ML20217F007
| ML20217F007 | |
| Person / Time | |
|---|---|
| Site: | Arkansas Nuclear  |
| Issue date: | 08/15/1997 |
| From: | Bauer A, Jaquith R, Weston R ABB COMBUSTION ENGINEERING NUCLEAR FUEL (FORMERLY, ENTERGY OPERATIONS, INC. |
| To: | |
| Shared Package | |
| ML20217E904 | List:
|
| References | |
| A-PENG-CALC-018, A-PENG-CALC-018-R00, A-PENG-CALC-18, A-PENG-CALC-18-R, NUDOCS 9710070339 | |
| Download: ML20217F007 (65) | |
Text
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, l Arkansas Nuclear One - Unit 2 Pi ot Plant Study l Risk-Informed Inservice Inspec1: ion Evaluation for the l Main Feedwater System September 1997
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N A-PENO-CALC-018. Rwision 00 II Design Analysi.s Title Page Page 1 of 39 (3
V
Title:
Implementation of the EPRI Risk Informed Inservice Inspection Evaluation Procedure for the MFW System at ANO-2 Document Number: A-PENG-CALC-018 Revision 00 Number:
Quality Class:
O QC 1(Safety Related) O QC 2 (Not Safety Related) @ QC 3 (Not Safety Related)
- 1. Approvalof Completed Analysis This Design Analysis is complete and verified. Management authorizes the use ofits results.
Printed Name Signature Date Cognizant Engineer (s) R. A. Weston [j [h]
A. V. Bauer h gh qq Mentor 9 None /j j f Independent Reviewer (s) R. E. Jaquith
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Management Approval B. T. Lubin g .
RI)5 c)}
Project Manager
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- 2. Package Contents (this section may be completed aner Management approval):
Total page count, including body, appendices, attachments, etc. 64 List associated CD-ROM disk Volume Numbers and path names: g None Note: CD-ROM are stored as separate Quality Records CD-ROM Volume Path Names (to lowest directory which uniquely applies to this document)
Numbers Total number of sheets of microfiche: g None Number of sheets-Other attachments (specify):
- 3. Distribution:
p B. Boya (2 copies)
U .
I:\d ata\lu bin \rbifinal\apeng018. doc
Ak I D'% FlklFIk Calculation No. A PENG CALC-018, Rev. 00 Page 2 of 39 RECORD OF REVISIONS Rev Date Pages Changed Prepared By Approved By 00 g l$ 61 ') Original R. A. Weston R. E. Jaquith A. V. Bauer k -
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1 1
g ABB Combustion Engineering Nuclear Operations 4
1 i
N N II 7"E BPIF Calculation No. A PENG CALC-018. Rev. 00 Page 3 of 39 TABLE OF CONTENTS SECTION PAGE 1.0 PURPOSE...............................................................................................................5
- 2. 0 SCOPE..................................................................................................................5
- 3. 0 SYSTEM IDENTIFICA TION AND BOUNDARY DEFINITION ............................................ 6
- 4. 0 C0NSEO UENCE EVA L UA TlON . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . I 1 4.1 CONSEO UENCE A SSUMP TIONS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 13 4.2 CONSEQ UENCE IDEN TIFICA TION . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 14 4.3 SHUTDOWN OPERA TION AND EXTERNAL EVENTS........................................... 14
- 5. 0 DEGRA DA TION MECHA NISMS EVA L UA TION . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 19 5.1 DAMA GE GRO UPS.. .... .... . ........ . ... ...............................................................20 5.2 DEGRADA TION MECHANISM CRITEr/A AND IDENTIFICA TION ........................... 21 5.3BASICDATA.................................................................................................27 6.0 SER VICE HIS TOR Y A ND SUSCEP TIBILITY REVIEW . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 29 7.0 RISK E VA L UA TIO N. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 32
- 8. 0 EL EMEN T SEL EC TlON . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 34
- 9. 0 REFERENCES........................................................................................................38
\
( LIST OF TABLES
% NUMBER PAGE 1 MFW S YS TEM BO UNOA RIES . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 7 2 MFW SYSTEM CONSEQUENCE ASSESSMENT
SUMMARY
......................................... 16 3A MFW SYSTEM CONSEQUENCE. FIGURES AND ISOMETRIC DRA WINGS....................... 17 3B MFW PIPING AND APPLICABLE CONSEQUENCE & LOCA TION.................................... 17 4 DA MA G E CR O UPS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 19 5 DEGRADA TION MECHANISM CRITERIA AND SUSCEPTIBLE REGIONS......................... 22 6 MAIN FEEDWA TER SYSTEM LINES AND OPERA TING CONDITf0NS............................. 28 7 SERVICE HISTORY AND SUSCEPTIBILITY REVIEW - MAIN FEEDWA TER SYSTEM ......... 31 8 RI SK SEGMEN T IDENTIFICA TION . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 33 9 RISK INSPEC TlON SC0PE. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 34 10 EL EMENT SELEC TION RISK CA TEG OR Y 3 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 35 m \
- f G
ABB Combustion Engineering Nuclear Operations
SI IIII A'%IFIF Calculation tvu. $PE G CALC 018, Rev. 00 l
Page 4 of 39 UST OF FIGURES NUAfBER EAGE 1A ANO-2 POWER CONVERSION SYSTEM (CONDENSA TE PORTION)..................... ............ 8 18 ANO-2 POWER CONVERSION SYSTEM (MFW PUMP PORTION)............................... .... 9 IC ANO-2 POWER CONVERSION SYSTEM (AfflV POR TION)................................... ........ 10 2 AfAIN FGED WA TER (MFW) FL O W PA THS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 18 LIST OF APPFNDICES A FMECA . CONSEQUENCE INFORMA TION REPORT B FA1ECA . DEGRADA TION AfECHANISMS C FAfECA SEGMENT RISK RANKING REPORT D OUALITY ASSURANCE VERIFICA TION FORMS O
O ABB Cornbustion Engineering Nuclear Operations
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- 'tIFIF Calculation No. A PENG-CALC-018, Rev. 00
- Page 5 of 39
- 1. 0 PURPOSE The purpose of this evaluation is to clocument the implementation of the Electric Power Resear:h Institute (EPRI) Risk-Informed Inservice Inspection Evaluation Procedure (RISI) of Reference 9.1 for the Main feedwater (MFWI systern at Arkansas Nuclear One, Unit 2 (ANO-2), Entergy Operations, Inc. The RISI evaluation process provides an alternative to the requirements in ASME Section XI for selecting inspection locations. The purpose of RISI is to identify risk-significant pipe segments, define the locations that are to be inspected within these segments, and identify appropriate inspection methods.
7his evaluation is performed using the guidelines of the EPRI Risk-Informed Inservice Inspection Evaluation Poocedure of Reference 9.1 and in accordance with the requirements of the ABB Combustion Engineering Nuclear Operations Quality Procedures Manual (OPM 101).
2.0 SCOPE This evaluation procedure applies to the MFW system at ANO-2, and utilizes the ISIS Software (Reference 9.2), which has been specifically developed to support and document this procedure.
,m \ As part 3 the procedure, the system boundaries and functions are identified. A risk
\'
evaluation is performed by dividing the system into piping segnoents which are determined to have the same failure consequences and degradation mechanisms. The failure consequences and degradation mechanisms are evaluated by assigning the segmer't to the appropriate risk category and identifying the risk significant segments. Finally, the inspection locations are selected. The guidelines used in determining the degradation mechanisms, the failure consequences and the risk-significant segments are those described
- in Reference 9.1.
O V
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- % FIN
( Calculation No. A PENG CALC-018, Rev. 00 O) Page 6 of 39 3.0 SYSTEM IDENTIFICA TION AND BOUNDARY DEFINITION
3.1 System Description
The Main feedwater (MFW) Systerr is designed to provide continuous feedwater supply to the two steam generators at the required pressures and temperatures under all anticipated steady state and transient conditions. In addition, the system provides isolation of the steam generators during main steam and feedwater line breaks, it has also been designed with the capability to maintain unit operation when the system has to operate under a.% normal conditions due to a feed pump, a condensate pump, a heater drain pump, or a heater or string of heaters, being out of service, it can continuously operate et approximately 80 percent plant output with Just one train of pumos and low pressure feedwater heaters.
3.2 System Boundary The Main Feedwater (MFW) system is described consistent with the FSAR (Reference 9.3).
The scope of this analysis includes all Class 2 piping in this system which is currently examined in the ANO 2, ASME Section XI Inservice Inspection (ISI) Program (Reference
- 9. 6). The code and non code lines which are part of or interface with the MFW system were evaluated to determine their risk significance. The system boundaries are defined in 0 Table I and Figures 1A,1B and IC. Certain line segments contain welds that were not O entered in the database (Reference 9.2) as outlined below:
3.2.1 Emergency feedwater lines upstream of emergency feedwater nozzles (2DBB-3-4",
2DBB-4-4")
These line segments are used to provide emergency feedwater tc the intact steam generator (s). The emergency feedwater line segments from upstream of the emergency feedwater nozzles are included as part of the Emergency Feedwater (EFW) system and are therefore not included in the evaluation of the MFW system.
3.2.2 Main feedwater lines upstream of valves 2CV-1023-2 and :CV 1073 2 (2DBD 24", 2DBD-2-24", 2DBD-3-24", 2DBD-4-24", 2FBD-3-24", 2FBD-4-24", 2FBD 24", 2FBD-2-24", 2FBD-1 1B", 2FBD-2-18", 2HBD-85-30", 2HBD-BS-30", 2HBD-85-34", 2HBD-86 34", 2HBD-85-24")
The above line segments are used to supply feedwater to the steam generators during normalpower operation. These line segments are cross-connected at various locations to facilitate the delivery of main feedwater to the steam generators following equipment malfunction or frilure, or a line break. Following a line break, continued delivery of feedwater to the steam generators is possible for a period of time. This would depend on operator actions and secondary system response, thus making the consequence of the transient less severe than the consequence caused by a break downstream of the main feedwater check valves.
A break in any of the above line segments is mitigated by the continued delivery of
\p) feedwater to the steam generators by the Emergency Feedwater (EFW) and Auxiliary ABB Combustion Engineering Nuclear Operations
A Ik R MIFIF Calculation No. A PENG CALC-018, Rev. OC Page 7 of 39 Feedwater systems via the EFWpiping. The above line segments are susceptible to Flow Accelerated Corrosion (FAC) and are already addressed by the existing plant FAC Program at ANO-2. Because the Risk Informed ISI methodology does not support granting inspection relief to elements within the plant FAC Program, the welds for these lines were not includedin the database.
3.2.3 Lines with Nominal Diameter of 1' or f.ess
, Piping with a nominal diameter of I' or less was not explicitly evolvated to determine its risk significance. Since volumetric examination of this piping is not practicable, the most effective means to ensure its integrity is via conduction of a system leakage tast. Consequently, since this piping is already subject to system leakage testing by the ASME Code, a risk assessment of this piping is not warranted, f
t TABLE 1 MFW SYSTEM BOUNDARIES Line Line ISI Pope Pope Nominal Number Desenption Drawing Code Diameter (in.)
Number Class 2DBB-l.18" Main Feedwater Supply Line to Steam Generator 2DBB-12 2 18 2E-24A - 18' piping 2DBB-1-24" Main Feedwater Supply Line to Steam Generator 2DBB-1-1 and 2 24 2E-24A - 24' piping 2DBB l 2 2DBB-2-18" Main Feedwater Supply Line to Steam Generator 2DBB-2-1 2 18 2E 24B - 18' piping i
2DBB-2 24" Main Feedwater Supply Line to Steam Generator 2DBB-21 and 2 24 2E-24B - 24" piping 2DBB-2 2 l
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FROW FROW 2CV-0748/0753 2CV-0740/0744 FIGURE IC: ANO 2 POWER CONVERSION SYSTEM (MFIV PORTION) (Reference 9.15)
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Calculation No. A-PENG. CALC-018, Rev, 00 O]
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Page 11 of 39
- 4. J CONSEQUENCE EVALUA TION The Main Feedwater (MFW) system provides continuous feedwater supp4 to the two steam generators at the required temperatures and pressures under all steady state conditions.
The system also has the capability to maintain plant operation with certain equipment out or service. It can continue to operate at approximately 80% power output with a single train of main feedwater pump and one string oflow pressure heaters. The MFW system consists of two trains which are interconnected at the main suction lines of the condenser hotweHs, at the condensate pump discharge, and at the suction and discharge sides of the main feedwater pumps. During normal operation, the condensate pumps take suction from the hotweHs and discharge the condensate into the two strings of low pressure feedwater heaters to supply the main feedwater pump suction. The main feedwater pumps then discharge the heated condensate to a common header before going through two independent strings of high pressure heaters. After exiting the high pressure heaters, the heated condensate is dischargedinto the steam generators.
The consequence evaluation for the MFW was performed based on the guidance provided in the EPRI procedure (Reference 9.1). The evaluation focused on the impact of a pipe segment failure on the capability of MFWsystem to perform its design functions, and on the overaH operation of the plant. Impacts due to direct and indirect effects were considered.
GeneraHy, the effects of a direct impact are confined to the MFW system itself. An indirect impact resulting from the failure of a pipe segment would affect neighboring equipment within the ^'bW system or other system (s). Indirect impacts would generaHy be caused by flooding, spraying, or jet impingement of neighboring equipment. Determination of the O consequences of a segment failure considers the poiential of losing affected mitigating systems, or trains thereof, and the consequentialimpact on the safety functions.
The spatial effects of a segment failure are primarily associated with flooding, spraying, or jet impingement. Plant locations as defined in the Internal Flood Screening Study (Reference 9.14) were used in this evaluation. The locations are summarized in Section 4.2.1 of Reference 9.17.
On November 19 and 20,1996, a walkdown was performed at ANO-2 to assess potential spatialinteractions associated with splashing, spraying, and flooding, including propagation paths. The foHowing individuals participated in the walkdown and meetings at ANO-2:
Jim Moody (YAEC Consultant)
Pat O'Regan (YAEC)
The plant was in an unexpected outage and radiological controls would not aHow access to the south piping penetration rooms (Rooms 2084 and 2055) and elevation 317'-0" (Rooms 2006, 2011, 2014, 2007, and 2010). However, this is not judged to have an impact on the analysis since spMial questions were answered for these areas during the visit. The focus of the walkdown was in those areas where analysis scope piping exists and their propagation paths andimpacts. The foHowing summarizes the walkdown observations:
p i 1 Y)
ABB Combustion Engineering Nuclear Operations
k Calculation No. A PENG-CALC-018, Rev. 00 .
Page 12 of 39 (a) North Piping Penetration (Room 2081) was walked down. EFW and main feedwater valves were identified in the room. Grating and a spiral stair ensures easy propagation to elevation 335 where there is a door with its latch removed (required due to high energy line break analysis) opening out to Room 2040. A door at elevation 354'O' opens into the room and provides access to a stairwell and the turbine auxiliary building. It would be difficult for floods to access Room 2081 from the turbine auxiliary building side. Propagation inta 2081 from 2040 would not affect any equipment as the EFWand MF valves are above the floor.
(b) Elevation 335' O' (Room 2040) is a very large area containing general access, corridors, and several large non safety related rooms. Several floor drains were noted. EFW and containment spray piping is located here, including RWT suction MOVs. The MOVs are located high off the floor in the tank room, protected from floods. Also, the EFW steam admission valve 2CV 0340-2 and 2SV-0205 is located behind a wall and sufficiently off the floor to be protected. Several rooms connect to this room from elevation 354' O' (Room 2073) and the piping penetration rooms (Rooms 2055 and 2084) at elevation 335'-O'. The most critical component identified in this area is MCC 2BS2 which powers several train A components.
Although the MCC is not near analysis scope piping, it is at the east stairway entrance (the propagation path to elevation 317' 0'). If six inches of water could be accumulated at elevation 335'-O' or if a very large pipe break occurred, it is considered likely that the MCC could fail.
The type of inputs used and the assumptions made in pstforming this evaluation are documented in Reference 9.17. Key inputs and assumptions, which are extracted from Reference 9.17, are provided in Section 4.1. All six consequence segments identified for MFW are assigned as ' MEDIUM". The consequence assessment summary for these segments is provided in Section 4.2. The bases and justifications for each category assignment are provided in Appendix A. This appendix contains reports obtained from the ISIS software (Reference 9.2) for the MFW system.
O ABB Combustion Engineering Nuclear Operations
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ABB Calculation No. A PENG CALC-018, Rev. 00 (Y Page 13 c? 39 4.1 CONSEQUENCE ASSUMPTIONS The following assumptions and inputs were extracted from Reference 9.17. The type of initiating events and mitigating capabilities considered in this evolvation are described in detailin Sections 4.3 and 4.4 of Reference 9.17.
4.1.1 Pipe follore can occur at anytime; three configurations have been defined as shown in Table 4-1 of Reference 9.17. These are normal (operating or standby), test, and accident demand. This table also summarizes j'adoments and assumptions regarding which configurations are most important. If pipe failure does not cause a direct initiating event, it is assumed that pipe faulJte Occurs during the accident demand configuration, if applicable. This assumes pipe inih.re occurs during the most conservative exposure time and accounts for the higher stress placed on the operators with resultant delay in operator response.
4.1.2 l'or main feedwater line breaks postulated outside containment, it is assumed that ERV supply valves in the north piping penetration room (2081) and AfCC 2BS2 at
\ the for end of the corridor in Room 2040 are effected. ERY and ECCS equipment l are in protected rooms and assumed not to be affected. The turbine ERV steam supply isolation valve in Room 2040 is protected behind a well and is assumed to function lopen successfully to initiate ERV A).
4.1.3 AfCCs at elevation 354' O' (Room 2073) and elevation 335' 0* (Room 2040) are assumed to fail if water accumulates to a height of 6 in:hes et the AfCCs. At
& elevation 354' O*, there is a large grated opening to ?levation 335' 0* on the west end of the building. Thus, it is assumed that 'eaks in this analysis scope can not accumulate 6 inches at this location. It takes a significant time to flood to a level of 6 inches at elevation 335'-O', but it is assumed that fal lure in isolate results in MCC failure.
4.1. 4 The IPE internal flooding study identifies impacts in rooms fron' cable terminal po:nts. Since mostjunction boxes, terminalboxes, etc., noted during the wuokdown were at least a few feet off the floor, these impacts were ignored in the analysis.
Also, junction boxes appeared to be t'pht and sealed, therefore, even if water reached them, an electrical fault appeared unlikely.
4.1.5 AfCC 2852 impacts are assumed to occur wather isolation occurs or not for RWT suction (20 inch pipe) and feedwater line brecks; this is conservative.
4.1. 6 Steam generator tube integrity is no* assumed to be lost during main steam and feedwater transients or pipe breaks.
4.1. 7 The IPEEE (extemal hazards analysis) assessment neglects (1) the potential impacts of relay chatter from relays with unknown capacity (possible optimism, although this is scheduled to be resolved), (2) en improvement if the seismic capacity of EDG tanks is increased, cnd (3) a detailed review of fire scenarios (not provided in the IPEEE). These are notjudged to significantly impact the analysis results.
p 4.1. 8 The main steam isolation signal (A1 SIS if either steam generat:t pressure decreases Q to 751 psis) isolates main steam and feedwater. It also removes EFAS allowing ABB Combustion Engineering Nuclear Operations i
ABB Calculation No. A PENG CALC 018, Rev. 00 Page 14 of 39 EF'V discharge paths to close. Once the plant protection system (PPS) has determined the affected steam generator, EFAS to unaffected steam generator returns. The unaffected steam generator is the one that has a 90 psis higher pressure. If the pressure in both generators increases to > 751 psis then the ERV volves willcycle on steam generatorlevelonly.
4.1.9 According to Table 3 2 of Reference 9.1, the unreliabHity of unaffected backup trains is as follows:
zero backup train
- 1.0 one backup train
- 1.0E 2 two backup trains
- 1.0E 4 three or more backup trains
- 1.0E 6 The probabHity of not performing corrective actions based on adequate information in the control room is typically 1.0E.2 (Reference 9.18). Therefore, failun of the operators to isolate a segment is treated as equivalent to one backup train.
- 4. 2 CONSEQUENCE IDENTIFICATION l The consequence summary assessment is provided in tabular form in this section. Simplified l schematics are provided in Figure 2 to iHustrate the boundaries for each of the MAV l consequences. Dotted lines are used to identify the boundaries for each consequence.
Afajor MRV equipment are shown on this figure for ease of identification. Table 2 summarizes the consequence evaluation for the A1RV system. (Refer to Section 5.2 of Reference 9. t 7.)
- The bases and justifications for each of the assigned consequences are documented in Appendix A. The ISIS (Reference 9.2) software was used as a tool to prepare the documentation in this appendix. The documentation of the spatial effects are currently based on a review of the Intemal flood Screening Study (Reference 9.14) and the walkdown that was conducted for the A1AV system. The walkdown captured subtle interactions which could not be readily identified using only the Intemal flood Screening study.
Observations from the wcikdown are factored into the consequence evaluation. Table 3A presents the A1RV consequences, their corresponding figure numbers and Isometric Drawings. In addition, Table 3B identifies the pipe line numbers and their corresponding locations.
4.3 SHUTDOWN OPERATION AND EXTERNAL EVENTS Shutdown Operation The consequence evaluation is on assessment assuming the plant is at-power. GeneraHy, the at-power plant configuration is considered to present the greatest risk for piping failures since the plant requires immediate response to satisfy reactivity control, heat removal, and inventory control. By satisfying these safety functions, the plant win be shut down and maintained in a stable state. At-power, the plant is critical, and is at higher pressure and tempe?sture in comparison to shutdown operation. The current version of the methodology (Reference 9.1) provides no guidance on consequence evaluation during shutdown operation. This limitation is assessed herein to gain some level of confidence that the consequence ranking during shutdown would not be more limiting.
ABB Combustion Engineering Nuclear Operations
ABB p Calculation No. A PENG CALC 018, Rev. 00 G Page 15 of 39 Pipe segments that are aircady ranked.~s 'HIGH' consequence from the evaluation at power need not be evaluated for shutdown. Those that are already 'AfEDIUAf* require some confidence that 'HIGH' would not occur due to shutdown configurations. However, the
- LOW" consequences for power operation requirn more confidence that a 'HIGH' would not occur and some confidence that a *A1EOlUA1' consequence would not occur. Recognising this, a review and cornparison of system consequence .*esults for power operation versus potential consequence during shutdown operation was conducted.
The results of the comparison Indicate that during at power operation, the AfRV segments are ranked as *A4EOlUA1*. During shutdown operation, the A1AV system is not in operation andis not required for decay heat removal. The Ainv consequence ranking during at power operation is therefore considered to be bounding.
Lxtemel Events Although extemal events are not addressed in the current version of the methodology (Reference 9.1), the potential importance of piping failures during extemal event is also considered. The ANO 2 IPEEE was reviewed to determine whether extemal initiating events, with their potential common cause impacts on mitigating systems, could affect consequence ranking. This information, along with information from other extemal event PRAs, is considered to derive insights and confidence that consequence ranking is not more significant during an extemal event. The following summarizes the review for each of the 1 major harords (seismic, fire and others).
' O O Seismic Challenges - The potential effects of seismic initiating events on consequence ranking is assessed by considering the frequency of challenging plant mitigating systems and the potentialimpact on the existing consequence ranking. The most likely scenario is seistnic induced loss of offsite power which also results in loss of the AinV system. A 7 seismic failure of MAY piping would be of little or no significance. Therefore, the MnV l consequence ranking at power is considered to be bounding for the ranking during a seismic event.
Fire Challenges The ANO 2 IPEEE indicates that the fire core damage frequency is dominated by fires initiated outside the containment. The most likely fire induced core damage scenario involves a loss of offsite power. For this scenario, the AfRV system would also be lost. Fallare of a AfRV pipe segment would be of little or no significance.
Therefore, t*w at power ranko for AfRVbounds the ranking during a fire event.
Other External ChaHcngos Ott.cr hasards were screened in the ANO 2 IPEEE and are assumed to have little or no risk s!gtnificant in1 pact on ERV.
O O _
ABB Combustion' Engineering Nuclear Operations 1
NNS MWW Calculation No. A-PENG-CALC-018, Rev. 00 Page 16of 39 TaNe 2 MFW Consequence Assessment Summary K1 Descrption Spetief Configureren kwtietor iscletion System BecAsp Centamment &posaae TeNe Used Rarat locetion knveces Trekan Yew tref. 9.171 FW-C-01A MF to SG *A" Contenuwent Operating 75 No PCS (T51 & NA (Jheffected NA 2-1 MEDel/M downsueern of ETW to SG 2FW5A *A
- doe to 75 FW-C-01B MF to SG ~B" Contemment Operating 75 No PCS(TSI& NA (knoffected NA 2-1 MEDIVnt dowrnsweem of ETH to SG 2FW58 *B" doe to 75 FW-C-02A MF to SG 'A* Contomment Operateg 75 til 2FW5A & PCS (T51 NA 2CV-1024 NA 2-1 MEDetint osaeorn of 2CV-!O24 oartsade 2FW5A FW-C-02B MF to SG *B" Contomment Operating T5 til 2FW 5B & PCSLT51 NA 2CV-1074 NA 2-1 MEZWnt wstreern of 2CV 1074 oartsade 2FW58 FW-C-03A MF to SG *A~ 2O61 Operating 7 5 (11 2FWSA & PCS (T51. 2 (EFW & 21W-5A NA 22 MEDIUnt cartside 2CV-1024 Trein A ECCS AFWto SG humide dscharge "A ~ end vm^res Treks B Chee W
- c.% ,.
veAery FW-CC38 MF to SG ~B" 2081 Operating 75 til 2FW58 & PCS (T51. 2 (EFW & 2FW-58 NA 24 MEDIUM outside 2CV-1074 Train A ECCS AFW to SG haside dscherge "A
- and
& Treen B Cnce W
dscherce verres9 ill Brooks answeens of chect volves 2FW-5A and 58 condd be classifn' d es T2 er T6 initiating events, but the conse:1uence remets are ret effected-ABB Combustion Engin ng Nuclear Operations
ABB Co/Culation No. A PENG CALC 018, ReV. 00 Page 17 of 39 Table 3A MAY System Consequence, figures and Isometric Drawings ConsequenceID figure Number isometric Drawings nV C-01 A 2 2D8812 RV C-018 2 2D88 2 2 AV-C-02A 2 2D8812 nV-C-028 2 2D88 2 2 nV C-03A 2 2088 1 1 n V-C-038 2 2D88 21 Table 38 Mnv Piping and Applicable Conseovence & Location Pope Deecreten Cornsequence Losotron 2DBB 1 from 2CV 1024.I to Steam Generator 2E-24A 01A,02A Contamment 2CV 1024-1 in 2081 03A 2081 HH 2088-2 from 2CV 1014 I to Strom Generator 2E 248 018, 028 Conteinment 2CV 1014-1 in 2081 038 2081 HH 2D801 trum 2CV 1023 2 to 2CV 10241 not m scope 2081 HH (19 2DBD-2 from 2CV.t073 2 to 2CV 1014-I not m scope 2081 HH til A
I \ Ill Mein feedwater snelysis scope is frorn 2CV 10241 and 2CV 10141 to the steem generators. 2CV 1023 2 and N 2CV 1073 2 are located outside 2081 m the turtdne buddmg and this piping as not in scope.
/^x V
ABB Combustion Engineering Nuclear Operations
ABB Calculation No. A PENG CALC 018, Rev 00 Page 18 of 39 r..........,...................,
' MFW T,F MFW-F e l l e i i
' i FW-C-02A i l...'....................'gg
' a .
FW-C-01A 'A' 2-; % >
2Cv.1023 2 2Cv10y1 : ****
v
- 2FvA.r*
~ * * *
. . f ^. MFW-L,T,F From EFW -
- l 2EFW 9A FW-C-028 l*****....
. FW-C 010 *B' 2CV 1073 2 2CV 10T41 *
. 2F#$B V FW-C-03B l * . , . l
, , MFW-L,T,F From E . . . . . d. i i 2kFW 90 l e ,
, MFW-F ' M FW-T,F '
c.........s....................a Figure 2 Main Feedwater (MFW) Flow Paths O
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'. Page 19 of 39
- 6. 0 DEGRADATION MECHANISMS EVALUA TION The purt v s of this section is to identify the degradation mechanisms that can be present in the piping within the selected system boundaries for the ANO 2 A1RV system, as described in Section 3.2 of this report. The conditions considered in this evaluation are: design characteristics, fabrication practices, operating conditions, and service experience. The degradation mechanisms to be identified (Reference 9.1) are:
- Thermal fatigue (TF)
- Thermal Stratification, Cycling, snd Striping (TASCS)
Thermal Transients (TT)
-
- Intergranular Stress Corrosion Cracking (IGSCC)
- Transgranular Stress Corrosion Cracking ((GSCC)
~ External Chloride Stress Corrosion Cracking (ECSCC)
- Primary Water Stress Corrosion Cracking (PWSCC)
- Lccalized Corrosion (LC)
- Microbiologically influenced Corrosion (AllC)
- Pitting (PIT)
Crevice Corrosion (CC) e flow Sensitive (FS)
Erosion Cavitation (E C)
Flow Accelerated Corrosion (FAC)
{}
V in performing this evaluation, some basic inputs were used. These inputs are discussed in Section 5.3. 1he criteria andjustifications are provided in Section 5.2. In accordance with Reference 9.1, degradation mechanisms are organized into three categories: 'Large leak',
"Small Leak", and *None'.
The results indicate that three degradation mechanisms are potentially present: thermal fatigue, flow sensitive attack, and localized corrosion. The damage groups (DA1 groups) were identified as A1RV T,F, AinV-F, and A1RV L,T,F and are defined in Table 4 below.
These DM groups result in one failure potential category: "Large Leak'.
Table 4 Osmage Groups Demare Demere Mechaniems Feoure Group Thermal fetique Stress Correelon Crocking Locekred Corrosion Row SenaltNe Potentiet to TAsCS TT lasCC TGsCC ECSCC PwsCC MIC PIT CC EC FAC Ceterary MF W T.F Yes Yes No No No No No No No No Yes large Leek MRV-F No No No No No No No No No No Yes largeleak MnVL. T,F Yes Yes No No No No No No Yes No Yes Lerge Leak n
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ABB Calculation No. A PENG CALC 018, llev. 00 Page 20 of 39
, 6.1 DAMAGE GROUPS 5.1.1 DM GROUPS: MRV T,F and MAV L, T,F The MAV T,F and MAVL,T,F damage groups are considered sub}ect to thermal fatigue as indicatedin the table below. The MnV T,F DM proup extends from check valves 2RV 5A and 2RV 58 to the steam generator Inlet norries and the MnV-l,T,F DM group consists of the norrie connections to the steam generators.
Extent of Affected Section Thermalfatigue Cause Plant Mode Basis horizontalsection from check volve 2EnV 5A Upon initiation of AnV, cold water is introduced into downstream to verticalsection this horizontalsection via the EnV flow path. TASCS is possible and TT wiH occur both downstream and TASCS and TT . Cooldown upstream (due to backful) of the EnVconnection.
verticalsection between horizontalsection Upon initiation of ARV, this vertical section, which is described above and horizontal section downstreans of the EnV connection. willbe subject to connected to Mnv nogale TT.
TT Cooldown horizontal section connected to MFW nottle Upon initiation of AnV, at low flow rates, this at steam generator 2E24A horizontal section win not be completely pu.ged. This section willstay warm due to convection heating from TASCS and TT Cooldown / Hestup the hot SG creating the poten.lat for TASCS. This section is also sub}ect to TT.
This section will continue to be subject to TT during hootup after the swap over from ARV to Mnv untH such time that the flow rate is sufficiently high (piping) or the water is preheated (nottle area).
horizontalsection florn check volve 2ERV 58 Upon initiation of AnV, cold water is introduced into downstream to verticalsection the vertical section downstreem of this horizontal section via the EnV flow path. TASCS is possible TASCS and TT Cooldown and TT will occur upstream Idve to backfin) of the EnV connection, vertics'section between horizontalsection Upon initiation of AnV, this vertical section, both described above and horizontalsection downstream and upstream (due to backfil!) of the connected tc Mnvnozzle Env connection, win be sub}ect to TT.
TT Cooldown horizontal section connected to Mnv nozzle Upon initiation of AnV, at low flow rates, this at steam generator 2E248 horizontal section win not be completely purged. This TASCS and TT Cooldown /Hestup
- " ###Y # # "#
the hot SG creating the potential for TASCS. This section is also subject to TT.
This section wiH continue to be subject to TT during hattup after the swap over from ARY to Mnv until such time that the flow rate is sufficiently high
. (piping) or the water is preheated (nortle area).
ABB Combustion Engineering Nuclear Operations
ABB (D Calculation No. A PENG CALC 018, Rev. 00 Page 21 of 39 Both of these DM groups are also considered susceptible to FAC since these piping sections are included in the ANO 2 FAC program (Reference 9.12) and the MAV L,T,F DM group is additionally considered susceptible to localized attack by crevice corrosion. Crevice regions are formed between the transition piece and the feedwater norrie, and the thermal sleeve on each steam generator.
5.1.2 DM GROUP: MAVF The MFw.' DM group extends from isolation valves 2CV 1024-1 and 2CV 10741 to check valves 2RV 5A and 2RV-58, respectively. This DM group is considered susceptible to FAC since these piping sections are included in the ANO 2 FAC program (Reference 9.12).
These piping sections are not considered subject to thermal fatigue. Per plant cooldown procedure 2101.010, a verification is performed to ensure the closure of isolation valves 2CV 10241 and 2CV 10741. Consequently, a leakage path should not exist to allow for backleakage by the check valves and the potential for therms.1 stratification or thermal transients is considered unlikely.
5,2 DEGRADA TION MECHANISM CRITERIA AND IDENTIFICA TION The degredation mechanisms and criteria assessed are presentedin Table 5.
C f
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ABB Calculation No. A PENG cal.C-018, Rev, 00 Page 22 of 39 Table 6 Degradation Mechanism Criteria and Susceptible Regions
- da on (,g,,,;, 3 ,,,,p,ggg,y ,gg ,,,
y ,,
TF TASCS -nps > I inch, and nostles, branch pipe
-pipe segment has a slope < 43*from horizontal (includes elbow or connections, safe ends, ter into a verticalpipe), and welds, heat glTected
-potential existsfor lowpow in a pipe section connected to a tones (llAZ), base component allowing muring ofhot and coldpuids, or metal, and regions of potential existsfor leakageyou past a valve (L e., in-leakage, out. stress concentration Irakage, cross-leakage) allowing mixing ofhot and coldpuids, or potential existsfor convection heating in dead sndedpipe sections connected to a source ofhotpuid, or potential existsfor two phase (steam / water) pow, or potential existsfor turbulentpenetration in branch pipe cor.nected to headerpiping containing hotpuid with high turbulentpow, and
-calculated or measvred AT > $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 coldpuid injection into hot pipe segment, or hotpuid smection into coldpope segment, and
-\ AT > 200*Ffor stainless steel, or AT > 150*Ffor carbon steel, or AT > ATallowable (applicable to both stainless andcarboni SCC IGSCC -evaluated ir accordance with existing plant IGSCCprogram per austentric stainless steel (lHill) NRC Generic Letter 88-01 welds andflAZ IGSCC -operatmg temperature > 200*F, and (Pil'R) -susceptible maternal (carbon content 2 0 033%), and
-tensile stress (includmg residual stress) is present, and
-omgen or oxidating species are present OR
-operatmg temperature < 200*F, the attributes above apply, and
-initiating contaminants (e.g., thiosulfate,fuoride, chloride) are also required to be present TGSCC -operatmg temperature > 150*F and austenitic stainless steel
-tensile stress (includmg residual stress) is present, and base metal, welds, and
-halides (e.g., fuoride, chloride) are present, or flAZ cas,stic (NaOll)is present, and
-omgen or oxidstmg species are present (only required to be present in conjunction w halides, not required weaustic)
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Calculation No A PENG cal.C 018, Rev, 00
- Page 23 of 39 TaNe 5 (cont'd)
Degradation Mechanism Criterlo and Susceptilde Regions I,E'",'"'f*,"
- g cg Criteria Susceptible Regions SCC ECSCC -operatmg temperature > 130*F, and austentric stainless steel
-tensile stress is present, and base metal, utids, and
-an outside piping surface is withinfive diameters ofa probable llAZ leak path (e.g., valve stems) and is covered with non metallic insulation that is not in comphance with Reg. Guide 1,36, or an outside piping surface is exposed to wettongfrom chloride bearing environments (e.g., seawater, brackish water, brane)
IM5CC -piping materialis inconel (Alloy 600), and noules, uelds, andflAZ
-exposed to primary water at T > 620*F, and without stress relief
-the materialis mill-annealed and cold worked, or cold worked and welded without stress relief 1.C blIC -operating temperature < 130*F, and fittings, welds, llAZ,
-low or intermittentflow, and base metal, dissimilar
-pil < 10 and metaljoints (e.g., welds, ;
-presence intnaston oforganic material (e.g., raw water system), or flanges), and regions water source is not treated w blocides (e g., refuehng water tank) containing crevices G PIT -potenttal existsfor lowflow, and
-ox1 gen or oxidmng spectes are present, and
-inttiating contaminants (e.g., fluoride, chloride) are present CC -crevice condation exists (e.g., thermal sleeves), and
-operatmg temperature > 130*F, and
-oxsgen or oxidating species are present FS E-C -operating temperature < 250*F, and fittings, utids, ilAZ, and
-flow present > 100 hrs yr, and base metal
-velocity > 30fl's, and
-(Pc PJ / AP < $
FAC -evaluatedin accordance with extstang plant FACprogram perplant FACprogram i \
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ABB Calculation No. A.PENG CALC 018, Rev. 00 Page 24 of 39 5.2.1 Thermal fatigue (TF)
Thermal fatigue is a mechanism caused tv attemating stresses due to thermal cycling of a component which results in accumulated fatigue usage and can lead to crack initiation and growth, 5.2.1.1 %ermal Stratification, Cycling, and Stn,ung (TASCS)
During plant bestup and normal cooldown conditions, the auxiliary feedwater system is normally used to provide flow to the steam generators at low reactor power operation ($ 2%), generally via the ERV flow path. As such the horizontal piping section connected to the feedwater nortles, which remains hot due to convection heating from the steam generator, is potentially subject to thermal stratification during low flow conditions. Thermocouples were instaHed on the horizontal piping section connected to the 'A' steam generator to monitor for top to bottom ATs experienced during ARV operation as the plant was heating up from refueling outage 2R10. Top to-bottom ATs in excess of 400*F were recorded at low ARV flow rates. The thermal stratification transients were cyclicalin nature and varied with ARY flow rates. A potential also exists for thermal stratification in the horizontal sections downstream of check valves 2RV5A and 2RV.SB. During a normalplant cooldown, cold ARY flow (100*.140*F) is injected into, or wHI backfiH, piping with an initial temperature of approximately 455*F.
5.2.1.2 Thermal Transients (TT)
During plant hestup and normal cooldown conditions, the auxiliary feedwater system is normally used to provide flow to the steam generators at low reactor power operation (S 2%), generally via the EFW flow path. As such the horizontal piping section connected to the feedwater nozzies, which remains hot (~ 500*F) due to convection heating from the steam generator, is subject to thermal transients upon injection of cold ARV flow (100*.140*F). AdditionaHy, during plant hestup, this section win continue to be suojected to TT siter the swap over from AFW to MAV linitiaHy 100*.140*F) until such time that the flow rate is sufficiently high (piping .
estimated to be over 300*F at time of swap over based on thermocouple detal or the water is preheated inorile area estimated a ~ 500*). During a normal cooldown, the other piping sections downstream and upstream (due to backfiH) of the EFW connection will be subject to thermal transients when cold ARV flow (100*.140*F) is introduced into piping with an initial temperature of approximately 455*F.
5.2.2 Stress Corrosion Cracking (SCC)
The electrochemical reaction caused by a corrosive or oxygenated media within a piping system can lead to cracking when combined with other factors such as a susceptible material, temperatura, and stress. This inechanism has several forms with varying attributes including intergranular stress corrosion cracking, transgranular stress corrosion cracking, external chloride stress corrosion cracking, and primary water stress corrosion cracking. .
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ABB Calculation No. A*PENG CALC 018, Rev. 00 (3
Page 25 of 39 5.2.2.1/ntergranular Stress Corrosion Cracking flGSCC)
The main feedwater system consists of carbon steelpiping. Carbon steelpiping is not susceptible to IGSCC.
5.2.2.2 Transgranular Stress Corrosion Cracking (TGSCC)
The AfRV system consists of carbon steel piping. Carbon steel piping is not susceptible to TGSCC.
5.2.2.3Extctnal Chloride Stress Corrosion Cracking (ECSCC)
The AfRVsystem consists of carbon steelpiping which is not susceptible to ECSSC.
Also, ANU-2 complies with the requirements of Regulatory Guide 1.36 for non.
metallic thermalinsulation and consequemly the potential for ECSCC to occur does not exist.
5.2.2.4 Primary Water Stress Corrosion Cracking (PWSCC)
PWSCC is not applicable as a potential damage mechanism for the A4RV system due to the fact that there is no inconel(Alloy 600) present in the system.
5.2.3 Localized Corrosion (LC) in addition to SCC, other phenomena can produce localized degradation in piping components. These phenomena typically require oxygen or oxidizing environments and are often associated with low flow or " hideout
- regions, such as exists beneath corrosion products or in crevices. This mechenism includes microbiologically influenced corrosion, pitting, and crevice corrosion.
5.2.3.1 Aficrobiologically influen:ed Corrosion (AflC)
All of the piping in this system operates at temperatures greater than the 150*F upper threshold for A4/C. Therefore these lines are considered non susceptible to AflC.
5.2.3.2 Pitting (PIT)
The A4RV system contains water that is continuously flowing at a high flow rate.
Therefore, the AfrN system is not considered susceptible to pitting.
5.2.3.3 Crevice Corrosion (CC)
There is a poten.'ist for crevice cc.rosion due to the fact that there are crevice regions in the attachment of the thermalsleeve to the feedwater nozzle downstream of the feedwater piping to nozzle transition piece, p .
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ABB Calculaflon No. A PENG CALC 018, Rev. 00 Page 26 of 39
- 5. 2.4 Flow Sensitive (FS)
When a high fluid velocity is combined with various other requisite factors it can result in the erosion and/or corrosion of a piping materialleading to a reduction in wall thickness. Mechanisms that are flow sensitive, and can create this form of degradation include erosion cavitation and flow accelerated corrosion.
5.2.4.1 Erosion Cavitation (E C)
All of the piping in this system operates above the E C upper temperature limit of 250'F. Consequently, this system is not considered susceptible to E-C.
5.2.4.2 Flow Accelerated Corrosion (FAC)
The MFN system is comprised of carbon steel piping. FAC is a phenomenon that i only affects carbon steelpiping. Per Reference 9.12, the MFW system is included in the ANO 2 FAC program. The MFW lir'es are categorized by Reference 9.12 as Susceptible Modeled (S M). These line segments are therefore considered susceptible to FAC d mage.
5.2.5 Vibration Fatigue Vibration fatigue is not specifically made part of the EPRI risk informed ISI process.
Most documented vibrational fatigue failures in power plants piping Indicate that they are restricted to socket welds in small bore piping. Most of the vibrational
[
fatigue damage occts's in the initiation phase and crack propagation proceeds at a 1 rapid rate once a crack forms. As such, this mechanism does not lend itself to typicalperiodic inservice examinations (i.e., volumetric, surface, etc.) as a means of managing this degradation mechanism.
Management of vibrational fatigue should be performed under on entirely separate l program taking guidance from ths EPRI Fatigue Management Handbook (Reference l 9.10). If a vibration problem is discovered then corrective actions must be taken to l either remove the vibration source or reduce the vibration levels to ensure future component operability. Frequent system walkdowns, leakage monitoring systems, l and current ASME Section XI system leak test requirements are some of the practical measures to address this issue. Because these measures are employed either singly or in combination for most plant systems it is not necessary to use a l risk informed inspection selection process for vibration fatigue.
t l
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ABB O. Calculation No. A PENG CALC-01B, Rev. CO '
Page 27 of 39 5.3 BASIC DA TA
- 5. 3.1 Under normal plant operating conditions, the AfRV system, as defined by the boundaries in Section 3.2, functions as indicated in Table 6.
l 6.3. 2 Ove to the cyclic nature of thermal transients, only those transients which occur during the initiating events Categories I and ll as described in Reference 9,1, Table j 3.1 are considered in the evaluation of degradation mechanisms due to thermal i fatigue. Category I consists of those events which occur during routine operation, e.g., startup, shutdown, standby refueling. Category ll consists of those events which have anticipated operational occurrence, e.g., reactor trip, turbine trip, partial loss of feedwater. Therefore, the transients to be evalvsted are those transients which occur under normal operating and upset conditions.
l O
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NkS PkW W Calculation No. A-PENG-CALC-018. Rev. 00 Page 28 of 39 Table 6 Main Feedwater System lines and Operating Conditions Damage Cw.wsnt Comlonent Descnption Material Design Specirmation Group ID Line [9.16] (I) [9.5]
No.
Design Operating Press Temp Press Temp psig *F rdg 'F MFW-F 2DBB-1-24" From rahr 2CV-1024-1 to check tsht 2FW-5A CS 1360 460 985 455 N M V-T F 2DBB-1-24" From check vahr 2FW-5A to 24* x 18" 90' Reducing Elbow CS 1360 460 985 455 MFW-T,F 2DBB-1-18" From 24* x 18* 90' Reducins EIbow to 2E24 A Steam Generator CS 1360 460 985 455 MFW-F 2DBB-2-24" From vahr 2CV-1074-1 to check value 2FW-5B CS 1360 460 985 455 MFW-T,F . 2DBB-2-24 From check vahr 2FW-5B to 24* x 18* 90' Reducing Elbow CS 1360 460 985 455 MFW-T.F 2DBB-2-18* From 24* x 18* 90" Reducing Elbow to 2E24B Steam Generator CS 1360 460 985U 455 MFW- 2DBB-1-18* Connection to 2E24 A Steam Gmerator Inlet Nozzle CS 1360 460 985 455 L.T.F MFW- 2DBB-2-18* Connection to 2E24B Steam Generator Inlet Nozzle CS i360 460 985 455 L,T.F Notes:
- 2. During fullpower plant operation continuous water flow would be present from both pumps. (From Reference 9.7)
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l A.BB (p)
Calculation No. A.PENG CALC 01B, Rev. 00 Page 29 of 39 6.0 SERVICE HISTORY AND SUSCEPTIBillTY REVIEW An exhaustive review was conducted from mid '96 to Spring '97 of databases (plant and industry) and station documents to characterire ANO 2's operating experience with respect to piping pressure boundary deprodation. 1he results of this review are provided in a condensed form in Table 7 for the Afain feedwater System.
Although several pre-commercial references are included for completeness, the timeframe for identifying items applicable to this effort was focused on post commercial operation (Commercial Operation date of Afarch 26, 1980). This was done to avoidinclusion ofitems primarily associated with construction deficiencies as opposed to inservice degradation.
The following databases and other sources were queried to accomplish this review:
- Station Information Management System (SIA45)
The SIAfS database was queried for all ANO 2 lob orders on Code Class 1, 2, and 3 components which involved corrective maintenance (CAf) or modifications (A100).
Additionally, a separate query was performed in order to capture certain non Code, O component failures. Th.'s query was for non Code 0 and SR (safety related) components. This database contains information from approximately 1985 to the present, g) - Condition Report (CR) Database The CR database was queried for any pipe leak / rupture events or other conditions associated with identified damage mechanisms at ANO 2. The keywords searched under were; pipe, piping, line, water hammer, leak, leaking and leakage. CR's are written on 0, F or S equipment failures or other conditions potentially adverse to safety.
This database contains information from 1988 to the present.
- Uconsing Research System (LRS)
The LRS database was queried using a keyword search specific to ANO 2. The key.vords searched under were: therrnal cycling, thermal stratification, thermal fatigue, defect, flaw, indication, fatigue, cavitation and corrosion. This search captured all communication between ANO and the t'RC, both plant specific and generic inaustry, associated with these topics. However, for the purpose of this review, only communication from ANO to the NRC was reviewed. Additionally, this search system was used to cusry industry Events Analysis files (captures INPO documents) for ANO 2 events or conditions relevant to this review. The keywords searched under for this portion of the query were: pipe & stratification, thermal & fatigue, thermal & transient, pipe & leak, vib?ation & thtigue and pipe & rupture. ' Furry
- search logic was employed to reduce the possibility of failing to identify a pertinent document. This database coratains information from prior to commercial operation to the present for ANO-2.
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ABB Calculation No. A PENG CALC 018, Rev. 00 Page 30 of 39
- Nuclear Piont Rehability Database System (NhtDS)
NpRDS was queried for ANO 2 entries for pipe failures. The keywords searched under were: pipe. This database contains Information from 1991 to the present.
~
ANO 2 ISIhosrom Records The ISI program findings were compiled and reviewed for all outage and non-outage insemice inspections conducted at ANO 2 since commercial overstion.
-- ControlMoom Station log The station log was utilized as a source of Informstlon for recent operational events.
The log exists in electronic format from early 1994 to the present and has search capabilities which allowed a review for events of Interest. The keywords searched under were: water hammer, leak and leskoge.
- System Upper levelDocument (ULD)
The ULD was reviewed as a source for historical perpective of issues related to the system and identification of modifications made to the system or changes to operational procedures to address those issues (e.g., water hammer, corrosion or vibrational fatigue).
- Other Station Documents Thss source of information consists of such documents as the SAR. Technical Specifications, operationalprocedures and the damage mechanism analysis done as part of this effort.
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& G D 9%WW Calctdatiots No. A-PENG-CALC-C18. Rev. 00 Page 31 cf 39 7a&Af 7 Service History and Suscep6 batty Review - Mairt Feedwater Systerts Seeree DocterperBre / Metsbesse NeeBeWed for Oefr9spe MeChefogrete #_'__I'_ _
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&sdence of Mieteriest Muminy Meessee " - . '_ r
- Thermerfetiyae Streee Cerroriser CrocAny Lecesseef Corressee neue Serisser,, m ,eg ,,,,g nr ,,, h Depredeoen oecersw cee et ANo-2 TAscs 77 nscc TcScc tcscc rwscc anc rrr Cc l EC FAC VF Nanuwwr Rm6we 4
Steriorr krformetsor Menegemorrt Systerry None %ne %ne %ne %ce None %ne %ne None %ne Nene None %ne U
%re CaridFaion Report Derebese PEit) hew %ne %ne %ne None %ne None %ne %ne PD(2) None %ne Morse Lee nse.g Research Syrrem PE{1) %ne %=w %ne %ew None %ne %ne %ne %ne PD(2I Eno %ew Nam Nuc/cor Merrt Re&nbsWr;r Dete6sse System None %ne Mene None None %ne %ne %ce None %ne %ne None Nona N:ne ANot2 tST Progrern Records %ne %ne %ne %ne None Nnne %ne %ne None %ne %ne i %ne %ne i tti Correrot Roorpr Stenan top None None None None %ne None None None None %ne None Mene None %n, Systerre L@per Lees /Cm . ..s %ne %ne None None None None %ne %ne %ne %ne %ne %ne %ne %ne Other Ste son Documents ~ Pl4) - Pl4) None %ne %ne None None %ne P14$ - %ce Plel : %ne hne f*>ne tcoand:
P (Procuroor) - Tise category includes idenefication of postule*ed demoge a echenseme end loetnge tteeugh it ,..14r,s of operstmg peremeters, water cL- =,. etc. % phys.cel evidence of pecoeure boundary de,,. 4c,a. currentfy emets. Ttwo category includes postuleted mechermome idonefied se e result of ttue rewow.
PE (Ptent Event) - The category includes identifice*,on of postuieted demoge M .--. end loodmge es e result of en observed or potenner plant event te g., water hommert %
phyeecal evidence of preesure boundary deg-edetron currentfy emete.
PD (Phys cel Demogel - TPwe category includes identification cf eboerved pres
- are boundary degradetion se evidenced by croclung, pittmo, westege, t*=nreng, phvoscal defo*metson er other detenoretxm.
PSF '(Procoure Boundary Feevee) - Ttwo category meludes identifiestion of through-was fle m reeuttmg from ttw effects of en identrfied demoge .. J._.a.~
Notee:
- 1. Refwence NRC Bulletin 79-13, CR 2 94-0272 end NRC Informet on %tsee 93-20 e ~ a., thermal stratificotson end creciting of mem feedwater nonles
- 2. Refere/sco CR 2-91-0227 and 2CANO49105 h.~caev FAC demoge to the main feedwetar ponie tharmel elaevee
- 3. Reference ISI program recorde wtuch document the identificat6on of one subourface ird ..L. in the MFW System. TNe ;.J, n. wee evolueted and deteemmed to be Code ecceptatde.
TNe indication wee not ettnbuted to en inservice demoge mecherwom and is balieved to beve been noe eerwece inducee 6.e., fabacetsers ongel.
- 4. Reference Section 5 of ttmo document wtuch identfees the potential for TASCS. TT, CC and FAC in speerfic portrone of the MFW System.
ABB Combustion Engineering Nuclear Operations
O JL lB Calculation No. A PENG CALC 018, Rev. 00 Y Page 32 of 39
- 7. 0 RISK EVALUATION The first step in the risk evaluation is the defining of the risk segments. Risk segments consist of continuous runs of piping that, if failed, have the same consequences II.e.,
consequence segments), and are exposed to the same degradation mechanisms (i.e.,
damage groups). The next step in the risk evaluation is the determination of the segment risk categories. This is accomplished by combining the consequence cnd damage mechanism categories to produce a risk category for escl .,egment. ApplicMion of the above criteria results in the formation of 8 risk segments of which all 8 are high risk (risk category 31. The risk segments are identified in Table 8 below.
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ABB Combustion Engineering Nuclear Operations
MWW l Calculathwr No. A-PENG-CALC 418. Rev. 00 Page 33 Of 39 Tabk 8 Risk Segmerrt Iderr66icadors msk Segment to conseneencelo Demoge Group D Msn Region Phung thw Mos. Msk Segment start Pbint Msk Segment EMP% int colegary resure Pt tenew ' ask colegery isometric Drewings (11 Steen C--- ; 2DU un 'RDWA-1 rW C 01A MfWTJ Mgh 2DBS 1-24* (11 Downsteam of 2FW 6A 2DBW118* MFW Nonfo Medium large les* 3 1112D86-1-2 StL 1 MFWR 01A-2 FWC 01A MFW L.TJ High 2CBS * .d* 111 Steem Genersect 2O48 (11 Steam Generetar 2O4A MFWNonio MFW Nante MecGum large Leek 3 ft: A BB 1-2 StL 1 M F W R D 18-1 FWCD ** MFWTJ High 2DB8224' ill Downsteem ot 2FW6:1 (11 Saeem Generetor 2O48 20B6 2-18* MFWNonie (11 2DB9 2-2 MF W R C18-2 FWC-018 MFWL.TJ High 20B6 2-18* (11 Steam Generator 2O48 fil Steem Generener 2E248 ;
Mediwn large Leek 3 ill 2D88 2-2 MFWR-02A FWC-O2A MFWF High 2DBW 1-24* 1117.. ;.6. 2F3 ill L&seeem ot 2FW SA Mediium large Leak fil 2DBB-1-2 StL 1 MF W R-O28 1W-CD2B MFW F Hmh 2DB8-2-24* (11 Penneetion 2P4 (19 Oneteemof 2FW$8 Medium large im 3 (1) 2DSB Z2 MFWRD3A FW C-03A MFWF High 2006-1-24* (11 Domensteem r4 2CV-1074- (1) Peneeeean 2P 3 1
Medium large Leek 3 fil 2DBS1-1 StL 1 MFWRD38 FWCD3B MFWF Wh 2DBd 2 24* (11 C . a -. of 2CV-1674- (117- ;.---. 2P4 1
Madiurn large Leek 3 fil 208921 S!L 1 y W
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p ABB Calculation No. A.PENG CALC 018 Rev. 00
() Page 34 of 39 To facilitate application of the sampling percentages to determine the inspection scope, ISIS combines like segments (i.e., same consequence category and damage group) into segment groups. A total of 3 segment groups have been identified and are summarized in Table 9 below.
Table 9 Risk inspection Scope Segment Consequence fouure Risk Risk Total Selections Selections Groups Category Potentiel Repton Category Welds Required Mode MnV-001 001 4 41 Medium large leak O High 3 17 6 (Note 1)
Ef MnV-003 totale 65 Note 1: All 65 of the total welds, which are distributed amongst the 3 segment groups as indicated above, are susceptible to FAC. Segment groups MAV 002 and mrv.
003 are also subject to thermal fatigua and MAV 003 is additionally susceptible to localized cerrosion. The 6 selections reflected above have been made in these two segment groups. All other required element selections will he as determined by the existing ANO 2 Flow Accelerated Corrosion program. The FAC program however, does not lend itself to the establishment of a long term plan with a predetermined inspection scupe. Generally, the scope of a planned FAC inspection is determined prior to each scheduled inspection, normally in ccnjunction with a refueling outage. As such, no additional element selections are included.
- 8. 0 ELEMENT SELECTION The number of elements to be examined as part of the risk-informed developed program depends upon the risk categories for the risk significant segment groups as indicated in Table 9 above. An element is defined as a portion of the segment where a potential degradation mechanism has been identified according to the criteria of Section 5.0. The selection of individualinspection locations within a risk category depends upon the relative severity of the degradation mechanism present, the physical access constraints, and radiation exposure. In the absence of any identified degradation mechanisms (i.e., risk Category Al, seleClions are focused on terminal ends and other locations (i.e., strvClural discontinuities) of high stress and/or high fatigue usage. An inspection for cause process shall be irnplemented utilizing examination methods and volumes defined specifically for the degradation mechanism postulated to be active at the inspection location.
Table 10 depicts the element selections and other pertinent information (e.g., examination methods and volumes, b3 sis for selectioni for risk-significant segment groups MFW-001, MAV-002 and MAV 003. As indicatedin the Risk Inspection Scope of Table 9, a total of 6 elements have been selected for examination from segment groups MAV002 and mrv.
003. The examination methods and volumes specified in Tob!e 10 (risk category 3) are defined in Reference 9.1 and are based upon the degradation mechanism (s) pcstulated to be active at each selected element.
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ERIl l B"K W W Calculation No. A-PENG-CALC-018. Rev. 00 l
Page 35 of 39 Table 10 E7ernent Selection - Risk Cetegory 3 S_erment croupe C- n _ e Feaure retender Rien Catesory Mek Rosien Tord e of elemente l 2 n ofetene te MFW 001.002.003 MeSum large Leek 3 Ingh 65 17 Bements Setected Lke No. Esem Method Kesk Segment D Desa:ripelen be Dwg No. Euem Vehene Consequence /DM Groep D's Reeeen for Seleceen 17 001 2DBB 18* Voksnetric MFW R 01A-2 TNs risk segment is sutpected to therme! stratincerion (TASC51 duning plant modes 2 and 3 when nNs sey nent, wNch remems wom, ekoe no Transition Mece to-SG 2DBB-1-2 Sh.1 Rgure No. 7.1-2 n V C O !A / M n V L.T J _ , y,,g,,,, y ,,,, 4,,4 y g, ggy y ,,,,
2E24A Feedwater Nozzie Rgure No. 7.2-2 TNs s is dse to (TT1 w Mie n W" of Anv now and the norrie aree wm concnue to be suetect to TT dway plant beengs efter the swap over from AFW to MFW ente the water is preheeted. AddFrioneEy, GNs segment is susceptrble to keeKred corrosion (thermet siereel and How ecceWwred corrosen.
TNs tenrenet end element bes been selected since it is suNected so the Nghest ben &ng moment in tNs risk segment (due to TASCSI.17-002 2DBB- 1-18' Voksnettne MnvRD1A-1 TNs rist segment is sutvected e thermet stretreeten (TASCSI durha plant modes 2 and 3 when tNs segment. wNch remems worm due to Pfpe-to- Transition Moce 2DBB-1-2 Sh.1 Rgure No. 7.1-2 DV CO!A / MAV- TJ _
- ;,,,,, 4 py y y, pyw gg, ,,,,,_
- 'd This segment is also saheet to therme! conssents (TTI avon hirierion of ARV now and the pking wM contmue to be sukoct e TT darkg plant how after the sway over from AFW so 19FW smtR the fnow rete is sufficsonsty twgh. Ad&6cneMr. this segment is suecoctGe so i tsow occelerered corroemer. TNs ernment hos been selected smco it is sutvected to the 2nd twghest stress htensfocation h e%is vist segment due to stre . w .%a arrip6ficeeion of the local TASCS stressee et the cc~. ; ~.+ region Ke sNckness 6scontmnstyi 17-003 2D88 1-18* Volumetric MFY'RDTA-1 TNs risk segmentis sutrected so therme!s;..;L.;~.. (TASCSI chrma plant modes 2 and 3 when sNs segment which remeens monum due no Reducing Dbow-twPipe 2DBB 1-2 Sh.1 Rgure No. 7.1-2 FW COTA / MnV-TJ . ;~.. heethg. is emot .n%; *w purged et kw AFW fkw retes.
Weld TNs segment is also sutvect to thermer tronomers (TTI soon krtisticn of AFW tkw and the pipkg wM contmue to be sutvect so TT dwing plant heetnes efter the swep coor from AnV to MnY arrtR the Row rets is sufficientty Ngh. AddFrionePy. tNs segment is susceptbe no trow eccelerered corrosen. This o!* ment hos been sofected smce it is sutyected to the twghest stroes brensFreciore h tNs risA segment dse to the ;w.L ; Wocedon of the local TASCS stressee et the ac~..; :~.e region Ke tNetness dscorrtinuityl.
L ABB Combustion En ring Nuclear Operations h
m , n aoo AY% W W UJculatiors No. A-PENG-CALC-018. Rev. 00 Page 36 Of 39 Table 10 Eternent S ' ,.GO,i - Risk Category 3 (Cottt'd)
Seement Greenpo C..--. . .a feaure Potencial Wret Ceterary Rid Recien Tetaf 9 ereiennente 25% ofademente MFW 001.002.003 Med%m large leet 3 PGgh 65 of Bements Selected Lir'e No. Esem Method Rieh Segment D C 4 -_ be Dws No. Esem Verume **- - - _ _ - /DM Groep D*s Reneen for Selecean 19D01 2DB8 2-18' Vokarwaric MFWR C18 2 T? mis risk segment as sehected to themmet cas sficeeon (TASCSI during plant modes 2 eruf 3 when this segment. which remeens worm doe to Transition Pfere-to-SG 2DBB 2-2 figure No. 7.12 FW C 018 / MFW L.TJ ;_; _
, y,,,,;,,,, g, ,,g 4,4 pg7 ,, g,,, y py gy,, ,,,,,_
2E248 feedwerer Non,, tipure No. 7.2-2 7fg, ,,gm,,,, g, og,, ,,4,eg ,, sp,,m,,4,,,isents (TT7 en on iretnecon Wold of AnVIFow ered tfse esonde aree awuT m to be sakeet to TT darirog plant heetups efter the swap over f orn AFW to MfW sets the water is preheered. AMtiornery, this segrnent is sansceptbe to benEred corresson (thern er sfeewel and fbw secederered corrosmrt Timit terremnet end element hos been sedected since it is se4ected e the tughest bendvrog enornent he this risk segment idoe to TASCSI.
19 D02 2DB8 2- 18* Voksnetric MnVRD18-1 This vist segment is sakected to thermalswerrficenen (TASCsl doing pdent modes 2 erid 3 where this segrrier-t. which romans warm due so Pfpe-to-Transrten Mece 2DB8 2-2 Rgure No. 7.1-2 nY CD1B/MFW TJ ;,,_ _ _ A , g,,,,,, y ,,,, ,,,,g,,,47 pu,y,4 ,, y, p py gy, ,,,,,_
Wald This segmorrt is also sa4*ct to thermal eensenes ITTI upon iraderers of AFW tkw and the v4*eg erai evnenue no be sakeet to TT daring plant beentes efter the swep ower from AFW to MnV emtR the ficw rete is sufficeerrtfy high. AMaonedy. tfmis segment is sansceptible :la thw accelerered cor osson. Ttis ehement. estich hos nho been repeived once inV12R1} and rowe& fed (creetmg ligher ressdkser stressest, hos beerr selected shsce it is sa4*cted to the 2nd tig*.est stress intensificatiorr is this risk segment die to the u.&.",. ;
enp&fication of the keel TASCS stresses et the courrterbore regiorr K.e,tricAness h .;_ t,1.
19D03 2DB8 2-18* Vokmeetric MFWR C181 This risk segmentis se4ected % thermalsawtificatierr (TASCSI dairmy pteret modes 2 and 3 when this egment, which ver neens mserm doe to Redkocmg EJbow to-Pfpe 2DBB 2-2 figure No. 7.1-2 FWC418/ MFW-TJ ; _
, y,,,,,,,, ;, ,,g 4,4 4 ,, g,, p yy fy, ,,,,,_
- This segment is also saheet to thermed trans*ents ITTI apon iritiation s of AFWIEow and the p4*sg ewiW conte've to be naheet to TT during pierrt heetups efter the sweg over from AFW to MFW urrto the thw rete is sufficierrtfy tigh. AdSteneWy stis segment is susceptMe to ftow secederered corrossort This elemer t fwe been selected sirsce it is sahected to the toghest stress interesification hv tres risk segment doe to the y..w,; erv t?cetiour s of the keer TAsCS stresses et etw e,- .; b- region Ke., aficArsess dscorrtinurtyf ABB Combustion Engineering duclear Operations '
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i ARn MIFW Calculation NO. A-PENG-CALC-018, Rev. 00 Page 37 Of 39 Table 10 Elemettt Selectiots - Risk Ca*egory 3 (Cortt'd!
Seement Greenpo C--- FeMure Potentie! Mek Catoneer Riot Recien Totaf I of enanoonte 26% et elemerate MFW001.002.003 Modum terge teet 3 Ingh 65 17 Bemeents Selected Lir.a No. Enem Method Risk Segment D C- ; L ane Dwg Me. Esam Volume C , _ -. /DM Greage D's Reneen for Seleceen AM ether reordred element 2DBB- 1-24' Vokanetric MMR OTA-1 A pnmery obt ecove of FAC ene& sus k to identsfy -.,~. :. thet are selectiorw walt be es most susceptble to FAC demoge in carbon-steelofpkg. kckadng both detemuned by the existing 2DBB t-t8".
208B 2-24 Rgwe %. 7.7-1 M CDIA /MWJ single and two phase I+ energy systems. The chonce of truvectierr ANO-2 Fkw Accelerered 2D88 2-18 Me h. 7.7 2 ye,,g,,, go, ,,,e;q;, ,,g,y,, ,,g ,, ;,, ,,,,,;g,,,,;,, ,,,,,,,
Corrosion progrem. The W* W I I'# yyy g aga.g Weion resents, kdusey esponence. EPRI 04ECWORKS prodceee FAC program however, 2DB8t-1 % 1 Rgaae M. 7.74 connparter model renkkg and enannenng judgmerrt, each of wNeh is 6:ves not lend itself to the 2DBB t-2 Sh.1 We M. 7.75 FWC-01A /MAYI 'l descrbed k greeter dete@ bekw.
,,,,sg,w g , .
2DBB 2- 1 Sh.1 Rgure No. 7.7 6 term pim evith a 2DBB 2-2 Rgure No. 7.7-7 .
ym . g g ,y m predetemened kspection #'## O'# ' premous kspecnons are remewedprior to each scheduled mspecoort scope. Genereny, th* FW C-019/ MFW-TJ The predcted e u, Efe, co n cedered by weer retes of the scope of a planned FAC premoansty inspected ~.,~.-.:., k artmzed to identWy futwo mspection is deterrraned & bceaans.
pric e to each schedded M-W-R 018 2 lG ueM normeMy in ~ #
corf.cwtion with a e ennueble semplement to plant anetyses and essocnered *.na:W .
refueGng oestege. As such, Some sJurces of informenon kchade the 04UG het Ene. Entergy FAC no element selections are MFWR C2A p**' 9'o'v, OfUG-**eported Prent Eeents Detebase and t%e ANO kcheded. kdustry EenMs Ane& sis croav. AI 4- L. ; industry exponences FWC-02A /MFW-F refering to FAC. are fecered into the progrerrt AE - ,- a trwt vedere so en mdustry eeeM wM generacy be kspected at the nest MFW-R 078 #***
FW-C-02B / MFWF CNECWORKS Renkky - The EPRI OfECMMKS pro 6cuer connparter model is amed for geodence k detemerang specific beetw to be enemmed. Wthh the OfD%ORKS corrparter model e weer rete MMRM calcadorion for sech modeleer sne is performed. A senple of the FW-CD3A /MFWF "9^**' '*""*# ** *" *"*'***"*
renking-MF W R-038 Espneerirs ~ :-: - AppEceoon of good engmesmng judgenent le en k,portant - J.. .ea , in the selecean of -.,~.-.:. Ist FW-C-03Bf MFWF w p.,,,,,,ess d .= %.; system operation es weg es iredusey exponence are factored hoe the coeponent selecean. Also, foo&ect from system engmeermg. plant operenons, memtenance and the secondary chemese, deparements are feden irree corssedecesort h ABB Combustion Engh. ring Nuclear Operations h
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Calculation No. A PENG CALC 018, Rev. 00 G Page 38 of 39
- 9. 0 REFERENCES l 9.1
- Risk Informed Inservice inspection Evaluation Procedure,' EPRI Report No. TR-106706, Interim Report, June 1996.
9.2 EPRIInservice Inspection Software (ISIS'),1996.
9.3 Arkansas Nuclear One Unit 2, ' Safety Analysis Report,' Amendment No.13.
9.4 ' Design Specification for ASAfE Section til Nuclear Piping for Arkansas Nuclear One Unit 2, Arkansas Power and Ught Company,' Specification No. 6600-Af.2200, Revision 9.
9.5 'ANO 2 SIA4S Components Databas6,* IPlant Piping Une Ust (A4 2083), dated 3 31 96).
9.6 'ANO 2 ISI Plant Piping Une Ust,' from Revision 4 of ANO 2 Inservice Inspection Plan.
- 9. 7 " Design Configuration Documentation Project, ANO 2 feedwater and Steam Generator Blowdown Systems", ULL 2 54513, Rev. dated 7/93.
9.8
- Technical Specification for Insulaticn for Arkansas Nuclear Orte Unit 2 of the (GD. Arkansas Power and Light Company, ' Specification No. 6600-Af 2136. Revision 9.
9.9 ' Primary Chemistry Afonitoring Program,' Procedure No. 1000.106, Revision 4.
9.10 *EPRI fatigue Afanagement Handbook,' Report No. TR 104534 VI, V2. V3, V4, Project 332101, Final Report, December 1994.
9.12 *ANO 2 Flow Accelerated Corrosion System Susceptibility Report *, ANO Report No.
95-R 2004 01, Rev. O, dated 8/18/95.
9.13 Arkansas Nuclear One Unit 2 " Technical Specifications, Appendix A to Ucense No.
NPF.6, Amendments Nos.173 and 174.*
9.14 Gaertner, J. P., et. al. ' Arkansas Nuclear One Unit 2 Internal Flood Screening Study," prepared for Entergy Operations, Inc. Calculation No. 89 E 0048 35, Rev. O, Afay 1992, 9.15 ' Arkansas Nuclear One Unit 2 Probabilistis Risk Assessment, Individual Plant Examination Submittal,' 94 R 2005-01, Rev. O, August 1992.
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A BB Calculation No. A PENG. CALC 018, Rev. 00 g Page 39 of 39 W 9.16 Entergy, Arkansas Nuclear One Unit 2, Isometric Drawings:
l 1. 0 Drawing No. A4 2206, Sheet t. Rev.122; ' Piping & Instrument Diagram Steam Generator Secondary System.'
2.0 Drawing No. 208811, Sheet 1, Rev.10; 'Large Pipe isometric Alain Feedwater Supply to Steam Generator 2E 24A.'
3.0 Drawing No. 208812 Sheet la Rev.19, Sheet 2, Rev. 2; *Large Pipe Isometric A4ain feedwater Supply to Steam Generator 2E 24A.'
- 4. 0 Drawing No. 2088 2 1, Sheet to Rev.13; *Large Pipe Isometric A4ain Feedwater Supply to Steam Generator 2E 248 from 2CV 1074-1. '
5.0 Drawing No. 2DBB 2 2, Sheet 1, Rev. 18; 'Large Pipe isometric A4ain Feedwater Supply to Steem Generator 2E 248 from 2CV 1074-1."
9.17
- Consequence Evaluation oV ANO 2 EFW, Containment Spray, and Alain Steam and Feedwater System Piping,* Arkansas Nuclear 1 Unit 2, Yankee Nuclear Services Olvision Calculation Na. NSD-018 Rev. O, August 1997, 9.18 Swain, A. D. and Guttmann, H. E.; " Handbook of Humats Reliability Analysis with Emphasis on Nuclear Power Plant Operations', NUREG CR 1278, August 1993.
9.19 Interoffice Correspondence from A. V. Bauer to Quality Records, letter No. PENG-l 97140, ' Submittal of SIA Calculations,' dated July 21,1997.
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Calculation No, A.MNG CALC 018, Rev. 09 Page Al of A7 0
APPENDIXA
'FMECA CONSEOUENCE INf'RMA TION REPORT *
(Attachment Pages Al . A7) h O
ABB Combustion Engineering Nuclear Operations
l FMECA - Consequence Information Report Calculation No A PENG-CALC 018. Rev. 00 is-sep-91 Page A2 of A7 Consequence ID: FW-C Ol A Consequence
Description:
Degradation of main feedwater flow to steam generator 2E 24 A inside containment during v: mal operation (line 2DBB 1 between 2FW 5A and steam generator ).
Ercak Stre: Large Isolability el Break: No ISO Comments: Feedwater isolation and feedwater pump trip will occur on low steam generator pressure. Also.
EFW will remain isolated to the faulted steam generator via a differential pressure between the faulted and good steam generators. However, blowdown of the faulted steam generator can not be isolated. g Spatial Effects: Containment Effected Instion: Containment Building Spatial Effects Comments: Feedwater line breaks are within the design basis and the necessary safety components located inside the containment building are qualified for such events.
Initiating Event: I Initiating Event ID: T5 initiating Event Recovery: No recovery from an unisolable feedwater line break. This results in an immediate plant trip due to low steam generator level.
Imss of System: SDM 2 System IPE ID: PCS, EFW System Recovery: MSIV isolation, feedwater isolation and pump taips occur on low steam generator pressure, it is possible to recover a condensate pump and makeup to the unfaulted steam generator. EFW discharge to the faulted steam generator is isolated and unavailable. Howwer, there is a discharge path frr :r each EFW pump to the unfaulted steam generator.
Loss of Train: N Train ID: N/A Train Recovery: N/A Cunuquence Comment: Cort'aur.- is ' Medium" based on Table 2 1 of Ref. 9,17. Containment isolation is ur:3,a@
Conscauence Category: MEDIUr 1 O Consequence Rank O O
. - . - .- _ . - . .-.-.- - . __ . _ - ~ _ . - . . - - . -
FMECA - Consequence Information Report Calca' oa xo A erNo.cate ola.Ru oo
] i4-ser 91 Page Al of A7 Consequence ID: FW C OlB f
Consequence
Description:
Degradation of main feedwater flow to steam geners:or 2E-24B inside containment during normal operation (line 2DBB-2 between 2FW 5B and steam generator).
Break Size: Large Isolability of Break: No ISO Comments: Feedwater isolation and feedwater pump trip will eccur on low steam generator pressure. Also, l EFW will remain isolated to the faulted steam generator via a differential pressure between the faulted and good steam generators. However, blowdown of the faulted steam generator can not
, be isolated.
Spatial Effects: Containment Effected Location: Containment Building i
Spatial Effects Comments: Feeduter line breaks are within the design basis and the necessary sdety components located inside the containment building are qualified for such estats.
Initiating Event: I Initiating Event ID: T5 Initiating Event Recoscry: No recovery from an unisolable feedwater line break. This results in an immediate plant trip due to low steam generator lestl.
.l Loss of System: SDM 2 System IPE ID: PCS, EFW
. System Recovery: MSIV isolation, feedwater isolation and pump tnps occur on low steam generator pressure. It
- is possible to recover a condensate pump and makeup to the unfaulted steam generator. EFW j discharge to the faulted steam generator is isolated and unavailable. Howestr, there is a 4
, discharge path from each EFW pump to the unfaulted steam generator.
1a.4 MTrain: N Train ID: N/A Train Recovery: N/A Consequence Comment: Conseqw,cc is " Medium" based on Table 2 1 of Ref 9.17. Containment isolation is unaffected.
l Consequence Category: MEDIUM O Consequence nank O O
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Calculation No A PENG-CA(,C-018 Rev. 00 i%fECA - Consequence Information Report 14-sep 97 Page A4 of A7 Consequence ID: FW-C 02A Consequence
Description:
Degradation of snain feedwater flow to steam generator 2E-24 A inside containment
& upstream of 2FW 5A (line 2DBB 1 between containment penetration and 2FW.
5A).
Break Size: Large holability of Break: Yes ISO Comments: Feedwater isolation and pump trip will occur on low steam generator pressure, but check valve 2FW 5 A isolates faulted steam generator and prevents immediate depressurization. Operater action or emptying of the condenser hot well will likely preside feedwater pump trip.
Spatial Effects: Containment Effected Imcation: Contaimnent Building Spatial Effects Comments: Feedwater line breaks are within the design basis and the necessary safety components located inside the containment building are qualified for such events. ,
Initiating Event: I laitiating Event ID: T5 laitiating Event Recovery: No recovery from a feedwater line break. Reactor trip (T6) will occur immediately on low steam generator level and 2FW 5A prevents steam generator blowdown and low pressure isolation in the short term.
Loss of System: SDM 2 System IPE ID: PCS, EFW System Recovery: Loss of PCS is assumed due to fdwater pump trip (operator or loss of suction from condenser hot well) or eventual MSIV isolation on low steam generater pressure or operator action. It is possible to recover a condensate pump and provide makeup to the unfaulted steam generator. Also, EFW could be isolated from the faulted steam generator before FW isolation occurs. However, this is recoverable and there is a discharge path from each EFW pump to the unfaulted steam generator.
Loss of Train: N Train ID: N/A Train Recovery: N/A Consequence Comment: Consequence is " Medium" based on Table 2 1 of Ref. 9.17. 2CV-1024-1 prosides containment isolation.
Consequence Category: MEDIUM O Consequence aank O O
FMECA - Consequence Information Report Calculatum No A PENG-CALC 018, Rev. 00 14 ser 97 Page A3 of A7 1
1 Consequence ID: FW-C 02B 4
Consequence
Description:
Degradation of main feedwater flow to steam generator 2E 24B inside containment
& upstream of 2FW 5B (line 2DBB-2 between containment penetration and 2FW-
- 5B).
Break Size: Large Isolability of Break: Yes
, ISO Comments: Feedwater isolation and pump trip will occur on low steam generator pressure, but check valve 2FW 5B isolates faulted steam generator and prevents immediate depressurization. Operator action os emptying of the condenser hot well willlikely provide feedwater pump trip.
I Spatial Effects: Containment Effected Location: Containment Building
, Spatial Effects Comments: Feedwater line breaks are within the design basis and the ne:essary safety components located inside the containment building are qualified for such events.
Initiating Event: I InitiatVg Event ID: T5 Initiating Event Recovery: No recovery from a feedwater line break. Reactor trip (T6) will occur immediately on low steam generator level and 2FW.5B prevents steam generaer blowdon and low pressure isolation in the short term.
Loss of System: SDM-2 System IPE ID: PCS, EFW System Recovery: Loss of PCS is assumed due to feedw1 ster pump trip (operator or loss of suction from condenser hot well) or eventual MSIV isolation on low steam generator pressure or operator
,T action. It is possible to recover P. condensate pump and provide makeup to the unfaulted steam generator, Also, EFW could be isolated from the faulted steam generator before FW isolation 1 occurs. However, this is recoverable and there is a discharge path from each EFW pump to the unfaulted steam generator, Loss of Train: N Train ID: N/A
, Train Recovery: N/A Consequence Comment: Consequence is " Medium" based on Table 2-1 of Ref. 9.17. 2CV 1074-1 provides containment isolation.
Consequence Category: MEDIUM O- Consequence Rank O 4
1
't 1
- O i
I l
FMECA - Consequence Information Report Cablanon No. A-PENG CAM-018. Rev. 00 W" Page A6 of A7 Consequence ID: FW-C-03A Consequence
Description:
Degradattor. of main feedwater flow to steam generator 2E-24 A outside containment dunng normal operation (line 2DBB 1 between 2CV 1024 1 and containment penettstion).
Break Size: Large Isolability of Break: Yes ISO Comments: Feedwater isolation and pump trip will occur on low steam generator pressure, but check vaht 2FW-5A isolates faulted steam generator and prevents immediate depressurization. Operator action or emptying of the condenser hot well will likely provide feedwater ptunp trip.
Spatial Effects: Propagation Effected location: Room 2081 Spatial Effects Comments: Propagation is into Room 2040, then into the east stairwell and down to El 317 (Rooms 2006 and 2011) into the auxiliary building sump (high level alarm in control room). EFW at El 335 (Rooms 2024 and 2025) and ECCS at El 317 (Rooms 2007, 2010, and 2014) are protected by watertight doors. It is assumed that EFW discharge tahts in Room 2081 and MCC 2B52 in Room 2040 are affected by this high energy pipe break. Even if FW continued unisolated and emptied a CST into the building, ECCS is unaffected at El 317.
Initiating Event: I laitiating Event ID: T5 Initiating Event Recos cry: No recovery from a feedwater line break. Reactor trip (T6) will occur immediately on low steam generator lesti and 2FW 5A prevents steam generator blowdown and low pressure isolation in the short term.
Loss of System: SDM 5 System IPE ID: PCS, EFW, HPSI, CSS, LPSI System Recovery: Loss of PCS is assumed due to feedwater pump trip (operator or loss of suction from condenser hot well) or esentual MSIV isolation on low steam generator pressure or operator action. It is possible to recover a condensate pump and makeup to the unfaulted steam generator. Also, EFW could be isolated from the faulted steam generator or the supply valves in Room 2081 could fail from the harsh environment. However, there is a discharge path from each EFW pump to the unfaulted steam generator. Failure of MCC 2B52 in Room 2040 is also assumed to affect train A supply valves in the containment spray, HPSI, and LPSI systems.
14ss of Train: N Train ID: N/A Train Recovery: N/A Consequence Comment: Consequence is " Medium" based on Tables 2 1 and 2 3 of Ref. 9.17 (2 backup trains, including EFW and AFW to SG A, FW recovery, and train B of once through cooling supply vehrs). 2FW 5A provides containment isolati on.
Consequence Category: MEDIUM O Consequence Rank O O
_ _ _. - _ ~ . _ _ _ _ .
1 FMECA - Consequence I.1 formation Report Cahtauem No. A PEAGCALC-018, Rev. 00 14-sep-91 3
Page A7 of A7 Consequence ID: FW C-03B Consequence
Description:
Degradation of n.ain feedwater flow to steam generator 2E 24B outside i
containment during normal operation (line 2DBB 2 between 2CV.1074 1 and containment penetration).
Break Size: Large Isolability of Break: Yes ISO Comments: Feedwater isolation and pump trip will occur on low steam generator pressure, but check vahr 2FW 5B isolates faulted steam generator and prevents immediate depressurization. Operator action or emptying of the condenser hot well will likely provide feedwater pump trip.
Spatial Effects: Propagation Effected Location: Room 2081 4
Spatial Effects Comments: Propagation is into Room 2040, then into the east stairwell and down to El 317 (Rooms 2006 and 2011) into the auxiliary building sump (high level alarm in control room). EFW at El 335 (Rooms 2024 and 2025) and ECCS at El 317 (Rooms 2007, 2010, and 2014) are protected by watertight doors. It is assumed that EFW
' discharge valves in Room 2081 and MCC 2B52 in Room 2040 are affected by this high energy pipe break. Even if FW continued unisolated and emptied a CST into the building, ECCS is unaffected at El 317.
Initiating Event: I Initiating Event ID: T5 Initiating Event Recovery: No reco try from a feedwater line break. Reactor trip (T6) will occur immediately on low steam generator level and 2FW-5B prevents steam generator blowdown and low pressure isolation in the short term.
Less of System: SDM 5 System IPE ID: PCS, EFW, HPSI, CSS, LPSI System Recovery: Loss of PCS is assumed due to feedwater pump trip (operator or loss of suction from condenser hot well) or eventual MSIV isolation on low steam generator pressure er operator action. It is possible to recover a condensate pump and makeup to the unfaulted steam generator. Also, EFW could be isolated from the faulted steam generator or the supply valves in Room 2081 could fail from the harsh emironment. Howe /er, there is a discharge path from each EFW pump to the unfaulted steam generator. Failure of MCC 2B52 in Room 2040 is also assumed to affect train A supply valves in the containment spray, HPSI, and LPSI systems.
Loss of Train: N Train ID: N/A Train Recovery: N/A Consequence Comment: Consequence is " Medium' based on Tables 2-1 and 2 3 of Ref. 9.17 (2 backup trains, including EFW and AFW to SC A, FW recovery, and train B of once through cooling supply valves). 2FW 5B provide.: n tainment isolation.
Consequence Category: MEDIUM O Consequence aank D O
i
- Calculation No. A PENG CALC-018, Rev. 00
- . Page 81 of 811 d
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s l APPENDIX B
'FMECA - DEGRADA TION MECHANISMS" (Attachment Pages 81 - 811) i 4
1 4
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{
, ' ABB Combustion Engineering Nuclear Operations
'## C"I"'#"""" #" NG""18 ^" 88 FMECA - Degradation Mechanisms Page B2 of Bil Weld System ID Segment Line Number Line Description Number Weld Location T C P I M E F 0 MFW MFW-001 2DBB-1-24" Main Feedwater Supply 17-010 Upstream orcheck valve No No No No No No Yes No Line to Steam 2FW-5A (ISO 2DBB-1-2)
Generator 2E-24A -
24" piping MFW MFW-001 2DBB-1-24" Main Feedwater Supply 17-010A Downstream of piping No No No No No No Yes No Line to Steam section #7 (ISO 2DBB-1-2)
Generator 2E-24 A -
24" piping MFW hMV-001 2DDB-l-24" Main Feedwater Supply 17-011 Downstream of elbow #17 No No No No No No Yes No Line to Steam (ISO 2DDB-1-2)
Generator 2E-24 A -
2J" piping MFW MFW-001 2DBB-t-24" Main Feedwater Supply 17 012 Upstream ofelbow #17 No No No No No No Yes No Line to Steam (ISO 2DBB-1-2)
Generator 2E-24A -
24" piping MFW MFW-001 2DBB-1-24" Main Feedwater Supply 17-013 Downstream orpiping No No No No No No Yes No Line to Steam section #6 (ISO 2DBB-1-2)
Generator 2E-24A -
24" piping MFW MITV-001 2DBB-t-24" Main Feedwater Supply 17414 Downstream of c!bo v #16 No No No No No No Yes No Line to Steam (ISO 2DBB-1-2)
Generator 2E-24A -
24" piping MFW MFW-001 2DBB-1-24" Main Feedwater Supply 17-015 Upsacam orelbow #16 No No No No No No Yes No Line to Steam (ISO 2DBB-1-2)
Generator 2E-24A -
24" piping Deeradation Mechamsms T-Thermal Fatigue P- Pnmary water Stress Corrosaan Cracking (PWSCC) M - Microtnologically Inbenced Cerros en (MIC) F-Flow AccelerrwlCerrosen C- Cerrosion Cracting I - Intergranutar Stress Cerrassen Cracking (IGsCC) F - Lamien - Cavitation 0- Other O O O
O O O
' **7 C*"'"" " #" A"G"#8 #" 88 FMECA - Degradation Mechanisms -
Page B3 of Bil W eld Systest ID Segment Line Number Line Description Number Weld IAestion T C P 1 M E F 0 !
MFW MFW-001 2DBB-t-24" Main Feedwater Supply 17-016 Downstream of piping No No No No No No Yes No Line to Steam section #49 (ISO 2DBB Generator 2E-24A - 2) !
24" piping MFW MFW-001 2DBB-1-24" Main Feedwater Supply 17-017 Downstream ofelbow #15 No No No No No No Yes No Line to Steam (ISO 2DBB-l-2)
Generator 2E-24 A - ,
24" piping i MFW MFW-001 2DBB-1-24" Main Feedwater Supply 17-018 Upstream of elbow #15 No No No No No No Yes No Line to Steam (ISO 2DBB-1-2)
Generator 2E-24A -
24" piping MFW MFW-001 2DBB-1-24" Main Feedwater Supply 17-018A Downstream cfpiping No No No No No No Yes No Line to Steam section #4 (ISO 2DBB-t-2)
Generator 2E-24A -
24" piping MFW MFW-001 2DBB-1-24" Ma in Feedwater Supply 17-019 Downstream ofelbow #14 No No No No No No Yes No l Line to Steam (ISO 2DBB-1-2)
Generator 2E-24A - t 24" piping MFW MFW-001 2DBB-1-24" Main Feedwater Supply 17-020 Upstream ofelbow #14 No No No No No No Yes No Line to Steam (ISO 2DBB-1-2)
Generator 2E-24A -
24" piping MFW MFW-001 2DBB-1-24" Main Feedwater Supply 17-021 Downstream of piping No No No No No No Yes No Line to Steam section #47 (ISO 2DBB Generator 2E-24A - 2) 24" piping C_' _ Mechanimes T-Dermal Fatigue P - Pnmary Water Strees Carrossen Cracking (PWSCC) M - Microbiologically Innuenced Cerroman (MIC) F- Flow Accelerseed Carramen C-Carromen Cracking I-Insergranular strees Corroman checking (losCC) E - Eraman- Caviestien O-Odur
14497 FMECA - Degradation Mechanisms Calculation No. A-PENG-C4LC-018. Rev. 00 7,y, 3, of 3,, l Weld System ID Segment Line Number Line Desca;;stion Number Weld Location T C P 1 M E F 0 MFW MFW-001 2DBB-1-24" Main Feedwater Supply 17-022 Downstream ofelbow #19 No No No No No No Yes No Line to Steam (ISO 2DBB-t-2)
Generator 2E-24A -
24" piping MFW MFW-001 2DBB-1-24" M-ir. Feedwater Supply 17-023 Downstream of cibow #13 No No No No No No Yes No Line to Steam (ISO 2DBB-I-2)
Generator 2E-24A -
24" piping MFW MFW-00' 2DBB-1-24" Main Feeduster Supply 17-024 Upstream orelbow #13 No No No No No No Yes No Line to Steam (ISO 2DBB-1-2)
Generator 2E-24A -
24" piping MFW MFW-001 2DBB-1-24* Main Feedwater Supply 17-025 Downstream ofcibow #12 No No No No No No Yes No Line to Steam (ISO 2DBB-1-2)
Generator 2E-24A -
24" piping MFW MFW-001 2DBB-1-24" Main Feedwater Supply 17-025A Upstream ofcibow #12 No No No No No No Yes No Line to Steam (ISO 2DBB-1-2)
Generator 2E-24A -
24" piping MFW MFW-001 2DBB-1-24" Main Feedwater Supply 62-002 Upstream of flued head No No No No No No Yes No Line to Steam #* I (ISO 2DBB-1-1)
Genuator 2E-24A -
24" piping MFW MFW-001 2DBB-1-24" Main Feeduster Supply 62-003 Downstream orelbow #1 No No No No No No Yes No Line to Steam (ISO 2DBB-1-1)
Generator 2E-24A -
24" piping Deeradatiori Mecharusrrs T-Thermal Fatigue P-Ihmary Water Stress Corrosion Crecimg (PWSCC) M - Micratmologmally Ldluenced Cerrosion (MIC) F-Fkm Accelermed Cerroman C-Cerronica Cracking I- Irmergranular Stress Corrown Craciang (IOSCC) E- Domion -Cavitanice 0-Other O O O
f'\
V f, (d
'## C"I'" lad n A*a ANMW/8, Rev 00
. FMECA - Degradation Mechanisms Page B5 of Bil . t W eld System ID Segment Line Number Line Description Number Weld 1mation T C P I M E F 0 '
MFW MFW-001 2DBB-1-24" Main Feeduster Supply 62-004 Upstream orelbow #1 No No No No No No Yes No !
Line to Steam (ISO 2DBB-1-1) i Generator 2E-24 A -
24" piping MFW MFW-001 2DBB-1-24" Main Feedwater Supply 62405 Downstream of cibow #2 No No No No No No Yes No Line to Steam (ISO 2DBB-1-1)
Generator 2E-24A - t 24" piping MFW MFW-001 2DBB-1-24" Main Feedwater Supply 62-006 Downstream of motor- No No No No No No Yes No Line to Steam operated valu: CV-1024-1 Generator 2E-24A - (ISO 2DBB-1-1) 24" piping MFW MFW-001 2DBB-2-24" Main Feedwater Supply 19-016 Downstream of elbow #16 No No No No No No Yes No Line to Steam (ISO 2DBB-2-2)
Generator 2E-24B -
24" piping MFW MFW-001 2DBB-2-24" Main Feedwater Supply 19417 Upstream of elbow #16 No No No No No No Yes No t Line to Steam (ISO 2DBB-2-2) i Generator 2E-24B -
24" piping MFW MFW-001 2DBB-2-24" Main Feedwater Supply 19418 Downstream ofpiping No No No No No No Yes No i Line to Steam secuon #5 (ISO 2DBB-2-2) !
Generator 2E-24B - t 24" piping MFW MFW-001 2DBB-2-24" Main Feedwater Supply 19-019 Downstream orpiping No No No No No No Yes No ;
Line to Steam section #4 (ISO 2DBB-2-2) ;
Generator 2E-24B -
24" piping Draradetsae Mechanums T-Thmnat Fatigue P - Pnmary Water Stress Corressen Cracking (PWSCC) M-ML ' j "y bdhnencedCorreusen(MIC) F-Flow AccelsessedCorressen C CerrossenCracking I-beergranularStressCarresia Crackmg(IGSCC)
. E - ksessen-Cavitation 0 -Oent 3
FMECA - Degradation Mechanisms Col dadon No. A-MWWIS. Rm 00 Page B6 of Bil Weld System ID Serment Line Number Line Description Number Weld Imation T C P I M E F O MFW MFW-001 2DBB-2-24" Main Feedwater Supply 19-020 Domistream orelbow #15 No No No No No No Yes No Line to Steam (ISO 2DBB-2-2)
Generator 2E-24B -
24" piping MFW MFW-001 2DBB-2-24" Main Feedwater Supply 19-021 Upstream ofelbow #15 No No No No No No Yes No Line to Steam (ISO 2DBB-2-2)
Generator 2E-24B -
14" piping MFW MFW-001 2DBB-2-24" Main Feedwater Supply 19-022 Upstream ofelbow #14 No No No No No No Yes No Line to Steam (ISO 2DBB-2-2)
Generator 2E-24B -
24" piping MFW MFW-001 2DSB-2-24" Main Feedwater Supply 19-023 Downstream ofelbow #13 No No No No No No Yes No .
Line to Steam (ISO 2DBB-2-2) !
Generator 2E-248 -
24" piping MFW MFW-001 2DBB-2-24" Main Feedwater Supply 19-024 Upstream of cIbow #13 No No No No No No Yes No Line to Steam (ISO 2DBB-2-2)
Ger.erator 2E-24B -
24" piping MFW MFW-001 2DBB-2-24" Main Feedwater Supply 64-001A Upstream of flued head #1 No No No No No No Yes No Line to Steam (ISO 2DBB-2-1)
Generator 2E-2 tB -
24" piping MFW MFW-001 2DBB-2-24" Main Feedwater Supply 64-002 Downstream ofelbow #6 No No No No No No Yes No Line to Steam (ISO 2DBB-2-1)
Generator 2E-248 -
24" piping Dearadatum Medianums T-Thermal Fatigue P - Pnmary Wster Stress Carmseen Cracking (PWSCC) M - Micratmokycally Infh.enced Cerresum (MIC) F-Flour Accelerased Cerroews C- Cerrasson Cracking I- Intergranular stress Cerrosum Cracting OGsCC) E- Erossen-Cantarson 0-Other e - - -
G
r r\
b> b O .
'# # # C"I"d"" " ^'" Ad^MW'8 R" 88 FMECA - Degradation Mecitanisms Pag
- B7 of BiI W eld System ID Segment Line Number Line Description Number Weld 1 mention T C P I M E F 0 MFW MFW-001 2DBB-2-24" Main Feedwater Supply 64-003 Upstream of elbow #6 No No No No No . No' Yes No Line to Steam (ISO 2DBB-2-1)
Generator 2E-24B -
24" piping MFW MFW-001 2DBB-2-24" Main Feedwater Supply 64-004 Downstream ofcibow #5 No No No No No No Yes No Line to Steam (ISO 2DBB-2-1)
Generator 2E-24B -
24" piping MFW MFW-001 2DBB-2-24" Main Feedwater Supply 64-005 Upstream ofelbow #5 No No No No No No Yes No Line to Steam (ISO 2DBB-2-1)
Generator 2E-24B -
24" piping MFW MFW-001 2DBB-2-24" Main Feedwater Supply 64-006 Downstream of elbow #4 No No No No No No Yes No Line to Steam (ISO 2DBB-2-1)
Generator 2E-24B -
24" piping MFW MFW-001 2DBB-2-24" Main Feedwater Supply 64-007 Upstream ofelbow #4 No No No No Ne No Yes No Line to Steam (ISO 2DBB-2-1)
Generator 2E-24B -
24" piping MFW MFW-001 2DBB-2-24" Main Feedwater Supply 64 @ 8 Downstream of motor- No No No No No No Yes No Line to Sicam operated valve 2CV-1074-Generator 2E-24B - 1 (ISO 2DBB-2-1) 24" piping MFW MFW-002 2DBB-1-18" Main Feedwater Supply 17-002 Item 54 at 2E-24A Yes No No No No No Yes No Line to Steam Generator 2E-24 A -
18" piping Dearadshan Mechanians T-Thmnal Fatigue P - Pnmary Wseer Stress Cerrosien Oncbng (PwSCC) M - MM.la", Innuenced Corresson (MIC) F-mw Accelerused Common C-Corvoeien Cracbng I - Irmergranular Stress Corressen Cracbng (IGSCC) E-Ercasen-Cavitafien 0-Other
'# # 7 C""'"" " #^ NG-C'iWl8, Rm 00 FMECA - Degradation Mechanisms Page BS of Bil W eld System ID Segment Line Number Line Description Number WelJ Location T C P I M E F 0 MFW MFW-002 2DBB-l-18" Main Feedwater Supply 17-003 Downstream of 24* x 18" Yes No No No No No Yes No Line to Steam reducing cibow #21 (150 Generator 2E-24 A - 2DBB-1-2) 18" piping MFW MFW-002 2DBB-l-24" Main Feedwater Supoly 17-004 Upstream of elbow #21 Yes No No No No No Yes No Line to Steam (ISO 2DBB-!-2)
Generator 2E-24A -
24" piping MFW MFW-002 2D53B-1-24" Main Feedwater Supply 17-005 Dowitstream of elbow #18 Yes Nc No No No No Yes No Line to Steam (ISO 2DBB-1-2)
Generator 2E-24A -
24" piping MFW MFW-002 2DBB-1-24" Main Feedwater Supply 17-006 Upstream of cibow #18 Yes No No No Nc No Yes No Line to Steam (ISO 2DBB-1-2)
Generator 2E-24A -
24" piping MFW MFW-002 2DBB-1-24" Main Feedwater Supply 17-006A Item 41 branch connection Yes No No No No No Yes No Line to Steam Generator 2E-24A -
24" piping MFW MFW-002 2DBB-1-24" Main Feedwater Supply 17-007 Dumouwn of elbow CO Yes No No No No No Yes No Line to Steam (ISO 2DBB-1-2)
Generator 2E-24A - ,
24" piping MFW MFW-002 2DBB-1-24" Main Feedwater Supply 17-008 Upstream of elbow #20 Yes No No No No ,No Yes No !
Generator 2E-24A -
24" piping Deeradaten Mechanums T-Thermal Fatigue P - Prunary Water Stress Cecrecen Cracking (PWsCC) M - Microtmolopeally Inomenced Common (MIC) F-Flow AcxulerstedCerroman C- Carraman Cracking I-Intergranular Stres Common Cracing (IOSCC) E - Eranon - Cantauon 0-Other O O O
. -. - -- - . _ - =_- - . - . . . - - ,
O O O
"*" FMECA - Degradation Mechanisms C""'"" e N . A-l'SMW18. Rex 00 Page 89 of B1i W eld System ID Segment Line Number Line Description - Number Weld IAestion T C' F- I M E F 0 t MFW MFW-002 2DBB-1-24" Main Feedwater Supply 17-009 Downstream orcheck Yes No No No No NJ Yes No Line to Steam valve 2FW-5A (ISO Generator 2E-24A - 2DBB-1-2) 24" piping MFW MFW-002 2DBB-2-18" Main Feedwater Supply 19-002 Item 52 at 2E-24B Yes No No No No No Yes No ,
Line to Steam Generator 2E-24B - '
18* piping MFW MFW-002 2DBB-2-18" Main Feedwater Supply 19-003 Downstream of 24" x 18" Yes No No No No No Yes No Line to Steam reducing elbow #20 (150 Generator 2E-24B - 2DBB-2-2) ,
18" piping MFW MFW-002 2DBB-2-24" Main Feedwater Supply 19-004 Upstream ofelbow f20 Yes No No No No No Yes No i Line to Steam (ISO 2DBB-2-2)
Generator 2E-24B -
24" piping .
MFW MFW4P]2 2DBB-2-24" Main Feedwater Supply 19-005 Downstream of piping Yes No No No No No Yes No Line to Steam secten #11 (ISO 2DBB t Generator 2E-?tB - 2) 24" piping MFW MFW-002 2DBB-2-24" Main Feedwater Supply 19-006 Downstream of piping Yes No No No No No Yes No Line to Steam section #10 (ISO 2DBB !
Generator 2E-24B - 2) 24" piping l MFW MFW-002 2DBB-2-24" Main Feedwater Supply 194X)8 Downstream of piping Yes No No No No No Yes No :
Line to Steana section #9 (ISO 2DBB-2-2)
Generator 2E-24B -
24" piping DW- Mechanum_s T-11wrinal Fatigue P - Pnmary Water Stress Corr.mion Cracking (PW3CC) M-MR
- J y Induenced Cerrassen(MIC) F- Flow Accelerseed Carrassen C-Carramen Cracting 1 - :..:-. - Swess Corronen Craciung OGSCC) E - Eremen-Caviestien 0-Other
' #*" C"#'"'"" " #" " " '#-##8 # " 88 FMECA - Degradation Mechanisms Page B10 of Bil Weld System ID Segment Line Number Line Description Number Weld Iecation T C I P M E F 0 MFW MFW-002 2DBB-2-24" Main Feedwater Supply 19-009 Item 21 branch connection Yes No No No No No Yes No >
Line to Steam Gencrator 2E-24B -
24" piping MFW MFW-002 2DBB-2-24" Main Feedwater Supply 19-010 Dornstream ofcibow #19 Yes No No No No' No Yes No Line to Steam (ISO 2DBB-2-2)
Generator 2E-24B -
24" piping MFW MFW-002 2DBB-2-24" Main Feedwater Supply 19-01I Upstream ofcibow #19 Yes No No No No No Yes No Line to Steam (ISO 2DBB-2-2)
Generator IE-248 -
24" piping MFW MFW-002 2DBB-2-24" Main Feedwater Supply 19-012 Downstrearn ofelbow #18 Yes No No No No No Yes No Line to Steam (ISO 2DBB-2-2)
Generator 2E-24B -
24" piping MFW MFW-002 2DBB-2-24" Main Feedwater Supply 19413 Upstream ofelbow #18 Yes No No No No No Yes No Line to Steam (ISO 2DBB-2 2)
Ge1erator 2E-24B -
24" piping MFW MFW-002 2DBB-2-24" Main Feedwater Supply 19414 Upstream ofelbow #17 Yes No No No No No Yes No Line to Steam (ISO 2DBB-2-2)
Generator 2E-24B -
24" piping MFW - MFW-002 2DBB-2-24" Main Feedwater Supply 19-015 Downstream ofcheck Yes No No No Nc No Yes No Line to Steam valve 2FW-513 (ISO 2DBB-Generator 2E-24B - 2-2) 24" piping Deeradaten Mechamsms T "IhermalFatigue P - Pnmary Water Stress Cerrosion Cracbng (PWSCC) M - Micratnologicany Innuenced Common (Mh3 C- Carassen Cracking F- Nw #ai..a " Carroman 1 -Intergranular Stress Cerros ac Crecimg (10 SCC) E- Eremon - Cavitatson 0 -OCaer e G #
aa +
O O O
' * " ' Ncularms Na AMMWI& Rm 00 FMECA - Pegradation Mechanisms Page Bf1 of Bil Weld System ID Segment Line Number Line Description Nmber Weld Imcation T C P I M E F O L
MFW MFW-003 2DBB-1-18" Main Feedwater Supply 17-001 Downstream of piping Yes No No No Yes No Yes No Line to Steam section #54 (ISO 2DBB-I-Generator 2E-24 A - 2) 18" piping MFW MFW-003 2DBB-2-18" Main Feedwater Supply 19-001 Dowinstream of piping Yes No No No Yes No Yes No Line to Steam section #52 (ISO 2DBB Generator 2E-24B - 2) 18" piping l
l Dearadshon Mechanisms T-Thermal Fatigue P - Pnmary Water Stress Common Cracking (PWSCC) V - Micrainologsemity bdhsenced Carroman (MIC) F- Flow Accelerseed Cerrassen C-Cerrasson Cracking 1 - Intergranular Stress Cerroman Cracking OGSCC) E-Eronen-Cavenhan 0 -Oiher
_ - _. . _ _ . _ _ _ _ . _ _ _ . _ _ _ . _ . . . . _ _ _ _ . _ . _ _ _ _ _ _ _ . . . _ - _ _ _ _ . _ _ _ ~ . . _ _ _ _ _ . . . . _
,r l- Calculation No. A PENG CAL 5018, Rev. 00-Page C1 of C2 l
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APPENDIX C l
'FMECA - SEGMENT RISK RANKING REPORT" (Attachment Pages C1 - C2) i 4
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8 *'7 FMECA - Segment Risk Ranking Report ch
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r.,, a of a Degradation Number Lines in Welds in Degradation Degradation Mechanism Consequence Risk Risk Segment ID of We'ds Segment Segment Mechanisms Group ID Category Category Category Rank -
MFW-001 41 2DBB-1-24",17-010,17-010A,17- F MFW-F LARGE hEDIUM CAT 3 HIGH 2DBB-2-24" 01I.17-012,17-013, LEAK 17-014,17-015, 17-016,17-017,17 018, 17-018A,17-019,17-020,17 ,21,17 4 22,17-023, 17-024,17-025,17-025 A,62-002,62-003, 62 4 04,62-005, 62 4 06 // 19-016,19-017, IG 'J18,19-019, 19-020,19-021,19-022, 19-023, I9-024,64-001 A,64-002,64-003, 64 4 04,64-005, 64-006,64-007,64-008 MFW-002 22 2DBB-1-18",17-002, 17-003 // 17- T. F MFW-T F LARGE hEDIUM CAT 3 HIGH 2DBB-1-24", 004,174 05,17-006 LEAK 2DBB-2-18", 17-006A,17-007,17-2DBB-2-24" 008,17-009 // 19-002,19-003 // 19-004,19-005,19-006,19-008, 19-009,19-010,19-011,19-012, 19 4 13,19-014, 19-015 MFW-003 2 2DBB-t-18*, 17-001//19 # 1 T,M.F MFW-T L F LARGE hEDIUM CAT 3 HIGH 2DBB-2-18" LEAK O O O
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- Calculation No. A-PENG CALC 018, Rev. 00
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!- QUAllTY ASSURANCE VERiflCA TION FORMS 6
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C:Icui: tion No. A PENG CALC 018 R:v. 00 Page D2 of D5 Verification Plan
Title:
Implementation of the EPRI Risk-Informed Inservice Inspection Evaluation Procedure for the MFW System at ANO-2 Document Number: C-PENG-CALC-018 Revision Number: 00 instruetbns: Describe the method (s) of verification to be employed, i.e., Design Review, Altemate Analysis, Qualification Testing, a combination of thcse or an attemative. The Design Analysis Verification Cbecklist is to be used for all Design Ana!ytes. Other elements to consider in formulating the plan are: methods for checking calculations; comparison of results p tirnilar analyses, etc.
Descrintion of Verification Method; An independent review will be conducted as appropriate with the work activities described in Project Plan PP 2OOOB39, Revision 00. The verification willinclude:
- 1. Verification of a Design Analysis by Design Review (per QP 3.4 of the Quality Procedures Manua .
- 2. Verification that the appropriate methodology is selected and correctly implemented
- 3. Verify all design input (as applicable) is appropriately and correctly obtained from traceable sources.
4, Review that the assumptions, results, conclusions, report format, ... etc. are made in accordance with Design Analysis Verification checklist.
Verification Pla prepared by: Approved by:
R.e Jn,n independent Reviewer prmted name and sagrpude'/
A v.w. M .
e Management approver t itmted name and signature ABB Combustion Engineering Nuclear Operations
1 C:Icul2 tion N2, A-PENG-CALC 018, R:v, 00 Page D3 of DS Other Design Document Checklist f))
\ (Page1of3) incructione ne Independent Re,iewer is to complete this checklist for each Other Design Document. This Checklist is to be made part of the Quality Record package, although it need not be made a part of or distdbuted with the document itse!f. He second section of this checklist lists potential topics which could be relevant for a particular"Other Design Du:ument', If they are applicable, then the relevant section of the Design Analysis Verification Checklist shall be completed and attached to this checklist.
, (Sections of the Design Analysis Verification Checklist which are not used may be leil blank.)
Titic: - Implementation of the EPRI Risk Informed Inservice Inspection Evaluation Procedure for the i MFW System at ANO-2 Document Number: Revision Number:
4 A-PENG-CALC-018 00 Section 1: To be completed for all Other Design Documents Yes N/A 1
Overall Assessment 1 Are the results/ conclusions correct and appropriate for their intended use? O
- 2 Are all limitations on the results/ conclusions documented? 8 Documentation Requirements
- 1. Is the documentation legible, reproducible and in a form suitable for filing and retrieving as a Quality @
Record?
U 11. Is the document identified by title, document number and date? @
lit, Are all pages identiGed with the document number including revision number? @
IV. Do all pages have a unique page number? @
V. Does the content clearly identify, u applicable:
1 A. objective O O B. design inputs (in accordance with QP 3.2) E O C. conclusions - S O VI. Is the veriGeation status of the document indicated? @
Vll. If an Independent Reviewer is the supervisor or Project Manager, has the appropriate approval been E D documented?
4 Assumptions
- 1. Are all assumption identified, justified and documented? O O II. Are all assumptions that must be cleared listed? 8 O A. Is a process in place which assures that those which are CENO responsibility will be cleared? O E B. Is a process in place wHch a<sures that those which are the customer's responsibility to clear will O O j be indicated on transmhtals to the customer 7 g
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C:Iculati:n N3. A PENG CALC-Ol8, R:v. 00 Page 04 of DS ,
1 Other Design Document Checklist fPage 2 of 3) g' Assessment of Significant Design Changes Yes N/A
- 1. llave significant design-related changes that might impact this document been considered? O II. If any such changes have been identified, have they been adequately addressed? O O Selection of Design inputs
- 1. Are the design inputs documented? @
. .re the design inputs correctly selected and traceable to their source? g
!!!. Are references as direct as possible to the original source or documents containing collection /tabutations of G inputs?
IV. Is he reference notation appropriately specific to the information utilized? g V. Are the bases for selection of all design inputs documented? g VI. Is the verification status of design inputs transmitted from customers appropriate and docurnented? O O VII. Is the verification status of design inputs transmitted from ABB CENS appropriate and documented? O G Vill. Is the use of customer controlled sources such as Tech Specs, UFSARs, etc. authorized, and does the S O authorization specify amendment level, revision number, etc.?
References l
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- 1. Are all references listed? g II. Do the reference citations include sufTicient information to assure retrievability and unambiguous location g of the referenced material?
Section 2: Other Potentially Applicable Topic Areas-use appropriate sections of the Design Analysis Verification Checklist (QP 3.4, Exhibit 3.4 - 5) and attach.
Yes N/A
- 1. Use of Computer Software O O
- 2. Applicable Codes and Standards O O
- 3. Literature Searches and Background Data O 8
- 4. Methods O E
- 5. Hand Calculations O O
- 6. List of Computer Software O 8
- 7. List of Microfiche O E
- 8. List of optical disks (CD-ROM) O S
- 9. List of computer disks O B m ABB Combustion Engineering Nuclear Operations
Calcul: tion No. A PENG CALC 018, R:v. 00 Page D5 of DS
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C/ Other Design Document Checklist (Page 3 of 3)
Independent Reviewer's Comments Comment Reviewer's Comment Response Author's Response Response Number Required? Accepted?
I Page 16 refers to consequence Yes The consequence Yes numbers as FW-XXX numbers were vs. changed to match the numbers in Appendix A 2 Page 17, Table 3A refers to Yes Same as response Yes MF-XXX to item 1 Vs.
3 Pages 35-37, Table 10 refers to Yes Same as response Yes MFW-XXX to item 1 4 Appendix A refers to consequences Yes Same as response Yes O
%J nubmers as FW XXX to item 1 5 I recommend using"FW" in all Yes Same as response Yes places to item 1 6 Page 33, Table 8 refers to Yes Same as response Yes MFW-XXX to item 1 4
Checklist completed by:
Independent Reviewer go668T E.
Pnnted Name h4Goa7N '
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hignatp/r/'
8///97 Date O
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