ML20217E967
| ML20217E967 | |
| Person / Time | |
|---|---|
| Site: | Arkansas Nuclear  |
| Issue date: | 08/08/1997 |
| 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-016, A-PENG-CALC-016-R00, A-PENG-CALC-16, A-PENG-CALC-16-R, NUDOCS 9710070326 | |
| Download: ML20217E967 (180) | |
Text
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i i
l Arkansas Nuclear One - Unit 2 4
i Pilot Plant Study l
Risk-Infonned Inservice Inspection Evaluation for 1:he l
Low Pressure Safety Injec1: ion l
l and Shutdown Cooling Systems l
September 1997
= Enter 5 1
i I A.PENO CALC-016 Revisi:n 00 Page1of70 Design Analysis Title Page r'
()h
Title:
Implementation of the EPRI Risk Informed Inservice Inspection Evaluation Procedure for the Low Pressure Safety Injedion and Shutdown Cooling Systems at ANO-2 Document Number:
A PENG-CALC-016 Revision 00 Number:
Quality Class:
O QC 1(Safety Related)
O QC 2 (Not Safety Related) @ QC-3 (Not Safety-Related)
- 1. Approvalof Completed Analysis nis Design Analysis is complete and verified. Management authorizes the use ofits results.
Printed Name f
Signature Date Cognizant Engineer (s)
R. A.Weston
[
[g h'I9 U /4 i
A. V. Bauer I
- h 66 M
Mentor g None
[
L
, fj Independent Reviewer (s)
R. E. Jaquith
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g/8[97 U
s
$ % c)}
Management Approval D. T. Lubin 3
\\
Project Manager
- 2. Package Contents (this section may be completed after Management approval):
Total page count, including body, appendices, attachments, etc.
177 List associated CD-ROM disk Volume Numbers and path names:
@ 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 attachments (specify):
- 3. Distribution:
O B. Boya (2 copies)
N.)
L\\ data \\lubin\\rbitinaNapeng016. doc
Sk IIlI F%Ir ty Calculation No. A PENG-cal.C-016, Rev. 00 Page 2 of 70 BECORD OF REVISIONS Rev Date Pages Changed Prepared By Approved By 00 6 6%
Original R. A. Weston R. E. Jaquith A. V. Bauer O
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ABB Combustion Engineering Nuclear Operations
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Calculation No. A PENG cal.C 016 Rev. 00 x
Page 3 of 70 TABLE OF CONTENTS SECTION PAGE 1.0 PURPOSE...............................................................................................................5
- 2. 0 SCOPE...................................................................................................................5 3.0 SYSTEM IDENTIFICA TION AND BOUNDARY DEFINITION............................................. 6 4.0 C0NSEQ UENCE EVA L UA TlON................................................................................ 13 4.1 CONSEQ UENCE A SSUMP TIONS...................................................................... 20 4.2 CONSEQUENCE SEGMENT LINE BREAK DETECTION CAPABILITIES..................... 23 4.3 C0NSEQ UENCE IOENTlFICA TlON.................................................................... 28 4.4 SHUTDOWN OPERA TION AND EXTERNAL EVENTS........................................... 28
- 5. 0 DEGRA DA TION MECHA NISMS EV/ ^ - A TION........................................................... 4 7 5.1 DA MA GE GR O UPS........................................................................................ 4 8 5.2 DEGRADA TION MECHANISM CRITERIA AND IDENTIFICA TION........................... 48 5.3 BASICDATA.................................................................................................54
- 6. 0 SERVICE HISTOR Y AND SUSCEPTIBILITY REVIEW.................................................... 55
- 7. 0 RISK EVA L UA TION............................................................................................... 58
- 8. 0 EL EMEN T SELEC TlON............................................................................................ 63
9.0 REFERENCES
........................................................................................................68
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LIST OF TABLES NUMBER PAGE 1
L PSUSD C B O UNDA RIES........................................................................................ 10 2
LPSUSD C CONSEQ UENCE A SSESSMENT SUMMA R Y................................................. 34 3
LPSUSDC CONSEQUENCE, FIGURES AND ISOMETRIC DRA WlNGS.............................. 38 4
DA MA GE G R O UPS................................................................................................ 4 7 5
DEGRADA TION MECHANISM CRITERIA AND SUSCEPTIBLE REGIONS......................... 49 6
SERVICE HISTORY AND SUSCEPTIBILITY REVIEW LOW PRESSURE SAFETY INJEC TION / SHUTD O WN CO OLING SYS TEM........................................................... 5 7 7
RISK SEGMENT IDENTIFICA TION............................................................................ 59 8
RISK INSPEC TlON SCOPE....................................................................................... 63 9
EL EMENT SELEC TION RISK CA TEG OR Y 4............................................................. 64 10 ELEMENT SELEC TION - RISK CA TEGOR Y 2.............................................................. 6 7
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LJ ABB Combustion Engineering Nuclear Operations
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Calculall. r
- o. A PENG CALC-016, Res. 00
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Page 4 cf 70 UST OF FIGURES NUMBER EA_G_E G
1 A N0 2 L PSI /SD C S YS TEM.................................................................................... I 2 2
SCHEMA TIC OF SCC SUCTION FLOW FA TH INSIDE CONTAINMENT........................... 39 3
SCHEMA TIC OF SDC SUCTION FLOW PA TH OUTSIDE CONTAINMENT A ND A 3 0 VE EL EV. 33 S '- 0 ".................................................................................... 4 0 4
SCHEMA TIC OF LPSI PUMP 2P-60A SUCTION AND DISCHARGE FLOWPA THS........... 41 5
SCHEMATIC OF LPSI PUMP 2P-608 SUCTION AND O/SCHARGE FLOW PA THS........... 42 6
SCHEMA TIC OF LPSI PUMP DISCHARGE FL OW PA TH.............................................. 43 7
SCHEMA TIC OF LPSIINJECTION PA THS LOOPS 2P-32A & 2P 32C............................ 44 8
SCHEMA TIC OF LPSIINJECTION PA THS LOOPS 2P-32B & 2P-32D............................. 45 9
SCHEMA TIC OF SD C RETURN FL O W PA THS............................................................ 4 6 llST OF APPENDICES A
FMECA CONSEOUENCE INFORMA TION REPORT B
eMECA DEGRADATION MECHANISMS C
FMECA - SEGMENT RISK RANKING REPORT D
QUAlllY ASSURANCE VERIFICA TION FORMS O
r ABB Combustion Engineering Nuclear Operations
A kR MIFIF Calculation No. A-PENC CALC-016, Rev. 00 Page S of 70
- 1. 0 PURPOSE The purpose of this evaluation is to document the implementation of the Electric Power Research Institute (EPRI) Risk-Informed Inservice inspection Evaluation Procedure (RISI) of Reference 9.1 for the Low Pressure Safety injec* ion (LPSI) and Shutdown Cooling (SOC) systems at Arkansas Nuclear One, Unit 2 (ANO-2), Entergy Operations, Inc.
The RISI evaluation process provides an attemative to the requirements in ASME Section XI for selecting inspection locations.
The purpose of RiSI is to Hentify risk-significant pipe segments, define the locations that are to be inspected within these segmsnts, and identify appropriate inspection methods.
This evelvation is performed using the guidelines of the EPRI Risk Informed laservice inspection Evaluation Procedure of Reference 9.1 and in accordance with the raquirements of the ABB Combustion Engineering Nuclear Operations Quality Procedures Manual (QPM-101).
2.0 SCOPE This evaluation procedure applies to the LPSI and SDC systems at AN.)-2, and ritilizes the ISIS Software (Reference 9.2), which has been specifically developed to support and document this procedure.
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As part of the procedure, the system boundaries and functions are identified. A risk evaluation is performed by dividing the system into piping segments which are determined to have the same failure consequences and degradation mechanisms.
The faioure consequences and degradation mechanisms are evaluated by assigning the segment to the appropriate risk category and identifying the risk :ignificant segments.
Fonally, the inspection locations are selected. The guidelines used in determining the degradstion mechanisms, the failure consequences and the risk significant segments are those desco;hed in Reference 9.1.
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Calculation No. A PENG-CALC-016, Rev. 00
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Page S of 70
- 3. 0 SYSTEM IDENTIFICA TION AND BOUNDARY DEFINITION
3.1 System Description
The Low Pressure Safety injection (LPSI) system is part of the Emergency Core Cooling System (ECCS) designed to provide core cooling in the unlikely event of a Loss of Coolant Accident (LOCA).
The cooling prevents significant fuel damage and removes energy generated in the reactor for extended periods of time following a LOCA. The two (2) Low Pressure Safety Irdection (LPSI) trains function by injecting borated water from the Refueling Water Tank (RWL, into the Reactor Coolant System (RCS) to prevent fuel damage and to increase the shutdown margin of the core (Reference 9.3).
The Shutdown Cooling (SDC) System, in conjunction with the main steam and main or emergency feedwater systems, is designed to reduce the temperature of the reactor coolant in post-shutdown periods from normal operating temperature to the refueling temperature.
During shutdown cooling, reactor coolant is urculated by the LPSI pumos through the shutdown cooling heat exchangers to the LPSI headers and retumed to the RCS through the four safety injection nozzles. The circulating fluid flows through the core, out the shutdown cooling nozzle in the reactor vessel outlet (hot leg) pipe, and back to the LPSI pumps. SDC is initiated when the RCS conditions drop below the design pressure and temperature of the shutdown cooling equipment.
O The LPSI/ SOC primary functions are:
i Upon a Safety injection Actuation Signal (SIAS), the LPSIpumps take suction from the RWT and deliver large quantities of borated water to the LPSI header, then into the RCS via the safety injection nozzles when the RCS pressure has decreased sufficiently as would be expected with a large &CS pipe rupture; Provide Shutdown Cooling (SOC) flow through the reactor core and SDC heat exchangers; In conjunction with the RWT, used to filland drain the refueling canal.
l 3.2 System Boundary The Low Pressure Safety injection (LPSI) system and the shutdown cooling system are described consistent with the FSAR (Reference 9.3). The scope of this analysis includes all Class 1 and Class 2 piping in this system which is currently examined in the ANO-2, ASME Section XI Inservice inspection (ISI) Prog am (Reference 9.6). All code and non-code lines which are part of or interface with the LPSI/SDC system 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 in the database (Reference 9.2) as outlined below:
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ABB Combustion Engineering Nuclear Operations N
l ABB Calculation No. A PENG CALC 016, Rev. 00 Page 7 of 70 3.2.1 Lines Downstream of LPSIInjection Une Check Valves (2CCA 218', 2CCA 21 12",
2CCA 22 8', 2CCA 2212", 2CCA-23-8', 2CCA 2312', 2CCA 24-8', 2CCA 24-c 12")
These line segments are common to both the LPSI and HPSI systems. They have already been addressed as part of the HPSI System and are therefore not included in 2
the evaluation of the LPSI system.
3.2.2 Lines Upstream of LPSI Pump Manual Suction Valves 2SI 2A and 2S12B (2HCB 13-14', 2HCB 15-14')
These lines provide suction for the LPSIpumps from the RWT. They are included as part of the Conteinment Spray System and are therefore not included in the evaluation of the LPSIsystem.
3.2.3 Lines Downstream of LPSI Pump Mini-flow Isolation Valves 2CV 5123-1 and 2CV-51241 (2DCB-504-2", 2DCB-505 2')
These lines recirculate LPSI pump mini-flow to the Refueling Water Tank (RWT). A failure during normal power cperation (i.e., standby or periodic testing) would not cause an initiating event. The failure would be detecad shortly after its occurrence based on levelinstruments in the ECCS pump rooms or walkdowns conducted on a regular basis. The failed segment can be isolated by closing the appropriate mini-flow isolation valve from the controlroom. During a demand for LPSI, the segment failure wiH not impact RCS injection. It should also be noted that the miniflow isolation valves are automaticaHy closed and the LPSI pumps are tripped by a Recirculation Actuation Signal (RAS). The line segments are isolated during SDC alignment and a failure would have no impact on SDC operation. Because the line segments downstream of the mini-flow isolation valves are not needed to support LPSI or SDC function, a LOW consequence category is assigned to the segment failure. No degradation mechanisms were identified for the welds in the above line segments.
Based on this assigned consequence category and no degradation potential, the risk significance of the segment failure is 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.
3.2.4 SDC Purification Inlet Line Downstream of 2SI-34 (2FCB-22 3')
This line provides a flow path for purifying reactor coolant during extended SDC operation. The flow path is connected downstream of the shutdown cooling heat excharager outlet piping and discharges to the Chemicaland Volume Control System (CVCS) purification filters.
During normal operation (i.e., standby or periodic testing), the line segment is isolated by a normaHy closed manual valve. The failure would not cause an initiating event. The line segment remains isolated during a LPSI demand andis therefore not needed to support LPSI. During SDC operation, a sman amount of flow is diverted to the CVCS purification filters in order to maintain the purity of thr RCS. Shutdown purification flow is manusHy initiated during the latter stages of shutdown cooling operation when RCS temperature and pressure are at refueling conditions. Because of he localmanualalignment that is required, a failure t
of the segment is considered to be detectable and isolable..Hence, the safety ABB Combustion Engineering Nuclear Operations
A kR M NIF h
Calculation No. A PENG CALC-016, Rev. 00 (0
Page 8 of 70 significance of the segment failure is LOW. No degradation mechanisms were identified for the welds in this segment. Because of the LOW ccnsequence end no degradation potential, the rick significance of the line segment is LOW (i.e., CAT. 7).
8 Since no element selections are needed for low risk significant segments, the welds for this line segment were not entered in the database.
3.2.5 SDC Purification Return Une Upstream of 2SI-35 (2HCB 179 3')
This line provides a flow path for returning the purified reactor coolant to the shutdown cooling system. The flow path is connected upstream of the Volume Control Tank (VCT) inlet line. During normal power operatiors (i.e., standby or puriodic testing), the line segment is isolated by two stormally closed manual valves.
A failure would root cause an initiating event. The line segment remains isolated during a LPSI demand and is therefore not needed to support LPSI. During SDC cperation, a smidt amount of flow is diverted to the CVCS where it is purified and retumed to the SDC system. Shirtdown purification flow is manually initiated during the latter stages of shutdown cu ling operation when RCS temperature and pressure are at refueling conditions. Because of the local manual alignment that is required to establish a return path, the segment failure would immediately be detected and isolated.
Hence, the safety significance of the segment failure is LOW.
No degradation mechanisms were identified for the welds in this segment. Becatase of the LOW consequence and no degradation potential, the risk significance of the line segment is LOW (i.e., CA T 7). Since no element selections are needed for low risk
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significant segments, the welds for this line segment were not entered in the
! L database.
3.2.6 SDC Return Lins to RWT Downstream of Manual Valve 2SI-18 (2GCB 16')
This line provides a flow path from the SDC heat exchanger discharge line to the RWT This line is used to transfer water from the refueling canai to the RWT, itsing the LPSI system. During normal operation (i.e., standby or periodic testing), the segment is isolated by two normally closed manual valves. A failure would not cause an initiating event. The line segment remains isolated during a LPSI or SDC demand. Since the line segment is not needed to support LpSi or SDC, a failure would be of low significance. Alignment of the path for transferring water from the refueling canal to the RWT requires localmanual actions to open the isolation valves.
The failure would be immediately detected and isolated, thus the safety significance is LOW. No degradation mechanisms were identified for the welds in this segment.
Because of the LOW consequence and no degradation potential, the risk significance of the segment is LOW (i.e., CAT. 7). Since no element selections are needed fo:
low risk significant segments, the welds in this line were not entered in the database.
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ABB Combustion Engineering Nuclear Operations
N N II PRIFIF Calculation No. A PENG CALC-016, Rev. 00
~ Page 9 of 70 3.2.7 Inlet and Outlet Lines for SDC Heat Exchangers Downstream of Manual valves 2SI-4A and 251-4B to Upstream of Manual Valves 2SI-SA and 2S1-58 (2GCB 10-12",
2GCB-1612', 2GCB 11-12", 2GCB 1712')
These line segments are common to both the Containment Spray System (CSS) and the LPSUSDC system. They have already been included as part of the CSS and are therefore not includedin the evolvation of the LPSUSDC system.
3.2.8 Lines Downstream of Hot leg Injection Check Valves 2SI-28A and 251-28B (2CCA-25-3')
This fino is connected to the SDC.tuction line and is used for providing hot leg injection to the Reactor Coolant Syrtem in the unlikely event of a large loss of Coolant Accident. This line is included as part of the HPSI system and is therefore not includedin the evaluation of the LPSUSDC system.
3.2.9 Lines with Nominal Diameter of 1" or Less Piping with a nominal diameter of 1' or less was not explicitly evaluated to determine its risk significance. Since volumetric examination of this piping is not practicable, the most effective means to ensure its integrity is via conducticn 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.
O ABB Combustion Engineering Nuclear Operations
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3 Calculation No. A-PENG CALC 016, Rev. 00 (O
Page 10 of 70 TABLE 1 i
LPS.YSDC BOUNDARIES Line Line ISI Pipe Pope Nominal Number Description Drawing Code Diameter (in.)
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Number Class 2CCA 2514" Shutdown coohng suction hnefrom RCS hot leg 2CC4-251 1
14 (Loop B) to shutdown cooling suction valve 2Cl'-
50 % 2 2CC4 25-8" Shutdown coohng suctionfrom refuehng canal 2CC4-25 2 1
8 (from valve 2S1-19 to 14" shutdown coohne hne) 2CCB-3-6" Safety injection pipingfrom control valve 2Cl*-
2CCB-3-1 2
6 5037-) to containment penetration 2P-10 2CCB k-1.5" Vent stack connection upstream ofcontainment 2CCB-3-1 2
1.5 penetratwn 2P 10 2CCB-4-6" Safety injectionfrom 2CV 5017-1 to contcinment 2CCB-4-1 2
6 penetration 2P-15 2CCB-4-1.5" Vent stack connection upstream ofcontainment 2CCB-4-1 2
1.5 penetration 2P-15 2CCB-5 6" Safety injection pipingJhom control valve 2CV-2CCB-5-1 2
6 5077 2 to containmentpenetration 2P-24 2CCB-6-6" Safety injectionfrom 2CV 5057-2 to containment 2CCB-6-1 2
6 penetration 2P-29 f3 2DCB-504-2" LPSIpump 2P-6a4 mini-flow recirculation hne 2DCB-5041 2
2 Q
(from downstream ofcheck valve 2SI 2241 2DCB-305-2" LPSIpump 2P-60B mini-flow recirculation hne 2DCB 5031 2
2 (from dow ustream ofcheck valve 2SI 22B) 2GCB-l-12" Low pressure safety injection pump 2P-6a4 inlet at 2GCB-1-1 2
12 pump inlet
_2GCB-l-14 "
Low pressure saferv iniection pump 2P-6a4 inlet 2GCB-1-1 2
14 2GCB-17-12" Discharge headerfrom shutdown coohng heat 2GCB-17-I 2
12 exchance" 2GCB-2 12" Low pressure safety injection pump 2P-60B inlet 2GCB-2-1 2
12 piping at pump inlet 2GCB-2-14" Low pressure safety injection pump 2P-60B inlet 2GCB-2-1 2
14 paping 2CCB-3-12" Low pressure safety injection pump 2P-6a4 2GCB-3-1 2
12 discharge 2GCB-3-14" Low pressure safety injection pump 2P-604 2GCB-3-1 2
14 discharge 2GCB-3 8" Low pressure safety u,1::%n pump B discharge 2GCB-3-3 2
8 2GCB-5-14" From 2CC4 25 to low pressure safety injection 2GCB-5-4 2
14 pump 2P-604 & B inlet 2GCB-5.t" Connecting hne to CVCS ofSDC drop hne outside 2GCB-5-1 2
3 containment Sh. I 2GCB-5-4" Connecting line to LPSIinjection header ofSDC 2GCB-5-1 2
4 drop hne outside containment t
,mV)
ABB Combustion Engineering Nuclear Operations
A &B Calculation No. A PENG CALC-016, Rev. 00 Page 11 of 70 TABLE 1 (Cont'd)
LPSI/SDC BOUNDARIES Line Line ISI Pope Pope Nominal Number Description Drawing Code Diameter P t.)
Number Class 2GCB-308-2" LPSIpump 2P-60A mini-flow recirculatson line 2GCB 308-1 2
2 (from pump discharee to check valve 2SI 22A) 2GCB-509 2" LPSIpump 2P-60B mini-flow recirculation hne 2GCB-3091 2
2 (from pump discharge to check valve 2SI 22B) 2GCB 710" Low pressure sqfety injection discharge header (10" 2GCB-71 2
10 piping) 2GCB 714" Low pressure safety injection discharge header (14" 2GCB-7-1 2
14 piping) 2GCB-7-4" Connecting line to SDC drop hne ofLPSIinjection 2GCB-71 2
4 header outside containment 2GCB-7-6" Low pressure safety injection discharge header (6" 2GCB-71 2
6 piping) 2GCB-7-8" Low pressure safety injection discherge header (8" 2GCB-7..'
2 8
piping) 2GCB-78-2" Connecting piping to 2PSV-5087 on SDC drop hne 2GCB-7-1 2
2 2GCB 78-1.5" Connecting piping to 2PSV-5087 on SDC drop line 2GCB-7-1 2
1.5 2GCB-8-12" cDC heat exchanger return hne 2GCB-8-1 2
12 2GCB-8-14" Shutdown coohng heat exchanger discharge header 2GCB-8-1 2
14 to low pressure safety infection header 2GCB-8-3" Connecting hnefrom SDC to CVCS 2GCB-8-1 2
3 2GCB-8-6" SDC heat exchanger discharge header to LPSI 2GCB-8-1 2
6 header (C" piping) 2GCB-8-8" SDC heat exchanger discharge header to LPSI 2GCB-81 2
8 header (8" piping)
O ABB Combustion Engineering Nuclear Operations
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Calculatinn No. A-PENG-CALC-016, Rev. 00 Page 12 of 70 ns soet
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Calculation No. A PENG CALC 016, Rev. 00 bage 13 of 70 4.0 CONSEQUENCE EVALUATION The low thessure Safety injection (LPSI) system is a subsystem of the Emergency Core Cooling System (ECCS) and is used to accomplish two major safety related functions, (1)
Low Pressure Safety injection and (2) Shutdown Cooling. A brief description of each function is provided below.
lo)
Low PressurtJafetv Inlectigry To accomplish this function, the LPSI pumps are used to inject large quantities of borated water into the Reactor Coolant System (RCS)In the unlikely event of a loss of Coolant Accident (LOCA) or any other event which involves a rapid decrease in RCS pressure. The LPSIpumps are normally aligned to take suction from the RWT and start automaticaHy upon receipt of a Safety injection Actuation Signal (SIAS).
Both LPSI pumps discharge into a common header which supplies four Individual injection lines. Each of these injection lines contains an isolation valve which also opens automaticaHy upon receipt of an SIAS. Each low pressure injection line is connected to a high pressure injection line which in tum is connected to the one of the RCS cold legs.
Operation of the LPSI pumps is not required during the recirculation mode of ECCS operation.
Consequently, the LPSI pumps are automaticaHy secured by a Recirculation Actuation Signal (RAS) once the Inventory in the RWT is depleted.
The LPSI pump mk.1 flow recirculation lines are also automaticaHy isolated by RAS.
(b)
Shutdown Coolino The shutdown cooling ISOC) function is accomplished by utilizing the LPSIpumpts),
shutdown cooling heat exchanger (s) and the serv't.e water system to remove decay heat from the RCS during the latter stages of plant couldmv.t. The SDC mode of operation is manuaHy initiated when the RCS temperature and pressure decrease to approximately 300'F and 300 psig, respectively. SDC is initiated by unlocking and opening the LPSIpump suction valves to the shutdown rooling wction line, closing the LPSIpump suctiot' valves to the RWT, unlocking and opening the LPSI crossover valves to and from the SDC heat exchangers, closing the containment spray header Isolation valves and closing the LPSIpump mini flow line isolation valves. The LPSI pumps and SDC heat exchangers are then warmed to reactor coolant temperature to minimite thermalshock on these co.r.oonents.
During SDC operation, the LPSI pump (si take suction from RCS hot leg "B' via the SDC suction line. The reactor coolant is then circulated through the SDC heat exchanger (s) and retumed to the RCS throwh the four safety injection nozzles. The heat from the RCS is rejected to the Service Water System as the coolant flows through the heat exchanger (s). This is an ongoing process, with flow through the SDC heat exchanger (s) adjusted periodicoHy, in order to bring the RCS to refueling conditions.
V ABB Combustion Engineering Nuclear Operations
ABB Calculation No. A PENG CALC-016, Rev. 00 Page 14 0f 70
'ho consequence evaluation for the LPSIportion of the ECCS was performed based on the guidance provided in the EPRI procedure (Reference 9.1).
The evaluation focused on the impact of a pipe segment failure on the capability of LPSI or SDC system to perform its 1
design functions, and on the overall operation of the plant. Impacts due to direct and
?
Indirect effects were considered. Generally, the effects of a direct impact are confined to the LPSI or SDC system itself. An Indirect impact resultlag from the failure of a pipe segment would o!!ect neighboring equipment within the LPSI/SDC system or other systemis).
Indirect impacts would generally be caused by flooding, 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 the consequentialimpact on the safety functions.
The spatial effects of a segment failure are primarily associated with flooding, spraying, or jet Impingement. The Internal Flood Screening Study (Reference 9.14) and Plant Design Drawings were used to identify the locations of major ECCS equipment that could be Irmacted by a segment failure. Based on a review of the ANO 2 Intertul Flood Screening Study and Plant Design Drawings, the major equipment of the LPSI system is located in the following rooms of the Reactor Auxiliary Building IRAB):
Room 2084 Upper South piping penetration room and equipment ares Roorn 2055 Lower South piping penetration room and equipment area Rooms 2009 & 2007 East HPSI, LPSI, & containment spray pump area and gallery Rooms 2013 & 2014 West HPSI, LPSI, & containment spray purnp area and gallery Room 2010 HPSIpump *C' area Room 2011 Tendon Gallery Access aren Since the LPSI system interfaces with the RCS, certain piping segments for the LPSI/ SOC system are also located inside the containment.
The walkdown of the LPSI/SDC system was conducted on February 11 & 12,1997 to assess the spatial effects of LPSI or SDC pipe failures, and to supp4 ment the analysis that was performed using the Plant Design Drawings.
Individuals who participated in the walkdown are as follows:
Rick Fougerousse (ANO 2)
Tim Rush (ANO 2)
Randy Smith (ANO-2)
Rupert Weston (ABB-CEI The walkdown (Jcused on the LPSI/ SOC piping locatedin the above rooms and the potential interactions with other equipment resulting from the failure of a LPSI or SDC pipe segment.
The following summarizes the key observations andjudgments that were noted during the walkdown.
O ABB Combustion Engineering Nuclear Operations 1
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Calculation No. A PENG CALC 016, Rev. 00 Page 15 of 70 West HPSr LPSI, & CS Pumo Room Area (flood Zone RAB 2014 ll)
(a)
This area is located at elevation 317' O'In the Reactor Auxiliary Building (RAB). The pumps for train 'A' of the Emergency Core Cooling System (ECCS) (i.e., HPSI pump 2P 89A, LPSIpump 2P-60A, end CS pump 2P 35A) and shutdown cooling heat 2E.
35A are locatedin this ares.
(b)
Two of the pump toom coolers (2VUC 1A & 2VUC 18) are mounted at one end of the shutdown cooling exchanger 2E 35A.
Room cooler 2VUC 1B Is mounted approximetely two feet above floor level with room cooler 2VUC 1A directly above it.
(c)
There are no motor control centers or switchgear within this area.
(d)
The motor for HPSIpump 2P-89A is mounted in the same horizontalplane as the Impeller itself and approximately one foot from the floor level.
(e)
A draln is included in this area. The area is equipped with water level detectors which annunciate in the controlroom. The capacity of the drain is not sufficient to accommodate a limiting break in the service water line for the shutdown cooling heat exchanger. Flooding of HPSIpump 2P 89A is judged to be very likely if a break occurs in the service water line for the shutdown cooling hest exchanger within the room. Because the HPSI pump motor is mounted close to the floor, it is very V) unlikely that the control room operators will be able to terminate the flow through i
the break before HPSIpump 2P-89A is flooded.
(f)
Access to this area is through a water tight door which is maintained closed.
(g)
Propagation of water from this area to the adjoining flood tones at elevation 317' 0*
is very unlikely because of the elevated pathway via the ventilation shaft. The ventilation damper is located several feet above floor level, and the access door is closed.
Epst HPSI LPSI, & CS Pumo Room Area (Flood Zone RAB 2007-LL)
(a)
This area is located at elevation 317'0* In the RAB. The pumps for train 'B' of ECCS fl.e., HPSI pump 2P-898, LPSI pump 2P 608, and CS pump 2P-358) and chutdown cooling heat 2E 358 are located in this area.
(b)
The layout of this area is similar to the West Pump Room Area, however this area is much larger.
(c)
Room coolers 2VUC If) & 2VUC tF are mounted close to the floor approximately 15 feet apart, whereas room cooler 2VUC 2E is mounted directly above cooler 2VUC-IF. Au room coclurs are located at one end of shutdown cooling heat exchanger 2E.
359.
(d)
Similar to tht' West Pump Room Area, a floor drain is included in this flood tone.
f'~'y The area is equipped with water level detectors which annunciate in the control
)
room. The capacity td the drain is not sufficient to accommodate a limiting break in ABB Combustion Engineering Nuclear Operations
ABB Calculation No. A.PENG. CALC-016, Rev. 00 Page 16 of 70 the service water line for the shutdown cooling heat exchanger. Flooding of HPSI pump 2P-898 is judged to be very hkely if a break occurs in the service water line for the shutdown cooling heat exchanger within the room. Because the HPSIpump motor is mounted close to the floor, it is very unkkely that the control room operators wiH be able to terminate the flow through the break before HPSIpump 2P.
898 is flooded.
(e)
There are no motor control centers or switchgear within this pres.
(f)
Access to this area is through a water tight door which is maintained closed. This operating practice, in addition to the elevated pathway via the ventilation shaft, makes it unkkely that water wiH propagate from this flood zone to adjacent flood zones at elevation 317' 0*, before corrective actions from the control room are performed.
HPSI Pumo 'C' Area (Flood Zone RAB 2010-LL)
(a)
This area is also located at elevation 317'-0* in the RAB. Unhke the other two pump rooms, this area contains the ' swing
(b)
The room coolers for this area are mounted in the ceiling.
(c)
Rigid supports / restraints are in place for HPSIpump 2P 89C piping.
(d)
Access to this area is through a water proof door which is maintained closed.
General Access Area (Flood Zone RAB 2006-LL)
(a)
This area is located at elevation 317' O' of the RAB. There are no safety related equipment located within this area.
(b)
The propagation of water from elevations 33S' O' and above wiH eventuaHy drain to the General Access Area. Propagation to this area wiH occur via the floor drains at higher elevations and stairway No. 2001, Because of the large open area, this flood zone and the adjoining Tendon GaHery Access Area can accommodate a significant amount of water before spiuov tr to the ECCS pump room is threatened.
(c)
Entrance to the General Access Area is via a non water tight door from stairway <?o.
2001. It is believed that this door wiH not be able to prevent the propagation of water from the stairway to the General Access Area.
(d)
The propagation of water fivm the General Access Area to the ECCS pump rooms via the ventilation shaft wiH occurif the ventilation dampers are not closed and there is an unlimited amount of water from the flood source.
(e)
The ventilation dampers are located several feet above the floor at elevation 317' O'.
Automatic closure of these dampers will occur l' a Safety injection Actuation Signal (SIAS)is generated.
ABB Combustion Engineering Nuclear Operations
ABB Calculation No. A PENG CALC 016 Rev. 00 Page 17 of 70 (f)
The RAB sump which is located in the General Access Ares is equipped with level detectors which annunciate in the controlroom. Because the General Access Area can accommodate a significant amount of water and the RAB sump water levelis directly indicated and annunciated in the control room, it is judged that there wiH be sufficient time available for the control room operators to close the ventilation dampers in order to prevent spillover into the ECCS pump rooms.
Lower South EaviDmerQggm (Flood Zone RAB.2055 JJ)
(a)
This area is located at elevation 335' 0"In the RAB. There are no motor control centers or switchgear within this area.
(b)
The HPSI orifice bypass valves (2CV 51031 & 2CV 5104-2) are located within this room, but for apart from each other (i.e., at opposite sides of the room) so that spraying or Jet Impingement resulting from a LPSI or SDC pipe failure would not affect both valves simultaneously. Because of its closeness, bypass valve 2CV.
51031 would be more likely to be effected by spraying orlet Impingement caused by a pipe failure in the LPSI or SDC line segment. It was also observed that there are pipe supports / restraints and other larger piping which provide some amount of protection from spatialinteractions.
(c)
Entrance to this area is through a non water tight door..cor the limiting break sita considered, it is judged that the force exerted on the door wiH cause it to open and provide a propagation path to flood zone itAB 2040-JJ. The water wiH eventually G
propagate to the RAB sump at elevation 317' 0*.
(d)
Although this room is equipped with a floor drain, the capacity of the drain cannot accommodate the outflow from a limiting pipe break in the LPSI or SDC line in this area.
(e)
The LPSI header isolation valve (2CV 5091) and the SDC heat exchanger return valve (2CV 5093) are located within close proximity of each other within this area.
It is judged that both valves will be effected by spraying orjet impingement caused by a break in the LPSI or SDC piping in this area.
If)
All valves within this flood zone are located three or more feet above the floor and are not expected to be affected by flooding (see item (c) above).
14?pJtr South Pipino and Penetration Room (Flood Zone RAB-2084-DD)
(a)
This area is located at elevation 360' O'. Although there are two floor drains within this area, the capacity of the drains cannot accommodate the outflow from a limiting pipe break in the SDC line in this area.
(b)
The HPSI, LPSI, and CS line injection valves are located within this area. Other safety related valves located within the area include the 11 PSI hot leg injection valves, the service water line isolation valves for containment cooling units 2VCC.
2A and 2VCC-28, the shutdown cooling line isolation valve, and the EFW Q
distribution valves to steam generator 2E 24A.
t/
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ABB Calculation No. A PENG. CALC-016, Rev. 00 Page 18 of 70 (c)
Entrance to this area is through a non water tight door. For the limiting break size considered, it is judged that the force exerted on the door will cause it to open and provide a propagation path to flood zone RAB 2073 DD at elev.ation 354'-0".
one entrance door willbe forced open before significant amount of water can accumulate inside the room and flood the vake motors. The outflow of water willpropagate to lower elevations and eventually to the RAB sump via open grating, stairway No.
2001, and the floor drain system.
(d)
There are no motor control centers or switchgear within this area.
(e)
The limitorque actuators for the motor operated vakes are scaled to minimite the effects caused by spraying orjet impingement.
(f)
A failure in LPSIline containing isolation volve 2CV 50171 may cause spraying or jet impingement to the following vakes:
2CV 50151 2CV-5016 2 2CV 60381 The affected valves are within close proximity of the line segment where the failure is postulated. Other valves within the room are judged to be unaffected by spraying orjet impingement because they are located at a far enough distance fmon the break location, or because of rigid supports / restraints or large piping i e path of trajectory.
(g)
A failuto in LPSIline containing isolation vake 2CV 5057 2 may cause spraying or jet impingement to the following valves; f
2CV 56121 2CV 50551 2CV 5056 2 2CV 1511 1 The affected valves are within close proximity of the line segment where the failure is postulated. Other valves within the room are judged to be unaffected by spraying orjet impingement because they are located at a for enough distance from the break location, or because of rigid supports / restraints or large piping in the path of trajectory.
(h)
A failure in LPSIline containing isolation valve 2CV 50371 may cause spraying or jet impingement to the following valves.*
2CV 1511 1 2CV 15191 The affected valves are within close proximity of the line segment where the failure is postulated. Other vanes within the room arejudged to be unaffected by spraying orjet impingement because they are located at a for enough distance from the break ABB Combustion Engineering Nuclear Operations l
ABB i
Calculation No. A PENG CALC 016, Rev. 00 Page 19 of 70 location, or because of rigid supports / restraints or large piping in the path of trajectory.
(i)
A failure in LPSIline containing isolation valve 2CV 5077 2 may cause spraying or jet impingement to the following valves:
2CV 5613 2 2CV 4040-2 2CV 10371 2CV 1038 2 2CV 10251 2CV 1026 2 The effected valves are within close proximity of the line segment where the failure is postulated. Other valves within the room are judged to be unaffected by spraying orjet impingement because they are located at a for enough distance from the break location, or because of rigid supports / restraints or large piping in the path of trajectory.
(j)
A failure in SDC suction line containing isolation valve 2CV 50381 may cause spraying orjet impingement to the following valves:
,m 2CV 50151 I
)
2CV 5016 2 2CV 50171 2CV C6121 The affected volves are within close proximity of the line segment where the failure is postulated. Other valves within the room are judged to be unaffected by spraying orjet impingement because they are located at a far enough distance from the break location, or because of rigid supports / restraints or large piping in the path of trajectory.
In performing this evaluation, several types of inputs were used and several assumptions were made. These inputs and assumptions are discussed in Section 4.1.
Twenty three consequence segments were identified for LPSI/SDC for the lines entered in une detabase.
Of ti.e twenty three, fourteen were assigned as "HIGH" and nine as ' MEDIUM *.
The consequence assessment summary for these segments is provided in Section 4.3.
The bases and justifications for each category assignment are provided in Appendix A.'
This appendix contains reports obtained from the ISIS software (Reference 9.2) for the LPSl/SDC system. For the LPSI/SDC lines not entered in the database, four were assigned as " LOW" (see Section 3.2).
(*
id
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ABB Combustion Engineering Nuclear Operations
ABB Calculation No. A.PENG. CALC-016, Rev. 00 Page 20 of 70 4.1 CONSEQUENCE ASSUMPTIONS 4.1.1 It is assumed that aH pipe degradation processes are relatively slow and that pipe failure tends to occur randomly in time. This assumption of randomness implies that the failure of a piping segment can occur during any of the various configurations (i.e., standby, testing and respond to a LPSI or SDC demand). For the purpose of this evaluation, OH configurations were considered to determine the limiting consequences of segment faHures in the LPSI system.
At the time a segment failure is detected, the LPSI system wiH be in one of its vadous configurations (i.e., standby, periodia testing nr response to a demand).
While in any of these configurations, the existence of a segment failure would be detected within a short time after its occurrence, if it is detectable. 11 a LPSI segment failure is detected dudng the standby or periodic testing configuration, the Technical Specification requires that the effected LPSI train be restored to operability within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or plant shutdown must be initiated. Because of the 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> AHowed Outage Time (ACT), the fault exposure time would be designated as s
'Short AOT'(Reference 9.1) for the standby or periodic testing configuration.
The fault exposure time of a segment failure which is detected during a LPSI or SDC demand would be much longer. The length of the exposure time would depend on whether the piping segment is exposed to the operating pressures on a quarterly basis or during a one year period. According to Reference 9.1_ piping segments exposed at least once during the year or not at sH are treated as having an 'AH Year" fault exposure time. Otherwise, piping segments exposed on a quarterly basis are treated as having a 'Between Test' fault exposure time. For both cases, the existence of a pipe degraded condition is assumed to be discovered (if it exists) during the last time the piping was exposed to LPSI operating pressures.
Table 3.2 of Reference 9.1 shows that the consequence of a segment failure depends on ths frequency of chaHenge, the number of equivalent backup trains, and the fault exposure time which is linked to the system configuration. The system is used to accomplish LPSI or SDC During LPSI operation, the system is designed to mitigate a large loss of Coolant Accident (LOCA), and the frequency of challenge is classified as a Design Basis Category IV event (Table 1 of Appendix A). For the segments common to LPSI and SDC operations, the consequence evaluation demonstrates that there are no equivalent backup trains available whether or not the failed segment is isolated during LPSI operation. Since there are no equivalent backup trains, the resulting consequence is 'HIGH' which is the most limiting.
A consequence assessment of piping segments used to perform only shutdown cooling operation is provided in Section 4.4.
The assessment considered the resulting consequences of a segment failure for various conditions or configurations including potential LOCAs where applicable, initiation of shutdown cooling, midloop operation and shutdown operation during the latter stages of the outage (i.e.,
refueling). The assessment demonstrates that the bounding consequence resulting from a failure of the shutdown cooling piping is ' MEDIUM".
O ABB Combustion Engineering Nuclear Operations
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Calculation No. A PENG CALC-016, Rev. 00
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Page 21 of 70 4.1.2 Piping segments that are used to accomplish both the LPSI end SDC functions are ess;gned the more limiting consequences associated with LPSI operation. Other piping segments were cor.sidered to determine the consequences during normal power or SDC operation. SpecificaHy; la)
Piping that is within the Reactor Cooling System (RCS) pressure boundary can only failin such a way as to cause an initiating event (i.e., a LOCA).
(b)
Piping that is downstream of the SDC isolation volve and upstream of the LPSI pump manualisolation valves (in the SDC suction line) is usually not exposed to PCS operating pressures.
The SDC suction volves are interlocked with RCS pressure to prevent them from opening during normal power operation. Piping segments downstream of the SDC heat exchanger manual outlet valves to upstream of the SDC throttle valve are not exposed to LPSI system operating pressures on a quarterly basis. Therefore failures of these piping segments are treated during shutdown operation (see Sections 4.4.2 through 4.4.7).
(c)
Piping in other segments of the LPSI/ SOC system is exposed to about the same pressure during periodic testing or during an actualdemand. For these segments, the limiting configuration (i.e., during LPSI demand) is addressed in the evaluation.
(
4.1.3 A segment failure in any system has the potential to (1) lead to an initiating event G
that might require a protective response from the Reactor Protection System (RPS) and/or the Engineered Safety Features (ESF), (2) degrade the reliability of the ESF, or (3) both I and 2. SpecificaHy, it is assumed that:
(a)
For pipe segment failures that occur during an actual LPSI demand (i.e., in response to a large LOCA), it is assumed that it is unlikely the failed segment wiH be isolated because of the short duration during injection for a large LOCA even though the failure is annunciated in the controlroom. Note also that the operators are highly stressed during this period. According to Table 3.2 of Reference 9. t, the untlliability 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.
(b)
For pipe segment failures that occur during SDC operation, the potential for break isolation is evaluated andif the break is readily detectable and isolable, then the isolation capability is treated as one additional success path, or the equivalent of having one backup train available.
This is based on the operators being less stressed because shutdown cooling is manusHy initiated during the latter stages of plant cooldown and is accomplished by performing certain local actions.
The probability of the operators not performing
(~h corrective actions based on adequate information in the control room is
()
typicaHy 1.0E 2 (Reference 9.18).
Therefore, failure of the operators to ABB Combustion Engineering Nuclear Operations
ABB Calculation No. A PENG CALC 016, Rev. 00 Page 22 of 70 isolate a faHed segment dunno SDC operation is treated as equivalent to having one backup train.
4.1.4 ft is assumed that the configuration of the LPSI system is normany in the standby state. In this state, the LPSIline injection isolation valves are closed and the LPSI pumps are aligned to take suction from the Refueling Water Tank (RWT). The LPSI headerisolation valve andits associated bypass volve are assumed to be in the fumy open posluon during normalpower operation. During the latter stages of cooldown, the SDC system is manusHy aligned to hot leg *B' of the RCS with reactor coolant circulating through the SDC heat exchanger (s). The LPSI header isolation volve and the SDC throttle valve are modulated to achieve and maintain the desired cooldown rate.
4.1. 5 Motor operated valves which are normaHy open, and faHas is, are assumed to be unaffected by spraying orlet impingement caused by the faHure of a ploe segment.
4.1. 6 Based on the initiating event trees that were developed and used in the Individual Plant Examination for ANO 2 (Reference 9.15), there are no available backup trains to mitigate core damage foHowing loss of the LPSI function in response to a large LOCA.
4.1. 7 The LPSI pumps are periodicaHy tested on e quarterly basis to comply with the inservice testing requirements (Reference 9.13, Tech Spec 4.5.2(f)) for these pumps.
4.1.8 Because there is an overlap period where the reactor coolant pumps are stiH running with the steam generators being fed and SDC is in operation, it is assumed that a SDC line break in the Upper South Piping Penetration room would have no significant impact on the abHity to deliver emergency feedwater (EFW) to the steam l
Generator 2E 24A. The EFW distribution valves in this room would have already been opened and failin this position due to spatial effects caused by the segment failure.
4.1.9 for pipe segments normaHy isolated from the RCS (i.e., between SDC suction valves i
2CV 5084-1 and 2CV 5086 2), a potential LOCA initiating event was considered in 1
the evaluation. A potential LOCA is defined as a failure of the motor operated valve l
to remain closed foHowed by the failure of the associated SDC segment. Since the l
probability of passive failure of the motor operated valve is approximately 3.0E 3, the combined effect of this failure and the condit!cnal core damage probability (CCDP) for a large LOCA (Table 1 of Appendix Al results in a MEDIUM consequence. The combined effect of two motor operated valves failing and the CCDP for a large LOCA would be less risk significant and the resulting consequence is LOW.
G 4
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Calculation No. A PENG CALC-016. Rev. 00 V
Page 23 of 70 4.2 CONSEQUENCE SEGMENT LINE BMEAK DETECTION CAPABILITIES This section summarises the unexpected alarms and indication available to the operator in the control voorn for detecting the failure of a LPSI/SDC segment. Segments that exhibit similar detection capabilities are grouped and shown below. Additionalinformation for each segment is providedin Appendix A.
(PSI Conseguence S gment LPSI-C-01 t
A line break in this segment of the SDC drop line upstream of 2CV50841 is a LOCA initiating event and the consequence is therefore driven by that classificadon.
LULQ1'1Lt90tGgy Seament LU! C 02 for a line break downstream of 2CV 50841 and upstream of containment penetration 2P.
27 during a demand for shutdown cooling, several unexpected alarms and indications would be encountered. Specifically; Low pump discharge flow (2FT 5091) would be annunciated.
Containment sump high level would be annunciated.
Unexpected motor amp reading (erratic or low from 2/l 5018A).
Decreasing RCS Inventory which is specifically monitored when isolation valves opened.
Decreasing RCS pressure.
p\\
{
Discharge pressure Indication (2PT 5092) much lower than expected.
l Suction pressure Indication (2PT 5039) much lower than expected or zero.
Because of the focused attention associated with the initiation and maintenance of shutdown cooling, the various indicat!ons and alarms associated with this event would cause an immediate response, it is therefore considered reasonable to conclude that this line break would be detected and mitigated by isolation of the failed line segment.
LP_S!_Co_nyggyence Seament LPSI C-03 For a line break downstream of containment penetration 2P 27 and upstream of the elevation 360' 0* floor penetration during a demand for shutdown cooling, several unexpected alarms and indications would be encountered. Specifically; Low pump discharge flow (2FT 5091) would be annunciated.
Auxiliary building sump high level would be annunciated.
Weste drain tank high level would be annunciated.
Auxiliary building high radiation would potentially be annunciated.
Unexpected motor amp reading (erratic or low from 211-5018A).
Decreasing RCS inventory which is specifically monitored when isolation valves opened.
Decreasing RCS pressure.
Discharge pressure indication (2PT 50921 much lower than expected.
Suction pressure indication (2PT 5039) much lower than expected or zero.
Because of the focused attention associated with the initiation and maintenance of
(,)
shutdown cooling, the various indications and alarms associated with this event would v
ABB Combustion Engineering Nuclear Operations
ABB Calculation No. A.PENG CALC-016, Rev. 00 Page 24 of 70 cause an immediate response. It is therefore considered reasonable to conclude that this line break would be detected and mitigated by isolation of the isHedline segment.
IPSI ConjgevenceJoyttent LPSI-C 04 For a line break downstream of the' elevation 360' 0* fioor penetration and upstream of the floor elevation 335' O'penet. Atic ts, including lines to 2SI 20 and 2SI 35, during a demand for shutdown cooling, several unexpected alarms and indications would be encountered.
SpecificaHyl Low pump discharge flow (2FT 5091) would be annunciated.
Auxiliary building sump high level would be annunciated.
Weste drain tank high level would be annunciated.
Auxiliary building high radiation would potentiaHy be annunciated.
Unexpected motor emp reading (erratic or low from 2115018A).
Decreasing RCSInventory which Is specificsHy monitored when isolation valves opened.
Decreasing RCS pressure.
Discharge pressure Indicstion (2PT 5092) much lower than expected.
suction pressure Indication (2PT 5039) much lower than expected or zero.
Because of the focused attention associated with the initiation and maintenance of shutdown cooling, the various indications and alarms associated with this event would cause an immediate response. It is therefore considered reasonable to conclude that this line break would be detected and mitigated by isolation of the failed line segment.
t (PSI Conseover'cv Srgments LPJi C 05 /10 for line breaks downstream of the elevation 335' O' floor penetration and upstream of LPSI pump suction isolation va!ves 2SI 1Ai2SI 1B during a demand for shutdown cooling, several unexpected alarms and indications would be encountered. SpecificaHy; Low pump discharge flow (2FT 5091) would be annunciated.
ECCS pump room high level would be annunciated.
Auriliary building high radiation would potentially be annunciated.
Unexpected motor amp reading (erratic er low from 2115018A).
Decreasing RCS inventory which is specificaHy monitored when isolation valves opened.
Decreasing RCSpressure.
Discharps pressure indication (2PT 5092) much lower than expected.
Suction pressure Indication (2PT 5039) much lower than expected or zero.
Because of the focused attention associated with the initiation and maintenance of shutdown cooling, the various indications and alarms associated with this event would cause an immediate response, it is therefore considered reasonable to conclude that this line break would be c'stected and mitigated by isolation of the failed line segment.
LPSI Consecuence Seaments LPSI C 06 /11 Forline breaks downstream of the LPSIpump shutdown cooling suction isolation valves 2St.
1A/2St 18 and RWT suction isolation valves 2SI-2A/2SI28, and upstream of pump ABB Combustion Engineering Nuclear Operations
ABB C
Calculation No. A PENG CALC 016 Rev. Ob Page 25 of 70 dischstge valves 2SI 3A/2SI 38 and recirculation isolation volves 2CV 5123 l/2CV 51241, in cerVunction with a LOCA demand, the foHowing unexpected clarms and Indications would be encountered. SpecificsHy; ECCS pump room high level would be annunciated very quickly after break initistion.
Fluoding and subsequent breaker opening of HPSI, LPSI and/or CSS, pumps would be annunciatedin the controlroom.
Low LPSIpump discharge flow (2FT 5091) would be Indicated.
Dischstge pressure indicotlon (2PT 5092) v ould be much lower than expected.
It le therefore considered reasonable to conclude that this line break would be detected and mitigated by isolation of the failedline segment.
LPSI Conrntuence Sooments LPSI C 07 /12 For line breaks downstream of the LPSIpurno discharge valves 2SI 3A/2SI 311 and upstream of ECCS room penetrations and 2Sl4A/2SI48, in con]vnction with a LOCA demand, the foHowing unexpected alarms and lodications would be encountered. Specificstly; ECCS pump room high level would be annunciated very quickly alter break initiation.
Flooding and subsequent breaker opening of HPSI, LPSI and/or CSS pumps would be annunciated in the controlroom, Low LPSIpur.sp discharge flow (2FT 5091) would be Indicated.
(Vp)
Discharge pressure Indication (2PT 5092) would be much lower than expected.
Although considered reasonable to conclude that this line break could be detected and mitigated by isolation of the failed line segment, no credit is taken due to the limited timeframe available for detection.
LP11 Conseevrxn& ament tPSI C-08 For line breaks in the connecting header between the LPSI pumps into the general sies of elevation 317' 0*, in conjunction with a LOCA demand, the foHowing slarms and indications would be encountered. SpecificsHy; Auxiliary building sump high level would be annunciated.
Weste drain tank nigh level would be annunciated.
Low LPSIpump discharge flow (2FT 5091) would be Indicated.
Discharge prescure indication (2PT 5092) would be much lower than expected.
Although considered reasonable to conclude th2t this line break could be detected and mitigated by isolation of the failed line segment, no credit is taken due to the limited timeframe available for detection.
(Pj]_Qonstg.vence Senment LPSI-C 13 For line breaks downstream of the 335' 0* elevation and upstream of elevation 360'-O'in the LPSI pump discharge piping, in conjunction with a LOCA demand, the following
(,)
unexpected alcrms and indications would be encountered. Specifically;
- t/
ABB Combustion Engineering Nuclear Operations
ABB Calculation No. A PENG CALC 016, Rev. 00 Page 26 of 70 AuxHiery building sump high level would be annunciated.
Waste drain tank high level would be annunciated.
- Dependent on break location, low pump discharge flow (2FT 5091) would be annunciated.
Discharge pressure indication (2PT 5092) would be much lower than expected.
Although considered reasonable to conclude that this line break could be detected and mitigated by isolation of the faHed line segment, no credit is taken due to the fitnited timeframe avaHable for detection.
LPSI ConMLmnce Segments LPSI C 14A /148 /14C /14D for a one break downstream of the elevation 3M'O' floor penetration and upstream of conteinment penetration 2P 15/2P 10/2P 29/2P 24 during a LOCA demand, the foHowing uncxnected alarms andindications would be encountered. SpecificaHy; AuxHiery building sump high level would be annunciated.
Waste drain tank high level would be annunciated.
Although considered reasonable to conclude that this line break could be detected and mitigated by isolation of the failed line segment, no credit is taken based on the limited indications available and the timeframe available for detection.
LPSI ConseQLoence Seaments LPSI-C 15A / ISE / ISC / ISD for a line break downstream of containment penetration 2P 15/2P 10'2P 29/2P-24 and upstream of 2St 14A/148/14C/140 during a LOCA demand, no unexpected alarms and indications would be encountered. As a result it is considered reasonable to conclude that this line break would not be detected and mitigated by isolation of the failed line segment.
LPSI Consoavence Legment LPSI-C 16 For line breaks during a shutdown cooling demand in the return header from the shutdown cooling heat exchangers, downstream of 2SI 5A/2SI 58 to upstream of floor elevation 335'-
O'into the generalarea of elevation 317'-O', the foHowing sistms andindications would be encountered. SpecificaHy; Auxiliary building sump high level would be annunciated.
Waste drain tank high level would be annunciated.
Auxiliary building high radiation would potentially be annunciated.
Unexpected motor amp reading (erratic or low from 2II 5018A).
Decreasing RCS inventory which is specifically monitored when isolation valves opened.
Decreasing RCS pressure.
Low LPSI pump discharge flow (2FT 5091) would be indicated when flow to the heat exchangers is established.
Discharge pressure indication (2PT 5092) would be much lower than expected.
O ABB Cembustion Engineering Nuclear Operations
ABB (h
Calculation No. A PENG CALC 016, llev. 00 Page 27 of 70 ft is therefore considered reasonable to conclude that this line break would be detected and mitigated byisolation of the failedline segment.
LPSI Congsgyp_nge Segment LPSI-C 17 for line breaks downstream of the 335'0* elevation and upstream of shutdown cooling heat exchanger outlet valves 2CV 5093 and 2515093 3 in the LPSI pump discharge piping, In conjunction with a shutdown cooling demand, the following unexpected alarms and Indications wouldbe encountered. Specifically; low pump discharge flow (2FT 5031) would be annunciated.
e Auxiliary bu:Iding sump high level would be annunciated.
Auxiliary building high radiation would potentially be annunciated.
Waste drain tank high level would be annunciated.
Unexpected motor amp reading (erratic or low from 2115018Al.
Decressing RCS Inventory which lu specifically monitored when isolation valves opened.
Decreasing RCS pressure.
Olscharge pressure Indication (2PT 50921 much lower than expected.
Because of the focused attention associated with the initiation and maintenance of shutdown cooling, the various indications and alarms associated with this event would cause an immediate response. it is therefore considered reasonable to conclude that this
,q line break would be detected and mitigated by isolation of the failedline segment.
V V
ABB Combustion Engineering Nuclear Operations
A1B Calculation No. A PENG. CALC-016, Flev. 00 Page 28 of 70 4.3 CONSEQUENCEIDENTIFICATION The consequence summary assessment is provided in tabular form in this section.
Simplified schematics are provided in Figures 2 through 9 to lHustrate the boundaries for each of the LPSVSDC consequences. Dotted lines are used to identify the boundaries for each consequence. Major LPSI equipment along with floor and wsH penetrations are shown on these figures for ease ofidentification. Table 2 summarises the consequence evaluation for the LPSVSDC system, given successfulisolation where applicable. This table contains the foHowing information for each of the consequences iderntified:
Consequence ID A unique number assigned to the consequence Boundary The figure number that (Hustrates the boundaries for the consequence Description A brief description of the effects of the consequence DB Event Category The category of design tasis initiating event that the LPSVSDC system is designed to mitigate, based on Table 1 of Appendix A Direct Effects The immediate or direct effects caused by a failure of the pipe segment Spatial Effects The indirect effects caused by a failure of the pipe segment impact Group The impact of a pipe segment failure on the LPSVSDC and other mitigating system (s) or train (s)
Available Trains The number of trains available for performing the intended design functions of the LPSI and SDC systems Consequence Cat.
The assigned consequence category based on the application of the methodology provided in the EPRI procedure (Reference 9.1)
For a herge LOCA where the faded segment remains unisolated, the LPSI system would be lost die to flow diversion. Because of the short duration of LPSI foHowing a large LOCA, operator actions to isolate the failed segment are considered to be unlikely. Hence, the operators are not credited as an equivalent backup tro!n for mitigating the failed segment foHowing a large LOCA. When sufficient time is available during shutdown operation, the operators ability to isolate the failed segment is credited as an equivalent backup train.
The bases and justifications for each of the assigned consequences are documented in Appendix A.
The ISIS (Reference 9.2) software was used cs a tool to prepare the documentation in this appendix. The documentation of the spatial effects are currently based on a review of the intern.s Flood Screening Study (Reference 9.t4) and Pbnt Design Drawings, and the womdown that was conducted for the LPSUSDC system. The walkdown captured subtle interactions which could not be readily identified using the screening study or the plant drawings. Observations from the walkdown are factoredinto the consequence evaluation.
Table 3 presents the LPSUSDC consequences, their corresponding figure numbers and Isometric Drawings.
4.4 SHUTDOWN OPERATION AND EXTERNAL EVENTS Ehutdown Oneration 7;e consequence ev&luation is an assessment assuming the plant is at-power. GeneraHy, the at power plant configurat.'on is assumed to present the greatest risk for piping failures ABB Combustion Engineering Nuclear Operations
ABB Calculation No. A PE' G CALC 016, Rev. 00
(
N
- )
Page 29 of 70 since tl.e plant requires immediate response to satisfy reactivity control, heat removal, and inventory control. By satisfying these safety functions, the plant will be shut down and maintained in a stable state. At power, the plant is critical, and is at higher pressure and temperature in comparison to shutdown operation. The current version of the methodology provides no guidance on consequence evaluation during shutdown operation. This limitation is assessed herein to gain some level of confidence that the consequence ranking during shutdown would not be more limiting, Pipe segments that are already ranked as 'HIGH' consequences from the evaluation at.
power need not be evaluated for shutdown. Those that are already ' MEDIUM" require some confidence that 'HIGH' would not occur due to shutdown configurations. However, the ' LOW" consequences for power operation requires more confidence that a 'HIGH' would not occur and some confidence that a 'A1EDIUM' consequence would not occur.
Recognizing this, a review and comparison of system consequence results for power operation versus potentialconsequence during shutdown operation was conducted.
The following generalassumptions or inputs were considered in performing the consequence assessment of the shutdown cooling (SDC) system.
During most of shutdown operation EFW and ECCS are manually actuated if not already operating and is required.
EFW is automatically actuated until cold shutdown (Afode 5, < 200*F) and ECCS is automatically actuated until RCS pressure is less than 361 psia in Afode 4 (?t02.010, Rev 30 *Piant Cooldown'). Risk p;}
management guidelines for shutdown operation in place and outage management philosophy at ANO-2 provide assuunces that loss of SDC will be detected and mitigated (e.g., 2203.23, Rev 9 ' Loss of Shutdown Cooling).
Unavailability of mitigating systems is higher due to planned maintenance during outages.
However, guidelines and procedures are in place to assure sufficient redundancy and to account fu higher risk configurations (i.e., midloop) which requires additional compensation provisions and or contingencies (i.e.,
2R12 Shutdown Operations Protection Plan, Afay 8,1997).
For the majority of class I piping, the exposure time associated with operation in a shutdown configuration is approximately 10% of the year. Also, the operating conditions are much less severe than during power operation. The frequency of being in a more risk significant configuration could be even lower depending on the system and function being evaluated. Operation of the SDC system is an important exception.
For the majority of class 2 piping, the frequency of challenging important mitigating systems is judged to be on the same order of magnitude or lower. Operation of the SDC system is an important exception.
The reactor is shutdown, depressurized, and decay heat is lower than for at power operation. The reactivity control function is not a concern because the rods are inserted. Re criticality during shutdown is unlikely and not judged to effect the presem ranking. The inventory makeup function (safety injection) is conside6ed the
()'
most important function during shutdown, given a class 1 or 2 pipe break occurs G
during shutdown causing loss of SDC.
ABB Combustion Engineering Nuclear Operations
ABB Calculation No. A.PENG. CALC 01C, Rev. 00 Page 30 of 70 During shutdown, the RCS and connected piping are not pressurized nor at high e
temperatures. Since the SDC system is sligned to the RCS during a significant portion of the duration for shutdown operatic,n in order to accomplish and *naintain shutdown cooling function, thh system is evaluatedin more deteH.
Because of low decay hest loads curing shutdown operation, the time for recovery of shutdown cooling or inventory makeup is usuaHy longer even though equipment may require manual actuation and may also be in maintenance LOCAs during shutdown operation (consideredless likely due to reduced pressure and temperature) would exhibit much less severe environmental conditions (e.g., hot or warm water versus steam) untH decay heat starts to hestup reactor after Icss of SDC.
LPSI pipe segments that are challenged during at powee operation due to a large LOCA 1
demand are ranked as 'HIGH' or
- MEDIUM' consequence. The segments that are already j
ranked as 'HIGH' consequence need no further evaluation, and are considered to envelop
\\
the consequences during shutdown operation.
Those that are ' MEDIUM" (i.e.,
Consequence Segments LPSI C 00 /11) can be detected and isolated prior to recirculation to prevent continuad diversion. Thus, the focus is on piping not chauenged in support of the LPSI function.
If a *HIGH" consequence should result from this review, other
- MEDIUM
- consequences determined for the LPSI function during st power operation would i
have to be evaluated further.
No 'HIGH' consequencs segments were identified for shutdown operation.
O The foHowing sumanarizes the assessment of the line segments for the SDC system.
4.4.1 The consequence of a segment failure ISOC suction piping) upstream of 2CV 5084-1 is already renked as "HIGH' due to a large LOCA during at-power operation, and is considered to envelop the shutdown risk.
4.4.2 SDC suction piping downstream of 2CV 50841 and upstream of 2CV 5086 2 inside conteir' ment is assigned a " MEDIUM' consequence during at power operation because passive failure of the normaHy closed 2CV 5084-1 is necessary to challenge this piping. Failure of this piping during shutdown operation is assessed below by considering two states of plant shutdown operation. The first one is associated with initial alignment of SDC because this is when pipe failure would most likely occur due to the demand chauenge. The second considers tne case where the plant is in cold shutdown and depicssurized because pipe rupture can be assumed to be less likely at this point given the successful demand.
For the first state, SDC is aligned to continue the removal of decay heat from the RCS. When demanded, SDC is manusHy aligned flocaHy) after RCS pressure and temperature decrease below 275 psla and 245 'F, respectively, per step 8.15 of the plant cooldown procedure (2102.010, Plant Cooldown, Rev. 301.
During this 1
important evolu, ion, operators are alerted to ensuring proper alignment and inventory per thL procedure. During the initial alignment and for some time thereafter, it is assumed that the operators are c.vefuoy monitoring RCS inventory and would detect a*sy unexpected decrease in RCS inventory dudng this important change'in plant configuration. Also, steam generators should stHI be available or recoverable since they were being used until the plant was aligned for SDC operation (assuming 2CV-ABB Combustion Engineering Nuclear Operations
ABB f3 Calculation No. A PENG CALC 016, Rev. 00 Page 31 of 70 50841 is isolated quick enough). For the case where the operators successfuHy isolate the break early, the EFW and AFW systems are available for decay hlat removal, thus there are more than two backup trains for mitigation. Therefore, the resulting consequence is
- LOW'(Table 3.4. of Reference 9.1). For the case where 2CV 50841 faHs to close ieither due to equipment or human), the reactor could drain to the Invert of the hot and cold leg nozzles. The loss of shutdown cooling procedure (2203.029 ' Loss of Shutdown Cooling
- Rev 9) Identifies entry conditions, including alarms, and directs the operators to stop the SDC pump and isolate SDC at step 4, if RCS level is decreasing rapidly.
The procedure exits to lower mode functional recovery (2202.011
- Lower hiode functional Recovery' Rev. 0) for both S!)C isolation success (step 4.0) and faHure (step 4.G).
This procedure repeats steps to secure and Isolate SDC, as weH ss other attempts to locate and Isolate inventory loss, it also provides steps for restoring level and/or heat removal depending on the configuration. Because of the avoHable indications and alarms, the
\\
capability of the operators to isolate the failed segment is treated as an equivalent backup train. The capability to provide inventory makeup after failing to isolate (e.g., before boH off and core damage) is also treated as another equivalent backup train. Thus there are two equivalent backup trains, and the resulting consequence is "A4EDIUA1'(Table 3.4 of Reference 9.1).
For the second state of plant shutdown operation (i.e., RCS at refueling conditions),
it is assumed that steam generators are not ovaHable or the RCS is vented or open (vessel head is off). Since SDC is lost due to the boesk, inventory inskeup is the
(
)
primary concem (f.
e., makeup must at least satisfy decay heat boH off).
By G
isolating the failed segment, the time to provide makeup is sacreased significantly, especiaHy if the refueling cavity is fuH. During n"d loop operation, the operators are also carefvHy monitoring RCS Inventory due o the importance of this shutdown configuration, but isolation is almost irrelevant because RCS levelis s!:eady near the invert of the reactor nozzles. Procedures require :edundant inventory makeup during mid-loop (2R12 Shutdown Operations Protection Plan, Afay 8,1997). There would be at least twc equivalent backup trains for mitigation and the resulting consequence is 'AfEDIUAf * (Tabis 3.4 of Reference 9.1).
Based on the above, a *A4EDIUA1' consequence ranking is assigned for this SDC pipe segment. This ranking is bounding for the various plant shutdown states and associated cases that were considered.
4.4.3 During power operation, the consequence of a segment failure (SDC suction piping) downstreem of 2CV 5086 2 inside containment is a " LOW" because two passive valve failures of the normaHy closed valves 2CV 50841 and 2CV 5086 2 are necessary to chauenge ciping. During shutdown operation, the consequence is
- A1EDIUA4' based on the rationalprovided in Section 4.4.2. Although there are two redundant valves for isolation, operator error is considered to be the dominant contributor in failing to isolate the segment. Hence, only one equivalent backup train is assumed.
4.4.4 During power operation, the resultir.g consequence of the segment failure ISOC suction piping) downstream of containment penetration 2P27 and upstream of 2CV-O 50381 is 'AfEDIUAf'. This is based on passive failure of the two barriers causing
!\\&
the containment to be bypassed and the potential consequences of a LOCA outside ABB Combustion Engineering Nuclear Operations
ABB Calculation No. A PENG CALC 016, Rev. 00 Page 32 of 70 containment with containment bypass. During shutdown operation, the resulting consequence of the segment foHure is 'MEDILIM' for similar reasons described Section 4.4.2. The LOCA environmentalconditions (failure of two MOVs to close or human error) outsido containment during shutdown are }ueled to be much less severe than the power operation case with regard to potentialimpact on mitigating systems. The fact that water is being lost outside containment is stiu a contem.
StHI, foHure to close two redundant MOVs is considered a
- Medium
- consequence.
4.4.5 Because there are three normsHy closed vs:,es in SDC suction pip.hg downstream of 2CV 50381 and outside containment, the consequence of the segment foHure Is a
" LOW' due to a potential LOCA.
During shutdown operation, the resulting consequence is ' MEDIUM
- for the simHar reasons given in Section 4.4.4.
4.4.6 Piping downstream of manual valves 2SI-4A and 2SI-49 and upstream of manual valves 2St 5A and 25158 is also only used during SDC operation. The resultbg consecuence is ' LOW" during power operation.
These lines provide redundant paths through the two heat exchangers. Isolation of a break in one of these lines would aHow SDC to be recovered with the other heat exchanger path. However, a
- MEDIUM
- consequence is stiH essumed for this segment.
4.4.7 Piping downstream of manual valves 2SI 5A and 25158 and upstream of throttle valve 2CV 5093is also only used during SDC operation. The resulting consequence is ' LOW" during power operation because this piping is not used to support or accomplish LPSI. A failure here would lead to loss of SDC because the piping is common to both LPSIpumps. A ' MEDIUM' consequence is assigned based on the above.
Based on the consequence assessment provided above for the LPSI and SDC pipe tegments, the at power consequence ranking of piping segments used to support both the LPSI and SDC functions envelops the consequence ranking during shutdown operation. The piping segments used to provide only the SDC function is bounded by a
- MEDIUM" consequence.
Extemal Events Although external events are not ads ssed in the current version of the methodology (Reference 9.1), the potential importance of piping faHures during exismal event is also considered.
The ANO 2 IPEEE was reviewed to determine whether extemal Initiating events, with their potential common cause impscts on mitigating systems, could affect consequence ranking. This information, along with information from other extemal event PRAs, is considered to derive insights and confidence that consequence ranking is not more significant during an extemal event. The following summarises the review for each of !u major horards (seismic, fire and others).
Saismic Challenges. The potential effects of seismic initiatong events on consequence ranking is assessed by considering the frequency of chaHenging plant mitigating systems i
and the potentialimpact on the existing consequence ranking. The foHowing summarises this assessment:
GeneraHy, the LPSI/ SOC piping considered 'n this evaluation has a seismic fragility capacity much greater than the 0.3p screto ing value and is not considered likely to fail during a seismic event.
l ABB Combustion Engineering Nuclear Operations 1
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Calculation No. A PENG CALC 016, Rev. 00 (O
Page 33 of 70 i
With regard to the impact on mitigating systems, the most likely scenario is selsmic inducedloss of offsite power. Based on a typical fragility forloss of offsite power, a High Confidence of low Probability of Falture (HCLPF) of 0.lg (Reference 9.19) is assumed. Combining this fragility with the ceismic harards developed for the ANO site (References 9.20 and 9.21), the frequency of seismic induced loss of offsite power is less than 1.0E 4 per year.
Considering the scenario where an mduced LOCA and non selsmic failure of one d;esel generator occur, there is only one LPSI pump avaHable. In response to the event, the LPSI piping in the common header is sssumed to faH when demanded, thus there are no backup trains for mitigation. Assuming a probabHity of 0.1 for faHure of the diesel generator and an *aH year
- exposure time, the Conditional Core Damage Probability (CCDP) for this scenario is less than 1.0E 05. Therefore, the resulting censequence is ' MEDIUM *.
For the nenario where on Induced LOCA occurs and both diesel generators are available, both LPSIpumps are inittany available. Howevst, in response to the event, the LPSI p.,Ing in the common header is assumed to faH, thus there are no backup trains avaHaNe for mitigatlun. With no tackup trains and an JH year
- exposure time, the CC 1P is less than 1.0E 4.
The resulting consequcnce would also be
' MEDIUM *.
(~'
Based on the above, the consequence ranking for LPSI during a selsmic event is enveloped by the at power consequence ranking.
Fire Challenges The ANO 2 IPEEE indicates that the fire core damage frequency is domi.wted by fires initiated outside the containment. Similar to a seismic event, fires do not impact reactivity control :r cause a LOCA. The resulting consequence is expected to be
't.O W. Since the at power consequence ranking is already 'HIGH" or
- MEDIUM *, the reau'iong consequence during a fire would not be of greater significance.
Other External Challenges Other hazards were screened in ths ANO 2 IPEEE and are assumed to have little or no risk significant imposs' on LPSI.
o ABB Combustion Engineering Nuclear Ooerations
SSS 9% WW CalCulatiort No. A-PENG-CALC-016. Rev. 00 Page 34 of 70 Table 2 LPSI/SDC Consequence Assessment Summary - Successfulisolation Consequence Boundary Descreten DB Event Drect Effec:s
.SipermalEfrects krpect Groso I Available Mrtigenny Consequence D
Ceregory l
Trains Categewy LPSI-C-01 Roure 2 Loss of reector coolerrt vue N
LOCA irwoetmg event None loss of shutdows AR four ECCS HKV4 SDC suction Ene ocews during coc6ty end het
& kops power everetion due to e Ene leg irpechan break LPSICD2 Rgure 2 Loss of reactor coolent occws IV Shutdowrr LOCA None loss of shutdowrr Two nowces for MEDIUM during imid kop operation due initionierg event and evoEng RCS metone es to SDC Ene break inside bss of shutdown requred by the
.v.a-.. ; ^
coorung ISOC1 Shutdown sucDon for LPSI Cperations pornes
- L, even LPSt-C-03 Rgwe3 Loss of reector coolant occws IV Shutdowrr LOCA Spreyirsgorjet loss of shutdown Two eeurces for MEDR,lM diurmy unid-bop operation due husting event end
- 44-,.....: of cooEng RCS menonp en to e break ire itse SDC suction kss of SDCsuction venres 2CV-5015-requeod by she Kne irr the Upper Scuth Pfping for LPSIpurrps
- 1. ICV-5016-2.
Shutdown and Penetration Room.
2CV 5017-1, 2CV-Operesums 6612-1 M; ;L,Meer LPSTC-04 Rgure 3 Loss of reactor cootent occurs IV Shutdown LOCA Spreymg urjet loss of shusdown Two sosecee for MEOKIM L
durirsg mid4aop operethm due initietmg event end 6 4.,
..u: of cor: cog RCS oneheng es 4
to e breek ire the SDC suction icss of 3DC sucDorr LPSL'SDC vehres reguredby she Ene irr Lower South PTping for LPSIpurrys 2CV-5091 and shutdowrr Penetration erre.
2CV-SO93, eruf Cpersoons HPSIorifice bypess
.", v;x ^L, Men sehre 2CV-6103-1 LPSTC-05 Rgure 4 Loss of reactor cooient occurs N
Shutdown LOCA Rooding or loss of shutdows Tew se aces for MEDR,rM dbrirsg niid doop operseion die irifiating event and spreye*g of coodng and dreirr RCS tristecy as to e Gree break in ECCS purry kss of SDC suction equpment irr ECCS
- A*of she ECCS requredby she roons*A*.
for LPSIpaarps purro roorn "A*
purrve Shutdown Operecons
- w. Men LPSI-CM Roure 4 Loss of trein *A*of ECCS IV Drversen of LPSI Rooding or loss of one trairr Orne redundant trearr MEDIUM purrps occw due to e Ene fkw to ECCSpur, spreymg of of HPSf, LPSt and for each of the breat ar, ECCSpur, room A-room A*
equpment kr ECCS CS eWected systems fosaw,ng a LOCa m roorrr A
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O EEE HWW Calculatiott No. A-PENG-CALC-016. Rev. 00 i
Page 35 of 70 Table 2 (Cont'dJ LPSUSDC Consecuence Assessment Surninary - Successfulisolation Consequence Boumfory Descrption DB Event Dweet Effects Spetief Ofects irrvect Graana AvenToble A6tigeeng Consequence JO Ceregory Treme Category LPSI C-07 Rgure 4 Loss of both LPS!purms.
N Doversson of LPS1fkrw Rooding or loss of LPSI None for LPSI FOGN HPSIpurm 2P-89A and CS to ECCSpurre roons spreymg of systern erut one purrp 2P35A acews dUe to "A
- egenprnent M ECCS eek of HPSt and a Ene break k ECCSpurm paarp room *A*
CS roorrr *A* fon6wkg a tOCA LPSFC-06 Roure 4 Loss of both LPSIpurms IV Drversson of LPSIflow Spreymg orjet loss of LPSI None HIGH occws dUe to e Jine break irr to Tendorr GeFory ammgement of systern Tendon GeRe*y Access eroe Access eroe wake 2CV-140(> 1 foabsvng a LOCA LPSFC-10 Rgure 5 Loss of reactor coolant IV Shutdown LOCA FJ6cdVng or loss of shueduwrr Tamo sources for MEDettht occurs dshgtridloop irvrimeing event and spreymg of cooEng and erek RCS onedeegr as operatiorr dUe to e Ene breet has of EDC sucean equpment ire ECCS
for LPSIpanw' psam roorrr "B*
purrps Shutdown Operebone r,; ;w pron LPSFC-11 Roure 5 Loss of trairr ~B*of ECCS W
Dversson of LPSI now fkoding or loss of one trak of Che redundant steer MEDetint purrps occw dUe to e Ene to ECCSpurm room spreymg of HPSI. LPSIand CS for oech of the breek k ECCSpeanp voorn "B'
equpment tre ECCS effected erstome
~3* foRowkg a LOCA peam room *B*
LPSFC-12 Rgure 5 Loss of both LPS!purrps.
IV Diversio*s of LPSt fTow Rooding or loss of LPSI Norn for LPS!
MIGH HPS! parry 2P-898 eruf CS to ECCS pasrp roorn spreyes of system and one purry 2P-358 occurs due no
- B*
w irrECCS amirs of HPSIand o &ne break in ECCS purrp purrp room *B" CS roorn ~B* fonowmg a LOCA LPSFC-13 Rgure 6 Loss of LPSIoecurs due to e IV Overssort of LPSInow Spreymg orjet loss of LPst Nana H!GH Tme bres& in the lower Soesth to the generet access irrvegemorrt of system Pipeg Penetration eres eree LPSVSOC veNes foWowng a LOCA.
2CV-60st cruf 2Cv-SO93. emf HPSIor# ice bypest veke 2CY5 t03-T ABD Combustion Engineering Nuclear Operations
R,,E MEE i
Calculation NO. A-PENG-CALC-016, Rev. 00 Page 36 of 70 Table 2 (Cont'dJ LPSUSDC Consequence Assessment Summary - Successfulisolation Consequence Boundsry Desception DB Event Deect Effects Spatiet Effects
>pect Grone Avesetk l Consequence D
Category Meyermg Treke Coregnry LPS!C14A Rgures 7 Loss of LPSIoccurs doe to e W
Drversion of LPSIflow Sprerng orjet loss of LPStand None for LPSI HIGH
&8 Kee break irr irpectionpth to to the generetaccess i uv.a.= of UW d*Uwery of EFW to cokileg 2P32A in stw D>per eree dstnibutson vehes to SG2E24A Soutn Pipeg Penetra tion aren SG 2E24A foBowmg e LOCA.
LPSIC-148 Rgure 8 Loss of LPSIoccurs due to a N
Drversen of LPS! flow Spreymg orjet loss of LPSIand Norw for LPST HG Ene break irs krecDon path to to the generet access ervesgement of UW deEveryof EFWM cokileg 2P32B in the Upper erne dstrbutson ve%s to SG 2E24A Snuth 1\\pkg Penetretion aree SG 2E24A fonowing e iOCA.
LPSIC-14C Roure 7 Loss of LPSIocews due to e W
Diversion of LPSt thw Spreyng orjet loss of LPStand None for LPSt HM s
Tme break k drpochonpath to to the generet access irrve,gemorrt of EFW deEvery of EFW to centleg 2P32C k the b>per eree dstr@artion verres no 33 2E-24A Scarth P(ping Penetretion aree SG 2E24A f& u ;eLOCA.
LPSIC14D Rgure 8 Loss of LPSIoccurs doe to a N
Doversen of LPSIflow Spraying orjet loss of LPSIarnt NoneIbrLPS!
HIGH nne breet irr iry'ection path to to the generet access orpergement of EFW de6very of UW to eroe dstrburnon vakes to SG 2E-24A cokf leg 2P32D in the L&per Soards Prphsg Penettetion aree SG 2E24A foRowmg e LOCA.
LPSI-C-15A Rg.ars 7 Loss of LPSIocews die to a N
Deversson of LPSIthw None loss of LPSIsystem None HM break ire irpecDon path to conf to contemenent sunp reg 2P22A M contenwnarrt fone--N e iOCA.
LPSFC158 Roure 8 Loss of LPS!occws doe to e W
Diversion of LPSI fkw None loss of LPS1 system None HG breet in irpection path to coAf to conteevnent survp leg 2P22B in conteswnent fogowmg e LOCA.
LPSI-C-15C Rgure 7 Loss of LPSIocews dro to e N
Diversen of LPSIfkw None loss of LPSIsystern None HIGH breet in irweefsors path to co&f to conteevnent osary leg 2P22Cits cor:temrnent fA --.;eLOCA.
l LPS!C15D Rgure 8 Loss of LPSIrecws doe to e W
Deversen of LPS!tkw None loss of LPStsystern None Hm breaa k irrection path to coks to contenwnent sunp kg 2P22D k cont.;ewnerrt fonowmg e iOCA.
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s C\\
C]'
'd d
NSS nWW Calculation No. A-PENG-CALC-016. Rev. 00 Page 37 of 70 Table 2 (Cont'd.:
LPSUSDC Consequence Assessment Summary - SuccessM isolation Consequence soundary Descr&eion DB Event Deect meets Spesef M octr krp ect Gro w AveneNe Consequence D
Catogewy nnvegermg Treew Cerespry LPSFC-16 Ronae 9 Loss of reactor coceent IV Shutdown LOCA
&ne loss of shutdown Two sources for httlNUnt i
occurs during r=mid kop irw*tiating event
'cooEng RCS onehener se operez. shutdown coosna renmared by the ocews due to e Gne breat k Shuedown she Tendon Genery Access C>ernoons ere.
Meteernm %
LPSFC 17 Roure 6 Lots of reactor c&st IV Shnordown LOCA Spreymg erjet loss of shuedowrr Teen scarcos for AGEINUni ocews durMg rnid kop eveeeng event 66
..,:of cooEng RCS onenese er coereeon due so breea h the Lpstsr. verres 2CV-rieered by the SDC heet exchanger retnan 5091 and 2CV-5093 Snartdown Kne h the lower Soorth Mphy and NPST orrTre Cherwoone
.".... ; L. eree.
bypass ve/re 2CV-Mor h, Men
$103-1 i
I ABB Combustion Engineering Nuclear Operations i
ABB Calculation No. A PENG CALC-016, Rev. 00 Page 38 of 70 Table 3 LPSI/SDC Consequence. Figures and Isometric Orswings Consequence ID figure isometric Orswings Number IPSI-C 01 2
2CCA 251 LPSI C-02 2
2CCA 251 2CCA 25 2 2GCB-5 4 2GCB 781
- f. PSI-C-03 3
2GCB.5 1 LP9-C-04 3
2GCB 51 2GCO 5 3 LPSI C 05
-4 2GCB 5 3 LPSI C-OS 4
2GCB11 _2GCB31 2GCB 5081 2DCB 5041 lPSI C-07 4
2GCB 31 20CB 71 LPSI-C CB 4
2GCB 31 2GCB 3 2 LPSI C 10 5
2GCB 51 2GCB 5 2 LPSI C 11 5
2GCB21 20C8 3 3 2GCB 5091 2DCB-505 1 iPS! C 12 5
2GCB 3 3 2GCB 3 2 LPSI C 13 6
2GCB71 2GCB 7 2 2GCB B 1 LPSI C 14A 7
2GCB 7 2 2CCB 31 LPSI C 14C 7
2GCB 7 2 2CCB&1 LPSI C 14D B
2GCB 7-2 2CCB51 LPSI C 15A 7
2CCB-4 2 LPSI-C 15h B
2CCB 3 2 LPSI-C 15C 7
2CCB 5 2 LPSI-C 16 9
2GCBB1 2GCB 171 LPSI C 17 6
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- 6. 0 DEGRADA TION MECHANISMS EVALUA TION The purpose of this section is to identify the degradation mechanisms that can be present in the piping within the selected system boundaries for the ANO-2 LPSVSDC system, as described in Section 3.2 of this report. The conditions considered in this evaluation are:
design characteristics, fabrication practices, operating conditions, and service experience.
The degradation mechanisms to be identified (Reference 9.1) are:
Thermal Fatigue (TF)
Thermal Stratification, Cycling, and Striping (TASCS)
Thermal Transients (TT)
Stress Corrosion Cracking (SCC)
Intergranular Stress Corrosion Cracking (IGSCC)
Transgranular Stress Corrosion Cracking (TGSCC)
- Extemal Chloride Stress Cctrosion Cracking (ECSCC)
- Primary Water Stress Corrosion Cracking (PWSCC)
Localized Corrosion (LC)
Microbiologically influenced Corrosion (MIC)
Pitting (PIT)
Crevice Corrosion (CC)
Flow Sensitive (FS)
- Erosion-Cavitation (E C)
(%
Flow Accelerated Corrosion (FAC) s"]
In performing this evaluation, some basic bputs 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: thermal fatigue.
Using ISIS (Reference 9.2), two damage groups (DM groups) were identified as LPSI-T and LPSI N and are defined in Table 4 below. These DM groups results in two failure potential categories: "Small Leak" and *None".
The FMECA - Degradation Mechanisms for each segment and each element are presented in Appendix B.
Table 4 Damage Groups Damage Damage Mechanisms FeRure Group ThermalFatigue stress Corrosoon Crackine locaksed Corrosion Row senaltive Potential 83 TAsCS TT IGsCC TGsCC ECsCC PWsCC MC PIT CC E-C FAC Category LPsi-T Yes No No No No No No No No No No sman Leak LPsi-N No No No No No No No No No No No None
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ABB Combustion Engineering Nuclear Operations
ABB Calculation No. A PENG-CALC-016, Rev 00 Page 48 of 70 6.1 OAMAGE GROUPS 5.1.1 DM GROUP: LPSIT The LPSI T damage group is considered subject to thermal fatigue due to the potential for thermal stratification (TASCS) in the SDC system It includes the welds in the shutdown cooling line 2CCA 2514", downstream of I... hot leg nozzle up to the verticalsection upstream of motor operated valve 2CV 5084-1 (Figure 2).
5.1.2 DM GROUP: LPSI-N The LPSI-N damage group is not considered susceptible to any damage mechanism, and includes the entire Low Pressure Safety injection (LPSI) system as defined in Section 3.2, and the remainder of the SDC system not included in group LPSI T (See Figures 2 9).
5.2 DEGRADA TION MECHANISM CRITERIA AND IDENTIFICA TION The degradation mechanisms and criteria assessed are presentedin Table 5.
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A ItIt F1IMF Calculation No. A PENG CALC 016, Rev, 00
<J Page 49 of 70 Table 3 Degradation Mechanism Criteria and Sesceptible Regions f*"
Criteria Susceptible Regions TF TAJCS
-nps > 1 inch, and nozzles, branch pipe
-pipe segment has a slope < 43*from horizontal (includes elbow or connections, safe ends, tee into a verticalpipe), and welds, heat alTected
-potential existsfor lowpow in a pipe section connected to a zones (HAZ), base component allowing mixing ofhot and coldpuids, or metal, and regions of potential existsfor leakagepaw past a valve (i.e., in-leakage, out-stress concentration leakage, cross-leakage) allowing mixing ofhot andcoldpuids, 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 turbulentpenetration in branch pipe connected to header piping containing hotpuid with high turbulentpow, and
-calculated or measured AT > $0*F, and
-Richardson number > 40 TT
-operating temperature > 270 Ffor stainless steel, or operating temperature > 220*Ffor carbon steel, and (q
-potentialfor relatively rapid temperature changes including
,")
coldpuid injection into hot pipe segment, or hotpuid injection into coldpipe segment, and AT
> 200*Ffor stainless steel, or i
AT
> 130*Ffor carbon steel, or AT
> AT allowable (applicable to both stainless and carbon)
-evaluatedin accordance with existing plantIGSCCpn. gram per austenitic stainless steel (BWR)
NRC Generic Letter 88-0) welds andHAZ 1GSCC
-operating temperature > 200*F, and (PWR)
-susceptible material (carbon content h 0.033%), and
-tensile stress (including residual stress) is present, and
-oxygen or oxidi:ing species are present OR
-operating temperature < 200*F, the attributes above apply, and
-initiating contaminants (e.g., thiosulfate,fuoride, chloride) are also required to be present TGSCC
-operating temperature > 130*F, and austentric stainless steel
-te,rsile stress (including residual stress) is present, and base metal, welds, and
-halides (e.g., fuoride, chioride) are present, or HAZ caustic (NaOH)ispresent, and
-oxygen or oxidt:ing species are present (only required to be present in conjunction wihalides, not required w/ caustic)
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Calculation No. A PENG CALC 016, Rev, 00 Page 49 of 70 Table 3 Degradation Mechanism Criteria andSusceptible Regions
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Criteria Susceptible Regions TF TASCS
-nps > 1 inch, and nonles, branch pipe
-pipe segment has a slope < 43*from horizontal (includes elbow or connections, safe ends, tee into a verticalpipe), and welds, heat afected
-potential existsfor lowfow in a pipe section connected to a zones (HAZ), base component allowing mixing ofhot and coldfuids, or metal, and regions of potential existsfor leakagepow past a valve (i.e., in-leakage, out-stress concentration leakage, cross-leakage) allowing mixing ofhot and coldfulds, or potential existsfor convection heating in dead-endedpipe sections connected to a source ofhotfuld, or potential existsfor two phase (steam / water) pow, or potential existsfor turbulentpenetration in branch pipe connected to header piping containing hotfuld with high turbulentfow, and
-calculated or measured AT > $0*F, and
-Richardson number > 4.0 TT
-operating temperature > 270*Ffor stainless steel, or operating temperature > 220*Ffor carbon steel, and p
-potentialfor relatively rapid temperature changes including coldpuidinjection into hot pipe segment, or hotpuid injection into coldpipe segment, and
- l AT: > 200*Ffor stainless steel, or lAT
> 130*Ffor carbon steel, or l AT
> ATallowable (applicable to both stainless andcarbon)
-evaluatedin accordance with existingplantIGSCCprogram per austenitic stainless steel (BHR)
NRC Generic L.etter 88-01 welds andHAZ IGSCC, -operating temperature > 200*F, and (PWR)
-susceptible material (carbon content k 0.033%), and
- ensile stress (includmg residual stress) is present, and
-exygen or oxidning species are present OR
-operating temperature < 200*F, the attributes above apply, and
-initiating contaminants (e.g., thiosulfate,fuoride, chloride) are also o equired to be present TGSCC
-operat..ug temperature > 150*F, and austenitic stainless steel
-tensile cress (includmg residual stress) is present, and base metal, welds, and
-halides (e.g., fuoride, chloride) are present, or HAZ caustic (NaC'n ispresent, and
-oxigen or oxidcing species are present (only required to be present in ccnjunction wthalides, not required w/ caustic)
/N t
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v ABB Combustion Engineering Nuclear Operations t
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k MIFIF Calculation No, A PENG-CALC-016, Rev, 00 Page 50 of 70 Table 3 (cont'd)
Degradation Mechanism Criteria and Susceptihte Regions r{,dat on
(,g,,,g, 5,,,,p,g$g, g,,g,,,
SCC ECSCC
-operating temperature > 150*F, and austenitic stainless steel
-tensile stress ispresent, and base metal, welds, and
-an outside piping surface is withinfive diameters ofa probable HAZ leakpath (e.g., valve stems) and is covered with non-metallic insulation that is not in compliance with Reg. Guide 1.36, or an outside piping surface is exposed to wettingfrom chloride bearmg environments (e.g., seawater, brackish water, br:ne)
PilSCC
-piping material is inconel (411oy 600), and no=les, welds, and HAZ
-exposed to primary water at T > 620*F and without stress relief
-the material is mill-annealed and cold worked, or cold worked and welded without stress relief LC MIC
-operating temperature < 150*F, and fittings, welds, HAZ.
-low or intermittentflow, and base metal, dissimilar
-pH < 10, and metaljoints (e g., welds,
-presence! intrusion oforganic material (e.g., raw water system), or Jianges), and regions water source is not treated w! biocides (e.g., refueling water tank) containing crevices PIT
-potential existsfor lowflow, and
-oxrgen or oxidi:ing spectes are present, and
-initiating contaminants (e.g., fluoride, chloride) are present CC
-crevice condstion exists (e.g., thermal sleeves), and
-operatmg temperature > 150 F, and
-oxsgen or oxid :ing species are present FS EC
-operatmg temperature < 250*F. and fittings, welds, HAZ, and
-flowpresent > 100 hrs'sr, and base metal
-velocity > 30)l's, and
-(Ps - PJ / AP < $
-evaluatedin accordance with existingplant FACprogram per plant FACprogram O
ABB Combustion Engineering Nuclear Operations
ARR A"EIF W Q
Calculation No. A PENG CALC-016, Rev. 00 Page 51 of 70 S. 2.1 Thermal Fatigue (TF)
Thermal fatigue is a mechanism caused by alternating stresses due to thermal cycling of a component which results in accumulated fatigue usage and can lead to crack initiation and growth.
S.2.1.1 Thermal Stratification, Cycling, and Striping (TASCS)
The piping sections in the LPSI T damage group, as defined in Section 5.1, are considered subject to thermal stratification. Thermocouple data revealed ATs of up to 345'F (Reference 9.7) due to turbulent penetration in horizontal sections of the shutdown cooling line 2CCA kd 14', which exceeds the allowable AT criteria of 50*F.
Under normal plant operating conditions, these piping sections contain essentially stagnant water at a containment ambient temperature of 120*F, and are conneued to piping containing hot water at RCS hot leg temperature (i.e., 611*F).
The section of the shutdown cooling line 2CCA 2514', downstream of the hot leg injection nozzle up to the vertical section upstream of motor operated valve 2CV.
5084-1, is considered subject to thermal stratification. The upper vertical section near the hot leg injection nozzle is not directly subjected to thermal stratification.
Indirectly however, the effects of the thermal stratification in the downstream horizontalsections couldimpact this piping section with respect to the magnitude of global bending stresses it is subjected to. Consequently, the potential for failure is considered to exist.
S.2.1.2 Thermal Transients (TT)
The potential for thermal transients was identified in the shutdown cooling lines 2CCA 25-14" and 2CCA 25-8" downstream of the hot leg injection nozzle during normalplant cooldown, with an i.sitial temperature of 120*F and a final temperattere of 350*F (Reference 9.4, Figure 4).
However, a review of the plant cooldown procedure (Reference 9.17) revealed that the RCS temperature is less than 245*F before shutdown cooling is placed in service. Consequently, no piping segments were itlentified where a potential exists for relatively rapid temperature changes that would exceed the AT allowable of 200*F.
S.2.2 Stress Corrosion Cracking (SCC)
The electrochemical reaction caused by a corrosive or oxygenated media within a piping system can lead to cracking when combined with other factors such as a susceptible material, temperature, and stress. This mechanism ha;. several forms with varying attributes including intergranular stress corrosion cracking, transgranular stress corrosion cracking, extemal chloride stress corrosion crecking, and primary water stress corrosion cracking.
OV ABB Combustion Engineering Nuclear Operations
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9%lHF Calculation No. A.PENG CALC-016, Rev. 00 Page 52 of 70 E.2.2.1/ntergranular Stress Corrosion Cracking (IGSCC)
The piping section in the LPSI T damage group, as defined in Section 5.1, is subjected to temperatures in excess of 200*F (300*F). However, this piping is not exposed to oxygen or oxidizing species (RCS backpressure maintains hot leg injection isolation valves closed, preventing oxygenste<l RWT water supgly intrusion) andis therefore not considered susceptible to IGSCC. The rema'nder of this system, which is at less than 200*F during normal plant operating conditions, is also not considered susceptible to IGSCC since plant chemistry controls ensure that initiating contaminants (e.g., thiosulfate, fluoride, chloride) levels are negligible.
5.2.2.2 Transgranular Stress Corrosion Cracking (TGSCC)
Plant chemistry controls ensure that the levels of halides or caustics present in the system are maintained extremely low and this piping is therefore not considered susceptible to TGSCC.
5.2.2.3 External Chloride Stress Corrosion Cracking (ECSCC)
ANO-2 complies with the requirements of Regulatory Guide 1.36 for non-metallic thermalinsulation and consequently the potential for ECSCC to occur does not exist.
5.2.2.4 Primary Water Stress Corrosion Cracking (PWSCC)
PWSCC is not applicable as a potential damage mechanism for the LPSI/SDC system due to the fact that there is no inconel(Alloy 600) present in the system (Reference 9.4) and the range of operating temperatures is below the PWSCC required temperature threshold of 620*F (Reference 9.5).
5.2.3 Localized Corrosion (LC)
In addition to SCC, other phenomena can produce localized degradation in piping components. These phenomena typically require oxygen or oxidizing environments and are often associated with low flow or " hideout' regions, such as exists beneath corrosion products or in crevices. This mechanism includes microbiologically influenced corrosion, pitting, and crevice corrosion.
5.2.3.1Microbiologically influenced Corros;on (MIC)
The portions of this system that are less than 150*F during normalplant operating conditions are considered potentially susceptible to MIC. The RWT is a potential source of microbes since biological controls (i.e., biocides) are not utilized and the temperature range is appropriate for MIC to exist.
MIC has not, however, ever been observed to exist in the LPSI/SDC system at ANO-2.
On occasions when the system has been opened for maintenance (e.g., valve disassembly), no evidence of MIC has been discovered. Also, prior volumetric examinations of this system have not revealed the presence of any degradation attributable to MIC attack. Furthermore, from an overallindustry standpoint, MIC has not historically been a source of degradation in LPSI/SDC systems.
ABB Combustion Engineering Nuclear Operations
A &B D
Calculation No. A.PENG CALC-016, Rev. 00 Page 53 of 70 Consequent 4, although the system conditions fall to preclude MIC attack, the potentialis consideredlow on the basis of the lack of ANO-2 or industry historical evidence, and this mechanism is therefore not c0nsidered active for the LPSUSDC system.
5.2.3.2 Pitting (pit)
The essentially stagnant flow conditions and tha oxygenated water supply from the RWTprovide an environment for pitting to occur, however, the absence ofinitiating contaminants (e.g., fluoride, chloride) in the system indicate the likelihood is extremely low.
Additionally, similar to the observations made in the MIC assessment above, pitting has not historically been a source of degradation in the LPSVSDC system for ANO-2 or in the industry.
Consequently, the potential for pitting attack is considered low due to both the absence ofinitiating contaminants in the system and the lack of ANO 2 or industry historic &l evidence, and therefore this mechanism is not considered active for the LPSVEDC systcm.
5.2.3.3 Crevice Corrosion (CC)
Crevice corrosion is not applicable due to the fact that there are no crevice regions i
included within the boundaries of the LPSUSDC system evaluation.
5.2.4 Flow Sensitive (FS)
When a high fluid velocity is combined with various other requisite factors it can result in the erosion and/or corrosion of a piping materialleading to a reduction in wall thickness. Mechanisms that are flow sensitive, and can create this form of degradation include crosion-cavitation and flow accelerated corrosion.
5.2.4.1 Erosion Cavitation (E-C)
All the piping in this system with the exception of the piping section in the LPSI T damage group is less than the E-C temperature limit of 250*F during normal plant operating conditions.
However, there are no regions downstream (within 5 diameters) of pressure reducing orifices or valves in the LPSUSDC system that experience flow greater than 100 hrs /yr or fluid velocities greater than 30 ft/sec (References 8.4, 8.5, and 8.16).
Consequently, this system is not considered susceptible to E-C.
5.2.4.2 Flow Accelerated Corrosion (FAC)
The LPSVSDC system is comprised entirely of austenitic stainless steel piping (Reference 9.4). Since FAC is a phenomenon that only affects carbon steelpiping, O
the LPSUSDC system is not susceptible to this degradation mechanism (Reference
()
9.12).
ABB Combustion Engineering Nuclear Operations
ARR FLIFIF
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1 Calculation No. A PENG-CALC-016, Rev. 00 Page 54 of 70 5.2. 5 Vibration Fatigue VibrotJn fatigue is not specifically made part of the EPRI risk informed ISI process.
l Most documented vibrational fatigue failures in power plants piping indicate that they are restricted to sockst 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 an entirely separate program taking guidance from the EPRI Fatigue Management Handbook (Reference 9.10). If a vibration problem is discovered, then corrective actions must be taken to either remove the vibration source or reduce the vibration levels to ensure future component operability. Therefore, frequent system walkdowns, leakage monitoring systems, and current ASME Section XI system leak test requirement.s are practical measures to address this issue. Because these measures are employed either singly or in combination for most plant systems it is not necessary to use a risk informed inspection selection process for vibration fatigue, 5.3 BASIC DATA 5.3.1 All piping in the LPSI/SDC system is austenitic stainless steel. Under normal plant operating conditions, the LPSI/SDC system, as defined by the boundaries in Section 3.2, os in a standby mode and stagnant. The temperature of this standby system during normal plant operating conditions, with the exception of the piping section (i.e., portion ofline 2CCA 2514") in the LPSI-T damage group, which is heated due to tt,rbulent penetration or convection, will range trom a containment building embient of 120*F to an auxiliary building ambient of BO*F (winter) to 100*F (summer).
5.3.2 Due to the cyclic nature of thermal transients, only those transients which occur during the initiating events Categories I and ll as described in Reference 9.1, Tab'e 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, loss of feedwater. Therefore, the transients to be evaluated are those transients which occur under normal operating and upset conditions.
O ABB Combustion Engineering Nuclear Operations
A It R MIFIF l~
Calculation No. A PENG CALC-016, Rev. 00
(
Page 55 0f 70 6.0 SERVICE HISTORY AND SUSCEPTIBlUTY 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 6 for the Low Pressure Safety injection / Shutdown Cooling System.
Although several pre-commercial references are included for completeness, the timeframe for identifying items applicable to this effort was focused on post. commercial operation (Commercial Operation date of March 26,1980). This was done to avoidinclusion ofitems primarily associated with construction deficiencies as opposed to inservice degradation.
The following databases and other sources were queried to accomplish this review:
Station Information Management System (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 csptura certain non-Code, O component failures.
This query was for non-Code O and SR (safety related) components.
This database contains information from opproximately 1985 to the present.
O
- Condition Report (CR) Database The CR database was queried for any pipe leak / rupture events or other conditions associated with identified damaga mechanisms at ANO-2.
The keywords searched under were; pipe, piping, line, water hammer, leak, leaking and leakage. CR's are written on O. 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 queried using a keyword search specific to ANO-2.
The keywords searched under were: thermal cycling, thermal stratification, thermal fatigue, defect, flaw, indication, fatigue, cavitation and corrosion.
This search 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, pipe & 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.
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O ABB Combustion Engineering Nuclear Operations
N lI N A'%IFBF Calculation No. A PENG cal.C-016, Rev. 00 Page S6 of 70
- Nucient Plant ReliabHity Database System (NPROS)
NPRDS was queriM ior ANO-2 entries for pipe failures. The keywords searched under were: pipe. Tors 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 log exists in electronic format from early 1994 to the present and has search capabilities which allowed a review for events of interest.
The keywords searched under were: water hammer, leak and leakage.
1
- System Upper LevelDocument (ULD)
The ULO was reviewed as a source for historical perspective of issues related to the system and identification of modifications made to the system or changes to operational procedures to address those issues (e.g., water hammer, corrosion or vibrational fatigue).
- Other Station Documents 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.
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D'% 9 EF Calculation No. A-PENG CALC-016, Rev. 00 Page 58 of 70
- 7. 0 RISK EVALUATION The first step in the risk evalvasbn 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 exposed to the same degradation mechanisms (i.e.
damage groups). The next step in the risk evaluation is the determination of the segment risk categories.
This is accomplished by combining the consequence and damage mechanism categories to produce a risk category for each segment. Application of the above criteria results in the formation of 23 risk segments of which 1 is high risk frisk category 21,13 are medium risk frisk category 4) and 9 are low risk (risk category'6). The risk segments are identifiedin Table 7 below.
I J
' OV ABB Combustion Engineering Nuclear Operations
t.
h F%FEF Calculation NO. A-PENG-CALC-016, Rev. 00 Page 59 of 70 Table 7 Risk Segenent identifh adon Msk Segment ID ConsequenceID Damage GroupID Msk Region Piping Une Nos.
Msk Segment Start Pbint Msk Segment EndP6fnt Category fsRure P6tential Risk Category isometric Drawings iPSIR 01-1 LPSI C-01 LPSI-T High 2CCA-25-14 *
(1) *B*Hotleg ill Dow.asurem sWe of 14*
High Smen teek 2
til 2CCA-25-1 Sh. I
~ "
lPSI R-01-2 LPSI C-01 LPSIN Mecrnan 2CCA-25-14 ~
(1) Downstreem side of 14' (1) L&streem of 2CV$084-1 elbow - Item 2 High None 4
til 2CCA-25-1 Sh.1 LPSIR-02 LPSICO2 LPSIN low 2CCA 14 *
(1) Downstream ot 2CV5084-(31 Penetration 2P 27 2CCA-25-8*
1 I41 lbstreem of 2PSV5067 Medium None 6
2GCB 5-14*
(21 Downstroom of 251-19 2GCB-78-2
- 2GCB 1 %
- Ill 2CCA-25-1 Sh.1 (2) 2CCA-25-2 Sh.1 (3) 2GCB 5-4 Sh.1 (4) 2GCB-78-1 Sh.1 LPSIR 03 LPSIC-03 LPSIN Low 2GCB 5-14*
(1) Penetration 2PL27 til Roor Sevetion Q 360'O' Medienn None 6
(1) 2GCB-5-1 Sh.1 LPSIR-04 LPSI-C-04 LPSIN Low 2GCB-5-14 *
(11 Roor Besetion Q 360'0" (2) L& stream of 25120 2GCB 5-4*
111 Downstroom of 2SI25
- 13) Roor Bevetion Q 335'O' Medium None 6
2GCB 5-3*
(41 Roor Bevetion Q 335'O' (1) 2GCB-5-1 Sh.1 (2) 2GCB-5-1 St :
(3) 2rCB-5-1 Sh. 3 (412GCB 5-3 Sh.1 LPSIR-05 LPSIC-05 LP.'l N Low 2GCB-5-14*
111 Roor Bevetion O 335'O' (1) L&stroom of 2SI-1A Medium Nons 6
ill 2GCB-5-3 Sh.1 1
ABB Cor ibustion Engh ing Nuclear Opc ations
p, k.
b ARR 91WEF Calculation No. A-PENG-CALC-016, Rev. 00 Page 60 of 70 Table 7 Risk Segtnent identificador1 (Cont'd)
Risk SegmentID Coruequence ID Demoge Group 10 Msk Region Mping Line Nos.
Mak soy ~.ans stat P6lat Msk Segment Endpoint 1
Category Faawc Potential Msk Category isometric Drawings LPSIR 06 LPSIC-06 LPSIN Low 2DCB-504-2*
(2) Downsweem ot 2SI2A (19 2*x %"Reduchsg kss*st -
(2) Downstreem of 2SI1A Item 22 Mednan None 6
2GCB 14" (112*x 1* Reduensa hsert -
(19 Upsweem of 2CV51231 2GCB 3 8' (3) l4psweem ot 2Sf 3A (41 l% stream of 2*x 1*
(1) 2DCB 604-1 Sh.1 g
w _ g, y (2) 2GCB-1-1 Sh.1 (312GCB 31 Sh.1 (4) 2GCB-508-1 Sh.1 LPSI R-01 LPSIC-07 LPSIN Medn^ an 2GCB-314" (1) Downstroem ot 2SF3A (19 Upsurem of 231M 2GCB-312" (21 kreccessMe Penetration (3) Roor Deverion O 335' G*
High M
4 2GCB-7 8*
(1) 2GCB-31 Sh.1 (2) 2GCB-31 Sh. 2 (3) 2GCB-7-1 Sh. 2 LPSIR-OS LPSIC-08 LPSIN Med%rn 2GCB-312" (21 Foamed Penetration 2007-(11 kseecessMe 1%netration 0039 High None 4
(112GCB-31 Sh. 2 (2) 2GCB-32 Sh.1 LPSI-R-10 LPSIC-10 LPSIN Low 2GCB-5-14*
(11 Roor Bevation Q 335'O*
(21 Upstream of 2ST1B Medium None 6
(1) 2GCB 5-1 Sh. 3 (2) 2GCB 5-2 Sh.1 LPSIR-11 LPSIC-11 LPSIN low 2DCB 505-2*
(21 Downstreem ot 2SI28 (19 Upstrenm of 2CV51241 2DCB-509-2*
(21 Downsweem of 25!-18 (112*x 1* Reducing kssert -
Mediurn None 6
- herss 31
fil 2DCB 506-1 Sh.1 (2) 2GCB-2-1 Sh.1 (3) 2GCB-33 Sh.1 2GCB SO9-1 Sh. !
ABB Combustion Engineering Nuclear Operatiorts
k A FV7
- 'tIFIF Calculation No. A-PENGOLC-016. Rev. 00 Page 61 of 70 Table 7 Risk Segment identification (Cont'd)
Msk SegmentID ConsequenceID Damage GroupID Msk Region Mping Line Nos.
Msk Segment Start IWnt Msk Segment EndIWnt Category FaRurc Putential Risk Category isometric Drawings
+
LPSIR-12 LPSIC-12 LPSI-N Mediurn 2GCB-3-12" (29 Oownstreem of 2SF3B (1) Foamed Penetration 2007-High None 4
(1) 2GCB-3-2 Sh.1 (2) 2GCB33 Sh. I LPSIR-13 LPSI C-13 LPSIN Medium 2GCB 7-14*
fil Downstream of 2SF20 til Roor Bevation O 360'O*
(2) Roor Beverion O 335'O' (3) Roor Bevetions O 360'O' High None 4
- Idi Downstreen of 2315093 3 2GCB-7-6*
141 Downstream of 2CV-CO93 2GCB-7-4
- 2GCB B 14*
2GCB 8-8" (1) 2GCB-7-1 Sh.1 (2) 2GCB-7-1 Sh. 2 1
(3) 2GCB-7-2 Sh.1 (412GCB-8-1 Sh.1 LPSFR-14A LPSIC-14A LPSIN Medium 2CCB-4 6" (2) Roor Beverion @ 360'0*
111 1%"x 1* Redue:ing kneert -
2CCB41%*
Ite,r 19 High None 4
(1) Pwnetration 2F15 Ill 2CCB41 Sh.1 (2) 2GCB-7-1 Sh.1 LPSTR-148 LPSIC-148 LPSI-N Mediurn 2CCB-3-6" (2) Roor Bevation Q 360'O' fll 1%
- x 1
- Reducing husert -
2CCB-3-1 %
- Item 8 High None 4
2GCB-7-6 *
(11 Penetration 2P 10 (1) 2CCB-3-1 Sh.1 (2) 2GCB-7-2 Sh.1 LPSI-R-14C LPSIC-14C LPSTN Medium 2CCB G-6*
(2) Roor Beverion Q 360'0*
(11 Penetration 2P-29 2GCB-7-6*
High None 4
til 2CCB-6-1 Sh.1 (2) 2GCB-7-2 Sh.1 ABB Combustion Engh ing Nuclear Operations
rT 3
/
0 Y
ADR F% FIF Calculation No. A-PENG-CALC-D16, Rev. 00 Page 62 of 70 Table 7 Risk Segrnent identificadon (Cont'd)
Msk SegmentID ConsequenceID Demoge GroupID Risk Region Pfping Une Nos.
Msk Segment Start P61nt Msk Segment EndPhint Category FoRexe P6tential Msk Category isometric Drawings LPSI R-14D LPSI C-14D LPSIN Metfnun 2CCB 5-6*
(2) Roor Beverion @ 360' O~
(1) Penetretion 2P-24 2GCB-7-6*
High None 4
(Il 2CCB 5-1 Sh.1 (212GCB-7-2 Sh.1 LPSIR-15A LPSIC-15A LPSIN Medium 2CCB46*
(1) Penettstion 2P-15 (1) Wsweem of 2ST14A High None 4
(112CCB42 Sh.1 LPSIR-ISB LPSIC-16B LPSIN Medium 2CCB-2 6*
{1) Penetretion 2P 10 (1) @seenm of 2SF148 High f+>ne 4
(1) 2CCB12 LPSI R-15C LPSIC-15C LPSIN Medium 2CCB 6-6*
- 11) Penetration 2P-29 (1) Wsweem of 2SF14C High None 4
(112CCB 6-2 Sh.1 LPSTR-15D LPSFC-15D LPSIN Medium 2CCB-5-6 *
(1) Penetration 2P-24 (1) Wstream ot 2SI-14D High None 4
(1) 2CCB-3-2 Sh.1 LPSIR-16 LPSI C 16 LPSTN Low 2GCB 8-14*
(1) Downstroem of 2SF5A (11 Wsweem ot2ST18 2GCB-8-!2 *
(2) Downstream of 2SI58 til Roor Govetion O 335'O' Medium None 6
(1) 2GCB-8-1 Sh. 5 (2) 2GCB-17-1 SIL 2 LPSIR-17 LPSFC-17 LPSIN low 2GCB-8-14
- ill Roor Beverion O 335' O*
(1) Wserem of 2SI5093-3 2GCB-8-8*
(11 Wsweem of 2CV-5033 Medium None 6
2GCB C-3*
(1) Wsweem ot2ST34 lif 2GCB-8-1 Sh.1 ABB Combustion Engineering Nuclear Operations
()
Calculation No. A PENG-CALC-016, Rev. 00 O
Page 63 of 70 To facilitate applicador of the sampling percentages to determine the inspection scope, ISIS combines like segmenta (i.e., same consequence category and damage group) into segment groups. A total of 3 segment groups have been identified and are summarized in Table 8 below.
Table 8 Risk inspection Scope Segment Consequence Degradet'on Risk Risk Total Selections Selections Groups Category Mechanism Region Category Welds Requked Mode LPSI-01 High None Medium 4
187 19 19 LPSI-02 High Small Leak High 2
11 3
3
{
LPSI03 Medium None low 6
176 0
0 8.0 ELEMENT SELECTION The number of elements to be examined as part of the risk-informed developed program depends upon the risk categories for the risk significant segment groups as indicated in Table 8 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 ofindividualinspection locations within & risk category depends upon the relative seve:ity 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 focused on terminal ends and other locotions (i.e., structural discontinuities) of high stress and/or high fatigue usage. An inspection for cause process shall be implemented utilizing examination methods and volumes defined specifically for the degradation mechanism postulated to be active at the inspection location.
Tables 9 and 10 depict the element selections and other pertinent information (e.g.,
examination methods cnd volumes, basis for selection) for risk significant segment groups LPSI-001 and LPSI 002. As indicated in the Risk Inspection Scope of Table 8, a total of 22 elements have been selected for examination, including 19 elements from segment group LPSI-001 and 3 elements from segment group LPSI-002. The examination methods and volumes specified in Table 10 (risk category 2) are defined in Reference 9.1 and are based upon the degradation mechanism (s) postulated to be active at each selected element.
Currantly, no specific guidance is provided in Reference 9,1 regarding appropriate examination methods and volumes for risk category 4 (i.e., no failure potentialidentdied) element selections. Consequently, the examination methods and volumes specified in Table 9 (risk category 4) are based upon the requirements defined in Reference 9.1 for thermal
- fatigue, p
ABB Combustion EngineerinD Nuclear Operations
ARR
- "E W W Calculation No. A-PENG-CALC-016, Rev. 00 Page 64 of 7G Table 9 Element Selection - Risk Category 4 Seement Groeg Coneequence Facure Potentist itset Careeery Riot Reelen Tetaf I of annnente 10% elelemente LPSI001 High None identified 4
htedium 185 19 Samente 3dected Line No.
Exam Method Riek segment D Deecmpelen
%e Dwg No.
Esem Vohene C:: :
ae /DM Groep D's Reneen tier Senecaron 53 004 2GCB-7-8*
Vohnnerne LPSI-R-07 h the absence of any identified demoge e6%~..
, the Inchest stress (eg.101 element (node point 590, Calc M. 66002-9601 h this Reducer-to Mpe WeM 2GCB-3-1 Sh. 2 Rgure No. 7.1-2 LPSFC-07 / LPSIN
,;,g,,,,,,, y,, y,,,,,g,,,,4, 53 012 2GCB-3-12
- VoAnnetric LPSIR 07 h the obsence of any Mentified demoge mechemsms, the 2nd toghest etress (eq. 101 element snode poht 650C Cole No. 66002-9609 k Bbowto-Dbow Weld 2GCB-3-1 Sh.1 Rgure No. 7.1-2 LPSIC-07 / LPSTN this risk segment has been selected.
54 006 2GCB 3-12" Vohnnetric LPSIR-08 h the absence of any identified demoge em:
6.
, the tughest stress (eq.101 element (node poht IFB, Cafe No. 91 E-0016-1649 h Bbow-to-Plipe WeM 2GCB-3-1 Sh. 2 Rgure No. 7.12 LPSI CD8 / LPSI-N ys;,,;,g,
g g,, y,g,c,g 54-011 2GCB-3-12
- Vohnnetric LPSI-R 08 h the obsence of any Mentified demoge em: u the 3rd tughest stress (eq.101 element (node point 4408, Calc No. 6600 24741 h Pye-to-Dbow WeM
- GCB-3-2 Sh.1 Rgure No. 7.1-2 LPSIC-06 / LPSIN gy,,;,g,
g g,, y,g, egg 54-021 2GCB-3-12" Vohnnetric LPSIR-12 h the ebeence J eny Hentified demoge mni u. the 2nd Nghest abess (eq.109 ehrment inn >de point 50, Cnic No. 660f}2 9611 h this Tee-to-Mpe Weld 2GrB-3-2 Sh.1 Rgure No. 7.1-2 LPSIC-12 / LPSTN rist segrnent has beerr sdectd 55-034A 2GCB-7-6*
Vokanetric LPSI-R 13 h the obsence of any Mentified demoge.u:-~.
, tt:w twghest
}
stress (eq.101 element (node poht 88A, Cole No. 91-DD069-011 h Weldolet-to-Pipe WeM
- GCB-1-1 Sh.1 Rgure No. 7.12 LPSI-C-13 / LPSIN ys;,,.;,g,
w w,g,,,g 55036 2GCB-7-14" Vohnnetric LPSI-R-13 h the absence of any Mentifieddamege em: 6, the 4th tut; hest sWess (eq.101 element (nodo point 92AL Cole No. 91-D 0069D19 h Mpe-to-Dbow Weld 2GCB-7-1 Sh.1 Rgure No. 7.1-2 LPSI-C-13/LPSIN this risk segment hos been selected.55-043 2GCB-7-14*
Volumetric LPSIR-13 h the obsence of any Mentified demoge mechenbms, the 5th Nghest stress (eq.101 element (node point 18B, Cole No. 91-D 0069411 h i
Tee-to-Mpe Weld 2GCB-7-1 Sh. 2 Rgure No. 7.1-2 LPSFC-13 /LPSFN gy,,g,g,,,,,,,,, g,, y,g,,,,g_
55457 2GCB-7-8
- Vohnnetric LPSIR-13 h the obsence of any Mentified demoge mechemsme, the 6th twghest stress (eg.101 element (node point 168, Calc No. 91-D-0069-01) in Mpe-to-Weldotet WeM 2GCB-7-1 Sh. 2 Roure No. 7.1-2 LPSIC-13 /LPSTN tNs risk a t hos & sdectd ABB Combustion En ring Nuclear Operations h
h
V V
L b
- % FIF Calculatio:s No. A-PENG-CALC-016. Rev. 00 Page 65 of 70 Table 5 Eternent Selection - Risk Category 4 (Cont'dl Seement Group Consequence fenure Potential Riek Cateeery Rieh Realen Total 9 er elemente 10% of einmente LPSI001 High None Identified 4
Medium 185 19 Bemente Selected Line No.
Exem Method Riek Segment JD Description leo Owg No.
Euem Vohene Coneequence / DM Graanp JO's Reeeen for SeleceTon
$7-CD4 2GCB 3-12*
Volumetric LPSIR-07 kr the obsence of any identified demoge m+d--6,,
the 3rd fughest stress (eq.101 element (node poht 790E. Calc No. 66002-9601 in Elbow-to-Mpe WeM 2GCB-31 Sh.1 Figure No. 7.1-2 LPSI C-07 / LPSI h gy, gg,
,, w
,g,,g 58-001 2GCB'3-12' VoAnnetric LPSIR-12 k the obsence of any iden'ified demoge mechanisms, the Nghest stress (eg.101 element (node point 1208. Colc No. 66002-9611 h-Mpe-to-Elbow WeU 2GCB-3 3 Sh.1 Rgure No. 7.1-2 LPSIC-12 / LPSIN gy,,;,g,,,,,,, g,, g,,,,,g,eg,g 61-002 2 GCB 6
- Vokanetric LPSIR-140 h the obsence of any identified damage mecherusens, the Nghest stress (eg.101 element (node point 3OA, Cole No. 91D-6009D11 in Mpe-to-Velve Weld 2GCB-7 2 Sh.1 Rgure No. 7.1-2 LPSF C-140 / LPSV N tNs risk segment, wNch hos also been cut out once (FwSC11 and reweMed (creethq Ngher resahoolstressest, hos been selected.61-011 2GCB-7-6*
Vohnnetric LPSIR-14C k the obsence of any identified demego,in.:_L.,s, the highest l
stress (eq. 101 element (node point 320 Cole No. SID 6009D11 ks Mpe-to-Velve Weld 2GCB 7-2 Sh.1 Rgure No. 7.1-2 LPSIC-14C/LPSIN tNs risk segment, wNch hos oise been cut cut once (FW9C11 and roweided (creating Ngher residuel stresses), hos been selected.61-021 2GCB-7-6*
Vohnnetric LPSIR-148 k the obsence of any identified demoge
,n: -Li. the Nghest stress (eg.101 e?oment (nodo point 20E, Cole No. 91D-6009D11in Mpe-to-Velve WeM 2GCB-7 2 Sh.1 Rgure No. 7.1-2 LPSIC-148/ LPSTN t%is risk segment, wNch has also been cut out once (FW12C1) and rewelded (creating Ngher residuelsWessesi, has been selected.61-025 2GCB-7-6*
Vokor=retric LPSI R-14A kr the absence of any Mertified demoge
,1, :
6,
, the Nghest etress (eg.101 element (node poest 28F, Cole No. 91-D-6009D11 h Mpe-to-Velve Weld 2GCB-7-1 Sh.1 Rgure No. 7.1-2 LPSIC-14A /LPSTN gy,,;,g,,,,
.m g,,
,g,, s,,,,s ed out once MM M roweMed (creating Ngher residdel stressesi, hos been selected.66-002 2CCB 6-6*
Volumetric LPSIR-158 kr the obsence of any identified demoge ;,n: L,
she Nghest stress (eq.101 element (rmde point 310E, Cole No. 66002-1047) in Mpe-to-Elbow Wald 2CCB 6-2 Sh.1 Rgure No. 7.1-2 LPSI-C-158 / LPSI-N ens M sem has & sdectd 66-009 2CCB 5-6*
Vohnnetric LPSI-R 160 kr the obsence of anyidentifieddemoge
-l 6..
the 2ndlughest stress (eg.101 element (node point 3108. Cole No. 66002-10461 hs Dbow-to-Mpe WeM 2CCB S-2 Sh.1 Rgure No. 7.1-2 LPSIC-150 /LPSTN gy,,;,g,,,,,,, y,, y,,,,,y,,g,g, ABB Combustion Engineering Nuclear Operations
l NI
- 'tIFIF Calculation No. A.PENG-CALC-016, Fev. 00 Page 66 of 70 Table 9 Element Selection - Risk Category 4 (Cottt'd)
Seement Grose Consequence FeMure Potential Riek Cateson Riek Renien l TotalI of elements 10% of elemente LPSI001 High None kientified 4
Medium 185 19 Bements Selected Line No.
Esem Method Riot Segment D Deectfption feo Dwp No.
Esem Vehene Coneequence /DAt Grom D's Reenen for Seleceien 66-013
.?CCB 3 6*
Vokanetric LPSIR-15C k the obsence of any identified demoge em:
6.
, the leghest stress leg.101 e* ament inode point 190C Calc No. 66002-10481 in Penetration-to-f%pe Wekt 2CCB 3 2 Rgure No. 7.12 LPSI C-15C/ LPSI-N tNs risk segment, which hos also been repaired once (FW1R11 and rowelded (creating Ngher residual stressesl. hos been selected.66-014 2CCB 4 6' Vokanetric LPSTR-15A k the obsermee of any klentified demoge n.echerusms, the lughest
.; tress leg.101 element inode poet 255L Cale No. 6600210501 k FWretration-to-Pipe Weld 2CCB42 Sh.1 Rgure No. 7.1-2 LPSFC-15A /LPSIN this risk segrnent hos been selected.
No odditionef rish estegory 2CCA-25-14' N/A LPSIR 01-2 Risk category 4 segment LPSIRD1-2 is kceted mwnedvarey downstroom of risk category 2 segment LPSIR 01-1. A feauce in 4 element selections are 2CCA-25-1 Sh.1 N/A LPSI C01/ LPSIN either of these non risk segments, which together form consequence
,4,g.
segment LPSIC-01, witresult k the same plant impact (i.e., LOCAL Hence, the eternent selectiuns prowded h Table 10 for the porten of LPSTCD1 subject to a demoge mecherusm (i.e., risk category 21 are considered to erwebpe the porvon of LPSFC01 not sukoct to e demoge aml
=h, (Lo., risk category 41. As a result, no eternent selections have been mode h risk segment LPSIR 01-2.
h ABB Combustion Engh ring Nuclear Operations h
7s o
l Ann
- 't WIF 1
Calculation No. A-PENG-CALC-016, Rev. 00 Page 67 of 70 Table 20 Eternent Selectiort - Risk Category 2 Somment Gmse
_Corweguence fearre Potendd Riek Category niek Manien Totaf I of e6eneenes 269G ereoemerne IPSI 002 High SmaR leek 2
High 11 3
Bements Seiected Lkse No.
Eron Method Riek Segment D Desenpden Are Dwg Me.
Eram Volume C-.z
- /DM Groep D*s Reneen for Senecaron 25-001 2CCA 2514' Vokmsetric LPSIA011 The norrte aren is stais risk segment is es4ected to Ngh bendrrsg stress due to the dowrastroom occurrence of therme! stratification Hot leg Shutdown Conting 2CCA-25-1 Sh.1 Rgure No. 7.1-2 LPSFC-01/ LPSI-T (turbulent penetrationi. 71w terment end element has been selected Norrie-to-Sofe End Weid
,;,,ce it is as4ected to the Nghest bending moment hs this risk segment.25-003 2CCA-25-14*
Vohmsetric LPSIR 01-1 The honzontal section of this risk segment is subtected to thermal
.tratification due to turbulent penetration. TNs element has been 11pe-to-Show Weld 2CCA-25-1 Sh.1 Rgwe No. 7.1-2 LPSI-C-01/ LPSF T sdected sksce due & beweers the %tet ered wM M sections may expersence some thermet cycGng 25-004 2CCA-25-14
- Vokanetric LP~IR-01-1 The horirontal section of this risk segment is sagected to thermet stratification due to terbulent peneestion. TNs element hos been Dbow-to-Mpe Weld 2CCA-25-1 Sh.1 Rgure No. 7.1-2 LPSIC-01/LPSI T selected shsce itis sub ected to e Ngher bending stress.
l 8r ABB Combustion Engineering Nuclear Operations 1
l
ABB D
Calculation No. A PENG CALC 016, Rev. 00 (0
Page 68 of 70
9.0 REFERENCES
I 9.1
- Risk Informed Inservico Inspection Evaluation Procedure,* EPRI Report No. TR-l 106706, Interim Report, June 1996.
9.2 EPRIInservice Inspection Software (151S*),1996, 9.3 Arksasas Nuclear One Unit 2, ' Safety Analysis Report,' Amendment No.13.
9.4
' Design Specification for ASME Section ill 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 Une Ust (M 20831, dated 3 31 9 61.
9.6
- 9. 7
- Evaluation of Thermal Stratification Effects on the Shutdown Cooling Une for 1
Arkansas Nuclear One, Unit 2,* ABS CE Report No. A MECH ER-009 Revision 00, August 1993.
(
)
9.8
- Techn/Lal Specification for Insulation for Arkansas Nuclear One. Unit 2 of the Nu Arkansas Power and Ught Compsny,' Specification No. 6600 M 2136, Revision 9.
9.9
' Primary Chemistry Monitoring Program,' Procedi ~s No. 1900.106, 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 I.ow Pressure Borsted Water Systems,' EPRI Report No NP 3320, 9.12
' Flow Accelerated Corrosion Prevention Program,' HES 05, Revision 1.
9.13 Arkansas Nuclear One Unit 2 " Technical Specifica' ions, Appendix A to Ucense No.
NPF 6, Amendments Nos.173 and 174.*
9.14 Geertner, J.
P., et. al. ' Arkansas Nuclear Ono Unit 2 Internal 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 Assess,nent, Individual Plant Examination Submittal,' 94-R 2005 01, Rev. O, August 1992.
(3
(
)
v ABB Combustion Engineering Nuclear Operations
ABB Calbulation No. A PENG CALC 016, Rev. 00 Page 69 of 70 9.16 Entergy, Arkansas Nuclear One Unit 2, Isometnc Drawings:
1.0 Drawing No M4232, Sheet la Rev.103; 'Puping & Instrumentation Diagram Safety iniection System.'
2.0 Drawing No. 2CCA 251, Sheet 1, Rev.10; 'Large Pipe Isometric 3hutdown Cooling Discharge from Reactor Coolant loop to 2CV 5086 2.'
3.0 Drawing No. 2CCA 25 2, Sheet I, Rev. 6; 'Large Pipe Isometric Safety injection Piping from Valve 2SI 19.*
- 4. 0 Drawing No. 2CCB-31, Sheet 1, Rev. 7; "Large Pipe Isometric Safety injection Piping from Control Valve 2CV 50371 to Containment Penetration 2P 10.
- 5.0 Drawing No. 2CCB 3 2, Sheet 1, Rev. 7; " Lame Pipe isometdc Safety injection System Piping from Containment Penetration 2P 10 to Valve 2St.
148.*
6.0 Drawing No. 2CCB-41, Sheet 1, Rev. 8; 'Large Pipe Isometric Safety injection Piping from 2CV 50171 to Containment Penetration 2P 15 '
7.0 Drawing No. 2CCB-4 2, Sheet 1, Rev. 7; "Large Pipe isometric Safety injectiot' from Containment Penetration 2P 15.*
8.0 Drawing No. 2CCB 51, Sheet 1, Rev. 7; "Large Pipe Isometric Safety injection Piping from Contre
- Valve 2CV 5077 2 to Containment Penetration 2P-24. "
9.0 Drawing No. 2CCB 5 2, Sheet 1, Rev. 6; ' Lame Pipe isometric Safety injection from Flued Head 2P-24 to Reactor Coolant Pump 2P-320."
10.0 Drawing No. 2CCB 6-1, Sheet 1, Rev. 7; 'Large Pipe Isometric Safety injection from 2CV 5057 2 to Containment Penetration 2P 29.*
11.0 Drawing No. 2CCB 6 2, Sheet 1, Rev. 4; *Large Pipe Isometric Penetration Piping from 3. sed 2P 29 to Valve 2SI 14C.'
12.0 Drawing No. 2GCB 1 1, Sheet 1, Rev. 12; " Lame Pipe isometdc Low Pressure Safety injection Pump 2P 60A Inlet.'
13.0 Drawing No. 2GCB 21, Sheet 1, Rev. 16; *Large Pipe Isometric Low Pressure Safety injection Pump 2P 608 Inlet Piping.'
14.0 Drawing No. 2GCB 31, Sheet 1, Rev.19, Sheet 2, Rev.1; 'Large Pipe isometric Low Pressure Safety Injection Pump 2P-60A Discharge.'
15.0 Drawing No. 2GCB 3-2, Sheet la Rev. X; 'Large Pipe isometric Low Pressere Safety injection Pumps Discharge to Shutdown Cooling Heat Exchangers.*
16,0 Drawing No. 2GCB 3 3, Sheet 1, Rev. 12; *Large Pipe isometric Low Pressure Safety injection Pump 2P 608 Dischargs to Shutdown Cooling Heat Exchanger 2E 358.*
17.0 Drawing No. 2GCB 51, Sheet 1, Rev.19, Sheet 2, Rev. 2, Sheet 3, Rev.1;
'Lorge Pipe Isometric RC Loop 2CCA 25 to LPSI Pump kP 60A & 2P 608 Inlet. "
18.0 Drawing No. 2GCB 5 2, Sheet 1, Re t 6; "Large Pope Isometric Reactor Ccolant loop to low Pressure Safety infection Dump 2P-608."
19.0 Drawing No. 2GCB 5 3, Sheet 1, Rev. 8; 'Large Pipe isometric Safety injection from Reactor Coolant loop to LPSI Pump 2P 60 A & B Inlet."
20.0 Drawing No. 2GCB 5-4, Sheet 1, Rev. 7; "Large Pipe isometric 2CCA 25 to low Pressure Safety injection Pump 2P-60A & B Inlet."
21.0 Drawing No. 2GCB 71, Sheet 1, Rev.15, Sheet 2, Rev. O. Sheet 3, Rev. 0; "Large Pipe Isometric Low Pressure Safety injection Discharge Header.'
ABB Combustion Engineering Nuclear Operations
("')
Calculation No. A PENG CALC 016, Rev. 00 Page 70 of 70 22.0 Drawing No. 2GCB 7 2, Sheet 1, Rev. 5; *Large Pipe Isometric Low Pressure Safety injection Header to Containment Penetrations 2P 10, 2P-24 & 2P.
- 29. '
23.0 Drawing No. 2GCB 81, Sheet 1, Rev.14, Sheet 2, Rev. 2; *Large Pipe isometric Shutdown Cooling Heat Exchanger Discharge Hender to low Pressure Safety injection Header.'
24.0 Drawing No. 2GCB.171, Sheet 2, Rev. 0; *Lorge Pipo isometric Discharge from Shutdown Cooling Heat Exchanger 2E 358.*
25.0 Drawing No. 2GCB 5041, Sheet 1, Rev. 6; *Small Pipe Isometric Boric Acid Atake-up Pump 2P-60A Discharge to Line DCB 2 4*.*
26.0 Drawing No. 2GCB 5051, Sheet 1, Rev. 11; *Small Pipe isometric Low Pressure Safety injection Pump 2P 60B to Refueling Water Tank 2T 3.*
27.0 Drawing No. 2GCB 5081, Sheet 1, Rev. 7; "Small Pipe Isometric Bypass
\\
l from low Pressure Safety injection Pump 2P 60A Discharge to Refueling Water Tonk 2T 3.'
28.0 Drawing No. 2GCB 5091, Sheet la Rev. 2; *Small Pipe Isometric 2P-60B Low Pressure Safety injection Pump Discharge to Refueling Water Tank 2T-
- 3.
- 9.17
- Plant Cooldown' Procedure No. 2102.010, Revision 28, Entergy Operations, Arkansas Nuclear One.
9.18 Swain, A. D. and Guttmann, H.E.; ' Handbook of Human Reliability Analysis with
(
Emphasis on Nuclear Power Plant Operations', NUREG CR 1278, August 1983.
V 9.19 North Atlantic Energy Services Cntp. ' Individual Plant Examination Extemal Events',
Report for Seabrook Stasjon, fesponse to Generic letter 88 20, Supplement 4, September 1992.
9.20
- Probabilistic Seismic Hazard Evaluations at Nuclear Plant Sites in Central and Eastem United States: Resolution of the Charleston Earthquake issue *, EPRI NP-6395 D, April 1989, Prepared by Risk Engineering, Inc., Yankee Atomic Electric Company, and Woodward Clyde Consultants.
9.21
' Revised Livermore Seismic Hazard Estimates for 69 Nuclear Power Plant Sites East of the Rocky Afountains", NUREG 1488 fFinalReport), April 1994.
9.22 Interoffice Correspondence from A. V. Bauer to Quality Records, letter No. PENG.
97140, 'Submittalof SIA Calculations,' dated.luly 21,1997.
/3 V
ABB Combustion Engineering Nuclear Operations
Calculation No. A PENG CALC 016, Mov. 00 Ptge A t of A42 O
APPENOlX A
- FMECA CONSEQUENCEINFOMMATION MEPOMT*
(Attachment Pages A t. A42)
O ABB Combustion Engineering Nuclear Operations -
FMECA - Consequence Information Report CaNd*
- d FNGCW8A N l4-s 91 m
Page A2 of A42 Consequence ID: LPSI C 01 Consequence
Description:
less of reactor coolant via SDC suction line occurs due to a line break Break Slie:
Large Isolability of Break: No ISO Comments: The break is postulated to occur during normal power operation, and in the piping from downstream of RCS hot leg (Loop B) to upstream of SDC suction valve 2CV 50841. This consequence includes the utids in the applicable portion ofline 2CCA 2514*.
A failure in this segment would result in a large Loss of Coolant Accident (LOCA). This would be characterized by a rapid decrease in RCS pressure, followod by automatic plant shutdown in order to bring the plant to a stable state. The lost RCS inventory would drain to the containment sump as expected and would then be recirculated by the HPSI pumps. A failure in the segment would also pretnt hot leg injecdon from being accomplished. The failed segment cannot be isolated during a LOCA because there are no isolation vahts in the RCS.
Spatial Effects: Containment Affected location: Containment Building Spatial Effects Comments: This segment of the SDC suction line is connected to Loop B of the RCS hot leg. A dynamic analysis of the SDC suction line inside the containment was performed.
The analysis concluded that uncontrolled pipe whip of the line segment would not impair the ability to nitigate the consequences of the ntnt and reach cold shutdown. The location of structures and safety related components also ensure that j
jet impingement will not impair safety features actuation (SAR Section 3.6.4.2.7.2).
In addition all ECCS components and associated electrical equipment have been designed to withstand the LOCA emironmental v.nditions inside the containment (SAR Section 6.3.2.12.1). Hence for this postulated break location, it is assumed that spatial effects are negligible.
Initiating Event: 1 Initiating Event ID: A Initiating Esent Recovery: Based on the ANO 2 IPE (Report 94 R 2005-01, Rev. 0) two of four SITS, one of three llPSI pumps, and one of two LPSI pumps are required for successful mitigation of a farge LOCA during RCS inventory control (i.e., injection mode).
For long term inventory control and heat removal (i.e., recirculation mode), one of three llPSI pumps and one CS pump with an associated SDC heat exchanger or one CS pump and two containment cooling units are required for mitigating a large LOCA. Automatic actuation of the Factor protection system and the enginected safety features actuation system occurs in response to this initiating mtnt.
Loss of S stem: N System IPE ID:
N/A 3
Sy stem Reemcry: Durirg a large LOCA, the HPSI, LPSI, SITS, CS and Containment Cooling System will not be a!Iected. No operator actions or automatic isolation are needed to recover from the segment failure. All engineered safety featwes required for mitigating this initiating event are actuated automatically.
Loss of Train: N Train ID:
N/A Train Recovery: N/A Consequence Comment: A large LOCA occurs due to a failure in the pipe segment. There are no valves in this segment to isolate it from RCS hot leg 'B', thus the potential for isolating the failed segment is not applicable. A large LOCA is classified as a limiting fault ntnt in
i 4
FMECA Consequence Inforenation Report Catalauon A rrm cuc 016,h oo
}
IW1 Page A.f of Au l
Table 1. Bec.ause the segment failure causes an initiating event, and based on Table I w hich was developed specifically for ANO 2 using the guidelines prmided in Tables l
3.1 and 3.4 of the EPRI procedure (EPRI TR.106706), a HIGH consequence category is assigned.
Consequence Category: HIGH C
Conseguemee Rank O
i I
4 i
I l
l 1
)
i 1
I 1
i 4
4 k
i e
t 9
FMECA - Consequence Information Report ColManw No. A FM CM MJn 00 la-stt91 Pop A4 of A42 Consequence ID: LPSI C-02 Consequence
Description:
less of reactor coolant occurs during shutdown operation due to a SDC line break inside containment.
Break Stre:
Large isolability of Break Yes 150 Comments: The break is postulated to occur during any of the various states of plant shutdown operation, and in the piping from downstream of SDC suction valve 2C%5084 1 to penetration 2P27 inside containment. 'Ihe piping from SDC suction line to upstrearn of manual vaht 251 19 in the refueling canal line is also included. Of the various plant shutdown states considered (see Sections 4.4.2 and 4 2 3), mid loop operation is the most risk significant because of the relatively short time it *vould take to uncmtr the core. Therefore, only the consequence resulting from a pipe L cak during mid loop operation is described herein. This consequence l
includes all welds in lines 2CCA 25 8" 2GCD 781.5" and 2GCD 78 2" and the utids in the applicable portions oflines 2CCA 2514" and 2GCB 514".
1 A failure in this segment would cause reactor coolant to drain to the containment sump during mid loop operation. Several unexpecte l alarms and indications would be encountered following the segment failure. These include increasing or high water level in the containment sump, decreasing RCS level, low LPSI pump discharge flow and pressure indications and fluctuating LpSI pump motor current. Because the operators are carefully monitoring RCS level during mid loop operation, it is likely that the failed segment would be identified and isolated.
Spatial Effects: Containment Affected location: Containment Building Spatial Effects Comments: This segment of the SDC suction line is connected to Loop B of the RCS hot leg. A dynamic analysis of the SDC suction line inside the containment was performed.
The analysis concluded that uncontrolled pipe whip of the line segment would not impair the ability to mitigate the consequences of the event and reach cold shutdown. The location of structures and safety related components also ensure that jet impingement will not impair safety features actuation (SAR Section 3.6.4.2.7.2).
In addition, all ECCS components and associate electrical equipment have been designed to withstand the LOCA emiromnental conditions inside the containment (SAR Section 6.3.2.12.1). Hence for this postulated break location, it is assumed that spatial efIccts are negligible.
Initiating Egent: 1 Initiating Event ID: A initiating Event Recovery: One of the redundant makeup sources is required for mitigation and to accommodate decay heat boil off. Manual initiation c.f makeup is required.
Loss of System: S System IPE ID:
SDC System Recovery: No recovery of SDC is expected in the near term. By closing SDC suction valve 2C%50841 from the control room the failed segment can be isolated it is assumed that the operators ability to isolate the break during mid loop operation is equivalent to having one backip train.
less of Train: N Train ID:
N/A Train Recovery: N/A Consequence Commtat: The segment failure causes a loss of reactor coolant and the unavailability of SDC during mid-loop operation. During this mode of shutdown operation, the operators are l
l
i l
FM ECA - Consequence information Report Calmlenm Na A PENG44LC-016, An,00 l
l&s 91 Page AS of A42 w
l carefully monitoring RCS inventory due to the importance of this configuration, but isolation may be irrelevant because RCS level is aliendy near the invert of the reactor noules. An incicase in the level of operator awareness is rW ne Shutdown Operadons Protecdon Plan utilized at ANO 2 requires that redundant sources and associated paths are in place to provide makeup during mid-loop operation. hus, there would be at least two equivalent backup trains available for midgating this type of LOCA.
Because there are two backup trains avail.ble, the resuldng consequence is MEDIUM.
This is based on the Table 3.4 of the EPRI procedure (EPRI TR.106706).
Conseque.ee Category: MEDIUM O
Co.seg.e.e an.k O
O 9
FMECA - Consequence lufortnation Report Ca h 'a'= '
Description:
Loss of reactor coolant occurs during shutdown operation due to a SDC line break in the upper south piping and penetration imm.
Break Sizer Large Isolability of Breakt Yes ISO Comments: The break is postulated to occur during any of the various states of plant shutdown operation, and in the piping from downstream of containment penetradon 2P27 to the floor penetration at elevation 360' 0". Of the various plant shutdown states considered (see Sections 4.4.2,4.4.4 and 4.4.5), mid-loop operation is the most risk-significant because of the relatively short time it would take to uncover the core. Therefore, only the consequence resulting from a pipe break during mid loop operation is deset2+d herein. This consequence includes the welds in the applicable portion ofline 2GCD 514",
A failure in this segment would cause reactor coolant to be diverted to the general access area of the "teactor Auxiliary Building (RAB)(Calc. 89 E-0048 35, pg. 28) during mid-loop operation. Several unexpected alarms ated indications would be encountered in the control room following the segment failure. These include increasing or high water level in the RAB sump, decreasing RCS level, low LPSI pump discharge flow and pressure indications and fluctuating LPSI pump motor current. Because the operators are carefully monitoring RCS level during mid-loop operation, it is likely that the failed segment would be identified and isolated.
Spatial Effects: Local Affected location: Room 2084 i
Spatial Effects Comments: A review of Plant Design Drawings M 2044 and M 2063, and Figure 3.6 2 at l
clevation 354' 0" of the ANO 2 IPE (Report 94 R 2005 01, Rev. 0) indicates that this line segment is located in flood zone RAB 2084 DD. The IIPSI, LPSI and CS injection line isolation valves, SDC suction valve 2CV 5038 1, senice water isolation valves for containment cooling system (CCS) units 2VCC 2A and 2VCC.
j 2B, EFW distribution valves to steam generator 2E 24 A are located in this flood zone. The ANO 2 Internal Flood Screening Study (Cale. 89 E 0048 35. Pd 13) assumes the failure of all components in the flood inutiation zone. This assumption is too conservative for this evaluation. For a limiting line break during a demand for SDC, it was observed during the walkdown that the force exerted on the entrance non water tight door would cause the door to open before a significant amount of water can accumulate inside the room and flood the valve motors. It was also observed that spraying orjet impingement may affect certain vahts (i.e.,2CV5015 1,2CV 5016 2,2CV 50171 and 2CV 56121) because of the close proximity of these valves to the line segment where the failure is postulated. It should be noted that the HPSI, CCS and CS valves are not needed to support SDC. The LPSI and EFW valves would be in their desired positions at the time the break is postulated to occur. Spatial effects of these valves would cause them to fail as is. IIence, flooding or other spatial effects due to the segment failure are considered to be insignificant.
The walkdown revealed that the outflow of water from the flood initiation zone can propagate through the non water tight door to the Access Area, Tank Area and passageway (i.e., flood zone RAB 2073 DD) at elevation 354' 0", From this flood zone, the water can eventually propagate to the RAB sump of the General Access Area of elevation 317' 0" via stairway No. 2001 and the floor drain system. Because
FMECA - Consequence Infor nation Report C*h* law Na A FMM 016.Rn 00 t4-ser e1 Page A7 of A42 the General Access area is large and there are direct indications of RAB sump level and fluctuation of LPSI pump motor current in the Control Room, it is assumed that corrective actions will be taken (i e., closure of SDC suction valves inside containment)in a timely manner. %erefore for this SDC line break scenario, spillover from the General Access Area to the ECCS pump room is very unlikely but suction to the LPSI pumps would be lost.
talttating Event: I taltiating Event ID: A laitiating E,ent Recovery: One of the redundant makeup sources is required for midgation and to accommodate decay heat boll off. Manual initiauon of makeup is required.
- 1. ass of System: S System IPE ID:
SDC System Recogeryl do recovery of SDC is expected in the near term. By closing SDC suction valve 2CV.3084 1 or 2CV.$086 2 from the control room, the failed segment can be isolated. It is assumed that the operators ability to isolate the break is equivalent to having one backup train.
3 1ms of Trsie N
TralaID:
N/A Train Reemery: N/A Consequence Comment: The segment failure causes a loss of reactor coolant and the unavailability of SDC during mid loop operadon. During this mode of shutdowti operation, the operators are l
carefully monitoring RCS inventory due to the importance of this configuration, but isolation may be irrelevant becasue RCS level is already near the invert of the reactor nonles. An increase in the level of operator awareness is assumed The Shutdow1:
O' Operations Protection Plan utilized at ANO 2 requires that redundant sources and associated paths are in place to provide makeup during mid loop operation. Thus, there would be at least two equivalent backup trains available for mitigating this type ofLOCA.
Because there are two backup trains available, the resulting consequence is MEDIUM.
This is based on the Table 3.4 of the EPRI procedure (EPRI TR.106706).
Consequence Category: MEDIUM O
Consequence Rank O
O nm
cohLs= No A.PENG C4M 016 Rev. 00 FMECA Consequence Inforrnation Report 14 Sep-97 Page A8 of An Consequence ID: LPSI-C 04 Conwquence
Description:
Loss of reactor coolant occurs during shutdown operadon due to a SDC suction line break in the lower south piping penetration area.
Hrcak Slie:
Large isolability of Breakt Yes ISO Comments: The break is postulated to occur during any of the various states of plant shutdown operadon, and in the piping from ceiling penetradon at elevation 360' 0" to floor penetradon at elevation 335' 0". Of the various plant shutdown states considered (see Sections 4.4.2 and 4.4.5), mid.
loop operadon is the most risk significant because of the reladvely short Ome it would take to uncover the core. Therefore, only the consequence resulting from a pipe break during mid loop operation is described herein. This consequence includes all the welds in lines 2GCD 5 4" and 2GCD 5 3" and the welds in the appilcable pordon ofline 2GCD 514",
A failure in this segment would cause reactor coolant to be diverted to the general access area of the Reactor Auxiliary Building (RAB)(Calc. 89 E 0048 35, pg. 28) during shutdown operadon. Several unexpected alarms and indications would be encountered in the control room following the segment failure. These include increasing or high water level in the RAB sump, decreasing RCS level, low LPSI pump discharge flow and pressure indications and fluctuating LPSI pump motor current. Because the operators are carefully monitoring RCS level during mid-loop operadon, it is assumed that the failed segment would be identified and isolated.
Spatial Effects: Local Affected Location: Room 2055 Spatial Effects Comments: A review of Plant Design Drawing M 2045 and Figure 3.6 2 at elevadon 335' 0" of the ANO 2 IPE (Report 94.R 2005 04, Rey,0) indicates that this line segment is located in flood zone RAB 2055 JJ in the Reactor Auxiliary Buildir.g (RAB). The lip 51 orifice b> pass valves (2CV 5103 1 and 2CV 5104 2), LPSI header isolation valve 2CV 5091, SDC heat exchanger throtdc valve 2CV 5093 and EFW pump flush valve 2CV 0714 1 are located in this zone. The ANO 2 Internal Flood Screening Study (Calc. 89 E 0048 35, pg.13) assumes the failure of all L,mponents in the flood initiation zone. This assumption is too conservative for this evaluation.
For a limidng line break during a demand for SDC, it was obsen ed during the walkdown that the force exerted on the entrance non water tight door would cause
,the door to open before a significant amount of water can accumulate inside the room and flood the valve operators. It was also observed that spraying orjet impingement may affect cenain valves (2CV 51031,2CV 5093 and 2CV 5091) because of the close proximity of these valves to the line where the failure is postulated. It should be noted that the IIPSI valves are not needed in support SDC.
Since SDC will be lost because of the segment failure, the impact of spatial effects on the SDC va ves within this flood zone is of no significance.
The walkdown revealed that the outflow of water from the flood initiation zone can propagate through the non water tight door to the adjacent tank rooms, pump room and corridor (i.e., flood zone RAB 2040 JJ) at elevation 335' 0". From utis flood zone the water can eventually propagate to the RAB sump of the General Acces" Area at elevadon 317' 0" via stairway No. 2001 and the floor drain system. W%r can also propagate from the flood initiation zone to the RAB sump via the floor drain system. Because the General Access area is large and there are direct indications of RAB sump level and fluctuation of LPSI pump motor current in the
i FMECA - Consequence Inforanation Report Ca'r*' ado No. A PENO CAM.0ld. Rev. 00 O
ls.5er91 Page A9 of A41 Centrol Room, it is assumed that correcths actions will be taken (i.e.. closure of SDC vaht 2CW50381) in a timely manner. Therefore for this SDC line break scenario, spillmtr from the General Access Areas to the ECCS pump room is very unlikely but suction to the LPSI pumps would be inst.
lultiating Event: 1 1mitiatleg Event ID: A teltleting Event Recovery: One of the redundant makeup sources is required for mitigation and to accommodate decay heat boil +ff. Manual initiation of makeup is required.
Ims of System S System IPE ID:
SDC Systesa Recovery: No reemtry of SDC is expected in the near term. By closing SDC suction valve 2C%5084 1, 2C%5086 2 or 2CW5038 1 from the control room, the failed segment can be isolated,11 is assumed that the operators ability to isolate the break during SDC operation is equivalent to havmg one backup train.
14es of Trale N
Trale ID:
N/A Trale Recovery: N/A Consequence Comment: The segment failure causes a loss of reactor coolant and the unavailability of SDC during mid-loop operation. During this mode of shutdown operation, the operators are carefully monitoring RCS inventory due to the importance of this configuration, but isolation may be irrelevant because RCS lesti is already near the invert of the teactor nonJes. An increase in the level of operator awareness is assumed The Shutdown Operations Protection Plan utilized at ANO.2 requires that redundant sources and O
associated paths art in place to provide makeup during mid-loop operation. Thus, there would be at least two equivalent backup available for mitigating this type of LOCA.
Because there are two backup trains available, the resulting consequence is MEDIUM.
This is based on the Table 3.4 of the EPRI procedure (EPRI TR 106706).
Consequence Category: MEDIUM O
Co. seq e.c. Ra.k O
O
FMFCA - Consequence Infonnation Report Cabla'* Na A f"G C4M 88. Arx 00 14.s 97 Page A10 </ AC e
Consequence ID: LPSI C 05 Consequence
Description:
less of reactor coolant and train 'A' of the ECCS pumps occurs due to a SDC line break in ECCS pump room "A".
Break Stre:
Large Isolability of Break Yes 150 Comments: The break is postulated to occur during any of the various states of plant shutdous operation, and in the piping from floor penetration at elevation 335' 0" to upstream of manual valve 2SI.
I A. Of the various plant shutdown states considered (see Sections 4.4.2 and 4.4.5), mid loop operation is the most risk significant because of the relatively short time it would take to uncover the core. Therefore, only the consequence resulting from a pipe break during mid loop operation is described herein. Tids consequence includes the welds in the applicable portion of l
line 2GCD 514".
l A failure in this segment would cause reactor coolant to be divertad to ECCS pump room "A",
Several unexpected alarms and indications would be encountered in the control room following I
the segment failure. These include increasing or high water level in the ECCS pump room "A",
dec:caslag RCS level, low LPSI pump discharge flow and pressure indicauuns and fluctuating LPSI pump motor current. Because the operators are cuefully munitoring RCS level, it is likely that the failed segment would be identified and isolated.
Spatial Effects: Local Affected location: Roorn 2014 Spatial Effects Comments: A review of Plant Design Drawing M 2046 and Figure 3.6 2 at elevation 317 0" of the ANO 2 IPE (Report 94 R 2005-04, Rev. 0) indicates that this line segment is located in flood zone RAB 2014 LL. IIPhl pump 2P 89A, LPSI pump 2P 60A, CS pump 2P 35A, SDC heat exchanger 2E 35A and associated valves are located in this flood zone. The ANO 2 Internal Flood Screening Study (Calc. 89 E 0048 35, pg.13) assumes the failure of all components in the flood initiation zone, llence for this SDC line break scenario, it is highly probable that the pumps in train 'A' of the ECCS uill be affected by flooding, spraying orjet impingement.
Dunng the walkdown it was observed that the outflow of water from the failed segment can be contained within the flood zone because the entrance door is water tight, and it is maintained closed. Although a propagation patinvay via the ventilation duct exists, it is located several feet above the floor level. The strategic location of the ventilation damper and the ECCS pump room level indication in the Control Room ensure that there uill be sufficient time for detecting and isolating the failure or containing the water in the flood initiation zone. Therefore for this SDC line break scenario, propagation of water from the flood initiation zone to the other ECCS pump rooms is very unlikely but suction to the LPSI pumps would be lost.
Initiating Event: 1 initiating Event ID: A Initiating Esent R v.9very: One of the redundant makeup sources is required for mitigation and to accominodate decay heat boil off. Manual initiation of makeup is required.
less of System: S System IPE ID:
SDP System Recovery: No recovery of SDC is expected in the near term. By closing SDC suction valve 2CV 50841, 2CW5086 2 or 2CV 50381 from the control room, the failed segment can be isolated. It is assumed that the operators ability to isolate the break during mid loop operation is equivalent to having one backup train.
FMECA Consequence Infonnation Report Ca'a'Ishcm No. A.rENG C4&0ld, Rev. 00 144ep97 Page All o/A41 laas of Trala
'I%3 Trale ID:
l LPSI pump 2P40A Trals Reco$en: Once shutdown cooling is initiated, CS function is not needed Although HPSI pump "A" is lost, the redundant source for makeup would not be aNocted. No recovery of the asected trains is not expected in the near term.
Consequence Cominent: The segment failure causes a loss of reactor coolant and the unavailability of SDC during mid-loop operation. During this mode of shutdown operation, the operators are carefully monitoring RCS imentory due to the importance of this configuration, but isolation may be irrelevant because RCS level is already near the invert of the reactor nor21es. An increase in the level of operator awareness is assumed. The Shutdowri Operations Protection Plan utilfred at ANO 2 requires that redundant sources and associated paths are in place to provide makeup during mid-loop operation, Thus, there would be at least two equivalent backup available for mitigating this type of LOCA.
Because there are two backup trains available, the resulting consequence is MEDIUM.
This is based on the Table 3.4 of the EPR1 procedure (EPRI TR.106706).
Consequence Category: MEDIUM O
Coaaequence Rank D
l-O
FMECA - Consequence Information Report Calcularkm Na A PENG.CM0ld. Ret 00 l4 sep 91 Page All of A42 Consequence ID: LPSI-C 06 Consequence
Description:
Loss of LPSI pump 2P-60A,IIPSI pump 2P 89A, and CS pump 2P 35A occun due to a line break in ECCS pump room "A" following a LOCA.
Break Slict Large Isolability of Break Yes ISO Comments: The break is postulated to occur either during normal power operation (i.e., periodic testing of LPSI pumps) or during a response to a LOCA demand. Because the operators are highly stressed during a LOCA event and the longer fault exposure time preceding the detection of the failure, it is assumed that the limiting consequence described herein is associated with a LOCA demand. The piping from downstream of manual vah es 2SI 1 A and 2SI 2A to LPSI pump 2P.
60A discharge chock valve 2SI 3 A is included in tids segment. The piping in the pump mini-flow path from tM ; mp discharge to upstream of vahr 2CV 5123 1 is also included in this l
segment. Tids consequence includes all wcids in lines 2GCB 1 12",2GCB 1 14", 2GCD 508-2" and the welds in the applicable portions oflines 2GCB 3 8",2GCD 312" and 2DCB 504 2".
A failure in this segment would cause the diversion of LPSI flow to ECCS pump room "A".
LPSI flow diversion would result in an increasing water level in ECCS pump room "A" which is annunciated in the control room. Other unexpected alarms and indications include low LPSI pump discharge flow and low LPSI purnp discharge prescure. Depending on the break location, fluctuation in LPSI pump motor current may also be s' sed in the control room. Although unexpected alarms and indications would provide sum ;nt information to identify and isolate the break, the timeframe available for detection is limited such that detection may not occur.
Ilowever, train isolation is provided by the respective pump discharge check valve.
Spatial Effects: Local Affected location: Room 2014 Spatial Effects Comments: A review of Plant Design Drawing M 2046 and Figure 3.6 2 at rievation 317' 0" of the ANO 2 IPE (Report 94 R 2005-04, Rev. 0) indicates that this line segment is located in flood zone RAD 2014 LL llPSI pump 2P 89A, LPSI pump 2P-60A, CS pump 2P 35 A, SDC heat exchanger 2E 35A and associated valves are located in this flood zone. The ANO 2 Internal Flood Screening Study (Calc. 89 E 0048 35, pg.13) assumes the failure of all components in the flood initiation zone, llence for this LPSI line break scenario, flooding of the pumps in train "A" of the ECCS is highly probable.
During the walkdown it was observed that the outflow of water from the failed segment can be contained within the flood zone because the entrance door is water tight, and it is maintained closed. The propagation pathway via the ventilation duct will be isolated by the closure of the ventilation damper upon receipt of SlAS.
Therefore for this LPSI line break scenario, the propagation of water from the flood initiation zone to the other ECCS pump rooms will not occur.
Initiating Event: N Initiating Event ID: N/A Initiating Event Recoverv: N/A Loss of S stem: N System IPE ID:
N/A 3
System Recovery: N/A IAss of Train: TM 3 Train ID:
FMECA Consequence Inforenation Report Calaslaman No. A PDGC4LC Old. Aw,00 O
te.ser-el Pese Al. of A42 Trala Recovery: No recovery of the purnps in train 'A' of the ECCS is expected prior to a plant cold shuidowrt Although one train of ECCS is lost, the redundant train is still capable of performing its intended design function (i.e., mitigation of a LOCA). The failed segment can be isolated b' closing R%T discharge valve 2CW56301 and the associated containment sump isolation i
valve in train 'A' to prevent flow diversion during the recirculation mode.
Consequesee Coasment: For the case where the failed segment is negrer,ated from train 'B' by the reclosure of check valve 2SI 3 A, train 'A' of ECCS (l.c., !! PSI, LPSI, and CS) would become unavailable due to flooding. The remaining train of ECCS will not be affected and is l
s'tallable for mitigating a LOCA, thus for this caec there is one backup train.
For the case where the failed segment remains unisolated (i.e., check nive 2513 A fails to reclose), train 'A' of ECCS would be lost due to flooding of the pumps in ECCS pump toom 'A*. The discharge from LPSI pump *B" would also be lost due to flow diversion. Since the probability of the check valve failing to reclose is approximately 2.0E 4 (ANO 2 IPE), failure to isolate is treated as an equivalent I
backup train. 'Iherefore, for either of the cases considered there is one equivalent backup train available for mitigating a large LOCA.
Periodic testing (i.e., pressurir.ing to operating pressure) of this line segment is performed on a quarterly basis during normal power operation. A between test
' exposure time'is therefore assumed Based on quarterly testing and the availability of one equivalent backup train of ECCS, and based on Table I and the guidance provided in Table 3.2 of the EPRI procedure (EPRI 'Ibl06706), a MEDIUM consequence category is assigned.
There are two active barriers (2CW5649 1 and 2C%$647 1) to protect against containment bypass. Both valves must fail to close in order for the containment to be bypassed Thus, the potential impact of the failure on containment performance is also a MEDIUM conasequence (Table 3.3. of EPRI procedure TR.106706).
Consequence Category: - MEDIUM C
Consequesee Rank O
FMECA - Consequence Information Report Cahtaam Na A. Pac. calc 016. /tcr oo l4-s 91 Page A14 of A41 m
Consequence ID: LPSI C 07 Consequence
Description:
less of LPSI pumps 2P 60A & 2P 60B, HPSI pun f 2P-89A and CS pump 2P 35A occurs due to a line break in ECCS pump room "A" following a LOCA.
Break Size.
Large Isolability of Break: No 150 Comments: The break is postulated to occur either during normal power operation (i.e., periodic testing of LPSI pumps), shutdown cooling or in response to a LOCA demend. Because the operators are highly stressed during a LOCA event and the relatively short time for which LPSI is needed, it is assumed that the limitics consequence described herein is associated with a LOCA demand.
The piping from downstream of check valves 2SI 3 A to upstream of manual valve 2SI-4 A, and to the inaccessible wall penetration and the ceiling penetration at elevation 335'-0" is included in this segment. This consequence includes all welds in line 2GCD 3 14' and the welds in the applicable portions oflines 2GCB 312" and 2GCD 7-8".
A failure in this segment would cause the diversion of LPSI flow from both pumps to ECCS pump room " A". LPSI flew diversion would be indicated by an increasing water level in ECCS pump room 'A' which is annunciated in the control room. Other unexpected alarms and indication include inappropriately low LPSI pump discharge flow and inappropriately low LPSI pump discharge pressure. Because of the short duration of LPSI following a large LOCA, the operator response time for identifying the location and isolating the failed segment may not be adequate even though unexpected alarms and indications are encountered in the control room.
For a smaller LOCA, the duration of LPSI would be longer which makes isolation of the failed segment more likely. It is very likely that HPSI pump " A" would be flooded because the pump motor is mounted close to the Door.
Spatial Effects: Local Affected 14 cation: Room 2014 Spatial Effects Comments: A review of Plant Design Drawing M 2046 and Figure 3.6 2 at elevation 317'-0" of the ANO 2 IPE (Report 94.R 2005 04, Rev. 0) indicates that this line segment is located in Dood zone P AP 2014 LL. HPSI pump 2P-89A, LPS! pump 2P 60A, CS pump 2P 35A. SDC heat exchanger 2E 35A and associated valves are located in this ficod zone. The ANO 2 Internal Flood Screening Study (Calc. 89 E 0048 35, pg.13) assumes the failure of all components in the flood initiation zone. Hence for this LPSI line break scenario, ficoding of the pumps in train "A" of the ECCS is highly probable, During the walkdown it was observed that the outflow of wates fron. the failed segment can be contained within the Dood zone because the entrance door is water tight, and it is maintained closed. The propagation pathway via the vertilation duct will be isolated by the closure of the ventilation damper upon receipt of SIAS.
Therefore for this LPSI line break scenario, the propagation of water from the flood initiation zone to the other ECCS pump rooms will not occur.
Initiating Event: N Initiating Event ID: N/A Initiating Event Recovery: N/A less of Sy stem: S System IPE ID:
LPSI/SDC System Recovery: Both the LPSI and the shutdown cooling functions are lost due to the break. Based on the ANO-2 IPE, there are no backup trains available to perform LPSI function. No recovery of the LPSI pumps is expected.
l
FMECA Consequence Infor1mation Report Cah n A rEW CUf 016.h. 00 rw AU of Au Laos of Trais: 1%2 Train ID:
HPSI train A, CS train A Trale Recovery: No recovery of HPSI train 'A' and CS train 'A' is expected. Although one train of those systems is lost, the redundant train would still be available. The failed segment can be isolated by closing RWT discharge valve 2CW56301 LPSI pump *B" manual suction rahe, and the associated containment sump isolation valves in train 'A",
Consequence Comanent: Because of the common discharge header for both LPSI pumps. LF31 mill be lost in response to a large LOCA whether or not the failed segment is isolated. Therefore, there will be no available backup trains (see Section 4.1.3) to prmide LPSI mitigation.
According to the ANO 2 IPE, a loss of LPSI will cause core damage. Although loss of LPSI results in core damage, it is still desirable to isolate the failed segment in order to minimize the impact on containment performance. For smaller LOCAs where the RCS preneure remains above the shutoff head of the LPSI pumps, the failed segment must be isolated to ensure the operability of HPSI during recirculation. The longer time for switching over to teritculation makes it highly hkely that the failed segment would be isolated. However, the resulting consequence would be less risk signtAcant than the case for a large LOCA.
Because there are no backup trains to accomplish the LPSI function in response to a large LOCA, and based on Table I and the guidance prmided in Table 3.2 of the EPRI procedure (EPRI TR.106706), a HIGH consequence category is assigned.
Consequesee Category: HIGH C
Consequence Rank O
O
FMECA - Consequence Infortnation Report Calcalana V '4 Pf*G C4M 018 R'" 00 16-ber 91 Page A16 of A42 Consequence ID: LPSI C 08 Consequence Desedption: Loss of LPSI pumps 2P40A and 2P40B occurs due to a line break in Tendon Gallery Access Arra following a LOCA.
Break Slee:
Large l>olability of Break No ISO Lomments: The break is pouulated to occur either during normal power operation (i.e., periodic testing of LPSI pumps) or during a response to a LOCA demand. Because the operatori are highly stressod during a LOCA event and tiu. longer fault exposure time for detecting the faihtre, it is assumed that the limiting consequence desciibed herein is associated with a LOCA demand.
The LPSI piping between the wall penetrations in the Tendon Gallery Access area is included in this segment. This consequence includes the welds in the applicable portion ofline 2GCB 3-12".
A failure in Utis segment would cause the diversion of LPSI flow from both pumps to the Tendon Gallery Access Area LPSI flow diversion would be indicated by an increasing water level in the Reactor Auxiliary Duilding sump which is annunciated in the control room. Other unexpected alarms and indications include inappropriately low LPSI pump discharge flow and inappropriately low LPSI pump discharge pressure. Because of the short duration of LPS!
following a large LOCA, the operator response time for identifying the location and isolating the failed segment may not be adequate even though unexpected alarms and indications are encountered in the control room. For a smaller LOCA, the duration of LPSI would be longer w hich makes isolation of the failed segment more likely.
Spatial Effects: Local Affected imention: Room 2011 Spatial Effects Comments: A review of Plant Design Drawing M 2046 and Figure 3.6 2 at elevation 317 0* of the ANO 2 IPE (Report 94 R 2005 04, Rey,0) indicates that this line segment is located in flood rone RAB 20ll LL Smice water loop I to ESF equipment isolation valve 2CV 14001 is located it this flood zone. The ANO-2 Internal Flood ecreening Study (Calc. 89 E 0048 35, pg.13) assumes the failure of all components in the flood initiation zone it should be noted that valve 2CV 14001 is normally open and receives a confirmatory open signal following a LOCA. If this valve were impacted by spraying orjet impingement it would therefore fail in its safe shutdon por.ition. Because the ventilation paths are isolated following a LOCA, the propagation of water from the flood initiation zone to the ECCS pump rooms is unhkely. Hence for this LPSI line break scenario, flooding of the ECCS pumps is not a concern.
As observed during the wa!Ldown, the outflow from the break can propagate from the flood initiation zone to the General Access Area. These two flood zones are adjacent to each other, and there are no physical barriers between them. The combmed area of these (bod zones can accommodate a significant amount of water without threatening the ventilation dampers, initiating Frent: N Initiating Event ID: N/A Initiating Esent Recovery: N/A Loss of System: S System IPE ID:
LPSI/SDC System Reem cry: Both the LPSI and the shutdown cooling functions are lost due to the break. Based on the ANO-2 IPE, there are no backup trains available to perform LPSI function. No recmtry of the o
FMECA Consequence laformation Report Colndesan h A rENG C4M 0/6. M 00 O
tstwn rese Ali of An LPSI pumps is expected in the near term. Flow through both LPSI pumps must be isolated to maintain the operability of HPSI and CS The failed segment can be isolated by closing the manual suctiot. valves for LPSI pumps 'A' and 'B".
Less of Trains N Train ID:
N/A Train Recovery:
Consequesee Coenasent: Because of the common discharge header for both !?St pumps, LPSI will be lost in response to a large LOCA wl ether or not the failed segment is isolated. Therefore, there will be no availabt? backup trains (see Section 4.1.3) to prmide LPSI mitigation.
l According to the ANO 2 IPE, a loss of LPSI will cause core damage Although loss of i
LPSI results in core damage, it is still desirab'- M inolate the failed segment in order to minimize the impact on containment performance. For smaller LOCAs where the RCS pressure re.nains above the shutoff head of the LPSI pu nps, de failed segment must be isolated to ensure the operability of HPSI during recirculation. The longer time for switching mer to recirculation makes it highly likely that the failed segment would be isolated. However, the resulting consequence would be less risk +1gnificant than the case for a large LOCA.
Because there are no backup trains to accomplish the LPSI function in response to a large LOCA, and based on Table I and the guidance prmided in Tabee 3.2 of the EPRI procedure (EPRI TR.106706), a HIGH consequence category is assigned.
Conseguemee Category: HIGH _
C Consequesee Rank O
O O
FMECA - Consequenet information Report Caledaa'n No A PENG CALC 0/Utrv. 00 l4.s 91 Page AIB qf A42 w
Consequence ID: LPSI C 10 Consequence
Description:
- 1. css of reactor coolant and train 'B" of the ECCS pumps occurs due to a SDC line break in ECCS pump room 'B",
Break Stre Large isolability of Break: Yes ISO Comments: The break is postulated to occur during a demand for shutdown cooling, and in the piping tom floor penetration at elevadon 335' 0" to upstream of manual vahe 2S1 18. Of the various plant shutdown states considered (see Sections 4.4.2 and 4.4.5), mid loop operation is the most risk.
significant because of the relathcly short time it would take to uncover the core. Therefore, only the consequence resulting from a pipe break during ndd loop operation is described herein. This consequence includes the welds in the applicable portion ofline 20CB 5 14".
A failure in this segment would cause reactor coolant to be diverted to ECCS pump toom *B" during mid loop operadon. Several unexpected alarms and indications would be encountered in the control room following the segment failure. These include increasing or high water level in the ECCS pump toom 'B", decreasing RCS level, low LPSI pump discharge flow and pressure indications and fluctuating LPSI pump motor current. Because the operators are carefully monitoring RCS level, it is likely that the failed segment would be identified and isolated.
Spatial Effects: local Affected Location: Room 2007 Spatial Effects Comments: A review of Plant Design Drawing M 2046 and Figure 3.6 2 at elevation 317' 0" of the ANO 2 IPE (Renort 94 R 2005 04, Rev. 0) indicates that this line segment is located in flood zone RAB 2007 LL. IIPSI pump 2P-89B, LPSI pump 2P 60B, CS pump 2P 35B, SDC heat exchanger 2E 35B and associated vahts are located in this We flood zone. The ANO 2 Internal Flood Screening Study (Calc. 89 E 0048 35, pg.
- 13) assumes the failure of all components in the flood initiation zone. llence for this SDC Ime break scenario, it is highly probable that the pumps in train 'B" of the ECCS will be affected by flooding, spraying orjet impingement.
During the walkdown it was observed that the outflow of water from the failed segment can be contained within the flood zone because the entrance door is water tight, and it is maintained closed. Although a propagation pathw1 y sin the ventilation duct edsts, it is located several feet above the floor lestl. The strategic location of the ventilation damper and the indication of ECCS pumr oom les el in the Control Room ensure that there will1,e sufficient time for detectQtg and isolating or containing the water in the flood initiation zone. Therefore for th's SDC line break scenario, propagation of water from the flood initiation zone to the other ECCS pump rooms is very unlikely but suction to the LPSI pumps would be lost.
Initiating Esent: 1 initiating Event ID: A Initiating Event Reemcry: One of the redundant makeup sources is required for mitigation and to acconunodate decay heat boil off. Manual initiation of makeup is required.
less of System: S System IPE ID:
SDC S stem Recovery: No recostry of SDC is expected in the near term. By closing SDC suction valve 2CV-5084 1, 3
2CV 5086 2 or 2CV 50381 from the control room, the failed segment can be isolated. It is assumed that the operators ability to isolate the break during mid-loop operation is equivalent to having one backup train.
l i
l
FMECA ConsequenceInfonnation Report Ca.blane h A PENG CGC Old, fin 00 O
ls-sw91 Page Al9 of A42 Lees of Train: TM 3 Trale ID:
HPSI train B. CS train B, LPSI pump 2P 60B Trale kecovery: Once shutdoun cooling is initiated. CS function is not needed Although HPSI pump *B" is e
lost, the redundant source for makeup would not be affected. No recovery of the affected trains i
is expected in th:near term.
Consequence Comment: The segnwnt failure causes a loss of reactor coolant and the unavailability of SDC during mid loop operation. During this mode of shutdown operation, the operators are I
carefully monitoring RCS inventory due to the importance of this configuration, but isolation may be irrelevant because RCS level is already near the imtat of the reactor noules. An increase in the level of operator awareness is assumed. The Shutdown Operations Protection Plan utilized at ANO 2 requires that redundant sources and associated paths are in place to provide makeup during mid loop operation. Thus, there would be at least two equivalent backup available for mitigating this type of LOCA.
Because there are two backup trains available, the resulting consequence is MEDIUM.
This is based on the Table 3.4 of the EPRI procedure (EPRI 1R.106706).-
Consequence Category: MEDIUM O
C.eque ce na.k O
O O
FMECA - Consequence Infortnation Report Cahlows & A.PENG CAM 0M. &v 00 14 sty 91 Page A20 cf A42 Consequence ID: LPSI C Il Consequence
Description:
less of LPSI pump 2P 608,IIPSI pump 2P 89B, and CS pump 2P 35B occurs due to a line break in ECCS pump room "B" following a LOCA.
Break Size:
Large Isolability of Break: Yes ISO Comments: The break is postulated to occur either during normal power operation (i.e., periodic testing of LPSI pumps) or during a response to a LOCA demand.13ecause the operators are highly l
stressed during a LOCA event and the longer fault exposure time preceding the detection of the failure, it is assumed that the limiting consequence described here!. u associated with a LOCA demand. 'The piping from downstream of manual vahrs 251 1B and 2SI 2B to LPSI pump 2P.
60B discharge check valt: 2SI 3B is included in this segment. The piping in the pump mini-flow path from the pump discharge to upstream of valve 2C%5124 1 is also included in this segment. This consequence includes all welds in lines 2GCB 212",2GCB 214",2GCD 509-2" and the welds in the applicable portions oflines 2GCB 3 8",2GCB 312" and 2DCD 504 2".
A failure in this segment would cause the diversion of LPSI flow to ECCS pump room "B".
LPSI flow diversion would cause an increasing water level in ECCS pump room "B" which is annunciated in the control roon,. Other unexpected alarms and indications include low LPSI pump discharge flow and low LPSI pump discharge pressure. Depending on the break location, fluctuation in LPSI pump motor current may also be alarmed in the control room. Although unexpected alarms and indications would prmide sufficient information to identify and isolate the break, the timeframe available for detection is limited such that detection may not occur.
Ilowever, train isolation is prmided by the respective pump discharge check vahr.
Spatial Effects: Local Affected location: Room 2007 Spatial Effects Comments: A review of Plant Design Drawing M 2046 and Figure 3.6 2 at elevatior. 317'-0" of the ANO 2 IPE (Report 94 R 2005~4, Rev. 0) indicates that this lity, s gment is located in flood zone RAB 2007 LL. HPSI pump 2P-800, LPSI re mp 2P-60B, CS pump 2P 35B, SDC heat exchanger 20 358 and associated valves are located in this flood zone. The ANOV Internal Flood Screening Study (Calc. 89 E-0048 35, pg.
- 13) assumes the failure cf all components in the flood initiation zone, llence for this LPSI line break scenario, Dooding of the pumps in train "B" of the ECCS is highly probable.
During the walkdown it was observed that the outflow of water from the failed segment can be contained within the flood zone because the entrance door is water tight, md it is maintained closed. The propagation pathway via the ventilation duct will in isolated by the closure of the s entilation damper upon receipt of SIAS.
rherefore for this LPSI line break scenario, the propagation of water from the flood initiation zone to the other ECCS pump rooms will not occur.
Initiating Event: N Initiating Event ID: N/A laitiating Event Recovery: N/A Loss of S stem: N Systcm IPE ID:
N/A 3
System Recovery: N/A less of Train: 'IW3 Train ID:
FMECA - Consequence Inforenatien Report Cahlaaaa N" d PENG CAM 018 d'" 00 14 sep91 Pegt A21 of A42 Train Recovery: No recovery of the pumps in train 'B' of the ECCS is expected prior to a plant cold shutdown.
Although one train of ECCS is lost, the redundant train is stil; capable of performing its intended design function (i.e., mitigation of a LOCA). The failed segment can be isolated by closing R%T discharge valve 2C%56312 and the associated containment sump isolation vaht in train 'B' to provent dh ersion during the recirculation mode.
l Consequence Coestment: For the case w here the faut.d segment is segregated from train "A" b/ the reclosure of check valve 2St.3B, train "B" of ECCS (i e., HPSI, LPS1, and CS) would become unavailable due to flooding The remaining train of ECCS will not be affected and is available for mitigating a LOCA. thus for this case t'.cre is one backup train.
For the case where the falted sepnent remains unisolateJ (i.e., check valve 2SI 3B fails l
to recmse), train 'B' of ECCS wuld be lost due to floodirig ff the pumps in ECCS pump room "B". 'Ihe discharge frorn LPSI pu:np "A" would also be lost due to flow diversior. Sir.cc the probability of the check valve failing to reclovs is approximately 2.0F-4 ( ANO 2 IPE), failure to isolate is treated as an equivalent backup train.
Thereforc, for either of ti.e cases consid: red there is ont e,gh ilcnt tackup train available for mitigating a large LOCA.
Periodi: testing (i c., pressurirJng to openiing presure) of this line regmet,t is performed on a quarterly basis during normal power t.peration. A between test
" exposure time" is ther: fore assumed. Based on quarictly testing and the availability cf one equivalent backup train of ECCS, and based on Table I and the guidance provided in Table 3.2 of the EPRI procedure (EPRI TR.106706), a MEDIUM consequence category is assigned.
There are two active barriers (2C%5648 2 and 2C%5650 2) to protect against contalament bypass. Both valves must fail to close in order for the containment to be bypused. Thus, the potential impwt of the failure on containment perfomiance is also a MEDIUM consequemx (Table 3.3. of EPRI procedure TR.106706).
Consequcace Catepr): MEDIUM O
Conseque.ec na.k O
O
FMECA Consequence Information neport Cahla'*' Na A PENG-C4LC OH. hs 00 te-scr l Pay A22 of M2 a
Consequence ID: LPSI C 12 Conner,uence
Description:
less of LPSI pumps 2P40A & 2P 60D, llPSI pump 2P-89B and CS purnp 2P 350 occurs due to a line break in ECCS pump room "D" following a LOCA.
Break Size:
Large Isolability of Break No ISO Comma,s: The break is postulated to occur either during normal power operation (i c., periodic testing of LPSI pumps) or dudng a response to a LOCA demand. Because the operators are highly stressed during a LOCA event and the relatively short time for which LPSI is needed, it is assumed that de limiting consequence described herein is assoctated with a large LOCA demand. The piping from downstream of check valves 2SI 3D to upstream of manual valve 2S1-4D, and to the wall penetration is included in this segment. 'this consequence inch &s the welds in the applicable portion ofline 2GCB 3 12".
A failure in this segment would cause the dh crsion of LPSI flow from both pumps to ECCS pump room "B", LPSI flow diversion would be indicated by an increasing water level in ECCS pump room "B" which is annunciated in the control room. Other unexpected alarms and indication include inappropriately low LPSI pump discharge flow and inappropriately low LPSI pump discharge pressure. Because of the short duration of LPSI following a large LOCA, the operator response time for identifying the location and isolating the failed segment may not be adequate even though unexpected alarms and indications are encountered in the control room.
For a smaller LOCA, the duration of LPSI would be longer which makes isoladon of the failed segment more likely, it is very likely that ilPSI pump "B" would be flooded because die pump motor is mounted close to the floor, Spatial Effects: Local Affected location: Room 2007 Spatial Effects Comments: A review of Plant Design Dr. ming M 2046 and Figure 3.6 2 at elevation 317'-0" of the ANO-2 IPE (Report 94 R 2005-04, Rev. 0) indicates that this line segment is located in flood zone RAB 2014 LL. IIPSI pump 2P-89A, LPSI pump 2P 60A, CS pump 2P 35 A, SDC heat cxchanger 2E 35A and associated valves are located in this Rood zone. The ANO-2 Internal Flood Screening Study (Calc. 89 E 0048 35, pg.13) assumes the failure of all components in the flood initiation zone. Ilence for this LPSI line break scenario, flooding of the pumps in train "A" of the ECCS is highly probable.
During the walkdown it was observed that the outflow of water from the failed segment can be contained within the flood zone because the entrance door is water oght, and it is maintained closed. The propagation patinvay via the ventilation duct will be isolated by the closure of the ventilation damper upon receipt of SI AS.
'Iherefore for this LPSI line break scenario, the propagation of water from the flood initiation zone to the other ECCS pump rooms will not occur.
Initistlag Event: N Initiating Event ID: N/A Initiating Esent Recovery: N/A Lass of System: S System iPE ID:
LPSI/SDC System Recovery: Both the L"Si and the shutdown cooling functions are lost due to the break. Based on the ANO 2 IPE, there are no backup trains r.vailable to perform LPSI function. No recovery of
(
the LPSI pumps is exnected.
FMECA - Consequence Infersnation Report Cablemas h A#ENG4G0/6. h. 00 O
ts-s ei Pese AH era 42 w
taas of Train: "I%2 TralsID:
HPSI train B, CS train B Yrale Recover)1 No roomery of HPSI train 'B' and CS train 'B' is expected. Although one train of those systems is lost, the redundant train would still be available. The failed segment can be isolated by closing RWT discharge valve 2CV.36312 LPSI pump *A' manual suction valve, and the associated containment sump isolation valm in train 'B'.
Consequence Comment: Because of the common discharge header for both LPSI pumps, LPSI will be lost in response to a large LOCA whether or not the failed segment is teolated. Therefore, l
there will be no available backup trains (see Section 4.1.3) to prmide LPSI mitigation.
According to the ANO 2 IPE, a loss of LPSI will cause core damage. Although loss of l
' LPSI results in core damage, it is still desirable to isolate the failed segment in order to minimize the impact on containment performance. For smaller LOCAs where the RCS pressure remains abm, the shutoff head of the LPSI pumps, the failed segment must be isolated to ensure the operability of HPSI during recirculation. The longer time for switching over to recirculation makes it highly likely that the failed segment would be isolated. Hmmer, the resuhing consequence would be less risk-sigalAcant than the case for a large LOCA.
Because there are no backup trains to accomplish the LPSI function in response to a large LOC A, and based on Table I and the guidance prmided in Table 3.2 of the EPRI procatwe (EPRI 1R.106706), a HIGH consequence category is assigned.
Consequence Categor>1 HIGH O
C.eq.e.ce Ra k O
O
'O
Cal'viemm Na A.rna catc-i un oo j
FMECA Consequence Information Report ss.s -91 Page A24 of A42 m
)
I Consequence ID: LPSI C 13 Consequence
Description:
Loss of LPSI occurs due to a line break in the lower south piping penetration area following a LOCA.
Break Slic:
Large Isolability of Breakt No ISO Comments: The break is postulated to occur either during normal power operadon, shutdown cooling operation, or in response to a LOCA demand. Because the operators are highly stressed during a LOCA event and the relath cly sheet time for which LPSI is needed, it is assumed that the "miting consequence described herein is associated with a large LOCA demand. The piping from the floor penetration at elevation 335' 0" to the ceiling penetrations at elevadon 360' 0" within the lower south piping room is included in this segment. The SDC return line from downstream of valves 2CV 5093 and 2515093 3 to the LPSI header is also included. This consequence includes all welds in line 20CD 7 4" and the welds in the applicable portions of lines 2GCD 814",2GCB 8 8",2GCD 714",20CB 710",2GCB 7 8" and 2GCB 74".
A failure in this segment would cause the diversion of LPSI flow from the discharge header to the general access area of the Reactor Auxiliary Building (RAD)(Calc. 89 E 0048 35, pg. 28).
The diversion of LPSI flow would be indicated by an increasing RAB sump water level which is annunciated in the control room. Depending on the break location, LPSI header low flow and low pressure may also be indicated in the control room. Because of the short duration of LPSI following a large LOCA, the operator response time for identifying the location and isolating the failed segment may not be adequate even though unexpected alarms and indications are encountered in the control room. For a smaller LOCA, the duration of LPSI would be longer w hich makes isolation of the failed segment more likely, Spatial Effects: Local Affected location: Room 20$$
Spatial Effects Comments: A review of Plant Design Drawing M 2045 and Figure 3.6 2 at elevation 335' 0" of the ANO 2 IPE (Report 94 R 2005-04, Rev. 0) indicates that this line segment is located in flood zone RAB 2055 JJ in the Reactor Auxiliary Building (RAB). The l{ PSI orifice bypass valves (2CV 5103 1 and 2CV 5104 2), LPSI header isolation valve 2CV 5091, SDC heat exchanger return valve 2CV 5093 and EFW pump flush valve 2CV 07141 are located in this zone. The ANO 2 Internal Flood Screening Study (Calt 89 E 0048 35, pg.13) assumes the failure of all cor.iponents in the flood initiation zone. For a limiting line break during a demand for LPSI following a LOCA, it was observed during the walkdown that the force exerted on the entrance non water tight door would cause the door to open before a significant amount of water can accumulate inside the room and flood the valve operators. It was also observed that spraying orjet impingement may affect certain valves (2CV 5103 1, 2CV 5093 and 2CV 5091) because of the close proximity of these valves to the line where the failure is postulated. It should be noted that the HPSI and LPSI valves within this flood zone are not required to change their positions in order to support ilPSI and LPSI functions. Thus the impact of spatial effects on these vah es is insignificant.
The walkdown revealed that the outflow of water from the flood initiation zone can l
propagate through the non water tight door to the adjacent tank rooms, pump rooms i
I and corridor (i.e., flood zone RAB 2040-U) at elevation 335' 0". From this flood zone the water can eventually propagate to the RAB sump of the General Access Area at elevation 317'-0" via stairway No. 2001 and the floor drain system. Water l
FMECA Consequence Information Report C*bina= Na A.trua-cate.on an. oo as see s1 Pese A23 of A42 can also propagate from the flood inidation zone to the RAB sump via the floor drain system. Because the ECCS pump room ventiladon dampers are closed by SIAS, spillmcr from the General Access Area to the ECCS pung rooms will hat occur for the LPSI line break scenario considered. However, LPSI function will be lost due to the break.
Ir:tleting Event: N laitiating E.ent ID: N/A initiating Event Reemery: N/A Ims of System: S
$ steen IPE ID:
LPSI 3
S) stem Recovery: Both LPSI and SDC tbnetions are lost due to the line break. Based on the ANO 2 IPE, there are no backup trains available to perform LPSI funedon within the required time period. No recovery of LPSI/SDC system is expected in the near term. The failed segment can be isolated by closing the manual suction valves for LPSI pumps "A" and "B",
Ims of Trala: N Train ID:
N/A Train Recovery: N/A Consequence Comment: Because of the common discharge header for both LPSI pumps, LPSI will be lost in response to a large LOCA whether or not the failed segment is isolated. Therefore, there will be no available backup trains (see Section 4.1.3) to prmide LPS! mitigation.
According to the ANO 2 IPE, a loss of LPSI will cause core damage. Although loss of LPSI results in core damage, it is still desirable to islate the failed segment in order to minimize the impact on containment performance For smaller LOCAs where the O
RCS pressure remains above the shutoff head of the LPSI pumps, the failed segment must be isolated to ensure the operability of HPSI during recirculation. The longer time for switching over to recirculadon makes it highly likely that the failed segment would be isolated. However, the resulting consequence woted be less risk significant than the case for a large LOCA.
Because there are no backup trains to accomplish the LPSI function in response to a large LOCA, and based or Table I and the guidance prmided in Table 3.2 of the EPRI procedure (EPRI TR 106700), a HIGH consequence category is assigned.
Consequemee Category: HIGH O
Co.se, eeRa.k O
O 4
E
FMECA - Consequence Information Report Cablata h. A PENG CEC 016, b. 00 t4-ser 91 Pagt A26 of A42 Consequence ID: LPSl C 14A Consequence Descr'ption: loss of LPSI occurs due to a line break in injection path to RCS cold leg 2P32 A in the upper south piping penetration room following a large LOCA.
Break Size:
Large Isolability of Breakt No ISO Comments: The break is postulated to occur either during normal power opennie, t hutdown cooling or in response to a LOCA demand. Because the operators are highly stressed during a LOCA esent and the relatively short time for which LPSI is neuled, it is assumed that the limiting consequence for the three configtrations is associated with a large LOCA demand. The piping for from the floor penetration at elevation at 360' 0" to containment penetration 2P15 is included in this segment. This consequence includes the welds in the applicable portions of lines 2CCB-41.5",2GCB 7 6" and 2CCB-4 6".
A failure in this segment would cause the diversion of LPSI flow from the irdection hne to the general access areas of the Reactor Auxiliary Building (RAB) (Calc. 89 E-0048-35, pg. 28).
The diversion of LPSi flow would be indicated by an increasing RAB sump water lestl which is annunciated in the control room. A low LPSI 1.eader pressure would also be indicated and annunciated in the control room. Because of ti c short duration of LPSI following a large LOCA, the operator response time for identifying the location of the failed segment and isoladng the failed segment may not be adequate even though unexpected alarms and indications are encountered in the control room. For a smaller LOCA, the duration of LPSI would be longer which makes isolation of the failed seg nent more likely, Spatial Effects: Local Affected Location: Room 2084 Spatial Effects Comments: A review of Plant Design Drawings M 2044 and M 2063, and Figure 3.6-2 at elevation 354' 0" of the ANO-2 IPE (Report 94 R 2005 01, Rev. 0) indicates that this line segment is located in flood zone RAB 2084 DD. The HPSI, LPSI and CS header valves, SDC suction vahr 2C%5038 1, senice water isolation vahts for CCS units 2VCC 2A and 2VCC-2B, EFW distribution valves to steam generator 2E-24A are located in this flood zone. The ANO 2 Internal Flood Screening Study (Calc. 89 E-0048 35 pg.13) assumes the failure of all components in the flood i itiation zone. This assumption is too conservative for this evaluation. For a limiting line break during a demand for LPSI following a LOCA, it was obsened during the walkdown that the force exerted on the entrance non-water tight door would cause the door to open before a significant amount of water can accumulate inside the room and flood the valve motors. It was also observed that spraying orjet impingement may affect those valves which are in close proximity to the line segment where the failure is postulated. It should be noted that the HPSI, LPSI and CS injection line isolation valves and CCS senice water isolation valves are normally closed and open automatically upon receipt of an ESF actuation signal.
These "alves would have already been opened at the time LPSI line break is postulated to occur. The impact of spatial effects on these valves would cause them to fail as-is in the safe shutdown position. Flooding of all valves within the flood initiation zone is considered to be insignificant, and the impact of spraying orjet irnpingement on the HPSI, LPSI, CCS and CS vahrs is also considered to be insignificant. However, the EFW distribution vahts to steam generator 2E 24 A which are closes may be impacted because if needed they would be required to open after the segment failure occurs
- FMECA - Consequence Infornsation Report Calcalem No. A.PENG<ALC-016. Rev. 00 14-ser 91 Page A27 q( A42 The walkdown revealed that the outflow of water C # O Dood initiation zone can propagate through the non water tight door to the L-M e ?A, tank area and passageway (i.e., flood zone RAB 2073 DD) at elevaun 314' 0". From this flood zone, the water can eventually propagate to the RAB sump of the Gencial Access Area at elevation 317' 0" via staltway No. 2001 and the floor drain system. Because the ventilatio:t dampers are closed by SIAS, propagation of water from the General Accet: Arer. to the ECCS pump rooms will not occur.
laitiating Event: N Initiating Ewat ID: N/A laitiating Event Recovery: N/A Ims of System: S System IPE ID:
-LPSI l
System Recovery: The LPSI function will be lost due to the line break. Based on the ANO-2 IPE, there are no backup trains available to perform LPSI function within the required time period. Recovery of LPSI in the near term is not expected. The failed segment can be isolated by closing the manual suction valves for LPSI pumps "A" and "B".
Loss of Train: N Train ID:
N/A Train Recovery: N/A Consequence Comment: Because of the common discharge header for both LPSI pumps, LPSI will be lost in response to a large LOCA whether or not the failed segnent is isolated. Therefore, there will be no available backup trains (see Section 4.1.3) to prmide LPSI mitigation.
- According to the ANO-2 IPE, a loss of LPSI will cause core damage. Although loss of LPS! results in core damage, it is still desirable to isolate the failed segment in order to minimize the impact on containment performance Fc smaller LOCAs where the i
RCS pressure renmins above the shutoff head of the LPSI pumps, the failed segment must be isolated to ensure the operability of HPSI during recirculation. The longer time for switching over to recirculation makes it highly likely that the failed segment would be isolated. However, the resulting consequence would be less risk-significant than the case for a large LOCA.
Because there are no backup trains to accomplish the LPSI function in response to a large LOCA, and based on Table I and the guidance prmided in Table 3.2 of the EPRI procedure (EPRI TR 106706), a HIGH consequence category is assigned.
Consequence Category: HIGH O
Consequence na k D
FMECA - Consequence Information Report Calc"la'en h A PENG CALC Old.Rev 00 14-sep 91 Page A28 of A42 Consequence ID: LPSI-C-14B Consequence
Description:
Loss of LPSI occurs due to a line break in injection path to RCS cold leg 2P32B in the upper south piping penetration room following a LOCA.
Break Size:
Large Isolability of Break: No ISO Comments: The break is postulated to occur either during normal power operation, shutdown cooling or in response to a LOCA demand. Because the operators at highly stressed during a LOCA ennt and the relatively short time for which LPSI is needed, t is assumed that the limiting consequerme for the three configurations is associated i ith a LOCA demand. The piping from the floor penetration at elevation at 360' 0" to contaim ent penetration 2P10 is included in this segment. This consequence includes the welds in the i pplicable portions oflines 2GCB 7-6",
2CCB 3 6" and 2CCB 31.5" A failure in this segment would cause the diversion of LPSI flow from the injection line to the general access assas of the Reacter Auxiliary Building (RAB)(Calc. 89-E 0048 35, pg. 28).
The diversion of LPSI flow would be indicated by an increasing RAB sump water level which is annunciated in the control room. A low LPSI header pressure would also be indicated and annunciated in the control room. Because of the short duration of LPSI fo!!owing a large 1,0CA, the operator response time for identifying the location of the failed segment and isolating the failed segment may not be adequate even though unexpected alarms and indications are encountered in the control room. For a smaller LOCA, the duration of LPSI would be longer which makes isolation of the failed segment more likely.
Spatial Effects: Local Affected Location: Room 2084 Spatial Effects Comments: A review of Plant Design Drawings M 2044 and M 2063, and Figure 3.6-2 at elevation 354' 0" of the ANO-2 IPE (Report 94 R 2005-01, Rev 0) indicates that this line segment is located in flood zone RAB 2084-DD. The HPSI, LPSI and CS header valves, SDC suction valve 2CV-5038-1, senice water isolation valves for CCS units 2VCC 2A and 2VCC 2B, EFW distribution valves to steam generator 2E-24 A are located in this flood zone. The ANO-2 Internal Flood Screening Study (Cale. 89-E-0048 35, pg.13) assumes the failure of all components in the flood initiation zone. This assumption is too consenutive for this evaluation. For a limiting line break during a demand for LPSI following a LOCA, it was observed during the walkdown that the force exerted on the entrance non-water tight door would cause the door to open before a significant amount of water can accumulate inside the room and flood the valw motors. It was also observed that spraying orjet impingement may affect those valves which are in close proximity to the line segment where the failure is postulated. It should be noted that the HPSI, LPSI and CS injection line isolation valves and CCS senice water isolation valves are normally closed and open automatically upon receipt of an ESF actuation signal.
These valves would have already been opened at the time LPSI line break is postulated to occur. The impact of spatial effects on these valves would cause them to fail-as-is in the safe shutdown position. Flooding of all vahrs within the flood initiation zone is considered to be insignificant, and the impact of spraying orjet impingement on the HPSI, LPSI, CCS and CS vahrs is also considered to be insignificant. However, the EFW distribution valves to steam generator 2E-24 A which are closed may be impacted because if needed they would be required to open after the segment failure occurs.
a
FMECA - Consequence information Report Calculanon No. A PENG CALC 016, Rn. 00 14-ser 91 Page A29 of A42 The walkdowti revealed that the outflow of water from the flood initiation zone can propagate through the non water tight door to the access area, tank area and passageway (i.e., flood zone RAB 2073 DD) at elevation 354'-0". From this flood zone, the water can eventually propagate to the RAB sump of the General Access Area at elevation 317'-0" via stairway No. 2001 and the floor drain system. Because
- the ventilation dampers are closed by SIAS, propagation oiwater from the General Access Area to the ECCS pump rooms will not occrr, initiating Event: N Initiating Event ID: N/A laitiating Event Recovery: N/A Loss of System: S System IPE ID:
LPSI System Recovery: The LPSI function will be lost due to the line break, Based on the ANO-2 IPE, there are no l
backup trains available to perform LPSI function within the required time period. Recovery of LPSI in the near term is not expected. The failed segment can be isolated by closing 'he manual suction valves for LPSI pumps "A" and "B",
Loss of Train: N Train ID:
N/A Train Recovery: N/A Consequence Comment: Because of the common discharge header for both LPSI pumps, LPSI will be lost in response to a large LOCA whether or not the failed segment is isolated. Therefore, there will be no available backup trains (see Section 4.1.3) to provide LPSI mitigation.
According to the ANO-2 IPE, a loss of LPSI will cause core damage. Although loss of LPSI results in core damage, it is still desirable to isolate the failed segment in order to minimize the impact on containment performance. For smaller LOCAs where the RCS pressure remains above the shutoff head of the LPSI pumps, the failed segment must be isolated to ensure the operability of HPSI during recirculation. The longer time for switching over to recirculation makes it highly likely that the failed segment would be isolated. However, the resulting consequence would be less risk-significant than the case for a large LOCA.
Because there are no backup trains to accomplish the LPSI function in response to a large LOCA, and based on Table I and the guidance provided in Table 3.2 of the EPRI procedure (EPRI TR 106706), a HIGH consequence category is assigned.
Consequence Category: HIGH O
Consequence Rank O
FMECA - Consequence Information Report Calculanon & A.PENG C4LC 016, Rev 00 le-ser 97 Page A30 of A42 Consequence ID: LPSI-C 14C Consequence
Description:
Loss of LPSI occurs due to a line break in injection path to RCS cold leg 2P32C in the upper south piping penetration room following a LOCA.
Break Size:
Large Isolability of Break: No ISO Comments: The break is postulated to occur either during normal power operation, shutdown cooling or in response to a LOCA demand. Because the operators are highly stressed dunng a LOCA esent and the relatively short time for which LPSI is needed, it is assumed that the limiting consequence for the three configurations is associated with a LOCA demand. The piping from the floor penetration at elevation at 360' 0" to containment penetration 2P29 is included in this segment. This consequence includes the welds in the applicable portions oflines 2CCB-6-1.5",
2GCB 74" and 2CCB-6-6".
A failure in this segment would cause the diversion of LPSI flow from the injection line to the general access areas of the Reactor Auxiliary Building (RAB)(Calc. 89 E-0048 35, pg/28).
The diversion of LPSI flow would be indicated by an increasing RAB sump water level which is annunciated in the control room. A low LPSI header pressure would also be indicated and annunciated in the control room. Because of the short duration of LPSI following a large LOCA, the operator response time for identifying the location of the failed segment and isolating the failed segment may not be adequate even though unexpected alarms and indicatioes are encountered in the control room. For a smaller LOCA, the duration of LPSI would be longer which makes isolation of the failed segment more likely.
Spatial Effects: Local Affected Location: Room 2084 Spatial Effects Comments: A review of Plant Design Drawings M 2044 and M 2063, and Figure 3.6 2 at elevation 354'-0" of the ANO 2 IPE (Report 94 R-2005-01, Rev,0) indicates that this line segment is located in flood zone RAB 2084 DD. The HPSI, LPSI and CS header valves, SDC suction valve 2CV 5038 1, senice water isolation vah es for CCS units 2VCC 2 A and 2VCC 2B, EFW distribution valves to steam generator 2E-24A are locateiin this flood zone. The ANO-2 Internal Flood Screening Study (Calc. 89-E 0048 35, pg.13) assumes the failure of all components in the flood initiation zone. This assumption is too conservative for this evaluation. For a limiting line break during a demand for LPSI following a LOCA, it was obsentd during the walkdown that the force exerted on the entrance non-water tight door would cause the door to open before a significant amount of water can accumulate inside the room and flood the vahr motors. It was also obsened that spraying orjet impingement may affect those vahts which are in close proximity to the line segment where the failere is postulated. It should be noted that the HPSI, LPSI and CS injection line isolation valves and CCS senice water isolation valves are normally closed and open automatically upon receipt of an ESF actuation signal.
These valves would have already been opened at the time LPSI line break is postulated to occur. The impact of spatial effects on these vahts would cause them to fail-as-is in the safe shutdown position. Flooding of all valves within the flood initiation zone is considered to be insignificant, and the impact of spraying orjet impinrment on the HPSI, LPSI, CCS and CS vahts is also considered to be insignhicant. However, the EFW distribution valves to steam generator 2E-24 A which are closed may be impacted because if needed they would be required to open after the segment failure occurs.
p FMECA - Consequence Infoimation Report Calculation No. A PENG-CALC 016, Rev. 00 l4-$er 91 Pagt A3I of A42 The walkdown revealed that the outflow of water from the flood initiation zone can propagate thr:4 gh the non water tight door to the access area, tank area and passageway (i.e., flood zone RAB 2073-DD) at elevation 354' 0". From this flood zone, the water can eventually propagate to the RAB sump of the General Access Area at elevation 317' 0" via staitway No. 2001 and the floor drain system. Because the ventilation dampers are closed by SIAS, propagation of w1 ster from the General Access Area to the ECCS pump rooms will not occur.
Initiating Event: N Initiating Event ID: N/A Initiating Event Recovery: N/A Loss of System: S System IPE ID:
LPSI System Recovery: The LPSI function will be lost due to the line break. Based on the ANO-2 IPE, there are no backup trains available to perform LPSI function within the required time period. Recovery of LPSI in the near term is not expected. The failed segment can be isolated by closing the i
manual suction valves for LPc! aumps "A" and "B".
Loss of Train: N Train ID:
N/A Train Recovery: N/A Consequence Comment: Because of the common discharge header for both LPSI pumps, LPSI will be lost in response to a large LOCA whether or not the failed segment is isolated. Therefore, there will be no available backup trains (see Section 4.1.3) to proside LPSI mitigation.
According to the ANO-2 IPE, a loss of LPSI will cause core damage. Although loss of s
LPSI results in core damage, it is still desirable to isolate the failed segment in order to minimize the impact on containment performance. For smaller LOCAs where the RCS pressure remains above the shutoff head of the LPSI pumps, the failed segment must be isolated to ensure the operability of HPSI during recirculation. The longer time for switching over to recirculation makes it highly likely that the failed segment would be isolated. However, the resulting consequence would be less risk significant than the case for a large LOCA.
Because there are no backup trains to accomplish the LPSI function in response to a large LOCA, and based on Table I and the guidance provided in Table 3.2 of the EPRI procedure (EPRI TR l%706), a HIGH consequence category is assigned.
Consequence Category: HIGH O
Consequence Rank O
%J
FMECA - Consequence Information Report Cahlatwn No A PEECEC 016, Rev. 00 14-ser 91 Page A32 of A42 Consequence ID: LPSI-C 14D Consequence
Description:
Loss of LPSI occurs due to a line break in itsection path to RCS cold leg 2P32D in the upper south piping penetration room following a LOCA.
Break Stre:
Large Isolability of Break: No ISO Comments: The break is postulated to occur either during normal power operation, shutdown cooling or in response to a LOCA demand. Because the operators are highly stressed during a LOCA event and the relatively short time for w hich LPSI is needed, it is assumed that the limiting consequence for the three configurations is associated with a LOCA demand. The piping from the floor penetration at elevation at 360' 0" to containment penetration 2P24 is included in this segment. This consequence includes the welds in the applicable portions oflines 2GCB 7 6*
and 2CCD 5-6" A failure in this segment would cause the diversion of LPS! flow from the injection line to the general access areas of the Reactor Auxiliary Building (RAB)(Calc. 89-E-0048 35, pg. 28).
The diversion of LPSI flow would be indicated by an increasing RAD sump water level which is annunciated in the control room. A low LPSI header pressure would also be indicated and annunciated in the control room. Because of the shon duration of LPSI following a large LOCA, the operator response time for identifying the location of the failed segment and isolating the failed segment may not be adequate even though unexpected alarms and indications are encountered in the control room. For a smaller LOCA, the duration of LPSI would be longer which makes isolation of the failed segment more likely, Spatial Effects: Local Affected Location: Room 2084 Spatial Effects Comments: A review of Plant Design Drawings M 2044 and M 2063, and Figure 3.6-2 at elevation 354'-0" of the ANO-2 IPE (Report 94 R 2005-01, Rev,0) indicates that this line segment ir, located in flood zone RAB 2084 DD The HPSI, LPSI and CS header valves, SDC suction valve 2CV 5038 1, senice water isolation vahts for CCS units 2VCC 2 A and 2VCC-2B, EFW distribution valves to steam generator 2E-24A are located in this flood zone. The ANO-2 Internal Flood Screening Study (Calc. 89 E 0048-35, pg.13) assumes the failure of all components in the flood initiation zone. This assumption is too conservative for this evaluation. For a limiting line break during a demand for LPSI following a LOCA, it was obsened during the walkdown that the force exerted on the entrance non-water tight door would cause the door to open before a significant amount of water can accumulate inside the room and flood the valve motors. It was also observed that spraying orjet impingement may affect those valves which are in close proximity to the line segment where the failure is postulated. It should be noted that the HPSI, LPSI and CS injection line isolation valves and CCS wnice water isolation valves are normally closed and open automatically upon receipt of an ESF actuation signal.
These valves would have already been opened at the time LPSI line break is postulated to occur. The impact of spatial effects on these valves would cause them to fail-as-is in the safe shutdown position. Flooding of all vahts within the flood initiation zone is considered to be insignificant, and the impact of spraying orjet impingement on tl e HPSI, LPSI, CCS and CS vah es is also considered to be insignificant. Hccever, the EFW distribution vahts to steam generator 2E-24 A which are closed may be impacted because if needed they would be required to open af er the segment failure occurs
FMECA - Consequence Information Report Calculation Na A PENG C4LC-0M. Rn 00 Q
l4.ser.91 Page A33 of A42 The walkdown revealed that the outflow of water from the flood initiation zone can propagate through the non water tight door to the access area, tank area and passageway (i.e., flood zone RAB 2073 DD) at elevation 35t' 0". From this flood zone, the water can eventually propagate to the RAB sump of the General Access Area at elevation 317' 0" via stairway No. 2001 and the floor drain system. Because the ventilation dampers are closed by SIAS, propagation of water from the General Access Area to the ECCS pump rooms will not occur.
Initiating Event: N Initiating Event ID: N/A InitiatingI': cat Recovery: N/A Loss'of System: S System IPE ID:
LPSI Systein Recovery: The LPSI function will be lost due to the line break. Based on the ANO-2 IPE, there are no backup trains available to perform LPSI function within the required time period. Recovery of LPSI in the near term is not expected. The failed segment can be isolated by closing the l
manual suction valves for LPSI pumps "A" and "B",
Loss of Train: N Train ID:
N/A Train Recovery: N/A Consequence Comm*nt: Because of the conunon discharge header for both LPSI pumps, LPSI will be lost in response to a large LOCA whether or not the failed segment is isolated. Therefore, there will be no available backup trains (see Section 4.1.3) to proside LPSI mitigation.
According to the ANO-2 IPE, a loss of LPSI will cause core damage. Although loss of fg f
LPSI results in core damage, it is still desirable to isolate the failed segment in order to minimize the impact on containment performance. For smaller LOCAs where the RCS pressure remains above the shutoff head of the LPSI pumps, the failed segment must be isolated to ensure the operability of HPSI during recirculation. The longer time for switching over to recirculation makes it highly likely that the failed segnwnt would be isolated. However, the resulting consequence would be less risk-significant than the case for a large LOCA.
Because there are no backup trains to accomplish the LPSI function in response to a large LOCA, and based on Table I and the guidance provided in Table 3.2 of the EPRI procedure (EPRI TR 106706), a HIGH consequence category is assigned.
Consequence Category: HIGH O
Consequence Rank O
FMECA - Consequence Information Report Cahlad nNo A PENG-C4LC-016.Rev 00 14-sep-91 Page A34 of A42 Consequence ID: LPSI C 15A Consequence
Description:
Loss of LPSI occurs due to a line break in injection line to RCS cold leg 2P32A inside the containment following a LOCA.
Break Size:
Large Isolability of Break: No ISO Comments: The break is postulated to occur either during normal power operation, shutdown cooling or in response to a LOCA demand. Because the operators are highly stressed during a LOCA esent and the relatively short time for which LPSI is needed, it is assumed that the limiting consequence for the three configurations is associated with a LOCA demand. The piping from downstream of containment penetration 2P15 to upstream of check vaht 2SI 14 A is included in this segment. This consequence includes the utids in the applicable portion ofline 2CCB-4-6".
A failure in this segment would cause the diversion of LPSI flow from the injection line to the containment sump, thus bypassing the reactor core. No unexpected alarms and indications would be encountered because of the location of the failed segment. It is therefore assumed that the failed segment would not be identified and isolated.
Spatial Effects: Containment Affected Location: Containment Building Spatial Effects Comments: The lines in this segment are included as part of the LPSI injection paths to the RCS cold legs. A dynamic analysis of the above line inside the containment was performed. The analysis concluded that there would be no failure of safety related ce mponents caused by the dynamic effects of the line break (S AR Section 3.6.4.2.8.2). In addition, all safety irgjection components and associated electrical equipment inside the containment have been designed to withstand LOCA emironmental conditions (SAR Section 6.3.2.12.1). It is therefore assumed that spatial effects caused by the line break are negligible, initiating Event: N Initiating Event ID: N/A Initiating Event Recovery: N/A Loss of S stem: S System IPE ID:
LPSI 3
System Recovery: LPSI function will be lost due to the segment failure and the occurrence of a large LOCA in any of the RCS cold legs, except 2P32 A (ANO-2 IPE). Based on the ANO-2 IPE, there are no backup trains available to perform LPSI function within the required time period. Recostry of LPSI in the near term is not expected.
Loss of Train: N Train ID:
N/A Train Recovery: N/A Consequence Comment: LPSI function is lost due to flow diversion caused by the line break following a large LOCA. Because the segment failure is not detectabic and there are no equivalent backup trains to accomplish the LPSI function in response to a LOCA, and based on Table I and the guidance provided in Table 3.2 of the EPRI procedure (EPRI TR-106706), a HIGH consequence category is assigned.
Consequence Category: HIGH O
Consequence nank O
4 l
FMECA - Consequence Information Repott Cahlahon No A FMG-C41f-016. Rev. 00
. p)
WSW Page A33 of AU q
Consequence 13: LPSI C 15B Consequence
Description:
Loss of LPSI occurs due to a line break in injection line to RCS cold leg 2P32B inside the containment following a LOCA.
Break Size:
Large Isolability of Batak: No ISO Comments: The break is pos+'ilated to occur either during normal power operation, shutdown cooling or in respons to a LOCA demand. Because the operators are highly stressed during a LOCA event and the relatively short time for which LPSI is needed, it is assumed that the limiting consequence for the three configurations is associated with a LOCA demand. The piping from downstream of containment penetration 2P10 to upstream of check vahr 2SI 14B is included in this segment. This consequence includes the welds in the applicable portion ofline 2CCB-3-6",
A failure in this segment would cause the diversion of LPSI flow from the injection line to the containment sump, thus bypassing 11-. reactor core. No unexpected alarms and indications would be encountered because of the location of the failed segment. It is therefore assumed that the failed segment would not be identified and isolated.
Spatial Effects: Containment Affected Location: Containment Building Spatial Effects Comments: The lines in this segment are included as part of the LPSI injection paths to the RCS cold legs. A dynamic analysis of the above line inside the containment was performed. The analysis concluded that there would be no failure of safety related components caused by the dynamic effects of the line break (SAR Section p/
3.6.4.2.8.2). In addition, all safety injection components and associated electrical y
equipment inside the containment have been designed to withstand LOCA emironmental conditions (SAR Section 6.3.2.12.1). It is therefore assumed that spatial effects caused by the line break are negligible.
Initiating Event: N Initiating Event ID: N/A Initiating Event Recovery: N/A Loss of System: S System IPE ID:
LPSI System Recovery: LPSI function will be lost due to the segment failure and the occurrence of a large LOCA in any of the RCS cold legs, except 2P32B (ANO 2 IPE). Based on the ANO-2 IPE, there are no backup trains available to perform LPSI function within the required time period. Recovery of LPSIin the near term is not expected.
Imss of Train: N Train ID:
N/A Train Recovery: N/A Conseqrence Comment: LPSI function is lost due to flow diversion caused by the line break following a large LOCA. Because the segment failure is not detectable and there are no equivalent backup trains to accomplish the LPSI function in response to a LOCA, and based on Table I and the guidance provided in Table 3.2 of the EPRI procedure (EPRI TR.
106706), a HIGH consequence category is assigned.
Consequence Category: HIGH O
Consequence aank O
Ch
FMECA - Consequence Information Report calcularmn Na A PENG-Calf 016. Rev. 00 l 4-Sep-91 Page A36 of A42 Consequence ID: LPSI-C-15C Consequence
Description:
Loss of LPS! occurs due to a line break in injection line to RCS cold leg 2P32C inside the containment following a LOCA.
Break Size:
Large Isolability of Break: No ISO Comments: The break is postulated to occur either during normal power operation, shutdow11 cooling or in response to a LOCA demand. Because the operators are highly stressed during a LOCA event and the relatively short time for which LPSI is needed, it is assumed that the limiting consequence for the three configurations is associated with a LOCA demand. The piping from dow tstream of containment penetration 2P29 to upstream of check vahr 2SI 14C is included in this segment. This consequence includes thwelds in the applicable portion ofline 2CCB-6-6",
A failure in this segment would cause the diversion of LPSI flow from the injection line to the containment sump, thus bypassing the reactor core. No unexpected alarms and indications would be encountered because of the location of the failed segment. It is therefore assumed that the failed segment would not be identified and isolated.
Spatial Effects: Containment Affected Location: Containment Building Spatial Effects Comments: The line1 in this segment are included as part of the LPSI injection paths to the RCS cold legs. A dynamic analysis of the above line inside the containment was performed. The analysis concluded that there would be no failure of safety related components caused by the dplamic efIects of the line break (SAR Section 3.6.4.2.8.2). In addition, all safety injection components and associated electrical equipment inside the containment have been designed to withstand LOCA emironmental conditions (S AR Section 6.3.2.12.1). It is therefore assumed that spatial effects caused by the line break are negligible.
Initiating Event: N Initiating Event ID: N/A Initiating Event Recovery: N/A Loss of System: S System IPE ID:
LPSI System Recovery: LPSI function will be lost due to the segment failure and the occurrence of a large LOCA in any of the RCS cold legs, except 2P32C (ANO-2 IPE). Based on the ANO-2 IPE, there are no backup trains available to perform LPSI function within the required time period. Recovery of LPSI in the near term is not expected.
Loss of Train: N Train ID:
N/A Train Recovery: N/A Consequence Comment: LPSI function is lost due to flow diversion caused by the line break following a large LOCA. Because the segment failure is not detectable and there are no equivalent backup trains to accomplish the LPSI function in response to a LOCA, and based on Table I and the guidance provided in Table 3.2 of the EPRI procedure (EPRI TR-j 106706), a HIGH consequence category is assigned.
Consequence Category: HIGH O
Consequence Rank O
O
v---,.---,
a p)
FMECA - Consequence Information Report Cakulana L. A.PENG CALC 016, Rev 00 14-5*p-91 Page AH of Au Consequence ID: LPSI-C 15D -
Consequence
Description:
Loss of LPSI occurs due to a line break in injection line to RCS cold leg 2P32D inside the containment following a LOCA.
Break Size:
Large Isolability of Break: No ISO Comments: The break is postulated to occur either dunng normal power operation, shutdowTi cooling or in response to a LOCA demand. Because the operators are highly stressed during a LOCA event and the relatively short time for which LPSI is needed, it is assumed that the limiting consequence for tne three configurations is associated with a LOCA demand. The piping from downstream of containment penetration 2P24 to upstream of check vah e 2SI 14D is included in this segment. This consequence includes the welds in the applicable portion ofline 2CCB 54".
A failure in this segment would cause the diversion of LPSI flow from the injection line to the containment sump, thus bypassing the reactor core, No unexpected alarms and indications would be encountered because of the location of the failed segment. It is therefore assumed that the failed segment would not be identified and isolated.
l Spatial Effects: Containment Affected IAcation: Containment Building l
Spatial Effects Comments: The lines in this segment are included as pan of the LPSI injection paths to the RCS l
cold legs. A dy:uunic r.nalysis of the above line inside the containment was performed. The analysis concluded that there would be no failure of safety related components caused by the dynamic effects of the line break (SAR Section G
3.6.4.2.8.2), in addition, all safety injection components and associated electrical V
equipment inside the containment have been designed to withstand LOCA emironmental conditions (SAR Section 6.3.2.12.1). It is therefore assumed that spatial effects caused by the line bre 1 armegligible.
Initiating Event: N Initiating Ev.at ID: R A Initiating Event Recovery: N/A Loss of System: S System IPE ID:
LPSI System Recovery: LPSI function will be lost due to the segment failure and the occurrence of a large LOCA in any of the RCS cold legs, except 2P32D (ANO-2 IPE). Based on the ANO-2 IPE, there are no backup trains available to perform LPSI function within the required time period. Recmtry of LPSIin the near term is not expected.
Loss of Train: N Train ID:
N/A Train Recovery: N/A Consequence Comment: LPSI function is lost due to flow diversion caused by the line break following a large LOCA. Because the segment failure is not detectable and there are no equivalent backup trains to accomplish the LPSI function in response to a LOCA, and based on Table I and the guidance provided in Table 3.2 of the EPRI procedure (EPRI TR-i 106706), a HIGH consequence category is assigned.
Consequence Category: HIGH O
Consequence aank O
OO
l l
FMECA - Consequence Information Report Cohtahon No. A PENG-CEC-0M, Rn. 00 14-sep-91 Page A38 of A42 Consequence ID: LPSI C 16 Consequence
Description:
Loss of reactor coolant occurs due to a SDC line break in the Tendon Gallery Access area Break Size:
Large Isolability of Break: Yes ISO Comments: The break is postulated to occur during any of the various states of plant shutdown operation.
The SDC piping from upstream of either manual valve 2SI 5A or 2SI 5B to the ceiling penetration at elevation 335'-0" is included in this segment. Of the various plant shutdown states considered (see Sections 4.4.2 and 4.4,7), mid-loop operation is the most risk-significant because of the relatively short time it would take to uncover the core. Therefore, only the consequence resulting from a pipe break during mid-loop operation is described herein. This consequence includes all the welds in line 2GCB-8-6" and the welds in the applicable por* ions oflines 2GCB-814",2GCB-812" and 2GCB 17-12".
A failure in this segment would cause the diversion of reactor coolant to the Tendon Gallery Access Area. This would be indicated by an unexpected increase in the Reactor Auxiliary Building (RAB) sump water level which is annunciated in the control room. Other unexpected alarms and indications include decreas A RCS inventory, low LPSI pump discharge flow and inappropriately low LT31 discharge pra are. Because the operators are carefully monitoring RCS level during mid-loop operation, it is assumed that the failed segment would be identified and isolated.
Spatial Effects: Local Affected Location: Room 2011 Spatial Effects Comments: A review of Plant Design Drawing M-2046 and Figure 3.6-2 at elevation 317'-0" of the ANO-2 IPE (Report 94 R 2005-04, Rev 0) indicates that this line segment is located in flood zone RAB-20ll LL. Service water loop I to ESF equipment isolation valve 2CV 1400-1 is located in this flood zone. The ANO-2 Internal Flood Screening Study (Calc 89 E-0048-35, pg.13) assumes the failure of all components in the flood initiation zone, it was observed during the walkdown that valve 2CV.
1400-1 would not be impacted by spraying orjet impingement because the valve is located sufficiently far away from the line segment where the failure is postulated.
The elevated position of the valve also makes it very unlikely to be impacted by flooding. It should be noted that valve 2CV 1400-1 is normally open and would therefore fail-as is in the open position. Hence for this SDC line break scenario, the impact of spatial effects on valve 2CV-1400-1 is negligible.
The walkdown revealed that the outflow of water from the flood initiation zone can
~
propagate to the General Access Area. These two flood zones are adjacent to each other, and there are no physical barriers between them. Because the combined area of these flood zones is very large, they can accommodate a significant amount of water before spillover to the ECCS pump rooms becomes a threat. With the General Access Area capable of accommodating a large volume of water, the control room operators will have sufIicient time to identify and isolate the break. Hence for this SDC line break scenario, the impact of flooding or other spatial effects are of no significance, laitiating Event: I Initiating Event ID: A Initiating Event Recovery: One of the redundant makeup sources is required for mitigation and to accommodate decay heat boil-off. Manual initiation of makeup would be required.
O FMECA - Consequence Information Report Cablanon No. A PENG CALC Old, Rn. 00 14-5p91 Page A39 of A42 less of System: S System IPE ID:
SDC System Recovery: Reactor coolant and shutdown cooling are lost due to the line break. Flow through the break can be terminated by securing the LPSI pump (s) followed by localized isolation of the affected lines, it is assumed that the operators ability to isolate the break during mid-loop operation is equivalent to having one backup train.
less of Train: N Train ID:
N/A Train Recovery: N/A Consequence Comment: The segment failure causes a loss of reactor coolant and the unavailability of SDC during mid-loop operation. During this mode of shutdown operation, the operators are carefully monitoring RCS inventory due to the importance of this configuration, but isolation may be inclevant because RCS level is already near the invert of the reactor nozzles. An increase in the level of operator awareness is M=M The Shutdown Operations Protection Plan utilized at ANO-2 requires that redundant sources and associated paths are in place to provide makeup during mid-loop operation. Thus, there would be at least two equivalent backup trains available for mitigating this type ofLOCA.
Because there are two backup trains available, the resulting consequence is MEDIUM.
This is based on the Table 3,4 of the EPRI procedure (EPRI TR 106706).
Consequence Category: MEDIUM O
Consequence Rank O
O O
l FMECA - Consequence Information Report Calculanon No. A PENG-CALC-016. Rev. 00 14.ser-91 Page A40 of A42 Consequence ID: LPSI-C 17 Consequence
Description:
Loss of shutdown cooling occurs due to a break in the SDC heat exchanger return line in the lower south piping penetration area.
Break Size:
Large Isolability of Break: Yes ISO Comments: The break is postulated to occur daring any of the various states of plant shutdown operation, and in the piping from the floor penetration at elevation 335'-0" to upstream of either the SDC heat exchanger discharge throttle valve 2CV 5093 or the SDC heat exchanger discharge throttle bypass valve 2515093 3. The SDC purification line from SDC heat exchanger return header to upstream of valve 251-34 is also included in this segment. Of the various plant shutdown states considered (see Sections 4.4.2 and 4.4.7), mid-loop operation is the most risk significant because of the relatively short time it would take to uncover the core. 'Iherefore, only the consequence resulting from a pipe break during mid-loop operation is described herein. This consequence includes the welds in the applicable portions oflines 2GCB-8 8" and 2GCB-814".
A failure in this segment would cause the diversion of reactor coolant flow from the LPSI header to the general access area of the Reactor Auxiliary Building (RAB)(Calc. 89-E-0048 35, pg. 28) during mid loop operation. This would be indicated by an unexpected increase in the Reactor Auxiliary Building (RAB) sump water level which is annunciated in the control room.
Other unexpected alarms and indications include decreasing RCS inventory, low LPSI pump discharge flow and inappropriately low LPSI discharge pressure. Because the operators are carefully monitoring RCS level during mid loop operation, it likely that the failed segment would be identified and isolated.
Spatial Effects: Local Affected Location: Room 2055 Spatial Effects Comments: A review of Plant Design Drawing M 2045 and Figure 3.6-2 at elevation 335'-0" of the ANO-2 IPE (Report 94-R-2005 04, Rev. 0) indicates that this line segment is located in flood zone RAB 2055-JJ in the Reactor Auxiliary Building (RAB). The HPSI orifice bypass valves (2CV 5103 1 and 2CV 5104 2), LPSI header isolation valve 2CV 5091, SDC heat exchanger throttle valve 2CV-5093 and EFW pump flush valve 2CV 0714 1 are located in this zone. The ANO-2 Internal Flood Screening Study (Calc. 89-E 0048-35, pg.13) assumes the failure of all components in the flood initiation zone, For a limiting line break during SDC, it was obsentti during the walkdown that the force exerted on the entrance non-water tight door would cause the door to open before a significant amount of water can accumulate inside the room and flood the valve operators. It was also obsened that spraying or jet impingement may affect certain valves (2CV-5103-1,2CV 5093 and 2CV 5091) because of the close proximity of these valves to the line where the failure is postulated. It should be noted that the impact of spatial effects on the HPSI bypass valves is of no significance bxause HPSI is not needed to support SDC.
The walkdown revealed that the outflow of water from the flood initiation zone can propagate through the non-water tight door to the adjacent tank rooms, pump rooms and corridor (i.e., flood zone RAB 2040-JJ) at elevatica 335' 0". From this flood zone the water can eventually propagate to the RAB sump of the General Access Area at elevation 317'-0" via stairway No. 2001 and the floor drain system. Water can also propagate from the flood initiation zone to the RAB sump via the floor drain system. Because the General Access area is large, there are direct indications of RAS sump level in the Control Room and the pathways to the ECCS pump room l
___m_-__
q FMECA - Consequence Information Report Calculation No. A PENG CALC 016. Rev. 00 t 4-ser-91 Page A41 of A42 s
are elevated, it is assumed that corrective actions will be taken (i.e., securing the LPSI pumps) in a timely manner. Therefore for this SDC line break scenario, spillover from the General Access Area to the ECCS pump room is very unlikely but SDC function would be lost.
Initiating Event: I laitiating Event ID: A Initiating Event Recovery: One of the redundant makeup sources is required for mitigation and to accommodate decay heat boil off. Manual initiation of makeup would be required.
Loss of System: S System IPE ID:
SDC System Recovery: The shutdown cooling function is lost due tt :he line break. Flow through the break can be terminated by securing the LPSI pump (s) followed by localized isolation of the affected line.
It is assumed that the operators ability to isolate the break during mid-loop operation is equivalent to having one backup train.
Loss of Train: N Train ID:
N/A Train Recovery: N/A Consequence Comment: The segment failure causes a loss of reactor coolant and the unavailability of SDC during mid-loop operation. During this mode of shutdoun operation, the operators are carefully monitoring RCS inventory due to the importance of this configuration, but isolation may be irrelevant because RCS level is already near the invert of the reactor nozzles. An increase in the level of operator awareness is==" mad The Shutdown Operations Protection Plan utilized at ANO-2 requires that redundant sources and associated paths are in place to provide makeup during mid-loop operation. Thus, there would be at least two equivalent backup trains available for mitigating this type of LOCA.
Because there are two backup trains available, the resulting consequence is MEDIUM.
This is based on the Table 3.4 of ti c EPRI procedure (EPRI TR-1067%).
Consequence Category: MEDIUM O
Consequence aank O
a
l FMECA - Consequence infuirnetion Report rahaavite.MmGCALC01G h. 00 14&p97 Age At? of A42 Table 1 ASSIGNED CONSEQUENCE CATEGORIES FOR ANO-2INTI1A'HNGEVENIS In6unues Ewat inau ung Ennt tatueung twat %scriptha IE CDF ccDP ceaugmence rYp (ty)
(CDritt Freq) i Routme Startg RA MA WA N'A Shutine.
Sam @y Refuehag II Anucipated Reactor inp. (T6) 2 03
$ 95E-06 2 93E-06 LEDIUM Operauoral Occwrence Loss of Power Cormruien Synern.(T2) 0 25 8 99E47 3.59E-06 MEDIUM Twbme Tnp. (TI) 0 76 1.27E46 2 98E46 LEDTUM 111 Infrequent Lass of 0f!sne Power.(T3)
$ 54E42 172E46 195E-05 MEDIUM Low of SWPwnp 2P4A.(T8) 738E42 2.14E47 2 90E 06 MEDIUM 88 Lass of 3W Pwnp 2P48.(79) 7.38E 02 2 04E-07 277E 06 MEDIUM '
8 IV Lenna 1Feuha Excessrve Feedwater. (T4) 9 40E44 187E49 199E m MEDIUM or At adents Swan Lee /Feedwaar Lee Break.(TS) 1.10E 03 1.13E49 103E 06 MEDIUM Tctalloss of 3W.frf)
$ 45E43 214E 4 3 92EM IDOH Loss of DC Bus 2 Dol -(TIO) 3 94E44 9 80E-06 2 49E-02 100H
- Loss of DC Bus 2002. (Tl1) 3 NEet 1 IIE46 2 83E 03 100H
Loss of AC Bus 2A3 -(T12) 3 94EM 3.23E 06 8 20EC3 ICOH
- Loss of AC Bus 2A4.(TI))
3 WE44 5 7tE 04 I 47E-N HIGH" Lees of 4s0V tmd Cenis: 2B5 -(T14) 104E43 I 90E47 I 83E 04
}DOH" Loss of 480V 14ed Centar 286 (T15) 104E-03 120E47 815E44 ICOH" Small LOCA.(3)
$ 00E-03 171E-06 3 43E4:
lOOH MahwnLOCA.(M) 100E-03 174E46 174E43 FUGH Large LOCA.( A) 1 005-04 139E 06 139E42 IDG1
(
Stearn Genemor Tube Ruptwo.(R) 9 77E 03 9 54E48 9 76E46 MEDIUM NOTE:
1.
This intunung ewne results la e reacter trtp and less erone trate of servtce water to ESF leads.
2.
This initiatlag event results la e reacter trip and perualless of effatte power.
s.
This initwo s ennt enuits in a re.cter tre, and een of p.wer te one trui of sneu uas sys===
s "IllGH" CCDP>1P 4
d
" MEDIUM" 10 < CCDP < 10
" LOW" CCDP < 104 The above table was devrioped for the ANO 2 spectSc inhisting events. It is based on the inferination provided in Tables 3.1 and 3.4 of the EPRI RISI procedure (EPRI TR-196706). The initiating event descriptions (for event categories 11.111. A IV) and associated initiating event frequencies and Core Damage Frequencies (CDFs)were estracted frans Tables 3J-4 and 3.14-7A of the ANO-2 IPE (Report 9d-R.
2005 01. Rev. 0),
O
Calculation No. A PENG-CALC 016, Rev. 00 Page 81 of 854 O
APPENDIX B
'FMECA - DEGRADA TION MECHANISMS" (Attachment Pages B1 854)
I O
ABB Combustion Engineering Nuclear Operations
'##7 '
FMECA - Degradation Mechanisms C"'"'"" " ^' Ad^"C-8'6 #" 88 Page B2 of B34 Weld System ID Segment Line Number Line Description Number Weld Imcation T
C P
I M
E F
0 LPSI LPSI-001 2CCA-25-14" Shutdown cooling 25-011 Upstream of elbow #3 No No No No No No No No suction line from RCS hot leg (Loop B) to shutdown cooling suction valve 2CV-5086-2 LPSI LPSI-001 2CCA-25-14" Shutdown cooling 25-012 Downstream orelbow #3 No No No No No No No No suction line from RCS hot leg (Imop B) to shutdown cooling suction vaht 2CV-5086-2 LPSI LPSI-001 2CCA-25-14" Shutdown cooling 25-013 Upstream of motor-No No No No No No No No suction line from RCS operated ulve 2CV-5084-1 hot leg (Imop B) to shutdown cooling suction ulve 2CV-5086-2 LPSI LPSI-001 2CCB-3-l.5" Vent stack connection 61-018A Item 8 reducinginsert No No No No No No No No LPSI LPSI-001 2CCB-3-6" Safetyinjection piping 61-012 Downstream of motor-No No No No No No No No from control valve operated vaht 2CV-5037-2CV-5037-1 to I (Item #16) containment penetration 2P-10 Dearadation Mecharusms T-Thmnal Fatigue P - Pnmary Water Stress Cernmion Cracking (PWSCC)
M-ML C e"fbdluenced Corremen (%HC)
F-flow Ameiermed Cernmen C-Cerrosson Cracking I - Ictergranular Stress Cerrosion Cracimg (IGSCC)
E-Erosen-Cavitation 0-Other O
O O
t O
O O
l C"" lad n A'a MMC 416 Rm 00 FMECA - Degradation Mechanisms
. age B3 of B54 W eld Systein ID Segment -
Line Number Line Description Number Weld tacation T
C P
I M
E F
0 LPSI LPSI-001 2CCB-34" Safety injection piping 61-013 Upstream ofelbow f5 No No No No No No No No from control nhe j
2CV-5037-1 to
. containment penetration 2P-10 l
1 LPSI LPSI-001 2CCB-34" Safety injection piping 61-014 Don 1 stream orelbow #5 No No No No No No No No from control nht -
2CV-5037-1 to containment -
I penetration 2P-10 LPSI LPSI401 2CCB-34" Safety injection piping 61-015 Upstream ofelbow #6 No No No No No No No No from control nhe 2CV-5037-1 to
~
containnoit penetration 2P-10 LPSI LPSI-001 2CCB-34" Safety injection piping 61-016 Downstream ofelbow #6 No No No No No No No No from control nhe 2CV-5037-1 to containment penetration 2P-10 LPSI LPSI-001 2CCB-34" Safety injection piping 61-017 Upstream ofelbow #7 No No No No No No No No from controlvahr 2CV-5037-I to containment penetration 2P-10 Desredshon W T-Thermal Fatigue P - Pnmary Water Stres Cemnaen Crecbng (PWSCC)
M - Micreemologically InAmenced Corresson (MIC)
F-mw Accelerused Cerroman C-Carmeen oncking I-Irmergranular stre= Corronen Cracbrig OGSCC)
E-Erosion-Canishen 0-Other I
_...-...,.,,_,m...
.. ~.
C"##"lati n No. A-PEVG-CtLC-0/6. Rev. 00 FMECA - Degradation Mechanisms Page B4 of B54 Weld System ID Segment Line Number Line Description Number Weld Location T
C P
I M
E F
0 l
LPSI LPSI-001 2CCB-34" Safety injection piping 61-018 Downstream ofelbow #7 No No No No No No No N-from control vahr 2CV-5037-1 to containment penetration 2P-10 LPSI LPSI-001 2CCB-34" Safetyinjection piping 61-019 Upstream of Flued Head No No No No No No No No from control vahr
- 17 2CV-5037-1 to containment penetration 2P-10 LPSI LPSI-001 2CCB-34" Safety injection piping 66-012 Upstream of LPSI check No No No No No No No No from control vahr vahr 2SI-14B 2CV-5037-1 to containment penetration 2P-10 LPSI LPSI-001 2CCB-34" Safetyinjection piping 66-013 Downstream of Flued No No No No No No No No from control vahr Head (item #17 on dwg 2CV-5037-1 to 2CCB-3-1 containment penetration 2P-10 LPSI LPSI-001 2CCB-4-1.5" Vent stack connection 61430B Item 19 reducing insert No No No No No No No No LPSI LDSI-001 2CCB-44" Safetyinjection from 61-026 Downstream ormotor-No No No No No, No No No 2CV-5017-1 to operated vahr 2CV 5017-containment I (Item #22) penetration 2P-15 Desradation Mecharusms T-Thmnal Fatigue P - Pnmary Water Stress Cemsson creclung (PWSCC)
M - Micrebsologically Innuenced Comisens (MIC)
F-Flow AccelerusedCermsson C-Cemsson Cracking I-IniergranularStressCorra scaCraclung(IGSCC)
E - Eresson-Carstatma 0-Other O
O O
O O
O
'N' Catalan n Na AMMW16. Rn. M FMECA - Degradation Mechanisms Page B5 of B54 l W eld System ID Segment Line Number Line Description Number Weld IAcation T
C P
I M
E' F
0 i
LPSI LPSI401 2CCB44" Safety injection fmm 61-027 Upstream ofelbow #16 No No No No No No No j
2CV-5017-1 to containment penetration 2P-15 LPSI LPSI-001 2CCB44" Safety injection from 61-028 Dowinstream ofelbow fl6 No No No No No No No No 2CV-5017-1 to containment penetration 2P-15 LPSI LPSI-001 2CCB44" Safety injection from 61-029 Upstream ofelbow #17 No No No No No No No 2CV-5017-1 to containment penetration 2P-15 LPSI LPSI-001 2CCB44" Safety injection from 61-030 Dom 1 stream ofelbow fl7 No No No No No No No 2CV-5017-1 to and upstream ofelbow #18 containment paa6vii 2P-15 LPSI LPSI-001 2CCB44" Safety injection from 61-030A Dos 1: stream ofelbow #18 No No No No No No No No 2CV-5017-1 to containment penetration 2P-15 LPSI LPSI401 2CCB-44" Safetyinjection from 61-031 Upstream of Flued licad No No No No No No No No 2CV-5017-1 to
((tem #25) contamment penetration 2P-15 LPSI LPSI-001 2CCB44" Safety injection from 66414 Downstream ofFlued No No No No No No No No 2CV-5017-1 to Head (Item #25 on dwg containment 2CCB4-1) penetration 2P-15 Deerndemanu.a T-Thennel Fatigue P - Pnniary Weser Stress Cerreman Crachng (PwSCC)
M-L2 '. ", homenced Corressen(MIC)
F-How Acceiwesed Carrossen C-Correseen Cradting I-A Stress Commen Oechng00 SCC)
E - Esossen-Cavitesen 0 -Oiher
C"I'"lanon A~a A-l'FCi-CsiK-016. Rm W FMECA - Degradation Mr :hanisms Page B6 of B54 Weld System ID Segment Line Number Line Description Number Weld Iacation T
C F
I M
E F
O LPSI LPSI-001 2CCB-4-6" Safety injection from 66-015 Upstream of LPSI check No No No No No No No No !
2CV-5017-1 to vahr 2SI-I4A containment penetration 2P-15 LPSI LPSI-001 2CCB-5-6" Safety injection piping 61-003 Du-mticam of motor-No No No No No No No No from control vahr operated vahr 2CV-5077-2CV-5077-2 to 2 (Item #15) containment penetration 2P-24 LPSI LPSI-001 2CCD-5-6" Safety injection piping 61-004 Upstream ofelbow #4 No No No No No No No No from control vahr 2CV-5077-2 to containment penetration 2P-24 LPSI LPSI-001 2CCB-5-6" Safety injection piping 61-005 Downstream ofelbow f4 No No No No No No No No from control vahr 2CV-5077-2 to containment penetration 2P ~4
)
LPSI LPSI-001 2CCB-5-6" Safety injection piping 61-006 Ups'. ream of elbow #5 No No No No No No No No from control vahr 2CV-5077-2 to containment penetration 2P-24 LPSI LPSI-001 2CCB-5-6" Safety injection piping 61-007 Downstream ofelbow #5 No No No No No No No No from control valve and upstream ofcibow #6 2CV-5077-2 to containment penetration 2P-24 Deeradation Mec!wasms T-Thennal ratigue F - Pnmary Water Stress Cerronce Cracbng (P%%T)
M - Micrabsoirwcally Innuenced Cerrosenn (MIC)
F-flow AccelerededCerrassen C-Cerrenon Cracbng I-Irmergranular Stress Cc rosion CracUng HGSCC)
E-Erossen-Cavitation 0-Other O
O O
~
~
e C"'"" "Na AGMC-014 Rn. 017 FMECA - Degradation Mechanisms Page BT of B54 Weld Syst-m ID Segment Line Number Line Description Number Weld imaties T
C P
I M
E F
0
~
LPSI LPSI-001 2CCB-54" Safety injecten piping 61-008 Downstream ofelbow #6 No No No No No No No No from control vahr 2CV-5077-2 to containment penetration 2P-24 LPSI LPSI-001 2('CB-54" Safety injecten piping 61-099 Upstream of Flued Head No No No No No No No No from control vahr (Item #14) 2CV-5077-2 to containment penetration 2P-24 LPSI LPSI-001 2CCB-54" Safetyinjection piping 66 007 Downstream of eitmw #4 No No No No No No No No from control vahr 2CV-5077-2 to containment penetration 2P-24 LPSI LPSI-001 2CCB-54" Safety injection pipi:ig 66-008 Upstream of cibow #4 No No No No No No No
' No '
from control nive 2CV-5077-2 to containment
. penetration 2P-24 LPSI LPSI-001 2CCB-54" Safety injection pipiaig 66-009 Downstream ofelbow #3 No No No No No No No No
' from control valve 2CV-5077-2 to containment penetration 2P-24 Desradstoon Medienssna T-Thennel Fatigue
' P-Prvnary Water Stress Cerronen Craciung (PWSCC)
M-ML-? ' J "y bdhseneed Carressan(MIC)
F-r*.aw Accelerused Corressen C CorrosionCracking I - beersranular Stress Carremen Crachng (IGSCC)
E-Eressen-Caviestion 0-Oesr
O'#"'"" " ^'" ##'8N"16 #" 88 FMECA - Degradation Mechanisms Page B8 of B54 l Weld System ID Segment Line Number Line Description Number Weld IAcation T
C P
I M
E F
0 j
LPSI LPSI-001 2CCD-5-6" Safety injection piping 66-010 Downstream of elbow #2 No No No No No No No No from contml valve and upstream ofcibow #3 2CV-5077-2 to l
l containment l
perdration 2P-24 LPSI LPSI-001 2CCB-5-6" Safety injection piping 66-011 Upstream of cIbow #2 and No No No No No No No No from control valve downstream of Flued llead 2CV-5077-2 to (Item #14 on dwg 2CCB containment 1) penetration 2P-24 LPSI IPSI-001 2CCD-5-6" Safety injection piping 66-016 Pipe Weld in between No No No No No No No No 1
from control vahe Pipes #1 and #5 2CV-5077-2 to containment penetration 2P-24 LPSI LPSI-001 2CCB-64" Safety injection from 61 022 Downstream of motor-No No No No No No No No 2CV 5057-2 to operated valve 2CV-5057-containment 2 (Item #16) penetration 2P-29 LPSI LPSI-001 2CCB4-6" Safety injection from 61-023 Upstream of Flued IIcad No No No No No No No No 2CV-5057-2 to (Item #13) containment penetration 2P-29 LPSI LPSI-001 2CCB-6-6" Safety injection from 66-001 Dumscam of elbow #3 No No No No No No No No 2CV-5057-2 to containment penetration 2P-29 Dearadation Mecharunrrs T-Thermal Fatigue P - Pnmary Water Str:s Carrosion Cracking (PWSCC)
M-M; 24MyInnuencedCarressen(MIC)
F-flow Accelerseed Cerrosion C-Conesson Cracking I-IMergranular Stress Carrosion Crackmg(IGSCC)
E-Erosian-Cavitatson 0-Other e
O O
m p
V V
C"latan n uo. AMG-CetWl6.Ren 00 FMECA - Degradation Mechanisms Page B9 of B54 Weld System ID Segment Line Number Line Description Number Weld Imation T
C F
I M
E F
O LPSI LPSI001 2CCB-64" Safety injection from 66-002 Upstream ofelbow #3 No No No No No No No No 2CV-5057-2 to containment penetration 2P-29 LPSI LPSI-001 2CCB-6-6" Safetyinjection from 66-004 Donstream ofelbow #2 No No No No No No' No No 2CV-5057-2 to containment l wtration 2P-29 LPSI LPSI-001 2CCB-6-6" Safety injection from 66-005 Upstream orelbow #2 No No No No No No No No 2CV-5057-2 to containment penetration 2P-29 LPSI LPSI-001 2GCB-3-12" Low pressure safety 53 008 Upstream ofconcentric No No No No No No No
- No injection pump 2P-60A reducer #26 discharge LPSI LPSI-001 2GCB-3-12"
' Low pressure safety 53 009 Dowrstream ofcibow #21 No No No No No No No No injection pump 2P40A discharge LPSI LPSI-001
' 2GCB-3-12" Low pressure safety 53-010 Upstream ofelbow #21 No No No No No No No No injection pump 2P-60A discharge LPSI LPSI-001 2GCB-3-12" Low pressure safety 53-011 Du..ms. of elbow #20 No No No No No' No No No injecten pump 2P40A discharge LPSI
'LPSI-001 2GCB-3-12" Low pressure safety 53-012 Downstream ofelbow #19 No No No No No No No No
'njection pump 2P40A and upstream ofelbow #20 discharge wm Mechen=ns i
T-Thmnal Fatigue P - Pnmary Water Stress Commsen Craciung (PWSCC)
M - Microisiologicany bdhsenced Cerroman (MIC)
F-Flow Accelerused Carramen C-Comm.encv cking I-Leery n=larserenCarmue CracLhis(losCC)
E-Eromen-Cavitseian 0-Oew I
cola,ianon Na AMWQld, Rn. 00 FMECA - Degradation Mechanisms Page BIO of B54 Weld System ID Segment Line Number Line Description Number Weld Location T
C P
I M
E F
0 LPSI LPSI-001 2GCB-3-12" Low pressure safety 53-013 Upsiream ofelbow #19 No No No No No No No No injection pump 2P-60A discharge LPSI LPSI-001 2GCB-3-12" Low pressure safety 53-014 Downstream of elbow #18 No No No Ne No No No No injection pump 2P-60A discharge LPSI LPSI-001 2GCB-3-12" Low pressure safety 53-015 Downstream ofelbow #17 No No No No No No No No injection pump 2P-60A and upstream ofelbow #18 discharge LPSI LPSI-001 2GCB-3-12" Imw pressure safety 53-016 Upstream cf elbow #17 No No No No No No No No injection pump 2P-60A discharge LPSI LPSI-001 2GCB-3-12" Low pressure safety 53-017 Downstream ortee #30 0n No No No No No No No No injection pump 2P-60A pipe #4 discharge LPSI LPSI-001 2GCB-3-12" Imw pressure safety 53-018 Downstream ofelbow #13 No No No No No No No No injection pump 2P-60A and upstream of tee #30 discharge LPSI LPSI-001 2GCB-3-12" Imw pressure safety 53-019 Downstream orstop check No No No No No N:
No No injection pump 2P-60A valve 2SI-3A (Item #59) discharge and upstream of eltow #1" LPSI LPSI-001 2GCB-3-12" Low pressure safety 53-023 Downstream ortee #30 No No No No No No No No injection pump 2P-60A discharge LPSI LPSI-001 2GCB-3-12" tow pressure safety 54-002 Upstream of CNCTRC No No No No No No No No injection pump 2P40A reducer #27 discharge Dearndation Mechamans T-Thmnal Fatigue P - Pnmary Water Stress Common Cracking (PWsCC)
M - Minbeologica!!y Influenced Corrasson (MIC)
F-Flow Accelerated Common C-Common Cracking I - Interprannhr Stress Ccerossen Cracking (10 SCC)
E-Eromen-Cavitation 0 - Other e
O O
)
FMECA - Degradation Mechanisms Gerlad n A*a AMMQ14 Ra 00 Page Bil of B54 W eld System ID Segment IJoe Number Line Description Number Weld lecation T
C P
I M
E F
0 LPSI LPSI-001 2GCB-3-12" IAw pressure safety 54-002A In between Pipes #8 and No No No No No No No No injection pump 2P-60A
- 9 after 1" plug (Item #71) discharge I
LPSI LPSI-001 2GCB-3-12" Low pressure safety 54-002B In between Pipe #9 and No No No No No No No No injection pump 2P-60A Pipe #10 discharge I
LPSI LPSI@l 2GCB-3-12" Low pressure safety 54-003 Downstream ofelbow f22 No No No No No No No No injection pump 2P-60A discharge LPSI-LPSI-001 2GCB-3-12" Low pressure safety 54-004 Upstream ofebow #22 and No No No No No No No No injection pump 2P-60A Domitstream of Elbow #24 discharge LPSI LPSI-001 2GCB-3-12" Low pressure safety 54-005 Upstream ofelbow #24 No No No No No No No No injection pump 2P-60A discharge LPSI LPSI-001 2GCB-3-12" Low pressure safety 54-006 Downstream ofelbow #23 No No No No No No No No injection pump 2P-60A discharge LPSI LPSI-001 2GCB-3-12" Low pressure safety 54-007 Upstream ofelbow f23 No No No No No No No No injection pump 2P-60A discharge LPSI LPSI-001 2GCB-3-12" Low pressure safety 54 008 Downstream orelbow #23 No No No No No No No No injection pump 2P40A discharge LPSI LPSI-001 2GCB-3-12" Low pressure safety 54-009A Upstream ofelbow f23 No No No No No No No No injection pump 2P40A discharge Dearadem Medenn=s T ~Ihe malFatigue P - Pnmary Water Stres Cerrasson Cradung (f"# SCC)
M - MicrotnologicmHy benenced Carresson (MIC)
F-flow Amelerseed Carremen C-Cononce Cracking I-:.%
Stress Cerremon Craciung(losCC)
E-Ernsson-Cavismuon 0-Other
C"'c"I*'"
- A N " ' W #
FMECA - Degradatia-Mechanisms Page B1: cf 354 W eld System ID Segment Une Nam'ver Line Desaiption Number Weld Iecaties T
C P
I M
E F
Low pressure safety 54-010 Dv-s-of cIbow #22 No tb No No No No Nc No injection pmp 2P-60A discharge ISSI 1ESI401 2GCB-3-12*
Imw pressure safety 54-01I Upstream ofc! bow #22 No No No Ib No No No 74 injection pump 2P-60A I
discharge LPSI 11 S1-001 2GCB-3-12*
Low pressure s;.fety 54-Olla Between Pipes #3 and 87 No No No No No No No No injection pump 2P-60A discharge LPSI LPSI401 2GCB-3-12*
Iew pressure safety 54-011B Between Pipes #7 and 86 No No No No No No No No injection pump 2P-60A discLarge LPSI LPSI-001 2CCB-3-12*
Low pressurc safety 54-012 Dv-smc orelbow #21 No No No No No No No No injecuon pump 2P-60A discharge LPSI LPSI401 2GCB-3-12*
tow p: essure safety 54413 Upstream orcibow #21 No No No No No No No No injecten pump 2P-60A and in,-dw of elbow I
discharge
- 20 LPSs LPSI-001 2GCB-3-12*
Low pressure safety 54-014 Upstream orelbow s20 No No No
.4 No No No Na I
injection pump 2P-60A discharge LPSI LPSI-001 2CCB-3-12*
Iow pressure safety 54-015 Dv-sms. ofelbow fl9 No No No No No No No No injecuan pump 2P-60A discharge LPSI LPSI-w.
Low pressure safety 54-016 Upstream ofelbow fl9 No No No No No No No No injection pump 2P-60A discharge Desradsbon Medisasses T. Thermal Fabgoe F - Pruriary Water Stress Cerramon Crackmg (TRW M - haboologicaDy Indl.sericed Corresson (M1C)
F-fle= Acce3erunedCarmuses c-cerromen omims
!-Irmersr===Isr sire Cvmean crxk OosCC)
E-Eramen -Caveneem O -odier e
O O
V d
V,a
(
l
'** k A M N -Cl4Rn.M FMECA - Degradation Mechanismes Page Bl3 of B.tt Wild Systems ID Segument Line Number IJee Description Number Weld Iacaties T
C F
I M
E F
0 LPSI LPSI401 2GCB-3-12' Low pressure safety 54-017 Don 1 stream ofcibow #18 No No No No No No No No injection pump 2P40A discharge LPSI LPSI-001 2GCB-3-12" low pressure safety 54-018 Upstrum ofelbow 818 No No No No No No No No injection pump 2P40A discharge LPSI LPSI-001 2GCB-3-12*
Low pressure safcty 54-019 Du s-n ofelbow 817 No No No No No No No No injection pump 2P40A discharge LPSI LPSI-001 2GCB-3-12*
Iow pressure safety 54-020 Upstream orelbow fl7 No No No No No No No No injection pump 2P40A discharge LPSI LPSI-001 2GCB-3-12*
Iow pressure safety 54-020A Betzten pipes 2 and l No No No No No No No No injection pump 2P40A disch.rge LPSI LPSI-001 2GCB-3-12" tow pressure safety 54-021 Upstream of tee *25 No No No Na No No No No injection purg 2P40A did-p LPSI LPSI-001 2GCB-3-12" Low pressure safety 54-022 Dominstream of tee #25 on No No No No No No No injection pump 2?40A Pipe #13 di d - p LPSI LPSI-001 2GCB-3-12" low pressure safety 54-023 D-&m. of elbow #6 No No No No No No No No injection punp 2P40A and uymme of t:e #25 discharge LPSI LPSI-001 2GCB-3-12*
Low pressure safety 54424 LW afelbow f6 No No No No No No No No injection pump 2P40A discharge T-Thwesi Fenigme F - Pnenary Weser Stress Cerec=an Craciung (FWsCC)
M - Mm:retwaleepenEy Indhsenced Commem (MIC)
F-mw Acizierened Curr smen c-Corrw== tracking I-6,
- seems Cerro-ce cruisig OGsCC)
E-Eresem-Cavenhan 0-06er
- '""*" "* " " # #6 8 " #
FMECA-Degradation Mechanisms Fage Bit of B54 W eld System ID Segment une Number Une Description Number WeldImaties T
C P
I M
E F
Low pressure safety 54-025 Domustream ofstop check No No No L
No No No No injection pump 2P-60A valve 2SI-3B (Item f32) discharge LPSI LPSI-001 2GCB-3-12*
Low pressure nfety 57-001 Upstream ofelbow f14 No No No No No No No No mystron pump 2P-60A discharge LPSI LPSI-001 2GCB-3-12*
Low pressure safety 57-002 Du-n-u of erbow f14 No No No No No No No No injection pump 2P-60A discharge LPSI LPSI-001 2GCB-3-12*
Lew pressure safety 57-003 Upstream orcibow fl5 No No No No No No No No injection pump 2P-60A discharge LPSI LPSI-001 2GCB-3-12*
Low pressure safety 57-004 Downstream ofcIbow fl5 No No No No No No No No injection pump 2P-60A discharge LPSI LPSI-001 2GCB-3-12*
Low pressure safety 57-005 Upstream ofelbow #16 No No No No
?b No No No injection pump 2P-60A discharge LPSI LPSI-001 2GCB-3-12*
Low pressure safety
~7-C")6 Dv-s-- of cIbow fl6 No No No No No No No No inyection pump 2P-60A and upL-- of manual discharge valve 2SI-4A (Item #58)
Imr pressure safety 58-001 Upstream ofelbow f5 No tb No No No No No No injection pump 2P-60A Add =su LPSI LPSI-001 2GCB-3-12*
Low pressure safety 58-002 Du-ioL-- of elbow f 5 No No No No No No No No injection pump 2P-60A dbd-su Deeradshem Medamems T-11mmal Tatigue P - Prsnary Waw stess Cermmon Cracting (FWsCC)
M-M na_':i nneeredCarrassen(MIC)
F. Flow AccelerusedCarrwmen i
C-Carwman Cractsig I " e.,
A-Stress Carmsven Cradung(IGSCC)
E-Eressen -Cavess.en 0-other O
O O
O O
O C"'c"'"'""' ##"N#* R" 88 FMECA - Degradation Mechanisens Page BIS of B54 i W eld System ID i,a :
Line Number Line Descripties Moseber Weld Locaties T
C F
I M
E F
Low pressure safety 58-002A Dv-i&enh ofpiping No Ne No No No A
No No irgeden pump 2P-60A secuen #12
Iow pressure safety 58-003 Upstream ofcIbow M No No No No No Na No No injection pump 2P-60A discharge LPSI LPSI-001 2GCB-3-12" Low pressure safety 58-004 Downstream orelbow M No No No No No No No No injecuon pump 2P-60A and sph-o of cibow #3 discharge LPSI LPSI 001 2GCB-3-12*
Low pressure safety 58 005 Dv-==-- of elbow f3 No No No No No No No No injecuon pump 2P-QA Edmp LPSI LPSI-001 2GCB-3-12*
Low pressure safety 584X4 Upstream ofcibow #2 No No No No No No Fo No irgecuon pump 'P-60A discharge LPSI LPSI4)01 2GCB-3-12*
Iow pressure safety 58-007 Don 1: stream ofelbow #2 No No No No No No No No irgecuon pump 2P-60A and upL-.. ofcibow #I discharge LPSI LPSI 001 2GCB-3-12*
Iow pressure safety 58-008 Dyms-- of elbow #I No No No No No No No injechon pump 2P-60A and upstream ofmarur I discharge valn: 2SI-4B (Item #31)
Low pressure safety 53-005 Downstream ofelbow f32 No No No No No No No No injecuon pump 2P-60A and oph e ofconcentnc discharge reducer #28 LPSI LPSt-001 2GCB-3-14*
Iow pressure safety 53-006 Upsueam ofelbow #32 No No No No No No No No injechon pump 2P-60A
&Gwy Dewedmas h T-nermal Famigue P - Pnmary Waner Sims Cermoene Ciudang (FHW M-I:---
' J ",InnuencedCarreann(MIC)
F Flow AmienedCarremen c-Corremo Cr duas 1-tmery nasersire=Cerra Cr chng(IGsCC)
E-Eremen-Cseman 0-Oew
C"'"'"" Aa Ammni6. Rm o0 FMECA - Degradation Mechanisms Page B16 of R 4 W eld System ID Segment Line Number Line Description Number Weld leention T
C P
I M
E F
I.sw pressure sarety 53-007 Du-a-ofconcentnc No No No No No No No No injection pump 2P40A reducer F26 discharge LPSI LPSI-001 2GCB-3-14" tow pressure safety 54-001 Domistream ofconcentnc No No No No No No No No l
injection pump 2P40A reducer #27 discharge LPSI LPSI-001 2GCB-7-10" Low presstEe safety 55-012 Upstream ofconcentnc No No No No No No No No injection discharge rertrer #28 header (10* piping)
LPSI LPSI-001 2GCB-7-10" Low pressure safety 55-018 Upstream of tee f25 No No No No No No No No injection discharge header (10* piping)
LPSI LPSI-001 2GCB-7-10" Low pressure safety 55-019 Du-swo of elbow f27 No No No No No No No No inject.on discharge header (10* piping)
LPSI LPSI-001 2GCB-7-10" Low pressure safety 55-020 Du=&
ofcIbow #26 No No No No No No No No injection discharge and upsma, ofelbow #27 beader (10" piping)
LPSI LPSI-001 2GCB-7-10" Low pressure safety 55-021 Downstream of tee 824 No No No No No No No No injection discharge and upstream of cibow #26 header (10* piping)
LPSI LPSI-001 2GCB-7-10" Lot. pressure safety 55-027 Upstream of tec #24 No No No No No No No No injection discharge header (10* piping)
Imw pressure safety 55-028 Du-smn orconcentne No No No No No No No No injection discharge reducer #28 beader (10* piping)
Dearadssace Medienums T-Therr ini Fatigue F - Pnrnary Waser Stress cerrassan Crackmg (F%W M - Macrobeeiogicany bdimenced Cer essee (MIC)
F-Dour AccelerseedCerrames C-Cerremen Crociang I - beergrunnier Stress Cerroman Cncing (IGsCC)
E - Ernema-Cavitsamen 0-Other O
O O
I O.
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V v
v FMECA - Degradation Mechanismus NIN A AMMWliRm W Pne BI7 of B54 Weld
. Syseesa ID Segnsent Iline Nuneber Lise Description Noesber WeldIscaties T
C F
I M
E F
Low pressure safety 5M29 Dontistream ofelbow f22 No No No No No No No No injection im*wg and up -.ofconcentne o
header (14" piping) reducer #28 LPSI LPSI-001 2GCB-7-14" Low pressure safety SM30 Upstream ofelbow f22 No No No No No No No No injection &x:eg header (l4* piping)
LPSI LPSI-001 2GCB-7-14" Low pressure safety SM34A At Weldolet #46 No No No No No No No No injection discharge header (14" piping)
LPSI LPSI-001 2GC'3-7-14" Low pressore safety 5M35 Dv.-u-orelbow #23 No No No No No No No No injectson &nimy l
header (14* piping)
Low pressure safety 55-036 Upstream ofcibow #23 No No No No No No No No injectson discharge header (I4* piping)
Low pressure safety SM37 Dowinstream of few orifice No No No No No No No No injection discharge 2FE5091 header (14* piping)
Low pressure safety 55-038 Upstream of flow orifice No No No No No No No No l
injecison &s! :p 2FE 5091
[
i header (14* piping)
}
Law pressure safety SM39 Downstream ofebow f24 No No No No No No No No injection &w: g
[
header (14* peptag)
Low pressure safety 55-040 Upstream ofcibow F24 No No No No No
'No No No injection discharge header (14* piping)
Dsam namW e
T-lliernant Fengue P - Prsnary werer Stress Carvessee Creduas (FWsCC)
M
J27Indlmanzad Carrousen(MIC)
F-F*seur Ameternaed Commmen f
c-Caerusine onduns I-Isinerrenmist Stress Cervemoon CW 00 SCC)
E-Eressee-Coventina 0-Odier l
i.
p.
l N "'" * *""W #"#
FMECA-Degradation Mechanisms Page BIS cf B54 W eld System ID Segment Line Number Line Description Number Weld Iecation T
C P
I M
E F
Low pressure safety 55441 Dom stream ofelbow #25 No No No No No No No No injection drscharge header (14* pqnng)
Low pressure safety 55442 Upstream ofcibow s25 No No No No No No No No injection discharge header (14* piping)
LPSI LPSI-001 2G(.B-7-14" Low pressure safety 55442A Dumm.-ofwtidolet No No No No No No No No injection discharge
- 45 header (I4* piping)
Low pressure safety 55442B Downstreamofutidolet No No No No No No No No injectien discharge
- 44 header (14* piping)
Low pressur: safety 55443 Doulistream ortec #39 No No No No No No No No inyecuon discharge header (14* piping)
Low pressure safety 55444 W of tee #39 No No No No No No No No injection discharge header (14* piping)
LPSI LPSI-001 2GCB-7-14" Iow pressure safety 55-045 Domitstream ofelbow f21 No No No No No No No No injection discharge header (I4* piping)
LPSI LPSI.001 2GCB-7-14' Low pressure safety 55446 Upstream ofcIbow f21 No No No No No No No No inject.% discharge header (14* piping) pen asoa.uea.===
T-neemmt Fatigue F - Phmary Wate Streis Carmaan Cradung(FWsCC)
M - Micretneicpently InAmenced Commsen (MIC)
F-fleur Accelerened cammene c-car e e=creding I-Insersre mier swess comme.cr chng poscc)
E - E,
- c o-ce,r e
e e
I V(.
(
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V C"'~'^"~r A7e AMMC416 Rn. M FMECA - Degradation Mechanisms Page B19 of 854 Weld Syst as ID Segeneet use Number Line Descripties Nasuber Weld Location T
C P
I M
E F
Connecting line to 50448 66-.cf weldolet No No No No No No No No SDC droplineoff LPSI
Connecting linc to 50449 Upstream ofelbow s38 No No No No No No No No SDC drop line off LPSI injecten header outside contamment
)
LPSI LPSI-001 2G(. >7-4*
Connecting line to 50-050 Downstream ofelbow #38 No No N
No No No No No SDC dropline offLPSI injecten header outside containment LPSI LPSI-001 2GCB-74*
Low pressure safety 55-001 L &-- ofelbow f37 No No No No No No No No injection discharge header (6* piping)
Low pressure safety 55-002 Upstream ofcibow #37 No No No No No No No No injecten discharge header (6* piping)
Iow pressure safety 55-003 Du= &-. of elbow #36 No No No No No No No No injection discharge header (6 piping)
Low pressure safety 55-0C4 Upstream ofelbow #36 No No No No No No No No injecten discharge header (6* piping)
Low pressure safety 55-005 Downstream ofcibow #35 No No No No No No No No injecten hp header (6* piping)
De W
. T-Thermal Fatigme F-Pnmary Waner stream Cermenn Cracking (FWSCC)
M-:.:; -.
- 4 :,ImAmencedCorremen(MIC)
F flew AccelurusedCarremen C-Cerresum Crecimag I -!.e,
Stress Commen Osckes(IGSCC)
E-Eremse-Cownesme 0 -Oemr t
FMECA - Degradation Mechanisms N " Na RMWI6. Rm RJ Page B:0 of B54 W eld System ID Segment Line Number Line Descri tion Number Weld Locaties T
C P
I M
E F
0 P
Low pressure safety 55-006 Upstream ofcibow #35 No No No No No No No No injection discharge header (6* piping) i LPSI LPSI-001 2GCB-74*
Low pressure safety 55-007 Douttstream ofelbow f34 No No No No No No No No injection discharge header (6" piping)
Low pressure safety 55-003 Upstream ofcibow #34 No No No No No No N
No injection discharge header (6" piping)
Low pressure safety 55-009 Do=1tstream orcibow f33 No No No No No No No No injection discharge header (6* piping)
Low pressure safety 55-010 Upstream of cibow #33 No No No No No No No No injection dsharge 3
header (6* piping)
Low pressure safety 55-0II Douttstream ofconcentnc No No No No No No No No injection discharge reducer #23 header (6* piping)
Low pressure safety 55-013 Dominstream cfelbow #32 No No No No No No No No in,ection discharge header (6" piping)
Imw pressure safety 55-014 Upstream ofelbow f32 No No No No No No No No injection discharge header (6* piping)
Low pressure safety 55415 Dun uww=. ofelbow #31 No No No No No No No No injection discharge header (6* piping)
_W T-Thmani Fatigue F-Pnenary Water stress Common Cradung(Fum M-ML 24 j bdhsmced Comme (%CC)
F-Flow AccelerenedCarrearm c-Cane==a Cr=ck-g I - Iniersranatar stre= Corr==>an ondung (IGsCC)
E-Eressam-Cavitsham O-Deur O
O O
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O FMECA-Degradation Mechanismas C"'c"'"' " hMMWid Rm 00 Page B I of B54 W eld Syseese sD Segument Line Nember Line Descripties Nonsber Weld Imcation T
C P
I M
E F
Iew pressure safety 55-016 Upstream ofelbow f31 No No No No No No No No injection discharge header (6* piping) -
Low pressure safety 55-017 D - =h - oftee#25 No No No No No No No No injection discharge header (6* piping)
Low pressure safety 55-022 Dominstream ofelbow f30 No No No No No No No No irijection discharge header (6* piping)
LPSI LPSI-OL!
Iow pressure safety 55423 Upstream ofelbow #30 No No No No No No No No injecten discharge header (6* piping)
Low pressure safety 55424 Downstream ofelbow f29 No No No No No No No injection discharge he::Jer(6 piping)
Low pressure safety 55-025 Upstream ofcibow #29 No No No No No No No No injecten discharge header (6" piping)
Low pressure safety 55426 Downstream of tee #24 No No No No No No No No injection discharge header (6* piping)
Low pressure safety 55-03I Downstream ofelbow #36 No No No No No
'No No No injection discharge header (6* piping)
Iow pressure safety 55-032 Upstream ofcibow #36 No No No No No No No No injection discharge hesJer(6* piping)
Desadsham W T-Thennel Fusigue P-Pnrnary Weser Stress Corrensen Crackmg (FWsCC)
M *&. "
- J ",InnuencedCorressen(nGC)
F-flew AccelerenedCerranse C.C ra.ionCr cking I-L, Stres Cerremen cractang(IGsCC)
E-Eressen-Cavennen 0-Odier
m.
..mm
l C#"'"*"r uome ai6. Rn. m FMECA - Degradation Mechanisms Page B:2 of B54 W eld System ID Segment Line Number Line Description Number Weld Escation T
C P
I M
E F
low pressure safety 55-033 biswo orcIbow s35 No No No No No No No No l injection discharge header (6" piping)
Iew pressure safety 55-034 Upstream ofelbow #35 No No No No No No No No injection discharge header (6* piping)
Iew p: essure safety 61-002 Upstream of LPSI header No No No No No No No No injection discharge vaht 2CV-5077-2 tw wier (6" piping)
low pressure safety 61-011 Upstream of LPSi header No No No No No No No No injection discharge vahe 2CV-5037-1 header (6" piping)
Iow pressure safety 61-021 Upstream ofLPSI header No No No No No No No No injecuon discharge vahr 2CV-5057-2 header (6" piping)
tow pressure safety 61-023 In between Pipe #19 and No No No No No No No No injection discharge Pipe #20 header (6" piping)
Low pressure safety 61-025 Upstream of LPSI header No No No No No No No R
injection discharge vahe 2CV-5017-1 header (6" piping)
LPSI LPSI-001 2GCB-7-8" tow pressure safety 53404 D-sm-of concentne No No No No No No No No injecuon discharge reducer #28 header (8* piping)
low pressure safety 55-053 Upstream of tee #39 at No No No No No No No No injechon discharge Pipe #42 header (8* piping)
Desredsoon Medumesnm T-Thermal Famigue
? - Prvnary Water stress Cerrosym Cradtmg (FWSCC)
M - MicrdeolopcmDy Irasenced Cerrowne (MIC)
F-flow Accelerened Comsman C-Carreman Cr=*ina I-Iraersrerwier stres Cerremen Cracbng OGSCC)
E - Eremen -Cavestaan 0-06er e
9 9
o o
U V
U N
FMECA - Degradation Mechanisms Nalenwr WMW'-016 RnE Page B23 of B54 WeM System ID L,,a.;
Line Number 11rne Description Neanber WeM1mcaties T
C P
I M
E F
Low pressure safesv 55453A bas--ofmanual No No No No No No No No inject.on discharge valve 251-5093-3 (ISO header (8" piping) 2GCB-7-1)
Imw pressure safety 55-057 LW ofweldolet #44 No No No No No No No No injection discharge header (8* piping)
LPSI LPST-001 2GCB Low pressure safety 55-058 bis-.. ofelbow #31 No No No No No No No No inject;on discharge header (8* piping)
Low pressure safety 55-059 Upstream ofelbow f31 No No No No No No No No injection discharge header (8* piping)
Iew pressure safety 55-060 b as-- ofelbow #48 No No No No No No No No injection discharge header (8" piping)
Iow pressure safety 55-061 bas-n ofbutterfly No No No No No No No No injectson discharge vahr 251-5091-3 (Item header (8* piping)
Iew pressure safety 55-062 Upstream ofbutterfly No No No No No No No No injection discharge sahr 2SI-5091-3 (Item header (8" piping)
- 54)
Low pressure safety 55-063 Downstream ofc! bow #30 No No No No No No No No inyection discharge header (8* piping)
Dearedmine h 7-Tnenal Tsaigne F-e5 mary Weser stress Carrossen Cracksng (PEW M - Microbioingscally Inome= iced remesen (MIC)
F.How AccelerusedCompaesa C-Cerressee Cracking 1
- L,,
stres Cerremenn Cracksig OGSCC)
E-Ereason-Caneshan 0-Odier
l l
C"I""*" ^a mmW16. Ra. 00 FMECA - Degradation Mechanisms Page B24 of B34 Weld System ID Segment Line Number Line Description Number Weld lecation T
C P
I M
E F
Low pressure safety 55-oG4 Upstream ofcIbow #30 No No No No No No No No injection discharge j
header (8* piping) l LPSI LPSI-001 2GCB-74*
Irw pressure safety 5545 Upstream ofwridolet #45 No No No No No No No No injection discharge header (8" piping)
Low pressure safety 55-066 Downstream orgate raht No No No No No No No No injection discharge 2SI-509I-2 (Item #5tt) header (8" piping)
Iew pressure safety 554 7 Downstream ofelbow #33 No No No No No No No No injection discharge and upstream ofgate raht header (8" piping) 251-5091-2 (item #81)
Low pressure safety 5548 Upstream ofcibow #33 No No No No No No No No injection discharge header (8" piping)
Low pressure safety 55 4 9 DouT stream ofbutterfly No No No No No No No No injection discharge vahr 2CV-5091 (Item #55) header (8* piping)
inw pressure safety 554170 Upstream ofbatterfly No No No No No No No No injection discharge vahr 2CV-5091 (Item #55) header (8" piping)
Low pressure safety 55-071 Dv..ohme of manual No No No No No No No No injection discharge vahr 2SI-5091-1 (item header (8* piping)
- 53)
Desredstme Wm T-Thmnal Fatigue P - Pnrnary Water stress Cerrassen Creding (FWSCC)
M - MicrabsekyscaDy IrJbenced Cerveusen (hDC)
F-Flow Accelerened Commsen c-Cam.weCmurs I. beergrusmier stress Carnman Cracong OosCC)
E-Eseason-Cavststram O-Oiher O
O O
O U
.Om m
C""##" *""##' #" 8' FMECA - Degradation Mechanisms Page B25 of B54 W eld System ID Seguient une Number Une Description Number Weld Escaties T
C P
I M
E l'
Low pressure safety 55472 h us.-- of elbow s32 No No No No No No No Ne injection discharge and vym-u of manual header (8" p ping) valve 251-5091-1 (Item
- 53)
Low pressure safety 55-073 Upstream orelbow f32 No No No No No No No injection discharge header (8* piping)
Low pressure safety 55-074 Downstream of tec #34 No No No No No No No No injecuon discharge header (8* piping)
Low pressure safety 55-075 Upstream of tee #34 No No No No No No No No injection discharge header (8" piping)
Iow pressure safety FW-36 6.=>-- of piping No No No No No No No No injecuon discharge secuen #9 header (8* piping)
. Shutdown cooling hest 55-047 Downstream ofelbow #14 No No No No No No No No exchangerdiid p and upstream of manual header to low pressure valve 251-5093-2 (Item safety injection header
- 27)
Shutdown moling heat 55-048 Downstream ofconcenanc No No No No No No No No exchanger discharge reducer #13 and upstream header to low pressure ofelbow #14 safety injecuan header Desrome== Me<*--
T-nsnnel remisee r-rnrnarywesersire cerwaancreding(rum M - :i..
- . Inamene dcarre sen(MIC)
F-Flow AcxxierusedCamusam c-carmeen creding I-L; seress carre=en Craing(10 SCC)
E-Eressen-Caveman 0-Odur
l l
1 1
caic=torm.Na mmr-oit Rw. oo l FMECA - Degradation Mechanisms Page B26 of B34 Weld Sydem ID Segment Line Number Une Descriptica Number Weld Leesties T
C P
I M
E F
SDC heat exchanger 55-049 Downstream ofbutterfly No No No No No No No No discharge header to valve 2CV 5093 and LPSI header (8* piping) upstream ofconcentne reducer #13 LPSI LPSI-002 2CCA-25-14*
Shutdown cooling 25-001 Adpcent to hot leg Yes No No No No No No No sucuan line from RCS hot leg (Loop B) to shutdown cooling suction valw 2CV-50 % 2 LPSI LPSI-002 2CCA-25-14" Shutdown cooling 25-002 Weld at RCS hot leg Yes No No No No No No No suction line from RCS (Loop B) hot leg (Loop B) to shutdown cooling suction valve 2CV-5086-2 LPSI LPSI-002 2CCA-25-14*
Shutdrma cooling 25 003 Upstream ofcIbow #1 Yes No No No No No No No suction line from RCS hot leg (Loop B) to shutdown cooling sucuon va!w 2CV-5086-2 LPSI LPSI-002 2CCA-25-14" Shutdown coolinr,25-004 Dumhe ofelbow f t Yes No No No No No No No suchon line frtra RCS hot leg (Loop B) to shutdown cooling
^
sucuon valve 2CV-5086-2 Denndesa Mediarmes T-Hermal Fatigee P-Prsnary Wakr stress Carauca Craciung N)
M-E _ '
",inomencedCorremen(MIC)
F-Flow Accelerseed Commune c-cerroman Cracbng I-Inergma=I r seres.Commen Cradmg F4 E -Eremen-Cavesasse 0-Oiher e
9 9
O O
O FMECA-Degradation Mechanisnas Page B27 ef B54 Weld System ID Segment Line Number Line Descripties Numeber T:Id IAcation T
C P
I M
E F
0 LPSI LPSI@2 2CCA-25-14*
Shutdown cooling 25-004A At 3"weldolet Yes No No No No No No No suction line from RCS hot leg (Loop B) to shutdown cooling section nIve 2CV-5086-2 LPSI LPSI-002 2CCA-25-14" Shutdown cooling 25-005 Upstream ofcibow #4 Yes No No No No No No No suction line from RCS hot leg (Loop B) to shutdown cooling suction taht 2CV-5086-2 LPSI LPSI-002 2CCA-25-14" Shutdown cooling 25-006 Downstream ofelbow #4 Yes No No No No No No No sucten line from RCS hot Icg (Loop B) to shutdemi cooling sucoon vahr 2CV-5086-2 LPSI LPSI-002 2CCA-25-14" Shutdown cooling 25-007 Upstream ofelbow #5 Yes No No No No No No No sucten line from RCS hot Icg (Loop B) to shutdown cooling suction vaht 2CV-5086-2 Deandmuan " > ___
T-Hennel Fmigue P-Prunary Weer stres common Chring (Pum M-L
- ;,IndImencedCarreusen(MIC)
F-flow Amsteraned Carrassem c-corressan creadng I%
E-Erossen-casemasse 0-Oeur
l l
1 N'"##' "Na A&MQi6.Rm M FMECA - Degradation Mechanisms j
Page B:8 of B54 l Weld l
I System ID Segment Line Number Line Description Number Weld Location T
C F
I M
E F
0 LPSI LPSI-002 2CCA-25-14" Shutdown cooling 25-008 Downstream ofcibow #5 Yes Ne No No No No No No suction line from RCS hot leg (Loop B) to shutdown cooling suction vahr 2CV-5086-2 LPSI LPSI-002 2CCA-25-14*
Shutdown cooling 25-009 Upstream ofcIbow #2 Yes No No No No No No No suction line from RCS j
hot leg (Laop B) to shutdown cooling suction uhr 2CV-5086-2 LPSI LPSI-002 2CCA-25-14*
Shutdown coohng 25-010 Du-uhm ofcibow #2 Yes No No No No No No No suction line from RCS hot leg (Ieop B) to shutdown cooling suction uhr 2CV-5086-2 LPSI LPSI-003 2CCA-25-14*
Shutdown cooling 25-014 Downstream of motor-No No No No No No No No suction line from RCS operated uht 2CV-5084-I hot leg (teop B) to shutdown cooling suction vahr 2CV-5086-2 Desradmasan Mahansms T-lhennel ratigue P - Prunary Waner stress Cerrwason Omdung (PWSCC)
M-EJ "
, he'luenced Cerreusen(%GC)
F mw AcceleraardCorressen c-Cerramen Cracking I-keergran=Isr s=== Cerroman Camiing OGSCC)
E - Erence-Cavneteen 0-Oeer e
G G
(y e
a j
V
\\
V 1
N'"#"8 "Na AGMWI6 Rn 00 FMECA-Degradation Mechanisms Page B19 of B54 w eld System ID Segment Line Number line Descripties Noesber Weld IAcation T
C F
I M
E F
0 LPSI LPSI403 2CCA-25-14" Shutdoun cooling 25415 6.s-- of elbow f6 No No No No No No No No sucten line from RCS hot leg (Loop B) to shutdown cooling suction vahr 2CV-5086-2 LPSI LPSI-001 2CCA-25-14" Shutdomm cooling 25416 Upstream of l4* X 8*
No No No No No No No No suctmn line from RCS reducing tee 815 hot leg (Imop B'L shutdown cooling
)
suction vaht 2CV-5086-2 e
LPSI LPSI403 2CCA-25-14" Shutdown cooling 25-017 6.s mo of l4* X 8*
No No No No it No No No sucten line from RCS reducing see #15 hot leg (Imop B) to shutdown cooling sucten vahr 2CV-5086-2 LPSI LPSI-003 2CCA-25-14*
Shutdown cooling 25-018 Upstream of motor-No No No No No No No No suction line from RCS operated valve 2CV-5086-2 hot leg (Irop B) to shutdown cooling sucten set 2CV-5086-2 LPSI LPSI-003 2CCA-25-8*
Shutdown cooling 25-019 Upstream of I** X 8' No No No No No No No No section from refueling 4 tee #15 canal (from vahr 2SI-19 to 14" shutdown cooling line)
Dundess N T-Unrnial Famigue P - Pnnery weser Swese Comunen Oracksng (PWsCC)
M - MicrabsoleyceGy Inamenced Commmen (MIC)
F-flo. Accelermaed C,,emen C-Cerrassen Cracksng I-Irmerymenier Stress Correene Cradtung (10 SCC)
E-Ereason-Cowmana 0 -Odsw
C""#" 'V" A*"##6 #" 8' i
FMECA - Degradation Mechanisms Page B30 of BS4 W eld System ID Segment Line Number Line Description Number Weld lacation T
C P
I M
E F
O LPSI LPSI-003 2CCA-25-8' Shutdown cecling 25420 Du-swui of cibow S4 No No No No No No No No suction from refueling canal (from vahr 251-19 to 14* shutdown cooling line)
Shutdown cooling 25421 Upstream ofcIbow #4 No No No No No No No No suction from refacting
]
canal (from vahr 2SI-19 to 14* shutdown cooling line)
Shutdomi cooling 25-022 Dv &-.ofcIbow #3 No No No No No No No No suction from refueling canal (from vahr 2SI-19 to 14* shutdown cooling line)
Shutdown cooling 25-023 Downstream of manual No No No No No No No Ne siscison from refueling uht 251-19 canal (from nht 251-19 to 14* shutdown coolingline)
LPSI LPSI-003 2DCB-504-2*
LPSI pump 2P-40A 53446 Du.um-ofcheck No No No No No No No No mini-flow recuculation uhr 2SI-22A (ISO 2DCB-line (from du.om-.
504-1) ofcheck vahr 2SI-22A)
LPSI LPSI-003 2DCB-504 IPSI pump 2P-60A 53-047 Upstream ortee #24 (ISO No No No No No No No No mini-flow recuculation 2DCB-504-1) line (from downstream ofcheck vahr 2SI-22A)
Desradene Mafwm!m T-Thermat Fatigue F - Pnenary Water Stress Carresean Crachng (F%W M - Micad=ologscaDy Isomenced Cerrassen (MIC)
F-Flew AccelerusedCarrenen C-Correman CracUng 1 -lesergrwesir Stress Cemiace Crechng OGSCC)
E-Eressen-Cavesenen 0-Oder O
O O
^N C's V(N (V
V
)
" ~ "^" * " M N 8" 88 FMECA - Degradation Mechanismos Foge E31 of B54 WeW Systems ID Segiment IJee Number Line Descripties Noseber WeM IAcntesa T
C P
1 M
E F
0 LPSI LPSI-003 2DCB-504 LPSI pump 2P40A 53448 Downstream of tee #24 -
No No No No No No No No mini-flow recirculation vent path 650 2DCB-504-line (from downstream I) ofcheck valve 2SI-22A) l LPSI LPSI-003 2DCB-504-2*
LPSi pump 2P40A 53-049 Downstream of tee #24 -
No No No No No No No No mini-flow reciruni.uw.
flow path (ISO 2DCB-504-line (from downstream I) ofcheckvalve 2SI-22A)
LPSI LPSI-003 2DCB-504-2*
LPSI pump 2P40A 53-050 Upstream of tec #23 (ISO No No No No No No No No mini-flow wo: anon 2DCB-504-1) line (from downstream ofcheck valve 251-22A)
LPSI LPSI-003 2DCB-504-2*
LPSI pump 2P40A 53451 Dowinstream of tee f23 -
No No No No No No No No mini-flow recirculation flow path OSO 2DCB-504-line (from dom 1: stream 1) ofcheck valve 2SI-22A)
LPSI 13'SI-003 2DCB-504-2*
LPSI pump 2P40A 53-052 Dowinstream of tee #23 -
No No No No No No No No mini-flow recirculacon drain path (ISO 2DCB-line (from dcwnstream 504-1) ofcheck vahr 2SI-22A)
LPSI LPSI-003 2DCB-504-2*
LPSIpump 2P40A 53453 Upstrea'n ofmotor No No No No No No No No mini-Gow recirculacon operated vahr 2CV-5123-line (from downstream I (ISO 2DCB-504-1) ofcheck valve 2SI-22A)
LPSI LPSI-003 2DCB-505-2*
LPSI pump 2P40B 54-052 Downstream ofcheck No No No No No No No No
~
mini-flow.m.u.:&.w vahr 2SI-22B (ISO 2DCB-line (from dominstream 505-1) ofcheck vahr 2SI-22B)
Desmessai Medm===es T-Thenne! Fahgoe F - Pnrnary Weser Smens Cornswun Cruisng (FRW M - MicretmobycmBy Indimanced Cmemen (hGC)
F-flour AccesureaseCommen C-Carreseen Cradung I. Inneryanaler Sims Cmence Oasing OGSCX')
E-Eremse-Cavemame 0-Oe=r
" "' *
- A- * * " # " 8' FMECA - Degradation Mecharsisms Fage B32 of B54 W eld System ID Segment Une Number Line Description Number Weld Locaties T
C P
I M
E F
LPSi pump 2P-608 54-053 Upstream oreIbow 81 No
?b No No No No No No mini-flow recirculation (ISO 2DCB-505-1) line (from downstream of check nhr 251-22B) i LPSI LPSI-003 2DCB-505-2*
LPSi pump 2P-60B
$4-054 Dun swm oreRnw #1 No No No No No No No No mini-flow recirculation (ISO 2DCB-505-1) line (from d<miistream ofcheck vahr 251-22B)
LPSI LPSI-003 2DCB-505-2*
LPSi pump 2P-60B 54-055 Upstream ortee #33 (ISO No No No No No No No No mini-flow recirculation 2DCB-505-1) line (from downstream ofcheck nhr 2SI-22B)
LPSI LPSI-003 2DCB-505-2*
LPS1 pump 2P-60B
$4-056 Domitstream of tec 833 -
No No No No No No No No mini-flow recirculatmn flow path (ISO 2DCB-505-line(from du. sum I) ofcheck nhe 2SI-22B)
LPSI LPSI-003 2DCB-505-2*
LPSi pump 2P-60B 54-057 Dv-iswm of tee #33 -
No No No No No No No No mini-flow recirculation drain path (ISO 2DCB-line (from downstream 505-1) of check vahr 2SI-22B)
LPSI LPSI-003 2DCB-505-2*
trSI pump 2P-60B 54-058 Upstream of motor No No No No No No No No mini-flow recirculation operated nht 2CV-5124-line (from downstream I (ISO 2DCB-505-1) ofcheck nive 2SI-22B)
tow pressure safety 51-014 Downstream of I4" x 12*
No No No No No No No No injecten pump 2P-60A reducmg cibow #16 inlet at pump inlet n
_.Ma$mnrems T-Thenal Fatigue P - Pnmary Weser Stress conessee Oscking(FWSCC)
M - Microbielegicalfy Isoleerical Cerreries (MIC)
F-Nur Acerterused carre====
c-Cance an Cracking I-: &,, - -sirmcorrueceCr.cimeOosCC)
E-Enessee-Canamanan 0-Odier O
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-m a
l U
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Cdcolatierr Na A-FDUC4LC-01S Rer 00 1 FMECA-Degradation Mechanismas
.,,,a33yauj weid l
Syseen ID Segneemt Liec Nomeber Line Desenption Nanter Weld Loessies T
C P
I M
E F
Low pressure safety 51-015 Dowinstream of flange #21 No No No No No No No No injechon pump 2P-60A inlet at pump inlet LPSI LPSI-003 2GCB-t-12*
In pressure safety 51-016 Upstream ofcibow #17 No No No No No No No No injechon pump 2P-60A inlet at pump inlet LPSI LPSI-003 2GCB-t-12*
Iow pressure safety 51-017 Downstream ofelbow #17 No No No No No No No No injecuen ptunp 2P-60A inlet at pump inlet LPSI LPSI-003 2GCB-t-12" tow pressure safety 51-018 Upstream of flange #22 at No No No No No No No No injecuon penp 2P-60A pump 2P40A Inlet inlet at punp inlet LPSI LPSI-003 2GCB-1-14*
Iow pressure safety 51409 Downstresm ofelbow f14 reo No No No No No No No injecuen pump 2P-60A inlet LPSI LPSI-003 2GCB-I-I4*
Low pressure safety 51-010 Upstream of tee #19 on No No No No No No No No iniechen pump 2P-60A Pipe #1 inlet LPSI LPSI-003 2GCB-1-14" Low pressure safety 51-011 Downstream ofpipe #2 No No No No No No No No injecuon pump 2P60A and upstream of tec #19 inlet LPSI LPSI-003 2GCB-t-14*
Low pressure safety
$1-Olla Downstreamofelbow fl5 No No No No No No No No injecuan pump 2P-60A inlet
. i ndesacedCarremmeOGC)
F-Fhm Acr-%rused Comassa "I
T-Thanent Fatipse F - Prweary Waeer Swess cermeem crackmg (PWSCC)
M "J J c-c,romenosams I - : --,,
Swess cemmen cWsymCc)
E - sw==e - ca.emm==
o-omer
FMECA - Degradation Mechanisms C"*" "^'* " " # ##R" "
Page B34 of B34 Weld System ID Sepnent Line Number Line Description tiember Weld Imcaties T
C F
I M
E F
Im pressure safety
$1-012 Downstream ermanual No No No No No No No No inyection pump 2P-60A uhr 2SI-2A (Gate Vaht inlet
Low pressure safety 51-013 Du-mstream of tec #19 an No No No No No No No No inject on pump 2P-60A pipe #3 inlet LPSI LPSI403 2GCB-1-14*
Low pressure safety SI-013 A Upstream of 14* x 12*
ito No No No Nc No No No injection pump 2P-60A reducing cibow #16 in!ct LPSI LPSI-003 2GCB-17-12*
Discharge beader from FW-15 Downstream orrahc 2SI-No Fo No No No No No No shutdoni cooling heat SB u.Isr@ r LPSI LPSi403 2GCB-2-12*
Low pressure safety 52429 Downstream of 14* x 12*
No No No No No No No No injection pump 2P-60B reducing cibow #2 inlet piping at pump inlet LPSI LPSI-003 2GCB-2-12*
Low prest r safety 52-030 Downstream of flange #22 No No No No No No No No injection mamp2P-60B inlet pipin g at pump inlet LPSI LPSI-003 2GCB-2-12*
Low pressur t safety 52-031 Upstream ofelbow s3 No No No No No No No No injection pumy 2P-60B inlet piping at pimp inlet Dearadaban Mechenrsms T-Therinal Fatigue P - Prunary Waser Stress Cerronne Cracinng (PwSCC)
M - MeebeelegicaDy Innemced Carrousen (MIC)
F-flow Accelerskd Carrammen C-Cerrensen Cracking I *.a.,
a Stress Cerrwoon Cracking (rGSCC)
E-Erwesese -Cantatsen 0-other O
O O
N"""*A *
- W'**"88 FMECA - Degradation Mechanisms Page B35 of B34 W eld Sy:: ens ID Segment Line Number Line Description Number Weld Emestion T
C P
I M
E F
0 LPSI LPSI403 2GCB-2-12" Low pressure safety 52-032 Downstream ofelbow #3 No No No No No No No No injection pump 2P-60B in~et piping at pump inlet LPSI 13'S1-003 2GCB-2-12*
Low pressure safety 52-033 Upstream of flange #21 at No No No No No No No No g
injection pump 2P-60B Pump 2P-60B Inlet inlet piping at pump inlet LPSI LPSI-003 2GCB-2-14*
Low pressure safety 52-022 Downstream ormanual No No No No No No No No injection pump 2P-60B nlw 251-1B (Gate Valve inlet piping
- 5)
Low pressure safety 52-.43 Downstream ofpipe #9 No No No No No No No No injechon pump 2P-608 and upwn of tee #24 inlet piping LPSI LPSI-003 2GCB-2-14*
Low pressure safety 52-024 Du.onwm of pipe #3 No No No No No No No No injection pump 2P-60B and upwusi of tee #24 inlet piping LPSI LPSI-003 2GCB-2-14" Low pressure safety 52-025 Downstream of manual No No No No No No No No injection purep 2P-60B vain 2SI-2B (Gate Vale inlet piping
- 32) and uym-- of elbow #1 LPSI LPSI-003 2GCB-2-14" Low pressure safety 52-027 Du-nou-o of tee #24 No No No No No No No No injection pump 2P-60B inlet piping LPSI LPSI-003 2GCB-2-14*
Low pressure safety 52-02E Upstream of 14* x 12*
No No No No No No No No injection pump 2P-60B reducing cibow #2 inlet piping Desr=5*= Maf=nams T-Thennel Fatigue F - Pnrnary Water Stress C-Cracksne (PWsCC)
M - Micr*ioiryceDy influenced Cmenem (MIC)
F-I'io= AccelmsedCorremen C-Cmasson Cradung I
- e,~1 Stress Ceomen Cruiing 00 SCC)
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C""*'^~" N " # '68"
- FMECA - Degradation Mechanisans Page B37 of B54 wen Systema ID Segament une Number Line Description Nesser WeM Imation T
C P
I M
E F
From 2CCA-25 to low 30 001 Downstream ofelbow fl2 No No No No No No No No Pressure safety injection pump 2P-60A
& B inlet LPSI LPSI-003 2GCB-5-14" From 2CCA-25 to low 30-002 Upstream ofcibow #1I No No No No No No No No Pressure safety injection pump 2P-60A
& B inlet LPSI LPSI-003 2GCB-5-14*
From 2CCA-25 to low 30-003 Downstream ofcibow #11 No No No No No No No No Pressure safety injection pump 2P-60A
& B inlet LPSI LPSI-003 2GCB-5-14" From 2CCA-25 to low 30-005 Don 1: stream ofelbow #10 No No No No No No No No Pressure safe injection pump 2P40A
& B inlet LPSI LPSI-003 2GCB-5-14*
From 2CCA-25 to low 30 006 Upstream ofcibow #10 No No No No No No No No Pressure safe injection pump 2P-60A
& B inlet LPSI LPSI-003 2GCB-5-14" From 2CCA-25 to low 30 007 Dom 1: stream ofelbow #14 No No No No No No No No Pressure safe injection pump 2P-60A
& B inlet LPSI LPSI-003 2GCB-5-14" From 2CCA-25 to low 30-008 Upstream ofelbow #14 No No No No No No No No Pressure safe injection pump 2P-60A
& B inlet Desradsa.,a W T-Thermal Fatiger P-Pesvary Water Stress Corremme Creding (F%W ht - hdoolorgmEw bdhsered Cemenin (MIC)
F-Flow AccriermeedCemenu c-Corromen Crecug I a, sirewCarm caCraing(IGsCC)
E-Erassen-Ca viession 0 -Odser
i C"' "" """ A**K-Ot6. nm os FMECA - Degradation Mechanisms Page B38 of B34 Weld Systesa ID Segreent Line Number Line Description Neinber Weld Leestion T
C F
I M
E F
From 2CCA-25 to low 30 8109 Du-awm ore! bow #13 No No No No No No No No j Pressum safety injection pump 2P-60A
& B inlet LPSI LPSI-003 2GCB-5-14*
From 2CCA-25 tolow 30410 Upstream ofelbow fl3 No No No No No No No No Pressure safety injection pump 2P-60A A B inlet LPSI LPSI-003 2GCB-5-14" From 2CCA-25 to low 30-012 Upstream ofelbow #9 No No No No No No No No Pressure safety injection pcmp 2P40A
& B in!ct LPSI LPSI-003 2GCB-5-14" From 2CCA-25 to low 30-013 Downstream ofcibow #8 No No No No No No No No pressure safety injection pump 2P40A
& B inlet LPSI IPSI-003 2GCB-5-14*
From 2CCA-25 to low 30-014 Upstream ofelbow #8 No No No No No No No No Pressure safety injection pump 2P-60A
& B inlet LPSI LPSI-003 2GCB-5-14*
From 2CCA-25 to low 30-017 Du-umwo ofefbow f9 No No No No No No No No pressure safety injecten pump 2P-60A
& B inlet LPSI -
LPSI-003 2GCB-5-14" Fruct 2CCA-25 to low 50-001 Du-uwam cf Flued No No No No No No No No pressure safety Head #79 (2P-27) irtjection pump 2P-60A
& B inlet Deeradam Mederees T "DwernalFatigme P-Prunary Weser stress Cerroman Crschng(FWSCC)
M - Miant=ciogscalfy Irdluenced Car euen (MIC)
F-Flow AccekrusedCenmmen c Cerramo Cracming I-Irey
- sure e Carmaa Q% 00 SCC)
E Eresw=-Canishoe 0-OGur O
O O
O O
O N '* * #" A * " # #8 #" 88
' N
FMECA - Degradation Mechanisans Pmee B39 of B34 Weld System ID Segument UneNeuiber Line Deecription Number Weld IAcaties T
C P
I M
E F
From 2CCA-25 to low 50 002 Upsta..am of motor-No No No No No No No No pressure safety operated valw 2CV-5938-injection pump 2P40A I (Gase Vahe #78)
& B inlet LPSI LPSI-003 2GCB-5-14*
From 2CCA-25 to low 50-003 Dominstream of motor-No No No No No No No No pressure safety operated vaht 2CV5038-1 injection pump 2P40A (Gase Valw #73)
& B Unlet LPSI LPSI-003 2GCB-5-14" From 2CCA-25 to low 50-004 Downstream of sockolet No No No No No No No No pressure safety
- 52 injechon pump 2P40A
& B inlet LPSI LPSI-003 2GCB-5-14*
From 2CCA-25 tolow 50-005 Upstream ofelbow f23 No No No No No No No No Pressure safety injecuen pump 2P40A
& B inlet LPSI LPSI-003 2GCB-5-14*
From 2CCA-25 to low 50-006 Downstream orelbow f23 No No No No No No No No pressure safety and upstream of tee #39 injecuen pump 2P40A
& B inlet
^
LPSI LPSM03 2GCB-5-14" From 2CCA-25 to low 50 @ 7 D - h.oftee#39 No No No No No No No No Pressure safety injecuan pump 2P40A
& B inlet LPSI LPSI-003 2GCB-5-14" From 2CCA-25 to low 50-008 Upstream ofelbow #24 No No No No No No No No Pressure safety injecnon pump 2P40A
& B inlet Demadman -
T-Thermal Famigme F-Phsmary Weser sarcus corressen Creams (F%W M - M=. _
" huomenced Cereseen(W3C)
F-flour Acce4arenedCerremen
.1 c-Carremen Creding I-:..- y
- stress Cerrm:m Craciung OG5CC)
E-Desen-Cavesse 0-Oemr
-.. _ ~
Narlod n A'a A-PDMW16. Rm 00 FMECA - Degradation Mechanisms Page B40 4 B34 Weld System ID Segment Line Number Line Description Number Weld Location T
C P
2 M
E F
0 LPSI LPSI403 2GCB-5-14" From 2CCA-25 to low 50-009 Dowitstream orcibow #24 No No No No No No No No pressure safety and upstreaan ofcibow #25 ir.jection pump 2P40A
& B inlet LPSI LPSI-003 2GCB-5-14" Fram 2CCA-25 to low 50-010 Downstream of elbow #25 No No No No No No No No pressure safety injection pump 2P40A
& B inlet LPSI LPSI-003 2GCB-5-14' From 2CCA-25 alow 50-011 Upstream ofelbow #26 No No No No No No No No I pressure safety injection pump 2P40A I
& B inlet LPSI 1/111 413 2GCB-5-14" From 2CCA-25 to icw 50412 Downstream of elbow #26 No No No No No No No No pressure defy injection pump 2P40A
& B inlet LPSI LPSI-003 2GCB-5-14" From 2CCA-25 to low 50-013 Upstrearl ofelbow #27 No No No No No No No No pressure safety injection cump 2P40A
& B inlet LPS:
LPSI-003 2GCB-5-l i" From 2CCA-25 to low 50-014 Dcwnstream ofelbow #27 No No No No No No No No pressure safety injection pump 2P40A
& B inlet LPSI LPSI-003 2GCB-5-14" From 2CCA-25 to low 50-015 In beturen Pipe 5 and No No No No No No No No pressure sar.-ty Pipe 6 injection pump 2P40A
& B inlet Denradetion Mecherusa; T-Thermal Fatigue P - ftirnary Watar Stress Carramen Cracimg (PWSCC)
M - Microtsale;;pc.Dy influenced Carramen (WC)
F-Tiow AcceleresedCarronen C-Cormion Cricies I - Intergranstar stresc Cerroman onchng 00 SCC)
E - Lesson -Cavumeson 0- Ottar e
9 9
l FMECA - Degradation Mechanisms
"#"'#*" ^'" "j[3j
' d*"
W eld System ID Segment Line Number Line Description Number Weld Iacaties T
C P
I M
E-F 0
LPSI LPSI-003 2GCB-5-14" From 2CCA-25 to low 50416 Upstream ofelbow #23 No No No No No No No No Presme safety injection pump 2P-60A
& B inlet LPSI LPSI-003 2GCB-5-14" From 2CCA-25 to low 50-017 Downstream ofelbow #28 No No No No No No No No
~
Pressure safety j
injection pump 2P-60A i B inlet LPSI LPSI-003 2GCB-5-14" From 2CCA-25 to low 51 001 Upstream ofelbow #5 No No No No No No No No pressure safety injection pump 2P-60A
& B inlet LPSI LPSI-003 2GCB-5-14*
From 2CCA-25 to low 51402 Domistream ofelbow #5 No No No No No No No No Presme safety injection pump 2P-60A
& B inkt LPSI bSI-003 2GCB-5-14" Frore 2CCA-25 to low 51-003 Upstream of elbow #6 No No No No No No No No Press saret7 injection pump 2P-60A
& B inlet LPSI LPSI403 2GCB-5-14" From 2CCA-25 to low 51-004 Downstr -un ofelbow #6 No No No No No No No No Pressure safety injection pump 2P-60A
& B inlet LPSI LPSI-003 2GCB-5-14" From 2CCA-25 to low 51-005 Upstream ofelbow #7 No No No No No No No Pressure safety injection pump 2P-60A
& B inlet Desudenna Mechenmes T-Thmnal Fatigue F - Franary Waser Stress Common Cracisng (FWSCC)
M - ; ;~->., hunmenced Common (MIC) 1r -flow Accelerssed Commen c-Corremen onriing I-heerpennier seen Common Oraciang (IGMr)
E-Eroenn-Centsasen O-Oqher
(
FMECA - Degradation Mechanisms C"" lad n A'a A6M-Ol6, Rm 00 Page B42 of BS4 Weld System ID Segment Line Number Line Description Number Weld Location T
C P
I M
E F
From 2CCA-25 to low 51-006 Donstream ofelbow #7 No No No No No No No No pressure safety injection pump 2P-60A
& B inlet LPSI LPS1403 2GCB-5-14*
From 2CCA-25 to low 51-007 Upstream of manual vahr No No No No No No No No i pressure safety 2SI-l A (Gate Vahr #8) injection pump 2P-60A
& B inlet LPSI LPSI-003 2GCB-5-14" From 2CCA-25 to low 51-003 Downstream of manual No No No No No No No No pressure safety nhe 2SI-1 A (Gate Vahr i
injection pump 2P-60A
%)
& B inlet LPSI LPSI-003 2GCB-5-14*
From 2CCA-25 to low 52-001 Upstream ofelbow #24 No No No No No No No No pressure safety injection pump 2P-60A
& B inlet LPSI LPSI-003 2GCB-5-1 **
From 2CCA-25 to low 52402 Dumohwn ofelbow #29 No No No No No No No No pressure safety injection pump 2P-60A
& B inlet LPSI LPSI-003 2GCB-5-14" From 2CCA-25 to lew 52-003 Upstream of elbow #38 No No No No No No No No pressure safety injection pump 2P-60A
& B inlet LPSI LPSI-003 2GCB-5-14*
From 2CCA-25 to low 52-003A Downstream ofelbow #3h No No No No No No No No pressure safety injection pump 2P-60A
& B inlet Desradsuon M "
T-Thermal Fatigue P - Pnmary Water Stress Corrosion Cracksng (PWS6 X')
M - Maik"y Influered Cer omen (MIC)
F-Flow Accelerened Carresam C CarromonCracking I - Intergrsrmist Stress Common Craciing (IOsCC)
E-Eroman-Cavitmien 0 -Other O
O O
O O
O FMECA - Degradation Mechanisnas Weld System 3D Segment Line Number.
Line Description Number Weld IAestion T
C P
I M
E F
0 i
LPSI LPSI-003 2GCB-5-14" From 2CCA-25 to low 52-004 Upstream ofelbow #30 No No No No No
' No No No.
Pressure safety injection pump 2P40A
& B inlet LPSI LPSI-003 2GCB-5-14" From 2CCA-25 to low 52-005 Downstream ofelbow No No No No No No
'No No.
pressure safety
- 30/ upstream ofelbow #31 injection pump 2P-60A
& B inlet LPSI LPSI-003 2GCB-5-14" From 2CCA-25 to low 52-006 Dowiestream ofelbow #31 No No No No No No No No p-essure safety injection pump 2P-60A
& B inlet LPSI LPSI-003 2GCB-5-14" From 2CCA-25 to low 524)o7 Upstream ofelbow #32 No No No No No No' No No, pressure safety injection pump 2P-60A
& B inlet LPSI LPSI-003 2GCB-5-14" From 2CCA-25 to low 52-008 Dow1: stream ofelbow #32 No No No No No No No No Pressure safety injection pump 2P-60A
& B inlet LPSI LPSI-003 2GCB-5-14"
. From 2CCA-25 to low 52-009 Upstream ofc! bow #33 No No No No No No
-No No Pressure safety injection pump 2P40A
& B inlet LPSI -
LPSI-003 2GCB-5-14" From 2CCA-25 to low 52-010 Downstream ofelbow #33 No No No No No No No No Pressure safety injecten pump 2P40A
& B inlet Deseassumi N T-ThermalFati ue F - Pnmary Water Stress Carrossen Cracking (FwSCC)
M-:1. '
- J -b bdimenced C-(MIC)
F-How AccelermedCeressen t
C-Carre senCracking I. :.a,
seress Carrossen Cradung(IG9CC)
E-Eressen-Cavemason 0-Oiher
FMECA - Degradation Mechanisms
- [j#"[
C"'"""*"
HM7 g
W eld System ID Segment Line Number Line Description Number Weld Exation T
C P
I M
E F
0 LPSI LPS1403 2GCB-5-14" From 2CCA-25 to low 52-01I Upstream ofelbow #34 No No No No No No No No pressure safefy injection pump 2P-60A
& B inlet LPSI LPSI-003 2GCB-5-14" From 2CCA-25 to low 52-012 Downstream of elbow No No No No No No No No pressure safety
- 34/upwwi. of cibow #35 injection pune 2P-60A t
& B inlet LPSI LPSI-003 2GCD-5-14" From 2CCA-25 to low 52-013 Downstream ofelbow #35 No No No No No No No No wsure safety injection pump 2P-60A
& B inlet LPSI LPSI-003 2GCB-5-14" From 2CCA-25 to low 52-014 Upstream ofelbow #36 No No No No No No No No pressure safety injection pump 2P-60A
& B inlet LPSI LPSI-003 2GCB-5-14" From 2CCA-25 to low 52-015 Downstream ofcIbow #36 No No No No No No No No pressure safety injection pump 2P-60A
& B inlet i
1 LPSI LPSI-003 2GCB-5-14" From 2CCA-25 to low 52-016 Upstream ofcIbow #37 No No No No No No No No pressure safety injection pump 2P-60A
& B inlet LPSI LPSI-003 2GCB-5-14" From 2CCA-25 to low 52 017 Dowitstream ofelbow #37 No No No No No No No No pressure safety injection pump 2P-60A
& B inlet Dear dation W T-Thmnal FMigue P - Pnmary Wata Strees Cerrosion Cracting (PWSCC)
M - Microlmolopeally Inonenced Commen (MIC)
F-now Accelerated Common c-Com=wn Cracking I. Intagranular stres Cemmen Creding OGSCC)
E-Ereseen Cavitauen 0-Oiher O
O O
O C
D C"'""#"^'**""'**"
FMECA - Degradation Mechanisms Page B45 of B54 Weld System ID Segment Line Number Line Description Number Wele Imaties T
C P
I M
E F
0 i
LPSI LPSI-003 2GCB-5-14" From 2CCA-25 to !ow
$2-018 Upstream orelbow #3 No No No No No No No No pressure safety injection pump 2P-60A
& B inlet LPSI LPSI-003 2GCB-5-14" From 2CCA-25 to low 52-019 Downstream ofelbow #3 No No No No No No No No pressure safety and upstream of elbow #4 5
injection pump 2P-60A
& B inlet LPSI LPSI-003 2GCB-5-14" From 2CCA-25 to low 52-020 Dom 1 stream ofelbow #4 No No No No No No No No Pressure safety injection pump 2P-60A
& B inlet LPSI LPSI-003 2GCB-5-14" From 2CCA-25 to low 52421 Upstream of manual valve No No No No No No No No pressure safety 2SI-IB (Gate Valve #5) 1 injection pump 2P40A
& Binlet LPSI LPSI-003 2GCB-5-14" From 2CCA-25 to low A
Upstream ofelbow #12 No No No No No No No No Pressure safety injection pump 2P-60A
& B inlet LPSI LPSI403 2GCB-5-14" From 2CCA-25 to low D
Downstream of p' ping No No No No No No No No pressure safety section #80 injection pump 2P40A
& B inlet LPSI LPSI-003 2GCB-5-14" From 2CCA-25 to low FWI Dowsntream of vahe 2CV-No No No No No No No No
~
pressure safety 5086-2 injection pump 2P-60A
& B inlet Desadenan Mechan==s T-Thennel Fatigue F - Pnmary Waser Stress Corroman Cracket (PWSCC)
M-MR
- J 5,basewedCommsen(MIC)
F-Flow Accelerseed Compman C-Commsen Cracking I-beergrunrlar Stress Common Cradung (IGSCC)
E-Eresson Cavitshan 0- Other
_~
..w
.................................r i.,
14-Sep C"Ic=I"uon so. A-rexacttc-Of6, n,,, og FMECA - Degradation Mechanisms Page B46 of B54 W eld System ID Segment Line Number Line Description Number Weld Location T
C P
I M
E F
0 LPSI LPSI403 2GCB-5-3" Connecting line to 50-031 Doumstream of tee #49 No No No No No No No No CVCS ofTSDC drop line outside containment LPSI LPSI-003 2GCB-5-3" Connecting line to 50-032 Downstream ofelbow #50 No No No No No No No No CVCS ofTSDC drop line outside containment LPSI LPSI-003 2GCB-5-3" Connecting line to 50-033 Upstream of manual valw No No No No No No No No l CVCS ofTSDC drop 2SI-35 line outside containment l
LPSI LPSI-003 2GCB-5-4" Connecting line to 50434 Downstream ortee #47-No No No No No No No No LPSIinjection header shown in 2GCB-5-I-2 ofISDC dropline i
outside containment LPSI LPSI-003 2GCB-5-4" Connecting line to 50-035 Downstream of piping No No No No No No No No LPSIinjection header section #16 ofTSDC drop line outside containment LPSI LPSI-003 2GCB-5-4" Connecting line to 50-036 Upstream of cibow #40 No No No No No No No No LPSIinjection header off SDC dropline outside containment LPSI LPSI-003 2GCB-5-4" Connecting line to 50-037 Downstream ofcibow #40 No No No No No No No No LPSIinjection header ofTSDC dropline outside containment Dearadsuon Mechamms T.Hmnal Fatigue P - Pnmary Water Stms Cerrosen Cracking (PWSCC)
M-M W W fInnuenced Carressen(MIC)
F-Flow Acrelerused Carrossen C-Cerressen Qacking I Ireergranular Stress Corream.a Cracking (IGSCC)
E-Ernsson-Cavitehen 0 - Other O
O O
o O
O l
C**'*" " ^'" A*""I' R" 88 FMECA - Degradation Mechanisms Page B47 of B34 j w egd System ID Segment Line Number Line Descripties
- Number Weld Location T-C P
I M
E F
0 LPSI LPSI-003 2GCB-5-4" Connecting line to 50-038 Upstream ofelbow MI No No No No No
. No
~ No No LPSIinjection header l
I offSDC dropline outside containment LPSI LPSI-003 2GCB-5-4" Connecting line to 50-039 Downstream ofelbow M1 No No No No No No No No LPSIinjection header off SDC dropline outside containment LPSI LPSI-003 2GCB-5-4" Connecting line to 50-040 Upstream ofelbow M2 No No No No No No No No LPSIinjecten header offSDC drop line outside containment LPSI LPSI-003 2GCE-5-4" Connecting linc to 50-041 Downstream of elbow M2 No No No No No No No No LPSIinjection header off SDC dropline outside containment LPSI LPSI-003 2GCB-5-4" Connecting line to 50-042 Upstream ofelbow M3 No No No No No No No No LPSIinjecten header off SDC dropline outside containment LPSI LPSI-003 2GCB-5-4" Connecting line to 50-043 Downstream ofcibow M3 No No No No No No No No LPSIinjectson header off SDCdropline outside containment LPSI LPSI-003 2GCB-5-4" Connecting line to 50-044 Upstream ofelbow M4 No No No No No No No No LPSIinjection header off SDC drop line outside containment Desradeen th T-normal Fatigue P - Pnmary Waser Stress Cerrasson Cracking (PWSCC)
M-ML'
. -, InAmanced Carrenses(MIC)
F-Nw AccelermandCerveuseo C-Cerronen crecting I-:.c.
Stress Corremen Credung 00 SCC)
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Calculation No. A-PENG-CALC-016. Rev. 00 FMECA - Degradation Mechanisms Page B50 of BS4 Weld System ID Segment Line Number Line Description Number Weld Location T
C P
I M
E F
0 LPSI LPSI-003 2GCB-508-2" LPSI pump 2P-60A 53-043 Upstream ofelbow #7 No No No No No No No.
No mini-flow recirculation (ISO 2GCB-508-1) line (from pump discharge to check valve 2SI-22A) l LPSI LPSI-003 2GCB-508-2" LPSi pump 2P-60A 53-044 Domitstream of elbow #7 No No No No No No No No mini-flow recirculation (ISO 2GCB-503-1) line (from pump discharge to check valve 2SI-22A) l LPSI LPSI 003 2GCB-508-2" LPSi pump 2P-60A 53-045 Upstream orcheck valve No No No No No No No No mini-flow recirculation 2SI-22A (ISO 2GCB-508-line (from pump 1) discharge to check vahr 2SI-22A)
LPSI LPSI-003 2GCB-509-2" LPSI pump 2P-608 54-048 Upstream ofpiping section No No No No No No No No mini-flow recirculation
- 1 (ISO 2GCB-509-1) j i
line (from pump discharge to check valve 2SI-22P)
LPSI LPSI-003 2GCB-509-2" LPSI pump 2P-60B 54-049 Upstream of flow orifice No No No No No No No No mini-flow recirculation 2FO-5119 (ISO 2GCB-line (from pump 509-1) discharge to check valve 2SI-22B)
Dessa.asson Mechar esms T-Thermal ratigue P - Pnmary Water Stress Corrosion Craclung (PWSCC)
M - Microbiologicapy Innsenced Corronen (MIC)
F-flow Accelerened Cerramen C-Corronen Oracking I - Intergranular Stress Corresion Crackmg OGSCC)
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FMECA - Degradation Mechanisms Page B32 of B34 Weld System ID Segment Line Number Line Description Number Weld Location T
C P
I M
E F
0 LPSI LPSI-003 2GCB-8-14" Shutdoun cooling heat 55-052 Upstream of manual vaht No No No No No No No No exchanger discharge 2S1-5093-1 (Item #25) header to low pressure safety injection header LPSI LPSI-003 2GCB-8-14" Shutdown cooling heat 55-055 Upstream ofconcentric No No No No No No No No j
exchanger discharge reducer #1I header to low pressure safety injection header LPSI LPSI-003 2GCB-8-14" Shutdown cooling heat 55-056 Upstream ortee #10 No No No No No No No No exchanger discharge header to low pressure safety injection header LPSI LPSI-003 2GCB-8-14" Shutdown cooling heat 56-001 Downstream of manual No No No No No No No No exchanger discharge vahr 2SI-6 (1 tem #24) header to low pressure safetyinjection header LPSI LPSI-003 2GCB-3-14" Shutdown cooling heat 56 002 Upstream of manual vaht No No No No No No No No exchanger discharge 2SI-6 (Item #24) header to low pressure safetyinjection reader LPSI LPSI-003 2GCB-8-14" Shutdown cooling heat 56-003 Downstream ofelbow #4 No No No No No No No No exchanger discharge header tolow pressure safetyinjection leader LPSI LPSI-003 2GCB-8-14" Shutdown cooling heat 56-004 Duwmbsm of fee #7 and No No No No No No No No exchanger discharge upstream ofcibow #4 header to low pressure safety injection leader Desradation Medarams T-Thermal Fatigue P - Pnmary Water Stress Carroman Cracking (PWSCC)
M - Micratmotopcany InReenced Cerremen (MIC)
F-Flow AcceleratedCommen C-Corrosion Cracking I - traersranatar Streu Cerremon Crackmg (1GSCC)
E - Eremen - Cavenhan 0 -Other O
O O
3 OV (V
C""'i " No A-PENGC4LC-CIS. Rev. 00 FMECA - Degradation Mechanisms Page B33 of B54 W eld l
System ID Segment Line Number Line Description Number Weld Leestion T
C P
I M
E F
0-l LPSI LPSI-003 2GCB-8-14" Shutdown cooling heat 56-005 Upstream of tee #7 No No No No No No No No I exchanger discharge I
header tolow pressure safety injection header 1
LPSI LPSI-003 2GCB-8-14" Shutdown cooling heat 56-006 Upstream ofconcentric No No No No No No No No exchanger discharge reducer #5 header to low pcssure safety injection header LPSI LPSI-003 2GCB-8-3" Connecting line from FW29 Dom 1tstream of weldolet No No No No No No No No SDC to CVCS
' LPSI LPSI-003 2GCB-8-3" Connecting line from FW30 Upstream ofelbow #19 No No No No No No No No SDC to CVCS (ISO 2GCB-8-1)
LPSI LPSI-003 2GCB-8-3" Connecting line from FW31 Upstream ormamial vahe No No No No No No No No SDC to CVCS 2SI-34 (ISO 2GCB-8-1)
SDC heat exchanger 56-007 Don 1: stream ofconcentnc No No No No No No No No discharge header to reducer #5 LPSI header (6" piping)
LPSI LPSI 003 2GCB-8-6" SDC heat exchanger 56-008 Upstream ofelbow #3 No No No No No No No No discharge header to LPSI header (6" piping)
LPSI LPSI-003 2GCB-8-6" SDC heat exchanger 56 009 Dumman ofelbow #3 No No No No No No No No discharge header to LPSI header (6" piping)
LPSI LPSI-003 2GCB-8-8" SDC heat exchanger 55-050 Don 1 stream ofconcentric No No No No No No No No discharge header to reducer #12 and upstream LPS! header (8" piping) ofbutterfly vahr 2CV-5093 (Item #26)
Desradm== Mechamans T-Thmnal Fatisme P - Prenary Water Stress Cerrosen Cracking (PWSCC)
M-ML-f
- J "y hosenced Cerroman(MIC)
F-Flow Acerlermoed Cencema C-Canaman Cracking 1 -Inser, seen Cenomen Cr.ckms 00 scc)
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Calculation No. A PENG CALC 016, Rev. 00 Page C1 of C7 O
APPENDIX C
'FMECA SEGMENT RISK RANKtNG REPORT *
(Attachment Pages C1 - C7) 1 ABB Combustion Engineering Nuclear Operations
l i
N"a dN* *a "
8 *r" FMECA - Segraent Risk Ranking Report re aera Degradation Number Lines in Welds in Degradation Degradation Mechneirm Consequence Risk Risk Segment ID of Welds Segment Segment Mechanisms Group ID Category Category Category Rank LP51-001 187 2CCA-25-14",25-011,25-012, 25-LPSI-N NONE HIGH CAT 4 MEDIUM 2CCB-3-1.5",
Ol3 // 61-018A // 61-l i
2CCB-34*,2CCB- 012,61-013,61-014, 4-1.5", 2CCB 61-015,61-016,61-6", 2 CCB-54",
017,61-018,61-019, t
2CCB-64",2GCB- %-012,66-013 // 61-3-12", 2GCB 030B // 61-026,61-14", 2 GCB-7-10", 027,61-028,61-029, 2GCB-7-14",61-030,61-030A,61-2GCB-7-4",
031,66-014,66-015 2GCB-74",
//61-003,61-004,61-2 GCB-7-8",
005,61-006,61-007, 2GCB-8-14",
61 008,61-009,66-2GCB-8-8" 007,66-008,66-009,66-010,66-011,66-016//61-022,61-023,66-001,66-002,66-004,66-005 //53-1 008, 5.i-009,53-010,53-011,53-012,53-013,53-014, 53-015,53-016, 53-017,53-018,53-019,53-023,54-002,54-002A,54-002B,54-003,54-004,54-005,54-006, 54 4 07,54 008,54-009A,54-010,54 Ol l, 54-Ol l A,54-011B,54-012,54-013,54-014,54-015,54-016, 54-017, 54-O O
O
,O O
D.
U C
V
' *-Ser'7 FMECA - Segment Risk Ranking Report ch * *Scuc-'m h o' re es <cr Degradation Number Lines la Welds in Degradation Degradaties Mechanissa Consequence Risk Risk I
Segmeut ID of Welds Segment Segment Mechanismes Group ID Category Category Category Raek j
I 018,54-019,54-020 54420A,54-021,54-022,54-023, 54-024,54-025,57-00I,57-002,57-003, 57 004,57-005,57-006,58-001,58-002,58-002A,58-003,58-004,58-005,58-006,58-007,58-008//53-005,53-006,53-007, 54-001//55-012,55-0I8,55-019,55-020,55-021,55-027,55-028 //55-029,55-030,55-034A,55-035,55-036,55-037,55-038,55-039,55-040,55-041,55-042, 55-042A,55-042B,55-043,55-044,55-045,55-046//50-048,50-049,50-050
//55-001,55-002,55-003,55-004,55-005,55-006,55-007,55-008,55-009,55-010,55-011,55-013,55-014,55-015,55-016,55-017,55 022,55-023,55-024,55-025,55-026,55-031,55
}{
k l
'N" FMECA - Segment Risk Ranking Report Nh"* *mm 8a
- rna co 4cr Degradation Number Lines in Welds in Degradation Degradation Mechanisme Consequence Risk Risk Segment ID of Welds Segment Segment Mechanisms Group ID Category Category Category Rank 032,55-033,55-034,61-002,61-011,61-02I,61-023,61-025
//53-004,55-053,55-053 A,55-057,55-058,55-059,55 460, 5541,5542,55-063,5544,5545,55-066,55 4 7,55-068,55 4 9,55-070,55-071,35-072,55-073,55-074,55-075 FW-36 // 55-047,55-048//55-049 LPSI-002 11 2CCA-25-14"25-001,25 4 02,25-T LPSI-T SMALL IIIGH CAT 2 HIGI 003,25-004,25-LEAK 004 A,25-005, 25-006,25-007,25-008,25-009, 25-010 e
9 9
s
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FMECA - Segment Risk Ranking Report c** * "m* *=
- r a cs 4cr n
Degradation Number Lines in Welds in Degradation Degradation Methmeism Co w Risk Risk Segment ID of Welds Segment Segment Mechanisms Groep ID Category Category Category Rank LPSI-003 176 2CCA-25-14",25-014,25-015, 25-LPSI-N NONE NEDIUM CAT 6 LOW 2CCA-25-8",
016,25-017,25-018 2DCB-504-2",
//25-019,25-020,25-
. 2DCB-505-2",
//53-046,53-047, 53-2GCB-1-14",
048,53-049,53 4 50, 2GCB-17-12",
53451,53-052,53-2GCD-2-12",
053 // 54-052, 54-2GCB-2-14",
053,54-054,54-055, 2GCB-3-12",54-056,54-057, 54-2GCB-3-8",
058//51-014,51 -
015,51-016,51-017, 2GCB-5-3",
51-018//51-009,51-2 GCB-5-4",
010,51-011,51-2GCB-508-2",
011 A,51012,51-2GCB-509-2",
013,51-013 A // FW-2GCB-78-1.5",
15 //52-029,52-030, 2GCB-78-2",52-031,52-032,52-2GCB-8-12",
033//52-022,52-2GCB-8-14",
023,52-024,52-025, 2GCB-8-3",
52427,52-028, H //
2 GCB-8-6",53-020,53-021, 54-2GCB-8-8" 026,54-027,54-028
//53-022,54-029 //
30-001,30-002,30-003,30-005,30-006,30-007,30-008,30-009,30-010,30-012,30-013,30414,30-017,50-00I,50-002, 50-003,50-004,50-005,50-006,50 4 07,
FMECA - Segment Risk Ranking Report c m ua m ocuc e w =
r.e ce orc?
Degradation Number Lines in Welds in Degradation Degradation Mechanism Consequence Risk Risk Segment ID of Welds Segment Segment Mechanisms Group ID Category Category Category Rank 50 408,50-009,50-010,50-011,50 4 12,50-013,50-014,50-015,50-016,50-017,
$1-001,51-002,51-
)
003,51 004,51-005, 51 006,51-007,51-008,52 4 01,52-002,52-003, 52-003 A,52-004,52-005,52-006.52-007,52-008,52-009,52-010,52-011,52-012,52-013,52-014,52-015,52-016,52-017,52-018,52-019,52-020,52-021 A. D, FW1// 50-031,50-032,50-033 //50-034,50-035, 50-036,50-037,50-038,50-039,50 040,50-041,50-042,50 443,50-044,50-045,50-046, 50-047//53-035,53-036,53-037,53-033,53-039,53-040,53-041,53-042,53-043, 1
53-044,53-045 //54-048,54-049, 54-050, s
54-051// 30-014A, 30-014B // 30-014C
//56-010,56-012, 9
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Calculation No. A PENG-CALC-016, Mov. 00 Page O1 of D5
- O i
i
?
4 4
i-.
t i
APPENDIX 0 1
QUALITY ASSURANCE VERIFICA TION FORMS O
O ABB Combustion Engineering Nuclear Operations r
C:Icut: tion No. A.PENG CALC 016. R:v. 00 Page 02 of Ob Verification Plan
Title:
Implementation of the EPRI Risk-Informed Inservice Inspection Evaluation Procedure for the LPSI/SDC at ANO 2 Document Number:
C-PENG-CALC-016 Revision Number: 00 instr uctions:
Describe the method (s) of verincation to be employed, i.e., Design Review, Attemate Analysis, Qualification Testing, a combination of these or an attemative. The Design Analysis Veri 6 cation Checklist is to be used for all Design Analyses. Other elements to consider in formulating the plan are: methods for checking calculations; comparison of results with similai analyses, etc.
Descrintion of Verification Method:
An independent review will be conducted as appropriate with the work activities described in Project Plan PP 2OOG839, Revision 00. The verification willinclude:
1.
Verification of a Design Analys:s by Design Review (per OP 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 fo* mat,... etc. are made in accordance with Design Analysis Verliication checklist.
.s Verincation Plan prepared by:
Approved by:
N w S E 040stTH
/b/
% T. N8ih> h independent Reviewcr pnnted name and Mgnature
~
Management appmver pnned narhg and signature i /
T ABB Combustion Engineering Nuclear Operations
1 C:Icul: tion No. A-PENG CALC 016, R;v. 00 Page D3 of DS AQ Other Design Document Checklist (Page1of3)
Instructions: The Independent Reviewer as to complete this checklist for each Other Design Document. This Checklist is to be made part of the Quality Record package, although it need not be mede a part of or distributed with the document itself. The 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 Ansk sis Verification Checklist which are not used may be left blank.)
/
Title:
Implementation of the EPRI Risk-Informed Inservice Inspection Evaluation Procedure for the LPSI/SDC at ANO-2 Document Number:
Revision Number:
A-PENG-CALC-016 00 Section 1: To be completed for all Other Design Documents -
Yes N/A Overall Assessment 1
Are the results/ conclusions correct and appropriate nor their intended use?
E 2
Are sillimitations on the results/ conclusions documented?
E Documentation Requirements g
Is the documentation legible, reproducible and in a form suitable for filing and retrieving as a Quality 1.
Record?
II.
Is the document identified by title, document number and date?
111.
Are all pages identified with the document number including revision number? -
IV.
Do all pages have a unique page number?
V.
Does the content clearly identify, as applicable:
A.
objective B
O B.
design inputs (in accordance with QP 3.2)
E O
C.
conclusions S
O VI.
Is the verification status of the document indicated?
Vll.
If an independent Reviewer is the supervisor or Project Manager, has the appropriate approval been E
O documented?
Assumptions 1.
Are all assumption identified, justified and documented?
O O
11.
Are all assumptione Gat must be cleared listed?
O A.
Is a process in place which assures that those which are CENO responsibility will be cleared?
O S
O B.
Is a process in place which assures that those which are the customer's responsibility to clear will O
2 V
be indicated on transmittals to the customer?
ABB Combustion Engineering Nuclear Operations
C:Icul: tion No. A PENG CALC-016, R:v. 00 Fage D4 of DS Other Design Document Checklist (Page 2 of 3)
Assessment of Significant Design Changes Yes N/A 1.
llave significant 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 111.
Are references as direct as possible to the original source or documents containing collection / tabulations of g
inputs?
IV.
Is the reference notation appropriately specific to the information utilized?
O 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?
3 O
Vll.
Is the verification status of design inputs transmitted from ABB CENS appropriate and documentedi O
S
- Vill, is the use of customer-controlled sources such as Tech Specs, UFSARs, etc. authorized, and does the S
O authorization specify amendment level, revision number, etc.?
References 1.
Arc all references listed?
g 11.
Do the reference citations include sufficient information to assure retrievability and unambiguous location g
of the teferenced 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
S 2.
Applicable Codes and Standards O
S 3.
Literature Searches and Background Data O
E 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 S
9.
List of computer disks O
S
.IsoM c d w -/k bd evu/dle ABB Combustion Engineering Nuclear Operations
i C:Icu:: tion NJ. A-PENG CALC 016. R:v. 00 Page 05 of D5
)
V Other Design Document Checklist (Page 3 of 3)
Independent Reviewer's Comments Comment Reviewer's Comment
- Response Au, hor's Response -
Response
Number Required?
Accepted?
1 See changes to table of contents.
Yes Table of contents y,f Update as needed has been updated 2
Wrong pipe code, class as shown Yes Tae correct pipe yg on Table 1 code class for the indicated line is now shownin Table 1 3
Typos pages 14,17,1819,21,22, Yes Typo's have been y,y 24,31,32,34,39,43, A0, A3, A4, corrected A9,A10,A17,A20,A21,A32 4
Page 34 refers to three damage Yes There are indeed y,y groups - should be two?
two damage groups and the correction has been made accordingly.
5 Where is consequence LPSI-C-Yes Consequence LPSI-y,f 09?
C-09 was included with Consequence LPSI-C-08 because of the open penetra-tion now shown on Figure 4 Checklist completed by:
Independent Reviewer [,g Jgg gf.
/![
j g/g/g
(
)
Pnnted Name S
c Date ABB Combustion Engineering Nuclear Operations
____A