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Journal Articles

Analyses with latest major nuclear data libraries of the fission rate ratios for several TRU nuclides in the FCA-IX experiments

Fukushima, Masahiro; Tsujimoto, Kazufumi; Okajima, Shigeaki

Journal of Nuclear Science and Technology, 54(7), p.795 - 805, 2017/07

 Times Cited Count:1 Percentile:64.68(Nuclear Science & Technology)

A series of integral experiments was conducted in FCA assemblies with systematically changed neutron spectra covering from the intermediate to fast ones. The experiments provide systematic data of central fission rates for TRU nuclides containing minor actinides, $$^{237}$$Np, $$^{238}$$Pu, $$^{239}$$Pu, $$^{242}$$Pu, $$^{241}$$Am, $$^{243}$$Am, and $$^{244}$$Cm. Latest major nuclear data libraries, JENDL-4.0, ENDF/B-VII.1, and JEFF-3.2, were tested using benchmark models regarding the fission rate ratios relative to $$^{239}$$Pu. For all the libraries, the benchmark tests by a Monte Carlo calculation code show obvious overestimations particularly for the fission rate ratios of $$^{244}$$Cm to $$^{239}$$Pu. Additionally, a large discrepancy about by 20% between the libraries is revealed for the fission rate ratio of $$^{238}$$Pu to $$^{239}$$Pu measured in the intermediate neutron spectrum. The cause of discrepancy is furthermore clarified by sensitivity analyses.

Journal Articles

Fission rate ratios of FCA-IX assemblies as integral experiment for assessment of TRU's fission cross sections

Fukushima, Masahiro; Tsujimoto, Kazufumi; Okajima, Shigeaki

EPJ Web of Conferences (Internet), 111, p.07002_1 - 07002_5, 2016/03

 Times Cited Count:2 Percentile:7.84

In the IX-th series at FCA, central fission rate ratios for TRU such as $$^{237}$$Np, $$^{238}$$Pu, $$^{242}$$Pu, $$^{241}$$Am, $$^{243}$$Am, $$^{244}$$Cm and $$^{239}$$Pu were measured in the seven uranium-fueled assemblies with systematically changed neutron spectra. The FCA-IX assemblies were constructed with simplicity both in geometry and composition. Taking their advantage, benchmark models with respect to central fission rate ratios had been recently established for the assessment of the TRU's fission cross sections. By virtue of these FCA-IX assemblies where the simple combinations of uranium fuel and diluent in their core regions were systematically varied, the neutron spectra of these benchmark models cover those of various reactor types, from fast to epithermal reactors. As an application of these benchmark models, JENDL-4.0 was utilized by a Monte Carlo calculation code. Several results show obvious discrepancies between calculation and experimental values. The benchmark models would be well suited for the assessment and improvement of the nuclear data for TRU's fission cross sections.

Journal Articles

Benchmark models for criticalities of FCA-IX assemblies with systematically changed neutron spectra

Fukushima, Masahiro; Kitamura, Yasunori; Kugo, Teruhiko; Okajima, Shigeaki

Journal of Nuclear Science and Technology, 53(3), p.406 - 424, 2016/03

 Times Cited Count:6 Percentile:17.95(Nuclear Science & Technology)

Journal Articles

Void reactivity evaluation by modified conversion ratio measurements in LWR critical experiments

Yoshioka, Kenichi*; Kikuchi, Tsukasa*; Gunji, Satoshi*; Kumanomido, Hironori*; Mitsuhashi, Ishi*; Umano, Takuya*; Yamaoka, Mitsuaki*; Okajima, Shigeaki; Fukushima, Masahiro; Nagaya, Yasunobu; et al.

Journal of Nuclear Science and Technology, 52(2), p.282 - 293, 2015/02

 Percentile:100(Nuclear Science & Technology)

We have developed a void reactivity evaluation method by using modified conversion ratio measurements in a light water reactor (LWR) critical lattice. Assembly-wise void reactivity is evaluated from the "finite neutron multiplication factor", $$k^ast$$, deduced from the modified conversion ratio of each fuel rod. The distributions of modified conversion ratio and $$k^ast$$ on a reduced-moderation LWR lattice, for which the improvement of negative void reactivity is a serious issue, were measured. Measured values were analyzed with a continuous-energy Monte Carlo method. The measurements and analyses agreed within the measurement uncertainty. The developed method is useful for validating the nuclear design methodology concerning void reactivity.

Journal Articles

Development of a calculation system for the estimation of decontamination effect

Satoh, Daiki; Kojima, Kensuke; Oizumi, Akito; Matsuda, Norihiro; Iwamoto, Hiroki; Kugo, Teruhiko; Sakamoto, Yukio*; Endo, Akira; Okajima, Shigeaki

Journal of Nuclear Science and Technology, 51(5), p.656 - 670, 2014/05

 Times Cited Count:3 Percentile:61.51(Nuclear Science & Technology)

A calculation system for the estimation of decontamination effect (CDE) has been developed to support planning a rational and effective decontamination. The method calculates the dose-rate distribution before and after decontamination, according to the distribution of radioactivity and the decontamination factor (DF), and uses a dose rate reduction factor (DRRF) to estimate the decontamination effect. The results that were calculated by using the CDE were compared with the results of measurements as well as with the results of calculations that were performed using a Monte Carlo particle transport code PHITS. It was found that the CDE successfully reproduced the measured as well as the calculated dose-rate distributions, requiring less than several seconds of calculation time.

Journal Articles

Recent activities in OECD/NEA/NSC and NEA databank

Okajima, Shigeaki

Kaku Deta Nyusu (Internet), (107), p.40 - 46, 2014/02

Outline and structures of Nuclear Science Committee and Databank in OECD/NEA are presented. Recent activities in the committee and the databank are also introduced through the report of NSC meeting and NSC Bureau meeting in 2013.

Journal Articles

Calculation system for the estimation of decontamination effect

Satoh, Daiki; Kojima, Kensuke; Oizumi, Akito; Matsuda, Norihiro; Iwamoto, Hiroki; Kugo, Teruhiko; Sakamoto, Yukio*; Endo, Akira; Okajima, Shigeaki

Transactions of the American Nuclear Society, 109(1), p.1261 - 1263, 2013/11

A computer software, named CDE (Calculation system for Decontamination Effect), has been developed to support planning the decontamination. CDE is programed with VBA (Visual Basic for Applications), and runs on Microsoft Excel with a user friendly graphical interface. It calculates dose rate distributions in a target area before and after the decontamination from a radioactivity distribution and DF (Decontamination Factor), which is a ratio of original radioactivity to remaining one after the decontamination. DRRF (Dose Rate Reduction Factor) is also derived to express the decontamination effect. All the calculation results are visualized on an image of the target area with color map. Owing to its quick calculation speed, CDE is able to investigate the decontamination effect in various cases for a short period. This is very useful to establish a rational decontamination plan before an action.

Journal Articles

Intra-pellet neutron flux distribution measurements in LWR critical lattices

Yoshioka, Kenichi*; Kikuchi, Tsukasa*; Gunji, Satoshi*; Kumanomido, Hironori*; Mitsuhashi, Ishi*; Umano, Takuya*; Yamaoka, Mitsuaki*; Okajima, Shigeaki; Fukushima, Masahiro; Nagaya, Yasunobu; et al.

Journal of Nuclear Science and Technology, 50(6), p.606 - 614, 2013/06

 Times Cited Count:1 Percentile:84.29(Nuclear Science & Technology)

We have developed an intra-pellet neutron flux and conversion ratio distribution measurement method. A foil activation method with special foils was used for the neutron flux distribution measurement. A $$gamma$$-ray spectrum analysis method with special collimators was used for the conversion ratio distribution measurement. Using the developed methods, intra-pellet neutron flux distributions and conversion ratio distributions were measured in critical experiments on a reduced-moderation LWR. Measured values were analyzed with a deterministic method and a Monte Carlo method. The neutron flux distribution measurements and analyses agreed within the range of 1% to 2%. The conversion ratio distribution measurements and analyses were consistent with each other. We found that the measurement methods are useful for the validation of neutron behavior in a fuel pellet, which is known as micro reactor physics.

Journal Articles

Summary of OECD/NEA/NSC expert group on integral experiments for minor actinide management

Okajima, Shigeaki; Fougeras, P.*; Gil, C.-S.*; Glinatsis, G.*; Gulliford, J.*; Iwamoto, Osamu; Jacqmin, R.*; Khomyakov, Y.*; Kochetkov, A.*; Kormilitsyn, M. V.*; et al.

NEA/NSC/DOC(2013)3, p.265 - 278, 2013/04

The Expert Group on Integral Experiments for Minor Actinide Management (EG on IEMAM) was established under OECD/NEA/NSC. The objectives are to review integral experiments for validating MA nuclear data, to recommend additional integral experiments and to propose an international framework to facilitate them from view points of the MA management. The paper summarized the discussion results in the EG on IEMAM as follows: (1) Requirement of nuclear data for MA management, (2) Reviewing existing integral data and identifying specification of missing experimental work to be required, (3) Identifying the bottlenecks and considering possible solutions to them and (4) Proposal of action program for international cooperation.

Journal Articles

Decontamination planning based on computer simulation code CDE

Satoh, Daiki; Oizumi, Akito; Matsuda, Norihiro; Kojima, Kensuke; Kugo, Teruhiko; Sakamoto, Yukio*; Endo, Akira; Okajima, Shigeaki

RIST News, (53), p.12 - 23, 2012/09

Decontamination planning based on a computer simulation code CDE is discussed in this paper. Large amount of radionuclides had been discharged to environment in the accident of the Tokyo Electronic Power Corporation Fukushima Dai-ichi Nuclear Power Plant. CDE has been developed to support planning the decontamination. From the present study, it is validated that the computer simulation is very useful to predict the effect of the scenario before actions, and to plan the decontamination.

JAEA Reports

Development of calculation system for decontamination effect, CDE

Satoh, Daiki; Kojima, Kensuke; Oizumi, Akito; Matsuda, Norihiro; Kugo, Teruhiko; Sakamoto, Yukio*; Endo, Akira; Okajima, Shigeaki

JAEA-Research 2012-020, 97 Pages, 2012/08

JAEA-Research-2012-020.pdf:7.32MB

A computer software, named CDE (Calculation system for Decontamination Effect), has been developed to support planning the decontamination. CDE calculates the dose rates before the decontamination by using a database of dose contributions by radioactive cesium. The decontamination factor is utilized in the prediction of the dose rates after the decontamination, and dose rate reduction factor is evaluated to express the decontamination effect. The results are visualized on the image of a target zone with color map. In this paper, the overview of the software and the dose calculation method are reported. The comparison with the calculation results by a three-dimensional radiation transport code PHITS is also presented. In addition, the source code of the dose calculation program and user's manual of CDE are attached as appendices.

Journal Articles

Development of reactivity feedback effect measurement techniques under sub-critical condition in fast reactors

Kitano, Akihiro; Nishi, Hiroshi; Suzuki, Takayuki; Okajima, Shigeaki; Kanemoto, Shigeru*

Proceedings of International Conference on Physics of Reactors; Advances in Reactor Physics; Linking Research, Industry, and Education (PHYSOR 2012) (CD-ROM), 14 Pages, 2012/04

The "Synthesis Method", a systematic and sophisticated method of sub-criticality measurement, is proposed in this work to ensure the safety margin before operation. The "Synthesis Method" is based on the modified source multiplication method (MSM) combined with the noise analysis method to measure the reference sub-criticality level for MSM. As a result of numerical simulation, it was suggested that a neutron detector located above the core center and three or more neutron detectors located above the radial blanket region enable the measurement of sub-criticality within 10% uncertainty from -0.5 to -2 and within 15% uncertainty for the deeper sub-criticality.

Journal Articles

Reactor physics

Okajima, Shigeaki; Kugo, Teruhiko; Mori, Takamasa

Genshiryoku Kyokasho "Genshiro Butsurigaku", 258 Pages, 2012/03

This textbook is prepared for a lecture of the reactor physics in Nuclear Professional School of The University of Tokyo. In Chapter 1, the characteristics of nucleus and the interaction between nucleus and neutron were described. In Chapter 2, the fission reaction, its chain reaction and a concept on criticality were described. Chapter 3 described about the fundamental study on a diffusion equation to treat the neutron spatial distributions in a substance. In Chapter 4, the application of a diffusion equation was shown to learn the feature of neutron distribution in a substance with homogeneous composition. Chapter 5 described the physics of neutron slowing-down in a substance and Chapter 6 investigated the neutron slowing-downed to become in thermal equilibrium state in the substance. Chapter 7 considered the effect of the practical heterogeneous composition from the homogeneous one to the criticality. The JENDL-4.0 library was referred for the nuclear reaction cross section data.

Journal Articles

Benchmark calculations of sodium-void experiments with uranium fuels at the fast critical assembly FCA

Fukushima, Masahiro; Kitamura, Yasunori; Kugo, Teruhiko; Yamane, Tsuyoshi; Ando, Masaki; Chiba, Go; Ishikawa, Makoto; Okajima, Shigeaki

Progress in Nuclear Science and Technology (Internet), 2, p.306 - 311, 2011/10

Journal Articles

JENDL-4.0 integral testing for fission systems

Okumura, Keisuke; Sugino, Kazuteru; Chiba, Go; Nagaya, Yasunobu; Yokoyama, Kenji; Kugo, Teruhiko; Ishikawa, Makoto; Okajima, Shigeaki

Journal of the Korean Physical Society, 59(2), p.1135 - 1140, 2011/08

 Percentile:100(Physics, Multidisciplinary)

The new version of Japanese evaluated nuclear data library, JENDL-4.0, is tested with integral data of fission systems. This data testing is carried out with a wide range of integral data including the critical benchmarks preserved in the International Handbook of Evaluated Criticality Safety Benchmark Experiments, the experimental data of MOX-fueled critical assemblies relating to the plutonium aging effect, the critical data of various fast critical assemblies and the fast reactors JOYO and MONJU, and the post-irradiation examination data of the pressurized-water reactor Takahama-3 and the fast reactor JOYO. The benchmark calculations are performed with a continuous-energy Monte Carlo code MVP-II or a sophisticated deterministic neutron transport code system. Benchmark calculations with other libraries, such as JENDL-3.3, ENDF/B-VII.0 and JEFF-3.1, are also performed, and differences in performance of these libraries are discussed with a help of sensitivity profiles to nuclear data.

Journal Articles

Measurement and analysis of reactivity worth of $$^{241}$$Am sample in water-moderated low-enriched UO$$_2$$ fuel lattices at TCA

Sakurai, Takeshi; Mori, Takamasa; Suzaki, Takenori*; Okajima, Shigeaki; Ando, Yoshihira*; Yamamoto, Toru*; Liem, P. H.*

Journal of Nuclear Science and Technology, 48(5), p.816 - 825, 2011/05

Journal Articles

JENDL-4.0 benchmarking for fission reactor applications

Chiba, Go; Okumura, Keisuke; Sugino, Kazuteru; Nagaya, Yasunobu; Yokoyama, Kenji; Kugo, Teruhiko; Ishikawa, Makoto; Okajima, Shigeaki

Journal of Nuclear Science and Technology, 48(2), p.172 - 187, 2011/02

 Times Cited Count:28 Percentile:5.33(Nuclear Science & Technology)

Benchmark testing for the newly developed Japanese evaluated nuclear data library JENDL-4.0 is carried out by using a huge amount of integral data. Benchmark calculations are performed with a continuous energy Monte Carlo code and with the deterministic procedure which has been developed for fast reactor analyses in Japan. Through the present benchmark testing using a wide range of benchmark data, significant improvement in the performance of JENDL-4.0 for fission reactor applications is clearly demonstrated in comparison with the former library JENDL-3.3. Much more accurate and reliable prediction for neutronic parameters for both thermal and fast reactors may become possible by using the library JENDL-4.0.

Journal Articles

Experimental reactor physics "past, present and future"; Towards establishment of safety basis in next generation, 6; Status and outlook of fast reactor physics experiments in JAEA

Okajima, Shigeaki

Nippon Genshiryoku Gakkai-Shi, 53(1), p.50 - 53, 2011/01

Fast reactor physics experiments carried out in Bee, FCA, Joyo and Monju has been reviewed. The outlooks of the experiments for the future are also discussed.

Journal Articles

Benchmark calculations of sodium-void experiments with uranium fuels at the fast critical assembly FCA

Fukushima, Masahiro; Kitamura, Yasunori; Kugo, Teruhiko; Yamane, Tsuyoshi; Ando, Masaki; Chiba, Go; Ishikawa, Makoto; Okajima, Shigeaki

Proceedings of Joint International Conference of 7th Supercomputing in Nuclear Application and 3rd Monte Carlo (SNA + MC 2010) (USB Flash Drive), 6 Pages, 2010/10

Journal Articles

Measurement and analysis of reactivity worth of $$^{237}$$Np sample in cores of TCA and FCA

Sakurai, Takeshi; Mori, Takamasa; Okajima, Shigeaki; Tani, Kazuhiro*; Suzaki, Takenori*; Saito, Masaki*

Journal of Nuclear Science and Technology, 46(6), p.624 - 640, 2009/06

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