An aging management approach for pressurized water reactor pressure vessels

Int. J. Pres. Ves. & Piping 54 (1993) 317-340

An Aging Management Approach for Pressurized Water Reactor Pressure Vessels* V. N. S h a h Principal Engineer, Idaho National Engineering Laboratory, PO Box 1625, Idaho Fails, Idaho, USA 83415-2409

& W . L. S e r v e r Vice President, ATI Consulting, Suite 140, San Ramon, California, USA 94583-1344

ABSTRACT This paper presents a generic approach, based on known and established results of aging damage assessments, for evaluating structural integrity and managing aging damage of pressurized water reactor (PWR) pressure vessels during extended operation. The approach addresses specific technical and regulatory issues that will arise in the assessment of aging damage during extended operation. Three major issues related to vessel integrity are addressed: pressurized thermal shock events, low upper-shelf energy, and revised pressuretemperature limits. Finally, the approach identifies several options available to manage aging. These options include use of advanced inspection methods to reliably detect and size flaws, flux reduction and thermal annealing to mitigate radiation embrittlement, and supplemental surveillance and material sampling to enhance the database for irradiated reactor pressure vessel materials. 1 INTRODUCTION The potential problems of assessing and managing aging in the older nuclear p o w e r plants have b e c o m e a m a j o r focus of the research sponsored by the U S Nuclear R e g u l a t o r y Commission ( U S N R C ) . A n * Work supported by the U.S. Nuclear Regulatory Commission under DOE Idaho Field Office Contract DE-ACO7-761D1570. 317 Int. J. Pres. Ves. & Piping 0308-0161/93/$06.00 ~ 1993 Elsevier Science Publishers Ltd, England. Printed in Northern Ireland

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important part of the USNRC research effort is the Nuclear Plant Aging Research (NPAR) Program being conducted at several national laboratories, including the Idaho National Engineering Laboratory (INEL). ~ One of the NPAR Program tasks at the INEL is to develop aging management approaches that will provide comprehensive understanding of aging damage to the major light water reactor (LWR) components and that will evaluate current and emerging methods to quantitatively estimate and manage aging damage. The major LWR components include the primary pressure boundary components, feedwater and steam lines, reactor pressure vessel internals and supports, safety-related cables and connections, emergency diesel generators, and primary containments. 2 Aging management approaches for five major components are being developed: PWR pressure vessels, reinforced concrete containments, cast stainless steel components, PWR steam generator tubes, and metal containments. This paper presents a generic approach for aging management of PWR pressure vessels. Figure 1 presents a ten-step approach to managing the aging of the PWR pressure vessels. The approach can be divided into four parts: Part I, estimating current damage (Steps 1 through 5); Part II, assessing 1 Review Design and Fabrication Records 1.1 Design configuration 1.1.1 Specifications 1.1.2 As-built dimensions and drawings 1.1.3 Applicable version of ASME Code 1.1.4 Weld locations 1.1.5 Core/flux design 1.1.6 Stress report 1.2 Fabrication history 1.2.1 Fabricator(s) 1.2.2 Fabrication drawings and procedures 1.2.3 Weld qualification and data 1.2.4 Thermo-mechanical treatment (including postweld heat treatment) 1.2.5 Repair maps 1.2.6 1.2.7

Cladding

2.2.2 2.2.3 2.2.4

Weld wire and flux As-deposited weld Cladding

Quality assurance records Supplemental stress analyses 2 Review Original Material Data for Base Metal, Weld Filler and Cladding 2.1 Specifications 2.2 Chemical composition 2.2.1 Base metal 1.3

2.3

Unirradiated mechanical properties Reference nil-ductility-transition temperature (RTNoT)

2.3.1 Fig. 1.

A n aging management approach for pressurized water reactor pressure vessels.

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Charpy V-notch impact results --estimate upper shelf energy (USE) --temperature at 41-J (30 ft Ib) Charpy V-notch energy (T30) 2.3.3 Tensile test results 2.3.4 Fracture toughness (if available) 2.4 Techniques to acquire missing/supplemental information 2.4.1 Material sampling 2.4.2 Archival and surrogate materials 2.4.3 Small specimen testing and verification 2.4.4 In-situ measurements of materials properties 2.4.5 Estimation of mechanical properties 3 Review Surveillance Program 3.1 Capsules 3.1.1 Number, types and locations 3.1.2 Lead factors of capsules 3.1.3 Number, types and locations of test specimens 3.1.4 Test specimen materials --limiting materials ----correlation monitor materials -----other materials 3.1.5 Correlation monitor materials 3.1.6 Temperature and fluence monitors 3.1.7 Withdrawal schedule 3.1.8 Comparing surveillance results with RG 1.99, Rev. 2 predictions 3.2 Supplemental surveillance program 3.2.1 In- and ¢x-vessel dosimetry 3.2.2 Temperature verification 3.2.3 Reinsertion capsules 3.2.4 Additional capsules 3.2.5 Archival, reconstituted, or surrogate materials 3.2.6 Thermal anneal considerations 3.2.7 Integrated program with other nuclear plants 3.2.8 Fluence projections (including through-wall) 4 Review In-service Inspection (ISI) Records 4.1 Construction radiographs and pre-service inspection records 4.2 Description of past ISI results 4.2.1 Location of inspected regions 4.2.2 Documentation of inspection methods used 4.2.3 Size, orientation, and distribution of detected indications 4.2.4 Use of advanced inspection methods beyond those required 4.2.5 Maximum undetected flaw size and sizing accuracy of inspection methods used 4.3 Disposition of detected flaws 4.3.1 Stress analysis 4.3.2 Minimum toughness at end-of-life 4.3.3 Fatigue crack growth evaluation 4.3.4 Pressurized thermal shock analysis 4.3.5 In-service repair 4.4 Augmented inspection and monitoring 4.4.1 Flaw size larger than ASME Code acceptance standards 4.4.2 Increased inspection frequency and additional inspections 4.4.3 Inaccessible inspection sites

Fig. 1,--cont.

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5 Review Operating History 5.1 Initial and updated pressure-temperature limits 5.1.1 Technical specifications related to P - T curves 5.1.2 Low-temperature-overpressure protection (LTOP) set point(s) 5.1.3 Definition of administrative controls 5.2 Pressure-temperature excursions 5.2.1 Overcooling events 5.2.2 Actual cooldown and heatup rates 5.2.3 P - T limit violations 5.2.4 Overpressurization events 5.3 Other operating transients 5.3.1 Hydrotests and leak tests 5.3.2 Actual versus design basis transients 5.4 Temperature history 5.5 Hours of operation 5.6 Neutron flux and core changes 5.6.1 Low leakage core 5.6.2 Other flux reduction measures 5.6.3 Structural changes in vessel internals 5.7 Updated RTNDT and USE 5.8 In-service damage and repairs 5.9 Plant operating procedures and changes in these procedures 5.9.1 Normal operating procedures --heatup and cooldown procedures --in-servitSe test procedures 5.9.2 Abnormal and emergency operating procedures 6 Assess Radiation Embrittlement Damage at the End of Next Operating Period (EONOP) 6.1 Estimate total fluence ( E > I M e V ) at EONOP at sites of limitng materials and weldments in beltline region (use RG 1.99, Rev. 2 attenuation model) 6.1.1 Vessel inside surface 6.1.2 1 thickness 6.1.3 3 thickness 6.1.4 Vessel outside surface 6.2 Estimate shift in RTNDT at EONOP using RG 1.99, Rev. 2 at sites identified in 6.1 6.3 Estimate adjusted RTNDT, RTvr s and USE at EONOP using RG 1.99, Rev. 2 and 10 CFR 50.61 at sites identified in 6.1 6.4 Revise P-Tlimits based on adjusted RTNo T at EONOP 6.5 Estimate fracture toughness (Krc and K~a) at EONOP at sites identified in 6.1 6.5.1 Shift in ASME Section XI fracture toughness curves 6.5.2 Upper shelf fracture toughness 6.5.3 Surveillance fracture toughness results if specimens are included in the surveillance capsules 6.6 Measure current fracture toughness, if needed 6.6.1 In situ measurements 6.6.2 Testing small samples removed benignly from vessel 7 Assess Fatigue Damage at the End of Next Operating Period 7.1 Identify size and locations of existing flaws from ISI records 7.2 Estimate simplified conservative loadings from review of operating history 7.3 Perform stress analysis 7.3.1 Identify regions of highest stress 7.3.2 Estimate stress history at known or postulated flaws 7.4 Update ASME Section III cumulative fatigue usage factors, if design basis has been violated Fig. L--cont.

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Compute growth of known and postulated flaws during next operating period 7.5.1 Cycle-dependent model (ASME Section XI) 7.5.2 Time-dependent model 8 Evaluate Vessel Integrity at the End of Next Operating Period 8.1 Pressurized thermal shock (PTS) 8.1.1 Estimate RTvrs for axial welds, circumferential welds and base metal 8.1.2 Probabilistic fracture mechanics analysis according to R G 1.154, if PTS criteria are violated --RTvrs for axial welds and base metal greater than 132 °C (270 °F) --RTvrs for circumferential welds greater than 149 °C (300 °F) 8.2 Revised pressure-temperature ( P - T ) limits 8.2.1 Ensure that the plant can be operated safely --incorporate appropriate LTOP set points 8.2.2 Minor violations of P - T limits - - A S M E Section XI, Appendix E screening criteria 8.3 Upper shelf energy at EONOP 8.3.1 Estimate USE --weldments --plate material --forging 8.3.2 Perform elastic-plastic fracture mechanics evaluation if USE is less than 68J (50 ft-lb) 8.3.3 Consider interaction between PTS event and low upper shelf toughness at EONOP if warranted 9 Establish Actions to be Taken 9.1 Continued operation to EONOP if acceptable safety margins on RTI~DTand USE exist 9.2 Physical modifications to the plant 9.2.1 Flux reduction 9.2.2 Thermal annealing -----evaluate re-embrittlement rates --implement a modified surveillance program 9.2.3 Augmented heating of safety injection water 9.3 Modified operating procedures 9.4 Revised operator training 9.5 Supplemental surveillance 9.5.1 Mechanical and chemical properties 9.5.2 Radiation environment 9.6 Enhanced ISI 9.6.1 Improved detection and sizing capability --automatic ultrasonic inspection methods --focused transducer --time-of-flight diffraction method 9.6.2 Acoustic emission technique to monitor crack growth in inaccessible regions 9.6.3 More frequent ISI if warranted 9.7 Weld repair if flaws of unacceptable size are expected 10 Re-evaluate Vessel Integrity at EONOP if Significant Changes are Recommended in Step 9 10.1 Update information in any of Steps 1 through 5 10.2 Return to Steps 6 and 7 to estimate damage at EONOP 10.3 Follow Step 8 to evaluate vessel integrity 10.4 Follow Step 9 to re-evaluate the actions to be taken

Fig. 1.---cont.

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the damage expected during the next operating period (Steps 6 and 7); Part III, evaluating vessel integrity at the end of the next operating period (Step 8); and Part IV, identifying the options available to manage aging during the next operating period (Step 9). 3 The paper identifies the important concerns that need to be addressed in each step of the approach. It then presents several conclusions for reliable evaluation of vessel integrity and for managing vessel aging damage. 2 ESTIMATING C U R R E N T D A M A G E (PART I) Part 1 (Steps 1 through 5) of the approach identifies the necessary information that needs to be reviewed for estimating current damage, including a thorough review and documentation of design and fabrication information, material data, operating history, in-service inspection results and surveillance program records.

2.1 Review design and fabrication records (Step 1) To assess reactor pressure vessel aging damage, it is first necessary to establish the baseline by which the vessel was designed and built. This requires a review of design and fabrication records which include the information used to originally license the plant. Depending on the plant and the contractual arrangements made with the nuclear steam supply system (NSSS) vendor, some of this detailed information may be scattered between the utility, the NSSS vendor, and the vessel manufacturer (if other than the NSSS vendor). The most important design and construction details are the final as-built dimensions, including cladding thicknesses, exact locations of welds, basic design assumptions (transients, fatigue analyses, temperature, etc.), and fabrication procedures and verification records (heat treatments, material formings, and welding and cladding applications). Cladding thickness is an important detail because it controls the heat transfer to base metal during the pressurized thermal shock event; thicker cladding reduces the magnitude of the thermal shock on the base metal. Typical cladding thickness varies between 3-29 to 9-5mm (0-1295 to 0-375in). The minimum cladding thickness for US PWR vessels is 2-77 mm (0.109 in.) Cladding over the vessel inner surface was applied in one or two layers by multiple- or single-wire process or by strip-cladding. Another important detail is whether a given vessel is of ASME Section VIII design or Section III design (with the applicable Code version); Section VIII vessels generally have thicker walls and larger nozzle corner radii because of lower allowable stresses. The design of the core and its projected neutron flux distributions are important for estimating

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accumulated fluence at critical locations at the end of the license period. Alterations to the core design might have been necessary to reduce radiation embrittlement effects. Information on repairs to welds should be collected if available; weld repair information could prove useful if any defects (NDE indications) were discovered during an in-service inspection. Weld repairs are common during fabrication and involve removing bad portions of welds (discovered by review of radiographs), with subsequent weld repairs made generally using a shielded metal arc welding process. Typically, the properties of a repaired portion are as good or better than those of an original, sound weldment. Information on supplemental stress analyses also needs to be collected; such analyses were performed to ensure that early production vessels complied with Section III of the ASME Code. Information on other similar design verification checks that may not have been performed by the NSSS vendor also needs to be collected.

2.2 Review original material data (Step 2) As in Step 1, much of the materials information may be scattered among several sources. A detailed review of all of the materials information and test data are necessary, including any additional work performed to evaluate the vessel materials and increase the quantity or quality of available materials property data. Of course, the chemical analyses for copper, nickel, phosphorus and sulfur are very important, as is the distribution of these elements (if extensive metallography has been performed). For example, the distribution of sulfur as sulfides may be important for fatigue crack growth considerations. Complete details on the weld wires, flux types and as-deposited weld properties are needed because the welds are most likely the most limiting materials of the vessel. Details on baseline mechanical properties are essential; they are the basis upon which radiation embrittlement effects are judged. Ideally, fracture toughness data would be the most helpful, but generally the data are restricted to Charpy V-notch impact energy, tensile and drop-weight nil-ductility-transition temperature (NDT-F) test results (and not all of these results are available for all of the vessel materials, especially for older vessels); the requirements for measuring unirradiated reference nil-ductility transition temperature (RTNDT) were added in the 1972 Addenda to Section III. In cases where plant-specific data were not generated, typical and bounding data from past and ongoing research projects can sometimes be used. Another approach is to use material sampling techniques, 4'5 subsize test specimens, 6'7 in-situ

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measurement 8"9 and archival or surrogate materials to generate missing or supplemental mechanical property information.

2.3 Review of the surveillance program (Step 3) The review of the surveillance program is limited by the original capsule design and the use of the life-limiting vessel materials in the capsules. The reporting of the surveillance capsule program results is covered by a plant's Final Safety Analysis Report and Technical Specifications. Many of the important background details may be missing from the reports submitted to the NRC for review. Therefore, it is important that utilities develop and maintain an accurate and up-to-date package of this information. The information on the orientation of the surveillance capsules should be reviewed. Section III, of the ASME Code requires transverse orientation, but the surveillance capsules for many vessels were fabricated before the code requirements were implemented, and some of them have a longitudinal orientation. According to the NRC Standard Review Plan, Section 5.3.2, the upper shelf energy (USE) for the specimens with transverse orientation is about 0.65 times the one for the longitudinal orientation. Correlation monitor materials that are tested need to be compared to past results from other power reactors, and all surveillance results need to be compared with RG 1.99, Rev. 2 predictions to ensure uniformity of results and to identify potential problems. Decisions as to supplemental surveillance options can then be assessed at any time in the future. Ex-vessel dosimetry and use of results from other nuclear plants' surveillance programs for similar, surrogate or the same materials can provide a significant benefit. Other techniques and data may help resolve future issues if a suitable supplemental surveillance program is developed. Temperature verification in capsules relative to the reactor vessel wall, insertion of new capsules with new or reconstituted specimens, consideration of thermal annealing response and reembrittlement rates, and integration with other plant surveillance programs can lead to meaningful results that can enhance a utility's regulatory posture.

2.4 Review of the in-service inspection program (Step 4) The results of the in-service inspection programs are currently documented through ASME Code and NRC reporting requirements. However, an integrated package of the details and many of the specifics not directly requested by the NRC need to be developed and maintained; this package would include information on any additional inspections, or enhancements performed that were not required by the

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ASME Code or the NRC. Proper storage and preservation of original weld radiographs are important for background comparison if a flaw is discovered and requires an evaluation. Details on all aspects of discovered flaws and their disposition (whether the flaws meet Code acceptance criteria, past fracture mechanics evaluations, or required repair procedures) need to be properly maintained. Changes in inspection frequency and additional inspections (IWB-2420 and -2430, ASME Section XI) might have been required if flaws larger than ASME Code Section XI acceptance standards (IWB-3400, -3500, ASME Section XI) were detected during past in-service inspections of a given vessel. Recently, the A S M E Section XI Code has developed more stringent requirements for demonstrating performance of ultrasonic examination procedures, equipment, and personnel used to detect and size flaws at the susceptible sites in the pressure vessels. The sites include the clad-base metal interface, nozzle inside radius section, reactor vessel structural welds, nozzle-to-vessel welds, and bolts and studs. These requirements are needed to ensure that inspectors apply the appropriate ultrasonic inspection techniques in the field to correctly characterize the flaws at the susceptible sites in the vessel. The requirements are presented in the two Mandatory Appendices of ASME Section XI: Appendix VII, Qualification of Nondestructive Examination Personnel for Ultrasonic Examination; and Appendix VIII, Performance Demonstration for Ultrasonic Examination Systems. Implementation of Appendices VII and VIII will take several years. Results from this enhanced inspection program need to be reviewed in the future. This program will provide more reliable in-service inspection data on US reactor pressure vessels than now available. These data can be used in the development of vessel flaw distribution more representative of the operating PWR vessels and likely to be less conservative, as far as fracture mechanics analyses are concerned, than the currently used distributions such as the Marshall distribution.

2.5 Review of the operating history (Step 5) Review of the operating history is crucial in evaluating a vessel's residual life because some of the design assumptions might have been violated during actual plant operation. Step 5 involves the consolidation of details on all operating transients that have occurred over the life of the plant. The most important parameters are temperature and pressure because they relate directly to the integrity of the reactor vessel. Documentation of transients that violate the plant's Technical Specifications for pressure-temperature limits needs to be maintained

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and cataloged; evaluations of such excursions may follow the guidance in the A S M E Code, Section XI, Appendix E. Specific information on any damage and repairs, physical plant or vessel modifications, and changes in fuel loading patterns that might have resulted in flux reduction need to be reviewed and d o c u m e n t e d in a form useful for future life assessment. Vessel modifications include changes in core structures and in vessel internals such as removal of the thermal shield. Because of some of these modifications and changes, surveillance capsules may not contain the limiting materials. Information on changes in normal, abnormal, and emergency operating procedures also needs to be reviewed and documented. 3 ASSESSING D A M A G E E X P E C T E D D U R I N G T H E N E X T O P E R A T I N G P E R I O D ( P A R T II) Part II (Steps 6 and 7) of the approach guides the assessment of what the radiation embrittlement and fatigue damage will be at the end of the next operating period. This assessment is based on the comprehensive information reviewed in Part I. Part II does not address the fatigue and stress corrosion cracking damage to closure studs because these components can easily be replaced when needed. Nor does Part II address the primary water stress corrosion cracking damage to control rod drive nozzles, because that is addressed in another task and is reported elsewhere.

3.1 Assessment of radiation embrittlement damage (Step 6) Assessment of radiation embrittlement damage includes estimating the adjusted reference temperature (RTNDT), USE, and fracture toughness at the end of next operating period. The shift in the adjusted reference temperature, ARTNDT, at the vessel inside surface is estimated using RG 1.99, Rev. 2. The ARTNDT and RTNDT at ~- and 3-thickness in the vessel wall are estimated using the dpa-based attenuation model in RG 1.99, Rev. 2; the estimated adjusted reference temperatures are employed in revising the P - T limits in Step 8. RTvrs is calculated using the correlation provided in the PTS rule, which is now equivalent to that used in the R G 1.99, Rev. 2.t RTvrs is now equal to the adjusted reference temperature. The fracture toughness values, K~c and Kta, are estimated by shifting the corresponding curves in Appendix A of ASME Section XI according to ARTNDT. Upper-shelf fracture toughness may be estimated if its correlations with USE are available. If the surveillance program does not include the limiting vessel t An amendment was adopted in 1991 that changed the PTS Rule (10 CFR Part 50.61) to be consistent with RG 1-99, Rev. 2.

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materials, the current fracture toughness of these materials may be estimated using an available field technique or by testing small samples removed benignly from the vessel. The damaged area remaining in the vessel after sample removal can be weld-repaired (if necessary) by an ASME Code-accepted procedure. Techniques for sampling small amounts of material from an actual vessel wall can aid in evaluating both material chemistry and mechanical properties. For example, the Belgian BR3 reactor vessel was sampled by making small drillings to varying depths into the vessel wall; the shavings were collected and chemically analyzed. Similar sampling of the thermal shield assisted in evaluating the accumulated fluence. 4 If larger boat-type or core-type sampling techniques were developed, subsized test specimens could be fabricated for direct measurement of mechanical properties. Development and implementation of any remote sampling technique for an irradiated pressure vessel is, however, not a trivial undertaking. Electric Power Research Institute has funded a proof-of-principle project for developing a core-type sampling tool; 5 however, this tool is not tested in the field. Recently, an apparatus called a field indentation microprobe was developed for nondestructive, in situ determination of mechanical properties (yield strength, flow properties, fracture toughness) of metallic structures. 8'9 The apparatus consists of an automated ball indentation unit that performs multiple indentations at the same site on a polished metallic surface and measures penetration depths during loading and unloading. The measured load-displacement data are used to determine the yield strength and the true-stress, true-plastic-strain curve up to 20% strain, and to determine the fracture toughness using the correlations developed for the particular material. The yield strength and flow properties for both irradiated and unirradiated reactor pressure vessel materials measured with this apparatus agree well with the measurements made with conventional methods. 8 These measured results have been used to estimate fracture toughness of the materials.

3.2 Assessment of fatigue damage (Step 7) Step 7, assessment of fatigue damage during the next operating period, aids in estimating growth of known and hypothetical flaws. If fatigue is ever an issue for a PWR vessel, the effects of accumulated fatigue can be evaluated. In terms of crack initiation, the fatigue usage factors for all vessel regions are relatively low except for closure studs (which are likely to be replaced at the end of the initial license period) and some nozzle regions. If the existing fatigue usage factor calculations

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cannot be fine-tuned to produce an acceptably small value, the susceptible sites need to be included in the in-service inspection program and, if the usage factor is greater than, say, 0.7, the inspection frequency may need to be increased. This degree of sophistication is less likely to be necessary for PWR pressure vessels than for attachment piping. The real fatigue concern is that of crack growth of either a postulated flaw or a flaw found in service. The primary element that can affect the rate of crack growth in a water environment is sulfur. "~ Sulfur (S) in combination with manganese tends to form inclusions in ferritic steels. The morphology (shape) and distribution of the sulfides can cause crack growth test results to differ by a factor of two, depending on the crack plane orientation. 11 The higher the concentration of sulfides in a particular crack propagation direction, the more likely the sulfides can accumulate at the crack tip, resulting in increased crack growth rates. Results from one segment of the Heavy Section Steel Technology (HSST) program conducted at Westinghouse show that high sulfur steels have higher crack growth rates than do low sulfur steels.12 Results from some of the low sulfur steels are worth noting. ~3 The effects of sulfides in plates, welds, and forgings appear to be different; even low levels of sulfur in plates still tend to produce higher growth rates. Also, there appears to be an enhanced environmental effect at high load ratios in the range of 0.7, typical of reactor vessel loadings. A fatigue crack growth evaluation can be performed using the ASME Code, Section XI, crack growth curves (cycle-dependent model) or using a time-dependent model if high sulfur (>0.01 wt%) is present in the vessel materials. ~4,~5 4 E V A L U A T I N G VESSEL I N T E G R I T Y AT T H E E N D OF T H E NEXT O P E R A T I N G P E R I O D (PART III) Step 8, evaluation of vessel integrity at the end of the next operating period, addresses three issues: pressurized thermal shock (PTS) events, low upper-shelf toughness, and revised P - T limits. The issue of pressurized thermal shock is addressed by the NRC in the 10 CFR 50.61 screening criteria (PTS Rule). Plant-specific probabilistic fracture mechanics analyses outlined in RG 1.154 are required when the screening criteria are violated before the end of the current license. This issue probably affects only about 10 to 15% of the PWR vessels. Recently, as discussed in Section 3.1, the NRC adopted an amendment that makes the embrittlement correlation in the PTS Rule the same as the one in RG 1.99, Rev. 2. A cursory assessment of the impact of the amendment indicates that for plants having low and

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moderate copper content welds and having plates as controlling materials with low nickel content, a decrease in the RTvrs at the end of licensed life is expected (compared to the original PTS Rule predictions). However, a number of plants with moderate copper, high nickel content welds show an increase in R Tvrs. The increase in the predicted RTzrs causes about five US plants to exceed the PTS screening criteria limits before the expiration of their 40-year operating licenses, and four more are within 0.5 °C (1 °F) of the screening criteria limits; the 0.5 °C difference is statistically insignificant.t The low-upper-shelf toughness concern affects about 20% of the PWRs, and plant-specific evaluations using proven elastic-plastic fracture mechanics methodologies are required to show continued safe operation below an upper shelf Charpy V-notch energy of 68J (50ft-lb). The A S M E Code Section XI Working Group on Flaw Evaluation has developed the appropriate acceptance criteria for Level A and B service conditions (normal and upset). These criteria ensure the retention "of adequate safety margins in reactor vessels where the upper-shelf toughness is less than 68 J (50 ft-lb). These criteria are based on the material J - d a curve and are as follows: For P = l'15Paec (accumulation pressure) + thermal loads, J0.1 applied < J0.1 material and

For P = l'25Pacc + thermal loads and Japplied= Jmaterial, (dJ/da)applied < (dJ/da)materia I.

The criteria assume a conservative (bounded) J - d a curve, postulated flaw with an aspect ratio of six, and a depth of one-quarter thickness of vessel wall. The first part of the criteria ensures the ductile behavior of the material, while the second part ensures that the postulated flaw is stable. Similar requirements are being developed for Levels C and D (emergency and faulted) conditions. The Babcock & Wilcox Owners' Group, Reactor Vessel Working Group, has a program (Integrated Reactor Vessel Surveillance Program) that comprises supplementary surveillance capsules containing Linde 80 welds and fracture toughness specimens. ~6 Seventeen PWR vessels fabricated by B&W are included in the program which is developing a unique Charpy V-notch impact energy and J-resistance fracture toughness database for Linde 80 welds. The screening criteria in the PTS Rule are based on linear elastic fracture mechanics and, therefore, do not account for the drop in the USE below 68 J in some of the PWR vessel materials. However, as t The proposed ammendment was adopted in 1991.

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discussed earlier, USE is likely to drop below 68 J in several P W R vessels. During a PTS event, low values of U S E could adversely affect the integrity of reactor vessels having low and moderate R TNDT values. For example, one analysis shows that the critical flaw depth decreases linearly as upper shelf toughness decreases for RTNDT< 120 °C (250 °F), and is independent of upper shelf toughness for RTNDT> 120 °C. 17 The concern for pressure-temperature restrictions is far more reaching (affecting almost all PWRs at some time in the future) and requires plant-specific evaluation of the radiation-induced shifting of the vessel pressure-temperature operating curve as related to other restrictions, such as the p u m p operating curve and low-temperature overprotection requirements. The implementation of R G 1.99, Rev. 2, has an impact on the calculation of operating limits for P W R plants. Operating plants are expected to incorporate the changes resulting from RG 1.99, Rev. 2, into their operating pressure and temperature limit curves within 3 years, or sooner, when revisions to existing curves are required in accordance with the plant Technical Specifications. A significant impact of R G 1.99, Rev. 2, for PWRs may be a narrowing of the pressure and temperature limit window for heatup.18 This narrowing would result from a predicted increase in RTNDT at the 3-thickness (3-T) vessel wall position. An upward shift in RTNoT at the 3-T location, corresponding to a reduction in the allowable pressuretemperature heatup limits, would occur for all PWRs, with the a m o u n t of the shift depending on the copper and nickel content and fluence for each plant. This increase in 3-T RTNDT is a consequence of the larger shift in RTNDT at the vessel inside wall, plus the use of the new attenuation model using dpa rather than fluence for E > 1 MeV in R G 1.99, Rev. 2, for predicting the RTNDT gradient through the vessel wall. The general trend indicates a greater than 28 °C (50 °F) increase in 3-T RTNDT on average. 19 Plant cooldown and heatup will have to be done at slower rates as the operating window narrows significantly. Minor violations of P - T limits can be evaluated with the screening criteria of Appendix E (ASME Section XI): for cooling rates greater than 5.6°C/h (10°F/h), the minimum coolant temperature, T~, during a transient is greater than RTNDT by 31°C (55 °F) or more, and the maximum pressure is less than the design pressure; and for a cooling rate smaller than 5-6 °C/h, the maximum allowable pressure is specified for different values of Tc- RTNDT. If the criteria are not satisfied, additional analyses are required. Structural integrity analyses of the pressure vessel need to take into account the interaction between austenitic cladding and ferritic base metal resulting from different thermal properties. TM Differences in thermal conductivities, which are accounted for in the analysis, affect

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the magnitude and distribution of thermal stresses resulting from thermal transients. Cladding thickness affects the heat transfer to base metal; thicker cladding reduces the heat transfer and, therefore, the thermal shock on the ferritic base metal. The differences in thermal expansion coefficients, which are generally neglected in the analysis, affect the behavior of cracks in the vessel at the cladding-base metal interface. Nonmandatory Appendix A of ASME Section XI (1989 version) recommends including the cladding-induced stresses for fracture mechanics analysis of pressure vessels. 5 I D E N T I F Y I N G T H E OPTIONS A V A I L A B L E TO M A N A G E A G I N G (PART IV) Once the damage assessments (Steps 6 and 7) and integrity evaluations (Step 8) are completed, and if there appears to be a problem in meeting the requirements at the end of the next operating period, other options and actions are available for mitigating and managing radiation embrittlement damage. This section discusses three such options: flux reduction, thermal annealing, and enhanced in-service inspection.

5.1 Flux reduction (Step 9) Reduction in the intensity of the neutron flux field impacting the critical vessel materials is a possible utility action to mitigate embrittlement damage so that exceeding the PTS screening criteria before the end-of-license period can be avoided and the current license requirements can be met. Neutron flux can be reduced by several techniques. One of the easiest and sometimes most cost effective approach is improved fuel management, using a low-leakage loading pattern (L3p) instead of a standard loading pattern. 21'22 The low-leakage loading pattern includes placement of fuel assemblies with high burnup (burnup reached at the end of one fuel cycle) in peripheral core locations adjacent to critical weld and base metal sites in the vessel. Besides a reduced flux (of the order of 25 to 50% flux compared to a standard loading pattern), a low-leakage loading pattern can produce fuel cycle cost savings over the long term. 23 A low-low-leakage loading pattern (L4P), which is similar to LaP but includes fuel assemblies with higher burnup (burnup reached at the end of two fuel cycles), may be used for further reduction in flux. 24 About 95% of the operating PWR reactors in the USA have implemented or planned some form of low-leakage fuel management. Additional flux reduction of up to 90% at longitudinal vessel welds can be achieved by placing highly burned fuel in assemblies across from

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welds. Several utilities have placed highly burned fuel across from welds and expect a 50% reduction in fast neutron flux at the welds. Other more drastic techniques, such as shielding techniques 24,25,26 or power reduction can be considered; these techniques can produce flux reductions of greater than 50%, but flux reduction generally needs to be implemented fairly early in life to significantly reduce the overall fluence exposure. Shielding techniques include stainless steel rods or pellets inserted in place of fuel rods or pellets, neutron poisons in peripheral guide tubes, and reactor internal changes. Several foreign PWR plants have used dummy stainless steel assemblies for flux reduction. At least one domestic PWR plant has used stainless steel pellets inserted in peripheral fuel rod assemblies at an elevation corresponding to the location of a circumferential weld to reduce the flux at the weld. Reactor internal changes and replacement provide the largest flux reduction. 24 Examples of reactor internal changes include modification of the existing thermal shield, replacement of the existing thermal shield with a neutron pad, and placing stainless steel patches on a core barrel. Vessel-specific analyses are essential before selecting a particular flux reduction technique. The benefits of flux reduction depend on the time of implementation, original flux level, and the radiation sensitivity (as correlated with chemical composition) of vessel materials. Some flux reduction techniques may result in power derating for the plant; therefore, implementation of these techniques is influenced by external economic concerns as well as plant safety concerns. Any flux reduction program should include verification by physical measurement that the technique is working as predicted. Therefore, a baseline flux distribution should be measured before a flux reduction technique is implemented, a3 Neutron monitors can be installed in the reactor vessel cavity (outside the vessel wall) at locations where maximum flux reduction is desired. 27 These monitors can be readily installed (even in plants with neutron shield tanks 2z) and removed during refueling outages; the results from these monitors can be compared with the surveillance capsule dosimetry results or other in-vessel dosimeters.

5.2 Thermal annealing (Step 9) Thermal annealing of the irradiated beltline region can fully or partially eliminate the transition temperature shift experienced during operation and restore the uppershelf energy properties lost during the same operation. This option has been assessed as being feasible for US

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commercial power reactor v e s s e l s . 28'29 Annealing in the temperature range of 425 to 475°C (806 to 887 °F) for 150h in air has been performed on at least nine VVER-440 (Soviet PWR) reactor vessels; seven with no cladding and two with weld-deposited c l a d d i n g . 3°m'32 Annealing at or below 345 °C (650 °F) has been accomplished using water as a heating medium (hence, a 'wet' anneal), 29 but the degree of recovery of mechanical properties is significantly less than the higher temperature 'dry' anneal in air, because the temperature difference between operating and annealing temperatures is less. It appears that the minimum of about 79°C (150°F) difference in temperatures is needed to obtain substantial recovery of the mechanical properties. Dry annealing involves the removal of the core internal structures and primary water so that a radiant heating source can be inserted near the vessel wall to locally heat the embrittled beltline region. Wet annealing is not as complicated from an engineering viewpoint because the primary water temperature is controlled by pump heat up to the vessel design temperature of 345 °C (650 °F). Engineering difficulties for the dry anneal are quite complex and require plant-specific evaluations to ensure that other portions of the plant such as concrete structures are not harmed by the high annealing temperatures. ASTM E509 provides guidance for thermally annealing a reactor p r e s s u r e vessel. 33 Early planning is essential so that a revised and supplemental surveillance program can be implemented not only to assess the degree of recovery from the anneal, but also to measure the rate of re-embrittlement (under power reactor conditions) for the critical materials. Measurement of re-embrittlement rates is essential for determining remaining life after annealing; no trend formulas for transition temperature shift or upper-shelf toughness drop exist for irradiated-annealed-reirradiated materials. Published test results on reirradiated materials have relied on test reactor conditions, 28"34 which include high fluence rates, and, therefore, the results may not be valid for power reactors. The degree of thermal anneal recovery and the rate of loss of the upper shelf toughness upon further irradiation is not as well established, primarily because there is a disparity between postannealed Charpy V-notch upper shelf energy and measured upper shelf Jic toughness. 2s Charpy V-notch upper shelf energy values show a high degree of recovery, whereas the J-resistance curve properties do not show the same level of recovery; this situation then carries over into the reirradiation response. Westinghouse has recently completed a study for Electric Power Research Institute that includes power reactor initial irradiations, but the postanneal reirradiation was under test reactor conditions. 35 In the

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Westinghouse study, transition temperature shifts were derived from microhardness tests (through a linear correlation with yield strength giving T30= 4-10 D P H for welds and ~0 = 3-15 DPH for base metal, 4° where DPH is the diamond pyramid hardness, and temperature is in degrees Fahrenheit). 36 These tests appear to provide reasonable results, although data scatter (in the range of at least 10%) and possible interactions between irradiation temperature and fluence should not be ignored. 37 The NRC is also sponsoring research to evaluate the re-embrittlement of thermally annealed RPV materials using test reactor data.t The evalution of the re-embrittlement of thermally annealed RPV materials is one particular area where small (miniature) specimen testing may prove adequate when coupled with the proper supporting documentation. Another important consideration after annealing is the through-wall property gradient in the vessel and the gradient change after reirradiation. Apparently, the Russians have removed 100-mm-diameter samples containing materials from the critical weld in the core region of the decommissioned Novovoronezh Unit 1 reactor pressure vessel to investigate through-thickness embrittlement and annealing response. 3' Further work is needed to provide a more realistic model for predicting re-embrittlement after annealing, when only a limited baseline of information is available from supplemental surveillance testing. If annealing is selected as an option, the need for early planning of a supplemental surveillance program is obvious. Broken Charpy specimens, if saved, may be reconstituted and used in the supplemental surveillance program. The reconstituted specimens would provide very useful information for the surveillance program during the postannealing period. 38 Material sampling and small specimen testing are other possibilities for enhancing the knowledge of either baseline or irradiated behavior information. Decisions as to whether to implement these options can be made only if the data collection steps (Steps 1-5) are completed, so that the most accurate projections of radiation embrittlement can be made in Step 6. Other plant modifications (e.g. use of heated safety injection water) or operational changes may provide further benefits that would have to be quantified. Needs for revising operator training to mitigate operating transients and accidents need to be evaluated.

5.3 Enhanced in-service inspection (Step 9) Automated ultrasonic inspection systems have shown improved detection and sizing capability over manual techniques. 39 These inspection t Unpublished handouts from the Vessel Integrity Review Group, Oak Ridge, Tennessee, December 1986.

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systems are capable of improving N D E reliability and, as a result, developing more appropriate flaw distributions. One such system is the Ultrasonic Data Recording and Processing System, which features automatic flaw detection software and can be used with many different ultrasonic inspection techniques, for example, amplitude-based techniques, and tip-diffraction techniques. 4° The amplitude-based technique using focused transducers is better suited to detect and size flaws in nozzle-to-vessel welds than using conventional, unfocused transducers. Similarly, the time-of-flight diffraction technique is better suited for sizing near-surface flaws normal to the clad-base metal interface. 4~ An acoustic emission technique may be used to detect flaw growth during hydrotests if the outside surface of the vessel is accessible.

5.4 Other actions (Step 9) Available solutions to restrictive P - T limits include hardware changes and administrative procedure changes. An example of a hardware change is modifying the power-operated relief valves to allow for variable opening setpoints so that the P - T curve can be closely followed while providing a low-temperature overpressure protection. An example of an administrative procedure change is prohibiting the starting of a reactor coolant pump during heatup when the reactor coolant system is in a water solid condition, which exists prior to drawing of the steam bubble in the pressurizer. Fracture specimens may need to be included in the surveillance program if the USE is expected to fall below the 68 J level. Weld repair may be needed if flaws of unacceptable size are expected during the extended period. A shielded metal arc welding process using the half-bead technique can be used for weld repair; this procedure is accepted by the ASME Code and it does not require any hightemperature postweld heat treatment. 42 However, the half-bead technique is not well-suited for remote operation, which is essential for repair in high-radiation environment present in the vessel. In addition, recent test results show that the weld metal deposited using the half-bead technique, and associated heat-affected-zone, are subject to a strain aging mechanism and exhibits significant reduction in fracture toughness at elevated temperatures. 43 A new automated weld repair procedure that employs a gas tungsten-arc weld process does not have these disadvantages. This procedure uses a temperbead technique that is well-suited for remote operation for repairs in the high-radiation environment. 43'44 Deposits of weld layers following the first layer temper the weld heat-affected zone in the base metal to ensure toughness equal to or greater than that of

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the original base metal. Therefore, this weld repair process does not require any high-temperature postweld heat treatment. The weld metal deposited by this procedure is less likely to subject to strain aging. This repair procedure is accepted by the ASME Boiler and Pressure Vessel Code and recognized under Code Case N-432. The final action (Step 10) is a re-evaluation of the vessel condition when an action or change is implemented in Step 9. This re-evaluation includes working back through Steps 6 and 7, unless the option or action warrants a significant design change resulting in a revised design basis (i.e. new stress analysis), a materials change, a change in projected operating h~story, better inspection results, or supplemental surveillance information. These cases would involve updating Steps 3 to 5 also. The safety and economic consequences of the various options and actions are plant-specific and need to be addressed in detail through the re-evaluation process. 6 CONCLUSIONS R E L A T E D TO R E S E A R C H N E E D S This paper has presented a generic aging management approach for estimating the current condition of a reactor pressure vessel and for evaluating the vessel's integrity during the next operating period. The approach accounts for the known degradation mechanisms of irradiation embrittlement and fatigue. Structural integrity is ensured through the use of fracture mechanics evaluations keyed to meet current ASME Code and the USNCR regulatory acceptance criteria. These criteria are generally based on fairly conservative assumptions and estimation techniques. Current research activities being conducted by the NRC and the utility industry are focused on projects that can help provide a better understanding of actual vessel safety margins and provide better tools for utility use. Several technical areas are discussed in which additional research and development work is needed to perform more accurate assessment and reliable management of aging damage. Some of the most important technical areas are as follows: • •

• •

Shifted K~R and K~c curves for power reactor irradiation conditions need to be validated. Further refinement of mechanistic modeling of irradiation embrittlement (including the effects of thermal annealing and reirradiation after annealing) is warranted. Better understanding of the effects of irradiation temperature and flux level is needed. Increased use of advanced ultrasonic techniques (e.g. automatic ultrasonic techniques, focused transducers, and time-of-flight

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•

•

•

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diffraction techniques) can provide more reliable and accurate characterization of vessel flaws. Better understanding of the effect of irradiated cladding on vessel integrity is needed. Qualified in-situ measurement techniques and small and miniature specimen testing techniques for generating mechanical property values can be used to estimate material properties of irradiated pressure vessel materials. Probabilistic risk and fracture mechanics analyses can be used for assessing safety margins and for evaluating the impact of remedial measures (to the vessel and to the overall plant) that can affect overall failure risk. Qualified weld repair techniques to repair an irradiated vessel would be needed if flaws of unacceptable size are detected during operation. Common databases, coordinated research activities, and cooperative exchange of information need to be encouraged between all nuclear power industry participants to eliminate duplicate efforts and to maximize the nuclear industry's resources. REFERENCES

1. Vora, J. P., Nuclear Plant Aging Research (NPAR) Program Plan, NUREG-1144, Rev. 1, September 1987. 2. Shah, V. N. & MacDonald, P. E., Residual Life Assessment of Major Light Water Reactor Components--Overview, NUREG/CR-4731, EGG2469, Vol. 1, 1987, Vol. 2, 1989. 3. Server, W. L., Amar, A. S. & Shah, V. N., Insights for Aging Management of Major Light Water Reactor Components, Volume 1: Pressurized Water Reactor Pressure Vessels, NUREG/CR-5314, Vol. 1 (Draft), October 1991. 4. Fabry, A., Motte, F., Stiennon, G., Debrue, J., Gubel, P., Van de Velde, J., Minsart, G. & Van Asbroeck, Ph., Annealing of the BR3 reactor pressure vessel. 12th Water Reactor Safety Research Information Meeting, NUREG/CP-0058, Vol. 4, January 1985, p. 144. 5. Reactor Vessel Material Sampling Tool, Technical/Brief, Electric Power Research Institute, Palo Alto, 1986. 6. Standard methods of tension testing of metallic materials. Annual Book of ASTM Standards, ASTM E 8-89, Vols. 01.01-01.05 and 03.01, American Society for Testing and Materials, Philadelphia, pp. 197-217. 7. Standard methods for notched bar impact testing of metallic materials. Annual Book of ASTM Standards, ASTM E 23-88, American Society for Testing and Materials, Philadelphia, pp. 277-300. 8. Haggag, F. M. & Nanstad, R. K., Estimating fracture toughness using tension or ball indentation tests and a modified critical strain model. In Innovative Approaches to Radiation Damage and Failure Analysis, eds D.

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This paper was prepared as an account of work sponsored by an agency of the United States G o v e n m e n t . Neither the U n i t e d States Government nor any agency thereof, or any of their employees, makes any warranty, expressed or implied, or assumes any legal liability or responsibility for any third party's use, or the results of such use, of any information, apparatus, product or process disclosed in this paper or represents that its use by such third party would not infringe privately owned rights. The views expressed in this paper are not necessarily those of the U.S. Nuclear Regulatory Commission.