A GFR benchmark: Comparison of transient analysis codes based on the ETDR concept

• Published in: Progress in nuclear energy (New series) 2008, Vol.50 (1), p.37-51
• Main Authors: Bubelis, E., Castelliti, D., Coddington, P., Dor, I., Fouillet, C., de Geus, E., Marshall, T.D., van Rooijen, W., Schikorr, M., Stainsby, R.
• Format: Article
• Language: English
• Subjects: Applied sciences; Benchmark; Energy; Energy. Thermal use of fuels; Exact sciences and technology; Fission nuclear power plants; Fuels; Gas-cooled fast reactor; Generation IV; Heat transfer; Installations for energy generation and conversion: thermal and electrical energy; Nuclear fuels; System codes comparison; Theoretical studies. Data and constants. Metering; Transient analysis
• ISSN: 0149-1970
• DOI: 10.1016/j.pnucene.2007.11.090

In preparation for the transient analysis of the Generation IV gas fast reactor (GFR) and experimental technology demonstration reactor (ETDR) designs, a transient benchmark exercise (in the frame of the Generation IV GFR project) was performed to compare the capabilities and limitations of the different code systems to analyse these new reactor concepts. The benchmark was based on the ETDR concept and was performed in three phases, i.e. a loss-of-flow (LOF) transient with reactor scram for phases 1 and 2 and a small-break loss-of-coolant accident for phase 3. The organizations which participated in the benchmark were AREVA, France; CEA, France; NRG and TUD, The Netherlands; AMEC, United Kingdom; INL, USA (phase 1); CIRTEN, Italy; JRC-IE EURATOM and PSI, Switzerland. Phase 1 of the benchmark was performed in a “blind” manner in that all participants were provided with the same information, but were free to make their own judgement on the use of this information. Following a review of the results from phase 1, the conclusion was made that the agreement between all the partners was acceptable, but better agreement was to be preferred. The reason for the discrepancies was identified, namely: different heat transfer correlations for the main and DHR heat exchangers and reactor core, modelling of the vessel wall thermal capacity, core and vessel pressure drop calculations and flow resistance in the DHR helium and water loops. The phase 1 transient was repeated as (phase 2) but with more ‘strict boundary conditions’ to identify which of the above gave the largest contribution to the differences in the phase 1 submissions. The second phase results were much improved. However, to demonstrate the capabilities of the codes at low pressures a SB-LOCA transient with reactor scram (phase 3) was defined to model the ETDR DHR system during the depressurization from 70 bar to the 3 bar containment pressure. The phases 1 and 2 results showed that the use of different heat transfer correlations for the main heat exchanger does not significantly affect the core coolant temperatures, while the modelling of ETDR heat structures has a major impact on the core inlet temperatures. Even though three participants used the RELAP5 code and two participants used CATHARE, there was limited consistency between the results for each specific code. This was particularly true for phase 1, where the wide spread in the results came from different input and boundary assumptions by the codes us