Benchmark Analysis of FFTF Loss of Flow Without Scram Test
Closed for Proposals
Project Type
Coordinated Research ProjectProject Code
I32011
CRP
2225
Approved Date
06.03.2018Project Status
ClosedStart Date
26.07.2018Expected End Date
31.12.2022Completed Date
27.06.2024Participating Countries
Switzerland, China, Germany, Spain, France, India, Italy, Japan, Republic of Korea, Kingdom of the Netherlands, Russian Federation, Sweden, United States of AmericaDescription
The IAEA Coordinated Research Project (CRP) will focus on benchmark analysis of one of the unprotected passive safety demonstration tests performed at the Fast Flux Test Facility (FFTF). The dynamics analysis of FFTF reactor core with complex reactivity feedback mechanisms and primary and secondary coolant loops using system codes will provide an excellent opportunity for validation of the physical and mathematical models and reactor simulation codes using actual experimental data.
Objectives
The overall objective of the CRP is to improve the Member States' analytical capabilities in the field of fast reactor simulation and design. A necessary condition towards achieving this objective is a wide international validation and qualification effort of the analysis methodology and codes currently employed in the fields of fast reactor neutronics, thermal hydraulics and plant dynamics to achieve enhanced safety. It will aim to enable Member States to make informed decisions on the development of new or advanced fast reactor designs, and to increase cooperation between Member States in achieving advances in fast reactor technology development through international collaborative R&D.The CRP will be implemented as a programmatic activity of the IAEA Project 1.1.5.3 “Advanced technology for fast and gas-cooled reactor” starting with the IAEA Program and Budget Cycle 2018 – 2019. The Project 1.1.5.3 has the objective, among others, to enable Member States to make informed decisions on the development of new or advanced fast reactor designs, and to increase cooperation between Member States in achieving advances in fast reactor technology development through international collaborative R&D. Given its aforementioned overall objective, the CRP clearly responds to the objectives of the IAEA Project 1.1.5.3.
Specific Objectives
Analyse Uncertainties and Results Qualification (Optional) Timing: Max Temperature, Final Shutdown Flowrates Key Temperatures, Natural Circulation Balance, and Others
Publish papers in journals and proceedings as well as a standard and evaluated benchmark report as an IAEA technical publication
Jointly develop a benchmark specification, analysis methods, including: Detailed Input Data on: FFTF Plant Layout; Core Layout; Detailed Description of Fuel/Reflector/Control Rod Assemblies plus Gas Expansion Modules; Description of Cycle 8C including Fuel Isotopic Composition and Pre-Transient Power History; Primary Coolant System including Pump, Intermediate Heat Exchanger, and Piping Specifications; Secondary Coolant System including Pump, Dump Heat Exchanger, and Piping Specifications; LOFWOS Test #13 Test Description including initial and transient boundary conditions; Instrumentation and Available Experimental Data Measured during LOFWOS Test #13. Physical Processes to be simulated in Benchmark: Primary and secondary loop fluid flow and heat transfer Core neutronics. Output Values and Data to be compared: Steady-State Core Physics (k-eff, Reactivity Feedbacks, Decay Heat, etc…); Steady-State Power Distributions; In LOFWOS Transient, History (vs. time) of: Reactor Power, Reactivity; Temperatures in Primary/Secondary Hot/Cold legs as well as within reactor core and at the outlet of the two fast response PIOTAs, etc…; Flow Rates.
Perform benchmark analysis by simulation of specific FFTF LOFWOS test using different codes, methods, and models in the participating institutions and jointly analyse the results of calculations, in particular: Evaluation of the key parameters that predict the transient behaviour of the reactor core and the primary coolant loop, including the reactor core flow rate and sodium coolant temperatures (against measurements obtained from the instrumented fuel assemblies) and intermediate heat exchanger temperatures; Calculation and benchmark comparison of key reactivity effects, including the Doppler feedback, sodium density, fuel axial and core radial expansion, and control-rod-driveline expansion. The modelling and analysis of the selected test will require steady-state conditions prior to the test, boundary conditions such as the primary pump speeds and dump heat exchanger sodium outlet temperatures, and core physics model of oxide core during the cycle in which the test was conducted.
Collect, evaluate and share experimental data obtained in FFTF during LOFWOS experiments
Impact
One of the primary lessons learned from the CRP was the importance of face-to-face interactions for a project of this size and scale. The project was initiated shortly before the COVID-19 pandemic, which necessitated converting the second and third meetings from in-person to virtual where large open discussions with numerous voices were more challenging. This was also proved true for the hybrid fourth and final meeting, for which approximately half of the participants were able to attend in person. At this meeting, more in-depth discussions took place but generally only among the in-person participants. These discussions occurred at the end of the project after most participants had concluded their modelling and simulation efforts. As a result, many of the discussions to investigate modelling successes and challenges and resolve discrepancies that participants would have benefitted from did not take place.
The FFTF CRP was one of the largest fast reactor CRPs ever conducted by the IAEA. Despite the limitations imposed by the COVID-19 pandemic, the CRP was an extremely valuable opportunity for validating fast reactor safety analysis codes, methods, and models. Participants in the project have gained confidence in their approaches to modelling and simulations while the CRP provided opportunities for the next generation of researchers and analysts to simulate a real world SFR transient scenario. The experiences and lessons learned that participants of this CRP have gained from analysing FFTF LOFWOS Test #13 will help them perform even more accurate safety analyses and continue to refine and improve SFR simulation tools in the future
Relevance
Relevant to the objectives of project 1000154 in as far as supporting MS interested in fast reactor development, licensing and deployment and specifically benchmarking software.
CRP Publications
Sapienza University of Rome - DIAEE
Nuclear Engineering and Design
2021
Thermal-hydraulic transient analysis of the FFTF LOFWOS Test #13
Argonne National Laboratory
FR22 Conference
2022
Simulation of FFTF Individual Reactivity Feedback Tests with SAS4A/SASSYS-1 Code
Nuclear and Industrial Engineering (NINE), Italy
HND2022 Conference
2022
Thermal-Hydraulics Analysis of the IAEA CRP FFTF LOFWOS Test #13
Argonne National Laboratory
NURETH-19 Confertence
2022
Impact of Assembly-to-Assembly Heat Transfer on the Simulation Results of FFTF LOFWOS Test
Helmholtz-Zentrum Dresden-Rossendorf (HZDR), Germany
Progress in Nuclear Energy
2024
Analysis of loss of flow without scram test in the FFTF reactor – Part I: Preparation of neutronics data
NRG, Netherlands
NURETH-20 Conference
2023
Thermal-hydraulic Analysis of the FFTF LOFWOS Transient with the SPECTRA Code