Long-term aging of light water reactor concrete containments

This paper evaluates the aging of light water reactor concrete containments and identifies three degradation mechanisms that have the potential to cause widespread aging damage after years of satisfactory experience: alkali–silica reactions; corrosion of reinforcing steel, steel liner, and prestressing steel; and sulfate attack. The aging evaluation is based on a comprehensive review of the relevant technical literature. Low-alkali cement and slow-reacting aggregates selected according to ASTM requirements cause deleterious alkali–silica reactions. Low concentrations of chloride ions can initiate corrosion of the reinforcing steel if the hydroxyl ions are sufficiently reduced by carbonation, leaching or magnesium sulfate attack. Magnesium sulfate attack on concrete can also cause loss of strength and degradation of cementitious properties of the containment concrete after long-term exposure. The techniques for inspecting, mitigating and repairing these long-term aging effects are discussed.     

Application of steam injector to improved safety of light water reactors

Steam injector (SI) is a simply designed passive jet pump which does not require external power source or internal mechanical parts. The SI utilizes direct contact condensation between steam and water as an operational mechanism and is capable of producing higher pressure water than the inlet fluid pressures. The accident in Fukushima Daiichi Nuclear Power Plant caused setback to the credibility and reliability of nuclear power. One way to regain its trust from the global community, it is suggested to develop and install passive coolant injection systems that are operable even during the station black out. In this review paper, thorough and complete review of the SI system was completed and applicability of the SI system as the passive core cooling system is discussed in details. Due to its high heat removal capability, the system can possibly be applied as a high efficiency heat exchanger as well. Its design and operational mechanisms, and fundamental thermal-hydraulic theory utilized in the analysis and experimental work are reviewed. In addition, its possible application towards existing nuclear power plant systems is reviewed.     

Selected safety aspects of containments for nuclear power plants with WWER-440 reactors

Considerable attention has been and continues to be focused on the design and operational features that prevent the release of radioactive materials to the environment for a spectrum of accidents for the two classes of WWER-440 reactors: the older 230 model and the more recently designed 213 models.This paper, based on published and unpublished information, aims to clarify the perceptions of the Russian WWER-440 models 230 and 213 Nuclear Power Plant containment system designs and their relevance to selected aspects of accident mitigation. It should be noted that these are unclearly and often negatively perceived, primarily because of a lack of reliable information and a poorly assembled experimental database. Conflicting statements have been made regarding the nature and the features of the plant's containment system. The paper presents a brief outline of the design of both WWER-440 models with respect to their confinement functions. Selected safety-related aspects of the accident localization systems are discussed, and the recognized shortcomings and safety merits are pointed out. The older 230 units experience high leak rates and are designed to withstand medium-size pipe breaks. The possible implications for safety are pointed out in the paper. The on going studies that concentrate on improving the system are highlighted. Some of the proposed modifications of the system, which would significantly decrease risks associated with accidents that are beyond the original design basis, are discussed. The design of the newer 213 model differs in many aspects. It incorporates the simple and original application of passive natural processes to limit the large-break loss-of-coolant accident post accident pressure. Other features of this containment design, such as complicated geometry, dependence on several mechanical devices and interlocks, and insufficient experimental evidence, lead to doubts concerning the operation of this containment under accident conditions. For the newer 213 model, current work is devoted mainly to safety assessment and verification of the containment design. Some information concerning the on-going work is provided in the paper.     

Spherical steel containments of pressurized water reactors under accident conditions: Investigation program and first results

The investigations will deal with the mechanical behavior of a free standing spherical containment shell built for the latest type of a German pressurized water reactor. The diameter of the containment shell is 56 m. The wall thickness is 38 mm. The material used is the ferritic steel 15MnNi63.The investigation program includes theoretical as well as experimental activities and concerns four different accidents which are beyond the scope of the common design and licensing practice: containment behavior under quasi-static pressure increase up to containment failure; containment behavior under high transient pressures; containment vibrations due to earthquake loadings (consideration of shell imperfections); containment buckling due to earthquake loadings. First results concerning the containment behavior under quasi-static pressure increase are presented. It turns out that the mechanical failure of the containment shell is controlled by plastic instability. A computer program to describe this problem has been developed and membrane tests to check the computational methods have been carried out.     

US NRC research on the integrity of piping in light water reactors

The paper describes the activities underway in NRC on the subject of LWR piping integrity as of the summer and fall of 1983. The paper is necessarily vague on certain topics of policy because they are either under review or are under development. Particularly in the area of BWR pipe cracking, events are very rapid so that positions and actions described in this paper may well be obsolete by the time it is published. Nevertheless, this paper is useful to show the intentions of NRC in the area of research for LWR piping, and it is also useful to document the status of the regulations on piping for which the research is being performed.     

On the structural reliability of reactor safety containments

A method for reliability based design of reactor safety containments is suggested, and a brief review of classical reliability analysis is presented. Seismic and climatic load occurrences are modeled by uniform Poisson processes. Extreme value distributions are assumed to represent the seismic, climatic, external and internal pressure load intensities. Reliabilities are calculated for various design loads and load combinations.     

Overview of containment venting as an accident mitigation strategy in US light water reactors

This paper summarizes the results of previous analyses of containment venting at US light water reactors. The focus of the paper is on the risk aspects of containment venting as a severe accident mitigation strategy; therefore, past risk analyses of venting are critically reviewed and conclusions are drawn where possible concerning the risk and efficacy of this strategy. Because the accident mitigation issues vary from one reactor and containment type to another, the paper examines five containment types separately.     

Plutonium breeding of light water cooled fast reactors

Light water cooled fast reactor with new fuel assemblies (FA) has been studied for high breeding of fissile plutonium. It achieves fissile plutonium surviving ratio (FPSR) of 1.342 (discharge/loading), 1.013 end and beginning of equilibrium cycle (EOEC/BOEC), and compound system doubling time (CSDT) of 95.9 years at the average coolant density of pressurized water reactor (PWR). It is further improved for reduced moderation boiling water reactor (BWR) (RMWR) coolant density. Fissile plutonium surviving ratio reaches 1.397 (discharge/loading), 1.030 (EOEC/BOEC) and CSDT is 37 years. The present study has shown the possibility of breeding at the PWR coolant density and meeting the growth rate of energy demand of advanced countries at the RMWR and Super FR coolant density for the first time. The new FA consist of closely packed fuel rods. The integrity of welding of fuel rods at the top and bottom ends is maintained as the conventional fuel rods. The coolant to fuel volume fraction is reduced to 0.085, one-sixth of that of RMWR. The volume fraction remains unchanged with the diameter of the fuel rod. The thermal hydraulic design of the cores remains for the future study.     

A study on the hydrogen recombination rates of catalytic recombiners and deliberate ignition

A study is being carried out by the Department of Nuclear and Mechanical Constructions at the University of Pisa on catalytic recombiners and on deliberately induced weak deflagration. These are the most practical methods for recombining hydrogen released into large nuclear containments during severe accidents. The recombination rates of different types of catalytic device were obtained from a thorough analysis of published experimental data. The main parameter that affects the effectiveness of these devices seems to be the molar density of the deficiency reactant rather than its volumetric concentration. The recombination rate of weak deflagrations in vented compartments has been assessed with experimental tests carried out in a small-scale glass vessel. Through a computerized system of analysis of video recordings of the deflagrations, the flame surface and the burnt gas volume were obtained as functions of time. These values of flame surface and burnt gas volume were used as inputs for a computer code to calculate the recombining rate, the burning velocity and the pressure transient in the experimental test. The code is being validated with a methodology principally based on a comparison of the measurements of pressure with the calculated values.     

Double-strata high burnup fuel performance in light water reactors

This study is aimed at the development of a fuel cycle concept for host countries with a lack of nuclear infrastructure. Two interrelated criteria, proliferation resistance and high-burnup, form the general framework of the fuel management scenario with the highest priority given to light water reactor technology and plutonium-free fresh fuel. Logically it implies the use of uranium oxide with enrichment close to 20%, whose effective utilization forms the main subject of the present paper. A sequence of two irradiation cycles for the same fuel pins in two different light water reactors is the key feature of the advocated approach. It is found that the synergism of PWR and pressure tube graphite reactor offers fuel burnup up to 140 GWd/tHM. Being as large as 8% in the final isotopic vector, the fraction of ²³⁸Pu serves as an inherent protective measure against plutonium proliferation.     

Western and Asian light water reactors at the millennium threshold

This paper presents the most advanced Western and Asian light water reactor (LWR) designs. The following pressurized water reactor (PWR) and boiling water reactor (BWR) designers are covered: Westinghouse ( ), Babcock and Wilcox (B&W—now part of Framatome), Combustion Engineering (CE—now ABB CE), Siemens (PWR), Framatome, Mitsubishi, General Electric (GE), Asea Brown Boveri (ABB), Siemens (BWR), Hitachi and Toshiba. The motivations that led to the design of the next generation of LWRs are discussed. The technical bases for evolutionary and innovative plants are summarized. Important safety features of some of the most complete (in operation, under construction or certified) evolutionary designs are described detail. Analogous implementations of systems into other advanced designs are given.     

Review of thermal-hydraulic researches in severe accidents in light water reactors

The Tohoku Region Pacific Coast Earthquake and subsequent severe accident (SA) in Fukushima Daiichi Nuclear Power Station caused unprecedented disaster in Japan. Before this accident, considerable researches on SAs had been carried out in Japan. However, unfortunately, such researches could not prevent the accident due to the unexpected huge Tsunami. However, the researches on SAs become more and more important in order to make clear the causes of the accident in Fukushima and improve the safety of nuclear power plants in Japan. In view of this, review on researches on thermal hydraulics in SAs in light water reactors was carried out. Important thermal-hydraulic phenomena in SAs were identified. Research activities on each phenomenon were surveyed mainly based on the articles published in Journal of Nuclear Science and Technology of Atomic Energy Society of Japan.     

Thermal-hydraulic transient analysis of light water reactors using moving boundaries

This paper examines the applicability of a mathematical dynamic model developed here for the simulation of the thermal-hydraulic transient analysis for light water reactors (LWRs). The thermal-hydraulic dynamic modeling of the fuel pin and adjacent coolant channel in LWRs is based on the moving boundary concept. The fuel pin model (FUELPIN) with moving boundaries is developed to accommodate the core thermal-hydraulic model, with detailed thermal conduction in fuel elements. Some results from transient calculations are examined for the first application of the thermal-hydraulic model and the fuel pin model with moving boundaries in a boiling water reactor (BWR). An accurate minimum departure from nucleate boiling ratio (MDNBR) and its axial MDNBR boundary versus time within the fuel channel are predicted during transients. Transient analysis using a known thermal-hydraulic code, COBRA and FUELPIN linked with a PWR systems analysis code show that the thermal margin gains more by a transient MDNBR approach than the traditional quasi-steady methodology for a pressurized water reactor (PWR). The studies of the overall nuclear reactor system show that moving boundary formulation provides an efficient and suitable tool for thermal transient analysis of LWRs.     

Calculations of hydrogen detonations in nuclear containments by the random choice method

Computer codes were developed for the prediction of pressure histories at different points of a nuclear containment wall due to postulated internal hydrogen detonations. These pressure histories are required to assess the structural response of a nuclear containment to hydrogen detonations. The compressible flow equations including detonation, which was treated as a sharp fluid discontinuity, were solved by the random choice method which reproduces maximum pressures and discontinuities sharply. The computer codes were validated by calculating pressure profiles and maximum wall pressures for plane and spherical geometries and comparing the results with exact analytic solutions. The two-dimensional axisymmetric program was used to calculate wall pressure histories in an actual nuclear containment. The numerical results for wall pressures are presented in a dimensionless form, which allows their use for different combinations of hydrogen concentration, and initial conditions.     

Risk assessment in setting inservice inspection priorities at light water reactors

As part of the Nondestructive Evaluation Reliability Program, sponsored by the U.S. Nuclear Regulatory Commission, Pacific Northwest Laboratory is developing a method that uses risk-based approaches to establish inservice inspection plans for nuclear power plant components. The method first uses probabilistic risk assessment (PRA) results and Failure Modes and Effects Analysis (FEMA) techniques to identify and prioritize the most risk-important systems and components for inspection. The acceptable level of risk from structural failure for important systems and components is then apportioned as a small fraction of the total PRA estimated risk for core damage. This process determines the target (acceptable) risk and failure probability values for individual components. The Surry Unit 1 Nuclear Power Station was selected for pilot applications of the method. The specific systems addressed are the reactor pressure vessel, the reactor coolant, the low-pressure injection, and the auxiliary feedwater. The results provide a risk-based ranking of components within these systems and relate the target risk to target failure probability values for individual components. These results will be used to guide the development of improved inspection plans for nuclear power plants.     

SBLOCA with boron dilution in pressurized water reactors. Impact on operation and safety

When the water level in the reactor pressure vessel (RPV) of a pressurized water reactor (PWR) is low enough and the core temperature is such that the coolant in that region boils, reflux-condensation conditions are established. Under such conditions, almost boron-free water is collected in a region of the primary system forming a non-borated slug. If subsequent natural circulation is established or a reactor coolant pump (RCP) is restarted, the slug could be transported to the core. This scenario configures an important part of the so-called boron issue. The Energy Systems Analysis Group at the Institute of Energy Technologies (INTE) of the Technical University of Catalonia (UPC) has studied the boron issue in three different stages. The steps were the following: participation in OECD-related projects, code improvement and investigation at nuclear power plant (NPP) scenarios. The third step is the main aim of this paper and consists of a continuation of the previous projects in the field of NPP analysis. The aim of this paper is to study SBLOCA transients with boron dilution in PWR. The chosen NPP was Ascó-2 which is a 3-loop-2940,6 MWth Westinghouse PWR. The paper contains some references to OECD/SETH and OECD/PKL experimental projects and analyses an established scenario including features of boron transport and sensitivity calculations for relevant parameters.