JAEA-Conf 2011-002 - 日本原子力研究開発機構

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In the MALIBU program, isotope concentration of one sample was measured by two or three laboratories independently, then, the measured results were checked each other. Finally “recommendation value” for isotope concentration of each nuclide was evaluated for each sample. Sample Fuel Type Assembly U235 [wt%] Puf [wt%] Burnup [GWd/t] Histrical Void Frac. Labo. GGM1 MOX DU 5.5 -70 - SCKCEN SCKCEN PSICEA PSICEA Table 1 MALIBU sample matrix PWR GGU1 UOX GOESGEN 1515 1515 4.3 -70 - PSI, CEA JAEA-Conf 2011-002 GGU2 UOX 4.3 -50 - SCKCEN, SCKCEN, PSI GRM1 MOX GUNDREMMINGEN-C 99 99 DU 5.5 -80 - 20% SCKCEN, SCKCEN, PSI BWR GRM2 MOX DU 5.5 -75 -50% SCKCEN, SCKCEN, PSI GRM3 MOX DU 5.5 -50 -70% SCKCEN, SCKCEN, PSI *GRM1 is focused on the following discussion. A continuous energy Monte-Carlo code "MVP-BURN" is employed for isotope concentration prediction to reduce both of geometry approximation and effective cross section approximation. 104 fp chain model which includes 21 heavy metals and 104 fission products was applied. Neutron history was 0.5 million per each burnup step with Predictor-Corrector method. Fuel assembly geometry was simulated without any approximation. Irradiation history of the sample, which was evaluated in the In Core Fuel Management, ICFM, was provided in MALIBU program. It was used for the isotope concentration prediction calculation. Spectrum interference is large because MOX fuel assembly is adjacent to UOX fuel assemblies. Spectrum interference was taken into account for the Pu isotopes concentration prediction. Four neutron cross section libraries, JENDL-3.2, JENDL/AC-2008, ENDF/B-VII.0 and JEFF-3.1 9) , are used for the calculation. JNDC v2 10) was used as the FP yield.
“Calculated burnup” which is Realistic history of Power, Void frac. etc. Measured burnup evaluated in ICFM and “Measured burnup” which is evaluated by radio-chemical analysis can be Calculated burnup different naturally. This sample burnup difference must considered appropriately. be An Averaged history through sample irradiation life example of difference between the "Calculated burnup" and "Measured sample [GWd/t] burnup [GWd/t] burnup" is shown schematically in Fig. 1. Schematic explanation of irradiation history Fig.1. Our approach to take the burnup difference into account is followings. / JAEA-Conf 2011-002 The time dependent irradiation conditions from the beginning of sample life, BOL, to the end of sample life, EOL, which is given in the irradiation report supplied from ICFM, linear power rate, etc. are considered in burnup calculation directly. All irradiation conditions are averaged through BOL to EOL, and then burnup calculation is performed with the averaged irradiation condition. 11) The irradiation history effect coefficient is a ratio of each isotope concentration in fuel calculated with averaged burnup history calculation over the concentration evaluated with realistic burnup history calculation at the EOL given in the report. Once the irradiation history coefficient was obtained, the effect of the change of the irradiation condition through the sample irradiation can be reflected to the isotope concentration evaluated for the arbitrary sample burnup with averaged burnup history calculation. Therefore the isotope concentration at the burnup evaluated by radiochemical analysis, "Measured burnup", can be obtained with consideration of the change of the irradiation condition correctly. Finally the calculated isotope concentration can be compared with measured isotope concentration to obtain the C/E value. For highest burnup BWR MOX sample of MALIBU, C/E values for heavy metals are shown below. Calculation results with all libraries are agreed to measurements within about 5% for most major actinides. On the other hand, difference between calculation and measurements are over 10% for
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JAEA-Conf 2011-002 1. The recent ac
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JAEA-Conf 2011-002 4. Conceptual de
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The Institute's staff of about 280
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3. Nuclear Theory Laboratory JAEA-C
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JAEA-Conf 2011-002 References [1] P
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JAEA-Conf 2011-002 reaction 6 Li +
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Using the approximated distribution
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momentum of the constituents of P.
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dσ/dΩ (mb/sr) c.m. scattering ang
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JAEA-Conf 2011-002 [2] N. Austern,
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moderating fast neutrons emitted fr
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JAEA-Conf 2011-002 3. Results and d
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JAEA-Conf 2011-002 the detection de
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Fig.1. Experimental arrangement for
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in this work was confirmed. (2) 153
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Neutron Capture Cross Section of 15
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ENDF resonance parameters are based
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sample changer. The measured flux s
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REFERENCES JAEA-Conf 2011-002 [1] W
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JAEA-Conf 2011-002 Fig. 4: The rela
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Fig. 10: The PHITS2 calculation geo
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for such a low energy region should
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MCNPX calculation was performed bas
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JAEA-Conf 2011-002 We measured dou
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1mm-thick carbon target 22mm in dia
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JAEA-Conf 2011-002 Figure 2 shows e
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The double-differential (n,xp) and
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JAEA-Conf 2011-002 forward angles.
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Fig.3. Two-dimensional plot of PI v
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JAEA-Conf 2011-002 Fig. 7. Deuteron
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JAEA-Conf 2011-002 The new detector
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4. Off line data analysis Double di
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eports of data of carbon-ion incide
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Count [-] 4 10 3 10 2 10 10 1 Charg
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Acknowledgement We are grateful to
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Beam Line BM1 Copper Target BM2 5×
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φ = φbm · ρtof · ρmulti (2) E
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References [1] K. Ishibashi, H. Tak
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JAEA-Conf 2011-002 hint at the orig
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Neutron Dose Rate [Sv/hr/src neutro
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JAEA-Conf 2011-002 radiation protec
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Ztarget, Atarget are the numbers fo
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We described our formalism for calc
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2. Model
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JAEA-Conf 2011-002 reactions. Figur
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JAEA-Conf 2011-002 Fig. 3. The angu
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analysis of integral experiments [6
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A modified SRAC library was prepare
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JAEA-Conf 2011-002 approximately 0.
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JAEA-Conf 2011-002 constructing the
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Fig.9 Spectrum of gamma-ray followi
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JAEA-Conf 2011-002 in the same way
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JAEA-Conf 2011-002 [7]. For example
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JAEA-Conf 2011-002 For
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JAEA-Conf 2011-002 (5.0-9.0MeV), a
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components, the 0.56Wt% for hydroge
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Sensitivities of curium isotope con
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decreases of (n,g) reaction rates,
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5. Conclusion JAEA-Conf 2011-002 Wi
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2. Foundation of sensitivity analys
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JAEA-Conf 2011-002 where f g and f
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Fig.1 Sensitivity coefficient of th
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An outline of core configurations o
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0.010% 0.000% -0.010% -0.020% -0.03
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To clarify what nuclides and reacti
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Calculation was performed by the MV
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among these libraries, and the diff
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JAEA-Conf 2011-002 - 日本原子力研究開発機構

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