Experience in Neutronic/Thermal-hydraulic Coupling in Ciemat

Transcription

1 Madrid 2012 Experience in Neutronic/Thermal-hydraulic Coupling in Ciemat Miriam Vazquez (Ciemat) Francisco Martín-Fuertes (Ciemat) Aleksandar Ivanov (INR-KIT)

2 Outline 1. Introduction 2. Coupling scheme 3. BWR pin benchmark 4. Other applications 5. Conclusions 2nd Serpent International Users Group Meeting, 2

3 Introduction This work is framed in nuclear reactor modern analyses: High accuracy requirements Trend towards multi physics simulations Coupled n transport (MC)/ thermal-hydraulics developed worldwide: TH feedback Exact geometry description Energy point-wise cross sections, angular dependency MC (+TH) codes are usually the reference for deterministic codes MCNP coupled with subchannel codes at pin-by-pin level, but none for large geometries: 3

4 Neutronic TH Applied to Who XS temperature dependence MCNP STAFAS (subchannel) MCNP STAR-CD (CFD) 3x3 PWR fuel pins HPLWR FA (Waata, 2005) Pseudo material (Seker et al., 2007) MCNP COBRA-EN 4 PWR FA (Capellan et al., 2009) MCNP COBRA-TF PWR FA (Sanchez and Al- Harmy, 2009) MCNP ATHAS (subchannel) CANDU- SCWR FA MCNP5 SUBCHANFLOW LWR pin and FA MCNP MCNP and TRIPOLI4 THERMO(subch annel) SUBCHANFLOW FLICA4 1/8 PWR core 3x3 BWR pins (Shan et al., 2010) (Ivanov et al., 2011,2012) (Kotlyar et al.,2011) (Hoogenboom et al., 2010) 5K temperature interval Pseudo materials Psuedo materials Pseudo materials Pseudo materials Polynomial expansion Pseudo materials 2nd Serpent International Users Group Meeting, 4

5 MCNPX & COBRA MCNPX is a stochastic multi-particle Monte Carlo transport code. Detailed geometry description, point-wise cross sections. Neutron flux estimations for critical and also subcritical systems, as it includes spallation models. k eff in critical systems with the KCODE transport mode. COBRA-IV is a deterministic code: Fluid conservation equations: flow & enthalpy distribution in rod bundles or cores. Heat transfer equation is used to calculate fuel temperature. Steady state and transients calculations. Different coolants: water, liquid metals or gas. 5

6 COBRA-IV Update for Fast Reactors Applications 6

7 MCNPX/ COBRA Coupling The coupling scheme is based on the iterative use of both codes, steady state condition: MCNPX computes cell power distribution with a flux tally (F4) multiplied by fission cross section. COBRA calculates fuel, cladding and coolant temperature and coolant density in each cell. Data are exchanged by means of external file sharing. 7 7

8 Computational Method Driver program written in C++ I/O interface COBRA I/O interface MCNP Requirements: Input with the name of the files and cross section libraries. MCNP and COBRA input decks with the same discretization. Fuel, cladding and coolant of each pin/fa and axial level is described with a cell and have a different material. 8

9 Approach for Temperature Effects (i) Thermal effects in MCNP: Density Cell density (coolant, ) Temperature Material cross section σ(fuel) * MCNP adjusts thermal scattering cross section with the TMP card. * Doppler broadening of absorption and fission cross section: the library should be compiled at the suitable temperature. (ii) Thermomechanical effects (fast reactors): Cell dimension Surfaces. We are working on the dependence of reactor geometry with temperature to obtain the real reactivity in hot steady state and to calculate expansion reactivity coefficient (important to license fast reactors under unprotected transients) 9

10 Cross Section Handling Alternatives to update the cross section libraries: Generate explicit cross section libraries with NJOY Time consuming and memory problems. On-The-Fly Doppler broadening Is not yet included in the MCNP release. Approach using pseudo materials (our choice) Accuracy depends on the interpolation interval. The new material composition is a weighted combination of nuclide X cross section at lower T1 and higher T2 temperature. w T T 1 2 = w1 = 1 w2 T2 T1 In fast reactors, this approach provides deviations 10 times lower with an interpolation interval of 200 K than with a library 100 K different. JEFF cross section libraries compiled at CIEMAT each 100 K 10

11 Convergence Since MCNP adopts a stochastic method, each computed power has an uncertainty. The convergence criterion must be more robust than the uncertainty to guarantee the convergence It is calculated in the first step as the maximum relative uncertainty among all the MCNP tally results. After the first step the temperature convergence is checked in each cell. An internal exercise showed that the temperature uncertainty due to the power uncertainty is 5 times larger in average at pin level and 2 times larger at core level. Ti Ti T i 1 1 < ε 11

12 BWR pin benchmark problem Benchmark data was obtained from: A.Ivanov, V.Sanchez, J.E. Hoogenboom, Single BWR pin coupled Monte Carlo -Thermal Hydraulic analysis, PHYSOR 2012, Knoxville, April 15-20, 2012 Comparison of results between MCNP-SUBCHANFLOW and MCNP-COBRAIV coupled codes. 12

13 Geometrical model Single fuel with reflective radial surfaces. 13

14 Diferences in the coupling scheme Initial conditions Convergence criterion Temperature dependence of fuel cross section Themal scattering data MCNP- SUBCHANFLOW Converged TH solution from KENO/SUBCHANFLOW ε=max ΔTi <0.15% Δk<σ Adjust N to satisfy this criterion Using pseudo-materials 50K interval Interpolation between data files MCNP-COBRA Averaged TH values max ΔTi <ε ε=max σqi=0.3% Using pseudo-materials 100K interval No interpolation. Use the closest library available. 14

15 Model comparison MCNP- SUBCHANFLOW Number of iterations MCNP-COBRA Histories per cycle/ Active cycles Fuel temperature Coolant and cladding temperature Two phase correlation / /350 Using pseudo-materials 50K interval Constant Chexal-Lellouche drift flux model Using pseudo-materials 100K interval Closest available XS library Armand model Sub-cooled boiling Bowring correlation No sub-cooled boiling model Fuel conductivity λ=f(t) λ=a+b*t+c*t 2 Fits to f(t) 15

16 Power and k during iterations 16

17 Coolant density 17

18 Fuel temperature 2% higher peak temperature 18

19 Relative Power 12 % greater peak power predicted by MCNP-COBRA due to the higher density in the upper part. 19

20 Analysis of SFR FA and core The coupled code has been used for fast reactors applications with a detailed geometry at pin level in the ESFR FA and at channel level in the ESFR and MYRRHA cores. In the full core MCNP model, FAs have been grouped in rings of similar power to complete the calculation in reasonable amount of time (3-4 h using 128 processors) and not exceeding our machine memory limitations (2GB). Coupling calculations with the working horse ESFR: ESFR fuel assembly with 271 pins, 30 axial levels. It converges in 3 iterations Δkeff=182 pcm. Max ΔPower=0.6% ESFR core with 11 rings, 10 axial levels. It converges in 3 iterations Δkeff=182 pcm. Max Δpower=3%. 2nd Serpent International Users Group Meeting, 20

21 Power density evolution in the central pin of the ESFR FA. 21

22 Power density in the lower part 22

23 Power density in the upper part 23

24 CR movement with TH feedback in the ESFR core to achieve criticality Movement of the 24 Control and Shutdown rods (CSD) CR insertion length (cm) K eff (σ pcm) (3) (3) (3) (3) Linear power per FA in the upper axial level (10th) 24

25 Conclusions A computational tool has been developed for N-MC/ TH coupled analysis Static conditions, 3D temp map effects on keff. Flexible geometry modeling. Can be used for full core & subchannel analysis. It has been compared with others MC/TH coupled codes. The discrepancies are because of the TH model. Quick and robust behavior for fast reactors. A method to accelerate convergence is needed for LWR applications..memory limitations in full core applications with burned fuel due to the use of pseudo-materials 2nd Serpent International Users Group Meeting, 25

26 Acknowledgments - Haileyesus Tsige-Tamirat and Luca Ammirabilie (JRC-IET) - The co-authors of the BWR pin benchmark problem: V. Sanchez (KIT) and Prof. Hoogenboom (Delf University of Technology) 2nd Serpent International Users Group Meeting, 26

27 Thank you for your attention!! 27

28 References [1] C.L. Waata, Coupled Neutronics/Thermal-hydraulica Analysis of a High-Performance Light- Water Reactor Fuel Assembly, FZKA 7233, Forschungszentrum Karlsruhe, [2] V. Seker, J.W. Thomas, T.J. Downar, others, Reactor Physics Simulations with Coupled Monte Carlo Calculation and Computational Fluid Dynamics, in: Proceedings of the International Conference on Emerging Nuclear Energy Systems (ICENES-2007), Istambul, Turkey, 2007: pp [3] N. Capellan, J. Wilson, S. David, O. Meplan, J. Brizi, 3D coupling of Monte Carlo neutronics and thermal-hydraulic calculations as a simulation tool for innovative reactor concepts, in: Proceedings of Global 2009, Paris, France, [4] V. Sanchez, A. Al-Hamry, Development of a Coupling Scheme Between MCNP and COBRA- TF for the Prediction of the Pin Power of a PWR Fuel Assembly, in: Proceedings of International Conference on Mathmatics, Computational Methods & Reactor Physics (M&C 2009), Saratoga Springs, New York, [5] J. Shan, W. Chen, B.W. Rhee, L.K.H. Leung, Coupled neutronics/thermal hydraulics analysis of CANDU SCWR fuel channel, Annals of Nuclear Energy. 37 (2010) [6] J.E. Hoogenboom, A. Ivanov, V. Sanchez, C. Diop, A flexible coupling scheme for Monte Carlo and thermal-hydraulics codes, in: International Conference on Mathematics and Computational Methods Applied to Nuclear Science and Engineering (M&C 2011), Rio de Janeiro, Brazil, 2011: p nd Serpent International Users Group Meeting, 28

29 References [7] A. Ivanov, V. Sanchez, U. Imke, Develpment of a Coupling Scheme between MCNP5 and Subchanflow for the Pin- and Fuel Assembly-Wise Simulation of LWR and Innovative Reactors, in: Proceedings of International Conference on Mathmatics, Computational Methods Applied to Nuclear Science and Engineering (M&C 2011), American Nuclear Society, Rio de Janeiro, Brazil, [8] A. Ivanov, V. Sanchez, U. Imke, Optimization of a Coupling Scheme between MCNP5 and Subchanflow for high Fidelity Modeling of LWR Reactors, in: Proceedings of International Conference on M, American Nuclear Society, Knoxville, Tennessee, USA, [9] D. Kotlyar, Y. Shaposhnik, E. Fridman, E. Shwageraus, Coupled neutronic thermo-hydraulic analysis of full PWR core with Monte-Carlo based BGCore system, Nuclear Engineering and Design. 241 (2011) [10] A. Ivanov, V. Sanchez, J.E. Hoogenboom, Single Pin BWR Benchmark Problem for Coupled Monte Carlo- Thermal Hydraulics Analysis, in: Proceedings of PHYSOR 2012, Knoxville, Tennessee, USA, nd Serpent International Users Group Meeting, 29