1. 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

2. 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

3. 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

4. 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

5. 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

6. 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

7. 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)

8. 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)

9. 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

10. Output: Material Flow Results
Sample
Sample

11. Output: Isotopic Composition Results
Sample

12. 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

13. 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

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

15. 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

16. Benchmarking Exercises (2)
MESSAGE
Case (see Figure)
Results: (e.g. for )

17. 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:

18. 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 | | T: (+43-1) |  F: (+43-1) |