Improvements of Nuclear Fuel Cycle Simulation System (NFCSS) at IAEA Ki Seob Sim* (IAEA), R. Yoshioka (International Thorium Molten-Salt Forum, Japan), H. Hayashi (IAEA – Retired, Japan), T.S.G. Rethinaraj (National Institute of Advanced Studies, India) 3rd fuel cycle workshop, Paris, 9-11 July 2018

Who We Are NE NS NA TC SG Div. of NEFW RRS Uranium resources and production WTS Nuclear power reactor fuel NFCMS Spent fuel storage Spent fuel recycling

Contents of Presentation Introduction Overview of chronological developments Overall features Premise for development Implementation in NFCSS Improvements (selected) Thorium fuel cycle Decay heat calculation Radiotoxicity calculation Benchmarking exercises Conclusions

Introduction - History of NFCSS, 1997-2007 Developed as a tool to support the International symposium on Nuclear Fuel Cycle and Reactor Strategies: Adjusting to New Realities, Vienna, 3-6 June 1997 Converted to the web-based system; available to MSs in 2005 Documented in IAEA-TECDOC-1535, published in 2007

Introduction - History of NFCSS, Since 2007 Significant improvements in implementation of NFCSS since the publication of IAEA-TECDOC-1535 in 2007, which includes: Cross-section data and verification Relation between specific power and neutron flux Effect of operation mode Initial core loading and final core discharge Concern on Am-241 cross-section Thorium fuel cycle analysis routines Decay heat calculation methodology Radiotoxicity calculation methodology Extended application to innovative reactors, e.g. INPRO, FR/FBR A new TECDOC is under development

Overall Features – Premise for Development Answer to strategic questions related to fuel cycle year by year over a long period of time (e.g. hundreds years): what are the amounts of demanded resources at each stage of the front-end fuel cycle? what are the amounts of used fuel, actinide nuclides and high level waste to be stored? what is the impact of introducing recycling of used fuel on the amounts of resource savings and waste minimization? Fast running & Easy to use

Implementation in NFCSS Minimally specified nuclides: 14 nuclides (for UO2, MOX) + additional 4 (for thorium) Built-in burnup models; no need of using reactor physics codes Minimally required inputs (see Figure) Built-in nuclear parameters (X-sections, initial enrichment-discharge burnup relation, spec P-n flux relation, Pu vectors)

Simplified Transmutation Model for UO2 Assumptions: U-235, U-238 as initial nuclides Short-lived nuclides (<8d) are skipped, i.e. U-237 (7d), Np-238 (2d), Pu-243 (5h), Am-242 (16h), Am-243 (10h), Am-244 (26m) Long-lived nuclides (>400y) as stable ones, i.e. Am-241 (432y) – no further transmutation Transmutation terminated for certain nuclides 14 nuclides only (see Figure)

Assumptions – Cont’d Effect of operation mode: Am-241 100% Load Factor is assumed in burnup model Impact on discharged amount of Pu-241 and thus Am-241 Ignore because several % errors after 1-2 decades Am-241 Two reactions, leading to Am-242 (0.8836), Am-242m (0.1164) Fixed branch fraction, valid only for thermal reactors

Output: Material Flow Results Sample Sample

Output: Isotopic Composition Results Sample

Thorium Fuel Cycle Burnup chains LEU (U-235, U-238), U-233, Pu-239 in ThO2 as the initial nuclides Short-lived (<8d) nuclides, Th-233, Pa-234, U-237 are skipped Two paths from Pa-233 to U-234 are considered: Pa-233 (decay)  U-233 (capture)  U-234 Pa-233 (capture)  [Pa-234; skipped]  U-234 Long-lives (>400y) nuclides as stable ones; no decay for Th-232, U-234, U-235, U-236, Np-237 Only 7 nuclides are considered (see Figure). Th-232 Pa-233 U-233 U-234 Np-237 U-236 U-235

Decay Heat Calculation ORIGEN2 analysis: Different trends between U and Th fuels, due to possibly Library U-233 concentration Burnup Fast neutron spectrum Only considers 18 nuclides and not all FPs  Use of a lookup table based on ORIGEN2 analysis and interpolation as a function of e.g. burnup and specific year

Radiotoxicity Calculation A lookup table based on ORIGEN2 analysis and interpolation as a function of e.g. burnup and specific year UO2 ThO2

Benchmarking Exercises (1) Comparison with independent solutions HIMMEL Cases of PWR-UO2, PWR-MOX, WWER-UO2, PHWR-UO2, LMFR (MOX, axial blanket) Reasonably agree for discharged fuel compositions (~1.5% difference); systematic differences for MAs COSAC Case 1 – PWR fleets with UO2 Case 2 – Mixed PWR and FBR (see Figure) Good agreement for annual fresh fuel consumptions

Benchmarking Exercises (2) MESSAGE Case (see Figure) Results: (e.g. for 2040-2150)

Conclusions With the features of fast running and easy to use in addition to reliable fuel cycle assessments, NFCSS has been well recognized as a public tool that serves the interests of a wide range of professionals in academia, research and policy arena in Member States. Since 2007 (when the first publication on NFCSS was made), a number of improvements have been implemented in NFCSS based on users’ feedback and requirements. Maintaining and updating NFCSS continue. Any issue or user support, please contact: k.s.sim@iaea.org

Thank you! Mr Ki Seob SIM, Ph.D. | Nuclear Fuel Engineering Specialist |Nuclear Fuel Cycle and Materials Section | Division of Nuclear Fuel Cycle and Waste Technology| Department of Nuclear Energy | International Atomic Energy Agency | Vienna International Centre, PO Box 100, 1400 Vienna, Austria | Email: k.s.sim@iaea.org | T: (+43-1) 2600-21921 |  F: (+43-1) 26007 |