IAEA CONTRIBUTION TO ASSESSMENT AND MANAGEMENT OF STEAM GENERATOR AGEING

Transcription

1 IAEA CONTRIBUTION TO ASSESSMENT AND MANAGEMENT OF STEAM GENERATOR AGEING Introduction L. Kupca, K. S. Kang, International Atomic Energy Agency, Wagramer Strasse 5, PO Box 100, 1400 Wien, Austria Ageing management of main primary components of a nuclear power plant has been in major focus for many years in the past, recent and current IAEA efforts and activities. The objective of the efforts to update the IAEA-TECDOC-981 Assessment and management of ageing of major nuclear power plant components important to safety: Steam generators with the latest relevant information to reflect the state-of-the-art situation and to produce a publication which addresses PWR, WWER and CANDU type of plants/reactors. The steam generators in the pressurized water reactor (PWR), Canada deuterium-uranium (CANDU) reactor, and Russian water moderated, water cooled energy reactor (WWER) plants are large heat exchangers that use the heat from the primary reactor coolant to make steam in the secondary side to drive turbine generators. A typical plant has two to six steam generators per reactor. The steam generators are shell-and-tube heat exchangers each with several thousands of tubes. The primary reactor coolant passes through the tubes and boils water on the outside of the tubes (secondary side) to make steam. The design confines radioactivity from neutron activation or fission products to the primary coolant during normal operation. The objective of the IAEA-TECDOC publication, which will be published as an update of IAEA-TECDOC-981 Assessment and management of ageing of major nuclear power plant components important to safety: Steam generators, is to document current practices for the assessment and management of ageing of the following types of steam generators used in water cooled nuclear power plants: Vertical tubesheet boiling steam generators, commonly known as "recirculating vertical U- tube steam generators;" Vertical/tubesheet super heated steam generators, commonly known as "once-through steam generators," and Horizontal/collector boiling steam generators used in WWER reactors. The TECDOC emphasizes safety and engineering aspects and also provides information on ageing degradation, current inspection, monitoring and maintenance practices for managing ageing of steam generators. The underlying objective of this IAEA-TECDOC is to ensure that the information on the current assessment methods and ageing management techniques is available to all involved, in the operation of nuclear power plants in the IAEA Member States. The steam generator subcomponents discussed in the publication are those susceptible to ageing damage and whose consequence of failure has a significant safety impact: the steam generator tubes tubesheets or collectors plugs (tube and tubesheet) 1

2 sleeves. These are subcomponents whose failure impairs the primary to secondary pressure boundary. In addition, the IAEA-TECDOC also discusses two other subcomponents: feedwater nozzles, and shell girth welds. These subcomponents have experienced significant degradation in some plants but their failure is a secondary side pressure boundary failure (affects conventional safety) and does not immediately lead to any release of radioactivity. In the IAEA-TECDOC, which is currently being updated, are discussed: steam generator designs design bases for the components of interest to steam generator ageing stressors, susceptible sites, and failure modes associated with the various steam generator degradation mechanisms, including detrimental effect of: primary water outside diameter transgranular stress corrosion cracking fretting wear thinning pitting denting high-cycle fatigue wastage erosion-corrosion corrosion-fatigue. operational guidelines aimed at preventing or minimizing age related degradation of steam generators are discussed tubing inspection requirements and technologies fitness-for-service guidelines in various countries. mitigation, repair, and replacement technologies guidelines for a systematic steam generator ageing management programme. Steam generator designs Different designs for currently operating steam generators are described. Recirculating steam generators designed by Westinghouse (USA), Combustion Engineering (USA), Framatome/AREVA NP (France), Mitsubishi Heavy Industries (Japan), Siemens(Germany), Doosan(Korea). The Canadian designs are discussed next with the Babcock & Wilcox (USA) once-through steam generator design and the Russian (WWER) designs completing the section. 2

3 Emphasis is placed on the design aspects and fabrication methods which may affect steam generator degradation. Fig. 1. Example of PWR Recirculating Steam Generator (Doosan, Korea) Steam generator design basis, fabrication and materials Most of the information concerns the heat exchanger tubing, including its fabrication, materials of construction and installation into the tube sheets (or collectors). The tube support, feedwater nozzle and steam generator shell designs and materials are also discussed. Chemical composition of used steels in steam generators is given in Table 1. 3

4 Table 1. Chemical composition of used steels and alloys in steam generators 4 Materials Content of elements [mass %] C Mn Si P S Cr Ni Mo W Ti V Cu Co Al Fe 08Ch18N10T 0,08 2,0 0,8 0,035 0, GN2MFA 22K 0,08 0,15 0,19 0,26 0,60 0,90 0,1 0,37 0,025 0,025 0,30 1,70 2,70 0,40 0,60 5xC 0, ,75 1,0 0,2 0,4 0,030 0,030 0,3 0, , ,08 1,0 1,0 0,040 0,030 11,5 14,5 ChN35VT-VD 0,12 1,0 2,0 0,60 0,025 0, Incoloy Alloy 800 0,10 1,5 1,0-0,015 Inconel 600 0,15 1,00 0,50 0,020 0,015 Inconel 690 0,05 0,50 0,50-0,015 19,0 23,0 14,0 17,0 27,0 31, ,0 36,0 30,0 35,0 0,10 0,30-2,8 3,5 1,1 1, ,15 0,60-0,75-72, , , , Monel 400 0,30 2,0 0,50-0,024-63, SA-106 Grade B 0,30 SA-508 Grade 1 0,35 SA-533 Grade B 0,20 SA-533 Type A Class 1 0,25 0,29 1,06 0,40 1,05 1,15 1,50 1,15 1,50 0,10 0,40 28,0 34,0 0,15 0,60-39,5 6,0 10,0 7,0 11,0 1,0-2,50 0,025 0,025 0,40 0,40 0, ,08 0, ,40 0,025 0,025 0,25 0,40 0, ,05 0,20-0,025-0,15 0,40 0,15 0,40 0,035 0, ,035 0, ,45 0,60 0,45 0,

5 Materials and methods used to fabricate steam generator components significantly affect their susceptibility to corrosion, especially to stress corrosion cracking. Degradation of the steam generator tubing is also influenced by other aspects of the steam generator design and construction, such as the tube support design and the method of tube installation. Also the design of the balance of plant systems and components have a decisive influence in the corrosion behaviour of the steam generator. Steam generator degradation mechanisms The steam generator (SG) tubing corrosion depends of the simultaneous influence of three factors: SG and balance of plant (BOP) system design SG and balance of plants materials Chemistry Fig. 2. Factors affecting corrosion The most frequent cases of SG degradation are due to incompatibility between these three factors. Plants having a good design and material concept can afford a less restrictive chemistry. It is not possible to speak about secondary side chemistry without establishing how the plant design is and with which materials it has to be confronted. In the same way, the behaviour of a certain material cannot be assessed without considering the chemistry applied and in which design this material is inserted. A direct consequence of having an adequate design and material concept is the possibility of having more relaxed, less stringent chemistry. An inadequate SG design having closed, narrow geometries, will cause corrosion products to accumulate. The section in the IAEA-TECDOC discusses: the main reasons of SG degradation, including SG and BOP design, materials and chemistry 5

6 susceptible sites and failure modes associated with the various steam generator degradation mechanisms. Summary of PWR recirculating steam generator tube degradation processes is shown in Table 2. Steam generator ageing management: operational guidelines The operational procedures are grouped into the following topic areas: Primary coolant system water chemistry control parameters. Secondary coolant system water chemistry control and diagnostic parameters Control of corrosion product ingress into the SGs Control of SG reducing conditions Control of impurity ingress into the SGs Condenser integrity Parameters of condensate polishing system Recycle of blowdown water Control of lead contamination Removal of copper from secondary coolant system Measures to control steam generator deposits Reduction of turbine/condenser/steam air in-leakage Steam generator lay-up Steam generator blowdown Balance of plant corrosion Water lancing Pressure pulse and water slap Measures to ensure reducing conditions Measures to control secondary-side impurity incursions Measures to remove secondary-side impurities The section in the TECDOC describes a set of operational guidelines which will help nuclear power plant operators prevent, or at least minimize, steam generator ageing degradation and thereby maximize component life. Implementation of these measures is expected to be considerably less costly than repairing or replacing steam generators and may provide the additional assurance required to operate some steam generators for additional time. 6

7 Table 2. Summary of PWR recirculating steam generator tube degradation processes Rank Degradation Mechanism 1 ODSCC Tensile stresses, impurity concentrations, sensitive materials Stressor Degradation Sites Potential Failure Mode ISI Method 2 PWSCC Temperature, residual tensile stresses, sensitive materials (low mill anneal temperature) 3 Fretting, Wear 4 High-cycle fatigue Flow induced vibration, aggressive chemicals High mean stress level and flow induced vibration, initiating defect (crack, dent, pit, etc) 5 Denting Oxygen, copper oxide, chlorides, temperature, ph, crevice condition, deposits 6 Pitting Brackish water, chlorides, sulphates, oxygen, copper oxides 7 Wastage Phosphate chemistry, chloride concentration, resin leakage Tube-to-tubesheet crevices Sludge pile Tube support late Free span Inside surface of U-bend Roll transition w/o kiss rolling Roll transition with kiss rolling Dented tube regions Contact points between tubes and the antivibration bars (AVBs), or tubes and the preheater baffles Contact between tubes and loose parts Tube-to-tube contact At the upper support late if the tube is clamped At the tube support plates, in the sludge pile, in the tubesheet crevices Cold leg in sludge pile or where scale containing copper deposits is found, under deposit pitting in hot leg Axial or circumferential crack Circumferential crack Axial crack Axial crack Mixed Crack Mixed Crack Axial Crack Circumferential Crack Local wear Depends on loose part geometry Axial Wear Transgranular circumferential cracking Flow blockage in tube, may lead to circumferential cracking (see PWSCC), decreases the fatigue resistance Local attack and tube thinning, may lead to a hole MRPC MRPC/Cecco5 Bobbin coil/cecco 5 Bobbin coil (in absolute mode) MRPC MRPC MRPC Bobbin coil or MRPC Bobbin coil Bobbin coil Bobbin coil Leak detection or by detection of precursor Profilometry, bobbin coil Bobbin coil, ultrasonics Tubesheet crevices, sludge pile, tube support plates, AVBs General thinning Bobbin coil 7

8 Steam generator inspection and monitoring requirements and technologies This section in the IAEA-TECDOC, which is currently being updated, identifies inspection and monitoring requirements and techniques for steam generators, with emphasis on examining the tubing, feedwater nozzle and shell welds. The probability and consequences of steam generator tube failures can be reduced through appropriate and timely inspections. Tubing inspection practices in the United States, Canada, China, the Czech Republic, France, Germany, Japan, South Korea, Russia, Slovakia, Slovenia, Spain, Sweden, and Switzerland are summarized in Table 3. Table 3. Steam generator tubing inspection requirements USA Canada Czech Republic France Baseline Inspection Number of Tubes to be Inspected Inspection Intervals All tubes prior to service and after any major change in secondary water chemistry 25% of the tubes prior to service All tubes prior to service All tubes prior to service All tubes every ten years (1st after 30 months) First inspection, 3% of the total steam generator tubes at a unit Subsequent inspections, see Table 6.2 Suggestion :All American plants follow the EPRI Guidelines for Examination of SG tubes maybe the document could be mentioned as a reference and used for definition of number of tubes At least 10% of the tubes in one steam generator per unit At least 10% of the tubes in each steam generator must be inspected full length Usually inspect all the tubes from the not collector and 50% of the tubes from the cold collector If susceptible tubing all of the tubes are inspected in the hot leg roll transition, tube support plate and sludge pile regions, and the U-bend region of the first row in service, with an appropriate probe. If less susceptible tubing: Sample of tubes inspected full length. All tubes in service with a previous defect indication. First inspection, 6-24 months Subsequent inspections, months If less than 5% of inspected tubes with indications and no defective tubes, 40 months If more than 10% degraded and more than 1% Every 5 years Every four years Every outage for roll transition and small radius U-bend regions Every other outage for TSP and sludge pile regions Sample every two years Each outage Germany All tubes prior to service 10% of the tubes per steam generator per inspection Every five years all steam generators Every two years, one half of the steam generators Japan All tubes prior to service Insertion depth of antivibration bars If no leakage and no defects: 30% If any leakage or defects 100% If no leakage and no defects, every other year If leakage or defects, every year Korea All tubes prior to service 20 % of total number of tubes each SG If a Potential Degradation area (expansion region, U-bend, Dent/Ding or inside the tubesheet) is not verified by the bobbin probe, an additional inspection by RPC is required. The periodical inspection depends on materials (alloy 600, alloy 690), operating years, degradation status. Each refueling outage 8

9 Slovakia All tubes prior to service At least 10% of the tubes m each steam generator must be inspected full length Usually inspect all the tubes from the both hot collectors Slovenia All tubes prior to service 100% using bobbin coil and all reported indications, roll transitions and inner bends with pancake coil. I think that they have replaced SGs in 2000 and changed the scope of inspection Every four years Each refueling outage Spain All tubes prior to service If susceptible tubing: 100% using bobbin coil and all indications and roll transition legions with rotating pancake coil If less susceptible tubing 9 to 20% Each refueling outage Sweden All tubes prior to service Random sample of 15-17% full length 100% hot leg tubesheet % of other selected regions Each year Switzerland All tubes after one year of operation If susceptible tubing: inspect the hot leg side up through the U-bend region to the top tube support plate on the cold side -full inspection If less susceptible tubing: random sample of 5.5% of all tubes Every outage Every three years Every three years Steam generator fitness-for-service assessment Number of countries have found the original ASME criterion overly conservative and inflexible and have developed revised or new fitness-for-service criteria, often in conjunction with revised inspection requirements. Although the new fitness-for-service criteria used in most countries follow the general technical basis, there are substantial differences in implementation. Currently implemented repair criteria can be grouped into two families: generic and defect type and location specific criteria. The objectives of the steam generator fitness-for-service assessment include: 1. Identification and characterization of material degradation within steam generators. 2. Implementation of steam generator inspection program to provide sufficient information concerning specific degradations present in the steam generators, and 3. Application of steam generator tube fitness-for-service assessment methods to evaluate condition of steam generators at the end of an inspection interval and to ensure integrity during the subsequent operation period. Successful implementation of the above objectives should provide reasonable assurance that steam generator integrity is being maintained consistent with the licensing basis. The steam generator integrity assessment consists of three key elements: degradation assessment, condition monitoring, and operational assessment, as shown on Figure 3. The eye of steam generator integrity assessment represents the integration of critical elements of steam generator fitness-for-service assessment. 9

10 Fig. 3. The Eye of Steam Generator Fitness-for-Service Assessment The purpose of the degradation assessment is to ensure that appropriate inspections are planned for the upcoming outage, by identifying all active degradation mechanisms and for each mechanism: choose NDT method and technique or techniques to test for degradation based on the probability of detection and sizing capability; establish the inspection sample population size (number of tubes to be inspected); establish the extent of inspection (exact location of tubes, section, row, column, to be inspected); establish of the part(s) of the tube for confirmatory examination with advanced techniques; establish the structural limits; and establish the flaw growth rate or a plan to establish the flaw growth rate, establish inspection expansion criteria if new degradation is identified, or if the measured parameters change, such as degradation growth rate. Key elements of a condition monitoring accident leakage assessment by analysis should include the following for each flaw type: The as-found frequency distribution of indications for each active flaw type is established as a function of indication size. The distribution should be adjusted statistically to consider the percentage of tubes sampled to address the subject flaw type. Models relating the magnitude of leakage rate as a function of actual flaw size or eddy current indication size measurement for each flaw mechanism are established. The leakage calculation for each flaw and for total steam generator leakage rate is performed deterministically or probabilistically (e.g., with statistical sampling methods such as Monte Carlo), accounting for all significant uncertainties. Potential sources of uncertainty include eddy current indication size measurement error or variability, material properties, and leakage models. Leakage models may be empirical or analytical (i.e., idealized models based on engineering mechanics). 10

11 Considerations for performing the operational assessment against the probabilistic performance criteria structural integrity should include the following for a given flaw type. The probabilistic approach should only be used when in-service inspection techniques and personnel are validated for detection and sizing, The calculation of the frequency distribution of flaws or indications should be by the size projected to exist immediately prior to the next scheduled inspection. The specific details for projecting the distribution of flaw or indication sizes are to be developed by plant operators. The performance of the predictive methodology that projects a distribution that results in a conservative estimate of conditional probability of burst should be evaluated based on the results of future in-service inspections and appropriate adjustments made to the methodology as necessary to ensure this objective is met. The empirical burst pressure should be established as a function of flaw or indication size. These empirical models should account for data scatter and model parameter uncertainties, The projected distribution of flaw or indication sizes, the calculated frequency of burst, and the calculated conditional probability of burst during postulated accidents should include a rigorous statistical treatment of all significant sources of uncertainty affecting the calculation, including growth rate, indication size measurement, and burst-pressure model. Statistical sampling methods such as Monte Carlo may be used. The frequency and conditional probability of burst should be evaluated at the one-sided, upper 95% confidence level. Steam generator maintenance: mitigation, repair and replacement The section in ther IAEA-TECDOC discusses mitigation and repair techniques for degradation mechanisms in tubes, tubesheets, feedwater nozzles and shells. Table 4 summarizes countermeasures for tube failures in PWR steam generators. Several modifications in the design of the feedwater system have also been made in the existing or new steam generators minimize or prevent thermal fatigue damage to the feedwater piping and nozzles. Some of these modifications have been also employed for repairing the thermal fatigue damage as discussed in the next section. The modifications include: Replacing the sharp counterbore with a blend radius to reduce stress concentrations. Installing a separate nozzle for injecting the auxiliary feedwater directly into the steam generator. Use of a spraying device located in the feedwater piping upstream of the nozzle to mix the cold auxiliary feedwater with the hot water in the pipe. Installing a long thermal sleeve to protect the feedwater piping from thermal stresses and fatigue damage induced by thermal stratification. Welding the thermal sleeve to the feedwater nozzle to reduce stratification in the annulus region between nozzle and thermal sleeve. 11

12 Use of a destratification loop in the feedwater piping (either just inside or just outside the steam generator shell). Table 4. Countermeasures for tube failures in PWR steam generators Mechanism Mitigation of damage in existing tubes Improvements in new/replaced steam generators Primary side stress corrosion cracking Intergranular stress corrosion cracking, intergranular attack Pitting Denting Wastage High cycle fatigue and fretting Erosioncorrosion and corrosion fatigue in oncethrough steam generators Rotopeening or shot peening to residual stresses, stress relieving of the U-bends and control of the denting problem. Control of the alkaline impurities, chlorides, sulphates, and carbonates; flush tubesheet crevices; use of hot soak, sludge lancing, and chemical cleaning; neutralization of crevice alkalinity; addition of boric acid; and fulldepth roll expansion of tubes to eliminate crevices. Elimination of condenser leakages and ingress of air/oxygen, chlorides, and sulphates; removal of copper from the feedwater train. Elimination of condenser leakages and ingress of air/oxygen and chlorides; use of hot soaks; removal of copper from the feedwater train. Use of all-volatile treatment water chemistry; elimination of hideout chemical concentrations; use of sludge lancing, chemical cleaning, hot soaks, and hot blowdowns and flushing; preclusion of resin ingress. Control of the chemistry and entrained solid particle content of the secondary side coolant Alloy 690 tubes with an optimum strength of about 380 MPa, little or no residual stresses. Alloy 690 tubes with an optimum microstructure, no tubesheet crevices, improved access for lancing and cleaning, increased blowdown capacity, and flow patterns that minimize sludge accumulation. Titanium or stainless steel condenser tubes, no copper alloys in the feed train, and corrosion resistant tube materials (Alloy 690) Strict water chemistry control, stainless steel support structures, support plates that preclude stagnant water in the annuli, and titanium condenser tubes Flow patterns that minimize hide-out and chemical concentrations and sludge formation; improved access for cleaning; increased blow-down capacity. Additional antivibration bars (AVBs) and insertion of the AVBs deeper into the bundle; minimum tube-to-avb clearances and wear matching of the AVBs to the tubes. 12

13 Steam generator ageing management programme Existing programmes relating to the management of steam generator ageing include operations, surveillance and maintenance programmes as well as operating experience feedback, research and development and technical support programmes. Safety authorities increasingly require licensees to define ageing management programmes (AMPs) for selected systems, structures and components by documenting relevant programmes and activities and their respective roles in managing SSC ageing. A definition of a steam generator AMP includes also a description of mechanisms used for programme co-ordination and continuous improvement. The continuous AMP improvement or optimization is based on current understanding of steam generator ageing and on results of self-assessments and peer reviews. A systematic evaluation of the ageing management requirements for steam generators should be performed in order to acquire information and knowledge about the following four elements: 1. Understanding ageing 2. Prevention of ageing 3. Detection and monitoring of ageing 4. Mitigation of ageing effects. A recommended methodology, which consists of the evaluation of relevant information and documentation of the findings, is illustrated in a flowchart shown in Figure 4. The results of operating experience, research and development, and available previous ageing evaluations (both generic and plant-specific) should be used in the evaluations. Relevant applicable ageing management reviews (e.g., prepared by the vendor/ owners group, suppliers or technical support organizations) should be used to minimize duplication of effort, if available. At the initiation of the AMP, and at defined intervals, the utilities should perform an assessment of the actual condition of steam generators to: 1. Determine the current performance and condition of the steam generators, including the assessment of any age-related failures or indications of significant material degradation; 2. Compare the current performance and condition against predictions for the identified ageing mechanisms and acceptance criteria; 3. Based on current performance and conditions, predict the future performance, ageing degradation, and, if possible, the residual service life of the steam generator (i.e., the length of time the steam generator is likely to meet its function and performance requirements); and 4. Determine whether the ageing degradation assumptions made regarding the design of the SG remain valid, and recommend appropriate follow-up corrective actions and preventive measures for input to the AMP. 13

14 Design and specifications Materials and material properties Service conditions Steam Generator Understanding of Ageing Performance requirements Operation and maintenance histories Generic operating experience Relevant R&D results Documentation of: 1) Current understanding of steam generators ageing (e.g., ageing mechanisms and effects, sites of degradation, any analytical/empirical models for predicting SG degradation, any gaps in the understanding of ageing); 2) Acceptance criteria including applicable regulatory or code requirements, and safety analysis limits & conditions; 3) List of data requirements for the assessment of SG ageing (including any deficiencies in the availability and quality of existing records). Prevention of Ageing Evaluation of the effectiveness of methods and practices for prevention of SG ageing degradation. Documentation of the information, including: 4) Design, materials, fabrication and construction, operations and maintenance methods and practices to prevent ageing degradation; 5) Operating conditions and practices that prevent or minimize the rate of ageing degradation. Monitoring of Ageing Evaluation of monitoring methods, taking into account relevant operating experience and research results. Documentation of the information, including: 6) Functional parameters and condition indicators for detecting, monitoring, and trending ageing degradation; 7) An assessment of the capability and practicability of existing monitoring techniques to measure these parameters and indicators with sufficient sensitivity, reliability, and accuracy; 8) Data evaluation techniques for recognizing significant degradation and for predicting future performance of SGs. Mitigation of Ageing Evaluation of the effectiveness of existing methods and practices for mitigating SG s ageing degradation. Documentation of the information, including: 9) Maintenance methods and practices, condition monitoring (including refurbishment and periodic replacement of parts and consumables) to control ageing degradation; 10) Operating conditions and practices that minimize the rate of SG s ageing degradation; 11) Possible modifications to design and materials of the SG to control ageing degradation. Report on Ageing Management Review Steam generators -specific information on understanding, monitoring, and mitigating ageing. Recommendations for the application of results of the ageing management review in plant design, operation and maintenance, and for R&D to address gaps in knowledge and technology. Fig. 4. A sample methodology for ageing evaluation, including review and evaluation of relevant information and documentation of the findings (adapted from IAEA NS-G-2.12 [2]). 14

15 References [1] INTERNATIONAL ATOMIC ENERGY AGENCY, Draft on Assessment and management of ageing of major nuclear power plant components important to safety: Steam generators, IAEA-TECDOC-xxxx, IAEA, Vienna, to be published in [2] INTERNATIONAL ATOMIC ENERGY AGENCY, Ageing Management for Nuclear Power Plants, Safety Standards Series No. NS-G-2.12, IAEA, Vienna (2008). [3] INTERNATIONAL ATOMIC ENERGY AGENCY, Strategy for Assessment of WWER Steam Generator Tube Integrity, IAEA-TECDOC-1577, IAEA, Vienna (2007). 15