experience in neutronic/thermal-hydraulic coupling in...

Experience in Neutronic/Thermal-hydraulic

Coupling in Ciemat

Miriam Vazquez (Ciemat)

Francisco Martín-Fuertes (Ciemat)

Aleksandar Ivanov (INR-KIT)

2nd Serpent International Users Group Meeting,

Madrid 2012

Outline

1. Introduction

2. Coupling scheme

3. BWR pin benchmark

4. Other applications

5. Conclusions

2nd Serpent International Users Group

Meeting, Madrid, September 20, 2012 2

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:


Madrid, September 20, 2012 3



Neutronic TH Applied to Who XS temperature dependence

MCNP STAFAS (subchannel)

HPLWR FA (Waata, 2005) Pseudo material

MCNP STAR-CD (CFD) 3x3 PWR fuel pins

(Seker et al., 2007)

5K temperature interval

MCNP COBRA-EN 4 PWR FA (Capellan et al., 2009)

Pseudo materials

MCNP COBRA-TF PWR FA (Sanchez and Al-Harmy, 2009)

Psuedo materials

MCNP ATHAS (subchannel)

CANDU-SCWR FA

(Shan et al., 2010)

Pseudo materials

MCNP5 SUBCHANFLOW LWR pin and FA

(Ivanov et al., 2011,2012)

Pseudo materials

MCNP THERMO(subchannel)

1/8 PWR core

(Kotlyar et al.,2011)

Polynomial expansion

MCNP and TRIPOLI4

SUBCHANFLOW FLICA4

3x3 BWR pins

(Hoogenboom et al., 2010)

Pseudo materials

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.

• keff 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.



COBRA-IV Update for Fast Reactors

Applications



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.



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.



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)



Cross Section Handling


Madrid, September 20, 2012

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.

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 3.1.1 cross section libraries compiled at CIEMAT each 100 K

21

12

12 1 w=w

TT

TT=w

10

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.

ε<T

TT

i

ii

1

1



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.



Geometrical model



Single fuel with reflective radial surfaces.

Diferences in the coupling scheme

MCNP-

SUBCHANFLOW MCNP-COBRA

Initial conditions Converged TH solution from

KENO/SUBCHANFLOW Averaged TH values

Convergence criterion ε=max|ΔTi|<0.15% Δk<σ Adjust N to satisfy this criterion

max|ΔTi|<ε ε=max σqi=0.3%

Temperature dependence of fuel cross section

Using pseudo-materials 50K interval


Themal scattering data Interpolation between data files

No interpolation. Use the closest library available.



Model comparison

MCNP-

SUBCHANFLOW MCNP-COBRA

Number of iterations 14 13

Histories per cycle/ Active cycles

100000/350 500000/350

Fuel temperature Using pseudo-materials 50K interval


Coolant and cladding temperature

Constant Closest available XS library

Two phase correlation Chexal-Lellouche drift flux model

Armand model

Sub-cooled boiling Bowring correlation No sub-cooled boiling model

Fuel conductivity λ=f(T) λ=a+b*T+C*T2 Fits to f(T)



Power and k∞ during iterations



Coolant density



Fuel temperature



2% higher peak temperature

Relative Power

12 % greater peak power predicted by MCNP-COBRA due to

the higher density in the upper part.



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%.



Power density evolution in the central pin of

the ESFR FA.



Power density in the lower part



Power density in the upper part



CR movement with TH feedback in the

ESFR core to achieve criticality



CR insertion length (cm)

Keff (σ pcm)

0 1.01353 (3)

37.673 0.99844 (3)

34.9133 0.99977 (3)

34.5172 0.99998 (3)

Movement of the 24

Control and Shutdown

rods (CSD)

Linear power per FA in the upper axial level (10th)

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



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)





Thank you for your attention!!

References

• [1] C.L. Waata, Coupled Neutronics/Thermal-hydraulica Analysis of a High-Performance Light-

Water Reactor Fuel Assembly, FZKA 7233, Forschungszentrum Karlsruhe, 2005.

• [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.

15–19.

• [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, 2009.

• [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, 2009.

• [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) 58–65.

• [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.

22.



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,

2011.

• [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, 2012.

• [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) 3777–3786.

• [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, 2012.



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