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JAEA Reports

Evaluation of nuclides migration for trench-type disposal by a calculation method taking leaching into consideration

Totsuka, Masayoshi; Kurosawa, Ryohei*; Sakai, Akihiro; Nakata, Hisakazu; Hayashi, Hirokazu; Amazawa, Hiroya

JAEA-Technology 2017-001, 40 Pages, 2017/03

JAEA-Technology-2017-001.pdf:2.24MB

Japan Atomic Energy Agency is planning for the near surface disposal of low level radioactive wastes generated from research, industrial and medical facilities industry in Japan. This document provides the values of radioactivity concentrations equivalent to dose criterion for trench-type disposal. These values are derived based on the safety assessment for ground water scenarios by using a model which describes the release of radionuclides from wastes to a cover soil caused by elution. These concentrations are compared with the one calculated by a model that describes the nuclide release mechanisms as solid-liquid partitioning equilibrium. Additionally, the change in the concentrations is evaluated when the amount of water percolating into a disposal facility varies.

JAEA Reports

Rearrangement works of unbalanced waste packages by influence of the Great East Japan Earthquake

Ishihara, Keisuke; Kanazawa, Shingo; Kozawa, Masachiyo; Mori, Masakazu; Kawahara, Takahiro

JAEA-Technology 2017-002, 27 Pages, 2017/03

JAEA-Technology-2017-002.pdf:21.88MB

At radioactive waste management facilities in the Nuclear Science Research Institute, solid radioactive wastes are stored by using containers such as 200L drums and pallets to tier containers in 2 to 4 stacks in the height direction in waste storage facilities (Waste Storage Facility No.1, Waste Storage Facility No.2 and Waste Size Reduction and Storage Facility). On March 11, 2011, the Great East Japan Earthquake was happened, and some waste packages dropped from their pallets and large number of waste packages moved from their original position and inclined due to the influence of the earthquake in the waste storage facilities. There was no experience of rearrangement works to set those dropped and unbalanced waste packages in their original position and it was necessary to prepare detailed work procedures and progress for this task to prevent the occurrence of industrial accidents. Therefore, we prepared detailed work manual and repeatedly carried out mock-up test. And then, we started rearrangement work from April 2011 after confirmation of workers skill and adequacy of the work manual. Finally, all rearrangement works for stored waste packages took about four and half years and were completed in September 2015 without any accident and shutdown of storage function. This report summarizes the countermeasures to reduce exposure doses of workers and to prevent the occurrence of industrial accidents during the rearrangement works.

JAEA Reports

Technical design report on J-PARC Transmutation Experimental Facility; ADS Target Test Facility (TEF-T)

Nuclear Transmutation Division, J-PARC Center

JAEA-Technology 2017-003, 539 Pages, 2017/03

JAEA-Technology-2017-003.pdf:59.1MB

JAEA is pursuing R&D on volume reduction and mitigation of degree of harmfulness of high-level radioactive waste based on the "Strategic Energy Plan" issued in April 2014. Construction of Transmutation Experimental Facility is under planning as one of the second phase facilities in the J-PARC program to promote R&D on the transmutation technology with using accelerator driven systems (ADS). The TEF consists of two facilities: ADS Target Test Facility (TEF-T) and Transmutation Physics Experimental Facility (TEF-P). Development of spallation target technology and study on target materials are to be conducted in TEF-T with impinging a high intensity proton beam on a lead-bismuth eutectic target. Whereas in TEF-P, by introducing a proton beam to minor actinide loaded subcritical cores, physical properties of the cores are to be studied, and operation experiences are to be acquired. This report summarizes results of technical design for construction of one of two TEF facilities, TEF-T.

JAEA Reports

Confirmation of feasibility of fabrication technology and characterization of high-packing fraction fuel compact for HTGR

Mizuta, Naoki; Ueta, Shohei; Aihara, Jun; Shibata, Taiju

JAEA-Technology 2017-004, 22 Pages, 2017/03

JAEA-Technology-2017-004.pdf:2.71MB

Confirmation of feasibility of fabrication technology and characterization of the high-packing fraction fuel compact of High Temperature Gas Reactor (HTGR) fuel were carried out. Fuel compacts were fabricated with CFP packing fraction targeted at 33 percent by the same manufacturing condition of HTTR fuel compact. SiC-defective fraction, compressive strength and internal CFP distribution of the compact, important parameters to guarantee its integrity, were evaluated. The high-packing fuel compacts showed as same level of SiC-defective fraction as that of HTTR first loading fuel, 8$$times$$10$$^{-5}$$, and larger compressive strength than the HTTR fuel criteria, 4,900N. The feasibility of fabrication technology and the performance for the high-packing fraction fuel compact was confirmed.

JAEA Reports

Study on engineering technologies in the Mizunami Underground Research Laboratory (FY 2015); Development of design and construction planning and countermeasure technologies (Contract research)

Toguri, Satohito*; Kobayashi, Shinji*; Tsuji, Masakuni*; Yahagi, Ryoji*; Yamada, Toshiko*; Matsui, Hiroya; Sato, Toshinori; Mikake, Shinichiro; Aoyagi, Yoshiaki

JAEA-Technology 2017-005, 43 Pages, 2017/03

JAEA-Technology-2017-005.pdf:4.4MB

The study on engineering technology in the Mizunami Underground Research Laboratory (MIU) project roughly consists of (1) development of design and construction planning technologies, (2) development of construction technology, (3) development of countermeasure technology, (4) development of technology for security, and (5) development of technologies regarding restoration and mitigating of the excavation effect. In FY2015, as a part of the important issues on the research program, water-tight grouting method has been developed. Grouting methods utilized in the MIU were evaluated and the post-excavation grouting at the -500m Access/Research Gallery-South was planned based on these evaluation results. Also, technology development from the viewpoint of geological disposal was summarized, and information on the alternative method to the grouting method was collected and organized.

JAEA Reports

Examination of decontamination of various materials at houses in difficult-to-return zone

Mori, Airi; Tanabe, Tsutomu; Wada, Takao; Kato, Mitsugu

JAEA-Technology 2017-006, 38 Pages, 2017/03

JAEA-Technology-2017-006.pdf:2.98MB

Large quantities of radioactive materials were released into the environment as a result of the Fukushima Daiichi Nuclear Power Station accident. Residential areas and forest areas near the power station were contaminated with the radioactive materials. Outside of the houses, schools and the other buildings are being decontaminated by national authority and local government. On the other hand, the materials (such as walls, floors, or windows) which constitute the houses are not decontaminated officially. In order to prepare decontamination methods that can be applied easily, we conducted examinations of decontamination for various materials in houses. Fibrous materials, woods, glasses, concretes, plastics, vinyl chloride materials, metals and synthetic leathers were used in our examinations. These materials were collected from houses in difficult-to-return zone, and were contaminated by radioactive materials released by the accident. Dry methods (suction, wiping, adsorption and peelable coating), wet methods (wiping, brushing, polishing and washing) and physical method (peeling of materials) were used for decontamination. As a result of our examinations, materials with low water permeability, such as glasses, concretes, vinyl chloride materials and metals, were able to be decontaminated efficiently (about 90% reduction) by using wet methods. Materials with high water permeability like woods were relatively well decontaminated by peelable coating (about 60%-70% reduction). In addition to the examination described above, the difference of contamination reduction effect between chemical properties of detergents and the effect of rubbing of peelable coating were also examined. Finally, the most effective method was summarized based on these examinations.

JAEA Reports

Criticality safety evaluation of the fresh fuel storage in NSRR; Under consideration of earthquake and tsunami occurrence

Motome, Yuiko; Murao, Hiroyuki

JAEA-Technology 2017-007, 18 Pages, 2017/03

JAEA-Technology-2017-007.pdf:2.16MB

Nuclear Safety Research Reactor (NSRR) facility have been utilized for fuel irradiation experiments to study the behaviors of nuclear fuels under reactivity initiated accident conditions. Unirradiated test fuels used in fuel irradiation experiments and flesh driver fuel elements for reactor operation are stored in the fuel building of the facility. In response to the 2011 off the Pacific coast of Tohoku Earthquake, the impact of NSRR's nuclear fuel material usage facilities on external events beyond design requirements was evaluated. The subcriticality of the flesh fuel storage was confirmed in consideration of earthquake and tsunami as superimposed event.

JAEA Reports

Report on analytical activities in potentially hazardous materials mitigation measures at the Plutonium Conversion Development Facility; 2015.12 $$sim$$ 2016.10

Horigome, Kazushi; Taguchi, Shigeo; Ishibashi, Atsushi; Inada, Satoshi; Kuno, Takehiko; Surugaya, Naoki

JAEA-Technology 2017-008, 14 Pages, 2017/05

JAEA-Technology-2017-008.pdf:1.15MB

The plutonium solution had been converted into MOX powder to mitigate the potential hazards of storage plutonium solution such as hydrogen generation at the Plutonium Conversion Development Facility. The plutonium conversion operations had been started in April, 2014, and had been finished in July, 2016. With respect to the samples taken from the conversion process, about 2,200 items of plutonium/uranium solutions and MOX powders had been analyzed for the operation control in the related analytical laboratories at the Tokai Reprocessing Plant. This paper describes the reports on analytical activities and related maintenance works in the analytical laboratories conducted from December, 2015 to October, 2016.

JAEA Reports

Gas-strip methods of dissolved inorganic carbon in groundwater for radiocarbon analysis

Kato, Toshihiro; Iwatsuki, Teruki; Nishio, Tomohiro*

JAEA-Technology 2017-009, 30 Pages, 2017/06

JAEA-Technology-2017-009.pdf:2.6MB

Groundwater age is an important information to infer the groundwater flow. The radiocarbon ($$^{14}$$C) dating of the groundwater is primary method for the evaluation of groundwater flow. The carbon in the groundwater generally exist as a dissolved inorganic carbon (DIC). Though DIC in groundwater samples is usually collected by chemical precipitation method, the method requires lots of preparation to sample the carbon. Furthermore there are problems with the reproducibility on precipitation and measurement value. This study newly examined the application of gas-strip method to collect DIC in groundwater sample by using JAEA-made gas-strip system. The performance of the CO$$_{2}$$ gas-stripping from groundwater and the influence of sulfide are investigated. Based on these results, the operation procedures of gas-strip system and preparation method for the groundwater samples were summarized in this report.

JAEA Reports

Study on the evaluation method to determine the radioactivity concentration in radioactive waste generated from post-irradiation examination facilities, 2

Tsuji, Tomoyuki; Hoshino, Yuzuru; Sakai, Akihiro; Sakamoto, Yoshiaki; Suzuki, Yasuo*; Machida, Hiroshi*

JAEA-Technology 2017-010, 75 Pages, 2017/06

JAEA-Technology-2017-010.pdf:2.31MB

It is necessary for reasonable disposal to be studied on evaluation methods to determine radioactivity concentrations in the radioactive wastes, which is generated from post-irradiation examination (PIE) facilities, for establishment of reasonable confirmation methods concerning radioactive wastes generated from research, industrial, and medical facilities. It has been chosen the PIE facilities of NUCLEAR DEVELOPMENT CORPORATION as a model for this study. As a result, it has been confirmed that the theoretical methods are applied for the important nuclides (H-3, C-14, Co-60, Ni-63, Sr-90, Tc-99, Cs-137, Eu-154, U-234, U-235, U-238, Pu-238, Pu-239, Pu-240, Pu-241, Am-241 and Cm-244).

JAEA Reports

Backfilling test in drilling pits as part of Groundwater REcovery Experiment in Tunnel (GREET) Project

Takayasu, Kentaro; Onuki, Kenji*; Kawamoto, Koji*; Takayama, Yusuke; Mikake, Shinichiro; Sato, Toshinori; Onoe, Hironori; Takeuchi, Ryuji

JAEA-Technology 2017-011, 61 Pages, 2017/06

JAEA-Technology-2017-011.pdf:9.15MB

The Groundwater REcovery Experiment in Tunnel (GREET) was put into effect as development of drift backfilling technologies. This test was conducted by making the Closure Test Drift (CTD) recovered with water after carrying out a plug around 40m distance from northern edge face of horizontal tunnel of depth 500m, for the purpose of investigation of recovering process of rock mass and groundwater under the influence of excavation of tunnel. This report presents the efforts of backfilling investigation using bentonite composite soil and execution of backfilling into borehole pits excavated in the CTD which were carried out in fiscal 2014 as a part of GREET, and succeeding observation results inside pits from September 2014 to March 2016.

JAEA Reports

Development of temperature measurement technology for control rod using melt wire in High Temperature engineering Test Reactor (HTTR)

Hamamoto, Shimpei; Sawahata, Hiroaki; Suzuki, Hisashi; Ishii, Toshiaki; Yanagida, Yoshinori

JAEA-Technology 2017-012, 20 Pages, 2017/06

JAEA-Technology-2017-012.pdf:7.9MB

A melt wire was installed at the tip of the control rod in order to measure the temperature of High Temperature engineering Test Reactor (HTTR). After experience with reactor scram from the state of reactor power 100%, the melt wire was taken out from the control rod and appearance has been observed visually. In this study, an exclusive device for taking out the melt wire was prepared. The take-out device functions as expected, and the melt wire was safely and reliably taken out using a remote manipulator. And because the visual observation of the melt wire was clearly carried out, we were successful in developing the control rod temperature measurement technology. It was confirmed that the melt wires with a melting point of 505$$^{circ}$$C or less were melted, and the melt wires with a melting point of 651$$^{circ}$$C or more were not melted. Therefore, it was found that the highest arrival temperature of tip of the control rods where the melt wires are installed reaches within the range of 505 to 651$$^{circ}$$C. And it was found that the control rod temperature at the time of reactor scram does not exceed the using temperature criteria (900$$^{circ}$$C) of Alloy 800H of the control rod sleeve.

JAEA Reports

Applicability confirmation test of optimum decay heat evaluation method for HTGR with HTTR (Non-nuclear heating test); Validation of residual heat evaluation model

Honda, Yuki; Inaba, Yoshitomo; Nakagawa, Shigeaki; Yamazaki, Kazunori; Kobayashi, Shoichi; Aono, Tetsuya; Shibata, Taiju; Ishitsuka, Etsuo

JAEA-Technology 2017-013, 20 Pages, 2017/06

JAEA-Technology-2017-013.pdf:2.52MB

Decay heat is one of an important factor for a safety evaluation of depressurized loss-of-forced cooling accident, a representative high consequence accident, in high temperature gas-cooled reactor (HTGR). Traditionally, a conservative decay heat curve is used for safety analysis according to the regulatory standards. On the other hand, there is growing interest in obtaining test data related to decay heat for the use of uncertainty analysis. However, such data has not been obtained for prismatic-type HTGR. Therefore, we have launched a test program to obtain the decay heat data from the HTTR. As an initial step, an applicability confirmation test of decay heat evaluation method for HTGR was conducted in February 2017 without non-nuclear heating condition. This report introduces an estimation method for the decay heat based on test data using HTTR and shows the results of validation of the reactor residual heat evaluation method which will be used to obtain the decay heat data based on test data.

JAEA Reports

Application of controlled-potential coulometry as a primary method for the characterization of plutonium nitrate solutions being used for reference materials (Joint research)

Yamamoto, Masahiko; Holland, M. K.*; Cordaro, J. V.*; Kuno, Takehiko; Surugaya, Naoki

JAEA-Technology 2017-014, 63 Pages, 2017/06

JAEA-Technology-2017-014.pdf:4.38MB

In this study, the controlled-potential coulometry has been applied as a primary method for characterizing the Pu master solutions being used as alternative source material for IDMS spikes. The coulometry system compliance with ISO12183 has been used for measurement. It has been calibrated using equipment traceable to the SI units. Plutonium standard samples have been measured to confirm the accuracy. The relative standard deviation is below 0.05%. The results agree with the reference value within $$pm$$0.05%. It is found that the Pu can be precisely analyzed by the coulometry system. Then, the Pu nitrate solution, which has been purified from mixed oxide powder containing relatively high $$^{239}$$Pu, has been measured. The relative standard deviation is below 0.05%. The relative expanded uncertainty is less than 0.074% at the 95% confidence interval (k=2). It is indicated that coulometric assay of Pu is fit for the purpose of characterizing reference materials.

JAEA Reports

Diffusion experiment using block sample of the Toki granite

Hama, Katsuhiro; Iwasaki, Riyo*; Morikawa, Keita*

JAEA-Technology 2017-015, 45 Pages, 2017/07

JAEA-Technology-2017-015.pdf:16.57MB

Tono Geoscience Center of Japan Atomic Energy Agency has been carrying out the Mizunami Underground Research Laboratory Project. The goal of mass transport study is to obtain a better understanding of mass transport phenomena in the geological environment as well as to develop technologies for measurement of the mass transport parameters, model construction, numerical analysis and validation of those technologies. This experiment was planned to understand the influence of the microscopic structure in the rock mass on the mass transport property. The diffusion experiment using rock sample was carried out. The macroscopic and microscopic observations were carried out to understand the distribution of tracer (uranine) after the diffusion experiment. The uranine was observed in the plagioclase, in the grain boundary and in the microfracture in the mineral grains. These results suggested that distribution of mineral and of microfracture could affect the diffusion property of uranine.

JAEA Reports

Preparation of uranium and plutonium mixed spike optimized for MOX analysis by isotope dilution mass spectrometry

Horigome, Kazushi; Taguchi, Shigeo; Yamamoto, Masahiko; Kuno, Takehiko; Surugaya, Naoki

JAEA-Technology 2017-016, 20 Pages, 2017/07

JAEA-Technology-2017-016.pdf:1.68MB

Mixed spikes of uranium and plutonium have been prepared for the determination of uranium and plutonium in dissolved MOX solution by isotope dilution mass spectrometry. Enriched uranium metal NBL CRM116 and plutonium metal NBL CRM126 were accurately weighed and then dissolved in nitric acid, respectively. Their dissolved solutions were mixed in a mass ratio of 1 to 2. The preparation values of uranium and plutonium were 1.0530 $$pm$$ 0.0008 mg/g (k=2) of uranium with a $$^{235}$$U relative mass fraction of 93.114 wt% and 2.0046 $$pm$$ 0.0019 mg/g (k=2) of plutonium with a $$^{239}$$Pu relative mass fraction of 97.934 wt%, respectively. The concentrations of uranium and plutonium in spike were confirmed by reverse isotope dilution mass spectrometry using tracer of $$^{233}$$U and $$^{242}$$Pu. Finally, the prepared spike was validated by parallel analysis of simulated sample of dissolved MOX solution. This spike was applied to measure the uranium and plutonium amount content of dissolved MOX solutions using isotope dilution mass spectrometry.

JAEA Reports

Clearance of concrete generated from modification activities of JRR-3; Results for measuring and evaluating radioactivity concentration

Ogoshi, Yurie; Satoyama, Tomonori; Kishimoto, Katsumi; Nanri, Tomohiro; Suzuki, Takeshi; Tomioka, Osamu; Takaizumi, Hirohide*; Kanno, Tomoyuki*; Maruyama, Tatsuya*

JAEA-Technology 2017-017, 152 Pages, 2017/08

JAEA-Technology-2017-017.pdf:15.97MB

At Nuclear Science Research Institute, clearance works for about 4,000 tons of extremely low-level radioactive concrete debris, which were generated from the modification activities of JRR-3 from FY 1985 to FY 1989 and stored in the waste storage facility NL, carried out. First of this clearance works, method for measuring and evaluating radioactivity concentration was approved by Minister of MEXT on July 25, 2008. And then, clearance works were started from FY 2009. Measuring and evaluating radioactivity concentration was achieved by using the approved method, and was confirmed by government. And then, clearance works were completed in FY 2014. The clearance concrete was recycled as a material for restoration works of the 2011 off the Pacific coast of Tohoku Earthquake. This report summarizes the results of measuring and evaluating radioactivity concentration, achievement of confirmation by government, recycling of cleared concrete and cost for clearance works.

JAEA Reports

Phenomenon elucidation experiment for target wastage caused in steam generator of sodium-cooled fast reactor; Corrosion experiment in flowing high-temperature sodium hydroxide environment

Umeda, Ryota; Shimoyama, Kazuhito; Kurihara, Akikazu

JAEA-Technology 2017-018, 70 Pages, 2017/08

JAEA-Technology-2017-018.pdf:9.67MB

In case of the water leak into sodium in a SG of SFRs due to tube failure, reaction jet is formed by sodium-water reaction with exothermic heat. The reaction jet forms highly alkaline environment with high temperature and high pressure, which cause local thinning of adjacent heat transfer tubes (target wastage). In this report, for the purpose of elucidation of target wastage, the authors developed the experimental apparatus and experimental technique which enable the separate evaluation of wastage influence factors, including temperature, impingement velocity, reagent ratio and so on by using high temperature sodium hydroxide as major reaction product and sodium monoxide as secondary reaction product. In addition, the impingement corrosion experiments have been conducted by using high temperature reagents (NaOH and Na$$_{2}$$O). Based on the corrosive data, authors quantitatively evaluated the influence factors of wastage and formulated the average corrosive equations.

JAEA Reports

Analysis of meteorological observation data for the atmospheric diffusion calculation; FY2005-2015

Nishimura, Tomohiro; Onuma, Toshimitsu*; Mizutani, Tomoko; Nakano, Masanao

JAEA-Technology 2017-019, 60 Pages, 2017/09

JAEA-Technology-2017-019.pdf:3.2MB

The meteorological observation has been performed since 1969's in the Nuclear Fuel Cycle Engineering Laboratories, JAEA after 1974. The meteorological observation data has been applied for the calculation of the atmospheric diffusion of radioactive wastes since the hot run was started 1977. This report presents statistical results of meteorological observation based on the decadal data from fiscal year 2005 to 2015. The characteristics of atmospheric diffusion related to the meteorological parameter are also discussed in this report.

JAEA Reports

System analysis for HTTR-GT/H$$_{2}$$ plant; Safety analysis of HTTR for coupling helium gas turbine and H$$_{2}$$ plant

Sato, Hiroyuki; Yan, X.; Ohashi, Hirofumi

JAEA-Technology 2017-020, 23 Pages, 2017/08

JAEA-Technology-2017-020.pdf:1.23MB

JAEA initiated a nuclear cogeneration demonstration project with helium gas turbine power generation and thermochemical hydrogen production utilizing the HTTR. This study carries out system analysis for the HTTR gas turbine hydrogen cogeneration test plant. The evaluation was conducted for the events newly identified corresponding to the coupling of helium gas turbine and hydrogen production plant to the HTTR. The results showed that loss of load event does not have impact on temperature of fuel and reactor coolant pressure boundary. In addition, reactor coolant pressure does not exceed the evaluation criteria. Furthermore, it was shown that reactor operation can be maintained against temperature transients induced by abnormal events in hydrogen production plant.

JAEA Reports

Investigation and measures of abnormal events of helium refrigerator for cryogenic hydrogen system at J-PARC

Aso, Tomokazu; Teshigawara, Makoto; Hasegawa, Shoichi; Aoyagi, Katsuhiro*; Muto, Hideki*; Nomura, Kazutaka*; Takada, Hiroshi; Ikeda, Yujiro

JAEA-Technology 2017-021, 75 Pages, 2017/08

JAEA-Technology-2017-021.pdf:33.03MB

Liquid hydrogen is employed as a cold neutron moderator material at the spallation neutron source of Materials and Life science experimental Facility of Japan Proton Accelerator Research Complex (J-PARC). From January 2015, it became observable that the differential pressure between heat exchangers and an 80 K adsorber (ADS) in a helium refrigerator system increased with operating time. In November 2015, the differential pressure rise became more significant, leading to degrade the refrigerating performance in cooling liquid hydrogen. In order to investigate the cause of the abnormal differential pressure rise between the heat exchangers and the ADS, we carried out visual inspection inside the heat exchangers and analyzed the impurities contained in the helium gas. Unfortunately, we could not identify the impurities causing the performance degradation, but observed a trace of oil in the inlet piping of the heat exchanger. Based on investigations of the abnormal events occurred in the refrigerators with similar refrigerating capacity at other facilities, we took measures that cleaning the heat exchangers with Freon and replacing the ADS with new one. As a result, the differential pressure rise phenomenon was removed to recover the performance. We have detected oil from the Freon used for cleaning the heat exchangers and at a felt supporting charcoal packed in the ADS. In particular, oil was accumulated in membranous form onto the felt at the entrance side in the ADS. The amount of oil contained in the helium gas was about 10 ppb or so, less than the design value, in the helium refrigerator. However, the oil accumulated onto the felt in the ADS through long operating period may cause abnormal differential pressure rise, leading to the performance degradation of the helium refrigerator. Further study is needed to specify the cause more clearly.

JAEA Reports

Evaluation items to attain safety requirements in fuel and core designs for commercial HTGRs

Nakagawa, Shigeaki; Sato, Hiroyuki; Fukaya, Yuji; Tokuhara, Kazumi; Ohashi, Hirofumi

JAEA-Technology 2017-022, 32 Pages, 2017/09

JAEA-Technology-2017-022.pdf:3.59MB

As for the design of commercial HTGRs, the fuel design, core design, reactor coolant system design, secondary helium system design, decay heat removal system design and confinement system design are very important and quite different from those of LWRs. To contribute the establishment of the safety standards for commercial HTGRs, the evaluation items to attain safety requirements in fuel and core designs were studied. In this study, the excellence features of HTGRs based on passive safety or inherent safety were fully reflected. Additionally, concerning the core design, the stability to spatial power oscillation in reactor core of HTGR was studied. The evaluation items as the result of the study are applicable to the safety design of commercial HTGRs in the future.

JAEA Reports

Cutting operation of simulated fuel assembly heating examination by AWJ

Abe, Yuta; Nakagiri, Toshio; Watatani, Satoshi*; Maruyama, Shinichiro*

JAEA-Technology 2017-023, 46 Pages, 2017/10

JAEA-Technology-2017-023.pdf:8.01MB

This is a report on Abrasive Water Jet (AWJ) cutting work carried out on specimen, which was used for Simulated Fuel Assembly Heating Examination by Collaborative Laboratories for Advanced Decommissioning Science (CLADS) molten core behavior analysis group in February 2016. The simulated fuel assembly is composed of Zirconia for the outer crucible/simulated fuel, stainless steel for the control blade and Zircaloy (Zr) for the cladding tube/channel box. Therefore, it is necessary to cut at once substances having a wide range of fracture toughness and hardness. Moreover, it is a large specimen with an approximate size of 300 mm. In addition, epoxy resin has high stickiness, making it more difficult to cut. Considering these effects, AWJ cutting was selected. The following two points were devised, and this specimen could be cut with AWJ. If it was not possible to cut at one time like a molten portion of boride, it was repeatedly cut. By using Abrasive Suspension Jet (ASJ) system with higher cutting ability than Abrasive Injection Jet (AIJ, conventional method) system, cutting time was shortened. As a result of this work, the cutting method in Simulated Fuel Assembly Heating Examination was established. Incidentally, in the cutting operation, when the cutting ability was lost at the tip of the AWJ, a curved cut surface, which occurs when the jet flowed away from the feeding direction, could be confirmed at the center of the test body. From the next work, to improve the cutting efficiency, we propose adding a mechanism such as turning the cutting member itself for re-cutting from the exit side of the jet and appropriate traverse speed to protect cut surface.

JAEA Reports

Inspection and repair techniques in reactor vessel of the experimental fast reactor JOYO; Development of devices for retrieving bent MARICO-2 subassembly and completion of retrieval work

Ashida, Takashi; Nakamura, Toshiyuki*; Ito, Hideaki

JAEA-Technology 2017-024, 198 Pages, 2017/11

JAEA-Technology-2017-024.pdf:55.8MB
JAEA-Technology-2017-024-appendix(CD-ROM)-1.zip:298.09MB
JAEA-Technology-2017-024-appendix(CD-ROM)-2.zip:210.77MB

In the experimental fast reactor Joyo, the disconnecting of an irradiation test subassembly MARICO-2 (Material Testing Irradiation Rig with Temperature Control) from its holding mechanism was conducted in May 2007. After the operation, the rotating plug was rotated despite the fact that the test subassembly was not disconnected completely. Consequently, top of wrapper tube of the MARICO-2 subassembly was bent onto the in-vessel storage rack. Since the overhanging part of the subassembly was in the height in which contacts with the upper core structure, it had damaged the bottom surface of the upper core structure. As the result, it was necessary to replace the damaged upper core structure and to retrieve the bent MARICO-2 subassembly for Joyo restart. Retrieval devices for MARICO-2 subassembly consist of a gripper mechanism to lift subassembly together with transfer pot, a guide tube built-in a pantograph mechanism to adjust lifting axis and safety mechanisms to prevent or mitigate falling of MARICO-2 subassembly, a retrieval cask and so on. Design of the retrieval devices have been verified in ex-vessel partial or full-scale mock-up tests and in-vessel function tests. In 2014, MARICO-2 subassembly was successfully retrieved from the reactor vessel by applying these retrieval devices. Then, retrieved subassembly was transported to a hot-cell facility for post-irradiation examinations. Devices have demonstrated expected performance under the actual environmental conditions of a sodium cooled fast reactor. This is a synthetic report about the retrieval work of the deformed and irradiated test subassembly in Joyo. This report includes the detail design and fabrication of the special retrieval device, results of tests for confirmation including the mock-up tests in manufacturer's factory, and results of MARICO-2 retrieval work from the reactor vessel.

JAEA Reports

Development of $$^{93}$$Zr, $$^{93}$$Mo, $$^{107}$$Pd and $$^{126}$$Sn analytical methods for radioactive waste from Fukushima Daiichi Nuclear Power Station

Aono, Ryuji; Sato, Yoshiyuki; Shimada, Asako; Tanaka, Kiwamu; Ueno, Takashi; Ishimori, Kenichiro; Kameo, Yutaka

JAEA-Technology 2017-025, 32 Pages, 2017/11

JAEA-Technology-2017-025.pdf:1.45MB

We have developed analytical methods for $$^{93}$$Zr, $$^{93}$$Mo, $$^{107}$$Pd and $$^{126}$$Sn, which are considered important in terms of the safety assessment of radioactive waste disposal. The methods are specialized for the wastes left after Fukushima accident. As the main analytical sample, we assumed accumulated water / treated water collected at Fukushima Daiichi Nuclear Power Station. As for $$^{93}$$Zr, $$^{93}$$Mo, $$^{107}$$Pd and $$^{126}$$Sn contained in this accumulated water / treated water, we have worked on the development of separation and purification method of target nuclide and improvement of recovery, and summarized these results in this report.

JAEA Reports

Degradation behavior of optical components by gamma irradiation (Contract research)

Takeuchi, Tomoaki; Shibata, Hiroshi; Hanakawa, Hiroki; Uehara, Toshiaki*; Ueno, Shunji*; Tsuchiya, Kunihiko; Kumahara, Hajime*; Shibagaki, Taro*; Komanome, Hirohisa*

JAEA-Technology 2017-026, 26 Pages, 2018/02

JAEA-Technology-2017-026.pdf:4.0MB

Under severe accidents, high-integrity transmission techniques are necessary so as to monitor the situation of the nuclear power plant. In this study, effects of gamma irradiation up to 10$$^{6}$$Gy on properties of optical devices were evaluated toward the development of a radiation-resistant in-water wireless transmission system using visible light. After the irradiation, for the LEDs, the total luminous flux decreased and the browning of resin lenses occurred. Meanwhile, the current-voltage characteristics hardly changed. For the PDs, the light sensitivity decreased and the browning of resin window occurred. The dark currents of PDs did not become large enough to adversely affect transmission. These results indicated that both the decreases of the total luminous flux of the LEDs and the light sensitivity of the PDs were mainly caused by not the degradation of the semiconductor parts but the browning of the resin parts by the irradiation. In addition, basic decrease behaviors of light transmission of several different types of glasses by gamma irradiation were also obtained so as to select the suitable optical windows and filters for the developing radiation-resistant in-water wireless transmission system.

JAEA Reports

Development plan of austenitic Fe and Ni based alloys with improved corrosion resistance to sulfuric acid and HI fluids of industrial processes

Hirota, Noriaki; Iwatsuki, Jin; Imai, Yoshiyuki; Yan, X.

JAEA-Technology 2017-027, 19 Pages, 2017/12

JAEA-Technology-2017-027.pdf:2.08MB

In this study, austenitic Fe-based alloys and Ni based alloys was developed as candidate structural materials for equipment operated in sulfuric acid and hydrogen iodide (HI) environment, which exists in various industrial processes including iodine-sulfur (IS) hydrogen production process and geothermal power generation process. The objectives of the study are to achieve the corrosion resistance performance sufficient under the working condition of these processes and to overcome the practical scale-up difficulty of the ceramic (SiC) material that is presently used in the processes due to the manufacturing size limitation of the ceramic. The chemical composition development plan for the austenitic Fe-based alloys is threefold: reinforcement of matrix by addition of Cu and Ta, strength compensation of the surface film by addition of Si and Ti, and prevention of peeling of surface oxide by addition of rare earth elements. Because addition of Cu and Si is known to reduce the ductility of the material and thus manufacturability of the component, it is important to determine the allowable amount of each element to be added. On the other hand, the chemical composition development plan for the Ni based alloys is reinforcement of matrix by addition of Mo, W and Ta, strength compensation of the surface film by addition of Ti, and prevention of peeling of surface oxide by addition of rare earth elements. In particular, the addition of Mo and W to the Ni based alloy is expected to be effective in preventing dimensional deviation of structures from increasing during heating and cooling of process equipment. Various material specimens will be fabricated based on the above chemical composition development plans and tests on these specimens will then be carried out to confirm the corrosion resistance performance under the fluid conditions simulating each industrial processes.

JAEA Reports

International Noble Gas joint measurement at Mutsu City, Aomori for CTBT verification

Kijima, Yuichi; Yamamoto, Yoichi; Oda, Tetsuzo

JAEA-Technology 2017-028, 33 Pages, 2018/01

JAEA-Technology-2017-028.pdf:38.85MB

The International Noble Gas Experiment related to monitoring network for radioactive noble gas (xenon) has been carried out all over the world, as part of the International Monitoring System (IMS) of CTBT. Thirty IMS radionuclide stations including the Takasaki station in Japan are monitoring radioxenon. The past measurement results show that several stations often detect radioxenon and the major emission source of these radioxenon is medical radioisotope production facilities. And nuclear power plants and medical institutions used radioxenon for nuclear medicine diagnosis, and so on are also considered as the possible sources of radioxenon, and it is therefore important to understand the background behavior of radioxenon originated from above facilities for enhancement of monitoring capability for nuclear tests. The international joint measurement was conducted in 2012 by the Preparatory Commission for the CTBT Organization, US Pacific Northwest National Laboratory, Japan Chemical Analysis Center and JAEA at the Ohminato site of JAEA Aomori Research and Development Center in Mutsu city, Aomori, as part of investigation on radioxenon background in East Asia region. In 2014, the additional measurement was carried out at the same place for further investigation. A high sensitive Transportable Xenon Laboratory developed by PNNL was used for this measurement. This paper describes the outline and the results of the joint measurement conducted in 2012 and 2014.

JAEA Reports

Assessment of lead-bismuth-eutectic leak at ADS Target Test Facility in Transmutation Experimental Facility of J-PARC

Iwamoto, Hiroki; Maekawa, Fujio; Matsuda, Hiroki; Meigo, Shinichiro

JAEA-Technology 2017-029, 39 Pages, 2018/01

JAEA-Technology-2017-029.pdf:2.68MB

Under an assumption that an incident of lead-bismuth eutectic (LBE) leak from an LBE circulation system occurred during a 250-kW beam operation, an estimation of radiation dose at the site boundary for the ADS Target Test Facility (TEF-T) in Transmutation Experimental Facility (TEF) of J-PARC was conducted using various conservative assumptions. As a result, the radiation dose at the site boundary was estimated to be about 660 $$mu$$Sv, which were dominated by mercury, noble gas, and iodine produced as spallation products from the LBE. Even though the incident scenario was made conservatively, it was shown that the estimated total dose was lower than the annual radiation dose due to natural sources, and the TEF-T has sufficient safety margin for the leak of radioactivity.

JAEA Reports

Calculation of radioactivity concentration of Cs-137 corresponded to the reference dose for the trench disposal facility according to the design condition

Sakai, Akihiro; Nakata, Hisakazu; Amazawa, Hiroya

JAEA-Technology 2017-030, 176 Pages, 2018/02

JAEA-Technology-2017-030.pdf:4.09MB

At present, the reuse method for the contaminated soil generated from the decontamination of radioactive materials caused by the accident of the Tokyo Electric Power Cooperation Fukushima Daiichi Nuclear Power Plant after intermediate storage is being discussed. The radioactivity concentration of contaminated soil with about 20 million cubic meters within total arising volume of the soil is less than 100 kBq/kg. Therefore, when it is assumed that contaminated soil was disposed of in the trench facility, exposure doses to public at the various exposure pathways resulting from Cs-134 and Cs-137 contained in the removal soil were calculated. From the dose calculation results, the radioactivity concentrations corresponded to reference doses that are assumed to be 0.01 mSv/y or 0.3 mSv/y were evaluated. Then, variation of the radioactivity concentrations was evaluated when the volume of disposal facility was increased taking into account variation of the volume of contaminated soil.

JAEA Reports

Waste acceptance criteria for waste packages destined for near surface disposal containing radioactive waste from research, industrial and medical facilities; Study on a method that fills voidage in waste package with sandy soil

Nakata, Hisakazu; Hayashi, Hirokazu; Amazawa, Hiroya; Sakai, Akihiro

JAEA-Technology 2017-031, 41 Pages, 2018/01

JAEA-Technology-2017-031.pdf:5.27MB

JAEA plans to install disposal facilities for radioactive waste arising from research institutes. It must meet the technical standards specified in the relevant rule. One technical standard is that the disposal facilities shall be performance so as not to be left with the voids after the backfilling with soil. Additionally, the rule also requires this radioactive waste be enclosed in a container in which no harmful voids remain. In order to contribute to the development of a method that adapts the disposal facilities to these technical standards, JAEA adopts a waste conditioning artifice that aims for reducing a quantity of voidage in each waste container by a vibration filling method using sandy soil, providing with average void ratios inside the disposal facilities being adequately controlled. In this reports, filling property tests are conducted in the light of filling sand characteristics, types of metal waste and vibration conditions.

JAEA Reports

Corrosion test of Fugen pressure tube (Zr-2.5wt%Nb alloy) under the sub-surface disposal environment, 2; Examination of long-term corrosion rate by 5 years keeping sample

Sugaya, Toshikatsu; Nakatani, Takayoshi; Sakai, Akihiro

JAEA-Technology 2017-032, 21 Pages, 2018/01

[The article has been found to have a problem about reliability of the corrosion data acquisition, and thus it is unavailable to download the full text in accordance with authors' intentions to retract the report.] For the purpose of the setting of the rate of nuclide elution necessary to safety assessment, we planned the gas-accumulating type corrosion test on Zr-2.5wt%Nb alloy in order to obtain long-term corrosion rate under low temperature, low oxygen and alkaline conditions assuming the disposal environment. A corrosion rate over a testing period of 5 years is acquired with the aim to grasp a long-term corrosion rate behavior in this report. This corrosion rate is compared with the same data that was previously acquired over a testing period of 2 years. As a result, it is confirmed that an evaluation method that is proportional to the minus cubic root of corrosion time squared can be applicable to the corrosion rate behavior acquired this time over a testing period of 5 years, which is the same result in evaluating the corrosion rate behavior acquired over a testing period of 2 years.

JAEA Reports

Safety design report on J-PARC Transmutation Physics Experimental Facility (TEF-P)

Partitioning and Transmutation Technology Division, Nuclear Science and Engineering Center

JAEA-Technology 2017-033, 383 Pages, 2018/02

JAEA-Technology-2017-033.pdf:28.16MB

JAEA is pursuing research and development (R&D) on volume reduction and mitigation of degree of harmfulness of high-level radioactive waste. Construction of Transmutation Experimental Facility (TEF) is under planning as one of the second phase facilities in the Japan Proton Accelerator Complex (J-PARC) program to promote R&D on the transmutation technology with using accelerator driven systems (ADS). The TEF consists of two facilities: ADS Target Test Facility (TEF-T) and Transmutation Physics Experimental Facility (TEF-P). Development of spallation target technology and study on target materials are to be conducted in TEF-T with impinging a high intensity proton beam on a liquid lead-bismuth eutectic target. Whereas in TEF-P, by introducing a proton beam to minor actinide loaded cores, reactor physical properties of the cores are to be studied, and operation experiences of ADS are to be acquired. This report summarizes results of safety design for establishment permit of one of two TEF facilities, TEF-P.

JAEA Reports

Radiation monitoring using manned helicopter around the Nuclear Power Station in the fiscal year 2016 (Contract research)

Sanada, Yukihisa; Mori, Airi; Iwai, Takeyuki; Seguchi, Eisaku; Matsunaga, Yuki*; Kawabata, Tomoki; Toyoda, Masayuki*; Tobita, Shinichiro*; Hiraga, Shogo; Sato, Yoshiharu; et al.

JAEA-Technology 2017-034, 117 Pages, 2018/02

JAEA-Technology-2017-034.pdf:25.18MB

By the nuclear disaster of Fukushima Daiichi Nuclear Power Station (FDNPS), Tokyo Electric Power Company (TEPCO), caused by the East Japan earthquake and the following tsunami occurred on March 11, 2011, a large amount of radioactive materials was released from the NPS. After the nuclear disaster, airborne radiation monitoring using manned helicopter was conducted around FDNPS. This result of the aerial radiation monitoring using the manned helicopter in the fiscal 2016 were summarized in the report. In addition, we developed the discrimination technique of the Rn-progenies. The accuracy of aerial radiation monitoring was evaluated by taking into consideration GPS position error.

JAEA Reports

Background radiation monitoring using manned helicopter for establishment of technique of nuclear emergency response in the fiscal year 2016 (Contract research)

Sanada, Yukihisa; Mori, Airi; Iwai, Takeyuki; Seguchi, Eisaku; Matsunaga, Yuki*; Kawabata, Tomoki; Toyoda, Masayuki*; Tobita, Shinichiro*; Hiraga, Shogo; Sato, Yoshiharu; et al.

JAEA-Technology 2017-035, 69 Pages, 2018/02

JAEA-Technology-2017-035.pdf:32.92MB

By the nuclear disaster of Fukushima Daiichi Nuclear Power Station (FDNPS), Tokyo Electric Power Company (TEPCO), caused by the East Japan earthquake and the following tsunami occurred on March 11, 2011, a large amount of radioactive materials was released from the NPS. After the nuclear disaster, airborne radiation monitoring using manned helicopter was conducted around FDNPS. We carried out the background monitoring around the nuclear power stations of the whole country to apply a technique of the airborne radiation monitoring that is cultivated in Fukushima as a technology of nuclear emergency response. This result of the aerial radiation monitoring using the manned helicopter around Ooi, Takahama and Ikata Nuclear Power Station and in the fiscal 2016 were summarized in the report. In addition, technical issues were described.

JAEA Reports

Measurement and analysis of in-vessel component activation and gamma dose rate distribution in Joyo, 2

Yamamoto, Takahiro; Ito, Chikara; Maeda, Shigetaka; Ito, Hideaki; Sekine, Takashi

JAEA-Technology 2017-036, 41 Pages, 2018/02

JAEA-Technology-2017-036.pdf:7.86MB

In the experimental fast reactor Joyo, the damaged upper core structure (UCS) was retrieved into the cask in May 2014 The dose rate on UCS surface was quite high due to the activation for over 30 years operation. In order to attain the optimum safety design, manufacture and operation of equipment for UCS replacement, the method to evaluate UCS surface dose rate was developed on the basis of C/E obtained by the in-vessel dose rate measurement in Joyo. In order to verify the evaluation method, the axial gamma-ray distribution measurement on the surface of the cask, which contained UCS, was conducted using a plastic scintillating optical fiber (PSF) detector. This paper describes the comparison results between calculation and measurement as follows. (1) The measured axial gamma-ray distribution on the cask surface had a peak on proper location with considering the cask shielding structure and agree well with the calculated distribution. (2) The C/E of axial gamma-ray distribution on the cask surface was ranged from 1.1 to 1.7. It was confirmed that the calculation for UCS replacement equipment design had a margin conservatively. Then, the results showed that the developed evaluation method for UCS replacement equipment design was sufficiently reliable.

JAEA Reports

Radioactive material analysis and research facility construction of first stage design report

Ichimura, Takahito; Takahashi, Iku; Iwasa, Kaoru; Nozawa, Yoshihiko

JAEA-Technology 2017-037, 322 Pages, 2018/03

JAEA-Technology-2017-037.pdf:69.64MB

The facility construction of the Okuma Analysis and Research Center is being promoted on the basis of the first phase facilities and the second stage facilities in accordance with the road map prescribed by the government in order to promote the decommissioning of the Fukushima Daiichi Nuclear Power Station is there. As the first phase facility, we designed "Radioactive Material Analyses and Research Facility Laboratory-1" which analyses mainly low-medium radiation doses radioactive waste (rabble, logging tree (incineration ash), dismantled waste etc), "Administrative Building" which is an office building, Site decontamination, Exterior and Utility equipment, etc. In addition to radiation protection and prevention of pollution at the time of construction, it is designed to reduce radiation exposure from outdoor pollution sources of residents after the facility starts to operate, shielding for the reliability of the analysis results, suppression and prevention of outside air inflow, etc.

JAEA Reports

Mock-up test of the modified STACY (Performance check of water feed and drain system)

Seki, Masakazu; Maekawa, Tomoyuki; Izawa, Kazuhiko; Sono, Hiroki

JAEA-Technology 2017-038, 52 Pages, 2018/03

JAEA-Technology-2017-038.pdf:4.6MB

The Japan Atomic Energy Agency is conducting a reactor modification project of the Static Experiment Critical Facility (STACY). In the modification, STACY is to be converted from a thermal reactor using solution fuel into that using fuel rods and light water moderator. Reactivity of the modified STACY core is controlled by the water level fed in the core tank as well as the present STACY. In order to verify the basic design of the water feed and drain system of the modified STACY, we constructed a mockup test apparatus with almost the same structure and specifications as the modified STACY. In the mockup test, performance checks were pursued regarding limitation of maximum flow of water feeding, adjustment of the flow rate of water feeding, stop of water feeding and others. This report describes the outline and results of the mock-up test of the water feed and drain system of the modified STACY.

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