1- Fast reactor basic features

Transcription

1 1- Fast reactor basc features M. Salvatores The physcs of fast vs thermal neutrons Flexblty: breedng and/or burnng for dfferent mssons n the fuel cycle CONSULTANTS MEETING: EDUCATION & TRAINING SEMINAR ON FAST REACTOR SCIENCE AND TECHNOLOGY ITESM CAMPUS SANTA FE, MEXICO CITY 29 JUNE 03 JULY 2015

2 Classfcaton of nuclear power systems based on system technology a wde offer!

3 Fsson and other neutron-nucle nteracton reactons. A remnder Fsson Or: (n+nucleus wth A nucleons)= (A+1) excted (A+1) +γ: (n, γ) capture reacton A+n+γ: (n,n ) nelastc scatterng Reacton channels etc (A-1) +2n: (n,2n) scatterng

4 Energy dstrbuton of neutrons emtted n fsson

5 Thermal or fast neutrons? The most fundamental technologcal dfference between nuclear fsson reactors concerns the means by whch the problem of sustanng a chan reacton s acheved. One soluton s to slow down neutrons to so-called "thermal" energes (around ev) by usng a «moderator»: Moderator can be placed around a fuel lump to slow-down fsson neutrons from fast, MeV/keV, to thermal, below ev, energes Ths has the advantage of allowng a chan reacton to be sustaned usng natural or slghtly enrched uranum, and almost all of the world's operatng power reactors employ ths soluton -these are known as "thermal reactors". If water s the moderator, we speak of «Lght Water Reactors», LWR The dsadvantage of ths approach s that only 0.7% of uranum produces useful energy. Ths can be overcome by ncreasng the proporton of fssle atoms by enrchment, or by usng plutonum, and by constructng the reactor wthout a moderator. In ths case the average energy of the neutrons n the core s much greater than n thermal reactors (they are known as "fast" neutrons).

6 Neutron Moderaton and Cross Sectons Comparson Sgnfcant elastc scatterng of the neutrons n both spectra However, n FRs neutron moderaton s much less snce hgh A materals are used If sodum s chosen as coolant n FR, t s also the most moderatng materal In LWRs, neutrons are moderated prmarly by hydrogen Slowng-down power n FR s ~1% that observed for typcal LWR: Mnmum energy of a neutron after elastc collson s determned by the parameter α: E mn = αe where: Hydrogen 0 2 Oxygen α A A 1 1 Na U Thus, fast neutrons are ether absorbed or leak from the reactor before they can reach thermal energes

7 Comparson of LWR and SFR neutron spectra (Fsson neutrons are or are not slow down)

8 Fast reactor moderatng materals Slowng down power α A A 1 1 2

9 Comparson of fast reactor spectra

10 Energy dependence of neutron cross-sectons: Pu-239

11 Energy dependence of neutron cross-sectons: U-238

12 Fsson/Absorpton Integral cross sectons comparson 235 U 239 Pu 233 U Thermal Fast Thermal Fast Thermal Fast f (barn) 582 1, , ,79 2,42 2,43 2,87 2,94 2,49 2, PWR SFR The fsson/absorpton ratos are consstently hgher for the fast spectrum SFR. Thus, n a fast spectrum, actndes are preferentally fssoned, not transmuted nto hgher actndes FISSION / ABSORPTION f E EdE E E E c f de

13 Implcatons of fast spectrum physcs

14 Neutron balance comparson SFR CR Converson rato defned as TRU producton/tru destructon

15 Consequences of the neutron balance features For the equlbrum actnde concentraton n the core, the neutron producton fuel (per fsson) D eq n the fuel can be computed for a specfed composton consstng of -components wth proportons x : fuel D eq ( f a) / f 1 where: x x f f ; x x c f

16 The «neutron excess» N ex (n neutrons/fsson) can be obtaned as follows: N ex = fuel D eq -C par - C FP - L where: C par s the total «parastc» captures (per fsson) n the core (structures, coolant); C FP s the total captures (per fsson) of the fsson products and L s the total neutrons leakng out of the core (per fsson) In the case of a PWR, N ex s ~0 In the case of an FR, N ex s >1.0,.e. there s an excess of neutrons that can be used: to convert the non-fssle U-238 sotope (99.3% of uranum) nto Pu-239, whch s fssle and/or to «transmute» specfc elements (e.g. nuclear wastes)

17 Fast reactors flexblty As frst dscovered by Enrco Ferm back n 1944, the nuclear characterstcs of TRU cross sectons n a fast neutron spectrum, as dscussed prevously, allow a great FR flexblty: Breed (Converson Rato,.e. rato of TRU producton/tru destructon, CR>1) Burn TRU,.e. CR<1 Breed (e.g. Pu) and burn (wastes: e.g. MA) CR~1: Self-sustanng cycles (.e. fssle producton=fssle destructon). Wde coolant and fuel type choce accordng to the objectve, e.g. short Doublng Tme CTD (.e. the tme requred for a breeder reactor to produce enough materal to fuel a second reactor) : Na and dense (e.g. metal) fuels Wde range of MA content and dfferent Pu vectors or TRU compostons can be handled. FRs have a unque potental to keep a large range of fuel cycle optons open leavng a lmted legacy of hghly radotoxc and radoactve materal. Fuel cycle ssues should however be carefully analyzed

18 Breeder FR (CR>1) The nternal breedng gan s defned as the rato of the net gan (.e. producton mnus destructon) of fssle materal to the net destructon of fssle materal n the core: core n C 1 A INTERNAL BREEDING GAIN IBG n F n F F s the total fsson rate n the core n regon n n A s the total absorpton rate of fssle sotope n core regon n n : fssle sotope ndex n: core regon ndex n C 1 s the total capture rate of fssle sotope (-1) n core regon n The ω values characterze the reactvty of each sotope n a scale where the value of Pu-239 s 1 and that of U-238 s zero. n n f a f a 8 n n f a 9 f a 8

19 Addng an external blanket n order to capture the neutrons n excess, one can reach a total breedng gan (TBG) wthn a wde range of values. TBG = IBG + blanket n A good breedng translates (together wth hgh power densty) nto short doublng tmes CDT,.e. the tme requred for a breeder reactor to produce enough materal to fuel a second reactor : C n 1 A F n TD CDT M P th 1 T c f / T f 365 TBG M mass of Pu P th thermal power T c out-of-ple tme T core resdence tme f loadng factor In any case, a core wth IBG~0 wll help to acheve very long rradaton tmes

20 Normalzed TRU consumpton rate Burner fast reactors CR<1 (or IBG<<0): Varyng the rato TRU/(TRU+U) one can reach the maxmum theoretcal consumpton of TRU: Relaton between TRU consumpton rate and TRU fracton n crtcal Advanced Burner Fast Reactors: % of max. theoretcal consumpton can be obtaned wth: 0.7 TRU/(U+TRU) ~ Metal, MA/Pu~1 feed Oxde, MA/Pu~1 feed Metal, LWR-TRU feed Oxde, LWR-TRU feed both for metal or oxde fuelled cores and for a wde range of Pu/MA ratos TRU fracton

21 Alternatves: the ADS Potental safety problems n the case of a crtcal core loaded wth only TRU and wth a hgh content of Mnor Actndes. In these types of cores, the absence of Uranum produces both a very low fracton of delayed neutrons and a very low Doppler reactvty coeffcent (n general, mostly due to U-238 capture). Moreover, a hgh content of Mnor Actndes lke Am sotopes and Np nduces a deteroraton of the vod reactvty coeffcent (n case of lqud metal coolants). Sub-crtcal systems (or Accelerator Drven Systems ADS), were redscovered (~1985), snce they could provde a possble way out from these potental dffcultes. Concept stll to be demonstrated.

22 Cf-252 kg n the core Cf-252 kg n the core Alternatves (other than ADS): Deep burners: HTRs; IMF-LWRs and Increased Moderaton LWRs: In all cases, the neutron spectrum can be consderably softer than a standard PWR, and consequently low fsson-to-absorpton ratos for «gateway» sotopes (Pu-242, Cm-244 etc.) wll nduce very sgnfcant buld-up of hgher mass actndes (up to Cf sotopes) As a remnder, there s a hgh Cf-252 producton durng the rradaton of TRU fuel n a standard LWR wth respect to a FR: E E E E E number of cycles 0.0E number of cycles Cf-252 nventory n the core. Case of full TRU multrecycle n a LWR Cf-252 nventory n the core. Case of full TRU multrecycle n a FR

23 Fuel cycle ssues for FR wth low converson rato Feasblty ssues can arse when consderng not only the core feasblty but also fuel cycle performances. E.g. n the case of decay heat and neutron producton after post-rradaton coolng (at fuel fabrcaton) Reactor type PWR FR Fuel type Parameter MOX (Pu only) Homog TRU recycle Pu only Homog. TRU recycle, CR=1 and MA/Pu~0.1 Homog.TRU recycle, CR=0.5 and MA/Pu~1 Decay heat 1 x3 x0.5 x2.5 x38 Neutron source 1 x8000 ~1 x150 x4000

24 Isogenerator or break-even FRs: CR~1 (or TBG~0) At present, ths opton s a reference for FRs n some OECD countres wth nonprolferaton concerns Alternatve: LWRs wth CR~1 (harder spectrum). Many studes n the past. Dffculty to overcome problem of postve vod coeffcent, n partcular for degraded Pu vectors. New studes n Japan; the RBWR of Htach (CR~1 and negatve vod coeffcent) Moreover, lttle hope to burn MA, snce they can degrade further the vod coeffcent value.

25 Fast Reactors and close fuel cycle Fast reactors and multple recycle allow sustanablty n terms of resources optmal utlzaton. Uranum utlzaton wthout reprocessng has been envsaged snce an early proposal by Teller, and more recently by the Travellng Wave Reactor proposal of Terra Power However, no mracle soluton can be found wth any once-through cycle Moreover, used fuel f put n a repostory wll have comparable characterstcs (.e. actvty, resdual heat, radotoxcty etc.) as the used fuel of a standard PWR based once through cycle. A comprehensve analyss has been performed at ANL

26 Once-Through Nuclear Systems PWR TWR Travellng Wave Reactor, TWR Uranum utlsaton % 0.6 ~2-5

27 Conclusons (1) Fast reactors allow a great flexblty n the choce of nuclear energy deployment strateges That flexblty can be used to desgn evolutonary fast reactor cores that can burn or breed TRU accordng to the objectve. Ths can be done n prncple n the same vessel (reversblty concept) and wthout degradng the core safety characterstcs Practcally any type of Pu vector and Pu/MA composton rato can be accepted n the core Dfferent fuel forms (oxde or dense fuels) can be used, accordng to the objectve Fast reactors should be conceved wthn close cycle strateges, n order to maxmze benefts wth respect to sustanablty and waste mnmzaton

28 Conclusons (2) Fuel cycle ssues are crucal n order to assess the feasblty and the economy of a specfc strategy: Fuel reprocessng wth very small losses n the TRU recovery s mandatory (e.g. 99.9% recovery of any TRU sotope) Buld-up of hgher mass actndes (Cm, Bk, Cf sotopes) can be a heavy burden at fuel handlng, fuel fabrcaton etc., wth a potental mpact on reactor avalablty and fuel cycle optmzaton. Ths should be nvestgated n practcal applcatons. Multrecycle can hardly be avoded: any once-through approach wll be lmted by the maxmum achevable fuel burn-up Molten salt systems wth fast neutron spectrum and on-lne fuel processng or other moble fuel concepts (not dscussed here) could offer extra gans n terms of potental fuel cycle smplfcatons and t could stll be worthwhle to (re)-explore them

29 Back-up

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

31 Thermal and fast reactors: SFR desgned before 1990s Source: A. Walter and A. Reynolds, Fast Breeder Reactors Fertle blankets surround the core to ncrease BR

32 V m PWR wth dfferent moderator-to-fuel ratos Vm/Vf = 3 (overmoderated) = 2 (standard) = 1.1 (hgh CR PWR) FR V f Rato of capture-to-fsson average cross sectons for dfferent types of spectra U U Np Pu Pu Pu Pu Pu Am Am Am Cm Cm Cm Cm