Purdue Advanced Reactor Core Simulator Research Articles

Neutronics–Thermal-Hydraulics–Coupled Transient Analysis for Reactor Power Change in an Inclined Offshore Floating Boiling Water Reactor

An application of the boiling water reactor (BWR) to an offshore floating nuclear power plant (OFNP) in Japan is discussed. The BWR-type OFNP has some challenges for practical use, although it has high economic efficiency because of downsizing and simplification. One challenge is understanding reactor kinetics under conditions specific to the marine environment. This study quantitatively clarifies the total and spatial changes in power when the BWR is inclined during regular operation. Therefore, the TRACE (TRAC/RELAP Advanced Computational Engine) and PARCS (Purdue Advanced Reactor Core Simulator) codes were used to perform a three-dimensional neutronics–thermal-hydraulics–coupled transient analysis. The calculation model is based on Peach Bottom II. This study clarifies the changing trend in total and local BWR power by inclination with simplified modeling and conditions. The reasons for such changes are discussed based on changes in several thermal-hydraulic parameters. The difference in BWR power against the inclinations is small. Thus, it is implied that the BWR-type OFNP is expected to have a stable power supply capability during natural disasters. However, to confirm the power stability of the BWR reactor under a full range of offshore external conditions, further research is required. A description of additional research needs that would further support the safety case for this reactor design are discussed.

Verification of the Coupled Code PARCS/TWOPORFLOW with Rod Ejection Accident Calculations for Small Modular Reactors

Research on small modular reactors (SMRs) is gaining importance since they are key for addressing energy challenges in various sectors. These types of reactors integrate novel technologies that rely heavily on passive safety systems. Among the most developed light water reactors, SMR designs are SMART and NuScale. This work analyzes the academic boron-free Karlsruhe Small Modular Reactor (KSMR), which may fit in SMART, and a core design resembling NuScale. The research aims to explore the potential of new multiphysics tools under development to predict safety parameters of SMRs during normal operation and transients. For this purpose, the Purdue Advanced Reactor Core Simulator (PARCS) and TWOPORFLOW (TPF) codes have been coupled using the Interface for Code Coupling (ICoCo), orchestrating code execution through a C++ Supervisor program. In this work, a nodewise rod ejection accident (REA) has been analyzed for two different scenarios. The first is a hot-zero-power scenario for the KSMR core, and the second is initiated at 75% of nominal power for the NuScale core. Verification of results has been done though code-to-code comparison. Comparison of the PARCS/TPF results obtained for both KSMR and NuScale with reference cases shows acceptable differences. Key safety parameters predicted by the codes for the REA analysis of both cores have also been evaluated against the latest U.S. Nuclear Regulatory Commission regulation for reactivity initiated accidents showing that all parameters fulfill core coolability acceptance criteria and fuel rod cladding failure thresholds. The presented work highlights the transition from the coupling of PARCS with a mixture model code, such as SCF, to a coupling with a two-phase model code, such as TPF. These findings contribute to a better understanding of SMR phenomenology during accidental sequences and demonstrate the capabilities of the coupled codes.

The evaluation of TRACE/PARCS model for BWR-4 nuclear power plant by startup test transient analyses

Abstract Generally, the thermal hydraulic (TH) codes need the results of Neutron Kinetics (NK) codes providing the reactivity properties to calculate neutron flux. Then the TH codes perform the safety analyses obtaining the responses of pressure, temperature, or water level. Two kinds of different codes calculate different physical behaviors sequentially and separately. Simultaneously computing thermal hydraulic and neutron kinetics behaviors can enhance the accuracy of the analysis. Hence, it is crucial to develop the TH-NK coupled model. This study presents the capability of the TH-NK coupled model, developed by TRACE (TRAC/RELAP Advanced Computational Engine) and PARCS (Purdue Advanced Reactor Core Simulator), for the BWR-4 nuclear power plant. The establishment of the TRACE/PARCS model presented the nodal and component modeling methodologies. This model was used to simulate two startup tests of high power level system transients. Principal system responses, calculated by the TRACE/PARCS model, were compared with the measured data in startup tests and the results of the point kinetic calculation of the TRACE (TRACE/PK) to evaluate the model. The evaluation shows that the TRACE/PARCS model can simulate the interaction between thermal hydraulic and neutron kinetics phenomena and predict the transients suitably. Through the comparison, the TRACE/PARCS model can be confident doing the analyses of normal and abnormal operational transients to predict the transient responses.

Core Modeling and Simulation of Peach Bottom 2 Turbine Trip Test 2 Using CASMO5/TRACE/PARCS

The Purdue Advanced Reactor Core Simulator (PARCS) three-dimensional neutron kinetics code and the TRACE nuclear systems analysis code were interfaced. This provides a best-estimate coupled code system for performing transient plant calculations with reactivity feedback from a detailed core model, significantly contributing to nuclear power plant safety analyses. This study performed steady-state and transient simulations of Peach Bottom 2 Turbine Trip Test 2 (PB2 TT2) using the CASMO5/TRACE/PARCS coupled code. Consequently, CASMO5/TRACE/PARCS simulates the rapid positive reactivity addition caused by the sudden closure of the turbine stop valve. Specifically, the discrepancy in the maximum total power during the transient condition was within 3% compared with the PB2 TT2 experimental data. Furthermore, the sensitivity of the thermal-hydraulic channel (CHAN) component modeling in the coupled CASMO5/TRACE/PARCS code revealed that the number of CHAN components influenced the assembly radial power peaking factor in the PB2 TT2 transient calculation.

A Rigorous Method for Selecting the Most Limiting Exposure During Cycle for Anticipated Transient Without SCRAM with Instability Analysis

The U.S. Nuclear Regulatory Commission (NRC) staff often perform confirmatory analyses using the TRAC/RELAP Advanced Computational Engine (TRACE) and Purdue Advanced Reactor Core Simulator (PARCS) codes to assist in regulatory decision making. Recently, the NRC staff have performed numerous such analyses of anticipated transient without SCRAM (ATWS) with core instability (ATWS-I) scenarios for boiling water reactor license amendment requests to expand the power/flow operating domain. In the conduct of these confirmatory analyses, the staff have simulated oscillatory conditions in the reactor core under certain ATWS conditions that result in regional mode (or out-of-phase mode) power oscillations. The nature of these regional oscillations may present a challenge to fuel damage limits. Therefore, there has been interest in methods to identify the most limiting point in cycle exposure. It has been conventional wisdom that the core is most susceptible to regional mode oscillations when the fission cross section is greatest, leading to the common practice of analyzing these events at the peak hot excess (PHE) exposure point in the cycle. The staff have found some limitations in applying the PHE concept in a consistent manner. In the current work, the NRC staff have developed a more rigorous method for identifying the most limiting cycle exposure by directly considering the core flow rate, the axial power distribution, the first harmonic mode shape, and the eigenvalue separation between the fundamental and first harmonic modes. This method is a more rigorous method to screen the various exposures between beginning and end of cycle. An example case is shown to demonstrate the application of this methodology.

Electromagnetism Mechanism for Enhancing the Refueling Cycle Length of a WWER-1000

Abstract Increasing the operation cycle length can be an important goal in the fuel reload design of a nuclear reactor core. In this research paper, a new optimization approach, electromagnetism mechanism (EM), is applied to the fuel arrangement design of the Bushehr WWER-1000 core. For this purpose, a neutronic solver has been developed for calculating the required parameters during the reload cycle of the reactor. In this package, two modules have been linked, including PARCS v2.7 and WIMS-5B codes, integrated in a solver for using in the fuel arrangement optimization operation. The first results of the prepared package, along with the cycle for the original pattern of Bushehr WWER-1000, are compared and verified according to the Final Safety Analysis Report and then the results of exploited EM linked with Purdue Advanced Reactor Core Simulator (PARCS) and Winfrith Improved Multigroup Scheme (WIMS) codes are reported for the loading pattern optimization. Totally, the numerical results of our loading pattern optimization indicate the power of the EM for this problem and also show the effective improvement of desired parameters for the gained semi-optimized core pattern in comparison to the designer scheme.

A novel multi objective Loading Pattern Optimization by Gravitational Search Algorithm (GSA) for WWER1000 core

In this paper, a novel multi objective optimization algorithm, Gravitational Search Algorithm (GSA), is developed in order to implement in the Loading Pattern Optimization (LPO) of a nuclear reactor core. In recent decades several metaheuristic algorithms or computational intelligence methods have been expanded to optimize reactor core loading pattern. Regarding the coupled behavior of Neutronic and Thermal-Hydraulic (NTH) dynamics in a nuclear reactor core, proper loading pattern of fuel assemblies (FAs) depends on NTH aspects, simultaneously. Thus, obtaining optimal arrangement of FAs, in a core to meet special objective functions is a complex problem. Gravitational Search Algorithm (GSA) is constructed based on the law of Gravity and the notion of mass interactions, using the theory of Newtonian physics and searcher agents are the collection of masses. In this work, for multi objective optimization, the NTH aspects are included in fitness function. Neutronic goals include increasing multiplication factor (Keff), decreasing of power picking factor (PPF) and flattening of the power density, also thermal–hydraulic (TH) goals include increasing critical heat flux (CHF) and decreasing average of fuel centers temperature. Therefore, at the first step, GSA method is compared with other metaheuristic algorithms on Shekel's Foxholes problem. In the second step for finding the best core pattern and implementation of the objectives listed, GSA algorithm has been performed for case of WWER1000 reactor. For the NTH calculations, PARCS (Purdue advanced reactor core simulator) and COBRA-EN codes are implemented, respectively. The results demonstrate that GSA algorithm have promising performance and can propose for other optimization problems of nuclear engineering field.

A novel optimization method, Gravitational Search Algorithm (GSA), for PWR core optimization

In-core fuel management optimization (ICFMO) is one of the most challenging concepts of nuclear engineering. In recent decades several meta-heuristic algorithms or computational intelligence methods have been expanded to optimize reactor core loading pattern. This paper presents a new method of using Gravitational Search Algorithm (GSA) for in-core fuel management optimization. The GSA is constructed based on the law of gravity and the notion of mass interactions. It uses the theory of Newtonian physics and searcher agents are the collection of masses. In this work, at the first step, GSA method is compared with other meta-heuristic algorithms on Shekel’s Foxholes problem. In the second step for finding the best core, the GSA algorithm has been performed for three PWR test cases including WWER-1000 and WWER-440 reactors. In these cases, Multi objective optimizations with the following goals are considered, increment of multiplication factor (Keff), decrement of power peaking factor (PPF) and power density flattening. It is notable that for neutronic calculation, PARCS (Purdue Advanced Reactor Core Simulator) code is used. The results demonstrate that GSA algorithm have promising performance and could be proposed for other optimization problems of nuclear engineering field.

The Modeling of Advanced BWR Fuel Designs with the NRC Fuel Depletion Codes PARCS/PATHS

The PATHS (PARCS Advanced Thermal Hydraulic Solver) code was developed at the University of Michigan in support of U.S. Nuclear Regulatory Commission research to solve the steady-state, two-phase, thermal-hydraulic equations for a boiling water reactor (BWR) and to provide thermal-hydraulic feedback for BWR depletion calculations with the neutronics code PARCS (Purdue Advanced Reactor Core Simulator). The simplified solution methodology, including a three-equation drift flux formulation and an optimized iteration scheme, yields very fast run times in comparison to conventional thermal-hydraulic systems codes used in the industry, while still retaining sufficient accuracy for applications such as BWR depletion calculations. The capability to model advanced BWR fuel designs with part-length fuel rods and heterogeneous axial channel flow geometry has been implemented in PATHS, and the code has been validated against previously benchmarked advanced core simulators as well as BWR plant and experimental data. The modifications to the codes and the results of the validation are described in this paper.

Optimization of refueling cycle length by an enhanced PSO with novel mutation operator

In this paper, a new method for optimizing the fuel arrangement in a WWER-1000 reactor core during refueling cycle is presented. Finding the best configuration corresponding to the desired pattern, an enhanced PSO with a Novel Mutation operator is applied. WIMS and PARCS (Purdue Advanced Reactor Core Simulator) codes are used to calculate the neutronics cross sections and multiplication factor of core with corresponded power peaking factors (PPFs) during burn up cycles, respectively. Cross sections and burn up during cycle length were calculated by WIMS code, then core parameters were calculated by PARCS and finally hybridization of intelligent PSO (Particle Swarm Optimization) method and novel mutation were used to obtain optimal arrangement. The purposed algorithm is based on increasing burn up value and refueling cycle length and by keeping power peaking factor in safe margins. In this way, neutronic parameters of the reactor during operation cycle from BOC (Begin Of Cycle) to EOC (End Of Cycle), were calculated. Implementation of this algorithm has been done in MATLAB. In this case, Bushehr WWER-1000 NPP reactor was studied. The comparison between results and Final Safety Analysis Report (FSAR) data shows good agreement.

The development and assessment of TRACE/PARCS model for Lungmen ABWR

This study consists of two steps. The first step is the development of a TRACE (TRAC/RELAP Advanced Computational Engine)/PARCS (Purdue Advanced Reactor Core Simulator) model of Lungmen nuclear power plant (NPP) which includes the vessel, reactor internal pumps (RIPs), main steam lines, and important control systems (such as the feedwater control system, steam bypass & pressure control system, and recirculation flow control system), etc. Key parameters were identified to refine the TRACE/PARCS model further in the frame of a steady state analysis. The second step is the performance of Lungmen NPP TRACE/PARCS model transient analyses. The transient data of Final Safety Analysis Report (FSAR) chapter 15th were used to compare with the results of Lungmen NPP TRACE/PARCS model. The trends of TRACE/PARCS analysis results were consistent with the FSAR data. It indicated that there was a respectable accuracy in the Lungmen NPP TRACE/PARCS model and it also depicted that the Lungmen NPP TRACE/PARCS model was satisfying for the purpose of Lungmen NPP safety analyses.

Nuclear reactor core optimization with Parallel Integer Coded Genetic Algorithm

The operation and fuel management of reactors are of utmost importance. Core performance analysis constitutes an important stage in core fuel management optimization. In this article, a new algorithm for core optimization on safety limitation of nuclear reactors is introduced. A Parallel Integer Coded Genetic Algorithm, PICGA, is applied to obtain the best configuration corresponding to the maximum multiplication factor. The PICGA may utilize two or more search algorithms along sharing their data, synchronously. This procedure makes the optimization search faster and more efficient. The PICGA is coupled with WIMSD4 and PARCS codes. The WIMSD4 and PARCS (Purdue Advanced Reactor Core Simulator) codes are used to calculate the group constants, fuel depletion, power peaking factors (PPFs) and corresponding effective multiplication factor, Keff. Implementation of the algorithm has been carried out in MATLAB for two reactor test cases, Tehran Research Reactor (TRR) and Bushehr Nuclear Power Plant (BNPP). Using the PICGA, we obtained new core pattern for TRR. Obtaining a configuration for Bushehr Nuclear Power Plant, which maximizes effective multiplication factor increment, is a test case to validate the procedure.

Progress in the development of an implicit steady state solution in the coupled code TRACE/PARCS

This paper describes the implementation of an implicit steady state solution method in the TRAC/RELAP Advanced Computational Engine (TRACE) thermal-hydraulics system code and Purdue Advanced Reactor Core Simulator (PARCS) code with the goal of improving solution stability and efficiency. The implicit steady state solution method has been implemented within the framework of the existing psuedo-transient solution method in TRACE and includes time-dependent thermal-hydraulic and heat transfer equations and time-independent neutron diffusion equations. The implicit solution method uses Newton's method to solve the thermal-hydraulic, heat transfer, and neutron diffusion equations during each psuedo-time step with an analytic construction of the Jacobian matrix. The linear system associated with each iteration of Newton's method is solved using a preexisting LU decomposition algorithm in TRACE. The implicit steady state solution was evaluated using two different meshes overlaid on a two-phase pipe model closely matching a boiling water reactor hydraulic channel. The implicit solution method reproduces the correct steady state solution for varying time step sizes for each mesh. An evaluation of the CPU runtime required to complete a steady state calculation using the implicit method shows that the well-developed and optimized explicit solution method currently requires less CPU runtime than the implicit solution method which has yet to be optimized. These results direct future development of the implicit solution method towards optimization strategies to reduce CPU runtime.

Investigation of PWR core optimization using harmony search algorithms

This work addresses applications of the classical harmony search (HS), improved harmony search (IHS) and the harmony search with differential mutation based pith adjustment (HSDM) to PWR core reloading pattern optimization problems. Proper loading pattern of fuel assemblies (FAs) depends on both neutronic and thermal–hydraulic aspects; obtaining optimal arrangement of fuel assemblies, FA, in a core to meet special objective functions is a complex problem. In this paper, in the first step HS, IHS and HSDM methods are implemented and compared with other meta-heuristic algorithms on Shekel’s Foxholes problem. In the second step to evaluate proposed techniques in PWR cores, maximization of multiplication factor, keff, decreasing of power picking factor (PPF) as much as possible and power density flattening are chosen as neutronic objective functions for two PWR test cases although other variables can be taken into account. In the third step, obtaining maximum core average critical heat flux (CHF) along no void generation throughout the cores are two thermal–hydraulic objective functions which are included to the desired neutronic objective functions. For neutronic and thermal–hydraulic computation, PARCS (Purdue Advanced Reactor Core Simulator) and COBRA-EN codes are used respectively. Coupling the harmony search with the PARCS code and the COBRA-EN code, we developed a core reloading pattern optimization code. The results, convergence rate and reliability of the techniques are quiet promising and show that the harmony algorithms perform very well. Furthermore, it is found that harmony searches have potential for other optimization applications in other nuclear engineering field.

The MSIV Closure Direct Scram Transient Analysis of Lungmen ABWR Using TRACE/PARCS

This study consists of two steps. The first step is the development of a TRACE (TRAC/RELAP Advanced Computational Engine)/PARCS(Purdue Advanced Reactor Core Simulator) model of Lungmen nuclear power plant (NPP) which includes the vessel, reactor internal pumps (RIPs), main steam lines, and important control systems (such as the feedwater control system, steam bypass & pressure control system, and recirculation flow control system), etc.. Key parameters were identified to refine the TRACE/PARCS model further in the frame of a steady state analysis. The second step is the performance of Lungmen NPP TRACE/PARCS model transient analyses. The MSIV closure direct scram (MSIVCD, MSIV = Main Steamline Isolation Valve) transient data of Final Safety Analysis Report (FSAR) is used to verify the Lungmen NPP TRACE/PARCS model. The trends of TRACE/PARCS analysis results are consistent with the FSAR data. It indicates that there is a respectable accuracy in the Lungmen NPP TRACE/PARCS model.

Demonstration of Implicit Coupling of TRACE/PARCS Using Simplified One-Dimensional Problems

Historically, large physics problems have been divided into smaller problems based on the individual physics, typically referred to as operator splitting. The analysis of a nuclear reactor for design-basis accidents is performed by a handful of computer codes each solving a portion of the problem, based on the physics involved. The reactor thermal-hydraulic response to an event is determined using a system code like TRAC RELAP Advanced Computational Engine (TRACE). The core power response to the same accident scenario is determined using a spatial neutron kinetics code like Purdue Advanced Reactor Core Simulator (PARCS). The drive of industry to uprate power for reactors has motivated analysts to move from a conservative approach to design-basis accidents toward a best-estimate method. To achieve a best-estimate calculation, efforts have been aimed at coupling the individual physics models to improve the accuracy of the analysis and reduce margins. The current coupling techniques are sequential in nature (i.e., they treat shared data explicitly in time). During a calculation time-step data are passed between the two codes. The individual codes solve their portions of the calculation and converge to a solution before the calculation is allowed to proceed to the next time step. This paper presents a fully implicit method of simultaneously solving the neutron balance equations, heat conduction equations, and constitutive fluid dynamics equations. The paper also outlines the basic concepts behind the nodal balance equations, heat transfer equations, and thermal-hydraulic equations, which will be coupled to form a fully implicit nonlinear system of equations. It presents a monolithic method for the solution of the implicit equation set. The coupling technique described in this paper was implemented into the TRACE/PARCS coupled code system and is applicable to other similar coupled thermal-hydraulic and core physics reactor safety codes. This technique is demonstrated using coupled input decks to show that the system is solved correctly and then verified by using simple one-dimensional coupled problems. These simplified problems demonstrate the ability of this method to solve nonlinear coupled systems and maintain accuracy while removing time-step dependency of the coupled calculation.

Consistent Comparison of the Codes RELAP5/PARCS and TRAC-M/PARCS for the OECD MSLB Coupled Code Benchmark

A generalized interface module was developed for coupling any thermal-hydraulic code to any spatial kinetic code. In the design used here the thermal-hydraulic and spatial kinetic codes function as independent processes and communicate using the Parallel Virtual Machine software. This approach helps maximize flexibility while minimizing modifications to the respective codes. Using this interface, the U.S. Nuclear Regulatory Commission (NRC) three-dimensional neutron kinetic code, Purdue Advanced Reactor Core Simulator (PARCS), has been coupled to the NRC system analysis codes RELAP5 and Modernized Transient Reactor Analysis Code (TRAC-M). Consistent comparison of code results for the Organization for Economic Cooperation and Development/Nuclear Energy Agency main steam line break benchmark problem using RELAP5/PARCS and TRAC-M/PARCS was made to assess code performance.

Making use of new ideas in examining data

In-core fuel management optimization (ICFMO) is one of the most challenging concepts of nuclear engineering. In recent decades several meta-heuristic algorithms or computational intelligence methods have been expanded to optimize reactor core loading pattern. This paper presents a new method of using Gravitational Search Algorithm (GSA) for in-core fuel management optimization. The GSA is constructed based on the law of gravity and the notion of mass interactions. It uses the theory of Newtonian physics and searcher agents are the collection of masses. In this work, at the first step, GSA method is compared with other meta-heuristic algorithms on Shekel’s Foxholes problem. In the second step for finding the best core, the GSA algorithm has been performed for three PWR test cases including WWER-1000 and WWER-440 reactors. In these cases, Multi objective optimizations with the following goals are considered, increment of multiplication factor (Keff), decrement of power peaking factor (PPF) and power density flattening. It is notable that for neutronic calculation, PARCS (Purdue Advanced Reactor Core Simulator) code is used. The results demonstrate that GSA algorithm have promising performance and could be proposed for other optimization problems of nuclear engineering field.