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T-OZON{sub TM} is the chemical decontamination method developed by Toshiba for the decontamination of BWR system piping. To apply T-OZON{sub TM} to PWR decommissioning, the decontamination application conditions have been optimized on actual PWR plant samples and ozone concentration behavior in the water have been studied. (authors)
Research reactor BR2 is undergoing a thorough maintenance and modernisation operation until mid-2016. In order to be able to use the reactor safely and efficiently in the coming period of 10 years, a preventive replacement of the beryllium matrix was required as a major step in the refurbishment. At the same time, a number of preventive maintenance and modernisation activities have been defined, including the dismantling of the CALLISTO (Capability for Light Water Irradiation in Steady State and Transient Operation) PWR experimental facility for in-pile studies installed into BR2 for more than 20 years. The CALLISTO end of life decommissioning strategy aims for waste, cost and dose uptake minimisation and includes: - the unloading and removal of the in-pile sections; - a closed-loop full system decontamination for dose reduction; - the complete dismantling of the loop; - the further sorting and segmentation in a dedicated workshop; - a hard chemical batch-wise decontamination for unconditional clearance. The operation is planned in the period of mid-2015 until mid-2016. The paper will describe in detail these subsequent steps, with a focus on decontamination and dismantling. Lessons learned and potential benefits towards the D and D of commercial PWR will be highlighted. (authors)
A nuclear reactor is a complicated system that is comprised of many systems and components. In order to fully model a reactor, several approximations and correlations must be used for a real-time simulation. In 1984 as part of a doctoral dissertation at Massachusetts Institute of Technology (MIT), a FORTRAN 77 program was developed to simulate a four-loop, Westinghouse pressurized water reactor (PWR). The program was later called PRISM and extended to include a comprehensive set of features accessible in a DOS-based user interface. This present work seeks to update the original PRISM program for cross platform deployment using the Python language. The goal of this present work is to update and improve a teaching tool for both students and reactor operators. Originally, PRISM was created as part of a PhD project in FORTRAN 77, but the code no longer works readily on modern operating systems. Python was chosen to rewrite the program because of the ease of use of the Python modules including Kivy as well as the portability of the program to various operating systems. This project aims to develop desktop and mobile versions of the software, allowing the user flexibility in choice of platform. This simulator is scheduled to be deployed in time for a live demonstration at the 2016 Annual ANS meeting. The project aims to produce a reactor simulator that will showcase the dynamics of a PWR system in real-time under various conditions. Some potential scenarios include startup conditions, scram operations, or failure conditions such as a break in a coolant leg. The user will be able to induce various failure scenarios and the program will simulate how the reactor will behave. The simulation will include a suite of monitoring functions that will allow automatic systems to attempt to adapt to the changing conditions, e.g., control rod movement. This program will provide an educational resource for anyone learning about nuclear reactor kinetics or dynamics of a reactor system. (authors)
The sustainable energy over a long-term period can be supplied by ultra-long-life fast reactor cores that are designed by fully utilizing breeding by fast neutrons to be operated over several tens of years without refueling. But the fast reactor cores of this type usually have high positive coolant void reactivity worth resulted from the high TRU (Transuranics) contents in the high burnup, which can degrade the inherent safety through high positive coolant expansion reactivity coefficient. Recently, the authors have designed new small ultra-long-life cores that can be operated over 50{approx}60 EFPYs (Effective Full Power Years) and satisfy the self-controllability conditions under the unprotected accidents. The cores use axial blanket-driver-blanket configuration to achieve ultra-long operational cycle length where the fuel burnup propagates from the outer core to the inner core and from the axially central driver to the outer blanket regions. The objective of this work is to suggest the different design options with thorium addition and TRU and uranium from PWR spent fuels for effectively utilizing the large PWR spent fuel stocks and to analyze the performances of the cores. (authors)
In this work we studied the sensitivity of different computational fluid dynamics (CFD) numerical solutions on turbulence model, domain discretization mesh type, and inlet boundary conditions. The case under examination is a turbulent one-phase flow (water) through a scaled nuclear pressurized water reactor (PWR) fuel assembly. The main feature of this geometrical configuration is the presence of a spacer grid with mixing vanes. The design of a spacer grid with mixing vanes plays a key role for the core heat transfer efficiency, increasing the coolant heat transfer coefficient. The main reason for this is the introduction of the top mixing vanes, which consist of an ordered repetitive succession of wings that generate secondary swirling flows. The increase of heat transfer efficiency also increases the critical heat flux (CHF), making it possible to increase the design and operational limits, far from boiling crises (departure of nucleate boiling in case of a PWR). The main focus of this work was to analyze the simulation capabilities of Reynolds-averaged Navier-Stokes (RANS) models (a cubic non-linear k-epsilon model formulation) for this complex flow geometry. The CFD code used was CD-Adapco STAR-CCM+.
The present work illustrates one example of a computational fluid dynamics (CFD) application in the nuclear engineering field. Since CFD is one of the most powerful tools available at the actual state of the art that could in principle help and assist the design process, a compromise should be established between the accuracy of the CFD numerical model (RANS - LES - DNS) and the computational cost of the simulations. In this spirit, this work will present the computational results from two different CFD model categories, Large Eddy Simulations (LES) and Reynolds-averaged Navier-Stokes (RANS) models. Both treat the flow turbulence issue in two separate ways, which differ mostly in the resolution of the turbulence statistics and the computational cost. The single-phase hydraulic problem presented involves the steady-state turbulent flow field modelling and resolution around a scaled PWR mixing vane grid, designed by Westinghouse and tested by Texas A and M University. The role of the mixing vane/spacer grid is to enhance the turbulent mixing and hence heat transfer of the core coolant around fuel bundles, thus increasing the departure of nucleate boiling-critical heat flux (DNB-CHF) point, as well as important structural functions for the rods bundles. The authors numerically studied the flow field characteristics at the region downstream of the spacer grids and validated their results to particle imagery velocimetry (PIV) measurements. Dominguez and Hassan performed two-dimensional time resolved PIV (TR-PIV) measurements within a 5 x 5 PWR rod bundle with spacer-grids that utilized a matching index of refraction (MIR) approach. Their averaged data showed a rather uniform direction velocity profile after the spacer grid while the downstream illustrated a no slip wall profile. The available PIV flow field experimental results from Texas A and M show complex structures of the flow field downstream the mixing vanes with the formation of secondary flows that tend to enhance the turbulent heat transfer, due to the mixing vanes geometry itself. Computational results will be validated with the available experimental data.
The burnup credit approach for a spent fuel transportation cask (SFTC) with a high density basket for 32 PWR fuel assemblies required credit not only for actinides but also for fission products, otherwise the required minimum burnups of fuel assemblies would have been too high to be practical. At the time of the development of the burnup credit methodology for this condition, no established approach existed to quantitatively validate credit for fission products. A validation approach was therefore developed specifically for this application, based on Commercial Reactor Criticals (CRCs). In this approach, hot critical state points of PWR plants were evaluated using the same codes that were used for the burnup credit evaluations for the transportation cask, namely CASMO-4 and MCNP4. A total of 36 of those state-points were analyzed, from 5 plants with two different assembly types. For each axial section of each fuel assembly, a separate CASMO-4 depletion analysis was performed, which required a total of about 4000 individual CASMO runs. The spent fuel compositions were then extracted from the CASMO runs and transferred to MCNP, where each core was modeled, considering the existing 1/8 symmetry. The core operating conditions were taken from work previously performed and documented for the Yucca Mountain project. The results of the MCNP calculations were then analyzed to establish bias and bias uncertainty values. Additional studies were performed to compare the in-core condition for the fuel with the condition that exists for the fuel in the spent fuel transport cask. The paper presents and discusses the results of the calculations, and the analysis of those results to establish bias and bias uncertainty values. The paper also presents results of the studies comparing in-core with transport cask conditions, and discusses the resulting limitations of this approach. (author)
Effects of temperature and flow gradients in Westinghouse designed three-loop Pressurized Water Reactor (PWR) cold legs, the piping between the main coolant pump and the reactor vessel, were evaluated using Computational Fluid Dynamics (CFD) code STAR CCM+ and coupled neutronic and thermal-hydraulic (T/H) code system VERA. In the parametric study, several symmetric and asymmetric temperature gradients that were significantly larger than those observed from plant measurements were applied to the cold leg inlets for comparison with the base case without any temperature gradient. A CFD model using the STAR-CCM+ code was developed for a portion of the RCS region between the Reactor Coolant Pump (RCP) and the core inlet based on previously validated modeling approach. The CFD simulation results were processed for the temperature and flow rate distributions at the core inlet as input to the VERA calculations. The VERA code system consists of COBRA-TF (CTF) for thermal-hydraulics, MPACT for reactor physics and neutron transport, and ORIGEN for isotopic depletion. The VERA model was for depletion calculations of a high-burnup loading pattern with the reactor core in pin-by-pin and subchannel resolution. The results of the study indicate that the postulated temperature gradients within the PWR cold legs do not result in any significant changes in the core inlet temperature distributions and the core power distributions during the reactor operation. (authors)
This paper investigates the new CAMELOT coupling route developed in WIMS, which uses prepared tabulated cross-sections combined with the flux solver MERLIN module and the integrated sub-channel thermal hydraulics solver ARTHUR module to solve for the coupled core state. This coupling route is applied to the OECD/NEA pressurized water reactor (PWR) core transient benchmark exercise, involving 3-D neutronics and thermal-hydraulics analysis of a full-sized PWR core featuring MOX fuel at various core states, as well as simulating a control rod ejection transient. Analysis of the results for each part of the benchmark demonstrate that the high-resolution pin-by-pin analysis used in WIMS closely matches the other participants in the benchmark, with differences of between 20-200 pcm for k-effective, 1-2.5% RMS difference for assembly-averaged powers, 2-8% difference in rod worth calculations, < 200 ppm for calculations of critical boron concentration, zero difference in delayed neutron fraction calculation and a very close match for power response to the rod ejection transient. This is the first application of the WIMS-CAMELOT approach for the transient analysis of a full-sized reactor and as such further improvements such as reducing computational cost through parallelization, optimisation and memory reduction, are currently in active development. This paper demonstrates the accuracy and flexibility of the CAMELOT coupling route, providing a straightforward and consistent user image that can be easily used for a variety of modelling problems involving coupled neutronics and thermal-hydraulics. (authors)
In this paper, a data-assimilation method has been proposed and applied for the burnup distribution of PWR. The burnup distribution is significant to the safety and economy of the reactor, as it is essential for the fuel-reloading design and optimization. Due to the burnup distribution cannot be measured directly during the reactor operation, the numerical simulation is widely applied to determine the burnup distribution. However, there is a deviation between the numerical simulation and the actual core due to some unavoidable factors, such as component manufacturing deviation, uneven flow distribution and so on. These differences would induce the errors to the simulation values of power distributions and hence to the burnup distribution. To address this problem, a data-assimilation method for the burnup distribution has been proposed with the application of power-distribution measurements. In our research, the three-dimensional variational (3DVAR) algorithm was applied for burnup-distribution calibration and the artificial neutral network (ANN) was applied to establish the relation between power distribution and corresponding burnup distribution. As engineering verification, the proposed data-assimilation method has been applied to the CNP1000 PWR operated in China. The numerical results indicated that the burnup-distribution errors can be reduced notably, as the maximum value of relative errors for power distribution can be reduced from 5.35% to 3.96%. (authors)
Here we describe, by means of Chebyshev polynomials, a parameterized representation of the homogenized macroscopic cross section for PWR fuel element as a function of soluble boron concentration, moderator temperature, fuel temperature, moderator density and U-235(92) enrichment. The cross-section data analyzed are fission, scattering, total, transport, absorption and capture. The parameterization enables a quick and easy determination of problem-dependent cross-sections to be used in few group calculations. The methodology presented in this paper will allow generation of group cross-section data from stored polynomials to perform PWR core calculations without the need to generate them based on computer code calculations using standard steps. The results obtained by the proposed methodology when compared with results from the SCALE code calculations show very good agreement.
This paper presents a TRACE model of a postulated large-break loss-of-coolant accident (LBLOCA) event in a four-loop pressurized water reactor (PWR). The input files have been made publicly available and serve as a starting point for researchers to improve upon or apply to their purposes. The model, based on a legacy PWR model, contains numerous improvements and modifications consistent with US Nuclear Regulatory Commission LBLOCA analysis guidelines. The model is applied to analyze high-burnup (>62 GWD/MTU) fuel behavior during LBLOCA, improving on previous work to provide realistic thermal hydraulic predictions for future fuel performance analyses and experiments related to fission fragment, relocation, and dispersal (FFRD). Sensitivity studies are presented to quantify the impact of the vessel-modeling approach, core radial discretization, and other phenomena. In conclusion, the model is intended for research purposes and contains sufficient plant geometric, operational, and safety system details to represent the main physical phenomena pertaining to LBLOCA.
A project to research the application of nondestructive assay (NDA) to spent fuel assemblies is underway. The research team comprises the European Atomic Energy Community (EURATOM), embodied by the European Commission, DG Energy, Directorate EURATOM Safeguards; the Swedish Nuclear Fuel and Waste Management Company (SKB); two universities; and several United States national laboratories. The Next Generation of Safeguards Initiative–Spent Fuel project team is working to achieve the following technical goals more easily and efficiently than in the past using nondestructive assay measurements of spent fuel assemblies: (1) verify the initial enrichment, burnup, and cooling time of facility declaration; (2) detect the diversion or replacement of pins, (3) estimate the plutonium mass, (4) estimate the decay heat, and (5) determine the reactivity of spent fuel assemblies. This study focuses on spectrally resolved gamma-ray measurements performed on a diverse set of 50 assemblies [25 pressurized water reactor (PWR) assemblies and 25 boiling water reactor (BWR) assemblies]; these same 50 assemblies will be measured with neutron-based NDA instruments and a full-length calorimeter. Given that encapsulation/repository and dry storage safeguards are the primarily intended applications, the analysis focused on the dominant gamma-ray lines of 137 Cs, 154 Eu, and 134 Cs because these isotopes will be the primary gamma-ray emitters during the time frames of interest to these applications. This study addresses the impact on the measured passive gamma-ray signals due to the following factors: burnup, initial enrichment, cooling time, assembly type (eight different PWR and six different BWR fuel designs), presence of gadolinium rods, and anomalies in operating history. As a result, to compare the measured results with theory, a limited number of ORIGEN-ARP simulations were performed.
The declared inventory of 235 U in fresh, low enriched uranium, nuclear fuel assemblies is routinely verified by nuclear safeguards inspectorates using the UNCL (Uranium Neutron Collar – Light Water Reactor Fuel) nondestructive assay instrument. The trend in modern fuels is towards higher initial enrichment, which in turn requires a larger number of burnable poison pins with higher Gadolinia concentration to hold down the initial reactivity. UNCL assay error for modern fuel assemblies typically exceeds 10%. The traditional algorithm used to correct the response for burnable poison content needs revision to achieve accuracy comparable to the $\sigma_\text{R}$ of 4.5% expected for poison-free assemblies. Here we review the problem and use a large set of 287 Monte Carlo simulations based on an experimentally benchmarked model corresponding to the 16×16 PWR array used in Brazilian Angra type II and III fuel. Simulated relative responses are used to evaluate the functional form of the poison correction and determine parameters optimal for 16×16 PWR assemblies. This update is conducted using between 4–24 Gadolinia pins ranging 2–11 wt% Gd 2 O 3 with assemblies having mean enrichments ranging 2.5%–5%. Experimental validation of the updated coefficients using seven measurements shows bias is reduced by an order of magnitude and $\sigma_\text{R}$R reduced from 10% to less than 2%. Two updated sets of coefficients are provided, the most accurate of which is directly useable in the existing analysis code (INCC) in use by inspectorate at fuel fabrication facilities.
VERA-CS (Virtual Environment for Reactor Applications, Core Simulator) is a coupled neutron transport and thermal-hydraulics subchannel code under development by the Consortium for Advanced Simulation of Light Water Reactors (CASL). VERA-CS was used to simulate a typical pressurized water reactor (PWR) full core response with 17x17 fuel assemblies for a main steam line break (MSLB) accident scenario with the most reactive rod cluster control assembly stuck out of the core. The accident scenario was initiated at the hot zero power (HZP) at the end of the first fuel cycle with return to power state points that were determined by a system analysis code and the most limiting state point was chosen for core analysis. The best estimate plus uncertainty (BEPU) analysis method was applied using Wilks’ nonparametric statistical approach. In this way, 59 full core simulations were performed to provide the minimum departure from nucleate boiling ratio (MDNBR) at the 95/95 (95% probability with 95% confidence level) tolerance limit. The results show that this typical PWR core remains within MDNBR safety limits for the MSLB accident.
The reactor performance and safety characteristics of mixed thorium mononitride (ThN) and uranium mononitride (UN) fuels in a pressurized water reactor (PWR) are investigated to discern the potential nonproliferation, waste, and accident tolerance benefits provided by this fuel form. In this paper, we present results from an initial screening of mixed ThN-UN fuels in normal PWR operating conditions and compares their reactor performance to UO 2 in terms of fuel cycle length, reactivity coefficients, and thermal safety margin. ThN has been shown to have a significantly greater thermal conductivity than UO 2 and UN. Admixture with a UN phase is required because thorium initially contains no fissile isotopes. Results from this study show that ThN-UN mixtures exist that can match the cycle length of a UO 2 -fueled reactor by using 235 U enrichments greater than 5% but less than 20% in the UN phase. Reactivity coefficients were calculated for UO 2 , UN, and ThN-UN mixtures, and it was found that the fuel temperature and moderator temperature coefficients of the nitride-based fuels fall within the acceptable limits specified by the AP1000 Design Control Document. Reduced soluble boron and control rod worth for these fuel forms indicates that the shutdown margin may not be sufficient, and design changes to the control systems may need to be considered. The neutronic impact of 15 N enrichment on reactivity coefficients is also included. Due to the greatly enhanced thermal conductivity of the nitride-based fuels, the UN and ThN-UN fuels provide additional margin to fuel melting temperature relative to UO 2 .
We report that absorbing heat from the fuel rod surface, water as coolant can undergo subcooled boiling within a pressurized water reactor (PWR) fuel rod bundle. Because of the buoyancy effect, the vapor bubbles generated will then rise along and interact with the subchannel geometries. Reliable prediction of bubble behavior is of immense importance to ensure safe and stable reactor operation. However, given a complex engineering system like a nuclear reactor, it is very challenging (if not impossible) to conduct high-resolution measurements to study bubbly flows under reactor operation conditions. The lack of a fundamental two-phase-flow database is hindering the development of accurate two-phase-flow models required in more advanced reactor designs. In response to this challenge, first-principles–based numerical simulations are emerging as an attractive alternative to produce a complementary data source along with experiments. Leveraged by the unprecedented computing power offered by state-of-the-art supercomputers, direct numerical simulation (DNS), coupled with interface tracking methods, is becoming a practical tool to investigate some of the most challenging engineering flow problems. In the presented research, turbulent bubbly flow is simulated via DNS in single PWR subchannel geometries with auxiliary structures (e.g., supporting spacer grid and mixing vanes). The geometric effects these structures exert on the bubbly flow are studied with both a conventional time-averaging approach and a novel dynamic bubble tracking method. Lastly, the new insights obtained will help inform better two-phase models that can contribute to safer and more efficient nuclear reactor systems.
A plethora of theoretical, numerical, and experimental investigations have relied on updraft forced convection of operating fuel assemblies for light water-cooled fuel assemblies. The inertial scales of turbulence in updraft flow through PWR fuel assemblies are primarily influenced by upstream mixing patterns within the lower plenum and consequent acceleration of flow through the lower core and support plates into the bottom nozzles. In secondary heat transfer applications such as steam generation, recuperation and economization, downdraft flow is utilized to retrieve maximal sensible heat from the primary coolant. Sub-channel volumes aligned vertically could also benefit from force of gravity in developing turbulent flows. A large reduction in pumping power, associated cost of operation and maintenance, and improved balance of plant is suggested to be possible through downdraft forced convection in PWR-type reactors. This article explores the impact of downdraft forced convection within the well-studied 3x3 sub-channel within a range of 20 hydraulic diameters across a representative intermediate flow mixing grid. A range of inflow Reynolds number from 10,000 to 60,000 is simulated over a range of equivalent heat loads of 20% to 100% spanning 20 hydraulic diameters. The fuel rods are represented in true geometric detail, and a constant-value heat profile is assumed for the investigation. At each inflow condition, the turbulence intensity, peak velocity, local heat transfer coefficient, peak wall heat flux, and peak wall temperature are collected across the sub-channel and compared with an updraft configuration evaluated at the same conditions.