Numerical Simulation of Molten Salt Reactor Experiment ... - CiteSeerX

Thermal-Hydraulic Simulation of the Molten Salt Reactor Experiment with the SIMMER Code S. Wang, A. Rineiski, W. Maschek, M. Flad (D.T.I. GmbH) Forschungszentrum Karlsruhe, Institute for Nuclear and Energy Technologies Postfach 3640, D-76021 Karlsruhe, Germany

Introduction The Molten Salt Reactors (MSRs) are currently studied as nuclear waste burners and Th/U thermal breeders. To simulate transient and accidents in the MSRs, the SIMMER [1] code is under development. Recent benchmarks have shown that effects of fuel circulation (with respect to effective number of delayed neutrons emitted in the core) for an experimental MSR (MSRE) of ORNL (that operated during 1960’s [2-4]) can be simulated by SIMMER [5], but minor deviations between experimental and calculation results are observed (see Fig.1). These deviations may relate to some simplifications in modeling. In particular, the salt flow profile was assumed constant at different salt velocities while simulating the transport of delayed neutron precursors. 21

modeling:pump up experiment:pump up modeling:pump down experiment: pump down

Control Rod position (in)

20 19 18 17 16 15 14

0

10

20

30 40 Time (s)

50

60

70

Fig. 1 Experimental and modeling (SIMMER) results related to the effect of fuel circulation In this paper a new 2D thermal-hydraulics model for MSRE is investigated for nominal and close to zero-power conditions. The results show that the flow profile may change appreciably due to reactor power and salt flow rate variations, therefore a tighter coupling between neutronics and thermal-hydraulics parts of SIMMER with respect to the delayed neutron precursor transport modeling may be needed.

MSRE Model Geometrical details of the MSRE reactor vessel and internals are shown in Fig. 2 . The core is made up of 586 graphite bars, 2 inch in square (5.08 × 5.08 cm2) and 64 inch (1.6637 m) tall, exposed directly to the fuel salt, which flows in passages machined into the faces of the bars. The core diameter is about 54 inch (1.4 m). There are lower and upper plena under and above the core. The SIMMER model in RZ consists of 5 × 12 thermo-hydraulic meshes. It is shown in Fig. 3. The neutronics and material properties modeling is described elsewhere [5].

Reactor Outlet Outlet

~65.5 in.

~1.6 in.

0.2134

Inlet

Effective Core Height: ~79 in.

~8.5 in.

0.3423

Salt Graphite Matrix Diameter ~54 in.

Inlet

0.1774

Upper Plenum

Core α Molten Salt = 0.2578 α Graphite = 0.7422

0.13

~5.5 in.

Salt

Lower Plenum 0.7097 (m)

Fig. 2 MSRE vessel and internals

0.04

Fig. 3 SIMMER model

Variations of the graphite temperature may induce significant reactivity changes, therefore an accurate modeling of the both graphite and salt temperatures is important. A graphite bar is represented by two nodes: a surface node (temperature = Tsurf) and an interior node (temperature = Tinter). The thickness of surface nodes is defined as the thermal penetration lengths, 2δsurf , by considering the transient thermal response of the surface nodes. It is determined from the (SIMMER input parameter) structure time constants, τsurf, as:

2δ surf = ψ

τ surf ⋅ k graphite , ρ graphiteC graphite

ψ =2 3,

(1)

where kgraphite, ρgraphite, Cgraphite are the graphite thermal conductivity, density, and thermal capacity.

Numerical Results Some experimental and calculated MSRE parameters are presented in Table I. To get these values, SIMMER started with an approximate initial temperature distribution and run until the steady-state conditions were reached. Therefore, the reactor power computed by SIMMER is slightly less (by about 0.68%) than the experimental value. Table I Comparison of MERE and SIMMER-III Calculation Reactor Power (MW) Salt Inlet Temperature (°C) Salt Outlet Temperature (°C) Fuel Temperature of Cell (1,6) (°C) Graphite Surface Node Temperature of Cell (1,6) (°C) Graphite Interior Node Temperature of Cell (1,6) (°C)

MSRE 7.34 633.7 654.1 ----------

SIMMER-III 7.29 633.7 653.17 647.76 663.96 668.29

In Fig. 3, 4 the calculated temperature and velocity distributions of the fuel are given for nominal conditions. Fig. 4 shows an axial fuel temperature distribution at R=0.08 m. The fuel temperature reaches the highest value of 656 °C at the upper boundary of the graphite. In the upper plenum, the higher and lower temperature salt volumes are mixed together, therefore the salt outlet temperature is slightly smaller than the maximum value. In Figs. 5 and 6 the fuel velocity and temperature distributions are shown at a close to zero reactor power of 208 kW. The flow profile is not stable: it varies with time (e.g. from time of 650 s to time of 690 s as shown below) while the salt flow and reactor power remain constant. That may influence the build-up and transport of delayed neutron precursors at close to zeropower regimes. Therefore, we plan in the future to establsih a tighter coupling between the neutronics and thermal-hydraulics models of SIMMER. It may hopefully explain the mentioned deviations between computation and experimental results on the precursor transport and its effect on the neutron (reactivity) balance in the core.

Fig. 4. 2D fuel temperature (K) and velocity (m/s) distribution at nominal power

Fig.5 Axial temperature distribution of the fuel at R=0.08 m (first radial mesh)

Fig. 6. 2D (RZ) fuel temperature and velocity distribution at power of 208 kW, time=650 s

Fig. 7 2D (RZ) fuel temperature and velocity distribution at power of 208 kW, time=690 s

Conclusions A 2D SIMMER model was developed for the MSRE. The experimental results for nominal conditions can be accurately reproduced. The calculations show that the flow profile may oscillate appreciably at close to zero-power conditions. Therefore a tighter coupling between neutronics and thermal-hydraulics parts of SIMMER may be needed to better simulate the delayed neutron precursor transport for which experiments were performed at close to zeropower conditions. AcknowledgementThis work has been supported by the EU Program MOST, contract No: FIKW-CT-2001-00096.

References 1. 2. 3. 4. 5.

S. KONDO, K. MORITA, Y. TOBITA, N. SHIRAKAWA, SIMMER-III: An Advanced Computer Program for LMFBR Severe Accident Analysis, ANP'92, Tokyo, Japan,1992. B.E. Prince, J.R. Engel, S.J. Ball, P.N. Haubenreich, T.W. Kerlin: Zero Power Physics Experiments on the Molten Salt Reactor Experiment, ORNL-4233, 1968. M.W. Rosenthal, R.B. Briggs, and P.R. Kasten: Molten Salt Reactors Program Semiannual Progress Report, ORNL-4396, 1969. H.G. MacPherson: The Molten Salt Reactor Adventure, Nuclear Science & Engineering, Vol. 90, pp.374-380, 1985. S. Wang, M. Flad, A. Rineiski, and W. Maschek: Extension of the SIMMER-III Code for Analysing Molten Salt Reactors, Jahrestagung Kerntechnik, Berlin, 20-22, Mai, 2003.