Typical Nuclear Power Plant Research Articles

Critically weighing the costs and benefits of a nuclear renaissance

This article was submitted as a runner up and introduction to JIES 7.3, a special issue on the policy, science and dilemmas of nuclear energy in the 21st century. It provides an overview of some of the key challenges surrounding the so-called ‘nuclear renaissance’. It provides a broad context for the more specific concerns with the social and political aspects of radioactive waste which will be considered in the next issue. What are the likely consequences of a global nuclear power renaissance? This article answers that question by exploring six categories of costs and benefits associated with modern nuclear power plants: capital and production costs, safety and reliability, fuel costs, land degradation, water use, and greenhouse gas emissions. After weighing arguments on both sides, when costs and benefits are normalized across studies and different reactor types, the article finds that the typical nuclear power plant has 8.6 cents of damages attached to every kilowatt-hour of electricity it produces, and the industry as a whole has $223.7 billion worth of net damages every year. These costs are so large (and unavoidable) that in most countries investments in nuclear power do not occur, and they raise doubts as to whether a nuclear renaissance will produce net benefits to society.

Experimental and theoretical study of steam condensation induced water hammer phenomena

We investigate the steam condensation induced water hammer (CIWH) phenomena and present experimental and theoretical results. The experiments were performed in the PMK-2 facility, which is a full-pressure thermo-hydraulic model of the primary loop of the VVER-440/312 type nuclear power plant and located in the Atomic Energy Research Institute Budapest, Hungary. The present experimental setup is capable to measure CIWH phenomena in a wide range of steam pressure, cold water temperature and mass flow rate at a high level of accuracy. On the theoretical side CIWH is studied and analyzed with the WAHA3 model based on two-phase flow six first-order partial differential equations that present one-dimensional, surface averaged mass, momentum and energy balances. A second order accurate high-resolution shock-capturing numerical scheme was applied with different kind of limiters in the numerical calculations. Our study clearly shows that Relap5 and Cathare which are used in the nuclear industry to simulate nuclear power plant accidents cannot resolve the narrow pressure peaks created during a CIWH event. Only WAHA3 can model CIWH properly. Experimentally measured and theoretically calculated pressure peaks are in good agreement, however simulations always show additional pressure peaks. As a new feature in this study we present calculations without additional unphysical reflections caused by boundary conditions.

Simulation of non-condensable gas flow in two-fluid model of APROS – Description of the model, validation and application

In Loviisa VVER-440 type nuclear power plant the nitrogen used to pressurize hydro-accumulators and other passive safety systems is gradually dissolved to the accumulator water during the long period of normal plant operation. If a primary circuit leakage takes place, the accumulator water is injected into the primary circuit, where lower pressure is prevailing and as a consequence the dissolved nitrogen is released from the liquid phase to gas phase. It is also possible that after the liquid has run out of the accumulator the gaseous nitrogen may flow into the primary system and may thus disturb the circulation in the primary circuit. It is important that the system codes that are used in safety analysis work are capable to simulate flows of non-condensable gases and that they can take into account the release of the dissolved gases. In this paper the non-condensable gas model of the APROS two-fluid safety analysis system code is described. The model has been validated using one experiment carried out in the PACTEL VVER-440 test facility, where the release of the nitrogen dissolved in the accumulator water has been studied. The model has been used to analyze the primary–secondary leakage (PRISE) in the Loviisa nuclear power plant. In this leakage incident the dissolved nitrogen from the accumulator was assumed to flow into the primary circuit of the plant.

Low-Power Chirp Spread Spectrum Signals for Wireless Communication Within Nuclear Power Plants

There are two major barriers in deploying wireless communication systems in nuclear power plants (NPPs): (a) the electromagnetic compatibility (EMC) between the wireless devices and the existing plant instrumentation and control systems, and (b) the high levels of electromagnetic noise and interference from high-powered devices and ionizing radiation sources. In a typical NPP there exist strict regulations that limit transmission power levels to avoid interfering with the sensitive safety systems inside the containment such as ion chambers. This will result in performance degradation of wireless communication systems. This paper proposes a wireless communication scheme based on low-power chirp spread spectrum (CSS) signals, which meet with the EMC requirements of NPPs and also are capable of providing interference rejection. The advantage of such a scheme is that satisfactory performance can be obtained using low levels of transmission power. The structure of the optimal receiver for low-power binary CSS signals and a closed-form expression for asymptotic bit error rate of this receiver are derived. The electromagnetic environment within an NPP is modeled as a Gaussian-Gaussian mixture process, which is based on the measurement data published in a U.S. Nuclear Regulatory Commission Regulation (NUREG). The parameters in the model can be adjusted to suit a particular NPP site.

Development of a digital turbine control system in a nuclear power plant

A digital turbine control system (TCS) has been developed for retrofitting an old analog TCS in a nuclear power plant. The developed TCS, which controls the speed of a turbine and the power load of a generator, is based on a triple modular redundant structure to ensure the system reliability. In addition, a turbine simulator has been developed to verify the perfection of the TCS prior to its actual installation. The simulator is composed of a graphic editor, a component model builder, and a system simulation solver. The tested TCS has been successfully applied to a CANDU type nuclear power plant. This paper describes major features of the developed TCS and the turbine simulator including thermal-hydraulic models. Also, the simulation result in a laboratory is compared with the pre startup simulation and the actual operation result.

동적 무한요소를 이용한3차원 지반-구조물 상호작용계의 지진응답해석

본 연구에서는 지반-구조물 상호작용을 고려한3차원 지반-구조계의 지진응답 해석을 수행하고 그 기법의 적용성과 타당성을 검토한다. 이를 위해 구조물과 구조물 주변의 근역지반을3차원 유한요소로서 모델링하고 원역지반에 대해서는 기 개발한 3차원 동적 무한요소를 적용한다. 모든 입사 성분P, SV 그리고 SH파가 고려되었을 때, 등가 지진하중은 무한요소에 의해 구해진 무한 지반의 동적 강성과 자유장 해석을 통해 구해진 지반의 응력과 변위응답을 이용하여 구해진다. 검증 및 적용 예제는 적층 자유장의 지반응답해석과 전형적 원자로 격납건물의 지반-구조물 상호작용을 고려한 층응답 스펙트럼을 구하는 것으로 하였다. 해석 결과는 다른 기법에 의해 구해진 값들과 비교하였으며, 본 기법의 정확성과 정밀성을 확인할 수 있다. This paper presents a seismic analysis technique for a 3D soil-structure interaction system in a frequency domain, based on the finite element formulation incorporating frequency-dependent infinite elements for the far field soil region. Earthquake input motions are regarded as traveling P, SV and SH waves which are incident vertically from the far-field soil region, and then equivalent earthquake forces are calculated using impedances of infinite soil by dynamic infinite elements and traction and displacement from free field response analysis. For verification and application, seismic response analyses are carried out for a multi-layered soil medium without structure and a typical nuclear power plant in consideration of soil-structure interaction. The results are compared with the free field response using a one-dimensional analytic solution, and a dynamic response of an example structure from another SSI package.

Response of proposed passive safety injection system for an intermediate size break LOCA on CHASNUPP-1

A passive safety injection system (PSIS) is proposed for Chashma nuclear power plant-1 (CHASNUPP-1) type nuclear power plants, for the simplification of their safety systems. This system is based upon passive components and is proposed in place of the existing safety injection system, for safety enhancement. The functionality of the proposed system is analyzed using reactor simulation. For this purpose an intermediate size break LOCA is simulated using the simulation software APROS. For this transient, different thermal-hydraulic parameters of the proposed and other safety related systems are presented and discussed. The results obtained show that the proposed system works properly by performing its role in the transient, leading to cold shutdown conditions.

An overview of instrumentation and control systems of a Korea standard nuclear power plant: A signal interface standpoint

This paper presents an overview of instrumentation and control (I&C) systems of a pressurized water reactor (PWR) type nuclear power plant (NPP) in Korea. Yonggwang unit 3, which was constructed as a basis model for a Korea standard nuclear power plant (KSNP), is selected as an example for the presentation. This overview is derived from analyzing the I&C systems based on a top-down approach. The I&C systems consist of 30 systems. The 183 I&C cabinets are also analyzed and mapped to the systems. The overview is focused on an interface between the systems and the cabinets. This information will be used to understand the implementation of the I&C systems and to group the systems for an upgrade.

A practical application of an evaluation model for the restraint effect of pressure-induced bending on a plastic crack opening

This paper presents an evaluation model for the restraint effect of pressure-induced bending (PIB) on the opening of a circumferential through-wall crack (TWC) and a result of its application to the calculation of crack-opening displacement (COD) of postulated cracks for a practical leak-before-break (LBB) analysis. Three-dimensional finite element analyses with different crack lengths, restraint conditions, pipe geometries, magnitudes of internal pressure, and material tensile properties were used to investigate the influence of each parameter on the PIB restraint for the plastic COD. From these investigations, we proposed an evaluation model based on elastic–perfectly plastic behavior. Comparison with finite element analysis results demonstrated that the proposed model reliably estimated the PIB restraint effect on the plastic crack opening of a circumferential TWC and properly reflected the effect of each parameter within the range over which the analytical expression was derived. The model was then used to calculate restrained CODs of postulated cracks for a practical LBB analysis. When plastic crack behavior was considered, the PIB restraint effect was considerable for some LBB analysis cases of the primary piping systems in a typical nuclear power plant. This effect was estimated to be negligible by existing linear elastic-based models.

An application of risk-informed evaluation on MOVs and AOVs for Taiwan BWR-type nuclear power plants

Implementing a risk-informed inservice testing (RI-IST) program provides a good aspect to the nuclear power plant licensee as an alternating program in the current ASME Section XI and 10 CFR 50.55a relevant testing programs. RI-IST concentrates testing resources on highly significant components, reduces excess testing burden, increases plant's availability, decreases dose rate on the plant's staff and also reduces cost on plant's operation and maintenance under nuclear safety expectations. Furthermore, RI-IST also gives a feature on prospective licensing change basis to a nuclear power plant's licensee. This study will focus on safety-related and PRA-molded motor-operated valves (MOVs) and air-operated valves (AOVs) under the inservice testing program in boiling water reactor (BWR)-type nuclear power plant. As MOVs and AOVs have crucial safety functions throughout the nuclear power plant's safety systems, the steady operation and performance of MOVs and AOVs will definitely ensure that the nuclear power plant operates under safety expectations; therefore, this is the key reason to implement risk-informed evaluation for MOVs and AOVs in this study and being able to provide the safety significance classification for MOVs and AOVs under the current IST program to the plant's management. As a pilot study of RI-IST, the methodology of qualitative assessment will incorporate with probabilistic risk assessment (PRA) analyzing MOVs’ and AOVs’ safety significance within the current PRA model. The evaluating result will then classify its safety significance into a high-safety significant component (HSSC) and a low-safety significant component (LSSC) for MOVs and AOVs based on relevant regulatory criteria. With this initiating achievement, it can provide a cornerstone for further studies on the other types of valves and pumps in RI-IST program and also provide a valuable reference as proposing license change to the licensee.

Chemical Tracer Method for Measurement of Flow Rates in Closed Conduits

Thermal output in a nuclear power plant is verified by a calorimetric heat balance on the secondary system of the power plant. The calorimetry involves the precise measurement of the feed water flow rate which should be designed to have ±1:0% of uncertainty. However, the indication of feed water flow rate obtained by a differential pressure measurement across a venturi can be affected by instrument errors, fouling, or a poorly developed velocity profile. These factors can lead to an inaccurate mass flow rate and consequently, an inaccurate estimate of power. The purpose of this study is to develop verification methods with accuracy better than ±0:5% for high-precision flow measurement to be used for measuring feed water flow rate. Such an improvement saves electric power. For a typical Korean nuclear power plant of 1,000MW, 10MW would be potentially saved. This chemical tracer method is a testing process using a tracer, which can be applied to quantify losses in electrical output caused by incorrect measurement of feed water flow rate. This method has a good response to changes in the flow rate. An accuracy better than 0.5% is expected for feed water flow measurement, provided that the feed water system is stabilized during the test.

Methods for a Fire PSA exemplarily applied to a German BWR-69 type nuclear power plant

Abstract In the following, an advanced German approach for state-of-the-art probabilistic fire risk assessment (Fire PSA) to be performed within comprehensive safety reviews of nuclear power plants is presented on the background of the current German PSA Guide. This approach has been successfully applied in the frame of a Fire PSA recently performed for a nuclear power plant with boiling water reactor (BWR-69 type). The set of all compartments in a nuclear power plant is the starting point for determining the annual frequency of fire induced plant hazard and core damage states. Fire PSA has to be carried out in several steps. First step of the analysis is a selection process (‘screening’) providing critical fire zones, where a fully developed fire has the potential to both cause an initiating event and impair the function of at least one nuclear safety related component or system. Detailed analyses providing compartment specific fire induced hazard state frequencies are performed only for those compartments, where in case of fire a relevant contribution to the overall hazard state frequency can be expected. Both analytical steps, selection of relevant compartments as well as detailed analyses, are based on a comprehensive compartment related compilation of fire specific data and information.

Chemical Tracer Method for Measurement of Flow Rates in Closed Conduits

Thermal output in a nuclear power plant is verified by a calorimetric heat balance on the secondary system of the power plant. The calorimetry involves the precise measurement of the feed water flow rate which should be designed to have ±1:0% of uncertainty. However, the indication of feed water flow rate obtained by a differential pressure measurement across a venturi can be affected by instrument errors, fouling, or a poorly developed velocity profile. These factors can lead to an inaccurate mass flow rate and consequently, an inaccurate estimate of power. The purpose of this study is to develop verification methods with accuracy better than ±0:5% for high-precision flow measurement to be used for measuring feed water flow rate. Such an improvement saves electric power. For a typical Korean nuclear power plant of 1,000MW, 10MW would be potentially saved. This chemical tracer method is a testing process using a tracer, which can be applied to quantify losses in electrical output caused by incorrect measurement of feed water flow rate. This method has a good response to changes in the flow rate. An accuracy better than 0.5% is expected for feed water flow measurement, provided that the feed water system is stabilized during the test.

Application of artificial neural networks to nuclear power plant transient diagnosis

A study on various artificial neural network (ANN) algorithms for selecting a best suitable algorithm for diagnosing the transients of a typical nuclear power plant (NPP) is presented. NPP experiences a number of transients during its operations. These transients may be due to equipment failure, malfunctioning of process systems, etc. In case of any undesired plant condition generally known as initiating event (IE), the operator has to carry out diagnostic and corrective actions. The objective of this study is to develop a neural network based framework that will assist the operator to identify such initiating events quickly and to take corrective actions. Optimization study on several neural network algorithms has been carried out. These algorithms have been trained and tested for several initiating events of a typical nuclear power plant. The study shows that the resilient-back propagation algorithm is best suitable for this application. This algorithm has been adopted in the development of operator support system. The performance of ANN for several IEs is also presented.

Development of accumulator computational aid for determining RCS injection volume

Abstract Reactor coolant system (RCS) injection using accumulator is an important strategy for both emergency operating procedure (EOP) and severe accident management guideline (SAMG) of pressurized water reactor (PWR) nuclear power plant. Once accumulator injection starts, the operator is requested to close the accumulator isolation valve to avoid nitrogen gas flow into RCS as the water level is low. Current accumulator water level indication system is not designed for this purpose. In emergency operating procedure, it relies on the steam generator pressure to close the accumulator isolation valve. The purpose of this paper is to develop a computational aid for estimating RCS injection volume of accumulator. First of all, simple accumulator model is verified using the plant data during a station blackout incident of Maanshan nuclear power plant. An isentropic expansion model is found better than adiabatic expansion model. Then, a computational aid is developed based on this model. Using this computational aid, the accumulator water level can be judged directly from the accumulator pressure. This computational aid can be applied for typical PWR nuclear power plants in both emergency operating procedure and severe accident management guideline.

05/02494 Adaptive control of a PWR core power using neural networks

This work proposes a new hybrid control strategy for a pressurized water type nuclear power plant by integrating linear quadratic integrator (LQI), linear quadratic Gaussian (LQG), and loop transfer recovery (LTR) approaches. The multi-input multi-output nuclear plant model adopted in this work is characterized by 38 state variables. The nonlinear plant model is linearized around steady-state operating conditions to obtain a linear model for the controller design. The proposed LQGI/LTR technique designs state-feedback assisted output control using the estimated states. The control architecture offers robust performance and tracks the reference set-point with zero steady-state error in the presence of uncertainties and disturbances. The effectiveness of the proposed technique is demonstrated by simulations on different subsections of a pressurized water nonlinear nuclear power plant model. The control performance of the proposed technique is further compared with other classical control design schemes. Statistical measures are employed to quantitatively analyse control performance.

Application of finite element model updating to a feed water pipeline of a nuclear power plant

This article is concerned with application of finite element model updating to a part of a secondary feed water pipeline of a VVER type nuclear power plant. A mathematical model of the pipeline was produced with finite element method. The pipeline was subjected to modal testing in three different conditions. The results of the modal testing were then compared with those predicted with the model. This comparison revealed discrepancies between these two sets of results. It is a commonly known fact that incorrect values of parameters in a model, among other things, do cause these discrepancies. In order to improve the correlation between the measured and the predicted results and hence improve the reliability of the model, the model was subjected to finite element model updating. In this updating process, incorrect values of parameters in a model are adjusted, so that a desired level of correlation between measured and predicted results is achieved. As a result, the correlation between the measured and the predicted results was improved. The bulk of the improvement was achieved by adjusting the values of those parameters that were associated with boundary conditions of the model.

Investigation of Containment Flooding Strategy for Mark-III Nuclear Power Plant with MAAP4

Containment flooding is an important strategy for severe accident management of a conventional boiling water reactor (BWR) system. The purpose of this work is to investigate the containment flooding strategy of the Mark-III system after a reactor pressure vessel (RPV) breach. The Kuosheng Power Plant is a typical BWR-6 nuclear power plant (NPP) with Mark-III containment. The Severe Accident Management Guideline (SAMG) of the Kuosheng NPP has been developed based on the BWR Owners Group (BWROG) Emergency Procedure and Severe Accident Guidelines, Rev. 2. Therefore, the Kuosheng NPP is selected as the plant for study, and the MAAP4 code is chosen as the tool for analysis. A postulated specific station blackout sequence for the Kuosheng NPP is cited as a reference case for this analysis. Because of the design features of Mark-III containment, the debris in the reactor cavity may not be submerged after an RPV breach when one follows the containment flooding strategy as suggested in the BWROG generic guideline, and the containment integrity could be challenged eventually. A more specific containment flooding strategy with drywell venting after an RPV breach is investigated, and a more stable plant condition is achieved with this strategy. Accordingly, the containment flooding strategy after an RPV breach will be modified for the Kuosheng SAMG, and these results are applicable to typical Mark-III plants with drywell vent path.

Validation of coupled codes using VVER plant measurements

A data set of five transients at different VVER type nuclear power plants was collected in order to validate neutron kinetics/thermal hydraulics codes. Two of these transients ‘drop of control rod at nominal power at Bohunice-3’ of VVER-440 type and ‘coast-down of 1 from 3 working MCPs at Kozloduy-6’ of VVER-1000 type, were then utilised for code validation. Eight institutes contributed to the validation with 10 calculations using 5 different combinations of coupled codes. The thermal hydraulic codes were ATHLET, SMABRE and RELAP5 and the neutron kinetic codes DYN3D, HEXTRAN, KIKO3D and BIPR8. The general behaviour of both the transients was quite well calculated with all the codes. Even an elementary modelling of coolant mixing in reactor pressure vessel under asymmetric transients improved correspondence to the measurements. Some differences between the calculations seem to indicate that fuel modelling and treatment of VVER-440 control rods need further consideration. The simultaneous validation interacted with the data collection effort and thus improved its quality. The complexity of data collection systems and sometimes conflicting data, however, called for compromises and interpretation guides that also taught the analysts balanced plant modelling.

Applications of Thermoelectric Power Measurement to Deterioration Diagnosis of Nuclear Material and Its Principle

In order to non-destructively evaluate changes in toughness of cast duplex stainless steel, which is frequently used in main coolant pipes of PWR type nuclear power plants, due to thermal aging, we tried to apply thermoelectric power (TEP) measurement. because TEP is sensitive to microstructural material changes, and to clarify the mechanism behind TEP changes due to thermal aging. As a result, TEP of cast duplex stainless steel increased with aging time, and good correlations were found between TEP and toughness. Concerning the mechanism, TEP of high and low Cr content alloys was higher than that of intermediate Cr content alloys. Because high and low Cr areas are created in the ferrite phase due to thermal aging. TEP of the entire material increased. Furthermore, when each Cr fluctuating area acted in parallel, the increase in TEP became larger. According to the Mott-Jones theory, TEP is largely related to the electron density of states at the Fermi level. The electron density of states of Fe–Cr–Ni alloys in the valence band was measured with X-ray photoelectron spectroscopy (XPS). As a result, there was a high correlation between the TEP calculated from the XPS spectrum and the measured TEP. Therefore, we showed experimentally that the electron density of states changed due to variations in Cr concentration, which also affect TEP.