15 research outputs found
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Guidelines for the verification and validation of expert system software and conventional software: Validation scenarios. Volume 6
This report is the sixth volume in a series of reports describing the results of the Expert System Verification and Validation (V&V) project which is jointly funded by the US Nuclear Regulatory Commission and the Electric Power Research Institute. The ultimate objective is the formulation of guidelines for the V&V of expert systems for use in nuclear power applications. This activity was concerned with the development of a methodology for selecting validation scenarios and subsequently applying it to two expert systems used for nuclear utility applications. Validation scenarios were defined and classified into five categories: PLANT, TEST, BASICS, CODE, and LICENSING. A sixth type, REGRESSION, is a composite of the others and refers to the practice of using trusted scenarios to ensure that modifications to software did not change unmodified functions. Rationale was developed for preferring scenarios selected from the categories in the order listed and for determining under what conditions to select scenarios from other types. A procedure incorporating all of the recommendations was developed as a generalized method for generating validation scenarios. The procedure was subsequently applied to two expert systems used in the nuclear industry and was found to be effective, given that an experienced nuclear engineer made the final scenario selections. A method for generating scenarios directly from the knowledge base component was suggested
Advanced Reactor Physics Methods for Transient Analysis of Boiling Water Reactors
The safety assessment of nuclear power plants requires the consideration of the several physical phenomena taking place in the reactor core. Since the last decade, the simulations are performed using the so-called Best-Estimate (BE) thermal-hydraulic system codes coupled with 3D nodal diffusion reactor dynamics solvers. These numerical tools are well developed and validated for PWR but for BWR there are still improvements and developments to be done. In addition, homogenized cross sections libraries, depending on the individual thermal-hydraulic state parameters, e.g. fuel temperature, moderator density, etc., must be supplied to take into account the interdependencies of the involved physics.
In this PhD a new coupled code system named TRADYN was developed that consists of the novel and innovative integration of the core simulator DYN3D into the code system TRACE/PARCS using compiler preprocessor directives. This has the advantage to preserve the original TRACE/PARCS system “untouch” and to facilitate the maintenance, modification and debugging. In order to manage the transfer information between TRACE and DYN3D, a General Interface and Specific Data Map routines in FORTRAN were developed. To properly describe the tightly-coupled neutronics and thermal-hydraulic phenomena within the core of a BWR, different physical models were extended and new ones implemented in DYN3D and PARCS.
Another contribution of this doctoral thesis is the development of a new in-house methodology called GENSIM-XS for the generation of nodal cross sections for BWRs considering history effects of control rods and void. GENSIM-XS is able to greatly simplify the number of the neutronics regions present in the reactor core. This new methodology uses the AUDIT option of SIMULATE-3 (S3) to report the cross sections on the output file. Then, they are extracted and written in multi-group tabulated cross sections in NEMTAB format in an automatized manner using Python scripts.
The validation of TRADYN is based on Boiling Water Reactor Turbine Trip (TT) benchmark data, where TRADYN has demonstrated its capability to predict the stationary plant conditions as well as the temporal evolution of the main plant parameters, showing a good agreement with the measurements e.g. core power and dome pressure.
The static core of the cycle 4 of BWR Laguna Verde nuclear power plant was selected as second case for the validation not only of the GENSIM-XS methodology but also of TRADYN. The TRADYN results such as k-eff, axial power profile, axial void fraction exhibit an excellent agreement with the reference values. This underpins the capabilities of the new methodology to generate cross-sections for coupled thermal-hydraulic/neutronics calculations accounting also for history effects for BWR core analysis
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Requirements for advanced simulation of nuclear reactor and chemicalseparation plants.
This report presents requirements for advanced simulation of nuclear reactor and chemical processing plants that are of interest to the Global Nuclear Energy Partnership (GNEP) initiative. Justification for advanced simulation and some examples of grand challenges that will benefit from it are provided. An integrated software tool that has its main components, whenever possible based on first principles, is proposed as possible future approach for dealing with the complex problems linked to the simulation of nuclear reactor and chemical processing plants. The main benefits that are associated with a better integrated simulation have been identified as: a reduction of design margins, a decrease of the number of experiments in support of the design process, a shortening of the developmental design cycle, and a better understanding of the physical phenomena and the related underlying fundamental processes. For each component of the proposed integrated software tool, background information, functional requirements, current tools and approach, and proposed future approaches have been provided. Whenever possible, current uncertainties have been quoted and existing limitations have been presented. Desired target accuracies with associated benefits to the different aspects of the nuclear reactor and chemical processing plants were also given. In many cases the possible gains associated with a better simulation have been identified, quantified, and translated into economical benefits
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An on-line advanced plant simulator (OLAPS)
A PC based on-line advanced plant simulator (OLAPS)
for high quality simulations of Portland General
Electric's Trojan Nuclear Facility is presented. OLAPS
is designed to simulate the thermal-hydraulics of the
primary system including core, steam generators, pumps,
piping and pressurizer. The simulations are based on a
five equation model that has two mass equations, two
energy equations, and one momentum equation with a drift
flux model to provide closure. A regionwise point
reactor kinetics model is used to model the neutron
kinetics in the core. The conservation equations,
constitutive models and the numerical methods used to
solve them are described. OLAPS results are compared
with data from chapter 15 of the Trojan Nuclear
Facility's final safety analysis report