ML20217H366

NPP Generic Aging Lessons Learned, Presented at 970817-22 14th Intl Conference on Structural Mechanics in Reactor Technology in Lyon,France

ML20217H366
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Issue date:	08/17/1997
From:	Christopher Regan NRC
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NUDOCS 9804290353
Download: ML20217H366 (10)

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14th International Conference on Structural Mechanics in Reactor Technology (SMIRT-14)

Lyon, France, August 17-22,1997 Nuclear Power Plant Generic Aging Lessons i.eamed (GALL)

Christopher M. Regan U.S. Nuclear Regulatory Commission, U.S.A.

l ABSTRACT The purpose of the generic aging lessons learned paper was to provide a systematic review of plant aging information to assess materials and component aging issues related to operation and license renewal of nuclear power plants. The results were documented using a standardized tabular and electronic database format and definitions of aging-related degradation mechanisms and effects. Results reveal that all significant aging issues are currently being addressed by the U.S. Nuclear Regulatory Commission regulatory process.

1. INTRODUCTION j Approximately 110 nuclear electrical power generating plants operating in the United Sates of America generated roughly 20% of the nations electrical demand. Some of these plants have been in operation for many years. It is well established that many of the critical components in nuclear power plants are subject to time-dependant degradation, or aging, as a result of normal plant operations. In recognition of the potentially adverse effects of the aging process on plant safety, the United States Nuclear Regulatory Commission's (USNRC) Office of Nuclear Regulatory Research began by establishing the Nuclear Plant Aging Research (NPAR) Program. The principal objective of this program was to develop a basic understanding of age-related degradation (ARD) processes and their effect on nuclear power plant systems, structures, and components. In addition, the Nuclear Energy Institute (NEI), formerly the Nuclear Management and Resources Council (NUMARC),

developed a series of license renewal industry reports (irs) to support a prospective applicant when submitting an application for a renewed license under the requirements of

Title 10, Part 54, of the U.S. Code of Federal Regulations (10 CFR Part 54) [1]. The irs l describe the NUMARC assessment of plant aging issues and management strategies for j several components and structures.

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To develop appropriate technical criteria for addressing the aging issues related to nuclear power plant license renewal, the USNRC initiated an activity to assess and integrate the age-related information from all USNRC documentation to include NPAR reports, generic communications, and License Event Reports (LERs) and to use the results of this assessment to supplement and update license renewal guidance previously developed. This activity was called the Generic Aging Lessons learned (GALL) Program, the results of which have been documented in NUREG/CR-6490 (ANL-96/!3), " Nuclear Power Plant Generic Aging Ixssons 12arned (GALL), Main Report and Appendices A and B."

l NUREG/CR-6490 presents the results of the GALL review program. The GALL cffort was sponsored by the USNRC with significant input provided through a joint effort involving 12 technical experts from Argonne National I.aboratory (ANL) and Idaho National Engineering Laboratory (INEL). ANL reviewed information on mechanical, l structural, and thermal-hydraulic components and systems and INEL reviewed information l

on electrical components and systems. The results of these reviews were compiled using a standardized tabular format and standardized definitions of ARD effects. All tabulated review information is contained in Appendices A (Volume 1) and B (Volume 2) of NUREG/CR-6490. The information is also available in a computerized data base format based on the software program FoxPro. The data base allows rapid queries and sorts of the large amount of information generated by the review.

2. DESCRIPTION OF REVIEW PROCESS More than 550 documents containing nuclear power plant information were reviewed for

" GALL" information. The USNRC staff performed searches for current operating experience documents covering the 5 year period, 1989-1994, using the USNRC's Nuclear l Documents Managements System (NUDOCS). The period preceding 1989 was documented in the NPAR program reports which are also summarized in NUREG/CR-6490. The searches used the following terms: aging, degradation, and failures. A total of 163 NPAR I reports, 31 USNRC Generic Letters, 265 USNRC Information Notices, 82 LERs, 5 l USNRC Bulletins, and 10 NUMARC Industry Reports (irs) containing mechanical, structural, thermal-hydraulic, and electrical systems and components were reviewed under the GALL program. The results of these reviews were compiled by using a standardized tabular format and standardized definitions of ARD and effects. A standardized and consistent set of definitions and descriptors for all aging mechanisms encountered during the review was developed for the GALL effort. Individual aging mechanisms and effects were identified and defined, and these are listed and described in Table 8 of NUREG/CR-6490. This list will help focus and systematize future reactor aging studies.

l The reports, notices, letters, and bulletins reviewed are listed in Tables 2 though 7 of i

NUREG/CR-6490. The results from each reviewed document are summarized in the I GALL tables contained in Appendices A (volume 1) and B (volume 2) of the two volume l 1

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c l NUREG/CR-6490 report. A separate table was prepared for each of the NPAR reports and NUMARC irs; findings from the USNRC Generic Ixtters, USNRC Information Notices, USNRC Bulletins, and LERs are tabulated by year in separate tables. All of the GALL table information has alto been entered into a FoxPro data base software program that can i be used on IBM PC-compatible systems to retrieve and categorize information on structures j and components and their related aging effects.

1 The information contained in the GALL tables is a summary of that provided in the l

reviewed reports. It was found that not all of the reports, notices, and bulletins reviewed contained relevant information on Age-Related Degradation (ARD) processes. A number of the NPAR reports described programs, methodologies, computer codes, etc., for l studying and analyzing aging processes in nuclear components, but did not provide detailed j information on the processes themselves. The tables for these reports contain a standard i statement indicating this fact. Almost all of the USNRC Generic Letters, USNRC L Information Notices, LERs, and USNRC Bulletins reviewed contained detailed information on the failure of specific components, but the failures were sometimes judged not to be aging-related by the reviewer or by the author of the reviewed document. For example, failures caused by improper heat treatment, preexisting defects introduced during manufacturing, or overloads or other abuse during operation were not considered aging-related by the reviewer, even though the failure might not have occurred until the component had been in service for some time. GALL table entries are not provided for USNRC Generic letters, USNRC Information Notices, LERs, and USNRC Bulletins

, judged not to contain detailed information on specific aging effects and their impact on specific plant components. The structure of the information documented in NUREG/CR-

! 6490 and in the database is in such format that enables a researcher to easily access i

information on a particular aging issue.

In the future, information documented in NUREG/CR-6490 may be used as a reference tool for reviewers performing license renewal application reviews or for the USNRC staff reviewing a licensee submittal. Indeed, a prospective license renewal applicant or nuclear power plant licensee may use this document as one of the sources for determining applicable aging effects to be addressed in a license renewal application.

3. OBSERVATIONS AND FINDINGS l More than 550 documents comprising 163 NPAR reports,31 USNRC Generic Letters,265 USNRC Information Notices, 82 Licensee Event Reports, 5 USNRC Bulletins, and 10 l NUMARC Industry Reports (irs) were reviewed under the GALL program. The results l of these reviews were systematically summarized in a tabular format, using standardized i definitions of ARD mechanisms and effects developed for this study (see Exhibits 1 & 2).

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The revi:w reveals (1) that there are no new issues with respect to the components subject

, to ARD and the degradation mechanisms responsible and (2) that all ongoing significant

! issues are currently being addressed by the regulatory process. However, (3) the aging of passive components has been high-lighted for continued scrutiny.

Included in NUREG/CR-6490, are recommendations of the referenced report's author for each aging issue and which are contained in Appendices A and B of NUREG/CR-6490.

l The GALL cffort then added the ANL/INEL judgement of the validity of the author's

! recommendations based on present day understanding of the issue. The characterization of the issue is noted by one of four possible aging issue current relevance categorization indicators with the all-important "may potentially need further evaluation," indicating the possibility of emerging aging issues. This characterization of particular issues, however, appears only a few times and where it does appear the issue is currently being addressed by the regulatory process. A summary of general observations concerning specific aging l issues and the components affected are presented below.

I l

3.1 Mechanical, Structural, and Thermal-Hydraulic Components and Systems As expected, corrosion and corrosion-related processes were the dominant mechanism of age related degradation (ARD) la coolant piping and steam generators. Where high- l velocity fluids were present in piping, erosion / corrosion was also a significant mechanism. !

Additionally for piping, feedwater nozzles, and interfacing tanks and other components, nonuniform water temperature fields aggravated by thermal buoyancy can cause large induced structural thermal stresses of either quasi-steady, low-cycle, or thermal shock nature and can lead to cracking or significant structural distortion. These thermal stresses are usually not accounted for in component design and are highly plant and mode-of-operation-dependant. They can occur under normal or intermittent operation of plant systems and tend to be worse under low flow conditions. For reactor internals, irradiation-assisted stress corrosion cracking was an important source of degradation where high radiation fields were present. Other forms of corrosion, as well as vibrational fatigue, also contributed to internals degradation.

Pump and valve casings were likewise found to be subject to corrosion and erosion / corrosion related degradation. Thermal embrittlement was an important mechanism in cast stainless steel pump and valve components. Moving parts in pumps and valves suffered from ARD produced by wear. vibration, fatigue, and erosion / corrosion. Valve and pump seals and other elastomer components were subject to degradation by physical and chemical degradation at elevated temperature and/or prolenged exposure to the service environment.

The principle degradation mechanism affecting concrete structures was leaching and breakdown of cement phases under the action of aggressive chemicals, degradation due to freeze-thaw cycles, and corrosive attack of the embedded rebar. The responsible mechanism (s) for some concrete wall cracking was found to be not well understood.

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Diesel generators, air compressors, and ventilating and air conditioning equipment suffered principally from wear, vibration, and fatigue associated with reciprocating motion, as well as corrosion and wear induced by contamination. Heat exchangers and steam generators were subject to contamination and corrosion, as well as biofouling, thermal fatigue, and vibrational fatigue. Vibrational fatigue, wear, and elevated temperature degradation of damping fluids commonly caused degradation in snubbers.

Table 1 lists aging issues found to occur almost equally in boiling water reactor (BWR) and pressurized water reactor (PWR) plants and tend to center on various forms of corrosion and fatigue. Another important commonality of the components listed in this table is that they are all what are termed passive components as described in 10 CFR Part 54 [11].

This may be of considerable significance because the literature reviewed seems to indicate that passive components are not as extensively or thoroughly covered by current plant maintenance procedures. Furthermore, surveillance and monitoring methods and instrumentation and procedures have not been as extensively developed or employed for passive components subjected to the highlighted aging mechanisms, nor are some of the passive component aging mechanisms as well understood. Thus, plant life extension by employing component replacement and maintenance could be more tenuous for the passive components. Furthermore, passive components tend to be some of the most costly in a plant and are frequently not as easy to replace. For these reasons, the knowledge base for i

predicting relevant aging effects behavior and significance, which is essential to the l

development of robust plans for aging reduction, monitoring procedures, and maintenance, l is very important for passive components.

1 3.2 Electrical Components and Systems Breakers and relays were usually covered together in the same report; the predominant aging-related failure mechanisms were contact wear, sticking linkage, loss of lubrication, or elevated temperature. Normally energized relay coils were frequently mentioned as high i failure-rate items because of the insulation breakdown caused by elevated temperature due to self-heating from the continuous current. Breakers are routinely refurbished on periodic schedules. Instrumentation and control (I&C) systems, including breakers and sensors, are made up of many small components that are routinely replaced after a number of years of service, as determined by qualification programs. Thus aging is controlled by scheduled l

maintenance and periodic replacement. Redundancy in the Reactor Protection System and 4 Engineered Safety Features Actuation Systems allows for taking a channel out of service for maintenance.

Degradation of cable insulation and jackets is the major effect of cable aging, due primarily to radiation and elevated temperature. Despite sizable efforts to develop electrical and mechanical methods of detecting cable insulation degradation, there are no reliable methods of detecting degradation of electrical cable insulation in a reactor containment. Electrical parameters, while relatively easy to measure, were found not to give a good indication of mechanical degradation of the cable insulation. The mechanical indentor method was successful only for some of the jacket insulation types.

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TABLE 1. Selected Examples of Issues Significant to Passive Structures & Components Reactor Component Material Degradation Process References Type PWR Instrumentation low-alloy steel Environmentally assisted fatigue. NUREG/CR-and control rod (A533D, A508) Appropriate design rules do not yet exist 5490 [2]

[ drive (CRD) with Type 308 or in the American Society of Mechanical housing nozzles 309 SS clad Engineers (ASME) Boiler & Pressure Vessel Code BWR Closure Studs A-540, B22, B2, Environmentally assisted fatigue, NUREG/CR-and or B24 fretting, and boric acid corrosion if 5490 [2]

PWR leakage present PWR CRD system Various Dropped or stuck rod due to failure by NUREG/CR-components fatigue, mechanical wear, or stress 5555 [3]

corrosion cracking BWR Jet pump & Inconel X-750 Cracking and possible failure form NUREGICR-holdown beams vibrational and/or environmentally 5754 [4] ,

assisted fatigue and stress corrosion l cracking l BWR Reactor Various Crack initiation, growth, and possible NUREG/CR-Internals failure from irradiation-assisted stress 5754 [4]

corrosion cracking (IASCC)

PWR Lower core Type 304 stainless Cracking and possible failure from NUREG/CR-support steel (SS), A- vibrational fatigue and IASCC 6048 [5l structure 286, Inconel X-components 750, and others BWR Pressure vessel low-alloy steel Cracking, possibly stress corrosion Information upper head (A533B, A508) cracking (SCC), of weld clad, with Notice (IN) with Type 308 or cracks penetrating into underlying base 90-29 [6]

309 SS clad metal BWR Core Shroud Type 304 SS SCC (or IASCC) leads to circumferential IN 93-79 [7], l cracking of core shroud and concerns IN 94-42 [8]

about possible structural failure in an accident or seismic event BWR Recirculating Cemented WC in Preferential corrosive dissolution on Ni coolant pump Ni birer binder under certain undefined conditions seals leads to excessive seal leakage and possible eventual pump failure BWR All piping and Commonly used large thermally induced stresses, either NUREG/CR-and feedwater materials, low quasi-steady or low-cycle transient, 4731 PWR nozzles and alloy steels thermal fatigue, induced by nonuniform Vols.1 & 2 interfacing coolant temperature fields aggravated by [9]

tanks and thermal-buoyancy-caused stratification components under no-flow / low-flow levels, cause wall cracking / gross abnormal component distortion, usually not accounted for in component design, highly plant and mode-of-operation dependant BWR Shielding wall Reinforced Actual process and mechanisms unclear; NUREG/CR-and concrete and concrete shows up as large surface cracks not 4652 [10]

PWR other locations caused by structural loading 720-9 1

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Motors and generators occasionally fait due to bearing wear caused by vibration and winding insulation breakdown from elevated temperatures. Brushes also age due to wear.

Battery chargers and inverters are small electrical systems made up of many electronic components titat, like the instrumentation and control (I&C) system, can be taken out of service for maintenance because of redundancy. Many of the electrical I&C components I

are included in plant quality assurance (QA) programs that require periodic replacement. l Inverter failures have caused numerous problems. Many of the electrical I&C components are included in plant quality assurance programs that require periodic replacement. A more detailed analysis may be carried out at a later date to assess the significance of these mitigative practices and the aging processes.

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4. CONCLUSIONS l

l A preliminary assessment of the GALL tables reveals that all significant issues with respect to structure and component aging are currently being addressed by the USNRC regulatory process. Nonetheless, the aging of certain structures and components and the resulting aging effects, particularly in the category of what are termed passive structures and .

components, have been high-lighted for continued scrutiny and evaluation. Among these l issues erosion / corrosion and fatigue of mechanical and structural components and l degradation of electrical insulation of cables and other electrical items are the most notable.

The material cocumented in NUREG/CR-6490 may be referenced by both operators of l currently licensed nuclear power plants and those wishing to extend their current operating license for a additional operating period.

l 5. REFERENCES

1. United States Nuclear Regulatory Commission. January 1,1996. Code of Federal Regulations, Title 10, Part 54.

2. Werry, E. V. October 31,1990. NUREG/CR-5490, Regulatory Instrument Review:

Management of Aging of LWR Major Safety Components.

3. Gunther, W. and Sullivan, T. March 31,1991. NUREG/CR-5555, Aging Assessment of the Westinghouse PWR Control Rod Drive System.

4. Luk, K. H. September 30,1993. NUREG/CR-5754, Boiling-Water Reactor Internals Aging Degradation Study-Phase I.

5. Luk, K. H. September 30, 1993. NUREG/CR-6048, Pressurized-Water Reactor Internals Aging Degradation Study-A Phase I Report.

6. United States Nuclear Regulatory Commission. April 30, 1990. Information Notice 90-19, Cracking of Cladding and Its Heat Affected Zone in the Base Metal of Reactor Vessel Head.

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7. United States Nuclear Regulatory Commission. September 30, 1993. Information Notice 93-79, Core Shroud Cracking at Beltline Region Welds in Boiling-Water Reactors.

l

8. United States Nuclear Regulatory Commission. June 7,1994. Information Notice 94-

! 42, Cracking in the lower Region of the Core Shroud in Boiling-Water Reactors.

l 9. Shah, V. N. and MacDonald, P. E. June 30,1987, Volume 1, and November 30, I

' 1989, Volume 2. NUREG/CR-4731. Residual Life Assessment of Major Light Water Reactor Components.

10. Naus, D. J. September 30,1986. NUREG/CR-4652, Concrete Component Aging and Its Significance Relative to Life Extension of Nuclear Power Plants I

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