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Journal Articles

Dogleg duct streaming experiment with 14 MeV neutron source

Morimoto, Yuichi*; Ochiai, Kentaro; Nishio, Takashi*; Wada, Masayuki*; Yamauchi, Michinori*; Nishitani, Takeo

Journal of Nuclear Science and Technology, 41(Suppl.4), p.42 - 45, 2004/03

no abstracts in English

Journal Articles

${it k}$$$_{0}$$ standardization for neutron activation analysis

Yonezawa, Chushiro; Matsue, Hideaki

Bunseki, 2004(2), p.75 - 82, 2004/02

no abstracts in English

JAEA Reports

Investigation on the JRR-3 high performance cold neutron source

Kumai, Toshio; Liem, P. H.*; Horiguchi, Yoji

JAERI-Tech 2002-023, 49 Pages, 2002/03

JAERI-Tech-2002-023.pdf:2.1MB

no abstracts in English

Journal Articles

Benchmark experiment on LiAlO$$_{2}$$, Li$$_{2}$$TiO$$_{3}$$ and Li$$_{2}$$ZrO$$_{3}$$ assemblies with D-T neutrons; Leakage neutron spectrum measurement

Murata, Isao*; Nishio, T.*; Kokooo*; Kondo, T.*; Takagi, H.*; Nakano, D.*; Takahashi, Akito*; Maekawa, Fujio; Ikeda, Yujiro; Takeuchi, Hiroshi

Fusion Engineering and Design, 51-52(Part.B), p.821 - 827, 2000/11

 Times Cited Count:7 Percentile:53.59

no abstracts in English

JAEA Reports

Experimental analyses results on the BFS 58-1-I1 critical assemblies

; Sato, Wakaei*; Iwai, Takehiko*

JNC-TN9400 2000-096, 113 Pages, 2000/06

JNC-TN9400-2000-096.pdf:3.1MB

This report describes the updated analyses results on the BFS-58-1-I1 core. The experiment was conducted at BFS-2 of Russian Institute of Physics & Power Engineering (IPPE). The central region is "non-Uranium fuel zone", where only Pu can induce fission reaction. The non-U zone is surrounded by MOx fuel zone, which is surrounded by U0$$_{2}$$ fuel zone. Sodium is used for simulating the coolant material. As it was found that the lattice pitch had been incorrectly understood in the past analyses, all items have been re-calculated using the corrected number densities. Furthermore, significantly softened neutron spectrum in the central region caused problems in applying the plate-stretch model that has been established for fast reactor cores through JUPITER experimental analyses. Both keeping the pellet density and using SRAC library for the elastic cross section for lighter nuclides allow us to obtain reasonable analysis accuracy on the spectral indices that were measured at the center of the core. Application of such a cell model was justified through comparison among various cell models using continuous energy Monte-Carlo code MVP. It is confirmed that both the MOX zone and the U0$$_{2}$$ zone can be correctly evaluated by the plate-stretch model. Based on the updated cell calculation, both the effective multiplication factor (k-eff)and the spectral indexes agree well with the measured values. The transport and mesh-size correction is made for the k-eff evaluation. Those results also agree well within reasonable difference between those obtained by IPPE and CEA, which were obtained by using sub-group method or continuous-energy Monte Carlo code. Evaluation by the nuclear data library adjustment confirmed that the analyses results of the BFS-58-1-I1 core have no significant inconsistency with JUPITER experimental analyses results. Those results are quite important for starting BFS-62 cores, which will be analyzed in the framework of supporting program for Russian ...

Journal Articles

A Monte Carlo program for estimating characteristics of neutron calibration fields using a Pelletron accelerator

Yoshizawa, Michio; Saegusa, Jun; Yoshida, Makoto; Sugita, Takeshi*

Proceedings of 10th International Congress of the International Radiation Protection Association (IRPA-10) (CD-ROM), 4 Pages, 2000/05

no abstracts in English

JAEA Reports

ComparaUve analyses on nuclear charaderistics of water-cooled breeder cores

; Sato, Wakaei*;

JNC-TN9400 2000-037, 87 Pages, 2000/03

JNC-TN9400-2000-037.pdf:3.48MB

ln order to compare the nuclear characteristics of water-cooled bleeder cores with that of LMFBR, MOX fuel cell models are established for boiling and non-boiling LWR, non-boiling HWR and sodium-cooled reactor. Frst, the comarison is made between the heterogeneous cell calculation results by SRAC and those by SLAROM. The results show some differences as for neutron energy spectrum, one-grouped cross section and conversion ratio due to the different grouped cross section library (both are based on JENDL-3.2, though) used for each code, however, the difference is acceptably small for grasping the basic characteristics of the above-mentioned cores. Second, using the SLAROM code, main core parameters such as mean neutron energy, ratio of fast neutron and $$eta$$-value, are analyzed. The comparison between the cores show that softened neutron spectrum by the scattering effect of hydrogen or heavy hydrogen increase the contribution of nuclear reaction (especially for neutron capture reaction rather than fission reaction) in lower energy region comparing with LMFBR. ln order to overcome the effect, tighter lattice than LMFBR is necessary for water-cooled cores to realize the breeding of fissile nuclides. Third, effects of Pu isotopic composition on the breeding ratio are evaluated using SRAC burnup calculation. From the results, it is confirmed that degraded Pu (larger ratio of Pu-240) show the larger breeding ratio. At last, sensitivity analyses are made for k-effective and main reaction ratios. As for k-effective, using a temporary covariance data of JENDL-3.2, uncertainty resulting from the cross sections' error is analyzed for a boiling LWR and a sodium-cooled reactor. The boiling LWR core shows larger sensitivity in lower energy region than the sodium-cooled reactor (especially for the energy region lower than 1kev), And, 18-group analysis that is considered acceptably good for LMFBR analysis, should not be enough for accurate sensitivity estimation of ...

JAEA Reports

JASPER Experimental data book (VII); Gap streaming Experiment

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.

JAEA Reports

Measurement of fission cross section and fission neutron spectrum of Np-237 by an advanced technique

*

JNC-TJ9400 2000-007, 46 Pages, 2000/03

JNC-TJ9400-2000-007.pdf:2.16MB

For fission cross section and prompt fission neutron spectrum, which largely influence core characteristics of a fast reactor, we have performed experimental and analytical studies for developing an advanced technique to measure absolute fission cross section and neutron fission spectrum for actinide nuclides such as Np237. As the results, we could develop an advanced technique, which combines a normalization technique for the well-known differential cross section and a correction method by a Monte-Carlo code for sample effects. This advanced technique accurately provides both absolute fission cross section and prompt fission neutron spectrum individually. By employing this technique, in this study, we have measured for three actinides ($$^{237}$$Np, $$^{232}$$Th and $$^{233}$$U), then, have obtained the fission cross sections and fission spectrum parameter data for those nuclides. Furthermore, we have also performed an analytical study to examine sensitivity of fission spectrum parameter to core multiplication factor by using the standard calculation code for a first reactor.

JAEA Reports

Study on improvement of reactor physics analysis method for FBRs with various core concept

*; 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$"o$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 ...

JAEA Reports

Analysis of the Rossendorf SEG experiments using the JNC route for reactor calculation

Dietze, K.

JNC-TN9400 99-089, 20 Pages, 1999/11

JNC-TN9400-99-089.pdf:0.66MB

The integral experiments performed at the Rossendorf fast-thermal coupled reactor RRR/SEG have been reanalyzed using the JNC route for reactor calculation JENDL3.2/SLAROM / CITATION / JOINT / PERKY. The Rossendorf experiments comprise sample reactivity measurements with pure fission products and structural material in five configurations with different neutron and adjoint spectra. The shapes of the adjoint spectra have been designed to get high sensitivities to neutron capture or the scattering effect. The calculated neutron and adjoint spectra are in good agreement with former results obtained with the European route JEF2.2/ECCO/ERANOS. The C/E-values of the central reactivity worths of samples under investigation are given. Deviations in the results of both routes are due to the different libraries, codes, and self-shielding treatments used in the calculations. Results outside of the error are discussed.

Journal Articles

Steady state neutron source by reactor

Morii, Yukio; Isshiki, Masahiko

Kessho Kaiseki Handobukku, p.111 - 114, 1999/09

no abstracts in English

JAEA Reports

Evaluation of covariance for fission neutron spectra

Kawano, Toshihiko*; Osawa, Takaaki*; Shibata, Keiichi; *

JAERI-Research 99-009, 43 Pages, 1999/02

JAERI-Research-99-009.pdf:1.36MB

no abstracts in English

Journal Articles

Benchmark experiment on vanadium assembly with D-T Neutrons; Leakage neutron spectrum measurement

Kokooo*; Murata, Isao*; Nakano, D.*; Takahashi, Akito*; Maekawa, Fujio; Ikeda, Yujiro

Fusion Technology, 34(3), p.980 - 984, 1998/11

no abstracts in English

Journal Articles

Users' requirements for IFMIF

Noda, Kenji; Ehrlich, K.*; Jitsukawa, Shiro; Moeslung, A.*; Zinkle, S.*

Journal of Nuclear Materials, 258(263), p.97 - 105, 1998/10

 Times Cited Count:19 Percentile:17.19

no abstracts in English

JAEA Reports

Measurement of fast neutron induced fission cross section of minor-actinide

*

PNC-TJ9601 98-002, 115 Pages, 1998/03

PNC-TJ9601-98-002.pdf:2.57MB

In fuel cycles with recycled actinide, core characteristics are largely influenced by minor actinide (MA: Np, Am), Accurate nuclear data of MA such as fission cross section are required to estimate the effect of MA with high accuracy. In this study, fast neutron induced fission cross section of MA was measured using Dynamitron Accelerator in Tohoku University. New or improved techniques and tools with high precision and fast timing capability were developed for this study. Those are as follows: (1)Development of a sealed fission chamber,(2)Intensification of Li neutron target, (3)Improvement of time-resolution of Time-of-Right (TOF) electronic circuit, (4)Introduction of MA (Np237, Am241 and Am243) samples with large sample mass and (5)Introduction of a U235 sample with high purity. Using these improved tools and samples, fission cross section of Np237 was measured between 10 to 100 keV. On the other hand, averaged fission cross section for Maxwell distribution spectrum with kt=25.3 keV was measured for Am241 and Am243.

Journal Articles

Source neutron spectra of radiation streaming measurement in a labyrinth

Sakamoto, Yukio; Tanaka, Susumu; ; Nakane, Yoshihiro; Meigo, Shinichiro; Takada, Hiroshi; Tanaka, Shunichi; Takada, M.*; Kurosawa, Tadahiro*; Nakamura, Takashi*; et al.

Genshikaku Kenkyu, 41(3), p.95 - 99, 1996/06

no abstracts in English

Journal Articles

Measurements of the neutron spectrum from a thick target

Meigo, Shinichiro; Takada, Hiroshi; ; Sasa, Toshinobu; Tanaka, Susumu; Minato, Kazuo; *

Genshikaku Kenkyu, 41(3), p.49 - 57, 1996/06

no abstracts in English

Journal Articles

An Intranuclear cascade model analysis of the proton incident experiment

Takada, Hiroshi

Genshikaku Kenkyu, 41(3), p.39 - 47, 1996/06

no abstracts in English

JAEA Reports

Irradiation test of HAFM and tag gas samples at the standard neutron field of "YAYOI"

*

PNC-TJ9602 96-004, 49 Pages, 1996/03

PNC-TJ9602-96-004.pdf:1.53MB

To check the accuracy of helium accumulation neutron fluence monitoring (HAFM) technique and the applicability of tag gas activation analysis to fast reactor failed fuel detection, these samples were irradiated at the standard neutron field of the fast neutron source reactor "YAYOI" (Nuclear Engineering Research Laboratory, University of Tokyo). The HAFM samples of 93% enriched B powder of 1 mg contained in a V capsule were set up in the reactor core center (Glory hole :Gy) and another samples including V encapsulated natural B powder of 10 mg, natural B chips and 96% enriched $$^{6}$$LiF thermoluminescence dosimeters were in the leakage neutron field from the reactor core (Fast column :FC). These neutron fields were monitored by the activation foils consisting of Al, Fe, Co, Ni, Cu, Ti, In, Au, $$^{235}$$U, $$^{237}$$Np etc., which were used to derive the neutron flux and spectrum. At the end of March in 1996, the irradiated neutron fluence in the energy more than 0.1Mev reached up to $$sim$$ 1.1$$times$$10$$^{17}$$ n/cm$$^{2}$$ at Gy and $$sim$$5.4$$times$$10$$^{14}$$ n/cm$$^{2}$$ at FC, respectively. Two kinds of tag gas samples, which are the mixed gases of isotipically adjusted Xe and Kr contained in stainless steel (SUS) capsules, were irradiated twice at Gy; one is up to $$sim$$9kWh and the other $$sim$$7kWh, and after the irradiation, $$gamma$$-ray spectra were measured for each sample. Through comparison with the $$gamma$$-ray spectrum from dummy capsule of no tag gas, the $$gamma$$-ray peaks of $$^{79}$$Kr, $$^{125}$$Xe etc., which would be produced from tag gas activation, were able to be clearly identified.

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