Watts Bar Unit Research Articles

A Multilevel Approach for Direct Core Calculation Schemes

A method of dynamic homogenization was recently proposed as an alternative technique for three-dimensional (3D) core calculations using a direct approach. This technique allows for producing homogenization parameters for a subdomain within the core at its actual state, and offers the advantage of avoiding off-line calculations, cross-section interpolations, and expensive 3D transport calculations. The methodology has shown a remarkable reduction in computational cost compared to 3D transport. However, the application of direct calculation schemes in multiphysics and multicycle problems still demands an extensive use of computational resources. Within the framework of traditional two-step calculations, several multilevel schemes have been developed to accelerate lattice calculations for the generation of the multiparameter cross-section libraries. With a similar objective, this work investigates how the multilevel approach can be applied within the framework of direct calculation schemes and discusses how depletion calculations may be performed. The objective is to enhance their performance and reduce memory requirements in favor of a high-fidelity model of the matter behavior in the reactor. We have then identified suitable flux solvers for each level and explored various homogenization options. We tested the methodology in a 3D pressurized water reactor core problem inspired by the TVA Watts Bar Unit 1 Multi-Physics Multi-Cycle OECD/NEA Benchmark, and performed a comparative analysis to assess the accuracy and computational efficiency of these new calculation schemes against direct single-level calculations and a two-step calculation based on pin-by-pin homogenization, as well as Monte Carlo simulations. Our analysis proves that the computational resources required to solve the full-core problem using a multilevel scheme are significantly reduced, and the best computational features are provided by multilevel dynamic homogenization. Furthermore, we observed that the multilevel approach not only allows for speedup, but can also be advantageously applied in order to improve the accuracy of the final solution of a calculation scheme when the core solver relies on an approximate transport operator.

High Fidelity Whole Core Reactor Eigenvalue Calculation using the Hybrid Monte Carlo and Deterministic RAPID Code System

This paper examines the accuracy and performance of the RAPID (Real-time Analysis for Particle transport and In-situ Detection) code system for whole core reactor simulation. The RAPID code is an implementation of the Multistage Response function Transport (MRT) methodology resulting in a hybrid Monte Carlo and deterministic technique. This paper utilizes the Watts Bar Unit 1 benchmark as the computational model to exam the accuracy and efficiency of RAPID compared to Serpent. The RAPID calculated eigenvalue and 3-D fission density distribution are compared to the results from Serpent. This paper shows that RAPID is able to obtain pin-wise, axially-dependent fission density in 7 minutes using one computer core, as opposed to assembly-wise, axially dependent fission density in 8 days on 40 computer cores.

Determination of the spent fuel decay heat with the VERA core simulator

VERA core simulator capabilities to predict spent fuel decay heat have been explored. The comparison with the Serpent2 and NEWT/TRITON sequence from the SCALE package, performed on the NPP Krško fuel test cases, has enabled independent verification of the successful implementation of the VERA depletion/decay calculation sequence applied on the extended nuclide set. Since the VERA is a 3-D core simulator, the focus in the rest of the paper was on the investigation of the 3-D effects and non-linearity of the soluble boron concentration averaging process, which are difficult to assess with most of available computational tools. A difference in the decay heat prediction between the 3-D and 2-D model approaches was analyzed. In addition, the impact of the Inconel grids on the decay heat was determined. Averaging the concentration of soluble boron over the fuel irradiation period has been found to be a viable approach, since it results in slight decay heat overprediction, which is considered conservative for most applications. Finally, demonstration on the Watts Bar Unit 1 cycle 1 depletion has shown formidable VERA capabilities to accurately predict spent fuel decay heat.

The World Casts New Eyes on Nuclear

Images conjured by the mere mention of the term “nuclear energy” are almost always negative. From Chernobyl to Fukushima, the power source has a historically bad reputation and the stigma of catastrophe. Winston Churchill knew those that failed to learn from history were doomed to repeat it, and perhaps it is lessons learned that have so many countries once again casting a keen eye over conventional nuclear energy as a potential key player in the low-carbon future. For the first time in more than a decade, most Japanese are in support of restarting idled nuclear facilities, according to a poll conducted by Nikkei. Rising power prices and the potential for electricity shortages in Tokyo are said to have swayed many back to being in favor of nuclear power just 10 years after the Fukushima disaster. The poll found 53% of respondents in favor of restarting the reactors if safety could be assured, while 38% said they should remain shut. With Russia continuing its invasion into neighboring Ukraine, many European countries are looking to rid themselves of dependence on Russian natural gas imports. Many have grown so dependent they cannot fully shut the stream off for fear of economic collapse, however several are attempting to rework their own energy policies and favoring nuclear as a solution. Both France and the UK have reversed course on nuclear in recent months. Just ahead of Russia’s invasion of Ukraine, French President Emmanuel Macron called for a nuclear renaissance in his country after earlier vowing to cut its reliance on nuclear power. The UK, which gets 15% of its electricity from nuclear power, had vowed to shutter all but one of its plants ahead of the Russian invasion. More recently, Prime Minister Boris Johnson is favoring the construction of new plants and boosting Britain’s electrical generation from nuclear to 25%. As for the US, at the end of 2021 there were 55 commercially operating nuclear power plants with 93 nuclear power reactors in 28 states. The newest nuclear plant to enter service in the US was Watts Bar Unit 2, which began commercial operations in 2016. Two new nuclear reactors are actively under construction: Georgia Power’s Vogtle Units 3 and 4 in Waynesboro, Georgia. A recent Pew survey found that while a majority of Americans (69%) favor the US taking steps to become carbon neutral by 2050, only around a third of US adults (35%) say the federal government should encourage the production of nuclear power, while about a quarter (26%) say the government should discourage it (Fig. 1). This survey was fielded before the Russian invasion and the renewed discussions about nuclear power it has prompted. According to the US Energy Department’s Office of Nuclear Energy, nuclear power currently provides 52% of the US’s carbon-free electricity, and the Biden administration has identified the current fleet of 93 reactors as a vital resource to achieve net-zero emissions economywide by 2050.

OVERVIEW OF THE OECD-NEA EXPERT GROUP ON MULTI-PHYSICS EXPERIMENTAL DATA, BENCHMARKS AND VALIDATION

The OECD Nuclear Energy Agency (NEA) Nuclear Science Committee (NSC) established the Expert Group on Multi-physics Experimental Data, Benchmarks and Validation (EGMPEBV) in 2014 to bridge the gap between advanced, multi-physics simulation capabilities and the relatively low availability of dedicated, high-fidelity experimental data and benchmarks specifically for multi-physics modelling and simulation tools. The EGMPEBV was mandated to establish mechanisms for the certification of experimental data and benchmark models and to establish the processes and procedures for the validation of multi-physics modelling and simulation tools. The EGMPEBV oversees three task forces, covering (1) experimental data qualification and benchmark evaluation, (2) validation guidelines and needs and (3) example application of validation experiments. These have generated numerous reports surveying the state-of-the-art in multi-physics validation, challenge areas and recommendations for the evaluation of multi-physics benchmarks, while in parallel developing the specifications for multi-physics benchmarks. Three benchmark specifications are in active development, including a reactivity compensation scenario in the Rostov Unit 2 VVER-1000, multi-cycle depletion of the TVA Watts Bar Unit 1 and study of pellet cladding mechanical interaction within ramp tests performed at the Studsvik R2 reactor. We provide an overview of the recent progress in these areas and a summary of the future activities of the EGMPEBV in establishing international multi-physics benchmarks.

Secondary-Source Core Reload Modeling with VERA

The CASL reactor simulation package VERA has been adapted to provide high-fidelity simulation capabilities for modeling source range detector response during subcritical reactor configurations. New features include the activation and shuffling of secondary-source assemblies, use of burned fuel neutron emission data from the ORIGEN depletion solver to the MPACT deterministic neutron transport solver, allowance of user-defined sources in MPACT based on material composition, ability to solve the subcritical source-driven system with neutron multiplication using the MPACT diffusion solver, and transfer of the calculated fission source from MPACT to the continuous-energy Monte Carlo solver Shift for final detector response evaluation using the CADIS methodology for variance reduction. These new capabilities were benchmarked against Watts Bar Unit 1 plant operating data for the first few fuel loading steps and were found to demonstrate excellent agreement with the measured data.

Coupled fuel performance calculations in VERA and demonstration on Watts Bar unit 1, cycle 1

As the core simulator capabilities in the Virtual Environment for Reactor Applications (VERA) have become more mature and stable, increased attention has been focused on coupling Bison to provide fuel performance simulations. This technique has been a very important driver for the pellet-clad interaction challenge problem being addressed by the Consortium for Advanced Simulation of Light Water Reactors (CASL). In this article, two coupling approaches are demonstrated on quarter core problems based on Watts Bar Nuclear Plant Unit 1, Cycle 1: (1) an inline approach in which a one-way coupling is used between neutronics/thermal hydraulics (through MPACT/CTF) and fuel performance (through Bison) but no fuel temperature information is passed back to MPACT/CTF, and (2) a two-way approach (coupled) in which the fuel temperature is passed from Bison to MPACT/CTF. In both approaches, power and temperature distributions from MPACT/CTF are used to inform the Bison simulations for each rod in the core.The demonstrations presented here are the first integrated fuel performance simulations in VERA, which opens many possibilities for future work, including applications to accident-tolerant fuel efforts and transient simulations, which are of critical importance to CASL. These demonstrations also highlight the potential to move away from the current Bison-informed fuel temperature lookup table approach, which is the default in MPACT/CTF simulations, if performance improvements are made in the near future.

Assessment of the analysis capability for core-wide PWR pellet-clad interaction screening of Watts Bar Unit 1

This article demonstrates the application of the Virtual Environment for Reactor Applications simulation environment to screen for pellet-clad interaction amongst the core-wide fuel rods within a pressurized water reactor, with emphasis on the multi-dimensional analysis capabilities of the BISON fuel performance code. The VERA codes utilized in this effort include the BISON fuel performance code and the VERA sub-element, which combines the MPACT neutron transport code and the COBRA-TF thermal-hydraulics code. This effort has involved screening Cycles 6 and 7 of Watts Bar Unit 1, which were selected on the basis that pellet-clad interaction failures had occurred during reactor start up in both cycles, despite the moderate power levels. These failure instances at Watts Bar Unit 1 offered an interesting test of the code’s veracity and a particularly challenging condition for this assessment.

Pellet-clad mechanical interaction screening using VERA applied to Watts Bar Unit 1, Cycles 1–3

The Consortium for Advanced Simulation of Light Water Reactors (CASL) aims to provide high-fidelity multiphysics simulations of light water nuclear reactors. To accomplish this, CASL is developing the Virtual Environment for Reactor Applications (VERA), which is a suite of code packages for thermal hydraulics, neutron transport, fuel performance, and coolant chemistry. As VERA continues to grow and expand, there has been an increased focus on incorporating fuel performance analysis methods. One of the primary goals of CASL is to estimate local cladding failure probability through pellet-clad interaction, which consists of both pellet-clad mechanical interaction (PCMI) and stress corrosion cracking. Estimating clad failure is important to preventing release of fission products to the primary system and accurate estimates could prove useful in establishing less conservative power ramp rates or when considering load-follow operations.While this capability is being pursued through several different approaches, the procedure presented in this article focuses on running independent fuel performance calculations with BISON using a file-based one-way coupling based on multicycle output data from high fidelity, pin-resolved coupled neutron transport–thermal hydraulics simulations. This type of approach is consistent with traditional fuel performance analysis methods, which are typically separate from core simulation analyses. A more tightly coupled approach is currently being developed, which is the ultimate target application in CASL.Recent work simulating 12 cycles of Watts Bar Unit 1 with VERA core simulator are capitalized upon, and quarter-core BISON results for parameters of interest to PCMI (maximum centerline fuel temperature, maximum clad hoop stress, and minimum gap size) are presented for Cycles 1–3. Based on these results, this capability demonstrates its value and how it could be used as a screening tool for gathering insight into PCMI, singling out limiting rods for further, more detailed analysis.

MC21/CTF and VERA multiphysics solutions to VERA core physics benchmark progression problems 6 and 7

Abstract The continuous energy Monte Carlo neutron transport code, MC21, was coupled to the CTF subchannel thermal-hydraulics code using a combination of Consortium for Advanced Simulation of Light Water Reactors (CASL) tools and in-house Python scripts. An MC21/CTF solution for VERA Core Physics Benchmark Progression Problem 6 demonstrated good agreement with MC21/COBRA-IE and VERA solutions. The MC21/CTF solution for VERA Core Physics Benchmark Progression Problem 7, Watts Bar Unit 1 at beginning of cycle hot full power equilibrium xenon conditions, is the first published coupled Monte Carlo neutronics/subchannel T-H solution for this problem. MC21/CTF predicted a critical boron concentration of 854.5 ppm, yielding a critical eigenvalue of 0.99994 ± 6.8E-6 (95% confidence interval). Excellent agreement with a VERA solution of Problem 7 was also demonstrated for integral and local power and temperature parameters.

VERA Core Simulator Methodology for Pressurized Water Reactor Cycle Depletion

This paper describes the methodology developed and implemented in the Virtual Environment for Reactor Applications Core Simulator (VERA-CS) to perform high-fidelity, pressurized water reactor (PWR), multicycle, core physics calculations. Depletion of the core with pin-resolved power and nuclide detail is a significant advance in the state of the art for reactor analysis, providing the level of detail necessary to address the problems of the U.S. Department of Energy Nuclear Reactor Simulation Hub, the Consortium for Advanced Simulation of Light Water Reactors (CASL). VERA-CS has three main components: the neutronics solver MPACT, the thermal-hydraulic (T-H) solver COBRA-TF (CTF), and the nuclide transmutation solver ORIGEN. This paper focuses on MPACT and provides an overview of the resonance self-shielding methods, macroscopic-cross-section calculation, two-dimensional/one-dimensional (2-D/1-D) transport, nuclide depletion, T-H feedback, and other supporting methods representing a minimal set of the capabilities needed to simulate high-fidelity models of a commercial nuclear reactor. Results are presented from the simulation of a model of the first cycle of Watts Bar Unit 1. The simulation is within 16 parts per million boron (ppmB) reactivity for all state points compared to cycle measurements, with an average reactivity bias of <5 ppmB for the entire cycle. Comparisons to cycle 1 flux map data are also provided, and the average 2-D root-mean-square (rms) error during cycle 1 is 1.07%. To demonstrate the multicycle capability, a state point at beginning of cycle (BOC) 2 was also simulated and compared to plant data. The comparison of the cycle 2 BOC state has a reactivity difference of +3 ppmB from measurement, and the 2-D rms of the comparison in the flux maps is 1.77%. These results provide confidence in VERA-CS's capability to perform high-fidelity calculations for practical PWR reactor problems.

Long-stalled U.S. nuclear plant set to come on-line

For the first time in 20 years, a new commercial nuclear power plant will come on-line in the U.S. when Watts Bar Unit 2 in southeastern Tennessee begins operations later this summer. The plant will be the first to meet new regulations from the Nuclear Regulatory Commission after the 2011 earthquake and tsunami that damaged reactors at the Fukushima Daiichi nuclear power plant in Japan, the Energy Information Administration (EIA) says. After a decades-long delay, construction of the plant, which began in 1973, concluded in 2015 at a total cost of $4.7 billion. The plant is expected to add 1,150 MW of electricity generating capacity to the area. Together with Watts Bar Unit 1, which began operating in 1996, Watts Bar Unit 2 will generate enough electricity to supply 1.3 million homes daily, according to Tennessee Valley Authority, a federally owned electric utility. EIA says that within the next four

Optimization and parallelization of the thermal–hydraulic subchannel code CTF for high-fidelity multi-physics applications

This paper describes major improvements to the computational infrastructure of the CTF subchannel code so that full-core, pincell-resolved (i.e., one computational subchannel per real bundle flow channel) simulations can now be performed in much shorter run-times, either in stand-alone mode or as part of coupled-code multi-physics calculations. These improvements support the goals of the Department Of Energy Consortium for Advanced Simulation of Light Water Reactors (CASL) Energy Innovation Hub to develop high fidelity multi-physics simulation tools for nuclear energy design and analysis.A set of serial code optimizations—including fixing computational inefficiencies, optimizing the numerical approach, and making smarter data storage choices—are first described and shown to reduce both execution time and memory usage by about a factor of ten. Next, a “single program multiple data” parallelization strategy targeting distributed memory “multiple instruction multiple data” platforms utilizing domain decomposition is presented. In this approach, data communication between processors is accomplished by inserting standard Message-Passing Interface (MPI) calls at strategic points in the code. The domain decomposition approach implemented assigns one MPI process to each fuel assembly, with each domain being represented by its own CTF input file. The creation of CTF input files, both for serial and parallel runs, is also fully automated through use of a pressurized water reactor (PWR) pre-processor utility that uses a greatly simplified set of user input compared with the traditional CTF input.To run CTF in parallel, two additional libraries are currently needed: MPI, for inter-processor message passing, and the Parallel Extensible Toolkit for Scientific Computation (PETSc), which is used to solve the global pressure matrix in parallel. Results presented include a set of testing and verification calculations and performance tests assessing parallel scaling characteristics up to a full-core, pincell-resolved model of a PWR core containing 193 17×17 assemblies under hot full-power conditions. This model, representative of Watts Bar Unit 1 and containing about 56,000 pins, was modeled with roughly 59,000 subchannels, leading to about 2.8 million thermal–hydraulic control volumes in total. Results demonstrate that CTF can now perform full-core analysis of a PWR (not previously possible owing to excessively long runtimes and memory requirements) on the order of 20min. This new capability not only is useful to stand-alone CTF users, but also is being leveraged in support of coupled code multi-physics calculations being done in the CASL program.

Acoustic emission/flaw relationships for inservice monitoring of LWRs : Pacific Northwest Laboratory, Richland, Washington (United States), NUREG/CR-5645, PNL-7479, 120 pp. (Oct. 1991)

The program concerning Acoustic Emission/Flaw Relationships for Inservice Monitoring of LWRs was initiated in FY76 with the objective of validating the application of acoustic emission (AE) to monitor nuclear reactor pressure-containing components during operation to detect cracking. The program has been supported by the US Nuclear Regulatory Commission, Office of Nuclear Regulatory Research. Research and development has been performed by Pacific Northwest Laboratory, operated for the Department of Energy by Battelle Memorial Institute. The program has shown the feasibility of continuous, on-line AE monitoring to detect crack growth and produced validated methods for applying the technology. Included are relationships for estimating flaw severity from AE data and field applications at Watts Bar Unit 1 Reactor, Limerick Unit 1 Reactor, and the High Flux Isotope Reactor. This report discusses the program scope and organization, the three program phases and the results obtained, standard and code activities, and instrumentation and software developed under this program.

Acoustic emission system calibration at Watts Bar Unit 1 nuclear reactor

An acoustic emission system has been installed on TVA's Watts Bar Unit 1 reactor to monitor selected areas of the pressure boundary during hydrostatic testing, hot functional testing, and ultimately, during reactor startup and operation. This is a part of a cooperative effort between TVA and NRC Research to test and demonstrate AE technology. This technology has been developed under an NRC Research program to validate the application of AE techniques for continuous, on-line monitoring of reactor pressure boundaries to detect cracking. This report discusses the performance of and results from a special calibration test of the AE system using simulated AE signals to evaluate the accuracy of signal source location by the system.