Development of a three-dimensional two-fluid code with transient neutronic feedback for LWR applications

The development of a three-dimensional coupled neutronics/thermalhydraulics code for LWR safety analysis has been initiated. The transient neutronics code QUANDRY has been joined to the two-fluid thermal-hydraulics code THERMIT with the appropriate feedback mechanisms modeled. A literature review of the existing coupled neutronics/thermal-hydraulics codes is presented. It indicates that all of the known codes have limitations in their neutronic and/or thermal-hydraulic models which limit their generality of application and accuracy. It was also found that a tandem coupling scheme was most often employed and generally performed well. A detailed steady-state and transient coupling scheme based on the tandem technique was devised, taking into account the important operational characteristics of QUANDRY and THERMIT. The two codes were combined and the necessary programming modifications were performed to allow steady-state calculations with feedback. A simple steady-state sample problem was produced for the purpose of testing and debugging the coupled code

Phenomenological modelling of molten salt reactors with coupled point nuclear reactor kinetics and thermal hydraulic feedback models

The Molten Salt Reactor Experiment (MSRE) was a small circulating fuel reactor operated at Oak Ridge National Laboratory (ORNL) between 1965 and 1969. To do date it remains the only molten salt reactor (MSR) that has been operated for extended periods, on diverse nuclear fuels. Reactor physics in MSRs differs from conventional solid-fuelled reactors due to the circulation of hot fuel and delayed neutron precursors (DNPs) in the primary circuit. This alters the steady state and time-dependent behaviours of the system. A coupled point kinetic-thermal hydraulic feedback model of an MSRE-like system was constructed in order to investigate the effect of uncertainties in the values of key physical parameters on the model’s response to step and ramp reactivity insertions. This information was used to determine the parameters that affected the steady state condition and transient behaviours. The model was also used to investigate features identified in the frequency response, in particular a feature corresponding to fuel recirculation. Greater than expected mixing in the primary circuit has been previously proposed as an explanation for the lack of observation of this feature. A velocity-dependent turbulent dispersion term is proposed to increase dispersion of the fuel temperature field in order to suppress the recirculation feature in the frequency response. An additional semi-analytical model was constructed as a part verification of the mixing hypothesis - this model was also used to examine the stability of an MSRE-like design. Finally the validated coupled system model was used to establish the just-safe combination of intrinsic source and ramp rate that does not exceed an estimated maximum permissible vessel temperature. The CALLISTO-SPK stochastic point kinetics code is used to demonstrate that the intrinsic source in an MSRE-like design is sufficient to reduce the probability of a rogue startup transient to an acceptably small value from the point of view of regulatory safety analysis. Such analyses may be used to support the case for extrinsic source deletion in future MSR designs.Open Acces

Continuous Order Identification of PHWR Models Under Step-back for the Design of Hyper-damped Power Tracking Controller with Enhanced Reactor Safety

This is the author accepted manuscript. The final version is available from Elsevier via the DOI in this record.In this paper, discrete time higher integer order linear transfer function models have been identified first for a 500 MWe Pressurized Heavy Water Reactor (PHWR) which has highly nonlinear dynamical nature. Linear discrete time models of the nonlinear nuclear reactor have been identified around eight different operating points (power reduction or step-back conditions) with least square estimator (LSE) and its four variants. From the synthetic frequency domain data of these identified discrete time models, fractional order (FO) models with sampled continuous order distribution are identified for the nuclear reactor. This enables design of continuous order Proportional-Integral-Derivative (PID) like compensators in the complex w-plane for global power tracking at a wide range of operating conditions. Modeling of the PHWR is attempted with various levels of discrete commensurate-orders and the achievable accuracies are also elucidated along with the hidden issues, regarding modeling and controller design. Credible simulation studies are presented to show the effectiveness of the proposed reactor modeling and power level controller design. The controller pushes the reactor poles in higher Riemann sheets and thus makes the closed loop system hyper-damped which ensures safer reactor operation at varying dc-gain while making the power tracking temporal response slightly sluggish; but ensuring greater safety margin.This work has been supported by Department of Science and Technology (DST), Govt. of India, under the PURSE programme

Investigation of the Coupled Nuclear, Thermal-Hydraulic, and Thermomechanical Response of a Natural Circulation Research Reactor Under Severe Reactivity-Initiated Accident Transients

Research reactors play an important role in higher education, scientific research, and medical radioisotope production around the world. It is thus important to ensure the safety of facility workers and the public. This work presents a new reactor transient analysis code, referred to as Razorback, which computes the coupled reactor kinetics, fuel element heat transfer, fuel element thermal expansion and thermal stress, and thermal-hydraulic response of a natural circulation research reactor. The code was developed for the evaluation of large rapid reactivity addition in research reactors, with an initial focus on the Annular Core Research Reactor (ACRR) at Sandia National Laboratories. Razorback has been validated using ACRR pulse operations, and the simulation results are shown to agree very well with measured reactor data. Razorback is also used to examine the response of a natural circulation research reactor (i.e., the ACRR) to large rapid reactivity additions. The reactor kinetic response, the thermal-hydraulic response of the fuel and coolant, and the thermomechanical response of the fuel element materials are each examined separately. Safety analysis and operational implications are discussed

Towards multi-physics description of fuel behaviour for accidental conditions

In the present document, the development of well-structured multi-physics simulation environments to complement fuel performance analysis is presented. The simulation environments are based on information from the sub-channel / reactor scale, i.e., initial and boundary conditions for the fuel pin simulations in off-normal conditions. The environments are developed based on the codes TRANSURANUS, OpenFOAM, SIMMER-III, and BELLA, focused on satisfying the requirements of the code/module to fuel behaviour, with a strong perspective towards the BPJ simulations of concern for the MYRRHA sub-critical core. The results obtained using the multi-physics simulation environments support the design optimization and safety assessment of the MYRRHA fuel pin during normal irradiation and transient scenarios. As well, it will be used in the activity associated with Task 6.2 of the PATRICIA Project, focused on the in-depth, complete analysis of multiple BPJ scenarios, to identify the worst case and hence draw conservative conclusions on the MYRRHA pin safety under irradiation

Reactivity changes during startup in large nuclear rockets

Large uranium 235 nuclear rockets - reactivity changes during startup by hydrogen density and core temperature variation

A High Fidelity Multiphysics Framework for Modeling CRUD Deposition on PWR Fuel Rods.

Corrosion products on the fuel cladding surfaces within pressurized water reactor fuel assemblies have had a significant impact on reactor operation. These types of deposits are referred to as CRUD and can lead to power shifts, as a consequence of the accumulation of solid boron phases on the fuel rod surfaces. Corrosion deposits can also lead to fuel failure resulting from localized corrosion, where the increased thermal resistance of the deposit leads to higher cladding temperatures. The prediction of these occurrences requires a comprehensive model of local thermal hydraulic and chemical processes occurring in close proximity to the cladding surface, as well as their driving factors. Such factors include the rod power distribution, coolant corrosion product concentration, as well as the feedbacks between heat transfer, fluid dynamics, chemistry, and neutronics. To correctly capture the coupled physics and corresponding feedbacks, a high fidelity framework is developed that predicts three-dimensional CRUD deposition on a rod-by-rod basis. Multiphysics boundary conditions resulting from the coupling of heat transfer, fluid dynamics, coolant chemistry, CRUD deposition, neutron transport, and nuclide transmutation inform the CRUD deposition solver. Through systematic parametric sensitivity studies of the CRUD property inputs, coupled boundary conditions, and multiphysics feedback mechanisms, the most important variables of multiphysics CRUD modeling are identified. Moreover, the modeling framework is challenged with a blind comparison of plant data to predictions by a simulation of a sub-assembly within the Seabrook nuclear plant that experienced CRUD induced fuel failures. The physics within the computational framework are loosely coupled via an operator-splitting technique. A control theory approach is adopted to determine the temporal discretization at which to execute a data transfer from one physics to another. The coupled stepsize selection is viewed as a feedback control problem, and a controller of the type integral is utilized. The temporal discretization adapts with the problem solution to maintain a user-prescribed tolerance of specific solution variables. Predictor-corrector algorithms enable convergence error estimates. The highly nonlinear precipitation rate of solid boron phases, and their dependence on the local thermal hydraulic conditions, is the primary motivation for seeking an automated and adaptive stepsize selection algorithm.PhDNuclear Engineering and Radiological SciencesUniversity of Michigan, Horace H. Rackham School of Graduate Studieshttp://deepblue.lib.umich.edu/bitstream/2027.42/120638/1/djwalter_1.pd

Coupled point neutron kinetics and thermal-hydraulics models of transient nuclear criticality excursions in wetted fissile uranium dioxide (UO2) powders

This thesis describes a phenomenologically based mathematical and computational methodology for the simulation of a postulated transient nuclear criticality excursion initiated by the incursion of water, from a fire-sprinkler system, into a bed of dry UO2 powder. These potentially hazardous multi-phase dispersed particulate systems may form as a result of a fire or explosion in a nuclear fuel fabrication facility. The models proposed in this thesis aim to support nuclear criticality safety analysis and assessment. In addition, the development of these models aims to support emergency planning and preparedness. The point neutron kinetics equations are coupled to phenomenological models of water infiltration, sedimentation, fluidisation, nuclear thermal hydraulics, radiolysis and boiling, through the use of multivariate reactivity feedback components. The spatial and temporal solution of this set of equations enables the modelling of postulated transient nuclear criticality excursions in highly dispersed three-phase particulate systems of UO2 powder. The results indicate that there is the potential for large positive reactivities to be added to a UO2 powder system as pores become filled with water. Generally, thermal expansion and Doppler broadening are insufficient to control the transient, leading to significant radiolysis and boiling on the surface of the UO2 powder particles. Radiolytic gas and steam bubble induced fluidisation and sedimentation significantly alters the characteristics of a transient nuclear criticality excursion and should be carefully considered. Research has also been undertaken examining transient nuclear criticality excursions in weak intrinsic neutron source UO2 powder systems by solving the forward probability balance equation and using a Gamma probability distribution function to estimate mean wait-time probability distributions. Significant variations in the potential initial peak power are predicted for highly enriched, wetted, UO2 powders as a function of the stochastic behaviour associated with criticality excursions in low neutron population systems.Open Acces

Fission gas effects in fast reactor dissassembly

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