ML20217F029
| ML20217F029 | |
| 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-019, A-PENG-CALC-019-R00, A-PENG-CALC-19, A-PENG-CALC-19-R, NUDOCS 9710070346 | |
| Download: ML20217F029 (93) | |
Text
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i Arkansas Nuclear One - Unit 2 i
l Pilot Plant Study 4
Risk-Informed Inservice Inspection Evaluation for the
- Main Steam System t
i l September 1997 i
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(gg) A PENO CALC 019 Revision 00 Page1 Design Analysis Title Page j C 's l
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Title:
Implementation of the EPRI Risk Informed Inservice Inspection Evaluation i Procedure for the Main Steam System at ANO 2 Document Number: A PENG-CALC-019 Revision 00 l
_ Number:
Quality Class:
O QC 1(safety Related) O QC 2 (Not Safety Related) @ QC-3 (Not Safety-Related)
- 1. Approvalof Completed Ana!ysis This Design Analysis is complete and verified. Management authorizes the use ofits results.
Printed Name Signature Date Cognizant Engineer (s) R. A. Weston gg 7[lkf7 A. V.1.aer h 6 g q Mentor a None [j/ l Independent Reviewer (s) R. E. Jaquith
[ /[ [ $[/f/g CN s Management Approval B. T. Lubin g } jgQ-)
Project Manager h
- 2. Package Contents (this section may be completed after Management approval):
Tof al page count, including body, appendices, attachments, etc. 92 List associated CD-ROM disk Volume Numbers and path narnes: g None Note: CD ROM are stored as separate Quality Records CD-ROM Volume Path Names (to lowest directory which uniquely apphes to this document)
Numbers Total number of sheets of microfiche: g None Number of sheets:
Other attamments (specify):
- 3. Distributien:
6 g B. Boya (2 copies)
N i:\ data \lubin\rbifinal\apeng019. doc
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- 't BF IF Calculation No. A PENG-CALC-019, Rev. 00 Page 2 of 41 RECORD OF REVISIONS Rev Date / ages Changed Prepared By Approved By 00 A#g Original R. A. Weston R. E. Jaquith jlf A. V. Bauer i
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- % FIF f,} Calculation No. A.PENG. CALC-019, Rev. 00 V, . Page 3 of 41 TABLE OF CON 7ENTS SECTION PAGE
- 1. 0 PURPOSE...............................................................................................................5 2.0 SCOPE..................................................................................................................5
- 3. 0 SYSTEM IDENTIFICA T.'ON AND BOUNDARY DEFINITION ............................................ 6
- 4. 0 C0NSEQ UENCES EVA L UA TlON . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 10 4.1 CONSEQUENCE A SSUMP TIONS/ INPUTS ..... .......... .. ................... ..... ................ 12 4,2 C0NSEO UENCE IDENTlFICA TlON . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 13 4.3 SHUTDOWN OPERA TION AND EXTERNAL EVENTS........................................... 13 5.0 DEGRA DA TION MECHANISMS EVAL UA TION ........................................................... 22 5.1 OA MA GE GR O UPS. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 3 5.2 DEGRADA TION MECHANISM CRITERIA AND IDENTIFICA TION ........................... 23 5.3 BASICDATA................................................................................................28
- 6. 0 SERVICE HISTG Y AND SUSCEPTIBILITY REVIEW.................................................... 30
- 7. O RISK E VA L UA TlON . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 33
- 8. 0 EL EMEN T SEL EC TION . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 7
- 9. 0 REFERENCES........................................................................................................39 LIST OF TABLES
%) NUMBER PAGE I MSS B O UNDA RIES . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8 2 MSS CONSEQ UENCE A SSESSMENT SUMMA R Y . ...... ... ............. . .... ............ .. ....... ... ... 17 3A MSS CONSEQUENCE. FIGURES AND ISOMETRIC DRA WINGS .................................... 19 3B MAIN STEAM (MS) PIPING LOCA TION & CONSEQUENCES ..................... .................. 19 4 DA MA GE GR O UPS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 22 5 DEGRADA TION MECHANISM CRITERIA AND SUSCEPTIBLE REGIONS........ . . . . . . . . . . . . . 24 6 MAIN STEAM SYSTEM LINES AND OPERA TING CONDITIONS.................................... 29 7 SERVICE HISTORY AND SUSCEPTIBILITY REVIEW- MAIN STEAM SYSTEM................. 32 8 RISK SEGMENT IDENTIFICA TION . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 34 9 RISK /NSPEC TlON SCOPE. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 7 10 ELEMENT SELEC TION - RISK CA TEGOR Y 3 ........ .................. ....... ............ ....... ... ....... 38
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ABB Combustion Engineering Nuclear Operations
ABB Calculation No. A.PENG. CALC 019, Rev. 00 Page 4 of 41 ilST OF FIGURQS NVAMER P. AGE 1 A NO .2 A1A IN S TEA Af S YS TEAf . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 9 2 Af A IN S TEA Af FL O W PA THS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 0 3 S TEA Af FL O W PA TH TO EAV l'UAfP. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 21 llST OF APPENDICES A FAfECA . CONSEQUENCE INFORAfA TION REPORT B FAfECA DEGRADATION AfECHANISAfS C FMECA . SEGAfENT RISK RANKING REPORT O QUAUTY ASSURANCE VERIFICA TION FORMS O
O ABB Combustion Engineering Nuclear Operations
C\ Calculation No. A PENG CALC 019. Rev. 00 Y Page 5 of 41
- 1. 0 PURPOSE The purpose of this evaluation is to document the implementativ of the Electric Power Research Institute (EPRll Risk Informed Inservice inspection Evaluation Procedure (RISil of Reference 9.1 for the Afsin Steam System (AfSS) at Arkansas Nuclear One, Unit 2 (ANO 2),
Entergy Operations, Inc. The RISI evaluation process trovides an alternative to the requirements in ASAfL Section XI for selecting inspection locetlons. The purpose of RISIis to identily risk srgnificant pipe segments, define the locations that are to be inspected within these segments, and identify appropriate inspection methods.
TiJs evaluatior, is performed using the guidelines of the EPRI Risk Informed Inservice Inspection Evaluation Procedure of Reference 9.1 and in accordance with the requirements of the ABB Combustion Engineering Nuclear Operations Quality Procedures Afsnual(CPA1101).
2.0 SCOPE This evaluation procedure applies to the MSS at ANO 2, and utilfres the ISIS Software (Reference 9.21, which has been specifically develnoed to support and document this procedure.
As part of the procedure, the system bour.daries and functions are identified. A risk evaluation is performed by dividing the system into piping segments which are determined (G) to have the some failure consequences and degradation mechanisms. The failure consequences and degradation mechanisms are evaluated by assigning the segment to the appropriate risk category and iden*"ying 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.
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ABB Combustion Engineering Nuclear Operations
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13 ABB Calculation No. A PENG CALC 019, Rev. 00 Page 6 of 41 3.0 SYSTEM IDENTIFICA TION AND ROUNDARY DEFINITION
3.1 System Description
The Afain Steam System (AfSSI is designed to convey steam generated by the steam generefors to the turbine generator and to other auxiliary equipment for power generation. it supplies steam to the high pressure turbine and to the moisture separator rehesters during normal plant operation, to the turbine gland seals during low loads, and to the main feedwa tv pumps steam turbine drivers during low loads or whenever low pressure steam is not sufdcient.
The AfSS is designed to remove the heat generated in the Nuclear Steam Supply System (NSSS) during normalload, plant startup, hot standby, hot shutdown and normal cooldown, and to permit load reductions of up to fullload. This function is accomplished by means of the Steam Dump end Bypass Control System (SDBCS) In conjunction with the main condenser, Emergency Feedwater System (EFS) and/or main steam safety valves.
The AfSS is designed to provide isolation of the steam generators from other components of the system by means of the steam block valves following a hypotheticalinsin steam line break. Additionally, the AfSSis designed to provide an assured source of steam to operate the emergency feedwater pump turbine driver.
3.2 System Boundary h
G The Afoin Steam system is described consistent with the FSAR (Reference 9.3). The scope of this analysis includes all Class 2 and Class 3 piping in this system which is currently examined in the ANO 2, ASAfE Section XI Inserv'o.e Inspection (ISI) Program (Reference
- 9. 6), in addition, all Class 2 and 3 piping which as in the AfSS flow path is also included in this evaluation. The code and non code lines which are part of or interface with the MSS were evaluated to determine their risk significance. The system boundaries are defined in Table 1 and figure 1. Certain line segments contain welds that were not entered it, the .
- tatabase (Reference 9.2) as outlined below:
- 3. 2.1 Lines downstream of the Afain Steam Safety Valves (AfSSVsl (2GBD 7710', 2GBD-78 10')
These Ine segments provide pressure protection for the AfSS by relieving steam directly to the atmosphere. A failure in any of these line segments during normal power operation would not cause an initiating event and would not prevent safety related equipment from performing their design functions. The ability to relieve main steam pressure svovid not be affected due to a segment failure. Because a segment failure downstream of the AfSSV does not impact pressure relief, a NONE consequence category would be assigned to these segments. Based on this assigned consequence category, the risk significance of a segment failure would be LOW (i.e., CAT 7). Since no element selections are needed for low risk significant segments, the welds for these lines were not enteredin the database.
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ABB Calculation No. A PENG. CALC 019, Rev. 00 Page 7 of 41 3.2.2 Lines downstream of the Atmospheric Dump Vs/ves (ADVs) (2GBD 7512", 2GBD.
76 12')
These line segments provide en attemate means of removing hest from the secondary side by relieving steam directly to the atmosphere In a controHed manner if the primary heat sink (i.e., the condensert becomes unsysHable. During normal power operation, these line segments are isolated by the closed ADVs. A failure in any of the above line segments would not cause en initisting event and would not prohibit the removal of heat from the primary and secondary sides of the plant.
Because the removal of RCS and core heat can stin be accomplished, a NONE consequence category would be assigned due to a segment failure. Based on this assigned consequence category, the risk significance of a segment isHure would be LOW (i.e., CA T 7). Since no element selections are needed for low risk significant segments, the welds for these lines were not entered in the database.
1 3.2.3 Lines downstream of the Main Steam Isolation Valves (MSIVs) (2EBD 138', 2EBD-2 38', 2EBD 3 6', 2EBD-4 6', 2EBD 1910", 2EBD 2010*, 2EBD 21 10", 2EBD.
- 2210', 2EBD 2310', 2EBD 4414', 2EBD 810', 2EBD 15 24", 2EBD 16 24')
1 The steam lines downstream of the main steam isolation valves are used to supply steam to the turbine generator, moisture separator rehesters , turbine gland seals, main feedwater pump turbines, and the valves in the steam dump and bypass system. None of this equipment are required for shutting down the plant or maintaining any of the safety functions. Steam is supplied to this equipment for power generation. A failure in any of the above line segments is mitigated by the rapid isolation of steam flow from the steam generators and feedwater flow to the steam generators. For this condition, a Main Steam Isolation Signal (MSIS) is generated which then signals the main feedwater isolation valves and Main Steam Isolation Valves (MSIVs) to close. The Emergency feedwater (EFW) System is then used to maintain the water levelin the steam generators. Because of the rapid isoration of the steam generators, the consequences of a line break downstream of the MSIVis less severe than the consequence of a line break occurring upstream of the MS/V.
The above lines are fabricated of carbon steel material and are susceptible to flow Accelerated Corrosion (FAC). L'ecause of this degradation mechanism, the main steam piping downstream of the MSIVs is already addressed in the existing plant FAC program at ANO2. 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 entered in the database.
3.2.4 Lines with Nominal Diameter of 1' or less Piping with a nominal diameter of I' or less was not explicitly evaluated to determine its risk significance. Since volumetric examination of this piping is not practicable, the most effectivo means to ensure its integrity is via conduction of a system leakage test. Consequently, since this piping is already subject to system leakage testing by the ASME Code, a risk assessment of this piping is not warranted.
ABB Combustion Engineering Nuclear Operations
ABB f~ Calculation No. A.PENG CALC 019, Rev. 00 O Page 8 of 41 TABLE 1 MSS BOUNDARIES Lane Lme 151 pipe pipe Nominal Number Description Drawong Code Diameter (in )
Number Class 2EBBt10' Steam generator 2E 24A 10* steam Ime ptptng 2 EBB 12 2 10 2 EBB 136" Afatn Steam Ltnefrom Steam Generator 2E 24A to 2 EBB.11 2 36 AfSil'2CI'1010-1 36* piping 2 EBB 138* Atatn Steam Linefrom Steam Generator 2E 24A to 2 EBB 11 and 2 38 Afsli'2Cl%I0101 2EBBl2 2EBU 16-2* Aiain Steam Bypass imefor 2CI' 10101 2 EBB 161 2 2 2 EBB 17 2" Atain Steam Bypass linefor 2Cl' 1060 2 2 EBB 171 2 2 2 EBB-210" Steam generator 2E 24B I0' steam ime plo . .ig 2 EBB 2 2 2 10 2 EBB 2 36" Alam Steam Linefrom Steam Generator 2E 24B to 2 EBB-21 2 36 AISii'2Cl%)060-2 36'piptng 2 EBB 2 38" Aiam Steam Linefrom Steam Generator 2E 24B to 2 EBB 21 and 2 38 AfSil'2Cl%1060 2 2 EBB 22
, 2 EBB 6-4' AIam Steam Supply to 2Ci'l000-1 2 EBB-61 2 4 Q 2 EBB-7-4 " EFit' pump Steam Supply to 2CI' 1030-2 2 EBB-71 2 4 2 EBB-8 10~ Alam Steam Du.np to Atmosphere thru 2Cl* 1001 2 EBB-81 2 10 2 EBB 9-10' Alam Steam Dump to Atmosphere thru 2CI'1051 2 EBB 91 2 10 2EBC I 3* Branch Im 2EBC I 2 3 3 2EBC14* Alam Steam Supply to EFil' pump Turbme Drn er 2EBC I I and 3 4 2K 3from Alain Steam fleader #1 2EBC l 2 2EBC 2 4* Alam Steam fleader v2 Supply to EFirpump 2EBC 21 3 4 Turbine Dria r 2K 3 (3
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% 33 FIGURE 1: ANO-2 MAIN STEAM SYSTEM (Raference 9.151 l
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O ABBCalculation No. A PENG CALC 015, Rev. 00 D Page 10 of 41
- 4. 0 CONSEQUENCE EVALUA TION The Alain Steam System (A1SS) is used to remove the heat generated in the reactor core during the various modes of plant operation. The heat removed from the RCS and reactor core by the A1SS is in the form of dry saturated steam. The rejected heat is supplied to the high pressure turbine and moisture separator rehesters during normal power operation.
Steam is also supplied to the main feedwater pump turbine drivers, the turbine gland seals, and the emergency feedwater pump turbine driver under various loads or equipment malfunction conditions. If the condenser becomes unavausble, RCS and core heat removal can stiH be accomplished by opening the Atmospheric Dump Valves (ADA). The Afain Steam Safety Valves (A1SSVs) provide over pressure protection for the steam generators and main steam lines. The opening .etpoint for the A1SSVs assur.as that pressure wiH not exceed the 110% design pressure cn erion for the steam generators. Afoin steam isolation capability is provided by the A1SIVin each of the two steam lines.
The consequence evaluation for the A1SS was performed based on the guidance providedin the EPRI procedure (Reference 9.11. The evaluation focused on the impact of a pipe segment failure on the capability of A1SS 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 A1SS itself. An indirect impact resulting from the failure of a pipe segment would affect neighboring equipment within the A4SS or other system (s). Indirect impacts would generaHy be caused by spraying, or jet impingement of neighboring equipment. Determination of the consequences of a segment
(~) failure considers the potential of losing affected mitigating systems, or trains thereof, and Q the consequentialimpact on the safety functions.
The AfSS consequence evaluation wts performed separately and was documented as a quality assured report, listed as Reference 9.17. Plant locations as defined in the Internal Flood Screening Study (Reference 9.14) were used in that evaluation. The plant locations are summarizedin Section 4.2.1 of Reference 9.17. The report contained herein is one of a series of reports which documents the implementation of the EPRI Risk Informed ISI methodology at ANO 2. As such, appropriate sections of Reference 9.17 were extracted and are provided herein in order to maintain the level of detail, the format structure and consistency with the other reports in the series. The remainder of this section (Section 4.0) along with Sections 4.1 and 4.2 in their entirety provide the information obtained from 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 following individuals participatedin the walkdown and meetings at ANO 2:
Jim Afoody (YAEC Consultant)
Pat O'Regan (YAEC)
The plant was in an unexpected outage and radiological controls would not allow access to m the south piping penetration rooms (Rooms 2084 and 2055) and elevation 317' 0" (Rooms (V) 2006, 2011, 2014, 2007, and 2010). However, this is not judged to have an impact on ABB Combustion Engineering Nuclear Operations I
JL lB Calculation No. A PENG. CALC 019 Rev. 00 Papa 11 of 41 the analysis since spatial 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 following summarises the walkdown observations:
(a) Alain steam piping ares (Room 2155) was investigated where main steam and ERY steam lines were noted along with the main steam isolation valves and ERV steam isolation and check valves. Although unlikely, it is possible for certain large main steam line breaks to impact both steam lines supplying ERV. The main steam lines are near the common well with the refueling area (Room 2151), thus, it is considered possible that a very large steam break could damage and/or vent through this siding. There also is siding and doors that open out to an adjoining building roof and stairway (turbine auxiliary building).
(b) Fuel handling and spent fuelpool area (Room 2151) was walked down. The ERV steam piping was obsetted coming vertically down from the steam piping ares and into a pipe chase which comes out at elevation 335' 0'(Room 2040). This is a very
, large area and it was judged unlikely that main steam piping break propagation to l this area would impact safety equipment which is mostly located at lower elevations. Also, ERY steam line break was judged unlikely to impact safety equipment.
(c) Elevation 335' 0* (Room 2040) is a very large area containing general access, corridors, and several large non safety related rooms. Several floor drains were noted. ERV 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 ERV 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 315'-O'. The most critical component identified in this area is MCC 2852 which powers several train A components.
Although the A1CC is not near analysis scope piping, it is at the east stairway i entrance (the propagation path to elevation 317' O'). If six inches of water could be l
accumulated at elevation 335' 0* or if a very large pipe break occurred, it is consideredlikely that the A1CC could fail.
The type of inputs used and the assumptions made in performing this evaluation are i documented in Reference 9.17 Key inputs and assurnptions, which were extracted from Reference 9.t7 are provided in Section 4.1. Fourteen consequence segments were l
identified for MSS lines entered in the database. Of the fourteen, four were assigned as l
" MEDIUM' and ten as 't OW'. The consequence assessment summary for these segments i is providedin Section 4.2. The bases andjustifications for each of the assigned categories j are providedin Appendix A. This appendix contains reports obtained from the ISIS software 1 (Reference 9.2) for the MSS. For the MSS lines not entered in the database, the l consequence segments were assigned as 'NONE", except those lines which are included in the FAC Program (see Section 3.2).
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- 3 ABB Calculation No. A PENG CALC 019, Rev. 00 4.1 CONSEQUENCE ASSUMPTIONS / INPUTS The following assumptions and input were extracted from Reference 9.17. The type of initiating events and mitigating capabilities considered in thh evaluation are described in detailin Sections 4.3 and 4.4 of Reference 9.17.
4.1.1 Pipe failure can occur et 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 summarites judgments and assumptions regarding which configurations are most important. If pipe failure does not cause a direct initiating event, it is assumed that pipe failure occurs during the accident demand configuration, if applicable. This assumes pipe failure ' 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 Main steam line breaks outside the containment building (Room 2155) are acknowledged as having the potential to propagate into the fuel handling ares (Room 2151), as weil as impact both EFW steam supply lines. These are considered conservative assumptions since such an extremely severe event is considered unlikely. On the other handpropagation into the spent fuelpool area is not assumed to affect safe shutdown equipment.
4.1.3 ANO 2 analysis (Reference 9.18) concludes that an EFW steam line break during Q normal operation will not cause a direct plant trip. However based on more recent Q analyses, this is judged to be marginal; it is assumed that pipe failure will cause a plant trip initiating event. Piping failure in normally pressurized piping up to "CV 0340-2 is assumed to occur during normal standby configuration (i.e., it is assumed that pipe degradation is just as likely to revealitself in standby as during a demand).
Also, it is assumed that environmental impacts are minor relative to safety equipment at elevation 335'O' (Room 2040) If the break is successfully detected and isolated and/or a controlled shutdown is initiated. This is lvdged to be r reasonable assumption given the site of the areas andlocation of key components.
4.1. 4 The IPE Internal Flooding Screening Study (Reference 9.14) identifies impacts in rooms from cable terminalpoints. Since most junction boxes, terminal boxes, etc.
noted during the walkdown were at least a few feet off the floor, these impacts were ignored in the analysis. Also, junction boxes appeared to be tight and sealed, therefore, even if water reached them, an electrical fault appeared unlikely.
4.1. 5 Steam generator tube integrity is not assumed to be lost during main steam and feedwater transients or pipe breaks.
4.1. 6 The IPEEE (external hazards analysis) assessment neglects (1) the potentialimpacts of relay chatter from relays with unknown capacity (possible optimism, although this is scheduled to be resolved), (2) an improvement if the seismic capacity of EDG tanks is increased, and (3) a detailed review of fire scenarios (not provided in the IPEEE). These are not judged to significantly impact the analysis results.
4.1. 7 High energy lines (HEL) such as the main steam and feedwater lines have automatic (mv) redundant instrumentation for break detection. Low steam generator level willlead ABB Combustion Engineering Nuclear Operations
l ABB Calculation No. A PENG. CALC-019, Rev. 00 Page 13 of 41 to a reactor trip as wiH overcooling events Inst impact RCS conditions. A low steam generator pressure signal willisolate main steam and feedwater lines as weH as trip the main feedwater pumps. A break in the ERY steam supply line win likely cause an initiating event.
4.1. 8 The main steam isolation signal (A1 SIS if either steam generator pressure decreases to 751 psla) isolates main steam and feedwater. It also removes EFAS allowing EAV discharge paths to close. Once the plant protection system (PPS) has deteronined the affected steam generator, EFAS to unaffected steam generator returns. The unaffected steam generator is the one that has a 90 psia higher pressure, if the pressure in both generators increases to > 751 psis then the EnV valves win cycle on steam generator level only.
4.1.9 According to Table 3.2 of Reference 9.1, the unreliability of unaffected backup trains is as foHows:
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 probability of not performing corrective actions based on adequate information in the control room is typicaHy 1.0E 2 (Reference 9.19). Therefore, failure of the operators to isolate a segment is treated as equivalent to one backup train.
4.2 CONSEQUENCE IDENTiflCA TION O
The consequence summary assessment is provided in tabular form in this section. A simplified schematic is provided in Figure 2 to illustrate the boundaries for each of the MSS consequences. Dotted lines are used to identify the boundaries for each consequence.
Afajor A1SS equipment is shown on this figure for esse ofidentification. Table 2 summarizes the consequence evaluation for the A1SS. (Refer to Section S.2 of Reference 9.17.)
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 A1SS. The walkdown captured subtle interactions which could not be readily identified using the screening study. Observations from the walkdown are factored into the consequence evaluation. Table 3A presents the MSS consequences, their corresponding figure numbers and Isometric Drawings. In addition, Table 3B identifies the pipe line numbers and their corresponding locations.
4.3 SHUTDOWN OPERA TION AND EXTERNAL EVENTS Shutdown Overation The consequence evaluation is an assessment assuming the plant is at-power. GeneraHy, 7 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 i inventory control. By satisfying these safety functions, the plant wiH be shut down and ABB Combustion Engineering Nuclear Operations
l r3 ABB Calculation No. A PENG CALC-019, Rev. 00 l
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.Page 14 of 41 l maintained in a stable state. At power, the plant is critical, and is at higher pressure and l temperature 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 contsquence ranking during shutdown would not be more limiting.
Pipe segments that are already ranked as *HIGH' consequence frorn the evaluation at-power need not be evaluated for shutdown. G.ose thot are already 'AfEDIUA4' require some confidence that 'HIGH' would not occur due to shutdown configurations. However, the
' LOW
- consequences for power operation require more confidence that a 'HIGH' would not occur and some confidence that a *AfEDIUM" consequence would not occur. Recognizing this, a review and comparison of system consequence results for power operation versus potential consequence during shutdown operation was conducted.
The results of the comparison indicate that during at-power operation, the A4SS segments are ranked as 'A1EDIUAf' for the large steam lines and
- LOW' for the smauer steam line for the EFW turbine-driven pump. During shutdown operation, the AfSS is not in operation and is not required for decay heat removal. The MSS consequence ranking during at power operation is therefore considered to be bounding.
Extemal Events Although external cvents are not addressed in the current version of the methodology (Reference 9.1), the potentialimportance of piping failures during extemal events is also
{} considered. The ANO 2 IPEEE was reviewed to de! ermine whether external initiating
() events, with their potential common cause impacts on mitigating systems, could affect consequence ranking. This information, along with intsrmation from other external event PRAs, is considered to derive insights and confidence that consequence ranking is not ment significant during an extemal event. The foHowing summarires the review for each of the major hazards (seismic, fire and others).
Seismic ChaHenges 7he potential effects of seismic initiating events on consequence ranking is assessed by considering the frequency of chaHenging plant mitigating systems and the potentialimpact on the existing consequence ranking. The frequency of a seismic event causing a large steam line failure is low and the probability of a seismic event occurring simultaneously with a large steam line failure is also low. The at power ranking for the large steam lines is considered to be bounding for the ranking during a seismic event.
With regard to impact on EFW as a mitigating system (i.e., smauer steam line not causing an initiating event), the most likely scenario is seismic induced loss of offsite power. Based on a typical fragility for loss of offsite power, a High Confidence of Low Probability of Failure (HCLPF) of 0.1g (Reference 9.20) is assurned. Combining this fragility with the seismic hazards developed for the ANO site (References 9.21 and 9.22), the frequency of seismic induced loss of offsite power is less than 1.0E-4 per year.
Considering the scenario where on induced loss of offsite power occurs and both dieselgenerators are available, both trains of EFWare initiaHy available. Howevct, in response to the event, the turbine-driven EFW train is assumed to fail with the motor-driven EFW train providing backup. Once through ecoling would also be available as a additional backup train. Assuming a probability of 1.0E 2 for either p\
\ backup train and an "all year" exposure time, the Conditional Core Damage G
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ABB Calculation No. A PENG CALC 019, Rev, 00 Page IS of 41 g
w Probability (CCDP) for this scenario is approximately 1.0E 6. Thus the resulting consequence is
- LOW'.
For the scenario where on induced loss of offsite power and non seismic failure of one diesel generator occur, only the turbine driven Env train is initially available (assuming that the motor 4 riven Env train is powered by the failed diesel).
However, in response to the event, the turbine 4 riven EFW train is assumed to foH with once through cooling providing backup. Assuming a probabHity of 0.1 for failure of the dieselgenerator and 1.0E 2 for failure of once through cooling and on
'au year" exposure time, the CCDP for this scenario should be less than 1.0E 6.
Thus the resulting consequence is ' LOW'.
For the scenario where on induced loss of offsite power and foHure of both dissel generators occur, only the turbine 4 riven ERV train is initiaHy available. However. in response to the event, the ERV turbine 4 riven train is assumed to fail. Assuming a probability of 1.0E 2 for failure of both dieselgenerators and an "sH year' exposure time, the CCDP for this scenario is approximately 1.0E 6. The resulting consequence is therefore ' LOW'.
Based on the above, the consequence ranking for the smsHer steam lines during a seismic event is enveloped by the al-power consequence ranking.
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. Based on fire core damage frequency of 3.SE 5 per year in the ANO 2 I?EEE, a fire induced loss of offsite power is assumed to be less than 1.0E 2 per year. For this scenario, the ElN system l-likely to be chouenced.
Considering the scenario where an induced loss of offsite power occurs and both diesel generators are available, both trains of EFW are initiaHy available. However, in response to the event, the turbine 4 riven ERV train is assumed to fail with the motor 4 riven ERV train providing backup. Once through cooling would also be available as a additional backup train. Assuming a probability of 1.0E 2 for either backup train and an 'all year
- exposure time, the Conditional Core Damage Probability (CCDP) for this scenario is approximately 1.0E 6. Thus the resulting consequence is ~ LOW'.
For the scenario where an induced loss of offsite power and non seismic failure of one diesel generator occur, only the turbine 4 riven EFW train is initiaHy available (assuming that the motor 4 riven EFW train is powered by the failed diesel). However, in response to the event, the turbine-driven ERV train is assumed to fail with once through cooling providing backup.
Assuming a probability of 0.1 for failure of the diesel generator and 1.0E 2 for failure of once through cooling and an 'aH year
- exposure time, the CCDP for this scenario should be less than 1.0E S. Note that the assumed faHura probability for the diesel generator is very conservative, thus a ' LOW" cansequence is assigned.
For the scenario where an induced loss of offsite power and failure of both diesel generators occur, only the turbine 4 riven ERV train is initiully available. However, in response to the event, the ERV turbine 4 riven train is assumed to fail. Assuming a probability of 1.0E 2 for \
failure of both dieselgenerators and an "all year
- exposure time, the CCDP for This scenario j 1
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l
_._ . _ . __ __ ~ ___. ___ __ _ _ _ _ . _ _ _ _ . _ _ . _ _
ABB O Calculation No. A PENG CALC 019, Rev. 00 Page 16 of 41 is less than 1.0E 4. Since the failure probability for the diesel generators is very l conservative, a
- LOW" consequence is assigned. 1 Since the at power consequence ranking is already ' LOW', the resulting consequences during a fire would not be of greater significance.
Other External Challengen Other horards were screened in the ANO 2 IPEEE and are assumed to have little or no risk significant impact on ERV.
~N (G
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9% EE Calculation NO. A-PENG-CALC-019 Rev. 00 Page 17 cf 41 Table 2 MSS Consequence Assessment Summary JD Descrpeian Svetief Configuraten kutistor isolatiews System Bectser Cantomment Espesmee TeNe Used Rent locetion kvvects Trauw Towe tref. 9 til MS-COTA MS to SG *A
- Cami .,, .^ Cperetmg 75 No PCS & ERY NA MSN eutsnie NA 2-1 MEOMJnt steem due ter T5 MSC018 MS to SG ~B* Contamment Operatmg 75 No PCS & EnY NA MSN cutsafe NA 2-1 MEDKJnt steem due w T5 MS to SG "A* 2155 CVeretmg No PCS & EFW 2 (EFW Possrve bemer NA 23 MEDRJnq MSCO2A 75 steam due to ~8*, AFW, insiste T5. EFW *A* Chee duetopke Thnwght whp-M CD2B MS to SG ~B" 2155 Caerermg T5 No PCS & EFW 2 (EFW Passnre bemer N& 23 MEDA/Kt \
steem due to *A *, AFW, insitie
- 75. EFW ~B* Cnce dO* to P >e Thmught wh&.
MSCD3A EFWsteem "A
- 2155 Stendby 76 No PCS & EFW 2 IEFW, Possrve bemer N'A 23 LOW UHisoleNo steem AFW, hueuide Once Throught MS-CO38 ETWsteem ~8~ 2155 Stendby T6 No PCS & EFW 2 (EFW, Possore bemer N'A 2-3 LOW UnisolaNo steem AFW, inside Osco Throught MSC40A EFWsteem "A
- 2155 Standby 76 2CV- PCS & EFW 2 tETW, 2CV-1000 wf N*A 24 LOW IsolaNe 1000 steem AFW, Pessrve bemer Chee inside M
MSC-048 EFWsteem ~B* 2155 Standby 76 2CV- PCS& ESW 2 (ErW, 2CV-10$O wi N'A 23 LOW isoisNe 1050 steem AfW. Possrve bemer Once kunide Thror W MSCOS EFWsteem 2155,2151, Stendby T51 2CV- PCS & EFW 2 (EFW 2CV-1000 & K'A 23 LOW above B 354 p&echose 1000 steem ~8*, AFW, 1050 wf and Osce fossive bemer 1050 Throught inside ABB Combustion Engin g Nuclea.- Operations
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%) , % )f A F,1 R MWW CalCtdation NO. A-PENG-CALC-019, RN. 00 Page 18 Of 41 Table 2 (Cont'd1 MSS Consequence Assessment Summary a Desenip& Spetw Conhoweten kuriere hosecon System seene Con:neumnt Eweewe Totte Used Reen location k vects Treins Tir e tref. S. t 71 MSC-06 EFWsteem to 2040 Stendby T61 2CV- PCS & EFW 3 (PCS, UrneWected t\flA 2-3 LOW 2CV-0340 1000 steem EFW *B*,
and AFW, 1050 Ohee Treougby '
MSC-07 EFW steem .**om 2040 Demend M iT5 2CV- PCS & EFW > 2 (EFW Uketfacted Betweer~ 2-2 LOW 2CVD34v chenenger 1000 steem *B*, AFW, test and Once 1050or Throomh 2CV- or 0340 isokaan and Once Treought M S C-OR EFW Steam in 2024 Demand M IT5 2CV- PCS & EFW >2 (EFW Uketfected Between 2-2 LOW twtune room chnRengel 1000 creem *B*, AFW, test and Orace 1050 or T?eought 2CV-0340 MS-CD9A M51V Bypass 2155 Standby M (T5 No PCSIT2) & >2 (UW, Passrwe bemer long ACT 2-2 LOW Lirse chenangel partier Asss of AFW, irmaide ow ,,- o,ee thmught MSCD98 MSIV Bypass 2155 Standby M (T5 M PCSIT21 & >2 IEFW, Pessowe bemer Lory ACT 2-2 LOW ur,e cher.nge Pers.1 ss, .1 AFw. arnside EFw , n trwemmht I
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ABB Calculation NO. A PENG CALC 019r Rev,00 Page 19 Of 41 Table 3A Afain Steam System Consequence, figures and Isometric Orswings Consequence 10 figure Isomaltre Drawings Number AfS C-Ol A 2 2 EBB t 1 AfS C-OlB 2 2 EBB 21 AfS C-02A 2 2E8812, 2 EBB-B 1 A4S C-028 2 2 EBB 2 2, REBB 91 AtS C-03A 2 2 EBB 61 AtS C 038 2 2 EBB 71 A1S C 04A 2 2EBC11 AfS C 048 2 2EBC 21 A45 C 05 2 2EBCl.1 A15-C 06 2 2EBCII A4S C-07 2 2EBC. l .2 MS C-08 2 2EBC12 MS C-09A 2 2 EBB 161 A4S C-098 2 2 EBB 111 Table 38 Main Steam (AfSI Piping location & Consequences Pope Desctvton Consequence Loretoon 2 EBB-1 Ms from Steam Generefor 2E 24A to 2CV.101O1 OIA Conternment 2CV 1090 I is in 2 966 02A 2166-A 2 EBB 2 M5 from Steam Generator 2E 248 to 2CV 10602 018 Contamment 2CV-10602 is in 2166 02B 2166-A 2EBU 6 from 2EB3-1 to 2CV 10001 (EfM OJA 2166 A 2CV 1000 I is in 2166 l 2EBD 8 from 2 EBB 1 to 2CV-1001 (A0V) 02A 2166 A l
2 EBB 1 from 2 EBB 2 to 2CV 10602 (EfW) 038 2166 A l 2CV 10$O 2 is in 1966 2 EBB B from 2 EBB-2 to 2CV 106I LA0V) 028 2166 A 2EBC1 from 2CV-1000I to ?P7A 04A, 06 2166 A 2MS 39A is M 2166 05 2161 A 2MS 398 is in 2166 06 ppechose Stop velve 2CV 340 a 2SV 260 in 2040 06,07 2040JJ Turbone & remaining velves k 2024 08 2024JJ 2EBC 2 from 2CV.10602 to 2MS 398 (2EBC 1) 048 2166-A 2MS 398 is in 2166 2 EBB 16 MSIVbwass to 2CV 1040 OSA 2166 A 2 EBB-11 MSIV bwess to 2CV 1090 098 2166 A AM steem durre end safety valve piping and velves are in 2166, Single EfWsteem lone leevne 2166 to supply turtyne dnven purnp in 2024.
O ABB Combustion Engineering Nuclear Operations
s SUS Calctdation No. A4EVG-CALC-019. Rev. 00 OWW Page 20 of 41 to amOSPME*E
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ABB Calculation No. A.PENG.CALCC!9, Rev. 00
[O Page 22 of 41
- 5. 0 DEGRADA TION A4ECHANISMS EVALUA TION The purpose of this section is to identify the degradation mechanisms that can be present, during normal operation, in the piping wittin the selected system boundaries for the ANO 2 AfSS, as described in Section 3.2 of this report. The conditions considered in this evalJetion 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, and Striping (TASCS)
Thermal Transients (TT)
Intergranular Stress Corrosion Cracking flGSCC)
Transgranular Stress Corrosion Cracking (TGSCC)
External Chloride Stress Corrosion Cracking (ECSCC)
Primary Water Stress Corrosion Cracking (PWSCC)
- Localized Corrosion llc)
Aficrobiologically influenced Corrosion (A1/C)
Pitting (PIT)
Crevice Corrosion (CC)
- Flow Sensitive (FS)
Erosion Cavitation (E C)
( Flow Accelerated Corrosion (FAC)
\%) in performing this evaluation, some basic inputs were used. These inputs are discussed in Section 5.3. The 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 one degradation mechanism is potentially present: flow sensitive attack. Two damage groups (DA1 groups) were identified as A1SS F and AfSS N, as defined in Table 4 below. These OM groups result in two failure potential ca:egories: *Large leak *,
and *None".
Table 4 Damage Grmrps Demsee Dameen Mecheriamo fogure Group Thermal fatieue Stress Cononion Crackine Loeekred Corrodon Row Senehke h>.entiel ID TAsCS 17 losCC TGsCC ECSCC PwSCC Mic PIT CC EC FAC Catecery MSS.F No No No No No No No No No No Yes largeLeak MSSN No No No No No No No No No No No None
(^')
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ABB Calculation No. A.PENG CALC 019, Rev. 00 l
Page 23 of 41 l 6.1 DAMAGE GROUPS 5.1.1 DM GROUP: AfSS F I The AfSS F DM group includes lines 2 EBB 138', 2 EBB 136', 2 EBB 2 38' and 2 EBB 2 36* These lines are includedin the ANO 2 FAC program (Reference 9.121.
5.1.2 DA4 GROUP: AfSS N The AfSS N DA1 group is not considered susceptible to any damage mechanism.
6.2 DEGRADA TION MECHANISM CRITERIA AND IDENTIFICATION The degradation mechanisms and criteria assessed are presentedin Table 6.
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ABB Calculation No, A PENG CALC 019, Rev, 00 L/ Page 24 of 41 Table 3 Ikgradation Mechanism Criteria and Susceptible Regions
',C'
, Criteria Susceptible Reglot s l TF TASCS -nps > I inch, and nottles, branch pipe
-pipe segment has a slope < 43*from hortiontal Oncludes elbow or connections, safe ends, tre into a verticalpipe), and welds, heat qfected
-potential existsfor lowpow in a pipe section connected to a tones (HAZ), base component allouing mixong ofhot and coldpuids, or metal, and regions of potential existsfor leakagepow past a valve (i.e., in leakage, out, stress concentra!)on leakage, cross leakage) allowing mixing ofhot and coldpuids, or potential existsfor convection heating in dead-endedpipe sections connected to a source ofhotpuid, or potential existsfor two phase (steam / water) pow, or potential existsfor turbulent penetration in branch pipe connected to headerpiping containing hotpuid with high turbulentpow, and
-calculated or measured AT > 30*F, and
-Richardson number > 4.0 TT -operating temperature > 270*Ffor stainless steel, or operating temperature > 220*Ffor carbon steel, and
, -potentialfor relatively rapid temperw are change: Including coldpuid injection into hot pipe segment, or (G) hotpuid trqjection into coldptpe segment, and AT > 200*Ffor stamless steel, or AT > 130*Ffor carbon steel, or AT > ATallowable (appiscable to both stainless and carbon)
SCC IGSCC -evaluated on accordance wtth existing plant IGSCCprogram per austenitic stainless steel (illVR) NRC Generic Letter 88 01 welds andflAZ IGSCC -operating temperature > 200*F, and (PlVR) -susceptable material (carbon content 2 0 033%), and
-tenssle stress (mcluding residual stress) is present, and
-ongen or oxida:mg species are present OR
-operatmg temperature < 200*F, the attributes above apply, and
-instiatmg contaminants (e g., thiosulfate, fuoride, chloride) are also requored to be present TGSCC -operatmg temperature > 130*F. and austentric stainless steel
-tensile stress (mcludmg residualstress) is present, and base metal, welds, and
-halides (e g., fuoride, chloride) are present, or flAZ caustic (NaOH)is present, and
-omgen or oxidazmg species are present (only required to be present in cortjunction w halides, not required wtaustic) fh
( )
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A &B Calculation No A PENG CALC 019, Rev, 00 Page 25 of 41 Table 3 (cont's0 Degradation Mechanism Criteria and Susceptible Regions Ih',C'",dat g on (,;,,,,, 3,,,,,,;;y, y,,g,,,
SCC ECSCC -operatsng temperature > 130*F, and austentric stainless steel
-tensile stress is present, and base metal, welds, and
-an outside piping surface is withinpre diameters ofa probable llAZ leak path (e.g., valve stems) and is covered with non-metallic insulallon that is not in compliance with Reg. Guide 1.36, or an outside piping surface is exposed to wettingfrom chloride bearing environments (e.g., seawater, brackish water, brine)
PilSCC -ptping materialis inconel(Alloy 600), and nonles, ueids, and flAZ
-exposed to primary water at T > 620*F, and uithout stress relief
-the material is mill-annealed and cold worked, or cold worked and u tided without stress relief LC MIC -operating temperature < 130*F, and pttings, welds, ilAZ,
-low or intermittentpow, and base metal, dsssimilar
-pil < 10, and metal)oints (e.g., utids,
-presence / intrusion oforganic material (e.g., raw water system), or fanges), and regions water source is not treated w4tocides (e.g., refueling water tank) containing crevices PIT -potential existsfor lowpow, and
-oxsgen or oxidissng species are present, and
-intttating contaminants (e.g., fuoride, chloride) are present CC -crevice condisson exists (e g., thermalsleeves), and
-operatong temperature > 130*F, and
- orsgen or oxid sing species are present FS E-C -operating temperature < 250*F, and fstings, welds, ilAZ, and
-pow present > 100 hrs 3r, and base metal
-velocoty > 30ft s, and
-(Pc P.) / AP < $
FAC -evaluated on accordance with extstang plant FACprogram perpan* FA Cprogrnm 5.2.1 Thermal fatigue (TF)
Thermal fatQue is a mechanism caused by alternating stresses due to thermal cycling of a component which resul*s in accumulated fatigue usage and can lead to crack initiation and growth.
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l A1B f) Calculation No. A.PENG CALC 019, Rev. 00 V . Pope 26 of 41 5.2.1.1 Thermal Stratification, Cycling, and Striping (TASCS)
TASCS is not considered to be apolicable to the AfS lines because there is no likelihood of mixing of hot and cold fluids arising from low flow from a connecting pipe segment or leakage flow past a valve. Even varing startup conditions, the rates of temperature change are low such that there should be no stratification in the AfS piping. Startup conditions with no flow past the main steam isolation valve bypass lines (MSIV) were considered, and it was concluded that if any condensation were to collect, it would be in the verticalpiping just upstream of the AISIVs. In addition there is no potential for desd-ended pipe sections connected to a a urce of hot fluid and there is no potential for turbulent penetration of a branch pipe connection.
5.2.1.2 Thermal Transients (TT)
Thermal trans'ents are not considered applicable to the MS lines because there is no potential for rapid temperature changes during plant operation including startups and shutdowns. Per Reference 9.7, the two-inch MSIV bypass valves provide enough steam flow during startup to slowly warm the MS lines thus prevuting any significant throughwall thermal stresses. It is not expected that the thermal response would be significantly differe:nt even if the bypass valves were not open.
5.2.2 Stress Corrosion Cracking (SCC)
^
(~) The electrochemical reaction caused by a corrosive or oxygensted media within a Q piping system can lead to cracking when combinod with other factors such as a susceptible material, temperature, and stress. This mechanism has several forrns with varying attributes including intergranular stress cor,*osion cracking, transgranular stress corrosion cracking, external chloride stress corrosion cr.acking, and primary water stress corrosion cracking.
5.2.2.1/ntergranular Stress Corrosion Cracking (IGSCC)
The MS system cunsists of carbon steel piping. Carbon steel piping is not susceptible to IGSCC.
5.2.2.2 Transgranular Stress Corrosion Cracking (TGSCC)
The MS system consists of carbon steel piping. Carbon steel piping is not susceptible to TGSCC.
5.2.2.3Extemal Chloride Stress Corrosion Cracking (ECSCC)
The MS system consists of carbon steelpiping which is not susceptible to ECSSC.
In addition, ANO-2 complies with the requirernents of Regulatory Guide 1.36 for non-metallic thermalinsulation and consequently the potential for ECSCC to occur does not exist. .
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ABB Calculation No. A PENG CALC 019, Rev. 00 Page 27 of 41 5.2.2.4 Primary Water Stress Corrosion Cracking (PWSCC)
PWSCC is not applicable as a potential damage mechanism for the MS system due to the fact that there is no inconel(Alloy 600) present on 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 licw or
- hideout' regions, such as exists beneath corrosion products or in crevices. This mechanism includes microbiologically influenced corrosion, p'tting, and crevice corrusion.
5.2.3.1 Microbic!ogically influenced Corrosion (MIC)
All of the piping on this system operates ct temperatures greater than the Ibo*F upper temperature threshold for MIC. Therefore, these lines are considered non.
susceptible to MIC.
5.2.3.2 Pitting (PIT)
During normal plant operation, the MS system contains dry saturated steam. As such there is no condensed water in the system. Steam traps and line slope also prevent standing water in the system. During plant outages, a nitrogen blanket is normally established (Reference 9.7). Because of the absence oflow flaw water or moist conditions, the MS system is not susceptible to pitting.
5.2.3.3 Crevice Corrosion (CCI Crevice corrosion is not applicable due to the fact that there are no crevice regions included within the boundaries of the MS system evaluation.
5.2.4 Flow Sensitive (FSI When a high fluid velocity is combined with various other requisite factors it can resu!t 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)
AII 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.
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A1B Calculation No. A PENG CALC-019, Rev. 00
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O Page 28 of 41 5.2.4.2 Flow Accelerated Corrosion (FAC)
The MS system is comprised of carbon steelpiping. FAC is a phenomenon that only affects carbon steelpiping. Per Reference 9.12, lines 2 EBB.I.38' and 2 EBB 2 38' of the MS system are included in the ANO 2 FAC program. These MS lines are c.stegorized by Reference 9.12 as Susceptible Modeled (S-M). These line segments w therefore considered susceptible to FAC damage. All of the remaining line segments of the MS system are not considered to be susceptible to FAC since usage is less than 2% of the plant operating time.
- 6. 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 occurs in the initiation phase and crack propagation proceeds at a 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 en entirely separate program taking guidance from the EPRI Fatigue Management Handbook IReference 9.10). If a vibration problem is discovered, then corrective actions must be taken to
()
G' either remove the vibration source or reduce the vibration levels to ensure future component operability. Frequent system walkdowns, leakage monitoring systems, and current ASME Section XI system leak test requirements are some of the practical measures to address this isset. Because these measures are employed either singly or in combination for most ,nlant systems it is not necessary to use a risk informed inspection selection process for vibration fatigue.
6.3 BAS,'C DA TA 5.3.1 Under normal plant of'etating conditions, the MS system, as defined by the boundaries in Section 3.2, functions as indicated in Table 6.
S.3.2 Oue to the cyclic nature of thermal transients, only those transients which occur i
during the initiatng events Categories I and ll as described in Reference 9.1, Table '
3.1 are considered in the evaluation of degradation mechanisms due to thermal 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 evaluated are those transients which occur under normal of'etating and upset conditions, r
O v -
ABB Cornoustion Engineering Nuclear Operations
k I!
J'1 W W Calculation No. A-PENG-CALC-019. Rev. 00 Page 29 of 41 Table 6 Main Steam System Unes and Operating Conditions Damage Component Component Descnption Material Design Speafk:ation Group Line 19.16] (1) [9.4]
ID No.
Desien Operating Press Temp Press Temp Psig T psig T MSS-F 2 EBB-1-36/38' F.um Steam Generator 2E24 A to Main Steam Isolation Vahr 2CV-1010-1 CS 10E5 5% 185 529 MSS-N 2 EBB-1-10' From Line 2 EBB-1-38'(IIcader I) to Main Steam Safety VrkqQlines) CS 1085 5% 885 529 MSS-F 2 EBB-2-36/38' From Steam Generator 2E248 to Main Steam Isolation Vr.hr ?ry-MYo-1 CS IC35 550 850 520 MSS-N 2 EBB-2-IO~ From Line 2 EBB-2-38'(Header 2) Main Steam Safety Vahes # knes) CS 1085 550 180 520 MS3-N 2 EBB-6-4' From Line 2 EBB-1-38' to E.imiswy Feedwater Pump Ste:w. Supply Vahr 2CV-1000-1 CS 1085 5% 185 529 MSS-N 2 EBB-7-4' From Line 2 EBB-2-38' to E.i, ig..c. Feedmiter Pump Steam Supply Valve 2CV-1050-2 CS 1085 5% 885 529 MSS-N 2 EBB-8-10' From Line 2 EBB-t-38' to Atmospheric Dump Vahr Isolation Vahr 2CV-1002 CS 1085 5% 885 529 MSS-N 2 EBB-9-10' From Line 2 EBB-2-38' to Aliiiesphaic Dump Vahr Isolation Vahr 2CV-1052 CS 1085 5% 885 529 MSS-N 2 EBB-16-2" From Line 2 EBB-1-38' to Main Steam Isolation Vahr B3pass Vahr 2CV-1040-1 CS 1085 5% 885 529 MSS-N 2 EBB-17-2' From Line 2 EBB-2-38' to Main Steam Isolation Vahr B3 pass Vahr 2CV-1090-2 CS 1085 5% 885 529 MSS-N 2EBC-I-4' From Vahr 2CV-1000-1 to EFW Pump Turbine During Trip and Hrottle Vahr CS 1085 5% 885 529 2CV4336 MSS-N 2Ef;C-1-3' From Line 2EBC-I-4' to Flange CS 1985 5% 185 529 MSS-N 2EBC-2-4" From Vahr 2CV-1050-2 to Line 2EBC-1 CS 1085 5% 885 529 Notes: 1. Material (From Reference 9.4) CS = Carbon steel Alllines contain steam traps
=
ABB Combustion Engin ng Nuclear Operations
e ABB Calculation No. A.PENG CALC 019, Rev UO i I
Page 30 of 41
- 6. 0 SERVICE HISTORY AND SUSCEPTIBILfTY REVIEW An exhaustive review was conducted from mid '96 to Spring '97 of databases (plant and industry) and station documents to characterize ANO 2's operating experience with respect to piping pressure boundary degradation. The results of this review are provided in a condensed form in Table 7 for the Main Steam System.
Although several pre-commercial referencer are included for completeness, the timeframe for identifying items applicable to this effcrt was focused on post commercial operation (Commercial Oneration date of March 26,1980). This was done to avoidinclusion ofitems primarily a:sociated with construction deficiencies as opposed to inservice degradation.
The following databases and other sources were queried to accomplish this review:
Station information Management System (SIMS)
The SIMS database was queried for all ANO 2 Job orders on Code Class 1, 2, and 3 components which involved corrective maintenance (CM) or modifications (MOD).
Additionally, a separate query was performed in order to capture certain non Code, O component failures. This query was for non Code 0 and SR (safety related) components. This database contains information from approximately 1985 to the present.
p -
Condition Report (CRI Database i !
The CR database was queried for any pipe leak / rupture events or other conditions associated with identified damage mechanisms *t 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.
Ucensing Research System (LRS)
The LRS database was overied using a keyword search specific to ANO 2. The keywords searched under were: thermal cycling, thermal stratification, thermal fatigue, defect, flow, indication, fatigue, cavitation and courosion. This scorch captured all communication between ANO and the NRC, both plant specific and generic industry, 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 query 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, pipv & leak, vibration & fatigue and pipe & rupture. " Fuzzy
- search logic was employed to reduce the possibility of failing to identify a pertinent document. This database contains information from prior to commercial operation to the present for ANO 2.
4 O
ABB Combustion Engineering Nuclear Operations
ABB Calculation No. A PENG. CALC 019, Rev. 00 Page 31 of 41
- Nuclear Plant Reliability Data System (NPROS)
NPROS 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 ISIProgram Records The ISI program findings were compiled and reviewed for all outage and non outage inservice inspections conducted at ANO 2 since commercial operation.
- ControlRoom Station log The station log was utilized as a source of information for recent operational events.
The 100 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 leakage.
System Upper levelDocument (ULD)
The ULO was reviewed as a source for historical perspective of issues related to the system andidentification 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 This source of information consists of such documents as the SAR, Technical Specifications, operationalprocedures and the damage mechanism analysis done as part of this effort.
O 1
ABB Combustion Engineering Nuclear Operations j l
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,e Calculation No. A PENG CALC 019, Rev. 00
() Page 33 of 41 7.0 RISK EVALUA TION l
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 (i.e.,
consequence segments), and are expo: sed to the same degradation mechanisms (i.e.,
damage groups). The next step b the :ssk evaluation is the determination of the segment risk categories. This is accom,oli@ed by combining the consequence cnd damage mechanism categories to produce a risk category for each segment. Application of the above criteria results in the formation of 26 risk segments of which 4 are high risk (risk category 3) and 22 are low risk (12 are risk category 6 and 10 are risk category 7). The risk segments are identifiedin Table 8 below.
(o v)
APB Combustion Engineering Nuclear Operations
N F%RFEF Calculation NO. A-PENG-CALC-019 Rev. 00 Page 34 Of 41 Table 8 Risk Segment identification N
Risk Segment 10 ConsequenceID Damage Group (D Msk Re.afon Mping Une Nos. Msk Segment Stut l'Oint Msk Segment Emf Pbint Category faHure Pdtential Msk Category Isometric Drawings MS-R-01A MS-C-01A MSS-F High 2 EBB- 1-38' (11 Steem Generator 2E24A (11 Penetration 2P-1 2 EBB- 1-36* MS Norrie Medium large Leek 3 (112 EBB-1-1 MS-R 018 MS-C-018 MSS F High 2 EBB 2-38* til Steam Generator 2E248 (11 Penetration 2P-2 2 EBB 2-36* MS Norrie Medium large Leek 3 (112 EBB 21 MS-R 02A-1 MS-C-02A MSSF High 2 EBB-1-38' (1) Penetration 2P-1 til Wsweem of 2CV-1010L1 2 EBB- 1-36*
Medium large Leek 3 (1) 2 EBB 1-2 Sh.1 MS R 02A-2 MS-C-02A MSS-N Low 2 EBB 8-10" (11 Sweepolet Connection - 111 10" Cap - trom 11 Medium None 6 til 2 EBB-8-1 Sh.1 MS-R-02A 3 MS-C-02A MSS N Low 2 EBB-1-10" (11 Sweepolet C --;;i- - (11 Wstreem of 2PSV-1002 2 EBB- 1-8
- Item 13 Msdi' s.sn None 6 til 2 EBB-1-2 Sh.1 MS-R C2A-4 MS-C-02A MSS-N Low 2 EBB-1-10* (11 Sweapolet Connection - f.') Wstroom of 2PSV-1003 2 EBB-18* Item 14 Medium None 6 til 2 EBB-1-2 Sh.1 MS-R 02A-5 MS-C-02A MSS N Low 2 EBB-1-10* (11 Sweepolet Connection - (1) Wsweem of 2PSV-1004 2 EBB-1-8* It*~ **
Medium None 6 til 2 EBB-t-2 Sh.1 MS-R-02A-6 MS-C-02A MSS N Low 2 EBB-1-10* (11 Sweepo set C-. -e;i- - ill Wsweem of 2PSV-100$
2 EBB-1-8* Item 16 Medium None 6 (112 EBB-1-2 Sh.1 MS-R-02A-7 MS-C-02A MSS-N Low 2 EBB 10* (11 Sweapolet C--,ea;i-, - (11 Wstroom of 2PSV-1006 2 EBB-1-8* Item 17 Medium None 6 (112 EBB-1-2 Sh.1 ABB Combustion Eng ring Nuclear Operations
ry ,-~ im_
ADR 9%FW Calculation No. A-PENG-CALC-019, Rev. 00 Page 35 Of 41 Table 8 Risk Segment identification (Cont'd)
Msk SegmentID Consequence ID Demoge GroupID Msk Region Piping line Nos. Msk Segment Start Pbint Msk Segment EndPbint Category FaRure Pdtential Msk Category Isometrie. Drawings MS-R 0281 MS C-028 MSS F High 2 EBB 2-38' (1) Penetration 2P-2 til Wstroom of 2CV-10602 2 EBB 2-36*
Medium large Leek 3 fil 2 EBB 2-2 Sh.1 l 1
MS-R 02B 2 MS-C-02B MSS-N Low 2 EBB 910* (11 Sweepolet Cor.nection - (1) Wstream of 2CV-1051 l
Medium None 6 1112 EBB 9-1 SIL 1 MSRO283 MS-C-028 MSS N Low 2 EBB 210" (1) Sweepolet Cormection - ill Wstream of 2PSV-1052 2 EBB 2 8* ltem 11 Medium None 6 (112 EBB 2-a Sh.1 MS R 028-4 MS-C-028 MSS N Low 2 EBB 2-10* (11 Sweepolet Carmection - (11 Wstreem of 2PSV-1O53 2 EBB-2 8* Item 12 Medium None 6 1112 EBB 2-2 Sh.1 MS-R O2B 5 MS-C-02B MSS-N Low 2 EBB 2-10* 111 Sweepolet Carmection - (1) Wstream of 2PSV-1054 l 2 EBB-2 8' trem 13 Medium None 6 fil 2ESB 2-2 Sh.1 MS-R 02B 6 MS-C O28 MSS N Low 2 EBB 2- 10' (1) Swe*polet Carmection - (tl Wstream of 2PSV-1055 2 EBB-2 8* Item 14 Medium None 6 (19 2 EBB 2-2 Sh.1 MS-R-02B-7 MS-C-028 M SS-N Low 2 EBB 10" (1) Sweepolet Cormection - 111 Wstreem ot2PSV-1056 2 EBB 2-8* Item 15 Medium None 6 (1) 2 EBB-2-2 Sh.1 MS R-03A MS C-03A MSS N Low 2 EBB 6-4* fil Weldolet Connection - from (1) 4* Cap - trem 4
~ ~
Low None 7 (1) 2 EBB-6-1 Sh.1 MSP 03J MS-C-03B MSS-N Low 2 EBB-1-4 * (11 Weldolet Ca -a-2. -Item (1) Wstreem of 2CV-105O2 10 Low None 7 (112 EBB-7-1 Sh.1 MS-R-04A MS-C-04A MSS-N Low 2EBC- 1-4
- fil Dcwnsweem of 2CV-10C# 111 Wstream ot2MS-39A 1
Low None 7 1112EBC-1-1 Sh.1 ABB Combustion Engineering Nuclear Operations
n .
A R1WEF Calculation NO. A-PENG-CALC-019, Rev. 00 Page 36 Of 41 Table 8 Risk Segenent identification (Cont'd1 Risk SegmentID Conseqc:mce ID Damage Group 10 Msk Region Piping IJne Nos. Msk Segment Start P61nt Esk Segment EndPbint l Category Tsaure P6tential Risk Category isometric Drawings ,
MS R-04B MS-C048 MSS-N low 2EBC-24 * (1) Downsurem of 2CV-1056 ill Wstream of 2MS-398 2
low None 7 til 2EBC2-1 Sh.1 MSR05 MS-C05 MSS-N low 2EBC-1-4* (1) DownsWeem of 2MS-39A til Roor Beverion @ 354* 0*
Low None 7 til 2EBC-1-1 Sh.1 MS R 06 MS C 06 AfSS N Low 2EBC- 1-4 * (1) Roor Beverion @ 354'0* (21 ($pstream of 1% *x 1*
Reducing hert - trem 51 low None 7 (1) 2EBC-1-1 Sh.1 (2) Wsvem of 2CV-03402 (2) 2EBC-1-2 Sh. 2 gy; g,y,, ,g y. m Range - from 100 MS-R 07 MS-C07 MSS-N tow 2EBC- 1 -4 * (2) Downstream of 2CV-0340- til Boot SeelPenetration 2 2024-0011 Low None 7 (112EBC-1-2 Sh.1 (212EBC-1-2 Sh. 2 MS-R-OS MS-C 08 MSS-N tow 2EBC-1-4* (19 Boot SeelPenetretion (1) Wsurem of 2CV-0336 Low None 7 til 2EBC-1-2 Sh.1 MS-R-09A MS-CO9A MSS-N Low 2 EBB- 16-2* II) 2* ConvEng - Stem 20 til Reducing hert - $ tem 14 Low None 7 (1) 2 EBB-16-1 Sh.1 MS-R 098 MS-C09B MSS-N Low 2 EBB 17-2* (112* CoopHng - trwn 22 (1) t)pstream of 2CV-10962 low IVone 7 til 2 EBB-17-1 Sh. I h ABB Combustion Eng ring Nuclear Operations
A kR MIF BB Calculation No. A PENG CALC 019, Rev. 00
.x ) Page 37 of 41 To facilitate application of the sampliny oercentages to determine the inspection scope, ISIS combines like segments (i.e., same consoquence category and damage group) into segment groups. A total of 3 segment groups have been identified and are summarized in Table 9 below.
l Table 9 Risk Inspection Scope Segment Consequence '
Fallwe Risk Risk Total Selections Selections Groups Category Potential Region Category Welds Requked Afede MSS-00 t low None low 7 109 0 0 A1SS-002 Afedium large leak High 3 57 t5 (Note t)
AfSS 003 Medium None low 6 24 0 0 Note 1: AH required element selections win te 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 scope.
Generally, the scope of a planned FAC inspection is determined prior to each scheduled inspection, normally in conjunction with a refueling outage. As such, no element selections are included.
- 8. 0 ELEMENT SELECTION p& The number of elements to be examined as part of the risk-informed 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 individual inspection 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 4), selections are focuscd on terminal ends and other locations (i.e., structural discontinuities) of high stress and/or high fatigue usage. An inspection for cause process shah be implemented utilizing examination methods and volumes defined specificaHy 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, basis for selection) for risk significant segment group MSS-002. As indicated in the Risk Inspection Scope of Table 9, all required e!ement selections wiH be as determined by the existing ANO 2 Flow Accelerated Corrosion program. The examination methods and volumes specified in Table 10 (risk category 3) are defined in Reference 9.1 and are based upon the degradation mechanism (s) postulated to be active at each selected element.
A t s wt ABB Combustion Engineering Nuclear Operations
h Y D%WW Calculation No. A-PENG-CALC-019, Rev. 00 Page 38 of 41 Table 10 Eternent Selection - Risk Category 3 Segment Group Conom FeMure Potential Riek Cetenery Rieh Region Total R of etenente 26% of elemente MSS-002 Mediurn large Leek 3 High 57 15 Demente Selected line No. Eram Method Riek Segment D l Deecrfp6on leo Dwg No. Exam Vohene Ca.- ,- /DM Group D's Reneen for Selecoon AR required element 2 EBB-1-38' VoAnnetric MS-R-01A A primary objective of FAC enetysis is to identify corrnponents that are l selections wiR be es most susceptble to FA C demoge k carbon-steelpking, mchadVng both I d:termined by the existing 2 EBB- 1-36'.
2 EBB-2-38 Rgure No. 7.7-1 MS-C-01A / MSS F single and two phase lig% snergy systems. The choice of kspection ANO-2 Row Accelerered 2 EBB 2-36 Rgwe W. 7.7-2 ge,,,,,, g,, ,,,,;yy, ,,,,,,, ,,g ,, g,,, ,,,,;g,,,,;,, ,,,,;,,,
Corrosion progrem. The Rg w e M . 7. 7 3 Inspection resofts, industry esponence, EPRI OfECWORKS pred;ctive 2 EBB 1 Rgure No.1.7-4 k S R-018 FAC program however, cormuter model renking and erg!.;...i,,, judgment, each of which is does not lend itsett to tfn 2 EBB-1-2 Sh. I figure No.1.7 5 MS C-01B / MSS-F describedin greater deter bekw.
esteb5stwnent of a kng- 2 EBB 1 Rgure No. 7.7 6 plan with a 2 EBB 2-2 Sh.1 Rgure No. 7.7-7 WL Rde - kspection rds of mR eveR&
term MSR 02A-1 predetermined inspection ,,,,;,,, y,,,,,,;,,, ,,, ,,,;,,,4,,;,, ,, ,,,y ,,g,gu;,4 g,,,,,,;,,_
scope. Gerwretty, the MS C-02A / MSS-F The prodcred rememong Efe, cotculated by weer rates of the scope of a planned .'4C prov ously kspected iw,w-;ts, is uto? ired to identify future inspection is determineo kspection locations.
pikr to each schesksled MS-R 028-1 inspection, normeMy in MS-C 02B / MSS-F
- "'" '# W """ ""
coriunction with a e vekseble amplement to plant anstysis and essociated inroections.
refueEng outage. As such, Some sources of hformation kchsde the CHUG hot Sne, Entergy FAC no element selections are peer grow, OfUG-smported Prent Events Detebese end the ANO kcluded. kdkastry Events Analysis grow. At significant kdustry experiences reinthg to FAC are factored klo the program. ARuw.e.;ts ther l
rotate to en kdustry event wm genereRy be inspected et she next outage.
CNECWoRKS Renake - The EPRI OfECWORKS predctive computer model is used for guidance k deterrrmamg specific heations to be enemmed Within the OfECWORKS corrputer model a weer rate calculation for each modeled Kne is performed. A semple of the fighest rar=*ed .- w e is then chosen based on a refetive renking.
Engineermy _". ', ..a.: - AppNestion of good enginoormg judgmoret is en irrportant consideration in the selection of .w,+.e. ts for inspection. Awareness of ren-twical system operation es weR es industry experience are factoredinto the connponent selection. Also, feedback from system a,y-.e .i,g, plant operations, maintenance end the secondary chemistry departments are taken into consideratiott
__ _ . _ _ _ _ __s h ABB Combustion Enghring Nuclear Operations
A Ik R MIPIr 7s Calculation No. A PENG CALC 019, Rev. 00
& 9. 0 REFERENCES Page 39 of 41 9.1 ' Risk Informed Inservice Inspection Evaluation Procedure," EPRI Report No. TR.
106706, Interim Report, June 1996.
9.2 EPRIInservice Inspection Software (ISIS'I,1996.
9.3 Arkansas Nuclear One Unit 2, " Safety Analysis Report," Amendment No.13.
9.4 "Desip Soecification for ASME Section lll Nuclear Piping for Arkansas Nuclear One Unit 2, Arkansas Power and Ught Company,' Specification No. 6600-M-2200, Revision 9.
9.5 "ANO-2 SIMS Components Database,"(Plant Piping Line Ust (M 2083), dated 3 31-96).
- 9. 7 Arkansas Nuclear One, Unit 2 System Training Manual, " Steam Generator Main Steam System," STM215, Rev. 3, February 7,1996, 9.8 " Technical Specification for Insulation for Arkansas Nuclear One Unit 2 of the 73 Arkansas Power and Light Company,' Specification No. 6600-M-2136, Revision 9.
V 9.9 ' Primary Chemistry Monitoring Program," Procedure No. 1000.106, Revision 4 9.10 'EPRI Fatigue Management Handbook," Report No. TR-104534-V1, V2,-V3.-V4, Project 332101, Final Report, December 1994.
9.11 " Pipe Cracking in PWRs with Low Pressure Borated-Water Systems," EPRI Report No. NP-3320.
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 Intemal Flood Screening Study," prepared for Entergy Operations, Inc. Calculation No. 89 E 0048-35, Rev. O, May 1992.
9.15 " Arkansas Nuclear One Unit 2 Probabilistic Risk Assessment, Individual Plant Examination Submittal," 94-R-2005-01, Rev. O, August 1992.
ABB Combustion Engineering Nuclear Operations
A R Ik MINIP Calculation No. A PENG-CALC-019. Rev. 00 Page 40 of 41 i 9.16 Entergy, Arkansas Nuclear One Unit 2, Isometric Drawings:
1.0 Drawing No, M-2206, Sheet la Rev.122; " Piping & Instrument Diagram Steam Generator Secondary System.'
2.0 Drawin; No. M 2202, Sheet 4, Rev.16; " Piping & Instrument Diagram Lube Oil, Lube Oil Cooling, Electro / Hydraulic Controls & Main Steam."
3.0 Drawing No. 2 EBB-1 1, Rev. 20, "Large Pipe isometric Main Steam from Steam Generator 2E 24A to Containneent Penetration."
- 4. 0 Drawing No. 2 EBB-12, Sheet 1, Rev.10, "Large Pipe Isometric Main Steam Header from Penetration 2P 1 to MSIV 2CV 1010-1."
- 5. 0. Drawing No. 2 EBB-21, Rev.14, "Large Pipe Isometric Main Steam Header from Steam Generator 2E-248 to Containment Penetration 2P-2.'
6.0 Drawing No. 2 EBB-2 2, Sheet 1, Rev 9, 'Large Pipe isometric Main Steam Header #2 from Penetration 2P-2 to MSIV 2CV 1060-2 ~
- 7. 0 Drawing No. 2 EBB-6-1, Sheet 1, Rev.12, "Large Pipe Isometric Main Steam Supply to 2CV 1000-1."
- 8. 0 Drawing No. 2 EBB-7-1, Sheet 1, Rev. II, "Large Pipe isometric Emergency Feedwater Pump Steam Supply to 2CV 10CO-2."
9.0 Drawing No. 2 EBB-B 1, Sheet 1, Rev. 9, "Large Pipe Isometric Main Steam Dump to Atmosphere through 2CV-1001."
10.0 Drawing No. 2 EBB-9-1, Sheet 1, Rev 9, "Large Pipe isometric Main Steam Dump to Atmosphere through 2CV-1051.'
11.0 Drawing No. 2 EBB-161, Sheet 1, Rev. 13, 'Small Pipe isometric Main Steam Bypass. ,or 2CV-1010-1."
12.0 Drawing No. 2 EBB-171, Sheet 1, Rev. 8, 'Small Pipe Isometric Main Steam Intercept Valve 2CV 1060-2 Bypass."
13.0 Drawing No. 2EBC-1-1, Sheet 1, Rev. 24, "Large Pipe Isometric Main Steam Supply to Emergency Feedwater Pump Turbine Driver 2K 3 from Main Steam Header #1."
14.0 Drawing No. 2EBC 12, Sheet 1, Rev. 33, "Large Pipe isometric Main Steam Supply to Emergency Feedwater Pump Turbine Driver from Main Steam Header # 1. "
15.0 Drawing No. 2EBC-1-2, Sheet 2, Rev. 2, "Large Pipe isometric Main Steam Supply to Emergency Feedwater Pump Turbine Driver from Main Steam Header #1."
16.0 Drawing No. 2EBC-21, Sheet 1, Rev.13, "Large Pipe isometric Main Steam Header #2 Supply to Emergency Feedwater Pump Turbine Driver 2K-3."
9.17 " Consequence Evaluation of ANO-2 EFW, Containment Spray, and Main Steam and Feedwater System Piping " Arkansas Nuclear 1 Unit 2, Yankee Nuclear Services Division Calculation No. NSD-018, Rev. O, August 1997, 9.18 "ANO-2 Steam Generator Pressure Drop Due to an EFW Steam Supply Une Break' ANO-2 Calculation No. 96-E-0022-01, April 1996.
9.19 Swain, A. D. and Guttmann, H. E.; " Handbook of Human Reliability Analysis with Emphasis on Nuclear Power Plant Operations", NUREG-CR-1278, August 1983.
O ABB Combustion Engineering Nuclear Operations
A Ik Ik ,
d'% IFIF l Calculation No. A PENG CALC-019, Rev. 00 Q Page 41 of 41 9.20 North Atlantic Energy Services Corp. " Individual Plant Examination External Events", l Report for Seabrook Station, Response to Generic Letter 88 20, Supplement 4, September 1992.
I 9.21 "Probabilistic Seismic Hazard Evaluations at Nuclear Plant Sites in Central and Eastern United States: Resolution of the Charleston Earthquake issue", EPRI NP-6395-D, April 1989, Prepared by Risk Engineering, Inc., Yankee Atomic Electric Company, and Woodwar; Clyde Consultants.
l 9.22 " Revised Uvermore Seismic Hazard Estimates for 69 Nuclear Power Plant Sites East of the Rocky Mountains", NUREG 1488 (Final Report), April 1994.
9.23 Interoffice Correspondence from A. V. Bauer to Quality Records, Letter No. PENG-97-140, " Submittal of SIA Calculations," dated July 21,1997.
O) u ABB Combustion Engineering Nuclear Operations
i Calculation No. A PENG CALC 019, Rev, 00
- Page A t of A15 l i.
O
)
APPENDIX A
'FMECA CONSEQUENCEINFORMATION REPORT" (Attachment Pages A1 A15)
O ABB Combustion Engineering Nuclear Operatior,e-
FMECA - Consequence Inforrnation Report Calculatkm No. A PENG CALC-019, Rev. 00 nssn 91 Page A2 of A13 Consequence ID: MS-C-01 A Consequence
Description:
Degradation of main steam flow from steam generator 2E 24 A inside containment during normal operation (line 2 EBB 1 inside containment).
Break Size: Large Isolability of Break: No ISO Comments: Blowdown from faulted steam generator is not isolable. Steam line break detection (low steam generstor pressure) will isolate MSIVs and feedwater to the faulted steam generator. Also, EFW flow to the faulted steam generator will be isolated and remain isolated via differential pressure between faulted steam generator and good steam generator.
Spatial Effects: Containment Effected location: Containment Building Spatial Effects Comments: Steam line breaks are within the design basis and the n cessary safety components located inside containment are qualified for such ennts.
Initiating Event: I Initiating Event ID: T5 Initiating Event Recovery: No recovery from an unisolable steam line break. This will result in an immediate plant trip.
Loss of System: SDM 2 System IPE ID: PCS, EFW System Recovery: MSIV isolation, feedwater isolation, and feedwater pump trip 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. However, there is a discharge path from each EFW pump to the unfaulted steam generator. Also,1 of 2 steam supply paths to the turbine EFW pump is unavailable, but the other steam supply is more reliable than the turbine EFW pump itself. Also, the auxiliary feedwater pump could be used. Loss of the steam dump capability from the faulted ste .:n ;-nerator is neglected since the event essentially provides successful depressurization.
Loss of Train: N Train ID: MA Train Recovery: N/A Consequence Comment: Consequence is " Medium" based on Table 2 lof Ref. 9.17. Containment isolation is unaffected.
Consequence Category: MEDIUM C Consequence Rank O I
O
FMECA - Consequence Information Report Calculation Na A PENG-C4LC .49. Rev. 00 16-ser-91 Page A3 of A!3 Consequence ID: MS C-01B Consegmence
Description:
Degradation of main steam flow from steam generator 2E 24B inside containment j during normal operation (line 2 EBB 2 inside containment).
Break Size: Large Isolability of Break: No ISO Comments: Blowdown from faulted steam generator is not isolable. Steam line break detection (low steam generator pressure) will isolate MSIVs and feedwater to the faulted steam generator. Also, EPN
- flow to the faulted steam generator will be isolated and remain isolated sia differential pressure
- between faulted steam generator and good steam generator.
4 5
Spatial Effects: Containment Effected I4 cation: Containment Building Spatial Effects Comments: Steam line breaks are within the design basis and the necessary safety components located inside contaimnent are qualified for such events.
Initiating Event: 1 Initiating Event ID: T5
) Initiating Event Recovery: No recovery from an unisolable steam line break. This will result in an immediate
! plant trip.
Loss of System: SDM 2 System IN ID: PCS, EFW System Recovery: MSIV isolation, feedwater isolation, and feedwater pump trip 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. However, there is a discharge path from each EFW pump to the unfaulted steam generator. Also,1 of 2
, - steam supply paths to the turbine EFW pump is unavailable, but the other steam supply is more reliable than the turbine EFW pump itself. Also, the auxiliary feedwater pump could be l used. Loss of the steam dump capability from the faulted steam generator is neglected since the event essentially provides successful depressurization.
Loss of Train: N Train ID: N/A Train Recovery: N/A Consequence Comment: Consequence is " Medium" based on Table 2-lof Ref. 9.17. Containment isolation is unaffected.
Consequence Category: MEDIUM C Consequence Rank O
- O
FMECA - Consequence Information Report Cakulanon No. A PENGCALC 019. h. 00 l4-sep-91 Page A4 of A13 Consequence ID: MS-C-02A Consequence
Description:
Degradation of main steam flow from steam generator 2E 24A outside containment during normal operation (line 2 EBB 1 between penetration 2P 1 and 2CV1010-1, and connected lines >4 inch diameter such as 2 EBB-8).
Break Size: Large Isolability of Break: No ISO Comments: Blowd;wn from faulted steam generator is not isolable. Steam line break detection (low steam generator pressure) will isolate MSIVs and feedwater to the faulted steam generator. Also, EFW flow to the faulted steam generator will be isolated and remain isolated via differential pressure between faulted steam generator and good steam generator.
Spatial Effects: Propagation Effected Location: Room 2155 Spatial Effects Comments: Propagation is through access doors and room siding to outside (roof of adjoining buildings), down through main steam line chase toward the turbine building, and through a stairway door into the turbine auxiliary building. For a severe main steam line break it is possible to damage and/or push the siding out into the adjacent fuel handling area (Room 2151). However, this is a large area and safety equipment is located at lower elevations.
Initiating Event: I Initiating Event ID: T5 Initiating Event Recovery: No recovery from an unisolable steam hne break. This will result in an immediate plant trip.
Loss of System: SDM 2 System IPE ID: PCS, EFW System Recovery: MSIV isolation, feedwater isolation, and feedwater pump trip occur on low steam generator pressure. It is possible to recover a condensate pump and snakeup to the unfruited steam generator. EFW discharge to the faulted steam generator is isolated and unavailable. However, there is a discharge path from each EFW pump to the unfaulted steam generator. Also, I of 2 steam supply paths to the turbine EFW pump is unavailable, but the other steam supply is more reliable than the turbine EFW pump itself. Also, the auxiliary feedwater pump could be used. Loss of the steam dump capability from the faulted steam generator is neglected since the event essentially provides successful depressurization.
less of Train: T Train ID: EFW "A" Train Recovery: it is possible for large main steam line breaks to affect both EFW turbine steam supply lines which are in close proximity and smaller piping. s Consequence Comment: Consequence is " Medium" based on Tables 2-1 and 2 3 of Ref. 9.17 (2 backup trains, including motor EFW, AFW, and once through cooling). Passive closed barrier provided by the steam generator is crulited for containment isolation.
Consequence Category: MEDIUM O Conseqtence Rank O O
p FMECA - Consequence Information Report 14-ser 91 Calculaten No. A PENG-CEC 019, b. 00 Page AS of A13 Consequence ID: MS-C-02B Consequence
Description:
Degradation of main steam flow from steam generator 2E 24B outside containment during normal operation (line 2 EBB 2 between penetration 2P 2 and 2CV1060 2, and connected lines >4 inch diameter such as 2 EBB 9).
Break Size: Large Isolability of Break: No ISO Comments: Blowdown from faulted steam generator is not isolabic. Steam line break detection (low t; team generator pressure) will isolate MSIVs and feedwater to the faulted steam generator. Also, EFW flow to the faulted steam generator will be isolated and remain isolated via differential pressure between faulted steam generator and good steam generator.
Spatial Effects: Propagation Effected Location: Room 2155 Spatial Effects Comments: Propagation is through access doors and room siding to outside (roof of adjoining buildings), don through main steam line chase toward the turbine building, and through a stairway door into the turbine auxiliary building. For a severe main steam line break it is possible to damage and/or push the siding out into the adjacent fuel i
' handling area (Room 2151). However, this is a large area and safety equipment is located at lower elevations, i
Initiating Event: I laitiating Event ID: T5 Initiating Event Recovery: No recovery from an unisolable steam line break. This will result in an immediate less of System: SDM 2 System IPE ID: PCS, EFW 3' System Recovery: MSIV isolation, feedwater isolation, and feedwater pump trip 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. However, there is a discharge path from each EFW pump to the unfaulted steam generator. Also, I of 2 steam supply paths to the turbine EFW pump is unavailable, but'the other steam supply is more reliable than the turbine EFW pump itself. Also, the auxiliary feedwitter pump could be used. Loss of the steam dump capability from the faulted steam generator is neglected since the event essentially provides successful depressurization.
Loss of Train: T Train ID: EFW "A" Train Recoveg: It is possible for large main steam line breaks to affect both EFW turbine steam supply lines -
which are in close proximity and smaller piping.
Consequence Comment: Consequence is " Medium" based on Tables 2 1 and 2-3 of Ref 9.17 (2 backup trains, including motor EFW, AFW, and once through cooling). Passive closed barrier provided by the steam generator is credital for containment isolation.
Consequence Category: MEDIUM O Consequence aank O r
i t
FMECA - Consequence Information Report Cokulation Na A PENGMC-019, Rev. 00 14-ser-91 Page A6 of A13 Consequence ID: MS-C-03A Consequence
Description:
Degradation of EFW steam supply from steam generator 2E-24 A during normal operation (line 2 EBB-6 from main steam line 2 EBB-1 to 2CW10001).
Break Size: Large Isolability of Break: Yes ISO Comments: Blowdown from faulted steam generator is not isolable. This line is not luge enough to cause an automatic plant trip or MSIV isolation.
Spatial Effee;s: Propagation Effc*4 Location: Room 2155 Spatial Effects Comments: Propagation is through access doors and rmm siding to outside (roof of adjoining buildings), down through main steam line i.hase toward the turbine building, and through a stairway door into the turbine auxiliary building.
Initiating Event: I Initiating Event ID: T6 Initiating Esent Recovery: An automatic plant trip is assumed. Otherwise, detection is expected from a high power alarm ud a controlled shutdown is assumed due to standing orders to do so as a result of any unknown or uncontrolled loss of 50 Mwt.
Loss of System: N System IPE ID: N/A System Recovery: N/A Imss of Train: TDM 2 Train ID: PCS, Turbine EFW Train Recovery: The affected steam generator would be isolated once identified during the controlled shutdown. I of 2 steam supply paths to the turbine EFW pump is unavailable, but the other steam supply path is more reliable than the turbine EFW pump itself.
Consequence Comment: Consequence is ' Low" based on Tables 2 1 and 2 3 of Ref. 9.17 (3 backip trains available, including PCS, EFW, AFW, and once through cooling). Passive closed barrier prmided by the steam generator is credited for containment isolation.
Consequence Category: Low 0 Consequence Rank O O
, FMECA - Consequence Information Report Calculation No A PENG<ALC 019, Rev. 00
- \ I4-s*V91 Page A7 of A13 Consequence ID
- MS-C-03B l
Consequence
Description:
Degradation of EFW steam supply from steam generator 2E 24B during normal operation (line 2 EBB 7 from main steam line 2 EBB 2 to 2CV 1050-2).
Break Size: Large Isolability of Break: Yes ISO Comments: Blowdown f.om faulted steam generator is not isolable. This line is not large enough to cause an automatic plant trip or MSIV isolation.
Spatial Effects: Propagation Effected 14 cation: Room 2155 Spatial Effects Comments: Propagation is through access doors and room siding to outside (roof of adjoining buildings), down through main steam line chase toward the turbine bul' ding, and through a stairway door into the ttubine auxiliary building.
Initiating Event: 1 Initiating Event ID: T6 initiating Event Recovery: An automatic plant trip is assumed. Otherwise, detection is expected from a high power alarm and a controlled shutdown is assumed due to standing orders to do so as a result of any unknown or uncontrolled loss of 50 Mwe.
Loss of System: N System IPE ID: N/A System Recovery: N/A Less of Train: TDM 2 Train ID: PCS, Turbine EFW Train Recovery: The affected steam generator would be isolated once identified during the controlled shutdown. I of 2 steam supply paths to the turbine EFW pump is unavailable, but the other steam supply path'is more reliable than the turbine EFW pump itself.
Consequence Comment: Consequence is " Low" based on Tables 2 1 and 2-3 of Ref. 9.17 (3 backup trains available, including PCS, EFW, AFW, and once through cooling). Passist closed barrier provided by the steam generator is credited for containment isolation.
Consequence Category: Low 0 Consequence Rank O e
V
FMECA - Consequence Information Report Calculatum & A FMG CALC-OlP, Rev 00 14-ssy97 Page A8 of A13 Consequence ID: MS-C-04A Consequence
Description:
Isolable degradation of EFW steam supply from steam generator 2E 24 A during normal operation (line 2EBC 1 between 2CW1000-1 and check valve 2MS 39A).
Break Size: Large Isolability of Break: Yes ISO Comments: 2CV 1000-1 can be closed by the operators to isolate this break and check valve 2MS 39A prevents backflow from line "B." This line is not large enough to cause an automatic plant trip or MSIV isolation and there is no automatic isolation of 2C%l000-1.
Spatial Effects: Propagation Effected location: Room 2155 Spatial Effects Comments: Propagation is through access doors and room siding to outside (roof of adjoining buildings), down through main steam line chase toward the turbine building, and through a stairway door into the turbine auxiliary building.
Initiating Event: 1 Initiating Event ID: T6 Initiating Event Recovery: An automatic plant trip is assumed. Otherwise, detcction is expected from a high power alarm and a controlled shutdown is assumed due to uanding orders to do so as a result of any unknown or uncontrolled loss of 50 Mwe.
Loss of System: N System IPE ID: N/A 1
l System Recovery: N/A Loss of Train: TD Train ID: Turbine EFW Train Recovery: 1 of 2 steam supply paths to the turbine EFW pump is unavailable, but the other steam supply path is more reliable than the turbine EFW pump itself.
Consequence Comment: Consequence is " Low" based on Tables 2-1 and 2-3 of Ref. 9.17 (3 backmp trains available, including PCS, EFW, AFW, and once through cooling). 2C%1000-1 provides containment isolation.
Consequence Category: Low 0 Consequence aank O l
l l
O
_. _ _ _. ___ _ _ _ _-__ _ _ . . _ . _-_ __ . . - m_-.
1 FMECA - Consequence Information Report Cale='a'""' Na AN C4LC-018 Ra M i4 s.r 97 pare AP of A15 Consequence ID: MS-C-04B
, Consequence
Description:
Isolable degradation of EFW steam supply from steam generator 2E 24B during normal operation (line 2EBC 2 between 2CV 1050 2 and check valve 2MS 39B).
Break Size: Large Isolability of Break: Yes 150 Comments: 2CV 1050 2 can be closed by the operators to isolate this break and check valve 2MS 39B prevents backflow from line 'A." This line is not large enough to cause an automatic plant trip or MSIV isolation and there is no automatic isolation of 2CV 1050-2.
Spatial Effects: Propagation Effected location: Room 2155 Spatial Effects Comments: Propagation is through access doors and room siding to outside (roof of adjoining buildings), down through main steam line chase toward the turbine building, and through a 4 airway door into the tuttine auxiliary building.
Initiating Event: I laitiating Event ID: T6 Initiating Event Recovery: An automatic plant trip is assumed. Otherwise, detection is expected from a high power alarm and a controlled shutdown is assumed due to standing orders to do so as a result of any unknown or uncontrolled loss of 50 Mwe.
less of System: N System IPE ID: N/A System Recovery: N/A less of Train: TD Train ID: Turbine EFW Train Recovery: 1 of 2 steam supply paths to the turbine EFW pump is usuwailable, but the other steam supply path is trore reliable than the turbine EFW pump itself.
Consequence Comment: Consequence is " Low" based on Tables 21 and 2 3 of Ref 9.17 (3 backup trains available, including PCS, EFW, AFW, and once through cooling). 2CV 1050 2 provides containment isolation.
Consequence Category: Low 0 Consequence Rank O 1
FMECA - Consequence Information Report Calculanon Na A PENG CEC-019. Rev 00 is Sm 01 Page A10 of AIS Consequence ID: MS-C-05 Consequence
Description:
Degradation of EFW steam supply from both steam generators dunng normal operation (line 2EBC-1 from check valves 2MS-39A & B to El 354).
Break Size: Large Isolability of Break: Yes ISO Comments: 2CV.10001 and 2CV.1050-2 can be closed by the operators to isolate this break. This line is not large enough to cause an automatic plant trip or MSIV isolation and there is no automatic isolation of 2CV 10001 and 2CV 1050 2.
Spatial Effects: Propagation Effected Imcation: Room 2155 Spatial Effects Comments: Piping is located in Room 2155,2151, and vertical pipe chase to El 354.
Propagation from Room 2155 is through access doors and room siding to outside (roof of adjoining buildings), down through main steam line chase toward the l turbine building, and through a stairway door into the turbine auxiliary building.
l Propagation from Room 2151 is into a very large area and at an elevation away from safe shutdown equipment. It is assumed that due to the torturous pathway that breaks in the pipe chase do not easily propagate into Room 2040 at El 335.
Initiating Event: I Initiating Event ID: T6 Initiating Event Recovery: An automatic plant trip is assumed. Otherwise, detection is expected from a high power alarm and a controlled shutdown can be assumed due to standing orders to do l
so as a result of any unknown or uncontrolled loss of 50 Mwe.
Loss of System: N System IPE ID: N/A System Recosery: N/A Loss of Train: T Train ID: Turbine EFW Train Recovery: Steam supply to the turbine EFW pump is unavailable.
Consequence Comment: Consequence is " Low" based on Tables 2 1 and 2 3 of Ref. 9.17 (3 backup trains
- available, including PCS, EFW B, AFW, and once through cooling). Containment I isolation is unaffected.
Consequence Category: low D Consequence nank O l
l l
l 9
( ,
FMECA - Consequence Information Report 14-Sep 97 Catalanon Na A PENG CUf Olp Rev. 00 Page All of A13 Consequence ID: MS.C-06 Consequence
Description:
Degradation of EFW steam supply in Room 2040 during normal operation (line 2EBC 1 from El 354 to normally closed 2CV 0340).
Break Size: Large Isolability of Break: Yes ISO Comments: 2CV.10001 and 2CV 1050 2 can be closed by the operators to isolate this break. This line is not large enough to cause an automatic plant trip or MSIV isolation and there is no automatic isolation of 2CV 10001 and 2CV 1050 2.
Spatial Effects: Propagation Effected Location: Room 2040 Spatial Effects Comments: The fuel pool cooling and purification pumps are in the vicinity of this piping r.nd are likely impacted (judged to be low consequence due to significant time for recovery and makeup). The most important component, MCC 2B52, is at the other end of the room. It is assumed to fail for the case where operators fail to detect and initiate a controlled shutdown.
Initiating Event: I laitiating Event ID: T6 Initiating Event Recovery: An automatic plant trip is assumed. Otherwise, detection is expected from a high power alarm and a controlled shutdown can be assumed due to standing orders to do so as a result of any unknow or uncontrolled loss of 50 Mwe.
Loss of System: N System IPE ID: N/A System Recovery: N/A Loss of Train: TM-4 Train ID: Turbine EFW, HPSI A, LPSI, CSS A Train Recovery: Steam supply to the turbine EFW pump is unavailable. For failure to detect and isolate case, it is assumed MCC 2B52 is a!Tected which contains breakers for key valves. These valves include normally closed valves in containment spray train A, HPSI train A cold leg supplies, and LPSI.
Consequence Comment: Consequence is " Low" based on Tables 2 1 and 2-3 of Ref. 9.17 (3 backup trains available, including PCS, EFW B, AFW, and once through cooling). For the failure to isolate case, the consequence is still "lew" (isolation, EFW B and once through cooling). Conte.inment isolation is unaffected.
Consequence Category: low D Consequence Rank O O
FMECA - Consequence Information Report Cakulahm Na A-PENG-CALC-Olg Rev 00 1&ser 91 Page Al2 g! A13 Consequence ID: MS-C-07 Consequence
Description:
Degradation of EFW steam supply on demand in Room 2040 (line 2EBC 1 from normally closed 2CV 0340 to Room 2024).
Break Size: Large Isolability of Break: Yes ISO Comments: 2CV 1000-1 and 2CV 1050 2 can be closed by the operators to isolate this break. Normally closed 2CV-0340 2 can also be reclosed. Since this line is normally in standby and 2CV 0340 2 opens on an EFW demand, pipe break is assumed to occur during a test or accident demand.
Spatial Effects: Propagation Effected Iecation: Room 2040 Spatial Effects Comments: The fuel pool cooling and purification pumps are in the vicinity of this piping and are likei fimpacted (judged to be low consequence due to significant time for recovery and makeup). The most important component, MCC 2B52, is at the other end of the room. It is assumed to fail for the case where operators fail to detect and isolate.
Initiating Event: N Initiating Event ID: N/A Initiating Event Recovery: Pipe break during standby isjudged less likely than during a pump test demand. An independent loss of PCS (T2) is assumed to challenge piping.
Loss of System: S System IPE ID: PCS System Recover:/: PCS loss is assumed to be the initiator, Less of Train: TDM-4 Train ID: Turbine EFW, HPSI A, LPSI, CSS A Train Recovery: Steam supply to the turbine EFW pump is unavailable. For the failure to detect and isolate case, it is assumed MCC 2B52 is affected which contains breakers for key vahta. These valves include normally closed containment spray train A, HPSI hot leg supplies, and LPSI.
Consequence Comment: Consequence is " Low" based on Table 2 2 of Ref. 9.17 ( Anticipated challenge, between test exposure, and 2.5 backup trains available, including motor EFW, AFW, and once through cooling). For the failure to isolate case (loss of MCC 2B52), the consequence is also " Low" because there are still 2.5 trains with a train of once through cooling, motor EFW, and isolation failure. Containment isolation is unaffected.
Consequence Category: low O Consequence Rank O O
FMECA - Consequence Information Report Cain,ioum Na A,rmm-olum oo
(
\ l4-Ser-91 rage A13 of AIS Consequence ID: MS-C48 Consequence
Description:
Degradation of EFW steam supply on demand in Room 2024 (line 2EBC 1 in
- Room 2024).
Break Size: Large Isolability of Break: Yes ISO Comments: 2CV 10001 and 2CV-1050 2 can be closed by the operators to isolate this break. Normally closed 2CV 0340 2 can also be reclosed. Since this line is normally in standby and 2CV 0340-2
- opens on an EFW demand, pipe break is assumed to occur during a test or accident demand.
. Spatial Effects: Propagation Effected Location: Room 2024 Spatial Effects Comments: The turbine EFW pump room is watertight (except for floor drain) with heasy doors that open into the room. Steam leakage through penetrations (to Room 2040) and the floor drain (to auxiliary building sump at El 317) is expected, but this is not judged significant enough to impact safety equipment.
Initiating Event: N laitiating Event ID: N/A
, Initiating Event Recovery: Pipe break during standby is judged less likely than during a pump test demand. An
+
independent loss of PCS (T2) is assumed to challenge piping.
, Loss of System: S System IPE ID: PCS System Recovery: PCS loss is assumed to be the initiator, Loss of Train: T Train ID: Tu:bine EFW Train Recovery: Steam supply to the turbine EFW pump is unavailable.
Consequence Comment: Consequence is " Low" based on Table 2 7 of Ref. 9.17 (Anticipated challenge, between test exposure, and 2.5 backup trains available, including motor EFW, AFW, and once through cooling). For the failure to detect case, the consequence is judged to be similar since Room 2024 contains most of the steam. Containment isolation is unaffected.
! Consequence Category: low D Consequence Rand 0
}
l O
FMECA - Consequence Information Report Calculation No. A.PENG C4LC-OlP, Rev. 00 l4-ser 91 Page A14 of A13 Consequence ID: MS-C-09A Consequence
Description:
Degradation of MSIV bypass line during normal operation (line 2 EBB 16 from main steam line 2 EBB 1 to 2CV.10401).
Break Size: Large Isolability of Break: No ISO Comments: Blowdown from faulted steam generator is not isolable. This line is not large enough to cause an automatic plant trip or MSIV isolation.
Spatial Effects: Propagation Effected Iacation: 2155 Spatial Effects Comments: Propagation is through access doors and room siding to outside (roof of adjoining buildings), down through main steam line chase toward the turbine building, and through a stairway door into the turbine auxiliary building.
Initiating Event: N Initiating Event ID: N/A Initiating Event Recovery: Piping is not large enough to cause an automatic plant trip. Detection is expected from a high power alarm and a controlled shutdown is assumed due to standing orders to do so as a result of any unknown or uncontrolled loss of 50 Mwe.
Loss of System: S System IPE ID: PCS S3stem Recovery: PCS loss is assumed to be the demand.
Loss of Train: TM 2 Train ID: PCS, Turbine EFW Train Recovery: The affected steam generator would be isolated once identified during the controlled shutdown. I of 2 steam supply padu to the turbine EFW pmnp is unavailable, but the other steam supply path is more reliable than the tvMnc EFW pump itself.
I Consequence Comment: Consequence is " Low" based on Table 2 2 of Ref. 9.17 (anticipated challenge, long AOT, and at least 2 backup trains available, including EFW, AFW, and once through cooling). Passive closed burier provided by the steam generattir is credited for cor.tainment isolation.
Consequence Category: Low 0 Consequence Rank O 9
p g
FMECA - Consequence Information Report t4-sw91 Calculation No. A PENG-CtLC 019, Rev. 00 Page Al3 ofA13 Consequence ID: MS-C 09B Consequence
Description:
Degradation of MSIV bypass line during normal operation (line 2 EBB 17 from
- main steam line 2 EBB 2 to 2CV 1090-2).
il Break Size: Large Isolability of Break: No
'l ISO Comments: Blowdov/n inom faulted steam generator is not isolable. This line is not large enough to cause an automatic plant trip or MSIV isolation.
Spatial Effects: Propagation Effected 14 cation: 2155 d
Spatial Effects Comments: Propagation is through access doors and room siding to outside (roof of adjoining buildings), down through main steam line chase toward the turbine building, and through a stairway door into the turbine auxiliary building.
Initiating Event: N Initiating Event ID: N/A laitiating Event Recovery: Piping is not large enough to cause an automatic plar.t trip. Detection is expected 1 from a high power alarm and a controlled shutdown is assunux! due to standing orders to do so as a result of any unknown or uncontrolled loss of 50 Mwe.
less of System: S System IPE ID: PCS Sptem Recovery: PCS loss is assumed to be the demand.
J Less of Train: TDM 2 Trala ID: PCS, Turbine EFW l, Train Recovery: The affected steam generator would be isolated once identified during the controlled i
shutdown. I of 2 steam supply paths to the turbine EFW pump is unavailable, but the other steam supply path is more reliable than the turbine EFW pump itself.
l Consequence Comment: Consequence is " Low" based on Table 2 2 of Ref. 9.17 (anticipated cbslienge, long AOT, and at least 2 backup trains available, including EFW, AFW, and once through i
cooling). Passive closed barrier provided by the steam generator is credity for containment isolation.
! Consequence Category: low D Consequence Rank O
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- ' Calculation No. A PENG-CALC 019, Rev. 00
- Page 81 of 826 J
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4 APPENDIX B
! *FMECA - DEGRADA TION MECHANISMS" 4
l j (Attachment Pages 81 826) i t
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O l ABB Combustion Engineering Nuclear Operations i
i
'* FMECA - Degradation Mechanisms Caladanon NoJIMM-019. Rm 00 l Page B2 of B26 1 W eld System ID Segment Line Number Line Description Number Weld 1mation T C P I M E F 0 l
MSS MSS-001 2 EBB-16-2" Main Steam Bypass63-127 Upstream of 2" coupling No No No No No No No No line for 2CV-1010-1 #20 shown in ISO 2 EBB 2 (ISO 2 EBB-16-1)
MSS MSS-001 2 EBB-16-2" Main Steam Bypass63-128 Upstream ortee #13 (ISO No No No No No No No No line for 2CV-1010-1 2 EBB-16-1)
MSS MSS-001 2 EBB-16-2" Main Steam Bypass63-129 Downstream ortee #13 No No No No No No No No line for 2CV-1010-1 (ISO 2 EBB-16-1) 2 EBB-16-2" Main Steam Bypass63-130 Downstream ortee #13 No No No No No No No No MSS MSS-001 line for 2CV-1010-1 (ISO 2 EBB-16-1) 2 EBB-16-2" Main Steam Bypass63-131 Upstream ofelbow #Il No No No No No No No No MSS MSS-001 line for 2CV-1010-1 (ISO 2 EBB-16-1) 2 EBB-16-2" Main Steam Bypass63-132 Downstream of elbow #11 No No No No No No No No MSS MSS-001 line for 2CV-1010-1 (ISO 2 EBB-16-l) 2 EBB-16-2" Main Steam Bypass63-133 Upstream of reducing tee No No No No No No No No MSS MSS-001 line for 2CV-1010-1 #7 (ISO 2 EBB-16-1) 2 EBB-16-2" Main Steam Bypass63-134 Domistream of reducing No No No No No No No No MSS MSS-001 line for 2CV-1010-1 tee #7 (ISO 2 EBB-16-1) 2 EBB-16-2" Main Steam Bypass63-135 Upstream of cibow #5 No No No No No No No No MSS MSS-001 line for 2CV-1010-1 (ISO 2 EBB-16-1) 2 EBB-16-2" Main Steam Bypass63-136 Downstream of elbow #5 No No No No No No No No MSS MSS-001 line for 2CV-1010-1 (ISO 2 EBB-16-1) 2EP'3-16-2" Main Steam Bypass63-137 Upstn:am orcibow #3 No No No No No No No No MSS MSS-001 line for 2CV-1010-1 (ISO 2 EBB-16-1)
Main Steam Bypass63-138 Downstream of elbow #3 No No No No No No No No MSS MSS-001 2 EBB-16-2" line for 2CV-1010-1 (ISO 2 EBB-16-1)
Dearadarwn Mecharisms P - Prunary Water Stnss Corrosion Cracking (PWSCC) M - Microbiologically InDoenced Common (MIC) F- flow Accelerated Cerroman T-Thermal Fatigue I-Intergrarsalar Stress Common Cracking (IGSCC) E- Erosson-Cavitation 0 -Other C- Cerrosion Cracking O O O
0v 0 v [v
'N' Catalad n A'a AMGM-019. Rn. Di7 FMECA - Degradation Mechanisms Page B3 of B26 W eld System ID Segment Line Number Line Description Number Weld IAcation T C P 1 M E F 0 MSS MSS-001 2 EBB-16-2" Main Steam Bypass63-139 Upstream of motor- No No No No No No No No line for 2CV-1010-1 operated vahr 2CV-104C-1 (ISO 2 EBB-16-1)
MSS MSS-001 2 EBB-17-2" Main Steam Bypass65-068 Upstream of half coupling No No No No No No No No line for 2CV-1060-2 #22 shomiin ISO 2 EBB 2 (ISO 2 EBB-17-1)
MSS MSS-001 2 EBB-17-2" Main Steam Bypass65-069 Upstream orelbow #5 No No No No No No No No line for 2CV-1060-2 (ISO 2 EBB-17-1)
MSS MSS 401 2 EBB-17-2" Main Steam Bypass65-070 Downstream of eltow #5 No No No No No No No No ,
line for 2CV-1060-2 (ISO 2 EBB-17-1)
MSS MSS-001 2 EBB-17-2" Main Steam Bypass65-071 Upstream ofelbow #6 No No No No No No No No line for 2CV-1060-2 (ISO 2 EBB-17-1)
MSS MSS-001 2 EBB-l'-2" Main Steam Bypass65-072 Domistream of elbow #6 No No No No No No No No '
line for 2CV-1060-2 (ISO 2 EBB-17-1)
MSS MSS-001 2 EBB-17-2" Main Steam Bypass65-073 Upstream ofelbow #7 No No No No No No No No line for 2CV-1060-2 (ISO 2 EBB-17-1)
MSS MSS-001 2 EBB-17-2" Main Steam Bypass65-074 Downstream of elbow #7 No No No No No No No No i line for 2CV-1060-2 (ISO 2 EBB-17-1)
MSS MSS-001 2 EBB-17-2" Main Steam Bypass65-075 Upstream of motor- No No No No No No No No !
line for 2CV-1060-2 operated wahr 2CV-1090-2 (ISO 2 EBB-17-1)
MSS MSS-001 2 EBB-6-4" Main Steam Supply to 63457 Upstream of piping section No No No No No No No No 2CV-1000-1 #9 (ISO 2 EBB-6-1) .
MSS MSS-001 2 EBB-6-4" Main Steam Supply to 63-058 Upstream of tee #14 (ISO No No No No No No No No 2CV-1000-1 2 EBB-6-1)
Dearadaten Mecharusms j T-Thermal Fatigue P - Primary Water Stress Common Cracking (PWSCC) M -ML Mw"y Innuenced Cerronen(MIC) F- Fler Accelerused Common C- Cenosma Cracting I- freergranular Stress corrosinn Cracking (IGsCC) E - Erosion -Cantation 0-Other
i4-seP97 .
CdculatsorrNo. A-FDSCALC-019.Rev 00 \
FMECA - Degradation Mechan.is m s <
ry u g 36 W cld '
System ID Sepnent Une Number Une Description Number Weld Leestion T C P I M E F 0 MSS MSS-001 2 EBB-6-4* Main Steam Supply to 63459 Dv- :- maoftce#14 No No No No No l '.c No No 2CV-1000-1 8* M-1)
MSS MSS-001 2 EBB-6-4" Main Steam Supply to 63-060 Upstream of 4* cap -item No No No No & No No No 2CV-1000-1 #4 GSO 2 EBB-6-1)
MSS MSS-001 2 EBB 4-4* Main steam Suppfy to 63461 D u- u.,L - n oftce f14 No No No No No No No No 2CV-1000-1 GSO 2 EBB 4-1)
MSS MSS-001 2 EBB 4-4* Main Steam Supply to 63-062 LW ofcIbow #3 No No No No No No No No 2CV-1000-1 GSO 2 EBB-6-1)
MSS MSS-001 2 EBB 4-4* Main Steam Supply to 63-063 Du-mL- ofelbow #3 No No No No No No No No ,
2CV-1000-1 (ISO 2 EBB 4-1) l MSS MSS-001 2 EBB 4-4* Main Steam Supply to 63-064 Upstream ofelbow f2 No No No No W No No No 2CV-1000-1 GSO 2 EBB 4-1)
MSS MSS-001 2 EBB 4-4* Main Steam Supply to 63-065 Dv-u L-o ore! bow s2 No No No No No No No No 2CV-1000-1 GSO 2 EBB 4-1)
MSS MSS-001 2 EBB 4-4* Main Steam Stapply to 63466 LW ofelbow dl No No No No No No No No 2CV-1000-1 GSO 2 EBB 4-1)
MSS MSS-001 2 EBB 4-4* Main Steam Supply to 63467 Downstream ofelbow #1 No No No No No No No No 2CV-1000-1 GSO 2 EBB-6-1)
MSS MSS-001 2 EBB 4-4* Main Steam Supply to 63-068 Upstreant of motor- No No No No No No No No 2CV-1000-1 cperated valw 2CV-1000-1 GSO 2 EBB 4-1)
MSS MSS-001 2 EBB-7-4* EFW Pump Steam 65-054 Upstream of 34* x 4" No No No No No th No No Supply to 2CV-1050-2 weldolet #10 shown ISO 2 EBB-2-2 '150 2 EBB-7-1) e u=*= us=
T-Thermet Faaw F - Pnmary Water Serns Commm Cradung (FWSCC) M - M.-- . , trasenced Commsen(MIC) F-T'.e. Accelerused Comme =
c-Corre nmCrudung 1- beers uneler stras Corresum Credung(tGSCC) E- Ernseen -Caminame 0-Oiher O O O
l c
V J V l
'" FMECA - Degradation Mechanisms N'*"r kMMW19. Rm R7 l l'qce B3 of B:6 weu Systesa ID Segiment Line Nonsber une Descriptbe Neseber Weu IAcaties T C P I M E F 0 MSS MSS-001 2 EBB-74* EFW Pump Steam 65-055 Upstream of elbow #5 No No No No No No No No I Supply to 2CV-1050-2 (ISO 2 EBB-7-1)
MSS MSS-001 2 EBB-7-4* EFW Pmnp Steam 65-056 Dv .smo of elbow #5 No No No No No No No No Supply to 2CV-105r.,-2 (ISO 2 EBB-7-1)
, MSS MSS-001 2 EBB-74* EFW Pump Steam 65-057 Upstream ofcibow M No No No No No No No No Supply to 2CV-1050-2 (ISO 2 EBB-7-1)
MSS MSS 4C1 2 EBB-74* EFW Pemp Steam 65-058 Downuream ofelbow M No No No No No No No No Supply to 2CV-1050-2 (ISO 2 EBB-7-1)
MSS MSS-001 2 EBB-74* EFW Pump Steam 65-059 Upstream orelbow #7 No No No No No No No No Supply to 2CV-1050-2 (ISO 2 EBB-7-1)
MSS MSS-001 2 EBB-74* EFW Pump Steam 65-060 Du-as-- ofelbow f7 No No No No No No No No Supply to 2CV-1050-2 (ISO 2 EBB-7-1)
MSS MSS-Col 2 EBB-74* EFW Pump Steam 65-061 Upstream of motor- No No No No No No Ne No Supply to 2CV-1050-2 operated vaht 2CV-1050-2 (ISO 2EEB-7-l)
MSS MSS 401 2EBC-I-3* Bline llange connection 63-099A Branch line tec #98 to No No No No No No No No flange #99 shown in ISO EBC-I-2 MSS MSS-001 2EBC-14" Main Steam Supply to 63-069 Dv..s-- of motor- No No Ne No No No No No EFW Pmg Iurbine operated wahr 2CV-1000-Drher 2K-3 from Main I (ISO 2EBC-I-1)
Steam Header #I MSS. MSS-001 2EBC-14* Main steam Supply to 63-070 Upstream ofcheck raht No No No No No No No No EFW Pump Turbine 2MS-39A (LW 2EBC-I-1)
Drhrr 2K-3 from Main Steam Header #I
[
Desraduinen Me:hmusem T-umnal Famigue F- thmery Weser Stress Comsnm Crachne(PEW M - brarebeelegnemDy imamenced Carrensen (MIC) F. Flow AccelerusedCarreeria c-Cerre enCrocune 1 a, - stressCerros a.Cracles(tosCC) E-Eroene-Comisseum O-Oderr
tosep.97 '
FMECA - Degradation Mechanisms Cak=ldo"
- A-FDGCALC-oIP. Rn 09 Paki B6 of B:6 Weld System ID Segment line Number Line Description Number Weld IAcaties T C P I M E F 0 MSS MSS-001 2EBC-14* Main steam Supply to 63-071 b a m oofcheck No No No No No No No No EFW Pump Turbine valve 2hG-39A (150 Drim 2K-3 from Main 2EBC-I-1)
Steam Header #1 MSS MSS-001 2EBC-14* Main Steam Supply to 63-072 Upstream of tee #26 (ISO No No No :.o No No No No EFW Pump Turbine 2EBC-1-1)
Driver 2K-3 from Main Steam Header #I MSS MSS-001 2EBC-14" Main Steam Supply to 63-073 Upstream of tee #26 (ISO No No No No No No No No EFW Pump Turbine 2EBC-1-1)
Driver 2K-3 from Main Steam Header #1 MSS MSS-001 2EBC-14' Main Steam Supply to 63-074 Downstn.am ofcheck No No No No No No No No EFW Pump Turbine vahr 2hG-37B (ISO Drher 2K-3 from Main 2EBC-I-1) ;
Steam Header #I MSS MSS-001 2EBC-14* Main Steam Supply to 63-075 Downstream of tee #26 No No No No No No No No EFW Pump Turbine (ISO 2EBC-I-1)
Drher 2K-3 from Main Steam Header #I MSS MSS-001 2EBC-14* Main Steam Supply to 63-076 LW ofcIbow #17 No No No No No No No No EFW Pump Turbine (105 2EBC-I-1)
Driver 2K-3 from Main Steam licader #1 MSS MSS-001 2EBC-14* Main Steam Supply to 63-077 bsmo ofeIbow fl7 No No No No No No No No EFW Pump Turbine (ISO 2EBC-I-1)
Drhtr 2K-3 from Main Steam Header #1 Duradenen Medanens T-Timnnel Feigne P - Pnmary Weer Stress Commian Cradmg (F%W M - Erobsoleycmay influenced Cerremen (MIC) F-Flow AccelerseedCarressem c-Common Crackms I- besynnetar Siress Comm.en cradmg OGSCC) E - Ereman-Canesamsn 0-Other O O O
(3 O O V V O
'*7 FMECA - Degradation Mechanism.s Galan r A~a MMM'-019. Ra. M i Page B7 of B:6 ,
W eld ;
System ID Segment une Number Une Description Naseber Weld Locaties T C P I M E F 0 '
MSS MSS-001 2EBC-14* Main Steam Supply to 63-078 Upstream ofcIbow fl8 No No No No No No No No EFW Pump Turbine (ISO 2EBC-I-1)
Driwr 2K-3 from Main Steam Header #1 MSS MSS-001 2EBC-14" Main steam Supply to 63-079 Dums-m ofelbow #18 No Nr No No No No No No EFW Pump Turbine (ISO 2EBC-1-1)
Drntr 2K-3 from Main Steam Header #1 MSS MSS-001 2EBC-14* Main Steam Supply to 63-080 Upstream ofcibow #19 No No No No No No No No EFW Pump Turbine (ISO 2EBC-1-1)
Driwr 2K-3 from Main Steam Header #1 MSS MSS 401 2EBC-14* Main steam Supply to 63481 Dv-isme. ofeltsw #19 No No No No No No No No EFW Pump Turbine (ISO 2EBC-1-1)
Driver 2K-3 from Main Steam Header #1
. MSS MSS-001 2EBC-14" Main Steam Supply to 63482 Upstream orelbow f20 No No No No No No No No t EFW Pump Turbine (ISO 2EBC-1-1)
Driver 2K-3 from Main Steam Header #1 MSS MSS-001 2EBC-14* Main Steam Supply to 63-083 Ov-. sum ofelbow f20 No No No No No No No No EFW Pump Tmbine (ISO 2EBC-1-1) ,
Driwr 2K-3 from Main Steam Header #1 MSS , MSS-001 2EBC-14* Main Steam Supply to 63-084 Upstream ofcibow #21 No No No No No No No No EFW PumpTurbine (ISO 2EBC-I-1)
Driver 2K-3 from Main Steam Header #1 Deerahnseeu -4.--as T-Thenal Tseysse F- Prunary Water Stress Commsen Camd4 &%W M-E J _ _ c-4 Influenced Commen(MIC) F-Flow Accelerened C-C-Commen Cracking I-Inneryncalar Stress Commen Credsng(IGSCC) E - Erwenn-Cavenhen 0-Other i
t
'# C"' """ N* A ** W R" "
FMECA - Degradation Mechanisms Page B8 of B26 Weld System ID Segment Line Number Line Description Number Weld Imation T C P I M E F O MSS MSS-001 2EBC-14* Main Steam Supply to 63-085 Downstream ofetbow #21 No No No No No ?b No &
EFW Pump Turbine (ISO 2EBC-I-1)
Dm-r 2K-3 from Main Steam Header #1 MSS MSS-001 2EBC-14* Main Steam Supply to 63486 Dv .smo ofpiping No No No No No No No No EFW Pump Turbine secuon #8 (ISO 2EBC-I-1)
Dmer 2K-3 from Mam Steam Header #1 MSS MSS-001 2EBC-14* Main steam Supply to 63-087 Dv-is-- of piping No No No No No No No No EFW PumpTurbine secuon #7 (ISO 2EBC-I-1)
Dristr 2K-3 from Main Srcam Header #1 MSS MSS-001 2EBC-14* Main Steam Supply to 63-088 Du .s-n ofpiping No No No No No No No No EFW Pump Turbine secuon #6 (ISO 2EBC-1 I)
Driser 2K-3 from Main Steam Header #1 MSS MSS-001 2EBC-14* Main Steam Supply to 63-089 Upstream ofelbow #22 No No No No No No No No EFW ? ump Turbine (ISO 2EBC-I-1)
Driver 2K-3 fmm Main Steam Header #1 MSS MSS-001 2EBC-14* Main Steam Supply to 63-090 Du- smi ofelbow #22 No No No No No No No No EFW Pump Turbine (ISO 2EBC-1-1)
Driser 2K-3 from Main Steam Header #1 MSS - MSS-001 2EBC-14* Main Steam Supply to 63-091 Dv- s- of piping No No No No No No No No EFW Pump Turbine secuon #4 (ISO 2EBC-14)
Driver 2K-3 from Main Steam Header #1
Dearadmues:
Mat =wsr=
T-ner uni Fatigue P. Prunary Waner Sarns Cermerse Cracturig (FWSCC) M - Microbiologpcmay Irdhnenced Cerressan (MIC) F-Flow AccelersnedCerrouseus C-Cerrasson Crucis g I- Intergranular Stress Correaren Crecimig (IGSCC) E- Ereusen - Cavanhan 0 -Other 9 9 O
O O O
'" FMECA - Degradation Mechanisms C*"'"'** EMMWIP. Ra. M l' age 89 of B26 W eld System ID Segment une Number Line Desenpties Neseber Weld teestsee T C F I M E F 0 MSS MSS-001 2EBC-14* Main Steam Supply to 63-092 Upstream ofcibow #23 No No No No No .W No No EFW Pump Turbine (ISO 2EBC-I-1)
Dmtr 2K-3 from Main l Steam 11cader #1 MSS MSSM)I 2EBC-I-4* Main Steam Supply to 63-092A Upstream ofelbow #24 No No No No No No No No EFW Pump Turbine (ISO 2EBC-I-1)
Driwr 2K-3 from Main Steam 11eader #1 MSS MSS @l 2EBC-14" Main Steam Supply to 63-093 Downstream ofcIbow #24 No No No No No No No No EFW Penp Tmbine (ISO 2EBC-I-1)
Driver 2K-3 from Main Steam IIcader #1 MSS MSS-001 2EBC-14* Main Steam Supply to 63-094 Downstream of piping No No No N No No No No EFW Pump Tmbine secten #2 (ISO 2EBC-I-1)
Driwr 2K-3 from Main Steam Header #I MSS MSS-001 2EBC-14* Main Steam Supply to 63 #15 Upstream ofcibow #25 No No No No No No No No EFW Pmnp Turbine (ISO 2EBC-I-1)
Driver 2K-3 from Main Steam Header #1 MSS MSS-001 2EBC-14* Main Sacam Supply to 63 #16 Dumou-n ofelbow f25 No No No No No No No No EFW PumpTurbine (ISO 2EBC-1-2)
Driver 2K-3 from Main Steam Header #1 MSS MSS-001 2EBC-14* Main Steam Supply to 63-097 Upstream ofcibow #54 No No No No No No No No EFW Pump Turbine (ISO 2EBC-I-2)
Driwr 2K-3 from Main Sacam Header #I Desadmen Medmamme T-TI w=Iraisee P - Pn wy her stress cum.a. Cr= ting (ruxr) M - M- a * . -5, homenced Cemeen (MIC) F-flew Accelerused Cerve=en C-Corremenn Cra% t - beerynnular Stress Common Crackmg (IGSCC) E-Emmen-Caneessen 0-Odier
1e Uqt.97 FMECA - Degradation Mechanisms Cak*la*** No MEVG-CALC-017.Rev. 00 Page B10 of B:6 W eld System ID Segment Une Number Line Description Number Weld Iecation T C P M E F I 0 MSS MSS-001 2EBC-I-4* Main Steam Supply to 63-098 Downstream ofelbow #84 No No No No No No No No EFW Pump Turbine (ISO 2EBC-I-2)
Driver 2K-3 from Main Steam E.:ader #1 MSS MSS-001 2EBC-14* Main steam Supply to 63499 Upstream ortee #98 (ISO No No No No No No No No EFW Pump Turbine 2EBC-1-2)
Driver 2K-3 from Main Steam Header #1 MSS MSS-001 2EBC-14* Main Steam Supply to 63-100 Dumsme of tee #98
. No No No No No No No No EFW Pump Turbine (ISO 2EBC-I-2)
Driser 2K-3 from Main Steam Header si MSS MSS-001 2EBC-14" Main Steam Supply to 63-101 Upstream or tee #33 (ISO No No No No EFW Pump Turbine No No No No [
Driver 2K-3 from Main Steam Header #1 MSS MSS-001 2EBC-14* Main Steam Supply to 63-102 Dvmsum of tee #83 No No No No No No No No EFW Pump Tmbine (ISO 2 EBC-I-2)
Driwr 2K-3 from Main Steam Header #1 MSS MSS-001 2EBC-14* Main steam Supply to 63-103 Upstream of 4* cap (ISO No No No No No No No No EFW Pump Turbine 2EBC-1-2)
Driwr 2K-3 from Main Steam Header #1 MSS MSS-001 2EBC-14* Main steam Supply to 63-103A Dvmsmu of 4* cap No No No No No No No No EFW Pump Turbine (Item 82 -ISO 2EBC-1-2) ;
Driwr 2K-3 from Main Steam Header #1 l Dearndsam Medannms T-Thmnal resigne r- Pwnery were sarns Co remen Credag (rum M.MR2- - A- InnammiCarressen(MIC)
, F-flew AccelernsedCarremen C-Cerrassa Credirg I- Innerywas!=r Stree Cerresson Cradmg (IGSCC) E- Eremen-Caviestico 0-Oder 9 _
O O
'# C&sidser A~a MMWIP. Rn. 00 FMECA - Degradation Mechanisms Page Bil of B:6 W eld System ID Segveemt Line NumS er Line Descrip: ion Number Weld IAcation T C P I M E F 0 MSS MSS-001 2EBC-I-4" Main Steam Supply to 63-1038 Upstream ofi l/2* x t" No No No No No No No No EFW Pump Turbine reducer (Item 49 -ISO Drher 2K-3 from Main 2EBC-I-2)
Steam Header #1 MSS MSS-001 2EBC-I-4* Main Steam Supply to 63-104 Dounstream of tee #83 No No No No No No No No EFW Pump Turbine (ISO 2EBC-I-2)
Driver 2K-3 from Main Steam Header #1 MSS MSS 4)01 2EBC-I-4* Main Steam Supply to 63-105 Upstream ofmotor- No No No No No No No No ,
EFW Pump Turbine operated vahr 2CV-0340-Drntr 2K-3 from Main 2 (ISG 2EBC-I-2)
Steam Header #1 MSS MSS-001 2EB u-I-4* Main Steam Supply to 63-106 Dommstream of motor- No No No No No No No %
EFW Pump Turbine operated uht 2CV-0340-Drntr 2K-3 from Main 2 (ISO 2EBC-I-2)
Steam Header #1 ,
MSS MSS-001 2EBC-1-4* Main Steam Supply to 63-107 Demistream ofpiping No No No No No No Na No l EFW Pump Turbine section #10 (ISO 2EBC-I-Drher 2K-3 from Main 2)
Steam Header #1 I
MSS MSS-0GI 2EBC-14" Main Steam Supp!y to 63-108 Upstream orelbow #85 No No No No No No No No '
EFW Pump Turbine (ISO 2EBC-1-2)
Drhtr 2K-3 from Main Steam Header #1 i
MSS MSS-001 2EBC-I-4* Main Steam Supply to 63-109 D-.oL- of elbow #85 No No No No No No No No EFW Pump Turbine (ISO 2EBC-I-2)
Drhtr 2K-3 from Main Steam Header #1 Desredsnan Mediensens T-Therinal Fangue P - Prunary Water Swess Carrasian Credting (FWSCC) M - Micrebooleycally Irdimenced Cerressen (MIC) F-Flew AcceterusedCarmasme c-Cerro Cradung 1- beergresinter Sireim Cerres.cn Cradtsng (1GSCC) E- Erossen-Cavempson 0-Odier l
t
'" FMECA - Degradation Mechanisms C"'##"" * " " # ###" 8'?
l' age B12 of B26 Weld System ID Segment une Number Line Description Number Weld Location T C F I M E F 0 MSS MSS-001 2EBC-14* Main Steam Supply to 63-110 Upstream ofelbow #86 No No No No No No No No EFW Pump Turbine (ISO 2EBC-I-2)
Driver 2K-3 from Main Steam IIcader #1 MSS MSS-001 2EBC-14* Main Steam Supply to 63-111 Dowitstream ofelbow #36 No No No No No No No No l EFW Pump Turbine (ISO 2EBC-I-2)
Driver 2K-3 from Main Steam Header #1 MSS MSS-001 2EBC-14* hhin Steam Supply to 63-112 Upstream ofcibow #87 No No No No No No No No EFW Pump Turbine (ISO 2EBC-I-2)
Driver 2K-3 from hhin Steam Header #1 MSS MSS-001 2EBC-14* Main Steam Supply to 63-113 Dvmobmis of cibow #87 No No No No No No No No -
EFW PumpTurbine (ISO 2EBC-I-2) l Driver 2K-3 from hhin Steam Header #1 MSS MSS-001 2EBC-14" Main Steam Supply to 63-114 Upstream ofcibow #88 No No No No No No No No l EFW Pump Turbine (ISO 2EBC-I-2)
Driver 2K-3 from Main Steam Header #1 MSS MSS-001 2EBC-14* Main Steam Supply to 63-115 Downstream ore; bow #88 No No No No No No No No EFW Pump Turbine (ISO 2EBC-I-2)
Driver 2K-3 from Main Steam Header #1 MSS MSS-001 2EBC-I-4* Main steam Supply to 63-116 Upstream ofelbow #89 No No No No No No No No EFW PumpTurbine (ISO 2EBC-1-2)
Driver 2K-3 from hhin Steam Header #1 Dearedstrm Mahannas T-11ermal Fatigue F - Pnrnary Water Stress Cerrousen Cracksig (FWSCC) M - Micrdeeles,cany Irdmenced Cermoen (MIC) F-flow AceelerusedCarraien c CarmuenCrecime 1 - Iraergranular stres Cerroman Crecing OGSCC) E-Eramen-Caventen 0 -06er O O O
p d p V
n U
'# C""ladon Na A&MW19. Rm 00 FMECA - Degradation Mechanisnes Page B13 of B:6 W eld System ID Segment Liec Noneber Line Description Number Weld Imation T C P I M E F 0 MSS MSS @l 2EBC-1-4" Main Steam Supply to 63-117 Upstream ofcibow #90 No No No No No No No No EFW Pump Turbine (ISO 2EBC-1-2)
Dnter 2K-3 from Main Steam Header #I MSS MSS-001 2EBC-1-4" Main Steam Supply to 63-118 Dontistream ofelbow #90 No No No No No No No No EFW Pump Turbine (ISO 2EBC-I-2)
Driwr 2K-3 from Main Steam Header #1 MSS MSS @l 2EBC-I-4" Main Steam Supply to 63-119 Upstream ofelbow f91 No No No No No No No No EFW Pump Turbine (ISO 2EBC-1-2)
Driver 2K-3 from Main Steam Header #I MSS MSS-001 2 ESC-1-4" Main Steam Supply to 63-120 Dom 1: stream of elbow #91 No No No No No No No No i EFW Pump Turbine (ISO 2EBC-I-2)
Driver 2K-3 from Main Steam Header #I MSS MSS-001 2EBC-1-4" Main Steam Supply to 63-121 Upstream ofcibow #92 No No No No No No No No EFW Pump Turbine (ISO 2EBC-I ~ j Dr'ver 2K-3 from Mam Steam Header #1 I MSS MSS-001 2EBC-I-4" Main Steam Supply to 63-122 Dontistream ofelbow f92 No No No No No No No No l EFW Pump Turbine (ISO 2EBC-1-2) l Driver 2K-3 from Main ,
Steam Header #1 MSS MSS-001 2EBC-I-4" Main Steam Supply to 63-123 Dowlistream orpiping No No No No No No No No EFW Pump Turbine section #2 (ISO 2EBC-I-2)
Driver 2K-3 from Main Steam Header #1 Desradmason N T-T1wrmal Fasigue P - Pnmary Wseer Stress Caminen Crackeg (PW3CC) M "A *
- J y InnuencedCeminen(MIC) F-Flow Amelermand Cerrouse C-Carmoon Craimg I ".L ,, drew Common Oschng OGSCC) E- Eremen -Cavnsessa 0-Oder
'" FMECA - Degradation Mechanisms C"'"" 'r No. A-PENGC4LC-019. Rcr. 00 Page Bit of B26 W eld System ID Segment Line Number Line Description Number We!d tecation T C P i M E F 0 MSS MSS @l 2EBC-14" Main Steam Supply to 63-124 Upstream ofcibow #94 No No No !b No No No No EFW Pump Turbine (ISO 2EBC-I-2)
Driver 2K-3 from Main Steam Header #1 MSS MSS-001 2EBC-14" Main Steam Supply to 63-125 Downstream of elbow #94 No No No No No No No No EFW Pump Turbine (ISO 2EBC-I-2)
Drhtr 2K-3 from Main Steam Header #1 MSS MSSml 2EBC-14" Main Steam Supply to 63-126 Du..u-. of elbow #93 No No No No No No No No EFW Pump Turbine (ISO 2EBC-1-2)
Driver 2K-3 from Main Steam Header #1 MSS MSS-001 2EBC-24" Main Steam Header #2 6542 Du-.s e.i or motor- No No No No No No No No Supply to EFW Pump operated vahr 2CV-1050-Turbine Driver 2K-3 2 OSO 2EBC-2-1)
MSS MSS-001 2EBC-24" Main Steam Header #2 6543 Upstream orelbow f2 No No No No No No No No <
Supply to EFW Pump (ISO 2EBC-2-1)
Tmbine Driver 2K-3 MSS MSS-001 2EBC-24" Main Steam Header #2 65-064 Downstream ofelbow #2 No No No No No No No No ,
Supply to EFW Pump (ISO 2EBC-2-1)
Turbine Dristr 2K-3 MSS MSS-001 2EBC-24* Main Steam Header #2 65-065 Upstream ofcIbow #1 No No No No No No No No Supply to EFW Pump (ISO 2EBC-2-1)
Turbine Drhrr 2K-3 MSS MSS-001 2EBC-24* Main Steam Header #2 65-4)66 Downstream ofcibow #1 No No No No No No No No Supply to EFW Pump (ISO 2EBC-2-1)
Turbine Drhtr 2K-3 Desredeemn Mahanisms T-Therweal Farisme P - Pnmary Waser Swess Cerrence Cract ing (PRM M - Microbeelonycally Infhmenced Carressen (MIC) F-Fkne AccelerseedCarrineem c-Cenesien Cruciung I- Irmergrer=lar Stress Cervooon Crechng (IGSCC) E - Ertmuon - Centstane O.Odier O O O
p c o
/
'N C" alan n A'a MMDl9. Rw. 00 FMECA - Degradation Mechanisms Page BIS of B26 W eld System ID Segnest Line Number Line Descripties Number Weld Location T C P I M E F 0 MSS MSS 4M)I 2EBC-2-4* Main Steam Header #2 65-067 Upstream ofcheck ulve No No No No No No No No Supply to EFW Pump 2MS-39B (ISO 2EBC-2-1)
Turbine Driver 2K-3 MSS MSS 4)02 2 EBB-t-36' Main Steam Line from I8-001 Upstream cf 38* x 36' No No No No No No Yes No Steam Generator . '- concentric reducer #18 24A to MSIV 2CV- (ISO 2 EBB-t-1) 1010 36" piping MSS MSS-002 2 EBB-1-38* Main Steam Line from 184)03 Dommtream of 38* x 36' No No No No No No Yes No Steam Generator 2E- concentric re&xer #18 24A to MSIV 2CV- (ISO 2 EBB-I-1) 1010-1 MSS MSS-002 2 EBB-1-38* Main Steam Line from 18-004 Upstream ofcibow #10 No No No No No No Yes No Steam Generator 2E- (ISO 2 EBB-1-1) 24A to MSIV 2CV-1010-1 MSS MSS-002 2 EBB-I-38" Main Steam Line from 18-005 Dv==Lmn of cibow #10 No No No No No No Yes No Steam Generator 2E- (ISO 2 EBB-1-1) 24A to MSIV 2CV-1010-1 MSS MSS 4)02 2 EBB-I-38* Main Steam Line from 18-006 Upstream ofcibow #9 No No No No No No Yes No Steam Generator 2E- (ISO 2 EBB-1-1) 24A to MSIV 2CV-1010-1 MSS MSS-002 2 EBB-1-38" Main Steam Line from 184)07 Du-mb-. ofelbow f9 No No No No No No Yes No Steam Generator 2E- (ISO 2 EBB-1-1) ,
Dormdssion Medan==s T-Thermal Fasigue F- Pnrnary Waser Stress Carrences CracGg (F%W M - MicrdeeleycaRy Indleenced Comissen (MIC) F-Flow Accelerseed Carroman c-Cerro onCr cEms 12, sanns Carros.on Creams (losCC) E-Eressen -Cavession 0-Oher
is-s,97 FMECA - Degradation Mechanisms Calc *la'= w. A-r2xc-catc-oi9. ner. oo Page B16 of B:6 Weld System ID Segmcat Line Number Line Description Number Weld Imation T '
C 7 I M E F 0 MSS M'no02 2 EBB-I-38* Main Steam Line from 18-008 Upstream ofcibow #15 No No No No No No Yes No Sicam Generator 2E- (ISO 2 EBB-t-1) 24A to MSIV 2CV-1010-1 MSS MSS-002 2 EBB-1-38* Main Steam Line from I8409 Dumo^u mo of cibow #15 No No No No No No Yes No Steam Generator 2E- (ISO 2 EBB-I-1) 24A to MSIV 2CV-1010-1 MSS MSS 402 2 EBB-1-38* Main Steam Line from 18-010 Downstream of flow No No No No No No Yes No Steam Generator 2E- clement 2FE-1030 (ISO 24A to MSIV 2CV- 2 EBB-1-1) 1010-1 MSS MSS-002 2 EBB-1-38" Main Steam Line from 1841I Upstream ofelbow #14 No No No No No No Yes No Steam Generator 2E- (ISO 2 EBB-I-1) 24A to MSIV 2CV-1010-1 MSS MSS-002 2 EBB-1-38* Main Steam Line from 18-012 Dumoumu of cibow #14 No No No No No No Yes No Steam Generator 2E- (ISO 2 EBB-1-1) 24A to MSIV 2CV-1010-1 MSS MSS-002 2 EBB-t-38* Main Steam Line from 18412A Domtstream ofpiping No No No No No No Yes No Steam Generator 2E- sectmn #3 (ISO 2 EBB-I-1) 24A to MSIV 2CV-1010-1 MSS MSS-002 2 EBB-1-38* Main Steam Line from 18413 Upstream ofcibow #13 No No No No No No Yes No Steam Generator 2E- (ISO 2 EBB-1-1) 24A to MSIV 2CV-1010-1 pgradame Mechmourre T-Thenner reusoe r - Pnmary wara stress Carrmion Crncing (rwscC) u-un 1 :fyinfluencedCommon(MIC) F-Flow AccelerusedCommen C-Correewn % I- beerpaneler Stress Common Cradung (IGSCC) E - Eressen -Cavissens 0-Other e O O
3 O n
~(V O V
'*" FMECA - Degradation Mechanisms " '"#"'"'" ^'" A * " " #' 8 " "
Pqce BIT of B:6 W eld Systems ID Segment use Number Une Description Number Weld Locaties T C P I M E F 0 MSS MSS-002 2 EBB-I-38' Main steam une from 18-014 hs- ofelbow fl3 No No No No No No Yes No Steam Generator 2E- (ISO 2 EBB-1-1) 24A to MSIV 2CV-I 1010-1 l
MSS MSS-002 2 EBB-1-38* Main Steam une from 18-015 Upstreari ofelbow #8 No No No No No No Yes No Steam Generator 2F- (ISO 2 EBB-t-1) 24A to MSIV 2CV-1010-1 MSS MSS 402 2 EBB-l-38* Main Steam Une from 18416 hs-.. ofelbow #8 No No No No No No Yes No Steam Generator 2E- (ISO 2 EBB-1-1) 24A to MSIV 2CV-1010-1 MSS-002 2 EBB-I-38' Main Steam Une from 18-018 Upstream of flued head No No No No No No Yes No MSS Steam Generator 2E- #16 (ISO 2 EBB-I-1) 24A to MSIV 2CV-1010-1 MSS 402 2 EBB-I-38* Main Steam Line from 63-001 A Upstream ofelbow #6 No No No No No No Yes No MSS Steam Generator 2E- (ISO 2 EBB-I-2) 24A to MSIV 2CV-1010-1 2 EBB-1-38' Main Steam Line from 63-002 Downstream ofcibow #6 No No No No No No Yes No MSS MSS 402 Steam Generator 2E- (ISO 2 EBB-1-2) 24A to MSIV 2CV-1910-1 2 EBB-1-38* Main Steam Line from 63-003 Upsticam of 38* x 10' No No No No No No Yes No MSS - MSS-002 Steam Generator 2E- -@ #13 (ISO 2 EBB-24A to MSIV 2CV- I-2) 1010-1 Desradame W "
P - Prunary Waser Seess Cermaan Crecinng (FRW M - Micre4=olog.cmily homenced Commen (MIC) F-flew AszelerenedCerreene T-T1useni Tetigue
- neery.neser sires. Cerres.en Cradung (TGSCC) E- Ermeen-Caveemes 0-Odwr c-Cerro== Cracams i
'*" FMECA - Degradation Mechanisms C""'#*"' #" d**C-## # #" "'
Page BIS of B:6 W eld System ID Segment Line Number Line Description Number Weld Location T C P I M E F 0 MSS MSS-002 2 EBB-1-38* Main Steam Line from 63-005 Upstream of 38* x 10* No No No No No No Yes No Steam Generator 2E- sweepolet #14 (ISO 2 EBB-24A to MSIV 2CV- 1-2) 1010-1 MSS MSSE 2 EBB-1-38* Main Steam Line from 63-007 Upstream of 38" x 10* No No No No No No Yes No Steam Generator 2E- sweepolet #15 (ISO 2 EBB-24A to MSIV 2CV- 1-2) 1010-1 MSS MSS-002 2 EBB-1-38* Main Steam Line from 63409 Upstream of 38* x 10' No No No No No No Yes No Steam Generator 2E- sweepolet #16 (ISO 2GB-24A to MSIV 2CV- I-2) 1010-1 MSS MSS-002 2 EBB-t-38' Main Steam Line fmm 63-009A Dum& ream ofpiping No No No No No No Yes No Steam Generator 2E- section #1 (ISO 2 EBB-I-21 24A to MSIV 2CV-1010-1 MSS MSS-002 2 EBB-1-38" Main Steam Line from 63-0II Upstream of 38* x 10* No No No No No No Yes No Steam Generator 2E- sweepolet #17 (ISO 2 EBB-24A to MSIV 2CV- 1-2) 1010-1 MSS MSS-002 2 EBB-1-38" Main Steam line fmm 63-013 Dom stream orpiping No No No No No No Yes No Steam Generator 2E- section #2 (ISO 2 EBB-1-2) 24A to MSIV 2CV-1010-1 MSS MSS 002 2 EBB-1-38* Main Steam Line fmm 63-014 Upstream ofcibow #5 No No No No No .No Yes No Steam Generator 2E- (ISO 2 EBB-t-2) 24A to MSIV 2CV-1010-1 Deersdataan Medenums T 11mn. F tic e r - rnmary water stres cerramen osame (rwsec) u-uscreioseye.ity14 :.4 corr (unc) r-rio. Acc ier.nedc.rro c-corronen cruciung 1-traersranminswe=carm oncr.chngposcC) E- Eroman -Cavnetsam O-Odier O O O
O O O
'*" FMECA - Degradation Mechanisms N '"'"'" ^'" A * * # #' # " 8' Page B19 of B:6 Weld System ID Segment Line Number Line Description Number Weld Imation T C F I M E F O MSS MSS-002 2 EBB-1-38" Main Steam Line from 63415 Du-sum of cibow #5 No No No No No No Yes No Steam Genera:or 2E- OSO 2 EBB-1-2) 24A to MSIV 2CV-1010-1 MSS MSS 402 2 EBB-1-38* Main Steam Line from 63-016 Upstream of airgerated No No No No No No Yes No Steam Generator 2E- vahr 2CV-1010-1 OSO 24 A to MSIV 2CV- 2 EBB-I-2) 1010-1 MSS MSS 402 2 EBB-2-36* Main Steam line from 20-001 Upstream of 38' x 10' No No No No No No Yes No Steam Generator 2E- concentne reducer #13 24B to MSiv 2CV- (ISO 2 EBB-2-1) 1060 36" piping MSS MSS 402 2 EBB-2-38* Main Steam Line from 20 003 Du. sum of 38* x 36* No No No No No No Yes No Steam Generator 2E- concentnc reducer #13 24B to MSIV 2CV- OSO 2 EBB-2-1) 1060-2 MSS MSS-002 2 EBB-2-38" Main Steam Line from 20-004 Downstream ofcibow #7 No No No No No No Yes No Steam Generator 2E- (ISO 2 EBB-2-1) 24R to MSIV 2CV-1060-2 MSS MSS-002 2 EBB-2-38* Main Steam Line from 20-005 Du.s- ofcibow #8
. No No No No No No Yes No Steam Generator 2E- OSO 2 EBB-2-1) 24B to MSIV 2CV-1060-2 MSS ,
MSS-002 2 EBB-2-38" Main Steam Line from 20-006 Upstream ofcibow #9 No No No No No No Yes No Steam Generator 2E- OSO 2 EBB-2-1) 24B to MSIV 2CV-1060-2 m- -w T-Thermal Fatigue P- Prunary Water Stress Cermasco Crackmg (P%W hl - nGcrotnelogicarry Infhsenced Carrossen (MIC) F-Flew AccelerusedCommenn C-Cemnion Cradtag a 1-Ir -rgranular Stree Coreassen Cradung (IGSCC) E- Eresson - Cavastian 0-Odier
'" FMECA - Degradation Mcchanisms Garlai n A~a AMMQIP,Rn 00 l' age B:0 of B:6 W eld System ID Segment Line Number Line Description Number Weld Imcation T C P I M E F 0 MSS MSS-002 2 EBB-2-38" Main Steam Line from 20-007 Downstream ofcibow #9 No No No No No No Yes No Steam Generator 2E- (ISO 2 EBB-2-1)
, 24B to MSIV 2CV-1060-2 MSS MSS-002 2 EBB-2-38* Main Steam Line from 20408 Upstream ofcIbow #10 No No No No No No Yes No Steam Generator 2E- (ISO 2 EBB-2-1) 24B to MSIV 2CV-1060-2 MSS MSS-002 2 EBB-2-38* Main Steam line from 20 009 Downstream ofelbow fl0 No No No No No No Yes No Steam Generator 2E- (ISO 2 EBB-2-1) 24B to MSIV 2CV-1060-2 MSS MSS-002 2 EBB-2-38* Main Steam Line from 20-010 Doumstream of flow No No No No No No Yes No Steam Generator 2E- clement 2FE-Il30 (ISO 248 to MSIV 2CV- 2 EBB-2-1) 1060-2 MSS MSS 402 2 EBB-2-38" Main Steam Line from 20-01I Upstream ofelbow #1i No No No No No No Yes No Steam Generator 2E- (ISO 2 EBB-2-1) 24B to MSIV 2CV-1060-2 MSS MSS-002 2 EBB-2-38* Main Steam Line fmm 20412 Downstream ofcibow #11 No No No Na No No Yes No Steam Generator 2E- (ISO 2 EBB-2-1) 24B to MSIV 2CV-1060-2 MSS - MSS-002 2 EBB-2-38* Main Steam Lire from 20-013 Upstream ofc! bow #12 No No No No No No Yes No Steam Generator 2E-(ISO 2 EBB-2-1) 24B to MSIV 2CV-1060-2 Dearedstmi Mecharmens T-Thermal Fatigue P- Phrnery Water Strns Cerressen Cradurig (PWSCC) M-M J J1y influenced Cerressen(MIC) F- flaw Accekrused Care =en c-corremen Creating I Ireersranmier Sire = Cerroman Crachng (ICSCC) E- Enesen -Carna:nen 0-Oeier O O O
\,_- ( V 14- 4 97 ^
FMECA - Degradaten Mechanisms Calculatierr No. A-PENG-C4LC-019. Rev. 00 y,,, 32, ,f 32, w eid System ID Segneemt Line Number Une Description Number Weld IAcstion T C P I M E F 0 MSS MSS-002 2 EBB-2-38* Main Steam lin: fmm 20-014 Downstream ofelbow #12 No No No No No No Yes No Steam Generator 2E- (ISO 2 EBB-2-1) 24B to MSIV 2CV-1060-2 MSS MSS-002 2 EBB-2-38* Main Steam Line from 20-015 Upstream ofelbow #6 No No No No No No Yes No Steam Generator 2E- (ISO 2 EBB-2-1) 24B to MSIV 2CV-1060-2 MSS MSS-002 2 EBB-2-38" Main Steam line from 20-016 Dumsmii ofelbow #6 No No No No No No Yes No Steam Generator 2E- (ISO 2 EBB-2-1) 24B to MSIV 20-1060-2 MSS MSS-002 2 EBB-2-38" Main Steam Line from 654)02 Upstream ofelbow #7 No No No No No No Yes No Steam Generator 2E- (ISO 2 EBB-2-2) 24B to MSIV 2CV-1060-2 MSS MSS-002 2 EBB-2-38" Main Steam Line from 654)02A Dums r.of flued head No No No No No No Yes No Steam Generator 2E- #9 (ISO 2 EBB-2-2) 24B to MSIV 2CV-1060-2 MSS MSS-002 2 EBB-2-38* Main Steam Line fmm 65-003 Dums- ofcibow #7 No No No No No No Yes No Steam Generator 2E- (ISO 2 EBB-2-2) 24B to MSIV 2CV-1060-2 MSS MSS-002 2 EBB-2-38" Main Steam Line from 65-004 Upstream of 38' x 10- No No No No No No Yes No Steam Generator 2E- emyoid #11 (ISO 2 EBB-24B to MSIV 2CV- 2-2) )
1060-2 M w M ahanamin T-Therinal Fatigue P- Pnmey Waner Stree Car esma Cradmg(PWSCC) M-ML" . y Irdluenced Car emme(MIC) F-Flow Accelerused Commen i C-Cerrassan Creding 1-bar grunularStressCer omanCradung(tGSCC) E-Ereman-Cavanesom O-Oeer B
'" FMECA - Degradation Mechanisms N"#"'" Na A&MWIP. Rn. 00 l' age B22 of B26 W eld System ID Segment Une Number Une Description Number Weld facetion T C P M I E F 0 MSS MSS-002 2 EBB-2-38* Main Steam Une from 65-006 Upstream .sf 38 x 10' No No No No No No Yes No Steam Generator 2E- sweepolet #12 (ISO 2 EBB-24B to MSIV 2CV- 2-2) 1060-2 MSS MSS-002 2 EBB-2-38* Main Steam Une fmm 65407A Dumumaofpiping No No No No No No Yes No Steam Generator 2E- section #6 (ISO 2 EBB-2-2) 24B to MSIV 2CV-1060-2 MSS MSS-002 2 EBB-2 Main Steam Une from 65-003 Upstream of 38* x 10" No No No No No No Yes No Steam Generator 2E- sweepolet #14 (ISO 2 EBB-24B to MSIV 2CV- 2-2) 1060-2 MSS MSS 402 2 EBB-2-38" Main Steam Une from 65-010 Upstream of 38* x 10' No No No No No No Yes No Steam Generator 2E- satepolet #15 (ISO 2 EBB-24B to MSIV 2CV- 2-2) 1060-2 MSS MSS 402 2 EBB-2-38* Main Steam Une from 65412 Upstream of 38* x 10' No No No No No No Yes No Steam Generator 2E- sweepolet #16 (ISO 2 EBB-24B to MSIV 2CV- 2-2) 1060-2 i
MSS MSS-002 2 EBB-2-38* Main Steam Une from 65-016 Upstream ofcibow #3 No No No No No No Yes No Steam Generator 2E- (ISO 2 EBB-2-2) 24B to MSIV 2CV-1060-2 MSS MSS 402 2 EBB-2-38* Main steam line from 65417 Dowiistream ofelbow f3 No No No No No No Yes No Steam Generator 2E- (ISO 2 EBB-2-2) l 24B to MSIV 2CV-1060-2 Destadmoon '"- "
T-Thermal Fatigue P - Pnenary Waser Sims Camoan Cradung (PWSCC) M - MiaminnlegicmDy Iremenced Comes ms (1CC) F-Fler= AccelerusedComunes c-Common Crudung I . e ,, sirew Cem= san Cradung(10 SCC) E -Erasen -Cunemhew O-Oser O O O
n r L) a L)
'# FMECA - Degradation Mechanisms Cd'"'"*" V* A-FEVG'C4LC-8 E'" 88 Page B23 of B:6 WeW Systems ID Segment Line Number IJee Description Naseber WeW Locaties T C F I M E F 0 MSS MSS-002 2 EBB-2-38* Main Steam line from 65-CI7A Du- s-u ofpiping No No No No No No Y's No Steam Generator 2E- section #2 (ISO 2 EBB-2-2) 24B to MSIV 2CV-1060-2 MSS MSSm2 2 EBB-2-38* Main Steam line from 654)18 Upstream of air-operated No No No No No No Yes No Steam Generator 2E- vahe 2CV-1060-2 (ISO 24B to MSIV 2CV- 2 EBB-2-2) 1060-2 MSS MSS-003 2 EBB-1-10" Steam generator 2E- 63-004 Domistream of 38* x 10~ No No No No No No No No 3
24A 10* steam line . J #13 (ISO 2 EBB-piping I-2)
MSS MSS-003 2 EBB-I-10" Steam gene ator 2E- 6M06 Domistream of 38* x 10* No No No No No No No No 24A 10* steam line sweepolet #14 (ISO 2 EBB-piping 1-2)
MSS MSS-003 2 EBB-1-10* Steam generator 2E- 63-008 Domistream of 38* x 10" No No No No No No No No 24A 10" steam line e-myotG #15 (ISO 2 EBB-piping 1-2)
MSS MSS-003 2 EBB-t-10* Steam generater 2E- 63-010 Domestream of 38* x 10* No No No No No No No No 24A 10* steam line sweepoict #16 (ISO 2 EBB-piping 1-2)
MSS MSS-003 2 EBB-I-10* Steam generator 2E- 6M12 Dv-.s-.. of 38* x 10' No No No No No No No No 24A 10* steam line . mA #17 (ISO 2 EBB-PIi P ng I-2)
MSS MSS-003 2 EBB-2-10* Steam generator 2E- 65-005 Du 6- of 38 x 10* No No No No No No No No 24B 10* steam linc >@ #II (ISO 2 EBB-piping '. 2)
Daredesen Mech =====
T-Theremi Fatigue P - Prvnery Weser Sems Cemnsen Crackh (F1GCC) M - Mirreheelegpenity hdhneieced Cemesen (MIC) F-Fless AccelerenedCam==e c-Comoson Cracking I-traersrann!sr sims Cemmen Cracbas OGSCC) E - Eremen -Cavessee 0-Oder
FMECA - Degradation Mechanisms C"# d" " *MMW'5' E'r 80 Page B:4 of B:6 W eld System ID Segment Line Number Line Description Number WeldIecation T C P I M E F 0 MSS MSS-003 2 EBB-2-10" Steam generator 2E- 65-007 Downstream of 38* x 10* No No No No No No No No 24B 10" steam line sweepolet #12 (ISO 2 EBB-piping 2-2)
MSS MSS-003 2 EBB-2-10* Steam generator 2E- 65-009 Downstream cf 38* x 10' No No No No No No No No 24B 10* steam line sw@ #13 (ISO 2 EBB-ppng 2-2)
MSS MSS-003 2 EBB-2-10* Steam gerierator 2E- 65-011 Du.. sam of 38' x 10' No No No No No No No No 24B 10" steam line surepolet #14 (ISO 2 EBB-piping 2-2)
MSS MSS-003 2 EBB-2-10" Steam generator 2E- 65-013 Downstream of 38' x 10' No No No No No do No No 24B 10* steam line omyv!G #15 (ISO 2 EBB-piping 2-2)
MSS MSS-003 2 EBB-2-10* Steam generator 2E- 65-014 Downstream of 38 x 10" No No No No No No No No 24B 10" steam line sweepoiet #16 (ISO 2 EBB-piping 2-2)
MSS MSS-003 2 EBB-8-10* Main Steam Dump to 63-019 Upstream of tee #10 (ISO No No No No No No No No Air susy .cic thru 2CV- 2 EBB-8-1) 1001 MSS MSS-003 2 EBB-8-10" Main Steam Dump to 63 019A Downstream of 38* x 10* No No No No No No No No A^u anglec thru 2CV- sweepolet #18 shown in 1001 ISO 2 EBB-1-2 (ISO 2 EBB-8-1)
MSS MSS-003 2 EBB-8 Main Steam Dump to 63-020 Dumamni of tec #10 No No No No No No No No Aiunoy!ee thru 2CV-(ISO 2 EBB-8-1) 1001 Dearean= Mecturens T-11rsmal Fatigue P- Prnary Water stre Carmosan Cincbng(FWsCC) M - MscreealessenIfy Infhranced Car =vesan (M7C) F-Flow AcerlermordCarme==
c-Carmison Cracbng I- sraers unular s== Cer==an Cr chng OGSCC) E-Eressam-Caventsan 0 -oder O .
O O
b
'# C"' "' " Na A&MT-019. Rm 00 FMECA - Degradation Mechanisins Pece B25 of B:5 W eld System 'D Segneemt Line Number Line Description Number WeldImation T C P I M E F 0 MSS MSS-003 2 EBB-8-10* Main Steam Dump to 63-021 Upstream of 10* cap #1I No No No No No No No Nc Atu,uy:sc thru 2CV- (ISO 2 EBB-8-1) 1001 MSS MSS-003 2 EBB-8-10" Main Steam Dump to 63-022 Du..s--of tee #10 No No No No No No No No Atnivy ee thru 2CV- (ISO 2 EBB-8-1) 1001 MSS MSS-003 2 EBB-8-10" Main Steam Dump to 63-023 LW of motor - No No No No No No No No Atn,ug, sc thru 2CV- operated vahr 2CV-1002 100I (ISO 2 EBB-8-1)
MSS MSS-003 2 EBB-8-10* Main Steam Dump to 63-024 Downstream of motor- No No No No No No No No Atmuy:sc thru 2CV- operated vahr 2CV-1002 1001 (ISO 2 EBB-8-1)
MSS MSS-003 2 EBB-8-10* Main Steam Dump to 63-025 Du-mu-- of cibow #6 No No No No No No No No Aiunrpec thru 2CV- (ISO 2 EBB-8-1) 1001 MSS MSS-003 2 EBB-8-10" Main Steam Dump to 63-026 LW of air-operated No No No No No No No No AtmuMcc thru 2CV- vahr 2CV-100I (ISO 1001 2 EBB-8-1)
MSS MSS-003 2 EBB-9-10" Main Steam Dump to 65-015 Upstream ofcibow #1 No No No No No No No No Atuivy:sc thru 2CV- (ISO 2 EBB-9-1) 105I MSS MSS-003 2 EBB-9-10" Main Steam Dump to 65-020 Upstream of motor- No No No No No No No No At6.ugi ee thru 2CV- operated vaht 2CV-1052 ,
1051 (ISO 2 EBB-9-1)
MSS MSS 4303 2 EBB-9 Main Steam Dump to 65-021 Dv-1s-. of motor- No No No No No No No No Atu uyime thru 2CV- operated vahr 2CV-1052 1031 (ISO 2 EBB-9-1)
Des =derson Meannmes T-Thermet Fatigue F- Pnenary Weser Swess Commsen Crackmg (P%W M- KmWM Isasemmt Caereursm @GC) F-TLsur AczeierusedCemiasse ;
c-commen crackmg 1-lasersranmiersweisCem esecrackmgOGSCC) E-Essesse- Toissaan o.osser e
'*" G *"'""#" N" A * " # ^ " "
FMECA - Degradation Mechanisms Page B:6 of B:6 W eld System ID Segment Line Number Line Descriptica Number Weld Location T C F I M E F O MSS MSS-003 2 EBB-9-10" ntain Steam Dump to 65-022 Upstream of air-cprxated No No No No No No No Ne Aturwisie thru 2CV- ta!ve 2CV-1051 (ISO 1051 2 EBB-9-1) ww T-Tbmnet Fatigue F - Primary Water Stress Cerrween Crmir4 (FWSCC) M - Micretme4opcmRy bdhamced Ceressen (MIC) F-flow AcceleresedComme=
C-Cerre-on Craime I-Interar=nular stress Cecman Crucimg OGSCC) E - Demum-Cavitunas 0-Other O O O
Calculation No. A MNG CALC 019, Rev. 00 Page C1 of C5 0 .
l APPENDIX C
'FMECA SEGMENT RISK RANKING REPORT' (Attachment h q: 2 C1 CS)
O ABB Combustion Engineering Nuclear Operations
'*" FMECA - Segment Risk Ranking Report ca'=== 5+ d - * **
rv aqa Degradation Number Lines in Welds in Degradation Degradaties Meckseise Consegeence Risk Risk Segiment ID of Welds Segment Segment Mechanisms Group ID Category Category Rank Catererv MSS-001 109 2 EBB-16-2",63-127,63-128,63- MSS-N NONE LOW CAT 7 LOW 2 EBB-17-2", 129,63-130,63-131, 2 EBB-6-4",2 EBB- 63-132,63-133,63-7-4*, 2EBC-I-3*, 134,63-135,63-136, 2EBC-1-4",2EBC- 63-137,63138,63-2-4* 139//65-068,65-069,65-070,654 7I,65-072,65-073,65-074,65-075//63-057,63-058,63-059,63-060,63-061,63- l 062,63 4 63,63-064, 634 65,63-066,63-067,63-068//65- t 054,65-055,65-056, i 65-057,65-058,65-059,65-060,65-061
// 63-099A // 63-069,63-070,63-071, 63- ;
072,63-073,63-074, !63-075,63-076,63- l 077,63-078,63-079, 63 430,63-081,63-082,63-083,63-084,63-085, 63-086,63-087,63-088,63-089,63-090, 63-091,63-092,63-092A,61-093,63-094,63-095,63-096,63-097,63-098,63-099,63-100, O 9 9
.O O O
'*" FMECA -Segment Risk Ranking Report c~ w -m *=
- rva .r a Degradaties Noseber Lines in Welds in Degradation Degradation Mechanisme Consequence Risk Risk Segenest ID ofWelds Segweent Segnicet Mechanisers Group ID Category Category Caterary Rank 63-101,63-102,63-103,63-103A 63-103B,63-104,63-105,63-106,63-107,63-108,63-109,63-110,63-111,63-112, 63-113,63-114,6?-
115,61-116,63-117,63-118,63-119,63-120,63-121,63-122, 63-123,63-124,63-125,63-126//65-062,654 43,65-064,65-065,65-066,65-067
'*" FMECA -Segment Risk Ranking Report ch vo*Gcuc*" 8 "
re co g a Degradaties Number unes in Welds in Degradation Degradaties Mechanisme Con w Risk Risk Segment ID of Welds Segment Segment Mechanisses Groep ID Category Category Category Rask MSS-002 57 2 EBB-1-36*, 184 01//18-003,18- F MSS-F LARGE hEDIUM CAT 3 HIGi 2 EBB-1-38 , 004,18-005,18-006 LEAK 2 EBB-2-36*,18-007,18-008,18-2 EBB-2-38* 009,18-010,18-011, 18412,18-012A,18-013,18-014,18-015,
- 18-016,18-018,63-001 A.63-002,63-003,63-005,63-007,63-009,63-009A 63-011,63-013,63-014,63-015,63-016//20-0;1 //20-003,20-004,20 005,20-006,20-007,20-008,20-009,20-010,20-011,20-012,20 4 13,20-014,20-015,20 4 16,65-002,65-O(PA 65-003,65-004,65-006, 65-007A 65-008,65- ,
010,65-012,65-016,65-017,65-017A,65-l 018 j l
l O O O
s .
'*" FMECA - Segment Risk Ranking Report N-Ne 8- " ,
r, a ya Degradation ,
Nannher Uses in Welds in Degradaties Degradaties Mechanisme Cee p Risk Risk l Segment ID of Welds Segment Segment Mechanismes Group ID Category Category Category Rank l MSS 403 24 2 EBB-I-10*,63-004,63-006.63- MSS-N NONE MEDIUM CAT 6 LOW !
2 EBB-2-10", 008,63-010,63-012 -
2 EBB-8-10", //65-005,65-007,65- '
2 EBB-9-10" 009,65-011,654 13 i 654 14//63-019,61-Ol9A,63-020,63- I 02I,63-022,63-023,63-024,63-025, 63- !
026 //65 4 I5,65- l 020,654 21,65-022 t
r i
I t
t i
4 .
Calculation No. A.PENG-cal.C-019, Rev. 00 Page Of of D5 O
i APPENDIX D i
1 QUAll1YASSURANCE VERIFICATIONFORMS l
iO O -
ABB Combustion Engineering Nuclear Operations
C:Icul: tion N2. A PENG-CALC 019, R:v. 00 Page D2 of DS Verification Plan
Title:
Implementation of the EPRI Risk Informed Inservice Inspection Evaluation Procedure for the MSS at ANO 2 A
Document Number: C PENG-CALC-019 Revision Number: 00 4
Instriretions: Describe the method (s) of veri 0 cation to be employed, i.e., Design Review. Attemate Analysis, Qualification Testing, a combination of these or an attemative. '!he Design Analysis Verification Checklist irto be used for all Design Analyses Other elements to consider in formulating the plan are: methods for checking calculations; comparison of results with similar analyses etc.
Descrintion of Verification Method:
An independent review will be conducted as appropriate with the work activities described in Project Plan PP 2000839, Revision 00. The verification willinclude:
- 1. Verification of a Design Analysis by Design Review (per QP 3.4 of the Quality Procedures Manual).
- 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.
b Ver' ication Itlan prepared by: Approved by: '
lY Y )
Independet t Reuewer pnnted narne an(sig6diurc ' Man,gement appmver pr.ntehame and signature
/'
ABB Combustion Engineering Nuclear Operations l
1 I
C:Iculation No. A PENG CALC 019, R:v. 00 Page D3 of DS Other Design Document Checklist (PageIof3) instructionc 'lhe independent leviewer is to complete this checklist for each Other Design Document. His Checklist is to be made part of the Quality Record package, although it need not be made a part of or distributed with the document itself, ne second section of this checklist lists potential topics which could be relevant for a particular"Other Design Document'. 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 left blank.)
Title:
Implementation of the EPRI Risk-Informed Insenice Inspection Evaluation Procedure for the MSS at ANO-2 Document Numtw: Revision Number:
A PENG-CALC-019 00 Section 1: To be completed for all Other Design Documents Yes N/A Overall Assessment 1 Are the results/ conclusions correct and appropriate or their intended use? O 2 Are all limitations on the results/ conclusions documented? G Documentation Requirements
- 1. Is the documentation legible, reproducible and in a form suitable for filing and retrieving as a Quality g Record?
- 11. Is the document identified by title, document number and date? @
lit. Are all pages identified with the document number including revision number? @
IV. Do all pages have a unique page number? O V. Does the coraent clearly identify, as applicable:
A. objective S O B. design inputs (in accordance with QP 3.2) S O C, conclusions a O VI. Is the verincation status of the document indicated? @
Vil. if an Independent Reviewer is the supervisor or Project Manager, has the appropri ie approval been S O documented?
Assumptions I. Are all ssumption identified,justiGed and documented? E O
- 11. Arc all assumptions that must be cleared listed? O O A. Is a process in place which assures that those which are CENO responsibility will be cleared? O O f) B. Is a process in place which assures that those which are the customer's responsibility to clear will O 8 V be indicated on transmittals to the customer?
ABB Combustion Engineering Nuclear Operations
6 C:Icul: tion No. A PENG CALC-019. R:v 00 Page D4 of OS Other Design Document Checklist (Page 2 of 3) h Assessment of Signincant Design Changes Yes N/A
- 1. llave sign 10 cant design related changes that might impact this document been considered? g
- 11. If any such changes have been identified, have they been adequately addressed?
O g Selection of Design inputs
- 1. Are the design inputs documented? g
- 11. Are the design inputs correctly selected and traceable to their source? g
!!!. Are references as direct as possible to the original source or documents containing collectiodtabulations of g inputs?
IV. is the 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 documented?
G O Vll, is the verification status of design inputs transmitted from ABB CENS appropriate and documented?
O E Vill. Is the use of customer controlled sources such as Tech Specs, UFSARs, etc. authorized, and does the authorization specify amendment level, revision number, etc.?
S O Referenees
- 1. Are all references listed?
G 11.
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 E
- 2. Applicable Codes and Standards O S
- 3. Literature Searches and Background Data O G 4, Methods O S 5, Hand Calculations O E
- 6. List of Computer Software O S ,
- 7. List of Microfiche O S
- 8. List of optical disks (CD ROM)
O G 9 List of computer disks
- g. O B g
ABB Combustion Enginee-ing Nuclear Operations i
, Cciculation N1 A PENG CALC 019. R3v. 00 Page 05 of DS O
() 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 37, Sectior. 8.0, first line - Yes Concur Yes delete the word " developed"?
O t.
Checklist completed by:
Independent Reviewer g p, yg,y -
g/jp/p7 Prmted Name Sipydd Date Ci V
ABB Combustion Engineering Nuclear Operations