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Oizumi, Akito; Fukushima, Masahiro; Gunji, Satoshi; McKenzie, G.*; Amundson, K.*
International Criticality Safety Benchmark Evaluation Project (ICSBEP) Handbook (2022/23 edition) (Internet) , 313 Pages, 2024/11
This benchmark report was compiled to register a critical experiment using the lower-enriched uranium (LEU) system core to the International Criticality Safety Evaluation Project (ICSBEP). The LEU experiment was one of a series of joint experimental project with the Los Alamos Laboratory in the United States from 2015 to 2019 aimed at improving the design accuracy of the accelerator driven system (ADS). This core was loaded alternating highly-enriched uranium (HEU) and natural uranium (NU) to simulate LEU. In addition, a fast neutron spectrum system was constructed with not only HEU and NU but also lead which is part of coolant in the ADS. In this evaluation, it was clarified that the experimental uncertainty for the effective multiplication factor was almost 100 pcm. Moreover, the C/E-1 values of almost -70 pcm and -145 pcm were obtained by the calculation with the continuous energy Monte Carlo code MCNP and the nuclear data ENDF/B-VIII.0 and JENDL-4.0, respectively.
Takahashi, Tone; Yamaguchi, Ikuto*; Hironaka, Kota*; Mochimaru, Takanori*; Koizumi, Mitsuo; Yamanishi, Hirokuni*; Wakabayashi, Genichiro*
Dai-44-Kai Nihon Kaku Busshitsu Kanri Gakkai Nenji Taikai Kaigi Rombunshu (Internet), 4 Pages, 2023/11
To enhance the counter capability against terrorism in major public events, the Japan Atomic Energy Agency (JAEA) has started developments of monitoring technique that covers wide-area and rapidly detects nuclear and radioactive materials. In this project, we are developing widely applicable radiation mapping device by combining with positioning sensors and a network device. Mapping tests inside and outside of the building by using the device to locate areas with high radiation count has been performed. Also, fast neutron detection technique using plastic scintillation detectors is being developed for detection of neutron sources including nuclear materials. Measurement with two rod shaped detector that placed into a cross showed better localizing capability compared with single detector. In this report, recent progresses of the project are given.
Koizumi, Mitsuo; Mochimaru, Takanori*; Hironaka, Kota; Takahashi, Tone; Yamanishi, Hirokuni*; Wakabayashi, Genichiro*
Nuclear Instruments and Methods in Physics Research A, 1042, p.167424_1 - 167424_6, 2022/11
Times Cited Count：4 Percentile：53.47(Instruments & Instrumentation)no abstracts in English
Nakamura, Shoji; Hatsukawa, Yuichi*; Kimura, Atsushi; Toh, Yosuke; Harada, Hideo
Journal of Nuclear Science and Technology, 58(12), p.1318 - 1329, 2021/12
Times Cited Count：0 Percentile：0.00(Nuclear Science & Technology)The present study performed fast-neutron capture cross-section measurement of 
Tc by an activation method using a fast-neutron source reactor "YAYOI" of the University of Tokyo. Technetium-99 samples were irradiated with reactor neutrons using a pneumatic system. Reaction rates of Tc were obtained by measuring decay gamma rays emitted from 
Tc. The neutron flux at an irradiation position was monitored with gold foils. The fast-neutron capture cross section of Tc at neutron energy of 85 keV was derived as 0.432
0.023 barn by using the reaction rates of Tc, evaluated cross-section data and the fast-neutron flux spectrum of the YAYOI reactor. The present study agreed with the evaluated nuclear data library JENDL-4.0.
Katabuchi, Tatsuya*; Iwamoto, Osamu; Hori, Junichi*; Kimura, Atsushi; Iwamoto, Nobuyuki; Nakamura, Shoji; Shibahara, Yuji*; Terada, Kazushi*; Rovira, G.*; Matsuura, Shota*
EPJ Web of Conferences, 239, p.01044_1 - 01044_4, 2020/09
Times Cited Count：2 Percentile：80.29(Nuclear Science & Technology)Komeda, Masao
Bunseki, 2019(10), p.459 - 461, 2019/10
no abstracts in English
Chiba, Satoshi*; Wakabayashi, Toshio*; Tachi, Yoshiaki; Takaki, Naoyuki*; Terashima, Atsunori*; Okumura, Shin*; Yoshida, Tadashi*
Scientific Reports (Internet), 7(1), p.13961_1 - 13961_10, 2017/10
Times Cited Count：51 Percentile：97.91(Multidisciplinary Sciences)Transmutation of long-lived fission products (LLFPs: 
Se, 
Zr, Tc, 
Pd, 
I, and 
Cs) into short-lived or non-radioactive nuclides by fast neutron spectrum reactors without isotope separation has been proposed as a solution to the problem of radioactive wastes disposal. Despite investigation of many methods, such transmutation remains technologically difficult. To establish an effective and efficient transmutation system, we propose a novel neutron moderator material, yttrium deuteride (YD
), to soften the neutron spectrum leaking from the reactor core. Neutron energy spectra and effective half-lives of LLFPs, transmutation rates, and support ratios were evaluated with the continuous-energy Monte Carlo code MVP-II/MVP-BURN and the JENDL-4.0 cross section library. With the YD moderator in the radial blanket and shield regions, effective half-lives drastically decreased from 10
to 10
years and the support ratios reached 1.0 for all six LLFPs. This successful development and implementation of a transmutation system for LLFPs without isotope separation contribute to developing a self-consuming cycle of LLFPs using fast spectrum reactors to reduce radioactive waste.
:Eu
storage phosphors for wide-range neutron detectionSakasai, Kaoru; To, Kentaro; Nakamura, Tatsuya; Ochiai, Kentaro; Konno, Chikara
Proceedings of 2014 IEEE Nuclear Science Symposium and Medical Imaging Conference; 21st International Symposium on Room-Temperature Semiconductor X-ray and 
-ray detectors (NSS/MIC 2014), Vol.3 , p.1834 - 1839, 2016/05
no abstracts in English
Kaji, Yoshiyuki; Matsui, Yoshinori; Kita, Satoshi; Ide, Hiroshi; Tsukada, Takashi; Tsuji, Hirokazu
Nuclear Engineering and Design, 217(3), p.283 - 288, 2002/09
Times Cited Count：4 Percentile：28.64(Nuclear Science & Technology)In the Japan Atomic Energy Research Institute (JAERI), in-pile strain measurement techniques have been developed using the Japan Materials Testing Reactor (JMTR). In order to evaluate the performance of fiber optic grating sensors under irradiation environment, heat-up and performance tests at elevated temperatures before irradiation and in-pile tests were performed in JMTR. It was determined that it is possible to measure strain under irradiation environment below 1
1023n/m (E
1MeV) by a fiber optic grating sensor, because in-pile temperature characteristics were in good agreement with out-of-pile test results.
Oigawa, Hiroyuki
Nihon Butsuri Gakkai-Shi, 56(10), p.749 - 754, 2001/11
no abstracts in English
Shinohara, Nobuo
AIP Conference Proceedings 561, p.223 - 231, 2000/09
no abstracts in English
JNC TN9440 2000-008, 79 Pages, 2000/08
JNC-TN9440-2000-008.pdf:2.33MB
This report summarizes the operating and irradiatlon data of the experimental reactor "JOYO" 35th cycle. Irradiation tests in the 35th cycle are as follows: (1)C-type irradiation rig (C4F) (a)High burnup performance test of advanced austenitic stainless steel cladding fuel pins (in collaboration with France) (2)C-type irradiation rig (C6D) (a)Large diameter fuel pins irradiation tests (3)Core Materials Irradiation Rig (CMIR-5) (a)Cladding tube materials irradiation tests for "MONJU" (4)Structure Materials Irradiation Rigs (SMIR) (a)Decision of material design base standard of structure materials for prototype reactor and large scale reactor (5)Upper core structure irradiation Plug Rig (UPR-1-5) (a)Upper core neutron spectrum effect and accelerated irradiation effect (6)SurVeillance un-instrument Irradiation Rig (SVIR) (a)Confimation of surveillance irradiation condition for "JOYO" (b)Material irradiation tests (based on a contract with universities) The maximum burnup driver assembly "PFD253" reached 67,600 MWd/t (pin average).
JNC TN1440 2000-005, 214 Pages, 2000/08
JNC-TN1440-2000-005.pdf:13.81MB
no abstracts in English
; ; Sakamoto, Naoki; ; Akasaka, Naoaki;
JNC TN9400 2000-095, 110 Pages, 2000/07
JNC-TN9400-2000-095.pdf:13.57MB
The effects of high fluence irradiation and swelling on the transient burst properties of austenitic steel fuel claddings; PNC316 and 15Cr-20Ni stcel, which were irradiated as the MONJU type fuel assemblies (MFA-1&MFA-2) in the FFTF reactor, were investigated. The temperature-transient-to-burst tests were conducted on a total of eight irradiation conditions. Fractographic examination and TEM observation were performed in order to evaluate the effect of high dose irradiation on the transient burst property and the relation between failure mechanism and microstructural change during rapid (ramp) heating. The results of the PIE showed that there was no significant effect of irradiation on the transient burst properties of these fuel claddings under the irradiation conditions examined. the results obtained in this study are as follows; (1)The rupture temperature of the irradiated PNC316 fuel cladding of MFA-1 was as same as that of our previous works for the fluence range up to 2.1310
n/m. There was no noticeable decrease in rupture temperature with increasing fluence in lower hoop stress region(
100MPa). (2)The rupture temperature of the irradiated 15Cr-20Ni fuel cladding of MFA-2 was almost as same as that of as-received cladding for the hoop stress range up to about 200MPa. The rupture temperature did not decrease significantly with fluence. (3)The rupture temperature of the irradiated PNC316 cladding tested at hoop stress 69MPa, which was the design hoop stress for MONJU fuel, was 1055.6
C. This suggested that the design cladding maximum temperature limit for MONJU (830C) was conservative. (4)There was no obvious relation between rupture temperature, swelling and microstructural change during transient heating under the irradiation conditions examined.
Matsubayashi, Masahito; Yoshii, Koji*; Hibiki, Takashi*; Mishima, Kaichiro*
Kashika Joho Gakkai-Shi, 20(Suppl.1), p.325 - 328, 2000/07
no abstracts in English
Takemura, Morio*
JNC TJ9450 2000-002, 112 Pages, 2000/03
JNC-TJ9450-2000-002.pdf:2.55MB
This report is intended to make it easier to apply the measured data obtained from the Gap Streaming Experiment, which was conducted at the Oak Ridge National Laboratory (ORNL) during about two months beginning at the start of March, 1992 as the sixth one of a series of eight experiments planned for the Japanese-American Shielding Program for Experimental Research (JASPER) which was started in 1986. For this reason. the information presented includes specifications and measurement data for all configurations, compositions of all materials, characteristics of the measurement system. and daily-basis records of measurements. The Gap Streaming Experiment was planned to obtain the data of neutron streaming characteristics in the inclosure system above the core of an advanced fast reactor for verification and improvement of the analysis method to be applied to the shielding design. A iron-lined solid or slit concrete assembly was placed, with or without a spectrum modifier forming soft incident neutron spectrum, behind the TSR-II reactor of Tower Shielding Facility. Inserting central cylinders and cylindrical sleeves gave various gap width and offset in the slit concrete assembly. Neutron flux was measured behind the configurations with various types of detectors. The integral neutron flux in wide energy region was measured on radial traverse and on the axis behind the concrete assembly in almost all configurations. Neutron spectrum and fine radial distribution in high energy region was measured further in case of hard incident neutron spectrum, Information presented in this report is based mainly on a report issued by ORNL (ORNL/TM-12140. "Measurements for the JASPER Program Gap Streaming Experiment"). Additional information reported by the assignee is utilized also.
Mori, Tomoaki*; Takemura, Morio*
JNC TJ9450 2000-001, 96 Pages, 2000/03
JNC-TJ9450-2000-001.pdf:2.04MB
This report is intended to make it easier to apply the measured data obtained from the Special Materials Experiment, which was conducted at the Oak Ridge National Laboratory (ORNL) during about a month beginning at the end of June, 1992 as the last one of a series of eight experiments planned for the Japanese-American Shielding Program for Experimental Research (JASPER) which was started in 1986. For this reason. the information presented includes specifications and measurement data for all configurations, compositions of all materials, characteristics of the measurement system. and daily-basis records of measurements. The Special Materials Experiment was planned to obtain the data of neutron attenuation characteristics of selected shielding materials for use in advanced fast reactors. The material of particular interest for the experiment was zirconium hydride that is rich in hydrogen. The mockup slabs for the special materials were preceded by the spectrum modifier behind the TSR-II reactor of Tower Shielding Facility. The layer of zirconium hydride was simulated with a combination of zirconium and polyethylene slabs. The thick layer of polyethylene with no zirconium was installed in some configurations.Neutron flux was measured behind the configurations with various types of detectors. The integral neutron flux in wide energy region was measured in eight configurations and neutron spectrum in high energy region was measured also in almost all configurations. Information presented in this report is based mainly on a report issued by ORNL (ORNL/TM-12277. "Measurements for the JASPER Program Special Materials Experiment"). Additional information reported by the assignee is utilized also.
Shiba, Tsuyoshi*; Kamezaki, Hiroshi*; Yuyama, Tomonori*;
JNC TJ9400 2000-012, 92 Pages, 2000/02
JNC-TJ9400-2000-012.pdf:3.18MB
This research aims to develop a system in which aspects necessary for FBR cycle and overall comparison of evaluation items (economy, safety etc.) are evaluated quantitatively and objectively as a part of Nuclear Cycle development's research project of the FBR cycle for practical use. There are various methods in the decision-making support. In this particular situation, features of each method were evaluated based on the analysis of cases with each method. Subsequently we constructed overall evaluation method by combining Analytic Hierarchy Process (AHP), Multi-attribution Utility Function Method (MUF) and Cut-off Method. This method has variation in evaluation items, transparency in evaluation process and uncompensation. The six aspects of evaluation are economy, effectiveness of resource use, proliferation resistance, environmental effectiveness, safety, and research and development. The evaluation items and the evaluation index of each aspect were hierarchized and the evaluation structure was constructed. In the present study effect function for each evaluation index and pair comparison for examining significance of each item were utilized to select prospective systems for FBR cycle experimentally. The result confirmed reliability of our general assessment system as a decision-making support system for FBR system.
; Kitada, Takanori*; Tagawa, Akihiro; ; Takeda, Toshikazu*
JNC TJ9400 2000-006, 272 Pages, 2000/02
JNC-TJ9400-2000-006.pdf:9.69MB
Investigation was made on the follwing three themes as a part of the improvement of reactor physics analysis method for FBR with various core concept. Part 1: Investigation of Error Estimation of Neutron Spectra in FBR and Suggestions to Improve the Accuracy. In order to improve the spectrum unfolding method used in fast experimental reactor JOYO, a trial was made to evaluate the error in the estimated neutron spectrum, cause by cause. And the evaluated errors were summed up to obtain the most probable and reasonable error as possible. The summed up error was found relatively small compared to the error caused by the uncertainty of cross section data: most of the error in the spectrum unfolding method can be attributed to the error in cross sections. It was also found that the error due to the fission spectrum causes a considerable error in the high energy neutron spectrum which is over several MeV. Part 2: Study on Reactor Physics Analysis Method for Gas-Cooled FBR. In gas-cooled FBR, the portion of coolant channels in core volume is larger than sodium-cooled FBR. This leads to strong neutron streaming effects. For sodium-cooled FBR, several methods were proposed to evaluate the neutron streaming effect, however, these methods can not be used directly to gas-cooled reactor because the direction dependent diffusion coefficient becomes infinitive along the direction pararel to the coolant chammel. In this study, a new method is proposed to evaluate the neutron streaming effect, based on the method taking the axial buckling into consideration, which method was originally proposed by K
hler. Part 3: Study on Reactor Physics Analysis Method for Water-Cooled FBR An investigation was made on low-moderated water-cooled FBR, on the point that the ordinary used analysis method for FBR may give considerable difference in results in such core. In light water reactors, it is well known that the space dependence of self-shielding effect of heavy nuclides are considerably ...
/activation and 3-D Monte Carlo analysesSato, Satoshi; Iida, Hiromasa; Plenteda, R.*; Valenza, D.*; Santoro, R. T.*
Fusion Engineering and Design, 47(4), p.425 - 435, 2000/01
Times Cited Count：9 Percentile：53.37(Nuclear Science & Technology)no abstracts in English
