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JAEA Reports

Cost-benefit analysis on FR cycle R&D for the world

Kawasaki, Hirotsugu; Yasumatsu, Naoto*; Kubota, Sadae*; Shiotani, Hiroki; Ono, Kiyoshi

JAEA-Research 2006-001, 60 Pages, 2006/02

JAEA-Research-2006-001.pdf:4.37MB

no abstracts in English

JAEA Reports

Current profiles and major disruptions in a lower-hybrid current drive tokamak

Uehara, Kazuya; Nagashima, Takashi*

JAEA-Research 2006-002, 15 Pages, 2006/02

JAEA-Research-2006-002.pdf:0.69MB

The radial rf current profile of the lower-hybrid current drive tokamak is estimated using an rf power damping model for various Gaussian type nz spectrum, and the stability of the major disruption is examined using the criterion of the tearing mode instability. The value of nzc determines the spatial damping of the rf power and in resultant the rf current profile. The value of hz determine the extent of the damping of the rf power and in resultant the absolute value of the rf current. The peaking and flattening rf current profile to be free from the disruption are obtained when the rf power does not damp at the boundary region and when the extent of damping is relatively small even if the rf power begins to damp at relatively boundary region, whose rf spectrum must be having relatively large phase velocity and must be not so broad.

JAEA Reports

Dissolution of Sr precipitated with platinum group metals at the precipitation step by denitration in the 4-group partitioning process

Fujiwara, Takeshi; Morita, Yasuji

JAEA-Research 2006-003, 22 Pages, 2006/02

JAEA-Research-2006-003.pdf:1.25MB

The precipitation method by denitration with formic acid is adopted in the 4-Group Partitioning Process for recovery of Tc and platinum group metals (PGM) from high-level liquid waste. A part of Sr and some other elements are precipitated with Tc and PGM at the precipitation step by denitration. As Sr is one of the target elements of the 4-Group Partitioning Process, it is necessary to recover the precipitated Sr from Tc and PGM fraction. The present study deals with the process for dissolving the precipitate of Tc and PGM at the precipitation step to recover the precipitated Sr from Tc and PGM fraction. It was possible to dissolve the precipitated Sr thoroughly by 0.010 mol/dm$$^{3}$$ nitric acid without dissolving PGM precipitate. Ba and Ni were contained in dissolved Sr solution. Any Sr was not appeared in the dissolved solution of PGM precipitate which remained after the Sr dissolution by 0.010 mol/dm$$^{3}$$ nitric acid. The pH of the dissolved Sr solution was about 2. As the appropriate pH is 5 or higher to adsorb Sr on inorganic ion exchangers, it is demanded to adjust the pH of dissolved Sr solution for next treatment. Precipitation of Sr did not occur during adjusting the pH of dissolved Sr solution to neutral region with sodium hydroxide solution.

JAEA Reports

Japan Sea expeditions for studies on water circulation and transport processes of radionuclides (Contract research)

Togawa, Orihiko; Ito, Toshimichi; Kobayashi, Takuya; Otosaka, Shigeyoshi; Suzuki, Takashi

JAEA-Research 2006-004, 132 Pages, 2006/02

JAEA-Research-2006-004.pdf:7.84MB

no abstracts in English

JAEA Reports

Water experiment on gas entrainment in reactor vessel using 1/1.8th scaled model; Evaluation of onset condition and mechanism

Kimura, Nobuyuki; Ezure, Toshiki; Nakayama, Okatsu; Tobita, Akira; Ito, Masami*; Kamide, Hideki

JAEA-Research 2006-005, 45 Pages, 2006/03

JAEA-Research-2006-005.pdf:15.66MB

An innovative sodium cooled fast reactor has been investigated in a frame work of the FBR feasibility study. One of the thermal hydraulic issues in this design is gas entrainment at free surface in the reactor vessel. Dipped plates (D/P) are set below the free surface in order to prevent the gas entrainment. We performed an 1/10th scaled model water experiment for the upper plenum of reactor vessel and flow optimization was done to reduce flow velocity near the free surface. However, previous studies showed that the gas entrainment depends on model scale. Then an 1/1.8th scaled model was also planned to confirm the phenomena in an enough large model. As a test section, 90 degree sector and region between the free surface and the D/P was modeled by 1/1.8th scale. Boundary conditions at D/P gaps and radial cross sections of sector ends were obtained by the 1/10th scaled full sector model. The gas entrainment was not observed in the model under the velocity condition of reactor full power operation at water levels higher than 3% of the normal height from the D/P in the case of double D/Ps geometry (current design). As for the case of single D/P geometry, it was found by the visualization and the velocity measurement that the gas entrainment occurred as the circumferential velocity increased at the water level higher than 50% of the normal height condition. It is shown that the gas entrainment in the reactor vessel will be eliminated in the current design approach.

JAEA Reports

Design study on sodium-cooled reactor; Results of the studies in 2004 (Joint research)

Hishida, Masahiko; Murakami, Tsutomu*; Kisohara, Naoyuki; Fujii, Tadashi; Uchita, Masato*; Hayafune, Hiroki; Chikazawa, Yoshitaka; Usui, Shinichi; Ikeda, Hirotsugu; Uno, Osamu; et al.

JAEA-Research 2006-006, 125 Pages, 2006/03

JAEA-Research-2006-006.pdf:11.55MB

In Phase I of the "Feasibility Studies on Commercialized Fast Reactor Cycle Systems (F/S)", an advanced loop type reactor has been selected as a promising concept of sodium-cooled reactor, which has a possibility to fulfill the design requirements of the F/S. In Phase II, design improvement for further cost reduction and the establishment of the plant concept has been performed. In this study, reactor core design and large-scale plant design have been performed by adopting the modified fuel assembly with inner duct structure and double-wall straight tube steam generator (SG), which concepts were chosen at the interim review of FY 2003. For this SG, safety logics have been studied and the structural concept has been established. And the plant designs improving the in-service inspection (ISI) and repair capability have been performed. Furthermore, elaborate confirmation of the design has been performed reflecting the development of elemental technology, back-up concepts have been proposed. Besides, cost reduction measures have been studied by reducing reactor grade materials, introducing autonomous standardizations, simplifying the design due to deregulation and adopting systemized standards for BOP and NSSS. From now on, reflecting the results of elemental experiments, in-depth design studies and examination of critical issues will be carried out and the plant concept will accomplish in preparation for the final evaluation in Phase II.

JAEA Reports

Conceptual design study of Cu bonded steam generator; Pressurized crack propagation experiments at the room temperature

Chikazawa, Yoshitaka; Aizawa, Kosuke; Konomura, Mamoru

JAEA-Research 2006-007, 114 Pages, 2006/03

JAEA-Research-2006-007.pdf:15.6MB

In the feasibility study of commercialized fast reactor cycle systems of JAEA, we make a concept of a sodium cooled reactor without secondary sodium circuits. And a sodium cooled reactor with Cu bonded steam generator is one of promising concept has been investigated. In the FY2004 study, pressurized tube experiments using the 3$$times$$3 array specimen imitated the reference tube geometry are carried out aiming to clarify crack propagation behavior in the reference steam generator. Compared with the results of the bend experiments in the FY2003, it is understood that the crack initiation and the crack propagation are greatly obstructed by the effect of the specimen shape. The reference tube geometry was optimized to pursue the economic. As a result of optimization, the construction cost of reactor cooling system with Cu bonded steam generators is 0.7 times as much as that of an ordinary sodium cooled reactor with secondary sodium circuits. Condition for crack propagation in the reference tube geometry is analyzed by using the result of the bend experiments in the FY2003. From the results of the bend experiments and the analysis, it was understood that there was a possibility that the crack reaches the Cu layer. But, when the crack reaches the Cu layer, it was shown that the crack propagation stopped.

JAEA Reports

Evaluation of prediction accuracy for $$^{238}$$U Doppler effect measured in FCA LWR simulating cores; Analysis with JENDL-3.3 library and SRAC system (Joint research)

Kawasaki, Kenji*; Ando, Masaki; Okajima, Shigeaki; Fukushima, Masahiro; Nakano, Makoto*; Matsumoto, Hideki*

JAEA-Research 2006-008, 40 Pages, 2006/03

JAEA-Research-2006-008.pdf:3.6MB

Analysis was performed to evaluate prediction accuracy of a neutronics code system for thermal reactor; the SRAC system with the use of the latest nuclear data library JENDL-3.3 for the $$^{238}$$U Doppler effect measured in the uranium fueled (FCA-XXI-1D2) and MOX fueled (XXII-1 series) cores. The results of the analysis with the diffusion theory showed overestimation by up to +11%. In relatively soft neutron spectra, however, the calculated values agreed with the experimental ones within the experimental errors.

JAEA Reports

Development of measuring apparatus for noncondensable gas concentration

Takemoto, Masafumi; Nakamura, Hideo; Owada, Akihiko; Osaki, Hideki

JAEA-Research 2006-009, 48 Pages, 2006/03

JAEA-Research-2006-009.pdf:3.8MB

Nitrogen (N$$_{2}$$)gas used for the pressurization of accumulator (ACC) tanks may flow into the PWR primary system during LOCAs after the ACC coolant injection is completed. Since N$$_{2}$$ gas may travel to and accumulates in reactor vessel top and U-tubes in steam generators (SGs), primary cooling and depressurization via SG secondary-side depressurization would become ineffective because of the degradation of condensation heat transfer in SG U-tubes. Quantitative measurement of gas accumulation is necessary to clarify the influences of gas onto such degradation in the heat transfer and thus the primary depressurization. However, direct measurement method of non-condensable gas concentration in steam has not been established. An apparatus to directly measure gas concentration in high-temperature steam was developed to measure gas concentration in vessel upper head and SG U-tubes during LOCA experiments using ROSA/LSFT. The developed apparatus is primarily composed of zirconia oxygen sensor and turbine meter, enabling to deal with small amount of steam-gas mixture at high temperature when air is used to pressurize ACC tanks instead of N$$_{2}$$ gas. This report describes the developed apparatus and its operation method with several test results for the confirmation of oxygen gas measurement capability and applicability of the apparatus to the ROSA/LSTF experiments.

JAEA Reports

Development of tissue substitutes for absorbed dose measurements in neutron dosimetry

Tsuda, Shuichi; Endo, Akira; Yamaguchi, Yasuhiro

JAEA-Research 2006-010, 47 Pages, 2006/02

JAEA-Research-2006-010.pdf:3.27MB

no abstracts in English

JAEA Reports

Study on applicability of DIDPA solvent to Talspeak method

Fujiwara, Takeshi; Morita, Yasuji

JAEA-Research 2006-011, 24 Pages, 2006/03

JAEA-Research-2006-011.pdf:1.43MB

In the 4-Group Partitioning Process developed in Japan Atomic Energy Research Institute (presently: Japan Atomic Energy Agency), the transuranic element and the rare earth elements are extracted from high-level liquid waste by diisodecylphosphoric acid (DIDPA). The concentration of the rare earth elements are about 50 times higher than that of Am and Cm in high-level liquid waste. It is, therefore, necessary to separate each other for the transmutation of Am and Cm, or volume reduction of the waste form of the long-lived nuclide. On the other hand, the Talspeak method is a separation method by the solvent extraction that gives selective stripping of Am and Cm from the solvent that contains the rare earth elements using di-2-ethylhexylphosphoric acid (DEHPA) as an extractant. In the present study, application of the DIDPA to the Talspeak method was examined for various conditions to separate Am from the rare earth elements by the batch examination.

JAEA Reports

Effect of initial heat treatment on tensile properties and charpy impact properties of reduced-activation ferritic steel F82H irradiated by neutrons

Wakai, Eiichi

JAEA-Research 2006-012, 51 Pages, 2006/03

JAEA-Research-2006-012.pdf:3.38MB

Effects of initial heat treatments on irradiation hardening and embrittlement of F82H steel were mainly examined by the experiments of neutron irradiations. From the analysis of the changes of yield stress and ductile-brittle transition temperature due to irradiation, it was found that the controll of heat treatment of tempering before irradiation was very useful for the improvement of irradiation hardening and embrittlement.

JAEA Reports

Analyses of core Shroud materials by three dimensional atom probe (Contract research)

Kondo, Keietsu; Nemoto, Yoshiyuki; Miwa, Yukio; Kaji, Yoshiyuki; Tsukada, Takashi; Nagai, Yasuyoshi*; Hasegawa, Masayuki*; Okubo, Tadakatsu*; Hono, Kazuhiro*

JAEA-Research 2006-013, 39 Pages, 2006/12

JAEA-Research-2006-013.pdf:4.57MB

There has been an increasing number of stress corrosion cracking (SCC) incidents on low carbon austenitic stainless steels used in boiling water reactor (BWR) environments. To reveal the acceleration factor of intergranular crack propagation from the viewpoint of solute distribution in stainless steels, the material extracted from a core shroud of Japanese BWR was analyzed by the three dimensional atom probe (3DAP), which has the highest spatial resolution among the various microanalytical techniques. It was revealed by statistical analysis on 3DAP data that solute elements, such as Fe, Cr, Ni, Mo, Mn, Si, are randomly distributed in matrix of the shroud material. This result means that solute was not segregated or precipitated and was not form spinodal decomposition during the service. The concentration profile in the vicinity of grain boundary obtained from 3DAP dataset showed the random distribution of Cr. This result shows that degradation of the corrosion resistance induced by depletion of Cr was not responsible for the crack propagation along grain boundaries in low carbon stainless steel. On the other hand, enrichment of Mo and Si was observed at grain boundary. The width of the enriched zone was about 2 nm across the grain boundary, and the concentration of those elements could be much higher than the concentration obtained by field emission transmission electron microscopy/energy dispersive X-ray spectroscopy (FE-TEM/EDS). Therefore, it is necessary to study about the effects of enrichment of Mo and Si as a potential contributor to SCC.

JAEA Reports

Study on the spatial resolution of position sensitive neutron gas detector with individual readout

Tanaka, Hiroki; Nakamura, Tatsuya; Yamagishi, Hideshi; Soyama, Kazuhiko; Aizawa, Kazuya

JAEA-Research 2006-014, 17 Pages, 2006/03

JAEA-Research-2006-014.pdf:1.3MB

The position sensitive neutron detector for neutron scattering experiments using intense pulsed neutron source is needed. To meet the detector performance requirements such as fast temporal response $$<$$ 1$$mu$$s, spatial resolution $$<$$ 1mm and detection efficiency $$>$$ 80 % for thermal neutron, we have been developing the micro-strip gas chamber with individual readout. In this paper, we report the simulation results of the spatial resolution of the chamber. This simulation can calculate the spatial resolution by using the distribution of neutron beam and gas condition. The simulation results agreed with the experimental data using the collimated neutron beam. In addition, the results of good uniformity about the spatial resolution and the peak counts were described.

JAEA Reports

Status of assessment tools on the performance guarantee contents of backfill, bulkhead, tunnel and pit

Kawakami, Susumu; Fujita, Tomoo; Yui, Mikazu

JAEA-Research 2006-015, 25 Pages, 2006/03

JAEA-Research-2006-015.pdf:2.66MB

no abstracts in English

JAEA Reports

Development of measurement technique for charged-particle emission double-differential cross section using pencil-beam neutron source (Cooperative research)

Kondo, Keitaro; Ochiai, Kentaro; Kubota, Naoyoshi; Nishitani, Takeo; Murata, Isao*; Miyamaru, Hiroyuki*; Takahashi, Akito*

JAEA-Research 2006-016, 50 Pages, 2006/03

JAEA-Research-2006-016.pdf:6.25MB

Charged-particle emission double- differential cross section (DDXc) is quite important to estimate nuclear heating, material damages of a fusion reactor. We have developed a new technique for detailed measurement of DDXc. The technique overcomes fundamental difficulties of DDXc measurement with a pencil-beam neutron source and a counter telescope consisting of silicon surface barrier detectors. A superior S/N ratio, fine energy and angular resolutions, a wide detection energy range, and an excellent particle discrimination are realized together with a reasonable measurement time. In order to confirm the validity of the spectrometer, measurements of the emitted $$alpha$$-particle from $$^{27}$$Al(n,x$$alpha$$) reaction and the recoiled proton from $$^{1}$$H(n,n) reaction are carried out. Based on the results, we conclude the validity and the superiority of the present spectrometry technique.

JAEA Reports

Study of hydraulic behavior for reactor upper plenum in sodium-cooled fast reactor; Verification analysis of water experiment and applicability of vortex prediction method

Fujii, Tadashi; Chikazawa, Yoshitaka; Konomura, Mamoru; Kamide, Hideki; Kimura, Nobuyuki; Nakayama, Okatsu; Ohshima, Hiroyuki; Narita, Hitoshi*; Fujimata, Kazuhiro*; Itooka, Satoshi*

JAEA-Research 2006-017, 113 Pages, 2006/03

JAEA-Research-2006-017.pdf:14.98MB

A conceptual design study of the sodium-cooled fast reactor is in progress in the Feasibility Study on Commercialized Fast Reactor Cycle Systems. Reduced scale water experiments are being performed in order to clarify the flow pattern in the upper plenum of the reactor which has higher velocity condition than the past design. In this report, the hydraulic analyses of the water experiments using the general-purpose thermal hydraulic analysis program were executed; and the applicability to evaluation of flow pattern and vortex cavitations for the designed reactor was examined. (1) Steady-state analyses under the Froude number similar condition were carried out for the 1/10th reduced scale plenum experiments. Analyses results reproduced the characteristic flow patterns in the upper plenum, such as gushed flow from the inside of the upper internal structure to reactor vessel wall and the jet flow from the slit of the upper internal structure. Further, it was confirmed that the calculated flow pattern of a designed reactor system agreed with that of the water experiment qualitatively. Moreover, the influence which setting of numerical solution and boundary condition etc. in analyzing causes to flow pattern in the plenum became clear. (2) The distribution of the vortices under the dipped plate region in the 1/10th plenum model was evaluated using the prediction method of a submerged vortex which is based on the stretching vortex theory. In case of the same velocity condition as the reactor, it identified the two vortices which were sucked into the hot leg piping from the cold leg piping wall as the submerged vortex cavitations. From this analysis result, it confirmed that the submerged vortex cavitations, which may occur in the reactor upper plenum steadily, could be identified using this prediction method.

JAEA Reports

An Experimental study on effective depressurization actions for PWR vessel bottom small break LOCA with HPI failure and gas inflow, ROSA-V test SB-PV-04

Suzuki, Mitsuhiro; Takeda, Takeshi; Asaka, Hideaki; Nakamura, Hideo

JAEA-Research 2006-018, 140 Pages, 2006/03

JAEA-Research-2006-018.pdf:7.14MB

A small break loss-of-coolant accident (SBLOCA) experiment was conducted at the Large Scale Test Facility of ROSA-V program to study effects of accident management (AM) measures on core cooling, which are important in case of total failure of high pressure injection system during an SBLOCA at a pressurized water reactor (PWR). The experiment (SB-PV-04) simulating a PWR vessel bottom SBLOCA with a rupture of ten instrument-tubes clarified that rapid depressurization action by full-opening of relief valves and supplying auxiliary feedwater were effective to avoid core uncovery through actuation of low pressure injection system irrespective of significantly degraded depressurization by non-condensable gas inflow from the accumulator tanks. It is clarified that the effective core cooling was established by the rapid primary cooling which contributed to preserve larger primary coolant mass than in the previous experiment (SB-PV-03) which was conducted with smaller primary cooling rate of -55 K/h as another AM action and resulted in core heatup.

JAEA Reports

Quantitative analysis on the surface of Lithium-6 enriched lithium titanate using proton RBS (Cooperative research)

Kubota, Naoyoshi; Fujiwara, Yoshio*; Okumura, Kazutaka*; Ochiai, Kentaro; Kitamura, Akira*; Furuyama, Yuichi*; Taniike, Akira*; Nishitani, Takeo

JAEA-Research 2006-019, 15 Pages, 2006/06

JAEA-Research-2006-019.pdf:0.75MB

Surface analysis of lithium-6 ($$^{6}$$Li) was performed for both 95 % and 40 % $$^{6}$$Li-enriched lithium titanate (Li$$_{2}$$TiO$$_{3}$$) samples as the candidate tritium breeding material using 2.6 MeV-proton Rutherford Backscattering Spectroscopy (RBS). The depth resolution of this method was enough to measure the $$^{6}$$Li depth profile in terms of the evaluation of thermal neutron transportation. The atomic density of $$^{6}$$Li could be measured within the error of 27 % for both samples although the accuracy of 5 % was not achieved for the evaluation of tritium breeding. It was confirmed that the sample atomic compositions agreed with Li$$_{2}$$TiO$$_{3}$$ within errors of 30 % for Li/Ti and 15 % for O/Ti. The whole errors were caused by the data accuracy of scattering cross sections predominately. Providing more accurate cross section data can lead to the utilization of the RBS method for tritium breeding materials analysis.

JAEA Reports

Data acquisition of groundwater flow and mass transport properties in sedimentary rocks, 2; Considerations of mass transport properties based on pore size distribution, chemical composition and air permeability of the Horonobe specimen

Hara, Akio

JAEA-Research 2006-020, 137 Pages, 2006/03

JAEA-Research-2006-020.pdf:26.13MB

Pore size distribution, chemical composition and air permeability are analyzed on diatomaceous mudstone specimens of Koetoi and Wakkanai formation sampled from boreholes HDB-1, HDB-2 and HDB-5 which drilled by Japan Nuclear Cycle Development Institute in Horonobe area Hokkaido Japan. Permeability of Koetoi and Wakkanai formation is related to diagenetic facis changes of diatom fossils. Diatom fossils are belonging to Opal-A facies in the specimens of Koetoi formation. Pore size distribution of Koetoi formation concentrated around 1000 angstrom and permeability is higher than Wakkanai formation. In Wakkanai formation, diagenetic facies of diatomaceous fossil is changes into Opal-CT facies and pore size distribution concentrates between 20 to 40 angstrom and permeability declines abruptly. The specimen which includes high SiO$$_{2}$$ content tend to have low permeability and high concentration of pore radius distribution between 20 to 40 angstrom. The degree of concentration should be caused to reducing permeability. A specimen sampled at 600m depth the formation boundary has high porosity and permeability than the specimen sampled around 200 to 300m depth below the formation boundary. The lowest permeability within the Koetoi and Wakkanai formation is observed in the Wakkanai formation at 200 to 300m depth of the formation boundary between Koetoi and Wakkanai formation. These strata have possibility of functioning as a seal rock, which can keep the high pressure and disturb ground water flow.

JAEA Reports

The Synthesis of B-C-N hybrid by ion beam deposition with aromatic molecules

Shimoyama, Iwao; Shigezumi, Kazuya*; Baba, Yuji; Sekiguchi, Tetsuhiro; Hirao, Norie*; Nagano, Masamitsu*

JAEA-Research 2006-021, 34 Pages, 2006/06

JAEA-Research-2006-021.pdf:1.47MB

Extreme low-energy ion beam deposition (IBD) method is devoted to synthesize boron carbon nitride (B-C-N) hybrid thin film. Several combinations of source gases, i.e., borazine (B$$_{3}$$N$$_{3}$$H$$_{6}$$), triazine (C$$_{3}$$N$$_{3}$$H$$_{3}$$), and benzene (C$$_{6}$$H$$_{6}$$), are used as precursor for the method in order to study the precursor effect. The characterization of the thin films is done by X-ray photoelectron spectroscopy (XPS). The precursor effect is observed at three points, (1) composition ratio (2) bond formation and (3) layered structure. XPS spectra show the composition ratio basically reflects the element ratio of each precursor. The existence of various kinds of bonds is observed in the B, C, and N 1s photoelectron peaks of the films and the formation of the bonds depends on the combination of source gas. The mixture of borazine and benzene precursor derives C-C, C-N, B-C, and B-N bonds formation. On the other hand, other gases scarcely derive B-C bond formation. We propose that this is caused by a layered structure caused by the viscosity difference of the precursor molecules. Finally, we conclude that the mixture of borazine and benzene is the most preferable for B-C-N synthesis by the IBD method among the precursor gases.

JAEA Reports

$$alpha$$-mode containment failure probability assessment for BWR by ROAAM application

Mayumi, Masami; Moriyama, Kiyofumi; Muramatsu, Ken

JAEA-Research 2006-022, 94 Pages, 2006/03

JAEA-Research-2006-022.pdf:5.35MB

In-vessel steam explosion-induced containment failure (alpha-mode containment failure) following core melt in nuclear power plants has a potential of large early release of radioactive materials. Therefore, it is an important issue to estimate the outcome frequency with the involved uncertainty in phenomena in PSA. There has been a methodology, called as ROAAM, proposed for resolving this type of issue. In this paper, application method based on ROAAM is studied and the estimation is carried out by this method for alpha-mode containment failure in BWR, which has less studied until now. This analysis verifies the practicability and capability of supplying the process parameter distributions. Analysis results show 95, 97.5 percentile, and expected (average) values to be 3.2$$times$$10$$^{-4}$$, 0.03, and 0.012 respectively for containment failure probability (conditional on explosion triggering). In addition, CCDF curves of various process parameters give a good representation for a grasp of whole event.

JAEA Reports

Study on high-performance fuel cladding materials; Joint research report in FY 2001-2005 (Phase 2) (Joint research)

Kiuchi, Kiyoshi; Ioka, Ikuo; Tanabe, Makoto*; Nanjo, Yoshiyasu*; Ogawa, Hiroaki; Ishijima, Yasuhiro; Tsukatani, Ichiro; Ochiai, Takamasa; Kizaki, Minoru; Kato, Yoshiaki; et al.

JAEA-Research 2006-023, 173 Pages, 2006/03

JAEA-Research-2006-023.pdf:20.51MB

The research concerning new cladding materials for ultra-high burnup of fuel elements with MOX fuels aiming at 100 GWd/t of BWR was pursued for 5 years from 2001 to 2005. On the Phase 1, the modified stainless steel of Fe-25Cr-35Ni-0.2Ti as fuel claddings and Nb-Mo alloy as a liner for inhibiting the pellet- clad interaction were selected as candidate materials, by evaluating fundamental properties required to BWR cladding materials, that are the nuclear economy, radioactivity, mass-transfer, irradiation properties, mechanical properties so on. On the present study, the making process of cladding tubes, lining by diffusion bonding, end plug by laser welding were developed and optimized, by considering the practical use of fuel elements consists of these candidates. The practical applicability was basically examined by irradiation tests using the accelerator of TIARA and the research reactor of JRR-3, for mainly confirming the resistance to IGSCC as one of the current important issues of BWR core materials of low carbon grade stainless steels. Creep and fatigue testing data were also obtained for evaluating the long performance of candidate materials. The behavior as fuel elements was analyzed with the safety calculation code for BWRs. The obtained results were established as a data base system, by considering the applicability to the fuel design and in-pile loop tests.

JAEA Reports

Corrosion behavior of steels in liquid lead-bismuth with low oxygen concentrations

Kurata, Yuji; Futakawa, Masatoshi; Saito, Shigeru

JAEA-Research 2006-024, 47 Pages, 2006/03

JAEA-Research-2006-024.pdf:15.54MB

no abstracts in English

JAEA Reports

Investigation of thermal hydraulic mixing mechanism in T-junction pipe with a 90-degree bend in upstream side for mitigation and controlling of thermal-striping phenomena (Joint research)

Yuki, Kazuhisa*; Hashizume, Hidetoshi*; Tanaka, Masaaki; Muramatsu, Toshiharu

JAEA-Research 2006-025, 47 Pages, 2006/03

JAEA-Research-2006-025.pdf:4.26MB

In T-junction pipe, temperature fluctuation is induced due to unstable fluid mixing. Development of the relaxation and control techniques for the thermal fatigue is one of the most important issues in the future plant design. If a 90-degree bend exists in the upstream of the mixing tee in the piping system of power plants, a secondary flow formed in the bend makes the fluid mixing phenomena even more complex. This study aims at clarifying effects of curvature ratio of the bend on the non-isothermal fluid mixing in the Tee-junction area and the temperature fluctuation induced by the unstable mixing, by visualizing the flow fields with PIV and measuring fluid-temperature fluctuation in the vicinity of wall. From the visualization, it is clarified that a high-temperature jet flowing out from a branch pipe swings and sways near the wall, which leads to higher temperature fluctuation than that in a case without the 90-degree bend. In addition, in the case that there exists a separation in the bend, the fluid mixing and the temperature fluctuation characteristics including its damping in the downstream direction are completely different from those without the separation. Furthermore, in the case using the bend that doesn't produce the separation, there are cautionary conditions in which the temperature fluctuation is maximized in a transition regime of a stratified flow and a turn-jet flow. It seems that the principal cause for this is repetition generation and disappearance of a circulating flow formed behind the jet due to an interaction between unsteady behavior of a secondary flow in a decay process after the bend and the wakes formed behind the jet, which leads to the vigorous oscillation of jet near the wall. For the prediction of the temperature fluctuation, it's possible to predict it in each fluid mixing pattern.

JAEA Reports

Development of thermal transient stress charts for screening evaluation of thermal loads

Furuhashi, Ichiro*; Kasahara, Naoto; Shibamoto, Hiroshi

JAEA-Research 2006-026, 178 Pages, 2006/03

JAEA-Research-2006-026.pdf:12.78MB

Thermal transient stress charts were developed for screening evaluation of thermal loads. Summay of obtaned results are as follows. (1) Thermal stress was theoretically analyzed on the plate subjected to thermal transient on both surfaces, and the design charts were proposed for evaluation of thermal transient stress. Compared with conventional design charts for the plate under single surface heat transfer, their applicable area is further extended. (2) Developed design charts can predict temperature and stresses responses to step or ramp change of fluid temperature. Utilizing these charts, surface temperature, average temperature in thickness, surface stress, bending stress and peak stress at arbitrary time can be obtained. (3) Non-dimensional temperature $$phi$$ and stress $$beta$$ were introduced, and reading errors can be reduced compared with the conventional ones. (4) Design charts were also proposed on the maximum thermal stresses and their arising times. It was revealed that the maximum thermal stresses never exceed 2 times of steady-state stress under the fixed back surface temperature. (5)Green functions of transient temperature and thermal stresses were developed. Temperature and thermal stresses can be predicted within 1.4% error. These charts will contribute to the screening evaluation of thermal loads with their locations, and will be employed for sensitive analyses for design and understanding of thermal stress mechanisms.

JAEA Reports

Study on magnetic property of surface hardening zone of stainless steel

Takaya, Shigeru; Nagae, Yuji; Aoto, Kazumi

JAEA-Research 2006-027, 17 Pages, 2006/03

JAEA-Research-2006-027.pdf:4.45MB

no abstracts in English

JAEA Reports

Report of the collaboration project for research and development of sphere-pac fuel among JNC-PSI-NRG (II); Irradiation tests and post irradiation examinations (Joint research)

Nakamura, Masahiro; Ozawa, Takayuki; Morihira, Masayuki; Kihara, Yoshiyuki

JAEA-Research 2006-028, 146 Pages, 2006/03

JAEA-Research-2006-028.pdf:37.95MB

The collaboration project concerning sphere-pac fuel among JNC (Japan Nuclear Cycle Development Institute, now Japan Atomic Energy Agency), Swiss PSI (Paul Scherrer Institut) and Dutch NRG (Nuclear Research and Consultancy Group) was performed from 1996 till 2005. The target of this project is comparative irradiation tests of sphere-pac fuel in the HFR (High Flux Reactor) in Petten in the Netherlands with pellet fuel and vipac fuel. Total 16 fuel segments (8 pins) containing 5%Np-MOX sphere-pac segments were irradiated. No fuel failure was occurred. Restructuring of sphere-pac fuel was quickly progressed in early stage of irradiation, and formation of the central hole was almost completed within 48 hours steady state irradiation. According to the results of the power-to-melt test, the power to melt linear heat rates were estimated as 60kW/m for the sphere-pac fuel and as 73kW/m for the pellet fuel under HFR irradiation conditions. Irradiation behaviors of the vipac fuel and Np-MOX sphere-pac fuel were basically similar to that of the MOX sphere-pac fuel. However, the central hole of the Np-MOX sphere-pac fuel was larger than that of MOX sphere-pac fuel in the restructuring test. It suggests that the thermal conductivity of Np-MOX fuel is smaller than that of MOX fuel.

JAEA Reports

Study of sub-surface disposal concepts for uranium waste, 2

Tsujimura, Seiichi; Funabashi, Hideyuki; Ishibashi, Makoto*; Takase, Toshio*; Kurosawa, Mitsuru*

JAEA-Research 2006-029, 96 Pages, 2006/07

JAEA-Research-2006-029.pdf:3.97MB

Uranium waste has characteristics that it is rarely expected to decay its radioactivities and it is not almost necessary to consider external exposure to radiation from waste package. We studied reasonable sub-surface disposal concepts for uranium waste in 2004 and 2005 considering the characteristics. In 2005, we studied necessity of engineered barrier for the disposal of uranium waste, considering change of chemical condition around disposal facilities over long periods of time. Safety assessment was made to analyze effect of difference in sorption parameters at reduction and oxidation conditions. The assessment showed that change from reduction to oxidation around disposal facilities did not lead to increase dose to the public. The assessment with realistic sorption parameters showed that dose to the public was not more than 10 $$mu$$Sv/y. The results proved that it was not necessary to keep reduction conditions around disposal facilities. This two-year- study showed that there was possibility of sub-surface disposal system without engineered barrier for uranium waste.

JAEA Reports

U, Pu and Np co-recovery in the simplified solvent extraction process; The Extraction behavior of Np at the condition of high HNO$$_{3}$$ concentration feed solution and scrubbing solution

Nakahara, Masaumi; Sano, Yuichi; Miyachi, Shigehiko; Koizumi, Tsutomu; Koyama, Tomozo; Aose, Shinichi

JAEA-Research 2006-030, 43 Pages, 2006/06

JAEA-Research-2006-030.pdf:1.93MB

Concerning the advanced aqueous reprocessing system, the simplified solvent extraction process for U, Pu and Np co-recovery has been investigated. We carried out the counter-current experiment, which aimed for Np oxidation and extraction by high [HNO$$_{3}$$] condition. For preventing Np leakage to the raffinate, feed solution and scrubbing solution with high [HNO$$_{3}$$] were used, which would bring high [HNO$$_{3}$$] in the extraction section and efficient Np oxidation and extraction in this section. In addition, high [HNO$$_{3}$$] in the feed solution could help the pre-oxidation of Np to extractable Np(VI). In the steady state, the Np leakage to the raffinate could be kept under about 1%. The stage efficiencies for these elements were estimated by fitting the concentration profiles calculated by MIXSET-X into the experimental results. The stage efficiency of U, Pu and Np were evaluated 100%, 100% and 98.5% in the extraction section and 95%, 90% and 89% in the stripping section respectively.

JAEA Reports

Study on the corrosion assessment of overpack welds, 2 (Joint research)

Mitsui, Hiroyuki*; Taniguchi, Naoki; Otsuki, Akiyoshi*; Kawakami, Susumu; Asano, Hidekazu*; Yui, Mikazu

JAEA-Research 2006-031, 88 Pages, 2006/06

JAEA-Research-2006-031.pdf:4.72MB

The corrosion experiments for welded carbon steel were planed to contribute to an assessment of long-term integrity of carbon steel overpack welds considering corrosion damage specific to overpack welds. Based on this plan, electrochemical tests for welded carbon steel using the samples welded by EBW and TIG were carried out, and the corrosion behavior of welded zone was compared with that of base metal. The results of anodic polarization tests in 0.01M and 0.1M carbonate aqueous solutions for base metal, heat affected zone and welded metal indicated that; -As for EBW, the anodic polarization curves were not affected by welding although the metallurgical structures vary with base metal, heat affected zone and welded metal. -As for TIG, the current density of welded metal was larger than that of base metal and of heat affected zone, and local dissolution with immediate increase in current density was observed in 0.01M-pH10 carbonate aqueous solution.

JAEA Reports

Design study on a fuel handling system in a sodium cooled reactor; Study in FY2004 (Joint research)

Chikazawa, Yoshitaka; Usui, Shinichi; Konomura, Mamoru; Ikeda, Hirotsugu

JAEA-Research 2006-032, 202 Pages, 2006/04

JAEA-Research-2006-032.pdf:38.12MB

In the feasibility study on commercialized fast breeder cycle system, fuel handling systems for sodium cooled reactors has been studied. In FY 2004 study, a fuel handling system with an EVST for a twin large scale reactor power plant is designed and key issues about the system are identified. A manipulator type fuel handling machine suitable for the upper internal structure with a slit designed and seismic analyses show that it can treat spent fuels without interaction with upper internal structure in earthquakes. Fuel handling time is reduced adopting a sodium pot which can carry 2 subassemblies in onetime. Spent fuels are stored at an EVST while their decay heat are reduced to be 5kW/subassembly. A new fuel handling system for fuels with minor actinide is designed considering 1kW/subassembly heat and shielding. A innovative concept without an EVST is also studied. A fuel handling system adopting fuel transfer without a sodium pot is constructed to reduce material mass. A fuel handling system for a metal fuel reactor plant has been design. From the result of a survey on a gas storage, a water pool storage with helium cans and EVST, a system with EVST is selected because of its economical and safety advantage. Fuel handling condition is briefly reviewed considering commercialized reactor fuel specifications such as minor actinide content and ODS cladding.

JAEA Reports

Investigation of transmutation target for long-lived fission products; Basic examinations of iodide candidates, 2

Donomae, Takako; Tachi, Yoshiaki; Matsumoto, Shinichiro

JAEA-Research 2006-033, 35 Pages, 2006/07

JAEA-Research-2006-033.pdf:8.59MB

no abstracts in English

JAEA Reports

Preparation and characterization of B-C-N hybrid thin films

Uddin, M. N.*; Shimoyama, Iwao; Sekiguchi, Tetsuhiro; Nath, K. G.*; Baba, Yuji; Nagano, Masamitsu*

JAEA-Research 2006-034, 72 Pages, 2006/06

JAEA-Research-2006-034.pdf:4.05MB

no abstracts in English

JAEA Reports

Status of assessment tools on the performance guarantee contents of buffer material

Tanai, Kenji; Jintoku, Takashi*; Kikuchi, Hirohito*; Nishimura, Mayuka; Matsumoto, Kazuhiro*; Aoyanagi, Shigeo; Yui, Mikazu

JAEA-Research 2006-035, 32 Pages, 2006/06

JAEA-Research-2006-035.pdf:3.46MB

In order to contribute to the safety standards and guidelines which a regulator decides, state-of the art assessment method is investigated and summarized in the table about performance guarantee contents of buffer material related to the mechanical support and protection of the overpack and rock matrix, and the retardation of radionuclide. In addition, examples of the assessment tool are described. In this report, summary of (1) basic properties of bentonite, including swelling properties, mechanical properties and hydraulic properties, (2) long-term behavior of bentonite, including creep deformation, penetration into host rock, erosion and alteration, (3) gas permeability, (4) colloid filtration and (5) mechanical stability of the near-field is described. Check points, assessment methods (based on the data obtained from the experimental results, the estimation value obtained from empirical equations and database, and the modeling calculations) and latest results of these R&D programs were also summarized.

JAEA Reports

A Study on finding suitable parameters for constitutive models of evaluating long term mechanical behavior of buffer material

Nishimura, Mayuka; Tanai, Kenji; Takaji, Kazuhiko*; Hirai, Takashi*; Shiratake, Toshikazu*

JAEA-Research 2006-036, 82 Pages, 2006/06

JAEA-Research-2006-036.pdf:4.09MB

In order to evaluate the long term mechanical behavior of the Enginnered Barrier System (EBS), it is essential to set suitable parameters of constitutive models of the buffer material. In this report, the parameters of the Sekiguchi-Ohta model and the Adachi-Oka model are examined. The results are as follows; it is impossible to set the sole parameter which can evaluate both the deformation and the stress of the buffer material. But when selecting a suitable parameter for the object of evaluation, the mechanical behavior of buffer material can be approximately described. On the other hand, the range of the secondary consolidation coefficient of the bentonite ore is estimated to corroborate to the viscous parameter of the buffer material, using a natural analogue study. Finally, the mechanical behavior of the buffer material is analyzed using these parameters, and the maximum amount of settling of the overpack is estimated.

JAEA Reports

Research work for utilizing technology of the lead-bismuth eutectic, 2; Research on corrosion resistance of ODS-Al steels in high temperature lead-bismuth eutectic under oxygen concentration control (Joint research)

Furukawa, Tomohiro; Nishi, Yoshihisa*; Aoto, Kazumi; Kinoshita, Izumi*

JAEA-Research 2006-037, 36 Pages, 2006/06

JAEA-Research-2006-037.pdf:11.89MB

In 2002, the Japan Atomic Energy Agency (past organization name: Japan Nuclear Cycle Development Institute) was made a contract with the Central Research Institute of Electric Power Industry on the research work for utilizing technology of the lead bismuth eutectic. In the contract, research on corrosion of FBR materials in high temperature lead bismuth eutectic was performed. This work was composed of two stages. In the first stage, corrosion test of high chromium martensitic steel, which was one candidate material for structures of advanced fast reactor, was performed in oxygen controlled lead bismuth eutectic at 923K. Effect of chromium on corrosion in the lead bismuth eutectic was estimated. In this second research, corrosion test of oxide dispersion strengthened ferritic steels whose chemical compositions of chromium and aluminum were differed has been performed in the lead bismuth eutectic for up to 4,000 hours. As the results, although chromium effect on corrosion has not been observed, good corrosion resistance by aluminum oxide formation on the surface has been obtained.

JAEA Reports

Comparison of the thermodynamic databases for radioactive elements in application to the calculation of the solubilities in the porewater

Doi, Reisuke; Shibata, Masahiro

JAEA-Research 2006-038, 30 Pages, 2006/07

JAEA-Research-2006-038.pdf:3.22MB

To calculate the solubility of radioactive elements which is the important parameter for assessing performances in geological environments, the thermodynamic database must be reliable and based on the latest information. In this research, we compared the solubilities of the representative radioactive elements in the porewater compositions of the compacted bentonite which were set up in the second progress report (H12). The solubility was calculated with the thermodynamic databases of JAEA, OECD/NEA, Nagra/PSI. And we investigated the causes of the differences among three thermodynamic databases.

JAEA Reports

Oxidation resistance of silicon coated by boron nitride ultra thin film

Shimoyama, Iwao; Miyauchi, Hideo*; Baba, Yuji; Sekiguchi, Tetsuhiro; Hirao, Norie*; Okuno, Kenji*

JAEA-Research 2006-039, 33 Pages, 2006/06

JAEA-Research-2006-039.pdf:2.98MB

Boron nitride (BN) ultrathin film attracts much attention as a coating material for Si cathode due to the chemical stability, heat resistance, and negative electron affinity. In order to study the oxidation resistance of BN ultrathin film coating, thermal dry oxidation is applied to BN coated Si and non-coated Si at various temperatures. X-ray photoelectron spectroscopy is devoted to clarify the modification of chemical state of the samples. The XPS spectra change by the thermal oxidation for the non-coated Si. On the other hand, it scarcely change by the thermal oxidation for the BN coated Si. The oxidation resistance in ambient air is also investigated for BN coated Si. After the several days exposure of air, the O 1s photoelectron peak is drastically enhanced. These results mean that BN ultrathin film works as protective coating for dry thermal oxidation, however it does not work in ambient air.

JAEA Reports

Study on applicability of low alkaline shotcrete in Horonobe URL project

Konishi, Kazuhiro; Nakayama, Masashi; Mihara, Morihiro; Yoshida, Yasushi*; Iriya, Keishiro*; Akiyoshi, Kenji*; Noda, Masaru*

JAEA-Research 2006-040, 53 Pages, 2006/06

JAEA-Research-2006-040.pdf:12.6MB

no abstracts in English

JAEA Reports

Environmental research for Horonobe Underground Research Project (Contract research)

Ichiyasu, Kenji; Ueda, Noritaka*; Ito, Takahisa*; Nakadate, Fumiyuki*; Yamakoshi, Chizuru*

JAEA-Research 2006-041, 71 Pages, 2006/06

JAEA-Research-2006-041.pdf:10.97MB

Since a development scale of Horonobe Underground Research Project is small, this research would permit under the Environmental Impact Assessment Low and the Basic Environment Ordinance of Hokkaido Prefecture. However, because of consideration of recent general social trends, environmental monitoring research was carried out to know this influence. Noise and vibration measurement was conducted at four points around the research area. Water quality measurement was conducted at two points. Fisheries research was conducted throughout the river. Plant community research according to the Blaun-Blanquet technique (1964) was conducted at two points in the area where rectangular divisions (quadrate) were set. And also, growing condition survey on Hai-dojoh-tsunagi was conducted. This Hai-dojoh-tsunagi is regarded as endangered species in Hokkaido, had been transplanted as an environment protection measure in 2003. In addition, the research places and the methods were made to be the same as that of the past (2004 and 2005) researches. From these results, the influence on the environment after construction and the progress after conduction of measure for protecting environment were checked. Consequently, the influence of construction was not found at present time. Therefore the measure was judged enough to protect environment. After this construction, the environmental monitoring research should be carrying out to check the influence and it would be necessary to conduct the quick and suitable measure to avoid or mitigate the influence.

JAEA Reports

Feasibility study on commercialized fast reactor cycle systems technical study report of phase II, 1; Fast reactor plant systems

FBR System Engineering Unit; FBR Systems Reliability Research Unit; FBR Safety & Innovative Technology Unit; FBR Cycle Synthesis Unit; Innovative Water Reactor Design Group; Nuclear Science and Engineering Directorate

JAEA-Research 2006-042, 36 Pages, 2006/06

JAEA-Research-2006-042.pdf:2.34MB
JAEA-Research-2006-042-Incl(CD).pdf:84.0MB

no abstracts in English

JAEA Reports

Feasibility study on commercialized fast reactor cycle systems technical study report of phase II, 2; Nuclear fuel cycle systems

FBR Fuel Cycle Unit; FBR Cycle Synthesis Unit

JAEA-Research 2006-043, 26 Pages, 2006/06

JAEA-Research-2006-043.pdf:2.12MB
JAEA-Research-2006-043-Incl(CD).pdf:43.84MB

A joint project team of Japan Atomic Energy Agency and the Japan Atomic Power Company (as the representative of the electric utilities) started the feasibility study on commercialized fast reactor cycle systems (F/S) in July 1999 in cooperation with Central Research Institute of Electric Power Industry and vendors. On the major premise of safety assurance, F/S aims to present an appropriate picture of commercialization of fast reactor (FR) cycle system which has economic competitiveness with light water reactor cycle systems and other electricity baseload systems, and to establish FR cycle technologies for the future major energy supply. In the phase-I of F/S from Japanese fiscal year (JFY) 1999 to 2000, representative FR cycle concepts were screened out. With regard to fuel cycle systems, fuel reprocessing methods such as advanced aqueous, oxide electrowinning and metal electrorefining, and fuel fabrication methods such as simplified pelletizing, sphere-packing, vibro-packing, metal casting and coated particle were selected. In the phase-II (JFY 2001-2005), the design study of several fuel cycle systems combined reprocessing methods with fuel fabrication ones, and the development of significant technologies necessary for the feasibility evaluation have been performed to clarify the promising candidate concepts suited for maximizing the FR ability. Further, key technical issues for the commercialization of fuel cycle systems are clarified and their R&D plans until around 2015 are made.

JAEA Reports

Effective application of partitioning and transmutation technologies to geologic disposal

Ikegami, Tetsuo; Ahn, J.*

JAEA-Research 2006-045, 17 Pages, 2006/07

JAEA-Research-2006-045.pdf:1.55MB

Environmental Impact, which is a newly developed measure in stead of the conventional radio-toxicity, has been evaluated for both the PWR cycle and the FBR cycle in order to clarify what kind of radio-nuclides and how much level of partitioning and transmutation are desirable. Bounding analysis for uncertainty of parameters relevant to radionuclide transport in a repository has also been performed. The evaluated results imply that the targets of partitioning and transmutation can be set; (1)In the case of PWR cycle, the release rate of $$^{237}$$Np and $$^{243}$$Am should be controlled under 1 %, in addition to the conventionally assumed release rate of 0.604 % for U and 0.297 % for Pu. (2)In the case of FBR cycle, recovery rate of 99.9 % for all actinide nuclides is appropriate.

JAEA Reports

Numerical simulation of turbulence mixing at T-junction piping system with elbow pipe in upstream side; Effect of secondary flow on temperature fluctuation

Tanaka, Masaaki

JAEA-Research 2006-046, 36 Pages, 2006/07

JAEA-Research-2006-046.pdf:5.82MB

In this report, numerical simulation using quasi-direct simulation code (DINUS-3) was carried out to investigate the effect of secondary flow upon the temperature fluctuation characteristics in the mixing region. The parameter was the angle of the elbow pipe flow direction to the branch pipe flow direction. By the numerical simulation, it found that the secondary flow had strong influence to the fluid mixing. Secondary flow changed the flow pattern of branch pipe jet and changed thoroughly the temperature fluctuation characteristics on the pipe surface, according to the angle. Attention should be paid to the arch-shaped large-scale eddies formed around the branch pipe jet in the thermal fatigue evaluation, because they played an important role for the fluid mixing. Furthermore, the range of about 2.0Dm from mixing point to the downstream should be focused for the evaluation of the thermal fatigue.

JAEA Reports

High temperature oxidation test of oxide dispersion strengthened (ODS) steel claddings

Narita, Takeshi; Ukai, Shigeharu; Kaito, Takeji; Otsuka, Satoshi; Matsuda, Yasushi*

JAEA-Research 2006-047, 100 Pages, 2006/07

JAEA-Research-2006-047.pdf:53.38MB

In a feasibility study of ODS steel cladding, its high temperature oxidation resistance was evaluated. Although addition of Cr is effective for preventing high temperature oxidation, excessively higher amount of Cr leads to embrittlement due to the Cr-rich $$alpha$$' precipitate formation. In the ODS steel developed by the Japan Atomic Energy Agency (JAEA), the Cr content is controlled in 9Cr-ODS martensite and 12Cr-ODS ferrite. In this study, high temperature oxidation test was conducted for ODS steels, and their results were compared with that of conventional austenitic stainless steel and ferritic-martensitic stainless steel. Following results were obtained in this study. (1)9Cr-ODS martensitic and 12Cr-ODS ferritic steel have superior high temperature oxidation resistance compared to 11mass%Cr PNC-FMS and even 17mass% SUS430 and equivalent to austenitic PNC316. (2)The superior oxidation resistance of ODS steel was attributed to earlier formation of the protective alpha-Cr$$_{2}$$O$$_{3}$$ layer at the matrix and inner oxide scale interface. The grain size of ODS steel is finer than that of PNC-FMS, so the superior oxidation resistance of ODS steel can be attributed to the enhanced Cr-supplying rate throughout the accelerated grain boundary diffusion. Finely dispersed Y2O3 oxide particles in the ODS steel matrix may also stabilized the adherence between the protective alpha-Cr$$_{2}$$O$$_{3}$$ layer and the matrix.

JAEA Reports

Water corrosion test of oxide dispersion strengthened (ODS) steel claddings

Narita, Takeshi; Ukai, Shigeharu; Kaito, Takeji; Otsuka, Satoshi; Matsuda, Yasushi*

JAEA-Research 2006-048, 52 Pages, 2006/07

JAEA-Research-2006-048.pdf:29.48MB

As a part of feasibility study of ODS steel cladding, its water corrosion resistance was examined under water pool condition. Although addition of Cr is effective for preventing water corrosion, excessive Cr addition leads to embrittlement due to the Cr-rich $$alpha$$' precipitate formation. In the ODS steel developed by the Japan Atomic Energy Agency (JAEA), the Cr content is controlled in 9Cr-ODS martensite and 12Cr-ODS ferrite. In this study, water corrosion test was conducted for these ODS steels, and their results were compared with that of conventional austenitic stainless steel and ferritic-martensitic stainless steel. Following results were obtained in this study. (1) Corrosion rate of 9Cr-ODS martensitic and 12Cr-ODS ferritic steel are significantly small and no pitting was observed. Thus, these ODS steels have superior resistance for water corrosion under the condition of 60$$^{circ}$$C and pH8$$sim$$12. (2) It was showed that 9CR-ODS martensitic and 12Cr-ODS ferritic steel have comparable water corrosion resistance to that of PNC316 and PNC-FMS at 60$$^{circ}$$C for 1000h under varying pH of 8, 10. Water corrosion resistance of these alloys is slightly larger than that of PNC316 and PNC-FMS at pH12 without significant differenceof appearance and uneven condition.

JAEA Reports

Study of an electromagnetic pump in a sodium cooled reactor; Design study of secondary sodium main pumps (Joint research)

Chikazawa, Yoshitaka; Kisohara, Naoyuki; Hishida, Masahiko; Fujii, Tadashi; Konomura, Mamoru; Ara, Kuniaki; Hori, Toru*; Uchida, Akihito*; Nishiguchi, Yohei*; Nibe, Nobuaki*

JAEA-Research 2006-049, 75 Pages, 2006/07

JAEA-Research-2006-049.pdf:4.55MB

In the feasibility study on commercialized fast breeder cycle system, a medium scale sodium cooled reactor with 750MW electricity has been designed. In this study, EMPs are applied to the secondary sodium main pump. The EMPs type is selected to be an annular linear induction pump (ALIP) type with double stators which is used in the 160m$$^3$$/min EMP demonstration test. The inner structure and electromagnetic features are decided reviewing the 160m$$^3$$/min EMP. Two dimensional electromagnetic fluid analyses by EAGLE code show that Rms (magnetic Reynolds number times slip) is evaluated to be 1.08 which is less than the stability limit 1.4 confirmed by the 160m$$^3$$/min EMP test, and the instability of the pump head is evaluated to be 3% of the normal operating pump head. Since the EMP stators are cooled by contacting coolant sodium duct, reliability of the inner structures are confirmed by temperature distribution and stator-duct contact pressure analyses. Besides, a power supply system, maintenance and repair feature and R&D plan of EMP are reported.

JAEA Reports

Effects of tungsten on microstructure and high-temperature strength of oxide dispersion strengthened (ODS) martensitic steel

Narita, Takeshi*; Ukai, Shigeharu; Kaito, Takeji; Otsuka, Satoshi; Fujiwara, Masayuki

JAEA-Research 2006-050, 85 Pages, 2006/10

JAEA-Research-2006-050.pdf:133.32MB

In 9Cr ODS martensitic steel, tungsten(W) is a solid solution strengthening element, whose addition increases high-temperature strength by the combined effect with oxide dispersion strengthening. However, its excessive addition results in the increase of ferrite phase causing precipitation of intermetallic compound (Laves phase) under high temperature irradiation condition and thus ductility degradation. The amount of W addition therefore should be as low as possible. In this report, the effects of W on microstructure and high temperature mechanical properties of 9Cr ODS martensitic steels were examined for obtaining insights into optimum W concentration in terms of high-temperature strength and ductility. The results obtained are as follows: (1)In the 9CrODS martensitic steel, addition of W exceeding 2mass% is shown to cause precipitation of Laves phase which degrades the ductility and fracture toughness. It can be said that the current specification of W concentration, i.e. 2mass%W, is appropriate. (2)Hardness and tensile strength is shown to increase with W concentration. This increase is caused by the increase of solid solution strengthening and residual-alpha ferrite. The retainment of residual-alpha ferrite is enhanced by the addition of W (ferrite former element). The improvement of tensile strength at 973K provided by the solid solution strengthening is shown to be equivalent to that provided by the retainment of residual-alpha ferrite. (3)It would be open task to explorer an improved alloy design concept, i.e. decrease of W as low as possible and increase of residual-alpha ferrite. The degradation of high-temperature strength by decreasing W addition can be made up by the increasing fraction of residual-alpha phase that is provided by reduction of austenite former elements and increasing addition of ferrite former elements.

JAEA Reports

Preliminary study on the corrosion behavior of carbon steel in Horonobe groundwater environment

Taniguchi, Naoki; Kogawa, Noritaka*; Maeda, Kazuto*

JAEA-Research 2006-051, 60 Pages, 2006/08

JAEA-Research-2006-051.pdf:6.45MB

It is necessary to understand the corrosion behavior of candidate overpack materials to plan the in-situ engineered barrier test at underground laboratory constructing at Horonobe and to design the overpacks suitable to Horonobe environment. The preliminary corrosion tests of carbon steel which is a candidate material for overpacks were carried out using artificial groundwater and actual groundwater sampled at Horonobe. As the results of anodic polarization experiments, the anodic polarization curves of carbon steel in buffer material were active dissolution type, and the corrosion type of carbon steel in Horonobe groundwater environment was expected to be general corrosion. The results of immersion test under air equilibrium condition and anaerobic condition showed that the corrosion rate and the degree of corrosion localization were not exceeded the data obtained in previous studies. Based on the experimental results, it was confirmed that the corrosion assessment model and assumed corrosion rate in second progress report (H12 report) can be applied to the assessment for Horonobe groundwater condition.

JAEA Reports

Evaluation of the ($$alpha$$,xn) reaction data for JENDL/AN-2005

Murata, Toru*; Matsunobu, Hiroyuki*; Shibata, Keiichi

JAEA-Research 2006-052, 63 Pages, 2006/07

JAEA-Research-2006-052.pdf:5.11MB

Neutron emission data of the $$(alpha,xn)$$ reactions were evaluated in the incident $$alpha$$-particle energy region below 15 MeV for nuclides important mainly in nuclear fuel-cycle applications, namely, $$^{6,7}Li$$, $$^{9}Be$$, $$^{10,11}B$$, $$^{12,13}C$$, $$^{14,15}N$$, $$^{17,18}O$$, $$^{19}F$$, $$^{23}Na$$, $$^{27}Al$$ and $$^{28,29,30}Si$$. The evaluation was performed on the basis of available experimental data and nuclear model calculations. The evaluated nuclear data were compiled in the ENDF-6 format, and released in June 2005 as JENDL $$(alpha,n)$$ Reaction Data File 2005 (JENDL/AN-2005) which is one of JENDL special-purpose files. This report describes evaluation methods and the results on evalauted cross sections and angular- and energy-distributions of emitted neutrons.

JAEA Reports

Development of chemical equilibrium analysis code "CHEEQ"

Nagai, Shuichiro

JAEA-Research 2006-053, 184 Pages, 2006/08

JAEA-Research-2006-053.pdf:14.36MB

"CHEEQ" code which calculates the partial pressure and the mass of the system consisting of ideal gas and pure condensed phase compounds, was developed. "CHEEQ" was consisted of following 3 parts, (1) analysis code, zc132.f (2) thermodynamic data base, zmdb01 and (3) input data file, zindb. "CHEEQ" code can calculate the system which consisted of elements (max.20), condensed phase compounds (max.100) and gaseous compounds. (max.200) Thermodynamic data base, zmdb01 contains about 1000 elements and compounds, and 200 of them were Actinide elements and their compounds. This report describes the basic equations, the outline of the solution procedure and instruction to prepare the input data and to evaluate the calculation results.

JAEA Reports

Study on evaluation of containment capability of grove-box under fire accident

Abe, Hitoshi; Watanabe, Koji; Tashiro, Shinsuke; Uchiyama, Gunzo

JAEA-Research 2006-054, 39 Pages, 2006/09

JAEA-Research-2006-054.pdf:2.91MB

no abstracts in English

JAEA Reports

Development of next generation code system as an engineering modeling language, 5; Investigation on restructuring method of conventional code into two-layer system

Yokoyama, Kenji

JAEA-Research 2006-055, 60 Pages, 2006/10

JAEA-Research-2006-055.pdf:4.65MB

A proposed method for gradually restructuring to the two-level system of next generation analysis system by reusing the conventional analysis system, called "incremental method", was applied and evaluated. The following functions were selected for the evaluation of the restructuring: Neutron diffusion calculation for the three-dimensional XYZ system based on finite differential method, and input utilities of the cross-section data file. In order to evaluate the effect of the restructuring, "Module Coupling Index(MCI)" and "McCabe's Cyclomatic Complexity (MCC)" were used for quantifying the quality of the modular design and the complexity of the program sequence. The incremental method could reduce MCIs from 6$$sim$$7 degrees to under 4 degrees in most module. And, it is found that the modules under 4 degrees of MCI can be easily combined with different programming languages. In the meantime, MCCs in most module before restructuring wereover 20 and some were over 50. The incremental method could reduce them to under 10 in most module. It is correspondent to reduction of the error frequency from 20$$sim$$40% to 5$$sim$$10%. The total number of MCC was able to be reduced to 1/2. By using the restructured functions in the present study and some previously developed functions, a reactor analysis tool was systematized and applied to criticality analysis of the Experimental Fast Reactor "JOYO" MK-I. In addition, it is confirmed that additional functionality expansions were carried out satisfying the condition that one can extend it only with input data and functions fornormal users (the user extendibility) and one can extend it without any modifications of existing programs (the open-closed principle).

JAEA Reports

Long-term groundwater pressure monitoring in deep boreholes in the Horonobe Underground Research Laboratory Project

Yabuuchi, Satoshi; Kurikami, Hiroshi; Seno, Shoji*; Hara, Minoru; Kunimaru, Takanori; Takeuchi, Ryuji

JAEA-Research 2006-056, 32 Pages, 2006/09

JAEA-Research-2006-056.pdf:1.87MB

Long-term groundwater pressure monitoring has been performed in HDB-1,2,3,6,7 and HDB-8 boreholes in the Horonobe Underground Research Laboratory Project. Groundwater pressure in many levels in the boreholes shows an almost steady state at present, however it is still recovering since the beginning of the observation in some levels. Relatively high groundwater pressure is observed in HDB-2 borehole, about 7km away from the URL area. According to the groundwater pressure monitoring in deep boreholes so far, it is inferred that hydraulic head becomes higher with the increase of the depth and hydraulic head in the east is higher than in the west around the URL area. Through the groundwater monitoring, performance of the long-term groundwater monitoring systems could be examined and some problems of the parts could also be found.

JAEA Reports

Numerical simulation system for environmental studies: SPEEDI-MP

Nagai, Haruyasu; Chino, Masamichi; Terada, Hiroaki; Harayama, Takaya*; Kobayashi, Takuya; Tsuzuki, Katsunori; Kim, K.; Furuno, Akiko

JAEA-Research 2006-057, 67 Pages, 2006/09

JAEA-Research-2006-057.pdf:13.49MB

A numerical simulation system SPEEDI-MP has been developed to apply for various environmental studies. SPEEDI-MP consists of dynamical models and material transport models for the atmospheric, terrestrial, and oceanic environments, database for model inputs, and system utilities for file management, visualization, etc. As a numerical simulation tool, a model coupling program (model coupler) has been developed. A coupled model system for water circulation has been constructed with atmosphere, ocean, wave, hydrology, and land-surface models using the model coupler. System utility GUIs are based on the Web technology, allowing users to manipulate all the functions on the system using their own PCs via the internet. In this system, the source estimation function in the atmospheric transport model can be executed on the grid computer system. Performance tests of the coupled model system for water circulation were also carried out for the flood and the storm surge events.

JAEA Reports

Investigation of the long term corrosion resistance of the overpack (Contract research)

Tachikawa, Hirokazu*; Kawakubo, Fumie*; Shimizu, Akihiko*; Shibata, Toshio*; Sugimoto, Katsuhisa*; Seo, Masahiro*; Tsuru, Toru*; Fujimoto, Shinji*; Inoue, Hiroyuki*

JAEA-Research 2006-058, 80 Pages, 2006/10

JAEA-Research-2006-058.pdf:10.86MB

The Japan Nuclear Cycle Development Institute submitted "Second Progress Report on Research and Development for the Geological Disposal of HLW in Japan" to the Japanese government. This report contains investigations of the corrosion life time of the overpack on the basis of experimental data and past research, assuming the ranging geological environment of Japan. However some subjects, such as the behavior of the overpack under high pH conditions and the behavior of the engineering barrier with change of near-field environmental condition with time for promoting reliability have still been left. To take into account these conditions, expert committee composed of metal corrosion science experts were established in the Nuclear Safety Research Association and past research outcomes and the theory of safety assessment were investigated from the view points of long term stability and corrosion resistance of engineering barrier.

JAEA Reports

Analyses of SEFOR experiments; Isothermal temperature coefficient (Core II) and power coefficient

Hazama, Taira; Numata, Kazuyuki*

JAEA-Research 2006-059, 133 Pages, 2006/09

JAEA-Research-2006-059.pdf:3.43MB

The SEFOR (South-West Experimental Fast Oxide Reactor) Doppler reactivity experiments have been analyzed on the isothermal temperature coefficient measured in SEFOR Core II, and on the power coefficient measured in SEFOR Cores I and II. Nominal values and uncertainties of the experimental data were re-evaluated, starting from a review of raw data in the original experimental reports. In particular, the power coefficient data were thoroughly re-evaluated, including fuel temperature data and associated uncertainties. The latest data on fuel thermal conductivity correlation was employed in the evaluation. As a result, experimental uncertainty of the power coefficient was reduced from 11% of reported value to 8%. In the temperature coefficient analysis, the calculated value overestimates the experimental value by 9%. The discrepancy exceeds sum of the experimental (3%) and analytical (4%) uncertainties. In the power coefficient analysis, the calculated values agree with experimental values within an experimental uncertainty of 8%. Benchmark data for the Doppler constants has been prepared based on the Power Doppler relativities. The resulting Doppler constants are different form existing benchmark data by about 4%. The change is mainly attributed to the update of fuel thermal conductivity correlation. The new values are more reasonable than the existing values in that C/E values do not depend on the core types.

JAEA Reports

Design studies on small fast reactor cores, 5; Research results in JFY2005

Uto, Nariaki; Okano, Yasushi; Naganuma, Masayuki; Mizuno, Tomoyasu; Hayashi, Hideyuki

JAEA-Research 2006-060, 68 Pages, 2006/09

JAEA-Research-2006-060.pdf:3.98MB

A design study on "Long-life Type Concept" of a 50MWe sodium-cooled metal-fueled reactor core was performed with more emphasis on irradiation results regarding fuel smear density. The concept aims at no refueling in a core life time, and achieving higher core outlet temperature such as 550$$^{circ}$$C which is advantageous to hydrogen production. The restriction of upper fuel smear density limit to 75% along with adjustments of fuel specifications showed feasibility of attaining core life time of 30 years and core outlet temperature of 550$$^{circ}$$C. No indication of occurrence of absorber-cladding mechanical interaction (ACMI) was found in the evaluation of ACMI for a control rod element. A shielding with Zr-H was selected in view of enhancement of shielding performance, and the feasibility was shown to satisfy the target allowance level of the ratio of hydrogen to zirconium, more than 1.53, with PNC316 used as the cladding material.

JAEA Reports

Study on reactor core and fuel design of sodium cooled fast reactor, Mixed oxide fuel core; Results in JFY2005

Ogawa, Takashi; Sato, Isamu; Naganuma, Masayuki; Aida, Tatsuya*; Sugino, Kazuteru; Hayashi, Hideyuki

JAEA-Research 2006-061, 54 Pages, 2006/09

JAEA-Research-2006-061.pdf:3.86MB

Sodium cooled fast reactor with mixed oxide fueled core is one of the promising candidates in "Feasibility Study on Commercialized Fast Reactor Cycle System" in Japan. The results of the study on the reactor core and fuel design in the JFY2005 are reported. (1)Design studies of high internal conversion (HIC) type core: (i)Influence of TRU composition variation on the HIC type core and fuel designs was evaluated. (ii)In adopting PNC-FMS steel as alternative cladding material of ODS steel, influence to the reactor core and fuel design was evaluated for the large-scale HIC type core. (iii)Shielding property of the large-scale HIC type core was evaluated. (iv)Some measures to extend the lifetime of control rod were studied for the large-scale HIC type core. (2)Design study on high breeding performance: The core design corresponding to a requirement of high breeding performance was studied based on the large-scale compact type core designed in the JFY2004.

JAEA Reports

High-efficiency improvement for high energy resolution experimental mode of DIANA spectrometer at Materials and Life Science Facility (MLF) of J-PARC

Takahashi, Nobuaki; Shibata, Kaoru; Sato, Taku*; Arai, Masatoshi

JAEA-Research 2006-062, 22 Pages, 2006/09

JAEA-Research-2006-062.pdf:2.16MB

DIANA is an indirect-geometry time-of-flight spectrometer which is planed to install at MLF, J-PARC. It has three exchangeable analyzer crystals, such as PG(002), Ge(311) and Si(111) for different energy transfer, momentum transfer, energy resolution experiments. Normal experimental mode, either PG(002) or Ge(311) analyzer is used, shows moderate energy resolutions of 15 micro eV or 40 micro eV, respectively. We are especially aiming very high energy resolution of 2 micro eV by using Si(111) analyzer crystal together with high speed counter-rotating pulse-shaping choppers with each rotation frequency of 300 Hz as an optional setting for the spectrometer. Although such a high energy resolution is attained, it is considerably inefficient having a very narrow incident energy (Ei) band if the pulse shaping chopper has only one slit. Therefore, we have designed multi-slit chopper and have performed Monte-Carlo simulation to study Repetition Rate Multiplication (RRM) capability.

JAEA Reports

Study on transmutation technology of Long-Lived-Fission-Products (LLFP) using commercial fast reactors; Loading type of LLFP target assembly and transmutation performances of cores designed in FS phase-II

Naganuma, Masayuki; Aida, Tatsuya*; Hayashi, Hideyuki

JAEA-Research 2006-063, 97 Pages, 2006/09

JAEA-Research-2006-063.pdf:9.19MB

In the Feasibility Study in Japan (FS), transmutation technology of LLFP using commercial fast reactors has been studied to reduce the environmental burden. In this report, loading type of LLFP target assembly, transmutation performances of FS designed cores and capability of transmutation core with high SF (transmutation / production ratio) are studied. Design studies for two loading types cores (in-core and ex-core loading type) were conducted for comparison. The in-core loading type core was found to decrease LLFP inventories significantly, thus, that was selected as the reference of FS. For FS phase-II designed cores, LLFP transmutation performances were evaluated. Every core was confirmed to have the capability to attain SF $$>$$ 1.0. Then, we conducted sensitivity evaluations of design conditions to SF for transmutation core with high SF. Since sensitivities of every condition were found to be small, we concluded that large SF may be impossible for the commercial reactors.

JAEA Reports

Hydrothermal experiments using Fe-bentonite; Identification of change of Fe-bentonite under low oxygen and high temperature conditions

Suyama, Tadahiro*; Shibata, Masahiro; Sasamoto, Hiroshi

JAEA-Research 2006-064, 21 Pages, 2006/10

JAEA-Research-2006-064.pdf:4.13MB

In previous experiment, we studied about interaction bentonite and Fe by the laboratory experiment that bentonite mixed with iron powder was kept for 6 years at room temperature in the glove box (O$$_{2}$$ $$<$$ 1 ppm). It was shown that bentonite was not influenced about its property, but change to Fe-bentonite. In this experiment, it is aimed to confirm mineralogical change of start substance, which is Fe-bentonite changed from Na-bentonite beforehand, by accelerating reactions on high temperature. Fe-bentonite was prepared by mixing Na-bentonite (Kunipia F) and FeCl$$_{2}$$ solution in the glove box (Ar, O$$_{2}$$ $$<$$ 1 ppm). Fe-bentonite samples were put into Au-tubes, and they were kept at 250$$^{circ}$$C in autoclave for 1 month and 6 months. 6 months sample was confirmed the peak of a iron-containing 7AA non-expandable clay mineral (i.e. berthierin). For the confirmation of reversibility of ion-exchange, 6 months sample was mixed with 1M-NaCl solution. XRD analysis for the treated sample under humidity controlled condition showed a same result with that of Na-bentonite. This means that a change of the sample was only ion-exchange reaction in the bentonite inter-layer, and bentonite properties was not changed.

JAEA Reports

Development of cylindrical type proton-recoil proportional counter and its use for absolute measurements of neutron fluences at 144, 250 and 565 keV monoenergetic calibration fields

Saegusa, Jun; Tanimura, Yoshihiko; Yoshizawa, Michio

JAEA-Research 2006-065, 40 Pages, 2006/10

JAEA-Research-2006-065.pdf:4.6MB

A proton-recoil proportional counter has been developed as a standard instrument for evaluating neutron fluence at the accelerator-based neutron calibration field. The counter consists of a cylindrical cathode and an external housing in which hydrogen is filled as counting gas. For neutrons in the energy range between 50 keV and 1 MeV, the fluence of the neutron field can be determined by an absolute measurement with the counter. The developed counter was used for determining the reference neutron fluence of the monoenergetic neutron fields at the Facility of Radiation Standards of JAEA. For the 144, 250 and 565 keV neutron fields, it has been able to measure the fluences with the standard uncertainties less than 2 %. The measured fluences for 144 and 565 keV have showed good agreement with the values estimated by another measurements with a transfer instrument traceable to primary standards.

JAEA Reports

Study on design method for gas entrainment prevention from a liquid surface based on a computational fluid dynamics method; 1st proposal of the design guideline

Sakai, Takaaki; Ito, Kei; Uchibori, Akihiro; Kimura, Nobuyuki; Ezure, Toshiki; Kamide, Hideki; Ohshima, Hiroyuki

JAEA-Research 2006-066, 43 Pages, 2006/10

JAEA-Research-2006-066.pdf:8.31MB

Japan Atomic Energy Agency has conducted a conceptional design study of a sodium-cooled fast reactor in a frame work of the FBR feasibility study. The plant system concept for a commercial step is intended to minimize a vessel diameter to achieve an economical competitiveness. Therefore, the coolant in the vessel has relatively higher velocity than conventional designs. Because of the high velocity, gas entrainment prevention from a liquid surface in the reactor vessel becomes one of important issues for the thermal-hydraulic design. Gas entrainment may cause the core power fluctuation or heat transfer reduction. Therefore, it is necessary to clarify the avoidance of the phenomena in the design conditions with sufficient allowance. The prevention of gas entrainment phenomena was prospectively confirmed by a 1/1.8 scale model water experiment. The large scale sodium experiment, however, needs very high costs to validate the design. The design method by a utilization of a computational fluid dynamics (CFD) method is one of possible choices for the gas entrainment prevention design. In this study, the gas entrainment prevention from vortex dimples at the liquid surface was investigated by a working group that consists of members from Universities, Research institutes, Utilities and Manufacturers, in order to establish a design method for the gas entrainment prevention by a CFD method. The research work was commenced from 2002 and performed for four years. This report is the first proposal of the design guideline for the gas entrainment prevention using CFD methods, from the achievements of the working group activity.

JAEA Reports

Experimental study of gas entrainment at free surface; Development of circulation and gas core of surface vortex

Ezure, Toshiki; Kimura, Nobuyuki; Kobayashi, Jun; Ito, Masami*; Kamide, Hideki

JAEA-Research 2006-067, 35 Pages, 2006/10

JAEA-Research-2006-067.pdf:7.49MB

A sodium cooled reactor has been investigated in the feasibility study of FBR cycle. In the study, a compact reactor vessel was designed, and the cover gas entrainment (GE) at the free surface is one of the significant issues. It is required to clarify the criterion of GE at free surface. GE at the free surface could be categorized into following three types, wave break, submerged flow, surface vortex. However, there was no clear quantitative evaluation method and criteria regarding the onset condition of GE by the surface vortex. In the present study, some experiments were performed focusing on the transient phenomena of GE by surface vortex. The relationship between circulation and length of gas core were measured by the particle image velocimetry and visualization. From the results of this study, the relationship between gas core length and probability of GE was clarified, and time-delay between the increase of circulation and the increase of gas core was found.

JAEA Reports

Evaluation on gas entrainment in reactor vessel using 1/1.8th scaled model; Investigation on dominant factors based on occurrence map and mechanism for gas entrainment

Kimura, Nobuyuki; Ezure, Toshiki; Tobita, Akira; Ito, Masami*; Kamide, Hideki

JAEA-Research 2006-068, 56 Pages, 2006/10

JAEA-Research-2006-068.pdf:23.42MB

For an innovative sodium cooled fast reactor, a compact reactor vessel is designed to reduce the construction cost, where sodium flow velocity increases. One of the thermal hydraulic issues in this design is gas entrainment (GE) at free surface in the reactor vessel (R/V). Dipped plates (D/P) are set below the free surface in order to prevent the GE. To evaluate GE, we made the partial apparatus modeled the 90 degree sector of the R/V with a central focus on the hot leg (H/L) pipe circumferentially and region above the D/Ps vertically. We obtained an occurrence map of the GE in order to evaluate the dominant factor for the GE. In the map, it was found that there were two kinds of the GE phenomena. One of the GE occurred at the downstream region of the cold leg pipe. Other one broke out at the region between the H/L pipe and the R/V wall. The mechanisms of the GE at the two regions were clarified by applying the PIV.

JAEA Reports

Uncertainty analysis of solubility for selenium and neptunium by Monte Carlo simulation of geochemical modeling code

Takeda, Seiji; Kimura, Hideo

JAEA-Research 2006-069, 16 Pages, 2006/11

JAEA-Research-2006-069.pdf:3.15MB

Solubility of radioactive element in waste glass is a key parameter for safety assessment of geologic disposal of radioactive waste. JAEA has developed the probabilistic analysis code of uncertainty of solubility limit for radioactive substances in a geological disposal (PA-SOL) to estimate the uncertainty of solubility associated with the uncertainties in both groundwater chemistry and thermodynamic data. The selenium and neptunium solubilities are calculated for the uncertainty of groundwater chemistry in repository environment, based on previous data of pore water chemistry in buffer materials, using the representative thermodynamic database, EQ3/6-TDB and JNC-TDB. The results of uncertainty analyses show that the highly effective parameters of groundwater chemistry are identified as the Eh-pH condition and iron concentration for selenium and the Eh-pH condition and carbonate concentration for neptunium.

JAEA Reports

Reliability improvement of hydro-geochemical and hydrogeological model on geological environment in Horonobe Underground Research Laboratory project; Report on the 2005 FY research (Joint research)

Hama, Katsuhiro; Kunimaru, Takanori; Kurikami, Hiroshi; Sasamoto, Hiroshi; Takahashi, Yasuhiro*; Haginuma, Masashi*; Ishii, Tomoko*; Matsuo, Yuji*

JAEA-Research 2006-070, 93 Pages, 2006/09

JAEA-Research-2006-070.pdf:23.48MB
JAEA-Research-2006-070(errata).pdf:0.08MB

Japan Atomic Energy Agency and Institute of Research and Innovation have started collaborative study in order to enhance the reliability of technology for the geological disposal of High-level Radioactive Waste (HLW) since 2005 fiscal year. In this collaborative study, the analysis has carried out to estimate long-term evolution of groundwater chemistry and groundwater flow. The study items and results are summarized in this report. (1) Groundwater Chemistry. The following items have been carried out: 1. Multivariate analysis. 2. Estimation of distribution of groundwater chemistry. 3. Construction of geochemical model. (2) Groundwater Flow: 1. Construction of geological model. 2. Construction of hydrogeological model. 3. Groundwater flow analysis. The modeling work will be continued in 2006 fiscal year. The applicability of the modeling methodology will also be evaluated.

JAEA Reports

Evaluation of deuteron-induced activation for IFMIF accelerator structural materials

Nakao, Makoto*; Hori, Junichi*; Ochiai, Kentaro; Kubota, Naoyoshi; Sato, Satoshi; Yamauchi, Michinori; Ishioka, Noriko; Suto, Hiroyuki*; Nishitani, Takeo

JAEA-Research 2006-071, 37 Pages, 2006/11

JAEA-Research-2006-071.pdf:6.72MB

In the design of IFMIF, long-term operation with total facility availability of at least 70 % is required. However, activation of structural materials by deuteron beam limits maintenance, which causes lower facility availability. Thus it is essential to prepare deuteron-induced activation cross section database and to select low activation materials based on it. In this work, we measured deuteron-induced activation cross sections of aluminum, vanadium, chromium, manganese, iron, nickel, copper, tantalum, tungsten and gold. The measured cross sections were compared with other experimental data and calculations. Deuteron-induced activities of nuclides produced in SS316 and F82H alloys used as the accelerator structural material were also measured to validate the measured cross sections comprehensively. It demonstrated that the measured activities of almost all the nuclides were in agreement with evaluated ones based on the measured cross sections within error.

JAEA Reports

A Study on effective system depressurization during a PWR vessel bottom break LOCA with HPI failure and gas inflow prevention (ROSA-V/LSTF test SB-PV-05)

Suzuki, Mitsuhiro; Takeda, Takeshi; Asaka, Hideaki; Nakamura, Hideo

JAEA-Research 2006-072, 144 Pages, 2006/11

JAEA-Research-2006-072.pdf:17.16MB

A small break loss-of-coolant accident (SBLOCA) experiment was conducted at the Large Scale Test Facility (LSTF) of ROSA-V program to study effects of accident management (AM) measures on core cooling, which are important in case of total failure of high pressure injection (HPI) system during an SBLOCA at a pressurized water reactor (PWR). The LSTF is a full-height and 1/48 volume-scaled facility simulating a 4-loop Westing-house-type PWR (3423 MWt). The experiment, SB-PV-05, simulated a PWR vessel bottom SBLOCA with a rupture of nine instrument tubes, which is equivalent to 0.18% cold leg break. It is clarified that AM actions with steam generator (SG) depressurization to achieve a primary loop cooling rate at -55 K/h and auxiliary feedwater supply for 30 minutes are effective to avoid core uncovery by actuating the low pressure injection (LPI) system. It is also shown through the comparison with the previous experiment of SB-PV-03 that prevention of non-condensable gas inflow from the accumulator injection system (AIS) is very important to actuate the LPI to achieve adequate core cooling. This report presents experiment results of SB-PV-05 in detail and shows the effects of gas inflow prevention on core cooling through the estimation of primary coolant mass and energy balance in the primary system.

JAEA Reports

Horonobe Underground Research Laboratory Project; Investigation report for the 2005 fiscal year

Matsui, Hiroya; Niizato, Tadafumi; Yamaguchi, Takehiro

JAEA-Research 2006-073, 72 Pages, 2006/11

JAEA-Research-2006-073.pdf:16.91MB
JAEA-Research-2006-073(errata).pdf:0.08MB

The investigations in 2005 fiscal year (2005/2006) were focused on the Hokushin area, which was selected as the area for laboratory construction. The main investigation region extends over approximately 3km$$times$$3km. Geophysical, geological and surface hydrogeological investigations are carried out to acquire the geoscientific data needed to develop techniques for investigating the geological environment. And the borehole investigation at HDB-11 was finished in 2005.

JAEA Reports

Horonobe Underground Research Laboratory project investigation program for the 2006 fiscal year

Matsui, Hiroya; Niizato, Tadafumi; Yamaguchi, Takehiro

JAEA-Research 2006-074, 36 Pages, 2006/11

JAEA-Research-2006-074.pdf:12.62MB

The Horonobe URL Project is planned to extend over a period of 20 years. The investigations will be conducted in three phases, namely "Phase 1: Surface-based investigations", "Phase 2: Construction phase" (investigations during construction of the underground facilities) and "Phase 3: Operation phase" (research in the underground facilities). This report summarizes the investigation program for the 2006 fiscal year (2006/2007), the second year of the Phase 2 investigations. The investigations in the 2006 fiscal year are focused on the Hokushin area of Horonobe, which was selected as the area for URL construction. The main investigation region extends over approximately 3km$$times$$3km. Construction of the underground facilities, which was initiated in the 2005 fiscal year, is ongoing and Phase 2 investigations are underway. A progress report on the surface-based investigations (Phase 1) is also being prepared. Regarding the surface facilities, construction of the Research and Administration Facility and the Test Facility will be completed in May 2006. Construction of the Public Information House is still continuing and preparation of the exhibits has started. A preliminary design will be drawn up for the International Communication House.

JAEA Reports

Study on coated layer material characteristics of coated particle fuel FBR, 3; Thick layer coating process of TiN and gas-nitriding treatment of Ti metal

Naganuma, Masayuki; Mizuno, Tomoyasu

JAEA-Research 2006-075, 65 Pages, 2006/12

JAEA-Research-2006-075.pdf:30.16MB

Helium gas cooled FBR is one of attractive core concepts, so that the design studies have been performed in the Japanese feasibility study. Here, the coated particle fuelling core is considered to be promising. One of key issues of that fuel is the coated layer material, and TiN is regarded as one of possible materials. Therefore, tests of thick coating of TiN and gaseous nitriding of Ti metal were conducted in this study. In the thick coating tests, PVD (Physical Vapor Deposition) and CVD (Chemical Vapor Deposition) are selected as coating methods. As a result, both methods are found to have capability to form 100 micrometer that is the target of the practical design. Then, as an alternative method to form thick layer, authors contrived the method to nitride Ti metal by nitrogen gas and conducted tests. As a result, the specimen with 100 micrometer thickness was found to be nitrided entirely in 1,200$$^{circ}$$C and 48 hours. However, the nitrided specimen has tendency to be brittle.

JAEA Reports

Study on extraction process for U-Pu separation using N,N-di-(2-ethyl)hexylbutanamide; Analysis of separation behavior of U and Pu by a simple calculation code

Ban, Yasutoshi; Asakura, Toshihide; Morita, Yasuji

JAEA-Research 2006-076, 22 Pages, 2006/11

JAEA-Research-2006-076.pdf:2.89MB

Extraction behavior of U and Pu by N,N-di-(2-ethyl)hexylbutanamide (D2EHBA) in a mixer-settler type extractor was analyzed by a simple calculation code. Analysis was carried out with the consideration of acid and element concentration effects on the distribution ratios of U and Pu. A flow-sheet that gives the separation of U and Pu by adjusting nitric acid concentration without the use of Pu reductant was obtained. According to the analysis based on this flow sheet, U/Pu ratio in the Pu product stream, the ratio of Pu in raffinate, and the ratio of Pu in the Pu product stream were 1.06, less than 0.1%, and more than 99.9%, respectively.

JAEA Reports

Study on reactor core and fuel design of sodium-cooled fast reactor (Metal fuel core); Results in JFY 2005

Oki, Shigeo; Sugino, Kazuteru; Ogawa, Takashi; Aida, Tatsuya*; Hayashi, Hideyuki

JAEA-Research 2006-077, 86 Pages, 2006/11

JAEA-Research-2006-077.pdf:6.34MB

Core and fuel design study of ${it the sodium-cooled metal fuel core with high reactor outlet temperature}$ was performed. The reference specification of the large-scale (1,500 MWe) and the middle-scale (750 MWe) cores were proposed as a final result of ${it FS phase-II}$. Since the local conversion ratio of any of the core points is made close to unity with single Pu enrichment, it is possible to minimize the necessary coolant flow rate for the core region and then, accept high reactor outlet temperature of 550$$^{circ}$$C. By the rationalization of hot spot factors, the coolant flow distribution design can be optimized to 5 regions for the large-scale core and 8 regions for the middle-scale core, respectively. It was also confirmed that the core specification met the criteria of fuel-assembly integrity, as well as those of shielding design. For further improvement on the reactor outlet temperature condition, the reduction of the maximum cladding inner-wall temperature was investigated with the reflection of the actual control rod insertion depth and the rationalization of the excess-reactivity uncertainty. An alternative core design was investigated by adopting the PNC-FMS steel as the cladding material instead of the ODS steel. As a result of the investigation of extending the control rod lifetime, three-cycle lifetime (which is the same as fuel assemblies) could be possible by means of the reductions in $$^{10}$$B enrichment and B$$_{4}$$C pellet diameter.

JAEA Reports

Fundamental concept of knowledge management system for geological disposal technology

Umeki, Hiroyuki; Osawa, Hideaki; Naito, Morimasa; Nakano, Katsushi; Makino, Hitoshi

JAEA-Research 2006-078, 45 Pages, 2006/12

JAEA-Research-2006-078.pdf:5.75MB

Geological disposal technology is a multi-disciplinary field and needs a wide range of relevant information to develop its safety case. Knowledge is defined here in the very widest sense to include all of the information, both explicit and tacit, which underpins a repository project. Knowledge management covers all aspects of the development, integration, quality assurance, communication and maintenance/archiving of such knowledge - including data, information, understanding and experience. In order to ensure that required knowledge is accessible to all stakeholders, including the implementer, the regulator, political decision-makers and general public and that gaps can be identified and prioritised, it is important that knowledge base are structured in a clear and logical manner. In this report, the fundamental concept of the knowledge management system (KMS) for the Japanese high-level radioactive waste (HLW) disposal programme is discussed by illustrating characteristics of an ideal formal KMS and associated knowledge base which could be supported by sophisticated information technology tools, based on the knowledge vision described in the research report "H17: Development and management of the technical knowledge base for the geological disposal of HLW; Knowledge Management Report". In addition, a mid-term plan up to the fiscal year 2009 is discussed for development and installation of the KMS.

JAEA Reports

Measurement of groundwater table in shallow boreholes

Seno, Shoji; Kurikami, Hiroshi; Yabuuchi, Satoshi; Hara, Minoru

JAEA-Research 2006-079, 22 Pages, 2007/03

JAEA-Research-2006-079.pdf:5.57MB

Horonobe Underground Research Laboratory of Japan Atomic Energy Agency has been investigating surface hydrogeological features in and around the Horonobe Underground Research Laboratory (URL) area as a part of Horonobe URL project. The objective of measurement of groundwater tables in shallow boreholes is to understand the distribution of the groundwater table and its seasonal fluctuation that will be boundary conditions of a groundwater flow analysis. This report shows the results of the measurement since December 2003 to October 2005 every few month and discussion on the surface hydrogeological features. The results are as follows: (1) The real watershed between the Shimizu river basin and the Penke-ebekorobetsu river basin exists a little to the south of the geographical watershed among the monitoring line. (2) Snow fall/melting is the largest impact on the annual fluctuation of the groundwater table, while influence of precipitation is temporal. (3) The amplitude of the fluctuation of groundwater table depends on the location. (4) Peak level of groundwater table after precipitation is gentler than that of the river water table and appears two or three days later.

JAEA Reports

Study on the corrosion assessment of overpack welds, 3 (Joint research)

Mitsui, Hiroyuki*; Takahashi, Rieko*; Taniguchi, Naoki; Otsuki, Akiyoshi*; Asano, Hidekazu*; Yui, Mikazu

JAEA-Research 2006-080, 322 Pages, 2006/12

JAEA-Research-2006-080.pdf:90.52MB

There is some possibility that the corrosion resistance of overpack welds is different from that of base metal due to the differences of material properties. In this study, corrosion behavior of welded joint for carbon steel was compared with base metal using the specimens taken from welded joint model fabricated by TIG, MAG and EBW respectively. The corrosion tests were performed for following four items. (1) Passivation behavior and corrosion type, (2) Propagation of general corrosion, pitting corrosion and crevice corrosion under aerobic condition, (3) Stress corrosion cracking susceptibility, (4) Propagation of general corrosion and hydrogen embrittlement under anaerobic condition. The results of these corrosion tests indicated that the corrosion resistance of welded metal by TIG and MAG was inferior to base metal for general corrosion, pitting corrosion and crevice corrosion. It was implied that the filler materials used for welding affected the corrosion resistance. No deterioration of corrosion resistance was observed in any corrosion modes for EBW, which does not need filler material. The susceptibility to stress corrosion cracking of welded metal and heat affected zone was lower than that of base metal.

JAEA Reports

Analysis on introduction scenario of partitioning-and-transmutation cycle and its benefit based on mass-balance calculation of actinides

Nishihara, Kenji; Oigawa, Hiroyuki

JAEA-Research 2006-081, 91 Pages, 2006/12

JAEA-Research-2006-081.pdf:8.21MB

The double-strata fuel cycle consisting of the commercial fuel cycle and Partitioning-and-Transmutation (P-T) cycle is proposed to reduce the burden for conditioning and disposal of the high level radioactive waste (HLW). The P-T cycle can be adapted to various fuel cycles for the commercial power plants. In the present report, introduction scenarios of P-T cycle were obtained based on mass balances analysis of actinide nuclides for several types of commercial nuclear fuel cycle during the next 200 years. Possible contributions to radioactive waste management were also quantified. The SCENARIO code was developed to calculate the mass balance.

JAEA Reports

Performance evaluation of CPF shredder type mechanical crusher with simulated core fuel pin

Nakahara, Masaumi; Sano, Yuichi; Aose, Shinichi

JAEA-Research 2006-082, 56 Pages, 2006/12

JAEA-Research-2006-082.pdf:14.3MB

The shredder type mechanical crusher was developed for using in a hot cell in Chemical Processing Facility, and the first crush experiment with this crusher was carried out at July 2004 using the simulated core fuel pin. This experiment showed that the crushed fragments could not be grinded efficiency because screen blade vibrated up and down during the operation. Additionally, the strength of screen blade block was insufficient to crush the sheared fuel pieces stably. Therefore, about 70% of fuel was recovered in maximum. Based on the results of the first experiment, screen blade was fixed up mainly and the second experiment was carried out with improved apparatus at September 2005. In this experiment, about 96% of fuel could be recovered in maximum because screen blade was stable during the operation.

JAEA Reports

Study on transient temperature measurement at fuel clad surface in NSRR experiments

Vincent, B.*; Sugiyama, Tomoyuki; Fuketa, Toyoshi

JAEA-Research 2006-083, 39 Pages, 2007/01

JAEA-Research-2006-083.pdf:3.1MB

In NSRR experiments, fuel clad temperature is measured by thermocouples welded on the clad surface. This report presents evaluation of temperature measurement error due to thermal conduction to the thermocouple. Analyses using the CASTEM code showed that a transient capacitive phase occurs while clad temperature remains below 400 $$^{circ}$$C. The temperature error reaches the maximum in this phase; -100 $$^{circ}$$C (underestimation) without zirconia layer at the clad outer surface and +150 $$^{circ}$$C (overestimation) with 100$$mu$$m zirconia layer. A "fin effect" phase starts when the clad temperature exceeds 400 $$^{circ}$$C and the film boiling regime is clearly established, during which the error stabilizes between -20 $$^{circ}$$C without zirconia layer and +50 $$^{circ}$$C with 100$$mu$$m zirconia layer. The influence of the thermocouple is limited within a distance of about 0.5 mm. A transfer function was determined from the calculation results to estimate the accurate clad temperature from the thermocouple records.

JAEA Reports

Study on groundwater flow system in a sedimentary rock area; Case study for the Yoro river basin, Chiba Prefecture

Sakai, Ryutaro; Munakata, Masahiro; Kimura, Hideo

JAEA-Research 2006-084, 16 Pages, 2007/01

JAEA-Research-2006-084.pdf:1.74MB

Japan Atomic Energy Agency (JAEA) has started to investigate a sedimentary rock area in the Yoro river basin, in Chiba Prefecture. Hydro-chemical conditions of the regional groundwater were discussed based on temperature, chemical compositions, isotopic ratios of hydrogen and oxygen, and the isotopic age of radioactive carbon for water samples collected from wells, rivers and springs in the Yoro river basin. It was found that the groundwater system in this basin consists of types of water: Ca-HCO$$_{3}$$ type water, Na-HCO$$_{3}$$ type water and NaCl type water. The Ca-HCO$$_{3}$$ type water is meteoric water cultivated several thousand years or after, the Na-HCO$$_{3}$$ type water is meteoric water cultivated under cold climates several to twenty thousand years ago. The NaCl type water is fossil brine water formed twenty thousand years ago. It was also observed that the Na-HCO$$_{3}$$ type water upwelled at the surface originates from GL-200m to -400m.

JAEA Reports

Research on the state-of-the-art of probabilistic safety assessment for non-reactor nuclear facilities (I)

Yoshida, Kazuo; Abe, Hitoshi; Yamane, Yuichi; Tashiro, Shinsuke; Muramatsu, Ken

JAEA-Research 2006-085, 81 Pages, 2007/02

JAEA-Research-2006-085.pdf:5.42MB

Japan Atomic Energy Agency (JAEA) entrusted with research on the state-of-the-art of probabilistic safety assessment (PSA) for non-reactor nuclear facilities (NRNF) to the Atomic Energy Society of Japan (AESJ). The objectives of this research is to obtain the basic useful information related for establishing the quantitative performance requirement and for risk-informed regulation through qualifying issues needed to be resolved for applying PSA to NRNF. A special committee of "research on the analysis methods for accident consequence in NFRF" was organized in the AESJ. The research activities of the committee were mainly focused on the analysis method for upper bounding consequences of accidents such as events of criticality, explosion, fire and solvent boiling postulated in NRNF resulting in release of radio active material to the environment. This report summarizes the result of research conducted by the committee in FY 2004.

JAEA Reports

The Results of the investigation on rock mechanics in HDB-3$$sim$$8 boreholes and study about the rock mechanical model around the Horonobe URL construction area

Niunoya, Sumio; Matsui, Hiroya

JAEA-Research 2006-086, 97 Pages, 2007/01

JAEA-Research-2006-086.pdf:11.55MB

This paper shows the result of the rock mechanical investigations on HDB-3 to HDB-8 which have been carried out from 2002 to 2003 as a part of the investigations of Horonobe URL(Underground Research Laboratory) project. The objectives of the investigation are as follows, to confirm the distribution of rock mechanical properties by laboratory and in-situ tests, and to obtain the rock mechanical data in order to understand the deference of state of stress between the west and east side area of Omagari fault. Omagari fault separated the URL construction area between west and east side. So that, we arranged the rock mechanical data which were obtained from each laboratory tests and in-situ measurements. According to the results of the investigations, we found that it is good agreement with the distribution and the state of rock mechanical properties in each borehole with regardless of east or west area of Omagari fault. So, we suppose that the rock mechanical model of sedimentary rock around the URL construction area is nearly continuum, because that the intact part of sedimentary rock around here is weak character, and that the scale of fracture confirmed by geological investigation is too small to influence rock mechanical properties. And according to the model which is constructed by the results of the physical and mechanical investigations, we can explain about the rock physical and mechanical properties in common through 3 zones with regardless of location.

JAEA Reports

Study on influence to core and fuel design by adopting vibro-packed fuel and sphere-pac fuel

Naganuma, Masayuki; Aida, Tatsuya*; Hayashi, Hideyuki

JAEA-Research 2006-087, 72 Pages, 2007/02

JAEA-Research-2006-087.pdf:4.5MB

In the core and fuel design of sodium-cooled MOX fuel FR of FS, core designs with pellet fuel were mainly evaluated. However, vibro-packed fuel and sphere-pac fuel are also considered one of the candidates. Besides, the design must be affected by difference of the fuel behavior, however, the influence had not yet been evaluated adequately. Thus, authors examined the fuel thermal design model and evaluated the influences to the core and fuel design quantitatively. As a result, in the fuel thermal design model, selection of restructuring conditions was found to be important. Proper values were evaluated from the viewpoint of fitting PIE results. In applying this model, limitation of the stationary LHR was reduced to 390 W/cm. For FS phase-II reference cores, though it is required to modify specifications due to the decrease of LHR, the influence to nuclear performance is found to be benign. Therefore, the design that meets the requirements and targets of FS is possible for those fuels.

JAEA Reports

Development of fuel microspheres fabrication by the external gelation process

Tomita, Yutaka; Morihira, Masayuki; Tamaki, Yoshihisa*; Nishimura, Kazuhisa*; Shoji, Shuichi*; Kihara, Yoshiyuki; Kase, Takeshi; Koizumi, Tsutomu

JAEA-Research 2006-088, 95 Pages, 2007/01

JAEA-Research-2006-088.pdf:23.02MB

JAEA has developed sphere-pac fuels in the feasibility study on commercialized FBR cycle systems as one of the candidates for low decontamination TRU fuels. Optimization of the fabrication condition for coarse spheres, development of an improved external gelation process, and examination of peculiar problems for the low decontamination fuel were carried out in Phase II. The results are shown as follows. (1) Fabrication condition of coarse spheres was optimized. (2) Feasibility of the improved external gelation process was confirmed. (3) Rare earth elements did not bring any problem for the characteristic of spheres and fabrication condition. (4) Radiation resistant data of the feed solution was acquired. Results of tests show the feasibility of the external gelation process for the low decontamination TRU fuel microsphere fabrication.

JAEA Reports

Kinetic simulations of electrostatic plasma waves using Cubic-Interpolated-Propagation scheme

Lesur, M.*; Idomura, Yasuhiro; Tokuda, Shinji

JAEA-Research 2006-089, 29 Pages, 2007/01

JAEA-Research-2006-089.pdf:1.67MB

Kinetic properties of small amplitude waves in an electron plasma are studied using analytic and numerical calculations based on the Vlasov-Poisson system. The dispersion relation of plasma waves in a Maxwellian plasma is solved using Laplace-Fourier transform, and it is shown that waves decay in time by Landau damping. A simulation code for solving the Vlasov-Poisson system in phase space is developed using the Cubic-Interpolated-Propagation (CIP) scheme, and Landau damping is successfully calculated numerically. Finally, the stability of an electron plasma with a beam component is discussed by applying these analytic and numerical approaches.

JAEA Reports

Measurements of spatial distribution of intensity and energy spectrum of neutrons from coupled hydrogen moderators by using a position sensitive detector (Cooperative research)

Kai, Tetsuya; Kamiyama, Takashi*; Hiraga, Fujio*; Kato, Takashi; Kiyanagi, Yoshiaki*

JAEA-Research 2006-090, 35 Pages, 2007/02

JAEA-Research-2006-090.pdf:13.67MB

Spatial distributions of neutron intensities were measured based on the pinhole camera principle to confirm a peculiar one for a para-hydrogen (H$$_2$$) moderator by neutron transportation calculations. Para- H$$_2$$ and ortho-rich-H$$_2$$ (ortho/para ratio was 60:40-70:30) were utilized. The results are consistent with the calculations. Position dependences of energy spectra were measured for further discussions. Both moderators exhibit the intensity enhancements at the fringe part between 15 and 100 meV. Below 15 meV, the para-H$$_2$$ moderator exhibits higher intensity at the fringe part, while a intensity decrease was recognized at the fringe part of the ortho-rich-H$$_2$$ one. The reasons for the peculiar distribution are due to the characteristics of para- H$$_2$$: the effective slowing-down reaction for thermal neutrons and the higher transmission rate for cold neutrons. The present result is one of the basis confirming the reliability of design calculation used in the J-PARC project.

JAEA Reports

Research on advanced technology of performance assessment for geological disposal of high-level radioactive waste (Joint research)

Geological Isolation Research and Development Directorate; Advanced Waste System Research Project, Radioactive Waste Management Funding and Research Center*

JAEA-Research 2006-091, 140 Pages, 2006/12

JAEA-Research-2006-091.pdf:9.66MB

Research on advanced technology of performance assessment for geological disposal of high-level redioactive waste has been performed by a joint research program between JAEA and RWMC using their own technology and know-how. The following 5 items have been considered in the program. (1)Planning of basic strategy for development of analysis technology on nuclides migration for various scales, (2)Development of analysis technology for a scale of vitrified waste, (3)Development of analysis technology for a scale of disposal facility, (4)Development of integration technology for geochemical information, (5)Development of understanding promotion technology for logic to explain safety Based on the overall plan defined in this year, it would be important to continue the concrete research and development to resolve the subjects in the future.

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