Light Water Reactor Nuclear Power Research Articles

Nondestructive Characterization and Monitoring of Cracking in Pressurized Piping Using Surface Strain Measurements

Primary loop piping integrity is a regulatory and safety concern for the long-term operation (LTO) commercial light water reactor (LWR) nuclear power plants (NPPs). Typically, primary loop piping in existing LWRs was manufactured from cast or wrought austenitic stainless steel alloy material and operates at a pressure of ~2250 psi (16MPa) for a pressurized water reactor (PWR) design type and ~1000 psi (7 MPa) for a boiling water reactor (BWR). The objective of this research was to determine if measurement of surface strains can be used to characterize inner diameter flaws. For this feasibility study, wrought carbon steel Alloy A106 Grade B material was selected in lieu of stainless steel. This research was performed on pre-notched, 10 in. (25 cm) nominal outer diameter (OD) Schedule 120 (0.844 in. [21.4 mm) wall thickness) pipe samples of 24 in. (61 cm) length. The notches were made using electrical discharge machining (EDM) on the inner diameter (ID) of the pipe samples, with notch features two-dimensional (2D) and planar in nature to simulate cracking. The surface strains in the samples were evaluated under increasing hydrostatic pressure, up to a maximum of 2300 psi (16MPa) to determine their response to flaws with variable depth. Reusable end caps were utilized in lieu of welded caps for the pressure testing. This article describes the test procedure for the pressure testing and results of the feasibility study, which is the first phase of research with the end goal of the development of online crack monitoring capability for use at operating NPPs.

A Possibility for Large-Scale Production of 238Pu in Light-Water Reactor VVER-1000

This paper considers the possibility for large-scale production of plutonium isotope 238Pu in the light-water nuclear power reactor VVER-1000. 238Pu is a unique source of long-term autonomous energy supply in various devices for remote regions of the Earth and in outer space. The design of the irradiation device with 237NpO2 as a starting material is proposed, which is placed in the central zone of the VVER-1000 reactor core and makes it possible to achieve 8% of the specific Pu production (Pu/237Np) by optimizing the pitch of NpO2-rod lattice. The computations showed that the scale of 238Pu production in the irradiation device was remarkably larger (2 ÷ 7 times more) than similar values in research reactors. At the same time, the use of heavy neutron moderators with low neutron absorption (natural lead or lead isotope 208Pb) around the NpO2 fuel assembly (FA) made it possible to obtain high-purity 238Pu with the content of 236Pu below 2 ppm. The paper also shows that if the irradiation device is placed in central zone of the VVER-1000 reactor core, then the displacement damage dose in the reactor vessel remains low enough to conserve its strength properties throughout the entire period of the reactor operation (60 years).

A Numerical Simulation of Radiation Chemistry for Controlling the Oxidising Environment in Water-Cooled Nuclear Power Reactors

Maintaining the integrity of materials of light-water nuclear power reactors requires the development of effective methods to control and minimise the corrosive environment associated with the radiolysis of a coolant. In this study, the behaviour of the oxidising environment is simulated using a hybrid method. The hybrid method has advantages in that the production of radiolytic species under exposure of the coolant to ionising radiation is simulated while providing material and charge balances. Steady-state concentrations of stable and transient oxidising agents are calculated as a function of radiation composition and dose rate by numerical integration of the system of kinetic equations describing radiation chemistry of neutral water, the alkaline solution, and the hydrogenated systems at 300 °C. The importance of the reactions and equilibria constituting the radiolysis scheme of the coolant is assessed. The influence of the presence of a base and the injected H2 on the yield of key reactions responsible for the formation of the main oxidants H2O2 and O2 are discussed. Simulation indicated the synergic effect of H2 gas and base added to the coolant on diminishment of the steady-state concentration of oxidants.

Numerical analysis for FP speciation in VERDON-2 experiment: Chemical re-vaporization of iodine in air ingress condition

In the late phase of severe accident in light water reactor nuclear power station, re-mobilization of fission products (FPs) has a significant impact on the source term because most portion of FPs is retained in reactor coolant system and/or containment vessel. Recently, VERDON-2 experiment showed noticeable re-vaporization, which was one of the re-mobilization phenomena, of iodine under air ingress condition, but this mechanism has not been identified yet. The present study numerically investigated the FPs behaviors in VERDON-2 experiment with the mechanistic FPs transport analysis code incorporating thermodynamic chemical equilibrium model in order to further understand nature for FPs behavior, especially iodine re-vaporization under air ingress condition. Consequently, this analysis reproduced the deposition profile of cesium, one of important FPs in the source term, along the thermal gradient tube (TGT) in the experiment, which revealed that cesium was transported as CsOH in early phase of FP release from fuel, and then formed Cs2MoO4 and Cs2Te after the release of molybdenum and tellurium was activated. Regarding iodine as another important FP, formation of CsI was predicted in steam condition. The CsI was transported and partly deposited and condensed onto the TGTs and other components of the VERDON facility. Under the air ingress condition, the present analysis showed the agreement for iodine re-vaporization in the experiment and revealed its mechanism; the deposits of iodide were chemical re-vaporized as molecular iodine (I2) gas by redox reaction with competitive elements such as molybdenum, chromium and tellurium.

External surface temperature measurements for the heat transfer analysis of internally heated cylindrical clad-tubes subjected to external forced convection bulk water coolant thermal-hydraulic conditions

In a worldwide pursuit for more Accident Tolerant nuclear Fuel (ATF), the quest to obtain and certify alternative nuclear fuel cladding tubes for light-water nuclear power reactors is still a key challenge. One of the facets in this program to develop more ATF is the heat transfer evaluation between the various proposed clad tubes manufactured from suitable replacement materials and the current zirconium-alloy based clad tubes used in nuclear power reactors. For the heat transfer analysis, the accurate measurement of the temperature on the heat transfer surface of heated tubes to be tested was one of the important objectives for the effective analysis of the heat transfer characteristics to the water coolant. After extensive investigations, a suitable technique was developed and validated against recognised forced convection heat transfer correlations. The results showed that this technique was well suited for external forced convection heat transfer studies from heated surfaces exposed to forced convection water coolants.

Light water reactor nuclear power plant commodity cost driver analysis

Light water reactor nuclear power plant commodity cost driver analysis

Light water reactor nuclear power plant commodity cost driver analysis

Construction costs constitute the largest outlay of a nuclear power plant. Direct costs, such as the cost of materials and installation of materials and equipment, make up the majority of construction expenditures. The purpose of this study is to quantify the cost of the two most abundant materials: concrete and steel. This study shows that a 1% reduction in concrete installation rate results in a 0.6% reduction in concrete costs, and a 1% reduction in the steel installation rate creates a 0.3% reduction in steel costs. The additional cost paid for nuclear-quality concrete is shown to be 23% of total concrete budget. The additional cost paid for nuclear-quality steel is shown to be 41% of the total steel expenditures. These results show that focus on decreasing commodity expenditures (through increasing installation rate) can improve the cost of constructing nuclear power plants.

Irradiation-assisted stress corrosion cracking susceptibility and mechanical properties related to irradiation-induced microstructures of 304L austenitic stainless steel

304L SSs are used for structural components of Light Water Reactor (LWR) nuclear power plants and have been shown to face Irradiation Assisted Stress Corrosion Cracking (IASCC). This degradation phenomenon has been of great concern regarding structural integrity of reactors for lifetime extension of LWRs but is still not completely understood. This study evaluates the evolution of mechanical properties (hardening) and susceptibility to stress corrosion cracking regarding different radiation induced microstructures. These microstructures were chosen to explore representative conditions encountered in Pressure Water Reactor (PWRs) conditions but also the effect of various populations of cavities as extreme case of those reported at higher flux and temperature. Irradiation microstructures are produced by heavy ions irradiation under two different temperatures (450 °C and 600 °C) and with or without helium implantation. The evolution of microstructure, hardness and cracking susceptibility are characterized after irradiation. Calculations based on defects population are found to be in agreement with hardening measurement for irradiations with no helium implantation only. Cracking susceptibilities are characterized after slow strain rate tests performed at 4% plastic strain under PWR environment. Cracking susceptibility changes for different irradiations and quantitative assessment of the effect of irradiation microstructure on IASCC phenomenon are discussed.

Modular system for probabilistic fracture mechanics analysis of embrittled reactor pressure vessels in the Grizzly code

Abstract In light water reactor (LWR) nuclear power plants, the reactor pressure vessel (RPV) plays an essential safety role, and its integrity must be ensured during a variety of transient loading conditions. These can include off-normal conditions such as a pressurized thermal shock (PTS), as well as transients encountered during normal startup, shutdown, and testing of the reactor. Exposure to irradiation and elevated temperatures embrittles the RPV’s steel over time, making it increasingly susceptible to failure due to propagation of fractures that could initiate at the locations of flaws introduced during the manufacturing process. As long-term operation scenarios are being considered for LWRs in the United States, it is important to have a flexible simulation tool that can be used to perform probabilistic evaluations of RPV integrity under a wide variety of conditions and incorporate improved predictive models of RPV steel embrittlement. The Grizzly code is being developed to meet these needs. This paper describes Grizzly’s modular architecture, provides results of benchmarking studies of various components of Grizzly, and demonstrates the application of Grizzly on a model that includes plume effects that are difficult to represent in other codes being used in current practice.

Numerical analysis of different break direction effect on hydrogen behavior in containment during a hypothetical LOCA

Large amounts of hydrogen could be generated and released into the containment during the course of a severe accident in a Light Water Reactor nuclear power plant. Getting detail hydrogen distribution in containment is important for analysis of hydrogen risk. In this study, a three dimensional CFD analysis model is established to analyze hydrogen behavior during hypothesized severe accident induced by LOCA in advanced passive PWR containment. Hydrogen distribution in containment simulating with algebraic and k-ε turbulence models is studied, the result shows that when k-ε model is adopted, hydrogen stratification is established in whole upper space and the peak concentration in the dome is 8%. Then the effect of source injection direction with vertical and horizontal injection cases is investigated. In vertical injection case, the local high concentration of hydrogen reaches over 25% in source compartment. In horizontal injection case, it occurs in SG-2 and the channel connecting two steam generator compartment i.e. vertical compartment. In upper space, comparing with horizontal source direction, when source direction is vertically towards upper space, hydrogen flows into the dome driven by momentum and stratification can be established more possible. In long term, with hydrogen distribution tending to be uniform, the effect of direction is not obvious. Due to more steam transporting into local compartments (SG-2, vertical and CMT compartments) when direction is horizontal, hydrogen concentration in these compartment is lower. In addition, source direction change can also lead to the change of location where hydrogen combustion risk probably occurs. When direction is horizontal, combustion risk happens in SG-2, vertical or CMT compartments rather than source compartment.

Lessons Learned in Protection of the Public for the Accident at the Fukushima Daiichi Nuclear Power Plant.

What insights can the accident at the Fukushima Daiichi nuclear power plant provide in the reality of decision making on actions to protect the public during a severe reactor and spent fuel pool emergency? In order to answer this question, and with the goal of limiting the consequences of any future emergencies at a nuclear power plant due to severe conditions, this paper presents the main actions taken in response to the emergency in the form of a timeline. The focus of this paper is those insights concerning the progression of an accident due to severe conditions at a light water reactor nuclear power plant that must be understood in order to protect the public.

Investigation of Fe-1.0% Cu surrogate specimens with nonlinear ultrasound

This research explores the sensitivity of nonlinear ultrasound (NLU) to the changes in microstructure of heat treated Fe-1.0at.% Cu binary alloy. Results from experimental measurements and a theoretical model each show that the acoustic nonlinearity parameter, β increases with increasing Cu-precipitate radius. The iron-copper material investigated in this research serves as a surrogate material to simulate the evolution of radiation damage as a result of copper nanocluster precipitation that occurs in reactor pressure vessels (RPV) of light water nuclear power reactors. With RPVs seeing more neutron cycles than originally anticipated, it is imperative to develop nondestructive evaluation (NDE) techniques capable of evaluating the integrity of these structures.

Online Monitoring and Prognostics for Passive Components in Nuclear Power Plants

This paper describes research toward developing prognostics technologies for light water nuclear power reactor components. The focus of this paper is on passive components (those that do not need to change state or move to perform their function), although the technologies are applicable to other classes of components as well. A prototypic failure mechanism (high-cycle fatigue) is used to focus the efforts and provide context for the development effort. A Bayesian framework is proposed for the prognostics of remaining useful life and applied to simulated data sets representing nondestructive measurements of high-cycle fatigue damage. The initial results of the prognostics based on simulated data sets are presented.

Combined numerical and experimental study of temperature pulsations in the fragment of header unit of heat exchanger of nuclear power unit clean-up and cooldown system

Experimental model of fragment of header unit of light-water nuclear power reactor clean-up and cooldown system was developed and manufactured. Experimental studies of temperature conditions were performed using the developed experimental model.Experimental distributions of temperature in characteristic zones of the header unit under study were obtained. The most thermally stressed zones of heat-exchanging surface were determined. Analysis of intensity of temperature pulsations on the heat-exchanging surface and coolant flow in different zones was performed, statistical and spectral characteristics of temperature pulsations were represented. Solutions were suggested aimed at the reduction of intensity of thermal pulsations.Calculation model of the fragment of header unit was developed and recommendations were given on the development of calculation models. Results of numerical modeling of transient temperature conditions and characteristics of temperature pulsations for different regimes of flow streamlining the model obtained using ANSYS CFX 14.0 CFD-code are presented here.Comparative analysis of experimental and calculated data was performed. It was demonstrated that calculated data are in agreement with experimental data with sufficient accuracy which gives the possibility to use the developed calculation model in the future for subsequent substantiation of heat exchanger design.

A REVIEW ON DEVELOPING INDUSTRIAL STANDARDS TO INTRODUCE DIGITAL COMPUTER APPLICATION FOR NUCLEAR I&C AND HMIT IN JAPAN

A comprehensive review on the technical standards about human factors (HF) design and software reliability maintenance for digital instrumentation and control (I&C) and human-machine interface technology (HMIT) in Japanese light water reactor nuclear power plants (NPPs) was given in this paper mainly by introducing the relevant activities at the Japan Electric Association to set up many industrial standards within the traditional framework of nuclear safety regulation in Japan. In Japan, the Fukushima Daiichi accident that occurred on March 11, 2011 has great impact on nuclear regulation and nuclear industries where concerns by the general public about safety have heightened significantly. However for the part of HF design and software reliability maintenance of digital I&C and HMIT for NPP, the author believes that the past practice of Japanese activities with the related technical standards can be successfully inherited in the future, by reinforcing the technical preparedness for the prevention and mitigation against any types of severe accident occurrence.

Large hybrid energy systems for making low CO2 load-following power and synthetic fuel

Hybrid energy systems using nuclear heat sources can economically produce load-following electrical power by exploiting their surplus generation capacity, available at night or seasonally, to make synthetic fuel. Vehicle fuel is the only current energy use large enough to absorb all the energy capacity that might be diverted from the power industry, and its ease of storage obviates problems with discontinuous synfuel production. The potential benefits and challenges of synfuels integration are illustrated by the production of methanol from natural gas (as a source of carbon) using steam from a light water nuclear power reactor which is assumed to be available in accord with a year's worth of power demand data. The synthesis of methanol is easily adapted to using 300 °C heat from a light water reactor and this simple compound can be further processed into gasoline, biodiesel (to esterify vegetable oils), or dimethyl ether, fuels which can be used with the current vehicle fleet. A supplemental feed to the methanol process of natural gas (for energy) allows operation at constant full rate when the nuclear heat is being used to produce electrical power. The higher capital costs of such a system are offset by a lower cost of heat and power production from a large base load type of plant and by reduced costs associated with much lower CO2 emissions. Other less tangible economic benefits of this and similar hybrid systems include better use of natural resources for fuels and greater energy supply reliability from the domestic production of vehicle fuel.

CORROSION FATIGUE OF NUCLEAR--GRADE STAINLESS STEEL IN HIGH TEMPERATURE WATER AND ITS ENVIRONMENTAL FATIGUE DESIGN MODEL

The high safety of light water reactor nuclear power plants (NPPs) requires very strict design standard and service property of pressure boundary components materials. The service degradation and life assessment of the components materials primarily depend on the understanding of environmentally assisted failure mechanism, accumulation of service property data and construction of evaluation models. Currently domestic NPPs are relying on foreign design, operation and life assess- ment standards. However, recent experimental data indicate that even the ASME design fatigue code may be deficient in safety margin under certain conditions of loading and environment. In the present work, based on the corrosion fatigue tests in simulated NPPs' high temperature pressurized water, the corrosion fatigue behavior and environmentally assisted failure mechanism of domestic nuclear-grade stainless steel have been investigated. The factors affecting fatigue life of nuclear grade stainless steel in high temperature water were evaluated. A design fatigue model was constructed by taking envi- ronmental degradation effects into account and the corresponding design curves were given for the convenience of engineering applications.

Challenges in coupled thermal–hydraulics and neutronics simulations for LWR safety analysis

The simulation of nuclear power plant accident conditions requires three-dimensional (3D) modeling of the reactor core to ensure a realistic description of physical phenomena. The operational flexibility of Light Water Reactor (LWR) plants can be improved by utilizing accurate 3D coupled neutronics/thermal–hydraulics calculations for safety margins evaluations. There are certain requirements to the coupling of thermal–hydraulic system codes and neutron-kinetics codes that ought to be considered. The objective of these requirements is to provide accurate solutions in a reasonable amount of CPU time in coupled simulations of detailed operational transient and accident scenarios. These requirements are met by the development and implementation of six basic components of the coupling methodologies: ways of coupling (internal or external coupling); coupling approach (integration algorithm or parallel processing); spatial mesh overlays; coupled time-step algorithms; coupling numerics (explicit, semi-implicit and implicit schemes); and coupled convergence schemes. These principles of the coupled simulations are discussed in details along with the scientific issues associated with the development of appropriate neutron cross-section libraries for coupled code transient modeling. The current trends in LWR nuclear power generation and regulation as well as the design of next generation LWR reactor concepts along with the continuing computer technology progress stimulate further development of these coupled code systems. These efforts have been focused towards extending the analysis capabilities as well as refining the scale and level of detail of the coupling. This article analyses the coupled phenomena and modeling challenges on both global (assembly-wise) and local (pin-wise) levels. The issues related to the consistent qualification of coupled code systems as well as their application to different types of LWR transients are presented. Finally, the advances in numerical and computation techniques for coupled code simulations are summarized with outlining remaining challenges.

A Light-Water-Cooled Nuclear Reactor of Natural Uranium and Beryllium

Abstract A light-water nuclear power reactor core configuration is described, which does not require any isotopic enrichments (e.g., D2O or U-235) without the use of graphite. This reactor design utilizes plates of natural uranium (NU) metal and Beryllium. The coolant is natural water, which flows between two plates of NU and is required to add additional moderator to attain criticality. Only square reactor shapes of uniform plate thickness are evaluated. Commercial viability is not demonstrated but the proposed design does warrant further study. The configuration shows potential to meet GEN IV sustainability and GNEP closed fuel cycle goals. Aspects covered include neutronics, criticality, costs and nonproliferation.

A MELCOR Application to Two Light Water Reactor Nuclear Power Plant Core Melt Scenarios with Assumed Cavity Flooding Action

The MELCOR 1.8.4 code Bottom Head package has been applied to simulate two reactor cavity flooding scenarios for when the corium material relocates to the lower-plenum region in postulated severe accidents. The applications were preceded by a review of two main physical models, which highly impacted the results. A model comparison to available bibliography models was done, which allowed some code modifications on selected default assumptions to be undertaken. First, the corium convective heat transfer to the wall when it becomes liquid was modified, and second, the default nucleate boiling regime curve in a submerged hemisphere was replaced by a new curve (and, to a much lesser extent, the critical heat flux curve was slightly varied).The applications were devoted to two prototypical light water reactor nuclear power plants, a 2700-MW(thermal) pressurized water reactor (PWR) and a 1381-MW(thermal) boiling water reactor (BWR). The main conclusions of the cavity flooding simulations were that the PWR lower-head survivability is extended although it is clearly not guaranteed, while in the BWR sequence the corium seems to be successfully arrested in the lower plenum.Three applications of the CFX 4.4 computational fluid dynamics code were carried out in the context of the BWR scenario to support the first modification of the aforementioned two scenarios for MELCOR.Finally, in the same BWR context, a statistic predictor of selected output parameters as a function of input parameters is presented, which provides reasonable results when compared to MELCOR full calculations in much shorter CPU processing times.