JAEA-Conf 2011-002 - 日本原子力研究開発機構

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JAEA-Conf 2011-002 Table 1 Comparison of Three Major Libraries Library ENDF/B-VII.0 JEFF-3.1.1 JENDL-4.0 Developed by US EU Japan Released Year 2006 2009 2010 No. of Nuclides 393 381 406 No. of Nuclides with Gamma-ray Data 206 139 354 No. of Nuclides with Neutron DDX 171 83 318 No. of Nuclides with Covariances 26 37 95 Main Code(s) Evaluation GNASH EMPIRE TALYS CCONE POD Purity* 60% 20% 96% *: “Purity” is defined as the ratio of the number of nuclides originally evaluated (not adopted from other file) to total number. Table 2 The Main Concept* of Next Generation Light Water Reactor in Japan Item Feature Electric Power Output 1.7-1.8 GW (1.0-1.4 GW optionally) Fuel Over 5% Enriched U Plant Life Time 80 year Others Shorter Construction Period (Balanced with Economy) Higher Rate of Operation Top Level Passive and Active Safety Equipments Earthquake-proof Construction *Japan Atomic Energy Commission, for example, (in Japanese); (http://www.aec.go.jp/jicst/NC/iinkai/teirei/siryo2010/siryo43/siryo3-2.pdf) The most urgent item is “covariance business” and this must be solved at least its direction of preparation in near future. The covariances in JENDL-4.0 have been mainly evaluated with the CCONE+KALMAN Method. The covariance data for 110 nuclides of ENDF/B-VII.1 will be provided for the Advanced Fuel Cycle Initiative (AFCI) project (BNL, LANL). The motto for covariance is “Might not be perfect but must be sensible”. The JEFF covariance data is being prepared by using “Total Monte-Carlo Method”. It must be stressed that covariance data and nuclear data qualities are the different items. 2.2 Medical, Space and Accelerator Related Applications For medical applications (radiation therapy, medical RI production, accelerator-driven BNCT, etc.) and space engineering (error of semi-conductor, dose estimation from cosmic-ray, etc.), nuclear data for particle transport calculation, energy
JAEA-Conf 2011-002 deposition, RI production rate and so on are necessary. Those are the neutron and charged particle spectra and activation cross sections below 250 MeV. JENDL Intermediate Files are also useful for accelerator related applications such as accelerator driven system (ADS) for nuclear waste transmutation, International Fusion Material Irradiation Facility (IFMIF). The neutron production cross section and neutron spectrum are important. 2.3 Safety Researches The safety researches can be categorized into “frontend” (Nuclear Fuel Supply), “reactor” (Core Characteristics and Operation), “down-stream” of nuclear fuel cycle (Spent Fuel Treatment) and “backend” (Nuclear Waste Management). At the frontend stage, criticality safety for handling of processing, storage and transport of fuel should be considered. If fail to control, critical accident, ex. the JCO Accident (1999.9.30), is happened. For the reactor safety, the operation by grasping power distribution, control rod worth, reactivity coefficients, etc. is important (ex. Chernobyl (1986.4.26), and TMI Accident (1979.3.28)). For this purpose, the prediction of burn-up characteristics and evaluation of delayed neutron effects should also be considered. At the down-stream stage of nuclear fuel cycle, the criticality safety for storage, transport and reprocessing, shielding and decay heat is important. At the backend stage, the waste management such as the long-term radio-toxicity (HLW) and clearance level (LLW) should be considered. The “Criticality Safety” is one of the key items for safety researches and well established. Though recent experiments are getting more expensive relatively than before, some databases have been prepared, ex. the OECE/NEA International Criticality Safety Benchmark Evaluation Project (ICSBEP), the Criticality Safety Handbook (JAEA-Data/Code 2009-010) with the critical limits for minimum mass, size, concentration, etc. at keff = 0.98. For the criticality safety research, nuclear data such as fission and capture cross sections, fission product yields (FPY) are the most important. Safety researches at the down-stream and backend stages are getting more important from my point of view, since the replacement period of current nuclear power plants is coming near future and competition of international sales of power reactors by developed countries is harder so as to reduce CO2 emission. The critical safety is still important for spent fuel storage and cask design; keeping “Subcritical when immersed in water” and “Subcritical when piled up in numbers”. For this purpose, neutron absorbers, B-SUS, B-Al, B-Resin, Cd-Alloy etc. are considered to keep keff < 0.95 even for unirradiated fuel with initial 235 U enrichment. The “Burn-up Credit” is also considered to estimate the reactivity loss and the nuclide concentrations. The Post Irradiation Experiments (PIE) is one of the good benchmark tests for analyzing accuracy of calculations and data (FPY, capture cross section and decay data). Important nuclides are, for example, 12FPs ( 95 Mo, 99 Tc, 103 Rh, 133 Cs, 143,145 Nd, 147,149,150,152 Sm, 153 Eu, 155 Gd) by JAERI-Tech 2001-055, 15FPs ( 95 Mo, 99 Tc, 101 Ru, 103 Rh, 109 Ag, 133 Cs, 143,145 Nd, 147,149,150,151,152 Sm, 153 Eu, 155 Gd) by OECD BUC WG and 13FPs ( 99 Tc, 103 Rh, 131 Xe, 133 Cs, 143,145 Nd, 147 Pm, 147,149,151,152 Sm, 153 Eu, 155 Gd) by SAND87-0151 for casks. The clearance (= radioactive nuclide productions) level estimation is needed for dispose waste, especially at the reactor replacement period. For this purpose, radioactive isotope (RI) production should be estimated as accurate as possible and activation cross sections and decay data are necessary. The nuclide list included in the IAEA Safety Guideline for the clearance level estimation is shown in Table 3 as an example. The similar list is also enacted in Japan and is given in Table 3. Other long-lived RIs, for example 182 Hf
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JAEA-Conf 2011-002 Cowley et al. [1
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4. Conclusions We have compared nuc
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of MeV region with highest cosmic-r
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JAEA-Conf 2011-002 events can be id
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JAEA-Conf 2011-002 models and its p
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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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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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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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