Advanced Reactor Core 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 Effect of Replacing Ni with Mn on the Microstructure and Properties of Al2O3-Forming Austenitic Stainless Steels: A Review.

Al2O3-forming austenitic steel (AFA steel) is an important candidate material for advanced reactor core components due to its excellent corrosion resistance and high temperature strength. Al is a strong ferrite-forming element. Therefore, it is necessary to increase the Ni content to stabilize austenite. Ni is expensive and highly active, and so increasing the Ni content not only increases the costs but also damages the radiation resistance. Mn is a low-cost austenitic stable element. Its substitution for Ni will not only help to improve the irradiation resistance of austenitic steel, but also reduce the cost. In order to explore the feasibility of Mn-substituted Ni-stabilized austenite in AFA steel, this paper summarized the research progress of Mn-added AFA steels, whilst the research status of traditional Mn-added austenitic steels are also referred to and compared herein. The effect of the addition of Mn on the microstructure and properties of AFA steel was analyzed. The results show that Mn can promote the precipitation of the M23C6 phase and inhibit the precipitation of the B2-NiAl phase and secondary NbC phase. With the increase in Mn content, the strength of AFA steel at room temperature and high temperature decreased slightly, the room temperature elongation increased slightly, while the high temperature elongation and creep resistance decreased obviously. In addition, for austenitic steel free of Al, the addition of Mn will destroy the oxide layer of Cr2O3, which will decrease the oxidation resistance of the steel. But the preliminary study shows that Mn has little effect on the Al2O3 oxide layer. It is worth studying the effect of Mn-substituted Ni on the oxidation resistance of AFA steel. In summary, more efforts are necessary to investigate the optimal Mn content to balance the advantages and disadvantages of introducing Mn instead of Ni.

Comparison study on thermo-hydraulic performance of circular and airfoil shaped wire-wrapped rod bundle

The turbulent flow behavior and heat transfer performance of wire-wrapped rod bundles have a significant impact on the thermal–hydraulic characteristics and safety of Generation IV nuclear reactors. In this study, the airfoil-shaped wire-wrapped bundle is proposed to improve the comprehensive performance of advanced reactor cores, and the thermal–hydraulic performance has been compared with the circular-shaped wire-wrapped bundle. Special attention has been paid to the temperature variation and friction pressure loss caused by the wire wraps. Specifically, the generation of the secondary flow is the main factor. The airfoil-shaped wire-wrapped bundle exhibits a half friction pressure loss, a more uniformed temperature distribution, and a reduction of the hot spot temperature compared to the circular-shaped wire-wrapped bundle. Moreover, the airfoil-shaped wire shows a 16.3% increase in Nusselt number and a 2.0% increase in friction factor compared to circular-shaped wire, resulting in an 8.7% increase in performance evaluation criteria (PEC). The airfoil-shaped wire-wrapped bundle demonstrates better comprehensive performance than the circular-shaped wire-wrapped bundle in terms of PEC.

Seismic response prediction using intensity measures: Graphite nuclear reactor core model case study

Seismic response analyses of structures have conventionally used the peak ground acceleration or spectral acceleration as an intensity measure to estimate the engineering demand parameters. An extensive shaking table test program was carried out on a quarter-sized advanced gas-cooled reactor (AGR) core model to investigate the global dynamic behavior of the system with degraded graphite components while subjected to seismic excitation. Evaluation of the most widely considered intensity measures, with respect to their capability for predicting the seismic response of an AGR core–like structure, is performed. Twenty intensity measures of 16 distinct seismic input motions are formulated and correlated, with experimental measurements describing the dynamic response of the reactor core model. Linear correlations are constructed for each intensity measure to statistically determine the best metric for predicting the seismic response of the AGR core model, and statistical analysis indicates that the acceleration spectrum intensity (ASI) is best suited to characterize and describe the structural demand of an AGR core-like structure when subjected to seismic loading. A response prediction tool is developed, based on empirically derived linear correlations, to estimate column distortions and determine the critical input motion for further experimental and numerical studies. Statistical analysis indicates that predicted column distortions, compared against direct experimental displacements, are significant, repeatable, and accurate.

The development of the combined fission matrix method in small advanced reactor design and optimization

The Monte Carlo neutron transport method is widely used in small advanced reactor core designs and optimizations but with a high computational cost. As an efficient and accurate neutron transport method, the combined fission matrix method is promising in small reactor design and optimization. However, it cannot well predict the strong reflector impact on the fission matrix in small reactors. A predictor–corrector correction ratio method is proposed in this work to solve the problem. The method will be validated in the Holos and Simons transportable small reactors. Compared to the Monte Carlo result, the combined fission matrix method has a 36 pcm keff error with 1.7% 2D RMSE in the Holos core and 9 pcm keff error with 0.3% 2D RMSE in the Simons core. With a similar accuracy, the combined fission matrix method is much faster than the Monte Carlo method.

A mathematical framework for evaluation of column shape profiles for an advanced gas-cooled reactor core model subject to seismic excitation

The safe shutdown of Advanced Gas-cooled Reactor (AGR) nuclear power stations in response to a seismic event is vital to their safety case. The tubular graphite bricks used to moderate the neutrons within an AGR core are arranged in columns whose bores provide channels for either fuel or control rods. Earthquake-induced distortion of the channels could impede the insertion of the control rods and compromise the safe shut-down, maintenance and servicing of the reactor. This paper presents a mathematical framework, utilising Euler mechanics, to evaluate the column shape displacement profiles of fuel and control rod channels within a state-of-the-art quarter-sized physical model of an AGR core when subjected to seismic loading. The data obtained from sensors installed within the model bricks, and configured to monitor interface displacements, are used to infer the global behaviour of a multi-stacked brick column subject to seismic excitation. Directly measured displacements of the top of the brick columns, obtained using a motion capture vision system, are compared with the displacements calculated using the framework presented, verifying the validity of the procedure. Statistical analysis is employed to quantify and characterise the performance of the Euler mechanics method. For multiple model build configurations, which represent different brick-cracking scenarios in an aged core, the Pearson correlation factor between the direct and indirect measurements is evaluated for the top of the column displacements giving an average value of 0.96 in the direction of the input motion. This shows that good agreement is achieved for the column shape displacement time-histories. The seismic responses are shown to be significantly larger in amplitude in the presence of large numbers of cracked bricks.

Inherent safety enhancement by design optimization of a floating absorber for safety at transient (FAST) in an advanced burner reactor

A floating absorber for safety at transient (FAST) is a passive safety system to enhance the inherent safety of sodium-cooled fast reactors (SFRs). This study optimized and improved the original FAST design in terms of the initial position, barycenter, eccentricity, reactivity worth, and diameter ratio of FAST to pin (k) in advanced burner reactor (ABR) cores. In contrast to previous studies based on average assembly analysis, a comprehensive design optimization study was conducted through transient simulations based on multiassembly analysis. Applying a damped FAST, which is an improvement over the original FAST design, enhanced the negative reactivity feedback in the oxide cores depending on the diameter ratio. Oscillatory behaviors were observed when the k of the damped FAST system was within 0.1 ∼ 0.55 for $1 of the FAST reactivity under an unprotected loss of flow event (ULOF). A damped FAST with a k of 0.7 maintained the lowest coolant temperature while suppressing the oscillatory behavior for $1 of the FAST reactivity. The results showed that damped FAST systems support negative reactivity feedback for the ABR oxide core and should be properly designed to avoid power and temperature oscillations depending on the FAST reactivity worth.

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.

Corrosion and high temperature compressive strength behaviour of 17Cr ferritic ODS steel with addition of aluminium through vacuum hot pressing

ABSTRACT The structural materials are essential for the present important advancements in economics, for reliability, safety, sustainability, etc., over presently operating reactor technologies of Generation IV reactors, advanced fast reactor core and fusion reactors. In order to avoid the release of radioactive fission products in the environment, the consistency of the cladding tube is necessary. Thus, increasing the operating temperature of these steels by oxide dispersion strengthening (ODS) makes them promising candidate materials. For the present study, the samples were prepared from the 430 L pre-alloyed powder using mechanical alloying through vacuum hot pressing. The hot-pressed alloys A, B, and C structural changes were examined through transmission electron microscope-energy dispersive spectroscopy (TEM-EDS) analysis and the mechanical properties, such as sintered density(g/cc), Vickers hardness (HV), compressive strength (650°C), and elongation (%) were measured. The corrosion resistance was measured through an electrochemical corrosion test of the alloys A, B, and C. From the TEM-EDS analysis, it was observed that the nano oxide particles and complex oxides particles were uniformly distributed. Alloy A has a higher compressive strength of 1307 MPa when compared to alloy Band alloy C. Moreover, alloy Chas higher corrosion resistance was observed than alloys A& B.

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.

Simulations of advanced reactor cores in research light water reactor LR-0

A design verification by basic neutronic experiments is an integral part of maturing a reactor concept. Before building a demonstrator unit, it is useful to design and carry out verification and validation work concerning the modeling tools as well as nuclear data in a mock-up simulating the environment in an advanced reactor. The focus of this paper is to show how the experimental research light water reactor LR-0 is used in development work connected to a wide range of advanced reactor concepts: High Temperature Gas Cooled Reactor (HTGR), Molten Salt Reactor (MSR), Molten Salt Fast Reactor (MSFR) Lead Cooled Fast Reactor (LFR), and Gas Cooled Fast Reactor (GFR). Results show that a large channel in a light water research reactor with appropriate neutron filter can be used for studies of basic reactor physics in the fast part of the neutron spectrum of some more advanced systems.

Neutronic design and analysis of advanced long-cycle boron-free operation of a small modular reactor core with particle type burnable poison rods

Soluble boron in the reactor coolant plays an important role for operation of pressurized water reactors (PWRs), but it leads to several issues such as generation of a huge amount of liquid waste, complications in the primary coolant system, and corrosion of system components. This work introduces an advanced small modular reactor (SMR) core concept that can operate over a long cycle without soluble boron and with inherent safety features. The large portion of excess reactivity for compensating for fuel depletion is suppressed by using advanced particle-type burnable poison (BP) rods, and the remaining ones by the control rods with their insertion strategies. The suggested SMR core is designed to operate over a long cycle of 3.8 effective full power year (EFPY) and is coupled with a suggested control rod map and the associated insertion strategy. The superb safety characteristics of this core were confirmed using a simulation of a rod ejection accident and by showing that the core has sufficient shutdown margin and can maintain its subcriticality even at cold zero power (CZP) conditions.

Advanced Burner Reactors with Breed-and-Burn Thorium Blankets for Improved Economics and Resource Utilization

This paper assesses the feasibility of designing seed-and-blanket (S&B) sodium-cooled fast reactor (SFR) cores to generate a significant fraction of the core power from radial thorium-fueled blankets that operate in the breed-and-burn (B&B) mode. The radiation damage on the cladding material in both seed and blanket does not exceed the presently acceptable constraint of 200 displacements per atom (dpa). The S&B core is designed to have an elongated seed (or driver) to maximize the fraction of neutrons that radially leak into the subcritical B&B blanket and reduce the neutron loss via axial leakage. A specific objective of this study is to maximize the fraction of core power generated by the B&B blanket that is proportional to the neutron leakage rate from the seed to the blanket. Since the blanket feed fuel is very inexpensive and requires no reprocessing and remote fuel fabrication, a larger fraction of power from the blanket will result in a lower fuel cycle cost per unit of electricity generated by the SFR core. It is found possible to design the seed of the S&B core to have a lower transuranics (TRU) conversion ratio (CR) than a conventional advanced burner reactor (ABR) core without deteriorating core safety. This is due to the unique synergism between a low CR seed and the B&B thorium blanket. The benefits of the synergism are maximized when using an annular seed surrounded by inner and outer thorium blankets. Two high-performance S&B cores are designed to benefit from the annular seed concept: (1) an ultra-long-cycle core having a CR = 0.5 seed and a cycle length of ~7 effective full-power years (EFPYs) and (2) a high-transmutation core having a TRU CR of 0.0. The TRU transmutation rate of the latter core is comparable to that of the reference ABR with a CR of 0.5, and the thorium blanket can generate close to 60% of the core power. Because of the high blanket power fraction along with the high discharge burnup of the CR = 0 seed, the reprocessing capacity per unit of core power required by this S&B core is only approximately 1/6th that of the reference ABR core with a TRU CR of 0.5. Although the seed fuel CR is nearly zero, the burnup reactivity swing is low enough to enable a cycle length of more than 4 EFPYs. This is attributed to a combination of reactivity gain in the thorium blankets over the cycle and the relatively high heavy metal inventory. Moreover, despite the very low leakage, the S&B cores feature a less positive coolant reactivity coefficient and large enough negative Doppler coefficient even when using nonfertile fuel for the seed, because of the unique physics properties of the 233U and Th in the thorium blankets. With the long cycles, the S&B SFR is expected to have a higher capacity factor, and therefore a lower cost of electricity, than conventional ABRs. The discharge burnup of the thorium blanket fuel is typically 70 MWd/kg such that the thorium fuel utilization is approximately 12 times that of natural uranium in light water reactors. A sensitivity study is subsequently undertaken to quantify the trade-off between the core performances and several design variables: amount of zirconium in the inert matrix seed fuel, active core height, coolant pressure drop, and radiation damage constraint. The effect of the criterion used for quantifying acceptable radiation damage is evaluated as well. It is concluded that a viable S&B core can be designed without significant deviation from typical SFR core design practices.

Thermal-hydraulic comparisons of 19-pin rod bundles with four circular and trapezoid shaped wire wraps

Coolant in wire-wrapped fuel assembly flows not only in axial direction but also in a transverse direction. This special flow is critical for momentum exchange and heat transfer in fuel rod bundle. To obtain a better performance wire-wrapped fuel assembly for advanced reactor cores, in this paper, the characteristics of turbulent flow and heat transfer inside 19-pin heated rod bundles where each pin is wrapped with four circular shaped and four trapezoid shaped wires are numerically studied, respectively. Considering the geometric complexity, the advanced hybrid-grid technique that combines the structured hexahedron grids and unstructured pentahedral grids is adopted to divide the modeling assembly. The validity of both RSM and SST k-ω model is assessed by comparing calculated pressure loss with the available experimental data to determine a proper turbulence closure. Special attention has been paid to understand the effect of secondary flows created by the wire wraps on the reduction of coolant temperature variation. Additionally, in order to assess the effectiveness of the different shaped wires, comparisons on both global heat transfer and associated penalty in pressure loss are evaluated. Results indicate that the frictional pressure loss of the rod bundle with circular shaped wires is smaller than that with trapezoid shaped wires, while the global heat transfer is equivalent. However, the trapezoid shaped wires give a lower maximum temperature as well as a uniform temperature distribution within the subassembly in comparison with circular shaped geometry.

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.