyuki; Yamano, Hidemasa; Kurisaka, Kenichi; Futagami, Satoshi; Fukasawa, Tsuyoshi*

Transactions of 26th International Conference on Structural Mechanics in Reactor Technology (SMiRT-26) (Internet), 7 Pages, 2022/07

The seismic evaluation of key components such as reactor vessel is important for the Seismic Probabilistic Risk Assessment (S-PRA) in a Sodium-Cooled Fast Reactor (SFR). Many components were damaged by cumulative damage like fatigue damage during seismic ground motion. However, general evaluation method for key components under seismic ground motion has been based on static loads and elastic region of materials. More accurate evaluation method for S-PRA, which can evaluate the failure of key components such as reactor vessels, has been actually required. In this study, failure probability evaluation method with integrated energy was developed by comparing the energy with vibration tests and fatigue tests. Vibration tests were performed to evaluate integrated vibration energy at failure by energy balance equation and fatigue tests were performed to evaluate integrated vibration energy at failure based on experimental results of fatigue tests.

Journal Articles

France-Japan collaboration on thermodynamic and kinetic studies of core material mixture in severe accidents of sodium-cooled fast reactors

Quaini, A.*; Goss$'e$, S.*; Payot, F.*; Suteau, C.*; Delacroix, J.*; Saas, L.*; Gubernatis, P.*; Martin-Lopez, E.*; Yamano, Hidemasa; Takai, Toshihide; et al.

Proceedings of International Conference on Fast Reactors and Related Fuel Cycles; Sustainable Clean Energy for the Future (FR22) (Internet), 10 Pages, 2022/04

CEA and JAEA defined new sub-tasks under the current implementing arrangement: Kinetics of interaction in core material mixtures- Physical properties of core material mixtures, High temperature thermodynamic data for the UO$$_{2}$$-Fe-B$$_{4}$$C system, Experimental studies on B$$_{4}$$C-SS kinetics and B$$_{4}$$C-SS eutectic material relocation (freezing), B$$_{4}$$C/SS eutectic and kinetics models for SIMMER code systems, Methodology for the modelling of mixtures liquefaction kinetics. The paper describes major R&D results obtained in the France-Japan collaboration under the previous implementing arrangement as well as experimental and analytical roadmaps under the current arrangement.

JAEA Reports

Effect of nitrous acid on migration behavior of gaseous ruthenium tetroxide into liquid phase

Yoshida, Naoki; Ono, Takuya; Yoshida, Ryoichiro; Amano, Yuki; Abe, Hitoshi

JAEA-Research 2021-011, 12 Pages, 2022/01

JAEA-Research-2021-011.pdf:1.49MB

In boiling and drying accidents involving high-level liquid waste in fuel reprocessing plants, emphasis is placed on the behavior of ruthenium (Ru). Ru would form volatile species, such as ruthenium tetroxide (RuO$$_{4}$$), and could be released to the environment with coexisting gases, including nitric acid, water, or nitrogen oxides. In this study, to contribute toward safety evaluations of these types of accidents, the migration behavior of gaseous Ru into the liquid phase has been experimentally measured by simulating the condensate during an accident. The gas absorption of RuO$$_{4}$$ was enhanced by increasing the nitrous acid (HNO$$_{2}$$) concentration in the liquid phase, indicating the occurrence of chemical absorption. In control experiments without HNO$$_{2}$$, the lower the temperature, the greater was the Ru recovery ratio in the liquid phase. Conversely, in experiments with HNO$$_{2}$$, the higher the temperature, the higher the recovery ratio, suggesting that the reaction involved in chemical absorption was activated at higher temperatures.

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