1 / 39

By M.Khoroshev, IAEA/INPRO on behalf of the Benchmark team :

First & Second INPRO Benchmark studies to validate capabilities of different nuclear energy modelling tools to perform analysis of open and closed nuclear fuel cycle for the developing nuclear energy system. By M.Khoroshev, IAEA/INPRO on behalf of the Benchmark team :

talon
Télécharger la présentation

By M.Khoroshev, IAEA/INPRO on behalf of the Benchmark team :

An Image/Link below is provided (as is) to download presentation Download Policy: Content on the Website is provided to you AS IS for your information and personal use and may not be sold / licensed / shared on other websites without getting consent from its author. Content is provided to you AS IS for your information and personal use only. Download presentation by click this link. While downloading, if for some reason you are not able to download a presentation, the publisher may have deleted the file from their server. During download, if you can't get a presentation, the file might be deleted by the publisher.

E N D

Presentation Transcript

  1. First & Second INPRO Benchmark studiesto validate capabilities of different nuclear energy modelling tools to perform analysis of open and closed nuclear fuel cycle for the developing nuclear energy system By M.Khoroshev, IAEA/INPRO on behalf of the Benchmark team: A.Jalal, PESS/IAEA, G.Fesenko, E.Fedorova for MESSAGE M. Ceyhan, NEFW IAEA for VISTA V. Tsibulskiy, RRC KI, Russia, D. Mohapatra, IGCAR, India, for DESAE. A.Vasile, CEA, France for COSI L.Durpel, LISTO, Belgium, D.Wade, ANL, USA, for DANESS

  2. General objective of theINPRO benchmark study • To compare different computer codes that can be used for INPRO assessment studies by analyzing a given nuclear energy scenario, in particular: • To provide reliable materials flow codes to the INPRO Toolbox. • To compare the tools, provide Assessors with the results of validation, which could be used for verification of their own tools.

  3. Specific objectives and scope First Benchmark study • To define capabilities of different models to perform material flow analysis for the selected nuclear energy system.

  4. Participants • MESSAGE: Mr. Ahmed Jalal, PESS IAEA, Ms. G.Fesenko* Ms. Elena Fedorova* (contracted by PESS) • VISTA: Mr. Mehmet Ceyhan, NEFW IAEA • DESAE: Mr. Victor Tsibulskiy, RRC KI, Russia Mr. Dina Mohapatra, IGCAR, India • COSI: Mr. Alfredo Vasile, CEA, France • DANESS Mr. Van Den Durpel, LISTO, Belgium, Mr. David Wade, ANL, USA

  5. Benchmark definition A hypothetical case/scenario: • Present standard PWRs using UOX and MOX fuels • Sodium cooled Fast Reactors (FR) using MOX fuels with Plutonium recycling only. • The proposed scenario, to be calculated on a 120 years period

  6. Fisrt INPRO Benchmark Study

  7. Input data Nuclear energy scenario: definition of power demand (time dependant total installed capacity of NE system, GWe): Description of nuclear fuel cycle models including material flowcharts Reactor databases were provided and agreed among the participants • Reactor characteristics for different types of reactors used in the study included: • General characteristics including reloading scheme; • isotopic composition of initial fuel loading; • isotopic composition of equilibrium state reloading; • isotopic composition of discharged fuel; • other required information (e.g., neutron-physics data); • cooling time; • reprocessing time. • Characteristics of various fuel cycle facilities used in the study: • Assumed limited capacities of spent fuel reprocessing plants; • Included time legs in different stages of NFC.

  8. Table 2. Data compilation for the benchmark study  Depleted U [1]Values corresponding to Pu compositions given in Table 3.

  9. Assumptions All codes except VISTA currently do not perform in-core fuel isotopic compositional calculations. Therefore the input data were based on tabled fresh/spent fuel compositions prepared by COSI.

  10. Assumptions (example) • DANESS: discrete introduction/withdrawal of NPPs. Linear rate simulated as a step-function, with each step at least equal to one unit-size NPP. • DANESS does include the isotopic evolution out-of-core and this for a total of 68 isotopes and two fission-product groups, i.e. short-lived fission products SLFP (half-life < 30 years) and long-lived fission products (half-life > 30 years) LLFP. This isotopic evolution out-of-core simulates the decay-chain for these 70 isotopes allowing to take into account, for instance, build-up of 241Am due to decay of 241Pu in storage.

  11. Assumptions (example) • MESSAGE: the initial stock of spent fuel will be used even after year 10 and based on a First-in First-out principle, the discharged spent fuel from PWR UOx, after necessary cooling, will be reprocessed after the initial stock of spent fuel is finished. • MESSAGE: PWR UOx plants with 60 GW capacity already exist at year-0 and that their initial cores have already been loaded. The PWR MOx plants will come on-line as envisaged and their initial core has to be provided, but from some initial stock not mentioned in the case specifications. • MESSAGE restricted the computations to 60 years only to get more precise material accounting, taking into account radioactive decay of various elements. • A subsequent 60 year simulation would permit evaluation of the full 120 years specified. MESSAGE can have 60 distinct time steps, each of one or more years. For more precise accounting of radio-isotopes, annual calculations are necessary. • The reprocessing of MOx fuel is not considered. The plutonium storage capacity is calculated in terms of fissile Pu only. Other isotopes of Pu are also present in this storage, but are not reported here as the “Case” suggests only to account for fissile Pu.

  12. Fuel cycle chart in the first INPRO Benchmark study

  13. Additional assumptions • PWR park is at equilibrium at t = 0. • The given “Fabrication time” includes previous reprocessing delay when applicable. • Reprocessing plants working at their maximum capacity if fuel to be reprocessed available.PWR and FR MOX fuel compositions correspond to an “average Pu during the corresponding utilization period” previously calculated by COSI/CESAR. • Reactors life is 60 years. • It is assumed that all the electricity generation is delivered to the grid (400 TWhe per year). • Time periods needed for mining, conversion, enrichment, interim storage and conditioning of spent fuel or HLW are not considered.

  14. RESULTS Template for presentation of results

  15. RESULTS

  16. RESULTS

  17. RESULTS

  18. RESULTS

  19. RESULTS

  20. RESULTS

  21. Second INPRO Benchmark Study

  22. Second INPRO Benchmark Study New features to be studied compared to the 1st Benchmark: • All existing reactors – real data (capacity, etc.) • Three reference scenarios of global NE development: through 1990 – 2050 - 2100 • New innovative reactors: • FBR (BR 1.05, 1.6) • FBR (BR 1.05, 1.6) with Th blankets • HTGR in Th fuel cycle • Thorium fuel cycle

  23. 1st Scenario • Open Fuel cycle: • 2050 – 1000GW, 2000GW • 2100 3000GW, 5000GW Reactors shares !!!!: PWR BWR PHWR RBMK AGR WWER • Burnup increasing stepwise

  24. 2nd scenario • Closed Fuel cycle: • 2050 – 1000GW, 2050 – 2000GW • 2100 3000GW, 2100 – 5000GW Reactors shares !!!!: PWR BWR PHWR AGR WWER with U5 From 2030 FBR 1.05 or 1.6, with Pu HTGR with U5 Pu from SNF from LWR is used in the core of FBR NFC facilities reprocessing SNF are introduced starting from 2020 Commissioning of FBR with Pu core starts from 2030 FBR with: top/bottom blankets with U238 side blankets with Th OR side blankets with U238 Surplus U233 will be used in thermal reactors All Pu to be used in FBR (will Pu be repeatedly used in FBR???) By surplus of Pu PWR with MOX is introduced (see 1st Benchmark) • Burnup increasing stepwise (e.g. 5 steps by 2100 from 35 to 70GWd/tHM

  25. 3rd scenario • 3rd scenario: the same for closed fuel cycle, but without increasing burnup. • We need to compare closed fuel cycle with and without increased burnup. It may appear not beneficial to increase burnup because it will require higher consumption of fissile material. Instead the necessary fissille material can be obtained more effectively through recycling.

  26. Assumptions For the purpose of this study, below assumptions are made in order to compare the different capabilities and characteristics of the codes: • ·The study will be based on the actual fleet of commercial nuclear power plants in the world; • ·The scenario starts from 1990 and ends in 2100; • ·The input values for the past operational history of the nuclear power plants are based on the atual values reported by the IAEA member states to the different databases of the IAEA such as PRIS, to the extent possible; • ·The future projection is based on the experts opinions; • ·It is assumed that the existing commercial nuclear power plants can be grouped into 7 main types based on their nuclear characteristics. They are PWR, BWR, PHWR, RBMK, AGR, GCR and WWER. • ·It is assumed that there will be new type (FBR) of nuclear power plant in the period of scenario calculation; • ·There will be no reprocessing of MOX fuel; • ·There will be no processing of second generation FBR fuel; • ·There will be no capacity limitations on any of the fuel cycle facilities; • ·There will be no process losses in the fuel cycle facilities;

  27. MOX fuel will only be used in PWR and BWR for the calculation period. Scenario 1 is the Low MOX (LM) case and represents the existing MOX use ratios in the current reactors and then decrease to zero in 2035. Scenario 2 is the High MOX case (HM) and represents the current MOX use ratio will be kept constant at 3% until the end of scenario period.

  28. In combination with power, MOX use and Reprocessing scenarios there will be five different scenarios for the study. These are: ·Scenario (1) : OT-LP-LM-LR (No FBR) ·Scenario (2)a : RC-LP-LM-HR (FBR) ·Scenario (2)b : RC-LP-HM-HR (FBR) ·Scenario (3)a : RC-HP-LM-HR (FBR) ·Scenario (3)b : RC-HP-HM-HR (FBR)

  29. Average discharge burnups are given for existing reactor types in below chart. FBR discharge burnups are constant and 136 GWd/t for fuel region, 15 GWd/t for axial blanket and 25 GWd/t for radial blanket.

  30. Initial enrichment is defined as the initial fissile material content in the fresh fuel for Uranium fuels. Initial U235 amounts for existing reactor types are given in chart.

  31. Total Plutonium content in MOX fuel is defined as the percentage of total plutonium in the fresh MOX fuel. It is calculated based on the fissile material content of uranium fuels and MOX fuels. Below chart gives the total Pu content in fresh MOX fuel for LWRs. For fresh FBR fuel the total Pu content is 25% and constant for scenario period. Pu vector in fresh FBR fuel is Pu238 4.00% Pu239 48.88% Pu240 32.39% Pu241 2.98% Pu242 11.75% Pu Vector in fresh LWR MOX fuel Pu238 13.65% Pu239 60.10% Pu240 21.63% Pu241 11.70% Pu242 5.2%

  32. Load factor is defined as the ratio of the power produced in a reactor in a year to the power ot be produced in case the reactor operats at full power for 365 days and 24 hours in a year. Past data is taken from PRIS database and the future is projected by the experts.

  33. Thermal efficiency is defined as the ratio of the net electricity output to the total thermal power produced in the reactor.

  34. Enrichment tails assay is defined as the amount of U235 left in the depleted uranium after the enrichment process.

  35. Reactor_Fuel Type Fuel Residence Time in Core (Year) Cooling Time (Year) Reprocessing Time (Year) Power Density (kW/kg) PWR 4 6 1 37.5 BWR 4 6 1 25.9 PHWR 1 18.8 RBMK 1 15.75 AGR 3 6 1 10.9 GCR 1 1 1 3.3 WWER 4 6 1 45.8 FBR Core 5 2 1 16.0 FBR Axial 5 2 1 1.78 FBR Radial 8 2 1 1.16

  36. Output Parameters Below parameters should be calculated annually or at least each 10 years: Natural Uranium consumption (annual and Integral=total) • UF6 Conversion requirements • Enrichment SWU requirements and Depleted Uranium Accumulation • Fuel Fabrication requirements for Uranium and MOX fuels • Spent Fuel Discharged every year from the reactors • Spent Fuel going to reprocessing route (to the cooling) • Spent Fuel going to the interim storage • Spent fuel reprocessed (after cooling) • Fissile and Total Pu in discharged and stored spent fuel (this will be annual and ignores the decay during storage of the spent fuel) • Separated total Pu from reprocessed fuel (taking into account the decay during cooling and reprocessing) • Total Minor Actinides (Np, Am and Cm) in Discharged, Stored and HLW from reprocessing (this excludes the amounts in spent fuel waiting for the cooling) (amount in HLW takes into account the decay during the cooling and reprocessing) • Investments’ requirements • Operational costs per kWh • Amounts of Fission products for dumping (per year) • Decay heat

More Related