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overpressurization in the circuit, which could open further leakage paths. Normally, steam generator tube failures have a high enough probability occurrence to be considered in the licensing procedure. It is noteworthy that a steam-generator failure was the cause of a LOCA and radioactive contamination in a Russian lead/bismuth-cooled nuclear submarine in 1982 [76]. Results for a transient test case study of a postulated steam generator tube rupture event leading to extensive cooling voiding were presented in Paper I. The core model and fuels are the same as the previous study of the unprotected loss-of-flow event. In the following analysis, it is assumed that that the coolant is swept upwards through the core, beginning at the lower cold-leg region, and the void front moves at the average coolant velocity through the core (2.5 m/s). Since the total height of the core plus plenum regions is 2.5 m, the passage occurs in 1 second. The transient calculation uses a reactivity history based on progressive axial voiding of the core. It is assumed that the void spreads axially and simultaneously in all subassemblies. The reactivity effect, as function of axial void level, is illustrated in Fig. 21a. Note that the reactivity effect is strongest for the CerCer fuel and most positive when the core has been voided up to slightly below the top of the active fuel region. The reactivity insertion rate is highest at core midlevel. As a coincidence, the maximum reactivity insertion due to coolant void corresponded to the initial subcritical reactivity of the CerCer core. (a) (b) Fig. 21 Transient results for a postulated steam generator tube rupture accident. (a) shows the reactivity effect ($) following progressive axial voiding of coolant beginning at the lower plenum.transient power. (b) Illustrates the transient power for the same accident. The resulting power history is presented in Fig. 21b. The steam bubble reaches the lower plenum at 1 second after the steam generator failure. Initially, the bubble passage produces a negative reactivity effect due to increased neutron leakage, as the lower plenum is voided first. The power will find its peak as the reactivity reaches its maximum. The reactivity at peak power, is –0.2$, -12.3$, and –6.8$, respectively for the CerCer, CerMet, and nitride cores, the corresponding peak power is 15.3, 1.3, and 2.1, times the initial power. The power rise in the nitride and CerMet fueled cores is quite modest. The CerCer core, on the other hand, suffers from a sharp power peak. Except from coolant void, axial fuel expansion is the only feedback effect that has some impact on the transient. At peak conditions, the contribution from axial expansion provided an extra reactivity margin, which was sufficient to 51
maintain the reactor in the subcritical state, thereby limiting the magnitude of the peak. Radial core expansion is too slow to be of any significance. It was found that the flux shape in the voided state was similar to the initial shape, and power peaking factors were even lower. The power rise is halted when the void has extended to the top of the core and begins to void the upper plenum region. Judging from Fig. 21b, voiding of the upper plenum plays a vital role in reversing the accident. It is assumed that the beam is shutdown after 2 seconds. Due to the strong positive reactivity effect, the CerCer system is subject to a sharp power peak, while the power rise in the nitride and CerMet fueled cores is quite modest, which simply confirms the importance of a having a low coolant void reactivity value in a lead/bismuth system, despite of its high boiling temperatures. Safety performance in super-prompt critical transients Even if the probability for prompt critical transients can be made very low, the safety analyses usually require them to be investigated. Considering the high reactivity potentials available by core compaction (see TABLE 13) it is always possible to postulate an accident that will proceed into a prompt critical state. The resolution for the large FBR’s was to demonstrate a low energetic potential (within the capabilities of the containment) of such accidents [66]. The large oxide cores relied on the presence of a strong Doppler effect to limit the energy yield (and pressure buildup) in a prompt critical reactivity transient. In such cores, the negative reactivity is provided by the Doppler broadening of resonances in primarily 238 U resulting in a relative increase in resonance capture over resonance fission. Since the Doppler effect is inherent in the fuel, the feedback mechanism is completely passive. The resulting reactivity effect is applied immediately since the Doppler broadening occurs simultaneously with the temperature change. As is well-known, the Doppler effect is small for fuels dominated by MA. Hence, the prompt negative effects, might be insufficient to stop an excursion before energetic disassembly occurs. Here we treat a classical problem of reactor excursions, induced by a rapid reactivity ramp in systems with different prompt negative feedbacks. It may be postulated that the accident is caused by a loss-of-coolant (e.g. as a result of a major break in the primary system) and that some fuel is melted and the core collapses by gravity forming a supercritical configuration (traditional Bethe-Tait scenario). Fig. 22 illustrates the essential characteristics of the transient in a MA-fueled ADS. The results are compared with calculations for a conventional sodium-cooled and MOX-fueled fast reactor (U0.8Pu0.2) presented in the same graph. Consider at first the ADS with initial keff=0.97. The transient is driven by a linear reactivity insertion, 100 $/s (0.167 Δk/s), a value which is frequently related to core compaction/meltdown phenomena. It is further assumed that this reactivity ramp continues unterminated. The value for the Doppler constant (T*dk/dT) as well as the effective delayed neutron fraction are chosen to simulate the conditions in the respective cores. The Doppler constant in the fast spectrum MA core (Pu0.4Am0.6) is taken as 2.0·10 -4 with a βeff=0.167 %. The Doppler constant for the oxide fast reactor is 8.1·10 -3 (see TABLE 11), which is about 40 times larger than the MA core. The corresponding ramp rate for the fast reactor is 50 $/s since the effective delayed neutron fraction (βeff=0.34 %) is about two times larger than the MA core. The Doppler effect contribution is assumed proportional to 1/T, as in the case of large oxide-fueled fast reactor, where T is the average fuel temperature in the core. As comparison the Doppler coefficient decreases somewhat faster with temperature (T -3/2 ) in the smaller, highly enriched FR’s. For the LWR’s it approaches T -1/2 . The fuel heat-up is treated for the simple case of adiabatic boundary conditions for any additional heating above steady-state levels. This is a reasonable approximation considering the overall speed of the transient. The steady-state operational power density is ~1.3 W/mm 3 , which is fairly typical for the FBR as well an ADS employing 52
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System and safety studies of accele
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Preface The research on safety of A
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coolant fraction allows for decay h
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To dispose the spent fuel as it is
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SAMMANFATTNING Resultaten från for
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Detaljerade analyser av neutronkine
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Modellering av de termo-kemiska ege
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LIST OF APPENDICES 1. Carl-Magnus P
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ABBREVIATIONS AND ACRONYMS EUROPEAN
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FCC Face Centered Cubic - type of a
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1 SYSTEM AND SAFETY STUDIES OF ADS
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pin. The full core void worth was c
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Figure 2.1 The 1180-loading of Yali
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Counts Counts 10 5 10 4 10 3 10 2 1
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Figure 2.4 Ion source interruption
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New 1 g 232 Th, 238 U (ultra pure)
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3.1.2 System descriptions The PDS-X
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4 NUCLEAR FUEL CYCLE ANALYSIS 4.1 E
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will be calculated using MCB, the M
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irradiation, the spent fuel assembl
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4.1.3 Results The calculations has
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4.2 ANALYSIS OF THE SWEDISH NUCLEAR
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Mass [tons] 300 250 200 150 100 50
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5 POTENTIAL OF REACTOR BASED TRANSM
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due to a partial insertion, the inf
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irradiation to maintain the core as
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configuration allowed the reactor t
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6.2 THERMOCHEMISTRY OF ADVANCED FUE
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Figure 7.1 Positions of Cr atoms in
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8 DEVELOPMENT OF CODES AND METHODS
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Let us assume the neutron flux is p
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waste repository notes which year t
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9 INTERNATIONAL INTERACTIONS, SEMIN
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Janne Wallenius Oct 2005 Participat
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[11] Sylvie Pillon and Janne Wallen
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Nuclear Instruments and Methods in
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376 fraction, beff, have been calcu
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378 Fig. 3. Fit of exponentials to
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380 Table 6 Experimental estimation
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382 dominating closer to criticalit
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12th International Conference on Em
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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
Page 199 and 200: “A grad student in procrastinatio
Page 201 and 202: specify the meaning of the term “
Page 203 and 204: that three fast breeder reactors (F
Page 205 and 206: emitting properties and its relativ
Page 207 and 208: TABLE 8 Effective dose coefficients
Page 209 and 210: Fig. 1 Radiotoxic inventory of the
Page 211 and 212: prior to disposal. The most commonl
Page 213 and 214: often used in repository safety ana
Page 215 and 216: separations process is also known a
Page 217 and 218: Spent Fuel U Pu PUREX FP + An Np Tc
Page 219 and 220: Fig. 6 Schematic picture of the pyr
Page 221 and 222: From the viewpoint of reducing the
Page 223 and 224: Chapter 3: Transmutation Strategies
Page 225 and 226: from reprocessing is sent into stor
Page 227 and 228: adiotoxicity in the spent fuel afte
Page 229 and 230: In order to overcome the safety iss
Page 231 and 232: 79 Se (T1/2=6.5·10 4 years) and 12
Page 233 and 234: Source multiplication in a subcriti
Page 235 and 236: given by the source multiplication
Page 237 and 238: Fig. 16 Delayed neutron spectra vs.
Page 239 and 240: offers low void worths. For tight l
Page 241 and 242: smaller than classical MOX-fuel sur
Page 243 and 244: Responsiveness of the ADS The kinet
Page 245 and 246: Fig. 19 compares the response in a
Page 247 and 248: design parameters. The reference co
Page 249: indefinitely. The burst limit is ap
Page 253 and 254: value (-$18.5), the power increases
Page 255 and 256: () β ( ) Λ () dck t k t = p t −
Page 257 and 258: (a) (b) (c) (d) Fig. 25 Source over
Page 259 and 260: Accelerator reliability In accelera
Page 261 and 262: Conclusions The thesis analysed and
Page 263 and 264: Appendix A: Reactor Kinetics Equati
Page 265 and 266: The definition of F(t) is: ρ () t
Page 267 and 268: Bibliography 1 R. G. Cochran and N.
Page 269 and 270: 38 M. Bunn, S. Fetter, J. P. Holdre
Page 271 and 272: 75 R. J. Puigh and M. L. Hamilton,
Page 273 and 274: 74 Paper I
Page 275 and 276: chemical reactions. An “inherentl
Page 277 and 278: temperature with irradiation time e
Page 279 and 280: and compressible pool volumes with
Page 281 and 282: comparison, the FFTF reactor, which
Page 283 and 284: CerMet and nitride. The reason is t
Page 285 and 286: failure, which leaves small safety
Page 287 and 288: XII.B. Transient results The result
Page 289 and 290: programme,” Nucl. Sci. Eng., Acce
Page 291 and 292: 277, American Nuclear Society, LaGr
Page 293 and 294: Neutronics of minor actinide burnin
Page 295 and 296: TABLE I: Fuel Doppler constant KD (
Page 297 and 298: nitric acid. Therefore, we may not
Page 299 and 300: (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
Page 559 and 560:
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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