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3 ADS SOURCE EFFICIENCY STUDIES 3.1 INTRODUCTION Efficient incineration of americium, curium and even higher acitindes require fast spectrum cores to yield a reduction of the overall radiotoxicity, avoiding production of even heavier nuclides. However loading a fast reactor with large fractions of minor actinide fuels is lead to cumbersome safety issues. As an example, the Doppler effect vanishes with the removal of 238 U, and the situation is further worsened by introduction of americium. Further the effective delayed neutron fraction is significantly reduced. Subcritical core are commonly discussed as a means of overcoming the safety issues. Such a core would be operated with a large enough margin to criticallity to allow for sudden accidental reactivity insertions. External spallation neutron sources are envisioned to supply the extra neutrons needed to sustain a chain reaction. Typically, 1 GeV neutrons impinging on a liquid metal target are foreseen. At impact, the protons produce neutrons via spallation reactions. These multiply in the target through further spallation reactions and through (n, xn)-reactions. The efficiency of the spallation source is an important parameter for the transmutation system as a whole. The multiplication production of neutrons in the target and the subsequent multiplication of these neutrons in target and fuel region depends on a set of parameters such as target size and geometry, target material, proton beam energy et cetera. In our work designing subcritical system, the impact of the core coolant and fuel on source efficiency have been recurrent questions. To study the impact on source efficiency of changes in fuel and coolant, we have modeled the two core designs suggested by the PDS-XADS project. Helium and leadbismuth eutectic (LBE) have been tested as coolants. MOX and a transuranium fuel have been tested as fuels. 3.1.1 Source efficiency Seltborg suggested proton source efficiency to be defined as [13]: ⎛ ⎞ ⎜ 1 ⎟ < Fφs > ψ * = − 1 ⋅ . (1) ⎜ ⎟ ⎝ keff ⎠ < S p > For a given keff, is approximately proportional to the total power produced in the core, and thus to the beam power amplification. The neutron source efficiency on the other hand relates the multiplication of “source neutrons” to fission multiplication. However, the definition of what is considered as source neutrons may vary inbetween different systems. This is why we choose to discuss effects on the proton source efficiency, which is uniqely defined. 32
3.1.2 System descriptions The PDS-XADS-project, funded by the European fifth framework programme, presented two concepts of sub-critical cores [14][15]. They were both 80 MWth experimental core designs. On of them was cooled by lead-bismuth eutectic and the other by helium. A 600 MeV proton beam impinging on an LBE-target was assumed. Our modelling was made with MCNPX 2.5e [16]. Two different fuel options have been tested. MOX fuel has been assumed as the reference case. In a more innovative setup, a transuranium (TRU) nitride fuel, with a Pu/Am/Cm composition of 40/50/10% respectively is assumed. This fuel has been chosen as an eligible candidate for fuel in a full-scale transmutation reactor, as the Pu/Am ratio correspond to the equilibrium nuclide vector that would eventually be reached in a transmutation fuel cycle with multi-recycling [17]. Together with two options for coolant, this gave us four cases, for which the proton source efficiency have been studied. 3.1.3 Results The proton source efficiency for the four different coolant-fuel combinations is displayed in Table 3.1 There are significant differences in the proton source efficiency in the four cases, both with respect to the choice of fuel and to the choice of coolant. Changing coolant from helium to LBE lowers ψ* by 10%. Replacing MOX fuel for TRU fuel further degrades the proton source efficiency by 10%. Table 3.1: ψ*, for the four cores studied and the relative difference to the He-MOX case He-MOX LBE-MOX He-TRU LBE-TRU ψ* 19.4 16.8 16.8 15.2 -13% -13% -22% In a previous study by Pelloni [18], the neutron source efficiency ϕ* was calculated for the He- and LBE-cooled XADS core designs respectively by deterministic methods, assuming a spallation neutron source modelled with Monte Carlo methods [19]. The absolute values for ψ* found in the present study may be recalculated to neutron source efficiency by dividing the numbers by Z, representing the number of source neutrons per source proton. Of course Z is dependent of the source neutron definition. In this case as well as in the reference [18], the cut off neutron source definition is applied. A neutron is defined as a source neutron as its energy falls below 20 MeV, regardless of its geometrical position. It shows our values are higher than those of reference [18]. The qualitative effects of changing the coolant, observed in the two studies, are similar though. An analysis of this discrepancy showed that the difference in geometries of the spallation targets in the two studies is important for the results. Increasing the radius, from the 20 cm of the present study to the 31.5 cm used by Pelloni, reproduces the previous results with good agreement. The comparative study is summarised in Table 3.2. 33
Page 1 and 2: System and safety studies of accele
Page 3 and 4: Preface The research on safety of A
Page 5 and 6: coolant fraction allows for decay h
Page 7 and 8: To dispose the spent fuel as it is
Page 9 and 10: SAMMANFATTNING Resultaten från for
Page 11 and 12: Detaljerade analyser av neutronkine
Page 13 and 14: Modellering av de termo-kemiska ege
Page 15 and 16: LIST OF APPENDICES 1. Carl-Magnus P
Page 17 and 18: ABBREVIATIONS AND ACRONYMS EUROPEAN
Page 19 and 20: FCC Face Centered Cubic - type of a
Page 21 and 22: 1 SYSTEM AND SAFETY STUDIES OF ADS
Page 23 and 24: pin. The full core void worth was c
Page 25 and 26: Figure 2.1 The 1180-loading of Yali
Page 27 and 28: Counts Counts 10 5 10 4 10 3 10 2 1
Page 29 and 30: Figure 2.4 Ion source interruption
Page 31: New 1 g 232 Th, 238 U (ultra pure)
Page 35 and 36: 4 NUCLEAR FUEL CYCLE ANALYSIS 4.1 E
Page 37 and 38: will be calculated using MCB, the M
Page 39 and 40: irradiation, the spent fuel assembl
Page 41 and 42: 4.1.3 Results The calculations has
Page 43 and 44: 4.2 ANALYSIS OF THE SWEDISH NUCLEAR
Page 45 and 46: Mass [tons] 300 250 200 150 100 50
Page 47 and 48: 5 POTENTIAL OF REACTOR BASED TRANSM
Page 49 and 50: due to a partial insertion, the inf
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Page 53 and 54: configuration allowed the reactor t
Page 55 and 56: 6.2 THERMOCHEMISTRY OF ADVANCED FUE
Page 57 and 58: Figure 7.1 Positions of Cr atoms in
Page 59 and 60: 8 DEVELOPMENT OF CODES AND METHODS
Page 61 and 62: Let us assume the neutron flux is p
Page 63 and 64: waste repository notes which year t
Page 65 and 66: 9 INTERNATIONAL INTERACTIONS, SEMIN
Page 67 and 68: Janne Wallenius Oct 2005 Participat
Page 69 and 70: [11] Sylvie Pillon and Janne Wallen
Page 71 and 72: Nuclear Instruments and Methods in
Page 73 and 74: 376 fraction, beff, have been calcu
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Page 77 and 78: 380 Table 6 Experimental estimation
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2.2 Yalina Booster Yalina and Yalin
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Yalina and Yalina Booster Λ 1 ⎛
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Yalina and Yalina Booster Table III
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Yalina and Yalina Booster 7 REFEREN
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1720 A. Talamo, W. Gudowski / Annal
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Journal of NUCLEAR SCIENCE and TECH
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1042 A. TALAMO and W. GUDOWSKI Tabl
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1046 A. TALAMO and W. GUDOWSKI 237
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1048 A. TALAMO and W. GUDOWSKI 1) 2
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1052 A. TALAMO and W. GUDOWSKI IX.
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Abstract Annals of Nuclear Energy 3
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Table 1 Technical data of the power
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Abstract Technical note Managing th
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86 A. Talamo / Annals of Nuclear En
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Westlén and Wallenius HEAT REMOVAL
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Accelerator-driven Systems: Safety
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Abstract The accelerator-driven sys
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Acknowledgements I owe thanks to ma
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“A grad student in procrastinatio
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specify the meaning of the term “
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that three fast breeder reactors (F
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emitting properties and its relativ
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TABLE 8 Effective dose coefficients
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Fig. 1 Radiotoxic inventory of the
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prior to disposal. The most commonl
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often used in repository safety ana
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separations process is also known a
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Spent Fuel U Pu PUREX FP + An Np Tc
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Fig. 6 Schematic picture of the pyr
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From the viewpoint of reducing the
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Chapter 3: Transmutation Strategies
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from reprocessing is sent into stor
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adiotoxicity in the spent fuel afte
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In order to overcome the safety iss
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79 Se (T1/2=6.5·10 4 years) and 12
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Source multiplication in a subcriti
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given by the source multiplication
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Fig. 16 Delayed neutron spectra vs.
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offers low void worths. For tight l
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smaller than classical MOX-fuel sur
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Responsiveness of the ADS The kinet
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Fig. 19 compares the response in a
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design parameters. The reference co
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indefinitely. The burst limit is ap
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maintain the reactor in the subcrit
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value (-$18.5), the power increases
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() β ( ) Λ () dck t k t = p t −
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(a) (b) (c) (d) Fig. 25 Source over
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Accelerator reliability In accelera
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Conclusions The thesis analysed and
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Appendix A: Reactor Kinetics Equati
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The definition of F(t) is: ρ () t
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Bibliography 1 R. G. Cochran and N.
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38 M. Bunn, S. Fetter, J. P. Holdre
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75 R. J. Puigh and M. L. Hamilton,
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74 Paper I
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chemical reactions. An “inherentl
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temperature with irradiation time e
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and compressible pool volumes with
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comparison, the FFTF reactor, which
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CerMet and nitride. The reason is t
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failure, which leaves small safety
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XII.B. Transient results The result
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programme,” Nucl. Sci. Eng., Acce
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277, American Nuclear Society, LaGr
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Neutronics of minor actinide burnin
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TABLE I: Fuel Doppler constant KD (
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nitric acid. Therefore, we may not
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(e.g. SS316) on the other hand have
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FIG. 4: Example of core map with 24
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TABLE X: BOL temperature coefficien
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[18] M.A. Smith, E.E. Morris, and R
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76 Paper III
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II. REACTOR KINETICS EQUATIONS In t
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The model pertains to an accelerato
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V.A. Variations in Source Strength
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TABLE VI Calculations of the core c
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pressure, starting at t=1 sec. The
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14. R. E. ALCOUFFE, F. W. BRINKLEY,
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Safety Analysis of Na and Pb-Bi Coo
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Neutronics analysis Table 1. Design
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stoichiometric oxide fuel. The fuel
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4000 3500 3000 2500 2000 Peak fuel
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Figure 7. Sodium (left) and lead/bi
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78 Paper V
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1690 M. Eriksson, J.E. Cahalan / An
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RELIABILITY ASSESSMENT OF THE LANSC
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calculations are based on operation
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Figure 3 is interesting in that sen
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The database is for practical reaso
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Failure analysis Analysis of failur
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Source efficiency and high-energy n
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Abstract Transmutation of plutonium
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The following papers are not includ
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ERALIB1 ERAnos LIBrary ERANOS Europ
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TRU TRansUranic elements TRUEX TRan
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Contents 1 Introduction 1 2 Transmu
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Chapter 1 Introduction The growing
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objectives of the present thesis ha
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n+ 238U → 239U + γ → 239Np + e
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natural uranium is typically needed
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ate step when an even-N nuclide is
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the atomic mass of a specific eleme
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χ (Ε) 0.4 0.3 0.2 0.1 0.0 0 1 2 3
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UOX- LWR 1. Once-through 2. Pu-burn
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2.2.4 Radiotoxicity reduction The t
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onstrated that recovery yields of 9
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advantage, thanks to the radiation
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As was mentioned earlier, a fast ne
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3.3 Coolant options Since the neutr
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3.4.1 Heavy liquid metal target Two
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In a single spallation reaction, ab
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for this purpose, since it allows f
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technical reasons. One is that larg
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concepts, developed by different EU
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Evaluated Nuclear Data Libraries (e
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SATURNE experiments that the Bertin
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4. Number of neutrons appearing as
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The eigenfunctions of hermitian ope
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( , , ) + ∫∫∫ Φ ( , , ) =
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ks FˆΦ FˆΦ = = AˆΦFˆΦ + S ,
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S Fˆ Φ AˆΦ − FˆΦ AˆΦ 1−
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the second-step calculation. This i
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⎛1−kˆ eff ⎞ FΦ ψ * = ⎜
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Moreover, replacing ϕ * ⋅ Z by
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6.3.3 Varying the target radius Bot
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6.3.4 Required proton beam current
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multiplication process from the ext
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Three sub-critical configurations,
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have energy close to 14.1 MeV. Howe
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and can from this point of view be
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1.E+00 Neutron flux (normalised) 10
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7.2.3 Distribution of the spallatio
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equal or slightly larger than 1, wh
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eactivity indicator can be obtained
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Chapter 8 The Yalina experiments In
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8.2 Dynamic experiments performed a
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fundamental decay rate, determined
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ρ n1 − n0 = β n eff 1 . (64) Th
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for this is the fluctuations of the
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above the bottom of the active part
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perimental results. Experience from
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Core power [kW] 300 200 100 0 0 0.9
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29). As expected, it appears that t
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ficients together with the specific
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personnel is 12 µSv/h. However, pr
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High-energy contribution As was sho
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Fig. 46. Vertical cross-sectional v
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there is a fraction of leakage neut
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ψ∗/Ep 40 30 20 10 0 Energy gain
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ψ∗ 40 30 20 10 ZrN matrix YN mat
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Table 20. Proton source efficiency,
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Fig. 56. Horizontal cross-sectional
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Chapter 11 Abstracts of papers 11.1
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11.5 Paper V Different reactivity d
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source response have been investiga
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Table A2. Production facts for oper
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Eq. (B.5) expresses the energy prod
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Table C1 (cont.) Neutrons Photons P
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13 R. Soule, “Neutronic Studies i
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45 H. A. Abderrahim, P. Kuoschus, M
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78 F. Atchison, Proc. of a Speciali
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108 I. Koprivnikar, E. Schachinger,
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The basic idea of ADS is to supply
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equilibrium (Prael, 1998) Models ar
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two dips in the neutron fluxes caus
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4.3 Calculations of ϕ* for the MUS
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Instead of plotting the ratio ϕ* /
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have already been multiplied in the
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APPENDIX A Table 3 Material Composi
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Definition and Application of Proto
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is the best way to define the neutr
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TABLE I Relative Fraction of Actini
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compared to 0.8% for r=50 cm). When
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Proton Source Efficiency ψ * 40 30
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16. L. S. WATERS, “MCNPX TM User
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314 spallation reactions. The produ
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316 current of3.0 mA (1.9 10 13 pro
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318 between 0.1 and 10 MeV, while m
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320 would be very thin, only the un
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322 Relative effective dose 1.E+01
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324 12 mSv h 1 , which is about 15
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326 originates from neutrons with e
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328 [17] Internal SAD document at t
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affecting ψ* - the macroscopic fis
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nected to an increase in ψ*. Preli
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fission and capture cross-sections
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Relative fission probability, σf/(
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F φ >=< Fφ > + < Fφ > + < Fφ >
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tions. Three parameters - the avera
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Nuclear Instruments and Methods in
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fraction, beff, have been calculate
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satisfactory statistical agreement,
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and combining the values of a obtai
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dominating closer to criticality, t
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Abstract Elsevier Science 1 Journal
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adjusted to give a keff close to 0.
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Elsevier Science 5 f Σ Σ c Core F
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Table 6 shows how the differences i
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Transmutation of nuclear waste in g
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ii Abstract The actinides in spent
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iv VI D. Westlén, J. Cetnar and W.
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vi MONJU Japanese experimental fast
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viii CONTENTS 6.5 Summary of the se
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2 CHAPTER 1. BURDENING FUTURE GENER
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10 CHAPTER 2. NUCLEAR REACTIONS Fig
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12 CHAPTER 2. NUCLEAR REACTIONS Fig
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14 CHAPTER 3. FAST REACTORS Reactor
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16 CHAPTER 3. FAST REACTORS reactor
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18 CHAPTER 3. FAST REACTORS Generat
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20 CHAPTER 3. FAST REACTORS Two ent
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22 CHAPTER 3. FAST REACTORS a liqui
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24 CHAPTER 4. PARTITIONING AND TRAN
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34 CHAPTER 5. FAST SUB-CRITICAL COR
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48 CHAPTER 6. DESIGNING A GAS-COOLE
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Bibliography [1] T. Ginsburg. Die f
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[29] G. Melese-d’Hospital. Perfor
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[58] L.A. Lys et al. Gas turbine po
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Acknowledgements Even though my wor
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PROOF COPY [LY8910B] 088545PRB OLSS
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£ ¤¥¦ §¨¨© ¡¢ ¨§¨ ¢¦
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List of Papers I. J. Wallenius, P.
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Chapter 1 Introduction About 16% of
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Figure 1.1: The length and time sca
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2.2. Primary Damage: Vacancies and
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2.2. Primary Damage: Vacancies and
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2.3. Primary Damage Evolution 9 the
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2.3. Primary Damage Evolution 11 20
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2.3. Primary Damage Evolution 13 Im
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3.1. Ab Initio 15 The problem is ad
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3.2. Interatomic Potentials 17 wher
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3.2. Interatomic Potentials 19 3.2.
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3.3. Molecular Dynamics 21 be of th
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3.4. The Monte Carlo Method 23 sele
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4.1. Potential Construction 25 func
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4.2. Verification of Potentials 27
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4.2. Verification of Potentials 29
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4.3. Cascades Using Molecular Dynam
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Chapter 5 Summary and Outlook Befor
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Bibliography [1] http://www.iaea.or
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Bibliography 37 [24] P. Olsson, J.
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Modeling of chromium precipitation
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MODELING OF CHROMIUM PRECIPITATION
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etween EAM and FS type of potential
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leaving other properties untouched.
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1. Introduction Displacement cascad
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(BCA) study [44], to correlate with
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correspondence with the maximum num
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however, that only above 20 keV can
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As explained in what follows, Figs.
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4.2 Clustered fraction While the st
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metals, for example, it has been sh
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Acknowledgements This work required
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[52] W.J. Phytian, A.J.E. Foreman,
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Table 2 - Summary of recoil energie
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Figure captions Figure 1 - Fe-Fe pa
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Number of FP at peak time cascade d
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Number of sub-cascades 4 3 2 1 0 Fi
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SIA clustered fraction Vacancy clus
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SIA clustered fraction Vacancy clus
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* Corresponding author, phone #: +3
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time and the different mechanisms o
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[35]. The use of a large distance c
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SIA clusters have been counted both
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and depleted in SIAs. Isolated vaca
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Figure 3 shows the number of recomb
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potentials predict between 30% and
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To summarize, we expect to see the
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properties. But while in the latter
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[24] H.-E. Schaefer, K. Maier, M. W
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fraction 0.5 0.4 0.3 0.2 0.1 Contri
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Number of defects 20 15 10 5 SIA cl
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Abstract This study is a first step
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Abbreviations ADS Accelerator-drive
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9 ADS and MOX Transmutation Strateg
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Important technological assumptions
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2 Generating nuclear power Fission
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coolants are low neutron capture cr
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steam-generators and other equipmen
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3.3 Conversion At the conversion fa
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3.5.2 Uranium oxide (UOX) The urani
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Table 1. Main elements present in s
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Fig. 4. Fission product yield as fu
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corresponding radioactive decay hal
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4 Swedish Spent Fuel Storage and Di
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Fig. 6. Swedish nuclear waste manag
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handling spent fuel at CLAB. It als
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Secondly it performs an operation o
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6 Nuclear Fuel Cycles studied in th
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7 Advanced Nuclear Fuel Cycles sele
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8 Characterization of the ADS Conce
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9 ADS and MOX Transmutation Strateg
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10.2 Scenario II - LWR + ADS Stabil
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ecycling in Scenario III resulted i
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12 Results - Phase-Out Scenarios 12
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the time of shut-down only 7 tons o
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LWR-park which is doomed to be phas
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22. Working Party on the Physics of
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nuclear power reactor under normal
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fission in LWRs and other reactors
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238 U microscopic cross sections [b
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Appendix B - Transmutation of Nucle
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In a fast neutron spectrum the neut
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understood it is important to exami
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Elsevier Science 1 From Once-throug
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Phase-out Reference Scenario The ph
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varied from 1800 to 3300 [MWt] depe
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plutonium in the system 4 though, w
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Elsevier Science 1 The Subcritical
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3. Subcritical assembly The sub-cri
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6. Conclusion The SAD project is in
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2 2. Key parameters of SAD and supp
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4 Fig. 5 Distributions of the 210 P
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O FFICE PARLEMENTAIRE D ’ ÉVALUA
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Nouvelles Les recherches technologi
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OPECST,Public Hearing, January 20,
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Some pre-declarations for this talk
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Pu can be pretty efficiently (but s
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I am leaving conclusions to the Hon
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