Meeting fuel temperature limits in an HTR-Module Reactor during depressurized core heat-up

The thermal behaviour of an HTR-Module Reactor is discussed for the design basis event of core heat-up after fast depressurization taking into account the most unfavourable initial state and uncertainties of input data. The reactor is designed to retain its fission products inside the fuel coatings even if all active components for decay heat removal and reactivity control should fail. To meet this goal maximum fuel temperatures during core heat-up should not exceed the technological limit of 1620°C, for which the integrity of the fuel coatings has been proven experimentally.Two-dimensional thermal-hydraulic calculations show that the maximum fuel temperature during core heat-up is expected to be 1472°C taking into account nominal full power operation as an initial state, a sudden depressurization in the beginning of the event, and nominal input data. The most unfavourable initial state is the steady state operation close to the scram set points, i.e. 105% power and increased cold and hot gas temperatures. Accounting for this leads to a maximum fuel temperature of 1522°C. Relevant uncertainties of input data are those of decay heat production, power distribution and core thermal conductivity and specific heat capacity. Their individual standard deviations can be combined to an integral uncertainty margin of ±86 K which covers two standard deviations. Hence the maximum fuel temperature taking into account unfavourable initial state and uncertainties is 1608°C.     

Simulation of the mixing process in FCIs with hydrodynamic fragmentation model

Fuel Coolant Interactions (FCIs) are important issues in nuclear reactor severe accident analysis. In FCIs, fragmentation model of molten droplets is a key factor to estimate degree of possible damage. In this paper, the mixing process in FCIs is studied by the simulation of MIXA experiment with hydrodynamic fragmentation model. The result shows that hydrodynamic fragmentation model underestimates the fragmentation rate of high temperature molten droplets under the condition of low Weber numbers. It is concluded that models based on thermal fragmentation mechanism should be adopted to analyze the FCI process and its consequence.     

Simulation of basic gas mixing tests with condensation in the PANDA facility using the GOTHIC code

The paper reports the results of the assessment of the GOTHIC code using the data of four basic tests with condensation performed in the PANDA large-scale facility. Three of these experiments featured vertical injection, and in one the transient response due to a high-momentum horizontal injection (jet) was investigated. The injected fluid was either saturated steam or a superheated mixture of steam and helium, and the fluid initially present in the vessels was pure air. The simulations were carried out using a detailed three-dimensional representation of the vessels. In general, the results of the simulations, which used a relatively coarse mesh, were in good agreement with the data. Limitations in modelling local phenomena controlled by complex flow patterns (e.g. heat transfer in the region of an impinging jet) and the need for refined meshes to reproduce certain aspects of the transients (e.g. erosion of the interface between layers of different gas composition) were also identified. Finally, the analyses indicated that in two tests the role played by re-vaporisation of the condensate film was unexpectedly large, and this effect should be more carefully considered in the containment analyses and future model developments.     

Neutron-temperature noise model for a wwer core

I. V. Kurchatov Institute of Atomic Energy. Translated from Atomnaya Énergiya, Vol. 72, No. 5, pp. 500–510, May, 1992.     

Simulation of mixing effects in a VVER-1000 reactor

This work has been performed in the framework of the OECD/NEA thermalhydraulic benchmark V1000CT-2. This benchmark is related to fluid mixing in the reactor vessel during a MSLB accident scenario in a VVER-1000 reactor. Coolant mixing in a VVER-1000 V320 reactor was investigated in plant experiments during the commissioning of the Unit 6 of the Kozloduy nuclear power plant. Non-uniform and asymmetric loop flow mixing in the reactor vessel has been observed in the event of symmetric main coolant pump operation. For certain flow conditions, the experimental evidence of an azimuthal shift of the main loop flows with respect to the cold leg axes (swirl) was found.Such asymmetric flow distribution was analyzed with the Trio_U code. Trio_U is a CFD code developed by the CEA Grenoble, aimed to supply an efficient computational tool to simulate transient thermalhydraulic turbulent flows encountered in nuclear systems. For the presented study, a LES approach was used to simulate turbulent mixing. Therefore, a very precise tetrahedral mesh with more than 10 million control volumes has been created.The Trio_U calculation has correctly reproduced the measured rotation of the flow when the CAD data of the constructed reactor pressure vessel where used. This is also true for the comparison of cold leg to assembly mixing coefficients. Using the design data, the calculated swirl was significantly underestimated. Due to this result, it might be possible to improve with CFD calculations the lower plenum flow mixing matrices which are usually used in system codes.     

Large-Eddy Simulation study of turbulent mixing in a T-junction

A potential cause of thermal fatigue failures in energy cooling systems is identified with cyclic stresses imposed on a piping system. These are generated due to temperature changes in regions where cold and hot flows are intensively mixed together. A typical situation for such mixing appears in turbulent flow through a T-junction, which is investigated here using Large-Eddy Simulations (LES). In general, LES is well capable in capturing the mixing phenomena and accompanied turbulent flow fluctuations in a T-junction. An assessment of the accuracy of LES predictions is made for the applied Vreman subgrid-scale model through a direct comparison with the available experimental results. In particular, an estimation of the minimal mesh-resolution requirements for LES is examined on the basis of the complementary RANS simulations. This estimation is based on the characteristics turbulent scales (e.g., Taylor micro-scale) that can be computed from LES or RANS simulations.     

Preliminary tests of particle image velocimetry for the upper plenum of a scaled model of a very high temperature gas cooled reactor

The Very High Temperature Reactor (VHTR) is a Generation IV nuclear reactor that is currently under design. During the design process multiple studies have been performed to develop safety codes for the reactor. Two major accidents of interest are the Pressurized Conduction Cooldown (PCC), and the Depressurized Conduction Cooldown (DCC) scenario. Both involve loss of forced coolant to the core, except the latter involves a pressure loss in the main coolant loop. During normal operation a circulator pumps the coolant into the upper plenum and down through the core, but following a loss of forced coolant the natural convection causes the flow to reverse to go through the core into the upper plenum. Computer codes may be developed to simulate the phenomenon that occurs in a PCC or DCC scenario, but benchmark data is needed to validate the simulations; previously there were no experimental test facilities to provide this. This study will cover the design, construction, and preliminary testing of a 1/16th scaled model of a VHTR that uses Particle Image Velocimetry (PIV) for flow visualization in the upper plenum. Three tests were run for a partially heated core at statistically steady state, and PIV was used to generate the velocity field of three naturally convective adjacent jets; the turbulent mixing of the jets was observed. After performing a sensitivity analysis the flow rate of a single pipe was extracted from the PIV flow field, and compared with an ultrasonic flowmeter and analytic flow rate. All the values lied within the uncertainty ranges, validating the test results.     

Validation of the mixing model

This paper contains a discussion of the work being performed in the UK to validate the coarse mixing model. Attention is focused on the experiments performed at Winfrith Technology Centre in which 3 kg of molten fuel simulant were released into water. The validation of against one of these experiments ( ) is discussed in detail. It is concluded that can reproduce some features of the experiment (such as the existence of a steam chimney around the mixture and the steam production rate within a factor of two) but it does not predict the observed mixture development (the radial spreading and the deceleration of the first melt arriving at the surface) well. Additional model development and experimental analysis underway to resolve these differences is discussed.     

LMR core temperature calculation based on implicit formulation of the ENERGY model and a Krylov subspace method

A computational method was developed for predicting the steady-state temperature field in an LMR core based on the simplified energy equation mixing model and the subchannel analysis method. The θ-method was employed for discretizing the energy equation in the axial direction, and the interassembly coupling was achieved by interassembly gap flow. For an implicit scheme, a Krylov subspace method based on the BiCGSTAB algorithm and the MILU preconditioning was employed to solve the resulting linear system at each axial level. Numerical test results suggest that at least an order of magnitude reduction in the computing time can be achieved with the implicit solution scheme employing the Krylov subspace method when compared to the explicit scheme.     

基于多孔介质模型的压水堆堆芯温场数值模拟

针对压水堆的复杂结构特点,对堆芯采用多孔介质模型,建立完整的压力容器堆芯模型,使用商用软件CFX对压力容器堆芯的热工水力特性进行数值模拟,得到偏环运行和典型事故工况下冷却剂的热工水力响应特性。计算结果表明:应用多孔介质模型能有效正确直观显示堆芯的冷却剂温度分布情况,在偏环运行工况下堆芯会出现偏心现象,而通过瞬态事故工况计算结果表明堆芯中上部冷却剂温度最高,对压水堆的热工安全具有一定指导作用。     

Experimental tests on the applicability of a radioactive tracer technique for measuring coolant mixing in nuclear reactor fuel subassemblies

A previously reported radioactive tracer method for measuring coolant mixing hot channel factor is developed here. Feasibility of the technique is demonstrated with a test section simulating part of a reactor sub-channel. The test section consists of a tube containing three hollow rods simulating the fuel rods, mounted in a small hydraulic loop. Tests were carried out by injecting a ³²P activated phosphate into the main flow at Reynolds number ranging from 20 000 to 70 000. The concentration of the tracer was measured by detecting the β-rays of ³²P with three windowless Geiger-Müller counters, placed inside the hollow rods, β-particles reach the counters through openings in various axial positions along the length of the rods, covered by a thin stainless steel foil. Measured values of the concentration are reported and corresponding values of coolant cross-flow are calculated.     

Theoretical analysis of the fluid mixing at the core inlet in rolling motion

The fluid mixing at the reactor core in rolling motion and steady state is investigated numerically with CFX12.0. The CFD results are validated with experimental data in steady state. In rolling motion, the fluid mixing factor at the center of the core oscillates in a cosine function, but the variation of the fluid mixing factor surrounding the core is not regular. The variation amplitude of the fluid mixing factor next to the boundary line of fluid mixing is the most significant. The variation of fluid mixing factor increases with the increasing of additional force. The increasing of Reynolds number could depress the effect of rolling motion on the fluid mixing.     

Simulation of pulsed accidental energy release in a reactor core

All-Union Scientific-Research Institute of Engineering Physics. Translated from Atomnaya énergiya, Vol. 77, No. 2, pp. 112–118, August, 1994.     

A one-dimensional Lagrangian model for large-volume mixing

In this paper, we describe the main features of a new Lagrangian model for mixing in large stably stratified enclosures. This new modeling approach eliminates artificial diffusion in strongly convectively dominated flows. The hyperbolic behavior of the system of PDEs requires a numerical method with no artificial diffusion to preserve the very strong gradients that can be present. We present the rudiments of the model and discuss an important aspect of the discretization error analysis. The new BMIX code is first validated against an analytical model which has been shown to model experimental data very well. Finally, a comparison is presented against experimental data, gathered from a two-enclosure exchange flow set-up. Both comparisons show good agreement and verify the suitability of this new modeling approach and the correctness of the BMIX code. The computer code can simulate mixing in any stably stratified large enclosure containing a multi-component fluid.     

Equivalent circuit model for the insulated core transformer

The design of the insulated core transformer(ICT)needs to consider the flux leakage effects.An equivalent linear circuit model is proposed based on the principle of duality.It is composed by two types of leakage inductances:conventional leakage between windings and special leakage introduced mainly by the insulation gaps.The values of leakage inductances depend on the dimensions of the core,gaps,or windings and the property of magnetic materials.The circuit allows for quantitatively evaluating influences of ICT internal parameters on its output properties.The winding self- and mutual inductance matrix is mathematically converted to derive the inductance formula.As an example,the leakage parameters of a sixstage two-dimensional(2D) ICT are calculated and analyzed.     

Thermal mixing processes and RPV wall loads for HPI-emergency core cooling experiments in the HDR-pressure vessel

This paper reports about experimental and analytical results of a first series of three thermal mixing experiments at HDR with high-pressure cold water injection (20°C) of a complete 3-D, large scale, thick-walled PV at 11 MPa. This experimental setup leads to a localized, stripe-like asymmetric cooldown of downcomer and vessel wall for the conditions examined. With respect to this asymmetric thermal loading, a first unique data set of wall temperatures and surface strains has been generated as decision basis for code validations and future fracture mechanic oriented HDR experiments. The paper summarizes the experimental results of the Preliminary Test Phase of HDR TEMB (thermal mixing experiments) consisting of the three experiments T32.15, T32.17 and T32.18.Major findings with respect to fluid mixing behavior, the decrease of fluid/wall temperature in the HPI-nozzle/cold leg region, the cold leg nozzle and along the downcomer are reported. Also, transient axial and azimuthal strains and deduced stresses at the inside RPV surface are reported in- and outside the plume. In addition, comparisons between measured data and blind pretest predictions by best-estimate codes COMMIX-1B and SOLA-PTS as well as engineering models REMIX, VOLMIX and JETMIX are presented and discussed. Measured strains and stresses are compared with VISA predictions at different axial positions.     

Analytical investigation of a one-dimensional homogenized model for a pressurized water reactor core

A one-dimensional homogenized model for dynamic fluid-structure interaction in a pressurized water reactor core is used to study the influence of the virtual density and spacer's stiffness. The model consists of a linear system of partial differential equations for fluid velocity, rod velocity and pressure. For these equations analytical solutions are deduced for boundary conditions prescribing either periodic wall oscillations or linearly growing wall accelerations from rest. The theoretical model for the virtual density is verified by comparison to an experiment. For zero spacer stiffness, purely acoustic oscillations appear. For positive spacer stiffness, additional oscillations arise with relative rod motions. The wavelengths of the latter oscillations are small for weak spacers. Large numerical effort would be required in a more complete three-dimensional core-model to resolve such short wave lengths. In fact in a typical core the spacer's stiffness c_s is small in comparison to the fluid bulk modulus K. For it might be appropriate to neglect the influence of the spacers.     

Simulation of reactivity transients in a miniature neutron source reactor core

Computer simulation was carried out for reactivity induced transients in a HEU core of a tank-in-pool reactor, a miniature neutron source reactor (MNSR). The reactivity transients without scram at initial power of 3 W were studied. From the low power level, the power steadily increased with time and then rose sharply to higher peak values followed by a gradual decrease in value due to temperature feedback effects. The trends of theoretical results were found to be similar to measured values and the peak powers agreed well with experimental results. For ramp reactivity equivalent of clean core cold excess reactivity of 4 mk (4×10⁻³ Δk/k), the predicted peak power of 100.8 kW agrees favourably with the experimental value of 100.2 kW. The measured outlet temperature of 72.6 °C is also in agreement with the calculated value of 72.9 °C for the release of the core excess reactivity. Theoretical results for the postulated accidents due to fresh fuel replacement of reactivity worth 6.71 mk and addition of incorrect thickness of Be plates resulting in 9 mk reactivity insertion were 187.23 and 254.3 kW, respectively. For these high peak powers associated with these reactivity insertions, it is expected that nucleate boiling will occur within the flow channels of the reactor core.