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Journal Articles

Treatment technology of highly radioactive solid waste generated by experimental tests and sample analysis in reprocessing facilities

Goto, Yuichi; Inada, Satoshi; Kuno, Takehiko; Mori, Eito*

Nippon Hozen Gakkai Dai-16-Kai Gakujutsu Koenkai Yoshishu, p.221 - 224, 2019/07

Test equipment, containers, and analytical wastes, generated by experiments using spent fuel pieces in hot cell of Operation Testing Laboratory and by analysis of highly active liquid wastes in hot analytical cell line of Tokai Reprocessing Plant, are treated as highly radioactive solid wastes. These wastes are stored in specific shielded containers called waste cask and then transport to the storage facility. The treatment of these highly radioactive solid wastes have been carried out for 40 years with upgrading waste taking out system and transportation device. As a results, automation of several procedures have been achieved utilizing conventional equipment, and work efficiency and safety have been improved.

Journal Articles

Dose reduction measure in exchange work of valves used for agitation of highly active liquid waste in storage tank

Isozaki, Naohiko; Morimoto, Kenji; Furukawa, Ryuichi; Tsuboi, Masatoshi; Yada, Yuji; Miyoshi, Ryuta; Uchida, Toyomi; Ikezawa, Kazumi*; Kurosawa, Kenji*

Nippon Hozen Gakkai Dai-16-Kai Gakujutsu Koenkai Yoshishu, p.225 - 228, 2019/07

Highly active liquid waste, which is generated by the reprocessing of spent nuclear fuel, is stored in storage tank of Tokai Reprocessing Plant until it is vitrified. The waste solution in the tank is periodically agitated to avoid the precipitation of insoluble residues during the storage. Three way valves and ball valves have been located at the tank for agitation. Radiation dose rate at the valve location is high and operator's radiation exposure become a problem. Therefore, measures to reduce radiation exposure are performed and reported in this presentation.

Journal Articles

Spectrochemistry of technetium by liquid electrode plasma optical emission spectrometry and its applicability of quantification for highly active liquid waste

Yamamoto, Masahiko; Do, V. K.; Taguchi, Shigeo; Kuno, Takehiko; Takamura, Yuzuru*

Spectrochimica Acta, Part B, 155, p.134 - 140, 2019/05

The emission spectra of technetium (Tc) by liquid electrode plasma optical emission spectrometry have been investigated in this study. From the spectra, 52 emission peaks of Tc were observed in the 250-500 nm wavelength range. All peaks were assigned to the neutral state or singly ionized state. The relative intensities of these peaks were similar to those excited by an electric spark. The strongest intensity peaks were found at 254.3 nm, 261.0 nm, and 264.7 nm. Spectral interferences of coexisting elements in highly active liquid waste of reprocessing stream on those three strongest peaks were investigated using simulated sample. No spectral interferences were observed around the 264.7 nm Tc peak. Therefore, analytical performance using 264.7 nm peak was evaluated. The detection limit, estimated on standard and blank samples in 0.4 M nitric acid, was 1.9 mg/L. The relative standard deviation of Tc standard sample (12.0 mg/L) was 3.8% (N = 5, 1$$sigma$$).

Journal Articles

Study on hydrogen generation from cement solidified products loading low-level radioactive liquid wastes at Tokai Reprocessing Plant

Sato, Fuminori; Matsushima, Ryotatsu; Ito, Yoshiyuki

QST-M-16; QST Takasaki Annual Report 2017, P. 60, 2019/03

Hydrogen gas generation by $$gamma$$-radiation from cement solidified products loading low-level radioactive liquid waste generated at LWTF in Tokai Reprocessing Facility was studied.

Journal Articles

Effect of surface conditions of the filament used in thermal ionization mass spectrometry on an uranium isotopic measurement

Taguchi, Shigeo; Miyauchi, Hironari*; Horigome, Kazushi; Yamamoto, Masahiko; Kuno, Takehiko

Bunseki Kagaku, 67(11), p.681 - 686, 2018/11

In thermal ionization mass spectrometry, de-gassing is one of the important treatments to release impurities of filaments and to minimize the influence of background. In this work, the effect of the surface change in the tungsten filament induced by the conductively heating treatment on uranium isotopic ($$^{235}$$U/$$^{238}$$U) measurement has been investigated. It was found that the conductively heating treatment of the filament has the effect of smoothing the surface of the filament and also has the effect of improving the deposition of the sample on the filament surface. As a result of either these effects, the precision of uranium isotopic ($$^{235}$$U/$$^{238}$$U) measurement was improved.

Journal Articles

Security measures at nuclear fuel facilities, 2; Internal threat countermeasure in cyber-security

Kono, Soma; Yamada, Hiroyuki; Goto, Atsushi*; Yamazaki, Katsuyuki; Nakamura, Hironobu; Kitao, Takahiko

Nihon Kaku Busshitsu Kanri Gakkai Dai-39-Kai Nenji Taikai Rombunshu (Internet), 2 Pages, 2018/11

no abstracts in English

Journal Articles

Separation technique using column chromatography for safeguards verification analysis of uranium and plutonium in highly-active liquid waste by isotope dilution mass spectrometry

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

Proceeding IAEA Symposium on International Safeguards; Building Future Safeguards Capabilities (Internet), 8 Pages, 2018/11

In this study, the single-column extraction chromatographic separation has been developed for analysis of U and Pu in highly active liquid waste by isotope dilution mass spectrometry (IDMS). The commercially available TEVA$$^{textregistered}$$ resin is selected as an extraction chromatography resin. The U is chromatographically separated from fission products (FP) elements by nitric acid while Pu(IV) is adsorbed on the resin. After that, Pu is eluted by reducing to Pu(III). The method has been successfully achieved the separation with yielding the enough recovery and sufficient decontamination factors for subsequent IDMS analysis. The column dose rate after the FP removal is decreased to the background. The analytical results obtained by the developed method are in a good agreement with those of the conventional method. It provides simple and rapid separation and expected that the method can be applied to join IAEA/Japan on-site analytical laboratory.

Journal Articles

Development of cement based encapsulation for low radioactive liquid waste in Tokai Reprocessing Plant

Matsushima, Ryotatsu; Sato, Fuminori; Saito, Yasuo; Atarashi, Daiki*

Proceedings of 3rd International Symposium on Cement-based Materials for Nuclear Wastes (NUWCEM 2018) (USB Flash Drive), 4 Pages, 2018/10

At TRP, LWTF was constructed as a facility for processing low radioactive liquid waste and solid waste generated at TRP, and a cold test is been carrying out. In this facility, initially, nitrate waste liquid after separation of nuclides generated with treatment of low radioactive liquid waste was to be solidified by using borate. However, at present, it is necessary to decompose the nitrate in the liquid waste to reduce the environmental burden. For the reason, as a plan to replace the nitrate with the carbonate and to make it as a cement based encapsulation, we are studying for the introduction of the facility. Currently, as a cement solidification technology development for this liquid waste, we are studying the application of cement material based on blast furnace slag (BFS) as a main component. In this report, we show the results of the test conducted on the actual scale (200 L drum can scale).

Journal Articles

Introduction and implementation of physical protection measures including trustworthiness program at Tokai Reprocessing Facilities

Nakamura, Hironobu; Kimura, Takashi; Yamazaki, Katsuyuki; Kitao, Takahiko; Tasaki, Takashi; Iida, Toru

Proceedings of International Conference on Physical Protection of Nuclear Material and Nuclear Facilities (Internet), 9 Pages, 2018/09

After the accident of Fukushima Daiichi Nuclear Power Station, to develop effective security measures based on the lesson learned from such crisis and to meet the IAEA Nuclear Security Recommendations (INFCIRC/225/Rev.5), NRA in Japan made a partial amendment of the regulations concerning the reprocessing activity in 2012. The Tokai reprocessing facility implemented all of those security measures by the end of March 2014. Those new measures help us to keep high degree of security level and contributed to our planned operations to reduce the potential risk of the plant. On the other hand, the trustworthiness program was newly introduced in 2016, based on the trustworthiness policy determined by NRA. The implementing entity of the program is JAEA for the Tokai Reprocessing Facility and is required for both the persons afford unescorted access to Category I and II, CAS/SAS, and the persons afford access to the sensitive information. Those who are involved this program will be judged before engaging the work whether they might act as insider to cause or assist radiological sabotage or unauthorized removal of nuclear material, or leak sensitive information. The program is expected as a measure against insider at reprocessing facilities, and is expected to be enforced around the autumn of 2017. As well as the establishment of security measures, the promoting nuclear security culture for all employees was a big challenge. The Tokai reprocessing facility have introduced several security culture activities, such as case study education of security events done by a small group and putting up the security culture poster and so on. This paper presents introduction and implementation with effectiveness of security measures in the Tokai reprocessing facilities and the future security measures applied to the reprocessing facilities are discussed.

Journal Articles

Physical property evaluation of valve seal material at analytical radioactive liquid waste storage tanks in reprocessing facility

Goto, Yuichi; Yamamoto, Masahiko; Kuno, Takehiko; Inada, Satoshi

Nippon Hozen Gakkai Dai-15-Kai Gakujutsu Koenkai Yoshishu, p.489 - 492, 2018/07

Radioactive liquid waste from the Tokai Reprocessing Facility Analytical Laboratory is temporarily stored in intermediate waste storage tank by using receiving valves. Then, the liquid waste is transferred to liquid treatment facility by using liquid feed valves. The deterioration of the gasket part of these valves (leakage of waste liquid) was confirmed in 2004. Since then, the material of gaskets was changed from polyethylene to Teflon. In 2016, the gaskets were replaced by periodical update. Therefore, physical properties of used gaskets were investigated, and the relevance between radioactive level and degradation degree was evaluated.

Journal Articles

Discussion of effective insider threat mitigation method at reprocessing plant

Nakamura, Hironobu; Kitao, Takahiko; Yamada, Hiroyuki; Kono, Soma; Kimura, Takashi; Tasaki, Takashi

Proceedings of INMM 59th Annual Meeting (Internet), 9 Pages, 2018/07

JAEA Reports

Development of separation technique of Pu, Am, Np using solid phase extraction resin for the determination of impurity metal elements in plutonium nitrate solution by inductivity coupled plasma optical emission spectrometry

Taguchi, Shigeo; Yamamoto, Masahiko; Furuse, Takahiro*; Masaki, Yuji*; Kuno, Takehiko

JAEA-Technology 2018-005, 14 Pages, 2018/06

JAEA-Technology-2018-005.pdf:0.94MB

The method to remove Pu, Am and Np from plutonium nitrate solution recovered from spent nuclear fuel prior to ICP-OES measurement has been developed for the determination of 18 impurity metal elements (Fe, Cr, Ni, Mn, Al, Cd, V, Cu, Si, Zn, Mo, Sn, Ca, Mg, Na, Ag, Pb, B). In this method, two TRU resin packed columns were used for separation. In the first column, Pu and Am were mainly removed by adsorption. The recovered solution from the fist column was added to the second column after reduction of Np, and Am(III) and Np(IV) were removed by adsorption. The Pu nitrate solution (22g/L) of 2mL were treated by proposed method. The alpha emission nuclide was decreased to $$<$$5.8 Bq/mL in a solution diluted to 100mL. As a result of ICP-OES measurement, the recoveries of impurity metals separated by proposed method were almost 100%. This separation scheme can apply to the metal impurity elemental analysis in Pu nitrate solution recovered from spent nuclear fuel.

Journal Articles

Quantitative determination of total cesium in highly active liquid waste by using liquid electrode plasma optical emission spectrometry

Do, V. K.; Yamamoto, Masahiko; Taguchi, Shigeo; Takamura, Yuzuru*; Surugaya, Naoki; Kuno, Takehiko

Talanta, 183, p.283 - 289, 2018/06

 Times Cited Count:1 Percentile:62.19(Chemistry, Analytical)

We develop a novel analytical method employing liquid electrode plasma optical emission spectrometry for measurement of total cesium in highly active liquid wastes. Limit of detection and limit of quantification are 0.005 mg/L and 0.02 mg/L, respectively. The method is validated and applied to the real samples.

Journal Articles

Outline of decommissioning plan of Tokai Reprocessing Plant

Okano, Masanori; Akiyama, Kazuki; Taguchi, Katsuya; Nagasato, Yoshihiko; Omori, Eiichi

Dekomisshoningu Giho, (57), p.53 - 64, 2018/03

The construction of Tokai Reprocessing Plant (TRP) was initiated in June 1971, and its hot test using spent fuel started in September 1977. Thereafter TRP had been operated to reprocess 1,140 tons of spent fuel for approximately 30 years until May 2007, according to the reprocessing contract with domestic electric power companies. JAEA announced a policy of TRP in report of JAEA reform plan published in September 2014. The policy shows that TRP will shift to a decommissioning stage by economic reasons. Based on the policy, application of approval for TRP decommissioning plan was submitted to Nuclear Regulation Authority (NRA) in June 2017. This plan provides basic guidelines such as procedures for decommissioning and specific activities for risk reduction, and implementation divisions of decommissioning, management of spent fuels and radioactive wastes, decommissioning budget, and decommissioning schedule. The process of TRP decommissioning is planned to continue for approximately 70 years until the release of controlled areas of approximately 30 facilities.

Journal Articles

Study on hydrogen generation from cement solidified products loading low-radioactive liquid wastes at Tokai Reprocessing Plant

Ito, Yoshiyuki; Matsushima, Ryotatsu; Sato, Fuminori

QST-M-8; QST Takasaki Annual Report 2016, P. 69, 2018/03

no abstracts in English

Journal Articles

Replacement of the glove port equipped with glove box in Nuclear Fuel Reprocessing Facility

Horigome, Kazushi; Taguchi, Shigeo; Nishida, Naoki; Goto, Yuichi; Inada, Satoshi; Kuno, Takehiko

Nippon Hozen Gakkai Dai-14-Kai Gakujutsu Koenkai Yoshishu, p.381 - 384, 2017/08

no abstracts in English

Journal Articles

Design and application of greenhouse on the maintenance of analytical machineries in Tokai Reprocessing Plant

Suzuki, Yoshimasa; Tanaka, Naoki; Goto, Yuichi; Inada, Satoshi; Kuno, Takehiko

Nippon Hozen Gakkai Dai-14-Kai Gakujutsu Koenkai Yoshishu, p.385 - 389, 2017/08

Greenhouse is used in order to prevent diffusion of radioactive materials on the maintenance of machineries and decomposition of the analytical equipment such as glove box in Tokai Reprocessing Plant (TRP). The specifications of the greenhouse change depending on a risk of the radiation exposure, operation and environment. Design and application of original greenhouses in the analytical laboratory of TRP is summarized.

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

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

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.

362 (Records 1-20 displayed on this page)