Fast neutron reactors: principal features and experience

Size: px
Start display at page:

Download "Fast neutron reactors: principal features and experience"

Transcription

1 Fast neutron reactors: principal features and experience Andrei Rineiski WORKSHOP ON FUSION FOR NEUTRONS AND SUB-CRITICAL NUCLEAR FISSION, FUNFI, Villa Monastero, Varenna, Italy, September 12 15, 2011 KARLSRUHE INSTITUTE OF TECHNOLOGY (KIT), INSTITUT für KERN UND ENERGETIK (IKET) KIT University of the State of Baden-Wuerttemberg and National Research Center of the Helmholtz Association

2 Outline 1. Thermal and fast reactors 2. Breeding 3. Uranium availability 4. Burning (transmutation) 5. Reactivity effects and kinetics parameters 6. Safety 7. Core optimization: ESFR 8. Fuels, structure materials 9. Experience, issues 10. Recent trends 11. Conclusions 2

3 3 1. Thermal and fast reactors: background U is the only available natural fissile material Fraction of fissile U235 ( 0.7% in natural U) can be increased: enrichment Moderator (water, graphite, etc.) can be placed around a fuel lump to slow-down fission neutrons from fast, MeV/keV, to thermal, below ev, energies Major cross-sections higher at low energies (except, e.g., threshold fission) => neutron leakage lower in systems with softer neutron spectra: e.g. 2-3% in a thermal system, 10-30% in a fast one. Neutron production-to-absorption ratio increases in natural and enriched U at low energies => lower enrichment is needed to achieve criticality in systems with moderator, i.e. in Thermal Reactors (TR) Fast Reactors (FR) contain no moderator, need higher enrichment; ratio of U238 neutron capture cross-section to total fission crosssection in FR is higher than that in TR=> higher Pu production

4 1. Thermal and fast reactors: background FR can be cooled by liquid metals (sodium, lead, lead-bismuth eutectic, LBE) or gas with U OXide fuel (UOX): 20-25% enriched (BN-350) with Mixed U/Pu OXide fuel (MOX): enrichment lower by a few % for similar cores as Pu239 is more efficient than U235 in fast spectrum with U-Pu-Zr metal alloy, U/Pu carbide and nitride: enrichment lower than for MOX due to faster spectrum, higher fuel density TR are mainly cooled by light/heavy water or gas (CO 2, He) LWR with UOX/MOX (moderated by water): 3-5% enriched U CANDU (moderated by heavy water) and RBMK (moderated by graphite, cooled by water) with UOX: no or lower enrichment By now, mainly TR are employed: fuels with lower enrichments more experience for water than for liquid metals in gas cooled systems, graphite is a sink for decay heat 4

5 1. Thermal and fast reactors: EPR (thermal) Cutaway of reactor pressure vessel of EPR (AREVA). EPR fuel SA (AREVA) 5

6 1. Thermal and Fast reactors: FR types Loop- vs. Pool design Vessel is smaller, less heavy: better for seismic, inspection, fabrication Piping is longer: higher probability for coolant leaks, pipe ruptures SFR (sodium): 3 loops (suggestion to use CO 2 in secondary loop=>2 loops) LFR (lead, LBE): intermediate loop excluded, steam generator in the pool GFR (He), SCWFR (Supercritical water): no secondary/tertiary loop MSR (liquid fuel in the primary loop): with/without graphite in the core Almost all GEN-IV (future) reactors are FR (sustainability) 6

7 1. Thermal and fast reactors: BN-800 SFR Pool Design, Based on BN-600, MOX fuel to be used Source: IPPE, Obninsk Sodium, instead of water for PWR, in primary and secondary loops 7 Low pressure in the vessel Higher coolant inlet/outlet temperatures => higher thermal efficiency than LWR

8 1. Thermal and fast reactors: electricity In 2008, ca. 440 reactors were used for electricity production worldwide (small/experimental reactors being not considered) LWR: More than 80% of the reactor fleet, (PWR and VVER) to BWR ratio: 3 to 1 Almost all others: Heavy-water and graphite moderated. In 2008, 2 FR in operation BN-600 (UOX), 600 MWe, Russia, since 1980 Phénix (MOX), 250 MWe, France (2 loops of 3), Monju (MOX), 280 MWe, Japan, since 1994, preparations for restart. 2 FR under construction in 2011 PFBR (500 MWe) in India, scheduled for 2012 BN-800 (880 MWe) in Russia, scheduled for

9 1. Thermal and fast reactors Medium and large FR for electricity production constructed in the past: Fermi 1, US (66 MWe), PFR, UK (270 MWe), Superphénix, France (1240 MWe), BN-350, Kazakhstan (135 MWe and desalination), SNR-300, Germany (300 MWe), 1985, not put into commercial operation due to non-technical reasons Large number of experimental FR in all major countries, from Clementine, US (25 kwth), in 1940s to CEFR, China (25 MWe), connected to grid in SFR only mentioned before. LBE-cooled FR employed by now only in Russian submarines: higher power in a compact design compared to LWR. FR often considered commercially unattractive for electricity production due to assumed higher fuel and construction costs, higher risks due to lower construction and operation experience. Why may we need FR? 9

10 2. Breeding Reactors with U breed Pu: U238-> ->Pu239->Pu240->Pu241-> Also Minor Actinides, MAs, are produced: U235->U236-> ->Np237, Pu241->Am241-> If Th is put in the reactor => U breeding (Th232-> ->U233-> ) Bred Pu239, Pu241, U233 are fissile materials and can be used as nuclear fuel (after adding to e.g. depleted U and/or Th), e.g., if the spent nuclear fuel is reprocessed. Even Pu isotopes (e.g. Pu240), many MAs: fission by fast neutrons only, their presence reduces quality of transuranic mixture, TRU, as concerns its employment as nuclear fuel, safety, etc. Fissile Generation to Destruction Rate Ratio: conversion ratio, CR If the Ratio>1: breeding ratio, BR. CR/BR depends on spectrum, fissile/fertile choice, fuel enrichment (not a fully free parameter in a critical system). Fertile blankets around FR cores: to increase BR 10

11 2. Breeding: SFR designed before 1990s LFR: higher coolant volume fraction, lower coolant velocity than in SFR Source: A. Waltar and A. Reynolds, Fast Breeder Reactors Fertile blankets surround the core to increase BR 11

12 2. Breeding CR in LWRs with UOX is slightly higher than 0.5, up to 0.7 (U-Pu cycle) or 0.8 (Th-U) cycle. Amount of U is limited => critical thermal reactors are not sustainable in long term (only a minor fraction of U238 can be used) BR in FR >1.2 (oxide), >1.4 (metal, due to faster neutron spectrum) in case of using fertile blankets around a fissile core FR allow to achieve a sustainable U-Pu cycle (all U238 can be used as fuel)! Th-U: lower breeding performance in FR (not considered in the following) A higher BR leads to a lower Doubling Time (DT), after which the amount of fuel available for nuclear energy production doubles due to breeding. DT from 10 to 15 years was targeted before 1990s. TRUs (Pu, MAs) produced in a fast system are of high quality: larger relative content of Pu239 as compared to thermal systems, smaller amount of MAs as the fission-to-absorption ratio for TRUs increases with increasing neutron energy High TRU quality facilitates their use: fuel fabrication, enrichment, safety 12

13 3. Uranium availability: prospects in 1970s FR were developed for breeding high quality re-usable Pu assuming shortage of U supply in the future. Uranium resources assumed to be small compared to demand in nuclear energy, even in the short term. Demand for electricity production in the future (after 2000) and for nuclear energy in particular was strongly overestimated, by several times. FR were considered as the main option to meet this projected demand. According to evaluations made in Germany, prices of electricity production (including fuel cycle costs), in FR with sodium, LWR, and FR with gas were comparable: SFR: 1.6 Pf/kWh LWR: 1.9 Pf/kWh GFR: 2.2 Pf/kWh 13

14 3. Uranium availability: price in The price was higher than in only for a few years shortly before

15 3. Uranium availability Uranium demand in 2007: ca. 70kt to increase to ca. 90 to 120 kt by 2030 (IAEA, OECD) In 2008: global known conventional resource ca kt at price less than 130 USD/kg. Sufficient time (several tens of years) is available now before introduction of breeders may be needed. Now, U price is a minor fraction of the electricity cost (unlike fuel for fossil power plants). German Fuel Tax introduced in 2011 before Fukushima accident, 145 EUR per g of U235/Pu239/, is a much higher burden for utilities Are FR important also for other issues than breeding? 15

16 4. Burning (transmutation) FR can breed good quality TRU from UOX or MOX. BR>1 in FR is mainly due to presence of fertile blankets surrounding the core=> FR without blankets can burn Pu/TRU LWR can also allow limited use of TRU-bearing (MOX) fuels. Central SA of a PWR reactor with 14x14 Fuel SA loaded with MOX 16

17 4. Burning (transmutation) TRU quality after irradiation in LWR deteriorates: the produced TRU may be difficult to re-use in LWR again (fabrication, safety reasons). Example: Pu with 55% of Pu239 -> Pu with 49% of Pu239 ->...with 46%... Produced TRU may provide a high load (radio-toxicity, decay heat) to repository, even if the mass is decreased. FR design can be changed (no blankets, higher TRU enrichment) to reach a low CR. TRU quality after irradiation can be better than in LWR. Incineration of TRU by multiple recycle in FR is an attractive option (if fuel reprocessing available)! Major FR employment scenarios assume fuel reprocessing and refabrication associated with partitioning and transmutation for spent fuel management; small actinide reprocessing losses being already demonstrated at laboratory or pilot plant (MAs) or industrial (Pu) scales Note, the maximum possible reduction of the repository volume is achieved by recycling of all actinides and a few long-lived fission products, provided that the losses during reprocessing are negligible 17

18 18 4. Burning (transmutation) In 1990s, priorities on FR utilization changed: breeding is a long term issue, but burning (transmutation) of nuclear waste (spent fuel) is an urgent problem, partly due to (near) zero fraction of FR in the current nuclear fleet. CAPRA project in early 1990s (EU cooperation lead by France, links to non-eu countries): high Pu enrichment, no fertile blankets, using of diluent (without actinides) pins to design a critical system. FR is a flexible concept that allows both (1) TRU production and reuse and (2) TRU destruction for spent fuel management The smaller the number of FR, the higher should be their burning capability and the worse is the TRU quality in their fuel. A very high TRU content and/or TRU of poor quality may not be allowed in FR due to safety reasons (reactivity feedbacks) Subcritical systems such as Accelerator Driven Systems, ADS, relieve limits in TRU quality and content => ADS development in 2000s, for utilizing U-free fuels with high MA content (also for phasing-out of nuclear strategy)

19 5. Reactivity effects and kinetics parameters Power variations in the initially critical reactor depend mainly on variations in reactivity, ρ. Also on the effective delayed neutron fraction (β-eff) and mean neutron lifetime (Λ). Reactivity is the balance of nuclear reaction rates (with positive contributions for neutron production, negative for absorption, leakage) normalized by the production rate. In a critical reactor at steady-state, the reactivity (1-1/k) is zero. If the reactivity is negative, reactor goes to stop: discontinues chain reaction (if no external neutron source) In reactivity computations, different neutrons are of different importance : with respect to their ability to produce new neutrons (after fission) in the future β-eff is the contribution of delayed neutrons to reactivity, while β is the fraction of delayed neutrons in the total amount of fission neutrons. Λ: a fraction of ms in TR, a fraction of μs (much smaller!) in FR 19

20 5. Reactivity effects and kinetics parameters In UOX-fuelled systems, thermal neutrons are more important (except in over-moderated systems). In MOX-fuelled systems with reactor Pu, fast neutrons are often more important. The lower TRU quality, the higher relative importance of fast neutrons: due to fast fissions Coolant void (or density reduction effect): negative in a thermal system with UOX, strongly positive in FR with TRU and small neutron leakage Doppler effect (neutron absorption higher -> reactivity lower for higher T) comes mainly from moderated neutrons (~ in kev region) interacting with nuclei of fertile materials The lower TRU quality, the smaller the Doppler effect. Negative coolant void/density and Doppler effects are major ones decreasing reactivity if power increases: reactivity feedbacks In critical reactor, the total reactivity feedback must be negative 20

21 5. Reactivity effects and kinetics parameters If reactivity is positive, the power increases exponentially. If reactivity exceeds β-eff, the power may increase dramatically within a very short period of time, with time constant (ρ β-eff)/λ. Therefore, the reactivity is often measured in $ (β-eff units). Lower TRU quality => lower β-eff due to 2 reasons TRUs, in particular MAs have lower β values than U (e.g. by a factor of 2) βeff/β ratios >1 (e.g. 1.15) for good fuels (e.g. LWR with U) and <1 (e.g. 0.85) for TRU-bearing fuels because delayed (softer) neutrons are more important than fission neutrons in reactors with good fuels Lower TRU quality => lower Λ: faster neutrons are more important, stronger transients possible Lower TRU quality => deterioration of major safety parameters 21

22 5. Reactivity effects and kinetics parameters Improving Doppler effect => changing material composition. e.g., by avoiding MAs in the fuel Improving void/density effect => also by increasing the leakage by modifying core geometry Other major reactivity effects in FR near nominal: Axial expansion of Clad/Fuel (depends on fuel and clad thermal expansion coefficients and whether or not fuel sticks to clad) Radial expansion of grid plate, subassembly, SA, bowing (strongly depends on how a SA is fixed and interact with others) Control Rod driveline expansion (effects can be enhanced by employing special devices that amplify this expansion) 22

23 5. Reactivity effects and kinetics parameters Transient Reactivity Feedback in Response to a Unprotected Loss Of Flow, ULOF, Accident for a Large Metallic-Fueled Fast Reactor Source: R. Wigeland, J. Cahalan, Accident Consequences Using Inherent Safety Principles, International Conference on Fast Reactors and Related Fuel Cycles(FR09), December 2009, Kyoto, Japan Radial Expansion Effect: smaller in in SPX, BN

24 24 6. Safety: CDA in FR In the following, FR with MOX/TRU are considered only Fundamental safety issue: FR may not be in the most reactive configuration at nominal conditions: unlike, e.g., LWR. Fuel compaction, e.g., due to fuel melting, may lead to over-criticality Fuel compaction should be prevented by design and accident management measures After coolant and/or steel removal and fuel compaction due to a hypothetical accident (pump stops, Control Rod withdrawn, SA blockage etc.), the reactivity may increase dramatically The power may increase until core disintegrates (CDA, core disruptive accident) Low Doppler, β-eff and Λ (if bad TRU fuel) may lead to a high mechanical energy release during CDA A small sub-criticality in an ADS may be insufficient to prevent CDA

25 6. SAFETY: CDA in FR TRANSITION PHASE : Progression of core-melt after end of initiation phase with already disrupted but unstable material configuration In case of (1) insufficient fuel release from the core and (2) coherent material motion (sloshing) potential of recriticality with energetics Typical nuclear power trace with multiple excursions until final core dis-assembly Z r 25

26 6. SAFETY: validation of employed simulation tools in an IAEA CRP Experimental and calculation analyses of SVRE, Sodium Void Reactivity Effect Experimental BFS-62, mock-up of a proposed hybrid UOX/MOX BN-600 core for Pu burning Experimental values adjusted for taking into account heterogeneity corrections k eff ± KIT results JEFF 3.0, 560- group data: Calculation- Experiment, C-E ± SVRE in LEZ, pcm (pcm=1.e-5) -20 ± 9 +4 ± 9 SVRE in MEZ, pcm -11 ± 4 +2 ± 4 SVRE in MOX, pcm -14 ± 6-6 ± 6 SVRE in HEZ, pcm -57 ± 6 +5 ± 6 SVRE TOTAL, pcm -102 ± ± 10 C-E values do not exceed uncertainties! 26

27 6. Safety: Validation of employed simulation tools: SPERT The experimental transient induced by 2.72 $ reactivity insertion with a period of 4.6 ms was simulated Reasonable agreement with experiment p p E6 position * Liu Ping, F. Gabrielli, A. Rineiski, W. Maschek, G. Bruna, Development of a plate-type fuel model for the neutronics & thermalhydraulics coupled code -SIMMER-III- and its application to the analyses of SPERT, Nuclear Engineering and Design, 240, (2010), pp

28 6. Safety: safety approach for CDA Optimize design to reduce the probability of CDA, e.g., try to exclude strong over-criticality caused by ULOF, which may initiate CDA (initial phase of a transient) Assess the maximum possible mechanical energy release, Optimize design again (after analyzing the transition phase, after a molten pool is created), Prove containment integrity Problems: High uncertainty margins in evaluations of the mechanical energy release => strong effort on simulation and validation by experiments needed Optimizations for initial phase of the transient may not always be optimal for the expansion phase: if an appreciable part of fuel (e.g., 30%) is ejected during the initial phase, no re-criticality may occur later. 28

29 6. Safety: JAEA safety approach for CDA New safety approach by JAEA (Japan): recriticality-free core with FAIDUS SAs Issue: enhancement of the void effect due to higher Na content Facilitates fuel ejection during accident 29

30 7. Core optimization: ESFR ESFR is the most recent EU project on SFR. Steel reflectors (blankets) around the core (1m high) Basic design with a fat pin (inner clad radius ca. 9.7 mm) by CEA, France Core void effect: ca pcm (near 4$), lower than in SPX Extended (core + upper structures) void effect: 1000 pcm Reduction of the sodium void effect is a major issue. OXIDE core characteristics Reactor Power (MWth) 3600 Core Inlet Temperature ( C) 395 Core Outlet Temperature ( C) 545 Ave. Core Structure Temperature ( C) 470 Ave. Fuel Temperature ( C) 1227 Inner Fuel S/As 225 Outer Fuel S/As 228 Control and Shutdown Device (CSD) Diverse Shutdown Device (DSD) Targeted Fuel residence time (efpd)

31 7. Core optimization: ESFR Attractive modification options: Large Na upper plenum with Absorber layer above (remove Upper axial blanket, lower Upper gas plenum; large Na plenum being first proposed for BN-800) Adoption of a Lower Fertile blanket (optionally with 5% vol. Am => increased proliferation resistance, more than 10% of Pu238 in Pu, discussed later) Adoption of empty SAs (discussed later) Adoption of an internal fertile layer Reduction of the core height. Extended void effect Na plenum and Absober (Boron) Layer above Near zero Boron layer Upper axial reflector Upper plenum +Shorter Upper gas plenum +Lower Fertile Blanket or Absorber +Empty SAs +Internal fertile layer +Core height reduction with extra radial SAs 100 pcm lower 100 pcm lower 200 pcm lower 200 pcm lower 200 pcm lower X Outer core Inner core Lower axial reflector Fission gas plenum Radial reflector 31

32 7. Core optimization: ESFR ULOF starts at 50s L3: L3: Na Na ULOF transient in reference ESFR at BOL: Reactivity up after boiling onset: possibly the initial phase for a CDA xx means reactivity in $ 32

33 7. Core optimization: ESFR ULOF starts at 50s ULOF transient in optimized ESFR at BOL: Reactivity drops down at Na boiling onset! 33

34 34 7. Core optimization: ESFR Using of empty SAs Reduces core void effect Strong own negative void May facilitate early fuel discharge May prevent fuel radial motion Conventional compared to FAIDUS, but can-wall failure assumed Addition of ~4% Am to the ESFR core and fertile blanket leads to the following mass balance, in kg per TWh(th): Core Reactor Np Pu Am Cm Optimized ESFR with 4% Am burns more MAs than it produces. Without fertile blanket (Core only) Pu production is minor Cm production is unfavorable, but Cm decays to Pu; Cm242, Cm244 half-lives being 160 days, 18 years, respectively. After Am addition, the safety parameters are deteriorated. A possible remedy is to reduce the core height and reactor power by 20%. This modification makes the void effect negative.

35 8. Fuels, structure materials First FR designs: metal fuel High metal fuel swelling under irradiation -> go to oxide in 1960s All large and medium (>300 MWth) reactors use MOX or UOX fuel: Higher Doppler effect due to softer spectrum (moderation by O). Achieved burnup in oxide: 200+ GWd/t (20+% of heavy atoms, 200+ DPAs, (Displacements Per Atom). In metal: a similar value reached in EBR-II (US) after increasing fuel porosity (to achieve smear density of fuel near 70%) and fission gas plenum volume Carbide (largest experience in FBTR, India) and nitride fuels: 10+% burnup achieved Highest values for experimental reactors/pins. In existing FR 10+% is now standard; limit: clad damage by irradiation New ODS steels to allow higher burnup limits. 35

36 8. Fuels, structure materials Fast Reactor Fuel Type Metal, U-20Pu-10Zr Oxide U02-20PuO2 Nitride UN-20PuN Carbide UC-20PuC Heavy Metal Density, g/cm^ Melting Temperature, K Thermal conductivity, W/cm-K Operating centerline Temperature at 40 kw/m, K (T/T_melt) 1060 (0.8) 2360 (0.8) 1000 (0.3) 1030 (0.4) Source: Robert N. Hill, NRC Topical Seminar on Sodium Fast Reactors, Two White Flint, Rockville, MD, May 3, 2007 Carbide fuel: highly pyrophoric, susceptible to oxidation and hydrolysis Nitride fuels: smaller oxidation and hydrolysis, N15 to be used 36

37 9. Experience, issues: common issues for SFRs Discontinuous operation for new medium/large designs at beginning of life, mainly because they were first-of-its-kind Sodium leaks, Steam Generator Tube failures Fuel clad integrity, can-wall swelling Successful operation, if enough time and budget for modifications allowed (reactors built since 1970s) Higher construction costs compared to conventional LWR 20 to 50% or more often assumed (on the basis of first-of-itskind experiences) +11% reported for BN-800 as compared to VVER-1000 per unit power (but smaller unit power for BN) Similar to LWR costs: targeted now in Japan and other countries MOX fuel reprocessing and fabrication in large scale for FR required 37

38 9. Experience, issues Specific problems: for particular reactors Positive reactivity feedbacks (EBR-I) SA blockage (Fermi 1) Sodium fires (BN-350, Monju) Damage during refuelling (Monju) 38

39 9. Experience, issues: for particular reactor types SFR (Na): Na interaction with air/water activation of Na (Na24) core inspection (opaque medium) LFR (Pb, Pb-Bi) : steel corrosion (coating needed) SG in the pool may lead to ingress of water/steam Po generation if Pb-Bi coolant core inspection (opaque medium) GFR (He): heat removal if pressure lost low power if steel cladding (clad surface roughening needed) dust if SiC cladding 39

40 9. Experience, issues: proliferation Proliferation problems exist also for conventional UOX fuel/lwr reactors: Relatively easy to enrich U further if already more than a few % (U232 presence makes it hardly possible) Using of re-enriched U: U235->U236-> Np237; the latter posing a proliferation concern Particular issue of FR: very high quality Pu in U blankets Replace fertile blankets by non-fertile (steel) reflectors Add a few% of Am to fertile blankets so that Pu would contain 6-8% or more of Pu238, a strong neutron and decay heat emitter leading to melting of a not very small bulk of Pu General comment: Nuclear weapon main developments in the past were done largely independently of the peaceful use of nuclear energy, both in major and de facto nuclear weapon states. 40

41 10. Recent trends European Union: Plan: construction of a larger ~600 MWe SFR (ASTRID in France) and smaller LFR (MYRRHA in Belgium, ALFRED in Romania), and GFR (ALLEGRO in Czech Republic, Slovakia or Hungary) reactors with MOX fuel India: Construction of PFBR (500 MWe), SFR with MOX fuel by 2012 Plan: several similar PFBRs with MOX in near term, metal fuel for short DT in future Japan: Review of different reactor coolants, fuels, reprocessing options led to that SFR with aqueous reprocessing and pellet-type MOX fuel was selected as the most preferred option for the future Plan: JSFR, a loop-type 1500 MWe SFR with MOX fuel Russia: Construction of BN-800, SFR with UOX/MOX by 2014 Plan: BN-1200 (SFR, MOX), SVBR-100 (Pb-Bi coolant), BREST-300 (Pb coolant) USA: Advanced metal or MOX burner reactor (CR<<1) proposed a few years ago China: Plan: Chinese projects and BN-800 with MOX in near term, metal fuel in the future 41

42 10. Recent trends: Travelling wave reactor (TWR) Most recent studies by a US private company. Similar concepts investigated earlier, since 1950s, aiming at avoiding MOX reprocessing and fabrication. Fertile SAs with depleted or natural U fuel are irradiated and then used as driver fuel for irradiating new fertile fuel SAs. Very high fuel burn-up is needed => New cladding materials for extremely high DPA values are needed, fuel re-cladding may have to be considered. Recent benchmark calculations at KIT Fertile fuel SA under irradiation. Conventional SFR power density and volume fractions of fuel, steel, Na. Computation of k-inf and isotopic composition vs. burnup. Metal Fuel: max. k-inf 1.25, Pu239 content 7.6%, Pu %, at 260 GWd/t Oxide Fuel: max. k-inf 1.07, Pu239 content 9.2%, Pu %, at 150 GWd/t Conclusion of the benchmark: in line with earlier results, TWR is hardly possible with oxide fuel, the maximum criticality is too low if leakage is taken into account 42

43 11. Conclusions Fast reactors and associated fuel cycle facilities are under development since 1950s in major nuclear countries. Initially as breeders in U-Pu cycle: for energy production, not limited by available U resources. Since 1990s also as burners: for managing nuclear waste. Major FR employment scenarios assume fuel reprocessing and refabrication associated with partitioning and transmutation for spent fuel management, small actinide reprocessing losses being already demonstrated at laboratory or pilot plant (MAs) or industrial (Pu) scales MA content in FR fuel is limited due to safety constraints therefore an appreciable fraction of FR in the nuclear fleet may be needed to manage spent fuel. Subcritical FR (ADS) may alleviate these limits. Higher construction/operation costs compared to conventional LWR often assumed, partially based on small unit powers and/or first-of-its-kind experience. Successful operation for reactors constructed since 1970s, if and after they were allowed to work for a sufficiently long time. The largest experience is for sodium-cooled reactors with oxide fuel, the only fast reactor type producing electricity now. Heavy liquid metal (Pb, Pb-Bi), gas are major alternatives to Na coolant. Metal, nitride and carbide fuels are major alternatives to oxide fuel. New FR designs are optimized for safety, economics, envisaged fuel cycles 43

Generation IV Fast Reactors. Dr Richard Stainsby AMEC Richard.Stainsby@amec.com

Generation IV Fast Reactors. Dr Richard Stainsby AMEC Richard.Stainsby@amec.com Generation IV Fast Reactors Dr Richard Stainsby AMEC Richard.Stainsby@amec.com Contents The Generation IV international research programme on advanced reactors The case for fast reactors The technology:

More information

Introduction to Nuclear Fuel Cycle and Advanced Nuclear Fuels

Introduction to Nuclear Fuel Cycle and Advanced Nuclear Fuels Introduction to Nuclear Fuel Cycle and Advanced Nuclear Fuels Jon Carmack Deputy National Technical Director Fuel Cycle Technology Advanced Fuels Program February 27, 2011 The Evolution of Nuclear Power

More information

DESIGN, SAFETY AND FUEL DEVELOPMENTS FOR THE EFIT ACCELERATOR DRIVEN SYSTEM WITH CERCER AND CERMET CORES

DESIGN, SAFETY AND FUEL DEVELOPMENTS FOR THE EFIT ACCELERATOR DRIVEN SYSTEM WITH CERCER AND CERMET CORES DESIGN, SAFETY AND FUEL DEVELOPMENTS FOR THE EFIT ACCELERATOR DRIVEN SYSTEM WITH CERCER AND CERMET CORES W. Maschek 1, C. Artioli 2, X. Chen 1, F. Delage 3, A. Fernandez-Carretero 4, M. Flad 1, A. Fokau

More information

THE COMPARISON OF THE PERFORMANCE FOR THE ALLOY FUEL AND THE INTER-METALLIC DISPERSION FUEL BY THE MACSIS-H AND THE DIMAC

THE COMPARISON OF THE PERFORMANCE FOR THE ALLOY FUEL AND THE INTER-METALLIC DISPERSION FUEL BY THE MACSIS-H AND THE DIMAC THE COMPARISON OF THE PERFORMANCE FOR THE ALLOY FUEL AND THE INTER-METALLIC DISPERSION FUEL BY THE MACSIS-H AND THE DIMAC Byoung-Oon Lee, Bong-S. Lee and Won-S. Park Korea Atomic Energy Research Institute

More information

ADS (Accelerator Driven Systems) e LFR (Lead Fast Reactor)

ADS (Accelerator Driven Systems) e LFR (Lead Fast Reactor) ADS (Accelerator Driven Systems) e LFR (Lead Fast Reactor) L. Mansani luigi.mansani@ann.ansaldo.it Workshop: La Trasmutazione ed i Cicli Innovativi del Combustibile Nucleare Roma 20 Dicembre 2010 Origin

More information

DEMONSTRATION ACCELERATOR DRIVEN COMPLEX FOR EFFECTIVE INCINERATION OF 99 Tc AND 129 I

DEMONSTRATION ACCELERATOR DRIVEN COMPLEX FOR EFFECTIVE INCINERATION OF 99 Tc AND 129 I DEMONSTRATION ACCELERATOR DRIVEN COMPLEX FOR EFFECTIVE INCINERATION OF 99 Tc AND 129 I A.S. Gerasimov, G.V. Kiselev, L.A. Myrtsymova State Scientific Centre of the Russian Federation Institute of Theoretical

More information

THORIUM UTILIZATION FOR SUSTAINABLE SUPPLY OF NUCLEAR ENERGY S. BANERJEE DEPARTMENT OF ATOMIC ENERGY

THORIUM UTILIZATION FOR SUSTAINABLE SUPPLY OF NUCLEAR ENERGY S. BANERJEE DEPARTMENT OF ATOMIC ENERGY THORIUM UTILIZATION FOR SUSTAINABLE SUPPLY OF NUCLEAR ENERGY S. BANERJEE DEPARTMENT OF ATOMIC ENERGY INDIA DHRUVA CIRUS 1 PLAN OF TALK Introduction Three Stage Nuclear Power Programme Thorium Utilisation

More information

Lead-Cooled Fast Reactor BREST Project Status and Prospects

Lead-Cooled Fast Reactor BREST Project Status and Prospects 資 料 4-3 第 11 回 GIF-LFR pssc (イタリア ピサ 2012 年 4 月 16 日 ) ロシア 側 発 表 資 料 V. S. Smirnov State Atomic Energy Corporation ROSATOM Open Joint-Stock Company N.A. Dollezhal Research and Development Institute of

More information

A short history of reactors

A short history of reactors A short history of reactors Janne Wallenius Reactor Physics, KTH Objectives of this meeting The origin of nuclear power was considerably more diversified than the existing variation in commercial reactor

More information

Fast reactor development program in Russia

Fast reactor development program in Russia International Conference on Fast Reactors and Related Fuel Cycles FR13 4-7 March 2013 Paris, France. Fast reactor development program in Russia Presented by Valery Rachkov ITC PRORYV Project, Moscow, Russia

More information

Chapter 6 Impact of Fukushima Daiichi Accident on Japan s Nuclear Fuel Cycle and Spent Fuel Management

Chapter 6 Impact of Fukushima Daiichi Accident on Japan s Nuclear Fuel Cycle and Spent Fuel Management Chapter 6 Impact of Fukushima Daiichi Accident on Japan s Nuclear Fuel Cycle and Spent Fuel Management Joonhong Ahn Abstract This chapter briefly summarizes the current status of spent nuclear fuel and

More information

Review of Generation IV Nuclear Energy Systems R E P O R T

Review of Generation IV Nuclear Energy Systems R E P O R T Review of Generation IV Nuclear Energy Systems R E P O R T REPORT SUMMARY According to many energy foresight studies carried out in the late 1990s, a shortage of uranium can be expected during the 21st

More information

The Fuel Cycle R&D Program. Systems Analysis

The Fuel Cycle R&D Program. Systems Analysis The Fuel Cycle R&D Program Systems Analysis Bradley Williams Systems Analysis Federal Program Manager Office of Systems Engineering and Integration Office of Nuclear Energy U.S. Department of Energy July

More information

On the neutronics of European lead-cooled fast reactor

On the neutronics of European lead-cooled fast reactor NUKLEONIKA 2010;55(3):317 322 ORIGINAL PAPER On the neutronics of European lead-cooled fast reactor Jerzy Cetnar, Mikołaj Oettingen, Grażyna Domańska Abstract. The perspective of nuclear energy development

More information

V K Raina. Reactor Group, BARC

V K Raina. Reactor Group, BARC Critical facility for AHWR and PHWRs V K Raina Reactor Group, BARC India has large reserves of Thorium Critical facility Utilisation of Thorium for power production is a thrust area of the Indian Nuclear

More information

The Physics of Energy sources Nuclear Reactor Practicalities

The Physics of Energy sources Nuclear Reactor Practicalities The Physics of Energy sources Nuclear Reactor Practicalities B. Maffei Bruno.maffei@manchester.ac.uk www.jb.man.ac.uk/~bm Nuclear Reactor 1 Commonalities between reactors All reactors will have the same

More information

BWR Description Jacopo Buongiorno Associate Professor of Nuclear Science and Engineering

BWR Description Jacopo Buongiorno Associate Professor of Nuclear Science and Engineering BWR Description Jacopo Buongiorno Associate Professor of Nuclear Science and Engineering 22.06: Engineering of Nuclear Systems 1 Boiling Water Reactor (BWR) Public domain image by US NRC. 2 The BWR is

More information

Russian Nuclear Power Program (past, present, and future) Dr. Igor Pioro Senior Scientist CRL AECL

Russian Nuclear Power Program (past, present, and future) Dr. Igor Pioro Senior Scientist CRL AECL Russian Nuclear Power Program (past, present, and future) Dr. Igor Pioro Senior Scientist CRL AECL Nuclear Power Units by Nation No. Nation # Units Net MWe 1 USA 104 100,460 2 France 59 63,363 3 Japan

More information

Fukushima 2011. Fukushima Daiichi accident. Nuclear fission. Distribution of energy. Fission product distribution. Nuclear fuel

Fukushima 2011. Fukushima Daiichi accident. Nuclear fission. Distribution of energy. Fission product distribution. Nuclear fuel Fukushima 2011 Safety of Nuclear Power Plants Earthquake and Tsunami Accident initiators and progression Jan Leen Kloosterman Delft University of Technology 1 2 Nuclear fission Distribution of energy radioactive

More information

The Future of the Nuclear Fuel Cycle

The Future of the Nuclear Fuel Cycle The Future of the Nuclear Fuel Cycle An Interdisciplinary MIT Study Ernest J. Moniz Cecil and Ida Green Professor of Physics and Engineering Systems Director, MIT Energy Initiative Study Participants Co-chairs:

More information

Retrieval of Damaged Components form Experimental Fast Reactor Joyo Reactor Vessel

Retrieval of Damaged Components form Experimental Fast Reactor Joyo Reactor Vessel Retrieval of Damaged Components form Experimental Fast Reactor Joyo Reactor Vessel June. 8 th, 2010 Yukimoto MAEDA Japan Atomic Energy Agency (JAEA) Experimental fast reactor Joyo Joyo (Oarai R&D Center)

More information

RESULTS OF R&D ON FUTURE FUEL CYCLE AND ASSOCIATED HL WASTE DISPOSAL: THE FRENCH CASE

RESULTS OF R&D ON FUTURE FUEL CYCLE AND ASSOCIATED HL WASTE DISPOSAL: THE FRENCH CASE IGDTP_SNETP_EF4, October 29-30, 2013, Praha RESULTS OF R&D ON FUTURE FUEL CYCLE AND ASSOCIATED HL WASTE DISPOSAL: THE FRENCH CASE Dominique Warin CEA / Nuclear Energy Direction 29 OCTOBRE 2013 14 juin

More information

LEAD-COOLED FAST REACTOR (LFR) DEVELOPMENT GAPS

LEAD-COOLED FAST REACTOR (LFR) DEVELOPMENT GAPS LEAD-COOLED FAST REACTOR (LFR) DEVELOPMENT GAPS M. TARANTINO Italian National Agency for New Technologies, Energy and Sustainable Economic Development, C.R. ENEA Brasimone Italy mariano.tarantino@enea.it

More information

Evaluation of Sodium-cooled Fast Reactor Neutronic Benchmarks

Evaluation of Sodium-cooled Fast Reactor Neutronic Benchmarks Evaluation of Sodium-cooled Fast Reactor Neutronic Benchmarks N.E. Stauff a, T.K. Kim a, T. Taiwo a, L. Buiron b, F. Varaine b, J. Gulliford c a Argonne National Laboratory, Nuclear Engineering Division,

More information

Fission fragments or daughters that have a substantial neutron absorption cross section and are not fissionable are called...

Fission fragments or daughters that have a substantial neutron absorption cross section and are not fissionable are called... KNOWLEDGE: K1.01 [2.7/2.8] B558 Fission fragments or daughters that have a substantial neutron absorption cross section and are not fissionable are called... A. fissile materials. B. fission product poisons.

More information

Fuel Cycle R&D to Safeguard Advanced Ceramic Fuel Skills Strategic Options

Fuel Cycle R&D to Safeguard Advanced Ceramic Fuel Skills Strategic Options Fuel Cycle R&D to Safeguard Advanced Ceramic Fuel Skills Strategic Options Fuel Cycle R&D to Safeguard Advanced Ceramic Fuel Skills The Nuclear Renaissance and Fuel Cycle Research and Development Nuclear

More information

Japan s current Nuclear Energy Policy

Japan s current Nuclear Energy Policy Japan s current Nuclear Energy Policy Hirobumi Kayama Agency for Natural Resources and Energy, METI December, 2014 Evaluation and Policy Timeframe of Nuclear Power

More information

Flow distribution and turbulent heat transfer in a hexagonal rod bundle experiment

Flow distribution and turbulent heat transfer in a hexagonal rod bundle experiment Flow distribution and turbulent heat transfer in a hexagonal rod bundle experiment K. Litfin, A. Batta, A. G. Class,T. Wetzel, R. Stieglitz Karlsruhe Institute of Technology Institute for Nuclear and Energy

More information

Heterogeneous world model and collaborative scenarios of transition to globally sustainable nuclear energy systems

Heterogeneous world model and collaborative scenarios of transition to globally sustainable nuclear energy systems EPJ Nuclear Sci. Technol. 1, 1 (2015) V. Kuznetsov and G. Fesenko, published by EDP Sciences, 2015 DOI: 10.1051/epjn/e2015-50031-2 Nuclear Sciences & Technologies Available online at: http://www.epj-n.org

More information

2012 by Rijan Prasad Shrestha. All Rights Reserved.

2012 by Rijan Prasad Shrestha. All Rights Reserved. 2012 by Rijan Prasad Shrestha. All Rights Reserved. STUDY OF SPACE-TIME EVOLUTION OF FLUX IN A LONG-LIFE TRAVELING WAVE REACTOR BY RIJAN PRASAD SHRESTHA THESIS Submitted in partial fulfillment of the requirements

More information

SCDAP/RELAP5-3D : A State-of-the-Art Tool for Severe Accident Analyses

SCDAP/RELAP5-3D : A State-of-the-Art Tool for Severe Accident Analyses SCDAP/RELAP5-3D : A State-of-the-Art Tool for Severe Accident Analyses D. L. Knudson and J. L. Rempe 2002 RELAP5 International Users Seminar Park City, Utah, USA September 4-6, 2002 Objective and Outline

More information

Prospect of Hitachi Nuclear Business (Boiling Water Reactor)

Prospect of Hitachi Nuclear Business (Boiling Water Reactor) Prospect of Hitachi Nuclear Business (Boiling Water Reactor) 42 Prospect of Hitachi Nuclear Business (Boiling Water Reactor) Masahito Yoshimura Shoichiro Kinoshita Hiroshi Arima Nobuo Tada OVERVIEW: To

More information

Progress status of the FUTURIX-FTA PIE programme for pins CEA-7 & CEA-8. F. Delage, G. Cécilia, J. Lamontagne, L. Loubet

Progress status of the FUTURIX-FTA PIE programme for pins CEA-7 & CEA-8. F. Delage, G. Cécilia, J. Lamontagne, L. Loubet Progress status of the FUTURIX-FTA PIE programme for pins CEA-7 & CEA-8 F. Delage, G. Cécilia, J. Lamontagne, L. Loubet FAIRFUELS final workshop May 20, 2015 20/05/2015 CEA 10 AVRIL 2012 PAGE 1 He bonded

More information

Plutonium Watch. Tracking Plutonium Inventories by David Albright and Kimberly Kramer. July 2005, Revised August 2005

Plutonium Watch. Tracking Plutonium Inventories by David Albright and Kimberly Kramer. July 2005, Revised August 2005 Plutonium Watch Tracking Plutonium Inventories by David Albright and Kimberly Kramer July 25, Revised August 25 Plutonium is a key ingredient in nuclear weapons, making it one of the most dangerous materials

More information

Technology Roadmap Update for Generation IV Nuclear Energy Systems

Technology Roadmap Update for Generation IV Nuclear Energy Systems Technology Roadmap Update for Generation IV Nuclear Energy Systems Preparing Today for Tomorrow s Energy Needs Technology Roadmap Update for Generation IV Nuclear Energy Systems January 2014 Issued by

More information

Thorium for Use in Plutonium Disposition, Proliferation-Resistant Fuels for Developing Countries, and Future Reactor Designs

Thorium for Use in Plutonium Disposition, Proliferation-Resistant Fuels for Developing Countries, and Future Reactor Designs Thorium for Use in Plutonium Disposition, Proliferation-Resistant Fuels for Developing Countries, and Future Reactor Designs Brian Johnson Oregon State University WISE 2006 e Sponsored By Executive Summary

More information

DESIGN FEATURES OF BREST REACTORS AND EXPERIMENTAL WORK TO ADVANCE THE CONCEPT OF BREST REACTORS

DESIGN FEATURES OF BREST REACTORS AND EXPERIMENTAL WORK TO ADVANCE THE CONCEPT OF BREST REACTORS DESIGN FEATURES OF BREST REACTORS AND EXPERIMENTAL WORK TO ADVANCE THE CONCEPT OF BREST REACTORS A.I. FILIN, V.V. ORLOV, V.N. LEONOV, A.G. SILA-NOVITSKI, V.S. SMIRNOV, V.S. TSIKUNOV STATE SCIENTIFIC CENTER

More information

Chapter 1 The Development of Nuclear Energy in the World

Chapter 1 The Development of Nuclear Energy in the World Chapter 1 The Development of Nuclear Energy in the World Abstract In 2011 there were about 436 commercial nuclear power reactors operating in the world with a total capacity of 370 GW(e). Nuclear energy

More information

Global Nuclear Energy Partnership Technology Demonstration Program

Global Nuclear Energy Partnership Technology Demonstration Program Global Nuclear Energy Partnership Technology Demonstration Program Dave Hill Deputy Director, Science and Technology Idaho National Laboratory August 10 2006 Global Nuclear Energy Partnership Goals of

More information

TECHNICAL WHITE PAPER MARCH 2014 V 1.0.1

TECHNICAL WHITE PAPER MARCH 2014 V 1.0.1 TECHNICAL WHITE PAPER MARCH 2014 V 1.0.1 Transatomic Power s advanced molten salt reactor consumes spent nuclear fuel cleanly and completely, unlocking vast amounts of cheap, carbon-free energy. It solves

More information

CFD Topics at the US Nuclear Regulatory Commission. Christopher Boyd, Ghani Zigh Office of Nuclear Regulatory Research June 2008

CFD Topics at the US Nuclear Regulatory Commission. Christopher Boyd, Ghani Zigh Office of Nuclear Regulatory Research June 2008 CFD Topics at the US Nuclear Regulatory Commission Christopher Boyd, Ghani Zigh Office of Nuclear Regulatory Research June 2008 Overview Computational Fluid Dynamics (CFD) is playing an ever increasing

More information

10 Nuclear Power Reactors Figure 10.1

10 Nuclear Power Reactors Figure 10.1 10 Nuclear Power Reactors Figure 10.1 89 10.1 What is a Nuclear Power Station? The purpose of a power station is to generate electricity safely reliably and economically. Figure 10.1 is the schematic of

More information

DEVELOPMENT OF A DYNAMIC SIMULATION MODE IN SERPENT 2 MONTE CARLO CODE

DEVELOPMENT OF A DYNAMIC SIMULATION MODE IN SERPENT 2 MONTE CARLO CODE International Conference on Mathematics and Computational Methods Applied to Nuclear Science & Engineering (M&C 2013) Sun Valley, Idaho, USA, May 5-9, 2013, on CD-ROM, American Nuclear Society, LaGrange

More information

PHENIX: THE IRRADIATION PROGRAM FOR TRANSMUTATION EXPERIMENTS

PHENIX: THE IRRADIATION PROGRAM FOR TRANSMUTATION EXPERIMENTS PHENIX: THE IRRADIATION PROGRAM FOR TRANSMUTATION EXPERIMENTS J. Guidez, P. Chaucheprat, B. Fontaine, E. Brunon, L. Martin (PHENIX) D. Warin (DDIN) A. Zaetta (DEN/CAD/DER) F. Sudreau (DEN/CAD/DEC) Abstract

More information

Boiling Water Reactor Systems

Boiling Water Reactor Systems Boiling Water (BWR) s This chapter will discuss the purposes of some of the major systems and components associated with a boiling water reactor (BWR) in the generation of electrical power. USNRC Technical

More information

DOE s Fuel Cycle Technologies Program - An Overview

DOE s Fuel Cycle Technologies Program - An Overview DOE s Fuel Cycle Technologies Program - An Overview William Boyle Director, Used Fuel Disposition R & D (NE-53) Office of Nuclear Energy U.S. Department of Energy ECA Las Vegas, Nevada Agenda Program Mission

More information

ADVANCED NUCLEAR ENERGY SYSTEM FOR THE TWENTY-FIRST CENTURY

ADVANCED NUCLEAR ENERGY SYSTEM FOR THE TWENTY-FIRST CENTURY ADVANCED NUCLEAR ENERGY SYSTEM FOR THE TWENTY-FIRST CENTURY Yoon Il Chang Argonne National Laboratory 9700 South Cass Avenue Argonne, IL USA 60439 ychang@anl.gov ABSTRACT The world needs more energy to

More information

Conservative approach for PWR MOX Burnup Credit implementation

Conservative approach for PWR MOX Burnup Credit implementation International Conference on the Physics of Reactors Nuclear Power: A Sustainable Resource Casino-Kursaal Conference Center, Interlaken, Switzerland, September 14-19, 2008 Conservative approach for PWR

More information

Microwave absorbing tiles:

Microwave absorbing tiles: On the basis of the results obtained from the first project activities, the grinding conditions on a larger scale were determined. As regards the sintering, an adjustment has been made to the roller furnaces

More information

Nuclear Energy: Nuclear Energy

Nuclear Energy: Nuclear Energy Introduction Nuclear : Nuclear As we discussed in the last activity, energy is released when isotopes decay. This energy can either be in the form of electromagnetic radiation or the kinetic energy of

More information

Fuel Development, Testing, and Schedule for the Traveling Wave Reactor

Fuel Development, Testing, and Schedule for the Traveling Wave Reactor Fuel Development, Testing, and Schedule for the Traveling Wave Reactor Kevan Weaver DOE-NRC Second Workshop on Advanced Non-LWR Reactors North Bethesda Marriott, June 7-8, 2016 The Traveling Wave Reactor

More information

DOE FUNDAMENTALS HANDBOOK

DOE FUNDAMENTALS HANDBOOK DOE-HDBK-1019/2-93 JANUARY 1993 DOE FUNDAMENTALS HANDBOOK NUCLEAR PHYSICS AND REACTOR THEORY Volume 2 of 2 U.S. Department of Energy Washington, D.C. 20585 FSC-6910 Distribution Statement A. Approved for

More information

Research and Development Program of HTGR Fuel in Japan

Research and Development Program of HTGR Fuel in Japan Research and Development Program of HTGR Fuel in Japan Kazuhiro SAWA, Shouhei UETA, Tatsuo IYOKU Masuro OGAWA, Yoshihiro KOMORI Department of Advanced Nuclear Heat Technology Department of HTTR Project

More information

Appendix 6.0 Molten Salt Reactor

Appendix 6.0 Molten Salt Reactor Appendix 6.0 Molten Salt Reactor A6-1 This page intentionally left blank. A6-2 CONTENTS ACRONYMS... 5 A6.1 INTRODUCTION AND BACKGROUND... 7 A6.1.1 System Description... 7 A6.1.2 Overall Systems Timeline...

More information

Japan s Long Term Nuclear Power Policy and Rising Need for Energy in East Asia 1. Shinzo SAITO Vice Chairman Atomic Energy Commission of Japan

Japan s Long Term Nuclear Power Policy and Rising Need for Energy in East Asia 1. Shinzo SAITO Vice Chairman Atomic Energy Commission of Japan Japan s Long Term Nuclear Power Policy and Rising Need for Energy in East Asia 1 Shinzo SAITO Vice Chairman Atomic Energy Commission of Japan Thank you Mr. chairman for your kind introduction. Good Morning

More information

Accidents of loss of flow for the ETTR-2 reactor: deterministic analysis

Accidents of loss of flow for the ETTR-2 reactor: deterministic analysis NUKLEONIKA 2000;45(4):229 233 ORIGINAL PAPER Accidents of loss of flow for the ETTR-2 reactor: deterministic analysis Ahmed Mohammed El-Messiry Abstract The main objective for reactor safety is to keep

More information

COMPARISON BETWEEN TWO GAS-COOLED TRU BURNER SUBCRITICAL REACTORS: FUSION-FISSION AND ADS

COMPARISON BETWEEN TWO GAS-COOLED TRU BURNER SUBCRITICAL REACTORS: FUSION-FISSION AND ADS International Conference on Mathematics and Computational Methods Applied to Nuclear Science and Engineering (M&C 2011) Rio de Janeiro, RJ, Brazil, May 8-12, 2011, on CD-ROM, Latin American Section (LAS)

More information

Dario Manara, Rudy Konings, Philippe Raison

Dario Manara, Rudy Konings, Philippe Raison LFR fuel overview and perspectives Dario Manara, Rudy Konings, Philippe Raison European Commission Joint Research Centre Institute for transuranium materials (ITU) P.O. Box 2340 76125 Karlsruhe Germany

More information

PROPOSALS FOR UNIVERSITY REACTORS OF A NEW GENERATION

PROPOSALS FOR UNIVERSITY REACTORS OF A NEW GENERATION PROPOSALS FOR UNIVERSITY REACTORS OF A NEW GENERATION Introduction R.P. Kuatbekov, O.A. Kravtsova, K.A. Nikel, N.V. Romanova, S.A. Sokolov, I.T. Tretiyakov, V.I. Trushkin (NIKIET, Moscow, Russia) Worldwide,

More information

MCQ - ENERGY and CLIMATE

MCQ - ENERGY and CLIMATE 1 MCQ - ENERGY and CLIMATE 1. The volume of a given mass of water at a temperature of T 1 is V 1. The volume increases to V 2 at temperature T 2. The coefficient of volume expansion of water may be calculated

More information

Technical Challenges for Conversion of U.S. High-Performance Research Reactors (USHPRR)

Technical Challenges for Conversion of U.S. High-Performance Research Reactors (USHPRR) Technical Challenges for Conversion of U.S. High-Performance Research Reactors (USHPRR) John G. Stevens, Ph.D. Argonne National Laboratory Technical Lead of Reactor Conversion GTRI USHPRR Conversion Program

More information

Economics of Thorium and Uranium Reactors

Economics of Thorium and Uranium Reactors Sherman Lam HSA 10-05 The Economics of Oil and Energy April 30, 2013 Economics of Thorium and Uranium Reactors In February 2012, the Nuclear Regulatory Commission (NRC) approved a license for two new nuclear

More information

Basics of Nuclear Physics and Fission

Basics of Nuclear Physics and Fission Basics of Nuclear Physics and Fission A basic background in nuclear physics for those who want to start at the beginning. Some of the terms used in this factsheet can be found in IEER s on-line glossary.

More information

How To Understand The Sensitivity Of A Nuclear Fuel Cask

How To Understand The Sensitivity Of A Nuclear Fuel Cask IAEA Conference Paper Thursday, February 16, 2006 Nuclear Science and Technology Division Application of Sensitivity/Uncertainty Methods to Burnup Credit Validation D. E. Mueller and J. C. Wagner Oak Ridge

More information

BEST-ESTIMATE TRANSIENT ANALYSIS WITH SKETCH-INS/TRAC-BF1, ASSESSMENT AGAINST OECD/NEA BWR TURBINE TRIP BENCHMARK ABSTRACT

BEST-ESTIMATE TRANSIENT ANALYSIS WITH SKETCH-INS/TRAC-BF1, ASSESSMENT AGAINST OECD/NEA BWR TURBINE TRIP BENCHMARK ABSTRACT BEST-ESTIMATE TRANSIENT ANALYSIS WITH SKETCH-INS/TRAC-BF1, ASSESSMENT AGAINST OECD/NEA BWR TURBINE TRIP BENCHMARK Hideaki Utsuno, Fumio Kasahara Nuclear Power Engineering Corporation (NUPEC) Fujita kanko

More information

TECHNICAL MEETING ON IN-PILE TESTING AND INSTRUMENTATION FOR DEVELOPMENT OF GENERATION-IV FUELS AND STRUCTURAL MATERIALS

TECHNICAL MEETING ON IN-PILE TESTING AND INSTRUMENTATION FOR DEVELOPMENT OF GENERATION-IV FUELS AND STRUCTURAL MATERIALS 651-T1-TM-42785 TECHNICAL MEETING ON IN-PILE TESTING AND INSTRUMENTATION FOR DEVELOPMENT OF GENERATION-IV FUELS AND STRUCTURAL MATERIALS I. BACKGROUND Halden, Norway 21 24 August 2012 INFORMATION SHEET

More information

A Strategy for Nuclear Energy Research and Development

A Strategy for Nuclear Energy Research and Development INL/EXT-08-15158 Rev. 0 A Strategy for Nuclear Energy Research and Development December 2008 Electric Power Research Institute and Idaho National Laboratory DISCLAIMER OF WARRANTIES AND LIMITATION OF LIABILITIES

More information

NUCLEARINSTALLATIONSAFETYTRAININGSUPPORTGROUP DISCLAIMER

NUCLEARINSTALLATIONSAFETYTRAININGSUPPORTGROUP DISCLAIMER NUCLEARINSTALLATIONSAFETYTRAININGSUPPORTGROUP DISCLAIMER Theinformationcontainedinthisdocumentcannotbechangedormodifiedinanywayand shouldserveonlythepurposeofpromotingexchangeofexperience,knowledgedissemination

More information

Achim Beisiegel Fouad El-Rharbaoui Michael Wich. AREVA GmbH, Technical Center, 63791 Karlstein, Seligenstädter Strasse 100, Germany

Achim Beisiegel Fouad El-Rharbaoui Michael Wich. AREVA GmbH, Technical Center, 63791 Karlstein, Seligenstädter Strasse 100, Germany Achim Beisiegel Fouad El-Rharbaoui Michael Wich AREVA GmbH, Technical Center, 63791 Karlstein, Seligenstädter Strasse 100, Germany 1 Introduction AREVA GmbH operates a unique Thermal-hydraulic platform

More information

Strategic Research Agenda of the MTA Centre for Energy Research related to the Atomic Energy Research Version 2013

Strategic Research Agenda of the MTA Centre for Energy Research related to the Atomic Energy Research Version 2013 Strategic Research Agenda of the MTA Centre for Energy Research related to the Atomic Energy Research Version 2013 The Strategic Research Agenda related to the Atomic Energy Research is determined by the

More information

Polyvalent fuel treatment facility (TCP) : shearing and dissolution of used fuel at La Hague facility

Polyvalent fuel treatment facility (TCP) : shearing and dissolution of used fuel at La Hague facility Polyvalent fuel treatment facility (TCP) : shearing and dissolution of used fuel at La Hague facility Frédéric LELIEVRE AREVA, Recycling Business Unit, Paris, France INTRODUCTION AREVA s current La Hague

More information

NEW NUCLEAR POWER PLANTS FOR ONTARIO - THE CONTENDERS - 2008 October

NEW NUCLEAR POWER PLANTS FOR ONTARIO - THE CONTENDERS - 2008 October NEW NUCLEAR POWER PLANTS FOR ONTARIO - THE CONTENDERS - 2008 October (THIS OVERVIEW IS AN UPDATED VERSION OF AN EARLIER ONE THAT WAS PUBLISHED IN THE BULLETIN OF THE CANADIAN NUCLEAR SOCIETY, MARCH 2008,

More information

Physics and Engineering of the EPR

Physics and Engineering of the EPR Physics and Engineering of the EPR Keith Ardron UK Licensing Manager, UK Presentation to IOP Nuclear Industry Group Birchwood Park, Warrington UK, November 10 2010 EPRs in UK EPR is Generation 3+ PWR design

More information

Uncertainty Estimation of Target System Multiplication Factors with the new COMMARA Covariance Matrix

Uncertainty Estimation of Target System Multiplication Factors with the new COMMARA Covariance Matrix Uncertainty Estimation of arget ystem Multiplication Factors with the new COMMARA Covariance Matrix Gerardo Aliberti, Won i Yang, and R. D. McKnight Nuclear Engineering Division Argonne National Laboratory

More information

Dynamic Behavior of BWR

Dynamic Behavior of BWR Massachusetts Institute of Technology Department of Nuclear Science and Engineering 22.06 Engineering of Nuclear Systems Dynamic Behavior of BWR 1 The control system of the BWR controls the reactor pressure,

More information

GNS Activities - Solutions for Russian Spent Fuel -

GNS Activities - Solutions for Russian Spent Fuel - GNS Activities - Solutions for Russian Spent Fuel - Bulgarian nuclear energynational, regional and world energy safety June 09-11, 2010 Varna, Bulgaria Dr. Felix Thomas GNS Gesellschaft für Nuklear-Service

More information

Nuclear Hydrogen Production: Re-Examining the Fusion Option

Nuclear Hydrogen Production: Re-Examining the Fusion Option Nuclear Hydrogen Production: Re-Examining the Fusion Option sbaindur@ottawapolicyresearch.ca http://ottawapolicyresearch.ca May 30th, 2007 Canadian Hydrogen Association Workshop on Hydrogen Production

More information

Closing the CANDU Fuel Cycle

Closing the CANDU Fuel Cycle WiN Canada Conference 2013 Closing the CANDU Fuel Cycle with Modified PUREX Recycling CANDU Spent Fuel Authors Juhx Pellazar Jinah Kim Alexander Koven Sameena Mulam (Presenter) Marty Tzolov Leon Wu Closing

More information

The soot and scale problems

The soot and scale problems Dr. Albrecht Kaupp Page 1 The soot and scale problems Issue Soot and scale do not only increase energy consumption but are as well a major cause of tube failure. Learning Objectives Understanding the implications

More information

Nuclear Physics. Nuclear Physics comprises the study of:

Nuclear Physics. Nuclear Physics comprises the study of: Nuclear Physics Nuclear Physics comprises the study of: The general properties of nuclei The particles contained in the nucleus The interaction between these particles Radioactivity and nuclear reactions

More information

Fundamentals of Nuclear Power

Fundamentals of Nuclear Power Fundamentals of Nuclear Power Juan S. Giraldo Douglas J. Gotham David G. Nderitu Paul V. Preckel Darla J. Mize State Utility Forecasting Group December 2012 Table of Contents List of Figures... iii List

More information

Nuclear Power Policy in Russia

Nuclear Power Policy in Russia Theocharis Grigoriadis FB Wirtschaftswissenschaft, Osteuropainstitut Nuclear Power Policy in Russia 19th REFORM Group Meeting Salzburg, 5th September 2014 Key Questions What is the role of nuclear power

More information

Development Study of Nuclear Power Plants for the 21st Century

Development Study of Nuclear Power Plants for the 21st Century Development Study of Nuclear Power Plants for the 21st Century Hitachi Review Vol. 50 (2001), No. 3 61 Kumiaki Moriya Masaya Ohtsuka Motoo Aoyama, D.Eng. Masayoshi Matsuura OVERVIEW: Making use of nuclear

More information

ADDITIONAL INFORMATION ON MODERN VVER GEN III TECHNOLOGY. Mikhail Maltsev Head of Department JSC Atomenergoproekt

ADDITIONAL INFORMATION ON MODERN VVER GEN III TECHNOLOGY. Mikhail Maltsev Head of Department JSC Atomenergoproekt ADDITIONAL INFORMATION ON MODERN VVER GEN III TECHNOLOGY Mikhail Maltsev Head of Department JSC Atomenergoproekt February 12, 2015 Introduction The main intention of this presentation is to provide: -

More information

Structure and Properties of Atoms

Structure and Properties of Atoms PS-2.1 Compare the subatomic particles (protons, neutrons, electrons) of an atom with regard to mass, location, and charge, and explain how these particles affect the properties of an atom (including identity,

More information

This article appeared in a journal published by Elsevier. The attached copy is furnished to the author for internal non-commercial research and

This article appeared in a journal published by Elsevier. The attached copy is furnished to the author for internal non-commercial research and This article appeared in a journal published by Elsevier. The attached copy is furnished to the author for internal non-commercial research and education use, including for instruction at the authors institution

More information

INTRODUCTION. The ANSWERS Software Service WIMS. 1998 AEA Technology plc

INTRODUCTION. The ANSWERS Software Service WIMS. 1998 AEA Technology plc INTRODUCTION INTRODUCTION 1 Overview 1 is the most extensive and versatile software package currently available for neutronics calculations. has an open structure which comprises a set of methods linked

More information

CESAR: A Code for Nuclear Fuel and Waste Characterisation

CESAR: A Code for Nuclear Fuel and Waste Characterisation CESAR: A Code for Nuclear Fuel and Waste Characterisation J.M. Vidal, J.P. Grouiller Commissariat à l Energie Atomique (CEA) Cadarache, 13108 Saint Paul lez Durance Cedex France A. Launay, Y. Berthion

More information

A MTR FUEL ELEMENT FLOW DISTRIBUTION MEASUREMENT PRELIMINARY RESULTS

A MTR FUEL ELEMENT FLOW DISTRIBUTION MEASUREMENT PRELIMINARY RESULTS A MTR FUEL ELEMENT FLOW DISTRIBUTION MEASUREMENT PRELIMINARY RESULTS W. M. Torres, P. E. Umbehaun, D. A. Andrade and J. A. B. Souza Centro de Engenharia Nuclear Instituto de Pesquisas Energéticas e Nucleares

More information

Overall Strategy for the Technology Development as the World's Leading Nuclear Company

Overall Strategy for the Technology Development as the World's Leading Nuclear Company Overall Strategy for the Technology Development as the World's Leading Nuclear Company KIYOSHI YAMAUCHI* 1 KEIZO OKADA* 1 (MHI) has been the sole manufacturer of pressurized water reactors (PWRs) in Japan

More information

CFD simulation of fibre material transport in a PWR core under loss of coolant conditions

CFD simulation of fibre material transport in a PWR core under loss of coolant conditions CFD simulation of fibre material transport in a PWR core under loss of coolant conditions T. Höhne, A. Grahn, S. Kliem Forschungszentrum Dresden- Rossendorf (FZD) Institut für Sicherheitsforschung Postfach

More information

Master Degree in Nuclear Engineering: Academic year 2007-2008

Master Degree in Nuclear Engineering: Academic year 2007-2008 Master Degree in Nuclear Engineering: Academic year 2007-2008 Number of students 2007-2008: University Politehnica Bucharest Total: 17 Language: Romanian For further information: www.cne.pub.ro Dates Title

More information

NUCLEAR POWER PLANT SYSTEMS and OPERATION

NUCLEAR POWER PLANT SYSTEMS and OPERATION Revision 4 July 2005 NUCLEAR POWER PLANT SYSTEMS and OPERATION Reference Text Professor and Dean School of Energy Systems and Nuclear Science University of Ontario Institute of Technology Oshawa, Ontario

More information

The TVEL Fuel Company of Rosatom CREATING FUTURE TODAY

The TVEL Fuel Company of Rosatom CREATING FUTURE TODAY The of Rosatom CREATING FUTURE TODAY For complete and current information, visit the company s website: www.tvel.ru TVEL_Nuclear 2 3 TVEL Fuel company Russian nuclear fuel designer, manufacturer and supplier

More information

Nuclear Design Practices and the Case of Loviisa 3

Nuclear Design Practices and the Case of Loviisa 3 Nuclear Design Practices and the Case of Loviisa 3 Harri Tuomisto Fortum Power, Finland Third Nuclear Power School, 20-22 October 2010, Gdańsk, Poland 22 October 2010 Harri Tuomisto 1 Objectives The objective

More information

Re: Deep Geologic Repository Project for Low and Intennediate Level Waste SUSTAINABLE NUCLEAR POWER AND NUCLEAR WASTE DISPOSAL

Re: Deep Geologic Repository Project for Low and Intennediate Level Waste SUSTAINABLE NUCLEAR POWER AND NUCLEAR WASTE DISPOSAL XYLENE POWER LTD. 20190 Kennedy Road, Sharon, Ontario LOG 1VO June 11 2014 J Dr. Stella Swanson Chair, Joint Review Panel Deep Geologic Repository Project C/O Canadian Nuclear Safety Commission 280 Slater

More information

Waste management. The failure of the rational waste management policy

Waste management. The failure of the rational waste management policy The failure of the rational waste management policy Another essential outcome of processing-recycling is to facilitate the management of radioactive waste. Compared to the direct disposal of spent fuel

More information

7.1 General 5 7.2 Events resulting in pressure increase 5

7.1 General 5 7.2 Events resulting in pressure increase 5 GUIDE YVL 2.4 / 24 Ma r ch 2006 Primary and secondary circuit pressure control at a nuclear power plant 1 Ge n e r a l 3 2 General design requirements 3 3 Pressure regulation 4 4 Overpressure protection

More information

Minor Actinide Transmutation. Position Paper

Minor Actinide Transmutation. Position Paper Minor Actinide Transmutation Position Paper Minor Actinide Transmutation A review For many years there has been a sustained international interest in partitioning and transmutation of the minor actinides

More information

Preliminary validation of the APROS 3-D core model of the new Loviisa NPP training simulator

Preliminary validation of the APROS 3-D core model of the new Loviisa NPP training simulator Preliminary validation of the APROS 3-D core model of the new Loviisa NPP training simulator Anssu Ranta-aho, Elina Syrjälahti, Eija Karita Puska VTT Technical Research Centre of Finland P.O.B 1000, FI-02044

More information