Ulrich Rohde

Coolant Mixing in a Pressurized Water Reactor: Deboration Transients, Steam-Line Breaks, and Emergency Core Cooling Injection

Abstract The reactor transient caused by a perturbation of boron concentration or coolant temperature at the inlet of a pressurized water reactor (PWR) depends on the mixing inside the reactor pressure vessel (RPV). Initial steep gradients are partially lessened by turbulent mixing with coolant from the unaffected loops and with the water inventory of the RPV. Nevertheless the assumption of an ideal mixing in the downcomer and the lower plenum of the reactor leads to unrealistically small reactivity inserts. The uncertainties between ideal mixing and total absence of mixing are too large to be acceptable for safety analyses. In reality, a partial mixing takes place. For realistic predictions it is necessary to study the mixing within the three-dimensional flow field in the complicated geometry of a PWR. For this purpose a 1:5 scaled model [the Rossendorf Coolant Mixing Model (ROCOM) facility] of the German PWR KONVOI was built. Compared to other experiments, the emphasis was put on extensive measuring instrumentation and a maximum of flexibility of the facility to cover as much as possible different test scenarios. The use of special electrode-mesh sensors together with a salt tracer technique provided distributions of the disturbance within downcomer and core entrance with a high resolution in space and time. Especially, the instrumentation of the downcomer gained valuable information about the mixing phenomena in detail. The obtained data were used to support code development and validation. Scenarios investigated are the following: (a) steady-state flow in multiple coolant loops with a temperature or boron concentration perturbation in one of the running loops, (b) transient flow situations with flow rates changing with time in one or more loops, such as pump startup scenarios with deborated slugs in one of the loops or onset of natural circulation after boiling-condenser-mode operation, and (c) gravity-driven flow caused by large density gradients, e.g., mixing of cold emergency core cooling (ECC) water entering the RPV through the ECC injection into the cold leg. The experimental results show an incomplete mixing with typical concentration and temperature distributions at the core inlet, which strongly depend on the boundary conditions. Computational fluid dynamics calculations were found to be in good agreement with the experiments.

Dynamics of Molten Salt Reactors

Abstract The dynamics of the molten salt reactor (MSR), one of the Generation IV International Forum concepts, was studied. The graphite-moderated channel-type MSR was selected for numerical simulation. MSR, a liquid-fueled reactor, has specific dynamics with two physical peculiarities: The delayed neutron precursors are drifted by the fuel flow, and the fission energy is released directly into the coolant. Presently, there are few accessible numerical codes appropriate for MSR simulation; therefore, the DYN1D-MSR and DYN3D-MSR codes were developed based on the light water reactor dynamics code DYN3D. These allow calculation of one-dimensional and full three-dimensional transient neutronics in combination with parallel channel-type thermal hydraulics. The codes were validated with experimental results of the Molten Salt Reactor Experiment from Oak Ridge National Laboratory and applied to several transients typical for a liquid fuel system. Those transients were initiated by reactivity insertion, by cold or overfueled slugs, by the fuel pump start-up or shutdown, or by the blockage of selected fuel channels. In these considered transients, the response of MSR is characterized by the immediate change of the fuel temperature relative to the temperature at that power level. This causes fast insertion of feedback reactivity, which is negative for power-related temperature increase. On the other hand, the graphite response is slower, and its feedback coefficient depends on the core size and geometry. The addition of erbium to the graphite can ensure negative feedback and inherent safety features also for big low leakage cores. The DYN1D-MSR and DYN3D-MSR codes have been shown to be effective tools for MSR dynamics studies. The MSR response to the majority of transients is considered acceptable within safety margins as long as the graphite feedback coefficient is negative. A transient that is possibly an exception is a local channel blockage.

The modeling of fuel rod behaviour under RIA conditions in the code DYN3D

Abstract A description of the fuel rod behaviour and heat transfer model used in the code DYN3D for nuclear reactor core dynamic simulations is given. Besides the solution of heat conduction equations in fuel and cladding, the model comprises a detailed description of heat transfer in the gas gap by conduction, radiation and fuel-cladding contact. The gas gap behaviour is modeled in a mechanistic way taking into account transient changes of the gas gap parameters based on given conditions for the initial state. Thermal, elastic and plastic deformations of fuel and cladding are taken into account within 1D approximation. A creeping law for time-dependent estimation of plastic deformations is implemented. Metal–water reaction of the cladding material in the high temperature region is considered. The cladding–coolant heat transfer regime map covers the region from one-phase liquid convection to dispersed flow with superheated steam. Special emphasis is put on taking into account the impact of thermodynamic non-equlibrium conditions on heat transfer. For the validation of the model, experiments on fuel rod behaviour during RIAs carried out in Russian and Japanese pulsed research reactors with shortened probes of fresh fuel rods are calculated. Comparisons between calculated and measured results are shown and discussed. It is shown, that the fuel rod behaviour is significantly influenced by plastic deformation of the cladding, post crisis heat transfer with sub-cooled liquid conditions and heat release from the metal–water reaction. Numerical studies concerning the fuel rod behaviour under RIA conditions in power reactors are reported on. It is demonstrated, that the fuel rod behaviour at high pressures and flow rates in power reactors is different from the behaviour under atmospheric pressure and stagnant flow conditions in the experiments. The mechanisms of fuel rod failure for fresh and burned fuel reported from the literature can be qualitatively reproduced by the DYN3D model. However, the model must be extended and improved for a proper description of burned fuel behaviour. The realistic simulation of the fuel rod behaviour is important not only under RIA conditions, but also for the analysis of operational transients. This has been shown in a calculation of an operational transient with power decrease after switching-off one from the two working feed water pumps in the NPP Balakovo (VVER-1000).

Analysis of the Boiling Water Reactor Turbine Trip Benchmark with the codes DYN3D and ATHLET/DYN3D

Abstract The OECD/NRC Boiling Water Reactor (BWR) Turbine Trip Benchmark was analyzed by the code DYN3D and the coupled code system ATHLET/DYN3D. For the exercise 2 benchmark calculations with given thermal-hydraulic boundary conditions of the core, the analyses were performed with the core model DYN3D. Concerning the modeling of the BWR core in the DYN3D code, several simplifications and their influence on the results were investigated. The standard calculations with DYN3D were performed with 764 coolant channels (one channel per fuel assembly), the assembly discontinuity factors (ADF), and the phase slip model of Molochnikov. Comparisons were performed with the results obtained by calculations with 33 thermal-hydraulic channels, without the ADF and with the slip model of Zuber and Findlay. It is shown that the influence on core-averaged values of the steady state and the transient is small. Considering local parameters, the influence of the ADF or the reduced number of coolant channels is not negligible. For the calculations of exercise 3, the DYN3D model validated during the exercise 2 calculations in combination with the ATHLET system model, developed at Gesellschaft für Anlagen- und Reaktorsicherheit for exercise 1, has been used. Calculations were performed for the basic scenario as well as for all specified extreme versions. They were carried out using a modified version of the external coupling of the codes, the “parallel” coupling. This coupling shows a stable performance at the low time step sizes necessary for an appropriate description of the feedback during the transient. The influence of assumed failures of different relevant safety systems on the plant and the core behavior was investigated in the calculations of the extreme scenarios. The calculations of exercises 2 and 3 contribute to the validation of DYN3D and ATHLET/DYN3D for BWR systems.

Comparative study of a boron dilution scenario in VVER reactors

Abstract Subsequent studies have identified many scenarios, which can lead to reactivity excursions due to boron dilution. The comparative study, presented in this paper, deals with the so-called “restart of the first reactor coolant pump’’ scenario and its reactor-dynamic consequences for both Russian designed VVER reactor types, VVER-440 and VVER-1000. The transient simulations were performed using the three-dimensional core dynamics code DYN3D. The DYN3D modeling features, including recent developments, as well as the cross-section methodology involved in these calculations, are described. The analyzed accident scenario is outlined together with the assumptions made. The results of core response in this boron dilution accident for both VVER reactors are compared within the ranges, determined by the two reactivity values of interest: the criticality limit and the reactivity initiated accident (RIA) limit.

On the use of a molten salt fast reactor to apply an idealized transmutation scenario for the nuclear phase out.

In the view of transmutation of transuranium (TRU) elements, molten salt fast reactors (MSFRs) offer certain advantages compared to solid fuelled reactor types like sodium cooled fast reactors (SFRs). In the first part these advantages are discussed in comparison with the SFR technology, and the research challenges are analyzed. In the second part cycle studies for the MSFR are given for different configurations – a core with U-238 fertile, a fertile free core, and a core with Th-232 as fertile material. For all cases, the transmutation potential is determined and efficient transmutation performance for the case with thorium as a fertile material as well as for the fertile free case is demonstrated and the individual advantages are discussed. The time evolution of different important isotopes is analyzed. In the third part a strategy for the optimization of the transmutation efficiency is developed. The final aim is dictated by the phase out decision of the German government, which requests to put the focus on the determination of the maximal transmutation efficiency and on an as much as possible reduced leftover of transuranium elements at the end of the reactor life. This minimal leftover is achieved by a two step procedure of a first transmuter operation phase followed by a second deep burning phase. There the U-233, which is bred in the blanket of the core consisting of thorium containing salt, is used as feed. It is demonstrated, that transmutation rates up to more than 90% can be achieved for all transuranium isotopes, while the production of undesired high elements like californium is very limited. Additionally, the adaptations needed for the simulation of a MSFR, and the used tool HELIOS 1.10 is described.

Two-Way Coupling Between the Reactor Dynamics Code DYN3D and the Fuel Performance Code TRANSURANUS at Assembly Level

In the last two decades the reactor dynamics code DYN3D was coupled to thermal hydraulics system codes, a sub-channel thermal hydraulics code and CFD codes. These earlier developed code systems allow modeling of the thermal hydraulics phenomena occurring during reactor transients and accidents in greater detail. Still these code systems lack a sufficiently sophisticated fuel behavior model, which is able i.e. to take into account the fission gas behavior during normal operation, off-normal conditions and transients. To our knowledge a two-way coupling to a fuel performance code hasn’t so far been reported in the open literature for calculating a full core with detailed and well validated fuel behavior models.A new two-way coupling approach between DYN3D and the fuel performance code TRANSURANUS is presented. In the coupling, DYN3D provides the time-dependent rod power and thermal hydraulics conditions to TRANSURANUS, which in turn transfers parameters like fuel temperature and cladding temperature back to DYN3D. The main part of the development is a general TRANSURANUS coupling interface that is applicable for linking of any other reactor dynamics codes, thermal hydraulics system codes and sub-channel codes to TRANSURANUS. Beside its generality, other features of this interface are the application at either fuel assembly or fuel rod level, one-way or two-way coupling, automatic switching from steady to transient conditions in TRANSURANUS (including update of the material properties etc.), writing of all TRANSURANUS output files and the possibility of manual pre- and post-calculations with TRANSURANUS in standalone mode. The TRANSURANUS code can be used in combination with this coupling interface in various scenarios: different fuel compositions in the reactor types BWR, PWR, VVER, HWR and FBR, considering time scales from milliseconds (i.e. RIA) over seconds/ minutes (i.e. LOCA) to years (i.e. normal operation) and thence different reactor states.Results of DYN3D-TRANSURANUS are shown for a control rod ejection transient in a German PWR. In particular, it appears that for all burn-up levels the two-way coupling approach systematically calculates higher maximum values for the node fuel enthalpy (max. difference of 46 J/g) and node centerline fuel temperature (max. difference of 181 K), compared to DYN3D standalone in best estimate calculations. These differences can be completely explained by the more detailed TRANSURANUS modeling of fuel thermal conductivity, radial power density profile and heat transfer in the gap. As known from fuel performance codes, the modeling of the heat transfer in the gap is sensitive and causes also larger differences in case of low burn-up.The numerical convergence for DYN3D-TRANSURANUS is quick and stable. The coupled code system can improve the assessment of safety criteria, at a reasonable computational cost with a CPU time of less than seven hours without parallelization.Copyright

Development and Verification of Dynamics Code for Molten Salt Reactors

To perform transient analysis for Molten Salt Reactors (MSR), the reactor dynamics code DYN3D developed in FZR was modified for MSR applications. The MSR as a liquid fuel system can serve as a thorium breeder and also as an actinide burner. The specifics of the reactor dynamics of MSR consist in the fact, that there is direct influence of the fuel velocity to the reactivity, which is caused by the delayed neutrons precursors drift. This drift causes the spread of delayed neutrons distribution to the non-core parts of primary circuit. This leads to a reactivity loss due to the fuel flow acceleration or to the reactivity increase in the case of deceleration. For the first analyses, a 1D modified version DYN1D-MSR of the code has been developed. By means of the DYN1D-MSR, several transients typical for the liquid fuel system were analyzed. Transients due to the overcooling of fuel at the core inlet, due to the reactivity insertion, and the fuel pump trip have been considered. The results of all transient studies have shown that the dynamic behavior of MSR is stable when the coefficients of thermal feedback are negative. For studying space-dependent effects like e.g. local blockages of fuel channels, a 3D code version DYN3D-MSR will be developed. The nodal expansion method used in DYN3D for hexagonal fuel element geometry of VVER can be applied considering MSR design with hexagonal graphite channels.Copyright

Influence of System and Neutron-Kinetic Parameter Variations on an Anticipated Transient Without SCRAM in a PWR

The complete failure of the reactor scram system upon request during an operational transient is called anticipated transient without scram (ATWS). According to the German regulatory guidelines, postulated ATWS events have to be analyzed with regard to their consequences on the safety of nuclear power plants. Since the course of ATWS transients is determined by a strong interaction of the neutron kinetics with the thermal hydraulics of the system, coupled 3D neutron kinetic/thermal hydraulic code systems are adequate tools for the analysis of such transients. The coupled code system DYN3D/ATHLET developed at FZD is applied to the analysis of an ATWS transient. The objective of the present work is to perform a best-estimate analysis with consequent use of a 3D neutron kinetic code (DYN3D) in combination with an advanced thermal hydraulic system code (ATHLET) together with a quantification of differences in the course and the results of transients, which arise from the uncertainties of thermal hydraulic and neutron-physical conditions. Typically, the complete failure of the main feed water supply is assumed to be the bounding ATWS event with regard to the maximum primary coolant pressure, which can be reached during the transient. The limitation of the coolant pressure is a precondition for the integrity of the primary circuit. The situation is aggravated if the main coolant pumps remain in operation. For this particular transient, the influence of different thermal hydraulic and neutron-physical conditions on the course of the transient was analyzed. In a number of code runs all systems which have an influence on the course of the transient were varied. These are the auxiliary boration system and the auxiliary feed water supply. Further, the influence of the modeling of the pressurizer safety and relief valves as well as the steam bypass system on the secondary side was assessed. The variation of the pressurizer relief and safety valve behavior has the biggest influence on the primary circuit coolant pressure. In the second part, two different core loading patterns were generated for the analyses by varying the number of MOX (mixed oxide) fuel assemblies (FA) in the core. The basic core loading contains 64 MOX FA. All these MOX FA were replaced by standard uranium oxide FA. The presence of MOX in the core has a remarkable influence on the reactivity coefficients of the fuel temperature and the moderator density. These two parameters mainly influence the behavior of the coolant pressure in the first part of the transient. It has been demonstrated that the pressure maximum decreases with growing MOX portion in the core. The maximum pressure determined in the calculations with variation of system and neutron-physical boundary conditions does not reach the allowed limit for the primary circuit.Copyright

Coupling of the 3D Neutron Kinetic Core Model DYN3D With the CFD Software ANSYS CFX

This article presents the implementation of a coupling between the 3D neutron kinetic core model DYN3D and the commercial, general purpose computational fluid dynamics (CFD) software ANSYS-CFX. In the coupling approach, parts of the thermal hydraulic calculation are transferred to CFX for its better ability to simulate the three-dimensional coolant redistribution in the reactor core region. The calculation of the heat transfer from the fuel into the coolant remains with DYN3D, which incorporates well tested and validated heat transfer models for rod-type fuel elements. On the CFX side, the core region is modelled based on the porous body approach. The implementation of the code coupling is verified by comparing test case results with reference solutions of the DYN3D standalone version. Test cases cover mini and full core geometries, control rod movement and partial overcooling transients.Copyright