Safety-related Structures Research Articles

Radiation effects in concrete for nuclear power plants – Part I: Quantification of radiation exposure and radiation effects

A large fraction of light water reactor (LWR) construction utilizes concrete, including safety-related structures such as the biological shielding and containment building. Concrete is an inherently complex material, with the properties of concrete structures changing over their lifetime due to the intrinsic nature of concrete and influences from local environment. As concrete structures within LWRs age, the total neutron fluence exposure of the components, in particular the biological shield, can increase to levels where deleterious effects are introduced as a result of neutron irradiation. This work summarizes the current state of the art on irradiated concrete, including a review of the current literature and estimates the total neutron fluence expected in biological shields in typical LWR configurations. It was found a first-order mechanism for loss of mechanical properties of irradiated concrete is due to radiation-induced swelling of aggregates, which leads to volumetric expansion of the concrete. This phenomena is estimated to occur near the end of life of biological shield components in LWRs based on calculations of estimated peak neutron fluence in the shield after 80 years of operation.

Effects of soil-structure interaction on fragility and seismic risk; a case study of power plant containment

Abstract Safety-related structures are designed to provide a safe environment for the occupants and equipment during and after earthquakes. This is due to the fact that any damage imposed to the systems might lead to catastrophic consequences. Seismic probabilistic risk assessment (SPRA) is a systematic approach for the quantification of the seismic risk. One of the crucial steps in this assessment is to determine the seismic capacity of the structures by fragility method. After a review of available methodologies, this article analyzes the seismic fragility for a typical power plant containment considering the effects of soil-structure interaction (SSI). The structure and underneath soil profile are analyzed as a unified model by the subtraction method. Two steps are considered for the assessment of seismic response: In the first step, a fixed-base hypothesis framework is implemented to the computational problem. The second step covers computations taking into account the SSI effects. Using the results of seismic response analysis and safety factor method, seismic fragility of the structure is computed and related fragility curves are developed. Finally, by comparing the fragility curves, the effects of SSI are quantified on the overall seismic risk.

Simulation analysis of impact tests of steel plate reinforced concrete and reinforced concrete slabs against aircraft impact and its validation with experimental results

Abstract The steel plate reinforced concrete and reinforced concrete structures are used in nuclear power plants for protection against impact of an aircraft. In order to compare the impact resistance performance of steel plate reinforced concrete and reinforced concrete slabs panels, simulation analysis of 1/7.5 scale model impact tests is carried out by using finite element code ANSYS/LS-DYNA. The damage modes of all finite element models, velocity time history curves of the aircraft engine and damage to aircraft model are compared with the impact test results of steel plate reinforced concrete and reinforced concrete slab panels. The results indicate that finite element simulation results correlate well with the experimental results especially for constitutive winfrith concrete model. Also, the impact resistance performance of steel plate reinforced concrete slab panels is better than reinforced concrete slab panels, particularly the rear face steel plate is very effective in preventing the perforation and scabbing of concrete than conventional reinforced concrete structures. In this way, the thickness of steel plate reinforced concrete structures can be reduced in important structures like nuclear power plants against impact of aircraft. It also demonstrates the methodology to validate the analysis procedure with experimental and analytical studies. It may be effectively employed to predict the precise response of safety related structures against aircraft impact.

Local impact effects on concrete target due to missile: An empirical and numerical approach

Abstract Concrete containment walls and internal concrete barrier walls of a Nuclear Power Plant safety related structures are often required to be designed for externally and internally generated missiles. Potential missiles include external extreme wind generated missiles, aircraft crash and internal accident generated missiles such as impact due to turbine blade failure and steel pipe missiles resulting from pipe break. The objective of the present paper is to compare local missile impact effects on reinforced concrete target using available empirical formulations with those obtained using LS-DYNA numerical simulation. The use of numerical simulations for capturing the transient structural response has become increasingly used for structural design against impact loads. They overcome the limits of applicability of the empirical formulae and also provide information on stress and deformation fields, which may be used to improve the resistance of the concrete. Finite element (FE) analyses of an experimental impact problem reported by Kojima (1991) are carried out that are able to capture the missile impact effects; in terms of local and global damage. The continuous surface cap model has been used for modelling concrete behaviour. A range of missile velocity has been considered to simulate local missile impact phenomenon and modes of failure and to capture the concrete response from elastic to plastic fracture. A comparison is then made between the empirical formulations, numerical simulation results, and available experimental results of slab impact tests. While the numerical simulation is able to capture the experimental trend and results, a comparison of penetration depth and scabbing and perforation limits as per different empirical formulation shows substantial divergence.

A computational set of tools developed for fluence calculation as support for BWR's surveillance programme

The determination of a neutron source for the RPV fluence calculation is a key factor to detect and prevent degradation of safety-related structures of a NPP. A 3D neutron fluence calculation can be very challenging if all the source details are considered. Several options are possible in order to address the problem, being the most direct the use of 3D transport codes; however, strategies that reduce calculation times are sometimes the preferred. A fast methodology has been developed in order to simplify the calculations without losing accuracy. The methodology relies on the use of DORT together with the synthesis method for having a 3D neutron flux distribution. A set of auxiliary modules for the calculations of the fixed source distribution has been developed. Comparisons of the calculated neutron fluxes against the measurements from the surveillance capsules of the NPP Laguna Verde show maximum relative differences up to 6.81% by far within the 20% difference required by regulation.

FACTORS OF GROUNDWATER FLUCTUATION IN SHIN KORI NUCLEAR POWER PLANTS IN KOREA

To establish an aging management plan considering seawater influx and changes in groundwater within nuclear power plant sites, the characteristics of groundwater flow must be understood. This study investigated the characteristics of groundwater flow within the site and analyzed groundwater level recorded by monitoring wells to evaluate groundwater flow characteristics and elements that affected these characteristics for supplying the information to conduct the appropriate aging management for ensuring the safety of the safety-related structures in Shin Kori Unit 1 and 2. The increase in groundwater level during the wet season results from high sea-level conditions and the large amount of precipitation. As a result of the analysis of groundwater distribution and change characteristics, the site could be divided into a rainfall-affected area and a tide-affected area. First, the rainfall-affected area can further be divided into areas that are affected simultaneously by excavation, backfill, and a permanent dewatering system. Secondly, areas that are not affected by excavation, or the dewatering system, or by structure arrangement and excavation. Analysis of the spectrum for wells affected by tides resulted in confirmation of the M2 component (12.421 hr) and S2 component (12.000 hr) of the semidiurnal tides, and the O1 component (25.819 hr) of the diurnal tides. In the cross-correlation results regarding tides and groundwater levels, the lag time occurred diversely within 1-3 hours by the effect of the well location from sea, the distribution of the backfill material with depth, and the concrete structure.

Assess of the Nuclear Power Plant Structures Residual Life and Earthquake Resistance

The paper deals with the seismic analysis of safety related structures of an operating nuclear power plant. At present time the nuclear power plants of VVER-400/213 type operate for over thirty years and there are arising requirements to verify the actual state of structures in order to assess their residual life in general. A sophisticated computation model has been developed for the seismic structural analysis using the ANSYS program package. The model involves the complex of all constrained structures of two main production blocks with equipment. In order to get a general view at the seismic load effects, seismic response analysis has been performed using direct integration of equations of motion in 25 sec interval at 0.01 sec step with excitation described by accelerograms. Combinations of dead loads and seismic loads have been considered in the stress assessment of the structures. The results of the performed analyses form a base for residual life prediction of selected structures

Strength and ductility of RC beam-column joints of non-safety related structures and recommendations by national standards

► The paper presents results of experiments performed on RC beam-column joints. ► Specimens were replicas of the joints of real life non safety related structures. ► It was found that confinement, in general, improves the seismic behavior of joints. ► Confinement gives only marginal improvement in case of full-scale joints. ► National standards must prescribe limits on joint shear stresses. ► Reduction factors should consider the period of the structure. Every nuclear facility consists of various structures of varying importance. Standards and design guidelines recommend categorization of these structures based on their importance, safety function that they have to perform, hazard class, etc. In order to economize the designs, the structures that pose only industrial risk are recommended to be designed following the national practice. This paper presents the results of experiments performed on the beam-column joints of such structures to verify their seismic performances. The specimens were full-scale replicas of the joints of the original non safety related structures such as turbine buildings, intake structures and office buildings. The beam-column connections in the original structures either had non-seismic detailing or seismic detailing, but in order to have a comparison, every joint was tested with both types of detailing. The strength and ductility obtained from the experiments for both types was compared with the codal recommended values by different national standards. Conclusions and comments on the suggested values of allowable stresses, assumed ductility values and corresponding strength reduction factors have been made.

Impact of soft missiles – evaluation of load-time functions

The threat spectrum of safety-related structures has severely changed since 11 September 2001. The deterioration capability of large aircrafts impacting on structures has become aware in a dramatic manner. Simplified calculation models for aircraft impact are available but have not been verified for large commercial aircrafts. Based on this background, experimental studies are performed at the Materials Testing and Research Institute (MPA) Karlsruhe of the Karlsruhe Institute of Technology (KIT) to examine the shape and the peak loads of load-time functions induced by an impacting soft missile where the mass and stiffness distribution is similar to a commercial aircraft. The characterisation of the impact phenomena and their variation due to changing test parameters is the goal of these experiments. The experimental set-up, measurement results and their evaluation will be presented. First numerical simulation results will be shown and briefly compared to the prediction using the analytical Riera method.

Fast neutron fluence calculations as support for a BWR pressure vessel and internals surveillance program

Materials surveillance programs are required to detect and prevent degradation of safety-related structures and components of a nuclear power reactor. In this work, following the directions in the Regulatory Guide 1.190, a calculational methodology is implemented as additional support for a reactor pressure vessel and internals surveillance program for a BWR. The choice of the neutronic methods employed was based on the premise of being able of performing all the expected future survey calculations in relatively short times, but without compromising accuracy. First, a geometrical model of a typical BWR was developed, from the core to the primary containment, including jet pumps and all other structures. The methodology uses the Synthesis Method to compute the three-dimensional neutron flux distribution. In the methodology, the code CORE-MASTER-PRESTO is used as the three-dimensional core simulator; SCALE is used to generate the fine-group flux spectra of the components of the model and also used to generate a 47 energy-groups job cross section library, collapsed from the 199-fine-group master library VITAMIN-B6; ORIGEN2 was used to compute the isotopic densities of uranium and plutonium; and, finally, DORT was used to calculate the two-dimensional and one-dimensional neutron flux distributions required to compute the synthesized three-dimensional neutron flux. Then, the calculation of fast neutron fluence was performed using the effective full power time periods through six operational fuel cycles of two BWR Units and until the 13th cycle for Unit 1. The results showed a maximum relative difference between the calculated-by-synthesis fast neutron fluxes and fluences and those measured by Fe, Cu and Ni dosimeters less than 7%. The dosimeters were originally located adjacent to the pressure vessel wall, as part of the surveillance program. Results from the computations of peak fast fluence on pressure vessel wall and specific weld locations on the core shroud are also presented.

Foundation Floor Geological Mapping of a Nuclear Reactor

Foundation floor mapping is essential for very important structures to provide permanent data set for geological interpretations. Prototype fast breeder reactor (PFBR) is being constructing at Kalpakkam, India. Nuclear Island Connected Building (NICB) is the main nuclear safety related structure of (PFBR), and is resting on a raft of 101.5 m × 93 m size at about 20 m below the existing ground level. Geological floor mapping was prepared after the excavation. 2 m × 2 m grids were used for sampling study and mapping of the walls and floor. The floor region of this raft foundation was seen to contain certain weathered zone with fractures of thickness of in the order of 2 m. No evidences of faulting/shearing were observed on the surface of the floor area. The entire floor area was made up of compact and sound rock mass. However, there were some isolated patches of soft weathered/altered rock and small fractures in the rock. The structural features observed during the mapping exercise indicated the need for consolidation grouting so that the entire floor area function as single rock mass. This foundation geological map will assist in making better interpretation of post-construction foundation instrumentation data.

Realistic seismic qualification using the updated finite element model for in-core components of reactors

Abstract It is now mandatory to seismically qualify the safety-related structures and components used in the nuclear power plants. Among several qualification approaches the qualification by the analysis using finite element (FE) method is the most common approach used in practice. However, the estimated dynamic behaviour by FE model of a structure is known to show significant deviations from the dynamic behaviour of the ‘as-installed’ structure in many cases. Considering such limitations, few researchers have advocated re-qualification of such structures after their installation at site to enhance the confidence in qualification vis-a-vis plant safety. For such an exercise, validation of FE model with experimental modal data is important. A validated FE model can be obtained by the model updating methods in conjugation with the in situ experimental modal data. Such a model can then be used for qualification. However, for the reactor in-core components such a modal testing and FE model updating may not be straightforward. Hence, the complication involved in the reliable seismic qualification of in-core components and the advantage of using the FE model updating has been brought out in the paper through an example of a typical in-core component—a perforated horizontal tube recently installed in a nuclear reactor in India.

Probabilistic methods for assessing current and future performance of concrete structures in nuclear power plants

Concrete safety-related structures and components in nuclear plants may be exposed to aggressive service and environmental effects that may cause their strength and stiffness to deteriorate over an extended period of service. A performance assessment of an existing structure subject to deterioration requires a consideration of likely degradation mechanisms, a framework for managing the inevitable uncertainties in operating or environmental loads, capacities, and nondestructive evaluation technologies, and tools for determining whether the observed degradation impacts the remaining service life or safety margin of the plant to any significant degree. Research is in progress to support the development of condition assessment tools and guidelines using modern structural reliability analysis methods.

Seismic safety of nuclear power plants in eastern Europe

Abstract This paper summarizes the work performed by the International Atomic Energy Agency in the areas of safety review and applied research in support of programmes for the assessment and enhancement of seismic safety in Eastern Europe and in particular, WWER type nuclear power plants during the past seven years. Three major topics are discussed; engineering safety review services in relation to external events, technical guidelines for the assessment and upgrading of WWER type nuclear power plants, and the Coordinated Research Programme on Benchmark study for the seismic analysis and testing of WWER type nuclear power plants. These topics are summarized in a way to provide an overview of the past and present safety situation in selected WWER type plants which are all located in Eastern European countries. The main conclusion of this paper is that even though there is now a thorough understanding of the seismic safety issues in these operating nuclear power plants, the implementation of seismic upgrades to structures, systems and components are lagging behind, particularly for those cases in which re-evaluation indicated the necessity to strengthen the safety related structures or install new safety systems.

Aging management of containment structures in nuclear power plants

Research is being conducted by Oak Ridge National Laboratory under US Nuclear Regulatory commission (USNRC) sponsorship to address aging management of nuclear power plant containment and other safety-related structures. Documentation is being prepared to provide the USNRC with potential structural safety issues and acceptance criteria for use in continued service evaluations of nuclear power plants. Accomplishments include development of a Structural Materials Information Center containing data and information on the time variation of 144 material properties under the influence of pertinent environmental stressors or aging factors, evaluation of models for potential concrete containment degradation factors, development of a procedure to identify critical structures and degradation factors important to aging management, evaluations of non-destructive evaluation techniques, assessments of European and North American repair practices for concrete, review of parameters affecting corrosion of metals embedded in concrete, and development of methodologies for making current condition assessments and service life predictions of new or existing reinforced concrete structures in nuclear power plants.

Reaction-time response of aircraft crash

Abstract The safety related structures of a Nuclear Power Plant (NPP) are required to be designed for the impact of aircraft. The reaction-time response is usually obtained assuming the target to be rigid. The assumption is claimed to be conservative in the case of stiff structures. The effect of target yielding has been studied by considering an aircraft crash upon the outer containment of an NPP. It has been observed in this investigation that the assumption of a rigid target is unconservative for some of the cases. It is usually assumed that variation in crushing strength has very little effect on the reaction from the target. However, the sensitivity analysis for reaction-time response indicates that both linear mass density and crushing strength are sensitive in affecting the reaction from the target, depending upon the characteristics of the aircraft and the striking velocity. The concept of confidence level has been investigated in relation to inclined targets also and confidence curves are obtained for aircraft.

Soil/structure seismic investigation of safety-related structures: Serhan, S J; Chen, C Proc 6th ASCE Speciality Conference on Probabilistic Mechanics and Structural and Geotechnical Reliability, Denver, 8–10 July 1992P396–399. Publ New York: ASCE, 1992

This paper promotes an improved understanding of the type of earthquake motions felt at V.C. Summery, Perry and Krsko nuclear power plants. An analytical investigation is performed to describe how the Krsko earthquake might be influenced by soil/structure interaction effects including several parametric studies to account for uncertainties in the soil properties. The analysis takes into account nonlinearity of the soil material, radiation and hysteretic damping, structural embedment, and structure/structure interaction. Results of analysis show clear evidence of soil/structure interaction, nonlinear softening of the soil material and encouraging qualitative and quantitative agreement with the recorded measurements. The structural response motions display high rocking mode.

Design basis documentation program-a critical overview

It is noted that many nuclear power plant owners have undertaken a massive effort of reviewing and reconstituting the design basis documentation (DBD) for the safety-related systems, components, and structures of their plants. In spite of the Nuclear Management and Resource Council's (NUMARC's) guideline on this issue, a significant amount of confusion exists in the industry about the need, requirements, extent, and the depth of the review and reconstitution efforts. A well-developed DBD program, developed on the basis of a plant's unique design, licensing, and operation history, can be very cost-effective and can pay back the investment within a short time. Ingredients necessary for developing such a program are discussed. A critical overview of the various parameters, both technical and programmatic, affecting the DBD issue is also presented. The author addresses two documents that he considers to be the keys to an effective DBD program: (i) a road-map-type plant-level DBD that will establish the plant design bases through tables and matrices uncluttered by lengthy design change and licensing histories, and (ii) a comprehensive document establishing the design margin of individual components. >

Ageing management of safety-related concrete structures to provide improved bases for continuing the service of nuclear power plants

On decrit les structures en beton se rapportant a la securite des installations nucleaires, et on determine les facteurs de degradation susceptibles de compromettre la conformite de ces structures aux criteres de fonctionnement et de performance qui leur sont assignes. On donne ici, de facon resumee, les exigences d’inspection en service. On examine l’evaluation des performances des elements en beton dans les installations nucleaires. Enfin, on trace les grandes lignes d’un programme pour gerer le vieillissement des structures en beton en rapport avec la securite des installations nucleaires, et ameliorer leur capacite de service. On resume egalement les differents aspects du programme.

Probabilistic descriptions of resistance of safety-related structures in nuclear plants

Abstract Calculations of reliability of safety-related nuclear plant structures require a knowledge of the probability distributions that describe their resistance. This study considers the applicability of existing statistical data for describing the resistance of steel and reinforced concrete nuclear plant structures. Probability distributions are recommended which can be used in assessing the reliability of containments and other Category I structures, developing fragilities, and selecting appropriate resistance criteria for probability-based structural design.