PWR Core Research Articles

Exergy Analysis of a PWR Core Heat Transfer

Exergy Analysis of a PWR Core Heat Transfer

Exergy Analysis of a PWR Core Heat Transfer

Exergy Analysis of a PWR Core Heat Transfer

Benchmarking of Monte Carlo model of PWR (CNPP-II) core against theoretical and experimental results

The purpose of this study is to develop a complete Monte Carlo model of Chashma Nuclear Power Plant Unit-2 (CNPP-II) core and benchmark against the theoretical (Final Safety Analysis Report FSAR) and experimental results. For this purpose, a comprehensive model of CNPP-II core is developed using Monte Carlo N-Particle (MCNP5) radiation transport code which is equipped with temperature dependent ENDF/B-VII.1 cross-section library. The model is validated against the reference results of CNPP-II at seven different operating conditions of the initial core. For Cold Zero Power (CZP) condition, the overall temperature is taken as 40 °C, while for Hot Zero Power (HZP) condition, the temperature is taken as 280 °C. In Hot Full Power (HFP), the fuel, cladding and coolant temperatures are taken as 873 °C, 340 °C and 302 °C respectively. To enhance the model validation from global level (keff) to local level (flux or power distribution), the simulated pin & assembly wise neutron flux and power distribution is compared with the available experimental results of CNPP-II. The comparison demonstrates a reasonable agreement between computational and reference results. This study infers that the criticality value and excess reactivity are within design limits of CNPP-II and the reactor has a symmetrical flat power distribution, ensuring the safe reactor operation. The deviations between the calculated and experimental results have been explained in this paper. This model is interesting in the sense that it presents a new full core benchmark with experimental validation. This research also reveals that CNPP-II, which is in SMR category, is a safe and reliable reactor.

Zinc corrosion after loss-of-coolant accidents in pressurized water reactors – Thermo- and fluid-dynamic effects

Within the framework of the German reactor safety research, generic experimental investigations were carried out aiming at thermal-hydraulic consequences of physicochemical mechanisms, caused by dissolution of zinc in boric acid during corrosion processes at hot-dip galvanized surfaces of containment internals at lower coolant temperatures and the subsequent precipitation of solid zinc borates in PWR core regions of higher temperature. This constellation can occur during sump recirculation operation of ECCS after LOCA. Hot-dip galvanized compounds, which are installed inside a PWR containment, may act as zinc sources. Getting in contact with boric acid coolant, zinc at their surfaces is released into coolant in form of ions due to corrosion processes. As a long-term behavior resp. over a time period of several days, metal layers of zinc and zinc alloys can dissolve extensively.First fundamental studies at laboratory scale were done at the Helmholtz-Zentrum Dresden-Rossendorf (HZDR). Their experimental results were picked up for the definition of boundary conditions for experiments at semi-technical scale at the Hochschule Zittau/Görlitz (HSZG). Electrical heating rods with zircaloy cladding tubes have been used as fuel rod simulators.As near-plant core components, a 3×3 configuration of heating rods (HRC) and a shortened, partially heatable PWR fuel assembly dummy were applied into cooling circuits. The HRC module includes segments of spacers for a suitable representation of a heating channel geometry. Formations of different solid zinc compounds (mainly borates) were observed at the heatable zircaloy surfaces and characterized in detail during the heating-up to several coolant temperatures. As a strict consequence of their proven influence on heat removal and coolant flow behavior in the PWR core, preventive water-chemical methods were defined and tested.

Validation of the Serpent-ARES code sequence using the MIT BEAVRS benchmark – HFP conditions and fuel cycle 1 simulations

This paper continues a series of studies in which the Serpent 2 Monte Carlo code is used for producing homogenized group constants for the ARES core simulator. The test case is the MIT BEAVRS benchmark, which involves the detailed description of a 1000MW Westinghouse PWR core and the operating data for the first two cycles. Previous initial core hot zero-power calculations are extended to full power conditions and fuel cycle simulation. The results of the Serpent-ARES code sequence are compared to a reference Serpent 3D calculation and experimental data provided with the benchmark specification. It is concluded that the results are in good agreement. This study also demonstrates that the Monte Carlo method can be a viable, albeit computationally expensive option for group constant generation, even if the procedure involves accounting for fuel burnup and covering the full range of reactor operating conditions.

Improving PWR core simulations by Monte Carlo uncertainty analysis and Bayesian inference

The Monte Carlo-based Bayesian inference model MOCABA is applied to the prediction of reactor operation parameters of a PWR nuclear power plant. In this non-perturbative framework, high-dimensional covariance information describing the uncertainty of microscopic nuclear data is combined with measured reactor operation data in order to provide statistically sound, well founded uncertainty estimates of integral parameters, such as the boron letdown curve and the burnup-dependent reactor power distribution. The performance of this methodology is assessed in a blind test approach, where we use measurements of a given reactor cycle to improve the prediction of the subsequent cycle. As it turns out, the resulting improvement of the prediction quality is impressive. In particular, the prediction uncertainty of the boron letdown curve, which is of utmost importance for the planning of the reactor cycle length, can be reduced by one order of magnitude by including the boron concentration measurement information of the previous cycle in the analysis. Additionally, we present first results of non-perturbative nuclear-data updating and show that predictions obtained with the updated libraries are consistent with those induced by Bayesian inference applied directly to the integral observables.

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 impact of heavy reflectors on power distribution perturbations in large PWR reactor cores

The paper is intended to provide a simple – but general – overview of the question of neutronics coupling in large PWR cores. As an example, we have investigated the sensitivity of the radial power distribution in the core to the nature of the reflector (conventional vs. a “heavy” stainless-steel one). This was done by applying a perturbation to some peripheral assemblies of the UAM GEN-III benchmark with a fresh fuel core. The perturbation was selected to simulate possible assembly bowing effects, causing a variation of the inter-assembly water gap, which can generate an azimuthal asymmetry in the power distribution. This is conventionally referred to as a ‘power tilt’. The amplification effect of the reflector on the power tilt for fresh fuel has been addressed in two steps. Firstly, we have compared the power tilt, computed with the deterministic code CRONOS2 (the neutronics diffusion module of the HEMERA chain) to that found with the probabilistic transport code MCNP, so as to validate the CRONOS2 response against a Monte Carlo reference and also obtain a quantification of the amplification effect at zero power. Then the thermal feedback and burn-up effects have been accounted for in the diffusion calculations so as to evaluate the sensitivity of the power tilt to the reactor operating conditions. The difference in the behavior of the cores with a conventional and a heavy reflector has been investigated through the examination of the power distributions and the dominance ratio of the systems, as well as of the assembly-wise infinite multiplication factors. Whilst the results help to explain the higher sensitivity to a peripheral perturbation of a fresh fuel heavy-reflected core compared to a conventionally-reflected one, they do not enable fully comprehensive conclusions to be drawn, especially towards End Of Cycle, due to the model assumptions and the limitation in the eigenvalue expansion.

A numerical study of isotropic and anisotropic ablation in MCCI by MPS method

Anisotropic and isotropic ablation in molten corium-concrete interaction (MCCI) phenomenon was studied with the Moving Particle Semi-implicit (MPS) method by carrying out numerical simulations of CCI-2 and 3 experiments. The interaction of the fully oxidized PWR core melts with specially-designed two-dimensional limestone and siliceous concrete test sections was analyzed, focusing on investigating the two-dimensional ablation behavior with both limestone and siliceous concrete. The phase transition of molten corium and concrete was modeled based on a phase transition model for mixture. Slag film model and crust dissolution models were incorporated in the current MPS code to simulate the effect of gas generation and crust dissolution phenomena in limestone concrete. The effects of gas generation and aggregates on the concrete ablation behavior were investigated by simulating different specially designed cases. The simulation results by MPS method reproduced the isotropic and anisotropic cavity ablation profile and the overall axial and lateral ablation rates agreed well with the experimental measures. The experimental and MPS results both indicate that the crust on the corium-concrete interface can play an important part in concrete ablation process. The simulation results by MPS method also provide evidence to support the theory that aggregates are part of the cause of anisotropic ablation profile in cavity with siliceous concrete because aggregates could delay the axial basemat ablation more significantly than the lateral one and influence the power split in the melt pool.

Implementation of the batch composition preserving genetic algorithm for burn up extension of a typical PWR

Nuclear reactor core is the heart of a power plant producing power from fissile fuel fission. Refueling is needed periodically when it becomes impossible to maintain the reactor operating at nominal power as a result of fuel burn up. In PWR core reloading, attention is drawn to the configuration that meets safety requirements and minimizes energy cost. This paper focuses on finding the best core configuration for a typical two-loop, 300 MWe PWR satisfying the objectives of power peaking factor minimization to enhance safety of the reactor and maximization of multiplication factor to increase fuel burn up. Multi-objective optimization of the first core has been accomplished by implementing the batch composition preserving genetic algorithms (GA). Neutronic calculations and burn up analysis of the optimized loading patterns have been carried out using available reactor physics codes. It is found from this study that burn up of the optimized core has been extended by 48 effective full power days (EFPD's) while satisfying safety criterion by keeping power peaking factor below the reference value.

Burnable absorber-integrated guide thimble (BigT) – II: application to 3D PWR core design

ABSTRACTThis paper is a companion to its immediate predecessor in the “Burnable absorber-integrated Guide Thimble” (BigT) series. It aims to demonstrate potential applications of the BigT concepts in a three-dimensional (3D) commercial pressurized water reactor core, which is based on the AP1000 first core design. The study specifically compares neutronic characteristics of the reference core against a BigT-loaded design. In this study, reactivity depletion patterns of nine fuel assembly lattices in the reference core were first evaluated. Corresponding sets of BigT-loaded assemblies that yield neutronically similar characteristics (i.e., initial reactivity suppression and depletion trend) with those of the reference assembly lattices were determined next. These BigT-loaded fuel assemblies were later loaded in place of the reference fuel assemblies for the subsequent high-fidelity 3D Monte Carlo core simulations. Results of the study clearly demonstrate that the BigT-loaded AP1000 first core performs as well as the reference core since all neutronic parameters are comparable, especially in terms of reactivity depletion, power peaking factors and shutdown margin. All simulations were completed using the Monte Carlo Serpent code with ENDF/B-VII.0 library.

PWR few-group constants parameterization analysis

To provide few-group constants for the assembly-homogenized PWR core neutronics simulation based on the two-step approach, few-group constants parameterization is required by using the discrete relationship between those few-group constants and the assembly state parameters. As numerical methods for both lattice and core calculations have been more and more advanced, few-group parameterization tends to be the main error source for reactor physics calculations. In addition, there are also other effects such as history effect expected to be treated during parameterization process. Currently there are several parameterization methods but there is no clear conclusions drawn on how to select when dealing with new problems or developing new codes. Thus, this study characterized the entire few-group constants parameterization into three main aspects, namely the combination of state parameters, functionalization and attached effects, and developed a common link code named NECP-Lilac to answer those questions. Firstly, by optimizing state parameter combination, error of few-group constants is reduced by half. Secondly, the two most widely employed functionalization methods including linear interpolation and least-square fitting are compared based on typical PWR problems. Linear interpolation is recommended for new reactors since it needs almost no experience on discrete point generation process as long as there are sufficient numbers of them. Least-square fitting is suggested for routine calculations considering the fact that it needs only half discrete points for each state parameter to provide the same precision with linear interpolation. Thirdly to treat the history effect, two methods including macro-correction and micro-depletion are implemented and compared. It has been found that the micro-depletion method can always reduce the corresponding error while the performance of the macro-correction method appears case dependent.

Hydraulic lift-off issues for application of high performance annular fuels in pressurized water reactors

In the PWR core, the fuel assembly is firmly seated on the lower core plate during operation. However, if the hydraulic force exerted on the fuel assembly by coolant flow is too large and the fuel assembly is lifted-off from the lower core plate, the excessive vibration will cause fuel failure. Therefore, the hydraulic lift-off issue needs to be addressed when the advanced fuel assembly is developed. It has been shown that the advanced annular fuel design with internal cooling allows power uprating up to 50% while the peak temperature of the fuel can be reduced and the MDNBR can be maintained. However, if the coolant condition in the core is kept unchanged, increasing the core power by 50% requires the core flow rate also increase proportionally, which will give rise to the hydraulic lift-off, an important issue to be addressed. In this paper, taking the 17×17 solid fuel design as the reference, the hydraulic lift-off issue is investigated for proposed 12×12 and 13×13 annular fuel designs. Both the steady-state and start-up operating conditions are evaluated. It is found that the hydraulic lift-off indeed is an issue for annular fuel design which requires careful analysis. By comparison, the lift-off forces and hold-down forces required for the externally and internally cooled annular fuels (13×13 and 12×12 arrays) are several times larger than that of the referenced solid fuel (17×17 array). Therefore, the hold-down mechanism for annular fuel needs to be carefully designed.

Drift-flux model for rod bundle geometry

In view of an important role of a one-dimensional drift-flux correlation in nuclear thermal-hydraulic system analysis codes, several drift-flux correlations such as Lellouche–Zolotar, Chexal–Lellouche, TRAC-BF1 and Ozaki correlations have been reevaluated by rod bundle test data taken in FRIGG and NUPEC test facilities. The mean absolute error of void fraction representing a correlation bias of the Lellouche–Zolotar, Chexal–Lellouche, TRAC-BF1 and Ozaki correlations are, respectively, −1.0, 0.5, −6.3 and −3.3% for the FRIGG test data and 2.0, 2.3, −0.4 and −0.7% for the NUPEC test data. The effects of unheated rods, axial and radial power distributions, large unheated center rod and geometry of a shroud or casing on void fraction are identified. The presence of unheated rods with similar size of other heated rods tends to increase a distribution parameter in a drift-flux correlation, whereas the presence of a large unheated center rod tends to decrease the distribution parameter. The axial and radial power distributions do not have significant effect on void fraction within the tested axial and radial power distribution range. The Ozaki correlation is recommended for predicting void fraction in a BWR core but it is suggested to reduce the distribution parameter in the Ozaki correlation if a large unheated center rod exists in the core. It is indicated that drift-flux correlations developed based on bounded rod bundle test facility data may overestimate the distribution parameter for a PWR core.

High-fidelity coupled Monte Carlo neutron transport and thermal-hydraulic simulations using Serpent 2/SUBCHANFLOW

Efforts to develop high-fidelity, in silico or ab initio, high performance multi-physics tools are undertaken by many groups due to the availability of relatively cheap, large-scale parallel computers. To this end, an internal coupling between the Monte Carlo reactor physics code Serpent 2 and the sub-channel code SUBCHANFLOW has been developed. The coupled code system is intended to serve as reference for deterministic reactor dynamics code developments in the future. It exploits the fact that Serpent was conceived as a lattice code for such deterministic tools. The coupling utilizes Serpent’s recently introduced universal multi-physics interface. With the multi-physics interface enabled, Serpent treats temperature dependence of nuclear data using the target motion sampling method. Since the target motion sampling methodology cannot be applied to thermal bound-atom scattering or unresolved resonances, a stochastic mixing fall back algorithm to enable the simulation of thermal reactors has been implemented. The developed coupled code is verified by code-to-code comparison with an external coupling of the Monte Carlo tool TRIPOLI4 and SUBCHANFLOW as well as the internally coupled code MCNP5/SUBCHANFLOW. Simulation results of all code systems were found to be in good agreement. Thereafter, the second exercise of the OECD/NEA and U.S. NRC PWR MOX/UO2 core transient benchmark is studied to demonstrate that Serpent 2/SUBCHANFLOW may be employed to analyze realistic, industry-like cases such as a full PWR core under hot full power conditions in a reasonable amount of time. The obtained simulation results are compared to known benchmark solutions and the numerical performance of Serpent 2/SUBCHANFLOW is analyzed to assess the feasibility of routine application. While Serpent 2/SUBCHANFLOW’s performance in terms of physics and numerical efficiency is found to be generally satisfactory, options to further improve the coupled tool concerning both aspects are discussed. Afterwards, first efforts to validate Serpent 2/SUBCHANFLOW using the hot zero power state of the cycle 1 of the BEAVRS benchmark are presented.

3-D numerical investigation of water/CuO nanofluid critical heat flux phenomenon in a PWR core channel during LOCA

Forced convection boiling and critical heat flux have been under considerable attention in variety of areas due to high heat removal capacity. However, once the heat flux exceeds a certain high level (CHF), the heated surface can no longer support continuous liquid contact, associated with substantial reduction in the heat transfer efficiency. One way to increase the level of the CHF is to add certain nanoparticles to the base fluid. The present paper investigates the effects of the addition of copper oxide nanoparticles on CHF phenomenon within the general-purpose computational fluid dynamics (CFD). The governing equations solved are generalized phase continuity, momentum and energy equations. Wall boiling phenomena are modeled using the baseline mechanistic nucleate boiling model developed in Rensselaer Polytechnic Institute (RPI). To simulate the critical heat flux phenomenon, the RPI model is extended to the departure from nucleate boiling (DNB) by partitioning wall heat flux to both liquid and vapor phases considering the existence of thin liquid wall film. It was shown that the presence of copper oxide nanoparticles in the base fluid, delays the dryout phenomenon dramatically and in specific concentration, CHF threshold would be enhanced, therefore, raising the upper limit of CHF could allow for higher safety margins.

Investigation of the ex-core noise induced by fuel assembly vibrations in the Ringhals-3 PWR

The effect of cycle burnup on the ex-core detector noise at the frequency of the pendular core barrel vibrations in the Ringhals-3 PWR core was investigated using a neutron noise simulator. The purpose of the investigations was to confirm or disprove a hypothesis raised by Sweeney et al. (1985) that fuel assembly vibrations could affect the ex-core detector noise and cause the corresponding peak in the auto power spectral density (APSD) to increase during the cycle due to the effects of fuel burnup, the change of boron concentration, flux redistribution etc. Numerical calculations were performed by modelling the vibrations of fuel assemblies at different locations in the core and calculating the induced neutron noise at three burnup steps. The APSD of the ex-core detector noise was evaluated with the assumption of vibrations either along a straight-line or along a random two-dimensional trajectory, with two different representations of the cross section perturbations caused by the vibrations. The results show the obvious effect of in-core fuel vibrations on the ex-core detector noise, but the monotonic increase of the APSD does not occur for all fuel elements, vibration types and cross section perturbation models. Such an increase of the of APSD occurs predominantly for peripheral assemblies with one of the perturbation models. However, assuming simultaneous vibrations of a number of fuel assemblies uniformly distributed over the core with random vibrations and the more realistic perturbation model, the effect of the peripheral assemblies will dominate and hence the surmised monotonic increase of the amplitude of the ex-core neutron noise during the cycle can be confirmed.

Simplified model to predict the maximum temperature locations in PWR core

Multi-group diffusion depletion 3D code TRITON was used to calculate the power density profile for a 2775 MW PWR core. TRITON was run for different control rod axial positions at full power: (a) all the control rods are outside the core and (b) control rod group D position at 220 cm with length 213 cm in core. Depending on the axial power density shape, a suitable function was chosen to fit the power density. This function was used to evaluate the temperature distribution in the core. A MATLAB script was used to predict the temperature distribution. The maximum temperatures for the coolant, clad and fuel centre were determined for each control rod position. The maximum temperature locations, when all the control rods were outside the core, were found to exist in assembly (F6), and the vicinity of the centre of the core in assemblies that are burnable absorber free. When control rod group D was inserted in the core in steps, a significant number of high temperatures were found on the same axial position or became closer to the centre of the core.

Parameterized representation of macroscopic cross section for PWR reactor

The purpose of this work is to describe, by means of Chebyshev polynomials, a parameterized representation of the homogenized macroscopic cross section for PWR fuel element as a function of soluble boron concentration, moderator temperature, fuel temperature, moderator density and U92235 enrichment. The cross-section data analyzed are fission, scattering, total, transport, absorption and capture. The parameterization enables a quick and easy determination of problem-dependent cross-sections to be used in few group calculations. The methodology presented in this paper will allow generation of group cross-section data from stored polynomials to perform PWR core calculations without the need to generate them based on computer code calculations using standard steps. The results obtained by the proposed methodology when compared with results from the SCALE code calculations show very good agreement.

Minor actinide transmutation on PWR burnable poison rods

Minor actinides are the primary contributors to long term radiotoxicity in spent fuel. The majority of commercial reactors in operation in the world are PWRs, so to study the minor actinide transmutation characteristics in the PWRs and ultimately realize the successful minor actinide transmutation in PWRs are crucial problem in the area of the nuclear waste disposal. The key issues associated with the minor actinide transmutation are the appropriate loading patterns when introducing minor actinides to the PWR core.We study two different minor actinide transmutation materials loading patterns on the PWR burnable poison rods, one is to coat a thin layer of minor actinide in the water gap between the zircaloy cladding and the stainless steel which is filled with water, another one is that minor actinides substitute for burnable poison directly within burnable poison rods. Simulation calculation indicates that the two loading patterns can load approximately equivalent to 5–6 PWR annual minor actinide yields without disturbing the PWR keff markedly. The PWR keff can return criticality again by slightly reducing the boric acid concentration in the coolant of PWR or removing some burnable poison rods without coating the minor actinide transmutation materials from PWR core. In other words, loading minor actinide transmutation material to PWR does not consume extra neutron, minor actinide just consumes the neutrons which absorbed by the removed control poisons.Both minor actinide loading patterns are technically feasible; most importantly do not need to modify the configuration of the PWR core and composition of nuclear fuel. Furthermore, the two minor actinide loading patterns completely isolate minor actinide transmutation materials from nuclear fuel, so it also facilitates fuel reprocessing after discharged from PWR core. We believe that loading minor actinide to PWR burnable poison rods for transmutation is an optimal minor actinide loading pattern.