ML20154B750
| ML20154B750 | |
| Person / Time | |
|---|---|
| Site: | Arkansas Nuclear  |
| Issue date: | 11/30/1996 |
| From: | Cofie N, Deardorff A, Giannuzzi A STRUCTURAL INTEGRITY ASSOCIATES, INC. |
| To: | |
| Shared Package | |
| ML20154A031 | List: |
| References | |
| CON-EPRI-110A EPRI-110A-401, SIR-96-097, SIR-96-97, NUDOCS 9810060005 | |
| Download: ML20154B750 (25) | |
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1 Repon No.: SIR-96-097 Revision No.: 0 1
Project No.: EPRI-l10A l
File No.: EPRI-l10A-401 November 1996 Review of Degradation Mechanisms in EPRI Risk - Informed Inservice Evaluation Procedure i
Prepared for:
Electric Power Research Institute Prepared by:
StructuralIntegrity Associates,Inc.
"!1 95 Reviewers:
Date:
N. G. Cofie 9A Date:
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hd. J. Licina 9810060005 980925 ATTACHMENT 5
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a Table of Contents Section Eagg
1.0 INTRODUCTION
1-1 2.0 TECHNICAL APPROACH.......
.. 2-1 3.0 RECOMMENDATIONS.....
. 3-1 3.1 Thermal Fatigue.
3-1 3.2 Vibrational Fatigue....
3-2 3.3 Corrosion-Related Mechanisms......
3-2 4.0 RESOLUTION OF COMMENTS AND CONCLUSION
. 4-1
5.0 REFERENCES
. 5-1 APPENDIX A Resume's ofReviewers..
. A-0 APPENDIX B Degradation Mechanisms Table Presented in EPRI TR-106706...
. B-0 APPENDIX C Supplement to Degradation Mechanism Table Presented in EPRI TR-106706 C-0 APPENDIX D SI Recommended Table of Degradation Mechanisms....
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1.0 INTRODUCTION
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This repon documents the findings of an independent third party review performed by Structural Integrity Associates (SI) on the degradation mechanisms considered in the EPRI risk-informed in service inspection procedure for application to nuclear power plant piping components. The degradation mechanisms and other aspects of the EPRI risk-informed process are discussed in EPRI-TR-106706 [1]. Recognizing the wide range and different forms of degradation mechanisms associated with nuclear plant piping, a team of experts with various backgrounds in the nuclear industry were assembled to review the various attributes of the degradation mechanisms considered in Reference 1 for accuracy, completeness and applicability. The resumes of the reviewers are presented in Appendix A of this repon and their expenise is briefly summarized below, N. G. Cofie, Ph.D. - Over 15 years of experience in the nuclear industry. Expert in fatigue, stress and fracture mechanics analysis. Active panicipant on AShE Section XI on issues relating to pipe and vessel flaw evaluation.
A. F.
Deardorff,
' M. S, P. E - Over 20 years experience in the nuclear ir.dustry. Expert in thermal hydraulics, fatigue, stress and fracture mechanics. He is well knov a in AShE Section XI for its contributions on fatigue, pipe wall thinning and vessel integrity issues. He was an active participant in the preparation of the EPRI fatigue management handbook and the EPRI TASCS program.
A. J. Giannut.::1, Ph.D. P.E - Over 20 years experience in the nuclear industry. Expert in corrosion, metallurgy, material selection, repair and replacement. Authority on imergranular stress corrosion cracking (IGSCC) and various mitigating practices to prevent IGSCC. Active panicipant in AShE Section XI on repair of degraded piping components.
G. J. Licina, RS, - Over 20 years experience in the nuclear industry. Expen in corrosion, metallurgy and material selection. Authority on microbiologically influenced corrosion (MIC) and other forms oflocalized corrosion cuch as pitting.
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Appendix B contains a copy ofTable 4-1 from TR-106706, listing all of the degradation mechanisms considered in the document, the criteria for assessing whether the mechanism is potentially active for the piping system being evaluated, and the materials, product forms, or specific locations where the mechanism is likely to be operative. The list was constructed from the listing of pipe failures in Commercial U.S. Nuclear Power Plants compiled by Jamali [2, 3]. The list is presented in a slightly different form in Appendix C.
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2.0 TECHNICAL APPROACII The independent review was based upon the degradation mechanisms that are known to affect piping systems in nuclear plants using a similar " coarse screen" as was constructed for EPRI report NP-5461 (4]. The review thus provided a look at the degradation mechanisms "from the ground up" and incorporated expert experience in dealing with the gamut ofdegradations that occur in nuclear plants.
The review also evaluated the listing in EPRI TR-106706 base:1 upon its implementation to selected systems at ANO, the PWR pilot plant in the Risk Informed Inspection study. The utility's attempts to apply the list and criteria to their plant systems indicated that some categories were too broad,.
some criteria were unclear, and that the only mechanism deemed capable of producing a large leak was Flow Accelerated Corrosion (FAC). This review also compared the EPRI TR-106706 results to those from the so-called "B-J Code Case", which has been approved by ASME Code,Section XI and has been subjected to a very extensive review by the industry at large. That Code Case had similar objectives to the work described in both NP-5461 and TR-106706. Finally, the reviewers interacted on several occasions with EPRI, Sartrex (one of the preparers of EPRI TR-106706) and personnel from one of the on-going pilot plant studies (ANO) soliciting their input so as to produce a list of degradation mechanisms that had some level of agreement from all of the involved parties.
The method provided in TR-lM706 for evaluating the applicability of the listed mechanisms are essentially binary in nature. That is, a degradation mechanism is considered potentially operative or not; no " shades ofgray" are permitted. Because of the binary nature of the degradation mechanism assessment, the criteria provided should be, and in most cases, are conservative to assure that a potentially operative mechanism will not be overlooked during the initial screening. We find this approach to be reasonable and meaningful for an iniial screening of the degradation mechanisms, where the primary objective is completeness.
Criteria and methodology that consider all components in a system and all of the potential operating conditions, both normal and off-normal, would render that first cut or coarse screening too complicated. As such, our evaluation of the criteria took the binary nature of the procedure at this first level into consideration.
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The EPRI process, however, goes beyond this binary approach in the element selection process in that a more " continuous" type process is used to select the most susceptible locations for inspection after i
j the initial degradation mechanism and consequence evaluation. This approach was found to be very reasonable and practical because it limits the degree of damage mechanism assessment on the " front end" of the evaluation for systems or subsystems where those mechanisms will eventually be found to be inconsequential to failure after the initial evaluation. However, the approach allows the inspection to be focussed at areas of a system that are considered the most susceptible to any identified damage mechanism. The review also took into consideration this element selection aspect of the EPRI risk-informed ISI process.
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l 3.0 RECOMMENDATIONS l
On the whole, the review by the various participants concluded that the degradation mechanisms outlined in EPRI TR-106706 capture most of the active mechanisms that potentially affect nuclear power plant piping. A few additional observations are discussed below with respect to thermal fatigue, vibrational fatigue and corrosion-related mechanisms. In addition, a reorganization of some j
of the mechanisms is proposed.
3.1 Thermal Fatigue l
The thermal fatigue section of the degradation mechanisms in EPRI-TR-106706 was appropriately subdivided into thermal stratification, cycling and striping (TASCS), and thermal transients. We believe that this subdivision is appropriate to distinguish the relatively " low cycle" thermal transient events that have typically been designed for in the piping Stress Reports and the TASCS events that are associated with high cycles and generally not accounted for in the original piping design. The basis for the proposed TASCS criteria is that from the EPRI fatigue management handbook [5] and the EPRI TASCS report [6]. These criteria arejudged to be sound. Based on input received from the ongoing pilot programs, the criteria in this section were clarified, taking guidance from the fatigue 1
management handbook.
l We find the section on thermal transients to be reasonably conservative for this binary approach.
During the element selection process, however, the user can use the severity of the thermal transients and the frequency ofoccunence to determine the most susceptible locations to inspect and this could, in fact, provide the basis for eliminating some thermal transients as potential sources of degradation.
We find the criteria based solely on the temperature differential for the thermal transients acceptable.
It is believed, however, that there may be a few cases where there is potential for hot water injection or reflood into a cold component which will initially result in compression stress on the inside of the pipe. Fatigue usage is insensitive to the sign of the stress. Moveover, if the temperature is large l
enough, this will cause compressive yielding which will subsequently result in tensile residual stresses on the inside of the pipe once the temperature differential is removed. This, of course, will involve SIR-96-097, Rev. 0 3-1 StructuralIntegrity Associates, Inc.
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l a much higher AT than that currently specified in EPRI TR-106706. For this binary approach, we recommend that the AT specified for the thermal transient be converted to absolute numbers to account for the possibility ofinside surface tensile stresses developing from a hot fluid on a cold pipe.
This approach is conservative for the initial screening. However, the user may choose to use the actual temperature differential to provide realistic assessment of the level of compressive stresses compared to the yield strength of the component.
i 3.2 Vibrational Fatigue Though a very common failure mechanism in nuclear power plant piping, vibrational fatigue was not specifically made part of the evaluation process in the EPRI risk-informed procedure. Most documented vibrational fatigue failures in power plant piping, however, indicate that they are restricted to socket welds in small bore piping (less than 2 inch nominal pipe size) which does not fall under current ASME Section XI volumetric inspection programs. Vibrational fatigue failures have not been observed in large bore piping welds. It is also well documented that ' ost of the damage m
in vibrational fatigue failures occurs in the initiation phase and once a crack forms, the propagation is so fast that failure of the component can occur very rapidly. As such, vibrationai i me failures cannot be avoided by the risk-informed ISI process being considered by EPRI or for tha;..atter by any ISI program. We, therefore, agree with the observation in the EPRI TR-106706 that vibrational fatigue should not be included in the risk-informed program but that it should be treated as an entirely separate program taking guidance from the work documented in the EPRI fatigue management handbook.
3.3 Corrosion-Related Mechanisms The majority of mechanisms listed in EPRI TR-106706 are related to corrosion. We believe that most of the corrosion-related mechanisms that are associated with nuclear plant piping components have been included in the report. However, for completeness, three additional degradation mechanisms were considered for possible inclusion into the list of mechanism. These mechanisms are pitting, general corrosion and galvanic corrosion. After carefulfonsideration of these mechanisms,
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only pitting was included in the potential list of mechanisms as discussed below.
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Even though this mechanism had presumably been considered under TR-106706 as part of microbiological influenced corrosion (MIC), pitting can occur without the presence of living organisms or organic material. This mecharlsm should be included in'the list of potential degradation mechanisms and can be combined with MIC and crevice corrosion under one general title called Localized Corrosion" 3.3.2 General Corrosion General corrosion occurs in ferritic piping and results in an essentially uniform wall loss around the circumference of the pipe. Though this is an active mechanism for ferritic piping, we consider it to
. be too broad to be included in the EPRI risk-informed process ' That is, general corrosion will be operative for all ferritic piping, thus, including it as a degradation mechanism would not really accomplish anything.- In general, Class 1 piping is designed with an allowance for general corrosion.
The major concern for a risk-informed ISI program would be to demonstrate that that corrosion allowance is adequate. Examinations of piping for pitting, crevice corrosion and flow assisted corrosion (FAC) would uncover general corrosion ifit occurs. As such, we believe that no special treatment of this mechanism is further required in the EPRI risk-informed process.
3.3.3 Galvanic Corrosion Galvanic corrosion occurs as a result of the potential difference developed in a conductive solution if two dissimilar materials are in contact either physically or through an external electrical circuit. The potential difference produces electron flow between the two materials and the corrosion rate for the less corrosion resistant material will be increased. Galvanic corrosion occurs in the vicinity of connection ofcarbon steel with stainless meel or other more noble metals in conducting solutions and as such, it is a potentially achieve mechanism for piping systems with dissimilar metal joints. Though an active mechanism, we believe that galvanic corrosion will be adequately addressed under other i
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forms oflocalized corrosion since the susceptible regions for localized corrosion due to galvanic i.
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a effects addresses fittings, welds, heat affected zone (HAZ), base metal and dissimilar metaljoints.
Hence no further specific evaluation is required for this mechanism.
3.3.4 Rearrangement of Corrosion-Related Mechanisms In order to streamline all the coirosion related mechanisms, a new organization of these mechanisms is proposed. This new organization is somewhat consistent with that used in ASME Code Case N-560 which was recently been approved by the ASME Boiler and Pressure Vessel Code Committee for Category B-J welds in class 1 piping. This Code Case has similar objectives to the EPRI risk-informed ISI process and has gone through extensive review by the industry.
Corrosion cracking, primary water stress corrosion cracking and intergranular stress corrosion cracking (from Table 4-1 ofEPRI TR-106706) are proposed to be combined under one general topic called " Stress Corrosion Cracking" with the following subsections:
Intergranular Stress Corrosion Cracking (IGSCC) -BWR IGSCC - PWR Transgranular Stress Corrosion Cracking (TGSCC)
External Chloride Stress Cracking Corrosion (ECSCC)
Primary Water Stress Corrosion Cracking (PWSCC)
A gene. category called " Localized Corrosion"is proposed with the following subsets:
MIC Pitting Crevice Corrosion l
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4.0 RESOLUTION OF COMMENTS AND CONCLUSION l:
Based on the observations made in Section 3.0 of this rcport, several discussions were held among i-l EPRI, SARTREX and some of the participants in the on-going pilot plant studies. Based on these discussions, and valuable input from these organizations, a final list of degradation mechanisms to be l
considered in the EPRI risk-informed process was established. This list shown in Appendix D of this report includes the resolution of all the issues discussed in Section 3 of this report.
This table, though different in arrangement, is not very different in content than the original !ist of mechanisms in Table 4-1 of TR-106706. The only new mechanism that appears in this new table is pitting which for most practical purposes was covered by MIC. All other mechanisms listed in this new table were also covered in Table 4-1 of TR-106706, albeit under a different general heading.
The attributes for the mechanisms remain similar to those identified ir. EPRI TR-106706. Additional clarification is provided to differentiate between oxidizing conditions and initiating contaminants which tend to exacerbate the corrosion phenomenon and potentially produce corrosive effects outside the range of parameters generally accepted for oxidizing conditions alone.
The Criteria and Susceptible Regions sections of the revised table are very similar to those in EPRI TR-106706 and are intended to be sufficiently general that all potentially active degradation mechanisms are considered in the binary, coarse screening but not so broad that degradation will be identified where it is unlikely to occur. The specific operating conditions for the system provide especially important information for comparison to the criteria defmed for corrosion mechanisms.
We conclude from this review that the mechanisms contained in EPRI TR-106706 are adequate for the purpose of the risk-informed ISI process. The new recommended table agreed upon with all participants of this project provides only an enhanced rearrangement of these mechanisms and does not materially change any of the attributes and criteria. Hence, even though some of the on-going pilot studies have been performed under the TR-106706 table, the use of the new recommended list of mechanisms will not invalidate any of the completed work performed under the original TR-r 106706 table.
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5.0 REFERENCES
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Gosselin, S. R., et al., " Risk-Informed Inservice Inspection Evaluation Procedure", EPRI-TR-106706, June 1996.
2.
Jamali, K., " Pipe Failures in U.S. Commercial Nuclear Power Plants", EPRI TR-100380, Electric Power Research Institute, Palo Alto, CA, July 1992.
3.
Jamali, K., " Pipe Failures Update Study", EPRI TR-100380, Electric Power Research Institute, Palo Alto, CA, April 1993.
4.
Copeland, J. F., et al.," Component Life Estimation: LWR Degradation Mechanisms", EPRI NP-f 461, September 1987.
1 5.
Riccardella, P. C., et al., "EPRI Fatigue Management Handbook", EPRI TR-104534, Electric Power Research Institute, Palo Alto, CA, December 1994.
6.
Roarty, D.H, et.al, " Thermal Stratification, Cycling, and Striping (TASCS)", EPRI TR-103581, Electric Power Research Institute, Palo Alto, CA, March 1994.
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l APPENDIX A 1
Resume's ofReviewers N. G. Cofie 1
A. F. Deardorff A. J. Giannuzzi G. J. Licina l.
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Dr. Nathaniel G. Cofie Associate l~
Education BS, Civil Engineering, University of Science & Technology (1975)
MS, Civil Engineering, Stanford University (1977)
Degree ofEngineer, Stanford University (1979)
PhD, Civil Engineering, Stanford University (1983)
Professional Associations -
Member - Amedcan Society of Mechanical Engineers (ASME)
Member - ASME Section XI Working Group on Pipe Flaw Evaluation Professional Experience 1990 to present Structural Integrity Associates, San Jose, CA Associate 1981 to 1990 NUTECH, San Jose, CA StaffConsultant 1979 to 1981 Stanford University, Palo Alto, CA Research Assistant 1977 to 1979 URS/ John A. Blume & Associates, San Francisco, CA Engineer 1975 to 1976 University of Science & Technology, Ghana Research Engineer Summary Dr. Cofie has been involved in engineering for nuclear power plant components and conventional structures since 1975. He is an expert on the inelastic modeling of materials for structural applications. He has considerable experience in the application of finite element analysis, fracture mechanics, leak-before-break analysis and fatigue analysis. He is well versed in the requirements of the ASME, AISC, ACI and UBC codes.
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3 N. G. Cofie Page 2 While at NUTECH, Dr. Cofie was the technical leader of the fracture mechanics group. He was involved with several aspects of stress corrosion problems in the BWR industry, including development and implementation ofinduction heating stress improvement (IHSI) and weld overlay repairs design and implementation.
He worked on leak-before-break analyses, feedwater nozzle cracking evaluations, and fatigue analyses of several nuclear power plant components. Beforejoining the fracture mechanics group, Dr. Co6e worked in the structural engineering group as a project engineer and individual contributor on several projects. Examples are BWR Mark I hydrodynamic and seismic loads evaluation of the vent system and suppression chamber, finite element analyses of reactor vessel and high pressure injection nozzles, structural analyses of flued head penetrations, and structural evaluation of spent fuel casks and canisters. He also served as project engineer in charge of several test programs, including a program to evaluate materials to be used in the design of a tension leg platform for off-shore stmetures.
As a research assistant at Stanford University, Dr. Cofie was involved with several projects in dealing with full-scale and component testing of structures under severe inelastic cyclic loading.
He was also involved with various material testing programs.
As an engineer with URS/ John A. Blume & As'sociates, Dr. Cofie was involved with development of finite element models for static and dynamic analyses of piping systems for nuclear power and chemical plants. He also gained experience in the structural analysis and design of pipe supports.
As a research engineer at the University of Science & Technology, Dr. Cofie participated in the design of various reinforced concrete, steel and timber structures. He performed geotechnical analysis and design of building foundations and retaining walls. He also served as a teaching instructor in structural analysis and design.
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l Arthur F.
Deardorff,
P. E.
Associate l
Education L
BS, Mechanical. Engineering, Oregon State University (1964)
MS, Mechanical Engineering, University of Arizona (1966)
Professional Associations l
Registered Mechanical Engineer, State of California American Nuclear Society American Society ofMechanical Engineers ASME Section XI Subcommittee Member - Working Group on Erosion-Corrosion Member - Task Group on Fatigue in Operating Plants Member - Task Group on Implementation ofRisk-Based Inspection Member - Sub-Group Water Cooled Systems Professional Experience 1987 to present.
Structural Integrity Associates, San Jose, CA L
Associate 1976 to 1987 NUTECH, San h se, CA i
Supervising Engir eer 1970 to 1976 General Atomic Co.mpany, San Dieno. CA SeniorEngineer 1966 to 1970 The Boeing Company, Seattle, WA Engineer Summary Mr.
Deardorff has been involved in specification,
design, analysis and testing of nuclear power plant l
. systems and structures since 1970. At StructuralIntegrity Associates, he has been actively involved in piejects related to fatigue monitoring, fatigue and fracture mechanics, erosion-corrosion, thermal-hydraulics, expert systems, ASME Section III design analysis, and other related topics. He has made y
major contributions in developing ASME Section XI methods and criteria for evaluating thinned piping and for assessing fatigue in operating nuclear plants.. He is actively involved in other l
= Section XI committee activities relating to inspecting and evaluating nuclear plant power plant L
components and systems. He has directed several fati ue monitoring projects and had developed H
l-many of the enhancements to the FatiguePro fatigue monitoring system. He has consulted to the Electric Power Research Institute in several major fatigue-related projects.
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A. F. DeardorfT I
Page 2 Prior to joining SI, he was involved in design, analysis and testing of nuclear plant vessels, piping and containment systems. He performed evaluations of piping systems for effects ofintergranular stress corrosion cracking and performed testing and analysis to develop new Induction Heating Stress Improvement (IHSI) techniques. He developed methodology for predicting leakage through i
i pipe cracks, and has performed leak-before-break evaluations. He has also performed loading and structural evaluations of containment structures for hydrodynamic effects. In the late 70's, he was deeply involved in the containment structure analysis and testing and in formulation of the industry's approach to resolution of the Boiling Water Reactor Mark I Containment issue.
Mr.
Deardorffs areas of expertise lie in the areas of fracture mechanics,
fluid mechanics and heat transfer, stress analysis, dynamics, ASME Boiler and Pressure Vessel Code applications, and reactor systems evaluation, with a strong academic background in thermal-hydraulics and fluid system. He t
has a good understanding ofboth Pressurized and Boiling Water Reactor systems and struciures and has been involved in several projects related to fossil-fired power plants. Over the years, he has developed the reputation of being able to provide practical engineering solutions to complicated problems involving mechanical / structural integrity issues.
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Dr. Anthony J. Giannuzzi, P. E.
Associate 1
Education BS, Physics, LeMoyne College (1964)
MS, Solid State. Science and Technology, Syracuse University (1967)
PhD, Solid State Science and Technology, Syracuse University (1969) 4 Professional Associations i
Professional Corrosion Engineer, State of Califomia American Society of Mechanical Engineers ASME Section XI Subcommittee Member - Working Group on Welding and Other Special Processes l
Professional Experience
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1983 to present StructuralIntegrity Associates, San Jose, CA Vice President 1979 to 1983' Electric Power Research Institute, Palo Alto, CA Project Manager 1978 to 1979 NUTECH, San Jose, CA Project Manager 1972 to 1978 General Electric Company, San Jose, CA PrincipalEngineer 1969 to 1972 Aerojet Nuclear Systems Company, Sacramento, CA Summary Dr. Giannuzzi has been involved in solving materials and corrosion problems for the nuclear industry since 1969. One of the world's leading authorities on intergranular stress corrosion cracking of stainless steel in aqueous systems, Dr. Giannuzzi was employed by the Electric Power Research Institute in the Nuclear Systems and Materials Department for three-and-one-half years
. prior tojoining StructuralIntegrity Associates in 1983. At EPRI, Dr. Giannuzzi was task leader and principal investigator involved in development and quali5 cation of all the Boiling Water Reactor IGSCC piping remedies. This activity included primary responsibility for qualifying and producing material specification for the alternative materials (Types 316NG and 304NG stainless steels),
qualifying the induction heating stress improvement (IHSI) remedy, qualifying heat sink welding, last pass heat sink welding and the weld overlay, and performing the investigations to determine the causes of and remedies to IGSCC in Type 304 stainless steel piping.
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In addition to his BWR IGSCC responsibility at EPRI, Dr. Giannuzzi has had the lead responsibility for investigating the causes oflow pressure large steam turbine stress corrosion cracking in nuclear l
ar.d fossil steam turbines and has been involved in projects associated with bolt and fastener reliability, steam and water piping erosion-corrosion and has been active in projects related to i
primary-side and secondary-side corrosion of steam generators. Dr. Giannuzzi has also been the l
lead project manager responsible for all materials-related failure analysis activities in the Nuclear Systems and Materials Department and was a member of the EPRI Three Mile Island Unit 2 task force.
Prior to his employment at EPRI, Dr. Giannuzzi was employed as a senior consultant at NUTECH.
While at NUTECH, he formed the stress corrosion cracking group and developed the methodology used to estimate likely locations ofIGSCC in stainless steel piping systems. He also w~as involved in the earliest investigations involving PWR boric acid corrosion and assisted in the final formulation of the NRC I-E Bulletin 79-02 which established criteria for inspection of the boric acid 4
system piping.
From 1972 to 1978, Dr. Giannuzzi worked as a principal development engineer at the General Electric Company Nuclear Energy Division. His resporsibilities while at GE involved investigation of alternative materials and processes to alleviate the IGSCC problem in stainless steel piping. He managed the initial weld residual stress measurement'and analyses activities which lead to the development of the residual stress remedies to IGSCC.
From 1969 to 1972, Dr. Giannuzzi worked for the Aerojet Nuclear Systems Company developing materials for use in the nuclear rocket engine (NERVA).
In 1983, Dr. Giannuzzi founded StructuralIntegrity Associates with Dr. P. C. Riccardella and Dr.
T. L. Gerber. His activities at Stmetural Integrity have included nuclear plant life extension studies, temper bead welding development on low alloy steels, and selecting of remedies to IGSCC in BWRs. Dr. Giannuzziis a member of the ASME Section XI Working Group on Welding and Other Special Processes and has chaired a Task Group on AJternative Repair Methods for Erosion-Corrosion Damage in Carbon Steel Piping. He is currently chairman of a Task Group on " Laser Welding of Steam Generator Tubes" for structural repairs of SG tubes.
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George Licina Associate l
Education BS, Metallurgical Engineering, University ofIllinois (With High Honors)
Graduate Work, Materials Science, San Jose State University Professional Associations & Awards Alpha Sigma Mu - Metallurgy Honorary Society Tau Beta Pi - Engineering Honorary Society Patent No. 4166019 - Electrochemical Oxygen Meter Patent No. 4139421 - Method ofDetermining Oxygen Content Patent No. 5246560 - Apparatus for Monitoring Biofilm Activity General Manager's Award - General Electric Company, Advanced Nuclear Systems Technology Operation,1984 Professional Experience 1986 - present Structural Integrity Associates, San Jose, CA Associate 1972 - 1986 General Electric Company, San Jose, CA Senior Engineer Summary Mr. Licina's experience at Structural Integrity Associates has dealt primarily with the degradation and environmental compatibility of power plant materials under a variety of operating conditions.
These degradation mechanisms include corrosion and environmentally assisted cracking in BWR, PWR, and various raw water environments and embrittlement of pressure vessel steels and high performance alloys. Mr. Licina is a recognized authority on microbiologically influenced corrosion and has authored reference documents on that topic for the Electric Power Research Institute and numerous utilities. Plant-specific activities include metallurgical and fracture mechanics evaluations of nuclear steam generators, heat exchangers, valve stem cracking, BWR pipe replacement. and irradiation embrittlement of reactor pressure vessels, and the use of electrochemical methods for predicting and monitoring corrosion in power plant environments.
Mr. Licina has also integrated technical and regulatory requirements into guidelines for field certification of materials in nuclear plants and developed a methodology and approach for nuclear life extension issues.
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G. J. Licina Page 2 He has authored more than thirty publications in these technology areas and is the author of two patents involving the determination of oxygen levels in liquid sodium systems and a third for an on-line method for monitoring bio 61m activity in cooling water environments.
Mr. Licina served as lead engineer and program manager on a number ofimportant development programs at the General Electric Company including:
Control Rod Blade Surveillance and Lifetime Evaluation Stress Corrosion Cracking of Cr-Mo Steels Carbon Transport Effects on Steels in Liquid Sodium Systems 1
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APPENDIX B Degradation Mechanisms Table Presented in EPRI TR-106706 i
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Degradation Mechanism Evaluation l
I Table 4.1 Degradation Mechanisms Degradation Mechanism Criteria Susceptible Regions Thermal fatigue (a) Thermal stratification' (a) Areas where hot and (a) Nozzles, branch pipe cold fluid can mix connections, safe cycling, stripping UASCS)
.where: Operating temp ends, welds, heat-
> 220*F(CS) or affected zones (HAZ),
i 270'F(SS), NPS >1 base metal, regions of I
inch stress concentration l
Vertical rise <45'F, and AT> 50*F or Richardson number >
4.0 (b) Nozzles, branch pipe (b) Thermal transient connections, safe (b) Operating temp >
ends, welds, HAZ, 220*F(CS) or base metal, regions of 200*F(SS), and AT>
stress concentration 150*F(CS) or p00*F(SS), and AT > T allowable Corrosion cracking (a) Chloride cracking (a) Areas exposed to (a) Base metal, welds and chloride contam,ination HAZ where temperatures
>150*F and tensile stresses (b) Crevice corrosion (b) Areas that contain (b) Base metal, welds and cracking crevices that can result HAZ in oxygen depletion and concentration of impurities Primary water stress Mill annealed Alloy 600 Nozzles, welds, HAZ corrosion cracking Cold worked or cold without stress relief, worked and welded thermowells
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Table 4.1 (cont.)
De5radation Mechanisms Degradation Mechanism Criteria Susceptible Regions Intergranular stress corrosion cracking (IGSCC)
(a) Generic Letter 88-01 (a) Austenitic steel welds (a) IGSCC - BWRs and HAZ (b) IGSCC - PWRs (b) High oxygen, stagnant (b) Austenitic steel welds flow and HAZ Microbiologically
. Presence or intrusion of Fittings, welds, HAZ, and influenced corrosion (MIC) organic material base metal, especially Untreated water regions containing Low flow crevices Operating temperatures of 20 to 120'F pH <10.
Erosion-cavitation
( p - p,)/ Ap < 5, and V Fittings, welds, HAZ, and
> 30ft/sec. and fluid base metal i
temperature < 250'F l
. EPRI TR-10318, T2, l
provides additional l
guidance Flow-accelerated Evaluated in accordance Evaluated in accordance corrosion (FAC) with plant FAC program with plant FAC program 1
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i APPENDIX C Supplement to Degradation Mechanism Table Presented in EPRI TR-106706 i
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1 SIR-96-097, Rev. O C-0 '
StructuralIntegrity Associates, Inc.
DEGRADATION MECHANISM WORKSHEET AprG3,1996
- 1. TMAMAL FATIGUE (TASCS & Tramients):
i Base metal and weld regions ARE NOT SUSCEPTIBLE to degradation from thermal fatigue if any of the criteria in both A and B are true:
A. ThermalStratulcadon, Cycling and Striping CrASCS)
There is no potential for low flow, and no pipe segments containing hot fluid connected to segments containing cold fluid, and no pipe segments corm-~~i to components containing steam, or Tbc pipe segment has a slope > 45* from horizontal (or elbow into a VERTICAL iPP*),
or NPS < 1 inch, or The calculated or measured AT < 50 F, or Re < 4, AND B. Transients The design or operating teraperature T < 270 F (S.S.) or T < 220 F (CS.),
or There is no potential for cold water injection onto a hot component, or AT < 200 F (S.S.),
or AT < 150 F (CS.),
or AT < AT allowable (per Fatigue Handbook)
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- 2. STRESS C01tROSION CRACKING (SCC):
A. GeneralSCC(Interna 0:
Welds and weld heat affected zones at the inner surface of austenitic stainless steel pipe ARE SUSCFIIIBLE to degradation from SCC if all of the following are true:
Stagnant, intenninent, or low flow, and Operating temperanues T > 120 F, and Water chemistry IS NOTmonitored.
and There is the potential for in leakage from + w] systems containing brackish or untreated water, or there is a history of or potential for contamination by chlorides,
- fluorides, sulfides,etc.
De outer surface of austenitic stainlas steel pipe within 3D of probable leak paths (e.g.
valves with stems) IS SUSCEITBLE to degadation from chloride SCC if the pipe in that region is covered with non-metallic insulation that IS NOT in compliance with Reg.
Guide 1.36.
The outer surface of austenitic stainless steel pipe IS SUSCE1TBLE to degradation from corrosion cracking if it is " ;-:=d to wering from chloride bearing environments such as sea or brackish water.
C. Crwice SCC:
Regions that ARE SUSCEFI'IBLE to crevice SCC include thermal sleeves as shown in Figures 5.2.2.2 & 5.2.2.3 of EPRI Report TR-106218.
- 3. PRBfARY WATER STPISS CORROSION CRACKING (PWSCC)
Inconel (alloy 600)IS SUSur.nusLE to degradation from PWSCC if both of the following are true:
De materialis mill-annealed and either cold worked or cold and welded without suess relief, and Is exposed to primary water at T > 620 F.
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INTERGRANULAR STRESS CORROSION CRACRTNG (IGSCQ A. BWRs Welds and weld heat affected zones in BWR piping ARE SUSCElmBLE to degradation
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from IGSCC if they are examined as part of the existhig plant IGSCC program, which was defined in acconiance with NRC Generic Letter 88-01, "NRC Position on IGSCC in BWR Austenitic m=inless Steel Piping'.
1 B. FWRs i
affected : tones in austenitic stainicas steel pipe in PWRa are==-wiible :o dWa%
from IGSCC. Accordingly, welds and weld heat m#er'ed zones in austenitic stainless steel pipe in PWRa ARE SUSCEPTIBLE to degradation from IGSCC if the following are true.
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The pipe segments are in sta6nant oxygenated borated water systems.
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The carbon content of the austenitic steel pipe material is equal to or greater than j
0.05 wt %, as determined form the material certification reports.
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- 1. The term " stagnant, oxygensted borated water system" refers to those systems serving j
as engmeered safeguards havmg no nortnal operating functions and contam:ng essend=11y air saturated borated water where dynamic flow conditions do not exist on a continuous basis. However, these systems must be maintained ready for actuation during normal l
power operation. Systems or portions of systems that are flushed at least once every
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three months need not be classified as stagnant for purpose of this evaluadon.
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- 5. MICROBIOLOGICALLYINFLUENCED COPJtOSION(MIC)
Carbon steel welds, weld heat =W zones and base metal, and austenitic steel welds and heat affec:cd zones ARE SUSCr.rumLE to degradation from MIC if the following are true:
The matenals are in raw water systems, transport systems, or storage tanks, or other systems containing untreated water where pH < 10, and There is low or intermittent flow, especially in regions of geometric discontinuities, and
'Ibe operating temperature is between 20 F and 120 F
- 6. EROSION-CAVTTATION Regions within 5D dews.=s of throttling or pressure reducing ralves or orifices, ARE SUSA.rsumu to degradation from e:osion-cavitation if all the following are true:
Operatmg temperature < 250 F, and Flow > 100 hrs /yr (approximately 2% of plant operating time),
'and V > 30 ft/s, and (P4 - P,) / AP < 5, where V = flow mean velocity at the inlet of the unit, Pa = static pressure downstream of the unit, P, = vapor pressure, and AP = pressure differemial ac:oss the unit.
- 7. FLOW ACr'wrFRATED CORROSION (FAC)
Regions in piping segments ARE SUSCEFIBG to degradation from FAC if they are
~=inM as part of the existing plant FAC program, which was :lefined in accordance with NRC generic leuer 89-08.
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i, 1-APPENDIX D l-SI Recommended Table of Degradation Mechanisms i
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t SIR-96-097, Rev. O D*0 f StructuralIntegrityAssociates,Inc.
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Revised Table 4.1for EPRI TR-106706 Degradation Mechanism Criteria and Susceptible ' 'egim Degradation Mechanism Crkeria Susuptible Regions TF TASCS
- nps > 1 inch, and no:-les. branch pipe
-pipe segment has a slope < 45*from hori:ontal(includes elbow or connections, safe ends.
tee into a verticalpipe), and welds, heat affected i
-- potential aistsfor lowflow in a pipe section connected to a
- ones (HAZ), base metal.
component allowing mixing ofhot and coldfluids. or andregions ofstress potential existsfo* leakageflowpast a valve (i.e., in-leakage out-concentration leakage, cross-leakage) allowing mixing ofhot and coldfluids. or potential aistsfor convection heating in dead-endedpipe sections connectedto a sor.rce ofhotfluid or potentialexistsfor two phase (steam / water) flow, or potential aistsfor turbulentpenetration in branch pipe connected
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to headerpiping containing hotfluid with high turbulentflow, and
-calculatedor measuredbT> $0*F, and
-Richardson number > 4.0 TT
- operating temperature > 270*Ffor stainless steel, or operating temperature > 220*Ffor carbon steel, and
-potentialfor relatively rapid temperature changes including coldfluidinjection into hotpipe segment, or hotfluid injection into coldpipe segment, and bT > 200*Ffor stainless steel, or AT > 150*Ffor carbon steel, or AT
> bTallowable (applicable to both stainless andcarbon)
- evaluated in accordance with exist.ng plant IGSCCprogram per austenitic stainless steel (BWR)
NRC Generic Lener 88-01 welds andHAZ IGSCC
- operating temperature > 200*F, and (PWR)
-susceptible material (carbon content 2 0.035%), and
- tensile stress (including residual stress) is present, and
- oxygen or axidi:ing species are present OR
- operating temperature < 200*F, the attributes above apply, and
-initiating contaminants (e g., thiosulfate, fluoride, chloride) are also regt:iredto bepresent TGSCC
-operating temperature > IS0*F, and austenitic stainless steel
- tensile stress (including residualstress) is present, and base metal, selds, and
- halides (e.g., fluoride, chloride) are present, or HAZ caustic (NaOH) is present, and
- oxygen or oxidi:ing species are present (only required to be present i:s conjunction w/ halides, not required w/ caustic)
Table Leyend ThermalFatigue (TF) locali:ed Corrosion TLC)
- ThermalStranpcanon. Cpimg. andSenpmg (TASCS)
- Alicrobiologically influenced Corrosson allC)
- Thermal Transiems ITT)
- Pittmg IPID Stress Corrosson Crackmg (SCC)
- Crevice Corrosuon tCC)
Intergranular Stress Corrosson Crackmg (IGSCC)
FlowSensarve tFS)
- Transgranular Stress Corrosion Crackmg (TGSCC)
- Erossant.nuanon (E-C)
- External Chlorude Stress Corrosion Crackmg (ECSCC)
- Flow Acceler erosson tFAC)
- Prtmarv Water Stress Corrosion Crackmg (PWSCC)
StructuralIntegrity Associates, Inc.
.S Revised Table 4.1for EPRI TR-106706 Degradation Mechanism Criteria and SusCentible Reaions Degradation Mechanison
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SCC ECSCC
-operating temperature > 130*F and austenitic stainless steel
-tensile stress is present, and base metal, welds, and
- an outside piping surface is withinfive diameters ofa probable leak HAZ path (e.g., valve stems) and is covered with non-metallic insulation that is not in compliance with Reg. Guide 1.36. or an outside piping surface is expossd to wettingfrom chloride bearing environments (e.g., seawater, brackish water, brine)
PWSCC -piping materialis inconel(Allov 600), and no::les, welds. andHAZ
-aposedtoprimary water at T> 620*F, and without stress relief
-the materialis mill-annealedandcold worked or cold worked and welded without stress relief l.C MIC
- operating temperature < 130*F, and fittings. welds. HAZ,
-low or intermittentflow, and base metal, dissimilar
-pH < 10, and metalfoints (e.g., welds.
-presence / intrusion oforganic material (e.g., raw watersystem), or flanges), andregions I
water source is not treated w/ biocides (e.g., refueling water tank) containing crevices PIT
-potentialaistsfor lowflow, and
- oxygen or oxidi:ing species are present, and
- initiating contaminants (e.g., fluoride, chloride) are present CC
-crevice condition aists (e.g., thertralsleeves), and
-operating temperature > 130*F, and
- oxygen or oxidi:ing species are present FS E-C
-operating temperature < 250*F, and fitsings, welds. HAZ, and
-flowpresent > 100 hrs /vr, and base metal
-velocity > 30ft/s, and
-(Ps - P,) / AP < $
-evaluatedin accordance with aistingplant FACprogram perplant FACprogram 1
i Table Lesvend ThermalFatigue (TF) localized Corroston (LC)
- krmalStratsfication. Cyctmg. andStrepung (TASCS)
- MicrobiologucalIv infuenced Corrosson (MIC)
- krmal Transwnts (TI)
Pitting (PIT)
Stress Corrosson Crachng (SCC)
- Crevice Corroston (CC)
- Intergranular Stress Corrosion Crackmg (IGSCC)
Flow Sensstive (FS)
Trantgranular Stress Corrosson Crachng (IGSCC)
- Erosson-Cavaatson (E C)
Esternal Chlonde Stress Corrosion Crackng (ECSCC)
- Flow decelera Corrosion IFAC)
Pnmary Water Stress Corrosson Crochng (PWSCC) gg,y,9y,,y g,y,,,jyy gggg, jpg,,g, jgg.
__ 3,SEP-14-SI_16:39_. _. _From:ANO GSB 1 _ _
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.. _. _ _ _ _ _.T-973 P.D UD3 M RAI Open Issues from Meeting on September 9-10,1998 RAINo.
lasues Identified / AdditionalInformation or Clarification Requested 1.0 More clearly derme alternative requesand to existmg Code requiremems (i.e., subsutution of Code Case N-578 using BPRI TR-106706) and specify timeframe (i.e., remainder of 2" Interval and ydInterval) for wtncts its applicable; Clarify content of Table 1 1 incloding crediting of existing plant FAC program for MFW and MSS 2.0 More clearly define contents of Table 21 and provide basis for scope ofsystems conandered 3.0 No changes identif=d 4.0 Provide a more explicit description of ths "cahancements" utfitzed in the ANO-2 applicaSza versus EPRI TR-106706 and Code Case N-578 5.0 Editorial comment - change " risk analysis" to "RI ISI" in last sentence Clarify meaning of " initial screening" versus " fall malysir; justify exclusion from ' full analysis" based upon 6.0 inclusian in existing FAC proaram and ensure considerat'on of other damage mechanisms 7.0 No changes identified Review critical floodmg water level in penetration room area; review pressure at which outside door latch is 8.0 muumed m fail, review timina of valve osotion; review whether door failures are credited for flood prevention elsewherein the analysis 9.0 No changes identified 10.0 Editorial comment - remove everythmg after crit notence Determme shther the ventilation dampers are credited as a flood prevention measure (propagation from General 11.0 Access area to ECCS pump rooma)in any of the consequence evaluations 12.0 Provide Sketch of penetration room area and addren impacts of pipe breaks in this area; verify 5 minute operator response time for breaks in ECCS pump rooms 13.0 Check recoveries forloss of SW eystem G7)
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14.0 No chmagesidernified 15.0 No ekaatesidentified i
16.0 Provide limited validation ruults Editorial comment-in the 2 paragraph rep! ace "ANO-2" evaluation with "$W" evaluation d
17.0 18.0 Provide limited vall%n results 19.0 Provide limited validation results I
20.0 No changes identified 21.0 Documem assumptions for support dependencies 22.0 ik i-;n a..^2'd revg3e,,3c,3s.\\o b we,rdi k
23.0 No chages identified 24.0 Describe controls for backup trains during SDC operation for midloop and other configurations; confum SOPP meets Rl ISI assumptions 25E Evaluate industry water hammer events fof applicability to ANO-2 using EPRUNRC guidelina 26.0 No changesidentified 27.0 Need input from staff ATTACHMENT 6 l of 2
, 3. SEP-14-98 16:39 From:ANO G5B 1 5018514815 T-973 P.03/03 Job-434 e
RAI Open Issues from Meeting on September 9-10,1998 Issues Identified / AdditionalInformation or Clarification Requested
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RAI No.
Editorial comment - in the " defense io depth" discus 6 ion clari.fy that "high" refers to the comegaance category 28.0 and not the risk rank; clarify that" poor performing equipment" refers to item with a hgh failure rase; clarify that high consequence or high failure potential always result in a nsk significant segm-st 29.0 No channesidentifled Bener define review process utilised including; 1) fau: tion performed by plant project team in the perfbrmance 30.0 and review of system calculations, and 2) independent, integrated plant review performed on completed sys calculations; expand on meaning of"all aspects"; remove implication that EPRI guidelines are not sutticient t casure identification of risk significant segments: reference internal plant validation / verification proceu 2
,i 31.0 No changesidentified 32.0 identify any existmg relief regnests for dinimilar metal welds i
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33.0 No e6 ages idad 34.0 No changes identified l
35.0 No chnges hiGed i
36.0 Expand upon reliance on SIR 96-097 for damage mechanism assessment f
37.0 No changes idea 6 fled i
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T-973 P.O FAX i,,,e gg, l Number ofpecesincluding centsheet 8 To:
BiLt Rccxzey rRou:
/rtis'c R 1a;u x' Arkansas Nude'ar One Entemy Operations 1448 S.R. 333 Russellville, AR 72801 Phone Phone (501) 858- @ 2 2 rax Phone 30/~ Y/.r-306}
FaxPhone (501) 858-4685 l CC:
REMARKS:
O Ument C Foryourreew Q ReplyASAP Q PleaseComment 3on y Si<.
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1 4-l ANO-2 RI-ISI Questions Clarifications orovided by the NRR staff to the licensee by facsimile on 09/17/98 I
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Below are the talking points regarding the two RAls whose resolution was not clarified Thursday the 10'th with ANO-2.
Information for RAI 27 - Delta Risk Methodology 1
- 1) What process did ANO-2 itself rely upon to review the method? Did you bring to bear any of your own internal expertise to approve the methodology?
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- 2) Request the executive summaries of the EdF benchmark and UM review.
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- 3) Request that the licensee state that they have found the work to be of sufficient quality and accuracy to support the conclusions drawn in the submitta!.
Information for RAI 12 - Human interactions
- 1) Please confirm that for all isolations credited that 1) there are Control Room alarms to which the operators respond by investigations or actions which would identify or confirm the leak,2) the response and/or isolation is directed by procedure, and 3) the isolation manipulations can be taken from the control room.
- 2) Please confirm that the, interaction between responding to an unrelated initiating event and an on-demand pipe rupture is considered during the evaluation to determine the appropriateness of crediting isolation as one full train.
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ATTACHMENT 7 I
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