Acoustic emission surveillance for improving availability of nuclear power plants

An experimental study of advanced acoustic emission monitoring of BWR components is being performed at Philadelphia Electric Company's Peach Bottom Atomic Power Station. The study addresses the feasibility of using continuous acoustic emission monitoring for increasing availability of LWR power units. Here we present a summary of a “Valve Surveillance” program and we describe the installation and performance of AE instrumentation for the “Pipe Surveillance” program.     

The development of a diagnostic and monitoring system for vibrating pipe branches

This paper deals with a diagnostic and monitoring system for assessing the integrity of pipe branches, during the operation of the nuclear power plant. This system have been developed under the concept of “easy to use without any sophisticated analysis” and “portable”. The accuracy of the diagnosis is based on the model optimization subsystem, which automatically modifies the numerical vibration model so as to fit its natural frequency to the actual natural frequency. The information obtained by this system may be reflected to a maintenance program of the plant to assure more reliable operation of the plant.     

Acoustic emission monitoring of ZB-1 intermediate scale vessel test

A program to develop the use of acoustic emission (AE) flaw detection methods for continuous surveillance of reactor pressure boundaries is in process in the United States. Evaluation of laboratory developed relationships for data verification and interpretation was performed by participation in a German intermediate scale vessel (ZB-1) test. The test sequence consisted of repeated blocks of a hydrostatic test followed by two sets of cyclic loading at different R-ratios. Testing was performed in cooperation with the German Materialprüfungsanstalt at the Grosskraftwerk facility in Mannheim, West Germany. This paper discusses preliminary results obtained during the first half of the test which was performed at 70°C. The AE system detected crack growth from machined flaws and also spontaneous crack growth in a fabrication weld. AE signals from cracking were consistently high amplitude and occurred at or near peak load. Crack growth rates estimated from AE data were consistent with values derived from crackopening-displacement gauges. The test produced unique and important data needed to develop reliable application of AE methods for continuous monitoring of reactor pressure systems.     

Acoustic emission for on-line reactor monitoring: Results of intermediate vessel test monitoring and reactor hot functional testing

This article discusses a program designed to develop the use of acoustic emission (AE) methods for continuous surveillance to detect and evaluate flaw growth in reactor pressure boundaries. Technology developed in the laboratory for identifying AE from crack growth and for using that AE information to estimate flaw severity is now being evaluated on an intermediate vessel test and on a reactor facility.A vessel, designated ZB-1, has been tested under fatigue loading with simulated reactor conditions at Mannheim, West Germany, in collaboration with the German Materialprüfungsanstalt (MPA), Stuttgart. Fatigue cracking from machined flaws and in a fabrication weld were both detected clearly by AE.AE data were measured on a US nuclear reactor (Watts Bar, Unit 1) during hot functional preservice testing. This demonstrated that coolant flow noise is a manageable problem and that AE can be detected under operational coolant flow and temperature conditions.     

Some conclusions with regard to the safety assessment of cracked components drawn from the research programme “integrity of components” (FKS II) atthe present state

The continuation of the research program “Integrity of Components”, Phase II, mainly deals with further evaluation and assessment of material properties and the application of data from small standard specimens to large scale specimens and components. This includes the use of advanced numerical methods to check the transferability of fracture mechanics parameters with regard to the type of load and degree of multiaxiality on the failure behaviour of fracture mechanics specimens with component-like dimensions. Further points of interest are the relationship between upper shelf toughness and load-bearing capacity, the influence of neutron irradiation on the properties, and the effect of corrosion on cyclic crack growth.     

The use of energy absorbers to protect structures against impact loading

When a flying missible impacts a fixed structure, the interface loading is dependent on the deformation characteristics of both impacting and impacted bodies. If both are too rigid to accommodate the amount of gross deformation required to neutralize the incoming kinetic energy, or if such energy absorption has a chance to proceed in uncontrolled and unreliable ways, then there is a need to interpose a specifically designed “energy absorber” between missile and structure, from which a well-defined load time history can be derived during the course of impact.

The required characteristics of such an energy absorption material are:

• the capability to accommodate large permanent deformation without structural failure; and

• the reliable and controlled “load-deformation” (or “stress-strain”) behaviour under dynamic conditions, with an aim at an optimal square shape curve.

Consideration must also be given to environmental or other disturbing effects, like temperature, humidity, and “out of plane” loading. A short survey is presented of the wide range of energy absorbers already described in technical papers or used in a number of practical safety applications within varied engineering fields (from vehicle crash barriers to high energy pipe whipping restraints). However, with such open a literature, information is usually lacking in the specific data required for design analysis.

The following “energy absorption” materials and processes have thus been further experimentally investigated, with an a aim at pipe whipping restraint application for nuclear power plants:

1. (1) plastic extension of austenitic stainless steel rods;

2. (2) plastic compression of copper bumpers; and

3. (3) punching of lightweight concrete structures.

Dynamic “stress-strain” characteristics have been established for stainless steel bars at several temperatures under representative loading conditions. For this purpose, a test rig has been specifically designed to incorporate a number of adjustable parameters and to behave as a representative “slice” of an actual pipe whipping restraint; typical strain rates are in the 10 sec⁻¹ range. The behaviour of copper bumpers has been compared under static and dynamic conditions (using a conventional drop weight test (DWT) machine); as no significant strain rate effects were emphasized, only static tests have been further developed. The DWT rig was used again to investigate crushing or punching of cellular concrete under varying geometries and loading conditions. To remedy certain deficiencies of the regular commercial grades of cellular concrete, special lightweight mixtures have been studied to optimize material toughness and provide a wider range of specific resistance.Results of this experimental program are presented and discussed. The use of energy absorbers is then illustrated for a few typical pipe whipping restraints. The design of restraints is based on real dynamic characteristics of “energy absorption” material as produced by the test program. To derive design loads of restraints, a number of methods can be used ranging from a simplified “energy balance” graph to sophisticated plastodynamic computer analysis. Typical results are presented and discussed to compare the efficiency of these alternative methods.     

A two-dimensional model for fluid-structure interaction in curved pipes

A “channel” model was developed for the purpose of simulating the interactive fluid-structural response of curved pipes to pressure pulses. Simulation is shown to have been achieved analytically in both the axisymmetric (“breathing”) and transverse (“bending”) modes of interactive behavior.An experimental program which was aimed at the validation of the model is also described. Tests were run in both straight and curved pipe configurations. Comparisons between measurements and model calculations demonstrate the validity of the model within the range of parameters under consideration.The model was implemented into the DISCO code for nonlinear fluid-shell interaction.     

SCALE calculation for evaluation of spent fuel condition during long-term storage

Experiences with an advanced spent nuclear fuel management in Slovakia are presented in this paper. The evaluation and monitoring procedures are based on practices at the Slovak wet interim spent fuel storage facility in NPP Jaslovské Bohunice. Since 1999, leak testing of WWER-440 fuel assemblies were completed using a special leak tightness detection system developed by Framatome-anp, “Sipping in Pool”. This system utilized external heating for the precise defects determination.Optimal methods for spent fuel disposal and monitoring were designed. A new conservative factor for specifying of spent fuel leak tightness is introduced in the paper. Limit values of leak tightness were established from the combination of SCALE4.4a (ORIGEN-ARP) calculations and measurements from the “Sipping in Pool” system. These limit values are: limiting fuel cladding leak tightness coefficient for tight fuel assembly – k_FCT(T) = 3 × 10⁻¹⁰, limiting fuel cladding leak tightness coefficient for fuel assembly with leakage – k_FCT(L) = 8 × 10⁻⁷.     

An assessment of acoustic emission for nuclear pressure vessel monitoring

Recent research has greatly improved our understanding of the basic mechanisms of deformation and fracture that generate detectable acoustic emission signals in structural steels. A critical review of the application of acoustic emission (AE) to the fabrication, proof testing and in-service monitoring of nuclear pressure vessels is presented in the light of this improved understanding. The detectability of deformation and fracture processes in pressure vessel steels is discussed, and recommendations made for improving source location accuracy and the development of quantitative source assessment techniques.

Published data suggest that AE can make an important contribution to weld fabrication monitoring, and to the detection of defects in lower toughness materials during vessel proof testing. In high toughness materials, however, the signals generated during ductile crack growth may frequently be too weak for reliable detection. The feasibility of AE for continuous monitoring has not yet been adequately demonstrated because of high background noise levels and uncertainty about AE signal strengths from the defect growth processes that occur in service. In-service leak detection by AE shows considerable promise.

It is recommended that further tests are carried out with realistic defects, and under realistic conditions of loading (including thermal shock and fatigue) and of environment.     

The susceptibility of SA-508-II and SA-533B to reheat cracking-evaluation of empirical and theoretical predictive methods

The primary objection of this program was to evaluate, by means of actual welding tests, the various empirical and theoretical methods for predicting reheat cracking in 508-II type and 533B type nuclear pressure vessel steels. The program concluded that the “Threshold Value Criterion” method is a generally good way to evalute 508-II materials. The program also found that SA 533B materials are less prone to reheat cracking than 508-II type materials, but 533B is not immune to this type of cracking.     

On-line monitoring the incore power distribution by using excore ion-chambers

For most of nowaday pressurized water reactors (PWR), excore ion-chambers (EIC) are the sole real-time sensors responding to the incore power distribution (IPD). Conventionally, it is supposed that the EIC could not carry out information of the IPD in radial direction. As a result, there is an excess conservatism when they are being used for the core monitoring so far, and some flexibilities of the reactor operation are sacrificed. Though progresses have been made, there are still some inevitable obstacles about newly developed monitoring methods: mounting and maintaining a large amount of fixed incore detectors are costly; and the IPD calculated by on-line simulators is not the “measured” one. Fortunately variations in the IPD must be driven by reasonable physical causes, and the corresponding variations in the readings of the EIC are notable enough to be identified. So with a specific referenced IPD measured from periodical flux mappings, it is possible to on-line monitor such variations by using the EIC. Aimed at this, a harmonics grouping method and an algorithm of higher order perturbation variables are developed, and an influence equation for control rods is also proposed. Then a precise restriction equation describing the reasonable variations in the IPD is deduced for example of the Heating Reactor. Combined with a proper-configured EIC, a systematic “excore-to-incore” method is realized finally.     

The use of acoustic emission to determine characteristic dynamic strength and toughness properties of steel

The dynamic J integral at crack initiation (J_id) and dynamic yield stress (σ_yd) are useful parameters to characterize elastic-plastic material behaviour under rapid loading rates. The critical step for evaluating J_id and σ_yd under the condition of the three point impact bending test is the detection of the crack initiation and of the yield point in the impact load–deflection curve, respectively. This paper presents an acoustic emission (AE) based method to determine the ductile crack initiation and additionally the beginning of yield. The experimental techniques used to evaluate σ_yd and J_id include both instrumented pendulum impact tests with the AE transducer within the striker (tup) and medium rate three point bend (TPB) tests with additional AE transducers on the specimen surface. Results obtained from the tests indicate that the AE method is capable of detecting general yielding and the onset of ductile crack growth (initiation). Different types of pulse shaped AE signals can be observed. They were connected with characteristic features during the loading process.     

Characterization of industrial NbTi strands at variable field for JT-60SA toroidal field coils

The mission of the JT-60SA Tokamak, to be built in Japan, is to contribute to the early realization of fusion energy by its exploitation in support of the ITER program. JT-60SA project is an important part of the “broader approach” activity as a satellite program for ITER. The toroidal field (TF) coils are a European “in kind” contribution and they will partly be built by France. JT-60SA TF coil uses the Cable In Conduit Conductor (CICC) with NbTi superconductor strands. TF conductors will have to operate at 5.7 T, 5 K and at current density of 450 A/mm² with sufficient margins. In the framework of JT-60SA TF coil manufacture, the variable temperature characterization is an important step to select NbTi strand. At an early stage of design, we had to choose the strand with acceptable performances. During the design qualification and validation stage, it is important to qualify strands in conditions close to the operation conditions. The industry has proposed various strands manufactured with different processes. This work and publication examines a strand with an internal CuNi barrier, which is expected to lead to better current distribution between strands, by more precise calibration and control of the inter-strand resistance. The strands were tested at the Grenoble High Magnetic Field Laboratory facility. The domain (B, T, J) explored was in the range of 4.5–11 T for the magnetic field intensity, 4.2–6.5 K for the temperature and between 40 A/mm² and 1200 A/mm² for the current density. For each strand, “critical current density” and “current sharing temperature” measurements have been carried out, with a temperature precision of few tens of mK. Once the measurements performed, the fitting parameters (of type J_C = f(B, T)) of each strand have been found, by performing regression analysis. This work will lead to select the strand with the best characteristics. In this paper, we present the results of this measurement task, the data and regression analysis (fits, T_cs, etc.) and the conclusion about the strand choice.     

The safety of Modular High Temperature Gas-Cooled Reactors

Modular High Temperature Gas-Cooled Reactors (MHTGRs) have an unprecedented degree of safety due primarily to coated-particle fuel retaining radiologically-significant nuclides within the fuel coatings. The practical ability of coated fuel particles to quantitatively retain important fission products up to 1600–1700°C has been demonstrated in the Federal Republic of Germany (FRG), and the United States (USA) is proceeding with similar demonstrations. The USA program has placed emphasis on development of fuel performance models, using experimental data for that development. The mathematical models predict high retention of fission products for reference-quality MHTGR fuel, and the predictions are supported by the FRG experimental data. The performance of USA fabricated fuel has not been as high as expected, and additional development is needed. Specific safety analyses and results obtained from various “bounding” type accidents based on reference-quality fuel support the claimed high degree of inherent safety of MHTGRs.     

Main R&D issues for fast reactor structural design standard

For realization of economical and reliable fast reactor (FR) plants, the Japan Atomic Energy Agency (JAEA) and the Japan Atomic Power Company (JAPC) are cooperating on the “Feasibility Study on Commercialized FR Cycle Systems”. To certify the design concepts through evaluation of the structural integrity of FR plants, the research and development of the “Elevated Temperature Structural Design Guide for Commercialized Fast Reactor (FDS)” is recognized as an essential theme. The FDS focuses on particular failure modes of FRs such as ratchet deformation and creep-fatigue damage due to cyclic thermal loads. For precise evaluation of these modes, the research and development for three main issues is in progress. First, the “Refinement of Failure Criteria” needs to be addressed for particular failure modes of FRs. Secondly, the development of “Guidelines for Inelastic Design Analysis” is conducted to predict elastic plastic and creep deformation under elevated temperature conditions. Lastly, efforts are being made toward preparing “Guidelines for Thermal Load Modeling” for the design of FR components where thermal loads are dominant.     

Potential application of NRC aging research to license renewal

A strong technical base, when developed and implemented, to manage aging in plant safety related systems, support systems, structures, and components will give confidence to all of us in the nuclear community with regard to our ability to maintain continuous safe operation of nuclear power plants of all ages. This technical base for managing aging must be built over the foundation of reviews and analyses of original designs, operating experience of over 20 years, experts' opinions, and the utilization of research results. The key elements of managing aging are the understanding of risk significance of aging phenomena and the licensee program(s) for inspection, surveillance, condition monitoring, trending, record-keeping and maintenance to mitigate the influence and effects of aging.An overview is provided of the intended application of plant aging research results. In the process it delineates some key steps recommended for the development of appropriate regulatory guides and review procedures involving the technical issues for license renewal considerations. This application of aging research results is based primarily upon the NPAR theme of “Understanding Aging-A Key to Ensuring Safety, and Managing Aging-A Necessity to Ensuring Safety.”     

Fuel coolant interaction experiments in TROI using a UO2/ZrO2 mixture

This paper discusses the results of steam explosion experiments using reactor material carried out under “Test for Real cOrium Interaction with water (TROI)” program. About 4–9 kg of corium melt jet is delivered into a sub-cooled water pool at atmospheric pressure. Spontaneous steam explosions are observed in four tests among six tests. The dynamic pressure, dynamic load, and morphology of debris clearly indicate the cases with steam explosion. The initial conditions and results of the experiments are discussed.     

Intermediate size pressure vessel programme (ZB 1)

One of the focal points in the discussion about the safety of nuclear power plants is the integrity of the reactor pressure vessel.In order to prove its integrity tests are in progress in an underground test facility of the main power station in Mannheim with an intermediate size vessel from the research programme “Integrity of Components”. Patches of A 533 B and modified A 508 B material were welded into the vessel ZB 1, the test temperatures are approximately 70 and 290°C. The main goal of the tests is to measure the behaviour of artificial and natural flaws during static hydrotests and simulated operational (cyclic) conditions.In the first half of the research programme the objective is to produce a crack growth of some centimetres by cyclic loading between a variable minimum pressure and a maximum pressure of about 24 MPa. The total number of load cycles will be approximately 30 000.In the second half of the tests the vessel will be loaded by a number of pressure cycles which correspond to the loading a reactor pressure vessel experiences during 40 years of operation.During the static and cyclic loading acoustic emission monitoring is being made by German and American laboratories.This paper presents details of the vessel, the test loop, results of the nondestructive examinations conducted to quantify the crack depths and results of the acoustic emission monitoring.     

The SQUG program for resolution of USI A-46 — Status and implementation plans

In response to U.S. Nuclear Regulatory Commission (NRC) Unresolved Safety Issue (USI) A-46, “Seismic Qualification of Equipment and Operating Nuclear Power Plants”, the Seismic Qualification Utility Group (SQUG), with the support of the Electric Power Research Institute (EPRI), developed a comprehensive program to verify the seismic adequacy of equipment in operating nuclear power plants. The primary thrust of the program has been the development of procedures, criteria, and data to apply actual experience on the performance of equipment during earthquakes to the verification of seismic ruggedness of similar equipment in nuclear plants. While the use of such experience data continues to play a primary part in the SQUG program for resolution of USI A-46, the overall SQUG program includes a number of other significant elements which, taken together, provide a comprehensive approach for verification of the seismic adequacy of equipment in nuclear plants. These elements of the SQUG program include the assimilation and use of seismic shake table data in a generic way; the development of simplified analytical tools and criteria for evaluation of equipment anchorage, tanks, heat exchangers and cable trays; and the development of procedures for identifying and evaluating electrical relays, which are essential to plant shutdown in response to an earthquake. Procedures and data bases for performing and documenting the various seismic evaluations and plant walkdowns, and a program for training the large number of engineers who will be required to implement the SQUG methodology, have also been developed. This paper describes the main elements of the SQUG program for resolution of USI A-46 and provides a status report on the plans for their implementation in SQUG member plants.     

The retardation of the released fission products deriving from graphite structures during a hypothetical HTR-500 accident

A comprehensive computer model is presented describing the transport paths of the released fission products in the coolant gas. The transport mechanisms within the graphite are discussed in detail.An experimental assembly for the verification of the computer model is described and the measurements carried out on the retardation (retention capability) of cesium in reflector graphite are presented and compared with the calculations from the “PATRAS-CORE” program.First reliable statements under realistic conditions can be made with the example of a core heat-up accident in the HTR-500.