Technical Meeting on WP Regional integration of R&D on sustainable nuclear fuel cycles Part 2 Assessment of regional nuclear fuel cycle options

Transcription

1 Technical Meeting on WP Regional integration of R&D on sustainable nuclear fuel cycles Part 2 Assessment of regional nuclear fuel cycle options Rita Plukienė CENTER FOR PHYSICAL SCIENCES AND TECHNOLOGY Savanorių 231, LT Vilnius, Lithuania Vilnius

2 According to the agenda Assessment of regional nuclear fuel cycle options SNF management scenarios for Baltic region; Identification of the optimal scheme of SNF management on regional level (open, one-way reproc., transmutation Scenario); Milestones, deliverables, visits; Aspects important to other WP2 Sub-tasks.

3 According to the agenda Assessment of regional nuclear fuel cycle options SNF management scenarios for Baltic region; Identification of the optimal scheme of SNF management on regional level (open, one-way reproc., transmutation Scenario); Milestones, deliverables, visits; Aspects important to other WP2 Sub-tasks.

4 SNF management strategies Open fuel cycle - Geological disposal of SNF. Partially closed fuel cycle MOX fuel. Closed fuel cycle - Partitioning and transmutation.

5 SNF and high-level radioactive waste Spent nuclear fuel (SNF) from NPPs: t SNF by % - U, 1% - Pu, ir 0.1% - Np, Am, Cm 2.4% Fission products ( 137 Cs, 129 I, 99 Tc, 90 Sr) Nuclear weapons: t U 250 t 239 Pu. Radionuclide toxicity characterizes alpha nuclides (dangerous in the possible inhalation case) Radiotoxicity, Sv/t Np 238 Pu Time, y 239 Pu 240 Pu 241 Pu 242 Pu 241 Am 243 Am 244 Cm nat U all FP alltru

6 SNF management strategies Deep geological isolation concepts are : mined repositories in three geologic media salt, clay/shale rocks, and crystalline (e.g., granitic) rocks and deep borehole disposal in crystalline rocks. Weakness: uncertainties of the safety and impregnability of the storage over hundreds of thousands of years SNF and nuclear weapons grade Pu partitioning and transmutation in the thermal or epithermal/fast neutron flux. LWR CANDU GT-MHR AMSTER Fast breeder (FBR) Accelerator driven nuclear systems (ADS) Fusion-Fission hybrid system Strength of PU and MA transmutation: conversion to short-lived isotopes, easer surface-storage with little/no proliferation value fuel for transmutation systems

7 Neutron-induced fission cross sections SNF transmutation in the neutron flux Looking for a new efficient and economically viable transmutation system the main consideration is on the neutron flux and spectrum characteristics of a particular system Fast neutrons thermal neutrons Life-time of a certain nucleus in the system flux: T1/2 ~ ln2 cross-section depends on neutron energy spectrum and neutron fluxes of the system: ( E) ( E) de ( E) de

8 According to the agenda Assessment of regional nuclear fuel cycle options SNF management scenarios for Baltic region; Identification of the optimal scheme of SNF management on regional level (open, one-way reproc., transmutation Scenario); Milestones, deliverables, visits; Aspects important to other WP2 Sub-tasks.

9 Management of SNF and radioactive nuclear waste in Lithuania Ignalina NPP (Lithuania): 2580 kt SNF ~22 kt Pu and ~2 kt MA Temporary storage for 50 y. SNF and RNW management strategy depends on nuclide composition of waste. Before decision making, the radionuclide vector should be established. For this: Modeling Direct measurements Indirect evaluations are performed.

10 Modeling tools available at CPST (FTMC) Microscopic cross sections and neutron fluxes, criticality and neutron balance: MCNP6 (LANL) MCNPX SCALE 6.1 (ORNL) Burnup calculations, time-dependent formulation, creation/destruction and decay: ORIGEN CINDER

11 Numerical modeling procedure Initial composition Code coupling 1 group x-sections MCNP, Neutron transport ORIGEN, Nuclide evolution Updated composition

12 Modeling of nuclide generation in the reactor Initial fuel composition Neutron cross sections, neutron fluxes, criticality and neutron balance 1 v t X, E,, t X, E,, t X, E,, t X, E, t Neutron cross sections data libraries t, texs EEX texdedt t,,,,,,,,,,, E Nuclides evolution in the reactor: burnup, creation/destruction and decay dn dt i ji j, i N j i ri N i N i No Last cycle? Yes Nuclide activity Nuclear data libraries: ENDF (ENDF/B-VII), JEF, JENDL...

13 Applications Modeling of nuclear systems, nuclide composition of spent nuclear fuel analysis: RBMK, GTMHR, Fusion-Fission Hybrid system, etc. Neutron activation analysis in the reactor construction materials: Graphite, Fuel channels, CPS rods, biological shielding Criticality safety: FA in the storage pools of INPP, CASTOR and CONSTOR casks

14 6.86 m Validation for RBMK SNF modelling (1) (3D description of fuel assembly) Axial region h A, A2 B, B2 h, cm (H 2 O) g/cm C, C2 D, D % enrichment 235U fuel explicitly modelled inner and outer fuel pellet rings vertically varying coolant density along the fuel channel R. Plukienė, A. Plukis, V. Remeikis and D. Ridikas, Lith. J. of Physics 45 4(2005).

15 Validation for RBMK SNF modelling (2) ( 239 Pu - comparison of calculated and [1] experimental data) M, kg/t U Pu inner ring outer ring M CN P [1] M CN P [1] A A 2 2 B 3 B2 4 C 5 C2 6 D D 2 7 & Pu average deviation ~5% Exp. points Ring inner outer inner outer inner outer outer outer h, m Burnup, MWd/kg 1 E.В. Бурлаков et all., Нуклидный состав образцов отработавшего топлива РБМК-1000, ИАЭ-6266/3 (РНЦ- КИ, Москва, 2003).

16 Validation for RBMK SNF modelling (3) ( 241 Pu and 244 Cm - comparison of calculated and [1] experimental data) M, kg/t U Pu inner ring M CN P [1] A B 3 C 5 D Burnup, MWd/kg outer ring M CN P [1] A 2 2 B2 4 C2 6 D 2 7 & 8 M, g/t U Cm 0.01 inner ring outer ring M CN P [1] M CN P [1] 1E-3 A A 2 2 B 3 B2 4 1E-4 C 5 C2 6 D D 2 7 & 8 1E Burnup, MWd/kg 241 Pu average deviation ~10% 1 E.В. Бурлаков et all., Нуклидный состав образцов отработавшего топлива РБМК-1000, ИАЭ-6266/3 (РНЦ- КИ, Москва, 2003). 244 Cm average deviation ~50% (at low burnup 244 Cm has small influence on the neutron and is not so important)

17 CASTOR cask modelling with MCNP5 (3D geometry description) 3.43m 4.4m CASTOR container No half-assemblies located at specified places of CASTOR cask using hexagonal lattice option A. Plukis et al., Lith. J. of Physics Vol.46,No.3,pp (2006)

18 786 cm 33.2 cm cm 457 cm Graphite activation analysis for INPP solid radioactive waste characterization CARBOWASTE (EU-FP7 project) RBMK-1500 FC, graphite and construction materials modelling scheme with 14 fuel assemblies and 2 control rods extracted CPS channel filled by water and graphite displacer. Continuous power 4200MW 2,4% U with Er, average fuel burnup 12,4MWd/kg R inserted CPS rod with neutron absorbent Dy 2 TiO 5 10 R a b R. Plukienė et al., Progr. Nuc. Science and Tech. 2 (2011).

19 3D GT-MHR geometry description Basic GT-MHR parameters Table I: Basic GT-MHR parameters. Power, MW th 600 Active core size: - height, cm - area, m Active core volume, m 3 : 91.9 Graphite mass in reactor, t: Averaged temperature, o C: - active core - inner reflector - side, top and bottom reflectors central reflector side reflectors Core contains : fuel particles burnable poison particles borated zone active core 310m R. Plukiene and D. Ridikas (2003), Annals of Nuc. Energy, 30/15, pp R. Plukienė et al., (2003) Lith. J. of Physics, Vol. 43, No. 6, pp Pu burnup up to 90% 200m

20 Fusion fission hybrid system simplified neutronic model Molten salt blanket with TRU (LiF-BeF 2 -(HN)F 4 ) HN (Pu + MA) Garphite reflector Molten salt liquid wall Fusion system D + T= 4He + n ( 14MeV) External neutron source (1MW ~ n/s ) System parameters Zone name Radius (cm) R i - R i+1 Material composition Fusion device Void Liquid wall Flibe ( 6 Li 0.1% in Li ) Metallic wall SS316, 50% Graphite graphite Blanket Region 1 Region 2 Region 3 Region Flibe = 2g/cm 3 TRU = g/cm 3 6 Li content 0.6% in Li Reflector graphite Envelope SS316, 50%

21 Fusion-Fission hybrid system (TRU) Initial conditions molten salt loaded with TRU vector (Pu and MA from SNF) replenished only by MAs (~3.1kg/day) working at constant fission power (3 GW th ) Zone n, n/scm 2 n E n >0,1 MeV % region 1 2, ,2 region 2 1, ,4 region 3 1, ,4 region 4 7, ,6 regions 1-4 1, ,9 Fast n flux (E n >0,1MeV) ~10 22 n/cm 2 / year n (E) is constant Neutron flux, n/(cm 2 * s* letargy) (Neutron flux in the system: n ir n (E)) 239 Pu 242 Pu 240 Pu 19 F 6 Li 1 region 2 region 3 region 4 region average in all regions E, MeV Transmutation speed 1.13 TRU/y Most effective for 237 Np, 239 Pu, 241 Am up to 90%, 240 Pu up to 76%, 241 Pu - ~60%, 243 Am - ~40%, accumulation of 244 Cm.

22 According to the agenda Assessment of regional nuclear fuel cycle options SNF management scenarios for Baltic region; Identification of the optimal scheme of SNF management on regional level (open, one-way reproc., transmutation Scenario); Milestones, deliverables, visits; Aspects important to other WP2 Sub-tasks.

23 Identification of the optimal scheme of SNF management on regional level (1) Reactor accidents as Three Mile Island (Pennsylvania 1979), Chernobyl (Ukraine 1986) and Fukushima (Japan 2011) undermined public confidence in reactor technologies. The regional disposal option would be one option to reduce negative aspects. If combined with the SNF reprocessing on regional level the cycle (with MOX) results in better use of fissile material, and partial optimisation of disposal. The introduction of transmutation of MA could further reduce overall risk, increase energy throughput and optimise final disposal. Lithuania (RBMK-1500), Finland (VVER-440, BWR) Sweden (BWR, PWR).

24 Identification of the optimal scheme of SNF management on regional level (2) I reference scenario - conventional open fuel cycle without reprocessing will be assessed by evaluating SNF of different types of reactors (RBMK, VVER, PWR and BWR) existing (and planed GenIII+ NPP) in the region. II reference scenario - estimates of SNF to be reprocessed (Mixed Oxide I reference scenario (U+Pu)) and disposed of in geological facilities. III reference scenario - SNF reprocessing & transmutation of minor actinides and disposal in geological facilities. Comparative analysis, conclusions and recommendations of closed fuel cycle perspectives in the region.

25 I reference scenario Estimation of radionuclide activity and toxicity (m 3 air) during the long term decay (ORIGN-ARP) Pu-239 Cm-244

26 II reference scenario SNF reprocessing and MOX fuel (case U with : 7 % Pu from SNF (70 % - Pu 239) from spent UOX fuel). Comparison standard and MOX fuel burning within 1000 d. typical burnup for each reactor type after 5 years of cooling. Estimation of radionuclide activity and toxicity (m 3 air) during the long term decay (ORIGN- ARP)

27 III reference scenario HTR case (GT-MHR) central reflector active core Actinides transmutation in GenIV reactors or transmutation systems and subsequent geological disposal option. Analysis of better fitting system for transmutation (thermal and epithermal/fast neutron spectrum, etc.). Assessment what benefit will be in terms of waste amount/decay time and generated energy in case of scenario III side reflectors borated zone Fusion-fission hybrid system with molten salt LiF-BeF 2 -(HN)F 4

28 Identification of the optimal scheme of SNF management on regional level (3) Estimates of SNF, HLW to be reprocessed and disposed off in geological facilities using IAEA NFCSS Material Flow approach for standart ractors. Comparative analysis, conclusions and recommendations of closed fuel cycle perspectives in the region. About.cshtml

29 According to the agenda Assessment of regional nuclear fuel cycle options SNF management scenarios for Baltic region; Identification of the optimal scheme of SNF management on regional level (open, one-way reproc., tramut. Scenario); Milestones, deliverables, visits; Aspects important to other WP2 Sub-tasks.

30 Milestones, deliverables, visits Milestone MS14: Case study for open cycle without reprocessing (FTMC) MS26: Case study for cycle with SNF reprocessing/transmutation (FTMC) Due Date (in months) DELIVERABLES: Report on Regional competences for R&D on sustainable nuclear fuel cycles (24 th Project Month, TARTU). Analysis of the open and closed NFC options with NFC simulator codes and more dedicated codes for SNF composition, radio-toxicity analysis. several peer-reviewed papers (e.g., Progress in nuclear energy etc.)? VISITS: Meetings of WP2 (Regional integration of R&D sustainable NFC) Meetings during the International conferences (e.g., NuMat, GLOBAL etc.)?

31 According to the agenda Assessment of regional nuclear fuel cycle options SNF management scenarios for Baltic region; Identification of the optimal scheme of SNF management on regional level (open, one-way reproc., tramut. Scenario); Milestones, deliverables, visits; Aspects important to other WP2 Sub-tasks.

32 Aspects important to other WP2 Sub-tasks (1) Information about spent nuclear fuel and nuclear fuel cycle options on regional level: Lithuania (Ignalina NPP 2-RBMK-1500, new Visaginas NPP ABWR?), Sweden (Ringhals NPP 1- BWR, 3-PWR, Oskarshamn NPP 3 BWR, Forsmark NPP 3-BWR). Finland? (Loviisa NPP 2-VVER-440, Olkiluoto NPP 2- BWR, new 1 EPWR? Hanhikivi VVER-1200 PWR? ) Poland? The greatest care has to be taken in collecting the actual data for historical operation to the extent possible.

33 Aspects important to other WP2 Sub-tasks (2) Existing assumptions for NFC scenarios and relevant parameters for the estimation of nuclear fuel cycle material and service requirements: different NFC models different reactor types and fuel types U and Pu recycling in PWR, BWR, VVER? NFC calculation period ( )? MOX fuel from SNF (initial enrichment, burnup, cooling time) Pu and TRU separation (initial composition, dedicated transmutation system)?

34 Aspects important to other WP2 Sub-tasks (3) Different aspects of analysis of different NFC options with varios NFC simulation codes: Initial fuel composition Nuclear capacity (Gwe) Reprocessing Ratio (%) MOX/TRU fuel use ratio (%) Average Discharge Burnup (MWd/kg) Final SNF composition to be disposed. Radiotoxicity analysis?

35 Aspects important to other WP2 Sub-tasks (4) Evaluation of geological disposal and reprocessing options for Baltic region: Two potential candidates for burial of high-level waste in Sweden Oskarshamn and Östhammar. Interim storage at Ignalina NPP site in Lithuania Onkalo spent nuclear fuel repository for Finland.? reprocessing options outside of Baltic region e.g. EU (France)?

36 Thank you for you attention!!! CENTER FOR PHYSICAL SCIENCES AND TECHNOLOGY Institute of Physics Department of Nuclear Research