ML20127D412
| ML20127D412 | |
| Person / Time | |
|---|---|
| Issue date: | 05/31/1985 |
| From: | Anderson W, Sterner P NRC OFFICE OF INSPECTION & ENFORCEMENT (IE) |
| To: | |
| References | |
| IEB-82-02, IEB-82-2, NUREG-1095, NUDOCS 8506240221 | |
| Download: ML20127D412 (75) | |
Text
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s NUFIEG-1095 Evaluatiori of Re' ponses to:
. Degradation of Threade'd-Fa'steners in Reactor' Coolant DPressure Boundary.of Pr'essurized-Water-Reactor Plants i
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NOTICE Availability of Reference Materials Cited in NRC Publications Most documents cited in NRC publications will be available from one of the following sources:
- 1. The NRC Public Document Room,1717 H Street, N.W.
Washington, DC 20555
- 2. The Superintendent of Documents, U.S. Government Printing Of fice, Post Office Box 37082, Washington, DC 20013 7932
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Referenced documents available for inspection and copying for a fee from the NRC Public Docu-ment Room include NRC correspondence and internal NRC memoranda; NRC Office of Inspection and Enforcement bulletins, circulars, informatior notices, inspection and investigation notices; Licensee Event Reports; vendor reports and correspondence; Commission papers; and applicent and licensee documents and correspondence.
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mission, Washington, DC 20555.
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Copies of industry codes and standards used in a substantive manner in the NRC regulatory process are maintained at the NRC Library, 7920 Norfolk Avenue, Bethesda, Maryland, and are available there for reference use by the pubhc. Codes and standards are usually copyrighted and may be purchased from the originating organization or, if they are American National Standards, from the American National Standards Institute.1430 Broadway, New York, NY 10018.
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NUREG-1095 Evaluation of Responses to IE Bulletin 82-02 Dtgradation of Threaded Fasteners in Reactor Coolant Prcssure Boundary of Pressurized-Water-Reactor Plants Manuscript Completed: May 1985 Date Published: May 1985 W. Anderson, P. Sterner Division of Emergency Preparedness and Engineering Response Office of Inspection and Enforcement U.S. Nuclear Regulatory Commission W:shington, D.C. 20555
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ABSTRACT IE Bulletin 82-02 was issued by the NRC on June 2,1982, to notify licensees about incidents of severe degradation of threaded fasteners. The bulletin required appropriate action including submittal of information from pressurized water reactors having an operating license. Responses from 41 licensees included their recent experience with degradation of threaded fasteners in primary system components. Data from recent regular inspections of reactor coolant pressure boundary component connections of 6-in. size and larger are compiled.for technical evaluation. Statistical analysis is used to deter-mine significant factors related to frequency of leakage incidents in connec-tions, occurrence of degradation of bolts and studs, and the need for bolt replacement. Factors examined include the age of the plant, types of components, use of lubricants and sealants, and differences between plants. The compiled data indicate that, on the average,10% of the bolted connections show evidence-of leaking during an 18-month period. Also, 80% of the connections that show evidence of leakage undergo some degradation of the bolting. Results of the analysis show a significant decrease in the occurrence of bolting degradation events as the age of the plant increases. The data also show that valves are less subject to bolting corrosion. A group of 5 of the 41 plants accounted for about one-half of the reported leakage and corrosion events. The common charac-teristic found for four of these five plants was the lubricant used.
The use of nickel-graphite based lubricants appears to offer a significantly reduced incidence of leakage and corrosion, based on late corrections to the reported data. The data also permits the conclusion that the use of molybdenum-disulfide-based lubricants and graphite-based lubricants results in a signifi-cantly increased incidence of leakage and corrosion.
Reporting of data on lub-ricants was of poor quality and detracted from the value of the bulletin responses.
NUREG-1095 iii
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TABLE OF CONTENTS Page ABSTRACT.............................................................
iii ACKNOWLEDGMENTS......................................................
vii
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1 INTRODUCTION...................................'[...r............
1-1 1.1 Objective.................................................
1-1
- 1. 2 Scope.....................................................
1-1 1.3 Background....................................'............
1-1 2
RESPONSE TO IE BULLETIN 82-02....................................
2-1 2.1 Data Compilation...........................................
2-1 2.2 Evaluation of Data.........................................
2-3 2.3 Stati stical Signi ficance Study.............................
2-6 3
DISCUSSION OF RESULTS...........................................
3-1 3.1 Rate of Occurrence of Leaks...................... 5 3-1 3.2 Type of Degradation Found..................................
3-1 3.3 Occurrence of Damage Relative to Leakage...................
3-2 3.4 Bolt Replacement..........................................
3-2 3.5 Comparison of Reports of Recent Events to Reports of Past Events................................................
3-2 3.6 Relation of Leakage and Degradation to the Age of the Plant......................................................
3-3 3.7 Leakage and Degradation Events Relative to Type of Component..................................................
3-3 3.8 Leakage and Degradation Events Relative to the Types of Lubricants Used............................................
3-4 3.9 Plant-to-Plant Variations..................................
3-4 3.10 Combined Effect of Plant Age and Lubricant Used.............
3-5 3.11 Physical Basis for Observed Effects of Age and Lubricants...
3-6 4
CONCLUSIONS.....................................................
4-1 5
RECOMMENDATIONS.................................................
5-1 6
REFERENCES......................................................
6-1 APPENDICES A
IE Bulletin No. 82-02, Degradation of Threaded Fasteners in the Reactor Coolant Pressure Boundary of PWR Plants B
Tabulated Data from Responses to IE Bulletin 82-02 y
C Contingency Table Analysis - Step by Step NUREG-1095 v
i
FIGURE P. ag 1
Percent of Inspections With Leakage vs. Age of Plant............
2-7 TABLES
.1 Summary of Information Compiled for Recent Events...............
2-8 2
Comparison of Response Data Compiled for Past and Recent Events for 41 Plants...................................................
2-8 3
Bolting Degradation Event Totals for 41 Plants..................
2-9 4
Response Information From Recent Events for 41 Plants Categorized by Age of Plant.....................................
2-9 5
Response Information From Recent Events for 41 Plants 1
Categorized by Type of Component................................
2-10 6
Response Information From Recent Events for 41 Plants Categorized by Type of Lubricant................................
2-11 7
Comparison of Response Information for Two Groups of Plants for Both Recent and Past Events.................................
2-12 8
Response Information From Recent Events for 36 Plants Categorized by Age of Plant.....................................
2-13 9
Response Information From Recent Events for 36 Plants Categorized by Type of Component................................
2-14 10 Response Information From Recent Events for 36 Plants Categorized by Type of Lubricant................................
2-15 11 Statistical Contingency Analysis Results........................
2-16 NUREG-1095 vi
ACKNOWLEDGMENTS
,1 The authors thank the members of the Engineering and Generic Communications Branch, Office of Inspection and Enforcement, who contributed to the develop-ment of this report, in particular Mr. W. J. Collins who developed 1E Bulietin 82-02, initiated the compilation of the responses, and provided valuable guid-ance during development of this report; Mr. C. D. Sellers of the Office of.
Nuclear Reactor Regulation who contributed valuable guidance on bolting practices and problems; and Dr. D. Lurie of the Office of Resource Mapagement for his guidance in the use of statistical evaluation methods.
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NUREG-1095 vii
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l 1 INTRODUCTION This report provides a summary of the responses to the Nuclear Regulatory Commission (NRC), Office of Inspection and Enforcement (IE)Bulletin 82-02,
" Degradation of Threaded Fasteners in the Reactor Coolant Pressure Boundary of PWR [ Pressurized Water Reactor] Plants." The bulletin is attached as Appendix A to this report.
1.1 Objective The objective of this report is to present the technical information included in the responses to the bulletin and the evaluation of the data relative to the concerns that prompted the bulletin.
Such an evaluation of the data will pro-vide justification and guidance for future actions, such as standards develop-ment programs to reduce the potential for failure of bolted connections in nuclear power plants.
This report should also provide a basis for closure of the bulletin.
1.2 Scope This report is limited to presentation and technical evaluation of the data on bolting that were reported in response to the bulletin.
The impact of the bulletin on plant staff time or radiation exposure is not included.
Technical evaluation is limited to observations on the reported data after it was inter-preted to permit compiling and statistical evaluation.
Comparison to the results of other NRC-sponsored studies is included also.
1.3 Background
A significant increase in the number of reported incidents of severely degraded bolts as a result of boric acid corrosion wastage and stress corrosion cracking mechanisms became apparent through licensee event reporting in 1980 and 1981.
This resulted in IE Bulletin 82-02 being issued on June 2, 1982.
The bulletin was addressed for action to all pressurized water nuclear power facilities holding an operating license.
The scope of action was limited to the reactor coolant pressure boundary (RCPB) and included the threaded fasteners (studs or bolts) in (1) steam generator and pressurizer manway closures, (2) valve bonnets and pump flange connections installed on lines having a nominal diameter of 6 inches or greater, and (3) control rod drive (CRO) and pressur-izer heater connections that do not have seal welds to provide leak-tight integrity. The reactor head closure studs were excluded for licensees committed to Regulatory Guide (RG) 1.65, " Materials and Inspection for Reactor Vessel Closure Studs."
NUREG-1095 1-1
The bulletin contained three action items to be performed and provisions for two action items for written reports to be submitted to the NRC as summarized in the following:
(1) Where procedures do not exist, develop and implement maintenance pro-cedures for threaded fastener practices.
Establish quality assurance measures for use of lubricants and sealants to minimize susceptibility to stress corrosion cracking environments.
(2) Threaded fasteners of closure connections covered by the bulletin shall be examined in accordance with Section XI of the American Society of Mechanical Engineers (ASME) Code when the connection is opened for inspection or maintenance.
(3) Identify those bolted closures of the RCPB that have experienced leakage, particularly during the most recent plant operating cycle.
Identify and describe materials and experience with lubricants for fasteners and sealants, particularly any instances of stress corrosion cracking of fasteners.
(4) A written report to include (a) a statement that Action Item 1 has been completed (b) identification of the sp7cific connections examined as required by Action Item 2 (c) the results of the examinations performed as required by Action Item 2 (5) A written report to provide the information requested by Action Item 3.
Additionally, the bulletin requested information on the cost of complying with these five items and costs to be considered in staff time and in radiation exposure attributed to the requested inspections.
Additional background and details are contained in the attached bulletin (Appendix A).
In the interim there have been other ongoing efforts by NRC and by industry to develop the basis for improvements in the performance of bolted connections.
Among these efforts are those by NRC in resolving Generic Issue No. 29, " Bolting Degradation or Failure in Nuclear Power Plants," and programs sponsored by the Electric Power Research Institute (EPRI), the Atomic Industrial Forum (AIF),
and Materials Properties Council (MPC).
NUREG-1095 1-2
2 RESPONSE TO IE BULLETIN 82-02 The 51 plants that the bulletin applied to responded to Action Item 3 of the bulletin a timely manner as prescribed by Action Item 5.
Data from these responses are referred to as "past events" in the evaluation of data.
Past leakage incidents, if any, were not reported in the licensee responses on a consistent basis.
Some older plants reported incidents occurring throughout most of the life of the plant, while some other older plants reported only those from the last refueling cycle.
Some newer plants, having recently received their operating license (0L), reported no past incidents of leaks at bolted connections.
Responses to Action Item 2 as prescribed by Action Item 4 have been received from 41 of the 51 plants.
Data from these responses are referred to as "recent events" in the evaluation of data.
The inspections to provide this information were to be performed no later than the completion of the next refueling outage that was initiated after 60 days from the date of the bulletin.
For eight plants, recent prolonged periods of down-time precluded their timely response to the bulletin. Another two plants had received their OL, but had not started operation when the bulletin was issued.
The licensees reporting under Action Items 4 and 5 was not consistent in format, content, and level of detail.
It appears that criteria for conclusions on ex-tent of damage and continued acceptability of bolting also were not consistent.
As an example, some licensees limited their response on results of examination to statements of no reportable indications (NRI).
The latest edition of the ASME Code in use, the 1980 Edition, does not provide any acceptance or reporting criteria for corrosion ir.dications such as pitting or wastage and, because the bulletin emphasized stress corrosion cracking, such degradation may have gone Utilizing the data, despite apparent inconsistencies unreported in those cases.
in the basis for reports, required interpretation to allow compilation and eval-uation of results.
Reasonable concerns with regard to the validity of any con-clusions drawn may arise as a result of the necessary interpretation.
2.1 Data Compilation The data compiled and evaluated are limited to those that were included in responses to the bulletin.
Incidents of past corrosion that were available but were not in the responses are not included.
The information on bolting problems provided by the licensees' responses was interpreted and compiled in tabular form for each plant. The tabulated data from the 41 plants that provided complete responses are contained in Appendix 8.
In developing the tables of Appendix B, several assumptions and clarifications were made in an attempt to provide a basis for consistent interpretation of the responses. These are as follows:
NUREG-1095 2-1
(1) Basis for the Selection of Connections and the Number of Inspections (a) One major assumption used in the compilation and evaluation of the licensees' reported data concerned the basis for inspections performed under Action Item 2 of the bulletin.
It was assumed that the inspec-tions were planned on the basis of random selection of connections.
This excludes consideration that inspections may have been directed primarily at bolted connections that required maintenance because they had experienced leakage.
(b) One inspection for steam generator manways consists of an inspection of one manway.
(c) One inspection for a reactor coolant pump consists of an inspection of either the pump case flange or the seal flange.
(2) Number of Leaks (a) Reports of leakage of components other than those contained in responses to the bulletin are not included.
(b) Packing or body-to-bonnet leaks are recorded as valve leaks.
(c) Leaks in pump case flanges or seal flanges are recorded as reactor coolant pump leaks.
(d) Where significant bolting corrosion was reported, it was assumed that leakage had taken place.
Significant corrosion includes wastage and heavy pitting that would warrant replacement.
Because light pitting may result from occasional wetting of the bolt by sources other than leakage, reports of light pitting were not interpreted as leakage incidents unless leakage was reported.
(3) Attributed Cause The attributed cause for leakage is shown in Appendix B only when the cause was specified in the licensee's report.
(4) Bolting Degradation Events (a) Data under the heading of " Number of Fastener Degradation Events" in Appendix B have been quantified by the number of joints with the appropriate degradation event, not by the number of bolts involved.
In those cases where a number in parentheses is included in the table under those subheadings, that number shows the number of bolts involved.
(b) Where responses indicated that corrosion had taken place but the responses failed to specify whether it had involved pitting or wastage, it was assumed to have been wastage.
(c) Axial indications were not considered to be stud cracking and have not been recorded as bolting degradation events.
Acceptance NUREG-1095 2-2
l standards for axial indications are not as restrictive as circum-l ferential cracks and it was assumed that such cracks were not induced in service.
Numbers recorded under the subheading of "No. Bolts Replaced" indi-(d) cate the total number of individual bolts replaced for all of one or more connections inspected in that type of component.
Entire sets of bolts that were replaced because of redesign or (e) replacement of a component are not included as bolt replacement.
(5) Lubricants Used (a) Lubricants are recorded as being used for a component if they were specified for use or approved for use for that type of component.
In cases where a licensee's response listed several lubricants that (b) were used or approved for use at that plant, with no details given with regard to individual components, it was assumed that all listed lubricants were being used or were approved for use for all types of components.
(c) Lubricantt, although generally reported by trade name, have been tabulated by generic composition.
(6) Sealant Used If a licensee's re:ponse indicated that a sealant had been used or was approved for use on a particular component a "yes" has been recorded in the appropriate block.
2.2 Evaluation of Data The data shown in Appendix B from the 41 plants that responded to Action Items 2 and 4 of the bulletin serve as the basis for further data evaluation.
The following questions and possibly useful relations were postulated to guide the compilation and the evaluation of the data:
(1) What is the rate of occurrence of (a) leaking bolted connections among the plants reporting data (b) amount and type of bolting damage among plants reporting data (c) fastener damage relative to leakage (d) bolt replacement (2) What is the relation of leaking connections and bolting damage to various parameters including (a) age of the plant (b) types of components (c) types of thread lubricants (d) use of sealants NUREG-1095 2-3
(3) Can any plant-to plant variations or other trends be observed?
Before performing any tabulation for use in evaluation, the data were examined.
Groups that were judged to be of limited significance were eliminated or com-bined with other groups.
Data for components with a limited total number of inspections iere not included in any further evaluation.
Not included in further evaluation were data for steam generator handholes, pressurizer manways, reactor coolant system (RCS) valves, and control rod drive flanges and pressurizer heater flanges.
The elimination of data from steam generator handholes, pressurizer manways and heater flanges, and RCS valves resulted in a loss of less than 10% of the data from recent inspections.
Data related to other valves, isolation valves (including safety / relief valves) and safety injection valves, were combined in order to produce a significant group of data for valves.
This combined group of valves has the common character-istic that the valves are shut during normal power operating conditions.
RCS valves were not included because they are not used in most plants and they re-main open during normal power operating conditions.
Control rod drive flanges are usually seal welded and, although bolted flanges were inspected for only a few plants, the large numbers inspected in each of those cases tends to dis-tort the data base.
The component groups that were included in the final eval-uation are steam generator manways, valves (including isolation valves, safety / relief valves, and safety injection system valves) and reactor coolant pump case and seal flanges.
Other data groups eliminated from further evaluation include the attributed cause of leakage and the use of sealants.
Only two licensees included in their responses the specific cause for any recent leaks.
Nine licensees indicated that they had used a sealant in the past or that it was approved for use for a particular component.
It was not clear in some reports if the sealant approved for use had actually been used.
Where the past use of sealant was reported, it coul not be related to recent incidents of corrosion that were reported.
There were no uses of sealant reported in recent events.
The data shown in Appendix B were reduced and consolidated as described above and then tabulated in various arrangements guided by the previously outlined questions.
Table 1 presents a summary of information compiled from recent events for the 41 plants reporting.
This table illustrates the rate of occurrence for leaking connections, corrosion events, and bolt replacement, along with the relation between these items.
Table 2 presents a comparison of response data compiled for past and recent Averages of leaks per reactor year were calculated by dividing the events.
number of leaks by the number of reactor operating years involved.
The number of reactor years for past data was calculated by adding the ages of all plants involved.
The age of a plant was defined by the number of years that had expired between the year in which the plant received its operating license and the year in which the bulletin was issued.
The number of reactor years for recent data was calculated by multiplying the number of plants (41) by 1.5.
This is based on an assumed 18-month interval between outages.
NUREG-1095 2-4
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Table 3 presents data from past and recent events categorized on the basis of type of bolting degradation events.
Table 4 presents the response information from recent events categorized by age Plants were grouped by age as determined by the time groups of the plants.
that had expired between the year in which a plant received an operating license, including a provisional operating license, and the year in which the bulletin was issued.
These age groups are 4 years or less, 5 through 7 years, 8 through 10 years, and greater than 10 years.
These age groups relate to the 3 year inspection periods and 10 year inspection intervals of the ASME Code Section XI.
Table 5 presents the response information from recent events categorized by the three groups of components:
steam generator manways, valves, and reactor coolant pump flanges.
Table 6 presents a summary of response information from recent events cate-gorized by type of lubricant. There were several limitations involved in this evaluation.
Categories are based on type of lubricants specified as follows:
(1) only nickel graphite (2) only molybdenum disulfide (3) graphite-alcohol (also includes graphite-oil for three plants)
(4) two lubricants approved - MoS2 and graphite-alcohol but excluding category (4) (may include (5) several lubricants including MoS2 graphite-alcohol)
(6) several lubricants including graphite-alcohol but excluding MoS2 (7) either copper graphite only or both copper graphite and nickel graphite Eighteen inspections could not be placed into any of these categories, because the lubricant (s) used on that type of component could not be determined from the licensees' reports.
Categories (4), (5), (6), and (7) contain data related to components for which more than one lubricant was reported as approved for use.
It was not clear in these cases whether these lubricants were considered interchangeable or if specific lubricants were used on specific components within that type orAlso, whether the lubricant used had been changed through time in operation.
it was not clear which of the lubricants had been used on components that had leaked.
If one of the lubricants denoted was molybdenum disulfide, the data were included in group (4) or (5).
If molybdenum disulfide was not present on a multiple listing of lubricants, the data was included in group (6) or (7).
In reviewing the data for plant-to plant variations or trends, examination of the data charts included in Appendix B, showed that data sets from five plants accounted for 50% of the total number of recent leaks and over 50% of the recent corrosion events.
Each of these data sets accounted for more than 3 leaks, an average of 4.6 leaks per plant, with a percentage of leaks per inspection NUREG-1095 2-5
greater than 30% with a total of 49 inspections, 23 leaks and 23 corrosion events.
Additional review showed that 24 plants reported no leaks in the components in the three groups selected for further evaluation though they may have reported leaks in other components.
Data from these 24 plants reported on 225 such con-nections examined.
The remaining 12 plants reported 23 leaks in the examination of 140 connections, with a percentage of leaks per inspection of 16.6%, an aver-age of about 1.9 leaks per plant.
The following parameters were examined for those plants with the five distinct data sets:
age of plant, owner of plant, lubricents used, NRC region, plant constructor, nuclear steam system supplier (NSSS), and architect / engineer.
The only significant characteristic was that four of the five plants indicated use of MoS or graphite alcohol lubricant for 29 of the 49 connections.
Table 7 2
presents a comparison between data from all of the 41 plants, the 5 distinct data sets and the remaining 36 plant subset.
Tables 8, 9, and 10 are directly related to Tables 4, 5, and 6, respectively, but are based on the data from the 36 plant subset.
Table 8 presents a summary of data categorized by age group for the 36 plants.
Figure 1 illustrates the percentage of leaks per inspection relative to plant age for all 41 plants and also for the 36 plant subset.
This figure clearly shows the downward trend of leakage incidents with increasing plant age.
Table 9 presents a summary of data for the 36 plant subwt categorized by type of component. The percentage of leaks per inspection for steam generator man-ways decreases from 12.4% on Table 5 to 4.2% on Table 9.
The 5 distinct data sets account for 17 of the 24 total reported steam generator manway leakage incidents.
Table 10 presents a summary of response information from recent events for the 36 plant subset for the 36 plant subset categorized by type of lubricant.
2.3 Statistical Significance Study Before deriving any conclusions, a study was made in order to determine the statistical significance of postulated relationships.
A chi-square contingency table analysis was performed for the comparisons of interest.
The procedure for this analysis is included in Appendix C.
Analysis calculations were per-formed using a simple program on a microcomputer.
Table 11 shows the results of that study listing relationships of interest, tables referenced, chi square values, degrees of freedom, and determinations of whether or not differences were judged to be significant and met a criterion for analysis of significance that was arbitrarily established for this report.
Because little action or impact is anticipated based on conclusions of significance, significance is based on an alarm threshold of a = 0.10 rather than the usual a = 0.05 and the analy-sis must pass the criterion that expected values for each cell in the contin-gency table are greater than 3.0.
When either of these is not satisfied, a notation is made in Table 11.
Satisfaction of the criterion of expected values greater than 3.0 can be checked by examination for many cases with two degrees of freedom to avoid further analysis.
In cases where recommendations for actions could result in recognizable impact, this report would require an a = 0.05 level of significance and also require expected values of 5 or more as a criterion for analysis of significance.
NUREG-1095 2-6
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Totals for all response data excluding five 4
distinct sets
-- Totals for all response 3
data 2
1 I
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0 04 5-7 8 10
>10 Age of plant at time of bulletin (years)
Figure 1 Percent of inspections with leakage vs. age of plant NUREG-109s 2-7
Table 1 Summary of information compiled for recent events Response Area Finding 4
Total:
Plants responding 41 Connections inspected 415
=
Leaks 46 Corrosion events 39 (j
r Corrosion-related bolt replacement
- 27 Leaks per plant (average) 1.1 Corrosion events per plant (average)
- 1. 0 Percent inspections with evidence of:
Leakage 11.1 Corrosion 9.4 Percent leaks with:
Evidence of corrosion 84.8 Bolt replacement 59
)
- This is the number of connections that had one or more bolts replaced.
The total reported number of bolts replaced exceeded 200.
Table 2 Comparison of response data compiled for past and recent events for 41 plants Response Area Past (0L-1982)
Recent (1982 present)
Leaks 184 46 Corrosion events 23 39 Percent leaks with evidence 12.5 84.8 of corrosion Leaks / reactor year (average) 0.6 0.7 NUREG-1095 2-8
Table 3 Bolting degradation event totals for 41 plants Bolting Degradation Events Recent Past Corrosion wastage 25 22 Stud cracking 0
2*
Pitting 15 0
Mechanical damage 18 1
Destroyed by removal 4
0 Steam erosion 2
2
- Two additional events of stud cracking, not reported in bulletin responses but recorded in NUREG-0943 are not included.
Table 4 Response information from recent events for 41 plants categorized by age of plant Age of plant at time of bulletin (years)
Response Area 0-4 5-7 8-10
> 10 Total Total:
Plants responding 8
9 18 6
41 Connections inspected 101 58 184 72 415 Leaks 16 8
18 4
46 Corrosion events 10 8
17 4
39 Corrosion-related bolt replacement events 8
7 8
4 27 Leaks per plant (average) 2.0 0.9 1.0 0.7 1.1 Corrosion events per plant (average) 1.3 0.9 0.9 0.7
- 1. 0 Percent inspections with evidence of:
Leakage 15.8 13.8 9.8 5.5 11.1 Corrosion 9.9 13.8 9.3 5.5 9.7 Percent leaks with:
Evidence of corrosion 62.5 100 34.4 100 84.8 Bolt replacement 50 87.5 44 100 59 NUREG-1095 2-9
1 Table 5 Response information from recent events for 41 plants categorized by type of component Component Steam Generator Reactor Coolant Response Area Manways Valves Pumps Total:
Connections inspected 194 142 79 Leaks 24 13 9
Corrosion events 22 8
9 Leaks per plant (average) 0.59 0.32 0.22 Corrosion events per plant (average) 0.54 0.20 0.22 Percent inspections with evidence of:
Leakage 12.4 9.2 11.4 Corrosion 11.3 5.6 11.4 Percent leaks with evidence of corrosion 91.7 61.5 100 NUREG-1095 2-10
1 Table 6 Response information from recent events for 41 plants categorized by type of lubricant Lubricant
- Response Area (1)
(2)
(3)
(4)
(5)
(6)
(7)
Total:
Connections inspected 137 14 16 29 60 105 36 Leaks 11 6
8 3
8 8
0 Corrosion events 9
6 8
2 8
4 0
Corrosion events with bolt replacement 6
2 4
2 7
4 0
Percent of inspection with:
Leaks 8.0 43 50 10.3 13.3 7.6 0
Corrosion 6.6 43 50 6.9 13.3 3.8 0
Percent of leaks with:
Evidence of corrosion 82 100 100 67 100 50 0
Bolt replacement 55 33 50 67 88 50 0
- Legend for columns:
(1) Only nickel graphite type lubricants were specified.
(2) Only molybdenum disulfide lubricants were specified.
(3) Only graphite-alcohol type lubricants were specified.
(4) Two lubricants specified:
molybdenum disulfide and graphite alcohol.
(5) Several lubricants were specified, including molybdenum disulfide.
(6) Several lubricants were specified, not including molybdenum disulfide but including graphite-alcohol.
(7) Either copper graphite type lubricants only, or two lubricants, copper-graphite and nickel graphite, were specified.
NUREG-1095 2-11
Table 7 Comparison of response information for two groups of plants for both recent and past events 41 Plants 36 Plants 5 Plants Response Area (1)
(2)
(3)
Total:
Plants responding 41 36 5
Connections inspected 415 366 49 Recent leaks 46 23 23 Recent corrosion events 39 16 23 Recent corrosion-related bolt replacement events 27 14 13 Past leaks 185 104 81 Recent leaks per plant (average) 1.1
- 0. 6 4.6 Recent corrosion events per plant (average) 1.0 0.4 4.6 Past leaks / reactor year (average) 0.6 0.4 2.3 Recent leaks / reactor year (average) 0.7 0.4 3.1 Recent corrosion / reactor year (average) 0.6 0.3 3.1 Percent recent inspections with evidence of:
Leakage 11.1 6.3 46.9 Corrosion 9.4
- 4. 4 46.9 Percent recent leaks with:
Evidence of corrosion 88.6 69.6 100 Bolt replacement 61 67 57 (1) Data from all 41 plants providing complete bulletin responses.
(2) Data from all plants providing complete bulletin responses, except those five plants providing distinct data sets.
(3) Data from the five plants providing distinct data sets.
i NUREG-1095 2-12
Table 8 Response information from recent events for 36 plants categorized by age of plant Age of plant at time of bulletin (years)
Response Area 0-4 5-7 8-10
> 10 Total:
Plants responding 7
8 16 5
Connections inspected 88 54 168 60 Leaks 11 4
8 0
Corrosion events 5
4 7
0 Corrosion-related bolt replacement events 5
3 6
0 Leaks per plant (average)
- 1. 6 0.5 0.5 0
Corrosion events per plant (average) 0.7 0.5 0.4 0
Percent inspections with evidence of:
Leakage 12.5 8.0 4.8 0
Corrosion 5.7 P.0 4.2 0
Percent leaks with:
Evidence of corrosion 45.5 100 87.5 Bolt replacement 55.6 75 75 NUREG-1095 2-13
Table 9 Response information from recent events for 36 plants categorized by type of component Component Steam Generator Reactor Coolant Response Area Manways Valves Pumps Total:
Connections inspected 168 126 72
~
Leaks 7
9 7
Corrosion events 5
4 7
Corrosion-related bolt replacement events 5
4 5
Leaks per plant (average) 0.19 0.25 0.19 Corrosion events per plant g
L (average) 0.14 0.11 0.14 Percent inspections with evidence of:
Leakage 4.2 7.1 9.7 Corrosion 3.0 3.3 9.7 Percent leaks with:
Evidence of corrosion 71.4 44.4 100 Bolt replacement 71.4 44.4 71.4 NUREG-1095 2-14
Table 10 Response inform 6 tion from recent events for 36 plants categorized by type of lubricant Lubricant
- Response Area (1)
(2)
(3)
(4)
(5)
(6)
(7)
Total:
Connections inspected 121 5
8 29 48 105 32 Leaks 6
0 0
3 4
8 0
Corrosion events 3
0 0
2 4
4 0
Corrosion events with bolt replacement 1
0 0
2 3
4 0
Percent of inspection with:
Leaks 5.0 0
0 10.3 8.3 7.6 0
Corrosion 2.5 0
0 6.9 8.3 3.8 0
Percent of leaks with:
Evidence of corrosion 50 0
0 67 100 50 0
Bolt replacement 17 0
0 67 75 50 0
- Legend for columns:
(1) Only nickel graphite type lubricants were specified.
(2) Only molybdenum disulfide lubricants were specified.
(3) Only graphite-alcohol type lubricants were specified.
(4) Two lubricants specified:
molytidenum disulfide and graphite alcohol.
(5) Several lubricants were specified, including molybdenum disulfide.
(6) Several lubricants were specified, not including molybdenum disulfide but including graphite-alcohol.
(7) Either copper graphite type lubricants only, or two lubricants, copper-graphite and nickel graphite, were specified.
NUREG-1095 2-15
Table 11 Statistical contingency analysis results 2
Table Degrees X
9 Test Correlation Tested Reference of Freedom Value Significance 1
Age of plant and leaks, 4
3 5.44 a >0.1, not 4 age groups, significant leaks vs. no leaks 2
Age of plant and leaks, 4
2 5.24 a <0.1, 3 age groups, significant
{
0-7, 8-10, >10, leaks vs. no leaks 3
Age of plant and corrosion, 4
1 0.04 a >0.1, not 2 age groups, significant 0-4, >4, corrosion vs. no corrosion 4
Components and leaks, 5
1 0.64 a >0.1, not steam generator vs. Other, significant leaks vs. no leaks 5
Components and leaks, 5
1 0.84 a >0.1, not valves vs. Others, significant leaks vs. no leaks 6
Components and corrosion, 5
1 7.59 valves vs. others, leaks with corr. vs. no corr.
7 Components and corrosion, 5
1 3.63 a <0.1, valves vs. others, significant corrosion vs. no corrosion 8
Lubricants and leaks, 6
1 15.42
- nickel vs. MoS,
2 col. 1 vs. 2, leaks vs. no leaks 9
Lubricants and leaks, 6
1 23.20
- nickel vs. graphite, col. 1 vs. 3, leaks vs. no leaks 10 Lubricants and leaks, 6
1 14.55 **
nickel vs. MoS2 and/or a <0.01, graphite, col. 1 vs.
significant 2 + 3 + 4, leaks vs.
no leaks NUREG-1095 2-16
Table 11 (Continued) 2 Table Degrees X
t Test Correlation Tested Reference of Freedom Value Significance 11 Lubricants and leaks, 6
1 4.10 and a <0.05, nickel vs. MoS2 mult, with MoS,
significant 2
col. 1 vs. 2 + 4 + 5, leaks vs. no leaks 12 Lubricants and leaks, 6
1 8.59 nickel vs. graphite or a <0.01, graphite + MoS,
significant 2
col. I vs. 3 + 4, leaks vs. no leaks 13
'.ubricants and leaks, 6
1 6.41 nickel vs. MoS2 apportioned, a <0.02, col. 1 vs [2 + 1/2(4 + 5)],
significant leaks vs. no leaks 14 Lubricants and leaks, 6
1 7.95 a <0.01, nickel vs. MoS2 apportioned, significant col. 1 vs. [2 + 1/2(4)],
leaks vs. no leaks 15 Lubricants and leaks, 6
1 12.41 a <0.01, nickel vs. graphite apportioned, significant col. 1 vs. [3 + 1/2(4)],
leaks vs. no leaks 16 Plants and lubricants, 6,10 1
9.89 36 plants vs. 5 plants, a <0.01, nickel and copper vs.
significant MoS and graphite, 2
col 1 + 7 vs. 2 + 3 + 4 17 Plants and lubricants, 6,10 1
25.44
2 col. 1 + 7 vs. 2 18 Plants and lubricants, 6,10 1
15.91
- 36 plants vs. 5 plants, nickel vs. graphite, col. 1 vs. 3 19 Plants and lubricants, 6,10 1
3.42 36 plants vs. 5 plants, a <0.10, nickel vs. MoS and significant 2
mult. w/MoS,
2 col. 1 vs. 2 + 4 + 5 NUREG-1095 2-17
Table 11 (Continued)
Table Degrees X2 7
Test Correlation Tested Reference of Freedom Value Significance 20 Plants and leaks, 7
1 72.33 **
36 plants vs. 5 plants, a <0.01, leaks vs. no leaks significant 21 Plants and corrosion, 7
1 8.26 a <0.01, 36 plants vs. 5 plants, significant leaks with corr. vs. no corr.
22 Age of plant and leaks, 8
1 10.79 a <0.01, 4 age groups, significant leaks vs. no leaks 23 Age of plant and leaks, 8
1 7.60 **
0-4 years vs. >4 years, a <0.01 leaks vs. no leaks 24 Age of plant and corrosion, 8
1 5.79 a <0.02, 2 age groups:
0-4, >4, significant leaks with corr. vs. no corr.
25 Components and leaks, 9
1 2.32 a >0.1, not steam generator vs. Others, significant leaks vs. no leaks 26 Lubricants and leaks, 10 1
2.13 a >0.1, not nickel vs. MoS2 + mult.,
significant col. 1 vs. 2 + 4 + 5, leaks vs. no leaks 27 Lubricants and leaks, 10 1
1.27 a >0.1, not nickel vs. Others, significant col. 1 vs. 2 + 3 + 4 + 5 + 6, leaks vs. no leaks tall correlation tests are based on numbers of connections with various attributes.
- Does not meet criterion of expected values greater than 3 for contingency table analysis.
- Exceeds criterion of expected values greater than 5.
- Leaks signifies the number of connections inspections with leaks.
NUREG-1095 2-18
3 DISCUSSION OF RESULTS 3.1 Rate of Occurrence of leaks Preliminary safety studies have been reported (NUREG-0933) which estimated the potential safety significance of leaking and corroded bolted connections in the primary coolant system.
In NUREG-0933 the occurrence of 0.13 corrosion events par plant year is considered in determining that improvements in the inspection program of bolted connections would be justified on a value/ impact basis.
That study also assumed that 1 in 10 corrosion events would lead to a small loss-of-coolant accident (52 LOCA).
Table 7 shows a rate of 0.6 recent corrosion events p r plant year for inspection of an average of 10 bolted connections per plant.
Considering that there may be 20 to 40 bolted connections per plant capable of producing an 52 LOCA, it appears that the actual rate of corrosion events may be 10 times that assumed in NUREG-0933.
However, as noted in the assumptions for compilation of data [Section 2.1(1)a], it must be recognized that the recent inspections may have been directed at components where leakage had been observed or suspected and such extrapolation may not be justified.
Examination of Table 7 shows that the five distinct data sets would provide an estimate of 3.1 corrosion events per year from inspection of 10 connections per plant.
On the basis of extrapolation of those data sets, the rate of corrosion events could be 100 times that assumed in NUREG-0933.
The assumption of the probability of one 52 LOCA event for each 10 corrosion events appears to be quite conservative.
Of the 62 corrosion events reported in the recent and past data, shown in Table 2, there have been no reported loss of coolant accidents.
The rate of 52 LOCAs per corrosion event could be lower by a factor of 10 or more, thus compensating somewhat for the increase in the rate of corrosion events.
3.2 Type of Degradation Found Some previous studies (NUREG-0943, NUREG/CR-2827) considered cracking to be the cajor problem with bolting.
Consequently, the Bulletin 82-02 emphasized interest in stress corrosion cracking problems.
However, no cracking incidents were re-ported from recent inspections, as shown in Table 3.
It is possible that cracking occurred in some bolts where significant wastage was reported, but the reports seem to indicate that such bolts were discarded without further examination for hidden cracks.
Past data shown in Table 3 indicate that only about 10% of the bolting damage in the primary coolant system involves cracking.
The dominant form of corrosion reported from both past and recent inspections is wastage.
For recent inspections, wastage or pitting developed in 39 of the 46 incidents (Table 1) of degradation that were reported.
Wastage can be even more common than these reports would indicate because the bulletin could have been interpreted in a manner that would not have required reporting of some pitting or wastage corrosion.
Action Item 2 of the bulletin required examina-tion in accordance with IWA-2210 and IWA-2220 of the ASME Code (XI). The Code NUREG-1095 3-1
editions in use provide acceptance standards for cracks in bolting, but do not provide acceptance or recording standards for corrosion damage.
3.3 Occurrence of Damage Relative to Leakage There is almost a one-to-one relationship between leakage and corrosion wastage or pitting.
Examination of Tables 5, 7, and 11 (Tests 6, 7, 21, 24) suggests that, except for the five distinct data sets and valves relative to other com-ponents, this relation does not tend to dif fer significantly with any of the parameters investigated, including the use of different lubricants.
The lower value of percent of leaks with evidence of corrosion that was reported for past data in Table 2 has been discounted for reasons to be discussed subsequently.
NUREG/CR-2827 and NUREG/CR-3766 show that corrosion rates for ferritic steel bolts can be as high as 0.1 in per year in the presence of boric acid.
This suggests that leakage that per.,ists undetected or disregarded for more than a year or two can develop a potentially seriously degraded condition of any joint.
The basic solution to this problem was addressed in Item 1 of the bulletin, which was directed at eliminating reactor coolant leakage during operations.
3.4 Bolt Replacement Replacement of some bolts is common where leakage has occurred.
More than two thirds (27 of 39) (see Table 7) of the joints where corrosion occurred required some replacement of bolting.
More than 200 bolts were replaced during recent inspections.
This suggests that, for those cases where there is concern about the quality of bolting material, the concern could be gradually alleviated by replacements with improved material, or alternatively complete, prompt replace-ment of all bolting in a problem joint might not be a significant hardship.
In addition to bolt replacement because of corrosion, there were 18 (see Table 3) reported events where some bolting was replaced because of mechanical damage to the bolts, and four reported events where some bolting was destroyed on removal.
This suggests a need for improvements in maintenance procedures and in inspection procedures to reduce the need for disassembly of joints.
3.5 Comparison of Reports of Recent Events to Reports of Past Events In comparing reports of past leakage and degradation events in response to Action Item 3 of the bulletin to reports of recent leak and degradation events in response to Action item 2 of the bulletin, it was concluded that reports of l
l past events are not of sufficient reliability to justify meaningful conclusions.
i Table 2 shows that the leaks per reactor year for both past and recent events l
appear reasonably consistent.
However, this can be misleading because the recent l
data are limited to reports from a limited sample for inspection while the past data should include reports from the entire scope of the bulletin.
Also, some plants reported past leakage events over a defined period of less than the plant age.
In addition, Table 2 also shows a much lower percentage of reported corrosion events for reported past leak events: 12.4% for past events relative to 84,8% for recent events.
The difference in percentage of reported leaks with evidence of corrosion between data from recent events and past events is 1
NUREG-1095 3-2
i too large to explain on any basis other than inconsistency in records or reporting.
It is possible that while observed leakage events were recorded, evidence of corrosion or bolt replacement was not recorded or reported.
In i
i reviewing reports of incidents of past leakage events, it appears that many j
leaking joints were merely tightened to stop the leak with no examination of l
the bolts for evidence of corrosion.
As a result, this evaluation does not propose to evaluate past data to any significant extent.
i 3.6 Relation of Leakage and Degradation to the Age of the Plant Table 4 shows a decrease in leakage events per inspection from 15.8% for newer l
plants to 5.5% for the older plants.
Corrosion events per inspection, leaks per plant, and corrosion events per plant during the recent inspection also show the same trend of reduction for older plants.
This effect is statistically signifi-cant when data from plants are grouped by plant age as 0-7 years, 8-10 years, and greater than 10 years (Table 11, Tests 1, 2).
The data plotted on Figure 1 sug-gests that a learning curve exists for good bolting practice and that good prac-tice reduces the number of leaking joints as the plant ages and the maintenance i
crews gain experience.
Other factors, also related to age of the plant and expe-I rience of the crews, could be that bolted joints, which show repeated problems j
early in the plant life, are modified, have materials replaced, or otherwise have i
special procedures applied that reduce the subsequent number of leaks.
- However, the percentage of leaking joints that experience bolting corrosion remains high with no significant difference for all plant age groups (Table 11, Test 3).
Table 8 evaluates this experience for the 36 plants excluding data from the five J
distinct data sets.
Again the major parameters of leakage events per inspection, i
corrosion events per inspection, and corrosion events per plant show a strong reduction in rate of occurrence as the age of the plant increases.
f With the five distinct data sets deleted, the difference between age of plants j
is significant for the four age groups (Table 11, Tests 22, 23).
Also, the pro-i portion of corrosion in leaking connections for plants of 4 years of age and less is a' statistically significant. difference (Table 11, Test 24).
These data from Tables 4 and 8 also are plotted on Figure 1.
i i
3.7 Leakage and Degradation Events Relative to Type of Component i
The results shown in Table 5 for comparison of leakage and corrosion events l
between types of components suggest that very little difference exists between types of components as to rate of occurrence of leakage (Table 11, Tests 4, 5, and 6).
Data from Table 9, making the same comparison for 36 plants, without the data from the five distinct data sets, appears to indicate a lower rate of occurrence of leaks and corrosion in steam generator manways.
This difference is not significant (Table 11, Test 25).
The data of Table 5 show that for valves there is a significantly lower percentage of reported corrosion problems in con-nections that experience leakage (Table 11, Test 7).
It has been suggested that j
this may be attributed to the selection of more corrosion resistant bolting mate-rial for some valves, but detailed information is not available to support this I
hypothesis at this time.
I l
NUREG-1095 3-3
~
3.8 Leakage and Degradation Events Relative to the Type of Lubricants Used Table 6 was developed to examine the relation of lubricants, especially molybdenum disulfide (MoS ), on the occurrence of leakage and corrosion events 2
for 41 plants.
The limited amount of data for events where only MoS2 was speci-fied as the lubricant used show that 43% of the joints inspected have reported evidence of leakage and corrosion.
However these data, for MoS2 only, do not meet the criterion for statistical analysis of significance (Table 11, Test 8) because of the small number of connections inspected, not because of the small number of leaking connections.
This same relation is found for the correlation of nickel graphite and graphite-isopropyl alcohol lubricants with leaks (Test 9).
[
However, by combining the results from reported use of MoS2 only, graphite-alcohol only, and the use of either of these two combined (Columns 2, 3, and 4, respectively, of Table 6), both the number of connections inspected and the number of leaks reported are large enough that a valid test can be made.
The correla-tion has a very high level of significance, a <0.01 (Table 11, Test 10), when tested against the results reported for the use of nickel graphite lubricants.
l l
Several additional tests were made on the basis of apportioning the reported i
results from approved multiple lubricants to either MoS 2 or graphite-alcohol, as appropriate.
Apportioning was on the basis of all, one-half or one quarter of the results as shown in Table 11, Tests 11, 12, 13, 14, and 15.
All the combi-nations had sufficient data for a valid test, and all correlations found a significant difference from that of nickel graphite lubricant.
3.9 Plant-to-Plant Variations f
Significant differences were found in the response from groups of plants as shown 4
in Table 7.
The difference is consistent between past and recent leakage events.
The five distinct data sets have 50% of the total recent leakage events and 59%
i of the total recent corrosion events for 12% of the total inspections and 12%
of the 41 plants reporting.
In attempting to analyze for statistical signifi-cance in the difference between these plants and the others, the X2 value was found to be significant at the 0.01 level for each of the five individually com-l pared to the remaining 40 plants.
However, none of the five plants reported i
enough inspections performed to satisfy the criterion for a valid test with
)
greater than three expected leaks, though they all had four or more leaks re-ported.
Table 11 shaws that, as a set, the combined data from these plants is significantly different from the data from the remaining 36 plants and the vali-l dity test is satisfied (Table 11, Tests 20, 21).
A sixth plant was considered for inclusion in these distinct data sets.
This t
sixth plant reported 41 inspections with seven leaks, about average for the l
other 12 plants that reported leaks.
The comparison of the data from this plant with combined data from the other 40 plants satisfied the criterion for analysis and showed that the differences were not significant at the 0.10 level.
Data from this plant was not included in the distinct data sets.
Data from one of the five plants was corrected during the development of this report as the result of an NRC inspection.
Correction of the data removed six connection inspections, five leakage incidents, and five corrosion events from
(
the nickel graphite column.
Four inspections, four leaks, and four corrosion l
t l
NUREG-1095 3-4 l
~
d events were attributed to the use of graphite-alcohol lubricant and the based lubricant.
Tables 6 and 10 show the corrected data.
remainder to MoS2 Uith the changes to the data noted above, there appeared a correlation of the data on leakage incidents in these plants with the lubricants reported as approved for use.
These correlations were evaluated in Tests 16, 17, 18 and 19 (see Table 11).
Correlations were found to be significant for all lubricants except nickel graphite.
However, the one plant in the set of five which reported leaks and corrosion problems with the use of nickel graphite is also one of the plants listed as a new plant (the 0-4 years operation).
The data from the 36 remaining plants shows that the age of the plant correlates significantly with leaks and with corrosion (Table 11, Tests 23 and 24).
However, tests for corre-lation of leaks with lubricants for these 36 remaining plants show no significant differences (Table 11, Tests 26 and 27).
Most data supporting the correlation of I
leaks with lubricants is not included "ith these 36 plants but is deleted with the 5 distinct data sets.
Also no sign'
' difference is found between leaks and steam generators relative to other
.ents for these plants (Table 11, Test 25).
3.10 Combined Effect of Plant Age and Lubricant Used The significant correlation of plant age and of lubricants with the number of leaks and corrosion events reported can be examined for combined results.
As shown on Table 6, only 44 of the 46 leaking connections can be related to a specific lubricant.
Of these 44 leaks, 16 leaks were in new plants, 7 leaks i
from two new plants authorizing use of nickel graphite as the lubricant on 27 con-nections, 8 leaks from two new plants authorizing use of either nickel graphite or graphite-alcohol (Table 6, Column 6) on 46 connections, and 1 leak from one new plant authorizing use of several lubricants including MoS2 on 7 connections.
Of the remaining 28 leaks, Table 6 shows 6 leaks related to use of MoS lubricants i
2 only (Column 2), 8 leaks related to graphite-alcohol lubricants only (Column 3),
and 3 leaks related to use of either MoS or alcohol lubricants (Column 4).
Also 2
seven leaks may be related to the authorized use of several lubricants including MoS -
2 Overall 40 of the 44 leaking connections can be related to new plants or to the authorized use of MoS2 or graphite-alcohol as a lubricant.
Further, 33 of the 44 leaking connections can be related to lubricants and new plants without the question on the use of MoS2 (Table 6, Column 5).
Also, 33 of the 44 leaks can be related to approved use of MoS2 or graphite-alcohol lubricants, including use with other lubricants, without including data from new plants.
This same trend may be observed in the reported data for past events.
Overall 75% of the 184 past leaks reported may be related to the use of molybdenum disulfide, or graphite-alcohol, or the age of the plant; 55 past leaks have been recorded in new plants (1977 or later); 41 past leaks reported with use of only MoS ; 7 leaks with the use of only graphite alcohol; 2 leaks in connections where 2
MoS2 or graphite-alcohol have been approved for use; 15 leaks where multiple lubricants have been approved including MoS2 or graphite-alcohol.
The remaining 46 past leaks (25%) occurred on connections where only nickel graphite and/or copper graphite lubricants have been used, and the plants were more than 4 years old.
NUREG-1095 3-5
(
b 3.11 Physical Basis for Observed Effects of Age and Lubricants The close correlation between the number of leaks and both the age of a plant and the lubricants used does not appear to be based on any obvious cause and effect.
The correlation with age, as an, inverse relation, can be postulated as a learning curve.
It is to be expected that plant maintenance crews in new plants will improve their performance and reduce the probability of leakage in any given connection, and other similar connections,'as they gain experience.
The correlation between lubricants and leaks leads to conjecture as to'the domi-nant causative physical characteristics and mechanisms.
No basis can be seen in the corrosive tendencies of the lubricant for causing leaks to initiate.
One characteristic of lubricants that can have a strong effect on joint tight-ness is the lubricating quality that= establishes the coefficient of friction, or friction factor, between the bolt and the other connection components.
The friction factor is inversely related to the tightening torque.
The resulting tension in the bolt and the joint tightness are thus affected for bolting pro-cedures based on torque setting.
Use of a low value for the friction factor in establishing torque can result in lack of adequate tension in the bolts.. Use of a lubricant with wide variations in the values of friction factor, dependent' on the supplier or the surface condition of the threaded and bearing areas, can result in wide variation in joint tightness.
Section 3.0 of NUREG/CR-3766 presents results of coefficient of friction testing
of 11 lubricants, 4 of these were MoS2 from four sources, 2 were graphite based, 3 were copper graphite based, and 2 were nickel graphite based.
The results of y
the tests show a range of values of greater than 10 for the static coefficient of friction of the MoS lubricant.
The range of values was,almost 3 for the 2
graphite based lubricants and almost 2 for both copper-based and nickel-based lubricants.
The greater variation in values cf the coefficient' of friction for MoS2 and griphite-based lubricants could be an explanation for the greater pro-portion of leaks reported for these lubricants.
f i
/
7
. e NUREG-1095 3-6
l e
i 4 CONCLUSIONS (1) Reporting in response to IE Bulletin 82-02 should be considered complete as regards any further reporting or evaluation of data.
There is no intent to reevaluate the total set of data as more responses to the bulletin are received.
t (2)
In general, the quality of the data considered herein cannot justify i
recommending actions with significant impact.
Any recommendations or j
actions proposed on the basis of results developed and presented in this report should be evaluated to ensure that a sound technical basis exists to predict the results of such action.
Also, impact in plant costs, personnel time, and radiation exposure incurred should be commensurate with the improvements in safety.
(3) The dominant mechanism for degradation of threaded fasteners in the primary l
coolant system of pressurized water reactors is general corrosion wastage and pitting rather than stress corrosion cracking.
Given sufficient time there appears to be almost a 100% probability of corrosion of bolting in a leaking connection.
(4) On the basis of data evaluated herein, the rate of occurrence of leaking 4
and corroded connections with threaded fasteners in primary coolant systems is much higher than has been assumed in previous risk studies for most plants.
However, the rate of occurrence of small loss-of-coolant accidents relative to the occurrence of leaks appears to be lower than has been assumed.
No conclusions are drawn on the basis of this data as to the overall probable rate of occurrence of a small LOCA.
(5) With 24 of the 41 plants reporting no recent leakage incidents and 5 of the 41 plants reporting one-half of the events, it appears that any safety prob-lems associated with leaking and corrosion of bolted connections is not widet,pread.
Much of industry appears to be able to get this problem under control.
i (6) On the basis of the significant differences of the five distinct data sets, there appears to be significant differences between plants as to the rate of occurrence of leaking and corroded connections.
No qualitative compar-ison of plants should be based on the data in Appendix B because of the 1
quality of the data reported in response to the bulletin.
The data contains a potential for misinterpretation based on such unreported factors as the selection of connections for the inspection and standards for accepting and reporting of corrosion of bolting.
The single reported factor which i
relates these 5 plants is the dominant use of MoS2 and graphite-alcohol based lubricants.
1 (7) There is a significant decrease in the rate of occurrence of leaking and corroded connections as the plants accumulate operating service time.
This 3
may be attributed to improvements in design and maintenance practice as plant crews gain experier.ce with their plant.
If the basis for this v
4-1 NUREG-1095
1
/
c improvement is made available as guidance for both manufacturers and plant maintenance staffs, the benefits of experience might be attained earlier in new plants.
Considering the cumulative effect of 24 of 41 plants reporting no recent leakage incidents, and the apparent learning curve, it appears that much of the industry has something worthwhile to share.
4 (8) There is some difference between types of components in the r$ ported data on leakage and corrosion.
Data on valves show a lower rate of occurrence of corrosion of fasteners relative to leakage.
Licensees may benefit from reviewing the types of bolting materials in valves to determine if improved alloys for replacement bolting can reduce their overall problems with bolting corrosion in other components.
(9) The data reported'from recent inspections do not show a significant rela-tion between the lubricant used and the incidence of corrosion in connec-tions that have developed leaks.
That data do not provide a basis for conclusions regarding the effect of molybdenum disulfide (MoS ) based 2
lubricants in contributing to problems of corrosion wastage and pitting relative to the incidence of leaks.
It appears that, except for some valves, most joints that develop leaks also will experience corrosion of the fasteners.
(10) The use of molydenum disulfide and graphite-alcohol based lubricants is concluded to be the dominant factor in the development of leaks in bolted connections. Without specific data on the lubricants and the bolting pro-cedures used for the plants which make use of these lubricants, no recom-mendations for change of lubricants or procedures can be justified.
How-ever, a recommendation can be justified that licensees perform a detailed critical review of their practices associated with the use of these lubricants.
(11) Bolt replacement is a common occurrence in connections that are inspected.
It does not appear that replacement of bolts with bolts having improved mechanical properties or corrosion resistance would have additional impact.
Replacement because of mechanical damage resulting from handling and disassembly and reassembly of the component connection appears to be a significant factor in the overall need Xor spare or replacement bolting.
Removal of bolts for the sole purpose of inspection should be avoided, if possible, by use of improved nondestructJve examination procedures.
(12) Consideration should be given to requiring all PWR plants to incorporate into their inspection programs, on an accelerated basis, the recent ASME Code provisions on inspection of bolted joints, including the acceptance standards for corrosion wastage.
The present requirements of the Code of Federal Regulations and the ASME Code allow as much as 10 years delay before implementing such provisions.
(13) The lack of inclusion herein of analysis of reported data on components such as pressurizer heater and control rod drive flanges should not be mis-interpreted in regard to the potential for problems associated with those components.
(14) The lack of reported data on the recent use of sealants does not allow evaluation of its effects on halting corrosion.
However, it appears that NUREG-1095 4-2
I any procedure that can obscure leakage, justify its continued acceptance, or otherwise allow prolonged exposure of, threaded fasteners to highly con-centrated boric acid solutions has the potential for developing an unsafe situation.
(15) The dominance of pitting and wastage incidents reported, relative to the
-lack of reports of stress corrosion cracking in coolant boundary connec-tions, should not be construed as a basis for avoiding consideration of the cracking problem.
l' l
I l
l NUREG-1095 4-3
5 RECOMMENDATIONS (1) Licensees that use molybdenum disulfide or graphite-alcohol based lubricants on bolted pressure boundary connections should perform a detailed critical review of their practices with the use of these lubricants with the aim of reducing the number of leaks which develop in these bolted connections.
(2) On the basis of the significant differences in bolting corrosion problems between groups of plants and the appearance of the beneficial effect of the learning curve, it is recommended that representatives from licensees share information on design and on maintenance procedures and practice.
This should be done with the objective of developing publicly available guidance or standards that could be used by both equipment manufacturers and plant maintenance staff to minimize the occurrence of leaking joints.
(3)
Industry groups should encourage the development and adoption of improved nondestructive examination methods that would allow in place inspection of bolts for damage from corrosion wastage as well as cracks.
(4) NRC should consider requiring licensees to include in their in-service inspection programs, at an accelerated pace, the ASME Code Section XI provisions for inspection of bolting and the associated acceptance standards.
(5) Licensees should continue to improve methods of detecting leakage at bolted connections to minimize the number of connections where leakage is allowed to persist for prolonged periods.
l NUREG-1095 5-1
6 REFERENCES American Society of Mechanical Engineers (ASME), " Boiler and Pressure Vessel C6de,"Section XI, " Rules for Inservice Inspection of Nuclear Power Plant Components."
U. S. Nuclear Regulatory Commission, NUREG-0933, "A Prioritization of Generic Safety Issues," December 1983.
U. S. Nuclear Regulatory Commission, NUREG-0943, Koo, W. H., " Threaded Fastener Experience in Nuclear Power Plants," January 1983.
U. S. Nuclear Regulatory Commission, NUREG/CR-2827, Czajkowski, C., " Boric Acid Corrosion of Ferritic Reactor Components," July 1982.
U. S. Nuclear Regulatory Commission, NUREG/CR-3766, Czajkowski, C., " Testing of Nuclear Grade Lubricants and Their Effect on AS40 B24 and A193 B7 Bolting Materials," Draft Report March 1984.
NUREG-1095 6-1
I APPENDIX A IE BULLETIN 82-02 NUREG-1095
1 SSINS No.:
3150-0086 Expiration Date:
5/30/86 l
UNITED STATES NUCLEAR REGULATORY COMMISSION OFFICE OF INSPECTION AND ENFORCEMENT WASHINGTON, D.C.
20555 June 2, 1982 IE BULLETIN N0. 82-02:
DEGRADATION OF THREADED FASTENERS IN THE REACTOR COOLANT PRESSURE COUNDARY OF PWR PLANTS Addressees:
All pressurized water nuclear power reactor facilities holding an operating license (0L), for action.
All other nuclear power reactor facilities holding an operatingslicense or construction permit (CP), for information.
Purpose:
The purpose of this bulletin is to:
(1) notify licensees and construction permit holders about incidents of severe degradation of threaded fasteners (bolts and studs) in closures in the reactor coolant pressure boundary (RCPB),
and (2) to require appropriate actions.
A response to this bulletin is required from pressurized water reactors (PWRs) holding an operating license as discussed below.
Description of Circumstances:
In May 1980, Omaha Public Power District (OPPD) submitted a special maintenance report to the NRC about the significant corrosion wastage experienced with closure studs in the reactor coolant pumps at its Fort Calhoun facility.
The corrosion wastage was attributed to boric acid attack as a result of leakage at flexita!1ic gasketed joints between the pump casing and pump cover.
These closure studs are 3.5 inches in diameter, and are manufactured of SA 193-B7 (AISI 4140) low-alloy, high strength steel.
Accordingly, the NRC issued Information Notice No. 80-27 on June 11, 1980 to all PWR licensees about the potential for undetected boric acid corrosion wastage and emphasized the need for supplemental visual inspection of pressure-retaining bolting in pump and
[
valve components.
Subsequently, similar occurrences of corrosion wastage from borated water leakage have been identified at other PWR plants, as discussed below.
l On March 10, 1982, the NRC was notified by Maine Yankee Atomic Power Company i
and Combustion Engineering (C-E) that during routine disassembly of a steam l
generator primary manway at Maine Yankee, 6 of the 20 manway closure studs failed and another 5 were found, by ultrasonic examination using specialized techniques, to be cracked.
Leakage had been noted from this manway during the current operating cycle and several efforts were made to eliminate the leakage.
These efforts involved increasing the joint operating compression through NUREG-1095 A-1 r
IE8 82-02 June 2, 1982 Page 2 of 5 torquing the studs to hydrotest levels and repeatedly injecting Furmanite sealant.
Normal plant operation continued until a planned maintenance outage.
Preliminary results of a metallurgical analysis C-E performed on the affected studs have indicated that the failure mode was stress-corrosion cracking (SCC).
By Information Notice No. 82-06 (issued March 12, 1982), the Office of Inspec-tion and Enforcement notified all licensees and construction permit holders about this degradation to emphasize the increased potential for studs to fail by the joint action of stud preload, material conditions and a corrosive environment generated by the presence of primary coolant leakage.
As a follow-up to the information notice, the utility established that the root cause of leakage was due to an interference contact between the gasket retainer lip and vessel cladding which prevented proper compression of the flexitallic gasket during reinstallation of the manway cover.
This problem was corrected and all 20 studs were replaced.
Magnetic particle and ultrasonic examinations of the studs in manways of the other two steam generators identified no other failures.
In the last several years a significant number of incidents have been reported of bolts and studs that have failed or become severeif degraded because of boric acid corrosion wastage or SCC mechanisms.
Prel minary results of an NRC staff review of threaded fastener experience in operating nuclear power plants have identified that specific generic actions need to ce taken before the study is complete.
The staff review identified 44 incidents of threaded fastener degradation since 1964.
From Table 1 it can be seen that since 1977, 15 incidents related to primary coolant pressure boundary application have been recorded.
These incidents involved 9 PWR plants.
Of concern is that degrada-tion and failure of such threaded fasteners constitute a potential loss of RCPB integrity and, in the extreme case, a loss-of coolant accident could occur, should extensive fastener failures in a pressure retaining closure not be detected.
In some instances, it has been reported that sealant compounds have been injected into bolted closures in the RCPB as a means of convenient maintenance to control leakage.
A review of the limited chemical analysis available on Furmanite indicates it has a variable composition with respect to concentration of chlorine, fluorine, and sulfur which are leachable and well recognized promoters of SCC.
Consequently, prolonged exposure of this sealant to leakage and high temperature conditions causing a gradual release of its potentially corrosive ions must be taken into account.
Also, certain lubricants may be formulated with molybdenum disulfide (MoS )
2 which contains a significant level of sulfide constituent.
Experience suggests that MoS has a pronounced tendency to decompose in the presence of high 2
temperature and moisture conditions to release sulfide which is a known promoter of SCC.
Therefore, care should be exercised in the selection and application of lubri-cants and injection sealants to minimize the risk of SCC from potentially corrosive ions due to the gradual breakdown and/or synergistic interaction of such materials with prolonged exposure to leakage conditions.
This would be of NUREG-1095 A-2
IEB 82-02 June 2, 1982 Page 3 of 5 particular concern for fastener materials made of high-strength low-alloy steels and, austenitic and martensitic stainless steels (i.e., 304, 316, 416, 17-4 PH, etc.) which are known to be susceptible to halogen / sulfide SCC degrada-tion.
The above concerns are further compounded by the fact that under the present ASME Code Section XI inservice inspection rules ultrasonic examination is not required on threaded fasteners in sizes 2 inches and less in diameter (e.g.,
Table IWB-2500-1).
However, except for the reactor coolant pump stud wastage, most failures have occurred in fastener sizes 2 inches and smaller. Further-more, experience has clearly shown that Code-specified ultrasonic testing (UT) methods are not singularly adequate to detect corrosion wastage conditions.
Moreover, the present Code UT procedures are not sufficiently sensitive to detect initiation of stress corrosion cracking (SCC) but requires the use of specialized UT techniques and calibration standards based on notch reflectors simulating critical flaw parameters to enhance reliability of detection.
At the present time, visual examination (e.g., IWA 2210, VT-1) appears to be the only method to detect borated water corrosion wastage or erosion-corrosion damage and may require insulation removal and/or disassembly of the component, in some cases, in order to have direct visual access to the threaded fasteners.
Therefore, degradation could go undetected when there is no clea evidence of leakage in the surrounding area.
Similarly, the reliability of visual examina-tion alone is questionable in detection of SCC initiation of t'areaded fasteners either in-situ or removed.
Accordingly, it is necessary that a combination of nondestructive examination techniques (UT, VT-1, MT, PT) be r.mployed to the maximum extent practical to enhance detection of the degradation mechanisms discussed above.
Actions To Be Taken by PWR Facilities Holding Operating Licenses:
The scope of action items listed below is limited to the RCPB.
Included are the threaded fasteners (studs or bolts) in (1) steam generator and pressurizer manway closures, (2) valve bonnets, and pump flange connections installed on lines having a nominal diameter of 6 inches or greater and (3) control rod drive (CRD) flange and pressurizer heater connections that do not have seal welds to provide leak-tight integrity. That is, CRDs having an omega seal weld design are excluded from this bulletin action.
The reactor vessel head closure studs are also excluded for those PWR licensees committed to the provisions of Regulatory Guide 1.65, " Materials and Inspection for Reactor Vessel Closure Studs."
Action Item 1 is to be completed prior to the performance of the subsequent action' items.
Action Item 2 is to be performed within the next cycle, but no later than the completion of the next refueling outage that is initiated after 60 days from the date of this bulletin.
The report requested by Action Item 3 is to be submitted within 60 days from the date of this bulletin.
1.
Where procedures do not exist, develop and implement maintenance procedures for threaded fastener practices.
These procedures should NUREG-1095 A-3
IEB 82-02 June 2, 1982 Page 4 of 5 include, but not limited to the following: (1) maintenance crew training of proper bolting / stud practices, tools application, specifications and requirements, (2) detensioning and retensioning practices (torque itera-tion), specified tolerances, and other controls for disassembly and reassembly of component closure / seal connections, (3) gasket installation and controls, and (4) retensioning methods and other measures to eliminate reactor coolant leakage during operations.
Quality assurance measures should also be established for proper selection, procurement, and application of fastener lubricants and injection sealant compounds to minimize fastener susceptibility to SCC environments.
2.
Threaded fasteners of closure connections, identified in the scope of this bulletin, when opened for component inspection or maintenance shall be removed *, cleaned, and inspected per IWA-2210 and IW4-2220 of ASME Code Section XI (1974 edition or later) before being reused.
3.
NRC Information Notice Nos. 80-27 and 82-06, and similar INP0 (Institute of Nuclear Power Operations) correspondence (with recommendations) have been issued in regard to corrosion problems associated with bolts / studs in RCPB closures (INP0/NSAC SER 81-12).
To assist the Nuclear Regulatory Commis-sion in its ongoing review and assessment of the scope of the problem you are asked to provide the following information for closures and connections within the scope of this bulletin:
Identify those bolted closures of the RCPB that have experienced a.
leakage, particularly those locations where leakage occurred during the most recent plant operating cycle.
Describe the inspections made and corrective measures taken to eliminate the problem.
If the leakage was attributed to gasket failure or its design, so indicate.
b.
Identify those closures and connections, if any, where fastener lubricants and injection sealant materials have been or are being i
i used and report on plant experience with their application particu-larly any instances of SCC of fasteners.
Include types and composi-tion of materials used.
4.
A written report signed under oath or affirmation under provisions of Section 182a, Atomic Energy Act of 1954 as amended, shall be submitted to the Regional Administrator of the appropriate NRC Regional Office within 60 days following the completion of the outage during which Action Item 2 was performed.
The report is to include:
A statement that Action Item 1 has been completed.
a.
- Fasteners " seized" or designed with interference fit, may be inspected in place.
NUREG-1095 A-4
IEB 82-02 June 2, 1982 Page 5 of 5 b.
. Identification of the specific connections examined as required by Action Item 2.
The results of the examinations performed on the threaded fasteners c.
as required by Action Item 2.
If no degradation was observed for a particular connection, a statement to that effect, identification of the connection and, whether the fasteners were examined in place or removed is all that is required.
If degradation was observed, the report should provide detailed information.
5.
A written report signed under oath or affirmation under provisions of Section 182a, Atomic Energy Act of 1954 as amended, shall be submitted to the Regional Administrator of the appropriate NRC Regional Office within 60 days of the date of this bulletin.
The report is to provide the information requested by Action Item 3.
Potential occupational exposure of personnel as a result of the above requirements should be considered in the program formulation process in an effort to maintain incurred exposures as low as reasonably achievable.
Personnel exposure-savings techniques such as use of steam generator primary manway cover-handling fixtures offer substantial time and man-rem savings.
This request for information was approved by the Office of Management and Budget under clearance number 3150-0086.
Comments on burden and duplication should be directed to the Office of Management and Budget, Reports Management, Room 3208, New Executive Office Building, Washington, D.C.
20503.
While no specific request or requirement is intended, the following information would be helpful to the NRC in evaluating the cost of this bulletin:
1.
Staff time to perform requested inspection.
2.
Radiation exposure attributed to requested inspections.
3.
Staff time spent to prepare written responses.
If you have any questions regarding this matter, please contact the Regional Administrator of the appropriate NRC Regional Office, or this office.
Richard C. DeYoung, Director Office of Inspection and Enforcement Technical
Contact:
W. J. Collins 301-492-4780 Attachments:
1.
Table 1 2.
List of Recently Issued IE Bulletins NUREG-1095 A-5
Attachment z
June 2, 1982 e
1 j
TABLE 1.
SufflARY OF DEGRADED TilREADED FASTENERS IN REACTOR COOLANT PRESSURE 8OUNDARY Degraded Reactor Coolant Pressure No. of Reported Plants (Year Incident Boundary Threaded Fasteners incidents Reported) & Reactor. Vendor Mode of Failure
- Pressurizer manway closure studs 2
Calvert Cliffs 2 (1981) C-E BC St. Lucie 1 (1978) C-E BC Steam generator manway closure 7
Maine Yankee (1982) C-E SC L
studs Oconee 3 (1980) B&W SC T
Arkansas 1 (1978) B&W BC Arkansas 1'(1980) B&W SC Calvert C11ffi I f1980) C-E BC St. Lucie 1 (1977j C-E BC l
San Onofre 1 (1977) W_
SC Reactor coolant pump closure 5
Ft. Calhoun (1980? C-E 8C studs Calvert C11ffs 1 BC Calvert Cliffs 2 ((1980) C-E i
1981) C-E 8C Oconee 3 1981 B&W BC Oconee 2 1981 B&W BC Safety injectjon check valve 1
Calvert Cliffs 2 (1981) C-E BC I
studs
- SC = stress corrosion; BC = borated water corrosion.
I
N IEB.82-02 June 2, 1982
.i.
' LIST OF RECENTLY ISSUED IE BULLETINS Bulletin Date of No.
Subject Issue Issued to Alteration of Radiographs of 05/07/82 All power reactor 82-01 4
Rev. 1 Welds in Piping Subassemblies facilities with an OL or CP 82-01 Alteration of Radiographs of 03/31/82 The Table 1 Welds in Piping Subassemblies facilities for action and to all others for information 02 Failure of Gate Type Valves 08/18/81 All power reactor Supplement to Close against Differential facilities with an
- 1 Pressure OL or CP 81-03 Flow Blockage of Cooling Water 04/10/81 All power reactor To Safety System Components by facilities with an CORBICULA SP. (ASIATIC CLAM)
and MYTILUS SP. (MUSSEL) 81-02 Failure of Gate Type Valves 04/09/81 All power reactor to Close Against Differential facilities with an Pressure OL or CP
' 81-01 Surveillance of Mechanical 03/04/81 Specific power reactor Rev. 1 Snubbers
- facilities with a CP 80-17 Failure of Control Rods to 02/13/81 To all specified BWRs Supp. 5 to Insert During a Scram with an OL & All at a BWR BWRs with a.CP 81-01 Surveillance of Mechanical 01/27/81 All power reactor Snubbers facilities with an OL and selected
[
power reactor facilities with a CP
_ 0L = Operating License CP = Construction Permit 1
NUREG-1095 A-7
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APPENDIX C L
CONTINGENCY TABLE ANALYSIS - STEP BY STEP a
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1.
Define task.
Test the hypothesis that two criteria (rows and columns) are independent.
2.
Decide on the risk you are willing to take in claiming the rows and i
columns are not independent when they are independent indeed.
This risk, written as probability (or percentage, when multiplied by 100) is denoted as u.
(e.g., a = 0.10.)
3.
With the help of a statistician, determine how many observations (n) you must take (e.g., n = 100).
4.
Collect data.
Each observation is tallied in a table of rows (r) and columns (c).
5.
Complete the contingency table with its marginal sums.
6.
For each of the rc cells determine the observed count 0, and the expected count E.
The expected count is computed as row sum multiplied by column sum and divided by the grand sum.
7.
Complete the following calculation for each of the rc cells:
0,E,(0-E),(0-E)2, (0-E)2/E 8.
Add up the values in the (0-E)2/E of Step 7 above.
Call this quantity 2
chi-square (X ),
9.
Determine the degrees of freedom for this chi-square as df = (r-1)(c-1).
2 under row df and column a.
Interpolation may 10.
Look up table value for X
~be necessary.
2 11.
If your computer X exceeds the table value, you may claim that rows and columns are not independent (p < 0.10).
2 12.
If your computed X does not exceed the table value, at best you can say that you have no evidence to claim that rows and columns are dependent.
t NUREG-1095 C-1
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Row r
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13.
Beabsolutely}surethatnoindividualiscountedinmorethanonecell 14.
Let 0 and E93 represent the count and the expected count in the cell of 93 th th the i row and the j column, respectively.
Then, mathematically write:
r c (0
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X"2 = y y
where v = (r-1)(c-1)
E i=1 j=1 ij 2
15.
The X ca be computed also as 0 2 1
0 i,j ij 2
16.
If r=c=2, there is a simpler formula for'X.
Denote the elements and the marginal sums of the contingency table as follows:
Column 1 Column 2 Sum Row 1 a
b e
Row 2 c
d f
Som g
h n
statistic is computed as X2=n
-bc)2 Then the X2 17.
Many authors caution against analyzing contingency tables wiiere the expected count of cell (s) is less than 5.
In many cases this situation can be corrected by a reasonable combination of rows and/or columns.
I NUREG-1095 C-2
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EVALUATION OF RESPONSE To lE BULLETIN 82-02 Degradation of Threaded Fasteners in Reactor Coolant Pressure Boundary of Pressurized Water-Reactor Plants
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A IE.Bulletin 82-02 was issued by the NRC oit June 2,1982 to notify licensees about incidents ia insta.ct am....o g
s of severe degradation of threaded f asteners[ Responses to the Bulletin from 41 PWR licensee included data from recent regular inspectionst of reactor coolant pressure boundary Statistical analysis is used to components connections of six-inch size and larger. factors related to frequ'ency of leakage inc d2termine significant Factors occurrence of degradation of bolts and spuds, and the need for bolt replacci ant.
types of lomponents, use of lubricants and sealants, examined include the age of the plant,The ep*mpiled dat'a indicate that, on the average,10% of and dif ferences between plants.
3 the bolted connections which were ins %cted show e%dence of leaking and 80% of those A significant decrease in the occurrence of undergo some degradation of the bolting.
Valves appear te bolting degradation eve.nts as the age' of the plant iricreases is observed.
A grup of 5 of the 41 plants accounted for about be less subject to bolting corrosiod.
t one-half of the reported leakage afd corrosion events. \\ The common characteristic found The usa of nickel-graphite based for 4 of these 5 plants was the lubricant used.
g lubricants appears to offer a significantly reduced incidence of leakage and _ corrosion; while use of molybdenuni disulfidejbased lubricants and gra'phite-based lubricants !!PPe --
e to result in a significantly incJeased incidence of leakagh and corrosion.
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