JAEA-Conf 2011-002 - 日本原子力研究開発機構

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JAEA-Conf 2011-002 the detection depth distribution with the response matrix using the Bayes theorem. Fig.7 Schematic view of parallel incidence experiment. 5. Conclusion We have been carrying out the series studies concerning the thermal/epithermal spectrum measurement especially for BNCT facilities. In the present study, we have developed a 3 He position sensitive proportional counter for the thermal/epithermal neutron spectrometry. The measurement system with a multi parameter MCA was designed and developed for the measurement. A thermal neutron field was designed by MCNP5 calculation and constructed for the measurement of the detection position of the 3 He counter. From the measurement, the detection position distribution could be obtained as a two dimensional spectrum. But the detection depth distribution needed to have a strict one-to-one correspondence with the real detection position. So a cadmium collimator with an open window was put on and it was confirmed that the position information could correctly be obtained. From the measurement, the detector was confirmed to work appropriately in Z=6~44cm out of the detector depth signals. It means in other words that the measurable energy range was thus limited to 0.005eV~1keV. In the next phase, we will carry out the real detection depth measurement with incident neutrons being parallel to the detector axis. For that purpose, several neutron sources are taken into consideration to be utilized for confirming one-to-one correspondence between the point of the direction depth distribution and the detection position in the detector, together with the response of the detector for the incident neutron energy. References [1] X-5 Monte Carlo Team: MCNP-A General Monte Carlo N-Particle Transport Code, 2003 X-5 Monte Carlo Team: MCNP—A general Monte Carlo N-particle transport code, version 5, Los Alamos National Laboratory, Los Alamos, New Mexico, USA, April 2003. [2] I.Murata et al. “Low-energy neutron spectrometer using position sensitive proportional counter ―Feasibility study based on numerical analysis” Nuclear Instruments and methods in Physics Research A 589 (2008) 445-454 [3] S.Iwasaki: “A new approach for unfolding PHA problems based only the Bayes’ Theorem”, Proc. 9 th International Symposium on Rector Dosimetry, Prague, Chezk Republic, pp. 245-252 (1996).
JAEA-Conf 2011-002 18. Measurement of 151,153 Eu Neutron Capture Cross Sections using a pair of C6D6 Detectors Jae Hong LEE*, Jun-ichi HORI, Ken NAKAJIMA Research Reactor Institute, Kyoto University Kumatori-cho, Sennan-gun, Osaka 590-0494, Japan * E-mail: zelloseve@rri.kyoto-u.ac.jp In the present study, we have measured the neutron capture cross sections of 151 Eu and 153 Eu by the time-of-flight (TOF) method in the range of 0.03 eV to keV region using a 46 MeV electron linear accelerator at the Research Reactor Institute, Kyoto University (KURRI). We employed a pair of C6D6 liquid scintillators for the capture -ray measurement. The energy dependent neutron flux was derived with the standard cross sections of the 10 B(n,) reaction. A pulse-height weighting technique was applied to observed -ray spectra to determine the capture yields of 151, 153 Eu. The weighting function was obtained using the response function of a pair of C6D6 liquid scintillators that were calculated with a Monte-Carlo simulation code EGS5 (Electron Gamma Shower Version 5). The present results were compared with the previous experimental data and the evaluated values of JENDL-4.0. 1. Introduction Recently, a great interest has been taken in burn-up credit for criticality safety in the transportation, storage and treatment of spent nuclear fuel. Burn-up credit is a concept in criticality safety evaluation that takes into account for the reduction in reactivity of spent fuel due to the composition change during irradiation. Neutron capture cross section data of fission product (FP) play an important role in burn-up credit. According to the reference of [1], twelve FP isotopes ( 95 Mo, 99 Tc, 103 Rh, 133 Cs, 143, 145 Nd, 147, 149, 150, 152 Sm, 153 Eu, 155 Gd) are recommended to be considered in burn-up credit. The objective of this work is to measure neutron capture cross sections of 153 Eu and 151 Eu. 153 Eu is one of the most important FPs for burn-up credit application. 151 Eu has a large capture cross section and is contained in the sample of enriched 153 Eu that used in this work. Therefore, the experimental data of 151 Eu are also necessary to correct the 153 Eu capture yield including the effect of 151 Eu as an impurity in the sample. 2. Experiments The capture cross section measurements were carried out by the TOF method using the linac at the KURRI. A photo-neutron target of Ta was adopted as a pulsed neutron source for the neutron TOF measurement. The experimental arrangement with a pair of C6D6 scintillators is shown in Fig. 1. The distance between the sample and the neutron source was 12.1±0.02 m. Output signals from the scintillators were summed up and stored with the Yokogawa’s WE7562 multi channel analyzer as a two dimensional data of pulse height (PH) and TOF. The accelerator was operated in two different modes as shown in Table 1: one was for the measurement below 1 eV with a repetition rate of 50 Hz and another was for the measurement above about 1 eV with a repetition rate of 200 Hz. In the latter case, a Cd sheet of 0.5 mm in thickness was inserted into the TOF beam line to suppress the overlap components of low energy neutrons from the previous pulsed due to the high frequency. Pulse width was 100 ns for each mode. The characteristics of the samples are summarized in Table 2. The samples of 151 Eu and 153 Eu were packed in aluminum foils 20 mm in diameter and 0.08 mm in thickness. The enriched 10 B sample was used for the measurement of the incident neutron flux on the sample. The sample of 10 B was packed in an aluminum case (25 mm in diam., 0.4mm in thickness).
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JAEA-Conf 2011-002 Proceedings of t
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JAEA-Conf 2011-002 31. Preliminary
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6.2 Measurement of neutron-induced
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JAEA-Conf 2011-002 Table 1 Comparis
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JAEA-Conf 2011-002 (T1/2=8,900,000
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JAEA-Conf 2011-002 Though words too
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JENDL-3.3 showed better applicabili
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ENDF. Thus this cross section store
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Acknowledgement JAEA-Conf 2011-002
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In the MALIBU program, isotope conc
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some minor actinides. For major act
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JAEA-Conf 2011-002 code MVP-BURN an
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JAEA-Conf 2011-002 [3] Measurements
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3.1 Experimental Set up JAEA-Conf 2
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References JAEA-Conf 2011-002 [1] N
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JAEA-Conf 2011-002 was large. The a
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JAEA-Conf 2011-002 4. Results and D
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Acknowledgments Present study inclu
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Fig. 1 A conceptural picture of the
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4. Theoretical studies Theoretical
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JAEA-Conf 2011-002 Fig.9 Comparison
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Ion source JAEA-Conf 2011-002 LE
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JAEA-Conf 2011-002 3. Readiness
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JAEA-Conf 2011-002 Figure 1. An exa
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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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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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