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Transactions of the Atomic Energy Society of Japan
- Vol.16(2017)
- Vol.15(2016)
- Vol.14(2015)
- Vol.13(2014)
- Vol.12(2013)
- Vol.11(2012)
- Vol.10(2011)
- Vol.9(2010)
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- Vol.6(2007)
- Vol.5(2006)
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- Vol.3(2004)
- Vol.2(2003)
- Vol.1(2002)

Predecessor

Journal of the Atomic Energy Society of Japan / Atomic Energy Society of Japan

Transactions of the Atomic Energy Society of Japan Vol. 16(2017) No. 3

Article

Distributions of ²⁴²Cm and ²⁴⁴Cm Nuclide Compositions as Neutron Sources and Their Neutron Emission Rates in BWR Spent Fuels based on Three-Dimensional Neutron Transport and Burnup Calculations

Hiroki WATANABE, Takanori KAMEYAMA

Released: August 15, 2017

[Advance Publication] Released: July 21, 2017

p.107-118

　Distributions of the TRU nuclide composition in spent fuels are important for radiation-shielding analyses in their storage and transport. In this research, detailed three-dimensional distributions of ²⁴²Cm and ²⁴⁴Cm compositions as the main neutron sources in spent fuel assemblies (BWR 8×8 and 9×9 types) were calculated with the three-dimensional neutron transport and fuel burnup code system. The distributions of ²⁴²Cm and ²⁴⁴Cm compositions become heterogeneous as fuel burnup increases. Horizontally, the distributions of both compositions in the fuel assemblies always show peaks at the corner rods. Vertically, in the corner rod, the distributions show peaks on the bottom nodes at the initial burnup, and the peak node gradually moves upward as fuel burnup increases until the fuel assemblies are unloaded. For example, in a 9×9 type fuel assembly at the final burnup (55 MWd/kgU), the horizontal peak ratio of the ²⁴⁴Cm composition was 1.78 at the corner rod and its vertical peak ratio was 1.67 on the upper node. ²⁴²Cm and ²⁴⁴Cm compositions mainly contribute to neutron emission rates in spent fuel. Accordingly, it is rational to consider those heterogeneous distributions for the neutron-shielding evaluation of spent fuels of higher burnup.

A Study on Enhancing the Effectiveness of Safety Culture in Nuclear Power Plants

Akihiro YAMAMOTO, Naoto SEKIMURA

Released: August 15, 2017

[Advance Publication] Released: June 27, 2017

p.119-138

　The enhancement of safety culture is an issue for both plant operators and regulators working in fields related to the safety management of nuclear power plants. Plant operators have been collecting safety culture data through a broad range of tools and methods such as observations, interviews and other surveys. However, many issues remain regarding the effectiveness of safety culture activity. A new regulatory authority was established in 2012 after the Fukushima Dai-ichi accident to enforce nuclear safety regulations but its activity is still weak in terms of monitoring the performance of plant operators’ safety culture. In order to promote and strengthen safety culture, plant operators need to collect detailed event information, even when events do not directly affect plant safety, and accumulate information related to human, organizational and technical factors through dialogue among the parties responsible for coping with events. Both operators and regulators should all be working in the same direction to assess information that will help their periodic reviews, and the involvement of local government is a key to enhancing their safety culture.

Simulation of Fuel-Coolant Interaction SERENA2 Test based on JASMINE Version 3

Akitoshi HOTTA, Akinobu MORITA, Mitsuhiro KAJIMOTO, Yu MARUYAMA

Released: August 15, 2017

[Advance Publication] Released: June 27, 2017

p.139-152

　Among twelve FCI cases examined in the OECD/NEA/CSNI/SERENA2 test series using two facilities, six steam explosion cases, five from TROI and one from KROTOS, were analyzed by JASMINE V.3. The major model parameters were categorized into the “focused zone”, i,e., the core part of interest, and the “peripheral zone”, corresponding to the initial and boundary conditions given intentionally for each test case. For the former, base values established through past validation studies of JASMINE V.3 were applied. The code was modified to implement the measured distribution of the entrained droplet size acquired in TROI–VISU. For the latter, melt release histories were given as a combination of time tables of jet diameter and release velocity that were estimated from image data and transit timing data of the melt leading edge. The base values were shown to predict impulse responses of SERENA2 systematically with a reasonable error band. A statistical analysis based on the LHS method was performed. Uncertainty ranges were given on the basis of measurement errors and past validation studies in the development of JASMINE. Underlying mechanisms causing apparent differences in the mechanical energy conversion ratio between two facilities were studied from the viewpoint of breakup length and trigger timing.

Preparation of Sodium Uranates

Masayoshi UNO, Keisuke YOKOYAMA, Yukihiro MURAKAMI

Released: August 15, 2017

[Advance Publication] Released: July 06, 2017

p.153-160

　In order to estimate the fuel behavior during a severe accident in a fast reactor, it is necessary to make pure sodium uranates and to measure their properties. In the present study, sodium uranates (Na_xU_yO_z) were synthesized at various reaction temperatures for various durations using UO₂ and Na₂O or Na₂CO₃. We succeeded in synthesizing nearly pure sodium uranates such as Na₂U₂O₇, α–Na₂UO₄, NaUO₃, and Na₄UO₄. Information for making α–Na₂UO₄ and NaUO₃ with higher purity was obtained. The phase transition temperature from the α–phase to the β–phase of a sample of Na₂U₂O₇ was clarified to be 658.2 K by thermal expansion measurement.

Technical Material

Synergistic Extraction of Radionuclides from Fuel Debris with a [C₂mim][Tf₂N]–based Extractant

Yuji YADA, Tsuyoshi ARAI

Released: August 15, 2017

[Advance Publication] Released: June 27, 2017

p.161-167

　Water contamination with radionuclides has been found to occur during the processing of fuel debris by the submersion method. In this study, we have investigated the collective removal of radionuclides by the solvent extraction method with the ultimate goal of developing a process for treating such contaminated water. Co(II), Ni(II), Sr(II), Zr(IV), Mo(VI), Cs(I), and Nd (III), which are assumed to be present in the fuel debris, are chosen as the target species for the solvent extraction process. Octyl(phenyl)–N,N–diisobutylcarbamoylmethylphosphine oxide (CMPO)–tributyl phosphate (TBP)–[C₂mim][Tf₂N], which serves as the extractant, exhibits a strong ability to extract Co(II), Ni(II), Sr(II), and Nd (III) at a NaCl concentration of 2.0 mol/dm³, whereas it exhibits high extraction ability towards Co(II), Ni(II), Sr(II), Zr(IV), Cs(I), and Nd(III) at a H₃BO₄ concentration of 0.1 mol/dm³. This extractant also quickly extracts the radionuclides, allowing extraction equilibrium to be reached within 3 min in the cases of Co(II), Ni(II), Sr(II), Zr(IV), Cs(I), and Nd(III). The results of the present study suggest that the radionuclides present in contaminated water can be extracted with high efficiency using CMPO–TBP–[C₂mim][Tf₂N] as the extractant.

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Volume And Issue List
Transactions of the Atomic Energy Society of Japan

Predecessor

Transactions of the Atomic Energy Society of Japan Vol. 16(2017) No. 3

Article

Distributions of ²⁴²Cm and ²⁴⁴Cm Nuclide Compositions as Neutron Sources and Their Neutron Emission Rates in BWR Spent Fuels based on Three-Dimensional Neutron Transport and Burnup Calculations

A Study on Enhancing the Effectiveness of Safety Culture in Nuclear Power Plants

Simulation of Fuel-Coolant Interaction SERENA2 Test based on JASMINE Version 3

Preparation of Sodium Uranates

Technical Material

Synergistic Extraction of Radionuclides from Fuel Debris with a [C₂mim][Tf₂N]–based Extractant

System Maintenance

Announcement From J-STAGE

Journal Tools

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Volume And Issue List Transactions of the Atomic Energy Society of Japan

Predecessor

Transactions of the Atomic Energy Society of Japan Vol. 16(2017) No. 3

Article

Distributions of 242Cm and 244Cm Nuclide Compositions as Neutron Sources and Their Neutron Emission Rates in BWR Spent Fuels based on Three-Dimensional Neutron Transport and Burnup Calculations

A Study on Enhancing the Effectiveness of Safety Culture in Nuclear Power Plants

Simulation of Fuel-Coolant Interaction SERENA2 Test based on JASMINE Version 3

Preparation of Sodium Uranates

Technical Material

Synergistic Extraction of Radionuclides from Fuel Debris with a [C2mim][Tf2N]–based Extractant

System Maintenance

Announcement From J-STAGE

Journal Tools

Share this Title

Volume And Issue List
Transactions of the Atomic Energy Society of Japan

Distributions of ²⁴²Cm and ²⁴⁴Cm Nuclide Compositions as Neutron Sources and Their Neutron Emission Rates in BWR Spent Fuels based on Three-Dimensional Neutron Transport and Burnup Calculations

Synergistic Extraction of Radionuclides from Fuel Debris with a [C₂mim][Tf₂N]–based Extractant