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Health, Safety & Environment

DYN3D ­ Advanced Reactor Simulations in 3D

a report by

U l r i c h R o h d e , U l r i c h G r u n d m a n n and S ö r e n K l i e m

Institute of Safety Research, Forschungszentrum Dresden-Rossendorf

The Computer Code DYN3D In nuclear reactors, transient processes with a significant reactivity insertion may occur, leading to an increase and 3D re-distribution of the fission power density in the reactor core in a very short period of time. These transient processes can be induced by local changes in the temperature or density of the materials (fuel, moderator, absorber) or by the movement of the control rods. To ensure safety by means of a proper reactor design, and to estimate and minimise the consequences of hypothetical accidents, accident scenarios have to be modelled by adequate simulation tools. A best-estimate tool for simulating the dynamics of water-cooled reactors is the computer code DYN3D, which was developed at the FZD Institute of Safety Research. It comprises a 3D neutron kinetics model, a thermal­hydraulic calculation module and a fuel rod model. The neutron kinetics model is based on the solution of the 3D twogroup neutron diffusion equation by nodal expansion methods. By means of the thermal­hydraulic model, the density and temperature of the coolant are obtained under one- and two-phase flow conditions. By solving the corresponding heat conduction equations,

fuel and cladding temperatures are calculated. These parameters are important for thermal­hydraulic feedback on the neutron kinetics, as well as for assessing the safety of the reactor. The nuclear cross-sections in the neutron diffusion equations depend not only on the feedback parameters, but also on the

To ensure safety by means of a proper reactor design, and to estimate and minimise the consequences of hypothetical accidents, accident scenarios have to be modelled by adequate simulation tools.

nuclide concentrations and the neutron spectrum, which change with the burn-up of the fuel. These dependencies must be taken into account, providing corresponding cross-section libraries. Figure 1 shows the interaction of the neutron kinetics, thermal­

Ulrich Rohde is Head of the Safety Analysis Department of the Institute of Safety Research, Forschungszentrum Dresden-Rossendorf. Previously, he worked as a research scientist in the Central Institute for Nuclear Research in Rossendorf, near Dresden. His main interests are reactor physics, neutron kinetics, thermal hydraulics and fluid dynamics.

hydraulics and neutron cross-sections. The DYN3D core model was linked to thermo-hydraulic system codes1 to provide boundary conditions for the core, such as coolant inlet temperature and pressure and coolant mass flow-rate distribution. System codes model the thermo-fluid dynamics of the primary and secondary circuit, including all major components of the plant.

Ulrich Grundmann, prior to retirement in 2007, worked in the Department of Accident Analysis at the Institute of Safety Research, Forschungszentrum Dresden-Rossendorf. He previously worked at the Central Institute for Nuclear Research in Rossendorf in the field of reactor dynamics, and in the Department of Reactor Physics and System Engineering of the Paul Scherrer Institute in Switzerland. His main task was the development and validation of the computer code DYN3D for transient analysis of thermal reactors and the coupling of DYN3D with the plant analysis code ATHLET. Until his retirement, he was a Board Member of the section of Reactor Physics and Calculation of the German Nuclear Society.

DYN3D is undergoing continuous development with respect to the improvement of physical models and numerical methods. Recently, a multigroup approach was implemented in order to improve the description of spectral effects, which are increasingly important for mixed-oxide reactor core loadings of light-water reactors, but also for innovative reactor concepts. A neutron transport method was developed for DYN3D to overcome the limitations of the diffusion approximation. Within this approach, even a pin-wise calculation of the power distribution is offered. This was recognised internationally, receiving large amounts of interest.2,3 The extended DYN3D version was integrated into the European Platform for NUclear REactor SIMulations (NURESIM). An interface to the platform was developed, which is based on the NURESIM software environment SALOME. This allows DYN3D to be linked to

Sören Kliem is a senior research scientist in the Institute of Safety Research of Forschungszentrum Dresden-Rossendorf, Germany. His background includes the development, validation and application of coupled neutron kinetics and thermal hydraulics system codes. He is also concerned with the experimental and analytical investigation of coolant mixing phenomena.

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© TOUCH BRIEFINGS 2007

DYN3D ­ Advanced Reactor Simulations in 3D

other components of the platform, e.g. advanced fluid dynamics simulation tools (computational fluid dynamics (CFD) codes). In 2006, eight research projects related directly to the improvement, validation and application of DYN3D were running or completed. Moreover, licences for DYN3D, including commercial ones, were provided to 12 users in Germany and other European countries. Within a PhD project, a version of DYN3D for dynamic studies of molten salt reactors (MSRs), which belong to the `Generation IV' concepts, was developed.4 Analyses were performed for a number of specific MSR transient processes, demonstrating the inherent safety of this reactor.

Cross-section of nodes Neutron kinetics · 2-neutron groups · Diffusion theory · 3-dimensional · Nodal methods · Version for quadratic assemblies hexagonal assemblies Thermal hydraulics · 1D 4-equations th. model for two phase flow · Radial heat conduction in fuel clad · Heat transfer model · Safety parameters · Boron mixing methods

Figure 1: Scheme of Interaction of the Different Models in the DYN3D Code

DYN3D 3D core model Steady state and transient

Nodal powers

Analysis of a Hypothetical Boron Dilution Scenario A reactivity-initiated transient process can be induced by the perturbation of the boron concentration in the core of a pressurised water reactor. Boron is added to the reactor coolant as a neutron absorber, compensating for the excess of reactivity in the fresh core at the beginning of the fuel cycle. Due to different mechanisms or system failures, for instance as a consequence of a small break loss of coolant accident, slugs of lowborated water can accumulate in the primary cooling system. The boron concentration in the reactor core results from turbulent During the start-up of the coolant circulation ­ after refilling the primary circuit with emergency cooling water ­ or by switching on the first main coolant pump (MCP), these slugs are transported into mixing along the flow path in the reactor vessel. In this case, mixing the de-borated condensate with borated water is the only mitigative mechanism to prevent re-criticality of the shut-down reactor and

· Library of nuclear group constants · Burn-up distribution Calculation of cross-sections Fuel temperatures Coolant densities Coolant temperatures Boron concentrations

the reactor core, causing a reactivity insertion by decreasing the amount of the neutron absorber.

DYN3D ­ Advanced Reactor Simulations in 3D

Figure 2: Boron Concentration Distribution over the Core Inlet Cross-section at the Moment of Minimum Boron Content

up of the first main coolant pump for German Konvoi-type reactors was performed using DYN3D. This analysis aimed at showing the integrity of the fuel rods even under such extreme conditions. Analysing the boron dilution scenarios is a challenge because the distribution of the boron concentration, depending on space and time, is crucial for the induced reactivity insertion. The boron distribution is obtained by modelling the mixing of the de-borated slug with the ambient coolant in the reactor by means of CFD. Figure 2 shows the distribution of the boron concentration

The investigations have shown large margins if a realistic approach to the modelling of the coolant mixing is applied.

at the reactor core inlet when the boron content is at its

Figure 3: Reactor Power (Relative to Nominal) During a Boron Dilution Transient Process

10,000 20 m**3 36 m**3 8,000

minimum. DYN3D transient calculations were performed for different slug volumes. Considering the bounding scenario with a slug size of 36m3, the reactor quickly becomes super-critical (maximum reactivity of about 2$), leading to a very fast power excursion with a peak power of about 7,000MW (about twice nominal power). Figure 3 shows the time behaviour of the reactor power. However, the power excursion is limited by the very effective Doppler feedback of the fuel temperature.

Core power (MW)

6,000

4,000

The width of the power peak in time is only about 25ms. Therefore, the integral energy release in the power peak is limited. The fuel

2,000

temperature remains below 800ºC, i.e. far below melting point. There is local coolant boiling in the hottest fuel assemblies, but

12 14 16 18 Time (s) 20 22 24

0

no cladding superheating occurs. It can be concluded from the analyses that even in the case of a conservatively large volume of un-borated water, the safety criteria are met and the integrity of the core is not endangered. The investigations have shown large margins if a realistic approach to the modelling of the coolant mixing is applied. Project Partners CEA (France); Vattenfall (Sweden/Germany); Fortum Nuclear Engineering (Finland); TÜV Süddeutschland (Germany); VGB PowerTech (Germany); AREVA (France/Germany); Nuclear Research

impermissible power excursion.5 CFD methods are applied, allowing the time-dependent boron concentration to be ascertained at each position of the fuel element.6 Experiments performed at the Rossendorf coolant mixing test facility ROCOM ­ a test facility for investigating the mixing of the coolant in a linear scale of 1:5 ­ were used for the validation of the code.7 The analysis of a boron dilution scenario with an inadvertent start-

Institute Rez (Czech Republic).

1.

2.

Kliem S, Kozmenkov Y, Höhne T, Rohde U, Analyses of the V1000CT-1 benchmark with the DYN3D/ATHLET and DYN3D/RELAP coupled code systems including a coolant mixing model validated against CFD calculations, Progress in Nuclear Energy, 2006;48:830­48. Beckert C, Grundmann U, A nodal expansion method for solving the multigroup SP3 equations in the reactor code DYN3D, M&C+SNA 2007, Monterey, US.

3.

4.

5.

Beckert C, Grundmann U, Development and verification of a nodal approach for solving the multigroup SP3 equations, Annals of Nuclear Energy, 2007, doi:10.1016/j.anucene.2007.05.014. Krepel J, Grundmann U, Rohde U, Weiss F-P, DYN3D-MSR spatial dynamics code for molten salt reactors, Annals of Nuclear Energy, 2007;34:449­62. Kliem S, Rohde U, Weiss F-P, Core response of a PWR to a slug of under-borated water, Nuclear Engineering and Design,

6.

7.

2004;230:121­32. Höhne T, Kliem S, Bieder U, Modeling of a buoyancy-driven flow experiment at the ROCOM test facility using the CFDcodes CFX-5 and TRIO_U, Nuclear Engineering and Design, 2006;236:1309­25. Rohde U, Kliem S, Höhne T, Fluid mixing and flow distribution in the reactor circuit ­ Measurement data base, Nuclear Engineering and Design, 2005;235:421­43.

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NUCLEAR ENERGY REVIEW 2007

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