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JAEA Reports

PROMILLE Database As a Part of JNC Reactor Physics Analytical System for BFS-2 Fast Critical Facility Experiments Analysis

Bedniak, S.

JNC-TN9400 2001-040, 97 Pages, 2000/12

JNC-TN9400-2001-040.pdf:6.6MB

The PROMILLE database for experimental data from the BFS-2 fast critical facility (Institute of Physics and Power Engineering (IPPE), Russia) has been developed and embedded into the JNC reactor physics analytical system to provide a strict documentation format, a common data source for different analytical tools and a unique export interface with different reactor codes. PROMILLE should be considered not only as a database but also as a bank of interfaces because of its dynamic role in the analytical process. The database currently accepts data from the simulation materials (pellets, tubes and bars) as well as full cores descriptions. A core description involves all different unit cells forming loading elements, all types of the loading elements forming a layout and the layout itself. In fact it is a description of criticality experiments. Export interfaces for the CITATION-FBR code and the SLAROM and CASUP codes have been developed. The PROMILLE software was developed with MS visual Basic 6.0 and the data is kept in the data tables generated with the MS Access database management system. Data for eight BFS-2 assembly configurations have been incorporated. They include BFS-58-lil uranium-free plutonium assembly with inert material included in its fuel matrix and also seven BFS-62 modifications simulating different stages of investigation of MOX fuel based BN-600 core.

JAEA Reports

Development of a standard data base for FBR core nuclear design (XIII); Analysis of small sample reactivity experiments at ZPPR-9

Sato, Wakaei*; Fukushima, Manabu*;

JNC-TN9400 2001-026, 90 Pages, 2000/09

JNC-TN9400-2001-026.pdf:2.61MB

A comprehensive study to evaluate and accumulate the abundant results of fast reactor physics is now in progress at O-arai Engineering Center to improve analytical methods and prediction accuracy of nuclear design for large fast breeder cores such as future commercial FBRs. The present report summarizes the analytical results of sample reactivity experiments at ZPPR-9 core, which has not been evaluated by the latest analytical method yet. The intention of the work is to extend and further generalize the standard data base for FBR core nuclear design. The analytical results of the sample reactivity experiments (samples: PU-30, U-6, DU-6, SS-1 and B-1) at ZPPR-9 core in JUPITER series, with the latest nuclear data libraly JENDL-3.2 and the analytical method which was established by the JUPITER analysis, can be concluded as follows: The region-averaged final C/E values generally agreed with unity within 5% differences at the inner core region. However, the C/E values of every sample showed the radial space-dependency increasing from center to core edge, especially the discrepancy of B-1 was the largest by 10%. Next, the influence of the present analytical results for the ZPPR-9 sample reactivity to the cross-section adjustment was evaluated. The reference case was a unified cross-section set ADJ98 based on the recent JUPITER analysis. As a conclusion, the present analytical results have sufficient physical consistency with other JUPITER data, and possess qualification as a part of the standard data base for FBR nuclear design.

JAEA Reports

Decay heat-measurement of minor actinides at YAYOI; Measurement results of gamma-ray decay heat

Okawachi, Yasushi;

JNC-TN9400 2001-001, 100 Pages, 2000/08

JNC-TN9400-2001-001.pdf:2.83MB

Gamma-ray decay heat released from fission products has been measured for fast neutron fissions of U-235 and Np-237 using the radiation spectrometry method. The samples were irradiated at fast neutron source reactor "YAYOI" of the University of Tokyo. Gamma-ray energy spectra were measured using a NaI(TI) scintillation detector. And, the number of fission was evaluated from measured gamma spectra by Ge detector. For the measured gamma-ray, the background count was subtracted from the pulse height distribution of 1024 channels measured. The results were grouped by 340 channels to match the response matrix of the detector. This distribution was converted to energy spectra using the FERDO code and the response matrix of the detector. Normalized decay heat by the number of fission was obtained by integration of the energy spectra for each time step. The finite irradiation decay heat that is directly obtained by experiments can not be compared with experimental results and calculational results obtained under various irradiation conditions. So, the finite irradiation decay heat was converted to the fission burst decay heat. These results were compared with summation calculations using JNDC-V2 dacy data file. The present results on U-235 were compared with other experimental data using the same method. The present results agreed with other experimental data using the same method within 10%, suggesting the repeatability of experimental method. The present results on Np-237 were compared with the results of summation calculations using JNDC-V2 decay data file. As the result, the present results agreed with summation calculations within 8%. Probelms to be solved for the future are estimation of the experimental error, re-evaluation of the number of fission using updated nuclear data. To improve accuracy of decay heat data in shorter cooling time range, less irradiation experiment will be useful. Furthermore, to improve accuracy of decay heat data in longer cooloing ...

JAEA Reports

Analyse on the BFS critical experiments; An analysis on the BFS-62-1 assembly

; Iwai, Takehiko*;

JNC-TN9400 2000-098, 182 Pages, 2000/07

JNC-TN9400-2000-098.pdf:5.74MB

In order to support the Russian excess weapons plutonium disposition, the international collaboration has been started between Japan Nuclear Cycle Development Institute (JNC) and Russian Institute of Physics and Power Engineering (IPPE). In the frame of the collaboration, JNC has carried out analyses on the BFS-62 assemblies that are constructed in the fast reactor critical experimental facility BFS-2 of IPPE. This report summarizes an experimental analysis on the BFS-62-1 assembly, which is the first core of the BFS-62 series. The core contains the enriched U0$$_{2}$$ fuel surrounded by the U0$$_{2}$$ blanket. The standard analytical method for fast reactors has been applied, which was used for the JUPITER and other experimental analyses. Due to the lack of the analytical data the 2D RZ core calculation was mainly used. The 3D XYZ core calculation was applied only for the preliminary evaluation. Further in terms of the utilization of the BFS experimental analysis data for the standard data base for FBR core design, consistency evaluation with JUPITER experimental analysis data has been performed using the cross-section adjustment method. As the result of analyses, good agreement was obtained between calculations and experiments for the criticality and the reaction rate ratio. However, it was found that accurate evaluation of the reaction rate distribution was impossible without exact consideration of the arrangement of the two types of sodium (with and without hydrogen impurity), which can be accommodated by the 3D core analysis, thus it was essentia1. In addition, it was clarifie that there was a room for an improvement of the result on the reaction rate distribution in the blanket and shielding regions. The application of the 3D core calculation improved the result on the control rod worth because 3D core model can more exactly consider the shape of the control rod. Furthermore it was judged that the result of the analysis on the sodium void reactivity .....

JAEA Reports

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

; Sato, Wakaei*; Iwai, Takehiko*

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

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

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

JAEA Reports

Analysis Results of Samples Reactivity Measurements at BFS-58-1i1 Assembly on the Basis of JNC Analytical System and BFS Traditional Approach

Bedniak, S.

JNC-TN9400 2000-081, 59 Pages, 2000/05

JNC-TN9400-2000-081.pdf:1.14MB

BFS-58-1i1 assembly is a critical configuration constructed at BFS-2 fast critical facility, IPPE/Obninsk/Russia, containing uranium-free Plutonium fuel in its central zone. Sodium void reactivity effect, spectral averaged cross-sections, Doppler reactivity effect, material samples central reactivity worth (CRW) have been measured there. The results obtained by means of the last experimental technique is a subject of the current report. ln order to make adequate a comparison conditions for calculation and experiment a code for heterogeneous and bilinear corrections for central reactivity worth ratios based on IPPE developments has been written and applied. Some extrapolated experimental data was also examinated from the point of view of their consistency and reliability and a conclusion about it has been made. The previous analysis results made by different laboratolies have been included in the current report and the discrepancies have been discussed.

JAEA Reports

Evaluation of linear heat rates for the power-to-melt tests on "JOYO" using the Monte-Carlo code "MVP"

;

JNC-TN9400 2000-061, 72 Pages, 2000/04

JNC-TN9400-2000-061.pdf:2.29MB

The linear heat rates of the power-to-melt (PTM) tests, performed with B5D-1 and B5D-2 subassemblies on the experimental fast reactor "JOYO", are evaluated with the continuous energy Monte carlo code, MVP. We can apply a whole core model to MVP, but it takes very long time for the calculation. Therefore, judging from the structure of B5D subassembly, we used the MVP code to calculate the radial distribution of linear heat rate and used the deterministic method to calculate the axial distribution. We also derived the formulas for this method. Furthermore, we evaluated the error of the linear heat rate, by evaluating the experimental error of the reactor power, the statistical error of Monte-Carlo method, the calculational model error of the deterministic method and so on. 0n the other hand, we also evaluated the burnup rate of the B5D assembly and compared with the measured value in the post-irradiation tests. The main results are following: (1)B5D-1(B5101, F613632,core center) Linear heat rate: 600W/cm$$pm$$2.2% Burnup rate: 0.977 (2)B5D-2(B5214, G80124,core center) Linear heat rate: 641W/cm$$pm$$2.2% Burnup rate: 0.886

JAEA Reports

LLFP transmutation in various fast reactors using a moderating target

; *

JNC-TN9400 2001-027, 62 Pages, 2000/03

JNC-TN9400-2001-027.pdf:1.66MB

Transmutation property of long-1ived fission product (LLFP) was analyzed for various fast reactor cores having different fuel and coolant types. The inves tigated fast reactor cores were (1)sodium cooled and oxide fueled core, (2)sodium cooled and nitride fueled core (He-bond), (3)sodium cooled and nitride fueled core (Na-bond), (4)sodium cooled and metal fueled core, (5)lead cooled and nitride fueled core (BREST-300), (6)lead cooled and oxide fueled core, and (7)carbon dioxide gas cooled and oxide fueled core (ETGBR). Selected seven LLFPs were $$^{79}$$Se, $$^{93}$$Zr, $$^{99}$$Tc, $$^{107}$$Pd, $$^{126}$$Sn, $$^{129}$$I and $$^{135}$$Cs. Taking them into moderating target subassemblies loaded on the radial blanket region, the transmutation amount and rate of LLFPs for each investigated core were evaluated and compared. Followings are the main results: (1)LLFP transmutation property in a moderating target has little dependence on the fuel and coolant types of the fast reactor core. (2)A middle-sized core has larger LLFP transmutation amount per reactor power than a large-sized core.

JAEA Reports

MA transmutation in various fast reactor core concepts

; Iwai, Takehiko*; Jin, Tomoyuki*

JNC-TN9400 2000-080, 532 Pages, 2000/03

JNC-TN9400-2000-080.pdf:14.98MB

Transmutation Property of minor actinide nuclides (MA) was analyzed for fast reactor cores having different coolant and fuel types in order to obtain basic data for evaluating an ability of the efficient use of resource and reducing the effect on the environment. The investigated fast reactor cores were (a) sodium cooled and oxide fueled core, (b) lead cooled and nitride fueled core (BREST-300), (c) carbon dioxide gas cooled and oxide fueled core (ETGBR), (d)lead cooled and oxide fueled core, (e) sodium cooled and nitride fueled core (both He-bond and sodium-bond), and (f) sodium cooled and metallic fueled core. Followings were observed for the relation between MA transmutation property and the different types of coolant and fuel of fast reactor core. (1) For the MA transmutation rate, the relation "Oxide $$<$$ Metal $$<$$ Nitride" was found out for difference of fuel type. A main reason of the increment of MA transmutation rate is that the neutron flux level rises on nitride and metallic fueled cores in comparison with oxide core. (2) The relation "Lead $$<$$ Sodium and Carbon dioxide" can be seen for the MA transmutation rate in the difference of coolant, but it is not clear whether the cause is driven from the difference of coolant itself on the difference of core design. (3) The changes of MA transmutation property mentioned above are comparatively small.

JAEA Reports

ComparaUve analyses on nuclear charaderistics of water-cooled breeder cores

; Sato, Wakaei*;

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

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

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

JAEA Reports

Development of a standard database for FBR core nuclear design (XI); Analysis of the experimental fast reactor "JOYO" MK-I start up test and oparation data

; Numata, Kazuyuki*

JNC-TN9400 2000-036, 138 Pages, 2000/03

JNC-TN9400-2000-036.pdf:10.16MB

Japan Nuclear Cycle Development lnstitute (JNC) had developed the adjusted nuclear cross-section library in which the results of the JUPITER experiments were renected. Using this adjusted library, the distinct improvement of the accuracy in nuclear design of FBR cores had been achieved. As a recent research, JNC develops a database of other integral data in addition to the JUPITER experiments, aiming at further improvement for accuracy and reliability. ln this report, the authors describe the evaluation of the C/E values and the sensitivity analysis for the Experimental Fast Reactor "JOYO" MK-l core. The minimal criticality, sodium void reactivity worth, fuel assembly worth and burn-up coefficient were analyzed. The results of both the minimal criticality and the fuel assembly worth, which were calculated by the standard analytical method for JUPITER experiments, agreed well with the measured values. 0n the other hand, the results of the sodium void reactivity worth have a tendency to overestimate. As for the burn-up coefficient, it was seen that the C/E values had a dispersion among the operation cycles. The authors judged that further investigation for the estimation of the experimental error will increase the applicability of the integral data to the adjusted library. Furthermore, sensitivity analyses for the minimal criticality, sodium void reactivity worth and fuel assembly worth showed the characteristics of "JOYO" MK-l core in comparison with ZPPR-9 core of JUPITER experiments.

JAEA Reports

Investigation of equilibrium core by recycling MA and LLFP in fast reactor cycle(II); lnvestigation of LLFP confined in eEquilibrium core with element separation

Mizutani, Akihiko; ;

JNC-TN9400 2000-013, 66 Pages, 2000/02

JNC-TN9400-2000-013.pdf:1.97MB

Feasibility study on a self-consistent fuel cycle system has been performed in the nuclear fuel recycle system with fast reactors. ln this system, the self-generated MAs (Minor Actinides) and LLFPs (Long-Lived Fission Products) are confined and incinerated in the fast reactor, which is called the "Equilbrium Core" concept. However, as the isotope separations for selected LLFPs have been assumed in this cycle system, it seems that this assumption is far from realistic one from the viewpoint of economy with respect to the fuel cycle system. ln this study, the possibility for realization of the "Equilibrium Core" concept is evaluated for three fuel types such as oxide, nitride and metallic fuels, provided that the isotopic separation of LLFPs is changed to the element one. This study provides, that is to say, how many LLFP elements can be confined in the "Equilibrium Core" with element separation. This report examines the nuclear properties of the "Equilibrium Core" for various combinations of LLFP incineration schemes from the viewpoints of the risk of geological disposal and the limit in confinable quantity of LLFPs. From the viewpoint of the risk of geological disposal estimated by the retardation factor, it is possible to confine with element separation for T$$_{c}$$, I and Se even in the oxide fueled core. From the standpoint of the limit of confinable amounts of LLFPs, on the other hand, T$$_{c}$$, I, S$$_{e}$$, S$$_{n}$$ and Cs can be confined with element separation in case that the nitride fuel is chosen.

JAEA Reports

Comparative study for minor actinide transmutation in various fast reactor core concepts (1)

JNC-TN9400 2000-007, 77 Pages, 1999/12

JNC-TN9400-2000-007.pdf:2.17MB

Comparative study for various core concepts is being carried out in a frame work of the study for minor actinide (MA) transmutation using a fast reactor. Different fuel types (Oxide, Nitride, Metal) and coolants (Sodium, Lead) were investigated. It is found that neither nitride nor metal-fueled core has significantly more excellent efficiency for MA transmutation comparing with an oxide-fueled core when the basic performance of these cores as a power reactor are fixed. The MA transmutation Properties of lead-cooled fast reaetor (BREST-300) and sodium-cooled fast reactor (3800MWth large core) were compared. The sodium-cooled reactor surpasses BREST-300 on the MA transmutation rate. Meanwhile, it is found that BREST-300 is excellent from the viewpoint of loading much more MA in the core to attain larger MA transmutation amount. The effect of MA to coolant void reactivity is considered by the sensitivity analysis. It is found that the lead void reactivity has different sensible energy regions on MA nuclides from those for the sodium void reactivity.

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