Main Steam Line Break Calculations Using a Coupled RELAP5/PARCS Model for the Ringhals-3 Pressurized Water Reactor

A Main Steam Line Break (MSLB) is an important transient for Pressurized Water Reactors (PWR) due to the strong positive reactivity introduced by the over-cooling of the core. Since this effect is stronger when the Moderator Temperature Coefficient (MTC) has a large amplitude, a conservative result will be obtained for a high burnup of the fuel due to the more negative MTC late in the cycle. The calculations have been performed at a cycle burnup of 12.9742 GWd/tHM. The Swedish Ringhals-3 PWR is a three loop Westinghouse design, currently with a thermal power of 3000 MW. The PARCS model has 157 fuel assemblies of 8 different types. Four different types of reflector are used. The cross sections, and kinetic data were obtained from CASMO-4 calculations, using a cross section interface developed at the department. There are 24 axial nodes, and 2×2 radial nodes for each assembly. The transient option for calculating the effect of poisoning was used. The PARCS model has been validated against steady-state measurements from Ringhals-3 of the Relative Power Fraction (RPF) and of the core criticality. The RELAP5 model has 157 channels for the core which means that there is a one to one correspondence between the thermal hydraulics model and the neutronics model. There is eight axial nodes. Originally, the intention was to have 24 axial nodes but this proved not to work because of some limitation in RELAP5. There is currently no mixing between the different channels in the core. The feedwater, and turbines are modelled as boundary conditions. The stand-alone RELAP5 model has been validated against steady state measurements from Ringhals-3. A number of different cases were considered. In the first case, both the isolation of the feedwater for the broken loop, and all the control rods were assumed to work properly. For the second case one of the control rods was assumed to be stuck. The stuck rod was located in the fuel assembly with the highest power. This rod has also one of the highest rod worths. In the final case, the feedwater control valve for the broken loop was fully open. None of the cases led to any recriticality. The increase in power for each fuel assembly was also investigated. With the control rod located in the assembly with the highest power, the maximum power increase before scram turned out to be about 25% compared to the initial power.

On Voltage-Sensitive Managing the Redox-Potential of MSR Fuel Composition

The principles for optimal managing a composition of base solutions for the molten-salt reactor are formulated here for ensuring the given properties and exchange processes as a selective extracting of salt components. The correction of melt properties can be carried out by means of impurity additives parallel with the forced and controllable variation of reduction-oxidation (redox) potential of the non-stoichiometric salts. The accent is done on a possible application of the potentiometer for monitoring and managing of the properties of MSR fuel compositions. For this, one can use the precision methods of e.m.f and the coulomb-metric titration of sodium (lithium) in a galvanic cell upon the base of Na+(Li+)-β″-Al2O3 solid electrolyte with cation conductivity.

Conceptual Design of a MEDE Treatment System for Sodium Bonded Fuel

Unirradiated sodium bonded metal fuel and casting scrap material containing highly enriched uranium (HEU) is stored at the Materials and Fuels Complex (MFC) on the Idaho National Laboratory (INL). This material, which includes intact fuel assemblies and elements from the Fast Flux Test Facility (FFTF) and Experimental Breeder Reactor-II (EBR-II) reactors, as well as scrap material from the casting of these fuels, has no current use under the terminated reactor programs for both facilities. The Department of Energy (DOE), under the Sodium-Bonded Spent Nuclear Fuel Treatment Record of Decision (ROD), has determined that this material could be prepared and transferred to an off-site facility for processing and eventual fabrication of fuel for commercial nuclear reactors. A plan is being developed to prepare, package, and transfer this material to the DOE HEU Disposition Program Office (HDPO), located at the Y-12 National Security Complex in Oak Ridge, Tennessee. Disposition of the sodium bonded material will require separating the elemental sodium from the metallic uranium fuel. A sodium distillation process known as MEDE (Melt-Drain-Evaporate), will be used for the separation process. The casting scrap material needs to be sorted to remove any foreign material or fines that are not acceptable to the HDPO program. Once all elements have been cut and loaded into baskets, they are then loaded into an evaporation chamber as the first step in the MEDE process. The chamber will be sealed and the pressure reduced to approximately 200 mtorr. The chamber will then be heated as high as 650 °C, causing the sodium to melt and then vaporize. The vapor phase sodium will be driven into an outlet line where it is condensed and drained into a receiver vessel. Once the evaporation operation is complete, the system is de-energized and returned to atmospheric pressure. This paper describes the MEDE process as well as a general overview of the furnace systems, as necessary, to complete the MEDE process.

Visual Observation of Fragmentation Behavior on Molten Material Jet Surface in Coolant

For the safety design of the Fast Breeder Reactor (FBR), it is strongly required that the post accident heat removal (PAHR) is achieved after a postulated core disruptive accident (CDA). In the PAHR, it is important that the molten core material is solidified in sodium coolant which has high boiling point. Thus it is necessary to estimate the jet breakup length which is the distance that the molten core material is solidified in sodium coolant. In the previous studies (Abe et al., 2006), it is observed that the jet is broken up with fragmenting in water coolant by using simulated core material. It is pointed out that the jet breakup behavior is significantly influenced by the fragmentation behavior on the molten material jet surface in the coolant. However, the relation between the jet breakup behavior and fragmentation on the jet surface during a CDA for a FBR is not elucidated in detail yet. The objective of the present study is to elucidate the influence of the internal flow in the jet and fragmentation behavior on the jet breakup behavior. The Fluorinert™ (FC-3283) which is heavier than water and is transparent fluid is used as the simulant material of the core material. It is injected into the water as the coolant. The jet breakup behavior of the Fluorinert™ is observed by high speed camera to obtain the fragmentation behavior on the molten material jet surface in coolant in detail. To be cleared the effect of the internal flow of jet and the surrounding flow structure on the fragmentation behavior, the velocity distribution of internal flow of the jet is measured by PIV (Particle Image Velocimetry) technique with high speed camera. From the obtained images, unstable interfacial wave is confirmed at upstream of the jet surface, and the wave grows along the jet-water surface in the flow direction. The fragments are torn apart at the end of developed wave. By using PIV analysis, the velocity at the center of the jet is fast and it suddenly decreases near the jet surface. This means that the shear force acts on the jet and water surface. From the results of experiment, the correlation between the interfacial behavior of the jet and the generation process of fragments are discussed. In addition, the influence of surface instability of the jet induced by the relative velocity between Fluorinert™ and coolant water on the breakup behavior is also discussed.

Experimental and Numerical Modeling of Transition Matrix From Momentum to Buoyancy-Driven Flow in a Pressurized Water Reactor

The influence of density differences on the mixing of the primary loop inventory and the Emergency Core Cooling (ECC) water in the cold leg and downcomer of a Pressurised Water Reactor (PWR) was analyzed at the ROssendorf COolant Mixing (ROCOM) test facility. This paper presents a matrix of ROCOM experiments in which water with the same or higher density was injected into a cold leg of the reactor model with already established natural circulation conditions at different low mass flow rates. Wire-mesh sensors measuring the concentration of a tracer in the injected water were installed in the cold leg, upper and lower part of the downcomer. A transition matrix from momentum to buoyancy-driven flow experiments was selected for validation of the CFD software ANSYS CFX. A hybrid mesh with 4 million elements was used for the calculations. The turbulence models usually applied in such cases assume that turbulence is isotropic, whilst buoyancy actually induces anisotropy. Thus, in this paper, higher order turbulence models have been developed and implemented which take into account for that anisotropy. Buoyancy generated source and dissipation terms were proposed and introduced into the balance equations for the turbulent kinetic energy. The results of the experiments and of the numerical calculations show that mixing strongly depends on buoyancy effects: At higher mass flow rates (close to nominal conditions) the injected slug propagates in the circumferential direction around the core barrel. Buoyancy effects reduce this circumferential propagation with lower mass flow rates and/or higher density differences. The ECC water falls in an almost vertical path and reaches the lower downcomer sensor directly below the inlet nozzle. Therefore, density effects play an important role during natural convection with ECC injection in PWR and should be also considered in Pressurized Thermal Shock (PTS) scenarios. ANSYS CFX was able to predict the observed flow patterns and mixing phenomena quite well.

Research on Two-Phase Flow Instability in Parallel Multi-Channel Under Rolling Motion Condition

Two-phase flow instability of the parallel multi-channel system has been studied under rolling motion condition in this paper. Based on the homogeneous flow model with considering the rolling motion condition, the parallel multi-channel model is established by using the control volume integrating method. Gear method is used to solve the system equations. The influences of the inlet and upward sections and the heating power on the flow instability under rolling motion condition have been analyzed. The marginal stability boundary (MSB) under rolling motion condition is obtained and the unstable regions occur in both low and high equilibrium quality regions. The region with low inlet subcooling is also instable. In high equilibrium quality region, the multiplied period phenomenon is found and the chaotic phenomenon appears at the MSB. The oscillation part of mass flow rate (amplitude) may be averaged into other channels so that the influence of rolling motion is weakened. But the stability of multi-channel system is independent of the channel number and the increase of the channel number could only make the amplitude more uniformity in channels.

Evaluation Analysis of Prediction Methods for Two-Phase Flow Pressure Drop in Mini-Channels

2092 data of two-phase flow pressure drop were collected from 18 published papers of which the working fluids include R123, R134a, R22, R236ea, R245fa, R404a, R407C, R410a, R507, CO2, water and air. The hydraulic diameter ranges from 0.506 to 12mm; Relo from 10 to 37000, and Rego from 3 to 4×105. 11 correlations and models for calculating the two-phase frictional pressure drop were evaluated based upon these data. The results show that the accuracy of the Lockhart-Martinelli method, Mishima and Hibiki correlation, Zhang and Mishima correlation and Lee and Mudawar correalion in the laminar region is very close to each other, while the Muller-Steinhagen and Heck correlation is the best among the evaluated correlations in the turbulent region. A modified Chisholm correlation was proposed, which is better than all of the evaluated correlations in the turbulent region and its mean relative error is about 29%. For refrigerants only, the new correlation and Muller-Steinhagen and Heck correlation are very close to each other and give better agreement than the other evaluated correlations.

Effect of Inlet Configuration on Moderator Velocity and Temperature Distribution Inside the Calandria of a Heavy Water Reactor

Computational study of the moderator flow in calandria vessel of a heavy water reactor is carried out for three different inlet nozzle configurations. For the computations, PHOENICS CFD code is used. The flow and temperature distribution for all the configurations are determined. The impact of moderator inlet jets on adjacent calandria tubes is studied. Based on these studies, it is found that the inlet nozzles can be designed in such a way that it can keep the impact velocity on calandria tubes within limit while keeping maximum moderator temperature well below its boiling limit.

Analysis on the Rotary Inertial of the Primary Pump for China Advanced Research Reactor

The rotary inertia of the reactor’s primary pump is an important factor related to the reactor’s safety under some accidental conditions. If the rotary inertia of the primary pump is big, that the pump’s inertia time will be long and the coolant mass flow rate will decrease slowly. So it is helpful to remove the residual heat and enhance the safety of the reactor. On the other hand, a bigger rotary inertia increases the cost, and causes some other inconvenience. In this paper, our research object is China Advanced Research Reactor (CARR) which is under construction. Systemic mathematical and physical models were set up and a program was developed using FORTRAN language with GEAR numerical method to analyze the transient of CARR. The primary thermal-hydraulic parameters of the reactor core were obtained in case of loss of power accident under different rotary inertia of the pump. These parameters include the temperature of the coolant the clad and the fuel, the maximal quality of the reactor core and MDNBR, etc. On the basis of the calculate result, it can be found that when the rotary inertia of the pump is larger than 150 Kg·m2, all of the thermal parameters in loss of power accident meet the safety requirement. And if the rotary inertia of the pump is 450 Kg·m2 which is the design value of CARR, the delay shutdown time in loss of power accident must not exceed 20 seconds.

Spent Nuclear Fuel Radiotoxicity Evaluation in the Context of Minor Actinides Management With ADS Systems

The most effective way to reduce the time to reach a radiotoxicity reference level is, probably, the transmutation of Minor Actinides with fast fission processes. The definition of a critical approach to justify GEN.IV nuclear facilities, as minor actinides transmuters, is the main topic of this work. In this direction, firstly the evaluation of the reference level best estimation was performed, followed by the qualitative and quantitative radiotoxicity evolution in the ICRP Recommendations, their impact on P&T chosen strategy and on US and European rules in order to preliminarily analyze industrial ADS burning capability. Computational tools used are MCNPX-2.5.0 and ORIGEN-2.2 codes in synergy with the state of art of internal dosimetry in ICRP Publications, 10 CFR Part.20 and 96/29/EURATOM. One order of magnitude difference between the radiotoxicity evaluated by the 10 CFR Part.20 (1994) coefficients and the radiotoxicity evaluated by the actual European rules coefficients was identified.