Advanced First Core Design for the Westinghouse AP1000

As the nuclear renaissance is now upon us and new plants are either under construction or being ordered, a considerable amount of attention has also turned to the design of the first fuel cycle. Requirements for core designs originate in the Utilities Requirements Document (URD) for the United States and the European Utilities Requirements (EUR) for Europe. First core designs created during the development of these documents were based on core design technology dating back to the 1970’s, where the first cycle core loading pattern placed the highest enrichment fuel on the core periphery and two other lower enrichments in the core interior. While this sort of core design provided acceptable performance, it underutilized the higher enriched fuel assemblies and tended to make transition to the first reload cycle challenging, especially considering that reload core designs are now almost entirely of the Low Leakage Loading Pattern (LLLP) design. The demands placed on today’s existing fleet of pressurized water reactors for improved fuel performance and economy are also desired for the upcoming Generation III+ fleet of plants. As a result of these demands, Westinghouse has developed an Advanced First Core (AFCPP) design for the initial cycle loading pattern. This loading pattern design simulates the reactivity distribution of an 18 month low leakage reload cycle design by placing the higher enriched assemblies in the core interior which results in improved uranium utilization for those fuel assemblies carried through the first and second reload cycles. Another feature of the advanced first core design is radial zoning of the high enriched assemblies, which allows these assemblies to be located in the core interior while still maintaining margin to peaking factor limits throughout the cycle. Finally, the advanced first core loading pattern also employs a variety of burnable absorber designs and lengths to yield radial and axial power distributions very similar to those found in typical low leakage reload cycle designs. This paper will describe each of these key features and demonstrate the operating margins of the AFC design and the ability of the AFC design to allow easy transition into 18 month low leakage reload cycles. The fuel economics of the AFC design will also be compared to those of a more traditional first core loading pattern.

Agents and Processes Design for TRU Elements Back Extraction in TRPO Process

To reduce the long-term risk of the high-level liquid waste (HLLW) and the waste disposal cost, transuranium (TRU) elements should be removed from HLLW. A so-called TRPO process has been developed by Chinese scientists to partition HLLW. In this process, the extractant, trialkyl phosphine oxide (TRPO), is able to extract TRU elements into organic phase completely, which makes the treatment and disposal of raffinate HLLW much easier. However, the treatment of extracted TRU elements in organic phase, in return, becomes new troublesome issue. Generally, there are three promising ways to treat the extracted TRU elements: (1)transmutation; (2)conditioning; (3)recycling U+Pu in Purex-TRPO Integrated Process. In any of the three ways, the back extraction agents and processes play significant roles. In this paper, the investigations on back extraction agents for TRU elements, such as TTHA, DTPA, AHA, HEDPA, DOGA, and carbonates are introduced. The corresponding back extraction processes and experimental results are reviewed.

Numerical Flow Simulation of Spacer Grids With Mixing Vanes in a 5 × 5 PWR Rod Bundle

The spacer grids exert great influence on the thermal hydraulic performance of the PWR fuel assembly. The presence of the spacers has two antagonistic effects on the core: an increase of pressure drop due to constriction on the coolant flow area and increase of the local heat transfer downstream the grids caused by enhanced coolant mixing. The mixing vanes, present in most of the spacer grid designs, cause a cross and swirl flow between and in the subchannels, enhancing even more the local heat transfer at the cost of more pressure loss. Due to this important hydrodynamic feature the spacer grids are often improved aiming to obtain an optimal commitment between pressure drop and enhanced heat transfer. In the present work, the fluid dynamic performance downstream a 5 × 5 rod bundle with spacer grids is analyzed with a commercial CFD code (CFX 11.0). Eleven different split vane spacer grids with angles from 16° to 36° and a spacer without vanes were evaluated. The computational domain extends from ∼10 Dh upstream to ∼50 Dh downstream the spacer grids. The standard k-ε turbulence model with scalable wall functions and the total energy model were used in the simulations. The results show a considerable increase of the average Nusselt number and secondary mixing with the angle of the vane up to ∼20 Dh downstream the spacer, reducing greatly the influence of the vane angle beyond this region. As expected, the pressure loss through the spacer grid also showed considerable increase with the vane angle.

Hydrodynamic Modeling of Mineral Wool Fiber Suspensions in a Two-Dimensional Flow

A consequence of a loss of coolant accident is that the local insulation material is damaged and maybe transported to the containment sump where it can penetrate and/or block the sump strainers. An experimental and theoretical study, which examines the transport of mineral wool fibers via single and multi-effect experiments is being performed. This paper focuses on the experiments and simulations performed for validation of numerical models of sedimentation and resuspension of mineral wool fiber agglomerates in a racetrack type channel. Three velocity conditions are used to test the response of two dispersed phase fiber agglomerates to two drag correlations and to two turbulent dispersion coefficients. The Eulerian multiphase flow model is applied with either one or two dispersed phases.

Software FMEA Analysis for Safety Software

This paper describes a software safety analysis for a software code that is installed at an Automatic Test and Interface Processor (ATIP) in a digital reactor protection system. For the ATIP software safety analysis, an overall safety analysis is at first performed over the ATIP software architecture and modules, and then a detailed safety analysis based on the software FMEA (Failure Modes and Effect Analysis) method is applied to the ATIP program. For an efficient analysis, the software FMEA is carried out based on the so-called failure-mode template extracted from the function blocks used in the function block diagram (FBD) for the ATIP software. The software safety analysis by the software FMEA, being applied to the ATIP software code which has been integrated and passed through a very rigorous system test procedure, is proven to be able to provide very valuable results (i.e., software defects) which could not be identified during various system tests.

Advanced Control Design, Optimal Sensor Placement, and Technology Demonstration for Small and Medium Nuclear Power Reactors

The efficient and safe performance of nuclear power plants of the future requires remote monitoring, control, and condition-based maintenance in order to maximize their capacity factor. Small and medium reactors, in the 50–500 MWe power range, may become commonplace for certain applications, with a design features for remote deployment. Such a reactor may be part of a smaller electrical grid, and deployed in areas with limited infrastructure. Typical applications include power generation, process heat for water desalination, and co-generation. There are other considerations in the deployment of these reactors: development of effective I&C to support nuclear fuel security monitoring, longer than normal fuel cycle length, and increased autonomy in plant operation and maintenance. A Model Predictive Controller (MPC) for the IRIS (International Reactor Innovative and Secure) system has been developed as a multivariate control strategy for reactor power regulation and the control of the helical coil steam generator (HCSG) used in IRIS. A MATLAB-SIMULINK model of the integral reactor was developed and used to demonstrate the design of the MPC. The two major control actions are the control rod reactivity perturbation and the steam control valve setting. The latter is used to regulate the set point value of the superheated steam. The MPC technique minimizes the necessity of on-line controller tuning, and is highly effective for remote and autonomous control actions. As an important part of the instrumentation & control (I&C) strategy, sensor placement in next generation reactors needs to be addressed for both control design and fault diagnosis. This approach is being applied to the IRIS system to enhance the efficiency of reactor monitoring that would assist in a quick and accurate identification of faults. This is achieved by solving the problem from the fault diagnosis perspective, rather than treating the sensor placement as a pure optimization problem. The solution to the problem of sensor placement may be broadly divided into two tasks: (1) fault modeling or prediction of cause-effect behavior of the system, generating a set of variables that are affected whenever a fault occurs, and (2) use of the generated sets to identify sensor locations based on various design criteria, such as observability, resolution, reliability, etc. The proposed algorithm is applied to the design of a sensor network for the IRIS system using multiple design criteria. This enables the designer to obtain a good preliminary design without extensive quantitative information about the process. The control technique will be demonstrated by application to a real process with actuators and associated device time delays. A multivariate flow control loop has been developed with the objective of demonstrating digital control implementation using proportional-integral controllers for water level regulation in coupled tanks. The controller implementation includes self-tuning, control mode selection under device or instrument fault, automated learning, on-line fault monitoring and failure anticipation, and supervisory control. The paper describes the integration of control strategies, fault-tolerant control, and sensor placement for the IRIS system, and demonstration of the technology using an experimental control loop.

Development of Code, PNFENT, Based on Using Finite Elements for Neutron Transport

A variational treatment of the finite element method for neutron transport is used based on a version of the even parity Boltzman equation for the general case of anisotropic scattering and sources. The theory of maximum principles is based on the Cauchy-Schwartz inequality and the properties of a leakage operator G and a removal operator C. For system with extraneous sources a maximum principle is used in boundary free form to ease finite element computations. The global error of an approximate variational solution is shown. The energy dependence of the angular flux is treated by the multi-group method. In this paper the spatial dependence of the angular flux is given in a finite element representation. The directional dependence of angular flux is represented preferably by a spherical harmonic expansion. The above method has been developed and implemented in the finite element program PNFENT. A homogenous slab of a pure absorber along edge-cell and a two dimensional problems are solved with an accuracy as good as the best problem techniques.

Analysis of ABWR Critical Control Operation by TRACG Code

Hitachi utilized three-dimensional transient analysis to design and verify a critical-control mode algorithm of an automatic power regulator (APR). TRACG has a three-dimensional neutron kinetics model based on diffusion theory and a six-equation two-phase flow model. To verify the APR critical-control mode algorithm, an external-neutron-source model that makes possible to simulate a sub-critical initial core, and an APR system model were developed and added on TRACG. The code was verified by comparison of measurements and calculation results of ABWR start-up operation under the critical-control mode. The modified TRACG could simulate neutron count rates of start-up-range neutron monitors (SRNMs), reactor period, control rod operation timing, CR withdrawal length, and time of criticality declaration, well.

Treatment of Noise in Experimental Transport Measurements Plots With Discrete Wavelet Transforms

In this work it is applied the wavelet transform method [2] in order to reduce diverse type of noises of experimental measurement plots in transport theory. First, suppose that a problem is governed by the transport equation for neutral particles, and an unknown perturbation occurs. In this case, the perturbation can be associated to the source, or even to the flux inside the domain X. How is the behavior of the perturbed flux in relation to the flux without the perturbation? For that, we employ the wavelet transform method in order to compress the angular flux considered as a 1D, or n-th dimensional signal ψ. The compression of this signal can be performed up to some a convenient order (that depends of the length of the signal). Now, the transport signal is decomposed as [9, 11]: ψ=〈am|dm|dm−1|dm−2|⋯|d2|d1〉 where ak represents the sub signal of k-th level generated by the low-pass filter associated to the discrete wavelet transform (DWT) chosen, and dk the sub signal of k-th level generated by the high-pass filter associated to the same DWT. It is applied basically the Haar, Daub4 and Coiflet wavelets transforms. Indeed, the sub signal am cumulates the energy, for this work of order 96% of the original signal ψ. A thresholding algorithm provides treatment for the noise, with significant reduction in the compressed signal. Then, it is established a comparison with a base of data in order to identify the perturbed signal. After the identification, it is recomposed the signal applying the inverse DWT. Many assumptions can be established: the rate signal-to-noise is properly high, the base of data must contain so many perturbed signals all with the same level of compression. The problem considered is for perturbations in the signal. For measurements the problem is similar, but in this case the unknown perturbations are generated by the apparatus of measurements, problems in experimental techniques, or simply by random noises. With the same above assumptions, the DWT is applied. For the identification, it is used a method evolving statistical and metric techniques. It is given some results obtained with an algebraic computer system.

Modernization of the Safety I&C at the Dukovany VVER 440/213 Nuclear Power Plant

The modernization of the Dukovany nuclear power plant (four VVER 440 MWe reactor units owned by CˇEZ, the Czech national utility) is presented with a special focus on the digital safety instrumentation and control (I&C) system. The first Unit has been successfully modernized in compliance with the initial schedule. The following matters are further discussed in this paper: • Goal, scope and industrial organization of this modernization, • Main design criteria and I&C architecture, • Digital technologies in use, • Design and Licensing processes, • On-site installation strategy and main milestones, • Progress of work.