Evaluation of decay heat used for effectiveness evaluations of countermeasures against severe accidents in the prototype FBR Monju

Usami, Shin; Kishimoto, Yasufumi*; Taninaka, Hiroshi; Maeda, Shigetaka

JAEA-Technology 2018-003, 97 Pages, 2018/07

JAEA-Technology-2018-003.pdf:12.54MB

The decay heat used for effectiveness evaluation of the prevention measures against severe accidents in the prototype fast breeder reactor Monju was evaluated by applying the updated nuclear data libraries based on JENDL-4.0, reflecting the realistic core operation pattern, and setting the rational extent of uncertainty. The decay heats of fission products, the actinide nuclides such as Cm-242, and radioactive structural materials were calculated by FPGS code. The decay heat of U-239 and Np-239 was evaluated based on ANSI/ANS-5.1-1994. The calculation uncertainty of each decay heat was evaluated based on summation of uncertainty factors, C/E values of reaction rates obtained in Monju system startup test, and so on. Furthermore, the decay heat evaluation method based on the FPGS90 was verified by the comparison of the results of the decay heat measurement of the two spent MOX fuel subassemblies in the experimental fast reactor Joyo MK-II core.

A Refined analysis on the power reactivity loss measurement in Monju

Taninaka, Hiroshi; Takegoshi, Atsushi; Kishimoto, Yasufumi*; Mori, Tetsuya; Usami, Shin

Progress in Nuclear Energy, 101(Part C), p.329 - 337, 2017/11

https://doi.org/10.1016/j.pnucene.2017.06.004

The present paper describes the evaluation of the power reactivity loss data obtained in the Japanese prototype fast breeder reactor Monju. The most recent analysis on the power reactivity loss measurement (Takano, et al., 2008) is updated considering the following findings: (a) in-core temperature distribution effect, (b) crystalline binding effect, (c) logarithmic averaging of the fuel temperature, (d) localized fuel thermal elongation effect, (e) updated Japanese Evaluated Nuclear Data Library, JENDL-4.0, and (f) refined corrections on the measured value. The influences of the updates are quantitatively identified and the most precise and probable C/E value is derived together with a thorough uncertainty evaluation. As a result, it is revealed that the analysis overestimates the measurement by 4.6% for the measurement uncertainty of 2.0%. The discrepancy is reduced to as small as 1.1% when the core bowing effect is considered, which implies the importance of the core bowing effect in the calculation of the power reactivity loss.

Validation of decay heat evaluation method based on FPGS cord for fast reactor spent MOX fuels

Usami, Shin; Kishimoto, Yasufumi; Taninaka, Hiroshi; Maeda, Shigetaka

Proceedings of International Conference on the Physics of Reactors; Unifying Theory and Experiments in the 21st Century (PHYSOR 2016) (USB Flash Drive), p.3263 - 3274, 2016/05

The present paper describes the validation of the new decay heat evaluation method using FPGS90 code with both the updated nuclear data library and the rational extent of uncertainty, by comparing the results of the decay heat measurement of the spent fuel subassemblies in Joyo MK-II core and by comparing with the calculation results of ORIGEN2.2 code. The calculated values of decay heat (C) by FPGS90 based on the JENDL-4.0 library were coincident with the measured ones (E) within the calculation uncertainties, and the C/E ranged from 1.01 to 0.93. FPGS90 evaluated the decay heat almost 3% larger than ORIGEN2.2, and it improved the C/E in comparison with the ORIGEN2.2 code. Furthermore, The C/E by FPGS90 based on the JENDL-4.0 library was improved than that based on the JENDL-3.2 library, and the contribution of the revision of reaction cross section library to the improvement was dominant rather than that of the decay data and fission yield data libraries.

Operation and commissioning of IFMIF (International Fusion Materials Irradiation Facility) LIPAc injector

Okumura, Yoshikazu; Gobin, R.*; Knaster, J.*; Heidinger, R.*; Ayala, J.-M.*; Bolzon, B.*; Cara, P.*; Chauvin, N.*; Chel, S.*; Gex, D.*; et al.

Review of Scientific Instruments, 87(2), p.02A739_1 - 02A739_3, 2016/02

https://doi.org/10.1063/1.4936248

IFMIF is an accelerator based neutron facility having two set of linear accelerators each producing 125mA/CW deuterium ion beams (250mA in total) at 40MeV. The LIPAc (Linear IFMIF Prototype Accelerator) being developed in the IFMIF-EVEDA project consists of an injector, a RFQ accelerator, and a part of superconducting Linac, whose target is to demonstrate 125mA/CW deuterium ion beam acceleration up to 9MeV. The injector has been developed in CEA Saclay and already demonstrated 140mA/100keV deuterium beam. The injector was disassembled and delivered to the International Fusion Energy Research Center (IFERC) in Rokkasho, Japan, and the commissioning has started after its reassembly 2014; the first beam production has been achieved in November 2014. Up to now, 100keV/120mA/CW hydrogen ion beam has been produced with a low beam emittance of 0.2 .mm.mrad (rms, normalized).

Development of a fast reactor for minor actinides transmutation, 1; Overview and method development

Takeda, Toshikazu*; Usami, Shin; Fujimura, Koji*; Takakuwa, Masayuki*

Proceedings of 21st International Conference & Exhibition; Nuclear Fuel Cycle for a Low-Carbon Future (GLOBAL 2015) (USB Flash Drive), p.560 - 566, 2015/09

The Ministry of Education, Culture, Sports, Science and Technology in Japan has launched a national project entitled "technology development for the environmental burden reduction" in 2013. The present study is one of the studies adopted as the national project. The objective of the study is the efficient and safe transmutation and volume reduction of minor actinides with long-lived radioactivity and high decay heat contained in high level radioactive wastes by using sodium cooled fast reactors. We are developing MA transmutation core concepts which harmonize efficient MA transmutation with core safety. To accurately design the core concepts we have improved calculation methods for estimating the transmutation rate of individual MA nuclides, and estimating and reducing uncertainty of MA transmutation. The overview of the present project is first described. The method improvement is presented with numerical results for a minor-actinide transmutation fast reactor.

Tensile mechanical properties of a stainless steel irradiated up to 19 dpa in the Swiss spallation neutron source

Saito, Shigeru; Kikuchi, Kenji*; Hamaguchi, Dai; Usami, Koji; Endo, Shinya; Ono, Katsuto; Matsui, Hiroki; Kawai, Masayoshi*; Dai, Y.*

Journal of Nuclear Materials, 431(1-3), p.44 - 51, 2012/12

https://doi.org/10.1016/j.jnucmat.2011.11.028

To evaluate the lifetime of the beam window of an accelerator-driven transmutation system (ADS), post irradiation examination (PIE) of the STIP (SINQ target irradiation program, SINQ; Swiss spallation neutron source) specimens was carried out. The specimens tested in this study were made from the austenitic steel JPCA (Japan primary candidate alloy). The specimens were irradiated at SINQ Target 4 (STIP-II) with high-energy protons and spallation neutrons. The irradiation conditions were as follows: the proton energy was 580 MeV, irradiation temperatures ranged from 100 to 430C, and displacement damage levels ranged from 7.1 to 19.5 dpa. Tensile tests were performed in air at room temperature (R.T.), 250C and 350C. Fracture surface observation after the tests was done by SEM (Scanning electron microscope). Results of the tensile tests performed at R.T. showed the extra hardening of JPCA at higher dose compared to the fission neutron irradiated data. At the higher temperatures, 250C and 350C, the extra hardening was not observed. Degradation of ductility bottomed around 10 dpa, and specimens kept their ductility until 19.5 dpa. All specimens fractured in ductile manner. The result from a microstructure observation on a specimen irradiated to 19.3 dpa at 420C indicates that some agglomeration of bubbles on grain boundaries was observed in the specimen irradiated to 19.3 dpa at 420C. However the tensile specimen irradiated up to 18.4 dpa at 425C still exhibited little loss of ductility. Since He/dpa was very high on SINQ target irradiations, the formation of highly dense small bubbles in the matrix consequently avoided the accumulation of He on grain boundaries, which might have resulted in avoiding grain boundary embrittlement.

Decommissioning five facilities in Nuclear Science Research Institute

Terunuma, Akihiro; Naito, Akira; Nemoto, Koichi; Usami, Jun; Tomii, Hiroyuki; Shiraishi, Kunio; Ito, Shinichi

JAEA-Review 2010-038, 96 Pages, 2010/09

JAEA-Review-2010-038.pdf:5.9MB

Japan Atomic Energy Agency (JAEA) has midterm plan for decommissioning the facilities being finished their role and the facilities that became unnecessary by shifting their functions to other facilities. In the first midterm plan (from the latter half of fiscal year 2005 to fiscal year 2009), decommissioning of five facilities (Ceramic Research Facility, Plutonium Research Facility No.2, Metallurgy Research Facility, Isotope Separation Research Facility and Reprocessing Test Facility) had been carried out in order to release controlled area and dismantle the facilities in Nuclear Science Research Institute (NSRI), JAEA. The decommissioning activity for each facility had been reported to the regulatory body and municipalities. On this report, we summarize the each activity for five facilities by reviewing the reports to the regulatory body and municipalities. We also added the knowledge obtained through the activity.

Replacement technology for front acrylic panels of a large-sized glove box using bag-in / bag-out method

Sakuraba, Naotoshi; Numata, Masami; Komiya, Tomokazu; Ichise, Kenichi; Nishi, Masahiro; Tomita, Takeshi; Usami, Koji; Endo, Shinya; Miyata, Seiichi; Kurosawa, Tatsuya; et al.

JAEA-Technology 2009-071, 34 Pages, 2010/03

JAEA-Technology-2009-071.pdf:21.07MB

As a part of maintenance technology of a large-sized glove box for handling of TRU nuclides, we developed replacement technology for front acrylic panels using the bag-in/bag-out method and applied this technology to replace the deteriorated front acrylic panels at Waste Safety Testing Facility (WASTEF) in Nuclear Science Research Institute of Japan Atomic Energy Agency (JAEA). As a consequence, we could safely replace the front acrylic panels under the condition of continuous negative pressure only with partial decontamination of the glove box. We also demonstrated that the present technology is highly effective in points of safety, workability and cost as compared to the usual replacement technology for front acrylic panels of a glove box, where workers in an air-line suit replace directly the front acrylic panels in a green house.

Mechanical properties of austenitic stainless steels irradiated at SINQ target 4

Saito, Shigeru; Hamaguchi, Dai; Usami, Koji; Endo, Shinya; Ono, Katsuto; Matsui, Hiroki; Kikuchi, Kenji*; Kawai, Masayoshi*; Yong, D.*

Proceedings of 1st International Workshop on Technology and Components of Accelerator-driven Systems (TCADS-1) (Internet), 9 Pages, 2010/03

The research and development for an accelerator-driven system (ADS) to transmute minor actinide (MA) have been progressed. The target beam window of ADS submerged in the reactor will be subjected to high-energy proton and spallation neutron irradiation. To evaluate mechanical properties of irradiated materials, post irradiation examination (PIE) of the STIP (SINQ target irradiation program) specimens was carried out at JAEA. In the present study, PIE on austenitic steels JPCA and Alloy800H irradiated at SINQ target 4 (STIP-II) was conducted. Austenitic steels are preferable as the material for the target beam window of ADS from the view point of DBTT shift, which should be taken into consideration for ferritic / martensitic steels. The irradiation conditions were as follows: proton energy was 580 MeV, irradiation temperatures were ranged from 100 to 450C, and displacement damage levels were ranged from 6.5 to 19.5 dpa. Tensile tests were performed in air at R.T. to 350C. Results of the tensile tests performed at R.T. indicate that irradiation hardening occurred with increasing displacement damage level up to 10 dpa. At higher doses, irradiation hardening seemed to tend to saturate. Degradation of ductility was bottomed around 10 dpa and specimens kept its ductility until 19.5 dpa. All the specimens fractured in ductile manner.

Feasibility study on an upgraded future Monju core concept with extended operation cycle length of one year and increased fuel burnup

Kinjo, Hidehito*; Kageyama, Takeshi*; Kitano, Akihiro; Usami, Shin

Nuclear Technology, 167(2), p.254 - 267, 2009/08

https://doi.org/10.13182/NT09-A8962

A conceptual design study has been performed on upgrading the core performance of the Japanese FBR Monju. The main aim of this study is to investigate and demonstrate the feasibility of an upgraded core with an extended refueling interval of 365 EFPD and increased average fuel burnup of 150 GWd/t, which are expected in future commercial FBRs. Two design measures have been taken to accommodate the largely increased burnup reactivity for the longer cycle: (1) A modified fuel pin with increased pin diameter, pellet density and active core height has been introduced to improve the burnup reactivity, (2) The control rod specification has been modified to enhance the reactivity worth by increasing the B content to assure sufficient shutdown margin. The evaluation results show that even a medium sized core of about 2.5 m could achieve the target, without causing significant drawbacks to the core characteristics. The feasibility is thus demonstrated.