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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.

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

Examinations of performance for utilization of radiation portal monitors in contamination inspection for motor vehicles in nuclear emergency (Contract research)

Hiraoka, Hirokazu; Komatsuzaki, Joji*; Hanaka, Norihiko*; Okamoto, Akiko; Saito, Yoshihiko*; Munakata, Masahiro; Togawa, Orihiko

JAEA-Technology 2022-003, 70 Pages, 2022/07

JAEA-Technology-2022-003.pdf:2.53MB

In contamination inspections for public in nuclear emergency, it is assumed that a large number of motor vehicles and evacuees will be inspected. At present, first measurement points of vehicles are tires and around a wiper, and they are basically inspected by persons using portable radiation surface contamination meters such as GM survey meters. However, in order to efficiently inspect contamination of vehicles, utilization of portable radiation portal monitors is being considered for the inspection. In this study, examination of performance of the portal monitor was conducted in order to obtain basic data. In this examinations, sealed radiation sources, $$^{133}$$Ba, were substituted for the contamination of Operational Intervention Level 4 (OIL4), the evaluation criteria to conduct simple decontamination in the contamination inspection. The radiation source was attached practically to a tire and around the wipers of a vehicle, and the counting rate of $$gamma$$-rays from the radiation sources was measured using the portal monitor. Three examinations were conducted: static examination with a vehicle stationary, moving examination to mock the actual inspection, and high back ground examination to investigate performance of the equipment in high back ground environment. The vehicle mainly used in the experiments was a sedan, which is generally used as standard vehicles. And, a van whose front body is like that of a bus was used at this experiment. In addition, the "Gamma Pole" manufactured by the Chiyoda Technol Corporation was used as the portal monitor. As the result, it was estimated that the case of contamination equivalent to 40,000 cpm, the default value of the OIL4, deposited on the tread on a vehicle tire, when the vehicle passes through the portal monitor at 10 km/h or less, the contamination would be detected with a probability of 99% or more. Similarly, when the contamination deposited on around the wipers and the vehicle speed is 5 km/h or less,

JAEA Reports

Impact assessment for internal overflow in JRR-3

Tokunaga, Sho; Iguchi, Shintaro; Kawamura, Sho; Hirane, Nobuhiko

JAEA-Technology 2022-004, 74 Pages, 2023/02

JAEA-Technology-2022-004.pdf:2.46MB

In JRR-3, in response to the new regulatory standard for research reactors that is enforced December 2013, we submitted the application document of reactor installation license for the JRR-3 on September 2014, and acquired the permission on November 2018. Thereafter, we carried out impact assessment for internal overflow based on the design principles as described in the application document of reactor installation license for the JRR-3. There are two legal requirements for internal overflow. The first is to ensure that the necessary safety functions are not lost due to internal overflow that occur in the facility. The second is to prevent leakage of liquid containing radioactive materials outside the radiation controlled area in the event of an internal overflow. For these requirements, assuming each overflow source, it was confirmed that the necessary safety functions would not be lost and that liquid containing radioactive materials would not leak outside the controlled area. Regarding these assessments, design and construction plans were submitted in installments, and the approvals were obtained sequentially. This report shows the result that is impact assessment for internal overflow in JRR-3.

JAEA Reports

Completion of waste removal work from the hot cell of Operation Testing Laboratory in Tokai Reprocessing Plant

Goto, Yuichi; Suzuki, Yoshimasa; Horigome, Kazushi; Miyamoto, Toshihiko*; Usui, Masato*; Mori, Eito*; Kuno, Takehiko

JAEA-Technology 2022-005, 42 Pages, 2022/07

JAEA-Technology-2022-005.pdf:4.48MB

Radioactive wastes were generated and stored in the hot cell of Operation Testing Laboratory of Tokai Reprocessing Plant due to the experiments related to the reprocessing technology development from 1974 to 2014. Waste removal work was strengthened by the shift work in the past, however another wastes were generated by the equipment dismantling. From 2006, an improved waste removal method was established by using bag-out technique and wastes were taken from the glove-box connected to the hot cell. The removal period, estimated from the conventional method using Cask No. 10, was reduced from 14 to 5 years. From 2016, upgrade of worker's awareness including related departments was performed by various software and hardware improvements. Also, the worker's skills were improved and equipment in Cask No.10 was checked for preventive maintenance. The prevention measures for past troubles were discussed with Radiation Control Department. In addition, transportation schedule including safety operation with Transportation Department and Waste Receiving Department was optimized to maintain the waste removal cycle. The removal period was reduced from 5 to 3 years by the above efforts. Finally, the work was completed in March 2020.

JAEA Reports

Continuous improvement activities on nuclear facility maintenance in Nuclear Science Research Institute of Japan Atomic Energy Agency in 2021

Task Force on Maintenance Optimization of Nuclear Facilities

JAEA-Technology 2022-006, 80 Pages, 2022/06

JAEA-Technology-2022-006.pdf:4.24MB

The Task force on maintenance optimization of nuclear facilities was organized in the Nuclear Science Research Institute (NSRI) of Japan Atomic Energy Agency (JAEA) since November 2020, in order to adequately respond to "the New nuclear regulatory inspection system since FY 2020" and to continuously improve the facility maintenance activities. In 2021, the task force has studied (1) optimization of the importance classification on maintenance and inspection of nuclear facilities, and (2) improvement in setting and evaluation of the performance indicators on safety, maintenance and quality management activities, considering "the Graded approach" that is one of the basic methodologies in the new nuclear regulatory inspection system. Each nuclear facility (research reactors, nuclear fuel material usage facilities, others) in the NSRI will steadily improve their respective safety, maintenance and quality management activities, referring the review results suggested by the task force.

JAEA Reports

Development of dissolved hydrogen concentration control apparatus by solid polymer electrolyte water electrolysis method

Nakano, Hiroko; Fuyushima, Takumi; Tsuguchi, Akira*; Nakamura, Mutsumi*; Takeuchi, Tomoaki; Takemoto, Noriyuki; Ide, Hiroshi

JAEA-Technology 2022-007, 34 Pages, 2022/06

JAEA-Technology-2022-007.pdf:3.35MB

In order to investigate the phenomenon of stress corrosion cracking (SCC) for structural materials at the light water reactor (LWR), it is important to manage a water quality for simulating high-temperature and high-pressure water. Generally, dissolved hydrogen (DH) concentration in water loop has been controlled by the bubbling method of pure hydrogen gas or standard gas with high hydrogen concentration. However, it is necessary to equip the preventing hydrogen explosion in the area installed experimental apparatus. In general, in order to prevent accident by hydrogen, it is required to take measures such as limiting the amount of leakage, eliminating hydrogen, shutting off the power supply, and suppressing combustion before an explosion occurs. Thus, the dissolved hydrogen concentration control apparatus by electrolysis method has been developed which has two electrolysis cells to control DH concentration by electrolyzing water loop. In this study, small basic experimental devices were set up. The preliminary data were acquired regarding the simple performance of two electrolysis cells and the change of DH concentration in circulation. Based on the preliminary data, the dissolved hydrogen concentration control apparatus was designed to be connected to the high-temperature and high-pressure water loop test equipment. This report describes the test results with the small basic experimental devices for the design of the dissolved hydrogen concentration control apparatus.

JAEA Reports

Preliminary evaluation of environmental uranium concentration originated from trench disposal facilities

Ogawa, Rina; Abe, Daichi*; Sugaya, Toshikatsu; Sakuma, Kota; Saito, Tatsuo; Sakai, Akihiro

JAEA-Technology 2022-008, 46 Pages, 2022/05

JAEA-Technology-2022-008.pdf:3.09MB

Japan Atomic Energy Agency (JAEA) has planned to dispose of the Uranium-bearing waste, whose radioactivity concentration is low, in trench disposal facility. In Japan, uranium is a material to impact on human health, therefore Environmental quality standards for water pollution for uranium has been established, and the standard value is 0.002mg/L. Safety of trench disposal facilities will be assessed that radionuclides contained in the radioactive waste are transferred to the biosphere by seepage water and groundwater. Therefore, JAEA considers that not only dose evaluation but also environmental pollution evaluation is needed as a safety assessment. In this report, we examined whether the concentration of uranium leaching from the trench facility in the aquifer can meet the Environmental quality standards. In addition, parameter study under various conditions of disposal facility were done. Based on the results, conditions and issues of future basic design of trench disposal facility were discussed. The uranium concentration in the aquifer was calculated by the one-dimensional dose evaluation code "GSA-GCL2" for the disposal of LLW. As the result, the uranium concentration in the aquifer significantly changed depending on the conditions of design of disposal facility and so on. However, if the shape and arrangement of the trench facility to groundwater flow direction, the distribution coefficient of uranium of the waste layer, the specification of the impermeable layer and their combination are appropriately designed we consider that the uranium concentration of aquifer can made to adapt the environmental quality standard.

JAEA Reports

Study on radioactivity evaluation method of research reactors using DORT and MCNP codes

Kochiyama, Mami; Sakai, Akihiro

JAEA-Technology 2022-009, 56 Pages, 2022/06

JAEA-Technology-2022-009.pdf:4.15MB

It is necessary to evaluate radioactivity inventory in wastes before disposal of low-level radioactive wastes generated from dismantling research reactors. It is efficient for owners of each research reactor to use a common radioactive evaluation method in order to comply with the license application for disposal facility. In this report, neutron transport and activation calculations were carried out for the Rikkyo University research reactor in order to examine a common radioactivity evaluation method for burial disposal of radioactive wastes generated by dismantling. We adopted the neutron transport codes DORT and MCNP and the activation code ORIGEN-S with cross-section libraries based on JENDL-4.0 and JENDL/AD-2017. The radioactivity concentrations obtained by the radiochemical analysis and both calculation codes were in agreement by 0.4 to 3 times. Therefore, by appropriately considering this difference, the radioactivity evaluation method by DORT, MCNP and ORIGEN-S can be applied to the radioactivity evaluation for buried disposal. In order to classify wastes from dismantling by clearance or buried disposal method according to their radioactivity levels, we also created radioactivity concentration distributions in the concrete area and graphite thermal column area.

JAEA Reports

Design study on cover soil in the trench disposal facility for very low-level radioactive waste generated from research facilities and other facilities

Ogawa, Rina; Nakata, Hisakazu; Sugaya, Toshikatsu; Sakai, Akihiro

JAEA-Technology 2022-010, 54 Pages, 2022/07

JAEA-Technology-2022-010.pdf:11.07MB

Japan Atomic Energy Agency has considered trench disposal as one of the disposal methods for radioactive wastes generated from research facilities and other facilities. The trench disposal facility is regulated by "Act on the Regulation of Nuclear Source Material, Nuclear Fuel Material and Reactors". In particular, the design of the trench facility is regulated by a rule under the law. When the rule was amended in 2019, the design of the trench disposal facility required equipment to reduce ingress of rain water and groundwater. In the report, studies on the design of a trench disposal facility to adapt to the amended rule were performed. The trench disposal facility has considered being established in a place lower than groundwater level. Therefore, it was decided to study covering soil at the upper part of the trench facility, because the ingress water in the facility is mainly derived from rain water. In this study, it was decided to evaluate the design of covering soil of the radioactive waste categorized into chemically stable materials. Therefore, as the examination method, a parameter study on varying the permeability coefficient and thickness of the layers composing cover soil. In the parameter study, the velocity of the water infiltrating into the trench facility was evaluated. Based on the results, more efficient design of the layers composing the covering soil was considered. The result showed that the impermeable efficiency of the covering soil was different depending on the thickness and the permeability conductivity of each layer. As a result, it was possible to understand the impermeable performance of covering soil by the permeability coefficient and thickness of each layer. We will plan to decide the specification of the cover soil while examination of future tasks and cost in the basic design.

JAEA Reports

Safety design philosophy of HTTR Heat Application Test Facility

Aoki, Takeshi; Shimizu, Atsushi; Iigaki, Kazuhiko; Okita, Shoichiro; Hasegawa, Takeshi; Mizuta, Naoki; Sato, Hiroyuki; Sakaba, Nariaki

JAEA-Technology 2022-011, 60 Pages, 2022/07

JAEA-Technology-2022-011.pdf:2.08MB

Japan Atomic Energy Agency is planning a High Temperature Engineering Test Reactor (HTTR) heat application test producing hydrogen with the HTTR which achieved the highest reactor outlet coolant temperature of 950$$^{circ}$$C in the world to realize a massive, cost-effective and carbon-free hydrogen production technology utilizing a high temperature gas cooled reactor (HTGR). In the HTTR heat application test, it is required to establish its safety design for coupling a hydrogen production plant to HTGR through the licensing by the Nuclear Regulation Authority (NRA). A draft of a safety design philosophy for the HTTR heat application test facility was considered taking into account postulated events due to the plant modification and coupling of the hydrogen production plant based on the HTTR safety design which was authorized through the safety review of the NRA against New Regulatory Requirements. The safety design philosophy was examined to apply proven conventional chemical plant standards to the hydrogen production plant for ensuring public safety against disasters caused by high pressure gases. This report presents a result of a consideration on safety design philosophies regarding the reasonability and condition to apply the High Pressure Gas Safety Act for the hydrogen production plant, safety classifications, seismic design classification, identification of important safety system.

JAEA Reports

Effect of repairing refractory material of main reactor in steam reforming system

Kijima, Jun; Koyama, Hayato; Owada, Mitsuhiro; Hagiwara, Masayoshi; Aoyagi, Yoshitaka

JAEA-Technology 2022-012, 14 Pages, 2022/07

JAEA-Technology-2022-012.pdf:1.51MB

Steam reforming system has been developed for the treatment of organic wastes which are not suitable materials (halogenated oil) for the incineration due to generation of corrosive compounds and plugging materials. The refractory material is cast inside the main reactor, which is a part of the steam reforming system. Since the surface of this refractory material has deteriorated over time, the main reactor was replaced. If the refractory material surface of the used main reactor can be repaired, the used main reactor can be reused as a spare. The refractory material surface was repaired using two types of repair materials ("S" and "P"). Combustion tests were conducted on samples simulating organic wastes to evaluate each repair material. As a result of the combustion test, it was concluded that the repair of the main reactor was possible to use the repair material "P" because no cracks or flakes were observed.

JAEA Reports

Evaluation of radioactivity concentration corresponding to dose criterion for near surface disposal of radioactive waste generated from research, medical, and industrial facilities, Volume 2

Sakuma, Kota; Abe, Daichi*; Okada, Shota; Sugaya, Toshikatsu; Nakata, Hisakazu; Sakai, Akihiro

JAEA-Technology 2022-013, 200 Pages, 2022/08

JAEA-Technology-2022-013.pdf:8.41MB

Japan Atomic Energy Agency has aims to carry out near surface disposal of low-level radioactive waste generated from research, medical, and industrial facilities. Therefore, Radioactivity Concentration Corresponding to Dose Criterion for near surface disposal for nuclides in the waste were calculated for the purpose of discussion for radioactivity limits between trench and concrete vault disposal, and key nuclides related to them. This report uses the results of sensitivity analysis and evaluation of the amount of leachate from the disposal facility for concrete vault disposal, and incorporates a new assessment pathway and exposure form that widely assume the conditions of the disposal facility. This trial calculation was carried out and compared with the trial calculation in the previous report, "Evaluation of Radioactivity Concentration Corresponding to Dose Criterion for Near Surface Disposal of Radioactive Waste Generated from Research, Medical, and Industrial Facilities, Volume 1". The results of Radioactivity Concentration Corresponding to Dose Criterion calculated in this report will be used as reference values when selecting key nuclides and for classification into concrete vault disposal when the location has not been decided. After deciding the location of the site, it is necessary to evaluate the dose based on the location conditions.

JAEA Reports

Improvement of load control unit in material irradiation test system (Contract research)

Okada, Yuji; Magome, Hirokatsu; Matsui, Yoshinori

JAEA-Technology 2022-014, 113 Pages, 2022/09

JAEA-Technology-2022-014.pdf:15.79MB

Material irradiation test system had been newly installed in JMTR (Japan Materials Testing Reactor) with taking 5 years which was from 2008 through 2013. The aim of material irradiation test system is to conduct IASCC (Irradiation Assisted Stress Corrosion Cracking) evaluation study. This system is mainly consist of water control unit, which can simulate elevated temperature and pressure of the light water reactor environment in the reactor, and load control unit, which can perform the crack propagation examination under irradiation. This load control unit gives a load to CT (Compact Tension) specimen, and perform the crack propagation examination. The principle of loading to CT specimen is using pressure difference between pressure generated by high temperature and high pressure water by water control unit in capsule and pressure generated by load gas pressure supplied by helium gas cylinder in bellows installed in load control unit. In 2013, the commissioning of material irradiation test system was carried out for adjustment. During this commissioning, the correlation between the differential pressure in load control unit and the load was confirmed by using the test container connected to load control unit with load cell. From the results of commissioning, the problem, which the load change speeds at loading and unloading were different due to different pressure change speeds by the piping resistance performance in the periodic loading test in which load from minimum to maximum repeatedly applied, was confirmed. This report summarizes the problem of load change speed due to the piping resistance performance, which was confirmed in 2013, the improvement and performance test of load control unit for solving the problem described above, which were carried out from 2014 to 2015, and operating procedure.

JAEA Reports

Calculation of nuclear core parameters for HTTR; Report of summer holiday practical training 2021

Isogawa, Hiroki*; Naoi, Motomasa*; Yamasaki, Seiji*; Ho, H. Q.; Katayama, Kazunari*; Matsuura, Hideaki*; Fujimoto, Nozomu*; Ishitsuka, Etsuo

JAEA-Technology 2022-015, 18 Pages, 2022/07

JAEA-Technology-2022-015.pdf:1.37MB

As a summer holiday practical training 2021, the impact of 10 years long-term shutdown on critical control rod position of the HTTR and the delayed neutron fraction ($$beta$$$$_{rm eff}$$) of the VHTRC-1 core were investigated using Monte-Carlo MVP code. As a result, a long-term shutdown of 10 years caused the critical control rods of the HTTR to withdraw about 4.0$$pm$$0.8 cm compared to 3.9 cm in the experiment. The change in critical control rods position of the HTTR is due to the change of some fission products such as $$^{241}$$Pu, $$^{241}$$Am, $$^{147}$$Pm, $$^{147}$$Sm, $$^{155}$$Gd. Regarding the $$beta$$$$_{rm eff}$$ calculation of the VHTRC-1 core, the $$beta$$$$_{rm eff}$$ value is underestimate of about 10% in comparison with the experiment value.

JAEA Reports

The Preliminary tests of the plasma melting treatment for homogenization of low-level radioactive wastes

Nakashio, Nobuyuki*; Osugi, Takeshi; Kurosawa, Shigenobu; Ishikawa, Joji; Hemmi, Ko; Iketani, Shotaro; Yokobori, Tomohiko

JAEA-Technology 2022-016, 47 Pages, 2022/08

JAEA-Technology-2022-016.pdf:2.23MB

The Nuclear Science Research Institute (NSRI) of the Japan Atomic Energy Agency (JAEA) started operation of the Advanced Volume Reduction Facilities (AVWF) for production of waste packages for disposal of low-level radioactive solid wastes (LLW). To clarify the operating conditions for homogenization of non-metallic LLW, preliminary tests were carried out using the plasma melting furnace of the non-metal melting unit. The fluidity of molten waste influences homogenization conditions of solidified products. It was clarified that the viscosity, which is determined by the chemical composition and the melting temperature, influence the fluidity of molten waste greatly through previous literature review and the small-scale melting tests. In the preliminary tests, the simulated waste with a cold tracer loaded in 200 L drums were melted. Using the waste chemical components (basicity, iron oxide concentration) as an experimental parameter, the homogeneity of the chemical components of the solidified product was investigated and the homogenization conditions of melting tests were examined. The retention ratio of the tracer in the molten bath was also confirmed. The viscosity of the molten wastes was measured and the correlation with homogeneity was examined. In addition, the technical requirements that should be concerned in advance for future actual operation were discussed.

JAEA Reports

Evaluation on activation activity of radioactive materials remaining in JMTR Reactor Facility

Nagata, Hiroshi; Otsuka, Kaoru; Omori, Takazumi; Ide, Hiroshi

JAEA-Technology 2022-017, 113 Pages, 2022/08

JAEA-Technology-2022-017.pdf:6.17MB

Japan Materials Testing Reactor (JMTR) was decided as a one of decommission facilities in April 2017. The activation activity of radioactive materials remaining in the reactor facility was evaluated in order to submit the decommissioning plan to the Nuclear Regulation Authority. Total activation activity was 9.3$$times$$10$$^{18}$$ Bq after the permanent shutdown of reactor, 2.7$$times$$10$$^{16}$$ Bq after 21 years, 1.0$$times$$10$$^{16}$$ Bq after 40 years and 2.4$$times$$10$$^{15}$$ Bq after 100 years. The structure with high activation activity was the core structural materials in JMTR such as beryllium frame, aluminum reflector, etc., and the material was stainless steel, beryllium, etc. The ratio of nuclides to the total amount of activated radioactivity was highest in H-3 until about 40 years after the reactor shutdown, and then in Ni-63. For reference, the radioactivity level was classified based on the results of the obtained radioactivity concentration. The ratio of the weight of each radioactivity level to the total weight was 0.3-0.4% (10-13t) for L1, 0.0-0.4% (0-14t) for L2, 1.0-1.2% (32-39t) for L3 and 98.0-98.7% (about 3200t) for CL until 100 years after the reactor shutdown. It was found that those classified as CL account for more than 90% of the total. When treating and disposing of radioactive waste, evaluation will be carried out based on appropriate methods, including evaluation results of secondary pollutants.

JAEA Reports

Technical note for the cavitation damage inspection for interior surface of the mercury target vessel, 1; Development of specimen cutting machine for remote handling

Naoe, Takashi; Kinoshita, Hidetaka; Wakui, Takashi; Kogawa, Hiroyuki; Haga, Katsuhiro

JAEA-Technology 2022-018, 43 Pages, 2022/08

JAEA-Technology-2022-018.pdf:7.84MB

In the liquid mercury target system for the pulsed spallation neutron source of Materials and Life science experimental Facility (MLF) at the Japan in the Japan Proton Accelerator Research Complex (J-PARC), cavitation that is generated by the high-energy proton beam-induced pressure waves, resulting severe erosion damage on the interior surface of the mercury target vessel. The erosion damage is increased with increasing the proton beam power, and has the possibility to cause the leakage of mercury by the penetrated damage and/or the fatigue failure originated from erosion pits during operation. To achieve the long term stable operation under high-power proton beam, the mitigation technologies for cavitation erosion consisting of surface modification on the vessel interior surface, helium gas microbubble injection, double-walled beam window structure has been applied. The damage on interior surface of the vessel is never observed during the beam operation. Therefore, after the target operation term ends, we have cut out specimen from the target nose of the target vessel to inspect damaged surface in detail for verification of the cavitation damage mitigation technologies and lifetime estimation. We have developed the techniques of specimen cutting out by remote handling under high-radiation environment. Cutting method was gradually updated based on experience in actual cutting for the used target vessel. In this report, techniques of specimen cutting out for the beam entrance portion of the target vessel in high-radiation environment and overview of the results of specimen cutting from actual target vessels are described.

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.

JAEA Reports

Experiment of incineration for Trans-Uranic (TRU) wastes containing chlorides

Yamashita, Kiyoto; Yokoyama, Aya*; Takagai, Yoshitaka*; Maki, Shota; Yokosuka, Kazuhiro; Fukui, Masahiro; Iemura, Keisuke

JAEA-Technology 2022-020, 106 Pages, 2022/10

JAEA-Technology-2022-020.pdf:4.77MB

Radioactive solid wastes generated by Fukushima Daiichi Nuclear Power Station disaster may contain high levels of salt from the tsunami and seawater deliberately released into the area. It is assumed that polyvinyl chloride (PVC) products may be used for decommissioning work and for containment of radioactive wastes in the future. Among the method of handling them, incineration is one method that needs to be investigated as it is good method for reduction and stabilization of wastes. But in order to dispose of Trans-Uranic (TRU) solid waste containing chlorides, it is necessary to select the structure and materials of the facility based on the information such as the movement of nuclides and chlorides in the waste gas treating system and the corrosion of equipment due to chlorides. Therefore, we decided to get various data necessary to design a study of the incineration facilities. And we decided to examine the transfer behavior of chlorides to the waste gas treatment system, the corrosion-resistance of materials in the incineration facilities, and the distribution survey of plutonium in them obtained using the Plutonium-contaminated Waste Treatment Facility (PWTF), Nuclear Fuel Cycle Engineering Laboratories, which is a unique incinerating facility in Japan. This report describes the transfer behavior of chlorides in the waste gas treatment system, the evaluation of corrosion-resistance materials and the distribution survey of plutonium in the incineration facilities obtained by these tests using the Plutonium-contaminated Waste Treatment Facility, Nuclear Fuel Cycle Engineering Laboratories.

JAEA Reports

Development of technologies for enhanced analysis accuracy of fuel debris; Summary results of the 2020 fiscal year (Subsidy program for the project of decommissioning and contaminated water management)

Ikeuchi, Hirotomo; Koyama, Shinichi; Osaka, Masahiko; Takano, Masahide; Nakamura, Satoshi; Onozawa, Atsushi; Sasaki, Shinji; Onishi, Takashi; Maeda, Koji; Kirishima, Akira*; et al.

JAEA-Technology 2022-021, 224 Pages, 2022/10

JAEA-Technology-2022-021.pdf:12.32MB

A set of technology, including acid dissolving, has to be established for the analysis of content of elements/nuclides in the fuel debris samples. In this project, a blind test was performed for the purpose of clarifying the current level of analytical accuracy and establishing the alternative methods in case that the insoluble residue remains. Overall composition of the simulated fuel debris (homogenized powder having a specific composition) were quantitatively determined in the four analytical institutions in Japan by using their own dissolving and analytical techniques. The merit and drawback for each technique were then evaluated, based on which a tentative flow of the analyses of fuel debris was constructed.

JAEA Reports

Demonstration of the groundwater observation network system in backfilled underground facility

Murakami, Hiroaki; Takeuchi, Ryuji; Iwatsuki, Teruki

JAEA-Technology 2022-022, 34 Pages, 2022/10

JAEA-Technology-2022-022.pdf:3.47MB

Japan Atomic Energy Agency (JAEA) has been conducting the hydro-pressure and hydrochemical monitoring for more than two decades to understand the hydrochemical disturbance due to the excavation of tunnels at Mizunami Underground Research Laboratory (MIU). To understand the environmental influence due to the backfilling of research tunnels that started in 2019, environmental monitoring of groundwater has been performed and recovery status of groundwater is being confirmed. In order to observe the deep-groundwater environment from the ground, the groundwater pressure monitoring and sampling, which have been performed in the research tunnel, are to be performed from the ground. However, backfilling of a large-scale underground facilities such as MIU is globally unprecedented, thus it was necessary to develop a new observation system. Accordingly, we developed a new observation network to observe the environment around the research tunnels of the MIU. This system enables monitoring of groundwater pressure and water sampling of the backfilled tunnel from the ground while utilizing the existing-monitoring system installed in the tunnels. Accordingly, we demonstrated its technology through the environmental monitoring of groundwater. The results of the environmental monitoring and the existing groundwater data of MIU indicate that this system is able to monitor the groundwater environment in the backfilled tunnels.

JAEA Reports

Optimization of mercury flow with microbubbles in the target-vessel design by means of machine learning

Kogawa, Hiroyuki; Futakawa, Masatoshi; Haga, Katsuhiro; Tsuzuki, Takayuki*; Murai, Tetsuro*

JAEA-Technology 2022-023, 128 Pages, 2022/11

JAEA-Technology-2022-023.pdf:9.0MB

In a mercury target of the J-PARC (Japan Proton Accelerator Research Complex), pulsed proton beams repeatedly bombard the flowing mercury which is confined in a stainless-steel vessel (target vessel). Cavitation damage caused by the propagation of the pressure waves is a factor of the life of the target vessel. As a measure to reduce damages, we developed a bubbler to inject the gas microbubbles into the flowing mercury, which can reduce the pressure waves. To operate the mercury target vessel stably with the 1 MW high-intensity proton beams, further reduction of the damage is required. The bubbler setting position should be closer to the beam window to increase the bubble population, which could enhance the reduction effect on the pressure waves and damage. However, the space at the beam window of the target vessel is restricted. The bubbler design and setting position as well as the vane design for the mercury flowing pattern are optimized by means of a machine learning technique to get more suitable bubble distribution, increasing in bubble population and optimizing bubble size nearby the beam window of the target vessel. The results of CFD analyses performed with 1000 cases were used for machine learning. Since the flow rate of mercury affects the temperature of the target vessel, this was used for the constraint condition. As a result, we found a design of mercury target vessel that can increase the bubble population by ca. 20% higher than the current design.

JAEA Reports

Accident evaluation during transfer of irradiated fuel specimen

Morita, Hisashi; Iimura, Koichi; Matsui, Yoshinori; Takemoto, Noriyuki

JAEA-Technology 2022-024, 73 Pages, 2022/11

JAEA-Technology-2022-024.pdf:1.56MB

JMTR was positioned as a decommissioning facility in the facilities mid- and long-term plan (formulated in April 1, 2017) of the Japan Atomic Energy Agency. On September 18, 2019, we applied for approval of the decommissioning plan for the JMTR reactor facility, and received the approval on March 17, 2021. This made it impossible to conduct irradiation tests with nuclear fuel material at the JMTR using facility. Therefore, on August 7, 2020, in order to delete the description about irradiation test and to change accident evaluation, we applied for change of permission to use nuclear fuel material regarding JMTR facility (Facility No. 1) at the Oarai Research and Development Institute (North Area), and received the permission on May 26, 2021. As the accident evaluation, radiation exposure evaluation was performed at the boundary of the surrounding monitoring area assuming a damage accident during transfer work of the irradiated fuel specimen to the hot laboratory. As a result, it was confirmed to satisfy the standards such as the dose notification concerning about external exposure due to atmospheric diffusion, internal exposure due to atmospheric diffusion, external exposure due to direct $$gamma$$-rays and skyshine $$gamma$$-rays. This report summarizes the methods and results of the accident evaluation related to permission change of JMTR using facility.

JAEA Reports

Skyshine dose evaluation of trench disposal facilities for waste generated from research, industrial and medical facilities

Nakamura, Mizuki; Izumo, Sari; Ogawa, Rina; Nakata, Hisakazu; Amazawa, Hiroya; Sakai, Akihiro

JAEA-Technology 2022-025, 73 Pages, 2022/12

JAEA-Technology-2022-025.pdf:1.64MB

Japan Atomic Energy Agency (JAEA) has been establishing as the implementing body of the near surface disposal of low-level radioactive waste (LLW) generated from research facilities and other facilities in order to actualize the near surface disposal. It is necessary to evaluate the effective doses by direct and skyshine $$gamma$$-rays from disposal facilities and reduce the doses below the target dose (50$$mu$$ Sv/y) at the site boundary for the safety assessment during operation. It was shown at the results of conceptual design that the distance from the trench disposal facilities to site boundary needed to be kept more than 120m in order to satisfy the target dose. However, the design of trench disposal facilities was changed because of increasing amount of waste subject to the trench disposal. Therefore, the dose by skyshine $$gamma$$-rays from trench disposal facilities was recalculated by use of two-dimensional discrete ordinates Sn code DOT 3.5. As a result, it was evaluated that the dose by skyshine $$gamma$$-rays from each trench facility at the site boundary whose distance was 120m from a trench facility was lower than 50$$mu$$ Sv/y, respectively, and the dose added up the doses from trench facilities was also lower than 50$$mu$$ Sv/y. In addition, it was suggested to reduce the target skyshine dose by thickening the covered soil on the top layer.

JAEA Reports

Investigations on distribution of radioactive substances owing to the Fukushima Daiichi Nuclear Power Station Accident in the fiscal year 2021 (Contract research)

Group for Fukushima Mapping Project

JAEA-Technology 2022-026, 152 Pages, 2023/01

JAEA-Technology-2022-026.pdf:20.14MB

This report presents results of the investigations on the distribution-mapping project of radioactive substances owing to TEPCO Fukushima Daiichi Nuclear Power Station (FDNPS) conducted in FY2021. Car-borne surveys, a flat ground measurement using survey meters, a walk survey and an unmanned helicopter survey were carried out to obtain air dose rate data to create air dose rate distribution maps, and temporal changes of the air dose rates were analyzed. Surveys on depth profile of radiocesium and in-situ measurements as for radiocesium deposition were performed. Based on these measurement results, effective half-lives of the temporal changes in the air dose rates and the deposition were evaluated. Score maps to classify the importance of the measurement points were created for Fukushima Prefecture and the 80 km zone from the FDNPS, and the factors causing changes in the score when monitoring data from multiple years were used were discussed. Monitoring data in coastal area performed owing to the comprehensive radiation monitoring plan until 2020 was summarized, and temporal changes in cesium-137 were analyzed. Using the Bayesian hierarchical modeling approach, we obtained maps that integrated the air dose rate distribution data obtained in this project with respect to the region within 80 km from the FDNPS and Fukushima Prefecture. The measurement results for FY2021 were published on the "Expansion Site of Distribution Map of Radiation Dose", and measurement data were stored as CSV format. Radiation monitoring and analysis of environmental samples owing to the comprehensive radiation monitoring plan were carried out.

JAEA Reports

Radiation monitoring via manned helicopter around the Nuclear Power Station in the fiscal year 2021 (Contract research)

Futemma, Akira; Sanada, Yukihisa; Nagakubo, Azusa; Kawasaki, Yoshiharu*; Iwai, Takeyuki*; Hiraga, Shogo*; Haginoya, Masashi*; Matsunaga, Yuki*; Akutsu, Yuichiro*; Urabe, Yoshimi*; et al.

JAEA-Technology 2022-027, 148 Pages, 2023/02

JAEA-Technology-2022-027.pdf:19.64MB

By the nuclear disaster of Fukushima Daiichi Nuclear Power Station (FDNPS), Tokyo Electric Power Company (TEPCO), caused by the Great East Japan Earthquake and the following tsunami on March 11, 2011, a large amount of radioactive material was released from the FDNPS. After the nuclear disaster, airborne radiation monitoring via manned helicopter has been conducted around FDNPS. The results of the airborne radiation monitoring and the evaluation for temporal change of dose rate in the fiscal 2021 were summarized in this report. Analysis considering topographical effects was applied to the result of the airborne monitoring to improve the accuracy of the conventional method. In addition, technique for discriminating gamma rays from the ground and those from the airborne Rn-progenies was also utilized to evaluate their effect on airborne radiation monitoring.

JAEA Reports

Background radiation monitoring via manned helicopter for application of technique of nuclear emergency response in the fiscal year 2021 (Contract research)

Futemma, Akira; Sanada, Yukihisa; Sasaki, Miyuki; Kawasaki, Yoshiharu*; Iwai, Takeyuki*; Hiraga, Shogo*; Haginoya, Masashi*; Matsunaga, Yuki*; Akutsu, Yuichiro*; Hokama, Tomonori; et al.

JAEA-Technology 2022-028, 127 Pages, 2023/02

JAEA-Technology-2022-028.pdf:15.21MB

A large amount of radioactive material was released by the nuclear disaster of Fukushima Daiichi Nuclear Power Station (FDNPS), Tokyo Electric Power Company, caused by the Great East Japan Earthquake and the following tsunami on March 11, 2011. After the nuclear disaster, airborne radiation monitoring via manned helicopter has been utilized to grasp rapidly and widely the distribution of the radioactive materials surrounding FDNPS. We prepare the data of background radiation dose, geomorphic characteristics and the controlled airspace surrounding nuclear facilities of the whole country in order to make effective use of the monitoring technique as a way of emergency radiation monitoring and supply the results during an accident of a facility. This report has summarized the knowledge noted above achieved by the aerial radiation monitoring around Ohi and Takahama nuclear power stations. In addition, the examination's progress aimed at introducing airborne radiation monitoring via an unmanned plane during a nuclear disaster and the technical issues are summarized in this report.

JAEA Reports

Evaluation on secondary radioactive contamination remaining in JMTR Reactor Facility

Nagata, Hiroshi; Otsuka, Kaoru; Omori, Takazumi; Hanakawa, Hiroki; Ide, Hiroshi

JAEA-Technology 2022-029, 55 Pages, 2023/02

JAEA-Technology-2022-029.pdf:2.77MB

Japan Materials Testing Reactor (JMTR) was decided as a one of decommission facilities in April 2017. The activation activity of secondary radioactive contamination remaining in the reactor facility was evaluated in order to submit the decommissioning plan to the Nuclear Regulation Authority. Total activation activity was 2.73$$times$$10$$^{12}$$Bq after 12 years, 1.46$$times$$10$$^{12}$$Bq after 21 years, respectively. The system with high activation activity was the primary cooling system in JMTR. The relatively large radionuclide was H-3, Fe-55, Co-60, Ni-63, Sr-90 and Cs-137. The radioactivity level was classified based on the values of the obtained radioactivity concentration. As a result, the primary cooling system and the drain system was classified as L2, and others was classified as L3. The nuclide that affected classification result was only Co-60 in irradiation facility of HR-1 and OSF-1. H-3, Co-60, Sr-90, Cs-137 and so on were affected classification in other system. When treating and disposing of radioactive waste, evaluation will be carried out based on appropriate methods.

JAEA Reports

Nuclear criticality benchmark analyses on TRIGA-type reactor systems by using continuous-energy Monte Carlo code MVP with JENDL-5

Yanagisawa, Hiroshi; Umeda, Miki; Motome, Yuiko; Murao, Hiroyuki

JAEA-Technology 2022-030, 80 Pages, 2023/02

JAEA-Technology-2022-030.pdf:2.57MB
JAEA-Technology-2022-030(errata).pdf:0.11MB

Nuclear criticality benchmark analyses were carried out for TRIGA-type reactor systems in which uranium-zirconium hydride fuel rods are loaded by using the continuous-energy Monte Carlo code MVP with the evaluated nuclear data library JENDL-5. The analyses cover two sorts of benchmark data, the IEU-COMP-THERM-003 and IEU-COMP-THERM-013 in the International Criticality Safety Benchmark Evaluation Project (ICSBEP) Handbook, and effective neutron multiplication factors, reactivity worths for control rods etc. were calculated by JENDL-5 in comparison with those by the previous version of JENDL. As the results, it was confirmed that the effective neutron multiplication factors obtained by JENDL-5 were 0.4 to 0.6% greater than those by JENDL-4.0, and that there were no significant differences in the calculated reactivity worths by between JENDL-5 and JENDL-4.0. Those results are considered to be helpful for the confirmation of calculation accuracy in the analyses on NSRR control rod worths, which are planned in the future.

JAEA Reports

Development of "MOX weighing and Ball-mill blending" based on experience in operation and maintenance of MOX fuel manufacturing equipment

Kawasaki, Kohei; Ono, Takanori; Shibanuma, Kimikazu; Goto, Kenta; Aita, Takahiro; Okamoto, Naritoshi; Shinada, Kenta; Ichige, Hidekazu; Takase, Tatsuya; Osaka, Yuki; et al.

JAEA-Technology 2022-031, 91 Pages, 2023/02

JAEA-Technology-2022-031.pdf:6.57MB

The document for back-end policy opened to the public in 2018 by Japan Atomic Energy Agency (hereafter, JAEA) states the decommissioning of facilities of Nuclear Fuel Cycle Engineering Laboratories and JAEA have started gathering up nuclear fuel material of the facilities into Plutonium Fuel Production Facilities (hereafter, PFPF) in order to put it long-term, stable and safe storage. Because we planned to manufacture scrap assemblies almost same with Monju fuel assembly using unsealed plutonium-uranium mixed-oxide (hereafter, MOX) powder held in PFPF and transfer them to storage facilities as part of this "concentration" task of nuclear fuel material, we obtained permission to change the use of nuclear fuel material in response to the new regulatory Requirements in Japan for that. The amount of plutonium (which is neither sintered pellets nor in a lidded powder-transport container) that could be handled in the pellet-manufacturing process was limited to 50 kg Pu or less in order to decrease the facility risk in this manufacture. Therefore, we developed and installed the "MOX weighing and blending equipment" corresponding with small batch sizes that functioned in a starting process and the equipment would decrease handling amounts of plutonium on its downstream processes. The failure data based on our operation and maintenance experiences of MOX fuel production facilities was reflected in the design of the equipment to further improve reliability and maintainability in this development. The completed equipment started its operation using MOX powder in February 2022 and the design has been validated through this half-a-year operation. This report organizes the knowledge obtained through the development of the equipment, the evaluation of the design based on the half-a-year operation results and the issues in future equipment development.

JAEA Reports

Development of freezed seal valve by using test stand for LBE technology

Saito, Shigeru; Yamaguchi, Kazushi*; Yoshimoto, Hidemitsu*; Obayashi, Hironari; Sasa, Toshinobu

JAEA-Technology 2022-032, 51 Pages, 2023/03

JAEA-Technology-2022-032.pdf:4.51MB

In the Accelerator Driven System (ADS) being studied by the Japan Atomic Energy Agency (JAEA) for transmutation of long-lived radioactive waste, lead-bismuth eutectic alloy (LBE) is used as a spallation target and subcritical core coolant. A proton irradiation facility in J-PARC is considered to prepare a material irradiation database for ADS development. The proton irradiation facility is equipped with an LBE loop, which enables material irradiation tests in spallation environment under flowing LBE condition. The slow leakage of LBE is one of critical issue to use LBE safety. The slow leakage is caused by the biting slag and/or other materials at valve seal of drain valve. To solve this problem, JAEA examined the application of freeze-seal valve (FSV), which seal the piping by freezing LBE in specific position. Water-cooled and air-cooled freeze-seal valve test modules were fabricated, installed in the test section of the existing test stand for LBE technology development, and tested to confirm their operation and performance. As a result of the tests, it was confirmed that the water-cooled FSV test module worked well along to the design values. This report describes the outline and details of the test stand for LBE technology and each FSV test module, as well as the results of operation and performance verification tests.

JAEA Reports

Decommissioning of uranium handling facility for development of nuclear fuel manufacturing equipment

Kageyama, Tomio; Denuma, Akio; Koizumi, Jin*; Odakura, Manabu*; Haginoya, Masahiro*; Isaka, Shinichi*; Kadowaki, Hiroyuki*; Kobayashi, Shingo*; Morimoto, Taisei*; Kato, Yoshiaki*; et al.

JAEA-Technology 2022-033, 130 Pages, 2023/03

JAEA-Technology-2022-033.pdf:9.87MB

Uranium handling facility for development of nuclear fuel manufacturing equipment (Mockup room) was constructed in 1972. The Mockup room has a weak seismic resistance and is deteriorating with age. Also, the original purpose with this facility have been achieved and there are no new development plans using this facility. Therefore, interior equipment installed in this facility had been dismantled and removed since March 2019. After that, the Mockup room was inspected for contamination, and then controlled area in the Mockup room was cancelled on March 29th 2022. A total of 6,549 workers (not including security witnesses) were required for this work. The amount of non-radioactive waste generated by this work was 31,300 kg. The amount of radioactive waste generated by this work was 3,734 kg of combustible waste (103 drums), 4,393 kg of flame resistance waste (61 drums), 37,790 kg of non-combustible waste (124 drums, 19 containers). This report describes about the dismantling and removing the interior equipment in the Mockup room, the amount of waste generated by this work, and procedure for cancellation the controlled area in the facility.

JAEA Reports

R&D and maintenance management of the WASTEF Facility (FY2021)

Sano, Naruto; Yamashita, Naoki; Hoshino, Kazutoyo*; Tsukada, Manabu*; Sawauchi, Fumiya*; Otake, Yoshinori; Ichise, Kenichi; Tagami, Susumu

JAEA-Technology 2022-034, 47 Pages, 2023/03

JAEA-Technology-2022-034.pdf:2.81MB

The Waste Safety Testing Facility (WASTEF) was established in 1982 as an experimental facility for long-term storage of solidified high-level radioactive waste generated in the reprocessing of spent light water reactor fuel and the subsequent safety assessment of geological disposal. It is a historic facility that started operation in 1982. This facility consists of 5 concrete cells, 1 lead cell, 6 glove boxes, and 7 hoods, and is a large-scale facility that can use nuclear fuel materials including uranium and plutonium and radioactive isotopes including TRU. In this facility, research and development requested by the research department is carried out in the Hot Material Examination Section. In addition, patrol inspections, self-inspections, etc. are also carried out as maintenance management based on safety regulations. This report summarizes the overview of WASTEF facilities, the results of operation, maintenance and management work in FY2021, and the future outlook.

JAEA Reports

Development of a test method for electrochemical measurements of stainless steel in nitric acid solution containing neptunium-237 under gamma-ray irradiation

Yamashita, Naoki; Irisawa, Eriko; Kato, Chiaki; Sano, Naruto; Tagami, Susumu

JAEA-Technology 2022-035, 29 Pages, 2023/03

JAEA-Technology-2022-035.pdf:2.54MB

In the treatment process of the current commercial reprocessing plant (Rokkasho Reprocessing Plant), the high-level liquid waste concentrator is the equipment that treats the most corrosive solution. In the high-level liquid waste concentrator, the extracted liquid waste after separation of uranium and plutonium is heated, concentrated, and reduced in volume. Therefore, the amount of gamma- rays emitted from fission products and the concentration of corrosive metal ion species such as neptunium-237 ($$^{237}$$Np) are the highest in the reprocessing process, and the amount of corrosion in the high-level liquid waste concentrate canner is expected to be large. In this study, in order to clarify the effect of gamma-rays on the corrosion reaction of stainless steel in nitric acid solutions containing $$^{237}$$Np from the electrochemical viewpoint, the corrosion test apparatus for heat transfer surfaces in an airtight concrete cell at the Waste Safety TEsting Facility (WASTEF) of Nuclear Science Research Institute was modified to enable electrochemical measurements under gamma-ray irradiation. The effect of gamma-rays on the corrosion reaction taking place on the stainless steel surface was discussed from the electrochemical test results obtained. As a result, changes in the immersion potentials of stainless steel and the polarization curves due to chemical species caused by radiolysis of gamma-ray irradiation were confirmed.

JAEA Reports

Precautions of capacitor inspection and its treatment based on the PCB Special Measures Law

Ono, Ayato; Takayanagi, Tomohiro; Sugita, Moe; Ueno, Tomoaki*; Horino, Koki*; Yamamoto, Kazami; Kinsho, Michikazu

JAEA-Technology 2022-036, 31 Pages, 2023/03

JAEA-Technology-2022-036.pdf:8.77MB

In the Japan Atomic Energy Agency (JAEA), many electrical facilities such as power receiving equipment and power supply units are installed in experimental facilities such as the Nuclear Science Research Institute (NSRI) and the Japan Proton Accelerator Research Complex (J-PARC). However, some facilities have been in operation for more than half a century since they were manufactured, some have already been closed or deactivated, and others are still in operation while replacing parts and taking other aging measures. In these facilities, materials that were used because of their excellent properties at the time of manufacture are now designated as hazardous substances and require special management when disposed of. One of them is polychlorinated biphenyl (PCB). PCB were used in a very wide range of fields because of their stability against heat, high electrical insulation, and chemical resistance. However, it was found that PCB have persistent properties and may cause damage to human health and the living environment, and the government has enacted the "Act on Special Measures for Promotion of Proper Treatment of PCB Wastes (PCB Special Measures Law)" to promote reliable and proper disposal. JAEA has almost completed the excavation survey of high-concentration PCB waste and is in the process of excavating low-concentration PCB waste. However, there are still new relevant items to be discovered. This report summarizes and reports the knowledge necessary for identifying PCB waste and points to be noted when handling capacitors, etc., based on examples of actual disassembly and investigation work conducted on power supply units and other electrical equipment, such as capacitors attached to power supply units, etc.

JAEA Reports

Development of bearing units used in the rotating shaft system for power generation in the JAEA-Tokai Tandem accelerator

Otokawa, Yoshinori; Matsuda, Makoto; Abe, Shinichi

JAEA-Technology 2022-037, 23 Pages, 2023/03

JAEA-Technology-2022-037.pdf:5.38MB

Bearing units of rotating shaft system for power generation in the JAEA-Tokai Tandem accelerator had a short operating life from the beginning of the accelerator installation, and even after replacing the bearing unit, early failures often occurred. Therefore, the quantity and frequency of replacement are large, and a lot of time is spent in regular maintenance performed by opening the accelerator pressure vessel, and solving this problem has been a long-standing concern. As a result of considering the cause of this early failure, it was considered that the load caused by parallel, angular, and axial misalignment of the upper and lower bearing units was the cause. In order to solve this problem, we have developed a bearing unit having metal disc-spring couplings in the upper and lower flanges to allow for parallel, angular, and axial misalignment of bearing units. As a result, it is now possible to allow variations in the distance between castings and parallel, angular, and axial misalignment of the upper and lower bearing units. By installing the developed new bearing unit on the rotating shaft system and making improvements while continuing to use it in actual operation, we succeeded in reducing the early failure and extending the operating life by more than double. With this development, the maintenance time has been reduced to one week by reducing the number of replacing the bearing unit. In addition, we have realized one regular maintenance that was carried out about three times a year, and as a benefit, we were able to reduce the amount of sulfur hexafluoride (SF$$_{6}$$) gas, which is a greenhouse gas, to about 33$$sim$$50% per year. We describe about development of new bearing units and these maintenance status from 2006 to 2020.

JAEA Reports

Historical changes and Correspondence to Research and Test Reactors New Regulatory Standards for Monitoring Post in Oarai Research and Development Institute, JAEA

Hamaguchi, Takumi; Yamada, Junya; Komatsuzaki, Naoya*; Hatakeyama, Takumi; Seya, Natsumi; Muto, Yasunobu; Miyauchi, Hideaki; Hashimoto, Makoto

JAEA-Technology 2022-038, 65 Pages, 2023/03

JAEA-Technology-2022-038.pdf:4.3MB

New regulatory requirements were developed taking into account the lessons-learnt from the accident at Fukushima Daiichi Nuclear Power Station on March 2011. The new regulatory standards required that monitoring posts should be diversified in transmission systems and equipped with backup power supply equipment for design basis accidents. In this report, we look back on the history of monitoring posts in Oarai Research and Development Institute, explained the application for the permission of reactor installment license, application for approval of the design and construction method, pre-use operator's inspection and improvement design of monitoring posts. This report also includes about inspection based on act on special measures concerning nuclear emergency preparedness and the installation of KURAMA-II, which was carried out in conjunction with the improvement of monitoring post for new regulatory standards. As an appendix, application document for approval of the design and construction method are included.

JAEA Reports

Decommissioning state of Plutonium Fuel Fabrication Facility; Dismantling the glove box W-9 and equipment interior, and a part of tunnel F1

Nagai, Yuya; Shuji, Yoshiyuki; Kawasaki, Takeshi; Aita, Takahiro; Kimura, Yasuhisa; Nemoto, Yasunori*; Onuma, Takeshi*; Tomiyama, Noboru*; Hirano, Koji*; Usui, Yasuhiro*; et al.

JAEA-Technology 2022-039, 117 Pages, 2023/06

JAEA-Technology-2022-039.pdf:11.96MB

Japan Atomic Energy Agency (JAEA) manages wide range of nuclear facilities. Many of these facilities are required to be performed adjustment with the aging and complement with the new regulatory standards and the earthquake resistant, since the Great East Japan Earthquake and the Fukushima Daiichi Nuclear Power Station accident. It is therefore desirable to promote decommissioning of facilities that have reached the end of their productive life in order to reduce risk and maintenance costs. However, the progress of facility decommissioning require large amount of money and radioactive waste storage space. In order to address these issues, JAEA has formulated a "The Medium/Long-Term Management Plan of JAEA Facilities" with three pillars: (1) consolidation and prioritization of facilities, (2) assurance of facility safety, and (3) back-end countermeasures. In this plan, Plutonium Fuel Fabrication Facility has been selected as primary decommissioned facility, and dismantling of equipment in the facilities have been underway. In this report, size reduction activities of the glove box W-9 and a part of tunnel F-1, which was connected to W-9, are presented, and the obtained findings are highlighted. The glovebox W-9 had oxidation & reduction furnace, and pellet crushing machine as equipment interior. The duration of activity took six years from February 2014 to February 2020, including suspended period of 4 years due to the enhanced authorization approval process

JAEA Reports

Irradiation test using foreign reactor, 1; Study of irradiation test with capsule temperature control system (Joint research)

Takabe, Yugo; Otsuka, Noriaki; Fuyushima, Takumi; Sayato, Natsuki; Inoue, Shuichi; Morita, Hisashi; Jaroszewicz, J.*; Migdal, M.*; Onuma, Yuichi; Tobita, Masahiro*; et al.

JAEA-Technology 2022-040, 45 Pages, 2023/03

JAEA-Technology-2022-040.pdf:6.61MB

Because of the decommission of the Japan Materials Testing Reactor (JMTR), the domestic neutron irradiation facility, which had played a central role in the development of innovative nuclear reactors and the development of technologies to further improve the safety, reliability, and efficiency of light water reactors, was lost. Therefore, it has become difficult to pass on the operation techniques of the irradiation test reactors and irradiation technologies, and to train human resources. In order to cope with these issues, we conducted a study on the implementation of irradiation tests using overseas reactors as neutron irradiation sites as an alternative method. Based on the "Arrangement between the National Centre for Nuclear Research and the Japan Atomic Energy Agency for Cooperation in Research and Development on Testing Reactor," the feasibility of conducting an irradiation test at the MARIA reactor (30 MW) owned by the National Centre for Nuclear Research (NCBJ) using the temperature control system, which is one of the JMTR irradiation technologies, was examined. As a result, it was found that the irradiation test was possible by modifying the ready-made capsule manufactured in accordance with the design and manufacturing standards of the JMTR. After the modification, a penetration test, an insulation continuity test, and an operation test in the range of room temperature to 300$$^{circ}$$C, which is the operating temperature of the capsule, were conducted and favorable results were obtained. We have completed the preparations prior to transport to the MARIA reactor.

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