Large Pressurized Water Reactor Research Articles

Improvement of Nodal Accuracy by Using Albedo-Corrected Parameterized Equivalence Constants

This paper introduces the albedo-corrected parameterized equivalence constants (APEC) method, a new method for correcting the homogenized two-group cross sections of the pressurized water reactor (PWR) fuel assemblies (FAs) by taking into account the neutron leakage. First, an analysis was performed of the position dependence of the assembly-homogenized two-group cross sections in an actual core. In order to eliminate the two-group cross-section error in the conventional homogenization method, the APEC method is proposed which parameterizes the homogenized two-group cross sections in terms of an integrated albedo information current-to-flux ratio (CFR). Also, small color-set models are introduced to obtain physically meaningful CFR boundary conditions for the APEC method and their characteristic features are discussed. In the case of FAs with neighboring baffle, slightly modified APEC functions are introduced to deal with the strong spectral interaction between the FA and the baffle-reflector region in PWRs. In addition, an improved APEC function is developed by explicitly accounting for the neutron spectrum change in a FA in terms of a spectral index defined as the fast-to-thermal-flux ratio. For the test of the proposed APEC functions, a small modular reactor (SMR) core was chosen and comparative analyses were performed in detail for each type of homogenized two-group cross section. In this work, the transport lattice code DeCART2D was used for the analysis of the benchmark problems. In the comparative analyses, the APEC-corrected cross sections were compared with the conventional two-group constants and reference ones for several representative FAs. The APEC algorithm was implemented into an in-house nodal expansion method code in conjunction with a partial-current CMFD (p-CMFD) acceleration. The nodal analyses of an SMR initial core and a large PWR core were performed to evaluate the performance of the APEC method. In order to show the generality of the APEC functions obtained from lattice calculations, several modified core configurations were also analyzed. In addition, a rodded SMR initial core problem was also analyzed to test the APEC method in an extremely abnormal core configuration. The nodal analyses showed that the APEC method can improve the nodal accuracy significantly with a small amount of additional computing cost.

Model-free adaptive control law for nuclear superheated-steam supply systems

Nuclear steam supply system (NSSS), which provides saturated or superheated steam to the secondary loop system for electricity or cogeneration, is central in a nuclear plant. The control of NSSS is important for the safe, stable and efficient plant operation. Nuclear superheated-steam supply system (Su-NSSS), which is equipped with an under moderated fission reactor and a once-through steam generator (OTSG) for producing superheated steam flow, is an important type of NSSS that widely applied in the nuclear plants based on the common large pressurized water reactors (PWR) and small modular reactors (SMRs) such as integral PWR (iPWR) and modular high temperature gas-cooled reactor (MHTGR). There is still very limited results for the control of Su-NSSS. Motivated by the gap between the importance and the lack of results in the field of Su-NSSS control, a model-free adaptive power-level control law is developed for Su-NSSSs with forced primary circulation in this paper. This new control law is free from physical and thermal-hydraulic parameters, is in the simple proportional-integral (PI) or proportional-differential (PD) form. To verify the feasibility of the newly-built Su-NSSS controller, it is applied to the NSSS control of a two modular high temperature gas-cooled nuclear plant with comparison to a specific MHTGR-based NSSS control law. Numerical simulation results show that this new control law can improve the transient performance of neutron flux and steam temperature in normal operation case, and can realize system resilience in some abnormal cases.

Preliminary study of applying adjoint-based mesh optimization method to nuclear power plant safety analysis

Although the number of meshes and their distribution are important parameters that affect both computation resources and the accuracy of the numerical results, meshing of most of all computational problems still depends highly on the users’ experiences. In this paper, a deterministic adjoint-based method of optimizing mesh distribution is proposed. The developed method is applied to a nuclear power plant safety analysis. The mesh optimization was performed with 1D steady state cylindrical nuclear fuel problem first. Radial and axial mesh distributions are optimized respectively. With no surprise, the optimized mesh system performs superior than the same number of uniformly meshed system. However, it was unexpected that the optimized mesh retains generality and therefore, the optimized mesh system can be still the best mesh system for given number of meshes under different condition or even during transient analysis. The authors applied the optimized mesh distribution to nuclear system safety analysis. A large pressurized water reactor cold leg guillotine break (LBLOCA) scenario was analyzed and the consequence of different mesh systems is investigated and discussed. From this preliminary study the usefulness and implication of the adjoint based mesh optimization method for the nuclear safety analysis is uncovered.

Improving power and desalination capabilities of a large nuclear power plant with supercritical CO2 power technology

To response to the increasing demands for clean water, a large pressurized water reactor (PWR) with a desalination capability has been studied and demonstrated its potential so far. However, the electricity production of the large nuclear reactor decreases by 10% due to steam bypass for desalination. In this study, the authors evaluate the possibility of a large PWR with a capability of producing both electric power and clean water by using the supercritical CO2 (S-CO2) Brayton cycle technology. The S-CO2 power technology is adopted to minimize the decrease in the electricity production capacity due to desalination process. Two concepts which replace the existing steam based power conversion system with a S-CO2 Brayton cycle were proposed. The first concept is that the low pressure steam turbine section of the power conversion system is replaced with the S-CO2 Brayton cycle. The second concept is that the whole steam based power conversion system is replaced with the S-CO2 Brayton cycle. Several S-CO2 cycle options were considered in terms of power production and the desalination capacity and conducted a comparative analysis of selected layouts and the optimal operating conditions of the suggested layouts were identified.

Core design options of an ultra-long-cycle sodium cooled reactor with effective use of PWR spent fuel for sustainable energy supply

In this work, 350MWe ultra-long-cycle sodium-cooled reactor cores are designed to supply electric energy over ~60 Effective Full Power Years (EFPYs) without refueling and with an effective use of Transuranics (TRU) and uranium from large pressurized water reactor (PWR) spent fuel stocks. The core employs the axial blanket-driver-blanket (ABDB) burning strategy, which was recently proposed by the authors to achieve an ultra-long-cycle length with self-controllability under unprotected accidents. In particular, a thorium–uranium fuel cycle is considered to remove the heterogeneity of the fuel assemblies for design simplification and to improve the core performance parameters by selectively adding thorium into both blanket and driver fuels. The results show that the use of TRU nuclides from PWR spent fuel leads to significant extension of the fuel cycle length, but considerable increase of burnup reactivity swing. In addition, these results also indicate that the uranium–thorium mixed fuels both in the lower blanket and driver considerably improve the inherent safety of the ultra-long-cycle core by reducing burnup reactivity and sodium void worth; this makes it possible to simplify the previous heterogeneous fuel assembly design with improved core performances. Copyright © 2016 John Wiley & Sons, Ltd.

MELCOR and GOTHIC Analyses of a Large Dry PWR Containment to Support Resolution of GSI-191

The thermal-hydraulic response of large dry pressurized water reactor containments under loss-of-coolant-accident conditions “particularly with respect to containment pressure and sump pool temperature” is crucial for risk-informed decision making about Generic Safety Issue 191. Texas A&M University has developed models with several computer codes including MELCOR and GOTHIC to model such scenarios.MELCOR is a best-estimate thermal-hydraulic and severe accident code created and actively maintained by Sandia National Laboratories for the U.S. Nuclear Regulatory Commission. GOTHIC is a thermal-hydraulic software package meant for design, licensing, and safety calculations for, among other systems, nuclear power plant containments. It was developed and is maintained by Numerical Applications Inc. for the Electronic Power Research Institute.The overarching goal of the analyses presented here is twofold: (1) produce best-estimate time profiles of sump pool temperature under double-ended guillotine-break conditions with MELCOR and GOTHIC and (2) investigate differences between the MELCOR and GOTHIC code results via a sensitivity study. The sump pool temperature was selected as a key parameter to compare because it has direct implications for sump pool chemistry, residual heat removal during recirculation, and pressure drop across sump screens.Aspects of the MELCOR and GOTHIC modeling strategies are discussed, and best estimates of the containment thermal-hydraulic response are presented. There are significant disagreements between code predictions. Hypotheses to explain the differences are tested through a comparative code sensitivity study. In this context, “sensitivity” refers to how containment thermal hydraulics respond to differences in code inputs or code phenomenological models. Sensitivity calculations are performed to exclude, individually, the model effects on comparative thermal-hydraulic responses of containment fan coolers, containment sprays, thermal surface condensation/films, and break source definition. Calculations are also performed with multiple models excluded. Using containment sump pool temperature as an indicator, the most impactful physics in terms of code agreement are those of thermal surfaces (condensation, film phenomena) whereas fan cooler models have a minimal effect. Containment spray exclusion results in disagreement in parts of the event sequence, while break source definition and/or break effluent flashing models lead to disagreement.

A two dimensional approach for temperature distribution in reactor lower head during severe accident

In order to evaluate the safety margin during a postulated severe accident, a module named ASAP-2D (Accident Simulation on Pressure vessel-2 Dimensional), which can be implemented into the severe accident simulation codes (such as ATHLET-CD), is developed in Shanghai Jiao Tong University. Based on two-dimensional spherical coordinates, heat conduction equation for transient state is solved implicitly. Together with solid vessel thickness, heat flux distribution and heat transfer coefficient at outer vessel surface are obtained. Heat transfer regime when critical heat flux has been exceeded (POST-CHF regime) could be simulated in the code, and the transition behavior of boiling crisis (from spatial and temporal points of view) can be predicted.The module is verified against a one-dimensional analytical solution with uniform heat flux distribution, and afterwards this module is applied to the benchmark illustrated in NUREG/CR-6849. Benchmark calculation indicates that maximum heat flux at outer surface of RPV could be around 20% lower than that of at inner surface due to two-dimensional heat conduction. Then a preliminary analysis is performed on the integrity of the reactor vessel for which the geometric parameters and boundary conditions are derived from a large scale advanced pressurized water reactor. Results indicate that heat flux remains lower than critical heat flux. Sensitivity analysis indicates that outer heat flux distribution is more sensitive to input heat flux distribution and the transition boiling correlation than mass flow rate in external reactor vessel cooling (ERVC) channel, and the correlation for molten vessel and ERVC coolant inlet temperature.According to the results achieved, the new developed module shows good applicability to simulate the pressure vessel behavior during melt pool formation. Thus it can be applied for the future study of the severe accidents relating to lower head integrity.

How to Produce a Reactor Neutron Spectrum Using a Proton Accelerator

A method for reproducing the neutron energy spectrum present in the core of an operating nuclear reactor using an engineered target in an accelerator proton beam is proposed. The protons interact with a target to create neutrons through various (p,n) type reactions. Spectral tailoring of the emitted neutrons can be used to modify the energy of the generated neutron spectrum to represent various reactor spectra. Through the use of moderators and reflectors, the neutron spectrum can be modified to reproduce many different spectra of interest including spectra in small thermal test reactors, large pressurized water reactors, and fast reactors. The particular application of this methodology is the design of an experimental approach for using an accelerator to measure the betas produced during fission to be used to reduce uncertainties in the interpretation of reactor antineutrino measurements. This approach involves using a proton accelerator to produce a neutron field representative of a power reactor, and using this neutron field to irradiate fission foils of the primary isotopes contributing to fission in the reactor, creating unstable, neutron rich fission products that subsequently beta decay and emit electron antineutrinos. A major advantage of an accelerator neutron source over a neutron beam from a thermal reactor is that the fast neutrons can be slowed down or tailored to approximate various power reactor spectra. An accelerator based neutron source that can be tailored to match various reactor neutron spectra provides an advantage for control in studying how changes in the neutron spectra affect parameters such as the resulting fission product beta spectrum.

A Feasibility Study of Load-Follow Operation Without Boron Adjustment for CPR1000

This paper presents mathematical modeling of dynamic phenomena in large pressurized water reactors to study load-follow capability. One of the main reactor types in China’s national nuclear development, CPR1000, uses a mode G control method, with G banks, N banks, R banks, and soluble boron to adjust reactor power changes and the axial power shape. In this paper, a new control mode is adopted that can follow the daily variation of power demand without changing the boron concentration. The control banks are regrouped to realize reactivity/temperature control by M banks and axial offset control by an AO bank. A two-node dynamic core model is constructed, taking into account the coupling coefficient and the mutual influence. The transient parameters are obtained by steady-state calculation of a single channel using the original design and operation parameters of CPR1000. Then, to adopt a control mode without soluble boron adjustment, the optimal control implementation is connected to the core simulation platform. Simulation results show that this optimal control policy can provide the capability for the CPR1000 to follow a daily load curve.

A Decoupling Algorithm for Large Pressurized Heavy Water Reactors

This paper presents a decoupling algorithm for a large pressurized heavy water reactor to facilitate the design of a decentralized control system. The reactor models are generally high-order systems, which increases the difficulty of designing control systems. A convenient method of model reduction while maintaining the important dynamic characteristics of the process is through decoupling. The new decoupling algorithm proposed in this paper is used to create a decoupled system for decentralized controller design. To demonstrate the performance of this algorithm, a 72nd-order system was decoupled into three partitions, each containing 20, 27, and 25 states. Both a centralized controller based on the original model and decentralized controllers based on the decoupled model are designed. The advantage of the decentralized controller is shown through a fail-safe study.

Transmutation of High-level Wastes in a Spherical-tokamak Reactor

We studied transmutation of high-level wastes in a spherical-tokamak reactor. We examined spent fuel, recovered fuel, natural UO2 fuel and reprocessed waste, though the natural UO2 fuel is not a high-level waste. We added a few percent of 239Pu to the former three fuels to increase the tritium breeding ratio and saw the change in the effective multiplication factor keff, number of fissions, multiplication of fusion energy and conversion ratio of 239Pu. We found that 2.5 and 5.5 fissions are needed to obtain the tritium breeding ratio of 1 and 2, respectively. High amplification of fusion energy occurs. Breeding of 239Pu is expected for the natural UO2 fuel even in a subcritical environment. For the reprocessed waste, 93Zr and 99Tc are transmuted efficiently by the (n,γ) reaction and so are 237Np and 241Am by the (n,f) reaction. Fusion neutrons of 100MW will transmute half of those nuclides produced from the generation of 1 GW(e) year in a large pressurized water reactor. However, tritium fuel may not be self-sustained in the reprocessed-waste burner.

Severe Accident Related Research and Development at Forschungszentrum Karlsruhe for Present and Future Needs

The research and development program at the Forschungszentrum Karlsruhe, performed within the Program Nuclear Safety Research, is centered around phenomena and processes that could possibly endanger the containment integrity of a large pressurized water reactor after a severe accident. The program includes three activities.The first activity is in-vessel steam explosion. Premixing phenomena are studied in the QUEOS and PREMIX test series. The efficiency of energy conversion is the subject of ECO tests. The BERDA experimental program investigates the load capacity of a reactor pressure vessel (RPV) in steam explosion events.The second activity is hydrogen behavior and mitigation. Advanced models and numerical tools are developed to describe hydrogen sources, distribution of gases in containment, the various modes of hydrogen combustion, and corresponding structural loads.The third activity is ex-vessel melt behavior. The release behavior of melt after RPV failure is studied in DISCO and KAJET tests. In support of core catcher development, interaction with sacrificial and refractory materials, further melt spreading and cooling phenomena are investigated in KAPOOL, KATS, and COMET tests.The goal is to describe and quantify the governing mechanisms and to develop verified models and numerical tools that are able to predict maximum possible loads for severe accident scenarios on full plant scale. The work supported the development and assessment of the safety design of the French-German European Pressurized Water Reactor (EPR). It led to a broader understanding of severe accident phenomena and of controlling and mitigating measures that can also be of benefit for existing plants.

Qualification of a Passive Catalytic Module for Hydrogen Mitigation

The advantage of passive catalytic modules for hydrogen mitigation during core-melt accidents, as compared with active devices like forced-flow recombiners or ignitors, is given by the high reliability of operation and the elimination of potentially violent combustion events. An important step in the qualification of a passive catalytic module system is the determination of the total required capacity and its distribution at various locations in the containment. Experiments and analytic modeling work were performed to qualify the installation of a system of catalytic modules for a large dry pressurized water reactor (PWR) containment. The operational capacity of a prototype catalytic module was determined experimentally, and a corresponding model correlation was developed and integrated into the GOTHIC containment code. This modified code was validated against experimental data. As an application, predictions of the effects, resulting from backfitting a large, dry PWR containment with 50 catalytic modules, were done using the modified code. The catalytic modules keep the hydrogen concentrations below a level of 10% where violent deflagrations could be expected. Local higher concentrations near the release location are inert due to associated low oxygen and high steam concentrations. A proper distribution of the modules in the containment guarantees full mixing ofmore » the atmosphere due to natural convective currents.« less

Severe Accident Containment Loads and Possible Design Concepts of Future Large Pressurized Water Reactors

The risk of present-day light water reactors is dominated by the consequences of core melt accidents followed by a failure of the outer containment. Although such events would have very low frequencies of occurrence, their risk cannot be neglected in the future. Therefore, specifications for mechanical loads and heat loads to the containment are analyzed, and design modifications are proposed, explaining how the containment can withstand the consequences of core melts. As a result, the radiological impact outside of the containment will be drastically decreased. Evacuation of the population outside of the reactor plant will no longer be necessary in the case of a core melt.

Application of the integrated safety assessment methodology to the emergency procedures of a SGTR of a PWR

This paper describes an application of the Integrated Safety Assessment (ISA) methodology to the safety and reliability assessment of emergency procedures of a nuclear power plant. The concept of ISA has been developed as a result of previous works on safety assessment and dynamic reliability. The method links the physical dynamics of the facility with its operating environment, subject to transitions between different time evolutions due to failures and/or system/operator interventions. For situations dominated by deterministic transitions (i.e. transitions upon deterministic demands as a result, for instance, of exceeding automatic actions or alarm setpoints), the methodology can be considered an extension of PSA and accident analysis techniques that replaces the static event tree with a deterministic dynamic event tree concept (DDET) based on the theory of probabilistic dynamics. In line with current studies carried out jointly by CSN and JRC-Ispra/ISEI, this paper reviews the main features of ISA and describes some of the details of its implementation in the case of a Westinghouse pressurized water reactor (PWR), in particular its application to the assessment of the emergency operating procedure (EOP) to mitigate the steam generator tube rupture (SGTR) initiating event. This application demonstrates the ISA feasibility for risk analysis of operating procedures (OP) by assessing a given set of OPs with a large PWR model of the TRETA-DYLAM-HOI software package, which is able to simulate recovery in a SGTR scenario. Some weak points in the SGTR EOP are identified and suggestions provided for their resolution.

Reconstructive Inverse Dynamics Control and Application to Xenon-Induced Power Oscillations in Pressurized Water Reactors

Axial flux shape control in large pressurized water reactors constitutes one of the most challenging control problems in the nuclear field. In commercial plants, the practical solutions are obtained at the expense of departure from the most economical operational conditions, often due to the difficulties in monitoring xenon-induced oscillations and inadequate control actions. In this paper the concept of inverse dynamics in control is introduced as an alternative approach for spatial control. The method is tested through computer simulations using a validated nonlinear model that successfully represents the limit-cycle behavior. Compared with the widely used half-cycling strategy or the proposed optimal control method in the literature, the use of inverse dynamics for partial-length rod control yields desirable stability characteristics. The return to target axial offset exhibits a smooth transition without any residual flux oscillations between the upper and lower halves of the core. The proposed approach consists of a set of nonlinear algebraic equations for control with single-step solutions. Thus, it is easier to implement compared with iterative or integral techniques.

軸方向出力分布偏差のオンラインデータ処理に基づくＰＷＲの軸方向キセノン振動の最適制御法

In a large pressurized water reactor (PWR), xenon oscillations can occur due to the variation of xenon distribution in the reactor. Horizontal and axial modes of oscillation are known. The former is of less concern because of its convergence but the latter can be divergent. Sometimes xenon oscillations are intentionally introduced to obtain recalibration data for the nuclear instrumentation system and in this case oscillations must be eliminated after the data acquisition as soon as possible. Many studies and experiments have been made on control of xenon oscillations, and this problem was solved clearly from the theoretical points of view on the reactor physics and control. However it is found that fine control is required within actual plant constraints to apply the principles and control laws to be indicated from the previous studies. From this point of view a new method has been developed for providing clear operating instructions to the reactor operator.Here, the new method is based on online data processing of axial offsets and the validity of the method has been evaluated through numerical simulations. Then the method has been applied to the experiments in actual PWRs and it has also been proved to be quite effective.

The contribution of international standard problem activities to containment code validation / Der Beitrag internationaler Standardproblemaktivitäten zur Validierung von Rechenprogrammen für Sicherheitsbehälter.

International standard problems have been performed in the field of containment thermohydraulics under the auspices of the OECD and the CEE. More than 25 codes were involved in these activities over the last ten years. The experiments serving as a comparative basis for the analyses cover a wide range of important phenomena expected inside the dry containments of large pressurized water reactors within or beyond the design basis accident concept. The specifications and results of standard problem activities are well documented and may serve future code validation activities. Some of the results are presented here. They show that there are some deviations between pre-test predictions and measured experimental parameters. A few more standard problems in this area are in progress

Seismic Isolation for Advanced Nuclear Power Stations

Seismic isolation offers an attractive approach for reducing seismic loads in nuclear structures, and more significantly, in reactor components. Isolation will lead to a simplification of designs, facilitate standardization, enhance safety margins, and may potentially reduce cost. To date, six large Pressurized Water Reactor units have been isolated in France and South Africa and several advanced nuclear concepts in the U.S., Japan, and Europe have incorporated this approach. It is recognized that to qualify and license an isolation system in the U.S. and in Japan, a comprehensive testing program of isolation components and systems would be required. A major seven year program was initiated in Japan in 1987 with the objective of establishing a qualified seismic isolation design for a large fast breeder reactor to be constructed at the end of this decade. In the U.S., two concepts which use steel laminated elastomeric bearings for seismic isolation have been developed. One of these concepts is a novel system which provides three-dimensional isolation. An extensive test program of scaled prototype bearings to demonstrate their feasibility and effectiveness has been carried out.

Suboptimal level controller for steam generators in pressurized water reactors

The authors consider the design of a suboptimal water-level controller for steam generators in large pressurized water reactors using linear output feedback control. A methodology for feasibility analysis and linear output feedback control design is developed through eigenvalue dynamics. The proposed controller is a linear, constant, partial state feedback law. The controlled system has the following desirable features: (a) the controller is independent of disturbance, (b) the water level returns to its prespecified level value following a step disturbance, (c) the controlled system is asymptotically stable, and (d) transient response is similar to the transient response of an optimal complete state feedback controller. An example illustrates the applicability and the effectiveness of the proposed techniques.< <ETX xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink">&gt;</ETX>