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.     

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.     

Acoustic monitoring of containment tendons

Assured safety and operational reliability of post-tensioned concrete components of nuclear power plants are of great significance to the public, electric utilities and regulatory agencies. Prestressing tendons provide the principal reinforcement for 40% of the containment structures in the United States. This paper briefly examines current in-service inspection requirements for prestressed containments and also reviews the feasibility of developing a passive surveillance system for identification of ruptures in tendon wires and its application to a one-tenth scale ring model containment structure.     

Acoustic emission measurements on large scale tensile specimens within the FKS-program (research program integrity of components)

Within the FKS-Program (Research Program Integrity of Components) acoustic emission measurements, mostly done in the new 100 MN-machine at the MPA Stuttgart, were carried out on large flat and cylindrical specimens manufactured from partly modified reactor steels with the following purposes (1) to distinguish the different origins of acoustic emission, (2) to recognize the onset of subcritical crack growth and (3) to predict fracture.     

PISC full scale reactor pressure vessel validation of non-destructive examination

The PISC III Programme involves validation of techniques and procedures and, within this programme, evaluation has now started on the ability to discriminate service induced defects from indications produced by fabrication defects in A 508 Class 2 material when sensitive techniques are used.Action No. 2 of PISC III: Full Scale Vessel Testing is designed for the performance demonstration of three groups of inspection procedures:

• - Mechanized ASME type procedures with variable recording level and complementary techniques

• - Industrial full ISI procedures (mechanized);

• - Several detailed evaluation procedures (generally mechanized) based on advanced techniques to be used on defective areas detected by usual inspection.

These procedures, typical for ISI in most of the cases, are applied in four situations which could be typical of old and new LWR pressure vessels:

• - vessel material and welds containing important service and fabrication defects but mixed with base material defects and small welding defects;

• - nozzle to shell welds with typical service defects, often well isolated and distant from other defective areas in rather clear material and/or welds;

• - nozzle inner radius defects;

• - artificially heat and unbranched fatigue defects in the test blocks assembled to simulate a PWR pressure vessel wall portion.

The paper summarizes the PISC II programme results which stress the characteristics of capable NDT techniques, in opposition to material characteristics like acceptable base material defects. It describes the full scale pressure vessel components available to conduct the PISC III exercise with improved ultrasonic techniques.     

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.     

Acoustic emission measurements during welding of a 250 mm testplate

The capability of the acoustic emission monitoring system of Prine (Gard Corporation, Illinois) should be demonstrated for a multi-layer submerged arc weld of a 250 mm test plate. The question had to be answered, if the measuring system developed by Prine and the evaluation models are suitable for production control of multi-layer submerged-arc welding, respectively if defect detection with interpretation of defect type, classification of size and triangulation is possible.The Prine system works with regard to electronic equipment as well as the interpretation quite satisfactory. The agreement of the defect indication with the defects visible during welding is very well. Hot and cold cracks as well as slag inclusions were indicated and normally evaluated correctly. Pores were not detected during welding.Partly a lot of low signals were realized, especially if residual stresses (closure seam) were expected. These small indications at the sensitivity level of the system could not be precisely explained. A clear evaluation of the results will be possible after completing nondestructive and destructive testing of the weld.     

Seismic proof test of a reinforced concrete containment vessel (RCCV): Part 1: Test model and pressure test

In Japan, the Nuclear Power Engineering Corporation (NUPEC), sponsored by the Ministry of Economy, Trade and Industry (METI), has been conducting a series of seismic reliability proving tests using full-scale or close to full-scale models to simulate actual important equipment that is critical for seismic safety of nuclear power plants. The tests are intended to validate the seismic design and reliability with a sufficient margin even under destructive earthquakes. A series of tests was carried out on a reinforced concrete containment vessel (RCCV) for advanced boiling water reactor (ABWR) from 1992 to 1999. A large-scale high-performance shaking table at Tadotsu Engineering Laboratory, the largest in the world, was used for this test. Part 1 reports the test model and the results of pressure and leak tests. Part 2 describes test procedures, input waves and the results of verification tests such as changes of stiffness, characteristic frequency and damping ratio, the failure of the model and the load deflection. Part 3 shows the seismic safety margin that was evaluated from the energy input during the failure test to a design basis earthquake. Part 4 reports simulation analysis results by a stick model with lumped masses.     

Acoustic method for monitoring the state of VVER vessels

The results of an investigation of the possibility of using acoustic characteristics to evaluate the state of the VVER vessel material are presented. The investigations were performed on template–samples cut from the VVER-440 vessel material and on control samples from VVER-1000 vessels. It is shown that the characteristics of elastic waves depend on the fluence and the radiation embrittlement of VVER vessel materials. Translated from Atomnaya énergiya, Vol. 106, No. 1, pp. 31–35, January, 2009.     

ZB2-tests — Description of the intermediate size vessel and AE results

To demonstrate and to extend the performance of acoustic emission testing as a method of detecting and classifying flaws, six institutes conducted acoustic emission measurements in the course of various loading tests on a medium-sized, thickwalled vessel (model of a reactor pressure vessel) containing natural flaws. This paper will present an outline of the working program of the MPA within the framework of this project. The presentation is concentrated on the vessel manufacturing, the preparation of the flaw patch with 14 natural flaws, the performance of the loading tests to simulate pressure test and operating conditions existing in the primary systems of pressurized water reactors, and especially on the conduction of acoustic emission measurements as a method of permanent vessel monitoring during the tests.In the pressure tests conducted with slowly rising pressures, only one flaw was detected unequivocally by the acoustic emission monitoring, although several flaws had grown in the test phases between the pressure tests. In fast pressure tests, flaws in principle were detected slightly better. Cyclic loading over prolonged periods of time produced clear signals of larger flaws, which calculations and subsequent destructive investigations showed to have grown. The small flaws which, most probably, had not changed, could not be detected.     

Reactor pressure vessel design of the high temperature engineering test reactor

The reactor pressure vessel (RPV) of the HTTR is 5.5 m (inside diameter), 13.2 m (inside height), and 122 mm (shell thickness). The RPV contains core components, reactor internals, reactivity control system, etc.2 1/4Cr–1Mo steel is chosen as the material for RPV. The temperature reaches about 400 °C at normal operation. The fluence of the RPV is estimated to be less than 1 × 10¹⁷ n/cm² (E > 1 MeV) and so irradiation embrittlement is negligible, but temper embrittlement is not negligible. For the purpose of reducing embrittlement, content of some elements must be limited in the 2 1/4Cr–1Mo steel for the RPV; embrittlement parameters, J-factor and are used.In this paper, design and structure of the RPV are reviewed first. Fabrication procedure of the RPV and its special feature are described. Material data on the 2 1/4Cr–1Mo steel manufactured for the RPV, especially the embrittlement parameters, J-factor and , and nil-ductility transition temperatures, T_NDT, by drop weight tests, are shown. In-service inspection and results of R&Ds are also described.     

Correlation analysis for forced vibration test of the Haulien large scale seismic test program

The correlation analysis for forced vibration test of a 1/4 scale containment SSI test model constructed in Hualien, Taiwan, was carried out for the case of backfilled foundation. Prior to the correlation analysis, structural property was revised so that calculated fundamental frequency of fixed base condition was adjusted to that derived from test results. Correlation analysis was carried out by ‘Lattice Model’ which was able to estimate soil-structure interaction effects with embedment. The analysis results coincide well with test results and concluded that the mathematical soil-structure interaction model established by the correlation analysis is efficient to estimate a dynamic soil-structure interaction effect with embedment. This mathematical model will be applied as a basic model of simulation analysis for earthquake observation records.     

Design and full scale test of the fuel handling system

In the 10 MW High Temperature Gas-cooled Reactor-Test Module (HTR-10) fuel elements move through the core driven by gravity. To reach their design burn-up the fuel elements are re-shuttled five times. This transportation outside the core is mainly achieved pneumatically. Although, adopting the international experience at design and operation of similar systems some key components were improved so that the fuel handling system (FHS) becomes simpler and more reliable. The improved components were tested in full-scale testing facilities. The debugging test and the first loading operation for the FHS indicate that the FHS meets the demands of the HTR-10. In this paper, the functions, design parameters, technological processes, main components and design characteristics of the FHS are described in detail. The flow schemes, design parameters of the full-scale testing facilities and the experimental results are briefly introduced.     

UPTF experiment: a synopsis of full scale test results

A series of 14 tests has been run at UPTF – a 1:1 scale test facility – to investigate the thermohydraulic phenomena in a PWR primary system during blowdown, refill and reflood phases. A synopsis of the most significant test results is given to characterize the controlling phenomena in a full scale primary system under LOCA conditions.     

Progress for on-line acoustic emission monitoring of cracks in reactor systems

This paper reviews accomplishments and planned tasks for the NRC-sponsored research program concerned with “Acoustic Emission/Flaw Relationships for Inservice Monitoring of Nuclear Reactor Pressure Boundaries”. The objective of the acoustic emission (AE) monitoring program is to develop and validate the use of AE methods for continuous surveillance of reactor pressure boundaries to detect flaw growth. Topics discussed include testing AE monitoring on reactors, refinement of an AE signal identification relationship, study of slow crack growth rate effects on AE generation, and activity to produce an ASTM standard for AE monitoring and to gain ASME code acceptance of AE monitoring.     

Creep—fatigue test of a thick-walled vessel under thermal transient loadings

To improve the damage evaluation methods in the design code for Fast Breeder Reactors (FBRs), a series of creep—fatigue tests of structural models under thermal transient loadings are going on at Oarai Engineering Center of the Power Reactor and Nuclear Fuel Development Corporation (PNC). Test models are designed to incorporate representative structures of components and pipings used in FBRs and are subjected to severer cyclic thermal transients than those experienced in FBRs. The test is planned to be continued until failure occurs. This paper describes the creep—fatigue test results and their damage evaluation for the first test model.A 40 mm thick vessel model made of SUS304 austenitic stainless steel was subjected to cyclic thermal transients, in which sodium at 600°C and 250°C flowed repeatedly. The period of each transient was 2 h. Cracks were observed at seven test portions in the model after 1002 cycles of the thermal transients.Elastic and inelastic analyses were performed to evaluate creep—fatigue damage and crack propagation. The safety margins included in the creep—fatigue design methods based on elastic analysis as well as those based on inelastic analysis are discussed. Finally fracture mechanics analyses were performed to explain the observed crack growth.     

Thermal anchoring of wires in large scale superconducting coil test experiment

Effective and precise thermal anchoring of wires in cryogenic experiment is mandatory to measure temperature in milikelvin accuracy and to avoid unnecessary cooling power due to additional heat conduction from room temperature (RT) to operating temperature (OT) through potential, field, displacement and stress measurement instrumentation wires. Instrumentation wires used in large scale superconducting coil test experiments are different compare to cryogenic apparatus in terms of unique construction and overall diameter/area due to errorless measurement in large time-varying magnetic field compare to small cryogenic apparatus, often shielded wires are used. Hence, along with other variables, anchoring techniques and required thermal anchoring length are entirely different in this experiment compare to cryogenic apparatus. In present paper, estimation of thermal anchoring length of five different types of instrumentation wires used in coils test campaign at Institute for Plasma Research (IPR), India has been discussed and some temperature measurement results of coils test campaign have been presented.