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Kubo, Hirotaka; Sugama, Hideo*; Murakami, Masakatsu*; Osawa, Yukiharu*
Purazuma, Kaku Yugo Gakkai-Shi, 77(12), p.1252 - 1253, 2001/12
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 2.13
10
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.
;
JNC TN9400 2000-031, 15 Pages, 2000/03
JNC-TN9400-2000-031.pdf:0.53MB
For the irradiation performance of metallic fuel, many of the analyses were conducted in USA using EBR-l and EBR-II. ln this study, based on the published data and papers on the above results, the appropriate methods to the evaluation of the irradiation performance of FBR metallic fuel for the design study were considered, as the fbasibility study for FBR. The followings are the targets in this work; (1)deformation of cladding (2)deformation of fuel slug (3)FP gas release (4)fluctuation of the bonding Na level in the fuel pin (5)FCCI
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.
Otagaki, Takao*; *
JNC TJ8420 2000-016, 427 Pages, 2000/03
JNC-TJ8420-2000-016.pdf:18.18MB
no abstracts in English
Ono, Takahiro*; Higuchi, Takanao*; Kazama, T.*; Hashimoto, T.*; Seito, Y.*; Hattori, Tomomi*; Kanamori, Miwa*
JNC TJ1420 2000-005, 257 Pages, 2000/03
JNC-TJ1420-2000-005.pdf:9.53MB
None
Savage, D.*; Arthur, R. C,*; Sasamoto, Hiroshi; Shibata, Masahiro; Yui, Mikazu
JNC TN8400 2000-003, 56 Pages, 2000/01
JNC-TN8400-2000-003.pdf:1.96MB
Geochemical as well as socio-economic issues associated with the selection of potential sites to host a high-level nuclear waste repository have received considerable attention in repository programs in Europe (Belgium, Finland, France, Germany, Spain, Sweden, Switzerland and the U.K.) and North America (Canada and the United States), The objective of the present study is to summarize this international experience with particular emphasis on geochemical properties that factor into the adopted site-selection strategies. Results indicate that the geochemical properties of a site play a subordinate role, at best, to other geotechnical properties in the international site-selection approaches. In countries where geochemical properties are acknowledged in the site-selection approach, requirements are stated qualitatively and tend to focus on associated impacts on the stability of the engineered barrier system and on radionuclide transport. Site geochemical properties that are likely to control the lomg-term stability of geochemical conditions and radionuclide migration behavior are unspecified, however. This non-prescriptive approach may be reasonable for purposes of screeing among potential sites, but a better understanding of site properties that are most important in controlling the long-term geochemical evolution of the site over a range of possible scenarios would enable the potential sites to be ranked in terms of their suitability to host a repository.
*
JNC TJ9400 99-001, 78 Pages, 1999/03
For studies on incineration of long-lived fission products (LLFPs) in a fast reactor, detailed characteristics of reactor core such as incineration performance have to be investigated. Therefore, accurate neutron cross section data of LLFPs become necessary. In the present study, in order to perform the precise measurements of keV-neutron capture cross sections of Tc-99, which is one of most important LLFPs, the details of the Tc-99 sample and the measurements with our experimental facilities were investigated.
Tsujimura, Norio; Shinohara, Kunihiko; Momose, Takumaro
PNC TN8510 98-001, 13 Pages, 1998/07
None
Robin E*
PNC TN1410 98-008, 39 Pages, 1998/05
The 3rd international forum on nuclear non-proliferation was held on 23-25 February 1998 at the Nadao hall of the Shin-kasumigaseki building, in Tokyo, Japan. The forum was organized by the nuclear non-proliferation study group. It was the goal of the nuclear non-proliferation study group to create a forum program that would be both interesting and informative. After some examination, the study group felt that the topics of most interest were the pursuit of the compatibility of the peaceful use of nucleare nergy and the maintenance of the non-proliferation regime. Another area receiving much attention as of late is the disposition of excess weapons plutonium. To address these areas, the forum was divided into three sessions and a special full day session which was devoted completely to the plutonium disposition issue. The first session tackled the peaceful use of plutonium. The second session discussed the issues and the recent situation of the non-proliferation regime. The third ...
Nemoto, K.*; *; Higuchi, Takanao*; Endo, H.*; Ono, Takahiro*; *; *
PNC TJ1250 98-002, 321 Pages, 1998/02
None
Peter, J. Collins
PNC TN9410 97-034, 35 Pages, 1997/04
While in the Reactor Physics Research Section of the Advanced Technology Division at OEC, I participated in the project to construct a data library for the demonstration fast breeder reactor (DFBR). This library would be produced using a combination of evaluated differential cross sections together with integral experimental data for fast reactors, so as to assure sufficiently accurate calculations for the DFBR designs. I had much experience of the design and use of experiments for the large-size cores at ZPPR under the title JUPITER which was performed under the USDOE/PNC joint agreement. My contribution here was mainly in extension of the experimental database to include the very-hard spectrum fast criticals from the Los Alamos National Laboratory (LANL). The data for these cores are described. Our work at ANLW with the GMADJ code, which is similar in effect to the ABLE code that we use at PNC, showed why many experiments are important in this project as well as those in the more obvious Pu/U oxide conventional cores which are of current interest for the DFBR. This point was not appreciated at PNC and is discussed here. The data from the fast spectrum critical experiments made at Los Alamos are described together with information that I have been able to find concerning the uncertainties. The main interest is these experiments has been for prediction of criticality. Consequently, the full covariance information that we would like has not been published. However, the uncertainty in the fuel content is, by far, the major contributor to the uncertainty. The LANL experiments have been a principal leg of the data testing for fast reactors for all versions of ENDF/B in the US. For our work, they provide measurements at Mev energies which are not available from the experiments in the softer-spectrum of the LMFBR.
Takemura, Morio*
PNC TJ9055 97-001, 112 Pages, 1997/03
With use of a standard groupwise shielding design library JSSTDL produced from the latest evaluated nuclear data library JENDL-3.2, experimental analyses for the Axial Shield Experiment (homogeneous and central blockage type shield configurations with B
C or stainless steel shield material) were performed. The results were compared with those obtained by the same analysis method and input data using JSDJ2 library that had been applied consistently to the JASPER experiment analyses. In general, the results with JSSTDL analyses are higher than those by JSDJ2 as were found in analyses in last year for the Radial Shield Attenuation Experiment and the Special Materials Experiment. Consideration was made on the discrepancies between JSSTDL and JSDJ2 analysis results of the Axial Shield Experiment and also those of the sodium configulation in the Radial Shield Attenuation Experiment. The former was done by exchange of macro cross section of each region, and the latter forcused on sodium cross section was done with use of cross section sensitivity analysis method. Compilation of the input data necessary for future reanalyses of important configurations in JASPER experiments, that were selected in previous study in last year, were continued and new data were added into the computer disk holding previous ones.
