Showing papers in "Progress in Nuclear Energy in 2009"

Deliberately Small Reactors and the Second Nuclear Era

Abstract: Smaller sized nuclear reactors were instrumental during the pioneering days of commercial nuclear power to facilitate the development and demonstration of early reactor technologies and to establish operational experience for the fledgling nuclear power industry. As the U.S. embarks on its “second nuclear era,” the question becomes: Will smaller sized plants have a significant role in meeting the nation's needs for electricity and other energy demands? A brief review of our nuclear history is presented relative to plant size considerations, followed by a review of several commonly cited benefits of small reactors. Several “deliberately small” designs currently being developed in the U.S. are briefly described, as well as some of the technical and institutional challenges faced by these designs. Deliberately small reactors offer substantial benefits in safety, security, operational flexibilities and economics, and they are well positioned to figure prominently in the second nuclear era.

Recent advances in nuclear power: A review

Abstract: The rise in oil prices and the increased concern about environmental protection from CO2 emissions have promoted the attention to the use of nuclear power as a viable energy source for power generation. This review presents the recent advances in the field of nuclear power and addresses the aspects of nuclear economics, safety, nuclear reactor design and spent fuel processing and waste management.

Nuclear heat for hydrogen production: Coupling a very high/high temperature reactor to a hydrogen production plant

Abstract: Hydrogen has been dubbed the fuel of the future. As fossil fuel reserves become depleted and greenhouse gas emissions are reduced inline with the Kyoto protocol, alternative energy sources and vectors, such as hydrogen, must be developed. Hydrogen produced from water splitting, as opposed to from hydrocarbons, has the potential to be a carbon neutral energy solution. There are several methods to extract hydrogen from water, three leading candidates being high temperature electrolysis, the SI thermochemical cycle and the HyS hybrid thermochemical cycle. All three of these processes involve a section requiring very high temperatures. The Very High Temperature Reactor (VHTR), a gas cooled Generation IV reactor, is ideally suited for providing this high temperature heat. Nuclear hydrogen production is being investigated around the world. The four leading consortiums are the Japan Atomic Energy Agency (JAEA), PBMR/Westinghouse, GA, and AREVA NP/CEA/EDF. There are also many smaller R&D efforts focussing on the development of particular materials and components and on process flowsheeting. A nuclear hydrogen plant involves four key pieces of equipment: the VHTR, the hydrogen production plant (HPP), the intermediate heat exchanger (IHX) and the power conversion system (PCS). The choice of all four items varies dramatically between programmes. Both pebble bed and prismatic fuel block VHTRs are being developed, which can be directly or indirectly coupled to a HPP and PCS placed either in series or parallel. Either a Rankine steam cycle or a Brayton gas turbine cycle can be employed in the PCS. This report details the choices made and research being carried out around the world. Predicted process efficiencies and plant costs are currently at a preliminary stage and are very similar, regardless of the options chosen. The cost of hydrogen produced from water splitting using nuclear technologies is around $2/kg H2. This is competitive with hydrogen produced by other methods, particularly if carbon emissions are regulated and costed. The technological feasibility and testing of key components will be one of the determining factors in plant viability.

Sustainability by combining nuclear, fossil, and renewable energy sources

Abstract: The energy industries face two sustainability challenges: the need to avoid climate change and the need to replace traditional crude oil as the basis of our transport system. Radical changes in our energy system will be required to meet these challenges. These challenges may require tight coupling of different energy sources (nuclear, fossil, and renewable) to produce liquid fuels for transportation, match electricity production to electricity demand, and meet other energy needs. This implies a paradigm shift in which different energy sources are integrated together, rather than being considered separate entities that compete. Several examples of combined-energy systems are described. High-temperature nuclear heat may increase worldwide light crude oil resources by an order of magnitude while reducing greenhouse gas releases from the production of liquid fossil fuels. Nuclear–biomass liquid-fuels production systems could potentially meet world needs for liquid transport fuels. Nuclear–hydrogen peak power systems may enable renewable electricity sources to meet much of the world's electric demand by providing electricity when the wind does not blow and the sun does not shine.

Particle Swarm Optimization applied to the nuclear reload problem of a Pressurized Water Reactor

Abstract: The concept of Swarm Intelligence is based on the ability of individuals to learn with their own experience in a group as well as to take advantage of the performance of other individuals, which are social–collaborative aspects of intelligence. In 1995, Kennedy and Eberhart presented the Particle Swarm Optimization (PSO), a Computational Intelligence metaheuristic technique. Since then, some PSO models for discrete search spaces have been developed for combinatorial optimization, although none of them presented satisfactory results to optimize a combinatorial problem such as the Nuclear Reactor Reload Problem (NRRP). In this sense, we have developed the Particle Swarm Optimization with Random Keys (PSORK) to optimize combinatorial problems. PSORK has been tested for benchmarks to validate its performance and to be compared to other techniques such as Ant Systems and Genetic Algorithms, and in order to analyze parameters to be applied to the NRRP. We also describe and discuss its performance and applications to the NRRP with a survey of the research and development of techniques to optimize the reloading operation of Angra 1 nuclear power plant, located at the Southeast of Brazil.

Assembly homogenization techniques for core calculations

Abstract: We have applied the black-box paradigm to assembly homogenization and introduced current discontinuity factors (CDFs) for an arbitrary low-order operator in the presence of boundary leakage. The CDFs preserve average reaction rates and the assembly partial currents in a given reference situation as well for full assembly as for pin-by-pin homogenization. In the presence of surface leakage, the CDFs depend on the discretization of the low-order operator but can be determined from a few calculations with the low-order operator without scattering. For diffusion-like, low-order operators, the CDFs can be advantageously replaced by flux discontinuity factors (FDFs), which also preserve partial currents. However, the effect of the FDFs is not equivalent to that of the CDFs in the final core calculation. Unlike the CDFs, that are double-valued for homogenization with surface leakage, the FDFs are always single valued. The cases when the low-order operator is diffusion, SP N or quasidiffusion are discussed in detail. We also show that, for full-assembly homogenization without boundary leakage, the FDFs are identical to Smith's discontinuity coefficients (DCs) only if the reference calculation has also been done with diffusion. In the case of diffusion, preliminary test calculations for small PWR motifs show that the FDFs and Smith's DCs give close results, with a better precision for the FDFs when transport effects are predominant.

Review on condensation on the containment structures

Abstract: In the last two decades condensation on the containment structures in presence of noncondensables has received substantial attention by nuclear scientists and engineers. The reason is that many Generation III and III+ reactors rely on passive systems operating under natural circulation. As a consequence, a vast number of publications have been made in the open literature. This paper reviews the specific physical phenomena that are involved in condensation and discusses how they have been considered in the different available models. In addition, it explores the data that have been used for validation and provides some insights on the effective suitability for this purpose. Finally, the paper summarizes the current codes' capabilities to deal with wall condensation in the presence of noncondensables according to the most recent available validation exercises.

A review of corrosion product transport and radiation field buildup in boiling water reactors

Abstract: Cobalt-60 is the major radiation source in the boiling water reactor (BWR) for personnel exposure during shutdown maintenance. The Co-60 activity is produced by neutron activation of cobalt with other corrosion products deposit on fuel surfaces, and is released into the coolant and deposited on primary system piping walls in the system. The transport phenomena of corrosion products in the primary system and radiation field buildup are reviewed separately in three different areas: the behavior of corrosion products in the BWR coolant, including the chemistry of corrosion products and formation of mixed metal oxides; the transport of corrosion products on fuel cladding surfaces, and the mechanisms of deposition and release are discussed; and the transport of Co-60 and radiation field buildup on out-of-core surfaces under various chemistry conditions, including normal water chemistry, hydrogen water chemistry and with chemical additives. It is concluded that with understanding the mechanisms of transport, the radiation field buildup in most operating BWRs has been considerably reduced in recent years. The major factors are reduction of cobalt source reduction, control of Co-60 release from fuel surfaces with zinc addition and improvement in water quality to minimize the corrosion product input and the material corrosion.

Multiprocessor modeling of parallel Particle Swarm Optimization applied to nuclear engineering problems

Abstract: Particle Swarm Optimization (PSO) is a population-based metaheuristic (PBM), in which solution candidates evolve through simulation of a simplified social adaptation model. Putting together robustness, efficiency and simplicity, PSO has gained great popularity. Many successful applications of PSO are reported, in which PSO demonstrated to have advantages over other well-established PBM. However, computational costs are still a great constraint for PSO, as well as for all other PBMs, especially in optimization problems with time consuming objective functions. To overcome such difficulty, parallel computation has been used. The default advantage of parallel PSO (PPSO) is the reduction of computational time. Master-slave approaches, exploring this characteristic are the most investigated. However, much more should be expected. It is known that PSO may be improved by more elaborated neighborhood topologies. Hence, in this work, we develop several different PPSO algorithms exploring the advantages of enhanced neighborhood topologies implemented by communication strategies in multiprocessor architectures. The proposed PPSOs have been applied to two complex and time consuming nuclear engineering problems: i) reactor core design (CD) and ii) fuel reload (FR) optimization. After exhaustive experiments, it has been concluded that: i) PPSO still improves solutions after many thousands of iterations, making prohibitive the efficient use of serial (non-parallel) PSO in such kind of real-world problems and ii) PPSO with more elaborated communication strategies demonstrated to be more efficient and robust than the master-slave model. Advantages and peculiarities of each model are carefully discussed in this work.

Virtual environments simulation for dose assessment in nuclear plants

Abstract: This work reports results obtained with the use of a game engine as a tool to create and to navigate in virtual environments, to perform simulations and training of workers in risky areas, for safety purposes. The game engine used can be used for non-commercial and educational purposes, and its source code is available for users. Thus, the engine can be modified and adapted for the modeling and simulation of any environment, including the development of new functionalities. A case study is shown, which has the purpose of supporting dose assessment in nuclear plants, for optimization of operational routines in these areas. The implemented modifications are explained, and simulations' results are shown.

Estimation of research reactor core parameters using cascade feed forward artificial neural networks

Abstract: The pattern of the core reload program is very important for an optimize use of research reactors Reactor safety issues and economic efficiency should be considered during pattern studies In order to find the best core pattern for a research reactor, its reloading program should be solved as a multi-objective and constrained optimization problem If considered objective functions of the optimization problem can be estimated in very short time, the optimal fuel reloading pattern can be used effectively In this research a very fast estimation system for suggested core parameters has been developed using cascade feed-forward type of artificial neural networks (ANNs) Four main core parameters are suggested to optimize reactor core adequately And also to get larger thermal fluxes in the desired irradiation box, a new flexible method was selected A Software package has been developed to prepare and reform required data for ANNs training The gradient descent method with momentum weight/bias learning rule has been used to train ANNs To get the best conditions for considered ANNs training a vast study has been performed It includes the effects of variation of hidden neurons, hidden layers, activation functions, learning and momentum coefficients, and also the number of training data sets on the training and simulation results Some experimental convergence criteria are used to study them A comparison selection rule has been used to adjust desirable conditions Final training and simulation results show that developed ANNs can be trained and estimate suggested core parameters of research reactors very quickly It improves effectively pattern optimization process of core reload program

The effect of carbon powder addition on the properties of hematite radiation shielding concrete

Abstract: Carbon powder was added to shielding concrete made of Hematite aggregates to investigate its effects on shielding properties. The powder was added in different percentages, and the mechanical and radiation attenuation properties of the prepared concretes were determined. It was found that, the addition of carbon powder by 6% (by wt.) of the concrete could increase the strength on concrete by about 15%. The shielding effectiveness decreased for both gamma and neutrons with the increasing carbon powder percentage. But the loss in shielding effectiveness for both gamma rays and neutrons were within measurements error range for carbon powder addition of 6%. Simulation for the experimental measurements was carried out using Monte Carlo MCNP code, to understand the effect of carbon powder on the shielding effectiveness against neutrons. The results of the simulation were in good agreement with the experimental results.

Three-dimensional h-adaptivity for the multigroup neutron diffusion equations

Abstract: Adaptive mesh refinement (AMR) has been shown to allow solving partial differential equations to significantly higher accuracy at reduced numerical cost. This paper presents a state-of-the-art AMR algorithm applied to the multigroup neutron diffusion equation for reactor applications. In order to follow the physics closely, energy group-dependent meshes are employed. We present a novel algorithm for assembling the terms coupling shape functions from different meshes and show how it can be made efficient by deriving all meshes from a common coarse mesh by hierarchic refinement. Our methods are formulated using conforming finite elements of any order, for any number of energy groups. The spatial error distribution is assessed with a generalization of an error estimator originally derived for the Poisson equation. Our implementation of this algorithm is based on the widely used Open Source adaptive finite element library deal.II and is made available as part of this library's extensively documented tutorial. We illustrate our methods with results for 2-D and 3-D reactor simulations using 2 and 7 energy groups, and using conforming finite elements of polynomial degree up to 6.

The role of the nuclear power generation in the Korean national economy: An input–output analysis

Abstract: The nuclear power generation has played an important role in the economic development of Korea and electric power has become a critical factor sustaining the well-being of the Korean people. This paper attempts to apply input–output (I–O) analysis to investigate the role of the nuclear power generation in the national economy, with specific application to Korea. A static I–O framework is employed, focusing on three topics in its application: the impact of nuclear power supply investment on the production of other sectors and the inter-industry linkage effect; the nuclear power supply shortage effect; and the impact of the rise in nuclear power rate on prices of other products. This paper pays particular attention to the nuclear power generation sector by taking the sector as exogenous and then investigating its economic impacts. Moreover, potential uses of the results are illustrated from the perspective of policy instruments and some policy implications are discussed.

The effective convectivity model for simulation of melt pool heat transfer in a light water reactor pressure vessel lower head. Part I: Physical processes, modeling and model implementation

Abstract: This paper, and its companion paper [Tran C.T., Dinh, IN. The effective convectivity model for simulation of melt pool heat transfer in a light water reactor pressure vessel lower head. Part II: Mo ...

Analysis on the neutron kinetics for a molten salt reactor

Abstract: The neutron kinetics of the molten salt reactor is significantly influenced by the fuel salt flow, which leads to the analysis methods for the conventional reactors using solid fuels not being applicable for the molten salt reactors. In this study, a neutron kinetic model considering the fuel salt flow is established based on the neutron diffusion theory, which consists of two-group neutron diffusion equations for the fast and thermal neutron fluxes and six-group balance equations for delayed neutron precursors. The temperature feedback on the neutron kinetics is considered by introducing a heat transfer model in the core, in which the group constants which are dependent on the temperature are calculated by the code DRAGON. The mathematical equations are discretized and numerically calculated by developing a code, in which the fully implicit scheme is adopted for the time-dependent terms, and the power law scheme is for the convection–diffusion terms. The neutron kinetics is conducted during three transient conditions including the rods drop transient, the pump coastdown transient and the inlet temperature drop transient. The relative power changes and the distributions of the temperature, neutron fluxes and delayed neutron precursors under these three different transient conditions are obtained in the study. The results provide some valuable information for the research and design of this new generation reactor.

Applying particle swarm optimization algorithm for tuning a neuro-fuzzy inference system for sensor monitoring

Abstract: A neuro-fuzzy inference system (ANFIS) tuned by particle swarm optimization (PSO) algorithm has been developed for monitoring the relevant sensor in a nuclear power plant (NPP) using the information of other sensors. The antecedent parameters of the ANFIS that estimates the relevant sensor signal are optimized by a PSO algorithm and consequent parameters use a least-squares algorithm. The proposed methodology to monitor sensor output signals was demonstrated through the estimation of the nuclear power value in a pressurized water reactor using as input to the ANFIS six other correlated signals. The obtained results are compared to two similar ANFIS using one gradient descendent (GD) and other genetic algorithm (GA), as antecedent parameters' training algorithm.

The effective convectivity model for simulation of melt pool heat transfer in a light water reactor pressure vessel lower head. Part II: Model assessment and application

Abstract: The paper reports detailed assessments and representative application of the effective convectivity model (ECM) developed and described in the companion paper ( Tran and Dinh, submitted for publication ). The ECM capability to accurately predict energy splitting and heat flux profiles in volumetrically heated liquid pools of different geometries over a range of conditions related to accident progression is examined and benchmarked against both experimental data and CFD results. Augmented with models for phase changes in binary mixture, the resulting PECM (phase-change ECM) is validated against a non-eutectic heat transfer experiment. The PECM tool is then applied to predict thermal loads imposed on the reactor vessel wall and Control Rod Guide Tubes (CRGTs) during core debris heatup and melting in the BWR lower plenum. The reactor-scale simulations demonstrate the PECM's high computational performance, particularly needed to analyze processes during long transients of severe accidents. The analysis provides additional arguments to support an outstanding potential of using the CRGT cooling as a severe accident management measure to delay the vessel failure and increase the likelihood of in-vessel core melt retention in the BWR.

Numerical tools applied to power reactor noise analysis

Abstract: In order to be able to calculate the space- and frequency-dependent neutron noise in real inhomogeneous systems in two-group theory, a code was developed for the calculation of the Green's function (dynamic transfer function) of such systems. This paper reports on the development as well as the test and application of the numerical tools employed. The code that was developed yields the space-dependence of the fluctuations of the neutron flux induced by fluctuating properties of the medium in the two-group diffusion approximation and in a two-dimensional representation of heterogeneous systems, for both critical systems and non-critical systems with an external source. Some applications of these tools to power reactor noise analysis are then described, including the unfolding of the parameters of the noise source from the induced neutron noise, measured at a few discrete locations throughout the core. Other concrete applications concern the study of the space-dependence of the Decay Ratio in Boiling Water Reactors, the noise-based estimation of the Moderator Temperature Coefficient of reactivity in Pressurized Water Reactors, the modeling of the beam- and shell-mode core-barrel vibrations in Pressurized Water Reactors, and the investigation of the validity of the point-kinetic approximation in subcritical systems driven by an external source. In most of these applications, calculations performed using the code are compared with at-power plant measurements. Power reactor noise analysis applications of the above type, i.e. core monitoring without disturbing plant operation, is of particular interest in the framework of the extensive program of power uprates worldwide.

Fire in the hole : A review of national spent nuclear fuel disposal policy

Abstract: Today, nuclear power plants operate in 31 countries and account for approximately 15% of the world's electricity production. Moreover, a number of additional reactors are expected to come on-line over the next several years as a result of a global resurgence in nuclear power. Since the amount of spent nuclear fuel generated by these plants is expected almost to double by 2020, the issue of how to dispose of highly radioactive waste properly is an international concern of growing importance. As of now, no country has yet solved the problem of what to do with the mounting inventories of spent nuclear fuel created as a by-product of nuclear power generation. This article focuses on national approaches for the disposal of spent nuclear fuel, and discusses the need for a global approach to cope adequately with the increasing inventories of highly radioactive waste.

Performance analyses of VHTR plants with direct and indirect closed Brayton cycles and different working fluids

Abstract: This paper investigated the performance of Very High Temperature Reactor (VHTR) power plants with helium working fluid and direct and indirect Closed Brayton Cycles (CBCs), and with binary mixture working fluids of He–Xe and He–N2 (molecular weight of 15 g/mole) and indirect CBCs. Also investigated are the effects of using low- and high-pressure compressors with intercooling, versus a single compressor, using bleed cooling the reactor pressure vessel in direct CBC helium plants, and varying the reactor exit temperature from 700 °C to 950 °C on the plant thermal efficiency, cycle pressure ratio and the size of and number of stages in the turbine and compressor. Analyses are performed for a shaft rotation speed of 3000 rpm, reactor thermal power of 600 MW and a temperature pinch of 50 °C in the intermediate heat exchanger (IHX) for the indirect CBCs.

An analytical solution for the point reactor kinetics equations with one group of delayed neutrons and the adiabatic feedback model

Abstract: The point reactor kinetics equations with one group of delayed neutrons and the adiabatic feedback model are solved analytically. The analytical solution is based on an expansion of the neutrons density in powers of the small parameter, the prompt neutrons generation time, into the second order differential equation in the neutron density. The relation between the time and the reactivity for reactor excursions near prompt critical is derived. Also, the neutron density and the average density of delayed neutron precursors as functions of reactivity are presented. The relations of reactivity, neutron density and temperature with time are calculated, drawn, and compared with other analytic method.

Sources: A code for calculating (alpha, n), spontaneous fission, and delayed neutron sources and spectra

Abstract: SOURCES is a computer code that determines neutron production rates and spectra from (alpha, n) reactions, spontaneous fission, and delayed neutron emission due to the decay of radionuclides in homogeneous media, interface problems, and three-region interface problems. The code is also capable of calculating the neutron production rates due to (alpha, n) reactions induced by a monoenergetic beam of alpha particles incident on a slab of target material. The (alpha, n) spectra are calculated using an assumed isotropic angular distribution in the center-of-mass system with a library of 107 nuclide decay alpha-particle spectra, 24 sets of measured and/or evaluated (alpha, n) cross sections and product nuclide level branching fractions, and functional alpha particle stopping cross sections for Z

Nuclear reactor dynamics on-line estimation by Locally Recurrent Neural Networks

Abstract: In this paper, the Causal Recursive Back-Propagation (CRBP) algorithm is employed to train on-line an Infinite Impulse Response–Locally Recurrent Neural Network (IIR–LRNN) for modelling the dynamics of a next-generation nuclear reactor. The results demonstrate the advantages of the on-line training over the batch-mode learning in the reconstruction of complex nonlinear dynamic relationships.

Oxidation properties of Zr-Nb alloys at 973-1273 K in air

Abstract: In order to investigate the oxidation behavior of LWR cladding materials under the condition of reactor accidents, e.g. LOCA, Zr–Nb alloys with 1–10 wt%Nb and Zircaloy-4 (0 wt%Nb) were oxidized at 973–1273 K in dry air. The weight gain due to oxidation increased with Nb content at 973 and 1073 K was the smallest for 2.5 wt%Nb at 1173 and 1273 K. The oxidation kinetics obeyed the parabolic rate law without a few cases, e.g. 6–10 wt%Nb and 1273 K. The parabolic rate constant at high temperatures had the somewhat low activation energy compared to that at low temperatures. These results implied that such oxidation behaviors of Zr–Nb alloys related to the lattice structures of oxide films as well as underlying metal during oxidation. Especially at high temperatures, 6ZrO 2 –Nb 2 O 5 compound might promote the oxidation of Zr–Nb alloys with high content of Nb.

A GoldSim model for the safety assessment of an HLW repository

Abstract: An assessment program for the evaluation of a high-level waste (HLW) repository has been developed by utilizing GoldSim, by which nuclide transports in the near- and far-field of a repository as well as a transport through a biosphere under various natural and manmade disruptive events affecting a nuclide release could be modeled and evaluated. To demonstrate its usability, three illustrative cases including the influence of a groundwater flow pattern through canisters associated with a flowing groundwater through fractures, and the possible disruptive events caused by an accidental human intrusion or an earthquake have been investigated and illustrated for a hypothetical Korean HLW repository. Even though all the parameter values applied to a hypothetical repository are assumed without any real base, the illustrative cases are very informative especially when seeing the result of the probabilistic calculation with the groundwater flow patterns which is turned out to be possibly important for nuclide release and further transport in and around the repository system.

Development of a ϕ20 mm annular centrifugal contactor for the hot test of the total TRPO process

Abstract: Centrifugal contactors have many advantages and are favored for reprocessing of spent nuclear fuel and partitioning of high-level waste. A new ϕ20 mm annular centrifugal contactor for laboratory scale has been developed for the hot test of the total TRPO process. A modular design, a multi-stage group housing design and ceramic bearings are adopted in the new centrifugal contactor. A sampling system and a rotor speed acquisition and monitoring system for ϕ20 mm annular centrifugal contactors have been developed for tests in the hot cell. The hydraulic performance of both a single-stage centrifugal contactor and a three-stage cascade has been studied. The nominal throughput of the single-stage centrifugal contactor can reach 13.5 l/h under suitable operation conditions for H2O–kerosene system. The hydraulic performance of the three-stage cascade is also good.

Use of poly(methyl methacrylate) in radioactive waste management: I. Radiation stability and degradation

Abstract: In this study, in order to understand the possible use of PMMA in radioactive waste management as a solidifying agent, radiation stability of the PMMA was studied by gamma irradiations at two different dose rates of 1485 and 82.8 Gy/h. The total dose of irradiation was up to 523 kGy. Degradation nature was tested by studying the changes in mechanical and thermal properties with rate and total dose of irradiation. Ultimate tensile strength and toughness first increased and then decreased with total irradiation dose. Half value dose (HVD) for elongation was 148 kGy and it was 178 kGy for tensile strength at the dose rate of 1485 Gy/h. Half value dose was found from the extrapolation of experimental data as 228 kGy for elongation and 205 kGy for tensile strength at the dose rate of 82.8 Gy/h. The FTIR spectral analysis showed depolymerization degradation of polymer with irradiation. It was concluded from experimental results that PMMA can be used for embedding radioactive wastes.

Control rod worth calculation for VVER-1000 nuclear reactor using WIMS and CITATION codes

Abstract: This paper summarizes efforts related to developing a technically justifiable approach for investigating the control rod worth of VVER-1000 reactor. For this assessment, computer simulation of nuclear reactor was needed. In this study nuclear reactor behavior was modeled by WIMS code, which solves transport equation for fuel assemblies' modeling at first step, and CITATION code that solves diffusion equation for core modeling. From these two codes, neutronic calculation of reactor was performed and control rod worth was calculated. Results of this study are comparable with the plant's FSAR. On comparing results of this study and reference some unacceptable errors were discerned. To find out the cause of these errors, some efforts had been performed and finally was discerned that the method of cell calculation, i.e., DSN method, was the important cause of errors. Therefore, some analysis had been performed by WIMS in PIJ + PERSEUS method and was shown that the results were improved.

Computation accuracy and efficiency of a power series analytic method for two- and three- space-dependent transient problems

Abstract: The establishment of solutions to large-scale three-dimensional (3-D) reactor benchmark problems is needed to serve as standards for the verification of design codes and for the detailed error analysis of calculational methods. A number of partially and fully inserted control rods, represented by absorber added to certain subassemblies, cause a strong nonseparable power distribution. In addition, the existence of a very large thermal flux peak in the reflector makes this a very difficult and challenging problem to solve. PWS code has been developed to include a numerical solution for the time-dependent neutron diffusion equations for the nuclear reactor analysis. The new technique employs a new parameter (α) which can reduce the rapid increase in magnitude of the power series coefficients. These coefficients, in turn, are determined by back substitutions in the non-linear canonical diffusion equations and treating terms of the same degree to obtain a modified recurrence relation which is valid for any type of the stiff non-linear kinetic diffusion equations. The validity of the algorithm was tested with three kinds of well-known two-group benchmark problems. The first one is the two-dimensional TWIGL seed-blanket reactor kinetics problem. The second is the two- and three-dimensional LAR BWR benchmark problem simulating a rod drop accident of a BWR core. The third is the three-dimensional LMW LRA transient problem which simulates an operational transient involving rod movements. The obtained results with the proposed PWS code are compared with those provided by other reference codes, indicating an overall agreement and excellent performance.