Boiling Water Reactor

An optimization system based on Genetic Algorithms (GAs), in combination with expert knowledge coded in heuristics rules, was developed for the design of optimized boiling water reactor (BWR) fuel loading patterns. The system was coded in a computer program named Loading Pattern Optimization System based on Genetic Algorithms, in which the optimization code uses GAs to select candidate solutions, and the core simulator code CM-PRESTO to evaluate them. A multi-objective function was built to maximize the cycle energy length while satisfying power and reactivity constraints used as BWR design parameters. Heuristic rules were applied to satisfy standard fuel management recommendations as the Control Cell Core and Low Leakage loading strategies, and octant symmetry. To test the system performance, an optimized cycle was designed and compared against an actual operating cycle of Laguna Verde Nuclear Power Plant, Unit I.

This scoping study proposes using mixed nitride fuel in Pu-based high conversion LWR designs in order to increase the breeding ratio. The higher density fuel reduces the hydrogen-to-heavy metal ratio in the reactor which results in a harder spectrum in which breeding is more effective. A Resource-renewable Boiling Water Reactor (RBWR) assembly was modeled in MCNP to demonstrate this effect in a typical high conversion LWR design. It was determined that changing the fuel from (U,TRU)O2 to (U,TRU)N in the assembly can increase its fissile inventory ratio (fissile Pu mass divided by initial fissile Pu mass) from 1.04 to up to 1.17.► RBWR assembly modeled in MCNP/MCODE with oxide and nitride fuels to calculate breeding ratios. ► Using high density nitride fuel can increase the breeding ratio from 1.04 to 1.14. ► Results should encourage further investigations to overcome materials challenges of nitrides.

Our aim was to evaluate the sensitivity and uncertainty of mass flow rate in the core on the performance of natural circulation boiling water reactor (NCBWR). This analysis was carried out through Monte Carlo simulations of sizes up to 40,000, and the size, i.e., repetition of 25,000 was considered as valid for routine applications. A simplified boiling water reactor (SBWR) was used as an application example of Monte Carlo method. The numerical code to simulate the SBWR performance considers a one-dimensional thermo-hydraulics model along with non-equilibrium thermodynamics and non-homogeneous flow approximation, one-dimensional fuel rod heat transfer. The neutron processes were simulated with a point reactor kinetics model with six groups of delayed neutrons. The sensitivity was evaluated in terms of 99% confidence intervals of the mean to understand the range of mean values that may represent the entire statistical population of performance variables. The regression analysis with mass flow rate as the predictor variable showed statistically valid linear correlations for both neutron flux and fuel temperature and quadratic relationship for the void fraction. No statistically valid correlation was observed for the total heat flux as a function of the mass flow rate although heat flux at individual nodes was positively correlated with this variable. These correlations are useful for the study, analysis and design of any NCBWR. The uncertainties were propagated as follows: for 10% change in the mass flow rate in the core, the responses for neutron power, total heat flux, average fuel temperature and average void fraction changed by 8.74%, 7.77%, 2.74% and 0.58%, respectively.

Cover Pictures: The micrograph shows the transition from a parallel to an inclined growth of columnar dendrites in a Sn-15wt.% Pb alloy. The morphology transition is caused by the initiation of convection in the liquid phase driven by a rotating magnetic field (see contribution page 53ff).

Cover Pictures: The micrograph shows the transition from a parallel to an inclined growth of columnar dendrites in a Sn-15wt.% Pb alloy. The morphology transition is caused by the initiation of convection in the liquid phase driven by a rotating magnetic field (see contribution page 53ff).

This scoping study proposes using mixed nitride fuel in Pu-based high conversion LWR designs in order to increase the breeding ratio. The higher density fuel reduces the hydrogen-to-heavy metal ratio in the reactor which results in a harder spectrum in which breeding is more effective. A Resource-renewable Boiling Water Reactor (RBWR) assembly was modeled in MCNP to demonstrate this effect in a typical high conversion LWR design. It was determined that changing the fuel from (U,TRU)O 2 to (U,TRU)N in the assembly can increase its fissile inventory ratio (fissile Pu mass divided by initial fissile Pu mass) from 1.04 to up to 1.17.

This article provides results from an assessment of low-frequency phased array ultrasonic testing (PAUT) for detection of inside surface-breaking defects in dissimilar metal welds (DMW), centrifugally cast stainless steel (CCSS) and reactor internal pump (RIP) piping segments as applied from outer surface of the components. The evaluated specimens are typical of configurations installed in pressurized water reactors (PWRs) or advanced boiling water reactors (ABWRs). The coarse-grained and anisotropic microstructure of CCSS material makes it difficult to inspect such components. Similar inspection problems exist for dissimilar metal welds. The large grains of these materials strongly affect the propagation of ultrasound by causing severe attenuation, change in velocity, and scattering of ultrasonic energy.

Flow accelerated corrosion (FAC) wear is a serious degradation problem especially for nuclear power plants since it may result in the plant damage as well as risk to personnel. In this paper, a methodology which includes the two-phase hydrodynamic CFD models and FAC models, is proposed to predict severe FAC wear sites. Based on hydrodynamic simulation results, the present CFD models can precisely capture the two-phase characteristics within the piping system, which include the centrifugal effect, the gravitation effect and the imbalance of phase and mass separation in a T-junction, etc. Coupled with these flow characteristics, the appropriate FAC indicators can predict the possible locations of severe FAC wear. This methodology was validated against the measured results of wear site distributions for the piping system in a boiling water reactor (BWR) power plant. Good agreement between measurements and predictions at severe wear sites indicate that the present models can capture the characteristics of severe FAC wear and can help assist in the pipe wall-monitoring program for a nuclear power plant.

Nomenclature a crack depth B specimen thickness or crack front length B 0 reference thickness for Master Curve analysis (B 0 = 25.4 mm) b ligament size E Young's modulus J Jintegral Jc critical value of J integral K stress intensity factor K0

The long term containment cooling of GE's passive BWR design is based on a new safety system called PCCS (passive containment cooling system). Performance of this system relies on the pressure difference between the drywell and wetwell in case of an accident and on the condensation of steam moving downward inside vertical tubes fully submerged in a water pool initially at room temperature. In this paper a model based on the resolution of momentum equations of both phases, the application of the heat and mass transfer analogy, and the consideration of the presence of a noncondensable gas by diffusion theory in a boundary layer is presented. Assumptions and approximations taken resulted in new methods to estimate film thickness and heat transport from the gas to the interface. Influence of phenomena such as suction, flow development, film waviness, and droplet entrainment has been accounted for. Based on this formulation, a computer programme called HVTNC (heat transfer in vertical tubes with noncondensables) has been built up. HVTNC results have been compared to the experimental data available. Experimental trends have been reproduced. Heat transfer has been found to be severely degraded by the presence of noncondensables whereas high Reynolds numbers of gas flow have been seen to enhance shear stress and therefore, heat transmission. The average error of HVTNC is essentially located at regions where only a residual fraction of heat remains to be transferred, so that minor deviations can be anticipated in the overall heat transfer in the tube. Comparison of HVTNC to other models show a substantial gain of accuracy with respect to earlier models. © 1997 Elsevier Science S.A.

Boron carbide (B 4 C) is widely used as an absorber material in many commercial reactors, such as boiling water reactors, Russian VVERs, later French pressurised water reactors, and planned to be used in the European Pressurised Water Reactor (EPR). Under hypothetical severe accident conditions, B 4 C reacts with its surrounding stainless steel cladding, producing eutectic melts above 1200°C; remaining bare B 4 C and B 4 C/metal mixtures are then exposed to steam and oxidize highly exothermically. As well as hydrogen, gases and aerosols containing boron and carbon compounds are produced, which affect the transport and deposition of radiologically important fission products such as iodine and caesium in the circuit, and subsequent behaviour in the containment.

Two-fluid model predictions of film dryout in annular flow, leading to nuclear reactor fuel failure, are limited by the uncertainties in the constitutive relations for the entrainment rate of droplets from the liquid film. The main cause of these uncertainties is the lack of separate-effects experimental data in the range of the operating conditions in nuclear power reactors. An air-water experiment has been performed to measure the entrainment rate in a small pipe. The current data extend the available database in the literature to higher gas and liquid flows and also to higher pressures. The measurements were made with the film extraction technique. A mechanistic model was obtained based on Kelvin-Helmholtz' instability theory. The dimensionless model includes the Weber number of the gas and the liquid film Reynolds number. Kataoka and Ishii's correlation (Kataoka, I., Ishii, M., 1982. NUREG/CR-2885, ANL-82-44) is modified based on this model and the new data. The new correlation collapses the present air -water data and Cousins and Hewitt's data (Cousins, L.B., Hewitt, G.F., 1968. UKAEA Report AERE-R5657) The effects of pressure and surface tension were considered in the derivation so it may be applied for boiling water reactor operating conditions. © 1997 Elsevier Science S.A.

The High Performance Light Water Reactor (HPLWR), how the European Supercritical Water Cooled Reactor is called, is a pressure vessel type reactor operated with supercritical water at 25 MPa feedwater pressure and 500 • C average core outlet temperature. It is designed and analyzed by a European consortium of 10 partners and 3 active supporters from 8 Euratom member states in the second phase of the HPLWR project. Most emphasis has been laid on a core with a thermal neutron spectrum, consisting of small fuel assemblies in boxes with 40 fuel pins each and a central water box to improve the neutron moderation despite the low coolant density. Peak cladding temperatures of the fuel rods have been minimized by heating up the coolant in three steps with intermediate coolant mixing. The containment design with its safety and residual heat removal systems is based on the latest boiling water reactor concept, but with different passive high pressure coolant injection systems to cause a forced convection through the core. The design concept of the steam cycle is indicating the envisaged efficiency increase to around 44%. Moreover, it provides the constraints to design the components of the balance of the plant. The project is accompanied by numerical studies of heat transfer of supercritical water in fuel assemblies and by material tests of candidate cladding alloys, performed by the consortium and supported by additional tests of the Joint Research Centre of the European Commission. Besides the scientific and technical progress, the HPLWR project turned out to be most successful in training the young generation of nuclear engineers in the technologies of light water reactors. More than 20 bachelor or master theses and more than 10 doctoral theses on HPLWR technologies have been submitted at partner organizations of this consortium since the start of this project.

After the NRC issued a revised license renewal rule in May 1995, the nuclear industry focussed on developing generic industry for implementing the rule and testing the guidance through various demonstration programs and work products in conjunction with the NRC. In addition, plant-specific programs also proceeded forward. These activities show that implementation issues continue to exist. Since the issuance of the rule, the NRC has issued a draft standard review plan for license renewal (SRP-LR), Working Draft, September 1997. Southern Nuclear Operating Company (SNC) has begun development work on a license renewal application for Plant Hatch Units 1 and 2. Plant Hatch Units 1 and 2 are BWR 4, Mark I plants whose operating licenses expire in 2014 and 2018, respectively. The Plant Hatch initiative also involves teaming with other Boiling Water Reactors (BWRs) to develop the license renewal technology within the BWR fleet, and to support Plant Hatch by providing an oversight role for the application process. The teaming effort involved two other utilities, each being assigned to prepare a common report on a mechanical system or a structure. The common report could be presented to the NRC with modifications to suit the individual plants, thereby saving time and money, and hopefully resulting in quicker approval by the NRC. The desired license renewal process end result is a renewed license with up to a 20 year extension (10CFR 54.31(b)).

Digby D. Macdonald (PI), Mirna Urquidi-Macdonald (Co-PI), John H Mahaffy (Co-PI) Amit Jain (Graduate Assistant)*, Han Sang Kim (Graduate Assistant***), Vishisht Gupta (Graduate Student**), Jonathan Pitt (Graduate Assistant*) * Graduate with a Master degree under ...

The Ignalina Nuclear Power Plant (NPP) consists of two units, commissioned in 1983 and 1987. Unit 1 of Ignalina NPP was shutdown for decommissioning at the end of 2004 and Unit 2 is to be operated until the end of 2009. Both units are equipped with channel-type graphite-moderated boiling water reactors RBMK-1500. The safety issues related with fulfilment of two fundamental safety functions-control of the reactivity and confinement of radioactive materials are addressed in this paper. The main systems, performing these safety functions at NPP-s, are Reactor Shutdown and Containment systems. The containment function at the Ignalina NPP is performed by the Accident localisation system. The performance and safety justification of Ignalina NPP additional and diverse shutdown systems, as well as of Accident localisation system, is discussed in this paper. The performed safety evaluations demonstrate that Ignalina NPP has adequate safety systems for reactor shutdown and confinement of radioactive materials.

An important source of uncertainty in boiling water reactor physics is associated with the precise characterisation of the moderation properties of the coolant and by-pass regions, with significant impact on reactor physics parameters such as the lattice neutron multiplication, the neutron migration area and the pin-by-pin power distribution. In this paper, the effects of certain relevant void-fraction uncertainties on reactor physics parameters have been studied for a BWR assembly of the type Westinghouse SVEA-96 using CASMO-4, HELIOS/PRESTO-2 and MCNP4C. The SVEA-96 geometry is characterised by the sub-division of the assembly into four different sub-bundles, by means of an inner by-pass with a cruciform shape. The study has covered: (a) the effects of different cross-section data libraries on the void coefficient of reactivity, for a wide range of void fractions; (b) the consideration of a water film inside the sub-bundle walls, and (c) the impact of partly inserted absorber blades producing very different void fractions in different sub-bundles.

An optimization procedure based on the tabu search (TS) method was developed for the design of radial enrichment and gadolinia distributions for boiling water reactor (BWR) fuel lattices. The procedure was coded in a computing system in which the optimization code uses the tabu search method to select potential solutions and the HELIOS code to evaluate them. The goal of the procedure is to search for an optimal fuel utilization, looking for a lattice with minimum average enrichment, with minimum deviation of reactivity targets and with a local power peaking factor (PPF) lower than a limit value. Time-dependent-depletion (TDD) effects were considered in the optimization process. The additive utility function method was used to convert the multiobjective optimization problem into a single objective problem. A strategy to reduce the computing time employed by the optimization was developed and is explained in this paper. An example is presented for a 10×10 fuel lattice with 10 different fuel compositions. The main contribution of this study is the development of a practical TDD optimization procedure for BWR fuel lattice design, using TS with a multiobjective function, and a strategy to economize computing time.

Three element Steam Drum (SD) Level Controller has been conventionally used for most of the boilers, Nuclear power plant steam generator & Boiling Water Reactor (BWRs). Based on the process dynamic studies it was found that this scheme does not work properly for an interacting, interconnected multiple loop boiling water system i.e., Advanced Heavy Water Reactor (AHWR). It is a pressure tube type light water cooled heavy water moderated Boiling Water Reactor (BWR). It has 4-inter-connected parallel loops with 113 4 = 452 boiling channels in the Main Heat Transport (MHT) system. These multiple (four) interconnected loops influences the Steam Drum (SD) level control adversely. Such a behavior has not been reported in the open literature. The open loop response is stable, non-oscillatory and non-diverging for a step change in the feed flow rates. Also it is not possible

This paper reports on the modeling and simulation of flashing-induced instabilities in natural-circulation systems, with special emphasis on natural-circulation boiling water reactors (BWRs). For the modeling the 4-equation two-phase model FLOCAL . Ein teoretisches Modell fur Zweiphasen-stromungen in wassergekulthen Kernreaktoren und seine Anwendung zur Analyse des Naturumlaufs im Heizreaktor AST-500. Ph.D. dissertation, Akademie der Wissenschaften der DDR, Dresden], developed at the Forschungszentrum Rossendorf (FZR, Germany), has been used. The model allows for the liquid and vapor to be in thermal non-equilibrium and, via drift-flux models, to have different velocities.

In this paper, we perform a parametric study of the nonlinear dynamics of a reduced order model for boiling water reactors (BWR) near the Hopf bifurcation point using the method of multiple scales (MMS). Analysis has been performed for general values of the parameters, but the results are demonstrated for parameter values of the model corresponding to the advanced heavy water reactor (AHWR). The neutronics of the AHWR is modeled using point reactor kinetic equations while a one-node lumped parameter model is assumed both for the fuel and the coolant for modeling the thermal-hydraulics. Nonlinearities in the heat transfer process are ignored and attention is focused on the nonlinearity introduced by the reactivity feedback. It is found that the steady-state operation of the AHWR mathematical model looses stability via. a Hopf bifurcation resulting in power oscillations as some typical bifurcation parameter like the void coefficient of reactivity is varied. The bifurcation is found to be subcritical for the parameter values corresponding to the AHWR. However, with a decrease in the fuel temperature coefficient of reactivity the bifurcation turns to supercritical implying global stability of the steady state operation in the linear stability regime. Moreover slight intrusion into the instability regime results in small-amplitude limit cycles leaving the possibility of retracting back to stable operation.

Findings from a four year study by an international benchmarking group in the comparison of computational methods for evaluating burnup credit in criticality safety analyses are presented in this paper. Approximately 20 participants from 11 countries have provided results for most problems. Four detailed benchmark problems for Pressurized Water Reactor (PWR) fuel have been completed and are summarized in this paper. Preliminary results from current work addressing burnup credit for Boiling Water Reactor (BWR) fuel will also be discussed as well as planned activities for additional benchmarks including Mixed-Oxide (MOX) fuels, subcritical benchmarks, international databases, and other activities.

An analytical model has been developed to study the nuclear-coupled density-wave instability in the Indian advanced heavy water reactor (AHWR) which is a natural circulation pressure tube type boiling water reactor. The model considers a point kinetics model for the neutron dynamics and a lumped parameter model for the fuel thermal dynamics along with the conservation equations of mass, momentum and energy and equation of state for the coolant. In addition, to study the effect of neutron interactions between different parts of the core, the model considers a coupled multipoint kinetics equation in place of simple point kinetics equation. Linear stability theory was applied to reveal the instability of in-phase and out-of-phase modes in the boiling channels of the AHWR. The results indicate that the stability behavior of the reactor is greatly influenced by the void reactivity coefficient, fuel time constant, radial power distribution and channel inlet orificing. The delayed neutrons were found to have a strong influence on the Type I and Type II instabilities observed at low and high channel powers, respectively. Also, it was found that the coupled multipoint kinetics model and the modal point kinetics model predict the same threshold power for out-of-phase instability if the coupling coefficient in the former model is half the eigen value separation between the fundamental and the first harmonic mode in the latter model. Decay ratio maps were predicted considering various operating parameters of the reactor, which are useful for its design.

In this paper we study the boundary conditions during out of phase oscillations, in a system formed by two parallel channels coupled to multimodal neutron kinetics. The fact that the pressure drop can change with time, but remains the same in all the parallel channels, leads us to analytical integration of the time derivative term of the channel momentum equation, in order to get a dynamic equation for the inlet mass flux to each channel. From these inlet mass flux dynamic equations plus the appropriate boundary conditions, we have obtained the expression of the time dependent common pressure drop to all the channels. Finally the oscillations in the pressure drop and in the inlet mass flux to the channels have been investigated when out of phase oscillations take place. #

This paper presents the training of an artificial neural network (ANN) to accurately predict, in very short time, a physical parameter used in nuclear fuel reactor optimization: the local power peaking factor (LPPF) in a typical boiling water reactor (BWR) fuel lattice. The ANN training patterns are distribution of fissile and burnable poison materials in the fuel lattice and their associated LPPF. These data were obtained by modeling the fuel lattices with a neutronic simulator: the HELIOS transport code. The combination of the pin U 235 enrichment and the Gd 2 O 3 (gadolinia) concentration, inside the 10 Â 10 fuel lattice array, was encoded by three different methods. However, the only encoding method that was able to give a good prediction of the LPPF was the method which added the U 235 enrichment and the gadolinia concentration. The results show that the relative error in the estimation of the LPPF, obtained by the trained ANN, ranged from 0.022% to 0.045%, with respect to the HELIOS results.

This paper presents some results of the implementation of several optimization algorithms based on ant colonies, applied to the fuel reload design in a Boiling Water Reactor. The system called Azcaxalli is constructed with the following algorithms: Ant Colony System, Ant System, Best-Worst Ant System and MAX-MIN Ant System. Azcaxalli starts with a random fuel reload. Ants move into reactor core channels according to the State Transition Rule in order to select two fuel assemblies into a 1/8 part of the reactor core and change positions between them. This rule takes into account pheromone trails and acquired knowledge. Acquired knowledge is obtained from load cycle values of fuel assemblies. Azcaxalli claim is to work in order to maximize the cycle length taking into account several safety parameters. Azcaxalli's objective function involves thermal limits at the end of the cycle, cold shutdown margin at the beginning of the cycle and the neutron effective multiplication factor for a given cycle exposure. Those parameters are calculated by CM-PRESTO code. Through the Haling Principle is possible to calculate the end of the cycle. This system was applied to an equilibrium cycle of 18 months of Laguna Verde Nuclear Power Plant in Mexico. The results show that the system obtains fuel reloads with higher cycle lengths than the original fuel reload. Azcaxalli results are compared with genetic algorithms, tabu search and neural networks results.

In this paper, we perform a parametric study of the nonlinear dynamics of a reduced order model for boiling water reactors (BWR) near the Hopf bifurcation point using the method of multiple scales (MMS). Analysis has been performed for general values of the parameters, but the results are demonstrated for parameter values of the model corresponding to the advanced heavy water reactor (AHWR). The neutronics of the AHWR is modeled using point reactor kinetic equations while a one-node lumped parameter model is assumed both for the fuel and the coolant for modeling the thermal-hydraulics. Nonlinearities in the heat transfer process are ignored and attention is focused on the nonlinearity introduced by the reactivity feedback. It is found that the steady-state operation of the AHWR mathematical model looses stability via. a Hopf bifurcation resulting in power oscillations as some typical bifurcation parameter like the void coefficient of reactivity is varied. The bifurcation is found to be subcritical for the parameter values corresponding to the AHWR. However, with a decrease in the fuel temperature coefficient of reactivity the bifurcation turns to supercritical implying global stability of the steady state operation in the linear stability regime. Moreover slight intrusion into the instability regime results in small-amplitude limit cycles leaving the possibility of retracting back to stable operation.

In his previous paper and in his recent letter, to which we are responding, Dr. Turnbull clearly indicates that his approach in analyzing the potential drop external to the crack works in the following manner. By assuming some value for the electrode potential at crack mouth, E mouth , he calculates the potential at crack tip, E tip , and the current density¯owing from the crack at the crack mouth, i mouth , (by solving the transport equations inside the crack) and subsequently obtains the corrosion potential, E corr , (by solving the LaplaceÕs equation outside the crack). In doing so, the relationship between E tip and E corr is obtained. Dr. Turnbull oers this approach as an alternative to convergence on i mouth , which we described a decade ago in developing the original coupled environment fracture model (CEFM) . We cannot agree with TurnbullÕs approach, because the corrosion potential, E corr , is fully determined by the properties of the external environment (including the substrate) at a point that is suciently remote from the crack that it cannot be in¯uenced by the potential at the crack mouth. In other words, the corrosion potential is not a crack variable and hence cannot be used as the variable in an iterative convergence cycle. More succinctly put, the point at which the local potential at the surface can be

This paper discusses the effects of truncation error on the stability of natural circulation in a single-phase loop as predicted by finite difference numerical methods. A short description of the characteristics of the selected problem is firstly given, also showing results obtained by a modal solution. Different numerical schemes are applied and the related effect on predicted stability is shown. The results obtained are then discussed in relation to the applicability of thermal-hydraulic system codes in the analysis of fluid-dynamic instabilities in real systems.

The aim of this paper is to explore the application of detrended fluctuation analysis (DFA) to study boiling water reactor stability. DFA is a scaling method commonly used for detecting long-range correlations in non-stationary time series. This method is based on the random walk theory and was applied to neutronic power signal of Forsmark stability benchmark. Our results shows that the scaling properties breakdown during unstable oscillations.

A new methodology is developed for the prediction of RPV embrittlement that utilizes a combination of domain models and nonlinear estimators including neural networks and nearest neighbor regressions. The Power Reactor Embrittlement Database is used in this study. The results from newly developed nearest neighbor projective fuser indicate that the combined embrittlement predictor achieved about 67.3% and 52.4% reductions in the uncertainties for General Electric Boiling Water Reactor plate and weld data compared to Regulatory Guide 1.99, Revision 2, respectively. The implications of irradiation temperature effects to the development of radiation embrittlement models are then discussed.

Our work involved the continued development of the theory of passivity and passivity breakdown, in the form of the Point Defect Model, with emphasis on zirconium and zirconium alloys in reactor coolant environments, the measurement of critically-important ...

Austenitic alloy weldments in nuclear reactor systems are susceptible to stress corrosion cracking (SCC) failures. SCC has been observed for decades and continues to be a primary maintenance concern for both pressurized water and boiling water reactors. SCC can occur if the sum of residual stress and applied stress exceeds a critical threshold tensile stress. Residual stresses developed by prior

A RETRAN-02 model was devised and benchmarked against the preliminary safety analysis report (PSAR) for the Lungmen nuclear power plant roughly 10 years ago. During these years, the fuel design, some of the reactor vessel designs, and control systems have since been revised. The Lungmen RETRAN-02 model has also been modified with updated information when available. This study uses the analytical results of the final safety analysis report (FSAR) to benchmark the Lungmen RETRAN-02 plant model. Five transients, load rejection (LR), turbine trip (TT), main steam line isolation valves closure (MSIVC), loss of feedwater flow (LOFF), and one turbine control valve closure (OTCVC), were utilized to validate the Lungmen RETRAN-02 model. Moreover, due to the strong coupling effect between neutron dynamics and the thermal-hydraulic response during pressurization of transients, the one-dimensional kinetic model with the cross-section data library is used to simulate the coupling effect. The analytical results show good agreement in trends between the RETRAN-02 calculation and the Lungmen FSAR data. Based on the benchmark of these design-basis transients, the modified Lungmen RETRAN-02 model has been adjusted to a level of confidence for analysis of pressure increase transients. Analytical results indicate that the Lungmen advanced boiling water reactor (ABWR) design satisfied design criteria, i.e., vessel pressure and hot shutdown capability. However, a slight difference exists in the simulation of the water level for cases with changes in water levels. The Lungmen RETRAN-02 model tends to predict the change in water level at a slower rate than that in the Lungmen FSAR. There is also a slight difference in void reactivity response toward vessel pressure change in both simulations, which causes the calculated neutron flux before reactor shutdown to differ to some degree when the reactor experiences a rapid pressure increase. Further studies will be performed in the future using Lungmen startup test data.

SCENARIO FOR THORIUM FUEL CYCLE WITH FREE 233 U IN BWR. This paper presents the scenario for thorium fuel cycle with free 233 U in boiling water reactor (BWR). In this study we have utilized plutonium and minor actinides (MA) as fissile nuclides instead of 233 U as one of the main scenario to obtain the 233 U free BWR core with thorium. Beside that, the void-fraction of the reactor is modified from 20% up to 70%. The results show that the standard BWR core can maintain its criticality when the loaded fuel is thorium with 11.16% and 1.24 % of plutonium and MA, respectively. Moreover, the lesser amount of plutonium and minor actinides in the fuel will results in the great degradation on the safety of reactor

The use of uranium-plutonium mixed oxide fuel (MOX) in light water reactors (LWR) is nowadays a current practice in several countries. Generally 1/3 of the reactor core is loaded with MOX fuel assemblies and the other 2/3 with uranium assemblies. Nevertheless the plutonium utilization could be more effective if the full core could be loaded with MOX fuel. In this paper the design of a boiling water reactor (BWR) core fully loaded with an overmoderated MOX fuel design is investigated. The design of overmoderated BWR MOX fuel assemblies based on a 10Â10 lattice are developed, these designs improve the neutron spectrum and the plutonium consumption rate, compared with standard MOX assemblies. In order to increase the moderator to fuel ratio two approaches are followed: in the first approach, 8 or 12 fuel rods are replaced by water rods in the 10Â10 lattice; in the second approach, an 11Â11 lattice with 24 water rods is designed with an active fuel length very close to the standard MOX assembly. The results of the depletion behavior and the main steady state core parameters are presented. The feasibility of a full core loaded with the 11Â11 overmoderated MOX fuel assembly is verified. This design take advantage of the softer spectrum comparable to the 10Â10 lattice with 12 water rods but with thermal limits comparable to the standard MOX fuel assembly. #

A very complex type of power instability occurring in boiling water reactor (BWR) consists of out-of-phase regional oscillations, in which normally subcritical neutronic modes are excited by thermal-hydraulic feedback mechanisms. The out-of-phase mode of oscillation is a very challenging type of instability and its study is relevant because of the safety implications related to the capability to promptly detect any such inadvertent occurrence by in-core neutron detectors, thus triggering the necessary countermeasures in terms of selected rod insertion or even reactor shutdown. In this work, simulations of out-of-phase instabilities in a BWR obtained by assuming an hypothetical continuous control rod bank withdrawal are being presented. The RELAP5/Mod3.3 thermal-hydraulic system code coupled with the PARCS/2.4 3D neutron kinetic code has been used to simulate the instability phenomenon. Data from a real BWR nuclear power plant (NPP) have been used as reference conditions and reactor parameters. Simulated neutronic power signals from local power range monitors (LPRM) have been used to detect and study the local power oscillations. The decay ratio (DR) and the natural frequency (NF) of the power oscillations (typical parameters used to evaluate the instabilities) have been used in the analysis. The results are discussed also making use of two-dimensional plots depicting relative core power distribution during the transient, in order to clearly illustrate the out-of-phase behavior.

DISCLAIMER Any opinions, findings, and conclusions or recommendations expressed in this material are those of the author(s) and do not necessarily reflect the views of the Department of Energy. design feasibility assessment reported in this set of papers justifies embarking upon an experimental investigation of material compatibility. * T c is the cycle length. A negative sign implies that the storage costs need to be referred back in time to the reference date.

The High Performance Light Water Reactor (HPLWR), how the European Supercritical Water Cooled Reactor is called, is a pressure vessel type reactor operated with supercritical water at 25 MPa feedwater pressure and 500 • C average core outlet temperature. It is designed and analyzed by a European consortium of 10 partners and 3 active supporters from 8 Euratom member states in the second phase of the HPLWR project. Most emphasis has been laid on a core with a thermal neutron spectrum, consisting of small fuel assemblies in boxes with 40 fuel pins each and a central water box to improve the neutron moderation despite the low coolant density. Peak cladding temperatures of the fuel rods have been minimized by heating up the coolant in three steps with intermediate coolant mixing. The containment design with its safety and residual heat removal systems is based on the latest boiling water reactor concept, but with different passive high pressure coolant injection systems to cause a forced convection through the core. The design concept of the steam cycle is indicating the envisaged efficiency increase to around 44%. Moreover, it provides the constraints to design the components of the balance of the plant. The project is accompanied by numerical studies of heat transfer of supercritical water in fuel assemblies and by material tests of candidate cladding alloys, performed by the consortium and supported by additional tests of the Joint Research Centre of the European Commission. Besides the scientific and technical progress, the HPLWR project turned out to be most successful in training the young generation of nuclear engineers in the technologies of light water reactors. More than 20 bachelor or master theses and more than 10 doctoral theses on HPLWR technologies have been submitted at partner organizations of this consortium since the start of this project.

An analytical model has been developed to study the nuclear-coupled density-wave instability in the Indian advanced heavy water reactor (AHWR) which is a natural circulation pressure tube type boiling water reactor. The model considers a point kinetics model for the neutron ...

A comprehensive mathematical model of boiling-water reactors (BWRs) for simulating nuclear coupled density wave oscillations is presented. A one-dimensional core dynamics program NARS has been developed based on this model. A General Electric BWR/6 system is analyzed using NARS. Our analysis shows that, even with very large power oscillations, the amplitude of heat flux oscillations is very low. The reduction in minimum critical heat flux ratio with large power oscillations is also observed to be quite low. Thus it can be concluded that the boiling transition does not occur even at very large power oscillations and hence fuel integrity is not compromised. It has been shown that void reactivity feedback has significant effect on the amplitude of the limit cycle oscillations and always has a destabilizing effect on the system. It is observed that the pressure loss coefficient in jet pumps has a significant effect on the reactor stability.

As part of a joint research programme between the Paul Scherrer Institute (PSI) and swissnuclear, with the co-operation of the Leibstadt nuclear power plant in Switzerland and fuel suppliers Westinghouse Sweden, measurements and calculations have been made of the axial and radial distributions of fission and 238 U capture rates in the fuel rods of a Westinghouse SVEA-96 Optima2 boiling water reactor assembly. The measurements, made in the zero-energy research reactor PROTEUS at PSI, have been compared with calculations carried out using the Monte Carlo code MCNPX. The results reported are for the regions near the ends of the part-length fuel rods, which are a feature of SVEA-96 Optima2 assemblies. The sudden increase in moderation above the ends of the part-length rods leads to power peaking in the adjacent rods. Careful attention needs to be given to this phenomenon in the deployment of such fuel, the present paper providing experimental evidence for the ability of a stochastic code to predict such effects.