JAEA-Conf 2011-002 - 日本原子力研究開発機構

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ENDF. Thus this cross section stored in ENDF has not been revised more than 36 years at least. The cross section of JEFF-3.1 was taken from ENDF/B-VI.8[7], and it evaluated based on as old measurement data as ENDF. Table 1 Neutron capture cross section of carbon at 0.0253eV JENDL-4.0 JENDL-3.3 ENDF/B-VII.0 JEFF-3.1 3.85mb 3.53mb 3.36mb 3.36mb 2.4 Calculation conditions and method The criticality calculations for the 12-30 fuel columns loaded core with all control rods withdrawn at room temperature were performed with the continuous energy Monte Carlo code MVP[8] to obtain the keff value for each core state. The core geometry was treated as much detail as possible, and a heterogeneous effect caused by a coated fuel particle was taken into account by using a Statistical Geometry (STG) model[9]. The history number was defined 8,000,000 for each calculation to reduce the standard deviation (1) to about 0.03%. 3. Results and discussion Figure 3 shows calculated keff curves for the several HTTR core 1.15 states, which were configured in JENDL-4.0 JENDL-3.3 the critical approach. JENDL-4.0 yields the keff values which are 1.10 ENDF/B-VII.0 JEFF-3.1 smaller than the other libraries for Excess reactivity the all core states. Specifically, the 1.05 for fully loaded core keff values JENDL-4.0 calculated with are 0.5-0.9%k 1.00 k = 1 eff smaller than JENDL-3.3, and are First criticality 1.0-1.5%k smaller than ENDFL/B-VII.0 or JEFF-3.1. The keff values calculated with 0.95 12 14 16 18 20 22 24 Number of loaded fuel column 26 28 30 ENDF/B-VII.0 and JEFF-3.1 are the same within 2 for the each core states. Fig. 3 keff curves for each core states k eff JAEA-Conf 2011-002 3.1 Loaded number of fuel columns achieving first criticality The loaded number of the fuel columns at the first criticality is determined by identifying the first column with which the keff value is more than 1.0. Thus, the keff curve with JENDL-4.0 indicates that the first criticality is achieved by loading 19 fuel columns, which agrees with the experimental result. Meanwhile, the keff curves with JENDL-3.3, ENDF/B-VII.0 and JEFF-3.1 indicate the first criticality is achieved by loading 18, 17, and 17
fuel columns, respectively. These libraries underestimate the loaded number of the fuel columns by one or two columns, which are caused by the overestimation of the keff value. The discrepancy of the loaded number of the fuel columns achieving the first criticality between the experiment and the calculations was resolved by replacing the nuclear data libraries with JENDL-4.0. 3.2 Excess reactivity of fully loaded core The excess reactivity of the fully loaded core ex, is defined by keff 1 ex . (1) keff Table 2 shows that JENDL-4.0 yields the excess reactivity of 12.0%k/k, which agrees well with the experimental result. Meanwhile, JENDL-3.3, ENDF/B-VII.0 and JEFF-3.1 give the excess reactivity of 12.4, 12.8 and 12.7%k/k, and they overestimates by 0.4-0.8%k/k. The problem of overestimating the excess reactivity at the fully loaded core was also resolved by replacing the nuclear data libraries with JENDL-4.0. Table 2 Comparisons between experimental[10] and calculation results Number of fuel columns achieving the first criticality Excess reactivity of the fully loaded core (%k/k) JAEA-Conf 2011-002 Experiment JENDL-4.0 JENDL-3.3 ENDF/B-VII.0 JEFF-3.1 19 19 18 17 17 12.0 12.0 12.4 12.8 12.7 4. Conclusions The criticality calculations of 12-30 fuel columns loaded core, which were constructed in the critical approach, were performed with JENDL-4.0, JENDL-3.3, ENDF/B-VII.0, and JEFF-3.1. As a result, JENDL-4.0 yields keff values which are 0.5-0.9%k smaller than JENDL-3.3, and 1.0-1.5%k smaller than ENDFL/B-VII.0 or JEFF-3.1. The keff values calculated with ENDF/B-VII.0 and JEFF-3.1 are the same within 2 for the all core states. The discrepancies between the experimental and the calculation results on the following two issues were resolved as expected by replacing the nuclear data libraries with JENDL-4.0. This fact shows that the applicability of JENDL-4.0 to the HTTR criticality calculations is more excellent than the other nuclear data libraries. (a) The loaded number of fuel columns achieving the first criticality. (b) Excess reactivity of the fully loaded fuel columns core. JENDL-4.0 promises to improve the accuracy of the HTGR criticality calculations, which allows the design of the commercial HTGR with lower cost and higher performance.
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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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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 (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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